ML20212M783

Exam Rept 50-237/OL-87-02 Administered on 870126-30.Exam Results:Four of Eight Senior Reactor Operators & Two of Four Reactor Operators Passed.Requalification Program Evaluation Rept & Exam Master Copy Encl

ML20212M783
Person / Time
Site:	Dresden
Issue date:	03/04/1987
From:	Bishop M, Bjorgen J, Burdick T, Clark F, Dave Hills, Keeton J, Lanksbury R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20212M753	List: ML20212M752 ML20212M783
References
50-237-OL-87-02, 50-237-OL-87-2, NUDOCS 8703120187
Download: ML20212M783 (79)

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{{#Wiki_filter:. _ _ _ _ - _ _. _. I, U.S. NUCLEAR REGULATORY COMI#SION REGION III Report No. 50-237/0L-87-02 Docket (s)No. 50-237; 50-249 LicenseNo(s).DPR-19;DPR-25 Licenswe: Commonwealth Edison Company Post Office Box 767 Chicago IL 60690 Facility Name: Dresden Nuclear Power Station Examination Administered At: Dresden Nuclear Power Station Examination Conducted: January 26-30, 1987 l $. hb 7/ J'7 Examiner (s): D. Hills Date )0 F. Clark f/87 DJ sbury) S/tl/E7 R a 1 Date m gf 'l 1/+k7 batt Da w n a,- M. 81 shop T/M 7 Approved By: [7 Operator Licensing Section Ifate r Examination Summary britten, ora', and simulator requ'a1Tfication examinaYfoiis were administered Examination administered en January 26-30.1987(ReportNo. 50-237/-0L-87-02 i to eight Senior Reactor Operators (SP0) and four Reactor Operators (RO). Results: Four SR0s and two R0s passes these examinations. All others failed one or more portions of the examination. The Requalification Program ~ Evaluation Report is contained in Attachment 1. K 37 V PDH

REPORT DETAILS .1. Examiners D. Hills, Chief Examiner R. Lanksbury M. Bishop J. Bjorgen J. Keeton F. Clark ,w 2. Examination Review Meeting Specific facility ccaments concerning written examination questions. followed by the NRC response, are enumerated in Attachatnt 2. 3. Exit Meeting t At the conclusion of the examinations, an exit meeting was conducted. 1he following personnel attended this exit meeting. Facility _ Representatives E. Eenigenbur ) R. Flessner, ke. Station Manager rvices Superintendent J. Wujciga. Production Superintendent B. Zank. Training Supervisor S. Stiles, Training Instructor R. Stols, QA Engineer NRC Representatives T. Burdick, Chief, Operator Licensing Section D. Hills, Operator Licensing Exeuiner R. Lanksbury, Operator Licensing Examiner J. Bjorgen, Operator Licensing hXaminer i F. Clark, Operator Licensing Examiner i c E. Hare Operator Licensing Examiner M. Bishop, Operatur Licesning Ex einer, INEL L. McCregor Dresden Senior Resident inspector lhe following items were discussed: a. During the simulator exauinations, the examinetw noticed a generic i reluctance to refer to both procedures and technical specifications when required. The candidates also failed to verify their actions against the procedure later. This was especially evident for. normal, abnorical, and annunciator operating procedures. b. During the oral and simulator examinations, the candidates had difficulty finding specific inforstation in technical specifications i although in inany cases, they knew of its existence. Several times the, candidates had f.o refer to the opJrating annunciator procedures to / 2 1g _ ;i

l identify the appropriate. technical specification. It was' evident .that the lack of an efficient index or other method to quickly identify and locate specific inforination interfered with the usage 1 of technical specifications.. 'J c. The simulator documentation provided to the examiners was inadequate. J Information was lacking in regards to a detailed plant response description, expected alarms, and expected operator actions. This - resulted in increased difficulty in developing simulator scenarios as well as additional last minute-changes in the scenarios. The-examiners reconscended that this infor1 nation be upgraded to a more acceptuble standard. f d. The examiners noted some unfamiliarity with the Dresden Safe Shutdown Procedures. Trairdng received in this area was discussed in the exit meeting, f, Y The examiners also noted procedure defic'iencies as noted below. e. j' l' (1) DOS-5600 Weekty Turbine Surseillancs does not contain any / precaut1ons concerningj tcwer level. 7 (2) DSSP-500-i2 - Describes start.ing of the' die'sel generator by momentarily holding the auto-start relap 4t'the associated 4 KV bus. However, a specific location or bis designation is not given in the procedure. I (3) Several'jnstances were noted in which the annunciator procedures do net refer to the appropriate abnormal procedures. (4) DGt 34 - Routine Power Changes does not contain any precaution ,e tu' increase load set on a power increase. As a resu'it, there / were 'siver/il instances during the simulator examina'tions in whichr he byrass valves 6rt.cked open due ti.-a failure to /*> c increaso the load set. t i / C I s' o e \\ t I f ,,Y f f a o' 3

ES-601-4 ATTACHMENT 1 REQUALIFICATION PROGRAM EVALUATION REPORT Facility: Dresden Nuclear Power Station Examiner! D. Hills t 6' 2'-30, 1987 Date(s) of Evaluation: January ' Area.s Evaluated: X Written X Oral X Simulator ,t ) Examination Results: R0 SRO Total Evaluation Pass / Fail Pass / Fail Pass / Fail (S, M. or U) Written Examination _ 2/2 7/1 9/3 K q Operating Examination Oral 3/1 7/1 10/2 S y 4 Simulator 3/1 6/2 9/3 M Overall Examination 2/2 4/4 6/6 U Evaluation of facility written examination grading p Overall Program Evaluation D' Satisfactory Marginal Unsatisfactory X (List major defi-ciency areas with brief descriptive t' comments) ( Major deficiency areas included the knowledge and usage o.f normal, abnormal, and annunciator operating procedures in addition to operating orders. Control board operations and understanding of instrument system response were also identified as major deficiency areas. Further weaknesses were noted dealing t with plant systens knowledge, usage of the technical specifications, and ~ communications / crew interactions. Individual weaknesses were noted across a wide variety of topics which also indicate an overall unsatisfactory requal-ification program. Submitted: Forwarded: Approve Ib sfAt} J u) v v , g Iection ef Branch Chief Examiner ll e

i J ATTACHMEh1 2 FACILITY COM4ENTS AND NRC RESOLUTIONS-FOR DRESDEN STATION OPERATOR EXAMINATION JANUARY 26, 1987 Question: 1.01b .R0 Requal 1.03b R0 Replacement 5.03b SRO Replacement Facility Coment: Answer key states that the mode of heat transfer from the fuel pellet to the cladding is conduction. The listed reference states " convection". However, this concept of heat transfer across the gas filled gap between the pellet and clad has been a point of discussion in the past. The mode would be considered conduction if the gaseous area were completely stagnant, but would be considered convection due to the pure definition of convectivr. heat transfer. This question was used on the Dresden 4 Requalification Exam of 1985 test #2, in which the answer p key allowed.either conduction or convection as an acceptable answer. Dresden requests that either answer be - accepted for this exam since each answer-is correct (See attached refere::ce.) U hRC Resolution: Coment is valid. Answer key will be changed to accept correction on conduction. Point valve remains unchanged. i j Question: l R0 Requal -- 1.04a 5.02a N SRO Requal 5.06a SR0 Replacement 1.06a R0 Replacement Facility Coment: This question was apparently written to be simple in nature but actually is vague and misleading. Control rod worth is a function of themal neutron flux, however, this L' statement is incoraplete. As explained by the Dresden Exam l Bank question / answer #12-020317, page 27: .... As power is increased, the flux also increases, this actually has no effect on the worth of the rod unicss the local flux is increased more than the core average. This explanation was also used on Dresden initial NRC licensee l examinttions: 5/24/83 question 1.03 and 5.05 and 6/25/84 question 5.03. Due to the question being misleading and incomplete, Dresden requests that either True or False be acceptable. (See attached reference.)

INtc Resolution: Comunent is not valid. The referenced question 12-020317 is in regard to the differential worth being "0NLY" dependent on the local flux. This question neither states nor implies this is the case but only that the CRW is a " function" of local flux, which is true. The answer key i remains unchanged. Question: 2.05 R0 Requai 2.07 R0 Replacement i There is an additional acceptable answer due to operational Facility Comment: experience. Operating with excessive number of RBCCW pumps could raise RBCCW system pressure above the service water pressure and therefore defeat the design purpose of the RBCCW system. (See attached reference) NRC Resolution: Comment is valid. It would be reasonable to expect this response from an operational point of view. This answer will be added to the answer key as a possible correct Point value remains unchanged. answer. Question: 2.07 R0 Regt.ti 2.12 R0 Repla:ement Were is an additional acceptable LP answer: Make-up to the Facility Comment: condenser Hotwell, Gland Seal Steam Loop Seal Fill, and Condensate Demineralizer Resin Transfer System. The answer The lesson key references the lesson plan for its list. These plan is not the controlling document for operation. additional answers should be acceptable due to the fact that it displays a knowledge of the system and interrelationships. (See attached reference) Comment is partially valid. The make-up to the Hotwell is NRC Resciution: from the " Condensate Make-up Pumps" which only start when the " Emergency Makeup Valve" OPENS and are separate from the Condensate Transfer System referred to in the question. This answer will not be added. The other two suggested answers are correct per P&ID's M-27 M-40, and M-35 and will be added to the answer key as possible correct answers. Point value remains unchanged. Questian: 2.09a R0 Rec.a1 2.15a R0 Replacement 2 i .,_,-,,.__.._,,,,._,.,,,,m,.,_.._,,,,__..,_.___.--,,,

L Facility Comment: Additional acceptable answers: 125 VDC and Pemanent - MagnetGeneratorLP(PMG) (See attached reference) NRC Resolution: Comment is partially valid. The answer key is wrong since the question asks for the " NORMAL" power supply for EHC Electronics which is the PNG. The answer key will be changed to "PMG" as the correct answer. The point value remains the same. The other suggested answers are redundant power supplies and as such are unacceptable for credit. Question: 2.09e R0 Requal 2.15e R0 Replacement Facility Comment: Check at facility: Answer is Essential Servico Bus (See attached reference) NRC Resolution: Answer verf fied per DOA 6800-1, Rev. 3 page 5 of E. This will be added to the answer sheet as the acceptable response. l l l i Question: 3.06 R0 Requal 3.10 R0 Replacement 6.12 SRO Replacement 6.08 SRO Requal Facility Coment: Additional acceptable answer: Manual operator action is a trip available to the operator (placing the control switch in pull-to-lock, PTL) Question is vague to the point that it does not mention how many trips are required for full credit. Additionally, since setpoints were not required, Dresden requests that "overcurrent" be accepted as a complete answer for " instantaneous and time delay ove murrent". NRC Resolution: Coment is partially valid. A MANUAL TRIP is available to the operator and will be added to the answer key as a re-l quired correct response. The question point distribution will now reflect four responses instead of three (4 at 0.25 each). The question is not vague as it quite plainly asks for the trips. Although the number of responses required were not specified it is assumed that this will not be a problem for the operators since Dresden's Exam Bank has numerous examples of questions where this is not specified. For example: 3

l 12-020153 12-020168 12-020157 12-020114 However, for future use of this q:;estion the number of responses will be specified. Question: 3.07a R0 Requal 3.12a 'R0 Replacement Facility Commsent:- The question and answer key is written to detemine the examinee's knowledge of the level setpoint reduction feature on a scram and from that standpoint the answer key is correct. However, the wording of the question is vague, and several examinees took a "real world" approach to the answer. Operational experience has prover. that a scram at high power will result in a Feedwater Pump auto trip on high level (+55"). This occurs due to leak-through of the Feedwater Regulating Valves and the fact that the Feed Pumps are not required to be manually tripped. The interpretation of steady state as written in the question is also vague and could afffect the examinee's interpretation of the question. " Steady state" could be interpreted to be a few minutes to a few hours or longer. NRC Resolution: Comment is partially valid. It is understandable for operators to take "real world" aporoaches to problems and respond accordingly. Credit will be given if the "real world" answer is given with an explanation of what should have happened. However, the question is not vague as plant l conditions are clearly set out and steady state is defined as "after scram" in the text of the question. The answer key remains unchanged. I Question: 3.07b R0 Requal 3.12b R0 Replacement I This question is vague. The expected answer is "The master Facility Comment: controller dual setpoint current output is returned to normal when the scram is reset". The question asked "what action is required by the operator to return vessel level to the master controller?" In the phrasing of the expected answer, vessel i level was never lost from the controller, just reduced. 4

This question could have been gnterpreted such that due 6 to feedwater flow at 12.5 x 10 (setpoint 11.2 x 10 ), the runout flow controller had taken over. Considering this, the action required by the operator to return vessel level to the master controller would be resetting of the runout relay using the pushbutton. (See attached reference) NRC Resolution: Comment is partially valid. The question is not vague but the answer is incorrect. Further analysis of the interaction of the master controller and run-out controller indicates that no operator action would be required to return vessel level control to the master controller. However, the question implies some operator action is required. Therefore, the answer key will be changed to accept either of the following: "No operator action required" or " Depressing the run out relay rest pushbutton." Point value unchanged. Verified by Lesson Plan Book 2 Chapter 6 Page 9-10 & 11. Question: 3.08 R0 Requal 3.14 R0 Replacement Facility Comment: Answer key is incorrect: 2301-8 would be closed. The 2301-8 (discharge to the vessel) would not have been open during the full flow test. Therefore, it would remain closed after the turbine is manually tripped. (See attached reference) NRC Resolution: Connent is valid. Answer key will be changed to CLOSED for part D. Verified by Table 1 of HPCI L.P. BK-3 #10 Table 1. pp.29 Section F.2.c.(b) and Figure 1. Point value remains unchanged. Question: 4.03 R0 Requal 4.06 R0 Replacement 8.06c, 8.06d, 8.13 and 8.14 SRO Replacement 8.04c and 8.04d SRO Requal Facility Comment: Dresden considers this question to be inappropriate. We do not require or expect our operators to commit the steps of j Operating Orders to memory unless they are immediate operator actions. l NRC Resolution: Comment is not valid. The operating orders referenced in this comment, such as 29-86, 38-86, and 39-86 are not in the form of procedural steps but rather in the fom of " General" or " Specific" CAUTIONS which must be observed by the operators during the perfomance of their assigned 5 j 1

6 b-duties. Since there was no documentation furnished to show that the procedure the operator would be using referenced or directed the operator to the Operating Orders the only method of complying with them is to " Remember" that they; are in place. The answer key remains unchanged. Procedure DAP 7-3 Rev. 5. OPERATING ORDERS specifically discoursges-the use of procedural steps (Section 2.a.) in operating orders. Questfon: 4.06 R0 Requal 4.10 R0 Replacement Facility Comment: Dresden considers this question to ba inappropriate. To correctly answer this question requires memorization of a particular list contained in a precaution of a normal procedure (DOP 5670-1). This is contrary to the examiner's standard (ES-202) which states the candidate is not expected i to have normal procedures committed to memory but should be i able to explain reason, cautions and liritations. In addition, this question does not cenform to the Dresden Station Requalification Program Objectives. (See ettached reference) NRC Resolution: Comment is not valid. No copy of the Dresden Requalification Program Objectives was supplied so no resolution can be made to this portion of the Facility Comment. The candidates are not required to respond word-for-word in 1 accordance with the answer key to receive full credit. The requirements for tripping EGC in DOP 5670-1 are general conditions and not generally specific events. Candidates would be given credit for answering with specific instances covered by Abnormal and Annunciator Procedures in the Immediate Operator Actions Sections. To make this clear the answer key will be modified by adding the following statement: "Will also accept specific events requiring the unit be removed from ESC". The following procedures are examples of DOA's with Immediate Operator Actions requiring the unit to be removed from EGC: 1. DOA 040-2 Localized Flooding in Plant Step C-4 2. DOA 202-1 Recirc Pump Trip Step C-1 3. DOA 300-5 Inop or Failed CR's Step C-1 4. DOA 300-6 RPIS Failure Step C-1 5. DOA 3300-2 Loss of Condenser Vacuum Step C-1 6. DOA 5650-3 Turbine Control Valve or Bypass Valve Failed Open Step C-4b 4 t 6 I . ~ w----,---.n ,w, -_---,--...,..,-_,.__---.,---n.

L. l l Ouestion: 4.09 R0 Recual 4.16 R0 Replacement 7.10 i SRO Requal i f Facility Comment: Dresden considers this question to have an inappropriate l assignment of point value. This one question is worth 19% of the section. Additionally, the question is incomplete regarding the I l expected response. The question asks for immediate operator actions for each unit upon a loss of 125 VDC. It does not specify which unit has experienced the loss of 125 VDC. The specified DOA 6900-2 is for the loss occurring on Unit 3. DOA 6900-3 concerns the loss of 125 VDC for Unit 2. 0perators are not required to have l all procedure designation numbers memorized, therefore, immediate actions referenced in DOA 6900-3 should be l allowed as additional acceptable answers. NRC Resolution: Comment is partially valid. The question point value is within the constraints of the Examiner Standards and will remain at its present value. It is reasonable to expect that the operators would not know which unit is addressed by DOA 6900-2 and may include the actions for either or both units'1oss of 125 VDC power. Full credit will be given if the Unit is specified and the correct actions listed. For future use of this question the procedure title will be added so there is no misunder-standing of the intent of the question. Question: 5.04 SRC Requal 5.09 SR; Replacement Facility Comment: Additional acceptable answer: Equalize control rod exposure. (See attached references) HRC Resolution: Comment is valid. This-answer is included in tte " Control Rod Blade and Drive Mechanism" Lesson Plan Section E.1.b.3.d. but not in the " Reactor Physics Review" Lesson Plan the question was generated from. The answer key will be changed to include " Equalize Control Rod Exposure" as an acceptable answer and the notation "Any 2 at 0.75 each" will also be added. 7 e

Guestion: 5.08b. SRO Requal 5.13b SRO Replacement Facility Commient: Answer could also be "Yes". Dresden requests that the candidate's responsa and explanation'be graded on its correctness, not solely by the expected response in 4 the answer key. NRC Resolu?. ion: Comment is valid. The following notation will be added to answer key: "Will also accept YES if proper explanation is included." For future use of this question it will include a note that the cooldown is terminated after 30 min:;tes. Question: 6.01b SRO Requal 6.Olb ' SRO Replacement Facility Comment: Additional acceptable answer: Unexplained in:rease in Isolation Condenser side water level. (Seeattached reference) NRC Resolution: Commer.t is valid. This answer is provided in DOA 1300-1 Revision 4. Isolation Condenser Tube Leak, Se: tion A.4. and net in the referenced document the question was generated from. The following will be added to the answer key as an acceptable answer: "7. Unexplained increase in i j Isolation Condenser side water level." Point value remains l the sane. I Question: 6.03c SRO Requal 6.03c SRO Replacement Additional acceptable answer: To prevent viclation of the Facility Comment: RPS design (failure of one component or powe-supply will ~ neither cause nor prevent a scram). (See at*. ached references) NRC Resolution: Comment is valid. This answer will be added to the answer key as a required correct response. The point distribution will be changed to (0.25) each for each of the tw: responses. Total Verified by Lesson Plan, A/C point value remains the same. Electrical Distribution System Revision 7 Section E.5.a.2) 8

L Guestion: 6.04a . SR0 Requal 6.05a SRO Repla:ement Facility Comment: Additional acceptable answer: The list contained in the answer key is correct. However, it is a list of conditions which automatically cause a " lock-up". Manual operator action is an additional cor.dition which will cause a " lock-up". (See attached references)- NRC Reselstion: Comment is valid. " Manual actuation" will be added to the answer key as an acceptable response. Point value remains the same. Verified per DOP 202-12 Revision 3 Section F.1.a and F.1.b. Question: 6.06 SR0 Requa~ 6.09 SRO Repin:ement ) Facility r-nt: Additional acceptable answers:

1) Main Steam line drain valves close, 2) Recirculation loop sample valves close,

3) Isolation Condenser vent valves close, 4) Scram valves i

open. (See attached references) NRC Resolrtion: Conment is valid. Verified by Lesson Plan Book 3 Chapter 16 Table 1 ISOLATION GROUPS - VALVE. CLOSURES and Lesson Plan These Book 3 Chapter 12 Section F.3.c, Isolation Condenser. additional answers will be added to the answer key as possible correct responses. Point value remains the same. Question:- 7.01 SRO Requz-7 7.01 SRO Repiz:ement Dresden considers this question to be inappropriate. To [ Facility Comment: correctly answer this question requires memorization of a particular step within the text of an emergency procedure (DEOP 100-2). This step is not an innediate action. This is contrary to the examiner's standard (ES-402) which states the candidate should be able to describe generally the objectives and methods of emergency operating procedures. L Additionally, this question does not conform to the Dresden Station Requalification Program objectives. NRC Resciution: Comment is valid. This question will be deleted from the exam and the point value adjusted accordingly. 9 L ~'

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a Guestion: 7.08 SR0 Requal 7.12 SR0 Replacement Facility Comment: For the exariner's information: High end of the Narrow Range Instrument is +60". (See attached references) NRC Resolution: This will be added to the answer key in parenthesis to show it is an acceptable response. Point value remains unchanged. Question: 8.02b SR0 Requal 8.03b SRO Replacement Facility Comment: Additional acceptable answer due to recent procedure ~ change, refer to attached procedure (EPIP 200-20 Rev. 3). (See attached references) NRC Resolution: Due to the very recent revision to this procedure (IC days since official approval) full credit will be given for answers based on either Revision 2 or Revision 3, b;t not if they are mixed. Future use will require answers based on Revision 3 and part a_ of this question will be deleted. Question: 8.03 SRO Requal 8.05 SRO Replacement Facility Comment: Dresden requests that full credit he granted for ES",A vs IESDA and DNS vs IDNS since this.s how these agen:fes are always referred to during actual and drill situations. NRC Resolution: These desigt.ators will be added to the answer key as acceptable correct answers since they are terms comonly used at Dresden Station. Question: 6.04b SRO Replacement Facility Comment: Drywell Pnentics are no longer used. The correct answer is Pumpback and N, Makeup Systems. (See attached reference) NRC Resolution: Comment is valid. Since this system is no longer used Drywell Pnetmatics will be deleted and these two systems, Pumpback and N, Makeup, will be the correct response on the answer key. Verified by telecon with Steve Stiles on 2/5/87. 10

Question: 2.08d R0 Replacement Facility Comment: Dresden considers this to be an inappropriate question because the HPCI Booster Pump discharges directly to the Main HPCI Pump (approximate.ly 10 feet of discharge piping) and there is no instrumentation on this pump. NRC Resolution: Coment is valid. Since there is no indication that the operator could refer to for Booster Pump Flow and the only instrumentation available is Booster Pump Suction Low Pressure, it is not reasonable to expect the operator to remember this parameter. Therefore, this portion of question 2.08 will be deleted and the question point value reduced by 0.5. Question: 3.04~ R0 Replacement Facility Comment: Additional acceptable answers: Oil Level, Coolant Level, Diesel Oil Day Tank Level. (Seeattachedreference) { NRC Resolution: Comment is valid. Parameters verified by High Voltage Operators Round Book, Unit 2 D/G. These three items will be added to the answer key as possible correct answers. Point value remains the same. ADDITIONAL DRESDEN CWWENTS GE TIE JANUARY 1987 NRC REQUALIFICATION EXAMINATION FACILITY Cof9ENT: In regards to the January 1987 requalification examinations, Dresden feels that the question difficulty and the assignment of point values should be approximately the same for all i questions. The shortened version of the exams necessitates this type of examination construction. It is not fair to the examinees or the station to have a failing criteria based on only two questions. Every section of both January 1987 requalification examinations had at least two questions l worth a minimum of 13% of that section. In three particular I i sections,19 - 20% of th2 section was based solely on one question. Dresden considers that an inappropriate weighting of questions that the point values be revised to make the examinations more representative of the examinee's knowledge level. NRC Resolution: ES-107 estsblishes criteria for question weighting per l section as follows: " Verify that no single question and/or topic is worth more than 20% of that category". i The January, 1987 Dresden Requalification Exam did not l deviate from these criteria. I 11

i i EXAMINER CHANGES T0 ANSWER KEY DURING GRADING Ouestion: (2.02) R0 Replacement (2.01) R0 Requal Change: Additional acceptable answers: HPCI Drain Pot Cooling Water Dresden Lesson Plan, Book 3 Section Figure E.

Reference:

Question: (2.06) R0 Replacement (2.d4) R0 Requal Answer key correct if candidate assumed manual pushbutton Change: initiation. However, if candidate assumed manual actuatice of CARD 0X valves, correct answer is: 1. Cardot flows 1sunediately (no time delay) (0.42) No alarm in D/G room (0.42) 2. Anyone inside room must be warned before initiation (C.41) 3.

Reference:

• Dresden Lesson Plan, Book 4, Chapter 5, pp. 6-7.

Question: (3.01) RO Initial Answer #2 is changed to: Change _: Inserts an SRM scram (5 X 10 ES cps) ES)OR provides a 2. method of bypassing the S8M scram. ( Dresden Lesson Plan, Book 1, Chapter 15, p. 25.

Reference:

Question: (4.08) R0 Replacement (4.05) R0 Requal Change: Corrected Typo 13

+.* Facility Comment:. The Dresden Station Training Department has done an analysis of the SR0 and R0 requalification examinations, comparing the questions contained in the exarinations to Dresden's Requalification Program Exambank and Requalifica-tion Program Objectives. The results of these analyses are contained in Table IA,18 and 2. NRC Resolution: The relationship of the questions to the Dresden Requal Objectives, R0 72%, SR0 835 is consistant with the re-quirements of ES-601 as follows: "The facility objectives for the requalification program shall be the PRIMARY subject areas tested on a requalification examination." There is no existing requirement (or is it desirable) to construct an examination solely from the facilities examination bank. i i

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r t U. S. NUCLEAR REGULATORY COMMISSION e REACTOR OPERATOR REQUALIFICATION EXAMINATION FACILITY: QBESDEN REACTOR TYPE: BWR-GE3 DATE ADMINSTERED: 87/01/26 EXAMINER: HISHOP. M. CANDIDATE INSTRUCTIONS TO CANDIDATE: Read the attached instruction page carefully. This examination replaces the current cycle facility administered requalification examination. Retraining requirements for failure of this examination are the same as for failure of a requalification examination prepared and administered by your training staff. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up four (4) hours after the examination starts. % OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY _15.00 _21.20 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 15.75 26.14 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 15.00 24.90 3. INSTRUMENTS AND CONTROLS 14.50 24.07 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 60.25 Totals Final Grade All work done on this examination is my.wn. I have neither given nor received aid. Candidate's Signature

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'1 sRC RULES'AND GUIDELINES FOR. LICENSE EXAMINATIONS- ' ~ During the administration of this examination:the following rules apply: 1. Cheating on the' examination means an automatic denial of your application: .and could result-in more severe penalties. z2. Restroom trips are to be limited and-only one candidate atta 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 sectic.4 of the answer sheet. 8. Consecutively numbar each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only on one. side. of the paper -and write "Last Page" on the last answer sheet. 9. Number each answer as to category and number, for example, 1.4, 6.3.

10. Skip at least-three lines between each answer.

.11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.

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

13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.

114. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.

15. Partial credit may be given.

Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK. 16.-If parts of the examination are not clear as to intent, ask questions of the examiner only. 17.~You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.

i e s -iS. When you complete your examination, you shall: a. Assemble your examination as follows: (1) Exam questions on top. '(2) Exam aids - figures, tables, etc. (3) Answer.pages including figures which are part of the answer. b. Turn in your copy of the examination and all pages used to answer the examination questions. c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions. d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked. O w w, ,, -,,,, -w ry- --r ,w ,w ge ,r w we "~wv- -, ~ we-, w-w-

a ' 4 -},; PRINCIELES OF NUCLEAR POWER PLANT OPERATION. Peca 4 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW LQUESTION 1.01 (1.00) Mat'ch the major form of heat transfer with:the condition described. Form of Heat Transfer (1) Radiation (2) Convection (3) Conductien CONDITION a. Across a fuel pellet. b. Fuel pellet to the cladding. c. From the cladding to the coolant. d. From fuel rod to fuel rod in a voided core. QUESTION 1.02 (1.00) When using the neutron monitoring instruments to identify a partially uncovered core during accident conditions, the neutron flux Might show a drop-off as the detector enters the uncovered region of the core -M2y MIGHT this indication occur on the nuclear instruments? . QUESTION 1.03 (1'50) a. What is the effect of core age on the doppler coefficient of reactivity (MORE NEGATIVE or LESS NEGATIVE) choose one? (0.5) b. Explain why this change takes place. (1.0) I T l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) l 1

['. I '.' PRINCIPLFR OF NUCLFAR POWER PLANT OPERATIQH2 Paca-5- -... THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW . QUESTION-11.04 (2.00) .TRUE or FALSE 7 Answer the following concerning factors affecting control rod worth. .a. The worth of a control rod is a direct function of the thermal neutron flux to which it is exposed.

b.

A control rods worth decreases as mo'derator temperature increases. c. As the void content inceases, the control rod worth' increases. d. ' Control. rod worth' increases as the fuel temperature increases.

QUESTION.

1.05 (1.50) Briefly explain how a control rod withdrawal of one or two notches can result in a decrease in bundle power. -QUESTION 1.06 '(2.00) Discuss how xenon changes for a reactor power increase from 50% to 100%. Assume steady state. operation at 50% power for two days prior to the maneuver. Include time frames and why they occur in your -answer. QUESTION 1.07 (2.00) Refer to the attached Figure 1, Keff vs Exposure and explain why the change in Keff occurs from the following points: (TWO reasons required for each set of points) 1m. A to B b. B to C 1 (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

.t 4. 1. ' PRINCIPLES'OF NUCLEAR POWER PLAE7 OPERATION. Pasa 6 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.08' -(2.50) a. Why is it necessary to operate with some Condensate Depre ssion? (0,5). . b. What would be.the affect.(INCREASE, DECREASE or REMAIN THE SAME) on' cycle efficiency if the amount of condensate depression were increased? Explain your answer. (1.0) c. List TWO methods of reducing condensate depression. (1.0) QUESTION 1.09 ~(1.50) What are THREE interlocks / trips associated with the recire pumps that insure adequate NPSH is available? d 4 n ' (***** END OF CATEGORY 1 *****) ,,.,,,,----,-----_-r-,.-

• f

~~ 4l . + ':s 7 '2; PLANT' DESIGN INCLUDING SAFETY AND EMERGENCY Pega 7; SYSTEMS 't ~ QUESTION 2.01- '(2.00)' ' List fourLliquid/ steam influents to'the HPCI gland seal condenser. h.- QUESTION

2. 02' ~

'(1.00)~ 2 .Which one of the below best describes the. location of the standby liquid-control pumps relief valve discharge to the-system? A. .the 55 gallon drum. B. the individual pump suction. C. the standby liquid control tank. D. the tes't tank. 1 i t . QUESTION -2.03 (2.00) The plant is' operating-at 100% power. A complete loss of instrument air i~ occurs. [. With NO Operator action, answer the below questions?-- J-A. The level in the hotwell would ( increase, decrease, remain the.same-). l B. Service water flow to the TBCCW heat exchangers would ( increase, decrease, remain the same ). 'i C'. Scram. Discharge Vol. drain valves would ( open, close, remain as is ). L _ D.

Main feedwater regulating valves would ( open, close, remain as is.).

j. QUESTION 2.04 (1.25) ,Breifly describe the operation of the CARDOX SYSTEM in the Unit 2/3 D/G Room when the system is MANUALLY INITIATED. L L t-(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) f

p 2. ' PLANT' DESIGN INCLUDING SAFETY AND EMERGENCY Pcca 8 SYSTEMS QUESTION 2'05 (1.50) List two reasons why the number of RBCCW pumps in service must equal the number of heat exchangers being used. Note: Procedural reqairements is an unacceptable answer! QUESTION 2.06 (2.00) Identify which group isolation would affect each of the below systems / components. A. Isolation condenser condensate return valves. B. Shutdown cooling isolation valves. C. HPCI auxiliary oil pump interlocks against (preventing) an auto start. D. Drywell air sample-isolation valves. QUESTION 2.07 (1.50) List three systems receiving makeup water from the Condensate Transfer System. (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)

4 2. PLANT DESIGN INCLUDING' SAFETY AND EMERGENCY Pcco. 9 EYSTEMS QUESTION 2.08 (2.50) Select the best purpose (s) from column B for each of the EHC. components listed in column A. (one component from column A will match up with more than one purpose from column B) Column A Column B 1. Master Trip Relay A. To provide a mechanism for most turbine trips. 2. Master Trip Solenoid Valve B. To allow for overspeed 3. Mechanical Trip Valve testing. 4. Lockout Valve C. To act as supervisory component for trips. D. To interrupt FAS fluid during overspeed conditions. E. To convert FAS fluid into ETS fluid QUESTION 2.09 (2.00) List the no.rmal power supplies (or buses) for each of the following systems or components. {be specific - i.e. if there are two buses state which bus.} A. EHC electronics B. Reactor Man. Control C. Channel B Primary Containment Isolation System Logic D. CRD Scram Pilot Solenoid Valves (117) E. Main Steam Line Rad Monitor Channel B. F. Recharging Motor for 4 kV Breakers on Bus 23. (***** END OF CATEGORY 2 *****)

c 5'. INSTRUMENTS AND CONTBQM Poco 10 I QUESTION 3.01 (2.50) List FIVE of the SIX conditions required for an auto start of a standby feed pump. (include applicable setpoints.)- QUESTION 3.02 (2.00) A 4kv breaker has neither red or blue lights illuminated, the control switch is not in pull to lock and the bulbs checkout good. ' List TWO probable CAUSES of this condition AND the RESULTS. QUESTION -3.03 (1.00) List three of_the signals capable of causing an "RPIS INOP" Alarm. QUESTION 3,04 (1.50) The APRM meters on the 902-37 have an function switch containing seven positions. For each of the positions below, describe the indication available when the function switch is in each position listed. A. AVERAGE B. COUNT C. FLOW. QUESTION 3.05 (1.00) The following question concerns the Isolation Condenser System. Briefly describe the switch operation required for the 3-valve to decrease the cooldown rate. Assume no automatic signal is present. QUESTION 3.06 (1.00) Assume a Core Spray initiation signal is present. list the trips available for the 2A Core Spray Pump. Setpoints are not required! (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) 4

d' 3. INSTRUMENTS AND CONTROLS P 03 11 QUESTION 3.07 (2.00) The reactor is operating at ~SO% thermal power, a spurious scram (full) is received..The reactor vessel level tape was set at 30 inches.and two RFW pumps were operating (the third pump is out of service). The total feedflow increased to 12.5 X10E6 lbs/hr and level decreased to r 15 inches. A. Assuming no operator action and all facility systems function properly, what will be the steady-state {after scram} vessel level? B. What action is required by the operator to return vessel level to the master controller? QUESTION 3.08 (2.00) Assuming the HPCI system is in operation for full flow testing, list the position of the below listed valves after the turbine is MANUALLY tripped. A. 2301 - 4 Steam supply inboard containment isolation valve B. 2301 - 36 Suction valve from torus C. 2301 - 14 Torus return minimum flow valve D. 2301 - 8 HPCI injection valve QUESTION 3.09 (2.00) List the AUTOMATIC initiation signals and setpoints for the Low Pressure Coolant Injection System. (***** END OF CATEGORY 3 *****)

' 4. PROCEDURES - NORMAL. AENORMAL. EMERGENCY P003 12 AND RADIOLOGICAL CONTROL QUESTION 4.01 (2.50) List FIVE automatic actions initiated by a-main' steam line High Radiation Alarm at 3 X normal. QUESTION 4.02 (1.00) A coupling check is performed when withdrawing control rods to position 48. A. EXPLAIN HOW this coupling check is performed. (0.5) B. WHAT indication would tell the operator the rod was uncoupled? -(0.5) x QUESTION 4.03 ~ 1.00) ( Why must control rod "L-2" of Unit 3 be only notched out during 4 reactor startup? QUESTION 4.04 (2.00) List the entry conditions for DEOP 100, Reactor Control. (Include applicable setpoints) QUESTION 4.05 (1.75) List the MINIMUM nuclear monitoring system requirements / restrictions during core alterations. Positions and locations are required for full credit. QUESTION 4.06 (2.00) List four of the conditions / requirements of DOP 5770-1, Economic ' Generation Control Operation, directing the NSO to remove the unit' from EGC. (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

( I,~],,.p [.3_, -n ;.. ey Idi PROCEDURES ' NORMAL. ABNORMAL. EMERGEfig f pga 13 /'

AND RADIOLOGICAL CQ* IQL il. -

it \\, / 1 -QUESTION 4.07. (0.50)

j. '.

Normally..operationofthe.drywellcoolersisdetermine[soleh(by interior drywell' temperatures. - What is:the normal or operating range of the drywell temperature? l 9 i ) I p r n 1 QUESTION 4.08' (1.00) e 2.,,a i List,the actions required by DGP 1-1, Normal Unit Startup, i$ during a<f.~ reactor startup a period.of 15 seconds is observed. T/ i 1 b* ,. s... 1 ' i j .. UESTION 4.09 (2.75) Q List the Immediate Operator Actions for EACH Unit spran a loss of 125v DC power supply. DOA 6900-2.,/.,,,p gys y p py jf. r e ,a i '; - e r ].,' l. [ + Y .- ( / b (***** END OF CATEGORY 4 *****) (********** END OF EXAMINATION **********) q g.,.- ,,--.------g~n,- ,.-...m.-g n.-g,,m,----,, ,-----n,-.,--.+e-,,,-n ,--,y, ,,.,,c,.,

• A

-4 /) f 'I el ~ 2v.

/

i,- sg 4 1 l / e m J h1 y d q p i /. a ~ %w3 / 4 pf-f O a C. .g x = i s1 N 3 h 3-D =. vw. e (s' s 1 m u, j. v'. = I w / , f,, / [ 1 l 3 I '\\ .e w I w /# a .I~ ~ e g 4 I d4 I f I f f S, ~ 2 C 2 C C a H*y v. _._r__-

~~ c. ['n,O a EQUATION SHEET f = se ~ V= s/t 2 Cycle efficiency = Net Work (out) - w = mg s = v,t + hat g .a = (vg - v )/t E = mC ,,[. E 2 A = AN A = A,e KE = mv yg,v g A = In 2/tq = 0.693/tg PE = agh w = 6/c W = v4P-e (eff) '= (t,;)(ts) N AE = 931Am -(Jg + t ) h 6 = 5c "AT -Ix I-Ie P ' o i Q = UAAT I=Ie -UX Pur = W' E' g I=I 10"* o SU(Cl TVI. = 1.3/u P=P IO e /T HVI. ' O.693/u t P=P ~ 'SUR - 26.06/T T = 1.44 DT SCR = S/(1 - K,gg) f1 o\\ /, SUR = 26 CR = S/(1 - K,g,) g g 1(

• gg)g CR C1 ~ Keff)2

' = ' T " '(1*/o ) + [(i 'o)/A,gg ] 2 e 7,= 1*/ (o - I") M " I/Cl - Kagg) = CR /CR0 g 7 = (I - 0)/ 1 o eff M = (1 _ geff)CIII ~ Keff)1 8 " ( eff~} eff

• A eftheff Sm - (1. geff)/geff l

[1*/TKffg ] + [I/(1 + Kfgg )] 1* = 1 x 10 seconds ~ T a= P = I4V/(3 x 1010) -I A,gg = 0.1 seconds A I = No Idgg=Id22 WATER PARAMETERS Id =Id g 2 2 i 1 gal. = 8.345 lba R/hr = (0.5 CE)/d g,,,,,3) I gal. = 3.78 liters R/hr = 6 CE/d (feet) I ft = 7.48 Ca b MISCEI.L\\NEOUS CONVERSIONS O Density = 62.4 lbm/ft 1 Curie = 3.7 x 10 dps Density - 1 gm/cm 1 kg = 2.21 lba 3 \\ ',0-Heat of varortt. scion = 970 Ecu/lbo 1 hp = 2.54 x 10 BTU /hr Hestoffusice[=144Stu/1L.u 1 Mw - 3.41 x 10 Beu/hr 0 11 Atm = 14,7 psi = 29.9 in. I g. 1 Btu = 778 f t-lbf 2 1 ft. H O = 0.f.333 lbf/in g. inch = 2.54 cm 2 F = 9/5 C + 32 "C = 5/9 ( T - 32) g - ~

Volume, ft9th Enthelpy, Stullb Entropy. Stutto a r p P Watee Evap $4eem Water Evep Steam Water

Eve, steem s,

s, vs se 's

• I h,,

h, se o e 22 0.06859 0.01602 3305 3305 -0.02 1075.5 1075.5 0.0000 2.1873 2.1873 32 35 0.09991 0.01602 2948 2948 3.00 1073.8 1076.8 0.0061 2.1706 2.1767 35 40 0.12163 0.01502 245 2446 8 03 1071.0 1079.0 0.0162 2.1432 2.1594 40 45 0.14744 0.01602 2.)31.7 2037.8 13.04 1068.1 1081.2 0 0262 2.1164 2.1426 45 SO 0.17795 0.01602 1734.8 1704.8 18 05 1065 3 1083.4 0.0361 2.0901 2.1262 50 80 0.2561 0.01603 1107.6 1207.6 28.06 1059.7 1067.7 0.0535 2.0391 2.0946 60 70 0.3629 0.01605 8G8.3 868.4 38.05 1054.0 1092.1 0.0745 1.9900 2.0645 70 80 0.5068 0.01607 633.3 633.3 48.04 1048.4 1006.4 0.0932 1.9426 2.0359 80 90 0.6981 0.01610 46E.1 468.1 58.02 1042.7 1100.8 0.1115 1.8970 2.0086 90 100 0.9492 0.01613 350.4 350.4 68 00 1037.1 1105.1 0.1295 1.8530 1.9825 100 %20 1.2750 0.01617 265.4 265.4 77.98 1031.4 1109.3 0.1472 1.8105 1.9577 110 220 1.6927 0.01620 203.25 203.26 87.97 1025.6 1113.6 0.1646 1.7693 1.9339 120 330 2.2230 0.01625 157.32 157.33 97.96 1019.8 1117.8 0.1817 1.7295 1.9112 130 140 2.8892 0.01629 122.98 123.00 107.95 1014.0 1122.0 0.1985 1.6910 1.8895 140 150 3.718 0.01634 97.05 97.07 117.95 1008.2 1126.1 0.2150 1.6536 1.8686 150 160 4.741 0.01640 77.27 77.29 127.96 1002.2 1130.2 0.2313 1.6174 1.8487 160 170 5.993 0.01645 62.04 62.06 137.97 996.2 1134.2 0.2473 1.5822 1.8295 170F 180 7.511 0.01551 50.21 50.22 148.00 990.2 1138.2 0.2631 1.5480 IJ111 180 190 9.340 0.01657 40.94 40.96 158.04 984.1 1142.1 0.2787 1.5148 1.7934 2D0 200 11.526 0.01664 33.62 33.64 168.09 977.9 1146.0 0.2940 1.4824 1.7764 200 210 14.123 0.01671 27.80 27.82 178.15 971.6 1149.7 0.3091 1.4509 1.7600 210 212 14.696 0.01672 26.78 26.80 180.17 970.3 1150.5 0.3121 1.4447 1.1568 212 220 17.186 0.01678 23.13 23.15 188.23 965.2 1153.4 0.3241 1.4201 1.7442 220 230 20.779 0.01685 19.364 19.381 198.33 958.7 1157.1 0.3388 1.3902 1.7290 230 240 24.968 0.01693 16.304 16.321 208.45 952.1 1160.6 0.3533 1.3609 1.7142 240 250 29.825 0.01701 13.802 13.819 218.59 945.4 1164.0 0.3677 1.3323 1.7000 250 260 35.427 0.01709 11.745 11.762 228.76 938.6 1167.4 0.3819 1.3043 1.6862 260 270 41.856 0.01718 10.042 10.060 238.95 931.7 1170.6 0.3960 1.2769 1.6729 270 280 49.200 0.01726 8.627 8.644 249.17 924.6 1173.8 0.4098 1.2501 1.6599 280 290 57.550 0.01736 7.443 7.460 259.4 917.4 1176.8 0.4236 1.2238 1.6473 290 300 67.005 0.01745 6.448 6.466 269.7 910.0 1179.7 0.4372 1.1979 1.6351 300 310 77.67 0.01755 5.609 5.626 280.0 902.5 1182.5 0.4506 1.1726 1.6232 310 320 89.64 0.01766 4.896 4.914 290 4 894.8 1185.2 0.4640 1.1477 1.6116 320 340 117.99 0.01787 3.770 3.788 311.3 878.8 1190.1 0.4902 1.0990 1.5892 340 l 360 153.01 0.01811 2.939 2.957 332.3 862.1 1194.4 0.5161 1.0517 1.5678 360 340 195.73 0.01836 2.317 2.335 353.6 844.5 1198.0 0.5416 1.0057 1.5473 380 400 247.26 0.01864 1.8444 1.8630 375.1 825.9 1201.0 0.5667 0.9607 1.5274 400 420 30S.78 0.01894 1.4S08 1.4997 396.9 806.2 1203.1 0.5915 0.9165 1.5080 420 i 440 381.54 0.01926 1.1976 1.2169 419.0 785.4 1204.4 0.6161 0.8729 1.4890 440 1 460 -466.9 0.0196 0.9746 0.9942 441.5 763.2 1204.8 0.6405 0.8299 1.4704 460 1 460 566.2 0.0200 0.7972 0.8172 464.5 739.6 1204.1 0.6648 0.7871 1.4516 480 f 500 680 9 0.0204 0.6545 0.6749 487.9 714.3 1202.2 0.6890 0.7443 1.4333 500 ( 520 812.5 0.0209 0.5386 0.5596 512.0 687.0 1199.0 0.7133 0.7013 1.4146 520 l 540 962.8 0.0215 0 4437 0.4651 536 8 657.5 1194.3 0.7378 0.6577 1.3954 540 SCO 1133.4 0.0221 0.3651 0.3871 562.4 625.3 1187.7 0.7625 0.6132 1.3757 560 550 1326.2 0.0228 0.2994 0.3222 589.1 589.9 1179.0 0.7876 0.5673 1.3550 580 600 1543.2 0.0236 0.2438 0.2675 617.1 550 6 1167.7 0.8134 0.5195 1.3330 Sco 620 1786.9 0.0247 0.1962 0.2208 646.9 506.3 1153.2 0.8403 0.46S9 1.3092 620 640 2059 9 0.0260 0.1543 0.1802 679.1 454.6 1133.7 0.8666 0.4134 1.2621 640 660 2365 7 0 0277 0.1165 0 1443 714.9 392.1 1107.0 0.8995 0.3502 1.2498 660 6s0 27085 0 0304 0.0808 0.1112 758 5 310.1 106P.5 0.9365 0.2720,. 1.2086 680 700 3034.3 0 0366 0.0386 0.0752 822.4' 172.7 995.2 0.9901 0.1490 1.1390 700 705.5 3203 2 0.0508 0 0.0508 906.0 0 906.0 1.0612 0 1.0612 705.5 PROPERTIES OF SATURATED STEM AND SATURATED TABLE A.2 WATER (TEMPERATURE) A.3

~ 1. - PRINCIPLES OF NUCrRAR POWER PLANT OPERATION. .Peca 14 IHERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWER 1.01 (1.00). a.- 3 b. - J' 2. o r 3. c. 2 d. .1' [4 @ 0.25 each = 1.0] REFERENCE Dresden, Lesson Plan Thermo. HTFF, PP. 14-16 ANSWER 1.02 (1.00) With a loss of water level, the instruments which indicate-the thermal flux would be in a fast flux field and indicating low. REFERENCE Dresden, Lesson Plan, Degraded Core Response, P. 5 ANSWER 1.03-(1.50) More Ne'gative (0.5) a. b. Because of the buildup of PU 240 (which is a large resonance absorber). (1.0) REFERENCE Dresden, Lesson Plan, Reactor Physics Review, P. 32 4 (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) 4 ,,,. _.. _ _. ~. _.

1. PRINCIPLES OF NUCLEAB POWER PLANT OPERATION. Pzca 15 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ' ANS*ER : 1.04 (2.00) a. TRUE b. FALSE c. FALSE d. FALSE REFERENCE Dresden, Lesson Plan, Reactor Physics Review, P. 36 ANSWER 1.05 (1.50) ~ The steam bubbles generated by withdrawal of a shallow rod increase the void fraction, which adds negative reactivity off-setting the positive reactivity effects of the rod withdrawal. REFERENCE-Dresden, Lesson Plan, Reactor Physics Review, P. 39 ANSWER 1.06 (2.00) The initial power increase will cause a decrease in the xenon concentration (0.5]. The low peak occurs 4-6 hours after the power increase [0.5]. The power production term takes over [0.5] and a new equilibrium level is reached in 40-50 hours [0.5]. REFERENCE. Dresden, Lesson Plan, Reactor Physics Review, P. 45 (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) ~

~1. JPRINCIPLER OF NUCf.FAR POWER' PLANT OPERATION. -Pega 16 TMERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ~ NSWER

1.07-(2.00)'

A

a.

.1. Buildup of xenon. 2. l Buildup of samarium. b. 1. Gadolina burnup. 2. PU-239 buildup. [4 @ 0.5 each = 2.0) RT/ERENCE Dresden, Lesson Plan, Reactor Physics Review, Figure 62,.P. 47 ' ANSWER-1.08 (2.50) a. To ensure sufficient NPSH for the Condensate Pumps. (0.5) b. DECREASE [0.5]. The heat loss from the feedwater must be made up by the nuclear fuel [0.5]. c. Reduce cire water flow Increase cire water temperature . Increase condenser pressure [any 2 @ 0.5 each = 1.0] REFERENCE Dresden, Lesson Plan Thermo, HTFF, P. 8 ANSWER 1.09 (1.50) 1. Minimum speed until feed flow > 20%. 2. Suction valve closed trip. 3. Low vessel water level trip. [3 @ 0.5 each = 1.5] REFERENCE-Dresden, Requal Q/A Thermo, P. 5 (***** END OF CATEGORY 1 *****)

,. ~ ~ 2 '. : PLANT DESIGN INCLUDING SAFETY AND EMERGENCY-Pcca.17 SYSTEMS LANSWER 2.01 -(2.00) 1. Turbine seals -2. Stop valve below seat. drains 3. Turbine first' stage drain and casing exhaust drains 4. Casing exhaust drains 6 p ?c: I * 'a~ PT (any 4 @ 0.5 ea = 2.0) .5. Turbine exhaust piping 7,g,,,, REFERENCE DRESDEN - Student text, Chapter 10, HPCI, page 8 Cr dw ksw /k, Bael 3, S-4 /O, A'y. 9. e ~ ANSWER 2.02 (1.00) C (1.0) REFERENCE DRESDEN - Student Text, Chapter 3, Figure 1. ANSWER 2.03 (2.00) A. Decrease B. Increase C. Close D. Remain as is (4 @ 0.5 ea = 2.0) REFERENCE DRESDEN - Student Text, Book 4, Chapter 3, rage 14. 1 (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) I.

1 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pcca 18 ~ SYSTEMS ANSWER-2.04 (1.25) l-s 1. Energize a siren in the room (0.25). 2.. Stops the vent fan in the room (0.25). 3. Opens the storage tank discharge valve (0.25). 4. Opens the selector valve for the room (0.25). 5. The pilot valves will remain open (for approximately 97 seconds) (0.25). ..- r 1. .,4 a... M,< 1 r re, 'o I,s e. c'. - , 7... (f,_@0.25ea=1.25; 1...;- 'r !..12') el; .\\n..+ 2: 3 .c REFERENCE s,. i, l- .,y DRESDEN - S'.,udent Text. Book 4, Chapter 5, page 6. Drv3J ~ leu - M -, Aak 4 CA,yh.s y 4i7 ANSWER 2.05 (1.50) 1. The use of two pumps with one heat exchanger would exceed the design capacity flow of the heat exchanger. ( excessive vibration). 2. The use of two heat exchangers with one pump would approach the load limits for the pump. J. gu.;t gocc.Jpme. A, ea<e*}y sv.a. s paews. (d**>~r,.) (2 @ 0.75 ea = 1.5) REFERENCE DRESDEN - Student Text, Book 3, Chapter 17, page 21. h ord ~ l *3se-- f % i Wh S k. Sht k.L.b. 2.). g /9 p ANSWER 2.06 (2.00) A. GROUP V B. GROUP III C. GROUP IV D. GROUP II (4 0 0.5 ea = 2.0) (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)

- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pcgo 19 ~ SYSTEMS REFERENCE DRESDEN ~ Student Text. Book 3. Chapter 16, page 60. ANSWER-2,07 (1.50) 1. Fuel Pool Cooling and Cleanup System. 2. Reactor Water Cleanup System. 3. Isolation Condenser. 4. The ECCS fill header. { alternative supply} s. et dS1%,1.et $sd 14 (any 3 @ 0.5 ea = 1.5) 6. d A s. A.,, Au: Te-s A-rydw. REFERENCE DRESDEN - Student Text, Book 2. Chapter 4, page 27. Orod w k na~ M ~, L d w. i fsd - tsv j*);.18, Drak P !I 0's - M-35, pt.no, 4 p.11. ANSWER 2.08 (2.50) 1. C 2. A&E 3. D 4. B (5 @ 0.5 ea = 2.5) ~ REFERENCE DRESDEN - Student Text, Book 2, Chapter 8, pages 13, 14, and 15. f (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) 7y,.,_ ,.,_.r,.,__,y y -..-r

2. PLANT DESIGN INCLUDING SAFETY ~AND EMERGENCY Pcga 20-EX.S.IIMS - ANSWER 2.09 (2.00) A. 1...N.".M..!.. T*# #* "' B. Instrument Bus C. Essential Service Bus D. RPS Bus A Ecww/S* wee As E.A " " "^' w -+~ ~ P.cf: cacc supplica d::: n:t identify currir +e s Ch nnel E. Ch::h at F:cility ^- F. 125V DC (T.B. Main Bus #2A-1 ESS Div I & Non-ESS) i (6 @ 0.33 ea = 2.0) REFERENCE DRESDEN - Student Text, Book 2, Chapter 13, Table 5. Chapter 14, page 4. pr.*sJ~ Ins-l'l, pfe.'a-T~rb6. y rs-d /3. Dr <>./~ d.c Ja.- &a 4 sca-L f.s. 3 /*n*A h # A S /**A / f~S J / Sa,,. % f /* r. T d l- ). l i (***** END OF CATEGORY 2 *****) I

53. INSTRUMENTS'AND CONTROLS -Prga 21 ' ANSWER 3.01 .(2.50) 1. Supply breaker trip on a running-feed pump. 2.

A standby RFP-is selected.

3. ~ Suction pressure above the trip setpoint_of 120 psig. 4. At least one ventilation fan is operating. 5. .011 pressure is greater than 20 psig 6. CRx water = level 11ess than 55" [5 required 9 0.5 each] -(2.5) REFERENCE-Dresden' Student Text #4, Book 2, PP. 23 and 24-ANSWER 3.02. (2.00) 1. 'A loss of control power preventing breaker operation. (1.0) 2. Loss of trip circuit-continuity which would prevent the breaker from being-tripped from the control. room. (1.0) ~ REFERENCE ~ DRESDEN - Student Text, Book 2, Chapter 14, page 12. 4 ANSWER-3.03 (1.00) 1. Invalid probe data 2. Card pulled 3. Loss of power supply 4. Internal logic stall (any 3 @ 0.33 ea = 1.0): ( REFERENCE DRESDEN - Student Text, Book 1, Chapter 7, page 14. ~ l i 4 (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) I l

7" 13.' INSTRUMENTS AND CONTROLS Para 22 ANSWER. 3.04 (1.50) A. The meter reads the core' bulk thermal power. B. The meter indicates the number of assigned LPRM's which-are in " operate"

C. ~The meter. reads the percentage of rated total-recirculation loop flow.

(3 0 0.5 ea = 1.5) s REFERENCE DRESDEN' Student Text, Book 1,_ Chapter 12, page 15. i ANSWER 3.05 (1.00) Throttling of the 3-valve is accomplished by placing the HAND-RESET Switch in the " HAND" position and operating the 3-valve Control Svitch -from CLOSE to AUTO as necessary to achieve the desired cooldown rate. ~ ~ REFERENCE DRESDEN - Stud'ent Text, Book 3, Chapter 12, page 7. ANSWER-3.06 (1.00) =1. Instantaneous Overcurrent 2. Time-delay Overcurrrent 4 0.25 3. Undervoltage ( $ 9 EH4HF ea = 1.0 ) 4. f4 ))y p/n nk l,;. Aff. J..),./c (M) n REFERENG DRESDEN - Student Text, Book 3, Chapter 13, page 4. L (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) i

a3. ' INSTRUMENTS AND CONTROLS Pega 23 ANSWER 3.07. (2.00) ( 1*. 0 ) A. ~15 (+/-10%) inches. (Upon full scram, the master controller dual setpoint current supply output is reduced by 50%. This effectively halves the the sepoint ad level as demanded by&ena a emer m.just tape. ) A o a J " so-r*Jpd W B.4% psk p% mle.

r. aster contre hal actp:1.t current -rrir ^"' rut i ^-.

(1.0) r;tm mt. tc ern:1 hen th: scr-- 4-r--et?- REFERENCE Book 2, Chapter 6, page 11. P y /O-DRESDEN - Student Text, ANSWER 3.08 (2.00) A. OPEN B. CLOSED C. OPEN ClasED D. 2 OrZ" (4 @ 0.5 ea = 2.0) REFERENCE DRESDEN - DOP 2300 -T1 and Student Text, Book 3, Chapter 10. .0pn k. b n -- l L., Hyt T. T~L!~ L - (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) t

-[3. INSTRUMENTS AND CONTRQLS Paca 24 ' ANSWER 3.09 (2.00) The system is initiated when one of the following conditions exist: 1. High Drywell Pressure-(Oi.25)' 2.0 psig (0.25)

• • OR **

2. Low Low Water-Level (0.25) -59 inches (0.25)

1. 1. and ##

Reactor Low Pressure (0.25) Less than 350 psig (0.25)

• • OR **

3. Low Low Water Level (0.25) Maintained for greater than 8.5 minutes. (0.25) (2.0) REFERENCE DRESDEN - Student Text, Chapter 14, page 13. (***** END 0F CATEGORY 3 *****)

y- - e 4. PROCEDURESi-NORMAL. ABNORMAL. EMERnFNCY Paco 25 AND RADIOLOGICAL CONTROL a= -ANSWER 4.01 (2.50) 1. Reactor scram 2. Group 1 isolation 3. Chimney isolation valve closure 4. Air ejector suction valve closure 5.- Mechanical vacuum pump trip (5 0 0.5 each) (2.5) REFERENCE DRESDEN - DGA 16, page 1. ANSWER' 4.02 (1.00) A. By attempting to withdraw the rod past position 48. (0.5) B. Rod overtravel alarm. (0,5)- REFERENCE DRESDEN - DOP 400-1, P 4 DOP 400-2 ANSWER 4.03 (1.00) The timing has been found to be out of specification (1.6 sec) and controling withdrawal by notch withdrawal will elimir. ate a possible "short period" startup. (1.0) REFERENCE DRESDEN - Operating Order #29-86 (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

F 7 4. PROCEDURM - NORMAL. AB 6 EX P co 26 AND RADIOLOGICAL CONTROL-o. ANSWER 4.04 (2.00) 1. RPV water level cannot be maintained above +8 inches OR cannot be determined. 2. RPV pres,sure above 1060 psig. 3. Drywell pressure above 2.0 psig 4. A condition which requires a reactor scram, and reactor power is either: a. above 6% OR b. cannot be determined. (4 9 0.5 ea = 2.0) REFERENCE DRESDEN - DEOP 100 ANSWER 4.05 (1.75) SRM's One in the core quadrant where fuel or control rods are being moved (0,5) and one in an adjacent quadrant (0,5) IRM's One in each reactor protection system trip channel may be bypassed with the following restrictions: (0.25) 1. at least one IRM must be in service in the core quadrant where refueling is taking place.(0.25) 2. Nomore5$dfoneofthefourcenterIRM'smay be bypassed during core alterations (0.25) (1.75) REFERENCE DRESDEN - DFP 800-1 (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

r D' -4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY .Pcco 27 AND RADIOLOGICAL CONTROL O' ANSWER 4.06 (2.00) 1. When unstable conditions exist. 2. If load must be dropped below the Low Megawatt Electric Limit. 3. If load must be raised above the High Megawatt Electric Limit. 4. During weekly Turbine Surveillance and Control Rod exercises. 5. Loss of Control Rod indication or any rod movement. 6. Whenever there is any unusual reactor steam system pressure changes. 7. If there are any unusual large condenser vaccuum problems. (u?/ lso **cyd 5p,#.4 u ds m.,,lh V4- +',4 (any 4 9 0.5 ea = 2.0) ea. re moed 4 m S&C) REFERENCE DRESDEN - DOP 5670-1, page 2. ANSWER 4.07 (0.50) 100 TO 135 F (any answer within this range is acceptable) REFERENCE DRESDEN - DOP 5750-11, page 1. ANSWER 4.08 (1.00) 1. IMMEDIATELY insert control rods until a period of 30 seconds or more is indicated. 2. Contact the Shift Supervisor before continuing control rod withdrawal. (2 @ 0.5 ea = 1.0) (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

o. -A. PROCEDUEES - NORMAL. ABNORMAL. EMERGEH.QX. Pega 28 MD RADIOLOGICAL COtiTROL REFERENCE-DRESDEN - DGP 1-1, Attachment A. ANSWER 4.09 (2.75) UNIT 2 1. Reduce the speed of the MG set 2A to the acceptable speed for single loop operation. (0.25) 2.- Trip recirculation MG set 2B locally at Bus 32 (0.25) 3.- As-long as the reactor remains in a safe condition, do not (0.75) attempt to shut down until 125v DC can be restored. UNIT 3 1. Manually scram the reactor and follow DPG 2-3. (0.75) 2. Trip all reactor building vent fans. (0.25) 3. Start either standby gas treatment train. (0.25) 4. Trip OCB's 9-10 and 10-11 locally at the 345 KV Switchyard. (0.25) ( w,II h.> . w rww. ,4 W., is 4 4. ua4 2. d sg4s/ ) REFERENCE DRESDEN - D.OA 6900-2, pages 2 and 7. (***** END OF CATEGORY 4 *****) (********** END OF EXAMINATION **********)

r o TEST CROSC REFERENCE Pc3a 1 CQUESTION VALUE REFERENCE 1.01 1.00-ZZZ0000062 1.02 1.00 ZZZ0000070 1.03 1.50 2ZZ0000063 1.04 2.00 ZZZ0000064 1.05 -1.50 ZZZ0000065 1.06 2.00 ZZZ0000066 ~1.07 2.00 ZZZ0000067 1.08 2.50 ZZZ0000068 1.09 1.50 ZZZ0000069 15.00 2.01 2.00 ZZZ0000071 2.02 1.00 ZZZ0000072 2.03 2.00 ZZZ0000073 2.04 1.25 -ZZZ0000074 2.05 1.50 2ZZ0000075 2.06 2.00 ZZZ0000076 2.07 1.50 ZZZ0000077 2.08 2.50 ZZZ0000078 2.09 2.00 ZZZ0000079 15.75 3.01 2.50 ZZZ0000080 3.02 2.00 ZZZ0000081 3.03 ~ 1.00 ZZZ0000082 3.04 1.50 ZZZ0000083 3.05 1.00-ZZZ0000084 3.06 1.00 ZZZ0000085 3.07 -2.00 ZZZ0000086 3.08 2.00 ZZZ0000087 3.09 2.00 ZZZ0000088 15.00 4.01 2.50 ZZZ0000089 4.02 .1.00 ZZZ0000090 4.03 1.00 ZZZ0000091 4.04 2.00 ZZZ0000092 4.05 1.75 ZZZ0000093 4.06 2.00 ZZZO900094 4.07 0.50 ZZZ9000095 4.08 1.00 ZZ40000096 4.09 2.75 ZZZ0000097 14.50 60.25

n s-e: U. S. NUCLEAR REGULATORY COMMISSION , SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION o FACILITY: DRESDEN REACTOR TYPE: EWR-GE3 -DATE ADMINSTERED: 87/01/26 EXAMINER: BISHOP. M. CANDIDATE INSTRUCTIONS TO CANDIDATE: Read the attached instruction page carefully. This examination replaces the current cycle facility administered requalification examination. Retraining requirements for failure of this examination are the same as for failure of a requalification examination prepared and administered by your training staff. Points for each question ~are indipated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up four (4) hours after the examination starts. % OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY zMti 2,.. S. THEORY OF NUCLEAR POWER PLANT 15.00

1. 8 OPERATION, FLUIDS,AND THERMODYNAMICS 24.11 15.00 2"

10" 6. PLANT SYSTEMS DESIGN, CONTROL, F# AND INSTRUMENTATION U is.s s 2S.zo 7. PROCEDURES - NORMAL, ABNORMAL, m, mm-75 EMERGENCY AND RADIOLOGICAL CONTROL 25.20 15.25 24.00" 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 40.So -6;.26 "- Totals 78 Final Grade All work done on this examination is my own. I have neither given nor received aid. Candidate's Signature e I i l l i i

m NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS buring the administratica of this examination the following rules apply: 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties. 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating. 3. Use black ink or dark pencil only to facilitate legible reproductions. 4. Print your name in the blank provided on the cover sheet of the examination. 5. Fill in the date on the cover sheet of the examination (if necessary). 6. Use only the paper provided for answers. 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet. 8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer sheet. 9. Number each answer as to category and number, for example, 1.4, 6.3.

10. Skip at least three lines between each answer.

11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.

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

13. The point va.lue for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.

14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.

15. Partial credit may be given.

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

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

17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination.

This must be done after the examination has been completed. I

r: v .? .18'. When you complete your examination, you shall: a. Assemble your examination as follows: (1) ~ Exam questions on top. (2) Exam aids - figures, tables, etc. (3) Answer pages including figures which are part of the answer. b. Turn in your copy of the examination and all pages used to answer the examination questions. c. Turn in'all scrap paper and the balance of the paper that you did not use for answering the questions. d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked. 6 6

c ,-t-- hEEORYOFNUCtrAR'POWERPLANT-OPERATION. Paco 4 '? 5. FLUIDS.AND THERMODYNAMICS [ QUESTION 5.01' (1.00) _ Wh'en using the neutron monitoring instruments to identify a. _ partially uncovered core during accident conditions,. the neutron flux MIGHT show a drop-off as the detector enters the uncovered region of the core f" Why'MIGHT this indication occur on the nuclear instruments? QUESTION 5.02-(2.00). TRUE or FALSE 7 Answer the following concerning factors affecting control rod worth. -a. The worth of a control rod is a direct function of-the thermal-neutron flux to which it is exposed. b.. A control rods. worth decreases as moderator temperature' increases. c. An'the void content inceases, the control rod worth increases. d. Control rod worth increases as the fuel temperature increases. -QUESTION 5.03 (1.50) Briefly explain how a control rod withdrawal of one or two notches can result in a' decrease in bundle Power. QUESTION 5.04 (1.50) What are.TWO reasons for performing rod pattern exchanges? QUESTION 5.05 (2.00) Discuss how xenon changes for a reactor power increase from 50% to 100%. Assume steady state operation at 50% power for two days prior to the maneuver. Include time frames and why they occur in your

• answer.

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

y. c y a

1. 5.

TernaY OF MUCfRAR' POWER PfANT OPERATION. Poco 5 FLUIDS.AND THERMODYNAMICS .A QUESTION 5.06-(2.00)-. Refer to the' attached Figure 1, Keff.vs. Exposure and explain why the change in Keff occurs from the following points: (TWO reasons. required for each set of points) a. A.to B .b. B to C ^ . QUESTION-5.07 (1.50) Match the failure mechanism in Column 1 and the limiting condition from Column 2 with the associated power distribution limits.(A-C) below. s. Linear Heat Generation Rate'(LHGR) b.- -Average Planer Linear Heat' Generation Rate (APLHGR) c. Minimum Critical Power Ratio (MCPR) Column:1 Column 2~ Failure Mechanism Limiting Condition ~ l. Fuel clad. cracking due to 1. 1% plastic strain ' lack of cooling caused by DNB. 2.

Fuel clad. cracking due to high 2.

Prevent transition boiling stress from pellet expansion. 3. Gross clad failure due to decay-3. Limit clad temp. to 2200 F. heat an'd stored heat following a LOCA. QUESTION 5.06 (2.00) During a cooldown of the reactor vessel following a control room evacuation, reactor pressure decreases from 800 pais to 450 pais in 30

minutes, a.

'What is the APPROXIMATE cooldown rate per hour? (1.5) b. Has the tech spec. cooldowr. limit been exceeded? Explain your answer. (0.5) (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

i 5. THEORY OF NUCfRAR POWER PLANT OPERATION. Pc 6 FLUIDS.AND THERMODYNAMICS . QUESTION 5.09 (1.50) Provide a brief discussion how natural circulation flow ~is achieved in the reactor. Include in your discussion the flow path and the motive force for the flow. (***** END OF CATEGORY 5 *****)

>:.s .i .'Pago 7 8. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION s. ~ QUESTION' ~6.01 (3.00). Concerning-the Isolation Condenser System: a. List THREE systems IN ORDER of preference which can supply makeup water..to the;shell side of the Isolation Condenser.. ~ (1.0) b., What are FOUR' indications of a possible leak'in the tube bundle of the Isolation Condenser? .(2.0) - QUESTION 6.02 (1.00) .What THREE conditions must be met in order for the Rod Worth Minimizer to be AUTOMATICALLY bypassed? QUESTION 6.03 (2.50) Concerning the Reactor. Protection System; a. The shift supervisor is approached by an electrician who wants-to transfer the RPS "A" bus to its Reserve power supply.in order to perform preventive maintenance on the MG set. The reactor is at .754 power and operating with a Half-Scram signal on RPS channel B due to an inoperable pressure transmitter. Should the shift supervisor authorize the transfer? EXPLAIN your answer.- (1.0) b. What THREE conditions will cause the Electrical Protection Assemblies (EPA breakers) to open? (1.0)- c. Why is the RPS Reserve Power interlocked to prevent supplying both busses at'the same time? (0.5) (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

n ,J, A a, pt.AMT SYSTEME MAIGNi CONTROL. AND INSTRUMENTATION POO3 8 ~ QUESTION 6.04 (2.00) In the recirculation flow' control system: a.~ What~are 4 conditions which will.~cause the scoop tube actuator to " lock up"'(setpoints not required.) [1.0].? ~b.-What e.ction must.be taken by the control room operator prior to-resetting a locked out scoop tube ? (Include WHY this must.be done in your answer.).[1.0). -QUESTION. 6.05 (1.50) What are five methods available to'the operator in-the control room to determine proper operation of the standby liquid control system after manual initiation 7 (Do NOT include annunciators or alarms) QUESTION 6.06 (2.00) List-four automatic valve or pump actions other than main steam -line isolation valve closure, which occur as a result of a 3' times normal main steam line radiation signal. QUESTION 6.07-(2.00) The plant is operating at 100% power. A complete loss of instrument air ' occurs. With NO Operator action, answer the below questions? A. The level in the hotwell would ( increase, decrease, remain the same ). B. Service water flow to the TBCCW heat exchangers would ( increase, decrease, remain the same ). l [ C. Scram Discharge Vol. drain valves would ( open, close, remain as is ). D. Main feedwater regulating valves would ( open, close, remain as is ). i ? l l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) 1

g 2 r; s t/[ / j. < Page 9 ' 8 '. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRQt!Etu'A? ION 1' a. / l -r vi, / I t CUESTION 6.08 (1.00) Assume a core' spray initiation signal is present, list the trips available for.the 2A Core Spray Pump. Setpoints are not requiredt- - rY I a ' J 4 i e 1 (***** END OF CATEGORY 6 *****) 1

(-~ f. 1

11... P2QCEDURFA - NORMAL. ABNORMAL. EMERGENCY Paco - 10 ID RADIOLOGICAL CObTROL L.y r

F QUESTION 7.01 ( ) ti Con }1. has the operators verify that Steam i Lt 't DEOP 100-2, RP Pre su Cooling is NO 're e criteria are used to determine that Steam C NOT REQUIRED 7 (Include applicable setpoints) l ,\\L QUESTION 7.02 (1.50) l' What are THREE of the four immediate operator actions per DOA 3700-1 for a total loss of Reactor Building Closed Cooling water? (Include I - applicable setpoints) l l . QUESTION 7.03 (1.64) l

a. Complete the following statements;

^ 1. Immediate operator actions for a Recirculation Pump trip (DOA 202-1) requ.i.re the Unit 2 NSO to reduce running pump speed to (b.1.dnk) percent of rated speed. (0.5) L 2. If the same incident occurred on Unit 3, the NSO would be required to reduce running pump speed to (blank) percont of' rated. (0.5) L

b. Fhs must,the Unit 2 recirculation pump be operated at a different speed l

that the Unit 3 pump during single loop operation? (0.5) QJESTION 7.04 (1.00) According to Operating Order #1-86, what immediate action is to be performed by the NSO when a Unit 2 Reactor Building Ventilation Radiation Monitor initiates the Standby Gas Treatment train? l i o (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) l +--

x i',' 7. - PROCEDUFFR NORMAL. ABNORMAL. EMERGENCY Peca 11 AND RADIOLOGICAL CONTROL 1 l. . QUESTION' 7405 (2.50) What are,the'FIVE entry conditions,for DEOP 200, Primary Containment Control?j(Includeapplicablesetpoints) ,p, H QUESTION 7.06 (2.00) Y Accordin's to DEOP 490-3 RPV Flooding, when is BORON injection required?-(Include applicable setpoints) QUESTION 7.07 (2.00)

a. During operation at 80% power an event causes feedwater l

temperature to decrease 25 F. What specific operator action is required by the feedwater temperature decrease 7 (Include i applicable setpoints) 1

b. Fill in the blanks with the correct choice from the lists provided:

b The cause of the event cannot be corrected and feedwater I temperature continues to decrease. If the temperature drops more than a F, then b Choices for a: Choices for b:

1. 70.

1. Scram reactor.

2. 100

2. Insert CRAM arrays 3.

130

3. Trip turbine

4. 160

4. Reduce power to < 25%

QUESTION 7.08 (1.00) Per DOA 600-1. Transient level cont $ol, when must the reactor be manually scrammed on an increasing level transient ? l l (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) 1p x' ! '.

~ 7.~ PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pasa 12 AND RADIOLOGICAL CONTROL e QUESTION' 7.03 (1.00) A requirement of_DOA 040-1, Slow Leak, is to shutdown hydrogen addition (Unit 2 only). Why is this done ? QUESTION 7.10 (2.75) List the Immediate Operator Actions for EACH Unit upon a loss of 125v.DC power supply. DOA 6900-2, /,u e/ v,14 J IM uA h* Syr'. / (***** END OF CATEGORY 7 *****) ~ -~,4,- e.., -, n.. m.

~ IDMINISTRATIVEPROCEDURES'. CONDITIONS. 8. Pace 13 ,AND LIMITATIONS QUESTION 8.01 (1.00) What-isLth'e' MAXIMUM valid length that each of the following may remain' in effect according to CECO Radiation Protection' Standards? a. Type 1.RWP.

b. Type 2 RM?

QUESTION 8.02 (2.75) According to EPIP 200-20, Control Room Evacuation / Shutdown, a. WHY is the mode switch left in RUN after scramming the reactor? (0.5)

b. What THREE additional-actions must be performed by the NSO before leaving the control reem when evacuation of the main control.

room is required? (2.25) QUESTION-8.03 (1.50) What three " state and federal" agencies are always notified if a GSEP-classified event occurs, per EPIP 100-C1, Station Director Checklist o.f Initial GSEP-Responsibilities? QUESTION 8.04 (3.00) a. For. unit 2 and unit 3 what are the Tech Spec limits and actions for identified and unidentified leakage from the primary system ? [1.0]

b. For the intent of the Tech Specs, how does the operator determine identified and unidentified leakage ? [0.5]

c. What additional limits and requirements are placed on unit 2 leak rates ? [1.0)

d. What additional limit and requirement are placed-on unit 3 leak rates ? [0.5]

l j '(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) l l i

I8. ~ ADMINISTRATIVE PRCCEDURES. CONDITIONS. Para 14 AND LIMITATIONS. ' QUESTION' 8.05 (2.00) Concerning control rods: a. List 3 conditions that, per the Tech Specs, will~cause the rod to ~be inoperable.

b. How many~ control. rods may be inoperable during power operations ?

QUESTION 8.06 -(2.00) Identify each of the following as TRUE or FALSE. a. Personnel protection cards alone are sufficient to protect personnel for locally operated equipment,

b. Personnel protection cards may be transfered to another crew but must not be transfered to another job.

.c. The " Competent Individual" signing the "After Outage" portion of an Equipment Outage Checklist confirms the outage checklist is adequate.

d. An outage may not be partially cleared.

QUESTION 8.07 (1.00) Per the Dre'sden Tech Specs, when must the isolation condenser be operable ? t (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)

g..-

4 jl. ADMINISTRATIVE PROCEDURES. CONDITIQH32 Peca 15

AND LIMIIATIONS QUESTION 8.08.

(2.00)' What are the Dresden radiation worker ADMINISTRATIVE limits for external whole body penetrating radiation for each of the~following time periods? [Do not consider 5(N-18) limits.] ~ a. Daily b. Weekly-c. Quarterly d. Yearly 1 (***** END OF CATEGORY 8 *****) (********** END OF EXAMINATION **********) L

I si p g g .w3 et0 G. E Y 44 m E D ~ e .I m ~ g 4w 35' { I 4 l s i i i O 2 = = = 2 H*M 6 4

EQUATION SHEET f = ma v = s/t 2 Cycle efficiency = Net Work (out) - w = as s = v,c + at kg W E = aC a = (vg - v,) /t 1 g=y +g A = AN A = A,e gg = hay y g PE = ash a = 6/t 1 = in 2/tg = 0.693/tg W = v&P- , t (eff) = (t,,)(e ) s AE = 931Am (k**b) k = k AT I = I,~h p .,4=uAAT z = z,,- = .~Pwr = U' a" go / m -x g I=I o M (t). T71. = 1.3/u y=y to e /T HVI. = 0.693/u t P=P '5175. = 26.06/T ~ 7 = 1.44 DT SCR = s/(1 - K,gg) fA oT = S/(1 - E,gg,) SUR = 26 CR g g 1( e f)1

• 2(

eff)2 T = (1*/p ) + [(f p)/Aeff,] ~ T,= 1*/ (p - fy M " 1/(1 - K gg) = CR /CR g O T = (I - p)/ 1,ggo M = (1 - K,gg)g/(1 - K,gg)g 8 * ( eff I aff " #eff eff SDM = (1 - Koff)/Keff /K [1*/TK,*gg.] + [I/(1 + 1,gg )] 1* = 1 x 10 seconds ~ T p= P = r$7/(3 x 1010)

• I A,gg = 0.1 seconds A

E = Na Idgg=Id22 UATER PARAMETERS Id =Id g 2 2 1 gal. = 8.345 lba R/hr = (0.5 CE)/d g,,,,,,) l 1 gal. = 3.78 liters R/hr = 6 CE/d (feet) I ft3 = 7.48 gal. MISCEI.L\\NEOUS CONVERSIONS 3 1 Density = 62.4 lbm/ft 1 Curie = 3.7 x 10 Odps 3 Density = 1 gm/cm 1 kg = 2.21 lba i Heat of varori:stion = 970 reu/lba 1 hp = 2.54 x 103 BTU /hr Hest of fusica = 144 Bru/lbm 1 Mw = 3.41 x 10 Beu/hr 6 1 Atm = 14.7 psi = 29.9 in, l'g. 1 Btu = 778 f t-lbf 1 ft. H O = 0.4333 lbf/in 1 ' inch = 2.54 cm 2 F = 9/5'c + 32

• C = 3/9 ('T - 32)

Vesume, ft'/m EMhelpy.Chr/Ils Emresy. Stas /us a F ) water Eves Steem Water Evep Steam Water

tve, geoem j

A A A e, sq s, t t s 't 'as 's at 0.08859 E01602 3305 3305 -0.02 1075.5 1075.5 0.0000 2.1873 2.1573 32 35 0.09991 0.01602 2948 2948 3.00 1073.8 1076.8 0.0061 2.1706 2.1767 35 40 0.12163 0.01602 2446 2446 8 03 1071.0 1079.0 0.0162 2.1432 2.1594 40 45 0.14744 0.01602 2037.7 2037.8 13.04 1068.1 1081.2 0 0262 2.1164 2.1426 45 50 0.17796 0.01602 1704.8 17044 18.05 1065.3 1083.4 0.0361 2.0901 2.1262 50 60 0.2H1 E01603 12074 12074 28.06 1059.7 1087.7 0.0535 2.0391 2.0946 40 TO 0.3629 0.01605 868.3 868.4 38.05 1054.0 1092.1 0.0745 1.9900 2.0645 70 to 0.5068 0.01607 633.3 633.3 48.04 1048.4 10 %.4 0.0932 1.9426 2.0359 30 90 0.6081 C.01610 448.1 468.1 58.02 1042.7 1100.8 0.1115 1.8970 2.0086 90 100 0.9492 0.01613 350.4 350.4 68.00 1037.1 1105.1 0.1295 1A530 1.9825 100 Ele IJ750 IL01617 265.4 265.4 77.98 1031.4 110BJ 0.1472 1A105 1.9577 110 130 1.6927 OA1620 203.25 203.26 37.97 1025.6 1113.6 0.1644 1.7693 1.9339 120 130 22230 0.01625 157.32 157.33 97.M 10192 1117A 0.1817 1.7295 1.9112 130 140 2.8892 0.01629 122.98 123A0 107.95 1014.0 1122.0 0.1985 1.6910 1.8395 140 150 3.718 0.01634 97.05 97A7 117.95 1008.2 1126.1 0.2150 1.6536 1.8646 150 360 4.741 OA1640 77.27 77.29 127.M 1002.2 1130.2 0.2313 1A174 13487 See 170 5.993 0.01645 62.04 62.06 137.97 996.2 1134.2 OJ473 1.5822 1A295 170" 130 7.511 0.01651 50.21 50.22 144.00 990.2 1138.2 0.2631 1.H00 12111 180 330 9.340 0.01657 40.94 40.M 158.04 984.1 1142.1 0.2787 1.5148 1.7934 190 300 11.526 0.01664 33.62 33.64 168.09 977.9 1146.0 0.2940 1.4824 1.7764 300 210 14.123 0.01671 27.80 27.82 178.15 971.6 1149.7 0.3091 1.4500 1.7400 210 212 14.696 0.01672 26.78 26.80 180.17 970.3 1150.5 OJ121 1A447 1.1388 212 320 17.184 0.01678 23.13 23.15 188.23 M S.2 1153.4 0.3241 1.4201 1.7442 320 230 20.779 0.01685 19.364 19.381 198.33 958.7 1157.1 OJ308 1.3902 1.7290 230 240 24.968 041693 16.304 16.321 205.45 952.1 1160.6 03533 1.3608 1.7142 340 250 29A25 OA1701 13.802 13.819 218.59 945.4 1164.0 0.3677 1.3323 1.7000 250 260 35.427 0.01709 11.745 11.762 228.76 938.6 1167.4 0.3819 1.3043 1.8082 3GO 270 41.854 0.01718 10.042 10.060 238.95 931.7 11704 03960 1.2700 14729 270 200 49.200 0.01726 S.627 8.644 249.17 924.6 1173A 0.4098 1.2501 1.6699 230 290 57.550 041736 7.443 7.440 259.4 917.4 1176 3 0.4236 1.2238 14473 250 300 57.005 0.01745 6.444 6.466 269.7 910.0 1179.7 0.4372 1.1979 14351 300 310 77.67 0.01755 5.609 5.626 280.0 902.5 1182.5 0.4506 1.1726 1.6232 310 320 39.64 0.01766 4.896 4.914 290.4 894.8 1185.2 0.4640 1.1477 1.6116 320 340 117.99 0.01787 3.770 3.788 311.3 878.8 1190.1 0.4902 1.0990 1.5892 340 360 153.01 0.01811 2.939 2.957 332.3 862.1 1194.4 0.5161 1.0517 1.5678 360 3a0 195.73 0.01836 2.317 2.335 353.6 844.5 1198.0 0.5416 1.0057 1.5473 380 400 247.26 0.01864 1.8444 1.8630 375.1 325.9 1201.0 0.5667 0.9607 1.5274 400 420 305.78 0.01894 1.4808 1.4997 396.9 806.2 1203.1 0.5915 0.9165 1.5080 420 440 381.54 0.01926 1.1976 1.2169 419.0 785.4 1204.4 0.6161 0.8729 1.4890 440 ' 440 466.9 0.0196 0.9746 0.9942 441.5 763.2 1204.8 0.6405 0.8299 1.4704 450 1 480 566.2 0.0200 0.7972 CA172 464.5 739.6 1204.1 0 6648 0.7871 1.4516 des [ 500 600.9 0.0204 0.6545 0.6749 487.9 714J 1202.2 0.8390 0.7443 1.4333 500 520 - 812.5 0.0209 0.5386 0.5596 512.0 687.0 1199.0 0.7133 0.7013 1.4146 520 540 962.8 0.0215 0.4437 0.4651 536.8 657.5 1194.3 0.7378 0.6577 1.3954 540 560 1133.4 0.0221 0.3651 0.3871 562.4 625.3 1187.7 0.7625 0.6132 1.3757 540 580 1326.2 0.0228 0.2994 0.3222 589.1 589.9 1179.0 0.7876 0.5473 1.3550 580 600 1543.2 0.0236 0.2438 0.2675 617.1 550.6 1167.7 0.8134 0.5196 1.3330 GOO $20 1786.9 0.0247 0.1962 0.2208 646.9 506.3 1153.2 0.8403 0.46S9 1.3092 620 640 2059 9 0.0260 0.1543 0.1802 679.1 454.6 1133.7 0.8666 0.4134 1.2821 640 640 2365.7 0.0277 0.1166 0.1443 71s.9 392.1 1107.0 0.8995 0.3502 1.2498 660 660 2700.6 0.0304 0.0008 0.1112 758.5 310.1 1068.5 0.9365 0.2720 1.2086 680 700 3094.3 0 0366 0.0386 0.0752 822.4 172.7 995.2 0.9901 0.1490 1.1390 700 705.5 2208.2 0.0508 0 0.0508 906.0 0 906.0 1.0612 0 1.0612 705.5 PROPERTIES OF SATURATED STEAM AND SATURATED TABLE A.2 WATER (TEMPERATURE) A.3

'5. THEQRY OF NUCLEAR POWER PLANT' OPERATION. Pesa 16 . FLUIDS.AND THERMODYNAMICS hn:RC0*Y ANSWER 5.01 (1.00) With a loss of water level, the instruments which indicate the thermal flux would be in a fast flux field and indicating low. REFERENCE Dresden, Lesson Plan, Degraded Core Response, P. 5 ANSWER 5.02 .(2.00) a. TRUE b. FALSE c. FALSE d. FALSE 4 @ 0.5 each = 2.0 ~ REFERENCE Dresden, Lesson Plan, Reactor Physics Review, P. 36 ANSWER 5.03 -(1.50) The steam bubbles generated by withdrawal of a shallow rod increase the void fraction, which adds negative reactivity off-setting the positive reactivity effects of the rod withdrawal. REFERENCE Dresden, Lesson Plan, Reactor Physics Review, P. 39 ANSWER 5.04 (1.50) 1. . Maximize fuel burnout (by flattening the flux). (0.'5 % 2. Limit the power peaking (that would occur from a subsequent rod withdrawal after the buildup of Pu-239). 3 0. ' 5 P- ~ f&d* O.4e*I d.,d /qur. 3. a (by 2@ D TS N (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) .~

5. THEORY OF NUCTRAR POWER PLANT OPERATION. Paga 17 FLUIDS.AND THERMODYNAMICS REFERENCE-Dresden, Lesson Plan, Reactor! Physics Review, P. 41 Dr s/~ o l"J'~ 1%, C&J /d 6/,JQ Aso+ /fasen

d. I. b. 3). d. y +

e ANSWER 5.05 (2.00) 'The initial power increase will cause a decrease in the xenon concentration [0.5]. The low peak occurs 4-6 hours after the power increase [0.5]. The power production term takes over [0.5] and a new equilibrium level is reached in 40-50 hours [0.5]. REFERENCE Dresden, Lesson Plan, Reactor Physics Review, P. 45 ANSWER 5.06 (2.00) a. 1. Buildup of xenon. 2. Buildup of samarium, b. 1. Gadolina burnup. 2. PU-239 buildup. [4 @ 0.5 each = 2.0] REFERENCE Dresden, Lesson Plan, Reactor Physics Review, Figure 62, P. 47 ANSWER 5.07 (1.50) Column 1 Column 2 Failure Mechanism Limiting Condition a. LHGR 2 1 i l b. APLHGR 3 3 c. MCPR 1 2 [6 @ 0.25 each = 1.5] (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) I

5. THEQRY OF NUCLEAR POWER PLANT OPERATION. Poco 18 FLUIDS.AND THEBMQDYNAMICS REFERENCE Dresden, Lesson Plant Thermo, HTFF, Figure 11 ANSWER 5.08 (2.00) 800psig+14.7=815psih'[0.25] a. 450 psig +-14.7 = 465 psig> [0.25] From-Steam Tables 815 psig = 520 F [0.25] ~ 465 psig = 460 F [0.25] 60 F 120 F 520 - 460 = ---- X 2 = [0.5] 1/2 hr. Hour b. No. [0.25] The cooldown rate limit of 100 F is averaged over a one-hour period. [0.25] if y W' hel*A=d) (w,*Il else euys YES f Dresden, Requal Q/A Thermo, P. 6 Dresden, Tech. Spec. 3.6.A.1 ANSWER 5.09 (1.50) As the water around the fuel is heated, it' expands and becomes less dense [0.5]. The (weight) or downward force on the water in the downcomer area exceeds the force exerted by gravity on the water in the core [0,5]. The water in the jet pumps is pulled downward by gravity forcing the water in the core upward [0.5]. REFERENCE Dresden, Lesson Plan Thermo, HTFF, P. 15 l (***** END OF CATEGORY 5 *****) c ~ f. m

c6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION -Paga 19 e

ANSWER 6.01

-(3.00) a. .1.: Clean demineralized: water-system

2. Condensate transfer system

~ -3. Service Water / Fire Protection system [ 3 @ 0.25 each + 0.25 for correct order = 1.0 total ] b. Any 4 of the following @ 0.5 each;

1. High temperature in the inlet. steam line

2. High shell water temperature

3. Increasing shell water temperature Increasinglohcalradiationlevels 4.

5. Increased noise in condenser area

~

6. Steam or fog coming out the vent on the side of the reactor building Y "*4fGJ S~ ia-7s,J,.b~ C J z s;J. w in /.

REFERENCE Dresden, Student Text #3, P. 12-3 thru 12-5, a 12-10 MA' And", Jod 13*o-i. fu.4 ; yy/ fia OA s**- T*J~ Anh. ~ f4 / 4.ff /S. / ANSWER 6.02 (1.00) l

l. Steam flow above 20% [0.33]

2. Feed flow above 10% [0.33]

[

3. 60 second timer timed out [0.33]

REFERENCE Dresden, Student Text #1, P. 8-7 and lesson plan pg. 5 l l l l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) l { 1

• 6.

PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Paga 20 ANSWER 6.03 (2.50) a. No [0'.25), power to the RPS bus will be interupted during the transfer and a full reactor scram will reult. [0.75] b.

1. Overvoltage

2. Undervoltage

3. Underfrequency

[ 3 @ 0.33 each ] Co.1s] c.1.To prevent overloading the reserve instrument transformer EO-5-]

1. To, f+4w*** % vI*/*%='~ *f //5 desop) (f1.1*~

• e
• ~s. + ~t o p s y & wly

e. w s or p w - t.* sm**,r

[o. z3) +> so REFERENCE Dresden, Student Text #1, P. 15-4 & 15-5 4 A.uol 4*a~ r%~, /C sk/rok/ As!. S sk gu, y, Q, g,f s.z), y 1 y ~ ANSWER 6.04 (2.00)

a. Any four of the following @ 0.25 ea.

Function generator signal failure Drive motor bus low voltage Coupling lube oil low pressure (16 second time delay) Coupling lube oil high temp. Loss of power to scoop tube positioner. Asua seduse.w

b. Null voltmeter is used to match the speed from the tachometer with the speed demand from the speed controller [0.5) in order to prevent a flow transient during reset [0.5).

REFERENCE Dresden Student Text, Book 2, Chap. 2, pg 22 and table 1. E* 0 r*' Opsc/w froese(to

• 00t* 2o1 / sche. Aray M4 5*
• SW S.d: l'..t. yjs-5f4 (/'* Y '

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

i. 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Peas 21 t ANSWER 6.05 (1.50) Any five of the following @ 0.3 ea. . Pump discharge pressure . Pum:; indicating light _ illuminates . Tank level decreases . Squib valve indicating lights extinguished flow indicator . power decreases reactor water cleanup isolation. REFERENCE Dresden Student Text #3, pg 3-12 thru 3-15 ANSWER 6.06 (2.00) Any four of the following @ 0.5 ea. . Chimney isolation valve closes Air ejector suction valve closes Mechanical vacuum pump trips The off-gas line drain valve closes The off-gas sample vacuum pump suction valve closes REFERENCE Dresden Student Text #2, pg 16-3 h*G A~*~ N~, D.Mik Q, ksn, Lr. f.s.c. 4~%d s M. &}L J. ys &*6 J us~ M*~ r /Ma6 Se%, & dmL v./m akt *- gec s e., lup S.,,Ja. /~ c./s<- .Tsok+i6 W+ vwf a /m a to.r Scree r A lu n 0,r w (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) l

6- [6. PLANT' SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Pcga 22-

AtlSWER 6.07 (2.00)

-L Decrease s B. Increase C. Close. 1 D. Remain as-is-(4 0 0.5 ea = 2.0) REFERENCE Dresden - Student-Text, Book 4, Chapter 8, page 14. ANSWER 6.08 (1.00) 1. Instantaneous Overcarrent 2._ Time-delay Overcurrrent- .+ o.zs 3. Undervoltage (V @ er&& ea = 1.0) /G 4 p. 4 4,* A,// L fel. - 4. REFERENCE 'Dresden - Student Text, Book 3. Chapter 13, page 4. i l (***** END OF CATEGORY 6 *****) t I r L

7. PROCEDURES - NOBdAL. ABNORMAL. EMERGENCY _ Peca 23-AND RADIOLOGICAL CONTROL s -{ ANSWER. 7.01-(1.50) 1. Any i et on sy em is running .2. React level e d-' ntained above -143 inches 3. - Reac rpe re is less than 80 psig q. [ 3 0 0.5 each ] i REFERENCE .Dresden, DEOP 100-2 RPV Pressure Control, P. 3 ' ANSWER 7.02 -(1.50) Any 3 of the following 9 0.5 each;

1. Attempt to restore RBCCW flow 2.

If flow is lost and cannot be restored within one minute, trip recirculation pumps A and B

3. Manually scram the reactor if equipment damage appears imminent 4.

If drywell pressure reaches +2 psig, verify a reactor scram, enter DEOP 100 and DEOP 200. REFERENCE Dresden, DOA 3700-1 Loss of RBCCW, P. 1 (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)

~.: 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pcco 24 -AND RADIOLOGICAL CONTROL ~ ~ - x ANSWER =7.03 (1.50) a. 1. 14 3 - 2. 60 [:2 0 0.5 each'] b. Unit 2 has a different jet pump riser bracket [0.25] which allows larger vibrations during single loop operation [0.25] REFERENCE Dresden, DOA 202-1 Recirculation Pump Trip, P. 4 ANSWER 7.04 (1.00) A high radiation. trip is inserted on the Unit 3 Reactor Building Ventilation (or the Reactor Building Fuel Pool) radiation indicator and trip unit. REFERENCE Dresden, Operating Order #1-86, July 1, 1986 ANSWER-7.05 (2.50) 1. Torus water temperature above 95 deg F 2. Drywell temperature above 200 deg F 3. Drywell pressure above 2.0 psig 4. Torus water level above -1.5 inches 5. Torus water level below Tech Spec minimum (fig. 2-9). [ 5 @ 0.5 each ] (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) i , - ~ _

'W J,n-. ~~*7; PROCEDURES'- NORMALi ABNORMAL.~ EMERGENCY- -Paco 25 ~ AND RADIOLOGICAL CONTROL ~ -e .- REFERENCE ' Dresden, DEOP. 200 Primary Containment Control,- P.11; ' ANSWER -7.06-(2.00) Reactor cannot be shutdown and reactor power is >6% or unknown [ 1 '. 0) before Torus water temperature reaches 110-deg F [1.0] (also accept; All1 control rods'are not inserted to or-beyond Position 04 [1.0] and Torus water temperature'cannot be maintained below 110 deg F [1.0] - ) . REFERENCE Dresden, DEOP 400-3 RPV Flooding, P. 4 ~ ANSWER '7.07. (2.00)

a. Reduce core flow 7.5'MLBM/hr (or Reduce' core flow 3'MLBM/hr per 10 F drop.) [1.0]

. (Scram reactor) [0.5]

b. a. 3 (130) [0.5] b. 1.

REFERENCE Dresden DOA 3500-2 I . ANSWER 7.08 (1.00)- When the water level exceeds the range of the narrow range. instrument. (-t GO (*clu ) l-REFERENCE l .-Dresden DOA 600-1, pg 3 l t I (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) l'o l

7. PROCEDURES'- NORMAL' ABNORMAL. EMERGENCY. Pact 26 AND RADIOLOGICAL CONTROL d ANSWER 7.09' (1.00) To prevent H2 buildup in the drywell. REFERENCE Dresden DOA 040-1, pg 4 ' ANSWER 7.10 (2.75) UNIT 2 -1. Reduce the speed of the MG set 2A to the acceptable speed for single loop operation. (0.25) 2. Trip recirculation MG set 2B locally at Bus 32 (0.25) 3. As long as the reactor remains in a safe condition, do not (0.75) attempt to shut down until 125v DC can be restored. UNIT 3 1. Manually scram the reactor and follow DFG 2-3. (0.75) 2. Trip all reactor building vent fans. (0.25) 3. Start either standby gas treatment train. (0.25) 4. . Trip OC'B's 9-10 and 10-11 locally at the 345 KV Switchyard. (0.25) ( L),II ira ug+ res se s. a4 N*s" U /m 4 &o4 2 O.1f*5b b REFERENCE Dresden - DOA 6900-2, pages 2 and 7. (***** END OF CATEGORY 7 *****) L

... ].. , f 8. = ADMINISTRATIVE PRQCEDURES. CONDITIONS. Paga 27L AND LIMITATIONS .o- _' ANSWER -8.01 (1.00) 'a. .1 year .[0.5] b.~ Length of the job [0.5] ~ REFERENCE - Dresden, CECO Radiation Protection Standards, P. 11 & 13 ANSWER 8.02 (2.75)

a. So the reactor will isolate on less than 850 psig

[0.5] b. 1. Verify all rods in. 2. Trip the control rod drive pumps 3. Trip the turbine [ 3 @ 0.75 each ] (tailj obe m e o m k,sd w /ss. J e 4 p 6 ) REFERENCE Dresden, EPIP 200-20, Control Room Evacuation / Shutdown, P. 1&2 ANSWER 8.03 (1.50)

1. Illinois Emergency Services and Disaster Agency (IESDAJ,ess4#)

2. Illinois Department of Nuclear Safety ( IDNS f., 49uJ )

3. Nuclear Regulatory Commision (NRC)

[ 3 @ 0.5 each ] (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) ~-/

a. 8 '. ADMINISTRATIVE PROCEDURES' CONDITIONSi Paco 28 AND LIMITATIONS .g' ' REFERENCE. .Dresden, EPIP 100-C1, Station Director Checklist of_ Initial GSEP Responsibilities, Rev 5. P. 1 thru 3 4 ANSWER-8.04 (3.00)

a. For unit 2 & 3 : 5 rpm unidentified [0.25] and 25 gpm unidentified.

and identified total [0.25]. If exceeded orderly shutdcwn to cold shutdown (in 24 hours) [0.5]. b. Identified leakage is leakage into the equipment sump [0.25]. Unidentified leakage is leakage into the floor drain sump [0.25]. .c. These limits apply to Unit 2 unidentified leakage: At > 1 gym increase from previous 4 hours or ) 3 gpm total perform investigation of cause. (i.e. Drywell air & water samples, plant evolutions) [0.5]. A leak rate > 4 gpm requires a containment entry and inspection. [0.5]

d. This limit applies to Unit 3 unidentified leakage:

If a 2 gpm increase in the unidentified leakage rate occurs within any 24 hour period, immediately initiate an orderly shutdown to cold shutdown. [0.5] (NOTE: The. specific time to reach Hot Shutdown and/or Cold Shutdown is not required for full credit in THIS question) REFERENCE Dresden Tech Specs 3.6.D & 4.6.D. pg 3/4.6-5 and 3/4.6-6 & Operating Order

1. 6-86.

I ANSWER 8.05 (2.00) j

a. Cannot be moved with normal drive pressure [0.5]

Excessive scram time [0,5) f Uncoupled rod [0.5]

b. 8 [0.5]

l l l (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) i l l ~

ADMINISTRATIVE PROCEDURES. CONDITIONS. Paca 29 4 AND LIMITATIONS o-REFERENCE-Dresden' Tech Specs 3.3, pg-3/4.3-1 ANSWER 8.06 (2.00)

a. false [0.5]

b. false [0.5]

c.-false [0.5]

d. true [0.5)

REFERENCE Dresden DAP 3-5, pg 9, 12, & 14. ANSWER 8.07 (1.00) When RPV is > 90 psig [0.5] with irradiated fuel in the vessel [0.5]. REFERENCE Dresden Tech Spec-3.5, pg 3/4.5-10 ANSWER 8.08' (2.00) a. 50 mrem / day b. 300 mrem / week c. 1250 mrem /qtr d. 5000 mrem /yr [ 4 @ 0.5 each ] (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) \\

1 L-j. ' f 4., 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. Para 30 0 AND LIMITATIONS. .. c.

REFERENCE

Dresden, CECO Radiation Protection Standards, P. 33 & 34 i

L l (***** END OF CATEGORY 8 *****) (********** END OF EXAMINATION **********) i i I .-.}}