Text

Exam Rept 50-237/OL-87-03 on 871026-30.Exam Results:Seven Senior Reactor Operators & One Limited Senior Reactor Operator Passed Exams.One Senior Reactor Operator Candidate Failed to Obtain 80% Overall Grade for Written Exam
	ML20236V628
Person / Time
Site:	Dresden
Issue date:	11/24/1987
From:	Bjorgen J, Burdick T, Dave Hills, Keeton J, Nejfelt G; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:	;
Shared Package
ML20236V620	List: ML20236V617; ML20236V628;
References
	50-237-OL-87-03, 50-237-OL-87-3, NUDOCS 8712040420
	Download: ML20236V628 (121)
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{{#Wiki_filter:- U.S. NUCLEAR REGULATORY COMMISSION REGION III i Report No. 50-237/0L-87-03 1 Docket Nos. 50-237; 50-249 Licenses No. DPR-19; DPR-25: j Licensee: Commonwealth Edison Company I Post Office Box 767 i Chicago, IL.60690 l l Facility Name: Dresden Nuclear Power Station j 1 l i Examination Administered At: Dresden Nuclear Power Station i Examination Conducted: 0 tober 26-30, 1987 //- 2 //-- 79 Examiners: fi s Date l l / / // - ff~ 87 J en f..o Date tO

l. Keeton

//[.29 7 { Date / 2~.N G. M. Nejf it / J7 Eate i [. f Approved By: T."M. Burdick, Chief A Operator Licensing Section Date Examination Summary Examination administered on October 26-30, 1987 (Report No. 50-237/0L-87-03) Written examinations were administered to eight-. Senior Reactor Operator (SRO) candidates. In addition,-a limited fuel handling SR0 examination was administered to one candidate. Results: Seven SR0 candidates and one SR0 limited candidate passed these examinations. One SR0 candidate failed to obtain an 80% overall grade for the written portion of the examination. 8712040420 871201 PDR ADOCK 05000237 V PDR l

REPORT DETAILS 1. Examiners D. E. Hills, Chief Examiner J..C. Bjorgen J. M. Keeton G. M. Nejfelt 2. Exit Meeting At the conclusion of the site visit, the examiners met with facility representatives. The following' personnel attended this exit meeting: Facility Representatives l E. D. Eenigenburg, Station Manager J. W. Wujciga, Production Superintendent C. W. Schroeder, Services Superintendent J. Kotowski, Assistant Operations Superintendent S. R. Stiles, Training Supervisor S. Mattson, Training Operation Group Leader R. J. Falbo, Regulator Assurance Group Leader NRC Representatives S. G. DuPont, Dresden Acting Senior Resident Inspector D. E. Hills, Operating Licensing Examiner G. M. Nejfelt, Operating Licensing Examiner The following items were discussed during the exit meeting: a. No generic training weaknesses or strengths were observed during the operating examinations. (This comment was further substantiated upon review of examination results in NRC Regional Offices). b. Security and radiological accesses for examiners to the plant did not result in any undue delays. c. Status of the suspension of NRC Requalification Examinations was discussed. d. Active versus inactive license status in accordance with 10 CFR 55 as it applies to a Senior Reactor Operator Limited to Fuel.. Handling License was discussed. 3. Examination Review On August 20, 1987, the NRC received Registered Copy No. 14 of the Unit 2 and Unit 3 Tachnical Specifications (TSs) with the following TS amendments not incorporated into the TS Registered Copies: 1 2 I

g. 1 I Amendment. Unit Numter Issued 2 94 11-10-86 2 95 03-31-87 3 j 90 11-10-86 On November 11, 1987, the Dresden Training Supervisor and the NRC Resident Inspector.were informed of this documentation control problem by G. M. Nejfelt. Specific facility comments concerning written examination questions are 4' followed by the NRC responses in Attachment 1 for the'SR0 examination and in Attachment 2 for the Limited SRG' examination. Additional answer key I expansion or clarification that was a result of. grading the SRD examination is enclosed as Attachment 3. l /- i .rt i l ) (.r es. E ' h i: 3

n i l j l' ATTACHMENT 1 i 3 RESPONSE TO FACILITY COMMENTS DT 0T7iTEIER 26,1987,DRESg SENIOR REACTOR OPyRATOR LICEME EXAMINATION 5.10 Facility Comment: Embrittlement is caused by the fis'sion products Iodine and Cadmium, not Cesium. See reference - GE Thermodynamics, Heat Transfer and Fluid Flow (BWR), Pages 9-124. ,1 NRC Rr.olution: Comment is accepted. In the answer kny, " cesium" was changed to " cadmium" to be consistent with example provided in the reference cited. 5.13.b J Facility Comment: Additional acceptable answer: " Reduced breadoning/ F with an increase in temperature and resonance peaks begin to overlap." 1 See re.erence - Reactor Operation, Book 3, Page 9.1 2. NRC Resolution: Comment is accepted. The answer key will be changed to include: " Reduced broadening / degree F with an increase in temperature and resonance peaks begin to overlap," as a j correct altervte answer. 1 6.05.c . I - (. i A Facility Comment: The question is m bleading. It could be interpreted as. The ACAD can only dilute the containment atmosphere by f injectfng compressed air (Answer true) or, it cannot dilute 1 4 by injecting another containment gas such as Nitrogen (Answer false). NRC 4esolution: Comment is not accepted. The Atmosphere Containment Atmosphere Dilution (ACAD) System is divided into two subsystems: the Dilution Air injection System, and the v Pressure Bleed System. No addition of compressed air is necessary for the ACAD to dilute the containment atmosphere using its Pressure Bleed System. [DLP Book 3, / Containment Systems, Section 15, Page 48.] !/ 6.11.d Facility t.:mment: The question is inacruratet;nd therefore potentially l aisleading. There is only one y-2125vdc battery, @deas I the question referreRto "Both Unit 2 '25vdc Battery Systems." This could easily be iderpreted toWe a 1 / I f r I 1,

l V J l l } ~ 8 r failure' of "both batteries supplying Unit 2," in which !cass no auto transfer could occur (as the answer key l ,' indicates) and the'2/3 Diesel would be considered- 'I inoperative. j NRC Resolution: Comment is accepted. Item "d" for Question 6.11 is deleted; and the point value for the question is reduced from 2.00 to.1.50 points. i \\ 6.14 ] Facility Comment: Total points in this question add up to 2.75 points versus 2.5 poin n as indicated. In the answer key, Part b is l worth.75, points, in the question it is worth 1.0 points. I i s NRC Resoluticn: Comment is accepted.' The indicated point value of Part "b" for Question 6.14 is. reduced from 1.0 to 0.75 point.to be consistent with tu answer key. 4 7.03.a i 9 Facility Comment: Additional correct answer: " Unbalanced thrust'." l l See reference - Condensate and Feedwater Lesson Plan, l Page 32, G.1.c. NRC Resolution: Comment is accepted. The answer key will be changed to include: " Unbalanced thrust," as a correct alternate-answer. I l 7.08 i Facility Comment: Should also acce'at 5.mr/hr as per D0A 902(3)-3 B-3. A' l procedure inquiry has 'acen initiated to correct this inconsistency. NR? Resolution: Comment is accepted. Full t.redit will be given for an answer of "5 +/- 1 mr/hr" to allow for the inconsistency of the operator's immediate action between D0A 902(3)-3 j B-3, Revision 4, with a sptpoint of 5 mr/hr; and 00A 1300-1, Revision 4, and D0P 1300-2, Revision 2 with setpoints of 20 mr/hr. Note: Four of the eight SR0 candidates specifically addressed both the annunciator and abnormal procedures in answering Question 7.08. 7.15.b Fac lity Comment: The use of the words " Primary Concern" is misleading. 00A 3700-1 does not specify any concern as being " Primary." The use of the words "does not involve individual components" is al.so misleading. The answer key response as written is due directly to the failure of the Drywell Coolers which are individual components. This question is not specific.enough in nature to achieve the-expected response, therefore other answers shou'1d be considered. 2 J mu______._________________

i J NRC Resolution: Comment is accepted in part. The answer key,has been expanded to include, as alternative answers for full credit: (a) heating damage to recirculation pumps; (b) neating damage to RWCU non-regenative filter bed; OR l (c) loss of drywell' coolers. l 8.01 Facility Comment: Although not listed specifically on Page B 3/4.5-32 of the Technical Specifications, there are additional design bases listed in the Lesson Plans for Low Pressure Coolant i Injection and Core Spray which should be included as j additional accept 3ble answers. Throughout the LC0's and i Technical Specifications Bases for LPCI, Core Spray and 1 l the Diesel Generators the topics of redundancy,. ) i testability and the ability to operate with a loss of off-site power are addressed. We therefore submit that any of the design bases included in the attached references be included in the answer key. I See reference - Dresden Lesson Plans Low Pressure Coolant Injection, Page 2 Core Spray, Page 2 NRC Resolution: Comment is accepted. The answer key will be changed to l include: (a) To prevent the temperature of the fuel cladding from becoming excessive. (b) In order to achieve redundancy. i (c) Each subsystem is designed to provide adequate core cooling. (d) The subsystems are designed to function with a j complete loss of offsite power. (e) The system components are to be testable during all modes of plant operation. 1 8.05.b i Facility Comment: Daily Orders are typically for a 24 hour period, however reference DAP 7-3 B.I.d states "24 hour time period or to 1 cover a weekend," therefore the expected answer could be J True or False, i NRC Resolution: Comment is accepted in part; and the answer key has been changed from " FALSE" to "TRUE," because'a weekend is a l longer period of time than 24 hours. Due to the information given in your comment, the answer cannot be considered false. 4 j 3

i .c ~'f 7 l 1 8.05.c l Facility Comment: This question is misleading: -1 (a) Operating Orders do have a maximum six month lifetime 1 (expires June 30 and December 31). The answer -{ anticipated an ongoing Operating Order which would expire in six months'. (Answer true). ] 1 (b) However, as an example an Operating Orde'r issued in May would expire in two months. (Answer false). See reference - DAP 7-3. 1 NRC Resolution: Comment is accepted. Item "c" for Question 8.05 is '] deleted; and the point value for the question is reduced u from 2.00 to 1.50 points-R 8.08 I Facility Comment: Question should be deleted. Paragraph B.1.a. of DAP 7-4 j defines'" temporary system alteration" but never refers to "non-emergency conditions." This procedure refers to. i several cases, each requiring different responses. This~ l questTon is not clear enough to expect ~a specific answer. See reference - DAP 7-4 i NRC Resolution: Comment is accepted. Question 8.08 is deleted. l l 8.16.b i Facility Comment: Question is invalid. Operation is permitted for > 24 hours I in single loop per Amendment 95 oT the Technical Specifications. See reference - T.S. 3.6.H.3 (Page 3/4.6-13).and 3.6.H.3.F (Page 3/4.6-15). NRC Comment: Comment is accepted. Your examination' critique documentation - Unit 2 Technical Specification. (TS), Section 3.6, Amendment No. 95, dated March 31, 1987 - supported your comment that one loop operation longer than 24 hours is not prohibited. Therefore, Item "b" for j-Question 8.16 is deleted; and the point value for the question is reduced from 2.25 to 1.50 points. The material sent to the NRC for examination preparation - Unit 2 TS, Section 3.6.H.3, Amendment No. 85, Page 3/4.6-12, dated February 27, 1985 - provided documentation that single loop operation was limited to 24 hours. 4

q The importance of providing up-to-date material, as required by the 90 day letter, is stressed. It is I requested that future examination reference material be verified to ensure its applicability, linit 2 TS Amendment i Ho,~95 was implemented approximately July 1, 1987 based I on a telephone conversation between Mr. S. R. Stiles end. Mr. G. M. Nejfelt on November 9, 1987. .j -l .] I h i 1 l i -) i s 5 ii-

t j ATTACHMENT 2 RESPONSE TO DRESDEN FACILITY COMMENTS FOR THE SENIOR REACTOR OPERATOR LIMITED i i l WRITTEN EXAMINATION OF OCTOBER 26, 1987 a 1 P.07.a Facility Comment: Additional acceptable answer: 30 mr/hr as per DFP 800-20, Page 3. l NRC Response: Agree with comment. Information given in DFP 800-20 I disagrees with Process Radiation Monitoring Lesson P'ian. Therefore, the answer key is changed to accept either 201 5 mr/hr or 301 5 mr/hr. P.07.c Facility Comment: Additional acceptable answer: " hoisting motion is disabled." NHC Response: Agree with comment. This would be an acceptable answer if a more liberal interpretation of the term " indication" is used. Therefore the answer key is changed to also accept " hoisting motion is disabled." Q.07.c Facility Comment: The answer found in the Dresden Radiation Protection Standards assumes the individual works only one shift per day. In the event that the individual works more than one I shift per day which is common for fuel handlers at Dresden l during an outage, the job Foreman is expected to initial the RWP each shift. NRC Response: Agree with comment. The answer I;ey is changed to also accept "once each shift." Q.11 Facility: Comment: Disagree with the answer. At Dresden station if a worker i has an NRC Form-4 on file, the 10.CFR 20 quarterly dose limit is 3 REM, in which case 5(N-18) applies. Answer should be Yes, 5(N-18) will be violated if the worker is permitted to receive a dose of 50 mrem. Life = 29,995 mr +50 mr 30,045 mr 5 (24-18) = 30 REM or 30,000 mrem 1 l

.. j j t NRC Response: Disagree with comment. _The' question emphasized 10 CFR 20' i limits.and not Dresden administrative limits. 'The question. I was designed to determine if the candidate had a' thorough understanding of 10 CFR 20 limits (i.e., more specifically that 5(N-18) limits apply on"ly if. exceeding 1.25 rent / quarter). Therefore,.the answer key remains unchanged. . I -l i l 2 I 4 I l l l 2

1 i l ATTACHMENT 3 i EXPANSION OR CLARIFICATION OF SRO EXAMINATION ANSWER KEY, AS A RESULT OF GRADING THE EXAMINATION l Question No. Answer Key Modification .j 5.06 Fuel Temperature Coefficient added as. synonymous ~ term for the Doppler. l 5.09 Statements that IRMs and SRMs measure neutrons and that these detectors are compensated to' discriminate for i gammas are respectively given 0.50 and 0.25 points. 6.03 For Part'a: "High enough or Low enough [0.25] to prevent I spurious operation [0.25]." For Part b: " Low enough or High enough [0.25] to prevent ' fuel cladding melting [0.25]." { Unit 3 TS, Amendment No. 33, page B 3/4.2-29, stated that'- l the Low-Low reactor water level trip setpoint was chosen to be high enough.to prevent s.purious operation;..and low -l enough to prevent ~ fuel cladding melting. Logicallys the "high enough" and " low enough" for these statements should j be reversed. j l 6.08 Added " Detector Wrong Position Interloc~k" rod block and I " Detector is not ' FULL.IN' when the reactor mode switch is i in 'STARTUP'." Cited reference was changed from page 17 1 to pages 17-18. i 6.12 Added as item (7), " Vent line radiation monitor." 1 6.17.6 Added"(CARD 0XManual)." 8.02 Added reference: DOP 400-2, Revision 5, Section E.6, l page 5. ) 8.14 Added to answer: "Also, full credit will be given for-other items stated in the Operating Order (i.e., circuity modification and NRC reporting)." i l E

MASTER G0QF U. S NUCLEAR REGULATORY'COMMIS SENIOR REACTOR OPERATOR LICENSE EXAMINATION s. FACILITY: QRESQg8_______,,,___,,______ REACTOR TYPEt @HB-GE@__________________ DATE ADMINISTERED: @Zd19226_________________ EXAMINER: N E J E E L I f _ G 2., _ _ _ _ _ _ _ _ _, _ _,,, CANDIDATE INSIBUGIIDNS_ID_GBURIDoIEi 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 } loist 80%. Examination papers will be picked up six (6) hours af ter the examination starts. % OF CATEGORY % OF CANDIDATE'S CATEGORY __Y9LUE_ _I9166 ___SG9BE___ _YoLVE... ______________GoIEG98Y M.44 _Ed ZE__ _ M p ___________ ______.5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS,AND THERMODYNAMICS M.cv .26,67 _itIff.4. _ E s s is h _ _ _ _ _ _ _ _ _ _ _ _______ 6. PLANT SYSTEMS DESIGN, CONTROL, ~ V AND INSTRUMENTATION .M.44 _E31ZE_ _E d ? ___________ ___ ____ 7. PROCEDURES'- NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL .W. ~74 24.fo _E s s 2 2.p. _s% _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ 8. ADMINIS'TRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS / 0 /. 2D S Q' M _is5Ai-K Totals ,A, ~ Final Grade t'ec.h.3 '...~ }p f.oy .. u LJ G. $ K,a N,* 7 All work done on this examin ion is my own. I have neither giv,_en w.w e ~w '"-.A c, g, g,

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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS Du ing 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, stort 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, 14, 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 tney are commonly used in facility literature.

1 1

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 arisumptions used to obtain an an'wer j

to mathematical problems whether indicated in the question or not.

15. Partial credit may be given.

Therefore, ANSWER ALL PARTS OF THE j QUESTION AND DO NOT LEAVE ANY ANSWER BLANK. 1 1 1

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 completet, l l l 1

b

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. Turn in all scrap paper and the balance of the paper that you did c. not use for answering the questions. \\ d. Leave the examination area, as defined by the examiner. If after { 1eaving, you are found in this area while the examination is still 1 in progress, your license may be denied or revoked. i l l I l 1 i 1

5. THgg8Y_gE_NugLg68_PQWER_PL6NT_QPgR6TIQUt Pasa 4 ELyID@t6HD_ISEBUgpINQUlg@ l ~ / DUESTION 5.01 (1.50) STATE the mode of furi failure prevented by each of the following thermal limits. a. Linear Heat Generation Ratio (LHGR) (0.5) i b. Average Power Linear Heat Generation Ratio (APLHGR) (0.5) c. Critical Power Ratio (CPR) (0.5) l l l QUESTION 5.02 (1.00) STATE why the centrifugal pump high point easing vents is used to remove noncondensible gases prior to starting the pump. i ) QUESTION 5.03 (2.25) l DESCRIBE how energy is transferred by the following and provide an l example of each within the reactor vessel. a. Conduction b. Radiation c. Convection l I QUESTION -5.04 (1.00) l EXPLAIN why deleyed neutrons have a larger effect on reactor period than i prompt neutrons. l l l QUESTION 5.05 (1.00) DESCRIBE TWO (2) ways Xenon 135 is removed from the reactor. 1 1 4 (***** CATEGORY 5 CONTINUED ON NEXT PAGE K****)

9, EA,_INE0RY'0F NUCLEAR POWER PLANT OPERATIgN - , Pose 5 1 - EkWIDE16HQ_IBEBdQD1BoblgS. GUESTION: 5 06 (1.00)- L STATE the reactivity coefficient that providesfimmediate feedback to. control reactor Power for a sudden. load increase.. The.reactorEremains at power; and no. automatic not operator action occurs. QUESTION 5 07 (1.50) Prior to performing a control rod' notch calibration, the reactor is1Just critical at 0.002% power, the control' rod is withdrawn one notch.and reactor power increases on a steady period with a doubling time,of'40. seconds. l a. CALCULATE the reactoi' period. b. HOW LONG will it take to reach 0.08% power? (Show all work _and assumptions') ) (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) lE__ ___1_:_

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ELUIQS 8NQ_ISEBbgQXU8 DIGS 1 '0UESTION 5.08 (1.50) ] i NATCH the following terms on the_left with their correct definition on the .right. (Only one correct answer each) a. Keff 1. Kex = 0. b. Critical 2. The fraction of~the thermal neutrons that had been born by the' decay of,a c. Supercritical delayed neutron precursor. \\ d. Excess Reactivity 3.. The fraction of fission neutrons born i from-the decay of a delayed neutron i precursor. j e. Effective Delayed 4. Multiplication factor dealing with an-J Neutron Precursor infinitely large reactor. f. Shutdown Margin 5. Neutron population is decreasing from-generation to generation. 6. A measure of how far.the reactor is below the critical condition. 7. Multiplication factor dealing with a realistic reactor. This includes leakage. B. The amount of additional reactivity built into a reactor to allow it to reach full power for its operating cycle life. ] 9. Koff > 1 1 10. A measure of the departure from criticality. QUESTION 5.09 (1.50) After a reactor trip from long high power run, thermal output of the reactor is significant for a couple of hours. EXPLAINt a. Where this power is coming from b. Why this power isn't visible on the nuclear instrumentation. (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) I

1 4-DI__IBEDBX_9E_HUGLE68_E9BEB_ELBNI_9EES$11981 Page .7 .] ELVIDE1689_IBEBbQQIN6dIgg I l L) DUESTION 5.10 (1 50) j STATE the pellet / cladding interaction-(PCI) failure mechanism'during a rapid power increase and EXPLAIN what causes this failure. ) l QUESTION 5.11 (1.00) q l STATE at what point during a reactor startup to full power, maximum core q void fraction is attained. i l GUESTION 5.12 (1 00) Of the following operations, which will have a negative effect on'the available net positive suction head (NPSH) of a given centrifugal' pump? a. Throttling open the pump suction va'1ve. b. Throttling open the pump discharge valve. I c. Throttling closed the pump discharge valve. d. Decreasing the temperatur e of the fluid being pumped 4 1 i l l QUESTION 5.13 11.00) l STATE why the fuel temperature coefficient (or dnppler coefficient) of reactivity becomes j i s. more negative over core life, b. less negative with an increased fuel temperature. ] i l QUESTION 5.14 (1 00) A centrifugal pump (i.e., ECCS pump) should not be operated with'its. discharge valve shut. STATE two (2) adverse consequences to pump operation, if a discharge flow path is not established.. I i (***** CATEGORY 5 CONTINUED ON NEXT'PAGE'*****)' l

5. THEORY _9E_guCLg68_EgsgB_eLegI_gggBeIIggi Pass 8 EbMIDS egD_IUgBd991gedIgS ~ t GUESTION 5.15 (1.00) EXPLAIN why an 80 second period is naturally established after every

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DUESTION 5.16 (2.00) } INDICATE for each of the following conditionse if the reactivity worth of l a single contro) rod will (INCREASE, DECREASE, or REMAIN THE SAME)! { a. Void content around the rod INCREASES. i b. Moderator temperature DECREASES. q i c. Adjacent control rod is WITHDRAWN. l i d. Xe-135 concentration around the rod DECREASES. l 1 l 1 1 GUESTION 5.17 (1.50) 1 Suppose beta effective over core life decreases from 0.0071 to 0.0055. With equal insertions of 0.001 delta K/K of positive reactivity a. CALCULATE the change in reactor perintd over core life. (1.0) i l l b. WHAT IS the cause for this change in beta effective over core i j life? (0.5) j GUESTION 5.18 (2.50) i Does a power change from 40% power to 30% power take LONGER, SHORTERr or the SAME, as a power change from 40% power to 50% power? (Assume on}y rod motion'and constant rod speed.) EXPLAIN your answer for BOTH condition,s. Also INCLUDE WHICH coefficients of reacitivity cause power to STABIL%BE and HOW for BOTH situations 2. l i 1 l l l l l l (***** END OF CATEGDRY 5 *****) l l

O 1..Ek6Hl.gISIENg_QEgIGN1_QgNIBQLt_6Hg_IN@lBgNENI611gd pass 9 QUESTION 6.01 (1.00) STATE'the TWO (2) design purposes provided by the torus'to drywell vacuum-breakers? QUESTION 6.02 (1.00) STATE why hotwell level decreases following a loss-af instrument air pressure? GUESTION 6.03 (1.00) STATE the TWO (2) primary considerations that were factored into'the selection of the LOW-LOW' reactor water level instrument trip setpoint. This trip setpoint is 444 inches abosn vessel zero-(e.g., 84 inches above the top of tha active fuel). j QUESTION 6.04 . (1.00) j 1 STATE why the Dcd Block Moni: Lot (RBM) rod block function IS-NOT required- ) below 30% pcwer? I QUESTION 6 05 11.50) 1 STATE if the following statements concer.ning the Atmosphere Containment Atmosphere Dilution (ACAD) System are TRUE or -FALSE. a. The ACAD is principally designed to deincet the drywell prior to entry. b. The ACAD can be operated only in the manual mode. The ACAD can only dilute the contsintent atrosphere by injecting c. compressed air. (***** CATEGORY 6 CONTINUED ON NEXT PAGE

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sk.;EleNI_EISIEMg_QggIGNt_ggNIBght_66Q_IUgIByd[NI611gd Pasa 10 GUESTION 6.06 (1 00) Single loop operation is allowed but procedurally limited to a maximum recirculation pump. speed of 43%'for Unit 2 and 60% for Unit 3. WHY IS this.llmit different for Unit 2 ar>d Unit 37 QUESTION 6.07 (2.00) STATE if the following questions concerning Masn Steam Isolation Valve (MSIV) losic are TRUE or F6LSE. (See Attachment 1 for Control logical schematic): a. With the ' Test

pushbuttoh ?PD) hsid down continuously, the MSIV would stop closing when it' reached.its '90% Open* Position, b.

If the MSIV is fully open, momentarily depressing the ' Test' PB would energize the 120 VAC colenoid. c. If the MSIV is fully open, momentarily depressing the ' Test' PB eould not affect the 125 VDC. solenoid. d. If the'MSIV is fully open, momentarily depressins the ' Test' PB the test valve solenoid would remain enersized. e. The contacts for relay 106 (e.g., 106A, 106B, 106C, and 106D) are open with the HSIV fully open. QUESTION 6.08 (1450B l STATE three (3) tod blocks that can be caused by the Intermediate Range Monitor (IRM') System. (Include setroints.for full credit) ) QUESTION 6-09 (1 00) LIST TWO conditions that casi initiate the Low Peessure Co91 ant Injection (LPCI) loop selection logic, given that both recirculation pumps are running with a 'B' recirculation loop line break. 1 (***** CATEGORY 6 CONTINUED ON NF.XT PAGE

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t 61._ELONI.SISIEdS_QSSIGNt_GQUIBQL1_68Q_INSIBQMENI61198. Post 11 QUESTION 6 10 (2 00) Givent -Reactor power at 80% -Load select set at 90% -Recircula(icn Flow Control System Master Contro11erLin AUTOMATIC -Lop,ic diagram provided as Attachment 2 Would the turbine contr61 valvas-(OPEN, CLOSE,'or REMAIN AS IS),. if the '- 10% bias' to the summing junction for load demand and pressure errc* steam demand signal fails to zero? This junction is clearly indicated ren Attachinent 2. [0.53 Justify your answer. C1.5 pts 3 f.h GUESTION 6 11 (W)? 6 Both Unit 2 125V Battery Systems fail. -STATE the significant operational: consequence (s) that would mesult, for each'o?lthe.following Unit 2 component or system. o. 4 KV breakers. j b. Emergency Core Cooling System (ECCS)-logic. I c. Automatic Depressurization System (AOS) ' logic. ,i -4v-- 0i:::1 C ncreter 'OC? 2/2, >c6fd4 b l 1 QUESTION 6.12 (1.50) s l 1 l Concerning the Isolation Condenser Gystem! h i ( The Unit-2 Reactor Operatcr reports to you, as the Shift Supervisor, J that the rounds person noticed.an increased noise levelLcoming from j ] l the isolation condenser. STATE what this increased noise level indicates and LIST four (4) ether indications that could be checked. to verify this diagnosis. 1 1 h i (exxxx CATEGORY 6 CONTINUED ON NEXT,PAGE

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} i t 41L_EkeBI_SI[TEdgQggIQHt_QQNI8QLziBUQ_IN@IBQUEHI611gd7" ' Pago l'2( Y o e .i J 1, i t; j -QUESTION-~ 6 13, '(2 50) 's ,1 W With regard to the Reactor. Protection System?(RPS)!

c,1 a.

You, as Shif t Supervisor r are approached by E th'e Electricalj Foreman'and. asked to transfer the RPS bus"'Ato?its1 RESERVE 1' POKER'EUPPLY.inLorder'to pictorm preventatiysimaintenar.ce;onithe t MG set. ;The' r eactor - is. oper ating: stL FS% hyaw'er : with a. HALF-SCRAM, ? signa 1Lon RPS Ch9nnel?'Bdue to an inoperable: Main Ste'am'Line' Radiation Monitor.. STATE if you:should.: authorize-the transfer'., (l 't' EXPLAIN the' answer. C; i, b. LIST three.(3) conditions that'will'cause~the Electrical o ~ ProLective Assemblies :(EPA Breakers) to open.. 1 STATE,two'(2) reasoni why,'the'RPS (ESERVE, POWER ~is inthrlockedh d c. to prevent supplying both RPS' busses at the same. time.. l g. g j .s 'j i 1 OdESTION 6.14 (2 50) 'k With tegard to the Autosatic Depressurization'Systeme(ADN): {l s. . STATE.the AOS initiation b gic for.a leakfinside~khe'drywell.- l '4 'etpoints.)' [1.03 -(Be specific and include. all s .j b.- STATE the. ADS. initiation'lasic'for alleak outsidethe"drywell'. 1 (Be specific and-inci,ude all setpoints. ) JLC1;^2 onhI j c: LIST three condition (s)/ action (s) that reset the' ADS tioer afterL l it has started. [0.75] j 1 1 1 J l QUESTION 6.15 (1.50) i With regard to the OFF-GAS Systsm!. 1 a. STATE why the PFeheaters use st'eam instead of electric heators. 1 i o i b. STATE the purpose of~the charcoel beds. 1 -s c. STATE how increased mnisture content. downstream of.the. moisture

l' i

sepe'rator will af fect tna chai coal bqds. o i. (***** CATEGORY 6-CONTINUED ON.NEXT PAGE *****Y l 1 l I r i t j ax -a.

s-41__ELBUI_STEIEdg_QgSIGNi_GQUIBQLt_6SQ_INSIBUdEHlellgy; pass 13 s. 1 QUESTION ~ 6.16 (1 50) i .The reactor is. operating at 95% power. The ReactorfVessel Select Switch-is selected.to Level "A.' The 'A' level instroment failst 1 SELECT'TRut or FALSE for each of the following statements'(assume no ) operator; action): a. -If-the failure is FULL UPSCALE, the deman'd level from the controller is'60 inches. b. If the failure is FULLLDOWNSCALE, Reactor. Feed: Pump (RFP) l runout will be' prevented by the runout relay energizing at 5.6E+6 lbm/hr per puinp. c. If the failure is FULL DOWNSCALE, the RFP runout ~ relay closes the Feed Water Rosulating Bypass Valve, if open. j QUESTION 6 17 (2.00) INDICATE for each of t.he below listed components / areas if it is protect'ed a by either the CARDOX SYSTEM or HALON SUPPRESSION SYSTEMt a. Statidb~y Das Treatment Area b. Auxilisvy Electric Room c. Computer Room 6. Excitor Room e. Recirculation MG Sets i QUESTION 6.1S (1.00) With respect to the Recirculation Flow Control Syrtete a. STATE TWO (2) conditions that must be satisfied te increase recirculation p>jmps speed above ;nir41 mum. ' t o. 5"I b. EXPLAIN why opetaling prnesdures probitit recirculation pumps speeds below 28%. [0.EJ (**n** END OF CATEGOPY 6 *****) s_.---

' Zl__EB9GEQWBgS_:_N9Bd6L1_6989806L1_EdEBgENQX Page 14. ONQ_B6DIQLQQIQ$(_ggNIBg6 V QUESTION 7.01 (1.00) During power operations at low recirculation flows and hish power,-an I unnecessary reactor scram mey be caused by large APRM oscillations. Answer the following questions in accordance with DGP 1-1, UNIT 2(3) l NORMAL STARTUP: How large can an APRH 'osci11ation become before: operator action. ic a. required? b. What is the IMMEDIATE ACTION required by the operator for such an APRM oscillation? GUESTION 7.02 (1.50) Refueling operations are in" progress Ahen the Reactor Building; Fuel Poc1 Radiation alarm sounds. a. Specifically indicate where the personnel on the Refueling Floor are supposed to evacuate (i.e., location of assembly area).. CO.753 b. STATE when the personnel can leave the' assembly area. [0'.753 QUESTION 7.03 (3.00) Unit 2 is operatins in Economic Generation Control (EGC) mode at 90% power, one of the 'D' high pressure feedwater heaters spuriously. trips (i.e., feedwcter heater extraction steam valve closes). a. EXPLAIN the effect(s), with one of the 'D' heaters out of service,' on the main senerator and DESCRIBE any potential problem this transient could create involving the turbine. E1.03 b. STATE the IMMEDIATE OPERATOR ACTIONS when the feedwater temperature drops 30 destess F. E1.53 c. Procedure DOA 3500-2, ' Loss of Feedwater Heaters,' requires'the Unit be manually scrammed if feedwater temperature drops by ____________ degrees F from the normal steady state value.EO.53 (FILL IN BLANK with correct temperature) (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)-

'l Zh__EBQgEDyggS__HQBdegz_6pNQBd8(t_EMEBGENQX Pa32113 '680_B69I96991GBL_G9 BIB 96 l i l d QUESTION 7.04 (1-.50) 1 Given: Unit 2'is' operating in Economic Generation Control (EGC) mode at. ?l the 100% rod line and recirculation pumps,at'the 78% flowecontrol line.- j Recirculation pump 'A' trips.

q.;

STATE the IMMEDIATE OPERATOR ACTIONS per DOA'202-1, ' Recirculation Pump (s) Trip.' + i j -QUESTION 7.05 (1.50)- STATE the reason (s) for the following recirculation pump' start I l i n. i t a t i o ns. 3 a. If' loop temperature is 50 degrees F.below reactor temperature DD, NOT START the MG set. I b. If the reactor. vessel bottom drain temperature is more than H 140 degrees F below the reactor water temperature, DO NOT START l the MG set. j 1 QUESTION 7.06 (1.00) Given: 1) Reactor Shutdown 2) Both Recirculation Pumps are off 3) 75%' flow through two shutdown. cooling loops For the given conditions, Procedure DOP 100-2, Shutdown Cooling Mode of j Operations, requires reactor vessel-level be maintained greater than + 40\\ inches; EXPLAIN WHY. l S (.- QUESTION 7.07 (1.00) e i ) Procedure DOP 2300-3, 'HPCI SYSTEM MANUAL STARTUP,' requires.that the l Standby Gas Treatment System (SGTS) be operable prior to starting-the HPCI j System. STATE WHY this requirement is imposed. ] f

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1 l l i (***** CATEGORY 7 CONTINUED ON NEXT:PAGE *****) l I E J

ZI__EBggggUBES__dQBd6(t_6QUg8d6(t_gdE8GEUQ1{ Pass 16 BND_86glg(Qglg6(_ggSIBQL hs,, l l QUESTION 7.08 (0.50) .] STATE the value on the Isolation Condenser ' Vent Radiation Mcnitor that would require manual isolation of the Isolation Condenser, i OUESTION 7.09 (1.00) l l STATE the reason for limiting the time Transformer 22 is in parallel with j Unit 2 per procedure DOP 6100-1,' REMOVING TRANSFORMER 22 FROM SERVICE 1 WITH UNIT 2 OPERATING." l I QUESTION 7.10 (1.30) I i Answer each of the following radioactive monitoring questions either TRUE or FALSE. l l a. Lieta samma survey instrument shall be checked for proper response prior to the start of the job. I b. Beta-samma survey instrument shall be checked for proper response I at least every 8 hours while in use on a job. i c. When a Geiger-Mueller (GM) type instrument is used in lieu of en ion chamber instrument, to determine personnel dose equivalent rates, the GM instrument reading cannot be used directly. i 1 1 GUESTION 7.11 (1.50) Answer each of the following questions concerning procedure DOP 2000-38, RADI0 ACTIVE WASTE DISCHARGE TO RIVER, either TRUE or FALSE. a. Only one sample is needed to be taken before a radiological release can be made. b. The observed dischotse rate CANNOT be set at the authorized rate. l c. NO radiological release can be made with the unit in OPEN CYCLE OPERATION. (*r*** CATEGORY 7 CONTINUED ON NEXT PAGE *****)

Zl__EBggggUggg_:_U9Bd862_6kN9BdBL2_Ed5BGEdgy Page 17 BMD_B89I96991986_G9dIB96 QUESTION 7.12 (2.00) Indicate either TRUE/ FALSE for tne following questions concerning equipment status to be left in the control room by procedure EPIP 200-20, INTERIM CONTROL ROOM EVACUATION / SAFE SHUT 00WN, if the control room is needed to be evacuated. a. The scram must be reset. b. Place the Automatic Depressurization System (ADS) inhibit switch to INHIBIT Fasition. c. Turn Target Rock Safety / Relief Valves control switches to 0FF Position. d. Leave Main Steam Isolation Valves (MSIVs) open. QUESTION 7.13 (1.00) 00A 040-2, ' LOCALIZED FLOODING IN PLANT,' lists the tiip of the Main Circulating Water Pump as an automatic action. EXPLAIN why the trip may occur. (Disregard possible electrical fault) (INCLUDE setpoint(s)). QUESTION 7.'14 (1.50) The Unit 2 NSO reports, to you as Shift Supervisor, that the generator is carrying less load. The Unit is stable and no alarms have annunciated. While verifying other instruments, indicated core flow is found to be higher than normal for the power level. a. What failure has occurred? b. EXPLAIN what caused the change in indicated: (1). core flow (2). generator output. [ (xxxx* CATEGORY 7 CONTINUED ON NEXT PAGE *****)

77 PROCEDURES'- NORMAL, ABNORMAL,' EMERGENGI Pasa 18, 6ND,,86QIg(gglg6(_ ggt!!8g( QUESTION 7.15 (2 50). During power. operation, the Unit #? NSO informs you-that the running RBCCW pump has tripped and the standby pump cannot be started. According to DOA 3700-1 " Loss of Cooling by Reactor Closed Cooling Water System'. a. STATE THREE (3) immediate operator actions. . (1,5) b STATE the PRIMARY concern if RBCCW is lost. (1.0)- (Does not involve individual components.) 00ESTION 7.16 (1.25) Personnel exposures under emergency conditions references dose limits;for life saving actions and less urgent actions. What are the dose limits for the following under those conditions? a. Whole body b. Extremities 1 c. An acute whole body dose equivalent in excess of _____ rem shall be limited to once in a lifetime. l GUESTION 7.17 (1.00) -s The Dresden Emergency Operating Procedures (DEOPs) require verifying / establishing LPCI flow through the heat exchangers as soon as, possible after an accident. Briefly EXPLAIN the basis / reason for this requirement. QUESTION 7.18 (0.50) l FILL IN THE BLANK. Before the reactor pressure vessel (RPV) is flooded i using DEOP 400-3, 'RPV FLOODING,' it should be.__________________ to increase the number of system available and decrease the stresses on the safety relief valves (SRVs) and associated discharge piping. (***** END OF CATEGORY 7 *****) l l

El__69dINIEIB6IIME_EBQGEQVBEEt_GQNQIIIQNEL Pago 19 689_LIMII6IIQNS QUESTION 8 01 (1.50) Technical Specifications list four criteria used for the design of the Core Spray and LPCI mode of RHR to assure that adequateLemergency cooling; capability is available. LIST THREE of these four criteria. QUESTION 8.02 (1 00) l Technical Specifications require a second independent operator ~when the Rod Worth Minimi ct (RUM) is inoperable in the run mode below 20% thermal power. What does this independent operator specifically verify.and when? QUESTION 8.03 (1.00) { In accordance with Procedure DAP 3-5, '0UT-OF-SERVICE AND PERSONNEL PROTECTION CARDS," special authorization must be obtained to remove certain equipment from service. STATE the additional signature, by titler; that is needed to remove the following equipment from service. a. a LPCI Pump b. CARDOX Systems GUESTION 8.04 (1 00) SELECT either TRUE or FALSE for each of the following questions concerning DAP 7-2, ' CONDUCT OF SHIFT OPERATIONS': s. The Shift Supervisor is the only watch standing member that has the authority to remove personnel from the control room during routine operations. b. The Unit Operator of either unit can leave his designated operating area, if he determines that his help would mitigate ~an emergency. (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)

g y i al__eDdIBISIBoIIME ESQQ5QL!BEQ1.QQNQIIIQNSI . Pass.20 BND_LIMIIeII9dB s O ;)) l I.O OUESTION 8.05 ( E v etr7 P - l 1 i SELECT either TRUE or FALSE for each of the following questions concerning j DAP 7-3, ' OPERATING ORDERS.', s. Daily Orders cannot supersede en appr.oved Operating Order. l l b. Daily Orders can be extend?d for more than 24 hours. pu. A, Crwi eiH a Cides o : : t o ; ;V' -e*H-Y-; 4 ir;* 7i: '6} 50Pt"E 2 tt**-- j p its initial i;=.::. e-+.7 'Each Shift Engineer will ter.eive a letter notifying them of a new Operating Order from the Pro'cedore Manager. 0

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i QUESTION B.06 [(1.50) ) SpecificationkT5) Limitir.s Conditions for ' Operation A number of Technical (LCOs) require the reactor to be in.s Cold' Shutdown Condition'if the operability of the system cannot be restored. COLD SHUTDOWN in accordance with Technical Specifications means (FILL IN THE BLANKS): } The reactor mode switch is in the< (a)._________,,,_____,_ position, no (b).__________________________ are\\bains performed,.and reactor coolant temperature is equal to or'less than s (c)+_____________ degrees F. I l I l .p } \\ (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)- 't ] \\ s \\ +

c y -- 8 + 1 . y f,f .e c ~ at. 2951NISTRA1.gi_PBgggQUREf t_QQNQIIIQNgt Pass 21 i W 9_LIMII6119dB l ^ 1 I'- ] i i 'J 4 ] QUESTION B.07 (2 50) y s l o SELECT.either TRUE or FALSE for each of the following quest:.ons concerning Dresden Generating Station Emergency Plan (GSEP)! I L The Station Director can delegat'e responsibility'for the a. classification of Emergency Action Level. b. The Station Director caN' delegate responsibility for' recommending protective actions for the offsite public. l ll The Shift Engineer would contact the Operations Dqd' Supervisor / c. i to obtain off-duty personnel to augment the.onsite' emergency [ organization. TheSt;tionDirectgv'iscitimatelyresponsiblefor[theaccountability-d. of r,i.te per sothel, sc A i Radiation Monit, ors (e.g., for main stacks and reactor building vents) e. can be used to determine actual radiological release during an accident. t V s q' GUCCTIO" 3.^" '1 00' h (h ,,e ST^TE the :er-e :e*;enty ra^d4 + 4 nn f e i +h=+ + h n Shi.f L M "c a " 4

0 " 45 b '

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QUESTION B.09 (1.20) 1 FILL IN THE BLANKS In accordance with DAP 12-4, ' CONTROL. OF llIGH RfDIATION AREAS,' the (a).________________ or (b).________________ hust M 9n the High Radiation Area Key Use Los within (c).__,_____ hour (s) ffn21owing shift change. as QUESTION 8.10 (1.00) l If's hi3h rediation area CANNOT be maintained LOCMI6 py9 DAP 12-4, t0NTROL OF HIGH RADIATION, AREAS,' STATE the Shi#t EdJ oeer's required l f J L action (s). /; / { i i.f ) I ./ ) \\ I r 1 (***** CATECOM B CONTINUED ON NENT PAGE *****) \\ ?) y; 1 1 .6 j L___-__ _ - -_-_______-____ N -___-__-_____-____-__ _ __________ Y f

8. ADMINISTRATIVE PROCEDURES, CONDITIQUEt Pagn 22 GU9_LIMII6119BS J QUESTION 8.11 (1.00) STATE the principal concern,of Operating Urder 49-87, ' INACTIVE SYSTEM STARTUP,' prior to placing any inactive system in service that has been isolated for more then one month. l QUESTION 8.12 (1.00) l As stated in the CECO Radiation Protection Standards, the use of the word ' DANGER" on radiological control signs has a unique meaning. CHOOSE the correct meaning from the following responses. a, The area is a high radiation area. i b. Airborne radiation is present. c. The area radiation rjetector is inoperable. d. The area is radiologically contaminated. ) DUESTION 8.13 (1.00) STATE the TWO (2) conditions that require establishing primary containment integrity with fuel in the reactor vessel per Technical Specification. QUESTION 8.14 (1.00) 1 On August 16, 1985, Dresden Unit 2 experienced an event where a design deficiency in an -ruxiliary power auto transfer circuit resulted in a failure to transfer buses (4 KV Bus 22 and 4 KV Bus 24) to the unit auxiliary transformer. Operating Order #14-87 was issued to provide an interim method of mitigating the consequences of this event. Briefly EXPLAIN the guidance provided by Operating Order (14-87. QUESTION 8 15 (1.00) Due to the number of motor failures on limitorque valve operators, guidelines are provided by Operating Order (19-87. Briefly EXPLAIN the guidance provided by this operating order to help minih.ize valve motor failures. (***** CATEGORY S CONTINUED ON NEXT PAGE *****)

B '. ADMINISTRATIVE PROCEDURES, CONDITIONS, Page 23 ] J 689_LIMII6110NS 1 l 60 (2.?5'epr GUESTION B.16 -ev-Technical Specifications require that recirculation pump speeds be -J maintained within 10% of each other'above 80% reactor power 1and within 15% of each other at less than 80% power. The bases.in Technical Specifications provide two ressens for the flow-mismatch restrictions. Briefly STATE these two reasons. (1 5) I b. Oper: tier uith cr:: ecirculatier leer is limited te ?^ Sevat by )W l l Technicial Opacification. Driefly ST^TE th:

Or provided inthwF^-)

Technicel C acificatica 0;;i; why ;u;tsir.ed CFer: tier with en 100F-fs F 65 c.whibited, '0.75 y s QUESTION 8.17 (3.00) y The Station Emergency Plan Implementing Procedure _(EPIP) 200-4 outlines each individual responsibilities in the event of a major plant fire. a. LIST the THREE (3). responsibilities of the Shift Fnsineer. I b. LIST THREE (3) of the responsibilities of the on-site Fire Chief (in addition to implementing fire drills). 1 i i L l OUESTION 8.18 (2.50) Following an Emergency Declaration, the state and local governments must l l be notified within 15 minutes using the Nuclear Accident Reporting System l (NARS), as listed in EPIP 300-1, ' INITIAL NOTIFICATIONS AND GSEP i RESPONSES.' The NARS reporting form requires six specific items of Information to be reported. 1 LIST FIVE (5) of the six items. ps i a i j GUESTION 8 19 (0.50) FILL IN THE BLANK. DEOP 300, ' SECONDARY CONTAINMENT / RADIATIVE RELEASE CONTROL,' provides authorization to bypass the reactor building ventilation system interlocks, however, the _______________________ interlocks must never be bypassed. (MORE THAN ONE WORD IS REQUIRED) 1 (***** END OF CATEGORY 8

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(********** END OF EXAMINATION **********)

5$ THEORY OF NUCLEAR POWER PLANT OPERATION, Page 24' E(Q1DgtBBQ ld[BdQQ186dICS .I ) i ANSWER 5.01 (1.50) a. LHGR is to prevent fuel clad cracking due to high stress. (0 5)- b. APLHGR is to prevent gross cladding failure due.to a' lack of cooling following a Loss of. Coolant Accident-(LOCA).- -(0.5) CPR is to prevent fuel clad ~ cracking due to inadequate' cooling of l c. the clad. (0.5) l i REFERENCE GE Thermodynamics Heat Transfer and. Fluid Flow (BWR), pg. 9-124 ) 293009K120 293009K112 293009K100 ..(KA's) -j i ~ ) ANSWER 5.02 (1.00) Centrifugal pumps are provided with high point easing vents which allow the pumps to be primed. ti.03 (Other reasonable answers will be accepted) ' REFERENCE GE Thermodynamics Heat Transfer and Fluid Flow (BWR), pg. 7-124 293006K112 ..(KA's) ANSWER 5.03 (2.25) s. ' Conduction heat transfer occurs as 9 flow of heat through a body, or between bodies in direct contset' EO.53. Cladding surface to coolant to.253. b. ' Radiation heat transfer occurs when energy is. exchanged between bodies through an intervening space by electromagnet waves' CO.53. Cladding surface with no cooling during a LOCA CO.253. c. ' Convection is the transfer of heat between a fluid and a surface by the circulation or mixing of the fluid' CO.53. Flow throvsh reactor vessel CO.253. Correct examples not stated will be considered. (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

.,c 5.__THEQRy_gE_NMGLEAR POWER PLANT OPERATION, _Posa 25 E' UIps egg _IggggggygegIgs z REFERENCE. HGE Thermodynamics Heat Transfer and Fluid Flow-(BWR), pg. 8-2 293007K101 ..(KA's). ANSWER "5.04 (1.00) Delayed neutrons are more likely to be thermalized than prompt neutrons'. REFERENCE Nuclear Energy Training, Book.3, pgs. 5.3-2, 5.3-3, 5.4-6, and 5 4-7.. 292003K106 ..(KA's) ANSWER 5.05 (1.00) Xe-135 is removed from the reactor byt 1. Decaying to Cs-135 2. Absorbing neutrons to form Xe-136 (2 0 0.5 each =1 0) REFERENCE { Nuclea? Energy Training, Book 3, pg. 10.2-1 292006K104 ..(KA's) ANSWER 5.06 (1.00) The BWR provides negative feedback to limit the transient due'to immediate i actionofthE[goppler) Coefficient. E1.03 ~ REFERENCE Nuclear Energy Training, Book 3, pg. 13 5-3 292008K122 ..(KA's) (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****).

1 5 THEORY'0F NUCLEAR POWER PLANT OPERATION, Pose 26 E691gstegg_IBEBbgg16edIgg y ~ 1 ANSWER 5.07 (1.50) a. P = P(0) exp (t/ period); and solve for period with P = 2P(0), j because doubling time is providede j Period = t/In(2) =~1.445 x t E0.253 1.445-x 40 E0.253 j = 58 +/ :2 sec. E0.253 = b. P = P(0) exp (t/ period) 1 0.08 = 0.002 exp (t/58)- EO.253 ] l Solve for timer t: 3.7 = t/58 E0.25] j t = 214 +/- 2 see EO.25] REFERENCE Dresden Requal Exambank, Question No. 12-020304, ps. 3-4, Reactor Theory 2 'l I 292003K108 ..(KA's) l ANSWER 5.08 (1.50) a. 7 6. 1 c. 9 d. 8 e. 2 j f. 6 I i (0.25 pts each) I I REFERENCE i Dresden Requal Exambank, Question No. 12-020276, pg. 9, Reactor Theory 1 l 292003K104 292002K110 292002K107 ..(KA's) i i ANSWER 5.09 (1.50) a. The thermal output after a reactor trip is from decay of' fission fragments and activated vessel structures by beta and samma decay. ,_ a,, to.vG g to 753-b. The IRM's snd SRM's^are compensated to discriminate for sammas. (The IRM instruments doinot respond because of the level and energy of the sammas). E0.753 ' -.y-um L--u w Ar. % s ,, - a :) nf., g g ~.-... -, (***** CATEGORY 5 CONTINUED ON NEXT PAGE

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~ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, Paga 27 FLUIDS,AND THERMODYNAMICS ~ ~ REFERENCE Dresden Requal Exambank, Question No. 12-020267, pgs. 4-5, l Reactor Theory 1 215003G005 292008K129 ..(KA's) l i l ANSWER 5.10 (1.50) l I The failure is due to stress corrosion crackinD E0.53 caused by high l localized pellet / clad thermal expansion EO.53 and the presence of l embrittling fission products (iodine and eeri" ' C0.53. CBdmium j i REFERENCE Y 1 GE Thermodynamics, Heat Transfer, and Fluid Flow (BWR), pg. 9-124. .j 293009K131 ..(KA's) I 1 l ANSWER 5.11 ( 1 0 ) l l 100% Rod Line T.O.53 with minimum recirculation flow E0.53. ) l 1 REFERENCE l Dresden Requal Exambank, Sec. 1, Duestion No. 12-020345, pg. 17 1 f 292000K102 ..(KA's) 1 ANSWER 5.12 (1.00) b. REFERENCE ) l Dresden Requal Exambank, Sec. 1, pg. 6, Question No. 12-020436 293006K110 ..(KA's) ANSWER 5.13 (1.00) a. - due to build up of Pu-240. [0.53 b. - due to reduction of self shielding as fuel temperature increases. E0.53 a< g c J b%dwo% l F WH n Aut:MA Q /tL5 W n bW 1 (***** CATEGORY 5 CONTINUED ON NEXT PAGE

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t: 5. ' THEORY OF NUCLEAR PQHgB_PLBNT_QEER6TIQHL Pasa 29 ELVIDE16SD IBEBBQQXBob1GL REFERENCE Book 3, E f Reactor Operation, timi 11.4.4, pg'11.4-4 for Part (a); and Dresden Requal Exambank,8 2-2,lfor Part.(b). Cectier, 0.2,-pas 8 2-1 4 l Section 1,-Question No. 12-020381, pg122. 292004K109 292004H108 ..(KA's)< 4 ANSHER 5.14 (1 00) (Heat generated by the pump will cause). cavitation E0.53 and pump internal-damage (by insufficient-cooling) E0.53 (1 0). Other correct answers will be accepted. j .y REFERENCE i GE Thermodynamics, Heat Transfer and Fluid Flow, pg 7-123. 291004K104 ..(KA's) ANSWER 5.15 (1.00) ] The grcup of delayed neutrons with the longest' half-life (Group 1, which is mainly Br-87) is controlling the rate of neutron production, j REFERENCE Reactor Operation, Book 3,'Section 14, pgs 14 2-4 to.14.2-6. 292008K125 ..(KA's) 'l 'I a l ANSWER 5.16 (2.00) i 3 a. Decrease j b. Decrease c. Increase d. Increase E4 items.each worth 0.5 point for a total of 2.03 REFERENCE Reactor Operations, Book 3, pgs 9.5-1 and 9.5-2. 292005K104 ..(KA's) (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

7'l 5$__INEQBI_QE_HQQLg68_EQS[Q_E(6BI_QEEBBIIQNr Pasa 29 j ELVIDSteHD_INEBdQQ1HedIGE i I

l j

ANSWER 5.17

1.50)

] a. 'Let B = beta effective t = reactor period E=3 second P = reactivity. insertion.E=3 delta K/K u = radioactive decay constant-(or lamda) = 0.1 per'second then t ='(B-p)/ ump At BOL: .] (0.0071 - 0.001)/0 1r0.001'm 61 T+/

2. seconds

'EO.333 4 t(BOL) = ] At EOL: 45 +/- 2' seconds [0.331 I (0.0055-0.001')/0.1*0 001 t(EOL) = = 61 - 45 = 16 +/- 4 seconds CO.33] Change = b. Build-up of pu-239 coupled with the burn-out of U-235 causes s' decrease in the effective delayed neutron. fraction-(Deff). -(0.5) j REFERENCE Reactor Operation, Book 3, pgs 11.4-3 and 11.4-4. 292003K106 ..(KA's) I ANSWER 5.18 (2.50) Lon3er (0.5) Because the power down transient is limited by the decay of the lens-lived delayed neutron precursors (0 5) whereas, the-power up 'j transient is dependent on prompt neutrons (0.25) and short-lived delay J neutron precursors. (0.25) The down transient is stabilized by the void coefficient.(alpha V) (0.25) and the decrease in doppler (alpha D) (0.25). The up transient is stabilized by the void coefficient (alpha V) (0.25) and the increase in doppler (alpha D) (0.25) j (Will accept and reasonable, correct discussion of the effect that. doppler and void coefficient have on stabilizing power.) l REFERENCE Dresden Lesson plan, Reactor Physics Review, pp. 21, 29, 32 292008K119 292003K106 ..(KA's) j 1 1 i l (***h* END OF CATEGORY 5 *****) { } i

v ip__P(681_gISIgdS_QESISHt_QQUIBght_66Q_INSIBgdgNI611gM .Page 30. ij 1 ' ANSWER 6.01. (1.00) j 1. To limit the dynamic water level variation in the torus vent pipes (e.g., to 2 ft. or less) during the postulated maximum credible -accident. [0.53 1 2. To prevent exceeding the design external pressures of the primary containment. CO.5J REFERENCE LPB No. 'J, Containment Systems, Section 15) Rev. 18, pg. 15 l 223001K501 ..(KA's) ANSWER 6 02 (1.00) 1 l The level in the hotwell would decrease because the makeup valve would close CO.UJ and the reject vhive.would open E0 5]. REFERENCE 1 J LPB No. 2, Instrument Air System, Section 3, pg. 14 256000K601 ..(KA's) j 4 ANSWER 6 03 (1.00) LOW-LOW reactor water level tr Q setpoint was chosen to bet -l l Vn Ic u G w y C o t Q f a. High enough to prevent spurious operation. -C^.5] Mf [,u.tD p Id Low enos sh^to prevent fuel cladding melting.(Alf[ EO.53 b. REFERENCE 4 Unit 3 Technical Specifications, Amendment No. 33, pg. B 3/4 2 l 212000G004 ..(KA's) I ANSWER 6.04 (1.00) I Below 30% power, the worst case withdrawal of a single control rod instrument rod block action will not violate the Minimum Critical Power Ratio (MCPR) fuel cladding integrity safety limit. (***** CATEGORY 6 CONTINUED ON NEXT PAGE

P:***)

41__EbeHI_HgIgdS_gggIgyz_ggNIEgLt_sug_INSIBgdgNIeIIg8 Poss'31 REFERENCE Unit 3 Technical Specifications, Amendment ~No. 87, pg. B 3/4.2-32 215002K403 ..(KA's). ANSWER 6.05 (1.50)' ~ a. False. b. True. c. False. REFERENCE DLP Book 3, Containment Systems, Section 15,.pgs. 47-48 223001K404 ..(KA*s) ANSWER 6.06 (1.00) Unit 2 has a jet pump riser vibration problem in. single -loop operatiori. REFERENCE DLP Book 1, Recirculation System, Section 15, Rev..17, ps. 18 202001K601 ..(KA's) ANSWER 6.07 (2 00) a. False b. False c. False d. True e. False (0.4 pts for each answer) REFERENCE DLP Book 2, Main Steam System, Section 1, Rev. 15, pas. 14 and 15 and Figure 10 239001A203 ..(KA's> (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

g[__PLANTSYSTEMSDESIGN,CONTR0(,ANDINSTRUMENTATIQN Paso 32

c l

ANSWER 6.08 (1.50) IRM Rod Block Rod Block Setpoint IRM Downscale 5/125 IRM High 100/125 Channel.A/B IRH 'Inop' condition bs"di"t Jm*" "Pa.; rgA4 bas 4 'g, gag _. ] 9 ,(0.25.pt for eacK/ rod block and for'esch M olnt ,,ggmf, e REFERNNCE3 <h ~Y0d. et00M m YA 6L I DLP Book 1, IRM System, Sec. 9, Rev. 14, p gf. 17 - /8. 215003K401 ..(KA's) { ANSWER 6.07 (1.00) ) 1. High drywall pressure (> 2 psis) CO.53 2. Low-Low Reactor Level (< -59') CO.53 REFERENCE DLP Book 3, LPCI, Sec. 12, Rev. 8, pgs. 14-15 1 219000K403 ..(KA's) j -1 I i I l 1 i ./ (xxxxx CATEGORY 6 CONTINUED ON NEXT PAGE xxxxx) 1

sh__PL6NI_@l@lgd@_QE@IQUt_GQUIBQLt_689_INSIBQdEMIBIIQB Paga 33 ANSWER 6.10 (2.00) a. Open E0.53 Justification: The output (e.g., a+ 10% signal increase) from this, summing junction will cause the recirculatiore pump speed to increase, thus resulting in higher core ficw and power C0.53. The power. increase causes reactor prassure.to increase, thus generating a larger pressure' error'CO.53. As the pressure error increases,.the sain unit outpuk callo'for more turbine steam flow and the control valve. operas wider [0.53. Grader's Note: The pressure /lcad low valve gate'(LVG) does not pass the increased load signal. REFERENCE DLP Bcok 2, EHC Pressure Control and Logic System, Section 8, Rev. 9, pgs. 10 and 11, and Figure 5 245000K303 ..(KA's) 9.70 ANSNER 6.11 (2.001 9-a. All 4KV breakers would fail 'as is.' CO.53 l b. All the ECCS would be disabled (because the ECCS logic is-DC powered and ' energize to function.' ECCS co21iponents would not start automatically and could not be started manually from.the control room). [0.5] ) l c. ADS logic would not function (nor would its solenoid valves). [0.53. 1 d. 30 1/2- :ald bc ptrable -(br::v:: it: centr:1 pru r i: :Up;?14 d b'/ J-either the Unit 2 er 'h: L'n i t 2 S : t t r y b y :.: n: O f e n ::.:t s - te*crofep hw t kii). [0.53 p REFERENCE DLP Boot 2, Emers PWR-D/G'and Station Batt., Section 13, Rev. 13, pg. 16 218000K606 264000K609 262001K601 ..(KA's) (x*n** CATEGORY 6 00t4TINUED ON NEXT PAGE *****)

f[i_ _P L 6 HI_ S I @IE N S _ Q g @lG N r _ Q Q UIB Q L t _6 N Q _IN EIBM d E NI61198Passe 34 ANSWER 6.12-(1.50) a. possible leak in the tube bundle CO.53. b(1).' high steam line inlet temperatus-e (2). high-shell water temperature (3). increasing shell; water temperature (4). increasing local radiation levels ( 5.). steam or fos coming out the side of..the Reactor Blds. (6). ' unexplained increaseLin-Icolation Condenser shell si4Je water. level. oint each3. d. d%. Canv 4 at 0.25 Q w:6m w (-f), y REFERENCE LPB No. 3, Section 14r and DOA.1300-1, 207000K405 ..(KA's) ANSWER 6.13 (2.50) ,,,. _ g < j e, ,,, g _, y 4 -howertotheRPSBus 'A' will be interupted during the n a. M;.44w CO.253 transfer and with a HALF-SCRAM on RPS Channci

B' a full scram will occur.

CO.53 b(1). overvoltage CO.253 b(2). undervoltage CO.253 b(3). underfrequency E0.253 l l c. To prevent violation of the RPS design that a single failure.cannot cause or prevent a scram CD.53 and to prevent overloading the reserve instrument transformer E0,53 REFERENCE LPB No-. 1, Section 14. 212000K404 212000K403 212000K402 212000N401: 212000K104 ..(KA's) 1 l l l l' (***** CATEGORY 6 CONTINUED OH NEXT PAGE

- - - - )

6 __P(BN1_$1@ led @_QEEI@di_G9818961_689_INE B9dEdlBTIQN-PageJ35- ' ANSWER 6.14 (2.50) Drydell pressure +2 psis CO,253 : and r eactor level less' than -59, a. inches E0.253 and 120 second timer timed'out.[0.253 and any Core Spray or LPCI pump running with-greater than 100 pcis disciierge pressure. [0.253 b. Reactor level less than -59 inches.CO.25? an'd 845: minute-timer: timed. out E0.253 and any Core Spray or -LPCI pump running lwith greater thare-100 psis discharge pressure. [0.253 c(1) loss of power CO.253 c(2) depressing the timer reset button E0.253 l c(3) reactor' level increasing to greater than '59 inches E0.253 REFERENCE LPB No. 3, Section 10,'pss 5, 6, and 7. 218000K403 218000K103 218000K102 218000K101 218000K201 ..(KA's) ANSWER 6 15 (1.50) a. To eliminate an explosion hazard E0.53 b. Adsorbs radioactive seses to allow the gases to decay before exhausting to the main stack E0.53 c. Moisture reduces the adsorption efficiency of the charcoal bed E0.53 1 l REFERENCE l LPB No. 2, Section 3, pas 4 & 5. 271000K404 271000K301 271000G007 ..(KA's) l l f l ANSWER 6.16 (1.50) a. Ff,LSE E0.53 b. FALSE E0.53 c. TRUE E0.53 REFERENCE LPB No. 2, Section 5, pgs 9, 10, 11, and 13. 259002K302 259002K301 259002G007 ..(KA's) (***** CATEGORY 6 CONTINUED ON NEXT'PAGE'*****) _____________d

i 'sh__PL6HI_SIgIEHg_QESIGHt_CgNIBgli_68Q_INSIBgNENI611gN. Paga 36-i ~ ANSWEP 6+17-(2.00) l 1 C ARDOX l ( Cdd* ###) - a. HALON... b. c. HALON-l 1 d. CARDDX e.~ CARD 0X .j -E0.4 points'for each item] l 4 i REFERENCE >1 l LPB No. 4, Section 5, Figures 1 and 7.. 294001K116 ..(KA's) l ANSWER 6.18 (1 00) .j j a. (1). Recirculation pump discharge. valve open.- E0.253 i] (2). Feed flow is above 20% CO.253 ~ b. Limits the possible thermal stress problems in theilower head region. j E0.53 J REFERENCE i ,/ L P P N o. 1, Section 16, Pages 5 and.13. i j 20iOO2K402 202002G005 ..(KA's) lc 1 ) I i l ld l 1 (**xxx END OF CATEGORY 6 xxxxx)-

U ZA__PBQGEQQBES_;_NQBb6(1_6BBQBdGLi_Edg5GENGX Pasa137 669_B6919699196k_9981896 J I J

AN5kER 7.01 (1.00) l a.

15 +/- 5%, peak-to-peak APRM ose'illation ~EO.53 l be INSERT control rods [0/53 l, or INCREASE core flow E0 253 at the direction 1of a' Qualified Nuclear Engineer E 0. 2 5 3 -. REFERENCE. 1 ] r. DGP 1-1, Unit 2(3) NORMAL ST6RTUP, Rsv. 25,'Section.6.1, ps. 5 j 215005K101 ..(KA's? !l ANSWER 7 02 (1.50) a. On the stound of the' Unit 2/3 Turbine Building E0.53 by the Unit 2 feedpumps E0.253. be Notified by Radiation Protection that it is safe to: leave.: E0~753 i REFERENCE EPIP 200-3,.ps. 2 295022K301 ' ..(KA's) 1 l 5 i i =l l q i i i i l (***** CATEGORY 7 CONTINUED ON NEXT PAGE xxxx*) ____ u._

ij Z1__E899E998ES_ _d98d9b1_&989806L1_gMEBGEMGX.

Pass 38-i 609_889I9609IG66_C931896 l

AN'6WER 7.03 .(3.00) I a. 1. Causes an increase in KH output because the steam:which?was q formerly extractsd now passes.through the: turbine.- -EO,53 q 2. This will overload the low pressure turbine stages because if-l 4 the larger d/p across them. C O.g5 gj b. 16 Reniove the unit from EGC1 operation, r0.53 1 1 8 2. Reduce core flow approximately 9 mlb/hr (i.e., 3 mlb/hr for-every 10 degrees F that feedwater temperature drops). .E0.52 I i 3. Insert the. CRAM arrays.per ths. Control ~ Rod Segoence Package to l dicar the Average Power Range Monitor Hi's. !EO.53' L c., 130 +/- 10 degrees F. E0.53 l REFERENCE Dresden Requal Exambank, Section 3, Question No. 12-020510,- Lesson Plan BLS00063 for CONDENSATE AND FEZDWATER, ps. 4 of 10/06/06' for Part' b L f" Oooll.2 OrbvkN A f"A N "2 i& N P i L DOA 3500-2, LOSS OF F'EEDWATER KEATERS, Rev. 4, Section C.2, pgs. 1 and 2 for Parts (b) and (c) i l 295014K301 275014A106 ..(KA's) 1 I ANSWER 7.04 (1.50) it 1. With the unit in EGC 1 'I a. Trip EGC E0.253 b. Transfer the Master Recirculation Flow Control to Manual E0.25] i 1 2. Reduce the running Recirculation Pump speed-to less than'43%'of rated I speed. [0.253 3. CLOSE the Recire Pump Discharge Valve (s) E0.253 MO-202-5A(B) and I MD-202-7A(B). l 4. Monitor the reactor water level E0.13, reactor pressu're CO*13, core i power E0.13, and off gas activity 00.133 Verify the; indications, if possible, by using more than one type of indication E0.13. i (***** CATEGORY 7 CONTINUED OH NEXT PAGE *****)- I

Z,$-_ES99EDUBES_:_N9850kt_0009Bdeb1_EdEBGENGX Page 39 680_B091960 GIG 66.C9f!IB9b REFERENCE DGA 202-1, Rev., 4, Section C, pgs. 1 and 2 Requel Exambank, Section 4, Question No. 12-020578, pg. 3 202002G001 ..(KA's) ANSWER 7.05 (1.50) a. 1. Prevent thermal stress of pump casing. CO.53 2. Limits cold water reactivity addition effect. [0 53 L b. Prevent thermal stresg of Control Rod Drive (CRD) stud tube welds and vessel skirt welds. CO.53 REFERENCE DOP 202-1, REACTOR RECIRCULATION SYSTEM STARTUP, Rev. 6, Sections E.3 and E.4, pg. 4 i Requal Exambank, Section 4, Question No. 12-020552, pg. 1 1 202002G001 ..(KA's) ANSWER 7.06 (1.00) To provide a natural circulation flow path E0.753 through the moisture separators E0.25]. REFERENCE Requal Exambank, Section 4, Question No. 12-020552, pg. 1 259002G001 ..(KA's) l l, ANSWER 7.07 (1.00) The HPCI System gland exhauster discharges to the SBGTS. REFERENCE l Requa3 Exambank, Section 4, Question No. 12-020564, pg. 5 206000G001 ..(KA's) (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) L -_ _ _

-I aj Z. PR0ggQWBES:- NQBtj6bz_BEUQBMBLt.,EdGBGENgY Page 40' ed9_BeQI960GIGok_G9 NIB 06 ANCWER 7.08 (0.50) 20 +/- 4 E0 253 mi lirem per hour.CO.253-MNb f J REFERENCE DOA 1300-1 ; /F]Mai-4, JSoc+7784 N h ScT 7d/M Lt* .l DM 1 DOP 1300-2,guTOMATIC OPERATION OF ISOLATION CONDENSE I Section E.3., pg. ' 1 - a 207000K405 ..(KA's) y C)g (5)_3 6 3, 0soisi.s 1., $$cikh*>J Wbth)StC AW H ANSWER 7 09 (1.00) (g 7,). jj ) Overloading of Transformer 22 due to circulating ' curt ent could occuv. and. trip its breaker ( nMt4 N /'dAteL ) l i REFERENCE DOP 6100-1, Rev. 2, Section D.2, pg. 1, REMOVING TRANSFORMER 27 FROM SERVICE WITH UNIT 2 OPERATING Requal Exambank, Section.4', Guestion No. 12-020570, pg. 6 1 262001G005 ..(KA's) i . ANSWER 7.10 (1 50) l a. True i b. Felse j c. True ~ REFERENCE Radiation Protection Standards, CECO, September 1985, Monitoring, pg. 38 294001K103 ..(KA's) l ANSWER 7.11 (1.50) 'l c a. False E0.53 b. True E0.53 c. False CO.53 (**x** CATEGORY 7 CONTINUED ON NEXT PAGE *****). -)

m. u, s ( j ABHOPMAL, EMERggdQL Paga141: 7d__ERgQEQURES'-'HgRMAL, ' { 'eHQ_BoDI9699ItoL_G9EIBOL i' REFERENCE 00F 2000-28, RADI0 ACTIVE WASTE DISCHARG5 TO.P.IVER, Rev'. 7,1 Sections E.& F.- 294001K103 ..(KA's) i

l ANSWER-7.12

-(2.00) o a. False b. True c. True d. Fkise 1 REFERENCE i j F. PIP 200-20, INTERIM CONTROL ROOM EVACUATION / SAFE'EMUTDOWN, Rev. 3, = Section F, pgs. 21and 3 295016K303 ..(KA's) \\ ANSWER 7 13 (1 00) i I tireviating. Water Pumps trip with a condenser pit high high water lovel at j ). 5 ft above condenser pit floor [1.03 l 1 I REFERENCE l Dresden Requal Exambank, Section 4, Guestion No. 12-020576, pg. 9. i 00A 040-3, Rev. 2, Sections B and E. i 295002K208 ..(ttA's) 1 s - I l ANSWER 7.14 (1.50) 0 a. Jet pump failure [0.53 b. (1). Reverse flow thrcush the failed jet remp-causes indicated' core flow to increase. [0.53 (2). With reduced power,' turbine control valves close to maintain header pressure thus reducing generator load.- [0.53 REFERENCE DOA 201-1 295001K302 202001K601 202001A201 (KA's) (***** CATEGORY 7 CONTINUED ON NEXT PAGE xmmxx) i --___-.__a

f .[ l I ) Z. PROGEEURgg_n,$g8dALL G@NQLdAl _Edg8Qg8G1. Paga 42-t -6dE_66QI9kEGIS8k_QQBIC06 .ANGWE 7.15 (2 503 i a. (1). Atte.mpt to restore RBCCW flou.. 'A' and i (2). If RBCCW flow is lost end.carmot be restored trip the

B' Recirculation Pumps.

3 (3). Manuvily scram the reactor, if. equipment. damage appears-i imminent. (4). If drywell pressiste To4ches +2Lpsig," verify reactor scrami and j i EHTER DEOP 100 and 240 series procedures.' Erny thtee at 0.5 each) l 'l l .b. Luss'of cooling to the drywell causes drywell pressure to increase J l causing a reactor s : r e m a n d E C Cf i n i t i a t i o n a t + 2 p/c-hi7~; sis. E1.03. 1 CVd Aly ci: 7W nt/vMr4 Fdt Fuc.L. C I (*9-). A'#h' L 'M igj 49 Y Al"'d M M- $g ^ REFERFNCE W \\ 497II+2- ) (<' d) /C< 15 ' g( K A .DOA 3700-2 ?t95018U202 295018G011 ) DLt' Arr 3, hehw coa,, '[ 7 T *' eHSWER 7.16 (1.25) a. Whole Bod'/ Life saving is 75 rem. [0.25] Less urgent actions are 25 rem. [0.253 b. Extremities _, Life saving is 200 rem.. CO.25] - )Less urgent actions is 100 rem. CO.253 c. 25 rem E0.25] REFERENCE Dresden Radiation Protection Standards, pg. 26. I ~ 294001K103 ..(KA's) i i ANSWER 7.17 -(1.00) The go,a1 is to preserve the heat capacity of the suppression pool by promptly removing core decay heat. l . REFERENCE DEOP Lerson Plan, Caution 12 295026G010 295026G007 ..(KA's) (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)

= - t , I -( 4. , -s 3

Z;.__BB9EEEMBEEi: _d9Bd661 6EdQBU661.5dEBGENG1 Pass 43

+... 689_B69196991G66_G9BIBQL-ANSWER 7.18 .(0.50) deprescurizo/ ] -i ' REFERENCE DEOP'. Lesson Plane DE0' 400-3. 2?:5031K205 (KA) s (***** END-0F CATEGORY 7 xxxxx) a_---2-.mu__s_2-_ m P

.8.- ' AD'AINISTR ATIVE. PR0p@QURE@f.,CgNQU,IgNEL l' asp:44... i g. 7;,

f g d.g

&~ ctut BNQ,(IdIIgIIQN@ _,ww Afwreh l / -p s:.A w< Aid &d" 4,,,, j ett4< '/v WM' -l g wa. g . A,a A, f -" e i ANSWE B.01 c2,50) t.FGcaledipted'peakcNMng-temperkte 1 o.1 s t, than 1. U acit

N g -

) 44%,h mM ~evi.( /, intact.U p p 270S.jecrees. F. 9h A ~ I 24. Assure the cci e geometry remains l j 3, Limit the core wide clad metal-water reaction to less than 1%'.' ( 4. Limit the calculated local, metal-water reaction to 'less-.than 17%.- V(3 9 Oe5 pts-each) l .2 "~ ~ f j h: vlM q%m~~ Cp wLy,,, ww l REFERENCE i Dresden Technical Speci tions, ps. B 3/4 5-31. 3 Iy.srtwr.1 I g g jy A fe,<J b gg ,2 203000G006 ..(KA's) cac swt yj au 3,,J It 5 ANSWER 8.02 (1.00) A second independent operator or engineer is needed to verify that all rod positions are correct E0.53 prior to commencing withdrawal of each rod group EO.53. 1 i REFERENCE Unit 3, TS 4.3.B.3.b., Amendment No. 75, pg. 3/4.3-8 01006G001 201006K301 ..(KA's) ] 'Dof 4m-2, 4+ f, f76 k d.c., g.r. l l ANSWER 8.03 (1.00) ) u i a. Operating Engineer of the affected unit E0.53 b. Station Fire Marshal E0 53 i REFERENCE l DAP 3-5, '0UT-OF-SERVICE AND PERSONNEL PROTECTION CARDS,' Rev. 16, Section B.1.c., pg. 3 294001K102 ..(KA's) l s l I 1 (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) i 1

)

. 1. l 8 ADMINISTRATIVE PROCEDURES, CD'NDITIONS, Pasa'45 j AND LIMITATIONS U ANSHER' 8.04 (1.00) i a. False 1 b. False REFERENCE q l 1 OAP.7-2,. 'CONDUCTEOF SHIFT CPERATIONS,' Rev. 13; Sections B.8.a(1),'ps. 4 1 and B.8.a.(3)(c)i pg. 5, t i:5Fectively.. 'l 194001AIS? t.M A ' s 5 <4 /,s o ANSWER 8.05 (2.007 7 a. True b. ";1 a r * 'T2a6 c[dhF Trbe l s (Each answer.is 0.5 pt) l REFERENCE l DAP 7-3, OPERATING ORDERS, Rev. 5, Sections B.1.a; B.1.di B.2.d(2); and B.2.f(3). 294001M116 (KA's) i ANSWER 8.06 (1.50) (a). shutdown (b). core alterations (c). 212 a REFERENCE U r.i t 3 T S 1. 0. A A, Amendment No. 87, pg. 1.0-5. 20100XG002 (KA's) (***** CATEGORY 8 CONTINUED ON NEXT PAGE

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'8 EADMINISTRATIQg_PRQQEQQRggt_QQNQIIIQHSt. Pass'46 BHQ_bIdIIBII9SS ' ANSWER 8 07 (2.50). a. False b. False c. True d. False e. False C5 0 0.5 pts each] u REFERENCE Dresden GSEP, Rev. 4, pages DA 6-1, 6-2, 6-5, and 7-9 294001A116 ..(KA's) l g f ANSWER-8.08 (1.00) If the Shit ervisor determines that has been terminated by a' s f time, the Sh Supervisor can remove a temporary ) 'significant" p(ers S/t;aig eg,/) BP.,vi / JNfpr 3. yyjfes), alteration. REFERENCE DAP 7-4, CONTRO TEMPORARY SYSTEM ALTERATI Rev. 9, Section B pg. 1 294001K ..(KA's) ' Tag oF 11-/o47 s>,m us>J 7w wisey,we. ANSWER 8.09 (1.20) (a). Shift Engineer I (b). Shift Foreman (c). One E0.4 point each item] REFERENCE DAP 12-4, ' CONTROL OF HIGH RADIATION AREAS,' Rev. 11, Section B.4.c(3)(b), PS. 3 294001K103 ..(KA's) l (***** CATEGORY 8 CONTINUED ON NEXT PAGE

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..1 Bh__69dIUISIBBIIME_EB99EDMBgS_ggggIIIgggt Pasa 47 t BND_LIMIIBII9BS j j e ANSWER 8.10 -(1.00) ] The. Shift Engineer must. assign a_ monitor to provide direct surveillance of the area to prevent unauthorized entry. I REFERENCE DAP.12-4, CONTROL OF HIGH RADIATION AREAS, Rev. 11, Section B.8.ar pg. 8 294001K103 ..(KA's) l ANSWER 8.11 (1.00) Potential system radioactive contamination. REFERENCE Operating Order 49-67 294001K103 ..(KA's) ANSWER 8.12 (1.00) a. REFERENCE CECO Radiation Protection Standards, Sept. 1985, pg. 8 294001K103 ..(KA's) ANSWER 8.13 (1.00) 1. Reactor is critical CO.53 2. Reactor water temperature is above 212 degrees F E0 53 REFERENCE Unit 3 TS, 3.7.A.2 223001G001 ..(KA's) I I 1 (***** CATEGORY 8 CONTINUED ON NEXT PAGE

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l l l l ______________J

1 ) - ), Be__6DdINISIBBIIME_EB99E99BES _GQ8911198Q1 s Paan 48 t BUD _!:IdIIBII9BE f ANSWER 8.14 (1.00) Operating Order #14-87 provides an alternate supply path of offsite and/or diesel generator AC power to assure avail.bility of adequate AC-power supply using the 4 KV cross-tie between Dresden Unit 2 and 3 (Bus 24-1 to [,4 ff@,y A /e Bus 34-1). vA, / ^ 24 0 S T - V v'A '" i Alternate wording is accepted. ww ' ^ [. N Q1.4% ( M., ci m#f8~,f, Nt'c [ h,"b. ) REFEREhCE MJV' d # 4,-* <l 44 Y 't, W Operatin3 O der 41 -87. g 294001K107 262001G010 262001G001 ..(KA's) \\ d* ANSWER 8.15 (1.00) Operating Order (19-87 limits the number and frequency of times valve operation should be attempted. Alternate wording is accepted. REFERENCE Operating Order 419-87e. 294001M116 291001K106 ..(KA's) ANSWER 8.16 (2.25) i

s. d Potential LPCI loop select logic problems.

b, -a.cf5y P o t e nt i a l jet pump vibration problems.. b ECCS perf:- r;r ' ;- i r:3 r p e r : t i e r-

it" a s leep hre 6et b e e r-

?.,_~J.. REFE t E 4 pf i ^ Technical Specification, Basis 3.6.H,#ps s.3 /4. 6-31 and 3/4.6-32. 202001G010 ..(KA's) ) (**xxx CATEGORY 8 CONTINUED ON NEXT PAGE *****) 1 i

I 1 8 ) / A B.' ADMINISTRATIVE PROCEDURES 3 CONDITIONS, Pesa 4;" 8'dD_~(IMIl6IIQt@ t e c j 1 e t:I ANSWER 8.17 (3 00) l a.- (1). Maintain control of the Control Room. Request status of Fire i Companies No. 1 and 2 as they are assembled. [0.53 j / (2). Initiate GMP as appropriate. [0.53 I AssistFireMarshal[inpreparationofappropritdefirereports. (3). [0.53 1 ,,m 'I t r ) b. (1).tY. rect all. F3refighter in the event of a fire. f (2). Consult with.bhe Off-Site Fire Department Officer te Jrding i methods and.vvipment to extinguish fires. l (3). Keep the Shifts Engineer and Fire Marshal! informed,of; fire's status when fi e dighting. 't d', { f1 f (4). Assist Shift En.epneer and Fir e Marsh.al. in r;rlpar dt iba of I< l / \\ appropriate fird reports. ( t, ) J [' (5). Will aise ace?pt.j report the use of any fire equbment to the Fire, Marshal as'soon as possible aftev the fire i <,. Is u t., 1 II l ? l [Any three at 0.5 each3; f r + REFERENCE ff( (b / s j ) i EPIP 200-4, page 3. f i' l 29 001K116 294001A116 ..(KA's) \\ 1, / ). ?> y/ ANSWF0f 8.13 12,50) 'p ' s as n a.n a at G,t-i%h o f p e r s o n r e p o r t i n g a. u i. location lp.d Lye of emergency { ~' d a te ay!- t i nq o f incident c. ? d. whetheti a radioactive relcrin,i.e'taking place t potentially affected popul)V mhy be necessary. [ ' j,

t Mertw.s-: ::; P e.

whethen wrotective measure i f. Cany 5 o p kt 0.5 point each] J LM4W. i ^ S. 4i44.4 +3 fe . j 'C q REFER 14CE K, gl p i f 1.4.cuf & l' Er1P 300-1, pgs 1 aHd 7. i i 295000G002 294001A116 ..(KA's) .t ( j< t 1 j (***** CATEGORY B CONTINUEDs0N NEXT FACE

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l ju ol l y t )

c- -1 .? o 'I 6__e9dINIEMOIME_EG99E99BEf t_G9BDIII9det essa 50 BNQ_(ZdII611gyg i. y l ANSWER 8.19 (0.50) high r adiation t s REFERENCE 1 DEOP Lesson Plani DEOP 300. 295034K304 295004M102 ..(KA's) i i l l l l ? (***** END OF CATEGORY B *****) (******n** END OF EXAMINAUON **********) j

I 1 1 0 c. ATTACHMENT 1 /. j l d i 112A d' TEST SEAL "~ tina ", RESET PE IN = = CM1 A I i LS1 A - s& Mi l moom OPEN l 10BA " ~ ",",21088 TEST 20MA gSOLENOID % CM1A VALVE on. 120 VAC ISOL INST DUS CONTACTS 10BC ~ ." "106D .Q fCRIA - CONTROt sw- ,,,, g 203 1A 120 V AC j [ SOLENOID 1 i i 1 111A 108A". ~ RESET SEAL IN 106A ~ ~ " ~106s GM.I 8 SOL 126 VDC CONTACTS ) 106C ",,*, " "106D i 1 OUTBOARD 4 i t> INBOARD j i d i l' 4 CR1A p l 111A = = CONTROL. %10BA N 1088 SWITCH 2031 A 125VDC i 1 f SOLENOID l i

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EQMTION SHEET 4 w f = ma v = s/t Cycle efficiency = (Network out)/(EnergyIn) 2 i w = eg s'= V,t + 1/2 at t e = ac I a = (V - V,)/t A = AN A = A e" KE = 1/2 av f g PE $ men Vf = V, + at w = e/t 1 = ant /tpg = 0.693/t1/2 ~ A =,,2 ifreff = ':(tm)(t,)] t i g,,,,, [($ g ) + (t,)] aE = 931 as "*V a av * ~** I = 1,e Q = aCpat 6=UAaT I = I,e~"" I = I,10-*/TV' Pwr = v ah f TVL = 1.3/s m(t) P = P,10 HVL =.4.693/u P=P,e# SUR = 26.06/T SCR=5/(1-Kgf) CR,=5/(1-K,ff,) SUR = 26p/s= + (s - p)T CR (1 - Kgfg) = CR I ~ Idf2) j 2 l T$ (a*/e) + [(s - syIs] M = 1/(1 - Kgf)=CR/CR, j T=s/(s-s) M =~(1 - Kgf,)/(1-Kgfj) T = (s - e)/(Is) SDN = ( -Kgg)/KWf ' e = (Kg,-1)/Kg f = aKg,/Kgf

= = 10 seconds I = 0.1 seconds-I e = [(a*/(T K Il*E8df (1 + IT)]-

/ df tdii=18 2,2 2 P = (s4V)/(3 x 1010) Id gd jj 22 2 E = eN R/hr=(0.5CE)/d(meters)' 2 R/hr = 6 CE/d gf,,g) piiscellaneous Conversions Water Parameters 1 gal. = 8.345 lem. curie = 3.7 x 1010dps 1 ga;. = 3.78 liters kg = 2.21 1ho ~ 1 ft' = 7.48 gal. hp = 2.54 x 103 8tv/hr' Oensity = $2.4 lbqi/ft3 1 aw = 3.41 x 106 Stu/hr Density = 1 ge/cm3 lin = 2.54 cm Heat of vaporization = 970 Stu/lem

F = 9/5'C + 32 Heat of fusion = 144 Stu/1bm
C = 5/9 (*F-32) 1 Ata = 14.7 psi = 29.9 in. Hg.

1 STU = 778 ft-lbf 1 ft.. H O = 0.4335 Inf/in. 2

i m ) MASTER COPY 2 1 U.S. NUCLEAR REGULATORY COMMISSION j SENIOR REACTOR OPERATOR LICENSE EXAMINATION (LIMITED TO FUEL HANDLING) Facility Dresden Reactor Type BWR-GE3 j Date Administered 87/10/26 i Examiner Hills. D. E. j Candidate Instructions to Candidate: 1 Use separate paper for the answers; staple question sheet on top of the answer sheets. Points for each question are indicated in parenthesis 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) i hours after the examination starts, i Category % of Applicant's % of l Value Total Score Cat. Value x 20.00 19.70 M. Reactor and Fuel I Characteristics j 21.75 21.43 N. Equipment, Instru-mentations, Design Description 19.50 19.21

0. Procedures and i

Limitations I \\ 20.25 19.95 P. Emergency Systems and Safety Devices 20.00 19.70 Q. Health Physics and Radiation Protection 101.5 TOTALS Final Grade All work done on this exam is my own, I have neither given nor received aid. ~ t,ancicate s Signature 3 _a

I ) I-A i NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS l ,a Du' ring the administration of this examination the following rules apply: ] 1 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties. l 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 i 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 e'ch a 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, f or exan.ple, 1.4, 8.3.

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, j

12. Use abbreviations only if they are commonly used in facility literature.

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.

Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSHER BLANK. l

16. If parts of the examination are not clear as to intent, ask questions of j

the examiner only. i l I

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. I A

-l l a 1Ei..When you complete your examination, you shall: a. Assemble your examination as follows: l (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. Turn in all scrap paper and the balance of the paper that you did c. not use for answering the questions. d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked. I

.w

MJ__EEAC'IQB_AND FUELCJARACTERISTICS Page ;4-

.V QUESTION M.01 (2.00)' Upon~removingLspent fuel from-the reactor, explain'whatJfactors. contributed by'the spent fuel affect the' change'ii core reactivity and attte whether they each-increase or decrease core' reactivity. '(Two factors required) (2.0)' I QUESTION M.02 (2.50) 1 Answer each of the following as TRUE or FALSE: During refueling a fuel loading chamber (dunking chamber)Elocated a. too close to the source of neutrons in the core'is considered a-conservative condition in1thatithe 1/M plot would predict criticality too early. (0.5) b. The higher the enrichment of th'e fuel rod, the. larger is'the diameter of the upper end plug. (0,5) Vertical' support of the peripheral fuelJbundles is'provided by c. the fuel support castings, the control rod guide tubes ~and the bottom head of the' reactor ves'sel. -(0.5) d. During refueling,:it is possible to. uncouple a. control rod blade'from its drive from either above or below'the core. (0.5) A core quadrant has a fixed boundary whether source range monitors e. or f uel loadit:g chambers are oused. '(0.5) (***** CATEGORY M CONTINUED.ON NEXT PAGE:*****)

7-N BEAD 1O_B_A@ FUEL _.QBABAGIEBISI19S Page 5 } .t QUESTION .M.03' -(2.50) Assume the reactor fuel had been completely unloaded and reloading was in progress. Given the following data: Number of fuel Elements Loaded SRM Reading (CPS) 0 35 10 54 l 20 67 30 Before Move 85-After Move 15 50 31 60 65 The dunking chamber was moved following loading of 30 fuel elements. Calculate and develop a 1/M plot and predict the number of fuel elements loaded at criticality. (2.0) Must show all work for full credit. l QUESTION M.04 (2.00) i The Nuclear Engineer indicates that the value of "R" in'regards to { shutdown margin given in the Nuclear Design Report is 0.10% delta k/k for .) the upcoming cycle. Explain what is meant by the term "R" and indicate what shutdown margin will need to be demonstrated to meet Technical Specification requirements under the stated conditions. QUESTION M.05 (1.00) Explain the major differences between an 8 x 8 fuel assembly and a P8 x 8R fuel assembly (i.e., address what the "P" and "R" designations represent.) (Two differences required.) (1.0) QUESTION M.06 (2.00) Describe four methods to verify correct fuel assembly orientation during refueling. (2.0) (***** CATEGORY M CONTINUED ON NEXT PAGE *****)

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) Mr BEAGIQR_A@ FUEh_QllABA91EBISTJCS Page 6 1 QUESTION M.07 (1.00) j During fuel loading, it is determined that it is not possible to provide a l minimum of 3 counts per second on a particular Source Range Monitor (SRM). Other than adding sources, what method can be used to achieve the required i 3 counts per second on the SRM? (1.0) I l 1 QUESTION M 08 (1.50) State two reasons why burnable poisons are.used in the core. (1.0) { a. b. What is the burnable poison that is used in the core? (0.5) 1 QUESTION M.09 (1.00) Choose all of the following which are purposes of the fuel channel. (MAY be more than one answer) l To provide a bearing surface for the control rod blades. a. h. To provide a fuel handling capability via a lifting bail. c. To direct the coolant flow upward through the fuel bundle. i d. To guide the fuel assembly into the fuel support piece during fuel loading. j i i QUESTION M 10 (1.50) Refueling is in progress. (The dryer and separator have already been removed from the core.) In order to replace a control rod guide tube, ) what additional items will need to be removed from the core? (1.5) { l QUESTION M.11 (2.00) a. Describe what is meant by the term " Fuel Cell." ( 1, 0.) b. Dresden Unit 2 uses the " Core Control Cell Concept." Explain what ) is meant by this term? (Include its purpose as well as how fuel assemblies are arranged in the core to achieve this effect.) (1,0) (***** CATEGORY M CONTINUED ON NEXT PAGE *****)

M REAM OB_AN_D_EMEL_GHABag E81sTICS Page 7 ) QUESTION M.12 (1.00) DFP 800-14 Replacement of a Fuel Support Piece" indicates that proper nwientutten cif the fuel support place (FSP)-is verified by checking that the orientation dog on the FBP is pc.mitioncd evur the alignmunt pin r-4n thw-core support plate. Choose which one of the following diiections describes the corner of the control cell on which the alignment pin is f i found. i I a. Southeast b. Northeast c. Northwest d. Southwest 1 i 1 (***** END OF CATEGORY M *****) _____s

N[ EQQlEMENT. INSTRUMENTATIOHl_AND DESIGH Page 8 .] DESCRIPTION ~ .) 1 QUEST. ION N.01 (1.50) Indicate TRUE or FALSE for each of the following: The Refuel Level Indication range overlaps with the top ofLthe active J a. fuel. (0.5) b. The Refuel Level Indication is calibrated ior cold plant conditions. (0.5) d Two Refuel Level Indicators are provided with signals from Rosement' c. pressure instruments. (0,5) QUESTION N.02 (2.50) During refueling there will normally be a net inflow of water from. a. the Control Rod Drive (CRD) System. How is this. excess water normally removed? (Two methods required) (Include both the system that removes the water and where it is sent for both methode) (2.0) b. What is the purpose of this water from the CRD System? (0.5) QUESTION N.03 (2.00) a. The air supply for the control rod grapple is provided through dual air hoses from the refueling platform. What is the source of air to the refueling platform? (1.0) b. In the event that the air supply is lost to the control rod grapple, what method is available to unlatch the grapple hook? (1.0) I 1 'l l 4 (***** CATEGORY N CONTINUED ON NEXT PAGE *****) ] 1 J J

I I N[ E9UIPMENT. IHSTEMMEDIATION AND DESIGE Page S DESgBIPTJQN

e 1

i QUESTION N.04 (1.75) l For each of the following set of conditions (a. through d.) select the appropriate resulting interlocks ( 1 through 6). (Some conditions may have more than one appropriate interlock - If a particular condition is needed to complete an interlock in addition to the given conditions, then assume that condition does not exist. If none of the interlocks will result, then state "none.") i Conditions l Trolley Mounted Hoist Loaded, not all rous in, Refueling Platform a. near or over core. b. Not all rods in, Refueling Platform near or over core, c. Refueling Platform near or over core, Mode Switch in "Startup." d. Frame mounted hoist loaded, refueling platform near or over core, 1 Mode Switch in " Refuel." Interlock Results l 1. Rod Block 2. Stop Refueling Platform travel toward core 3. Stop Frame Mounted Holst Operation 4. Stop Trolley Mounted Hoist operation 5. Stop Service Platform Hoist operation 6. Stop Fuel Grapple operation l 1 1 I (***** CATEGORY N CONTINUED ON NEXT PAGE *****)

,I Pa'ge 10: cE[ EQUIPMENT. INSTRUMENTATION AND DESIGN' DESCRIPTION QUESTION N.05 (2.00)- For each of the following e'quipment (a. through d.)~ associated'with'the Refueling Platform, select 1the available sources of. control (1 through 3); for that piece of equipment. (MAY be more than one source for each piece' of equipment.). Equipment a. Platform ~ Bridge- .b. Main Hoi'st (Fuel Grapple') c. Trolley Mounted Hoist d. Trolley. e. Frame Mounted' Hoist Control Source 1. Operator Control Console (Mounted to the Mast and operated from the Operator's Basket.)' 2. Frame Mounted Hoist Control Pendant'(Suspended from the. hoist'and operated from the Operator's Basket.) 3. ' Control Pendant operated'from the walkway deck. QUESTION N.06 (1.00) Select the one piece of equipment from the following list that11s used to remove or install the channel fasteners which hold the fuel channels to the fuel assemblies. 1 1 a. ' Channel Bolt Wrench.. j b. Channel Handling Tool c. Fuel Disassembly Tool d. Channel Manipulator 1 \\ l. 1 l l l l ,1 ) l (***** CATEGORY-N CONTINUED ON.NEXT.PAGE *****) j I 1 l

N. EQUJPMENIs_ INSTRUMENTATION AND DESIGN Page 11 DESCRIPTION QUESTION N.07 (2.00) With respect to a fuel support casting being removed from a control rod guide tube, answer the following questions: The fuel support grapple is lowered from the refueling platform a. and the grid guide is clamped to the monorail hoist cable. What I is the reason for clamping the grid guide to the cable? b. The fuel support grapple is seated into the fuel support casting and engaged. How does the fuel handler know that the grapple is 1 properly engaged? I QUESTION N.08 (2.00) j a. State the purpose of the incore detector cutter. (1.0) b. Wnat is the motive force for operation.of the incore detector cutter? 4 (0,5) l c. In what area of tne plant is the incore detect?r cutter used? Be specific. (0.5) QUESTION N.09 (1.00) i Choose which one of the following pieces of equipment are required to remove a spsnt LPRM string from the core and transfer it to the opent fuel pool. (More than one piece of equipment MAY apply.) (1.0) a. Instrument Otrongback b. Incore Manipulator c. Inntrument Handling Tool ds Thermal Sleeve Installation Tool I (***** OATEGORY N CONTINUED ON NEXT PAGE *****) i l i l

m 'EQUIPM NI. INSIRNMEHIAllON ANDiDESIGH .Page 12-t ' Ns DESCRIPTION , q, d QUESTION' N.' 10 (2.50) a., 'DFP'800-24 " Loading and. Handling loffthe1 General Electric (GE),Model' i / i

1600 Cask for Bundle HardwareLShipping"'requiresLthat the, Reactor-Building crane be operated in the '" Restricted Mode" duringfcabk handling on the'613' elevation.

What11s meant by the term " Restricted Mede?" ( l'. 0 ) - 3 b. . Explain how the Renetor Building. Crane is placed-in.the restricted mode. (0.5) Explain under what c' nditions handling of a cask on the refuel $r g h o c. floor in the Unrestricted Mode would1be allowed.. (1.0 ) ; l QU3ST10N N.11 (1.60) l Indicate TP'1E' or' FALSE for each of the following statements. The sensors for the Area Radiation Monitors are Proportional a. Counters. (0.6) b. The Refueling Floor Equipment Hatch Area Rediation~ Monitor contains a rad.tation source to provide a small upscale reading (" live 1zero"). ~ (0.5; If an Area Radiation donitor saturates'it wi1~1 stay ht. full c. scale. (0.5) l i RUESTION N,12 (1.00) -f Choose which one of the following describes how core load 3ng chambers y (aunkers) are installed in the core. l a. Detector chambar is placed in a dumny fusi assembly b. betector chamber is s9ated on a fuel support piece c. Detector chamber is suspended by a cable to.approximately fuel. center]ine d. Detector chamber is clamped to a control rodLblade guide I L i -(***** CATEGORY N CONTINUED ON N$XT-PAGE *+***) i 9 1 L.

l It E991L"JgBL 1NSTEMBENIAIf 0N AND_ DESIGN Page 13' DM_qMETlDN QUESTION N.13 (1.00)

DFP 800-12 " Replacement of.LPRM Strings" Section F.1 describes LPRM replacement without the use of a spring reel. Explain what equipment is available and how de;es it' function to make the use of -a spring reel unnecessary. (1.0) l l l l l 1 1 i l 1 l l l (***** END OF CATEGORY N *****) l _-__.____-__---__-___s

O[ ER2 GEE EES AEp_hlMlZAllQHS Page 14 [ l l QUESTION O.01-(3.00) The plant has gone into a ratueling outage. Fcr.sach of the following. 1 conditions, state whether actual movement of '2uel from and to the vessel' { is allowed. Surveillance testing shows that 0.25 inch of water vacuum in the 3 a. reactor building cannot ce ac.ieved with only one standby gas treetment system filter train. Technical Specifications require this capability for en operable Secondary Containment. b. The control room nuclear observer leaves the control room and is l enroute to the refueling floor. l The Source Range Monitor in the quadrant in which fuel is to 'ce c. moved cannot be rully inserted. A minimum of 3 counts per second l is still achlevable on the SRM. (Fuel Loading Chambers arc not l immediately available and the quadrant in question presently q contains more than two fuel assemblies:) d. A control rod blade has been withdrawn. f.11 fuel locations adjacent to the withdrawn control rod blade are occupsed by fuel cell caps. ] I e. All low pressure core and containment cooling subsystems are inoperable, f. All Intermediate Range Monitors (IRM) are inoperable. 1 QUESTION O.02 (1.50) J l Two control rods have already been withdrawn from the core for I maintenance during an outage. Describe three conditions under which the l l withdrawal of third control rod for maintenance would be allowed l l in accordance with Technical Specifications. (***** CATEGORY O CONTINUED ON NEXT PAGE *****)

f O[ PBQQEQQBES L AND LIMLTAllQUE 'Page 15 c. t QUESTION 0.03 (2.50) Answer each'of the'followingLas ThUE or FALSE: If.no agreement can:be: reached:between the Unit NSO~and the. Fuel-a. Handlers as to which move is.tc be performed per'the~ Nuclear. Materials' Transfer Checklist,sthe. Shift' Engineer is to be contacted to provide final directien'. (0,5) l b. The Fuel Handling Foreman.isfrequired.to update the' refuel floor tagboard of the core:after each fuel ~ move. (0.5). The Control Cell Status Checklist must'be completed:each time a-c, fuel assembly la removed from the core. (0. 5 )- d. Sipping can numbers.may be:specified or-changed:on the Nuclear. Materials Transfer Checklist ^at the discretion'of.the FuelhHandling Foreman. Transcription errots in'the Nuclear Materials.Transferfchecklist may e. be corrected by the' Fuel-Handling. Foreman with the stipulation that1 he notify a Nuclear Engineer.within 24' hours.from the= change. l l 1 ' QUESTION O.04 (1.00) Select the one statement which describes the. Technical Specification-definition of FUNCTIONAL TEST for a Source Range Monitor, A qualitative assessment of channel res.ronso when the channel sensor a. is exposed to a source of' increased neutrens. b. The injection of a sinulated signal into the' instrument primary sensor to verify the proper instrument response alarm, and/or' initiating action. The comparison by visual observation of.one neutron monitorings c. channel to two other independent neutron monitoring channels measuring the same' parameter, d. Adjustment of the neutron monitoring channel.such.that it. responds with the necessary range and accuracy to-known valves'of the neutron level which the channel monitors. 1 (***** CATEGORY-O CONTINUED ON NEXT PAGE *****L - _ - _ _ - _ - _ - _ _ - _ - - _ _. _ _ _ = - _ _

i ~ G. EBQCED_URES AND LlulIAIl0NS Page-le I l QUESTION O.05 (1.00) Technical Specification Figure 4.2.1 indicates that the Refueling Floor Radiation Monitors' calibration interval is once every 3 months. The Fuel Handling Foreman is told that it has been 105 days since these monitors-were last calibrated. Refueling is in progress. Select the statement below which describes the required action. Suspend refueling activities until the monitors are calibrated, a. b. No specific action is required at this time, May continue with refueling activities. Isolate the reactor building ventilation system and initiate the c. standby gas treatment system. d. Suspend refueling activities until radiation protection personnel are available on the refueling floor with portable radiation monitoring devices. QUESTION O.06 (2.00) a. DFP 800-13, " Unloading of the TN-9 Spent Fuel Shipping. Cask" indicates that immersion of the cask in the pool should bs minimised. Explain why it is desirable to minimize the l immersion time. (0,5) i b. Explain why when lowering the cask into the pool, DFP 800-13 requires the plastic protective skirt internal volume water level be kept approximately one foot above the pool water surface. (0.5) c. After the removal of the lid from the cask, DFP 800-13 indicates I that the front and rear drums should be observed for a steady stream of escaping air bubbles. If this occurrence was observed, of what would it be indicative? (0.5) d. When the cask is submerged in the pool, what is provided to handle any displaced water? (0.5) QUESTION O.07 (2.00) q List four locations designated as Item Control Areas for Dresden Unit 2. l l i (***** CATEGORY O CONTINUED ON NEXT PAGE *****) l

j O.. PBOCEDURES AND LIMITATIONS Page.17 ~ QUESTION O.08 (2.00) Refueling is in progress on the Refuel Floor when the Reactor Building i Fuel Fool Radiation alarm sounds, Specifically indicate to where the personnel on the Refueling Floor l a. _ 1.0) ( are supposed to evacuate (i.e., location of asser.bly area). b. Following evacuation from the Refueling Floor, wh(n can the personnel leave the assembly area? (1.0) I i QUESTION O.09 (1.00) l Explain why, upon re-establishing normal conditions fellowing refueling, nozzles in the dry-separator pit are directed so as to be continuously washing down the pit walls during draining of the pit. (1.0) I QUESTION O.10 (1.00) l Explain why DFP 800-25 " Handling and Loading of the NLI 1/2 Spent Fuel Ghipping Cash" requires that the cask be hung freely for five minutes when I it is being initially lifted off the cash trailer. QUESTION O.11 (1.50) Answer each of the following statements involving unirradiated fuel l receipt as True or False: a. The Nuclear Materials Custodian is responsible for inspecting the' i shipping containers against the shipping papers provided by the driver. b. At a minimum, shipping papers should include a packing list: Radioactive Material Shipment Record, and Bill of Lading. c. If containers are found damaged or missing, it should immediately be reported to the Security Supervisor. l l l (***** CATEGORY O CONTINUED ON NEXT PAGE *****) j l l

O. PROCEDURES'AND LIMITATIONS Page.18 ) i QUESTION O.12 (1,00) Describe two reasons why the Fuel Storage Pool Minimum Water Level-1 -prescribed by Technical Specifications is at 33 feet. i i 1 i -i i 1 i i a l l 1 i l l l \\ l 1 i (***** END OF CATEGORY O *****) 1 i .- J

i j b. EMERGENCY SYSTEMS AND SAFETY DEVICES 'Page 19 I l A -QUESTION' P.01 -(1.00) A. control 1 rod has been uncoupled:in conjunctionwith'the replacement,of' .the control rod blade...What. method is.available to ensure:an. inadvertent scram signal-during.the replacement will~not result in serious equipment. damage? (1.0) 1 QUESTION P.02 (2.00) Identify each of the components ("a"'through '"d") on-the attachedjControl-Rod Figure-. QUESTION P.03 (1.50) Describe the' changes that occur in the Source Range Monitor (SRM),Ethe Intermediate. Range Monitor (IRM),'and.the. Average Power Range Monitor: (APRM) scram logics upon removal of'the-SRM shorting links. (1.5) QUESTION P.04 (2.00) a. During a refueling outage, the' reactor well is inadvertently drained without the fuel pool gates in place. Explain how adequate. water j coverage above the active fuel.is'still ensured in'the. fuel pool. (1.0)- b. With the fuel pool gatesin place, explain how= leakage;between the gates would be' detected. '(1 0) QUESTION P.05 (2.00) Ventilation ducts are located around'the~ perimeter of the. dryer-separator l pit, reactor well, and the. fuel storage pool just above.the~ skimmer weirs., ] a. What are the purposes of having the ventilation ducts located there?- i (Two purposes required) (1.0) i b. To what systems can'this air be routed? (Two systems required)' (1.0) i -l R i (***** CATEGORY-P CONTINUED ON NEXT PAGE *****)- i 1 .j - - - - - _ _ _ _ - _ = _ _ _ _ _ _ _ _

9.08 in. e g MANDLE l -l. / 8# UPPER GUIDE ROLLER 1 L g TYPICAL 4 PLACES '/ p O WELDED END PLUG 6.5 v. q *p / 1 r ,y i s' ;e l i O i ' e I, 'I, NEUTRDN ABSORDER .N, b RODS P I I SHEATH h i d L Ikc I ]} I 103 in. ACTIVE p POISON LENGTH [ I O T%I 16 in. BETWE E N 4 f l B ALLS (TYPICALI I P l p I JC1 ; i C +y- / a.. s I) j dF If

p l*

h. L.. t.s 4 f *.' ). -Q m \\ ASSORBER ROD j i.- C. 4 =v-FlGURE 1. CONTROL ROD

i ~ P. EMEBgENCY SYSTEMS AND SAFETY DEylqES .Pege 20 l QUESTION P.06 (2.50) Define the terms " Maximum Normal Heat Load" and " Maximum Heat Load" a. as they pertain to the design of the Fuel Pool Cooling and Cleanup System. (2.0) b. Explain how the Maximum Heat Load of the Fuel Pool can be accommodated. (0.5) l l QUESTION P.07 (2.50) During refueling, the Reactor Building Overhead Crane Monitor exceeds its high radiation alarm setpoint. a. What is the alarm setpoint? (0.5) b. List the automatic actions or interlocks that occurs when the Reactor Building Overhead Crane Monitor exceeds the high alarm setpoint. (0,5) c. Other than the 4 decade range radiation indicator in the crane cab, list two indications that would cause personnel to become aware that the Reactor Building Overhead Crane Monitor had reached its high alarm point. (1,0) d. In what region of the Detector Current vs. Applied Voltage curve does the Reactor Building overhead crane monitor operate? (0.5) QUESTION P.08 (1.00) Explain why the Reactor Building Ventilation System is designed to maintain the reactor building pressure slightly less than atmosphere. l k 1 l i (***** CATEGORY P CONTINUED ON NEXT PAGE *****)

l a ~l P. EME8GENCY SYSTEMS _AND_faEEIY_ DEVICES _ Page 21 QUESTION P.09 (2.00) For each of the following set of conditions (a. through d.)-indicate whether a Reactor Building Ventilation System isolation and automatic start of the Standby Gas Treatment System will result. One Reactor Building Ventilation Exhaust Radiation Monitor reaches a. its high-high level trip setpoint. b. One Reactor Building Fuel Pool Radiation Monitor and..ane Reactor Building Ventilation Exhaust Radiation Monitor fail downscale, c. The Refueling Floor, New Fuel Storage Area, and Refuela'nh Sloor Equipment Hatch Area Radiation Monitors all reach their'high level trip setpoints. d. Loss of power to the Main Instrument' Bus with J. failure to transfer to the reserve power supply. QUESTION P.10 (1.00) Refueling preparations are in progress. A " cattle chute" has been installed between the fuel pool gates and the reactor vessel. Explain the purpose of this " cattle chute." (1,0) QUESTION P.11 (1.00) o Explain how breaching of secondary containment is prevented when moving a I fuel cash from outside into the Reactor Building. (1.0) l QUESTION P.12 (1.75) Indicate the order in which the following components in the Standby Gas Treatment System can be found starting from the inlet off the Reactor' Building Ventilation System. (1.75) a. Charcoal Bed Adsorber I b. Rough Prefilter c. High Efficiency Afterfilter d. Demister e. Electric Heater f. Fan i g. High Efficiency Prefilter ( l l l (***** END OF CATEGORY P *****)' l I p !'( ni Y

,i i e I G, y o 93.__HE610.'AlHY_SIVE_AND BADI6 TION PROTEQT,T.Q30, k' s 8pg6 22 4 -f"' s g" V >s j ] )) t p .( v [. s e, ,\\ l QUESTION Q.01 (2.00), While removing a shipping cask containing a spent fuel bundle from the fuel pool, the cask is dropped to,tha floor. As a result, one of the fuel handlers is stuck by debris and is unconscious. Both the cash and the 3 epent fuel bundle are severoly dipaged creating a 1000 rem /hr field at'the, injured fuel handler's body. i How long would it take f t.r an inWvidual to reach his Emerget cy a. Exposure Limit assumin{ that he elected to remove the irjured fuel ' e handler for medical assistaue? ,lAssume the radiation field the rescuer would be in averagev'iti0 'r em/hr while he is in it. Consic}ir , only whole body dose and show all : calculations f or f ull credit. ) b. If the rescuer decided to' d for help instd d, howlongwouldit'Nbe the injured fuel handler to reach.a definite lethal dose (i.e., I lethldoseforvirtuallyallexposedindividualsevenwith I treatment)? (Show all calculations for full credit.) .f s Q4ESTION Q.02 (1.50) j t \\, j 1 Choose the correct word in parenthesis for each of the follow .p Aneutronexposureorone.remwou'2dhaveailarger, smalldr, the a. same) biological effect as that of a gamma-Tay exposure of one [ l rem. (0.6) b. Due to their low penetration capabili t/, (alphas, betas, gamma-rayr., s neutrons) are considered to do biologi:al damage to the human body only when they are indested o2. inha3.eJ. (0.5) One curie of strontiun 90 would have (the aame, a different) biological effect as. hat of (ne curie of e bnit 60. (0.5) \\ QJESTION Q.03 (1.00) / Fill-in-the-blank. / When performing a whole body frisk, the frisker prcbe.(,nould oe held fa) inch (es) from the surface being surveyeci and the speyd at j which the probe is moved should be (b) inch (es) per secqnc.. (1,0) ,f ,r h /*u l y 'l (***** CATEGORY Q CONTINUED ON NEXT PAGE *****) ) !4 \\* d i

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T,

g,3, m./gg > 7w q 77

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c.g f ~ '~: Q. HEALTH PHEICS_AMD._ BAD 1hTION EBQILQ31QN .t Page 23' )+ r ? F

1

-QUESTION -Q.04 (2.00)- ] a> Describe the special precautionsHthat are. required to protect personnel during each of.the.following activities' occurring during refueling. SU (

1 Performarke'of a"Subor?itical Check

'/ a. .4 4 b '. Performance ofa Shutdown Margin; Demonstration F l n

s QUESTION

-Q.05 (1.00): t .] The.drainplugfora.:GE'Modelk1600' Cask (forbund1Aihardwareshipping)'is i removed to drain water fromithe cask cavity.. During the. draining-1 operation, the dose frate atJone foot:- from the. opening exceeds 100-millirem / hour. Choose which one of thelfollowidg types of areasnthat the

j area around the cask mu.-A: s now be considered.

/ t-4 a. Uncontrolled area l b. Radi.t. tion area ,l c. High RedJation Area l i d. Airborne Radioactivity Area l QUESTION Q.06 (1.00) / 1 The'use of the word " DANGER" cw radiological signs' uniquely means:which one of the following: a. The area is a high radiation area b. Airborne radioactivity is present c. The area radiation detector is inoperable 'd. The area is radiologically contaminated I j .I 1 i l(*****' CATEGORY Q. CONTINUED. Ork NEXT PAGE *****) . /: j \\ +l 'I 1 m-

Q[__ HEALTH PHYSICS' AND_RAQlAIION PROTEGIION-iPage:24' QUESTION. Q.07 (2.50) Describe under what conditions a Type 2 Radiation Work Permit is a. required. (1.0) b. .Under,what condition would a Type 2 Radiation Work Permit NOT be ~ required for entry into'a High Radiation Area?' (0.5) The Job Foreman is required to read and understand a Type;2 Radiation: c. Work ~ Permit and;then sign and initial for-the'first day'and: shift his crew is working. Following this, how often must the: Job Foreman initial wnile his crew works under that RWP?- (0.5). d. The work area dose rates shall:be verified at least once a dayLwhile work'is in progress for a Type 2 Radiation Work Permit. Who is authorized to perform the. working' dose rate verification <in accordance with Commonwealth Edison Radiation Protection: Standards?- l (0.5) QUESTION Q.08 (2.50) a. Prior to entry into a High Radiation Area, an individual must call the' Control' Room and be. logged in. List three items of~which the Unit Operator should be informed. (A local log-is not being used.) (1.5)- l b. If the High Radiation Area Access Point must remain' unlocked while. ] personnel are in the area, what additional controls must be 1 instituted? (1.0) ( QUESTION Q.09 (1.50) J An individual working on the refueling floor notices that his personnel j dosimeter is missing. What actions is he required to takenin.accordance j with Commonwealth Edison Radiation Protection. Standards?_ j .) l l'! .I 'l (***** CATEGORY. Q CONTINUED ON NEXT PAGE *****). l - - ~ _ _ - _ - - - - - _ - -

5 GL HEALTH.. PHYSICS AND RADIATION PROTECTION: Page'25: I QUESTION-Q.10 '(1.50) An individual.who normally works at-the Dresden: Station spends a. two weeks:workin, at Quad Cities. His whole body dose theLeurrent quarter prior to going to Quad Cities was 800' mrem. While at -Quad Cities, he received a whole body dose of.200' mrem. Upon'his return to Dresden, what additional whole. body dose may he receive-during the remainder of the quarter without obtaining.special approvals? (State all assumptions and show all work for full credit.) (1.0) l b. Whose approval must be received irr order to exceed Dresden's administrative quarterly whole body dose limit? (0,5). L QUESTION Q.11 (1.00) A condition arises that requires-the entry of a radiation worker into a High Radiation Area. The individual will receive an estimated whole' body dose of 50 mrem. The following data about the individual is.available. Age: 24 yrs. Exposure: Week - 5 mrem- -Quarter - 16 mrem Life - 29995 mrem NRC Form 4 is on file Indicate whether any 10 CFR 20 limits would be exceededLif the individual were ellowed to perform the task AND if any limits are exceeded, then explain how they are exceeded. } QUESTION Q.12 (1.00) l List two things an individual must have in accordance with Commonwealth Edison Radiation Protection Standards prior to performing tasks requiring the use of respiratory protective equipment. (1;0) QUESTION 1Q.13 (1.50) i List three reasons why a Radiation Work Permit may be terminated in accordance!with Commonwealth Edison Radiation Protection Standards. (***** END OF CATEGORY Q.*****) (********** END OF EXAMINATION **********) l 1 1 I

j ' EQUATION SNEET. "' f a es y a s/t Cycle efficiency'o (Network I out)/(Enugy'in)' s = v t + n att

w = ag e

E's ac2 KE = n av2 a = (Vf - Vo)/t A = AN A = Ag -At' e PE

mgh o

vf = v + at w = e/t-A = an2/tg = 0.693/ts.

t eff = [(tg)-(te))

v.F [(tg) + (t )3 - b AE = 931 am-g g,,-Ex k=iuCpAt 6=UAst I = lo *

e Pwr = Wrah I = Io'10~* N TVL = 1.3h P = Po10sur(t)

HVL = -0.693h P = P et/T o SUR = 26.06/T SCR = S/(1 - Keff) CRx = S/(1 - Keffx) { SUR = 26p/s* + ( s-p )T CRg (1 - K,ff t) = CR (1 - keff2) l 2 T = (2*/p) + [(p - p)/ip] M = 1/(1 'Keff) = CR /CRo 1 T = r/(p 's) M = (1 - Keffo)/(1 - Keffi) . T = (a - p)/(Ap) SDM = (1 - K,ff)/K,ff. p = (Keff-1)/Keff = Akeff/Keff A* = 10-5 seconds I ='O.1 seconds-1 p = [(1 /(T Keff)] + [s ff/(1 + IT)] e 'I ds =.1 d22 t 2 P = (reV)/(3 x 1010) 11d1 2 = 1 d22 2 I=N R/hr '= (0.5 CE)/d (meters) 2 R/hr = 6 CE/d -(feet) Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lbm. I curie = 3.7.x 1010dps 1 gal. = 3.78 liters 1 kg = 2.21 lbm i 1 ft3 = 7.48 galt - I hp = 2.54'x 103 Btu /hr Density = 62.4 lbm/ft I uw = 3.41 x 106 Btu /hr i Density = 1 gm/cm3 1 in = 2.54 cm Heat of vaporization = 970 Btu /1bm 'F = 9/5'C + 32 ) Heat of fusjon = 144 Btu /1bm 'C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. Hg. 1 BTU = 778 ft-1bf 1 ft H O = 0.433 Ibf/in2 q 2 l 1

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a 15 g a e ) Table t Saturated Steam: Temperature Table '1 ads Press. Specabc Volume [r thalpy Entran femp L0 per $4t. $st. Sat. $31 $at. Sat. Temp fahr 50 In. Li0uid Lap V8por Llewd Ivap V8por Li0md isto V8per f8hr 1 p v, vig - vg he h fg hg s, Srr 8e i 32 8 ' 0 08859 0 016022 3304 7 3J04 ? - 0 M 79 60M5, 075 5 0 0000 2 1873 2.1873 JJ.8

34 0 0 09600 0 016021 3061 9 3061 9 1 996 1074 4

.076 4 0 "M1 21f62 218c2 3t 1 38 8 0 10195 0 016020 2839 0 2839 0 4 008 1073 2 .077.2 09081 2 1651 2 1732 34 e j 38 8 0 11249 0016019 2634 1 2634 2 6 014 1072.1 l0781 6 0122 2.1541 2 2663 38 3 y -f 48 8 8 12163 0 016019 2445 8 24J5 8 8 027 Iti110 1079 0 ' 0 0162 2 1432 2 3561 48 8 42 9 0 13143 0 016019 22724 2272 4 10 031 ?069 8 1079 9 0 0702 21325 1.1527 42 8

3 44 8 0 14192 0 016019 ?!!2 8 ??l2C '

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+ 6 16 t. o on Table ?. Setwated Steamt Temperature Table-continued ~ Aos Press 50een~1aume Enthatpv latropt lemn tb per $at ' 5st set f8t $4t 58t Temp F8ter 50in Li0m0 fv40 v8por bow 0 frap. V8por leuio Ivap Vape 78hr 1 n W v e, vg he fi f t lie le l8s S 'I t l 180 8 7 5110 0 0165.0 50 21 ' 50 22 i43 00 990 2. 138.2 0201 5400 .8111 180 0 187 8 7 850 0 016522 48 172 48179 50 01 989 0 iWO 02662 .5413 .8075 982 8 - 144 6 8 203 0 0 h,534 46 232 86 24; .52 01 987 8 '31 8 : 0 2694 J346 8040 104 0 ' 188 0 2%8 0 016547 44 383 44400 5d d2 986 5 W- - 02125 .5279 8004 188 0 18G 8 8 947 0 016 % 9 42 t28 42618 56.03 9?5.1 ll41J 0.2756 4213 11969 188.8 1W B 9 340 0 016572 40 941 40 957 58 04 9M i i M21 02787 Sles 7934 toe e 192J 9 747 0 016585 39 97 39 354 6005 982 8. li42 9 0 2818 50e{ ,7900 192 8 194 8 10 168 0 016538 ' 37 408 34 824 62 05 981 6 1 43 7 0 2848 Seir .7865 its e Its 8 10 60s 0 0166k1 36 344 3639 64 06 980 4 1 44 4 0.2879 4952. 78.31 198 8 19tJ 81.058 0.016624 34 954 34 970 M 08 9711 1,45.2 - 0J910, 48st 11798 198 8 290 0 - 'l1526 0 016637 33 622 33 639 168 09 9779 - J60 02e40 : 4t24 1 7764 205 0 e.- s4.h.$f. h $ p p' ~ 204 0 ..? Sr2 0 016664 il 135. 31151 172 11 975 4 475 0 300, 4607 .7698 264 e i 2I28 ..3588 0 0lM94 28 862 28878 176 14 9728 49 0 0 306 4571 .7632 2es t 250 0 t e96 0 016719. 26 782 26 799 18037 970 a .50 5 0 312 ud7 ,7568 tit s 216.0 15 901 0 016747 24 818 24tS4 184J0 967.8 452 2 0.318. . 4323 - ,7505 218.0 270 0 17 186 0 016775 22 131 22 148 18823 965 2 l153 4 0 3241 4791 1 7442 ??O 8 224 8 14 % 6 0 016805 21 529 21 545 39222 ' 462 6. ,1H 9 0 3300 4081 1 7380 274 8 228 8 20 015 0 016834 ' 20 056 20 073 196 31 960 0 ,1% ) ' 0 3359 3961 1 7320 229 0 232J 21 H7 0 016864 18 701 - 18 718 20035-9574 .l57 8 0 3417 3842 47260 232 8 2M8 23.218 0 016893 17 454 17471 - 20440 954 8 .159 2 0 3476 J725 11201 2M 8 24f 8 24 968 0 016976 16 304 '8 321 20845 952 1 .160 6 0 3533 J609 ,1142 istI 244 9 26 826 0 016 % 8 15 ?43 5 260 21; 50 949 5 162.0 0 3591 3494

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5352 per s 3ttJ 236 193 0 01858 1 9291 1 9477 370 8 829 7

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, ~ >..... ~ =,. - . ~. - - - :..- --. ~.. y I o .e -j l '17 a t t E5 Table 1. Saturated Steam: Temperature Table-continued ADS Press Specific volume Enthalpv [ntro0y Temp t0 per $at $81 Set Sat Sal - Sat tem 0 Fahr Sg in LQu'd bap Vapor Liound Evap 9aDor lirluid [va9 Vapor Fahr t p we w eg vg he h ig he s, 3, 488 0 ' 4M 87 0 01961 0 97463 0 99424 441 5 763 2 204 8 0 6405 in s?99 I 4704 440 0 444 8 405 56 0 01969 0 93588 0 95557 446 1 758 6 '204 7 06454 odd 3 18W7 4e4 8 ' 444 9 504 83 0 01976 0 89885 0 91862 450 7 754 0 204 6 0 6502 0 8i27 1 4629 488 8 472 8 524 67 0 01984 0 86345 0 88329 455 2 749 3 - 704 5 0,6553 0 8042 145t2 472 8 478 9 M511 0 01992 0 82958 0 84950 4599 744 5 ,204J 0 6599 E7956 4.4555 476J - 488 0 566 15 0 0?o00 0 79716 0 81717 464 5 739 6 .204 1 0 % 48 0 7871 L4514 aos e 444 8 587 81 0 02009 0 76633 0 78622 4691 734 7 203 8 0 M96 0 7785 ,4481 444 8 488 8 61010 0 02017 0 73641 0 76658 4738' 729 7 203 5 0 6745 0 7700 .4444 aost 4:28 633 03 0 02026 0 70794 0N820 478 5 724.6 203 1 0 6793 0 1614 i4407 492 0

94.0 656 61 0.02034 - 0 68065 010100 443.2 719 5 2021 0 6442 0 7528 1 4370 stSA -

500 0 680 86 0 02043 0 65448 0 67492 4479 714 3 !?02 2 0 6890 0 7443 8333 Set t 804 8 70518 0 02053 0 62938 0 6499I 492 7 709 0 .2011 0 6939 01357 4296 604 0. M89 731 40 0 02062 0 60530 0 62592 4975 703 7

201-8 0 6987 0 7271 4258 les s

%ftg.'.. i M.- gg g.% S12 9 757 72 0 02072 0 58218 0 60289 502 3 698 2 ,200 5 0 7036. 0 7185 4221 312.8 816 8 7M 76 0 02081 0.55997 0.58079 507.1 6921 ,1918 0 7085 0 7099 i dl83 lies 3* 529 0 812 53 0 02091 0 53864 0 55956 512 0 687 0 ' 99 0 0 7133 0 7013 4146 528 I 124 0 841 04 0 02102 0 51814 0 53916 516 9 681 3 98 2 0 7182 0 6926 4108 324 0 528 0 870 31 0 02112 0 49843 0 51955 521 8 675 5 97 3 0 7231 06839 4070 126 8 532.0 900 34 0 02123 0 47947 0 50070 526 8 6696 96 4 0 7280 06752 4032 632 0 534 0 931 17 0 02134 0 46123 0 44257 531 3 663 6 95.4 0.2329 06u5 IJ993 Sat.e 548 0 962 79 0 02146 0 44367 0 46513 536 8 6575 1 943 0 7378 06577.l 3954 548 4 544 8 995 22 0 02157 0 42677 0 44834 541 8 651.3 8 93 1 0 7427 0 6489 1.3915 M40 tes t 1028 49 0 02169 0 41048 041217 546 9 645 0 1 91 9 0 7476 0 6400 .3876 648 8 182 8 1062 59 0 02182 0 39479 0 41660 55 0 638 5 1 90 6 07 5 0 6314 .3837 452A 55{i.2 158 5 1091 55 0.02194 0 37966 0 40160 632.0 1 892 E75is 04222 1.3797 668 8 558 8 !!33 38 0 02207 0 36500 0.38734 562 4 625 3 ' 187 1 01625 0 6132 1.3757 8e8.8 564 8 !!70 10 0 02221 0 35099 0 37320 5676 618 5 146 1 0 7474 0 6041 - 1.3716 564.0 Set t 1207 72 0 02235 0 33741 0 35975 5729 611 5 184 5 0 7725 0 5950 13675 tesJ $72 8 1246 26 0 02249 0 32429 0 34678 5783 604 5 182 7 0 7775 0 5859 1.3634 872.8 878 0 1285 74 0 02264 0.h tU 3.33426 5831 597J 140 9 01825 0.5766 IJS$2 876.I Set 0 1326 17 0 02279 0.29937 0 32214 589 I 549 9 l 79 0 03876 0 5673 13550 800 0 S44 8 1367 7 0 02295 0 28753 0.31048 594 6 582 4 ,169 '4 7927 05580 1 3507 SM S Sea t 1410 0 0 02311 027608 0 29919 600 1 574 7 .12 6 0 8030 0.5390 1.3420 ' 5s2.0 74 8 0 7978 0.5485 13464 let t 592.0 1453 3 0 02328 0 26499 018827 c51 566 8 448.8 14978 0.02345 025425 027770 611 4 558 8 16.2 0 8082 03293 1.1375 Se8A see s

5432 0 02364 0 24384 026747 6171 550 5 11871 0 8134 0 5196 1.3330 telt 844 I 589 7 0 02382 0 23374 0 25757 622 9 542 2 !!65 1 08187 0 $097 1.3284 864 8 les s 6373 0 02402 0 22394 0 24796 628 8 533.6 1162 A 0 8240 04997 1 3238 000 4 It2 9

. 686 1 0 07422 021442 0.23865 634 8 526 7 1159 5 0 8294 04896 1 3190 812 0 lit s 1735 9 0.02444 0.20516 0.22960 640 8 SIE6 1156 4 E6348 04794 JJ141 818.0 $29 s 17869 0 02466 0 lES 0 22081 646 9 506 3 Il572 0 8403 0 4689 l.3092 820 8 824 8 18390 0 02489 O !8737 0 21226 553 1 496 6 U49 8 0 8458 04583 .3041 824 0 521 9 18924 0 02614 0 17880 0.20394 659 5 486 7 832 8 19470 0 02539 0 17044 0 19583 ES9 4764 .,46 1 0 8514 0 4474 .29e8 528 8 438A 2028 00?$66 0 16226 0 18792 672 4 465 7 o 38 1 0 8628 04251 1.2879 836 4 42.2 0 8571 0 4364 11934 832.8 kt8 2059 9 0 02595 0 L5427 018C21 679 1 4548 I t33 7 0 3686 04134 2821 641e 448 2110 3 0 02625 0 4644 0 17269 685 9 443 1 644 4 21781 0 02657 0 3876 0 16534 692 9 4311 . 29 0 0 6746 0 4015 2761 M40 . u v, "-. 812J 2239 2 0 02691 0 3124 0158tt 700 0 411 2 . 24 0 0 8806 0 3893..2634812.0 'O- ^ 436J 23011 0.02728 0 2387 0.15115 7074 4051 13 1 0 3931 0.3637 1.2567 9s68 .2699 648 8 18 7 0 8868 0.3767. 'dow. 4 met 2365 7 0.02769 O !!663 0 14431 714 9 392 1 1107 0 0 8995. 0 3502 11498 0s88 Met 2431 1 0 02811 010947 0 13757 7*29 3771 1100 6 0 9064 0 3361 1 2425 564 8 654 8 24981 0 02858 0 80229 0 13087 73t 5 362 1 1093 5 0 9137 ' O.3210 11347 - as8J 6 72.s 2%66 0 02914 0 09514 0 12424 740 2 3wif 1085 9 0 1212 0.3054 1 2266 8 72.s 5763 26363 0 02970 0 08799 0.11769 7492 328 5 1077.6 0 9287 0.2092 1.2179 8764 let 3 2708 % 0 03037 0 08080 0 11117 758 5 310 1 1068 5 0 9365 0.2720 L2006 808 8 soe s 2782. 0 03114 0 0/349 0 10463 7682 290 2 1056 4 0 9447 Ch37 L1984 set s las t 28574 0 03?04 0 06595 0 09799 778 8 268 2 1047 0 0 9535 ( 2337 .1872 588 8 8 28 2934 5 0 03313 0 05797 0 09110 790 5 243 1 1033 6 0 9634 0 2110 tesJ 30134 0.03455 0 44916 0 08371 804 4 212A 1017.2 0 5749 0 1441 ..8744 esta 1.1591 se6A 198 8 30943 0 03662 0 03857 0 07519 822 4 172 7 995 2 0 9901 0 1490 1 1790 700 0 782 8 3135 5 0 03824 0 03173 0 06997 835 0 144 7 979 7

0006 0 1246. 1 1252 7e2.0 ins t al??J 0 04108 0 02192 0 06300 854 2 102 0 956 2 0169 0 0876 1 1046 ft4A 7s59 3198 3 0 04427 0 01304 0 05730 873 0 614 9344 0329 0 0527 l 70sJ 73347*

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~ .t, 1 P M. BEACTOR AND FUEL _9 BABA 9IERIST19S .Page:26-j: e-1

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1 x ;) ANSWER M.01 (2.00)' Removal of fission product poisons [0.5) increases reactivityf[0.5]. Rem' oval of fissionable material [0.5] decreases reac'ivity;[0.5]l c REFERENCE Reactor Physics' Review Lesson Plan, pgs. 41-46 234000A103 292002K502 ..(KA's). ANSWER M.02- ' ( 2. 50 ). l a. False. (0.5) b. True. (0.5)' c. False: (0. 5') d. True (0.5)- e. False' (0.5) REFERENCE Fuel Handling and Refueling Equipment Lesson-Plan, page 19 Fuel Lesson Plan, page 9 RPV and Internals Lesson Plan, page 17- ~ Control Rod Blade and Drive Mechanism-Lesson. Plan, page 6 DFP 800-1, page 4 215004K503 292002G007 292008K101- ..(KA's) ANSWER M.03 (2.50) No. of Elements 1/M: 0 '1 10 .65-20 . 52-30 Before Move.41-After Move 1 50 .48 60 .23 'i Criticality at-68 +/- 5 bundles (0.25 pts.for each value of 1/M for total.of 1.75' pts) (0.25 pts for plotting points correctly) U (0.5 pts for final answer for number.of fuel elements loaded at criticality)

1

(***** CATEGORY M CONTINUED ON NEXT PAGE *****) ) 'i l __ s J

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Rg&CIOR 8MQ_ FUEL' CHARACTERISTICS-Page127- .s-l REFERENCE Fuel HandlingLand Refueling Equipment Lesson Plan, page_20 215004K503

292008K101

..(KA's) F.NSWER M.04 (2.00) The value R is defined as the difference'in shutdown: margin at.the:most t limiting point'in the cycle and the'beginning of cycle. .(1~.0) Required shutdown margin - 0.35% delta h/k (0.25 O.10) (1.0). REFERENCE Fuel. Handling and Refuel Equipment Lesson Plan, pages 2 ano 16-Technical Specification 3/4.3.A.1:and Basis 3.3.A.1 DFP 800-1, page 4 292002K113 ..(KA's) ANSWER M.05 (1.00) The P8 x BR fuel assembly.has an additional water rod [0.5] (to help-flatten local flux peaking)'. The P8 x 8R fuel assembly has rods pre-pressurized to 3 atmospheres'with helium [0.5] (to improve gap conductivity between pellet and cladding). (Will also accept " Pre pressurized" for "P"'and '" Retrofit" for."R") REFERENCE Fuel Lesson Plan, page 5 l 292002K504 292009K132 292002G007 ..(KA's) I I I ' l ~ i ' i 1 L i l (***** CATEGORY M CONTINUED ON NEXT PAGE'*****) ) i ?

]

u J i M.,. REACTOR _AND' FUEL CHARACTERISTICS l Page 28 l 1 1 ANSWER M.06' ' ( 2. 0 0 ') i 1. The spring clips should 'be at tlie center. of "the four-bundle array ~. 2. The fuel: spacer blocks should be located on1the.inside surfaces ~off l the four bundle array. 3. The fuel orientation lugs on:the handle of'each' fuel bundle should point toward the center of the'four-bundle array. 4. The serial number on the fuel handle should read from the center of the four-bundle a" ray. 5. The gadolinia rod location, as. identified by. extended: fuel rod end plugs, should be correct. 6. There is cell to cell symmetry. (Any 4 of 6 @ 0.5 pts each) REFERENCE Fuel besson Plan, page 13' 234000Kb05 292002G013 ..(KA's) ANSWER M.07 (1.00) A fuel assembly scheduled to.be next to.the SRM may be temporarily-replaced by an assembly of higher burnup. (1.0) REFERENCE DFP 800-1 292003K110 215004G005 215004A202 ..(KA's), i ANSWER M.08 (1.50) a. To allow for loading of a larger mass'of fissionable material without. having to place more control rods in the core. (0.5) j To allow for better radial and axial flux shaping. (0.5) b. Gadolinium oxide (Gdo). (0.5) (***** CATEGORY M CONTINUED ON NEXT PAGE *****).

l i 5 M.' ' REAQTOR ' AN]L. FUEL _CHARACIEBISTICS :

Pa'ge529

. REFERENCE, .Dresden Facility 1986 Requal Exam Fuel-Lesson l Plan, page 11-292007K101 ..(KA's)' -l ANSWER M.09 (1.00)- 1

a. [0.5] and c-[0.5]

-(WillLsubtract 0.5' pts for each incorrect answer;given up to total value of' question) Ik REFERENCE Fuel Lesson Plan, page10 290002G007 ..'(KA's)

ANSWER

'M.10 (1.50) Four Fuel Assemblies (0.5) Fuel Support Casting (0.5) Control Rod Associated with the Guide Tube (0.5) REFERENCE DFP 800-1, Appendix A, page 2 201003K104 292002K111 ..(KA's) ANSWER M.11 (~2.00)- a. A control rod surrounded by four fuel bundles.. (1.0) -1 1 b. A control cell is a control rod and:four surrounding fuel. bundles j used to offset reactivity changes during power. operations [0;5]. 1 This is achieved by placing the low reactivity; fuel in the control j cells (such that control rod motion. occurs adjacent.to low power ] fuel). [0.5] (Helps-eliminate the need for periodic control rod-pattern changes.) i a ' l i '(***** CATEGORY M CONTINUED ON.NEXT PAGE *****) l { b

B.?.EACTOR AED FUEL CHARACTERISTICS Page 30 REFERENCE RPV and Internals Lesson Plan, page.18 Control Rod Elade and Drive Mechanism Leeson Plan, page 4 234000K505 201003K504 ..(KA's) ANSWER M.12 (1.00) q c. (1.0) REFERENCE j DFP 800-14, page 5 234000K102 234000K505 290002G007 ..(KA's) i l i (***** END OF CATEGORY M *****) l ]

= 9.- 'EQUIPMENTi-INST:10MEETATION AND _ DESIGN iPage 31:- 4 DESCRIPTION l ANSWER-N.01-(1.50)' a. False (0.5)- b. True (0.5) c. False (0.5)- REFERENCE Nuclear Boiler 1 Instrumentation Lesson Plan, page 7 216000K501 216000K122 (KA's) ANSWER N.02 (2.60) Reactor Water Cleanup Sytem [0.5] - Main Condenser.HotwellL[0.53-a. Fuel Fool Cooling System [0.5] 2/3 "A CST [0.5] b. To provide cooling water to.the CRD mechanism seals. (0.5) REFERENCE Fuel Pool Cooling and Cleaning System Lesson-Plan,.page.8 Control Rod Drive Hydraulic System Lesson Plan, page.19 201001K403 ?.01001K303 233000K103 ..(KA's). ANEWER N.03 (2 ^)0) Air Compresser mounted to the "A-Frame" of the Refueling: Platform. a. (1.0) b. The hook is f tted with a ring for manual unlatching. (1.0) REFERENCE a 1 DFP 800-1, Appendix A, page 5 Refueling Platfore Instruction Manual, page 2 234000K604 2'-4000A202 ..(KA's) 'l 1 i 1 (***** CATEGORY N CONTINUED ON NEXT'PAGE *****) -. _ _. _ _ _. _ _ _ _ _ _ _ m _. _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _. _. _ _..

N.' EQUIPMENI2 TNSTBUdENTATION ABD_DESIGB Page 32 D,ESCRJEIIQB l l ANSWER N.04' (1.75) a. 2, 4, 6 (0.75) b. 6 (0.25) c. 1, 2 (0.50) a d. 1 (0.25) l i (Subtract 0.25 pts for each additional incorrect answer given up to the j total value of question.) REFERENCE Fuel Handling and Refueling Equipment Lesson Plan, Figure 5 234000K502 234000K403 234000K401 ..(KA's) i ANSWER N.05 (2.00) a. 1, 2, 3 (0.75) b. 1 (0.25) c. 3 (0.25) d. 1, 2 (0.50) e. 2 (0.25) I REFERENCE Fuel Handling and Refueling Equipment Lesson Plan, pages 5-7 Refueling Platform Instruction Manual, pages 3-11 234000G009 ..(KA's) ANSWER N.06 (1.00) a. (1.0) 1 REFERENCE i DFP 800-5, page 2 234000G007 (KA's) 1 l i (***** CATEGORY N CONTINUED ON NEXT PAGE *****) l

4 M __E9QlEddNT. INSTRUMENTATION __AND-DESIGN .Page 33' I DESCRIPT1014 l ANSWER N.07 (2.00) To maintain the fuel support grapple in position. (1,0) a. b. Battery-operated position indicator energizes a light on the i grepple. (1.0) REFERENCE i i DFP 800-1, page 7 DFP 800-14, page 3 234000G007 ..(KA's) ANSWER N.08 (2.00) a. Incore Detector Cutter - To cut spent, irradiated incore detectors into lengths suitable for offsite shipment. (1.0) 1 b. Hydraulic (0.5) c. Fuel Storage Pool (0,5) 1 REFERENCE DFP 800-12, pages 1, 11, 12 234000G007 ..(KA's) ANSWER N.09 (1.00) c. (1.0) q (0.5 pts subtracted for each incorrect answer given up to. total value of question.) 1 REFERENCE l -l DFF 800-12, pages 3-6 234000G007 ..(KA's) i 1 i (***** CATEGORY N CONTINUED ON NEXT PAGE *****) 1

Es _EQHJ2 MENT. INSTRUMENTATION __AND DESIGN-Page 54 y i DESCRIZIlQN 1 l ) ) ANSWER N.10 (2.50) 1 J Autematic 13mits are instituted that ensure that the cask ~ remains a. within a controlled area. (1.0) 1 b. Mode Select Key switch on pendant placed in " RESTRICTED" position. J (0.5) Only in the case of equipment failures or emergency conditions when c. the cask is already suspended only to the extent necessary to get the i cask to a suitable stationary position so the required repairs _can l be made. (1.0) l f l REFERENCE I DFP 800-24, pages 3, 4 Tech Spec Basis, 3.10.F 234000K501 ..(KA's) j l I ) ANSWER N.11 (1.50) a. False (0.5) b. False (0.5) j c. True (0.5) { ) REFERENCE ) 1 i l Area Radiation Monitoring System Lesson Plan, pages 5, 7, 11, Table 1 272000A308 272000G007 291002K123 ..(KA's) ANSWER N.12 (1.00) l a. (1.0) REFERENCE DFP 800-23, page 1 L 215004K404 215004K106 234000K102 ..(KA's) i 1 l (***** CATEGORY N CONTINUED ON NEXT PAGE *****)

~ N. EQUIEME137. INSTRUMENTATION ANILDES19]3 Page 35 DESCRIPTION ANSWER N.13 (1.00) A drainable water seal cap allows water to be drained selectively through the LPRM lower seal region which then sucks the bottom of the LPRM i assembly into position. (1.0) REFERENCE DFP'800-12, pages 3, 8 234000G007 .(KA's) l l l J l 1 1 I i l l I l l***** END OF CATEGORY N *****) 1

t 98 '., t ,[, Page 7 36.: _.Qi PBQCEDURES AND LIMILA_TIQNS_ ; f .1 ~~ ~ l g, ANSWER 0.01 (3.00). . a. No. :f.0.5) 4.. m ~ b. Yes (0 5) c. .No-(0,5) d. Yes.- ( 0. 5.) c. Yes (O'5). f.< No ( 0. 5') . REFERENCE Dresden Tech Spec 3/4.7.c.1.c,'3.10.B.1,.3.SiF.3 -DTS 8473,~paFe 1.2 DFP 800-1,;pages 5 and 7. . 215004G005 215003G005 294001K103-' 201003G005 290001G005 ..(KA's) ANSWER O.02 (1.50)- 1. Reactor mode switch shall be locked in " refuel" 2. Fuel assemblies in the cell containing that control rod have been removed 3. All other ref ueling interlocks shall bd operable (except the' one bypassed to allow more than one. control rod from being withdrawn) 4. SRMs shall be operable in core qusdrant where fuel'or control rods are being moved and in an adjacent 1 quadrant.. (3 of 4 required @ 0.5 pts each) a REFFJ1ENCE -Dresden Tech Specs 3.10.E DFP 800-1, page 8 i f. '201003G005' 201003G006 .. ( KA ' s.Y 1 ANSWER O.03 -(2.50) .a. False (0.6) b. False (0,5) c. False-(0.5) d. True (0,5) e. True (0.5) (***** CATEGORY O CONTINUED ON NEXT.'PAGE *****) \\ '},\\ h 'm______E__._

1 O. EROCEQUREe_ABR_ LIMITATIONS Page 37 ) EFERENCE DTP 8472, page 2 DFP 800-1, pages 8 and 10 294001A106 294001A103 294001A102 ..(KA's) i' ANSWER O.04 (1.00) b a-r (1.0) REFERENCE Dresden Tech Spec 1.G DFP 800-1, pages 11 and 12 215004G013 215004G010 215004G005 .,(KA'e) 1 ANSWER O.05 (1.00) 1 b. (1.0) REFERENCE l Dresden Tech Specs 1.cc, 3.2.D.1 i 272000G005 .(KA's) ANSWER O.06 (2.00) l l a. To minimize contamination problems. (0.5) 1 b. To avoid contamination of the fins. (0,5) j c. Pcssible presence of an integrity defect. (0.5) l l i d Ski,mmer Surge Tanks. (0.5) i REFERENCE DFP 800-13, pages 15, 19, end 20 Fuel Pool Cooling and Cleaning System Lesson Plan, page 7 234000G01D 234000G013 2340003007 234000K103 ..(KA's) { I (***** CATEGORY O CONTINUED ON NEXT PAGE *****)

9 __PROCEDUBEG_hBD LIMITATIQHS Page 38 l l ANSWER C.57 (2.00) 1. Fuel Vault (0,5) 2. Fuel Pool (0.5) 3. Reactor (0.5) 4. Other Protected (Special) Locations (0.5) (Will also accept.any one of the following in place of item 4: J receivAnc/ shipping vehicle located in the reactor building or within the protected area of the yt.rd, chipping containers within the reactor building or fuel buildings, in the fuel handling fixtures, or other locations with the approval of the Nuclear Materials Safeguards Manager) REFERENCE CECO Nuclear Procedure No. 1, page 10 DFP 800-1, page 15 294001K103 ..(KA's) l ANSWER O.08 (2.00) On the ground floor of the Unit 2/3 Turbine Building by the Unit 2 l a. l feedpumps. (1.0) b. Notified by Radiation Protection that it is safe to leave. (1.0) REFERENCE EPIP 200-3, page 2 295023K301 294001A116 (KA's) ANSWER O.09 (1.00) To prevent the walls from drying up and some of the adhering contamination I from becoming airborne. (1.0) { l REFERENCE I Fuel Pool Cooling and Cleanup Lesson Plan, page 21 294001K103 233000K308 233000K306 ..(KA's) ) (***** CATEGORY O CONTINUED ON NEXT PAGE *****) l m

l b O. PSOCEDURES_bND LIMIIATIONE Page 39 l i ANSWER O.10 (1.00) This is a load test on the erane 12fting mechanism and brake system that io required by Technical Specifications. (1.0) l REFERENCE ] DFP 800-25, page 8 Tech Spec Basis 3.10.F ( 234000G005 ..(KA's) { ANSWER O.11 (1.50) l a. False (0.5) b. True (0.5) c. False (0.5) REFERENCE DFP 800-2, page 2 234000G001 ..(KA's) ANSWER O 12 (1.00) Above level to assure adequate cooling [0.5] (just above active fuel) Above level at which GSEP action is initiated [0.5] (5 feet uncontrolled 1 loss of level with level decreasing.) EEFERENCE i ~ Dresden Technical Specification Basis 3.10.C. 233000G006 ..(KA's) i (***** END OF CATEGORY O *****) l

R[ - EdgRGENCX_SYSIEUS AND -S&EETY ' DEVICgg .Page 40 ANSWER P.01 (1.00) Hydraulically isolate the CRD associated with thie' control rod to be removed. (3.0) REFERENCE LFP-800-16, page 2 201003G007 201002A205 201003K406 201001G007 201001K405 201'001K107 ..(KA's) ANSWER P.0L (2.00) a. Coupling Release Handle-b. Velocity Limiter c. Lower Guide Roller d. Coupling Socket REFERENCE Control Rod Blade and. Drive Mechanism Lesson Plan, page 4, and Figure 1 201003G007 .(KA's) ANSWER P.03 (1.50) SRM - Non-coincident scram - before no scram existed off SRMs (0.5) IRM - Non-coincident scram - coincident scram before (0.5) APRM-Non-coincident scram - coincident scram before (0.5] REFERENCE DFP 800-1, page 12 295023K304 215005K101 215003K101 215004K305 215004K101 ..(KA's) ) i l 1 (***** CATEGORY P CONTINUED ON NEXT PAGE *****) i

PI EMEBgEEg1_S1EIEde_AND SAFEIY DEVICES Page 41' ANSWER P.04 (2.00) The slot between the fuel pool and the reactor cavity is.only deep a. enough to permit passage of one fuel bundle when carried by the refueling bridge fuel grapple in the full-up position. (1.0) b. The drain line from the area between the gates is equipped with a flow switch. (1.0) REFERENCE Fuel Pool Cooling and Cleanup System Lesson Plan, page 6 233000A211 233000K406 233000K302 ..(KA's) ANSWER P.05 (2.00) Keep the airborne radiation level to a minimum (0.5) a. Keep the relative humidity of the refueling floor as low ac possible (0.5) I b. Reactor Building Ventilation System (0.5) StandbyGasTditmentSystem(0.5) REFERENCE Fuel Pool Cooling and Cleanup System Lesson Plan, page 8 233000K114 ..(KA's) ANSWER P.06 (2.50) i I a. Maximum Normal Heat Load - The sum of the decay heat released by the i average spent fuel batch discharged from the equilibrium fuel cycle l of the earliest closure of the pool gate, plus the heat released by ] l the batch discharged at the previous refueling. (1.0) I Maximum Heat Load - Decay heat of the full core load of fuel at the end of the fuel cycle plus the decay heat of the spent fuel discharged at the two previous refuelings. (2.0) i b. Assistance from shutdown cooling system. (0.b) 4 i l i (***** CATEGORY P CONTINUED ON NEXT PAGE *****) l 1 i l a

I l E. EMEBGENGy_ SYSTEMS AND_E6EEI1_pEVICES Fage 42 1 ~ \\ REFERENCE Fuel Pool Cooling and Cleaning Lesson Plan, page 18 233000K101 233000K507 233000K506 ..(KA's) 4 ANSWER P.07 (2.50) a. 20 +/- 5 mr/hr (0.5) As en a ls. o ccer& 3ofar.-r/Ar b. Inhibits the raising of either the main or auxiliary hoist. (0.5) ) c. Annunciator in crane cab. +Get7 64r51 Red light on the bridge. /4. ( 54 m m4eme es ecs Ltd A m y 1 e4 3 A+ o T/>h e = A d. Geiger-Mueller Region. (0.5) j 1 REFERENCE Process Radiation Monitoring System Lesson Plan, page 17 l DOP 1700-7, page 1 b /= P f o o - t o, pa3 e 3 291002K123 272000A310 272000A305 272000A101 272000K111 ..(KA's) l l ANSWER P.08 (1.00) l To ensure that all building leakage will be inwards. (1.0) l REFERENCE Reactor Building Ventilation System Lesson Plan, page 7 i 290001K601 290001G004 (KA's) 1 i ANSWER P.09 (2.00) i a. Yes. (0.5) b. No. (0.5) i c. No. (0.5) d. No. (0.5) ) (***** CATEGORY P CONTINUED ON NEXT PAGE *****)

.g. EMERGENCY SYSTEMS AHD SAEETY pgVICES Page 43 REFERENCE Process Radiation Monitoring System Lesson Plan, pages 14 and 15 Area Radiation Monitoring System Lesson Plan, page 10 AC Electrical Distribution System Lesson Plan, page 13 288000K204 288000K402 288000K401 261000fA01 ..(KA's) ANSWER P.10 (1.00) To ensure adequate shielding to protect personnel in the drywell against high radiation levels which otherwise might result from the dropping of an exposed bundle on top of the refuelinc bulkhead. (1.0) REFERENCE Fuel Handling and Refuel Equipment Lesson Plan, page 19 233000G007 ..(KA's) ANSWER P.11 (1.00) An equipment airlock can accommodate movement of the fuel cask. (1.0) REFERENCE Containment Gystems Lesson Plan, page 26 290001K401 ..(KA's) ANSWER P.12 (1.75) The following order: d, e, b, g, a, c, f (0.25 pts for each in correct order) REFERENCE Containment Systems Lesson Plan, page 27 261000G007 ..(KA's) (***** END OF CATEGORY P ****t)

1 9.' HEALTH PHYSICS AND RADIATION PROTECTION Page 44 ANSWER Q.01 (2.00) a. 75 rem (Emergency Dose Limit)/800 rem /hr : 0 03375 hr = 5.6 minutes (1.0) b. Assume 800 rem for definite lethal dose. (Will actually accept 800 +/- 100 rem for full credit.) \\ l a 800 rem /1000 rem /hr = 0.8 hr = 48 minutes (1.0) 2 REFERENCE g Health Physics Lesson Plan, pages 1, 21 EPIP 300-4, page 4 i CECO Radiation Protection Standards, page 35 295023A201 295023K101 ..(KA's) s ANSWER Q.02 (1.50) 3 a. The same (0.5) b. Alphas (0.5) c. Different (0.5) REFERENCE Heulth Physics Lesson Plan, page 15 294001K103 ..(KA*s) ANSWER Q.03 (1.00) a. 1/2 +/- 1/4 (0.5) b. One to two (Will accept either) (0.5) REFERENCE Radiation Protection 294001K104 (KA's) (***** CATEGORY Q CONTINUED ON NEXT PAGE *****) 3

l 1 Q. HEALTH PHYSICS 6HD_ RADIATION PROTECTION. Page 45 / ,y l i . I, / \\ L ANSWER Q.04 (2.00) e All personneE 'khquld be out of the line-of-site of the reactor or a. by having personn61 clear the refuel floor. (1.0) l s N b. All personnel shall leave the :rafuel floor. (1.0) REFERENCE / ,/., i-DTS 300-6, page 2 , e 'l 3 { DTP 800-1, page 3 !!e1(}G3G010 294001K104 (KA's) ') %) ) S. + J ANSWER / Q.05 (1.00) y t l s! ,,,p' c. o(1.0) ) REFERENCE, ( i, DFP 800-24, page 7 , CECO Rsdiation Protection Standards, ppge 3 (. j. 's.194001K104 29.4001K103 .. ( KA ' s )."

i. ' f l

f 3 7 AbSWER Q.06 (1.00) j ' I V(1.0) l-l a. ( ) ( REFERENCE 4i-C E C O R a d i a t i o n 'j? p t e c t i o n S h n d a r d s, page 8 ? '294001K104 if,1001K103 ..(KA's) I s 4 I / 3 t / '( y \\ ) \\ I f I t (3***** CATEGORY Q CONTINUED ON NEXT PAGE *****) l N i s s

G- "m

dbs y

4, E f 'm .i fd._,JEALIH_.EHMICS AND BAE $ TION PROTEQTIONL 'Page 46-

l 1

[' h 9 I u, d' ANSWER Q.07 (2.50) t Radiologically controlled' area's'where'whole body. dose'expectoduto. a. l exceed 50 mrem / day [0.5).. Also jobs involving significant1 i l contamination and/or airborne-radioactivity as specified-by-Radiation-Chemistry Supervision-[0.5].. (1.0) Rou:j%iue Operator High Rad., Rounds' (0,5)m b. On each'dNy.of that week his crew works under the RWP. (0.'5) ~ L c. l 4.). ti a,l s e o c.c.. c p 4 4 m t

e. =<_L sWh~

j d. Any' individual'who has received NuclearfBtation GeneralLEmployee I Training. (0,5) 3 REFEREECE a CECO Ra'dintion Protection. Standards, pages 13-18, 38, 294001K103 ..(KA's) i ) 3, % i ANSWER Q.08 (2.50) l,' I a. Location of the entry. (0,5):. k Department or Departments invol'ved. (0.5) Key user's security badge number. ' ( 0. 5 )- b. An individual must be present/in line of site of the entrance to positively control access, (110) i 4 .JFERENCE .lj DAP 12-4, page 6. 3 294001K105 294001K104 294001K103 ..(KA's) / ANSWER Q.09 (1.50) } l

1.,

Leave the controlled. area (0,5)' ~ f 2.4-Notify his immediate supervisor (0,5) 3. Notify Radiation-Chemistry (0.5): REFERENCE CECO Radiation Protection Standards,.page 39 294001K104 -294001K103 ..(KA's); 't j ) (***** CATEGORY Q CONTINUED ON NEXT'PAGE *****) r i 1 v i 3 1 i ?

HEALTH;EBYElqS AND_B6D16110N PROTECTION Page 47 e ANSWER Q.10 (1.50) 1250 mrem / quarter limit without special approval. (0,5) a. Dose thus far received = (800 + 200) = 1000 mrem Dose remaining = 1250-1000:250 mrem (0.5) b. Administrative and Support Services Assistant Superintendent. (0,5) REFERENCE CECO Radiation Protection Standards, pages 34, 42 294001K104 294001K103 ..(KA's) ANSWER Q.11 (1.00) 10 CFR 20 limits are not exceeded (1.0) (Even though 5(N-18) is exceeded, this only applies if applying 3 rem / quarter limit instead of 1.25 rem / quarter limit. Whole body dose will not exceed 1.25 rem in the quarter.) REFERENCE 10 CFR 20.101 CECO Radiation Protection Standards, page 33 294001K104 294001K103 ..(KA's) ANSWER Q.12 (1.00) i Adequate training on each type of respirator to be used. (0.5) Medical approval. (0,5) REFERENCE CECO Radiation Protection Standards, page 43 294001K304 294001K103 (KA's) l 1 i l (***** CATEGORY Q CONTINUED ON NEXT PAGE *****) i

1 Q[' BEALILLEBYSICS AND_E&pI6110N PROIECTION Pega 48 e O ANSWER Q.13 (1.50) . j Conditions Change 1 Job Coinplete { Job Cancelled j Expired (any three @ 0.5 pts each) ( REFERENCE CECO Radiation Protection Standards, page 19 294001K104 294001K103 ..(KA's) i 1 0 l l } 4 9 a l i ) (***** END OF CATEGORY Q *****) (********** END OF EXAMINATION **********) f

' TEST CROSS REFERENCE Page 1-j " QUESTION _1ALUE BEEERgjicg_ a M.01 2.00 ZZ20000001 i M.02 2.50 ZZZ0000002 i M.03 2.50 ZZZ0000003 M.04 2.00 ZZZ0000004 M.05 1.00 ZZZ0000005 1 M.06 2.00 ZZZ0000006 M.07 1.00 ZZZ0000007 M.08 1.50 ZZZ0000008 { M.09 1.00 ZZZ0000009 j M.10 1.50 ZZZ0000010 M.11 2.00 ZZZ0000011 3 M.12 1.00 ZZZ0000012 ) 20.00 N.01 1.50-ZZZ0000013 j N.02 2.50 ZZZ0000014 i N.03 2.00 ZZZ0000015 f N.04 1.75 ZZZ0000016 N.05 2.00 ZZZ0000017 N.06 1.00 ZZZ0000018 N.07 2.00 ZZZ0000019 N.08 2.00 ZZZ0000020 { N.09 1.00 ZZZ0000021 y N.10 2.50 ZZZ0000022 I N.11 1.50 ZZZ0000023 N.12 1.00 ZZZ0000024 N.13 1.00 ZZZ0000025 21.75 0.01 3.00 ZZZ0000026 0.02 1.50 ZZZ0000027 0.03 2.50 ZZZ0000028 l 0.04 1.00 ZZZ0000029 0.05 1.00 ZZZ0000030 0.06 2,00 ZZZ0000031 O.07 2.00 ZZZ0000032 0.08 2.00 ZZZ0000033 0.09 1.00 ZZZ0000034 0.10 1.00 ZZZ0000035 0.11 1.50 ZZZ0000036 I O.12 1.00 ZZZ0000037 j 19.50 P.01 1.00 ZZZ0000038' { P.02 2.00 ZZZ0000039 P.03 1.50 ZZZ0000040 P.04 2.00 ZZZ0000041 s P.05 2.00 ZZZ0000042 P.06 2.50 ZZZ0000043 P.07 2.50 ZZZ0000044 P.08 1.00 t'Z0000045 P.09 2.00 zZZ0000046 J

. TEST CROSS REFERENCE Page 2 Cl1EF1 TION _YAIME BEFERENCE-I P.10 1.00 ZZZ0000047 P.11 1.00 ZZZ0000048 P.12 1.75 ZZZ0000049 20.25 Q.01 2.00 ZZZ0000050 0.02 1.50 ZZZ0000051 Q.03 1.00 ZZZ0000052 Q.04 2.00 ZZZ0000053 Q.05 1.00 ZZZ0000054 0.06 1.00 ZZZ0000055 0.07 2.50 ZZZ0000056 Q.08 2.50 ZZZ0000057 Q.09 1.50 ZZZ0000058 Q.10 1.50 ZZZ0000059 Q.11 1.00 ZZZ0000060 Q.12 1.00 ZZZ0000061 Q,13 1.50 ZZZ0000062 20.00 MOMMMM 101.5 l I 1 i i f n __}}

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