ML20153F552
| ML20153F552 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 08/31/1988 |
| From: | Jordan M, Nejfelt G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20153F545 | List: |
| References | |
| 50-341-OL-88-01, 50-341-OL-88-1, NUDOCS 8809070336 | |
| Download: ML20153F552 (78) | |
Text
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U.S. NUCLEAR REGULATORY COMISSION REGION III Report No. 50-341/0L-88-01 j Docket No. 50-341 . License No. NPF-43 Licensee: The Detroit Edison Company 6400 North Dixie Highway i Newport, MI 48166 Facility Name: Fermi 2 Nuclear Plant Examination Administered At: Fermi 2 Nuclear Plant, Michigan Examination Conducted: August 2, 1988 Chief Examiner: d)#./' { d4 ////
'G. M. Nej f,eJ, Principal Date Examiner, sperator Licensing l Section No. 1 Approved By: J/
Michag)/ Jordan, Chief Date Operator Licensing Section No. 1 Examination Summary Examination administered on August 2, 1988 (Report No 50-341/0L-88-01)
One written examination was given to a reactor operator (RO) candidate as a retake.
Results: The candidate successfully passed this examination.
P"usa em PNU
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REPORT DETAILS.
- 1. Examiners G. Nejfelt, Chief Examiner
- 2. Exit Meetina At the conclusion of the site visit, the examiner met with facility
- training staff. The following personnel attended this exit meeting.
Facility Representatives G. R. Overbeck, Director, Nuclear Training R. W. Bovinet, Work Leader, Operator Training Program NRC Representative G. M. Nejfelt, Chief Examiner The following items are discussed during the exit meeting on August 3, 1988:
- a. Timely submission of examination comments from the facility training staff. These comments were received in the Regional Office on August 19, 1988.
- b. Materials that were prepared for the tentatively s;heduled Requalification Examinations starting the week of October 31, 1988 were discussed. Most of the written examination questions were found to be satisfactory, except for static simulator questions that could be answered without the use of the simulator.
No Job Performance Moasures (JPMs) were available for review.
- 3. Examination Review Responses to licensee's comments, concerning the written RO examination, are provided in Attachment 1. Minor changes made without facility comments by the examiner are provided in Attachment 2. Also, note that six comments were made concerning the examination materials provided (see Attachment 1 NRC Response to Questions 2.06.a. 2.07, 2.10.a. 3.02, 3.10, and 4.03).
Attachments:
- 1. RO Examination Comments and Resolutions
- 2. R0 Answer Key Modification Without Facility Comment 2
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, ATTACHMENT 1 ,
REACTOR OPERATOR (RO) f EXAMINATION COMMENTS AND RESOLUTIONS FERMI 2 R0 EXAMINATION OF AUGUST 2. 1988 The following represents the facility comments and the NRC resolution to those comments made as a result of the current examination review policy.
1.05.c Facility Comment: Comment answer for Part (c) is 518 seconds . . .
a math error was made in the answer key.
NRC Resolution: Comment is accepted. The answer key is changed from "318 seconds" to "518 seconds" to correct the typographical error.
2.01 Facility Comment: Correct answer for Part (c) of 2.01 should be 1153
+0/-20 psig and 441 +20/-0 psig. This licensing class was taught allowable values as listed in Technical Specifications.
References:
Reactor Recirculation System Handout, Page 18 Fermi 2 Technical Specification (TS),
Section 3.3.3 Fermi 2 IS, Section 3.3.4 Pump Trip System Instrumentation NRC Resolution: Comment is accepted, since the allowable maximum value agrees with the setpoint value and is more conservative than the setpoint value. The setpoint valves of 1133 psig and 441 have been changed to the allowable values and tolerances provided above.
2.02.a Facility Commen,t: In Part (a), the answer states that the backup air supply for the inboard main steam isolation valves (MSIVs) and J' normal air supply for the outboard MISVs is instrumeret c'. This answer could be misleading. The term used at f for this instrument air is called Control Air.
Specifically, INTERRUPTIBLE CONTROL AIR (IAS).
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References:
4 Student Handout, Main Steam System and Bypass '
System, Page 14.
Drawing 6m721-5730-5... f
, f.
NRC Resolution: , Comment is accepted. The use of the alternateinames ,
for instrument air (e.g., control air and Interrvptible ,
Control Air) are in addition to the answer as ' acceptable 7
alternate wording. : ,
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- 4 2.02.b j .
.l Facility Comment: Part (b) answer should be "130 VDC and 120 VAC '
l (RPS M-G Set)."
Reference:
L Student Handout, Main Steam System and Bypass i System, Page 14. !
i
! NRC Resolution: Comment is accepted. Typographical error is changed :
l from "125 VDC . . ." to "130 VOC . ." t t
2.02.d !
i Facility _ Comment: In Answer (d), the statement is made that the isolation will not reset. Assuming the isolation signal itself is' clear pushing the RESET pushbutton without first pushing l the close pushbuttons on the MISVs will cause the MISVs i to cpen.
References:
d' Student Handout, Main Steam and Bypass, t Page 14.
1 LER 50-341/87-037.
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61721-2095-14, 15, 17 and 18.
NRC Resolution: Comment is accepted, only if, the candidate stated that the solenoids for the MISV were deenergized and reenergized in an unorthodox manner (e.g. . LER 50-341/87 037).
2.06.a Facility Comment: On Part (a), the answer key should be changed to accept the power supplies to the RPS M-G Sets as correct actions for normal supplies. These would be:
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Attachment 1 3-
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- 1. Div I (EPS M-G Set A).- 480 VAC Hotor Control.
References:
Nuc. Prod. Sys. Operati, Proc. 23.316, Pages 8 and 9.
Deco Drawings 6I721-2151-1 and 2.-
r NRC Resolution: Comrant is accepted for both the primary and alternate RPS power supplies. if the candidate's' answer is in the same detail. Student Handout, RPS, PIC-C71, Revision 3, Page 4, suest ntiate the answer key as originally written.
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2.06.b Facility Comment: Part b should have four answe-s: !
- 1. Process Radiation Monitoring System. !
- 2. Power Range Monitoring System.
- 3. Nuclear Steam Supply Shutoff System. -
Reference:
Deco Drawing 6I721-2151-1 and 2. l NRC Resolution: Comment is not accepted. The question asked for systems ,
that received powr from RPS. Also, if schematics are l provided as supporting material, please ensure that the information is large enough and clear enough to be used withou* a tedious effort to extract the information.
I 2.07 !
I Facility C mment: Using the reference given on Pages 8 and 14, there is ;
no mention of High Rea-tor Pressure causing a Standby Gas Treatment (SSGT) System auto start. Thore is l
- e. a misprint on Page 14 that may lead one to believe r Part (ab High Radiation (ARM) on the Refuel floor. '
may be a correct response. 4.a.4 on Page 14 should ;
be Fuel Pool Ventilation Exhaust Radiation High as shown '
by Page 8 of the SBGT system, Student Handou+ and Alarm Response Drocedure 17D14. Therefore, the answer key should read: l 1 I l
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-, A- .,,,-.,, ,-,_ _,._ _ __ _. _ . . . _ _ , . _ . . _ _ _ _ , - , , , , . _ . _ _
Attachment 4
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., will: b, g will not: a,c,d,e,f,h A:
References:
Standby Gas Treatment System Student Handout, Pages 8 and 14.
Alarm Response Procedure 17014.
NRC Resolution: Comment is accepted and the answer key modified as above. Revision of Student Handout, SBGT, T46-00, Revision 4, Page 14, is needed, 2.08 Facility Comment: Even though not listed in the given reference, the word chugging is often used to describe the (..ndition of injecting the boron too fast. So an alternative phrase at the end of the answer would also be: Could Cause Power Chugging, i NRC Rs. solution: Comment is accepted. The word "chugging" is added to the answer key as an alternative wording.
2.10.a l Facility Comment: To be consistent with Question 2.01, the answer l
- for (a) should be 441 +20/-0 psig.
NRC Resolution: Comment is accepted. The answer key is changed to the allowable value and tolerances provided above.
Note that the asterisk used in Table 4 of Student l Handout, RHR, Revision 4 Page 48 is not defined. !
2.10.c [
Facili_ty Comment: Part (c) should read:
F015A/B: Remove initiation signal and push l
i Initiation Signal Reset Push Button i
and Leak Detection tine Break Push Button.
F017A/B: Five minutes after initiation signal j is received.
Attachment 1 5
Reference:
HRH Student Handout, Pages 32, 33, and 48.
NRC Resolution: Comments is not accepted. References did not support revising answer key.
2.10.d Facility Comment: Part (d) should be changed to read:
F017A/B can be closed (throttled) to control injection flow (Reactor Level).
This is because the push button for closure of the valve (F017A/B) wl'1 allow full closure if pushed long enough. .
Also, the way the question is worded (close inhibit) would lead one to think along the lines of open/close ;
not throttled, r
References:
RHR Student Handout, Pages 32, 33, and 48.
i Fermi 2 Technical Specification 3.3.3 Table 3.3.3-2.
NRC Resolution: Comment does not warrant changing answer key, since i "throttling" can be considered synonymous with "closing" for this question.
3.02 racility Comment: Using the Student Handout on RWM. referenced Fermi 2
! Technical Specifications and Proposed Design Change (PDC) 7050 it is possible to answer Part "a" several i different ways. The first way would be a word definition as stated in the answer key. The other way would be with specific numbers. Per the Student Text and Fermi 2 !
Technical Specification the MINIMUM allowable reset power level is 20% for the LPSP. The actual per PDC 7030 <
is 30%. the LPAP per the Student Handout is 35%. But per l PDC 7030, the actual value set in the Control Room is 40%. .
Therefore, a suggested addition to the provided answer I would be-i
- a. (1) . . . similar. Also accept 25 i 5%.
- b. (2) . . . similar. Also accept 35% to 40%. ,
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References:
Rod Worth Minimizer Student Handout, Page 4.
Fermi 2 Yechnical Specifications 3.1.4.1 Rod Worth Minimizer Proposed Design Change (PDC) 7030, Revision B.
NRC Resolution: Comment is partially accepted. The question elicited more ,
than a numerical values for Low Power Point (LPSP) and Low Power Alarm Point (LPAP). Partial credit will be given for range of values stated above. l Training material is needed to be revised with <
plant changes. l 3.04.c Facility Comment: Using the referenced Automatic Depressurization System Handout, Figure 2 and Detroit Edison Drawings 61721-2095 02 and 07 the correct answer for "c" should be ADS valves ;
remain as is. The reason is that BOTH Core Spray Pumps ;
in Division II must be running to satisfy the ADS logic. i i
Reference:
Automatic Depressurization Handout, Pages 7, 8, and Figure 2.
Detrott Edison Drawings 61724-2095-02 and 07.
NRC Wesolution: Comment is not accepted. With no 130 VOC, the solenoids for ADS valves will deenergize and close valves (e.g., ,
remove fuses to shut stuck open SRV). l 3.08 i Facility Comment: Using the Power Range Monitor and Technical Specification i reference plus Technical Specificatten Page 3/4 3-45 there are three parts of the answer that needs to be expounded. ,
- 1. For the information provided, Answer "c" could be
- YCS or NO. True eight inops LPRM inputs to CHANNEL 0 i APkM sill leave it with 14 operable inputs (minimum required to be operable). The questions of how man.y i operable LPRM inputs per level is not addressed. '
Therefore since eight LPRM's are inop, it could ,
4 be assumed there are < 2 operable per level which :
! would make that ARPM Tnop per Technical Specification. [
1 1
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Attachment 1 7 ,
Regarding Answers "b" and "f." Technical Specifications Page 3/4 3-44 lists the allowable value of Reactor Coolant System Recirculation Flow Comparator as < 11% flow deviation. Which is what the Student Handout on Power Range Monitor Page 13 used. The < means that value has been set at or below this value per Tech Specs. So if the actuation is set at 11% (the Tech Spac allowable value class was instructed to memorize), there would not be a rod
- block flow reference off normal. Therefore Answers 6 and 7 could be NO.
References:
Power Range Monitoe Student Handout Page 13 Technical Specification Pages 3/4 3-5 and 3/4 3-44.
NRC Resolution: For Item "c," your comment is accepted. Part "c" is deleted because of the question ambiguity; and the question point value is reduced from 3.00 points to 2.50 points.
For Items "b" and "f"'your comment is not accepted. As stated above, the greater than or equal to 11% difference is signal between Flow Unit "A" and Flow Unit "B" was memorized by the licensing class. Unless the candidate clearly stated or assumed flow deviation less than 11%,
no change to the answer key is made.
1 3.10 Facility Comment: Per the Recirculation Flow Control System Pages 10, 11, and 12 there are three flow limiters. No. 1 limiter first in line before the individual pump controller and Limiter 2 and 3 are af ter the Error Limiting Network and Individual Speed Controllers.
Their setpoints per Proposed Design Change 8294 are:
Limiter No. 1 - 30%
Limiter No. 2 - 42%
Limiter No. 3 - 48%
The indi.:idual pump controllers do not have any limits on them. The limiters are separate units. So Answer "a" should be as above. Also, Answer "b" for limiter No. I should be . . . not fully open OR feedwater . . . .
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Attachment 1 8
References:
Recirculation Flow Control Student Handout, Pages 10 through 12.
Detroit Edison Drawing 6I721-2105-7.
Proposed Design Change 8294. ,
NRC Resolution: For Part (a), comment is accepted. The answer key is revised to change the setpoint values and to added Limiter No. 3 based on Potential Design Change (PDC) No. 8294 that superseded Student Hardout, Recirculation Flow Control System (B31), Revision 4, which was used to prepare the question. Also, the point value.for Part (a) was reduced from 1.00 point to 0.75 point with each limiter setpoint with 0.25 point.
For Part (b), comment is accepted based upon Schematic Drawing, 61721-2105-7, Revision 01; and the answer key is changed from ". . . and . . ." to ". . . or . . ."
with either answer worth 1.00 point. Student Handout Recirculation Flow Control System (B31), Revision 4, Section 0.4.a(a) is in error.
3.11 Facility Comment: Using the Rod Sequenca Control System Student Handout Pages 4 and 6. The answer in "2" would more correctly be: "Neither Sequence Groups B 1-2 and B 3-4 Control Rods are fully out, and."
The answer is more complete because it recognizes that A and B Sequences are broken into Groups A 1-2, A 3-4, B 1-2, and B 3-4 on the Rod Sequence Select Switch.
Reference:
Rod Sequence Control System Student Handout, Pages 4 and 6.
NRC Resolution: Comment is not accepted because it is moot. Nothing is gained by specifying the subgroups for the "A" and "B" Rod Control Sequence Groups.
4.01 Facility Ccmnent: The answer in the above question was nnt specific.
Therefore, answers should also include:
L
Attachment 1 9 7
' 1. During plant startup (when RHR shutdown cooling !
is secured). ,
- 2. During plant shutdown (prior to RHR shutdown l cooling being started). '
not.able to be restored). ;
References:
l General Operating Procedures 22.000.03, Page 9 l and 22.000.10, Page 11. :
Abnormal Operating Procedures 20.205.01, Page 1. :
NRC Resolution: Comment is accepted. The question could be construed for plant situations when operation of recirculation pumps may be required to operate. The answer and key :
point distribution are revised to read:
- 1. Preventing temperature stratification or j in the Reactor Vessel (0.50). l
- 2. Retaining solids in suspension until they can ,
be removed (by the RWCU System to prevent their deposition in the bottom of the Reactor Vessel i of the CRD mechanisms) (0.50). j i 3. During plant startup (0.25), when RHR shutdown l cooling is secured (0.25). :
r
- 4. During plant shutdown (0.25), prior to RHR shutdown [
cooling being started (0.25). j
- 5. During a loss of RHR shutdown cooling (when RHR 'I cannot be restored) (0.50). [
t (Note: Only two of the items above are required I for full credit.)
4.03 f Facility Comment: Although not listed as an indication in Section 3.0 l
.i of AOP 20.106.04, there is another indication listed in j Step 2.1.1. This step has the operator check LRPM (POWER) as the Rod is moved to verify the rod is recoupled. This i method could also be used prior to the rod reaching i j Position 48 by looking for no LPRM chango during rod i movement. j i
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Attachment 1 10
Reference:
I AOP 20.106-04, Page 3.
NRC Resolution: Comment is partially accepted. POM 20.106.104, ;
Revision 4, Section 2.1.1 must include the condition of Section 2.1.2. Therefore, the answer key is j revised to read: ;
~
i "4. Observe a response to the rod movement through !
Nuclear Instrumentation (NI) response (0.25). i and Demonstrates that the Control Rod will not go to the withdraw overtravel position (0.25). !
(Any three of the above for full credit)." l Note: Use of the NI response as an indication should
, be added to Section 3.0 of POH 29.106.04.
! 4.11 r
Facility Comment: Using the reference (AOP 20.710.01 Pages 1 and 2) [
- the answer key should be expanded to include
If Reactor Building Vent Exhaust Radiation Monitor upscale trip occur, verify the following automatic ,
action occur:
{ 1. Reactor Building Ventilation System tripped. 'f I
- 2. Reactor Building Divisions I and II Supply and .
Exhaust isolation valves close. I
- 3. Pri Containment Purge and Vent Valves close, ,
- 4. SBGT System Auto Starts.
, 5. CCHVAC System aligns to Emergency Recirculation Mode.
Notify the NSS of the event, actions taken, and that it may be required to classify the event in accordance -
with Emergency Plan Implementing Procedure EP 101, ,
"Classification of Emergencies."
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9 Attachment 1 11 ,
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Reference:
AOP 20.710.01 Refueling Floor High Radiation, Revision 5, Pages 1 and 2. ,
NRC Resolution: Comment is partially accepted. The notification of the Nuclear Shift Supervisor (NSS) is added to tne answer ,
key as an alternative answer. However, the question asked for immediate cperator actions for the "Refueling Flow High Radiation alarm." Therefore no credit is given for the "Reactor Building Vent Exhaust Rod Monitor" upscale trip immediate operator actions.
4.13 Facility Comment: Per Tech Spec 1.36, Page 1-6 0-Rings Bellows, and Welds should be acceptable for "Sealing Mechanism."
Reference:
Tech Spec 1.36. Page 1-6.
NRC Resolution: Comment is acceptable. Answer key Item 5 is revised with these examples to read:
, "5. The sealing mechanism associated with each secondary containment penetration (e.g., welds, bellows, or 0-rings) is operable."
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.' ATTACHMENT 2 FERMI 2 REACTOR OPERATOR EXAMINATION AUGUST 2, 1988 ,
- CHANGES MADE-TO ANSWER KEY WITHOUT PROMPTING FR_0M FACILITY - l Question Number Comment 1.11 Answer key esterisk at point (0, 1) is deleted, as a typographical error. Also, the range for the allowable value on the x-axis is increased from "34-36" to "32-38," i because the graphical accuracy intended would not be '
acnieved without the usa of graph paper and straight ;
edge ruler.
1.12 Point value for Question 1.12 was errcneous given as "2.00 points." The point value for this question has been charged to "1.50 points" to agree with the answer i key. l I
1.13 Point values redistribuiwd between Parts (a) and (b) to !
reflect effort needed in response. Part (a) vslue was !
changed from "1.00 point" to "1.50 points" and Part (b) value was changed for "1.00 point" to "0.50 point."
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i ENo osaae(t U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION s
FACILITY: EFJQU _1 _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ .
REACTOR TYPE: BWB-qed __________... ____
DATE ADMINSTERED: BB4004Q2________________.
EXAMINER: E t _8 A _8 6 B E . _ . . _ _ _ _ . _ _ _ _ _
, CANDIDATE _____.. ________________.
IUSIBUCIl001_IQ.C6UQ1061El U30 separate paper for the answers. Write answers on one side only.
Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the Question. The passing grade requires at least 70% in each category and a final grade of at loest 80%. Examination repers will be picked up six (6) hours after tho examination starts. ,
% OF CATEGORY % OF CANDIDATE'S CATEGORY
__V8LUE _IDI8L ___SCQBE___ _YeLUE__ ______________C6IEQQBl 2 5.48 : . 73- ____________
fi 22._ _21201 . _ _ _ _ _ _ _ _ _ . . ____.... 1. PRINCIPLES OF NUCLEAR POWER 2 5'. f t PLANT OPERATION, THERMODYNAMICS,
,'l) f.T) gj,(g. HEAT TRANSFER AND FLUID FLOW
. . . ,y
_aitsg__ _2' 00 ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY s ,,'
g AND EMERGENCY SYSTEMS Er hb "
.g . . ____.______ ___.____ 3. INSTRUMENTS AND CONTROLS
_2diSQ_ .";i 22 ___________ ________ 4. PROCEDURES -
NORMAL, ABNORMAL, i-
EMERGENCY AND RADIOLOGICAL "inf. ai7_ UE/.'I5- CONTROL mv . * ( '
.12;&i__ __________. __ _____% Totalt 7J, 7 g, Final Grade All work done on this examination is My own. I have neither given nor received aid.
Candidate's Signature MAS ~EPs CC0Y 1
y .
. e s .
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS '
During the administration of this examination the following rules apply:
- 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all conteuts with anyone outside the examination !
room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil only to facilitate legible reproductions, '
- 4. Print your name in the blank provided on the cover sheet of the oxamination.
- 5. Fill in the date on the cover sheet of the examination (if necessary). l
- 6. Use only the paper provided for answers. ;
t i
- 7. Print your name in the upper right-hand corner of the first page of each i cection of the answer sheet.
9 8.
j Consecutively number each answer sheet, write "End of Category __" as copropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer sheet.
] 9. Number each answer as to category and number, for example, 1.4, 6.3.
s f10. Skip at least three, lines between each answer. l t
i i 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly used in facility literature. !
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- 13. The point value for each question is indicated in parentheses after the l Question and can be used as a guide for the depth of answer reoutred. j i
l 14 Shom all calculations, methods, or assumptions used to obtain an answer !
to rathematical problems whether indicated in the Question or not. '
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- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE l QUESTION AND 00 NOT LEAVE ANY ANSWER BLANK.
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- 16. If parts of the examination are not clear as to intent, ask questions of j the examiner only. '
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistence in completing the examination. This must be done after the examination has been completed.
O 4
- 18. When you complete your e x am i n e t' i o n , you shell:
- e. Assemble your examitiation so f ollows:
(1) Exam questions on top.
(2) Exem aids - figures, tables, etc.
(3) Answer pages including figures which are part of the answer.
- b. Turn in your copy of the examinetton and all pages used to answer the examination Questions.
- c. Turn in all scrap paper and the belance of the paper that you did ,
not use for answering the questions.
- d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
I li__EBINCIELES_QE_WUCLE8B_EQWEB.EkeUI_QEEBoIIQUt Page 4 IBEBdQQIbeb1 cst _BEel_IBeBSEEB_eUQ ELUID ELQW QUESTION 1.01 (1.50)
Increasing recirculation pump speed will cause WHAT change (INCREASE, DECREASE, or REMAIN THE SAME) in each of the following parameters?
(Assume normal operating conditions) (1.5)
- a. actual bundle power
- b. critical power
- c. critical power ratio l
i QUESTION 1.02 (1.98)
Match each of the power distribution limits (1, 2, 13) with its associated FAILURE MECHANISM in Column A and its associated LIMITING CONDITION in Column B. (2.0)
- 1. Linear Heat Generation Rate (LHGR)
- 2. Average Planer Linear Heat Generation Rate (APLHGR)
- 3. Minimum Critical Power Ratio (MCPR)
Column A Column B FAILURE MECHANISM LIMITING CONDITION i
A1. FUEL CLAD CRACKING DUE B1. 14 PLASTIC STRAIN TO LACK OF COOLING i A2. FUEL CLAD CRACKING OUE TO B2. PREVENT TRANSITION BOILING HIGH STRESS FROM PELLET
! EXPANSION A3. GROSS CLAD FAILURE DUE TO B3. LIMIT CLAD TEMP TO 2200 F
, DECAY HEAT AND STORED HEAT FOLLOWING A LOCA QUESTION 1.03 (1.00)
Using the steam tables, calculate a reactor cooldown rate (F/hr) for a reactor pressure decrease from 1000 psig to 2SO psig in one hour and forty five minutes (105 minutes total) SHOW ALL WORK.fcr full credit.
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
It__EBIUCIELES QE.UWCLE60 20 WEB.EL8bl.0EEB61100t Pego 5
,I B E B d Q DI U e dl G S & _ U E 61_ I B e O S E E B _6 b D.E L V I D.E L Q W QUESTION 1.04 (2.50)
For each of the following events, state which COEFFICIENT of reactivity (fuel temperature, moderator temperature, void) would act FIRST to change reactivity.
- a. Control rod drop at power (0.5)
- b. SRV opening at power (0.5)
- c. Loss of shutdown cooling when shutdown (0.5)
- d. One recirc pump trips while at 50% power (0.5)
- e. Loss of one feedwater heater at 100% power (extraction steam isolated) (0.5)
QUESTION 1.05 (2.50) tv. ...+c. 3 v , ,y v . ,,3+3,.i .+ ins m tew ...y e m 4+v +v.
et:-t;et e;r=leeibi; ,teb!: pe;itive peried elledd t, 7 OOP 22,000.00 "Sterte; "re 0;;d thette.r te n;;;d "; ;r". R;;;ter pewer i; deteraire
- v um 4 0 '. v ,t .. . O v7
- TJ , . A SHOW ALL WORK.
AssJn Josh & % k,, // ' fu&I, tenennin rease%r~ peered
- a. What is the s test permissible stable reactor period, as stated in the caution cf POM 22.000.03, STARTUP FROM COLD SHUTDOWN TO RATED POWER 7 (0.5)
- b. What is the doubling time for a constant reactor period of 2 minutes?
(lg., de') on WMel eS Wi b
- c. How long 111 it take for pomer to reach the point of adding heat ^1f a period of 75 seconds is maintained. (1.5) y,% eelvs fov% *s< ,'NN b%'d
-664VA N D*1y I $
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QUESTION 1.06 (2.00)
ANSWER each of the following questions given that the reactor is at 100%
power and 1000 psig, when a relief valve inadvertently opens.
- a. STATE the tailpipe temperatur3, assuming atmospheric pressure in the Suppression Pool and No Reactor Depressurization (0.5)
- b. If the Suppression Pool Pressure were to INCREASE, STATE whether i
the Te11 pipe Temperature would INCREASE, DECREASE, or REMAIN THE SAME. (0.5)
- c. If the reactor starts to depressurize when the valve is opened, STATE whether the Teilpipe Temperature will INITIALLY INCREASE, DECREASE, or REMAIN THE SAME. (0.5) t
- d. STATE the Reactor Pressure at which the Te11 pipe Temperature would be at its MAXIMUM value (during the depressurization). (0.5) i (ASSUME A SATURATED SYSTEM AND INSTANTANEOUS HEAT TRANSFER) 4 l
l QUESTION 1.07 (1.00) j The reactor trips from full power, equilibrium XENON conditions. Twenty.
four (24) hours later the reactor is brought critical and power level is 1 maintained on range 5 of the !RMs for several hours. Which of the f ollowing i statements is CORRECT. (1.0)
- a. Rods will have to be withdrawn due to XENON build-in.
1
- b. RnJs will have to be rapidly inserted since the critical reactor mill cause a high eate of XENON burnout.
- c. Rods mill have to be inserted since XENON will closely l follows its normal decay rete.
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- d. Rods will appromimately rematri as is as the XENON estab-
]
11shes its equilibrium value for this power level.
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It__tBIUCl2LES_0E_SUQLE68_20 WEB _2LeUI_0EEBell086 Pcgo 7 l IBEBbODIded1CS4_dE81_IB80SEEB_880_ELW10_ELQW QUESTION 1.08 (2.00)
Answer each of the following questions either TRUE or FALSE I
d A reactor heat balance was manually calculated during the midnight to 1 i 8 a.m. shift, because the Process Computer was inoporable. The gain adj ustment factors were computed, but the APRM gain adj ustmer,ts have not i been made.
- a. If the feedwater flow rate used in the heat balance calculation was LOWER than the actual feedwater flow rate, then the actual power is HIGHER than the currently calculated power. (0.5)
I
- b. If the reactor recirculation pump heat input used in the heat balance
- calculation was OMITTED, then the ac'tual power is LOWER than the
, currently calculated power. (0.5)
I c.
If the steam tiow used in the heat balance calculation was LOWER than the actual steam flow, then the actual power is LOWER than the
- currently calculated pcwer. (0.5) 1 d. If the RWCU return temperature used in the heat balance calculation was LOWER than the actual RWCU return temper ature, then the actual power is
, HIGHER than the cur rently calculated power. (0.5)
QUESTION 1.09 (2.00)
Concerning the Net Positive Suction Head (NPSH):
- a. DEFINE NP$H 1
- b. For each of the following, INDICATE whether the available NPSH at the suction of the recirculation pump would !NCREASE/ DECREASE /
REMAIN THE SAME: (1.5)
(1) The Feedwater Flow is INCREASED (2) The Recirculation Flow is INCREASED (3) The Vessel Pressure is INCREASED from 200 psig to 800 psig l
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QUESTION 1.10 (1.00) -
SELECT the answer below that would typically coincide with the MAXIMUM ;
Control Rod Worth, during a reactor startup (1.0) ,
i
- s. Cold Shutdown prior to the startup l
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- b. Heatup i n Progress (approximately 14 Reactor Thermal Power (RTP)) -
- c. Heatup Complete (approximately 1% RTP) i i
- d. 50% RTP !
- e. 100% RTP [
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It- EBIWCIELES DE NUCLEeB 20 WEB EleUI DEEB8IIQUt Pago 9 IBEBdDQ166dlC1&_BE61 18681EEB 680 ELVID ELQW QUESTION 1.11 (2.00)
USE the 1/M hiot and PREDICT the number of control rods required to be withdrawn to. achieve criticality, P4ef su4 de/a pinf g/d> 6 e/e.v en g 4 NOTES: 1. CR = Count Rate
- 2. USE THE FIGURE BELOW TO SKETCH YOUR SOLUTION Each CR reading is recorded following a 5 rod withdrawal with CR0 representing 100% rod density. ,
1l Number of Contr ol Rods Withdrawn Count Rate 0 40 5
50
,, 10 89 15 129 I 20 191 25 333 30 800 5 10 15 20 25 30 35 40 45 50 55 i 1.01----l----l----l----l----l----l----l----l----l----l----l---11.0 0.9- -0.9
- - l 0.8- -0.8 ;
0.7- -0.7 1/M 0.6- -0.6 0.5- -0.5 0.4- -0.4 0.3- -0.3 0.2 -0.2 0.1- -0.1 0.01----l----l----l----l----l----l----l--..-t----l----l----l---I O 5 10 15 20 25 30 ' 35 40 45 50 55 Control Rods Withdrawn
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It__EBINCIELES_DE UUCLEeB_EQWEB_EleNI_DEE86IIQNi Pcge 10 IBEBdQQ1Ned1 Cit _UE61_IB881EEB_8ND_ELU10_ELQW
/.'re QUESTION 1.12 ( )
An i ric r e as e in void content in an operating reactor causes a negative reactivity insertion. DESCRIBE THREE (3) effects which cause the void reactivity coefficient (alpha-v) to be negative.
QUESTION 1.13 (2.00)
Concerning the effects of Control Rods on reactor power: !
- a. EXPLAIN how a control rod withdrawal at certain power levelscould[j,r) ,
result in a reactor power decrease (reverse power effect).
Q<t ) l
- b. Which one of the following would be the rod movement sequence (<,,r')
most likely to cause the reverse power effect? )
i
( (1) Deep Rod - 10 notch movement (2) Deep Rod - 1 or 2 notch movement (3) Shallow Rod - 10 notch movement t
(4) Shallow Rod - 1 or 2 f.otch movement ,
i QUESTION 1,14 (1.00) -
MULTIPLE CHOICE (Select the ONE correct answer.)
i The Doppler Coefficient of Reactivity correlates the change in fuel j temperature to a reactivity insertion.
Which st atement is TRUE concerning Doppler Coefficient?
- a. The coefficient becomes lest negative with fuel burnup, and more negative with control rod withdtawel.
- b. The coefficient becomes more negative with fuel temperature l increase and less negative with void fraction increase.
- c. The coefficient becomes less negative with control rod withdrawal, l and more negative with fuel temperature increase.
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- d. The coe f f ic ie nd, becomes more negative with void fraction increase ;
and less negative with fuel temperature increase.
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It. 2BINCIELES_QE_NWCLE88_20 WEB _ELeNI_DEEB6I1QNa Pego ,11 I,WEBUQDINed101&_BE61_IB861EEB_8ND_ELVIQ.ELOW QUESTION l'.15 (1.50) ; ,
Suppose Beff over core life decrease from 0.0072 to 0.0055. With equal insertions of 0.001 dK/K of positive reactivity:
- a. Calculate the change in the reactor period over core life. (1.0)
(5.0)
- b. What is the cause for this change in Beff?
i p
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2t__EleNI_DESIQ0_INCLUDIND_S8EEII.8ND_EdEBGENCY Page 12 111IEd1 2.n QUESTION 2.01 (W
- e. What will cause an RECIRC SYS A (B) STARTUP SEQUENCE INCOMPLETE o annunicator when starting a recirculation pump? (
- b. What automatic action occurs, if an incomplete start sequence is detected? (0.$)
- c. Describe any automatic actions that occur in the recirculation system as a result of REACTOR PRESSURE. INCLUDE SETFOINTS/ SEQUENCES AS APPLICABLE. (1.0)
QUESTION 2.02 (3.00)
Answer the following questions cont.orning the Main Steam Isolation Valves (MSIV)
- a. State the applicable NORMAL and BACKUP pneumatic suoplies to the Inboard MSIVs and to the Outboard MSIVs. (1.0)
- b. What provides the power for the MSIV normal operating solenoids and logics? (1.0)
- c. How does the loss of power to one of the MSIV solenoids affect opening and closing ability of that MSIV? (0.$)
- d. An MSIV isolation (Group I) has occurred. The operator attempts to reset the isolation without depressing the MSIV close pushbuttons.
Does the isolation reset, and if so, what happen *. to the MSIVs? (0.$)
QUESTION 2.03 (2.00)
Describe the featurddaf the Safety / Relief Valves (SRV) that results in increasing the amount of energy released per SRV opening. (2.0)
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. . 1 24 2L8NI DES 198.INGLVDING 16EEII_eND_EbEBGENCY Pcgo 13 111IEb1 QUESTION- 2.04 (2.00)
, s. The RCIC system has received a valid initiation signal and is
, ramping up in speed when the RCIC oil pump fails. HOW will the turbine respond (assuming the turbine does not trip on low oil pressure) and WHY? (1.0)
- b. The HPCI system has received a valid initiation signal and its auxiliary oil pump fails to start. HOW will the turbine respond and WHY? (1.0)
QUESTION ,,
2.0$ (2.$0)
The Control Rod Drive Hydraulic System interfaces with (interacts with) several in-plant systems /eoutpment. EXPLAIN the interaction with each of tho following systems.
- a. Control Air System
- b. Reactor Building Closed Cooling Water
- c. Recirculation Pumps
- e. Reactor Manuel Control System QUESTION 2,06 (2.50)
Antwer each of the followino Questions concerning the Reactor Protection System (RPS)
- a. STATE the normal and alternate power supplies to RPS (1.0)
- b. List three (3) systems that receive electrical power from RPS (1.5)
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QUESTION 2.07 (2.00)
State whether the following conditions or signals WILL or WILL NOT cause initiation of the Standby Gas Treatment System (SBGT) system: (2.0)
NOTEi Do not consider setpoints. If the indicated parameters will initiate the sy? tem, assume the setpoint has been reached.
.a. High radiation (ARM) on the Refuel Floor
- b. High radiation in the reactor building ventilation exhaust
- c. High particulate activity in the drywell
- d. Low flow in the offges system
- e. High reautor pressure
- f. Low RPV pressure
- g. Low RPV water level
- h. Main Steam Line high r adiation QUESTION 2.08 (2.00)
The Standby Liquid Control System (SLC) inj ects sodium pentaborate solution into the rewctor coolant. The pumps are designed to limit the boron injection rate. WHY (i.e. what is the basis) are there limits at which the solution must be injected (both upper and lower limits must be discussed for full credit)? (2.0)
QUESTION 2.09 (2.00)
- a. What is the purpose of the Rod Block Monitoring System? (1.0)
- b. The RBM gain change is done so that the RBM output will be equal to or greater than the reference APRM output. VHAT is the reason for changing the gain? (1.0)
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QUESTION 2.10 (2.50)
In regards to the RHR loop, injection valves, F015A/B and F017A/B:
- a. WHAT interlock must be satisfied to open BOTH inj ect ion valves in a loop? (0.5)
- b. WHAT is the purpose of this opening interlock? (0.5)
- c. Following an automatic initiation, WHAT INTERLOCK ($) must be satisfied to close each of these valves? (1.0)
- d. WHAT function is allowed after satisfying the close inhibit interlock (s)? (0.5)
QUESTION 2.11 (2.00)
Concerning the RWCU System
- a. L!$T four (4) conditions that will automatically isolate the RWCU system. ($ETPOINTS NOT REQUIRED) (1.0)
- b. During startup, while blowing down the reactor, the rate of blowdown is limited. What limits the rate of blowdown and what is the reason for this limit. (1.0) i I
(***** END OF CATEGORY 2 *****)
24..INIIBUDEUIS.eUD.G0 NIB 0L1 Page 16 4
QUESTION 3.01 (2.00)
State whether each of the following parameters directly initiates a scram, rod block, both, or neither. Setpoints are not required. (2.0)
- a. Main steam line radiation
- b. Neutron flux
- c. Reactor vessel high level
- d. Recirculation flow 1 QUESTION 3,02 (2.50) n
- Answer each of the f ollowing questions concerning the f.od Worth Minimizer
! (RWM) system
- s. OEFINE or DESCRIBE (1.0)
J J
(1) Low Power Set Point (LPSP) i (2) Low Power Alarm Point ( L P A F' )
(1) How many withdraw errors will the RWM allow? (0.5)
(2) How many insert errors will the RWM allow? (0.5)
(3) What restrictions are imposed on rod movement if the allowable number of errors is exceeded? (0.5) i l
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24..INSIBW5ENIl_86D.GQUIBQLS Pago 17 E
QUESTION 3.03 (3.99)
Assume the feedwater level control system is being operated in 3. element control using reactor level detector channel "A". Reactor power is at 85%,
steady state. For each of the instrument or control signal failures listed below, state how reactor level will initially respond (increase, decrease, or remain constant) and briefly explain why, in terms of what is happening in the Level Control System and Feedwater System immediately following the fa11ure. Ale $4 4 4e+1 oc**es.
NOTE: A block diagram of the feedwater level control system is attached j (figure 1)
- s. Channel "A" reactor level detector signal fails low.
- b. Loss of signal to "B" f eedwater control valve M/A transfer station.
- c. "B" feedwater flow signal fails high.
l QUESTION 3.04 (2.50)
- l For the following situations, state whether the Automatic Depres.
] surization System (ADS) relief valves will OPEN, CLOSE or REMAIN AS !$.
Consider each set of conditions separately.
I i s. ADS initiating signal sealed in, ADS valves open . . . . reactor j water level then rises to 177 inches. (0.5)
! b. ADS initiating signal uealed in, ADS valves open . . . . ADS j timer reset buttons are then depressed. (0.5) l
- c. ADS initiating signal sealed in, ADS valves cpen . . . . then a DC power failure occurs that affects all busses supplying AD$
valves. (0.5)
- d. ADS inttleting par ameter s present, a loss of the pneumatic supply to the drywell has occurred, 120 second timer timing out . . . . then the 120 second timer times out. (0,5)
- e. ADS initiating parameters present, all ECCS pumps are secured except for CS pump B which is running with a discharge pressure of 195 psig. 120 second timer timing out . . . . then the 120 second timer times out. (0.53
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. QUESTION 3.05 (2.50)
The Core Level Indicator (L1-R610) located on Control Room Panel i H11.P601 has an indicating range from .150 to +50 inches. I
)
- a. WHAT is INSTRUMENT 2ERO for this level sensort t (Provide the core component located at
[
Instrument Zero.) 10.5) !
I b. Is this level sensor temperature compensated? (0.5)
- c. MATCH the core conditions in Column 1 with the correct response of the core level indicator (LI-R610) in Column 2 . (1.5) {
COLUMN 1 COLUMN 2
- 1. LPCI is the only system a. Full Scale l i nj ect i ng into the vessel i
- b. Downscale i
- 2. No recirculation flow exists, no systems are injecting to c. Actual Level [
the RPV, and the reactor is I i
at atmospheric pressure d. 14 inches HIGHER than actual level ;
- 3. Both recirculation pumps are at 45% speed e. 14 inches LOWER than !
actual level 1 i
QUESTION 3.06 (1.00) (
L HOW would an SRM detector respond to a pinho'e leak which !
caused a gradual decrease in A' gen gas pressure?
{
- a. Gamma and neutron sensitivity would DECREASE.
- b. Garma sensitivity would DECREASE but neutron sensitivity would REMAIN THE SAME. i I
- c. Gamma sensitivity would REMAIN THE SAME but !
neutron sensitivity would DECRE ASE.
J
- d. BOTH gamma and neutron sensitivity would REMAIN THE SAME.
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2t__INSIBUDENIS.eND_CONIBQLS 'Pegobl9: '
QUESTION 3.07 (1. 41 LIST three (?~ v icns and theli etpoints.(it applicable)-that will automatically station air compressors.- (1.5)
QUESTION 3.08 s# ,J Consider the following inf ormat ion:
- the Reactor is at 100% power
- APRM CHANNEL D is readiiig 1024
- F L 0tJ UNIT A is reading 90%
- FLCV UNIT B is reading 100% '
- 8 t.PRM inputs to APRM CHANNEL D are bypassed STATE whether each of the following will occur. (YES or NO) occur.
e) EPS DIV I trip (1/2 scram) (0.5) b) Control Rod Withdrawal Block (0.5) l g.02 ^ O I ' 0 :- cfdk/Ag (0.5) d) APRM D upscale high (0.5) e) APRM D upscale Hi Hi (0.5) l f) Flow Reference Off Normal (0.5) l l
QUESTION 3.09 (1.50)
- a. WHAT automatic actions occur when the scram discharge volume high level scram bypcss switch is placed in BYPAS$ and +he scram is reset.
- b. WHAT position (s) must the roactor mode switch be i to allow BYPASS o* the scram discharge volumo high level scram function?
- c. WHAT additienal protective function is inserted when the scram bypass switch is in the BYPASS position?
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-QUESTION;' f 3 '.'10, ' ( 3. 0 0 3 -
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- & There are THREE9(3)L.opeed.11 miters 71n the Recirculation flow control system
'T
- c which functionit'o~?1imit:.the' -
maximum'spe'ed demand.
~
a.
~
WHAT is the ma x imum i s p'o e d demand limit imposed by EACH limiter?
.,c '
(LIST EACH LIMITER'ANDcTHE APPLICABLE SETPOINT.) (1.0)
- b. During what pfantJconoitions are the #1 & #2 limiters in control and imposing limitetio'ns on Recirculation pump spoed?
~
<l (SETPOINTS REQUIRED.). (2.0) l.
-QUES 110N 3.11 (1.50)'.
T STATE three (3)Sconditi'ons that would cause the Rod Sequence Control System Annunicator, ALL"A/0l SEQUENCE RODS NOT FULL OUT, to alarm.
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(***Sv END OF CATEGORY 3 *****)
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st__EBOCEDUBES_ _NDBd6Lt_6BNOBd8Lt_EdEBGENQ1 Pcgo 21 l 8ND_BeQ10LQQ196L_CQUIBQL QUESTION 4.01 (1.00)
Operation of the Recirculation Pumps at a suction pressure below 300 psig should be minimized since such operations can contribute to shortening seal life. However, STATE two reasons that may require recirculation pump '
operations at low pressure.
QUESTION 4.02 (1.00)
STATE two (2) alternate methods of scramming the reactor per AOP 20.000.19, "Shutdown From Outside The Control Room".
QUESTION 4.03 (1.50) ,
During reactor startup, a control rod is to be withdrawn. LIST three possible indications that the control rod is uncoupled per AOP 20.106.04, "Uncoupled CR0".
QUESTION 4.04 (1.00)
SOP 23.202, "High Pressure Coolant Inj ection System", cautions not to defeat the automatic function of an ECCS by placing the controls in MANUAL or OFF unless confirmed by at least two independent indications. STATE these two (2) indications.
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7 ft__EBQQEDUBES_:_UQBdekt_6BUQBb6Lt_EbEBGEUQY Pcgo 22 AUQ 88010LQQ108L_QQUISQL QUESTION 4.05 (2.00)
For each of the following conditionc (a through d) INDICATE which of the corresponding procedures (1 through 4) would need to be entered. (More than one procedure may apply for each condition. If none of the procedures are required to be entered then state "NONE").
CONDITIONS
- a. Crywell pressure has increased to 3 psig ,
- b. Reactor water lovel cannot be determined ,
- c. Reactor water level decreasing to +160"
- d. Grywell Equipment Sump High Level Annunicator alarms PROCEDURES
- 1. 29.000.01- Level / Pressure Control
- 2. 29.000.02- Cooldown
- 3. 29.000.03 Primary Containment Control
- 4. 29.000.04- Contingency For RPV Flooding QUESTION 4.06 (2.00)
Per E0P 29.000.08, "Reactivity Control Procedure", if the MSIVs are open and the Maj, Condenser is available, the operator is to verify or manually runback ths Reactor Recirculation Pumps to minimum speed before tripping them. Explain why these pumps are runback prior to tripping them. (2.0)
QUESTION 4.07 (1.50)
Answer the following TRUE or FALSE with regards to the Criteria for
, Standing Orders.
- e. Standing Orders should be used to provide additional guidance on administrative matters.
- b. Standing Orders may conflict with procedural requirements.
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- c. Standing Orders can be issued prior to approval by the OE or A0E.
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di__EBQQEDuBES_:_UQBb6Lt_6tNQBd6Lt_EdEBGEUQY Pego 23
'eUQ_B6Q10LQQlC6L_CQUIBQL QUESTION 4.08 (2.00)
Match the following tag descriptions (a through d) with the correct type of tag (1 through 4)
DESCRIPTION
- e. May be placed to identify a short term condition or to explain limitations.
- b. Used to warn against the operation of electrical or mechanical equipment which could inj ury personnel.
., c. Used to mark the unsafe condition of equipment
- such as tools and ladders.
- d. Used for the protection of equipment or when determined to be required be the NSS.
TYPE OF TAG
- 1. Red Tag
- 2. Safety Tag l
- 3. Information Tag
- 4. Equipment Protection Tag QUESTION 4.09 (2.00)
Per GOP 22.000.03, "Startup From Cold Shutdown To Rated Power", when the reactor is critical, four items must be logged in the Control Room Log Book. LIST these four (4) items (some "items" may include more than one entry).
i 1
4
)
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. w At__EBQQEQUBES_ _NQBdelt_6?NQBd8Lt_EDEBQEUQ1 Page 24 600_B8010LQQ106L_QQUIBQL QUEETION 4.10 (2.00)
LIST the following syFtem/ equipment in the order they are taken ou'6 of service during a reactor shutdown from rated power with MSIVs open per GOP 22.000.10 Shutdown From Rated Power To Cold Shutdown.
- 1. Off Gas System
- 2. Lrst Reactor Feed Pump
- 3. Last Heater Feed Pump
- 5. Main Turbine QUESTION 4.11 (2.00) ,
A local Area Radiation Monitor ARM alarms on the refueling floor. Upon receiving indication or notification of a Refueling Floor high radiation condition, STATE the four (4) immediate operator actions in the control room?
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4 t__2800EQUBES_:_NQBb6Lt_6BNQBb6Lt_EDEBGENQ1 Pago 25 880_B601QLQQ196L_QQNIBQL l QUESTION 4.12 (2.00)
Match the following Emergency Classifications (a - d) with the appropriate description of that event (1 -
4). (2.0)
+
- a. General Emergency 1. Events which involve actual or imminent substantial core degradation or melting has occurred with a potential for loss ,
of containment integrity.
- b. Site Emergency 2. Any condition that involves an actual or potential substantial degradation of the level of safety of the plant.
- c. Alert 3. "Events which involve likely or actual major failures of plant functions needed for the protection of the public.
- o. Unusual Event 4. Any station related event which indicates a potential degradation of the level of safety of the plant, but which is not likely to effect onsite personrel or the public or result in radioactive releases requiring offsite monitoring.
QUESTION 4.13 (2.50)
List five of the six conditions that must be met to establish Secondary Containment Integrity. (2.5)
(***** CATEGORY 4 CONTINUED ON NF.XT PAGE *****)
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4 QUESTION 4.14 (2.00)
TRUE or FALSE- (2.0)
A General Radiation Work Permit (RWP) should be used for the following:
0 1. To enter posted Radiation Areas
- 2. To enter posted Contaminated Areas s
- 3. To' enter posted Airborne Radioactivity Area
- 4. To enter posted Neutron Radiation Area 4
st
(***** END OF CATEGORY 4 *****)
(********** END OF EXAMINATION **********)
i~ s li__ B16CIELE1_QE_UUCLE68_EQWEB_EL8UI_QEE86I1QUt- Pcgo 27 IHEBUQD16601Gi&_HE81_IB861EEB_86D_ELVID ELQW , s ANSWER 1.01 (1.50)
- a. increase (0.5)
- b. increase (0.5)
- c. decrease., (0.5)
REFERENCE GE Thermodynamic, Heat Transfer, and Fluid Flow Text pg.9-85, 9-86,' 9 92 FERMI Nuclear Power Plant Thermal Sciences pg.10 10, 10_15 K/A 293009 K1.23 (2.8/3.2) 293009K123 ..(KA's),
ANSWER 1.02 (1.98) f-FAILURE MECHANISM LIMITING CONDITION
- 1. LHGR A2 B1
- 2. APLHGR A3 83
- 3. MCPR A1 82 (6 6 0.33 each = 1.98)
REFERENCE FERMI Nuclear Power Plant Thermal Sciences pg 10-10, 10 15 K/A 293009 K1.08 (3.0/3.4), K1.12 (2.9/3.5), K1.20 (3.1/3.6) 293009K120 293009K112 293009K108 ..(KA's)
ANSUER 1.03 (1.00)
Obtain corresponding temp. from the steam tables by interpolation 1000 psig = 546.3 des F (.25) 250 psig = 406.0 deg F (.25)
Temp. change: 546.3 - 406.0 = 140.3 deg F (.25)
Rate of cooldown: 140.3/1.75 = 80.2 deg F/hr (.25)
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
V ^
Pago 28 lt__EBINGIELEE-QE_NUGLEeB_EQWEB_ELeUI_QEEB8I1 Qui IHEBdQQIUed1 Cit _BEeI_IB6HIEEB_eUQ_ELVID ELQW REFERENCE Steam Tables K/A 293003 K1.23 (2.8/3.1) 239003K123 ..(KA's) ,
s ANSWER 1.04 (2.50)
- a. Doppler or fuel temperature
% b. Void
'\ c. Moderator temperature
- d. Void
- e. Moderator temperature (5 at 0.5 each = 2.5)
REFERENCE FERMI Reactor Theory Fundamental og 7.5, 7.6 K/A 292004 K1.14 (3.3/3.3) 292004K114 ..(KA's)
ANSWER 1.05 (2.50) t
- a. From POM 22.000.03, shortest permissible stable period equals 50 (+/-5) seconds. (0.5)
- b. Doubling time = (2 min)(60 s/1 min)/1.44 = 83.3 4/- 4 seconds. .
(0.5)
- c. 40% range 2 is equal to 0.04% on range 8 (0.25)
P(0) = 0.04 P(t) = 40 Period = 75 seconds P(t) = P(0) e ^(t/ period) 40 = 0.04 e ^(t/75 sec)
Time = -&&arl + / - 2 0 ) seconds (1.25) 5/6 REFERENCE GOP 22.000.03 rev 15 pg 13 FERMI Reactor Theory Fundamentals pg 10.9 K/A 292003 K1.08 (2.7/2.8) 292003K108 ..(KA's) l l
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
s . -
y ,
IA__2BINCIELES_QE_NVCLE88_EQWEB_EL8NI_QEEB6I1QNi Pogo 29 3.UEBdQQ1NedICSt_BE61_IB8HSEEB eND_ELUID_ELQW
\q ANSWER .1.06 (2.00) s.i 296 deg F-(+/- 15 deg F) (0.5)
- b. ' Increase (0.5)
. A
- c. !peresse (0.5)
.d. 450 psia (+/- 50 psia) ,(0.5)
REFERENCE Mollier diagram Steam tables System Book 1, 03-15-05, MS an,d Bypass System,>Rev. 4, pg 31.
K/A 293003 K1.23 (2.8/3.1) 293003K123 ..(KA's)
ANSWER 1.07 (1.00) c.
REFERENCE ,
FERMI Reactor Theory Fundamentals pg 9.8 K/A 292006 K1.07 (3.2/3.2) 292006K107 ..(KA's)
. t
,a, ANSWER 1.08 (2.00)
- 4. FALSE (0.5)
- b. TRUE (0.5)
- c. FALSE (0.5)
- d. FALSE (0.5)
REFERENCE :
FERMI Nuclear Power plant Thermal Sciences pg 11-9 K/A 293007 K1.13 (2.3/2.9) 293007K113 ..(KA's)
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
y
\
\
It__EBINCIELESl0E_ NUCLE 8B_EQWEB_EL68I_QEE861108t Page 30 ,
IHEBbODIN001 Git _HE81_IBeNSEEB_6ND_ELVID_ELOW
.t I .
ANSWER 1.09 (2.00) e, P - P (or pressured nessured at the inlet of the pump) (0,5) set set
- b. '(1) INCREASE (More Subcooling at the pump suction) .
(2) DECREASE (Reduced pressure at the eye of the pump results in being closer to saturation pressure)
(3) DECREASE (Further to saturation temperature end increased discharged fluid density causing less static head) g (3 at 0.5 each = 1.5)
REFERENCE '
FERMI Nuclear Power Plant Thermal Sciences pg 17-20 K/A 293006 K1.10 (2.7/2.8, K1.08 (2.5/2.6) 293006K108 293006K110 ..(KA's)
ANSWER 1.10 (1.00) c (1.0)
REFERENCE General Electric Reactor Theory, Chapter 5 FERMI Reactor Theory Fundamentals pg 12.1 K/A 292005 K1.09 (2.5/2.6) 292005K109 ..(KA's)
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
N
. 1 li__EBIUCIELES_QE_ NUCLE 88 20 WEB _EL8NI DEEB6IIDUt Pcgo 31 IHEB00D1860101i_HEeI_IBeBSEEB eUD ELUID_ELQW ANSWER 1.11 (2.00) 31-4F Drawing a straight line between the last two s's predicts O' control rods must be withdrawn. 20 f (0.2 for each point plotted, 0.50 for line and prediction) 5 10 15 20 25 30 35 40 45 50 55 T 1
---l----I----I----I----I----1----1----I----1----I----1---11.0 0.9- -0.9 0.8- *
-0.8 0.7- -0.7 1/M 0.6- -0.6 0.5- -0.5 -
0.4- -0.4
, 0.3- * - 0 ,' 'J -
, 0.2-
- 0.2 0.1- *
-0.1
- s -
I O.OI----1----I----1----I--- '----I----X----1----1----I----I---1 0 5 10 15 20 25 30 35 40 45 50 55 Control Rods Withdrawn REFERENCE GE BWR Academic Series, Reactor Theory, Chap. 3, pg. 13-15, FERMI Reactor Theory Fund. pg 11.11 K/A 292008 K1.04 (3.3/3.4) 292008K104 ..(KA's)
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
r Iz__EBINQIELE1_QE_NWQLEeB_EQWEB_EleUI_QEEBeIIQUt Pege 32 IBEBdQQ1Ned1Q1t_BEeI_IBeUSEEB_6NQ.ELWID_ELQW X
/. Ce ANSWER 1.12 (M4 (1) More neutrons will be captured in the resonant peaks of the fuel (uranium and plutonium (0.5) (as the slowing down length increases).
(2) Thermal neutron leakage increases (0.5)
(3) Fast neutron leakage increases (0.5).
REFERENCE FERMI Reactor Theory Fundamentals pg 5.13 K/A 292002 K1.03 (2.0/2.1) 292002 K1.04 (1.9/2.0) 292004 K1.03 (2.6/2.7)
. 292004 K1.10 (3.2/3.2) 292004K110 292004K104 292002K104 292002K103 ..(KA's)
ANSWER 1.13 (2.00)
- a. The negative reactivity added by the increased voids generated by .,
the rod withdrawal is greater than the positive reactivity added by (/, 3'/
the reduced rod absorption. (1<0)
(C.D
- b. (4) M)
REFERENCE FERMI Reactor Theory Fund. pg 12.1 - 3 K/A 292005 K1.04 (3.5/3.5) 292005K104 ..(KA's) l ANSWER 1.14 (1.00) d (1,0)
REFERENCE FERMI Reactor Theory Fundamentals pg 8.22 K/A 292004 K1.05 (2.9/2.9) 292004K105 ..(KA's)
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
Pago 33 li__EBINCIELES_QE_NUGLE68_EQWEB EL881_QEEB6IIQNt JUEBdQQ1NedIGS&_BE61_IB6NSEEB_6NQ_ELUIQ_ELQW ANSWER' 1.15 (1.50)
- a. BOL: t=B- P/u P = (0.0072 - 0.001)/(0.1)(.001) m
= 62 seconds (0.33)
EOL t = (0.0055 - 0.001)/(0.1)(.001) = 45 seconds (0.33)
CHANGE 62 - 45 = 17 seconds (0.34)
- b. Buildup of Pu-239 coupled with the burnout of U-235 causes a decrease in the effective delayed neutron fraction (Beff). (0.5)
REFERENCE FERMI Reactor Theory Fundamentals pg. 10.4, 10.12 K/A 292003 K1.06 (3.7/3.7) 292003K106 ..(KA's) 1
(***** END OF CATEGORY 1 :: ar) e
2i__EL68I_DESIQU_IUCLUDINQ_18EEIX_8HD_EUEBGEUCI Page 34 31SIEt!S
.L OO ANSWER 2.01 ( Mr&)
e, C
- a. M-G set failed to complete startup sequence in <15 seconds 41.07
- b. The recirc M-G drive motor Lockout Relay trips (0.5)
- c. The recirc pumps trip (0.25) on high reactor pressure of /163 + 8/-18 ,p ,
y-La2.3.psig (0.25). The recirc pumps discharge valves shut (0.25) if a LPCI initiation signal is present (0.125) and reactor pressure decreases to 441 psig (0.125).
REFERENCE Recirculetion System, pg 10, table 3 and 4 K/A 202001 K1.12 (3.5/3.6), K1.27 (4.1/4.3), K4.07 (2.8/2.9)
K4.14 (4.0/4.1), K6.06 (3.1/3.1) 202001K606 202001K414 202001K407 202001K127 202001K112
..(KA'sj ANSWER 2.02 (3.00) WMU
??[Jf4et ww:% y Q fa c J d 4 % 4S b
- a. The pneumatic supply for the i r.b o a r SIVs is Nitrogen (0.34) with a backup supply from instrument air 0.33). The supply for the outboard MSIVs is instrument air (0.33).
/30 b . y 4M VDC (0. 5 ) and 120 VAC CO.5) (RPS acceptable for the 120 VAC)
- c. The loss of one solenoid will not affect the ability of the valve to open or close (0.5). Both solenoids must de_ energize to close the valve. If either is energized, the valve will open,
- d. The isolati not reset (0.5). ( CWh. h w *t> * /ns8. % ,,[ A M 'det? &on will M=: bt %4 (t.
T REFERENCEq gw % diihML4re44 4 Wnfa%,34f, f /g7,f./.j
)), & & x k t! d & &
Main Steam System and Bypnns System, Sys Bk 1, 03-15-05, Rev 4, pgs 14 & 20 K/A 239001 K1.12 (2.5/2.6), 2.01 (3.2/3.3), K4.01 (3.8/3.8),
K6.01 (3.1/3.3) 239001K601 239001K401 239001K201 239001K112 ..(KA's)
(***** CATEGORY 2 CONTINsiD s 1 M (T PAGE *****)
2t__EL88I_DESIGU_It4GLUDING_ 6EEll_eND.EDEBGENCX Page 35 111IEd3 ANSWER 2.03 (2.00)
The logic (Low-Low Set) is armed by actuation of any SRV (0.5) and i e high reactor pressure scram signal (0.5), the logic will lower the opening and closing setpoints of two SRVs (0.5). The valve blowdown is also increased (0.5)
REFERENCE Main Steam System and Bypass System pg 10 K/A 239002 K4.01 (3.9/4.0), K4.02 (3.4/3.6) 239002K402 239002K401 ..(KA's)
ANSWER 2.04 (2.00)
- a. The turbine will speed up and trip on overspeed (0.5) because the governor valve will fail open (0.5).
- b. The turbine will not start (0.5) because the governor and stop valves require oil pressure to open (0.5) which is normally supplied by the auxiliary oil pump.
REFERENCE RCIC, Figures 3 and 7 HPCI, Figures 4 and 7 K/A 217000 A2.07 (3.1/3.1), 206000 K4.14 (3.4/3.4),
206000 K5.05 (3.3/3.3) 206000K505 206000K414 217000A207 ..(KA's)
(ses** CATEGORY 2 CONTINUED ON NEXT P/,3E *****)
2i__EleUI_DESIQU_INGLUDIND_SeEEIl_eUQ_EbEBGEUG1 Pcg'o 36 31SIEd1 ANSWER 2.05 (2.50)
- a. Services the flow control valves, scram valves and scram dischargo 4 volume vent and drain valves.
- b. The CR0 pumps are cooled by the Reactor Building Closed Cooling water system,
- c. The CRD hydraulic system supplies recirculating pump seal purge '
water.
- d. The RPS provides signals to energize or de-energize scram pilot and scram valves and backup scram valves (to insert rods on a scram.) ,
- e. The Reactor Manual Control System provides signals to the hydraulic control unit, to position directional control valves to control rod motion.
(5 at 0.50 each = 2.5)
REFERENCE CROH System pg 23 and 24 K/A 201001 K1.03 (3.1/3.1), K1.06 (2.8/2.8), K1.07 (3.4/3.4)
K1.08 (3.4/3.4), K1.09 (3.1/3.2) 20a001K109 201001K108 201001K107 201001K106 201001K103
..(KA's)
( W T W t M.G W A) - 4 M "fCCs12646 Y ANSWER 2.06 (2.50) (hTVE (4t's #*f4 M d)- 400 ATC4, ncN se ). '
- a. Normal - RPS M-G sets (0.5)
Alternate - 480V MCC C'" m w ? -a te er ere
y (0.5)
- b. 1. Pr ocet s Radiation Monitoring System (0.5)
- 2. Power Range Monitoring System (0,5)
- 3. Nuclear Steam Supply Shutorf System (0.5)
REFERENCE RPS pg S, 28 K/A 212000 A2.02 (3.7/3.9) 212000A202 ..(KA's) (hyy1 Q/ICat.
gM ');gt 4//M int / ,,hh
- 3(. ** 33,1)
(b148{A/'SANt'!**HAEN'kl l$h
- g&V/}c hist. MiAnf 1,25 44 Post')
(***** CATEGORY 2 CONTINUE 0 ON NEXT PAGE *****)
F i
i s
2i__EL8NI_ DESIGN INCLUDING _38EEIl_8ND_EdEBGENCY Pago 37 111IEdi ,
t 1 i g j II ANSWER 2.07 (2.00)
- ddded. y. i will: b , .6, g wk11not: a, c, d Al,f , h' p. *
(0.25 each = 2.0)
/
REFERENCE i
SBGT system pg 14 .
K/A 261000 K401 (3.7/3.8) 261000K401 ..(KA's)
ANSWER 2.08 (2.00)
The system discharge boron inj ection is limited such that the rate of increase in the concentration of natural boron in the primary coolant water is fast enough to ensure a negative reactivity insertion rate, greater than positive react ivity addition rate due to plant cooldown (1.0), yet slow enough to ensure sufficient mixing so boron does not recirculate through the reactor core in uneven concentrations (could cause power cyclin ).
! (1.0). (2.0)
O REFERENCE ,cAu33 /r3 pg SLC System pg 5 K/A 211000 K4.05 (3.4/3.6), A1.07 (4.3/4.4) 211000A107 211000K405 ..(KA's)
ANSWER 2.09 (2.00)
- a. 1. To prevent local fuel damage that may result from a single rod j withdrawal error. (0.5)
- 2. Provides a signal used by the operator to evaluate the change in local relative power level during control rod movement. LO.5)
- b. The local power may be lower than the core average power. (1.0) l t
i
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
t
s, \
i
. t 2i__EL8SI_ DES 106_INCLUQ1NQ_S8EEIX_86D_EdEBGENCY Pcge 38
.SYSIEBS 8
h REFERENCE s
RBM pg 4 and 7 i : ,
K/A 215002 K1.02 (3.2/3.1),.G004 (3.3/3.4) 215002G004 215002K102 ..(KA's) l ANSWER 2.10 (2.50) -
41e/-#
- a. Reactor Pressure less than 441^psig >
(0.5)
- b. Prevent over pressurization of the low pressure piping upstream of the injection valves. "(0.5)
- c. F015A/B: . Remove initiation signal (0.5)
F017A/B: 5 minutes after initiation signal is received (0.5)
- d. F017A/B can be throttled to control injection flow (0.5)
REFERENCE RHR pg 32 K/A 203000 K4.02 (3.3/3.4), K4.10 (3.9/4.1) 203000K410 203000K402' ..(KA's)
ANSWER 2.11 (2.00)
- a. 1. Low reactor water level (level 2)
- 2. High temperature at outlet of the non-regenerative heat exchanger (140 deg F)
- 3. SLC initiation
- 4. High ambient temperature (>183 F) 5 .' High differential monitoring temperature (>53 F)
- 6. High differential flow comp ar ison (>63.4 gpm) and 60 second time !
delay (any 4 at 0.25 each a 1.0)
- b. Non-regenerative heat exchanger outlet temperature <130 deg F (0.5)
To prevent damage to the ion exchange resin (0.5)
REFERENLE RWCU pg 14 and 18 K/A 204000 K4.01 (2.5/2.5), K4.03 (2.9/2.9), K4.04 (3.5/3.6) 204000K404 204000K403 204000K401 ..(KA's)
(***** END OF CATEGORY 2 *****)
l
2i__1HSIBWUENI1_8ND_CONIBQL2 , Pcgo 39 i ,
ANSWER 3.01 (2.003'
- a. scram ,
- b. both
- c. neither
- d. rod block (4 at 0.5 each = 2.0)
REFERENCE ,
". RPS,,pg 6 RBM, pg 16 K/A 212000 K1.01 (3.7/3.9), K1.03 (3.4/3.6), K1.05 (3.3,3.6),
1, K1.14 (3.6/3.7) 212000K114 212000K105 212000K103 212000K101 ..(KA's) i i
ANSWER 3.02 (2.50)
(.2 f +/- i 7') p.
s s. (1) The LPSP is defined as the power level ^below'which the RWM program is enforcing adherence to the control rod movement as compared to the rod sequence (0,5) or similar. '
(pf+/.J.5 7d h (2) The LPAP is the power level above which all RWM blocks, alarms, j and error displays are discontinued (0.5) or similar.
- b. (1) 1 (0.5) l (2) 3 (0.5)
(3) Rod blocks will occur on all rods (0.25)except for the rod (s) required to correct the insert or withdraw errors (0.25).
REFERENCE Rod Worth Minimizer, pg 4, 15 K/A 201006 K4.01 (3.4/3.5), K4.02 (3.5/3.5), K4.06 (3.2/3.4),
K4.07 (3.1/3.2) 201006K407 201006K406 201006K402 201006K401 ..(KA's) i L
[
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
Ji__INSIBudENIS_8ND_CONIB0LS f Pogo 40 l
f ANSWER 3.03: (3.99)
- a. Causes reactor level to increase (0.33) due to the level control system having a level error, level set. indicated level (0.5) resulting in the feedwater control valves opening to match new higher level (0.5),
- b. Reactor level should rer!ain constant (0.333/becausethe"B"M/A
- transfer station will lock-up (0.5). The "A" feedwater control valve will control level (0.5).
- c. Causes reactor level to decrease (0.33) due to the level control system having a steam flow / feed flow error, steam flow < feed flow (0.5) resulting in the feedwater control valves to close to match new lower level (0.5).
REFERENCE Reactor Vessel Level Control System pg 12, 13, and figure 1 K/A 259001 K1.08 (3.2/3.2), K1.09 (2.0/3.0), K3.01 (3.8/3.8)
K3.02 (3.7/3.7) 259001K302 259001K301 259001K109 259001K108 ..(KA's)
ANSWER 3.04 (2.50)
- a. ADS valves remain as is (0.5)
- b. ADS valves close (0.5)
- c. ADS valves close (0.5)
- d. ADS valves open (0.5)
- e. ADS valves open (0.5)
REFERENCE Automatic Depressurization System, pg 2, A, 9, and figure 6 K/A 218000 K5.01 (3.3/3.8), K4.04 (3.5/3.t), A2.05 (3.4/3.6),
K6.02 (4.1/4.1), A1.05 (4.1/4.1) 218000K602 218000K501 218000K404 218000A205 218000A105
..(KA's)
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
l= ,i
, 2t__INSIBWdENIS_6HD_CQNIBQLS Pego 41 s
ANSWER 3.05 (2.50) 1
- e. Top of active fuel (0.50)
- b. Yes (0.50) ,
- c. 1. d '
- 2. c
- 3. e
, (3 at 0.5 each = 1.5)
'-REFERENCE FERMI RPV Process Inst.'umentation pg 6.
K/A 216000 K1.05 (3.7/3.9), K1.22 (3.6/3.8), K1.23 (3.3/3.4)
K5.01 (3.1/3.2), K4.14-(3.3/3.4) 216000K123 216000K105 216000K414 216000K501 216000K122
..(KA's)
ANSWER 3.06 (1.00) e REFERENCE SRM pg 17 K/A 291002 K1.22 (3.0/3.1), K1.19 (3.0/3.1), 215004 K5.01 (2.6/2.6) 715004K501 291002K119 291002K122 ..(KA's)
ANSWER '3.07 (1.50)
- a. Motor overload (0.25); 175 % (0.25)
- b. High frame oil t ernpe r at ur e (0.25) 150 deg F (0.25)
- c. Low lube oil pressure (0.25); 10 psig (0.25)
- d. Ground fault (0.50)
(any three = 1.50)
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
?
2A_ INSIBUt!ENIl_8ND_CQNIBQLS , Pcge 42 REFERENCE Compressed Air $> stem pg 7 K/A 295019G011 (3.9/4.1), G009 (3.4/3.4) 295019G009 295019G001 ..(KA's)
O'O Y' ANSWER 3.08 (' aa?
a) NO (0.5) b) YES (0.5)
- ) ""
/t$gfd y (0.5) d) NO (0.5) e) NO (0.5) s f) YES (0.5)
"EFERENCE
- Power Range Monitor (PRM) pg 13 FERMI Tech Specs pg 3/4 3-5 K/A 215005 K101 (4.0/4.0), K4.01 f.3.7/3.7), K4.02 (4.1/4.2) 215005K402 215005K401 215005K101 ..(KA's)
ANSWER 3.09 (1.50)
- a. Opens the scram discharge volume vent and drain valves (0.5)
- b. Shutdown (0.25) or Refuel (0.25) (0.5)
- c. Rod Withdrawal Block (0.5)
REFERENCE CR0 Hydraulics, pg 12 K/A 201001 A1.05 (3.3/3.4), K4.06 (3.8/3.9) 201001K406 201001A105 ..(KA's)
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
2i__INSIBUDENIS_eND_CONIBQLS Pago 43 ANSWER 3.10 )
7
^
- a. _tect af th: i.dicidu:1 ;; ; 1;-t : 111- [0.25) *0t !0 25)' (" t)-sg
- 1 Limiter <30% (0.25)
- 2 Limiter < P (2% (0.25) P" Wj lim N A 4 487.
- b. Limiter #1: Recirc. pump discharge valve (0.25) not fully open (g( .75 Q)b of.and$feedwater flow [0.25) <20% [0.25) (1.0)
Limiter #2: "A" or "B" Reactor feed pumps (0.25) not running rated flow (0.25) and a reactor vessel (0.25) level less than level 4 (192.5") (0.25) is received. (1.0)
P.EFERENCE Recirculation Flow Control System pg 10-12',A*4 7DC YIlid $~
K/A 202002 K4.02 (3.1/3.2), K4.07 (2.8/2.9) 202002K407 202002K402 ..(KA's)
ANSWER 3.11 (1.50)
- 1. Power less than 35%
- 2. Neither all A sequence control rods or all B sequence control rods are fully out, and
- 3. Power is greater than 30% (or the Reactor Mode Switch is in the REFUEL position)
(3 at 0.5 each a 1.5)
REFERENCE Rod Sequence Contrcl System pg 8 K/A 201004 G008 (3.7/3.4) 201004G008 ..(KA's)
(*a*** END OF CATEGORY 3 *****)
y
St__EB0GEDUBES_:_UQBdelt_eBUQBdelt_EUEBQEUG1 Page 44 AUQ_B6Q1Q(QQ1Q6L_QQUIBQL
.t.
yl <f / @ lo.tr], A AR .rLJ M ' a & Co.L k 4, by [0M]fAAA/M2 % '
ANSWER
- s. h 4 a h u 4.01(A(1.00)Of ft ik .run 9 c A su wh
) ID' k-3el k I & M AM Q Ndf y M Y],
- 1. Preventing temperature stratification in the R(a c ro r Vessel. f (0.5)
- 2. Retaining solids in suspension until they can be removed (by the RWCU
{
system to prevent their deposition in the bottom of the Reactor Vessel of the CRD Mechanisms). (0.5)
REFERENCE SOP 23.138.01 Rev. 21, Reactor Recirculation System, og 7
'J/ A 2 0 2 0 01 K 1. 0 5 ( 3. 4 / 3. 4 ), K4.12 (3.2/3.5) 202001K412 202001K105 ..(KA's)
ANSWER 4.02 (1.00)
- 1. Scram the reactor at H11-P608RR by taking one operable APRM's Mode switch out of OPERATE position in Div. I and DIV. II (0.5)
- 2. Trip the main turbine at H11-P632 by removing the relay cover from TTR1 or TTR2 and then push back on the relay trip coil bar at the top until the trip flag falls. (0,5)
REFERENCE AOP 20.000.19, Shutdown From Outside The Control Room K/A 295016 G006 (4.1/4.1), AK2.02 (4.0/4.1) 295016K202 295016G006 ..(KA's)
ANSWER 4.03 (1.50)
- 1. Control Rod Overtravel annunicator 2 Loss of pocition indication, (past position 48), when fully withdrawn
- 3. Control Rod Drift annunicator REFERENCE AOP 20.106.04, Uncoupled CRD I. OhW 4 KON hfd h K/A 201003 G015 (3.8/3.9) A I M M A t:
- h (A/J) 201003G015 ..(KA's) g [d.t r)
%up%Jo/A N p A wa % w m M 7 cs.,.Co.a.rJ.(
[4n 34 N A0n% /b1 i &%tW ),
(***** CATEGORY 4 CONTINUE ON NEXT FAGE *****
it__EBOCEDVBEl_:_UQBdeLA_eBSDBd8Lt_EdEBGENCY Pego 45 AND_B8019LQG1ceL_C961BQL
/'
ANSWER 4.04 (1.00)
- 1. Nisoperation in automatic was init iat e d
- 2. Adequate core cooling is! assured
/ (2 at 0 . 5 r. 1.0)
REFERENCE !
SOP 23.202, HPCI System pg 13 K/A 206000 A2.17 (3.9.'4.3) 206000A217 ..(KA's) .
ANSWER 4.05 (2.00)
- a. 3 i b, 4
- c. 1
- d. none !
- (4 at 0.5 each = 2.0)
REFERENCE !
E0P 29.000.01 pg 1, 29.000.03 pg 1, 29.000.04 pg i <
K/A 295024 G011 (4.3/4.5), 295031 G011 (4.2/4.6), 295036 0011 (3.8/4.1) l 2950360011 295031G011 2950240011 ..(KA's) ;
p I
! ' ANSWER 4.06 (2.00) t 1 1 1 Running back the recirculation pumps prior to tripping them minimizes the {
l heat load added to the suppression pool (1.0). If the pumps were tripped l
, at higher speeds it may cause a severe enough transient to trip the main t turbine and lift the safety / relief valves which would add heat to the torus !
(1.0). (2.00) l l
\ l 4 ,
i
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) r i
? E z
_t*
I
- > j.
at _EBQGEDWBEf_:_9f1Bdelt_etWDBd6Lt EdEBGENCY Pcge 46 AND_B6019LQG198L_CDUIBQL -
i
/ -
, l R2FERENCE /j s
E0P 29.000.08 rev 2, pt; 2 4 0C&P Course: ATWAS Study Guide Section X ,
K/A 294001 K1.09l/(3.4/3.8) ,
294001K109; ..(KA's) l i
ANSWER 4.07 (1.50) /
- s. True , r
- b. False .
t
- c. False REFERENCE Admin Proc.- 21.000.01, Conduct of Shift Operations, pg 12 K/A 294001 A1.03 (2.7/3.7) 294001A103 ..(KA's) l I
ANSWER 4.08 (2.00) '
- s. 3 '
i
- b. 1 t l c. 2 l
l l l
- d. 4 (4 at 0.5 each = 2.0) i REFERENCE 1
Admin Proc 12.000.012 rev 15, Tagging and Protective Barrier System, pg7 !
K/A 294001 K1.02 (3.9/4.3) '
.294001K102 ..(KA's) I i
f f
r T
l
(* esse CATEGORY 4 CONTINUED ON NEXT PAGE *****)
~
it__EBOCEDUBES_:_NDBd6Lt_6BUCS56Lt_EdEBGENCY Pago 47 aND_B6D10 LOGIC 8L_CQUIB0L ANSWER 4.09 (2.00)
.. Time
- 2. Rod Sequence, R'od Group, Rod, and Rod Position
- 3. Reactor Coolant Temperature (as indicated by RWCU inlet or Reactor Recirc. loop temperature)
- 4. Reactor period (4 at 0.3 cach = 2.0)
REFERENCE ~
GOP 22.000.03 rev 15, Startup From Cold shutdown To Rated Power pg 1 K/A 294001 A1.06 (3.4/3.6) 294001A106 ..(KA's)
ANSWER 4.10 (2.00) 5, 2, 4, 1, 3 (0.4 each, subtract 0.4 for each one out of order up to ths '61ue 2.0)
REFERENCE GOP 22.000.10 rev 6, pg 8', 10, 11 K/A 294001 A1.13 (4.5/4.3) 294001A113 ..(KA's) r ANSWER 4.11 (2,00)
J
- 1. Announce the event over the Hi-comm.
- 2. Sound the "klent Area Alarm".
- 3. Notify Health Physics.
- 4. Notify $ecurity.
r'-+ n t l
..rh = ?.01
- s. ne+f9 asa SLp Lpwr can ) ;
t A,, 4 y st 1%s k Mk M wr#d e.rof&.
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
! i
.i t _ _ t B O C E D tiB E 1_: _ N O B d 6 L & _8B U R B d e l t _ E U E B G E N C Y Page 48
- C-
.6SD_86P19LQQ106L.CQUIB9L
- x. ;
- w. ,
, f..
REFERENCE
.AOP 20.710.01. Refueling Floor High Radiation rev 5, og i K/A'295023 G010 (3.S/3.5) 295023G010. ..(KA's) ,
4 ;
l ANSWER 4.'12 (2.00) !
e !
a.- 1
- b. . 3 ,
- c. 2' ;
- d. .4' ;
(0.5 each)
REFERENCE t
t EP'102,-103, 104, 105 '
K/A 294001'A1.16 (2.9/4.7) '
294001A116 ..(KA'c) !
t l
L i
i i
l j
4 .
I i I
f i
t
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
r
m
$t__tB0CEDUBES_=_NQBd6Lt_ ABUQBd6L t_EdEB9ENCY Pcgo 49 AUD_B6019LQQ1G6L.CQNIBQL i ANSWER 4.13 (2.50) l
- 1. All secondary containment penetrations required to be closed dur'ing accident conditions are either:
- a. Capable of being closed by en operable containment automatic isolation system, or
- b. Closed by at least one manual valve, blind f16nge, or deactivated automatic damper secured in its closed position, Coxcept as provided in Table 3.6.5.2-1 of Spec 3.6.5.2)
- 2. All secondary containment hatches and blowout panels are closed and sealed
- 3. The standby gas treatment system is ope r able (or in compliance with the requirements of Spec 3.6.5.3)
- 4. At least one door 1.1 each access to the secondary containment is closed.
- 5. The sealing mechanism associated with each secondary contdinment penetration 41 s operable.
(t.es ., siildt, be)/e*Jto er 0*n'18b
- 6. The prissure within the secondary containment is less than or equal to the value 0.125 inch of vacuum water gauge (required by Spec 4.6.5.1.a)
(any 5 at 0.5 each = 2.5)
REFERENCE Tech Spec. 1.36 pg 1-6 K/A 290001 G011 (3.3/4.2) 290001G011 ..(KA's)
ANSWER 4.14 (2.00)
- 1. False
- 2. False
- 3. False
- 4. False (4 at 0.5 each = 2.0)
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
_ _ - - - -- - - - - - - - _------- _ --- -A
r it__EBQGEDUBE1.:_NQB06Li 6BUQBd6Lt EUEBGENCY -Pcgo 50 6ND_B6010LQGIG6L_CONIBQL REFERENCE Admin Proc 12.000.013 rev 8, pg 7 ,
K/A 294001 K1.03 (3.3/3.8) 294001K103 ..(KA's) '
W 4
4 1
d a
W (ese** END OF CATEGORY 4 *****)
( ::::: *** END OF EXAMINATION **********)
H/L 7 2OI.5 . .
A 4 192.5 LEVEL A
(2MA) RRS RUN9ACK REWN C Rot
- LARM a
/t,% -
LEVEL m
u=ir 1I LEVEL / FLOW 3
atw,c 1 f MODE sette-s wTCw LOGIC l 3
3f SULCV LEVEL l SELECT SWITCH COMPARATOR C0=T*0ttte O
LEVEL- N 1III COMPARATOR m gp g MASTER RwM LPSP 30*/. lI cot (TROLLER STEAM FLOW ALAP.!A O
LOSS OFHD E/P UNIT 73*/,
DUAL LIMITER E
h STEAM / ,3 pay
-> FIOw A O COMPARATOR d3
--> M/A E > ST AM M/A . STATION C > FLOW STATION hG
~
SULCV STEAM FLOW FUNCTION C FUNCTION GEN.
GEN. M-3 STM/ FEED x 3r SIGNAL STEAM FLOW I FLOW l I_.___l FAILURE o pcconocR [N.RFP g i iSRFP I TUR9INE g i TURBINE g g IEHC y I EHC g FEED FLOW FLOW L___J L _ _. _ J
- INTEG. '
l A E #Al RRS RUN9AOK 20*/.
UNIT FEED FLOW D
REF. DECO DWG.
61721-2126-1 REV. E 03-15-4 6 A I JC (von usn umi ODESMO!; 3.03) FIGURE I !
REV. 4
. o t
Pcg3 1 cf 5 Dele MEEI BEBCIQB IBEQBY EQBdWLOL's P = P ,t/v P=P O SUR(t)
O EI th p = --------------
SUR = 26.06/T
10 3.12 x 10 fissions /sec 8
1 2 1 elf p , ,-(B Lg 2) P = -- + ------
th 2 2 y 1+ v 1 + (B L th
-(B Lf 2) K~I P, =e A = -----
K Ap = In f!Ufl--
p , ,-CN3CI,,,3/tt, .
initial b e44 ~
U Cg (1-K,,gg) =C (1-K,jfp) T = ---- ---
2 1 C l' a . --- . tLoel-- , . --
~
initsal 1 of t op AL 2 2 2 , __i_ , 3 g__1b_3 ay = - -- + - -- - 0 f at p at at at e44 ~0 0 K,,f =( P, pP g fn Pg =P O
~~~~~~~~~
eff P1
- Pcg3 2 cf 5 ,
Dele _EOEEI IUEBdQQYNed1CE_8ND_ELUID_dECueWICS_EQBdWLGE:
0 = A ah , 2 w LaT Q = -----------------------
, 1 In R /R in R /R Q=UA (AT*) - + -------- + --- 3 2 3 ----
7 K K K 2 3
- p b=adAR4 6
n= !O:6 - 91 (n
n = - !0-- h991-)CE*l--
Q (h -h out ideal in in actual 11 22 W -
T T supplied 1 2
- A = PAV pAVg=PAV222 gg 3
A nc
= KA Q
E = KA AT Jal = KA op J oP ar m . eI_J101_:_eI_Jguti g,[i!!b___
AT tin) 9 '
a 8.Bx10 '
In (--------)
AT (out) f 2
Gr , kAoT l T -
T = --- G = -----
c1 ps 4g 3, ,
. t L
, tetal G = gg--- 3g-- oX
, a b n i
- - - + - - - + . . . + ---
K (
K K a b fa l
l l
i 1
. D8I6.5dEEI CENIBIEUG8L_EUdE_ LOWS Ng 4 (N )2 H (N g )
__ . __3 g
__g P
_____ . ___.._ . __3 N A H 2 2 (N2) 2 (N P 2 2 89DIGIlQN_3ND_CUEdISISY_EQBdWL85:
R/hr = 6CE/d 2 g , y ,-mx CV g g g=CY22 G = Dilyttgo_ Bate 1 = 1 C=C Volume 0 ill O' 10 A=A O ,- At A = AN CONVEBSIQN"*
1 gm/cm = 62.4 lbm/ft Density of water (20 C) = 62.4 lbm/ft 1 gal = 8.345 lbm 1 ft = 7.48 gal Avogadro's Number = 6.023 x 10 1 gal = 3.78 liters Heat of Vapor (H O) = 970 Btu /lbm 2
1 lbn. = 454 grams Heat of Fusion (ICE) = 144 Btu /lbm
~
e = 2.72 1 AMU = 1.66 x 10 grams u t' 3.14159 Mass of Neutron = 1.008665 AMU 1 kW = 738 ft-lbf/sec Mass of Proton = 1.007277 AMU 1 kW = 3413 Btu /hr Mass of Electron = 0.000549 AMU 1 HP = 550 ft-lbf/sec One atmosphere = 14.7 psia = 29.92 in. Hg 1 HP = .746 KW 'F = 9/5 'C + 32
- 1 HP = 2545 Btu /hr 'C = 5/9 (*F - 32) 1 Btu = 778 ft-lbf ' R = ' F + 460 1 NEV = 1.54 x 10 -6 Btu ' K = ' C + 273
~
h = 4.13 x 10 M-sec 10 1 W = 3.12 x 10 fissions /sec gg = 32.2 lbm-f t /l bf-sec 2 e = 931 MEV/ AMU 1 inch = 2.54 cm 0 C=3x 10 m/sec r = 0.1714 x 10 ~O Btu /hr ft R
4 . . . . . . . . . . . . . .Q
. Pcgo 4 of 5
. DOIO.5dEEI GVEB0GE_IUEBd8L CONDUCIIVIIL.lgt detecial E Cork 0.025 Fiber Insulating Board 0.028 Maple or Oak Wood 0.096 Building Brick O.4 Window Glass 0.45 Concrete 0.79 1% Carbon Steel 25.C) 1% Chrome Steel 35.00 Cluminum 110.00 Copper 223.00 Si l ver 235.00 Water (20 psia, 200 degrees F) 0.392 Steam (1000 psia, 550 degrees F) 0.046 Uranium Dioxide 1.15 Helium 0.135 Zircaloy . 10.0 MISCELLONEQUS_INEQBdGIl0N:
2 E = mc KE = 1/2 mv 2 PE = mgh V, =VO + at Geometric Object Area Volume Triangle A= 1/2 bh /////////////////
Square 2 A=S fjjjjjjjjjjjjjjjj Rectangle A=Lx W /////////////////
Circle A = nr 2 ffjjjjjjjjjjjjjjj Rectangular Solid A = 2(LxW + LxH + WxH) V=Lx Wx H Right* Circular Cylinder A= (2 nrz)h + 2(wr2) V = nr z h Sphere A = 4 nr 2 V = 4/3 (nr 3 2
Cube ///////////////////////////// V=S
5 coa 5 of 5 DOIe_SUEE.
UISCELL8NEQUS_INEQBdeI1ON_Jcoattowedt, 10 CFR 20 Appes. dix B Table 1 Table II Gamma Energy Col I Col II Col I Col !!
MEV per Air Water Air Water Material Half-Life Di si nt egr ati on uc/ml uc/ml uc/ml uc/ml 2x10 -6
~
Ar-41 1.84 h 1.3 Sub -----
4x10 ' ------
~3 1x10 ~0 5x10 ~3
~7 Co-60 5.27 y 2.5 S 3x10 1x10 1x10 ~IU
~
-D ~7 I-131 8.04 d 0.36 S 9x10 ' 6x10 3x10
-3 -7 Kr-83 10.72 y 0.04 Sub 1x10 -----
3x10 ------
~7 ~3 ~ ~
Ni-65 2.52 h 0.59 S 9x10 4x10 3x10 " 1x10 "
~A2 ~
~34 -6 Pu-239 2.41x10" y 0.008 S 2x10 1x10 " 6x10 5x10
~
~31 ~#
Sr-90 29 y ----- S 1x10 1x10'O 3x10 3x10
-6 ~#
Xe-135 9.09 h 0.25 Sub 4x10 -----
1x10 ------
Any single radionuclide with T > 2 hr -6 which does not decay by alpha dr 3x10" 9x10 -5 1x10 -10 3x10 l spontaneous fission 2
Neutron Energy (MEV) Neutrons per cm Average flux to duliver equivalent to i rem 100 mrem in 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />
~
6 thermal 970x10 6 670 0.02 400x10 280 (neutrons) 6 30
- 0.5 43x10 ,
g------ t to 24x10 17 cm x sec Linear Absorption Coefficients p (cm"3)
Energy (MEV) Water Concrete Iron Lead 0.5 0.090 0.21 0.63 1. 7 1.0 0.067 0.15 0.44 0.77 1.5 0.057 0.13 0.40 0.57 2.0 0.048 0.11 0.33 0.51 2.5 0.042 0.097 0.31 0.49 3.0 0.038 0.008 0.30 0.47
.'6 w e.w$ w' a v nw e, 4 *
- n. a .
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$41 Daws (vet Vepo'
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Is*
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gl 64c t its 4 l 741 ) 4 4.H43 64 ?4 64 4:
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$1P t CN: . Ps) I?H 177 4 4 6t = ti e 44 9 C h2 ? vbH I?;l 044 Ull Pik; $ t .ed.e h! )?
44 4' 6 4) b We } } )t t g?wl 44 :6 4,H t ?ll 1
$2 )t I4) H 96. 4 11)*4 C ?th hat 8:4' 048 l
% ~. a. e, - a -
'*d
e n.7.s ...s.. ... .n m.. w.g w+ wr. w 9. . ,4.m. M, ,, m,4 m.
. . . , . .. . . .v..
.. r.
6
.5.u_ut;:n
.. . . . . . ,. .+
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+
a m n.gg.ee..-m,s . ..g. s.m..v.,,,.4.g.
r.e n z - . , ;m w.,g. -
r - -
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. . t et Table 1. Satureted Steam: Tempereture TaWe-consenesed hi hm somn telee tatver talw:
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