ML20244A955
| ML20244A955 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 05/24/1989 |
| From: | Arildsen J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20244A952 | List: |
| References | |
| 50-424-OL-88-04, 50-424-OL-88-4, 50-425-OL-88-04, 50-425-OL-88-4, NUDOCS 8906120210 | |
| Download: ML20244A955 (323) | |
Text
{{#Wiki_filter:q a I i d ENCLOSURE 1 J l EXAMINATION REPORT 424/425/0L-88-04
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Facility Licensee: Georgia Power. Company - l P. O. Box 4545 l Atlanta, GA. 30302 { Facility Name: Vogtle Electric Generating Plant _ Facility Docket Nos.: 50-424 and 50-425 i Written examinations and , rating tests were administered at Vogtle Electric Generating Plant near a sboro, Georgia. Chief Exam.ner: 4 Jesse A.'
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. Date' Signed Approved By: 4t Charles A. Custo, Chief *
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A (M ~ f/ @ Date Signed Operator Licensing Section 1 Division of Reactor Safety Summary: Examinations were administered on December 5, 1988, and January 10.- 12, 1989. Operating tests were administered to nine candidates; eight of whom passed. Nine candidates were administered written examinations; seven of whom passed. Based on the results above, six of nine SR0s passed. Eight of twenty-seven (30 percent) of the changes made to the written examination as a result of facility comments were due to inadequate or incorrect reference material supplied by the facility to the examiners for ; examination preparation. Reference material submitted to the NRC should { accurately reflect the current plant configuration so that post-examination ! modifications to the examination are minimized. i 1 i 1 L
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i REPORT DETAILS
- 1. Facil_ity Employees Contacted:
l *K. Holmes, Plant Training and Emergency Planning Manager
*R. Dorman, Operations Superintendent of Training
'*D. Scukanec, Training Department Staff.
~* Attended Exit Heeting
- 2. Examiners:
*J. Arildsen, NPC, Region II M. Morgan, NRC, Region II G. Salyers, NRC, Region II' B. Picker, INEL
- Chief Examiner i
- 3. Examination Review Meeting At the conclusion of the written examinations, . the examiners provioed R. Dorman with a copy of the written e,tamination and answer key for review. The NRC resolutions to facility comments are listed below.
Question Number: 5.07B 1 NRC Resolution: Agree with facility comment. 1 l Lesson plan material provided to the NRC for examination preparation ' (LO-LP-33220-01, lessons Outline II.B.3.d.3) states, " effective fuel temperature for a given power level decreases over core life due to pellet swell and clad creep." This is in full agreement with the 3 original answer key, however, the answer Iey is changed to include ,
" fission gas in the gap" as a third acceptable answer due to facility comment references. Two of three factors required for full credit. The facility must ensure consistency in its training material.
Question Number: 5.08B l NRC Resolution: Agree with facility comment. 1 The facility is cautioned that their training material is, at best, , ( unclear i.e., it specifically states ". . . total worth of all rods decreases slightly over core life." The facility should update their training material. This question's Part B is deleted and point values appropriately altered. I
T 2 Question Number: 5.09 NRC Resolution: Agree with facility comment. The answer for Pressure Feedback out is added to the answer key but will be used Joni if the candidate specifically states the assumption that Pressure Feedback is out. This was specifically announced and written on the classroom dry-erase board during examination administration. No technical information to defend the facility recommended answer was included-in the submittal. References provided defend only the original answer key for Pressure Feedback in. The facility is reminded that complete support references are rTquired with the facility comment submittals. Question Number: 5.11 NRC Resolution: Disagree with facility comment. The facility comment addresses the " redistribution effects" as stated in the lesson plan (D.E.a-d). The question asks for the two principle l factors effecting the redistribution reactivity effect which is specifically addressed in the same referenced lesson plan (D.7.b). .It states ". . . principle factors affecting redistribution are moderator creativity defect at BOL and fuel burnup at E0L." This is consistent with the answer key which remains untbanged. l Question Number: 5.12 NRC Resolution: Agree with facility comment. Samarium cencertration is added to the answer key bs one of the six pcssible answers, five of the six are required for full credit. Question Number: 5.13b NRC Resolution: Agree in part with the facility comment. The doppler power coefficient and doppler-only power defect change due to I. the same phenomena over core life as stated in the reference sited by the The referenced Vogtle Lesson Plan specifically addresses facility. Vogtle learning objective #12 in stating that " factors affecting. doppler only power Defect are: a) Reactor Power, b) Coolant Tavg, c) Core Age and d) Clad Creep " This is located on a hand written insert (page 9a), and is consistent with the answer key, however, to eliminate potential _ confusion due to Vogtle training material inconsistency with the answer key is revised. The technical reference supplied by the facility does not defer ' recommendation #3 nor the first part of #1; a:cordingly, the arswer key is changed to read: 1 1 i 1 __
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- b. 1. Helium gap thermal conductivity
- 2. Depletion of U-238 and buildup of Pu-2C
- 3. Fuel densification
- 4. Clad Creep
- 5. Coolant Tave (Core age credited as one factor in' substitution of the first three listed). (three factors required at 0.5 points each)
Question !' umber: 5.17 NRC Resolution: Agree in part with facility comment. Page 10-55 of the facility-provided Westinghouse text gives the available NPSH equation, and in explanation states, ". . . . the available NPSH is dependent on static fluid pressure, fluid temperature, and fluid - velocity." This is fully consistent with the answer key. The facility's recommended second answer " static head due to elevation difference," is incomplete without mention of the fluid pressure (P static fluid + P elevation). Volumetric flowrate will be included as an acceptable l alternative to fluid velocity. Question Number: 5.19 : NRC Resolution: Agree in part with facility comment. 4 Inherent to the question are three very distinct consequences of- l cavitation. The initial answer key grouped the consequences consistent i with facility supplied reference material. The point breakdown was added to answer key and answer number 3 was altered to reflect a reduction in flow rate as stated in the reference material. Question Number: 6.03 I NRC Resolution: Disagree with facility comment. I Reactor Coolant Pump seal #3 has two sealing surfaces which develop a delta-P in two directions. In one direction, the standpipe is at ! containment atmospheric pressure. Therefore, the delta-P is the height of standpipe water only. In the second direction, it is . the standpipe interaction with the leakoff of f 2 seal. Therefore, distractor "d" is incomplete and is not the best alternative when compared to "a." No change to the answer key was made. Question Number: 6.09 NRC Resolution: Agree in part with facility comment. The Tavg-Tref recorded is added as an additional correct answer. Annunciators are not considered systems nor instruments.
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q 4-Question Number: '6.10 NRC Resolution: ' Agree'in part with-facility comment. I The - answer key is changed to' reflect all' of the facility's recommendations, however, P-4 is . listed - as one : answer . With all information in items 1 and 11 required:for full credit. ! Question Number: 6.11 NRC Resolution: Information on the' AMSAC system was not' initially sent for examination" preparation and was celivered to._ the examiners only after specifically requested by the NRC. The-logic diagram provided as' reference material for this fecility comment is . illegible : however, since AMSAC will! automatically close the S/G blowdown. containment-isolation valves, it has been added to the answer key. Question Number: 6.12-
-NRC Resolution: Agree in part with facility comment. i The reference material appears to ' support. the ' original answer as stated. . .
L The handwritten . notes on the supplied reference material .do not - .( constitute a- technical basis on which.- to alter the answer key. Due to 1 the confusion generated by the reference material and the question not specifying ' local or remote operation, the-' question. is deleted. -The facility is cautioned that due to the inaccuracy of: facility . training: . material. . uncontrolled training material will not be accepted .as reference material for correction of' training material errors. Question Number: 6.13 NRC Resolution: Agree in part with facility comment. The question already exempts many of the loads recommended for ! acceptance; yet, due to a lack of clarity the question is deleted. 'j Question Number: 6.14 ; NRC Resolution: Agree in part with facility comment, i The reference provided by the facility states the " conditions" satisfied for the DG breaker closure. Its listing adequately , describes! the . permissive and is acceptable. The answer key is changed for answer- i number four to read: "Close Permissive Signal from the ' Sequencer (indicates load shed as been completed)." ;
5' l Question Number: 6.15 NRC Resolution: Disagree with facility comment. The recommended answer as stated by the facility would' not require any difference in description of the DPb~ and DPc transmitters. To better distinguish acceptable answers the answer key is altered to incorporate. additional correct wording: DPa - (0.5 each) measures static (no RCPs running) narrow range level of the vessel from hot leg to top of head. DPb - measures static (no RCPs running) wide range level from bottom of vessel to top of vessel. DPc - measures dynamic differential pressure (RCP running) between bottom of the vessel and upper head. Question Number: 6.16 NRC Resolution: Agree in part with facility comment. Setpoint tolerances added to the answer key. Question Number: 6.18 l NRC Resolution: Agree in part with facility comment. I The UV sequencer on Unit I may be reset by any one of three methods. All ) three methods are required for full credit, and the answer key is j appropriately changed. Question Number 6.19 NRC Resolution: Agree with facility comment. Typographical errors corrected. Question Number: 6.20 NRC Resolution: Agree with facility comment. The answer key changed to delete #2 and redistribute point values. Question Number: 7.06 NRC Resolution: Agree with facility comment. The answer key is altered to indicate "C" as a correct answer.
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, l o ;1 Question Number: 7.07 NRC Resolution: Agree in part with facility comment.'
l The facility supplied. reference procedure (19211-C) indicates, that the - .' facility training material is in. error.- The sequence provided by the facility in recommendation for partial credit is the only. correct answer.
- and the answer key is. modified-to. reflect.the change.
' Question Number: 7.08 NRC Resolution: Acknowledge facility comment.
The alternate wording : recommended is L acceptable,. and no, change to- the ' answer key is . required.
- Question Number 7.09 NRC Resolution: Acknowledge facility comment.
The alternate wording recommended is acceptable,' and no change to the answer key is required. Question Number: 7.10 , NRC Resolution: Agree in part with facility comment. l The point distribution is corrected and additional correct' answers are - added to the answer key. Question Number: 7.11 , NRC Resolution: Agree in part with facility comment. i The original answer reflected - the facility -training material ' (LO-LP-37002-02-c, Objective 13) properly, however, specific examples of proper transition to plant procedures are acceptable. "After consulting TSC" is not acceptable. Question Number: 7.12 NRC Resolution: Agree in part with facility comment. The facility procedure does not state specifically how the restoration is performed and, thereby, assumes the operator. has knowledge of- the actions to be accomplished. i
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p t 7 Question Number: 7.14 NRC Resolution: Disagree with facility comment. The' facility recommendation is unclear and subjective with no concise technical recommendations. The answer key clearly lists the three alternative methods used to satisfy the requirements. Partial credit point distribution is annotated on the answer key. Question Number: 7.15 NRC Resolution: Agree with facility comment. The answer key is altered to include " loose contamination area" as an answer, Question Number: 7.16 NRC Resolution: Acknowledge facility comment. Specific criteria to terminate SI as may be given as paraphrased steps from the E0Ps are acceptable. No change to the answer key is required. ' Question Number: 7.17 , l NRC Resolution: Acknowledge facility comment. The facility's recommendation will be taken into consideration during examination grading. Question Number: 7.18 ; j NRC Resolution: Agree with facility comment. The applicable limitation (12003-C, 2.2.9) is an acceptable alternate 1 answer. ) Question Number: 7.20 NRC Resolution: Agree with facility comment. The answer "b" was eliminated from the answer key and the value of the j question reduced to I reint. Question Number: 7.21 NRC Resolution: Agree with facility comment, i I i
8 Question Number: 7.23 NRC Resolution: Agree with facility comment. The correct values are entered in the _ answer key. The facility is reminded that the training material supplied contained the incorrect values, and requires correction. Question Number: 8.05 NRC Resolution: Agree with facility comment. Typographical error corrected. Question Number: 8.09 NRC Resolution: Agree in part with facility comment. The question and answer were taken directly from the procedure, but, due to the potential for confusion of the candidate with respect to the numerous other General Work Practices, the question is delete. Question Number: 8.10 NRC Resolution: Agree in part with facility comment. The OSOS may designate an alternate only in the absence of a Support Shift Supervisor. The answer key is changed to include "0 SOS designee if Support Shift Supervisor is absent." Question Number: 8.12 NRC Resolution: Agree in part with facility comments. The additional titles were added to the answer key for General Manager and Vice President. However, the supplied reference material does not indicate the Plant Manager as anything other than the first alternate to the primaries for ED. There is no supporting documentation indicating any n conditions in which the Plant Manager would assume the ED while awaiting the primaries to arrive. No addition to the answer key was made. Question Number: 8.19 NRC Resolution: Disagree with facility comment. The reference material supplied by the facility clearly states " control and final approval" and the answer key clearly uses the words " control" and " final approval." " Control" is not analogous nor identical to " final approval . " No change to the answer key is required.
'9 4.- Exit Meeting Atithe conclusion of the' site' visit the examiners. met with representatives of. the plant. staff to. discuss 'the results of the' examination. -
The cooperation given to the . examiners and the effort to ensure an - atmosphere in :the. control room conducive ' to' oral examinations ~ was also E .noted and appreciated. 1
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1 The licensee did not identify as proprietary any of the material provided to.or reviewed by the examiners. The area (s) of below normal performance'were:?
- a. Incorrect control board manipulations.
- b. Inadequate familiarity with' boration/ dilution calculations.
c.- Vague andLincorrect communications-between. crew. d.- Poor interpretation of plant ~ alarms and indications.' I l i
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l i l l 1 i Nuclear Regulatory Commission ;) Operator Licensing Examination MASTBCCPY l This document is removed from Official Use Only category on l date of examination, l. 1 i 4 NRC Official-Use Only i 1 I l i
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's U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: VOGTLE 1&2 REACTOR. TYPE: P_WR-WEC4 DATE ADMINISTERED: 88/12/05 EXAMINER: REGION II CANDIDATE INSTRUCTIONS TO CANDIDATE:
Use separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after l the examination starts. ) % OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE' TOTAL SCORE VALUE CATEGORY-27.50 25.96 20.00 24.00 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS,AND' THERMODYNAMICS 25.25 23.65 l _20.25 25.30 6. PLANT SYSTEMS DESIGN,. CONTROL, i AND INSTRUMENTATION 26.00 24.36 00.00 -e6-03 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 26.23 28.00 90- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 115.26- ._
% Totals Final Grade All work done on this examination is my own. I have neither given l nor received aid.
l l Candidate's Signature
'- ', NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
- 1. Cheating on the examination means an automatic denial of your application end could result in more severe penalties.
- 2. Restroom trips are to be 13mited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil only to facilitate legible reproductions.
- 4. Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer sheet.
- 9. Number each answer as to category and number, for example, 1.4, 6.3.
- 10. Skip at least three lines between each answer. ,
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the question anc can be used as a guide for the depth of answer required. j
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has !
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- 18. When you complete your examination, you shall:
- a. Assemble your examination as follows:
(1) Exam questions on top. (2) Exam aids - figures, tables, etc. (3) Answer pages including figures which are part'of the answer,
- b. Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn in all scrap paper and the balance of-the paper that you did not use for answering the questions.
- d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
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,. .. S ' . ,' THEORY OF }{UCLEAR POWER PLANT OPERATION.. Pesa 2
~A ' FLUIDS.AND THERMODYNAMICS ;
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f QUESTION 5.01 (1.00) Which;one of the following is NOT correct concerning theLFuel' l Temperature Coefficient vs. Fuel Temperature? 1 a.. The change in atomic motion 'is more per degreefchange in'T-fuell at higher temperatures..
.. b . The change in broadening'of the' peaks is-.less for changes'in l T-fuel at higher. fuel temperatures. i a
The f ractional change in nun.ber ' of' neutrons absorbed is less at
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higher T-fuel for the same' delta-T; q l d.~ Overlap-of'b'roadened neighboring' peaks mayLreduce the magnitude- 4 for higher T-fuel changes. ) l 1 1 l l l 1 i
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, QUESTION 5.02 (2.50)
Compare the calculated-Estimated Critical Position (ECP) for'a startup to be performed 10. hours after'a' trip from 100% power, to ACTUAL. - )". control rod position if the following: events / conditions. occurred. Consider each independently. (State your. answer as~ Actual position is. HIGHER THAN the ECP the Actual position;is LOWER THAN the ECP, or the Actual' position is the SAME AS the ECP)
- a. One' reactor coolant pump-is stopped two minutes prior to criticality.
!- b. The S/U is conducted 4 hours earlier-than planned. l c. The steam dump pressure setpoint is reduced to.950 psig,
- d. Condenser. pressure is reduced.from 7 psia to 2 psia.
-e. .All S/Gilevels are allowed to decrease by 5%-as the rod withdrawal occurs during the startup.
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4 .'. QUESTION. 5.03 (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 change of 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 initially (INCREASES, DECREASES, STAYS THE SAME)? (0.50) q 1
- b. Why does indirect production of xenon change during this {
transient? .(1.00) ! 4
- c. The final steady-state xenon concentration will be (LESS THAN, GREATER THAN, EQUAL TO) the original concentration. (0.50) l l
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.1 QUESTION. '5.04 (2.00) ,
Will the Departure from Nucleate Boiling Ratio 1(DNBR) INCREASE,' DECREASE,.
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or REMAIN 'fBE SAME if. the following plant parameters 'are measurably REDUCED during power operation?- Consider each-parameter separately,
- a. Reactor Coolant System'(RCS)' Pressure
- b. RCS Tavg-Temperature
- c. Reactor Power l
- d. RCS Flow i
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-QUESTION. 5. 0!i (1.00)- !
Which one af the following conditions would cau'se the Moderator-Temperature _ Coefficient to be LESS NEGATIVE?
.a.- bbl,900.ppmboron,withplantheatupfrom5'7'Fto550 4 Fusing rods'only.
- b. ;BOL, 10% power,- and increasing boron concentration from 850 to-900. ppm.
- c. EOL,.50 FPm boron, with power decrease to 50% 'from 75% using control rods,
- d. EOL,>1200 ppm boron, plant heatup in progrees for startup from-
-350 F to 400 F on' Reactor Coolant Pumps.
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QUESTION 5.06 (1.00) Which one of the following is the LEAST significant contributor to the Power Coefficient during normal full ~ power operation?
- a. Void Coefficient
- b. Moderator Coefficient
- c. Pressure Coefficient
- d. Doppler Coefficient l
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s , QUESTION 5.07 (1.00) Answer the following regarding effective fuel temperature.
- a. How (INCREASE, DECREASE, or REMAINS THE SAME) does the EFFECTIVE fuel temperature change over core life for a given power level?
(0.5)
- b. What TWO factors cause the change in EFFECTIVE fuel temperature?
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i i' ; s , b 3 , 7 -QUESTIONL .5.08- ' i . 00 ) -A - ( 0. 5) . L:
'For each case below indicate;WHICH,-item 1 or item 2, would produce the- l l: highest-reactivity worth, o
- a. '1. Dropping a rod"at power.,
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- 2. A single rod stuck out following a reactor' trip. J Total worth of all rodsat:
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QUESTION 5.09 (2.00) The plant is operating a 30% power when one reactor coolant pump trips due to an electrical fault. The control systems are all in automatic. Bank D rods are initially at 165 steps. Assume the plant does not trip. Indicate how EACH of the following STEADY STATE parameters would change: INCREASE, DECREASE or wa hea or+ REMAIN THE SAME. [caad'Jafe ma'af 54aic asrumpflea q;,, + - s.,H6 feedhaeL - sen ,adicafe. c.heiae .)
- a. Turbine power
- b. Reactor power (final)
- c. Final rod height
- d. T-avg (affected loop)
- e. T-avg (unaffected loop)
- f. Core Delta-T I
- g. Delta-f (affected loop)
- h. Delta-T (unaffected loop) i I
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QUESTION 5.10 (1.50) i How would an INCREASE.in each of:the following affect the fuel centerline temperature? Answer with INCREASE, DECREASE, or REMAINS..THE SAME.
- a. Coolant Quality
. .b . Coolant Flowrate
- c. Coolant temperature
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9 QUESTION 5.11 (1.00) <
-What are the TWO principal factors affecting the redistribution i reactivity effect in the reactor core?
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QUESTION 5.12 (1.50) What FIVE reactivity groups (factors) are considered when calculating Shutdown Margin? 1 1 I 4 I l i l l (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
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' QUESTION-5.'13 (2.bO) ,
Answer the following related to Doppler-only Power Coefficient.
- a. How does Doppler-only Power Coefficient change as reactor power is increased _(MORE NEGATIVE, LESS NEGATIVE, NO CHANGE)? (0.5).
- b. Besides~ reactor power, WHAT THREE factors affect Doppler-only. Power Coefficient?- .(1.5) :
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QUESTION- 5.14 (1.50) Plot the Xenon curve vs.' time for.the. follow $ng transient history.' Plant operates at 50% equilibrium conditiens.for four hours, then the Power Dispatcher orders a load decrease"to'25% which'is-followed: four hours 'later with a power increase to 100% power. 100%
; power is maintained for 16 hours until a trip. occurs. Trip recovery is made'to.-50% power 20 hours after.the trip and held at 50%.
ASSUME ALL. POWER CHANGES ARE STEP CHANGES. DO NOT PLOT THE POWER . HISTORY. END YOUR PLOT SIX HOURS AFTER REACHING 50% POWER FOLLOWING. THE TRIP-RECOVERY. MAKE YOUR PLOT ON THE AXIS PROVIDED'BELOW. INDICATE THE STARTING VALUE FOR.50% EQUILIBRIUM XENON IN PCM ON THE- ' GRAFS. P 6-l C l M i 5-! X l e- !. n 4-l o ! n ! 3-! x l 1 l 0~2-l l 0 l 0 l 1-l l l
!. - 1 l l l l l 0 10 20 30- 40 50 time in hours 1 l
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QUESTION 5.15'- .(2.00) The' plant is in Hot Standby with the RCS pressure being maintained.at 985 peig. A pressurizer PORV.is slowly discharging-to the pressurizer relief tank which is at ~5 psig. The steam quality'in the pressurizer steam )
' bubble is 100 percent. I
' a. What is the enthalpy of:the fluid' entering the PRT?
.b. What is the temperature on the tail pipe downstream of the PORV?
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1 [ ' QUESTION 5.16: (1.00) , What'is the. adverse' affect of operating a centr $ fugal pump at-its l; shutoff head'AND what design feature is' incorporated in systems to )- preclude this adverse affect? , q l l l .1 J { , l l l l l 1 1
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QUESTION 5.17 (1.50) What THREE parameters in a fluid system determines the AVAILABLE NPSH7 l l l l l l l b i l l i 1 1 1 l i i I 1 l l l l l l l l l l (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) l l E-_-__________________________-_____ _
QUESTION 5.18 (1,ony Define the term Nucleate. Boiling. 1. l l l (***** CATEGORY 5 CONTINUED ON-NEXT PAGE *****)
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,; QUESTION: l5.19 (liS0) l What'are THREE consequences.of cavitation on a' centrifugal. pump?'
THESE ARE NOT SYSTEM WIDE EFFECTS. ,
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) l ! i 6, PLANT SYSTEMS DESIGN, CONTROM AND INSTRUMENTATION Paco 21 QUESTION 6.01 (1.00) Which one of the following is NOT a' design. feature-of the flow restrictors'in the? Auxiliary Feed lines to the steam generators?
'a. Equalize flow to'all steam generators.
H
- b. LPrevent contributing to' pressurizing the containment ^due to steam line breaks.
- c. Prevent runout of the AFW pumps.
- d. Prevent feeding the steam generator with a feedline: break.
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' QUESTION- 6.02 (1.00)
Which one of the following BEST describes ~the method ~of providing oil
-flow past the heat exchanger for the Reactor Coolant Pump-Thrust Bearing oil-system?
l a. The heat exchanger is submerged.in the oil' reservoir.
- b. Cooling flow'is from an auxiliary AC motor package.
- c. Cooling flow is from an auxiliary DC motor' package.
- d. 'A Viscous pump on,the thrust bearing.provides flow.'
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- QUESTION 6.03 (1.00)
Which one of the following is CORRECT regarding operation of the Reactor Coolant Pump seal system?-
- a. Seal. injection water temperature may be increased to 150~F if-RCS temperature-is limited toiless than 400 F.
Q ! b.. The RCP should not be started if 81 seal leakoff'is greater than:0.3 gym and less than 2.5 gpm. q
, c. :325 psig is the lowest pressure-at which the RCP can be operated and still maintain the required 100 psid across the #1 seal.
- d. #3 seal. differential _ pressure is' maintained by the}#3' injection l head tank and the atmosphere.
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QUESTION 6.04 ( 1. 00 -) Duria;g normal operation, a steam flow transmitter fails low. Which one of the following statements CORRECTLY describes the response to this failure?
- a. Alarm received for Lo FW flow, feed regulating valve goes'ful3y open, steam generator. level. increases f b. Alarm received on Hi FW Flow, feed regulating valve goes-fully I closed, steam generator level' decreases
- c. Alarm received for Hi FW flow, feed regulating valve fails as is, l
steam generator level remains stable
- d. Alarm received for Lo FW flow, feed regulating valve fails as is, feed water flow increases l
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. QUESTION 6.05 (l.00)
Which one of the following statements' CORRECTLY describes the effect' on rod motion of a Power Cabinet URGENT failure?
- a. Rods will not move in AUTO, rods will-move in MANUAL,. rods supplied by the unaffected power: supplies will move in-individual bank select.
- b. Rods will not move in AUTO, rods will not move in MANUAL, rods; l supplied by the unaffected power supplies will.not move in
' individual bank select.
- c. Rods will move in AUTO, rods will not move in MANUAL,' rods supplied by the unaffected power supplies will move in individual bank select.
- d. Rods will not move in AUTO, rods will not move in MANUAL, rods j supplied by the unaffected power supplies will move in individual ban?% select.
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-QUESTION, 6. 0'6-(1.00) i j Which one of the following is a function of the Steam Generator Blowdown I monitor RE-0021, other.than. producing panelpalarms? '
i
.a. Indicates activity in the~backflushable filter..
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- b. Closes discharge valve to the retention basin.
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- c. Closes blowdown system,' inlet valve.'.>
- d. Takes:no automatic action. '
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... ,, ' '. l q QUESTION 6.07 ,(1.00)
'Which one of the following ensures that'ONLY one'AC vital bus is powered from the regulated transformer. backup power supply at,any.one i time during normal routine maintenance? l l
a._ Administrative control'. ' l l
- b. Key interlock system. j
- c. ' Electrically operated interlocks.
- d. Direct mechanical interlocks. 1 1
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. s QUESTION 6.08 -(1.00)
Which one of the following power range instrument output signals passes through an ISOLATION transformer? f- a. Loop flow permissive l 1
- b. Power mismatch signal 1
- c. High flux rate trip
- d. NIS panel power meter l
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4
- QUESTION 6.09 (2.00)
List FOUR systems.or. instruments which are supplied with an auctioneered
.HIGH Tavg signal.
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i QUESTION 6.10 (2.00) List the TWO interlocks and permissives associated with the Steam Dumps including setpoints, coincidence and the reason for the interlock or permissive. i 1 i i l i i I (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) a I
-QUESTION 6.11 -(2.00)
List the FOUR signals which will automatically close the Steam Generator Blowdown containment isolation valves. l 1 l I i i i i l t I (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) E---_ _ _:_ _ _ _ _ _.
QUESTION % 1.50) What THREE conditions must e t t reset the Turbine Driven Auxiliary-Feedwater pump trip and throttle valve - P l i l i k i I i l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) i
1 l QUESTION . (2.00) i List FOUR major loads whic be affected by failure of the 125 VDC j
. class 1E bus 1AD1. DO NOT inclu (de 4 M 480 V, 13.6 kV switchgear; i misc. equipment panel, Hvac or termination cab n .
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l l QUESTION 6.14 (2.00) ! List FOUR permissives required'to allow the Safeguard Sequencer to shut the Emergency Diesel Generator output breaker. l l' i (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) l l, 1
1 l s . l QUESTION 6.15 (1.50) l What does each of the THREE DP tran.smitters (a, b, c) used for one i train of the Reactor Vessel Level Innica. tion System measure? I I l l 1 1 1 l I i i l l l l i 1 (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
i , QUESTION 6.16 (1.50) i B:leifly. describe the function or operation of each of the following j interlocks / components on the Fuel Handling Machine. Include the setpoint if applicable. 1
- a. Low Load Inhibit
- b. Upper Limit )
- c. Slack Cable i
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s , , I QUESTION 6.17 (1.00) ] During a slow pressure increase transient PORV 455 automatically opens I to reduce the pressure but PORV 456 doec not open. EXPLAIN the reason ] why PORV 456 did not open. ASSUME that both valves functioned correctly, neither are stuck and the I absolute setpoints of each valve are identical. (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) w____-___-____
s ., QUESTION 6.18 (1.00) Expla.in how the diesel sequencer FOR EACB UNIT is reset following an Undervoltage condition where the diesel breaker never closes. 4 I I 1 i 1 j l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) l
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-l QUESTION' 6.19 (1.00)
Answer the'following questions in reference to the LOW Steamline-Pressure Safety Injection Signal'.
~
- a. What criteria must'be met to cause the safety injection-signal?. j Give actual setpoints. State coincidences if applicable. 1 i
- b. What criteria must be met to block'the safety injection signal? l Give, actual'setpoints and state. coincidences if applicable. i t
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s . QUESTION 6.20 (1.50) ( t.c o) p. How is the Spent Fuel Pool designed to protect against gravity draining in the event of a downstream pipe break accident in the Spent Fuel Cooling system? I 1 i l I i (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
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; QUESTION. 6.21 (2.25)
What' conditions will cause-isolation of the Auxiliary Component Cooling
'? Water. system.to..the Reactor Coolant Pump thermal barrier?. Be specific !
and include any applicable setpoints? 1 l 1 1 1 1 1 k 1 l -J 1 J l 1 1 i
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[ (***** END OF CATEGORY 6 *****) ; i
7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY ~ Pada 42 '
AND-RADIOLOGICAL CONTROL l i
. 1 QUESTION 7.01 (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. l
.b. : Continuous warnings contained in'a caution (step) areLin j 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-l apply-either'to the whole procedure or just'to the first step. 1
- d. ' Unless otherwise specified, a required task needs tx) be: fully.
L completed before proceeding to,the next instruction; it is NOT L enough to begin the task and have some assurance-that it is progressing satisfactorily. 1; l-l . i-L p. 1 i l l-L i
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-QUESTION '7 02 -(1.00) .,
Which one of the following abnormal procedure does NOT require placing rod control in manual as an'immediate. action?:
- a. Failure of Turbine Impulse Pressure Instrumentation.
- b. Intermediate l Range Drawer,-N35 or N36 Malfunction, Failure of RCS Loop Narrow Range Temperature Instrumentation.
~
c.
- d. Uncontrolled Continuous' Rod Motion.
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s s QUESTION ~ 7.03 (1.00) 1 Which one of the following would be'the CORRECT use of E-0,.
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" Reactor Trip or Safety Injection", with'a loss of off-site power ~and.
> . failure of the Emergency Diesels to' recover the' vital buses?
- a. Go immediately:t'o ECA'O.0, Loss of all A/C powerfprior to entering E-0.
- b. Go immediately to-ECA O.0.from E-0 after, verifying reactor and turbine trip.
- c. Go immediately to ECA'O.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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il b QUESTION- - _7 . 0 4 - (1.00), The control room operators:are performing FRP-C.2, " Response to Degraded Core Cooling'f, in response to an orange path condition shown on the core cooling Critical Safety Function (CSF) status tree. Which ONE of the following statements is CORRECT with regard to-
' transitions out of this-procedure 7' a .. .The operators must leave this procedure before completion and go to FRP-S.2, " Response to. Loss of Core Shutdown", if the suboriticality status tree indicates:a yellow path condition,
- b. The operators may leave this procedure'at.any step if the core' cooling. status tree indicates'a green path condition.
- c. The operators must leave this procedure before completion'and.
go to FRP-H.1, "Reeponse to. Loss of Secondary. Heat Sink", if
~
the heat sink CSF status tree indicates a red path condition.
- d. The operators must, leave this procedure ~before completion'and go to FRP-J'1, " Response to High containment Pressure", if the containment CSF status tree ~ indicates an orange path condition.
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QUESTION -7.05 (1.00) l FILL IN.THE BLANKS
~
Complete the following statements concerning the Precautions'and. Limitations of procedure 93300-C, " Conduct of Refueling Operations." ll;) e
- a. During core loading / unloading, no more than. ~
fuel assembly / assemblies shall be out of its storage l rack at any given- ' '
' time in the fuel building. (
- b. No more than fuel assembly / assemblies shall.be'out of E the;
. reactor vessel at any.given, time in the' containment. -
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i . QUESTION' '7.06 (1.00) :l l Which one-~of the following applies to GENERAL:radiationjwork permits?'
-l a;i May.be issued for'a period of one year.
l May be placed in inactive status if not used'in 3-days. l 1 l b. s l ll c. Reviewed on routine basis to ensure adequate' radiol'ogical l controls. . .- Entry records' maintained for each day including time of: day and I, d. l:
' dose received.
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QUESTION 7.07 _(1.00) Place lthe following major action categories in-proper sequence according
'to procedure 19211 " Anticipated Transient Without. Trip Response."
- a. Verify suberiticality
- b. Verify automatic actions .,.
.c. Check sources'of positive reactivity..
- d. . Emergency borate i
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QUESTION 7.08 (2,00) List in order of priority, the THREE methods of regaining core cooling when using FR-C.1, Response to Inadequate Core Cooling? 1 1 1 l l j l l (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) l l
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;1 QUESTION, 7.09 (l.50) .!
.l Procedure 12000-C,." Refueling Recovery", states-for modes 4, 5. and ,
6 with the' Reactor; Vessel Head.on at least one' Cold' Overpressure Protection System shall be operable. .; List.the THREE.di2ferent; systems used for Cold Overpressure' Protection and provide, system setpoints. a-i l i l 1' 1 I (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)- 1
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$ QUESTION:. 7.10. .(1.00) '
..~ . .
Name the FOUR types of: CONTINUOUS-actioni: steps used;in.the Emergency Operating Procedures, , i
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. QhESTION- 7.11- (1.00)
State 700 possible'. conditions ~in which Emergency Operating Procedures -
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a
' QUESTION: . (1.00)
According to procedure 190. , actor Trip and Safe y In,iection", what action should you take to. restore p he.AC. Emergency Busses?. O W
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, ,4 QUESTI N' 7i13. .
( l'. 00 ) . List she1TWO meth'ods which would be used to. identify a RUPTURED Steam Generator as outlined'in-E-3, " Steam Generator: Tube Rupture." t. I. f f i i l l . 4 l l 4' l l l 1 1 l l r 1 (***** CATEGORY 7 CONTINUED ON NEXT PAGE
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s. q QUESTION' 7.14 (1.50)
-A given radiation area is identified as requiring a; lock, but due to the; physical dimensions of the area and obstructions, a lock canLNOT be l ?
L.- used.- WHAT are THREE alternative methods that can'be used to satisfy the requirements ~for locking the area? 1 i ! -1 ( l 1 l l l l l (*****~ CATEGORY 7 CONTINUED ON NEXT PAGE *****), _ nL_-_-----------_--_---_-_--__-------- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - _ _
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; QUESTION _: 7.15 .* :(1. 50')i 1 w
' Namel THREE types of. areas where a; Specific Radiation; Work Permit 'is.
"- crequired for, entry. --' .
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I QUESTION 7.16 (2.00) List the FOUR criteria which must be satisfied before procedure 19011, '
" Safety Injection Termination",-is entered.
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'QUESTI N . 7.17 - ( 1. 00 ). 't Why must the Hydrogen Monitor isolation. valves;' remain' closed exceptr ,!
64 1during' hydrogen monitor' operation, according to the System Operating
' Procedure, 13130-1,." Post-Accident Hydrogen Control?" l l,
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CATEGORY 7 CONTINUED ON NEXT PAGE w i j i _?____--.-_------__--_-----_.-.._-_.--_____.--.___ . - - . _
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,4 QUESTION 7.18 '(1.00) .
During a reactor startup using procedure,12003-C, " Reactor Startup",- .. there is'one PRECAUTION which requirestthe termination of rod withdrawal if a' certain Leondition' is: expected' to occur. Whatfis the condition that-will. require.the termination uf, rod withdrawal? . c , 1 t t i ,. Il i: .j l i I l t
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-IQUESTION. ,7,19 . ( 1 ~. 50 )
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c ; List all situations that-require'a Shutdown Margine calculationitolbe" -
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. performed according to Technical Specifications?. 1 t
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- QUESTION 7.20 (2.00)-. (Loo).A As the oncoming' Shift' Supervisor, you see the following. conditions existing while performing procedure 12001-C, " Unit.Heatup-to Hot Standby-(Mode 5 to Mode 4)."
.i
.A RCS heatup of 49 F/hr.'is in progress with. temperature of.190 F and' pressure is at 350 psig.. Reactor Coolant Pump number 1 is operating. Samples of.RCS and Pzr boron indicate 1200 and 1240 ppm
.respectively. 'A boron dilution is in progress to reduce-concentration-to 1050 ppm for the Estimated Critical' Concentration.
Reactor Trip breakers are-shut and-all shutdown and control banks are. withdrawn two steps. Both oxygen.and lithium are within'11mits-for plant temperature. INDICATE AND EXPLAIN any abnormal or irregular conditions represented in the above paragraph according to the Precautions and Limitations of the procedure 12001-C. l-1' , .t l \ l 1 l i 4 l 1 i (***** CATEGORY 7 CONTINUED ON:NEXT PAGE *****) i
' QUESTION.
1 (2.00) i Identify the Techn Specifications , by- parameter .name, limiting value .I
' and time limit for those- cal Specifications where ACTION is to-be taken in LESS.THAN ONE HOUR afte eeding the specification.
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QUESTION 7.22 -(1. 00 ) . . I
-According to procedure 93210-C, "New Fuel Elevator Operating.
Instructions", WHOSE authorization,-by title, is required':to OBTAIN-and-USE the bypass key to raise a' loaded' elevator?
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. QUESTION 7.23'. (1.00).
What is the minimum pressurizer level and reactor coolant system subcooling recommended-for a. Reactor Coolant ~ Pump restart.if a void is
- suspected in the vessel following a Natural Circulation Cooldown?-
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QUESTION 7 24 (1.00) While performing EOP-19231-1, " Response to Loss of Secondary Heat Sink". WHY are the Reactor Coolant Pumps tripped if Auxiliary Feedwater cannot be established? l, i 1 l i l l l 1 I l (***** END OF CATEGORY 7 *****)
8; ADMINISTRATIVE PROCEDURES. CGnpITIONS. P.?ge 66; AND-LIMITATIONS- ,
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; QUESTION' 8 .' 01 ' '(1.00)
^
Which one of'the following.is NOT. required'for each;on-coming operator to, preform prior;to assuming the. shift duties?- u a. Complete the Emergency Safeguards Equipment Checklist.
- b. Review: Narrative' log, round sheets and checklist.
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c; Initial'each document reviewed.. ,
- d. DiscussErelevant items affecting plant ^ operations with the off-going counterpart. -
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QUESTION 8.02 (1.00) , Which one of the following reator trips would NOT require a Reactor Trip Report Data Sheet 1, por Administrative procedure 10006-C,
" Reactor Trip Review?"
i A trip with...
- a. all shutdown rods withdrawn and one control rod withdrewn.
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- b. group 1 shutdown rods withdrawn.
- c. automatic openin6 of the Reactor Trip Breakers with one or more rods withdrawn.
I l d. opening of Reactor Trip Breakers during rod drop time testing. l l l l l l l < l i J l l l l l 4 (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)
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QUESTION 8.03 (1.00) . Which one of the following events does NOT require a ONE HOUR 1 f notification per 10CFR50. l i a. The plant is in a condition NOT covered by operating and l emergency procedures.
- b. The loss of the off-site notification system.
- c. A valid automatic initiation of the Reactor Protection System.
- d. A shutdown was commenced because the plant was in violation of the Technical Specifications.
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l i QUESTION 8.04 (1.00) 1 Disabling of an annunciator input by means other_than the pulling of. cards is controlled by which one of the following?
- a. Unit Logbook entry.
- b. Clearance issuance.
- c. Temporary Jumper log entry.
- d. Maintenance Work Order system.
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4 7 .- .;- . a QUESTION 8.05~ (1.50). 4 , r, . L Fill'in the blanks. A Fire Team' consisting of at least1' [a]' members shall'be maintained'on site at all:. times. : The [b] .(title) shall designate the' Fire' Team captain and members at the:beginning:of:each: . shift. The Fire Team may be reduced-bye.one' person'for'a period.not-- to exceed: [c] -hour (s). , l J P 4
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QUESTION 8.06 -(1.00)
. Which one of the following,is' CORRECT concerning the use offproce'dures as describedJin procedure.00054-C, Rules for Performing Procedures?.
t
- a. If a'section of a procedure is NOT' APPLICABLE, each step must
'be marked NA. I
- b. If an evolution is suspended for greater-than 12. hours, reverification;of initial conditions is required
- c. When time is'not!available to have the procedure in hand, then read the. procedure afterward.and. verify'the actions.
- d. Aniexampleof~aprocedbral'errorwhereworkwouldstopis; finding an incorrect personnel title in a step.
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f (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) L_ _.__1_________[_______-.____._m____ _ _ . . _ _ _ _ . _ _ _ _ _ _ _ _ . . . _ _ _ . _ _ _ _ _ _ _ __ . _ _ _
i QUESTION 8.07 (1.00)
.Before the beginning of the shift, what must the OSOS ensure has taken 1 l place prior to appointing a person as the STA? j l
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)
I l 4 i (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)
QUESTION ;8.08 '(1.00) Which.one of the following is correct concerning'those requirements which are COMMON to all logs, according.to Operations Administrative procedure 10001-C, '"Logheeping?"
- a. When making log entries use black or blue indelible ink, and leave no blank lines.between consecutive entries. ,
1
- b. When necessary to insert additional information after the fact, note the entry with the actual time of.the event'and mark it "LE."
The last completed logbook is kept in the. control room until
~
c. the current logbook is completed then it is forwarded to the i Plant Superintendent for filling. k d. Copies of the completed logbook pages are collected by the Shift Supervisor and after local review are discarded.
\ i
\
is i l i 1 i (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) '
).
~
i a_____._________________.__________________________._______________________________ _ _ _ _ _ _ j
l : . l QUESTION 8.09 (2.50) 1 List FIVE General Work Practices which all personnel' assigned to shift l operations are responsible to carry-out as detailed in procedure 10000-C, Conduct of Operations. 1 i j 1 ( l i l 1 l l i l I l (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) l j
QUESTION B.10 (1.00) State the TWO personnel, by title, given the responsibility to analyze a trip event for the Reactor Trip Report according to procedure 10006-C, Reactor Trip Review?" i l 1 1 (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)
QUESTION 8.11 (2.00) What FOUR actions shall the Shift Supervisor PERFORM, DIRECT or ENSURE if a locked valve is found in a position other than the required position? i
/
I i l (**'*** CATEGORY 8 CONTINUED ON NEXT PAGE *****)
QUESTION 8.12 (2.00) t Who are the first FOUR individuals, by title, who can assume the I duties of Emergency Director? Consider the worst case conditions. / l l l t t l I l
-l l
i i 1 i (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) II hm_____.__-._,m __m__._m ._. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ , - . _ _ . _ _ _
1 i QUESTION 8.13. (2.50) l' List FIVE decision-making responsibilities which may NOT be delegated by the Emergency Coordinator? j i l
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i i 1 (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)
a I
.' =
l QUESTION 8.14 (1.50) I f List THREE situations requiring follow-up notification to the NRC I Operations Center after the initial notification of an emergency has l been made. I l 1 I i, i i 4 I I l i i l i l ll j I l (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) ! 1
l .- QUESTION 8.15 (2.00) l' What are the FOUR submaster key sets and where are they located? Be specific about location. l 1 l 1 l I i 1 i (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)
E l (QUESTION 8.16 '( 2. 00 )' According to procedure 00052-C, " Temporary Procedure Changes", changes
.in intent are considered to occur if steps are added or deleted which-could cause.FOUR things to happen. NAME.the FOUR things which are considered.to change the intent of a-procedure.
l (***** CATEGORY ~8 CONTINUED ON NEXT PAGE *****)
.a QUESTION 8.17 (1.00)
What are the TWO purposes of the quarterly check performed on EXTENDED. ACTIVE CLEARANCES? (***** CATEGORY 8 CONTINUED ON IfEXT PAGE *****)
l QUESTION 8.18 (1.00) According to procedure 00308-C, " Independent Verification", what are the TWO exceptions allcwed to the policy requiring independent verification? l l l l l (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)
7. hUESTION~ 8,19 -(2.00) .a 1
'According'to procedure 00306, " Temporary Jumpers and-Lifted Wires"',
l what are the. responsibilities of the Shift Supervisor?- ) d 1
'1
(
'd
(***** END OF CATEGORY 8 *****) (********** END OF EXAMINATION ~**********) s ___ _.________m~_._E---------------------- - " ' - - - - - - - - ' - - - ^ - - - - ~ ~ - ~ ~ '
^ ~
Q EQUATION SHEET !
~~
t l f " ** * * 'IE
,,, ,.,g+ g,g 2 Cycle efficiency = Net Work ,
(out) - -- I E = aC - a = (vg - v,)/t ,gg l gg = my2 ., ,, , ,g A = AN A = A,e j PE = msh w = 6/t = in 2/tq = 0.693/tg . W = v&P- (t,)(ts) AE = 931Am ~ ( + ) 6=$C,AT ~I* _I - I,e k=UAAT I = I,e "*
~
Pwr = Wg a" I = I, 10 *D
~
y.y go SUR(t), TyI,= 1,3/g y,y ,t/T HVL 6 0.693/u ! o <
~
SUR = 26.06/T T = 1.44 DT SCR = S/(1 - K,gg) j eff SUR = 26 CR, = S/(1 - K,ggg) (A-o CR y (1 - K,gg)g = Ch G - 3 ,ggT2 T = '(1*/o ) + [(6-o)/1,gg) o T = 1*/ (o - E) !( ~
*ff} " 1 T
- N ~ 8)I leff" M = (1 - K,gg)C/Cl ~ Kaff)1 8 * (Esif~l}! eff " eff off SDM = (1 - K,gg)/K,gg p= (L*/TK,"gg ) + h/(1 + A,ggT )) ,
t* = 1 x 10 ~3 seconds y = I(v/(3 x 10 0) 1,ggA= 0.1 seconds
~I I = Na Idgg=Id22 WATER PARAMETERS Id =Idg g
1 gal. = 8.345 lba 2 R/hr = (0.5 CE)/d (meters) I gal, a 3.78 liters R/hr = 6 CE/d (feet) - 1 ft = 7.48 gal. MISCELL\NEOUS CONVERSIONS ! 3 Density = 62.4 lbm/ft 1 Curia = 3.7 x 10 dps 10 3 Density = L gm/cm 1 kg = 2.21 1ha l Heat of va;orization = 970 Etu/lba 'I hp = 2.54 x 103 BTU /hr Heat of fusica = 144 Btu /lba 1 Hw = 3.41 x 10 Etu/hr 0 ' 1 Atm = 14,7 psi = 29.9 in, l'g, 1 BLu = 778 f t-lbf 1 ft. H 2O = 0.4333 lbf/in2 g inch = 2.54 cm F = 9/5'c + 31- 1
'C = 5/9. ('T - 32)
O O i l _________._________._______m_____.___._..__ _ _
- 5. THEORY OF NUCLEAR-POWER PLANT OPERATION. Pcga 85 FLUIDS.AND THERMODYNAMICS' M ASTE R
-ANS'WER 5 . 01. - .(1.00) a REFERENCE s
-VEGP Lesson Plan LO-LP-33410-02, p. 8 Lesson Objective LO-LP-33410-02,- #7 -
192004K107 ..(KA's) ANSWER 5.02- (2.50)
- a. SAME AS
- b. LOWER THAN-c.- LOWER THAN
- d. SAME AS *
- e. SAME AS REFERENCE ,
h VEGP Lesson Plan, LO-LP-33610-01, p. 7, sec. II.F.3.b. Lesson Objective LO-LP-33510-01, #14. , , i 001010A201 001000K534 001000K104- ..(KA's) ANSWER- 5.03 (2.00)
- a. INCREASES (0,5)-
- b. Because the concentration of iodine (the xenon precursor) is decreasing. . (1.0)
- c. LESS THAN .(O.5) q l
l l l l l .(***** CATEGORY 5 CONTINUED.ON NEXT PAGE *****)- l r s
'Paga 86 Si THEORY'OF-NdCLEAR POWER' PLANT' OPERATION.
FLUIDS.AND THERMODYNAMICS: 1 5 REFERENCE 1 b VEGP Lesson Plan, LO-LP-33530-00,' Sec. II-D, pp. 7 &^8-
-LO-LP-33430-02,LSec. II-D, p. 7-
' 192000K108, . . (KA's) u ANSUER- !5.04 . (2.00) ,
* ^
(0.5-each) .
- a. DECREASE b', INCREASE- '
- c. INCREASE-
- d. DECREASE LREFERENCE- .
WEST T-H'PRIN PG 13-23,24 VEGP Lesson Plan, LO-LP-34510-00, pp 11-13 193008K105 . . (KA's) I i ANSWER 5.05 (1.00) , l b 4 REFERENCE VEGP Lesson Plan, LO-LP;33420-02',~pp 5-.8 Lesson Objective, LO-LP-33420-02, #7-9
;j 001000K509 . . (KA's) -j i
ANSWER 5.06' (1.00). e i i I (***** CATEGORY 5 CONTINUED ON NEXT-PAGE-*****) v
-mmm-____m__-.__m_m. m.__am__-.__m._____.-m ___ _ _ _ _ _ .
3 =
. .- u-
.5; THEORY OF'NUCMAR POWER ~ PLANT OPERATION. Paca 871 ,
FLUIDS.AND THERMODYNAMICS.
' REFERENCE 7VEGP Lesson Plan,:-LO-LP-33210-01,~pp 17-20
..;s .LO-LP-33420002. p. 7.
Lesson Objective, LO-LP-33420-02, #14
-192'004K108 ..(KA's)
ANSWER' 5.07) (1'. 00 ) . f
- a. DECREASE: .[0.5] #
b .' 1. = Clad creepL [0. 25] ? .
, 2. -Pellet (fuel) swelli -[0. 25]? ,
[ang 2 4'e'r~ '! o.15-l p c. enck
)
- 3. . ' V;sslon . gas t., h e g.p
. REFERENCE ,
VEGP Lesson Plan LO-LP-33220001, p.':4 -( . LO-LP-33410-02. R, Le'ason Objective LO-LP-33410-02, #10 i
'192004K107 ..(rA's) .
ANSWER 5,08 31.00)/" (0,5)
- a. 2
-t. . 1 )*' . l REFERENCE -t 4
VEGP Lesson Plan LO-LP-3345b-01, p. 10 Lesson Objective LO-LP-33450-01,;#4 ' 001000K511 ..(KA's) '! i
. .c i l
Q I
-j z
i 0 (***** CATEGORY 5 CONTINUED ON,NEXT PAGE *****) -l l i
- u. .__ _ _ _ _ _ . .-
- 5. THEORY;OF MUCLEAR POWER' PLM.T OPERATION. Paca 88.
~ ~
GUIDS.AND THERMODYNAMICS
~
ANSUER' 5.09' . ll,a,Qe. (2.00In ) (ae Me Assump-lo. o k. . p,,,,,,, Pres s6ee. F*eedback (0.25:each),, y,,)Qg gg (o})l n3 if Ae. ~ 1
- a. REMAINS.THE SAME- a. Jecrease. ,, g.j g, j decreaas b.. REMAINS THE SAME 6 .,pg f '
- c. REMAINS'THE SAME' c- Me e rca se- . .
.or d.. # +" #5 585" '##
d., DECREASE. decreas c.
- e. ' REMAINS-THE SAME e. decrease ; A=4 F re55d!Y.
- f. INCREASE- 4' incre,,, Meek is ss -
DECREASE-g.
.p. ecrease
- h. INCREASE h; inc ressw. .j$ l ;
REFERENCE VEGP Lesson Plan LO-LP-34210-00', various pages Lesson Plan LO-LP-34310-00, various pages. 000017K104 ..(KA's) , 4 ANSWER 5.10 (1.50) (0.5 each)
- a. INCREASE 1
- b. DECREASE )
- c. ' INCREASE {
j REFERENCE . VEGP Lesson Plan LO-LP-34210-00, pp 15, 16 Lesson Objective LO-LP-34210-00, #10 l i d 193008K114- ..(KA's) !
- l 1
' ANSWER 5.11 (1.00) !
l
- 1. Moderator. reactivity-Fuel ~ burnup. !
'2. .
(cere a3th
.i i
s l l (***** CATEGORY 5 CONTINUED.ON NEXT PAGE *****) I l
. l
;5: THEORY"OF NUCWAR POWER PLANT' OPERATION. '
Pzca 89' l FLUIDS.AND THERMODYNAMICS -l 1 1
. REFERENCE- .
- VEGP-LessonLPlan, LO-LP-33420-02,ipp'7a-b Lesson Objectives,.;LO-LP-33420-02, #13 ,
001'000K529 ..(KA's) i i 4 .
]
ANSWER: . 5 .12 --- (1.50)- Ans 5 Be , (0.3 each)L
- 1. Boron concentration 2
- 2. Control rod position
- 3. 'RCS:Tave 4
- 4. Fuel burnup:(core age) 5.
- 6. -: Sa Xenon concentrationbian m a r ,*u m con ce n +ra
-l H
l REFERENCE s VEGP Lesson. Plan, LO-LP-33510-01,~p. 6 !
' Lesson Objective, LO-LP-33510-01, #5 192002K110 ..(KA's)
ANSWFR' 5.13 (2.00)
- a. LESS NEGATIVE [0.5] b..l.McIlab y'f O c am* I /** d'c E*V 2, Octl e4len* ef U-238 and buiUe-f up.3 Pa-Ma
- b. 1. Ccciant Tave [0.5] 3, Fuel dens 4 Ace Men ~
-2. Cerc age [0.5] .5.
- 4. coelad Cl4J cre* P,Tave.
- 3. Cled crcep. [0,5; .
. Tve e.ase
((Core ecediled as erne. & c4 ca~ ln Subs +14ukow of $e "fo'es+ l-lls
-ra a+e ,, ce a fee) J o.5 p+. eaeh )
REFERENCE:
VEGP Lesson Plan L'O-LP-33410-02, pp 9-10 Lesson Objective LO-LP-33410-02, #9 & 12 0 192004K107 192004K113 ..(KA's) '
)
j i l l l ' d i (***** CATEGORY 5 CONTINUED ON NLxT PAGE *****). 8J i
- i. .
q
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. Pcco. 90 FLUIDS.AND THERMODYNAMICS y
l ANSWER 5.14 (1.50) , I I
-P 6-l C !. oo M l o '{
5-l o ! x l o 1 e- l- o n 4-l- o o o- ,l n o o 3-l o o-x ! oo oo o 'o 1 l~ o- o o o o 0 2- l - o o o 0 -l ooo 0 l 1-1 l l l l l l l l l l l l l 0 (4) (8) 10 20 (24) 30 (32) 40 (44) 50 time NOTE: DO NOT deduct points for values of Xenon curve peaks'except as i noted GRADING: k Start point Xenon Concentration 2250 +/- 100 pcm [0.5] i Break points on curve for' time reference (4, 8, ...) +/- 2 [0.5] I General Shape of curve [0.5] ] REFERENCE VEGP Lesson Plan LO-LP-33430-02,- various pages Lesson Objective LO-LP-33430-02, #8, 9, 10 192006K113 192006K112 192006K111 ..(KA's) i I (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) I I f
- 5. THEORY OF NUCLEAR POWER' PLANT OPERATION. Pcga 91 FLUIDS.AND THERMODYNAMICS ANSWER 5.15 (2.00)
- a. 1192 BTU /LBM (1190-1195 BTU /LBM acceptable)
- b. 500 F (290-310 F acceptable)
REFERENCE
.VEGP Lesson Plan LO-LP-33110-03, various pages-STEAM TABLES
- 193003K125 ..(KA's) l l
ANSWER 5.16 (1.00) l l (0.5 each)
- 1. Overheating l
- 2. . Recirculation lines (mini flow systems)
REFERENCE VEGP Lesson Plan LO-LP-34410-00-0, p. 18 191004K104 ..(KA's) l ANSWER 5.17 (1.50) 1 (0.5 each)
- 1. Fluid pressure (or c I'v"4I*" Ps s"'*
- PI " * #"U P'cs'a<< oc
- 2. Fluid temperature Pah4tc)
- 3. Fluid velocity (or volu, 4,-/<
floac44e)7 l REFERENCE VEGP Lesson Plan LO-LP-34410-00-0, p. 22 Lesson Objective LO-LP-34410-00-C, #10 191004K106 ..(KA's) L L (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) I l l
5; THEORYOFNbCLEARPOWERPLANTOPERATION.' ,
- Paga 92 FLUIDS.AND THERMODYNAMICS ANSWER- 5.18' . ( 1. 00 )1 Boiling thattoccurs when small-bubbles are formed'at the. heated surface
~
Land move off into the liquid. REFERENCE VEGP Lesson-Plan ~LO-LP-34210-00, p. 9 Lesson Objective LO-LP-34210-00, #4 9 193008K103' ..(KA's): 1 1 ANSWER 5.19 ,(1.50)
-(0.5 each)'
~
1.
- 2. -Excessiv(e noise (and vibration.)Erosionand.pittingof.pumpsurfaces.) -
- 3. Ex:r;; in gle::::
Reduth, 4)L>rde(and lower pump ef ficiency'.)) l pm REFERENCE VEGP Lesson Plan LO-LP-34410-00-C, p. 22 Lesson Objective LO-LP-34410-00-C, #10. ]l 191004K101. ..(KA's) i I l
-l q
i
~ ,
l (***** END OF CATEGORY' 5'*****) ! u_ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _
.l i
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, ' .$ . .? . PLANT SYSTEMS! DESIGN. CQNIRGId.lHD_.lESIBUtfElffhIIDE ; -y -T i , 4 F; d' 8.3 . e y t En , . ,1-
. _'i'
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' ANSWER; . 6 . 01 ' (1.00)' - n' ,
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.n.' 8
,I'0 d- '
l 4 PI-!'.h'[l
.- i -'~ . '
.Q,,, e 5
l ,
' REFERENCE ' '
;. i W' (
.v ,
VEGP Train'ing T' ext, Vol. /7, : Chap. ,13' d. , p. ' 13d-6 /- .jf. f '
'h .
i *
'0610f. 4K404 . .(KA's)' '
n , J i j
'4
;p L ,
i 4 fr . " M q . i ! . ANSWER 6.02- (1.00): ,
, , s.
C '
~'
d '
<. y, REFERENCK <
a.. . i l , , s .. l VEGP Training Text, Vol'. 5, Chap. 1.a.,'RCS, p. la-8 -
./c l 0 :
I 000000K603 003000K403 . . (KA's) ,1
.( , .i L .a y s
-)
ANSWER 6.03 (1.00)' ,
,.,, .q,
, j a
l l pl' 1 I REFERENCE 1 VEGP Training Text, Vol. 5, Chap. la,-RCS, pp la-23.through la-27s ' 003000K602 003000K404 < .003000K407 ..(KA's) ! i. i
.C.04 (1.00)
ANSWER j
< ;b .
1 L 4 t ( ****
- CATEGORY . 6 CONTINUED ON NEXT. PAGE *****)
4 i l',
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m+ a :ma p ..p; : yy ,s. + ,'
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v,;
, , .. + .-.
7
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- y. y ,
y y, y ;9. , ; y,. .; ,. - 4 .. 1 t -. . 4,y ' - 4 ,'
.. + -- . ,
y; Paga;94 3,:, M,g - / .. . r 1BW p. PLANI; q;;>iSYSTEMS- DESIGN. CONTROL. ~'AND INSTRUMKETA c,1 :
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w-
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,^,,i, H TREFERENCE. ; '
/. o g.gj je ,'
,. .w M .
.f'# w. o
. , ,',g,
'VEGP-l Lesson Plan LO-LP-21iO1 2 02fC, various YS '
+ .
.r~
035010K10b 0350idA204l -.
.(K *s)I,,, ,.
I ' '
.{ ; '/ -' i ,
- u. / -
.i 1 . . <-
.,. m .3, , ,
ANSWER 6.05LT (1.00) o 4 ,.
,a i .I
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, \
- ((
.s
<1-1 i' '.)..
, s '. s -
p , 5 tt
., 3
,l. ' \l 3 .
REFERENCE ,v . s
' . VEGP Lesson Plan 1LO-LP-27102-03, ' p'; 9 . , ,
Lesson Objective LO-LP-27102-03, 8tB s v , n o , .i i.. 1: s"' 001050A201 ' . . ( KA' s )1 - Y q E e' - f-.
<> -- s %
v
. i. ,
i .} il ;
-(1;00) J- /..
= ANSWER 6.06 -
y j .).
+ '
;4 v
4 g
'I .,- ( -
..J- s3 y
.. ; '., :o 3 ,
'e "* '
REFERENCE . w< ..A ,~
, VEGP Training Text, Vo,l.i7, Chap.'13.c., p. . .
13c-4 & SE , 000037A113 .. .. ( KA ' s ) - 5 .
' o
.l
.t ; i 3,
ANSWER 6.07 (1.00) w 1 a
'e . r
. REFER,ENCE .
+ s VEGP Training Text, Vol. 8,. Chap. 16.b'., p. ~16b-4 < .
063000K402 ..(KA's) > l ,
, s
\ . t
^
L r Lg 1 8 y 7 (*****' CATEGORY. 6 CONTINUED ON NEXT.PAGE *****) s
-._____.-_-______._________E__'l__._m______
' _ _ _ _ _ _ _i____.,__,,..___.______,____.______,,________________,_________ _ _ _ . . _ __ , _ _ _ _ _ _ _
n= _ ;- . , _ . C kh,':PLANTSYSTEMS4DECIGN'. CONTROL'. MANDf1 INSTRUMENTATION: -/- Paca 951 u ,, , ,. s ,, ,
<4 I
C ANSWER. , 6.08 (1.00) , b" , f U iREFERENCE. .
.4 . . . 4 M
,VEGP Lesson Plan LO-LP-17301-01, Figure 3-r Lesson Objective'LO-LP-17301-01,;#2 ,
s y ,015000K601 ..(KA's) - 4h J
, ANSWER 6.09 (2.90) 4 ,J An3 4 of ,
(0.5 each) '
- 1. . Rod controli. ,
- 2. Steam-dumps .
i s
- 3. Pressurizer level
'4. Control rod insertion limits. r St TAVE ~ Tacec r~ec** der' y ,
.REFERENCEi s VEGP Training Text, Vol.~5, Chap.-1;a,,.p.~1a-19 .
7 002000A109 016000K403 ..(KA's) T :. {' .. ANSWER- 6.10 (2.00) /Any 2atf.4eccQ 1
- 1. C-9. Condenser available [0.5] - . satisfied when scuum'> 17"'on 2/2. pts and 1//2Cire pumps running [0.25). Ensures condenser?
is capable'of condensing steam WO no'overpressurisation' occurs < I
,[0 25).
5 50 3 2. P-12 Low-Low'Tavg SI: [0. 5] - 2/4 loop Tavg ' < 4 rte F r[0.25) . i v , Prevents" inadvertent or. excessive'cooldown-lof RCS-[0.25]. l
- 3. P- 4 Reae+oe Teir [o.5] r %u A RTA BVA open [o.s2,]i Ak sleap damps &a}
REFERENCE - %;a 6 ~ RTB/slYB egen [o.89]. Achah I VEGP Training Text, Vol. 7, Chap.;12b, p. 12b-14 &'1'2b-15 j
, o., t 041020K409 ..(KA's)- 1 i
% C-7. t.oad Rejei% [o4]p : 10 2 :1o5): nLe% i i26 second$[o.15]and am !
s h ,., h - p [o.2.r}.~ ja. - l i
power. e
- 4. Trip.of both" Main Feed pumps.
l 5, i AMSA C se+ua4fon. M- '
.o , ",.
n.
?REFEhENCE: ,
a; , m VEGP Training Text,;Vol.-7, Chap.l13.c., p.513c-5 e. 1 ., 03501'0K403 '
..(KAs) 47 4
, s. ,
'ANSWE 12 (1.50). 3
- 1.
(0.5 each) X Local handswite rip'andIthrottle: valve in local.,
.; _2. 'Overspeed trips lnot pres a
- 3. Hechanical overspeed. reset at.va dei.sygg
REFERENCE e . VEGP Training Text, Vol. 7, Chap. 13.d., p. 13d-12_ 3 061000K407 ..(KA's)
'g.'.
"ANSWEP% :6.13' (2.00)-
(any.4 9 0. >
- 1. Diesel Generato Pi nerator Control power.(field flash).'
'2. '".. .
ntrol: Panel;A.
- 3. Solid State Protection System -I.
? Reactor Trip Switchgear Train A. b L4. S. Auxiliary relay cabinet Train ~A. 'Ng5c(. ~ N ;J ay , REFERENCE 1 VEGP Training Text f.Vol. 8, Chap. 16;b.,;p. 16b-5!, 063000K302 063000K3011 ..(KA's)~ )% , 4 (***** CATEGORY- 6 CONTINUED ON NEXT.PAGE.*****), m j v e y
,['
3 .] 'g, f 5 _ib..__ ._ .- _ _ _ _ _ . _ . _
._m.__
,__[____ __/_ _ _ _ _ _ _ _
'6. I PLANT' SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION, Pco 97 4
1
' ANSWER 6.14 (2.00).
(any 4: 9 0.5 each)
- 1. DG ready to-load ( > 440, RPM,.> 90% voltage).
- 2. Preferred incoming breaker open.-
- 3. NO. faults on. bus
- 4. .rB9-/ lose 2T ormissi . rom buslockouts)fi.f'cadesload*Ie2is.
sequencer cd /'l*.#c O
;5. No. lockout'on Diesel Engine or' generator.
4-REFERENCE VEGP. Lesson' Plan LO-LP-01401-01', p. 8' Lesson Objective LO-LP-01401-01,.#7al 064000K411' 064000K410 ..(KA's)
~ ANSWER 6.15 (1.50)
(0.5 each) (a* RC PI ** alan) DPa ' measures static ^ narrow range' level of the' vessel from.hotileg to top of head.(o, gcps c . ,3) DPb -measures'staticAwide range level from bottom'of' vessel'<to top of-vesse1. y,k,,,, b,m Avesse) { u , a. kea)(gcppavnni,3) DPc-measureI*81kferentialpressureAcf".CSflcu:::::pp,h: 4 cerc. REFERENCE , VEGP Training Text, Chapter id, p. 5 i 002000K603 ..(KA's)- ) i
- ANSWER 6.16 (1.50)
- a. Prevents bridge or trolley movement [0.3) Setpoint'1407 lbs [0.2] 'j
.. b. Stops hoist' withdrawal'[0.25] and clears low load inhibit interlock
.[0.253 'p i c. Stops downward. motion of hoist [0.3) Setpoint 20 *slbs.
ibs) [0. 2] . j l 1 j l 1 l I (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) L~- '
- 6. PLANT _ SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION: Paca 98 REFERENCE VEGP Training Text, Chapter 18a, p. 16 034000K403 034000K402 034000K401 ..(KA's)
ANSWER 6.17 (1.00) The master controller which controls PORV 455 has in integral portion [0.5] which.causes the controller output to increase as'the difference
~
between Pref and actual Pressure persist with time [0.5]. REFERENCE VEGP Training Text, Chapter 1.c, p. 16 000027A101 010000K403 ..(KA's) ANSWER 6.18 (1.00) [a,861 oc Unit 1 - must pull circuit card en,]d- g open supply breaker or 3[c.tVI] install jumper 00. 5] . [o.4673 Unit 2 --Reset pushbutton [0.5]. /* REFERENCE i VEGP Lesson Plan LO-LP-60301-01-C, Plant Differences 000056K121 ..(KA's) ! I ANSWER 6.19 (1.00)
- a. 2/f channels in 1/4 steam lines <= 585 psig. [0.5]-
4.
- b. P-11 met, 2/3 pressure channels / 1970 psig [0.50]
P' i l l l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****).
,6. PLANT SYSTEMS DESIGN. CONTROL. AND1 INSTRUMENTATION- LPags-99 il REFERENCEf
- VEGP Lesson Plan LO-LP-28103-03-C, p. 5 Lesson Objective LO-LP-28103-03-C, #1 013000K412: 013000K101 ..(KA's)
~
ANSWER 6.20 ' 1. 50-) .( l'.co) - A .
. 1. SFP cooling suction is located just below the surfaceiof.'the' pool (-4 ft). [0,53 '-
- 2. ; Return pipe ter=inates above the fuel rech (S ft) [0.5]f nd - fW s -B . - Anti-siphon holes just below the low water level of the pool [0,5].
REFERENCE VEOP Training Text,.Vol. 8, Chap. 18;b., p'. 18b-3 4
- 033000K403 ..(KA's)' J
; e ANSWER 6.21' (2.25) i
- 1. High Flow on individual pump'[0'.53 -- 69 gym [0.25]' ;
l
- 2. High Flow - common line for'all pumps [0.5) -- 200 gpm [0.25]. j High Pressure ocommon line for all pumps [0.5] -- 120-psig [0.25]- .4 eREFERENCE VEGP Lesson Plan LO-LP-04101-03-C, p. 14 Lesson Objective LO-LP-04101-03-C, #8d 008000A301 008010A301 ..(KA's)'
1 1 l . i 1 (***** END OF CATEGORY- 6 *****) i I
--------2. . _ . . ._. -.
m, -
I' a % ' ;}
yl'g: PROCEDURES ^ NORMAL. ABNORMAL. EMERGENCY ,. 4
- Pacold0- d i AND RADIOLOGICAL CONTROL- , C .,
'j 3
'~-
,-pl
, ANSWER- . 7.01: (1.00) H a
d' '
- , } ;
" ~
. REFERENCE Ij VEGP J Lesson Plan LO-LP-37002-02-C,-pp;7 & 8 '
Lesson Objective ~LO-LP-37002-02-C, #2 ;
+
Westinghouse ERG 1 Background documents.- 4
- ;7 :j
..(KA's)-
0000110012- ,
' ANSWER 7.02 .(1.00) b ,
~ REFERENCE
- VEGP Procedures, 18001-C, 18002-C, 18003-C 001000G015 ..(KA's)
ANSWER 7.03 (1.00) b 4 REFERENCE VEGP E-0 19001-1, Reactor Trip or Safety InSection" ECA-0.0 entry conditone. 194001A102 ..(KA's). 1NSWER 7.04' (1.00)- ' 1 a, c (***** CATEGORY 7. CONTINUED _ON NEXT PAGE *****) , u _ _
- 7. PROCEDURES - NORMAL, ABNORMAL. EMERGENCY Page101 AND RADIOLOGICAL CONTROL REFERENCE VEGP Lesson Plan LO-LP-37002-02-C, p. 10 Lesson Objective LO-LP-37002-02-C, #7 & 8 WOG EOP Background Document, Ch. F-0, 000054SG11 ..(KA's)
ANSWER 7.05 (1.00)
- a. I
- b. 2 REFERENCE VEGP Procedure, 93300-C, p. 2 l
034000G005 ..(KA's) ANSWER 7.06 (1.00) a or C, REFERENCE VEGP Procedure, 43007-C, pp 4 & 5 l 194001K103 ..(KA's)- ANSWER 7.07 (1.00) a,d,c b, 4, c, s g (MUST BE ALL CORRECT FOR ANY CREDIT) REFERENCE VEGP Lesson Plan LO-LP-37041-02-C, pp 8 & 9 Lesson Objective LO-LP-37041-02-C, #8 & 9 Student Handout LO-HO-37041-01-C-01, Self Test 000029G012 ..(KA's) (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)
e- . .
;q
~
17 . PROCEDURES NORMAL. ABNORMAL. EMERGENCY , PO4;2102 ^ I AND RADIOLOGICAL CONTROL s ; l
, ANSWER 7.08 ( 2 ~. 00 )L , ;
(0.5 each.plus 0.5=for correct order) g 'i
.1. Reinitiation of.high pressure safety injection 1) 2.3 Rapid secondary.depressurization 1
- 3. RCP: restart:and/or opening of pressurizer PORVs j i
REFERENCE VEGP Lesson Plan,1LO-LP-37061-01, p. 5 4 Lesson Objective, LO-LP-37061, #2 .] 000074K311 000074K103 ..(KA's)-
- , j
.M4SWER 7.09 .(1.50) .i l;
.t
-1. Two PORVs'[0.25] with variable lift settings lper Tech. Specs [0.253-
- 2. Two RHR suction relief-valves.[0.25] with setpoints of 450 psig 1
(=/-3%)-[0.25]
- 3. RCS depressurized with~ vent (0.25]' capable of' relieving 670 gpm @
' q 470 psig'[0.25]. j ll REFERENCE
] a VEGP Procedure 12000-C, Refueling Recovery, p. 2 1
010000K403 ..(KA's) I i ANSWER 7.10 (1.00)
-F&. 5 ccch)--(aag 4 a+ o,2 reach) .
- 1. :When (when / Then ) !
- 2. Then )
- 3. Control i
- 4. Maintain j Jr. Check H? ?
6 If (of /4hra) f> j I
'I l
l l
'l
.(***** CATEGORY' 7 CONTINUED ON NEXT PAGE *****) !
,/
,s
_ i Al =
g - -
; ;g ;-
7,
'P203103-
-7. PROctnURER NORMAL.-ABNORMAL. EMERGENCY' AND RADIOLOGICAL CONTROL" , .
o , b 1 , I
-REFERENCE; y.
4 VEGP Lesson Plan LO-LP-37002-02-C Lesson Objective LO-LP-37002-02-C, #4-L194001A102 - . . ('KA' s ) ,
~
ANSWER. 7.11 ~(1;00) ~
-(any.2 9lOl5 each)-
1 l. Transition to' plant. operating procedures. Transition'to plant operating procedures Ohile'on RHR 9'. cold'
- 2. ,
~ ,
-shutdown.1 .
- 3. _ On' RHR system ops at cold shutdown or on either cold ~1eg circul'ation or hotsleg recirculation.
(Sece;C.'a. ' ess ples ef pope,- 4,s.,hl4s** l an}}ecedea-esllc necep 4dle. Jo . p REFERENCE
, u LVEGP Lesson Plan LO-LP-37002-02-C, pp 12 & 13 Lesson' Objective LO-LP-37002-02-C,=#12. ,
i _000029G012_ ....(KA's)' ' ANSWER q : (1.00)
- 1. Start and Load' dies
- 2. Attempt.to re-energize bus fro A
REFERENCE < VEGP Lesson Plan LO-LP-37011-02-C Lesson Objective LO-LP-37011-02-C,-#4 , Student Handout LO-HO-37011-00-C-002, p._7, self' test 000007G012 ..(KA's)
.y ,
t- ANSWER ' 7.13 ' ,-.(1.00) , l' 1 '. ! Unexpected rise in SG level
- 2. - High radiation from any SG steamline 1
(***** CATEGORY. 7 CONTINUED ONLNEXT'PAGE *****)1 , o , Ii" +, i .
- 7. PROCEDURES - NORMAL. ABNO M IH.Q1 Pcco104 AND RADIOLOGICAL' CONTROL:
REFERENCE-VEGP Procedure 19030-C,-EOP E-3, Stes.m' Generator Tube Rupture,
- p. 2-
.000038A202 ..(KA's)
ANSWER 7.14 -(1.50)
. 1.. Rope the area off. (0 5) .
Flashing light installed at area boundary.:(o.f)~ 2.-
- 3. Postedras a HIGH radiation area. (o.r.)
REFERENCE VEGP Procedure 43005-0, p.~4 194001'K103 ..-(KA's) ANSWER' 7.15 (1.50) (aas 3 s+ 0 5 each )
- 1. Airborne Radioactivity Area.
- 2. High Radiation Area.
- 3. Lock High Radiation. Area.
- 4. Locse Co n 44- . n a+/e n A ,.e.. .
REFERENCE VEGP Procedure (3500-C, pp 3 &-4 194001K103 ..(KA's) ANSWER 7.16 (2.00) (0.5 each)
- 1. RCS Subcooled.
L 2. RCS pressure stable or' rising. l
- 3. Adequate Secondary Heat. Sink must exist.
- 4. Pressurizer level on scale (span).
l (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)
- 7. -PROCEDURES - NORMAL. ABNORMAL. EMERG5NCY Py_o105 AND RADIOLOGICAL CONTROL <
REFERENCE i VEGP Lesson Plan LO-LP-37022-02 Lesson Objective LO-LP-37022-02, #1 f Student Handout LO-HO-37022-01-001, Self Test j 000000A234 ..(KA's). 1 ANSWER 7.17 (1.00) To ensure containment integrity is maintained. REFERENCE ~ VEGP System Operating Procedure, 13130-1, p. 4 ] 028000A101 ..(KA's) l ANSWER 7.18 (1.00) If criticality is projected to occur below the rod insertion limit.
-l
(,, If e e' 4 n'ea I '45 n's peeies+e4 += occe a1,w e -lhe Rael Hli+h d ra w a l L l n (4. ) y
>i REFERENCE {
VEGP Procedure 12003-C, p. 2 001000G010 001000G005 ..(KA's) I a ANSWER 7.19 (1.50) (0.5 each)
-1. Mode 1, 2, 3, 4, 5 within i hour after detection of inoperable control rod (and each 12 hours thereafter).
- 2. Prior to initial operation above 5% RTP after each refueling with control bank at the maximum insertion limit.
- 3. Once per 24 hours in Mode 3, 4, 5.
i (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) f ,
m.. , y , .g g q l7." PROCEDURES - NORMAT,. ABNORMAL, EMERGENCY Pro 106 '
"i MD~ RADIOLOGICAL CONTROL a
5 , ,
, REFERENCE ,
' VEGP, ,esson Plan LO-LP-33510-01,.p. 6 ..
- Lesson Objective: LO-LP-33510-01', . #4 -
001000G001". ..(KA's) ,
;. 4 ['
ANSWER- 7.20 '2.00) (I; col
~1; ;[RCP #4 should be. operating [1'.03 ,
- 2. 'Shutden red b::h: cheuld be fullr-withdr== - =ct et tre stere.
.[1.03 /
REFERENCE'- , , VEGP< Procedure 12001-C,' Unit Heasup to. Hot Standby;:pp 1-11
~
001000G010 .1(KA's) - ANSWER 21 (2.00)
.1. Reactor Coola stem Pressure >:.2735' psig [0.5],'
be within. limits in minutes [0.5].
- 2. Tavg'<'551 F [0.53, be within 1 -
-in 15 minutes [0.5]. .
p. REFERENCE Al' VEGP Technical Specification, 2.1.2 and'3.1.1.4 002010G005 ..(KA's) , l ANSWER 7.22 (1.00) -
- 1. Shift Supervicor L s
- 2. Fuel" Handling Engineer 1
- j. ' I l i l
j t (*****: CATEGORY 7. CONTINUED ON NEXT PAGE *****) . l l I i.___._______.__.___._._____.____._._____.__._ __
L__PROCEDUNi - NORMAL'. ABNORMAL EMERGENCY , P2co107 - AND. RADIOLOGICAL CONTROLH c I -
- ' REFERENCE -
VEGP' Procedure 93210-C, p. 3 03400'G010[ 0 ..(KA's)J
~
' ANSWER , '7 :23: (1.00)
- 1. ~ PZR level- .-6-?%- 96 % -
- 2. Subcooling - -07 7-- ; (,0 *F. - j.
REFERENCE , ,
.VEGP: Lesson Plan ~LO-LP-37012-03-C Lesson Objective LO-LP-37012-03-C, #10 Student Handont LO-HO-3701'2-01-C-001,;p. 17 L000055K302 ..(KA's)L ANSWER 7.24' (1.00)~
The RCPs add heatito the'RCS water ~- therefore when the RCPs-are-
~
Ltripped:the remaining water.in the SG is effective'for a longer period. of time. . j [ REFERENCE <l VEGP: Lesson Plan LO-LP-37051-02-C,'p. 6 Lesson Objective LO-LP-37051-02-C; #2 Student Handout LO-HO-37051-00-C-002, Self Test.
.000054K304- ..(KA's) l 1
1 a
> I A
i l (***** END OF. CATEGORY 7.*****) l 1 i A.L..--_-_-__.---_. . . - - - . _ - _ . - . _ _ . . - _ _ - _ - _ _ _ _ - - . . . . _ . _ _ .
m '- '
-m x-s.
~
i8. ADMINISTRATIVE PROCEDURES.2CONDITIONSi , Pcca108?
- AND LIMITATIO E 4 5
' ANSWER: 8.01' (1.00) ,
a
~
K, , s 1 REFERENCE.- , tiVEGP. Procedure 10004-C, p. 2 t; '
"194001K103: ..(KA's). ,
ANSWER- ' 8'. 0 2 '. (1i00) , , d - o- , REFERENCE VEGP Procedure, 10006-C, p. 1_. -
~'
012000G002 ..(KA's)
~
s i ANSWER: 8.03 -(1.00) , f 2 ,
- c. .>. ~ l
. REFERENCE ..
.4 ,
10CFR50.72.
< 1
.J
-l
. ANSWER- 8.04 , (1.00) 'l
-f' l c - !
; r ;
.)' L L
.> D i j
1l ! (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) :l t Lk 4 [ j
.-__ '1_._-__
_ . _ _ _ . _ _ _ _ . _ _ . _._m.
m 4 - a 4
.' 8 .' ADMINISTRATIVE PROCEDURES. CONDITIONS. , Pada109 ' '
AND LIMITATIONS' :) y - s P
' REFERENCE ,
,VEGP Procedure 110018-C, p;-1 194001K102 ..(KA's) ,
f 4 ) ANSWER: 28.05 (1.50)) , , a.- 5. .
- b. . .On' Shift. Operations Supervisor '
- c. .+ _ 2. -p ,
4
' REFERENCE' VEGP. Procedure 10003-C, Manning'the Shifth P.'2 ANSWER 8'~06 (1.00)- ,
c ,
' REFERENCE. .
VEGP Procedure 00054-C,'various pages 194001A102 ..(KA's) ANSWER '8.07. (1.00) 1 (0.5 each) .
- 1. The' person has performed shift functions within.30" day. OR ~
- 2. The person has been briefed in accordance with' Licensed Operator.
- Requalification Program.
REFERENCE - fVEGP Operations ~ Admin; Procedure, 10003-C, p. 2 t f 194001A111 ..(KA's)- .. 4 (, s -u 1 l (***** CATEGORY 8 CONTIN 8ED'ON NEXT PAGE.*****). , l e ; 1:
, . , [- .' 2 -
4 ,
- 8. ADMINISTRATIVE' PROCEDURES. CON)6ITIONS;' . P @ l10-AND LIMITATIONS ,
. ANSWER? 8.08 (1.00)' ,
b REFERENCE l , y VEGP Operation Administrative Procedure, 10001-C,-p. 3- 4 194001A106 ..(KA's)- (2.50). ANSWER. 8. 0 9. ' L . .
- l. (any 5 9 0.5 each) .
Assure safe operation.of the plant under all conditions.
- 1. .
- 2. Protect plant personnel, the health and safety of the public, and plant equipment.
- 3. Conduct major plant. operations in accordance with approved written l -procedures.
- 4. Be attentive to conditions of the plant atiall times.
- 5. Believe and' respond ~to. instrument indications until they'are proven l to.be incorrect.
- 6. Not' bypass,. reset bypasses, defeat safety systems or interlocks'or l.
1 remove' Cat.'1, 2, 3 instrument channels from service unless allowed to do so by approved procedures. REFERENCE VEGP Procedure 10000-C-, p. 14 Peach Bottom Incident 194001K103 ..(KA's)
.i ANSWER 8.10' (1.00) ~
- 1. Shift Support Supervisor- (or osos Jes *3ne . If SSs is ,d3enf *)
;2. On - shift. Operations Supervisor
(* Accep+ "osos des's*3 nes W SSS is *bsend " enQ ln conjunedl~ aHb anwee
*2 for ' 1 full credl+.) /k i 1
I i (***** CATEGORY. 8 CONTINUED ON NEXT PAGE'*****)- ] i
e- -
.e
- . 8.
ADMINISTRATIVE PROCEDURES ~. CONDITIONS. ,, ,Pagi111: c AND Litf]TATIONS :
. REFERENCE ,,
EVEGP Procedure 10006-C, p. 2
, 012000G002' ..(KA's)
ANSWER 8.11 (2.00) , (0.5 each)
- 1. .Perf orm'eval uati on'to determine if the valve's current position has.
resulted in any adverse condition.
- 2. Perform evaluation to determine whether' repositioning 1to correct ' t position will result in any' adverse system' conditions.
- 3. Direct repositioning and locking'of.the valve or place'the system in-a condition where'the valve can.be-repositioned if.not' presently..in' in such a condition.
- 4. 1 Ensure 1 Deficiency Card has been initiated'.
REFERENCE. ' VEGP Procedure 10019-C, p. 3 194001K101 ..(KA's). ANSWER 8.12 (2.00) .
- 1. OSOS
- 2. SS
- 3. - Sr. Vice President Nuclear Operations ( V,ce. Pre 4Jeaf .- Nucleach
^
- 4. . General Manager, Vogtle Nuclear. Operations (Generol . Mao,3,r -
Mca:Je r}
- REFERENCE' VEGP' Procedure 91102-C, p.1 '
l L t-k'
~,
(*****' CATEGORY 8 CONTINUED ON NEXT PAGE *****) s h-.____._____m_ . _ _ _ - - -
- p. ADMINISTRATIVE PROCEDURESs_ CONDITIONS. Page112 AND LIMITATIONS ANSWER 8.13 (2.50)
(any 5 @ 0.5 each) l 1. Emergency classification and E-Plan implementation. l 2. Deciding to notify off-site authorities.
- 3. Protective action = recommendations to off-site authorities and content of messages.
- 4. Authorizing personnel radiation ex>osures in excess of 10CFR20 limits.
- 5. Deciding to evacuate non-essential personnel from the site at the ALERT classification level.
- 6. Deciding to request assistance from federal support groups.
REFERENCE VEGP Procedure 91102-C, p. 2 ANSWER 8.14 (1.50)
- 1. Further degradation in plant conditions.
- 2. Any change from one emergency class to another.
- 3. Termination of an emergency.
REFERENCE VEGP Procedure 91002-C, p. 3 ANSWER 8.15 (2.00)
- 1. Fire Brigade keys - Control Room
- 2. Rounds' keys - Control Room
- 3. Valve Train keys - Control Room Key Box or as specified by OSOS.
- 4. Priority keys OR Emergency keys - Maintained by Security Specialist or SNS l
(***** i CATEGORY 8 CONTINUED ON NEXT PAGE *****)
g_ _ ___ ,
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. PcC2113
%. AND LIMITATIONS e
- REFERENCE VEGP Procedure 00008-C, p. 9 194001K105- ..(KA's)
ANSWER 8.16 (2.00) (0.5 each)
- 1. Change in purpose or scope. .- .
- 2. ' Change in admin procedure or admin control'which reduces control already established.
- 3. Deviation from FSAR or Technical Specifications.
- 4. Acceptance criteria is less conservative than previously established.
REFERENCE VEGP Procedure 00052-C, pp 1 & 2 194001A101 ..(KA's) ANSWER 8.17 (1.00)
- 1. Equipment is still positioned and tagged as indicated.
- 2. Clearance is still valid and required by plant conditions.
REFERENCE VEGP Procedure 00304-C, p. 3 194001K102 ..(KA's) ANSWER 8.18 (1.00)
- 1. Where personnel can receive significant radiation exposure.
- 2. Cases involving containment entry, while containment integrity is established.
1 (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) l
-@. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS Page114 REFERENCE VEGP Procedure 00308-C, p. 3 194001K102 194001K101 ..(KA's) ANSWER 8.19 (2.00) (0.5 each)
- 1. Control of jumpers and lifted wires.
- 2. Final approval of jumpers and lifted wires.
- 3. Determination involvement in safety related system.
- 4. Review of log to ensure proper transfer of operating information from one shift to the next.
REFERENCE VEGP Administrative Procedure, 00306-C, sec. 3.2.1, p. 4 194001A106 ..(KA's) (***** END OF CATEGORY 8 *****) (********** END OF EXAMINATION **********)
g.- - Georg a Power Cornpany Nuciew opea:<;nt Post Oh Box 1600 EECLOSURE 3 Waynenboro, Georg:a 30830 Telephone 404 724 8114 404 654 9161 _ ; g ,, c n i.i , L. U U Vogtle Project , 9 r, Georgia Power p{ . 5p ,.4 s ,, n o:,m , , own
..: r,e.,fiON II M ANTA,GA.
Plant Vogtle - Units 1 & 2 SRO Written Examination Comments December 5, 1988 Log: NOT-01587 Security Code: NC ( United States Nuclear Regulatory Commission j Region II - Suite 2900 l 101 Marietta Street, N. W. Atlanta, Georgia 30323 ATTENTION: Mr. John Munro
Dear Mr. Munro:
Attached are the comments compiled by the Training regarding the SRO Written Examination conducted at Vogtle Electric Generating Plant on December 5, 1988. Your consideration of these comments in grading of the examination is appreciated. Very truly yours,
,.- f1f G. Bockhold, Jr.
General Manager SAB:kss ATTACHMENT xc: W. G. Hairston (w/o attachments) C. K. McCoy (w/o attachments)
QUESTION 5.07 (1.00) Answer the following regarding effective fuel temperature.
- a. How (INCREASE, DECREASE, or REliAINS THE SAME) does the EFFECTIVE fuel temperature change over core life for a given power level?
(0.5)
- b. What TWO; factors cause the change in EFFECTIVE fuel temperature?
(0.5) ANSWER 5.07 (1.00)
- a. DECREASE [0.5J
- b. 1. Clad creep [0.25]
2. Pellet (fuel) swell [0.25] REFERENCE VEGP Lesson Plan LO-LP-33220001, p. 4 LO-LP-33410-02 Lesson Objective LO-LP-33410-02, #10 192004E107 ..(KA's) 4 (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) l
?i; ,
. QUESTION NUMBER: 5.07B
~ STATEMENT: Answer key incomplete per VEGP references a
FACILITY RECOMMENDATION: - Add the following'as an' acceptable answer for Part B: " Fission gas in the gap" .l l
REFERENCES:
LO-LP-33410-00 : Materials II.E.
" Doppler Coefficient: By M. G. Woram
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turns around and after a few hundred full power : hcurs decreases to a small value. See Figure 12.
' . '9 .
- n. . Because the heat transfer coefficient changes, the AT between the fuel and the water must also change.
('"] ' Since the water temperature for a given power levelis l constant over core life, the fuel temperature must change. It first increases and then rums around and over most of core life it gets lower and lower for a given power level. See Figure 13. '- I
-. ,,- ~ ~ , , . . . - ,
THE M15 OFJIEAT TRA>5FER FROM THE, FVE1.15, AFFEC'iT.D BO)'PimilON GAS IM sei 1ME C&P! 2) g[f,3&ggggggL 3) - ' g ( g'A90 4) clap w,~>Acasar" i 1 : Cove tMe -
~
l t$g Etgo
" doppler cosmcientd Rememtsr ^ tidt" thE operator' k rigurr 13: Changes in fuel temperature for a sisen power W " "'* id*-
judges 'the doppler coefficient a function of power level (which he observes) not fuel temperature (which I he cannot obsene). At the begmning of life, fuel tmeperature at say 100% power is relatively high, so As corr life progresses, however, the fuel runs cooler reactivity loss due to doppler is also relatively high. and cooler and you should expect to see an enr decreasing loss of.remedvity per percent change in l power due to doppler and a smaller total reacddty defect for a given power. i [' - 1
,I ' D.C.25 Phttoalum Build Up I ,
i g There is, however, an additional complication that i an et .: p opposes this decrease. This is due to the build up of l [. f Pu". Some of U", after absorbing a neutron, is ultimately transformed to Pu'*. j j j e e / ' L '
)
d
- U* + N - -U* -!
- k. weer y J
#. 23.5m y* s
.l Np" + N-Np 2a
# 2.3d # 65m O cmae car.m . eteu. pen ,en.nn, i
Pu" t N Pu'* 1 1 Fisure 12: Temperature dainbution in a feel pin ai BOL and EOL. 11 3 9 _. . .
~ ' - ' ' ~ ~
1---
-~
.. _M Tr- - - - ~1
QUESTION 5.08 (1.00) For each case below indicate WHICH, item 1 or item 2, would produce the highest reactivity worth.
- a. 1. . Dropping a rod at power.
OR
- 2. A sin'gle rod stuck out following a reactor trip.
~
- b. Total worth of all rods at:
- 1. BOL~
OR
- 2. EOL ANSWER 5.08 (1.00)
- a. 2 :
P. 1 REFERENCE ; l YEGP Lesson Plan LO-LP-33450-01, p. 10 Lesson Objective LO-LP-33450-01, #4 '; 001000K511 ..(KA's) (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) l:
7 ,
.g i
QUESTION NUMBER: '5.08B _[ STATEMENT:. Both answers for Part B are correct... .. FACILITY RECOMMENDATION: Delete Part B'since:.
, . Total Rod Worth BOL IeFREE is more than EOL EQ XeT. ROD. WORTH-Total Rod Worth BOL'EQ.Xe.is less than EOL EQXcT. ROD WORTH REFERENCES. VEGP PTDB: TAB 1.5.1-F1 TAB 1.5.1-F2
. TAB l'.5.2 I
i I i i s i
~,
i 1 i 1 1
)
)
w_---_----a- u- - - - - - - - ---------- - - - - - - - , - - - - - _ - - - - _ - - , - - - - - - _ - - - _ - - - - - - - - - . - - - - . - - - - - - -
. ,,.m -
MEDURE NO.
~
REVISION PAGE NO. PTDB TAB'1.0 1 21 of 27 TAB 1.5.1-F1' JROD WORTH . , , , , . . , iHFP'%.,E,,ti,X., .. i.,.j'[BOLLA MOL b USE FOn BURNUP < 7500 MWD /MTU 2n - - - - a g * [ [ i i t f I I ' q 4 i i , i > t , 1 3 t i + * . 4 e 1 I i 6 e I 5 5 5 . { e i ( . I
% + r , 1 e f f i
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6 X e
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- I .
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- e Mm a' '
a, , O J ' I w i . wi i
.I 'A i N i i I
- u. w e .
x, ne pagefter (sies e Mang sua e Ie I ns ] us sua e 1in L Ias suae I lim Ias . DIFFERENTIAL AND INTEGRAL ROD WORTE VS. STEES WITEDRAWW, M W.S D, C, AND B MOVING WITE 113 STEP OVERLAP, BOL & MOL EFP, EQUILIBRIUM XENON l Q l i l I A A $A Reviewed By
/ Date 9h/n renas
- - m
EDURE NO. REVISION PTDB TAB 1.0 PAGE NO. R 1 23 of 27 TAB 1.5.1-F2 ROD WORTH ge tH n.FPiiEqXe!,NDl.?
~
A10l>, w-aa c USE FOR BURNUP 2,7500 MWD /MTU j A i i! i e ' c, =
- s. ,. .
a ( f W c: : ,'
== ' :, :,
s
,l i- =
t, , . s a x . . . . s,
]== : , : ' ', N.. ' . , ..
1 , s , , . , . : ', = I f I t F 1 , o. . . . . , , N, meo . s . . , . . i f %1 i x . , . a
. 18 4 , '
7 , ig 1" A t , I t I *I f i T I t i __i a 8# ,r, Ti X , 3~
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[s I us - aset c { 1me lles osse I lin ]as 1as l D, C, AND 8 WOVING WITE 113 STEP OVERLAP, XENON NN, BANKS DIFFERE MOL & EOL, EFP, EQUILIBRIUM ! l
)
d D LL ! TNr1 Reviewed By Date , 1
gouRgg, REVISION PTDB TAB 1.0 0 PAGE NO. 27 of 27 [ TAB 1.5.2 ROD WORTH
SUMMARY
sN55 87 0F cos?ROL NOD woaTNS
=
y*W**,,.c,. .
- l. ( B064:. NSF) 1 SOL. NFP i -+
.n. n.. ... #....
500, EPP
... r....
},, ' rf . , .,
- (,.,, set,y esp.'j.,
... .....\
8d 5..k Co.ul.tiv. 5.mt cm.u!.tav. x .. - gg,,, W.rth W.rth W.rts worth S..k cm.utatAw. s.at ..cm.ulatave worth w.rts w rth wort. 0.63 0.63 0.70 8 0.14 0.74 0.14 0.66 0.66 1.23 1.86 1.34 2.04 1.26 2.00 1.18 3.84
,, 0.97 2.83 1.04 3.12 1.18 3.18 1.03 2.87 i.
c...a 0.64 3.47 0.58. 3.4S e. ( i>m@'- *C OM1RWR
,,,,, ou .....,... .,. . 1..
. .. . 4 0 er ... 1 0.r...
.u ... .... ... . ... . .. .. ., . ....... ..-r...
m sacTIVE RDO: COE LOCATION F-10 Wo m 1200 PCM
\,~
f I { 4 J y
, Reviewed By '/Sb/k$f Date
w-- 1 QUESTION' 5.09 (2.00) ! The plant is operating a 30% power when one reactor coolant pump trips due to an electrical fault. The control systems are all in automatic.
- Bank D rods are initially at 165 steps. Assume the plant does'not trip.
Indicate how EACH of the following STEADY STATE parameters would change: INCREASE, DECREASE or REMAIN THE SAME. f.
- a. Turbine power
{
- b. Reactor power (final) !
I
- c. Final rod height
- d. T-ava (affected'ioop)
- e. T-avs (unaffected loop)
- f. Core Delta-T l t
- g. Delta-T (affected loop) l
- h. Delta-T (unaffected loop)
ANSWER 5.09 (2.00) (0.25 each) :
- a. REMAINS THE SAME
- b. REMAINS THE SAME
- c. REMAINS THE SAME
- d. DECREASE
- e. RDiAINS THE SAME l
- f. INCREASE
- g. DECREASE
- h. INCREASE REFERENCE VEGP Lesson Plan LO-LP-34210-00, various pages Lesson Plan LO-LP-34310-00, various pages 4
000017K104 ..(KA's) (***** CATEGORY 5. CONTINUED ON NEXT PAGE *****)
QUESTION NUMBER: 5.09 STATEMENT: Provide allowed assumption of stage pressure feedback g or out. Key answer is for Stage Pressure Feedback in. FACILITY RECOMMENDATION: Accept the following as the answer for Stage Pressure Feedback out:
- a. Turb Power 4 Steam Pressure I , Control Valves don't respond
- b. Reactor PWRt Reactor is tied to turbine load which 4
- c. Rod Height f TREFi due to Turb Ist stage Presst rods move in to lower TAVC.
- d. TAVG affected Loop Y YTAVG = p,H + $
2
- e. TAVG unaffected Loop i TREF h ===>. rods move in to lower TAVG.
m
- f. Core ATi 6 = ($C o Ti
- g. AT affected .
Loop i h4T = hTH -T C
- h. A T unaffected ,
Loop i idT= TH ~ C
REFERENCES:
Lo-IU-60305-00-01 Pages 1-2 & 1-3
- - - - - __ _ a :_ _ - _ _ _ _ _ _ _ - - _ _ _
Georgia Power- b 0 POWER GENERATION DEPARTMENT f't ' y' VOGTLE ELECTRIC GENERATI T g pv
- INSTRUCTIONAL UN ! J . l TITLE: RESPOND TO PARTIAL LOSS OF FLOW NUMBER: 'LO-!U-60305-00-001 PROGRAM: LICENSED OPERATOR TRAINING REVISION: o AUTHOR: U2ues DATE:
APPROVED g DATE: 7,9df[g
REFERENCES:
[ VEBP PROCEDURE 18005-1, REV 2, PARTIAL LOSS'0F FLOW
. r i
^
w " f ^
* ~W M %
- . ~ .
,, w
.-Q ^^
x U-c_ r_ - -
'" tow & 19
- Q
- i
= g-
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-:- =
., -- %T. ' '"
. 3. .- %;
j ,
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_ _ .x om ., . r p
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sTeDeNT e 1E MASTER CO PS .
m- .
.s ".f 'LO-IV-60305-00-001; Respond'to Partial Loss-of Flow.
PERFORMANCE OBJECTI'VE' Given the loss of 'on'e RCP with Dower between F (10 percent) and P-8 (49. h l: . percent), respond to cartial. loss of flow. 'I d Plant conditions must be stabilized -and the pump restarted within 6 hours.
'If tne RCP cannot ce restarted, the'LO must bring:the unit to.not.standoy,
~
witnin s'nours. The steam generator levels andfthe pressurizer pressure anc-level'must be controlled to orevent a' reactor trio. All communication: N and activities must be performed in accordance'with current, acproved_ procedures. 1 i
. INFORMATION . ,
?
1 i The LO will initiate Procedure 18005-1, " Partial Loss off Flow," upon-inidication of loss of flow. The following annunciators can indicate a partial loss of flow: 1 is
-- RCP TRIP
-- RCP LOOP 1(2.3,4) LOW FLOW ALERT ,
- LOOP 1(2,3,4) RTD BYPASS LOW FLOW
-- RCP 1(2,3,4) MTR OVERLOAD'
-- UNDERVOLTAGE RCP BUS ALERT
-- UNDERFRE00ENCY RCP BUS ALERT
-- 13.8 KV SWGR 1NAA(INAB) TROUBLE Tne first symotom of loss of partial flow may be theIRCP. TRIP alarm; the other alarms will be coincident'with the. trip alarmLorfwill. follow ~directly-after it. All of the listed alarms may not b'e present.
A reactor coolant pump trip may be caused by an electrical $ problem with one of the two series RCP breakers or a mechanical problem'with the' pump., A manual pump trip may be necessary if ' there is al problem with lubricating oil or cooling. water to the pump. If power is low (near 10 percent) when 'the RCP is - tripped, 'the ef f ects of Icsing one cump will be small. -If power is near P-B-(49 percent), there- - will be wicer effects in the secondary flows, steam generator-levels, and primary and secondary temperatures. In all likelihood,-with power near P-B (less. tnan or ecual to 49 percent), a plant trip will occur because of the transient's effect on the steam generator water levels. If' power is 'above.
- P-8. partial loss of flow is a DNBR concern and the RPS will. automatically trip the reactor on detection of less tqan 101cercent flow in one loco.
-_ ____--_ A
, w - -, ,
?- "
'.,. LD-IU-60305-00-001; Respond to Partial Los, of Flow WY($$k.WhW Y Y mik
~he,f ollowing discussion assumes that there is no reactor trio.
' When WCKisflosCflow through that 1000 will coast down until flow stoos; this will tare aoout 15 secones. There will be a slight increase in flow tnrough tne other loops oecause tne system flow resistence will De slightly reduced. Af ter the ' flow comoletely stoos, there will be oackflow tnrougn tne Idled cold leg because pressure will be slightly lower in this cold leg than in tne colb legs of the operating loops.
The hot leg temperature in the affected loop will be colder than the cold leg temperature in tnat loco, because some crimary-to-seconcary neat transf er wi,11.. occur. in,the steam generator./ beYEA'TMlilee"negeIl9e' abd willi 'eb o' O lessTmaanitbdpi.than delta T for the other locos. This input should be defeated.so tnat it will not affect control systems. N Delta 7T i Q reacn"coeratino loo ~p 'wi11 ? inhrease because oower and heat transfer will tn ' maintainea at the original value determined oy turoine loading. The average temperature (Tavo)L or theaffected' loco will Ibe-lower /tnan the f cold leg temperatures (Tc) f or tne coerating looos. This input shoulo also
'be def eated, tndIIItIT' avg f or the olant will start to rise f ollowing the loss of flow " rod control will drive roos to keep Tavg at tne procrammed value. Indicated Tavg in the other locos will initially rise because the reduced core flow will cause elevated not leg temperatures in the three coerating loops.
As the RCS flow in the affected loco coasts down, the cooling will cause tne water level in the idle loop's steam generator to snrink. The recuced temperature will result in a reduced steam generator pressure. subsequently reducing steam flow from that steam generator. The Steam Generator Water Level Control (SGWLC) system will temporarily increase the feed flow to i that steam generator because of the cominance of the level error signal in l the control circuit. Feed flow will exceed steam flow until the level is restorea to its setpoint. The result will be that the steam generator will j 4 nave more rnass than when the 1000 was operating, at the same level. The i mass flow rate of steam and f eedwater will ce less than it was originally anc less than that of the other steam generators, j The other steam generators will steam more to maintain the unchanged turbine load. Their levels will increase initially (caused by swell) anc then stabilire. The pressuri:er will respond to the actual Tavg. not the indicated Tavg. The short-term response of the olant to a loss of forced coolant flow in one loco is an initial rise in INDICATED Tavg. This effect is snort-11vec and cecreases as loop flows near their stacie values. Several factors aHect this response. but the initial cominant one is tne increase in temperature (T ) of the coolant exiting the core. This is a result of tne reactor ceing at a given oower (trermali aC anc n cw ceing recucec.
,f,p LO-IU-60~05-00-001: Respond to Partial Loss of Flow Several of the other f actors may be significant. However, the two major factors are:
-- time in core life, and
-- initial power level.
.Early in core liife, the moderator temperature coefficient-(MTC) will nave little eHoct. Late in core lif e, when the MTC is of greater magnituce, the ef f ect will ce a relatively rapid . decrease in neutron power : level caused oy increasing coolant temperature. (in the core) that resulted from the-flow reouction. If the MTC is hign,~ this ef f ect will, when coupled
.witn inward rod-motion,'act to furtner decrease core power.
If the power level is law (less than 15 percent), the magnitude of_the transient will ce greatly reduced. In aodition, ' the control rods should 'be in manual. If the Rod Control System is in automatic, the rods wil1~ step in (as a result of .the short-term increase in indicated Tavg) and'then step back out to approximately their original cosition. The resconse of the : pressurizer level to this transient-is greatly.af f ected. The pressurizer level is a' response to the actual plant Tavg. not the Tavg-temperature indicated by instrumentation. As stated earlier -(again, depending upon initial plant conditions), this indicated Tavg may be in direct disagreement with tne actual plant Tavg. This can (for' a short oeriod of time) result in a situation where inoicated plant Tavg is . increasing and the pressuri:er level is decreasing. All of these effects are transitory in nature and should dissipate rapidly. From steady-state to steacy-state condition, theM' ant' shou d st'abi'lfzel withiih'e' cimtrolcrodsD
,attapproximatelW the'= hafdht; the$urbineVl'oad'the"s5she (gc vernor! '
' valves slightlylfurthetppe_nl, and sthm 'oressure reduced. This discussion assumes no operator action.
Plant Vogtle Procedure 18005-1 is used for a loss of forced RCS flow in one loco during powec operation below the P-8 (49 percent) setooint. If one
]
RCP is lost with power above the P-8 setcoint, the reactor will trip. The reactor will also trip if more than one loop is lost with power.above the ; F-7 (10 percent) setpoint. If power is.celow the P-B setpoint, out above i the P-7 (10 percent) setpoint, the reactor will not trip if only one RCP is ! lost; Technical Specifications allow uo to 6 hours fro the plant to be I olated in hot standby. Depenoing upon the oower level. the steam cenerators may exhibit erratic level and flow responses. Manual control may be used if setpoints are l I cnallenged while the automatic control systems are stabili:ing the RCS ! temperature. 1 RESPOND TO PARTIAL LOSS OF FLOW i l 1-3
i I i QUESTION 5.11 (1.00) What are the TWO principal factors affecting the-redistribution { reactivity effect in the reactor core?
]
1 ANSWER' 5.11- (1.00)
- 1. Moderator reactivity
- 2. Fuel burnup REFERENCE .
VEGP Lesson Plan, LO-LP-33420-02, pp 7a-b Lesson Objectives, LO-LP-33420-02, #13 001000K529 ..(KA's) 1 l l j i (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) < i
QUESTION NUMBER: 5.11 STATEMENT: Credit for candidates understanding of the concept of redistribution. FACILITY RECOMMENDATION: Accept the following as additonal factors affecting~ redistribution:
- 1. Axial Flux Profile across the core
- 2. Temperature Change across the. core
- 3. Moderator Density Change across the core
- 4. MTC Change across-the core
REFERENCES:
LO-LP-33420-00 III.5 a-c l d
D G ill'
. c lEt3ON 00171NB:
N01BS 9 h LXI M9" }/9 " v- / gem 7 y'7g</"f ',t n-B n,)t 0.x ov rp1 v rifr -y a gem v rF y*1 iJ ;
> lyp-y y?n y g> Q 9% = < % 1 9 ( 7v 1
77 4 &3 ? 1O*11 qq ptc>O 04 7" t' '7 e.
" = WW vT i
f" MW I 3' 4TWMp:4 07"tT ; 0~'191 a -a . o M 5 l.f [, T/J M ,", . 3,?W[l57
/y M s* g " ",
~J;jf. f(f WW '
f5 '
} I
> r y a -y= .r~y vy.v 7e 3yrf m ,.s i
,nysme Q"YTI%
yypp y yupy y ~ v>y g y.]y . I ,o.g.gurc-oc.ous
-yy y~> y y; - y. ye7-f
. -n..y.as.ooo n m 3 wg oey m.a.ume v e lET!xupaM&Ma}5&Ll93m ,m h-a t40'W 7-yr n n p y n ,m T e:t7 v r 9 y
- m.w,no .w -o,6 7 ~ mi ' 9 l' T T T P T u- nn :ps aw- -
n' 'f M D"Y Tl" 5 G ,
'f" y M M, f %o-6-tthte cc ow
#' fi f TfD }9.Ld*'Q bt C '
y rg- ws m,s . ~,~ p.t- o m.semoo* w 5~=rP yysysor' f ,. ..u , w . m .o a v y .i. minr,.,;r-qc,sn +7:r ~ g~ - (
~ -
- sse n m e,
,~.v t~r~ r,.wym 77 4 A*' ""
i
QUESTION 5.12 (1.50) What FIVE reactivity groups (factors) are considered when calculating Shutdown Margin? ANSWER 5.12-(1.50) (0.3 each) 1.
- 2. Boron concentration Control rod position
- 3. RCS Tave 4.
- 5. Fuel burnup (core age)
Xenon concentration REFERENCE VEGP Lesson Plan, LO-LP-33510-01, 6p. Lesson Objective, LO-LP-33510-01, #5 192002K110
..(KA's) e
(***** CATEGORY .5 CONTINUED ON NEXT PAGE *****)
-QUESTION NUMBER: 5.12 STATEMENT: Answer key incomplete per VEGP' References FACILITY RECOMMENDATION: l Add: " Samarium Concentration", as an acceptable answer.
REFERENCES:
Tech Spec. -3/4.1.1.2 Surveillance Requirements 4.1.1;2.b 1)-6) attached LO-LP-33510-01 page 6 . Procedure 14005-1 Shutdown Margin Calculation l} I
lit. LESSON OUTLINE: NOTES
- 3) RIL ensures the SDM is maintained.
- 2. SDM Calculation as PNG Reactivity balance - this data y considered when calculating a SDM.
3)*x f
- 1) Boron concentration hC. ** ?f j g g, y s
- 2) Control rod position -
MWh Ql;[Q'sidIffN,
- 3) RCS Tave ' Qjf
- 4) Fuel burnup .. kk
- 5) Ie concentration : 8 '
- b. *,%a)s ;+r. w y concentrationisnotusedbecausethe-[
m ,1 y a, w . of'*S)6.is offset by the +pj%dded from the f Wfg4 p buildup of Pu 239. ' pMet
- c. A SDM is calculated when:
p00 % y%-q'% (
- 1) In modes 1,2,3,4 or 5 within I hour after detection of an inoperable control .
ro'd and each 12 hrs thereafter.
- 2) Prior to initial operation 5% RTP after each fuel load.
- 3) When in modes 3,4,5 one per 24 hrs
- 4) Pick the most appropriate fort of SDM to do for the given plant conditions.
- d. Sample Calculations:
uk: M+h m &[A6us. Sh;Js k fua H cos-t . W w+
'b c sM. > {td> h4 ;' ,
yk e dha 4 e <h md6 m 4 {/ Mauas cl k Fet+t )j' gggy . L<>f/-33samp
& Cuasd $'
h kk ?wc - WS 47% (c< - M /f" d gi hsh- . Mi s4 kl D h pg k& nk twHcm si mi Als 6 I n u,- - sac wo
. r, O atac11v1Tv coatao' svsTras SHUTD0WN MARGIN - MODES 3. 4 AND 5' LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to the limits shown in Figures 3.1-1,and 3.1-2.
APPLICA81LITY: MODE 3, 4 AND 5. ACTION: With the SHUTDOWN MARGIN less than the required vslue, immediately initiate and. continue boration at greater than or. equal to 30 gpa of a solution containing greater than or equal-to 7000 ppe boron or equivalent until the required SHUTDOWN MARGIN is restored.
^
SURVEILLANCE REQUIREMENTS' 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to the required value:
- a. Within I hour after detection of an inoperable control rod (s) and at least once per 12 hours thereafter while the rod (s) is inoperable.
If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable - control rod (s); and
- b. At least once per 24 hours by consideration _of therfollowing factors:
, 1) Reactor Coolant System boron concentration, - l 2) Control rod position,
- 3) Reactor Coolant System average temperature,
; 4) Fuel burnup based on gross thermal energy generation,
- 5) Xenon concentration, and
$6)[? Samarium! concentration.: }
O 3/4 l-3
_ _ _ _ = - . - - .-
--R ,,. Approva!
Pr C; dure No. Vogtle Electric Generating Plant NUCLEAR OPERATIONS 14005-1
'I -
Revision No. D.to
- /oh/g unit 1 '
GeorgiaPbwer e...~Iof8 - l I SHUTDOWN MARGIN CALCULATIONS ~ ' i d]. g 1.0 PURPOSE 1.1 The purpose of this procedure is to provide methods for ' determining SHUTDOWN MARGIN to verify it is greater than or the limits 4.1.1.2 of Technical Specifications 4.1.1.1.1 as applicable. 1.2 Verification of' SHUTDOWN MARGIN is required:
- a. When in Modes 1, 2, 3, 4 or 5 within one hour after detection of en inoperable Control Rod and each 12 hours thereafter while the Control Rod is inoperable.
b. Prior to initial operation above 5% Rated Thermal Power after each fuel loading. c. When hours. in Modes 3, 4, or 5 at least once per 24 1.3 This procedure also provides instructions for determining the Keff of the reactor in its present state or predicting its value at future plant conditions. 2.0 APPLICABILITY This 3, 4, procedure 5 and 6. is applicable in Operational. Modes 1, 2, 3.0 PRECAUTIONS AND LIMITATIONS 3.1 Xenon Worth provided by reference curves assumes equilibrium conditions prior to shutdown. If plant was in a transient condition, obtain Xenon Worth from plant computer or calculate Xenon Power using the form l provided by PTDB TAB 1.4.3. 3.2 If boration or dilution evolutions are or have been occurring RCS boron, samples are used for chemical analysis. care 3.3 JigI&s gnegativ($QgrakerEforiSasarlucQreac,t$pissihce thA eJeactiViffXifdF6BdEdd:bf"$amarium'149"bsildupy
;is" offset;byLan .equaLpositive ths*bdilduptof Plutonium :239.;
reactivitynintroddderbit
" *^~
,,e....
I h h ' QUESTION 5,13 (2.00) i Answer the following elated to Doppler-only Power coefficient. i
- a. How does Doppler-only Power Coefficient change as reactor power is increased (MORE NEGAT7.VE, LESS NEGATIVE, NO CHANGE)? (0.5) i
- b. Besides reactor power, WHAT THREE factors affect Doppler-only Power '
Coefficient? (1.5) l l ANSWER 5.13 (2.00)
- a. LESS NEGATIVE [0.5]
- b. 1. Coolant Tave [0.5]
- 2. Core age [0.5] )
- 3. Clad creep [0.5] ]
1 REFERENCE VEGP Lesson Plan LO-LP-33410-02, pp 9-10 1 Lesson Objective LO-LP-33410-02, #9 & 12 { 1 192004K107 192004K113 ..(KA's) i l (***** CATEGORY & CONTINUED ON NEXT PAGE *****)
QUESTION NUMBER: 5.13b' STATEMENT: Answer key incomplete per VEGP references. FACILITY RECOMMENDATION: Add'the following as acceptable answer for Part b:-
'?
- 1. Increase in fission gas'.in gap or change-in Helium Cap Thermal Conductivity. '
- 2. Pellet Shrink g Fuel Densification
- 3. Pellet Swell-
- 4. -Depletion of U-238 and buildup of Pu-240 f a
REFERENCES:
Westinghouse Reactor Core Control for Large PWRs. I Page 2-40 referenced in LO-LP-33410-00. 1 1 q l i
,1 i
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hi temperature increases, the amount of r of change in effective fuel temperature decreactivity ins eases. g\ Roooler-0niv Power Coefficient ect and Powe The poooler-only power effective fuel temperature ascoefficient is a combination g a function of power g temperatures (Doppler) coefficient. and of g ge ocm dP gg I5 Power) " dT Ipg}
- dl 'T
'F Doppler Only dP I % Power)
Power Coefficient Doppler or fuel Temp. Coefficient Variation in . Effective fuel Temperature with Reactor Power As shown in figure FND-RF-224, as react or power increases, tk effective fuel temperature increases and thus th coefficient j only power is defined a function of reactor powere Doppler-onl coefficient is . The Dopplet-always negative and its magnitude decreases as power level and effectiv Typical first cycle e fuel temperature increast. Doppler-only power d -15 pcm/% power and -11 pcm/% powerare coefficients at betweet BOL 9 . The Doooler-oniv Dower defect added The defectto is the core due to the Doppler effectis the net nega zero pcm at as power is increased. j approximately -1200 pcm at zero percent of full power and it Doppler-only power defect for BOL is100 percent power. of full A typical ->' given in Figure FND-RF-236. W 4) h riations in Dopoler Power Coefficient Over C ore life , l. powef defect change duent andlDoppler-only _ nomena: LtoJfdurjheJ0ve
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k N 1[h 5 N IEdh d M G MkaiMNM) , 5YA sils k . k 0465C 2-40 i
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QUESTION 5.17 (1.50) What THREE parameters in a fluid system determines the AVAILABLE NPSH7' ANSWER 5.17 , (1.50) (0.5 each)
- 1. Fluid pressureL
-2. Fluid temperature ,
- 3. Fluid velocity REFERENCE VEGP Lesson Plan LO-LP-34410-00-0, p. 22 Lesson Objective LO-LP-34410-00-C, #10 191004K106 ..(KA's)
(***** CATEGORY & CONTINUED ON NEXT PAGE *****)
. QUESTION NUMBER: 5.17
' STATEMENT: Answer key is incomplete per VEGP references.
FACILITY RECOMMENDATION: Accept the following as answers:
- 1. System flow; flovrete'= area x velocity
- 2. Static head due to elevation difference
REFERENCES:
k' westinghouse Thermal-Hydraulic Principles and' Applications to PWR II Page^10-55 l a 1 _ - - _ _ - _ _ - _ _ - - - _ _ . _ . -- -~
I I I.'
/
A E
"x <
s i r n
- VOLUMETRIC FLOW RATE i
l FIGURE FND-FF-47: PARALLEL OPERATON OF CENTRIFUGAL AND POSITIVE DISPLACEMENT PUMP (REV. 2) Local shock. waves caused by the rapid bubble collapse cause excessive noise and vibration within the pump. Cavitation leads to irreversible energy losses causing a large decrease in pump efficiency and a reduction in the flow rate maintained by the pump.
) Cavitation is also undesirable because of the mechanical pump damage caused by the erosion, pitting, and vibration.
t The point in the pump where cavitation is most likely to begin is the low-pressure suction area. In order to prevent pump cavitation, vapor bubbles can not be permitted to form. In order to prevent bubbles f rom being formed, the total head (pressure) at any point in the pump must be greater than the temperature-dependent ' saturation head (pressure, Psat) at that point. The difference between the total head and saturation head is called the available net positive suction head, NPSH,y,g). This term is a measure of is the head available to prevent cavitation. If NPSH If NPSH is negative, positive, cavitation will not occur. avail -, cavitation will occur. The^totalEhead=is"thelsumofthKfluikhead,[
~
-(velocitylhead " and' elevation 1 tie'adNMathematically,theavailablenet positive suction head is expressed as V t l Cxfy~ ~
?, ::" -
10-54 W-0502F 1 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ J
l-N av 11j= Hfluid hevet [" velocity " sat Since head is related to pressure and pressure is more often convenient, NPSH is usually expressed es k D d 5M MId b fdi/ N M $ f k k d Q $ [ d d : Since the pressure due to elevation contributes totally to the J static pressure of the fluid, the. static pressure term is expressed as ' P =P +P 'j static fluid elevation j 1 The dynamic pressure head (i.e., velocity head) is .directly proportional to the kinetic energy of the fluid and is expressed as i
'h 2
P velocity = Pdynamic " a 2gv g g' Substituting into the equation for available net positive suction head yields I il 1 a p NPSHavail = Pstatic -P sat + 2 g ' g The above equation implies that the available NPSH 17 dependent on g static fluid pressure, fluid temperature, and fluid velocity. As p! i fluid velocity through a system increases, the system head loss of i the system increases. Consequently, for higher system velocities, the static pressures at the pump decrease. The overall result of a il velocity increase, assuming a constant temperature, is a decrease in i the available net positive suction head. Figure FND-FF-83 is a graphical illustration of the relationship between NPSH, ,g) and volumetric flow rate. 10-55 w 0502F _ _ - - - - - - - - i'
QUESTION 5.19 (1.50)
# hat are THREE consequences of cavitation on a centrifugal pump?
THESE ARE NOT SYSTEM WIDE EFFECTS. ANSWER 5.19 - (1.50) (0.5 each)
- 1. Erosion and pitting of pump surfaces.
- 2. Excessive noise and vibration.
- 3. Energy losses and lower pump efficiency.
REFERENCE VEGP Lesson Plan LO-LP-34410-00-C, p. 22 Lesson Objective LO-LP-34410-00-C. #10 191004K101 ..(KA's) l (***** END OF CATEGORY 5'*****) )
QUESTION NUMBER: 5.19 STATEMENT: Answer key is incomplete'per VEGP references.
' FACILITY RECOMMENDATION:
' Accept the following as consequences of' cavitation on a centrifugal pump (any three):
erosion fpitting excessive noise vibration energy loss decrease in pump efficiency reduction in flow rate
REFERENCES:
Westinghouse ' Thermal-Hydraulic Principico and Applications to PWR 11 pages 10-53 &'10-54 Referenced in LO-LP-34410-02-C (attached).' t ( 6 1 1 1 I o g..
, 4-
\ [
l As contrasted with a centrifugal pump which is subject to frictional and turbulent energy losses in the fluid, the positive displacement pump is generally more efficient. Positive low eed displacement pumps are used when high pump heads are required, when ump low and accurate flow rates are required, or for various viscous fluids. In the PWR, centrifugal pumps are most commonly used to pump low-viscosity water. Positive displacement pumps are normally 1a < used for pumping lubricating oils or in high pressure water l cts applications (e.g., charging pumps for RCS makeup water). In the lure PWR, centrifugal pumps are either turbine-driven or motor driven. lter The PWR turbine-driven centrifugal pumps are usually variable speed. The flow rate from a variable speed centrifugal pump is controlled by discharge throttle valves or pump speed. Motor driven centrifugal pumps in the PWR are normally single speed and their flow rate is controlled using discharge throttle valves. Positive displacement pumps are normally motor driven in the PWR. The flow rate they produce is controlled by pump speed. Occasionally, a centrifugal pump is used in parallel with a positve displacement pump. In this case, for pressures less than the shut-off head of the centrifugal pump, the flow rate is the sum of the flows for the two pumps. As illustrated in Figu're FND-FF-47, for pressures above the shut-off head of the centrifugal pump, the total flow rate is due to the positive displacement pump alone. For this reason, most PWR systems do not use positive displacement and centrifugal pumps in parallel. Often, operational procedures are particularly written to prevent this parallel alignment. Ol PUMP CAVITATION AND NET POSITIVE SUCTION HEAD
~
i tation :is defined as the formation and subsequent collapse Cav' if I of vapor bubbles formed in a pump. Vapor bubbles form where the by . local pressure has . dropped below the vapor pressure of the fluid or (Psat). As the bubbles move into regions of higher pressure, they collapse causing e'rosion > and pitting < of adjacent pump surfaces., I 10-53 mx g
S b I ;
} ;
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f' ,d. VOLUMETRIC Flow RATE ff f- :. t FIGURE FND-FF-47: PARALLEL OPERATON OF CENTRIFUGAL AND [ ' POSITIVE DISPLACEMENT PUMP (REV. 2) O S o.
! h[ Local shock waves caused by the rapid bubble collapse cause j ;. 7 excessive ~ noise and vibration; within the pump. Cavitation leads to
[ [ irreversible CeneigyTT1osses _ causing a large decreaset in -pump pp E 'efficiiencyjand a redustion inD E the flow ratk_ maintained .by the pump. I bh Cavitation is also undesirable because of the mechanical pump damage-l I g caused by thej< erosion, pitding.~ and Wit ration'. l . 1
]C.
1 [ [ The point in the pump where cavitation is most likely to begin I is the low-pressure suction area. In order to prevent pump
!f ifl i cavitation, vapor bubbles can not be permitted to form. In order to h[. prevent bubbles f rom being formed, the total head '(pressure) at any
.] l point in the pump must be greater than the temperature-dependent saturation head (pressure, Psat) at that ' point. The difference J .
hetween the total head and saturation head is called the available fgfhi; net Dositive suction head, NPSH,yg). This term is a measure of the head available to prevent cavitation. If NPSH,y,g) is q;Q positive, cavitation will not occur. If NPSH,yg) is'. negative, b pl] cavitation will occur. The total head is the sum of the fluid head, f .; velocity head and elevation head. Mathematically, the available net l positive suction head is expressed as l
!{
-(
4 10-54 0369C-
e Georglalbwer Powe6M Ge#44 AAle0N D8PaA8&stMt VOOTLE ELECilllC GEllEllATit4G PLAf1T ga TMINiin IESSfH PIM g FLUID MECHANICS filNBEft: 58'M * * >' '>~ "' ~ "- Pit 06ftAti: LICENSED OPERATOR TRAIN!ns 2 - REVIStutt Atillidit L.' FIT 2 WATER
, , DATE: 10/13/88 APPil0VEllgd%Mg DAIE: g/yff IllSilllf.lfR Glliff IIES '
!. LESSON FORMAT A. Lecture with Visual Aids
!!. MATERIALS A. Overhead Projector B. Transparencies C. White Board with Markers
!!!. EVALUATION A. Written or Gral Exas in conjunction with other Lesson Plans IV. REMARKS b
l
- s s % ..
' jP" ~ .
LO-LP-34410-02-C
REFERENCES:
- 1. " HEAT TRANSFER, THERMODYNAMICS, FLUID FLOW FUNDAMENTALS," 6ENERAL
, PHYSICS CORPORATION
- 2. 4 E'
WESTINGHOUSE. THERM 0 HYDRAULICS; TEXT, f-
- 3. HAND 00T -
LO-HO-34410-001 FLUID MECHANICS
- 4. TRANSPARENCIES:
LO-TP-34410-001 OBJECTIVES (D-TP-34410-002 VISCOSITY OF WATER
'LO-TP-34410-003 LAMINAR FLOW LO-TP-34410-004 TURBULENT FLOW LO-TP-34410-005 REYNOLD*S NUMBER LO-TP-34410-006 REYNOLD'S NUMBER VARIATIONS LO-TP-34410-007 FRICTION FACTOR LO-TP-34410-008 SYSTEM HEAD LOSS CURVE LO-TP-34410-009 SYSTEM HEAD LOSS CHANGES LO-TP-34410-010 SERN00LL1'S EQUATION LO-TP-34410-011 REARRANSED BERNOULLI'S EQUATION LO-TP-34410-012 EXAMPLE ,
LO-TP-34410-013 MODIFIED BERNOULLI*S EQUATION
- LO-TP-34410-014 REARRANGED MODIFIED BERNOULLI*S EQUATION LO-TP-34410-015 CLOSED SYSTEM LO-TP-34410-016 OPEN SYSTEM LO-TP-34410-017 CENTRIFUGAL PUMP LO-TP-34410-018 CENTRIFUGAL PUMP CURVE
' LO-TP-34410-019 Hp, HEAD AND EFFICIENCY " CURVES" LO-TP-34410-020 VARIABLE SPEED PUMP LO-TP-34410-021, PUMP LAWS
- LO-TP-34410-022 SERIES PUMPS 1.0-TP-34410-023 PARALLEL PUMPS LO-TP-34410-024 POSITIVE DISPLACEMENT PUMP
- LO-TP-34410-025 POSITIVE DISPLACEMENT PUMP CURVE LO-TP-34410-026 SYSTEM CURVE AND POSITIVE DISPLACEMENT PUMP CURVE LO-TP-34410-027 CENTRIFUGAL AND POSITIVE DISPLACEMENT PUMPS IN PARALLEL LO-TP-34410-02B NPSH -
LO-TP-34410-029 NPSH PROBLEM LO-TP-34410-030 NPSH PROBLEM LO-TP-34410-031 CALCULATION LO-TP-34410-032 PUMP EFFICIENCY = LO-TP-34410-033 WATERHAMMER - PIPE NOT FULL LO-TP-34410-034 WATERHAMMER - VALVE CLOSES i - LO-TP-34410-035 WATERHAMMER - CHECK VALVE l LO-TP-34410-036 WATERHAMMER - PIPE BEND LO-TP-34410-037 WATERHAMMER - FEEDRINS - 1 LO-TP-34410-038 WATERHAMMER - FEEDRING - 2 LO-TP-34410-039 WATERHAMMER - FEEDRING - 3 LO-TP-34410-040 WATERHAMMER - FEEDWATER HEATER - 1 LG-TP-34410-041 WATERHAMMER - FEEDWATER HEATER - 2 4
l
- l
'J I
QUESTION 6.03 (1.00) f Which one of the following is CORRECT regarding operation of the Reactor Coolant Pump seal system? q
- a. Seal injection water temperature may be increased to 150 F if RCS temperature is limited to less than 400 F.
i
- b. The RCP should not be started if #1 seal leakoff is greater '
than 0.3 gym and less than 2.5 gym.
- c. 325 psig is the lowest pressure at which the RCP can be operated ,
and still maintain the required 100 paid across the #1 seal. {
.d. #3 seal differential pressure is maintained by the #3 injection I head tank and tne atmosphere.
ANSWER 6.03 (1.00) a ! REFERENCE - VEGP Training Text, Vol. 5, Chap. la, RCS, pp la-23 through la-27 003000K602 003000K404 003000K407 ..(KA's) I l l j i (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
i 1 QUESTION NUMBER: 6.03 i STATEMENT: ;The question asks whichtstatement is correct' ; regarding the.RCP seal system. . Answer (d)'is also j correct since # 3 seal differential pressure is 1
. maintained by the standpipe'and atmosphere. #' 3 !
seal is a double dam seal. The stand pipe (injection head tank) provides the pressure to-the inside of, the seal. .The water is drained to the containment-sump. '] - FACILITY RECOMMENDATION: Accept a and d.as correct.
REFERENCES:
P&ID IX4DB114 and student handout sketch' i
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.--" _ _ ~ . - - - ~ ~~ _ _ _ _ _ _ - _ _ -. - - - - - - [] __ -. . ^
QUESTION: 6.09 (2.00) List FOUR systems-or instruments which are supplied with an.'auctioneered HIGH Tavg signal. ANSWER 6.09 ' (2.00) (0.5 each)
- 1. Rod control
- 2. Steam dumps
- 3. Pressurizer level 4.
Control rod insertion limits j REFERENCE VEGP Training Text, Vol. 5, Chap.~1.a., p. la-19 ! 002000A109 016000K403 ..(KA's) i I i (***** CATEGORY S CONTINUED ON NEXT PAGE *****)
n_ - . i
'l 1
j QUESTION NUMBER: 6.09 , STATEMENT: The question asks 'for ~ systems or instruments whiche . are supplied with auctioneered high TAVG. FACILITY RECOMMENDATION: The following should be included'as correct in , addition to the answers cn1 the key. -
- 1. High T avg Annunciator 1
- 2. T,y -T rec rder ref
- 3. T- !
avg -Tref deviation annunciator
REFERENCES:
LO-H0-16101-00-01.
}
1 1 i l i l-
'l J
j l
e c, . $ 0-(10 4 foi D g ;
;) ,
~
. . , . 1.. .j ,
C
.' : High' r f- l TvG ,
(l Rod Rod ~- I~ Reactivity .,- Control B/S Computer - Auctioneef I"hNIt n .. l L. ' ' . -j l l ~' High1TAK l ;
),
.. , i Steam ,
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- a. AUCTIO!1EERED HIG11 T avg s
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Auctioneer Lag B/S - Low TAVG . u C - 16 B/S - ( 4 l ! L
- b. AUCTIONEERED LOW T avg Figure 27-7 T,yg SYSTEM: CONTROL -
04200:4 PAGE 11 0F 12
QUESTION 6.10 (2.00) List the TWO interlocks and permissives associated with the Steam Dumps including setpoints, coincidence and the reason for the interlock or permissive. 1 1 6
. ANSWER 6.10 (2.00) 3
- 1. C-9 Condenser ava ble [0.5) -- satisfied when vacuum.> 17" on 2/2 pts and 1 3 Cire pumps running [0.25]. Ensures condenser.
is capable of ndensing steam so no overpressurization occurs [0.25].
- 2. P-12 Low-Low Tava SI [0.5] -- 2/4 loop Tava < 540',F [0.25).
Preventsinadvertentorexcessivecooldow(n-ofRCS 3 1 REFERENCE VEGP Training Text, Vol. 7, Chap. 12b, p. 12b-14 & 12b-15 041020K409 ..(KA's) l (***** CATEGORY ti CONTINUED ON NEXT PAGE *****)
QUESTION: 6.10 STATEMENT: Several numbers in the~ answer key are' incorrect and there are three (3) additional correct answers. F.ACILITY RECOMMENDATION:
.n. . Change coincidence for CIRC pumps to h in answer 1.
'b. Change LO-LO T,y setpoint to 550*F in answer 2.,
- c. Add the following 3 answers:
- 1. P-4 reactor trip - train A arms steam dumps; RTA & BYA open-
- 11. P-4 reactor trip - train B activates trip controller; RTB &.BYB open 111. C-7 Load Rejection - arms steam dumps; 10% load reduction
REFERENCES:
Vogtle Text. Chapter 8a-Logic Diagrams i 4 I L L I
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REACTOR PROTECTION SYSTEM power. Functions-
'1. Block gutg rod withdrawal.
C-7A SUDDEN LOSS OF TURBINE LOAD-i Source - Turbine Impulse Pressure Channel PT-506 Coincidence - 1/1 pressure bistables 10 percent Dec.. 120 sec. .. Setpoint - Reduction of 10 percent full load with ~ a 120:second impulse time constant. Function:
- 1. Arms the steam dump valves. i
- 2. Manual reset by momentarily placing the steam dump mode selector switch to the reset position.
_ -,-,r,rg . [C-9 CONDENBER AVAILABLE ,a
. J Source - Condenser Pressure Channels and Cire. Pump Breaker Position. -i
- Coi'ncidencW{a2/2; condenser press,UreMa$nels%d'1/2f aam pumpibreakers;
~~w
- 7f-u, , .,q.gg , ,
n,.. ,
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- 1. Condenser pressure channel at 17 inches of mercury gag
- 2. Cire. pump breakers closed, q Function:
- 1. Provides a permissive to allow steam dump operation.
C-11 BANK D ROD STOP Source - Pulse-to-Analog Converter for Control Bank D. j Coincidence - 2/2 Bank D Rod Height bistables j Setpoint - Demanded bank position of 223 steps. Function l l
- 1. Blocks agig rod withdrawal.
C-16 STOP TURBINE LOADING Source - Auctioneered Low Tavg-Tref deviation or auctioneered low Tavg ) low. ; Coincidence - 1/1 Tavg-Tref deviation or 1/1 Auct. Low Tavg LO. - Setpoint--'Tavg-Tref Deviation of -20 degrees F (Tavg Low) et Auct. low Tavg of 553 degrees F. Functions
- 1. Stop turbine loading.
i 64-42 Revision 1 1
1
.. . , j REACTOR PROTECTION SYSTEM
'i.
- Two switches,. one for Train [A and one for Train B,: are provided for the purpose of. blocking the low setpoint power range trip. In. a f ashion '
s similar to the sour::e and intermediate ranges, both switches must be 1 placed in BLOCK to inhibit the trip..
- Four~ protection-bistable ststus lights are.on TSLB-4 (P-7/P-10).-
NUCLEAR: A,T POWER PERMISSIVE P-10-light is on the bypass and'permissives - a status' light box. '1
?
1 i
- 7. P-11'. Permissive ;)
Permissive P-11 allows. manual block of;1ow' pressurizer safety injection ~ actuation. : This permits normal plant cooldown and depressurization. When 2/3 pressurizer pressure' bistable: sense less than 1970.-psig,-thel low pressurizer pressure safety. injection signal may be blocked. Placing l
.the Train A and Train:B pressurizer safety injection block switches to 1 BLOCK will 'now prevent low pressurizer. pressure (1870 psig) from' j initiating. safety injection. Each switch will initiate the block j function'in its respective protective train..If pressure' rises.above thel P-11 setpoint on 2/3 channels, the-block automatically clears..If 2/3
~
channels exceed the P-11 setpoint, any shut ECCS accumulator isolation 1 ] ' valves will automatically open if'handswitch is in AUTO. knen 2/3 pressurizer bistable sense:1ess than the set' pressure, the low I steam line pressure SI and MSLI may be blocked (The blocking action.will ] enable the Hi Steam Line Pressure Negative Rate MSLI)'. When pressure , rises above setpoint, the~ steam-line low pressure SI and MSLI'will be automatically unblocked -(enabled). Additionally,'the Hi Steam Line' Pressure Negative Rate MSLI will be automatically blocked. ' The three protection bistable status lights are'on TSLB-3. ; a
- 8. W' Mmi.%
%kiW46,Wcu,
, ai2?PeralissiWTM .
to This permissive undesired automatically [ blocks, Steam, Dump, System operationI !
.4 prevent cooldown.*I7 TapinstrumentsibeloeIthDoblo skjoWiS50'dsipsesMaclunEftselP;11. When ihe tabperatuhe f alls M below "t'he' sethoint,"all open'sthidU$iEalves will . shut,'. and -all' !
- further operation will'be blocked. To allow for plant cocidown less:than {
P-12, the operator. places the Train ~ A and Train B ' steam' dump' interlock ! iswitches to BYPASS INTERLOCK. With temperature less:thanithe P-12 setpoint' (550 degrees F)-'and high steam flow, . main steam line isolation { will automatically initiate. ' The P-12' interlock allows.the steamline. I pressure safety. injection and steamline' isolation signal.to be blocked.. l This allows'a normal plant cooldown ta be performed. The three protection bistable status . lights are on TSLB-3. There is a : ! common alarm, REACTOR COOLANT LOOP 1A, IB, 1C OR 1D TAVG LO-LO.-'There js j 3also a LO-LO TAVG SAFETY INJ STEAM DUMP.INTLK.' The P-12 light-is on'the bypass'and permissives light box, which energizes when 2/4. loop Tavg. i decrease below 550 degrees F.- i Ba'-59 j Revision 1~ l
QUESTION 6.11 (2.00) a List the FOUR signals which will automatically-close.the Steam Generator Blowdown containment isolation valves. ANSWER 6.11 ' (2.00) (0.5 each) . .
- 1. Low-Low Steam Generator level on any Steam. Generator.
- 2. SI
- 3. Loss of off-site power.
- 4. Trip of both Main Feed pumps.
REFERENCE VEGP Training Text, Vol. 7, Chap. 13.c., p. 13c-5 035010E403 ..(KA's) k 1 l i l 1 (***** CATEGORY f CONTINUED ON NEXT PAGE *****)
QUESTION NUMBER:- _6.11 STATDIENT: A recent design. change added another' automatic closure signal to the.SGBD isolation valves. FACILITY RECOMMENDATION: Add AMSAC actuation (ATWS mitigation system); to list-of correct answers.
REFERENCES:
Logic Diagram LO-LP-37042-01C-
1 3 I GeorgiaPower en===, cm=== VOGTLE ELECTRIC GENERATING PLANT g ' TRAINING LESSON PLAN I TITLES g cAT"' ^!IISAL*N 71,47, orc 34ctgy sfgggAc 4YST* "h NUMBER': :.o-v- 57o+a -ge l~ PROGRAM: L/cre_ MJE A opg4Aro/( I AUTHOR:
. REVISION: af /
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l Lo LP-3'Jo4. t-spc lil. LESSON OUTLINE: NOTES l.'fdAblME TAlf
- a. /MGCHAA/lc A L kJ E LEOTdic A L
- 1. :AFv) ACTUA TI0 +I a T h L. 6c'TH tta e u m P r STALT V s4b C.S.o.ciAT!!$b s
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I QUESTION 6.12 (1.50) What THREE conditions must be met to reset the Turbine Driven Auxiliary Feedwater pump trip and throttle valve?~ l ANSWER 6.12- (1.50) 4 (0.5 each)
- 1. Local handswitch for trip and throttle valve in local.
- 2. Overspeed trips not present.
- 3. Mechanical overspeed reset at valve.
REFERENCE VEGP Training Text, Vol. 7, Chap. 13.d., p. 13d-12 061000K407 ..(KA's) ( l l l l l
)
1 J l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) ! l l _ _ _ _ _ __b
i ANSWER: 6.12' '! STATEMENT: -The answer key incorrectly states that the TDAFW pump trip and trottle valve must be in. LOCAL in order'to be. reset. -r
\
FACILITY-RECOMMENDATION: 1 Delete answer 1.
.a l
REFERENCES:
Vogtle Text Chapter 13d j
. .q
.q H,
l l I
- . AUXILIARY FEEDWATER SYSTEM NOTE: HS-45111C aust be in LOCAL to operate'with 151118 or A.
The limitorque motor will energize and drive the operator to 80 percent closed position, not far enough to relatch. The green'11ght is illuminated when~limitorqua is fully closed. It remains illuminated until limitorque is fully open. If the valve is' closed, depressing HS15111-A resets and opens the valve. NOTE: Mechanical overspeed trip aust have been. reset, if present. The signal generated by depressing HS15111A will cause liettorque to move to-fully closed position, and relatch the valve. The limitorque will then move to open valve. The red light will be illuminated when limitorque is fully.open. The red-light stays illuminated until the actor operatur is at 80 percentLclosed position. NOTE: These are operator indications, not valve positions. 1 The valve position is indicated by red T&T' valve open and green T&T "' valve closed indicating lights. j To perfore this task, complete the following step. ' - ' N A . . fM,W& , N ; To reset the T&T valve from the PAF , the following conditions must occurs '
] j d ' 1L s - J.-
~
- HS-15111C;sust'be'in,the LOCAL position. -
Overspeed trips aust not be present. Mechanical overspeed trip sunt have been reset at the valve. g .
/- -_
__/- By depressing HS-15111A, the signal will cause the liettoritue to move to~ the fully CLOSED position, relatch the valve, and then open the valve. To manually reset the sechanical trip linkage, pull the spring-loaded energency connecting rod against the spring force, moving the energency head lever away from the energency tappet and tappet nut. This action allows the tappet spring force to enve the emergency tappet in the downward direction and the tappet nut to locate in the TRIP RESET position on the head bracket and hold the energency connecting rod in position under its spring tension. Spring force, acting on the limit switch lever, will anye the lever-and contacts to the CLOSED position, completing the electrical circuit to the actor operator of the trip and throttle stop valve.'(Figure 1) To operate the limitorque actuator, push down handsheel to engage. If a the valve is latched, the handwhee1~can be operated to move the valve stes to opsn or close the T&T valve. I To latch with annual override, ensure that the overspeed trips are ' reset. Reset the mechanical trip linkage, if necessary. Then, turn the handwheel, clockwise, in the closed direction, until the latching anchanism on the valve latches. 13d-12 Revision 0
QUESTION 6.13 (2.00) List FOUR major loads which would be affected by failure of the 125 VDC class 1E bus 1AD1. DO NOT include 4.16 kV, 480 V, 13.8 kV switchgear; mise. equipment panel, Hvac or termination cabinet. i 1 ANSWER 6.13 (2.00) ! (any 4 0 0.5 each) 1.
- 2. Diesel Generator 1A Generator Control power-(field flash).
- 3. Engine Control Panel A.
4. Solid State Protection System Chan. I. Reactor Trip Switchgear Train A.
- 5. Auxiliary relar cabinet Train A.
REFERENCE VEGP Training Text, Vol. 8, Chap. 16.b., p. 16b-5 063000K302 063000K301 ..(KA's) (****~* CATEGORY -6 CONTINUED ON NEXT PAGE *****)
"'" , 3 0 .ij .y ,; , R .: ,__ , s I
, , , - +
,1.<,=, c. < i. ,
_ t;. e l j -) QUESTION NUMBER:' f6.13 L-STATEMENT: There are many.more~majorf(1.e.;'important) lo'ds a
'affected byLa loss of 1ADI than,those on the key for ' '
panel'IAD11 1 Especially.since 1ADl. supplies two train'A 120VAC busses through inverters.- ' FACILITY RECOMMENDATIONS: Delete the question or accept any' items listed'on
- the attachments to; 18034-1 and'18032-1.-
REFERENCES:
Vogtle' Text Chapter 16b 18032 18038-1 1 1 1 1
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l ( l l 3, .l
y g g e r ,
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t f) , DC. ELECT {(ICAL POWER SVsTEM< s . , 1 4 ) , !
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- 125 VDC dist?ibution switchgear V. C. 125 VDC DISTRIBUTION SYSTEM LOADS
- , t i
> _8 The.loaos supplied by tne 125 VDC system are as follnws:
- f 125 Volt Cless-1E Loads Switchgear: JAD1 Battery Chargers /400 VAC Supply: 1ADICA/1ABA,-1ADICB/1ABE A'
Batteryt., . .. IAD.tB Paneig $$ $ ADD.....TP - Capacity - 400 Ampere Locati on *, Control Building-Level B Leads: 4.16 kV Switchgear................ 1AA02 480 V Swi tchgear. . . . . . . . . . . . .'. . . . . 1 AB04, 1AB05, 1AB15 ' 13.8 kV Switchgear.................1.AAA, 1 DAD,,1BAB, 1CAC Diesel Generator 1A Generator Control F;ower(Field Flashing) w/ s Diesel Generator 14 Engine Control Panel A Solic-State Protection System (SSPS) Channel I ' Reactor Trip Switchgear - Train A I Auxiliary Relay Cabinet '- Train A \ I Miscellaneous System / Equipment Panel -] HVAC Panel IACQHVCt l Termination cabinet 1ACPT15 l Panel 41ADII.l' 1 Capacity: 400 Ampere Location: Control Building - Level B ! l Loads: Diesel Generator 1 A Engine Control Panel B 1 1 t Diesel Generator 1A Generator Control Panel Ter mi nati on Cabi net. . . . . . . .. . . . . . . 1 ACPTO3 ! 16t-5 Revision 1
e - PRocEDUME NO. . HEVISloN PACE No. 4
, VEGP '18034-1 0 2 of 59 S
A. LOSS OF 125V DC BUS 1ADl; )
' SYMPTOMS e 125V DC Vital Bua IAD1 voltage low.
e Loss of ' power to lAYlA and 1AY2A 120V AC Vital Instrument Panels. e- Loss of indicating lights on lAA02 and 1A304, IAB05 and 1AB15 Switchgear Controls. . t e Train A Main Steamline Isolation. j e Train A Main Feedwater Isolation. ! i NOTE l t See Attachment A for Equipment Responses, Breaker and Valve Control Loss, Valve Failures from loss of - instrument air, and Annunciator Failures. ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAIITED A1. Dispatch an omeratur to the IAA02 SWGL Room to:
- a. Verify offsite power is a. IF Diesel Generator IA is available to AC not running, it can not be EmerEency Bus 1AA02 started, o Ensure' Normal' Feeder e Loss of Train A.
Breaker lAA02-05 is CLOSED. s Go to Step A3.
- b. IF Emergency Diesel Uenerator DGlA is running, and no offsite power is available to 1AA02 Bus, THEN do not perform Step X2 until offsite power to Bus lAA02 is rastored.
- c. Go to Step A3.
l l 103aes
PROCEDURE No, REVIStolJ . IPAGE NO. l VEGP 18034.1 0 3 of 59 1 ACTION / EXPECTED RESPONSE, RESPONSE NOT OBTAINED NOTE Diesel Generator Electrical Protective Trips are ' inoperable. . ., A2. Dispatch an operator to locally remove. Diesel Generater DG1A from service,
- s. Ensure 1AA02 Bus Feed Bresker 1AA02-19 is OPEN.
- b. Stop the Diesel Generator by placing the " Pull-To-Run/
Push-To-Stoa" handewitch at the sout:2 end of the engine to STOP. A3. Dispatch an operator to A3. Initiate 18032-CF LOSS 0F , restore power to 120V AC 1120V AC INSTRLMENTl POWER,. InstrumentLPane11A"lA{ ySub-proca.ure d A'.- , a. Open Normal Feeder Breaker from Inverter 1AD1II: . . e 1AYlA-02. ! b. Ensure the Regulated Transformer Su? ply Breaker for 1A5C09X is CLOSED: l e 1ABC-09. .
- c. Close Alternate Source Breaker to lAYlA Instrument Panel:
e 1AY1A-01. F03445
WOOCEDURE No. ret /nSloN PAGE No. l VEGP 18034-1 0, 4 of 59. 1 . t i ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED A4. Dispatch an operator to A4. Initiate 718032-C;aLOSScOF rectore power to 120V AC e5120V!AC)flNSTRUMENT" PANEL. InstrumentPanelS1AY2AL}{ , Subhprocedufe B. ,
- a. Open Normal Feeder Breaker from Inverter 1ADlIllt e IAY2A-02.
- b. Ensure the Regulated Transformer Supply Breaker for IABB40X is CLOSED:
e 1ABB-40.
- c. Close Alternate Source Breaker to IAY2A Instrument Panel e 1AY2A-01.
AS. IF an RCP 1 undervoltage AS. IF the< sump does not trip, or underfrequency condition BIspatch an operator to PSDA arises, trip the pump using to trip the pump: 1-HS-0495B on QMCB. e 1-HS-0495D. A6. Dispatch an operator to A6. Initiate to 13405-1, 125V DC restore power to Bus lADI. IE ELECTRICAL DISTRIBUTION SYSTEM.
- a. Ensure Bus Feeder Breakers are CLOSED: o Continue attempts to restore power to lADl.
e 1AD1 From
- Battery 1AD13, l .. e 1ADI From Charger lADICA, o 1ADl From Charger LADICB.
- b. Ensure Panel Feeder Breakers are CLOSED:
a IADl Feeder to IADll, e lADI Feeder to IAD12, a lADI-ll - Feeder to IAD1M. FOM45
[ REVISloN PAGE No. '1 PROCEDURE No. . .
-VEGPL <
'18034-1 .0 10 of.59!
' Sheet'1:of 17 1 ATTACHMENT'A
, .nw-memwxmW TABLErl W:LOS5 OFE1A&O2eBUSJBREAKER' i CONTROL- q l
EQUIPMENT RESPONSE:(Failure of: Bus LAD 1). . ;}
.el Main Feedwater. Isolation, Isolation. Bypass, Main' Flow Control, .
+ -l and Startup' Flow; Control Valves;close resulting in.Feedwater i '
Isolation Main Steam-Isolation"and' Isolation Bypass Train A Valves Close
]
e ,
!resulting in ateamline. isolation.-
a e LReactor and Turbine trip' occurs from loss of. main feedwater.-L e .' Control Power is lostito 1AA'and'1AB SWGR Breakers. , e Diesel Generator DGIA~ control-power to Generator Control. Panel ! PDG1 and' Engine Control Panel PDG2 is' lost rendering-:the Diesel : Generator inoperable: if running, it willsfail as is with 4 loss 'J of electrical protective trips, frequency, Land voltage' control. Due to loss of power to the: Low Speed: Relay, the' generator space, i Engine Lube 011 and Jacket Wate.r Heaters and Lube Oil ~and Jacket-Water Keep-Warm Pumps will come on.' - j H e Power to Inverters lADlII.and. LAD 1111~is lost causing 120V AC Vital Busses lAYlA and'1AY2A to de-energize. e Instrument Air Containment Isolation Valve 1-HV-9378 closes resulting in loss of instrument. air inside Containment.. e- Power To Isolation Panel.lACQIPl is lost rendering the annunciators:in Train A: inoperable. NOTE Feeder Breakers must be manuallyfcontrolled in the-event the transfer to an alternate power supply-is required. ~IF the Diesel Generator-(DGlA).is not - running, it may not be selected as an alternata= power source. l '- ' yes . F
PROCEDURE No. REtflStoM PAGE Wo. VEGP 18034-1 0 al of 59 Sheet 2 of 17 ATTACHMENT A
~
TABLEc1T JLOSf 0F 1 02"' BUS BREAKEP CbNTROL The following 1AA02 Bus Feeder Breakers will fnil as aligned prior to ' loss of Bus IAD1: . . 1AA02-05 Normal Power from RAT XFMR Normally Closed INXRA 1AA02-19 Emergency Backup from Diesel Normally Open Generator (DGIA) NOTE Output Breakers must be manually controlled in order to START /STOP the following listed pumps or to ENERGIZE /DE-ENERGIZE 480V Load Centers 1AB04, 1AB05 and 1AB15. 1AA02-03 CCW PMP 1 1AA02-04 NSCW PMP 1 1AA02-06 CB ESF CHLR A 1AA02-07 CCW PMP 3 1AA02-08 NSCW PMP 3 1AA02-09 RHR PMP A 1AA02-11 CCW PMP 5 1AA02-12 NSCW PMP 5 1AA02-13 CCP A 1AA02-14 CS PMP A 1AA02-15 ACCW PMP A 1AA02-16 SIP A 1AA02-17 MDAFW PMP A 1AA02-10 Supply To Load Center IAB15 1AA02-20 Supply To Load Center LAB 04 1AA02-21 Supply To Load Center 1AB05 1AA02-22 Supply To Load Center 1NB01 703445
PROCEDURE No. REVISION PAGE No. VEGP- 18034-1 0 12 of 59 Sheet 3 of 17 ATTACHMENT A
'E 2
TABLEa2.t " LOSST Of21hBl04 BUS BRE KER CONTROL NOTE ' ' The following Feeder and Output Breakers must be manually controlled at IAB04 Bus. 1AB04-01 IAB04 Bus Feeder Breaker IAB04-02 1AB04 Feeder To MCC-1ABE l 1AB04-04 CC Fan 1 Hi Speed 1AB04-05 CC Fan 1 Low Speed 1AB04-08 CC Fan 2 Hi Speed 1AB04-09 CC Fan 2 Low Speed 1AB04-12 CC Fan 5 Hi Speed 1AB04-13 CC Fan 5 Low Speed 1AB04-16 CC Fan 6 Hi Speed 1AB04-17 CC Fan 6 Low Speed 6 i
PROCEDURE No. REVISloN PAGE NO. -,
'l 13 of.59 VEGP 18034-1' 0-Sheet 4 of 17
-ATTACHMENT A .
1 F ) TABLE.3I.4- nLOSSiOF21AB65 BishBNAEkkid3NNROL 1 LJ NOTE
~
The following Feeder and Outiput Breakers must be manually controlled at 1AB05 Bus. 1AB05-01 1AB05 Bus Feeder Breaker i 1AE05 1AB05 Feeder To MCC-1ABA 1AB05-04 CBCR Fltr Unit 1 ) i 1AB05-05 1AB05 Feeder To MCC-1ABC 1AB05-06 CBCR Htr 1 l 1AB05-08 CTB H 2 Recombiner 1 : 1AB05-09 CB Elec Pen Fitr Unit Fan 1
. i DGIA Vent. Fan 1 1AB05-12 1/.305-13 DGlA Vent Fan 3 :
1AB05-14 1AB05 Feeder To MCC-1ABF
1
;4 PROCEDURE NO.' REVISON PAGE NO.
.14 of 59-18034-1 0-
~
.VEGP Sheet 5 of 17 ATTACHMENT A TnE 4SEBLOSS"OF E 1 1XhiFBuFinzArrFE6wraot-NOTE The'following Feeder and Output Breakers must be-manually controlled at LAB 15 Bus.-
~1AB15-04 Pipe Pen Fitr & Exh Unit Htr l~
'1AB15-05 'NSCW C1g Twr Tr A' Fan'l 1AB15-06 'NSCW C1g Twr Tr.A Fan 2-
~1AB15-08 -NSCW C1g Twr Tr A Fan.4 -
1AB15-09 1AB15 Feeder to MCC-1ABB' 1AB15-10 1AB15 Feeder to MCC-1ABD 1AB15-12' Spend Fuel Pit Pump 2'TR.A 1AB15-13 NSCW C1g Twr Tr A Fan 3 1AB15-14 Pipe Pen Fitr & Exh Unit Fan l' . l
.. j
.- I i
'I l
ma.s - I . - l
4 l PROCEDURE No.' . REVISloN . PAGE No. .
-.VEGP '18034-1. . . 0: <
15)of 59.- ) Sheet-6 of 17-ATTACHMENT A ,
-LOSSTOFS125VfDCEBUS 21ADl % !v- 1 w
'l
['LOSSEONiNSTRUMENk'.I11R$TbIbbhTAk wDUExT0eLOSS10Fs125V+DCETO m LAD
;a a s a c u a w -.a m a s s m *j ~ "12- ^i g~ r.
~
FAILED- ' l POSITION
- ' REACTOR COOLANT SYSTEM 1-HV-8032 Reactor Vessel' Flange Leakoff ,
.Open .
WASTE PROCESSING SYSTEM - LIQUID ,
.l 1-HV-7127 <RCDT To Drain Pump-Suction Open
-)
'l-HV-7144 RCDT~Recirc' Isolation Valve .Open ' '
CONTAINMENT AND AUXILIARY BUILDING DRAINS.- RADI0 ACTIVE I 1-HV-17091 CTB Cooling Units To CNMT; Sump Open ' . 1-HV-17095 CTB Cooling Units To CNMT' Sump .Open; ;{ MAIN STEAM SYSTEM 1-PV-5238 Steam Generator N . Blanketing System Open 5 1-PV-5239 Steam Generator N Blanketing System! Open 1-PV-5240 Steam Generator N Blanketing-System .Open 1-PV-5241 Steam Generator N :. Blanketing-System Open-1-PV-15250 Steam Generator N -Blanketing-System Open 1-PV-15252 Steam Generator N -Blanketing; System Open. s } . . . 703445
PROCEDURE NO. . . REVISION - PAGE NO,
.VEGP 18034-1 0' '16 of 59 ' -
]
Sheet 7:of 17 l ATTACHMENT A~ j LOSS OF 125V DC BU'S'1AD1 .
' LOSSE6 N N5 5UMENTTAIRl W 56Eh I' N 1
'"DUE 'TO LOSS'"0F'"125V"DCETO 1AD12e '
FAILED-POSITION i REhCTOR COOLANT. SYSTEM- , 1 1-PV-0455B. ' Pressurizer Spray-Isolation Valves- : Closed ~j 1-PV-0455C. Pressurizer Spray < Isolation Valves . Closed l-HV-8145 Pressurizer Spray Isolation Valves . Closed-1-HV-8031' 'PRT Drain To The RCDT ~ Closed' i ll 1-HV-8047. :PRT Vent To The WGDT Closed 1-HV-8030 RMWST Fill To The PRT- Closed 4 l l 703aa5 , _a_________
- r.
REVISloN PAGE No. -
! PROC DU E No. 'i -. , .
-)
Sheet'8 ofLl7 :l ATTACHMENT:A
' LOSS'0F 125V~DCtBUS~1ADI- j
'LOSSOOFh1NSTRUMENTIA b bAINMENT .
deDUE LT05 LOSSIOFill23VkDCtT0ilAD12 b* - ,j m .. .- FAILED ,
!POSITIONz CHEMICAL'AND' VOLUME CONTROL' SYSTEM' 1-LV-0178- RCP Standpipe Fill Valves, Closed" l-LV-0179- RCP Standpipe Fill. Valves ' Closed-1-LV-0180 RCP Standpipe Fill Valves' : Closed ~
1-LV-0181 RCP Standpipe Fill Valves. Closed
~
1-HV-8141A Number 1 RCP Seal Leakoff Valves Closed 1-HV-8141B Number 1 RCP Seal Leakoff Valves Closed 1-HV-8141C Number 1 RCP Seal. Leakoff Valves Closed .. 1-HV-8141D Number.1 RCP Seal Leakoff' Valves Closed L' t 1-LV-0459 RCS Letdown Isolation Valves Closed j RCS Letdown Isolation Valves Close'd
~
1-LV-0460 1-HV-8149A RCS Letdown Isolation. Valves Closed
- 1-HV-8149B ~RCS Letdown. Isolation-Valves Closed 1-HV-8149C RCS Letdown Isolation Valves Closed 1 HV-15214 RCS Letdown Isolation Valves Closed 1-HV-8160 RCS Letdown Isolation-Valves Closed
.1-HV-8153 RCS Excess Letdown-Isolation Valves ' Closed 1-HV-8154 RCS Excess. Letdown Isolation Valves ' Closed-1-HV-0123 RCS Excess Letdown Isolation Valves Closed 1-HV-8143 RCS Excess Letdown Divert Valve. To VCT L 703445 .
___n____.r._ _ _ _ .
REVISloN: PAGE No.. PROCEDUZE No.
.VEGPM 18034-1 0' 18'of.59-Sheet 9'of 17' ATTACHMENT A LOSS OF 125V DC BUS-1ADI LOSS 70FEiE5iRUMENTMIROT0 CONTAINMENT
" DUE"T0TLOSS ' 0F*125V DC'T0"1AD12 FAILED
' POSITION.
SAFETY INJECTION SYSTEM' 1-HV-8878A ' Accumulator Fill Isolation Valves ' Closed. 1-HV-8878B Accumulator Fill. Isolation Valves . Closed
'l-HV-8878C Accumulator Fill Isolation Valves- Closed 1-HV-8878D Accumulator. Fill Isolation Valves Closed.
1-HV-8871 SIS Check Valve And System Test Line Isolation Valves 1-HV-8881 SIS Check Valve And System Test,Line Isolation Valves 1-HV-8823 SIS Check Valve And System Test Line
' Isolation' Valves-1-HV-8824 SIS Check Valve And System Test Line .
Isolation Valves'. 1-HV-8825 SIS Check' Valve And System Test Line Isolation Valves 1-HV-8843 SIS Check Valve And System Test Line-Isolation Valves 1-HV-8877A, SIS Check Valve And System Test'Line B,C, and D Isolation Valves 1-HV-8879A', SIS Check Valve And System Test Line. B,C, and D Isolation. Valves 1-HV-8882 SIS Check Valve And System Test Line Isolation Valves 1-HV-8889A, SIS Check Valve And System Test Line B,C, and D Isolation Valves 1-HV-8890A SIS' Check Valve And System Test Line and B Isolation Valves l 1 il m- _
'l
.1
________--_--__-_a_ _ i
r _ . _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _- - PROCEDURE No. REtflSlofd . PAGE Mo. VEGP 18034-1 0 19 of 59 Sheet 10 of 17 ATTACHMENT A LOSS OF 125V DC BUS 1AD1
-.::; : :7T .
LOSS.OFINSTRUMENTcAIRs W CONTAINMENT 2 i
D.UE -TO < LOSS ' OF 2125V2 DC mTO31AD12 FAILED POSITION WASTE PROCESSING SYSTEM - LIQUID 1-HV-7699 RCDT Pumps Discharge Containment Closed
- Isolation Valve 1-HV-1003 RCDT Level Control Valve Closed 1-HV-7126 RCDT Vent To GWPS Closed 1-HV-7143 RCDT Recire Valve Closed 1-HV-7141 RCDT To PRT Isolation Valve Close'd NUCLEAR SAMPLING SYSTEM - LIQUID 1-HV-3507 Pressurizer Liquid And Steam Space Closed Sample Isolation Valves 1-HV-3513 Pressurizer Liquid And Steam Space Closed Sample Isolation Valves i
1-HV-3501 RCD Hot Leg 3 To Failed Fuel Closed Detector Isolation Valve p 7034e5
O PROCEDURE No. REVISION PAGE Wo. VEGP 18034-1 0 20 of 59
. Sheet 11 of 17 ATTACHMENT A LOSS OF 125V DC BUS 1AD1 m . :" ' . . . _ .
LOSS?OF^ INSTRUMENT." AIR 1T0fl CONTAINMENT l DUE>TO LOSSq OF ;:125V :DC tT0ilAD12 me - . . FAILED POSITION CONTAINMENT AND AUXILIARY BUILDING DRAINS - RADIOACTIVE 1-HV-0780 Reactor Cavity And CNMT Sumps Closed Discharge Isolation Valve STEAM GENERATOR BLOWDOWN SYSTEM l-HV-15212A Steam Generator Blowdown Isolation Closed Valves 1-HV-15212B Steam Generator Blowdown Isolation Closed Valves 1 1-HV-15212C Steam Generator Blowdown Isolation Closed l' Valves 1-HV-15212D Steam Generator Blowdown Isolation Closed. Valves 1-HV-15216A Steam Generator Blowdown Isolation Closed Valves 1-HV-15216B Steam Generator Blowdown Isolation Closed Valves 1-HV-15216C Steam Generator Blowdown Isolation Closed - Valves 1-HV-15216D Steam Generator Blowdown Isolation Closed Valves CONTAINMENT PURIFICATION AND CLEANUP SYSTEM l-HV-2626B CNMT Mini-Purge Supply Isolation Closed Damper 1-HV-2628B CNMI Mini-Purge Exhaust Isolation Closed Damper
.n 1-HV-12985 CTB Preaccess Filter Unit Dilube Closed 1-HV-12987 CTB Preaccess Filter Unit Diluge Closed D40$
PROCEDURE NO. REVISION PAGE NO.
, .VEGP- 18034-1 O' 211of 59 Sheet 12 of 17 ATTACHMENT'A
.'; 27
- . . ,... wwLOSS10F+125V-DC T0" BUS *_1AD1#.
[. ; ENR $NN NNUhtCI Tbb IA$tk$$1DSTN![I DUE'lTO:,_ LOSS w .- 0F2125Vf DC.ET0..,Jc1ACQIPl!FROMt,1AD12
~- .
ALARM BOARD / WINDOW DESCRIPTION ALB02-B1 NSCW TRAIN A LO.HDR PRESS i ALB02-B2.- NSCW TRAIN A TRANSFcPMP LO DISCH PRESS ALB02-D2 NSCW TRAIN A CNMT CLR 1&2 LO. FLOW. ALB02-E2 NSCW TRAIN A CNMT CLR 5&6 LO FLOW-ALB02-F2 NSCW TRN A RX CVTY CLG COIL LOW FLOW ALB02-C3 NSCW TRAIN A DG CLR LO FLOW ALB02-E3 NSCW TRAIN A NORM /BYP VLV MISPOSITIONED ALB02-C4 NSCW TRAIN A RHR PMP & MTR.CLR LO FLOW ALB02-A5 CCW TRAIN A SURGE TK LO-LO LVL ALB02-E5 CCW TRAIN A RER PMP. SEAL LO FLOW ALB02-A6 CCW TRAIN A LO HDR PRESS ALB02-B6 CCW TRAIN A LO FLOW ALB02-D6 CCW TRAIN A RHR HX LO FLOW ALB04-El TRAIN A SYS STATUS MON ~PNL~ ALERT ALB04-F3 TRAIN A SHUTDOWN PNL ON LOCAL CNTL ALB04-F5 TRAIN C SHUTDOWN PNL ON LOCAL CNTL 70344S -
PROCEDURE NO. . . REVISION PAGE NO. VEGP.. 18034-1 0 22 of 59 Sheet 13 of 17 ATTACHMENT'A
. . ve==hik?fb5???@i::W',"==::
dbLOS S "0F'"125V ibC "T02 BUS J1ADl:bt;
. . . - ma n rau+~ ~~~-
nui..lTRAIN'm' ANNUNCIATOR ALARMS LOST DUE'TO LOSS OF_125V DC TO 1ACQIPl FROM 1AD12. , ALARM: BOARD / WINDOW DESCRIPTION, ALB05-A5 GROUP-1 MONITOR LIGHT COMP'0FF NORM ALB05-B5 . GROUP 2 MONITOR LIGHT COMP OFF NORM ALB05-C5 GROUP 3 MONITOR LIGHT COMP OFF NORM ALBG5-D5 GROUP 4 MONITOR LIGHT COMP OFF NORM
-ALB05-E5 GROUP 5 MONITOR LIGHT COMP OFF NORM {
ALB06-C1 RHR PMP OVERLOAD TRIP 1 i ALB06-F6' SI PMP OVERLOAD TRIP ; i ALB07-C6 CHARGING PUMP OVERLOAD TRIP' -I i ALB09-El MANUAL REACTOR TRIP I ALB09-F1 MANUAL SAFETY INJ REACTOR TRIP j ALB12-E4 PCV-455A OPEN SIGNAL
'ALB12-E6 A COLD OP ACTU VLV HV-8000A NOT FULL OPEN .
i l i i
)
703446
a PROCEDURE NO. ' REVISl@N , PAGE NO. VEGP' 18034-1 0- 23 of 59
-Sheet 14 of 17 ATTACHMENT Ao
$x. _. e s m t m :! m r p r .
- 61085 OF/a125VODCaToeBUS LAD 1-a - w : m 7 m ~ - . ..-._
-_-..,-TRAINTA# ANNUNCIATOR; ALARMS; LOST ~ '
DUE,lT0lLOSS"0F"125V DC:TO 1ACQIPl FROM 1AD12 ALARM' BOARD / WINDOW -DESCRIPTION LALB14-A4- MN STM LOOP 1.TRN'A; ISO-VLV GAS LO' PRESS. ALB14-B4- MN STM LOOP 2 TRN A' ISO VLV GAS LO PRESS ALB14-C4 MN'STM LOOP 3 TRN A ISO VLV GAS LO' PRESS ALB14-D4 MN STM LOOP 4 TRN A' ISO VLV GAS LO PRES l ALB14-A5 MN STM LOOP 1 TRN A' ISO VLV' TROUBLE-
'ALB14-B5. MN STM LOOP..2 TRN A ISO VLV TROUBLE ALB14-C5 MN STM LOOP 3 TRN A ISO VLV TROUBLE !
l ALB14-D5 MN STM LOOP 4 TRN A ISO VLV TROUBLE ALB14-F6 MN STM ISO VALVES.NOT FULL OPEN ALB16-F5 AFW AUTO START MFPT TRIP RLY CNTRL PWR LOSS ALB16-F6 AFW P-3 DISCH HDR LO PRESS. l l ALB34-D1 125V-DC SWGR 1AD1 TROUBLE ALB34-El BAT CHARGERS 1ADICA 1ADICB TROUBLE l ALB34-D2 125V DC MCC 1ADlM TROUBLE ALB34-E2 INVERTERS 1ADlIl 1ADlIll TROUBLE L ALB34-F2 125V DC PNL 1ADll TROUBLE 'j l l ALB34-D3 125V DC PNL 1AD12 TROUBLE i ALB34-E3 120V AC PANELS lAYlA LAY 2A TROUBLE ! I ALB34-A6 13.8KV SWGR 1AAA TROUBLE I i i i l 703845 l
j PAGE NO, PROCEDURE NO. . REVISION . ' VEGP 18034-1 0 24:of;59i Sheet 15 of:.17 ATTACHMENT.'A .) LOSSMFi$5bEYWEUS1ADll E e: ,.r..
-e?-
TRAINEATANNUNCIATORiALARMS9 LOST cDUE3301_1088tOF2125V"DC TO 1ACQIPl FROM 1AD12' I ALARM BOARD / WINDOW : DESCRIPTION .
'1l
.ALB34-B6 13.8KV SWGR'1BAB TROUBLE ,l ALB34-A7 13.8KV SWGR.1CAC-TROUBLE ALB34-BT '13.8KV SWGR 1 DAD-TROUBLE f
ALB36-Al' 4160V'SWGR'1AA02. TROUBLE-
-ALB36-Bl 4160V SWGR~1AA02 NEG PH SEQ BUS PT ALB36-Cl 480V MCC'1ABA TROUBLE
'ALB36-D1 480V MCC 1ABD TROUBLE ,
ALB36-El 4160V/480V SWGR TRN.A'. TRANSFER SW ON. LOCAL ,- 1 ALB36-F1 ISO DEVICE-PNL TRN A_.QIP5 TROUBLE ALB36-A2 480V.SWGR-1AB04 TROUBLE ALB36-B2 480V SWGR 1AB15 TROUBLE ALB36-C2 480V MCC 1ABB TROUBLE'- t ALB36-D2 480V MCC 1ABE TROUBLE ALB36-F2- BAT 1ADlB BRKR OPEN ALB36-A3- 480V.SWGR 1AB05. TROUBLE' ALB36-B3 SEQ A PNL DOORS 0 PEN ALB36-C3 480V MCC 1ABC TROUBLE. ALB36-D3 -480V MCC 1ABF TROUBLE ALB36-F3 ISO DEVICE PNL TRN A QIPl TROUBLE ~ 703446
~-
e, , w PROCEDURE NOc REVISION .. PAGE NO.
.VEGP; '18034-1 <
, 0- 25 of'59 w
' Sheet 16 of.17 ATTACHMENT A
. ... u._- 2--~% . . .
M _e1257!:DCI W;J3US;71AD1y
~~
. y;;ca,TRAINIA~ ANNUNCIATOR ' AUKRMSELOSTE '
. DUESTOT&OSS 20F"125V' DC "T. OTI ACQIP I'" FROM " 1AD 12
- ALARM BOARD / WINDOW- : DESCRIPTION ALB50-B3:- 'CR HI/LO DIFF; PRESS.
?ALB50-B4- ' LCR . TRAIN: A FLTR: TROUBLE:-
a-JALB50-A7: BAT RM TRN A SPLY FANLLO'AIRLFLOW;-
- ALB50-B9i 'ELEC PEN TRAIN:.A FLTR. TROUBLE:
ALB51-Al DG.TRN A CBLETUNNEL HI TEMP t , ALB51-B1 NSCW TRN-A CBL TUNNEL HI TEMP
.ALB51-C1 AB.CBL' TUNNEL HI. TEMP ALB51-A3 SWGR ABIS RM HI TEMP .
ALB51-B3 MCC~ABD.RM HI TEMP ALB51~-C3 MCC ABB RM HI TEMP ! ALB51-D3 RHR TRAIN A PMP'RM HI TEMP , ALB51-E3 CCS TRAIN A PMP RM HI TEMP :l ALB51-A4 CCW TRAIN A PMP RM HI. TEMP i ALB51-B4 CVCS TRAIN A PMP RM HI-TEMPL ALB51-C4 SI TRAIN A PMP RM HI TEMP > ALB51-D4 SFCS & HX TRAIN A RM HI TEMP q 1 ALB52-B2 PIPE PEN TRAIN A FLTR TROUBLE. ALB52-C8 CNMT POST LOCA F-1 LO AIR FLOW h 703445
.__m.ml___.___E_.I.__U-_...i.
1 l PROCEDURE NO. REVISION PAGE NO. -
'i VEGP 18034-1 0 26.of 59 .
Sheet 17 of 17 i I ATTACHMENT-A-f.,653 T op"l555YNEhiBNS7 1AD1]"lJ
. TRAIN.JA ANNUNCIATOR.. ALARMS., LOST.L.._..
DUEr OSS10Fjl2.5ViDCET011ACQIP1;FRG O AD12 ,j 1 ALARM BOARD / WINDOW DESCRIPTION' ALB53-A2. AFW TRN'A PMP HOUSE HI-TEMP ALB53-C2 AFW~TRN.A FMP HOUSE FAN LO AIR FLOW-ALB53-C7 CHLR TRN A EVAP WTR'HI/LO' TEMP ALB53-D7' CHLR TRN'A EVAP OR COND WTR LO FLOW ALB53-E7- :CHLR TRN A TROUBLE ALB54-Al FHB POST ACCIDENT FLTR 1 TROUBLE ALB54-F2 FHB AREA LO DIFF PRESS ALB61-D6 LVL B LEAK DETECTED' . ALB61-E6 LVL C LEAK DETECTED ALB61-F6 LVL D LEAK DETECTED ALB62-F5 CNMT H2 MON TRAIN A ALERT END OF ATTACHMENT A m3s43
c. REVISION PAGE No. 2 of.65-PROCEDURE NO. 2 VEGP 18032-1 qA.,,. LOSSrOFeVITAL* INSTRUMENT" PANEL-IAY1A.j
- ~s ac %
%wn; ..
f SYMPTOMS e All Channel 1 Trip Status Lights' energized. e- Simultaneous loss of Source Range.' Channel N31, Intermedia'te Range Channel N-35 and' Power Range. Channel N-41. e 120V AC INSTR' PANEL 1AYIA TROUBLE Alarm at EAB. e INVERTER IADIII TROUBLE Alarm at'.EAB.. ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED I NOTE SG1 and 4 ARVs will NOT operate following loss of~ Panel. IAY1A. .j A1. Verify Reactor - NOT A1. Perform the following steps 1 Tripped. concurrently with 19000-C, j E-0 REACTOR TRIP OR SAFETY j
-INJECTION:
- a. Position PRZR Pressure Control Selector Switch PS-455F to Channel 457/456. ;
- b. Position PRZR Level l Control Selector Switch l LS-459D to Channel Ji 461/.460.. l l
- c. Go to Step A13.. {
l l i l l l I l m..s
c REVISION - PAGE NO. . PROCEDURE NO. 18032-1 2- 6.of 65
.:VEGP
]
Sheet 1 of 1. l
. 1 1
l-1 i ATTACHMENT A. j
~ "
. _ : ~ r_ . . .--r-n r'27:' Fe....,v;- :: :::::~;-~~~;. Wy f . _.,2E;Jln., ' ?,7~* y-
.- TABLE T.WSGWLCE) SELECTOR' SWITCH ALIGIOtENT'FOR' LOSS OF PANEL LAY 1A Required Selector Position Switch No. Function
~
SG 1 Steam Flow. FT-513 FS-512C FT-523 FS-522C. .SG 2 Steam Flow' FT-533. FS-532C.' SG'3 Steam Flow FT-543 FS-542C- SG 4 Steam Flow SG 1 Feed Flow FT-511~ FS-510C FT-521 FS-520C SG 2 Feed Flow
.SG 3 Feed Flow FT-531-FS-530C .
FT-541 FS-540C SG.4 Feed Flow LS-529C SG'2 Level LT-552 LS-539C SG-3 Level .LT-553 r
- 703a85
- - - ~ _ . _ - _ -
?> ,
, ./ ;
, 1 REVISION - PAGE NO.~
. 7 of'65 PROCEDURE NO.
2'
. VEGP ' 18032-1 .
Sheet:1 of.2-J ATTACHMENT'A.
~
pa , nap - - .< # ; w w a ,. ,,;
*- T.ABLgs.2;5PISCELOADSAPANEL11AY1Ah. ,
Description < Alternate-Inst. No.-
~ Controllers PIC-3000A 'SG lLAtmospheric Relief PIC-3030A SG 4 Atmospheric. Relief .
HC-190A Charging Tirottle NC-190B HC-943A Accumulator Vent Throttle .HC-943B Nuclear Instrumentation = NI-31B SR N-31 Countrate NI-32B NI-35B IR N-35 Amis NI-36B NI-41B- PR N-41 I full Power NI-42B NI-41C PR N-41 % Delta Flux :NI-42C-NR-41-P1 PR N-41 Upper Flux Rec NR-41-P2-NR-43-P1 PR N-41 Lower Flux Rec NR-43-P2 Reactor Coolant System Instrumentation TDI-411A Loop 1 Delta T (I) 1 TDI-411B Loop 1 Overpower belta T (1) TDI-411C, Loop l'Overtemperature T (I) TI-412 Loop 1 Tava ] TI-413A Loop 1 WR Th Temp (I) TR-413A Loop 1 & 2'WR Th Rec TI-423A Loop 2 WR Th Temp (I) TI-433A Loop 3 WR Th-Temp (I) l*- TR-433A. Loop 3 E 4 WR Th Rec TI-443A Loop 4 WR Th Temp (I) . PI-405 Loop 4 WR Th Press-(I) .l~ l .PI-438 Loop 4 NR Th Press (I) . FI-414 Loop 1 Flow (I) FI-415.(II)' FI-424 Loop 2 Flow (I)' FI-425 (II). FI-435 (II), FI-434 Loop 3 Flow (I) .!- 1 FI-444 Loop 4 Flow-(I)- FI-445 (II) PI-455A PRZR Press (I) PI-457 (III) LI-459A PRZR Level (I) LI-461 (III)- i l l
)
m.o
REVISION FAGE NO.
~ PROCEDURE NO.
18032-1 2 8.of 65 VEGP Sheet 2 of 2 ATTACHMENT A.
- TABLEo2. TIECLLOADSh-kP M 1AY1A'. .( Cont ! d .-)9
,r
-Description Alternate Inst. No..
Steam Generator Instrumentation LR-501-P1 SG 1 WR Level Rec LI-501 SG 1 WR Level (I) .LI-528'(III) LI-529 SG 2 NR Level (I) LI-538 (III) LI-539 SG 3 NR Level (I) LI-551 SG 1 NR Level (I) LI-517 (IV) LI-554 SG 4 NR Level (I) LI-547-(IV) PR-514 SG 1 & 2 Press Rec PI-514A SG 1 Press (I) PI-516A_(IV)- PI-524A SG 2 Press (I) PI-526A (III) PI-534A SG 3 Press (I) PI-536A (III) PI-544A SG 4 Press (I) PI-546A (IV) Auxiliary Feedwater Instrumentation LI-5111A CST 1 Level (I) LI-5101 (II) LI-5116A CST 2 Level (I) LI-5104 (II) FI-5150A SG 4 AFW Flow FI-5152A SG 1 AFW Flow Miscellaneous Systems Instrumentation LR-990 RWST Level Rec LI-990A RWST Level (I) LI-991A (II) LI-102A Boric Acid Tk Level (I) LI-104A (IV)- CNMT Stunp Level. Rec LR-764 . PI-937 CNMT NR Press (I) PI-936 (II) PI-505 Turbine Impulse Press (I) PI-506 (II) PI-1636 NSCW Train A Supply Press PI-1637 l
_ _ _ _ ~ REVISION PAGE NOc 9 Of.65'- PROCEDURE NO.
. 18032 2- !
.VEGP..
l
. Sheet 1 Of 1 j
-1
.)
ATTACHMENT A.- 7- TINLE::323$ANELh1AY1IILOADiLiN 1 BRK LOAD ;
' I 03- NIS CHANNEL.I CONTROL POWER' I 04 SPARE 05 PROCESS RACK PROTECTION. SET'I' 06 SAFEGUARD TEST CABINET, TRAIN A ..
-07 SOLID' STATE PROTECTION SYSTEM OUTPUT CABINET:1, TRAIN A'
- 08. NIS CHANNEL I INSTRUMENT POWER 09 DIESEL GENERATOR 1A WHM PULSE AMPLIFIER-10> SOLID STATE PROTECTION CHANNEL I, TRAIN B. .i 11 BOP' PROTECTION PANEL CHANNEL I
.i 12 SPARE .
13 HVAC PANEL 1ACQHVC2 i 14 SPARE . 15 AUXILIARY FEEDWATER PANEL 1302-P5-AFP -l 16 SPARE 17 HVAC INSTRUMENT PANEL 1500-V7-001aCBA' 18 SWITCHGEAR 1ADI-' TRANSDUCER POWER: 19 PLASMA DISPLAY A > J 20 SPARE' ! 21 SPARE 4 22 DIESEL GENERATOR BLDG HVAC INSTRUMENT CHANNEL I .l 23 MISC EQUIPMENT PANEL XFMR DF ZLB ) 24 DISTRIBUTION PANELS 1ADll, IAD12 CONTROL POWER 1 END OF SUB-PROCEDURE TEXT , l 1 i j j i 7034 5 ,.
> j
REVISloN PAGE No. PFtoCE RE No. . 6 B.' LOSS OF VITALi1NSTRUMENT PANELLIAY2A i
.- SYMPTOMS' e SEQ A TROUBLE Alarm on,QEAB.
s RMS CHANNEL FAILURE Alarm on QMCB. e Loss of DRMS Safety Related Display. e Loss of Train A System Status Monitoring Panel. ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE s Train A ESF Sequencer will not operate following loss of Panel 1AY2A. Bl. Dispatch an operator to restore Panel LAY 2A by ) initiating 13431, 120V AC 1E VITAL INSTRUMENT DISTRIBUTION SYSTEM. B2. Refer to Attachment B. Table 1, IEC Loads - Panel LAY 2A to determine affected instrumentation. B3. Notify Health Physics of inoperable Radiation , Monitors. B4. Refer to Technical Specifications and j l complete any applicable : Action Statements. 703445
O , PROCEDURE NO. REVISION PA _E NO. VEGP 18032-1 2 11 of 65 ) Sheet 1 of l' l i i I ATTACHMENT B. s- ..
- r. . ..
. TABLEol'. 1.
. PANEL 1AY2A' LOAD LIST /. l I
BRK LOAD 1 03 SAFETY RELATED DISPLAY CONSOLE'DRMS 1 04 DATA MODULE 1RX0005 ,CNMT AREA MONITOR l 05 DATA MODULE 1RX0002 -'CNMT AREA MONITOR 06 DATA MODULE ARX-2532 . FUEL HANDLING BLDG HVAC MONITOR-07 DATA MODULE 1RX-12116 - CR AIR INTAKE MONITOR 08' SEQUENCER BOARD 1821-U3-001 09 BOP SAFETY ACTUATION CABINET J 10 TRAIN.A SYSTEM STATUS MONITORING PANEL 11- PREAMP 1RT-005 12 DATA MODULE'1RX-13119 - MAIN STEAMLINE MONITOR i 13 DATA MODULE 1RX-13120 - MAIN.STEAMLINE MONITOR 14 DISPLAY PROCESSING UNIT 15- REMOTE PROCESSING UNIT A1, CHANNEL I l 1 16 REMOTE PROCESSING. UNIT A2, CHANNEL'I 17 SERVO-AMP FOR ATM DUMP VALVE 1ATPY3000 18 SERVO-AMP FOR ATM DUMP VALVE lATPY3030 19 SPARE 20' SPARE 21 R.G. 1.97 NFMS CAB (DPU-A) 22 SPARE 23 SPARE 24 SPARE END OF SUB-PROCEDURE TEXT e l l l rom
QUESTION 6.14 (2.00)- List FOUR permissive.4 required to allow the Safeguard Sequencer to shut the Emergency Diesel Generator output breaker.
~
ANSWER 6.it (2.00) (any 4 0 0.5 each)
- 1. DG ready to load ( > 440 RPM, > 90% voltage).
- 2. Preferred incoming breaker open.
- 3. NO faults on bus (No bus lockouts).
- 4. DG close permissive from sequencer.
S. No lockout on Diesel Engine or generator. REFERENCE VEGP Lesson Plan LO-LP-01401-01, p. 8 Lesson Objective LO-LP-01401-01, #7a 064000K411 064000K410 ..(KA's) l l s (***** CATEGORY B CONTINUED ON NEXT PAGE *****)
QUESTION NUMBER: 6.14 STATEMENT: There is alternate wording which should be accepted for answer 4. s FACILITY RECOMMENDATION: Accept " Load Shed complete" as equivalent to answer 4.
REFERENCES:
Vogtle Text Chapter 8b. l
~l
n 4 f y
",.. *L 5AFETYlFEATURE SEQUENCER SiSTEM s, 1
,e
- ^
Shea all safety,anc nonsafety ous loacs.(except i loan V centeretransformers). l
-j H l, l c.' Time ='9.5 seconds iTh[65[63aTRUfTM16553MbilTdi18[W'"denc4t$h-[hEI7-wauawns-Y p k ed: M 141M Lyaw w aa,u '
1 , . . . . . ..
' ).
- DG at S440 RPM en ( 590 cercent Vcitece
-*. Pref errea feeder creakers to tus are coen lccm;iletec-within 1:seconc.cf receivinggopen c gnsl)1
- 4. 'No'ous Vaults es' indicated by no cus locicuts on the oreferrec breakers.
- No DG lockouts (IB6C. L:::ckout - Relay 'tr:es).
umwas
.u-~ ~Am - - . - #'
q oTne fellowing lanps' energi:e on .the Nanual Test Panelf D-G READY,FOR LOADING' D-G BRKR CLOSED
* ?, l -
Notet 'When the "DG BRAR CLO5ED" 1amo comes on 'the #- "# sequencer system will. resume operation' - Load center;transf ormers energi:eit.5 second delay allows transformers +o come ua=te voltage)~
- c. Time 110 seconds Sequencer er ergizes requirev loads: f erEUV conditten at 5 se:onc intervals.
- 1A thru 9A and 1C'thru 9C SEQ STEPS INDICATION lamps
- on tne Manual Test Panel.~will. sequentia11y. energi:e .!
and centinue to flash until the' entire sequence is. complete , (accroximately 31 secs. )= i 4
- d. Time'= 46.2 seconds
.)
All SEQ STEF INDICATION lamps;ce-energi:e (approximately: 1.2 secs..after 94 anc 9C lanps.e ergi:e)
- e. Time = 50.2 seconos '
.I 4
The BLOCK'AUTC/MNL SIGNAL 1amo ce-energizes
)
' The' UNDERVOLTAGE. and SEO LOGIC CAIt' aE .' lamps may ce reset (ce-energt:ec) =;
by decressing the' specific oushoutton. 5everal indications are orovicea fj toth locally and in the control recm to alert the coerator of a I se::vencer or other ecui; ment f ailure curing.the UV secuence. ;j 1 4
-i 00*;'- NeV1510n -l
..l
.__w__-____2__ __2_A
QUESTION 6.15 (1.50) I'That does each of the THREE DP transmitters (a, b, c) used for one train of the Reactor Vessel Level Indication System measure? i ANSWER 6.15 , (1.50) (0.5 each) DPa - measures static narrow range level of the vessel from hot les to top of head. DPb - measures static wide range level from bottom of. vessel to top of vessel. DPc - measures differential pressure of RCS flow across the core. REFERENCE VEGP Training Text, Chapter Id, p. 5 002000K603 ..(KA's) L 1 (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
-QUESITON NUMBER: 6.15 STATEMENT: There is alternate wording for all three answers.
FACILITY RECOMMENDATION: \ Accept the following alternate answers: DP befwe(upper range) en hot leg and measures upper headdifferential (no RCP'spressure running) DPh (full range) measures differential pressure between bottom of Reactor Vessel and upper head I (no RCP's running) DP beEwe(dynamic en bottom ofrange) measures Reactor Vessel differential and upper pressure head (with RCP's ninning)
REFERENCES:
LO-LP-16701-01
Of
' LO-LP-16 701-Te0' lil. LESSON OUTLINE:
- 1. INTRODUCTION NOTES __
A. The lesson will describe the purpose, construction, function, and indication performance of the reactor vessel lavel system.
- 11. PRESENTATION .
A. General Overview of System
- 1. LO-TP-16701-00-001 Uses core delta-p to measure vessel level or void Obj ectives content of "'T :::%t. ,#
Tai icnent Wste 4.
. 2.
. ..It is a redundant system (Train A, 3).
3. Delta-p outputs are density compensated for more accuracy. . 4 Operator aids a.
.To detect void or non-cond d C!// !
le ass bubblem
- b. in the reactorWih As -
To assist in identificati Frshc4l saisbr f (Inadequate Core Cooling) of$$Fj chauenge c. Detectvoidingin'coreunderforcedflow;condi-tions. - B. Description c i 1. Differentici Pressure Measurement
- a. LO-TP-16701-00-002 Each train utilises 3 dp c. ells of different !
rangas to measure different areas of the core 2 Trains
~ ~ ~
heiste with the RCP's ruaning or off. O b..
, - Reactoir'YesesMVpper-Rangs.(DFa)- ~'
. NJ1f./M a .
Dj 1) Measures vessel level between hot leg [ nossle and top of head. , Note: Pump running
- 2) Valid only if RCPs are off. in associated loop
- 3) will result in off-Not required by % if any pump is on.
. scale reading.
- c. R (**tittat1 eactor Vessel - Full Range (Kt4esuc (guadekrgg,g DFb')
% 1)
,,,,,g l d.
Measures vesselare top when RCPs level off..from boccom to
~
'/ Reactor Vessel - Dynamic Head Range (DFe) 1)
Provides delta-p of core and internals
- 2) for any cosibination of RCFs running.
Comparison of actual core delta-p to the known delta-p for normal single I phase flow provides a relative indication of voiding or density changes during forced flow conditions.
.s) d dw.b dusu(vyA 5) df
- h'm (N kill a/n .c/,,,,,g ---- /
QUESTION 6.16 (1.50) Breifly describe the function or operation of each of the following interlocks / components on the Fuel Handling Machine. Include the setpoint if applicable.
- a. Low Load Inhibit
- b. Upper Linrit
- c. Slack Cable ANSWER 6.16 (1.50)
^
a.
- b. Prevents bridge or troller movement [0.3) Setpoint 1407 lbs (0.2)
Stops [0.25]. hoist withdrawal [0.25) and clears low load inhibit interlock c. Stops downward motion of hoist [0.3) Setpoint 200 lbs. [0.2) REFERENCE VEGP Training Text, Chapter 18a, p. 16 034000K403 034000K402 034000K401 ..(KA's) J (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) I
-QUESTION NUMBER:' 6.16 .I
' STATEMENT: No tolerances'are allowed on setpoint values.
FACILITY RECOMMENDATION: Accept' answers vithin 10%.of'setpoint for full' credit.
- a. 1407 1 140 lbs C h. 200 20 lbs
REFERENCES:
N/A l
-l ,
1 1 1 d l ( l 4 l 1 i
QUESTION 6.18 (1.00) f Explain how the diesel sequencer FOR EACH UNIT is reset following-an Undervoltage condition where the diesel breaker never closes. ; i l ANSWER 6.18 - ( 1. 00 ) . '( s e Unit 1 - must pull circuit card End open supply breaker or install jumper [0.5). Unit 2 - Reset pushbutton (0.5). REFERENCE VEGP Lesson Plan LO-LP-60301-01-C, Plant Differences' 000056K121 ..(KA's) i l l i l l (***** CATEGORY B CONTINUED ON NEXT PAGE *****)
> U3 f t
QUESTION: 6.18 ' STATEMENT: The UV' sequencer on Unit 1 may be reset by'any one of: three methods.' Any one should be accepted for full credit.
?
FACILITY RECOMMENDATION: s Change Unit.I answer to: pull card or open' supply J breaker or install jumper [0.5 for any one] ~
REFERENCES:
RQ-LP-60301-01-C .! 1 1 1 1 j l. l I l e __________._m._. _ _ . _ ._ ___. -. __ _ _ _ _
1 f,. . RQ-LP-60301-01-C'. .4 l 6
- 5. s,*nc mode select switch-f l'
Uni t 2 sync mode . sel ect 'swi tch f unctions - opposite from unit 1. When the switth ! is in manual it -is functionally in auto
~
and visa versa. The name plate is to be c temporaril y repl aced. - At some point the s
, switch will be replace'with the correct {
switch. 'j
- a. unit 2 sync mode select switch turns .
in:the' opposite directionLfrom unit 1 b.' human factors problem
- 6. sequencer. LOP reset pushbutton 1
Tne unit 2 sequencers have a reset pushbutton to allow resetting of the sequences'if the DG breaker'should' fail to clase f or: any reason f ollowing a LOP. - 1
- This circuit will provide the means to 1
'i reset the sequencer if~offsite power'is restored and the diesel generator breaker has failed to close. It will allow the sequencer to be reset from a U/V condition. -
'i
- a. resetr-sequencer if DG breaker does j I
not close 'i
/q/ L5L MM
- b. allows closure of load i y breakers,.if;DG breaker did not close. 'l 3 lg/- QM&pQgjj '"L M dh % 'c d."
dhni t$ 11'aust[gsd115; dees card $"'L n5t affeictF2' hour-3
~
" t'i mmer# " *7dp. ]
/ ( d CL 3 , If the sequencer runs on an-UV condition I and the DG breaker never cloues then the l load breakers cannot be closed when power-is restored until the sequencer is reset.
When offsite power is restored and the I buss.is reenergized, the UV condition will still-be sealed in because the DG , breaker did not'close. Depressing ~the i reset will allow the.UV condition to be. > reset and deenergize the load shed relays. The load breakers may then be manually ., closed as necessary. ' l I WMW, 9/29/G8 PAGE 10 4 l
QUESTION 6.19 (1.00) Answer the following questions in reference to the LOW Steamline. Pressure Safety Injection Signal.
- a. What criteria must'be met'to cause the safety injection signal?
Give actual setpoints. . State coincidences if applicable.
- b. What crit'eria must be met to block the safety injection signal?
Give actual setpoints and state coincidences if applicable. ANSWER 6.19 (1.00)
-O a
hannels in 1/4' steam lines <= 585 psig. [0.5) b. P-11 met, 2/3 pressure channels > 1970 psig [0.50] REFERENCE VEGP Lesson Plan LO-LP-28100-03-C.-p. 5 Lasson Objective-LO-LP-28303-03-C, #1 q 013000K412 013000K101 ..(KA's) I I (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
QUESTION: 6.19 STATEMENT: Coincidence and setpoint values in answer key should be corrected. FACILITY RECOMMENDATION: Change the answer as follows:
- a. 2/3 channels in 1/4 steamlines less than or equal'to 585 psig
- b. P-11 met 2/3 pressure channels less than 1970 psig
REFERENCES:
Tech Specs 3.3-2 Vogtle Text Chapter 8a l' I ss { t . .s I
2 1 E 4 L 4 4 BS , , AE
'3 3 3
- CD 3 3 I O LM 2 2 2 P 2 2 P , ,
A 1 1 1 1 1 N e O i n I SE l T A MLL UEB m T N MNA a E INR e NAE t M IHP s U R MCO 2 2 2 / 3 2 T S N I M e E n T i S S S Y LP_ El l Ni i m N NT a C 2 G e, & A t 3
- I T
HO s 8, - CT / A 1 1 2 2 2 AAA 3 U 456 T 444 E C 4 555 L A S P 000 B A S
.L Ea q
- e. O - - -
O III OE T E MN l l L PPP R N s~
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U LA T AH a A TC it E O B F TF & O s- AA4 Y n 2 2 3 / 456 T l me ,i i no 4 3 333 E ossat 3 555 F A oent a 7n P 000 O - - - S Rianl r O III DFoie * , L_ PPP D l Ctt 7~ 1 A rrn ,ew na otg E - ) R s h6 7 "? E t airVd y 5 s-E ntl e e a g3 i9 w 4 B oSott e o&0 N rn
,C oenF l
e H0 L - s & AAA I G oo ,n ,CWe R
- - - 6I U 456 N ti r o t g n r my - I eP e4 5P -
222 E cteinnia aatt eiai n nn oo ii r, u s5 r0 &) u 4 s~ r 2 555 P 000 O - - - el aamltl . o tt s3 sI78 u O III RoWunoni) i aa sP55 e s s- L PPP y (' Is gcaCCupti ooxs t a uu e9 r0 r ,00 44 e-n eri nAt ne ,Am u i tt cc P-I PC- - r ml el eocndP t AA tP n r&II e 8, P P C
- tt odCion i
n cd e , zA ,& f Pn e
- n. 8, wCM ,rtvian ,ve in m4 i~
I T I ~)caooc N . net yrSlsea d ni l ta a n3 i9 i5A r568 u455 L ~. AAA 456 U enco .oor a u mc a0 s044 m. 1 111 555 _ L _Iy Fent neaaIt rsiD- n a oi t g t - nI s- 09 eI - - a. e P O 000 A . t , ogre ar M uo oP rPII t O III N _eppreltl o AL C( P(PP $. O L PPP
.fimencnot aromeueso (
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4 REACTCR PROTECTION SYSTEM Control System can adequately handle the load rejection and keep Tavg within its normal band. Four protection bistable status lights are on TSLB-4 TURB TRIP-RX TRIP BLOCK P-9 light is on the bypass and permissives status light box. \
- 6. F-10 Fermissive The P-10 permissive provides the f ollowing key functions 1
- a. Allows the operator to manually block the intermediate range high flux trip and C-1 rod stop.
- b. Allows the operator to manually block the low setpoint power range trip.
- c. Automatically restores intermediate range trip and C-1 rod stop when power falls below the F-10 setpoint.
- d. Autom0tically restores the low setpoint power range trip wl'en power falls below the 8 percent reactor power.
- e. Provides an input to the P-7 circuit.
- f. Serves as a backup to P-6 by preventing the operator from inadvertently reenergizing the source range high voltage and restoring the source range trip protection with power above F-10.
Inputs to the P-10 permissive circuit are from each of the four power range channels. The P-10 cetpoint 11 10 percent. When 3/4 power range .
- nar.nels drop halow 0 percent, the functions jast mer:tioned art automatically reinstated. Additionally, the operater could manually reinstate SR high voltage and trip protection. However, this is obviously undesirable with actual neutron level well above the P-6 setooint level.
Two switches, one f or Train A and one for Train B, allow blocking the IR trip and C-1 rod stop. As with the source range, both switches must be placed in BLOCK in order to achieve the desired result. Two switches, one f or Train A and one f or Train B, are provided for the purpose of blocking the low setpoint power range trip. In a fashion similar to the source and intermediate ranges,: both switches must be placed in BLOCK to inhibit the trip. Four protection bistable status lights are on TSLB-4 (P-7/P-10). , NUCLEAR AT POWER PERMISSIVE P-10 light is on the bypass and permissives status light box. >
- 7. P-11 Permissive Permissive P-11 allows manual block of low pressuri:er safety injection actuation. This permits normal clant cooldown and depressuri:ation.
When 2/3 pressuri:er pressure b1 stable sense [13ss[thahj,1970 pEig[ the low pressurizer pressure safety injection signal may be blocked. Placing Ga-58 Revision O-
n . n , ,- 9
, REACTOR PROTECTION SYSTEM the Train A and Train B pressuri:er safety injection block switches to BLOCK will now prevent low pressuri er pressure (1870 psig) from initiating safety injection. Each switen will initiate the block function in its respective protective train. If pressure rises above tne F-11 setpoint on 2/3 channels, the block automatically clears. I{ 2/3 cnannels exceed the F-11 setpoint, any shut ECCS accumulator isolation valves will automatically open if handswitch is in AUTO.
. . . _-m , _ . . . . _ . .. s m m G
Jhp.. %.Usddh:sQlstablegensede(sghan[fhe setj pressure,;the11osi st,eantyne; pressure;$Iland NSLI maysbeiblocked),(The blocking action will enable the Hi Steam Line Pressure Negative Rate MSLI). When pressure rises above setpoint, the steam line low pressure SI and MSLI will be automatically unblocked (enabled). Additionally, the Hi Steam Line Pressure Negative Rate MSLI will be automatically blocked. The three protection bistable status lights are on TSLB-3. B. P-12 Permissive This permissive automatically blocks Steam Dump System operation to ! prevent undesired cooldown. 2/4 Tavg instruments below the lo-lo setpoint (550 degrees F) actuate the P-12. When tne temperature falls < below the setpoint, all open steam dump valves will shut, and all " furthte operation will be blockee. To allow for plant cooldown less than , P-12, the operator places the Train A and Train B steam dump intsrlock switches to BYPASS INTERLOCK. With temperature less than the P-12 setpoint (550 degrees F)' ai d high steam flow, main steam line isolatiaa j will auton.htically initiate. The F-12 inter? mck allows the steamline ! pressure saicty injection ara steamline isolation signal to be blocked. 1 This allows a normal plant cooldown to be performed. 1 The three protection bistacle status lights are on TSLB-3. There is a ! common alarm, REACTOR CODLANT LOOP 1A, 15, 1C OR 1D TAVG LO-LO. There is also a LO"LO TAVO SAFETY LNJ STEAM DUMP INTLK. The P-12 light is on tre bypass and permissives light box, which energizes when 2/4 loop Tavg decrease below 550 degrees F.
- 9. P-13 Permissive Permissive P-13 senses turbine impulse chamber pressure. When 1/2 '
detectors senses the impulse pressure greater than 10 percent power, all the reactor trips associated with the P-7 are automatically unblocked. The two protection bistable status lights are on TSLB-3. If both turbine ; impulse pressure instruments dron below a pressure equivalent to 10 I per. cent reactor power, the LOW TURBINE IMPULSE PRESS. PERMISSIVE P-is ; light energi:es. l
- 10. P-14 Fermissive l
The P-14 signal protects the turbine from steam generator moisture carryover. The setpoint is 78 percent of the narrow range instrument Ba-59 Revision 0
l l l QUESTION 6.20 (1.50)
]
How is the Spent Fuel Fool designed to protect against gravity ] draining in the event of a downstream pipe break accident in the Spent { Fuel Cooling system? j i i ANSWER- 6.20 '(1.50)
- 1. SFP cooling suction is located just below the surface of the .l pool-(-4 ft). [0.5] i
- 2. !
- 3. Return pipe terminates above the fuel racks (6 ft) [0.5] and l
Anti-siphon holes just below the low water level of the pool [0.5] 1 REFERENCE VEGP Training Text, Vol. 8, Chap. 18.b., p. 18b-3 033000K403 ..(KA's) 1 I l l l l 1 (***** CATEGORY $ CONTINUED ON NEXT PAGE *****) i
- ]I I
~
. QUESTION NUMBER': ' '6.20 d 1
STATEMENT.: .'Only two'of the. answers in the key are design features to protect against gravity draining of the- ; spent fuel pool.. The two are;1'. (location of .. ] suction. piping) an'd 3. (anti-siphon holes). ,
,e FACILITY RECOMMFRDATION:
-Delete item 2 from the answer key.
REFERENCES:
Vogtle. Text Chapter 18b. . i j
.l 1
l a l
r e directly into the SFP' The SFP can be isolated from the fuel transfer canal by an air-operated, air-sealed gate supplied by the' Service' Air Systen. The gate is.. installed so that the-transfer canal say be. drained to allow maintenance of the Fuel Transfer Systes. equipment. . Water from the transfer _ canal is transferred to the SFP by temporary submersible pumps. The water is then transferred to the recycle holdup tanks for temporary storage. The borated water is later returned to the transfer canal directly by the recycle evaporator feed pumps. The Baron Recycle System supply to the transier' canal is the only direct piped water' supply to the transf er canal. The purification loop can be' isolated from the heat' removal portion of the SFPCPS to allow purification and cleanup of the RWST while SFP cooling is in service. Connections are provided to the purification loop . such that the refueling water'may be pumped fromlthe RWST through the filter and/or domineralizer and discharged either to the refueling cavity, the refueling water storage tank, spent fusi pool, or the recycle holdup tanks. Water clarity. in the refueling cavity is maintained by the.ure of a reactor 18b-2). cavati filtration unst during refueling' operations (Figure The reactor cavity filtration unit takts suction on the reactor. cavity, circulates the pater through a set of filters,'and discharges the water back into the refueling cavity.
- 3. System Flowpaths
- a. Cooling Loops When a cooling train is in operation, water flows from the spent fuel pool to the-suction of the train's SFP pump. It-is then pumped through the tube side of the train's SFP. heat exchanget and; returned;to..the SEP.
.;ProtectionlisprovidedctoVpresent?gravit'yidrsinageLofSFf/waterWTh'e7 Mcit'ionNtirairieYss
~
neioc$teMatidepth' V4~f eetTheion ih'e/ nors "waterylevel}(241(eetTab'oselthabtipYof)tdf"sel?esseabTiiitJTh s pipinjUehilth a feFeinstesW f tnbsve T ttiU 6 ele assembikesJcohainsi "*d, I e return (
^fsEiffen[5fr 'nt Psiphon151eI1 Acates. T~ 27f e'et' beiom3helloejd~er
#itT 9 ' 11evdl llinesTenser'isifTuciTWpT)Th~fandanti~sppheniheldanM@itreturne event.;ef:
" ' pipe breakwr -
A - - ' s;that[thMakstentail1$notfgravityMainithhnolfin[thQ
~at ~'
?
f wa_w 1r i
- b. Purification Loop While the~ cooling loop is in operation, a portion of the water any'be diverted through the purification loop. filter alone.or through the filter and the desineralizer. The maximum flow rate through the demineralized is approximately 100 gal / min. Higher. flow rates'could cause " channeling" in the demineralized, reducing its efficiency.
Channeling is the term used_to describe the effeet of high velocity fluid flowing into a deafneralizer resin bed. The water tunnels through the resin, ef f ectively bypassing some of the ion exchange process. Transfer canal-water can also be circulated through the same 1Bb-3' Revision.C'
l
.{
. il QUESTION 7.06. (1,00)
Which one of the following applies to GENERAL. radiation work permits?
- a. May be issued for a period of'one year.
- b. May be placed in. inactive status if not used in 3' days.
- c. Reviewed on routine basis to ensure adequate radiological controls.
d.- Entry' records maintained for each day including time of. day and
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dose received.. [ ANSWER 7.08 (1.00) a RETERENCE VEGP Procedure, 43007-C, pp 4 & 5 194001K103 ..(KA's) i i l i i 1 l l l (***** CATEGORY T CONTINUED ON NEXT PAGE *****)
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' QUESTION NUMBER: i7,06-i STATEMENT: There are two correct choices for.the question.>
FACILITY RECOMMENDATION:-
. Answer C is also correct. All RWP's are reviewed.on a routine basis to ensure adequate controls.
REFERENCES:
00930-c 5,2.7 6 5.2.12.2 l i
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, WoCEDU!.E No. LEVISloN PAGE No.
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- VEGP 00930-C 4 13 of 36 5.2.4.L- In this case, however, HP shall provide continuous coverage to ensure radiation safety practices are followed. The job shall be documented after its completion. .
5.2.4.2 Spill kits will be strategically located throughout the plant and will contain equipment and materials for (
' emergency response.
ll 5,2,5 General Radiation Work Permit i f5.2.5.1 The General RWP.is used for the performance of routine l duties such as. operator rounds, observation and j inspection, supervision, chemistry sampling, radiation. I exposure control activities, laundry operations, valve 1 lineups, work assignment evaluation,'and planned ] maintenance activities.- General RWP's will not
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be issued for work in high radiation areas, areas of significant loose. contamination, or areas requiring a , specific. job survey by Health Physics prior to entry. 1 The General RWP is called an extended RWP in.the Final- 1 Safety Analysis Report. 5.2.5.2 General RWPs specify basic radiation sefety requirements such as special training requirements, minimum protective clothing, dosimetry and respiratory protection, and &ccess restrictions, and the type or-extent of work that is authorized. 5.2.5.3 Records of entries shall be specified by name, elapsed
-- time and incremental dose, and maintained on a daily basis.
5.2.6 Specific Radiation Work Permits The Specific RWP is used.for the performance of a specific job in specific- locations or areas. The Specific RWP is active for the duration of the job and may be modified on the basis of surveys. Surveys shall be gerformed as required to support RWP authorized Wor t. S M " .HPishould rev'iew active RWPs on a~ routine basis. 5.2.8 When a job is in a high radiation area, HP supervision may determine that performing pre-job RWP surveys would not be consistent with ALARA. The RWP may then be initiated prior to surveys being taken. However. HP shall enter the areas with the work crew and perform the appropriate surveys while the job is being done. The survey data will be recorded on the RWP and radiological controls adjusted as necessary. 5.2.9 The RWP should be posted at the job location or main EP control point. 703445
," me VEGP }
, . 00930.-C- 4 14 of 36
~5.2.10 ~ Personnel authorized to make an entry on a Specific RWP, requiring a prejob briefing should be specified by name or section.
Records5.2.5.3. accordance with-step of entries will be maintained in l 5.2.11 Issuance of Radiation Work Permits 5.2.11.1 The Work Planning ' Group or the job supervisor should
- initiate an RWP by submitting a Radiation Work Permit Request similar to that shown in. Figure 1.
5.2.11.2 HP should, be notified at least 24 hours before job performance is required to allow sufficient time to-generate the RWP and aerform an ALARA review. o Exceptions to the 24 hours should be considered for planned maintenance. Planned maintenance RWP Requests should be submitted 60 days prior to the start of activities. 5.2.11.3 HP reviews the work co be performed-and specifies radiological controls to minimize exposure to personnel. 5.2.11.4 If pre-job planning is required, it shall be' performed in accordance with VIGP procedures. 5.2.11.5 HP shall approve Supervisor (SS). the RWP, and forward it to the Shift HP supervision shall Wign all RUPs. A copy of the RWF may be included in the Maintenance Work Order (MWO) package. However, the posted RWP shall be reviewed daily prior to use in order that changes to radiological conditions may be noted.
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5.2.11.6 The SS shall review the RWP for any operations or conditions that may adversely affect plant or personnel safety. Signature of the SS indicates this review. i 5.2.11.7 An individual (s) may be added as necessary to complete the work, to an existing specific RWP by the job supervisor after verifying the individual's remaining
" exposure limit and with the concurrence of HP.
supervision. 5.2.12 Termination of Radiation Work Permits 5.2.12.1 Saecific RWPs shall be terminated or suspended whenever the job is completed or cancelled, ur whenever l significant changes in radiological conditions occur. The job supervisor should inform HP of RWPs to be ' terminated. 5.2.12.2 General RWPs shall be terminated whenever significant changes in radiological conditions occur or at the end of each calendar year.~ General ~RWPs shallLbe reviewed by Health" Physics supervision 1each~ calendar quarter.
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QUESTION 7.07 (1.00) ! Place the following major action categories in proper sequence according
.to procedure 19211 " Anticipated Transient Without Trip Response." <
- a. Verify suberiticality I l b. Verify automatic actions
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- c. Check sources of positive reactivity
- d. Emergency borate ANSWER 7.07 (1.00) b, d, c, a (MDST BE'ALL CORRECT FOR ANY CREDIT)
REFERENCE VEGP Lesson Plan LO-LP-37041-02-C. pp 8 & 9 Lesson Objective LO-LP-37041-02-C, #8 & 9 Student Handout LO-HO-37041-01-C-01, Self Test 000029G012 ..(KA's,' J (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)
l l QUESTION NUMBER: : 7.07 STATEMENT: Although the question asks for, major. action- f categories for the ATWT procedureipartial credit should be given for'the. correct sequence of immediate actions. Student knowledge of the sequence-of steps in the procedure demonstrates some knowledge.of the basis. I l'ACILITY RECO<MMENDATION: Give partial credit for "b.a,'d.c"
REFERENCES:
19211-C, Revision 3: steps 1,2,3 constitute - (b); step 4' constitutes (a); step'5 constitutes (d); subsequent. action steps contain item (c). 1 t
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ACTION / EXPECTED RESPONSE _ RESPONSE NOT OBTAINED IMMEDIATE OPERATOR ACTIONS 11 ' Verify Reactor Trip: 1. Manually trip reactor. o Rod bottom lights - LIT. e Reactor tri p and bypass ,- IF reactor vill NOT trip, breakers OPEN. TREN manually opeii~ supply e Neutron flux - LOWERING. Tied breaker to INB08 (2NB08) and INB09 (2NB09): UNIT 1 _ UNIT 2 _-- o- INB08-01 2NB08-01 e INB09-01 2NB09-01 IF reactor NOT tripped, TREN manual W insert control T6Ti. 2.- Verify Turbine Tript ( a. All turbine stop valves.- SHUT a. Manually trip turbine. IF turbine will NOT trip, THEN manually run~"Eack turEine. i I i IF turbine can NOT be run liiick, i i THEN ehut main steam line Ti5Tation valves and bypass valves. i
- 3. Verify A N Pumps Running:
- a. MDA W pumps - RUNNING.
- a. Manually start MDAW pumps,
- b. TDA W pump - RUNNING IF NECESSARY. b. Manually open steam supply valve HV-5106.
- 4. Check Reactor Power: 4. Return to procedure and step o in effect, GREATER THAN 5%.
-OR-e IR SUR - GREATER THAN 0 DPM.
-- - ==v._-
$ pHOCEDURE NO. REVISION PAGE NO.
VEGP 19211-C 3 3 of 7
/- ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 5. ' Emergency Borate The RCS:
- a. Emergency borate the RCS by initiating 13009, CVCS REACTOR MAKEUP CONTROL SYSTEM.
- b. Check RCS pressure -
- b. Perform the following:
LESS THAN 2335 PSIG.
- 1) Verify PRZR PORVs and block valves open.
IF NOT THEN o'3en PRZR PURWa,nTEToc2 valves as necessary until PRZR pressure less than 2135 psig.
- 2) Verify containment ventilation isolation MLB indicators -
CORRECT FOR CVI. O 703445
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QUESTION 7.08 (2.00) List'in order of' priority, the.THREE methcds of regaining core cooling when'using FR-C.1, Response to Inadequate Core Cooling? ANSWER 7.08 (2.00) (0.5 each plus 0.5 for correct order) 1.-
- 2. Reinitiation of high pressure safety injection Rapid secondary depressurization
- 3. RCP restart and/or opening of pressurizer PORVs REFERENCE VEGP Lesson Plan, LO-LP-37061-01, p. 5 Lesson Objective, LO-LP-37061, #2 000074K311 000074K103 ..(KA's)
(***** CATEGORY ~T CONTINUED ON NEXT PAGE *****)
QUESTION: 7.08 STATEMENT.: A paraphrase of step 2 of the E0P should be accepted for answer 1. FACILITY RECOMMENDATION: Alternate acceptable wording for answer 1 is "Reinitiation of.ECCS-flow".
REFERENCES:
1921'l-C. Step-2 i i I 1 1 t, {
REVISloN - PAGE No. PROCEDURE No.' 19221-C 4 2 of 17 q VEGP RESPONSE NOT OBTAINED ACTIONrEXPECTED RESP 0'NSE 1 i i CAUTION 1 e If RWST level lowers to less than 39%,.the SI l system should be aligned for cold leg recirculation by initiating 19013-C, ES-1.3 TRANSFER TO COLD LEG l RECIRCULATION. e RHR pumps should not be run longer than 30 minutes without CCW to the RHR heat exchangers.
- 1. Verify ECCS Valve 1. Manually align valves as Alignment - PROPER necessary using Attachments INJECTION LINEUP INDICATED A, B, and C.
ON MLBs. (. 2 . ,_ Verify:ECCS Flow:
- a. CCP to BIT flow .
- a. Start pumps and align indicators - CHECK for valves as necessary FLOW. using Attachment A.
IF CCP flow NOT verified, THEN start the positive displacement pump. ,
- b. SI flow indicators - b. Start pumps and align CHECK for FLOW. valves as necessary using {
Attachment B.
- c. Start pumps and align
- c. RHR flow indicators - valves as necessary using ,
CHECK for FLOW. Act,achment C. i
- 3. Check RCP support 3. Establish support conditions conditions using using Attachment D.
Attachment D - AVAILABLE. O 7034e% i
QUESTION 7.09 (1.50) Procedure 12000-0, " Refueling Recovery", states for modes 4, 5, and 6 with the Reactor Vessel Head on'at least one Cold Overpressure Protection System shall be operable. List the THREE different systems used for Cold Overpressure Protection and provide rystem setpoints. ANSWER 7.09 (1.50)
- 1. Two PORVs (0.25] with variable lift settings per Tech. Specs [0.253
- 2. Two RER suction relief valves [0.25) with setpoints of 450 psig
(=/-34) [0.25)
- 3. RCS depressurized with vent [0.25] capable of relieving 670 rpm e 470 psig [0.25].
REFERENCE VEGP Procedure 12000-C, Refueling Recovery, p. 2 010000K403 ..(KA's) (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)
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. QUESTION: 7.09-STATEMENT: A paraphrase.of the' Tech Spec. statement for answer-1 ,l should be. accepted.
FACILITY RECOMMENDATION: Alternate acceptable wording for answer 1 is'"Two PORV's - with lift settings which vary with RCS temperature".
REFERENCES:
Tech Spec 3.4.9.3: i a 1 i
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REACTOR COOLANT SYSTEM ' COLD OVERPRESSURE PROTECTION SYSTEMS l LIMITING CONDITION FOR OPERATION a 3.4.9.3 At least,one of the following Cold Overpressure Protection Systems shall be OPERABLE: k '
- a. Two power-operated . relief valves (PORVs) with lift-settings which vary'with RCS temperature and which do not exceed the limits estab-lished in-F4 N e l
, b.
Two residual heat removal (RHR) suction relief valves each with a setpoint of 450 psig
- 3%, or
- c. The Reactor Coolant System (RCS) depressurized with an RCS vent capable of relieving at least 670 GPM water flow at 470 psig. l APPLICABILITY: MODES 4, 5, and 6 with t'he reactor vessel head on.
I ACTION:
- a. With one PORV and one RHR suction relief valve inoperable, either restore two PORVs or two RHR suction relief valves to OPERABLE status within 7 days or depressurize and vent the RCS as specified in 3.4.9.3.c, above, within the next 8 hours.
- b. With both PORVs and both RHR suction relief valves inoperable, depressurize and vent the RCS as specified in 3.4.9.3.c above, within 8 hours.
- c. In the event either the PORVs,-the RHR suction relief valves, or the RCS vent (s) are used to mitigate an RCS pressure transient, a Special R port shall be prepared and submitted to the Cosnission pursuant to Specification 6.8.2 within 30 days.. The report shall describe the circumstances initiating the transient, the effect of the PORVs, the RHR suction relief valves or RCS vent (s) on the transient, and any i corrective action necessary to prevent recurrence.
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- d. The provisions of Specification 3.0.4 are not applicable.
l l V0GTLE - UNIT 1 3/4 4-34 __ __ __- _ __ _ - _ - - l
QUESTION 7.10 (1.00)
.Name the FOUR. types of' CONTINUOUS action steps used in the, Emergency Operatina Procedures.
ANSER 7.10 - (1.00) (0.5 each)
- 1. When
- 2. Then
- 3. Control .
- 4. Maintain l
REFERENCE . VEGP Lesson Plan LO-LP-37002-02-C Lesson Objectivs LO-LP-37002-02-C, #4-194001A102 ..(KA's) (***** CATEGORY ~7 CONTINUED ON NEXT PAGE *****)
QUESTION: 7.10-STATEMENT: The answer key is incorrect regarding point distribution and the types of continuous action steps. FACILITY RECOMMENDATION: Change the points to 0.25 each. The choices should be (any 4):
- a. when/then'(or when)
- b. control
- c. maintain
- d. Check if l e. 'If/Then (or If)
REFERENCES:
LO-HO-37002-C-001 i
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' ACTION / EXPECTED RESPONSE = RESPONSENOT ~ 0BTAINED :
1.> ------> 19211^ 3a. ------> 19100, ' a.
.4. ------>L19001! '
v 18. ------>L19231 21,*......):19010-23a.--'----> 19020 ,
- 24. ------>:19030 ,
- 25. ------> 19010-
- 27. ------>o190111'
'29b.------> 190302 -
,30. ------> 19112' 131.1------> 19010-:
- 33. ------> 19030 36.:------> 19010, Transitions out a procedure generally occur to:
Respond to specific degrading conditions Move.to the correct procedure Return to a normal procedure It is important to distinguish between transitions out of a procedure and : conductingL another procedure at the same time. Position' transitions"are L indicated by "go ~to.. ." Simultaneous' procedures are generally indicated by
" initiate....." (see RNO 35 a.). - In this-instance, you would.: remain in Procedure 19000 while initiating Procedure 13710 to start an air; compressor. -,i
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;I Refer. to Procedure: 19001
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' Entry Conditions.
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- The first page of Procedure' 19001 lists the means by which Procedure 19000 l
is entered under; the heading, ENTRY CONDITIONS.. Step 4 of: Procedure' 190001 - is the only point from which Procedure 19001 can be entered. t
- All EOPs except Procedure 19000 list entry conditions (note lthat Procedure.
21910t lists its entry condition under the heading SYMPTOMS / ENTRY ;
' CONDITIONS). Some EOPs have several entry conditions b'ecause- they can be-entered from several other procedures'or steps of procedures. {
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['1 C6htinuous"ActionsCSteps j
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t Some statements in the procedure initiate actions which are then understood to be dona continuously from that point on through the rest of the procedure. These statements are called " continuous action steps."
' Continuous actions steps are not specifically identified in the EOPs (the procedure does not state, "this is a continuous action step"). Continuous action steps may be contained in NOTES, CAUTIONS, or in the procedure steps themselves. , The, following key. words generally. identify continuous action j
steps: "W1ENGTHEN," " CONTROL," " MAINTAIN," and " Check,if." Exanples of-the different types of continuous action steps are p'rovided below. The first example is from the RESPONSE NOT OBTAINED. column for step .10a, Procedure 19001. The remaining examples have been selected from other EOPs.
- WHEN, THEN Continuous Action Steps i
l Example: ! a. WHEN flux is less than' l j 10 i0 amps, ( THEN do Steps 10b and c. I Continue with Step 11. I AWHEN,THEN[continuousactionstepisineffectthroughouttheprocedure
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in~ whish it appears. If the condition specified in the WHEN-statement has I not been satisfied, continue in the procedure as directed. Frequently. ' check the condition specified in the WHEN statement while performing ' subsequent steps in the procedure. When the WHEN statement condition (s) is (are) satisfied, perform the action provided by the THEN statenent. l The continuous action of a WHEN, THEN step remains in effect when you are i l require to perform another procedure concurrently with the EOP in effect. The continuous action of a WHEN, THEN step is NOT in effect when the EOP transitions the operator to another EOP. If the WHEN, THEN step should be ' carried over to the "transitioned" EOP, it will be repeated in that procedure. i l j l l 1 l I 1-7
*' CONTROLDcontinuous action steps Examples:
,' Control Charging Flow to Maintain' PRZR level.
Control RCS Pressure and Charging Flow to Minimize. RCS-To-Secondary Leakage-You should control .the'specified parameter or device to maintain the -
-specified' condition'or p1' ant parameter; A CONTROL continuous'actionfstep~.
is in'effect throughout.the procedure unless a subsequent step instructs-Lyou to control the parameter in.some'other manner. Frequently ~ check the condition to be maintained while performing' subsequent steps in the procedure and adjust.the controls'as:necessary to maintain the condition.- When aLsubsequent step directs you'to control'the parameter.in another manner, perform the action speciffsd in the subsequent' step; the CONTROL continuous action step is.no longer in effect. The CONTR'db continuous action step remains in effect when you are-referenced to a procedure to be performed concurrently with'the EOP in effect, unless the concurrent procedure directs you to control-the j parameter in another manner,or within different limitations.- The ; continuous action of a CONTROL step is NOT.'in 'effect when the EOP .j transitions you to another EOP. If the CONTROL' step should be. carried over to the transitioned EOP, it will be repeated in.that procedure.
- MAINTAIN Continuous Action Steps i
Examples: Maintain RCS subcooling monitor indication - ri GREATER THAN (31) l ________________.. __'F. _______________ ._____________ i Maintain seal Injection 2 M T Ncn nu u act n is i e et brough h pr ed r Tha MAINTAIN statement is often used when a subsequent step or steps will- l
' directly affect the ~ parameter being maintained. - The MAINTAIN statement is i advice to make adjustments as required due to the interaction of plant- I systems. Frequently check the condition to be maintained while performing !
subsequent steps in the procedure. I t-e = _ _ _ _ - - - _ - - _- _
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. g; When a. subsequent step synytoms and diagnosis indicates that .the plant's condition requires actions that' prevent the parameter from being maintained and directs you to perform actions, perform the action specified inlthec current' step. The MAINTAIN continuous' action step is no longer in effect.
The MAINTAIN' continuous action step remains ine~ffect when' performing a concurrent EOP. The MAINTAIN continuous, action is NOT in effect'when the EOp transitions'to another EOP. . The continuous action 'will be repeated in the "transitioned" EOP, if it is required.
* "bick_if."ContinuousActionSteps- ,
Exanple: Check if RHR Punps Should Be stopped:
- a. Check RCS pressure:
- 1) Pressure - GREATER THAN 300 pSIG
- 2) Pressure - STABLE OR RISING
- b. Stop RHR punps.
1 Some " Check if" steps are continuous action steps and some are not. " Check if" steps that are continuous action steps are generally found in diagnostic loops, such as steps 20 - 30 of Procedure 19000. Some " Check if" statements are not continuous action steps because the action is performed in another step later ih the procedure. One way.to tell if a
" Check if" step is a continuous action step is to review subsequent steps-in the procedure to see if the action is repeated.
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Continuous' Action'Containe'dfwithin CAUTIG4s 'l Examples: ' q
, CAUTION '
When power is restored to any AC Emergency bus, recovery actions should continue starting with step 24. 'I CAUTION If'an SI signal is actuated during this d procedure, it should be reset to permit manual loading of equipment on an AC emergency bus. CAUTION 1 If SI actuation occurs, 19000-1, E-0 REACTOR TRIP OR SAFETY INJECTION should be performed. ' The continuous action contained'in a CAUTION is in effect throughout the procedure in which it occurs, unless the CAUTION specifically states applicability for a defined range of steps.: 1 i Frequently check the condition specified in;the CAUTION while performing .j subsequent steps in the procedure. When the condition specified in the CAUTION is satisfied or occurs, comply with~ the directions contained in the CAUTION. The continuous action contained 'in a CAUTION ~reneins in effect
. when you are referenced to a procedure to be performed concurrently with the EOP in effect.
The continuous action is NOT in effect when the EOP transitions the operator to another EOP. If the continuous actiol.'should be carried over to the "transitioned to" procedure, it will be esiterated in that procedure. . Exiting the EOPs - Summary
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l . As previously indicated, some EOPs (such as Procedure 19000) always lead to other EOPs: there are no direct exits to the UOPs from these EOPs. Other EOPs allow you to return to the UOPs if specific conditions are met. For. example, step'13 of Procedure 19001 directs you to go to UOP 12006 if you 1-10 i i
QUESTION '7.11 (1.00) i I State TWO possibl.e. conditions in which Emergency Operating Procedures can terminate. ANSWER 7.11 (1.00) (any 2 0 0.5 each)
- 1. Transition to plant operating procedures.
- 2. Transition to plant operating procedures while on RHR S cold shutdown.
- 3. On RHR system ops at cold shutdown or on either cold leg circulation or hot les recirculation.
l REFERENCE VEGP Lesson Plan LO-LP-37002-02-C, pp 12 & 13 Lesson Objective LO-LP-37002-02-C, #12 : 000029G012 ..(KA's) I l i (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)
. n ;---.- n T -A.j , y r, , I,',,* ) 'f 5.
c f' " ton *RHR~ System operation at"cSid" shutdown
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conditionsio rtomilitheiEAoidiisglirculation f or hotiles' recirculation'iwith: longer. termi redoverysactions:beingTdetermined by;the' s
" ~ ' " " ~
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individual'utilify!""'
%dM d;.w._ #
E. M- i Modes of E0P Applicability Objective as - i
- 1. E0P's' originally developed to acconanodate' emergency -
transients originating while at power. , a ' Operator action is based.upon the availability ' of equipment normally in operation in MODES 1 or 2. 1
- b. CSFST's are also based on MODES 1 and 2 as the originating condition.
- c. Example of how EOP usage was incorrectly applied while shutdown. '
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BEGIN IEN 83.030 , 5
- 1) Inadvertent SI at Summer (W-PWR) during ;
cooldown: 340*F, 1000 psig.
- 2) E0P guideline at the time did not allow SI termination until RCS pressure was j greater than 2000 psig. i
- 3) Operators allowed RCS to pressurize rather than terminating.SI knowing as they should that the E0Ps were based on Modes 1, 2, 3 operation.
- 4) Exceeded 1600 paid across S/G U-tubes.
- 5) Same thing happened the next day while plant was water-solid.
- 6) Operators terminated SI prior to-lifting PORVs at 750 psig (COPS)
- 7) VEGP E0Ps allow SI termination if RCS is
" stable or rising" and checks subcooling as well as other criteria.
- 8) RCS leakage AOP direct operator action when END IEN 83.030 shutdown /cooldown.
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QUESTION 7.12 (1.00) Accoraing to procedure 19000, "E-0 Reactqr Trip and Safety Injection'", what action should you take to restore power to.the AC Emergency Busses? ANSWER 7.12 - (1.00)
- 1. Start and Load diesel OR
- 2. Attempt to re-energize bus from RAT. .
REFERENCE VEGP Lesson Plan DD-LP-37011-02-C Lesson Objective LO-LP-37011-02-C, #4 Student Handout LO-HO-37031-00-C-002, p. 7, self test 000007G012 ..(KA's) 1 l l l l I ( ***** CATEGORY -7 CONTINUED ON NEXT PAGE *****) -mm
n, - - - - _ _ ._ .
.i QUESTION: 7.12 -
1 STATEMENT: The question ~ asks' '.'according to 19,000 . . . . what ' action . U
.should you take to restore power to AC emergency busses" The procedure does not tell-how to do it. 3 l
. . l FACILITY RECOMMENDATION:
Delete',the question.
REFERENCES:
19000 Step'3
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PROCEDURE No. REVISloN PAGE No. VEGP 19000-C 5- 2 of 26
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ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE Foldout page should be continuously monitored and applicable actions taken. IMMEDIATE OPERATOR ACTIONS
- 1. Verify Reactor Trip: 1. Manually trip. reactor.
e Rod bottom lights - LIT. IF reactor will.NOT trip. TREN manually open supply e Reactor tri p and bypass TeTd breakers to: breakers OPEN. UNIT 1 UNIT 2-e Neutron flux - LOWERING. e 1NB08 e 2NB08 e 1NB09 e 2NB09 IF reactor NOT tripped. THEN go to T!Plll-C, FR-S.1 l
. 0, IEEPONSE TO NUCLEAR POWER GENERATION /A W T.
. 2. Verify Turbine Trip:
- a. All turbine stop valves - a. Manually trip turbine.
SHUT.
- 3. Verify Power To AC Emergency Busses:
- a. AC Emergency Busses - a. Tryito" restore power to at AT LEAST ONE ENERGIZED: ::leastTone AC Emergency
-Bus.
UNIT 1 UNIT 2 IF power can NOT be e 1AA02 e 2AA02 e 1BA03 e restored to at least one 2BA03 AC emergency bus, THEN go to 19100-C, E0.0 LOSS OF ALL AC POWER.
- b. AC Emergency Busses - b. Try to restore power:to ALL ENERGIZED. 'de-energized AC Emergency,
" Bus'while~ continuing with
( Step 4. ' t k
'03486
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l 4 QUESTION 7.14 (1.50) 1 A given radiation area is identified as requiring a lock, but due to the { physical dimensions of the area and obstructions, a lock can NOT be { used. WHAT are THREE alternative methods that can be used to satisfy the requirements for locking the area? ANSWER 7.14. (1.50) i
- 1. Rope the area off. !
- 2. Flashing light installed at area boundary.
- 3. Posted as a HIGH radiation area.
REFERENCE VEGP Procedure 43005-C, p. 4 194001K103 ..(KA's) l l (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) l
QUESTION: 7.14. STATEMENT: .The answer key provides.three (3). elements of one' method' to satisfy the requirement for locking the' area.(per Technical, Specification 6.11). FACILITY RECOMMENDATION:, Other methods which demonstrate understanding'of ALARA
.principics should not cause students to. lose credit.
REFERENCES:
Tech Spec 6.11 and 00930-C l
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ADMINISTRATIVE CONTROLS 6.11 HIGH RADIATION AREA (Continued)
- b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is recalved. 1 Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them; or i
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' c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control.over the activities within the area and shall perform l
periodic radiation surveillance at the frequency specified by the , Health Physics Superintendent in the RWP. 6,11.2 In ' addition to the requirements of Specification 6.11.1, areas accessible to personnel with radiation levels greater than 1000 mR/h at 45 cm (18 in.) from the radiation' source or from any surface which the radiation penetrates shall be maintained under the administrative control of the shift Forem duty and/or health physics supervision. Doors shall remain locked except during periods of access by personne1' under an approved RWP which shall specify the-dose rate levels in the immediate work areas and the maximum allowable stay
;O time for individuals in that area. -In lieu of. the stay time specification of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures performed within to provide positive exposure control over the activities being the area.
For individual high radiation areas-accessible to personnel with radiation levels of- greater than 1000 mR/h that are located within large areas, such as PWR containment, where no enclosure exists for purposes of locking, and where no,enclosurecanbereasonablyconstructedaroundtheindivjdyalarea,that f ridividualtareaf shall .be. barricaded, conspicuously. posted [and'a- flashing light shall: be activiated as' a warning devicer " ~ 6.12 PROCESS CONTROL PROGRAM (PCP) i 6.1?. 1 The PCP shall be approved by the Commission prior to implementation. i 6.12.2 Licensee-initiated changes to the PCP: I i a. Shall be submitted to the Commission in the Semiannual Radioactive Effluent made. Release Report for the 3. lod in which the change (s) was j This submittal shall contain:
- 1) Sufficiently detailed information to totally support the rationale for the change without benefit of additional or !
supplemental information;
- 2) A determination that the change did not reduce the over'all conformance of the solidified waste product to existing criteria for solid wastes; and V0GTLE - UNIT 1 6-23
l '. FIEVISloN PAGE No.
- 00930-C 4 8 of 36 geogENo.
gp c i Posting boundaries are determined by radiation surveys. g 5.1.3 3- These surveys will be made at a distance of 18 inches from the surface of sources, unless the whole body will be exposed at a closer distance during a specific work task'. In this case, surveys will be made at contact with the source. _ High Radiation Area 5.1 4 Each high radiation area shall be conspicuously posted 5.1 4.1 with a sign, or signs, bearing the radiation symbol and the words: CAUTION HP Escort required for entry High Ra61ation Area 5.1.4.2 High radiation area boundaries will be determined by surveys as specified in 5.1.3.3. i 5.1.4.3 High radiation area doors that are equipped with security access control readers will be locked unless ! individual access is authorized by RWP. Tailgating is strictly prohibited. Individuals caught tailgating will receive disciplinary action. 5.1.4.4 Individuals who have successfully completed HP self-monitoring training are qualified to provide their own escort in high radiatior areas. 5.1.5 Locked High Radiation Area 5.1.5.1 Each locked high radiation area shall be locked in such a manner as to prevent unauthorized entry and not prevent exit from the area. Those areas where no enclosure can be reasonably < constructed; shall:hetroped" off4 and'atflashing light installed *as"a warning' device. Areas meeting the above requirements shall be conspicuously.postedewith assign, or signs, bearing the radiation symbol and the words: DANGER Keep Out High Radiation Area HP Escort Required for entry NOTE Only Health Physics personnel are qualified to perform escort duties in Locked High Radiation Areas. 1
4 QUESTION 7.15 (1.50) Name THREE types of areas where a Specific Radiation Work Permit is-required for entry. ANSWER 7.15 (1.50)
- 1. Airborne Radioactivity Area.
- 2. High Radiation Area.
- 3. Lock High Radiation Area.
REFERENCE- l 1 VEGP Procedure.43500-C, pp 3 & 4 194001K103 ..(KA's) i I I J l (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)
=. , . . . - . _ _ _ _ - .
)
j, QUESTION: 7.15 .! STATEMENT: Another area. requiring a specific RWP is:a loose '1 I contamination area. 1
)
FACILITY RECOMMENDATION:
<" Accept' loose contaminaiton area as'an additional correct ;
answer. ! I
REFERENCES:
.00930-C Revision 4 Page 13 l l
i .i l I l l
)
1 i i s y I. ' i
,H mmpa
-VEGP. 00930-CD 4
;\' 13.of 36
. c u> l 5. 2. 4.1-L In thisJcase', however, HP'shall' provide continuous-coverage to. ensure. radiation safety practices are followed. . The job shall be7 documented afterJits:
completion. : 5.2.~4.2
- Spill kits'will be strategically located'throughout the: '
plantiand.willscontain equipment,and materials,for emergency response. 5.2.5 General Radiation' Work Permit-5.2.5.1 The General'RWPiisiused for the performance of. routine duties such;as. operator rounds. observation.and. inspection, supervision, chemistry sampling, radiation'
- exposure control 1 activities, laundry ~ operations,. valve lineups, work: assignment evaluation, and planned-maintenance activities'. General.RWP's willJnot be:
si issued'for work in high' radiation areas, areas of s)gnificantTlooseTcontamination', or areas requiring;a. E a ecific j ob" survey' by 'He'alth; Physics : prior- to entry. -l The General'RWP
-Safety is called an extended RWP in~the' Final Analysis Report.
5.2.5.2 General RWPs specify basic. radiation; safety. requirements ~such:as special training ~ requirements,; minimum. protective clothing,idosimetry and respiratory. ; protection,1and access restrictions,.andsthe type:or 1' extent of work that is: authorized. ' 5.2.5.3 Records of entries'shall be~specified by name,' elapsed time and incremental dose, and maintained on a daily basis. ' 5.2.6 Specific Radiation Work Permits! l The Specific RWP is used for the performanc'e of a specific job in specific locations: or areas. - The Specific' RWP is active -for the- duration: of the job ~ and. e may be modified:on the basis of surveys. Surveys'shall i i ? be performed as required to support RWP authorized work. .; L 5.2.7 ^ HP should review active.RWPs on a: routine' basis.
- 5. 2. 8 , 1When a job .is in a. highl radiation area, HP supervision i
' may determine. that performing. pre-job RWP surveys would not be consistent < with; ALARA. he RWP ma i
' initiated prior to surveysubeing*taken. y However, then be HP j
~
i shall enter the-areas.with the' work' crew and perform the- i appropriate - surveys' while Lthe -j ob: is -being 'done. - ! The survey data will be recorded on the RWP'and radiological controls adjusted: as necessary. 1 5.'2. 9 The RWP should be posted at the job location or main EP control point.
~
' i l
( .. '.
I QUESTION 7.16 (2.00) j List the FOUR criteria which must be satisfied before procedure 10011, i
" Safety Injection Termination", is entered.
ANSWER 7.16 , (2.00) (0.5 each) I
- 1. RCS Subcooled.
- 2. RCS pressure stable or rising.
- 3. Adequate Secondary Heat Sink must exist.
- 4. Pressurizer level on scale (span).
I REFERENCE - , VEGP Lesson Plan LO-LP-37022-02 Lesson Objective LO-LP-37022-02, #1 Student Handout LO-HO-37022-01-001, Self Test 1
'I l
l l (***** CATEGORY -7 cot!TINUED ON NEXT PAGE *****)
QUESTION: 7.16
~
STATEMENT: Specific criteria to terminate SI should be accepted as equivalent answers. FACILITY RECOMMENDATION: Accept paraphrased steps from'the'E0P's.'-
REFERENCES:
19010-C page 6'of 16 i i
,s
-\
, (' ,
2-_.._ _
PROCEDURE No. REVISloM ' PAGE No. VEGP 19010-C 7 6 of 16 ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 6. Check If ECCS Flow Should Be Reduced:
- a. RCS s,ubcooling monitor a. DO NOT STOP ECCS PUMPS.
indication.- GREATER Go to Step 8. THAN 24*F [38'F FOR ADVERSE SMT).
- b. Secondary heat sink: b. IF neither condition sitisfied, o Total feed flow to THEN DO NOT STOP ECCS intact SGs GREATER PUMPS.
THAN 570 GPM. Go to Step 8.
-OR-e Narrow range IcVel in at least one intact SG - GREATER THAN 5% [27% FOR ADVERSE CNMT).
- c. RCS pressure - STABLE OR c..DO NOT STOP ECCS PUMPS.
RISING. Go to Step 8.
- d. PRZR level - GREATER d. DO NOT STOP ECCS PUMPS.
THAN 9% [36% FOR Try to stabilize RCS ADVERSE CNMT). pressure with normal PRZR spray. WHEN PRZR level greater than 9% or [36% FOR ADVERSE CNMT), i THEN do Step 7. ' Continue to Step 8.
- 7. Go to 19011-C, ES-1.1 SI TERMINATION.
ws
QUESTION 7.17 (1.00) Why must the Hydrogen Monitor isolation valves remain closed except during hydrogen monitor operation, according to the System Operating Procedure, 13130-1, " Post-Accident Hydrogen Control?" ANSWER 7.17 (1.00) To ensure containment integrity is maintained. REFERENCE VEGP System Operating Procedure, 13130-1, p. 4 028000A101 ..(KA's) 1 i l i (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)
QUESTION:. 7.17' STATEMENT: Hydrogen monitor isolation' valves should remain closed' except during operation because these-valves are required. to be closed during' accident conditions and do not receive a close signal ~on CIA. FACILITY RECOMMENDATION: Accept simular wording if student' displays that Containment Integrity is the concern without actually saying CONTAINMENT INTEGRITY.
REFERENCES:
, 1. T.S. 1.7 definition offCNMT integrity.-
2.- FSAR Table 6.2.4-1 sheet 6 of 10 i
)
i
)
i i I 1 I
.l
)
i l
c .' DEFINITIONS
-l
' CONTAINMENT INTEGRITY 1.7 CONTAINMENT-INTEGRITY.~shalbexist/when:
- a. -All penetrations required to be closed during accident conditions are either:
- 1) 'C'apable of being closed by~an OPERABLE containment automatic-isolation ' valve system,;org
- 2) . CloEed"bf m'anualTvalves, blind' flanges, or deactivated automatic valves secured in their closed positions.
- b. All equipment-hatches are closed and sealed, c.
Each' air lock is in compliance with the requirements of Specification 3.6.1.3,. , d. The containment 3.6.1.2, and leakage rates are within the limits of Specification .i e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE. . CONTROLLED LEAKAGE 1.8 coolant _ CONTROLLED pump seals. LEAKAGE shall be that seal water flow supplied to the. reactor CORE ALTERATIONS 1.9 CORE ALTERATIONS shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head remoyed.and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe conservative position. DOSE EQUIVALENT I-131 1,10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic ' mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid i dose conversion factors used for this calculation shall be those listed in Table E-7 of NRC Regulatory Guide 1.109, Revision 1, October 1977. E - AVERAGE DISINTEGRATION ENERGY 1.11 I shall'be the average, weighted in proportion to the concentration of each radionuclides in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration ~in MeV, for the isotopes with half lives greater than 14 minutes, making up at'least 95% of the total non-iodine activity in the coolant. V0GTLE - UNIT 1 1-2 _ - _ - _ _ _ - _ _ _ _ - - - - - _ - - - - - - - - - - - - - I
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QUESTION ~ 7.18 (1.00) .i During a reactor startup using procedure 12003-C. " Reactor Startup", there is one PRECAUTION which requires the termination of rod withdrawal , if a certain condition is expected to occur. What is the condition that will_ require the termination of rod withdrawal? .. . i s ! ANSWER 7.18 (1.00) If' criticality is projected to. occur below the rod insertion-limit. I REFERENCE VEGP Procedure-12003-C, p. 2 1 0010000010 001000G005 ..(KA's) 1 i I I I (***** CATEGORY Y CONTINUED ON NEXT PAGE *****)
. QUESTION: 7.18 STATEMENT: Although the_ question asks for a specific PRECAUTION which requires termination'of rod withdrawal, 12003-C, also contains a LIMITATION.-to terminate rod withdrawal if criticality is projected to occur _ above the rod' withdrawal.limic_.(12003-C; 2.2.9). Students are expected-to treat PRECAUTIONS &' LIMITATIONS in'a similar manner.
FACILITY RECOMMENDATION: The LIMITATION should be an acceptable alternate answer. l
REFERENCES:
12003-C 2.2.9 l l l , 1 1, 4 I i 3 _-_____m___ _ _. __ __m. _ __ _ _ _ _
VW W
~ == ~ -Q 7
PROCEQUAE No,- REVISION FAGE NO.
,/
VEGP 12003-C 11 -- ' 2 of 21. 2.1.7 Rod withdrawal shall be terminated if criticality is projected to occur below the rod insertion limit. 2.1.8 Welding in the vicinity of the NI cables may result in erratic noise spikes, so should be avoided during startup if possible. 2.2 LIMITATIONS 2.2.1 While in Modes-1 and .2, shutdown margin shall be.. greater than or equal to 1.3% delta K (1300 pcm). (Technical specification.3.1.1.1). 2.2.2 While in modes 3 and 4, shutdown margin shall be
. greater than or. equal to the limit specified Technical Specification 3.1.1.24-Figure 3.1-1.
2.2.3 Criticality shall not be achieved with the control banks _ positioned below the limitations shown~in Figure-3.1-1 of Technical Specification 3.1.3.6.- 2.2.4 Shutdown and. control rods shall be operable and . positioned within *12' steps of their Demand Positiott Indication System. (Technical Specification 3.1.3.2) 2.2.5 4 Prior to achieving criticality,.the RCS-lovest loop-O Tavs shall be greater than or equal to 551*F. Technical Specification 3.1.1.4) 2.2.6 The maximum sustained'startup rate (SUR) should not exceed 1.0 decade per minute (DPM). 2."2.7 All shutdown rods shall be fully withdrawn prior to entering Mode 2. (Technical Specification 3.1.3.5) 2.2.8 The boron concentration in the pressurizer should.not l be different from the Reactor Coolant System (RCS) by more than 50 ppa. Pressurizer backup heaters may be energized as necessary to equalize the boron . concentration. ' 2.2.9 To ensure moderator temperature coefficient remains h less positive than zero, Trod 7withdrsusl'shallibe L g ,. N 1 n 3~.1.1.3) :1 O PC34a5 ca.
QUESTION 7.20 (2.00) As the oncoming Shift Supervisor, you see the following conditions existing while performing procedure 12001-C, " Unit Heatup to Hot Standby (Mode 5 to Mode 4)." l A RCS heatup of 49 F/hr. is in progress with temperature of 190 F and pressure is at 350 psig. Reactor Coolant Pump number 1 is operating. Samples of RCS and Pzr boron indicate'1200 and 1240 ppm respectively. A boron dilution is in progress to reduce concentration to 1050 ppm for the Estimated Critical Concentration. Reactor Trip breakers are shut and all shutdown and control banks are withdrawn two steps. Both oxygen and lithium are within limits-for plant temperature. INDICATE AND EXPLAIN any abnormal or irregular conditions represented in-the above paragraph according to the Precautions and Limitations of the procedure 12001-C. ANSWER 7.20 (2.00)
- 1. RCP #4 should be operating [1.0]
- 2. Shutdown rod banks should be fully withdrawn - not at two steps.
[1.0) REFERENCE VEGP Procedure 12001-C, Unit Heatup to Hot Standby, pp 1-11 001000G010 ..(KA's) (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)
lj
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i l QUESTION: 7.20 .) l STATEMENT: Answer 2'is incorrect and an additional correct answer ,i may be elicited. ]! i
- a. Withdrawal of the shutdown banks is per USS- ;
discretion.(step A4.1.16). I
- b. If a positive MTC is assumed, then heatup and dilution should not take place simultaneously since :
. both would add positive reactivity. (step'2.1.4);-
]
FACILITY RECOMMENDATION: Accept answer 1 as a complete answer. Do not deduct credit for b. (above)
REFERENCES:
12001-C' l .. s
)
I i I I i l l 1 l a k LL__ - : _' . ._ ___ _ _ _ _ _ . _. _ _ _ _ _ _ . __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ - . -_3
e
) ^ ' * *'
j' j Vogtle Sectric Generating Plant ' Proceoure NA 12001-C. / NUCLEAR OPERATIONS
%,, go, COMON '
Unit Georgia Pbwer e a g. g a UNIT NO. k: DATE /- / f ( UNIT HEATUP TO HOT SHUTDOWN (MODE 5 TO MODE 4) MANUAL SET 1.0 PURPOSE NO. E This prea.edure provides instructions for taking the unit frem Cold Shutdown (Mode 5) with temperature between 80*F and 130'F, to Hot Shutdown (Mode 4) with temperature less than 350'F. 2.0 PRECAUTIONS AND LIMITATIONS 2.1 PRECAUTIONS 2.1.1 The Residual Heat Removal (RHR) pump suction line should not be isolated from the Reactor Coolant. System (RCS) unless there is a steam bubble in the Pressurizer. 2.1.2 Whenever the RCS temperature is above 160'F, at least one Reactor Coolant Pump (RCP) should be in o preferably Pump 4 to ensure best spray flow. peration, 2.1.3 Thc RCS pressure and temperature shall.be maintained within the operating region of Figure 1. 2.1.4 TDo not ?addi pas itive ;. re activity tby;nore; tf5ni onM Tcontrolled: Tsuberitical'.cae thod 4 f a t "a 'timet while T thelsac toEii [ 2.1.5 If the count rate on either source range channel increases unexpectedly by a factor of two or more during any operation, the operation must be suspended immediately until a satisfactory evaluation of the b j situation has been made. 2.1.6 Criticality must be anticipated any time the rods are being withdrawn or when dilution operations are in progress. 2.1.7 If the Reactor Trip Breakers are closed and the RCS-
..- temperature is to be changed more than 50'F, all rods o should be withdrawn at least 2 steps. This is to Q prevent thennal lockup of the rods.
F33101A MCE130 Q______-_--_-___----_-__.---------
?
PROCEDURE No. REvlslON . -- PARE No. . 10 of 40f
~
.10-VEGP- 12001-C .
INITIALS A4.1.9: PERFORM:a~ general RCS leak inspection. LOOK specifically at the~ vessel head,
.SG and Pressurizer manways, RCP flanges,; ,
and pressurizer safeties and PORVs.- . LLOOK also for water on the floor or sumps and bulging cr. wetness of
; insulation.
A4.1.10 INITIATE Lowering Steam Generator p levels for unit atartup per;13601, l _ Steam Generator ' And Main Steam System
' Operation".
A4.1.11 ' INITIATE placing:the CST in' operation per 13625, " Condensate Storage And < Degasifier System".. , A4.'1.'12 INITIATE placing the AFW System in standby _per 13610 " Auxiliary i Feedwater System". A4.1.13 INITIATE ' aligning the Main Steam System per 11601, " Steam Generator And Main Steam Alignment For Startup And Normal. Operation". . A4.1.14 INITIATE activitiesefor establishing L containment integrity.per Substep A4.2.5. A4.1.15 :If backfeeding through the: Main j Transformer, then NOTIFY Maintenance and INITIATE restoring normal service through the RATS. l A4.1.16.'gkb@discretioiitNb$E5bSkw
'VITEDRAWshutidown banksLbyaperfora $.w. ng
.thelfollowing: ,
- a. . PLACE at least one.CRDM Cooling Fan in operation per 13120,.
" Containment Building' Cooling ;
-System", ;
r a
- i. b. PLACE a't least one.-Control Rod .l l Drive M-G set and the: Digital '
- l. Rod Position Indication System l l
in operation per 13502,'" Control. -l
- Rod Drive And Position Indication
*~ !
System",-
]
3 e
' 703445 ,
f' , L. .
i QUESTION 7.21 (2.00) Identify the Technical Specifications, by parameter name, limiting value and ti.me limit for those Technical Specifications where ACTION is to be taken in IISS THAN ONE HOUR af ter exceeding the specification. I ANSWER 7.21 (2.00) 1. Reactor Coolant System Pressure > 2735 psig [0.5), be within limits in FIVE minutes [0.5]. i j 2. Tava < 551.F [0.5), be within limits in 15 minutes [0.5). REFERENCE VEGP Technical Specification, 2.1.2 and 3.1.1.4 002010G005 ..(KA's) 4 i
)
i (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)
l
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)
-QUESTION: 7.21 STATEMENT: 'The question'is much'too general. Action statements -l'ess :
than one (1); hour number about 73.: The interpretation of-parameter used in the key'is too' restrictive.- A. ._ parameter can be any' quantity;whose value. varies with'the' j circumstances. It does not need to be directly readable:
~
1 on'the main contro1 board:(i.e., it may.be a calculated-value, the number of operable channels, etc) j 1 FACILITY RECOMMENDATION: i
.l Delete the question. j
.)
. 1
REFERENCES:
-Technical Specifications (samples attached).
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I 3/4.1 REACTIVITY' CONTROL SYSTEMS-3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - MODES 1 AND 2 l 4 LIMITING CONDITION FOR OPERATION i 3.1.1.1 TheSHUTDOWNMARGIMshallbegreaterthanorequalto(3%lAk/k,. APPLICABILITY: MODES I and'2*. ACTION:
! With the SHUTDOWN MARGIN 1ess than 1.3% Ak/k.demedfe At' lf initiate and~ continue l boration at greater than'or equal to 30 gpm of~a solution containing greater . ! than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN l MARGIN is restored.
l . ,l l SURVEILLANCE REQUIREMENTS
) \ ! .' 4.1.1.1.1 The SHUTOOWN MARGIN shall be determined to be greater than or equal I to 1.3% Ak/k:
i
- a. Within 1 hour after detection of an inoperable control rod (s) and at least once per 12 hours thereafter while the rod (s) is inoperable. !
i If the inoperable control rod is immovable or untrippable, the above
! required SHUTDOWN MARGIN shall be verified acceptable with an increasecI ; allowance for the withdrawn worth of the immovable or untrippable control rod (s);
i
- b. When in MODE 1 or MODE 2 with K,ff greater than or equal to 1 at l 1 east once per 12 hours by verifying that control bank withdrawai is
{ within the limits of Specification 3.1.3.6;
- c. With K,ff less than 1, within 4 hours prior to achieving reactor i criticality by verifying that the predicted critical control rod
; position is within'the limits of Specification 3.1.3.6; and l! d. Prior to initial operation above 5% RATED THERMAL POWER after each
! fuel loading, by consideration of the factors below, with the control banks at the maximum insertion limit of Specification 3.1.3.6:
I
*See Special Test Exceptions Specification 3.10.1.
~
V0GTLE - UNIT 1 3/4 1-1
3 3/4.3 INSTRUMENTATION- i
]
3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING' CONDITION FOR OPERATION
, .,, -wmmy 3.3.1' As a, minimum, the Resstor.iTripisystemdm , y m%; instrumentation ~ channels and intg locks of Table 3.3-1 shal1~be CPERABLE with'their Trip Setpoints set consistent with the values shown in the Trip Setpoint column of Table 2.2-1 and with response times within.their limit value.
. APPLICABILITY: -As shown in Table 3.3-1.' h
'1 ACTION: )
q As shown in Table 3.3-1. -l l i SURVEILLANCE REQUIREMENTS 4.3.1.1 Each Reactor Trip System instrumentation channel and_ interlock and -) the automatic trip logic shall be demonstrated OPERABLE by the performance of l the Reactor Trip System Instrumentation Surveillance Requirements specified in Table'4.3-1. .
.I 4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE-TIME of each Reactor trip function i
shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one' train such that both trains are tested at 1 ~ 1 east once per 36 months e d one channel per function such that all channels y are tested at least.;;;e every N times 18 months where N~is the total number a of redundant channels in a specific Reactor trip function as shown in the
" Total No. of Channels" column of Table 3.3-1.
i ] l i e 4 V0GTLE.- UNIT 1 3/4 3-1 O______-_-___. ,
N b b b b O 1 1 2 2 2 2 I 1 J ~ 33 T C A E a L' 5 BS AE -', CD a . IO ' 4 - 2 -. LM 2 , 2 2 2 P , ., L P ,a
,d , ,
A 1 3 l' 1 1 1 d 12
~
. SE MLL UEB MNA 2 2 3 3 3 3 INR .
22 _ NAE _ N IHP O MCO _ I - - _ T A T _ N - E S M LP U EI R NR T NT 1 1 2 2 2 _ 1 S A - 2 11
- N HO 3 I CT 3 M
_ E _ E T L S B Y A S S T .L P OE I NN _ R N T LA 2 2 4 4 4 AH 1 4 22 R TC O O T TF C O A - E ,' R x u l
, F
)
, n G x x x o x &
u , u r uaD l C) u , t l F &C l C) l C) u F 8 B& F &C F &C e }2 n 28 8& 8& N n 3 p n 28 n n i o 44 r 04 o re 04 44 o : 4428 , o o 0 r 0 _ r t 00 t tt 00 re 04 e g i t - T u - 0 n t ua - 0 tt 00 t u I ua - 0 n e I - i n eR I - eR a a e N r N NI o o i N NI I - r N o p NI R e t , ,N t p , e v ,N N,v'e ,N. E) p 1, E c e &G Op e e C , e t ei C T a ei C , t & g &C a D, D, rt u i h e S e gt &C gt &C 0, I N U R n 8& a 18 R 43 h S g wR o 43 ni as 1B 8& na ag 18 R e 43 8& i 8 d 58 wa e 36 ot er a R 1 l 3 8 l 04 i o P 04 L A a u r 00 H L r . 00 r N 0400 m 03 PS r 00 e 0 5 N n e - 0 w I - eh - 0 eh - 0 e - 0 ~cr -0 I T O M a o NI P (N a b PH (N wg I - oi NI wg I - oi NI PH (N t I n NI I - (N ab
. . dN S
I _ ( C U N
$r.
- F 1 2 , 3 4 5 6 4Qrn , e34.- H y* Y " L 1;; ll
+ ,
3
~
y
~
]
,1 5
TABLE 3.3-1 (Continued) 1 d ACTION STATEMENTS (Continued): ,
-l ' -k ACTION 3 - With the number of channels OPERABLE one lessithan th'e Minimum- <
j Channels OPERABLE requirement and with the THERMAL POWER level': j 1 [ a.- Below the P-6 (Intermediate Range Neutron Flux-Interlock)
'Setpoint restore the inoperable channel to OPERABLE' i , status prior to" increasing THERMAL POWER above the P-6'.
Setpoint, and' i
, b. Above the P-6 (IntermediateLRange NeutroniFlux Interlock).
4 Setpoint but below 10%.of RATED THERMAL POWER,. restore the' .{o inoperable channel to OPERABLE: status prior to increasing 'l
. THERMAL POWER above 10% of RATED THERMAL POWER.' d n... m.gr w , m .y 4 ACTIONL4x-iWith4theinumber;m.. mm m
. g., , m% i Am. m
..._m m % .ofm0PERABLEichannels;onesless thantthe1 Minimum-d s
's
,Channelst0PERABLEureguirement l epositivej reactivitKchanges.' hsuspendfalixoperationsiinvolvisig 3
)
~
1 JCTION15Ma.- With~the number of OPERABLE channels-one less.than the Mini - 1 mum Channels OPERABLE requirement, restore:the inoperable I' channel to 0PERABLE status within 48 hours or.open the Reac- " tor Trip System breakers, ' suspend all operations involving positive reactivity changes and verify valves 1208-04-175,
, .1208-U4-177, 1208-04-183, and 1208-U4-176.are closed and secured in position.within the next. hour. - l r
m - m e.e w e m.u a A j b T XWithFno chann. ..- --_.-els? OPERABLE gopenitheiReactorrTrip y. Breakers ?suspankal t operations sinv'ol ving posi tive' ' ireactivityJchanges, and verify compliance.with the . SHUTDOWN MARGIN requirements'of Specification 3.1.1.1 or _j 3.1.1.2,.as applicable, within 11 hour and'every 12 hours ' thereafter. -Verify valves 1208-U4-175; 1208-04-177, 1208-U4-183, and 1208-U4-176 are closed and secured in-position within 4' hours and verified to be closed and secured in position every 14 days. . F 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- U
- a. The inoperable channel is placed in the. tripped condition ;
within 6 hours,'and .
- b. The Minimus' Channels OPERABLE requirement is met; however, 1 the inoperable channel may be bypassed for up to_4 hours for surveillance testing.of.other channels per:
Specification 4.3.1.1. ACTION 7 - (Not used) s 1, . . j 1- ,V0GTLE -' UNIT 1. 3/4 3-7 '
.q I
REFUELING OPERATIONS 4 LOW WATER' LEVEL i LIMITING CONDITION FOR OPERATION { i 3.9.8.2 Twf indepe6 dent /resiSsak heat'removalc'(RHR);tra{ns shall be OPERABLE, and at least~on'~RHR' e train shall be in" operation.*" .j 1 APPLICABILITY: MODE 6, when the water level above the top of the reactor lj vessel flange is.less than 23 feet. . j ACTION:
- a. WithlitssthantherequiredRHRtrainsOPERABLE,imme'diatnipliniifale corrective action to return the required RHR train's to'0PERABLE 1 status, or to establish greater than or equal to 23 feet of water
-above the reactor vessel flange, as soon as possible. i b;
With no RHR train in operation,~ suspendfalEoperationsiinvol.ving a reduction' in" boro'n"co'ncentration"of"the ~ Reactor Coolant ^ System and immediately initiate"correctiveiaction to'? return the required RHR { , train 'to" operation. ' IClose"all' containmerit penetrations providing direct access from the~ containment atmosphere to.the outside atmosphere within 4 hours. SURVEILLANCE REQUIREMENTS 4.9.8.2 At least one RHR train shall be verified in o I reactor coolant at a flow rate (FIC-0618A, FIC-0619A)ofperation greaterand thancirculating or equal to 3000 gpm at least once per 12 hours. -.
- Prior to initial criticality, the RHR train may be removed from operation for up to I hour per 2-hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor vessel hot legs.
V0GTLE - UNIT 1 3/4 9-9
N o. a,0ctogE 10000.C escu " ' " ~ 7 14 of 26 / ll 3.2.3 Notification of Absences Anyone expecting to be late or unable to report for shift duty at the scheduled time shall, at the earliest possible time, inform the OSOS or USS. l 3.2.4 , Call Out Authority The OSOS is authorized to call out anyone required for the safe plant operation, per Plant Administrative Procedure 00006-C, " Recall Of Off Duty Personnel". 3;3l .'
,, *1GENERAf WORK?pRICTiCES5R x e T ~ f~T . [-
~^
All personnel assigned to shift operations shall: f t
# a. Be aware'that the primary responsibility of the fi I o?erating shift is to assure the safe operation of 0
the plant under all conditions, f b. Protect plant personnel, the health and safety of g the public and plant equipment, i j c. Conduct major plant operations in accordance with j approved written procedures, ,
)
! d. Be attentive to the condition of the plant at all
! times. They must be alert to ensure that the
[ plant is operating safely and take action to 7 j prevent 3e unsafe,any progress toward a condition that might b i e. Believe and respond to instrument indications t
! until they are proven to be incorrect,,
I f. Not bypass, reset bypasses, defeat safety' systems i or interlocks or remove Category 1, 2, or 3 / instrument channels 'from service, unless allowed to do so by an approved procedure. ,
,~ ,. . . , .
e 4 h a 703445
%oCEDURE'No. REVISloN VEGP 10000-C 7 PAGE No.
15 of 26 ( S , '. p M [31s311L Shift Conduct g I. e3 3 thl- 3Eachtmembercof?theTshiftferew;shAlliperformiinla0i 7professionalEmanner.C Potentially; distracting 1.._ .P .. W -) 4
' . "lis Fl activities:-shall area,a Activitiesiprohibited not belconductied?
include'loiteringb l M int t teningito? mus ic? E hobb ie s ynon Lj ob-related c reading ". material, and horseplay.(The full! focus of thef shift ' icomplement':s? attention-'shalEbe thei safeLand efficienti operation 9 of "the d plant. s . 1 533fli27 OperationsTpersonrielIon.shlf te sustibeLaware isf a'ni
% responsible,for:the' plant; status;at~alletimes. This includes" supervisors being responsible for the performance of personnel assigned to their shift who could affect plant safety. ' '
. e 4 T.
4373! $ '. __ mwAll"Operatibns'personn'el must',be' alert and remain . . within their work area until properly relieved, o Operators are responsible for monitoring the 4 n instrumentation and controls located areas. They are responsible for taking timely and within their work. proper actions to ensure safe operation of the facility;.g .g, ,. , ,.
~
(3'.'3~ '.~ l!4 ControlsL plevel ofi that directly affectithe reactivity lorJpowerj aireactor; shall only be manipulated by yg; licensed;operatorsrexceptcforetraining. purposes. o
- 3.3.1.5:' ' l Mechanisms andlapparatus,.otherJtihanicontrols,fthatlmay
. indirectly affect the powerLlevelcor; reactivity'of A reactor shall"only be operated with the~ knowledge and' prior' consent:of a licensed operator.
+3;3 1 Abnormal Indications The OSOS, USS, and RO, and BOP have the authority and responsibility to perform the tasks necessary to limit l plant operations or to shutdown the unit when such action is warranted by unit conditions, unusual circumstances,_o When~analyzingr 1such situations,r unidentified events.
fshifttoperating: personnel'shall;.
- consider instrument > readings" and control; indications. to'.
? be' true unleasD the~y Lare proven ~ to' be t incorrecc.GWhen1 o
~abn'rmal indications; occur," operations:personnelEshall A- initlateLappropriateecorrective~
de tiermitie F thei caus e V o f : the ? abnormal ' indica tion Eand ' action. : { h
T 8 PR oCEDU'RE No' REVISloN "" PAGE No. VEGP 10000-C 7 16 of 26 I
) 93)3331' Instrument Setpoints t
4Shiftroperatingl personnel;r,h'all not~ manip'ul'te a ~ instrument? control #orfal' arm"setpointsF other than those availabledn the ?controli c'ossolelor ' thosernormsny?
.x requiredfdurinprouti'efoperation. n 'Setpoint changes-
'""shall b'e entered in the ~ Unit Control. Log or the Shift i
' Supervisor's Log._ j p _y y y 9 . . . <o wwm ; .
, n,g w _ ,, y 7,,.
i( < &Anyone2p'erforming operation or a Control a function Room indication that may affect shall notify a' unit's the i (; Control Room operators before initiating the function.,,,4 w' me--+-~~ ~ ~~~~~~% _,_,__,,_,cen ' r73_.3.4 "ControlfRoom-Access'
~
~. , .~.. ~ m. m ,c~: en ~ .. : . ~,,,,...u l
D,Controli Room ; acce s ss sha11..b'ellimitiedi tofd.ffi"cial? ;
.businessc o nly in.accordance.with. Plant Administrative ProcedureLOO301-C"
;PersonneliConduct9(Main Control Room'.AccessLAnd * ' ' ' ~
SEFreRggEs 3.3.5 Generator Load Changes yg@ggg jy- i Normally generator load changes will be made as requested by the System Operator. If approved by the { j OSOS, and if plant operating conditions and operational 3 orders permit, the Reactor Operator will comply with the request. The RO shall inform the USS when the l i l l requested load change is completed. Whenever plant conditions require a load. change, the System Operator - shall be notified as soon as possible of the proposed load and rate of change. j i Scheduled outage requests shall be initiated by the Operations Manager and approved by the General Manager
,<. and_ System Operator prior to scheduled plant shutdown. 1
..w.. - a .~
v { t323.6~ * -Control Room Housekeeping 0 I * > ,
}' .; + +< #
The Control Room will be maintained in a clean and
'q' orderly condition in the interest of safe and efficient y~ operations. Dustin instrument panels, and g and cleaning computer of control consoles willconsoles, be performed by shift operating personnel. 3
" ~
.l I
7C3*c5
.r.
PROCEDURE No. REVISloN PAGE NO. 00301-C- 2 2 of 3
'4.0 CONDUCT OF PERSONNEL'-
- . . . . . s. / a.:
JEverylperson:< entering 4the,(Control?Roomt Area;;will' // l
'^o bserve"the~following good conduct rules; For" the" purposes of this procedure " Control Room Area" includes
'those areas of the control panels, relay and. control cdbinets, rest room, Shift Supervisor's office, kitchen, instrument repair room, . and chart storage.
room.
#'a'f""~Keepnnoised.evels4< . atofa[ minimum.; %.,_
m ,a nau. we a:+ x - - .'.. %,ML?
. UO db $M, Maintainsari"
.nu : ~.a t t it6deV,o fu,m faler tnes ug sg..
'c.- Keep;the. area in ta;cleandandvorderlymcondition by, e ard;;nga:3 implementing and;maintainin'g Cgood ' housekeeping % i N pracriae~s'.alfJ you make3a r :me ssp;cleanuits
. ~ ,, a up .
l ddt : 70btaiti" operations z.supervisoryvapprova15 prior; to < V .performingianyiactivity? that would'bevdisruptivet I fand-)possibl.ytaffectiopeiatioiR ,"?
~~
g 'j
.e.: Do notLtakeefoodeand/oradrink!t'oltheicontrol
~ ~^"
m 7p'anelsY L : -x. ,w :- u
& :y+ 9a
5.0 REFERENCES
5.1 FSAR 13.5.1.1.D 5.2 NUREG-0578,'Section 2.2.2.a 5.3 Procedure 10000-C, " Conduct of Operations" I END OF PROCEDURE TEXT 1
]
1 l 1 I i l s 103445 - A 1 i ______________J
I 1 l
;I QUESTION 8.10 (1.00)
State the TWO personnel, by title, given the. responsibility to analyze 1 i a trip. event for the Reactor Trip Report according to procedure l 1 10006-C, " Reactor: Trip Review?" i 1 4 l ANSWER 8.10 (1.00).- I
- 1. Shitt Support Supervisor l'
- 2. i On shift Operations Supervisor '
j l RETERENCE - i l VEGP Procedure 10006-C, p. 2' 0120^60002 ..(KA's) l i l i 8 1 i .l 1 i 1 i l i i i ! l (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) i _mm.- _ . - _ _ _ _ _ _ _ _ . _ . _ _ _ _ . _ _ .
u ..-- , 7
' 1,
-QUESTION NUMBER: 8.10 i-STATEMENT: The answer key. lists.two people by name, but the-
"0 SOS" can designate someone to take.the place,of the Support Shift Supervisor.. Thiu should be included as an acceptable answer also.
, FACILITY RECOMMENDATION:
i '
' Accept OSOS Designee as an alternate answer.-
REFERENCES:
. Procedure 10006-C i i i
\
)
s l i I (
- 1 1
I l~ t ,' hi ' I 4; h
'E I'
4 c I S. PROCEDURE Wo. SEVIStoM VEGP- 10006-C PAGE No.
- g. .
i 9 2 of 19 i 3.3 REACTOR TRIP REPORT 3.3.1 The Support Trip. Report", Shift Supervisor shall complete a "Reacter-recorded. Data Sheet 1. Any abnormality should be the review. Its significance can be determined during For items that cannot be verified, indicate,so.on,.the, trip report and why. s
'of a SupportIShiftlSupervlsordthW!0n-Shlft OperationsVIn d the1ab'
'Supervisorfshallidesignatessomeone1to!completefthe(J ff*P,"#Ci .
i$ 3.3.2 The Support' Shift Supervisor and On-Shift Operations Supervisor'shall cause analyze the event to determine its and the following: a. Did all major safety-related and other important equipment $nvolved in the trip operate as I expected? _This review shall include a detailed review of the pre- and post-trip-computer data to i equipment operation and proper plant respons b. j i Is it acceptable to restart the reactor? shall look beyond the obvious indications toThey diagnose plant response.the cause of the event and evaluate the~ ! information throughly, looking for:They shall review availab j (1) abnormal ind performance,ications or degraded equipment
~
(2) evente occurring out anticipated sequence,of the normal
.(3) failed orsignals, control abnormal response of equipment te
,l (4) unusual chemistry readings, and reaults or radiation (5) unanticipated alarms.
3.3.3
..Any abnormality be analyzed discovered during the analysis should and corrected.
by Maintenance Work Order (MWO) number, Action ItemList any number, etc. on Data Sheet 1. 4 3.3.4 be completed by the persons involved with the O included with the Reactor Trip Report, Reacter TriporPersonnel Data Sheet 1. , individuals groups ofStatements individuals asmay be completed by lon promotes'an accurate event reconstruction. g as it cm
.),,
~~~
l I
)
QUESTION 8.12 (2.00) , Who are the first FOUR individuals, by title, who can assume the duties of Emergency Director? Consider the worst base conditions. ANSWER 8.12 , (2.00)
- 1. OSOS
- 2. SS
- 3. Sr. Vice President Nuclear Operations
- 4. General Manager, Vogtle Nuclear Operations REFERENCE VEGP Procedure 91102-C, p.i l i
I i l I i i I i I (***** CATEGORY B CONTINUED ON NEXT PAGE *****)
> 1 x
QUESTION: ;8.12 j
.n STATEMENT: Reorganization has.resulted in, title changes.. Also the
?' worst case" scenario could result in another individual 'j
'being amon'g the first' four individuals assuming E3 i' dutics. This is because'the SRVP, Nuclear Operations is based 'iit Birmingham, Alabama.
FACILITY RECOMMENDATION: y
-1
' Accept alternate titles for. General Manager and Vice -;
President. Accept Plant Manager as an additional ~ choice. l 1 l REFERENCES. . Tech Specs 6.2.1 'i \ 91102-C 4 1 y 1 1 1 1 i
~l J
d 4 1 l 1
a S .+ ADMINISTRATIVE CONTROLS-6.1 RESPONSIBILITY 6.1.1 The General Nanager - Nuclear Plant shall be responsible for overall plant operation and shall delegate in writing the succession to this respon- l
.sibility during his absence.
6.1.2 The General Manager - Nuclear Plant will annually reissue a directive [ Supervisor (or during his absence from the control room, the in designated to assume the command functions) for safe operation of the plant under all conditions on his shift and that clearly establishes his command duties. E.2 ORGANIZATION ONSITE AND OFFSITE ORGANIZATION 6.2.1 Onsite and offsite organizations shall be established for' plant operation and corporate management, respectively. The onsite and offsite organizations sT&TTplant. power include the positions for activities affecting the safety of the nuclear a. Lines of authority, responsibility, and communication shall,be estaba lished and defined for the highest management 1evels through inter-mediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the fom of organization charts, functional descriptions of-departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent' forms of documentation. These requirements shall be documented in the FSAR. b. MGMal NE-iblear:hient$sha11 Responsible for overail l
" plant safe operation and shall' have control over those onsite activ-ities necessary for safe operation and maintenance of the plant. 1 c.
The3Ne" Pees dehtYNuil'oedshall have corporate responsibility fo overall plant nuclear safety ~and shall take any measures needed to !' ensure acceptable performance of the staff in operating, maintaining, d. and providing technical support to the plant to ensure nuclear safety. The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the ! appropriate onsite manager; however, they shall have sufficient organizational pressures. freedom to ensure their independence from operating PLANT STAFF 6.2.2 The plant organization shall be subject to the following: a. Each on-duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1; l V0GTLE - UNIT 1 6-1 Amendment No. 6
)
Apprrv3 Proc; curs No.
, Vogtle Electric Cenerating Plant 91102-C NUCLEAR OPERATIONS Revision No.
D.te ' 4
/6 /ff/ uni, Coss0N Georgia Power esse Ng. eu I
DUTIES OF THE EMERGENCY DIRECTOR I 1.0 PURPOSE
]
The purpose of this procedure is to provide i instructions to the Emergency Director in fulfilling his responsibility for overall coordination and direction of the VEGP Emergency Response Organization. 2.0 RESPONSIBILITIES i 2.1 DESIGNEES 1 2.1.1 On-Shift Operations Supervisor:(OSOS)lor. Shift l Supervisor,i _m The OSOS shall assume the position of Emergency Director in the Control Room until relieved by the designated Emergency Director. If the OSOS is incapacitated, the Shift Supervisor shall assume the position of Emergency Director. ' j 2.1.2 Th'e'~ Sr.~ ' Vice ipr'esident!. NuclearjOperat-iods or Gener"al Manager,rVogtle'. Nuclear Operations)are the7 designated primaries:for the Emergency-Director. ' ' 2.1.3 Plant Manager The1Pl'antf theipossition Maiaber i.sitis designatedIfirstEal' o th,e, Emergency; Director.7 t ernatNfor
,;w 1 2.1.4 Plant Support Manager The Plant Support Manager is the designated second alternate for the position of the Emergency Director.
2.1.5 Manager, Operations The Manager, Operations is the designated third ( l alternate for the position of the Emergency Director. FC310lA MCst3D j
QUESTION 8.19 (2.00) According to procedure 00306, " Temporary Jumpers and Lifted Wires", what are the responsibilities of the Shift Supervisor?
~
ANSWER 8.19 (2.00) l (0.5 anch) 1. Control of jumpers and lifted wires.
- 2. Final approval of jumpers and lifted wires.
- 3. Determination involvement in safety related system.
- 4. Review of log.to ensure proper transfer of operating information from one shift to the next.
REFERENCE VEGP Administrative Procedure, 00306-C, sec. 3.2.1, p. 4 194001A108 ..(KA's) i l l l (***** END OF CATEGORY 8 *****) (********** END OF EXAMINATION **********) I
"; q; ., ,
~
T '
~
e
^
t 1 4 , J . i. .
. 'i t
; QUESTION NUMBER:. 8.19'
' -'^
'{
STATEMENT:: ~NRC answer breaks SS. responsibilities-_down;into 4.. ' q}
- parts /and J: separates . the .first1 sentence of- the" SS-- l. ;
. 1 responsibilities-to;implyc.thatLthese are separate) :j
# m rduties'. ; The procedure ' states The; Shif t1. Supervisors d
;is responsible:for, control'andl final- approval'of TJ;
- l
; and/or LW clearances'?.' : LThis 1s.a'singleiduty'and should:not'be' broken ldown into more than 1: activity." .
FullfereditLfor Parts-1;&L-2 oflthe: answer.in'the' key, Tshotild be allowed ' for.: paraphrasing thisl statement . asi followsi '" Reviews and-approves the-installaticu and-
. removal of gall TJ. 6 LW".; .
I-FACILITY RECOMMENDATION:
, ,. .. , . .. . . .. . l Combine' answers liand=2 on? answer, key:into_a single-
[ ' answer and accept: alternate wording; y tj
REFERENCES:
00306-C' Step 312.1 - s
~1
'N.
'I
- j ii s i
)
1 e 1
'l 1
>i
!. i
;f
_.__=_.__._______.m____.m.. m__-_m_ .__.mi._m_.:__._m_-_. __ .__._ _- _m__m_.__m_____.________mm-.,_._ --__.___m_ ,___________m____-____.____m_ _ _ . , _ _ _ _ .__=.'__._ ____.__-.___m_
_m _ _ _ _ - - 00306-C ((of 4 of 15
- 3. 2 ' OPERATIONS DEPARTMENT
- n-- w.mm-w wannwm m - w.e m,
~1TheiShiftiSupervisorsiesresponsiblasferic,.ontrolsandi 32.1-7finalSapprovalsof4TJPand/or>LWsclearances.:He vil1.
datermine when a TJ and/or LW involves a safety-related, s} stem. The Shif t . Supervisor, . after he assumes : the shift, will -
' review active Jumper and Lifted Wire Clearances, to ensure a proper; transfer of operating information from one shift to-the next.
The Manager Operations, or designee, shall perform a
~
3.2.2 . quarterly review of the extended active Jumper and Lifted Wire clearances to check for continued validity. - This review shall include a verification, by Operations personnel, that safety >related-TJs and LWs are properly. installed and. all tags are properly attached and legible. This review will be documented across the next line of the Clearance Index. Each active clearance sheet will be initialed and dated. For those clearances found to-be of questionable-validity, the initiator of the clearance will.give reasons why the TJs and/or LWs should stay installed and why the wiring change should not be made permanent. 3.3 ENGINEERING SUPPORT DEPARTMENT Engineering personnel shall perform a technical review: of proposed TJ and/or LW modifications controlled by this proc ~edure. 4.0 INSTRUCTIONS
4.1 EXCLUSIONS
4.1.1 This procedure is not required when a Maintenance Work Order (MWO) requires, as : art of a functional test, that all wires and linkt ie red-lined.- 4.1.2 The controls described in this procedure are not. applicable for Temporary Jumpers (TJ) 'and/or Lifted Wires (LW) that are controlled or positively identified in other approved procedures, provided that the other procedures clearly s:>ecify and provide methods for ocumenting all o the fo lowing: ,p
- a. Shift Supervisor notification when installing or removing TJs and/or LWs.
9 9 _.1_________ ___________ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
3/4.11'RADI0 ACTIVE EFFLUENTS: 3/4.11.1 LIQUID EFFLUENTS i' 1 a _ CONCENTRATION 1 LIMITING CONDITION FOR OPERATION ,, i 3.11.1.11: . }, The concentrationiof radioactive! material released in liquid effluents. ' to UNRESTRICTED AREAS specifiedf'in?10TCFRYPh.(see t Fjgure 5.1-1) shall be concentrations limited to-the[ 1 othefi rti20EAppendixf8 2TableRII.;Columni2rf.qr r than dissoluedM. entrained noble gases. foe dissolved or--entrained' noble' gases, the concentration shall' be' limited to 2$10-SmicNicurle/mlitet'al activity :
. APPLICABILITY::iAt all times. '
ACTION:
- a. l
. hith the concentration of radioactive material released;in_1fquid
-to UNRESTRICTED AREAS exceeding the above limits """
finesidiately^f ~ ~~ ~~" restoreithe. tconcoatrationito?withistheT~aboM11mits.J uLsuunsM"" " ~ " ~ ' - b. The. provisions of' Specifications 3.0.3 and 3.0.4'are not applicable. SJJRVEILLANCE REQUIREMENTS- .i
'4.11.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to.the; sampling and analysis program of Table 4.11-1. ]
4.11.1.1.2. Tne results of the radioactivityLanalyses shall be used in accordance i with the methodology and parameters in the ODCM to asstire that the concentrations: - at the point of release are maintained within-the licits of Specification 3.11.1.1. s j ' 1
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QUESTION 7.23 (1.00) What is the minimum pressurizer level and reactor coolant system subcooling recommended for a Reactor Coolant Pump restart if a void is ; suspected in the vessel following a Natural Circulation Cooldown? ; ANSWER 7.23 (1.00)
- 1. PZR level - 872
- 2. Subcooling - 67 F j
REFERENCE-I VEGP Lesson Plan LO-LP-37012-03-C Lesson Objective LO-LP-37012-03-C, #10 Student Handout LO-HO-37012-01-C-001, p. 17 000055K302 ..(KA's) .. ( l i
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. QUESTION: 7.23 STATEMENT: Values in answer do not. reflect those in the. procedure.
FACILITY RECOMMENDATION: PRZR Level - 96% Subcooling - 60*F
REFERENCES:
19003'C step 2
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,A PROCEDURE No. REVISloN PA2E No.
VEGP 19003-C 4 3 of 11 s ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED ' CAUTION e Two hot starts of an RCP are allowed if the pump coasts to a stop,between starts. (Approximately 2 minutes.) e Subsequent starts are allowed if the pump has been running for 20 minutes or idle for 45 minutes. e Start only one RCP at a time. NOTE e RCP 4 or RCP 1 should be run preferentially to ~ provide normal PRZR spray. e If conditions are established for starting an RCP during this procedure, Step 2 should be repeated.
- 2. Try To Restart A RCP:
- a. Check RVLIS u?per range a. Perform the following:
indication FULL.
- 1) Raise PRZR level to
* .9E _using charging and' letdown.
- 2) Establish subcooling greater than d60' y using steam dump,
- b. Try to restart a RCP b. Go to Step 3.
using Attachment A.
- c. Go to appropriate plant procedure with Shift Supervisors concurrence.
I I FfYWe$
QUESTION. 8.05 (1.50) Fill in the blanks. A Fire Team consisting of at least- [a] , members shall be maintained on site at all times. The [b]- (title) shall designate.the: Fire Team captain and members at the beginning of-each shift. The* Fire Team may be reduced by.one person for a period not to exceed- [c] . hour (s). l ANSWER; 8.05 .(1.50) I
- a. 5 b.- On Shift Operations Supervisor
- c. 4 \
i REFERENCE VEGP Procedure 10003-C, Manning the Shift, p. 2
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' ';lI QUESTION NUMBER: 8.05 STATEMENT: ' Answer.co Part-C concerning allowed time for which' the Fire Team may be,. reduced by.one~ person is 2 hours per procedure 10003-C. step.3.5.- l(Answer-stated 4' hours) ;
FACILITY RECOMMENDATION:
.J Change answer key to 2 hours.- j
REFERENCES:
10003-C. step 3.5 (attached) 1 s J t i i I
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-moCEpu8tE No REvtssoN ;
' VEGP 10003-C- PAGE No
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,- 7 ~'
2 of 5 J/ () 3.4 A Fire , Team consisting of at a team captain) least 5 members times. shall be maintained on site at all(including and members The atOSOS shall designate the Fire Team captain captain and at least the beginning of each shift. The team will know plant safety related systems well enough totwo
, understand on the plant's the effects of a fire and fire suppressants safe shall not include the OSOS, SS..utdown capability. The Fire Team are requiredp,,-r
- for the minimum on,-shift operating staffRO, BOP, A 3.5 ~e ~w mmww -~. .
TTh Eshifm$_ EFire;Teai ,sfc be idekminimumMred6frementsnof: Tablei fMddheb reducedsbyfons7personiiforj alt jeriodsfa s "unexpscred timelnotitoyax)ceedi2thours A This is to accommodate absence of on-duty shift personnel. the minimum requirements.Immediate action shall be taken to re This provision does not change due to tardiness or absence of on-comingp personnel. 3.6 During any absence of the SS from the Control Room while the unit is in Mode 1, 2, 3 or 4, an individual with a valid SRO license shall be designated to assume O the Control Room command function. of the SS from the Control Room while the unit is'inDuring any absenc Mode 5 or 6, an individual with a valid SR0 or R0 license shall be designated to assume the Control Room command function. 3.7 The balance of Plant (BOP) operator will normally the remain plant.in the Control Room when not needed elsewhere i 3.8 An operator shall be assigned to the Auxiliary Boiler in the Maintenance operation. Building whenever it is in responsibilities.He shall have no other concurrent 3.9 The OSOS shall designate'a qualified person to perform the Shift Technical beginning Advisor (STA) function at the of each shift. shift functEons for 30 days or longer shall be briefedAny STA who h in accordance with 00715-Requalification Program".C " Licensed Operator (5) l j
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l QUESTION 8.09 (2.50) ! List FIVE General Work Practices which'all personnel assigned to shift operations are responsible to carry-out as detailed in procedure 10000-C, Conduct of Operations. ANSWER- 8.09 -(2.50) ' t'any 5 0 0.5 each)
- 1. Assure safe operation of the plant under all conditions .
- 2. Protect plant personnel, the health and safety of the public, and plant equipment..
- 3. Conduct major plant operations in-accordance with approved written procedures.
- 4. Be attentive to conditions of the plant at all times. I
- 5. Believe and respond to instrument indications until ther are proven l to be incorrect. i I 8. Not bypass, reset bypasses, defeat safety systems or interlocks or l-remove Cat. 1, 2, 3 instrument channels from service unless allowed i to do so by approved procedures. I 1
REFERENCE VEGP Procedure 10000-C, p. 14 Peach Botton Incident 1 194001K103 ..(KA's) 1 i i 1 (***** CATEGORY -8 CONTINUED ON NEXT PAGE *****)
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- QUESTION NUMBER: ' 8.'09 )
1 I - STATEMENTi Procedure 10000-C Revision 7, Section 3.3 (General ! Work Practices) contains many more items than those ] listed.in'the. key.' Sections 3.3.1 through 3.3.6 are L all General Work Practices and should be included in :j 1 , the key. 'Also included in section 3.3.4 control L room access (which.is still part of section 3.3' General Work Practices) is a reference.to 00301-C..
-These General Work ' Practices should also be J l accepted. ;
FACILITY RECOMMENDATION: Please accept the;following additional answers per section 3.3 of 10000-C and 00301-C.
- 1. Perform in a professional manner.
- 2. Maintain full focus.on safe / efficient op'eration.
- 3. No distracting activities in the Control room. ;
- 4. No' loitering. music, hobbies, non-j ob reading, i
material or horseplay.
- 5. Be aware of and responsible for plant status at all times.
- 6. Be alert and remain in work area until relieved. .
- 7. Take timely and proper actions to ensure safe l Ops.
- 8. Licensed Operators shall be the.only one to affect reactivity / reactor power except for-training.
- 9. Manipulation of equipment that may (indire'etly) ;
affect power of reactivity must be with ( knowledge and consent of Licensed Operator. l
- 10. Believe instruments unless proven wrong.- ;
- 11. Determine cause of abnormal situation and cure.
- 12. Do not manipulate instrument control or alarm setpoints unless on console or normally required during routine ops.
- 13. Notify Control Room prior to initiating' actions that could affect Unit Ops..
- 14. Control Room Access conduct all personnel includes:
- a. Keep noise levels to a minimum
- b. Maintain an attitude of alertnese
- c. Keep area clean
- d. Obtain Ops' Supv. approval prior to disruptive activities
- e. Do not take food / drink to control board
- 15. Maintr'n clean / orderly control room.
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'I ENCLOSURE 4-I <
SIMOLATION FACILITY = FIDELITY' REPORT Facility' Licensee: Georgia. Power Company Facility. License Docket Nos.:- 150-424:and 50-425'.; - Facility' License ~Nos.: NFP-68 and NFP,-81~ Operating Tests: administered At: IVogtleElectric' Generating' Plant Training Simulator
-January,10 -'12, 1989
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Operating Tests'Given.0ni
- Duringethe conduct ofthe simulator-: portion ofLthe' operating tests identified 1 above, the following apparent performance and/or human. fcctors1 discrepancies L; were' observed'and discussed.with theLfacility's staff instructors.
- 1. .During a steam generator tube leak scenari6, the procedurally isolated steam- generator . exhibited' the isame pressureLdecrease. as . the intact, cooling-down steam generator.-
- 2. Periodically, the candidates tended to look. back at the staff instructor.
in the simulator control booth. 'With no visual' barrier isolating the candidates from the. staff ' instructor. operating -the' simulator, the. candidates. were ' distracted in the' performance of their' duties. 'The-candidates stated and demonstrated an awareness of the simulator's-artificiality in this regard. 4 n l l 1 J i
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