ML20126M293
| ML20126M293 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/17/1985 |
| From: | Lawyer S, Wilson B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20126M265 | List: |
| References | |
| 50-302-OL-85-01, 50-302-OL-85-1, NUDOCS 8506200284 | |
| Download: ML20126M293 (113) | |
Text
ENCLOSURE 1 EXAMINATION REPORT 302/0L-85-01 Facility Licensee: Florida Power Corporation P. O. Box 14042, M.A.C. H-2 St. Petersburg, FL 33733 Facility Name: Crystal River Unit 3 Facility Docket No. 50-302 Requalification examinations were administe d at Crystal River Nuclear Plant near Crystal River, FL Chief Examiner: M /7 #N Sandy Law er # ' Dats Signed Approved by: AA4
- i. %
B/uceA. Wilson,SectionChief n f T7 Datte Si@ned Summary:
1 Requalification examinations on March 4-7, 1985 Written and oral requalification examinations were administered to ten SR0s and seven R0s; three of the SR0s and two of the R0s passed these examinations.
The performance on the requalification examinations (29.4% pass rate) has resulted in a determination that Crystal River's requalification training program is unsatisfactory as of March 1985. Corrective actions are addressed under separate cover.
8506200284 850521 PDR ADOCK 05000302 G PDR
Enclosure 1 2 REPORT DETAILS
- 1. Facility Employees Contacted:
J. Alberdi, Manager Site Nuclear Operations and Technical Services, (E)
J. F. Belzer, Nuclear Operations Training Supervisor, (E)
R. C. Zareck, Nuclear Operations Instructor, (R/E)
V. R. Roppel, Manager Plant Engineering and Technical Services, (E)
L. C. Kelley, Manager Nuclear Operations Training, (E)
E. M. Howard, Director, Site Nuclear Operations, (E)
P. F. McKee, Plant Manager, (E)
G. L. Boldt, Plant Operations Manager, (E)
W. S. Wilgus, Vice President, Nuclear Operations, (E)
P. G. Haines, Licensing Engineer, (E)
W. L. Giles, Nuclear Training Instructor, (R)
J. P. Haerle, Nuclear Operator Instructor, (R)
C. D. Arbuthnot, Nuclear Operator Instructor, (R)
NOTE: "R" indicates present at examination review "E" indicates present at exit meeting
- 2. Examiners:
B. A. Wilson, NRC N. F. Dudley, NRC S. Lawyer, NRC*
B. F. Gore, PNL J. C. Huenefeld, PNL
- Chief Examiner
- 3. Examination Review Meeting At the conclusion of the written examination, the examiners met with facility representatives (identified in 1. above) to review the written examinations and answer keys. Specific facility comments and associated NRC resolution of those comments follow:
NOTE: Comments on questions duplicated between exams are only detailed once,
- a. SR0 Exam (1) Question 6.3 Facility Comment - This question is poorly worded. It could very logically be read such that any of the answers are correct.
i i
Enclosure 1 3 NRC Resolution - Review of the question in light of this comment revealed that the facility comment was accurate. The question was deleted.
(2) Question 6.4 Facility Comment - The facility reviewers were advised that a review of the exam key after final typing, but prior to administering the examination, revealed that the answer to 6.4 had to be changed from d to a.
NRC Resolution - The answer key was changed accordingly.
(3) Question-6.13 Facility Comment - The facility reviewers were advised that a review of the examination during its administration revealed that this question was ambiguous in that the cooling supply could be for the pump or motor. The same ambiguity exists in Crystal River operating procedure OP-605.
NRC resolution - The reviewers verified while onsite that the pump and motor cooling supplies are different. Answers a or d were accepted.
(4) Question 6.19 Facility Comment - The facility reviewers were advised that a review of the examination prior to its administration' revealed that both answers b and c were correct.
NRC resolution - Both answers b and c were accepted.
- b. R0 Exam (1) Question 2.14 - This is the same question as 6.13 addressed above.
(2) Question 3.9 Facility Comment - None NRC resolution - It was discovered during grading that the answer key contained a typo. The correct answer supported by the referenced material was d. The answer key was changed to indicate d as the correct answer.
4 (3) Question 4.15 Facility comment - The facility reviewers were advised that a review of the examination key prior to administration of the examination revealed the key was in error due to a typo.
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Enclosure 1 4 NRC resolution - The answer key was changed.
- 4. Exit Meeting At the conclusion of the site visit, the examiners met with representatives of the plant staff to discuss the results of the examinations. Those individuals who clearly passed the oral examination were identified.
.The following generic weaknesses were noted by the examiners during the oral examinations:
' System descriptions and control room reference material are inadequate in some areas.
Some non-operating shift personnel did not demonstrate as much familiarity with control room facilities as was desired.
Problems were identified in the knowledge of and ability to apply the required actions of abnormal and emergency procedures.
The cooperation extended to the examination team by the Operations Depart-ment was noted and appreciated. These assistances were conveyed to facility representatives at the exit meeting.
- 5. On April 3,1985, a meeting was held in the Region II office to discuss the written examination and the examination results. A summary of this meeting was issued by Region II on April 19, 1985. During the meeting, Florida Power Corporation (FPC) representatives noted that although they had concerns on a number of questions on each of the written examinations (approximately 15% of the questions) favorable resolution of those comments would not significantly change the overall pass / fail results. The following is a question by question commentary of the examinations with accompanying NRC resolutions. Most of the facility comments are stated verbatim from the April 3, 1985 meeting.
- a. R0 Exam (1) Question 1.9 Facility Comment - The utility raised a concern that although the question and answer were technically correct, the initial conditions were not specified and several answers would be logical depending on assumptions made by the candidate.
NRC Resolution - We concur with this finding and the question was deleted.
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Enclosure 1 6
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NRC Resolution - Question was deleted due to erroneous information ,
in facility supplied System Training Manual. '
(5) Question 3.2 Facility Comment - There are two correct answers to this question.
The answer key identified d as the correct response. Answer a is also correct. The ICS field power ABT is powered from VBDP-2 and VBDP-4. See Enclosure 2, STM chapter 504, Pg. 41.
NRC Resolution - Answer key changed to accept d and a.
Additional reference OP-7000, "120 Volt AC Vital Buses," supports answer a.
(6) Question 3.3 Facility Comment - There are two correct answers to this question.
The examiner drew this question from a sentence found in page 43-13 of the STM. Based on that sentence, answer c is correct.
A consideration of pages 43-12, and 43-15 will prove answer a to be correct as well. See Enclosure 3, STM-43-12, 43-13, 43-15.
NRC Resolution - Answer key changed to accept c and a based on additional information in STM-43.
(7) Question 3.5 No facility comment.
NRC Resolution - This question was deleted. Post exam analysis with NRC Test writing experts showed this question to violate several principles of correct multiple choice writing guidelines.
(8) Question 3.10 Facility Comment - The intent of this question is good, the way it is asked is unsatisfactory. The intent of the question is to determine if the operator can discriminate the Crank, Ready, and Run lights on the SSF section of the main control board. These indicators are required knowledge of all control room operators, they have not, however, been required to learn them by color. We require an operator to learn indicator colors only when the color is utilized as an indication. Some examples of indicators that require color recognition are the green / red lights on control switches, and the blue / amber lights on the ES status board. The
. color of the Crank, Ready, and Run lights have no meaning, the fact that they are illuminated is what provide the operator his required indication.
^
4 Enclosure 1 7 If we deal with this question as written, answers a, b, and c are all correct. Answer a.is correct per the answer key. Answer b is also correct, if you close a diesel output breaker, an amber light comes on indicating crosstie blocking is in effect on the other diesel's output breaker.
Answer c is correct because when you close the diesel output breaker a white light is lit at ANN window P-2-4 indicating "4 KV ES Bus 3A/3B Paralleled."
NRC Resolution - This question was deleted. The color of the lights in this case were determined to not be the sufficient indicator of equipment status.
(9) Question 3.~19 Facility Comment - All answers could be construed as incorrect.
Answer b is the required response per the answer key. Answer a discusses something called " safeguards control center starters."
i We have no component that fits that description. Answer c could be considered incorrect because the running make-up pump (HPI Pump) h tripped on a loss of voltage. Note that this assumption is encouraged by the use of the singular "it" in response c "until the generator comes up to speed it (indicating one pump) is not connected to the bus." Answer d is incorrect because a RB Spray pump requires both a HPI block 4 start permit, and >30 psi on 2 out of 3 RB pressure switches.
NRC Resolution - The terminology, " safeguards control center i starters" is contained in STM-15, page 6. Distractors c and d can be interpreted as being incorrect, only if the examinee makes assumptions that are not indicated in the question. Choice b is clearly incorrect without making any assumptions. No change to question or answer key.
(10) Question 3.20 Facility Comment - This question and answer are valid; however, it is not a common practice to require an operator to recall every start permit on every component in the plant. The control
- switches for the RB Purge fans are fitted with indicating lights i
that indicate the status of each of the fan start permits. The
- operator, by procedure, simply verifies that all permits are satisfied prior to starting the fan.
NRC Resolution - The RB Purge System is a system designed to limit I the release of radioactive materials during all modes of facility.
operations. We recognize, subsequent to the exam, that the Purge system at Crystal River is now only used in Modes 5 and 6.
, However, the NRC Draft NUREG-1122 "Kncwledges and Abilities i
4
Enclosure 1 8 Required of Nuclear Power Plant Operators: Pressurized Water Reactors," considers the RB Purge System relatively i=portant for both RO and SRO (distractor importance ratings generally between 2.5 and 3.5). No change to question or answer key.
(11) Question 3.23 Facility Cc=ent - This question requires an operator to cesorize detailed steps of a operating procedure. When shifting rods or
, rod groups, operators utilize a detailed procedure which lists each step and expected response. It is unreasonable to expect these steps to be co=mitted to me=ory.
NRC Resolution - We disagree. This question is intended to test knowledge of the Control Rod Syste= design features and inter-locks. This particular K/A has a relatively high Importance Factor in Draft NUREG-1122. No change to question or answer key.
(12) Question 3.24 Facility Co=ent - This question goes beyond the knowledge require-cents of a control rcos operator. There is no reason why an operator should co=it te recory the types of detectors utilized by the RDS. At CR-3 an operator is expected to be familiar with the function and ' operation of the RDS. He should be able to perform the required surveillance on the system and response to alar:s per AP-320 " Loose Parts Monitoring Syste=s." None of these responsibilities require a knowlecge of detector types.
NRC Resolution - Tnis cuestion was written from a recently developed Lesson Plan on the RDS. This Lesson Plan had apparently not been distributed to, nor read by licensed personnel. This question revealed a fundamental weakness in the Requal Progra= in that recently developed (or i= proved) training =aterials are not disseminated to licensed personnel until the subject is taught when these personnel are assigned to Requal training.
No change to question or answer key.
(13) Question 4.1 Facility Co cent - This answer is extremely misleading. The initiating event in this question puts the operator in EP-220,
" Pressurized Thernal Shock." The Im:ediate Action of this procedure is to " stabilize existing RC pressure and te=perature conditions" while the remedial action is to " reduce subcooling cargin to minimum." Nowhere does it say to " reduce RC pressure to 1000 psig". The question indicated that one of the answers would state the freediate or remedial action per the EP and it did not.
I agree that answer a does eeet the requirements of the remedial action, but it doesn't indicate the operators ability to handle a
Enclosure 1 9 PTS event. In fact, depressurizing would be in violation of the procedure since the RCS pressure / temperature combination fall within the thermal shock operating region. Based on conditions given, the proper procedure would be:
- 1. Stabilize " existing RC pressure and temperature."
- 2. Determine cooldown rate.
- 3. If rate was >100'F/hr. " maintain RC pressure and temperature at that point for >3 hours."
NRC Resolution - It was incorrect to assume that one of the choices would state the immediate or remedial actions per the EP.
The question asked what action should be taken. The question also assumed that the examinee would be knowledgeable of which EP to use based on the Entry Symptom (Tc <500*F). It was not apparent they knew which procedure to use. The facility comment seems to indicate that the procedure is deficient. No change to question or answer key.
(15) Question 4.13 i
Facility Comment - This question lists four valid methods of 4
determining which OTSG has suffered a tube failure and then requires the operator to state which one was omitted on the sixth page of follow up actions in EP-390. This is an unreasonable requirement, especially since all four choices are valid methods of making this determination.
NRC Resolution - While a candidate need not memorize steps of a procedure which are not immediate actions, the candidate must be able to describe conceptually the objectives and methods used to achieve those objectives for all emergency and off-normal procedures. Here, one method is conceptually in stark contrast to the other three in that the time it takes to get information is unreasonably long.
No change to question or answer key.
(16) Question 4.14 No facility comment.
NRC Resolution - Post exam review revealed the question to be confusing in that the stem of the question together with the correct answer utilized three negatives ("not" statements).
Question was deleted.
1:
- i. Enclosure 1 10 (17) Question 4.16 3 Facility Comment - This question requires the operator to . recall the step, setpoints, and. breaker numbers of a very infrequently performed operating procedure. The procedure in - question is i recuired to be. signed off stap by step as it is performed, and it -
is then . verified by the shif t supervisor. It is ridiculous to expect this level of memorization on a procedure that must be signed off' step by step.
1
. NRC Resolution - The correct answer was not intended to test
{ memorization but a concept, i.e., the pressurizer venting was via
, the Nuclear sample room sink hood or the . MUT gas space. A previous step in the procedure went unnoticed which rendered I answer d incorrect. Since there was no correct answer, question
! was deleted.
4 ,
(18) Question 4.19 4 Facility Comment - This question is invalid. Following a trip the j method of determining cooldown rate is not specified. The steps ,
j listed in question 4.19 deal with EP-390 and the calculation of
- j. cool down rate following an overcooling event when Tc is <500*F.
{ NRC Resolution - Since incore thermocouple readings are not
- i. continuously recorded, it would be highly unlikely that the information- in choice b would be available following a trip.
!- Therefore, choice b was clearly the correct answer regardless of
{ the procedure used. No change to question or answer key.
b (19) Question 4.20 l No facility comment.
NRC Resolution '- This question was deleted. - The usage of the-
- double negatives "except" and " incorrect" in the stem may have led i to candidate confusion.
- b. SR0 Exam (1) Question 5.2 Facility Comment - This question . is well- written and concise, l however, the answer on the answer key 'is incorrect. The answer ;
f key required d as the . correct' response when in reality a is i the correct answer.
NRC Resolution - Question was deleted since neither answer a or d could be adequately supported by available references.
. _ . . . . _ _ _ . . . _ _ . . . , . , _ . _ _ . , . . , . _ . - _ _ . . , _ , . _ . . . . . _ . , _ . ~ , _ . . . . _ . _ _ . - . _ _ _ _ _ _ _ _ ..... . _ _ ,-.._,..-_,m_
Enclosure 1 11 (2) Question 5.15 Same comment as on R0 Exam, Question 1.9 NRC Resolution - Question deleted.
(3) Question 5.20 Facility Comment - The correct answer to this question per the answer key is c, " Low quality steam." I can find no text book that defines the point at which steam quality changes from high to low. A far better choice could have been " wet saturated steam" or
" wet / saturated vapor."
NRC Resolution - We agree that the term " low quality steam" could be better defined. However, it is the only logical choice since none of the other choices could be interpreted as steam that contains some quantity of moisture.
(4) Question 6.5 Facility Comment - The manual operation of an atmospheric dump valve is not considered a normal operator function at Crystal River Unit 3. In our seven plus years of operation we have not yet had to utilize this method. Our procedure AP-990 " Shutdown from Outside Control Room" directs the Nuclear Operator to "obtain an available staff member" to perform this operation, therefore instructions on how to manually operate these valves have been provided at each valve. It is not reasonable to expect an operator to recall the specific steps of an operation he has never, and possibly will never perform when the steps are provided on a permanent instruction plate located at each valve.
NRC Resolution - Manual operation of the atmospheric relief valves is a required " follow-up" action to AP-990, " Shutdown From Outside Control Room". This action will probably be performed by an unlicensed staff member at the direction of a licensed staff member. 10 CFR 50.54(j) requires that when operating apparatus and mechanisms (such as atmospheric relief valves) it be done only with the knowledge and consent of a licensed person. We there-fore, believe it is reasonable to expect licensed personnel to be able to direct manual operation of these valves. No change to question or answer key.
(5) Question 6.7 Facility Comment - There is no correct answer to this question.
The ES system does not block load the battery chargers. If the breakers feeding the battery chargers were open, they would remain open. The battery charger is simply a load on an ES MCC.
NRC Resolution - Comment accepted. Terminology used in question was not appropriate. Question deleted.
i Enclosure 1 12 (6) Question 6.11 Facility Comment - Answer a is the correct response, however, c could also be construed as being correct if the examinee assumed the statement to infer that each RPS channel is powered from a separate vital bus.
NRC Resolutten - We disagree. Distractor c, in the context of distractors a and b should preclude making this assumption.
(7) Question 6.17 Same comment as RO Question 3.20.
NRC Resolution - Same as R0 Question 3.20.
(8) Question 6.19 Facility Comment - Answers b, c, and d are incorrect. Answers b and c were identified as acceptable on the answer key. Response d should also be accepted for the following reason. Answer d states "The RB spray pumps are not connected to the 4160 V ES bus until about 15 seconds after block loading begins." In reality, after 15 seconds (HPI block 4) an RB spray actuation permit is set.
In order to load the RB spray pump, RB pressure must exceed 30 psig on 2 out of 3 pressure switches.
NRC Resolution - We disagree. It would be unreasonable to make this assumption considering the acceptability of answers b and c.
(9) Question 7.3 Facility Comment - This question is totally unreasonable. It asks the examinee to recall which item out of a list of four items listed in OP-202 is not a requirement for escalating from mode 4 to mode 5. In a quick review of this section of OP-202 (section 6.4), I have identified no less than 21 items which must be completed prior to making this mode change. An operator should not be expected to recall from memory every required step of every operating procedure, especially when the procedure is required to be signed off step by step by the operator, and then verified by the shift supervisor prior to making the mode change. See OP-202, section 6.4.
NRC Resolution - On the surface it appears the question demands an unreasonable amount of memorization. The correct answer simply asks for an understanding of a concept, i.e., degassification cannot be completed while the RCS is still " cold". No change to question or answer key.
Enclosure 1 13 (10) Question 7.11 Facility Comment -
In reality, if there was an uncontrolled decrease in refueling canal water level, the level would probably stabilize at the seal plate. 'Therefore, choices a and c are only testing whether components should be stored 4 feet or 5 feet below the water level.
NRC Resolution - Comment accepted. This was not the intent of the question and the distractors were not sufficiently considered. Question was deleted.
(11) Question 7.12 Facility Comment - This question requires an unreasonable level of recollection. The condition stated has never, to my knowledge, occurred at CR-3. In the event it does, it most certainly would be corrected by use of a detailed procedure, not by memorized limits. To require an operator to recall every possible limit in every procedure is not realistic.
NRC Resolution - We disagree. The frequency of occurrence is not sole criteria upon which subject matter for questions can be judged. The event postulated in the question is covered by a Note or Caution statement in the Fuel Handling Procedure, FP-601, and would not be corrected by use of a detailed procedure.
(12) Question 8.1 No facility comment.
NRC' Resolution - Post-examination grading revealed that none of the four situations caused a Technical Specification action statement to be initiated within one hour. Question was deleted.
(13) Question 8.9 Facility Comment - Answer d was identified as the correct response on the answer key. In reality, since two makeup pumps are inoperable, and no mention is made as to why they are inoperable or when they will be returned to operability, answer b more nearly reflects the intent of the OSIM requirement. I feel that answer b or c should be considered correct since the initial conditions are not specific and are therefore open to interpretation by the examinee. See enclosure 2 " Policy Statement 84-1."
NRC Response - We disagree. Answer d is directly from the OSIM and does not infer intent or assumptions as does distractor b.
C' g Enclosure 1 14 (14) Question 8.15 Facility Comment - This question requires t.'1e operator to recall a Tech Spec action statement. This is a totally unreasonable requirement. Our operators are required to recall those items that are covered by Tech Spec LCO's and then given a copy of Tech Specs, be able to correctly interpret the required actions. By requiring an operator to recall this action statement you are in essence forcing us to memorize all action statements.
NRC Resolution - We believe this question is appropriate for several reasons:
- 1. At the time of the examinations the plant was beginning an extended outage where the postulated situation was most realistic.
- 2. There is no time limit specified in the action statement which infers that the action be taken te:ediately.
- 3. A fa=111arity of three separate Tech Specs (3.0.3, 3.8.1 and 3.9) would lead to the correct answer.
(15) Question 8.16 Facility Coneent - This question requires the operator to recall Tech Spec action statements. See co=ents on questions 8.15.
NRC Resolution - A knowledge of operator actions required within I hour by Technical Specifications will be-a continuing requirerent.
(16) Question 8.21 l Facility Comment - This question asked the examinee to state which of one of four listed operations did not meet the Tech Spec definition of a Core Alteration. The answer is based on a letter of clarification dated April 1, 1983. This letter was not
, reevaluated by the PRC and rereleased until March 18, 1985, approximately two weeks after the NRC exam was administered. See enclosure 3, inter 2ffice correspondence dated March 18, 1985.
NRC Resolution - We agree with the situation as stated above. The subject memo (4/1/83) was in effect at the time of the examination. It states "The following ... clarification of the 4
refueling Technical Specifications will apply to Refuel IV and future refuelings/ outages...". The fact that it was reviewed and reaffirmed by PRC just prior to head removal for Refuel V does not i
mean it had been rejected or had " expired". No change to question or answer key.
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Enclosure 1 15 (17) Question 8.24 Facility Comment - STS requires a maximum level in the OTSG based on degrees of superheat. The absolute maximum level allowed is 96% on the operate range (answer c). A more correct answer would be "per table 3.4-5, Maximum Allowable Steam Generator Level."
See enclosure 4 STS table 3.4-5.
NRC Response - The enclosure was the material upon which the question was based. While the quoted statement may or may not be a "better" answer, it was not one of the choices. Had it been utilized as the correct answer, there would have been serious objection to the question on other grounds. The question and cfted answer are valid and appropriate.
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CATEGORY 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERM 0 DYNAMICS, HEAT TRANSFER AND FLUID FLOW 1.1 d - Fuel heats up first Ref: NUS Module 3, Sec's 8.3-8.5 1.2 d Ref: NUS Module 3, Sec's 5.3 and 6.7 1.3 does not change 235 a-PubuildinreducesS'$f;OP-103,Rev.38,p.12U Ref: NUS Module 3, Sec 5 ; delayed neutrons 1.4 c Ref: NUS Module 3, Sec 10.1 1.5 a Ref: NUS Module 3, Sec 10.3 1.6 c - Warmer water in downcomer and core bottom; less power differential along core axis; more n leakage makes NIs read high Ref: NUS Module 2, Sec 16.5 NUS Module 3, Sec 8.4 1.7 c - Sm is stable (a); Sm burns out (b); independent of power level (d)
Ref: NUS Module 3, Sec 10.5 1.8 d - Restricting exit flow reduces head loss in the inlet piping, increasing pump suction pressure and NPSH Ref: NUS Module 4, Sec 6.5 g 1.9 a-T ave increases, increasing PZR level and pressure, PZR spray turns pressure before level increase stops Ref: STM-419, p. 30 1.10 d - Ref: AP-530, p. 4 1.11 c Ref: , STM-419, p. 15 1.12 b - Steam flow can entrain or push the water.
Ref: CR Fluids and Mechanics lesson, p. 47 1.13 b - Rhoi Dolta= Rho =-0.05263; 0.02594 Rho 2 = -0.02669 Final Rho = -0.02669 + 0.02594 = -0.00075 Ref: NUS Module 3, Sec 6.1 1.1
t
- o 0.050 kg = 0.952 1.14 b - Rhog Rho = 0.030
- kg = 0.971 CR,2= =50 x (1 - 0.952)7(1 - 0.971) = 50 x 1.67 = 83 Ref: MUS Module 3 Sec 12.1 1.15 a Ref: OP-103, plant curves 3.2 A&B 1.16 c Ref: SP-0312 STM-420-1.17 b Ref: CR Lesson RQ-84-7E Degraded Core Recognition and Mitigation 1.16 a - 420,000 scf; 26,000 scf(b); 140 scf(c); 1,320 scf(d)
Ref: C. R. Lesson Fundamentals of Natural Circulation 1.19 a - Neutrois travel further with reduced water density Ref: NUS Module 3, Sec's 9.4, 9.5 1.20 c - Would lead to similarity , not difference Ref: NUS Module 3, Sec's 7.2-7.5 l STM-1, pp. 21-28 1.21 a - Doppler is a 238 U cross section effect Ref: NUS Module 3, Sec's 8.2,11.3 1.22 d - Super heat region larger, plus 50% FP is closer to 15% FP (lov level limits) than 100% FP.
Ref: STM-24, p. 4 I 1.23 a - Rotameter is not based on Bernouli's principle.
Ref: CR Fluids and Mechanics Lesson, pp. 50-53 1.24 1) T, 2) F, 3) T, 4) F Ref: CR Core Power Of stribution Handout i
3 i
1.2
CATEGORY 2. PLANT DESIGN, INCLUDING SAFETY AND EMERGENCY SYSTEMS 2.1 d - Elevated temperature (a); trip is lo p, not hi T (b); only closure to 60% open trips FWP (c)
Ref: STM-27, pp. 3, 7, 14, 43 2.2 a - 6 starts max. (b); 3 hr supply in day tank (c); air operated booster piston supplies oil (d)
Ref: STM-10, pp. 1-8 2.3 b - Injection after blowout, before rapid clad T increase is most effective timing Ref: STM-4, p. 6 2.4 b - 7 gpm enters RCS (a); pressure equally divided between 3 seals (c);
i computer alarm at 970 psid (d)
Ref: STM-420, pp. 1-3, 16 2.5 b - Li-7(n.n')a,T Ref: C.R. Letter TRA 85-0013, 1/29/85 2.6 c Ref: STM-4, p. 8 2.7 d Ref: STM-23, p. 11 2.8 d l Ref: STM-4, p. 11 i
j 2.9 b - MFWPs supply EFW nozzles i
Ref: C.R. letter TRA 85-0013, 1/29/85 2.10 c Ref: OP-403, Sec 4.7.10, p. 4 4
2.11 d - 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> total (c)
- Ref
- STM-4, p. 16; OP-404 2.12 c Ref: CR Main Steam Requal Lesson, pp. 20, 69, 72, 69.9.1 2.13 c Ref: OP-202, pp 3, 22 FP-302-661, sheet 3 of 4 2.1
2.14 a A O Ref: OP-605, p. 29 2.15 b Ref: TS B 3/4 7-2 2.16 d Ref: OP-703, p. 9 Drawing EC-206-017 2.17 b - High heater level does not require more extraction steam Ref: C.R. Letter TRA 85-0013, 1/29/85 2.18 c - Main steam is normal supply to Gland Steam System Ref: STM-39, p. 4 Ag 2.19 c - Deluge must be manually operated
-- v Ref: STM-38, p. 3 2.20 a - Inoicates operation with <10 psig pressure,. too low; delta p alarm is 20 psi, OK (b); indicates control oil p > 80 psig, OK (c); 110 psig is normal p (d)
Ref: STM-27, pp. 22, 23 2.21 c - Separate pump during startup Ref: STM-420, pp. 6-12 l
2.22 d Ref: STM-18, p. 1 2.23 b - A float trap automatically drains tank Ref: STM-23, p. 5, 6 2.24 1) F, 2) F, 3) T, 4) F -
Ref: STM-4; OP-401 p. 2110 2.2
I l
t F
T i,
l CAT' G 3M 3. IMST:MYT5 #0 C3G3 5 t
3.1 0 - 5:x!!:s t:iesulf ate m im;er ccens :
i Ref: STN 3, p. 3 I SG- i l 3.2 d L 110 -2 is ;referre so;rce (a); A37 is :se: f:r fiele, ci -31 E i 4 s::;1y (2); lag lirts if 5-1 arx: 5-2 fail (c) !
Ref: 53-504, ;c. 43,11 i
3.3 c#w- Particulate. I, gas (a); gas m1y (wits filte s) ( ); ps ylf (:)
Ref: 575 13, ;o. 12-11 l
3.4 d - Lancs 11@: m1y if tista:1e is rese: (a}; L I is still Ofrasse: (a;;: :
rese: is at 9% ;si) ( ); bistatles trf; a- 533 :si;, a .: L '.
initiates een rese: (c) l Tef: STw-11, ;;. 21,25 ;
3.5 a - Cely 5 gro::s f teste e:r;1pnet: (2); E salves are ;a-:-st~.<e teste i (c); auto test s=iten sele::s cee test ;~x;: o:17 (:) !
l Ref: 57W 11, :c. 19-21 3.5 e - Char.sel O f ; ass causes 2 o;; cf 31 ;1: (a); m1y : e tri s:rt ; cm- r ta:: has cce* (t); all fo;r char els nay oe : late: 1-:: "S*ste:w- l Bf ass
- ces assisistrative rw;irenetts are me: (:) [
tef: Siv-510, p;. 25 13 [
3.7 : - 5:e+: ce:en:s ::ra ere : r e-: (a); :Icclis; is electrical, n:: re:rati-cal (2); differeitf al c: ere- Or te: tim is re e se sitive ese :: .
c: mal :ero c rrett (c) !
Ief: 575-15 ;;. 13-19, 23 l 3.E d - 5: r lui : avels re:;;f re (a); samas are sete:te: (:); m;; ires ;
NI-5 er 513: 41-7 cr 3 (c) i 7ef: 5 3-5, ;. 13 3.9 / el I
- ef: CF-232 Se: 5.1.29, ;.12 r
I 3.10 a '
Ref: STw-13, c. 42
[
j 1.11 c - False 1:=-teret is:ica:f o ; tip-letel 19:t: stir .oai: :a.s e f a}, it! !
l a-c (d) !
Ref: T -203, ; 3
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3.12 c !
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1 1
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3.13 c - Only two channels (a); by pass allowed below 725 psig, is not automatic (b); reset above 600 psig is manual (d)
Ref: STM-27, pp. 63-65 3.14 d - Fault reset resets faults and inhibits after faults are cleared Ref: STM-12, Sec. 2 3.15 c - Thermocouples are smaller and faster Ref: STM-7, pp. 2-7 3.16 a Ref: STM-420, p. 15 3.17 b - Turbine reset is not required Ref: STM-27; C.R. Letter TRA 85-0013, 1/29/85 3.18 d - Track requires both FW controls in manual Ref: STM-504, p. 77 3.19 b - Battery charger feeder breakers stay closed.
Ref: STM-15, pp. 6, 35 3.20 a - Exhaust fans must operate before start of supply fans Ref: STM-22, pp. 22-25 3.21 c - Signal is higher of ave of 586 or 788; hi flux, flux / delta flux / flow, power / pumps (a)
Ref: STM-6, pp. 20, 21 3.22 c Ref: OP-204, p. 11 3.23 d - SEQ lamp stays on Ref: OP-502, pp. 7, 9, 16, 33 3.24 1) E, 2) A and E, 3) E, 4) B
.i Ref: CR Reactor Diagnostic System Lesson i
i l .
3.2 i
CATEGORY 4 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTRC:
4.1 a - This is 50'F subcooled, as required.
Ref: EP-220, p. 2 4.2 b Ref: OP-210, Sec 6.3.5, p. 11 4.3 d - One SG lo level limited will not prevent (b); incorrect statement (c).
Ref: OP-204, p. 20 4.4 c Ref: AP-530, pp. 2, 3 4.5 d - Subcooling margin is normally less than 50*F during power operations (b); not required by TS 3.1.1.1.1 (c)
Ref: OP-204, p. 7 4.6 b Ref: RP-101, Rev. 20, p. 17 4.7 c Ref: RP-101, Rev. 20, p. 23 4.8 a Ref: OSIM, V-22 4.9 See attached pages , 3 Ref: AP-580 i
4.10 See attached pages -
Ref: EP-290 1
4.11 See attached pages ?'
Ref: AP-380 4.12 See attached pages
- Ref: EP-140 4.13 d - Not required--takes longer than others.
Ref: EP-390, p. 6 4.14 d - May move either in or out.
M Ref: OP-502, pp. 28-30 /
4.15/lb)OP-209 Ref: j 1
. 4.1
{ il, 4.16 d - Final venting is by gas space sample lines.
Ref: OP-202, pp. 18-20 4.17 b Ref: AP-542, p. 2 4.18 d Ref: AP-245 4.19 b - Only Te is used.
Ref: EP-220, p. 3 4.20 b h Ref: OP-209, Sec 4.12, p. 4 4.21 c Ref: OP-302, Sec 4.14, p. 5 OP-204, Sec 4.2.2, p. 5 OP-210, Sec 4.14, p. 3 I
4.2 l
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RPSA REV 02 Date 05-25-84 AP-580 ACTIONS
)
IMMEDIATE REMEDIAL
- 1. Ensure GRP 1-7 rods (.07 1. Open 480V BKRs inserted:
- 3305 e Depress"ReactorTrip"(.Q])
pushbutton - 3312.
e Observe " TRIP CONF" 2. Start boration:
light litondiamond[,0 panel. a. Open 8WST suction
- b. Start 2nd MVP
. c. Open MUV-24 Ca, d. Establish letdown path to RCBis.
- 2. Ensure main turbine TVs and 1. Close MSIVs.
GVs fully closed.
- 2. Select " ATMOS"on "TURB.
BYPASS VLV" switch.
- 3. Ensure main block valves 1. Trip both MFPs.
closed.
- 2. Refer to AP-450, Emergency Feedwater Actuation.
- 4. Ensure low load block valves 1. Trip both MFPs.
closed.
- 2. Refer to AP-450. Emergency Feedwater Actuation.
O j AP-580 Page 2 of 11 RPSA l
-- y Eh i
V 1 02 Date 05-25-84 AP-580 RP4A REV t
ACTIONS (Cont'd)
REMEDIAL IMMEDIATE O
- 1. Open suction from 8WST.
- 5. Ensure PZR level 1 50".
- 2. Start 2nd MVP.
- r Open MUV-24.
' 3.
i 4 Close MUV-51.
%s 1
Close MUV-51, Letdown Block Close MUV-49, Letdown
- 6. Containment Isolation.
Orifice Bypass.
i 4
Manually control STM HOR PRESS
i
' controlling at 1010 PSIG.
o HOR PRESS controller t
S.!
e Turbine Bypass Valves in j " HAND" l .
ER e Atmospheric Dump Valves in i " HAND".
i .
i ;
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! I RPSA I
L=s AP-580 Page 3 of 11
. t
HKC 10/15/84 RPSA REV 02 Date 05-25-84 AP-580 l >
ACTIONS (Cont'd)
I IMMEDIATE REMEDIAL
'8 . Ensure GEN output BKRs open:
e BKR 1661 e 8KR 1662.
2
%s f
- ,~
& y %
. .v: , l
, AP-580 . Page 4 of-11 RPSA
, V --
- ~. , ,
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J. ,_cu- jf.55 "'I*)
ICC REY 03 Date 09-20-84 EP-290 r
ACTIONS IMMEDIATE . REMEDIAL
- 1. Ensure full HP! flow. Ensure: .
- a. 2 HP! pumps are running. ,
- b. Open:
. o MUY-23 o MUV-25 o MUY-24 o MUY-25.
- c. HPI flow is > 500 GPM total flow.
- 2. IF LPI is delivering flow, Ensure:
Na THEN maintain maximum LPI o 2 LPI pumps are running.
TTiv.
o Open: -
- DHY-5 OHY-6.
- 3. Ensure OTSGs are at 95% on a. Select "CLOSE* on:
the operating range.
o FWY-34 o FWY-162 o FWY-35 o FWY-161.
- b. Trip both MFPs.
- c. Ensure both EFPs start.
- d. Slowly raise OTSG 1evels to 951 using:
.o FWY-162 (A-075G) o FWY-161 (3-0TSG)
~
1
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- o. ,
1 ESSA REY 03 Date 09-20-84 AP-380 ACTIONS IMMEDIATE REMEDIAL l, 1. If[ RC PRESS < 1500 PSIG, 1. Bypass ES actuation.
THEN depress "HPI Actuation" 2. Return ES equipment to STBY 7usEbutton "A" AND "B". status.
, 3. Go to VP-580.
Z, 2. Trip all RCPs. Open affected 6900V BXRs:
o 3101 o 3103 o 3102 o 3104 "I
- 3. Ensure HPI trains start: Notify AB operator to start affected pump (s) at 4160V ES y o 2 HPI pumps switchgear.
,} o SWPs S UI- /4 # # C f o RWPs. A WP. ut2.8 f 4. Ensure BWST suction valves Notify AB operator to open open: af fected valve (s) locally.
o MUV-58 o MUV-73.
T l i !
7 5. Ensure HPI valves open: Notify AB operator to open af fected valve (s) locally.
o MUV-23 o MUV-25 .
o MUV-24 o MUV-26.
I' AP-380 Page 3 of 25 ESSA I
W W
a ESSA REY 03 Date 09-20-84 AP-330 ;
ACTIONS (Cont'd)
IMME D I ATE REMEDIAL
'T
$ o DHPs fME~3N "3 o 4160V ES .
7 o DCPs DCI '#4 O o 480V ES fg o R W P s . g w a* - 3 4
- 3 8 o 4150V ES.
// 7. Ensure EDGs start. .
Motify AB operator to start ,
affected diesel locally.
5
/ J- S. Ensure diverse containment isolation actuation.
fjj 9. Place R5 susp pump in Motify AB operator to open
" PULL-TO-LOCK": affected SKG at MCC:
o WDP-2A o Reactor 3A 2 o WOP-23. o Reactor 332.
I i
AP-380 Page 4 of 25 ESSA W
/
tt
- 4. ng~p h w~a t
ERC . REV 00 Date 06-08-83 EP-140
! ACTIONS !
IMMEDIATE REMEDIAL .
1 i
! l
- 1. Start emergency boration: 1. Adjust batch controller to 1000 GAL.
- a. Establish letdown fb w to MUT > 40 GPN
- 2. Select RC8T with highest baron concentration.
- b. Open CAV-60
- 3. Establish flow to MUT.
c.
Start Boric Acid Pump
. 4. Open CAV-57.
e CAP-3A
- 5. Start Boric Acid Pump:
OR e CAP-3A e CAP-38.
S.E.
e CAP-38.
d l
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. ENCLOSURE 3 b U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR REQUALIFICATION EXAMINATION Facility: Crystal River - 3 Reactor Type: Babcock & Wilcox Date Administered: March 5, 1985 Examiner: 8.F. GORE Candidate:
INSTRUCTIONS TO CANDIDATE:
Use separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheet. Points for each question are indicated in parenthesis after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up four (4) hours after the examination starts.
Category % of Candidate's % of Value Total Score Cat. Value Category
- 24. Jet" 25 1. Principles of Nuclear Power Plant Operation, Thermodynamics, Heat Transfer and Fluid Flow 2!f.39' 25 2. Plant Design Including Safety and Emergency Systems i 23 H&t- 25 3. Instruments and Contrcis SCL-4MF 25 4. Procedures: Normal, Abnormal, Emergency, and Radiological Control
- f3400' TOTALS Final Grade %
All work done on this examination is my own; I have neither given nor received aid.
Candidate's Signature i
i f
CATEGORY 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW (25.0 Points) 1.1 In the event of a rod ejection accident, which will be the first reactivity coefficient to insert negative reactivity? (1.0)
- a. Moderator temperature coefficient.
- b. Pressure coefficient.
- c. Void coefficient.
- d. Doppler coefficient.
1.2 It takes less reactivity to go prompt critical at: (1.0)
- a. BOL because of the higher value of beta effective.
- b. BOL because of the lower value of beta effective.
- c. EOL because of the higher value of beta effective.
- d. E0L because of the lower value of beta effective.
1.3 Which of the following correctly describes the effect of increasing core life? (1.0) a
- a. decreases and SUR increases for a given reac-tfftyinsertion.
- b. Stuck rod worth at hot zero power increases.
l c. Overcooling transient becomes less severe.
1
- d. The SUR 5 minutes after a trip becomes more negative.
1.4 Which of the following radioactive isotopes, if found in (1.0) the reactor coolant, would NOT indicate a leak through the fuel cladding?
- a. I - 131
- b. Xe - 133
- c. Co - 60
- d. Kr - 85 i
1.1
1.5 The reactor is brought to 10-8 amps two hours after a trip (1.0) from 100% FP at equilibrium xen n c nditions. In order to maintain power level at 10-8 amps for the next hour, what ;
will have to be done with the control rods? l
- a. They will have to be withdrawn.
- b. They will have to be inserted.
- c. They will have to be withdrawn initially, then inserted to compensate for xenon burnout.
- d. They will have to remain at a constant position because
' the rate of xenon burnout is almost exactly matched with the post-shutdown indirect xenon production rate.
1.6 RCS boration may reduce power without CRDM motion. Which of the (1.0) following statements best describes the consequences of this method of power reduction?
- a. Imbalance becomes less negative and Power Range NI calibration becomes less conservative.
- b. Imbalance becomes more negative and Power Range NI calibration becomes less conservative.
- c. Imbalance becomes less negative and Power Range NI i calibration becomes more conservative.
t
- d. Imbalance becomes more negative and Power Range NI
- . calibration becomes more conservative.
1.7 Which of the following statements about Sm-149 is true? (1.0)
- a. It is removed from an operating reactor by burnout and radioactive decay,
- b. When a reactor is restarted after a temporary shutdown Sm-149 concentration increases for several days. '
- c. It .1as less effect on reactor operation than Xe-135 due to its' smaller fission yield and smaller microscopic neutron cross section.
- d. The equilibrium concentration of Sm-149 at 50% FP is about two thirds of the equi, librium concentration at 100% FP.
j !
1.2 1
1.8 Which of the following will increase the NPSH of the (1.0) condensate pumps?
- a. Increasing the vacuum in the condenser.
- b. Increasing the temperature of the condenser hotwell,
- c. Restricting the flow from the condenser to the condensate pumps.
- d. Restricting the ficw exiting the condensate purps.
1.9 Which of the following statements best describes parameter (1.0) changes in the pressurizer following a rapid load reduction of 15% FP?
- a. Pressurizer pressure and level will increase, with pressure increase stopping before level increase stops.
- b. Pressurizer pressure and level will increase, witn level increase stopping before pressure increase steps.
- c. Pressurizer level and pressure will decrease, with pressure decrease stopping before level decrease stops.
- d. Pressurizer level and pressure will decrease, with level decrease stopping before pressure decrease stops.
1.10 Why should RC cooldown on natural circulation not exceed (1.0) 10*F/hr for Tc > 280*F?
- a. to prevent exceeding brittle fracture limits of the reactor vessel.
- b. to ensure adequate mixing of HPI injection water with RC flow into the downcomer.
- c. to ensure that adequate heat removal tnrougn the OTSGs is possible without having to increase level above 50%.
- d. to prevent rapid and erratic changes in pressurizer level from occurring due to bubble formation in tre vessel head.
1.3
1.11 On reactor trip, the steam header pressure setpoint is (1.0) increased by 125 psi to reduce RCS shrinkage by elevating T Which of the statements below best explains the eff!c.t of this bias increase?
- a. After reactor trip T exceeds c OTSG T sat-
- b. After reactor trip Th decreases and approaches Tc '
- c. After reactor trip the OTSG Tsat is increased.
- d. After reactor trip OTSG 1evel is maintained on low level limits.
1.12 Which of the following actions or occurrences is likely to (1.0) cause water hammer?
- a. Maintaining the discharge line from an auto starting purp filled with fluid.
t
- b. Water collecting in a steam line,
- c. Pre-warming of steam lines.
- d. Slowly closing the discharge valve of an operating pump.
1.13 Reactivity is added to a shutdown reactor by rod withdrawal, (1.0) increasing keff from 0.950.to 0.974 if rods are again withdrawn, adding the same amount of reactivity as in the first withdrawal, the reactor will be:
- a. Subcritical by more than 0.1% delta k/k.
- b. Within plus or.minus 0;1% delta k/k of critical.
- c. Supercritical by more than 0.1% delta k/k, but not prompt critical.
- d. Prompt critical.
1.14 The reactor is shutdown by 5% delta k/k with a neutron (1,0) count rate of 50 CPS. Rods are withdrawn inserting 2%
delta k/k reactivity. Which of the following is closest to the resulting final neutron count rate?
- a. 70 CPS
- b. 80 CPS
- c. 90 CPS
- d. 100 CPS 1.4
1.15 Which one of the following statements is TRUE concerning (1.0) the change in differential baron worth (% delta k/k) with RCS boron concentration (range of 0 to 1800 ppm) and Tave (range of 532'F of to 579*F)?
1.16 The amount of heat being added by the reactor coolant pumps: (1.0)
- a. Is less than the RCS heat loss to ambient at operating temperature.
- b. Is less than the amount of heat being lost to letdown at operating temperature.
- c. Causes total OTSG thermal output to be greater than the thermal output of the core itself.
- d. Is insignificant at normal operating temperature.
1.17 During a LOCA with a resultant loss of subcooling margin, (1.0)
Reactor Coolant Pumps (RCPs) are secured for which one of the following reasons,
- a. To prevent pump damage resulting from operation under two phase conditions.
- b. To prevent core damage resulting from phase separation upon subsequent loss of RCS flow.
1.5 l l
J
1.18 Which of the following sources can potentially introduce (1.0) the largest (in standard cubic feet) amount of non-condensible gas into the RCS?
- a. Zirc-water reaction.
- b. Core Flood tanks.
- c. Pressurizer steam space.
- d. 100% failed fuel.
~
NOTE: The following questions ask you to select the incorrect, or negative, response from among a list.
1.19 Which of the following statements about control rod (1.0) worth is INCORRECT?
- a. Integral rod worth decreases with increasing Tave,
- b. Integral rod worth increases'toward ECL.
- c. Differential rod worth is least near full insertion and full withdrawal.
- d. The integral worth of two control rods can be either larger or smaller than the sum of their integral worth when inserted separately.
1.20 Which of the following is NOT a reason why the integral (1.0) rod worth curve for an APSR 1s shaped differently than
- the other rod groups?
- a. Active poison length is 3 feet.
- b. Neutron flux is higher near the center of the core.
- c. Neutron poison materials are the same in APSRs as in other control rods.
- d. Differential rod worth depends on flux level at the rod position.
t 1.6
1.21 Which of the following statements about burnable poisons (1.0) is NOT true?
- a. Including burnable poison in the fuel affects the doppler coefficient.
- b. Including burnable poison in the fuel affects the moderator tegerature coefficient.
- c. As core life increases burnable poison effects par-I tially compensate for fissile depletion of the fuel.
- d. As core life. increases burnable poison effects par-tially compensate for fission product buildup effects.
j 1.22 Which of the following statements about heat transfer regions (1.0) along the OTSG tubes is NOT true?
- a. The film boiling region is always the smallest region.
- b. The nucleate boiling region increases with increasing
! power.
- c. As power increases the superheat region decreases and the film boiling region remains the same.
l l d. At 50% power the superheat region is the sane size as the nucleate boiling region.
1.23 Which of the following flow ceasuring concepts is NOT based (1.0) on the square root of a measured pressure difference?
- a. Rotameter
- b. Orifice
- c. pitot tube i d. Venturi l
l l
l
! 1.7 l
l
o
'1.24 Answer the following statements concerning core power distribution and thermal design limits TRUE or FALSE.
- 1. Over core life, the limitations on negative axial (0.5) power imbalance become less restrictive.
- 2. Tech Spec limitations on quadrant power tilt only (0.5) apply in Mode 1 above 50% of Rated Thernal Power.
- 3. The linear heat rate is limited to prevent centerline (0.5) fuel melt, while DNBR is limited to prevent fuel clad failure.
4 There are Tech Spec safety limits on both axial power (0.5) imbalance and quadrant power tilt.
1 END OF CATEGORY 1 1.8
. .~. - - . _ _ _ _ .
CATEGORY 2. PLANT DESIGN, INCLUDING SAFETY AND EMERGENCY SYSTEMS (25.0 Points) 2.1 Which of the following statements about the Feedwater (1.0) system is true?
.a. Deaeration of condensate in FWT-1 is accomplished by spraying condensate into a deaerator section at reduced temperature.
- c. During operation with one main FWP, if the associated suction valve FWV 14 or 15 drifts 10% from full open, 1
the pump will trip in 30 seconds.
- d. Low pressure steam for each MFP turbine is supplied through its upper steam chest, passing through " poppet" 4
type valves to the upper turbine nozzles.
2.2 Which of the following statements about the Emergency Diesel (1.0)
System is true?
- a. Operator override valves parallel the air start solenoid valves, allowing manual starting of each engine, if necessary,
- b. The air start reservoirs for either diesel are designed to allow at least 8 starts without recharging.
. c. The fuel oil day tank for each diesel must be refilled once per day during operation at rated power.
- d. During startup oil is supplied to the main drive end bearing t by an auxiliary gear-driven pump.
2.3 Which of the following is the most important reason why Core Flood (1,0)
Tank pressure is carefully controlled?
- a. The borated water level in the CFTs can only be read on existing instrumentation to plus or minus one-half foot.
- b. During a large LOCA CFT injection will occur immediately af ter i " blowout" of water from the lower part of the core occurs.
! d. The LPI system could not be designed to inject at CFT pressure.
1 l 2.1
2.4 Which of the following statements about the RC Pump seals is true? (1.0)
- a. Under normal conditiens the seal injection flow to each pump divides, with 3 gpm passing the first seal and 5 gpm entering the RCS.
- b. If two pump seals fail the third can take the full RCS pressure,
- c. Two pressure reducing devices carry a small leakage flow in parallel with the seals so that under normal conditions the RCS pressure is equally divided between seals one and two.
- d. An alarm will scand if pressure in the third seal cavity exceeds 470 psig.
2.5 Lithium-7 is a concern in the RCS for which of the below listed (1.0) reasons?
- a. Too much of it gives an excessively low pH.
- b. It is a source of Tritium.
'I
- c. It adds to post shutdown radiation levels in the containment.
- d. It fouls heat transfer surfaces.
2.6 Which of the following statements is true of the High Pressure (1.0)
Injection System (HPI)?
- a. The HPI pumps nay be started at their local 480V breakers i located on ES MCC AB on the 119' elevation.
- b. The system has been designed such that during a low pressure i situation, as long as only two HPI pumps are running, pump runout is not an operational concern.
- c. Automatic initiation of HPI by the ES system starts both selected HPI pumps and all pumps necessary for LPI.
- d. The LPI pumps are automatically lined up to supply the suction of the HPI pumps from the RB sump when the BWST reaches a specified minimum level.
4 i
2.2 L
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2.7 The purpose of the 85 psig nitrogen overpressure in the NUCLEAR (1.0)
SERVICES SURGE TANK is to: ;
- b. Reduce the general corrosion rate.
- c. Meet ASME code standards,
- d. Prevent inleakage from the containment during a LOCA. ,
2.8- Direct cooling of the HPI pumps is provided by which of the (1.0) following combinations:
- a. Nuclear Services Cooling water and Decay Heat Seawater
- b. Decay Heat Closed Cycle Cooling water and Decay Heat Seawater
- c. Decay Heat Closed Cycle Cooling Water and Secondary Services Closed Cycle cooling water. ,
- d. Nuclear Services Cooling water and Decay Heat Closed Cycle cooling water.
2.9 Loss of all Reactor Coolant Pumps will directly cause which of (1.0) the following:
- a. Close the MFW and the Low load Block Valves and open the SU Block Valve.
"* Close the SU Block Valve and open the EFW Block Valve.
b.
, c. Close the EFW Block Valve and open the SU Block Valve.
- d. Start the EFW pumps.
2.10 Why shouldn't hydrazine be added to the RCS during (1.0) operation of the makeup demineralizers?
- a. Because the hydrazine will be removed by the demineralizers, and therefore wasted.
- b. The demineralizer resin does not perform satisfactorily at the low temperatures at which hydrazine is used.
- c. Hydrazine chemical reaction with the demineralizer resin could result in release of chlorides.
- d. If high 02 levels in the RCS warrant hydrazine addition, a potential source may be the demineralizer and therefore it should be off-service. .
2.3
i l
2.11 Which one of the following statements about the Low Pressure (1.0)
Injection (LPI) system is true?
- a. The reason Na0H is added to the discharge side of the LPI pumps is to prevent corrosion of the impellers.
- b. Actuation of LPI by the RC pressure reaching 500 psi l
opens BSV-36 and BSV-37 causing NA0H to be added.
- c. The DH pump operation in the recirculation mode (80-100 gpm) must be timed and limited to 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> total over the life of the pump.
- d. The maximum allowable flow rate per DH pump is 4000 gpm.
2.12 Which of the following statements about the Turbine Bypass (1.0)
Valves is correct?
- a. The bypass valves are air opened and spring closed, and operate off a +10V to OV signed at the EP convertor,
as a zero volt signal, thus an EP convertor output signal j of zero psig, causing the valves to fail as is,
- c. Local manual operation of the TBV requires isolation of t
instrument air, opening of the position controller handle valve, opening the equalizing valve, and inserting a pin to connect the handwheel to the valve stem.
- d. On loss of instrument air the lack of a close signal from j the electropneumatic positioner will result in the TBV l
remaining at its current position. ,
2.13 The Makeup valve, MVV-31, may be bypassed manually. The bypass (1.0) is normally throttled to maintain which one of the following?
- a. 8 gallons per minute Makeup flow to ensure that Fakeup i
never falls below that required for seal return flow,
- b. Greater than 15 gallons per minute to ensure that the running Makeup pump does not over-heat.
l c. Greater than 15 gallons per minute at all times when the RCS is above a minimum temperature specified by procedure.
l
- d. Equal to 1 gallon per minute at all times to minimize l
thermal shock to the Makeup nozzle.
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. 2.14 The motor driven EFW pump (EFP-3A) is cooled by which one (1.0) of the following?
- a. NSCCC
- b. SSCCC
- c. DHCCC
- d. Its own discharge.
2.15 The condensate water storage tank with minimum water volume (1.0) is sufficient to maintain the plant in HOT STANDBY for how nany hours with steam discharge to atmosphere?
- a. 8
- b. 24
- c. 50 ,
- d. 100 NOTE: The following questions ask you to select the incorrect, or negative, response from among a list.
2.16 Select the CORRECT statement regarding the 480V to (1.0)
! 120V AC distribution system,
- a. 120V AC Vital buses 3A and 3B are supplied from either DC Bus 3A or 480V ES MCC 3A-1 via dual input inverters.
- b. DC Bus 3A or 3B can be supplied from a standby battery i charger that is powered from 480V ES MCC 3AB.
- c. ICS 'X' power supply is from 120V AC Vital Bus 3A and ICS 'Y' power ,upply is from 120V AC Vital Bus 38.
i d. Each 120V AC Vital Bus can be transferred from its normal inverter feed to the 480V ES MCC alternate feed via a Static Switch, i
s 1
2.5
2.17 Which of the following statements is NOT a true statement (1.0) concerning the design and operation of a feedwater heater?
- a. A high level in FWHE 5A will open the heater dump valve
- and close the extraction non-return valve feeding it.
- b. The extraction steam flow rate must be increased as the heater level rises above normal.
- c. A high level in a feedwater heater will result in a decrease in its efficiency.
- d. The feedwater heater subcooler utilizes the heater inlet water as a cooling medium.
2.18 Which of the following statements is NOT true about the (1.0)
Auxiliary Steam System?
- a. During normal operation Aux Steam is supplied from Main Steam. ,
- b. Aux Steam may be supplied by Unit 1.
- c. Aux Steam is the normal steam source for the Gland Steam System.
- d. Aux Steam header pressure is maintained by supply regulators set for 140 psia by either the normal (ASV-27) or backup (ASV-26, ASV-184) regulators.
2.19 Which of the following plant areas is NOT automatically (1.0) protected by the Water Spray Deluge Fire Protection System?
- a. Startup transformers *
- b. Unit auxiliary transformer
- c. Charcoal plenum in Auxiliary Building
- d. Hydrogen seal oil unit 2.6
2.20 Which of the following conditions for the Main Feedwater pump (1.0) and turbine oil supply system is NOT indicative of proper operation?
- a. Lighted red light on oil test panel associated with the DC oil pump.
- b. Lube oil filter delta p less than 20 psi,
- c. Lighted red light on oil test panel associated with either AC oil pump.
- d. Control oil pressure equals 110 psig.
2.21 Which of the following statements about the RC pumps is (1.0)
NOT true?
- a. Inertia of the pump flywheel allows adequate coastdown to limit the minimum DNBR on loss of pump power,
- b. An impeller located below the lower pump seal pumps 70 gpm through the lower seal chamber and RC pump cooler,
- c. Oil pressure for the anti-reverse rotation device is automatically provided during startup by a pung vane near the thrust bearing.
- d. The pump flywheel is cooled by air whi:h transfers heac to the NSCCC System.
2.22 Which of the following chemicals are NOT added to the (1.0) indicated systems?
- a. Hydrazine is added to the EFW system.
- b. Lithium Hydroxide is added to the RCS.
i c. Sodium Hydroxide is added to the Building Spray System.
- d. Ammonia is added to the RCS.
I i
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, 2.7 i
2.23 Which of the following statements about the Instrument Air (1.0)
System is INCORRECT?
- a. Compressors are powered from Reactor Aux Bus 3A and 38.
- b. Condensation collected in each aftercooler must be drained regularly by manually opening the tank drain valve.
- c. Closure of the crosstie valve IAV-30 will not prevent service air from entering the IA system on low IA pressure.
- d. Although moisture, oil and foreign matter must be periodically drained from the air receiver tanks, subsequent treatment results in air meeting the quality of breathing air.
2.24 Answer the following statements concerning the Core Flood Tanks TRUE or FALSE.
- 1. Isolation valves CFV-5 and 6 receive an open signal (0,5) following ES actuation even though they are required to be open with their breakers in the " Locked Reset" position.
- 2. When the breakers for CFV-5 and 6 are in the " Locked (0.5)
Reset" position, they lose position indication in i the control room.
1 i 3. During plant operation, the CF tank levels nay be (0.5) i increased by adding from the makeup and purification l (MUP) system and decreased by draining to the Auxiliary Building Sump.
4 During plant operation, high CF Tank pressure may be (0,5) relieved by venting to the Reactor Building.
j i
END OF CATEGORY 2 2.8 i
CATEGORY 3. INSTRUMENTS AND CONTR01.S (25.0 Points) 3.1 Actuation of RB isolation results in the closing of most (1.0)
RB fluid penetrations. However, some valves are opened instead. Select the group below which includes all systems and supplies in which valves are opened when only ES Channel B actuates:
- c. NSCCC
- d. RB spray and sodium hydroxide supply 3.2 Which of the following statements about power supplies (1,0) i to the ICS is correct?
- a. The 118 V field power supply will automatically transfer to VBOP-2 if it is available.
- b. Power to the +24 V DC supply will be switched from VBDP-2 to VBOP-4 by an automatic bus transfer device if VBOP-2 is lost.
- c. If the "S-1" -24 V DC power supply in the ICS cabinets fail's the yellow " loss of ICS power" lanp on the redundant instrument panel will be extinguished.
- d. The 118 Y field power supply automatic transfer between vital buses may be immediate or delayed, depending on which bus is available.
3.3 Which of the following Atmospheric Monitoring System channel (1.0) groups consist of iodine and gas measuring channels preceded by a particulate filter?
- a. RM-Al and A2.
- b. RM-A3, A4, A7 and A8.
- c. RM-A5 and A6.
- d. RM-All, A12 and A13.
3.1
3.4 Which of the following statements correctly describes control (1.0) of the Low Pressure Injection System?
- a. If RCS pressure is raised from 250 to 350 psig the white LPI bypass / reset permit lights will illuminate.
- b. If RCS pressure is raised from 400 to 550 psig, and then is reduced to 450 psig LPI will initiate.
- c. If, with LPI bypassed, RCS pressure is lowered from 550 psig to 450 psig, then increased to 550 psig, resetting all LPI bypass / reset switches will not result in LPI initiation,
- d. If RCS pressure is lowered from 950 to 450 psig the white LPI bypass permit light lights and the blue LPI bistable tripped lights will illuminate.
3.5 Which of the following statements about ESAS control and (1.0) testing circuits is true?
- a. Testing of selected equipment groups can be done using manual switches and the manual actuation pushbutton, but return of tested equipment to,its normal lineup requires use of the manual reset switch.
- b. The RB Cooling and Isolation System has six group test switches, one for each group of equipment to be tested.
- c. Valves are assigned to two or more test groups so that all may be full-stroke tested without affecting plant operations or damaging equipment.
d) By proper use of the auto test select switch and the auto test switches, equipment in HPI test groups 2 and 3 can be tested simultaneously.
i f
3.2
~ _.
(1,0) 3.6 Which of the following statements about maintenance and testing of the Reactor Protective System is true?
- a. Placing a channel into " Channel Bypass" changes
.! RPS trip logic to a "one out of three" mode.
- b. During functional testing of a channel in " Channel Bypass"
~
a bright subsystem trip lag indicates that the channel trip relay has tripped.
- c. Maintenance of a single module in an RPS channel can be performed without keeping the entire channel in
" Channel Bypass" or " Trip" mode, but it must pass through one of these modes during module removal.
- d. Placing more than one channel into " Shutdown Bypass" will trip all CRD breakers.
3.7 Which of the following statements about Electrical System (1.0) protective relays is correct?
l a. The speed of an overcurrent relay trip is independent of the overcurrent magnitude,
- b. Lock-out relays prevent closing of a circuit breaker t by mechanically blocking the closure mechanism.
- c. A differential relay scheme balances current flows in each phase of a multi-phase line,
- d. Because no current normally flows through a differential relay coil, this protection is less sensitive than overcurrent relay protection.
3.8 Which of the following statements about the Source Range (1.0)
- NI System is true?
- a. High voltage supply is tugned off when one Intermediate Range Channel reaches 10~ amps,
- b. Detectors detect only neutrons, not gamas.
- c. Contains a rod withdrawal inhibit which is bypassed if any two power range channels indicate >101 FP,
- d. A SUR signal from the Source Range goes to a bistable to halt rod withdrawal when the SUR exceeds 2 DPM. This is reset at 1 DPM.
i 3.3
T 3.9 Which of the following pairs of conditions is needed to put (1.0) the RCS pilot-actuated relief valve in the " Overpressure Protection (550 psig)" position?
- a. RCV-10 in manual and " Low-Range" key switch in "0N".
- b. LPI 500 psig trip bistables reset and " Low-Range" key switch in "0N".
- c. LPI 500 psig trip bistables reset and RCV-10 in "AUT0".
< d. RCV-10 in "AUT0" and " Low-Range" key switch in "0N".
1 3.10 Which of the following statements is true regarding Control (1.0)
Room indication of Emergency Diesel Generator status?
olo$5 / a. A yellow light is lit when there is 1508 air being
! supplied to the air start control valve.
- b. A yellow light is lit when the generator output l breaker is closed.
- c. A white light is lit when the generator output I breaker is closed.
I
- d. A yellow light is lit when the generator is at rated voltage and frequency.
} 3.11 Selecting another OTSG startup range level transmitter for (1.0) j both OTSGs at the same time can potentially cause which
- one of the following results?
i
- a. Lock in both high level limits.
]
- b. Cause the Feedwater startup control valve to fail shut.
- c. Cause Emergency Feedwater to actuate.
- d. Cause the low-load block valve to open.
I
! I i
3.4 ,
3.12 Which one of the following statements is true regarding the (1.0)
RCP electrical distribution?
- a. Upon Icsing 6900 VAC power, the RCP supply breakers trip within 6 cycles.
- b. Upon losing 6900 VAC power, the RCP supply breakers fail as is.
- c. Upon regaining 6900 VAC power, any RCP supply breaker that tripped on undervoltage will automatically reclose.
3.13 Which of the following statements concerning the Main Steam (1.0)
Rupture Matrix system is true?
- a. It contains three actuation channels, "A", "B" and "AB".
- b. Each channel is automatically bypassed when MS pressure i
<725 psig.
- c. Each channel contains four pressure switches, two located on an A and two located on a B main steam line.
- d. Each channel automatically resets when MS pressure increases >600 psig.
NOTE: The following questions ask you to select the incorrect, or negative, response from among a list.
3.14 Which of the following statements about the Diamond Panel (1.0) is INCORRECT?
- a. The LATCH switch allows CRD motors to be driven past i
the In Limit.
I
- b. The MOTOR FAULT lamp identifies programmer operation l without a command, or out motion during an in command.
- c. The CLAMP / CLAMP RELEASE switch cross connects the Auxiliary power supply and the DC Hold bus.
- d. The FAULT RESET switch closes CROM trip breakers when groups are at the In Limit.
3.5 ,
3.15 Which of .the following statements about temperature measurement (1.0) is NOT true?
- a. If the sensing wire of an RTO breaks the instrument will read offscale high.
- b. If a thermocouple wire breaks the instrument will read off scale low,
- c. RTDs respond faster to temperature changes than thermocouples,
- d. The temperature range which can be measured by an RTD is smaller than that for a thermocouple.
3.16 Which of the following conditions will PREVENT starting (1.0) of an RC pump?
- a. NSCCC flow equals 240 gpm.
^
. b. Oil lift pressure equals 220 psig.
- c. Seal injection flow equals 4 gpm.
- d. Reactor power equals 25% FP.
3.17 Whi af the following is NOT a condition required to (1.0) al i vo "A" main feedwater block valve to open?
- a. Reactor reset !
4
- b. Turbine reset
- c. A and B main feedwater pumps operating
- d. Feedwater cross-over valve (FWV-28) closed 1 3.18 Which of the following will NOT cause the ICS to control (1.0) '
) the unit in " TRACK"?
- a. Reactor Trip i
- b. Turbine not in "ICS/AUT0"
- c. SG/Rx master control station in " MANUAL"
, d. Either feedwater loop contr'ol station in " MANUAL" 3.6
, - -~- -
-- , m-- , ,
4 3.19 Which of the following statcmrits about the design of the (1.0)
Emergency Diesel Generator ?nat ing control is NOT true?
- a. During' block loading sr.quincing output voltage dips by 25%
without tripping safeguards control center starters,
- b. EDG. start on loss of 4160 V ES bus voltage is accompanied by tripping of the battery' charger feeder breakers, with reclosure about 5 seconds later. ,
- c. Feeder breakers from a 4160 V ES bus to the HPI pumps are not tripped upon loss of bus voltage. However, until the generator comes up to speed it is not connected to the bus.
- d. The RB spray pumps are not connected to the 4160 V.
ES bus until about 15 seconds after block loading ~begins.
1 3.20 Which of the following statements about RB~ Purge control (1.0) is NOT true?
- a. Both purge supply fans must be operating to permit start of the exhaust fans.
- b. Exhaust duct temperature greater than 135'F will s shut down the exhaust fans.
- c. Dampers 0 - 93 and 94 automatically adjust to maintain vent flow rate about 50,000 CFM when purge valves are open and fans are operating,
- d. Purge valves are automatically closed by a HIGH radiation alarm, but supply and exhaust fans continue operating.
'- 3.21 Which of the following statements about the Power Range (1.0) t NI System is NOT true?
- a. Contains three bistables which input to the RPS.
- b. Each channel contains a gain adjustment for use -
in calibration.
- c. Provides the ICS power level signal as the output of the highest of the four channels.
- d. Detectors detect both neutrons and gammas.
~ .
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3.7
l 3.22 If the Reactor Control Station is in " Auto" and neutron (1.0) power level is >60% FP, the reactor will automatically run back to 60% neutron power and a rod out-inhibit l exist if any of the following conditions EXCEPT one '
exists. Select that one.?
- a. Lose safety rod out-limit.
I
- b. One or more rods are >9 inches out of alignment with the group average position for Group 5.
- c. Insufficient or excessive overlap exists between Group 5 and 6, or 6 and 7.
- d. One or more rods are >9 inches out of alignment with the group average position for Groups 6, 7, and 8, and an in-limit is actuated.
- 3.23 Which of the following indication would NOT be expected, (1.0)
! and might indicate an instrument failure?
i
- a. The CR0 " Travel" lamp does not indicate when' Group 8 rods are in motion.
- b. Group 7 out-motion is prevented past 91.4%. .
j d. During a transfer of a group from DC hold to t
Auxiliary, when you select "SE0-0R," the
" SEQ-0R" lamp is on and the " Seq" lamp goes off.
1 1
3.8 1
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3.24 Identify by letter from the list below the detector type l
used in the Reactor Diagnostic System (RDS) for the individual monitoring subsystens listed below. NOTE: some subsystens l may use more than one type of detector.
l Detector Types A. Displacement probes l B. Excore nuclear C. Incore nuclear D. Acoustic Noise E. Accelerometers Monitoring Subsystem
- 1. Loose parts (0.5)
- 2. External structural vibration (0.5)
- 3. Reactor coolant pump (0.5) 4 Reactor internals vibration (0.5) l l
l l
END OF CATEGORY 3 l
3.9 l
CATEGORY 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL (25.0 Points) 4.1 If an overcooling transient has reduced Tc to 495'F and (1.0)
RC pressure to 1450 psig, which of the following immediate or remedial actions should be taken according to the appropriate EP.
- a. Reduce RCS pressure to 1000 psig.
- b. Initiate a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> hold at that condition.
- c. Reduce feedwater flow until Teincreases above 500*F.
- d. Stabilize T eat 100*F subcooled and initiate engineering evaluation 4.2 According to OP-210, REACTOR STARTUP, how often should (1.G)
RCS boron concentration be sampled during deboration?
- a. Every hour.
- b. Every 30 minutes.
- c. Every 20 ppm B.
- d. Every 50 ppm B.
4.3 According to OP-204, POWER OPERATION, if, during a power (1.0) decrease, either OTSG goes on low level control, which of the following is CORRECT 7?
- a. The delta T, load ratio control setting in AUTO should be vesIfted to be at O.
t
, b. The Diamond or Reactor Demand stations shnuld NOT be placed in HAND since FW cannot accept T,yg control.
- c. The Steam Generator / Reactor master should be kept
- in AUTO to prevent overriding the OTSG low level control signal.
- d. The delta Te load ratio control station should be placed in HAND.
f 4.1
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g --
,-y y --m 5
4.4 Reactor Coolant Pumps have been lost because of a loss (1.0) of offsite power. Plant control is being maintained in accordance with the NATURAL CIRCULATION procedure, AP-530. According to this procedure, which one of the following statements accurately describes a correct course of action for maintaining 0TSG level?
- b. If PZR level is less than 50 inches, DO NOT exceed an OTSG level of 50%.
- d. If subcooling margin exceeds 100*F, OTSG levels may be decreased below 50%.
4.5 Which one of the following conditions is a procedural (1.0) requirement for manually tripping the reactor?
- a. Emergency Feedwater actuates.
- b. Subcooling margin drops below 50*F during power operation,
- c. Shutdown margin is determined to be less than 1.0% Delta K/K.
- d. Feedwater flow is lost.
4.6 Thermoluminescent dosimeters should be rezeroed prior to (1,0) reaching:
- a. 50% of full scale
- b. 75% of full scale
- c. 90% of full scale
- d. 100% of full scale 4.2
L 4.7 The background reading on a frisker used for whole body (1.0) '
l frisking should be no more than:
- a. 10 cpm
- b. 50 cpm j c. 100 cpm
- d. 300 cpm 4.8 According to the OSIM, during an einergency, when is it (1,0) permissible to use the PORV to prevent a high pressure j trip?
- a. When the block valve is operable, and a " dedicated I operator" is used. !
- b. When the cause of the highpressure has been determined to be an under cooling event. I
- c. When subcooling margin is > 60*F. l l d. It is never permissible to use the PORY for this pu rpose.
4.9 List all of the Inunediate Actions for a REACTOR PROTECTION (2.0)
SY5TEMACTUATION,(AP-580). (Do NOT include Remedial Actions.)
4.10 The first Immediate Action of EP-290, INADEQUATE CORE COOLING (1.0) is to ensure full HPI flow. List the Remedial Actions associated with this step.
4.11 List all of the Immediate Actions for an ENGINEERED (3.0) l TATEtiTTANbs SYSTEM ACTUATION, (AP-380). (Do NOT include Remedial Actions.)
l l 4.12 List all of your Inunediate and Remedial Actions for EP-140, (2.0) l Il M ilCf REACTIVfTY CONTROL.
l l \
l l
l 4.3
. - - , .-,- -.- ~., .- -
.-~.------eat. ,, -- . - - - - - ~ - - - - , , .
NOTE: The following questions ask you to select the incorrect, or negative, response from among a list.
4.13 When responding to an OTSG leak, which of the following (1.0) is NOT a method recorwended by EP-390, OTSG TUBE RUPTURE, :
for determining the affected OTSG during cooldown and !
depressurization?
- a. Observe MS radiation monitors,
- b. Notify chem / rad to survey main steam lines.
- c. Compare OTSG levels and feed flows.
- d. Notify chem / rad to sample the-0TSGs.
4.14 According to OP-502, CONTROL R00 DRIVE SYSTEM, which (1.0) of the following statements about a jammed CRDM is 3
NOT true?
- a. Stator cooling water must be flowing and the j stator thermocouple temperature must be continuously monitored during attempts to free the mechanism.
- b. Jamming usually occurs due to metal chips lodging between the motor tube and the segment arms.
, c. Attempts to free the rod may be made using l
controls in the Control Rod Drive Room.
, d. If the CROM will not move at run speed it
! is not permissible to attempt moving it at l
jog speed in either direction.
i 4.4
4.15 Which of the following limits and precautions (1.0) about PLANT C00LDOWN (OP-209) is INCORRECT?
NOTE: Do not judge the accuracy of the following statements (in this question only) based on information in parentheses,
- a. The pressurizer spray shall not be used if the delta T between the RCS and the pressurizer
, is >(410'F).
- b. When RCS pressure is <(150 psig) lock out and red tag the breakers for all makeup pumps and all HPI valves MUV-(23,24,25,26).
- c. The average shell temperature of the OTSGs shall not exceed (60*F) above or below RC temperature.
- d. Continuous blowdown of the secondary side of the steam generators via sample line is required whenever reactor power is <(157 FP) and RC temperature is >(250*F).
4.16 Which of the following statements about producing (1.0) a steam bubble in the pressurizer does NOT agree with 0P-202, PLANT HEATUP?
- a. Pressurizer temperature and RC wide range pressure must be placed on trend recorders. '
- b. Nitrogen heaters must be deenergized and breakers j #19 and #16 opened.
- c. When heating the pressurizer to saturation, maintain pressure between 50 and 150 psig.
- d. If chemical analysis indicates incomplete venting after i saturation is reached, reopen RCV-6 and RCV-5 to the
] waste gas system.
4.17 Which of the following is NOT an immediate action for (1.0)
AP-542, ASYMMETRIC R00 RU M 7
- a. Ensure NI power decreasing,
- b. Ensure Control Rods inserting.
I c. Ensure Feedwater flows decreasing,
- d. Ensure RC pressure stable.
4.5
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t 4.18 Which of the following issuediate actions related to (1.0) :'
an RMA-5 "HIGH" alarm is INCORRECT?
- a. Ensure dampers D-1 and D-2 are closed and 0-3 is open. l
- b. Ensure stopped or in slow speed: AHF-20A, AHF-208. j
- c. Ensure stopped: AHF-17A and B, and AHF-19A and B.
- d. Ensure stopped: AHF-30, and AHF-34A and 8. !
4.19 Which of the following are NOT required to determine (1.0) ,
cooldown rate after a trip? l
- a. Decimal equivalent of clock time since trip. !
- b. Pretrip average of 5 highest incore thermocouples.
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! c. Present Tc . ;
- d. Pretrip T g. ,
4.20 According to OP-209, PLANT C00LOOWN, Safety Rod (1.0)
Groups 1 through 4 should be at their fully withdrawn I position whenever reactivity is being changed by all !
EXCEPT one of the following. Select the INCORRECT statement. .
- a. Reduction in boron.
- b. Decrease in RC temperature.
- c. Motion of APSRs. r
- d. Motion of full-length control rod banks. l 4.21 According to Crystal River Operating Procedures, which (1.0) one of the following actions does NOT have to be taken '
prior to tripping an RCP at power? ,
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- a. Bypass the individual RCP power monitor. ;
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- b. Reduce power to below 60%.
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- c. Run AC oil lift pump for at least two minutes.
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- d. Ensure cold leg temperature selected for delta l Te control is for loop other than for the pump i tnat is to be secured. i END OF CATEGORY 4 l
END OF EXAM
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I 4.6 i
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.= .-
EQUATION SHEET CR g Q = mah M= y o
Q = UAAT x2(ft) = xw A* = 10~8 see xf=xw A = AN Q = mcpai in2 O
c A = g1 /2 DNGR =x7
? = Pa1050R(t) N = No e(*At)
P = Po et /T g /2
- 26.06 60En SUR = 7 R/h r *d T !
T=8~D ,
A = 0.1 sec-I 1*+8-p 9 1*2
- D2 - hg T=P AP q=haa: !
p = K'f f ~ l g.u Kerr 3 t
p= K2gg * *1 XEg+hg + 4 2
- KE2 CR1 1 - Keff 2 +h2 * *12 1 - Kerr 1 where:
RR = IfSth 1) KE is Xinetic energy ;
- 2) w is work done t
.c* 3) q is the heat transferred ;
SCR = 3 , Kerr .
- 4) h is the enthalpy t 1
m__ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _
_ . _ _ -~'
/ *
- ENCLOSURE 3 s' U. S. NUCLEAR REGULATORY COPNISSION SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION Facility: Crystal River - 3 Reactor Type: Babcock & Wilcox Date Administered: March 5, 1985 Examiner: J.C. HUENEFELD Candidate:
INSTRUCTIONS TO CANDIDATE:
Use separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheet. Points for each question are indicated in parenthesis af ter the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up four (4) hours af ter the examination starts.
, Category % of Candidate's % of Value Total Score Cat. Value Category S 3 45* 25 5. Theory of Nuclear Power Plant Operation, Fluids and Thermodynamics g3 E 25 6. Plant System Design, Control i and Instrurientation
, M# 25 7. Procedures - Normal, Abnormal, Emergency, and Radiological l Control MI 25 8. Administrative Procedures, Conditions, and Limitations
% -4etr TOTALS Final Grade %
All work done on this examination is my own; I have neither given nor receiveo ,
t aid.
! Candidate's Signature l
%g
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l 5.0 Theory of Nuclear Power Plant Operation, Fluids, and Therinodynamics (25 Potnts) 5.1 Voiding has occurred in the RCS, in the vicinity of the reactor vessel l during a natural circulation cooldown. Which of the following CORRECTLY I
characterizes the process of collapsing the void? (1.0) a) The void will collapse immediately upon increasing the pressure above the local saturation pressure; the main concern is water hammer. ,
b) The void will collapse at a rate equivalent to the rate of HPI flow; therefore, full HP1 should be run until the void is fully collapsed.
c) The void will be composed largely of hydrogen gas, and will therefore require degasifying of the RCS in order to begin collapsing it.
d) The void will superheat if an attempt is made to collapse it too fast. The rate of collapse will be governed largely by ambient heat loss from the void.
l 5.2 khile operating at 70s power with three reactor coolant pumps running, which of the following statements is CORRECT about the relationship between loop outlet temperatures? (1.0) co e lu$bkki bh or b) T n the loop with lower flow will be hi her because of the fE1: owing energy balance: Q=mcp (Th - c) c) The loop with lower flow is receiving less pump heat than is the loop with higher flow, and is therefore at a lower temperature.
d) The loop with lower flow is at a lower temperature because the neutron power is suppressed in the regions of lower flow.
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(Section 5.0 Continued on Next Page) l l
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3 5.3 Which of the following is TRUE regarding OTSG outlet pressure?
(i.e., the actual pressure of the steam just as it leaves the OTSG)
(1.0) a) OTSG outlet pressure increases with increasing power to overcome head loss in the main steam piping.
b) 0TSG outlet pressure decreases with increasing power because of the increasing effect of aspirating steam.
c) OTSG outlet pressure decreases with increasing power because of the decrease in length of the superheat region.
d) OT5G outlet pressure does not vary with power because turbine header pressure is maintained constant.
5.4 When the main generator is synchronized with the grid, excitation current is proportional to: (1.0) al reactive load (MVAA) b) real load (MW) c) output voltage (KV) '
d) generator speed (RPM) 5.5 The amount of heat being added by the reactor coolant pumps: (1.0) a) is less than the RCS heat loss to ambient at operating temperature. .
b) is less than the amount of heat being lost to letdown at operating temperature.
c) causes total OTSG themal output to be greater than the themal output of the core itself.
d) is insignificant at normal operating temperature.
(Section 5.0 Continued on Next Page)
.m
l* 4 I 5.6 During a LOCA with a resultant loss of subcooling margin, why are the l Reactor Coolant Pumps (RCPs) secured? (1.0) l i
a) To prevent pump damage resulting from operation under two phase conditions.
l b) To prevent core damage resulting from phase separation upon subsequent loss of RCS flow.
c) To reouce RCS pressure by removing the pressure head developed by the RCPs.
d) To remove the heat being added to the RCS by the operating RCPs.
5.7 Which of the following is the LARGEST (in standard cubic feet) potential source of RCS noe-condensable gas? (1.0) a) Zirc water reaction .
b) Core Flood tanks c) Pressurizer steam space d) 100% failed fuel.
5.8 The RCS is in Hot Standby with no Reactor Coolant / umps running. If OTSG pressure is decreased, accoroinq to the Plant Verification ;
procedure (VP-580), which of the fol' owing temperature responses best '
indicater the presence of natural circulation? (1.0)
! a) T h increases, Tc remains the same b) T h increases Tc decreases c) Th decreases, Tc decreases a) T h remains the same, Tc decreases.
i 5.9 According to the Bases for the Limiting Safety System settings, which l of the RPS trips will be reached first during a slow reactivity I
insertion accident from low or high power? (1.0) a) Nuclear overpower trip setpoint
! b) RCS pressure high setpoint c) ACS outlet temperature high d) RCS variable pressure temperature low l (Section 5.0 Continued on Next Page) l
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . ____.___2_ _ _ _ _ _ _ _ .___ _ _ _ _____ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _
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- 5 5.10 The reactor is brought to 10-8 amps 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> af ter a trip from 100% FP at equilibrium menon conditions. In order to maintain power level at 10-8 amps for the next hour, what will have to be done with control rods?. (1.0) a) They will have to be withdrawn.
b) They will have to be inserted.
c) They will need to be withdrawn initially, then inserted to compensate for xenon burnout.
d) The rods will have to remain at a constant position because the :
rate of xenon burn out is almost exactly matched with the post shutdown indirect xenon production rate.
i 5.11 Which of the following best describes the behavior of equilibrium menon reactivity over core life? (1.0)
- 4) It decreases, because of the increased fuel burn-up.
I b) It decreases, because of the decrease in plutonium-xenon yield.
c) It increases, because of the increase in thennal flux.
d) It increases, because of the decrease in boron concentration.
5.12 The reactor is critical at 10-8 amps. Which of the following best o describes the behavior of neutron power following a prompt insertion of negative reactivity? (1.0) a) Neutron power drops issnediately to " Beta" (delayed neutron '
fraction) times the neutron power prior to t'io prompt insertion of negative reactivity.
b) Neutron power decreases 11ne4rly with time af ter the initial prompt drop.
c) After the initial prompt drop, neutron power decreases on a constant negative period; the magnitude of the period determined by the amount of negative reactivity inserted.
! d) Because only delayed neutrons are lef t insnediately after a negative reactivity insertion, neutron power decreases on an 80 second period regardless of the size of the negative reactivity insertion.
(Section 5.0 Continued on Next Page) 1
6 5.13 Which of the following statements best describes the consequences of an RCS boration that reduces power without Control Rod motion? (1.0) a) Imbalance becomes less negative and Power Range N1 calibration becomes less conservative.
b) Imbalance becomes more negative and Power Range NI calibration bec wes less conservative.
c) Imba ance becomes less negative and Power Range NI calibration becor.es more conservative.
d) ImbJ1ance becomes more negative and Power Range N! calibration be;omes more conservative.
5.14 Whica of the following statements about Sm-149 is TRUE7 (1.0) '
a) It is removed from an operating reactor by burnout and radioactive decay, b) When a reactor is restarted af ter a temporary shutdown Sm-149 [
l concentration increases for several days.
l c) It has less effect on reactor operation than Xe-135 due to its l Smaller fission yield and smaller microscopic neutron cross l section.
l l
d) The equilibrium concentration of Sm-149 at 50% FP is about two thirds of the equilibrium concentration at 100% FP.
l l 5.15 Which of the following statements best describes parameter changes in l the pressurizer following a rapid load reduction of 15% FP7 (1.0) r a) Pressurizer pressure and level will increase, with the pressure increase stopping before the level increase stops, b) Pressurizer pressure and level will increase, with the level increase stopping before the pressure increase stops.
l c) Pressurizer level and pressure will decrease, with the pressure decrease stopping before the level decrease stops.
d) Pressurizer level and pressure will decrease, with the level decrease stopping before the pressure decrease stops.
l (Section 5.0 Continued on Next Page)
O 7
5.16 Why should an RCS cooldown on natural circulation NOT exceed 10*F/hr for Tc > 280*F7 (1,0) a) to prevent exceeding brittle fracture limits of the reactor vessel.
b) to ensure adequate mixing of HPI injection water with RC flow into the uowncomer.
c) to ensure that adequate heat removal through the OTSGs is possible without having to increase level above 50%.
d) to prevent rapid and erratic changes in pressurizer level from occurring due to void formation in the vessel head.
5.17 Which one of the following statements is TRUE concerning the change in differential boron worth (% delta k/k)W RCS boron concentration (range of 0 to 1800 ppm) and Tave (range of 532*F to 579'F)? (1.0) a) It cecreases as T aye and RCS boron concentration increase.
b) It decreases as RCS boron concentration increases but is constant 85 Iave increases.
c) It increases as Tave and RCS boron concentration increase,
' #U"I " "I U" con ntrat n nMsW'"""
5.18 On a reactor trip from 100% power and equilibrium menon conditions, peak xenon will be reached in approximately _ hours. (1.0) a) 4 to 6 b) 8 to 10 c) 12 to 14 d) 72 s.19 What is the effect of starting a large induction motor oi a bus while being supplied by a DG7 (1.0) a) It inCrdases the real load, but reactive load remains unchanged.
b) It decreases the power factor.
c) It increases the real load, but decreases the reactive load.
d) It increases both the real and the reactive load.
(Section 5.0 Continued on Next Page)
8 5.20 At normal operating temperature, a leek from the PZR water space to the containment atmosphere would consist of: (1.0) l
- 4) Super-heated steam, b) 55turatedsteam. l c) Low quality steam, d) Saturated water. l 5.21 hhtch of the fo11owf ng woJ1d tend to place the ICS closer to a BTU 11stt? (1.0) ,
a) A decreate in feedwater flow.
b) An increase in feedwater topperature.
c) An increase in OTSG pressure.
d) An increase in T ay,,
5.22 The ratto of Pu 239 ana Pu 240 atoms to U 235 atoms changes over core life. Which one of the pairs of parameters belcw are most affected by this change? (1.0) ,
al moderator temperature coefficient and doppler coefficient b) doppler coefficient And beta c) beta and moderator temperature coefficient d) moderator temperature coefficient and neutron generation time.
1 (Section 5.0 Continued on Next Page)
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5.23 A moeerator is necessary to slow neutrons dean to thermal energies. t mich of the following is the C0ARECT reason for operation with thermal instead of fast neutrons 7 (1.0) l
- 4) Increased neutron efficiency since thersal neutrons are less !
likely to leak out of the core than fast neutrons.
b) Reactors operating primarily on fast neutrons are inherently unstatie and have a t.igher risk of going prompt critical. ;
c) The fission cross section of the fuel is much higher for therwal I energy neutrons taan fast neutrons.
- 4) Doppler and moderator temperature coefficients become positive as neutron energy thcreases.
5.24 The reactor core Safety Limit Curve (Figure 2.11 in the Technical Specifications) is prevented from being onceeded by a comeination of ,
four APS trips. LIST the comeination of RPS trips tnat define the i reactor trip envelope. (g g incluee setpoints) (2.0) r t
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6.0 Plant Systems Design. Control. and Instrumentation (25 Points) j 6.1 The Makeup valve. MUV 31, may be bypassed manually. The bypass is normally throttled to maintain: (1.0) a) 8 gallons per minute Makeup flow to ensure that Makeup never falls below that required for seal return flow.
b) greater than 15 gallons per minute to ensure that the running Makeup pump does not over-heat.
c) greater than 15 gallons per minute at all times when the RCS is above a minimum temperature specified by procedure, d) equal to 1 gallon per minute at all times to minimize thermal shock to the Makeup nozzle.
6.2 Selecting another OTSG startup range level transmitter in both OTSGs at the same time can potentially: (1.0) a) lock in both high level 11.mits.
b) cause the Feedwater startup control valve to fall shut.
c) cause Emergency Fecdwater to actuate.
d) cause the low-load block valve to open.
t.3 I'r;; Of th; ;;;:iti;;; Mbefew-ehowM-elways-eewee :: ;;,T tM; r; n t::t. ':'i;t ;;; w+H-40T- necessarily- cowee-evt ..,..- . .. ,, . .-.
. . . . . . . , , M!.0) a) One or more Group 7 rods are > 9 in. out of alignment with the g/)/
J group average and an in-limit is actuated.
b) One or more AP$Rs are > 9 in. out of alignment with the group average and an in-limit is actuated, c) One safety rod is dropped.
d) One or more Group 5 rods are > 9 in. out of alignment with the group average position.
(Section 6.0 Continued on Next Page)
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6.4 In the mein turhine steam enters the steam chest by passing through I the throttle valves and exits the steam chest by passing through the governer valves. Which of the following statements is consistent i with the steam flow path: (1.0) +
r a) If both throttle valves in one steam chest are open, then the i 0T M's steam headers are cross connected. [
b) If any tue governor valves are shut then the OT5G steam headers l are no longer cross connectee. i c) If any two throttle valves are shut, then the OT5G steam headers are no longer cross connected.
If only one throttle valve is open in each steam chest, then the I d)
OT5G steam headers are cross connected, l
i i 6.h Which of the following is an IDWA0ptR action to perfons when taking manual control of an atmospherTE~TETs dump valve? (1.0) ,
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a) Secure instrument air to the position controller. !
t b) Place the position controller to the "8V-PA55* position.
c) Open the vent valve up stream of the position contro11er pressure l regulator. j d) Position the atmospheric dump valve by using the actuator I handsheel. [
6.o which one of the following load limiting conditions and corresponding j load listt is COAntCT7 (1.0) ;
e) Loss of 1 RC pump with 4 running - MS/ min to maximum Ilmit of 751 b) Loss of 2 AC pumps with 4 running - 301/ min to maximum Ilmit of 50%
c) Loss of a Feedwater looster pump - %l/ min to maximum Ilmtt of .
555 r d) Asymmetric Ro4 304/ min to monimum limit of 60%
i (Section 6.0 Continued on Nent Page) l t
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6.7 Which component is loaded on a safeguards bus during block one loading? (1.0) a) The motor driven EFW pump b) A battery charger c) A'Recctor Building fan d) A Reactor Building Spray pump 6.8 Which of the following is TRUE regarding the RCP electrical distribution? (1.0) a) Upon losing 6900 VAC power, the RCP supply breakers trip within 6 cycles.
b) Upon losing 6900 VAC power, the RCP supply breakers fail as is.
l c) Upon regtining 6900 VAC power, any RCP that tripped on undervoltage will automatically restart.
J) RCP supply breakers will trip open on UV af ter an 8 second time delay.
6.9 Which on the following actions will cause the "A" CROM breaker to open? (1.0)
- a) Placing more than one channel "A" module test switch in " test".
b) Racking out both the channel "A" low pressure bistable and the channel "A" high flux bistable, c) Placing the "A" channel output trip test switch on the "A" reactor trip module in " test",
d) Placing the "A" channel in " channel bypass" and then deenergizing the "A" RPS cabinet.
6.10 Which of the following RPS trips is NOT bypassed when the RPS is in
" Shutdown Bypass"? (1,0)
- 4) Low Pressure (1800 psig) b) High Flux c) Flux / Delta Flux / Flow d) Variable Pressure Temperature
($ection 6.0 Continued on Next Page)
13 6.11 All R'S equipment (sensors, recorders, modules, etc.) is powered from: (1.0) a) 115 VOC or 120 VAC from its associated channel.
b) i 15 VOC or 120 VAC from the other three channels.
c) 116 VDC of 120 VAC from a separate vital bus.
d) the DC distribution system. -
6.12 Which of the following is a direct result of a loss of NN! "X": (1.0) a) Teve fails to 570* F, control rods withdraw.
b) Diamond panel transfers to manual.
c) RC total flow fails to 751, and the ULD runs back to 751. ;
d) Turbine bypass valves fail to mid position.
6.13 fte motor driven EFW pump (EFP 3A) is cooled 'by: (1.0) a) N5CCC r
b) 55CCC c) DHCCC d) its own discharge. .
t 6.14 The concennate water storage tank with minimum water volume is suf ficient to maintain the plant in HOT STAND 8Y for ? hours with .
steam discharge to atmosphere. (1.0) j a) 8 b) 24 l
c) 50 d) 100 E
(Section 6.0 Continued on Next Page)
o 14 6.15 Which of the following statements about the Steam One Rupture Matrix isN01 correct? (1.0) a) The maintenance bypass key in "Maint" position and bypass pushbutton depressed will bypass the rupture matrix.
b) If the rupture matrix actuates when in the test mode, the M51Vs will close.
c) The bypass reset buttons will remove the bypass if depressed only if pressure is > 600 psig.
d) Each 5/G is protected by both matrix "1" and ";".
6.16 Under which condition will the Main Feedwater block valves NOT shut? -
(1.0) l a) Reactor trip b) Feedwater demand reaches 50% decreasing c) Main Feedwater pump trip d) Feedwater pump discharge crosstie is not shut.
6.17 Which of the following statements about RD Purge con',rol is NOT true?
(1.0)
, a) Both purge supply fans must be operating to permit start of the exhaust fans, b) Exhaust duct temperature greater than 135' F will shut down the exhaust fans.
c) Dampers 0 - 93 and 94 automatically adjust to maintain vent flow rate about 50,000 CFM when purge valves are open and fans are operating.
d) Purge valves are automatically closed by a HIGH radiation alarm, but supply and exhaust fans continue operating.
(Section 6.0 Continued on Next Page)
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I 6.18 The primary fire protection system protecting the Emergency Diesel j Gene *ators is at (1.0) t a) Wet pipe sprinkler system !
b) Deluge system i
c) pre-action sprinkler system (Dry pipe) '
d) Open neaale sprinkler system ,
l 6.19 Which of the following statements about Emergency Diesel Generator ;
control and Iceding is hof true? (1.0) L a) During block leading, system voltage may dip by several percent f without tripping safeguards loads. j b) While the 3A th bus is being supp11ed by the EDG, a condition I causing the initiation of EFW will result in a load shed of the l 3A El tus, followed by an auto start of the electric EFW pump. l r
c) 4160 V E5 breaker for a running Hp! pump is not tripped upon loss !
of bus voltage; therefore, it Ismediately automatically re starts l when its ES bus it reenergtaed by the diesel. j d) The R8 spray pumps are not connected to the 4160 V E5 but until !
about 15 seconds af ter block loading begins. , j 6.20 Which of the following statements is TR'JE regarding the Emergency Core Cooling System (ECC5)? (1.0) a) The Hp! pumps may be started at their local 440V breakers located I on ES MCC As on the 11g' elevation. ;
b) The system has been deshned such that during a low pressure ;
situation, as long as y two Hp! pumps are running, pump runout ;
is not an operational ( W ern.
i c) Automatic initiation of hp! by the E5 system starts both -
selected !
Hp! pumps and g pumps necessary for I,pl. j d) The LPI pumps are automatically lined up to supply the section of I the Hp! pumps from tne RB sump when the BWST reaches a specified
, minimum level. i l
l (Section 6.0 Continued on Neat Page) l t
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16 6.21 Why shouldn't hydrazine be added to the ACS during operation Jf the oakeup demiperal12ers? (1.0) a) Secause the hydrazine will be removed by the domineralizers, and therefore wasted. j b) The dominera112er resin does not perform satisfactorily at the low temperatures at which hydrazine is used.
c) Hydrazine chemical reaction with the domineralizer resin could result in release of chlorides.
d) If high 02 levels in the RCS warrant hydrazine addition a potential source may be the domineraltlter and therefore,it should be of f service.
6.22 Select the CORRECT statement concerning the Nuclear Services Booster Pumps and the cRo cooling System. (1.0) !
4)
One A droppump is normally(operated in line flow 4100 gpm) will with the the start other serving idle pump.as backup.
j b) On an E5 signal, the supply and return valves will close and the booster pumps w111 have to be manually secured.
c) Shp 2A is po.ered form ES MCC 3A2 and SkP 28 is powered from E5 MCC 382.
d) Maatsum allowable temperature of the cooling water is 120" F; q there are no limits on etntmum temperature.
6.23 With regard to overspeed protection of the main turtilne, select the one CORACCT statement: (1.0) a) There is a mechanical overspeed trip at 103% and a backup electrical overspeed trip at 111%.
b) With the Overspeed Protection Control (0.P.C.) switch in the
! " Test" position, the electrical overspeed trip is bypassed.
c) At approntmately 1031 shaf t speed only the governor and intercept valves will close, = nile at 1111 speed 411 four sets of valves will close.
d) In the "Overspeed Test" position on the 0.P.C. switch, only the Aeheat and Intercept valves close.
(Section 6.0 Continued on heat Page)
O 17 6.24 When conducting a plant coc1down, several operations are required to prevent inadvertent ES actuation. Which of the following statements is TRUE during a plant cocidown? (1.0) ,
a) When RC pressure is reduced to 1800 psi, the Hp! white bypass ,
permit lights will come on.
b) If'HP! was properly bypassed, the 1500 psi bistable tripped Ityhts will not come on when pressure is reduced below this va ue, i c) When RC pressure reaches 900 psi, the LP! white bypass permit -
Itghts will come on allowing the operator to bypass LP! and RB spray, d) When each channel was bypassed, its respective ember channel bypassed light would have come on, and the green channel function enabled lights and the green bypass / reset l'ghts would have gone out.
6.25 If all three makeup pumps are expected to be running following an E5 actuation and the pumps are powered from the [5 busses as shown ,
below, where should the selector switches be positioned? ; ,
HUP A running, powered from ES Out "A"
- 1) A B selector switch in ,7_. (0.5)
- 2) 6 C selector switch in ,J_. (0.5)
NOTE: For 1) and 2) above, write the apprcpriate letter "A", "B", or i "C" on your answer sheet.
i (End of Section 6.0)
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18 7.0 Procedures . Noma 1, Abnormal. Emergency, and Radiological Control (25 Points) 7.1 Reactor Coolant Pumps have been lost because of a loss of off-stte power. Plant control is being maintained in accordance with the Natural l Ctreulation procedure, AP 530. According to AP-530, which of the following is CORRECT regarding OT W 1evel c"6ntroH (1.0) a) If less than 2 hPI pumps are available, then OTSG 1evels should i
be estabitshed at 50%.
b) If PZR level is less than 50* then 00 40T exceed an OTSG 1evel of
- 50%.
1
l 1 d) If operating range level inalCatton is lost on one OT5G, then i matruin level in the other OT5G at 951 untti RCS flow is re estabit shed.
7.2 Prtor to reaching 200* F RC5 temperature during a neatup, a vacuum is drawn on the OT5Gs. This is done: (1.0) al because there is danger of onygen pitting of the OTSG tubes i
during heatup; therefore all air must be removed from the OT5G.
b) to ensure that all nitrogen is recoved from the OT5G prior to reaching 212* F.
c) to promote early botitr.g of tre OT5G water inventory; therefore, promoting even heating of the OT5G shell.
d) to mtntatze the possibility of water haver.
7.3 Three of the four items below must be ccepleted prior to entering Mode 4 from Mode 5. According to the Heatup procedure (0P 202),
whtCh item need M be Completed prior to entering Mode 47 (1.0)
- 4) Containment integrity must be established.
b) OT5G 1evels must be reduced to 1350 in, c) begastf tcation must be cceplete.
d) The DH system should be shutdown.
(Section 7.0 Continued on Next Page)
19 7.4 Which of the following conditions is a procedural requirement for manually tripping the reactor? (1.0) a) Emergency Feedwater actuates.
b) Subcooling margin drops below 50* F during power operation.
c) Shutdown margin is determined to be less than 1.0% Delta K/K.
d) Feedwater flow is lost.
7.5 If one MSIV drifts shut, according to the Power.0perations procedure
_(0P-204), what is the proper response? (1.0) a) Trip the reactor.
b) Manually reduce power to 60%.
c) Shut one MSIV on the other side to balance OTSG steam loads.
d) Manually open the MSIV.
7.6 Shoula a Xenon oscillation start, according to the Power Operation procedure (0P-204), it is best to: (1.0) a) increase or decrease power by 10% to 15% as soon as the oscillation is diagnosed.
b) immediately determine the oscillation period and position corrections 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the peak 1make rod deviation.
c) determine the oscillation period over a 2 to 3 day duration and then make rod position corrections 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> before the peak i deviation.
d) adjust the axial power shaping rods frequently to maintain imbalance between 0% and -10%.
WM
- eM M
(Section 7.0 Continued on Next Page{ _
~
-n - g - -
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O 20 7.7 According to the RCP Operation procedure (OP-302), which of the following statements is TRUE? (1.0) a) During cooldown, following transfer to the Decay Heat (DH) system, the RCPs should be sequentially stopped, about 5. seconds apart.
b) RCPs may be cperated in an emergency without seal injection flow provided NSCCCW is'in service.
c) If RCP start permissives are bypassed, the maximum allowable reactor power for starting the fourth RCP is 30%.
d) The AC or DC lift oil pumps should be run for at least 2 minutes prior to stopping an RCP.
7.8 During a Steam Generator Tube Rupture (EP-390), which of the following does NOT require the use of the emergency cooldown limits? (1.0) a) The main condenser is not available, b) HPI is required to maintain PZR level.
c) The affected 0TSG cannot be identified.
c) RCPs are not operating..
7.9 Thermoluminescent dosimeters should be re-zerced prior to reaching: (1.0) a) 100% of full scale.
b) 90% of full scale.
c) 75% of full scale.
d) 50% of full scale.
7.10 The background reading on a frisker used for whole body frisking should be no more than: (1.0) a) 10 cpm b) 50 cpm c) 100 cpm d) 300 cpm.
k (Section 7.0 Continued on Next Page)
21 l
7.11 During fuel or core internals handling operations at elevations above the seal plate, you must keep a careful watch over the level in the refueling canal. Upon recognition of an uncontrolled decrease in refueling canal water level, components being handled should be positioned: -H-*)--
a) to maintain at least 5 feet of water. shielding.
I b) to the level of the seal plate. j l
c) in a minimum of 4 feet below the seal plate. !
d) as is, since you are required to immediately evacuate the RB.
7.12 During refueling operations, if problems occur within the Control Rod mast that will not allow the Control Rod to be fully withdrawn, ANSWER the following TRUE or FALSE:
a) There is a special grapple release tool that can be used to disengage the control rod grapple. (0.5) b) A control rod (or burnable poison rod) shall not be released if it is more than the maximum of one foot up from full insertion. (0.5) 7.13 Seating of fuel assemblies has been a generic problem at Crystal River Unit 3. According to the Fuel Handling Equipment Operations procedure (FP-601), ANSWER the following TRUE or FALSE concerning the proper action to take it fuel assembly hang-up is experienced at the spacer grid level?
a) Rotate the fuel assembly and attempt to seat it again. (0.5) b) Reverse the motion of the assembly until the underload or overload is relieved, then shake the cable supporting the fuel assembly. (0.5) 7.14 List all of the Immediate Actions for a Reactor Protection System Actuation,(AP-580). (Do NOT include. Remedial Actions) (3.0) 7.15 The first Immediate Action of EP-290, " Inadequate Core Cooling" is to ensure full HPI flow. List the Remedial Actions associated with this step. (1.0) 7.16 List all of the Immediate Actions for an Engineered Safeguards System actuation, ( AP-380) . (Do NOT include Remedial Actions) (3.0)
(Section 7.0 Continued on Next Page)
22 i
7.17 List all of your Immediate and Remedial Actions for EP- 140,
" Emergency Reactivity Control". (2.0) 7.18 Which of the following is NOT an immediate action for AP-542,
" Asymmetric Rod Runback"? (1.0) a) En'sure NI power decreasing.
b) Clear the asymmetric condition.
c) Ensure turbine runback.
d) Ensure RC Pressure stable.
7.19 ANSWER the following TRUE or FALSE according to OP-412 " Waste Gas Disposal System":
a) An increase on a portable radiation detector (such as an E-102) is used to indicate that all water has been drained from the waste gas surge tank drain pot. (0.5) b) .The " Operator at the Controls" is responsible for verification of the radiation monitor setpoints as they are specified on the '
i release permit. (0.5) c) If meteorological conditions show a delta temperature of zero or positive, you are not allowed to proceed with a gaseous release. (0.5) c d) If the flowrate monitor is inoperable, Technical Specifications allow continuing the gaseous release (assuming the action statement IS[ met). (0.5) l l
l (End of Section 7.0) l l
23 8.0 Administrative Procedures, Conditions, and Limitations (25 Points) 8.1 Which of the following four situations causes entry of a technical specification action statement requiring action to be initiated within one hour? (1.0) ,
a) One air lock door exceeds the technical specification limit for leakage (0.05 La at Pa).
b) The containment average temperature is greater, by just a few degrees, than its technical specification limit (130' F).
c) The letdown isolation valve, MUV-49, is temporarily inoperable.
MUV-49 is one of the containment isolation valves listed in Table 3.6-1 of STS 3.6.3.1.
d) One containment cooling unit is found to be inoperable.
8.2 During power operation, according to the Power Operation procedure (OP-204), the Chem Rad department must be notified whenever power is changeo ? in any ? period: (1.0) a) 15%, I hr b) 50%, 2 hr c) 15%, 2 hr d) 50%, 6 hr 8.3 The operability of the Main Steam Code Safeties is addressed in the technical specifications, STS 3.7.1.1. Which of the following most accurately reflects the requirements for Main Steam Code Safety operability? (1.0) ,
a) All must be operable for continued operation, b) Up to four may be inoperable, provided that the atmospheric dump
! valves and the turbine bypass valves are operable.
c) Up to three may be inoperable, provided RPS setpoints are lowered appropriately.
d) Up to five may be inoperable provided power is reduced by 14% for each inoperable valve.
(Section 8.0 Continued on Next Page)
24 i
8.4 Which of the following is limited to ensure that " Power Peaking" limits are maintained: (1.0) a) Control rod speed b) Axial power imbalance I
c) RCS pressure l d) RCS flow 8 .'5 The fuel pin compression limit: (1.0) a) is more restrictive with forced RCS flow than with natural circulation, b) is more restrictive with no RCPs running than with 2 RCPs running, c) is the same regardless of the number of running RCPs.
d) is only applicable during RCS heatup.
8.6 According to the OSIM, the equipment out of service log is required to be updated and maintained when the plant is in: -(1.0) a) modes 1 and 2 only.
b) modes 1,2,and 3 only.
c) modes 1,2,3 and 4 only.
d) All Modes of Operation.
8.7 All short-term instructions shall automatically expire after ? if not previously cancelled. (1.0) a) one week b) one month c) 90 days d) 6 months.
(Section 8.0 Continued on Next Page) l l
j 25 8.8 According to the OSIM, in the interim between a' trip and the approval for recovery, the Nuclear Shift Supervisor: *
(1.0) a) must ensure that the SCRAM breakers remain open.
b) may authorize the withdrawal of all four Safety Groups provided that a 1% Delta K/K shutdown margin is maintained.
c) may authorize the withdrawal of Safety Group 1 provided that a 1%
Delta K/K shutdown margin is maintained.
. d) may take the reactor critical, holding at 10-8 amps.
8.9 According the OSIM, if two makeup pumps become inoperable: (1.0) a) the time for obtaining operability of one more pump may be extended beycnd the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed by Technical Specifications.
i
{ b) one additional MVP should be made operable within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or a
- plant shutdown should begin.
! c) operation may continue, but only if the "C" MVP is the operable pump, and its cooling water is supplied from both the NSCCCS and the DHCCCS.
d) operation may continue, not to exceed Technical Specification requirements, but consideration will be given to reducing power.
6.10 The Fire Brigade is composed of: (1.0) a) two operations personnel, two maintenance personnel, and one
- security guard.
b) three operations personnel, a team leader, and two maintenance personnel.
c) three operations personnel, including the team leader and two maintenance personnel d) three maintenance personnel, three operations personnel, and two security guards. -
P (Section 8.0 Continued on Next Page)
26 8.11 According to the OSIM, during an emergency, when is it permissible to use the PORV to prevent a high pressure trip? (1.0) a) It is never perinissible to use the PORV for this purpose.
b) When subcooling margin is > 60* F.
c) When the block valve is operable, and a " dedicated operator" is used.
d) When the cause of the high pressure has been determined to be an under cooling event.
8.12 Accoraing to the OSIM, which of the following must he used for proper tracking of "special valve line-ups"? (1.0) a) Placing an entry in the Operator's Log.
b) Listing on the shift relief check list.
c) Placing an entry in the Shift Supervisors Log.
d) Making a temporary procedure change.
l 8.13 Which of the following is NOT listed as an OSIM requirement for removing a decay heat train from service? (1.0) a) Only one decay heat train may be removed from service at any one time, b) Total decay heat load should be less than 25 MW.
c) The requirements of CP-115, In-Plant Clearance and Switching Orders, must be met.
d) The refueling transfer canal is flooded or one OTSG is available for heat removal.
8.14 What is the HIGHEST level of approval necessary prior to performing maintenance on systems that could trip the unit (as specified in the OSIM)? (1.0) a) Nuclear Operator b) Assistant Shift Supervisor c) Nuclear Shift Supervisor d) Operations Superintendent or- person on-call (Section 8.0 Conti~nued on Next Page) t
27 8.15 While conducting a cooldown in Mode 5, if both diesel generators i become inoperable: (1.0) a) no Technical Specification action statement is entered.
b) Technical Specifications require that insnediate actions should be ,
taken to establish Mode 6.
c) Technical Specification requires that positive reactivity changes
. be stopped.
d) Technical Specifications requires that OTSGs must remain operable as a means of heat removal.
8.16 Which of the following Technical Specification action statements require some action to be taken within one hour? (1.0) a) Primary containment internal pressure exceeds the Technical Specification limit.
I b) Two control rods are fully inserted and inoperable, c) One APSR becomes inoperable.
d) One Reactor Coolant Pump becomes inoperable while in Mode 1.
8.17 Which of the following MAY proceed given that a Technical Specification action statement has been entered requiring you to
" suspend all CORE ALTERATIONS"? (1.0) a) Removing a neutron source from the core or positioning the auxiliary neutron detector.
b) Using the bridge in the core is allowed, provided that the low load limit is jumpered out.
c) Control rods and burnable poison rods may be shuffled as long as KEFF < .95.
d) Completion of the movement of a component to a conservative j position.
l (Section 8.0 Continued on Next Page) l I
28 8.18 According to the Defueling and Refueling procedure (FP-203), boration of the RCS is to begin immediately if: (1,0) a) there is a perceptible increase in the audible countrate.
b) countrate on one neutron monitor doubles with no core geometry change.
c) countrate on more than one neutron monitor doubles with no core geometry change.
d) countrate on any two indicators shows a perceptible increase with no geometry change.
8.19 According to the Site Emergency Plan, the NRC must be notified of the declaration of a General Emergency within what time limit? (1.0) a) 15 minutes b) 30 minutes c) I hour d) 90 minutes 8.20 The Crystal River STS has an action statement which states, in part, "whenever the point defined by the combination of Reactor Coolant System flow, Axial Power Imbalance and Thermal Power has exceeded the appropriate safety limit, . Which one of the following CORRECTLY completes this action statement? (1.0) a) be in Hot Standby within one hour.
b) be in Hot Standby within 15 minutes.
c) reduce thermal power within its limit within one hour.
d) reduce the thermal power within its limit within 15 minites.
8.21 Which of the following are NOT considered " CORE ALTERATIONS" by Technical Specifications? (1.0) a) Movement of core internals b) Removal of reactor vessel head c) Exercising the internals vent valves d) Replacement of surveillance capsules.
(Section 8.0 Continued on Next Page)
r a
29 9
8.22 According to Technical Specifications, what is the DIFFERENCE between a high radiation area with levels of 800 mrem /hr, as opposed to one with levels of 1200 mres/hr? - (1.0) a) The 1200 mrem /hr area requires that a staff member entering have an accompanying HP representative, the 800 mrem /hr area does not.
b) The 1200 arem/hr area requires that a staff member entering take an audible dosimeter, the 800 mrem /hr area does not.
c) The 1200 mrem /hr area requires key-locked doors to prevent unauthorized entry, the 800 mrem /hr area does not.
d) The 1200 mres/hr area requires posting as an " Extremely High Radiation Area", the 800 mrem /hr area does not.
8.23 ANSWER the following TRUE or FALSE:
a) The Emergency Coordinator shall not delegate the responsibility for decisions related to, emergency classification, notification, and/or protective action recommendations. (0.5) b) At the time that the EOF is manned and operational, the EOF director will assume responsibility for emergency classification, notifications, and protective action recommendation. (0,5) c) The TSC and OSC must be activated for all emergency action levels except an Unusual Event. (0.5) d) The Site Director, Nuclear Plant Manager, or their designated altenatives are the only individuals who may relieve the Nuclear Shift Supervisor as Emergency Coordinator. (0.5) 8.24 According to the Technical Specifications, while operating in modes 1 through 3, the maximum level allowed in the OTSGs is: (1.0) a) 83% on the Operating Range b) 87% on the Operating Range c) %% on the Operating Range d) 98% on the Operating Range (End of Section 8.0) i I
(END OF EXAMINATION) l l
)
h r
,1 CRYSTAL RIVER-3 -
Answer Key, Section 5.0 5.1 d) Ref: SECY 82-475 Lesson on EP-260. EP-290 Fundamentals of National Circulation 5.2 d) Ref: Curve'1.5B, Curve Book 5.3 a) Ref: Curve 1.6, Curve Book ICS Analog and Digitals 5.4 a) Ref: OP-203 Rev. 43, Page 20 5.5 c) Ref: SP-0312, STM-420 5.6 b) Ref: Lesson No. RQ-84-7E Recognition / Mitigation of Degraded Core 5.7 a) Zr - H 2O 420,000 CFTs 26,000
~
Pzr 140 Fuel 1,130 Ref: Fundamentals of Natural Circulation 5.8 c) Ref: VP 580 Rev. 0-2, Page 19 -
5.9 b) Ref: STS Page B 2-6 5.10 a)- Ref: NUS Module 3 Sect.10.3 5.11 c) Ref: CR Study Guide Page 2 5.12 c) Ref: NUS Module 3 Sect. 5 5.13 c) Ref: NUS Module 2 Sec.16.5 NUS Module 3 Sec. 8.4 5.14 c) Ref: NUS Module 3, Sec.10.5 5.15 a) Ref: STM 419 Page 30 l 5.16 d) Ref: AP-530, Page 4 5.17 a) Ref: Plant Curve Book, Curve 3.2A 5.18 b) Ref: CR Training Ltr TRA 85-0013 5.19 d) Ref: STM-15-1
e-2 5.20 c) Ref: Any Mollier Diagram or Steam Tables 5.21 c) Ref: ICS Analog and Digitals 5.22 b) Ref: CR Training Ltr TRA 85-0013 5.23 c) . Ref: Duke Power Company, FNRE 5.24 Low RCS Pressure l
High RCS Pressure _ .- !
RCS Outlet Temp - High l Variable Pressure - Temperature Trip Ref: TS Figure 2.1-1 1
I l
3 CRYSTAL RIVER-3 Answer Key Section 6.0 l
6.1 c) Ref: OP-202 Rev. 56, Page 22, 3 FP-302-661, Sheet 3 of 4 ;
6.2 c) Ref: OP-203 Rev. 43, Page 3 )
6 3 CI M- 3?,[i vr n l Se ,M2i gl P: : 15 Rn. , , . , . .
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F Ref: Main Steam, Page 3, 8 6.4 JH CLe 6.5 b) Ref: Main Steam, Page 78 6.6 d) Ref: ICS Analog and Digitals, Page 6 6.7 b) Ref: Handout - Failure of both diesel generators to start, Page 3 6.8 d) Ref: CR Transient Assessment 6.9 d) Ref: CR-3 RPS Handout, Figure 2 6.10 b) Ref: CR-3 RPS, Page 8, Figure 2 6.11 a) Ref: CR-3 RPS, Page 14 6.12 a) Ref: NNI/ICS Power Supplies Handout 6.13 a)aA d) Ref: OP-605 Rev. 30, Page 29 6.14.b) Ref: TS 8 3/4 7-2 6.15 c) Ref: CR SLRM Handout, Page 6 6.16 b) Ref: OP-504, Page 8, 5.2.2.8 6.17 a) Ref: STM-22, Page 22-25 6.18 c) Ref: CR Training Letter TRA 85-0013 6.19 b) oh. d) Ref: STM-15, Page 6, 35 6.20 c) Ref: STM-4, Page 8 6.21 c) Ref: OP-403, Sec. 4.7.10, Page 4 6.22 b) Ref: STM 23-7 and OP-502, Page 3 l
i l
4 6.23 c) Ref: STM428-6/8 l 6.24 d) Ref: STM 11-24 6.25 1) B Ref: STGM 17-14 '
- 2) C STM 17-14 --
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0 1
5 CRYSTAL RIVER-3 Answer Key Section 7.0 7.1 c) Ref: AP-530 7.2 c) -
Ref: OP-202, Rev. 56, Page 23 7.3 c) Ref: OP-202 and TS 3.6.1.1, Pages 3/4 6-1 7.4 d) Ref: OP-204, Rev. 38 Page 7 7.5 b) Ref: OP-204, Rev. 38, Page 7 7.6 c) Ref: '0P-204, Rev. 38, Page 18 7.7 b) Ref: OP-302 7.8 c) Ref: EP-390, Page 5, 7 7.9 c) Ref: RP-101, Rev. 20, Page 17 1
7.10 c) Ref: RP-101, Rev. 20, Page 23 gfgjlgy, 7.11 c) Ref: FP-203, Rev. 12, Page 8 7.12 a-T Ref: FP-601, Rev. 19, Page 52 '!_ -
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7.13 a-F Ref: FP-601, Rev. 19, Page 28 f b-T ..f - .
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7.14 - 7.17 Ref: Attached Procedures - 2.4 ~
7.18 b) Ref: AP-542 /C 7.19 a-T Ref: OP-412, Page 3-10, 18 ."'
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- REACTOR PROTECTION SYSTEM ACTUATION '
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i, p RPSA REV 02 Date 05-25-84 AP-580 l
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ACTIONS
) 1MMEDIATE REMEDIAL '
U.
Ensu're GRP 1-7 rods l'. Open 480V 8KRs l' inserted:
e Depress " Reactor Trip" 3305
. pushbutton !
3312.
[o Observe " TRIP CONF" [
- r. , 1ight 1It on diamond t. Start boration: i I
-, ,' panel, I
\ L a. Open 8WST suction 3
u b b. Start 2nd MUP :
A i'
L c. Open MUV-24 4* d. Establish letdown path
- to RCBTs.
7 n GVs Ensure fully main turbine TVs and closed. 1. Close MSIVs.
J 2.
i Select " ATMOS"on "TURB.
. BYPASS VLV" switch. t
]j. Ensure main block valves .
closed. 1. Trip both MFPs.
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- 2. Refer to AP-450. Emergency
. Feedwater Actuation. ,
)
j Ensure low load block valves closed. 1. Trip both MFPs.
4
- j- 2. Refer to AP-450 Emer Feedwater Actuat1on. gency
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3 ACTIONS (Cont'd) j
! 1 IMMEDIATE REMEDIAL l l l l
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- l. Ensure PZR level > 50". 1. Open suction from BWST.
k' 2. Start 2nd MUP. .
I d il l 3. Open MUV-24. ' '
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! 4. Close MUV-51.
i I close MUV-51, t.etdown Block Close MUV-49,t.etdown j ! .l Ortfice Bypass. Containment Isolation. ,,
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g'. Ensure STM HDR PRESS Hanually control STM HDR PRESS ; ;
I . controlling at 1010 PSIG. at 1010 PSIG using: l I
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RPSA REV 02 Dato 05-25-84 AP-580 ,ll a '
ACTIONS (Cont'd) '
- IMMEDIATE REMEDIAL i
- h. Ensure GEN output BKRs open:
i b e BKR 1661 i t* i e BKR 1662. '
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1 INADEQUATE CORE COOLING b ENTRY TNPTOMS CONDITIONS
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ICC REV 03 Date 09-20-84 EP-290 ;
i s' MTIONS s
IMMEDIATE. REMEDIAL
- 1. Ensure full HPI flow. Ensure:
- a. 2 HP! pumps are running.
] b. Open:
l
! , o MUV-23 o MUV-25 I
o MUV-24 o MUV-26.
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- c. HPI flow is > 500 GPM total flow. -
'I t 2. IF LPI is delivering flow, Ensure:
i THEN maintain maximum LPI o 2 LPI pumps are running.
!l TT5ii.
'1
- o Open:
- DHV-5 4 -
DHV-6.
F.
'! 3. Ensure OTSGs are at 95% on a. Select "CLOSE" on: *
.' the operating range. *
', o FWV-34 o FWV-162 -
I o FWV-35 o FWY-161.
~
- b. Trip both MFPs.
- c. Ensure b'oth EFPs start.
I
- d. Slowly raise OTSG Ievels to.95% using:
o FWV-162 (A-0TSG) o FWV-161 (B-0TSG) i f EP-290 Page 2 of 20 y ICC l
l 1
.p ICC REV 03 Date 09-20-84 EP-290
.. ) . 0 Inadequate Core Cooling Sa' g! FOLLOM-UP --
M.. ACTIONS DETAILS
! U1 .1 i **, Ensure CFT isola Open:
valves are open. tion L
l o CFV-S f' '-
i
! o CFV-6. i 4
- f. I
.{ 1. 2 Lower and maintain OTSG i o
TSAT 90-110*F less than Lower and maintain OTSG n*
TSAT for the existing RC pressure using:
pressure.
iJ TBVs g . -
ADVs . \
L fP .
o l; Determine TSAT for existing 1 i y RC PRESS ANO required OTSG PRESS.
- ,f o Refer to saturated steam '
- ! tables.
I 1
- ps k
2s
(
li
- h 6
cEP-290 Page 3 of 20 ICC 9
w we - - - - - - -- - _ , _ _ -r -- -- -
i ESSA REV 03 Date 0g/20/44 f.P-300 j l reaf tfERED SAFEGUARDS SYSTEM ACTUATION 5 D a eroe- .... Q . , '
L N F O R ?.;; A y i ga g a gy g -
- c. r.. ::en.., )
sinrivna I
N 3tTIONS ,
{ ', Annunciator alares-
- o associated with RC PRESS has exceeded Engineered Safeguards ESFAS trip setpoint on 2 of i Systems Actuation: 3 channels.
4 i o o Automatic R8 PRESS has exceeded l
ESFAS trip setpoint on 2 of -
1 I
o Manual. 3 channels.
o Manual Engineered Safeguard System Actuation.
o Loss of Coolant Accident.
! 2. ES status 11ghts indicating .
- , actuation
, o LPI, RC PRESS < 500 PSIG 1
o
- R8 Isolation and Cooling. .
i R8 PRESS > 4 PSIG -
! .o R8 Spray, 28 PRESS -
i
> 30 PSIG.
eviewed By PRC _ M Da ;e _09-20-84 Mtg.# 84-36 l
pproved By NPM ,/ fp4Sbit Date /p//f/Pf ,
AP-380 -
Page 1 of 25 ESSA 9@M$ I
- - - , _ - , - - - . - . _ , _ - - , _ - . _ - - - . _ -- . . - - - , _ . _ _ _ . _ _ - - - _ _ - _ _ _ _ _ . . .-- - ___~ - . _ _ ,-_-- - - -_ __-
ESSA RE Y 03 Date 03-20-84 AP-380 I
l ENTRY (Cont'd)
- . SYMPTOMS ,
CONDITIONS
! 3. Subcooling meter or monitor I
indicating:
l l
RC SUSC00 LING l -
PRESSURE MARGIN i i PSIG 'F l
{
.\
i
(< ) 1500 < 20 l
4
< ~ 1500 < 50 .;
s 1
l l
1 'l i
i <
l 1
l . <
1
. -l
.\ i 1
i i
i t
i
- i j AP-380 Page 2 of 25 ESSA I
)i .
i i I
F l
c . . .. -
l,- ESSA RE Y 03 Bate 09-20 84 . AP-330 I ,
_ACT!9a5 IMME0! ATE LINE0!AL
{ 1.
i E AC PRESS < 1500 PSIG. 1. tysass E5 actsatics.
1 TWa depress 'MP! Actuatica* 2.
7iiiTsuttaa ' A' M *1'. Istars Es eqst;sest 13.5737 states.
i 1
. 3. Go to Y7-580.
i
- 2. Trip all ICPs.
Opes _affected 69007 Itts:
4 o 3101 o 1 3103 I o 3102 o i
3104 i
- 3. Ensure M71 tratas start: Ratify A3 0;eratcr to start o 2 M7! pumps affecte pas;(s) at 4150f ES ~
switcagear.
I,
. o SWPs
. o RWPs.
4 -
Ensure 0;es: SWSi suctica valves Estify 25 sperator to caes affecte: valve (s) Iccally, -
o MUV-53 1
j o I
MUY-73. t i
- 5. Ensure HPI valves open:
f Iotify A! operator to cpet j .
o MUY-23 o affected valve (s) locally, 1 MUY-25 j s MUV-24 c MUV-25. -
t
\
- AP-330 Page 3 of 25 j i ESSA 3
l
!. _ . ..--eN*
, - _ , v.- . _-_.,,..e.-----._-__m_.mW
_ _ . . = . . _ _ _ _ _ _ . _ _ _ _ _ __
Date 09-20-84 AP-380 RE V 03
. Essa -
ACTIONS (Cont'd)_
REME0!AL IMME0! ATE l
l Notify A8 operator to start I
- 6. Ensure LP! trains start: af fected pump (s) at switchgear:
o 4160V ES o OHPs o 480V ES o DCPs l
i RWPs.
o 4160V ES.
o l i
1 I
Notify AB operator to start ,
1 7.' Ensure EDGs start. ,
affected diesel locally.
I I
i
- 8. Ensure diverse containment i
isolation actuation.
I i
Place R8 sump pump in Notify AB operator to open
! 9. affected 8KG at MCC:
! " PULL-TO-LOCK":
o WDP-2A o Reactor 3A2 1
WDP-28. o Reactor 382. i o
)
i l
l i
I I
ESSA a AP-380 Page 4 of 25 l e
I i
l e w=
T l - L REV 00 Date 08-08-83 EP-140 I ^
_E E' ACTIONS ; ,
Ml IMME0! ATE REMEDIAL :l i
grt emergency horation: 1. Adjust batch controller to 1000 GAL. -
Establish letdown flow Z to MUT 140 GPM 2. Select RCST with highest _d__
boron concentration. 7 Open CAV-80
- 3. Establish flow to MUT. .-
Start Boric Acid Pump _
- 4. Open CAV-57. =I e CAP-3A d'
- 5. Start Boric Acid Pump: -~1
' ~
e CAP-3A '
e CAP-35. =,
t' e CAP-35. l k-
~
,d
== ,
e 98 -
. =
w C
5 -
$ -"'l
- ' -1 h E I ;
l _. =;l:ig 4 il E140 Page 2 of 3 ERC l
_m m
a
_ . . _ _ _ . . . =
l ( . , , -
f .
6 CRYSTAL RIVER-3 -
i Answer Key Section 8.0 l 8.1 a) ~Ref: Tech Specs 3.6.1.3,1.5,[3,3.1' 8.2 a) -
Ref: OP-204 Rev. 38, Page 8 ,
l 8.3 c) Ref: Main Steam Handout, Page 27 Tech Specs 3.7.1.1 Page 3/4 7-1
~
8.4 b) Ref: Tech Specs Page 8 2-2 '.
Requal Cycle 2,1984, J. P. Haerle EP-390 Encl 1 Page 11
~
8.5 b) Ref:
8.6 c) Ref: OSIM I III-9 8.7 c) Ref: OSIM I III-10 8.8 c) Ref: OSIM I IV-2 8.9 d) Ref: OSIM Gen 04 V-3 Policy Statement 84-1 8.10 c) Ref: GSIM V-21 8.11 c) Ref: OSIM V-22 8.12 d) Ref: OSIM V-2 8.13 b) Ref: OSIM VI-2 8.14 d) Ref: OSIM VI-3 i
8.15 c) Ref: Tech Specs STS 3.8.1.2 Page 3/4. 8-6 8.16 a) Ref: Ref: STS.3.6.1.4, 3.1.3.1, 3.1.3.2, 3.4.1 8.17 dl- Ref: FP-203, Page 18' 8.18 c) Ref: FP-203 Rev. 12, Page 13 8.19 c) Ref: EM-207, Page 6 8.20 a) Ref: STS 2.'1.2, Page 2-1 8.21 c) Ref: TS insert,1CR, 4/1/83 Refueling l
8.22 c) Ref: TS 6-19 8.23 a-T Ref: EM-202 ,6 b-F . ef c-T .C d-F ,(
t 7
8.24 c) Ref: Tech Specs 3.4.5, Page 3/4 4-6 Figure 3.4-5 4e
- I e
6 D
30 e
EQUATION SHEET Where my.2 m (density)1(velocity) area )1 = ( den si ty)2( vel oci ty)2( area )2
.----....--.--.1( ------ ---------.---------------------------------...
KE = mv2 PE = agh PEl +KEl+P1V1 = PE2+KE2+P2V2 where V = spuific 7 volume P = Pressure Q = &cp (Tout-Tin) Q = UA (Tave-Tstm)
...----------..---------------...--------------. Q = a(ht-h2) ----.------------
P=Po10sur(t) p = pg et/T SUR = 26.06 1
delta K = (Kef f-1)/Kef f CR1(1-Keff1) = CR2(1-Keff2)
M = (1-Keff1) SDM = (1-Keff) x 100%
(1-K
- --- - - . . .e f.f2. -) - - - - - -- - - - - - - - - - . . K e.f.f _ -- - - . . . - - - - - -- - - - . . . .. - - - - - - - - - - - -
decay constant = In (2) = 0.693 A=A ge (decay constant)x(t) t 1/2 t1/2 Water Parameters Miscellaneous Conversions a 1 gallon = 8.345 lbs 1 Curie = 3.7 x 1010 dps 1 gallon = 3.78 liters 1 kg = 2.21 lbs I ft3 = 7.48 gallons I hp = 2.54 x 103 Btu /hr Density = 62.4 lbs/f t3 1 Mw = 3.41 x 106 Btu /hr Density = 1 gm/cm3 1 inch = 2.54 centimeters-Heat of Vaporization = 970 Btu /lbe Degrees F = (1.8) x (Degrees C) + 32 Heat of Fusion = 144 Btu /lbm 1 Btu = 778 ft-lbf 1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 ft-lbm/lbf-sec2
--