ML20245B407
| ML20245B407 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 04/13/1989 |
| From: | Baldwin R, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20245B400 | List: |
| References | |
| 50-327-OL-88-01, 50-327-OL-88-1, NUDOCS 8904260094 | |
| Download: ML20245B407 (300) | |
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ENCLOSURE 1
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l EXAMINATION REPORT 327/0L-88-01 Facility Licensee: Tennessee Valley Authority 6N 38A' Lookout Place 4 1101 Market Street 1 Chattanooga, TN 37402-2801 Facility Name: Sequoyah Nuclear Plant i Facility Docket.Nos.: 50-327 and 50-328 Written examinations and operating tests were administered at Sequoyah Nuclear
. Plant near Soddy Daisy, Tennessee.
Chief Examiner: 8a A*2 - M/ 4-.-.=. v/jd/57 Richard S. Baldwin D6te Signed I Approved By: /tww /-/. h ( V./3 69 Qopn Munro, Chief Date Signed Qperating Licensing Section 1 Division of Reactor Safety Summary:
Examinations were conducted on December 12 - 15, 1988.
Written and operating tests were administered. to' 11 candidates; 7 of whom j passed. 1 Based on the results described above, 6 of 9 R0s passed and 1 of 2 SR0s passed.
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$38"1883AS$8$Ng7i.
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l REPORT DETAILS
- 1. Facility Employees Contacted:
- B. Lovelace, Manager, Sequoyah Operations Group
- M. J. Lorek, Sequoyah, Training Manager
- C. T. Benton, Sequoyah, Operator Training Manager
- C. H. Noe, Operations Training Manager
- W. R. Ramsey, Sequoyah, Senior Operator Instructor i
- E. Keyser, Sequoyah, Senior Operator Instructor
- 2. Examiners:
- Richard S. Baldwin Jim Moorman Glenn Salyers Clyde Shiraki Bobby Picker i
- Chief Examiner
- 3. Examination Review Meeting At the conclusion of the written examinations, the examiners provided
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Sequoyah Training Staff with a copy of the written examination and answer key for review. The NRC resolutions to facility comments are listed j below. !
R0 Exam (Applicable SR0 Question Numbers in Parentheses) I
- 1. Question 2.10 Agree. The recommended responses will be included as additional correct answers.
- 2. Question 2.16 (6.06) Facility comment acknowledged. The question will be graded based on the candidates assumptions.
- 3. Question 3.11(6.10) Agree. The facility is reminded to provide l technically correct reference material .
The answer key will be changed as ,
required. 1 Post Examination Review Changes: j R0 Exam
- 1. Question 1.10 Following post examination review by NRC graders, it was determined that a typographical error existed. The answer key will be changed to reflect "d" as the only correct answer.
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4 2.--Question 1.16(a.2) Following post examination review by ' NRC.
graders, -and .with follow-up ' with the facility it has. been determined that no definitive answer exists' for-. the given-conditions. This question will be deleted j
from the examination 'and the- total point.
value adjusted. .l Following post examination review by' NRC
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3., Question 1.16 (b)-
graders, it was determined that .this l question was a double. jeopardy 1 question. l The question will ' e deleted from the l examination and the total point value adjusted.
- 4. Question 1.20 Following post examination review by NRC graders, it was' determined that if the assumption that the rods were considered to withdraw . as far as the "All Rods Out Position" full credit would be allotted. The answer key will be changed to reflect this..
- 5. Question 2.05(6.02) Following post examination review by NRC graders, and with facility concurrence, it was determined that "c"'is also a correct answer. The answer key-will be changed to allow either "a"' or "c" to be correct.
- 6. Question 2,06 Following post examination review by ' NRC graders, and with concurrence with the facility, the correct answer is "a". The answer key will be changed to reflect this.
- 7. Question 2.07 Following post examination review by NRC graders, it was determined that the alternate answer "to relieve excess Reactor Coolant during heat up" was also acceptable. The answer key will be changed to reflect this.
- 8. Question 2.09(6.03) Following post examination review by NRC graders, the following will be added as additional correct answers: Containment Atmosphere Upper, Containment Atmosphere Lower,ContainmentSump(ReactorDrain) 1
3 9.- Question 2.11 Following post examination reviewL by NRC-graders, the following will be added as additional correct answers per L A01-6:
-1. Containment . Particulate Radiation-Monitors, Containment Total Gas Radiation
- l. Monitor; 2. Containment Pressure Increase.
and 3. Pressurizer Level.
.10. Question 2.14(a) (6.05) Following post examination : review by NRC graders, it wasi determined. that the r following- additional answers will be accepted as correct. 1. Pressurizer Low Pressure; 2. Pressurizer. High . Water Level;
- 3. Low Reactor Coolant Flow; 4. Reactor Coolant Bus Undervoltage.
- 11. ' Question 2.17 Following post examination review by NRC graders, partial credit will be redistributed based on facility discussions.
- 12. Question 2.18 Following post examination review by NRC graders, it was determined that tolerances would be allowed. The answer key will be changed to reflect this criteria.
- 13. Question 3.10 Following post examination review by NRC graders,' two additional answers will be added as correct answer: 1. Loop Low Flow -
Trip and 2. Turbine Trip.
- 14. Question 3.11 (6.10) Following post examination . review by NRC graders, . individual components started by the Safeguards Sequencer will be accepted as additional correct answers. The answer key will be changed to reflect this.
- 15. Question 3.14 Following post examination review by NRC graders, the following additional correct answers will be accepted under the category of No Lockout relay: 1. Shutdown board delta current or over current; 2. DC circuit problem; 3. Normal Feeder Breaker Over Current Relay. The answer key will be changed accordingly.
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- 16. . Question 3.17. Following post examination review' by NRC y graders, with facility concurrence it has been determined that the question and answer are inaccurate due to the interlock bistable setpoint. .This question will be deleted i- from the examination and the total 1 point.
l value adjusted.
- 17. Question 4.08(d) Following post examination review by NRC graders, it was determined that the answer should be 4. The answer key will be' changed accordingly.
- 18. Question 4.11 Following post examination review by NRC graders, partial credit will be redistributed, see examination for details.
- 19. Question 4.17 Following post - examination review by NRC graders, and with concurrence with the facility it has been determined that only one of the three required answers is correct.
The answer key will ta changed to reflect this and the point value of the question reduced accordingly.
One of the three changes (33%) made to the answer key following exam administration was due to incomplete reference material submitted to the NRC for exam development.
The total number of facility post examination comments was greatly reduced over previous exams due to a pre-exam review conducted by the facility.
However, following post examination grading, 19 questions on the R0 exam and 4 on the SR0 required answer key modifications, based on additional information supplied by the facility. The facility is reminded that a thorough pre-examination' review produces a better examination and precludes excessive post examination answer key modifications.
Several of the pre-exam review changes were due to inadequate reference material. The facility is encouraged to keep facility training material current with actual plant configuration.
- 4. Exit Meeting At the conclusion of the site visit the examiners met with representatives of the plant staff to discuss the results of the examination.
The facility recognized the benefits of pre-exam review of the written
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examination before the administration, and expressed enthusiasm over continuing this process.
The examination team expressed concern over simulator inadequacies and modeling prob'lems (See Enclosure 4). Problems with the exam material were addressed.
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There were no generic weaknesses noted during the oral examinations.
The cooperation given to the examiners and the effort to ensure an atmosphere in the control room conducive to oral examinations was also noted and appreciated.
The licensee did not identify as proprietary any of the material provided to or reviewed by the examiners.
.- 4 U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: SEOUOYAH 1&2 REACTOR TYPE: PWR-WEC4 DATE ADMINISTERED: 88/12/12 EXAMINER: REGION II I l
CANDIDATE I
INSTRUCTIONS TO CANDIDATE:
Use examination paper for the answers. Write answers on one side only, beneath the question. Points for each question are indicated I in parentheses after the question number. The passing grade requires at least 70% in each category and.a final grade of at least 80%. Examination papers will be picked up six_.(6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY
- 29. c o ft 30.00 ^" 24.90"' 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
- 25. v1 30.00 94-90-*$ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY GYSTEMS 28.so 55 30.00 4 2 4 .110# 3. INSTRUMENTS AND CONTROLS L% sc 24
-30.50^" 25.Gi#' 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 114.00 ft&-tW-M % Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.
Candidate's Signature MASTR:0PY i
NRC RULES AND GUIDELINES FOR. LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
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- 1. Cheating on the examinatin 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 time you START and STOP on the cover sheet of the examination.
- 6. Use only the paper provided for answers. l
- 7. Print your name (or initials) in the upper right-hand corner of each page of the exam.
- 8. The exam has one question per page. Write the answer beneath the question. Write only on one side of the exam and any extra sheets.
- 9. Number each answer continued on additional naper as to category and number, for example, 1.4 or 6.3.
- 10. 1 Attach continued answers to the back of the question to which it applies.
- 11. Place completed answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly used in facility ,
literature. Avoid using symbols such as < or > signs to avoid a simple transposition error resulting in an incorrect answer. Write it out.
- 13. The point value for each question is indicated in parenthesis after the question number and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an 'j answer to mathematical problems whether indicated in the question l 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.
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.17. Proportional grading will be applied. Any. additional wrong information that is provided will count against you. For example, p if a question is worth one point and asks for four responses, each L of which is worth 0.25' points, and you give five responses, each of your responses will be worth 0.20 points. If one of your five responses is incorrect, 0.20 points will be deducted and your total credit for that question will be 0.80 instead of 1.00 even though you got the four correct answers.
- 18. After the examination has been completed, you must sign the statement on the cover sheet.that indicates that the work is your I own and you have not received or been given assistance in L completing the examination. This must be done after you complete the examination.
- 19. When you complete your examination, you shall:
- a. Assemble your examination as follows:
(1) Exam questions / answers on top.
(2) Exam aids - figures, tables, etc.
.(3) Scratch pages used during the course of-the exam.
- b. Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn the balance of the paper that you did not use for ,
answering the questions.
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- d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OEERATION1 Page 2
' THERMODYNAMICS , HEAT TRANSFER _AND FLUID FLOW l
l QUESTION 1.01 (1.00) l l Complete the following statement concerning shutdown margin.
The actual shutdown margin DOES NOT ...
- a. ensure the reactor will be maintained sufficiently subcritical in the shut down condition at EOL with a double ended main steam i line break with Tavg at a no load operating temperature.
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- b. provide for allowing the reactor to be made subcritical from all normal operating conditions.
- c. ensure the reactivity transients associated with postulated accident conditions in the FSAR are controllable within acceptable limits.
r ~d . remain constant if the boron concentration changes, Tavg
\'s changes, or as fuel depletion occurs over core life.
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TAEEE66YUERIC5~~55ET TRd55EER An5 EC0io FE6w QUESTION 1.01 (1.00)
Complete'the following statement concerning the Technical Specification, SHUTDOWN MARGIN requirements.
CHUTDOWN MARGIN requirements DO NOT. . . . .
- a. ensure the reactor will be maintained suf ficiently suberitical in the shut down condition at EOL with a double ended main steam I
line break with Tavg at a no load operating temperature.
- b. provide f or allowing the reactor to be made subcritical from all normal operating conditions.
- c. ensure the reactivity transients associated with postulated accident conditions in the FSAR are controllable within acceptable limits, d*. vary throughout core life as a function of fuel depletion, ;
RCS boron concentration changes, and RCS Tavg.
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l QUESTION ~1.02 (1.00)
Indicate HOW (HORE NEGATIVE or LESS NEGATIVE), and WHY Fuel Temperature Coefficient changes over core life at 100%-power.
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1 QUESTION 1.03 (1.00)
During a reactor trip recovery, the initial 1/M data point was 1.0. l After a 1-hour delay, rod withdrawal was commenced. Upon stopping 1 l rod withdrawal to take 1/M data, you find that the second 1/M point' l L is-1.1. Which of one the following explains this increase'in the 1/M value?
- a. This is not possible, the RO must have made an error when taking count rate data.
l b. The buildup of Xenon during the 1-hour delay added more negative reactivity than the rod withdrawal had added in positive '
reactivity. f
- c. The source-detector geometry is incorrect.
d An inadvertent dilution is in progress. I 4
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o QUESTION 1.04 (1.00)
Which one of the following. statements describes the change in Moderator Temperature Coefficient (MTC) from BOL to EOL7
- a. The MTC becomes more negative due to increasing boron concentration, decreasing fission product inventory, and axial flux redistribution toward the edges of the core.
- b. The MTC becomes more negative due to decreasing boron concentration, increasing fission product inventory, and radial flux redistribution toward the edges of the core.
- c. The MTC becomes less negative due to increasing boron concentration, increasing fission product inventory, and axial {
flux redistribution toward the edges of the core.
- d. The MTC becomes less negative due to decreasing-boron j concentration, decreasing fission product inventory, and axial !
flux redistribution toward the edges of the core.
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~ QUESTION 1.05 (1.00)
During a reactor startup under xenon-free conditions, rod withdrawal is stopped at the -0.02% delta k/k position and the count rate is allowed to stabilize. In regard to the response of the count rate'in the hour after stabilization, which one of the following statements is correct? (Assume NO further operator actions are taken.)
- a. Count rate will remain essentially constant.
- b. Count rate will rapidly decrease to its pre-startup level,
- c. Count rate will slowly decrease because it is subcritical.
- d. Count rate will slowly increase due-to long-lived delayed neutrons.
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QUESTION 1.06 -(1.00)
-Which one of the following statements describes xenon concentration behavior following a power increase?
- a. The migration length changes as moderator density decreases,
- b. Concentration decreases due to more thermal neutrons being available to burn out xenon.
- c. The amount of xenon from fission does not increase until the reactor is at a new stable power.
- d. The amount of Xe-135 produced from Sm-149. increases for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after power changes, i
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QUESTION: 1.07 (1.00)
Which one of the following is the correct value for the cooldown rate of a y saturated steam system, if the initial pressure is 985 psig and the pressure 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> later is 385 psig?
- a. 80 degrees F/ hour
~b. '100 degrees F/ hour
- c. 120 degrees F/ hour
- d. 125 degrees-F/ hour i
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. QUESTION 1.08 (1.00)
Which one of the following statements is correct concerning boron worth changes?
I a. Control rod withdrawal will cause boron worth to increase due to I decreased competition for neutrons.
- b. Boron worth decreases over core life.
l c. An increase-in fission product inventory will cause boron worth l to increase due to decreased competition for neutrons.
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due to increased competition for neutrons.
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QUESTION 1.09 (1.00)
'Which one of the following will contribute to a lower fuel centerline l temperature over core life.
- a. Crud buildup on clad.
- b. Fuel densification.
- c. Clad creep.
- d. Fission gases released to the fuel pellet-clad gap.
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QUESTION 1.10 '(1.00)
Which one of the following is a correct statement when adjusting the
-power range channels to 100% based on a calculated calorimetric?
- a. If the feedwater temperature used in the calorimetric l calculation was lower than the actual feedwater temperature ;
actual power will be higher than indicated power.
- b. If the steam generator blowdown flow is increased prior to adjusting indicated power but after the calorimetric data was taken, actual power would be lower than the indicated power.
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- c. If the steam flow used in the calorimetric. calculation was lower I than actual steam flow, actual power will be less than indicated-power and is not an input to calorimetric calculation.
- d. If the reactor coolant pump heat input used in the calorimetric calculation was neglected, actual power will be less than indicated power.
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QUESTION 1.11 (1.00)
Which one of the following actions would help, rather than hinder, natural circulation?.
a Lowering steam generator level
- b. Lowering RCS pressure
- c. Increasing RCS temperature j
- d. Increasing pressurizer level l
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QUESTION 1.12 (1.00)
The reactor is operating at 100% power, all-rods-out, near the end'of cycle prior to a scheduled power reduction to 50% power for surveillance. The Unit Operator observes AFD (delta I) to be in the doghouse and decides to lower power and temperature by borating, while leaving rods fully withdrawn. Actual Tavg follows the programmed Tavg.
Which one of the following BEST describes the AFD initial change?
- a. AFD will change in the Negative direction; because the relatively more Negative reactivity is added to the top half of the core.
- b. AFD will change in the Negative direction; because relatively more Negative reactivity is added to the bottom half of the core.
- c. AFD will change in the Positive direction; because relatively more Positive reactivity is added to the top half of the core.
- d. AFD will change in the Positive direction; because relatively more Positive reactivity is added to the bottom half of the core.
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QUESTION 1.-13 (1.00)
In each of the following statements, SELECT the bracketed [] word that will make the statement correct.
- a. As Keff approaches unity, a.[ shorter / longer] period of time is required to reach the equilibrium neutron level for a given change in'Keff.
- b. As Keff approaches unity, a [ larger / smaller] change in. neutron level results from a given change in Keff.
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' QUESTION ,1.14 (1.50)
List the THREE reactivity coefficients which. makeup the Power Defect.
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QUESTION 1.15 (2.00)
A beginning-of-life core has been operating with all rods fully withdrawn at 100% power for 1 week. The bank D-control rods are inserted 100 steps with sufficient boron dilution to offset the reactivity added by the rods. Answer the following questions about the resultant xenon transient IN THE RODDED REGION OF THE CORE.,
- a. How does the xenon level change at first (INCREASES,' DECREASES,.
STAYS THE SAME)? (0.50)
- c. The final steady-state xenon concentration will be (LESS THAN, GREATER THAN, EQUAL'TO) the original concentration. (0.50)
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l OD QUESTION 1.16 (-1-5fr)#
l a. At normal hot standby conditions, in which direction will the pressurizer level indication change as a result of the following I
transients? (INCREASES, DECREASES, STAYS THE SAME).
Consider each transient separately. Assume the letdown and charging flows are equalized and the pressurizer level control system is in manual.
! 1. The reference leg heats-up from 120 F to 200 F due to the I
relocation of a ventilation duct. (0.50)
- 2. The pres d'~r e heaters fail and d essurizer ater ols from rjor operating tempera o 590 F. 0.50)
- b. F 2 above, followi he cooldown, is t d level REATER THAN, LESS THA EQUAL TO the a g i level?
(0.50) 4 1
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QUESTION) 1.17 (2.00)
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- The plant is operating a 30% power when one reactor coolant pump trips i due to an electrical fault. The control. systems are all'in automatic. j Bank D. rods are initially at.165 steps. Assume the plant does not trip.
Indicate whether the following parameters INCREASE, DECREASE or REMAIN THE SAME.
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- a. Turbine power
- b. Reactor power (final) l
- c. Final. rod height 1
- d. T-avg (affected loop)
- e. T-avg (unaffected loop)
- f. Core Delta-T
- g. Delta-T (affected-loop)
- h. Delta-T (unaffected loop) i l l l
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QUESTION 1;18 (1.50)
For each condition in COLUMN A find the correct heat transfer equation in COLUMN B that would be used to calculate the heat transferred.
-NOTE: ' Answers in COLUMN B maybe used more than once.
COLUMN A COLUMN B
- a. Across the reactor 1. k=UAdeltaT j (cold leg to hot leg) . .
- 2. Q = m delta T
- b. Across S/G U-tubes . .
(primary to secondary) 3. Q = m epdelta T
- c. Across S/G secondary 4. h=UAdeltah (feedwater to steam) . .
- 5. Q = m delta h l 4
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QUESTION 1.19 (2 00)
For each of the following, state WHICH condition given would produce the HIGHEST differential rod worth,
- a. Control rod is inserted adjacent to an inserted control rod OR adjacent to a withdrawn control rod.
- b. At BOL a control rod is inserted near the center of the core OR near the core periphery.
- c. Control rod is inserted into the core at a boron' concentration of 900 ppm OR a concentration of 600 ppm.
- d. Control rod is inserted into the core with RCS temperature at 540 F OR at 500 F.
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QUESTION 1.20 (1.50)
During a power ramp to 50 percent power with rod control in automatic an incorrect boron change was calculated and made which resulted in the plant stabilizing et the desired power but with control rods at the all out position and Tavg 5 degrees F below the target value. Given the following initial parameters, PROVIDE the final RCS boron concentration needed to INCREASE Tavg by 5 degrees F while returning control rod bank D to the 188 step position. Assume turbine power stays constant at 50 percent. Show all calculations.
Initial RCS boron concentration = 600 ppm j Total power coefficient = -20 pcm/ percent Moderator temperature coefficient = -15 pcm/ degree F Differential boron worth = -10 pcm/ ppm Control rod worth avg. (5-80) = 8.60 pcm/ step (95-170) = 4.16 pcm/ step (185-228) = 1.095 pcm/ step l
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QUESTION 1.21 (1.50) ,
I Unit 2 is still near the beginning of its fuel cycle; Unit 1, an identical reactor and fuel loading scheme, is near the end of its fuel cycle. Assume both plants have just started up following a 3-week shutdown period,
- a. Critical data has just been taken at 10 E-8 amps at both plants and the operators have added small, equal amounts of reactivity to continue the power. ascension. Which plant, Unit 1 or 2, will have the HIGHER steady startup rate from this egual reactivity insertion and EXPLAIN WHY? (0.75)
- b. Shortly after 50 percent power is reached during these startups, rod control is placed in manual at both plants.. Shortly afterward, a shutdown bank control rod worth -150 pcm drops into the core at both plants. Assuming that no operator action is taken for these casualties.and that neither reactor trips. Which plant, Unit 1 or 2 will stabilize with the HIGHER steady-state Tavg and EXPLAIN WHY? (0.75) 4 l
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QUESTION 1 22 (2.00) 1 The plant is in Hot Standby with the RCS pressure being maintained at
'985 psig. A pressurizer PORV is slowly discharging to the pressurizer l ' relief tank which is at 5 psig. The steam quality in the pressurizer l steam bubble is 100 percent.
- a. What is the enthalpy of the fluid entering the PRT7 b What is the temperature os.. the tail pipe downstream of the PORV?
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.. . QUESTION 1.23 (1.50)
Answer the following questions in reference to subcooling margin of the plant.
- a. What'is the subcooling margin of the plant if the following <
conditions exist': )
Thot = 587 F. Tavg = 572 F Tcold = 557F Ppar = 2235 psig Psg = 1033 psig
- b. If power is raised from 50% to 100%, why does the'subcooling margin decrease?
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- 2. Eh&NI_ DESIGN INCLUDING SAFETY AND EMERGENCY Page 25
'SYSTEME I
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h QUESTION 2.01 (1.00) l l Which one of the following provides the correct reasons for maintaining l a minimum spray bypass flow to the pressurizer?
- a. Prevent excessive cooling to the surge line.
Reduce the delta pressure across the spray valves.
I b. Reduce thermal shock to the spray no==le.
Ensure that the backup heaters cycle on.
- c. Prevent excessive cooling to the spray line.
Equalize boron between pressurizer and the RCS.
- d. Minimize stress to the surge line thermal sleeve.
Remove gases from the RCS.
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r -- . _ _ _ . - . _ _ _ . - - - - .
. QUESTION, -2.02 (1.00)
.The Component' Cooling Water system in conjunction with the RHR system is
' designed to reduce the RCS temperature to F within hours after shutdown to mode 4 conditions.
a, '120, 24 l
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- b. 140, 16
- c. 250, 18
- d. 330, 16
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QUESTION 2.03 (1.00)
For EACH of the following components, indicate whether they will receive an OPEN, CLOSE, or NO SIGNAL, as a result of a safety injection initiation signal with Phase "A".
- a. Main feed bypass valves
- b. Cold Leg accumulator discharge isolation valves
- c. . Component cooling isolation from letdown heat exchanger
'e. Seal water return isolation valve f -. Component cooling isolation valve from RHR system
- h. Normal charging header isolation valves
- 1. Steam supply valves to turbine-driven feed pump
- j. Control room supply ducts
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! l L QUESTION 2.04 ( l '. 00 )
L l Which one of the following is a reason for the Steam Generator level
{ program?
L a. Facilitates chemical control by minimizing steam generator mass L at all power levels,
- b. Controls turbine blading erosion by allowing a specific amount-of carryover.
- c. Controlling steam generator level following a load rejection such that inadvertent low-low level trips are min; sized,
- d. Termination of cooldowns of the Reactor Coolant System from steam breaks inside and outside containment.
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QUESTION 2.05 (1.00)
Which one of the following design features enhances the heat removal operation of the ice condenser and containment spray system?
- a. Pressure operated doors open to allow upper containment air to flow through the lower containment.
- b. Air return fans provide flow to return the air from the upper containment to the lower containment.
- c. Containment design, such that the delta P between upper and lower containment drives the air circulation.
- d. Ventilation coolers and recirculation fans are used to mix the air and provide additional cooling.
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QUESTION 2.06 -(1.00)
Which one of.the following is the method used to protect the Upper Head Injection System.from overpressurization?
- a. Relief valve
- b. Surge tank
- c. Rupture Disk
- d. Pressure Control valve k
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QUESTION 2,07 (1.00)
'What are the TWO reasons for using the Excess Letdown system? j I
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' QUESTION 2.08 '(1.50)
State the EE water sources to the Turbine' Driven Auxiliary Feedwater-Pump. Simila.r anroac
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QUESTION 2.16' (1.00)'
Briefly explain.the effect on pressurizer level indication following a {
break in the bellows separating the reference-leg fluid from the -
- s. pressurizer fluid. INCLUDE your reasoning for the effect. Assume steady state power and level conditions.
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QUESTION 2.17 (1.50)
~
What is the design basis for each of the following Pressurizer pressure
- operational setpoints or control' points?
- a. Normal operation pressure - 2235.psig.
- b. Low Pressure SI setpoint - 1870 psig,
- c. High Pressure setpoints - 2335 and 2385 psig.
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QUESTION ~ .2.18 (2.00)-
Provide the RCS pressure and system flowrate/ volume at which each of the following Emergency Core Cooling Subsystems will inject water into
'the'RCS injection point.
- a. Safety. Injection pumps.
- b. Cold Leg Accumulators.
- c. ' Centrifugal Charging pumps.
- d. Residual Heat Removal pumps.
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QUESTION 2.19 (2.00)
.W hich major system has a penetration to the-Reactor Coolant System located BELOW the piping center-line and WHY was this location chosen?
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- _ _ _ _ _ _ _ - _ = _ __ - 8
l QUESTION 2.20 (1.50) l l
Answer the following concerning the long term standby operation of the Auxiliary Feedwater System (AFW)
- a. Which component (s) of the AFW system are the most susceptible to i
problems which can result in steam vapor binding of the pumps?
- b. In accordance to a TVA memorandum, what action is required to to minimize the probability of steam vapor binding from occurring?
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e ~ QUESTION. 2.21 (1.50)
What motive forces are used for Main Steam Isolation' Valve:
i~ a. :for Opening?
'b. for Fast Closing?
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(***** END OF CATEGORY 2 *****)
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INSTRUMENTS AND CONTROLS Page 46' i (1.00)
QUESTION 3.01 Which one of the following is correct concerning the operation of the Reactor Trip (RT) and Reactor Trip Bypass (BY) Breakers?
- a. The Train B SSPS trip signal directly trips RTB and BYB.
- b. The Train A SSPS trip signal directly trips RTA and BYB.
- c. Tripping is accomplished by an undervoltage relay, normally held open by 48 volt de power from the logic panels.
- d. To allow testing of the RTs, BOTH BYs may be closed while the reactor is at power.
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QUESTION.. 3.02 ( 1 -. 0 0 ) . j i
Which one of the following Rod Control subsystems provides the function ]
of dampening the mismatch channel output at low rates of change, j allowing the Tavg/ Tref channel to control?
- a. Variable gain unit
- b. Rate comparator l
- c. Non-linear gain unit
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- d. ' Reactor Control unit
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QUESTION 3.03 (1.00)
Which one of the following describes the no> val status of the Steam Dump System with the unit at 100% steady state power?
- a. In the Tavg mode with the reactor trip controller selected to effect steam dump valve operation upon a reactor trip signal when Tavg increases to > or : high Tavg.
- b. In the Tavg mode wzth the load rejection controller selected and having any error signal between Tavg and Tngo-load present and sufficient to effect steam dump valve oper,af3on immediately upon receipt of a load rejection arming sigral.
- c. In the steam pressure mode with load rejection controller selected to effect steam dump valve operation upon receipt of an error signal between Tavg and Tgo-load of > or = 5 F.
Avg
- d. In the Tavg mode with the load rejection controller selected to effect steam dump valve operation upon receipt of an error signal between Tavg and Tref > or: 5 F if a load rejection arming signal is supplied.
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E_
' QUESTION 3.04 (1.00)
Which one of'the following electronic components would be~found in the INTERMEDIATE RANGE nuclear instrument channel?
.a. Two loss of voltage alarm functions per' channel
- b. ~ Preamplifier
- c. Summing and level amplifier
- d. Pulse shaper i
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i U___.-____-_--...__-.-. 1
QUESTION 3.05 (1.50) l State the THREE signals used to develop the Control Rod Insertion
- Limits?
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__. _..-ma-__ m_m.___---... __ m - -
QUESTION 3.06 (2.00)
TRUE or FALSE Indicate your response-(True or False) to each of the following concerning the 6 9 kV shutdown boards. Answer each separately,
- a. The alternate and normal feeder breakers sense each others position and both will trip if both breakers indicated shut at the same time.
- b. The 1A-A 6.9 kV Shutdown Board is fed by its normal supply bus of the 6.9 kV unit board 1B.
- c. An 86 relay actuation will lockout the normal feeder breaker, Diesel Generator breaker, Util1ty breaker and select the alternate feeder l
to close,
- d. The blackout relays BOX and BOY are reset prior to paralleling the Diesel Generator to the Unit board to prevent the DG from overloading after shutting the Unit board feeder.
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-QUESTION 3.07 (1.00)
Which one of the following is correct concerning the operation of the Emergency Core Cooling System equipment?
A Safety' Injection Signal will...
- a. start the charging pumps, SI pumps, RHR pumps, and Containment Spray pumps.
- b. initiate a reactor trip, isolate the feedwater system, and start the ERCW pumps.
- c. start the diesel generators and align them to the shutdown boards.
- d. open the valves in the boron injection tank recirculation lines.
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QUESTION- 3.08 '( 2. 00 )
J. List the FOUR. opening. interlocks for the letdown orifice isolation valves. NOTE: Similar items count as one interlock.
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QUESTION 3.09 (1.00)
State the.TWO signals _that will cause valves LCV-62-135 and'LCV-62-136, RWST to CCP suction, to open?, Setpoints.are not required.
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.c QUESTION 3.10 (1.50)
List' reactor, trips which are BLOCKED or DISABLED (manual or automatically) during a startup when reactor power is at approximately 12%. Do not include setpoints or coincidences.
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l QUESTION 3.11 (2.50)
\
1 l 'As a result of a Safety Injection signal, certain signals generated from I the SI will " latched in" (" Seal in"), while others will automatically J l reset or clear.
l List FIVE signals which are " latched in" (" Seal in") as a result of the SI l signal.
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QUESTION 3.12 (2.00) i List the TWO interlocks and permissives associated with the Steam Dumps L which must be satisfied for the availability of the system. Include in l
your answer'setroints, coincidence and reason for the interlock or
-permissive.
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) ;
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QUESTION 3.13 (1.50')
-List the THREE signals that will automatically start a standby Component Cooling Water pump. Setpoints are not required.
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QUESTION 3.14 (2.00)
List-EIGHT interlocks / conditions that must be satisfied for the Emergency Diesel Generator breaker to automatically shut following a degraded voltage or undervoltage condition.
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- QUESTION 3.15 (2.00)
Answer the following concerning the Main Feedwater System.
- a. List THREE Main Feedwater isolation signals. Setpoints are not required. (1.5)
- b. What automatically happens as main feed pump suction pressure approaches saturation (low Net Positive Suction Head)? (0.5)
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QUESTION 3.16 ~(3.00) l List the SIX control interlocks which will stop outward Control Rod l P i motion. INCLUDE the coincidence, origin and setpoint of the input signal and whether it is for manual and/or automatic rod ~ control.
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QUESTION 3.17 (1.50).
During a slow pressure: increase tr ent PORV 340A automatically opened to reduce the pressure but PORV 4 does not open. EXPLAIN the reason WHY.PORV 334 did not open.
ASSUME that both valv functioned correctly, neither are stuck and the absolute setpoints each valve are identical.
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1 QUESTION- 3.18 ._( 1.50)
WHAT are the THREE p>>tential causes of the Reactor - Turbine trip.due to failure of the 125 V DC vital battery board I, according to AOI-21.1,
" Loss of 125 V DC Vital Battery Board I?" 1
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i-QUESTION 3.19 (1.00)
Briefly describe HOW the Auxiliary Feedwater system is prevented-from feeding a faulted Steam Generator.
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- 4. PROCEDURES - NORMAL. ABNORMAL, EMERGENCY Page 65' AND RADIOLOGICAL CONTROL QUESTION 4.01 (1,00)
Which one of the following is NOT correct concerning the use conventions associated with EOP's?
- a. Even after a transition to another procedure, the steps started BEFORE the transition was made must still be completed, but not to delay the transition.
- b. Continuous warnings contained in a caution (step) are NOT in effect when an operator is referenced to a procedure to be performed concurrently with the EOP in effect.
- c. If a caution statement occurs before step one of an EOP it may apply either to the whole procedure or just to the first step.
- d. Unless otherwise specified, a required task need not be fully completed before proceeding to the next instruction; it is enough to begin the task and have some assurance that it is progressing satisfactorily.
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. QUESTION 4.02 (1.00)
With a fire in the auxiliary building, according to AOI-30, Plant Fire, Which'one of the following is the reason for transferring control of l the fire pumps to the shutdown board by placing the transfer switches to 1
.the auxiliary position?
- a. to regain control of the pumps if the fire is around the RCP relay board on Unit 1.
- b. the shutdown board is the desired location for control of the pumps.
- c. to ensure automatic control transfer does occur if circuits are jeopardized by the fire.
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- d. to place fire pumps in a continuous operating mode thus ensuring adequate fire water pressure.
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lQUESTICN 14 .'03 (1.00)
From the following, select the one situation uhat has the highest priority:for operator response,
- a. Core Cooling - Yellow
- b. Containment Red
- c. Suberiticality - Orange
- d. Heat Sink - Orange
- e. Inventory - Red
- f. Pressurized Thermal Shock - Orange
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
i QUESTION 4.04 (1.00) l l
.Which one of the following symptoms would require the initiation of a !
manual reactor trip AND safety injection if neither had occurred automatically?
- a. Containment pressure = 1.0 psig
- b. General Warning alarm on the Solid State Protection System B
- c. Pressurizer pressure = 1850 prig, pressurizer level = 40%
- d. Power = 33%, and loss of flow in one loop
- e. Power = 45%, pressurizer level 93%
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i QUESTION 4.05 (1.00)
Which one of the following symptoms would require initiation of ONLY a manual reactor trip if neither a reactor trip or safety injection had automatically occurred?
- a. Containment pressure = 1.0 psig
- b. General Warning alarm on the Solid State Protection System B
- c. Pressurizer pressure = 1850 psig, pressurizer level = 40% !
- d. Power = 33%, and loss of flow in one loop
- e. Power = 45%, pressurizer level 93% 1 i
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QUESTION. 4~.06 (1.00)
Which.one of the following would be the correct use of E-0,
" Reactor Trip or Safety Injection" with a loss of off-site power and failure of the Emergency Diesels to recover the vital buses?
i a.
Go immediately to ECA 0.0, Loss of al] A/C power prior to entering E-0.
- b. Go immediately to ECA 0.0 from E-0 after verifying reactor and turbine trip. !
- c. Go immediately to ECA 0.1, Loss of All A/C power recovery without SI required.
- d. Complete E-0, Reactor Trip or Safety Injection IMMEDIATE ACTIONS then go immediately to ECA 0.0, Loss of all A\C power.
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- _ _ _ _ _ _ _ _ _ . _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _____m ._
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-QUESTION- 4.07 (1.00)
MULTIPLE CHOICE When are the procedures' required by the Critical Safety Function Status Trees implemented during ECA-0.0, Loss of All AC Power, according to
.the Westinghouse Background Information Users Guide for the Functional Restoration Guides?
- a. Never implemented when in ECA-0.0
- b. Upon entry to ECA-0.0
- c. Upon reaching step 5 of ECA-0.0
- d. Upon exiting to subprocedures of ECA-0.0 l
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- c. .
QUESTION 4.08 (2.00) i Match the whole body radiation exposure terms in Column A to their. limit in Column B. Some answers in Column B could be used more than once.
Column A Column B
- a. 10CFR20 limit /qtr without NRC Form 4 1. 0.3 REM
- 3. 1.25 REM
- c. 10CFR20 limit /qtr with an NRC Form 4
- 4. 3.0 REM
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QUESTION 4.09 (1.00)
Which.one of the following conditions would REQUIRE fuel shuffle operations to be immediately stopped? Assume the initial nucleus of ten assemblies are loaded, and exclude ANTICIPATED change in count rates due to detector and/or source movement.
An increase in count rate by a factor of:
Any
- a. 2 on'ALL responding nuclear channel or by a factor of 1.5 on 4tdALL nuclear channels during any single loading step.
Aav
- b. 3 on-ALL responding nuclear channel or by a factor of 1.5 on ALL nuclear channels during any single loading step.
ANV
- c. 4 on ALL responding. nuclear channel or by a factor of 2 on ALL nuclear channels during any single loading step.
AH V
- d. 5 on ALL responding nuclear channel or by a factor of 2 on ALL nuclear channels during any single loading step.
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QUESTION 4.10- (2,00)
What are the FOUR conditions which must be met to open the RCP #1 seal bypass-valve if either a high pump bearing temperature or seal leakoff temperature alarm occurs according to SOI 68.2, Reactor Coolant 1 Pumps?
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
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. QUESTION 4.11 (2.00)
List the'TWO immediate actions for a steam generator tube LEAK according to AOI-24,- Steam Generator Tube Leak. . Include both the IF and THEN actions.
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QUESTION 4.12 (2.00)
Provide FOUR-of the five symptoms (conditions).where Emergency Shutdown is required according to AOI-32, Emergency Shutdown.
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. QUESTION 4.13 (1.50)
List the immediate actions for AOI-15A.1, Loss of. Component Cooling Water - Loss of Miscellaneous Equipment Header and Loss of Reactor Building Header. Assume there was.out leakage from the Component Cooling Water System. Include setpoints where applicable.
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. _ _ _ _ _ _ _ _ - . _ . = _ _ _ _ _ - _ _ _ _ _ __.
QUESTION 4.14 (2.00)
List the FOUR. normal (NOT ADVERSE containment) containment Safety-Injection termination criteria found on-the EOP Foldout page.
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-QUESTION 4.15 (1.50)
- i. List ~the FIVE checks which would be made to identify a RUPTURED Steam.
Generator as outlined in E-3, Steam Generator Tube Rupture.
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QUESTION 4.16 (1.50) l List THREE critical rod height limits as described in GOI-2, " Plant l: Startup from Hot Standby to Minimum Load."
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y QUESTION ~ 4.17 pt'. 50 )
During execution of AOI-27, Control Room Inaccessibility, when preparing to close the MSIVs, WHY are the SG PORVs. opened by lowering the controller setpoint enough to allow closure of the steam dumps?
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QUESTION- 4.18 (1.00)
Explain WHY any attempt to realign a dropped rod should be coordinated with Reactor Engineering according to a CAUTION statement in AOI-2, Malfunction of' Rod Control System, section C-Dropped Rod.
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p.- . ..
L QUESTION 4.19 (1.00)
- l. -
l State the precautions taken with both' safety injection pumps and the non-operating centrifugal charging pump to prevent overpressurization of the RCS due to an inadvertent safety injection signal according to GOI-1, Plant Startup from Cold Shutdown to Hot Standby.
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QUESTION- 4.20 (1 00)
.WHAT action-is required concerning the use of radiological' protective
- clothing per Radiological Control Instruction-RCI-1, Radiological Control Program, prior to donning the radiological protective
' clothing for entry'into contaminated areas? !
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. QUESTION 4.21. (2.00) j What is meant by the term "ALARA ZONE" and when is such a zone used?
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l QUESTION 4.22 (1.50)
Prior to a reactor startup, with the RCS at normal operating pressure and -
L . temperature, the following RCS11eakages exist. For each leak listed below, state whether you could STARTUP or would have to remain SHUTDOWN.
. Treat each case as an independent event.
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- a. A' leak from an unknown source of 1 5'GPM.
- b. 6.0 GPM from a manual valve packing gland.
- c. 0.4 GPM from one steam' generator i
(***** END OF CATEGORY 4 *****)
(********** END OF EXAMINATION **********)
/
EQUATION SHEET f = at v = s/t 2 Cycle efficiency = Net Work (out) - --
w = ms s = v,t + at 2
E = aC .
, , (,I , v ')jt 2 -lt EE = uv vg A = AN A = A,e
= v, + at PE = agh w - e/t A = in 2/tq = 0.693/tg W = v&P- ,
AE = 931Am eq(eff) = (e,,)(ty)
< . g) k = M pAT ,
I . g 4X k=UAAT I . r -ux ' '
' ' Pur = wi i , z.z to -x/Tvt P=P 10 5UR(t) . TVL = 1.3/u t
P=P e /T HVL
- 0.693/u
~
SUR = 26.06/T ,
T = 1.44 DT SCR = S/(1 - K,gg) fA
- SUR = 26 g fph, CR x = S/(1 - K,ggg)
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T = (1 /p ) + [(f _* p)/ A,gg ] p eff}l = CR2(1 ~ Xeff)2 T,= 1*/ (p - 7) M " I/(1 ~ Keff) = CR /CR0 g T = (I - p)/ A,ggo g , (l ,K gg) j(l ,K,gg) 8 " IEeff ~I)I aff " #eff/Kaff SDM = (1 - K,gg)/K,gg
~
p= (1*/TKygg.] + [I/(1 + A,ggT )] 1* = 1 x 10 ' seconds
~
P = I$V/(3 x 10 0) A,ggA= 0.1 seconds I = No Idgg=Id22 UATER PARAMETERS Idg =Id2 1 gal. = 8.345 lba R/hr = (0.5 CE)/d 2 (,,,,,,)
1 gal. = 3.78 liters R/hr = 6 CE/d (feet) -
1 ft3 = 7.48 gal. MISCELL\NEOUS CONVERSIONS ,
Density = 62.4 lbm/ft 3 1 Curie = 3.7 x 10 dps 10 Density = 1 gm/cm i kg = 2.21 1ha Heat of vaporization = 970 reu/lba 1 hp = 2.54 x 10 3BTU /hr Heat of fusica = 144 Btu /lba 1 Hw = 3.41 x 10 Btu /hr 1 Atm = 14.7 psi = 29.9 in. I'g. 1 Btu = 778 f t-lbf 1 ft. H O = 0,4333 lbf/in 2 1' inch = 2.54 cm F = 9/5 C + 32 ,
- C = 5/9 ( F - 32)
l '. PRINCIPLES OF NUCLEAR POWER PLANT OPERATIONt Page 87
< TEERMODYNAfflgS. HEAT TRANSFER AND-FLUID FLOW
. ANSWER .1.01 (1.00) d REFERENCE SNP Question-and Answer bank, Cert week 13, 6/6/86 192002K114 ..(KA's)
ANSWER 1.02 (1.00)
Le:m SNsMORE NEGATIVE [0.5] - due to lower effective fuel temperature at EOL
-[0.5]
REFERENCE SQN Lesson Plan, EGT222.005' Lesson Objective, J 001000K549 ..(KA's)
' ANSWER 1.03 (1.00) b REFERENCE SQN Lesson Plan EGT222.005 Learning Objective B 192008K106 192008K103 001000K517 ..(KA's)
ANSWER 1.04 (1.00) b
(***** CATEGORY 1 CONTINUED ' A NEXT PAGE ***** )
.1. PRINC.lELES OF NUCLEAR POWER PLANT OPERATION. Page 88 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW REFERENCE SQN Prelicense Training Program lesson plan, week 1, Reactor Physics Review, Training objective V.H 192004K106 ..(KA's)
ANSWER 1.05 (1.00) a REFERENCE SQN Prelicense Training Program lesson plan, week 1,. Reactor Physics 1.'evi ew , Training objectives V.B and V.C 192008K104 ..(KA's)
ANSUER 1.06 (1.00) b' REFERENCE SQN Operator Certification Training Reactor Theory Core Poisons V, week 1-5, training objective V.A Lesson Plan EGT222.005, p. 23 Lesson Objective, Q 192006K106 ..(KA's)
ANSWER 1.07 (1.00) l l
b '
REFERENCE SQN Prelicense Training Program lesson plan, week 1, Reactor Physics Review, Para X l l
193003K125 ..(KA's) j (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
I !
l l
1 PRINCIPLES-OF NUCLEAR POWEB PLANT OPERATION, Page'89 THERMODYNAMICS HEAT TBANSFER AND FLUID FLOW ANSWER 1.08 (1.00) a REFERENCE L SQN Question and Answer Bank Lesson Plan EGT222.005, p.-22 Lesson Objective, N 192005K107 192005K105 001000K522 ..(KA's)
ANSWER 1.09 (1.00) e REFERENCE SQN Question and Answer Bank Lesson Plan EGT222.005, p. 17 Lesson Objective, I 001000K549 ..(KA's)
ANSWER 1.10 (1.00)
/ d. aqpri REFERENCE SNP Question and Answer Bank, Prelicense Final Exam, 12/10/86, sect. 1.0 193007K108 ..(KA's)
ANSWER 1.11 (1.00) d i
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
1 l 1. PRINCIPLES OF NUCLEAR POWER' PLANT OPERATION. Page 90 l*~ -
THERMODYNAMICS. HEAT TRANSFER'AND FLUID FLOW l
1.
l . REFERENCE.
SQN Natural circulation / inadequate core cooling Question and Answer Bank, Prelicense Exam - RO, 11/21/86, Q. 1.20 1.
i 193008K123 ..(KA's) l l ANSWER 1.12 (1.00) c i
REFERENCE t l
SON Question and Answer Bank Lesson Plan EGT 222.005 192005K111 ..(KA's)
ANSWER 1.13 (1.00)
(0.5 each)
- a. longer
- b. ' larger REFERENCE
{
SQN Question and Answer Book Lesson Plan EGT222.005, PP 7 & 8 Lesson Objective, C 192002K108 ..(KA's)
ANSWER 1.14 (1.50)
(0.5 each)
- 1. Moderator Temperature Coefficient
- 2. Fuel Temperature Coefficient (Doppler)
- 3. Void Coefficient
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) ,
f
-~____.m__ _ _ _ _ _ . . _
-1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Page 91
.- . THERMODYNAMICS, HEAT - TRANSFER AND FLUII) FLOW REFERENCE Westinghouse Reactor Physics I-5.p.26/27 SQN Lesson Plan EGT222.005 Learning Objective K 192004K108 ..(KA's)
ANSWER 1.15 (2.00) l
- a. INCREASES
ANSWER 1.16 ( [06) l a.
- 1. INCREASES
- 2. INCRE S [') hn(I (1.00)
1 l
(*'**** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
- l. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION < Page 92
- THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWER 1.17 (2.00)-'
(0.25 each) '
- a. REMAINS THE-SAME
- b. REMAINS THE SAME
- c. REMAINS THE SAME
- d. DECREASE
- e. REMAINS THE SAME
- f. INCREASE
- g. DECREASE f
- a. 3
- b. 1
ANSWER 1.19 (2.00)
(0.5 each)
- a. Adjacent to withdrawn rod
- b. center
- c. 600 ppm
- d. 540 F k
I l
- ) l
(***** CATEGORY 1 CONTINUED ON NEXT PAGE q l
i
1
- 1. PRINCIELES OF NUCLEAR POWER PLANT OPERATION. Page.93
- d M '" ~ d -
- """"PCMchangeduetotemperature=5F*-15pcm/degreeF=-75pcml ,
[0.25] )
Reactivity change required by rods = 228 step to 188 steps = 40 steps
[0.25]
40 steps *.1.095 pcm/ steps : -43.8 pcm (inward rod movement)
[0.25] ;
-75 pcm (temperature) + -43.8 pcm (rods) = -118.8 pcm reactivity change.
[0.25]
-118.8 pcm / -10 pcm/ ppm = 11.88 ppm boron change.
[0.25]
800 ppm - 11.88 ppm = 588.12 ppm final concentration. [0.25]
i REFERENCE WEST RCC for Large PWRs pg 3-21/22, 5-33 l SQN Lesson Plan EGT222.005 004000A404 ..(KA's)
. ANSWER 1.21 (1.50)
- a. Unit 1 (0.25) due to a smaller Beff value (0.50)
- b. Unit 1 (0.25) due to a more negative MTC value (0.50)
REFERENCE WEST Fund of NRP PG 7-33; WEST RCC for Large PWRs PG 3-23 f SQN Lesson Plan EGT222.005, PP 11 & 12 Lesson Objective. F ,
l 192008K124 192008K114 192008K110 ..(KA's) l I
l l
3 l
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) '
L 1
1
- 1. PRIHglELES OF NUCLEAR POWER PLANT OPERATION. Page 94
' THERMODYNAMICS. HEAT TRANSEEB AND FLUID FLOW i i
l i
(
ANSWER 1 22- (2.00) 1192 BTU /LBM (1190-1195 BTU /LBM acceptable)
- b. 300 F (290-310 F acceptable)
REFERENCE !
} STEAM TABLES l
193003K125 ..(KA's) !
l ANSWER 1.23 (1.50)
- a. Tsat for 2250 psia (2235 psig)
= 652.67 F (0.50)
Subcooling margin : Tsat - Thot = 652.67 - 587 = 65.67 F + or - 0.05 F (0.50)
- b. Subcooling margin decreases because Thot will increase as power increases (0.50)
REFERENCE STA, licensed operator training, Natural circulation and inadequate core cooling, Para X.B 001000K556 ..(KA's)
(***** END OF CATEGORY 1 *****)
l s
L 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Paga 95 SYSTEMS l.
l ANSWER 2.01 (1.00)
I L
l REFERENCE SQN Lesson Plan, OPL2710019, pp 4, 5, 19 Lesson Objective, OPL2710019, #6 i
ANSWER 2.02 (1.00) b-REFFRENCE SQN Lesson I'lan, OPL2710023, p. 6 Lesson Oojective, OPL2710023, #1 System Descr$ption, RHR ANSWER- 2.03 (1.00)
(O.Icad1) a CLOSE
- b. OPEN
- c. NO SIGNAL
- d. NO SIGNAL
- e. CLOSE
- f. NO SIGNAL
- g. NO SIGNAL
- h. CLOSE
- 1. NO SIGNAL
- j. CLOSE REFERENCE SQN Lesson Plan OPL2710025, pp 7 & 8 Lesson Objective OPL2710025, #2
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
'2. PLKdT__ DESIGN INCLUDING SAFETY AND EMERGENCY Page 96 SYSTEtie 4
ANSWER 2.04 .(1.00)
I c 1 i
REFERENCE SQN Question and Answer Book, SQN Cert Wk 9, 5/9/86, Quest. 2.03 035010G007 035010G004 035010K401 ..(KA's)
ANSWER 2.05 ('1.00) b on c (fY111 REFERENCE SQN Question and Answer bank, Prelicense Exam, 9/25/87 025000G004 ..(KA's)
BJ R.C ( Wcd W r 3 migk om md ANSWER 2.06 (1.00)
)f L CNr\
REFERENCE t
SQN Lesson Plan, Emergency Core Cooling, OPL271C025, p. 15 1
006000K602 ..(KA's)
( C_ A F ( &;h% TenN.~) M ed. Teunsmct- ( C -t -, b ANSWER 2.07 (1.00)
'"W N7 p g ,. u ,}
l (0.5 each)
- 1. To relieve excess reactor coolant when drawing a bubble in the PZR)
C- --
l l 2. (Toletdownsealinjectionflow)whennormalflowpathisunavailable.
l l
l l I (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
L i
l L - - _ -- _ _ -
- 2. PLANT DESIGN ItLCLUDING SAFETY AND EMERGENCY Page 97 SYSTE_ tis _ -
REFERENCE SQN Lesson Plan, OPL2710022, p. 8 Lesson Objective, OPL2710022, #2.1 & 3 ANSWER 2.08 (1.50)
(0.15 each)
- 1. CST
- 2. ERCW
- 3. High Prcesure Fire System s REFERENCE SQN System Operating Procedure, SOI-3.2 i
ANSWER 2.09 (2.00) ;
(0.5 each) ' h ~ ~" ' # ~"
- 1. Hot leg loop 1 =
- 2. Hot leg' loop 3
- 3. RHR downstream of Deat Exchangers M1%
- 4. Containment Atmosphere w' m- u, G ab . v v r, ms, w ,,u 5, w m.
REFERENCE SQN Question and Answer Bank, prelicense exam 10/16/87 000011G006 ..(KA's) .
S o 5 - ist ~s.
1 ANSWER 2.10 (1.50)
(0.25 each) 1
- 1. Ensure core is flooded and decay heat is being removed. j
- 2. Minimize cladding failure. I 6 83 g 3. Maintain RCS integrity, (unless already lost). I
%a 4. Stop positive reactivity additions. 4
- 5. Maintain reactor in a subcritical condition.
s6. Restore core parameters to stable conditions.
Octcpt -these 5 deMc3n crNtnh. plus on,y w WA4. On bs t, t ekst pak ckdtwp. < zoco*F 1
z Lo;.;.t ekd widrdort1s < ax -Malclod'LAness. 1 3.tou t' m x.th 9ene 'on % <t 3; aHy&MiaJ a noud <1 sli (kducra h ull 8 tempt ths ~ ?
- )
(***** CATEGORY 2 CONTINUED ON NEXT PAGE
- 4. Ma dd, askble. yomely
- 5. pa,. dark Lov ferm coo /Q eqd,'ll7 1
l
- 2. ELANT DESIGN INCLUDING SAFETY AND EMERGENCY Page 98
.SYSTZUS
ANSWER E.11 (2.50)
(any 5 @ 0.5 each) n ' c p ; - m g p, t ;, g , 7 m
- 1. Containment Air monitors
- 2. Pocket sump level it. 'T'2 4t t u t. 4
- 3. Vessel Flange leakoff tp (p?vt
- 4. Humidity detectors G b ~' ' "5'^
- 5. Condenser Vacuum pump
- 6. Steam Generator Blowdown monitor
- 7. Component Cooling effluent monitor
- 8. Excessive ~ makeup volume (makeup system) 9, Co b ~* 7t 7-m.
sc. ,
REFERENCE n.
S g. m &
SQN System Description, Reactor Coolant System, OPL2710017, p. 8 002000K405 ..(KA's) b O'Pb 7 ANSWER 2.12 (2.00)
- 1. Containment pressure > 9.5 psig
- 2. At least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> since start of accident
005000K306 ..(KA's)
ANSWER 2.13 (1.00)
Steam lines 1 and 4
(*.:*** CATEGORY 2 CONTINUED CN NEXT PAGE *****)
l 1
i l
1
. R. PLANT _RKSIGN INCLUDING SAFETY AND EMERGENCY Page 99
. SYSTEMS-REFERENCE SQN System Operating Procedure, SOI-3.2
& Pze am f r e s s < <-
Poc- u: A Lv u Lev h C., .A e FLe v C~jT29 ANSWER 2.14 (2.00) g
- a. (O.5 each)-
- 1. The power range low flux trips will not be automatically reinstated.
- 2. The intermediate flux trips will not be automatically reinstated.
-p ge7.m c et ot 4o, car sI Wo - e , cht T w~ L-c REFERENCE bo Tcob ~ L ~5 W R\ % DL -
c (%'.d :t ch._ A m < X 4 6 I ue W F~ ~ @ 3 @01'c "
SQN Operator Certification Training, Solid State Protection System, Para X.F SQN Lesson Plan, OPL2710048, p. 27 NRC IE notice 86-105 012000K401 ..(KA's)
ANSWER 2.15 (1.00)
(0.5 each)
OT delta T -- Low DNBR protection OP delta T -- Excessive Reactor Power Output (Slow reactivity addition)
REFERENCE SQN Lesson Plan, Reactor Coolant Temperature Instrument, Objective F 012000K402 ..(KA's)
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
' E. ' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Page100 SYSTEMS l
L l ANSWER 2.16 .(1.00) l NO CHANGE [0.5] - since the bellows only separates the RCS/Pzr water l from the DP cell internals to prevent flashing of the reference leg fluid on transients. [0.5]
l l REFERENCE SQN Lesson Plan, Par and Control System, OPL2710019, pp 10-12 Lesson Objective, OPL2710019, #2.e 011000K604 ..(KA's) ooo oteAu2 ANSWER 2.17 (1.50)
(0.5 each) gzs] 5.rG
- a. Maintains 43 F subcooling marging to ensure adequate margin to DNB. 3 ean eneration-and
.for#s3- ( OG6M starts ECCS 3to prevent DNBR (from
- b. -GLves uncleer hes' 3 fallingbelow1.( Co.3 0 G.zO
- c. Protects RCS pressure boundary against rupture' and lifting of i safeties [*C
. REFERENCE SQN Lesson Plan, Par and Control System, OPL2710019, p. 16 010000K602 010000K403 ..(KA's)
ANSWER 2. . (2.00) ,
.y ( g o g _. p , q pD ert. (TcE7.-gum p O ;
O
- a. 425 gpm, 1500 psig f (3W T-%~1 g, b . 8,000 gallons each tankt 400 psig (4 tanks) e c. 90 gpm, 2250 psig
- d. 3000 gpm, 175 psig .
j + 5 *}, crn $k S. b~S -
l
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
'2. PLANT DESIGN INCLUDING SAFEIY AND EMERGENCY Page101
. . SYSTEMS
'I I
REFERENCE 1
'SQN. Lesson Plan,. Emergency Core Cooling, OPL271CO25,-p. 7 ]
Lesson Objective, OPL2710025, #9 l SQ M% s 006000K605 006000K602 ..(KA's)
ANSWER 2.19 (2.00) l RHR suction line.[1.0) -- To allow for RHR operation during Steam Generator tube repairs (partial drain down of RCS) [1.0]
. REFERENCE SQN System Description, Reactor Coolant System, OPL2710017, p. 14 005000K109 ..(KA's)
ANSWER 2.20 (1.50)
- a. ' Check valves [0.5]
. -;(- b . Monitoring pump discharge piping temperature routinely [0.5].and venting periodically (every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) if temperature is elevated
(> 200 F). [0.5]
L);\\ Tf% < c o ,A l 'S '~ thws w . t b__ ck CC G es +N ~ '" A N d REFERENCE Ib m c r 4 o ~ 'a i % v'~ 'Tos * # ""
(c. u n d &ccm ed y ( 4 SQN System Description, Auxiliary Feedwater, OPL271C035, p. 10 TVA Memorandum, Dated April 6, 1984 from P. R. Wallace (L82140403 800)
SQN Examination - Cert Wk 5, 3/25/88, question #7 061000K505 ..(KA's) a l
ANSWER 2.21 (1.50)
- a. Air pressure [0.5]
- b. Spring pressure [0.5] and steam flow [0.5]
1
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
l
2 ', PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Page102
- - . SYSTEMS-REFERENCE ,
SQN System Description, Main Steam System, p. 8 039000K405 ..(KA's) t I
l l i I
l 1
1 i
l l
(***** END OF CATEGORY 2 *****) I
ANSWER 3.02 (1.00) e REFERENCE SQN Lesson Plan, Rod Control, OPL2710046, p. 8 001000K403 ..(KA's)
~ ANSWER 3.03 (1.00) d REFERENCE SQN Lesson Plans, Steam Dump Control System 041020A408 ..(KA's)
ANSWER 3.04 (1.00) a
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
ANSWER 3.05 (1.50)
(0 5 each)
Auctioneered High Delta T Auctioneered High Tavg Rod Position Indication REFERENCE SQN Lesson Plan, Rod Control System, OPL271C018, p. 15 and Figure OTT-85-7 Lesson Objective, OPL2710018, #4 Lesson' Plan, OPL2710046, p. 15 001000G007 ..(KA's)
ANSWER 3.06 (2.00)
- a. TRUE
- b. TRUE
- c. FALSE
- d. FALSE REFERENCE SQN Lesson Plan, 6.9 Shutdown Boards, OPN220.021, pp 4, 6, 13 ,
Lesson Objective, OPN220.021, # 1, 2, 3, 9 062000K401 ..(KA's)
ANSWER 3.07 (1.00) ,
i b {
l 1
1
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) l i
F 3. INSTRUMENTS AND CONTROLS Paga105 REFERENCE SQN Lesson Plan OPL2710025, p. 11 Lesson Objective, OPL271CO25, #B-4 006000K405 ..(KA's)
ANSWER. 3.08 (2.00)
(0.5 each)
- 1. Pzr level > 17%
- 2. FCV-62-69 & 70 open (containment isolation valves)
- 3. Charging pump running
ANSWER 3.09 (1.00)
(0.5 each)
- 1. SI signal
ANSWER 3.10 (1.50)
_(0.5 each)
- 1. Source range high flux.
- 2. Intermediate Range Low Power Trip.
- 3. Power Range Low Power Trip.
No c=JE r dc(ku f(A i#
t be 6 1: L3-> ha i." T~ ^ s w Tr:c W PS ht i e r c.,4 < d p q
N5 cf.Schb4
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *4***)
3.- INSTRUMENTSLAND CONTROLS' Pcgo106 REFERENCE SQN: Lesson Plan OPL2720020, Rev. 2, p. 73 012000K406 ..(KA's) 1 ANSWER 3.11 (2.50)
(any 5-@ 0.5 each)
- 1. Containment Isolation phase A
- 2. Feedwater isolation
- 3. Diesel generator emergency start
- 4. Reactor Trip 5.
6.
Safeguard Sequencer - 4 h Mu-
- C-- ~ N -4 Containment Ventilation
@]
- 8. Am lia.3 Btd3 stouh
ANSWER 3.12 (2.00)
- 1. C-9 Condenser available [0.5] -- satisfied when vacuum > 17" on 2/2 pts and 1/3 Cire pumps running [0.25]. Ensures condenser is capable of condensing steam so no overpressurization occurs
[0.25].
- 2. P-12 Low-Low Tavg SI [0.5] -- 2/4 loop Tavg < 540 F [0.25].
Prevents inadvertent or excessive cooldown of RCS [0.25).
REFERENCE SQN System Description, Steam Dump System, OPL271C030, p. 7 Lesson Objective, OPL271CO30, #5 Question and Answer Bank 041020K418 041020K417 041020K402 041020K409 ..(KA's) l
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
1 i
)
- 3. INSTRUMENTS _AND CONTROLS Page107 'i i
l I
l
' ANSWER 3.13 (1.50) l (0.5 each) 4 I
- 1. SI
- 2. Blackout
- 3. . Low discharge header pressure.
REFERENCE i SQN Question and Answer Bank, Exam Cert Wk 5, 1/6/87, quest. #36 008000K401 ..(KA's) l ANSWER 3.14 .(2.00)
(any 8 @ 0.25 each)
- 1. Mode select switch NORMAL
- 2. Control switch A - AUTO '
- 3. UVX and UVY deenergized (load shed occurred) or uV c~ L ' b W 5
- 4. Voltage available from DG
- 5. Alternate feeder open
- 6. Normal feeder open
- 7. Utility feeder open '
- 8. No lockout relay - --
eg scc c A " rc' *
g h c , ge, ab m p/
3 p o ,t p o a_. 'a e t rne o c_ %
REFERENCE SQN Lesson Plan, 6.9 kV Shutdown board, OPN221.021, p. 22 Lesson Objectives, OPN221.021, #16 Fotw-064000A301 ..(KA's) l ANSWER 3.15 (2.00)
- a. (0.5 each)
- 1. High-High level in any steam generator
- 2. SIS
- 3. Reactor trip with Low Tavg
- b. Condensate booster pump selected for P-auto starts. [0.5]
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
i
ANSWER 3.16 (3.00)
(GRADING: 0.1 : Control' relay number: 0.1 = coincidence; 0.1 = origin; 0.1 = setpoint; 0.1 = manual / automatic determination.)
- 1. C-1, 1/2, Inter. Range, > 20% pwr current equiv., Auto / man.
L
- 2. C-2, 1/4, Power Range, > 103% power, auto / man.
- 3. C-3, 2/4, OTdeltaT, deltaT within 3% of trip setpoint, auto / man.
- 4. C-4, 2/4, OPdeltaT, deltaT within 3% of trip setpoint, auto / man.
- 5. C-5, 1/1, PT-1-73, < 15% turbine load, auto only.
- 6. C-11, no logic, Bank 'D' control rods, >220 steps, auto only REFERENCE SQN Question and Answer Bank, Prelicense Final Exam, 12/10/86, quest.
3.02 001000K402 ..(KA's)
D&M ANSWER 3.17 (1.50) 340A The master contr er which controls PORY M4 has an integral portion
[0.75] which ses the controller output to increase as the difference between P and actual Pressure persist with time [0.75].
T ELC"If 9 ~t<. Gusny cb c vo k P%< Si ~ s e REFERENCE w{ 1% nRZ w c w B h '3 QQ p 3 g.th SQN Lesson Plan OPL271C019, Rev. 1, fig. OTT-68-5.6 000027A101 010000K403 ..(KA's)
T %-4\-(oP
(**'*** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
- 3. INSTRUMENTS AND CONTROLS- Page109
- 2.. .
ANSWER 3.18 (1 50)
- 1. All 4 MFRVs fall closed [0.5].
- 2. All-4 MSIVs fail-closed [0.5].
- 3. #2 & #4 MFRV. bypass valves fail shut.[0.5].
C("wd :t - p 1 itt- g .' n - p L.o d - L L h " ' N 't REFERENCE SQN AOI-21.1, Loss of 125 V DC Vital Battery Board I,'p. 1 000058G011 ..(KA's)
ANSWER 3.19 (1.00)
The operator manually closes the AFW feed supply to the affected S/G.
REFERENCE
'SQN Question and Answer Bank, Prelicense exan., 10/16/87 0000/9K304 ..(KA's) l l
l l
l 1
l l
l' 1
l' l-(***** END OF CATEGORY 3 *****)
L
I -
- f. PROCEDURES -' NORMAL.' ABNORMAL. EMERGENCY Page110 N 4 AND RADIOLOGICAL CONTROL
. ANSWER 4.01 (1.00) b REFERENCE Westinghouse ERG Background documents 194001A102 ..(KA's)
ANSWER 4.02 (1.00) a REFERENCE SQN AOI-30, Plant Fires, p. 3 000067K304 ..(KA's)
ANSWER 4.03 (1.00) b
. REFERENCE SQN FR-0 Status Trees SQN, Operator Certification Training, LP: Status Trees (week 7-7),
objectives B and D 194001A102 ..(KA's)
ANSWER 4.04 (1.00) c
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
'4 '. PROCEDURES-- NORMAL, ABNORMAL, EMERGENCY Page111
..- .-AND RADIOLOGICAL CONTROL-REFERENCE-SQN Technical Specification, Table .3.3-4 000007A202- ..(KA's)
ANSWER 4.05 (1.00) e REFERENCE SQN Technical Specification, Table 3.3-4 000007A202 ..(KA's)
ANSWER 4.06 (1.00) b REFERENCE SQN E-0, Reactor Trip or Safety Injection ECA 0.0 entry conditons 194001A102 ..(KA's)
ANSUER 4.07 (1.00) a REFERENCE Westinghouse Background Information for ECA 0.0 Users Guide for ERG's and Background Documents p. 17/18 000055K302 ..(KA's) l (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
l l
l .
___-___ - _ l
FT -
- 4. PROCEDURES'- NORMAL. ABNORMAL. EMERGENCY Page112
- +- ' .AND-RADIOLOGICAL CONTROL ANSWER 4.08 (2.00)
(0 5 each)
- a. 3
-b. I
- c. 4
- d. !L/
)M7 REFERENCE SQN Radiological Control Instructions, RCI-1, p. 3
'194001K103 ..(KA's)
ANSWER 4.09 (1.00) d' REFERENCE SQN Fuel Handling Instruction, FBI-7, p. 4
, 034000G010 ..(KA's)
ANSWER 4.10 (2.00)
(0.5 each)
- 1. RCS pressure is >100 psig but < 1000 psig.
- 2. #1 seal leakoff valve is open.
- 3. #1 seal leakoff flow rate is less than one gpm.
- 4. Seal injection water flow rate to each pump is > six gpm.
REFERENCE SQN SOI 68.2, Reactor Coolant Pumps, p. 5 003000K401 003000A201 ..(KA's)
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
c 4. PROCEpDRES ~ NORMAL. ABNORMAL. EMERGENCY Page113 3
(. ,AND RADIOLOGICAL CONTROL i h !
l i l
l ANSWER 4.11 (2.00) (o.c1)
G 3D -g:sn V k
- a. IF par level is falling, THEN start charging pumps as necessaryA(to )
maintain levelfens3) [1.0] d777 j S.uJ 5.ub 5.zs3 l b. IF loss of pzr level is imminent THEN trip the reactor, initiate SIs '
j and go to E-0,go.t53 [1.0]
f ^
REFERENCE SQN AOI-24, Steam Generator Tube Leak, p. 1 1
L 000037G010 ..(KA's)
ANSWER 4.12 (2.00)
(any 4 @ 0.5 each)
- 1. Violation of Tech Spec safety limit or other conditions which requires rapid shutdown withou- trip.
2 Can not satisfy Tech Spec LCO and actions because of conditions in excess of specifications.
- 3. Expiration of time interval of Tech Spec LCO.
- 4. Serious condition which requires rapid load drop to prevent or minimize a more serious condition but does not require a trip.
- 5. Serious condition which requires a reactor trip.
REFERENCE SQN AOI-32, Emergency Shutdown, p. 1 001050G015 001050G006 ..(KA's)
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r- 7
- 4. PROCEDURES-- NORMAL. ABNORMAL, EMERGEECX Page114 ;
a- ,AND RADIOLOGIC &L' CONTROL {
1 1
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I L . ANSWER 4.13 (1.50)
D I Check auto actions completed:
.a. Pump selected for auto starts @ 40#. [0.5]
- b. Automatic makeup from demin water @ 75" (60%) in surge tank. [0.5]
- c. If not -- start the pump and add demin makeup. [0.5]
-REFERENCE SQN AOI-15A.1, Loss of' Component Cooling Water, p. 4 000026K303 ..(KA's)
ANSWER 4.14 (2.00)
C o, s eu.h)
- 1. Subcooling > 40 F
- 2. RCS pressure - stable or increasing
- 4. Pzr level > 20%
REFERENCE SQN, E-FOP, SI Termination 000009G012 ..(KA's)
ANSWER 4.15 (1.50) 3 (0. 25 each)
- 1. Unexpected rise in S/G 1evel
- 2. S/G blowdown monitors RM-90-120, 121, or 124 alarm
- 3. Rad Con survey of main steamlines
- 4. Rad Con survey of blowdown lines
- 5. Chem lab sample
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
- 4. PRQQEDU8ES - NORMAL. ABNORMAL. EMERGENCY Pagell5 .
- t.
.AND BADIQLQQlg&L CONTROL j
. REFERENCE
.SQN EOP E-3, Steam Generator Tube Rupture, p. 2 >
Operator Prelicense Training, OPL2720005, Objective B.1 I Question and Answer Bank 000038A202 ..(KA's)
. ANSWER 4.16 '(1.50)
- 1. > Zero power rod insertion limit
- 2. +/- 1000 pcm of estimated critical rod height
ANSWER 4.17 11. 5C P C o 8)
(0.5 each)
-To--ensure tha MRTVa close with-rn f'i nu evoss-the-velvee. <DM To ensure the Heat Sink is available through the SG PORV's (operability).
-To-prevent inadverterrt-1+f4;ing of the SG Safety Velvaa m REFERENCE SQN AOI-27, Control Room Inaccessibility, p. 2
- FACILITY TO PROVIDE FURTHER DETAILS ***************
000068K318 ..(KA's) Lj o-b t DMu' Q m. e s .n m ,. Tk L . A ;- 7, , + ve,l <^ E U- .' 3 d rsc~-ss.s ~
Wl & , % .3 T)E9Av h b CDA l
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- 4. PROCEDURES - NO3 MAL, ABNORMAL, EMERGENCY Page116 t
oAND RADIOLOGICAL CONTROL I
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- .. ANSWER 4.18 (1.00)
Prevents Localized power peaking [0.25], minimizes Xenon oscillations
[0.25] and(withdrawing the rod at normal rate)can lead to fuel 6777 cladding damage in the region of the dropped rod [0.5].
REFERENCE SQN AOI-2, Malfunction of the Rod Control System, p. 7 000003G003 ..(KA's).
ANSWER 4.19 (1.00)
Power is tagged out to both SIP's [0.50]
The non operating centrifugal charging pump is in pull to lock [0,50).
REFERENCE SQN GOI-1, Plant Startup from Cold Shutdown to Hot Standby, p. 4 006050G005 006050A201 ..(KA's)
ANSWER 4.20 (1.00)
Inspect clothing for damage o-c V Oli t" Swk he Q.% : ~ 3 \ % -; s t n .f5 WhV)
REFERENCE SQN Radiological Control Instruction, RCI-1, p. 10 194001K103 ..(KA's)
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.4. PEOCEDURES - NORMAL, ABNORMAL, EMERGENCY Page117
.. .AND RADIOLOGICAL-CONTROL ANSWER 4.21 (2.00)
This is a Low Dose waiting area inside radiation areas [0.5] where background is significantly lower than the general area radiation levels [0.5],
When there is a temporary work delay [0.5] and exit from the work i area is not an option [0.5].
REFERENCE SQN Radiological Control Instruction, RCI-10, p. 6 194001K103 ..(KA's)
ANSWER 4.22 (1.50)
(0.5 each)
- a. Remain Shutdown
- b. Startup
- c. Remain Shutdown REFERENCE SQN TS 3.4.6.2 002020G011 ..(KA's) 1 1
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(***** END OF CATEGORY 4 *****)
(********** END OF EXAMINATION **********)
I _ __-
MA sT ER U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: SEGUOYAH REACTOR TYPE: PWR-WEC4 DATE ADMINISTERED: 88/12/12 EXAMINER: REGION II CANDIDATE:
INSTRUCTIONS TO CANDIDATE:
U;a . separate paper for the answers. Write answers on one side only.
Stcplo question sheet on top of the answer sheets, ' Points for each quostion are indicated in parentheses after the question. The passing grcd3 requires at least 70% in each category and a final grade of at Iccet 80%. Examination papers will be picked up'six (6) hours after
.tha examination starts.
% OF 3ATEGORY %.0F CANDIDATE'S CATEGORY VALUE' TOTAL SCORE VALUE CATEGORY P3 31.00 25.62: 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 46 30 00 24.75 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 30,06 24.h ^ 7. PROCEDURES - NORMAL, ABNORMAL,
' EMERGENCY AND RADIOLOGICAL
/ CONTROL M.f 48 3Or&&# 24.79 # 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS
- 1.to # g 12r707 % Totals Final Grade All work done on this examination is my own. I have neither given nor rcceived aid.
Candidate's Signature
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this ev=htion, the following rules apply:
- 1. Chan on the examination means an automatic denial of your applica and.oculd result in more severe penalties.
I
- 2. Restroom trips are to be limited and only one candidate at a ties may leave. You must avoid all contacts with anyone outside the examination room to avoid evea the appearance or possibility of cheating.
- 3. Use black ink or dark pencil oA to facilitate legible reproductions. ,
- 4. Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the time you START and STOP on the cover sheet of the examination.
- 6. Use only the paper provided for answers.
- 7. Print your namo in the upper right-hand corner of each pag.: of the exam.
- 8. The exam has one question per page. Write the answer beneath the question. Write only o_n_ o_Il_e side of the exam and any extra answer sheets.
- 9. Number each answer continued on additional paper as to category and number, for example,1.4, 6.3.
- 10. Attach continued answers to back of question to which it arnlies.
- 11. 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 exammation 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.
- 18. When you complete your ev==ination, you shall: ,
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- a. . Assemble your examination as follows: l' (1) Exam questions with answers on top.
(2) Exam aids - figures, tables, etc.
(3) Scratch paper used during the examination.
l 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.
If after d.- Leave the examination area, as defined by the examiner.
leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
S I
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S. THEORY OF NUCLEAR' POWER PLANT' OPERATION. FLUIDS, 8NQ PAGE~ 2 THERMODYNAMICS e
{
i QUESTION 5.01 (1.00)
Complete the f ollowing statement. concerning the Technical t ~Speci fi cati on, ~ SHUTDOWN MARGIN requi rements. ,
1 SHUTDOWN MARGIN requirements DO NOT.....
'c). ensure the reactor will be maintained sufficiently suberitical in the shut down condition at EOL with a double ended main steam line break with Tavg at a no load operating temperature.
b). provide for allowing the reactor to be made suberitical from all normal operating conditions.
c). ensure the reactivity transients associated with postulated accident conditions in the FSAR are controllable within acceptable limits.
d). vary throughout core life as a function of fuel depletion, RCS baron concentration changes, and RCS Tavg.
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S .- IHEORY10F NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 3 THERMODYNAMIC QUESTION 5.02- (1.00)
Which DNE of the following statements describes the change in Modsrctor Temperature Coef ficient (MTC) from BOL to EDL7 c).- The MTC becomes more negative due to. increasing baron concentration, decremaing fission product inventory, and axial flux redistribution toward the edges of the core.
~b). The MTC becomes more negative due to decreasing boron concentration, increasing fission product inventory, and radial flux-redistribution toward the edges of the core.
c). The MTC becomes less negative due to increasing boron concentration, increasing fission product inventory, and axial flux redistribution toward the edges'of the core.
d). The MTC becomes less negative due to decreasing baron concentration, decreasing fission product inventory, and axial flux redistribution toward the edges of the core.
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
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- 5. THEDRY OF NUCLEAR' POWER PLANT OPERATION. FLUIDS. AND 'PAGE 4
-THERMODYNAMICS.
QUESTION 5.03 (1.00)
During a reactor startup under xenon-free conditions, rod withdrawal
!io ctopped at the -0.02%' delta k/k position and the count rate is
'cIlow:d to stabilire. In regard to the response of the count. rate in
.tha hour after stabilization, which one of the following statements is
'corrcc t?. (Assume N0 further operator actions are taken) a). Count rate will remain essentially constant.
b). Count rate will rapidly decrease to its pre-startup level.
c). Count rate will' slowly decrease because it'is subcritical.
d). Count rate will slowly increase due to long-lived delayed neutrons.
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'5 '. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 5
'(\ THERMODYNAMICS (UESTION. 5.04 (1.00)
Chich ONE of the following actions would help, rather than hinder, estural circulation?
- m). Lowering steam generator level
'b).- Lowering RCS pressure c). Increasing RCS temperature d). . Increasing pressurizer level 1
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JL. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 6 THERMODYNAMICS QUESTION 5.05 (2.00)
A b ginning-of-life core has been operating with all rods fully withdecwn at 100% power for 1 week. The bank D control rods are
'inscrtcd 100 steps with sufficient baron dilution to offset the rccctivity added by the rods. Answer the following questions about tha rccultant xenon transient IN THE RODDED REGION OF THE CORE.
c). How does the xenon level change at first (INCREASES, DECREASES, STAYS THE SAME)? (0.5) b). Why does indirect production of xenon decrease during this transient? (1.00) c). The final steady-state xenon concentration will be (LESS THAN, GREATER THAN, EQUAL TO) the original concentration (0.5)
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' 5.
THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 7 THERMODYNAMICS 1
I
- QUESTION S.06 (2.00) l
!Tha plcnt is operating at 30% power when one reactor coolant pump l
- tripa due to an electrical fault. The control systems are all in I cutomatic. Bank D rods are initially at 165 steps. Assume the plant I lDOES NOT trip. Indicate whether the following parameters INCREASE, l l DECREASE or REMAIN THE SAME.
l s). Turbine power j b). Rsactor power (final) l l
c). Final rod height l
d). T-avg (affected loop) j o). T-avg (unaffected loop)
- f). Core Delta-T l g). Dalta-T (affected loop) h). Dalta-T (unaffected loop) i l
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5.'
THEQRY OF NQGLEAR POWER PLANT OPERATION, FLUIDS. AND PAGE 8 THERMODYNAMICS QUESTION 5.07 (1.50)
During a pouer ramp to 50 percent power with rod control in automatic cn incorrect boron change was calculated and performed, which resulted in tha plant stabilizing at the desired power but with control rods at tha all out position and Tavg 5 degrees F. below the target value.
Givcn the following initial parameters, PROVIDE the final RCS baron concentration needed to INCREASE Tavg 5 degrees F. while returning control bank D to the 188 step position. Assume turbine power stays conctcnt at 50 percent.
Initial RCS baron concentration = 600 ppm Total power coefficient = -20 pcm/ percent Moderator temperature coefficient = -15 pcm/ degree F.
Differential baron worth = -10 pcm/ ppm Control rod worth avg. (5-80) = 8.60 pcm/ step (95-170) = 4.16 pcm/ step (185-220) = 1.075 pcm/ step Show all calculations l
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3 THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 9 THERMODYNAMICS
'i QUESTION 5.08 (1.50)
Unit 2 still near the beginning of its fuel cycle; Unit 1, an
- 1dsntical reactor and fuel loading scheme, is near the end of its fuel
, cyclo. . Assume both plants have just started up following a 3 week chutdown period. !
l c). Critical data has just been taken at 10 E-G amps at both plants cnd the operators have added small, equal amount of reactivity to l continue the power ascension. Which plant, Unit i or 2, will have the HIGHER steady startup rate from this equal reactivity insertion and EXPLAIN WHY? (0.75) b). Shortly after 50 percent power is reached during these startups, rod control is placed in manual at both plants. Shortly _
ofterward,.a shutdown' bank control rod worth -150 pcm drops into the core at both plants. Assuming that no' operator action is token for these casualties and that neither reactor trips. Which plant, Unit l'or 2 will stabilize with the HIGHER steady-state ;
Tavg and EXPLAIN WHY? (0.75) 1
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 10 THERMODYNAMICS GUESTICN 5.09 (2.00)
The plcnt is in Hot Standby with the RCS pressure being maintained at 985 poig. A pressurizer PORV is slowly discharging to the pressurizer roliof tank which is at 5 psig. The steam quality is the pressurizer otarm bubble is 100 percent.
a). What is the enthalpy of the fluid entering the PRT7 b). What is the temperature on the tail pipe downstream of the PORV7
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 11 THERMODYNAMICS j QUESTION 5.10 (1.00)
@ON, GOI-2, Plant Startup from Hot Standby to Minimum Load, requires Ghnt the critical rod position be taken at 10 E-8 amps on the Antarcadiate range. If, during a xenon-free startup at MOL, the ep; rotor " overshot" 10 E-8 amps and instead leveled off at 10 E-7 cmpa, which one of the following statements is CORRECT 7 a). At 10 E-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 would be higher. '
b). At 10 E-7 amps, there are little or no effects frcm nuclear heat, therefore, the critical rod position should be the same as at 10 E-8 amps.
c). At 10 E-7 amps, there are substantial effects from nuclear heat, therefore, the critical rod position will be higher than at 10 E-8 amps.
d). At 10 E-7 amps, nuclear heat, xenon, and being a decade higher in power level will all contribute to a higher critical rod position.
o). At 10 E-7 amps, there are substantial effects from nuclear heat but they are cancelled by the additional burnable poison depletion. Therefore, the critical rod position should be the same as at 10 E-8 amps.
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5.
'THEDRY OF NUCLEAR POWER PLANT OPERATION. FLUIDS, AND PAGE 12 THERMODYNAMICS
- 1 1
QUESTION- 5.11 (1.00) j
. i Which DNE statement below describes centrifugal pump. runout conditions?
a).. High. discharge pressure, low flow, high power demand.
b). High discharge pressure, low flow,7 1ow power demand.
c). Low discharge pressure, high flow, high power' demand.
d). Low discharge pressure, high flow,, low power. demand.
o). Low discharge pressure, low flow, high power demand.
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
'5.- THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 13 THERMODYNAMICS
{
l(EUESTION . 5.12- (1.00) l lIn tha event of a rod e.iection accident, which DNE of the following taill b3 the.first reactivity coefficient to insert negative lrocetivity?-
l c). Moderator temperature coefficient b). Pressure coefficient l
c). Void coefficient
!. .d). Doppler. coefficient I
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5.' THEDRY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. ANQ PAGE 14 THERMODYNAMICS
@UESTION 5.13 (1.00) dith.the plant operating at 85% power and all systems in a
@oracl/ automatic configuration, the operator borates 100 PPM. Which
$NE of the following. describes the effect on Shutdown Margin as Gofinsd in Technical-Specifications?
a). Increase-b). Increase until rods move c). Decrease d). Decrease until rods move a). Remain unchanged regardless of rod movement
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLULDS. ANQ PAGE 15 THERMODYNAMICS QUESTION 5.14 (1.00)
Which DNE of the following statements is CORRECT concerning the paralleling of electrical systems?
c). Although it is desirable to have speed and phase position matched it is much more important to have voltage matched.
b). If voltages are not matched at the time the synchronizing switch is closed, there will be a VAR flow from the lower voltage source to the higher one.
c). If the incoming machine is at synchronous speed but out of phase with the running bus when the breaker is closed, heavy currents will flow to either accelerate or retard the incoming machine.
d). -If the incoming machine is in phase but slightly faster than synchronous speed when paralleled, the system will tend to speed up to synchronous speed.
a). If the resistances are not matched at the time the synchronizing switch is closed, heavy currents will flow to tend to speed up the incoming machine to synchronous speed.
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. F(UIDS. AND PAGE 16 THERMODYNAMICS
@UESTION 5.15 (1.00)
Ehich DNE of the'following is the reason for the -1/3 DPM start up
- rcto following a reactor trip?
o). The decay constant of the longest-lived group of delayed neutrons.
b). The ability of U-235 to fission with source neutrons.
c). The amount of negative reactivity added on a scram being greater than the Shutdown Margin.
d). The doppler effect adding negative reactivity following a scram.
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'5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 17 THERMODYNAMICS
$UESTION 5.16 (1.00)
@or thn boiling phases listed below choose the answer indicating the Gorrcet order in which they would occur in a coolant channel with Rormal flow and high heat flux.
- 1. transition boiling
- 2. bulk boiling
- 3. film boiling
- 4. sub-cooled nucleate boiling a). 2, 4, 3, 1 b). 2, 4, 1, 3 c). 4, 2, 3, 1 d). 4, 2, 1, 3 a
i.
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- 5. THEORY'OF NUCLEAR' POWER' PLANT OPERATION. FLUIDS. fdA PAGE 10 THERMODYNAMIC.<
QUESTION 5.17 (1.50)
Ehe plcnt has.been operating at 100% power for the last ten days when St trips. For each of the times given below, state HOW xenon will Of fcct the Shutdown Margin . (INCREASE, DECREASE, or NO EFFECT).
Sonoidsr each case separately and assume all other factors are sonatent.
O) . . 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the trip
@).- 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the trip e ) .. - 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after the trip l
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l.. j l'5. THEDRY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 19 THERMODYNAMICS
'GUESTION 5.18 (1.50) l Indicate how each of the following parameters would have to change l(INCREASE, or DECREASE) to cause an increase in the Departure from Nuc1cate Boiling Ratio (DNBR).
o). Raactor. power J l
lb) . - RCS pressure !
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'8 ) . RCS flow-1 l
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l5. '
THEORY OF NUCLEAR POWER' PLANT OPERATION. FLUIDS. AND :PAGE 20.
-THERMODYNAMICS
$UESTION 5.19 (1.50)
Indicate how each of the following would affect differential control 1 pod worth. (INCREASE, DECREASE, or NO'EFFECT). Consider each case ccparctely.
d). Increase in moderator temperature G). Increase in baron concentration G). Increase in core age (fues. depletion)
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS, ANQ PAGE 21
-THERMODYNAMICS QUESTION 5.20 (1.50)
For EACH of the following, indicate whether the following will, INCREASE, DECREASE, er REMAIN THE SAME.
'o ) . Available NPSH for a MFP as volumetric flow rate increases.
b). Minimum required RCP NPSH as volumetric flow rate increases.
c). Available NPSH to condensate (hotwell) pumps as condenser cubcooling increases.
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5.' THEORY'Q~ NUCLEAR POWER'-PLANT DPERATION. FLUIDS, AND PAGE 22 THERMOQX69MICS QUESTION ~ 5.21 (1.00)
Which DNE of the following must the condenser remove the most heat from to condense?. (assume steam is of equal quality)-
a). one pound of steam at O psia b). one pound of steam at 300 psia c). two pounds of steam at 600 psia d). two pounds of steam at 1200 psia
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i5. = THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 23 THERMODYNAMICS L
QUESTION 5.22 (1.00)
Chich DNE expression.below describes the heat flux hot channel factor.
@q(Z)?
o). Maximum fuel assembly G at height Z/ Avg Q at-height Z b). Maximum fuel assembly G at height Z/ Avg Q in the core c). Average fuel assembly G/ Maximum G at height Z d). Average Q at height Z/ Avg G in the core I
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SL ' THEORY-OF NUCLEAR' POWER PLANT OFERATION. FLUIDS. ..AND PAGE 24
. THERMODYNAMICS QUESTION 5.23 (1.00)
Which ONE of the following would cause an inadvstrtent di11 tion cccid:nt?
c). Overfilling a S/G while in hot standby.
b). A regenerative heat exchanger laak.
c). Valving in a domineralizer that was not saturated.-
d). A VCT Lo-Lo level resulting in the RWST being used for charging.
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f,i THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS. ANE PAGE 25 THERMODYNAMICS' GUESTION 5.24 (1.00)
Indicate TRUE or FALSE for each of the following statements concerning tho offect that delayed neutrons have on reactivity.
a). _Bacause delayed neutrons are born at lower energies than prompt ncutrons,they are less likely to leak out of the core resulting in a positive effect.
b ) '. Dalayed neutrons are born at'an average energy incapable of cousing fast fission of U-238 creatin; a negative effect.
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5.' THEORY OF NUCLEAR *CWER PLANT OPERATION. FLUIDS. AND PAGE-26 THERMODYNAMICS GUESTION '5.25 (1.00)
Ch2t are the two reasons for shifting the SI mode from cold leg eccirculation to hat leg recirculation approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a (LOCA?
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- 6. ' PLANT SYSTEMS'DESIGNa CONTROL. AND INSTRUMENTATION 'PAGE 27 I
(
.)
'GUESTION 6.01 (1'.00)
Which DNE of theffollowing provides the correct reasons for cointaining a' minimum spray bypass flow to the pressurizer 7 a). Prevent excessive cooling to the surge line.
Reduce the delta pressure across the spray valves. ~j b). Reduce thermal shock to the spra/ nozzle.
Ensure that the backup heaters cycle on.
c). Prevent excessive cooling to the spray line.
Equalize baron between pressurizer and the spray line, d). Minimize stress to the surge line thermal sleeve.
Remove gases from the RCS.
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'6. PLANT SYSTEM 9 DES.IGN- CONTROL. AND INSTRUMENTATION PAGE 2]
QUESTION 6.02 (1.00,
.Which ONE of the following design features enhances the heat rcmovc1 operation of the ice condenser and containment spray
'cyatca?
c). Pressure operated doors open to allow upper containment air to flow through the lower containment.
b). Air return fans provide flow to return the air from the upper containment to the lower containment.
c). Containment design, such that the delta P between upper and lower containment drives the air circulation.
d). Ventilation coolers and recirculation fans are used to mix the air and provido additional cooling.
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A. 'PL. ANT SYSTEMS-DESIGN. CONTROL. AND' INSTRUMENTATION PAGE 29 QUESTION 6.03 (2.00)
@ toto the FOUR locations tha Post Accident Sampling System can
&cw cceples.
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61 -PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE:30 QUESTION' 6.04- (2.00) i LIST the FOUR conditions.that must exist to place one train of Racidual Heat Removal Spray in service.
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6.- PLANT' SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION { PAGE 31-QUESTION 6.05 (2.00)
Whilo conducting a plant shutdown, power range channel N41 la prop 3rly removed from service with the reactor at 50% power. As pow r drops'below 10%, a P-10 permissive interlock solid ste,te biotcble for power range channel N43 does not reset.
4). What are the THREE effects on the Solid State Protection System.(SSPS) due to the failure of the P-10 bistable switch to remet? Disregard otheir parmissives which may interact with P-10.
b). Dsscribe-TWO methods which should be used to monitor for this failure in tha control' room.
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l' fi PLANT: SYSTEMS DESIGN. CONTROL, AND INSTRUMENTATION PAGE 32 l-GUESTION 6.06 (1.00)
$riofly explain the offact on pressurizer level indication
- following a' break in the' bellows separating the reference leg
- fluid.from the pressurizer fluid. INCLUDE your reasoning for the.
Offcct.- Assume steady state power and level conditions.
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- 6. PLANT SYSTEMS DESIGN.' CONTROL.'AND INSTRUMENTATION PAGE 33 IUESTION '6.07 (1.00) dhich GNE ' of ' the following ~ in correct. concerning the operation of the.Rocetor Trip (RT) and. Reactor Trio Bypass (BY) Breakers?
a). The Train B SSPS trip signal directly trips RTB and BYB.
b). The Train A SSPS trip signal directly trips RTA and PYB.
c). Tripping is accomplished by an undervoltage relay, normally held open by'48 volt DC power from the logic panels.
d). To allow testing of the RT's, BOTH BY's may be closed while the reactor is at power.
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! l 6C PLANT SYSTEMS DESIGN. CQNTROL. AND INSTRUMENTATION PAGE 34
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1 QUESTION 6.08 (1.00) 1 Chich'ONE-of.'the1fo11owing Rod Control Subsystems provides the l function of dampening the mismatch channel output at. low rates of phinga, allowing the Tavg/ Tref channel to control?J c). Variable gain-unit b ) .- Rate comparator i c). Non-linear gain unit d). Reactor control unit l
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'6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 35 l,
l GUESTION 6.09 (2.00)
Indicate your response (TRUE or FALSE) to each of the following Gonecrning the 6.9 KV shutdown boards. ANSWER EACH SEPARATELY c). The alternate and normal feedor breakers sense each others position and both will trip if both breakers indicate shut et the same time.
G). The 1A-A 6.9 KV Shutdown Board is fed by its normal supply bus of the 6.9 KV unit board 1B.
i 8). An 86 relay actuation will lockout the normal feeder breaker, Diesel Generator breaker, Utility breaker, and colect the alternate feeder to close.
6). The blackout relays BOX and BOY are reset prior to paralleling the Diesel Generator to the Unit board to prevent the DG from overloading after shutting the Unit board feeder. i
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'6 . PLANT SYSTEMS DESIGN. CONTROL. AtJD. INSTRUMENTATION PAGE 36 QUESTION 6.10 (2.50)
As a rcsult of a Safety Injection signal, certain signals generated from the SI will " latch in" (seal in), while others will automatically rocat or clear.
Liot FIVE signals which are " latched in" (seal in) as a result of the SI cignal.
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J6 .' ' PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE-37.
QUESTION 6.11- (1.25)
Daccribe what happens to the coils associated with the CRDM's, end-otete 2-alarms /13dications actuated on an Urgent Failure of tho-rod control synts.w.
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- 6. PLANT SYSTEMS DESIGN. LONTROL. AND INSTRUMENTATION PAGE 38 GUESTION 6.12 (1.50)
Ohilo at 100% power with rod control in automatic, the Turbine Iepulco Channel supplying the rod control logic fails high. What 60 ths effect on the following rod control system components?
O . -- Verlable Gain Unit
@. Tavg-Tref mismatch 8 Rod speed
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- 6. ~ PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 39 1
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-QUESTION 6.13 (1.50)
What FOUR signals will automatically initiate the operation of the Auxiliary Building Gas Treatment System?
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- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 40
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I QUESTION 6.14 (1.00)
Chich ONE of the following describes how the AFW system is grcvanted from feeding a faulted S/G7 c). Operator action is required to isolate AFW when a faulted S/G is detected.
b). Pressure switches on the AFW discharge lines will automatically close the loop level control valves when low AFW discharge pressure is detected.
c). Level transmitters on the S/G wide range level instrument will automatically close the loop level control valve when a low level is detected in a S/G.
d). Flow transmitters on the AFW discharge lines will automatically close the loop level control valves when excessive flow is detected in AFW discharge piping.
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- 6. PLANT SYSTEMS DESIGN. CDNTROL. AND INSTRUMENTATION PAGE 41' g.
i OUESTION 6.15' .(1.50)
. Indicate whether the'following situations will ARM ONLY, ARM-AND CCTUATE, or HAVE NO-EFFECT on the steam dump system.
Q. . -COX power, 7.5%/ min ramp decrease in turbine load for.3 minutes, Tavg> Tref by'7 deg F, steam dumps'in Tavg mode of operation G.
Hot Zero Power,.Tavg=549 deg F, steam dumps in.STM PRESS code with 1005-psig set into the steam pressure controllers G. . Turbine trip, Tavg=542 deg F, steam dumps in Tavg mode
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- 6. ' LANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 42 GUE2 TION 6.16 (1.00)
[h:t'DNE set of signals below are sent to the Reactor Protection Cy; tea to indicate a Turbine Trip?
c). Throttle valves closed & Auto Stop 011 pressure low b). Throttle valves closed & EHC pressure low c). Governor valves closed & Auto Stop Dil pressure low d). Governor valves closed & EHC pressure low I
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h' L PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 43 i:
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GLE] TION 6.17 (1.50)
Conecrning the Solid State Protection System, state how the logic ccrd3 receive input signals and trip the reactor when the proper logic'is met?
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6 -PLANT SYSTEMS DESIGN. CONTROL.'AND INSTRUMENTATION PAGE 44 !
! QUESTION 6.18 (1.00) htatotheinterlocksassociatedwithautomaticclosureofthe' doin feed pump turbine condenser valves for the following lOituations:
l lo). If both MFP's are running and one trips above 40% feedwater l- flow.
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- b). If only one feed pump is running and it trips.
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6.. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 45 QUESTION 6.19 (1.50)
Conecrning.the rod control system, state the direction and speed of rod motion and the effects of the rod motion on the plant for the following-(assume parameter is a controlling instrument,'and tharo.is no operator action):
c). RTD fails upscale b). Power Range Instrument fails upscale
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- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATIQM PAGE 46 i
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'l GUESTION 6.20 (1.50)
Ctato_the three signals used to develop the control rod insertion licito?
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- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 47 QUESTION 6.21- (1.25) l (Dolative to the Main Turbine operations: 1
[hnt in-the reason (s) the generator is NOT immediately Gripp:d when the turbine trips.
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7.- PRDCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 43 RADIOLOGICAL CONTROL CUESTION 7.01 (1.00)
Chich DNE of the following is NOT correct concerning the use Gcnv:ntions associated with EDP's?
c). Even'after a transition to another procedura, the steps started BEFORE the transition was made must still be completed, but not to delay the transition.
b). Continuous warnings contained in a caution ! step) are i NOT in effect when an operator is referenced to a procedure to be porformed concurrently with the EDP in effect.
c). If a caution statement occurs before step one of an EOP it may apply either to the whole procedure or just to the first step.
d). Unless otherwise specified, a required task need not be fully completed before proceeding to the next instruction; it is enough to begin the task and have some assurance that it is progressing satisfactorily.
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- 7. ' PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND
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PAGE 49 RADIOLOGICAL CONTROL DUESTION .7.02. (1.00) !
Chich DNE of the following symptoms would require.the initiation ef10 mnnual. reactor trip AND safety injection if neither had occurrsd automatically?
a). Containment pressure = 1.0 psig-b). General Warning alarm on the Solid State Protection System B c). Pressurizer pressure = 1850 psig; pressurizer level = 40%
d). Power = 33%g and loss of flow in one loop a). Power.= 45%; pressurizer level = 93%
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8, PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 50 RADIOLOGICAL CONTROL GUESTION 7.03 (1.00)
Chen cre the procedures required by the Critical Safety
@ unction Status Trees implemented during ECA-0.0, " Lose of All AC
@ownr," according to the Westinghouse Background Information
@ caro Guide for the Functional Restoration Guides?
o). Never implemented when in ECA-0.0 b). Upon entry to ECA-0.0 c). Upon reaching step 5 of ECA-0.0 d). Upon exiting to subprocedures of ECA-0.0 1
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- 7. PROCEDURES - NORMAL'. ABNORMAL. EMERGENCY AND PAGE'51 RADIOLOGICAL CONTROL QUESTION 7.04 (2.00)
Liot the TWO'immediate actions for a steam. generator' tube LEAK lcccording to ADI-24, Steam Generator Tube Leak. " Include both tha IF and THEN actions.
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7; .' PROCEDURES -- NORMAL. ABNORMAL. EMERGENCY AND PAGE 52
[ RADIOLOGICAL CONTROL t
GUESTION- 7.05: (2.00)
Liot . the FOUR r.ormal (NOT adverse containment) containment Safety Injcction termination criteria found on the EDP foldout page.
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l7. PROCEDURES - NORMAL. ABNDRMAL. EMERGENCY AND PAGE 53 RADIOLOGICAL CONTROL E1JESTION 7.06 (1.00)
@xplcin WHY any attempt to realign a dropped rod should be l8cordinated with Reactor Engineering according to a CAUTION iOtotcr:nt in AOI-2, " Malfunction of Rod Control System," section l@ " Dropped Rod."
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- 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 54 RADIOLOGICAL CONTROL i
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GUESTION 7.07 (1.00) l
$uring normal CVCS operation, which ONE of the following is an j Obnormn1 condition and would require operator action tc correct? )
c). VCT pressure is 15 psig. l 1
The temperature of the fluid leaving the letdown heat I b).
exchangers is 126.5 deg F.
c). The RCP seal injection water temperature is 120deg F and flow to the seals is O gpm/ pump.
d). RCP seal differential pressure is 300 psid.
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- 7. ' PROCEDURES -- NORMAL, ABNORMAL. EMERGENCY AND PAGE 55 RADIOLOGICAL CONTROk-GUESTION, 7.08 (1.00)
During an inadvertent dilution accident while at 100% power, what reactor trip signal would be the most probable cause of a reactor
- trip?
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i 7, .- PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 56'.
i ~ RADIOLOGICAL CONTROL i i
' QUESTION 7.09 - ( 2. 50 ) '
dhat are five of the six symptoms =used to indicate entry into-
@-3,'" Steam Generator Tube Rupture?" ,
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- 7. - PROCEDURES - NORMAL. ABNORMAL. EMERGENCY ANE-PAGE'57'
.RA'DIOLOGICAL CONTROL a
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GUESTION 7.10- (1.00)
Q NOTE'in GOI-2, " Plant Startup from Hot Standby to Minimum i Lord," states that if control rods were withdrawn 5 steps during 'l hootup, the control rods must be fully inserted prior to t c)ithdrcwing rods.
d). Why are the rods withdrawn 5 steps during heatup? I b). Why must they be inserted prior to withdrawal?
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'7 . - PROCEDURES'- NORMAL'. ABNORMAL.' EMERGENCY AND PAGE 58-RADIOLOGICAL CONTROL p.
GUESTIDN. 7.11 (1.50)
Whnt:three conditions must be met prior to implementing ES3.3
." Post-SGTR.Cooldown by Ruptured'S/G Depressurization?"
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- 7. ' PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE~59 RADIOLOGICAL CONTROL
- QUESTION- 7.12 (1.00)
,In.the precautions section of SGN GOI-2, " Plant Startup from Hot
@t::ndby to Minimum Load,:" it states that:
"The turbine-generator unit shall not be motored for any oxtended periods."
Chnt:io the reason'for limiting motoring operation to less than leno minute?
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7.. 'PRDCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 60 RADIOLOGICAL CONTROL i
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GUESTION 7.13 (1.00)
'In cccordance with precautions in SON GOI-2, " Plant Startup from Hot Stcndby to Minimum Load," what must be verified in the event ef o fcedwater'line break in the east or west valve rooms during O ctertup?
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- 7. - PROCEDURES - NORMAL.' ABNORMAL. EMERGENCY AND PAGE'61 RADIOLOGICAL CONTROL QUESTION 7.14 (1.00)-
If'while performing SON GOI-5, " Normal Power Operations," it is
- found that a particular precaution can not be complied _with, what 1 must.ba done?
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- 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY ANQ PAGE 62 RADIOLOGICAL CONTROL l
- GUESTION 7.15 (1.00) l lIncecordancewithSONGOI-5, " Normal Power Operations," while
- cpercting at 100% power a " Rod Control Banks Limit Low-Low" alarm
'comro in, what must be done?
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7.- PROCEDURES - NORMAL 4 ABNORMAL.-EMERGENCY AND PAGE 63
. RADIOLOGICAL' CONTROL.
QUESTION 7.~16 (1.00).
On cccordance with SON GOI-5, " Normal Power Operations," what is Ghe rosson for NOT'holdit.g turbine load on Unit 1 at 93%
(indiented power on EHC. console) for long periods of time?
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- 7. PROCEDURES - NORMAL. ADNORMAL. EMERGENCY AND PAGE 64 RADIOLOGICAL CONTROL GUESTION 7.17 (1.50)
Antwar the following TRUE or FALSE concerning manual valve operations of motor operated valves.
- c. Wrenches or " cheaters" can be used on motor operated valve hnndwheels, or on any valve with a drive gear drive provided parmission is obtained from the Shift Engineer.
- b. When operating motor operated valves manually without line pressure, the final seating should be done manually with oxtreme care.
- c. When using a handwheel, turn the handwheel slowly when approaching either end of travel.
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7.'
PROCEDURES '- NORMAL.,- ABNORMAL. EMERGENCY AND PAGE 65 RADIOLOGICAL CONTROL l
GUESTION. 7.19 (2.00)
I'fl-o locked valve is to be checked closed and-it is found that
!ths hrnd operator is unable to be moved because of the locking-f(Ocvico, what ' action must be taken to check the valve shut, and
- how in it documented?
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- 7. PROCEDURES - NOBMAL. ABNORMAL. EMERGENCY AND PAGE 66 RADIOLOGICAL CQNTROL I
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QUESTION 7.19 (2.00) biot, in their order of preference, the four recovery techniques Ototsd-in FR-C.1, " Response to Inadequate Core Cooling." I;
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'7. PROCEDURES'- NORMAL. ABNORMAL'.-EMERGENCY AND' PAGE 67 RADIOLOGICAL CONTROL QUESTION 7.20 (1.00)
Why io the S/G atmospheric PORV NOT isolated while performing E-3, "SGTR", but is only verified shut <1040 psig, and that its controller is'in AUTO 7 t
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- 7. PROCEDURES - NORMAL. ABNORMAL.' EMERGENCY AND PAGE 68 88pl9 LOGICAL CONTROL-GUESTION 7.21L (1.50)
Fhe Racponse Not Obtained for step 1 of FR-H.1, " Response to Loss ef S condary Heat Sink," states:
"If ALL S/G Wide Range levels <25%, the STOP ALL RCP's and immediately initiate feed and bleed per steps 11 to 13."
CDhyn cro the RCP's tripped prior to initiating f eed and bleed, Ocida from the fact that heat input from the pumps will be Octovcd7
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- 7. PRONDURES '- NORMAL - ABNORMAL. EMERGENCY AND -PAGE 69 PtADIOLOGICAL CONTROL QUESTION 7.22 (2.00)
Chat cre ALL the actions contained in the RESPONSE NOT DBTAINED Golumn of FR-S.1, "ATWS", for the steps listed below? Ensure you G]iccusa any contingency actions stated within the RNO step itc311. ,
O. - Turbine is NOT verified as tripped.
G. Roactor Trip Breakers will not open, l
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B. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 70 QUESTION B.01 (1.00)
Which one of the following may proceed given that a Technical Specification Action Statement has been entered that you " suspend all
' CORE ALTERATIONS 7"
- o. Remove a neutron source from the core,
- b. Using the bridge in the core is allowed, provided that the low load limit is jumpered out,
- c. Control rods and burnable poison rods may be shuffled as long a K-effective is less than or equal to 0.95.
- d. Completion of the movement of a component to a safe conservative position within the reactor pressure vessel.
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I El . ADMINISTR671VE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 71 1
I l8UESTION b.02 (1.00)
. Unit 1 is in COLD SHUTDOWN with Tavg being maintained stable at 290 l@cgrosa F. by RHR. The following equipment is INOPERABLE:
-Centrifugal Charging Pump 19-B # (1hr)
-Reciprocating Charging Pump 1B # (1hr)
-DG 1A-A # (1hr)
O Thara is no estimate of repair time.
The Shift is directed to recommence the plant cooldown to a RCS Tavg ef 130 degrees F.. SHUTDOWN MARGIN calculations indicate compliance cith the TS LCO for " SHUTDOWN MARGIN - Tavg less than or equal to 200 6cgrcca F." throughout the full range of the anticipated cooldown.
Ohich DNE of the following actions most correctly detail the allowrnces and/or limitations imposed by the Technical Specifications in the instance?
NOTE: APPLICABLE TECHNICAL SPECIFICATIONS ARE ENCLOSED FOR REFERENCE.
- c. Plant cooldown may recommence; OPERATIONAL MODE 6 may be entered with no restrictions on plant operations.
- b. Plant cooldown may recommences OPERATIONAL MODE 6 may be entered BUT CORE ALTERATIONS are precluded.
- c. Plant cooldown is prohibited AND heatup to 200 degrees F. is required.
- d. Plant cooldown is prohibited.
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G- ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIORE PAGE 72 EUESTION 8.03 (1.50)
SIST the FIVE bases for the minimum temperature for criticality limit ef tha Technical Specifications, i.e. what does this limitation Oncura?
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G ADMINISTRATIVE ~ PROCEDUR_ES. - CONDITIONS. AND LIMITATIO}ls PAGE 73 GUESTION 8.04 (1.00)
'What ACTION (S) must IMMEDIATELY be initiated (per the TSs)-11 BHUTDOWN MARGIN decreases to less than 1.0% delta k/k in Mode 57 Be Specific.
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- 8. - -ADMINISTRATIVE 2 PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 74 (EUESTION 8.05 (1.00) 1 lAccording to Technical Specification 6.7, Safety Limit Violation, what two actions must take place within one hour in the event a Safety biait'is violated?
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ba B. ' ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE'75 GUESTION 8.06- (1.00)-
Ehst ie.the basis for the upper limit of containment temperature?
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- g. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 76 QUESTION 8.07 (1.00)
With reactor power above 50% and AFD within the " doghouse" limits (Tcch Spec figure 3.2-1; AFD limits as a function of rated thermal powsr):
What operator action is required by Technical Specification 3/4.2.1
- ( Axial Flux Dif ference if the AFD monitor alarm becomes inoperable?
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- 8. ADMINISTRATIVE:PROCE_DMBES. CONDITIONS. AND LIMITATIONS PAGE 77 QUESTION 8.08 (1.00)
-In cccordance with SON AI-16, " Fuse Control", if while replacing a-Q)1own circuit fuse, with the correct fuse, the replacement fuse blows.
Qhet sction(s) must take place if this occurs?
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L- ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 78~
QUESTION B.09 '(1.50)
In cecordance with AI-19, " Plant Modifications: After Licensing", what throa' plant personnel, by title, can act as the Plant Manager
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B.. ' ADMINISTRATIVE' PROCEDURES. CONDITIONS. AND LIMITATIONS.- PAGEl79 GUESTION 8.10 .(1.00)
In,cccordance with.AI-22, " Plant Duty Supervisor", - if while atteempting to contact-the Duty Supervisor.to report an unusual condition, the Duty Supervisor can not be. contacted, what shall be done?
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fa Si ADMINISTRATIVE' PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 80
$UESTION- 8.11 (1.00)
In eccordance with AI-37, " Independent Verification", what is the
$urposo for having Independent Verification for component restoration?
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B- 'ADMINI'BTRATIVE PROCEDURES.~ CONDITIONS. AND LIMITATIONS 'PAGE 81 QUESTION -- 8.12 (1.50)
In'cecordance with AI-37, " Independent Verification", name 3 of.tho'4 Gonditions.to waive independent. verification.
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S. ADMINISTRATIVE PROCEDURES. CONDITIONS.-AND LIMITATIONS PAGE 82-GUESTION- B.13. (1.00)
What'in the Technical Specification basis for the requirement to rcduca Tavg_to less.than 500 degrees F. when specific activity limit
.on tha.RCS are exceeded?
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lS? ADMINISTRATIVE PROCEDURES. CONDITIONS; AND LIMITATIONS PAGE G3-i-
- DUESTION. 8.14 (1.00)
@thto the configuration log entry requirements when clearances are lpicccd'on equipment.
l.
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(***** CATEGORY OB CONTINUED ON NEXT PAGE *****)
i
-- _______ _ _________-- _ _ _ _ _ __ _ j
i B. ADMINISTRATIVE PROCEDURES.' CONDITIONS. AND LIMITATIONS' PAGE C4 GUESTION 8.15 (1.50)-
Sict THREE critical rod height limits as described.in GOI-2, " Plant
@tcrtup from Hot Standby to Minimum Load."
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B. ' ADMINISTRATIVE PROCEDURES,' CONDITIONS. AND LIMITATIONS PAGE C5 i
)
1 QUESTION 8.16 (1.00)
Which one of the following is a condition which requires the i opnrctor to EMERGENCY BORATE per ADI-34, "Emergsncy Boration."
- o. failure of any full length rod to insert after a reactor trip.
- b. receipt of the ROD BANKS INSERTION LIMIT LOW annunciator.
- c. an uncontrolled cooldown following a reactor trip.
- d. an uncontrolled or unexplained increase in source range counts curing a reactor startup.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
l
(s - . . . . .
.31' ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE B6 EUESTION B.17 (2.00) t ,
LIST'the.Immediate Opersator Actions required to be taken if a reactor !
- trip cignal is generated.and no reactor trip occurs, per FR-S.1, l"R eponse to Nuclear Power Generation /ATWS."
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B. ADMINISTRATIVE PROCEDURES. CONDITIONS, AND LIMITATIONS PAGE 87 QUESTION 8.18 (1.50)
In cccordante with: SON AI-30, " Nuclear Plant Conduct of Operations",
ecro, by title,~6 of the B individuals that are permitted in the unit Gontrol room (other than the units horseshoe area) during plant trcnsients or trips.
i
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
B. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE.C8
- f. oo
(-id O) e GUESTIDN 8.19 a.- Whose prior verbal permission is required in order to assign a designated Unit Operator to the opposite unit?
b.' If assignment of an Unit Operator is changed to the. opposite unit, what must be done to allow the reassignment to take. place?
i
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____x__________________ . _ _ _ _ _ _ . _ _ _ _ _ _ __ _ _ _ _ . _ _ _ . _ _ . - _ _ _ _ _ _ _i
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 89 QUESTION 8.20 (1.00)
Whilo in MODE 1 operation of Unit 2 the following abnormal
' indications / conditions are identified:
A.;vicual inspection reveals that a pressurizer spray valve is leaking from the valve body.
Lack rate calculations indicate a leakage history of 0.16, 0.098, and 0.055 gpm, for the past two days.
Tha most recent leak rate calculation indicates a 1.115 gpm leak.
Rafor to the attached Technical Specification excerpts on RCS leak rota. State the applicable paragraph (s) that would address this condition.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
- 8. ' ADMINISTRATIVE' PROCEDURES,' CONDITIONS. AND LIMITATIONS PA8E 90 (Ei)ESTION 8.21 (1.00) dhile in MODE 1 operation of Unit 2, maintenance personnel are
@crforcing a degraded voltage test on 6.9KV Shutdown Board 2B-B. The Collowing data is collected from the tests
@cgredad Voltage Sensors Trip Value Response Time (secs) 27DAT 65BOv 9.5
'27DBT 6570v 10.2 27DCT 6540v 8.7 Bofor to the attached Technical Specification excerpts (3.3.2). State Gh2 cp;cific Technical Specification paragraph and action numbers chich cddress the results of this test.
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B! ADMINISTRATIVE PROCEDURES. CONDITIONS.'AND LIMITATIONS PAGE 91 1
1 i
GUESTION 8.22 (1.00)
Ohilo in MODE 5 operation _.of Unit 2 the ERCW Train "B" Radiation Monitor 0-RM-90-134_has-a failed power supply. The failure has ~
rccultsd in the disbling of the common instrument malfunction alarm cith 0-RM-90-141, ERCW Train "B" Radiation Monitor. Refer.to the ottech:d Technical. Specification 3.3.3.9 Table 3.3-12. No other plant Oquipssnt is failed. Which one of the following would be the Appropriate response to this condition? )
- n. Implement Action #32.
- b. Implement Actions 3.3.9.b. and #32.
- c. Implement'LCO 3.0.3.
- d. No Technical Specification action (s) are required.
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1
- 8. " ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 92 QUESTION 8.23 (2.00)
Whilo in MODE 1 operation of Unit 2 the following condition exista SI-90.72, " Reactor Trip Instrumentation Functional Test of delta T/TAVG Channel IV, is being performed. One alarm has been received, "Procacs Protect Racks Channel Test Sequence Violated," (see attached Annunciator Procedure). All trip bistable status lights are extinguished.
Anewsr the following questions:
- a. Which Technical Specification Action statement (s) of LCOs 3.3.1 and 3.3.2 (see_ attached) would be applicable under this condition?
- b. In accordance with AI-47 " Unplanned Test Stoppage, Exiting, Raentering," what action must be taken to return the instrument loop to service if the test cannot be reversed or completed as written?
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(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) !
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- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIQNE PAGE 93 6UESTION 8.24 (2.00)
Unit.2'is:in MODE 3 operation. .NO LCOs are in effect. Refer to the Ottcch:d excerpts of SI-2 " Shift Log" and answer the following:
G. The Evening Shift'(1500 - 2300) SRO reviews SI-2 and notes the Channel deviations on.page 4 of 20. Which Technical Specification LCO would be applied as a result.of the readings on this page?
@. .How would'the SRO complete the DATA PACKAGE COVER SHEET
-(cttached) for.the " Potential Reportable. Occurrence Initiated" blocks?
1 e
(***** END OF CATEGORY 08 *****)
(************* END OF EXAMINATION *************) ;
o %
ATTACHMENTS
. SEQUOYAH SRO EXAM REGION II 88/12/12 J
i
s' ..
s F-ANSWER KEY SEQUOYAH SRO EXAM REGION II 88/12/12
f THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 94 THERMODYNAMICS ISWERS -- SEQUOYAH -88/12/12-REGION II EWER 5.01 (1.00) d (1.00)
@ERENCE F Gucotion and Answer bank NEED TO GET MORE INFO FOR THIS QUESTION 1.8/3.9) 7~/.5 B4 SES 3. 4. t. t. I ~ 4 2/41. ;_ y m 6002K114 ...(KA*S)
EWER 5.02 (1.00) b (1.00)
PERENCE
@ Prolicense Training Program lesson plan, week 1, Reactor Physics elsw, Training objective V.H i.1/3.1-)
SOOO4K10 ...(KA*S)
SWER 5.03 (1.00)
(1.00)-
PERENCE DO Pralicense Training Program lesson plan, week 1, Reactor Physics vicw, Training Objectives V.B and V.C i.8/3.8)
@OOBK104 ...(KA*S)
).- THEDRY DF NUCLEAR POWER PLANT DPERATION. FLUIDS. AND PAGE 93-4- THERMODYNAMICS iNSWERS -- SEQUOYAH' -88/12/12-REGION II NSWER 5.04 (1.00)
!). (1.00)
@FERENCE ON Natural circulation / inadequate core cooling GN C rt. Wk 9 4/22/88 03008K123 ...(KA*S) iNSWER '5.05 (2.00) i) . Increases (0.5)
)) . Because the concentration of iodine (the xenon precursor) is d: creasing (1.00)
- ) '. LGOs than (0.5)
@FERENCE
@T222.OO5, p. 23
@Q 8.3/3.4)
@2OO6K108 ...(KA*S)
ZSWER 5.06 (2.00)
@.25 occh) i) . rcmains the same
)) . - rGmains the same
- ) . / rcmains the same
- 1) . dOcrease
)). ramains the same
'). increase I) . dccrease
)) . , increase
@FERENCE NP Quantion and Answer bank SON Week 1, 4/25/86
$O017K104 ...(KA'S)
THEDRY OF' NUCLEAR POWER PLANT OPERATION. FLUIDS.-AND PAGE 96 THERMODYNAMICS EWERS'-- SEQUOYAH -88/12/12-REGION II
@WER 5.07. (1.50)
Gi ch nge due to' temperature = 5 deg. F. X -15 pcm/ degree F = -75 pcm (0.25)
Octivity change required by rods = 228 step to 188 step = 40 steps (0.25)
) uteps X 1/095 pcm/ steps = - 43.8 pcm (inward rod movement (0.25)
-l
'5 pcm (temperature) + -43.8 pcm (rods)~= -118.8 pcm reactivity j Cnga. (0.25)
,18.8 pcm/-10pcm/ ppm = 11.88 ppm boron change. (0.25) i
@ ppm .11.88 ppm = 588.12 ppm final concentration '(0.25) l FERENCE l
@T RCC for Large PWRs pg 3-21/22, 5-33 l 8222.005 i.2/3.6)
COOOA404 ...(KA'S) i I
l
@WER 5.08 (1.50) i
.- Unit 1 (0.25) due to a smaller Beta eff value (0.5)
. Unit 1 (0.25) due to a more negative MTC value (0.5) l WERENCE l
@T Fund of NRP p 7-33, West Rcc_for Large PWRs p 3-23 i l
@222.005 p 11,12
)F l i.5/3.6)
@OOOK124 ...(KA'S)
)
l I
I l
1 i
1
_ _ - . . _ = _ - _ - _ - - - _ _ - ._
5 ' THEORY'OF NUCLEAR-POWER PLANT OPERATION. FLUIDS AND PAGE 97
-" IBEBNODYNAMICS NSWER3 --- SEQUOYAH -88/12/12-REGION II t ,
TSWER', 5.09 (2.'00)
- 3) . - 1192. BTU /LBM (1190-1191. ATU/LBM acceptable) (1,00)
)) . 3OO deg. F. (290-310 deg. F acceptable) ('.,00)
@FERENCE-Wocm tcbles 8.3/3.4) 93OO3K125 ...(KA*S)
ZSWER 5.1 (1.00) b). (1.00)
@FERENCE li)I-2 p. 17 8.4/3.6)
@2OO K113 ...(KA*S)
ZSWER 5.11 (1.00) c ) '.
GFERENCE
@T 201.100 p. 32 2.5/2.7)
@1004K112 ...(KA*S) 1 1
i
THEORY OF NUCLEAR' POWER PLANT OPERATION. FLUIDS. ANR PAGE 98 THERMODYNAMICS SWERS -- SEGUDYAH -88/12/12-REGION II EWER 5.12 (1.00) d PERENCE 6205.201 p.26 i.4/3.8) 8001K118 ...(KA*S)
EWER 5.13 (1.00) a).
PERENCE Z C rt. Wk 9, 4/22/88 8002K114 ...(KA'S)
)SWER 5.14 (1.00) c).
PERENCE Cd To Get A Ref.
i.3/3.5) )
600"K108 ...(KA*S) l MWER 5.15 (1.00) .
I c). 4 PERENCE S205.201 p. 18
- .2/3.3)
Soo3K106 ...(KA*S.) !
I i;
!.- ' THEORY OF NUN FAR P N R PLANT' OPERATION. FLUIDS. AND- 'PAGE 99'
- L THERMODYNAMICS NSWERS--- SEGUOYAH -88/12/12-REGION'II 4
ZSWER: 5.16 (!-00) d
@FERENCE. .
@T205.101 pp. 39,40,41:
@C,. Fundamentals of Nuclear Reactor Engineering p. 195 Cncral Physics HT&FF p 122-125
@300!K103.- ...(KA*S)
$$WER 5.17 (1.50)
>). Increase . (0.5 each)
)) . . R mains the'same
- ) . d: crease
@FERENCE lTG205.201.p. 32,33,34
& 1/3.1)
@2006K112 ...(KA*S)
ZSWER 5.18 (1.50)
- 1) . D: crease' (0.5 each)'
)).-- Increase
- ) . - Increase
@FERENCE-
@T205.201 p. 42,43 8.4/3.6)
$300BK105 ...(KA*S)
~
).- -THEORY'OF NUCLEAR-POWER PLANT OPERATION. FLUIDS. AND' PAGE-Z100, THERMODYNAMICS NSWERS'- .SEQUOYAH ' -98/12/12-REGION II ZSWER2 5.19- (1.50)
)) . Increase '(0.5'each) 1).L:D; crease
- ) .. . Increase
@FERENCE
@T205.201 p. 27,28,29
@2005K107 ...(KA*S):
NSWER 5.20- (1.50)
- 3) . d: crease (0.5 each)
)) . increase
- ) . . increase GFERENCE
@T201.OO p. 31, 32
@1004K115' ...(KA'S)
ZSWER 5.21 (1.00) c)., 1.00
@FERENCE
)tocm Tcbles 3.3/3.4)
@3OO3K125 ...(KA*S)
). - ' THEORY OF NUCLEAR POWER' PLANT ~ OPERATION. FLUJp3dEQi P A G E */,1 0 1-THERMODYNAMICS ZSWER3 - SEQUOYAH. -88/12/12-REGION II iNSWER- 5.22 (1.00) b).
lEFERENCE Cchnical Specification 3/4.2.2
- 8.9/3.1)
@2OO5K112 ...(KA*S)
ZSWER 5.23 (1.00) c).
GFERENCE FL271CO22 p. 20
@<020A213 ...(KA*S) l I
ZSWER. 5.24 (1.00)
- 3) . TRUE (0.5 ear;h) ,
l
)) . TRUE
@FERENCE ,
GMr 205.201 !
8.4/2.5) l O2OO1K102 ...(KA'S) ,
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. _ _ _ . .i
- 5. THEORY;OF NUCLEAR POWER PLANT OPERATION.' FLUIDS. AND PAGE %102-
- THERMODYNAMICS ANSWERS -- SEGUDYAH -88/12/12-REGION II m
. ANSWER' 5.25 (1.00)
To recove boric acid that has precipitated on upper core surfaces l( 0. 5 )
To t rminate any boiling or steam formation in the upper head region (0.5)
REFERENCE OPL271CO23 p.-16 l( 3. 2/3. 5 ) '
OO5000K402 ...(KA*S) 1 2
l
. PLANT SYSTEMS' DESIGN. CDNTROL. AND INSTRUMENTATION PAGE X103 iNSWERS -- SEQUOYAH -88/12/12-REGION II iNSWER 6.01 (1.00)
@FERENCE
'QN Lcccan Plan , OPL271C019, pp. 4, 5 & 19.
Loccon Objective, OPL271C019, #6.
B: 010000K401 (2.7/2.9)
?10000K401- ...(KA*S)
ZSWER 6.02 (1.00)
).
@FERENCE
'QN Quantion and Answer bank, Prelicense Exam, 9/25/87 B: 025000 GOO 4 (4.0/4.3) 85000 GOO 4 ...(KA*S)
ZSWER 6.03 (2.00) a h.p5 ccch)
- 1) Hot leg loop 1
- 8) Hot leg = loop 3
- 8) RHR downstream of Heat Exchangers O) Containment Atmosphere ufA C l
[5) cz,swa .uc~T M mas eetc t-cr -
@FERENCE LQN Quoation and Answer Bank, Prelicense Exam 10/16/87 B: 000011G006 (4.2/4.4)
@OO11 GOO 6 ...(KA*S) t j
PLANT SYSTEMS DESIGN, CONTROL. AND INSTRUMENTATION PAGE 7.104 ISWERS -- SEQUOYAH -88/12/12-REGION II
- SWER 6.04 (2.00)
).5 ccch)
>) Containment pressure > 9.5 psig
LOccon Objective, OPL271CO24, #19.
): OO5000K306 (3.1/3.2)
)5000K306 ...(KA*S)
EWER 6.05 (2.00)
(0.5 each)
(1) The power range low flux trips will not be automatically reinstated.
(2) The intermediate flux trips will not be automatically reinstated.
(3) Inability to reenergize the SRM's when below P-6 (because P-10 disables the power supply to the SRM's.)
(0.25 each)
(1) By the status lights for the P-10 bistable switches.
(2) By the status lights for the P-10 permissive interlocks.
@ERENCE Z Oparator Certification Training, Solid State Protection j Otcm, Para. X.F. j M Loccon Plan, OPL271C048, p. 27. !
@ IE notice 86-105.
)012OOOK401 (3.7/4.0) 3000K401 ...(KA*S) l 1
PLANT SYSTEMS' DESIGN CONTROL. AND INSTRUMENTATION 'PAGE %105 NSWERS -- SEQUOYAH. -88/12/12-REGION II NSWER 6.06 (1.00)
@ CHANGE.(0.5) since the bellows only separates the RCS/Pzr Otcr from the DP cell internals to prevent flashing of the pfarcnce leg fluid on transients. .(0.5)
@FERENCE QN Loccon Plan,'Pzr and Control System, OPL271C019, pp. 10-12.
Leccon Objective, OPL271C019, #2.e.
$s. 011000K604 (3.1/3.1) 61000K604 ...(KA'S)
ZSWER 6.07 (1.00)
@FERENCE QN Lcocon Plans, Solid State Protection System, OPL271C048, BguroLO3.
@s 012OOOK603-(3.1/3.5) 62OOOK603: ...(KA*S)
ZSWER 6.08 (1.00)
@FERENCE EN'Loccon Plans, OPL271C046, p. 8.
2: OO1000K403 (3.5/3.8)
@lOOOK403 ...(KA*S)
PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE X106
!SWERS -- SEQUOYAH -88/12/12-REGION II EFER 6.09 (2.00)
).5 occh)
TRUE TRUE FALSE FALSE PERENCE Z Loccon Plan, 6.9 Shutdown Boards, OPN220.021, pp. 4, 6 & 13.
Loacon Objective, OPN220.021, #1, 2, 3 & 9.
1: 062OOOK401 (2.6/3.2)
$O00K401 ...(KA*S)
@WER 6.10 (2.50) ay fivo; O.5 each)
) Containment Isolation phase A l} Focdwater isolation
') Diocel generator emergency start
>) Racetor Trip
)) Snfcguard Sequence
- 1) Containment Ventilation p ,2 8 4W -)
') Stcrt EGTS & ABGTS (d * (**^'/
i Gm./d x%.n e Ah FERENCE Z Syntcm Description, Solid State Protection System, L271C048, p.31.
): 012OOOK406 (3.2/3.5)
OO6020K406 (3.9/4.2)
SOOOK406 006020406 ...(KA*S) 1 1
). PLANT SYSTEMS DEST.GN. CONTROL, AND INSTRUMENTATION PAGE X107 NSWERS -- SEGUOYAF -88/12/12-REGION II ZSWER -6.11 (1'.25)
'ha otationary (receives a higher voltage signal) and movable grippers aargizo and'the lift coil deenergizes (0.75); ROD URGENT FAILURE 11crm , red lamp at power cabinet actuates, or urgent failure lamp on Egic cabinet lights 4(0.5 for 2 of 3.)
c4 ( cow /L:>' Room) Rob cMTAo' da rm u%wr fmmn--
@FERENCE EN " Rod Control' System", OPL271C046, p. 12.
2: -OO1050A201 (3.7/3.9)
@1050A201 ...(KA'S)
ESWER 6.12 (1.50)
@.5 oach)
- s. . Output remains at the low and of the gain
). O degree mismatch
- . Goss high to 72 steps per minute GFERENCE
!GN Raqual 1986, week 3, day 4, " Instrument Failures"; " Rod e
'ntrol" 2: OO1000K403
)@lOOOK403 ...(KA*S)
- i. PLANT SYSTEMS DESIGN, CONTROL. AND INSTRUMENTATION PAGE */,108 NSWERS -- SEUUDYAH -88/12/12-REGION II iNSWER 6.13 (1.50)
@.375 cach)
- 1) Phnse A Containment Isolation signal from either Unit
- 8) High Radiation signal from fuel handling b1dg area rad conitors (this can be split into RM-90-102, train.A, and RM-90-103, train B)-
$) High Radiation signal from Aux Bldg exhaust vent rad monitors O) High Temperature in Aux Bldg Supply Fan Suction
@FERENCE ENP Syo Descr. 4.4, " Coat Air Purif and Cleanup Sys," pp. 15.
'QN Dwg 45N630-4 2: OOOO60G010 (3.8/3.8)
$O060G010 ...(KA*S) 6.14 l ZSWER (1.00)
- 1. 1 i
@FERENCE DN Rnqual 1986 week 2, day 1, " Major Modifications," pp. 8 & 9. ;
2: '061000K404 (3.1/3.4) ]
61000K404 ...(KA*S) l 1
i I.
-_._____-_-_.m__ _ - _ _ _ _ - _ _ _ _ _
- i. ' PLANT SYSTEMS' DESIGN. CONTROL.'AND INSTRUMENTATION PAGE 7.109 NSWERS.-- SEQUOYAH .-88/12/12-REGION II NSWER 6.15 (1.50) 0.5 oceh)
L. . ARM AND ACTUATE ~
1 ARM AND ACTUATE
- . ARM ONLY
@FERENCE-bricy SD, " Steam D. S System," pp.-23-28 ;
LQNP Syctem Descrip. "dteam Dump System," pp. 6-8. l 8: 041020K411 (2.8/3.1) 041020K414-(2.5/2.8) 041020K417 (3.7/3.9) 01020K411 041020K414 041020K417 ....(KA*S)
ZSWER 6.16 (1.00) s.
@FERENCE
@s 012OOOK603 (3.1/3.5)
)12OOOK603 ...(KA*S)
_ = _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _
I y PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE X110
]
iNSWERS -- SEQUOYAH -88/12/12-REGION II RSWER 6.17 (1.50)
Chun the proper logic is met), the logic card shuts off the
.5VDC going to the UV driver card (0.3) which will cause the UV
)rivcr to cut off the 4SVDC gning to the Reactor Trip breakers i
@.3) cnd the UV trip coil will unlatch the Reactor Trip breakers
@.3), cnd the shunt trip coil (0.3) also receives 125 VDC power signol (0.3).
@FERENCE
@L271C048, p. 14.
@ B5 l 2: 012OOOK603 (3.1/3.5)
>12OOOK603 ...(KA'S)
MSWER 6.18 (1.00)
@.5 ccch)
- 1. Tha tripped FWP's condenser inlet and outlet valves will close.
). Itc condenser valves will remain open regardless of load.
@FERENCE FL271CO34, p. 5.
@ 03.
2 059001A3OS (2.4/2.7)
G9001A305 ...(KA*S)
PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATIQM PAGE %111 EWERS -- SEQUOYAH -88/12/12-REGION II SWER- 6.19 :
(1.50)
Rodo run in at' maximum. speed-(0.25), unit trips on low pzr procsure (0.25), SI on low pzr pressure (0.25)
Rods run in at max speed until the rate comparator signal d:ccys (0.25), unit trip on low pzr pressure (0.25), SI on iow pzr pressure (0.25)
@ERENCE L27C046, Rod Control, p. 14. :
)B-3
): OO1050K501 (3.3/3.6) 015000K103 (3.1/3.1)
>1050K501 015000K103 ...(KA*S)
EWER 6.20 (1.50)
).5 occh)
,) Auctioneered high Tave
!) Auctioneered high delta T i) - RPI (from P/A converter) i FERENCE L271C046, p. 15.
)B-3
):OO1000 GOO 7
>1000 GOO 7 ...(KA*C)-
I l
1
, I
m u . .
- 6. PLANT-SYSTEMS DESIGN.' CONTROL. AND INSTRUMENTATION PAGE 7,112 ANSWERS - SEQUOYAH -88/12/12-REGION II ANSWER 6.21 (1.25) 8.. It provides supply of voltage to the Reactor Coolant Pumps O< 3'/(Or25)"to keep' them running f or at ' least 30 seconds (Or25)M o $ 5 in ' the- event of a - blackout ("C.-25) and cooldown._by_ natural-ag '
circula tion _0CL.25 ),w 8.- Prevents overspeed of the turbine (0.25)
CCEFERENCE OPL271C019, p.19 '
?.O. V. E-2 (2.9/3.1)
@45050K301 ...(KA*S)
)
1
PROCEDURES"- NORMAL.' ABNORMAL, EMERSENCY AND PAGE X113
. 88RIOLOGICAL CONTROL' NSWERS:-- SEQUDYAH' -88/12/12-REGION II NSWER- 7.01 -(1.00)
)..-
EFERENCE .
@s :194001A102 (4.1/3.9) 94001A102 ...(KA*S)
NSWER .7.02 ('1.00)'
@FERENCE.
@L273C101 p.5, E-O
@s OOOOO7A202 (4.3/4.6)
@OOO7A202 ...(KA*S)
ZSWER 7.03 '(1.00)
L.
@FERENCE botinghouse Background'Information for ECA-0.0 users Guide for
@G's cnd Background Documents, pp. 17 & 18.
@s .OOOO55K302 (4.3/4.6)
@OO55K302 ...(KA'S)
l F. PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE %114 1 RADIOLOGICAL CONTROL lNSWERJ -- SEGUOYAH -88/12/12-REGION II 1
J ZSWER 7.04 (2.00)
- 1. IF pzr level is falling, THEN start charging pumps as !
nccessary to maintain level. (1.0)
). IF loss of pzr level is imminent, THEN trip the reactor, initiate SI, and go to E-0. (1.0)
@FERENCE DN ADI-24, Steam Generator Tube Leak, p. 1.
E): OOOO37G010 (3.7/3.9)
@OO37G010 ...(KA*S)
- ZSWER 7.05 (2.00)
,. Subcooling > 40deg F
?. RCS pressure - stable or increasing ~
- s. Pzr level > 20%
@FERENCE DN , E-FOP, SI Termination E): OOOOO9G012 (4.1/4.3)
)@OOO9G012 ...(KA*S)
'. PROCEDURES - NORMAL.-ABNORMAL. EMERGENCY AND PAGE 7.115-
. 86DJOLOGICAL CONTROL '
NSWER3'-- SEGUOYAH -88/12/12-REGION II ISWER 7.06 (1.00)
(Povcnto localized power peaking (0.25), minimizes Xenon Ccilletions.(0.25), and withdrawing the rod at normal rate can Ccd to fuel cladding damage in the region of the dropped rod
@.5).
@FERENCE
!GN AOI-2, Malfunction of.the Rod Control System, p. 7.
2: OOOOO3 GOO 3.(3.3/3.8)
@OOO3 GOO 3 ...(KA*S)
ZSWER 7.07 (1.00) t.
@FERENCE ENP'SOI62.1B, pp. B & 9.
Es . OO4020A404 (3.3/2.9) 84020A404 ...(KA'S)
RSWER 7.08 (1.00)
)vor-power del ta T.
@FERENCE ENP'ADI-3B, p. 4. ,
Ba 012OOOK603 (3.3/3.6)
>12OOOK603 ...(KA'S) e
~
ij." PROCEDURES'- NORMAL.' ABNORMAL; EMERGENCY ANDL PAGE */,116 RADIOLOGICAL CONTROL'
$5WERS,-- SEQUOYAH- -88/12/12-REGION II NSWER' ;7.09 (2.50)-
dny_5p'O.50 each)
- f. ;~ Candenser exhaust radiation-high'
).. S/G blowdown radiation high i Abnormal RADCON survey of main steam lines
).- Abnormal RADCON survey'of-'S/G blowdown lines
- 1. : Abnormal. chemistry-lab sample of S/G blowdown
'.. Increasing-S/G 1evel with-no AFW flow
@FERENCE 7L.2739835, p. 3. .
@.01.
2: OOOO38G012 (3.8/4.0)
SOO38G012- ...(KA*S)-
ZSWER. 7.10 (1.00)
- 6. .To' prevent thermal lock up (0.5).
).- LTo prevent bank overlap malfunction (0.5).
8FERENCE EN GOI-2, p. 14.
@s L 001000G010 (3.3/3.5)
@loooGolo ...(KA*S)
' PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE X117 RADIOLOGICAL CONTROL
- SWERS -- SEQUOYAH -88/12/12-REGION II EWER -7.11 (1.50) ould not use ES-3.3, unless (0.5 each):
) ES-3.2, " Post SGTR Cooldown Using. Backfill" is inadequate l} M negement Approval
- ) 10CFR2O Offaite dose evaluation completed.
FERENCE 53.3 6 2730835, p. 32.
l: 194001A102 (4.1/3.9) 0001A102 ...(KA'S)
@WER 7.12 (1.00)
> provant overheating of the turbine blading adag:2 (0.25) and lack of ventilation ( Or2&)). '
5fhause-ofr'##
Ms g%p FERENCE Z GOI-2, p. 5.
1: 045000G010 (2.6/2.8) 600G010 ...(KA'S)
EWER 7.13 (1.00) eify main steam warming valves are closed. (1.00)
PERENCE R GOI-2, p. 7.
): 059000G010 (2.9/2.9)
@OOOG010 ...(KA*S)
PROCEDURESL- NORMAL. ABNORMAL.' EMERGENCY AND PAGE.%110.
RADIOLOGICAL CONTROL-l CWERS -- SEGUOYAH -88/12/12-REGION II J (1.00)
BWER' f7.14 O Shift Engineer (0.25) .shall initial, date and w-it's N/A
).25 ) cnd give a brief explanation of why.the. precaution cannot
> complied with in 'the lef t hand margin besitte the precaution
). 5 ) ,
@ERENCE-QQ GOI-5, p. 2.
): .194001A102 (4.1/3.9)
@WER 7.15 (1.00) ecdiatoly borate (0.5) at > or.= 10gpm to clear the alarm (0.5)
PERENCE.
Z GOI-5, p. 2.
is -OO1000G010 (3.3/3.5)
OO1000A405-(3.7/3.7) 6000G010 OO1000A405 ...(KA'S)
@WER 7.16 (1.00) 60 in to prevent governor valve #3 vibration (1.0).
@ERENCE ZEGOI-5, p. 4.
!: 045000G010 (2.6/2.8) 045000A305 (2.6/2.9) ,
I GOOOG010 045000A305 ...(KA*S)
&- PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE %1191
. RADIOLOGICAL CONTROL i I
kNSWER3:-- SEQUOYAH. -88/12/12-REGION II i bNSWER' 7.17 (1.50)
'O'.5 ccch)
).- False-M. 'True.
l . -- True-8FERENCE- ,
IGN GOI-6, pp. 2 & 4. ,
Dr 191001K106 (3.3/3.7) !
91001K106 ...(KA*S) i ZSWER 7.18 (2.00)
- 1) R3 rove the locking device and attempt to move the operator in the closed direction (0.5). .
- 2) Rainstall the locking device and verify it is securely locked (0.5). j
- 3) Independent verification of the locking device is required !
(0.5), and a third block drawn which will be used for i independent verification of the locking device (0.5).
BFERENCE EN GOI-6, p. 7.
2: 191008K108 (3.4/3.4)
@1008K108 ...(KA'9) ,
____m_ -
. _ _ _ . . _m___-___._ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ .
72 PROCEDURES -- NORMAL. ABNORMAL 2_ EMERGENCY AND PAGE* */,120
'o RADIOLOGICAL CONTROL kNSWERS -- SEQUOYAH- -88/12/12-REGION II bNSWER 7.19 (2.00) l0.3 'for technique; ' O.2 for order) l1) Increase injection flow (1) Depressurize S/G's.
l'2 ) D3 pressurize S/G's OR (2) Increase injection flow l3) Start RCP's (3) Start-RCP's
?4 ) . Opsn RCS vent paths (PORV's) (4) Open RCS vent paths.
(PORV's)
HFERENCE FR-C.1 94: OOOO74K103 (4.5/4.9)
- oOO74K103. ...(KA*S) 2SWER 7.20 (1.00)
Fhic allows S/G pressure protection without having to depend on sina codo safeties-(0.75), which could lift and not reseat causing 3rf unisolable steam leak (0.25).
4 2EFERENCE 90etinghouse Background Document, (TPT OBNK E-160)
U4: OOOO38K306 (4.2/4.5)
DOOO38K306 ...(KA*S) l I
l i
b PROCEDURES - NDRMAL. ABNORMAL. EMERGENCY ANQ: PAGE %121 RADIOLOGICAL CONTROL J
ZSWERS -- SEQUOYAH -BB/12/12-REGION II 1 NSWER 7.21 (1.50) I
'l
@P'o will keep 2 phase. flow mixture (0.75) and the PORV's will )
et ba oble to release as much steam (energy) (0.75) t i
- OR Jigh2r pressure will reduce SI flow (0.75) and increase inventory eco out PORV*s (0.75).
@FERENCE botinghouse Background Document
@-H.1 p.2 B: OOOO74EK308 (4.1/4.2)
]@OO74K308 ...(KA*S)
ZSWER 7.22 (2.00)
- s. - (0.25 each)
(1) Close MSIV's and bypasses
-(2) Trip turbine from main turbine front standard (3) Stop and pull-to-lock both EHC pumps at the pump control station
). (0.25 each)
(1) Decrease Turbine load (2) Verify auto rod insertion or manually insert control rods to maintain Tavg at Tref.
(3) Open reactor trip breakers at MG set (ROOM Aux. Bldg. 759)
(4) Open breakers to control rod MG sets at 480V Unit Boards A and B (5) AND notify inst to close P-4 contact.
@FERENCE m-S.1 2- OOOO29EK312 (4.4/4.7) 4
@OO29K312 ...(KA*S)
F.
ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE X122 NSWERS -- SEQUDYAH -88/12/12-REGION II' NSWER 8.01_ (1.00) d.;(1.OO)
@FERENCE EN TS Section 1.0 8a7/4.1)
@1000G005 ...(KA*S) ,
ZCWER, 8.02 (1.00)
(1.00)
@FERENCE
- GN , TS's 3.05, 3.1.2.1, 3.1.2.2, 3.1.2.3, 3.1.2.4, 3.8.1.1, 3.8.1.2 l 8.3/3.8) (3.4/3.9)
@4000 GOO 5 064000 GOO 5 ...(KA'S) 1 ZSWER- 8.03 (1.50) ;
- s. MTC within analyzed range (0.3)
- i. Trip. instrumentation within operating range (0.3)
. Above P-12 setpoint (0.3) l.- Pzr capable of being operable (0.15) with a steam bubble (0.15)
). HRx vessel above its RT (NDT) i
@FERENCE at, TS p 3/4 1-17 ZP TS B3/4.1.1.4 ;
8 9/3.8)
@lOOOGOO6 ...(KA'S) i
= )
_ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . __ _ _ a
). ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE %123 iN!WERS -- SEQUOYAH -88/12/12-REGION II 94SWER 8.04 (1.00)
Oc dictely initiate and continue boration (0.3) at greater than or qual to 10 gpm (0.3) of a solution containing greater than or equal e 20,000 ppm or equivalent (0.3) until the required SHUTDOWN MARGIN O rostored (0.1)
@FERENCE EN TS 3.1.1.2 0.1/3.9)
@<OO1A102 ...(KA'S)
ESWER 8.05 (1.00)
@.5 ccch)
- s. tha unit shall be placed in a least Hot Standby within one hour.
). tha NRC Operations Center shall be notified (by telephone as soon as possible and in all cases) within one hour.
@FERENCE
@NP TS 6.7 8.4/3.9)
@OOO1A011 ...(KA'S)
ESWER B.06 (1.00)
Ompsroture does not exceed that temperature allowable (for continuous ety rating specified) for equipment and instrumentation. (1.00)
@FERENCE
@NP TS B3/4.6.1.5
@NP Rcqual. Training Inst Notes 8.8/4.3)
@OO25AOO6 ...(KA*S)
I
)
. _ _ _ _ - - - - _ - - _ _ _ - - _ ._ . _ - . _ _ ._ i
. ' ADMINISTRATIVE PRQREDURES. CONDITIONS. AND LIMITATIONS PAGE */,124 i
kSWERS -- SEGUOYAH -88/12/12-REGION II ZSWER 8.07 (1.00) 1
@D ~(for each operable channel) must be monitored and logged (0.5) at Ocot once per hour (0.5) (for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least once per
@ minutes thereafter).
@FERENCE QNP TS 3/4.2.1 9.6/4.0)
@1050A206 ...(KA*S) i kSWER 8.08 (1.00)
If the replacement fuse blows a second time), The appropriate cintencnce section must'be notified to check the circuit before erthar fuse replacement occurs. (1.00)
@FERENCE EN AI-16, Fuse Control, p. 7 8.6/3.7)
@4001107 ...(KA*S)
ZSWER 8.09 (1.50)
- 1. - Op2 rations Superintendent (0.5)
). Tcchnical Support Superintendent (0.5)
. Maintenance Superintendent (0.5)
GFERENCE-
'GN AI-19 P. 9
...(KA'S)
C ADMINISTRATIVE PROCEDURES. CONDITIONS.-AND LIMITATIONS PAGE %125 NSWERS -- SEQUOYAH -88/12/12-REGION II i
NSWER' 8.10 (1.00)
Entcct those persons normally responsible for providing direction and Epport.-(1.00)
@FERENCE'
)I-22 p. 1 8.6/3.8)
@4001A105 ...(KA*S)
SDSWER . 8.11 (1.00) t provides assurance that system configuration has been returned to erect (by the section responsible for performing the work prior to be rolosse of~the clearance). (1.00)
@FERENCE 11-37, p. 2 8.7/4.1)
@4001K102 ...(KA*S)
COSWER 8.12 (1.50)
@.5 cach 3 of 4)
- 1. Radiation exposure
). Parsonnel safety
- l. Plcnt emergency considerations
). Other operational considerations
@FERENCE 11-37, P. 3 2.8/3.4) (3.3/3.4) (3.1/4.4) 94001K103 194001K104 104001A116 ...(KA*S) l
_ _ _ - - - - - - - - - _ _ - _ - - _ --- - - -_ _ - _ _ . -- -A
ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS- PAGE 2126
- BWERS - SEQUOYAH -88/12/12-REGION II EWER- 8.13 (1.00)
> cncuro Psat for S/G's is reduced to below atmospheric relief valve Gpoint (0.5) to, ensure offsite release is limited in the event of a Ocm'ganarator tube rupture.(0.5)
@ERENCE 2.TS 3.4.8 p. 4/3 A-19 B3/4 4-5 2 Excm Bank, Test 12/5/86
@O30A214 ...(KA'S)
@WER 8.14 (1.00)
Gplo clearance - description of equipment removed from service.(0.5) eplox clearance - include a copy of the clearance sheet in the configuration log. (0.5)
@ERENCE R) A I ? - 58 p. 15 Z Excm Bank, Test, 12/5/86
- .7/4.1) 0001K102 ...(KA*S)
EWER B.15 (1.50) above the zerc power rod insertion limit. (0.5 each) muot be critical with in + or - 1000 pcm (in equivalent rod atops) of the ECP.
Balow the withdrawal limit to ensure a negative MTC.
@ERENCE II-2, p. 15 2 Quantion Bank, Exam 11/7/86 6000G001 ...(KA*S) www-_-_-__.-_-__-__--_- _ _ - . . - _ . - .
rs ' ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE X127 NSWERS -- SEGUOYAH -88/12/12-REGION II ZSWER 8.16 (1.00) l (1.00)
@FERENCE EN.A01-34
!GN Bcnk, Exam 11/7/86 0.1/.4.4).
@OO24k301 ...(KA*S).
ZBWERI 8.17' (2.00) s.- Op n Reactor trip-breakers (0.5 each)
).- Corate RCS-L. After Reactor Trip, Verify turbine trip J. LEncure all AFW pumps are running.
@FERENCE Q-S.1 p. 2,3,4
!GN Qusation Bank, Exam 11/7/86 0.5/4.5) 80029G010 ...(KA*S) 4
)2__ ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 2128 1
lNSWERJ -- SEQUOYAH -88/12/12-REGION II i
l ZSWER 8.18 (1.50) !
e of 8 8 0.25 each) 1
- s. Op3 rations section personnel specifically assigned to the unit centrol room.
). Ano'.stant Operations Group Manager
- . Oprrations Group Manager
- l. Op3 rations Superintendent
). Plcnt Manager
. NRC Personnel
- 1. - Shift Technical Advisor
). Radiological Control Supervisor.
@FERENCE EN AI-30 p. 21 8.1/3.4) 04001K105 ...(KA*S)
/. c ap44 RSWER 8.19 (.1-<507
- s. Plcnt Manager (0.5) ;
). Th3 UD is required to have at least a 4 hr break on the new unit i prior to assuming the shift. (0.5) This-period-wil-1-be used-to ^^' j review the cur tent-units status ~and-to-become--f ami-liar-with M 1 plcnned-activi. ties _on the-newly-assigned-unit.-(0.5)~- /s f
@FERENCE
!GN AI-30 p. 37 8.5/3.4) !
@4001A103 ...(KA'S)
ZSWER 8.20 ('. 00) j
@ 3.4.6.2.a.
@FERENCE bquoych Tech Specs. 3.4.6.2.a. (2.6/3.8)
$2000g006 ...(KA*S)
\
l l
_ - - - _ _ _ - _ - _ _ _ _ _ __ _ b
!. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS. PAGE %129 iNSWERS - SEQUOYAH -88/12/12-REGION II ZSWER. 8.21 .(1.00)
@ 3.3.2.a. . action '20. [*,3.3-4.7.b.2,3.3-5.12.a.]
@FERENCE tquoych TS 3.3.2.a.. (2.6/3.6,3.0/3.2)
G2OOOg006 062OOOg010 ...(KA*S)
ZSWER. 8.22 (1.00)
).
@FERENCE Oquoych TS 3.3.9. (2.1/3.3) l
@6000 GOO 6 ...(KA*S)
ZSWER 8.23 (2.00)
M A i.- 3.3-1. action 6#( W ) [0.5] and 3.3-2 action
). Initiate a procedural change to allow. returning the loop to carvice IAW AI-4.
@FERENCE Cquoych LER 88-36, TS 3.3.1/3.3.2, AI-47. (3.7/4.0,3.4/3.7,2.9/4.1) 62OOOk401 012OOOa203 012000g006 ...(KA'S)
ZSWER 8.24- (2.00)
- s. LCO 3.0.3
). you
@FERENCE Cquoych TS 3.0.3 and Section 6. (2.9/4.1) 62000g006 ...(KA*S)
SEQUOYAH 88/12/12 REGION II SECTION 5 5.01 1.00 5.02 1.00 5.03 1.00 5.04 1.00 5.05 2.00 5.06 2.00 5.07 1.50 5.08 1.50 5.09 2.00 5.10 1.00 5.11 1.00 5.12 1.00 5.13 1.00 5.14 1.00 5.15 1.00 5.16 1.00 5.17 1.50 5.18 1.50 5.19 1.50 5.20 1.50 5.21 1.00 5.22 1.00 5.23 1.00 5.24 1.00 5.25 1.00 31.00 SECTION 6 6.01 1.00 6.02 1.00 6.03 2.00 6.04 2.00 6.05 2.00 6.06 1.00 6.07 1.00 6.08 1.00 6.09 2.00 6.10 2.50 6.11 1.25 6.12 1.50 6 13 1.50 6.14 1.00 6.15 1.50 6.16 1.00 6.17 1.50 6.1B 1.00 6.19 1.50 6.20 1.50 l 6.21 1.25 )
)
30.00
_ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ . _ _ . _ _ _ _ . _ _ _.____ _ w
SEQUOYAH 88/12/12 REGION II l SECTION' 7 7.01 1.00 7.02 1.00 j 7.03 1.00 7.04 2.00 7.05 2.00 7.06 1.00 7.07 1.00 7.08 1.00 7.09 2.50 7.10 1.00 7.11 1.50 7.12 1.00 7.13 1.00 7.14 1.00 7.15 1.00 7.16 1.00 7.17 1.50 7.10 2.00 7.19 2.00 7.20 1.00 7.21 1.50 7.22 2.00 30.00 SECTION 8 0.01 1.00 8.02 1.00 8.03 1.50 8.04 1.00 8.05 1.00 8.06 1.00 8.07 1.00 8.08 1.00 8.09 1.50 8.10 1.00 B.11 1.00 8.12 1.50 8.13 1.00 B.14 1.00 B.15 1.50 B.16 1.00 8.17 2.00 B.18 1.50 B.19 1.50 8.20 1.00 8.21 1.00 8.22 1.00 8.23 2.00 8.24 2.00 30.00 Total 121.00
g,,--- - - . . - - - _ _ _ , . .,_ - - - . - , - , - - - , . _ , - - . . , , - - . .
r Y y 0- ,* ',
d? Ras,ci p.m. p
-u ;.LLL..uxuG'.25 L&xtdLO.:Ls.L.MLLL:. aw . . ~ = =: , u l'~ :: ,
Gku i-'L--i n . a.a_.L..c^.2. .in' ,
'S e
)4 ATTACHMENTS SEQUOYAH SRO EXAM REGION II 80/12/12 1
1 9
l 1
. . i
, /, I UNIT 2 ATTACHMENTS Table of Contents:
- 1. SQNP SOI 55-1-M-6 p. 17
- 2. T.S. 3.4.6.2, p. 3/4 4-18, 3/4 4-19, 3/4.4-20
- 3. T.S. 3.0.3, p. 3/4 O-1
- 4. T.S. 3.3.1, p. 3/4 3-1
- 5. T.S. TABLE 3.3-1, p. 3/4 3-2 thru 3/4 3-8
- 6. T.S. 3.3.2, p. 3/4 3-14
- 7. T.S. TABLE 3.3-3, p. 3/4 3-15 thru 3/4 3-23
- 8. T.S. TABLE 3.3-4, p. 3/4 3-24 thru 3/4 3-28
- 9. T.S. TABLE 3.3-5, p. 3/4 3-29 thru 3/4 3-33a
- 10. T.S. 3.3.3.9, p. 3/4 3-6B
- 11. T.S. TABLE 3.3-12, p. 3/4 3-69 thru 3/4 3-72
- 12. SQNP SI-2 p. 1, 6, 6a, Data sheet 4 l
l 1
l l
l l 1
T SQNP
- i SOI 55-1-N-6 l 1-XA-55-6A i . Page 17 oi 35 Rev. 3 i Process Protect ;
Racks Channel Annunciator I,ocation 17 Test Sequence Violated Setpoint: None i
- Origin: Protective process channel bistable
. not deliberately tripped with bistable trip switch before test-operate switch is placed in test.
Probable Cause: 1. Improper test switch operation sequence in the auxiliary instrument room. j t Aatomatic Action: 1. None Immediate Action: 1. Stop testing.
- 2. Place this channel in trip using the bistable trip switch.
Supplementary Action: 1. Review the procedure and determine where the
, sequence was violated before resuming the test procedure.
~
e .
References:
Prepared by: W. Muirhead ,
L2 W .
, q _
e' . 6 * * *' '
. " ar*
- l 1
OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to: . ,
I
- a. No PRESSURE BOUNDARY _ LEAKAGE, )
- b. 1 GPM UNIDENTIFIED LEAXAGE,
- c. 1 GPM total primary-to-secondary leakage through all steam generators and 500 gallons per day through any one steam generator, j
- d. 10 GPM IDENTIFIED LEAKAGE from the Reactor, Coolant Systes, and
- e. 40 GPM CONTROLLED LEAKAGE at a Reactor Coolant Systes pressure of f 2235 20 psig,
- f. 1 GPM leakage at a Reactor Coolant System pressure of 2235 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1. l l
APPLICABILITY: MODES 1, 2, 3 and 4 ACTION: (
- a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the followin'g 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, and leakage from Reactor Coolant System Pressure Isolation Valves, redu:e the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 h'ours and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. ,
- c. With any Reactor Coolant System Pr6ssure Isolation Valve leakige greater than the above limit, isolate the high pressure portion of the affected systes from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by ,
use of at least two closed manual or deactivated automatic valves, '
or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS i
4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within ;
i each of the above limits by:
SEQUOYAH - UNIT 2 3/4 4-18 i W
t lT e .~5"~~.7 {. c -
.Wh @e c . . q3. J.p g.:.je.'GI@lgfgy
} y
REACTOR COOLANT SYSTEM pMVEILLANCEREQUIREMENTS(Continued)
- a. Monitoring the lower containment atmosphere particulate radioactivity monitor at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- b. Monitoring the containment pocket sump inventory and discharge at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />,
- c. Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump seals when the Reactor Coolant System' pressure is 2235 1 20 psig at least once per 31 days with the modulating valve fully open. The provisions of Specification 4.0.4 are not applicable for entry into Mode 3 or 4.
~
- d. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />,
- e. Monitoring the reactor head flange leakoff system at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE pursuant to Specification 4.0.5, except that in lieu of any leakage testing requirements required by Specifica-tion 4.0.5, each valve shall be demonstrated OPERABLE by verifying leakage to be within its limit:
- a. At least once per 18 months.
- b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more and if leakage testing has not been performed in the previous 9 months.
- c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve.
- d. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following valve actuation due to automatic or manual action or flow through the valve.
The provisions of Specification 4.0.4 ace not applicable for entry into MODE 3 or 4.
3/4 4-19 SEQUOYAH - UNIT 2
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63-560 Accumulator Discharge (63-561
- Accumulator Discharge >
63-562 Accumulator Discharge 63-563 .
Accumulator Discharge 63-622 Accumulator Discharge 63-623 Accumulator Discharge i 63-624 Accumulator Discharge 63-625- -
Accumulator Oischarge 63-551- Safety Injection (CoTd Leg)63-553 Safety Injection (Cold Leg)63-557 Safety Injection (Cold Leg)63-555 Safety Injection (Cold Leg)63-632 Residual Heat Removal (Cold Leg)63-633 Residual Heat Removal (Cold Leg)63-634. Residual Heat Removal (Cold Leg)63-635 Residual Heat Removal (Cold Leg)63-641 Residual Heat Removal / Safety .
Injection (Hot Leg) ,63-644 Residual Heat Removal / Safety j Injection (Hot Leg) 1 63-558 Safety Injection (Hot Leg) {
63-559 Safety Injection (Hot Leg) /
Safety Injection (Hot Leg)63-543 63-545 Safety Injection (Hot Leg) j 63-547 Safety Injection (Hot leg)- <
63-549 Safety Injection (Hot Leg) l 63-640 Residual Heat Removal (Hot Leg)63-643 Residual Heat Removal (Hot Leg)87-558 UpperHeadInjection 87-559 Upper Head Injection 87-560 Upper Head Injection 87-561 Upper Head Injection 87-562 Upper Head Injection -
87-563 Upper Head Injection FCV-87-7* Upper Head Injection .
(charging header) i FCV-87-8* Upper Head injection g73 i (charging header) i FCV-74-1* Residual Heat Removal )
FCV-74-2* Residual Heat Removal .j
~ -
k >
- These valves do not have to be leak tested following manual or automatic actuation or flow through the valve. ',
SEQUOYAH - UNIT 2 3/4 4-20 Amendment No. 74
. September 21, 1988 ,
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3/4 LINITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY LINITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Condit' ions for Operation contained in the succeeding Specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.
3.0.2 Noncompliance with a Specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Conditions for Operation is restored prior to expiration.of the specified time intervals, completion of the ACTION requirements is not required.
3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within one hour action shall be initiated ,
to place the unit in a M00E in which the Specification does not apply by placing it, as applicable, in:
- 1. At least HOT STAN08Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,
- 2. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and
- 3. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for !
Operation. Exceptions to these requirements are stated in the individual Specifications.
3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the conditions for the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION requirements. This provision shall not prevent passage through OPERATIONAL N00ES as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual Specifications.
3.0.5 When a system, subsystem, train, component or device is determined to ,
be inoperable solely because its emergency power source is inoperable, or 1 solely because its normal power source is inoperable, it may be considered !
OPERA 8LE for the purpose of satisfying the requirement.s of its applicable Lietting Condition for Operation, provided: (1) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system (s),
subsystee(s), train (s), component (s) and device (s) are OPERABLE, or likewise satisfy the requirements of this Specification. Unless both conditions (1) and (2) are satisfied, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> action shall be initiated to place the' unit in a NODE in which the applicable Limiting Condition for Operation does not apply by placing it as applicable in:
- 1. At least H0T STAND 8Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,
- 2. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and .
- 3. At least COLD SHUTDOWN within tho subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
This Specification is not applicable in MODES 5 or 6.
- SEQUOYAH - UNIT 2 3/4 0-1 .
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3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor trip system. instrumentation channels and interlocks of Table 3.3-1 shall be OPERA 8LE with RESPONSE TIMES as shown in Table 3.3-2.
APPLICA8ILITY: As shown in Table 3.3-1.
ACTION:
As shown in Table 3.3-1.
SURVEILLANCE REQUIREMENTS 4.3.1.1.1 Each reactor trip system instrumentation channel and interlock shall be demonstrated OPERABLE by the performance of the CHANNEL CHECX, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-1. .
4.3.1.1.2 The logic for the interlocks shall be demonstrated OPERABLE prior to each reactor startup unless performed during the preceeding 92 days. The total interlock function shall be demonstrated OPERABLE at least once per 13 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.
4.3.1.1.3 The REACTOR TRIP SYSTEM RESPCNSE TIME of each reactor trip function shall be demonstrated to be within its limit at least once per 18 months.
Each test sna11 include at least one logic train sucn that both logic trains are tested at least once per 36 months anc one channel pc.r function such that all channels are tasted at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shewn in the " Total No. of Channels" column of Table 3.3-1.
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TABLE NOTATION I a
With the reactor trip system breakers in the closed position, the
, control rod dove system capable of rod withdrawal, and fuel in the reactor vesse's.
" The channel (s) associated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped condition. ,
The provisions of' Specification 3.0.4 are not applicableif .
"Hi hg voltage to detector may be de-energized above the P 6 (81o'ck of Source Range Reactor Trip) setpoint.
ACTION STATEMENTS I
ACTION 1 - With the number of OPERABLE channels one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable <
channel to OPERABLE stat ~us within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and/or open the reactor trip breakers.
ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
- a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. lR39
- b. The Minimum Channels OPERABLE requirement is met; however, ene additional channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> lR39 for surveillance testing per Specification 4.3.1.1.1.
- c. Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range, Neutron Flux trip setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />,
~
- d. The QUADRANT POWER TILT RATIO, as indicated by the remaining three detectors, is verified consistent with the normalized symmetric power distribution obtained by using the movable incore detectors in the four pairs of symmetric thimble locations at least.once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when THERMAL POWER is greater than 75% of RATED THERMAL POWER.
September 17, 1986 '
StQUOYAH - UNIT 2 3/4 3-5 Amendment No. 39 6
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... . TABLE 3.3-1 (Continued)
ACTION.3 - With the number of channels 0PERABLE-one less than required by.
the Minimus Channels OPERABLE requirement and with the THERMAL POWER level:
I a. Below the P-6 (Block of Source Range Reactor Trip) setpoint.
I restore the inoperable channel to OPERABLE status prior to L increasing THERMAL P0WER above the P-6 Setpoint.
- b. Above the P-6 (Block of Source Range Reactor Trip) setpoint, but below 5% of RATED THERMAL POWEA,.gestore the inoperable channel to 0PERABLE status prior.to increasing THERMAL l POWER above 5% of RATED THERMAL POWER.
- c. Above 5% of RATED THERMAL POWER, POWER OPERATION may continue.
- d. Above 10% of RATED THERMAL POWER, the provisions of Specification 3.0.3 are not applicable.
ACTION 4 - With the number of OPERABLE channels one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL p0WER level: ,
- a. Below.the P-6 (Block of Source Range Reactor. Trip).setpoint.
restore the inoperable channel to OPERABLE status prior to .
increasing THERMAL POWER above the P-6 Setpoint.
- b. Above the P-6 (Block of Source Range Reactor Trip) se operation may continue.
ACTION 5 - With the number of OPERABLE channels one less than required by the Minimum Channels OPERABLE requirement, verify compliance with the SHUTOOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable, within I hour and a* 1 east once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.
ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
- a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. lg39
- b. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> lR39 for surveillance testing per Specification 4.3.1.1.1.
ACTION 7 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed until performance of the next required CHANNEL FUNCTIONAL TEST provided the inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. l R3, September 17, 1986 SEQU0YAH - UNIT 2 3/4 3-6 Amendment No. 39
,e . *4 , e
TABLE 3.3-1 (Continued) i :'
- ACTION f ~- With less than the Minimum Number of Channels 0PE the interlock inoperable and verify that all affected channels of the functions listed below are OPERABLE or a priate ACTION statement (s) for those functions.pply the appro-be evaluated are: Functions to.
- a. Source Range Reactor Trip.
- b. Reactor Trip
' ~
Low Reactor Coolant Loop Flow ('2 loops)
Undervoltage Underfrequency
Pressurizer Low Pressure Pressurizer HQh Level
- c. Reactor Trip
^
low Reactor Coolant Loop Flow (1 loop)
- d. Reactor Trip Intermediate Range Low Power Range Source Range .
ACTION 9 - Deleted ACTION 10 - Deleted I ACTION 11 - Deleted ACTION 12 - With the number of OPERABLE channels one less th the Minimum Channels OPERABLE requirement,-be in at least HOT STANOBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> however, one channel may be bypassed for Rie up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surve;illance testing per Specification 4
- ' provided the other channel is 0PERABLE.
~
SEQUOYAH - UNIT 2 March 16, 1987 3/4 3-7 Amendment No. 46
. l
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- TABLE 3.3-1 (Continued)
ACTION 13 - With the number of OPERABLE channels one less than the Total Number of Channels and with the THERMAL POWER level above' the P-7 (enable reactor trips) setpoint place the inoperable channel in the tripped condition within 6 hourc, operation-may continue until performance of the next required CHANNEL' FUNCTIONAL TEST.
ACTION 14 - With the number of channels OPERA 8LE one less than required by the Minimum Channels OPERA 8LE requirement, be in at least HOT STAND 8Y within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
ACTION 15 - With one of the diverse trip' features (undervoltage or shunt trip attachment) inoperable, restore it to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or declare the breaker inoperable and apply ACTION 12. The breaker shall not be bypassed
- while one of the diverse trip features is inoperable !
except for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for performing maintenance to restore the breaker to OPERA 8LE status. R ACTION 16 - With the number of OPERABLE channels one less than the minimum channels operable. requirement, restore the inoperable channel to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or open the reactor !
trip breakers within the next hour.
l
\
March 16, 1987 l SEQUOYAH - UNIT 2 . 3/4 3-8 Amendment No. 46,
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- "'~7'**T*TYM**
L________----_------------. -- - - - - - - - - -
INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION -
3.3.2 The Engineered Safety Feature Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3'shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5.
APPLICABILITY: As shown in Table 3.3-3.
ACTION:
- a. With an ESFAS instrumentation channel or interlock trip setpoint less conservative than the value shown in the Allowable Values column of Tasle 3.3-4, declare the cnannel inoperable and apply the .
applicable ACTICN requirement of Taole 3.3-3 until the channel is l restored to OPERABLE status with the trip setpoint adjusted con-sistent with the Trip Setpoint value.
- b. With an ESFAS instrumentation cnannel or interlock inoperable, take ;
the ACTION shown in Table 3.3-3. '
SURVEILLANCE REQUIREMENTS 1
4.3.2.1.1 Each ESFAS instrumentation channel and interlock shall be demon-strated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION l and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies J snown in Table 4.3-2. I J
l 4.3.2.1.2 The logic for the interlocks snail be comonstrated OPERABLE curing I
! the automatic actuation logic test. The total interlock function shall be !
demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.
]
4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function l shall be demonstrated to be within the limit 3t least once per 18 months.
Each test shall include at least one logic train sucn that both logic trains are tested at least once per 36 months and one channel per function such that -
all channels are testeo at least once oer N times 18 months wnere N is the total numoer of reduncant :nannels in a specific ESFAS function as shown in tne " Total No. of Channels" Column of Table 3.3-3.
SEQUOYAH - UNIT 2 3/.4 3-14 1
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SEQUOYAH - UNIT 2 3/4 3-21a Amendment No.18, 55 December 31, 1987 k
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TABLE 3.3-3 (C+ntinued)
TABLE NOTATION
- T rip function may be bypassed in this MODE below P-11 (Pressurizer Block of Safety Injection) setpoint.
y,Injection)setpoint.
Trip function may be bypassed in this MODE
- Block below of Safety.P-1
,,,The channel (s) associated with the protective functions derived from th out of service Reactor Coolant Loop shall be placed in the tripped mode.
The provisions of Specification 3.0.4 are not applicable.
ACTION STATEMENTS ACTION 15 -
With the number of OPERABLE Channels one less than the Total Number of Channels, be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />; however, one channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance R55 l
. testing per Specification 4.3.2.1.1 provided the other channel is OPERABLE. R2 ACTION 16 -
With the number of OPERABLE Channels one less than the Total Number of Channels, operation may proceed until performance of the next required CHANNEL FUNCTIONAL TEST, provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
ACTION 17 -
With a channel assc:iated with an operating loop inoperable, restore the inoperable channel to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; howevar, one channel associated with an operating looo may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.3.2.1.1.
ACTION 18 -
With the number of OPERABLE Channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is demonstrated within I hour; one additional channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for R2 surveillance testing per Specification 4.3.2.1.1. 4 ACTION 19 -
With less than the Minimus Channels OPERABLE, operation may continue provided the containment ventilation isolation valves are maintained closed.
ACTION 20 -
With the number of OPERABLE Channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within a8 hours or be in at least HOT STANDBY within the i
next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. -
SEQUOYAH - UNIT 2 3/4 3-22 Amendment No. 55 December 31, 1987
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. TABLE 3.3-3 (Continued)
ACTION 21 -
With the number of OPERABLE Channels one less than the Number of Channels, STARTUP and/or POWER OPERATION may proce provided the following conditions are satisfied:
- a. The inoperable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.,channe,1 is placed in the tripped condition b.
The Minimum Channels OPERA 8LE requirements is met; however one additional channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.3.2.1.1.
ACTION 22 - It2 With less than the Minimus Number of Channels 0PERA8LE, declare, the interlock inoperable and verify that all affected channels of the functions listed below are OPERA 8LE or apply the
_ appropriate ACTION statement (s) for those functions. Functions to be evaluated are:
- a. Safety' Injection Pressurizer Pressure
- b. SafetyInjection High Steam Line Flow Steam Line Isolation High Steam Line Flow Steam Dump
- c. Turbine Trip Steam Generator Level High-High Feedwater Isolation Steam Generator Level High-High ACTION 23 -
With the number of OPERABLE channels one less than the Total Number of Channels, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to 2' hours for surveillance testing per Specification 4.3.2.1.1. g3~ I 8
ACTION 24 -
With the number of OPERABLE channels one less than the Total Number of Channels, restore the incperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STAN08Y within.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
ACTION 25 -
With the number of OPERABLE channels one less than the Total Number of Channels, restore the inocerable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or oeclare the associatea valve inoperable and take the ACTION required by Specification 3.7.1.5.
4 SEQUOYAH - UNIT 2 3/4 3-23 3%",gntg.g
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' ~ '
ENGINEERED SAFETY FEATURES RESPONSE TIMES -
INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONOS l
- 1. Manual {
- a. Safety Injection (ECCS) Not Applicable Feedwater Isolation Not Applicable Reactor Trip (SI) Not Applicable Containment Isolation-Phase "A" Not Applicable j Containment Ventilation Isolation 'Not Applicable I Auxiliary Feedwater Pumps Not Applicable Essential Raw Cooling Water System Not Applicable Emergency Gas Treatment System Not Applicable
- b. Containment Spray Not Applicable Containment Isolation-Phase "B" Not Applicable Containment Ventilation Isolation Not Applicable Containment Air Return Fan Not Applicable
- c. Containment Isolation-Phase "A" Not Applicable Emergency Gas Treatment System Not Applicable Containment Ventilation Isolation Not Applicable
.s
- d. Steam Line Isolation Not Applicable
- 2. Containment Pressure-High
- a. Safety Injection (ECCS) < 32.0(1) 147
- b. Reactor Trip (from SI) 13.0
- c. Feedwater Isolation <8.0(2)
- d. Containment Isolation-Phase "A"(3) 118.0(8)/28.00) )
- e. ' Containment Ventilation Isolation Not Applicable
- f. Auxiliary Feedwater Pumps 160(11)
- g. Essential Raw Cooling Water System 165.0(8)/75.0(9)
- h. Emergency Gas Treatment System 138.0(9)
SEQUL.'AH - UNIT 2 3/4 3-29 Amendment No. /f7// 68 August 5, 1988, i . .
~ * ~ * - - ' ' ~ ~ **'
,,, p% += - - - e, - = . _ - -
gy wr**7- . r* - * * **- -* ~e- ~'
_.- 1 TA8LE 3.3-5 (Continued) l ENGINEERED SAFETY FEATURES RESPONSE TIMES f
I INITIATING SIGNAL AND FUNCTION RESPONSE TINE IN SECONOS t '3. Pressurizer Pressure-Low
- a. Safety. Injection (ECCS)
I']
132.0(1)/28.0(7) f
- b. Reactor Trip (from SI) 1 3.0 l
- c. Feedwater Isolation' < 8.0(2) {
- d. Containment Isolation-Phase "A"I3) [18.0(8) j
- e. Containment Ventilation Isolation Not Applicable
- f. Auxiliary Feedwater Pumps 160(11) R61
- g. Essential Raw Cooling Water System 165.0(8)/75.0I8) !
- h. Emergency Gas Treatment System 128.0(8)
- 4. Differential Pressure 8etween Ste'am Lines-High R4'
- a. Safety Injection (ECCS) 128.0(7)/28.0(1)
- b. Reactor Trip (from SI) <3.0
- c. Feedwater Isolation 8.0(2)
- d. Containment Isolation-Phase "A"(3) 18.0(8)/28.0(9)
~
t
- e. Containment Ventilation Isolation Not Applicable
- f. Auxiliary Feedwater Pumps <60(11) R6f
- g. Essential Raw Cooling Water System 65.0(8)/75.0(9)
- h. Emergency Gas Treatment System I8) 138.0
- 5. Steam Flow in Two Steam Lines - High Coincident with T --Low-Low
- a. S ty Injection (ECCS) RU 130.0(7)/30.0(1)
- b. Reactor Trip (from SI) < 5. 0 c, .Feedwater Isolation <10.0(2)
- d. Containment Isolation-Phase "A"(3) {20.0(8)/30.0(I)
- e. Containment Ventilation Isolation Not Applicable
- f. Auxiliary Feedwater Pumps <60(11) R6E'
- g. Essential Raw Cooling Water System 67.0(8)/77.0(') -
- h. Steam Line Isolation < 10. 0
- 1. Emergency Gas Treatment System [40.0C8) ,
SEQUOYAH - UNIT 2 3/4 3-30 Amendment No. /M// 68 August 5, 1988-I.D9?#'Of;Wh5' k ? ^
~'
TABLE 3.3-5 (Continued) '
l ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION
- RESPONSE TIME IN SECONOS
- 6. Steam Flow in Two Steam Lines-High
, coincident with Steam Line Pressure-Low l
"'I l a. SafetyInjection(ECCS) 1 28:0I7)/28.0f1)
- b. Reactor Trip (from SI) < 3.0 '
- c. Feedwater Isolation 8.0(2)
- d. ContainmentIsolation-Phase"A"(3) [18.0(8)/28.0II)
- e. Containment Ventilation Isolation Not Applicable R68
- f. Auxiliary Feedwater Pumps -
160(11)
- g. Essential Raw Cooling Water Systes 5 65.0(8)/75.0(8)
- h. Steam Line Isolation < 8.0
- 1. Energency Gas Treatment Systes [38.0(8) .
- 7. Containment Pressure--High-High . It51
- a. Containment Spray < 208(9) i R73
- b. ContainmentIsolation-Phase"B"I12) 1 65(8)/75(9)
- c. Steam Line Isolation ~$ 7.0 l
- d. . Containment Air Return Fan > 540.0 and 1660 R55
- 8. Steam Generator Water Level--High-High l
- a. Turbine Trip i 2.5
- b. Feedwater Isolation i 11.0(2)
- 9. Main 5' as Generator Water Level -
Low-Low
- a. Motor-driven Auxiliary 1 60.0 Feedwater Pumps (N R68
- b. Turbine-driven Auxiliary 1 60.0 Feedwater Pumps (5)(11) ,
September 9. 1988 SEQUOYAH - UNIT 2 3/4 3-31 Amendment No. M ,5r, yt, N 73 ;
T ~ ~;; g.y;;. . F '"
. , . . g;qq.59.>g.y.7.g. , ;
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. . . TABLE 3.3-5 (Continued) l ,
l .
ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONOS
- 10. Station Blackout -
R68
- a. Auxiliary Feedwater Pumps 1 60(11)
- 11. Trip of Main Feedwater Pumps R68
- a. Auxiliary Feedwater Pumps 1 60(11) ,
- 12. Loss of Pcwer R68
- a. 6.9 kv Shutdown Board - Degraded < 10(10)
Voltage or Loss of ,
Voltage
- a. Automatic Switchover to Containment Sumo 1 250
- 14. Containment ource U r E.xhaust Radioactivity - Mich
- a. Containment Ventilation Isolation 1 10(6)
- 15. Containment Gas Monitor Radioactivity Hich
- a. Containmen- ventilation Isolation i 10(6)
- 16. Containment ? articulate Activity High
- a. Containment Ventilation Isolation 0) 1 10 R68 l
\
August 5, 1988 SEQUOYAH - UNIT 2 3/4 3-32 Amendment No. JK 68 i
~
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INSTRUMENTATION .
TABLE 3.3-5 (Continued)
TA8LE NOTATION (1) Diesel generator starting and sequence loading delays included. Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps, SI and RHR pumps.
(2) Using air operated valve (3) The following valves are exceptions to the response times shown in the l table and will have the values listed in seconds for the ini,tiating signals and function indicated:
Valves: FCV-26-240, -243 Response times: 8) 2.d.2221(C )f33(9) 3.d.
4.d. 21 (8) 5.d. 24 (8) 6.d. 21 (8)
Valves: FCV61-96, -97, -110, -122 -191, -192, -193, -194
- R8 Response times 2.d. 31(8) .
3.d. 32(8) 4.d. 31((8) 5.d.
6.d. 3134(8) 8)
Valve: FCV-70-143 .
Response times:
2.d.
3.d. 6261((8)f71(9) 8)
4.d. 61 8 f
(9) 5.d. 64 8 f
(9) 6.d. 61 8 j (9) .
(4) On 2/3 any Steam Generator (5) On 2/3 in 2/4 Steam Generator (6) Radiation detectors for Containment Ventilation Isolation may be excluded from Response Time Testing.
(7) Diesel generator starting and sequence loading delays not included.
RI.7 Offsite power available. Response time limit includes opening and closing of valves to establish SI path and attainment of discharge pressure for centrifugal. charging pumps. ,
(8) Diesel generator starting and sequence loading delays not included.
Response time limit includes operating time of valves.
(9) Diesel generator starting and sequence loading delays included. Response time Itait includes operating time of valves. u, M87 SEQUOYAH - UNIT 2 3/4 3-33 Amendment No. X 47 !
~. . . - .a m.,..~ :n ..- -
c ----~~-y'- - - " ~ "
TABLE 3.3-5 (Continued) . .
TABLE NOTATION (10) The response time for loss of voltage is measured from the time voltage '
is lost until the time full voltage is restored by the diesel. The- R68 response time for degraded voltage is measured from the time the load shedding signal is generated, either from the degraded voltage or the SI !
enable timer, to the time full voltage is restored by the diesel. The response time of the timers is covered by the requirements on their setpoints.
(11) The provisions of Specification 4.0.4 are not applicable for entry into a MODE 3 for the turbine-driven Auxiliary Feedwater Pump.
l R68 (12) The following valves are exceptions to the response times shown in the table and will have the values listed in seconds for the initiating signals and the function indicated:
Valves: 10 Response FCV-67-89,-SO,-10f8)85{9) times: 7.b, 75 /
R73 Valve: FCV-70-141 Response times: 7.b, 70(8)/80(9) ,
, September 9, 1988 SEQUOYAH - UNIT 2 3/4 3-33a Amendment No. Aff, M 73
-- -~
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-- - : --v-----
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L '
INSTRUMENTATION ,
RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION L_IMITING CONDITION FOR OPERATION 3.3.3.9 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The alarm / )
I trip setpoints of these channels shall be determined in' accordance with the R34 methodology and parameters in the OFFSITE 00SE CALCULATION MANUAL (ODCM).
APPLICABILITY: During releases via these pathways. lR34 i
ACTION:
- a. With a radioactive liquid effluent monitoring instrumentation charnel f alarm / trip setpoint less conservative than required by the above spe- ;
cification, without delay, suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel I O'
inoperable, or change the setpoint so it is acceptably conservative. ,
i
- b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE take the ACTION shown in Table 3.3-12. Exert best effort to return the instruments to OPERABLE status within 30 days and, if unsuccessful, explain in the next Semi-Annual Radioactive Effluent Release Report why the inoper- -
ability could not be corrected within 30 days.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. ,'.
l SURVEILLANCE REQUIREMENTS 4.3.3.9 Each. radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-8.
3/4 3-68 Amendment No. H , 72 SEQUOYAH - UNIT 2 '
September 1, 1988 e . .
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) SEQUOYAH - UNIT 2 3/4 3-69 O
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TA8LE 3,3-12 (Continued)
TA8LE NOTATION .
l ACTION 30 - With the number of channels OPERABLE less than required by the Minimum Channels OPERA 8LE requirement, affluent releases may continue provided that prior to initiating a release: 'ln34
- a. At least two independent samples are analyzed in accordance with Specification 4.11.1.1.1, and b'. At least two technically qualifie.d members of the Facility Staff independently verify the release rate calculations and discharge line valving; otherwise, suspend release of radioactive effluents via this pathway.
ACTION 31 - With the number of channels 0PERABLE less than required by the Minfaum Channels OPERA 8LE requirement, effluent releases via !
this pathway may continue provided grab samples are analyzed ]R34 for grosg7 radioactivity gamma at a limit of detection of at least 10 microcuries/ gram:
- a. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the specific activity of lR34 the secondary coolant is greater than 0.01 microcuries/ gram DOSE EQUIVALENT I-131.
- b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the specific activity of I the secondary coolant is less than or equal to 0.01 micro-curies / gram DOSE EQUIVALENT I-131.
ACTION 32 - With the number of channels OPERABLE less than required by the Minimum Channels OPERA 8LE requirement, effluent releases via this pathway may continue provided that, at least once per R34 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, grab samples are collected and analyzed for gros 3 7 radioactivity gamma at a limit of detection of at least 10 microcuries/e1.
ACTION 33 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway say' continue provided the flow rate is estimated lR34 at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump curves i may be used to estimate flow.
ACTION 34 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, liquid additions to this tank may continue provided the tank liquid level is estimated ]R34 during all liquid additions to the tank.
l January 14, 1986 SEQUOYAH - UNIT 2 3/4 3-71 Amendment No. 34
, J s* .
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+
TA8LE 3.3-12 (Continued)
TABLE NOTATION ;
i
. ACTION 35 - With the number of channels OPERA 8LE less than required by the j Ninimum Channels OPERA 8LE requirement, effluent releases via '
this pathway may continue provided representative batch samples of each tank to be released are taken prior to release and composited for analysis according to specification 3.11.1.1, footnote g.
l b.
4 t
SEQUOYAH - UNIT 2 3/4 3-72
, . . . . . . . - - - - - . - ~ - - . . . -
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SQNP SI_2
~
DATA PACKAGE COVER SHEET Page 1 of 1 Revision 55 SHIFT LOG
) Unit Mode 3 Mode 3 Mode (230G-0700) (0700-1500) (1500-2300)
Performed By / (AUO, 00, ASOS) Date
/
l 2300-0700 0700-1500 1500-2300 l List of data sheets attached.
Data Sheet No. No. of Pages Instruction No.
51 2 Shift 1.oo Did all SI data meet acceptance criteria? Yes No Verifled By:
If criteria were not satisfied, notify the Shift Operations Supervisor who completes the following: .
Was a Potential Reportable Occurrence Initiated? Yes No Mas Limiting Condition for Operation action required?
Yes (explain in remarks) No (explain in remarks) verified By Date Shift Operations Supervisor Time l Reason for test:
Required by schedule Plant condition (explain)
Other (explain) __
Review of Test Results Unit Operator 11-7 _, 7-3 , 3-11 j 1
Review of Test
.STA Date Review and Approval of Test SRO Date Remarks
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SQNP SI-2 Page 6 of 15 Revision 55
~
3.0 INSTRUCTIONS (continued) 3.3 (continued) 3.3.9 (SR 4.3.1.1.1. A.14, 4.3.1.1.1. A.15) (SR 4.3.2.1.1. A.5.a !
and 4.3.2.1.1.A.6.c and 4.3-2.6.C) l Steam Generator (S/G) Nater Level-Low-Low~
S/G Water Level-High-High S/G Water Level-Low The low-low S/G water level reactor trip systee, and high-high level for turbine trip feedwater isolation, and auxillary feedwater initiation instrumentation will be channel checked by verifying the operability of the three level indicators for each S/G on panel M-4.
i Each S/G 1evel is indicated on M-4 by three Indicators. l The level indicators for S/G #1 are LI-3-42, 39, 38; for S/G #2 LI-3-55, 52, 51; for S/G #3 LI-3-97, 94, 93; for S/G #4 LI-3-110, 107, 106. ,
I
, Comparing the three level indicators for each S/G to each other serves as a channel check for that S/G's level instrumentation. Operab111ty is vertfled by having 1 67, deviation between channels. l 3.3.10 (SR 4.3.1.1.1.A.15) Steam /Feedwater (FW) Flow Mismatch The steam /FW flow alsmatch reactor trip system instrumentation channel check will be performed by comparing the two FN flow Indicators and comparing the two SF indicator (Panel M-4) for each loop to varify operability.
l 1
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SI-2 Page 6a of 15
, ,',, Revision 55 3.0 INSTRUCTIONS (continued) 3.3.10 (continued)
Operability is verifled by observing the following:
With Reactor Power > 25%: acceptable deviation 13
< .315 X 106 lbe/hr for feedwater and < .36 X 10' Iba/hr for steam flow with Reactor Power < 25%,
acceptable deviation is < .63 X 106 .lbalhr for feedwater and 1 65 X 106 lbs/hr for stens flows.
NOTE: Steam /feedwater flow indication should change accordingly with known increase or decrease in system flows. At zero system flows, questionable indicators may be vertfled by IM's, Flow indicators on loop #1 are FI-3-35A and 8 for FW and FI-l'3A and 8 for steam.
Flow indicators on loop #2 are FI-3-48A and B for FW and FI-1-10A and B for steam.
Flow indicators on loop #3 are FI-3-90A and 8 for FW and FI-1-21A and B for steam.
Flow indicators on loop #4 are FI-3-103A and 8 for FM and FI-1-2BA and B for steam.
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UNIT 1 ATTACHMENTS Table of Contents:
- 1. T.S 3.0.5 p. 3/4 0-1
- 2. T.S 3.1.2.1 p. 3/4 1-7
- 3. T.S. 3.1.2.2 p. 3/4 1-8
- 4. T.S. 3.1.2.3 p. 3/4 1-9
- 5. T.S. 3.1.2.4 p. 3/4 1-10
- 6. T.S. 3.8.1.1 p. 3/4 8-1 thru 3/4 8-5 l
I l
l 1
l j
2 ;
1 C_______.__.___ _ _ _ _ _ _ _ _ _ _ _
3/a LIMITING CONDITIONS FOR OPERATION A4 $ SURVEILLANCE REQUIREMENTS .
3/4.0 APPLICA81LITY LIMITING CONDITION FOR OPERATION 3.0.1 Coupliance with the Limiting Conditions for Operation contained in the succeeding Specifications is required durin0 the OPERATIONAL 20E5 er other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be set.
3.0.2 Noncosoltance with a specification shall exist when the requirements of the Limitin0 Candition for Operation and associated ACTION requirements are not set within the specified time intervals. If the Limiting Candities for Operetten is restemd prior to expiration of the specified ttee intervals, completten of the ACTION requirements is not required.
3.0.3 When a Limiting Coneition for Operatten is not est,.except as provided in the. associated ACTION requirements, within ene hour action shall be initiated to place the unit in a 20E in editch the Specification does not apply by placing it, as applicable, in:
- 1. At least HDT STANDBY within the nort 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />,
- 2. At least NOT 3HLTTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and
- 3. At least COLD SHLffDOWN within the seseguent N hours.
Where carrective sensums are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time ifrits 37 as measured from the time of failure to meet the Limiting Candition for Operation.
Exceptions to these roovirements are stated in the individual Speirifications.
3.0.4 Entry into an OPERATIONAL 20E or other spe'cified condition shall not be made imless the conditions for the Limiting Candition for Operation are est without relianca en previsions contained in the ACTION requirements. This provision shall not prevent passage through OPERATIONAL MDE5 as required to comply with ACTION requirements. Exceptions ta these requirements are stated in the individual Specifications.
3.0.5 When 8 systas, subsystes, train, component er device is detoftined to be inoperable solely because its emergency power searce is inoperable, or solely because its normal power source is inoperable, it any he considered OPERA 8LE for the purpose of satisfying the requirements tf its applicable Limiting Condition for Operation, provided: (1) its corresponding normal er emergency power source is OPERABLE; and (2) all of its redundant systes(s), subsystes(s),
train (s), compenant(s) and device (s) are OPERARLE, or likewise satisfy the requirements of this Specification. Unless both conditions (1) and (2) are g satisified, within I hours action shall be init.tated to place the unit in a
- 4. MODE in which the applicable Limiting Condition for Operation does not apply by placing it as applicable in: *
! 1. At least NOT STAND 8Y within the next 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />,
! 2. At least NOT SHUTDOWN within the fe11 wing 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, and
- 3. At least COLD SHLITDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
This specification is not applicable in MODES 5 or 5. .
. SEQUQYAN - UNIT 1 3/4 0-1 $y7g
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REACT!Y!TY CONTROL SYSTEMS 3/a.1.2 BORAT10N SYSTEMS FLOW PATHS - $WTDOW
' LIMITING CONDITION FOR OP!1LATION l
3.1.1.1 As a sinimus, one of the following boron injection flow paths shall be OPERA 8LE:
- a. A flow path fme the boric acid tank via a boric acid transfer pump and charging pump to the Reactor Coolant Systes if only the boric acid storage tank in Specification 3.1.1.5a is OPERABLE, or
- b. The flow path from the refueling water storage tank via a charging pump to the Reactor Coolant System if only the refueling water storage tank in Specification 3.1.2.5b is 0PERABLE.
APPLICA81LITY: MODES 5 and 5.
gTI0lt:
With none of the above flow paths OPEAABLE, suspend all operations involving CORE ALTERAT!00t5 or positive reactivity changes.
1 1
SURVE!LLANCE REQUIREMElffs 4.1.2.1 At least one of the above required flow paths shall be demonstrated CPERABL1:
- a. At least once per 7 days by verifying that the temperature of the heat traces portion of the flow path is greater than or equal to 145'F when a flow path from the beric acid tanks is used.
- b. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured ir. position, is in its correct position. .
l -
$EQUQYAH - UNIT 1 3/4 1-7 y- ;,. _ -.
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l REACTIVITY CONTROL inism ,
I FLOW PATH 5 - OPERATING '
LIMITING CON 0! TION E OPERATION 3.1.2.2 At least two of the following three beren injection flet paths shall ,
l j
be OPERA 8LE: ,
.. g i
- a. The flew path from the beric acid tanks via a heric acid transfer j pump and a charging pump to the Reactor Coolant Systas. l
. l
- b. Two flew paths from the refueling water store 0s tank via charging '
pumps to the Reacter Coelant System.~
APPLICABILITY: 2 0E5 1, 2, 3 and 4. ,
sm: .
With only one of the above required boron injection flew paths t,a the Reacter Coolant Systes OPERASLE, restare at least two boren injection f1sw paths to f-the Reactor Coolant System ta OPERASLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least I HOT STAMMY end borated to a $ltlTDOW MARGIN equivalent to at least 15 delta k/k "
at 200*F within the next 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />; restore at least two flew paths to OPERABLE status within the next 7 days or be in COLO 5HLffD0'rM within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILL*FE REQUIREMENTS 4.1.2.2 At least two of the above required flow paths shall be desenstrated OPERABLE:
- a. At least once per 7 days by verifying that the temperature of the '
l heat traced portion of the flow path from the beric acid tanks is I
greater than or equal to 145'F when it is a required water source.
- b. At least once per 31 days by verifying that each valve (sanval, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
\
- c. At least once per 18 months during shutdown by verifying that each l I
automatic valve in the flow path actuates to its correct position en asafetyinjectiontestsignal.
- d. At least once per !$ sonths by verifying that the flow path required ,
by Specification 3.1.2.2a delivers at least 10 gpe to the Reacter Coolant Systes. ,,
WAR 251982 SEQUOYAH - UNIT 1 3/4 1-8 Amendment No. 12 f l
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l REACTIVTTY CONTROL SYST9t$
)
CwutGIC PUMP - SHUT 00WN s .
LIMITING CONDITION FOR OPERATION l
3.1.2.3 One charging pop in the boren injection flow path required by
$ specification 3.1.2.1 shall be OPERABLE and capable of being powered free an i OPERA 8LE shutdown board.
APPLICA81LITY: 2 0E3 5 and 6.
i M: .
l With no charging pep OPERABLE, suspend all sperations involving CDRE ALTERATIONS .
er positive reactivity changes.
]
)
SURVEILLANCE RE0uttetNTS i .
l .
4.1.2.3 The above requimd charging pump shall be demonstra:ad OPERABLE by verifying, that en recirevlation flow, the pop movelops a discharge pressum .
of gree.ar than er equal to 2400 psig when tested pursuant to specification 4.0.5.
i e
A $E000YAH - UNIT 1 3/4 1*9 9
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i REACT!vfTY CONTA0L SYSTEMS 3
= CHARGING PUMP 5 - OPEAATING -
^
LIMITING CO,NDITION FOR OPERATION ,
3.1.2.4 At least two charging pumps shall be 07 ERA 8LE. ;
APPLICA8!QQ: MODES 1, 2, 3 and 4.
.AE.I.E: ,
With only one charging pump OPERA 4LE, restore at least two chargint pumps to OPEAABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> er be in at least NOT STAN08Y and berated to a '
$NUTDOWN MAAGIN equivalent to at least 15 delta h/k at 200*F within the next i 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two charging pumps to OPEAA8LE status within the
- next 7 days er be in COLO LHUTOOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. .
SURVEILLANCE REOUfREMENTS ,
a l
~
a.1.2.a At least two charging pumps shall be demonstrated CPERASLE 1py verifying, that on recirculation flow, each pump develops a discharge pressure of greater than er equal to 2a00 psig when tasted pursuant to Specification a.C.5.
9 SEquCYAH - UNIT 1 3/a 1-10 !cencent No.1 -
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- I 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES
- OPERATING LIMITING CONDITION POR OPERATION q 3.8.1.1 As a sinism, the following A.C. electrical power sources shall be OPERABLE: 1
- a. Tw physically independent circuits between the offsite transmission
. netwet and the onsite Class 1E distrktion system, and
- b. Four separata and independent diesel generator sets each with:
j 1. Two diesels driving a common generster i
- 1. Two engine-sounted fuel tanks containing a sinimum volme of 250 gallons of fuel, per tank ;
{- A separate fuel storaes systas containing a minimum volume of (
I 3. I 62.000 gallons of fuel, f 4. A separete fuel transfer pump, and
~
- 5. A separata 125-volt D.C. distribution panel,125-volt D.C. !
battery bank and associated charger.
APPtlCA81LITY: MOES 1, 2, 3 and 4. , . .
.A G E:
. a. With either an offsita circuit er diesel generator set of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY ef the remaining A.C. sources by performing Surveillance Requirements 4.B.1.1.1.a and 4.8.1.1.La.4 within one hour and at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter; restore at least two effsite i
circuits and four dissal generstar sets to OPERA 8LE status within 72 i hours or be in at least IET STAN08Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in
' COLD $HUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />,
- b. With one offsita circuit and one diesel generator set of the above reouired A.C. electrical power sources inoperable, demonstrate the OPERA 81LITY of the remainin0 A.C. sources by performing Surveillance Requirements 4.8.1.1.1.a and 4.8.1.1.2.a.4 within one hour and at least once per 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> thereafter; restare' at least one of the inoperable sources to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> er be in at least MOT STAND 8Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within
' the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Restore at least two offsite circuits and L four diesel generator sets to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> free the time of initial loss or be in at least NOT STA28Y within the
~
next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD $HUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
't SEQUOYAH - UNIT 1 3/4 4-1 .
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' * ' l ELECTRICAL POWE1t SYSTEPs ACT!0n (Continued) .
- c. Vith too of the I
demonstrate theabove reoufred OPDtAI!LITY of 4offsfto diesel A.C. circuits generator setsinoperableMa8 Iry perfb ]
Surveillance Pequirment 4.8.1.1.2.a.4 w< thin one hour and at least
- once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter, unless the diesel innerator sets are already coersting; restore at least one of the ' noperable offsite .
sources to OPDtAEE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least WT STAN087 within the nort 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. With only one offsite source restored, restore at least tm offsito cfrtuits to OPDIAE! status j i
within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from time of initial less or be in at least NOT STM08Y withis the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and in COLD SHUTDOWN within the following JO hours. (
f
(. . . With either diesel simultanoon eenerator with 15-8 sets damnstrate aed/or 28-8 IA-A and/orthe 1 A-A iMreble of optPAa!LITY I i
two offsite A.C. circuits by perfomine Surveillance Pesufreuent 4.8.1.1.1.4 within one hour and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafters ,,
restore at least 1) 1A-A and tA-A or 2) 15-8 and 25 8 to nPtPAnLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> er be in at least NOT STANOBY within the next
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and in CDLD SHUT!WR within the 411owing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Destore at least four diesel ponerator sets to CPULAILE status wfein 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> fra time of initial loss or be in least NOT STANf9Y within ~
the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and in CDLD SWTDom within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. f.. ,
i
$UPVEILL4NCE PEDUIREPDiff I
- 4. 8.1.1.1 Each of De abeve reoutred independent circuits between the offsite !
transmission networt and De onsite Class 1E distribution systen shall be:
a, Detamined OPERAKE at least once per 7 days by verifying correct breakar alignrents, indicated powr availability, and
- b. Demonstrated OPERAEE at least once per 18 nonths during shutdown by transferring (sanually and autanatically) unit power supply from the I
normal circutt to the alternate cf reuf t.
4.3.1.1.2 Each diesel generator set shall be derenstrated CPEPA8LE:
- a. In accordance with the frecuency specified in Table a.8-1 an a
$TAGGEPG TEST 8A5!$ by:
- 1. Verifying the fuel level in the engine-counted day tants.
- 2. Verifying the fuel level in the 7 day tant.
- 3. Verifying the fuel transfer pump can he ' started and transfers fuel from the storace systen to the encine enunted %el tanks. -
$E000YAN - UNIT 1 3/a t-2 .
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ELECTRICAL POL *;A SYS1Ek, SURVEILUm;.i iiEJJistEMEi4I5 W. .tinceil
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.v,: n.. a. conoition anti accaterates t<.* A i n.:.t G eam in ic:.: 8. hen or egeal c.c IC* . ecs.W .
590;; i 690 vdha gwatw ts a?.d 60 ::voluigu ar.c hegaency
".2 Hr witt.in in* seconds shall be the af+.er start signal. 1har dian*1 generator shall be : tarted for this ' k R56 test by usina one of tne in110 wing signals with startup or. eacn '
sig; i v...i?*ed .t ;;ai c...:e pt- 10 oays.:
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Sist.iultsi io!.a ( f wif ah. [.s.usir t./ itself. , f atS6 c: M. CS' N.s.. t ion *.v, . ny. a % v.r.. . c -
- 5. Wri"fini; ti:t 9c. rut.v *; 'T /;i.ru *t.d, 'oaci+c to greater than
'*ll os v.p:1 to .;4 0 k.s in '. . thaq or co s' to 60 seconds *, and F-operates fot ,jrcate. tenan ar .u..* .e rio .M./.i.<.i. aad '
E. Veritying the dir.;c1 gener *wr li, 11gned to provite i tendby pcwer to the ass., .t..;.ied shutdown boards,
- b. At least once per 31 days and after each operation of the diesel where the period of operation was greater than or estual to I hoJr by checkitig for and removi..g accum.ulated water frosi the engine-mounted fuel tanks. I c At 'td at c ice ;.er 5; uaya ma frot. new fuel oli g rier t<> acchion i to the 7-nay tanks by ve.ifyirig that a samole octained in accorence 4 with ASTW D270-1975 has a water ano sediment content of less than or i egeal to .05 volume percent and a kinematic viscosity 6100*F of >
greater than or equal to 1.8 but less than or equal to 5.8 centi- I stokes when tested in accordance witn ASTM-0975-77, and an impurity level of less than 2 mg. of insolubles per 100 el when tested in accordance with ASTM-D2274-70.
- d. At least once per 18 months during shutdown by:
- 1. Subjecthg the dieset te er. inspection in accordaxe eith procedu es prepared in conjur.ction with its sansfactJ fer's recommendations for this class of stanty service.
- 2. Verifying the generator capability to reject a load of greater than or equal to 600 kw while maintaining voltage at 6900 1 690 volts and frequency at 6011.2 Hz.
- 5. Verifying the generator capability to cejtet a load of 44G kw R6q without tripping. The generator voltage shall not exceed -
7866 solts during and following the load rejection.
m-~f%' gerieratv s'.ar' (10 .cc) and 1c3d (60 *.ec) " roc .ita.~1b wd*e r; ,
chall be perfo med a+ east ove per lei c'tys ir these surve'11ance tests. l.1J other diesel Genere. tor enche starts and loading fo* the purrose of thde " " ~
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, ELECTRICAL POWER SYSTEMS -
SURVEILLANCE REQUIREMENTS (Continued)
< 4. Simulating a loss of offsite power by itself, and:
a) Verifying de-energization of the shutdown boards and load . .I shedding from the shutdown boards. -
I b) Verifying the diesel starts on the auto-start signal, i energizes the shutdown boards with permanently connected loads within 10 secor.ds, energizes the auto-connected ,
shutdown loads through the load sequencers and operates !
for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After energization, il the steady state voltage and frequency of the shutdown boards shall be maintained at 0900 1 690 volts and l 60 1 1.2 Hz during this test. l
- 5. Verifying that on a ESF actuation test signal (without loss of l offsite power) the diesel generator starts on the auto-start l signal and operates on staney for greater than or equal to ,
5 minutes. The generator voltage and frequency shall be !
6900 t 690 volts and 6011.2 Hz within 10 seconds after the auto-start signal; the steady state generator voltage and fre-quency shall be maintained within these Ifnits during this test.
- 6. Simulating a loss of offsite power in conjunction with an ESF U3 (~
actuation test signal, and a) Verifying de-energization of the shutdown boards and load shedding from the shutdown boards.
b) Verifying the diesel starts from ambient condition on the auto-start signal, energizes the shutdown boards with permanently connected loads within 10 seconds, energizes the auto-connected emergency (accident) loads through the load sequencers and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads. After energization, the steady state voltage and frequency of the emergency busses shall be maintained at 6900 1 690 volts and 60 1 1.2 Hz during this test.
c) Verifying that all automatic diesel generator trips, )
except engine overspeed and generator differential, are automatically bypassed upon loss of voltage on the shutdown board and/or safety injection actuation signal. '1
- 7. Verifying the diesel generator operatet for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. D3 '
During the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of this test, the diesel generator i shall be loaded to greater than or equal to 4840 kw and during %68 ,
the remaining 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> of this test, the diesel generator shall >
be loaded to greater than or equal to 4400 kw. f68 SEQUOYAH - UNIT 1 3/4 8-4 Amendment No. A$, 64 J4:uary 7, 1968 l
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ELECTRICAL POWE;t 3 STEMS SURVEItLid4CE RtQip,t,4Ehlb (cordioed , , , , . _
Within b ri.au'r af'.c :.W11'.icer; thi., 24 levur Last, periors 5pecificatian A.8 '. 3.:t d.4. Ine gentir. tor volt 48 anu fre- !R56 qum;./ Nil' 'w GO.8 i 6'/, vo'ts s..J K 1 1.2 H1 within i in it.. c.WJ hi .:r. see m . ; ri ,nas; & 5.teady state generator voltage ano f requency shall be maintained witnin t.)ese linica daring this test.
P. .'.:r d yt or t':' t* : . t< v: .s -.. hw. t. et.-% d%*1 nentv.-
do tsot er.ceep hn cort'io ous ti .eg of 4t00 W lR68
- f. Verif. ' .g ths diest r:enav+ts. '
. t.t t. 4' " *.,- tv -
a) Synchren 8 e wi'h the efitite powter soa:ce t.hi16 cne genera-tor is 1 sed titi. its ete?vem j Ivans u0s.1. simulated restoration ot celsite co .'
h) Transfer its loads to the o..'a-te power swrca, W c) Be restored to its shutdown status.
- 10. Verifying that the automatic load sequence timers are OPER4LE 1 witn che setpoint fcr taca s=;uence timar within ; 5 p.64.:t ct i its design setpoint. ,
11 Verifying that tbr. fc11owin; ."*.1 ger.or='.or lie at fatsen prevent diesel ga :"te st.'rt' .; only whee re:. sired: -
a) Engine overspeed ;
b) 56 GA lockout reiay
- s. At least once per 10 years or af ter an) s.odificati.,36 .tilch could affect diesel generator interdependence by starting the diesel generators simultaneously, during shutdown, and verifying that the diesel generetors accelerate to at least 900 rom in less 'J.an o-equel to 10 seconds. -
- f. At least once per 10 years" by:
- 1. Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypoclorite solution, and
- 2. Performing a pressure test of those portions of the diesel fuel cil .systes design to Section III, sbbsection ND cf the .
ASME Code at a test prenure equa' to 110 percent o' the syste destgri prassiere. ,
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EMERGENCY CORE COOLING SYSTEf.5 (ECCS) 3/4.5.3 ECCS SUBSYSTEMS - T,yg Less Than 350'F LIMITING CONDITION FOR OPERATION 3.5.3 As a rinimum, one ECCS subsystem comprised of the following shall be OPERABLE:
- a. One OPEF ABLE centrifugal charging pump,
- b. One OPERABLE residual heat removal heat exchanger,
- c. One OPERABLE residual heat removal pump, and -
APPLICABILITY: MODE 4. -
ACTION:
- a. hithnoECCSsubsystem.0PERABLEbecauseoftheinoperabilityof either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.
- b. With no ECCS subsystem OPERABLE because of the inoperability of either the residual heat removal heat exchanger or residual heat removal pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reactor _ Coolant System T less than 350*F by use of alternate heat removal methods, avg
- c. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a REPORTABLE EVENT shall be prepared and submitted to the Commission pursuant to Specification 6.6.1. This report shall include a description of the circumstances of the actuation and the R40 total accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be provided in this report whenever its value exceeds 0.70.
November 23, 1984 SEQUOYAH - UNIT 1 3/4 5-9 Amendment No. 36 J
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l EMERGENCY CORE COOLING SYSTEMS (ECCS) I r
SURVEILLANCE REQUIREMENTS 4.5.3 The ECCS subsystem shall .be demonstrated OPgRABLE,per the applicable .
Surveillance Requirements of 4.5.2.
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SEQUOYAH - UNIT 1 3/4 5-10
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