ML20235W242
| ML20235W242 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 10/09/1987 |
| From: | Burdick T, Damon D, Hare S, Nejfelt G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20235W233 | List: |
| References | |
| 50-315-OL-87-02, 50-315-OL-87-2, 50-316-OL-87-02, 50-316-OL-87-2, NUDOCS 8710160093 | |
| Download: ML20235W242 (147) | |
Text
{{#Wiki_filter:- _ _ _ _ _ _ - _ _ _ _ . _ _ - _ _ _ _ . _ e s U.S. NUCLEAR REGULATORY COMMISSION
" l REGION III l l
l Repcrt No. OL 87-02 Docket Nos. 50-315; 50-316 Licenses No. DPR 58; DPR 74 Licensee: Indiana Michigan Power Company Post Office Box 458 Bridgeman, MI 49106 Facility Name: American Electric Power Service Corporation Examination Administered At: American Electric Power Service Corporation Examination Conducted: August 4-7, 1987 Examiners: i i he
*S. M. Hh tb~~~
, /[l-7'87 Date Chief Examiner lob [f 7 D.(J. Damon Date Examiner
'G.
A+)(mk M. Nejf e)t// Date
/0 /or/P7 Examiner V Approved By: M[ / / 7/M T. M. Burdick, Chief Date / / /
Operator Licensing Section Examination Summary Examination administered on August 4-7, 1987 (Report No. 50-315/0L 87-02) Written and/or operating replacement examinations were administered to five senior reactor operators and 11 reactor operators. Results: Three senior operators and 10 reactor operators passed the examinations. 8710160093 871013 PDR ADDCK 05000315 V PDR
L .. j' DETAILS
- 1. Persons Contacted Licensee
*W. G. Smith, Jr., Plant Manager
*L. S. Gibson, Assistant Plant Manager, Technical Support
*W. A. Nichols, Training Manager .
*K. R. Baker, Operations Superintendent
*J. A. Stubblefield, Operation Training Supervisor ;
*W. J. Davidson, Senior Training Instructor R. L. Strasser, Training Instructor '
USNRC
*S. M, Hare, Chief Examiner
*D. J. Damon, Examiner l
*G. M. Nejfelt, Examiner j
*B. L. Jorgenson, Senior Resident Inspector, D.C. Cook
*J. K. Heller, Resident Inspector, D.C. Cook
*K. C. Parkinson, Sonalyst Contract Examiner
*F. M. Victor, Sonalyst Contract Examiner
- Denotes persons attending the exit meeting of August 6, 1987.
- 2. Examination Schedule Due to difficulties encountered with the licensee's requalification training program, the written exam that was originally scheduled for Monday August 3rd, was delayed to Friday August 7th, to coincide with a planned requalification exam. This was done to allow the NRC's participation in the requalification exam, the details of which are contained in NRC Operator Licensing Report No. 315/0L-87-01 and NRC Inspection Report No. 315(316)/87027(DRS).
The oral examinations were performed on August 4th and 5th with the , written examination being administered the 7th.
- 3. Examination Observations During the administration of the operating exams, the examiners observed I both generic strengths and weaknesses on the part of Senior Reactor Operator and Reactor Operator candidates.
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The following candidate strengths were observed:
- The SRO and RO candidates were very familiar with the Technical )
5 specifications and in general could locate all required LCO's and Action Statements.
-+ The SRO and R0 candidates with one exception were familiar with i the emergency procedures and were proficient'in their use during the integrated operations portion of the oral exam. !
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- The candidates knowledge of plant systems'in the auxiliary building was good.
The following candidates weaknesses were observed:
- Candidates in general, did not use procedures to perform normal evolutions such as power changes.
- When procedures were used for power changes, the candidates incorrectly used Procedures 4021.001.003 and .006, entitled " Power Reduction and Power Escalation" when Procedures 4021.011.002 and .003 entitled
" Automatic" and " Manual Reactor Control with Load Variations" should have been used.
- An Additional weakness was identified with the candidates knowledge of the Westinghouse Radiation monitors, specifically with regard to what indications to expect while using their function selector .
switches.
- 4. Exit Meeting The examiners met with licensee representatives (denoted in Paragraph 1) on-August 6, 1986, at the conclusion of the oral examinations prior to the written scheduled examinations. The examiners observations i regarding candidate performance during the oral examinations as denoted in Paragraph 3 were summarized. The licensee representatives acknowledged this information. <
i The logistics of the Senior and Reactor Operator replacement and i requalification written examinations scheduled for August 7, 1906, j were also discussed. The licensee's comments on these written exams were provided to the examiners on August 17, 1986, in accordance with agreements made between the Chief Examiner and the Operations Training j Supervisor, after the written examinations were administered. An j evaluation of the facility comments is included as Attachment 1 to this report. I l ! i l l ) i 1 1 ) l 1 l l 3 l 1
Attrchment#1
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SR0 EXAM COMMENTS I COMMENT 5.01
- b. ~This question is a good example of applying theoretical knowledge to actual (potential) plant transients. It should be noted that additional variables are added when this is done. In this case,.
the knowledge being tested is "How does Beta Bar Effective change from BOL to E0L, and how does that affect the responsiveness of the reactor." This question, however, addresses a transient condition described in the FSAR which adjusted many variables (such as rcd worth) from BOL to E0L and we request that the candidates not be
. penalized for justifying their answers using these variables.
REFERENCE FSAR 14.2.6, Appendix 14.c (attached). NRC Response: Concur. Candidates will not be penalized for a different justification as long as the justification can be supported by the FSAR. COMMENT 5.02 (a) (3) The generally accepted reason for control rod worth changing with boron concentration is " competition." We request that a discassion of competition be accepted for full credit. REFERENCE Attached Westinghouse internal letter on "Effect on Control Rod Worth of Changes in Soluble Boron, Xenon, and Power Level." NRC Response: Concur. Answer key modify. COMMENT 5.04 Rated thermal power for Unit 1 is 3250 MW and the conversion factor given on Page 2 of the Data Sheet is 1 kW = 3413 BTU /hr. We request the answer key be corrected as follows. Line 1: 5% decay heat = (.05) 3250 MW = 162.5 MW Line 6: 162.5 MW + 15 MW = 177.5 MW (3413 BTU /hr/KW) (1000 KW/MW)
= 6.06E8 BTU /hr i (1.01E7 BTU / min.) (
Line 7: 3.10E7/6.06E8 = .0051 hr ( .0001 hr for rounding) or 3.10E7/1.01E7 = 3.07 min. ( .1 min for rounding) , REFERENCE Unit 1 Technical Specification 1.3. I NRC Response: Concur. Answer key modified. COMMENT 5.05 The question as stated, "What happens to the margin to criticality?", does not elicit a quantity or factor by which the margin changes. We request that full credit be assigned for stating that the margin
" decreases."
NRC Response: Partially concur. Because the requested modification does not adequately test the concept referenced by the cited K/A's, the question is deleted. COMMENT 5.06 We feel that this question could be misinterpreted due to the fact that early in core life, the rate of change of critical boron concentration (decrease) is much larger than at my other time in core life due to fission product poison. buildup. Ti is rapid decrease precedes the plateau which the stated answer is apparentiy referring to. Please refer to the stated reference (attached), particularly Figure SNP-RF-14: Critical Baron Concentration Curve (Revision 2). We request that you also accept, for full credit, a discussion of wny 1 the critical boron concentration decreases at a much more rapid rate l at 00L than it does at E0L, specifically, at BOL, fission product poisons i are building up, requiring more rapid boron reduction to offset the buildup 1 of poisons and to offset fuel depletion. At E0L, boron reduction is only l required to offset fuel depletion. NRC Response: It is agreed that the question could be confusing if a candidate did not adequately understand the variables that effect critical boron concentration. If a candidate was confused by the question, he was instructed to question a proctor. No such questions arose during the exam. The rapid decrease in boron concentration referred to in the facility comment occurs extremely early in core life. When this amount of total core life is compared to the rest of core life, it becomes microscopic in detail . The question implies L. that this effect is not the one to be discussed. l l I l 2 t__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _
For these reasons, the answer key remains unchanged. No credit will be deducted for a discussion of the rapid initial decrease, as long as the required answer as stated in the key is included. COMMENT. 5.08 1 The question asks the candidate to "Show all work, including any applicable j formulas." The answer requires multiple written variations of the same formula [Q = UA(Tave-Tstm)] for full credit. We request that full credit be given for mental manipulation of the applicable formula as can be demonstrated by substitution of numerical values into the modified formula, i.e., A(1)[567-515] = 0.8a(1)[Tave-515] or an equivalent variation. We also request that the full credit value for the required Tave be accepted if it is within the range of a least 579.5 to 580.5 since the calculation involves steam table interpolation. NRC Response: Full credit will be given if all work is shown and the work method is obviously correct. The range of answers for Tave is modified as requested. COMMENT 5.09 The answer, as keyed, requires specific terminology to be used. We provide two of the several alternate reasonable explanations that should be acceptable. (1) As fluid flows around the bend, the velocity of the fluid increases in the shorter radius (pressure decreases) and the velocity of the fluid decreases in the larger radius (pressure increases), and thus the flow rate is proportional to the square root of the delta P. (2) As fluid flows through the pipe, it tends to maintain its direction unless acted upon by some force (First law of Motion). This force is exerted on the fluid by the wall of the pipe causing its momentum to change as determined by the radius of the bend. As the force changes, so does the pressure since . Flow rate is then proportionaltothesquarerootoftbe_d[eltaP. ~ REFERENCE Mitigating Core Damage Training Manual (Westinghouse) Chapter 6, Pages 6-23 through 6-25 (attached). Lesson plan R0-C-MC06, Vital Process Instrument Response. NRC Response: The alternate answers provided by the facility are considered l equivalent alternate wordings of the original answer key. As such, they will be accepted for full credit. I 3
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p' -{ . g .. .. !" COMMENT ' 5.'10 : L We-feel that candidate performance'on this question may have been
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'. diminished because of poo'r wording. FQ(Z)'is'not listed in the Definition Section of Tech. Specs. 'Also, if.the candidate'did not h
recognize-'FQ(Z) as the Heat Flux Hot Channel' Factor,'he would have also
~1ost; points'in Part (b). This method of questioning should not be used, as stipulated by.ES-202 Paragraph E.13.
A better way of wording the question would have been:
- a. Define'the term Heat Flux Hot Channel Factor,.FQ(Z)'
per Technical Specifications Basis Section.
, 1 b. List the FOUR items monitored-by the operator to ensure that hot channel factor limits are maintained.
We' request that this be taken into consideration while grading this question and in future' examination questions covering this topic. NRC Response: The wording of Part (a) of the question did not limit x j the range of possible answers-to'a specific area of the- ' Technical Specifications. The original answer is from Technical Specification Bases,=and therefore remains , unchanged. The Paragraph of ES-202 cited by.the facility refers to the use of double jeopardy questions. This question is'not double jeopardy in nature. To correctly answer'Part (b), the candidate had to simply know that FQ(Z) is a hot channel factor of some-kind, and not,the specific definition of FQ(Z) as required for Part a. Therefore, a correct answer for Part (a) is not required to obtain full credit for Part (b) and thus the question and answer remain unchanged. COMMENT 5.11
- a. The Unit of differential rod worth used at D.C. Cook is pcm/ step and we request that common units be used in exam questions. l The keyed answer is incorrect in that a value of .00767 was used in-the calculation for Beta Bar Effective when the question specifies .00596 be used. The correct answer is -0.187 dpm. q 1
Additionally, a candidate may try to account for the power change ~l during the 25 seconds that the rods are in motion (48 steps per minute, 20 steps). Since determining the power drop would require use of equations not provided, the candidate could reasonably assume
. a.6.5E-8 amps be accepted for full credit. ,
[P = (10E-8)(10E( .187)(1.417))] 4 i _ _ _ _ _ _ - _ _ _ _ _ _ - - _ _ _ _ - _ _ _ _ _ _ _ l
L NRC Response: While is reasonable to expect that candidates be able to perform simple math manipulations to place data given in a question into , units commonly used at the facility we concur that it is not a ! required knowledge. The section requiring manipulation of the value of differential rod worth remains unchanged for the purposes of this exam as it has no effect on the pass / fail rate of the candidates. This will not be a requirement for credit 'j on future exams. The answer key is modified to use the correct value of Beta Bar Effective. The facility is correct regarding the use of equations not
.provided for Part (b). However, it is not valid to assume a constant startup rate over the entire time span stated in the question. Based on the fact that not enough information was provided to correctly answer this question, Part (b) is deleted, and the point value is adjusted accordingly.
COMMENT 5.12 This question is subject to interpretation as to which charging flow rate is to be evaluated and the keyed answer states that you will accept other answers if supported by viable assumptions. Question 5.12, Attachment 1 identifies other possible answers and their calculations. We request that any of the following answers be accepted for full credit. NRC Response: As stated in the original answer, other methods will be accepted if supported by viable assumptions. The facility comment simply restates this. As such, the answer key remains unchanged. COMMENT 5.13 The introductory sentence of the question is a paraphrase of a sentence from T.S. Basis which states " SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RSC boron concentration and RCS Tavg." T.S. then goes on to say that the most restrictive condition is the steam line break at EOL and a minimum of 1.6% K/K is required to control this reactivity transient. The required SHUTDOWN MARGIN of 1.6% K/K is required at BOL as well as EOL and does not change over core life.
% better way to word the question may be:
"In determining the SHUTDOWN MARGIN by Tech. Specs, B0L and EOL conditions were analyzed. Which time in core life was more restrictive? Why?"
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This question requires the candidate to explain why a T.S. LC0 changes over. core life, when it doesn't change. This type of question should be' avoided as per Paragraph 4.5.1(5) of NUREG 1121 (DRAFT 9/85) and we
. request that it be deleted.
NRC Response: The clarification made during the course of the exam made it clear that the LC0 does not change over core life, Lut that the analysis did change over core life. It was also made clear that a discussion of the LC0 was not desired, and that a discussion of the Technical Specification Basis analysis was required. The question and answer remain unchanged. COMMENT 6.02
- c. Shutdown bank rod speed is "62" steps per minute. (From Page 11 of 24 of the referenced document).
NRC Response: Answer key will be modified. COMMENT. 6.03
- a. The following additional Unit 2 main feedwater pump trips are supplied in addition to those keyed. Unit 2 has two different diverse Lubricating 011 trips. Refer to attached functional ,
diagram (Dwg. 2-5627).
- 10. L.0. Pressure < 60% Normal
- 11. L.0. Pressure < 40% Normal
- b. The question is general in nature but appears to assume in its answer that all steps have been completed in Procedure CHP-4021.055.003 (Placing MFP and MFP Turbine in Service) up to Step 6.4 (Prepare the FPT for Service). If the SR0 does not make this assumption, he is liable to supply other additional correct answers found in previous steps in the procedure, particularly from Step 4.9 of the PRECAUTIONS which provide conditions that must be satisfied to enable turbine startup. (Resetting reactor trip breakers, emergency trip device, operating device, etc.) The procedure that governs the answer is much more in detail than the lesson plan used to develop the answer key. We request that any correct actions provided by the SR0s that answer the question and come from OHP 4021.055.003 should be allowed for full credit.
NRC Response: a. Answer key will be modified to accept Unit 2 FW Pump Trips. i
- b. Do not concur. Step 4.9.1 of Procedure OHP 4021.055.003 specifically delineates the three actions which must be taken to reset the turbine trips.
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COMMENT 6.04-L c. The erroneous startup rate'('SUR) listed as one of two problems in. L the answer key is an extremely insignificant problem when compared to the "two"' problems in answer,Part c.2 (Limited overlaps - inability-
.to block SR level trip). The error introduced by the overcompensation of:IR detectors will~ be below their " idling' current";for the majority of the startup;and will not be sensed by the SUR circuit. If no overlap- orivery little overlap is seen during the startup, taking.
K into account the low startup rates used by operators-(0.1 to 04 3 dpm , typically near criticality), the SUR' error will not be observable by using the meters in the control room.- In addition,.since the difference'in the-'! rate of change" in level of'a properly compensated channel.and an overcompensated channel is very small, the observable
- 'SUR would p.robably never be_ determined on control room meters.
Furthermore, if the SR trip could not be a blocked, IR range SUR
-would never indicate as it would be below the sensitivity of-IR level output due to '! idling current" of the IR.
.We request that these factors be taken into account during grading-of this question'and further recommend that the answer key be modified as follows:
- c. (Any two @ 0.375 each)
'1. Limited SR to IR overlap. '
- 2. Inability to block SR high level trip.
- 3. Possible erroneous IR SUR indication.
NRC_ Response: The answer key will be modified to not require " erroneous start-up rate" as an answer to the question. The point value of the question will.be reduced and will be; revised to accept any two of the above answers at .25'each. COMMENT- 6.05
- a. One purpose of the Containment Hydrogen Recirculation /" Hydrogen Skimmer" (CEQ) fans is to draw air from the upper containment and distribute this air to the vicinity of the hydrogen recombiners.
We request that the CEQ fans be an alternate answer to receive full credit for this question,
- b. Please note that the " Temperature Set Controller" is never used to adjust temperature (see referenced lesson plan). In all plant Surveillance and Abnormal Operating Procedure (OHP 4030.STP.013, OHP 4022.034.004 and THP 4030.STP.206 - Performance section STP) only the " Power Adjust Potentiometer" is used to adjust heater temperature.
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l l NRC Response: a. Answer key will be modified to also accept CEQ Fans,
- b. Answer key will be modified to reflect facility comment.
The facility learning objectives should be modified to better reflect actual plant use of the Temperature I Controller. l COMMENT 6.06 ) a. The question does not solicit the answer in the detail required by the answer key for full credit. A sufficient answer should be
" Generated by a reactor trip" or " Reactor trip breekers open." In order to obtain the keyed answer, two suggestions are provided for future question modification at this facility.
(1) How is Reactor Trip Permissive (P-4) generated? (BE SPECIFIC) (2) How is Reactor Trip Permissive (P-4) generated? (Included any logic or coincidence.)
- b. We request that answer key part (2), (3) and (4) have the following placed ir, parentheses and not be required for full credit.
(1) Feeds feedwater isolation (coincident w/ Low-Tavg) (2) Blocks closed FRVs (when closed by SI or Hi-Hi S/G 1evel) (3) Feeds SI (block / reset) logic NRC Response: a. Do not concur. Per Functional Diagram 98512-5, P-4 is generated only when a reactor trip breaker and its bypass breaker are open. Partial credit was already allowed for the answer " Reactor Trip."
- b. Concur. Answer key will be modified.
COMMENT 6.08 This question may be interpreted in two ways by the SR0 candidate. The first way to understand the question is to ask "What is the basis for 1) Normal pressurizer spray flowrate (* 800 gpm) and 2) pressurizer spray valve bypass flowrate (* 2gpm total). The second interpretation of this question would be that it asks "What is the basis for the minimum and maximum pressurizer spray flowrate (800 gpm nominal)?" The FSAR provides that the spray flow must be at least enough (minimum) to mitigate the step power reduction of 10% that is keyed in the answer. No basis is provided for the maximum flowrate but it is 8
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ob'vious that it 'would be limited to prevent extreme pressure reductions ! when the spray valve opens (prevent reactor trips and SI's on low- l pressurizer pressure). We request that the answer to.this_ question I be modified to. include answers for both interpretations of this question'. l
.The applicable FSAR pages are included and' highlighted for your - !
convenience. l Also the question does not specify_the number of response required for ! full credit concerning the basis for minimum spray flow in the first ' interpretation of the question. The examinees may assume that only one answer is required. We request that the answer key be modified to allow full credit (1.0 points) for listing one of the three basis statements j for minimum spray flow. 1 Note: Question 2.04 of the R0 licensing examination is very similar to this question in that it asks _for TWO reasons for pressurizer spray I bypass flow. That question does specify the number of responses i required. I 1' REFERENCE DCC FSAR, Pages 4.2-6 and 7 (attached). NRC Response: Partially concur. The answer key will be revised to accept the suggested alternate answers in the facility comment. Credit will be awarded based upon the assumptico the candidate 3 makes when answering the question. No extra credit wiii be awarded if the candidate supplies only one of the " minimum bases" in reference to the snray valve bypass flow of 2 gpm. COMMENT 6.09 I The answer key for this question specifically mentions " positioning air will be vented" which causes steam dump valve to clone. The question did not solicit this type of detail concerning steam dump valve operation. We request that full credit be given for statements similar to "Vhen ; temperature reaches 541 F (or P-12), all steam dump will close." ; NRC Response: No credit will be taken away for an answer that does not , specifically refer to positioning air.being vented. However, l for full credit the candidate must also indicate that if the f temperature increases above 541 F the steam dump valves will again open. l t 1 I 9
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[. 'g COMMENT l 6;10 - Because the: circumstances' surrounding'this' event are not.specifically-
. addressed'by the_ question,' examinees are not supplied enough information- J L , ;
. 'to determine which Technical / Specification' are complied with and which 4 were not. . We request that the answer key be. modified to' accept for full (
credit.similar wording to the following: q
~
"This constituted'an;unmonitored release' path which i.s.in..
( violation of Technical Specifications." NRC Response: Concur. Answer key will be modified, f COMMENT.- .'6.11
- a. Recent, labeling c d
- the' designation'o the 'nges made toposition UNBLOCKED Unit 2 to control room panel changed COLD OVERPRESSURE.
These-are scheduled for Unit I control room in the near future. > We request that either UNBLOCKED position or COLD.0 OVERPRESSURE-position be acct >pted for full' credit.
- c. We also request that the following tolerances be' accepted for the Unit 1 and 2 setpoints.
U1 385 psig (+/- 22 psig) - U2 420 psig (+/- 15 psig)
.NRC Response: Concur. Answer key is modified.
COMMENT- 6.12 Because the initial conditions specify the 1E ESW Pump is in lockout, omission of automatic actions related to this pump from answers provided by examinees indicates that they know the pump will not start. We-request that full credit be given for answers that-only identify actual automatic actions which occur as a result of an SI under these conditions. Because the question relates to a Unit 1 SI. signal, examinees may not specifically mention start signal to Unit.2 ESW pumps as implied by the "All" statement in the answer key. We request that no points be lost by examinees who listed only U1 ESW system responses. Other actions will occur that are indirectly related to those signals generated by the . equipment response to the SI signal, i.e., opening of the ESW Pp discharge i valve on a pump start, opening of the Emergency D/G'ESW supply valve in response to a D/G running signal. We request that points not be taken 1 off for listing these additional automatic actions. i 10
-NRC Response: An acceptable' answer to the question would be Units.1 and'2
-West and Unit 2 East ESW Pumps Start. Points will not be taken off.for not specifically stating that Unit 1 East Pump will not start, nor will points be taken off for listing additional automatic actions.
COMMENT- 6.14 The questions and answers, as written, require the examinees to make assumptions on the positions of the west ring header iso'iation valves RPV-11 and PRV-21. It appears that the question writer assumed the west ring header valves to be closed in Part (a) and open in Parts (b), (c), and (d). Without knowledge of the question writer's assumption, examinees could not answer all parts of this question correctly. The one exception to this would be Part (d) which would be FALSE. We request
.that Parts (a), (b), and (c) be deleted from the exam and a reassignment of point valve be made making Part (d) worth 1 point rather than 0.5.
This reassignment of point values is to avoid increasing the effective point values of the remaining questions in this exam category. REFERENCE OP-12-5120-6 Flow Diagram Compressed-Air System (Key Plan) (attached). NRC Response: After reviewing the supplied drawing, the answer for Part (a) of-the question has been changed to false. No part of this question will be deleted as this question was very explicit in that it specified the position of only PRV-20 and PRV-10 and no other valves. COMMENT 6.15 Part (c) of this question asked the reason for having a maximum power level above which lower ice condenser access is forbidden, The question does not ask for any specific type of radiation hazard. The answer key < contains " Neutron dose consideration." While neutron dose is of concern, the reason for having limited access is to avoid excessive radiation exposure. This includes neutron and gamma dose. We request the answer key be modified to allow full credit for any of the following: I Neutron Dose l Gamma Dose I Gamma Does for N M gamma l ALARA considerations i High Radiation Levels ] (Or similar wording) j l 1 I 11 l o
NRC Response: The answer key wn1 be modified.to accept "High Radiation Levels" or equivalent wording. In additici. the answer key will be revised to.also accept a power level of 10% for the first'part of (c) as was discussed between the chief examiner and your training staff. COMMENT '7.03 We request that "2 out of 4 power range channels > 10% power to satisfy P-10" also be accepted for full credit. NRC Response: Answer key modified.' COMMENT 7.04 For Parts (a), (b), and (c) inclusive, we request that an indication as l to whether the specified permissive allows a manual block of a trip or instates an automatic block of a trip should not be required for full credit. The question wording requires the students to list (state) all defeated trips (both manual and automatic), but does not request that the student enumerate which trips are blocked by which method.
- a. For Part (a), we request that you accept for full credit " Source Range Reactor Trip."
- b. For Part (b), we request that you also accept for full credit, wording similar to that employed in R0-C-NSil-SH03 (appended as-a reference to 7.03) and restated as follows:
(1) All low flow trips
- low RCS loop flow
- RCP breaker position trip
- RCP bus undervoltage (UV)
- RCP bus underfrequency (UF)
(2) turbine trip (3) pressurizer low pressure (4) pressurizer high level
- c. For Part (c), we request that you accept for full credit wording similar to that employed in R0-C-NS11-SH03 (appended) and restated below as follows: i P-8 blocks:
(1) low RCS loop flow (1/4) loops trip (2) RCP breaker position trip from 1/4 breakers j i 1 12 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ]
- r. ,
I a 1 i NRC Response: Answer key modified. COMMENT 7.05-
- a. No comment, ,
1
- b. .We request that you additionally consider as probable causes the following list:
(1) High lake (or high Non-Essential Service Water - NESW) temperature as an individual cause. 1 (2) High ambient temperature as an individual cause. (3) Insufficient-number of ventilation units in service ! (Reference-4021.028.001 " Upper Compartment Ventilation" k Step. 6.1.1. ) NOTE: A fourth fan may be started as r to maintain containment temperature. necessa'y (4) Upper containment ventilation' heating on. Refer to CHP 4021.028.001 6.1.2 Upper containment Ventilation heating-should be initiated if. upper containment temperature < 75 F (therefore, heaters may have been left on inadvertently) and 6.1.3 Place Upper Containment Ventilation heaters in AUTO (a malfunction of heater control thermostat may have caused the heaters to energize at a higher setpoint). (5) Upper compartment supplied with purge and heating steam valved in to purge supply units. NRC Response: The first three of the five comments simply reiterate the answers in the answer key. Comments four and five will be incorporated into the answer Key, COMMENT 7.06 We request that the allowable range for a full credit answer be extended to .6 - 1.0 Rad /hr for the .7 + .1 Rad /hr range allowed by the answer key. The actual scale width spanned by the .6 - 1.0 range is 2.5/32 inch (* 2 mm). The intersection of the 610 meter dose rate scale is the result of the third line drawn on the nomograph. Several five (5) attempts by the reviewer to ascertain replication of the keyed answer resulted in the following data: Three results - * .8 R/hr (Note: The keyed answer is lower as the point of origin for !ine 2 is slightly to the right of scale 3). 1 result - slightly less than .7 R/hr 1 result - 1.0 R/hr 13
l These.results were obtained using conditions similar to those which. the students would have been required to employ during the exam. (i .e, the use of the cardboard backing from a pad of writing paper as a straight-edge since no straight-edges were provided and a slightly dulled pencil point since this was the third sequential section of the exam.) Using
.8 R/hr as the midpoint (the most frequently obtained result), a variation of one scale graduation on either side results in the
.6 - 1.0 range. An answer on either end of this range represents a cumulative error of 1 mm from the midpoint. As Scale No. 7 represents the third sequential scale intersection determined in this problem, an error of .33 mm for each scale intersection would accumulate to a total error of 1 mm for the final reading.
The requested range does not impact the classification of the event. PMP 2080.EPP.101, " Emergency Classification" ECC 19 " Radiation Releases" requires that measured or projected dose rates of 250 mR/hr whole body at the site boundary be classified as a General Emergency. Thus, for all results within this' .6 - 1.00 Rad /hr range, a Protective. Action Recommendation (PAR) would be required to be developed based on the duration of the release and the resulting projected dose. If the 8-hour 4.8 Rem Whole Body (.6 x 8) to 8.0 Rem Whole Body (1.0 x 8). Projected dose rates above 5 Rems Whole Body require a precautionary evacuation for a two mile radius and sheltering five miles downwind. A projected dose rate just 5 Rem Whole Body requires sheltering in a two mile radius and five miles- downwind. However, the result obtained form the given nomograph does not constitute the 610 meter dose rate as, at a minimum, the 610 meter dose rates from Unit 2's plant vent, both Unit l's and Unit 2's steam jet air exhaust and both Unit l's and Unit 2's gland steam exhaust must be added to obtain the actual site boundary dose rate. Therefore, even at the low end of the range (.6 Rad /hr), once the other release points are added, and, taking into account the points listed on Page 1 of the protective action recommendation flowchart, the PAR would be appropriate to protect the public. NRC Response: The problem was designed to determine the methodology of calculating the initial dose assessment and was not wholly dependent upon an exact numerical solution for credit. The answer key was modified to accept 0.8 +/-02 R/hr. COMMENT 7.08 We request that (in addition to the keyed response) full credit be allowed for any of the following listed responses: Piping or equipment damage. Thermal stressing of piping. ; l 14
1 l j The potential problems resulting from a steam trap remaining isolated for 1 an excessive period of time can be divided into two areas of discussion: ) (1)- The problems which would ensue if this trap is unisolated too quickly. ] l (2) The problems which would ensue if this trap remained isolated l indefinitely with no: provisions made to bypass it. The answer key reference from the body of the procedure (Section 4.1.1.8) lists an immediate event (water hemmer) which would ensue in Case 1 above. The consequences of the water hammer are the problems which are to be I avoided. These consequences are downstream piping or component damage and/or failure resulting from the mechanical shock created by the water hammer. Furthermore, the piping immediately upstream of the trap which has collected the relatively cooler water would be subjected to thermal stress when it is purged by the flow of warmer steam when trap is quickly unisolated. Case 2 above is also addressed by this precaution which requires logging of isolated equipment on Table No. 1 of CHP 4021.001.001. This table is entitled " Equipment Out Of Service Due To Heatup." As initial condition (3.1) of this procedure requires completion of Data /Signoff Sheet 5.1. Line Item No. 14 on Data /Signoff Sheet 5.1 requires that the Main Steam System be lined up in accordance with OHP 4021.051.001 " Placing the Steam Generators and Main Steam System In Service." The key point states that this lineup must be completed prior to heating and pressurizing this system. Since this lineup requires steam traps to be unisolated, Table No. I serves as a record of which traps must be returned to service or manually bypassed in order to prevent water backup into main steam lines with resulting erosion problems. NRC Response: Alternate answers are acceptable. COMMENT 7.10 An alternative full credit answer is contained in a caution proceeding OHP 4021.FR-H.1, Response to Loss of Secondary Heat Sink. An entry condition is "from E-0, Step 8, when minimum Auxiliary Feedwater flow is not verified." The caution states that RCPs should be tripped (and bleed and feed immediately initiated) if: ,
- Any 3 S.G.'s < 8% W.R.
OR
- Pressurizer Pressure > 2335 psig due to loss of sec.ondary heat sink.
I l- 15 I l-
4 x .. NRC Response: These alternate answers are acceptsble provided the candidate- /- , specifies that'he is in the Loss.of Secondary Heat: Sink
, procedure.
COMMENTJ 7.13'
- a. We regimst that the answer key ~ forLitem a be modified to ' accept either.of the following titles'as an acceptable response:
Shift' Supervisor. Site Emergency Coordinator. and that the keyed answer of the Assistant Shift Supervisor be. deleted.
. Reference PMP 2080.EPP.112 Sections 4.1.2, 4.1.3, 4.4.5,-Exhibit A and Exhibit B.to substantiate modification; Sections'4.1.4 and 4.5 to' remove' Assistant Shift Supervisor as a. correct response.
-(Thej referenced Section of SR-C-EP09;' Item X, Page 10 of 10 does not contain information pertinent to any of the'above questions.')
NRC Response: -Answer key has:been modified to reflect facility comment. COMMENTL 7.14 The first sentence in the question states that pressurizer-pressure control-is'in MANUAL'. In this condition, there is no automatic control, as the controlling channel is effectively removed from the circuit for
. pressurizer control components. In the second sentence, the question-
' asks;for response to a failed controlling channel. The candidates may
<- interpret this question'in'either of two ways. If he assumes that the question is checking his ability to recognize no automatic action is required to be tended.to, he will answer accordingly. If he' assumes that the second sentence stands alone, he should answer according to.
the. key. .In. essence, there are two opposing questions contained in this one question. Therefore, we are requesting that, in addition to the keyed answer, the following alternative correct responses be accepted for full credit: !
- None. (The procedure does not apply to manual control as there are no inputs from the failed channel to the control system).
*' Trip bistables from failed channel. (This action is not required by the referenced procedure which only addresses a failed controlling channel, but would be required by the Tripping of Protection Set Bistables Procedure and by Technical Specifications).
NRC Response: Because the requested modification does not adequately test the I concept referenced by the cited K/A's the question is deleted. L 1 16 l
COMMENT 7.16 a, The question, as stated, asks why adverse conditions require the use of alternate setpoints to implement the ERGS. The question does not request a discussion of what those adverse conditions are. We request the following alternative correct response be awarded full credit (in addition to the keyed response).
" Instrument errors may become excessive due to radiation or temperature concerns."
- b. The wording in Question 7.16(b) implies that, since adverse conditions were declared due to radiation, containment pressure is not at its adverse setpoint. Therefore, we request that the answer key be modified so that " containment pressure less than 1.1 psig' not be required for full credit.
In addition, we request that the answer key be further modified to allocate full credit for either of the following correct response: ,
- For adverse conditions declared due to the Containment dose rate exceeding 10 5 R/hr, the plant engineering staff (Plant EvaluationTeam.-PET)mustassesstheintegrateddoserate.
If the integrated dose rate is less than 10 Rads, then normal instrument values may be used. (It is not necessary to have a dose rate < 105 R/hr to reinstate normal values), j
- If the integrated dose setpoint has been reached, then adverse conditions are used until the instrument is replaced.
NRC Response: a. Do not concur. The proposed alternate answer is simply a statement of why alternate setpoints are used. The question specifically asked for a brief explanation as to why alternate instrument setpoints are used. This request for ! an explanation should have indicated to the candidates that l a more in-depth answer was desired. ;
- b. Do not concur. Because adverse conditions were declared due to radiation and not containment pressure does not imply that the containment pressure did not exceed the limit at a later time. When exiting adverse containment conditions, all criteria must be evaluated prior to returning to the " Normal" instrument setpoints.
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COMMENT 8.04: The preface to these individual questions is incorrect. Controlled-
- g / leakage as defined in the D.C.' Cook T.S. is:. " Controlled leakage-
- shall; be that' seal' water flow supplied M the reactor coolant pump.
seals." If this: definition was' applied by the examinee, then the answer to all questions listed would be-Not Applicable (N/A) or False as the question
. stated that-controlled leakage is from the.RCP Seals.
_If trie ~ examinee assumed the preface was correct, then the: answer'to (c)
'< could be TRUE because this is one method for determining seal leakoff. 'l' It.should be noted that actual T.S. does not' stipulate what instruments should beiused.
Answer "e" is' FALSE. This facility;is obligated to' follow more stringent , requirements other than Technical Specifications unless we receive. specific-exemption by the.USNRC. _D.C'. Cook has received citations in the past for-
;taking the action; suggested by the answer key. Time does not permit gathering documentation to support this statement but Region III y Inspection and Enforcement personnel should be able to support our !
. stance .if asked.
'In short,'the. preface to the-question is technically incorrect which
.can-lead:to' confusion in answering the. individual questions. Therefore,
.we request that' answers to Questions (a), (c), and (d) should be allowed- 0
- to be answered as eitherlTrue or False. l Partially concur. The interchange of the prepositions "to" NRC Response:
and :"from" for describing the association between " controlled leakage"'to the RCP seals did not result in a fundamental problem with this question. During t'ne examination, no questions were j asked regarding this question, which indicated no confusion about i the' question intent. However, if the candidate states in his answer that the-question was incorrect in it's use of " controlled leakage"- and answers the questions based on that assumption, credit will be awarded. t i Regarding Part (c) of the question, it is being deleted from the exam because change in VCT level, not flow recorders, k are used in the surveillance procedure to determine controlled leakage. Part (e) is True, because a less restrictive code (e.g., ASME) ! cannot supersede a T.S. Surveillance requirement. If a more restrictive code standard is used it can only supplement a i p binding legal T.S. Surveillance requirement. l 18 !
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COMMENT: =8.05 The question,7as stated,:does not elicit the'T5ch.- Spec.~ bases as the '
; only answer for the. concern over an ' anticipated' ice bed loss. R0-C-NS14,
'Page 4, states'three specific: purposes for the ice; condenser and the loss of' ice from the bed would result in a degradation of these purposes as d well. Therefore, it is requested that the answer.be expanded to accept j the following additional responses:- <
- a. ensure post. LOCA containment pressure.is maint'ained below design. ;
b.- reduces the~ ability of the sodium tetraborate to remove iodine.
-(Containment Sump pH control)
- c. reduces the heat sink available during and after the LOCA. q Reduces on 1 absorption after LOCA. 1 l
(or similar wording) NRC Response: -Equivalent wording of an answer is always considered. Alternate answers a. and b. are also accepted for credit. q Alternate answer (c) simply reiterates'an'swers already in ' answer key.- Answer key be will be modified. COMMENT '8.09
- )
- c. Master' Surveillant.e Schedule lists the frequency as daily. Since 1 Part (c) of question did not specifically ask for a T.S. frequency, -.
we request that either "72 hours" or " daily" be acceptable as correct answers. l NRC Response: Answer key modified to also accept daily. COMMENT 8.10
- b. We request that " Plant Manager" be accepted for full credit and place c "or his designee" in parentheses since the responsibility rests with the Plant Manager even though he may designate someone else. We also .
' request that " Assistant Plant Manager," " Department Head Requesting Deviation," and " Senior Licensed Reactor Operator on Shift" be accepted as parenthesize answers, but not be required for full- .i credit as delineated in PMI-4010, Revision 4., Attachment 1. j NRC Response: Part (b). . Answer key, as written, would provide full credit with the " Plant Manager" provided as the answer. Also, the answer key, as written, would provide full credit for a designee (i.e. , Assistant plant Manager, Department Head .
Requesting Deviation, or Senior Licensed Reactor Operator ! on Shift) with the stipulation that the Plant Manager was absent. 19
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1 . . f
- COMMEtiT~ 8.03
' ~
.The cited bases are different between the two units. Unit 1 basis does not' j refer to multiple rod. drop scenarios ~specifically and Unit-2 states that lj ia single dropped rod'with auto rod control..could cause local.DNBR limits
~ !
ito be exceeded. Therefore, it-is requested that'the word " multiple" be j dropped from'the answer key and " rod drop' accidents" alone should be 6 acceptable for full' credit. !
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.l
'NRC' Response: Answer key modified, l
. COMMENT. 8.12 'i
- c. Per PMI-4010,' the access authorization to the upper and lower ' volumes of containment are the same for operating Modes 3, 4, and 5. The question could have'been written such that it is always false if_ !
the reference to. Modes 1 and 2 was added: 1 "The level of-authorization needed to obtain access to j containment upper volume while in Mode 1 and 2 is the'
~
same'as the level of authorization required to-enter the containment lower volume while in Modes 1 and 2." REFERENCE a i PMI-4010,,Section 3.1'3.1, Revision 4 l 1 NRC Response: Concur. - Since Part (c) can be answered either true or false,- this portion of the question does not adequately test the f concept referenced by the cited K/A's and is therefore deleted from the. exam.
-j COMMENT '8.14
- d. We request-that an answer of."TRUE" be' accepted for fully credit q if justified by stating that a striped tag clearance may be issued, i to the Shift Supervisor (SS).
REFERENCE PMI-2110, Section 3.2.2.2 NRC Response: Additional correspondence between the facility and the chief i examiner revealed that the questinn as stated could be answered i either true or false depending upon the candidate's interpretation i as to'the question intent. Since Part (d) can be answered by / either a'true or false, this portion of the question does not adequately test the concept referenced by the cited K/A's and , is therefore deleted from the exam. I a
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COMMENT- 8.15 We request that "SRO-CA Required" by accepted for full credit. The meaning .is identical to the answer given (1.e., Senior Reactor Operator in charge of CORE ALTERATIONS). NRC Response: Equivalent wording of an answer is always considered. Answer key clarified to also accept SRO-CA. 4 1
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r i i i l l I l 21
V DAttachment#1I
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. > R0 EXAM' COMMENTS
' COMMENT' 1.03
- c. .The keyed answer.contains two. equally correct brief'. explanations
- why SCM'is less. under natural circulation conditions. .However, y
- the-phrase' "will approach fu1111oad delta 'T" is.lnot elicited by ' i the question since no indication.is given as.to how long the' reactor
; 'has been shutdown. We.. request that " Core delta T during natural.
circulation is. larger" or "Thot is greater" be' accepted-for full
. credit.
REFERENCE-
. Thermal-Hydraulic. Principles and Applications to PWR II',1 Chapter 14.
'NRC Response: - Comment' accepted. Answer' key is expanded to include these
- alternate answers.
COMMENT- 1.04
- a. The keyed answer states that "A will-reach critical red height sooner than B" which.will: allow less time for subcritical multiplication to .
-take place. The.second part of the' keyed answer which refers to the '
time to reach stable neutron counts is greater as Keff gets' closer:to one does not really apply in this case. Since the reactors are being taken critical'on continuous rod' withdrawals, no stable neutron levels
-are achieved. In fact, the higher the reactivity addition rate, the l higher...the transient SUR. We request that "and the closer Keff is to one, the longer..it takes to reach a new steady state neutron level (more neutron. generations are required to achieve the larger ;
population)" not be required for full credit. NRC Response: Concur. Answer for Part a is modified as requested and point ; value appiocriately adjusted. However, the. facility is cautioned to include references to justify specific comments in accordance with ES-201, Attachment 1, Enclosure 3, Paragraph 3, Note 1. COMMENT 2.01 We request that Item No. 3 of the answer key be modified to read
.(QRV-111 OR 112). Either of these valves closed will cause QRV-160, 161 and 162 to close.
REFERENCE Print OP-98274-6 (attached). NRC Response: Comment accepted. The facility is requested to change the misleading section of the R0 Lesson plans submitted to the NRC prior to-the administration of the next NRC exam.
L l COMMENT 2.02 We request that Item a of the answer key be modified to allow full credit'for answers anywhere within a band of pressures rather than-
-the exact setpoint as follows: . . . less than 353'psig to 397 psig for Unit 1 .', . less than 410 psig to 440 psig for Unit 2.
Actual setpoint for Unit 1 is 375 1 22, Unit 2 setpoint is-425 1 15. I
~. REFERENCE i 1-0HP-4021.017.002, Placing In Service the RHR System, Page 3, Item 4.3 for Unit I and 2-)HP.-4021.017.002, Page 3, Item 4.3 for Unit 2.
NRC Response: Comment accepted. Answer key modified for Part (a) so that pressure setpoints are 375 1 22 psig for Unit 1 and 425 i 15 for Unit 2. The facility is requested to correct the lesson plans submitted to the NRC to make the pressure setpoints consistent throughout. COMMENT 2.07 We request that Item 1 of the answer key be modified to allow full credit as stated or "Either unit vent radiation monitor (VRS-1500 or VRS-2500) not' alarming (high)."- Rad monitors numbers not required. REFERENCE Print OP-98311 (attached). NRC Response: Comment not accepted. The reference submitted with the comment by the facility clearly shows that both unit. vent radiation monitors must not be alarming as a prerequisite to opening RRV-306. Therefore, the answer key is modified so that Part 1 reads "Both unit vent radiation monitors must not be alarming." COMMENT 2.13 Because the wording in the referenced lesson plan is somewhat misleading when taken out of context, we request that the answer key be modified to more accurately define conditions that will trip a Diesel Generator Output breaker. Modify the answer key to accept any three of the following at 0.5 points each:
- 1) Emergency Trip Pushbutton (Control Room or Local)
- 2) Operation of the Diesel Generator HEA 1 i
1 2 ) - _ _ - - - _ _ . . n
lCy 2 . , l E m3 ,, , ,
> \
h n , z. : , - L g.- 'c.- 4 3) "lD/G con'trolfswitch to the'Siue position. b I4) Safety Injection Signal-(if no blackbut signal'is present).
, 5);:. -Transformer:11C(21C) or 11B(218) _HEA (for T11C' and T118 1, N ' ifeed'only)
' 6)* D/G breaker overcurrent Similar wording. acceptable. ..
Information'in parenthesis is not required for full _ credit, r NRC; Response: Emergency Trip Pushbutton will be included'in the list of correc't-a . answers. For the. balance of_the answer, the reference supplied [I by the' facility does.not' support the facilities contention that their' answer is'more accurate ~than the' original answer. 1The
'~ facility!s answer is viewed as simply a rewording of.the' original answer, therefore,- the balance.of 'the answer . remains unchanged.'
'The facility is cautioned to include the reference'with the comment as-required by ES-201<
' COMMENT 2.14
. The intention of the matrix contained in the lesson plan referenced' in the answer is to show the NESW loads off each' units NESW header.
. It appears that this'was misinterpreted by the exam writer to indicate that some' loads on Unit 1 may be supplied from Unit'2:and some loads are only H supplied.from a specific unit. In actuality, because of the cross-tie capabilities.of the_ system,-any load on either unit can be supplied with !
water from the opposite unit.. Because of this confusion as to whether "both" means the load appears on both' Unit 1 and Unit 2' headers or.the load pay be supplied by both Unit 1 and. Unit 2,. we request that the; portion of the question requiring.which l unit or both the load is supplied from should be deleted.
- 19) Demineralized Water Retention Tank
- 20) Fire Protection Peggin0 Pump !
21)- Safety Showers and Eye Wash Stations i
- 22) Chemical Mixing Station H
- 23) Turbine' Room Sump neutralization
]
'24) Demin Wtr System Vacuum Degassified Seal Wtr
)
3 1 i
V 9 25 Ice Condenser Glycol Chiller Condensers
- 26) Plant Heating Boiler. Blowdown Tk
- 27) Reactor Coolant Pp Fire Protect
- 28) Containment Iodine Cleanup Fan Fire Protect REFERENCE Prints OP-1?-5115-34 OP-1-5105B-20 0P-1-5114A-15 OP-1-5114-38 OP-2-5114-22 OP-2-5114A-12 Are included for reference.
NRC Response: The question clearly asked for loads from the unit HEADERS. It then asks for which HEADER Supplies the load and gives the list of possible correct answer that will be accepted. As acknowledged in the facility comment, the cited reference lists loads from tea unit headers. Thus the original reference and question are consistent. The references provided with the facility comment support the original reference in showing that the original ar.swer key is correct. While it is.0. rue that the supplies for the unit NESW headers can be cross-connected, this Information is not required for credit on this question, nor was it elicited in the wording of the question. The question was very specific in requesting the header designation for supplying the load, and not which unit's LLSW pump supplies the load. Indeed, the word " pump" never appeared in the question. The list of additional loads requested by the facility to be included in.the answn; key are supplied from the NESW miscellaneous header. This is supported by both the original reference and the additional reference material submitted with the facility comment. Again, the question clearly asks for loads from the UNIT headers, which excludes loads from the miscellaneous header. For the above listed reasons, the antwer key remains
' unchanged.'
4 l
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b' ' 1.., , . , , l- , .- 31 l o : COMMENT 3'.01~
, 'The_ question,duetoiss. wording,asksforTWOconditionsthat.
generate ~an AUTOMATIC Containment Spray Signal. 'ThereLis only . ,
'one correct response'to this.~ While most. candidates willJprobably? '
provide the manual.actuationisignal also, it.should be~noted that 'i 1 they may omit the.second answer because there is~~only one. automatic-initiation signal.- We request that if only the. containment pressure response is provided, that full credit.be given'. REFERENCE 1
- g. ,
~
- Not applicable. .
NRC' Response: . Di'sagree. The.questionJasks.for two conditions that will .
. generate a containment spray / containment isolation Phase 8 actuation signal,'and'does not specify that this signal,is automatically or manually initiated. A consequence of this ..
signa 11being present (either manually.or automatically initiated)
'is the AUTOMATIC initiation of containment spray, along with other automatic actions; Thus it is incorrect to interpret this spray signal. However, because of possible misinterpretation of the question,.the manual initiation signal is deleted from the answer key _ and question value adjusted to 1.25.
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COMMENT 13.03 To make the. question a bit clearer, and more technically correctc
" Turbine trip system pressure . . ." should probably read as follows:
" Turbine trip (auto stop oil U-1 and safety oil U-2) systee pressure . . . ."
i Unit 2 Low S/G 1evel coincident with SF > FF is the same setpoint as Unit 1 and should be included as a possible trip in the Unit 2: trips.
-Also as an editoria1' comment, experience'will bear out that, for the same level of knowledge tested, the more straight forward the question is, the l higher the reliability factor will be in determining the candidates' level l of knowledge for that information you wish.to test on. Asking the candidate to provide, for example, the trip setpoints, coincidence and logic for a number of trips on both units is a more straight forward method of asking the same level of knowledge required-to answer this question. 4 LREFERENCE Unit 1 and 2 OHP-4023.E-0, SYMPT 0MS OR ENTRY CONDITIONS, Pages 1 and 2 of 23 (previously provided). .
a ': (
NRC Response: During the administration of the exam, candidates questioned the use of the words " Turbine trip system pressure." A clarification ,, was made-during the exam that these words were equivalent to auto l p, stop oil on Unit 1 and safety oil on Unit 2. 'he facility is . I requested to correct the misleading wording in reference I materials previously provided to the NRC for exam preparation. I The additional trip for Unit 2' requested in the facility comment is included in the answer key and the point value
'for.the question is appropriately adjusted. I ES-203 Paragraph F discusses structufe of test. questions. Test questions are written to be placed as high on this hierarchy as possible. Rewriting the question as suggested by the facility would move the question from Category 4 (analysis and deduction) to Category 1 (knowledge and recall). This is undesirable and i therefore the question will remain unchanged.
COMMENT 3.04 The question and answer are not technically correct, apparently due to a misunderstanding of the interrelationship between P-13 and P-7 in the reactor protection scheme. Only ONE input channel to P-13 is necessary to given an output. Only ONE input channel (P-13 _o_r P-10) to P-7 is necessary to given an output from P-7. An output from P-7 coincident with a standing turbine trip signal (from 4/4 stop valves closed in this scenario) will result in the reactor trip breakers opening. With C&I performing repairs on the FIRST failed channel, the withdrawal of Shutdown Banks and should never have been initially authorized by the SRO. Allowing C&I to test the SECOND channel would be totally unsatisfactory with a 1/2 coincidence. All candidates should exhibit an adequate knowledge of P-13 in their answer, but we highly recommend extreme latitude in grading this question based on the above discussion. REFERENCE R0-C-NS11 composite functional diagram of P-7, P-10, and P-13. (Previously supplied, attached for convenience.) NRC Response: Question deleted. COMMENT 3.07 1 As a general comment, the questien does_not solicit a particular detail of answer. A candidate with a lot of time, could write several pages - one limited in time could answer as follows: "S/G level: Remains unchanged at 44% throughout" "Feedwater: Pump speed increases to meet increased flow demand." The question apparently expects the candidate 6
E e l o ,' [ l- { to assure that a STEP load increase has occurred in order to get the response the answer key requires. Step load increase are avoided in g plant ' operations and only. occur at a rate of about once per year for l L both units a D.C. Cook - Step load increases would not be an expected 1 assumption. If a step load increase is assumed, the feedwater level j control response in the answer key is^still incorrect. The first ' indication of a step load increase is S/G level INCREASE due to level swell. - A suggested substitute answer, if a step increase in power -is assumed, is as follows:
" Steam flow increases (and steam pressure decreases which will cause S/G ,
1evel to swell). The increase in steam flow will cause the FRV to open (level error is " lagged" in the circuit to minimize the effect of shrink and swell). Steam flow and feed flow will become matched after a short period of time and S/G will remain at its program level of 44% (20 to-100% power program)." The answer for Main Feedwater system response in the key is viewed on a microscopic level. For a step change power increase, this answer is more - appropriate than it would be for a ramp up power increase. The Main Feed Pump response to a ramp up power would be as follows:
"As feed demand and feedpump AP program increasec, the feedpump control system will increase feedpump speed (and maintain the AP program)."
Under actual operating conditions at D.C. Cook, a ramp power increase is a " straight line" trend on S/G chart recorders and meters. No oscillations will be seen in a normal ramp up power. We ask that latitude in grading be given for this question based on the preceding discussion, especially in light of the question not specifying degree of detail required in the answer. REFERENCE R0-CO-PG11 (previously supplied) NRC Response: It is inappropriate to assume that this question assumed a st :p load change. In addition, during the administration of the exam, all candidates were told to disregard the effects of shrink and swell. However, it is agreed that the original answer provided was inaccurate. Therefore, the answer is changed to read:
" Total steam flow will increase (.5), causing program delta P to increase (.5), which causes MFP speed to increase (.5).
This will increase feed flow (0.5), causing steam generator level to remain constant at programmed level (.5)." The total point value of this question is adjusted as indicated. 7
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COMMENT '3;09 ~ '
-The question is' very straight forward. arid clear as to its. required answer. We'do,.however,-request that " VARIABLE" or " RAMPED with POWER".be an' acceptable setpoint for the Unit.1 steam flow value.
^
; REFERENCE lNot; applicable.-
NRC. Response: As acknowledged in the facility comment, the question clearly stated that a setpoint was required. "Setpoint" implies that a-number .is . required, thus " variable'.' .or " ramped" are inappropriate responses to the question. As indicated in the answer,'any. setpoint stated by a candidate that is within'the required range will be accepted for full credit. The answer key remains unchanged.
.L 'COMM:NT. 3.10
- b. Because'the question implies that.C-7A'and/or C-78 (the load rejection bistables) have been made up in stating "on a: load rejection," candidates may. assume they are not required to restate-in their answer what the question implies. Also,.during a load rejection and for a short time thereafter, until the rod control system acts to match Tave to Tref, this greater than 5 temperature deviation will exist. This again, from the candidates point of view, may be implied by the question and may not need to be restated.
'Considering both of'these points, candidates may assume the only additional requirement other than C-9 is that the ON/0FF switches
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.for both-trains of steam dump be in the ON position. Furthermore.
-the answer key lists three correct responses, yet the question does not specifically address how many responses are required to receive full credit for this question. .If the question were worded ". . .
what three other conditions in' addition to C-9 . . .," this would have cued the candidates that C-7A.and/or C-78, a temperature deviation'and Tave greater than 541 F would be the logical choices. The question did, however, ask for " conditions" which indicates'more
-than one response is required.
We request the answer. key be modified as follows:
- b. 1. C-7A (accept explanation)
- 2. C-78 (accept explanation)
L 3. Tave greater than 541 F-p. 8 i.
p- ). i .
- 4. 'Tave greater than Tref
- 5. Both Steam Dump switchesin "0N" E (Any 2'.of the above .375 pts each)'
- c. If a candidate. assumes that Tave is less than 541 F, then one of-the required conditions for steam dump operation would place the steam dump in bypass interlock. We request the answer key be modified as follows:
- c. 1. Tave greater than 541 F
- 2. Steam pressure (UPC-101) greater than setpoint
- 3. Bypass interlock if Tave is less than 541 F (Any 2 of the above .25 pts each)
NRC Response: Part 6: The question did not state the magnitude of the load rejection; therefore,-it is inappropriate to assume that C-7A or C-78 is present, since not all load rejections will trip the load rejection bistables. Also, not all load rejections will result in a temperature error greater than 5 degrees. Therefore,-the answer remains unch'anged. Part c: The question was meant to preclude bypass interlock as a way to get steam dumps open; howeve- it is agreed that this is not clearly stated. The answer key is modified as follows:
"1. Steam pressure greater than setpoint (.25) and (either of the following at .25)
- 2. Tave greater than 541 F
- 3. Bypass interlock if Tave is less than 541 F" COMMENT 3.11 j
- a. The answer for this question is dependent upon whether P-7 is viewed as a blocking circuit or a permissive circuit. Either approach is acceptable but the setpoints and coincidences are different.
If you assume P-7 blocks the seven reactor trips, then the following results: P-7: Blocks trip if it does not have input from both P-10 and P-13. ; 9 i
._ _______________________J
dI l' P-10: 3/4 Power. range NI's'< 10% P-13: -2/2 Impulse Pressures < 10%
.If you assume P.-7 (permissive) enables the seven Rx trips, then the following results:
P-7: Enables trips.if it has input from either f P-10'or P-13 P-10: 2/4 Power Range NI's > 10% P-14: 1/2 Impulse Pressures > 10% REFERENCE R0-C-NS11 SH03 Page 6. R0-C-NS11 Tp-12 (attached). NRC. Response: The question specifically stated " permissive P-7." Therefore it is inappropriate in this question.to view P-7 as a " blocking circuit." Also, the facility comment makes no. specific .i recommendation for changing the answer key. For these reasons, the answer key remains unchanged. COMMENT 4.01 It is suggested that full credit should be given for responses identifying the fan units by component designation nomenclature: Containment Purge Supply Units: CPS. Containment Purge Erbaust Units: CPX. Instrument Rm. Purge Supply and Exhaust: CIPS and CIPX REFERENCE Drawing No. OP-1-5147A (attached). NRC Response: Concur. Answer key modified. COMMENT 4.02 Respondents citing Shift Supervisor notification should not be penalized ,
'for including this in their answer. As indicated in the Immediate Actions, the fuel handling supervisor notifies the SS/ Control Room of the incident. If the Control Room receives the information, the .
Shift Supervisor must be notified immediately to address E-Plan ! implementation (see PMP 4050.029.004, Page 2 of 3 attached). l 1 10 I _-_-:---_-- _ ._ i
o, . . Respondents' indicating that Aux. Building evacuation would accompany the PA announcement of occurrence of a fuel handling accident in the spent fuel storage area should not be penalized for this' inclusion. Evacuation would be performed as a response to Immediate Action 4.2.2 (attached). Point value assigned to Question 4.02 (3.00) when compared with the value i of Question 4.03 (1.50) appears inappropriate for the Importance assigned l by the K/A catalog. In accordance with NUREG-1122, the immediate actions for which an R0 is responsible following a fuel handling accident in the SFP area has an importance of 2.8, while the immediate actions for a loss of all AC have an importance of 4.1. In the interest of fairness in appraising the required knowledges and abilities of the reactor operator candidates, we: request the point values of Q.4.02 and Q4.03 be adjusted to reflect the level of importance: Q.4.02 Adjusted from 3.0 points to a value of 1.8 pts. Q.4.03 Adjusted from 1.5 points to a value of 2.7 pts. This adjustment will not affect overall category point value. REFERENCE K/A Catalog: PWR Page 3.7 - 20.1 and 3.11-11 (attached). NRC Response: The question specifically asked for the immediate actions required to be carried out by control room personnel. The Shift Supervisor is not normally considered part of the control room staff. However, if Shift Supervisor immediate actions are included in the response, and are correct, credit will not be removed. In order to gain full credit for the answer, the candidate must include all answers listed on the answer key, in addition to any other answers included being correct in accordance with the cited procedure. The answer key. remains unchanged. K/A values are used to determine what arees are to be included in an exam, as well as a basis for point'value assigned to a question. However, the K/A value is not the sole criteria for assigning point value to a specific question. ES-202 l Paragraph E.14 gives guidance on assigning point values for ! questions. It is felt that the original point values assigned j to Questions 4.02 and 4.03 are appropriate based on the depth of answer required. The assigned point values remain unchanged. l^ l ! i ! ! l L l i 11 1
r ! e. . COMMENT 4.03 The reference should be changed to: OHP 4023.ECA-0.0, Pages 2, 3, and 4 (attached).
'NRC Response: Concur. Answer key modified.
COMMENT 4.04 Part (a) Responses 3, 4, and 5 as addressed in OHI-4014 identify the specific means of verification which should be employed. The use of i.e. , in the instruction indicates "that is." We request that consideration be given for responses which reflect the actual methods used to determine valve position during a lineup.
- 3. Mechanical position inJication (i.e.; stem travel indicators, butterfly disc position' indicators . . . )
- 4. Remote position indicators (i.e.; limit switch lights, meter position indication, . . .)
- 5. System response (i.e.; flow, pressure, temperature, . . .)
REFERENCE Page 2 of 3, OHI-4014 (attached). NRC Response: Concur. Answer key modified. COMMENT 4.09
- a. Heat Sink, FR-H.1, would be a red path on the Critical Safety Function Status Tree F-0.3, " Heat Sink." The answer cites FR-H.2 which is entered on a yellow path from F-0.3. It is suggested that responses identifying FR-H.1 as a red path procedure and answering (2, 3, 4, 6, 5) be given full credit. In addition, the question directs respondents to list the order that these procedures would be performed in if all the procedures were to be used. The wording may cause some respondents to list (1), Subcriticality, as the sixth and final procedure to be performed because of the direction provided in the question. We request that no credit should be lost by respondents as a result of listing (1) as the last to be performed.
REFERENCE OHP-4023.E-0.3, Heat Sink (attached). 12
--- _j
4s A ,- NRC Responsei Credit will be given if a candidate identifies.FR-H.1.~as a red path procedure and answers appropriately. No credit.will.be 4 deducted for including ~subcriticality 'sa the last procedure. Credit will be' deducted if. included elsewhere in the answer. COMMENT 4.10
.Part (e) might be better. understood on future' exams if worded:
Following a_ reactor trip, with RCS Tavg stable at 547 F, 2 steam generators become faulted."
NRC Response: Comment noted. There is no facility recommendation for change,:therefore, the answer key remains unchanged. COMMENT 4.11-This question requires several detailed responses specific to W. % and responsibilities at the SR0 level for full credit. The Shift Supervisor , .is the-sole-individua1'with the authority.to dispatch people to guard all control of isolation points. We request that no credit should be lost by respondents for citing the SS (Shift Supervisor) as having responsible authority. Additionally, we request consideration be given for responses indicating. that="a Clearance Permit should be hung as.soon as possible." This statement, included in Paragraph.3.1.1.1, might be interpreted as a required limitation to the exception-for emergencies. Responses addressing minor adjustments and troubleshooting may include
'.'or pressurized" together with " energized" and we request that no credit should be deducted for-this inclusion. Responses which reference " Job Orders" instead of "PMI-2290" thould be acceptable for credit as PMI-2290 is the Job Order instruction, and is cited on Page 2 of 18, PMI-2110 (attached).
NRC Response: R0's as well as SR0's should be aware of tagging requirements as indicated by the cited K/A. No additional reference 'is included to support the facility statement that this question is at the SR0 level. Therefore, the question and answer remain unchanged. No credit will be deducted for citing that the Shift Supervisor has responsibility. However, if a complete explanation of requirements is not given as indicated, credit will be I deducted. " Pressurized" is analogous to " energized", as are-
" Job Orders" and "PMI-2290", and thus will be accepted for full credit. No credit will be deducted for stating that .
a clearance must be hung as soon as possible. ! L 13 L - _ _ _ _ _ _ . _ ______ _ _____ __ _ l
- MASTER CUlti U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION
" FACILITY: _Qg g E _1k 2_ _ _ _ _ _ _ __ _ _ _ _ __ _,
#s* *'4t g?g REACTOR TYPE: _PNR-WEge________________
l .{% g E l e S g g DATE ADMINISTERED: _@ZZQ@f9Z________________
***** EXAMINER: _D6MQN 2 _Dz_______________
CANDIDATE: _ ______________________ 1NSIBUGIlDNS_19_GBNQlDBJE1 Use separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The grade passing of at grade requires at least 70% in each category up si:: (6) hours after and a final least 80%. Examination papers will be picked the examination starts.
% OF CATEGORY % OF C ANDI DATE' S CATEGORY
__YGLUE_ _IQI6L ___@GQSE___ _y@LUE_________________Q6IEQQSY____________
- 1. PRINCIPLES OF NUCLEAR POWER
_23tE9__ _2Ez1] ___________ ________ PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 252b3 ________ 2. PLANT DESIGN INCLUDING SAFETY _2Ez99__ ___________ AND EMERGENCY SYSTEMS ________ 3. INSTRUMENTS AND CONTROLS _2Ez99__ _22252 ___________ PROCEDURES - NORMAL, ABNORMAL, _2Er99__ _2Ezb3 ___________ ________ 4. EMERGENCY AND RADIOLOGICAL CONTROL _22.nEp_ _ ___________ ________% Totals Final Grade 1 Al a work done on this examination is my own. I have neither given l nor received aid. Candidate's Signature MASTER COP _Y l u_________ __
.. a 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 cnd could result in more severe penalties.
- 2. Restroom trios 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 facilitate legible reproductions. .
4 Print your name in the blank provided on the cover sheet of the examination.
'I
- 5. Fill in the date on'the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers. l
- 7. Print your name in the upper right-hand corner of the first page of gach section of the answer sheet.
- D. Consecutively number each answer sheet, write "End of Category __" as appropriate,. start each category on a ngw page, write gnly gn gng sidg 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 thcgg lines between each answer. i
- 11. Separate answer sheets from pad and place finished answer sheets face f
{ down on your desk or table.
- 12. Use abbreviations only if they are commonly used in f acili ty li tgratutg.
- 13. The point v6 se f or hath question is indicated in parentheses after the ]
question and~can be used as a guide for the depth of answer reqJired. l l
- 14. Show all cal cul ati on s , methods, or assumptions used to obtain an answer {
to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
16..If parts of the examination are not clear as to intent, ask questions of J the ggamiger only. j i
- 17. .You must sign the statement on the cover sheet that indicates that the work i s your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.
C-_-______.__ _.
f, P I
.l:
410. When.you complete your examination, you shall .! f$. Assemble your examination as follows: Exam questions on. top.
~
(1)
-(2) Exam aids - figures, tables, etc.
(3)- 'AnswerEpaees including figures which are part of the answer. .
, 1 j
- b. ; Turn in your copy'of the examination and all pages used to answer the examination questions. ]
i
- c. ' Turn in all. scrap paper and the balance of the paper that you did q
not : use f or' answering- the. questions. 4 Ld. Leave the examination area, as defined by the examiner. If after leaving,-you are found in this area while the examination is still j j in' progress, your license may be denied or revoked. . 4 1 i I l 1 1 I i f' t t -- - - - - . - _ - - _ - - _ - - _ _ _ _
' Pege i st 5 .,'
DATA SHEET l CTOR TEORY FORMA.AS: i P = Poet /T l P o Po10SLR(t) Po Ig lth V 3.12 x 1010 fissions /sec pa t*'+ .5eff ' 7 1 + AT ! l Tc 1* i p .I T = Ieff - p Ap j op = In K (final)
~
K.(initial) p = K-1 l K 1 SUR = 26.06 7 Kerr = c Pr p Pth f4 e 2 2 3
-(B Lth )
p 1 + (B2 Lth2) ) 2 Pr = e_(g2 Lf ) pae-[N}[Ierr) IIs i L C1 (1- Kerri) = C2 (1 - Keff23 oa 1 = C (final) ;I T4 t (initial)
- t-cry = 1 Af + 1 Ap -
B2 ( ALf2 + E.W) f At p At . At 'At
.____-_____________-__-_3
Pepe 4 Cf 5 g DATA DEET AVEpWE TEmRL CO@UCTIVITY (K) BTUjhr-ft *F , Material 0.025
' Cork _.
0.028
' Fiber Insulating Board : -
0.GP6' Maple or Dek Wood D.4 Building Brick O.45 1 tindow Glass' 0.79 Concrete. 25.00 1 Percent Carbon Steel '35.00
'l Percent Chrome Steel 118.00 Aluminum .
223.00 mr 235.00 ' Silver 0.392 Water (20 psia,200*F)
. Steam (1000 psia, 550*F) 0.046 1.15 Uranium Dioxide 0.135 Helium 10.0 Zircaloy.
MISCELLANEOUS IWORieTION:
. E = ac2 NE = 1/2 mv2 PE = agh Vf = Vo + at b
AE = 931 Am, AREA VOLLDE E.06t.inIC MECT TRIAPOLE A = 1/2 kh ////////////////// A' = S2 ////////////////// M
//////////////////
RECTANr3 F A=LxW
/////////////////j A = wr2 CIRQ.E A = 2(LxW + LWi + W40 V=Lxwxh RECTANGULAR SOLID !
' A = (2 Tr2)h + 2(vr2) V = wr2h !!
RIGHT CIRCLA.AR CYLIM A = 4 sr2 Y = 4/5 vr3 SPERE umE //////////////////////// V=S3 - 4 a
i ., Page 5 cf 5 .' e .< . . . - DATA SEET ] i NISCELLAENE IWIRM4 TION CONTINLED: 10 CFR 20 Appendix B I Table I Table 11 - S Gamma Col 1 Col II ColI Col II Energy EV per Air Water Air . Water f Disintegration ucAal ucAal uchl ucAal
. Haterial _ Half-Life Sub 2x10-6 4x10-8 - l Ar-41 1.83 h 1.3 e 00-60 5."I/ y 2.5 S 3x10-7 1x10-3 1x10-8 5x10-5 .f l S 9x10-9 6x10-5 --1x10-10 3x10 1-131 8.04 d 0.36 Sub 1x10-5 __ 3x10-7 _----
Kr-85 10.72 y 0.0L S 9x10-7 4x10-3 3x10-8 1x10-4 [t Ni-65 2.52 h' O.59 S 2x10-12 1x10-4 6x10-14 5x104 j Pu-239 2.41x104y 0.0N S 1x10-9 1x10-5 3x10-11 3x10-7 Sr-90 29 y - 9A 4x10-6 - - - - 1x10-7 ----- Xe-135 9.09 h 0.25 { Any single radionuclides with 3T fp 2 hr 3x10-9 9x10-5 1x10-10 3x10-6 j.. which does not decay by a or spontaneous l fission NeJtrons per an2 Average flux to deliver Neutron Energy (EV) 100 area in 40 hours e dvalent to 1 rem 970 x 106 h70 Thermal 280 (EUTRONS) 0.02 400 x 106 o=2 "c 185 #:18! 19 LIEAR ABSOFTION COEFFICIENTS y (cm-1) j. Energy (EV) Water Conc: rete Iron Lead
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2.CEP 4030.BTP.032 f.. ,' 'IATEST REV. CEECEED
' ' ' CALCD1ATION SEEET 4.1 QUADRANT POWER TILT RATIO CAICUIATION
=- l N 41 DET A.I DIVIDED BY N 41 DET A 1204 I
=
. N 42 DET A I DIVIDED BY N 42 DET A 1204 I
=
N 43 DET A I DIVIDED BY N 43 DET A 120% I
=
N 44 DET A I , DIVIDED BY N 44 DET A 120,4 I 4 UPPER TOTAL UPPER TOTAL DIVIDED BY 4 = AVE. UPPER DIVIDED BY AVE. UPPER = UPPER TILT RATIO MAX. UPPER
=
(4kDETBI DIVIDED BY N 41 DET B 120% I
=
N 42 DET B I- DIVIDED BY N 42 DET B 120% I
=
N 43 DET B I DIVIDED BY N 43 DET B 120% I
=
N 44 DET B I DIVIDED BY N 44 DET B 120% I LOWER TOTAL DIVIDED BY 4 = AVE. LOWER IDWER TOTAL DIVIDED BY AVE. LOWER = I4WER TILT RATIO NAX. LOWER QUADRANT POWER TILT RATIO = Max. P-250 QUADRANT POWER TILT TATIO TECHNICAL SPECIFICATION 3.2.4 LIMIT = 1.02 FLUEE INSTR 9 CAL DUE DATE CONDITION REPORT REQUIRED Yes No DATE TIME PERFORMED BY REVIEWED BY DATE SRO Page 1 of 1 Rev. 2
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PAGE 2 01!1a; EBING1ELEf_9E_NMELE98_EDWEB_EkeNI_9EEBBII9N2
-ISEBDDDYNed1GDa_UE9I_ISBNEEEB_9ND_ELVID_EL9W. :
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" 1. 01 f (2.50) J GUESTION u IThe Unit' 1 reactorf s-operatingi atl50% power, BOL', when a steam dump fails
. Assume rods:are-in manual, no operator action is taken, and no open.
reactor trip occurs'. Explain HOW and WHY reactor power and Tave;will ] change, and where.they will stabilize in relation .to the initial values. j 1 i
- QUESTION? '1.02 ' ( 2.~ 50 )
a.- Which one of the following descriptions best supports the reason why. Xenon ' reactivity increases _ sharply af ter a ' trip f rom 1000 hrs. at 100% power ? (0. 5) - e
/
'1) Xenon decays less rapidly due to. a reduction in the neutron flux.
'2) Iodine half-lif e is -much shorter than Xenon half-life.
3)- Iodine production is greatly reduced and Xenon production is greatly increased due to the reduction in neutron flux. increases,
- 4) Due to reduced-neutron absorption,. Iodine concentration and Xenon decays directly f rom Iodine, thus Xenon increases.
- b. .Give two reasons.why Sm-149-is.not;as much(1.0) o f-a concern to an operator.after a reactor trip as is Xe.
A Xe oscillation in a' reactor core might be produced by certain types c. of rod motion. How would the Xe oscillation'resulting from the
;f ollowing< two cases be dif f erent? Explain. (1.0)
- 1. :A turbine runback occurs with rods in auto.-Rods drive 60 steps.
II. Rods are driven 60 steps starting from the same position as in Case I, but slowly over 4 days time. [
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(***** CATEGORY 01 CONTINUED DN NEXT PAGE
I PAGE- '3 3,14__EBl*51ELEH_DE_N99LEBB_E9 WEB _ELANT OPERATION2 IBEBU99YNSDJQg3_SggI_IB9hgEg8_$Np_E(yJp_E(QW '/. QUEST 1DN. 1.03L ( 2.50) J c.- What'is thelsubcooling margin (SCM) of the plant if.the following conditions. exist 7-(1.0) Th,'= 580 F- Ppzr_= 2185-psig. Tave = 550 F' Psg = 850 psig Tc = 520 F
. I! plant. power is raised from 50 to 100%, how-will SCM
~
b.
. change . (increase, decrease, stay the same)? 'Why? (0.75).
c: . >
. Which of. the r f ollowing would give a smaller SCM? Assume normal expected temperature and pressure f or each case. Briefly explain why.
- 1) SCM during a stabilized natural circulation decay heat removal f ollowing a reactor trip from loss of flow. i 2)- SCM f rom continued operation at 5% power
- 3) SCM' produced when'all RCP's are operated at normal no-load temperature after extended. shutdown.
QUESTION 1.04 (2.00) Two l'dentical reactors are taken c.-itical using continuous rod withdrawal. ~ Reactor A has a rod speed of 48 steps per minute and reactor B has a rod speed of'24 steps per: minute.
- a. Which reactor will have the highest source range counts at (1.5) criticality? Explain.
- b. How will 1E-8 critical rod heights ' con: pare in the two reactors? (1.0)
Explain bri efly. i (***** CATEGDRY 01 CONTINUED ON NEXT PAGE
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PAGE 4
*i 1 ;1EBINQlELES_QE_NUQ(EQB_EQWE8_E(QUI _QEgBQIlQN m ,
IMEBdQQYNed1GEt UE81_IBONEEEB_9ND_ELWID_ELQW a .. QUESTION- 1.05 '(1.50) i
)Uning the f ollowing initial' conditions, calcu* ate the MINIMUM number of
-cteps of_ rod bank insertion'that are required to ensure the reactor is ,(1.2) l
' EXACTLY CRITICAL.- Show'all work.
-SUR i s . O. 5 DPM
-ef f ective. delayed neutron -f raction =. O. 005
-average' neutron precursor decay constant = O.OB sec -1
-rod bank worth = 5 pcm/ step QUESTION 1.06 (3.00)
A reactor has the following characteristics at 100% power: Tave is 573.5 F and Tstm i s 513. 0 F. The plant is shutdown for maintenance and 5% ofLthe steam generator tubes are then plugged. The plant returns to 100% power. Given.that Tave_is again 573.5 F, determine the pressure of the' steam
. leaving the, steam generator. State assumptions.and show all work.
QUESTION. 1.07 (2.00)
- a. If the' reactor is operating in the power range, how long will Show it takeall to raise power from 20% to 40% with a +0.5 DPM Startup rate? (1.0) work.
- b. Will it take the same amount of time to raise power from 40% to 60% if (1.0) j the same startup is maintained? EXPLAIN.
I f
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PACE 5
.12 _E810GIE6ES_9E_UUQLE86_EQWE6_EL6MI_QEE8611QUt ISE8dQ9YUBulCg,_SE61_I66MSEEB_6NQ_E(Q1Q_ELQW QUESTION 1.08 (2.50)
- c. Describe how the pressurizer thermodynamic 11y attempts to thaintain Ignore the RCS pressure during insurge AND outsurge transients. (1.5) operation of heaters and spray in your explanation.
(.5 ea)
- b. Answer the f ollowing TRUE or FALSE
- 1. The pressurizer steam space is ideally a little hotter than the liquid space.
than
- 2. The primary consideration f or spray flow being COLD leg rather HOT leg is to provide better pressure reduction because of the coldar fluid.
QUESTION 1.09 (2.50) The Unit 2 reactor plant is operating normally with the plant parameters indicated below. Plant power 50 */. Total RCS f low 130 million pounds per hour RCP PSID 90 RCP supply voltage 11.5 KV RCP motor power 4400 KW Grid frequency 60 Hz Due to an off-site problem the grid frequency decreasew to 55 Hz and (Show the RCP's continue to operate. Assume that the RCP's are ideal pumps. your work and state assumptions needed f or any calculations. ) (a) After the frequency decrease, what is the Total RCS flow ? (in million (0.75) pounds per hour) (b) After the frequency decrease, what is the RCP di f f erential press,ure? (0.75) (c) Assuming constant motor efficiency, what is the RCP motar power at( 1.the 0) lower frequency ? (in KW) l (***** CATEGORY 01 CONTINUED DN NEXT PAGE *****) ~ . _ _ - - -
i J / PAGE 6
,1g' PRINCIPLES OF NUCLEAR POWER PLANT' DPERATION S-IdEBdQDYNed1GE t_dE01_IBBNSEgB_eNg _ELylg_ ELgW A 10'
- s
'DUESTION '1,10 (1.50)'
i For each of the questions below,. select the correct answer. . (a) what reactivity addition is required to double the count rate if Keff (.75) is 0.98 ?
- 1. 1015 pcm
- 2. 5000 pcm
- 3. 101.5 pcm
- 4. 500 pcm^
(b) If a reactor core with a neutron source is exactly critical at 1000 CPS the source-range, over the next few minutes the count rate should: (.75)
- 1. Increase exponentially with time 2.. Increase-linearly with time
- 3. Remain constant at'1000 CPS
- 4. Increase geometrically with time DUESTION 1.11 (3.00)
(1.0)
- a. Define Sub:ritical Mul ti pl i cati on (M).
a criticality prediction.
- b. Briefly explain why M is not used to plot (1.0)
- c. If the count rate is 100 cps at a Keff of 0.95, what will the count (1.0) rate be at a Keff of 0.997 Show all work. !
l 1 l (***** END OF CATEGDRY 01 *****) _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ 1
r PAGE 7 P . 2c_sEL6MI_QEEl@N_lNGLUQ16@_g@Egly_9NQ_gd[RQENQY_gy@ led @. f OUESTION2.01 (1.50) Describe three (3) interlock or automatic control functions associated with,ORV-160, 161, and 162 (CVCS orifice isolation valves) that cause L these' valves to shut. Include setpoints, where applicable.
-QUESTION 2.02 (3.00)
.The f ollowing concern the RHR system;
- a. What interlocks or automatic control features are associated with ICM-129.(RHR pump suction f rom hot leg)? Give values for both units. i (1.5)
- b. What interlocks must be satisfied to OPEN IMO-340 (East Hx to SI pump (1.5) l suction)7
.l 1
QUESTION 2.03 (3.00) Briefly describe the flow paths f or high head, medium head, and low head injection. Include in the descriptions, all major components and valves that fms4 injection water passes through from suction source to all possib.a injection points into the RCS. Describe flow paths f or injection phase only. l QUESTION 2.04 (1.50) A bypass valve is provided around the pressurizer spray valves to maintain ; a constant 1 gpm flow even with spray valves shut. What are TWO reasons for~this bypass flow?
)
I 1 DUESTION 2.05 (1.50) I Concerning the containment spray system: l
- a. What are the bus power supplies for both units' CTS pumps? (1.0)
- b. When in the recirculation mode. whus is the source of water for the CTS pumps? (0.5)
(***** CATEGORY O2 CONTINUED ON NEXT PAGE *****)
l PAGE .O j
' %z;ickeNI_ pef 19N_INEbup]Ng_ggggly_gNp,gDER@gN9y,$1SIgD) i
\
q l l (1.50) i QUESTION '2.06 ) l List or describe 3 of-the 5 annunciators available in the Unit 2 control f room f or monitoring of the ice' condenser system.
.) ,
OUESTION 2.07 (1.50) I What 2 conditions must be satisfied i order to open val ve RRV-306 (Waste
)
Gas Release Valve)7 ! I
-QUESTION .2.08 (1.00)
Concerning the Liquid Waste Disposal Systems a.' What condition will cause. automatic closure'of valve RRC-2857 )
.b. What condition will cause valves RRV-286 and 287 (Effluent Valves) to close?
\
1 QUESTION 2.09 (1.50) ! The "SFP Cooling System Abnormal" annunicator in either cotrol room is l f ed by various alarms on the Spent Fuel Pit Subpanel. What are three of -{ the six conditions on the SFP subpanel that will actuate the control room ! annunicator? (Setpoints are not required) OUESTIDN 2.10 (1.50) Describe the sequence of events in the Essential Service Water System upon receipt of a saf ety injection signal . Assume no pumps in lockout, l' recirculation sump valves are open, and phase B' signal present. QUESTION 2,11 (1.50) What three signals will cause a component cooling water pump to auto start ! with the control switch in the " auto" position? l l l (***** CATEGORY O2 CONTINUED ON NEXT PAGE *****) i
~~ -- , ,_
PAGE- 9. - i M A _fLONI_PE!!99_INGkVPIND_fBEEIY_BNP_EDEB9EN9Y_fYSIED) o s i !. , l l tQUESTION 2.12- (2.00) l List-all sources of'. water f or. the Auxiliary Feedwater pumps. in their preferred order of use. QUESTION 2.13 (1150) List three (3) conditions that will cause an automatic trip of. the Emergency Diesel Generator OUTPUT BREAKER. QUESTION 2.14 (2.50) List 10 loads supplied by thr. NESW unit headers. Indicate for each load if_it is supplied by the Unit 1 header, the Unit 2 header, or both. 1 l 1 l 1 (***** END OF CATEGORY O2 *****) l
PAGE to
.ite INpIg g M Ig_8Np_QQNIB9k!
.i QUESTION 3.01 (1.25) <
'The Containment Spray System will automatically start upon receipt of a containment spray / containment isolation Phase B actuation signal . , List
-TWO conditions that will generate this actuation signal. Include coincidence and setpoint, if applicable.
QUESTION 3.02 (2.50) Describe five (5) conditions that will prevent withdrawal of control banks while in automatic rod control . Include setpoints and coincidence, where cpplicable. QUESTION 3.03 (3.75) , You are given the following data for' Units 1 and 2. - Power was initially 100% RCP Bus Voltage 3175 volts Steam Generator A NR Level 25.5% Steam Generator A steam flow-f eed flow mismatch .8x10E6 lb/hr Turbine Trip system pressure 92 psig Loop A flow 92.5% Steam Generato- B NR level 19% Pressurizer pressure 1975 psig Both units have tripped. What are all possible causes of the trips for EACH unit? Include in your answer the applicable setpoints. l QUESTION 3.04 ( .00) (deleted) (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****) L -- ________ _ ___
^b2k_;1sPIBQMgNJy_ANp_ppNISQk! PAGE' 11 'l oc '
i
-i QUESTION 3.05 (2.00) ,
. Given the f ollowing . data' concerning the- power range nuclear instruments:
N41 N42 N43 N44 Upper'Actus'.1 Reading- 52 56 58 57
^i Upper:120%. current 104 112 112 108
-I Lower Actual Reading 53' 55 56 54 Lower 120% current. 106 110 112 108 (Note: All readings are in milliamps)
What'is the' Quadrant Power Tilt Ratio (QPTR)?- DHP-4030.STP.032 Calculation sheet 6.1 is attached f or your use. Include'the completed calculation sheet with your answer. y QUESTION 3.06 '( 1. 50 ) Concerning the Auxiliary Feedwater Pumps: a., What conditions will cause an auto start of a MOTOR DRIVEN pump if (.75) the control switch is in NEUTRAL?
- b. What additional condition will cause a MOTOR DRIVEN pump to. auto
(.25) start if the control switch is in AUTD?
- c. What conditions will cause an auto start of the TURBINE DRIVEN' pump?
(0.5) 3
/
1 QUESTION 3.07 (2.50)
. Describe _the response of the steam generator level and main feed system
- when increasing turbine power. Assume that the plant is initially:at 70%
power and all control. systems are in automatic. l 4
-)
l t I (***** CATEGORY 03 CONTINUED DN NEXT PAGE *****) s
PAGE' -12 l
- ;2_c_INEI690ENIS_8NQ_CQNI6QLE i
l DUESTION 3.08 (2.50)
- 6. s What are the pressure setpoints at.which the cycling and backup
' heaters energize and de-energize while in automatic control? (1.5)
L b. When. using 'the PORVs f or cold. overpressure protection, what are.the (1.0) opening setpoints' f or BOTH units? E QUESTION 3'.09 '(3.00)
.There are five ECCS actuation signals for each unit, four of which'are the name f or both units. These are: manual, low pressurizer pressure, lower containment high pressure, . and high steamline delta P. l Describe the fifth ECCS actuation signal for EACH unit,. including setpoints and coincidence. i QUESTION 3.10 .(2.00)
Concerning the steam dump system:
- a. Describe the conditions necessary to satisf y the C-9 permissive. (.75)
- b. If.the mode control switch is in Tave mode, what conditions in addition to C-9 must be met in order to open steam dumps on a load rejection? Assume no turbine trip and steam dumps inautomatic. I
(.75)
- c. If the mode control switch is in STM PRESS mode, what conditions in addition to C-9 must be met'in order to open steam dumps? Assume l steam dumps are in automatic, and are not bypassed. (0. 5 )
l l QUESTION 3.11 (2.00) r Concerning the reactor protection system:
- a. Describe, in detail, the logic f or permi ssive P-7. Include coincidence and setpoints for all inputs to the P-7 circuit. (1.0) l
- b. What are f our (4) reactor trips that are enabled above P-7?
(1.0) Setpoints and coincidence not required. l, (***** END DF CATEGORY 03 *****) i
PAGE- 13 0c_ EBpggpuggS_ _NpBDSL2_BDNDSDDbx_gDEBGENCy_$Np 669196991G66_G9NISQL DUESTION 4.01 (2.00) Per PMP 4050.029.003, " Fuel Handling Accident in Containment Building", one of the immediate actions is f or control room personnel to verify that cutomatic actions have occurred. What are 4 of the 7 automatic actions that occur based on increased radiation levels f rom this accident? QUESTION 4.02 (3.00) Per PMP 4050.029.004, " Irradiated Fuel Handling Accident in Spent Fuel Storage Area", what are the three immediate actions that control room personnel must take during a fuel handling accident in the spent fuel ctorage area? QUESTION 4.03 (1.50) What are the immediate actions of OHP 4023-ECA-0.0, " Loss of ALL AC Power? Give major step headings only. QUESTION 4.04 (3.00) Answer the f ollowing per OHI-4014, " Conduct of Operations: Valve Lineups."
- a. What are FOUR acceptable means of verifying valve position during e valve lineup if you are the independent verifier? (2.0)
- b. How is the position of a sealed throttled valve verified? (1.0)
DUESTION 4.05 ( .50) TRUE.or FALSE? Per DHI-4013, "Auhorities and Responsibilities", in the event of an ! unplanned reactor trip or unschedul ed or unexplained change in power, it is permissible for personnel in training to manipulate the controls of the reactor provided they are under the direct supervision of a licensed reactor operator. l l 4 l 1 (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****) f i i i 1
g+ , n - 1 0"/e2_aEBDGEDWBEE:_NpBoeLi_epugBDebi_EDEB9hNQy_eNp: PAGE 14! 88DIDLD91G9L_GDNIBQL'
- 3,
'I L ! f '.
QUESTION: 4.06 (2.50)
.PERjo1-DHP 4023.FR~S.'1, " Response to Nuclear-Power Generation /ATWS", what
~
cre the five immediate actions? (Only - major headings of : the action steps '
- cre required.)
rQUESTION:- '4.07. ( 2.25) LPar~1-OHP 4022.012.003, " Continuous Withdrawal of a Control Bank", what f.ssume rod control'is in, are: the 1three - i mmedi ate : manual' acti ons? Ecutomatic.- DUESTION 4'. 08. (2.00) LPer Unit 1 Technical' Specifications, complete the f ollowing statements: (.5 each)
'a. 1 Minimum- boron concentration in .ttus RCS during ref ueling operations.
- shall=be _______ ppm.
.. b . ECCS accumulator-nitrogen cover gas pressure shall be at least
_______ psi 9-
- c. Primary containment i nternal pressure shall not exceed _______ psig.
- d. JThe maximum ice condenser' ice bed temperature shall be less than or !
equal to _______ degrees F. l' I l l. l\. I I l 1 (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****) I
- f. . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ I
- n. er
' C .',$AL_ES9GE90 BEE _:_N9BD962 9BN9BD962_EDEBEENGY_9Np PAGE 15'
;B99196991G96_G9 NIB 96
'DUESTION 4.09 (1.25)
'You are given the f ollowing summary of critical saf ety f unction (CSF)
'ctatus trees while in 02-OHP.4023.E-1, " Loss of Cool ant ":
- 1. Subcri ti cal i ty Green i
- 2. Heat Sink. Ysilow FR-H.1 3.- Containment Red FR-Z.1 l
- 4. - : Integrity Orange FR-P.1 ,
1 S. : Inventory : Yellow FR-1.1 l Core Cooling Yel l,ow FR-C.3 I 6. List the. order'that these procedures would be performed in if all the procedures were to be used. i 1 QUESTION 4.10 (1.50)
'Per 2-OHP 4022.005.001, " Emergency Boration", indicate whether or not smergency boration is required to be perf ormed f or each of the f allowing
( s. 25 each) conditions,
- a. 1-control rod-fails to insert on'a reactor trip
- b. during a reactor startup, a steam dump failure results in a cooldown to 541 degrees F c.- reactor power is 100% with control bank D at 78 steps
- d. deta.I is at -17. Target delta I is -3 e.- 2 steam generators become faulted with RCS temperature at 551 degrees f ollowing' a reactor trip
- f. the reactor is in mode 3 with shutdown margin at 0.5% delta K/K
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
o,- n ,
."4,. ' NORMALt' ABNORMAL t_EMERQENCy_ANQ PAGE '1<6
-PROCEQURES
- RQQlg6QQlqQL_QQNIROL; QUESTION 4.11 (3.00)
PMI-2110, " CLEARANCE PERMIT SYSTEM",' states'that-a clearance permit must Lbo obtained.for:any work which, for the safety of personne11or protection of equipment., .. requires a . certain definite arrangement or positions of '
' controls,. circuit breakers, switches, valves, etc.
Explain the two exceptions to the'above statement when a clearance' permit
'ig NOT required. Include any other limitations required. -l j
'DUESTION' 4.12 (2.50)
OHI-4012,. " Conduct'of Operations-Shift Turnover", requires the'on-comng Raactor' Operators.to perform eight' tasks pilor to discussing conditions of the unit with the off-going-Reactor Operators.-
.c. List FIVE ofJthe eight tasks to be perf ormed by on-coming Reactor-Operators. '(1.25) i
, b.- List FIVE of the seven items that should be included in the discussion wi th the of f going Reactor Oper ators. (1.25) i i
i. f L , L t' l (***** END OF CATEGORY 04 *****) I (************* END OF EXAMINATION ***************) l l-
.PAGE 17
!1b__E81NQlE(gg_QE_Nyq(g@BiEQWE8_EL8MI_QEg6@I1QNt
. IUE60QQYd@diq@g_dEGI_IB8N5EE6_@NQ_E(Q1Q_E(QW
-87/OB/07-DAMON, D.
ANSWERS -- COOK 1&2
-ANSWER 1. 01 l (2.50)
Tave decreases since.more energy is being removed. (0.7) Rx Power ' increases due to the positive reactivity added through MTC. (0.5)
. doppler would add negative reactivity. (0.3)
' Power' stabilizes at a higher value. (0.5)
'Tave stabilizes.at'.a lower value. (0.5)
REFERENCE W Thermal-Hydraulic Principles and Applications to the PWR,
.pp 1-9 to 1-12 W Reactor Core Control for'Large PWRs, pp 3-29 to 3-41 192OOBK117 192OOOK121 ...(KA'S)
ANSWER 1.02 (2.50)
- a. 2 (0.5) ;
b.-Sm-149 has a smaller absorption cross section and therefore less reactivity worth than Xe (0.5), and does not' decay away like Xe (0.5). (Will also accept explanation that half-lif e is much greater than xenon and reactivity effect is much slower ~ than xenon)
- c. Case I would be a more noticeable Xe transient (0.2) because the local power chances occurred rapidly with respect to Xenon's ability to maintain equilibrium with local power. (0.8)
REFERENCE W Reactor Core Control for Large PWRs, pp 4-21 to 4-23, 4-28 to 4-34 192OO6K106 192OO6K112 192OO6K122 ...(KA'S) L _m_.___________ _ _ _ _ _ _ _ _ _ _ }
)
;1L , PRINCIPLES OF NUCLEAR POWER PLANT OPERATIQN t PAGE 1G
-IUE6dQ QYN@diq @ t _Hg61_lB8N @E gB _@NQ _E(ylp _E(QW p;
~ ANSWERS -- COOK 11&2 -87/OB/07-DAMON, D.
L ANSWER' -1.03 (2.50)- l o. 2185 psig =L22OO psia (.25) 1Tsat for 2200 psia = 649.5 F-(0.5) n .. SCM= Tsat-Th= 649.5-580= 69.5 F +/- 3 F (.25)
.b. decrease (0.25) 0 - Th increases as' unit delta T increases with' power (0.5)
- c. 'l ( 0. '25 )
(Core delta T during natural circulation will approach fu'll load deltw
~
T. That' is greater thaniin the other 2 cases. <0.5) (accept " core delta :T is larger") eREFERENCE Steam Tables OO2OOOK509 OO2OOOK517 ...(KA'S)
. ANSWER 1 .~ 04 ( 2. 00 )'
- a. B (0.2S) because A will reach critical rod height sooner than B (0.25) thus B will allow its neutron ;
population more time toaachieve a higher'subcritical level' than A (0.5) I
' b '. Same (0. 5) , critical rod height is dependent only upon the' l reactivity characteristics of the core, not on neutron level. (0.5)
REFERENCE < W Fundamentals of Nuclear Reactor Physics, pp B-54 to 8-57 192OOOK104 ...(KA'S). A - - - - - - - - . - - - - . .
~ li;_EBlNQlE(g@ ige _NQQ(ESB_EQWEB_E(8NI_QEgg@l1QN s PAGE ~ 19 -
CINE 80QQ1N0dlG@u_SEGI_I8@NSEg8_@NQ_E(Q1Q_E(QW tc _ " ANSWERS -- COOK 1&2 -87/OB/07-DAMON, D. i (1.50) i ANSWER -1.05 compute the reactivity _ represented by the stable SUR: rho = beta eff/ '1 + lambda bar( 26.06/sur) (0.5) l
= O.005/ 1 + 0.08 (26.06 / O.5) (0.5)
=.96.7 pcm (0.2)
# Steps =:96.7 pcm/ 5 pcm/ step
= 19 steps -(+/ ' 1 step) (0. 3 ) -
(other varittions of the formula are acceptable as long as work method is correct. ) REFERENCE 'i W Fundamentals of Nuclear Reactor Physics, pp 7-12 to 7-17, 7-19 to 7-22, 7-44 to 7-46 . i OO1000K547 ...(KA'S) 1 i ANSWER 1.06 (2.00) O = U1 A1 (Tavg - Tstml) = U2 A2 (Tavg - Tstm2) U1 a U2 (.25) A2 = .95 A1 (.25) (Tavg - Tstml) = .95 (Tavg - Tstm2) (0.5) i (573.5 - 513.8) = .95 (573.5 - Tstm2) 573.5 - 59.7/.95 = Tstm = 510.6 F (0.5) From steam tables Pstm = 748.5 psia +/-5 psia (0.5) REFERENCE ; Steam Tables W' Thermal-Hydraulic Principles and Applications to the PWR, i pp 5-16 to 5-22 041020K502 ...(KA'S) I i
"7 , , '
l:.. W." y. '.. . .
'PAGE.! 2OJ J.ji(J[__ PRINCIPLES'OFJNUCLEAR POWER' PLANT OPERATION g ~
lIMEBUQQYN90lGEg_d[91-TRANSFER;ANQ,E(ylg_E(QW
. m; .
-87/OB/07-DAMON, D.-
LANSWERS --'C'OOK 1&2, m s: xe _f
'ANSWERJ ~ 1. 07. t (2.00) i P,, ..o.' t = .( (l og (P/PO) )- / SUR P=100% (.125) PO=20% -(.125) SUR=.5 DPM
(.75)
.t = (l og' 2) / . 5. = . 301/. 5 = . 6' mi n or 36 sec +/- 1 sec
- b. No.-(.25F Power escalation' is a log f unction and theref ore -increases att an increasing: rate.- . (.75)
. REFERENCE
' W Fundamentals of. Nuclear Reactor Physics, pp17-12 to'7-17 OO1010K537- ... (KA'S)
' ANSWER, 1.08 (2.50)
- a. '1. When reactor. coolanto expands into. the pressuri zer ' the vaporJ volume
~
This is compressed and the pressure of.the steam increases.' causes steam-to condense unti1~the'new set of equilibrium conditions is reached. ' (0.75). , 2.fDuringlan outsurge"the level' decreases and'the steam volume
' increases resulting in a system pressure decrease. This'results:in increased boiling, generating more steam to stop the pressure decrease and establish new equilibrium conditions. (0.75)
- b. 1. FALSE
- 2. FALSE (U.S ea.)
' REFERENCE W Thermal-Hydraulic Principles and Applications to the PWR, pp=4-69 to 4-70 010000K501 ...(KA'S) i
)
i 2 n_._.____.._
t PAGE 21 ! . le__EBING1 ELE @_QE_NQQ(EQB_EQWEB_ELONI_QEEB811QN ISE SdQ QYN @d lC @ t _dg 81_1686 @EE8_6N Q _E(Q 1Q _E(Q W l ANSWERS -- COOK 162 -87/08/07-DAMON, D.
- i ANSWER 1.09 (2.50)
(a) 55/60 x 130 = 119 million pounds per hour (flow is proportional to (0.75) pump speed) (b) (55/60)**2 x 90 = 76 PSID (pressure drop is proportional to speed squared) (0.75) (c) 3390 Kw (Power is proportional to speed to the third power (55/60)**3 (1.0) x 4400 = 3390) REFERENCE W Thermal-Hydraulic Principles and Applications to the PWR, pp 10-32 to 10-37 OO3OOOA203 191004K105 ...(KA'S) ANSWER 1.10 (1.50) (a) 1 (1015 pcm) (.75) (b) 2 (linear increase) (.75) REFERENCE W Fundamentals of Nuclear Reactor Physics, pp 5-22 to 5-25, 8-16 to B-19 8-54 to 0-56 OO4000K507 OO4000K508 ...(KA'S) ANSWER 1.11 (3.00)
- o. M: The ratio of the total number of fission and source neutrons to the total number of neutrons which would exist due to the source only.
(Will accept any valid description of the multiplying ef f ect on source (0.75) neutrons.)
- 6. As Keff approaches 1, M approaches infinity and cannut be used to predict criticality graphically. (0.75)
= 100(5) = 500 cps (1.0)
- c. CR = 100(1-0.95 / 1-0.99)
EEFERENCE W Fundamentals of Nuclear Reactor Physics, pp B-13 to 8-14, 8-27 to 8-28 192OOBK106 ...(KA*S)
PAGE s22 j' . 2c__EL ANI_DEE10N _1NGLVQ1NQ _E8EEIY_8ND _EdEBEENGY_EYEIEd5
-87/OS/07-DAMON, D.
- ANSWERS -- COOK 1&2 a
ANSWER 2.01 (1.50?' i Volves will' auto close on:
' ' ( Any 3 of; the f ollowing 4)-
containment isolation signal. (0,5) 1
- 1.:
(Credit .25'if answer is close on SI) All' (.25), charging pump breakers open. (.25) 1
- 2. '
3.- ' Letdown isolation valve.(QRV-111 or 112) shut. (0,5)
'4. Low pressurizer level ( . 25) .177. (.25) 1 REFERENCE RD-C-NSO6, Obj ective #4 RD-C-NSD6,-pg8 of 33 OO4020K401 ...(KA'S)- ..
yANSWER 2.02- (3.00)
- a. To open, RCS pressure must be'less than 375 +/-22 psig for Unit 1(0.5)
~
or?less than 425 +/-15'psig for' Unit 2-(O.)). If open, it'will auto-iclose if RCS pressure is greater than 600 psig.(0.5). b.- ICM-305 (RHR suction , f rom containment sump) open. (0.5) and , Either IMD-262 shut (0,5) or IMD-263 (SIS recire valves) shut (0.5). REFERENCE l RD-C-NSOB, Dbjective #6
.RD-C-NSOB, pgs 19 and 23 of 39 OO5000K401 OO5000K407 OO5000K411 ...(KA'S) l t
)
l E i L l i i J
~ ~
..]
PAGE 23 L " J2ru_6LONI_pggl@N_JNgbypjNQ_gBEEIy_9ND_gDEBgENgy_SYSTEDS
-87/08/07-DAMON, D.
ANSWERS -- COOK'1&2
, ANSWER 2.03 (3.00)
High. Head: RWST.- IMO-910/911 (RWST to CCP suction valves) - CCP's - BIT Inlet Val ves - BIT - BIT Outlet Valves - Cold Legs LMedium Head: RWST IMO-261 (RWST to SI suction valves) - SI pumps - Hot or Cold Legs
. Low Head: RWST - IMO-390 (RWST to RHR isolation valve) - Suction Valves - RHR pumps - RHR HX - Discharge Valve - low head injection lines - hot or cold legs
(.15 ea component, 3.0 total) REFERENCE RO-C-NS12, Objective #8 RD-C-NS12, pgs. 10 to 16 of 28 OO6000K103 OO6000K108- ...(KA'S) .' ANSWER 2.04 (1.50) (Any 2 of the following 3): (.75 each)
- 1. Keep piping, val ves, and spray nozzle hot to minimize thermal shock to PZR and internal components.
- 2. Circulate PZR water with RCS to prevent chemical / temperature stratification (maintain unif orm chemistry and temperature in PZR). :
^
- 3. - Ensure PZR and RCS boron concentration is the same.
l l
' REFERENCE RO-C-NSO3, Objective #10 RO-C-NSO3, pg. 11 of 53 010000K401 ...(KA'S) 1
! i l- i l l l L_____--_-_---___ -
'PAGE '24 h
.p__P(gNI_DESl@@_lNC(UDIN@_@@EGIf_6ND_EdE6@[N9Y_@X@I$df
. ANSWERS 1-- COOK 1&2
-87/OB/07-DAMDN, D.
i
. ANSWER. 2.05 (1.50)
-c. JUnit 1 West ~ spray' pump' T 11 A ,
' East spray pump T 11 D Unit 2 West spray pump T 21 A East spray pump T 21 D (0.251 pts each)
'b. Containment recirculation sump (0.5)
' REFERENCE RD-C-NS15, pg. 7 of :21 RO-C-NS15, Objecti've #4a l RD-C-NS15, pg. 12 of 21-RD-C-NS15, Objective #s 2 and 3
'026000K201' 026000K401 ...(KA'S) f ANSWER 2.06 (1.50)
( Any 3 of;the f ollowing 9 0.5 pts' each) -j
'1.
Ice condenser recorded ice temperature high/ low
- 2. Ice condenser equipment access doors open {
- 3. Ice condense inlet' doors open f
- 4. Ice condenser lower access door open 1
i
- 5. Ice condenser air handler abnormal REFERENCE
, .RO-C-NS14, pgs. 16 and 17 of 22 l j
RD-C-NS14, Objective #6b l 2RORO25000 ...(KA'S) 1 I l 1 i_i__. _ __i __
PAGE 3M5 ,.' ( Z e2 LEL8NI_QEl1GN_INGbWQ lng;SeEgIy _ egg _gdgR@ENQY_EY@l5d@ i
-8'7/OS/07-DAMON, D..
JANSWERS --l COOK 1&2, p l1 l ' b' ANSWER ' 2. 07.' '( 1. 50 ); 1.. Both unit - vent radiation monitors must not be alarming.
- 2. . / At ' l east ' 1= Uni t 1; Auxiliary Building exhaust. f an ' cont' r ol switch must'belin:the closed position.
-(.75 pts each)
REFERENCE RD-C-ASO7,.pg. 8 of 21 RD-C-ASO7, Objective'#4 071000K106 ...(KA'S) ( 1. 00 )-
. ANSWER 2.0B-
~
r a. , Hi gh' acti vi ty alarm on . R-18. - (0.5)
.b. Noncirculation water pumps running'on the appropriate' unit's circulation water system. (0. 5) -
' REFERENCE RO-C-ASO6 Objective-#4 RO-C-ASO6 pg 18 of'19
,068000A302 068000A404 ...(KA'S)
-ANSWER 2.09 (1.50)
( A',, / 3 of the f ollowing 9 0.5 pts each)
- 1. cSFP water level high
- 2. - SFP water level low 3.; SFP temperature high
[4 . , Ref ueling water purification pump f ailure (breaker trip)
- 5. SFP north' pump failure (breaker trip) 6.- SFP south pump. failure (breaker trip)
REFERENCE RO-C-ASO3, pg. 12 of 14 RD-C-ASO5, Objective #6 1
. . - . . . _ _ _ _ . _ _ . _ _ l
2e__P60NI_ DESIGN _lNQ(QQLG@_@dEgly_QNQ_gd[R@ENQY_EYEIEd@ PfGE 26
" ANSWERS - .COOKL1&2
-87/08/07-DAMON, D.
033OOOA203 033000G008 033OOOG015' 033OOOK303 ...(KA'S) ANSWER. 2.10 (1.50)
- 1. . .All ESW' pumps start
- 2. -ESW outlet valves'from affected unit CCS Hx throttle to mid position'-
- 3. Containment Spray Hx outlet will open.
~
(O.5 pts each) REFERENCE RO-C-ASO2,.pg.. 12 of 14. RD-C-4SO2, Objective M4 076000A302- 076000K106 076000K407 ...(KA'S) i ANSWER 2.11 (1.50.
- 1. . Saf ety' Inj ection 2.- Blackout
- 3. Low header pressure (0.5 pts each) ,
i REFERENCE RD-C-ASO1, pg.'10 of 13
'RO-C-ASO1, Objective #4 OOBOOOK401 ...(KA'S) ' ANSWER 2.12 (2.00)
- 1. That units' Condensate Storage Tank
- 2. Opposite unit CST
- 3. ESW System Noter ~ no credit will be given or deducted for " makeup to unit CST from I
hotwell s" or ~ " makeup to unit CST from make-up system". .25 deducted for cny additional source of water listed to the unit CST. l (.5 pts each answer, .5 for order) L ! i l l-
I," q PAGE- 27
*: .2ii_ELONI_QEE10N_INQ(QQ1NQ_E9EEIY_9NQ_EDE8@ENGY @y@lgd@ .
-87/08/07-DAMON, D.
ANSWERS -- COOK.1&2-1
' REFERENCE' y -.
RO-C-AS11, pg. 13 of 27' RD-C-AS11,lDbjective.#6 061000K107 061000K401 ....(KA'S) ANSWEi! 2.13~ (1.50) [ l l (Any 3 of the f ollowing & O.5 pts each)
- 1. HEA !
- 2. - D/G' start switch through D/G run relay
- 3. - D/G breaker overcurrent
<- D/G over ground current' lockout relay -
5, . SIS if - D/G paralleled and no blackout 1
-6. Emergency.Stop Pushbutton REFERENCE RO-C-ASIO,; pg. = 29 of 37-HD-C-AG10" Objective #19 ,
_064000K402 ...(KA*S) l I i 1 1 l l 1 I
. $ ~.
Es__P(ANI_DEG1GM_l@C(UQLNG_SQEEIY_6dQ_EdERGENGY_Sy@IEdg. .PAGE 28- .; y
- ANSWERS -- COOK l1&2 -87/OB/07-DAMON, D. )
j ANSWER 12. 1 4 (2.50) s ( Any 10 of ' the f ollows .125 f or each load; .125 for header designation) 1
- 1. Main Turbine Oil Tank Coolers - Both
'2.: ' Feed Pump Turbine EHC Fluid Coolers -
Unit 2
- 3. Feed Pump Turbine LO Coolers -
Both
- 4. Main Turbine EHC Fluid Coolers - Both I 5.- Upper Containment Ventilation Units -
Both
- 6. Lower: Containment Ventilation. Units -
Both
- 7. Instrument Roem Ventilation -
Both
'B. Tech Support Center Ventilation Units -
Unit 1
- 9. Control _ Air Compressor Coolers - ' Both ,
.10. Plant Air Comprensor Coolers -
Both
- 11. RCP Motor Air Coolers -
Both
- 12. S/G-Blowdown Heat Exchanger -
Both
- 13. Secondary Side Sample Heat Exchangers -
Both
- 14. S/G Blowdown Tanks -
Both
- 15. Auxiliary Feed Pump Bearings -
Both - i i
- 16. Ultrasonic Generator Cool er -
Unit i
- 17. Generator Seal 011 Coolers - Unit 2
' 18. . Blowdown Drain Suppression Pots -
Unit 1 REFERENCE l RO-C-ASO3, pgs. 4 and 5 of 15 RO-C-ASO3, Dbjective #1 ' 076000K109 076000K110 076000K118 076000K120 ...(KA'E) l L L n 4 I i
PAGE 29 i .gg ,JNgISgDENIS_gND_ggNIBgbg
-87/OB/07-DAMON, D.
I ' ANSWERS -- COOK 1&2 f ANSWER 3.01 (1.25) Containment pressure (0.5) greater than 2.9 psig (0.5) 2/4 coincidence (.25) REFERENCE RD-C-NS15, pg. 17 of 21 RO-C-NS15, Objective #10
^026000A301 ...(KA'S)
ANSWER 3.02 (2.SO) i (Any 5 of the following)
- 1. Power range NI overpower (C-2) (.25) greater than 103% (.125) 1/4 channels (.125) ;
- 2. Intermediate range NI overpower (C-1) (.25) greater than corrent equivalent to.20% (.125) 1/2 channels (.125)
- 3. OP Delta T (C-4) (.25) greater than 3% below trip setpoint (.125) j 2/4 channels (.125)
- 4. OT Delta T (C-3) (.25) greater than 3% below trip setpoint (.125) 2/4 channels (.125)
S. Urgent f ailure in rod control. (0.5)
- 6. Cont r ol Bank D withdrawal limit (C-11) (.25) 230 steps (.25)
- 7. Turbine power (C-5) (.25) less than 15% (.125) 1/2 impulse channels (.125)
REFERENCE RD-C-NSO4, Objective #7 RD-C-NSO4, pgs. 11 and 12 of 24 OO1050K401 ...(KA'S)
..n
' PAGE 30 m,fli_rNSIggdENIS_669_GQN1696S_
COOK'1&2' -87/OB/07-DAMON, D. '$$ ' ANSWERS - ANSWER, 3.03 -(3.75) Unit 1 Trip'could be caused by: Turbine trip , system low ' pressure (0.5) setpoint-BOO.psig (.25) Steam Flow-Feed Flow mismatch coincident with low -level on A steam generator'(0.5).
'Setpoint -~.6xE6 lb/hr with 26% NR level (.25)
' Uni t 2 Trip' could be caused by:
Steam generator B Low-Low level' (0.5) setpoint 21% (.25) Loss of LFlow Loop A (0.5) setpoint 93% (.25) Steam Flow-Feed Flow mismatch coincident with low level on A steam generator: (0.5)- Setpoint - .6xE6 lb/hr-with 26% NR level (.25)-
.(deduct .5 f or any other trip given that .is not present. Deduct. 25 if trip is given correctly f or the wrong' unit)
REFERENCE' RO-C-NS11, Objective Ms 4, 5
'RO-CeNS11-SHO3, pgs. 1-4 ;
T/S Table 2.2-1 ' 012OOOG012 -012OOOG015 012OOOK402 ...(KA'S) ANSWER 3.04 ( .00) t (del eted ) ' l; l l I 1
]
l I
p , . PAGE. 31- H
" 01i__INEI6WUENIE OND G9NIBQLE
-87/08/07-DAMON, D.
i f ANSWERSf- -COOK'i&2 ANSWER' 3.05 (2.00) F N42 N43 N44 Normalized Current N41 ll
' Upper 0.5 0.5 0.518 0.528 0.5 0.5 0.5 0.5 i Lower
(.15 pts f or each normalized ' current) Upper average normalized current = 0.5114 (.2) ,
' Lower average normali zed current = 0.5 (.2)
Upper; tilt ratio = 1.032 (.1) Lower tilt ratio = 1.00 (.1)
'QPTR = 1.03 (.2)
REFERENCE. RD-C-NSO9, Objective #11 OHP-4030.STP.032' , 015000A104 ...(KA'S) 1 ANSWER 3.06 (1.50)
'a . SIS (.25) 1 Blackout -(.25)
S/G 1evel l o-l o (.25)
- b. Both main feed pumps tripped (.25)
- c. low volts on RCP busses (.25)
S/G 1 eve 1 ~ lo-lo (.25) REFERENCE- j RD-C-ASil, pgs. 14 and 15 of 27 RO-C-AS11, Objective #7 061000K101 061000K402 ...(KA'S) f l F" L f i
a Q .-+L._,1NgIBlLMENIg_9Np_ppNM&g' PAGE 32
.s . , . ..
ANSWERS.--: COOK.1&2- -87/OS/07-DAMON, D. n..
.f
~
iANSWER- .3.07.. -(2.50) Totalf. steam flow will iincrease (.5)' causing' program' delta P to increase; .'.- (.M ,' which-causes MFP : speed to increase -(.5) . ' This will increase; feed ] flow ('.5) , . causing steam ! generator ; level s to remain constant at: program- ?
. l evel . . ( 5) ..
?' REFERENCE- .
. RD-C-PG11,.pg. 12 ofl15 RO-C-PG11, Objective #8 s
.035010A101' 035010K101 ...(KA'S) j .<
ANSWER', 3.08. (2.50)-. i.
' i- f c.' Backups l on atJ2210 psig off at 2218-.psig
, .. Cycl i ng onfat 2220.psig-off at 2250 psig
(.375, pts each) . _,
- b. Unit it' 385 + 22 psig ' '
Unit 2: 420 + 22 psig (.'5 pts each) ,
)
REFERENCE
'RO-C-NSO3, pgs. 13, 14, and 17 of 53 RD-C-NSO3, Dbj ective #s 17e, 179 l 010000A302 010000A403 ...(KA'S) I i
i q
)
.q s
l l 4 L__________ _ _ _ . . _ _ _ _ _ _ . .
V , PAGE 33 p " Ji_.1NSIBVUENIE_9ND_QQNIB965
~
COOK 1&2- -87/OS/07-DAMON, D. ANSWERS - y l.
, ANSWER: 3.09 (3.00)
UNitjla High steamline flow (.25) 1/2-per loop, 2/4. loops (.25)
-at 1.42c6-3.88E6 pph (accept any setpoint in range) (.25)
Coincident with
'1). Low steamline pressure (.25) 1/1 per-loop, 2/4 loops ( . 25) -
at greater than or. equal to 600 psig (.25)
. Ort
- 2) Lo-l o Tave' (.25) 2/4 1 caps ( . 25) ' ,
at 541 degrees F f.25) c Unit 2:- Low..steamline pressure t.25) 1/1- per loop, 2/4 loops'(.25) at 600 psig (.25). REFERENCE RO-C-NS11-SH03, pg. 5 of 9 013OOOK101' ...(KA'S) l 1 1
_ i
PAGE 34 d
.;l' (3 __ANEI6sMENIE_eNg_CQUI6969 ,
ANSWERS -- COOK 1h2. -87/08/07-DAMON, D. i 1
' ANSWER 3.20 (2.00) /
4 ( c. 1. Condenser vacuum greater than.10.6 inches in'all three condensers. - ( . 25)' ;
~
- 2. At.least one circulating pump'breakte closed. (.25)
- 3. ' Power available from CRID II. (.25).
1 b; . 1. C-7A od'C-7B (wil11 accept explanation of these) (.25)
- 2. -Tave' greater.than'541'F (.25) 3j- Ta've. -Tref greater than 5 F (.25) l c.. 1. - Steam pressure greater than controller setpoint (.25), AND
.(either of the following at .25)
- 2. Tave greater than 541 F
- 3. Dypass. interlock if Tave is less than 541 F (For ' b. 2 and c . 2, will accept "no P-12 signal present")
REFERENCE-RD-C-PG12,-pgs. 12 and 13 of 18
'RO-C-PG12, Dbj ective Hs 3. and 4 041020K105- -041020K402 041020K414 041020K418 ...(KA'S) l l
l- l l.
7 , . PAGE' 35-jl__lNEI!MLENJ$_9Np,QQUIRpby -
- ANSWERS -- COOK.1&2 -07/OB/07-DAMON, D.
l j-.
' ANSWER 3.11 (2.00)
- o. .c. P-7 is met if either P-10 is met (.25) or'if'P-13 is not met (.25)
P-10.is met i f ' 2/4 power range channels . (.125) are greater than or equal to 10%. (.125)
! P-13 is'not met if.1/2 turbine impulse channels (.125) is. greater than or equal: to 10%. (.125)
+
- b. L(any 4 of ~ the f ollewi'ng at .25 ea)
Pressurizer Water Level.High Turbine Trip ! Pressurizer Pressure Low Low RCS. loop flow RCP breaker trip i RCP' bus undervoltage,
.RCP bus under requency f
REFERENCE RO-C-NS11, pgs. 10 and 11 of 14 RO-C-NS11-SHO3, pg. 6 of 9 RD-C-NS11, Objective #6 012OOOK406 ...(KA'S) I 1
PAGE~ 36 T4d__E69EEEW6EE_ _N9Bd86t_0BN9Bdebt_EdEBQENEY_eNQ
-bed 196901G96_GQUIB96
- ANSWERS '-- COOK .1&2 -87/08/07-DAMON,- D.
' ANSWER: 4.01 (2.00) 4 (Any 4' of the lf ollowing at 0.5 pts each)
~
J1. Containment purge supply valves (VCR-103, 203, 105, and 205) close.
, 2. ' Containment purge exhaust valves (VCR-104, 204, 106, and 206) close.
3, Containment' pressure relief valves (VCR-107 and 207) close.
<4 . Instrumentation room purge supply valves .(VCR-101 and 201) close.
- 5. Instrumentation room purge exhaust valves (VCR-102 and 202) close.
- 6. Valve RRV-306 for GDT's to Auxiliary Building' ventilation closes.
'7. Containment purge supply (CPS) and exhaust fans (CPX), instrumentation ,
room purge suppl y-- (CIPS) and exhaust fans (CIPX), and containment
. pressure relief fan all stop.
REFERENCE PMP 4050.029.003, pg. 2 of 3. . OOOO36A101 OOOO36G010 034000A201 034000G014 072OOOK101 072OOOK102 072OOOK401 ...(KA'S) ANSWER 4.02 (3.00)
- 1. Announce on the page system that a fuel handling accident has occurred in the spent f uel ' storage area. (1.0)
- 2. Verify all automatic actions (.33), manually initiate any that have not occurred (.34), and ensure that one fuel handling area exhaust fan is running. (.33)
- 3. Notify Radiatiot Protection of the accident. (1.0)
REFERENCE PMP 4050.029.004, pg.- 2 of 3 034000G014 ...(KA'S) 1
.I
'it__88QQgQQBggi;_NQBd@(u_6QUQ8d@(g_gdE8QgNQy_6NQ PAGE 37 l i I 869196QQlg86_GQNTROL (lV COOK 1h2 -87/OB/07-DAMON, D. !'
[; (ANSWERS - ANSWER 4.03 (1.50) L _1. , Verify. reactor trip 21 ' verify turbine / generator trip
- 3. ' check:if'the RCS~is isolated ,
verify AFW flow (.375.ea)-
'4..
REFERENCE:
OHP 4023.ECA-0.0 steps 1-4 OOOO55G010 ...(KA'S) ANSWER 4.04 (3.00)
-(Any'4'oi the f ollowing 9 0.5 pts each)
- s. 1. . Attempt to move the valve in the closed direction.
- 2. Stem position on a. ri sing stem valve.
- 3. Mechani cal position i ndi cati on (ie stem travel indicators, butterfly-disc position indicators,...)
- 4. Remote. position indication (ie limit switch lights, meter position indicators,...)-
- 5. System response (ie flow, pressure, t e'mp er at ur e , . . . )
- b. Verify that theJseal is installed properly (0.5) and that there are no signs of physical tampering with the valve or seal (0.5).
REFERENCE OHI-f014, pg. 2 of 3 194001K101 ...(KA'S) ANSWER- 4.05 ( .50) False.. REFERENCE OHI-4013, pg. 2 of 17 194001A102 ...(KA'S)
I' *
,L., ,.
.I
'e-[Qi;_RROCEDURES: ~NORMALg_ABNQRMAl _QMER@ENQX_ANQ t
' PAGE 38 899196001GOL_QQNI6QL. ,
. ANSWERS 1-- COOK 1&2 -87/08/07-DANON, D.
l' f n ' ANSWER 4.06 (2.50)
' 1 :. Verify' reactor trip
- 2. LVerify turbine: trip 3s . Check'AFW. pumps running q
- 4. Initiate . BIT inj ection 'i S. Check PZR pressure.
(0.5 pts each) ; REFERENCE 1 01-OHP 4023.FR-S.1, Steps 1-5 OOOO29G010 ...(KA'S) ANSWER 4.07 (2.25)
' 1. Transfer reactor control.to manual.
~
2.- If control bank continues to withdraw, manually trip the reactor. 4
~ 3. If withdra'wal stops, manually insert control bank to restore power and temperature.
(0.75 pts each) i REFERENCE
;1-OHP 4022.012.003, pgs. 2 and 3 of 3 OOOOO1G010 001000G014 ...(KA'S)
~ ANSWER 4.08- (2.00) l i
- a. 2000 )
' b . 585
- c. O.3 d.. 27 l
-(0.5 pts each) '
i l
, ?
PAGE1 39
-SJ_,EBQGEQQBEE_:_NQ6 debt _eHNQBdebt_EUE6EENGX_0NQ 889196901GOL_GQNIBQb
~B7/OB/07-DAMON, D.
g ANSWERS ~ -- COOK 1&2-l I
; REFERENCE q Technical Specification 3.9.1, 3.5.1, 3.6.1.4, 3.6.5.~1.c 006000G005 022OOOGOO5 025000 GOO 5 ...(KA'S).
004000 GOO 5 l
. ANSWEN 4.09' (1.25)
FR-2.1, FR-P.1, FR-C.3, FR-H.2, FR-I.1 (3, 4,' 6, 2, 5) (.25 pts f or each procedure in proper order) ( (will accept 2,3,4,6,5 if FR-H.1 is identified as a red. path procedure) REFERENCE' RD-C-E011 objective #4 WDG ERG Executive Volume - Users Guide OOCS69G011 OOOOO9G011 ~O00009G012 OOOO29G011 OOOO29G012 OOOO69G012 OOOO74G011 000074G012 ...(KA'S) ANSWER 4.10 (1.50)
- a. no
- b. no c.- yes
- d. no
,e. yes
- f. yes (.25' pts each)
REFERENCE 2 OHP 4022.005.002, pp 1 & 3 of 5. OOOO24A205 000024K301 000024K302 ...(KA'S)
'$41,L' PROCEDURES"--NORMALi_QgNORMAL _EMERQENgy_QNQ' t
PAGE--40 L> BeQ1960Q1ceL_cQNIRQ('
;sa ANSWERS ---COOK 1&2.
~
-87/08/07-DAMON, D.
' ANSWER 4.11- (3.00)
-s
- 1. Work offan emergency. nature (0.5) such that a delay in obtaining.'a clearance 'would'#rolong termination of an undesirable eventL (0.5) .
Personne1'will: guard all c o'n trol or isolation points (0.5). [2.. ' Minor ' adjustments ~and troubleshooting on energized (pressurized) ' l may be conducted 1(.5) provided it is in accordance:with PMI-2290' l (Job. orders) _ or other approved procedure-(.5) and no personnel or. equipment hazard:is included (.5).
-: REFERENCE PMI-2110,-pg. 2'of 18 i 194001K102 ....(KA'S)
I ( l l :. .
' 4 t_,889GEQQQE@_ _UQBd66t_6BNQBd@bt_EME6@@Ngy_ANQ PAGE 41 d6Dighg@lGOL_GQNI696 ANSWERS -- COOK 1L2 -87/OS/07-DAMON, D.
1 ANSWER 4.12 (2.50) i
- c. 1. Review the Control Room Log.
- 2. Review Non Tech. Spec. Equipment Log.
- 3. Review temporary modifications logs: Lifted Wires, Electrical and Mechanical Jumpers and Mechanical stops.
- 4. Review Blocked Alarm Log.
- 5. Review surveillance schedule.
- 6. Make a thorough inspection of the unit condition, including a control panel walkdown.
- 7. Review Eberline monitoring system to determine status of monitors.
- 8. Review any Standing Orders, operating memos of instructions that have been issued since last reviewed.
(Any 5 at .25 pts each) i
- b. 1. Any problems with equipment or control s. j
- 2. Status of any Tech. Spec. equipment that is inoperable or out of service.
- 3. Any unusual standing alarms.
- 4. Any procedure that may be in progress.
- 5. Any planned change in the status of the unit during the oncoming Reactor Operator's tour of duty.
- 6. Any special tests, projects or evolutions that are in progress or about to be started.
- 7. Any unusual condition on the unit or in the plant that required special attention or consideration.
(Any 5 at .25 pts each) REFERENCE OHI-4012, pgs. 6 and 7 of 11 l 194001A106 194001A113 ...(KA'S) l l l 1 _ - _ - _ _ _ _ _ _ 1
- . o .
ANSW Elt. ~5,09 QUADfuurf POWER TILT RATIO CALCULATION y
. 0,9 N 41 DET A 1 , Sil DIvroED s u 41 DET A 12os I 10 4 Il 2. 0,y
. m 42 DET A I $6 DIVIDED BY N 42 DET A 12o% I .
6f D m DED BY u 43 DET A 12o% I 11 2. - 0,SIf a 43 MET A I 10( O<S2.8 M 44 DET A I 87 DIVIDED BY N 44 DET A 12o% I . UPPER TOTAL 2,0V6 0 , TlIf ave. UPPER UPPER TOTAL _ 2.09L D m DED BY 4 , MAX. UPPER D.T19 DIVIDED BY AVE. UPPER 0,EllT - f.031 UPPER TILT AATIO E3 l06 - 0.S m 41 DET s I DmDED sY w 4:. DET a 120s I w 42 DET a I 57 D m DED nY w 42 DET s 120s I (10 - 0. f x 43 DET a I FL D m DED sY w 43 DET a 120s I Ill - 0.9 w 44 DET B I DIVIDED BY N 44 DET B 120s I 10g - 0E LOWER torAL 2.0 LOWER TOTAL 20 DIVIDED BY 4 = Os I AVE. LOWER . MAX. LOWER 0.f DIVIDED sY ave. LOWER 0.C - /.0 r4WER TILT RATIO l QUADRANT POWER TILT RATIO - 1.03Nax. P-2so oUADRAnT POWER TItT RATIO , l TECHNICAL SPECIFICATION 3.2.4 LIAIT = 1.02 i l' LUKE INSTR # CAL DUE DATE CONDITION REPORT REQUIRED Yes No PERFORMED BY DATE TIME REVIEWED BY DATE SRO Page 1 of 1 Rev. 2
U. S. NUCLEAR REGULATORY COMMISSION
/ no%*, SENIOR REACTOR OPERATOR LICENSE EXAMINATION f- }i FACILITY: COOK 1&Z b l REACTOR TYPE: PWR-WEC4 i
% ....u /
DATE ADMINISTERED: 81ZQ8407 EXAMINER: EAEE, S. _ 4 CANDIDATE INSIRMITIONS TO CANDIDATE: Use separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF
. CA"EGORY % OF CANDIDATE'S CATEGORY
__YbkME_ _IQIAL bQQBE _VALUE__ CATEGQBY _23.75 24.36 S. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS,AND THERMODYNAMICS _B5.60 _ZE 15 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 24.75__ _gS m3e 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL _22&50 _ Hit 1Q 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS _gluSO % Totals Final Grade All work done on this examination is my own. I have neither given nor received aid. Candidate's Signature t TASTER COPY
(. ____ _ t NRC. RULES AND GUIDELINES FOR LICENSE EXAMINATIONS-During the' administration of'this examination the following rules apply:
' 1 '. Cheating;on the examination means an automatic denial of your-application and.could result in more severe penalties.
- 2. Restroom; trips are to be limited and only one candidate at'a time may
' leave. You must avoid'all' contacts with anyone outside the examination-
. room to avoid evenLthe appearanceLor possibility of. cheating.
13, 'Use black. ink or dark pencil only to facilitate legible reproductions. 4 Print your namenin the blank provided on the cover sheet of the examination.
'5. FillEi'n'the date on the cover sheet of the examination (if necessary).
6; UseLonly the: paper provided for answers. l
- 7. Print.your'name in'the upper right-hand corner of the first page of each section of the answer sheet.
8; Consecutively number each answer sheet, write "End of Category __" as appropriate, start.each category on a new page, write only on one side of.the paper, and write "Last Page" on the last answer sheet.
- 9. Number each answer as to category and number, for example, 1.4, 6.3.
- 10. Skip at least three lines between~each answer.
11'. Separate. answer sheets from pad and place finished answer sheets face-down on your desk or table.
- 12. Use' abbreviations only if they are commonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the question and.can be used as a guide for the depth of answer required.
l l
- 14. Show all~ calculations, methods, or assumptions used to obtain an answer
-to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE l
-QUESTION AND DO NOT' LEAVE ANY ANSWER BLANK. l 1
' 16 . If parts of the examination are not clear as to intent, ask questions.of the examiner only. j 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 1 completing the examination. This must be done after the examination has 1
.been completed.
1 _ _ _2__
ulbl When you complete your examination, you shall:
- a. Assemble your examination as follows:
(1) Exam questions on top. (2) Exam aids - figures,. tables, etc. 1 (3) Answer pages including figures which are part of the answer.
- b. Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn in al.1 cerap-paper and the balance of the paper that you did not use for answering the questions,
- d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
l l l l
i t
;5. THEORY _QE_HUCLEAR POWER'PLPJ:i OPEBATIOHu Page: ~4 iEkUIDS.AND_IBERMODYN6MICS. a l
1 . 1 3
'b 1 QUESTION. 5;01' (1.50) 1 Answer (ther following questions-1 j
- '( a )'; How does. Beta Bar change (increase, decrease, stay.the same), if C tak all, if: temperature.is raised.5 degrees F in'a short period 1 of time? ~ [0 25 ) Why-? '[0.50.') ']
(b). Which' condition would 1'sult in.a higher.startup; rate; a. rod- [0.25 ] Explain. .[0.50] ejection' accident at BOL'or EOL? l QUESTION 5.'02 ( 3 .. '0 0 ) .
'AnswerEthe following: questions:
(a). If core' power is increased from 50% to 100%,.how, if at.all, will differential rod worth change (increase, decrease, stay the
- same)lfor the following 3. cases? Why?
~
(1). Rod position and boron held constant, and temperature allowed to decreaue [0.'75'] (2).. Boron. constant, Bank D at 150 steps and is'then' fully
. withdrawn', temperature remains constant. [0.75 ]
(3). Rod position constant, boron dilution used, temperature constant. [0.75-] (b). Indicate whether the following. statement is either TRUE or FALSE [0.25 ); and explain your answer [0,50 ]: The differential rod worth'of Bank D rods at the moment of criticality during a startup will be the same as when the rods are at the identical height'during power range operations. i (***** CATECORY 5 CONTINUED ON NEXT PAGE *****) 1 4 1 l
- 5. 'THEQBi_QE_MMGLEAR POWEB PLANT OfEH6IIONx Page L5 FLUIDS AND THERMODYNAMICS
^ QUESTION 5.03 (2.00)
For the following 3 cases, state whether the reactor is critical or if criticality cannot be determined. If criticality cannot be determined, state the action that should be taken to determine, if the reactor is critical. ACTION Case 1 Case 2 Case 3 Rod motion in progress Yes No No Boron dilution in progress No No Yes SR level status (CPS) 4 EOS 4 E04 3 EOS increasing increasing increasing SUR status (DPM) +0.4 +0.3 +0.0 oscillating oscillating oscillating slightly slightly positive Clarification for Cases 1 and 3: The reactor is not tripped. Disregard SR trip and answer from a strictly theory standpoint. QUESTION 5.04 (1.50)- After Unit-1 has operated at 100% power for three weeks, it experiences a reactor ~ trip. Assume that no heat (e.g., no S/G safeties lift) is removed from the primary coolant AFTER Tave stabilizes at its no load value and the following parameters are constant. Calculate the length of time it will take for pressurizer safety valves to lift if primary pressure is maintained at exactly saturated conditions and no boiling occurs. Mass of water in the RCS = 214,000 lbm Cp = 1.2 BTU /lbm-degree F RCP heat input to primary = 15 MW Decay heat is 5% of full power QUESTION 5.05 (0.00) Deleted from examination. , I f I f ! I i (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) i I 1 l
.. 5. .. THEORY: OF NUGLEBR POWER ~ PLANT OEEBAIION. Page'
~
6 T' y ELMIDS_AND_-THEBdQDYNAMICS a u :l 4 l
.[.e
!' '-QUESTION. -5.06 -(1.00) 1
.Whyfdoes the critical boron concentration' decrease at'a much more rapid rate #at EOL:than11tEdoes at-BOL?-
H
' QUESTION 5.07 ' ( 1. 50 ) .
JWhat?is1 REFLUX-BOILING'and.under-what' core /RCS conditions would core
. cooling-be'provided by this mechanism? 1
- QUESTION 5.08' -(2.50)'
.i
.You are operating at 100% power with RCS Tave at 567 F and a steam. y
? pressure of 765:psig. . What must Tave be changed toxin. order:to maintain 1 these conditi~ons with 20% of the tubes in each steam generator' plugged ?
HShow all' work,' including.any applicable formulas. QUESTION '5.09 .(2.00) Pipe elbows are used'for flow measurements of the RCS. Describe how pressure' differences occur at the elbows and state how' differential pressure islrelated to flow. QUESTION 5.10- (2.75) la. Per' Technical Specifications,. define FQ(Z). (0.75)
- b. List the four conditions that ensure that FQ(Z)-is maintained within
-limits. (2.0) -)
i (***** CATEGORY S CONTINUED ON NEXT PAGE *****)
'h; THEQBY OE_HUGLEAR POWER PLANT OPERATION t Page 7 ELUIDS.AND THERMODYNAMICS QUESTION 5.11' (1.50)
You are in the process of starting up the plant with reactor power at 10E(-8) amps. The reactor operator inadvertently moves the control rods
'IN for 20 steps. Assume rod worth is 8 pcm/ inch, an average neutron precursor decay constant of 0.05 sec-1, and a weighted average delayed neutron fraction of- 0.00596; with no further operator action:
- a. What is the resulting Start-Up Rate immediately after rod motion stops? (Show all work) (1.5)
- b. Deleted from examination.
QUESTION 5.12 (1.00) During operation, indication for charging flow rate is lost due to a transmitter failure. Explain how the operator can accurately evaluate the charging flow rate using a calculational technique. QUESTION b.13 (1.50) Shutdown margin requirements change from the beginning of core life (BOL) to the end of core life (EOL). Does the required shutdown margin increase or decrease from BOL to EOL? (0.5) Explain why the required shutdown margin at EOL is different than BOL? (1.0) Clarification: TS Bases analysis says required SDM changes over core life. Discuss this analysis. QUESTION 5.14 (1.00) State how the following occurrences would affect Net Positive Suetien Head available to the Reactor Coolant Pumps. (Increase or Decrease) Assume that no operator action occurs and consider each occurrence independently and separately. Also assume steady high power operation.
- 1. Grid-frequency decreases to 59.6 HZ. (.33)
- 2. Pressurizer temperature increases. (.33) l
- 3. Turbine power increases slightly with rods in manual. (.33)
} l l (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
5 __IBEQRY OF BU_QLEAB POWER PLAN.T__OEEBATIOlh Page- 8 ELUlEAUD THEEMODYNAtiLQS i ( .... QUESTION 5.15 (1.00)
- At E01 after a power transient, xenon is oscillating axially in the core. With the control rods all the way out, the operator can force a very-negative delta-I toward the positive side by borating. Explain this phenomenon. .
i 4 1 i (***** END OF CATEGORY 5 *****)
Em_Ekatil_EYSIEtiS_DEnlGth COMIBQLudHD_INSTRUMEHI&IlQH Page 9 QUESTION 6.01 (3.00) Answer the following concerning the CCW system: :
- a. -List four loads that are included in each safeguards train. (0.5)
- b. What is the purpose of the baffle in the surge tank? (0.6) l
- c. Where is the location of the radiation' monitor that isolates CRV-412 (Surge Tank Vent). (0.5)
- d. What are three automatic pump start signals (setpoints, if appropriate) and which control switch position (neutral, auto, lockout) is required for them to be effective? (1.5)
QUESTION 6.02 (1.26) Concerning the Rod Control and Rod Position Indication Systems, complete the following statements. ( 25 each)
- a. The ________ generate the required sequence of current orders to the CRDM coils for rod motion.
- b. A logic ~ cabinet occurs when any one of two 120 VAC power supplies in the logic cabinet fails,
- c. Normal speed for manual operation of the shutdown banks is _
SPM.
- d. If a rod does not move when demanded, the step counter will/will not (choose one) reflect its true position.
- e. A Rod Position Deviation alarm is generated if a deviation of +/-
steps occurs between an" rod and its bank demand or between any two rods in the same bank. QUESTION 6.03 (2.75)
- a. List eight conditions that will cause either main feedwater pumps )
to trip,
- b. What three actions must be taken to reset the resultant turbine trip.
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) l l \ \ \ u_---__--_-_------- l
1 p .. ._ELABI._MMEtLS._DEMGN . COMIBOL. AND INSTRUMENTATION Page 10 QLcSTION 6.04 (1.50) Ten minutes after a reactor trip, the Intermediate Range Detectors decay I to 10[-11] amps,
- a. Is the detector likely to be overcompensated or undercompensated?
(0.5)
- b. What is the. purpose of the compensating voltage? (0.5)
- c. Briefly explain TWO potential problems you might expect during a subsequent startup associated with these intermediate range detectors !
as you pass through the source and intermediate ranges. (Assume no corrective action had been taken for part a.) (0.5) i . QUESTION 6.06 -(1.50) Answer.the following regarding the HYDROGEN RECOMBINERS:
- a. By what method is containment air supplied to the inlet of the recombiners? [0.5)
- b. Explain briefly how the recombiner reduces the containment hydrogen concentration. [0.F]
- c. State the functions of the " Temperature set control" and " Power adjust potentiometer" controllers. -[ 0 . ' 5 1 o,83 QUESTION 6.06 (2.00)
-Concerning the P-4 permissive:
- a. How is P-4 generated?
(.8)
- b. List any four functions performed by P-4. (1.2)
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
g 6; 2 L6dliSYSTEMSj ESIGNi3Q8IBOL."AHQ_INSIBUBENI&II,QH Page .'11: i f-g.
' QUESTION ~ 6.07 (1,00).
'The plantlistoperating at 80 percent power with all' systems in' automatic when the LOOP.IL(controlling) pressurizer' pressure channel FAILS'HIGH. 'l
'AfterLcompleting all; required operator actions, the= plant is stabilized e ~ withi an' alternate pressurizer pressure-channel selected for control'and Lthe ' appropriate bistables: tripped. At this time, the~ LOOP IV T-hot ~
, instrument. FAILS HIGH.
Explain what' automatic actions occur as a result of thisLadditional failure; . QUESTION 6.08 (2.00)
. What are'the; bases for maximum and minimum pressurizer spray flowrate? ! QUESTION 6.09- -(1,50)
- a. The reactor is' critical at 1E-8 amps during a reactor startup. A malfunctioning steam header pressure transmitter causes six steam Edump valves ~to open. Assuming the reactor does NOT trip, at what average temperature-(Tave) will'the Reactor Coolant' System-stabilize?
(0.5)
- b. Describe the feature (s) which cause(s) Tave to stabilize and remain at this value. (1.0)
QUESTION 6.10. (1.00) Due to a_ procedural deficiency on October 1, 1986, the Vent Stack SPING
~ Unit.was' removed from service for a total of 6 hours and 12 minutes without. backup.
U d Why.was this a problem? I l 4
.6. ELANT SYSTEUS_ RESIGN, CONTROLu_AHQ_lBSTRUMENTATION Page 12 l
QUESTION' 6.11 (1.50) Answer the follesing regarding the PORV pressure protection system used for low temperature operation.
- a. In which position must the PORV Pressure Protection Switch be placed in order to reduce the PORV(s) lift setpoints? (.025)
- b. How many PORV(s) lift setpoints are modified per unit when PORV Pressure Protection is being used? (0.25)
- c. At what pressure will the PORV(s) lift when the system is activated for Units 1 and 2? (0.5)
- d. What is the minimum temperature that may be attained prior to this l system being activated for Units 1 and 2? (0.5)
,. QUESTION 6.12 (1.00) State the Essential Service Water System response to a 01 SI signal assuming the following initial conditions.
. Unit 1 at 50% power
. Unit 2 at 100% power
. 1 East ESW Pump control Switch in Lockout 1AB Diesel Generator is running loaded for a 31 day TS Surveillance Unit i recirc sump valves closed QUESTION 6.13 (1.50)
State the conditions / actions that will trip a Diesel Generator under the following conditions (either during DG operation or DG startup).
- a. Station Blackout
- b. l.oss of All DC
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
- 16. PLANT SYSTElfS DESIGN. CONTROL,_AND INSTRUMENTS 11_ON Page:13 o
QUESTION 6.14 (2.00) TRUE or FALSE 7 l Plant Air Isolation Valves PRV-20 AND PRV-10 must be open to cross + connect:
- a. Unit 2 plant air.to Unit.1 plant air
- b. Unit 1 plant air to Unit 2 control air
- c. Unit 1 plant air to Unit 1 control air'
- d. Unit 1~ control air to Unit 2 control air (0.5 pts each)
QUESTION 6.15 (2.00) Answer the following regarding the ICE CONDENSER Lower Inlet doors.
- a. In what TWO locations outside of containment can the 3.sition of these doors be monitored? (0.5)
- b. How can an operator identify which door is open without a containment entry? (0.5)
- c. There is a maximum power at which an open door may be closed. What is that power level and what is the reason for having a maximum power level? (1.0)
(***** END OF CATEGORY 6 *****) l u_ __ __
' iI' i
- , K 7 .' 'PRQQEp3RES'- NORNALi ABNORMAL EMERGENCY Page 14-fAND RADIOLOGIC 8k CONTROL 1 ,,..
+ '
1
- ( I; ,j ;,
' ~
., iQUEdTION: '7.:.01 (1.00) )
+ , w j
. DuringEa fire, t' hat compensatory action must' the - Shif t Supervisor.
cm ." Jdirect when hanging clearance permits to perform work f s not feasible? 1 e
+ l } ~.
~ QUESTION.: 7.;02' (1.00)-
*- dnswer the following questions concerning' SHUTDOWN MARGIN with the
- i 'unitLoperating'jn'MODEL1:
(a).-An inoperable rod is detected. What IMMEDIATE Technical
' Specification (TS) ACTION is required within one (1) hour of inoperable rod detection? [0.5 Point).
.(b). What IMMEDIATE TS' ACTION is required, if the required SHUTDOWN MARGIN can not be achieved? [0.5 ] !
QUESTION: 7.03. (1.00)- I When may the reactor trip associated with the power range. low setpoint.be-manually bypassed per.2-OHP.4021.001.006, POWER ESCALATION 7 1 QUESTION 17.04 (3.00) State what reactor trip (s) could be prevented or defeated, either manually.or automatically, by each of the.following:
'(a). P-6 [0.5 );
'(b). P-7 [2.0 ]; and (c)..P-8 [0.5 ].
I ; ( , (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)
7- ;; ' r
,; r _g. : <
l:
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.. - . i b , f. iEEQQEDQEES - IlQRd6k.'ABNORtialu_EMERGENC1' Page 15 c 6HQ._R6H1QL_QQ198L . CONTROL L e.
t . c
- ry 1-t: - >i?',
05 . (3.00)
- QUESTION. -7.
Nhi'le~onwathh'as. Shift.. Supervisor, thel Unit-2 "CONTAIN AIR L ~ TEMP AIR" annuciator alarms. - The primary. containment upper H Temperatures of recorder, SG 18, were: 105l degrees F and-101
' degrees F-on the,-712 elevation;. and 91 degrees F on the L ~624' elevation'. Based on this information:
(a). Would'the primary-containment' upper compartment average L lairftemperature Technical. Specification limit of 100- T! degrees F be violated?- YES or:NO. [1.0]
- (b). What are:four.-(4) probable causes for the high containment:
(.f, temperature alarm, excluding spurious alarms? h l QUESTION ~ 7 06. I2.50)
-A-fuel. handling accident has occurred; .and an initial dose assessment is'needed to beLimplemented'by~the Shift. Supervisor per.
PM 2060 EPP,101,' EMERGENCY CLASSIFICATION. . Determine the site
' boundary-dose rate (i.e.,~at-610 meters, dose rate in RAD /HR) i givent i
PMP-2080 EPP.108, INITIAL DOSE ASSESSMENTS (GASEOUS) - (Provided as-L Examination ATTACHMENT 1); and the following-information: radiation monitor VRS 1507 reading '= 1 mircocuries/cc.
. unit ~ vent flow rate = 1E+5 cfm wind-speed = 3 mph
- meteorological tower.' temperatures at!30' elevation = 22.22 degrees C at 180' elevation = 21.40: degrees C For full credit, all work should be shown on nomograph provided.
\ -
(***** CATEGORY 7' CONTINUED ON NEXT PAGE *****)
;Jns7-h;EB9DENUBES:NOBMAL.ABNOBMALEMERGENCY. . Page,16-5J :: 6HD_BbDIOLD919AL_QQETROL-o- n ,
I h ' QUESTIdN- : 7.070 '( 2. 0 0 ) ' a
< I.
-Answe'r;the following' questions, concerning refueling, either'TRUE>
oriFALSE: i (a)).Afmore restrictive reactivity condition', Keff, is.
. required.in REFUELING.-(MODE.6) than in COLD SHUTDOWN
- :(NODE 5).
'(b).-With no. source range neutron flux monitor operable'in.
~ MODE 6, the' Reactor Coolant System must be. emergency'.
-borated.
F=, (c). The reactor is required to.be subcritical for'at 1 east.
~
100 hours:.before-irradiated fuel can be moved within-the spint fuel pool.
-(d). The Containment Purge and Exhaust. isolation system is required to'be operable during core alterations.
i: ,
. QUESTION. 7.08 ( l'. 00 )
2-OHP 2 4 0 21. 001. ,001', LPLANT,HEATUP FROM COLD SHUTDOWN TO. HOT
- STANDBY, requires the logging and periodical blowing down of isolated, steam traps. 'What' potential problem is this administrativeLprocedure intended to prevent 7 r.
j: . t I l (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) m h .m__________.. _ _ _ . , _ _
, 1 m , ,
M," 71 PROGEDURES' / NORMAL; ABNORMAL, EMERGENCY Page(17;
.-AND RADIOkQQICAL CONTROL-r I
I i LOUESTION. 7.09 (2.00)
. Answer each-of the following questions concerning- f 2-OHPi4021.011.003, MANUAL REACTOR CONTROL.WITH-LOAD VARIATIONS, i feither TRUE or FALSE.
(a). Reactor' coolant average' temperature is maintained within
'/-1.
+ 5 degrees F of :T(ref erence)fby comparing the measured-auctioneered. Reactor CoolantLSystem average
. temperature with the reference 1 average temperatures.
.(b). Following-a large load decrease-(i.e.,-greater than 20%) i l
with rods in manual control and the steam dump in
.T(average) control; no rod motion is needed, because the steam dumps will maintain the band for T(average).
(c). If a.' control bank reaches its low rod. insertion limit alarm setpoint, the boron concentration must be t increased immediately in~accordance with 2-OHP 4022.005.002, EMERGENCY BORATION. (d). .Before transferring from manual to automatic reactor control, T(average) must be brought within pC1 degree F of reference average temperatures.
' QUESTION 7.10 (1.00)
During a Safety Injection, reactor coolant pumps are tripped if two conditions are met. State these conditions.
..QUESTION- '7.11 (1.00)
At D.C. . Cook, how is a CONTAMINATION AREA defined per PMP 6010 RAD.001, " Radiation Protection Manual?" Include site
-radiological limits in answer.
(***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)
m - d D. *s Pagei18 s 7; 'PROCKQQBES 'NQBUAki_ABNQBMAL, EMEEGENCY CAHQ_B6DIOLOGICAL_p_QNTROL , o yc QUESTION 7.1'2 l : (1.50)l
. Answer 1the-following-questions-regarding. emergency notification, i either TRUE or, FALSE.
I
'(a)'. The: local: authorities-(i.e., state police) are notified
' before USNRC,'when an emergency classification.is declared. (b). After the initial. notification of an' emergency,
' classification to groups'off-site, the. Shift Supervisor-
- is relieved'from any. additional reporting requirement.to~
,off' site groups, unless the emergency classification 1s-
. upgraded.
-(c). Prior.to-declaring.an UNUSUALEEVENT, the Shift
-Supervisor is? procedurally required by.PMP-2080-EPP.101,
' EMERGENCY NOTIFICATION, to contact the Plant Manager or his designated representative.
' QUESTION 7.13- (2.00)
Answer the following questions concerning PMP 2080 EPP.112, PERSONNEL INJURY: (a). Who (by; title) would contact the hospital typically per the procedure? (b). An. injured worker is radiologically decontaminated before first aid is administered. (TRUE or FALSE)
.(c), Exposure in excess of 10 CFR 20 limits for personnel helping injured personnel can be authorized by an HP supervisor at the scene of the accident.
(TRUE or FALSE) (d). Who by title would initially act as the Site Emergency Coordinator? QUESTION 7.14 (0.00)
~ Deleted from examination.
(***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)
[7 . PRQQEQQBER_.iNQBM6LNABNQBMAhuEMEEGENCY~ Page 19 AHD_BADIOJOGICAL CONTROL
-OUESTION~ . 7 .15 -- -(1.00)
In' MODE 1,- the Volume Control Tank pressure is . required to be greater than or equal.to 15 psig; and less.than 50 psig by ' il-'OHP 4021'.003.005,eREACTOR COOLANT SYSTEM DEGASSING. What is ,
- thelprocedural' basis.for:
(a). low pressure.. limit; and
.(b). upper. pressure limit. o 1
-QUESTION' 7.16 (l'.75)'
As a result of a'small break' loss of coolant accident, adverse containment
. conditions have1been declared, a
- a. Explain briefly why adverse.containme'nt conditions require-the use.
sof' alternate instrument'setpoints'while implementing D. C. Cook Emergency. Procedures. (1.0)
- b. Assuming adverse containment was declared due to containment radiation; what conditions must be satisfied prior to using the normal 1 containment instrument setpoints? (0.75) i l
I i f (***** END OF CATEGORY 7 *****) p _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ -
1 8.. ADMINISTRAllYE PROCEDUBE6._GQHDITIONSi ' Page'20 J - AED__h1MlIAIl0ES i
- g. g 1
- ja EQUESTION 8.01- ,(1.50)
' Answer the^following questions, either.TRUEtor FALSE,.concerning'
. INDEPENDENT VERIFICATION,. as definc.t in PMI-4010,. PLANT OPERATIONS POLICY:
(a):. Independent verification is applicable'only for returning' safety-related equipment back into service (b). Remote indications can be used'during independent verification providedLthey are redandant ofLthe
'i ndications used.during the' original. lineup,
- (c). Independent verification may be waived for some activities'during outages, if a full system lineup.is scheduled to be conducted prior to returning the Unit to service.
QUESTION 8.02 (1.00) Fill !a'the'hianks for the following questions, concerning methods used to ensure.that the steady state maximum power'powerLlimit is o not exceede~d per PMI-4010, PLANT-OPERATIONS POLICY: (a). The average over.any hour period will not exceed the maximum licensed power'1evel. (u). " Excursions" where. power level may exceed the maximum by 2% are permissible for periods up to (time). QUESTION 8.03 (1.00) What is the basis for the POWER RANGE NEGATIVE RATE TRIP? l l' (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) L l g Y,
= ,,o7- == ;- -- -
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- oc ,
Tage:21
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8'." ADMINISTEAII.VE_I'BQCEILyBEjiidONDITIONS.
;, s AND-LIMITATIONS:
se ,
'(2.00)
Qi)ESTI'ON. ? ' 8. 04
! Answer each of.the following questions, either TRUE or FALSE, regardingLTechnical Specification (TS) surveillance requirements for Reactor: Coolant System-(RCS)' CONTROLLED LEAKAGE from the Reactor Coolant; Pump (RCP) seals:
- (a). The monthly monitoring ~ of : RCP seal leakage can not be .
extendedito exceed 25%'of'its surveillance. interval. (b). TS surveillance requirement has been exceeded - given that the total length of tima ior the last three (3) consecutive-monthly surveillance was:97 days. (c).' Deleted from examination. (d). The monthly surveillance for RCP seal leakage is defined with an' interval.that would be done at least once'every thirty (30). days. (e).-Nothing.in the ASME Boiler land Pressure Vessel Code can' supersede the RCP. seal leakage' monthly surveillance requirement,.TS.4.4.6.2.~1~.c. QUESTION. 8.05 ( 1. 50.) a l State!three (3) reasons why containment ice bed loss is a Technical Specification OPERABILITY concern. l
' QUESTION - 8.06 (1.00)
What CONTAINMENT Ih"EGRITY requirement is imposed by the Unit-2 Technical-Specification (TS) definition that is not imposed by the
' Unit-1 TS definition?
' QUESTION 8.07 (1.00)
What is the physical significance of the region under the curve i for a-given pressure, where reactor operation is permitted, in i I Unit-2 Technical Specification (TS) Figure 2.1-1 (Provided as ) Examination Attachment 2). i (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) ) m.__ . _ . . _ _ _ _ -_.m_ ___.m.-.
y
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J" d83 LADMINISTRATIME PROCEDUBES. CONDITIONS.. Page 22; K 'm ! AHD.' kldJTATIONS E'p > 1 e w
! )
P m l ,.
" QUESTION' J8408 :(l'.00)
E. L Who woul'd the Shift Supervisor contact, by title,.to'obtain L additional ^ personnel to respond to a' plant emergency during
?non-working. hours.per'PMP 20~0 EPP.107, NOTIFICATION OF PLANT
-PERSONNEL 7 QUESTION 8.09 (1.50)
Fill.in1the blanks'for-the following statements: (a). Technical 1 Specifications limits primary to secondary leakage to ~ gpm total ~through all steam generators (S/Gs) not' isolated from the RCS. (b). ... and _ gallons per day through any one.S/G.
# (c). Performance of a RCS water inventory balance is-hours during required'at least once every steady state operation.
.. q QUESTION 8.10 (1.00)
Answer the following' questions; (a) Personnel on the D.C. Cook plant staff should not be permitted to (Fill in the' blanks. NOTE - Do not include shift turnover time): (1)'. Work more than hours straight. [0.25 ] (2). Work more than _ hoursLin any 48 hour period. [0.25 ] (b).'Who may authorize deviations to the above restrictions? [0.5 ) t (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) l
- 8. Page.23 AQMlHLSIE&IIVE EBQCEDMBES. CONDITIONS _
6HD LIM 11&IIONS l 1 QUESTION 8.11 (2.00) s l Answer each of-the following questions, either TRUE or FALSE, concerning acce'ss to the Reactor Protection System (RPS) cabinets
;during p'lant operations:
(a). RPS cabinets keys are controlled by the Flant Control and Instrument Supervisor. (b). Duplicate RPS cabinet keys are controlled by the Shift Technical Advisor. (c). RPS cabinet keys are contained in emergency break glass boxes in each control room. (d). With the physical security established for MODES 1, 2,. and 3, RPS cabinets can be left unlocked. LQUESTION 8.12 (1.50) Answoc the following questions, either TRUE or FALSE, concerning containment access: (a). The only exception to the "two-man" safety rule for a person to be alone inside containment is a supervisor inspecting work completed. (b). It is the responsibility of the containment entry point personnel to maintain a log of containment entries and exits for all containment entries (i.e., entries made in MODES 1, 2, 3, and 4). (c). Deleted from examination. (d) Whenever the reactor is shutdown and a containment entry is made, a neutron monitor with an audible indication and alarm is needed within. containment to warn personnel if an evacuation is required. (***** CATEGORY 8 CONTIN 9ED ON NEXT PAGE *****)
) I . .
l.' g. ADMINISTRATIVE PROCEDURES. cot 4DITIONS, Pcge 24 L AND L.IMIl6TIONS QUESTION 8.13 (2.00) , i Answer each of the following questions, either TRUE or FALSE,
- concerning radiological controls: 1 (a). The Radiation Protection Section initiates a , ,
Radioactive Release Form, when it is desirable or necessary to make either a radiological liquid or gaseous release. (b). Requests'for Radiation Work Permits (RWPs) are limited to supervisory personnel (i.e., in Operations, Maintenance, etc.) and to Radiation Protection Section personnel. (c). The Shift Supervisor has the sole primary responsible to minimize his shift personnel radiation exposure. (d). It is the facility policy to strictly adhere to radiological requirements, although not necessarily i 5.ith the spirit of the requirement. I 1 QUESTION 8.14 (1.50) ( l
' Answer each of the following questions concerning clearance tags either TRUE or FALSE
- i (a). A RED CLEARANCE TAG can be hung over a STRIPPED TAG :
I CLEARANCE at the same isolation point. (b). A STRIPPED TAG CLEARANCE can be hung over a RED CLEARANCE j TAG at the same isolation point. ; (c). A clearance permit can be effective without the Shift j Supervisor's signature. 1 (d). Deleted from examination. QUESTION 8.15 (1.00) , What Technical Specification (TS) shift crew requirement is ) i imposed when CORE ALTERATIONS are performed in Mode 5 and Mode 6 that is not required in other modes? (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) l l 1
'18.,/ ADMINISTRATIVE PROCEDURES ' CONDITIONS; Page"25:
g, 4AND L1)LITAIJONSe e , , l
.g .:)" '
QUESTION 8.16 (*'.50)- lAnswerithe following questions, either TRUE or FALSE, concerning ~ i required administrative action following a Reactor Coolant System
~
.(RCS) overpressurization'that' violated the Technical Specification (TS)LSafety Limit::
, T a) . The. facility shall be "'. aced in-at least HOT STANDBY i within'one (1) hour.
(b).1The.USNRC is required to be notified by telephone' ; within four (4) hours,~ if the'RCS' Safety Limit overpressurization is considered a non-emergency.
~
(c). A. Safety: Limit. Violation shall'be reported to the Chairman 1of the. Nuclear. Safety and Design Review- '
. Committee (NSDRC)' within twenty-four-(24) hours. :
QUNSTION- ' 8'~.17 , (1.50)
-Answer? the following questions concerning Unit-1 primary Echemistry:
- (a). In. MODE 3, what'are the steady. state and transient limits for dissolved' oxygen in'the. Reactor Coolant "
System (RCS)? [0. 5 '] (b). With the unit in MODE 1,'it is reported to the Control : Room that the steady state dissolved oxygen limit;is exceeded. How long can the unit be operated before the reactor is required to be shut down by. Technical Specification (TS)? [0.25 ] o E .(c). What'is the normal sample frequency for~ dissolved , ocygen? [0.25 )
'(d'). What is the TS RCS chemistry control intended to minimize? [0.5 ). i, 1
1
- i 1
1 I (***** END OF CATEGORY 8 *****) l L (********** END OF EXAMINATION **********) l
~
\ , u wx __- _ - -
- 5. THEDBL9E NELEbH_EOWEE_ELANT OPERATION2 Page 26
- EL1!IDLA11D THEBMODYNAliLQS .
~ ANSWER 5.01 (1.50)
(a).. Stay the same. [0.25 ) AND Beta Bar's magnitude is strictly dependent on the concentrations of U-235, U-238, Pu-239, and Pu-241 and they change only with life and not with temperature. [0.5 ] (b). EOL [0.25 ] AND Since the beta of' the core is smaller, a larger SUR would result for a given reactivity insertion. [0,5 ) (Accept other answers if justification is supported by FSAR.)
' REFERENCE W Fundamentals of Nuclear Reactor Physics, pp 7-31 to 7-33, 7-44 to 7-45 MASTER CO?Y l
~!
I i
.1
-l 1
I L 1 l ! (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) ! _ 1 _ _ _ _ ____-- --- - - _ -
D
- 15. . IBEORY- OF NUCLEAB i POWER ^ ELBNT OPERAI19HL Page 27 FLUIDS _,.AND THEPJ9DYNAdlCD .
5 4 fhNSWER <5.02 .(3.00)J (a). . (1). . decrease-[0.25-] lower temperature reduces the rods
" sphere of influence" by reducing-the number.of neutrons 1
I _the.rodLsees. -[0,50 ]'
-(2). Decrease [0.25']
. Rods are moved from a high flux' region to a low flux region;o rod worth decreases. [0.50 )
(3). . Increase [0.25 )
- Decrease.in boron increases,the rods," sphere of. influence"
- .by. increasing the number of neutrons the rod sees.
[0.50 )
~
(Will'also accept explanation of " competition") l(b). TRUE' [0.25-). DRW isiessentially independent =of-power as long as the rel'ative-neutronLflux (local flux divided by total core-flux)Lremains constant. ~[0.50 ) (Will accept False as. correct if accompanied by an explanation of temperature, Doppler, and Xe effects that change relative flux from HZP to power operation). REFERENCE W' Reactor. Core Control for Large PWRs, pp 6-22 to 6-28 192005K107 192005K105 ..(KA's) q l ANSWER 5.03 (2.00) ) i
- 1) Criticality.is uncertain, (0.25); Stop rod motion (0.5)
- 2) Reactor is critical (0.5)
. 3) Criticality is uncertain, (0.25); Stop boron dilution (0.5)
L, (***** CATEGORY b CONTINUED ON NEXT PAGE *****) I t L. - 4 _11_ ~ - .
3 5 '. THEQRY OF NUGEAB_EOWEE_fLANT OPEBATION. Page 28 ELM E BED __T.BEEMODXNAMICS REFERENCE W-Reactor Core Control for Large PWRs, pg 9-15 W Fundamentals of Nuclear Reactor Physics, pp 8-57 to 8-60 , 192008K111 ..(KA's) ANSWER 5.04 (1.50) 5% decay heat = (0.05) 3250MW = 162.5 MW (0.1) ! Safeties lift at 2485 psig = 2500 psia (0.1) Tsat at 2500 psia = 668 F (0.3) delta T required = 668 - 547 = 121 F (0.1) 4 (1.2 Btu /lbm-F)(214,000lbm)(121F) = 3.10E+7 Btu (0.3) 162.5 MW + 15 MW = 177.5 MW (56.896 BTU / min /KW)(1000KW/MW)
= 1.01E+7 BTU / min (0.3)
'3.10E+7 / 1.01E+7 = 3.07 minutes (0.3)
REFERENCE Steam Tables Thermal Hydraulic Principles on Applications to the PWR 1, 42. 192008K127 010000K501 000027K102 000027K101 000027A211
..(KA's)
I i ANSWER 5.05 (0.00) l Deleted-from examination. 4 REFERENCE Deleted from examination. 192002K114 001000Kbb4 .(EA's) ANSWER b.06 (1.00) l
)
At BOL the poison rods are burning out, offsetting fuel depletion. At EOL l almost all of the poison rods have burned out, requiring boron dilution to offset fuel depletion. 1 (***** CATEGORY b CONTINUED ON NEXT PAGE *****) 1
-5. IDEQBLQE_tiUGLEAB_EQliEB_EhatiLQEEBAIlQL Page 29
'EkULMSt6ND THEBMQDYNAMICS REFERENCE W Reactor Core Control for Large PWRs, pp 2-11 to 2-12 001010K521 ..(KA's) 1 ANSWER 5.07 (1.50)
Reflux boiling is the process where steam exits the core and is condensed in the SG tubes, with the resulting condensate returning to the core via the hot leg to repeat the cycle. (1.0) This type of cooling occurs with a voided core when no reactor coolant pumps are running. (0.5) Alternate Answers will be accepted on a case-by-case basis. REFERENCE , LP-RO-C-MC01 Objective 3 193008K124 ..(KA's) ANSWER 5.08 (2.50) Q(1) = U(1)A(1)[Tave(1) - Tstm(1)] = Q(2) U(2)A(2)[Tave(2) - Tstm(2)] (0.50 point) Q, U, ar._ Tstm remain constant, so that: A(1)[Tave(1) - Tstm] = A(2)[Tave(2) - Tstm] (0.50 point) From steam tables, Tatm=Tsat for 780 psia =515 F (0.50 point) From question, A(2) = 0.8A(1) (0.25 point) i A(1)[567-515] = 0.8A(1)[Tave - 515] (0.25 point) 4 Tave = 580 F (13 degree increase) (accept 579.5 to b80.5)A (Give credit for alternate methods if work method is correct) (0. 50 point) REFERENCE Steam Tables W Thermal-Elydraulic Principles and Applications to the PWR, pp 5-16 to 5-22 l 035010K109 ..(KA's) l (***** CATEGORY 5 CONTINUED ON NEXT PAGE ** x** ) i
L .
- 5. : IH50RY OF N QLEAR POWER _ELANT OEEEATION 1 .Page-30 JLQlDS. ANDlIHERMQDYNAMICji I
l t
.I t
. ANSWER.
5.09: (2.00)
/As the fluid flows around'the bend in-the' elbow, its velocity increases (0.5), the centripetal forceLon the outside'of the radius will increase 1 .(0.5)fand-result-in a pressure increasecon the'outside radius ard'a' decrease in pressure on'the inside' radius-(0.5). TheLflow rate-is then o
proportional'to thelsquare root of the pressure. difference-(0,5)- . (Accept ~ alternate wordings) REFERENCE. WL Thermal-$ydraulic Principles-and Applications to theLPWR pp 11-17 to 11-24 191002K105 ..(KA's) LANSWER. 5.10 (2.75)
- a. Lthe~ maximum local ~ flux on theisurface'of a fuel rod (0.25) at' core elevation 1Z (0.25) divided-by the average' fuel rod heat fluxz(0.25), allowing for manufacturing ~ tolerances on fuel, pellets
-and rods'.
-b. Control rods'in a single 1 bank move together with no individual' rod
; insertion differing by more than +/-12 steps from the bank demand position'(0,5)
Control rod' banks are, sequenced with overlap (0.5). Control bank insertion limits are not violated-(0,5) Axial power distribution 'is maintairied within limits (0.5) l REFERENCE j i
, Technical Specifications' Bases, pp 3/4 2-1 and 3/4 2-4 1
-015020K505 ..(KA's) ;
i j I n (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****-) _ = = - - _ - _ _ _
a.
. sc
/5; ~ THEORY QE_NUHL_E68_EQWEB_ELANT OEERATIQNm Page 31
% , FLUIDS tAUD_IESBMODYd6MICS- _
L
? ANSWER l .5.11. (1.50):
ai one. step:= 5/8"'==> 8'pcm/ inch ==> 5 pcm/ step .(0.25)
. Beta'Bar. Effective = 0.00596- (0.25)
' Lambda Effective = 0.05 sec-11 _
(0;25), (0.25) Rho-='(5 pcm/ step)('-20. steps)'= -100 pcm o Lyo , Sur=[(26.06)(lambda)(rho)]/['(beta bar-rho)] (0.25)-
' =[(26 06)(.05)(-'.001)]/[( 00147+.001)]
p ( .001303)/(.00869)
% L.
- .- 19 7dpm ~PM (0.25) 1x Deleted.from M .
examination. l LREFERENCE
.W Fundamentals.of~ Nuclear. Reactor. Physics, pp 7-12:to 7 - 001000A106: . .- ( KA ' s )
1
-l LANSWER ' 5 ~.12 - - ( 1.~ 0 0 )
~
Perform'a heat balance calculation across the regenerative HX-(.5) by.using
- delta T across both~ sides of,the'HX (.25) and letdown flow. rate'(.25) to determine' charging flow rate. (Will allow 1/2-creditLfor: .chg flow = L/D
- flowi+ Seal Leakoff - Seal injection, with Par level constant)=
(Will accept other' answers if supported'by viable assumptions.) REFERENCE i W Thermel-Hydraulic Principles and Applications to the PWR, pp 5-16 to 5-381 191006K108 004000K607 ..(KA's) ANSWER. 5.1'S (1.50) j
, Increase. (0,5)
The: shutdown margin at EOL is based on the value used in analysis of the hypothetical steam' break accident [0,5], which is more severe at'EOL because the moderator temperature coefficient is more negative [0.5].
)
~(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) !
i
- e. s-y ; o
[ w:
~
lTHEORY'OF'NUCLEAB' POWER PLANT OPERATIOL.. .Pego32-h..E w
,f 5.ELEIDS.AND THEFJQDYNAMICS w
n REFERENCE
~D.;C.' Cook Tech Spec Basi's 3/4.1'.1.1.
s 001010A404 000007K102 ..(KA's).
" ANSWER-5.14 (1 00)
, 11. Increase.
-2. Increase
- 3. Increase. (0.33.ea.'). (1.0)
REFERENCE W Thermal-Hydraulic Principles and' Applications to the' PWR, pp 10-54 to 10-61 191004K106' 003000K201 003000K110 ..(KA's) fANSWER 5.15 ( l'. 00 )
-Temperature will be lowest in the lower ~part_of the core,.thereby making )
s in this section of the: core. More negative i boron: worth higho't reactivity will be added to the lower section of the core, forcing the flux-'pcak higher inLthe core, thus driving delta-I'toward the positive side of the: band. Alternate acceptable answer: ! Boration with no change in power or rod position will produce the same
-magnitude reduction in Th, Tavg and Tc. The mass' density change per degree-for Th'is greater than for Tc; thus the value of MTC for Th is greater (more.. negative) than for Tc. .Therefore the temperature reduction will add more positive reactivity to the upper section of the core than'to the lower i
~
.section, thus pulling delta-1 in the positive direction. ]
REFERENCE W Reactor Core Control for Large PWRs, pp 5-13 to 5-14 and pp 3-20 to 3-21
'iO2004K110 001050K502 ..(KA's) l
\
x , I (***** END OF CATEGORY 5 *****) >
l
'gr. EhdNT SYSIEMS DESIGN. CONTROL. AND INSTRUMENTATTOS Page 33-I L
f ANSWER 6.01 (3.00)
- a. 1) RHR Hx
- 2) CCP - (Gear oil cooler, bearing oil cooler, and seal plates) 1
- 3) SI. pumps - (Mechanical seal Hx's, bearing oil cooler)
- 4) RHR pump (mechanical seal Hx) 5)- CTS pump'(mechanical seal Hx)
(Any 4 @ .125 pts each)
- b. Splits surge tank [0.25] maintaining one-half for each saf~eguards train [0.25).
- c. On inlet to CCW Hx. (0,5)
,d. SI [0.3] with switch in neutral [0.1] or auto [0.1] !
Blackout (0.3] with switch in neutral [0.1] or auto [0.1] ; Low Header' Pressure [0.3] of 80 psig [0.1] in auto [3.1] ! REFERENCE R0-C-AS01 008010K403 008000K301 000028K302 000026A105 ..(KA's) l ANSWER 6.02 (1.25) l
- a. slave cyclers i
- b. non-urgent failure
- c. d2
- d. will not
- e. 12 (0.25 pts each)
REFERENCE RO-C-NSO4, pgs. 11, 12 001000K403 ..(KA's) (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
- 16. PLANT SYSTEMS _ DESIGN, CONTROL. AND INSTRilMENTATION Page 34 ANSWER 6.03 (2.75)
- a. 1. Vacuum trip Wear
- 2. Thrust bearing water
- 3. Overspeed
- 4. Low bearing oil pressure
- 5. Low pump suction pressure
- 6. Safety injection
- 7. Steam generator Hi Hi level
- 8. Reactor trip
- 9. Manual (local and remote)
?q. L.O. Pressure < 60% Normal
- 11. L.O. Pressure < 40% Normal (Any 8 @ 0.25 pts each)
- b. 1. The HP & LP stop valves must be closed.
- 2. The motor speed changer must be at the low speed stop.
- 3. The turbine speed controller must be down to the miniumum position.
(0.25 pts each) i REFERENCE RO-C- pg. 10 059000K419 059000K416 059000A411 059000A401 (KA's) , I i ! (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) l l L____________.___
cr r-- - 1
-6. PLANT SYSTEMS _ DES 19H. CONTROL 1 6ND INSTRUMENTATION Page 35 h;
i L , h i I L
' ANSWER 6.04 (1.50)
- a. Overcompensated. (0.5)
- b. It counteracts the signal produced by the gamma radiation so that -
only the neutron level is indicated. (0.5)
- 1. Erroneous startup rate.
- c. ,
I
- 2. Limiti.e IR-SR overlap.
- 3. Delay the P-6 permissive beyond the source range trip block setpoint. i (Any 2 @ 0.25 pts each) l REFERE NCE l ES-0.1, RO-C-NSO9-SH03
.015000A202 000033A211 .(KA's) I t
ANSWER. 6.05 (1.50)
- a. Natural convection [0.5] or CEQ Fans -
- b. Air te.nperature .is raased to the point where hydrogen and oxygen 3 spontaneously recombine, forming steam. [0.5] ;
i
- c. Temp set controller:
Provides d ndication of actual temperature (0.25) Power adjust potentiometer: Set the amount of power supplied to the heaters. [0.25] l 1 REFERENCE : LOGIC 98514-1 RO-C-NS15, pgs. 10, 16 028000K601 028000G004 028000A401 . (KA's) l i (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
-1 lx ' ). '
... llf b ELbHI._SYSIEMS DESJGN.> CONTROL.-AND INSTRUMENTATION ;Page~-36'-
ifl' ANSWER 6'.06 (2;00)
- a. !WheneverLa; reactor trip breaker (.4) and-itsl bypass. breaker.(.4)- -- ; -
are both open
.l
.('.4 for. stating " reactor. trip")
b' -. (any fcur of,the following) (~3 each)- 1., Trips turbine
.2. Feeds feedwater isolation (coinc w/ Low-Tavg) 4 3. Blocks ~ closed feed. reg, valves (when' closed by SI or Hi-Hi S/G 1evel)'
- 4. Feeds SI(block / reset)' logic
<5. Trips feedpumps ;
REFERENCE-RO-C-NS11-SH03- . . s i 012000K610t .i(KA's) ANSWER 6.07 (1.00)
.The' reactor trips [0,5) on OT delta T [0~.5].
-)
REFERENCE-
'RO-C-NSO9, NS11 016000A201 '012000K611 000027A215 .(KA's) l l
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) l l
$ 6 .' ? PLANT SYSTEMS' \ DESIGN.'CONTBQ k AND' INSTRUMENT H QF Page-37' i > '
' ANSWER. i6,08' : ( 2. 00 ) 1
~
Maximum: Prevents' reaching the operating-setpoint of the power operated relief valves:[<0.5] during'a step reduction in power of'10% of j f ull iload - [0. 5] '. ;
. Minimum: Minimize thermal shock'to PZR components [0.5). Maintain uniform chemistry [0.25] and temperature in PZR [0.25).
. Alternate acceptable _ answers if candidate assumes "800 gpm" is the nominal flowrate:
- Minimum: Prevent reaching the operating setpoint of the power operated' i
[0.5) relief. valves during a step reduction in power of 10% of full-load [0.5].
~
Maximum: Reduces severity and swiftness of stuck open spray valve transients. '(Prevent reactor trips or'SI's on low pressurizer-pressure. (1.0) REFERENCE i RO-C-NS03,.pg 11
.010000K603 010000K403 010000K401 ..(KA's)
ANSWER 6.09 '(1.50) a .- 541 degrees F (0.5)
- b. The increased steam dump will cause Tave to decrease. At the P-12 setpoint.(541 degrees F) (positioning air will be vented and) all steam' dump valves will close_(0.5). -The affected steam dump valves ,
will open when decay' heat increases Tave above 541 degrees F and will close again at 541 degrees F (thus maintaining Tave at approximately 541' degrees F) (0,5). REFERENCE
'SD B-7, p. 46
~
Westinghouse Logics, Sheet 11 0200041K40 ..(KA's) [ (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) l i j
L 6. PLANT SYSIEMS~ DESIGN. CORIRQht_AND INSTRUMENTATIQM Page 38 f-1 LANSWER 6.10 (1.00) The ability to monitor Vent Stack tritium release was lost. (Also accept - This constituted an unmonitored release path which is in violation of Technical Specifications.) l REFERENCE LER 86027 073000K301 .(KA's)
' ANSWER 6.11 (1.50)
- a. Unblocked position [0.25] (Also accept COLD OVERPRESSURE)
- b. 2/ unit .[0.25]
- c. U1-385 psig ( - +/-2 2 ) [0.25]
U2 420 psig ( +/1 15 ) [0.25]
- d. U1 170 degrees.F [0.25]
U2 152 degrees F [0.25] REFERENCE OHP 4021.001.004, pg. 11 010000K403 . (KA's) ANSWER 6.12 (1.00)
, Both unit's ESW pumps start [0.25] except 1 East ESW Pump [0.25] '
. Unit 1 ESW out1ct valves from CCW Hx [0.25] throttle to mid position [0.25]
RRr"RENCE RO-C-AS02, Learning Objective #4, pg. 12 076000K403 076000K402 013000K108 .(KA's) (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
)
6,.::fLANT SYSTEdS DESIGN. CONTBQL. AND IEEIBUL4ENIATION Paga 39
-ANSWER 6.13 (1.50)
- a. Electrical overspeed (110%) [0.25]
Generator Phase Differential [0.25] i Electric Shutdown Buttons [0.25] j Incomplete Start [0.25]
- b. Manual (MFU) pull button [0.25] ;
Mechanical Overspeed [0.25] REFERENCE RO-C-AS10, Learning Objective 13, pg. 24 064000K402 064000A401 064000A306 .(KA's) ANSWER 6.14 (2.00)
- a. FALSE
- b. FALSE
- c. FALSE
- d. FALSE REFERENCE RO-C-AS12 078000K402 078000K401 078000K303 078000K103 ..(KA's)
ANSWER. 6.15 (2.00)
- a. Control Room (0.25)
CAS (0.25)
- b. Operator to go to CAS to determine which door is open. (0.5)
- c. 20% FL 107, (0.5)
Neutron dose consideration (0.5) (Also accept high radiation levels) REFERENCE i RO-C-NS14, Learning Objective #6 194001K104 194001K103 025000K601 025000K402 025000A203
..(KA's) ,
!r L i L i l
(***** END OF CATEGORY 6 *****) j l
- \
/ 1 I
r
- 7. PROCEDURES - NQBd6Lt_6BNORMAkt_ EMERGENCY Page'40 AND_BADLQLOGIC6h___CONTRQL
) ANSWER 7.01 (1.00) The Shift Supervisor shall dispatch people to guard all control or isolation points (i.e., valves, switches, circuit breakers, etc.) as required. REFERENCE , PMI-2110, Sections 3.1.1.1 (pg 2 of 18) and 3.3.2 (pg 9 of 18), Rev. 12. 194001K102 ..(KA's) ANSWER 7.02 (1.00) (a). A SHUTDOWN MARGIN verification is required. [0,5 ] (b). The immediate action is to initiate and continue boration until the required shutdown margin is restored. [0.5 ] FEFERENCE Unit-2 TS 4.1.1.1.1.a, pg 3/4 1-1, Amend. No. 82 for Part (a). Unit-2 TS 3.1.1.1, pg 3/4 1-1, Amend. No. 82 for Part (b). 000024K301 000005K304 000005K105 . (KA's) l ANSWER 7.03 (1.00) The power range low setpoint may be manually bypassed when P-10 permissive is satisfied (by having 2 of 4 power range channels i indicateapowerlevelthatisgreaterthanorequalto10% power}. REFERENCE 2-OHP 4021.001.006, POWER ESCALATION, Step 6.16, Rev. 6. i Unit-2 TS 2.2.1, page B 2-3, Original Page. l 012000K610 022000A403 ..(KA's) (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) 1 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ l
i f EBQQEDMRES - NORMAL; ABN0BMAL.' EMERGENCY ,
'Page:41!
cAND_ BAD 19 LOGICAL CONTROL n, .- , 4 ANSWER 7.04. L(3.00') l
.(a)',5SourceTrange reactor-trip. [D. 5 > ]
(b). Reactor l trip'on:
-- p ' ( 1 ) .. All' low: flow trips'[0.~4].
- a. Low:RCS: loop flows
- o r~ - .
d e b.. RCP breaker position
'(21. reactor coolant pump under-vc1tage and under-frequency
-[0.4-];
(3)1. turbine trip,[0.4r];
'(4')e pressttrizer low-pressure [ 0., 4 );
4 > ( 5 ) .. . pressurizer high11evel [0.4.].
~
!(c). .P-8 blocks reactor. trip'on low coolant flow in a' single loop. [0. 25]'
~
.RCP breaker position trip from 1/4 breakers-{0.25]
.-REFERENCE
' Unit'-2 TS Table 3.3-1, page 3/4 3-6 with Amendment No. 86 and page 3/4 3-7.with. Amendment No. 82.
012000K610 .012000K406 ..(KA's) o T (
..(***** CATEGORY 7 LONTINUED ON NEXT PAGE *****) ,
= - __ , _ _ _ _ _ _ _ _ _
c:- un _
...s !.-
U: . . . . f 7 . PageX42
,,,s.
? R 7.~'PROCEDDBES w NOBMAL..ABEORMAL.' EMERGENCY r ;. A N D - R A D l Q L O G I C A L C O N T R O 1.' ,
+
l LANSWER- ' 7. 05' (3.00). (a). NO. [1.0') .
, ;(NOTER Obtain,the'arithmeticalLaverage, which is:
(106 +.-102 + 91)L/.3-= 99 degrees F. Since 99 <.100 degrees F, the.TS=limitLis not violated). (b). Probable causes are: in , (1).' malfunction of containment ventilation cooling units.
- c. . . .)
(2). loss'of cooling water to containment ventilation ; cooling units. '.
'(3). high lake temperature and'high ambient temperature.
(4)..RTD malfunction.' 1 (5) dpper-containment ventilation heats on. (6) Heating steam valved in. _(Any 4 @ 0.5. pts each)
'(Other examples will be considered on a case basis).
REFERENCE 2-ONP 4024 224.045, Rev. 1,. dated 04-20-82. Unit-2 TS 3.6.1.5 and 4 6.1.5 on page 3/4 6-7, Original Page. 103000G008~ .103000G005 103000A101 ..(KA's) !
, l 1
.1 b l '
L (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) l-i-
- = _ - _
$'l N. ' PROCEDUBES ~ ' NOFMALUAB)31QRMAL. EMERGENCY Page'43 fAHp_ RADIOLOGICAL ~CONTEROL
. s:
4 , I j ,' ' TANSWER 7.06) (2.50)-
.Using the. nomograph'for: radiation monitor VHS 1507,. Exhibit D, page,1 of:2,: Rev. O of PMP'208G EPP.108: '
(a).. Draw a st'raight lino connecting the data points on scale (1)1foro the monitor reading and on maala (2) for the unit ventiflow rate. [0. 5 ]'
-(b). DrawLa. straight line' connecting the intersection of scaled (3) for radioactive' vent; flow rate and with the wind. speed provided on scale.-(6). [ 0 5 -]
(c).'Determi'ne the Pasquill Category by: (1).. subtracting the' meteorological tower temperature at 30from-its' temperature at 180' would obtain -0.82 degrees C [0,5 ]; and (2). using PMP 2080 EPP.108, Section 4.4.4, Page:2 of 5,lRev. O, to'obtain the Pasquill Category ( e .' g . , with -0.82 degrees'C, the Pasquill Category would be "B"). [0.51].
- (d). Connect the intersection of scale (5), which is unitless, and.the Pasquill Category determined.
[0. 5 ' ]. If entire problem was done correctly,. scale (7)'.for the 610 meter dose rate would be 0.8 +/- .2 RAD /HR. REFERENCE- , 1 PMP 2080 EPP.101,. Exhibit A, ECC-18, Page 24.
'PMPc2080 EPP.108 Initial Dose Assessment.
SD-DCC-NE101, Section 6.1, page 30 000036A203 ...(KA's) z ANSWER' 7.07 (2.'00) (a).'TRUE. [0.5 Point] (b). FALSE. [0.5 Point]
..(c). FALSE. [0.5 Point]
.(d). TRUE. [0.5 Point]
L (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) l Em-- - ' -
x< ' J
- 17. PROCEDURES 0 NORMAL, AB'NORMAlu_EME8GENCY '
Paga 44 .l 6HQ._BAQ1QLQGLGAh_,CONTRQL j y_ L2. 2.> . 1 h'[ iU s I g
-REFERENCE' ', '
I s U-2':TS Table 1.'1', pg 1-9, Amend. No. 51. U-2 TS'3.9.1, pg-3/4 9-1, Original Page. l T. .0-2.TS 3.9.2,1 pg 3/4 9-2, Amendment No. 62 for Part (b). i i
'U-2'TS:3.9.3,jpg 3/4 9-3, Original Page for Part.(c).
- D-2iTS 3.9.9,' pg 3/4 9-9, Amendment No. 63 for~part (d). .
0330000005; _029000G005 002000G001' 000024G003', .(KA's) a
- l eANSWER: 7.08 (1.00) 'l
... to prevent water hammer, thermal stressing of piping or piping or b.
l equipment damage: -(Any accepted for full credit) REFERENCE D , > 2-OHP 4021.001.001,. Section 4.1.18, Rev. 9, pg 5 of 26. 1 039000K501- ..(KA's)_ l 1 s l ANSWER' 1 7.09- (2.00) (a). TRUE. [0.5 Point] 6 '(b) FALSE. [0.5 Point] (c). FALSE. [0'.5-Point] (d). TRUE. [0.5 Point] i REFERENCE 2-OHP 4021.011.003, Section 6.0, Rev. 1, pg 3 & 4 of 5. 001000A403 001000A101- ..(KA's) i T l I (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) A- _ _ . _ . - . _ _ __m__.- - - - -
W' m;w ; .
+
9 r s a, /.
+
g 7c PROCEDURES'- NQBMALL ASRQBMAL. EMERGENCF Page 45-W;i@x@eAND.RAR19LQGLQ6k_. S3 .
- CONTROL
'. e
- gw, : , ,
/,. t]
s s__ e
,' , " ANSWER' - 7 .10 L '(1 00)
? , (a)..!At,least one ECCS. pump. running (i.e., CCP or SI pump). % -[0,5 Point]. E V A* , p ,q/g w ...RCS pressure'less than 1250.psig. [0.5. Point]
-(b). "g ,
5, .7%,, .Also' accept: ca ' Any 3 SG'sL<;8%:WR'
- p. . .
or: i
'9 ' Pressurizer Pressure > 2235 psigi 4 N, ,
- Provi'dedfthat.it is'specified that:. secondary heat sink was. lost
[ ; REFERENCE.
"y
~02-OHP.'4023.EO,nStep 22, Rev. O, pg.13-of 20.
003000A202 . . ( KA '. n ) . y . ,J, ['2 ANSWER 7 11 "( 1. 0 0 ) ' ! CONTAMINATION AREA'is' defined'as any area in which the removable , contamination'on an. accessible' surface or equipment exceeds H 1[0.4' Point]: ( a ) '. 500 dpm/100 sq. cm. beta-gamma [0.3 ) ! (b)..t 50,dpm/100 sq. cm', alpha [0.3 ] l t REFERENCE , I I PMP 6010. RAD.001,-Section VIII.D.1.d.1.d, Rev. 7, pg 81 of 129.
-194001K103' .,.(KA's) 4
+: JANSWER ,
7.12 (1.50) L I
, '(a). TRUE. [0.S' Point]
J< '(b). FALSE. [0.5 Point] I t :(c). FALSE. -[0.5 Point] j i O .
, 1 l I n
y > j i' (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) t 1 I i
- 7. PROCEDURES =- NOBd6Lt ABNQBdAkt_ EMERGENCY Page 46 AND._BADlQLOGICAL CQN.1ROh REFERENCE PMP 2080 EPP.106, Section 4.4, Rev. O, pg. 2 of 3. l PMP 2080 EPP.106, Section 4.4.2, Rev. O, pg. 2 of 3. H PMP 2080 EPP.107, Section 4.1, Rev. O, pg. 1 of 3. l 194001A116 ..(KA's) l I
l ANSWER 7.13 (2.00) I i (a). Assistant Shift Supervisor. (Also accept Shift Supervisor, Site Emergency Coordinator) [0.5 Point] (b). FALSE [0.5 Point) (c). FALSE [0.5 Point] (d). Shift Supervisor. [0.5 Point) REFERENCE ] PMP 2080 EPP. 112, Rev. O, SRO Lesson Plans, Vol. 2, SR C EP09, item X, page 10 of 10, Rev.1. 194001A116 ..(KA's) 1 i ANSWER 7.14 (0.00) I Deleted from examination. REFERENCE l J Deleted from examination. 010000G014 ..(KA's) ANSWER 7.15 (1.00) I (a). to prevent possible damage to the reactor pump seals [0.5 Point] (b). to minimize the chances of accidental releases !' i [0.5 Point] 1 REFERENCE j 1 1-OHP 4021.003.005, Section 4.0, Rev. 7, pg 2 of 6. 004000A204 004000A203 .(KA's) ( I l (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) l l
- 7. PRQQEDUBES ' NQBUMu_aHMQBMAkEMEBGEllGY Page 47 61]D_B6DIOLOGIGAL CONTROL l
J ANSWER 7.16 (1.75)-
- a. Accuracy of instrumentation can be adversely affected [0.5] by temperature [0.25] and/or radiation [0.25). )
I
- b. Containment pressure less than 1.1 psig. (0.25)
Containment radiation less than 10E+5 R/hr (0.25) ! l Containment integrated dose verified less than 10E+6 R (0.25) REFERENCE i RO-C-MC06, pp. 9, 10, Objectives 5, 6 191002K111 000069G011 000011K312 ..(KA's) i i i (***** END OF CATEGORY 7 *****) j i
)
Page 48 E. 'AQUIE1HIBATIVE PROCEDURES. CONDITIONS& AED LIMITATIONS ANSWER 8.01 (1.50) (a). FALSE. [0.5 ] (b). TRUE. [0.5 ] (c). .TRUE. [0.5 ] REFERENCE' PMI-4010, PLANT OPERATIONS POLICY, Sections 3.8, Rev. 3, pg 11 to 14. 194001K101 ..(KA's) ANSWER 8.02 (1.00) (a). 8 hour period. [0.5 ) (b). 15 minutes. [0.5 ) REFERENCE PMI-4010, PLANT OPERATIONS POLICY, Section 3.9, Rev 3,pg 14 of 18. 001000G001 ..(KA's) 1 AllSWER 8.03 (1.00) i To ensure that the calculated DNBR is maintained above the design DNBR for control rod drop accidents. REFERENCE Unit-2 TS 2.2.1, page B 2-4, Amendment No. 82. 012000K402 012000G006 . (KA's) l f I I l (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)
- 8. ADMlHlEIRAILYE_EBOCEDUREE1 CONDITIONS t Page 49
'AUR_LIMII8Il0HS ANSWER 8.04 (2.00)
(a). TRUE. [0.5 ] (b), FALSE. [0.5 ] (. Note: 3,25 x 31 days = 101 . days; and this answer would not change if a 30 rather 1 than a 31 day monthly frequency was selected). (c). Deleted from examination. (d). FALSE. [0.5 ] ( Note: Correct answer is 31 days). (e). TRUE. [0.5 ] (Special consideration given for answers that reflect question prefaced incorrectly) REFERENCE Unit-2 TS 4.0.2, Page 3/4 0-2, Amendment No. 53 for Parts (a) and (b). Unit-2 TS 4.0.3, Page 3/4 0-2, Amendment No. 53; and SD-DCC-HP 108, Section 4.3, Rev. 6, Page 9 of 14 for Part (c). Unit-2 TS 4.0.5.b, Page 3/4 0-3, Amendment No. 78 for Part (d). Unit-2 TS 4.0.5.e, Page 3/4 0-3, Amendment No. 78 for Part (e). 003000K110 002000A301 .(KA's) I i I l l l I (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) i i
t 3 ADMIN _ISIBAI1VE'EBOCEDURE W DED1110MS2 Page 50 l: AHD._Idtil2ATIONS ('
' ANSWER 8.05 (1.50)
Three reasons are: ; (1), an even distribution of ice in containment bays is needed [0.5 ]; i (2). a sufficient amount of boron is required to preclude
' dilution of the containment sump following a LOCA
[0.5 ]; and (3). a sufficient amount of ice is required to contain sufficient heat removal capability to condense the reactor system volume released during a LOCA. [0 5 ] (4) Ensure post LOCA pressure below design (5) Reduces ability of sodium tetraborate to remove Iodine (pH control) REFERENCE Unit-1 TS Basis 3/4.6.5.1, Page B3/4 6-4, Original Page. 025000K301 025000G006 .(KA's) .) ANSWER 8.06 (1.00) l The Unit-2 CONTAINMENT INTEGRITY definition requires the j additional verification that the sealing mechanism associated with ' each penetration (e.g., welds, bellows or 0-ring) is OPERABLE. REFERENCE SR-C-0204, Section IV.F.5, Rev. O, Page 10 of 28. U-1 TS 1.8, page 1-2, Amendment No. 87. I U-2 TS 1.8, page 1-2, Amendment No. 73. j 10300G005 000069A201 .(KA's) (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)
'8. ADMINISTRATIVE _EBQQEQHBES. CONDITIQNE& Page 51 , AND_LldLT.ATIONS ANSWER 8.07- (1.00)
The curves of Figure 2.1-1 show the loci of points of THERMAL l POWER,-Reactor Coolant System pressure and average temperature . below which: the calculated DNBR is no less than the correlation DNBR
, limit value OR-the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid. 1
[ Full credit for either answer]. REFERENCE U-2 TS Figure 2.1-1, pg 2-2, Amendment No. 82 ; U-2 TS Basis, 2.1.1, pg 2-1, Amendment No. 82. RO-C-MC01, POST ACCIDENT COOLING, SRO Lesson Plan Vol 2., pg 5 of 16, Rev. 1. ; 002000G006 ..(KA's) i 1 ANSWER 8.08 (1.00) l Security Shift Supervisor REFERENCE i PMP 2080 EPP.107, Section 4.4.1.3, Rev. 1, pg. 2 of 3. 194001A116 . (KA's) 1 ANSWER 8.09 (1.50) (a). 1 gpm. [0.5 ) ) (b), 500 gallons per day. [0.5 ) (c). 72 hours. [0.5 ] (Also accept daily) l I i l t l (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) { l
.8. ADMldlSIBAIIVE PROCEDURES. CONDIIlONS. Page 52 AND LIMITAllQHS ' REFERENCE U-1 TS 3.4.6.2.c, pg 3/4 4-16, Order dtd 04-20-81.
U-2 TS 3.4.6.2.c, pgi3/4 4-15 Order dtd 04-20-81. U-1-TS 4.4.6.2.1.d, pg 3/4 4-17, Order dtd 04-20-81. U-2 TS 4.4.6.2.1.d, pg 3/4 4-16, Order dtd 04-20-81. 002000G005 .(KA's). ANSWER 8.10 (1.00) (a). Working limits are: (1). 16 hours. [0.25 ) (2). 24 hours. [0.25 ] (b). Plant Manager ur his designee. [0.5 ] REFERENCE ! PMI-4010, Sections 3.2.1.(a) and (b); and 3.2.2, Rev. 3. I 194001A103 ..(KA's) ANSWER 8.11 (2.00) (a). FALSE. [0.5 ] (NOTE: Shift Supervisor) (b). FALSE. [0.5 ] (NOTE: Security Supervisor) (a). TRUE. [0.5 ) (a). FALSE. [0,5 ] i REFERENCE PMI-4010, PLANT OPERATIONS POLICY, Section 3.7, Rev. 3. l 012000G001 (KA's) ANSWER 8.12 (1.50) (a). FALSE. [0.5 ] (b). FALSE. [0.5 ] (c). Deleted from examination. . I (d). FALSE. [0.5 ] (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)
l
- 8. ' ADMINISTRATIVE PB9CEDHEEEm_GONDITIQNS. Page 53 AND LIMITATIONS REFERENCE PMI-4010, PLANT OPERATIONS POLICY, Sections 3.4.2, 3.4.4, and 3.4.6, Rev 3 103000G001 ..(KA's) q l
i
' ANSWER 8.13 (2.00)
(a). FALSE. [0.5 ] (NOTE: Shift Supervisor or Assistant Shift Supervisor). i I (b). FALSE. '[0.5 ] (NOTE: Any individual may { request a RWP). 1 (c). 7ALSE. [0.5 ) (NOTE: radiation worker shares responsibility). j l, (d). FALSE. [0.5 ]
-REFERENCE PMP 6010. RAD.001, " RADIATION PROTECTION MANUAL,'
Section XI.D.6.b, Rev. 7, pg. 112 of 129 for question (a). l Section XIII.D.5, Rev. 7, pg. 123 of 129 for question (b). Section II.D.3, Rev. 4, pg. 22 of 129 for question (c). Section I.B, Rev. 7, pg. 2 of 129 for question (d). 194001K103 ..(KA's) l i ANSWER 8.14 (1.50) / (a). FALSE [0.5 ] (b). FALSE [0.5 ] (c). FALSE [0,5 ] (d). Deleted from examination. REFERENCE PMI-2110... Section 3.4, page 10 of 18; Sec. 3.4.5, pg 11 of 18; Sec. 3.2.2.2, pg 4 of 12; and Sec. 3.1.1, page 2 of 18, Rev 12. 194001K102 ..(KA's) ANSWER 8.15 (1.00) A licensed SRO is specifically required to supervise the CORE ALTERATION (SRO-CA). (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) t . _ _ _ _ _ _ _ _ - _ _ _ _ _ _
=8. AD_111HISIB6IIVE EEQCEDURES CQHDITIONS, Page 54 6ND_5IMIIATIONS l
i REFERENCE 1 0-1 TS Table 6.2-1, page 6-4, Amendment No. 63. U-2 TS Table 6.2-1, page 6-4, Amendment No. 34. 034000G005 ..(KA's) ) l i ANSWER 8.16 (1.50) (a). TRUE. [0.5 ] ; (b). FALSE. [0.5 ] ( NOTE: within one hour). i (c). TRUE. [0.5 ] ) 1 REFERENCE U-1 TS 6.7.1, pg 6-13, Amendment No. 72 for Parts (a) & (c). U-2 TS 6.7.1, . pg 6-13, Amendment No. 51 for Parts (a) & (c). 10 CFR 50.72(b)(ii)(B) for Part (b). 000027G003 000027G002 000027G001 (KA's) i ANSWER 8.17 (1.50) ; (a) Oxygen limits are: (1). Steady Stato: < 0.10 ppm [0.25 ) (2). Transient: < 1.00 ppm OR 10 times steady state limit [0.25 ) (b). Within 24 hours. [0.25 ) (c). 3 times per 7 days. [0.25 ) (d). Corrosion. [0.50 ] REFERENCE U-1 TS Table 3.4-1, pg 3/4 4-19, Original Page for (a) & (b). U-1 TS 3.4.7.a, pg 3/4 4-18, Original Page for (c). U-1 TS 3/4.4.7, pg B 3/4 4-4, Amend. No. 53 for (d). 004000K501 004000G006 004000G005 ..(KA's) (***** END OF CATEGORY 8 *****) (********** END OF EXAMINATION **********) _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -}}