ML20209E172
| ML20209E172 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 03/30/1987 |
| From: | Bill Dean, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20209E169 | List: |
| References | |
| 50-280-OL-87-01, 50-280-OL-87-1, NUDOCS 8704290448 | |
| Download: ML20209E172 (127) | |
Text
{{#Wiki_filter:' h, ENCLOSURE 1, EXAMINATION REPORT 280/0L-87-01 Facility Licensee: Virginia Electric and Power Company P.O. Box 26666 Richmond, VA 23261 Facility Name: Surry Nuclear Plant Facility Docket No.: 50-280, 50-281 Written and operating (oral and simulator) examinations were admini-stered at Surry Nuclear Plant near Surry, Virginia. Chief Examiner: / 3ttf / l ($L .h/ 7 bY William M. Dean - Date Signed Approved by: # ~7 2 !/7 Hohn F. MunrailSection Chief Uate Signed Summary: Examinations were administered on February 9-11, 1987. Written and Operator to 1 Reactor operatingRO) ((oral andand simulator) 1 Senior re-examinations Reactor Operator (SR0).were given 1 R0 and 1 SRO were given a written re-examination. 1 SR0 was given a written and operating examination. All R0s and SR0s passed all exams administered. Based on the results described above, 2 of 2 R0s and 3 of 3 SR0s passed the examination. 2 of 18 (11.1%) of the changes made to the written examination answer key were due to inadequate or insufficient training material provided by your staff to the NRC for examination development. 070429044G G70403 0 PDR ADOCK 0000 V
REPORT DETAILS
- 1. Facility Employees Contacted:
*H. McCallum, Operations Training Director
*B. Marshall, Training Staff
- Attended Exit Meeting i
I
- 2. Examiners:
*W. M. Dean D. J. Nelson
- 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 comments made by the facility reviewers are included as Enclosure 3 to this report. The NRC resolutions to these comments are listed below,
- a. R0 Examination (Applicable SR0 questions in parenthesis)
- 1. 1.04(5.08): Based on information contained in Technical Specifications, answer key will be modified to require Tavg vice Tcold. Lesson Plan should be changed to reflect the correct parameters which are monitored.
- 2. 1.08(b)(5.10(b)): Agree with facility comment. Answer key changed as recommended.
,l
- 3. 1.13(5.16): Agree with facility comment. Power rising I will still be included in the answer key, however, the recommended answer will also be i dCCepted for full Credit.
4, 1.14: Recommended answer is equivalent to existing answer key. No change required.
- 5. 1.18: Agree with facility comment. Due to similar-ity between choices "a" and "c", they will both be accepted as correct answers.
i l i I ! 6. 1.20(5.18): The answer key already addresses the fact that there will be additional answers that ! will answer the question that are not listed j in the key. The two recommended by the facility will be included. l
- 7. 2.06
- Agree with facility comment. Answer key will be modified to reflect the correct power sup-plies.
].
- 8. 2.14: Recommendea answer is equivalent to existing l answer key, No change required.
j 9. 2.19(a): Based on the information presented in Tech-nical Specifications as well as in Surry LP-
- 92.3, " Control Room Ventilation", the answer ,
i recommended by the facility will be added to ~ 1 the key as an additional correct answer. i ! j 10. 2.21: Agree with facility comment. The question , 1 is not worded to elicit the first part of the answer. Only the second part of the answer will be required for full credit. ) 11. 2.22(6.17): The facility's first recommended additional answer is a legitimate response during shut-3 down conditions and will be accepted as an i' additional correct answer. The second answer recommended by the facility is an extension j of the third answer in the answer key, and no , j change to the key is required. ' j 12. 2.23(a)(6.09(a)): Agree with facility comment. The structure ' ! of the question does not lend itself to a ! technically correct response. Clarification during the examination administration is in ! agreement with the facility recommended
- answer of a system drawing or sketch as the
; required answer. A suitable drawing will be j attached to the key.
I i i
- 13. 3.06(6.05): Do not agree with facility comment. The RWST i Low Level alarm actuates at 22.5% whereas the
. automatic swapover requires 2/4 level trans-i mitters below 18.5% level. No change to j answer key. l l 14, 3.09(6.19): Do not agree with facility comment. No i information was provided to support changing j the answer key, No change to grading. i l i i L __
- 15. 3.10(6.13): Agree with facility enmment. The recommended additional answers are all a function of the P-11 permissive and will be added to the key.
- 16. 3.11(6.14): Recommended additional answer is equivalent to part 3 of existing answer key. No change required.
17, 3.13: Agree with facility comment. Inaccurate extraneous information will be deleted from the answer key.
- 18. 3.14: Recommended additional answer is equivalent to existing answer key. No change required.
- 19. 4.03: Agree with facility coment. Due to differ-ences between Surry and the plant for which this question was initially generated, key will be modified as recommended.
- 20. 4.08(d)(7.11(d)): Agree with facility comment. Error in answer key will be corrected as recommended.
- 21. 4.10(2)(7.12(2)): Agree with facility comment. Question, as taken out of context, is essentially true.
- 22. 4.12(a)(7.13(a)): Do not agree with facility coment. Part(e) of the same question specifies the NRC limit with a Form 4, whereas part (a) does not.
A candidate will not be penalized if he answers both "2" and "6".
- 23. 4.12(b)(7.13(b)): Agree with facility coment. Question is not specific enough to elicit just one of the possible choices. Both answers will be required for full credit.
- 24. 4.14(7.16): Agree with facility comment. To be accurate, the procedural step refers to verifying the ,
pumps as running, vice verifying flow. I 25, 4.23(7.24): Agree with facility comment. Though the recommended answer is not contained in the referenced Abnormal Procedure (AP-39), it is a technically correct response and will be accepted as an additional correct answer,
- b. SRO Examination
- 1. 5.12: Answer inadvertantly omitted from the answer key. Correct answer should be "a" as recom-mended by the facility.
I
- 2. 5.70(a): agrea with facility coa =2nt. The additional recommended answer will be included as required information and the point value for that answer appropriately adjusted. The !
facility lesson plan should be corrected to reflect this additional basis. The noted typographical error will be corrected.
- 3. 6.10(a): Agree with facility comment. Answer key will be modified as recomended.
- 4. 8.07(b): Agree with facility coment. Due to the !
use of non-standard Technical Specifications, ' the modes of operation given as examples in [ this question do not apply to Surry as ; intended. Question will be deleted. ;
- 5. 8.09: Agree with facility coment. Additional '
recomended answers will be added to the key.
- 6. 8.17: intry conditions to EPIP-2.01 will not be !
accepted as a correct answer as the question clearly states that initial notification has , already occurred. However, the recommended t answer regarding event termination will be added to the key as an additional correct l answer. The recomendation that "the status of any notification item has changed" is just , a compendium of the original answer key and l will be graded accordingly.
- c. Other Examination Changes i
During the examination administration, Question 2.09(c) was deleted due to non-applicability to Surry Nuclear Plant. Post exam review , resulted in an increase in the point value of Question 3.16(6.16) I from 1.5 to 2.0 points to agree with the grading delineated in the l answer key.
- 4. Exit Meeting:
At the conclusion of the site visit, the examiners met with members of your staff to discuss items pertinant to the operating exams and l the examination process. There were no generic weaknesses noted during the oral and simula-tor examinations. The cooperation given to the examiners and the effort to ensure an atmosphere in the control room conducive to oral examinations was noted and appreciated. The licensee did not identify as proprietary any of the material provided to or reviewed by the examiners.
U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _SUBBY 1&2 REACTOR TYPE: _EEB-WEC3 DATE ADMINISTERED: 87/02/09 EXAMINER: _DEAHt_W M l CANDIDATE: INSIBUCTIQMS_IQ_CANDIDAIE1 Use separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY VALBE_ _IDIAL SCOBE _YALUE__ CAIEGQRY
&?. s l caGr90- _2h QQ _ _ S. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND
- THERMODYNAMICS
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. . . , 4 34T00~ _25.00 _ _ 6. PLANT SYSTEMS DESIGN, CONTROL, i AND INSTRUMENTATION
_2QuGQ__ _25 QQ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL
,. p CONTROL
-anted 25,00 8. ADMINISTRATIVE PROCEDURES.
CONDITIONS, AND LIMITATIONS 1 13 s-
-1BRrQ2"__ _= _% Totals Final Grado All work done on this examination is my own. I have neither given nor received aid.
Candidate's Signature
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
- 1. Cheating on the examination means an automatic denial of your applicatioQ and could result in more severe penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil QalZ to facilitate legible reproductions.
- 4. Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of cash section of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a n2W page, write OnlE Qn Qua Aida 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 thrgn lines between each answer.
- 11. Separato answer shoots from pad and place finished answer shoots face down on your desk or table.
- 12. Uso abbreviations only if they are commonly used in facility 1119raturn.
l
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Thereforo, ANSWER ALL PARTS OF Tile QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the cxaminur only.
- 17. You must sign the statement on the cover shoot that indicatos that the work is your own and you have not received or boon given assistanco in completing the examination.
been completed. This must bo done after the examination han
- 18. When you complete your examination, you shall:
- n. Ammambla ynne eynmination = f ell;;;;; j i
(1) Exam questions on top. (2) Exam aids - figures, tables, etc. r (3) Answer pages including figures which are part of the answer. i b. Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions,
- d. Leave the examination area, as defined by the examiner. If after i leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
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1 1 ! Q.___IUEg8y_(E TJ;LEQB.EQWE8_ELOUI_QEEBOI1991_ELUID$1_8ND PAGE 2s
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~ i i i s OUESTION 5.01 (1.50) j With respect tn reactor thermp3 1imits, indicate whether j each of the following statements. are TRUE or FALSE.
- a. The average linear power denoity in the core is expressed in units of tw/ft and in the total thermal power divided by ,
the active length of all the fuel rods. 1-
- h. The purpose of limiting the enthalpy rise hot channel 1 factor in to prevent bulk boiling from taking place during a [
l.DCA. r j
- c. The purpone of the limit on the heat f l u:t hot channel [
i tactor in to innure that fuel clad temperature does not j onceed 2200 deg F during normal operations. l i i l
- OtlE b f ! Of f S.02 (1.50) l Indicate whether each of the following will make the ,
moderator temperature coeff1cient l eLa neg.ati ve , more ! nr?o a t i v e , or havo no offcct.
- a. increase temperature ,
{ b, decr cano bor on concentr ation l e l c. incroaco coro age 1 J { UlJtGT 2 (1tl M.03 (1.60) i I fluc l ear reactor s are initially loaded with more fuel than in
- requirort to bring the roactor critical. The additional
; finnile matorial in the core tu said to reprecent built in !
j or au c ou ti rtattivitv. La ut ~3 t hinqu that encoun reactivity ! { in doSigned to overcne. i I I l l l , (***** UnitGURY 0*i CONilNtJID ON Ntxt pal;C *even) l i 1 > f L_______- - - _ _ _ _ - _ _ _ _ - _ - _ _ - _ _ _ _ - _ - _ . - _ _ _ _ _ _ - - _ _
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j y 5. THEORY OF NUCLEAR POWER PLANT OPERATION 3 _FLUIDL_AND PAGE 3
. ,IUERMODRMMICO i
i , OUESTION 5.04 (2.00) ) A. How doen Deff vary over the lifa of the core? I ) D. How lu Deff affacted au plutonium isotopos are produced { over the lifn of the core? l ! C. How is reactor response affected by a lower delayed neutron fraction? i UUtisTION 5.05 (2.00) l Given two pumps at equivalent design, operatinq at the name, i j canutant spoedt i j A. What will be the effect of placing the two pumps in seriou (with runpect to flow and head)?
- II. What will be the effuct cf placing the two pumps in oaral1ol (wi th r onpoct to flow and head)?
i 1 I DUliGTION S.06 (1.00)
- Utven
- Three reactor coolant (RCP) pumpr, operatiriq in
{ por allel . nach wtth a flow rato "m" anti a combined flow rato i "M". Out of tho four ponuabilitiew below, choot.e the one j t. h o t b rm t f1t u if un r- hcl i t, sncurud. ' I ) a. The e m,ul t a ne core t!nw (M) will incroaso. l i b. l'h o r t"uti t i nq core flow (M) wt11 i nr.r nann al ong wt th A ndi v i du.a t oporattog RCP flow (m). ]
- c. lhe renulting enre flow (ii) will decreano au individual
- operatinq hl
- P fIow (m) i nc r tw>ou . l tl. the renut t t tus core fIow (N) wl11 not change due to I l docruabo i n f(CI' b anl+ pro nure.
i i i i i UllEU T !!)N ii . 0 / (1.00) i j What. 15 t h t, dnal yn tuni n of h avi tig a UNDR ' or v4 to 1. I? I I 4 ( { (***** Ch il:buH Y O's CUN I Illlil'.D ON NL X l' l ' AM u****) I 4
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- 5. THEORY OF NUCLEAR POWER PLANI _9EER8Ilgth_ELylpg3_8NQ PAGE 4*
.. , THERMODYNAMICS l
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, OUESTION 5.08 (2.00)
I Lint the four (4) plant parametern observed to insure that ] CHF or DNDR are not onceeded. i ,' I DUESTION 5.09 (2.00) i j What are all the conditionn that must be prosent in order j for natural circulation to exist? : l 1 , 4 } UUESTION 5.10 (1.50) l < Write on your answer sheet INCREASES , DECREASES or DOES NOT l CHANGE for the followings j i The magnitudo of the fuel temperature coefficient (FTC): A. INCREASES / DECREASES / DOES NOT CHANGE with incronuo in , power. D. INCREASC3 / DECREASES / DOES NOT CHANGE with core age. l } r C. INCREAGES / DECREASEG / DOES NOT CHANGE with decroauo in I rnoderator temperatura coofficient (MTC). l , I i 1 l UUEST10N 5.11 (1.00) ' l The negative reactivity added when fuel temperature j 1 ric r e.w o r. in primarily causod by __,,__,.
. depimo on e, U-220
- h. dopplur broadoninq ;
i C. deplUI.lOn Of U-23D f I
- d. fHul pollut eiwulI thus docru.auing tho yap i l
1 1 l 1 4 1 i I i i I 4 j (***** CAILuuhY UD CONIINUCD ON NGXI PAGE **ma*) I 1
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I l h __IEE98Y_0E_.NUCLEGB_EgWEB_E66NI_gEgBOIlgN3 _ELglDL _6dD PAGE 5 l . , IUEBugDYN8dlCS f i ? 1 i f a ; I OUESTION 5.12 (1.00) l l l Which one of the following statements below is NOT correct
- regarding xenon behavior following a power increase? ,
j note: EXe] denotes xenon 135 concentration i l a. The minimum EXe] reached is independent of the magnitude j of the power level increase and initial power level. 1 1 i b. The time to reach equilibrium is also dependent on the j magnitude of the power change and final power level. j c. The time to reach the minimum CXu] is always < 11 hours. I j d. The time to reach equilibrium is approximately 40-50 i hours. 1 i , d ! j ! ! 4 . L'UESTION 5.13 (1.00) ' i ! GIVEN: Two identical control rodu, each absorb an equal ) amount of neutronn. The neutron flux at the center of the j core equals that at the edge of the core. Why do the control rods in the middle of the core (radially) have a greater f effect on Keff then the control roda at the edge of the core (radially). ] OllEGilON 5.14 (1.00) i ! What effect d o e c. rod chadowing have on the worth of control i ruds' I i l a i } ! i 1 4 1 i i (***** CATECORY 05 CONTINUED UN NEXT PAGE *****) i
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5 __IBEgBy_QE, NUCLE 88_EgWEB_ELONI_QEEB8IlgN2 _E691pS3 _8NQ PAGE 6'
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! ~ QUESTION 5.15 (1.00) Concerning subtritical multiplication, which one of the following statements is NOT correct?
- a. The neutron behavior per generation can be stated ,
mathematically.
- b. The neutron population will reach and maintain an equilibrium value.
- c. The f uel in the core effectively multiplies the source neutrons.
- d. As the source strength is increased, the magnitude of Keff is increased.
DUESTION 5.16 (1.50) On a reactor startup, what 3 conditions indicate the reactor is critical? h t [ QUESTION 5.17 (1.00) l l Give two reasons why 10 e>:p -8 amps is chosen as a standard ! reference for critical rod height data. l note: " standard reference" in NOT an acceptable answer ! 6 i UUESTION 5.10 (1.50) f List three things, in practice, that prevent water hammers From occurring ( W- * * *
- CATEGORY 05 CONTINUED ON NEXT PAGE *****)
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D.__IHEQBy_gE_ NUCLE 88_EgWEB_E68NI_QEEB8IIQN2 _ELulpS 2 _8Np PAGE 7
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OUESTION 5.19 (1.00) A centrifugal pump is started up with its discharge valve open. How would the following parameters differ (INCREASE, DECREASE, or REMAIN THE SAME) if the pump was started with j its discharge valve shut?
- 1 j a. Motor current i
- b. Discharge pressure OUESTION 5.20 (2.50)
What are the purposes of each of the following reactor thermal limits? If a specific accident or condition applies, state this in your answer.
- a. Reactor safety limits (1.0)
- b. Enthalpy rise hot channel factor (Fn (del ta H) ) (0.5)
- c. Nuclear flux hot channel factor (Fq(z)) (1.0)
QUESTION 5.21 (1.50) Attached is a typical boiling curve for water as it approaches, then exceeds, the DNB point. What are the thermodynamic conditions that cause: a) The decrease in heat transfer rate in Region III? b) The increase in heat transfer rate in Region IV? l l l (***** END OF CATEGORY 05 *****)
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i i 6.__PLGNI_SySIEdg_DEg1GN z_CONIBOL1 _@ND_INSIBUdENISIION PAGE 2 l , I I ! l 4 OUESTION 6.01 (1.00) 3 Which statement below regarding the Source Range Nuclear i Instrumentation System is INCORRECT. l j a) P-6 allows the source range high 1 vel reactor trip j i signal to be bypassed manually wher. one of the two i
! intermediate range instruments is above 10 E-10 ion I j chamber amps.
l l b) Placing BRTH source range blocking switches to the } BLOCK posi; ion de-energines the high voltage supply to both source range instruments. ] 1 c) The source range high level trip is blocked whcn P-10 f is present. I d) When P-6 is present and P-10 is not present, the 4 source range high level trip is automatically ] reinstated and the source range high voltage l re-energized when one of the two intermediate ranges j is below P-6 reset. t i OUESTION 6.02 (1.00) i Which of the following is NOT a design basis of the Steam j Dump System? { I l a) Accommcdate ramp load increases greater than { 10%/ minute. l ) f b) Pass 40% steam flow on a 50% turbine step rejection j without a reactor trip occurring. i c) Allow a turbine trip and a subsequent reactor trip from 100% power without lifting the S/G code safety j valves. i j d) Allow for a smooth shift of plant steam load from the I i steam dump system to the turbine on a plant startup. i f j t 4 i (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
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l 6.__PLGUI_SYSIEMS_ DESIGN1 _CgNIBQL1 _OND_INSIBUMENIGIION PAGE 3; DUESTION 6.03 (1.00) Which statement below reaarding the Main Generator Protection System is INCORRECT. a) Opening the generator output breakers always results in a turbloe trip when the generator is loaded. b) Once the generator is loaded, a turbine trip always results in a generatcr 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. QUESTIGN 6.04 (1.00) f Which valve listed below is used to throttle auxiliary spray f1ow? a, FCV-122 (Charging Flow Control Valve) b) HCV-311 (Aux Spray Valve ) c> PCV-455Et (Loop C Spray Va]ve) d) PCV-4554 (Laop A Spray Valve) e) You cannot throttle auxiliary spray OUESTION 6.05 (1.00) Which one of the f ol1owing conditions is required for au tcjna ti c swapover of the LHSI pumps to the Recirculation Modt,fallowiog a SI') (
*a) 2/4 RWST 1evel at Lo level setpoint e
*b)
, A LHSI pump recirc. isolation MOV closed for each
; pump t r. ) SI signal present
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SI Recirculation Modo signal present 1 s (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) 1 I i I
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- 6. PLANT SYSTEMS DESIGN t _GONIRQLt_GUQ_INSIRUdENIGIlgd PAGE 4 i
I ! 2 DUESTION 6.06 (1.00) i 1 } Concerning the Overtemperature Delta Temperature Setpoint j (OTSP) describe how (i nc r eases , decreases or remains the
- same) each of the following parameter changes will effect j the OTSP. '
i. j a) Increase in Tave i, i b) Decrease in Reactor Pressure l, ' t 1, - i o j QUESTION 6.07 ( .50) i TRUE/ FALSE i ! An urgent failure alarm could indicate that a slave q cycler failure has occurred in the logic cabinet. f i
! i l
6 OUESTION 6.08 (1.50) ' i i' Indicate whether each of the statements below regarding i permissive functions associated with the Excore Nuclear ! Instrumentation is TRUE or FALSE. I a) In order for the P-7 permissive (At Power Trips) to be DISABLED, both reactor power permissive P-10 and turbine
- power permissive P-13 must cl ear. i s
4 i b) The single loop loss-of-flow reactor trip is one of the j { trips ENABLED oy the P-7 permissive. - ! c) When actuated, the F-10 permissive will automatically i DE-ENERGIZE the high voltage to the Source Range i Instrument, but it will NOT RE-ENERGIZE the SR 1 Instrument high voltage when P-10 clears. i (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) l ! I
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6.__ELONI_sygIEMS_DEQ1GNt_CQUISQLt_GNQ_INSIBUMENIGIlON PAGE 5 QUESTION 6.09 (1.50) Concerning the Rod Control System: a) Place the following components in their proper flow path order. Start from the normal power supply and ending at the CRDM's j 1) DC hold cabinet
- 2) Power cabinet
; 3) Motor generator set i 4) Reactor Trip Breaker 1
- 5) Automatic Rod Control Unit l 6) Rod Position Indication Cabinet i
- 7) Logic Cabinet
; b) For the components in Part a), above, STATE the number j of each present in the system.
4 1 i OUESTION 6.10 (1.50) ' ) Match the RCL penetrations in Column A with the appropriate j RCS loop segment listed in Column B. (Answers may be used j more than once)
- Column A Column D a) Normal Letdown 1) Loop A cold leg
{ b) PZR Surge Line 2) Loop C hot leg d i c) CVCC Normal Charging 3) Loop A intermediate leg i d) PZR Spray Line 4) Loop B intermediate leg i j e) RHR Suction 5) Loop C hot leg j 6) Loop B cold leg
- 7) Loop C intermediate leg
- 8) Loop A hot leg t
i i j QUESTION 6.11 (2.00) i ! LIST 4 of the 5 Design bases for the ECCS Cooling I Performance following a LOCA as stated in 10CFR50.46. { l l t l (***** CATEGORY 06 CON TINUED ON NEXT PAGE * * * **) i I l \ -.
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1 1 6.__ELONI_SYSIEMS_DESIGU1_GONIBg61_GUD_lUSISUMENTATIgN PAGE 6 4 1 I i OUESTION 6.12 (1.50) What is the Technical Specification bases for the minimum , ! level requirements of the Condensate Storage Tank. ; 1 1 i ! i ! j QUESTION 6.13 (2.00) l l List 5 protection logic signals generated by the Pressurizer l l Protection System. (Include in your answer set points, !
- coincidence and associated interlocks, if any)
I l QUESTION 6.14 (1.50) List 3 purposes of Rod Insertion Limits. ! l ! OUESTION 6.15 (1.00) List the 4 requirements / control manipulati6n3 that will make up the logic to manually close the Diesel Generator output breakers (1GH3). I OUESTION 6.16 (1.50) I List the automatic start signals for the "A" and "C" charging pumps. (Assume "C" is on its normal power supply) l OUESTION 6.17 (1.50) l l State 3 reasons for having HCV-1142 (RHR letdown penetration from the RHR heat exchangers) kept about 15% open? i OUESTION 6.18 (1.00) l The Reactor trip breaker shunt trip coils have been modified I to also energize upon any trip signal to the Undervaltage l coils. What is the reason for this modification? l l (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) i i
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6.__ELGUI_SYSIEUS_ DES 10N 1 _GOUIBQLt_GUD_INSIEUUEUIGIlgd PAGE 7 QUESTION 6.19 (2.00) a) What are al1 the conditions which wi11 generate a #3 EI)G auto start signal? b) If the Unit #1 switch is in BYPASS and the Unit #2 switch is in NORMAL, for #3 EDG control, what will happen to 15J3 and 25J3 if a loss of off-site power occurs? QUESTION 6.20 (1.25) Describe how the High Steam Line Flow S1 input varies and the parameter on which this program is based. l OUESTION 6.21 (1.00) The Detector Current Comparator rece1%es input trom all 4 upper and lower power rango detectors. How are these inputs compared, and what conditions are needed to auto bypass circuitry while at p ower 1 DUESTION 6.22 (1.75) l a) What consequenceu could be expected in the Rod Control System's DC Hold Cabinet if 2 or more groups of rod drive mechanisms were placed on hold power (excluding control Bank D rods)9 b) Why in there both a 125 VDC and a 70 VDC power supply in the DC Hold Cabinet? l l l QUESTION 6.23 (1.00) Describe the dosign feature on the Main Steam Syst em that prevents reverse flow in the event of a Main Steam Line Dreak upstream of the Main Steam ' Trip valve. Include .t n your answer how it is a c c orop l i sh ed . (***** END OF CATEGORY 06 *****)
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a Z. __EBOCEDUBES_ _NOBdOL1_GENOBdOL1_EMESGENCLOND PAGE 2 EODIO.LOGICOL CONIBOL OUESTION 7.01 (1.00) On a lous of condenser vacuum where vacuum is greater than 20" Hg and decreasing, which of the following is NOT an immediate action?
- a. Place an additional not of air ej ec t or s in operatton
- b. Start a hogger
- c. Start an additional circ pump
- d. Reduce turbine load
- e. Sturt an additional condensate pump, if available DUESTION 7.02 (1.00)
Pr i or to operating Reactor Coolant Pumps in accordance with OP-5.2, Reactor Coolant Pump Operationu, the minimum seal flow should be ___ gpm and VCT prcanure should be a minimum of ____ psig.
- 1. O , 10
- 2. O.2 , 15
- 5. 2.O , 30
- 4. 5.0 , 20 OUESTION 7.nz (t.00)
Which e4 l- h e f011ouiog tr correct as 1t applies to NON-EMERGENCY procedural d e n a t i on C'
- a. The bNSOC must appr ove procedural deviations prior to performtnq the deviated step.
- b. A " Request to Change Procedure" form iu required for a temporary procedure deviation,
- c. A "F r ocedur e Deviation" form zu NOT required for typo-graphical errors in frequently used procedures.
- d. The ENDOC may pre-approve procedural deviationu.
OUESTION 7.04 (1.00) What epi r a tor activir nr e required upon evacuating the control room if the reattor could not be tripped befare exiting the control room? (164t* CATEGORY 07 CONTINUED ON NEXT PAGE $4**4)
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4 1 Zz__EBgCEDUBES_;_UgBUO61_8pNQBUGLt_EUESGENCy_8ND PAGE 3 f 1 80919b991C86_CgNIBQL i j i l OUESTION 7.05 (1.00) ' l Which of the following describes a temporary change which alters the INTENT of a procedure? ]l a. A change that corrects an incorrect valve lineup, i i b. A change that modifies the criteria by which a system's } operabi1ity is determined.
- c. A change that allows partial use of a procedure to test a subtrain without affecting remaining equipment in that train. !
1 1 { d. A change that allows you to change incorrectly specified [ { instruments for data taking. j QUESTION 7.06 (1.00) i . ] If you are in a 100 mrad / hour gamma field for 45 minutes, what is your dose f
; in mREt1 af ter 45 minutes? l 4
- a. 45 !
s , i
- b. 75
- c. 450 1
- d. 750 t
t i 1 OUESTION 7.07 (1.00) i One of the cource range channels falls on a reactor startup just above j the point where P-6 is a r. t u a t ed . Which one statement below describeu the
- correct action (s) that should be taken by the operator?
4
- a. Inuert the control banks to the fully inserted position and repair the source range instrument before increasing power abovo P-6 again, i
i b. Continue with the reactor startup, i l c. Insert control banks until below P-6 . then repair the malfunctioning i nource range channel before continueing with the startup. 1 1
- d. Dorate the RCG to the shutdown margin requirements of the l applicable Technical Specifications ucction.
i 1
- (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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, Zu_EBOGEQQBES._ _UQBdOLt_68UQBUGLt_EdEBQENQLONQ PAGE 4 E0010LQelGOL_GQNIBQL l QUESTION 7.08 (1.00) A hydrogen bubble formed in the reactor vessel is eliminated by
- a. increauing pressuricor temperature above core thermocouple readings.
l l b. .njecting oxygen into the reactor coolant nystem via the chemical and volume control system.
- c. maximizing coolant flow by running all reactor coolant pumps, increasing letdown flow to 120 gpm, and placing the cation bed demineral' err in service in parrallel with the mixed bed demineralizer.
l d. venting the reactor vessel head. QUESTION 7.09 ( 1. '50 ) r For each ;f the following, indicat.e YES or NO if the conditions violate
; critical safety function (CSF) red path criteria.
I a) Preusurizer level of 5% and RVLIS upper head GOX J b) Total AFW flow 400 gpm with all S/O lovelu < 6%(NA) c) Containment pressure 65 poig i OUESTION 7.10 (1.30) Answer the following questionn r egarding EOP usage TRUE or FALSE: i
; a) If a Function Restoration Procedure (FRP) is entered due to an ORANGE
{ Critical SMety Function (CSF) conditica, and a HIGHER priority ORANGE condition tu encountered, the artginal FRP must be completed prior to pr oceedi ng to the newly idontified FRP. b) Unless specifled, a tauk need not be fully completed before proceeding j to a nubcoquent stop as long as that tank is progrenning natisfactorily c) If a procedure transition occurn, ar y tasks still in progress from the
; procedure which was in of Fuct need r.at be completed.
l ($44** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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b i 7. PROCEDURES - NORMAL t_APNORMAL t_ EMERGENCY AND PAGE 5 , R @.I0,mLOGICAL CONTROL ' 1 i i OUESTION 7.11 (2.00) 1 j Prior to a reactor startup, with the RCS at normal operating pressure and ! i temperature, the following RCS leakages exist. For each leak listed below, ! indicate whether you could S TARTUP or would Lave to remain SHUTDOWN. ' (Treat each leak below as an independent event) t 1 a) A leak f rom an unknown source of 1.5 GPM. ! I b) 6.0 GPM from a manual valve packing gland. ; c) 0. 4 GPt1 from one S/G. d) 0.1 GPH from the reactor vessel head INNER seal. OUESTION 7.12 (1.00) Anuwer the following TRUE or FALSE concerning verification of tagouts:
- 1. Closed valves shall be verified closed by cracking them open and i
reclosing I i 1 2. Indirett verification methods uuch an use of indicating lights, or other I { indication will only be used if ALARA concept makes verification , 1 impractical. I OUESTION 7.13 (2.50) t Match the termu in column A to the values in column B for the radiation exposure guidelines. Assume whole body done unlers otherwise ststed. CAUTION: Some answers could be uned more than once. (0.5 ea) COLUt1N A COLUt1N B
- a. NRC limit 9/qtr 1. 0.5 REM 1
I
- b. Virginio Power limitn/qtr 2. 1.25 REM i
- c. NRC pregriant woman 1imit/ gestation 3. 1.O REM
' ( l
- d. NRC general public 1imi t / year 4. O.75 REf1 l
- e. NRC quarterly limit with a Form 4 5. 5 REM i
j 6. 3 REM l (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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; is ruptured. I i
d 7.15 ( .75)
- l QUESTION
} Complete the following concerning temporary modifications: ! 1 !
- 1. Modificatians not control 1ed by approved procedure wi11 not be used 1 within the station without prior knowledge of and approval
! from the ________ ____.,_. i l 2. Modificationu installed on safety related syatoms or systemn co: . taining ! radioactivity shall be approved prior to installation by the_ ______,__. I i i i
)
2 OUESTION 7.16 (1.00) ( j Following a valid reactor trip and safety injection, what are the Reactor i Coolant Pump Trip Criteria? (Ascume normal containment conditions) l } l ) DUESTION 7.17 (1.00) i
! List four of the critical conditionu r equired to be recorded during a j startup when i X 10E-8 amps in attained.
1 l a i , i
! OUESTION 7.18 (1.00) l J l Liut the immediate operator actions to initiate emergency baration if it in f required on an Anticipated Trannient Without Trip condstion. Asnume Gafety Injection has not accuated and in not desired.
i OUESTION 7.19 (1.00) Lint the 4 DISTINCT hazards to which personnel are exposed when an entry 1 into the reactor compartment in made during reactor operations. t i l u i 1 (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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i 2.__EBgggguagS_;_Nggd86t_8gNQBd862_EdgBGENGy_8Ng PAGE 7 80gIOLQGIC86_CgNIBg6 I QUESTION 7.20 (2.50) i !
- List FIVE indications of a loss of Component Cooling Water in accordance '
with AP-15, Loss of Component Cooling. 4 QUESTION 7.21 (3.00) j List all immediate operator actions for a loss of refueling cavity level
- during refueling in accordance with AP-22, Fuel Transfer Equipment 1 Malfunction and Loss of Refueling Cavity Level During Refueling.
! (Do not include system alignments or specific actions required '
i to complete the step) i ) } QUESTION 7.22 ( .50) 1 Unit 1 is operating at 307. power when a valid Reactor Coolant Pump ] vibration DANGER alarm is received for one pump. What are your immediate actions? 4 i QUESTION 7.23 ( .75) l What constitutes a Class II reactor trip? I i QUESTION 7.24 (1.00) During a natural circulation cooldown, it is desired to cooldown using the steam dumps. Which MODE is the steam dump system operated in and WHY7 i ( 1 1 i i 2 1 i f b t (***** END OF CATEGORY 07 *****) (************* ***************) b
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1 ez__8DMINISIB8IIME_BBgCEDUBES 2_CQNDII,'ONS2 _6NQ_LidIISIlgNg PAGE 2 l QUESTION S.01 (1.00) Which of the following require activation of both the TSC and OSC7 i
'; a. Either an unusual ev?nt, alert, site area emergency.or general '
4 emergency.
- b. Only an alert, site area emergency, or general emergency.
t
- c. Only a site area emergency or general emergency.
- d. Only a genera'l emerg ency.'
- ~.
._~.
. x QUESTION S.02- (1.00) j 1 -t During a reactor plant cooldown uc-1.;g the RHR system, the Safety /'l Injection System is automatically actuated. Since the normal EP SI termination criteria do not apply, which of the following would be the S1 termination criteria in thl's condition? i
. .c
- a. No criteria exists, terminate SI immediately.
1
\
- b. RCS pressure stable or increasing AND RCS subcooling greater than 10 degrees. -
- c. RCS pressure stable or increasing OR RCS subcooling greater ,
! than 50 degrees with 1 S/b'WR level > 6*J%. l
- d. Initiating condition cleared.
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O. ADdlNISIB8IlME_EBQGEDUBES 2 _GQNDlIlgNSg_8NQ_Lld118IlgNS PAGE 3 J ! OUESTION 8.03 (1.00) If control power is lost to a Unit 2 pressuriner power operated relief valve while in mode 1, which statement below is correct? I a. Tech specs require no action provided another PORV is operable and all pressurizer code safety valves are operable. , ., i j b. Tech specs require the power supply to be removed from the i associated block valve after verifying it to be open, if the PORV is not operable within 1 hour and contnuous operation is desired.
, c. Tech specs require the associated block valve to be shut and its
! power removed if the PORV is not made operable within one hour and continuous operation is desirable. i d. Tech specs require action to be initiated within one hour to place
; the plant in at least hot standby within the following hour if the PORV is not made operable.
QUESTION 8.04 (1.00) ]
- In Mode 1 a control rod is determined to be INOPERABLE and UNTRIPPABLE due to excessive friction. Tech Specs require which action listed below?
- a. Reactor operation can NOT be maintained. Shutdown immediately.
l b. Restore the rod to operable status in 6 hours or shutdown. 1 f c. Position the remainder of the ross in that group to within
"+" or "" 12 steps of the inoperable rod.
{ d. Determine that the tech spec shutdown margin requirement is satisfied. , e. Utilize more restrictive rod insertion limits contained within l Technical Specifications. d I i i i t a f 3 (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) I l ) 1
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- 9.__0DMINigIB8IlyE_BBgCEDUBES 2 _CgNQlligNS 2 _8NQ_LIMII611gNS PAGE 4 QUESTION 8.05 (1.00) 4
- Which one of the following statements is correct regarding the control and 4 issuance of Special Order Tags (Blue tags)?
I
- a. These tags may be used by all departments except Health Physics
- b. These tags may be used in lieu of a mechanical danger tag.
4
! c. The Control Room Operator may authorize tag removal.
- d. The tag indicates who must be contacted to operate the equipment.
i l I ) ] QUESTION 8.06 (1.00) i According to Tech Specs, which of the following is the correct action to be } taken if the Radwaste Effluent Monitoring Line Process Monitor is out of ,
, service?
1 l
- a. Effluent releases cannot be performed until the Monitor is back in service.
?
! b. Effluent releases may be performed if Grab Samples are analyzed
'. every twelve hours during the release. j c. Effluent releases may be performed provided two samples taken l prior to the release are analyzed and do not exceed 10CFR2O limits I and two qualified staff members verify the release rate i calculations and the discharge valve lineup. 1 ]
- d. The effluent release may be performed provided a sample prior ,
to the release indicates that the Lower Limit of Detection (LLD) ! i is not exceeded for all the analyses required and subsequent hourly samples during the release confirm this condition continues to exist. l ) I i l t i i ) , (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) l
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- 8. ADMINISTRATIVE PROCEDURES,_CgNDITIgNg2_AND_LIMITATIgNS PAGE 5 i
4 OUESTION O.07 (1.00) 4 Answer TRUE or FALSE to the f ollowing: l { a) IF a component's emergency power supply is INOPERABLE but all other i supporting equipment for that component is OPERABLE, than surveillance l requirements on that component must still be performed within the proper time frame. j b) If it is required by an LCO Action Statement to be in HOT STANDBY in 6 l hours and then HOT SHUTDOWN in the next 6 hours, it is permissable to be in HOT STANDBY in 3 hours then use the next 9 to be in HOT SHUTDOWN. OUESTION 8.08 (1.50) } a) During a non-emergency situation, who must authorize a temporary change ' to a operating procedure which does not change the procedural intent? (1.0) J b) What 2 forms are temporary changes and permanent changes to procedures ! documented on? (0.5) i 1 1
! OUESTION O.07 (1.00)
)j List 3 of the 4 pieces of information that the Shift Supervisor must give the Medical College of Virginia regarding a contaminated injured man being transported there, in accordance with EPIP 5.01. I i !
- QUESTION 8.10 (1.00)
What two individuals, by position, determine the applicability of a Dlanket Tag Authorization? 4 i OUESTION 8.11 (1.00) i List the three different panels / lockers for which keys must be available j on the key ring kept on the SS console for a Control Room Evacuation. i I , l 1 i J i j (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) 4
. -- -_ . . ~ ._ _ . . -_ . - - . ;i f "00g UNITED STATES
! NUCLEAR REQULATORY COMMISSION
' .#. .ot 3 .f. g REGION il 101 MARIETTA STREET, N.W., Sui 1E 2000 5 m
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az__09MINISIBOIlyE_BBgCEDUBES 3 _CgNQIllgNS 2 _BND_LidII611gNS PAGE 6 QUESTION 8.12 (1.50) List three administrative precautions that must be met to enter a Locked High Radiation Area (> 1r/hr). QUESTION 8.13 (1.00) What is the required complement, by department and number, of the Fire Brigade? OUESTION 8.14 (1.50) List the support equipment in ts 3.16, " Emergency Power System", required for a Diesel Generator to be considered operable (there are 6 different criteria that must be met). QUESTION 8.15 (1.50) List five hard copy sources of information that are referred to when performing a post trip review, following an unplanned reactor trip. QUESTION 8.16 (1.50) List the five conditions, as stated in SU-ADM-0-20, " Containment Evacuation During Refueling" that will initiate containment evacuation. QUESTION 8.17 (2.00) What are the 4 conditions listed in the EPIPs that dictate when updates should be given to offsite authorities regarding an emergency, subsequent to the initial 15 minute notification? (***** CATEGORY 08 CONTINUED ON NEXT PAGE n***)
sksErG0 UNITED STATES 77 9(o g, NUCLEAR REGULATORY COMMISSION 2 f, g REGION il l E 101 MARIETTA STREET. N.W.. SUITE 2900 o ! ATLANTA, GEORGI A 30323
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l Bs__ODULUISI68IIVE_PBQQEQUBES1_CQUDlI1QUS1_6UQ_blM1IGIlgNS PAGE 7 1 ! l J 3 UUESTION 8.18 (1.50) t
- a) What are the two emergency exposure limits addressed in EPIP 5.06,
" Emergency Radiation Exposure Authorization", for damage control considerations. (0.5) l, j b) List 4 criteria that should be considered by the Station Emergency j Manager in selecting personnel for emergency radiation exposure (1.0) 4 l
OUESTION O.19 (2.00) i j a) Why doesn't the Technical Specification for theOverpressureMitigationf l System require overpressure protection from the PORVs if the reactor ; j vessel head is unbolted? i i i b) Explain why the TS for the OPMS require that Pressurizer level be no i higher than 33% narrow range when Tavg < 350 degreen and the reactor , vessel head is bolted. ! c 1 1 j OUESTION O.20 (1.00) , As stated in 10CFR50.54, under what conditions may a tions be taken that depart from a license condition or a technical specification, and l who, as a minimum, must approve such action? l , !l i OUESTION O.21 (1.00) l
! Technical Specification 3.3, " Safety Injection System", requires that I neveral valven have their power removed when critical, and the valve in the l open or shut position. If testing is required on the Unit 2 Accumulator Isolation Valves, what criteria (include any time constraints) must be met r 4
in order to reretore power to these valven? OUESTION 8.22 (1.00) ' What is different in the forms and procedures that are required when performing a simple jumpur installation (e.g. lifted lead) an opposed to j a jumper inntallation requiring multiple steps (i e. beyond the skill of ' I the craft)? Ausume no approved procedure exists in either case initially. \ { (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) i l I
I UNITED STATES pkast%q'.o NUCLEAR REGULATORY COMMISSION
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j j 8z__0DdINigIB8Ilyg_BBQCEQUBES2_CgNp1IlgNS 2 _8ND_61dII8IlgNS PAGE 8 t I i j OUESTION 8.23 (1.00) I 1 4 Why, in the attached Tech Spec 3.1.D, is it a requirement to cooldown ' to less than 500 degrees as an action if the RCS activity limits are , t exceeded? i l OUESTION 8.24 (1.00) ! I The Shift SRO determines that an Emergency condition exists that l requires the generation of an Emergency Work Order, and the planning staff j is NOT available and the computer used to generate the f orms is out of service. What two forms must be generated, and to whom are they routed? J i i l OUESTION 8.25 (1.00) j Why does the Technical Specification for the CVCS require that a charging pump from the opposite unit be AVAILADLE when the applicable unit is ,. j critical? Include how this requirement was determined to be necessary. I 1 I t i i i i I ! l j l ! { i k I. i 1 I a i .i I i 1 i e (***** END OF CATEGORY 08 *****) 1 (**m*******w** END OF EXAMINATION ***************) I i
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i TS 3.2-1 10-12-84 3.2 CMDf! CAL AND V0 Ling C0!mtot SYSTDI Applicability Applies to the operational status of the Chemical and Volume Control Sys-tea. objective To define those conditions of the Chemical and Volume Control System necessary to ensure safe reactor operation. Specification A. Whas fuel is in a reactor, there shall be at least one flow path to the core for boric acid injection. The =4 ai== capability for boric acid injection shall be equivalent to that supplied from the refuel-
- ing water stortga tank.
- 8. For one-unit operation, the reactor shall not be critical unless the j following Chemical and Volume Control System conditions are met:
I l 1. Two charging pumps sha'll be operable' and one charging pump from , the opposite unit shall be available*.
- 2. Two boric acid tramafer pumps shall be operable.
- 3. The beric acid tanka (taak associated with the unit plus the comaos tank) together shall contain a minimum of 6000 gallons of at least 7.0% (but not >4.5%) by weight beric acid solution at a temperature of at least 112*T.
l
*Available means (1) operable except for automatic initiation instrumentation, j (2) offsite or emergency power source may be inoperable is cold shutdown, and (3) it is capable of being used for alternate shutdeva with the opening of the charging pump cross-cosaect.
t Amendment No. 99 and No. 98
]
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, TS 3.1-15 2-10-81 systes leakage. Radiation monitors which indicate primary system leakage in-l clude the containment air particulate and gas monitoring, the condenser air I ejector monitor, the component cooling water monitor, and the steam generator blowdown monitor. References i l FSAR, Section 4.2.7 - Reactor Coolant System I.eakage FSAR, Section 14.3.2 - Rupture of a Main Steam Pipe D. Maxieue Reactor Coolant Activity Specifications i
- 1. The total specific activity of the reactor coolant due to nuclides with half-lives of more than 15 minutes shall not exceed 100/E pCf/cc when- 71 ever the reactor is critical or the average. temperature is greater than 500*F. where I is the average sus of the beta and gamma energies, in Mov, per disintegration. If this limit is not satisfied, the reactor shall i
be shut down and cooled to 500*F or less within 6 hours after detection. Should this Itait be exceeded by 255, the reactor shall be made sub- ! critical and cooled to 500*F or less within 2 hours after detection. 1 4 i Ameneent No. 65, Unit 1 Amen eent No. 54, Unit 2
T.S. 3.1-23 3/ t./ 80 References (1) F3AA 4.2 (2) FSAR 9.2 , e G. Reactor Coolant System Overpressure Mitigation Specification
- 1. The Reactor Coolant systen overpressure aitigating systes shall,be operable as described below.
- a. Whenever the reactor coolant average temperature is greater than 350'F, a bubble shall exist in the pressuriser with the necessary sprays and heaters operable.
- b. 11henever the reactor coolant average temperature is 13,50'F and the reactor vessel head is bolted '
(1) A =amt=um of one charging pump operable. . (2) Two charging pumpe shall be demonstrated inoperable at least once per 12 hours by verifying the motor circuit breakers have been removed from their power supply or the benchboard 63 control switch is in the "FULI.-TO-LOCK" position. (3) Two operable Power Operated Relief Valves (PORV) with a lif t setting of 1433 peig, or (4) A bubble in the pressurizer with a maximum prdssurizar narrow range level of 333. Af tet a period of 72 hours, tve PotV's aust alas be operable 6 or (3) , The Raaetor Coolant systen vented through one opened FORV,
- or sa equivalent size opening.
- 2. The requiressats of Specification 3.1.C.1.b any be modified as follower
- s. One PORY may be inoperable for a period not to saceed 7 days. If the inoperable FORV, is not vestered te operable status within 7 days, then depressurite the RCS and open one FORY within the next 8 hours. ,
M endment No. 56, Unft 1 Mendment No. 55. Unit 2
- i. _ _10EQBLDE_UUGLEGB_ERWEB_ELGUI_0EE8011002_ELUIDS3_QNQ PAGE O IMEBt10DYNOt!1GS r.MaWERS -- SURRY 1&2 -87/02/09-DEAN, WM ANSWER 5.01 (1.50) 1
- a. true
- b. falso !
- c. faise ,
REFERENCE Surry lesson plan ND-86.3, pp3.4, 3.5, 3.7, 3.10, 3.12 193009: K1.05(3.5) ANSWER 5.02 (1.50)
- a. more nuqativo f
- b. more negative
- c. more negative REFEHENCE Stir r y lennon plan ND-06.2-LP-2, n2.11-2.17 ,
192004: U1.06(3.1) i ANSWER 5.03 (l.50) ; l (acceptable answers, 3 of any 4) 1
- a. fuel bitrnup i L
- b. f1eston product poinon bui1 dup
- c. power defoct
- d. heat up HEFERENCII Citr r y Iennon plan ND-06.2-LP-5, p5.5.
192002: K 1. O'? ( 2. 7 ) i i
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1 1 h _ _IUE Q8LQ E _UQC L EGB _EQWE 8_EL GU LQEE80I I Q th_ELQ 1 ph _OU Q PAGE 9l l' - THERMODYNAMIQQ l AN9WERS -- SURRY ! ?.< 2 -07/02/09-DEAN, WM ] l i i I i 4 l I l ANSWER 5.04 (2.00) [664 da , A. An the U-235/U-238 isotopes are depleted their fraction ! l of finnions decreaseu thun Doff decreason, l i- [ I j D. Production of plutonium isotoper utth smaller delayed ! neutron fractions decreases the ax , ojo delayed neutron i iraction over core lifo. I e l C. As the delayed noutron fraction decreases, one in likely l j to nuo a quickor response to change in power (i . e. more of a [ j prompt jump / prompt drop) t i ! i REFERENCE l tiurry l esson pl an ND-06.1-LP-7, p7.10 192003: V.1. 07 / K 1. 09 ( 3. 0/ 2. 9 ) I l l ANGWER 5.05 (2.00) 6 i A. It doubles (or increanuu) the head for a given masu Ilow
- rate. l i
j 11 It will double (or increases) the macs flow rate capacity for a oivon heiad. l HEFERENCE .j Surry 1 #"suon pl on ND--0 3 LP-0. Rov 1, pO.10 l 191004: E1.OY/1.10(2.S/2.4) , j l !l nNSWER 5.06 (1.On) J 4 C I i HEFEREN(:E j Gur ry l uuuuri pi no ND-0 3 -LP-0, Rev 1, j 191004: E1.14(2.5) , 1 i l l l l 1
- i 1 !
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r-UNITED $TATES
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L.__IMEQ8LQE_UQCLEG8 EQWE8_E66NI._QEEB01100t_ELUIDSi_0ND PAGE 10 INERMODYNOr'11GS ANSWERS -- GURRY 1&2 -07/02/09-DEAN, WM l ANSWER 5.07 (1.00) l With a DNDR of 1.3, duri ng norrnal operation and anticipated , I operational occurrencen, there is a 957. confidence that DND does not occur. When > 1.3, the liisulyhood of DND occurring decreason. , REF'ERENCE ; Surry lusson plan ND-06.3-LP-2, p2.10 193000: K1.10(3.1) ANSWER 5.00 (2.00)
- 1. reactor power
- 2. coolant flou rate :
- 3. R C 3 -- M ' ' ,,- - - ' ' ' TAmj 4 RCG preur4ure.
RLf'ERENCE i Surry 1eunon plan ND-06.3-LP-2, p2.10 i 193090 K1.05(3.6) { l L i ANSWER S.09 (2.00)
- 1. Dennity difference (or DELTA T) created by heat addition i by tho huat 5.nurco and huat romoval by the boat ninl' !
t I
- 2. The heat e, i nl: must t>e eluvoted phys,1 cully above the huat i nourco. l t
REFERENCC ! Gurry 1 onnon pl on ND -06. 3-LP-4, p4 D 1V3000 I:1.21(4.2)
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,*, NUCLEAR RECULATORY COMMISSION f.
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! N.__IBUOBY_GE_UUGLGOB_COWEB ELOUI 0EE60110Ni_ELUIRDt_0NQ PAGE 11 , THERMODYUGU1GH !
j ANSWERS -- GURRy i t<2 -07/02/09-DEAN, WM i 0 i ANSWER 'd , 10 (1.30) i n. DECREASES l i
- b. .m u u.e u a - 7MM(1(C.$
- c. DOEG NOT CHANGE i
REFERENCE Surry loucon plan ND-06.2-LP-1, p1.4, 1.11 192004: K1.07(2.9) i i j i l ANGWER L.11 (1.00) I : l b ! a 1 i i<EFCRENCE r l '
!iur r y l et.non p l an ND -06. 2-LP- 1, pl.16 1Y2004: K1.05(2.4) 4 r
AN:iWLR L.12 (1.00) l 0 : .I i REFERENCE ! ) Uurry luce.on p1an ND-06.2--LP-4 p4.11. i l 192006: Kt.02(3.t) I t I ; 1 ' I AN!iWER 5.1* (1.00) 4
- Nout.ronu at or nrnar thu utige of the core havo a htghor i j protnibili t y of loaling out than the onow, at the contre which l have .4 htghor probabt11ty of couwing f i ein i on . (i feric o s DhW at I j c en t er- in '
t h e-in a t edgo). l l 1 RLlEREilCL } Gor t y l o% tion p l 4n flD-06. 2- LP 6, p6.12. I 192005: Lt.14(3.0) l
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; , IU!itiURD10001Cg ! ANSWERG -- GlJRRY 162 -87/02/09-DEAN, WM i
J 1 j < i ANSWER 5.14 (1.00)
; The presence of adjacent control rods may cauce a l signtficant change in an individual control roo worth. l 1
i REFERENCE
! Ourry l ecuan pl an ND-06. 2--LP-6, p6.19 001/0003 f *J. 05 ( 3. 9 )
i l ANSWER G .1 *J (1.00' d REFLhENCE i Uurry leucon plan ND-06.2-LP-7, p7.20-7.34. I 1920tG l'.1. U 1 ( 2. O ) ANSWER 5.!6 (l.50) p.( r#{4(stad, haf Cd M hf Utart up rato in punitivo and constant, eactor power tu fMbuld (r .)rm increautn and there in no outward rod motion. ' NM 3d JWW() ! REl-E RLNCE
; Gurry lesuon plan ND-06.2-LP-7, p/.51.
, 19 28 tvd i l'.1. I 1 ( 3. U ) l l i 1 l ONOWLR D.17 (1.u0) (any 2 of the the 3 ) ! 1. Neutron production ir,relatively high, no power lu j conotant whun the ruattor is critical.
) 2. lWlou 10 eup - l) ampn the output of the intermodlate ranqu may not bo directly proportional to the neutron population.
- 3. React A vi t y han not yet been channett by the moderator or
! fool temperature. l l RLTEkFNCE Gur r y l esnion pl.as ND-H6.2-LP-/, p .' . 07 l'/2:10 0 : 1.1.12(3.6) i I i
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5.__ISEQBy_QE_NyCLEOB_EQWEB_E68NI_QEEB8IlgN2_ELQlDS 2 _6ND PAGE 13 ISE8MQDYN8t!1CS ANSWERS -- SURRY 162 -87/02/09-DEAN, WM ANSWER 5.18 (1.50) ($ f r c, br au-f 3
- 1. Gradual warm up of steam lines
- 2. Proper venting of tanks and components during warm up and operation.
- 3. Steam traps
- 4. Lines kept full
- r. A.,+ pq w/ docu r9a w/s s /m3 d
(others as appropriate) - - - G , o Q W, G- 5tjfhst REFERENCE Q &lG l A t h rh Surry lesson p l an ND-83-LP--0, pO-36 193006; K1.04/1.10(3.6/3.4) ANSWER 5.19 (1.00)
- a. Lower
- b. Higher REFERENCE Surr y lesson plan ND-83-LP-8, p8.9/10 191004;K1.04(3.4)
ANSWER 5.20
- a. Maintain DNDR > 1.3 and core eyit enthalp < ~ _ .
saturated (+1.O> L{$~'y ppd Oo (UCshX d( 6'IlEY MNf'f [0- 7
- b. Prevent bulk boiling during normal operations (+0.5)
- c. Ensure fuel clad temperature < 2200 deg F during a LOCA(+1.O)
REFERENCE Surry lesson plan ND-86.3-LP-3, p3.12 193009;K1.07(3.3)
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i l j Qu__ISEgBY_QE_NgGLE88_EQWE8_E68NI_ GEE 8611gN3_E6ylpS2_9NQ PAGE 14 i . IMEBdgQYN801GS l l ANSWERS -- SURRY 1&2 -87/02/09-DEAN, WM i 1 1 1 ANSWER 5.21 (1.50) { a) > DNB, have partial film boiling , where the fuel rod is alternately covered with steam and water (+.25). Steam has poor thermal conductiv-ity capabilities (+.25), so heat transfer rate drops and Delta T rises (+.25) J ) b) As fuel surface temperatures rise, stable steam layer forms (+.25)
! causing a further increase in fuel rod temperatures (+.25). Eventually, I significant radiative heat transfer occurs causing heat xfer rate to incarease (+.25)
REFERENCE Westinghouse Thermal / Hydraulic Principles II, pp 13-18/20 . EPE-074; EK1.02(4.6/4.8) I i-l l l 1 l l
i pS88004 UNITED STATES
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l 6.__PLGNI_SYSIEUS_ DES 10Nt _CgNIB06t_GND_10SIBUMGNIGIICN PAGE 9l ANSWERS -- SURRY 18< 2 -87/02/09-DEAN, WM ANSWER 6.06 (1.00) a) STSP decreases (0.5 EA) b) STSP decreases REFERENCE ND 93.3-LP-14 p. 7 012/000; A1.01 (2.9/3.4) ANSWER 6.07 ( .50) TRUE (0.5) REFERENCE NA NCRODP 93.5 Rod Control 001/000; K4.03 (3.5/3.8) ANSWER 6.08 (1.50) a) TRUE (0.5 ea) b) FALSE c) TRUE REFERENCE ND-93.2-LP-4 015/000; K4.07 (3.7/3.8) ANSWER 6.09 (1.50) r -- ( a) (3) motor generator set ' b) 2 (4) reactor trip breaker ' 2 (2) power cabinet 4 (7) logic cabinet 1 (6) rod position indication cabinet 4 (5) automatic rod control unit 1 (1) DC hold cabinet 1 (0.75 for a) fully correct, 0.75 for b) fully correct
-0.1 for each switch needed to place a omponent ._ i n proper order) ,
r, REFERENCE 8 8 ND 93.3-LP-3 Rod Control System 001/000 K4.01 (3.5/3.8)
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6.__PL8UI_SYSIQUS_ DESIGNt _CQNIBQLt_GUQ_IUSIBudENIGIlON PAGE 10 ANSWERS -- SURRY 1&2 -87/02/09-DEAN, WM ANSWER 6.10 (1.50) a) 2? I [ b) 5 (0.25 ea) c) 6 d) 1,2 e) 8 REFERENCE ND 88.1-LP-7 PCB, Valves, Piping & Instrumentation 002/000; K1.09 (4.1/4.1), K1.06 (3.7/4.0) ANSWER 6.11 (2.00) (any 4 of 5 at 0.5 ea)
- 1) Max. Fuel Element Cladding Temp. < 2200 Deg. F
- 2) Cladding Oxidation < 17% thickness
- 3) Hydrogen generated by Zirc-Water reaction <1% of max.
possible.
- 4) Core remains in a coolable geometry
- 5) Provides for long term decay heat removal REFERENCE 10CFR50.46 ND 91.1-LP-1 006/050; PWLi 4 (4.2/4.3)
ANSWER 6.12 (1.50) Sufficient water available to maintain the RCS for 8 hours of residual heat removal (1.0) following a trip (0.25) with a loss of of f si te power (0.25) . REFERENCE NA Technical Specifications Bases 3/4 7.1.3 026/000; PWG-5 (3.3/4.1) i l 1 i ___a
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. ANSWERS -- SURRY 162 -87/02/09-DEAN, WM ANSWER 6.13 (2.00)
- 1) PZR Hi Press. Trip (0.2) 2370 psi g (0.1 ) , 2/3(0.1)
- 2) PZR La Press. Trip (0.2) 1875 psi g (0.1) , 2/3(0.1)
- 3) PZR Lo-Lo Press. SI (0. 2) 1715 psig and not blocked (0.1),
2/3 (0.1 )
- 4) P-11(0.2) m (2000 psig (0.1) on 2/3 (0,1)
- 5) Press. input: to the OT Delta T (0.4)
REFERENCE #M ~ M'*'rgI/cf .W du/c 9/W Supc./(,'2] ND 93. 3--LP- 12 / LP- 14 -
/AtTC ['OI2 V E k< -T s tq da{ [< 2) 010/000; K1.01 (3.9/4.1) -
gp g/3cg /w h/oct t (9,%f (,2) ANSWER 6.14 (1.50)
- 1) Adequate SDM upon trip (0.5 ea)
- 2) To minimize the amount of positive reactivity inserted during a rod e.J ec t i on accident, and
- 3) To minimize radial f l u:. tilt (peaking)
REFERENCE ND 93.3-LP3 p. 25 001/000; US.04 (4.3/4.7) i ANSWER 6.15 (1.00)
- 1) Control switch to close (0.25)
- 2) Synchronizing selector switch is ON (0.25)
- 3) Overcurrent/ generator diff trips reset (O.25) i
- 4) Aux trip relay reset (0.25) l l
REFERENCE < ND 90.3-LP-3, p. 12 064/000; A4.01 (4.0/4.3) l l l l l l l l l l l
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6.__PL6NI_SYSIEMS_DESlQUx_CQUIBQLt_QUQ_lySIEy0EUIG110N PAGE 12 ANSWERS -- SURRY l !<2 -87/02/09-DEAN, WM Q.0 ANSWER 6.16 -
"O Charging Pump A
- 1) Low chaging header pressure (0.25)
- 2) SI Train A [0. Z. d
- 3) Undervoltage or degraded bus voltage on emergency bus J (0.25)
- 4) All other charging pump breakers (f or charging pumps B, C(alt) and C normal ) open. (0. 25)
Charging Pump B
- 1) Low discharge header pressure (0.25)
- 2) SI Train A or B (0.25)
- 3) Undervoltage or degraded bus voltage on emergency bus J (0.25)
- 4) Charging pump A R< B breakers open (0.25)
{ 5) C alt. supply must be open for any C normal start } REFERENCE ND 88.3-LP-5 p. 5.12/5.13 004/000; K4.04 (3.2/3.1) a Tfcu \ ANSWER 6.17 (1.50) {c t G q $ / To provide a path to keep the RHR system full (0.50) and to al l ow for e::p an si on of the system during heat up of the RCS (0.50) and thus ambi en tl y heating up RHR (0.50). c, t - S Kow" (* Pwik lor (< Q f '-I REFERENCE J <f(( W iidb 44 Y0 1 [O 50) ND 88.2-LP-2 p 2.8 004/000: K1.01 (3.4/3.9) ANSWER 6.18 (1.00) Tho design change resulted because of ex peri ences where the undervoltage trip signal alone was not sufficient to trip the breaker. (1.0) REFERENCE ND C73.3-LP-10 p.9/10 012/000; K6.03 (3.1/3.5) l
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. ANSWERS -- SURRY 1&2 -87/02/09-DEAN, WM
[ ANSWER 6.19 (2.00) i ! a) -Unit 1 or 2 Train B SI (0.33 ea)
-Unit 1 or 2 Train B Hi-Hi CLS
-1J or 2J Bus undervoltage b) Race between breakers to close (0.5) j Whichever closes first locks out the other breaker (0.5) i i
REFERENCE j ND 90.3-LP-7 l 062/000; K3.02 (4.1/4.4) l !
?
ANSWER 6.20 (1.25) I 38% setpoint from 0-20% (0.5) Turbine power (0.25) ! and linearly from 38-108% as Turbine Power goes from 20-100% l (0.5) REFERENCE ND 91-LP-3, p.9 013/000; K1.01 (4.2/4.4) I I l ANSWER 6.21 (1.00) The highest reading upper / lower detector is compared to the i average of the upper / lower detectors (0.5). The circuit ! ! auto defeats below 50% power on ALL channels (0.5). l l l REFERENCE I ND 93.2-4 015/000; K6.04 (3.1/3.2) & A1.04 (3.5/3.7) 1 I i I ) ANSWER 6.22 (1.75) l [ a) Cabinet has the capacity to support up to 6 stationary l gripper coils simultaneously (0.5). So with 2 groups or more, would overload / heat the cabinet (0.5). b) 125 VDC-Latching Rods ! 70 VDC-Holding Rods (0.5 for reasons, 0.25 for correctly ' associating voltages). I t i i
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-- SURRY 18<2 -87/02/09-DEAN, WM i
I ] REFERENCE j ND 93.3-LP-3. p.15/16 ) J I ANSWER 6.23 (1.00)
- Accomplished by the Non-Return valve. (O.25) l A reverse differential pressure is sensed across the valve 1
(0.25) j A sensing line on the down stream side of the valve which directs down stream pressure to the upper area of the valve ' 4 disc. (0.25) Which drives the disk down when up stream pressure is < downstream pressure.(0.25) 1 } REFERENCE ND 89.1-LP-2 039/000 K4.06 (3.3/3.6) I i I f l i 4 i c i I 1 i
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ANSWERS -- NORTH ANNA 1R<2 -87/02/09-DEAN, WM , l i i ! 1 i i l ANSWER 7.01 (1.00) ,
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i i REFERENCE i l McG, AP/2/A/5500/23, p. 2 l Surry AP-14, pp 4 ; EPE-051; PWG-11(3.7/3.7) l ! i ANSWER 7.02 (1.00) 2 i REFERENCE ,
- VCS, SOP-101 p1 l
NA OP-5.2 p 4 ; i SUR OP-5.2 p 2,4 l l i ) 4
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5 ANSWER 7.03 (1.00) I i 1 d
, REFERENCE l
Surry SUADM-0-10 p 17 i PWG-23: Shift Staffing and Activities (2.8/3.5) ANSWER 7.04 (1.00) 9 0.5 points each:
- 1. Trip turbine locally.
i 2. Manually open reactor trip breakers or the rod drive MG output breakers. I REFERENCE NAPS 1-AP-20, p.3. SUR 1-AP-20, p5
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i Z __EBOCEDUBES_ _NQBM86t_6BNgBM8kt_EMESQEtjCy_6tjD PAGE 9 RADIOLOGIGGL_GQNTRgL ' ANSWERS -- NORTH ANNA 1&2 -87/02/09-DEAN, WM l ANSWER 7.05 (1.00) l l b l REFERENCE NA ADM 5.8, pp 2/3 Sur SUADM-ADM-21 p 21 PWG-23: Plant Staffing and Activities (2.8/3.5) i < ANSWER 7.06 (1.00) i b 4 OF=1 for gamma } 100 (45/60) (1 ) =75 REFERENCE ! 10 CFR 20. PWG-15: Radcon Knowledge (3.4/3.9) 1 l ANSWER 7.07 (1.00) i 1 b t REFERENCE VCS, T/S p 3/4 3-2, 3/4 3-6 j NA T/S Table 3.3-1 l SUR T/S Table 3.7-1 < l i f I l ANSWER 7.08 (1.00) 1 i d t REFERENCE MNS EP/2/A/5000/16.3 { CNS EP/1//A/5000/2F3, p.7. l NAPS 1-FRP-I.3A, p.3. SUR F RF'- I . 3, p 9 i (
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, REFERENCE ! SDN TS 3.4.6.2 i NA TS 3.4.6.2 SUR TS 3.1-13 002/020; PWG-8 (3.5/4.4) i ANSWER 7.12 (1.00)
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-Unexpected rise in S/G 1evel (+.25 ea)
-High radiation on a S/G blowdown line
-High radiation on an MS line monitor
-High radiation as determined by sampling and analysis REFERENCE Surry EP-4.OO, pp 2 NA 2-EP-3, pp 2 j EPE-038; EA2.03 (4.4/4.6) i i
l ANSWER 7.15 ( .75)
- 1. Shift Supervisor [.3753
}
- 2. Superintendent of Operations E.375]
j REFERENCE SUADM-O-11 p 13 PWG-1, 3.5/3.9
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Z.__EBgCggy8gg_;_NQBM6Lt_6BUQBMOLt_EME80ENCY_6ND PAGE 12 RADIOLOGICAL CONTROL ANSWERS -- NORTH ANNA 12<2 -87/02/09-DEAN, WM ANSWER 7.16 (1.00) f u np (Lt Wl}
- 1) Verify Charging /SI vten (+.5 ea)
- 2) RCS Subcooling less than 25F REFERENCE SONP Foldout Page NA Foldout page for 2-EP-O Surry Foldout page for EP-1.OO 003/000; PWG-10 (4.1/4.4)
ANSWER 7.17 (1.00) Any 4 .D O.25 points each: North Anna Surry
- 1. Bank C position. 1. Date Critical
- 2. Bank D position. 2. Time Critical
- 3. Auct. High Tavg. 3. Average RCS temp.
- 4. IR N35. 4. RSC Baron concentration
- 5. IR N36. 5. Bank C Control rod position
- 6. RCS baron concentration. 6. Bank D Control rod position
- 7. Actual critical position within admin. requirements REFERENCE NAPS 1.OP-1.5, p.12.
SUR 1-OP-1C App. A p 10 of 10 001.010; K5.OB (2.9/3.3) ANSWER 7.18 (1.00) Surry (+.25 ea) North Anna (+.25 ea) Verify SI/CHG pumps running / flow 1. Verify 2 SI/CHG pumps running / flow Check RCS pressure <2335 psig 2. Switch BATP to fast speed Switch BATP to fast speed 3. Open MOV 2350 or Inject the BIT Open MOV-( )350 4. Check pzr press <2335 REFERENCE Surry FRP-S.1 p3 NA FRP-S.1 p 4 EPE-029; PWG-11 (4.5/4.7) .i
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Zs__EBgCEDUBES_;_NgBU66t_6BNgBM66t_gMEBGENCY_6ND PAGE' 13 RADIOLOGICAL CONTROL ANSWERS -- NORTH ANNA 1&2 -87/02/09-DEAN, WM > ANSWER 7.19 (1.00) icinizing radiation; heat stress; differenti.1 pressuFe; O2_ deficiency (+.25 ea) l~ REFERENCE ., Surry-SUADMO-19 p 3 NA ADM 20.9, pp 1 , PWG-18: Knowledge of Safety Procedures (3.0/3.1) ANSWER 7.20 (2.50) CC surge tank low level alarm Reactor containment air Motor protection alarm CC pump recirc coolers hi-temp. CCW low flow discharge header alarm CCW low pressure discharge header alarm 59 . 5 ea Reactor coolant pump flow / temp alarm Excess letdown HX flow / temp Non-regenerative HX (Mian Coolant) .high temp Primary shield water wall coolers low pressure l Primary shield penetration cooling coils low pressure
- Noutron shield tank coolers low flow /hi temp
~
REFERENCE
! SUR AP-15, p 1 4
1 008/030 PWG-10 3.8/4.2 i i ANSWER 7.21 (3.00) 1
- 1. Halt all operations of the equipment involved
- 2. Place fuel in the safest condition possible
- 3. Immediately commence make-up to the refueling cavity
- 4. Return any fudl assemblies to the core that are in the manipulator or ref uel i ng c'avi ty , ,
- 5. Close or verify clo,(ed the fuel transfer tube gate valve
- 6. Verify the integrity of'the fuel assemblies
.50 ea REFERENCE SUR AP-22 p5 s 1
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- 1. Place delta T and Tave Defeat switches to defeat the affected loopE.125]
- 2. Select delta T recorder to unaffected loopE.125]
- 3. Trip the affected pump and commence shutdown (IAW OP-3.1)[.25]
REFERENCE SUR AP-9 p 3 000/015 PWG-10 4.2/4.5 ANSWER 7.23 ( .75) Cause not clearly understood (+.25) or safety related/important equipment operated in an abnormal ar degraded manner (+.5) i REFERENCE Surry SUADM-O-02 p 3 NA ADM 19,18, pp i PWG-10: Recognizing abnormal indications (4.1/4.5) ANSWER 7.24 (1.00) Steam pressure mode EO.25] Tavg input to the steam dump control is not valid without forced flow in the loops. [.753 O E-REFERENCE
- Tnuj tMirholrto$ dfpl<Gr Wf h*]N*rltU /- F/ ] F (~' y g) i NA 1-AP-10, Att. 2, p 2 SUR AP-39, p4 1
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- (e) i j REFERENCE
! SUR TS 3.12-8.C.6 i i
- l. 001/050; PWG-5 (2.9/4.3) i i
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0.__0DMINISIBOIlyE_PBQCEDUBESx_CONDlI1QNSg _6dD_61MIIGIlgNS PAGE 10 At4SWERS -- SURRY 1&2 -87/02/09-DEAN, WM ANSWER 8.05 (1.00) d REFERENCE NA ADM 14.0, pp 6/7 SU-ADM-O-13 PWG-14(3.6/4.0) ANSWER 8.06 (1.00) a REFERENCE TPT TS 3.9 NA TS 3.3-12 Surry TS table 3.7-5a 073/000; PWG-5(3.0/3.8) ANSWER 8.07 _1( ru m C f) I a) TRUE (+.5 ea) ur~ 2=>"- W (th( REFERENCE TPT TS B3.0.1, 93.0.5 NA TS B3.0.3/B3.0.5/B4.0.3 Surry TS 3.01/3.02 PWG-5; (2.9/3.9)
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3 8.__8DMINISIB8IlyE_PBQCEDUBES 2 _GQNDlIlgNS 2 _8ND_61dlI8IlQNS PAGE 11 ANSWERS -- SURRY 1842 -87/02/09-DEAN, WM I I I I i I l' ANSWER 8.08 (1.50) l I A a) two SROs (+.5)of which one must be the shift supvr or Ops Supt.(+.5) b) Temp: Procedure Deviation (+.25 ea) i Perm: Request to Change Procedure ! REFERENCE { Surry SDM-60, pp 19/21 ! NA ADM 5.8, pp 4,5 1 PWG-23(2.8/3.5) Change Procedure i I } ANSWER 8.09 (1.00) i j 1) Time of accident (+.33 ea for any 3) j 2) Severity of injuries j 3) Dose received by victim j 4) Is victim neutron irradiated ] dM0 ' c) g. gp j t c y a REFERENCE NA EPIP 5.01, pp 2/3 Surry EPIP 5.01, pp 3/4 l PWG-36(2.9/4.7) I i i 4 j ANSWER 8.10 (1.00) i j Shift Supervisor and Operations Maintenance Coordinator (+.5 ea) j l REFERENCE SUADM-0-13, pp 15 PWG-14(3.6/4.0) l ANSWER 8.11 (1.00) l 1 App R locker, Aux S/D Panels, EDG break glass panel (+.33 ea) i 1 REFERENCE ! Surry SUADM-0-09, pp 27
. PWG-23(2.8/3.5) i
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- 1) Dose Rate Meter is required to be on/ monitored continuously (+.5 ea for any 3)
- 2) Use of buddy system is required (two people in constant contact or communication)
- 3) Two people must sign for key issue
- 4) The entrance is guarded while area is occupied (can be locked if egress is not hindered)
REFERENCE Surry HP Manual 2-12, pp 3/4 PWG-15(Radcon) (3.4/3.9) ANSWER 8.13 (1.00) 3 operations (includes the scene leader) and 2 security (+1.0) REFERENCE Surry ADM-29.2, pp 14 PWG-19(3.4/4.2) ANSWER 8.14 (1.50) i
- 1) Sufficient capacity in day tank (290 gal) (+.25 ea)
- 2) on-site supply (35,000 gal)
- 3) 2 operable fuel flowpaths
- 4) EDG battery operable 1
- 5) EDG Charger operable i
- 6) DC Control Circuitry for the EDG operable REFERENCE Surry TS 3.16 064/050; PWG-5(3.1/4.1) l I
I j
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8,__eQMINI@IBOIIME_BBQCEQUBES2_CQNQlllQNS2_8NQ_LIMlI@IlQNS PAGE 13 ANSWERS -- SURRY 1&2 -87/02/09-DEAN, WM ANSWER 8.15 (1.50)
- 1) Sequence of events recorder (+.3 ea for any 5)
- 2) P-250 Alarm Typewriter
- 3) Strip Charts
- 4) Logs
- 5) Completed Procedures
- 6) Post trip review printout REFERENCE SU ADM O-02, Attachment B
- NA ADM-19.18, Attachment 8 PWG-28(2.9/3.5)
. ANSWER 8.16 (1.50)
- 1) Loss of source range audible counts (+.3 ea) 3
- 2) High flux at shutdown alarm j 3) Station evacuation alarm j 4) Announcement of containment evacuation
- 5) Fire alarm
] 1
- REFERENCE i SU-ADM-0-20 i
! 103/000; A2.04(3.5/3.6) a I ANSWER 8.17 (2.00)
- 1) approximately 30 minute intervals ( +. 5 ea )[o r C.g
- 2) Significant change to meteorlogical data .
- 3) plant status
- 4) radiological data
& 2WuN Nc' nut ttJ~
REFERENCE Surry/NA EPIP Note following State / County Notification Step PWG-36(2.9/4.7)
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et__GQuLUISIBGIlyE_P8QCEQUBES 1 _CQNQlIlgNS 2 _80Q_61dlIGIlgNS PAGE 14 ANSWERS -- SURRY 1&2 -87/02/09-DEAN, WM t i j ANSWER 8.18 (1.50) j a) Whole Body- 25 Rem (+.25 ea) Thyroid- 125 Rem
- b) --volunteers (+.25 ea for any 4) 1 - professional rescue personnel
- - good physical health
--above the age of 45 I --should not be a woman capable of reproduction
--familiar with consequences of exposure REFERENCE j Surry/NA EPIP 5.06, pp 3/4
- PWG-36(2.9/4.7) a 4
4 i ANSWER 8.19 (2.00) i,
; a) If the head is unbolted, a RCS pressure of < 100 psig is sufficient to 1 provide the relieving capacity of a PORV (+1.0) ;i b) Gives sufficient time for an operator (approx 10 minutes) to respond f in case a malfunction resulting in man charging flow from one Chg j pump. (+1.0) i l REFERENCE l Surry ND-93.3-LP-6, pp 6.9/11 010/000; PWG-5(2.9/4.1)
ANSWER 8.20 (1.00) I j In an emergency when the action is needed to protect the health and safety of the public (+.75), approved by at least a licensed SRO (+.25). REFERENCE 10CFR50.54 PWG-36(2.9/4.7) i i 1 t I a l 2
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1, 8.__6DMirjlgIB611VE_BBQCEDUBES1 _CONQlIlgNS2 _8ND_61MlI8IlQNS PAGE 15 , ANSWERS -- SURRY 1&2 -87/02/09-DEAN, WM r i . I i l { ANSWER 8.21 (1.00) i Can only restore power to one valve at a time (+.75) and must remove
- power and complete testing in 4 hours (+.25)
I REFERENCE I Surry TS 3.3-5 3 006/050; PWG-1(3.8/4.0) & PWG-5(3.2/4.3) i l 1 ANSWER 8.22 (1.00) j Simple jumpers require a Jumper Log Form (+.5) whereas a more complex jumper operation would also require a controlling procedure (+.5) , REFERENCE r i NA ADM-14.1, pp 5 l SU-ADM-0-11, pp 14 ^ PWG-14(3.6/4.0) ANSWER 8.23 (1.00) i With Temperature < 500 degrees, the release of activity to the environment I due to a SGTR is precluded (+.7) since Paat is< SG PORV lift setpoint(+.3) , i REFERENCE j TPT TS D3.1-6 NA TS D 3/4.4.8
. Surry TS 3.1-17 i
I , 002/020; PWG-5(2.9/4.1) a 1 l ANSWER 8.24 (1.00) l A hand written W.O. must be generated and routed to the Cognizant j Department Supervisor (+.5) and a work request card must be forwarded
! to the Planning Dept (+.5)
REFERENCE l SUR SUADM-t1-13, pp 3 i I i, l
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l St__0DdlNISIB8I1Mg_EBQQEDUBES 2 _GQNQlIlQNS1_8NQ_(IdlI8IlgNS PAGE 16 i ANSWERS -- SURRY 13<2 -87/02/09-DEAN, WM , t l PWG-23(2.8/3.5) l ANSWER 8.25 (1.00) ' Identified during fire protection reviews (+.25). Required to bring a ] disabled unit to cold shutdown conditions (+.75) l i REFERENCE
! Surry TS 3.2-5 I
l 004/000; PWG-6(2.7/3.6) i l i l f 1 l 1 4 i j i : I s i I i i l i l 1 1 j
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TEST CROSS REFERENCE PAGE 1 'OUES, TION VALUE REFERENCE 05.01 1.50 WMD0001378 05.02 1.50 WMD0001379 05.03 1.50 WMD0001380 05.04 2.00 WMD0001381 05.05 2.00 WMD0001382 05.06 1.00 WMD0001383 05.07 1.00 WMD0001384 05.08 2.00 WMD0001385 05.09 2.00 WMD0001386 05.10 1.50 WMDOOO1388 05.11 1.00 WMD0001389 05.12 1.00 WMD0001390 05.13 1.00 WMD0001391 05.14 1.00 WMD0001392 05.15 1.00 WMD0001393 05.16 1.50 WMD0001394 05.17 1.00 WMD0001395 05.18 1.50 WMD0001396 05.19 1.00 WMD0001397 05.20 2.50 WMD0001399 05.21 1.50 WMD0001055 30.00 30.00
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TEST CROSS REFERENCE PAGE 1 QUESTION VALUE REFERENCE 06.01 1.00 BSR0000122 06.02 1.00 DSR0000126 06.03 1.00 BSR0000132 06.04 1.00 BSROOOO120 06.05 1.00 BSR0000141 06.06 1.00 BSR0000131 06.07 . 50 BSR0000123 06.08 1.50 BSR0000134 06.09 1.50 BSROOOO124 06.10 1.50 BSROOOO127 06.11 2.00 BSR0000119 06.12 1.50 BSR0000125 06.13 2.00 BSR0000120 06.14 1.50 BSROOOO130 06.15 1.00 DSR0000136 06.16 1.50 BSR0000140 06.17 1.50 BSR0000121 06.19 1.00 DSR0000129 06.19 2.00 DSR0000133 06.20 1.25 BSROOOO135 06.21 1.00 ESR0000137 06.22 1.75 BSR0000139 06.23 1.00 DSR0000139 30.00 M4MMwh 30.00 t
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TEST CROSS REFERENCE PAGE 1 9UESTION .VALUE REFERENCE 07.01 1.00 KTM0003178 07.02 1.00 KTM0003230 07.03 1.00 KTM0003176 07.04 1.00 KTM0003213 07.05 1.00 KTM0003218 07.06 1.00 KTM0003223 07.07 1.00 KTM0003229 07.08 1.00 KTM0003226 07.09 1.50 KTM0003215 07.10 1.50 KTM0003217 07.11 2.00 KTM0003224 07.12 1.00 KTM0003240 07.13 2.50 KTM0003222 07.14 1.00 KTM0003221 07.15 .75 KTM000323G 07.16 1.00 KTM0003207 07.17 1.00 KTM0003212 07.18 1.00 KTM0003214 07.19 1.00 KTM0003220 07.20 2.50 KTM0003233 07.21 3.00 KTM0003236 07.22 .50 KTM0003237 07.23 .75 KTMOOO3216 07.24 1.00 KTM0003231 30.00 hb-h-- 30.00
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TEST CROSS REFERENCE PAGE 1 QUESTION VALUE REFERENCE 08.01 1.00 WMDOOOO423 08.02 1.00 WMDOOO1002 08.03 1.00 WMDOOO1320 08.04 1.00 WMDOOO1321 08.05 1.00 WMDOOO1336 08.06 1.00 WMDOOO1344 l 08.07 1.00 WMDOOO1346 ! 08.08 1.50 WMDOOO1007 1 08.09 1.00 WMDOOO1332 08.10 1.00 WMDOOO1337 08.11 1.00 WMDOOO1351 08.12 1.50 WMDOOOO981 08.13 1.00 WMDOOO1005 ! 08.14 1.50 WMDOOO1009 l 08.15 1.50 WMDOOO1012 08.16 1.50 WMDOOO1330 08.17 2.00 WMDOOO1341 i 08.18 1.50 WMDOOO1342
- 08. Ir? 2.00 WMDOOOO976 08.20 1.00 WMDOOO1328 08.21 1.00 WMDOOO1333 08.22 1.00 WMDOOO1339 08.23 1.00 WMDOOO1345 08.24 1.00 WMDOOO1349 08.25 1.00 WMDOOO1350 30.00
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e U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: SURRY 1&2 REACTOR TYPE: PWR-WEC3 DATE ADMINISTERED: 87/02/09 EXAMINER: __DE6N. WM CANDIDATE: INSTRUCTIONS TO CANDIDATE: Use separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examinatton starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCOBE..__ VALUE CATEGORY 30.00 .25.00
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, gc773' HEAT TRANSFER AND FLUID FLOW 30-CCI _25.09 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 140-ftr _25.00 3. INSTRUMENTS AND CONTROLS
_20.00__ 25.00 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL e CONTROL [ 7.* 7L M % Totals Final Grade All work done on this examination is my own. I have neither given nor received aid. Candidate's Signature 1 l
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o NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
- 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil only to facilitate legible reproductions.
- 4. Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer sheet.
- 9. Number each answer as to category and number, for example, 1.4, 6.3.
- 10. Skip at least three lines between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.
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- 18. 'When you complete your examination, you shall:
- a. Assemble your examination as follows:
(1) Exam questions on top. (2) Exam aids - figures, tables, etc. (3) Answer pages including figures which are part of the answer.
- b. Turn in your copy of the examination and all pages used to answer the examination questions,
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions,
- d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
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( 1s__E81NCIE6ES_QE_NUC6E88_EQWEB_E68NI_gEEB8IlgN 2 PAGE 2 THERMODYNAMICS,_HE8I_IB8NSEEB_8NQ_E6Ulp_ELgW o e QUESTION 1.01 (2.00) Given two pumps of equivalent design, operating at the same, constant speed: A. What will be the effect of placing the two pumps in series (with respect to flow and head)? B. What will be the effect of placing the two pumps in parallel (with respect to flow and head)? QUESTION 1.02 (1.00) Given: Three reactor coolant (RCP) pumps operating in parallel, each with a flow rate "m" and a combined flow rate "M". Out of the four possibilities below, choose the one that best fits if one RCP is secured.
- a. The resulting core flow (M) will increase.
- b. The resulting core flow (M) will increase along with individual operating RCP flow (m).
- c. The resulting core flow (M) will decrease as individual operating RCP flow (m) increases.
- d. The resulting core flow (M) will not change due to decrease in RCP back pressure.
QUESTION 1.03 (1.00) What is the design basis of having a DNBR > or = to 1.3? DUESTION 1.04 (2.00) List the four (4) plant parameters observed to insure that CHF or DNBR are not exceeded. QUESTION 1.05 (2.00) What are all the conditions that must be present in order for natural circulation to e::ist? I (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****) I
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1.__EBINCIELES_gE_ NUCLE 88_EgWEB_ELANI_QEEBBIlgN 2 PAGE 3 THERMODYNAMICS2 _ HEAT _TRANSEER_AND_ELUID_ELgW t QUESTION 1.06 (1.00) With respect to reactor thermal limits which of the following statements is NOT correct.
- a. The ratio of the peak linear power density to the average linear power density in the core at a particular elevation is called the nuclear heat flux hot channel factor.
- b. The average linear power density in the core is expressed l in units of kw/ft and is the total thermal power divided by l the active length of all the fuel rods.
- c. The purpose of limiting the enthalpy rise hot channel factor is to prevent bulk boiling from taking place during normal operations.
- d. The rod bow penalty (RBP) accounts for the bowing of fuel rods as their burnup increases.
- e. The purpose of the limit on the heat flux hot channel factor is to insure that fuel clad temperature does not exceed 2200 deg F during norn.a1 operations.
l QUESTION 1.07 ( .50) Consider a fuel pellet at 70 deg F. A 6.7ev neutron coming in will be absorbed. The 6.7ev neutron will be absorbed in the outer part of the fuel. The inner fuel will not even see the neutron (low flux). This phenomenon is called QUESTION 1.08 (1.50) Write on your answer sheet INCREASES , DECREASES or DOES NOT CHANGE for the following: The magnitude of the fuel temperature coefficient (FTC): l A. INCREASES / DECREASES / DOES NOT CHANGE with increase in I power. B. INCREASES / DECREASES / DOES NOT CHANGE with core age. { C. INCREASES / DECREASES / DOES NOT CHANGE with decrease in moderator temperature coefficient (MTC). (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****) r
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It__EBIUCIE6ES_QE_UUCLE88_EQWEB_E68NI_QEEB8I1QN 2 PAGE 4 THERMODYNAMICS3 _ HEST _TBANSEEB_8NQ_E6UIQ_ELQW I QUESTION 1.09 (1.00) The negative reactivity added when fuel temperature increases is primarily caused by ______.
- a. self shielding of the fuel
- b. doppler broadening c..an increase in the GAMMA heating contribution
- d. fuel pellet swell thus decreasing the gap OUESTION 1.10 (1.00)
Which one of the following statements is correct? At the beginning of the Xe transient on a power decrease following 100 hours at 100% power: note: EXe] denotes xenon concentration
- a. Direct EXe] increases, indirect EXe] decreases, total EXe] decreases,
- b. Direct EXe3 increases, indirect EXe] increases, total EXe] increases.
- c. Direct EXe] decreases, indirect EXe] decreases, total
[Xe3 decreases.
- d. Direct EXe] decreases, indirect EXe] increases, total EXe] increases.
- e. Direct EXe3 decreases, indirect EXe3 increases, total EXe] decreases.
QUESTION 1.11 (1.00) GIVEN: Two identical control rods, each absorb an equal amount of neutrons. The neutron flux at the center of the core equals that at the edge of the core. Why do the control rods in the middle of the core (radi al l y) have a greater effect on Keff than the control rods at the edge of the core (radially). (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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1.__EBINCIELES_gE_ NUCLE 88_EQWEB_ELONI_CEEB8Ilgd2 PAGE 5' THERMODYNAMICS2 _IjgAT_TRANSEEB_8ND_ELUIp_ELOW 9 e QUESTION 1.12 (1.00) What effect does rod shadowing have on the worth of control rods? OUESTION 1.13 (1.50) On a reactor startup, what 3 conditions indicate the reactor is critical? QUESTION 1.14 (1.00) Give two reasons why 10 exp -8 amps is chosen as a standard reference for critical rod height data. note: " standard reference" is NOT an acceptable answer (*+*** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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.t.__EBINQIE6ES_QE_ NUCLE 88_EQWEB_E68NI_QEEB81IQN 1 PAGE 6 THERMODYNAMICS 1 _HE81_IB8NSEEB_8ND_ELUID_E6QW QUESTION 1.15 (2.00)
Match the term in column A with the correct definition in column B. column _A column _B a) Specific Entropy 1) BTU /deg F b) DNDR 2) Ratio of local Q to to CHF 051.30 c) Quality 3) Internal energy of a substance
- d. Enthalpy 4) % steam mass to total steam
& water mass
- 5) BTU / lbm-deg R
- 6) Ratio of critical O to local O
- 7) Internal Energy plus Flow Energy of a substance B) % steam volume to total steam and water volume QUESTION 1.16 (1.00)
What effect does adding an 800 ci source yielding 1 X 10 exp 8 neutrons /sec have on the magnitude of Keff in a subcritical reactor? NOTE: For simplicity assume the microscopic TOTAL cross st -t i on of the source equals zero.
- a. increase
- b. decrease
- c. no change
- d. insufficient data
( **
- x-* CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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1x__EBINCIE6ES_gE_NUQ6E88_EgyEB_E68NI_QEgB8IlgN 2 PAGE 7i IUEBUgpYN8dlCS,_Ug8I_IB8NSEgB_8NQ_E6UID_E6QW e QUESTION 1.17 (1.00) Given a SUR of 0.1 dpm, determine the final power P in terms of the initial power Po after O.1 hr. Show all work. QUESTION 1.18 (1.00) Choose the best answer for the definition of subcritical multiplication.
- a. The process of utilizing source neutrons to sustain the chain reaction for Keff < 1.
- b. The phenomenon where by source neutrons are used to measure the f racti onal curvature change of the flux for Keff < 1
- c. The manipulation of neutron sources to sustain the chain reaction until Keff = 1.
- d. The phenomenon where by source neutrons are used to stabilize reactor period /startup rate thus ensuring reactivity (rho) is<< Beff for Keff < 1.
QUESTION 1.19 (1.50) Calculate the heat transferred across one U-tube of a steam generator. Show all work. GIVEN: (for simplicity) U-tube heat transfer coefficient: 1.565 BTU /(sq ft-deg F) U-tube height: 25 ft U-tube outer radius: 1/2 inch primary coolant temperature: 550 deg F secondary water temperature: 480 deg F QUESTION 1.20 (1.50) List three things, that in practice, prevent water hammers from occurring (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****) L- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
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I 1.__EBINCIE65S_gE_NUC6E88_EgNgB_E68NI_gEEB8IlgN2 PAGE 8 IHEBMgDYU6MICS2 _HE81_IB8NSEEB_8ND_E6UID_E6gW 4 i QUESTION 1.21 (2.00) 1 If steam goes through a throttling process, i ndicate whether the following parameters will INCREASE, DECREASE, or REMAIN I THE SAME.
- a. Enthalpy .
- l i
- b. Pressure I
\ l i c. Entropy I d. Temperature ! QUESTION 1.22 (1.50) : I ! A motor driven centrifugal pump is operating at a low flow condition. You then start opening the throttle valve on the discharge side. How will each of the following be affected? (INCREASE, DECREASE, or NO CHANGE) I a. Discharge Pressure
- b. Available NPSH
- c. Motor Amps 4
QUESTION 1.23 (1.00) Unit A is at EOL while Unit B has just been started up after a refueling. Assuming a rod speed of 48 spm, both reactors are taken critical by aulling l 50 steps at a time, waiting until counts stabilize then pulling agai n. i Assuming all systems and parameters are identical at the commencement of I the startup, and both units are initially shutdown by 27. (delta k/k): a) Which Unit wil.1 have the highest source range counts'when criticality is reached? b) How will critical rod heights compare in the two Units? (***** END OF CATEGORY 01 *****) t
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2s__P68NI_DEgl@N_INC6UQ1N@_@@ Eely _8ND_EDE6@ENCy_@y@ LEU @ PAGE 2' I OUESTION 2.01 (1.00) 1 1 l Which one of the following is NOT a source of water to the PRT7 a) CVCS Regenerative Heat exchanger relief valve b) RCP Seal Water return line relief valve c) Reactor Vessel Flange Leakoff Detection Drain d) RHR removal system safety relief valve QUESTION 2.02 (1.00) Which one statement below regarding the Source Range Nuclear Instrumentation System is INCORRECT. a) P-6 allows the source range high l evel reactor trip signal to be bypassed manually when one of the two intermediate range instruments is above 10 E-lO ion chamber amps. b) Placing BOTH source range blocking switches to the BLOCK l position de-energizes the high voltage supply to both source range instruments. c) The source range high level trip in blocked when P-lO is present. d) When P-6 is present and P-lO is not present, the source range high level trip is automatically reinstated and the source range high voltage re-energized when one of the two intermediate ranges is below P-6 reset. (***** CATEGORY O2 CONTINUED ON NEXT PAGE *****) L .. .
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2.__P68MI_DE@lGN_ INCLUDING _@8Egly_98D_EDE8GENCy_SYSIEUS PAGE 3 o QUESTION 2.03 (1.00) Which one of the following describes the method of NaOH solution addition to the Containment Spray System? a) An eductor utilizing OS pump discharge draws NaOH solution from the Chemical Addition Tank (CAT) into the QS pump output. I b) Balanced Gravity feed from the CAT to the RWST near where the CS pumps take a suction. c) Balanced Gravity feed from the CAT to the area between the CS pump inlet isolation valve and the suction side of the pump, d) The CAT pump discahrges the contents of the tank into the QS pump suction with a pre-determined flow rate set by a manual throttle valve. I OUESTION 2.04 (1.00) Which location below is the discharge point for the pressurizer head vent? a) Containment refueling cavity b) Upper region of containment below quench spray rings c) Pressurizer Relief Tank d) Suction side of containment Hydrogen Recombiners QUESTION 2.05 (1.00) i 1 Which valve listed below is used to throttle auxiliary spray flow? a) FCV-122 (Charging Flow Control Valve) b) HCV-311 (Aux Spray Valve ) c) PCV-455B (Loop C Spray Valve) d) PCV-455A (Loop A Spray Valve) e) You cannot throttle auxiliary spray (***** CATEGORY O2 CONTINUED ON NEXT PAGE *****) t
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2 t__PL@NI_DE@lGU_INCLUDIN@_g8 Eely _9ND_EMEBGENCy_SYSIEME PAGE 4 QUESTION 2.06 (1.50) Match the following Pressurizer heater banks in Column A with their proper MCC in Column B. COLUMN A COLUMN B A) Back-up Heaters
- 1) Group I (0.25) a) 1A1
- 2) Group II (0.25) b) 181
- 3) Group IV (0.25) c) 1C1
- 4) Group V (0.25) d) 1D1 B) Control bank heaters Group III (0.25) e) 1G f) 1H g) IJ QUESTION 2.07 (1.00)
A "High CLS" Automatic Safety Injection signal will: (Choose one) a) cause a main steam line isolation. b) be initiated by 2/4 containment pressure instruments greater than 17 psig. c) be blocked whenever the reactor trip breakers are open. d) cause a feedwater isolation and a phase "A" isolation. (***** CATEGORY O2 CONTINUED ON NEXT PAGE *****)
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2 1__ELONI_DE@l@N_lNCLUDIN@_S@EEIy_6NQ_EME8@ENCy_Sy@IEMS PACE 5 , 4 QUESTION 2.08 (1.00) Which of the following does the operator MANUALLY adjust to reduce the RCS temperature when the RHR system is in service for a normal plant cooldown, per OP 14.l? a) Throttle open CCW from RHR Heat Exchanger outlet isolation valve, b) Throttle open RHR Heat Exchanger outlet isolation valve. c) Throttle closed RHR Heat Exchanger bypass valve, d) Throttle closed RHR mini-flow recirculation valve. QUESTION 2.09 g h2 Listed below are valves associated with the Recirculation Spray (RS) System. Indicate whether each of the valves listed are NORMALLY OPEN or CLOSED. a) MOV-SW-lO64 and B (Service Water supply header x-connects) b) MOV-SW-lO5A and B (Service Water B return header isolation valves) 6-- nuv-Mb-iviA ( C o ui . ,3
, , _a , , a Cocling Pump ^ FIRET ui uuhur yu / g;j d) MOV-RS-155B (Outside RS Pump suction valve)
(***** CATEGORY O2 CONTINUED ON NEXT PAGE *****)
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2 1__ PLANI _ DESIGN _lNCLUDING_SBEEIY_8ND_EdEBGENCY_SYEIEUS PAGE 6 QUESTION 2.10 (1.00) In regards to the Chemical and Volume Control System (CVCS), state what position (OPEN, 3LOSED, AS IS) the following valves fail upon a loss of air. a) Letdown i sol ati orn valves LCV-1460 A/B ' b) Orifice Isol ation' valves LCV 12OO'A/B/C c) i Pure Grade wated supgly valve FCV-1114A d) Boric Acid supply to blender FCV-1113A e) Emergency Borate valve MOV 1350 L i OUESTION 2.11 (1.50) Indicate whether the following statements regarding PCP seals are TRUE or FALSE. . ( a) The floating ring seal, will limit leakage to 50 gpm if the #1 seal fails, b) #3 seal is designed to withstand full RCS pressure.
?
c) Seal water injection from CVCS enttirs the RCP between the seal package and the pump radid1. bearing. t QUESTION 2.12 ( .50) TRUE/ FALSE; An urgent failure slarm could indicate that a slave cycler failure has ocr a ed in the logic cabinet. QUESTION 2.13 (1.50) The principle driving force for PZR normal spray flow is the differential pressure between _________and j the_________. i l
.t 4-
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l Ez__PL@NI_DgglGN_ lng 6UQ1Ng_S@EEIY_@NQ_EdgBGENgy_Sy@IgdS PAGE 7 QUESTION 2.14 (2.00) List the 4 flow paths within the reactor vessel which BYPASS the fuel rods. QUESTION 2.15 (2.00) LIST 4 of the 5 Design bases for the ECCS Cooling Performance following a LOCA as stated in 10CFR50.46. QUESTION 2.16 (2.00) List 5 parameters associated with the RCP's which are monitored after starting a RCP as stated in OP 5.2 "RCP Operation". Provide the required minimum values which must exist if applicable. QUESTION 2.17 (1.75) Aside from the manual trip and overcurrent/ ground trip, LIST 5 Main Feed Water pump trips. (Include coincidence, time delays and setpoints where applicable) QUESTION 2.18 (1.25) According to the plant emergency response procedures, LIST the 5 preferred sources of auxiliary feedwater in proper sequence. QUESTION 2.19 (2.00) a) List two emergency loads supplied by the Service Water System, b) List four of the normal loads supplied by the Service Water System. I (***** CATEGORY O2 CONTINUED ON NEXT PAGE *****)
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 8I a
. . 1 4
QUESTION 2.20 (1.00) Where is the source of power for the automatic field flash of the Emergency Diesel Generators generated ? QUESTION 2.21 (1.00) Sodium Hydorxide (NaOH) added during the injection phase after a LOCA will eventually be distributed by the ContainmentSpray System and raises the Containment sump pH approximately B. What are the two (2) reasons for establishing the elevated pH in the containment? QUESTION 2.22 (1.50) State 3 reasons for having HCV-ll42 (RHR letdown penetration from the RHR heat exchangers) kept about 157 open. QUESTION 2.23 (1.50) Concerning the Rod Control System: a) Place the following components in their proper flow path order. Start from the normal power supply and ending at the CRDM's
- 1) DC hold cabinet
- 2) Power Cabinet
- 3) Motor generator set
- 4) Reactor Trip breaker
- 5) Automatic Rod Control Unit
- 6) Rod Position Indication Cabinet
- 7) Logic Cabinet b) For the components in Part a), above, STATE the number of each present in the system.
- (***** END OF CATEGORY O2 *****)
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3.__INSIBUMENIg_6ND_CQNIBQLS PAGE 2 l
. l f OUESTION 3.01 (1.00)
Which of the following is NOT a function of the P-4 permissive (trip and bypass breakers open)? ! a) Allows bypassing a steam dump cooldown interlock, b) Causes feed reg. valve to shut for associated S/G with a Hi-Hi level. c) Causes feedwater isolation if low Tavg is also present. d) Causes a turbine trip. QUESTION 3.02 (1.00) Concerning the Overtemperature Delta Temperature Setpoint (OTSP) describe how (increases, decreases or remains the same) each of the following parameter changes will effect the OTSP. a) Increase in Tavg b) Decrease in Reactor Pressure OUESTION 3.03 (1.00) Which statement below regarding the Main Generator . Protection System is INCORRECT. 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. l 1 I (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
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3.__INSIBUMENIS_8ND_CONIBOLS PAGE 3 QUESTION 3.04 (1.00) Which of the following would be the INITIAL response of the feedwater flow due to the response of the S/G Water Level Control System if the steam pressure transmitter controlling the SGWLCS failed HIGH while at 50% power? a) Feed flow would INCREASE due to the maximum steam I pressure input to the steam flow signal. b) Feed flow would INCREASE due to the level mismatch error between actual and programmed level caused by the pressure instrument failure. c) Feed flow would DECREASE due to the mismatch between steam and feed flow signals caused by the pressure instrument failure. d) Feed flow would remain THE SAME due to the dominance of the level error signal over the flow error signal. e) Feed flow would remain THE SAME as steam pressure will not affect the steam flow signal. QUESTION 3.05 (l.00) With the pressurl er level control selector switch in position I/II, a failure causes the following plant events. (Assume no operator actions taken.)
- 1) Charging flow reduced to minimum
- 2) Pressurizer level decreases
- 3) Letdown secured and heaters off
- 4) Level increases until high level trip Which one of tNe f ollowing failures occurred?
a) Level channel I failed high b) Level channel I failed low c) Level channel II failed high d) Level channel II failed low (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
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a i OUESTION 3.06 (1.00) J Which one of the f ollowing conditions is required for j automatic swapover of the LHSI pumps to the Recirculation ] Mode following a SI? 1-a) 2/4 RWST level at Lo level setpoint b) A'LHSI pump recirc isolation MOV closed for each pump 1 c) SI signal present d) SI Recirculation Mode signal present i t-4 OUESTION 3.07 (2.00) 4 i ' Match the trips in Column A, with the correct purpose of the
- trip in Column B. Some answers may be used more than once, r s
Column A Column B a) OP Delta T 1) DNB for slow transients b) Undervoltage trip 2) High kw/ft (subsequent fuel damage) i i- c) PR High Flux 3) Fuel pellet mn!+ ira t ] (High Stpt) i j 4) DNB
- 5) Uncontrolled power I excursion in power range j l i
i 1 ! OUESTION 3.08 ( .50) ! 'l TRUE or FALSE j Pulling the control power fuses when the Source Range level
! trip switch is in " Bypass" will cause a trip signal to j occur.
i i 1 J l i (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****) f, S i L-_
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PAGE 5 3.__INSIBUMEUIS_AND_CONIBOLS QUESTION 3.09 (2.00) a) What are all the conditions which will generated a # 3 EDG auto start signal? b) If the Unit #1 switch is in BYPASS and the Unit #2 l switch is in NORMAL, for 83 EDG control, what will happen to 5J3 and 25J3 if a loss of off-site power occurs? QUESTION 3.10 (2.00) List 5 protection logic signals generated by the Pressuriner Protection System. (Include in your answer set points, coincidence and associated interlocks, if any) QUESTION 3.11 (1.50) List 3 purposes of Rod Insertion Limits. QUESTION 3.12 (1.00) List the 4 requirements / control manipulations that will make up the logic to enanually close the Diesel Generator output breakers (15H3). l QUESTION 3.13 (1.50) List the 6 reactor trips which are enabled / blocked by the reactor trip system interlock P-7. QUESTION 3.14 (2.00) Describe the three different paths of power to the 120 VAC Vital Buc l-III, Identifying al1 components. Start with the source of power. (***** CATEGORY 03 CONTINUED ON NEXT PAGE $4***) i
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' PAGE 6 3 t__INSIBUdENIS_GND_COUIBOLS QUESTION 3.15 (1.00) List the TWO conditions that will provide signals to automatical1y open the ECCS accumulator discharge valves (1865 A/B/C). QUESTION 3.16 # bl0) List the automatic start signals for the "A" and "C" charging pumps. (assume "C" is on its normal power supply) OUESTION 3.17 (1.50) In regards to the PZR Pressure Control System, Unit 2 has an alarm, NDT REQUIRED. What are all the conditions that will CL4UGO this alarm to actuate? OLIESTION 3.18 (1.00) The Reactor breaker shunt trip coi1s have been modified to 4 also energizo upon any trip signal to the Undervoltage coils. What is the reason for this modification? QUESTION 3.19 (1.00) 4 The Detector Current Coniparator receives input from all 4 upper and lower power range detectors. How are these inputs j compared, and what conditions are needed to auto bypass j circuitry while at power? QUESTION 3.20 (1.75) u) What consequences could be enpected in the Rod Control l System's DC Hold Cabinet if 2 or more groups of rod I drive mechanisms were placed on hold power (excluding
- Control Bank. D rods)' E>
- p l a i n you reasoning. (1,0) b) Why in there both a 125 VDC and a 70 VDC power supply in the DC Hold Cabinet?
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
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31__INSIEUMENIS_END_CgNISOLS PAGE 7 1 QUESTION 3.21 (1.25) Describe how the High Steam Line Flow SI input varies and the parameter on which this program is based. OUESTION 3.22 (1.00) Describe the design feature on the Main Steam System that prevents reverse flow in the event of a Main Steam Line Break upstream of the Main Steam T.ip valve. Include in your answer how it is accomplished, i j QUESTION 3.23 (1.50)
} Sketch the rod speed program by indicating rod speed versus ; error signal.
1 4 1
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; QUESTION 4.01 (1.00) i On a loss of condenser vacuum where vacuum is greater than 20" Hg and decreasing, which of the following is NOT an immediate action?
1 j a. Place an additional set of air ejectors in operation i b. Start a hogger } c. Start an additional circ pump ) d. Reduce turbine load i e. Start an additional condensate pump, if available i 1 l, QUESTION 4.02 (1.00) l Prior to operating Reactor Coolant Pumps in accordance with OP-5.2, j i Reactor Coolant Pump Operations, the minimum seal flow should be ____ gpm and VCT pressure should be a minimum of ____ psig. .
- 1. O , 10 1
1 4 1
- 2. O.2 , 15 i 3. 2.0 , 30 1
l 4. 5.0 , 20 t j DUESTION 4.03 (1.00) Liut FOUR indications of one dropped rod at 75% power. 4 i QUESTION 4.04 (1.00) $ What operator actions are required upon evacuating the control room if the !. reactor could not be tripped before exiting the control room? l l i I i i (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
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PRgggpUBgS_ _NgBMAL x_ADNgBMAbt_EMgBGENgy_AND PAGE 3 dz__58DIRLgGlg86_ggNIBg6 QUESTION 4.05 (1.00) Which of the following describes a temporary change which alters the INTENT of a procedure? I ) a. A change that corrects an incurrect valve lineup.
- b. A change that modifies the criteria by which a system's operability is determined.
- c. A change that allows partial use of a procedure to test a subtrain without affecting remaining equipment in that train.
- d. A change that allows you to change incorrectly specified instruments for data taking.
QUESTION 4.06 (1.00) If you are in a 100 mrad / hour gamma field for 45 minutes, what is your dose in MREM after 45 minutes?
- a. 45
- b. 75
- c. 450
- d. 750
]
OUESTION 4.07 (1.00) If a " Rod Control Urgent Failure" alarm occurs due to a failure in the logic cabinet, the Tave/Truf mismatch is immediately maintained by which of the following?
- a. controlling turbine load.
b, taking manual control of individual control rod banks.
- c. taking manual control of individual control rod groups,
- d. baration and dilution of the reactor coolant system.
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
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,86DigtO@lG66_GQNIBQL QUESTION 4.08 (2.00)
Prior to a reactor startup, with the RCS at normal operating pressure and temperature, the f ollowing RCS leakages e::ist. For each leak listed below, indicate whether you could STARTUP or would have to remain SHUTDOWN. (Treat each leak below as an independen'. event) a) A leak from an unknown source of 1.5 GPM. b) 6.0 GPM from a manual valve packing gland. c) 0.4 GPM from one S/G. d) 0.1 GPH from the reactor vessel head INNER seal. QUESTION 4.09 (1.00) List all condi ti ons that require the Control Rod Drive Mechanism Shroud Cooling Fans to be in operation. QUESTION 4.10 (1.00) Anuwer the following TRUE or FALSE concerning verification of tagouts:
- 1. Closed valves shall be verified closed by cracking them open and reclosing
- 2. Indiredt verification methodu such an use of indicating lights, or other indication will only be used if ALARA concept makes verification impractical.
QUESTION 4.11 (1.50) Natch the action listed in Column A with the approuimate power level in Column D at which this action is taken on a unit startup to 100% power. COLUMN A COLUMN P
- a. Place a second Main Feed pump in service 1) 15%
- 2) 30%
- b. Gtop increasing power and check ior a 3) 50%
chemistry hold 4) 60% ,
- 5) 70% 1
- c. Perform a calorimetric 6) 90%
(***te CATEGORY 04 CONTINUED ON NEXT PAGE *****) t .
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. QUESTION 4.12 (2.50) 3 Match the terms in column A to the values in column B for the radiation j exposure guidelines. Assume whole body dose unless'otherwise stated.
CAUTION: Some answers could be used more than once. (0.5 ea)
] COLUMN A COLUMN B i
l
- a. NRC limits /qtr 1. 0.5 REM'
; b. Virginia Power limits /qtr 2. 1.25 REM f
l c. NRC pregnant woman limit / gestation 3. 1.0 REM 1
- d. NRC general public limit / year 4. 0.75 REM fl
- e. NRC quarterly limit with a Form 4 5. 5 REM
- 6. 3 REM i
I I I OUESTION 4.13 (1.00) ! List the 4 methods given in the S/G Tube Rupture EOP to identify which S/G 1 is ruptured. I j QUESTION 4.14 (1.00) j Following a valid reacto- 1 rip and safety injection, what are the Reactor l Coolant Pump Trip Criteria? (Assume normal containment conditions) QUESTION 4.15 (1.00) l List four of the critical conditions required to be recorded during a i startup when 1 X 10E-G amps iu attained. i l l QUESTION 4.16 (1.00) List the immediate opurator actions to initiate emergency boration if it is j required on an Anticipated Transient W1thout Trip condition. Assume , i Safety injection hau not actuated and is not desired. I \ i (***** CATEGORY 04 CONTINU'ID ON NEX T PAGE * * * * * )
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- 4. PROCEDURES - NORMAL _1 ABNORMAL z_ EMERGENCY AND PAGE 6
,88DIRLOGIGOL_ COL 4IBOL <
l 1 i QUESTION 4.17 (1.00) List the 4 DISTINCT hazards to which personnel are exposed when an entry l into the reactor compartment is made during reactor operations. ) i ] QUESTION 4.18 (2.50) ] List FIVE indi cations of a loss of Component Cooling Water in accordance j with AP-15, Less of Component Cooling. 1 'r i QUESTION 4.19 (3.00) j List ALL imediate actons required by 1-AP-42, Loss of Reactor Coolant
; System Pressure.
1 1 i i i
.i CUESTION 4.20 ( .50) .
f Unit 1 is operating at 30% power when a valid Reactor Coolant Pump
! vibration DANGER alarm is received for one pump. What are your immediato ! actions?
i OUESTION 4.21 (2.00) List ALL immediate operator actions for a PARTIAL loss of main feedwater. i 1 j QUESTION 4.22 (1.00) i l Briefly explain the effect that placing an " unsaturated" mixed bed i demineralizer in service will have on the reactor coolant system and f on control of the Reactor. l ( OUESTION 4.23 (1.00) During a natural circulation cooldown, it is desired to cooldown using the steam dumps. Which MODE la the steam dump system operated in and WHY7 (***** END OF CATEGORY 04 *****) (********$**** END OF EXAMINATION $$*************) , e n - n- .e-e ....,<n-,n -a .,n-- m-,,-,_,.,.ne---., _ wm m - n n .w ., , _ r w n, ,-we,,n,. ,.m,--e
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1.__EBINCIE6ES_gE_UUCLEOB_EgWEB_ELONI_gEgBOIIgg2 PAGE. 9 IUEBMgDyU6MICS 2 _UE81_IB60SEEB_6NQ_E691p_ELgW At/SWERS -- SURRY 1842 -87/02/09-DEAN, WM i f ANSWER 1.01 (2.00) A. It doubles (or increase) the head f or a a t von mass flow rate. B. It will double (or increase) the mass w rate capacity for a given' head. REFERENCE Surry lesson plan ND-83-LP-8, Rev 1,-p8.18 191004; K1.09/1.10(2.4/2.4) ANSWER 1.02 (1.00) C REFERENCE Surry lesson plan ND,-83-LP-8, Rev 1, 191004; K1.14(204) _, i ANSWER 1.03 (1.00) With a DNDR of 1.3, during normal operation and anticipated operational occurrences, there is a 95% confidence that DNB does not occur. When > 1.3 likelihood of DNB occurring decreases. RECERENCE - Surry lesson plan ND-86.3-LP-2, p2.10 193008; K1.10(2.9)
)
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it__EBINCIELES_QE_NQCLE88_EQWEB_ELONI_QEEB8IlgN 2 PAGE 10 ISEBMQDYN8MICS1 _UE8I_IB8NSEEB_8ND_E6QlD_ELQW ANSWERS -- SURRY 1&2 -87/02/09-DEAN, WM ANSWER 1.04 (2.00)
- 1. reactor power
- 2. coolant flow rate
- 3. RCS cn'" * - - = + ' " = 'Tc) 7MJ 4 RCS pressure.
REFERENCE Surry lesson plan ND-86.3-LP-2, p2.10 193008; K1.05(3.4) ANSWER 1.05 (2.00)
- 1. Density difference (or DELTA T) created by heat addition by the heat source and heat removal by the heat sink.
- 2. The heat sink must be elevated physically above the heat source.
REFERENCE Surry lesson plan ND-86.3-LP-4, p4.5 193008; K1.21(3.9) ANSWER 1.06 (1.00) I e REFERENCE Surry lesson plan ND-86.3-LP-3, pp3.4, 3.5, 3.7, 3.10, 3.12 193009; K1.05(3.1) ANSWER 1.07 ( .50) Self shielding / Self shielding of the fuel pellet. REFERENCE Surry lesson plan ND-86.2-LP-1, p1.7 192001; K1.08(2.3)
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1.__E81NCIELES_QE_NQCLE88_EQWEB_E68NI_QEEB8IlON 1 PAGE 11 < IUEBdQDYN801CS1 _UE8I_IB80SEEB_8UD_E6QlD_E6QW ANbWERS--SURRY 1&2 -87/02/09-DEAN, WM ANSWER 1.08 (1.50)
- a. DECREASES
- b. J' CCREiiSF G - /dLTF M c"5
- c. DOES NOT CHANGE REFERENCE Surry lesson plan ND-86.2-LP-1, pl.4, 1.11 192004; K1.07(2.9)
ANSWER 1.09 (1.00) b REFERENCE Surry lesson plan ND-86.2-LP-1, p1.16 192004; K1.05(2.3) ANSWER 1.10 (1.00) d REFERENCE Surry lesson plan ND-86.2-LP-4, p4.4, 4.8. 192006; K1.06(3.4) ANSWER 1.11 (1.00) Neutrons at or near the edge of the core have a higher probability of leaking out than the ones at the center which have a higher probability of causing fission. (Hence: DRW at center is > than at edge). REFERENCE Surry lesson plan ND-86.2-LP-6, p6.12. 192005; K1.14(3.2) l L
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1.__EGINCIELES_gE_NQCLE88_EgWEB_ELONI_QEEB811gN2 PAGE 12 IMEBMQDYNOMICS 1 _UE8I_IBONSEEB_8ND_ELylD_ELQW ANSWERS -- SURRY 1842 -87/02/09-DEAN, WM ANSWER 1.12 (1.00) The presence of adjacent control rods may cause a significant change in an individual control rod worth. REFERENCE Surry lesson plan ND-86.2-LP-6, p6.19 001/000; K5.05(3.5) ANSWER 1.13 (1.50) - y st0 4Mfutml Start up rate is positive and constant,(reactor power is b ut accredo ble increasing and there is no outward rod motion. 44 d ecm foxe REFERENCE Surry lesson plan ND-86.2-LP-7, p7.51. 192OOS;K1.11(3.8) ANSWER 1.14 (1.00) (two of the three answers below required)
- 1. Neutron production is relatively high, so power is constant when the reactor is critical.
- 2. Below 10 e::p-8 amps the output of the intermediate range may not be directly proportional to the neutron population.
- 3. Reactivity has not yet been changed by the moderator or fuel temperature.
REFERENCE Surry lesson plan ND-86.2-LP-7, p7.57 192008; K1.12(3.5) l
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It__EBINGlELES_QE_Nyg6EGB_EQWEB_E68NI_gEEB811gN 1 PAGE 13 IUEBdQDYN8 digs 1_UE61_IB@NSEEB_6ND_E(QlD_E(QW At4bWERS -- SURRY 1&2 -87/02/09-DEAN, WM ANSWER 1.15 (2.00) a) 5 b) 6 c) 4 d) 7 REFERENCE Surry lesson plan ND-83-LP-(1-10) 193008; K1.10/K1.06(2.9/2.8) ANSWER 1.16 (1.00) C REFERENCE Surry lesson plan ND-86.1-LP-6, p6.35 000/015; K5.06(3.4) ANSWER 1.17 (1.00) 0.1hr = 6 min P = Po 10 exp SUR(t) (+.5) P = Po 10 exp O.1 dpm(6 min) (+.25) P = Po 10 exp O.6 (+.25) P = 3.98 Po REFERENCE Surry lesson plan ND-86.1-LP-8, p8.12 192003; K1.09(2.3) ANSWER 1.18 (1.00) a OF C. REFERENCE Glasstone & Sesonske. Nuclear reactor engineering third ed. New York: Van Nostrand Reinhold Co., 1981.
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is__EBINCIELES_QE_ NUCLE 6B_EQWEB_E66NI_QEEBBIlgN 2 PAGE 14 ISESMQDYNGMICS1 _dE81_IBONSEEB_8ND_E6UID_E6QW A,NhWERS--SURRY 1&2 -87/02/09-DEAN, WM 192003; K1.01(2.7) ANSWER 1.19 (1.50) O = UA DELTA T (+.5) A = 25' (2 pi r) (+.25) DELTA T = 70 deg F (+.25) 0.5" = .042' (+.25) O= 1.565 BTU /(sq ft-deg F) x 25ft x 2 pi r :: 70 deg F (+.25) O = 723 BTU REFERENCE Surry lesson plan ND-83-LP-1, p1.27 193007; K1.08(3.1) ANSWER 1.20 (1.50) [ 5~ e ct [or aug 3
- 1. Gradual warm up of steam lines
- 2. Proper venting of tanks and components during warm up and operation.
- 3. Steam traps
- 4. Lines kept full f, 3 6&E9 drackgr9( vet /K hp4rt S/o pomp (Others as appropriate) *
* '7 '^' " 'E'] 0"# 40 4 IY P/ v*
REFERENCE Surry lesson plan ND-83-LP-8, p8-36 193006; K1.04/1.10(3.4/3.3) ANSWER 1.21 (2.00)
- a. REMAIN THE SAME
. b. DECREASE
- c. INCREASE
- d. DECREASE REFERENCE Surry lesson plan ND-83-LP-(1-10), steam tables 193004; K1.15(2.8)
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Iz__EBINCIELES_QE_NQCLE88_EgWEB_EL6NI_QEEB6IlgN S PAGE 15 IHEBdQDYNOMICSg_UE8I_IBONSEEB_8ND_E(UID_E6QW A,NSWERS -- SURRY 1&2 -87/02/09-DEAN, WM ANSWER 1.22 (1.50)
- a. Decrease
- b. Decrease
- c. Increase REFERENCE Surry lesson plan ND-83-LP-8 191004; K1.15(2.6)
ANSWER 1.23 (1.00) a) Will be the same (+.5 ea) b) Unit B will be higher REFERENCE Westinghouse Reactor Core Control, pp 6-23/26 Westinghouse Fundamentals of Nuclear Reactor Theory, pp 8-48/60 001/010; K5.08(2.9/3.2) & 001/000: K1.05(4.5/4.4) I i
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,' ANSWERS -- SURRY 1&2 -87/02/09-DEAN, WM ANSWER 2.01 (1.00) c (1.0)
REFERENCE I NA NCORDP 88.1 Reactor Coolant p 2.24 007/000; A3.01 (2.7/2.9) ANSWER 2.02 (1.00) d (1.0) REFERENCE ND 93.2-LP-2 SR NIS 015/000; K4.01 (3.1/3.3) ANSWER 2.03 (1.00) b 1.00 REFERENCE ND91-LP-5-H-5.1 026/000; K4.01 (4.2/4.3) ANSWER 2.04 (1.00) a aaaaaaaaaaaaaaaa REFERENCE ND 88.1 H-3.2 ND 88.1-LP-3 Learning Objective C 002/000; K4.03 (2.9/3.2) l ANSWER 2.05 (1.00) I a (1,o) REFERENCE t ND 88.1-LP-3 p3.13 010/000; A4.01 (3.7/3.5)
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,' ANSWERS -- SURRY 1&2 -87/02/09-DEAN, WM ANSWER 2.06 (1.50)
A) 1) 1J Jat($ ) (0.25)
- 2) 1B1 (b) (0.25) 3)
plCl W(N c) (O.25)
- 4) 1ETid fe7( (0.25)
B) 1A1 (0.25) REFERENCE ND 88.1-LP-3 PZR & Press. Relief 010/000 K2.01 (3.0/3.4) ANSWER 2.07 (1.00) d REFERENCE ND 91.0 ESF 013/000 K1.01 (4.2/4.4) ANSWER 2.08 (1.00) b bbbbbbbbbbbbbbbb REFERENCE ND 88.2-LP-2 RHR System 1-OP-14.1 ( .3 ANSWER 2.09 ' '. 0 0 ? - a) Open (0.25 ea) b) Open _' C; -- Of l_ & d) Open REFERENCE ND 91-LP-6 Retirc Spray 026/000; K1.02 (4.1/4.1) L s.
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, ANSWERS -- SURRY IL2 -87/02/09-DEAN, WM ANSWER 2.10 (1.00) a) Fails closed (0.2 ea) b) Fails closed
! c) Fails closed d) Fails open e) Fails as is REFERENCE ND 88.3-LP-9 004/000; A2.04 (3.6/4.2) ANSWER 2.11 (1.50) a) True (0.5 ea) b) False c) False REFERENCE ND 88.1-LP-6 p6.13 002/000; K1.13 (4.1/4.2) ANSWER 2.12 ( .50) TRUE (0.5) REFERENCE ND 93.3-LP-3 Rod Control System 001/000; K4.03 (3.5/3.8) ANSWER 2.13 (1.50) RCP discharge and the PZR (0.75 ea) REFERENCE ND-88.1-LP-3 002/000; K1.09 (4.1/4.1) i 4
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, ANSWERS -- SURRY IL2 -87/02/09-DEAN, WM ANSWER 2.14 (2.00)
- 1) To upper head plenum via nozzles in core barrel flange.
(0.5)
- 2) Between hot leg discharge nozzles and upper core barrel outlets. (0.5)
- 3) Between baffle plates and core barrel. (0.5)
- 4) Around inserts in guide thimble tubes in the fuel assemblies. (0,5)
REFERENCE ND 88.1 LP-2 p. 2.32 002/000; A1.05 (3.4/3.7) ANSWER 2.15 (2.00) (any 4 of 5 at O.5 ea)
- 1) Max. Fuel Element Cladding Temp. < 2200 Deg. F
- 2) Cladding Oxidation < 17% thickness
- 3) Hydrogen generated by Zirc-Water reaction <1% of max.
possible.
- 4) Core remains in a coolable geometry
- 5) Provides for long term decay heat removal REFERENCE 10CFR50.46 ND 91-LP-2 p2.5 006/050; PWG 4 (4.2/4.3)
't ANSWER 2.16 (2.00)
- 1) Motor current (0.25 ea)
- 2) Bearing temperatures
- 3) Seal injection flow, 6-9 gpm
- 4) Seal leak off flow, 0.2-5 gpm
- 5) Seal differential pressure > or = 200 psi d REFERENCE ND 88.1-LP-6 RCP OP 5.2 p. 8/9 003/000; PWG-7 (3.5/3.9)
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2 __PL@NI_DE@lGN_fNCLUDING_@@ eel 1_@ND_ EMERGENCY _SYSIEMS PAGE 20 AN9WERS -- SURRY 1&2 -87/02/09-DEAN, WM ANSWER 2.17 (1.75)
- 1) Lube oil pressure (0.2), < 5 psig (0.1)
- 2) SI (Train A or B) (0.2),
- 3) Low suction header pressure (0.2), <58 psig (0.1)
"A" pump both have a 15 sec TD ( 0.1 )
"B" pump
- 4) <2800 gpm T;ci e flow (0.2), recirc. valve not open 15 set TD (O.1) MS
- 5) Bus undervoltage (0.2)
- 6) S/G Hi-Hi Level (0.2), 2/3 channels on 1/3 S/G's > 75%
(0.25) REFERENCE ND 89.3-LP-3 Main Feedwater system p. 3.12/13 059/000; K4.16 (3.1/3.2) ANSWER 2.18 (1.25) a) Vertical emergency condensate tank (CN-TK-1A) (0.25 ea) b) Horizontal emergency makeup tank via booster pumps c) Normal Condensate Storage Tank makeup (CN-TK-2) d) Unaffected units Aux. feedwater supply e) Fire water makeup supply REFERENCE ND 89.3-LP-4 Aux Feed Sys. p 4.13 061/000; K4.01 (3.9/4.2) ANSWER 2.19 (2.00) a) 1. Recirculating Spray (0.5E_cL40F4*/ 2)
- 2. Gharging pump seals and oil coolers 44rTT 3, fontrol Roon Cb liffi )
b) 1. Bearing Cooling (0.25 ea)
- 2. Component Cooling
- 3. Chilled water
- 4. Station vacuum priming.
REFERENCE l ND 89.5-LP-2 Service Water System p 2.4/2.5 l 076/000; K1.19 (3.6/3.7)
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2.__P6@NI_DSSIGN_lNCLUDING_@@EEIY_GND_@dEBQENCY_SYSIEUS PAGE 21 , ANSWERS -- SURRY 1&2 -87/02/09-DEAN, WM ANSWER 2.20 (1.00) From its own (0.5) 125 VDC Distribution system (0.5) REFERENCE ND 90.3-LP-2 064/000; K1.04 (3.6/3.9) ANSWER 2.21 (1.60)
- 1) To ioac i g n a d hs-Cont 6 m
- 2) Helps Chloride stress corrosion in the sump by kr eping the ultimate pH of the sump water slightly basic containment sump. (O. 5' REFERENCE (/ 0)
ND 91-LP-5 Containment Spray p5.13 026/000; K4.02 (3.1/3.6) ANSWER 2.22 (1.50) To provide a path to keep the RHR system full (0.50) and to allow for expansion of the system during heat up of the RCS (0.5) and thus ambiently heating up RHR (O.50). REFERENCE d"J G CC @ Ct+tJ w ( f [J . ND 88.2-LP-2 p2.8 004/000; K1.01 (3.4/3.9) f**d M PI#UU"Mo by predaqa/ph/g -[Q pf4(g]
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2 __E6@yI_QEQ1@y_1gg(UDINQ_@@E@IY_@ND_EDEBQEUQY_SY@IEd@ PAGE 22 ,' ANSWERS -- SURRY 1&2 -87/02/09-DEAN, WM ANSWER 2.23 (1.50) '
\s a) (3 motor generator set 32 (4) eactor trip breaker 12 (2) p er cabinet q?j k4 [G (h[
(7) logi% cabinet 4 1 C(fatJty yhg (6) rod pd%ition injd' tion cabinet (5) r Vc on t r ol unit
$ 94._, 4 automat \1 1 J (1) DC hold 4 inet /1 O W O d qcrit4 - "
(0.75 for a 1 y cor t, 0.75 for b) fully correct, b,
-0.1 for ep switch ne ed to place a component
- In proper order)
REFERENCE ND 93.3-LP-3 Rod Control System 001/000 K4.01 (3.5/3.8) l 1 I l
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" Stafni has a Magnitude and o Pet'arity (for birection)
La speed & b ir e c.tien Datermines Rod Speed and Aute Diceettea S
- nol Controlter Direc. tiers in A uT o Manual 5paed Auto Samad SISnal 1 F Aute 5 '"* ' 4 *
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232__1USIBUMENIS_0ND_CDUIBOLS PAGE 8 ANSWERS -- SURRY 1t<2 -87/02/09-DEAN, WM ANSWER 3.01 (1.00) a REFERENCE ND 93.3-LP-16, p7 012/000; K6.10 (3.3/3.5)
'9 ; ,
ANSWER 3.02 (1.00) a) STSP decreases (0.5 EA) b) STSP decreases I REFERENCE ND 93.3-1p-14, p.7 012/000; A1.01 (2.9/3.4) i ass ..ER 3.03 (1.00) a (1.0) REFERENCE ND 89.2-1p-8 045/010; K1.11 (3.6/3.7) ANSWER 3.04 (1.00) a aaaaaaaaaaaaaaaa REFERENCE ND 93.3-LP-6 035/010; A2.03 (3.4/3.6) I ANSWER 3.05 (1.00) i a (1.0) REFERENCE ND 93.3-LP-7 g 011/000 A2.10 (3.4/3.6) i
gha Kt4 UNITED STATES y, ,o,, NUCLEAR REGULATORY COMMISSION 3 f, o REGION il
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{ 3 3__INSIBQdENIS_ANQ_GQNIBQLS ANSWERS -- SURRY l t<2 -87/02/09-DEAN, WM i- ' e i i 1 l I, . !,
- ANSWER. 3.06 (1.00) {
l d dddddddddddddddd 4
- REFERENCE j ND-91-LP-3. p 20 i 005/000; K4.11 (3.5/3.9)
ANSWER 3.07 (2.00) j a) 2 l b) 4
- 3) 3-REFERENCE j j ND 93.3-LP-10 ,
1 012/000; PWG-4 (4.1/4.3) . i I 4 1 t l ANSWER 3.08 ( .50) 1 i 4 TRUE 1 I I I REFERENCE ND-93.2-LP-2 a 015/000 K1.01 (4.1/4.2) } 4 I i i l ANSWER 3.09 (2.00)
)
i a) -Unit 1 or _ Train B SI (0.33 ea) 1
! -Unit i or 2 Train B Hi-Hi CLS j -1J or 2J bus undervoltago
, b)- Race between breakers to close (0.5) Which ever closes first locks out the other breaker (0,5) 1 REFERENCE ND 90.3-LP-7 j 062/000 K3.02 (4.1/4.4) i i j l 1 1 i t i 1 u _ _ _ _ __ _ _ _ _ _ _ .._ _ _ . ... - . -
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3 t__INSIEUUENIS_GUR_GQUIEQLS PAGE 10 ANSWERS -- SURRY 162 -87/02/09-DEAN, WM s ANSWER 3.10 (2.00)
- 1) PZR Hi Press. Tri p (0. 2) 2370 psig(O.1), 2/3(0.1)
- 2) PZR La Press. Trip (0.2) 1875 psig (0.1) , 2/3(0.1)
- 3) PZR Lo-Lo Press. SI(0.2) 1715 psig and not blocked (O.1),
2/ 3 (0.1 )
- 4) y P-11(0.2) (2000 psi g (O.1) , 2/3(0.1)
D) 'Pr e s s . input to the OT Delta T(O.4) REFERENC
# N U 5 All*AQNOI NM 9$N0uCN" z)
ND 93.3-LP-12 / LP-14 - pt(2 00gv e(o&c rep,ac,g [o. L) 010/000; K1.01 (3.9/4.1) h ) (O. L )
- 5 L hlockfunl)(ocl' Et9 ANSWER 3.11 (1.50)
- 1) Adequate SDM upon trip (0.5 ea)
- 2) To minimize the amount of positive rear.tivity inserted during a rod ejection accident, and
- 3) To minimize radial flux tilt (peaking)
REFERENCE ND 93.3-lp-3, p. 25 001/000; K5.04 (4.3/4.7) ANSWER 3.12 (1.00)
- 1) Control switch to close (0.25)
- 2) Synchronizing selector switch is ON (0.25)
- 3) Overcurrent/ generator diff trips <eset (0.25)
- 4) Aux trip relay reset (0.25)
REFERENCE ND 90.3-1p-3, p.12 064/000; A4.01 (4.0/4.3)
.. . . . . . ~ . - . .-
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3 t__INSIBUMENIS_GUD_GOUIBOLS PAGE 11 ANSWERS -- SURRY 1!42 -87/02/09-DEAN, WM i. 4 ANSWER 3.13 (1.50)
-PZR high water level (0.25 en)
- -PZR la pressure ' ~ ' ~
I
-Lo primary coolant flow [7 Qd
-RCP breakers open (two pumps)
-Under voltage on 2/3 4 KV buses
-Turbine trip REFERENCE ND 93.3-LP-16, 012/000; K4.06 (3.2/3.5) 1 ANSWER 3.14 (2.00) l
- 1) MCC 1H1-1 to the Battery Charger 1A1 (or 1A2) to Inverter 1-III to App R isol. breaker to Transfer switch to Bus 1-III. (1.0)
- 2) Batt 1A to Inverter 1-III to same as 1 (0.25)
- 3) MCC 1H1-1 to SOLA transformer to Transfer switch to Bus 1-III (0.75) 4 REFERENCE ND 90.3-T-5.1 062/000; K4.09 (2.4/2.9) 4 4
ANSWER 3.15 (1.00) i 1) RCS pressure >2000 psig
') SI REFERENCE ND 91-LP-2 006/000 A3.01 (4.0/39) l
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f 3 t__IUSIBUMENIS_GUQ_CQUIBgLS PAGE 12 ' i j ANSWERS -- SURRY 1&2 -87/02/09-DEAN, WM i - l ANSWER 3.16 dO Fr. ' Sea l ! i t t j Charging Pump A l
- 1) Low charging header pressure (0.25) i 2) SI Train A (0.25) l 3) Undervoltage or degraded bus voltage on emergency bus l J (0.25) l 4)- All other charging pump breakers (for charging pumps l B, C(alt) and C normal) open. (0.25) {
Charging Pump C (Normal power supply) .
- 1) Low discharge header pressure (0.25) i
- 2) SI train A or B (0.25) I
- 3) Undervoltage or degraded bus voltage on emergency bus !
J (0.25)
- 4) Charging pump A & B breakers open (0.25)
{ ( 5) C alt supply must be open for any C normal start ) ; ,1 REFERENCE I ND 88.3-LP-5 p.5.12/5.13 l 004/000; K4.04 (3.2/3.1) ( i i J ANSWER 3.17 (1.50) ' 1 i } -Pressure ' 390 psig (0,5) AND either PORV Block valve shut (0.5) { OR Associated Key switch in disable (0.5) REFERENCE ND 93.3-LP-6 p. 6.7 l i 010/000; K4.03 (3.8/4.0) i {; ANSWER 3.18 (1.00) 1 i The design change resulted because of experiences where the ' undervoltage trip signal alone was not sufficient to trip l the breaker. (1.0) l REFERENCE ND 93.3-LP-10, p. 9/10 012/000; M6.03 (3.1/3.5) r l l i i i 1..
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3s__INSIBUMENI@_GNQ_GQNIBQLS PAGE 13 f ANSWERS -- SURRY 1&2 -87/02/09-DEAN, WM i, . f i l i ANSWER 3.19 (1.00) J j The highest reading upper / lower detector is compared to the I average of the upper / lower detectors (0.5). The circuit
) auto defeats below 50% power on ALL channels (0.5). !
1 REFERENCE l' ND 93.2-LP-4 i 015/000; K6.04 (3.1/3.2) & A1.04 (3.5/3.7) ANSWER 3.20 (1.75) j a) Cabinet has the capacity to support up to 6 stationary j gripper coils simultaneously (0.5). So with 2 groups or
- more, would overload / heat the cabinet (0.5).
l' b) 125 VDC-Latching Rods 70 VDC-Holding Rods (0.5 for reasons, 0.25 for correctly associating voltages) REFERENCE ND 93.3-LP-3, p. 15/16 001/050; PWG-1(3.6/4.1) ANSWER 3.21 (1.25) i 38% setpoint from 0-20% (0.5) Turbine power (0.23) and linearly from 38-108% as Turbine Power goes from 20-100% (0.5) i REFERENCE ND 91-LP-3, p.9 013/000; K1.01 (4.2/4.4) I k
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Gu__INSIEUdEUIS_GUQ_CQUIBQLS PAGE 14 ANSWERS -- SURRY 18<2 -87/02/09-DEAN, WM ANSWER 3.22 (1.00) I Accomplished by the Non-Return valve. (0.25) A reverse differential pressure is sensed across the valve (0.25) A sensing line on the down stream side of the valve which directs down stream pressure to the upper area of the valve d'sc. (0.25) i Which drives the disk down when up stream pressure is< downstream pressure.(0.25) REFERENCE ND 89.1-LP-2 039/000 K4.06 (3.3/3.6) ANSWER 3.23 (1.50) See attached sketch REFERENCE ND 93.3-LP-3 001/000; K4.03 (3.5/3.8) t l J
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r-4.__EB9CEDUBES_:_NQEdQ6t_GBUQ8066t_EUEBGEUGy_60Q PAGE 7 , BORIOL0tilGOL_G9dIB96 l ANSW$RS -- NORTH ANNA 18< 2 -87/02/09-DEAN, WM ANSWER 4.01 (1.00) e REFERENCE McG, AF'/ 2 / A /5500/ 23, p. 2 Surry AP-14, pp 4 EPE-051; PWG-11(3.7/3.7) ANSWER 4.02 (1.00) 2 REFERENCE VCS, SOP-101 p1 NA OP-5.2 p 4 SUR OP-5.2 p 2,4 ANSWER 4.03 (1.00) four @ 0.25 points each:
- 1. Rod bottom light
- 2. Computer alarm, power range tilt, rod deviation / sequence
- 3. Flux deviation alarm (3)
- 4. Rapid drop in Tavg and power level
- 5. Rapid drop in pr r level and pressure
- 6. Noa , _ , .;: ,s qL, '
p; ,$ d ,u p cf p g( g g y g g gg
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REFERENCE ' #' " 553P (1/M A NAPS 1-AP-1.4, p.3. 8 TgP,T_ ei is ,t ,, 001/050 PWG-10 4.3/4.5 9 78 @ ^ '
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g ANSWERS -- NORTH ANNA 1842 -87/02/09-DEAN, WM ; ] ANSWER 4.04 (1.00) ) G O.5 points each: I l 1. Trip turbine locally. j 2. Manually open reactor trip breakers or the rod drive MG output breakers. REFERENCE NAPS 1-AP-20, p.3. SUR 1-AP-20, p5 1 ANSWER 4.05 (1.00)
, b , REFERENCE NA ADM 5.8, pp 2/3 Sur SUADN-ADM-21 p 21 PWG-23: Plant Staffing and Activities (2.8/3.5) i I
- ANGWER 4.06 (1.00) b DF=1 for gainma 100(45/60)(1)=75
, REFERENCE 10 CFR 20. , PWG-15: Radcon Knowledge (3.4/3.9)
{ . i
; ANSWER 4.07 (1.00)
D REFERENCE
, MNS, AP/2/A/5500/14, Cane I, p.2.
CAT, AP/1/A/5500/15, Case I, p.2. Surry AP-1.OO pp 2,3 I, I
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4t__EBQQEDUBE5_;_UQBM86t_GEUQBMOL3_EMEBGEUCY_@UD PAGE 9 BBDIQL,0GICQ6_.CQUISQL ANSWER $i -- NORTH ANNA 1&2 -87/02/09-DEAN, WM NA, AP-1.0 p 4 001/050; PWG-11(4.4/4.4) ANSWER 4.08 (2.00) a) Shutdown (+.5 ea) b) Startup c) Shutdown d) 57M4 P REFERENCE SON TS 3.4.6.2 NA TS 3.4.6.2 SUR TS 3.1-13 002/020; PWG-8 (3.5/4.4) ANSWER 4.09 (1.00) Whenever CRDM'u are energized. or if CRDM's are energized when primary plant temp in 100F to 350F REFERENCE NA OP-21.1 p4 NA 1-OP-1.2 p5 SUR OP-21.3 p 9 ANSWER 4.10 (1.00)
- 1. FALSE
- 2. : ~ m. -T{zt)L REFERENCE SUADM-0-13 p 10 PWG-1 3.5/3.9
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80DIRLOGIG06_GQNIBQL ANSWNRb -- NORTH ANNA 12<2 -87/02/09-DEAN, WM ANSWER 4.11 (1.50) a) 4 (+.5 ea) b) 2 c) 6 (5 for Surry) REFERENCE NA OP-2.1, pp 9-13 Surry OP-2.1.1 pp 14-19 PNG-12: Perform Integrated Plant ops (3.5/3.4) ANSWER 4.12 (2.50) a 2 b 4 },*.dO 'L' c 1 d 1 o b REFERENCE Virginai Power GET pp 14-17 PWG-15 3.4/3.9 ANSWER 4.13 (1.00)
-Unexpected rise in S/G 1evel (+.25 ea) -High radiation on a S/G blowdown line -High radiation on an MS line moni tor -High radiation an determined by sampling and analysis REFERENCE Surry EP-4.OO, pp 2 1 NA 2-EP-3, pp 2 EPE-038; EA2.03 (4.4/4.6)
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r St__PBQGEQUBES_ _NOBd@Lt_O@ugBdOLt_EdE8QENGy_ANQ PAGE 11 B00I96901 COL _Cgu1Bg6 ANSWEFd3 - NORTH ANNA 162 -87/02/09-DEAN, WM ANSWER 4.14 (1.00) pam?%0%
- 1) Verify Charging /S1 N (+.5 ea)
- 2) RCS Subcooling less than 25F REFERENCE SONP Foldout Page NA Faldout page for 2-EP-O Surry Foldout page for EP-1.OO 003/000; PWG-10 (4.1/4.4)
ANSWER 4.15 (1.00) Any 4@ 0.25 points each: North Anna Surry
- 1. Bank C ponition. 1. Date Cri ti cal
- 2. Dank D position. 2. Time Critical
- 3. Auct. High Tavg. 3. Average RCS temp.
- 4. IR N35. 4. RSC Doron concentration
- 5. IR N36. 5. Bank C Control rod ponition
- 6. RCS baron concentration. 6. Dank D Control rod position
- 7. Actual critical position within admin. requirements REFERENCE NAPS 1.OP-1.5, p.12.
SUR 1-OP-1C App. A p 10 of 10 001.010; KS.08 (2.9/7.3) ANSWER 4.16 (1.00) Surry (+.2D ua) North Anna (+.25 ea) Verify SI/CHG pumps running /f1ow 1. Verify 2 SI/CHG pumps running /f1ow Check RCS pressure <2335 pnig 2. Switch DATP to fast speed Switch DATP to fant speed 3. Open MOV 2350 or Inject the DIT Open MOV-( )350 4. Chuck per presc <2335 REFERENCE Surry FfiP-S.1 p 3 NA FRP-S.1 p4 EPE-029; PWO-11 (4.5/4.7)
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ANSWERS -- NORTH ANNA 1&2 -87/02/09-DEAN, WM
- ANSWER 4.17 (1.00) icinizing radiation; heat stress; differential pressure; O2 deficiency
(+.25 ea) REFERENCE Surry SUADMO-19 p 3 NA ADM 20.9, pp 1 PWG-18: Knowledge of Safety Procedures (3.0/3.1) ANSWER 4.18 (2.50) CC surge tank low level alarm Reactor containment air ( , Motor protection alarm CC pump retirc coolers hi-temp.' CCW low flow discharge header alarm CCW low presuuru discharge header alarm 5 .i) .5 ea , i Reactor toolant pump flow / temp alarm Exceus letdown HX flow / temp Non-regenerative HX (Mian Coolant) high temp Primary shield water wall coolers lou pressure Primary shield penetration cooling coils low pressure Neutron shield tank coolers low flow /hi temp REFERENCE SUR AP-15, p 1 008/030 PWG-10 3.8/4.2 i AN3WER 4.19 (3.00) '
- 1. Inolato pzr por v's ,
- 2. For stuck open spray val ve onl y, stop RCP associated with the individual 3 >3 p r a y line
- 3. If the reactor trips or a trip is imminent, trip the reactor
- 4. If SI is initiated or is imminent, manually SI
- 5. If RCS subcooling cannot be maintained, manually SI /
- 6. Request initiation of Emergency Plan .50 ea
! REFERENCE SUR AP-42 p2 I /
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3:__P60GEQQEES_ _UQSUOLt_G@UQBdO(t_EUESQEUQy_ONQ PAGE 13 RAGIOLQGIGOL_QQNTRQL e . ANSWNRS -- NORTH ANNA 1&2 -87/02/09-DEAN, WM ANSWER 4.20 ( .50)
- 1. Place delta T and Tave Defeat switches to defeat the affected loopE.1253
- 2. Select delta T recorder to unaffected loopE.125]
- 3. Trip the affected pump and commence shutdown ( I AW OP--3.1 ) E . 25 J REFERENCE SUR AP-9 p3 000/015 PWG-10 4.2/4.S ANSWER 4.21 (2.00)
- 1. Reduce turbine load
- 2. Open MOV-CP-100 to bypass the condensate polisher
- 3. Start the standby condensate pumps
- 4. Trip the reactor if a trip is imminent , .5ea REFERENCE SUR 1-AP-21 p 3 059/000 PWG-11 3.9/3.9 ANSWER 4.22 (1.00) i An unsaturated m2xed bed demineralizer will remove boron from the reactor coolant system (.50) and add.poritive reactivity.(.5)
(Reasonable wording accepted) REFERENCE ' VCS, sop-102 p 1 NA OP-8.2, p 5 , SUR OP 8.2, p2 - Y 004/eOO K6.02 2.5/2.1 < f e I
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l St__ESQGEDUBES_,; _NQBdG(t_.6BNQBd6L3._EMERGENGLONQ PAGE 14 RADIOL.OGICAL CONTROL j ANSWdRS - NORTH ANNA l t<2 -87/02/09-DEAN, WM ANSWER 4.23 (1.00) Steam pressure mode CO.25] Tavg input to the steam dump control is not valid without forced flow in the 1 cops. E.753 0A Tev3 co ufa b6 7TV7# F . m.[ 57pf /}ciff ,4coc 7o de/qbw e fy/*g REFERENCE [*N) NA 1-AP-10, Att. 2, p2 SUR AP-39, p 4 t { { 1 \ _ _ - - . ..-. -- --, - . , . - - - - - - - - - - - - - -
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