ML20237K685
| ML20237K685 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 08/20/1987 |
| From: | Bjorgen J, Burdick T, Clark F, Keeton J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20237K678 | List: |
| References | |
| 50-254-OL-87-01, 50-254-OL-87-1, NUDOCS 8708270255 | |
| Download: ML20237K685 (93) | |
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.f i a ., l U.S. NUCLEAff REGULATORY COMMISSION l'
REGION III Report No.- 50-254/0L 87-01' j Docket Nos. 50-254/265 Licenses No. 50-254; 50-265
.l Licensee: Commonwealth Edison Company I' Post Office Box 767 Chicago, IL 60690' i l
Facility Name: -Quad Cities Nuclear Power Station-Examination Administered At: Morris and Cordova, Illinois I 1 Examination Conducted: July 13-23, 1987 0.bs ' w Examiners:
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Date Y$ g-8/}r/77 Ocfte ' l Clark E 7 i Date Approved By: urd Operating Licensing Section
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Date / l Examination Summary Examination administered on July 13-23, 1987 (Report No. 50-254/87-01)) Written, oral and simulator requalification exams were administered to eight Senior Reactor Operator (SRO) and five Reactor Operator (RO) licensed personnel. Results: One Reactor 0porator (RO) and three Senior Reactor Operators (SRO) failed the written examination. All other personnel passed all portions of ' the examination. l l 8708270255 870021 e PDR ADOCK 05000254 V PDR
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1 REPORT DETAILS
- 1. Examiners'
*J; C. Bjorgen,' Region III '
F. Clark, Region III J. Keeton, Region ~III
- Chief examiner ;
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- 2. Exit ~ Meeting ;
An exit meeting was held on July 23, 1987. The following personnel were present at this meeting. Commonwealth Edison Company R. L ' Bax,. Manager, Quad Cities Station M. Kooi,; Regulatory Assurance J. Neal, Training Supervisor , G. Tietz, Assistant' Superintendent of_ Operations ! J. Wethington, Quality Assurance R. Roby, Services Superintendent W. Graham, Training Department USNRC J. C. Biorgen, Chief Examiner. F. Clark, Examiner. Rt Higgins,. Senior Resident Inspector N. Chrissotimos, Deputy Director, Division of Reactor Safety' The following topics were discussed in the exit meeting: !
- a. The examiners identified no generic weaknesses at the exit meeting.
- b. A general discussion was held concerning the examination process and the estimated date that the exam results would be completed.
- c. Two items of concern noted during the examinations were referred to the Resident Inspector for investigation: The. operability of the Reactor Core Isolation Cooling system with the 1300-53 valve out of service and a concern of whether or not an operator was maintained "at the. controls" of. Unit 2.
- 3. Examination! Review Copies of the written examinations and answer keys were given to the facility personnel for review at the conclusion of the examination.
The~ facility comments were provided to the examiners on July 23, 1987. The comments and resolution are enclosed as Attachments 1 and 2 to I the' report. 2 . I
Facility: Quad Cities Examiner: J. Bjorgen Dates'of Evaluation: ' July 13-23, 1987 Areas Evaluated: X' Written X Oral X Simulator Examination' Results: o R0 SR0 Total- Evaluation Pass / Fall Pass / Fail Pass / Fail (S, M, or U)' Written Examination 4/1 5/3' 9/4 M
-Operating Examination Oral- 5/0 8/0 13/0 S Simulator. 5/0 8/0 13/0 S Evaluation of facility written examination grading N/A Overall Program Evaluation Satisfactory Ma rginal . X Unsatisfactory (List major deficiency areas with brief descriptivecomments)
The overall evaluation of marginal is' based on the results of the written examination which.is an-improvement over previous examinations. administered in.0ctober 14-16, 1986 (Report 50-254/0L-86-03). The results of the written examination and operating test indicated no specific major weaknesses. u itted: Fo rded: A Approv /.
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l , ( f I l J ATTACHMENT 1
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I Resolution of Facility comments for the Quad Cities Nuclear Station Reactor Operators Requalification Examination administered July 20, 1987. Comment: 1.05 Part "b" of this answer is looking more at a theoretical change l in reactor power than at a change that would be seen in an l operating power plant. The change in vacuum caused by the l 10 F change in Circ. Water Temperature would cause a much ' smaller temperature change in the actual feedwater temperature at the inlet to the reactor. Therefore, reactor power would i essentially remain the same. An acceptable answer to this j question should be either, decrease or remain the same. I Resolution: Comment acknowledged. The change in reactor power for the small I change in circulating water temperature would not have a major affect on an operating reactor plant. However, the intent of the question is to determine the candidates knowledge of the I theoretical relationship between decreased subcooling of the condensate and reactor power. The answer key was unchanged. Comment: 1.06 An examinee may misinterpret this question since the initial reactor pressure that is referenced is 350 psig. At 350 psig, the RHR injection valves are still interlocked closed. They will not open until reactor pressure is 325 psig. (refer to LIC 1000-1, page 20) With this interpretation, the correct , examinee response would be as follows: ! I
- a. Increase (.25) since no injection was occurring at a 1 reactor pressure > 325 psig. (0.5)
- b. Decrease (.25) due to the fact that the reactor pressure that the RHR pump is discharging to is lower than the pressure created in the RHR system due to flow resistance through the minimum flow valve. (0.5)
- c. Increase (.25) since more power is required to pump this !
increased flow rate of water to the reactor. (0.5) ! Therefore, either answer should be acceptable based on the various ways that the reactor pressure initial setpoint was judged. , I Resolution: Comment partially accepted. j Part a. The answer key changed to accept " Increase (0.5) Initially no injection path is available, as pressure decreases ! and the injection valve opens, flow increases due to centrifugal , pump head / flow characteristics." (0.5) 1 1 ___________U
ATTACHMENT 1 2 Part b. . The answer key changed to accept " Decrease (0.5) The point on the pump operating curve shifts to a lower discharge pressure due to the system pressure being less than the pressure due to the system pressure being less than the pressure the minimum the pump will develop flow valve." (0,5) at low flow rates through Part c. The answer key remains unchanged because the reason for the pump's increased power requirement does not change . Comment: 1.07 a. The answer to this question may vary according to what situation map. the examinee assumed.
- Reference the Power-Flow 40% power at minimum recirc pump speed would flow decrease due to reduced natural circulation. A control rod insertion with a starting condition of 100% power at maximum recirc pump speed would cause a core flow increase as stated in the answer key. Either answer is acceptable.
Resolution: Comment acknowledged. increase. The answer key remains unchanged.All However, the candidates stated examination comments. bank will be changed to incorporate the facilities Comment: 1.08 The coefficients listed in the answer key are correct for each part follows:of this question but the explanation for each are as a. and a reduced resonance escape probability. Adds ne b. and an increased resonance escape probability. Adds c. Adds negative reactivity due to a longer slowing down length and a reduced resonance escape probability. The large concept reactor. of neutron leakage plays a very insignificant role in a LIC THE0 lesson plan on page 16.The correct answers listed are referenced Resolution: Comment accepted. accept the recommended answers. Answer key for part a, c, and d changed to Comment: 2.01 a.1. The High Orywell Pressure setpoint has recently been changed Refer to 2.5 to Tech psigSection Spec and is therefore the correct number. 3.2/4.2-12.
ATTACHMENT 1 2 Part b. The answer key changed to accept " Decrease (0.5) The point on the pump operating curve shifts to a lower discharge pressure due to the system pressure being less than the pressure due to the system pressure being less than the pressure the pump will develop at low flow rates through the minimum flow valve." (0.5) 4 Part c. The answer key remains unchanged because the reason for the pump's increased power requirement does not change. Comment: 1.07 a. The answer to this question may vary according to what i situation the examinee assumed. Reference the Power-Flow map. A control rod insertion with a starting condition of 40% power at minimum recirc pump speed would cause a core flow decrease due to reduced natural circulation. A control rod insertion with a starting condition of 100% power at maximum recirc pump speed would cause a core flow increase as stated in the answer key. Either answer is acceptable. Resolution: Comment acknowledged. All candidates stated that core flow would increase. The answer key remains unchanged. However, the examination bank will be changed to incorporate the facilities comments. Comment: 1.08 The coefficients listed 'i the answer key are correct for each part of this question but the explanation for each are as follows:
- a. Adds negative reactivity due to a longer slowing down length and a reduced resonance escape probability.
- b. Adds positive reactivity due to the shorter slowing down length and an increased resonance escape probability.
- c. Adds negative reactivity due to a longer slowing down length and a reduced resonance escape probability.
The concept of neutron leakage plays a very insignificant role in a large reactor. The correct answers listed are referenced from the LIC THE0 lesson plan on page 16. Resolution: Comment accepted. Answer key for part a, c, and d changed to accept the recommended answers. Comment: 2.01 a.1. The High Drywell Pressure setpoint has recently been changed to 2.5 psig and is therefore the correct number. Refer to Tech Spec Section 3.2/4.2-12.
w <- ATTACHMENT 1 3
- 2. Lo Lo Reactor Water Level - 59" and reactor pressure LT325 psig is a Quad Cities _ LPCI initiation signal.
350 psig is not a Quad Cities-setpoint. Refer to L .LIC 1000-1, Page 34.
' Resolution: Part a.1. Comment accepted. Answer key. changed to also accept the 2.5 psig setpoint.
Part 2. Comment accepted. Answer key changed to the correct pressure setpoint of 325 psig. Comment: 2.03 This question refers to CR0 cooling water flow at a normal value of 75 gpm. In this particular use, I cannot foresee this altering an examinee answer but if this question is used in the NRC Exam bank, a 40 gpm flow rate should be used since this is the actual Quad Cities setpoint. Resolution: Comment acknowledged. The examination bank will be corrected to indicate the correct flow of 40 gpm.
. Comment: 2.05 The reason for sequential loading of.the diesel generator is well stated in the second half of this answer (sequential loading protects the generator form being overloaded due to equipment starting current). That statement alone answers-the question and, therefore, should be accepted for full credit.
If it must be stated that the diesel generator is a small power supply, then that portion should not be weighted so that one half of full credit'is based on that ctatement. Resolution: Comment partially accepted. The reason for sequential loading ' is because the diesel generator is a small power supply. If l the D/G were a large power supply, sequential loading would not be required. Answer key changed to accept "The D/G is a small power supply (0.25) and sequential loading protects the D/G from being overloaded due to equipment starting current (0.75)." l Comment: 2.07 This question is based on the reactor water level decrease and the automatic actions that occur as a result of this decrease. The trip of the RHR pumps is net a direct result of the level decrease and should not be required for full credit on this question. Resolution: Commer.t acknowledged. Answer key ur. changed. Comment: 2.08 Answer No. 4 in the list of possible answers should include the possible, but not required, explanation for decreased rod worth. , This being that the boron is in competition with the control l rods. Refer to LIC 1100, Page 2. ' l l
9 b c 6 ATTACHMENT 1- 4 Resolution: Comment accepted. Answer key changed to also include "(control rod competition with the injected boron.)"- l Comment: 3.01 The instrument referred to in the answer as Yarway Wide Range is a local indication instrument only that utilizes the jet pump taps. This same tap is also used to input to the Wide Range Rosemont which feeds the +57 to -243 indicators on the 90X-3 panel. The Wide Range GeMac (lower 400) also is from this tap. Any 2 of these 3 instruments is a correct answer. Refer to LIC 0263-1, Page RV1-7 for information on the Wide ; Range Rosemont. 1 This question also asks for the jet pump input to the level instruments. Knowledge of which jet pumps are full instrumented is required and it is sufficient to identify that jet pumps 1, 6, 11, and_16 are our fully instrumented jet pumps and not require that specific instrument line configuration be stated. l Therefore, an answer which includes jet pumps 1 and 11 in 4 addition to 6 and 16 should be accepted for full credit. ' Resolution: Part b. Comment partially accepted. The Quad Cities training material specifies what. jet pumps input to the wide range level detectors. The answer key remains unchanged for this portion of the answer. The answer key for the portion that addresses the detectors was changed to include two of the three detectors , at 0.25 points each. The value of the question was increased i to 1.75 points vice 1.5 points. Comment: 3.03 This question asks the examinee to state what actions will occur. with each stated situation. Part "a" will result in an APRM inop. which is a Rod Block as well as a half scram. Since the halp scram is the most severe, it is the correct answer as indicated on the answer key but credit should not be taken away if the examinee also mentioned that a rod block will occur. Refer to LIC 700-4, Page 8. Resolution: Comment acknowledged. Answer key changed to indicate that a rod block will also occur. Comment: 3.05 The answer to part "d" should not include the ATWS initiation since ATWS is not a part of RPS. Refer to LIC 0300-3, Page 2. The answer to part "e" should also include -59 inches and < 325 ; psig as a LPCI and CS initiation setpoint. Refer to LIC 1000-1, j Page 34. Resolution: Part d. Comment accepted. The portion of the answer referenced to the ATWS tip was removed from the answer key. I _.-_----__--______-__--____0
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' ATTACHMENT 1 5
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Part e. Comment accepted.- The answer key was changed to accept i the additional answer. j q Comment: 3.08 At Quad Cities Station,'either a Limited SR0 Fuel Handling Foreman or a SRO, Shift Foreman, will direct all' action connected with the actual refueling operation. Since an R0 does not operate or direct operation of the refueling platform, they have not been held accountable for this material at Quad i Cities Station. This' question should be deleted and the point value redistributed.
' Resolution: Comment partially. accepted. The question was deleted from the examination and the points removed from the section.
Comment: 3.09 Part "d" of this question ir misleading. The Flow Converter input to APRM's No. 1, 2 and 3.is Flow Converter No. 1. The i High' Flux Alarm would be the APRM Rod Block alarm, which is . ! calculated by the formulate (.58 50). Using the gitten Flow Converter No. 1 output of 90%, tb +high flux alarm would come up at 102.2% power. This calculated value is so close to the APRM No. 1 actual reading that is is hard to determine if an alarm would be up from this APRM or any other APRM in its channel. Due to this problem, either a true or a false response could be justified and therefore,. full credit should be given for either response. Resolution: Part d. Comment not addressed due to all candidates answering the question correctly. Comment: 4.04 Part "b" of this question requires that the examinee be so familiar with this particular procedure as to recognize that it is a note in this specific procedure that requires establishing' the drywell to torus dp. Since the inerting process would be done'in accordance with the procedure and due to the fact that this is a note in the procedure as opposed to a procedure step, this part of Question 4.04 should be deleted and its point value redistributed. Resolution: Part b. Comment partially accepted. Part b. of the question is deleted from the exam'ination and the points removed from the examination. Question 4.04 value changed to 0.5 points vice 1.0 points. Comment: 4.05 Part "c": appears to be saying that one minute following a Load Reject condition (scram condition), the reactor power is 93%. This would qualify as an entry into QGA 100 block because ] j of the fact that reactor power is >3% following a scram j initiation. This question uses the term " major load rejection" 1 1 I _ 1
ATTACHMENT 1 6 i which could mislead an examinee and he would answer the l question as if a turbine / generator run back was in progress. I Since a load rejection which is initiated from a turbine / generator runback is not a direct scram condition, the answer to this question could also be that no QGA entry condition was reached. Credit should be given for either answer. j Part "d": A Group I Isolation itself is not an entry condition for the QGA's however, depending on plant conditions, will a probably lead to a condition where the 100 block of the QGA's will be entered. The answer given identifying the 100 block should be considered correct however "NONE" should also be considered correct as a Group I Isolation is not an entry condition. Page 1 of GQA 100-1 shows entry conditions for the 100 block. Part "g": A suppression pool level of 13.7 feet corresponds to
, an indicated level of < -2 inches. This is an entry condition for the QGA procedures. Normal terminology at the station, procedural and indication, is in inches rather than in feet.
This added some room for misunderstanding the question, but the value given would be an entry condition for the QGA procedures. Refer to QGA 200 block entry conditions page. Resolution: Part c: Comment acknowledged. A load reject above approximately 45% power is a scram signal. Answer key unchanged. Part d: Comment partially accepted. Part d. of the question was removed from the examination. Part g: Comment accepted. Answer key changed to "yes" vice "none." Due to the deletion of Part d. from the examination, the value of Question 4.05 was changed from 1.8 points to 1.6 points. Comment 4.07 The answer to part "a" specifies that core submergence is to be verified by two means of level indication. There is no procedure or lesson plant that states a requirement for two means of level indication and therefore it should not be required for full credit on the question. Resolution: Comment accepted. Answer key changed to " verified by level indication" vice the answer as stated.
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' ATTACHMENT 1 7" Comment 4.09a: The Quad Station limit for the first quarter is 1250 mrem as stated'in the answer key. However, this-limit is not an absolute limit and may be raised to 3000. mrem with the proper approval, a Form 4 on file and not to exceed 5(N-18). Either answer.should be accepted. Refer to QRP.1000-1.
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Resolution: Part a. Comment acknowledged. Will accept 3000 mrem if the candidate states the assumptions that a NRC Form 4 is on file and personnel do not exceed the.5(N-18) limit. Answer key changed. Comment 4.10: Since the secondary. containment is the same area for both units,.the Reactor Building, the answer for 4.10 would be correct to state that an isolation of the Reactor Building Ventilation system should be initiated and SBGT system .! operated. -Any method used to affect a Reactor Building Ventilation system isolation will isolate the whole system and not an individual unit. The purpose of this would be to prevent the spread and/or release of radio nuclides, as stated in the answer key.-
. Resolution: Comment partially accepted. In accordance with Quad Cities l- Q0A 800-1 the immediate actions are:
- 1. Use the public address system to evacuate the reactor a building. l
- 2. Verify initiation or initiate standby gas treatment and affected unit ventilation isolation 3
- 3. If the other unit's ventilation has not isolated, trip
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L fans and isolate the other reactor building
.1 The answer key was changed to the following:
- 1. Use the public. address system to evacuate the reactor building (0.25) to minimize personnel exposure (0.25)
- 2. Verify or initiate SBGT system and affected unit ventilation isolation (0.25) to prevent the spread and/or release of radionuclides (0.25)
- 3. Verify or isolate the non affected units reactor building ventilation system (0.25) to prevelt the spread of contamination (0.25) l l _
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E J - i I l' '" 1 S' ATTACHMENT.2. l 1 EXAM COMMENTS AND RESOLUTION
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EXAM DATE 7/20/87 j l i Comment:- 5.01 a. ' This question asks how the available NPSH changes when
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i power is. increased with rods from 65% to 85%. The listed answer is decrease, which may be true if the increase in feed' flow is disregarded. However,'when power.is raised 20%, the: feed flow will increase'accordingly.and the available NPSH will increase (as described in Part c. of this question). Increase should also be an acceptable- 1
- answer for-this question. , Resolution: Comment accepted. The answer key was changed to " increase" .l as the correct answer. j Comment: 15.02 The first part of this answer " Period depends on delayed ] >
neutron fraction (beta).and reactivity" is not clearly asked . for'in the' question and should not be an expected part of the ! answer. The fact that beta changes over life, due to the . buildup and fission of Pu-239 and burnout of U-235, may better ! describe the why, in that it d2 scribes why beta changes ovec life. I Resolution: Comment partially accepted. Partial credit was given for a i description of Pu-239 buildup. But, full credit was not given unless the examinee explains the correct relationship between period and delayed neutrons. The answer key was changed. ' Comment: 5.04 a. FLCPR may also be described as the MCPR LC0 or MCPR limit
' divided by the actual CPR or bundle CPR or any other items I that have the same meaning. See Page 35 of 47 of the HTFF !
-REVIEW Lesson Plan. !
Resolution: Comment acc.epted. Other forms of describing FLCPR were I added to the answer key. , i Comment: 5.05 Reactor criticals may also occur in the IRM range. In the ! last part of this answer, "on the SRM," a reference to the IRM's should also be acceptable. Resolution: Comment accepted. The answer key was changed to include IRM. i r . Comment: 5.10 This question is somewhat confusing in that the choices available to the student are MORE NEGATIVE, LESS NEGATIVE or REMAINS THE SAME. Parts a and d of the question deal with rod ; worth, which is more commonly looked at as a positive coefficient. I . _. . _ - ~ _. - __ _._ - - - _ _ - _ _ . __ ____---_ _ __ --- _____-________-__ - -
-ATTACHMENT 2- 2
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Perhaps if this. question is to remain in the exam bank', the choices could be changed to how the absolute value changes, g INCREASE, DECREASE or REMAINS THE SAME, or some other. wording-to prevent any possible confusion. Resolution: Comment noted. Comment:- 6.01 One of the answers to this question is'to raise water level to above 47 inches. This is what was stated in the lesson plan and the procedure, however, it is currently'being changed to 90 inches. Regardless of the value, to raise water level. without identifying a level, should be sufficient for full credit on this-part of the answer. Resolution: Comment partially accepted.- Raise reactor water level is acceptable for partial credit, but the examinee must indicate that' level is being raised "above 47"" for full credit, - i.e.,. 90" is acceptable. Answer key was changed. Comment: 6.03.The setpoint of 1.3 psid should not be required. A dp is established to minimize.the~ water level in the downcomer pipes. If'a .value is required, the Technical Specification limit of 1.2 should be acceptable as well. See Page 3.7/4.7-6a of Unit 2 Technical Specifications. Resolution: Comment accepted.' Stating that drywell pressure is greater c' than torus (or suppression chamber) was acceptable for full credit. Answer key was changed.
' Comment: 6.05 a. 2 x. 10 8lbm/hr is commonly used as this setpoint and is considered equal to 20% feed flow. Either value
- should be acceptable. See 38 of 47 of LIC-0600 (VESSEL LEVEL CONTROL SYSTEM).
ResolutLn: Comment accepted. Either 20% or 2 x 106 is acceptable for full credit. Answer key-was revised. Comment: 6.05 b. .The setpoints identified in this answer are not asked for and should not be required. Full credit should be given for identifying the trips as high reactor pressure and low' low reactor water level, at , . Resolution: Comment not accepted. The examinees were given a correction l to the question during the exam to' indicate setpoints are required. The answer key was unchanged. L Comment: 6.06 Two units provide inputs to the recirc flow control while in EGC. These units are'the pressure control unit and the load control unit. The signals from these units are summed to develop
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ATTACHMENT 2 3 the signal to be sent to the recirc pump controls. Part c of this question identifies only the load control unit (1). The pressure control unit (2) should also be considered a correct answer since it will control recirc pump speeds if pressure changes and the EGC demand signal from the load control unit remains the same. See Figure 6 of LIC 5650-2 (EHC PRESSURE CONTROL AND LOGIC). Resolution: Comment not accepted. The referenced training material explicitly states the subsystem in control as identified by the answer key. Answer key unchanged. 6.08 c. The answer to this question is wrong. The question asks Comment: on what IRM range the IRM INOP trip is bypassed. The correct answer would be that it is not bypassed on any range. It seems that the examiner may have confused this l with the SRM INOP trip which is bypassed on range 8 of the IRM's. In any case, this questions does not have a correct answer and should be deleted and the points redistributed because the wording of the question leads the student to believe that a correct answer does exist. Resolution: Comment noted. Due to the confusing wording of the question, Part c was deleted. The point value was redistributed. Comment: 6.09 The upper limit of the Rod Block Monitor is .65 Wn + 42 per Tech Specs. See Page 3.2/4.2-14 of the Unit 2 Tech Specs Resolution: Comment accepted. The answer key was changed to accept either 41% as stated in the referenced training material or 42% as i stated in the Technical Specifications. 1 Comment: 7.01 This question asks about limitations dealing with continuous - I rod withdrawal and references QOP 280-1. While this is a valid reference for this question, requirements dealing with continuous rod withdrawal exist in other procedures as well. Responses dealing with requirements from other procedures should be considered acceptable because it is not so important as to what i procedures list various requirements as it is that an operator is aware of the requirements. Refer to Page 4 of QGP 1-1 which . shows additional requirements dealing with continuous rod { withdrawal. Also, in QOP 280-1, four conditions are listed ! that allow continuous rod withdrawal. In the answer key two i of these, timing CRD's and CRD friction testing, are listed i as one. These are two separate evolutions and should be j identified as two possible choices. I l
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ATTACHMENT 2 4 l l Resolution: Comment accepted. The question specifically identifled QOP E80-1. However, answers appearing in other procedures are acceptable. The answer key was changed to accept valid controls listed in QGP 1-1, Page 4 and QGP 4-1. Comment: 7.02 b. This question asks the " primary concern'.' for the Loss of RBCCW. Nowhere in the referenced procedure, or any other procedure, is the term " primary concern'.' used. While the answer of drywell pressure increase due to loss of cooling is obviously a major concern, it is not even mentiored until the discussion section of tha procedure. Explanations of other major concerns such as recirculation pumps or the cleanup system, which are addressed in the IMMEDIATE OPERATOR ACTIONS or SUBSEQUENT OPERATOR ACTIONS, should also be considered when grading this question. Resolution: Comment partially accepted. Answer key changed to accept either of the following: 1
- a. Loss of cooling to drywell (increased temperature and pressure).
- b. Loss of major component cooling that requires a shutdown (recirc pumps, RWCU pumps).
Alternate wording accepted. ! Comment: 7.03 a. The title of the individual listed in this answer, Administrative and Support Services Assistant Superintendent has recently been changed to Assistant Superintendent of Technical Services. This new title should be acceptable for this answer. j Resolution: Comment accepted. The answer key was changed to accept either title. Comment: 7.07 The values listed in this procedure, with the exception of the l cooldown limit of 100F/hr, are not values that are critical for Technical Specification conformance. Considering this, some tolerance should be allowed when evaluating student responses to this question. Resolution: Comment partially accepted. The answer key was changed to allow a tolerance on values for Part a only. Comment: 7.08 When identifying which QGA procedures should be entered, the exact procedure number listings should not be required. Instead, referencing the proper block of procedures, such as
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I I ATTACHMENT 2- 5' the 100 block or 200 block, should be acceptable as'that is how the procedures are utilized and how they are referred to at QCNPS. Resolution:- Comment accepteo. The answer key was changed to indicate tuse of " Block"'as acceptable. Ccmment: 7.08 d. A Grou'p I isolation itself is not an entry condition for: QGA's however, depending on plant conditions..will probably lead to a condition where'the 100 block of QGA's will be-
' entered.' The answer given, identifying the 100 block, should be considered correct, however, NONE should also be considered correct as the Group I is not an entry condition. Any other blocks of procedures listed as an answer should be considered wrong. Page 1 of QGA 100-1 is attached to show the entry conditions for the 100 block.
- Resolution: Comment partially accepted. Part d was deleted.
Comment: 7.08 g. 13.7 feet in the suppression pool is an entry condition to the 200 block of'the QGA's. This is not the normal method of monitoring suppression pool water level. The level instrument on the front of the 90X-3 panel, indicating water level in inches from a reference zero, is the normal method of monitoring suppression pool level. Page 1 of QGA 200-1 shows the entry conditions for this block of procedures. Resolution: Comment accepted. The answer key was changed to accept only "200 block." Comment: 8.02 a. One other possible answer that may appear on this question is " exceeds 16 hours in~a 24 hour period." -This may be seen in QAP 300-3 Step C.2.b. If this' answer is provided by examinees, it should not be considered wrong. Also, ) this response should not be a required answer if the limit " of "less than 8 hours between work period," as is already required b.' the answer key, was identified because the only time that 16 hours in 24 hours was exceeded was during the j less than eight hour rest period. ) Resolution: Comment accepted. The answer key was changed to accept either answer for full credit. Comment: 8.02 b. The answer to this question is in a manner, yes. The Shift Engineer does deviate from these guidelines on occasion and is allowed to do so as described.in QAP 300-3, Page 2, the paragraph below e. Since this does occur on l 1
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1 ATTACHMENT 2 6
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l occasion at the plant, and is allowed per the QAP's, a YES I' answer should be considered ccrrect. A NO answer, as shown in the answer key, may be considered correct since the Shift Engineer is not authorizing an exception, he is documenting and allowing a deviation. Resolution: Comment partially accepted. Part b deleted. Comment: 8.03 At Quad Cities, a Communication Center exists that is staffed by operating experienced personnel and functions mainly to support the Shift Engineer and Shift Foreman. The logs for these keys and' locks are maintained by the Communication Center personnel. Since some examinees may respond "the Communication Center" to this question, some consideration should be given to the actual operating department operation when evaluating these responses. Resolution: Comment not accepted. The questions asks for "Who is ! responsible." The only correct answer is Shift Engineer. Answer key unchanged. l Comment: 8.05 A normal shift at QCNPS is an eight-hour period. Responses of "8 hours" should be acceptable in place of "one shift" for this answer. { Resolution: Comment accepted. The answer key was changed to accept either eight hours or one shift. Comment: 8.08 This QEP lists six duties of the Shift Engineer, only four of which are identified in the answer key. The other two duties should also be acceptable as well as duties identified in the ! LIMITATIONS AND ACTIONS section of this procedure that would be performed by the Shift Engineer. These actions are all i identified in QEP 340-5. Acceptance of these answers also brings this answer key closer in line to the answer required by the NRC on previous exams dealing with this procedure. ] l Resolution: Comment partially accepted. Unlike previous NRC questions, the ! question does not ask to list the duties. It places the examinee in the situation of being Shift Engineer with a secondary containment fire in progress. Therefore, two of the duties, i.e., conducting fire drills and assisting the Fire Marshall with his report, are not acceptable. The answer key was changed , to include " arranging access for outside fire fighting equipment," l as an additional acceptable answers. 1 Comment: 8.09 This question is difficult to answer because it is a little vague in the information given. If the intent of the question is to identify the examinee's knowledge of the 1.25 times, the l 1
}
. . . . 4 ATTACHMENT 2 7 surveillance period and 3.25 times the surveillance period for three consecutive surveillance criteria, then the date of the surveillance completed before the July 1 due surveillance should have been given. If the examinee assumed it was exactly three morths before July 1 and assumed the system was operable the whole time, the surveillance inay have been legitimately completed from October 1 to October 23.
The wording that the surveillance was not completed because of an LC0 on the system could lead the examinee to the conclusion that the system was inoperable and had to be repaired, then was repaired and tested on July 15. In this case, the examinee would be justified in starting the surveillance period all over at July 15, similar to restarting a HPCI surveillance after an extended outage. In this case, the surveillance could have been legitimately completed from October 15 to November 7. Resolution: Comment not accepted. Supporting documentation for the comment was not provided. Any answer similar to the second paragraph of the comment could result in violation of the 3.25 for three consecutive surveillance periods. The answer key was not changed.
r : t, - .
.a. ,
U. S. NUCLEAR REGULATORY COMMISSION j SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION g/gr./ c!?M FI,C I LI TY : _QUAp_CITIE@_1Lg_________ { y/c / /D REACTOR TYPE: _@hlR-GE3_________________ DATE ADMINISTERED: _gZfgZfgg________________ EXAMINER: _KEETgN 3 _y3______________ i CANDIDATE: _________________________ i l JNgIRygIJpNg_IQ_gSNp1pSIEi i Raad the attached instruction page carefully. requalification This examination replaces l current cycle facility. admi ni st ered examination. the R2 training requirements for failure of this examination are the same as for failure of a requalification examination prepared and administered by in your' training staff. Points for each question are indicated parentheses after the question. The passing grade requires at least 70% grade of at least 80%. Ex ami nat i on papers. in each category and a final will be picked up four (4) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE
-------- CATEGORY l
________ 5. THEORY OF NUCLEAR POWER PLANT 15:35-- 35E2T ---------__ OPERATION,. FLUIDS, AND THERMODYNAMICS ; PLANT SYSTEMS DESIGN, CONT R.0L , 19:99-- ?dr22' -----_-.___ ________ 6. AND INSTRUMENTATION I4.MI PROCEDURES - NORMAL, ABNORMAL, ker@9-- 2@I22 ---------__ ________ 7. EMERGENCY AND RADIOLOGICAL CONTROL
/S. S ADMINISTRATIVE PROCEDURES, h@ETE-- Ffi9)' ---________ ________ B. CONDITIONS, AND LIMITATIONS f 6.0 -------_% Totals kerD9_- _------_---
Final Grade All work done on this examination is my own. I have neither given nor received aid. Candidate's Signature I i
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l _ < . , ; . _ . _ . m v . . i 2
v ,.--
, ;;a .; .,. ., , . , , _, ,, __
t . 4 l' NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS
/'
During the administration of this examination the f ollowing 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 gnly to f acilitate legible reproductions.
- 4. Print your name in the blank provided on the cover sheet of the ex ami nat i on.
- 5. Fill in the date on the cover sheet of the examination (if necessary). l
- 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. B. 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. Number each answer as to category and number, for example, 1.4, 6.3. 9.
- 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 f or each question is indicated in parentheses after the ;
question and can be used as a guide f or the depth of answer required. l
- 14. Show all cal c ul ati on s, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not. j
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DD 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 cone after the examination has been completed.
, e
_ . _ _ . _ ___ _ _ _ . _ _ _ _ . _ ___________________a
l l
- 10. When you- complete your examination, you shall
- a. . Assembl e your examination as f ollows:
Exam questions on top. .l
-(1).
(2) Exam aids - figures, tables, etc. 1 I (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 q i
the examination questions.
- c. Turn in all scrap paper and the balance of.the paper that you did j not use-for answering the questions.
- d. Leave the examination area, as defined by the examiner. If after leaving, you are f ound in this area while the examination is still in progress, your license may be denied or revoked.
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'PAGE 2 gi -THEORY OF NUCLEAR POWER PLANT-OPERATIONg, FLU 1ggg,ANQ IHggdQQyN951gg;
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i
'GUESTION- 5.01 (1.50)
STATE whether each'of.the following changes will (INCREASE, DECREASE, or NOT AFFECT) AVAILABLE recirculation pump Net Positive _ Suction Head
- (NPSH):
- e. Reactor power level is increased f rom 65% to 85% by withdrawing control rods.
- b. ' Reactor water l evel is stable and a high pressure f eedwater heater i sol ates,
- c. Recirculation pump speed at minimum and f eedwater flow increases.
QUESTION 5.02 (1.00) ! Dur!99 a given fuel cycle, the period will be longer at BOC.than at EOC I
-for the SAME AMOUNT of reactivity insertion.
" EXPLAIN WHY and HOW this occurs.
.OUESTION 5.03 (2.00)
Answer each of the following TRUE or FALSE:
- a. During equilibrium power conditions, the production rate of indirect Xenon from Iodine is faster than the decay rate of Xenon to Cesium.
- b. Slowing the rate of a power decrease, lowers the height of the resultant Samarium peak,
- c. The resultant Xenon peak due to a scram from 50% power is larger than one from 100% power. l l
- d. During an increase in power from equilibrium Xenon conditions, Xenon concentration initiall y decreases.
l (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) l
U
. . . . . , .o. . .: . . . . _
Ss . ' THEORY OF' NUCLEAR POWER PLANT OPERATIONg,FLt_JID@g,AND PAGE 3
-ISEBdQDXN951QS
~ QUESTION 5.04 (1.25)
DEFINE the' f ollowing ' terms and STATE HOW each term is used .during operation. c.. FLCPR (0. 5 ) .
- b. MFLPD (0.75)
I 1
. i DUESTION 5.05 (1.00)
(remec(j.fJhaaevuldws During a reactor startupV the reactor has just gone critical; STATE TWO, indications of criticality and on WHICH nuclear instruments the indications are observed. l l I T (***** CATEGORY 05 CONTINUED ON NEXT PAGE * * ** * )
c _ m. . . . . . ._ m._. _ . ... . PAGE 4
'5 ...
2 ISEQS1_QE_NUC(E66_EQWE6_P($NI_QEE6@llgNt _E(UlDS t_QND
..- IdEBdQDYN901Qg 1
4 QUESTION 5.06 (1.50) l
~
M11 A I 8 Core Reactivity, Keff e.co 10 [ O CORE EXPOSURE (mwd /T x 1000) Using the curve provided above, STATE WHY:
- a. Core reactivity DECREASES from point A to point B.
- b. Core reactivity INCREASES from point B to point C.
- c. Core reactivity DECREASES from point C to point D.
QUESTION 5.07 (2.00) You are on shift, reactor power is 95%, control rod manipulations are in progress. The Reactor Operator reports to you that LPRM readings on the four rod display can't be right, because he has just withdrawn a rod from notch 36 to 40 and power indication ; decreased. A check'of the LPRMs indicates no apparent malfunction. WHAT was the cause of the power decrease and EXPLAIN WHY it occurred. 9
> >=s-3
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
r 5 3,,, THEORY'OF NUCLEAR POWER' PLANT OPERATIO _% UIpSg_ D GE 5 I IdEBDQQyNSD1CS QUESTION 5.08 (1.00) l You are'the Shift Engineer and have just been forced to evacuate your l crew from the control room. You must now shutdown and cooldown the 'l i plant f rom outside the control room. Using the f ollowing pressures (available locally) and times, CALCULATE the reactor coolant cooldown rate using the attached steam tables: TIME initial: Reactor Pressure = 400 psig TIME initial + 1 hr: Reactor Pressure = 122 psig
'"*' *
- A" (kmd)5.09Skan labb0c" W (2.50)
QUESTION Does'a power change from 40% power to 30% power take LONGER, SHORTER, or the SAME, as a power change from 40% power to 50% power? ( Assume onl y rod motion and constant rod speed.) EXPLAIN your answer for BOTH conditions. INCLUDE WHICH coefficients of reactivity cause power- to STABILIZE and HOW f or BOTH situations. QUESTION 5.10 (1.50) STATE whether each of the f oll owi ng becomes MORE-NEGAT-IVE, LESS NEGATIVE, or REMAINS THE SAME:
- a. ROD WORTH (delta k/k/ rod) as fuel temperature INCREASES.
- b. ALPHA DOPPLER (delta k/k/ degree F change in fuel temperature) as voids DECREASE.
- c. ALPHA VOI,DS (delta k/k/% voids) as core age INCREASES.
- d. ROD WORTH (del ta k /k/ rod) as voids DECREASE.
- e. ALPHA DOPPLER (delta k/k/ degree F change in fuel temperature) as core age INCREASES.
. 1
(***** END OF CATEGORY 05 *****)
g , PAGE- 6
-su._P69dI_SYSIEDg_pEgiggi_ggyl896t, gyp _IygIRUDEyl@I199 QUESTION 6.01 (2.50)
.Following a scram near EOC on Unit 2 caused by electrical problems, chutdown: cooling with RHR has been established. The recirc pumps trip on loss of-power and cannot be restarted. There are SEVEN methods of minimizing temperature stratification. LIST FIVE. l l
i QUESTION 6.02 (1.50) With regard to the OFF-GAS Systems
- a. WHY does the Pre-heater use steam instead of electric heaters?
- b. WHAT would happen initially if the Off-Gas Condenser level control valve f ailed open and th,eAcondenser completely drained? l OH gar
- c. WHAT is the purpose of the Charcoal beds?
TUESTION 6.03 (1.00) STATE the purpose of the Pump Back AIR System and WHAT it is designed to prevent. QUESTION 6.04 (1.25) LIST THREE conditions and setpoints that will cause the Reactor Water Cleanup (RWCU) System inboard and outboard i sol ati on valves (MOV-1201-2 and MOV-1201-5) to close. QUESTION 6.05 (1.50) With respect to the Recirculation Flow Control System: , I'
- a. WHAT TWO conditions must be satisfied to increase recirc pump speed above minimum?
m ,g zy;~ p d pc.nt(~1
- b. WHAT'causqf)recircpumptripduringATWSconditions?
- c. The MASTER LIMITER lower limit is normally set at (1) ________%
and must be reset to (2) _______% when using the EGC. i s, , , (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
y - . _.. - _ _ ___ ______ L 6 -_P($N1_SYSIEUS_QESlGN g_CQN16Q(t,gNQ,1NS16UDENI@llgN PAGE 7~
- . v QUESTION 6.06 (2.00)
There are FIVE subsystems in the Electrohydraulic Control System (EHC). In each.of the f oll owing si tuations, IDENTIFY the subsystem that is in control,
- c. Valves are repositioning due to. increased steam flow demand; system is providing nulling signal .
.b. Steam flow is greater than control valves can handle.
- c. Providing a signal to the Recire flow control during EGC operation.
- d. Maintains constant relationship between steam flow and reactor pressure during a power increase.
Subsystems:' 1. Load control unit
- 2. Pressure control unit
- 3. Val ve flow control unit
- 4. Bypass control unit
- 5. Speed and acceleration control unit QUESTION 6.07 (1.00)
With the~1/2 Diesel, Generator Ke_yl oc k 'swi tch in the f ollowing positions, describe what happens(to th_e_p/G_ output if an ECCS initiation occurs in e og;fys E w/,af (J. II & @ both uni ts si mul taneousl y. (f,e,,h'*f c'L' do if tryf ac ,t 0 4 a f b r 4 ,* g jf
- a. Unit 1 - OFF and Unit 2 - OFF s
- b. Unit 1 - ON and Unit 2 - DN
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
PAGE O 6 .._P(901,@ySIEd!_pESigyz_CQUIB96t_SUD_JNSIBUdEUISIIDS 4 QUESTION 6.08 (1.25) A reactor startup is in progress ,your NSO is withdrawing rods, and IRMs are.on Range 5. IRM "11" is bypassed so the instrument. technicians can troubleshoot the power supply. The technician mistakenly takes the IRM "14" operate s' witch to standby and starts troubleshooting its power eupply,
- c. WHAT specific plant / system TRIP (s) did this cause? (0.75)
- b. HOW many channels are required for reactor startup? 40c251 .{)
- r. On WAL-4RM-eeftce-woth-the--above-.tr ip ha.vp hopn m '!t@m aticeW
.bypasssed? ' O 2M - gg /d TO QUESTION 6.09 (1.50)
Concerning the Rod Block Monitor (RBM) system. FILL IN THE BLANKS with an appropriate word, phrase, or number.
- a. The .THREE upscale rod blocks f or two loop recirc operation are 0.65 W5 + (1) __________.
0.65 Wd + (2) __________. 0.65 Wd + (3) __________.
- b. The TWO RBM auto-bypasses are (1) __________ and (2) __________.
- c. For RBM channel 7, the alternate reference for APRM Ch. 3 is APRM Ch. .__________.
QUESTION 6.10 (1.50) The reactor is operating at 80% power. The Reactor Level Select switch on the 90X-5 panel is selected to Level A . "A" l evel instrument fails: Answer each of the following TRUE or FALSE. (Assume no operator action)
- a. IF the failure is UPSCALE, the demand level from the controller is 60 inches vs. 30 inches.
- b. If the failure is DOWNSCALE, RFP runout will be prevented by the runout relay energizing at 5.6 x 10E6 lbm/hr per pump.
- c. An UPSCALE f ailure may be caused by steam flashing in the condenser chamber. , ,
l l u . .. . . (***** END OF CATEGORY 06 *****)
PAGE 9 j Za ,PRgCEDURES - NORMAL _ 1ABNORMAL _gEMERGENCY _AND
- 869196991996_QQNIBQL j
j l j QUESTIDN 7.01 (2.00) ! According to DDP 280-1, " Reactor Manual Control," a caution in Step 4.a. states that continuous rod withdrawal should only be used under certain ! conditions.
- o. STATE THREE conditions when use IS permitted. (1.5)
- b. STATE ONE condition when use is specifically PRDHIBITED. (0.5)
QUESTION 7.02 (2.50) During power operation, the NSO informs you that the running RBCCW pump l has tripped and the standby pump cannot be started. According to l ODA 3700-2, "RBCCW Pump Trip":
- a. STATE THREE immediate actions. (1.5)
(1.0)
- b. EXPLAIN the PRIMARY concern if RBCCW is lost.
(Does not involve individual components.) QUESTION 7.03 (1.50)
.In each of the following cases, IDENTIFY the minimum approval level required (Position Title) to receive the whole body dose: (Assume no prior ex posur e. ) ((f er yn r* I sfm I rwy dre-d si.< h hw e." ) i
- a. 2000 mrem in a quarter.
l b. 75 mrem in one week,
- c. 400 mrem in one week.
QUESTION 7.04 (1.00) According to DDP 1600-2, " Suppression Chamber Pressure Relief Through the l SBGTS," the SBGTS may be used to relieve suppression chamber pressures ' that result from normal plant operation. 1 WHY must drywell pressure be greater than or equal to suppression chamber pressure? _ l , (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) L________________.________________________ __ _
' PROCEDURES - NORMALg_ABNQRMAL _tEMERGENCY _AND PAGE 10 .7. . 689196991996_QQNIBQ6 QUESTION 7.05 (1.50)
While monitoring the control room panel s, the NSO discovers the generator is carrying less load. The unit is stable and no alarms have come in. While verifying other instruments, indicated core flow is f ound to be l higher than normal for that power level.
- a. WHAT f ailure has occurred?
- b. EXPLAIN what caused the change in indication of:
- 1) Core flow
- 2) Generator output 7.06 (2.00) !
DUESTION While moving a spent fuel bundle from the core, the grapple f ails releasing the f uel bundle. In accordance with DOA 800-1, " Irradiated Damage While Refueling," STATE TWO required immediate actions and EXPLAIN the reason for each action. l QUESTION 7.07 (1.50) l QOA 6900-7, " Loss of AC Power to the 125 Vdc Battery Chargers with ! Simultaneous Loss of Auxiliary Electric Power," states that if AC feed cannot be restored to the Unit 1 125 Vdc battery, then load shedding must be initiated immediately. , i
- a. Load shedding must be accomplished within (1) _____ minutes to reduce the battery load to (2) _____ amperes. ;
I l ! b. An accelerated cooldown must commence at a rate of greater than f l (1) _____ F/hr but less than (2) _____ F/hr. 1
- c. The procedure addresses loss of of f-si te power in conjunction with loss of AC to both unit battery chargers, but does not address these events plus LOCA. WHY NOT?
i m; . - o _. . l 1 (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) < E-
PAGE 11 Z ..P69QEQUBEg_ ,NQ6566g,$gGQ6d6(t,{d[6Q[GQy_QNQ Be91969GIGe6,QQN16Q( 1.T f
-QUESTION 7.08 (.GYSO7 For each of the f ollowing conditions, determine whether or not emergency procedure entry is required. If . entry is required, STATE WHICH If NO entry is required, state NONE. (Consider procedure (s) to enter.
cach item separately and assume no additional conditions)
- o. RPV level is 10 inches l
- b. Reactor power is 12% STARTUP MODE
- c. Reactor power is 93% one minute after load reject I
;d_ ower---operationg- C"OUP I -i sal-ati = ne< "r = --
h /g /'M
- o. Suppression pool level is -3 inches
- f. Drywell pressure is 2.5 psig
- g. Suppression pool level is 13.7 feet
- 1
*h. Reactor. building vent exhaust 3.2 mR/hr ,
- i. Reactor shutdown, reactor pressure 1090 psig l
J. Drywell temperature is 160 degrees F L - l (***** END OF CATEGORY 07 *****) I t- _ -
PAGE 12 Ez __9901NISI6811VE_ESQQgpyBESg_QQNQlligNSg_gNQ_61DJI$11QNp QUESTION 8.01 (1.50) 1 According.to QAP 300-2, " Conduct of Shift Operations," equipment may be placed in PULL-TO-LOCK for testing ONLY if THREE conditions are met. STATE THOSE THREE CONDITIONS. QUESTION 8.02 1.L-MD f..f You are called in f rom vacation because of a union walkout. You are scheluled as Shift Engineer as follows: Sunday 0800-1600 Monday 0800-2000 Tuesday 0800-1600 Wednesday 0000-1600 Thursday 0800-2200 Friday 0400-1200 Saturday 0800-1600
- a. STATE any exceptions to the guidelines of QAP 300-3, " Shift Manning," that would normall y apply. (1.50)
I
-h -+ m you as Shift t.ngi neTeMIltt1Drtzewxceptions t m +h=
qui. tis 44*es'7----( O.-257 g ggg l QUESTION B.03 (1.00) According to QAP 300-24, " Fire Suppressi on Val ve Locks," STATE WHO is responsible f or maintaining the logs f or keys and locks AND HOW OFTEN they are required to be inventoried.
. v... . V U-
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( (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) L_.
EL. 6EUld1E166112E.P@QQ[QU@@@g_QQUQlll@G@t_@@Q,Llyll@llgNS PAGE 13 H QUESTION B.04' (2.00)
.The Technical Specifications SAFETY LIMIT on Power Transients implies that an.APRM SCRAM on a power excursion will terminate the transient before a significant increase in surface heat flux occurs,
- c. DESCRIBE the inherent characteristic of the fuel that'makes this possi bl e.
- b. HOW LONG can an-APRM SETPOINT be exceeded WITHOUT SCRAM before vi ol ati on of a Safety Limit MUST be assumed:
- 1) WITH the process computer in operation?
- 2) WITHOUT the process computer in operation?
QUESTION B.05 (1.00) In accordance with GAP 300-14, " Equipment Out-Of-Service," " Temporary li f ts shoul d be no l onger than ____ (1) ______ normall y, with a maximum ti me l i mi t of . _____ (2) ______. QUESTION 8.06 (2.00) ! MATCH the f ollowing TAGS (CARDS) with the appropriate DESCRIPTION number :
- 1. Red with black letters
- a. Personnel Protection
- 2. White with green letters
- b. Caution
- 3. Yellow with black letters r
- c. Hold
- 4. White with blue letters l
- d. Special Order
- 5. Red and white
- 6. Orange with black letters (Letter / number sequence sufficient for answer, i.e., e-7, f-9, etc.)
l
-- w i_ m .
1 (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) t __ . - - _ - -a
E: 14
- PAGE sc__eQUIN1516SIlyE_P6QCEQUBEgg_CQNQlliQNS t _@NQ_(ldlI@Ilgdg QUESTION B.07 (2.00) l j
l , STATE WHICH Emergency Classification is appropriate for each of the j f ollowing definitions: J
.o. Events are in progress or have occurred which involve actual or potential ' substantial - degradation of the level of safety of the plant.
b.. Events are in progress or have occurred which indicate a 3 potential degradation of the level of safety of the plant. l
- c. Events are in progress or have occurred which involve actual.or
' i mmi n en t substantial ~ core degradation or melting with the i, potential for loss of containment integrity. I
- d. . Events are in progress or have occurred which involve an actual or likely major f ailure 'of plant functions needed for protection of the public.
l QUESTION 'B.08 (2.00) You are Shift Engineer and a' fire is reported in Secondary STATE Containment; FOUR actions you according to OEP 340-5, " Station Fire Fighting," must take. QUESTION B.09 (1.00) i j A Technical Specification Quarterly Surveillance f or an'ECCS system was !
- due July 1st. Because of an LCD on'the system, the surveillance was not completed until July 15th,
- a. WHEN will the next QUARTERLY surveillance become due? !
- b. DEFINE'" Grace Period." .
I i 4 (1.50) QUESTION B.10 { According to Technical Specifications, there are THREE ref ueling interlocks'that result in ROD BLOCK. LIST ALL THREE. s . - u . . .. - (***** END OF CATEGORY OB *****) (************* END OF EXAMINATION ***************)
~ . - - - _ _ - EQUATION $HEET 4 Cycle efficiem y o (Network f y o s/t foma- o n)/(Em q in) ,f i e = ag s = Vot + 5 att ) E = mc2 A = AN A = Ace - At , KE = 5 av2 a = (Vf - Vo)/t ; PE = agh I w = e/t x = an2/tg = 0.693/tg Vf = Vo + at tgeff = [(tg) (t t)3 W- 6 [(tg) + (t t)3 g , go,-Ix l AE = 931 Am k=iiCpat I = Ioe *
- 6=UAat I = Io 10'*/ M Pwr = Wrah TVL = 1.3/p HVL = -0.693/9 P = Po10su@ .
P = Po et/T SCR = S/(1 - Keff)
,SUR = 26.06/T CRx = S/(1 - Keffx)
CR (1 - Kef f g ) = CR 2 (1 - keff2) SUR = 26p/s* + ( s- p )T 1 , M = 1/(1. - Keff) = CR /CRo 1 T = (t*/p) + [(s - p)/ip)
' M = (1 - Keffo)/(1 - B,ffi)
T = 1/(p - 6) T = (6 - p)/(AP) SOM = (1 - Keff)/Keff s* = 10-5 seconds P = (Keff-1)/Keff = AKeff/Keff I = 0.1 seconds-1 P = [(1 /(T Keff)3 + [Seff/(1 + IT)]
'Ijdt=1d22 2
11d1 2=1d22 P = (reV)/(3 x 1010) 2 R/hr = (0.5 CE)/d (meters) I = .N R/hr = 6 CE/d2 (feet) Miscellaneous Conversions Water Parameters 1 curie = 3.7 x 10 10dps 1 gal. = 8.345 lbm. 1 kg = 2.21 lbm 1 gal. = 3.78 liters I hp = 2.54 x 103 Btu /hr 1 ft3 = 7.48 galt - 1 sw = 3.41 x 106 Btu /hr Density = 62.4 lbm/ft 1 in = 2.54 cm Density = 1 gn/cm3 Heat of vaporization = 970 Btu /1bm 'F = 9/5'C + 32
'C = 5/9 (*F-32)
Heat of fusion = 144 Btu /1bm 1 BTU = 778 ft-1bf 1 Ata = 14.7 psi = 29.9 in. Hg. ) 1 ft H 2O = 0.4331bf/in2
\ ._ .
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. ETHERMQQyN9t)1QE j w
' ANSWERS -- DUAD: CITIES 1&2 -87/07/20-KEETON, J. !
' ANSWER '5.01 '( 1. 50)
- a. DCCCC ^,3C INCR FA SE~
- b. INCREASE
- c. INCREASE (0.5' pts each)
REFERENCE Quad Cities: LIC-0202-1, LO 11 202OO1K101 202OO1K103 202OO1K105' 202OO1K122 ...(KA'S)
' ANSWER- 5.02 (1.00)
Period depends ~on-delayed neutron fraction (beta) and reactivity (0.5). Bet'a i s 11 arger at BDC than at EDC (0.25), therefore, period would be t l(/ls/ er at BDC than ong/smy?'estC cw5at EDC for the same reacti vity inserti on. gg) ,%,.;ffg," 3~
~ *
/'e' *%@e.g.aw
##I#"' adA'*4 mea dotma
. REFERENCE THEORY M' U h /LIC-THED, TPD 7 and B Quad Cities:
292OO3K106 ...(KA'S) ANSWER- 5.03 (2.00)
- c. True
- b. False
- c. False
- c. True
( O.5 pts each) REFERENCE Quad Cities: THEORY LIC-THEO, TPD 11
.292OO6K105 292OO6K106 292OO6K107 ...(KA'S) 4
\ a
. ~. . _. .
PAGE16 LEz__IdEQSI_QE_NyC(E@B_PQWE5_P(@NI_QEE6811QN t _((QlDSg_6NQ ISE60991N951Qg
-87/07/20-KEETON, J.
i ANSWERS -- QUAD CITIES 1&2 ANSWER 5 ..( 1. 25 ) m cnz a /.04'
- e. CF.7 /i;CPO (0.25) Quick method of, verifying MCPR is'within A
l i mi t s. (0.25) (QcLef Cf .n Cama WA]MC M L Cc> o e mc/A cmD
- b. Mai local LHGR/LHGR Limit (0.25) Quick method of verifying LHGL as within limits (0.25) and used with FPR to adjust APRM gai n. _ ,(0. 25)
(p;gisj REFERENCE Quad' Cities: THEORY LIC-THED page 77 293OO9K109 293OO9K128 ...(KA'S) ANSWER 5.05 (1.00) Wi th. no f urther ' rod moti on, period remains positive (0.25) on the source range period meter (0.25). Neutron level continues "to increase (0.25) on the SRMV(0.25) . REFERENCE Quad Cities: THEORY LIC-THED, page 76 and QGP 1-1, page 5 2900BK107 292OOBK106 ...(KA'S)
. ANSWER 5.06 (1.50)
(Will
.a. Decreases due to Samarium (C.25) and Xenon (0.25) 5uilding .
accept fission product poitons f or CO.253.)
- b. Increases oue to Gadolinia burnout (0.25) and Pu-239 building (0.25).
- c. Decreases due to fuel depletion. (0.5)
REFERENCE Quad Cities: THEDRY LIC-THEO, page 74 and Fig. 62 292OO7K101 292OO7K103 ...(KA*S) 1
. I
- - -- . s, ,,
"6- %
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sc._IdEQBy_QE_NQC(EQB_ggWEB_g(@N1_QEEQ@llgNt_E(Q[DSg_@NQ PAGE 17
. 1dE60QQYN@dlQS S
ANSWERS -- QUAD CITIES 1&2 -87/07/20-KEETON, J. ANSWER 5.07 (2.00) Caused by the reverse power effect. (0.5) When withdrawal.of a shallow rod increases power low in the bundle (0.5)
.this causes the void content to increase in the upper part of the bundl e . (0.5) . If the negative reactivity due to voids is larger than the positive reactivity- due to rod withdrawal, a power decrease will result. (0.5)
(Exact wording not critical, accept correct discussion of reverse power that contains all of the above elements.) REFERENCE Quad Cities: THEORY LIC-THEO, pages 62 and 63 l 292OO5K104 ...(KA'S) ANSWER 5.08 (1.00)
" Assume Patm = 14.7 psia at T initial; P = 400 psig = 414.7 psia at T initial; (0.5)
P= 122 psig = 136.7 psia at T initial + 1 hr; P sat of 414.7 psia = T sat of 448.14 F P sat'of 136.7 psia = T sat of 351.19 F Del ta T = 448.14 - 351.19 = 96.95 +/- 5 deg/ hour (0.5) REFERENCE Quad Cities: THERMODYNAMICS, TPO 6 and QGP 2-1, page 9 293OO3K123 ...(KA'S) ANSWER 5.09 (2.50) Longer (0.5) Because the power down transient is limited by the decay of the long-li ved del ayed neutron precursors (0.5) whereas, the power up transient is dependent on prompt neutrons (0.25) and short-li ved del ay neutron precursors. (0.,25) The down transient is stabilized by the void coefficient (alpha V) (0.25) and the decrease in doppler (alpha D) (0.25). The up transient is stabilized by the void coefficient (alpha V) (0.25) and the increase in doppler (alpha D) (0.25). (Accept any reasonable, correct discussion of the af f ect that doppler and void coefficient have on stabilizing power.)
, - 4 ,
THEORY OF' NUCLEAR,PQWER_PL@NT,,_QPE, RAT [QNg_F,LUlDg3,_AN_D PAGE 18 ( '5.
- IdE6dQQyNgdipg
~
ANSWERS -- QUAD CITIES 1&2 -87/07/20-KEETON, J. j 1 q REFERENCE
-THEORY LIC-THEO,_TPO 9 j QuadLCities:
.(KA'S) {
292OO3K106 292OOOK119 . . l l ANSWER 5.10 (1.50)
.o. Remain'the same i
- b. .Less negative _
Less' negative j c.:
- d. More negative
- o. More negative (0.3 pts each).
REFERENCE Quad Cities: THEORY LIC-THEO, Fig. 46 and pages 58-62 2*2OO5K109 292OO4K114 . ..(KA'S) e a a i 1 1 l i 1 < I
PAGE- 19' 6:. _P(@NI_SYSIEdg_QESl@N ,_CQUl@Q6g,_@NQ_lypl6QDEdl@llgd g
' ANSWERS - DUAD CITIES 1&2 -87/07/20-KEETON, J.
ANSWER 6.01 (2.50) l '. . Increase shutdown cooling flow. pg,0ft'veckkre up% vet 64
- 2. - Raise reactor . t essel water l evel to above 47 inches as read on LI-263-101 (upper 400") on 90X-4. This instrument is calibrated for cold conditions. The O-60" narrow range instruments on 90X-5 are calibrated hot, and will read non-conservatively high in the' cold condition. (45" a c ^.u al level'will indicate 60"). }
- 3. Use head spray.
- 4. Start up the-cleanup system.
- 5. Start up both.CRD pumps.
- 6. Bleed through cleanup system, and feed through feedwater system.
- 7. Fl ood the main steam lines, and drain through the main steam line drains if.the cle&nup system is not available.
'(5 @ 0.5 pts each)
REFERENCE Quad Cities: RHR LIC-1000-1, page 50, TPO 9 205000K303 205000K603 295021K102 ...(KA'S) ANSWER 6.02 (1.50) To eliminate an explosion hazard. (0.5) e.
- b. Gases would return to main condenser (vacuum would decrease)
(system flows would change) (0.5)
- c. Adsorbs radioactive gases to allow decay bef ore exhausting to the main stack. (0.5)
REFERENCE Quad Cities: Off-Gas LIC-5450, TPO 2 271000 GOO 7 271000K301 271000K404 ...(KA'S) I h a e I .=en e g g n
. . _ _ . _ _ _ _ . _ _ _ _ _ _ _ __ m____ _ _ _ _ .____________a
ww __
-87/07/20-KEETON, J.
ANSWERS -- QUAD CITIES 1&2 i ANSWER 6.03 (1.00) ,, y The purpose is to maintain h.3 psid drywella amd torus. (0.5) Prevents torus structural damage that could be caused by water hammer from the downcomer in the event of a blowdown during DBA. (0.5) REFERENCE Quad-Cities: Pump Back Air (LIC-1600-3), TPO 1 223OO1K307 ...(KA'S) ANSWER 6.04 (1.25) Rio L
- 1. GR III i sol at i on I'J . 25 ) @ + 8 inches-REGL (0.25)
- 2. SBLC initiation 'O.25)
- 3. High temperature on NRHX outlet (0.25) 140 degrees F (0.25)
REFERENCE Quad Cities: Reactor Water Clean-up System, LIC-12OO, TPO 5
*204000K108 204000K111 204000K404 ...(KA'S)
ANSWER 6.05 (1.50) (E' A/0/h/%e)
- a. Feed flow > 20%A(0.25) and pump discharge valve open (0.25)
- b. 1250 psi (0.25) or -59 inches RWL (w/9s TD) (0.25)
- c. (1) 45 (0.25)
(2) 65 (0.25) REFERENCE Quad Cities: Recirculation System (Flow Control) LIC-0202-2, TPO 2 and 8 202OO2 GOO 5 202OO2K402 25037K301 ...(KA'S) ANSWER 6.06 (2.00) C
- a. 3 v4W
- b. 4 BY(
- c. 1 t ony
- d. 2 (p epp ' ' * -
1
)
(0.5 pts each) i
Lh__Eb8NI_Sy@lEDS_QESlGNg ,CQGI6QLg ,@yQ_lNS169dEyl@IlgG PAGE 21 9
-87/07/20-KEETON, J.
ANSWERS -- DUAD CITIES 1&2 REFERENCE Quad Cities: EHC Pressure Control & Logic LIC-5650-2, TPO 2 241000 GOO 7 ...(KA'S) ANSWER 6.07 (1.00) ' [p.,;(fy. [e ,% euddh &O Asl~Arawn $wbd H c. Both buses (13-1) and (23-1) deenergize. - (0. 5 )
.b. The'first unit to experience LNP energizes. (0.5)
REFERENCE Quad Citiest Diesel Generator LIC-6600, TPO 7 264000K101 264000K506 ...(KA'S) ANSWER '6.08 (1.25)
- a. IRM Inop. Trip. (0.25) and a Rod Block (0.25) and 1/2 scram (0.25)
,b. . Thr ee channel s .LGr25') [C,5)
/ s 14.Oypnxek 7 ..;; e w d.c.e c.,"j eg ; G r.ge 9 7 ccept d M). ;O.25) O'e/ETC3 REFERENCE Quad Cities: Intermediate Range Monitor System (IRM), LIC-0700-12 LO 9 and 10.
215003K101 215003K301 215003K302 215003K401 215003K402
...(KA'S)
ANSWER 6.09 (1.50) p La$ g.[* n s/ W tNl'<"/'W h f * (0.25 pts each in any order)
- b. (1) APRM < 307.
(2) Edge Rod selected (0.25 pts each)'
- c. 2 (0.25)
REFERENCE Quad Cities: Rod Block Monitor System (RBM), LIC-700-5, LO 4 and 7 215002K104 215002K301 215002K604 ...(KA'S) I 9 b...._____.__._______,__,______________ _
- 6. --PLANT SYSTEMg_QEg[QN _CQNTRQLt_ANQ_lNgTR,UMgNTATiQN g
PAGE 22 L- . .
-87/07/20-KEETON, J.
ANSWERS -- QUAD CITIES 1&2. l* l i ANSWER- 6.10 (1.50) c.' False (0.5)
.b. TMs&r&> Tra e- (0 5)
- c. True (0.5)
REFERENCE Duad-Cities: Vessel Level Control System LIC-0600, TPD 3, 6, and 10 259002 GOO 7 259002K301 - 259002K302 ...(KA'S) g6 ,s
, J P-
, j
; ,. /._ f, L , - , ,,
PAGE 23 7.1. _P,,gg gE gg g g@ _ , _yggd@( t _@ggg60@( t _qyE,$@@GQ{ _@y g B89196gG1C06_CQNISQL
-87/07/20-KEETON, J.
' ANSWERS -- DUAD CITIES 1&2-ANSWER 7.01 (2.00)
'a. 1. During startup (0.5)
Timing CRDs (0. 5)J)(CRD Fri cti on Testi ng)[C.5) 2. f. Emergency sitt.*i6ns (0.5) 6* r .3 Q c. s n)
- b. Shutdown margin demonstration.(0.5) e4 RrNd A/orcecs, of / /q
& CA) *AfftY > 3 l & ort REFERENCE 66 76 b^) rt L 6A) <. JW4M VACM G CM) ,
% MND y x y pf,U)
' Quad Citiest DOP 280-1, page 3 [ GC# /-/ gg64 % [s9Ct'4A/47%
201002G010 ...(KA'S) t4.) d 43' /4) G 201002 GOO 1 fAUY 004 &o A cct f 7<Dh ANSWER 7.02 (2.50)
- a. 1. Commence a normal shutdown
- 2. Stop the Recirc Pumps
'3. Stop the RWCU pumps
- 4. Notif y the Load Di spatcher (Any 3 & O.5 pts each)
- b. /The primary concern i s l oss of cooling to the drywell.
(Drywell pressure will increase as the dr well heats up (f.0)
'^
L= cw/A JVL47.PQL4.)(p dchmea ~ .4fa., }<civh - ODULys~e.kb. .R
- REFERENCE . A <.A ' cA.&n.g di r L. c Q W9Pe&.f' Duad Cities: DDA 3700-2, page 1 (0o6 % .
295018G011 29501BK202 295028K202 ...(KA'S) ANSWER 7.03 (1.50) Adm' istrative and Superintendent. a. tnf 444<&A 5pu, Support nndenYl7: Servi cep Assi stantV/.au.'cs.J' G<'rs<&
- b. None.
- c. Rad-Chem Supervi sor or designee.
(0.5 pts each) REFERENCE Duad Cities Radiation Pretettion, LIC-RAD PRO, page 54 ! 294001K103 ...(KA*S) l l s . l l e.
i
. m._____-
. R@QlQLQG1G96_QQN2@QL
-87/07/20-KEETON, J.
~ ANSWERS -- GUAD CITIES 1&2 ANSWER 7.04 (1.00) l; i
of Torus-to-drywell vacuum breakers.
' Prevents cycling
(_K'Sl'yNe-j^ ' . . $dk.di v.Mc5utanispA n/A [ g,,, wm%p2j., . {c,
- REFERENCE d Quad Cities
- DOP 1600-2, page 1
.261000 GOO 5 .261000K111 ...(KA'S)
ANSWER 7.05 (1.50)
- a. Jet pump failure. (0.5)
- b. 1) Reverse flow through the failed jet pump causes INDICATED core flow to increase. (0.5)
- 2) With reduced power, turbine control valves close to maintafa l header pressure reducing generator load. (0.5)
REFERENCE Quad Cities: QOA 202-1 202OO1A201 202OO1K601 295001K302 ...(KA'S) ANSWER 7.06 (2.00) 1.- Evacuate the Reactor Building EO.53 to minimize radiation exposure to personnel CO.53.
- 2. Verify initiation of (or initiate) SBGTS [0.253 and vent i sol at i on
[0.25] to prevent release of radioactivity to the environment.EO.53. ( Al so accept isolate other unit ventilation and trip fans.) REFERENCE Quad Cities: 004 BOO-1, page 1 295023G010 295023K101 295023K301 ...(KA'S) J s
996D96t_@gNQ60@6g_E_dE_QQENgy_hNQ PAGE.- 25 - Z:..,_P69qEQUBE@_ . 609196991G06_G9dIB96 i
-87/07/20-KEETON, J.
' ANSWERS -- QUAD CITIES 1k2' i
ANSWER- 7.07 (1.50) g?f-3 fM
- c. (1)
(2) 30 minutes _(0.25) ,e 62 amperes (0.25) n~ -h p [gM8 60 "70 dY)
- b. (1) 70 F/hr (0.25) l (2) 100 F/hr (0.25) (
- c. Any LOCA would onl y aid in depressurization and reaching cold shutdown sooner. (0.5) (Within the life of the battery.)
REFERENCE Quad Cities:- GOA 6900-7, pages 1 and 2 263OOOG014 295003K106 295004G010 ...(KA'S) 2.15
' ANSWER 7.OS (.2. E n
- a. NONE
- b. NONE
- c. QGA 100-1, 100-7, 100-3 (RPV Level, Pressure, and Power Control Simultaneously)
[%"VONEmor
-4 . 9GA-100 iMOO-2T- 11:)P3MRPV Level , " essure,~antf--N er Conts,ul, Ermal t _: _ M y; ::_ c/f /4 g g
- e. QGA 200-1, 200-2, 200-3, 200-4 (SP Temp, DW Temp, Pri. Cont. Press, ,
and SP level Si mul t nneousl y) l
- f. QGA 100-1, 100-2, 100-3, 200-1, 200-2, 200-3, 200-4 (as above Simultaneously)
- g. -NCNC QQ/9,2E'h/fek
- h. QGA 300-1, 300-2, 300-3 (See Cont. Temp, Rad level, and Water level Simultaneously)
- i. QGA 100-1, 100-2, 100-3 (as above Simultaneously)
J. NONE (0.25 pts each, all that require simultaneous entry must be stated for full procedurey4W & b/s; a, /CO S/M$,2pc> Q1gg, g/c.. ) credi t) /110 l t - i
lNN1hEhhishi,[QQN3@Q!,
-87/07/20-KEETON, J.
' ANSWERS -- DUAD' CITIES 1h2 l
- REFERENCE ~
DGAs, Entry Conditions Quad Cities: 295024G011 295025G011 295030G021 295037G011' ...(KA'S) 2
# s 4 f L- _
et__eQUINIS16@llyE_P6QCEQUBESi_CQNQ11[QN@g_@NQ,(ldll@llgNS PAGE 27 ANSWERS -- QUAD CITIES 1&2 -87/07/20-KEETON, J.
' ANSWER B.01 (1.50)
'1. It-is impractical to modif y the equipment or procedure. '
2.. =The test is per a regular procedure.
- 3. - An operator is available to immediately return the switch to normal when test is completed or interrupted.
(0.5 pts each) REFERENCE
- Quad Ci ti es: QAP 300-2, page S 203OOOGOO1 203OOOK401 206000 GOO 1 206000K407 ...(KA'S)
' ANSWER .8,02 LW/ST [/, f C
- 1) Exceeds 72 hrs in 7 day period. (0.5)
- a. ,
- 2) Exceeds 24 hrs in 48 hrs. (0.5) (Wed-Thurs = 30 hrs) ~
- 3) Less than 8 hrs between work peri ods. jg (0.5) (Thurs-Fri = 6 hrs)
? . '-Y ,1 y/ E' M 5 N- A AYbf MNr
.. f
~
. j -
~ .,m,,,.-.
- -, -y. -.
REFERENCE / # _ ' 17' ^' ' ' Quad Cities: QAP 300-3, page 1 and 2 201003 GOO 1 201002 GOO 1 201001 GOO 1 ...(KA'S) 202OO2 GOO 1 I ANSWER 8.03 (1.00)
- 1. Shift Engineer (0.75) j i 2. Annual l y (at least) (0.25) 1 l
REFERENCE ! Quad Cities: QAP 300-24, page i 286000K303 286000 GOO 1 ...(KA'S) ANSWER 8.04 (2.00)
- a. The heat transfer time constant (8-9s) (0.5) is such that the neutron f;ux will be seen at detectors and terminate the transient before peak centerline temperature can be transferred to the L cladding surface. (0.5) (Accept any correct discussion of '
relationship between time constant and neutron flux.)
- b. 1) 1.5s (0.5)
(0,5)
- 2) Immediately upon exceeding the setpoint w/o scram.
~ \ -
u . N -
maw - - ~ _ _ -_--
~'
-87/07/20-KEETON, J.
ANSWERS -- QUAD CITIES.1&2
. REFCRENCC' Quad Ci ti es: - Technical Specifications, Saf ety Limit.1.1.C and Basis 215005G006 295015 GOO 3 295037 GOO 4. ... (KA'S)
ANSWER- 8.05. (1.00) c
- 1. - One: shift 2.- 24 hrs-(0.5 pts each)
REFERENCE Quad Cities: QAP 300-14, page 9 294001K102 ...(KA'S) ANSWERL B.06 (2.00)
- a. 4
- b. 3
'c. 6
- d. 2 (0.5 pts each)
REFERENCE Quad Cities: QAF 300-13, page 1 and 2 294001K102 . . . (K A' S) ANSWER 8.07 (2.00)
- a. . Alert
- b. (Notification of) Unusual Event
- c. General Emergency
- d. Site Area Emergency
( O.5 pts each) REFERENCE Quad Cities: QGA 5-2 295026G011 295025G011 295024G011 294001A116 ...(KA'S) a
- 02. _ePD191EI5ellyE_PBQCEQ(JBE@t_CQUpil199@t_@U211011@l199@ PAGE 29 ANSWERS - 'GUAD CITIES-1&2 -87/07/20-KEETON, J.
l l ANSWER 8.08 (2.00)
- 1. Report to the control room upon notification of fire.
- 2. Notify Chicago LD.
3.- Notif y Rock River LD in Rockf ord.
- 4. Classif y fire emergency as GSEP (per GEP 2OO-TI). .
(0.5 pts each) k!<W &tecwcrg. B Y ^< n ,.. . --! !L - " AAmy accue WM f y.!tj,iguy, REFERENCE Quad Cities: DEP 340-5, page 3 and 4 (/ AN
- M 294001K116 ...(KA'S)
. ANSWER 8.09 (1.00) c.. October 1st (92 days) (0.5)
- b. . Grace Period is +25% of the frequency not to exceed 3.25 for three consecutive surveillance. (0.5)
" REFERENCE Quad Cities: Technical Specifications, Section 1.0 206000 GOO 5 ...(KA'S) 1 i
ANSWER O.10 (1.50)
- 1. Mode switch in Startup/ Hot Stby and refuel platform over the reactor.
- 2. Fuel on any refuel hoist and refuel platform over the reactor.
Mode switch in Refuel with one control rod withdrawal permit. 3. (0.5 pts each) REFERENCE
' Quad Cities: Technical Specifications, 3.10.A.1 234000K401 234000K402 ...(KA'S)
O e I L
- s - b > , , , , ,
e-_ = .. - . . . . . . . . . . . . ... . . . . . ..
PAGE 1 0 TEST CROSS REFERENCE ~
. nR: '
QUESTION VALUE REFERENCE
.05.01 1.50 KMJOOO1227
, 05.02 1.00 'KMJOOO1228 05.03 2.00 KMJOOO1229 05.04 1.25 KMJOOO1230 05.05 1.00 KMJOOO1231 05.06 1.50 KMJOOO1232 05.07 2.00 KMJOOO1233 05.08 1.00 KMJOOO1234 05.09- 2.50 KMJOOO1235 05.10 1.50 KMJOOO1236 15.25 06.01 ;2.50 KMJOOO1237 '
06.02 1.50 KMJOOO1238 06.03 1.00 KMJOOO1239 06.04 1.25 KMJOOO1247 '
'06.05 1.50 KMJOOO1248 06.06 2.00 KMJOOO1249 06.07 1.00 KMJOOO1250 ( 9 06.08- 1.25 KMJOOO1251 06.09 1.50 KMJOOO1253 06.10 1.50 KMJOOO1254 15.00 i
07.01. 2.00 KMJOOO1240 07.02 2.50 KMJOOO1241 07.03 1.50 KMJOOO1242 , 07.04 1.00 KMJOOO1243 07.05 1.50 KMJOOO1244 07.06 2.00 KMJOOO1245 07.07 1.50 KMJOOO1246 07.08 2.50 KMJOOO1252 14.50 08.01 1.50 KMJOOO1255 08.02 1.75 KMJOOO1256 08.03 1.00 KMJOOO1257 08.04 2.00 KMJOOO1258 08.05 1.00- KMJOOO1259 08.06- 2.00 KMJOOO1260 08.07 2.00 KMJOOO1261 08.08 2.00 KMJOOO1262 08.09 1.00 KMJOOO1263 HOB.10 1.50 KMJOOO1264 t 15.75 60.50 i
. . .. .. . - - ____- _ ______._ ______________-_s ____ J
k.._ j
\
T 1 b ~U. S. NUCLEAR REGULATORY COMMISSION i ' REACTOR. OPERATOR REQUALIFICATION EXAMINATION FACILITY: QUAD CITIES 1&2 p aw:,, gWR-GE3,_________________ REACTOR TYPE: p,sW <,
@ Z lg Z Z '2 9_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
q h ,(. ;. '. DATE ADMINISTERED:
%p# .A ,y
#*t** EXAMINER: GLORK i _F ________________
CANDIDATE _________________________ JN@ISWCIJgN@_IQ,Q9Nplp@IEi This examination replaces
. Read . the attached instruction page facility administered car ef ully. requalification examination.
the ' current cycle as failure of this examination are the same Retraining requirements fora requalification examination prepared and administered in by for failure of for each question are indicated staff. Points 70% your ' training parentheses after the question.grade The passing grade requires at leastpapers in each category and a final of at least 80%. Examination will be picked up four (4)' hours after the examination starts.
% OF
% OF CANDIDATE'S CATEGORY CATEGORY
___ggg6E___ _y@kUE__ ______________g81EGQ6Y_____________ __M8 bye _ _19196
- 1. PRINCIPLES OF NUCLEAR POWER
_15:.ZE__ _Sf!I2 ___________ PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
- 2. PLANT DESIGN INCLUDING SAFETY
_1Er39__ _2S_ d2 ___________ ________ AND EMERGENCY SYSTEMS lt.S INSTRUMENTS AND CONTROLS _k M f_~_ _2dI92
- ___________ ________ 3.
IS. 8S 4. PROCEDURES - NORMAL, 4BNORMAL, _ M 5__ _2+r@@ ___________ EMERGENCY AND RADIOLOGICAL CONTROL 6 /. 9 Totals _Hr$7__ ___________ ________x Final Grade I have neither given All work 'done on this examination is my own. nor received aid. ______________ignature Candidate's S j
/""% n ,j t d . v s,f - ,_ ,--i Dl/
p r , a t ~ 6D v v! }
- f.g s
\fhbi)
Pegr 4 11__EBjNCJghgg_QE_ NUCLE 88_EgWEB_E69NI_ GEE 681]QN 2
.IbESDQQyNSDJG@i_dESI_I68N@EE6_8NQ_E(Ulp_E(QW QUESTION 1.01 (1.50)
Using the attached Figure 1, STATE why:
- a. Core reactivity DECREASES from point A to point B.
- b. Core reactivity INCREASES f rom point B to point C
- c. Core reactivity DECREASES f rom point C to point D QUESTION 1.02 (2.00)
TRUE or FALSE
- a. During equilibrium power conditions, the production rate of indirect Xenon from Iodine is faster than the decay rate of Xenon to Cesium.
lowers the height of the
. b. Slowing the rate of a power decrease, resultant Samarium peak.
- c. The resultant Xenon peak due to a scram from 50% power is larger than a scram f rom 100% power.
- d. During an increase in power from equilibrium Xenon conditions, Xenon concentration initially decreases.
QUESTION 1.03 (2.00)
- a. Consider two control rods, both at notch position 16. Rod A is locatedThe near the center reactor of the scrams aftercore and rod operating at Bfull is power locatedforat200 the hours.- core edge.
A hot startup is performed and 10% power is attained ten hours after the scram. STATE WHICH control rod (A or B) would add the most reactivity at this time with a one notch withdrawl and EXPLAIN WHY.
- b. STATE if a f ully inserted control rod has greater differential worth if located next to a FULLY WITHDRAWN or FULLY INSERTED control rod anc EXPLAIN WHY. ASSUME STEADY STATE POWER AND EQUILIBRIUM POISONS.
f 5;! MAv , t , h vvi c (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
Pegs 5 g It _ _ES'IN gig (E S _QE _NUC(E @B _EggE8,6(@NI_gCE 6@I1QN l IHEBdQQXU6diq@s_dg61_IB90EEE5.069_E6W19_E69W i l' i GUESTION 1.04 (1.50) STATE if the following become MORE NEGATIVE, LESS NEGATIVE, or REMAIN THE 1 l SAME: ) a. Rod worth (delta K/K/ Rod) as f uel temperature INCREASES l
- b. Alpha doppler (delta K/K/ degrees F change in fuel temperature) as voids DECREASE :
- c. Alpha voids (delta K/K/% voids) as core age INCREASES
- d. Rod worth (delta K/K/ Rod) as voids DECREASE as core age INCREASES l
- e. Alpha doppler . (delta ' K/K/ degrees F f or f uel)
QUESTION 1.05 (2.50) The reactor is operating at 100% power and flow. Condenser Ci r c ul at i on water temperature increases 10 degrees F over a ten (10) minute time period. STATE whether the f ollowing INCREASE, DECREASE, or REMAIN THE SAME and EXPLAIN why each is or is not affected.
- a. Condenser vaccum.
- b. Reactor power.
QUESTION 1.06 (2.25) 350 psig. A LPCI automatic initiation occurs with reactorINDICATE pressure ifatthe parameters Reactor pressure then decreases to 100 psig. REMAIN THE SAME and EXPLAIN why the i listed below, INCREASE, DECREASE, or I parameter changes or remains the same. l
- a. LPCI injection flow
- b. LPCI pump discharge head ( assume constant NPSH )
- c. LPCI pump power requirements
,~ e - ; ,
, 7-,
r b .s ...*,.? { .
, s j. - 3 ,
$i tb l I\ b .((
4 (***** CATEGORY 1 CONTINUED ON NEXT PAGE ****s) ,o
Page 6 PRINCIPLES OF NUCLEAR POWER' PLANT OPERATIONg 1.+
'- ISE5099%Ngd1GSt,dg@l_I69ygEgg,@NQ,E(ylp_ELQW QUESTION 1.07 (2.00) a.
Power is reduced using control rods. The Recirc. pumps remain at a constant speed. STATE if core flow will INCREASE, DECREASE, or REMAIN THE SAME and EXPLAIN why.
- b. At low power operation (< 20%) , with the Retirc. pumps at minimum speed, power is increased with control rod withdrawal. STATE if core flow will INCREASE, DECREASE, or REMAIN THE SAME and EXPLAIN why.
QUESTION 1.08 (2.00) CHOOSE whether the following will ADD NEGATIVE REACTIVITY or ADD POSITIVE l REACTIVITY during plant operations and EXPLAIN why. (INCLUDE IN THE EXPLANATION, WHICH COEFFICIENT CAUSES THE CHANGE IN REACTIVITY)
- a. Moderator temperature increases while below saturation temperature.
- b. Fuel temperature increases.
- c. Loss of a feedwater heater.
- d. A sudden fifty (50) psig reduction in reactor primary system pressure.
MASTF "Oir ' (***** END OF CATEGORY 1 *****) b
Pago 7
~
22,,E699I_DggJgy, jug 6UpJyg_gSEgIY_999_gyg69gygy SYSIEDS DUESTION 2.01 (2.00) Injection (LPCI) system. With ' regard to the Low Pressure Coolant
- a. STATE the LPCI initiation signals.
to
- b. STATE the interlocks which must be satisfied in order divert injection from the reactor to drywell spray with a LPCI initiation signal present.
QUESTION 2.02 (1.60) a. LIST all the conditions under which the IRM downscale trip is automatically bypassed.
- b. LIST all the conditions under awhich the RBM is automatically bypassed QUESTION 2.03 (1.80) flow is normally 75 gpm. EXPLAIN in Control Rod Drive (CRD) cooling water flow changes immediately after a SCRAM.
detail why CRD cooling water QUESTION 2.04 (1.50) The ATWS system inserts negative reactivity by performing two specific actions. STATE these actions and what SPECIFIC COMPONENTS operate to accomplish these actions? QUESTION 2.05 (1.00) EXPLAIN why diesel generator loads are designed to energize sequentially. al i-.
!i C O ?Y
(***** CATEGORY 2 CONTINUED ON NEXT PAGE sasss) A_------_-_-__
Pago 8 h_ @NI_QESigN_lyCgp1N@_S9 Eely,@Np_EdER@ENCY EYEIEd5 i OUESTION 2.06 (1.50) Instrument air has been lost to the Control Room HVAC system. STATE what instrument air.
. automatically happens to the f ollowing due to the loss of
- a. The Control Room HVAC isolation dampers. J
- b. The condensing unit cooling water supply valves.
OUESTION 2.07 (2.00) recirculation { Unit 1 is in a cold shutdown condition is.+ with 52 both reactorBoth RHR pumps are pumps operating, reactor vessel level inches. l operating in the shutdown cooling mode to maintain reactor vessel TT5 ' temperature. Reactor vessel level is decreasing due to a mispositioned valve. ASSUME NO OPERATOR ACTION until the LEVEL DECREASE is DISCOVERED and TERMINATED at -15 inches. LIST and EXPLAIN the automatic actions that will occur to the RHR System Shutdown Cooling Mode as the reactor vesselv is lowered. INCLUDE APPROPRIATE l SETPOINTS. y i QUESTION 2.08 (2.00) The Standby Li quid Control (SLC) System is designed to provide suf ficient reactivity compensation to shut down the reactor from rated power to zero power. This includes the 3% DELTA K/K shutdown margin and to allow the nuclear system tc be cooled to 125 degrees F with control rods wi thdrawn tc' a full power position. LIST four (4) additional reactivity gains that must be compensated for by SLC system as specified by the system's design bases. QUESTION 2.09 (2.00)
- a. STATE the purpose of the Extraction Steam Non Return Check Valves
- b. EXPLAIN how a f ailure of the B Non Return Check Val ve , to CLOSE on a turbine trip could cause the turbine to overspeed.
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(~"5~ I ;% . g\ v . { g i l (***** END OF CATEGORY 2 *****) i
page 9 31__1NSIBUMENIS_gND,CQNIRgLS QUESTION 3.01 (1.50) a a. EXPLAIN why the Mechanical vacuum pump automatically trips at 7X normal signal from the main steam line radiation detectors b. STATE which jet pumps have instrumentation taps that also serve RPV level instrumentation and which type (s) of RPV level instrumentation is/are f ed by these taps. QUESTION 3.02 (1.00) TRUE or FALSE a. With the reactor at 100% power the water level in the core region is six to eight inches higher than the water level in Only the downcomer the narrow region due to the pressure drop across the dryers. range GE/MAC and narrow range yarways are accurate during this condition. b. The Recirculation Flow Control System limiter prevents damage to the recirc pumps by ensuring the discharge valve is f ull open to limit the axial thrust on the recirc pump. QUESTION 3.03 (1.50) The plant is operating at 45% power when you notice that the " A" l evel LPRM f or the selected control rod (centrally located) is operating erratically.
"A" l evel LPRM is bypassed.
This LPRM is assigned to APRM Channel 5. The Based on this inf ormation, STATE if the following will cause a/ HALF SCRAM, mp . A'u) FULL SCRAM, or ROD BLOCK. (g / 4 ,e The LPRM's in ASSUME APRM Channel 5 has ,k2' OPERABLE LPRM inputs. a. the f our-rod display have the f ollowing distribution level, and 2 of"D"INOPERABLE level. LPRM's: 2 "A" level, 2 "B" level, 1 "C" IN The LPRM's in ASSUME APRM Channel 5 has }5 OPERABLE LPRM inputs.
- b. INOPERABLE the f our-rod display have the f ollowing distribution level, and 1 of "D" level.
LPRM'S: 2 "A" level, 2 "B" l evel , 2 "C" M,** (""" P O \
\
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L } [ -Q i u *s 1 I (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) 1
Paga 10 L.__]NSI69tLENJp,ANg_CgNI$pg) { QUESTION 3.04 (1.50) i TRUE or FALSE
- a. A HALF SCRAM will occur if MSIVs A and B <2re shut in the RUN raode and reactor power i s . equal to or less than 50%.
i With the reactor. operating at power, you are instructed by the f b. SRO to FULL SCRAM by placing the mode switch in SHUTDOWN and to l IMMEDIATELY reset the SCRAM. The circuit will not allow the scram-to be reset immediately to ensure the scram valves open. QUESTION 3.05 (3.00) The reactor is operating at 40% power with the Feedwater Control System in selected for the controlling single element control, and level channel "A" input. The channel "A" Narrow Range GEMAC develops a problem such that the measured differential pressure starts to slowly decrease and continues until at zero dif f erential pressure.- EXPLAIN the effect(s) on the systems listed in a. through f. below, that is/are CAUSED by the above failure. (INCLUDE APPLICABLE SETPOINTS) (ASSUME NO OPERATOR ACTION)
- a. ACTUAL Reactor Pressure Vessel Water Level
- b. Channel "A" INDICATED Reactor Pressure Vessel Water Level
- c. Feed Water Level Control System
- d. Reactor Protective System
- e. RCIC, HPCI, LPCI and LPCS Systems
- f. Primary Containment Isolation System
- g. Recirculation System y -}
t ( 1-D\! g - rs /\ - i l-l'\ R O 'i ~.. At l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) l l' w _ _ _ _ _ _
Pego 11 3:.__INSIg g EN]p_gyD,CgNJggby s QUESTION 3.06 (2.25)
"A" Electric Pressure The reactor is operating at 75% power when thethe EHC system shifts to the "B" EPR.
Regul ator . (EPR) detector fails and INITIALLY change and WHY7 How would the following parameters
- a. Reactor pressure
- b. Core Flow l
- c. Reactor power GUESTION 3.07 (2.00) to the recirc l The plant is at full power and the feedwater flow signal '
system f ails to zero.
- a. HOW and WHY will the speed of EACH recire pump be af*fected?
(INCLUDE ANY APPLICABLE SETPOINTS)
~
b. HOW and.WHY will the speed of EACH recirc pump be affected if the speed controller output on A MG set f ails to zero at the same time the feedwater flow signal fails. OlmsTIGN 5.ve ~ <1.50) hG WM4e-vncvTng the ref uel platform in the reverse direction the platf ur m s to;L%_--4.-IST711 iNitMLUCKS that_cpuld cause the refuel platform to stop.
@ St.1ME M FawEn raitunti MAS ~E. COoY
*****)
(***** CATEGORY 3 CONTINUED ON NEXT PAGE cE_-____---------
Pcg2 12, 3___INSIguyEyly_3hp_ggyIR9L@- (. QUESTION- 3.09 (1.50) Given the following information: the reactor;is at 100% power
--APRM CHANNEL 1 is' reading 102%
- FLOW SIGNAL CONVERTER UNIT 1 output is 90%
- FLOW SIGNAL CONVERTER' UNIT 2 output is 102%
- 3 LPRM signals to' APRM CHANNEL 1 are bypassed INDICATE If the following conditions are TRUE or FALSE:
- a. RPS CHANNEL A. tripped (1/2 scram)
- b. Control' Rod withdrawl block c.. APRM H; -HI' FLUX /INOP al arm
- d. APRM HI FLUX alarm
- e. FLOW REFERENCED OFF NORMAL alarm
-s I
i
? - .
4, f [\/I,/ks,'~: I b.i s L(ll E (***** END OF CATEGORY 3 *****) 4 m.____
Pag 9 13 EMERGENCY
- 4. PROCEDURES - NORMAL 3 ABNORMAL 2 ,
OND_5091969GIGOL_G9 NIB 96-
.(>
6 DUESTION 4.01 (1.00)
'TRUE or FALSE >
- a. Procedure DAP 300-12,." Placement of Plant Jumpers and Blocks,"
does not have to be followed when.the temporaryofalteration the DOS is. part of an DOS request and is controlled as part program and is returned to normal when the DOS is cleared.
- b. Once an Emergency Operating Procedure has been exited,.it need not be.re-entered even if an entry condition is reached.
QUESTION 4.02 (1.00) LIST four (4) of.the five (5) duties the oncoming NSO is required to do per after assuming the OAP 300-7, " Shift Change for Nuclear Station Operators", Unit 1 NSO duties. QUESTION 4.03 (1.75) Immediate Operator Actions of GOA 201-10, " Reactor Pressure Control Using Manual Relief Valve Actuation," step 1.c states, "DO.NCT actuate any singis. relief within ten (10) seconds of that valve being closed.
- a. EXPLAIN the reason for the above caution statement.
- b. EXPLAIN how the operator can determine if the 203-3B and 203-3C relief valves have been closed f or less than ten (10) seconds.
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6 (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) C-______ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
( ' Pcga 14 ,{ PROCEDURES - NORMAL g _ ABNORMAL 2 _ EMERGENCY' /
'.' 4 . -
ONQ,89Q1QLQGigS6_gQNISQL
'l QUESTION. 4.04 N [ 8,9 TRUE or FALSE a.
In accordance with QOP 207-2, " Declaring Rod Worth Minimizer (RWM)-can be Computer Inoperable,".the Rod Worth Minimizer bypassed if a startup is in progress and the following conditions. . exist;
- 1. All rods are at position 00
- 2. The.RPIS input to the RWM fails, i
- 3. A second verifier is present.
- t. M're:crdant: =tTn a nuTE in OOP 1600-10; " Nitrogen Inert 1DQ of i Unil--4--and 2 FFTmary Contai nment . Usi ng tne Unit CI::tric
-m YApsc.12mcrs-centrires tnat a 1.z ps1o oitterenti.1 Let-weert the_
j 6 er .5teb1i? d --!er +o i drywell--ancHetrpprunrETon cnamirar ter-mi-nat+ng-tn e inerting process. 922d g [
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I QUESTION 4.05 N ,- STATE whether or not. emergency procedure entry is required for each of the following conditions. If NO entry is required state "NONE", if ENTRY ! conditions are required state "YES". .! CONSIDER EACH ITEM SEPARATELY and ASSUME NO ADDITIONAL CONDITIONS. i i
- a. Reactor Pressure Vessel level i s 10. inches Reactor power is 12% in the STARTUP MODE
- b. $4D99&' l n 4' ..-
_c. R ea r- t er nm- i c 93*' c . -i nute-.after = maine lead reiection4 m
& n cc ap I-i uti - :::ur2 mr i r,g p- ---r m --: dlAA4
- e. Suppression pool level is -3 inches l
- f. Drywell pressure is 2.5 psig
- g. Suppression pool level is 13.7 feet
- h. Reactor buliding ventilation exhaust is reading 3.2 mR/hr ,
- i. The reactor is shutcown and reactor pressure is 1090 psig 4 C,0NTINUED ON NEXT PAGE *****)
(*{**fCATEGORY. ,~ n ; , 1: ..
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Pega 15 9 __E599EQUSEg_;_NQQU@6t_@@NQ60@bt_EDE6QENQy , GNQ_SSpig6QQ1Q@(_QQNIBQL !
' QUESTION 4.06 (1.50)
" Station Fire Answer the following questions pertaining to OEP 340-5, ,
Fighting." i
- a. You discover a fire while on a plant tour. LIST the information required to be reported to the control room, b.
STATE what actions are required by the control room operator upon notification of a fire. QUESTION 4.07 (2.00) While operatindag under accident conditions per the EOPs, adequate core cooling is a major concern.
- a. LIST the three (3) available methods, in their order of preference, to assuro adequate core cooling and STATE how each method of adequate core cooling can be verified.
- b. STATE why this is the preferred order.
QUESTION 4.08 (1.50) STATE three (3) of the four (4) items that should be checked to assure that I a procedure is a valid procedure before using it to operate plant equipment. l O' g 9 t- , s. 3 ,. l (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) l
~ -- -- -_________._ _ __
______________________________g
Pega 16-
- 4. PROCEDURES _ _NgRMAL3 _ABNQRM@L 1 _EMERgENgy
- ONQ_$@pJg(QQlG@b,QQNISQL QUESTION 4.09 (1.50)
Answer the f ollowing questions about a 45 year old individual with a current lifetime whole body exposure of 131 REM and a current NRC f orm 4 on file. (ASSUME THE DATE 15 JANUARY 1, 1987)
- a. LIST the individuals 10 CFR 20 and QUAD STATION allowable whole body exposure f or the first quarter.
- b. LIST the individuals 10 CFR 20 and QUAD STATION allowable whole body exposure for the year.
- c. LIST the 10 CFR 20 and QUAD STATION allowable whole body exposure per quarter for an individual (45 years old) who does not have a current NRC form 4 on file.
, QUESTION 4.10 (1.50)
While removing a spent fuel bundle from the core, the grapple-fails-releasing the fuel bundle. STATE the immediate actions to be taken Per QOF B00-1, " IRRADIATED FUEL DAMAGE WHILE REFUELING," and EXPLAIN WHY. J. b ,
, , . .: l;. u i 's. vs he L r_
1 1 (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) b- _ - - .
Paga 17
- c. PROCEDURES - NORMAL g _ t ABNORMAL _ EMERGENCY ANQ,68Q1Q(QQ1Q@L_QQNIBQL l
QUESTION 4.11 (1.00) TRUE or FALSE
" Normal Unit Startup," a sustained reactor
- a. According to QGP 1-1, not be allowed.
period of less than 45 seconds shall
" Unit Startup to Hot Standby,"
- b. During the performance of QGP 1-2, when reactor pressure reaches approximately 875 psig rods are notched.in to stop the pressure rise and pressure is maintained between 850 psig and 920 psig by the following; 1.- Operating RCIC
- 2. RWCU system reject flow
- 3. Vary power with CR's
- 4. SRV manual operation to prevent reaching 1060 psig if the other three methods fail to maintain pressure.
precaution c. Procedure GOP 202-2, " Recirculation System Startup,"
#4 states, during a loss of cooling water to the recirculation pumps, the pumps must be secured within one minute to prevent overheating the motor bearings.
from service, d.- The precaution f or RWCU System filter demin removal"Do not allow the flo QOP 1200-3, states, system to exceed 180 gpm with the filter domin byp;ss open to prevent RWCU pump runout." QUESTION 4.12 (1.00) LIST four immediate operator actions required by procedure QEP 340-3, (4)Dam
" Lock and # 14 Failure", upon notification that the dam has f ailed.
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**********)
l (********** END OF EXAMINATION l l 1 I 6 1
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- g = ga
. Idgg=1d22 Id g =Id2 WATER ?_ARAMETE11 1 gal. = 8.345 rom R/hr = (0.5 CE)/d (meters) ,
R/hr = 6 CE/d (feet) 1 gal. = 3.78 liters
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I., Cudi.a =.:j3h7 x f0 dps Density = 62 lb,s/Ic , i kg = 2.21 1ha Density = 1 gm/cm 1 hp = 2.54 5 103RTU/hr Heat of valorization - 970 Ecu/lba 1 N = 3.41 x 10 Stu/hr Heat of fusica = 144 Stu/lba 1 Ata = 14.7 Psi = 29.9 in. Ig. 1 Blu = 778 f t-lbf 1' inch = 2.54 cm I ft. H O 2
= 0.4333 lbf/in F = 9/5'C + 32
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- 100 0 S492 77.96 1031.4 1109.3 041617 265.4 265.4 5 110 12750 1.9339 120 1025.6 1113.6 0.1646 1.7693 203.25 203.26 87.97 1.9112 130 120 1.6927 041620 97.96 1019.8 1117.8 0.1817 1.7295 0D1625 157.32 157.33 0.1985 1.6910 1.8895 140 130 2.2230 107.95 1014.0 1122.0 0 ' N9 122.98 123.00 0.2150 1.6536 1.8686 150 140 2.8892 117.95 1008.2 1126.1 Os 134 97.05 97.07 1.6174 1.8487 180 150 3.718 127.96 1002.2 1130.2 0.2313 CDs640 77.27 77.29 150 4.741 170 1134.2 02473 1.5822 1A295 62.04 62.06 137.97 996.2 180 5.993 021645 1138.2 02631 1.5480 1.8111 170 50.22 148.00 990.2 7.511 041651 50.21 0.2787 1.5148 1.7934 190 180 40 96 158.04 964.1 1142.1 9340 0D1657 4034 02940 1.4824 1.7764 200 190 33 64 168.09 977.9 1146.0 0.01664 33.62 0.3091 1.4509 1.7600 210 300 11.526 178.15 971.6 1149.7 OD1671 27.80 27.82 210 14.123 1.7568 212 970.3 1150.5 0.3121 1.4447 26.78 26.80 180.17 1.7442 220 212 14.696 0.01672 965.2 1153.4 0.3241 1.4201 23.13 23.15 188.23 1.7290 230 220 17.186 041678 958.7 1157.1 03388 13902 19.364 19.381 198.33 1.7142 240 230 20.779 0A1685 952.1 1160.6 03533 1.3609 1
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't 's 0 1021J a.Oges
't 0 2.1872 2.1872 3302.4 0A0 1075.5 1075.5 4.10 3.03 1022J Sate 6 82.018 041602 3302.4 '
i 3 43 1073.8 1076A 00061 2.1705 2.1766 13.50 10253 S.15 2945 5 020 8.10 35423 041602 2945.5 2004.7 LY.50 1067.9 1081.4 00271 2.1140 2.1411 SJo 115 45.453 CD1602 2004J 1526.3 21.22 1063.5 10841 0.0422 24738 2.1160a.4021.22 1 32.54 1032.0 40.92 1034J SJO 53.160 041603 1526J 10393 32.54 1057.1 10893 00641 2.0168 2.0809 eJo 64.484 041604 1039.7 792.1 40.92 1052.4 1093.3 00799 1.9762 2.0562 47.62 1036.9 8.5 SAO 72A69 041606 792 4 47.62 1048 6 1096.3 0.0925 1.9446 2.0370 as 641.5 0.7 El 79 586 001607 641.5 540.1 5325 1045.5 1098 71A966 0.1028 2 0083 1.9186 5810 1040.32 4215' CA 5324 0.6 85.218 0 01609 540.0 466.94 58 10 10423 1100.8 03 CS 0.7 90.09 041610 466 93 411.69 62.39 1040.3 1102.6 0.1117 IJ775 1.9970 6239 1 SA 9438 0 01611 411.67 368 43 66.24 1038.1 1104.3 0.1264 IJ606 1.98701.06624 10 6913 1044.1 e.9 98.24 0.01612 368.41 2.0 333 60 69 73 1036.1 1105.8 0.1326 1.8455 94 1.9781 03 1051 A 84 2.0 101.74 0.01614 333 59 17376 94 03 1022.1 1136.2 0.1750 1.7450 1.9200 4A 24 126 07 0 01623 173.74 11813 109 42 1013.2 1122.6 02009 1.6854 1A864 EA 109 3A 141.47 0.01630 118 71 90 64 120.92 1006.4 1127.3 02199 1.6428 1A626 120 9 4.0 152 % 0.01636 0.01641 90.63 73.515 73.53 130.20 1000.9 1131.1 02349 1.6094 1.8443 , 5.0 130 16224 7.0 SA 0.01645 61.967 61.98 138.03 9962 1134.2 02474 1.5820 1.8294 144.81 992 l' 1136.9 025P1 1.5587 1A168 150A4 1069.2 1067.4 th 138 0 4.0 170.05 53.65 144.83 0.2676 1.5384 1A060 BA 0 01649 53 634 968.5 1139.3 156.28 1070A 7A 176.84 47.35 150.87 0.2760 1.5204 1.7964 10 0.01653 47J28 985 1 1141.4 161.23 10723 80 182.86 42.40 1 % .30 0J836 1.5043 11879 42J85 9A 18627 0.01656 38 404 38.42 16126 982.1 1143.3 180.12 1077.6 1 & s96 19321 0.01659 03121 1.4447 1.7568 10 26182 2620 180 17 970.3 1150.5 15 0.01672 181.16 1077.9 1L996 212.00 2629 181.21 969.7 1150.9 0.3137 1.4415 1J552 196 21 1082.0 20 , 213 03 0 01673 25274 196.27 960.1 1156 3 0.3358 1.3962 1J320 218.8 1087.9 30 40 15 to 227.96 0 01693 20 070 20 087 218 9 945.2 1164.1 0 3682 1.3313 1.6995 2360 1092.1 250.34 0 01701 13 7266 13 744 2361 933.6 1169A 0.3921 12844 1.6765 250.1 10953 50 30 40 267.25 0 01715 10 4794 10 497 8.514 250.2 923.9 1174.1 0.4112 1.2474 1.6586 80 0.01727 8.4%7 262.0 1098.0 70 50 281.02 7.174 262.2 915.4 1177.6 0.4273 1.2167 1.6440 272.5 1100.2 30 80 29211 0.01738 7.1562 6.205 272.7 907.8 1180.6 0.4411 1.1905 1.6316 281.9 1102.1 90 302.93 0.01748 6.1875 282.1 900.9 1183.1 0.4534 1.1675 1.6208 290 4 1103.7 70 5.4536 5.471 894.6 1185.3 0.4643 1.1470 1.6113 298.2 1105.2 100 80 312.04 0.01757 4.895 2901 90 32028 0.01766 4A777 4.431 298.5 888 6 1187 2 0.4743 1.1284 1.6027 120 32722 0.01774 4.4133 312.2 1107.6 140 100 3128 312.6 8772 1190.4 0.4914 1.0960 1.5879 324.5 1109.6 34127 0.01789 31097 3.219 325.0 868.0 1193.0 0.5071 1.0681 1.5752 335.5 1111.2 150 180 120 353D4 0.01803 3.2010 336.1 859.0 1195.1 0.5206 1.0435 1.5641 1A0 363.55 0.01815 22155 2 134 346 2 8507 1196.9 05328 1.0215 1.5543 345 6 111 300 150 130 373D8 0.01827 2.5129 22689 2.531 2287 355.5 842.8 1198J 0.5438 1.0016 1.5454 354A 250 111 200 381A0 0.01839 825 0 1201.1 0.5679 0 9585 1.5264 375.3 SFO 1 13432 3761 808 9 1202.9 0.5882 09223 1.5105 392 350 400 9 1 400 97 0.01865 11245 1.5427 394.0 794.2 1204.0 0 6059 0.8909 1.4968 408 450 6 1 250 1.5238 41735 0.01889 1.3255 409.8 300 431J3 0.01913 1.3064 1.1610 780.4 12046 0 6217 08630 1.4847 4221 1 1118.9 350 ace 444.60 0.0193 1.14162 .0318424.2 437J . ~ 767.5 1204A ~& .A.6360,,CA378 r . '. .
.1.4738< 4353 500 3 41224 4%.28 0.0195 ~.
450 l o a0.90787 , 0.9276 449.5 e 755.1 '12047? 0 6490 0A148 1.4639 4473 550 500 11 [ 900 467Al 0.0198 743J 1204.3. 0.6611. 0 1936 1.4547 458 700 91 M A2183 b 03418 460.9 550 900 476A4 0.0199 48620 0.0201 PE74962- 0.6556 491.6 0.7594 710.2471.7 1201A 0.6928'732A07377 "1203J 1.4304 D.6723 488 800 9 11-700 503 48 0.0205 063505 0.54809 0.5690 509.8 689.6 1199.4 03111 07051 1.4153 506 900 1 11 800 51821 0.0209 669 1 1196.4 0.7279 0 6753 1.4032 523.2 1 0.5009 526 7 900 531 95 0 0212 0.47968 0.4460 542 6 650 4 1192.9 0.7434 0.6476 13910 1107.5 1100 1104.3 1200
$38 6 1000 544.58 0.0216 E42436 0 4006 557.5 631.5 1189.1 0 7578 0 6216
- 1.3794 553 1 55L28 0.0220 OJ7863 613 4 1184.8 01714 0.5969 1.3683 566.9 1100.9 1300 1100 0.3625 571.9 1300 MU1 0 0223 0.34013 CJ0722 0.3299 585.6 5946 1180.2 0 3843 0.5733 1.3577 500.1 0.0227 1300 EM.42 0.3018 598 8 576.5 1175.3 01966 0.5507 1.3474 1093.1 1500592 587A7 0.0231 0.27871 0J772 411.7 558.4 1170.1 0A085 0.5288 1.3373 605.2 1400 50620 0 0235 0J5372 0.1883 672.1 46E2 1138.3 02625 0.4256 1.2811 662 6 1900 3000 E35A3 0.0257 0.16266 0.1307 731J 361.6 1093.3 0 9139 0.3206 1.2345 973.1 9000 718 2500 GER 11 0 4296 0.10209 045073 0.0850 801A 218.4 1020.3 0.9728 0.1891 1.1619 7f2A 475.9 320s 3000 e5.33 0 0343 0 1A612 875.9 CONS 906.0 0 906.0 ' 1.0612 L 0.0508 8 230L2 30L47 I
~ ~
F l
$4
- TeWe3 d . ;
~
Properties et eigerbested senem and compreened water se 9em9eratursF . aspnees. ' 700 WO 900 1000 1100 1800 1300 h400 1900 / 000 000 400 tilet is. Gusttemp) 14D W 300 0 0161l sta 462.3 511 3 871.5 W1.1 000.7 6 500' 11902 11M.7 1241A last4 1336.1 13s4.5 e 8 2J237 22708 23144 197.70 30BA2 221.53 233 45 2 5 00 2.1152 2.1722 1 9 9 4 18333 1748.0 1803 5 (101.74) e 0.1296
, 0A161 EM 90.24 10224 11421 1M.15 1380s 19001 161.M 17326 te57516 1JES.2 1335.9 154.3 14336 14837 l
3 4 mal 11E6 11MA 1241J 2A460 2A032 2.1369 2.17M 221599522521 34 104E 110.M 114.72 1 3716 12300 13943 40.00- 74.98 SO M 36S1 92 1639 37 5 1893 3 1747.9 1303 4 (142.24) s 0.1295
, 0.0161 BA4 44.98 51 23 57.04 63 1364.0 03 1433 4 1483 522101 2.1757 1534.6 1586 6 22430 22744 2J04 30 a 502 114L6 1193.7 12406 (19321) s 0.1295 1 1.7153 13583 13173 187A 1335.5ISES2 241M 2.0603 2.1011 2 37.905 41.986 45 375 49 964 e 0.0 161 ADIM 1192 J 29299 13 963 1383 8 14332 1443 4 1534.5 1506 5 1639.42.198 5 1239 9 1287.3 1335.2 2.0155 2 0%3 2.0946 2.1309 2.1M3 15 4 68 04 18809 1A134 1A720 13242 1.9717 43 435 46 420 49 405 52 21303 8 55.370 5 1883 1 1747A 3 (213 03) s 0.1295 42M O 31 AE6 34 465 37.455 40.447 1586 3 1639 3 22J56 25A28 28457 2.1336 2.1665 2.1979 2 2282 2.25 30 e 00161 SD146 a W O5 18B11 1191.4 1.7005 1239.2 13397 12921 1296.913397 1834 9 1383.5 1.983623 IM 2.0244 1432 9 24 689 26183 2.0628 1832 27476 29348 153 2.09f )
(227.96) s 0.1295 02940 18 699 20 199 21497 40 e 00161 E0166 4 641D le15 1196 6 IJ608
' 12361A143 11.036 12MA 14.165 15 685 17.195 14992 4 12850 13624 13331.9065 6 1382.5 1.M 1 (26725) s 0.1295 02MO
, 0A161 AD166 7.257 1181.6 1.71M 8.354 12335 1.7681 1283.2 9A00 1A168 1A612 1332.3 10 1.90241M1.5 425 13410,13774 11 438 1545.3 1638 4 1892 4 1747.1 180 2.0120 2.0450 1431.3 2A765 2.1068 1481.8 1215 4 2.1 40 6 4815 18520
. (292.71) s 01295 OJ53911.6492 6.218 7.018 7.794 8360 1.9800 9319 1544.9 1638.0 1820 17462 1802 2 A131 2A646 10.075 2.0750 2.11 30 6 e
6821 18B24 04161 E0166 0.0175 ' 0.1295 02B39 04371 1.6790 269 1.7349 74 1230.5 11842 1A289 1281.3 13702 1330.9 12009 9A60 10460 1380 13454 1637.6 1891.6 1746.5 11.060 11459 1802.2 5 14 (812.04) s 4.935 5.588 6 216 4 833 7.443 8050 130E3 2219986552 0602 9258 2.0794 68 26 15J9 269 77 1227A 1279.3 1329.6 1379 8.5 a 04161 04166 0.0175 100 6 0 1295 0 2939 0.43711 6516 1J008 IJ586 1 3036 1A451 1 At39 13205 1.9552 61928 6J006 7.2060 1637.1 73096 1891.3 1746 9 7130 S.2119 2 1402 0 (127 A2) s 4.0786 4 6341 5.1637 5.6831 1 3600 13996 2.0300 2 0592 e 0 4161 6 0166 0 0175 EL31 14833 26941 1224.1 1A296 14872 1277.41.7376 1328.111829 137841A246 1428 81A635 1479A13001 1531. 120 6 0.12E5 82539 04371 6.1709 46036 7Abe9 1745 7A652 9 1801 7 7A9 16361 1890.9 1 (341.27) s 3 4661 39526 4.4119 4AS45 52995 5.7364 13508 1AR25 2.0129 2.042 e 0A161 84106 0.0175 SJ7 IEJB 269 85 1220A 12753 14085 14606 1.71% 1.7652 1A071 1326A 1877A 1A461 1428.0 1AR28 1.9176 1479.1 1530A 1 140 6 0.1255 82R39 CA370 95 5.0132 5JD45 5U41 6.15223 65 1745.6 18014 (353.04) s 34060 34413 3A480 42420 4621427.2 1475.4 1982.9 1.9027 1530 16363 3 1600.51 335 5 0A141 84146 0.0175 1217.4 1273.3 1325 4 1376 4 IJ919 1A310 1A678 100 6 Se 14B 42 200A9 1.5906 1 4522 13039 1.7499 4E67 5A014 6.1363 6.4704 s 0.1394 8 2938 0 4370 42 (363 55) 2.6474 3.0433 3 4093 33621 4.1 5 4 4.4505 43907 5.1 1124.0 13753 1426.3 14MJ 1529.7 1582 4 13227 r DAl61 E0156 CD174 12712 1.UB4 1A176 1AS45 12094 5 5209 5 3219 130 & E 47 15 47 269 92 15743 12133 14376 14900 1J362 s 0.1294 0J938 04370 (373 08) 4 2J598 21247 3 0583 3.3783 3 6915 4 0008 430U 4.61 300 (38120) 6 ES2 148 51
, CA161 SalM 0017 s Statt 8J538 0 4369 i t.5593 1 4242 14776 269 1123996 11863 1210.1 2 4462 '2AB72' t.M10 3.190944382 38837 f
13057 13426 1269.0 5 3 709 4.4131 4.6 least 1744.2 18002 1A77546 7M 03t61 4A146 OA174 Opla6l2.19081263 5 1319.0 13714 14M 14233 1527A 41J4051500.6 L 1.7901' .1634 41A524.1 1A173
, 373.10 % %2. - 36746 32764 300 6 E46 18 63 270 05 1.5951 14502 3 1A076 g 1
,,, e s 1795 4 (400 97)s 83394 4 2937 0 4368 q ;,.43467 ,
407 2.460B 22585 3A643 SJW 3 1J686 9.0044 22263 2A e 4 341 0.0166 00174 40196 375.15 12573 1315.2 1368 1J9919 14213 13964 12317 14731AE52 6 1526.2 1AB72157 R79 18834 270.14 1.5703 1A274 14754 IJ192 300 m (417.35)s 432B4 42937 44307 0.5465 2J652 2 1575.2 44451432 2A219 3 1887.1 23950 17426 17954 2 13105 1.M OA186 1A913 11028 1 AB70 20B32 e 48141 601M 0.0174 1251.5 1311.4 13662 1J787 1419.2 13411 1A141 1471A 1A471 1524.7 13798 500 t E t2 lea 5 270J4 375J1 1.5443 1AOU 1A671 1.7009 (43113) e 412R3 42536 0436710.5464 13759 2.1339 1576D 18312 2J901 12Ae0 3955 1 2250 2.5 04162 541 1.4763 1245.1 1907.4 136314499 4 1 A1511417A 1JWN 1470.1 15233 13325 1A647 at0 e 8 $m 5 601M 04174mm 1 A32 1.500135971A406 270.331375.27 AB50 1J2551 Eft.1 1J632 122M 1 (444 GO) s 412W 42335 E4306 0.5M3 0.W19 1.1594 1.3037 1 4397 1.5708 1A092 1574 4 1484 4 1740 3 1 1A702 1 AW9f . 157J 1412J 1466 6 1520.3 1.U 30 1 Ass 1ABR3 e tale S4106 40174375JB 84186 12312 1399.1 1 3905 1A123 14678 1AS90 1J371 age 4 SR 300.19 270.51 ,
. (447
.01) 6e ,125 42334 04364 OA400 l1A321 l
W
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7 masse / flamedynsmiss of steem j Table 3
;'g iAe et supertiested steem and rg::M water (temperstwo and presswe)
Temperstwo, F me pose. Eh/sulla- 300 800 900 000 700 900 900 1000 1100 1200 1300 1400 1900 gest,tnany) 100 100 03944 0.94% 14726 1.1992 1.3008 1A093 1.6160 1A211 1.7252 12284 1.9309 e 00161 0.01 % 0 0174 0.0186 1215.9 1290.3 1351 A 14083 1463.0 1517.4 1571.9 1627.0 1526 17385 1795 6 ( 30 e 88.58 1 # 42 27030 375A9 1 A590 1.5329 1.5444 14351 1A769 1J1% 1.7517 1.7859 ? Alas 1A494 1A792 (406J0) s 01292 OJS13 0.4362 0.5467 l 01920 0.9072 1 8102 1.1075 1J023 12948 1J858 MA7 1.5647 1A530 e OA161 SA166 0A174 0A186 0.0204 1281.0 1345.6 1403.7 1459A 1514A 1569 4 1624.5 1907 1737.2 1794J 3g0 t 80 34 19 66 27039 375 61 487.93 1.5090 1.5673 1A154 14500 1.4970 1.7335 1.7679 13006 1A318 13617 (503.05) s 0 1291 0 3 32 0 4360 OM55 068a9 04774 0.7828 02759 0.9631 1.0470 1.1299 1.2093 1 JOB 5 13669 1A446 I e 0.0161 OAl% 0.0174 0.0186 0.0204 1271.1 1139.2 1399.1 1455A 1511A 1%6.9 1622.7 1675 9 1735.0 1792.9 goe 6 70 11 19 38 27147 3M.73 487.88 1.4869 1 5484 1.5900 1.6413 1.6007 1.7175 1J522 13351 1A164 13464 (518.21) s 0.1290 A2930 0A354 0 5662 0.6885 0 5869 0.6858 0 7713 0 8504 0 9262 0.9998 1.0720 1.1430 1.2131 1.2825 e 0 0161 04166 0 0174 0.0186 0.0204 1260 6 13323 1394 4 1452.2 1508.5 1%4.4 1620 6 1677.1 1734.1 1791.6 900 4 70 37 170.10 27126 3MA4 487.83 1.4659 1.5311 IM22 14263 1.6662 13033 IJ382 1.7713 1A028 12329 (531.95) s 0 1290 02929 0 4357 0.%49 0 6a81 0.5137 06080 06875 03603 08295 OA966 0.9622 1.0266 1.0901 1.1529 e 0.0161 0.0166 0.0174 0 0186 0.0204 1249.3 1325 9 1389 6 1448.5 15044 1%1.9 1618 4 16MJ 17325 17903 1000 6 70 63 17033 27144 375 96 48739 1 4457 1.5149 IM77 1.6126 1.6530 1.6905 1.72 % 1.M89 1.7905 1A207 (544.58) s 0 1289 02928 0 4355 0.5647 0 6876 e 0.4531 0.5440 0 6188 0 6865 0 7505 0A121 0 5723 0.9313 0 9894 1.0468 0.0161 0.01 % 0.0174 0.0185 0.0203 12373 1318A 13843 14443 1502 4 1559 4 1616 3 1673.5 1731.0 1799.0 1190 6 70 90 170 % 27143 376.0B 487.75 1.4259 1.4996 1.5542 1.6000 1.6410 1.6787 13141 1.7475 1J793 1A097 (5 % 28) s 0.1289 02927 0 4353 0.5644 0 6872 e 0 0161 0.01 % 0 0174 0 0185 0 0203 0 4016 0.4905 0 M15 0 6250 0 6845 01418 0 7974 02519 0 9055 0 9 % 4 1300 e 71.16 1703B 27122 37620 48712 1224.2 1 4061 1311.5 13793 1440 9 1449 4 15% 9 1614.2 1671.6 1729 4 1787. 1A851 1.5415 1M83 1 4298 1.6679 1.7035 1.7371 1.7691 1J996
.(567.19) s 01238 0326 0 4351 0 5642 0 6868 e
03176 0 4059 0 4712 0.5282 0 5809 0 6311 0.6798 0 7272 0 7737 CA195 0 0161 0 0166 0 0174 0.0185 0.0203 1194.1 1296.1 1369.3 1433.2 1493.2 15518 1609.9 1668 0 1726.3 1785 1400 6 71 68 171.24 272 19 376 44 487.65 1.3652 1 4575 1.5182 1.M70 1 6096 14484 1.6845 IJ185 1.M08 1J815 (587.07) e 0.1287 0 3 23 04348 0.506 06859 0.3415 0 4032 0 4555 0.5031 0 5482 0.5915 0 6336 0 6748 07153 e 0 0161 0 0166 0.0173 0 0105 0.0202 0 0236 12794 1358 5 1425.2 1486 9 1546 6 1605.6 1M4.3 1723.2 1782.3 1800 6 72.21 171.69 272.57 376 69 487.60 616 77 1.4312 1A968 1.5478 1.5926 14312 1.6678 13022 IJM4 13657 (604.87) s 01296 0321 0 4344 0 %31 0 6851 0A129 0 2906 0.3500 0.3988 0 4426 0 48M 0.5229 0 %09 0 5980 0 6343 e 0.0160 0.0165 0 0173 0 0185 0.0202 0.0235 1261.1 1347.2 1417.1 1480 6 1541.1 1601.2 1660 7 1720.1 1779.7 1800 4 72.73 172.15 272.95 376.93 487.M 615 58 1A054 1A768 1.5302 15M3 1.6156 1.6528 1 5876 1.7204 1J516 (621 A2) e 0 1284 02918 0A341 0.%26 04843 0A109 ; 0.2488 0.3072 0.3534 0 3942 0 4320 0 4680 0.5027 0 5365 0.5695 e 0.0160 E0166 0 0173 0 4184 0.0201 0 0233 1240.9 1353 4 14083 1447.1 1536.2 1596.9 1657 0 1717.0 1777.1 2000 h 7326 17240 273J2 377.19 487.53 614 48 13794 1.4578 15138 1.5603 1.6014 14391 14743 1.7075 1J389 (635 20) e 0.1263 42916 0 4337 0 % 21 0 6834 CAC91 0.1681 0.2293 0.2712 0J068 0.3390 03692 0J900 04259 0 4529 I I e 0.0160 0 0165 0.0173 0 0184 0 0200 0 0230 1176 7 1303 4 1366 7 14573 1522.9 1585 9 1647A 1709.2 1770.4 ' l 2500 6 74.57 173 74 27427 377A2 487.50 612 08 IJ076 1.4129 1.4766 1.5269 13703 1.6094 1 A456 1.6796 1J116 (448.11) s 0.1200 0.2910 0 4329 03609 0 6815 0 2048 0.0982 0.1759 0 2161 0.2464 0.2770 0.3033 6J32 0.3522 0 3753 e 0.0160 0.0166 0.0172 0 0153 0.0200 0.0228 1060.5 1267.0 1 % 3 2 14402 1509 4 1574 8 1638.5 1701.4 1761A 3000 b M AB 1742B 25 22 378 47 487.52 610 08 1.1966 1.3692 14429 1 4976 1.5434 1.5841 1A214 1 6561 1.6888 (995.33) e 0.1277 02904 0 4320 0.5597 0 6796 0A009 01588 41987 0.2301 . 0.2576 . 0.2827 43065 0 3291 OJ510 o 04160 0A145 00172 0D183> 0.0199 OA227 A0335 3300 4 76 4 175J 2MA 3M.7 487.5 809 4 800.8 1250.9 1353 4 1433 1 1503 A '1570.3 16343 1698 3 1M1.2 1.3515 1.4300 1.4e66 1.5335 1.5749 1A126 1.6477 1A806 (706.08) e 0.1276 0 3 02 0.4317 05592 0 6708 01994 0.9708 02995 0 3198 e 04160 SD164 0.0172 041 Ob1 0.02h5 0 0307" 0 k%4 "017'64 CMk6 okk26 'Oh563 0.27 3000 6 77.2 17E0 276.2 379.1 487.6 608 4 779 4 1224 6 1338 2 14222 1495 5 1%33 1 s 0.1274 0 3 89 04312 05585 04777 01973 0 9508 1.3242 1A112 1.4700 13194 1 e 001% SA164 80172 E0182 0.0198 0.0223 0.0287 0.1052 0.1463 0.1752 0.1994 D 8830 4 78 5 177.2 2 77.1 379 2 487J 406 9 763.0 1174.3 1311.6 1403 6 1481.3 1552.2 1619A 16 e 0.1271 0M3 0 4304 03573 0 6760 03940 0.9343 1.2M4 1.3007 1A461 1A976 1.5 e 0A159 SA164 0A171 60181 0.0196 0 0219 0 0268 0.0591 0.1038 0.1312 0.1529 0 9000 6 81.1 1793 279.1 M1.2 408.1 404 6 746.0 1042 9 1252.9 1364 6 1452.1 1529.1 1600 9 167 e 012% 021 R4257 03550 04726 E7900 0.9153 1.1%3 1.3207 1.4001 1A502 1.5061 0 1684 0.1817 e 0.0150 03163 RD170 60100 E0195 0.2216 0.0256 0.0397 00757 0.1070 0.1221 01391 E1544 E8 4 MJ 1813 51.0 3823 488 6 402 9 736.1 945.1 lies 8 13236 1422.3 1505 9 1982 0 16 8 41250 0370 E4271 RS628 0.4493 0J526 02026 1.0176 IJ615 IJ574 1A229 1 A748 1 e 0A158 Bale 44170 40100 0.0193 0.0213 Cards 0.0334 0 0573 0 0B16 0.1004 0.110 MB e 36J IM4 330 3842 400.3 401.7 729J 901 8 1124 9 1281 7 1392.2 1482 6 1%3.1 163 e R1252 taa8 E42% SM07 0.04&3 0 7777 08026 1A350 IJ055 1J171 IJ004 1 A446 1 A J- - - - - . - - _ _ _ _ _
rx Page 1.- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION t
.IdEBdQQYN@dlG@g_dg@l_IB@N@Eg6_6NQ,E(ylg_E(QM l l 1
I
' ANSWER 1.01' (1.50)
- a. Decreases due to Samarium CO.253 and Xenon CO.253 (will accept fission product poisons f or CO.253
- b. Increases due to Gadolinia burnout (0.253 ;
and Pu-239 building CO.253 , JI l f
- c. astr=2c=, r O rE63 jkse t o f uel depletion C .
REFERENCE QC Theory LIC-THEO, pp 74 and Figure 62 292OO7K101 ..(KA's) 292OO7K103 ANSWER 1.02 (2.00) ,
- a. True i
. b. False
- c. False
- d. True CO.5 each3 REFERENCE l
QC Theory LIC-THEO, TPO 11 ..(KA's) 292OO6K107 292OO6K106 292OO6K105 ANSWER 1.03 (2.00)
- a. Rod B (0.25) Following a scram, fission product poisons cause a severe the core.
flux depression in the highest power producingin region of of low poisor regions CO.253 This results in a higher relative flux concentration. CO.253 The shift in the flux distribution increases the1l worth of peripheral rods and decreases the worth of those in the ' center of the core.CO.53 CO.25] Flux is
- b. The rod next to the withdrawn rod has greater worth higher in this area, thus rod worth is greater. CO.53 REFERENCE / .,, , ,,,
[. p it ch. 6 pg 12 QCTheoryLIC-THEOp/i,GfE. Reactor Theory, [cb)$ kh'. IS',k 18. 292OO5K112 . . (K A? /s ) ib h i'
' u ' ,, (i.1 ,,5
[
*****)
(***** CATEGORY 1 CONTINUED ON NEXT PAGE r
Pcg3 19
.1 .- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION 3-l' -IUE6099Yb8DICg,_bE9I,166NgEgB_6Np_E(y]p,E(pW ANSWER 1.04 (1.50)
- a. Remain the same
- b. Less negative
- c. 'Less negative-
~d.- More negative
- a. More negative CO.3 each3 REFERENCE DC Theory LIC-THEO, Figure 46 and VII.C.1-6 pp 58, 60, and 62 292OO5K109 292OO4K114 ..(KA's)
ANSWcR 1.05 (2.50)
- a. decrease (0.5) the condenser i s a saturated system, if the cooling media temp increases, Tsat increases and Psat increases and vaccum decreases (0.75) b.
decrease (0,5) the temperature of condensate /feedwater increases and l inserts negative reactivity due to moderator temp. coefficient. (0.75) REFERENCE 7-40 to 7-45, SLO-QC Thermodynamics; G.E. HEAT , TRANSFER AND FLUID FLOW '7 50. P. 7.3.4, 7.3.5, 7.4.2, 7.4.3. y F/C 7-/O pgse 293OO7K108 293OO7K106 ..(KA's) 291006K100 s , / , t /~ ' i y , l* ** f. , f /g., , ,
' . g. i., a (2.25) JLw *j",f,} ',;7 re -l,,/(j [co .yj .J y,< ,7, '}, , ,4 7
,, p ,, ,
ANSNER 1.06 . , , s .,;/,si ,. . <
- w i ,, o. . + . . t' u ) '
? p es, 4 /u:n i
p
- a. increase CO.253, as the pressure of the system decreases characteristic. the flow CO.53 increases due to the centrifugal pump head / flow b.
decrease CO.253, as the pressure of the system decreases the pump
!cwer, operatingj point on the pump characteristic o ft.
curveP isshiftedtoa
- F A N+ l q
J dischar.ge ,_pressureJ. . C O... 5 3 (-
. c% ., -.
e r b h g,., I m
,1 t.
w m. - <s. e s ' o 2 . - . ,.t . increase CO.253, from the pump characteristic curve as the flow 7 . r f .,
- c. requirements increase. [0.53 /U (capacity) increases the po6 t
4
~- ! y
, , . , . ,, m , f, ., r .~
f' k.,4 L M Lg G. ' ':j [* j \',, i. / ' t ,, L u. . . ) (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) i t - -- )
q (' Pega 20
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION S
.IUE60QQYNedigSi_UgeI_IgeNSEEg_eNQ,E(gip _E(QW 1 l
l (
)
REFERENCE 1 j I i DC Thermodynamics, Sections L, G and R G.E. Heat Transfer and Fluid Flow, Ch. 6 pg. 6-95 & 6-96 25L1ACL"! ^5 ..(KA*s)
'? 930 C& NO O i
ANSWER 1.07 (2.00)
- a. Core flow will increase CO.53 due to less two phase flow resistance.
CO.53
- b. Core flow will increase CO.53 due to greater natural circulation.
CO.53 REFERENCE B-40 to 8-49. QC Thermodynamics; GE Heat Transfer and Fluid Flow pg. 293OOOK129 ..(KA's) t
. .: a a : A ANSWER 1.08 (2.00) {, w he;d -, (e .", ' t .
.a ~ ')(
eicege CO.253
- a. Adds negative reactivity due" <-
Moderator temperature coefficient. CO.253 ; Adds negative reactivity due to the increase in neutron CO.253
- b. !
capture in the fuel Doppler coefficient. 4 CO.253 edmA. + m- - T, "~, ,- ll / .g
. t s. .y A . . . L e ,4 r.
f 1, ... . 4 ir acute:r CO.d53
- c. Adds positive reactivi ty due%--th; d:;rae;;CO.253 - ,
leakage Moderator temperature coefficient. < b b " ' "j ~., , 4 ,, \ , e ;: < ; te ,lq As< \- r y ir,;c.... in :Wnn CO.253 L r*i-
~
- d. Adds negative reactivity due't the l l
1eakage Void coefficient. CO.253 REFERENCE pg. O, 9, 16, 24, 34, 35, 37; QC Theory LIC-THEO; G.E. Reactor Theory ch. 4, 292OOOK111 ..(KA's) 292OO4K110 292OO4K101 292OO6K122 i r*' - {l- 'Q. E ff \ .'fr [f,. 'i f*a E' b(* k.J[ L
,..-s %f n pi Ly .. i'.- L L.L .
(***** END OF CATEGORY 1 *****) 1 i i
l Paga 21 2:_ _ E L ONI_ pg gj @y_ J yCLUply@, @ @E gIy_ @y p_ g D E 69E N C y 5%5I502. 1 I j(y 7.( A' ANSWER 2.01 (2.00) .
)
- a. 1. Hi drywell pressure 2.0 psig' (0.34) )
I Lo Lo Rx water level -59" (0.33) and I 2. less than 450 psig Rx pressure or (-59"for 8.5 minutes l continuously 6 (0.33) b. With'an auto initiation signal presentt [pt.,re n)gre, wogpac jccepre,p in "ON" (0.25)
- 1. Containment cooling permissive switch (0.25)
- 2. Vessel water level is greater than 2/3 core height OR Containment cooling 2/3 level and ECCS initiation bypass switch in bypass (0.25)
- 3. Drywell pressure must be greater than 1 psig (0.25) .
REFERENCE j
- OC RHR LIC1000-1, pp 17, 18, 42 and 44 i 203OOOK401 ..(KA's) 226001K403 ANSWER 2.02 (1.60) a.
Reactor node switch in "Run" CO.43 or Range 1 on the IRM CO.43
<30% CO.43
- b. Edge rod selected CO.43 or Reference APRM REFERENCE QC RBM LIC-700-5, pp RBM-8; QC IRM LIC-700-2,215003K102 pp 4 and 20 ..(KA's) 215002K101 215001K103 215003K103 ANSWER 2.03 (1.80)
(or near O) CO.33 due to the CRD system Cooling water flow would be zero Flow through the flow being directed to recharge the accumulators CO.33. valve CO.33 and the charging flow element controls the flow control Sensed f low is greateri connection is downstream of the flow element CO.3]. than the set flow CO.33 causing the flow control valve to close and stop cooling water flow. CO.33 l f - ~ ,,,c r.: 7 3- [ f \. L f '_[ l*
9 L .%.. 4 , .g, 3
( YI ... ,. (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) .i - - _ _ _ _ _ _ - _ - _ _______________ - ______________a
.~ _--_
( Page.22-U 12 ._ELBNI,pgglGN,JNC69pjNg,53EgIy,9ND,gDgBggNgy. ' iEYSIEDS-I i l REFERENCE j QC CRD Hydraulic System LIC 0300-2, pp 58 '1 L 201001K412 ..(KA's)
. l.
; ANSWER 2.04 (1.50)
Alternate rod insertion (0.37) by opening ATWS scram valves (0.38) Retire pump trip (0.37) by opening the MG Set field breaker - (0.38) REFERENCE QC'ATWG LIC 0300-3, pp 4, 8 and 10 ..(KA's) 202OO1K306 202OO1K401 202OO1K127 2.05 (1.00)
-ANSWER 0,25 The D/G is 'a small power supply JOf5I'and sequential loading protects the generator from being overloaded due to equipment starting current. (5. 00rSi 7f REFERENCE QC' Diesel Generator LIC 6600, pp 16 264000K506 ..(KA's) 264000GOOB
' ANSWER 2.06 (1.50)-
a. The isolation dcmpers fail to the Train B mode- CO.753
- b. The cooling water supply valves shift to the emergency. supply position (opening the RHRSW header valves) CO.753 REFERENCE QC Ventilation LIC-5750, pp 12 and 13 290003K603 29000K401 ..(KA's) j
.. e p g i h .,
"o ' '
m i '. (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) + ~----- - - - - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
Pega 23 2..' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY 3YEIEUS' ANSWER 2.07- (2.00) the suction and discharge path for RHR the Group II At + B inches RPV level shutdown cooling mode isolate CO.753 due to the initiation ofand the RHR pumps trip isolation CO.753 (MO-1000-47, 50, 29A and 298 shut) due to loss of suction path. CO.53 REFERENCE QC RHR LIC-1000-1, pp 34 and 36 ..(KA's) 293OOOK135 293OOOK136 , 293OOOK137 ANSWER 2.08 (2.00)
- 1. Decay of Xenon inventory
- 2. Eliminating steam voids
- 3. Reduced doppler effect 'O
( -rl ~ q j .j- [ f{ m) I ('.G Decreased rod worth (as moderator cools) 4 4. (4 @ 0.50 each)
- 5. Imperfect mixing REFERENCE.
GC SBLC System L IC-1100, pp IG2 and 3 , 211000K301 ..(KA's) l ANSWER 2.09 (2.00) valves prevent water in the
- a. The extraction steam non return check following a turbine trip, due to ,
feed heaters from flashing to steam,condenser vacuum. C1.03 the turbine casing equalizing with in the feed b. A turbine overspeed could be caused by steam flashing heaters, flowing back into the turbine, through the turbine and into the condenser. C1.03 REFERENCE pp. 9 OC Main Turbine and Turbine Auxiliary LIC-5100/5600, 239001K110 ..(KA's) 239001A303 ' C e n (p,f l 4 f l't I $,.:,/;. ( ,, s. bd % 0 y L ' A-s t [, , . . 2 *****) 4 \ ( #(Cits* END OF CATEGORY
Pago 24-
-h__Is!IB!:!!1ENJg,$ND,CgN]BgLE
. ~) .
- 3. O l' L.1,40+-
ANSWER holdup line
- a. The pump discharges to the chimney via the gland seal i
[0.53'which bypasses the Offgas system filters and would. release radioactivity to the environment [0.53 I l 6 and 16 [0.253 Yarway Wide Range M- bewer 400= 0.233 i b. 4 pg c Gov a- o .k gf;de I
- M '^ ^ '
REFERENCE G g Q O,2 2 w e!-) QC OFF-GAS LIC-5450, pp. 18 and 22 QC Reactor Vessel Instrumentation - LIC-0263, 271000K408 Sec. B.4.e. ..(KA's) 216000K501 216000K122 271000K302 ANSWER 3.02- (1.00)
- a. False
- b. True [0.5 each3 REFERENCE QC Lesson Plan - Reactor Vessel Instrumentation - LIC-0263, Sec. E.a Quad Cities LIC - 0202-2. 216000K122 ..(KA's) 202OO.1K402 216000K501 ANSWER 3.03' (1.50)
- a. Half Scram k
- b. Rod Block [0.75 each3 REFERENCE Quad Cities LIC - 0700-3 and LIC - 0700-5.
..(KA's)
~215005K103 215005K101 ANSWER 3.04 (1.50)
- a. True
- b. False CO.75 each]
e '# hr t ' (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) .n ____ _ _ - _ _ _ _ _
V
- Pegs. 25
" , -. & __JNg]$ WEN]p_$N9,CgNJB9L) .j E
. REFERENCE Quad Cities LIC - 0500-1.
GC, Lesson Plan, LIC-0500-1,RPS, pp 44. 212OOOK114 201001K405 ..(KA's) I ANSWER 3.05 ~(3.00) L (0.3)
- a. Actual vessel l evel is decreases
- b. Channel "A" level will indicate an increasing water l evel . (0.3)
FWLCS will close the-FRVs to try to maintain channel "A" indicated c. reactor water level at +30" (0.3)
~~"
^c.'S i ,i ti at i si, O w th = ::c time
- d. Reacter scram 9 +8" :r.d
- d: lay (0.3)
(0.3)
- e. RCIC and HPCI initiation 9 -59" ( 0. 3 )- 6/c -
LPCI and LPCS ini ti ati or)_9 -59" conginous L.P ts ir it 'N u - b1 m for 8.5 minutes AR 3p j, , , e pt er ,, k PCIS Group 2 and 3 isolations +8" (0.3)
- f. 1. (0.3)
- 2. PCIS Group 1 isolation 9 -59" I
- g. Recirc pumps runback to 30% speed when feed flow goes < 20%
(2 x 10E6 LBM/HR) (0.3) Recirc pumps trip 9 -59" and due to ATWS signal (0.3) REFERENCE QC Reactor Vessel Instrumentation, LIC-0263, pp. 12; QC Scrams and Isolations LIC-Scrams, pp. 46 and 48; Recirculation System LIC-0202-2, pp 8 259001K108 259002K103 259002K101 216000K112 259002K508
..(KA's)
I l 3 l 1
. ,- -: ,r -
r r 8 h .- !
~
t,
; i, e
!. L q
, ;, ; g L. .u . V<(
1 (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) C..........-- - .
Pcgo 26 3:__INgIggDENIg_SyD_CgNIgg6S ANSWER 3.06 (2.25) Increases (0.25) due to the control valves throttling shut in response a. to the EHC system shif ting controlling pressure regulators CO.25] and the backup EPR. set 10 psi higher.CO.253
- b. Increases (0.25) due to the reduction in the void content of the'two phase mixture in the core (0.25) because of the pressure increase CO.253
- c. Increases (0.25) due to the collapse of voids from the higher pressure which adds positive reactivity. (0.5)
REFERENCE pg. 4-24 QC Thermodynamics; QC Theory LIC-THEO; G.E. Reactor Theory, ch. 4, pg. 8-41. G.E. Heat Transfer and Fluid Flow, ch. 8, 241000K102 241000K101 ..(KA's) 241000K302 241000K301 ANSWER 3.07 (2.00)
- a. Both pumps will runback to 30% (0.5) because the FW flow signal to the recirc system is less than 20% (0.5) (the flow limiter is not bypassed when this signal is less than 20%)
- b. B pump runs back as in part A (0.34). A pump speed remains the same (0.33) because of a scoop tube lockup caused by a loss of speed signal. (0.33)
REFERENCE QC Recirc System LIC 0202-2, pp 8, 12 and 22 ..(KA's) 202OO2 GOO 7 202OO2K402 202OO2K604 AN (1.5 "mv r =O'=1 i ng Ni Et f0 79) 1,- Arry-Y5nTr ol rod withdrawn and iws1 Or in stetup/hM **mndhy fo 7M)
- 2. Br i dge._nver--vessei- aTTH~ mode swi tch REFERENCE QC Refueling LIC Refuel, pp 42 . _ g 234000K401 234000K104,.,i. .. . -( K A's ) f /' ,
h l- t . ' h, ' - / qs ,4 m _ s ;, 3 CONTINUED ON NEXT PAGE *****) i (***** CATEGORY
Pago 27L !
,; h__lh!!IBWENI),$Np_CgNIB9k!, . \
ANSWER. 3.09 (1.50) !
)
i
' a. False b.- .True
- c. False j
- d. Fa1se.
True !
- e.
. CO.3 each)
REFERENCE QC RPS System LIC 0500-1, pp 40; APRM System LIC 0700-4,.pp APRM4-9' 215005A104 215005K506 . . .(KA's) e
'4 O
f Rp
- As (,,*'W ****"? t** f*? ,
i ~.g ',,e
,, f '\ 1, - f
'~^ J 's :, lI
_, '\l 't ,( ' ,, ',. ;l? :. c (***** END OF CATEGORY 3 *****)
~ _ _ - _ _ _ _ _ _
Pcga 20
- 4. ,P6995998gg,;,,ygBD662,,$gygBU961,gDg8@gygy 999_ bed 196991G06_G99IB96 e
i , ANSWER 4.01 M
- a. True (O.5)_ ,ygA e % p ar -1Gr3 T~ OOJra q --
be REFERENCE fcc( 6 C- [O6) QAP 300-12; QC QGA Caution #1 and Statement after entry conditions 294001K102 ..(KA's) 295000G011 ANSWER 4.02 (1.00)
- 1. Review " Operation weekly summary of daily surveill ance" sheets j (DOS 005-S1) for the previous three shifts.
2. Transfer to the log for the present shift, any information that should be passed on to the next shift.
- 3. Check the control room panels per the shift surveillance sheet, initial the surveillance sheet and report any off normal conditions to the shift engineer.
4. Read and initial the unit log f rom the previous time he filled that position. 5. Review temporary procedures in temporary procedure notebook and initial l the shift surveillance sheet.
- REFERENCE QC QAP 300-7 294001A113 ..(KA's) 201003 GOO 1 201001 GOO 1 218000 GOO 1 ANSWER 4.03 (1.75) a.
The time delay is used to prevent excessive containment loads CO.253 due to a high water leg in the relief discharge line CO.253 which could cause structural damage from a water slug forced into the suppression pool CO.253, if the relief were opened shortly af ter (within ten seconds) a previous operation CO.253. b. The 203-3B and 203-3C relief val ves have amber lights on the controls; ,. , CO.53 l, ? , I '. . .
- ;. 1 ,
n,)k[I
'( (. . L. l ,L \ l; s
' -(, !, [
j k (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) a
( Pcga 29: h__EB99EgyBEE_ _NQ8d b _6pNgBd h _EdgggENgy
.9Np_BSpjQkggig66_ggNIBQ6 REFERENCE QC DOA 201-10 219000G005 ..(KA's)
I ANSWER 4.04 -41 ,-O M [ c) . f)
- a. False g n -
_s. Tri - rGr5 =.ch] QM REFERENCE QC DOP 207-2 and GOP 1600-19 201000 GOO 5 ..(KA's) 1
' ANSWER 4.05 w )
- a. none
- b. none 6 -n " A -- M T A t N f; P c- v==
. _4.-- ,- dcCcG~/
- e. yes
- f. yes
- g. n nn =~ qc %
- h. yes
- 1. yes CO.2 each]
REFERENCE QC OGA Entry Conditions ..(KA's) 295030G011 295025G011 295024G011 295037G011 l ANSWER 4.06 (1.50)
- a. Your name (0.3), fire location (0.3), nature of the fire (0.3) b.
Sound the fire siren, stop the alarm, announce the location (0.3) and; then sound the fire alarm for one minute. (0.3) REFERENCE QC QEP 340-5, pp 1 and 2 294001K116g ..(KA's) e M f t
/ ,, / (, 6 j
%%,s 4 he
% k, ' ha $
J. (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
l Paga 30
- 4. ' PROCEDURES - NORMALg_ ABNORMAL 1_ EMERGENCY ANQ,8@QJQLQGigeL_ggNIBQL ANSWER 4.07 (2.00) l
- a. . 1. Core submergence CO.253 verify by 2 ::;..; cf level indication ;
that level is above TAF CO.253 !
- 2. Spray cooling CO.253 verify by one core spray system operating at or above design conditions CO.253
- 3. Steam cooling CO.253 verify by use of steam flow over fuel and out of vessel via SRV's as per EDP (cannot be verified by direct indication) CO.253 b.
This order preferred because this is the order from highest to lowest heat transfer coefficient (0.53 REFERENCE QC GGA EOP, pp 4; QC Thermodynamics ..(KA's) 295031K303 295031K302 295031K101 295031K304 ANSWER 4.08 (1.50)
- 1. controlled copy
- 2. approved
- 3. current revision
- 4. date of use is af ter the effective date or marked with the approved date and O.C.O.S.R. stamped on the procedure Cany 3 G O.5 each3 REFERENCE QC QAP 1100-8; QAP 1100-1, pp 3; QAP 1100-2; QAP 1100-4, pp 1; QAP 1100-5 and QAP 1100-6 294001A101 ..(KA's)
- '- ~- .
( , . (-
.s 4 CONTINUED ON NEXT PAGE *****)
1 (***** CATEGORY f
- Pcgo 31
,4 . PROCEDURES - NORMAL2, ABNORMAL 3_EMERQENCY 3NQ,6@Q,1QLQQJQ@Q,,QQNIBQL ANSWER 4.09 (1.50) ,. -/ ' .
QUAD STATION F-10 CFR 20 _(, GE/ i, ~ "d <
~'
~.
O h~ 5
- a. 3000 mrem 1250 mrem -c92- g:/g> p+v jf, ,
- b. 4000 mrem 4000 mrem l Y ,, -
/!*
, ~."
l' f,:
;O
, A) ' l 1250 mrom 1250 mrem ... a.
c. (O.25 each) REFERENCE
'QC QRP 1210-1 294001K103 ..(KA's) I 4.10
(' " " U ANSWER (:ng.50)sc(g ,, a '5 W
.c 9, sn.
- 1. bevacuate the reactor building CO.253 to minimize personnel exposure CO.253 ;
- 2. . verify or initiate the standpy,Sas ,geppment t system and affect '
unit ventilation isolationt CO.253',to prevent the spread and/or release of radionuclides CO.253 ~A Q ' , 1S ff%.i. 6' f R '.lf. V -V t :
- e is~ a *'
p ,. , 7 f; e 'hA T
~
.; q.. , 11ro11mTre7ther u s i i i. " = ventitatieystem C0,253 to
- 3. A verH y v,-
prevent the spread and/or release of radionuclides CO.253 REFERENCE i QC DOA 800-1 ..(KA's) 295023G010 295023A107 295023K301 ANSWER 4.11 (1.00) ' i > a. False
- b. True
- c. False ;
- d. False CO.25 each]
' c.
i b ? l l ' - t '
- - ..~ , . .
{ (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) 1
- _m-m -
Pego 3
.b.__ES9CE99BE),;,NQgd h ,9QNQ$D L ,EDE8@gNCY
*- ONQ,69QJQLQQlgg(,QQNIBQL F
REFERENCE
- a. GC OGP 1-1 Sec C.8 pg 3 j
- b. GC OGP 1-2, pp. 6 i c .~ GC GOP 202-2, precaution #4 202001K602 . . (K A's) I 239001K125 204000 GOO 5 .I 239001K126 i ANSWER 4.12 (1.00)
- 1. Verif y electrical power for bus 13 (23) and bus 14 (24) for the RHR service water pumps.
I 2. Open ice melting line.
- 3. Terminate radwaste discharge.
Conduct an orderly shutdown of Unit 1 and 2. 4. room and crib house.
- 5. Establish communications between the control
- 6. Close the discharge fiume gates to the river.
Cany 4 @ 0.25 each3 Note: Items d and h of the procedureJ not listed since they are'considere-f h fre I/3 cr ef. ,1.-) to be SRO duties. ( gig Ace tt'f @f REFERENCE QC QEP 340-3, Rev 4. 29501BG010 ..(KA's) l
~
.ir:=, y; ~.,r^
(( I: ') g a', ( h " u 4 *****) (***** END OF CATEGORY
**********)
(********** END OF EXAMINATION L" - - - - - - _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _}}