ML20151V999
| ML20151V999 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 04/15/1988 |
| From: | Bill Dean, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20151V985 | List: |
| References | |
| 50-395-OL-88-01, 50-395-OL-88-1, NUDOCS 8805030266 | |
| Download: ML20151V999 (208) | |
Text
{{#Wiki_filter:_ . i [ ENCLOSURE 1 EXAMINATION REPORT 395/0L-88-01 .l, 1 Facility Licensee: South Carolina Electric and Gas Company P. O. Box 88 i Jenkinsville, SC 29065 Facility Name: V. C. Summer Nuclear Station ! Facility Docket No.: 50-395 Written examinations and operating tests were administered at V. C. Summer Nuclear ! Station near Jenkinsvil e, South Carolina : Chief Examiner: I - # /8 , William M. Dean Date Signed Approved by: / [# Jonh F. Munro,1 Chief tg// W Uate Signed i I Operator Licensing Section 1 t 1 Summary: Examinations on March 7-10, 1988, f. Operating and written examinations were administered to six initial senior ! reactor operator (SRO) and two reactor operator (RO) retake candidates. All ! of the candidates passed the operating and written exams, i Four of the 14 changes (29%) made to the written examinations were a result of inadequate or incomplete reference material provided to the NRC for ! examination item development. . i i l 8805030266 080427 ' PDR ADOCK 05000395 V DCD ; i
E REPORT DETAILS
- 1. Facility Employees Contacted:
- Ken Woodward, Manager, Training
- Gene Soult, OPS Manager
- Randy Ruff, Supervisor, Training
- Victor Kelley, Senior Instructor
- Attended Exit Meeting
- 2. Examiners:
- William Dean Richard Baldwin Ron Aiello Pete Isaksen, INEL (PerryHopkins,R1,attendedexit)
- Chief Examiner
- 3. Examination Review Meeting At the conclusion of the written examinations, the examiners provided your training staff, with a copy of the written examination and answer key for review. The NRC Resolutions to facility coments are listed below,
- a. SR0 Exam (Applicable R0 questions are in parentheses)
(1) Question 5.07(b) l Disagree with facility coment. Due to the fact that a power level ; was not specified, there is no ONE correct answer for part "b". I Therefore, part "b" has been deleted from the exam. For future ' reference, the question will, specify a power level which will, in turn, define the fuel centerline temperature, thus making the question NON ambiguous. (2) Question 5.15(b) Facility coment accepted. 5.15(b) has been deleted from the exam. (3) Question 6.01(2.21) Facility coment accepted. Suggested additional answer (choice "c") will be accepted as one of the responses. I
?
I r 2 (4) Question 6.07 Facility coment accepted. The answer key has been changed to ' the recomended response. , (5) Question 6.10 [ Facility coment accepted. The answer key has been changed to the ! recomended response, j i (6) Question 6.14 (3.04) Facility comment accepted. 'The answer key has been changed to ; the recomended response, j (7) Question 6.15 j Facility coment accepted. "High level alarm at 70%" will be ; deleted from the answer key.
~
r (8) Question 6.19(1.15) Facility coment accepted. Suggested answers 1 & 2 paraphrase , the existing answers in the key. However, the answer key will
- be expanded to include the facility's suggested answer number !
three. The training material should be modified to reflect this information. (9) Question 7.11 ' Facility coment accepted. Due to the lack of specific procedural guidance, the facility's suggested answer will also be accepted. (10) Question 7.13 Facility comment accepted. Note, however, that ARP-001, XCP-636, P-12 specifically addresses this failure. Although operators should be familiar with what actions to be taken for a casualty of this nature, the question was not adequately worded to elicit the desired information. The question will be deleted from the exam. { (11) Question 7.17 Facility coment accepted. Number one of the answer key will be changed as recommended. l l
1 l 3 (12) Question 7.23 Facility coment accepted. The answer key will be changed as recomended. (13) Question 8.22 Disagree with facility coment. It is cleat' that taking "B" EDG 005 renders both RHR pumps inoperable, thereby placing one in an action statement. Furthermore, the RHR loop may be removed from operation for up to one hour per eight hour period only during the performance of core alterations in the vicinity of the reactor pressure vessel hot legs. The answer key remains unchanged.
- b. R0 Exam (1) Question 2,08 (a)
Facility coment noted. Forexamgradingconsistency.(.75) will be allotted for the protective feature and (.25) for the set point. (2) Question 2.13(a) Facility coment noted. For exam grading consistenc will be allotted for the automatic action and (.19) for y. (.56) the set point. (3) Question 2.17 Facility comment accepted. Note that the information supporting this answer was not made available to the author prior to the exam. (4) Question 3.06 (b) Facility's recommended answer is equivalent to existing answer key. No change required. (5) Question 3.18 (a) Facility coment accepted. The answer key has been changed to accept the recomended answer for full credit. (6) Question 4.17(a) Facility coment accepted. The yearly administrative limit is clearly steted in the reference utilized for question development. No change to answer key,
l l l 4
- c. Post-Exam Review Changes A detailed review of the examinations and associated answer keys resulted in the following additional changes:
(1) Question 2.12 (b) The TS basis for minimum spent fuel pool water level relative to removal of iodine gap activity was added as an additional correct answer. The facility reference material should be updated to reflect this information. l (2) Question 7.15 Tripping the turbine locally from the front standard was added as an additional correct answer as the qJestion Was not specific enough to elicit control room actions exclusively. l l l l 1 i l l l l l l l l 1 l j
1
- 4. Exit Meeting At the conclusion of the site visit the examiners met with representatives ,
of the plant staff to discuss the results of the examination. There were no generic weaknesses noted during the oral examination. The examiners did note that there may be some inconsistent training on requirements to restart condensate pumps after a loss of condensate flow. It was also noted that the facility's ongoing effort to update both their lesson plan learning objectives and simulator exercise guides has greatly improved these training documents and once fully implemented should be much more closely related to their job task analysis. . The cooperation given to the examiners and the effort to ensure an j atmosphere in the control room conducive to oral examinations was also l noted and appreciated. The licensee did not identify as proprietary any of the material provided te or reviewed by the examiners. 1 l 1 i l l 1 l
p. y s-
+ U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: SUMMER __,_,,,,_,_,,,_____
REACTOR TYPE: PWS-WgG3_________________ gj "*'.' &O 1) OATE ADMINSTERED: _8_8_/_0_3_/_0_7_________________ I L
- 00 EXAMINER: _I __S A_K __S E N_2_
_ P _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ , CANDIDATE _________________________ lySIgggIlgN@,IQ,g@yQ1Q@Ig1 Use separate paper for the answers. Write answers on one side only. Staole 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*/. i n each category and a final grade of at , least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY
__Ye6WE. _I9Ie6 ___EG9BE___ _y@6yg __ _ __ _ _ __ ___ _ _ __q@IEQQB y___ __ _ __ _ _ __ _ ;
.49z99._ .2Ez99 ____...____ ________ i. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, i HEAT TRANSFER AND FLUID FLOW ,
_29199.. 2E199 ..__.._____ ___..___ 2. PLANT DESIGN INCLUDING SAFETY , AND EMERGENCY SYSTEMS i 6 29199_. _2E199 _________.. _______ 3. INSTRUMENTS AND CONTROLS l t _}9299._ .2E199 ___________ ________ 4. PROCEDURES - NORMAL, ASNORMAL, { EMERGENCY AND RADIOLOGICAL t CONTROL _12919._ ___________ ________% Totals ' Final Grace All work done on this examination is my own. I have neither given , nor receivea aic. l l t m= J V , J1 aTDV b i u_________.____________________________
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules applyt
- 1. Cheating on the examination means an automatic don'ial of your application i and could result in more severe penalties. {
- 2. Restroom trips are to be limited and only one candidate at a time may i leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating. l
- 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 proviced for answers.
)
- 7. Print vour name in the uppur right-hand corner of the first page of each !
section of the answer sheet.
- 6. Consecutively number each answer sheet, write "End of Category __" as 4 appropriate, start each category on a new page, write only on one side i 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 abureviations only if they are commonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer reautred.
14 Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems wnether 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 nave not receivec or ceen given assistance in completing the examination. This must be done after the examination nas
] been completed. I l ) l 4 I I A a
~
E g' l 1t - i
. i I .
I 19. When you complete your examination, you shells i ! l l .a. Assemble your examination as follows: i (1) Exam questions on top. i l (2) Exam aids - figures, tables, etc. (3) Answer pages including figures which are part of the answer. l f b. Turn in your copy of the examination and all pages used to answer ( l the examination questions. l 1 l c. Turn in all scrap paper and the balance of the paper that you did i not use for answering the questions, i
- c. Leave the examination area, as defined by the examiner. If after ;
leaving, you are found in this area while the examination is still l , in progress, you- license may be denied or revoked. t i I i 1 i 1 l l i l l l
Pceo 4
'la__E61091 ELE 5_9E_SW96693_E9Ws0_E699I_92EB9II992 IbE6D99209019Ex dE9I_I69N}{[$,$yp,g(yJp,gypy QUESTION 1.01 (1.00)
Increasing the boron concentration at low temperature has little effect i on the Moderator Temperature Coefficient as compared to higher operating l temperatures becauset (choose the correct answer)
- 4. water density does not change as much at low temperature.
O. water density is-greater at lower temperatures so neutron leakage is less.
- c. Wate" density is greater at lower temperatures so parasitic neutron absorption is greater.
- d. boric acid is less soluble at lower temperatures.
QUESTION 1.02 (1.00) Which one of the following statements concerning Xenon-135 production and removal is correct?
- a. At full power, equilibrium conditions, about half of the Xenon is produced by lodine decay and the other half is produced as a direct fission product.
- b. Following a reactor trip from equilibrium conditions, Xenon peaks because delayed neutron precursors continue to decay to Aenon wnile neutron aesorption (burnout) has ceased. ,
- c. Xenon production and removal increases linearly as oower level !
increases: 1.e., the value of 100% equilibrium Xenon is twice ' that of 50% equiliorium xenon. l d. At low power levels, Xenon decay is the major removal method. At htgn power levels, burnout is the major removal method. i l l l l l teesse CATEGORY 1 CONTINtJED ON NEXT PAGE esess) l l
. l
i .- .- [ e
'12_ ES1091E6ES_gE_ygg6[98_Egy[0_E699I_QgES$I1ggi Pcgo 5 IdEBO99109019ft_bl9I_IB90EEEB_9SP.E6W19_EL9W
, , ?
GUEST 10N 1.03 (1.00) l t As the core ages, the ratio of PU239 atoms to U235 atcms increases. This_cnanging ratio causes thes (Choose the correct answer)
- a. reactor startup rate (SUR) to increase, for the same reactivity ;
addition. l l
- b. Void Coefficient to become less negative. ;
l
- c. Moderator Temperature Coefficient to become less negative. ;
- d. Celayed neutron fraction to increase.
QUESTION 1.04 (1.00) which one of the following statements concerning Shutdown Margin LSDM) - is correct.' l
- a. The maximum SDM requirement occurs at EOL and is based on a rod l ejection accident. l t
- e. The maximum SDM requirement occurs at EOL and is based on a steam i line break acetdent, i
- c. Tne maximum SDM requirement occurs at SQL anc is based on having a j positive moderator temperature coefficient.
J
- a. The maximum SDM requirement occurs at BOL and is based on a rod I withdrawal accident wnile in tne source range.
1 l l tesess CATEGORY 1 CONTINUED ON NEXT PAGE sessa) l l l i i
'Az__'ESibCIE6ES,gE_yyChE9E_EghgD_E69y1,96559Il992 PCgo 6 i Ibs50991090195t_b59I_I6905E56_909_E6919_E69d QUESTICN 1.05 (1.00)
The reactor is critical at 10,000 cps wnen a S./G PORV fails open. Assuming ECL conditions, no rod motion, and no reactor trip, choose the answer colow that best describes the values of Tavg and nuclear power for the resulting new steady state. (PCAH = point of adding neat),
- a. Final Tavg greater than initial Tavg, Final power aeove POAH.
- b. Final Tavg greater than initial Tavg, Final power at POAH.
- c. Final Tavg less than initial Tavg, Final power at POAH.
j d. Final Tavg less than initial Tavg, Final power above POAH. i l QUESTION 1.06 (1.50) l The roautor is taken critical with Xenon concentration at zero. Power l 1s raised to 50% at 5%/nr. 1 1 i Use one of the following choices to cascrice now Xenon concentration I will be trencing for each of the following situations (a,b, and c). Increasing Decreaoing At equilierium
- 4. One nour after a trip occurs as power reaches 50%.
D. Four nours after the trip occurs the reactor is taken critical and power raised back to 50%.
- c. If reactor operation continues and power level is maintained at 50%
for one nour. QUESTION 1.07 (2.00) l a. How is Shutcown Margin LSDM) affected (Increase, Decrease, or No , enange) by a 50 ppm coren accition wnile operating at 50% power? l to.br l o. List FIVE factors, other than RCS coren concentration and rod I position, which will affect SDM ano are useo in the SDM l calculation. (1.5) l 1 tosses CATEGORY 1 CCNTINUED ON NEAT PAGE esse ) l l l--_____. _ _ _ _ _ _ _ _
Pego '
' 1:__'E61NC186Eg_gE_NyC6g@B_EgWE8_E6@NI_gEgB@IlgN2
' IHg E DQQ Y N@d] C gz _ Hg @I_I S@ N@ E E 6_ @ NQ_ E(yJ Q, E 69 W QUESTION 1.08 (1.00) ,
Which one of the following conditions would cause a 1/M plot to be NON-conservative during fuel loading?
- a. Fuel being leaded closer to the neutron source than to the source range detector.
D. Loading fuel in the order of high reactivity worth to low react.vity worth,
- c. Loading poison rods between the source range detectors and spaces to be filled by fuel assemblies.
- d. Increasing the boron concentration in the moderator.
QUESTION 1.09 (1.00) During a Xenon-free reactor startup, critical data was inadvertently taken two decades below the required Intermediate Range (1R) level (1 E-10 amps). The critical data was then taken at the proper IR level (1 E-08 amps). Assuming RCS temperature and baron concentra-tion did not change, which one of the following statements is correct?
- a. The critical rod posi ti on taken at the proper IR level is LESS THAN the critical rod position taken two decades below the proper IR level.
- b. The cri tical rod position taken at the proper IR level is THE SAME AS the critical rod position taken two decades below the proper IR level.
- c. The critical rod position taken at nie proper IR level is GREATER THAN the critical rod posi t i on taken two decades below the proper IR level.
- d. There is not enough information given to determire the relationship between tne critical rod position tasen at the proper IR level and the critical rod position taken two decades below the proper IR level.
(us*** CATEGORY l CONTINUED ON NEXT PAGE *****)
' 1_ 'E81Ng186E@_QE_NyC6E98_EQWE8_E6@NI_9EEB8IlgN 2 Pcgo 8
'IbE8099XN@ digs _bggI_I@@NgEEB_@NQ_E691p_E69W 1
QUESTION 1.10 (1.50) Match the parameter change in Column A to the direction it will change the tiederator Temperature Coefficient (MTC) in Column B. Consider each case separatel y. COLUMN A COLUMN 9
- 1. Moderator temperature increases a. More Negative
- 2. Baron concentration increases b. Less Negative
- 3. All rods in vs. all rods out c. No Effect QUESTION 1.11 (2.00)
If the Source Range (SR) instruments indicate 50 cps wit.h Keff equal to O.9, what would tne SR instrument indicate if rods were withdrawn to bring Keff equal to 0.95? Assume EOL conditions.
- a. 1. 50 cps
- 2. 75 cps
- 3. 100 cps 4 200 cps (1.0)
- b. How much reactivity was addeG?
- 1. 0.0347
- 2. 0.0500 !
- 3. 3526 4 0.0585 (1.0) I l
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
'It__E61NCIE6ES_gE_GUC(E@6_EQWE6_E6@NI_QEEB@IlgN t Pago 9 ISE60QQ1N@d1CS t_dE@I_IB@NSEE6_@NQ_E6UlQ_E6QW
~
QUESTION 1.12 (1.50) Compare the calculated Estimated Critical Position (ECP) for a startup 15 hours after a trip from 100*/. power operation equilibrium conditions to the Actual Critical Rod Position (ACP) if the following events /condi-tions occurred. Consider each independently. Limit your answer tot
- a. ACP higher than ECP.
- b. ACP lower than ECP.
- c. ACP would not be significantly different than ECP.
- 1. One Reactor Coolant Pump is stopped one minute prior to criticality.
- 2. The steam dump pressure setpoint is increased to a value just below the code safties setpoints.
- 3. The startup is delayed 2 more hours.
QUESTION 1.13 (1.00) Which one of the f ollow1 - ;orrectly describes the observed reactor response for the same smais addition of reactivity, one positive and one negative?
- a. The response will be faster for the negative addition at all ti mes in core life.
- b. The response will be faster for the negative addition at BOL Dut faster for the positive addition at EOL.
- c. The response will be faster for the positive addition at all times in core life.
- d. The response will be faster for the positive addition at BOL but taster for the negative addition at EOL.
- e. The response will be the same for both the positive and negative addition.
(***** CATEGCRY 1 CONTINUED ON NEXT PAGE *****)
.- i i
-1_ _ _ E81 N CI E6 E S_ gE _ hUC 6 E 98_ E QWg B _ E 6 @N1_ gEg@@ll gN1 Pcgo 10
'IUE6DggyNgDICS 1_bg@l_l@@NSEgg,@Np,[6UJg_E699 l
- j QUESTION 1.14 (1.00)
Which one of the following statements concerning power defect is correct?
- a. The power defect is the difference between the measured power coefficient and the predicted power coefficient.
- c. The power defect increases the rod worth requirements necessary to maintain the desired shutdown margin following a reactor trip.
- c. Because of the higher boron concentration, the power defect is more negative at beginning of core life.
- d. The power defect necessitates the use of a ramped Tavg program to maintain an adequate Reactor Coolant System subccoling margin.
QUESTION 1.15 (1.00) State TWO reasons for establishing Rod Insertion limits. QUESTION 1.16 (1.00)
" centrifugal pump is operating at rated flow when the discharge valve is throttled towards the snut direction. For each of the following indicate whether the parameter will increase, decrease 7r remain the same, after the valve is throttled in the shut direction. I
- a. NPSH
- b. motor amperage I
i l i l (***s* CATEGORY 1 CCNTINUED ON NEXT PAGE ***ss) , i i i
'1___26]NCIB6Eg_QE_UUC6E@B_EQWEB_26@NI_gEEB@llgN t Pago 11 IUEBD99XN@d]C@g_bg@I_IB@NgEg6_@NQ_E6Ulp_E6QW QUESTION 1.17 (1.00)
TRUE or FALSE
- a. During~a RCS heatup, as temperature gets higher, it will take a smaller letdown flow rate to maintain a constant pressurizer level.
a
- b. Increasing condensate depression (subcooling) will cause BOTH a decrease in plant efficiency AND an increase in condensate (hotwell) pump available NPSH.
QUESTION 1.18 (1.50) Assume the plan' is operating at 85% power with normal / automatic system-lineup. Indicate how the following changes in plant conditions would-atfect DNBR (increase, decrease, remain the same). Consider each case separately.
- 1. AFD changes from 0% to +5%.
- 2. Steam Generator PORV f ail s open.
- 3. Pressurizer heaters are inadvertantly lett on and prsssure increases 50 psig. '
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
' 1___E61NCIE6ES_QE_NyC6E98_EgWE5_E68BI_gEEB911gN 1 Page 12 IdEBdQQyN@digSz_bE@I_I@@NSEE6_@NQ_E69]Q_E6QW 1
1 i I QUESTION 1.19 (2.50)
- a. What is the subccoling margin (SCM) of the RCS if the following conditions exist?
Th =580 F Pressurizer pressure =2185 psig Tc =520 F Steam Generator Pressure =850 psig (1.0)
- b. If power is raised f rom 50 to 100%, how AND why will SCM change (increase, decrease, stay the same)? (0.75)
- c. Which one of the following would result in the smal. ;t SCM? Briefly explain your choice. Assume identical RCS pressures. (0.75)
- 1. SCM during a controlled natural circulation cooldown following a reactor trip from loss of flow. ,
l
- 2. SCM during continued operation at 5% power.
I
- 3. SCM produced when all RCP's are operated at normal no-load l temperature after extended shutdown. l OUESTION 1.20 (1.50)
At each of the following leak locations, indicate the state of the l exiting flutd (subcooled, saturated, or superneated). Assume normal 100% power initial operating plant conditions. l
- a. PZR steam space to CTMT atmosphere.
D. Steam Dump to the condenser.
- c. Main steam header to turbine building atmosphere.
l l l l l l i 1 l i (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) i
1
. 1___EBINCIE6gS_gE_Nyg6E96_E99EB_E6@NI_ GEE 6@IlgN i Pego 13 l
.IbsBD99XN@digg1_bg@I_IB@NgEgB_@NQ_E691Q,E6gy 1
1 I CUESTION 1.21 (2.00) The reactor is operating at 30% power when one RCP trips. Assuming ; no reactor trip or turbine load change occur, indicate whether the j following parameters will INCREASE, DECREASE, or REMAIN THE SAME. j
- a. Flow in operating reactor coolant loops
- b. Core delta T
- c. Reactor vessel delta P
- d. Operating loop steam generator pressure l
i QUESTION 1.22 (1.00) Primary system flow rate is many times greater than secondary system , flow rate while the heat transferred by the two systems is essentially l the same. Explain how this is possicle. GUESTION 1.23 (1.00) List all of the conditions that must be present in order for natural circulation to exist. 1 (***** END OF CATEGORY l *****)
- 2 __E68Bl_pggigy_Jhg(Uplyg,@@ggly_8dp_gdg595b91 Pego 14
. *s15IE05 QUESTION 2.01 (1.50)
What signal (s) must be present for an AUTOMATIC switchover of the suction of the RHR system from the RWST to the reactor building recirculation sumps to occur? Give setpoint(s) and coincidence (s) if applicable. QUESTION 2.02 (2.00) Which components of the Reactor Building Spray system are affected by as
- a. Phase A Containment Isolation Signal?
- b. Spray Actuation Signal?
QUESTION 2.03 (1.50) The following questions are associated with the normal service gas decay tanks.
- a. How many normal service gas decay tanks are there?
- b. Normally, how often is the in-service tank switched?
- c. Why is the waste gas distributed among all normal service gas decay tanks instead of filling one tank at a time?
QUESTION 2.04 (1.50) List THREE systems +or which the Acoustic Leak Monitoring System provides indications of a leak. , (***** CATEGORY 2 CONTINUED ON NEXT PAGE $9s**)
a 3 2.- 2 E68NI_QESigN_lhC6UQ1Ng_S8 Eely _@NQ_EMEggENCY Pegs 15 SYSTEMS QUESTION 2.05 (1.00) Which one of the following statements describing the design of the fuel transfer tube is correct? 1 a. A blind flange is used to close the transfer tube on BOTH the containment side and the spent fuel side. D. A blind flange is used to close the transfer tube on the containment side and a valve is used on the spent fuel side.
- c. A valve is used to close the transfer tube on the contain-ment side and a blind flange is used on the spent fuel side.
- d. A valve is used to close the transfer tube on BOTH the con-tainment side and the spent fuel side.
GUESTION 2.0e (1.00) The #3 RC pump seal leakoff is normally collected in which one of the following?
- a. Containment Sump
- b. Pressurizer Rel i ef Tank
- c. Volume Control Tank
- d. Reactor Coolant Drain Tank QUESTION 2.07 (2.00) i The following questions concern the RHR System.
- a. What THREE interlocks must be met prior to opening the RHR inlet line isolation valves (8701A or 0702A)? (1.0) 1 1
1
- b. What interlock will automatically close the RHR inlet line j isolation valves? (include setpoint) (0.5)
- c. What is the bases of the RHR inlet line relief valve capacity? (0.5) 1 I
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
' 2___h66NI_pEglGN,1NCLUp]NG_3@EEIy_@Np_EMEBGENCy Pego le
. 'Sy@IEM@
QUESTION 2.08 (2.00)
- a. Other than the thermal barrier heat exchanger, list THREE reactor building loads supplied by the Component Cooling Water System.
- b. Other than relief valves, what feature prevents overpressuriza-tion of the CCW System if a thermal barrier heat' exchanger tube ruptures? (Include setpoint, if applicable)
QUESTION 2.09 (1.50) Match the RCP seal flow paths in Column A to the appropriate design flow rate in Column B. COLUMN A COLUMN B
- a. Down shaft into the RCS 1. 10 cc/hr
- 2. 100 cc/hr
- b. #2 seal leakage 3. 3 gph 4 5 gph
- c. #3 seal leakage 5. 3 gpm
- 6. 5 gpm QUESTION 2.10. (1.00)
List TWO relief valves that discharge into the Volume Control Tank. QUESTION 2.11 (1.50) The following questions concern the Emergency Feed Pumps.
- a. What is their normal scurce of water?
D. What is their backup source of water?
- c. What signal shifts the supply from the normal source to the backup source? Setpoints and coincidence NOT required.
(***** CATEGORY 2 CCNTINUED ON NEXT PAGE *****) _ . _ _ _ -, , _ - _ _ _ , . . . - .,.-_ - _ _ _ _ _ _ _ _ _ . . . , . . _ ~ - _ . . . _ . _ . . . . _ . _ . _ _ . - _
2 __P(@yI_QE@lgy_lNC6UglN@_@@Egly_@gg_EdE8@ENCy Pcgo 17 S_ Y_ S_ _T E_M_ S_ QUESTION 2.12 (2.00)
- a. State THREE of the four places that flow from the Spent Fuel Cooling and Transfer pump can be directed to.
- b. What are the TWO reasons f or establi shing the design low water level of the Spent Fuel Pit (SFP)?
1 QUESTION 2.13 (2.00)
- a. Describe TWO automatic actions associated with the instrument air system which serve to mitigate a loss of air pressure. Include '
any associated setpoints. (1.5)
- b. Assume the plant is operating at 100% power and NO operator action is taken. What will initially cause a reactor trip on a continued l loss of instrument air pressure? (0.5) !
1 QUESTION 2.14 (2.00)
- a. What TWO conditions will energize a lockout (86 relay) for an ESF transformer?
- b. List FOUR of the abnormal conditions which will cause an ESF 1 tr ansf ormer- trouble alarm, but will NOT cause the transformer l to be deenergized. (Setpoints not required) i i
OUESTION 2.15 (1.S0) State the purpose AND operation of the interlocks between the Letdown isolation valves and the orifice isolation valves. (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
'2_dE60NI_9ESIGN_INC6U91dG_g@EEIy_8N9_EDE8GENCy 2 Pego 18
. 51EIEDg GUESTION 2.16 (1.00)
Wnich one of the following statements is NOT true concerning the operation of "C" CVCS pump, if a blackout occurs on the ESF bus with "C" pump aligned to it.
- a. The "C" pump would be tripped and locked out provided the A/B pump on that reain was racked in and running,
- b. The "C" pump weteld continue to run l' it was the running pump on that train.
- c. If the A/B pump on that train is manually started the "C" pump would cor.tinue to run.
- d. If the A/B pump on that train starts on an SI signal the "C" pump will trip and be locked out.
QUESTION 2.17 (1.00) What automatic action (s) should ocdur to the Reactor Building Cooling Units upon receipt of an SIAS signal? Assume an initial normal at power lineup. QUESTION 2.18 (1.00) The S/G PORV's maximum capacity is limited by design to approximately 6% of rated steam flow. What is the reason for this limitation? t***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
l l 2 t__P6@NI_Qgg]GN_JNC6Up]NG_@@EEIy_@NQ_EdgBGENCy Pega 19
!X5IEDS QUESTION 2.19 (1.00)
An undervoltage on a 7.2 kV safeguards bus occurs 20 seconds after the receipt of a Safety Injection signal. Which of the following state-ments regarding sequencing of loads onto the safeguards bus is correct?
- a. All loads except Lead Block #1 are stripped and the EGF.
Leading Sequence is reinitiated once the DG output breaker is closed.
- b. Sequencing stops until the DG output breaker is closed at which time it continues from the point at which the under-voltage occurs.
- c. Sequencing stops until the DG output breaker is closed at which time only the ECCS-related equipment sequence will be reinitiated.
- d. All loads except the ECCS-related equipment are stripped and only the ECCS-related equipment sequence will be continued once the DG output breaker is closed.
QUESTION 2.20 (1.00) What is the main design purpose of the flow restricting nozzle in the Main Steam Lines? 1 QUESTION 2.21 (1.00) j i Which statement below regarding the Main Generator Protectica System I is NOT CORRECT. a) Opening the generator output breakers ALWAYS results in a turbine trip when the generator is loaded. l l b) Once the generator is leaded, a turbine trip ALWAYS results in a generator trip. l c) A turbine trip above the protection interlock P-7 (10% power) ALWAYS results in a Reactor trip. d) A reactor trip ALWAYS resul ts in a turbine trip. 1 (***** END OF CATEGORY 2 *****)
Pego 20 3__41N@l@UdENIQ_@ND_CQGI6Q6S QUESTION 3.01 (2.00) Indicate whether the Over Power Delta Temperature trip setpoint will INCREASE, DECREASE, or REMAIN THE SAME for each of the following parameter changes. Consider each separately.
- a. Increasing Tavg
- b. Tavg less than rated power Tavg
- c. Delta I becoming more negative
- d. Pressurizer Pressure decreasing QUESTION 3.02 (1.00)
Which one of the following flowpaths describing how power is normally supplied to a typical vital instrument bus is correct?
- a. 125 VDC from battery, supplied to battery bus, inverted to 120 VAC, and supplied to instrument bus.
- o. 480 VAC from vital bus, transformed to 120 VAC, and supplied to instrument bus. ,
- c. 480 VAC from vital bus, rectified to 125 VDC, inverted to 120 VAC, and supolied to instrument bus.
- d. 480 VAC from vital bus, rectified to 120 VDC, and supplied to instrument bus.
QUESTION 3.03 (1.00) List FIVE outputs of the Pressurizer Pressure master controller. NOTE: Redundant outputs count as one, i.e., Pump A and Pump B. QUESTION 3.04 (1.00) The Cold Overpressure Protection System (COPS) provides an alarm to warn the operator that overpressure protection is isolated. What plant conditions will cause this alarm? (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
3:__1NSI@UdENIS_AND_ CONI @O65 Page 21 QUESTION 3.05 (1.50) With the pressurizer level control switch in Position 2, describe the response for a high failure of LT-459, including components affected, alarms received, and the effect on actual pressurizer level. Assume normal charging and letdown system lineups and no operator actions are taken. Continue the description until pressurizer l evel is constant or a reactor trip occurs. Include setpoints were applicable. QUESTION 3.06 (1.50) l l
- a. What TWO interlocks must be satisfied prior to manually resetting j a safety injection signal? (1.0) l
- b. After resetting safety injection, automatic actuation is inhibited until what signal is cleared? (0.5) l l
QUESTION 3.07 (1.00) Describe the interlock associated with the reactor trip bypass breakers. ' QUESTION 3.08 (1.50) Provide the expected RVLIS indications for all THREE ranges (Upper, Narrow, and Wide) for a full vessel at 100% AND 0% flow. QUESTION 3.09 (1.00) Which one of the following malfunctions will result in both a low Tavg indication and a low delta T indication?
- a. Hot leg RTD failed high
- b. Hot leg RTD failed low
- c. Cold leg RTD failed high
- d. Cold leg RTD failed low
)
l l l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
\
3 __INSIGUMENI@_@ND_CgNIRgLg Pcgo 22 QUESTION 3.10 (1.50) Describe what each of the f ollowing Steam Dump System status lights indicate.
- a. "PERMISV C-9" status light dims.
- b. "STEAM DUMP CONTROL" status light illuminates.
- c. "PERMISV C-78 PB-447B" status light brightens.
QUESTION 3.11 (1.00) Indicate whether the f ollowing conditions will cause the Steam Dump System to ARM ONLY, ARM & DUMP, or HAVE NO EFFECT.
- a. Tavg Mode, Tavg channel fails high during a 3% per minute load reduction. 1
- c. Tavg Mode, first stage pressure (PT-447) fails low with Tavg l
1.5 degrees F greater than iref. QUESTION 3.12 (1.00) i Indicate whether the following malfunctions in the SGWLC System will result in a LOW, HIGH, or NO S/G level trip signal. Assume initially tne reactor is at 75% power with S/G 1evels at program.
- a. Steam pressure (density compensation signal) fails high. j
- b. N-44 fails low.
QUESTION 3.13 (1.50) The following questions are associated with the Reactor Control Unit of the Rod Control System.
- a. What signals are used to generate the temperature error signal?
1
- b. Wnat signals are used to generate the power error signal?
- c. What is the purpose of the variable gain applied to the power error signal?
i I l 1
*** CATEGORY 3 CONTINUED ON NEXT PAGE *****) 1 I
1
3:_.JySIBUMENI@_@yp_CQNI69Lg Pogo 23 QUESTION 3.14 (1.50) i List SIX rod control interlocks. Setpoints are NOT required. Redundant interlocks only count once, i.e., Channel A high voltage and Channel B high voltage. ,
.l QUESTION 3.15 (1.50)
Li st ALL input parameters used to calculate the Over Temperature Delta T trip setpoint. QUESTION 3.16 (1.50) j What THREE signals will cause a feedwater i sol ati on ? (Setpoints NOT l required) l GUESTION 3.17 (2.00) Which of the following radiation monitor channels have automatic actions (other than indication and alarm) associated with them. Briefly describe the automatic actions, if any.
- a. Reactor Building Manipulator crane (RM-G17A)
- b. Component Cooling Water (RM-L2A)
- c. Nuclear Blowdown Waste Effluent (RM-L7)
- d. Main Plant Vent Exhaust (RM-A3) l l
l (as*** CATEGORY 3 CONTINUED ON NEXT PAGE ****e) l l
Pago 24
'2_{1NSISUMENIS_@ND_CONISO(S OUESTION 3.18 (3.00)
Answer the following concerning the Core Cooling Monitor (CCM).
- a. What are the FIVE inputs to the CCM? Be specific. (1.25)
- b. For each of the THREE CCM status lights, state the color AND briefly describe what is meant when each of the status lights are lit. (0.75)
- c. Wnat does it mean if the status lights are flasning? If all the status lights are off? (1.0)
GUESTION 3.19 (2.50) The following concern the Rod Position Indication (RPI) system. t
- a. What is the effect of taking the "accuracy mode" switch out of its "normal" A + B position? (1.0)
- b. State TWO conditions which will cause a DRPI urgent alarm. (1.0)
- c. How are the affected rods displayed when a RPI urgent alarrn signal is received? (0.5)
OUESTION 3.20 (1.50) State the purpcse/ function for eacn of the following permissive signals. Include wnich instrument is used to develop the signal, setpoints are NOT required.
- a. P-6
- b. P-8
- c. p-9
(***** END OF CATEGORY 3 *****)
i . . ' j
' 3:_IE899g9Uggg_;_UgBU@62 ABNORMALg_EMERGgNgY Pego 25
'9Np_G@pJg69GIC@6_ggNIBg6 i
l QUESTION 4.01 (2.50)
- a. What is the MINIMUM number of operable excore channels indicating AFD outside the target band before AFD is considered outside its target band by Technical Specifications (TS)? (0,5)
- b. Assume the plant is operating at full power and the AFD has been outside the target band for the last 5 minutes. What are the TWO different actions specified which may be used to meet the TS requirements? Include time limitations, if any. (1.0)
- c. Assume that it is 0310 on 03/07/88 and the plant is presently at 45% power. Considering the AFD penalty history below, at what )
date and time may power be increased above 50%? Explain and ! show all work. Assume no deviation outside the band after 0310 , on 03/07/88. l TIME WENT OUT TIME BACK DATE OF BAND IN BAND % POWER 03/06/88 0310 0318 85 03/06/88 1557 1637 65 03/07/88 0148 0310 45 (1.0) QUESTION 4.02 (2.50)
- a. What are THREE symptoms / plant conditions which would require the RCS to be emergency borated, accorcing to EOP-11.0, Emergency Boration procecure? (1.5)
- b. What TWO operator actions are required to perform Emergency 8cration? (1.0) i QUESTION 4.03 (2.00) l l
According to SAP-2OO, Conduct of Operations administrative procecure, under what condition may the "operator at tne controls" leave the ' surveillance area of the control room curing Mode 1 operation? How does this change during Mode 5 operation? i l L***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) I i l
*4:__P89CEQUBES_;_Ng6[@6t_@BNg6d@61_EDE69ENCy Pcgo 26
. eNQ_6@Qlg69G1C96_CQNI6gL QUESTION 4.04 (1.00)
What are the Immediate Corrective Action (s) required if the charging flow control valve (FCV-122) fails closed while in automatic control' , and will not respond to a manual open signal? QUESTION 4.05 (1.50) What are the immediate Corrective Actions if Group 1 and Group 2 step counters of Bank D rods differ by more than 1 step? (Three required) GUESTION 4.06 (1.00) Which one of the below statements describe the RCP trip criteria following a SI with normal containment conditions?
- a. Less than 30 degrees subcooling AND pzr leve!, less than 4%.
- b. Less than 30 degrees subccoling AND S1 flow verified.
- c. RCS pressure less than 1380 psig AND SI flow verified.
- d. RCS pressure less than 1380 psig AND pzr level less than 4%.
QUESTION 4.07 (1.00) If a void exists in the reactor vessel with all RCPs stopped, which of the following actions is used to collapse the void according to EOP-18.2, "Response to Voids in Reactor Vessel"?
- a. Decrease temperature while maintaining system pressure.
- c. Start a SI pump to increase system pressure while keeping temperature constant.
- c. Increase system pressure using pressurizer neaters wnile maintaining pressurizer level.
- d. Fill pressurizer solid and vent the reactor vessel head.
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****;
l ,* 1 6:_(P60GEDUBEg_;_NO@d@6t_@@@g6d@6t_EdERGENCY Pcga 27
'AND RADIOLOGICAL CONTROL l
l 1
- i QUESTION 4.08 (2.00)
List the FOUR possible alternate actions that can be taken if during a response to abnormal power generation (ATWS) the turbine had not auto-matically tripped. 1 I QUESTION 4.09 (1.00) What plant condition determines when "ADVERSE" Containment conditions (brackets in EOP) are used in the Emergency Operating Procedures? ! QUESTION 4.10 (1.50)
- a. What is the Technical Specification Safety Limit for RCS pressure j while in Mode S? (0,5) '
- b. What are the required actions if this limit is violated? Only l include actions required witnin i hour. (1.0) !
l i I I QUESTION 4.11 (1.00) l Indicate the numerical value(s) associated with the following precautions.
- a. Maximum differential pressure between RCS and S/G. l l
- b. Maximum differential temperature between RCS loops,
- c. Minimum RCS flowr&te prior to and during RCS dilutions,
- d. Maximum rate of power increase above 20% reactor power without management approval.
QUESTION 4.12 (1.00) List the TWO actions required if the actual critical position is above l the low-low insertion limit but differs from the estimated critical position by more than 50 steps, and less than the maximum rod withdrawal limit, according to GOP-APPENDIX A, Generic Operating Precautions. l (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
.' l l
- S __EBOggDUBE5_:_NOBD961_@@UOBD66t_EDES9EN91 PcgG 28
'8NQ,8@Qlg69@lg@6_ggNIBQ6 l
QUESTION 4.13 (1.50) The f ollowing questi ons concern the requirements associated with performing a reactor startup, according to GDP-3, Rea: tor Startup from Hot Standby to Startup procedure,
- a. When must all Shutdown Bank rods fully withdrawn be verified?
- b. How many licensed operators must be present in the Control Room?
Be specific as to'what licenses are required. I
- c. If the startup is delayed, when must the Estimated Critical l Condition calculation be reviewed?
QUESTION 4.14 (1.50)
- a. Indicate the Immediate Corrective Action (s) required, if while operating at 50% power, the following alarms occur simultaneously.
"RCP A #1 SL LKOFF FLO HI/LO" and "RCP A #1 SL dP LO"
- b. How much time is allotted to take the above Immediate Corrective Action?
- c. For how long may the RCP be operated with a #1 Seal Failure?
QUESTION 4.15 (1.50) What THREE conditions would require that SI be reinitiated, according to EOP-1.2, Safety Injection Termination procedure? (1.5) i 1 i i 1 l l (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) i l l l 1 _ - y--
) '.
'3:_lEBQgEQUBgS_ _Ng8d@61_8gyggd@(1_gdgBgENgy Pego 29 8NQ_B8Qlg(Q@lg@6_ggNIBQ6 QUESTION 4.16 (1.50)
Answer the following according to EOP-10.0, Malfunction of Rod Control System,
- a. Other than verifying the reactor is not tripped for a dropped control rod condition with the plant at full power, what are the TWO remaining immediate operator actions? (1.0)
- b. TRUE or FALSE? i When commencing recovery of the dropped rod an "URGENT FAILURE" alarm ,
will occur due to the lift coils for the other rods in the group l being disconnected. (0.5) QUESTION 4.17 (1.50) Answer the following according to VC Summer Radiation Protection Fundamental s (HP Handout).
- a. What are the V.C. Summer Nuclear Station Administrative exposure I limits for Whole Body, Skin, and extremities for occupational l workers? (exclude fertile females) (1.0)
- b. (Fill in the blank on your answer sheet) i Any time your dosimeter reads _____ me or greater when entering the RCA it should be re-zerced by HP. (0.5)
QUESTION 4.18 (1.50) Caution III.1 of EOP 15.0 "Response to loss of Secondary Heat Sink" states: If S/G Wide Range levels in any 2 S/G is < 20% OR Pressurizer pressure is > 2335 psig then STOP ALL RCPs and immediately initiate bleed and feed per steps 7 through 14. Why are the RCPs tripped prior to initiating bleed and feed, aside from the fact that heat input from the pumps will be removed? 4 (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) i
31_;E60gggU6gg_ _UO5d@62_9pygBd@61_gdg6GENQY Page 30 eNp,@@Qlg6QGig@6_ggyl@Q6 QUESTION 4.19 (1.00) Emergency Operating Procedure EOP-13, "Response to Abnormal Nuclear Power Generation", has the operator trip the turbine as one of the immediate actions. However, one of the major concerns in the ATWS transient response evaluations is the excessive RCS pressure developed due to significant heatup of the primary coolant. Since keeping the turbine on the line would mitigate this temperature rise, why is the turbine tripped? QUESTION 4.20 (1.00) With the exception of electrical personnel acting in the capacity of a red tag, what are the TWO required qualifications of the individual responsible for second verification of danger tag placement? QUESTION 4.21 (1.00) What, AS A MINIMUM, should the temporary or unexpected relief turnover include? (***** END OF CATEGORY 4 ****s) (********** END OF EXAMINATION **********)
EQUATION SHEET w v = ms s = v,c + hac 2 Cycle efficiency = Net Work , (out)
- E = aC
- a = (v, - v,)/t 2
A = AN " KE = hav y, , y +g A = A,e PE = agh w = 6/c 1 = in 2/tg = 0.693/tg , { W = vaP-eq (eff) = (t.)(ts) i .. . AE = 931Am . (g + g) Q=[ncAT
, P , I=Ieo d*
', Q = UAAT I=Ieg -UX ,
Pvr = W' g I" , I = I, 10 */ M P=P 10 (th M = 1.3/u y=y t e /T o HVT. a 0.693/u
'SUR = 26.C6/7 ~
T = 1.44 07 SCR = S/(1 - X,gg) o SUR = 26 f gA [f , CR x = S/(1 - K,gg,)
~
T = '(1*/o ) + [(i ' o)/A,ggo ] IC eff)1
- 2(I ~Eeff }'2 T,= 1*/ (o - D M " 1/(I ~ Kaff) = CR /CR g 0 I*I ~ 8)! eff M = (1 - g,gg) /(1 - g,gg)
'"I aff ~I)IKeff " AKeff/E eff m = 0 - gg
~
p= [1*/TKygg -] + [3/(1 + A,ggT )) i* = 1 x 10 seconds P = I4V/(3 x 10 0) A,ggs= i 0.1 seconds I = No - Idgy=Id22 WATER PARAMETERS Idg =Id222 1 gal. = 8.345 lba R/hr = (0.5 CE)/d 2 g,,,,,,) I gal. = 3.78 liters R/hr = 6 CE/d (feet) . I ft = 7.48 gal. MISCEI.I.ANEOUS CONVERSIONS , Density = G2.4 lbm/ft 3 1 Curia = 3.7 x 10 dps 10 Density = 1 gs/cm 3 1 kg = 2.21 1ha j Heat of va;ori:stioni = 970 reu/lbm 1 hp = 2.54 x 103 BTU /hr ! Hest of fusica = 144 Btu /lbs 1 Hw = 3.41 x 10 Stu/hr 6 1 Atm = 14.7 psi = 29.9 in. I'g. 1 Stu = 778 f t-lbf l 1 f t. H2O = 0.4333 lbf/in 2 g inch = 2.54 cm ! F = 9/5*C + 32 i
*C = 3/9 (*r . 32) l
y. I
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION t Pego 31 ISE8dQQIN@d1CSt_dE@I_IB@NSEE8_@NQ_E(UlQ_E(QW IMB:CPY -
ANSWER 1.01 (1.00) A. REFERENCE , VCS RT-11, p 12-15. EO-5,10 KAI 3.1 192OO4K106 ..(KA's) ANSWER 1.02 (1.00) d. REFERENCE VCS RT-12, p 11-23. EO-3 1 KA1 3.1 192OO6K105 ..(KA's) [ ANSWER 1.03 (1.00) a. REFERENCE VCS RT-10, p 17-29. EO-15 l l KA1 3.2 192003K106 ..(KA's) ANSWER 1.04 (1.00? l l c. (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) W R ::PY
v s
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION z Pcgo 32 IUE5099XN@d1CS3 _bE91_I@@NSEE6,9NQ_E(ylp_E(gW REFERENCE VCS TS, p B 3/4 1-1; RT-15. EO-3,5 KAI 3.8 192002K114 ..(KA's)
ANSWER 1.05 (1.00) d.. REFERENCE VCS RT-11. p 24,25. EO-19 KAI 3.1 192008K114 ..(KA's) ANSWER 1.06 (1.50)
- a. Increasing.
- b. Decreasing,
- c. Increasing. Co.5 eacn3 (1.5)
REFERENCE VCS RT-12, p 20-24. EO-4,5 KAI 3.4,3.4 ; 19200eK106 192006K107 ..(KA's) ! ANSWER 1.07 (2.00) . i t
- a. SDM is increased. Co.53
- b. Cany 5, 0.3 each3
-RCS avg temp -Samarium l
-Fuel burnup -Power defect l
-Xenon concentration -Power level
]
(sses: CATEGORY 1 CONTINUED ON NEXT PAGE sasse)
iz_,E51NQ186E@_QE_Nyg(EeB_eQWE8_E(@NI_QEE8@IlgNt Pcgo 33
- ISEBdQDYN@ dig @t_bE8I_IB@NSEE6_@ND_E(gig _E(QW
~ REFERENCE VCS RT-15, p 7-10. EO-6,7 KAI 3.8 _
192OO2K114 ..(KA's). ANSWER 1.08 (1.00) c REFERENCE i VCS RT-8, p 21-26. EO-10 ; VCS, SOP-403, Rod Control and Position Indicating System, p 14. KAI 3.8,2.9 192008K106 ~ 192OO2K114 ..(KA's) ANSWER 1.09 (1.00) b REFERENCE VCS RT 17, p 11. EO-2 Westinghouse Reactor Physics, Sect. 3, Neutron Kinetics and Sect. 5, Core Physics KAI 3.3 1 001010K505 ..(KA's) 1 - ANEWER 1.10 (1.50) i
, 1. a
- 2. b
- 3. a Co.5 each)
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
l
- i !
- 1. PRINCI(lj@_QE_Nyg6g@R_PQWER,P(ANT _QPERATigN 1 Pcgo.34 i
- IbE50.1_N@digSg_bg@l,I@@NS((5_@NQ;[6QJQ,E(QW i
REFERENCE VCS RT-11, p 12-20. EO-5,7,8 Westinghouse Nuclear Training Operations, pp. 1-5.6 - 16 : KAI 3.1 192004K106 ..(KA's) ANSWER 1.11 (2.00)
- a. 3
- b. 4 C1.0 each]
REFERENCE VCS RT-8, p 15-17.- EO-6,7 SHNP, RT-HO-1.6. KAI 3.8 192008K104 ..(KA's) ANSWER 1.12 (1.50)
- 1. c (same)
- 2. a (ACP higher)
- 3. b (ACP lower) [0.5 each3 REFERENCE i
VCS RT-15. EO-4,7 ! Cook Theory, Pp. 1-36-45. l SHNP, RT-HO-1.14. KAl 3.6 I 001010A207 001010K207 . . ( K A ' s 's l l ANSWER 1.13 (1.00) i c j (ssas* C:4TEGORY 1 CONTINUED ON NEXT PAGE ***ss) i i
8 a -
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION3 Pcgo.35 4 - IbEBd9DXN@djg@g,bg@l,]@@Nj[ES,9NQ,[(yjD,[(QW REFERENCE VCS, RT-10, P 11.
RT-6, LO 13, 16, 21, 26, 29. KAI 3.3 192OJ8K110 192OO6K116 192OO2K114 ..(KA's) ANSWER 1.14 (1.00) b REFERENCE I VCS RT-11, p 24-26. EO-17 Westinghouse Reactor Physics, pp. I-5.26 & 2/ SHNPP RT-LP-1.10, p 13-15. KA1 3.8,2.9 l 192OO4K113 192OO2K114 ..tKA's) , 1 ANSWER 1.15 (1.00) I
- 1. To insure minimum shutdown margin is maintained.
- 2. Minimize the reactivity consequences of an. ejected r od . at M -M4 A s r-ur~ p e~ m _n_?-.,+===,= n r- p.s
- 3. Maintain acceptable axial flux di stri buti on.c% A/
- 6 w l-:=w) .
Cany 2, 0.5 each3 REFERENCE VCS RT-14, p 20 1 TG p 3/4 1-3. EO-10 i l j MAI 3.4 192OO5K115 ..(KA's) I l ANSWER 1.16 (1.00)
- a. Increase l
- b. Decrease C0.5 each)
(ssss* CATEGORY 1 CONTINUED ON NEXT F^GE *****) l l l l _.~.._-..I
li_ ESINQlE(gg_QE_NQQLg@g_EQWg8_g(@NI_QEg@@IlQN g Pcgo 36 IUEBd991N@DlQS2 _dg@l_IB@NSEg@,$NQ,E(gip _E6QW REFERENCE GP HTFF p.328 NEO KA1 2.9 l- 193OO6K105 ..(KA's) ANSWER 1.17 (1.00) . a. FALSE i
- b. TRUE CO.5 eachJ REFERENCE
^ GP HT&FF, p 155,320 and Subcool ed Liquid Density Tables. NEO KAI 2.5,2.4 193OO4K111 193OO5k1O3 ..(KA's) I ANSWER 1.18 (1.50) a
- 1. DNBR cecreases 4 2. DNBR decreases
- 3. DNBR increases CO.5 each3 REFERENCE I GP HTFF p 243-259. NEO ,
KA1 3.4 193OOOK105 ..(KA's) 4 4 1 l j T 3 (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) 1 1 l l
1:_ EBINQ166ES_QE_NQg6E@B_EQWEB_E6SNI_QEg@@IlgNt Pcgo 37
- IU55d991N$digSg_bg@l_IS@NSEgg,8NQ_E(ylp_E6QW i ANSWER 1.19 (2.50) :
- a. From tne C-E Stm Tables.
Tsat for 2200 psia = 649.5 F ; SCM= Tsat-Th= 649.5-580= 69.5 F (+/-1 F) CO.5 ea3 (1.0)
- b. decrease CO.253 t Th increases as unit delta T increases with power CO.53 (0.75) l
- c. 1 CO.253 Core delta T during natural' circulation cooldown will approach full load delta T. Thot i s greater than in the other 2 cases.CO.53 (0.75)
REFERENCE GP HTFF p 356; Steam Tables NEO KAI 3.6 193OO9K115 ..(KA's) i 4 1 ANSWER 1.20 (1.50) 4 l a. Saturated. i
- b. Superheated. ,
a
- c. Superheated.
I REFERENCE l Steam Tables, Mollier diagram NEO GP HTFF p 83,94 KAI 2.8 ; 193OO4K115 ..(KA's) I l l l l l l i i I (ssass CATEGORY 1 CONTINUED ON NEXT PAGE
- sass) )
] I l l 2
,_ _ - _ _ _ , , _ . . _ . . ~ . . _ - , _ - , ,
)
li_IEB]Ng1EbEg_gE_ygg6geB_EgegB_E69NI_gggB8Ilgyz Pcgo 38 IbgBD99XU601951_bg6I IBeNgEg@_gNg,E(919,[(gW ANSWER 1.21 (2.00)
- a. INCREASE ,
- b. INCREASE
- c. DECREASE
- d. DECREASE CO.5 each2 REFERENCE GP HTFF Section 3 Part B sections 2&3. NEO KAI 3.3 OO3OOOK501 ..(AA's)
ANSWER 1.22 (1.00) In the secondary system there is a phase change (0.53. A phase. change requires a large delta h. With the larger delta h of the seconcary, the same heat can be transferred with a lower flow rate LO.53. REFERENCE GP HTFF, Section 3.2 NEO KAI 2.0 193OOOK101 ..(KA's) ANSWER 1.23 (1.00)
- 1. Density difference (or DELTA T) created by heat addition by the heat source and heat removal by the heat sink. (0.5)
- 2. The heat sink must be elev.sted physically above the heat source.(0.5)
REFERENCE VCS, TH SCI, TS-14, P 27, LO 5. KAI 3.9. 193OO8K121 OO6020K304 ..(KA's) (s**** END OF CATEGORY 1 *****)
3 PL9NI,,QE@l@N_lNQ(yQ1NG_S@EgIY_8NQ_gdERGgNQy Pego 39 SXEIEDS +
. i s
ANSWER 2.01 (1.50) Safety Injection Signal CO.33 AND CO.33 2/4 CO.33 , RWST level CO.33 less than 18% CO.33 (1.5) REFERENCE VCS, AB-7, RHR System, p 16. EO-1.3.1 KA1 4.2 006020K304 ~005000K402 ..(KA's) ANSWER 2.02 (2.00)
- a. 1. NaCH isolation' valves (open)
- 2. Scray disenarge isolation valves (coen)
- c. 1. RWST suction valves tcpen)
- 2. Spray pumps (start) CO.5 each3 REFERENCE i
.i VCS, AB-8 RB Spray System, p 15. EO-1.1.4.1.2 , KA1 4.2 ? . 026000K101 ..(KA's)
- +
1 s / ! ANSWER 2.03 (1.50) , i
- a. 6 !
b, every 2 days l
- c. To reduce the site boundary dose that would' occur i f a si ngl e tank ruptured. CO.5 each3 l
REFERENCE i VCS, AB-12, Waste Gas System, p 16. EO-1.4
; KA1 2.5 l- 071000G007 071000K007 ..(KA's) l i !
(ssas* CATEGORY 2 CONTINUED ON.NEXT PAGE * * * **) , l i i
- , . . - - - _ - - - . - - ,_.m. . ._ _ _ . , - , __,-s.-. , _ ,. . . . _ . _ . - _ _ . , ...., - - - I
- 3_)_E6eNI_DEElhN_INQ(gQ16@_@@ Eely _@NQ_gdg6@EUQX Pcgo 40
~'f!SIges :
ANSWER 2.04 (1.50) (Any 3 at 0.5 pts. each)
- a. Rx Vessel Head Vent
- b. Main Feedwater
- c. Emergency Feedwater
- d. Pressurizer Safety Valves REFERENCE VCS, IC-14, Acoustic Leak Monitoring, p. 21 and Operation & Maintenance Manual-for TEC Model 1414-7-(4) EO-1.2 KA! 3.8 OO2OOOK405 ..(KA's)
ANSWER 2.05 (1.00) b t REFERENCE VCS, GS-4, Fuel Handling System, p.4 EO-1.2 KA1 2.4 033OOOK101 ..(KA's) ANSWER 2.06 (1.00) d I i REFERENCE l
)
VCS, AB-4, Reactor Coolant Pump, Figure A84.9 EO-1.3 1 l KAI 3.3 OO3OOOK103 ..(KA's) i i l (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) 1
- 2,* PLANT DESIGN INCLUDJNG_gAFgTY,AND,gMgRGENCY Pcgo'41
, *!!!IEMS
~
ANSWER 2.07 (2.00)
- a. 1. RWST-to-RHR suction valve closed (MVG-8809) (0.33)
- 2. RHRS-to-CVCS isolation valve closed (MVG-8706) (0.33) ;
- 3. RCS pressure less than 425 psig (0.33)
- b. RCS Pressure greater than 700 psig (0.5)
- c. Discharge combined flow of all ( t t' r ee ) charging pumps. (0.5)
REFERENCE VCS, A8-7, RHR System, p 13 & 14 EO-1.1.3 KAI 3.0,3.2,3.2 005000K407 OO5000K402' OO5000K401 ..(KA's) t ANSWER 2.08 (2.00) L
- a. 1. RCP bearing oil coolers tupper and lower) (0.33)
- 2. Excess L/D heat exchanger (0.33)
- 3. RCDT heat excnanger 75h6h (0.33) ,
- b. Thermal barrier isolation valve automatically closes-10.5 pts.)
wnen respective downstream flow exceeds 65 gpm :0." pts.) (1.0) e X gpvy REFERENCE l 1 VCS, 18-2, CCW System, p 4 & 14 EO-1.1,1.3 1 i KA1 3.3,3.0 W 003000K112 008000K102 008000K103 000000K301 ..(KA's) i 1 ' l ANSWER 2.09 (1.50) I
- a. 6 l
- b. 3 l
- c. 2 CO.5 each]
I (s**** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
3.' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pcgo 42
'. 'EY@IED@
REFERENCE VCS, AB-4, Reactor Coolant Pump, p 9, 12, & 13. EO-1.2,1.4 KA1 2.7 003000K602 ..(KA's) ANSWER 2.10 (1.00) CAny 2 at 0.5 each)
- 1. L/D Hx (tube side)
- 2. Seal Water Hx (tube side)
- 3. Letdown Reheat Hx (shell side)
REFERENCE VCS, AB-3, CVCS, p 22. EO-1.1 KAI 3.1 004010K403 ..(KA's) i ANSWER 2.11 (1.50) i a. Condensate Storage Tank
- b. Service Water System
- c. Low suction pressure in Emergency Feep Pump header Co.5 each3 REFERENCE VCS, 18-3 Emergency Feedwater System, p 9L 10. EO-1.1 KAI 3.9 061000K401 ..(KA's) 4
) t***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
. 2. ' ' PL ANT DESIGN INCLUDING SAFETY AND EMERGENCY Pcgo 43
*!YgIgy)
ANSWER 2.12 (2.00)
- a. RWST Cask-loading area Fuel transfer canal SFP Cany 3, 0.33 each3
- b. 1. Minimum water shieldirig depth (9.5 ft. over fuel assemblies limiting the dose rate to 2.5 mr/hr) 0.57-
- 2. To satisfy the Spent Fuel Shipping cask drop criteria (li mi t max velocity of a droppec SF cask to 44 ft/sec) {-OdFP- ,
- 3. )=-g#=:d%d a sc.~ # ; * +- z = : == n s m M 1 rs .et'~< = ' _-- c _ s o% & ' -
REFERENCE D^ ""+ ' " M
- P M *^ ^ ' ="
^#~^~~
C m 2 . o .' s% T -] VCS, GS-5, spent Fuel Pool Cooling Svstem, p 7-10. EO-1.4
-r 5 , f 6 V4 9 - 1 .
KAI 2.7,2.7 033OOOK404 033OOOK105 033OOOK401 ..(KA's) ANSWER 2.13 (2.00)
.SWp
'2
, Jf' 75$'l
- a. 1. Service air isolates 3.C.33 at 90 psig ro 253-(IPV-8324) '
- 2. Standby air compressor starts CO.53 at 70 psig CO 253 (i f lead compressor breaker is closed)
- b. Low S/G 1evel C o . 5 3 - A - - - r # h'W #4 ^ 't N REFERENCE VCS, ACP-220.1, p 1. NEO KAI 3.2,3,4 078000K302 078000K402 ..(KA's)
I i 4 4 (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
3t1_666NI_DggigG_lfjg(yQ1GQ_@@E[Iy,@UQ,[d{@Q[Ngy Pcgo 44
'sygIgus ANSWER 2.14 (2.00)
- a. Electrical CO.53 and Sudden Pressure faults CO.53
- b. Loss voltage High transformer winding temp High combustible gas w/in transformer Low oil level High N2 pressure Low N2 pressure High oil temp Loss of power to the cooling fans C4 required, 0.25 each]
REFERENCE VCS, GS-2, Safeguards Power, p 8. EO-1.1.1.3 KAI 2.6,3.1 062000G008 062OOOK401 ..(KA's) ANSWER 2.15 (1.50) Prevents excessive pressure surge damage to the CVCS regenerative heat ex changer . CO.53 The isolation valves will not open unless all three letdown orifice isolation valves are closed and will close only if all orifice valves are closee. (both letdown isolation valves must be open bef ore orifice valves cars be opened and orifice isolation valves will close if either letdown isolation valve closes). C1.03 REFERENCE VCS, AB-3, CVCS, p 9. EO-1.1 KAI 3.1 OO40 LOK 403 ..(KA's) ANSWER 2.16 (1.00) C (sses* CATEGORY 2 CONTINt'ED ON NEXT PAGE sesss)
2.' PLANT DES 1QN_JNCLUQ1t@_jAFgTY_ANQ,gMgRQgNQX j Pcgo 45
. '!!!IEDE REFERENCE VCS, AB-3, CVCS, p 25. EO-1.2 KAI 3.3,3.5 q 062OOOK301 OO4000K202 ..(KA's) l ANSWER 2.17 (1.00)
MJELb. f ans would start /shi f t to low speed CO.53 and service water system booster pump starts automatically CO.5J ana (4-_'-r u e ? - 9 A sq 'j r- R. 6c.o ash) REFERENCE VCS, TS 3.6.2.3 and A/B-92h NEO KAI 2.9,3.1 022OOOK402 022OOOK301 ...(KA's) ANSWER 2.19 (1.00) Limits plant cooldown rate CO.53 if any one PORV stickt open. CO.53 REFERENCE VCS, TB-2, Main Steam System, P 16 and 05-8, Accident Analysis, P 28. ' LO 1.1. KAI 3.1. 035010K602 012OOOK502 ..<KA's) ANSWER 2.19 (1.00) a REFERENCE VCS, GS-2, SAFEGUARDS POWER SYSTEMS P 36, LO 1.3, 1.4 KAl 3.1. 064000K410 015000K102 ..(KA's) (as*** CATEGORY 2 CONTINUED ON NEXT PAGE assse)
2,
- PLANT DESIGN INCLUDING SAFETY _ANQ_[MERQ[NQy Pego 46 .,
srsIstS l l l l ANSWER 2.20 (1.00) To limit the rate of 5/G blowdown during a main steam line break. REFERENCE VCS, TBS, TB-1, P 16, LO 1.1.1. KAI 2.9. 039000 GOO 7 016000K403 ..tKA's) ANSWER 2.21 (1.00) , a rz G REFERENCE VCS, I&C, IC-9, P 44,59, LO 1.1.2. TBS, TB-5, P 87, LO 1.2.1, 1.4. KAI 3.4 045010K423 016000K108 ..(KA's) l l I
- i. 1 l
l l ! 1 l 2 )
- (***** END OF CATEGORY 2 *****)
i I
- _ _ _ . _ . _ . _ ,._ --_ ,__. _ . - _ , _ __.D
i Peg? 47 j 3:_',1NgI6gMENIg_@NQ,QQNIBO(g l 1 ANSWER 3.01 (2.00)
- a. DECREASE
- b. REMAIN THE SAME
- c. REMAIN THE SAME
- d. REMAIN THE SAME CO.5 each]
REFERENCE VCS, IC-9, Reactor Protection and Logic, p 47,48. EO-1.4 KAI 2.9 012OOOK611 016000K201 ..(KA's) l i ANSWER 3.02 (1.00) C l 1 REFERENCE ; VCS, GS-2, Safeguards Power System, p 27 & 28. EO-1.1 KAI 3.4 015000K102 OO603OK406 ..(KA's) l ANSWER 3.03 (1.00) ANY FIVE AT O.2 POINTS EACH
- 1. PVC-4448 (PORV)
- 2. Hign Pressure Alarm
- 3. B/U Heater Centrol 4 Low Pressure Alarm
- 5. Proportional Heater Control
- e. PCV-444 C (D) (Spray)
REFERENCE VCS, IC-3, Prr Pressure and Level Control System, Figure IC3.8 EO-1.2 . I KAI 2.8,3.8 j 010000K403 016000K403 012OOOK401 ..(KA's) j i I (**s** CATEGORY 3 CONTINUED ON NEXT PAGE *****) ] i i j
~
- 4ti_10lIEWMEUIf_eNQ_QQUI6Q($ Pcgo 48 _;
l
~
.f q
ANSWER .3.04 (1.00) , 1 3co ! owauctioneered)widerangeTh or Tc less than 425' degrees AND any
- HR Suction Valve not fully open. ;
1 REFERENCE ; 1 VCS, IC-3, Par Pressure and Level Control System, p 33. EO-1.3,1.4 , KA1 3.4 016000K108 016000K302 ..(KA's) ANSWER 3.05 (1.50) ! l B/U heaters come on (0.1) and charging flow reduced to minimum (0.1). , 4 Level decreases (0.1). At 17% (0.1) level, letdown i isolates (0.1), heaters turn off (0.1), and alarm (0,1). Level increases (0.1). High level alarm (0.1) at 70*/. (0.1). ! High level trip (0.4) at 92% (0.1). REFERENCE , VCS, Control System Failure Analysis, p 7;IC-3 PZR Press & Lvl Control ; . system, p 37-39. EO-1.4 i $ r KAI 3.4 , 011000A210 016000K201 ..(KA's) I 1 4 i ANSWER 3.06 (1.50) i i j a. 1. 60 second time delay (0.5) >
- 2. Rx Trip Skrs open (P-4) (0.5)
- b. Rx Trip Skrs closed (P-4) (0.5) j i REFERENCE
- i
! VCS, IC-9, RPS, p 52. EO-1.3,1.4 i l j KAI 3.7 l 00603OK406 039000KO15 ..(KA's) ; j i i i ! (ssass CATEGORY 3 CONTINUED ON NEXT PAGE sesse) i , I
. . j
. i Is2,_INSIBWdENIE_eSQ GQUI6QLE Pc90 49 l l
l r s l ANSWER 3.07 (1.00) ; Shutting one sends a trip signal to the other. (Both cannot be shut ; at the same time.) REFERENCE VCG, IC-9, RPS, p 26. EO-1.4 4 KA! 3.7 012000K401 016000K403 ..(KA's) , 1 t ANSWER 3.08 (1.50) l 1- ! I CO.2$ POINTS EACH3 100*/. FLOW 0*/. FLOW I UPPER 65 (60-70) 100 (95-105) NARROW 115 (110-120) 100 (95-105) WIDE 100 (95-105) 30.(25-35) . I REFERENCE ! l .
- .l VCS, IC-13 RVLIS, Figure IC13.7 EO-1.2 l KA1 3.5 016000K303
~ 002000K402 ..(KA's) : i , ANSWER 3.09 (1.00) ! b f
~
) k l REFERENCE VCS, SCP-401, Reactor Protection and Control System, p 15. IC-9, EO-1.6 , KA! 3.0 ,
- 016000A201 001010K507 ..(KA's) !
A 3 l 1 ! I I l .I . > 1 E i i i l (esses CATEGORY 3 CONTINUED ON NEXT PAGE esess) ,
, 2 I i i )
- - - - - . _ - _ - - - - . _ , , _ . .. __. =_. , , , _ . . _ _ _ , _ _ _ _ . - , . _ ~ _ - , . _ - . _ . , .
Pego 50 3_$,_ INS 15gMEUIS_QNQ,CQNIgg(S ANSWER 3.10 (1.50) a a. The condenser is not available for steam dump (2/2 condenser ' pressure and 1/3 Circ Water pump breakers closed).
- b. At least one steam dump valve bank trip-open solenoid valve is energized.
- c. Load rejection arming signal is present (25%/ min). Co.5 each]
REFERENCE VCS, IC-1, Steam Dumps, p 24,28&32. EO-1.3 1 KA1 2.8 , 4 041020G008 001010K410 ..(KA's) ANSWIR 3.11 (1.00)
- a. HAVE NO EFFECT
- b. ARM ONLY CO.5 each]
. REFERENCE 4 VCS, Control System Failure Analysis, p 3&4; IC-1, SDS, EO-1.1,1.2 KA1 2.8 041020G007 012000K501 ..(KA's) ANSWER 3.12 (1.00)
- a. HIGH
- b. LOW Co.5 eacn3 REFERENCE ;
)
VCS, Control System Failure Analysis, p S&9; IC-9, EO-1.4 l l j KA1 3.4,3.6 j i 035000K401 035010A203 006000K107 ..(KA's) i 1 } l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) l 1 i
. . - - - . . ~ . . . , . _ _ , ,
- 3d_leSI69dEUIS_@yD_Cgyl8Q(S Pcgo 51-ANSWER 3.13 (1.50)
- a. Auctioneered High Tavg Co.253 and Tref (f rom Pimp) CO.253
- b. N-44 CO.253 and Pimp CO.253
- c. Reduce contribution of power error at high power (where reactor'is more-responsive).CO.53 REFERENCE VCS,-!C-5, Rod Control,-p 17L18. EO-1.1,1.2 KAI 3.2 001010K507 073000K401 ..(KA's)
ANSWER 3.14 (1.50) SIX AT 0.25 POINTS EACH
- 1. IRM High Flux
- 2. PRM Overpower
- 3. OT Delta T 4 CP Delta T
- 3. Turbine Power < 15 */.
- e. Bank D withdrawal limit (220 steps)
REFERENCE VCS, IC-5, Rod Control, p 37. EO-1.4 KAI 3.2 001010K410 017000G011 017020K401 ..(KA's) ANSWER 3.15 (1.50)
- 1. Tavg
- 2. Pze Pressure
- 3. Delta Flux CO.5 eacn3 I
REFERENCE VCS, IC-6, Temperature Indication System, p 20. EO-1.3 KAI 3.3 012000K501 014000G007 014000K403 ..(KA's) (esses CATEGORY 3 CONTINUED ON NEXT PAGE ssess) l l 1 l l l
3_k,_INSl@yMENIS_@NQ,CQNIQQ6S Pcga 52 ANSWER 3.16 (1.50)
- a. Hi-Hi S/G Level (P-14) (2/3 on 1 S/G)
- b. SI
- c. P-4 in coincidence with Low Tavg (2/3) CO.5 each)
REFERENCE VCS, IC-9, RPS, p 57. EO-1. 4 KA1 2.9 OO6000K107 015000K407 ..(KA's) ANSWER 3.17 (2.00)
- a. Yes CO.23, Closes purge supply and exhaust isolation valves CO.33.
- b. Yes CO.23, Closes surge tank vent valve CO.33.
- c. Yes CO. 23, Diverts flow to nuclear blowdown monitor tank CO.33.
- d. Yes CO.23, Closes waste gas decay tank discharge valve CO.33.
REFERENCE VCS, RMS, p 5,10,11,13,27. NEO l KAI 4.0 ! 073000K401 015000 GOO 5 ..(KA's) ] l l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) 1
3{_1NSI6WDENIS_6NQ,CQNI6Q63 t Pcgo 53 ANSWER 3.10 (3.00)
- a. Wide range Tc Wide range Th Incore TC's 4 ' , q e3 4)
Narrow range PZR press. py,402 Wide range RCS(pot leg) press.A Co.25 each3 (1.25)
- b. Green - margin to saturation greater tnan alarm or caution setpoints Yellow - margin to Saturation is between caution and alarm setpoints Red - margin to saturation is less than the alarm setpoint CO.25 each3
- c. The margin to saturation status has changed since the last - actuation of the "alarm acknowledge" pushbutton. CO.53 The sensor is disabled or out of range. CO.53 REF77ENCE VCS, IC-12, CCM, p 4,5,10. EO-1.2,1.4 KAI 3.4,3.8 017000G011 017020K401 000024K302 000024K301 ..tKA's)
ANSWER 3.19 (2.50)
- a. The roc position indication is reduced to one half its normal accuracy and/or data is supplied by either the data A or B cabinet.
(1.0)
- b. 1. DATA A AND DATA B failure (will also accept causes for data failures)
- 2. Combined rod neight data greater than 38 tequivalent to 228 steps)
- 3. Greater than one bit difference between the data A and B gray coces.
Cany 2, 0.5 each]
- c. On the bottom. CO.53 tasss CATEGORY 3 CONTINUED ON NEXT PAGE ssses)
- 2 __INj!ByMjNIj,@hp_QQNIBQLS Pcgo 54 REFERENCE ,
VCS, IC-4, RPI, p 12,13. EO-1.2,1.4 KA1 3.2,2.8 014000 GOO 7 014000K403 000060KOO5 ..(KA's) 4 ANSWER 3.20 (1.50) ! i
- a. IR CO.23 to allow the operator sufficient time to actuate the ;
SR reactor trip block CO.33 i
- b. PR CO.23 enables the single loop loss of flow reactor trip CO.33
- c. PR to.23 provides the reactor trip / turbine trip interlock CO.33 i REFERENCE i VCS, IC-8, NIS, p 51. EO-4 KAI 3.7 015000K407 OO4000KO14 ..(KA's) i i
l i , l 4 a < 1 s I 1 ) i I i r i 1 i (sesse END OF CATEGORY 3 seest) l l 1 l
dt
- gGQQEQW6(g_ _NQ6d@(t_@8NQ6d@(g_gdgRGgNQ1 Pcgo 55
.eNQ_@@Qlg(Q@lCQ(,QQN15Q6 ANSWER 4.01 (2.50)
- a. 2 Co.53
- b. Within 15 (or next 10) minutes Co.23 either ,
- 1. Restore the indicated AFD to within target band Co.41 or i
- 2. Reduce power to (90% of rated thermal power Co.43. l
- c. Accumulated penalty over the past 24 hrs is 89 min Co.53 The penalty will be reduced to 60 min at 1618 on 03/07/88 and then power may be increased CO.53 I
(OR)
! 85% 0318-0310 = 8 CO.23 65% 1637-1557 = 40 CO.23 45% 0310-0148 = 82/2 = 41 CO.33 89 min total ;
03/07/88, from 1.1571 81 min left - 60 = 21 min > 1618 03/07/88 CO.33 3 REFERENCE VCS TS, 3.2.1 NEO KAI 3.3 ; 015000G005 001050A305 ..(KA's) l 4 1 ANSWER 4.02 (2.50) i
- a. 1. Failure of the reactor makeup control system (such that l
, cypass is necessary to accomplish boration.
- 2. Uncontrolled cocidown NOT requiring SI 1
- 3. Any questionaole shutdown . margin. Co.5 each] )
I
- b. Open emerg. borate valve (MVT-8104) CO.753 I Verif y flow (on F1-110) Co.253 l
i REFERENCE VCS, EOP-11.0, p 1. KA1 4.1,3.9 i 000024A117 000024K301 003000K005 ..(KA's) i i 1 1 ! (sesse CATEGORY 4 CONTINUED ON NEXT PAGE esses; j i i I i i l l
. 4 PROCEDURES - NORMAlt_@BNQ] MALT,@MERGENCY Pcgo 56
' A_N_ D_ _R_ A_ D_ _I O_L_O_ G_ _I C_
_ A_L_ _C_O_N_ T R_O_L_ ANSWER 4.03 (2.00) To initiate corrective actions resulting from an emergency condition is tne only reason allowea during Mode 1 C1.03. In Mode 5 may also leave to verify receipt of annunciators C1.03 REFERENCE VCS, SAP-200, Conduct of Operations, p 12. AAI 3.o 00006SKOO5 000074K011 ..(KA's) ANSWER 4.04 (1.00)
- 1. Reduce letoown to 45 gpm (0.5)
- 2. Control p:r level using (local) FCV-122 cypass valve (XVT-8403)(0.5) i REFERENCE VCS, SOP-102, CVCS, p 37L38.
KAl 3.9 004000G014 000029K312 ..tKA's) ANSWER 4.05 (1.50) i
- 1. Step any changes in reactivity
- 2. Place red control system in MANUAL
- 3. Adjust Tavg by acjusting turbine load C0.5 eacn]
REFERENCE VCS, SOP-400, Rod Control anc Position Incicating System, p 14 KA1 3.1 014000G014 194001A102 ..tKA's) I I ANSWER 4.06 (1.00) I C l l tasses CATEGORY 4 CONTINUED ON NEXT PAGE sosts)
3 . 2 fad _BBQGE99 BEE _ _U98U96t_9EU9056kt_[d[@QgNgy Pcgo 57
.eNQ,69Q1QLQQ1Q@6,QQ$16QL ,
)
6 i i REFERENCE 1 i ! VCS, ECP-1.0, p 10. ! KA! 3.8 i 003000G015 010000K005 ..(KA's) i ANSWER 4.07 (1.00) c . t $ REFERENCE , VCS, EOP-18.2, p 5. KAI 4.0 ! } 000074K311 194001K102 ..(KA's) t >
; ANSWER 4.08 '(2.00) .
- 1. Manually trip turbine from.MCB f
. 2. Stop (and lock out) both'EHC pumpu j 1 3. Runback the turbine f
- 4. Close all MSIVs CO.5 each3 l i
i REFERENCE ; r I VCS, EC'-13.0, p 2. I i L KAI 4.4 j { 000029K312 006000K005 ..(KA's) . I i i ANSWER 4.09 (1.00) f "HIGH-1" signal present l REFERENCE j VCS, ECP Lesson Plan, p 4 J KAI 4.1 194001A102 002000K005 ..tKA's) l i 1 (esses CATEGORY 4 CONTINUED ON NEXT PAGE asses) 2
l
$t$.EE9GEQWEEE_.UQEU66t6EUQ6d6bt-{dgEQ[NQY Pcgo 58 eNQ,6@Q{Q(QQ1Q@(,QQNIQQ(
i
- i
- ANSWER 4.10 (1.50) l t
- a. 2735 psig (0.5) ,
4 b. Reduce pressure to within limit CO.43 within 5 min. CO.33 and notify i the NRC to.33. (1.0) l l ! REFERENCE l l VCS, TS, p. 2-1 l 4 KA1 3.6 l 002000G005 001010A207 ..(KA's) ANSWER 4.11 (1.00) ,
> c CO.25 POINTS EACH3 ')
- a. 1600 psid :
- b. 25 degrees F l
- c. 3000 gpm '
4
- d. 3*/./ h r REFERENCE VCS, GOP-Appendix A, GOP Precautions, p 2,3,&5. )
i l KA! 3.6 { 002000K005 001000K005 ..(KA's) i i ANSWER 4.12 (1.00) 9 l 1. Continue with S/U j 2. Recalculate ECC (if no error found, notify RE) Co.5 each3 I REFERENCE j VCS, GOP-APPENDIX A, p 5. 1 { KA1 3.6 i 001010A207 003000A201 ..(KA's) I 4 1 i i
- (seass CATEGORY 4 CONTINUED ON NEXT PAGE sesse) i i
l j i i
dt).660ggggygg_ _UgGU@(t_@@UQSd@(t_g((RggNgY Pcgg 57
.eNQ_(@Q1g(ggiq@(_CQNISQ(
ANSWER 4.13 (1.50)
- a. Within 15 minutes of commencing Control Bank red withdrawal.
- b. Two, one of whom in SRO licensed.
- c. Within 4 hours prior to criticality. Co.5 eacn3 REFERENCE VCS, GOP-3, p 8.9.
l KAI 3.7 001000K005 013000K015 ..(KA's) l ANSWER 4.14 (1.50)
- 4. Close the RCP A seal leakoff valve (PVT-8141A). CO.53
- b. 5 min. 2,0.53 l c. 30 min. CO.53 REFERENCE VCS, SCP-101, Reactor Coolant System, p 51&S2.
KAl 3.5 l 003000A201 000003K010 ..(KA's) l l l l ANSWER 4.15 (1.50)
- 1. RCS pressure unstable or cecreasing.
l 2. RCS suecooling (based on core exit TC's) <30-F. 1 l 3. PZR level cannot be maintained >4% C39%3 C0.5 eacn3 REFERENCE VCS EOP-1.2, p 4LS. NAl 4.1 013000G015 194001K103 ..iKA's)
)
is**** CATEGCRY 4 CONTINUED ON NEXT PAGE oev ) , I l l 1 \
i
$t'.PBQGEQWBEE.:.SQBdekt_9BSQBdekt.EDEBGESQY Pcgo &O
.eNQ,$$Q1Q(QQlq@(,QQNI@Q( )
i
~
.Ii ANSWER a.16 (1.50) f I
- 1. Decrease turbine load to match Tavg with Tref (+/- 3 to 5-F)
- a.
- 2. Take manual control control of rods (rotate Rod Cntel Sel switch to Man) Co.5 each3 ,
- b. True Co.53 l s
REFERENCE j VCS EOP-10.0, p 13k14 ; r KA1 3.9 ! 000003K010 ..(KA's) i l ANSWER 4.17 (1.50)
- a. WB 1 rem per Atr CO.253 and 4 rem per year CO.253 Skin 6 rem per otr CO.253 ,
Ex 12 rem per qte Co.253 j I
- b. 250 (hal f scale) CO.53 REFERENCE VCS Raciation Fundamentals, p 9,10 17.
KA1 2.8 194001K103 ..(KA's) ANSWER 4.18 (1.50) The RCPs will keep 2 phase flow mixture (0.75) and the FORVs will not ce able to release as much steam (energy). (0.75) OR Higner pressure will reduce 51 flow (0.75) anc increase the inventory flow out of the PORVs. (0.75) teses: CATEGORY 4 CONTINUED ON NEXT PAGE seses)
- i dz!_EE99E99 BEE _ _S96deki_9BygSD962_gd[ggghgf Pcgo 61
',009_5e9196991996_99S1696 .
f i TEFERENCE - VCS, EOP-15.0, P 1. (NO LO AVAIL) Westinghouse B/G document, ERG-HP, FR-S/C/H, FR-H, P 55. 7 KA! 4.1 000074K300 ..(KA's) i ANSWER 4.19 (1.00) ; The turbine is tripped so thet the neat sink will be maintained as ' l long as possible (0.5) on a total loss of feedwater ATWS (0.5). REFERENCE , ERG-HP, Westinghouse Background Information, FR-5.1, P 75-77. . (NO LO AVAIL) 1 KA! 4.4 000029K312 ..(KA's)
?
[ ANSWER 4.20 (1.00)
- 1. Qualified Danger Tagger to.5) ,
- 2. (Current) NRC License (0.5) l REFERENCE ;
VCS, SAP, SAP-201, DANGER TAGGING, P 5, (NO LO AVAILABLE) KA! 3.7. ; 194001K102 ..(KA's) ! { ANSWER 4.21 (1.00) l
- 1. A discussion of existing plant conditions ano anticipated evolutions during tne relief. (0. 5) ;
- 2. A review of the main control ocaro controls, instrumentation i and annunciators. ( 0. 5 )
REFERENCE , 3 VCS, SA?, SAP-200 P 7, (NO LO AVAILABLE) l KA! 2.5. 194001A103 ..tKA's) l l l (esses END OF CATEGORY 4 sanse) l (essssssses END OF EXAMINATION stesseessa)
, Yeo#'fd** Ic"dr"s'eE" Womg gi 29 M Nuckar Op~atons SCE&G m enn March 14, 1988 Mr. Bill Dean License Examiner U.S. Nuclear Regulatory Comission Region II, Suite 2900 101 Marietta Street, N.W.
Atlants, Georgia 30323
Subject:
Virgil C. Sumer Nuclear Station Docket No. 50/395 License No. NPF-12 Operator License Examinations
Dear Mr. Dean:
Enclosed are the utility coments to the NRC Operator and Senior Operator examinations administered at V.C. Sumer on March 7,1988. Your consideration is appreciated. Very truly yours,
!th utb u~-fd 0.A. Nauman GAL:0AN Enclosures cc: (without attachment)
- 0. S. Bradham M. B. Williams A. R. Koon NPCF (withattachment)
K. W. Woodward M. Morgan R. Aiello P.Isaksen(w/ attachment-d0only) File (814.04)
l I l 1 REACTOR OPERATOR EXAM QUESTIONS WITH NRC RESOLUTION
QUESTION: 1.15 List 3 purposes for the Rod insertion Limits. ANSWER KEY RESPONSE:
- 1. Adequate 5DM upon trip
- 2. To minimize the amount of positive reactivity inserted during an ejection accident, and
- 3. To minimize radial flux tilt (peaking)
REFERENCES. VCS, RT B K lh, RT- 14, P. 20 SUGGESTED CORRECT RESPONSE Either the above answer, or. To ensure that:
- 1) Acceptable power distnbution limits are maintained,
- 2) The minimum SHUTDOWN M ARGIN is maintained, and
- 3) The potential effects of rod misalignment on associated accident analyses are limited.
REASON: Although the answer key response lists the "standard" reasons for maintaining rods above the RIL, the alternate answer is found in V.C. Summer Plant Technical Specifications, page B 3/41-3 (attached) as the bases for control rod insertion limits. NRC RES01.IITION: Comment accepted answer key and references modified to include additional correct answer either accepted for full credit. 4 e
~--,,m-- - . --, , . -., n.- , - , -- - - ---,-
Qasbn 1.15 REACTIVITY CONTROL SYSTEMS BASES 80 RATION SYSTEMS (Continued) i MARGIN from expected operating conditions of 1.77% delta k/k or as required by Figure 3.1-3 after xenon decay and cooldown to 200*F. The maxieue expectact boration capability requirement occurs free full power equilibrium xenon candi-tions and is satisfied by 12475 gallons of 7000 ppe borated water free the coric acid storage tanks or 64,040 gallons of 2300 ppe borated water from the refuel-ing water storage tank. With the ROS temperature below 200*F, one injection systan is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single injection systes becomes inoperable. The limitation for a saximus of one centrifugal charging pump to be OPERA 8LE and the Sueve111ance Requirement to verify all charging pumps exu pt the required OPERACLE pump to be inoperable belcw 275'F prov< des assurar.ca that a mass addition pressure transient can be relieved by the operation of a single PORV. The boren capability required below 200*F is sufficiant to provide the required SHUTDOW MMIGIN of 1 percent delta k/k or as required by Figure 3.1-3 after xenon decay and cooldown from 200*F to 140*F. This condition is satisfied by either 2000 gallons of 7000 ppe borated water fror che boric acid storage tank. or 9690 gallons of 2300 ppe berated water free the refueling water storage tanks The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics. The OPERA 81LITY of one beron injection system duri REFUELING ensures i thatthissystemisavailableforreactivitycontrolwhqt.4inM00E6. 314.1.3 MOVA R E CONTRE AS$DSLIES The specifications of this section ensure that (ITectaptable power-distribution 1fstta"are'seintained. (2) the minimum SWTOOm MARGIN is main-tained, and (31.11ait the~petentfai~effectsWr~od afs'aTijiuibatron associated accident analyses. OPERASILITY of the control rod posir. ion indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion Itaits.
- - - - , , , ,-,,-,----.,,-,,,,,-,,.---,,,.,,,,_-w.,-.-.--,,,-,,,..,._n,n
QUESTION: 2.08(b) Other then relief valves, what feature prevents overpressurization of the CC'J/ System if a thermal barrier heat exchanger tube ruptures? .(include setpoint,if applicable). ANSVVER KEY RESPONSE: . Thermal barrier isolation valve isolation automatically closed (0.5 points) when respective downstream flow exceeds 65 gpm (0.5 pts.). (1.0)
REFERENCE:
VCS,18-2, CCW System, p 4 & 14 SUGGESTED RESPONSE. As given above with 0.8 points far action and 0.2 points for setpoint. REASON: Knowing that the CCW system is protected from RCS pressure upon thermal barrier failure by the closure of the isolation valve on high flow is far more important than knowing the associated setpoint. Only reactor trip, safety injection, permissive and control interlocks, steam and feedwater isolation signals, RB spray signal, and other 55P5 setpoints should carry such high weighting. An 80%/20% split between action and setpoint is more amenable for a feature such as this. NRC RESOI.llTION: Comment noted. It is also significant to know when an automatic feature should have or have not actuated. To be consistent with other sections of the exam the answer key will be modified to 0.75 for the protective feature and 0.25 for the setpoint. i
I QUESTION: 2.13 (a) Describe TWO automatic actions associated with the instrument air system which serve to mitigate a loss of air pressure. Include any associated setpoints. ANSWER KEY RESPONSE.
- a. 1. Service air isolated (0.5) at 90 psig (0.25)(IPV-8324)
- 2. Standby air compressor starts (0.5) at 70 psig (0.25) (if lead compressor breaker is closed).
REFERENCE:
VLS, AOP-220.1, p 1 SUGGESTED ADDITIONAL RESPONSE: As given above with 0.6 for action and 0.15 for setpoint. - REASON: Knowing that service air isolates is afar more significant than knowledge of the setpoint at which the action occurs. Only reactor trip, safety injection, permissive and controlinterlocks, steam and feedwater isolation signals, RB - spray signal, and other SSPS setpoints should carry such high weighting. An 80aA/20% split between action and setpoint is more amenable for a feature of t its nature. NRC RESOLUTION: Comment noted. It is also significant to know whkn an automatic feature should have or have not performed its protective action. No cha.'ge to answer key.
QUESTION: 2.17 What automatic action (s) should occur to the Reactor Building Couling Units upon receipt of an SIAS signal? Assume an initial normal at power lineup. ANSWER KEY RESPON5E. All fans would start / shift to low speed (0.5) and service water system booster pump starts automatically (0.S)
REFERENCES:
VCS, TS 3.6.2.3 and AB -9 SUGGESTED RESPONSE: Selected fans (1 in each train) would start / shift to low speed (0.S) and service water system oooster pumps start automatically (0.S) (Industrial cooling water supply to R 8 C U's isola tes). REASON: Main control board selector switches allow the operator to choose one fan in Train A and one in Train B for the RBCU's. Only these fans will shift to slow speed upon a safety injection signal. See attached drawing showing selector switches for each train. NRC RESOLllTIGl!: Comment accepted. This was a CAF (Check at Facility) since information was not supplied to prepare the examination. Answer key modified as suggested. - h t i 9
+
t F
,. . i. .. i ....,,..,,..._,.,.,;.,,,___,.._ .
._-,is..~m.,. n._--.
s
> sws *:, i
;1 I.,
.y9/. a.d g.1 *
- Je
}D . ..,'.-..e'q,;w.- t. a. "g y . l 5 9 .w ,.i gj i s .-.
., *- 5> %y(,i; 7 t
.. 0 315 ( . f. )^ , ,. ; s' f
. ..% ! d O l.$ $ j d.
l . ,l, [ c a er ev.),-=w w.ly , q:,,,,,J, I
, . c.. :, w !pg h . .y
'j ; -
;i.l'
. i5.
['N , ':,ss;(! it , s 0--*& -- i l4 e, Lc-v,bj . . e tqJ o
- s ..'
s 1 ...,v hi - :r..g . "q! lli*- o 9; i yy 4 e .
.. % 7 ,'
.._ . . l
'4 .f 1 7 c
5 '#
- g. ~'z ' - f' ' - '
u NlII ; es
. "f ue..y' g
CI g Il\ = WfYf iD. ' fg s p * .' q,,..ea-..gi....'p;f,f. d :- y
. . 6..o .
.y -
= 1. '
! epy--c
.- lii,:,,' pll9.,.:;. - ~'
- p . . ..4ng 3
';,! 3, . i ! vs jM t j,i .I -
f,l . = 0. svs'm i
! , .:a ! ; e I ;liN t ,I: .E4 .e a.
4 ups , g M/.. . . N g.li s
<i
-,d
'MW w,. w .
?~ r $sgg
[f; i '< Df'.' .'$ g..4.?.-.,7 s ., , i. y Mym&j N fj/ ,k, %p ,g!Mp*\\, NO 4 d 'h !, 'i s'M:ih ' h j i
/'n.,8r$.4}y U .' ' ".
5 T '
@b ej "/- m M
i s 4'
'e h
Lua.4% . (J . . i/ II. f in Tow *vi 1' ,
!/ .:l
~
! e 'l>q' V. --
- 4;!( ',, ch "iNA . i' y n%, + , :ll 'J u ~ u ,a WK,r. . .'.g.M *
,J5,inm f-D!
a
,' V l
!'l g..:Q l+ ;!_md,
! ' i~ a s '
I i
.= ~ g .. ,
4
,=a8 etr-----
a ri Ja
*v N's ' ' k['
kb ,.! h, ll( I
.,t k,.
r' ... -- s ,
. s
,1(,1 l f, ' '.
i KI .h , l , f *i ik, "
'; ,[ y. k t i ,
- f. I ,ki ti ,r j j ":j ,
- d. , %{i @(f% s.s. &, /, m ' gd.y' '
l I,(s
* /. Ng.-
- t A c M,* 8
' * ' t' .k4
- 4 s
8 ,, ynw'E$$, 3 .b . ,i d. -" -- 'd @ Qaig, x ;. j; - ( s i ?Ji s g) -
' G ',t';
,L j
de
/' ~ <N--r j
@ $f ,'f ,t. ,
y,. ,1 0 g til .
., [ If!! j-
/ ! f g r, y t e , w"t5 w
- j :: l
/ .
i I.. , /(/y l > 3 3.3 ?i f i= li.! e ,i\x.47.J-- -- ' i!h r g~"y" e .".i-
.w. . . .. .. . :
- ,s
.N
' , pl
, 4 .Ji 9 . e '3 , .59 p, a ..o
. 'iM'
~
.li #N I!'7" ;m . .( /)
-jlli-M'll' I ~
i 4-- e9-w d'i' hm 't'Ei
- e. 8 ,9w ; ,.g. 4,:w.m w.gtvf.4 - . .
.o,,".,
'"t
.,j L_d er;-. Nt i
i if Lg#is' 'nv c.w. r v i i R.i.., . a i .l. i
- 7 ,i i -c P -r {g a H:p,H iu l }4 5
.i ift + 'r 5
** L gO, gM l.<: T i(4 : f r b'T .i:1'.f iD-+'f~~ .,Y ' . ...%,
\
. . Q ,1*Mi 4
Ig m m,,o @. ,L i' a . e m , em %.,
~ .
1 i a f ikmM \ .- m i i 34 $. ~- % m
...-. .. m .
n..mMI. w
., .... w.n.
. . ~
_3 L.X 'l **4% .-
t QUESTION: 2.21 Which statement regaroing the Main Generator Protection System is NOT CORRECT. a) Opening the generator output breakers ALWAYS results in a turbine trip when the generator is loaded. b) Once the generator is loaded, a turbine trip ALWAYS results in a generator trip. c) A turbine trip above the protection interlock P-7 (10% power) ALWAYS results in a reactor trip. d) A reactor trip ALWAYS results in a turbine trip. ANSWER KEY RESPONSE. a) Opening the generator output breakers ALWAYS results in a turb ne trip when the generator is loaded.
REFERENCE:
VCS, IC-9, P. 44, 59, TB 5, P. 87 SUGGESTED ADDITIONAL RESPONSE. (c) A turbine trip above the protection interlock P-7 (109o power) ALWAYS results in a reactor trip. REASON: While choice (a)is certainly a correct answer to the question, choice (c) is also correct. The question asked which choice is NOT TRUE. At V,C. Summer, a turbine trip above P 7 will not always result in a reactor trip. At one time this was the case. However, now, the P 9 interlock (>50% power) enables the reactor trip upon turbine trip. This is documented by the attached references, IC-8, Nuclear instrumentation, p. 29, and IC 9, Reactor Protection and Logic, p. 45. NRC RESOLUTION: Comment accepted answer key modified to accept a or c. .
Qa.0 % L . *L \
' more 09 Ene four power range channels exceed the trio setooint, the bistable will cause a reactor trip and actuate the reaccor trip first-out annunciator "PR FLUX HI SETPT HI" on C8P 10. This trip function cannot be blocked.
o P-8 pemissive circuit. When - the outputs of two of four power range r.hannels exceed the P-8 setpoint (38 percent oower), a loss of flow, as sensed by two of three flow transmitters in any one loop, will cause a reactor trip. Between readings of 38 per::ent power (P-8) and 10 percent power (P-7), flow must be lost in at least two loops to cause a reactor trip. Selow P-7 there are no loss of flow trips. As each power range channel exceeds 38 percent power, the respective "CHAN I (II, III, IV) P8" light on the reactor pemissives panel energizes. When two of four channels exceed 38 percent power, the "P8" light on the reactor pemissives panel on CSP 9 energizes, o P-10 permissive circuit. This permissive is automatically enabled when two of four power range channels exceed 10 percent. Actuation of P-10 allows the manual blocking of the power range low setpoint trip, the intermediate range trip, and the intemediate range rod stop. It also blocks the source range trip and provides a portion of the P-7 signal . As each power range channel exceeds 10 percent _ power, the respective "CHAN I (II, III, IV) P10" light on the reactor pemissives panel energizes. When two of four channels exceed 10 oercent power, the "P10" light on the reactor permissives panel on C8P 9 energizes. The "P7" light also energizes because its logic is satisfied den either P-10 or P-13 (1/2 turbine impulse pressure greater than 10 percent) is satisfied. I o P-9 permissive circuit. As reactor power exceeds 50% indication for two of the four power range channels, a turbine trip will activate a reactor trip. At power levels less than 50% (3/4), the reactor trip resulting from a turbine trip will be blocked. The differential amplifier provides an output signal proportional to the rate of change of reactor power to two separate bistables. The first bistable will trip when a sudden increase in neutron flux occurs (+5 percent of full power within 2 seconds). When this bistable trips, it will actuate "CHAN I (II, III, IV) PR FLUX RATE HI" on the reactor bistable 29 1068S:4
Q - . W. , s 1. q A separate permissive, P-8, will automatically block the trip that occurs wnen flow is lost in ono loop. This occurs when three of four power ranga channels are below the setpoint of 38 percent. Again, the reacte .s adequately pro-tected below the setpoint without the trip. If allowed in the future, it would be possible to continue limited power operation with an inoperable loop. The AND gate of the P-8 permissive logic denoted with a 2 of 4 coincidence actually energizes above the P-8 setpoint and the NOT gate will extinguish the permis-sive st.atus light. To deenergize the AND bistable, 3 of 4 power ranga channels " " must be below the P-8 setpoint. The P-8 permissive status light energizes and the zero input to the AND gate of the low flow logic blocks a reactor trip fran a loss of flow in one loop. P-7 works in a similar manner. P-9 will prevent the reactor trip that normally results from a turbine trip. The setpoint for tnis additional permissive is 50f; power with a coincidence of 2 out of 4 power range channels. Rod Stops (Sh. 4 and Sh. 9) The C-1 high neutron flux rod stop will block both automatic and manual rod withdrawal. This occurs if the cperator does not block the rod stop prior to either a normal or an unexpected power increase. Any out motion is then stopped if one of two intermediate range channels exceeds an amperage output that is equivalent to 20 percent of rated thermal power ( =10-5 amps) . This rod stop is manually blocked at the same time the operator blocks the inter mediate range high flux reactor trip. In fact, the same switches that are used to block the trips are also used to prevent the rod stop. Again, attempts to block the C-1 function prior to 2 of 4 power range channels exceeding P-10 will be unsuccessful. The rod stop is automatically reinstated when 3 of 4 power range channels fall below 10 percent. The C.2 overpower rod stop also blocks automatic and manual control rod with-drawal. The block action occurs when one of four power range channels rises above 103 percent power. This is the only function associated with the Reactor Protection System that actuates on a 1 of 4 coincidence. Because of this, special switches called Rod Stop Bypass Switches are provided on the flux deviation, miscellaneous control and indication drawer of NIS. These switches allow continued normal rod withdrawal if a power range channel snould fail high or if a channel should require testing. 45 11555:4 l Rev 1 5/85 t
QUESTION . 3.04 What plant conditions will cause the Cold Overpressure Protection System (COPS) alarm to actuate? ANSWER KEY RESPONSE: Low auctioneered waae range T, or T, less than 350"F AND any RHR Suction valve not fully open REFERENCE. VCS,IC-3, PZR PRESSURE AND LEVEL CONTROL SYSTEM, p. 33 SUGGESTED CORRECT RESPONSE. Any wide range T. or T, less than 300*F AND any RHR Suction valve not fully open. REASON: Annunciator Response Procedure, ARP-001 XCP 610 (attached) clearly specifies the conditions which cause actuation of the alarm. This is backed up by Technical Specification 3.4 9.3 (attached), which is the basis for the alarm. NRC RES01.IITION: The answer key response was 325 F (not 350 F) and will be modified to 300 F and "Low Auctioneered" will be placed in parenthesis,
0-d.., 3,oq !
' ARP-001 REVISION 3 n
1/22/85 PANEL XCP-0610 (NSSS) ANNUNCIATOR. POINT 2-4 RCS TEMP LC AND RHR SUCT *' SETPOINT_: ORIGIN VLVS NOT OPN RCS temp <300*P and TY/410J any RER auction valve TY/413K not fully open 33X-RH04 PROBASLE CAUSE: 33X-RH06
- 1. Failed low temperature instrument.
2. RER suction valve (8701A, 87018, 8702A, 87023) not fully open. AUTOMATIC ACTIONS None IMMEDIATE ACTIONS: 1.
- 2. Verify RHR Verify RCS suction temperature valves<8701A, 300*P as read on ITI-410 and ITI-413.
8701B, 8702A, and 8702B fully open.
~
SUPPLEMENTAL ACTION:
- 1. Ensure at 2.
by an onservice RHR suction relier valve.least one train Cold Overpres Refer to Technical Specification 3.4 9 3.
REFERENCES:
- 1. GAI DWG. B-208-082, sh. 73
- 2. GAI DWG. B-208-084, RH-03
- 3. GAI DWG. B-208-084, RH-04
- 4. GAI DWG. B-208-084, RH-05
- 5. GAI OWG. B-208-084, RH-06
- 6. V.C. Summer Technical Specifications
- 7. SOP-115 l
l l PAGE 11 or 17 l 1
, com ., 3.o r 1
1 1 1
.......:..1 . .. . : :i.. :
, r.yg r yc .- sc. ...y :.24 .; :.*4 .". 9
- s. . .
3.43.3 A: 'eas: e .:::ag,..::ne
..- :' : e '::-ing :ve crossure : ::ect :
i 5:e s 5 1
- 3. .o awa e. e' ,a;,es .its.
- ; se:: ; :
.. i
'es s : an :r ecuai :: 450 :s ;. 1 : l 2.
e asso:~3:40 0'4 - el'e# ,aI ee o. iolati ' .ai,es ::em; :- '
!"e 4eac: :-an::::etart greater e:ua' 5 s:em:: 2.~ (403) Jec*essuri 4c ai scuare '9cnes. tn an RCi <ent :'
ADot!CABILI"Y-
*00L 4 wnen tne temoe-sture of any RC5 colo l 300*77 uc0E 5. se: eg is less than orrecue* to SOCE 5 mitn :ne eact:r vessel 9eac on.
..N:
i
- 3. < i te :.,e 3-4 e i e f , a l .e 'ne:erno ie res- e ne tc OP59ABLE 59Pougn 3 gr93*a 5:stus: an :e
.itnin 7 :ays :r c,ecre: . r':eme an'c :eracle . i .e Nes t 3 nours. . eat t e 3: 1
. u a l *. - 2. ' iC .1'e ' 7 c.9 .ent ai:ni1 e at:- OC:n
-4 #elief ,a!.e5 '** ern0'e
' mitr'- 3 :.ri d ' '. e - '
Iest:re at least one RhR relief '
- 2. .alve :: ::34A3LE itat.i . ;-
Secres3uriZe
.: and vent the ICS tnr3bgn 1
;"ea:ee : !
2.7 scuare inen vent. : & .a ' c.
- - . e e,ent an RHR relief valve or RC5 vent IC3 : essure transient, a $cecial Recort sna;' i .sec O :: - . ;a:e t-e
- :a,s. to the C0mmission pursuant to Scec:steCaee0 3"c -'-
i.:S'iteQ '
- 09 5. i. s a
- t s ent, e-The recort snali descrice tne cir:.mstances 131 - :e :
1 0 any tne ef *ectaction c:rrectt.e of tnenecessary RhR relief to val,er, or ,ent :n : e :;1 s-! t 1.
- re,ent ec.r eace.
e :rovis': s of 5:aci:stion 3.0.4 are act: s. 1:0'i:n ' i l I i i.vuga
.st
- 6 3: 1-34 .
Ameacment No. M
t QUESTION: 3.06(b) After resetting safety injection, automatic actuation is inhibited until what signalis cleared? (0.5) ANSWER KEY RESPONSE:
- b. Rx Trip 8krs closed (P-4)
REFERENCES:
VCS, IC-9, RPS, p 52. EOP- 12,14 SUGGESTED RESPONSE.
- b. P-4 must be cleared. (it is cleared by closing the reactor trip breakers).
REASON: The intent of the answer key appears to be correct. The question asked only "what signal must ce cleared". However, to provide clarification, a logic drawing is attacnea showing that the P 4 signal must be removed in order to enable a subsequent auto 51 signal. The P 4 signal is cleared by closing the reactor trip breakers NRC RESOLllTION: Ccmment noted. Either response will be given full credit since the action to clear P-4 is to close the reactor trip breakers. No change to the answer key is required.
n i l e s'1MS-41-01150-6 l I ld51je "j 21 .Mif i 5.
- OI*
3 _ , . f ig! -{!fl!!] i (i,1 h-]' 1b. Mi ,!l
,b'i 2I j
wi
/)
ai
, y'r;jrm! i l 'o '
< 1
. p&
f
.. t - a - ' .
~ '
.l
,lg r.
- .. I ai le lill!
= ..
- ,-: / .
i
'p! ,
, , f- , fi*Iilni}!\ig MEng!. '
, 'a
.l_
lg , r .v,_ Y/[
' 'f-c-c-
J ltD jll II!ljf !! lar i,H I
.l.
T - 2 . l cm s ii 11 3. i 1 , 6pg I ja M l o
- 6 4 !. .
ia. 3n I m. l t ar 4: iiiii,a w gr ;;s !;; .. i !,!D iT-gg Di-r 1 - L ikin.':i I ,glj!ll,i jlgNg Is iii i"
.IIj=lihrj,
- tir i i 800 5 .' ! I s i $i i: :i or iil ggj; .)
i ',7
! T ..
( fI ?IJ s3::
-r a
- p I 5 4 le .t .::
- 5. . ,M Fi!
! iTrz-Pr{ Ilj m
a . w- ,. '
- i pl .
Till!i l ecci_
.$t- l Il
[ m
.il.1'
- 'i
=st=
io lj. .
- l. 3-6ggd~
' b hin i,i :Esi j
'C; o,
i m '. l
- r:
---I Jij1* IIIUI'I! M -
g ia==:e t
!hj " bk.
? M_ 'i' I!'Illri i i:=
l] flh!!-hlj)1.l50
- lliD jnki!! ")R lJl Eid r a
g a 3, -**. p;,y IIjI . k7JU!' hLe iig , (It ll, ,_,- g g litt . h n [, (, - h
..ulb
.s;, [j g ... -
. ,!f ~~ S
. . , , . . . . .. . . =
l
,.E .
(mot " P"o _
QUESTION: 3.18 a. What are the FIVE inputs to the CCM? Be specific. ANSWER KEY RESPONSE; Wide range Tc Wide range Th incore TC s Narrow range PZR press. Wide range RCS hot leg press. (0.25 each)
REFERENCES:
VCS, IC-12, CCM, p 4.5,10. EOP- 1.2,1.4 SUGGESTED CORRECT RESPONSE. Wide range Tc Wide range Th Incore TC's Narrow range PZR press. Wide range RCS press. (0.25 each) REASON The only wide range pressure transmitters on the RCS (PT-40im, PT-403A) are located on the Loop A and C RCS hot legs (see attached drawings). Therefore, requiring the words "hot leg" in conjunction v'ith the wide range pressure instruments is noth redundant and unnecessary. NRC RESOLilTION: Co; cent noted. The reference uses the terms wide range and hot log prersure in several places. The answer key will be modified to place "hot leg" in parenthesis and include PT-402A, -403A in parenthesis and wide range RCS pressure will be accepted for full credit.
l QUESTION: 4.17 a. What are the V.C. Summer Nuclear Station Administrative exposure limits for Whole Body, Skin, and extremities for occupational workers? (exclude fertile females) ANSWER KEY RESPONSE:
- a. WB 1 rem per qtr (0.25) and 4 rem per year (0.25)
Skin 6 rem per qtr (0.25) Ex 12 rem per qtr (0.25) l
REFERENCES:
VCS Radiation Fundamentals, p. 9,1017 SUGGESTED CORRECT RESPONSE: WB 1 rem per qtr. (0 33) Skin 6 rem per qtr. (0.33) Ex 12 rem per qtr. (0.33) REASON: Plant limits are expressed in terms of maximum quarterly exposure. Requiring 4 rem per year as well as 1 rem per qtr. is redundant. Neither tne skin or extremities limits were extrapolated to yearly values. Quarterly limits . are more restrictive and should be all that is required. Deletion of this item ! and reassignment of points to one-third point for each limitis requested. I NRC RESOLUTION: l Comment not accepted. The reference specifically addresses both quarterly and whole body limits and no additional supportive reference material supplied with these utility comments justifies any change to the answer key. I
S Additional changes made to answer key: 2.12b Also accepted TS basis for water level concerning removal of 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly and reference added. 2.13b Clarified answer to accept Low-low S/G level trip in addition to low S/G level.
U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: SUMMER REACTOR TYPE: PWR-WEC3 DATE ADMINISTERED: 88/03/07 EXAMINER: AIELLO. RF CANDIDATE: JNSTRUCTIONS TO CANDIDATE 1 Use separate paper for the answers. Write answers on one side only. Staple question sheat on top of the answer sheets. Points for each question are indicated in parentheses after the que_._.. The passing grade requires at least 70% in each category and a final 7rade of at least 50%. Examination papers will be picked up six (6) hcurs after , the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY A(.T 11.C
( "^ ~~ c S. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THEFMODYNAMICS 3/ Q 2GV
~^ "" '" -"
4 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION gc .1 C L T. '] X ~* ~~ "I 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL ti. _6 , 29.75 K_"' 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS
>% 2'","5 % Totals Final Grade .u /L4 a g p 2 All work done on this examination is my own. I have neither given ner received aid.
Car.didate's Signature
. ~ _ - - - - - - - . - - -...- - -- _ _- _ ..-_ - .....~ . -.-.. _.-.....--. - -.-
NRC RULEG AND GUIDELINES FOR LICENSE EXAMINATION 9-During the administration of this examination the following rules apply:
- 1. Cheating on the examination means an automatic denial of your application !
and could result in more severe penalties. i i' 2. Restroom trips are to be limited and only one candidate at a time may l leave. You must avoid all contac ts with anyone outside the examination l i room to avoid even the appearance or possibility of cheating. I
- 3. Use black ink or dark pencil nly o to facilitate legible reproductions.
- 4. Print your name in the blank provided~on the cover sheet of the ;
examination, t q D. Fill in the date on the cover sheet of the examination (if necessary). !
- 6. Uuo only tne paper provided for a n *2 W a r n . '
- 7. Print your name in the upper right-hand corner of the first page of ca_q1' l ucction of the answer sheet. !
G. Consecutively number each answer sheet, write "End of Category ,,__" as ! 4 appropriate, utart each category on a new page, write gn_1y_ on one nJ,dr _ [ of the paper, and write "Last Page" on the last annwer sheet. [
- '/ . Number each answer as to category and number, for example, 1.4, 6.3.
l
- 10. Skip at least three 1ines between each answer, 1 11. Separate anuwer sheets from pad and place finished annwer sheets face down on your desk or table.
! 12. Une abbreviationn only if they are commanly used in facility literature.
- 13. The point value for each queution in indicatud in parentheuet after the question and can be used an a guide for the depth of anuwer required.
I
! 14. Show all calculationu, methods, or asuumptionu used to obtain an answer i to mathematical problems whether indf.cated in the quention or not.
l j 15. Partial credit may be given. Therefore, ANSWER ALL PARTC OF THE j QUESTION AND DO NOT LEAVE ANY ANSWER DLANK. 1 q 16. If parts of the examination are not cluar an to intent, auk questionu of a the q;1aginer only. i,
- 17. You munt nign the statement on the cover uheet that indicaten that the work in your ov'a and you have not received or been given asuistance in completing to examination. Thin mut>t be done after the examination hau I been completed.
l A I i _- - .-
. - - - _ . - . - - _ _ - . - . - - - - - ~. .- .. _ .. - . . . . -
1-_
- 10. 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 vour copy of the examination and all pages used to answer the examination questions.
l
- c. Turn in all scrap paper and the balance of the paper that you did ~ l 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 whilf> the examination is still in progress, your license may be denied or revoked.
I i l l t 4 1 i t i i l l i i l
-w w-*me--m-_--~-.-ve-m*
~
[ 5. -TH50RY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE- .2 [ 4
.-THERMODYNAMICS I
a ! QUESTION 5.01 (1.00) Given 3 reactor coolant (RCP) pumps operating in parallel, each with l 1 -a flow rate "m" and a combined flow rate "M". Which one of the i following best describes the system response to securing one RCP7 ^
- a. The resulting core flow (M) will_ decrease _as individual operating RCP flows-(m) decrease.-
y i b. The resulting_ core flow (M) will increase along with individual operating RCP flows (m). i c. The resulting core flow (M) will decrease as individual l operating RCP flows (m) increase. 1 i d. The resulting core flow (M) aill not change due to decrease in RCP back pressure. i l 1 l 1 1 4 l l l t l l l r f (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) I t l
S. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND -PAGE 3 l THERMODYNAMICS { l l l l i { GUESTION 5.02 (1.00) i l Shich one of the following statements BEST describes' Xenon. behavior, over the first few hours, on a power decrease following 100 hours at 100% power? NOTE: [Xe] denotes xenon concentration
- a. Direct [Xe] increases, indirect [Xe] decreases, total [Xe] l decreases,
- b. Direct [Xe] increases, indirect [Xe] increases, total.[Xe] !
increases. l
- c. Direct [Xe] decreases, indirect [Xe] decreases, total [Xe]
decreases. l
- d. Direct [Xe] decreases, indirect [Xe] increases, total [Xe] !
increases.
- e. Direct EXe] decreases, indirect [Xe] increases, total [Xe] !
decreases- ; I I i l i t i l I (***** CATEOORY 05 CONTINUED ON NEXT PAGE *****)
S. THEORY OF NUCLEAR-POWER PLANT OPERATION. FLUIDS. AND PAGE 4
-THERMODYNAMICS i
GUESTION S.03 (1.00) 1 Which ONE of the following expresses the relationship between difforential rod worth (DRW) and integral rod worth (IRW)? l
- a. DRW is the slope of the-IRW curve at that location. ;
I " l
- b. DRW is the area under the IRW curve at that location. !
i ! j c. DRW is the square root of the IRW at that location. i t
- d. There is no relationship between DRW and IRW. l l
1 i I i, ! i 1 i> l !
+
1 [ , l ii ! e ! I l 4 1 i 1 l l, 1 i i i l i 4
- (***** CATEGORY 05 CONTINUED ON NEXT PAGE t****)
l 4
i S. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE- 5 THERMODYNAMICS i
; OUESTION 5.04 (1.00)
The reactor trips from full power, equilibrium xenon conditions. Six hours later the reactor is brought critical at iOE-9 amps on the inter-mediate range. If power level is maintained at iOE-9 amps, which ONE of the following statements, concerning rod motion requirements for the next two hours, is correct? i a. Rods will have to be withdrawn since xenon will closely
' follow its normal build-in rate following a trip.
- b. Rods will have to be inserted since xenon will closely follow its normal decay rate following a trip.
- c. Rods will have to be rapidly inserted since the critical reactor will cause a high rate of burnout,
- d. Rods will have to be rapidly withdrawn since the critical reactor will cause a higher than normal rate of build-in.
l f (***** CATEGORY 05 CON *INUED ON IJEXT PAGE *****)
7 -.- -... - _ _ ._-. .- - . S. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 6 THERMODYNAMICS QUESTION 5.05 (1.00) i Which ONE of the following correctly describes the observed reactor
- response for the same small addition of reactivity, one positive and one negative?
[ l a. The response will be faster for the negative addition at all times in core life. ! b. The response will be faster for the negative addition at BOL. but faster for the positive addition at EOL. 1 l c. The response will be faster for the positive addition at all l times in core life,
- d. The response will be faster for the positive addition at DOL but faster for the negative addition at EOL.
I j e. The response will be the same for both the positive and negative addition. h h 2 ; i d i I I l i e i i-t (***** CATEGORY OS CONTINUED ON NEXT PAGE *****) i i ] L. _ _ .
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS.'AND PAGE 7 THERMODYNAMICS QUESTION 5.06 (1.00)
Indicate whether each of the following will make the moderator temperature coefficient LESS NEGATIVE, M O R E t?C T,^ r I V E , or have NO EFFECT.
- a. INCREASE moderator temperature
- b. DECREASE baron concentration
(***** CATECORY OS CONTINUED ON NEXT PAGE *****)
,S . THEORY OF. NUCLEAR FOWER PLANT OPERATION. FLUIDS. AN_D, PAGE 8 THERMODYNAMICS
^
QUESTION S.07 "') State the change (INCREASES , DECREP.SES or REMAIN THE SAME) in the magnitude of the fuel temperature coefficient (FTC) for each of the following: A. An :;r.c rea ne in power. E ~ 3: s
- o f t s' r i o /? fs 0 3/2 y/??
C. A decrease in magnitude of the moderator temooreture ) coefficient (MTC), i j 23GRY 05 CONTINUED ON NEXT PAGE *****)
- 5. THEORY t'
- 1UCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 9 .
THERMOI._ :ICS 4 1 I l-QUESTION S.08 (1.00)
\
A centrifugal pump is started up with its discharge valve open. How ! would this following parameters diffor(HIGHER, LOWER, THE SAME) if the ! punp had been started with its discharge valve shut?
- a. Motor current I
i b. Discharge prestsure l l l I l i 1 1 a 1 l j i !! l t ! ! l ) ) - l 1 i i i ) l 4 i 1 4 4 f ! (***** CATEGORY t.3 CONTINUED ON NEXT PAGE *****) J .1 ? i I' t j
% <r, , , - - - -y--.m--,w ,_n ,,_ ,.w,
,,,..,,-w-,.-%m-.--- , m, w. %.. . _. - en t-te - ..N**=-t'w5-
. - , . _ _ _ . . . - - . . . . - - - ...- .-.. ~ . . _ . _ _ _ . . . - . . . . - . . . _ . . - - . . . . - - .
- 5. THEORY OF' NUCLEAR' POWER PLANT OPERATION, FLUIDS. AND PAGE ..._.
10 THERMODYNAMICS- ) 1 QUESTION S.09 (i.50) : t Indicate whether the following w 11 INCREASE, DECREASE, or have NO ; EFFECT.on the available NPSH of a centrifugal pumps
- a. Throttle shut on the pump discharge valve.
I
- b. Increase the t.emperature of the < suction side fluid. !
i
- c. Increase the pressure of the N2 blanket on the auction side I supply tcnk, i 1
I i l l 1 i l I 4 f 1 i i i I l I t 6 I i i (*****. CATEGORY 05 CONTINUED ON NEXT PAGE *****) { i
- 5. THEORY OF NUCLEAR POWE_R PLANT 2 OPERATION. FL IDS. AND PAGE 11 THERMODYNAMTCS [
t i GUESTION 5.10 (2.90) ! t cor steam (at 1000 psia) going through a throttling process, inoicate how the following paramoters change as the steam passes through the , i valve (INCREASE, DECREASE, or REMAIN THE SANE). (NOTE: assume ideal conditions)
- a. Erithalpy i
- b. Pressure
, i j c. Entropy [
- d. Temperature .
r s a n [ t l r p 1 1 I r
\
i I E l
- i 9 t a
1 F j ! l 1 !
; r i
a t i . [ l 1 l 1 t a L j (***** CATEGORY 05 CONTINUED ON NEXT PAGE ***La) j a i i !
- t
- .__. _ _ _ , ~ _ .
j S. THEORY OF NUCLEAR POWER PLANT OPERATIO!!. FLUIDS. ANQ PAGE 12- ; THERMODYNAMICS' 4 i i QUESTION 5.11 (1.50) l i Nith respect to reactrr start up, indicate whether each of the e l following statements are TRUE or CALSE: 4 i l A. The purpose of fully withdecwing the shut down rods is to provide j adequate negative reactivity which may be inserted should a problem i ocr.r i 4 J { B. Counts doubling indicates that the margin to criticality has been
- j halved.
- i l C. If counts have doubled, adding the same amount of reactivity will l cause the reactor to go supercritical. ;
l I l 4 ! i I f 1 i i 'l i : l ! i ! l l 2 l 1 s i , i i i d 1 (***** CATEGORY OS CONTINUED DN NEXT PAGE *****) i l i 1 i
- - - - - . _ . _ .-.. . . ~ . . . -
S. ' THEORY OF NUCLEAR POWER PLANT l_l OPERATION - LUIDS. AND PAGE 13 THERMODYNAMICS , t OUESTION S.12 (2.00) The reactor is coerating at 30% power when one RCP tripn. Assuming i no reactor trip or turbine load change occurs, indicate whether the following parameters will INCREAGE, DECREASE, cr. REMAIN THE SAME. j l
- a. Flow in operating reactor coolant loops (0.5) l
- b. Core delta T ( 0. 5 ) ~
- c. Reactor vessel delta P (Om S)
- d. Operating loop steam generator pressure (0.5) :
1 1
)
1 l 1 l l (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) l i i
! 5 '. - THEORY OF NUCLEAR POWER PLANT OPERATION. F1UIDS. AND PAGE 14 ! l THERMODYNAMICS i ! -l l' , i i i 5 OUESTION S.13 (1.00) l l^ ! Liut all of the conditions that must be present in order for natural 'f l circulation to exist. ! ! I i 4 I i j i 1
)
i j i l i l i t a , \ 5 t 4 i i t i j l l i i (***** CATEGORY OS CONTJtv)ED ON NEXT PAGE ****t) 4 l 3
-S. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS. AND .PAGE . 15 (
THERMODYNAMICG j I OUISTION 5.14 (1.00) State TWO (2)-reasons why 10 exp -3% powir (1x10exp-O amps) is chosen as a standard reference for critical rod height data. , t NOTE: "standard reference" is NOT an acr:optable answer j i I i
}
t I I l i l t i 4 l l l t i r i (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) i
- - - - . . v :,
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 16 THERMODYNAMICS ,
, 1.*T GUESTION 5.15 G . - /7// 0 :/2 y'FF j a. When in MODE 2 with keff < 1, the Shutdown Margin (SDM) _must be verified within 4 hours prior to criticality. How in this veri-l fication accomplished without performing a GDM calculation? (1.0) ; 4 l 1s os...
. i os c ru.4 3 4. . . .n v. -
-nn g; . ,; , ,3 , e e e_> (0.5) !
c)/M 1,'lk gjp
- c. Name SIX factors that go into a SDM calculation. .(1.S) J l
l I i i I 1 1 l a 4 i T n a 1 t t i l e (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) , 1 !' 1
S. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE' 17
' THERMODYNAMIC _S i 'OUESTION 5.16 (2.00) l -List the FOUR bases for the minimum temperature for criticality.
l-i i l I I i i I i i l l J l i l l (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****) t
. ._ . _ . . _ __.. . _ _ . _ . -_. ~.. _ . . . _ . . _ . . _ . . - _ _ . _ _
S. THEORY OF NUCLEAR POWER PLANT OPERATION FLUIDS. AND PAGE id i THERMODYNAMICS , l 1 QUESTION S.17 (1.50) ! 2
- a. With reactor power greater than 50%, what in the maximum Quadrant-4 Power Tilt Ratio (OPTR) that can exist without putting limitations on plant operations? (O.S) l i r
- b. The Technical Specifications allow operations for a 2 hour time {
j period with a GPTR of greater than-that in part a. What is the i' reason for this allowance? (1,0) [ i r i f l . i I i 3 l l , j i l l I (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) l r l I l
l .5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 19 , i THERMODYNAMICS ll j' i l l t 1 l i QUEGTION S.10 (1.S0) i f l The enthalpy of the reactor coolant entering the steam generator is j j 650 Btu /lbm; its e n tha l p'f leaving is 500 Dtu/lbm. The enthalpy of the l feedwater is 400 Btu /lom. What in the enthalpy of the steam produced if the reactor coolan t flow is 6 x 10exp 7 lbm/hr'and the feedwatel- !
, flow is 6 x 10exp 6 lbm/hr? (SHOW ALL WORK!) !
i l (- ! 1 I i i i 1 ) 1 i i } i 6 2 l l l I l 1 P l . 1 t . I ! (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) l i _ _,__ _. _ _ _ _ .-. _
S. THEORY OF NUCLEAR POWER-PLANT OPERATION. FLUIDS. AND PAGE 20 THERMODYNAMICS t GUESTION 5.1o (1.00) l Calculate the condensate depression in a condenser.opa.ating at 1 psia 4 with a condensate temperature'of 95 deg F. (SHOW ALL WORK!) l l l i I i T I r I i i l 1 (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) l l l I i
u.' . 1 S. THEORY OF NUCLEAR POWER ' PLANT OPERAT ION. _ FLt! IDS. AND PAGE 21'
' THERM 0DYNAMICli, i-
- GUESTION S.20 (1.00)
Define subtritical multiplication. 1 (***** END OF CATEGORY 05 *****)
i 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMEN_T AT ION _ PAGE 22 I i, j QUESTION 6.01 (1.00) Which statement below .egarding ' the Main Generator Protection Syntom is NOT CORRECT. a) Opening the generator output breakers ALWAYS results j in a turbine trip when the generator is loaded. j b) Once the generator is loaded, a tur*.ine trip ALWAYG . res'il ts in ,1 generator trip. l d c) A turbine trip above the protection interlock P-7 (10% power)*ALWAYS results in a Reactor trip, d) A reactor trip ALWAYG results in a turbine trip. x c s a ., c l IGG ro $-9 C 5~ 0 % i M u' Ca ) l 4 l f i 4 1
- 1 1 l 1
j i I
, i 1 6
)
i : i ! 1 l I f 1 i 1 ! i t 4 i .I i 3 4 i i 1 l ) i 4 (***** I CATcr40RY 06 CONTINUED ON NEXT PAGE *44**) i
, we . - . . - - . _ - - - . - - . . - .
- 6. PLANT SYSTEMS DFSIGN. CONTROL. AND INSTRUMENTATION .PAGE 23 i,
GUESTION 6.02 (1.00) l l An undervoltage on a 7.2 kV safeguards bus occurs 20 secondu after the receipt of a Gafety Injection signal. Which of the following state-g
.aen ts regarding sequencing of loads onto the safeguards bus is correct?
I
- a. All loads except Load Block #1 are stripped and the ESF
! Loading Sequence is reinitiated once the DG output breaker ! is closed. l l b. Sequencing stops until the DG output breaker in closed at ! which time it continues from the point at which the under-i voltage occurs.
- c. Gequencing utopa until the DG output breaker is closed at
- which time only the ECCS-related equipment sequence will j be reinitiated.
? 4
- d. All loads except the ECCS-related equipment are stripped and only the ECCS-related equipment sequence will be cont.inued once the EG output breaker in closed.
1 (***** CATEGORY O!> CONTINUED ON NEXT PAGE *****)
l . - . . -6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATIOff .PAGE 24 L 6 i I i GUESTION 6, 3 (1.00) What set of signals below is sent to the Reactor Protection System to j l indicate a Turbine Trip? ; t
- i
- a. Stop valves closed & Auto Stop 011 pressure low
- b. Stop salves closed & EHC pressure low I
- c. Governor valves closed & Auto Stop 011 pressure low
- d. Governor valves closed & EHC pressure low i
i l 1 1 l l 1 l l i l I l i l ((41F4 CATEGORY 06 CONTINUED CN NEXT PAGE *****) j l 1 l l i a l
L6. PLANT SYSTEMS DESIGN. CONTROL, AND INSTRUMENTATION PAGE 25 OUESTION 6.04 (1.00) Which valve listed below in used to throttle auxiliary spray flow? a) FCV-122 (Charging Flow Control Valve) b) PVT-0145-(Aux Spray Valve ) c) PCV-444C (Loop C Spray Valve) d) PCV-444D (Loop A Spray Valve) e) You cannot throttle auxiliary spray (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 26 I f
GUESTION 6.05 (1.50) i Indicate whether the Over Power Delta Temperature trip setpoint will ! INCREASE, DECREASE, or REMAIN THE SAME for the following parameter j changes. Consider each separately. ; i
- a. Incruauing Tavg, (0.S) l
- b. Tavy less than rated power Tavg (O.S) i 3
l
- c. Delta I becoming more negative (0.S) l t
l (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) i l
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 27- +
OUESTION 6.06 (2.50) Concerning the Rod Control System: ; r a) Using a one line diagram, show the inter-relationships of the l following components. i 1) DC hold cab 4 net l j 2) Fower cabinet ;
! 3) Motor generator set '
j 4) Reactor Trip Breaker (s) S) Automatic Rod Control Unit ,
- 6) Logic Cabinet ,
l b) For the components in Part a), above, STATE the number of rach present in the system. l 1 1 h > f I h r : i I I ! i ! I
- t
) I n + I . t i f a f 1' i 4 i i (***** CATEGORY 06 CONTINUED ON NEXT PAGC *****) i ! l i i ii i i ! j _, - - _ , _ . . , , _ - - . - _ _ . _ - . _ _ . _ , _ ~ . _ . . _ _ _ . , _ . . _ , _ _ _ . - , _ , _ _ _ _ . _ _ , _ . _ , _ , , _ . , . - , _ - , . _ , , - , . . . _ . -
i t i 6. PLANT SYSTEMS DESIGN. CONTROL, AND INSTRUMENTATION PAGC 28 ! i i i t QUESTION 6.07 (1.S0) , Match the RCS penetrations in Column A with the appropriate RCS loop negment listed in Column D. (Answers may be.uned ; j more than once) , Column A Column D i' a) Normal Letdown 1) Loop A cold log b) PZR Surge Line 2) Loop A hot lag i {- c) CVCS Normal Charging 3) Loop A intermediate leg l [ d) PZR Spray Line 4) Loop D cold leg j e) RHR Suction 5) Loop D hot leg I
- 6) Loop B intermediate leg i
)
i 7) Loop C cold leg
- 8) Loop C hot leg '
- 9) Loop C intermediato leg I
[
.e l i
1 1 (***** CATEGORY 06 CONTINUED ON NEXT PAGC *****) l I l
- 6. PLANT SYSTEMS DE. SIGN. CONTROL. AND INSTRUMENTATION PAGE.. 29 j i I
; I t
i GUESTION 6.00 (1.00) ;
- t i t j What is the main design purpose of the flow restricting noz:le in the [
j Main Stuam Lines? ' i i i t a 4 ! l i ; i f I k - i ! a i k a j i i I- . } I, i I l, t i f i i i [ i t 2 1 2 J t 1 i l t l 4 ($4*** CATECDPY 06 CONTINUED ON NEXT PAGE *****) l l ! i i t 4
. _ _ _...__ . . _ _ . . . . . - . _ . . _ _ . -. _ . . ~. _ _ _ _ . _ _ . _ _ . _ . _ . __ . . _ _ .
I i-a
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 30 - ;
r i
}
I o f QUESTION 6.09 (1.00)
- i 5 1
The S/G PORV's maximum capacity is limited by design to approximately 6% of rated steam flow. What is the. reason for this l imi ta tic,n? I
- l 1 :.
1 i i l - i i l' ) I ' \ i
}
.- l 4 l 1 l
; 1 i
. 1 i i 1 i s ! 6 i } ! i 4 i 1 I 4 i l 4 i l d i l , (***** CATEGORY 06 CONTINUED ON NEXT PAGE ****'.) 1 1 J f 1
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUt1ENTATIg{ PAGE 31 ;
t i k t QUESTION 6.10 ^Y^' l (i.50) o3/a-.*'/97 f j . or arx r h A e> A.o 9: '> a n sL [
~
What signal (s)3must be present to initiate,an AUTOMATIC switchover of i the uuction of the RHR system from the RWST to the reactor building , recirculation sumpo? Give setpoint(s) and coincidence (s) if } 4 applicable. _j i i i
- l. 6 i
i r t 1 1 i ! s, d i i i } ( I i i i e l ! (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) i, I 1 s j I (_..._
- 6. PLANT SYSTEMS DEGIGN. CONTROL. AND INSTRUMENTATIO!! PAGE 32 OUESTION 6.11 (2.00)
For each of the following, list the components of the Reactor Building Spray system which are affected.
- a. Phase A Containment Isolation Signal (1.0)
- b. Spray Actuation Signal (1.0)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
4z_1PJ. ANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 33 GUESTION 6.12 (1.50) The following questions are associated with the normal service gas decay tanks:
- a. How many normal service gas decay tanks are there? (O.S)
- b. Normally, how often in the in-service tank switched? (0.S)
- c. Why is the waste gas distributed among all normal service gas decay tanks inutead of filling one tank at a time? (O.S)
(***** CATEGORY 06 CONTINUED ON NEXT FAGE *****)
/
- 6. PLANT SYSTEMS DESIGN. CONTROL.'AND INSTRUMENTATION PAGE- 34 QUESTION 6.13 (1.00)
List FIVE outputs of the Pressurizer Pressure mantor controller. NOTE: Redundant outputs count as one, i.e., Pump A and Pump D. (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
- o. PLANT SYSTEMS DESIONa CONTROL, AND-INSTRUMENTATION -PAGE- 35 l i
i ', t i
-!t
- 1 1 5
) GUESTION 6.14 (1.00) , j l 4 t < What plant conditions will cause the Cold Overpressure Protection : f System (CDPS) alarm to actuate? l 1 i 3 l I i l !
+ ,
) ! l I 1 ! 1 i l i I ;
- l
- 1. 1
) l L 4 i i i 1 ! i t
- l !
i i f ; } . a 1 i i ! 6 1 i i 4 3 ] 1 r k I I i a i 1 J j (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) l l 1 l l ..-~-_________.___-_.._.____._____.____._4
l 6. PLANT SYSTEMS DESIGN._ CONTROL, AND INSTRUMENTATION PAGE 36 l 4 i i i J l GUESTION 6.15 (1.25) With the pressuri er level control switch in Position 2, describe how a high failure of LT-459 will affect actual pressurizer-level. Assume norinal charcine and letdown system lineups exists and no operator i (.c tion s are taken. Continue the description until pressurizar level l is constant or a reactor trip caccurs. Include setpoints% here
- applicable.
l i i { 4 h t h ev> J L F a. 6 to o sv u t u s> E s t 4.s ,n s i ! ,1A4 oS/Ae/<'7 I i l I i j I 1 ! 3 i I 3 i 1 l 1 t I l l l i l 1 i ! 1 , I f (***** CATEGORY 06 CONTINUED ON NEXT PACE *****) i f l 0 l 1
-,n -
____---___w
6, PLANT SYSTEMS DESIGN._ CONTROL. AND ~ INSTRUMENTATION .PAGE 37 , l l 4
.)
. l l 1 . 6 GUESTION 6.16 (1.00) 4
. List 2 reasons for providing a void volume in a new fuel rod. ,
d i 1 I i i I Tr I 4 i j 4 t i > f , l l 1 . i l l r
> t ! l 1 i i I 4
l L i ; i i ! I a t 'l : 6 t t I e 4 ; j t (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) i i I l 4 l 1 i } t
l
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND_ INSTRUMENTATION PAGE 30 -
OUESTION 6.17 (1.00) Asido from a loone printed circuit board card, list 4 distinct causes of an "URGENT FAILURE" in the Power Cabinets of the Rod Control Gyutem. (**4** CATEGORY 06 CONTINUED ON NEXT PACE *****)
6 '. PLANT /SYSTEMG DESIGN. CONTROL. AND INGTRUMENTATION PAGE 39 l-OUESTION 6.18 (2.00) LIST 4 of the 5 Design bases for the CCCS Cooling Performance i following a LOCA, as stated in iOCFRSO.46. i (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
. _ _ . , _ _ . . _.. .m_... . _ . _ . . _ _ . _ . _ . . _ _ . .
- 6. -PLANT SYSTEMS DESIGN. CONTROL, AND INSTRUMENTATION PAGE' 40 i i
l QUESTION 6.19 (1.50) , t ! List 3 purposes for the Pod Insertion Limits. i I
- i 1 .
[ i i ' i 4 t 4 h ! l 1 ! $ I i e 1 . i ! i i i 1 - 2 j t ! i t j l , t , i i ! i i (
- 4 i
f; e I 1 . n t k i J l i i i ! a r i . I f r b ,,J j (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) [ j
- +
j f i i i h 1 i
,e - .
. .- . . - . = . - . - . . . . . _ . - . . . - - - . . . . . . - . - . - . . . . - . . . .
- 6. PLANT' SYSTEMS DESIGNe CONTROL. AND INSTRUMENTATION PAGE 41 i 6
l GUESTION 6.20 (1.50) ; f- Describe what causes a "Hard Dubble" in the prennur 1::er during normal I plant operations and how this affects the reactor on plant trannients. l i I i, i i k 4 4 h i I L a ! .i 1 l l l l l t l l ! I i J l, 1 i l 1 i i l i l l 1 i l I i 1 (***** CATEGORY 06 CONTINUED ON NEXT PAGE * * * * * ) f i i i i i
- - - - - - - -,e-----s. =e m vrw y --v - m e ew -wm ww-,-- a. w+m,,,e- ., . .. w+-pgy w w ww -* m . . . n , eg w evd
!- 6. PLANT SYGTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE. 42 ; i . l 1 : GUEU)1ON 6.21 (3.00) j All reactor controls arc in automatic with Dank D rods at 210 steps
- and reactor power at 70%. The output of the Auctioneered Tavg circuit i fails LOW. Describe the effect this would have on ALL the plant j control systems supplied with this signal until a utable condition is f f reached or a reactor trip occurs. (asuume no operator action) l
! i
- I
) i 4 l 1 i ) ! i ; 1 .i i 1 I [ i i i I j i i ! j i I l l t ) i i i l i 4
- (***** CATEGORY 06 CONTINUED ON NEXT F' AGE *****)
i ( L . , . - - .- - ._ - . . -
- 6. PLANT GYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 43-OUESTION 6.22 (1.00) I Hot, are the inputs to the Detector Current Comparator compared, and I what conditions are needed to automatically defeat the circuitry while j at-power? ;
i I l i l E l l t i 5 i i t i I t I t
)
i
}
I f I f (***** END OF CATEGCRY 06 *****) i o l i I
- 7. PROCEDURES - NORMAL. ADNORMAL. EMERGENCY AND PAGE 44 RADIOLOGICAL CONTROL i
l i k OUESTION 7.01 (1,00) ' p ! Which ONE of the following is NOT an input into determining , ! conform.ince with the reactor core Safety Limits? i 4
- a. Pressurizer Pressure f
! b. Lowest Operating Loop Flow l i i
- c. Thermal Power 'j i
- d. Highout Operating Loop Tavg l 4 !
l I s l I ! 4 i I l I ! l 1 i l 4 ) ! l s I l l I, 1 l (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) l l
i 7. PROCEDURES --NORMAL. ADNORMAL. EMERGENCY AND PAGE 45 j j RADIOLOGICAL CONTPQL j d, ! i i b, I ! OUESTION 7.02 (1.00) 1 1
.I i If the Subtriticality Critical Gafety Function indicates a dached !
path, which ONE of the f ollowing ac tions -is required? ] l i ! j a. Go immediately to the referenced procedure. l I
} ; b. Go to the referenced procedure as time permits.
l l 4
- c. Monitor other status trees and if no higher priority condi- '
] , tion exists, then go to referenced procedure. l j i
< d. Monitor other status trees and if no higher priority'condi-tion existu, then go to referenced procedure as time permits.
l ~ i 1 2 } l i i I t I i (t**** CATEGORY 07 CONTINUED ON NEXT PAGE *****) i I
j 7. PROCEDURES --NORMAL. ADNORMAL. EMERGENCY AND PAGE 46 l l E_ADICLOGICAL CONTROL, 4 l a !, l li i GUESTION 7.03 (1.00) i l While performing steps in EOP 6.0, "Loss of All A/C Power", an GI
- nignal may be generated when power is restored. If an SI signal in j generated, which action below is required by the procedure?
I } a. Reuet the SI to permit the EDG's to energize the emergency bunen. l b. Place SI in test to prevent an overpreunurization of the RCG. 4 j c. No action is necessary an it has no'effect, i i d. Reset the Si to permit MANUAL loading of equipment on an A/C emergency bun, i i-s 4 1 ! l 4 1 I 1 i t i 4 : i I l I i i i ; 5 1
- i
{ ) ! i l 1 (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) 1 1 I
. . _ - . . . . .-- . _. - . . . . . . . . _ . .--. - . ~ . . - . . - - - . - . - . _ . - _ . .
- 7. PROCEDUREG - NORMAL. ADNORMAL. EMERGENCY._AND 'PAGE 47.
RADIOLOGICAL CONTROL i
- .I 1 ;
1 I I; 4 I GUESTION 7.04 (1.50) j In accordance with EOP 12, Monitoring of Critical Safety Functions, : i list the Critical Safety Functions (CSF) in deucending order of l j priorityi ; i t 8 ; 3 i } i I I
-l
.f i 4 I i 4 i ' i i ! e 1 i i f t I I l l ). i 5 (***4* CATEGORY 07 CONTINUED ON NEXT PAGE *****)
l i 4 1 1
- 7. PROCEDURES - NORMAL. ADNORMAtst_EMERGF;N_CY AND PAGE 48 RADIOLOGICAL CONTROk i
DUESTION 7.05 (1.00) 5 i j Which'ONE of the following statements is correct concerning the status ! of the Nuclear Instrumentation Recorder prior to withdrawing control f bank rods for a reactor startup?
- a. The highest reading source range channel and the highest reading intermediale rdoge channel are selected.
1
- b. Only the highest reading source range channel is selected.
t 4 c The highest reading source range channel and the lowest reading intermediate range channel are selected,
- d. The highest reading source range channel and either intermediate ,
range channel are selected. i l 1 j l l l l l l (884t* CATEGORY 07 CONTINUED ON NEXT PAGE to$$*) i
. - _ - . . . . - , . _ _ . -_ ~ _ . _ . - - .--. -_ - . . - _ - .
4; 7. PROCEDURES - NORMAL2 ADNORMAL. EMERGENCY AND. PAGE 49 ! RADIOLOGICAL CONTROL l l f i QUESTION 7.06 (1.00) I Which ONE of the following reasons correctly. describes the basiu for l allowing RCP restart in EOP 14.0 (ERG-FR-C.1) "Rouponse to ; Inadequate Core-Cooling l
- a. Helps to mix the SI flow to protect the reactor vessel from cold-water i 1
- b. Once subcooling is establiuhed, restarting the RCPs helps to j collapso voids that may have formed in the reactor vescel I head,
- c. Allows reutoration of PZR pressure control uning normal ;
uprays. ; i
- d. Provides for cooling of the core when secondary depressur- !
12ation does not alleviate inadequate core cooling. ; I l i i i 1 t 3 i 1 (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) I
- 7. PROCEDURED - NORMAL. ADNORMAL. EMERGENCY _AND PAGE SO I RADIOLOGICAL CONTROL l 1 i i
l OUESTION 7.07 (1.00) ! i With respect to Emorgency Plan Procedure EPP-003, In Plant i i Radiological Surveying, indicate whether each of the following l ! utatements are TRUE or FALGE. l ) I
; 1. A Standing RWP must be used for an entry into an area with a f nevere or unknown exposure potential. i
'I
- 2. En tri6m in te, areau _ where a nevere or unknown potential for radiation exposure exists should be limited to Rescue attemptu ;
i to prenerve human life.
! i i i i,
I 4 l l t l L I i 1 i
- i
) I i ' i l ! 1 3 1 } i l a I 1 l i I (***** CATEGDRY 07 CONTINUED ON NEXT PAGE *****) i l 1 l
\
l 7, PROCEDUREG ~ NORMAL. ABNORMAL. EMERGENCY AND PAGE S1 RADIOLOGICAL CONTRQL-i 4 s I l i GUESTION 7.00 (1.50) ! 4 l t . j The reactor han been operating for 1 week at 100% power. RCS prensure j and temperature are in their normal bands. Indicate whether EACH of l the following conditions would be a symptom of a lous o1 containment j intugrity au listed in the TS for primary containment integrity? i l Answer YED or NO. ! I 1 : j a. Primary containment air temperaturen for elevations 412', 436' l
- and 463' are 114, 116, and 110 degrees F respectively, j J b. One air lock door as inoperable for more than 24 hours with l 1
the operable door claued and the inoperable door locked. j 1
- c. A section of the containment ventilation ducting exiting the i containment is removed and a thin metal plate has been taped l over the opening thus keeping air from leaking out.
i j ! I i 1 i i l i l i 1 1 1 l J l (***** CATEGORY 07 CON 11NUED ON NEX T PAGE *****) i 1 i i l
} 7. PROCEDURES - NORMAL. ADNORMAL. EMERGENCY AND PAGE 52 i i RADIOJyGICAL CONTROL l 3 i 4 i ! DUESTION 7.09 (1.C3) i 3 i FILL IN THC LLANKS: l i ? A loss of flow signal is nunt to the solid state protection system (SSPD). A Reactor Trip will occur only if _ of (0,5) Power Range Channels exceeds percent (O.S) , i i 1 l L ! i < 1 1 ! 4 l i i i ! 1 i 1 .l a ,l I I i ! i ! i ! .! l 1 1 l l l t 4 i a 1 I l l (6**** CATEGORY 07 CONTINUED ON NEXT PACE *****) l
.__._m.-,-_.._.__.--_.,-_.__._.-.___.___ . _ _ . ,,_ m .,_ _ -_.. - _ _ . _ ._ __ _ , , - _ , . , . - , .
- 7. PROCEDURES - NORMAL. ADNORMAL. EMERGENCY AND PAGE 15 3 -
RADIOLOGICAL CONTRQL l OUESTION 7.10 (1.23)
- Indicate the numerical values associated with the following precautions
- a. Maximum di f f erential pressure between RCS and S/G.
i
- b. Maximum differential temperature between RCS loops. l
- c. Minimum RCG flowrote prior to and during RCS dilutions,
- d. Maximum rate of pcwor increase above 20% reactor power without managemen t approval .
- e. Maxtmuu RCS temperature during RHR operations.
l ? l t i
? < t i
l 1 1 1 I i l (***** CATEGORY 07 CONTINUED ON NEXT PAGE ****4) l E
I.-..-.. - - . . . . - . - . . - - . - .. . - - . - . .- { 7. PROCEDURES - NORMAL. ADNORMAluEMERGENCY AND PAGE S4 l' RADIOLOGICAL CONTROL i 1 i: l 7
^!
- QUESTION 7.11 (1.00) !
i s , i j What are the 2 Immediate Corrective Actions required Af the charging ! i flow control valve ( FCV- 122 ) fails closed while in automatic control ! I l and will not retpond to a manual open signal? i i 5 1 I O f L L .- t ? >, , ,y Go' Woon fil e m il G m ;
'l N44 o s/1 p/ 7E l i
] l 1 l, i l i 4 1 : I
\
i l
} <
4 I i i i a > l i I I ! l i i i I ! l ! l ! } ! i I i, I, i k t 4 i (***8* CATEGORY 07 CONTINUED ON NEXT PAGE o<***) a j l 1 i.-,-.--.__
.._...-..__m __ __
- 7. PROCEDt! REG - NORMAL. ABNORMAL. EMERGENCY AND , PAGE' SS l RADIOLOGICAL CONTRQL I
OUESTION 7.12 (1,50) [ What are the plant icad limitations with the following number of feedwater boostar pumps in service?
- a. 3 l I
- b. 2 i
- c. 1 l
i i i i I t l I t [ I f l i l l (***** CATEGORY 07 CONTINUED ON NEXT PAGE t****)
- 7. PROCFDUftEG t - NORMAL. ADNORMAL. EMERGENCY AND PAGE 56 RADIOLOGICAL CONTROL oA(4P40 # # #
.GIEr.T I O!! ". I '* ( i , 5 t' '
Other than identifying and correcting the malfunction, what are the 3 Immodlato Corrective Actions for a diessel generator trip following an emergency start? nas.trl tw e.e si r. fi /7/A nl>vlPI I i l l ($**** CATEGORY 07 CONTINUED ON NEXT PAGE *****) _ - . _ ~ - . . . - - . _ _ . - -
- 7. PROGCDURES - NORMAL, ADNORMAL. EMERGENCY AND PAGE 57 (i'ADIOLOGICAL CONTROL QUESTION 7.14 (1.00)
What method is used to collapse the void, according to EOP-iO,2, "Response to Voidu in the Reactor Vessul", if a void exiutt in the , reactor vcusel with all RCPs stopped? j 1 (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
7, PROCED_URED - NORMAL. ADNORMAL_3 EMERGENCY _fLNH PAGE 50 , RADIOLOQLGAL CONTROL i 1 $ I i l OUESTION 7.15 (2.00) ! t I List the FOUR possible alternate actions that can be taken if, during a j response _to abnormal power generati(.n (ATWS), the turbine had not auto- i matically tripped. , t I i l i f i l t I i I I i i l l I i 1
- l
\
l (***** CATECORY 07 CONTINUED ON NEXT PAGE *****)
l 7. PROCEDURES - NORMAle ADNORMAL. EME6DENCY AND PAGE 59 RADIOLOGICAL CONTROL t OUESTION 7.16 (1.00) ! What plient conditions dotormine when adverse containment conditions i are ut.ed in the En,crgency Operating Procedurns? l i 5 1 l l l ) l ! i l I I I l l l i l i I 1 l l ) 1 i l l 1 I I i 1 i l l (****4 CATEGORY 07 CONTINUED ON NEXT PAGE *****) l l l
3 1 j 7. PROCEDURES - NORMAL. ADNOfMAL, CMERGENCY ANQ PAGE 60 l RADIOLOGICAL _C_QNE Ok 1 GUESTION 7.17 (2.00) In accordance with EOP-1.3, "Natural Circulation Cooldown", what are the five (S) parameteru (conditions) that uupport or indicate natural
- circulstica flow during cooldown? Which way should thety be trending?
I i I l I 1 l' ( i ) I i 1 i l i it f i i I i l i I i I (***88 CATEGORY 07 CONTINUED ON NEX1 PAGE 1**t) 1 l
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND - PAGE '61 RADIOLOGICAL CONTROL QUESTION 7.18 (2.00)
List'the 4 conditions, including appropriate setpoints, that must be enet prior to opening a Reactor Coolant Pump (RCP) neal water' bypass valve. > i i s k i J l i (***** CATEGORY 07 CONTINUED ON'NEXT PAGE *****) i
7 '. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 62 RADIOLOGICAL CONTROL GUESTION 7.19 (2.00) List all the immediate actions for a loss of all AC, as stated in EOP 6.0, Loss Of All AC. NOTE: It is NOT necessary to list the Alternative Action. (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) ( _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ - _ - _ _ - - - - _ _ _ _ - _ - - - - _ _ _ _ - - - - - - - - - _
- , - - . ,. .__. _. _ . . _ _ . . _ . _ _ _ . . = ..-_ - ..- .. _ -_ .- _ . - .___ -. _ . - - . _
! 7. ' PROCEDURES - NORMAL, ABNORMAL. EMERGENCY AND- PAGE. 63' j RADIOLOGICAL CONTROL j t t j GUESTION 7.20 (2.00) t l I LIST the 4 SI Termination Criteria ao stated in EOP 1.2, . Safety Injection Termination. l i l' 4 7
+
5 .I 4 a t I O i 1 > 1 l h d I i i 5 i (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) i 1 i i i ! 1 1 1 --,.=n-.,--n-- _.., ,,,, ~ ,,n.-,- ,- _ ,,,- ,..,,- ,-.n-..~,.,-,.~-.n,--n,.-,
_._ _ - . . _ _ . . . . _ . . . . ..._..m._ .. . . . _ , _ _ _ . . . . . . _ _ , . - . . _ . . ._ k l ( I < [
- 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 64 l RADIOLOGICAL CONTROL ;
I QUESTION '7.21 (1.00) t EOP-13.0, "Response to Abnormal Nuclear Power Generation /ATWS", has the operator trip the turbine as one of the immediate actions. i However, one of the major concerns in the ATWS transient response evaluations is the excessive RCS pressure developed due to significant heatup of the primary coolant. Since keeping the turbine on the line would mitigate this temperature rise, why is the turbine ' tripped? i
.l t
1 j i l i i i 1 2 i j (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) I
____-~ _- - _ . . _ _ . _ . . _ . . - _ . _ . _ . . - . _ - . _ . . _ _ . _ _ . . _ _ . _ _ _ _ _ - . _ _ . _ _ _ _ _ _ _ - _ . _ . - _ . . _ _ l
'7. PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE 65 RADIOLOGICAL CONTROL i ,
j ' QUESTION 7.22 (1.00) ! ! In J.ccordance with EOP 1.3, Natural Circulation, what determines the 1 i amount of RCS subcooling required during a natural circulation cooldown following a reactor trip? i 4 1 i i j s l l l l < l a . 3 L i , /
- i i i i :
1 i i 4 l t i j (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
- 7. PROCEDURES - NORMAL, ABNORMAL. EMERGENCY AND PAGE 66 RADI_OLOGICAL CONTROL _
QUESTION 7.23 (1.00) How are the trip bistables of a failed power range detector placed in the trip conditionl
& 14csa ct C= cir ct ?st n i.n I: t16 to >~ccon4 r+ t pi> s e s <) o n. g
/2/4 0 2 /L y-/s ?
((**** CATEGORY 07 CONTINUED ON NEXT PAGE *****) l 1 1 l \
- 7. PROCEDURES ~ NORMAL. ABNORMAL, EMERGENCY AND PAGE 67 RADIOLOGICAL CONTROL i
k QUESTION 7.24 (1.50) Caution III.1 of EOP 15.0 "Response to loss of Secondary Heat Sink" states: If S/G Wide Range levels in any 2 S/G is < 20% OR Pressurizer , pressure is > 2335 poig then GTOP ALL RCPs and immediately initiate i ble%' 'nd feed per steps 7 through 14. ' Why are the RCFs trippud prior to initiating bleed and feed, aside from the fact that heat input.from the pumps will be removed? l f i t a w l l i i ! i 4 ; k l I i I s i t 1 e f I i i i r I !
- i. !
i (***** END OF CATEGORY 07 *****) , i !
- r J t
] I a , ? i t
1 i
- ' (
O. ADMINISTRATIVE PRGCEDURES. CONDITIONS. AND LIMITATIONS. PAGE 68 i GUESTION 8.01 (2.00) 5 State WHETHER or NOT ther conditions below would meet the safety injection system TS OPERABILITY criteria, assuming all other conditions are normal? Answer YES or NO. 3
- a. "A" RHR pump fails to start and "B" RHR pt.mp mechanical seal-fails, i
- b. One of #2's Accumulator discharge check valvas fails to open.
- c. "D" cnarging pump trips ~and "B" RHR pump seals fail.
l '
- d. Two charging pumps fail, one due to a breaker malfunction, the l I
other due to bearing problems. i i
- i j !
5 i 4 : l [ > l F t !
, s e
i ! l l
- i
+
E i .I l i i (4**** CATEGORY 08 CONTINUED ON NEXT PAGE *****) l $ I _ . - - _ - - - _ _ ~
.t
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMIIATIONS PAGE 69 OUESTION O.02 (1.00)
The following plant conditions have existed-for the past twelve (12) hours: 100 pe"cent rated thermal power (RTP) Normal Operating Temperature / Normal Opa ating Pressure Residual Heat Removal Heat Exchanger (RHR Hx) "A" - INOPERADLE The maintenance supervisor reports that the suction from the containment sump to RHR Pump "B" is-INOPERABLE. You concur. Which one of the following most accurately describes the allowances
- and/or limitations imposed by the Technical Specifications?
NOTE: Technical Specifications 3.0, 3.4.1.3, 3.4.1.4, 3.5.2 and 3.5.3 are enclosed for reference, a) No limitations are imposed by the Technical Specifications. b) Restore the inoperable subsystem to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours. c) Suspend all operations involving reductious in Reactor Coolant System (RCS) baron concentration and immediately initiate corrective action to return loop to operation. d) Within i hour, action shall be initiated to place the unit in at least HOT STANDBY within the next 6 hours & at least HOT SHUTDOWN in the following 6 hours. e) Decause of the inoperability of either the RHR Hx or RHR pump, restore at least one ECCS subsystem to OPERABLE status or maintain RCS Tavg to less than 350 degrees by use of alternate heat removal methods. (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
,..... . .~ . . . . - . _ - . ..-. . . . - - - - . . . _- - - . - . - - . _ _ . -- - _ . - . , 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 70 5 1 1 l -QUESTION- G.03 (2.00) l l t For EACH of the following, indicate WHETHER or NO.T the Technical Specifications require action to be taken within one hour,
- a. In Mode i with one full-length rod immovable but within 12 steps of the group counter demand position.
'b. In Mode 5 with one pressurizer code safety valve inoperable.
.; c. In Mode 2 with one pressurizer PORV block valve inoperable. 1 4 ! d. In Mode 1.with only one operable charging pump. i ? I I i < d
)
{ 4 ! 4 i i i
)
I l 2 l l i l l 1
- (***** CATEGORY 00 CONTINUED ON NEXT PAGE *****)
i i i 3 l 1
- , . . ._. _--,,-_-..n--.-._ ._
. . . . . . . . - - . . . . ~ . - . - . - . . . .~ -. . - . - - . . - - - - . . . . _ . .-. . - --. - ,
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE. 71 4
QUESTION 8.04 (2.00) With respect to the EDG, state whether a VALID TdST, INVALID TEST, or VALID TEST FAILllRE would apply for each of the following:
- i. Malfunction of equipment that is not part of the defined 3
generator unit ( e . c; . ESF Lcad Sequencer). , l
- 2. Successful starts that are intentionally. terminated without' loading.
- 3. Any diesel trip that is the result of a valid trip condition.
I 4. Test performed in the process of maintenance troubleshooting. d i F I I J ! ! l I I,
! i i !
i l I l l i (***** CATEGORY OO CONTINUED ON NEXT PAGE t****) l 0 , l _ __ _ _ . . ~ . . _ . _ _ _ _
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 72 l t l :
i
+
l- GUESTION 8.05 (1.50) l i l With regard to Maintenance Work Requests (MWR), indicate whetter EACH , of the following are TRUE or FALSE. 4
- a. Emergency maintenance does NOT require the initiatiors of a MWR.
- b. MWRs for corrective maintenance are normally preparted by-the l' person identifying the problem. ,
- c. Priority 1 MWRs can ONLY be approved by the Shift Uupervisor. !
i i, i i l i I i l' { i ~ l 1 1 l I ! I l i 4 i i I d J i 4 5 l i (***** CATEGORY 00 CONTINUED ON NEXT PACE *****) I i 1 i-
. . .. . . - . . . . __ . - - - . . = .
O. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 73 GUESTION O.06 (2.00) For each of the following, state whether or not the condition (s) .below , will place the unit directly into an LCO. (state YES or NO)
- a. Loss of ONE (1) seismic monitoring instrument.
- b. Loss of ONE (1) meteorological monitoring instrument
- c. Loss of a PRNI channel at > 1% power.
- d. Loss of BOT'i SRNI channels at > 1% power.
I I d b 8 6 t i 1 l i f (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) 4 l l t
O. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 74 I J
\
QUESTION 8.07 (1.25) Match the RCS leakage types in column.A to~the TS limits in column D. (Assume plant is in mode 1) 4 i COLUMN A COLUMN D I
- a. Through PZR PORV to PRT 1. O gpm l
- b. From RCS into steam generators 2. 1 gpm ;
From unknown location 10 gpm
- c. 3.
- d. To RCP seal supply 4. 31 gpm
- u. Nonisolable fault on RTD bypass line 5. 33 gpm !
- 6. 21-gpm i
I k' h r i i I I l (***t* CATEGORY 08 CONTINUED ON NEXT PAGE *****) I l l i i
j 8. ADMINISTRATIVE: PROCEDURES . CONDITIONS. AND LIMITATIONS PAGE 75
-l f
OUESTION 8.09 (1.00) What is the minimum boron concentration requirements that must exist in the RCS prior to unbolting the. reactor vessel head? l 5 t i i f l 4 r
)
t 6 1 d j 4 I i I i .i .s i a J i l 1 4 . (***** CATEGORY OG CONTINUED ON NEXT PAGE *****) i 4
- 1 i
i
- 8. _. .. _ . . _ _ __ .-,-. .-,.n. .---,,n-,-
. . . . . ~ -.. . -. . - - . - - - . . - . . . . _ - - - - . . - . - - - . . . - - . _ - . . . . _ - - - - - - . - . . - - . 1 1 8. ADMINISTRATIVE' PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 76-l OUESTION O.09 (1.00) 4 What is the following a Technical Specification definition of: t The process of determining an instrument's accuracy by visually j comparing the indication to other independent instrument channels l measuring the same parameter. - i B t I h 6 7 4 5 4 I b i J r [ 1 . ] 1 I i
^
,l 1 8 9 a 1 (***** CATEGORY OG CONTINUED ON NEXT PAGE *****) t s i i l._ _ _ _ _ _ _ - _ . . _. __ _ _ . - _ . _ - - - _ , . _ _ _.
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE '77 .
QUESTION 8.10 (1.00) , What is the basis for the high pressurizer water level reactor trip? i i f I b i l I I l l l l (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) l r .. . . . .
. . _ . . - - . .. .-. _ .....-~_. --. - ..-.-.-...-.-.--
l G. ' ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONG PAGE 78
'f l'
I 4 ; l
- GUESTION 8.11 (1.00) l
> l j- What in the basis for the upper containment temperature limit? l j 'l I l: e i J r . i I l i ' I j i ) 0 i I i i i r 4 f f 4 f i a I i l l 4 i i I a ' 1 l (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) } i . I I
- i
, i
B. ADMINISTRATIVE PROCEDURES.-CONDITIONS. AND LIMITATIONS PAGE 79 QUESTION 8.12 (1.00) The most limiting condition f or GDM requirements occurs at EOL with , Tavg at no load operating temperature and is based on a steam line break accident.
- a. Why in EOL more limiting than BOL? (0.S)
- b. Why is no load temperature more limiting than full load temperature? (0.5) i l
t I L i I (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
D. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND-LIMITATIONGJ PAGE 80 l l i I GUESTION O.13 (1.50) t l ! a. What is the Technical Specification Safety Limit for RCS prescure i ! while in Mode 1? (0.S) l < l
- b. What are the 2 required actions that must be taken within i hour j if this limit is violated? (1.0) 4 i i
) l 1 J , j i . I f 1 4 i .
- I 1 !
i i l i ! l ! i i e j ! l, : I T i (***** CATEGCRY 00 CONTINUED ON NEXT PAGE *****) , i f
O. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 81 i QUESTION B.14 (1.00) Once a piece of equipment is danger tagged -for a particular maintenance work request, no other work can be performed on the isolated equipment unless either of two requirements are met. What are these TWO require- ; i ments? i s 1 I h 6 l
)
i i l 4 i 1 I I f ! i t i '
! l ) ,
4 , l l ! l
- l J
! (***** CATECORY 00 CONTINUED ON NEXT PAGE *****) l t i I
,. . - . . . .-. . ..-. _. . - - . . _ - . . . . . . - . . _- . - .. . __ ... .~. . - . . . . - .
-0.- ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE O2 l
- QUESTION 8.15 (1.00)
! With the exception of electrical personnel acting in the capacity of a l red tag, what are the TWO required qualifications of the individual responsible for second verification of danger tag placement? 4 1 1 1 i i, i f I i 1 4 i l i } - i 4 1 1 1 1 ) I l 1 : I l; l, 3 l 1 i i r 'i 1 ,I (***** CATEGORY 00 CONTINUED ON NEXT PAGE *****) L _ __. _ -. . - - . . . . .. . , --)
O. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONG PAGE 83 OUESTION G.16 (2.00) According to EPP-OO1 for each of the following indicate the LOWEST Emergency Action Level (EAL) that requires the action to be taken:
- a. Notify Fairfield Pump Storage Facility
- b. Emergency Log Established
- c. Sound Radiation Emergency Alarm
- d. Activate EOF l
(****t CATEGORY 08 CONTINUED ON NEXT PAGE *****)
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 84 OUESTION 8.17 (1.00)
List the SCE&G employees allowable planned emergency exposure limit for each of the following: l
- a. Save human life
- b. Mitigate damage to vital equipment l
t I h r i l r i 1 i (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 85 e ,
t QUESTION 8.18 (1.00) ! What, AS A MINIMUM, should the temporary or unexpected relief turnover include? I b L P J I l i i s . l i a i l 1 i i i 4 4 ) (*A*** CATEGORY 08 CONTINUED ON NEXT PAGE *****) 1 J f )
i 1 i
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 86 4
GUESTION G.19 (2.00) ! a. What are the Bases of the Undervoltage (UV) and Underfrequency (UF) RCP Bus Trip as well as the bases for their setpoints? [
- b. What is the significance of the time delay incorporated in the trips linked in part "a"?
i i i i
! i
, I
. r E. t I i 3 !
, i I i r i I f l l l l (***** CATEGORY 00-CONTINUED ON NEXT PAGE *****) i l i . - . ._ _ ___ . - _ . - - l
i l 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 87 l i l i QUESTION 8.20 (1.00) l l What is meant by the statement in TS 3.4.4, action C, attached, that j says "the provisions of specification 3.0.4 are not applicable"7 6 l l ! i i l j : }, i 1 I i i i 1 i i ; i i !T i 1 1 1 4 I I I 1 i d (**f** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
! 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONG PAGE 80-I i
. OUESTION O.21 (1.00) i I What la the definition 'or the Heat Flux Hot Channel Factor 7 i
i 4 l I I i i !
! i
,I 4 l - t i l i ! b ' i s i 1 I i I t l I i t
~
i i i I, i i l l 1 1 (***** CATenORY 00 CONTINUED ON NEXT PAGE *****) i l 1 ! l; I
l 8. ADMINISTRATIVE PROCEDURES J ONDITIONS. AND LIMITATIONS PAGE 09 i i i l i j DUESTION 8.22 (1.S0) i i Using the attached Technical Specifications (TS), answer the question f stated in the situation presen ced below: I l The reactor in being refueled ( ) 23 ft above Gu Vescul-Flange), loop ! "C" is isolated for maintenance and RCPn "A" oud "B" are out of ) service for breaker repairs. "D" RHR pump and.its Heat Exchanger are j being used to circulate reactor coolant, w i t.h "I'" RHR pump INOPERABLE ! for routine maintenance. A request to take the "lF EDG out of service i for about 30 minuten to do a uurveillance on the gerserator is made by i the electrical supervisor. Can this surveillance be performed? Support your answer by referring to the J:pp l icabl e - TS . i 1 i i t Il 4 l l 1 i l 1
- i I
I j ! i l 4 , i !. I, l
- s i
I ! t I I 1 i l I i ! (***** END OF CATECORY 00 *****) ! (************* END OF EXAMINATION ***************) l
- 5. -THEORY ~OF NUCLCAR POWER PLANT OPERATION. FLUIDS. AND -PAGE 90 THERMODYNAMICS ANSWERS -- SUMMER -88/03/07-AIELLO, RF ANSWER 5.01 (1.00) c (1.0)
REFERENCE VCS, TH SCI, TS-10, P 17-20, LO 4,5,0,10,11,14. fGTI 2.5. 191004K109 ...(KA'S) ANSWER 5.02 (1,00) d (1.0) REFERENCE VCS, RT DK III, RT-12, P 17-23, LO 9. KA1 3.4. ! 192OO6K106 ...(KA'S) I ANSWER 5.03 (1.00) a REFERENCE DK III, REACTOR THEORY, CH-14, CURVES FND-RF-46/47, LO 7. KAI 3.1, 3.4. OO1000K502 192OO5K105 ...(KA*S) ANSWER 5.04 (1.00) a REFERENCE VCS, DK III, REACTOR THEORY, CH-14, CORE CONDITIONS AFFECTING ROD WORTH, P-16, LO 9. KAI 3.4. 192OO6K106 ...(KA*S)
. . . . __ _ __ _ ._ ..~ .- . . _ _ _
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS. AND '
- 5. _PAGE 91 4 THERMODYNAMICS i I '
ANSWERS -- SUMMER -DD/03/07-AIELLO, RF t ANSWER S.05 (1.00)
- c i i
i REFERENCE ; VCS, RT-10, P 11. RT-6, LO 13, 16, 21, 26, 29. KAI 3.4 192OOSK110 ...(KA'S) ANSWER 5.06 (1.00) ,
- a. MORE NEGATIVE (0.5) a l
- b. MORE NEGATIVE (0.5)
REFERENCE VCS, RT DK III, RT-11, MTC & TPD, P 4,5, LO 5.7,0. KAI 3.1. f i 192OO4K106 ...(KA'S) ! i
/.O ANSWER 5.07 -(_.:' ff p? o ;/z 9Y [
- a. DECREASES (0.5) -
u ~ ve r w r- .
- c. n s , , ,: o pp s;/19l71' i
- c. DOES NOT CHANGE (0.5) i .
{ REFERENCE ! , VCS, RT BK III, RT-11, P 24, LO 10,19. j VAI 2.9. ; ] 192OO4K107 ...(KA'S) { i
?
ANSWER 5.00 (1.00) !
- a. LOWER (0.5) ('
1
- b. HIGHER (0.5) ;
I i-2 I ? I i 2 I
- t j !
l l [ d
S. THEORY-OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 92 i THERMODYNAMICS ANSWERS -- SUMMER -OG/03/07-AIELLO, RF l ; 1 ) l i REFERENCE : l VCS, TH SCI, TS-10, P 25-30, LO 11. l } KAI 2.9, 2.6 ! 191004K107 ...(KA*S) a j ANSWER S.09 (1.50) i i
- j. a. INCREASE (0,5) !
i i
- b. DECREASE (0,5)
) ! c. INCREASE- (0.5) i i REFERENCE l i; VCS, TH SCI, TS-10, P 16-30, LO 14. l 4 KAI 2.8.
- 191004k11S ...(KA*S) ,
I f s i
- ANSWER 5.10 (2.00) 4 !
1 i i
- a. REMAIN THE SAME (0.5) l
- b. DECREASE (0.5) !
- c. INCREASE (0,5) i i
- d. DECREASE (0,5) ;
4 j' REFERENCE i VCS, TH SCI, TS-8, P 1-14, LO 2,3. KAI 2.8. { } t l 193OO4K115 ...(KA'S) l I l 2 6 1 . 1 i ANSWER 5.11 (1.50) l A. TRUE (0.5) l l j D. TRUE (0.5) ! { C. TRUE (O.S) ! ) f l REFERENCE ! i VCS, It.C DK I, IC-5, P 12, LO 1.4. VCS, RT DK II, RT-0, P 36, LO 7. { l i KAI 3.9, 3.0. l P l I L ___ .J
~ _ - - - - _ - - . - . . -_ - - . .. -. . . .. - .
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 93' j
THERMODYNAMICS ANSWERG -- SUMMER -80/03/07-AIELLO, RF l OO1010KSO6 192OO2K114 ...(KA'S) ! i E i f ANSWER 5.12 (2.00) ! a. INCREASE (0,5) z
- i. b. INCREASE (0.5) !
l c. DECREASE ( 0.' 5 )
- d. DECREASE (0.5) q REFERENCE , a
- i. VCS, TH SCI, TS-10, FIG FND-FF-45, LO 4,0. !
General Physics, HT & FF - Fluid Flow Applicetinna for Systems and . Components . j KAI 4.0, 3.4. OO2OOCK501 OO3OOOK301 ...(KA*S) i ! ANSWER 5.13 (1.00) ;
- 1. Density difference (or DELTA T) created by heat addition by the i heat source and heat removal by the heat sink. (0.5) {
i ll 2. The heat sini: must be elevated physically above the heat ; .; source.(0.5) i 1 i l REFERENCE i 1 VCS, TH SCI, TS-14, P 27, LO 5. KAI 4.2. 193OOGK121 ...(KA*S) I ANSWER 5.14 (1.00) i t i ! ) 1. It is well above the source range. OR Source neutrons are negligible. (0.5) l if
! 2. It is below the point of adding heat. (0,5) 1 OR !
l Doppler effects are not present (0.25) l j AND l 1 MTC effects are not present (0.25) < ] j REFERENCE j
- VCS, I&C DK II, IC-G, P 45. j
- RT-17, P 1, LO 2. i j
i r i
..,.,._ , ,,_. _ _. _ , __ _ ._ , , _ ._.._ _, _ , _ ,... _. , _ , _ _ , _ _ .k
S. THEORY OF NUCLEAR POWER Pt. ANT OPERATION. FLUIDS. AND PAGE 94 THERMODYNAMICS ANSWERS -- CUMMER -GO/03/07-AIELLO, R F-1 O O 112 ...(KA*S) l 1.7 ANSWER S.15 'T_"' A ff; a jgrx pg?
- a. The predicted critical rod position (0.5 ptu) in within control rod insertion limits. (0.5 pts)
'~
~
2 y f) d t. t 7 11/.3 /? f4 o y gr;2 s> (0.5)
- c. (any 6 at 0.25 pts each)
- 1. RCS baron concentration 2. Control rod position (stuck rods)
- 3. RCS Tave 4. Fuel burnup
- 5. Xenon 6. Sanarium
- 7. Power level REFERENCE VCS, TS, P. 3/4 1 1-3, GOP-3, p-7, RT DK III, CH 1D, ECC & SDM P 10, LO 5,6.
KAI 3.6, 3 . ~7 , 3.9, 4.4. 001OOCKSOD 192OO2K110 192OO2K113 192OO2K114 ...(KA'S) ANSWER S.16 (2.00)
- 1. MTC within analyzed range (0.S)
- 2. Protective instrumentation within normal operating range (0.5)
- 3. Pressurizer capable of being operable (0.25 pts.) with a steam bubble (0.25 pts.)
- 4. Reactor presuure vessel above RT(NDT) (0.5)
REFERENCE VCS, TS, p. D 3/4 1-2 (NO LO AVAILABLE) KAl 3.0. 001000G006 ...(KA~S) ANSWER S.17 (1.50)
- a. 1.02 (0.5)
- b. Allow correction (and detection) of a dropped (or misa11gned) control rad.
REFERENCE VCS, TS, P D3/4 2-13, RT DK III, RT-14, CONTL ROD REACTIVITY, P-2, LO 10.
S. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 95 j' THERMODYNAMICS
- )
i j ANSWERS -- SUMMER -GO/03/07-AIELLO, RF l 1 l:
.KAI 4.1, 3.0. l l 001000G005 001000G006 ...(KA*S) ;
t i ! i P ] ANSWER S.10 (1.50) ! ) (from formula H/0) l l l m1(hout,1 - hin,1) = -m2(hout,2 - hin,2) ml = reactor coolant mass flow rate 1 I m2 = feedwater mass flow rate l (hout,2 - hin,2) = -ml/m2 (hout,1 - hin,1) l.
- haut,2 = -ml/m2 (hout,1 -hin,1) + hin,2 (0,5) I
! i j hout,2 = -(6 x 10exp 7 lbm/hr) / (6 x 10exp 6 lbm/hr) x i 1 (500 Dtu/lbm - 650 Dtu/lbm) + 400 Dtu/lbm (O.S) l 1 i I hout,2 = [(-10) x-(-70 Btu /lbm)] + 400 Dtu/lbm ! ! I hout,2 = 700 Dtu/lbm + 400 Btu /lbm ] j hout,2 = 1,100 Btu /lbm (0.S) ! i i j Thus, the enthalpy of the steam produced in 1,100 Dtu/lbm. j { r ! REFERENCE ! VCS, TH SCI, TS-S, P 4, LO G. f KA1 2.4, 3.4. [ 193OO3K119 193OO3K125 ...(KA'S) ' f 4 ANSWER S.19 (1.00) f i t i From the steam tables, 1 psia corresponds to a saturation temperature j 5 af 101.74 deg F. (O.S) ! i i
- The difference between the saturation temperature and the condensate
- temperature is the condensate deprension, thuss i 101.74 deg F - 95 deg F = 6.74 deg F (0,5) -
REFERENCE VCS, TH SCI, TS-9, P 20, LO 6. STCAM TADLES MAI 2.5, 3.4. ! 193OO3K12S 193OO4K111 ...(KA'S) I i P I I
. ._. . . . . _ _ . . . _ _ _ _ . . __ ._. _ . . ~ _ . m.
S. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. ANQ PAGE ~96 f 3 THERMODYNAMICS ; I j ANSWERG -- SUMMER -08/03/07-AIELLO,'R F . I j ANSWER D.20 (1.00) ! i ! j The process or phenomenon of utilizing source neutrons to sustain ~the ! '! chain reaction for Keff < 1.0, (1.0) i I i ii REFERENCE , i VCS, RT BK II, RT-8, P 12, LO 4. i KAI 2.8. ! j 192OO3K101 ...(KA*5) J t
; l a ,
t s a i , p p t l t l i ' i 2 r [
] f I ) i 1 i 1
If i s I l I i l
)
i d 1 i
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRhMENTATION PAGE 97 I ANSWERS -- SUMMER -00/03/07-AIELLO, RF 1 ANSWER 6.01 (1.00) a o tt c (1.O) A. /*7 o 3/2 //7Y REFERENCE VCS, I&C, IC-9, P 44,by, l.O 1.1.2.
TBS, TD-5, P 07, LO 1.2.1, 1.*. MAI 3.6. 04SO10K423 ...(KA'S) ANSWER 6.02 (1.00) a REFERENCE I VCS, GS-2, SAFECUARDS POWER SYSTEMS P 36, LO 1.3, 1.4. KAI 4.0. 064000K410 ...(KA'S) l ANSWER 6.03 (1.00) b REFERENCE VCS, I&C DK II, IC-9, P SD, LO 1.3.2, 1.4.1. KAI 3.7. 045010K111 ...(KA'S) ANSWER 6.04 (1.00) a (1.0) REFERENCE VCS, ABS, AD-2, P 34, LO 1.4. AD-3, P 28, LO 1.1.3. KAI 3.9. 010000A202 ...(KA*S) _ _ - - 'f
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 99
~
WiGWERS -- SUMMER -88/03/07-AIELLO, RF l ll ANSWER 6.05 (1.S0)
- 9. DECREASE (0.5)
- b. REMAIN THE SAME (0.5)
- c. REMAIN THE SAME (O.S)
REFERENCE VCS, I&C BK I, IC-6, RCS TEMP IND, LD'1.4.4. VCS, I&C DK II, IC-9, REACTOR PROTECTION & LODIC, P-47, LO 1.3.2. KAI 3.3, 2.9. l 012OOOKSO2 012OOOK611 ...(KA'S) i L I ANOWER 6.06 (2.SO) a) 1. Auto Rod Control---> Logic Cabinet---> Power Cabinet (0.5) i
- 2. MG Set---> Reactor Trip Breakers---> Power ;
Cabinet DC Hold Cabinet[ 4-->(0. S) ' b) motor generatar set 2 /7(fp g [ reactor trip breaker (s) 2 g//p p, , power cabinet 4 l logic cabinet i ! automatic rod control unit 1 , DC hold cabinet 1 ! (0.25 ea) REFERENCE VCS, I&C BK I, IC-S, Flu IC"i.1, ICS.10, LO 1.1.2. KAI 3.7, 3.1. j OO1000K202 010000K203 ...(KA'S) i f i l i
- _ _ _ ___.._..__,--_ -. ~ , . _ . ,
- 6. PLANT SYSTEf18 DESI N. C O N T R O L__._ A N D INSTRUMENTATION PAGE 99 i ANSWERS -- SUMMER -OG/03/07-AIELLO, RF t i
. /. } T
* "^
/2 /4 ANSWER 6.07 -, o3/cy/9?
2 a) 4- 3 ll l y; o + /v " d) 1,7 4 e) 2, H ! (0.25 ea) + a REFERENCE l VCS, P&ID, RCS, DWG F-21. , ADS, AD-2, FIG AD2.3. ; { AD- 2, P 16-17, LO 1.2.3. 7 KAI 4.0, 4.1, 3.9. ! 002OOOK106 OO2OOOK109 OO5000K109 ...(KA'S) l J ANSWER 6.09 (1.00)
- d
! TO limit the rate of S/G blowdown during a main steam line break i ; (+1.0) (
i 4 REFERENCE l VCD, TDS, TD-1 P 16 LO 1.1.1. i KAI 2.9. , ] O~MOOOGOO7 ...(KA*S) 1 1 > i ANSWER 6.09 (1.00) i , 1 Limits plant cooldow1 rate (0.5) if any one PORV sticks open. (0.5) ' ) i 4 REFERENCE ' l VCS, TD-2, Main Steem System, P 16 and 05-0, Accident Analysis, P 20. 2 LO 1.1. ]
** i i KAI .3. '
03SC10K602 ...(NA'S) I i l 1 l l J } 4 l i i 4 )
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND I t_1STRUMENT AT I ON PAGE 100 ANSWERS -- SUMMER -OG/03/07-AIELLO, RF 4
i y , ANSWER 6.10 (1.50) i l v. 2 c.g) . v. - 1 ; p2 , in j. s mm '^~p'- ' 2/4 (464 pta.) RWST level ( M pts.) lens than 10% ( M pts.) i 0 'i 07 / yay O % 2 V 9W , 1 REFERENCE I 1 VCS, AB-7, RHR SYSTEM, P 16, LO 1.2. , i MAI 4.3, 3.9. OO5000K411 OO6020A304 ...(KA'S) l ANSWER 6.11 (2.00) ,
- a. 1. NaOH isolation valves (open) (0.5) i j c. Spray discharge inolation valves (open) (O.S) 4
- b. 1. RWST suction valves (open) (O.S) 1
- 2. Spray pumps (start) (0.5)
I REFERENCE VCS, AL-0, RD SPRAY 3YSTEM, P 15, LO 1.1.4, 1.2, 1.3, 1.4. I j KAI 4.3, 3.6. I j O26000K401 026000K402 ...(KA'S) l i ANSWER 6.12 (1.50) > 5 a. 6 (0.5) i , b. ovary 2 days (0,5) l
- c. To reduce the site boundary dose that would occur if a ningle tank ruptured. (0.5) +
REFERENCE VCS, AD-12, Waste Gas System, P 16, LO 1.2, 1.4. ,
! KAI 2.5, 3.3. ;
071000A203 071000K610 ...(MA'S) ! J J a f f 1 J l 4 1 L ! l i
. 6. PLANT SYSTEtLS DESIGN. CONTROL. AND INSIRUMENTATION PAGE 101-2 ANSWERS - SUMMER -OO/03/07-AIELLO',-R F l' E i ANGWER 6.13 (1.00) .i l ( Any FIVE at 0.2 points each) { ! i
- 1. PVC-444B (PORV) !
! 2. High Pressure Alarm I j 3. D/U Heater Control
! 4. Low Pressure Alarm
- 5. Proportional Heater Control i
- 6. PCV-444 C (D) (Spray) i REFERENCE 1 VCG, I&C DK-1, IC-3, PZR PRESSURE AND LEVEL CONTROL SYSTEM, FIGURE
; IC3.8, LO 1.1.2.
KAI 3.4, 4.1. l I 010000K402 010000K403 ...(KA'G) i i i ! l l ANSWER 6.14 (1.00) l os soy ?00 Low auctioneered wide range h or Tc less than itt9 degreen AND any j RHR Suction Valve not fully pen, gg p;p pp y i REFERENCE t i VCS, IC-3, PZR PRESCURE AND LEVEL CONTROL SYSTEM, P 33, LO 1.2.2. KAI 4.1. l l 010000K403 ...(KA*S) I i n
- ANStJER 6.13 (1.25) !
I (D/U heaters come on). Charging flow reduced to minimum. Lovel l l decreases (0.25). At 17% Icvel, letdown isolates (0.25). (Heaters l ! turn off). Level increases. "E u!
~^% (0.25). High I
- level trip at 92% (O.S). /Db o 3/a 7/917 '
NOTE: setpoints worth O.1 pt ea. i f REFERENCE VCS, CONTROL SYGTEM FAILURE ANALYSIS, IC-15, P 27, LO D.1.b, B.2.a. ! AOP 401.6, P 2. KAl 3.6. 011000A210 ...(KA'S) ! 1 1 t l i ( l L
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGC 102 ANSWERS -- SUMMER -OO/03/07-AIELLO, RF ANSWER 6.16 (1.00)
(0.5 ea for any two) j 1) Accommodate release of f.p._gasen l 2) Differential thermal expansion between clad and fuel pellet j 3) Fuel density changes during burnup REFERENCE VCS, RT DK II, RT-9, P 14,15, LO 12. i ANSWER 6.17 (1.00) i
- 1) Phase Failure (+.25 ea) l
- 2) Regulation Failure
- 3) Logic Error j 4) fiultiplexing Error 1
l REFERENCE j VCS, I&C DK I, IC-5, P 32, LO 1.4. j KAI 3.6. l, 001000G000 ...(KA'S)
- ANSWER 6.18 (2.00) i
, (any 4 of 5 at 0.5 ca)
- 1) Max. Fuel Element Cladding Temp. < 2200 Deg. F l 2) Cladding Oxidation < 17% thickness
- 3) Hydrogen generated by Zirc-Water reaction <1% of max.
i possible, f 4) Core remains in a coolable geometry
- 5) Provides for long term decay heat removal REFERENCE
, 10CFR50.46 UAI 3.0 006000G004 ...(KA'S) l i
l l l
.i
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 103 ;
j ANSWERS - SUMMER -GO/03/07-AIELLO, RF l l
?
ANSWER 6.19 (1.50)
- 1) Adequate SDM upon trip
- 2) To minimize the amount of positive reactivity Annerted during a rod ejection accident,# and ,
- 3) To minimize radial flux tilt (peaking) i A IW # Y *A (0.3 ea) cn A $ c e r. o p rit 3 A r c o c o* ~ r Mo v 5r S HM t > dad }
REFERENCE AfA 6 5 fn.yj? ? VCS, RT BK III, RT-14, P 20, LO 10. , KAI 4.7. ; OO1000K504 ...(KA'S) .I ANSWER 6.20 (1.50) i i i Gancs, particularly Hydrogen from the VCT (and some f.p. gases) come l out of solution as they are sprayed into the PZR. (+1.0) this ! creates larger pressure oscillations during transients (+.S) REFERENCE ! VCS, ILC DK I, IC-3, P 11--13, LO 1.3. ! KAI 3.6. l 010000G010 ...(KA'S) ; L i ANCWER 6.21 (3.00) { l
- 1) Rod control- Due to Tavg < Tref, rods will withdraw until a rod l stop is reached (+1.0) l
- 2) Pzr level control- Low Tavg will cause Pze level control system to l shutdown on FCV-122 until level is 25% (+1.0) {
i
- 3) Gteam Dumpa- Tavg i Tref, so that even if armed in the Tavg mode, !
no du:np actuation would occur (+1.0) REFERENCE VCS, ILC DK I, IC-5, P 17-19, LD 1.4. IC-3, P 19 LO 1.2.2. I C- 3, P 30, LO 1.2.2. KAI 3.6, 3.5, 3.1. 016000K!O1 016000K302 016000K303 ...(KA'S)
l
- 6. PLANT'StSTEMS DESIGN. CONTROL. AND INSTRUMENTATION 'PAGE 104 l
i ANSWERS -- SUMMER -80/03/07-AIELLO, RF i i l 1 ANSWER 6.22 (1.00) The highest reading upper / lower detector is compared to the average of ! the upper / lower detectors (0.5). The circuit auto defeats below 50% l power on ALL channels (0.5). ! i REFERENCE l VCS, I&C DK II, IC-0, P 39, LO 4.b, S.a. KAI 3.2, 3.9. 01SOOOA303 01SOOOK604 ...(KA'S> I t l
?
i I I l
! 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 105 RADIOLOGICAL CONTROL , ANSWERS -- SUMMER -DS/03/07-AIELLO, RF _ l I I l l ANSWER 7.01 (1.00) i b i REFERENCE VCS, TS, 2.1, DK II. I&C-9, REACTOR PROT. A LOGIC, FIG IC9.1, LO 1.2. KAI 4.1. i 000000G005 ...(KA'S) 4 l l ANSWER 7.02 (1.00) C I REFERENCE l VCS, EDP-12.0, P 3. (NO LO AVAIL) KAI 4.2. 00002?G012 ...(KA'S) I ANSWER 7.03 (1.00) d REFERENCE VCS, EOP 6.0, P 13, (NO LO AVAIL) VAI 4.1. l OOOOSSA206 ...(KA'S) I l I ANSWER 7.04 (1.50) Subcriticality, Core Cooling, Heat Gink, RCS Integrity, Containment, RCS Inventory. (0.1S for CSF, 0.10 for correct order) REFERENCE VCS, EOP 12, P 1. (NO LD AVAIL) KAI 4.2, 4.4. 0000290012 000074G012 ...(KA*G)
, _ . - - . .- . _ , _ . - . . - ~ _ -
- 7. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY ANQ PAGE 106 MJLIOLOGICAL CONTROL ANSWERG -- SUMMER -GO/03/07-AIELLO, RF ANSWER 7.05 (1.00) d REFF'RENCE 1 VC5 GOP-1, INSTRUCTION 7. (NO LO AVAIL) l KAI 3.6. l 015000A401 ...(KA'S) !
l l I ANSWER 7.06 (1.00) d ; REFERENCE ERG-HP, Westinghouse Dackground In f ormati on , RCP Restart", P 4. 1 (NO LO AVAIL) ! KAI 4.4 l OOOO74K307 ...(KA'S) , I ANSWER 7.07 (1.00)
- 1. FALSE !
- 2. FALSE i
REFERENCE t VCS, EPP-OO3, In Plant Radiological Surveying, P 1-6. (NO LO AVAIL) l KAI 4.4 i l'74001A116 ...(KA'S) ! ANSWER 7.00 (1.50)
- a. NO l
- b. YES l
- c. YES l (0.5 ea) l 1
REFERENCE ! VCS, TS 3.6.1.3, 3.6.1.5, 3.6.1.7. (NO LO AVAIL) ! KAI 3.0, 3.9. j 0000690003 000069G004 ...(UA*S) ' i l i - - I
1 7. PROCEDURES - NORMAL. ADNORMAL, EMERGENCY AND PAGE 107 l EADIOLOGICAL CONTROL l ANSWERS - - SUMMER -GO/03/07-AIELLO, RF f i l 1 1 ANSWER 7.09 (1.00) i i c of 4, 30 i REFERENCE ! j VCS, I&C, IC-9, P 46, LO 1.4.1. 1.4.2. I j KAI 4.2. l 01SOOOK101 ...(KA'S) l
- i. 1 ANSWER 7.10 (1.25)
(0.25 Points each) j
.l
- a. 1600 psid
- b. 25 degreen F j c. 3000 gpm
- d. 3%/hr
- e. 350 dugreen F REFERENCE VCS, GOP-APPENDIX A, GENERIC OPERATING FRECAUTIONS, P c. 3, 5, & 6.
ADS. AB-2, LO 1.3.1. i KAI 3.9 l OO2000G010 ...(KA'S) l l- ANSWER 7.11 (1.00)
- 1. Reduce letdown to 45 gpm (0,5)
- 2. Contral PZR level using (local) FCV-122 bypass valve (XVT-0403) (
.o_4 (0.5) l REFERENCE /- T '>W d <Y l e r /> </ w ^<
! VCS, SO P-' 102 , CVCS. ' Bt W r d 4e(>T5 ta y c c . . ,- ,m ; el v ic g ADS, AD-3, LO 1.3. VCS, P&lD, CVCS. /' ' # 2 /' eMI KAI 3.0. j 004000G014 ...(LA*S) ; i l l 1 I i l l l i
- 7. PROCEDURES - NORMAL._ ADNORMAL, EMERGENCY AND PAGE 100 RADIOLOGICAL. CONTROL
! ANSWERS -- SUMMER -GO/03/O'l-AIELLO, RF i i' ANSWER 7.12 (1.30) i a. No restriction (0,5)
- b. 60 +/- t% (O.S)
- c. 30 +/- 1% (0 S) i REFERENCE I VCS, SOP-210, FEEDWATER SYSTEM, P 35,36.
! GOP-4, INST 12. 1 TURD PLD SYS, TD-7, LO 5,6. KAI 2.9 0S9000G010 ...(KA'S) i ANSWER 7.13 '1 "^' na a:S ujze i
- 1. Reset mechanical overspeed, if plicable (0.S) i
- 2. Momentarily place EXCITER e ch in SHUTDOWN (0.5)
- 3. Depress GEN RELAYG, ffED L pushbutton (0,5) i REFERENCE j) A
- t. V*
l VCS, SOP-3n6. ,-RGENCY DIESEL GENERATOR, P 4, S, 24, 25. (NO LD AVAIL) ARP - ni -OOi , XCP-636. j r, 2, f' 22. I V^ 4.3. 000055G010 ...(KA*S) ANSWER 7.14 (1.00) Increase system pressure using precnuriner heatern while maintaining pressurizer level. REFERENCE
- VCS. EOP-10.2, P 5. (NO LO AVAIL)
LAI 4.0. OOOOO9A201 ...(KA*8) 1 d j
- 7. PROCEDURES - NORMAL. ApfN,QRMAL. EMERGENCY AND PAGE 109 RADI_OLJ2GTCAL CONTROL, 1 ANSWERS -- SUMMER -OD/03/07-AIELLO, RF 1
,i i 1 1 ANSWER 7.15 (2.00) l (y , f 7 --gg,;j f
/M (1 fr' [/.[f j 1. Manually trip turbino from MCD (0.5)
]
- 2. Stop (and lock out) both EHC pumpo (0,5) i 3. Runback the turbine (0.S) {
i 4. Cloue all MSIVn (O.S) l l C. T w' TU'h se l (f2s m f liC f 't o !! ? 1 742. .lE/4 .9 y/jojf?
~
REFERENCC ' j VCC, EOP-13.0, P'2. (NO LO AVAIL) l" KA1 4.5. l 0000270010 ...(KA'G) , i < 4 ANSWER { 7.16 (1.00) i :
)
"HIGH-1" signal present !
I l i t CR i r e i
! Cont Temp CAF i j Presu CAF ; Rad CAF t ! t REFERENCr i VCS, EOPs. DOOK 1 &' '
3 AUX BLDG SYS, AD-0, LO 1.2. j i AI 4.1. (P 3.6-27). f
}
i 102000G015 ...(KA'S) ! l t b I' i
)
i k 1 i t I a l f j , l l I I l d : 1 1 k !
- i 1 t t
4 4 I
- n, - - ..-,-
f
t f i l l 7, PROCEDUPES - NORMAL, ABNORMAL, EMEROFNCY ANQ PAGE 110 i RADIOLOGICAL CONTRO1 : l ANSWERS -- SUMMER -80/03/07-AIELLO, RF l 1 l l ' ANSWER 7.17 (2.00) C ,> % ) #3 ."' ' c o t ; N 6 IM G 40 C rJ C o h' / f .s ir r c .> 2 n of 1.-}i'#-'
; '. r (0.2) 95-6(*% cf fu!I paa s t, "D ) ,
p/77 (0.2) l
- 2. S/G Pressure (0.2) Afg.; a g/) Stable or decreasing (follow RCS ;
TEMP) (0.2)
- 3. Hot Leg RTD (0.2) Stable or decreasing (0.2) t 4 Core exit T/C (- 2) Stable or decreasing (0.2)
- 5. Cold Leg Temp (0.2) Near Saturation Temp, for S/G press OR
- Constant or slowly decreasing (0.2) fi I
REFERENCE j VCS, TH SCI, TS-14, P 14-24, LO 3,5. j GENERAL PHYSICS, HTFF, P 356. t VCS, ECP 1.3. KAI 4.2. 193OOGK122 ...(KA'S) f i ANSWER 7.10 (2.00)
- 1. RCS pressure (0.13) is ? 100poig (0,5) but < 1000 psig (0.5), f
- 2. RCP seal leakoff valves are open (0.25). '
- 3. RCP No. 1 seal injection flow to each RCP (0.15) is > 6 gpm (0.15).
- 4. RCP No. 1 seal leakoff flow rate (0.15) is < 1 gpm (0.15).
REFERENCE I VCS, SOP-101, P 6. l ADS, AD-4, LO 1.1.2, 1.4. KAI 3.6, 3.6. 0000000010 OO3OOOK103 ...(KA*S)
7 IJ
- 7. PROCEDURES - NORMAL. ADNORMAL._F,MERGENCY AND PAGE 111 PADIOLOGICAL CONTRO1 ANSWERS -- SUMMER -OO/03/07-AIELLO, RF ANSWER 7.19 (2.00)
- 1. Trip the Reactor manually from the Main Control Daard.
- 2. Verify Turbino Trip.
- 3. Isolate RCS.
- 4. Verify total EFW flaw > 390 gpm.
(0.50 ca) REFERENCE VCS, EDP 6.0, LOSS OF ALL AC, P 2,3. (NO LD AVAILADLE) KAI 3.9. 000056G010 ...(KA'S) ANSWER 7.20 (2.00) a) RCS Presuure - Stable or Increasing (0.S) b) RCS Gubcooling - greater than 30 degrees F (0.5) c) PZR Lovel greater than 4 percent (0.5)
- 4) Either 1 OR 2 below in natisfied (for Heat Sink Critoria):
i) Total Feed Flow to intact S/G's - / 390 gpm (0.25). OR
- 2) Narrow Range Level in at least 1 S/G - > 30 percent (0.25).
REFERENCE VCS, EDP 1.2, CI TERMINATION, P 1. (NO LO AVAILADLE) KAI 4.3. 0060SOA401 ...(KA*S)
1. l 7. PROCEDURES - NORMAL. ADNORMAL._EMERCENCY AND PAGE 112 FADIOLOG_LCAL_ CONTROL ! ANSWERS -- SUMMER -80/03/07-AIELLO, RF l 1 s l ANSWER 7.21 (1.00) t The.turbino is tripped so that the heat sink will be maintained as long au pouutble (0.5) on a total loun of feedwater ATWS (0.5). l REFERENCE 1 l ERG- HP , Weatinghouse Dackground Information, FR-S.1, P 75-77. l l (NO LO AVAIL) [ MAI 4.7 l OOOO29K312 ...(KA'S) l l j ANSWER 7.22 (1.00) l 1 i j Number of CRDM fans running. (1.0) : l I j REFERENCE f j VCS, EOP -1. 3, p. 7. (NO LO AVAIL) l ) MAI 4.5. l j OOOOO9E326 ...(KA'S) 4 l I l , ANSWER 7.23 (1.00) , i Removing the applicable control and instrument power fusea on the ! power tange drawers, pg f R RENCE
# ~#' *'i ' '
#* #5 0# 'W VCS, ADP-401.10, POWER RANCE FAILURE, P 2. /'oJ v ' 4 / ?Artf s h ) */ 4 9 c - 4' I&C, IC-0, LO O.
M,+#
# EM 9/M
}/ A I 3.2, 3.9.
015000A403 oiS000K604 ...(KA'S) 4 J l 1
- 7. PROCEDURES - NORMAL. ADNORMAL. EMERGENCY AND PAGE 113 RADIO W ICAL CONTROL ANSWERS -- CUMMER -GB/03/07-AIELLO, RF ANSWER 7.24 (1.50)
The RCPs will keep 2 phase flow mixture (0.7S) and the PORVs will not be able to releaue as much steam (energy). (0.75) OR Higher pronsure will reduce SI flow (0.7S) and increase the inventory flow out of the FORVs. (0.7S) REFERENCE VCS, EOP-1S.0, P 1. (NO LO AVAIL) Westinghouse B/O document, ERO-HP, FR-G/C/H, FR- H , P SS. KA1 4.2 , 000074KC00 ...(KA'G) i i l l l 1 4 I i l l l l
. _ - - _ . . . ~ _ . - _ . -. -. __ .-_m.- . _ . - . _ - . . , . _ . . _ . _ - _ . _
a h i i ! ! O. . AQ11fJISTRATIVF PROQLDURES. CONDITIONS. AND L_ IMITATIONS PAGE 114 ! ANSWERS -- SUMMER -08/03/07-AIELLO, RF ; i ! i l '{ i i-d ANSWER 0.01 (2.00) 7 i i a. NO ! j b. NO ! +
- c. NO f d. NO ,
l (0.5 ea) ! ! REFERENCE l } VCS, ADS, AD-7, P 11, LO 1.3. l ! AB-10, P 10, LO 1.3. ! ! VCS, TS, LCO 3.5.1 & 3.5.2 e.nd bar>es for each. [ ] KAI 4.2. [ . 013000G011 ...(KA'S) ; I ! ! f 4 ANSWER 8.02 (1.00) t d) (1.0) [ li ! 4, i REFERENCE L VCS TS, Section 3/4, LCO and SURVEILLANCE REQ. (NO LO AVAIL) { i Specifications 3.0, 3.4.1.3, 3.4.1.4, 3.5.0 and 3.5.3. ! I KAI 4.2, 4.2. ; OO6000 GOO 5 OU6000G011 ...(VA'S) ! ! l 2 6 ! i ANSWER 8.03 (2.00)
- a. YES (0,5) !
j b. NO 9.3)
- c. YES (0.9) i d. NO (0.5) !
l ! 2 REFERENCE I l VCS, TS, pp. 3/4 1-10 & 14 and 3/4 4-7 & 10. (NO LO EXIST) l j KAI 3.6, 3.8, 3.8. ; j 00000$G003 OO4000 GOOS 010000C005 ...(KA'S) f l 1 J J l I ! i I l 3 , I ! l t l l .
O. ADMitMTRATIVE PROGEDURES. CONDITIONS. AND LIMITATIONQ PAGE 113 l ANSWERS - SUMMER -80/03/07-AIELLO, RF l f I s ANSWER 3.04 (2.00) VT-Valid Test, IT-Invalid Tout, VTF-Valid Test Failure
- 1. IT I
- 2. IT ;
- 3. VTF {
4 IT l f (0.5 ea) REFERENCE ! VCG, CAP, SAP-204, P 9. (NO LO AVAIL) l. KAI 3.9. ! l 0640000011 ...(KA'S) ! l ! { ANSWER O.05 (1.30) f i I a. TRUE l
- b. TRUE :
- c. TRUE i (0.5 ea) l l
l REFERENCE VCG, CAP, GAP-601, REV 3, P 7, LO: RECPONSIDILITY 5.7. i KAI 3.4. 194001A103 ...(KA'G) ANGWER 8.06 (2.00)
- a. YEG
- b. YEG
- c. YEG
- d. NO REFERENCE VCG T.G., LCO 3.3.1, 3.3.3.3, 3.3.3.4. (NO LO AVAIL)
LAI 4.3. 015000K301 ...(l'A'S) { l
-I . . - _ _ , _ _ _ - - _ . _ . - . . _ . - ._._.-._.m
i
- 8. ADMINISTRATIVE PROplDM[1ED. CONDIT10r19. AND LIMITATIONG PAGE 116 -[
l ANSWERS -- SUMMER -00/03/07-AIELLO, RF [ l i l ANSWER O.07 (1.20) -[
- a. 3
- b. O
\
- c. 2
- d. 5 e, 1 (0.25 ea)
REFERENCE VCS, TS 3.4.6.2. (NO LD AVAIL) KAI 4.1. 002000G005 ...(KA*S) ANSWER G.00 (1.00) l l 2000 ppm OR enough to ensure keff is less than 0.95, whichever is ! GREATER. REFERENCC VCS, TS, P 3/4 0-1. (NO LO AVAIL) , LAI 4.1. { 00."000G005 ...(KA*S) ANCWER 0.09 (1.00) Channel Check (1.0) REFERENCE VCG, TG, P 1-1. (NO LO AVAIL) V. A I 3.b. Ol6000 GOO 5 ...(KA'D) l ANSWER O.10 (1.00) Protects the pressurizer !>a f e ty va l ves egainst water relief. REFERENCE VCS, TS, p. D 2-6, EU II, I&C-9, REACTOR PROTECTION & LOGIC, P-49, LO 1.4.2. VAI 4.3. 012OOOK402 ...tMA'S) l l 1 l 1
G. ADMINISTRATIVE PROCEDURES. CONDITIONS.-AND LIMITATIONS PAGE 117 ANSWERS -- SUMMER .-GB/03/07-AIELLO, RF l I ANSWER 8.11 (1.00) l l Prevent exceeding design pressure during steam line. break. REFERENCE VCS, TS, P -B 3/4 6-2. (NO LD AVAIL) KA1 3.6. ; O22000G006 ...(KA'S) i ANSWER 8.12 (1.00)
- a. MTC more negative at EOL (O.S)
- b. More mass in S/G at no load temperature (0.5)
REFERENCE VCS, OS-8, ACCIDENT ANALYSIS, P 31. (NO LD AVAIL) VCS, TS, P D 3/4 1-1. KAI 3.0. OO1000 GOO 6 ... A'S) ANSWER G.13 (1.50)
- a. 2735 puig (0.S)
- b. Be in Hot Standby (0.7 pts.) and notify NRC (0.3 pts.) (1.0)
REFERENCE VCS, TS, P 2-1. (NO LO AVAIL) MAI 4.0 002000G011 ...(KA'S) ANSWER 8.14 (1.00)
- 1. Another series of tags are issued (0.5)
- 2. Applicable information is added to (original and yellow copy)
Danger Tag Log sheet already in effect. (0,5) REFERENCE VCS, SAP, SAP-201, DANCER TAGGING, P 15, (NO LD AVAILABLE) KAI 4.1. 194001K102 ...(KA'U)
j O. ADMINISTRATIVE' PROCEDURES. CONDITIONS.-AND LIMITATIONS. PAGE 110~ ANSWERS -- SUMMER -80/03/07-AIELLO,.R F-l ANSWER 8.15 (1.00)
- 1. Qualified Danger Tagger (0.5)
- 2. (Current) NRC License (0.5)
REFERENCE VCS, SAP, SAP-201, DANGER TAGGING, P 5, (ND'LO AVAILABLE) MAI 4.1. 194001K102 ...(KA*S) , i
- ANSWER 8.16 (2.00) i !
- a. ALERT (0.5) f
- b. NUE (0.5)
- c. ALERT (0.5) ,
< d. SITE EMERGENCY (0.5)
REFERENCE VCS, EPP-OO1, ATTACHMENT III. (NO LO AVAIL) KAI 4.4 194001A116 ...(KA'S) i 5 ; 4 ANSWER G.17 (1.00) I i !
- 1. Life - 75 REM I
- 2. Equip - 25 REM ;
~
I j (0.5 ea) i
- REFERENCE
] VCS, EPP-02C, P 2. (NO LO AVAIL)
- KAI 3.4 i l 194001K103 ...(KA*S) j I
! i i ; ! l i 9 5 l l s i
- I 4
- i i
4 I f
.. . . _ . . . .. . . . . . . . . . . ~ . . . . . . . . . . . . . . . - . . . . . . _
O. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 119 1 N ANSWERS -- SUMMER -GO/03/07-AIELLO, RF l b ! ANSWER 8.10 (1.00) 4 1 , As a minimum, the temporary or unexpected relief' turnover should l include the following: , 1. A discussion of existir.g plant conditions and anticipated evolutions during the reliuf. (0,5) 4
- 2. A review of the main control board controls, instrumentation and annunciators. (0.5) i l
+ REFERENCE i VCS, SAP, SAP-2OO, P 7, (NO LD AVAILABLE) KAI 3.4. 194001A103 ...(KA'S) ANSWER 8.19 (2.00)
- a. The UV & UF RCP Bus Trips provide reactor core protection against DNB as a result of complete loss of forced coolant flow. (0.5) The specific set points assure a reactor trip signal is generated
! before the low flow trip set point is reached. (0.5) d t
. 'me delays are incorporated in the UF & UV trips to prevent [
curious reactor trips from momentary electrical power transients. !
! (1.0) l i 8 REFERENCE .
i VCS, TS BASES, P B2-7. I j I&C, IC-9, LO 1.1.2, 1.2, 1.4.1, 1.4.2. ! ] KAI 4.3, 3.0. !
- OO3OOOK501 012OOOK402 ...(KA'S) 1
( l l ANSWER G.20 (1.00) i l Entry into an Operational Mode may be made (0.5) even tf the i conditions for an LCO are not met (0,5). l REFERENCE VCS, TS 3.0.4.. (NO LD AVAIL) j KAI 3.6 010000 GOO 6 ...(KA'S) { i 4 I
- 0. ADMINISTRATIV,E PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 120.
ANSWERS -- SUMMER -88/03/07-AIELLO, RF l k-ANSWER O.21 -(1.00) The maximum local heat flux on the surface of a fuel rod at core elevation Z divided'by the average fuel rod heat flux. REFERENCE l VCS, TS, P B~3/4 2-1. l TH SCI, TS-13, LO 2. KAI 3.3. 1 193OO9K107 ...'(KA'S) I , ANSWER 8.22 (1.50) 2 Maintenance can not be performed-(0.5), TS 3.9.7.1 applies (1.0) REFERENCE VCS, TS 3.9.7. (NO LO AVAIL) MAI 3.8. ! j OO5000 GOO 5 ...(KA*S) i , 2 1 ! 3 I J l a Y l ; l I _ _ _ . . _ . . _ . ~ . _ ._ _ . _ - . _}}