IR 05000410/1986034
| ML20215N494 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 10/20/1986 |
| From: | Crescenzo F, Keller R, Kister H, Lange D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20215N485 | List: |
| References | |
| 50-410-86-34OL, NUDOCS 8611060100 | |
| Download: ML20215N494 (105) | |
Text
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V. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO. 86-34 (OL)
FACILITY DOCKET NO. 50-410 CONSTRUCTION PERMIT NO. CPPR-112 LICENSEE: Niagara Mohawk Power Corporation 300 Erie Boulevard West Syracuse, New York 13202 FACILITY: Nine Mile Point 2 EXAMINATION DATES: Jul-2, 1986 i
CHIEF EXAMINER:
lo
\\% 8 G
'FrankC/egenzo Q
Date
'
/d[/yP4 REVIEWED BY:.
eed j,
Davit Lange, Le actor Edjineer (Examiner) Datd
/
REVIEWED BY:
/0//V /$4 Robert' M. Keller, Chief, Projects Section 1C Date'
[b OM APPROVED BY:
Harr/ B. Kistar, Chief Date/
/
Projects Branch No. 1 SUMMARY:
License examinations were administered to five Reactor Operator candidates and nine Senior Reactor Operator candidates at Nine Mile Point Unit 2 during the week of July 22, 1986. Three Senior Reactor Operator candidates failed the written examination and one Senior Reactor Operator
,
candidate failed the simulator and oral examination. The grading of the l
written examinations revealed a generally marginal performance in those examination sections regarding plant procedures. No significant generic weaknesses were noted during grading of the oral examinations.
OFFICIAL RECORD COPY OL NM 2 EXAM RPT - 0003.0.0 8611060100 RA1029 PDR ADOCK 05000410 V
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REPORT DETAILS TYPE OF EXAMS:
Initial X
EXAM RESULTS:
l R0 l
SR0 l
l Pass / Fail l
Pass / Fail l
I I
I I
I I
I l Written Exam-l 3/0
5/3 l
l l
l l
I I
I l Oral Exam l
5/0
7/1 l
l l
I I
I I
I l Simulator Examl 5/0 l
7/1 l
l
I I
I I
I I
l0verall I
5/0 l
5/4 l
1 I
I 1.
CHIEF EXAMINER AT SITE:
Frank Crescenzo 2.
OTHER EXAMINERS: David Lange A. H. Howe M. O. Bishop (EG&G)
1.
Summary of generic strengths or deficiencies noted on oral exams:
A.
The candidates demonstrated good usage of the abnormal operating procedures.
B.
The candidates did not routinely fill out surveillance or scram checklists during simulator scenarios.
This created problems for several of the candidates, and it was suggested that the facility emphasize use of the checklists during training in the future.
C.
It was noted that procedure E0P-SCT was not trained on during simu-lator training. During one scenario designed to test knowledge of this procedure, the SR0 candidate demonstrated he did not know that this procedure existed.
It was further noted that the procedure was not in the simulator control room and,. per discussion with the simulator operator, had not been trained on during simulator training sessions.
OFFICIAL RECORD COPY OL N!1 2 EXAM RPT
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2.
Summary of generic strengths or deficiencies noted from grading of written
'
exams:
A.
The class, in general, performed only marginally well in sections
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4 (R.0. Procedures), 6 (S.R.0. Plant design and Instrumentation),
and 7 (S.R.0. Procedures).
B.
The class performed poorly on the following specific questions:
Question No.
Topic Class Avg.
!
4.01 Administrative definitions 66.0%
4.02 CSO responsibilities 66.0%
4.10 Deviation from license 63.0"
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conditions 6.07 Flow biasing of 34.3%
Nuclear Instrumentation 2.11/6.08 Rod Worth Minimizer 41% R0/50% SRO 6.10 CSH Diesel and Service 65.6%
Water interlocks 7.03 Protective Action 57.5%
Recommendations 7.04 Level / Power Control 62.5%
procedure 7.09 Recirc Pump Interlocks 57.8%
7.11 DBA Hydrogen Recombiner 28.1%
procedure 8.01 Tech. Specs. regarding 58.3%
Service Water to EDG-2 8.03 Tech. Specs. regarding 68.5%
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Bypass Valves during startup 0FFICIAL RECORD COPY OL NM 2 EXAM RPT
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3.
Personnel present at exit interview:
NRC Personnel Frank Crescenzo, Reactor Engineer (Examiner)
David Lange, Lead Reactor Engineer (Examiner)
NRC Contractor Personnel M. Bishop (EG&G)
Facility Personnel K. Zollitsch, Superintendent of Training M. Dooley, Training Supervisor M. Jones, Operations Supervisor 4.
Summary of NRC comments made at exit interview:
A.
Items 1.A and B were discussed with facility personnel present.
B.
The facility was informed that preliminary results of simu-lator/walkthrough exams were favorable, C.
With a few exceptions, the simulator performed well.
D.
The training and operations staff were cooperative throughout the examination period.
5.
Summary of facility comments and commitments made at exit interview:
A.
The facility acknowledged the NRC comments noted above.
B.
The facility felt that the written examinations were.very comprehen-sive with Section 7 being " extremely difficult."
C.
The operating and simulator examinations were conducted in a competent and professional manner.
Attachments:
1.
Written Examination (s) and Answer Key (s) (SR0/R0)
2.
Significant Facility Comments and NRC Resolutions on written examinations made during Exam Review 3.
Facility Comments and NRC Resolutions on Written Examinations made after Exam Review.
_
0FFICIAL RECORD COPY OL NM 2 EXAM RPT
_
U. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO. 86-34 (OL)
FACILITY 00CKET NO. 50-410 CONSTRUCTION PERMIT NO. CPPR-112 LICENSEE: Niagara Mohawk Power Corporation 300 Erie Boulevard West Syracuse, New York 13202 FACILITY:
Nine Mile Point 2 EXAMINATION DATES: Jul-2, 1986 CHIEF EXAMINER:
la M 66
~
FrankCfe$enzo Q
Date
'
/d!/Y!P4 REVIEWED BY:
m e:
Davit Lange, Le M actor Edhineer (Examiner) Datv
/
REVIEWED BY:
)
/0//f//474 Robert' M. Keller, Chief, Projects Section 1C Date'
~
APPROVED BY:
[b OM Harry IF. Kist er, Chief Dat#
/
Projects Branch No. 1 SUMMARY:
License examinations were administered to five Reactor Operator candidates and nine Senior Reactor Operator candidates at Nine Mile Point Unit 2 during the week of July 22, 1986.
Three Senior Reactor Operator candidates failed the written examination and one Senior Reactor Operator candidate failed the simulator and oral examination.
The grading of the written examinations revealed a generally marginal performance in those examination sections regarding plant procedures.
No significant generic weaknesses were noted during grading of the oral examinations.
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.
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2.
REPORT DETAILS
.
TYPE OF EXAMS:
Initial X
EXAM RESULTS:
l RO l
SRO l
l Pass / Fail l
Pass / Fail l
l l
l l
l l
l l Written Exam I 3/0 l
5/3 l
l l
l l
I I
l l Oral Exam l
5/0 l
7/1 l
I I
I I
I l
I l Simulator Examl 5/0 l
7/1 l
l
I l
i I
I I
l0verall l
5/0 l
5/4 I
I I
I i
1.
CHIEF EXAMINER AT SITE:
Frank Crescenzo 2.
OTHER EXAMINERS: David Lange A. H. Howe M. O. Bishop (EG&G)
1.
Summary of generic strengths or deficiencies noted on oral exams:
A.
The candidates demonstrated good usage of the abnormal operating procedures.
B.
The candidates did not routinely fill out surveillance or scram checklists during simulu+.or scenarios.
This created problems for several of the candidates, and it was suggested that the facility emphasize use of the checa,ists during training in the future.
C.
It was noted that procedure E0P-SCT was not trained on during simu-lator training.
During one scenario designed to test knowledge of this procedure, the SRO candidate demonstrated he did not know that this procedure existed.
It was further noted that the procedure was not in the simulator control room and, per discussion with the simulator operator, had not been trained on during simulator training session.
Summary of generic strengths or deficiencies noted from grading of written exams:
' A.
The class, in general, performed only marginally well in sections 4 (R.0. Procedures), 6 (S.R.0. Plant design and Instrumentation),
and 7 (S.R.0. Procedures).
B.
The class performed poorly on the following specific questions:
.
Question No.
Topic Class Avg.
4.01-Administrative definitions 66.0%
4.02 CS0 responsibilities 66.0%
4.10 Deviation from license 63.0%
conditions 6.07 Flow biasing of 34.3%
Nuclear Instrumentation
.
2.11/6.08 Rod Worth Minimizer 41% R0/50% SR0 6.10 CSH Diesel and Service 65.6%
Water interlocks 7.03 Protective Action 57.5%
Recommendations 7.04 Level / Power Control 62.5%
procedure 7.09 Recirc Pump Interlocks 57.8%
7.11 DBA Hydrogen Recombiner 28.1%
procedure 8.01 Tech. Specs. regarding 58.3%
Service Water to EDG-2 8.03 Tech. Specs. regarding 68.5%
inoperable Main Turbine Bypass Valves during startup I
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j 3.
Personnel present at exit interview:
NRC Personnel Frank Cre'scenzo, Reactor Engineer (Examiner)
i David Lange, Lead Reactor Engineer (Examiner)
NRC Contractor Personnel M. Bishop (EG&G)
Facility Personnel
'
K. Zollitsch, Superintendent of Training M. Dooley, Training Supervisor
'
M. Jones, Operations Supervisor 4.
Summary of NRC comments made at exit interview:
A.
Items 1.A and B were discussed with facility personnel present.
j B.
The facility was informed that preliminary results of simu-
-
lator/walkthrough exams were favorable,
,
C.
With a few exceptions, the simulator performed well.
j D.
The training and operations staff were cooperative throughout the i
examination period.
5.
Summary of facility comments and commitments made at. exit interview:
)
A.
The facility acknowledged the NRC comments noted above.
I B.
The facility felt that the written examinations were very comprehen-sive with Section 7 being " extremely difficult."
C.
The operating and simulator examinations were conducted in a competent and professional manner.
Attachments:
1.
Written Examination (s) and Answer Key (s) (SRO/RO)
2.
Significant Facility Comments and NRC Resolutions on written examinations made during Exam Review 3.
Facility Comments and NRC Resolutions on Written Examinations made after Exam Review.
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Facility Comments and NRC Resolutions on Written Examination made during Exam Review
NOTE: The following represent facility comments made during the examination review which resulted in significant changes to the examination answer keys.
Question
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No.
2.0.7.a.
COMMENT:
"This also allows a plant cooldown using the bypass valves to depressurize per OP-101-C, page 5 and 6."
RESOLUTION: Comment accepted. Answer key changed to reflect additional correct answer.
l 2.0.7.c.
COMMENT:
"This is not the only unique feature of an MSIV isolation (i.e. valves are hydraulically operated, a reactor scram is a direct result when in run on mode switch, only group at level one, only group affected by mode switch).
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RESOLUTION: Comment accepted. Alternate answers will be considered during grading.
3.0.4.
COMMENT: " Color coding is same for both AC and DC, which is implied
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and should not be required for full credit".
l RESOLUTION: Comment accepted. Discussion of color coding for current
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type will be deleted from the answer key.
5.0.6.
COMMENT: "First part of question is never answered in exam.
Key, should be "No".
{
1) Pressure remaining high is false.
!
RESOLUTION:
"No" is implied by text of answer; however, the key has been changed to reflect this explicitly. Comment regarding pressure is accepted, key changed to reflect this."
6.0.6 COMMENT:
"Part a. Note, if transmitter power is lost, the recorder will fail downscale, if lose recorder power, it fails as is.
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Part b. Since part (a) of the question talked about
" instrumentation", the responses to part (b) may indicate that if power still exists to the transmitter, ECCS will initiate if level drops to the initiation setpoint. This should be acceptable as opposed to key answers."
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RESOLUTION:
Comment will be considered during grading.
Full credit will be awarded based on candidate's assumption of exact nature of power failure.
6.10.a.
COMMENT:
" Answer is misleading, the question stated that " circuit fuse failure has initiated logic for service water to the CSH Diesel".
This implies that valve 94A has opened since this is the only occurrence that takes place when SW is initiated.
Recommend changing answer to " valve 94A opens".
RESOLUTION: Comment accepted. Answer key changed to reflect correct answer.
7.0.5.b.
COMMENT: " Question asked for basis of staying above minimum alter-nate flooding pressure, not necessarily the definition.
Per C7 bases p. 14 of 23 11.a. "As long as RPV pressure remains above the Min. Alternate Flooding Pressure, the core is adequately cooled irrespective of whether any water is being injected into the RPV".
To ensure adequate core cooling should be sufficient answer."
RESOLUTION: Comment accepted with exception that a more detailed discussion than " adequate core cooling" must be included for full credit.
8.0.2 COMMENT: Answer key has Remote Shutdown Panel L.C.0. of 7 days.
This specification was not supplied to examinees in their packet, so answer should be restricted to Spec 3.5.1 14-day LCO only.
RESOLUTION: Comment accepted.
Reference to R.S.P. deleted from answe __
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Facility Comments and NRC Resolutions on Written Examination made after Exam Review Question No.
(SR0 6.08) 2.11b.
COMMENT:
Should be either True or False.
Not all rod blocks in these groups are affected by mode switch position but some are. Statement was not clear as to whether all or some were affected.
Not Affected:
NMS - APRM inop.
RBM - All but downscale Ref: RMCS Ops Tech, Table 1
,
RESOLUTION:
Comment accepted.
Part "b" deleted from the
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answer key.
(R0 4.09) 6.11 COMMENT:
a.
RHS system and LPCS are cross connected in several ways that the key does not reflect:
1)
Jockey Pump, a single pressure holding pump supplies RHS loop A and LPCS.
2)
LPCS test return returns to the suppression' pool via MOV30A which is the RHS loop A suppression pool return.
Note: there are no procedural steps for the line up discussed in this question, only cautions. So operators are not required to have an in depth knowledge of this non-routine lineup.
If person answered using either of the cross connects above, part b makes no sense.
Ref: FSK 27-7 (RHS System)
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Question No.
l 6.11 RESOLUTION:
Facility comment to part "a" will be considered i
during grading, however, full credit must include
a discussion of the cross connect as described i)
in the answer key. No credit will be allowed for discussion of the jockey pump as a system cross connect. Part "b" will be deleted since this
,
mode is infrequently used and as such operators need not maintain familiarity with these cautions.
7.04 COMMENT:
The answer in key is one of three valid answers.
i Others are:
1)
Per E0P-C7, the normal control band is 159.3
)
to 202.3.
Per page 19 of 23, the definition
,
of Hot Shutdown baron weight includes level assumed to be at "high RPV water level trip i
(202.3")" so answer could be sufficient
,
baron has been injected to allow operator to return water level to normal band.
2)
RQ is being utilized simultaneously with C7.
'l Action steps in RQ taken to insert rods include " resetting Rx scram" (Step 15.1)
!
which is the basis for setting the lower i
band limit of 159.3" (p. 13 of 23, C7 bases, 8.c and RL bases, p. 11 of 15 6c).
So valid answer would be " sufficient boron injected
to raise water level to above 159.3" in t
order to reset the alarm".
RESOLUTION:
Comments will be considered during grading and partial credit awarded as is appropriate.
Full
,
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credit answer must include discussion of boron
stagnation as described in the answer key.
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MCl7/71C/7f/%
Enshs u n o MASTER COPY
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U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
NINE MILE POINT 2 REACTOR TYPE:
BWR-GES DATE ADMINISTERED: 86/07/21 EXAMINER:
BISHOP.M.
APPLICANT:
INSTRUCTIONS TO APPLICANT:
Use separate paper for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%.
Examination papers will be picked up six (6)
hours after the examination starts.
% OF CATEGORY
% OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 05. 3:M 25.00-25. 00p#
1.
PRINCIPLES OF NUCLEAR POWER 25 50 PLANT OPERATION, THERMODYNAMICS, r*
<*
HEAT TRANSFER AND FLUID FLOW 24 50 M
{ 5. 00 *Q,b-25.00 2.
PLANT DESIGN INCLUDING SAFETY ce <*
AND EMERGENCY SYSTEMS 24 hhp 2f5#
25.00-2 5. 00#'
3.
INSTRUMENTS AND CONTROLS-t+e?$ 9 z4 0cro-25.00 5 45.00 4.
PROCEDURES - NORMAL, ABNORMAL, 2 5.50 /s EMERGENCY AND RADIOLOGICAL CONTROL 98.0D f*
4L 73 **
4 00.00 100.00 TOTALS FINAL GRADE
%
All work done on this examination is my own. I have neither civ n nor received aid.
APPLICANT'S SIGNATURE N * N h~ #!h Lt
<
r--
e
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this cxamination the following rules apply:
.
1.
Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
2.4 Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
~
3.
Use' black ink or dark pencil og to facilitate legible reproductions.
4.
Print your name in the blank provided on the cover sheet of the examination.
5.
Fill in the date on the cover sheet of the examination (if necessary).
6.
Use only the paper provided for answers.
7.
Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
8.
Consecutively number each answer sheet, write "End of Category
" as appropriate, start each category on a new page, write jon1 o g side of the paper, and write "Last Page" on th 7 east answer sheet.
9.
Number each answer as to category and nLmber, for example,1.4, 6.3.
10. Skip at least three lines between each answer.
.
11. Separate answer sheets free pad and place finished answer sheets face down on your desk or table.
12. Use abbreviations only if they are commonly used in facility literature.
13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
14. Show al) calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND 00 NOT LEAVE ANY ANSWER BLANK.
16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done af ter the examination has been completed.
18. When you complete your examination, you shall:
a.
Assemble your examination as follows:
(1) Exam questions on top.
(2) Exam aids - figures, tables, etc.
(3) Answer pages including figures which are a part of the answer.
b.
Turn in your copy of the examination and all pages used to answer the examination questions, Turn in all scrap paper and the balance of the paper that you did c.
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.
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1.
PRINCIPLES OF NUCrJAR POWER PLANT OPERATION.
PAGE
TWRMODYNAMICS. MAT TRANSFER AND FLUID FLOW
,
QUESTION 1.01 (1.50)
While withdrawing control rods to take the reactor critical, does the startup nuclear instrumentation require the same time to stabilize at each suberitical level?
Briefly explain your answer.
'
QUESTION 1.02 (2.00)
If the reactor power level is increased on a positive period from 50 MW to 370 MW in two minutes, what is the doubling time?
(Show cll work)
QUESTION 1.03 (3.00)
Assume that the reactor is being started up with the bulk coolant temperature less than the saturation temperature.
Excessive rod withdrawal causes the reactor to increase in power on a short period.
Of the void, doppler, and moderator temperature coefficients, which will come into effect first, second, and third?
BRIEFLY EXPLAIN!
(Assume no operator action to stop the power increase.)
QUESTION 1.04 (3.00)
With the reactor operating at 50% power and 50% rated flow, the flow is increased to 70% of rated.
a. Briefly explain the effect on power level. Include Void and Doppler effects and the effect on boiling boundry.
(1.5)
b. Briefly explain how the turbine EHC system reacts to this change.
(0.5)
c. Briefly explain WHY the generator load increases as a result of this change.
(1.0)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
PAGE
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW
.
QUESTION 1.05 (2.00)
c.
Give two reasons how feedwater heating improves power plant efficiency.
b.
If the highest pressure feed heater is removed from service (extraction steam isolated), WHAT happens to Megawatt output of the generator and WHY?
Assume no operator action.
QUESTION 1.06 (2.00)
In an emergency, a technique called " Breaking Vacuum" can be used to stop the turbine faster than normal.
BRIEFLY EXPLAIN how this helps stop the turbine and what hazard (s) exist for the turbine during this evolution.
QUESTION 1.07 (2.00)
Give ONE undesirable result for each of the following.(Be more specific than " pump failure"):
A. Operating a centrifugal pump for extended periods of time with the discharge valve shut.
B. Starting a centrifugal pump with the discharge valve full open.
QUESTION 1.08 (2.50)
Refer to the attached figure 15.3-2 from the Nine Mile Point 2 FSAR concerning a trip of both recirculation pump motors, n.
Explain the sequence of events which lead to the pressure spike noted from time T=7 to T=10 secs.
(1.5)
b.
Explain the behavior of core inlet subcooling from time T=7 to T=20 secs.
(1.0)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
PAGE
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW
.
QUESTION 1.09 (1.50)
A centrifugal pump is operating at 3600 RPM with a pump head of 160 ft.
Pump speed is then reduced so that pump head is 100 ft.
WHAT is the new pump speed?
SHOW ALL WORK.
QUESTION 1.10 (1.00)
During a rapid power increase, very short periods can be maintained, yet for rapid pcwer decreases, the period quickly becomes -80 sec.
Explain the reason for the difference.
QUESTION 1.11 (2.50)
a.
WHY are installed neutron sources needed in the initial core
/. 6 (
loading?
(4-01" l
b.
Assume no administrative limitations prohibited reactor startup I
without the neutron sources.
Could the reactor achieve criticality without these sources?
If so, HOW would the critical rod density
/. o change relative to a startup with a source installed?
(4-6-) "*
QUESTION 1.12 (2.00)
The reactor is started up after a refueling outage.
Rods are pulled to the 100% line and power is then increased to 100% with recirculation flow.
After approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, reactor power has decreased to about 98%.
Assume no operator action.
[
a.
What is the primary cause for this reduction in power.
(1.0)
b. When would you expect the power decrease to stop and WHY does it stop?
(1.0)
(***** END OF CATEGORY 01 *****)
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2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
.
l QUESTION 2.01 (3.00)
Answer the following in regard to the Recire. Pumps:
a.
In Low Speed, WHAT is the MAJOR contributor to the NPSH required for the pumps?
(0.5)
b.
In High Speed, WHAT is the MAJOR contributor to the NPSH required for the pumps?
(0.5)
c.
WHAT are four interlocks that will cause an Automatic High-to-low Speed transfer?
(Downshift)
(2.0)
QUESTION 2.02 (2.00)
WHAT are the suction sources for CSH.
Include the normal and alternate suctions and parameters that will initiate an automatic shift from the normal suction.
(Setpoints required for full credit.)
QUESTION 2.03 (2.00)
Answer the following in regard to the ADS system:
a.
How many SRV's are dedicated to the ADS system?
(0.5)
b.
What is the purpose of the 105 sec. time delay?
(1.0)
c.
What RPV water level instrumentation RANGE supplies the low (159.3 in.) level signal to the logic?
(0.5)
QUESTION 2.04 (2.00)
The RCIC (ICS) water leg pump maintains the discharge piping full of water up to the discharge isolation valve (MOV-126).
What two things does this accomplish that enhance system operation when an initiation signal is received.
(***** CATEGORY 02' CONTINUED ON NEXT PAGE *****)
.
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
.
QUESTION 2.05 (1.00)
What D/G automatic shutdowns are operable with the Standby D/G's operating in the LOCA mode?
QUESTION 2.06 (1.50)
What effect does a LOCA signal have on Drywell Cooling Systems components?
QUESTION 2.07 (3.00)
e.
WHAT operational evolution is accomidated by having the Group I Isolation on low steam line pressure ONLY in affect when the mode switch is in the RUN position?
b.
What is unique about the Group I isolation logic (MSIV only)
ARRANGEMENT in respect to the other group isolations?
c.
What is unique about an MSIV Isolation in respect to other system isolations?
QUESTION 2.08 (2.00)
Answer the following in regard to the RPS.
o.
What TWO conditions will automatically trip the RPS EPA's?
b.
What is the power supply for the RPS MG sets 1A and 1B7 QUESTION 2.09 (3.00)
Answer the following in regard to the Recire Pumps.
A.
In the event of gross failure of both seals with the RPV at operating pressure, what limits the leakage rate to the drywell?
B.
What indication would you have of a No. 1 seal failure?
C.
Why do the Recire Pumps ALWAYS start in Fast Speed?
D.
What system provides cooling to the pump seals?
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
-
_
.
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
QUESTION 2.10 (3.00)
Dascribe the interlocks associated with the 4.1 KV emergency bus or switchgear which will ensure the emergency bus remains energized given the following malfunctions:
a.
The diesel generator is running in test, paralleled to the grid, when a LOCA signal occurs followed by a loss of offsite power.
(1.5)
b.
The diesel generator is running in test, paralleled to the grid, when a loss of offsite power occurs.
(1.5)
2.co QUESTION 2.11 (4,,40)#
Indicate whether the following statements regarding the RMCS are TRUE or FALSE:
n.
The currently latched RWM group (i) is that group with a rod withdrawn past its insert limit and with no greater than 2 B+/+!'
insert errors in RWM groups 1 to 1-1.
_
pJ was b.
The rod blocks imposed by the NMS (including the RBM) and the service and refuel platforms are' dependent on the Mode Switch position.
_
_
c.
A RWM rod block will occur if more than one control rod is withdrawn and the Rod Test pushbutton is depressed.
d.
A double X, (XX), indication on the four rod display, indicates that the RPIS is receiving abnormal data.
System hardware malfunctions in the RWM will not cause rod e.
blocks when above the LPSP.
(***** END OF CATEGORY 02 *****)
.
3.
INSTRUMENTS AND CONTROLS PAGE
.
QUESTION 3.01 (2.00)
Briefly explain how RBCLCW temperature is controlled.
Include the COMPONENTS, HOW they function, and the Temperature maintained.
QUESTION 3.02 (3.00)
Explain the the normal lineup, loading, and unloading sequence of the station air compressors A, B, and C. Assume compressor CIA is to be selected for LEAD and include any applicable setpoints and/or control switch positions.
QUESTION 3.03 (1.50)
Answer the following in regard to the Radiation Monitoring system (RMS).
A.
What TWO types of radiation detector are used in the ARM's?
B.
What type of radiation detector is used for the Main Steam Line Rad.
Monitors?
QUESTION 3.04 (2.00)
Plant emergency power is color coded for easy recognization.
What is the color code for each Division of both Essential AC and Essential DC?
QUESTION 3.05 (3.00)
Concerning the Electrohydraulic Control System (EHC):
What THREE parameters are sensed and evaluated by the control a.
system?
(1.5)
b.
At 100% power, what controlling circuit is actually positioning the control valves?
(SYSTEMS AT NORMAL OPERATION)
(0.5)
c.
With the main generator " synched" to the grid, what control circuit is effectively out of the control scheme, AND WHY?
(1.0)
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
.
3.
INSTRUMENTS AND CONTROLS PAGE
QUESTION 3.06 (3.00)
State the calibration conditions for the following reactor water level rcnges:
a.
Shutdown Range (2 items)
,
b.
Narrow Range (2 items)
c.
Wide Range (4 items)
d.
Fuel Zone Range (3 items)
QUESTION 3.07 (2.00)
Explain how all control rods can be determined to be at the FULL IN (e.g.,00) position using the RWM Rod Test / Select pushbutton on the RWM operating panel.
QUESTION 3.08 (2.00)
The Redundant Reactivity Control System (RRCS) initiates actions if the ARI function fails to reduce power following a Reactor Vessel High Dome Pressure signal (1050 psig).
WHAT three actions occur to insert negative reactivity, QUESTION 3.09 (3.00)
What will the final reactor level be (Higher than, Less than, or No change) for the following feedwater level control system failures? Explain Why.
a.
loss of 1 steam flow signal input (3 element control)
b.
loss of 1 feed flow signal input (3 element control)
c.
loss of the selected NR water level input (downscale)
l (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
_ _ - _ _ _ _
.
3.
INSTRUMENTS AND CONTROLS PAGE
.
QUESTION 3.10 (1.00)
What are the feedback signals to each of the following RRFC controller outputs:
a.
Output of the master controller b.
Output of the flux controller c.
Output of the loop controllers QUESTION 3.11 (1.50)
What are THREE parameters, and Associated Setpoints, that will cause a primary containment automatic isolation of Group 5 (RHR/ shutdown cooling)
to occur.
QUESTION 3.12 (1.00)
In regard to Load Shedding and Load Sequencing, what is the difference between the Division I switchgear, 2 ENS * SWG 101 and the Division III switchgear 2 ENS * SWGy400 tin.the event of a loss of power to the buses?
/01 >#
,
'***** END OF CATEGORY 03 *****)
.
4.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE
RADIOLOGICAL CONTROL QUESTION 4.01 (2.00)
Dafine the following terms:
A.
Shall B.
Should C.
May D.
Will QUESTION 4.02 (2.00)
According to Procedure AP-1.2, " Comp. and Resp. of Unit Org."
WHAT are four responsibilities of the Chief Shift Operator (CSO)?
QUESTION 4.03 (2.50)
What are the FIVE entry conditions for Emergency Operating Procedure, N2-EOP-RQ, RPV Reactivity Control?
(Setpoints required for full credit.)
QUESTION 4.04 (1.50)
WHAT operator actions are required if the seal oil system fails with the plant at 75% power.
(Three actions required.)
QUESTION 4.05 (3.'00)
Procedure N2-IOP-1010
" Plant Shutdown," notes that while in shutdown cooling with both reactor recirculation pumps off, it is possible to observe pressurization of the vessel or venting off of steam, a.
Explain WHY this condition might occur and WHAT indications are available to the operator to assist in recognition of this condition.
(1.0)
b.
Explain two methods by which an operator can prevent the above condition from occuring.
(2.0)
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
-
.
4.
PROCEDURES - NORMAL. ABNORMAL EMERGENCY AND PAGE
RADIOLOGICAL CONTROL QUESTION 4.06 (2.50)
A.
Per N2-OP-21, Main Turbine, WHAT are FOUR indications of water induction to the Main Turbine?
(2.0)
B.
If accelerating below rated speed and water induction is apparent, what are the immediate operator actions?
(0.5)
QUESTION 4.07 (2.00)
What are FOUR actions that should be taken if a control rod drift alarm is received while at power?
QUESTION 4.08 (1.50)
During startup of the 24 VDC Battery Chargers, the DC output breaker must be shut PRIOR to the AC input breaker.
HOW is this verified during the startup and WHAT damage could result if the AC input breaker is closed first?
/. So QUESTION 4.09 ( M )#
o.
Briefly explain HOW and WHY LPCS can be cross-connected with the RHS system.
(1.5)
I b.
While LPCS and RHS are cross-connected as described above, the operator is cautioned to:
1. Ensure LPCS suction valve MOV 112 is open.
.
2. Ensure LPCS suction isolation valve 2CSL*V121 remains shut.
Explain the reasons for each of these cautions.
(1.5)
(
--
J
-
jet b c/c 2kL (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
l
.
4.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE
RADIOLOGICAL CONTROL QUESTION 4.10 (2.00)
n.
Under WHAT conditions may a plant operator deviate from an approved procedure and depart from a license or Tech Spec condition?-
b.
Whose approval is required to do this?
QUESTION 4.11 (3.00)
Adequate core cooling is defined to be heat removal from the reactor sufficient to maintain fuel clad temperature < 2200 deg. F.
According to the EOP's, three viable mechanisms of adequate core cooling exist. List these three mechanisms IN ORDER OF PREFERENCE and include HOW adequate core cooling is verified for each mechanism.
.
l (***** END OF CATEGORY 04 *****)
(************* END OF EXAMINATION ***************)
-.
-
_ - _
__
-
e
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.
3 8vPAS$ V ALVF F LOW 4 UlHF int I T Silts 4 Der F USE H F LOW I (%)
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.
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l t i V L t 1.sh eet sep 56 n,p 2 VI SSE L STI AMf LOW J.
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,
3 TisFistiPJF S f t AMF t (?n y
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4 10T AL HE ACTIVITY
_
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SO O
20
40 SO IDMtIveb
. T tui isn t FIGURE 15 3-2 TRIP OF BOTH RECIRCULATION PUMP MOTORS NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
- EQUATICN SHEIT
.
Cycle efficiency = (Net * -
v = s/t f = ma cut)/(Energy in)
s = V,t - 1/2 at
, = 3g
.
_,
E = mc A*A*
A = AII s
4E = 1/2 mv a = (Vf - 1 )/t
,
PE = agn A= Ln2/tijg = 0.693/t1/2
= e/t y
y. at 8#f * b(*' )( ~*)]
t
1/2
.,
g(g /2),(; ))
.
- "
A=
$=V,yAo g, g, -h
- 9'
.
.
Q = mCast
[=Ie g
h = UA A.T I = I,10"* O pwr = w,ah TYL = 1.3/u sur(t)
HVL = -0.593/u p = p lo o-
? = 7,e j '
SG = S/(1 - Kaff)
SUR = 25.06/T G = S/(1 - K,ffx)
x G (1 - K,ff1) = G (I ~ eff 2)
j
SUR = 25s/t* * (5 - o)T T = ( t*/a ) + ((3 - s'/ Ta]
M = 1/(1 - K,ff) = CR)/G 3 M = (1 - K,ff,)/(1 - K,ff;)
T = t/(s - s)
SCM = ( - K,ff)/K,ff f
T = (3 - o)/(Is)
L' = 10 seconcs a = (X,ff-1)/K,ff =.X,ff/K,ff I = 0.1 seconds *I
= ((t*/(T K,ff)] + (s,ff (1 + IT)]
/
l1*Id2 =2 2 Id Id 7 = (:sv)/(3 x 1010)
Idj
2 A/hr = (0.5 CE)/c (,,g,73)
R/hr = 6 CE/c2 (f,,g)
- = 2N -
.
Miscellaneous 0:nversiens Watar Partneters
I curie = 3.7 x 10 egs I gal. = 3.345 lom.
1 '4g = 2.21 lem $ Stu/hr 1ga}.=3.78litars 1 no = 2.34 x 10 1 f.
= 7.48 gal.
1 mw = 3.41 x 100 5tu/hr
Oensity = 62.4 lerg/ft lin = 2.54 cm Censity = 1 gm/c9.
'F = 9/5'C + 32 Heat of vacoritation = 970 Stu/ Tem
- C = 5/9 ('F-32)
Heat of fusion = 144 Scu/lem 1 STU = 778 ft-lbf 1 Aca = 14.7 asi = 29.9 in. Hg.
-
1 ft. H O = 0.4335 luf/in.
-
_
_.
_
-
- - -
.
1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW
.
ANSWERS -- NINE MILE POINT 2-86/07/21-BISHOP,M.
ANSWER 1.01 (1.50)
No.(0.5) The fractional change in neutron population becomes less as you approach criticality (0.5) and the time to stabilize becomes longer.(0,5) (any equilivant answer acceptable)
REFERENCE NMP-2, BWR Academic Series, reactor kenitics, page 3-0 and 3-0.
ANSWER 1.02 (2.00)
P = Poe t/T 370 = 50e 120 sec/T
[0.5]
T = 59.95 sec
[0.5]
T Doubling Time = ----------- 1.445
[0.5]
DT = 41.49 sec
[0.5]
(2.0)
REFERENCE
NMP-2, BWR Academic Series, Reactor Kenitics, PP. 3-19 ANSWER 1.03 (3.00)
a.
First:
Doppler deals with fuel temperature, and this will be the first parameter to change.
There is a lag from the power generated in the pellet until the heat is transferred to the
,
coolant.
Thus, doppler adds negative reactivity to turn the power excursion.
(1.0)
b.
Second:
Moderator temperature coefficient begins adding negative reactivity as soon as sufficient heat is transferred to the coolant to raise coolant temperature.
(1.0)
,
c.
Third:
Void coefficient will have little or no effect until saturation temperature is reached.
(1.0)
REFERENCE NMP-2, BWR Academic Series, Coefficients of Reactivity, PP. 4-1 thru 4-59 f
_ -,
l
.
1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
PAGE
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW
.
i ANSWERS -- NINE MILE POINT 2-86/07/21-BISHOP M.
ANSWER 1.04 (3.00)
o.
Power increase as recirc flow is increased.
The increase in power is due to the addition of positive reactivity as voids are i
swept away by the increase in recirculation flow.
As power increases, fuel temperature increases, inserting negative reactivity due to
doppler broadening.
As this new heat is transferred to the coolant, more void formation occurs, thus more negative reactivity.
Therefore, boiling boundary is back te-ite orginal 1 :: tion; approximately.
(1.5)
b.
Power increasing causes pressure to increase?" The pressure increase is sensed by the EHC system which admits more steam to the turbine.
(0.5)
c.
The turbine generator will try to turn faster, due to increased steam flow, thereby picking up more electrical load since the speed is fixed by the grid grequency. (v,v/ a <c.f e,-y muma opk /J--
(1.0)
E ol' stec G esee)
y M
REFERENCE NMP-2, OPS Tech. Turb. EHC, P. 2, Rev. 2 NMP-2, Academic Series, Reactor Operational Physics, PP. 7-18 and 7-19 ANSWER 1.05 (2.00)
a.
The energy recovered in feed heating would otherwise be lost to the main condenser (0.5) and less heat is required from the reactor to reach the desired conditions. (0.5)
(1.0)
b.
Megawatt output from the generator would increase (0.5).
Steam that was formerly being extracted now passes through the turbine to the condenser (0.5).
(1.0)
REFERENCE GE Reactor Fund. vol. 3 & GE Turbine Handout NMP-2, BWR Academic Series, Thermodynamic Cycles, PP. 5-48
__
1.
PRINCIPLES OF NUCr.rAR POWER PLANT OPERATION.
PAGE
I_RERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW
.
ANSWERS -- NINE MILE POINT 2-86/07/21-BISHOP,M.
ANSWER 1.06 (2.00)
As air enters the turbine the blades are no longer turning in an atmosphere of ~1 psia, but are now in an atmosphere of 14.7 psia.
This increases the back pressure or drag on the turbine blading.
With the increase in pressure results in an increase in temp due to the windage effect on the turbine blading which can cause damage'to the turbine.
(2.0)
l REFERENCE Dresden Thermodynamics NMP-2, OPS. Tech. Condenser Air Removal and Off-Gas System, P. 6, Rev. 2 ANSWER 1.07 (2.00)
A. The pump will eventually add a sufficient a ount of heat to the fluid to cause cavitation. Also will accept, erheating-cf th;",,
"
(1.0)
. ;u.;. o r fg c.JM4.
7*
B. Could cause excessively long starting currents or water hammer if the downstream piping was not filled.
(Will accept either answer.)
(1.0)
REFERENCE GE THERMO HT & FF pg 7-123, 124 1&T-2, BWR Academic Series Heat Transfer & Fluid Flow, Fluid Statistic, Dynamics and Delivery, PP. 6-108 and 6-109 ANSWER 1.08 (2.50)
9.
Reactor level increases due to loss of recirc suction from annulus and increased voiding in core region.
Increase in level causes turbine trip which in turn creates pressure spike to SRV opening. (1.5)
b.
As reactor pressure increases, core inlet temperature remains essentially constant thus inlet subcooling increases and decreases as a function of reactor pressure.
(1.0)
REFERENCE NMP2 FSAR Chap 15 Vol 27
-
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_ - _,
,
_
_
_ - - _ -,
1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
PAGE
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW
,
ANSWERS -- NINE MILE POINT 2-86/07/21-BISHOP,M.
ANSWER 1.09 (1.50)
According to centrifugal pump laws:
,
Head ~ (Speed)
[0.5]
Therefore,
100 ft/160ft = (x/3600 RPM)
[0.5]
=
12.96 x 10 E6 RPM (0.625)
ix!
x
=
2846 RPM
[0.5]
REFERENCE MNS Thermodynamics, p.
12-6.
NMP-2, Fluid Statics, Dynamics and Delivery, P.
6-96 ANSWER 1.10 (1.00)
The period during power increases is governed by how quickly the neutron population can increase.(0.5)
The same holds true on a power decrease however, the neutron' population is dominated by the longest. lived delayed neutron precursor.(0.5) (This decays with -80 sec. period.)
(1.0)
a'
se REFERENCE Millstone Reactor Theory pg. 3.45 Pilgrim Reactor Theory pg 3.45 Nine Mile 2 Reactor Theory pg 3.45, L.O.
5.6 ANSWER 1.11 (2.50)
o.
Installed sources are used to raise flux levels in the core to a point where it is on scale for the nuclear instrumentation.4A 7 E ? "- 76-IuiLimi luLiiusic sourcc iccol arc : t high nse.L Lv brin. the t in:tiamulation ou svale. (0.75) e_,,
(1.5)
b.
Yes, (0.5) criticality would be achieved with no change in critical rod density. (0.5)
(1.0)
REFERENCE NMPC Operations Technology pg.
2-3, Theory L.O..
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
PAGE
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- NINE MILE POINT 2-86/07/21-BISHOP,M.
ANSWER 1.12 (2.00)
Xw. ~ b.;the H a.a.^.: the.coctor Operater at rnupy Yneen build in tc equilibrigglt WW
--[0.52 cdding nomaLive acovLivity, vousins Peacr to decimaou [6.511'II.0)
b.
IN 40-50 hours when equilibrium Xe is reached.
(1.0)
REFERENCE Dresden General Physics BWR RX Theory j
NMP-2 Academic Series, Poisons, PP. 6-6 thru 6-9
<
_ _ _
_
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE-19 ANSWERS -- NINE MILE POINT 2-86/07/21-BISHOP,M.
- ANSWER 2.01 (3.00)
o.
Maintaining the water level in RPV.
.(0.5)
b.
Subcooling affect of the incoming feedwater.
(0.5)
c.
Auto High-to= Low Speed Transfer 1. Delta T between steam dome and recire, loop suction <10.7F for 15 sec.
2.
FW flow <30% rated and FCV <19% open for 15 sec.
3. Vessel water level < level 3.
4. EOC-RPT signal present 5. RRCS high dome pressure signal present ( StefoA h u,( ftpi.J ) g (4 @ 0.5 ea)
(2.0)
REFERENCE NMP-2, Ops. Tech., Reactor Recire. System, PP. 3 & Lesson Plan pg 17.
ANSWER 2.02 (2.00)
Normally from CST-B [0.5].
Alternate suction from suppression pool [0.5] will be automatically initiated an CST Low Level [0.25],
12.5' [0.25] or SP high level [0.25], 201' [0.25].
(2.0)
REFERENCE NMP-2, Lesson Plan ECCS Systems, High Pressure Core Spray, PP. 6 and 15 ANSWER 2.03 (2.00)
a.
(0.5)
b.
Provides opportunity for CHS system [0.50] to recover RPV water level [0.50].
(1,0)
c.
Narrow Range (0.5)
REFERENCE NMP-2, Lesson Plan ECCS's, ADS, PP.
4, 13 and 14
_ -
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.
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
ANSWERS -- NINE MILE POINT 2-86/07/21-BISHOP,M.~
'
ANSWER 2.04 (2.00)
Minimizes delay time [0.5] between system initiation and' injection
[0.5] and prevents water hammer [0.5] to the pump and discharge piping [0.5].
(2.0)
REFERENCE NMP-2, Lesson Plan, RCIC, P. 3 ANSWER 2.05 (1.00)
peed [0.5] and generator dif f erential [0. 5]. 04// */" '"'M (1.0)
Engine ove5brwei-1 c.s>>a
+,et,y /,e ot o,. d,ff a.~4sL/ care n t.
ge-<rst=~ d; s
REFERENCE NMP-2, Lesson Plan, Standby diesel generator and auxiliaries, P.
ANSWER-2,06 (1.50)
ed*ou~ ~4 G*ool"rO g
g RBCLC%., mm _ d, al?' isolation valves close [1.0] and the fens trip a
[0.5] (if LOCA override keylock switch not in OVERRIDE).
(1.5)
,v REFERENCE NMP-2, Lesson Plan, Drywell Cooling, P. 5 j
ANSWER 2.07 (3.00)
a.
To allow heatup and pressurization of the steam lines during a plant startup y se &~t e**lk" t*4 byt'"
us!" d* =4 **su uris s.
(1,g) "
/
b.
It is a one-out-of-two-taken twice arrangement. _p,
-(4-fM O C Both the IB and OB valves close. (l./,'//.cesr* -f /*-se-*F* *//&
(9. 5)(/ o c.
ausws,-) 4
,me REFERENCE NMP-2, Ops. Tech., Primary Containment Isolation System, P. 4
,
- -.
-
.. -
- -, ~
-.
-.--..---.n
,
--,n.-
,. -
.
.
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
.
ANSWERS -- NINE MILE POINT 2-86/07/21-BISHOP,M.
ovu nih '
a,, j,e o.s e.
undn voll'1'
ANSWER 2.08 (2.00)
/*" #" 4"'"'/
pd 9.Iheltsed.loLos free 123 celt: by er 10% [0.5] cr.d*
f;;qsmuvy drugs beluw 6 Es by 5% [0.5].t.
(1.0)
7*
b.
1A - 2NHS-MCC008 [0.5] og us-S
,,
IB - 2NHS-MCC009 [0.5] o, as G (1.0)
,,
( s&kis s.t teruas J ).pc REFERENCE NMP-2, Lesson Plan, RPS, PP. 7 and 8 ANSWER 2.09 (3.00)
A.
Breakdown bushing.
B.
No. 2 seal pressure will approach No. 1 seal pressure.
C.
LFMG's will not supply breakway torque.
D.
[0.75 each]
(3.0)
REFERENCE NMP-2, Ops. Tech., Reactor Recire. System, PP. 3 and 5 ANSWER 2.10 (3.00)
a.
DG bkr trips on LOCA 40.5) dicsel auc; tc crergency rede, (9. 5)"
when LOOP occurs, breaker closes and loads sequence as 7#
(1.5)
normal.(0.5)t-,
b.
The offsite bre,aker will stay closed (0.5) and the diesel will attempt to pick up the offsite test loads. (0.5)
Directional current trip will open offsite breaker and isolate bus with EDG. (0,5)
(1.5)
REFERENCE NMP2 FSAR pg. 8.3-18e LP E.O.
- 's 3,4
.
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
.
ANSWERS -- NINE MILE POINT 2-86/07/21-BISHOP,M.
2.00 (4rse)g ANSWER 2.11 O.
True j O.
True) D&M'
4. 7.
True
g c.,g.
True 4. c7.
False REFERENCE NMP2 RWM LP pgs. 3,5 of 26 and table 2 E.O. 5
\\
'
.
l
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3.
INSTRUMENTS AND CONTROLS PAGE
ANSWERS -- NINE MILE POINT 2-86/07/21-BISHOP,M.
ANSWER 3.01 (2.00)
The RBCLCW temperature is controlled by a temperature control valve (TV108)
which regulates the RBCLCW flow around and through the heat exchangers
[0.5].
The temperature control valve is a single pneumatic controller connected by linkages to two valve assemblies [0.5], one in the heat exchanger bypass line and one in the heat exchanger common discharge [0,5].
The assemblies are connected so that as one opens the other closes [0.25].
The control valve adjusts the positions of the valves to maintain system temperature at 90 Fv[0.25].
(2.0)
(+ or - S*M ) pt l
REFERENCE
'
NMP-2, Ops. Tech., RBCLCW, P.
l i
ANSWER 3.02 (3.00)
Control switch selected to CIA Lead, C1B Lag, ClO Backup.
(0.5)
.
'
l The station air compressor CIA stagts when the compressor control starts automatically when th,e(0.5];,,;Jhe station air compre switch is placed in " START"V 'edmpressor control switch is
'(MFMSTOP4(O.5)
400rpsig [0.5].
The station air compressor ClO starts automatically when the 'E' mpressor o
pg control switch is in NORMAL AFTER STOP, (0.5) and compressed air header pressure is low-low (less than 4NP psig) [0.5].
(3.0)
SS (+ o, - S pi s a s&h)
(gggy y
M N
n REFERENCE NMP-2, Ops. Tech., Instrument, Service, and Breathing Air Systems, P.
ANSWER 3.03 (1.50)
A.
G-M tubes and Ionization Chambers.
(1.0)
B.
Ionization Chambers.
(0.5)
REFERENCE NMP-2, Ops. Tech., Radiation Monitoring System, PP. 2 and 8
.
.
3.
INSTRUMENTS AND CONTROLS PAGE
ANSWERS -- NINE MILE POINT 2-86/07/21-BISHOP,M.
- ANSWER 3.04 (2.00)
(2.0)
DIV. I - Green Co " 9 -
-
DIV. II - Yellow 0.50 each]
(1.5)
DI_V. III - Purple
-
olor codius is same for both AC and DC.
(0.5)
(Will accept color designations for bother AC and DC.)
k h./y/,.
REFERENCE
NMP, Ops. Tech., Plant DC ELECT. DIST. SYSTEM, P. 2'
pg ANSWER 3.05 (3.00)
a.
Throttle (or Reactor) pressure Generator Load (MWe)
Turbine Speed
[0.5 each]
(1.5)
b.
The pressure control circuit (0.5)
c.
Speed control circuit [0.5].
This is because if the main generator is " tied" to the grid, the main generator cannot spin any faster or slower than grid frequency [0.5].
(1.0)
REFERENCE NMP-2, Ops. Tech., Turbine Electrohydraulic Control, Rev. 2, P.
5-9 of 14
.
3.
INSTRUMENTS AND CONTROLS PAGE-25 ANSWERS -- NINE MILE POINT 2-86/07/21-BISHOP,M.
'
ANSWER 3.06 (3.00)
9.
Shutdown Range 1.
120 in the vessel 2.
80 in the drywell b.
Narronw Range 1.
1000 psig in the reactor 2.
135 in the drywell c.
Wide Range 1.
1000 psig in the Reactor 2.
135 in the drywell 3.
No jet pump flow 4.
20 BTU /Lbm core inlet subcooling d.
Fuel Zone Range
"0" psig in drywell,,. zsz'M 4"% W
"0" psig in RPV e 2,2 'r /- t/v 1.
2.
3.
No jet pump flow
[0.27 each]
(3.0)
REFERENCE NMP-2, Ops. Tech., Reactor Vessel Instrumentation System, Rev.
2, P. 6-9 of 30 ANSWER 3.07 (2.00)
When the RWM rod test pushbutton is depressed, if all control rods are full in except the single rod selected, the " select" half of the pushbutton will illuminate [1.0].
If another control rod is_ selected, the select light will extinguish, and then relight if the first selected control rod is full in [1.0].
(2.0)
REFERENCE NMP-2, Ops. Tech., Reactor Manual Rod Control, Rev. 2, P. 12 of 26 ANSWER 3.08 (2.00)
1. Shuts the feed regulating valves.r d int;;10cks th;;. : hut?' PW (1.0)
2. Downshift or trip the recirculation pumps (accept either action)
(0.5)
3.
Initiates the SLS System (0.5)
REFERENCE NMP-2, Ops. Tech., Redundant Reactivity Control System, Rev. 2, P. 7 of 12
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3.
INSTRUMENTS AND CONTROLS PAGE
'
ANSWERS -- NINE MILE POINT 2-86/07/21-BISHOP,M.
i ANSWER 3.09 (3.00)
g fe a.
lower than, since SF/FFCmismat (25%)theFWflowwillbe (1.0)
since FF/SFCmismatchM(50%)rd reduced to compensate for it.
u FF will increase to b.
higher than, compensate for it.
(1.0)
c.
higher than, since circuit sees low level and will feed at max flow until high Iccci tri;/L,g (1.0)
REFERENCE NMP-2, Ops. Tech., Feedwater Control System, Rev. 2, P.
5 of 6 ANSWER 3.10 (1.00)
e.
APRM feedback (C, normal; E, backup)
(Also accept flux estimator feedbac o /. 6e /bar)/w b.
Loop elbe"? flow.
A<, c fled ye c.
Flow control valve position feedback (Also accept velocity feedback).
[0.33 each]
(1.0)
REFERENCE NMP-2, Ops. Tech., Recirculation Flow Control, Rev. 2, PP. 4, 5,
6 of 16 i
ANSWER 3.11 (1.50)
Parameter Setpoint
________
_________
A.
High RPV pressure 128 psig B.
Low RPV level, (level 3)
159.3 inches C.
RHR area high temperature 135 F D.
High Reactor Building Temp.
135 F E.
High Reactor Building Pipe Chase Temp.
135 F
[any 3 @ 0.25 for parameter 0.25 for setpoint) (1.5)
REFERENCE NMP-2, Ops. Tech., Primary Containment Isolation System, Rev. 2, Table 1
.
3.
INSTRUMENTS AND CONTROLS PAGE
'
ANSWERS -- NINE MILE POINT 2-86/07/21-BISHOP,M.
ANSWER 3.12 (1.00)
Y to3 There is no load shed or sequencing for 2 ENS * SWG 409.
(1,0)
( tu;// ah= ~~rt h-d sLJ nJ sy-~ A.?ns o swa lot as Aiks :
3d REFERENCE NMP2 Lesson Plan, Emergency AC Power Systems, page 7.
I. $MS funf "A sh Is (T> 1)
N. z. a s p,sf-fr (ra)
M y. savu. w tn p.y ss,h (r.zs)
.
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4.
PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE
RADIOLOGICAL CONTROL
,
ANSWERS -- NINE MILE POINT 2-86/07/21-BISHOP,M.
ANSWER 4.01 (2.00)
A.
Denotes a requirement.
B.
Denotes a recommendation.
If not acted uyvu, maplanation is O Av ou-required justifying the oviivu Lokou.. pg
^
C.
Denotes permission.
D.
What is to be done or what is expected.
Ncither a rcquir:rento pd rcccancndation.O.,j
[0.5 each]
(2.0)
ner REFERENCE NMP-2, Procedure AP-1.0, Procedure for Admin. Controls, PP. 3 and 4 ANSWER 4.02 (2.00)
1.
Responsible for general operation of the control room (Subject to SSS and ASS).
2.
Responsible for direct supervision of operators on shift.
3.
Responsible for starting and stopping all major equipment.
4.
Responsible for control of turbine.rrd rc::ter.2 pe 5.
Responsible for operation of major power board breakers.
6.
Responsible for operation of line breakers in switchyard.
7.
Responsible for determining if an ESF performs as required in the event of a LOCA or other abnormal incident.
9. ps?/a.4y JL cso (%b ) A*~) hf. u
[any 4 @ 0.5 each]
(2.0)
p,p.,v,;y. L cnbs /. /' r s* * be
.,,s REFERENCE NMP-2, Lesson Plan, Composition and Responsibility of Unit Org., Item 11.
NMP-2, Procedure AP-1.2, Composition and Responsibility of Unit Org., P. 5 ANSWER 4.03 (2.50)
1.
RPV water level < 159.3 inches 2.
RPV pressure > 1037 psig 3.
Drywell pressure > 1.68 psig 4.
A condition which requires an MSIV isolation.
5.
A condition which requires a Rx scram, and Rx power is above 4%
or cannot be determined.
[0.5 each)
(2.5)
REFERENCE
,
NMP-2, Emergency Operating Procedure N2-EOP-RQ, RP7 Reactivity
'
Control, P.
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.
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4.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE
RADIOLOGICAL CONTROL
,
ANSWERS -- NINE MILE POINT 2-86/07/21-BISHOP,M.
ANSWER 4.04 (1.50), y.jk.A.cL~fush'- W 1.
Dump H2 to atmosphere (to reduce H2 pressure)
a 2.
Trip the generator 3.
Trip the turbine if st: : 30% P.; ;: :re_ ve
[0.5 each]
(1.5)
REFERENCE NMP-2, N2-OP-27, Generator H2 and CO2 System, Section H.2.6 ANSWER 4.05 (3.00)
a.
Thermal stratification as noted by axial skin temperature variations.
p>o4r**8 pd (1.0)
b.
1. Raise level to above the steam separatorsa(202G),for natural circulation. (1.0)
2. Maintain SDC flow normal and throttle service water. (1.0)
(2.0)'
REFERENCE NMP-2, N2-IOP-101C pg 6 ANSWER 4.06 (2.50)
A.
1.
Rough starts 2.
Steam line water hammer 3.
High rotor eccentricity 4.
Large differential expansions 5.
Large vibration increase
[any 4 @ 0.5 each]
pe (2.0)
B.
Trip the turbine,-(9.25' and glovo iL :n th; tumulus gear.(0.25'; (0.5)
.
REFERENCE NMP-2, N2-OP-21, Main Turbine, Section D.15.0, D.1.
4.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE
RADIOLOGICAL CONTROL
,
ANSWERS -- NINE MILE POINT 2-8S/07/21-BISHOP,M.
ANSWER 4.07 (2.00)
A.
(Verify annuncia:cr) and rod drift alarm on full core display.
B.
Select rod so its position can be monitored.
C.
Check rod sequence sheet to determine proper position of the rod.
D.
Place rod in its proper position.
E.
If rod continues to drift, check Tech. Specs.
F.
Notify Reactor Analyst to correct possible flux problems.
G.
vu;/y M. 54 6;- d.f.-s<-.
[any 4 @ 0.5 each]
(2.0)
P/
REFERENCE NMP-2, N2-OP-96, Reactor Manual Control and Rod Position Indication,Section I.5.3 ANSWER 4.08 (1.50)
The DC volmeter on the charger must show battery voltage [0.5] before energizing the AC input to the charger [0.5], or the charger output filters could be damaged [0.5].
(Always connect the battery to the charger output before energizing the charger.)
(1.5)
REFERENCE NMP-2, N2-OP-73B, 24 Volt DC Distribution, Section E, Caution I. So pe ANSWER 4.09 (-3-ee)
a.
By cross connecting the LPCS suction (0.25) and RHS SD cooling suction (0.25) via isolation valves (0.25) and a removeable spoolpiece. (0.25)
This is done to allow testing of the LPCS system (0.25) with a suction from the reactor vessel.(0.25)
(1.5)
b.
1.
To prevent excessive D/P across the valve.
(0.76) 1
'-
2. To ensure the reactor does not drain to the suppression
'
pool.
(0.75)
REFERENCE NMP2 OP 33 and LP LPCS; E.O.
- 's 2.9 I
IY
-.-
-
-
uh M
.
4.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE
RADIOLOGICAL CONTROL
,
ANSWERS -- NINE MILE POINT 2-86/07/21-BISHOP,M.
ANSWER 4.10 (2.00)
a.
To protect the public health and if no other approved method exists.
4# y ss5 (1.0)
b.
Prior approval from a licensed SROjand if time permits SORC review and NRC notification.
(1.0)
REFERENCE NMP2 Procedure AP-4
,,
ANSWER 4.11 (3.00)
1. Submergance - water level > TAF
,(1.00)
2. Spray Cooling one core spray at or above design conditions.
(1.00)
3. Steam Cooling proper pressure and steam flow through SRV's.
(1.00)
REFERENCE NMP2 EOP structure page 3-1.
.
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,
...
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TEST CROSS REFERENCE PAGE
(
QUESTION VALUE REFERENCE
________
______
__________
01.01 1.50 TEX 0000365 01.02 2.00 TEX 0000366 01.03 3.00 TEX 0000367 01.04 3.00 TEX 0000368 01.05 2.00 TEX 0000369 01.06 2.00 TEX 0000370 01.07 2.00 TEX 0000371 01.08 2.50 TEX 0000372 01.09 1.50 TEX 0000373 01.10 1.00 TEX 0000374 01.11 2.50 TEX 0000375 01.12 2.00 TEX 0000376
______
25.00 02.01 3.00 TEX 0000377 02.02 2.00 TEX 0000378 02.03 2.00 TEX 0000379 02.04 2.04 TEX 0000380 02.05 1.00 TEX 0000381 02.06 1.50 TEX 0000382 02.07 3.00 TEX 0000383 02.08 2.00 TEX 0000384 02.09 3.00 TEX 0000385 02.10 3.00 TEX 0000386 02.11
_ _g g TEX 0000387 u.u -
w. nn 03.01 2.00 TEX 0000388 03.02 3.00 TEX 0000389 03.03 1.50 TEX 0000390 03.04 2.00 TEX 0000391 03.05 3.00 TEX 0000392 03.06 3.00 TEX 0000393 03.07 2.00 TEX 0000394 03.08 2.00 TEX 0000395 03.09 3.00 TEX 0000396 03.10 1.00 TEX 0000397 03.11 1.50 TEX 0000398 03.12 1.00 TEX 0000410
______
25.00 04.01 2.00 TEX 0000399 04'.02 2.00 TEX 0000400 04.03 2.50 TEX 0000401 04.04 1.50 TEX 0000402 04.05 3.00 TEX 0000403 04.06 2.50 TEX 0000404
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TEST CROSS REFERENCE PAGE
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NUESTION VALUE REFERENCE
________
______
__________
04.07 2.00 TEX 0000405 04.08 # 1.50 TEX 0000406 04.09 t.5 4. 0T TEX 0000407 04.10 2.00 TEX 0000408 04.11 3.00 TEX 0000409
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hf
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______
""
- ".
92. o o W l
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STTAC hmet?? /
/ I
-
Ib fos /nuse, o
.
U.
S.
NUCLEAR REGULATORY COMMISSION
.
SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
NINE MILE POINT 2
_
REACTOR TYPE:
BWR-GES
,
B6/07/21
_ _ _
DATE ADMINISTERED: _
_CRESCENZO _F.___________
EXAMINER:
t rre-~
__ gd[_f[3
-__ __
APPLICANT:
_
INSTRUCTIONS _Tg_ APPLICANT:
Write answers on one side only.
Use separate paper for the answers.
Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%.
Examination papers will be picked up six (6)
hours after
,
)
the examination starts.
% OF CATEGORY
% OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY
---
__
---
5.
THEORY OF NUCLEAR POWER PLANT
_ 25. 00__ _ __25. 00
_
___
____
OPERATION, FLUIDS, AND THERMODYNAMICS 6.
PLANT SYSTEMS DESIGN, CONTROL,
_25.g9__
25.00
._
__
AND INSTRUMENTATION
_25 99__
25.00
__
7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 25.00__
25.00
__
_ _ _
_ 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 199199__ 100 99
________ TOTALS
___________
FINAL GRADE
__
_%
All work done on this examination is my own. I have neither given nor received aid.
APPLICANT'S SIGNf.IURE
- _ - - _ _ _
.
s FLUIDS _AND PAGE
5.
THEGRY_OF_ NUCLEAR POWER PLANT OPERATION
t t
THERMODYNAMICS
..
,
QUESTIDN 5.01 (2.50)
a. At the beginning of a fuel cycle, control rod density is approximately 10 to 12% at equilibrium full power. Approximately one third into the cycle, the control rod density is about 15 to 16% at equilibrium full power. Explain why rod density (1.00)
changes.
b. What ef f ects does this increase in control rod density have on (1.50)
the void coefficient of reactivity?
DOESTION 5.02 (1.00)
During a rapid power increase, very short periods can be maintained, yet for rapid power decreases, the period quickly becomes -80 sec.
(1.00)
Explain the reason for the difference.
DUESTIDN 5.03 (2.50)
I. b0 Why are installed neutron sources needed in the initial core a.
' ;. 004 loading?
Assume no administrative limitations prohibited reactor startup b.
i without the neutron sources. Could the reactor achieve criticality without these sources? If so, how would the b'UC critical rod density change relative to a startup with ii.SGF a source installed?
.
DUESTION 5.04 (3.00)
Regarding the xenon transient following a significant DECREASE in reactor power from high power operation:
a. HOW will peripheral control rod worth be affected (INCREASE, DECREASE, REMAIN THE SAME) during the xenon peak?
BRIEFLY
!
(1.50)
EXPLAIN your answer.
b. If the decrease in reactor power was from 100% to 50%,
would the new (50% power) equilibrium xenon reactivity be MORE THAN, LESS THAN or EQUAL TO one half the 100%
(1.50)
equilibrium value?
Briefly, JUSTIFY your answer.
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
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5.
THEORY OF NUCLEAR PDWER PLANT OPERATION, FLUIDS AND PAGE
t
__ THERMODYNAMICS
.
QUESTION 5.05 (2.50)
The figure below represents six fuel assemblies within a critical rod Why would an operator expect a higher than normal reactor.
(2.50)
worth associated with control rod B-2?
X= control rod fully inserted O= control rod fully withdrawn 1.
X
X 2.
O X
O 3.
X
X A.
B.
C.
QUESTION 5.06 (2.00)
Current plant conditions suggest that a severly degraded core condition exists. Several SRV's have been, or are currently, open.
The STA reports that the temperature detectors in the SRV tailpipes are reading ERRONEOUS since the readings are significantly higher than those calculated from the Mollier diagram. Would you agree (2.00)
with this conclusion? Justify your answer.
DUESTION 5.07 (3.00)
List three parameters which contribute to AVAILABLE NPSH for a a. recirculation pump. Limit your answer to those parameters which (1.50)
are directly available in the control room.
b. Consider two RPV conditions: low power and low flow (< 10 %) OR high power and high flow ( >B57.).
1. During which condition is REQUIRED NPSH for a recirculation (0.50)
pump greater?
2. During which condition is AVAILABLE NPSH for a recirculation (1.00)
pump greater and why is it greater?
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
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PAGE
5.
THEdRY OF NUCLEAR POWER _PL. ANT OPERATIONt_ FLUIDS _AND
t THERMODYNAMICS
.
QUESTION 5.08 (2.00)
Match the descriptions (a-d) to the corresponding 1/M plots (1-4).
(2.00)
(e) tDE AL (b) DETECTOR TO CLOSE TO SOURCE (c) DETECTOR TO FAR PRots MURCE e
CR.
7'E (d) PREDICTON AFTER IDo BUNDLES (L)
l.0 -
.9 -
.a -
.7 -
(l)
.6 -
.5 -
3)
4-s
.s -
\\
\\
CRITICAllTY 2_
~
(\\,
ibo abo sbo abo soo sho
"
sammet or rutt swoLEs s narrn QUESTION S.09 (2.00)
level The term " critical power" refers to that bundle power corresponding to the onset of transition boiling (OTB) somewhere in that bundle.
State how critical power varies (ie. increases, decreases, or is not affected) by each of the following:
(O.50)
a. If coolant mass flow rate increases (0.50)
b.
If reactor pressure increases (0.50)
c.
If local power increases (0.50)
d.
If inlet subcooling increases (***** CATEGORY OS CONTINUED ON NEXT PAGE *****)
.
.
FLUIDS _AND PAGE
THEO'RY_OF,NtJCLEAR POWER _ PLANT OPERATION
'
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5.
THERMODYNAMICS
.
QUESTION 5.10 (2.50)
Refer to the attached figure 15.3-2 from the Nine Mile Point 2 FSAR concerning a trip of both recirculation pump motors.
Explain the sequence of events which lead to the pressure (1.50)
m. spike noted from time T=7 to T=10 secs.
b. Explain the behavior of core inlet subcooling from time (1.00)
T=7 to T=2O secs.
QUESTION 5.11 (2.00)
A heat balance has to be performed on your shift due to a computer software malfunction. Using the following information determine the (2.00)
core thermal power.
= feedwater flow
= 6400000 lb/hr 110000 lb/hr MCU
= cleanup demin flow
=
= steam flow
= 6423000 lb/hr 23000 lb/hr MCRD
= CRD flow to reactor
=
345 Btu /lb HFW
= enthalpy of feedwater
=
506 Btu /lb HCU,IN = enthalpy cleanup flow in
=
419 Btu /lb HCU,0UT= enthalpy cleanup flow out =
1194 Btu /lb HG
= enthalpy of steam
=
68 Btu /lb Hcrd
= enthalpy of CRD flow to reactor
=
26500000 Btu /hr Op
= recir. pump energy input
=
2040000 Btu /hr Of1
= heat losses from nuclear boiler
=
Qcore
= reactor core thermal energy input = solve for (***** END OF CATEGORY 05 *****)
.__
._
I AND INSTRUMENTATIDN PAGE
6.__ PLANT _ SYSTEMS DESIGN, CONTROL
'
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QUESTION 6.01 (3.00)
With the plant initially at 1007. steady state power, briefly describe the response of RPV level to the following input signal failures to the Feedwater Control Include in your answer a discussion of how the FWCS will System.
respond. Continue your discussion to a point where stability is reestablished. Assume FWCS is in three element control mode i
end no operator actions.
(1.00)
Feedwater flow signal fails low.
(1.00)
a.
b.
Steam flow signal fails low.
(1.00 c.
RPV level signal fails low.
QUESTION 6.02 (1.00)
SELECT which one of the following best describes the operation /
(1.00)
t performance of an IRM during a reactor startup.
a. When the IRM is reading full scale on Range 10, the APRM's should be reading approximately 10*/. power.
b.
Shifting from Range 4, indicating 75, to Range 5, will result in an indication of 24, en Range 5.
c. Reactivity feedback, due to the moderator temperature coefficient should begin at approximately Range 5.
d. When an IRM channel increases from 25 on range 2 to 25 on range 3, the indication has increased by one decade.
W (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
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CONTROL AND INSTRUMENTATION PAGE
6.__ PLAN'T SYSTEMS DES _I_GN,
~
.
QUESTION 6.03 (3.00)
With the plant operating at 1007. power, recirc in Flux Manual, inadvertently decreases the EHC pressure setpoint an operator by 5 psi. What will be the initial response and final status of the following due to this action? Briefly explain for initial response only. Refer to the attached EHC logic diagram if necessary. Assume load limit set at 100% and max (3.00)
combined flow set at 105%.
a.
TCV position.
b.
BPV position.
c.
Reactor power.
d.
Reactor pressure.
)
QUESTION 6.04 (2.00)
Describe the conditions necessary to cause the following alarms on the Interlock Status Display Module associated with the ref ueling platf orm. USE attached Figure 2 for reference.
(0.50)
a.
Back Up Hoist Limit (O.50)
b. Rod Block Interlock No. 1 (0.50)
c.
Fuel Hoist Interlock (O.50)
d.
Bridge Reverse Stop No. 1 i
QUESTION 6.05 (1.50)
Briefly explain how the leak detection system used for the High Pressure Core Spray System functions to indicate a break in the (1.50)
system.
l
(***** CATEGORY 06 CONTINUED ON NEXT PAGE
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PLANT _gYgTEM@_DE@IGN _CONTROLt_AND INSTRUMENTATION PAGE
3__
t QUESTION 6.06 (2.00)
e. How does the reactor vessel water level instrumentation fail upon a loss of power? Include both meters and recorders.
(1.00)
b. Will ECCS initiate upon a loss of power to the reactor (1.00)
vessel water level instrumentation? Why or why not?
QUESTION 6.07 (2.00)
Explain how the recirculation loop flow signals used to bias the APRM's and RBM are generated. Include those portions of the system from the recirculation loops to the APRM/RBM channels. (simplified diagram is adequate; discussion of recirc flow venturis NOT (2.00)
required)
QUESTION 6.08 (2.50)
Indicate whether the following statements regarding the RMCS tre TRUE or FALSE:
a. The currently latched RWM group (i) is that group with a rod withdrawn past its insert limit and with no greater than 2 (0.50)
insert errors in RWM groups 1 to i-1.
b. The rod blocks imposed by the NMS (including the RBM) and the service and refuel platforms are dependent on the Mode Switch (0.50)
position.
c. A RWM rod block will occur if more than one control rod is (O.50)
withdrawn and the Rod Test puslinutton is depressed.
d.
A double X, (XX), indication on the four rod display, indicates that the RPIS is receiving abnormal data.
(0.50)
System hardware malfunctions in the RWM will not cause rod e.
(O.50)
blocks when above the LPSP.
($$$$$ CATEGORY 06 CONTINUED ON NEXT PAGE *****)
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6.
PLANT SYSTEMS DESIGN CONTROL 1__AND INSTRUMENTATION PAGE
'
QUESTION 6.09 (3.00)
Describe the interlocks associated with the 4.1 KV emergency bus or switchgear which will ensure the emergency bus remains cnergized given the following malfunctions:
c. The diesel generator is running in test, paralleled to the grid, (1.50)
when a LOCA signal occurs followed by a loss of offsite power.
b. The diesel generator is running in test, paralleled to the grid, (1.50)
when a loss of offsite power occurs.
QUESTION 6.10 (2.00)
Assume an EDG-2 starting circuit fuse failure has initiated a. logic for Service Water to the CSH diesel. The CSH diesel does NOT start. Explain how Service Water valves MOV 95A,B (Water header valves) and MOV 94A,B (cooler return valves) will respond. State any assumptions made as to initial valve (1.00)
lineup; include setpoints and time delays.
How would these same valves respond with a valid start signal b. to the CSH diesel, successful start of the diesel, but low service water pressure to the CSH diesel. State any assumptions made as to initial valve lineup; include setpoints and time (1.00)
delays.
QUESTION 6.11 (3.00)
Briefly explain how and why LPCS can be cross-connected with a.
(1.50)
the RHS system.
b. While LPCS and RHS are cross-connected as described above, the operator is cautioned to:
1. Ensure LPCS suction valve MOV 112 is open.
2. Ensure LPCS suction isolation valve 2CSL*V121 remains shut.
Explain the reasons for each of these cautions (1.50)
.
(***** END OF CATEGORY 06 *****)
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7.
PROC'EDURES - NORMAL ABNORMALt _ EMERGENCY ANE)
t RADIOLOGICAL CONTROL DUESTION 7.01 (1.50)
Concerning PCIS, if a containment isolation had occurred due'
to an actual 10-10 RPV level or hi drywell pressure, what two administrative precautions must be taken prior to resetting (1.50)
or bypassing the isolation signal?
QUESTION 7.02 (1.50)
A precaution in N2-DP-92, Neutron Monitoring System, states that "BWR cores typically operate with neutron flux noise. Care chould be taken when operating in this area."
(0.50)
What problem can this noise create?
a.
b. In what specific operating condition is this applicable?
(0.50)
What actions are required if this conditions exists?
(0.50)
c.
QUESTION 7.03 (2.50)
I a. As the SSS, you have just declared a General Emergency. Per EPP-26, you must make Protective Action Recommendations to evacuate a 2 mile radius and shelter 5 miles downwind.
and a wind Using the attached figures from EPP-8 and EPP-26, direction as shown on figure 14, determine which ERPA must (1.50)
seek shelter.
B. Refer to the attached figure 1 of EPP-26, " Recommended Protective Actions for General Population and Emergency Workers." Explain the significance of the upper and lower (1.50)
dose rate limits.
j QUESTION 7.04 (2.50)
Procedure N2-EOP-C7, " Level / Power Control," directs the operator to restore RPV water level to between 159.3 and 202.3 once the SLC tank has lowered to 3000 gallons. Assuming all control rods remain withdrawn, why does the procedure have the operator restore (2.50)
,
'
water level at this point?
,
(*****
CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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7.
PROCEDURES - NORMAL, ABNORMAL EMERGENCY AND PAGE
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t RADIOLOGICAL CONTROL
'
-_
QUESTION 7.05 (3.00)
During an ATWS condition if RPV water level should become a. undeterminable, EDP-C7, " LEVEL / POWER CONTROL," directs the injection operator to maintain power above 8% by controlling to the RPV. What is the significance of 8% and why must (2.00)
I power be mintained above it?
b. If, in the above situation, the operator is unable to maintain reactor power above 8%, he will be directed to depressurize to below the " MINIMUM ALTERNATE FLOODING PRESSURE". Explain (1.00)
the basis for this pressure.
QUESTION 7.06 (3.00)
from the Adequate core cooling is defined to be heat removal reactor sufficient to maintain fuel clad temperature < 2200 deg.
F.
According to the EOP's, three viable mechanisms of adequate core cooling exist. List these three mechanisms IN ORDER OF PREFERENCE cnd include how adequate core cooling is verified for each (3.00)
mechanism.
I QUESTION 7.07 (1.00)
i Define " Maximum Safe Operating Parameter" as it applies to (1.00)
i the EOP's.
QUESTION 7.08 (2.00)
Contingency 6 of the EDP's, RPV Flooding," provides instructions
"
f or the operator to flood the RPV without level indication. These intructions direct the operator to inject to the RPV until RPV 80 psig is pressure is 80 psig above suppression chamber pressure.
known as the Minimum Flooding Pressure. Discuss the basis for this (See attached sheet 3 of 5, N2-EOP-C6)
(2.00)
pressure.
f (***** CATEGORY 07 CONTINUED ON NEXT PAGE
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.-..c--
.,,,,
,~,,.s___,-___,,n
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y
..__,yg_ _.., _...,, _., _ -,,, _ _ _
_. - - - _. _ - - -, - _.. - - -,, -.., <
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PAGE
PROCEDURES -_ NORMA k ABNORMAL _ EMERGENCY AND
~
'
t 7._
RADIOLOGICAL CONTROL QUESTION 7.09 (2.00)
Explain the basis f or the f ollowing prerequisites to transf er of a
rscirculation pump from low to high speed.
"Feedwater Flow greater than 4.25 million pounds per hour (1.00)
c.
(307. of rated)...."
b. " Differential temperature between recirculation pump SUCTION and STEAM DOME TEMPERATURE is greater than 10.7 deg F."
(1.00)
DUESTION 7.10 (2.00)
a. According to N2-OP-39, " Fuel Handling and Reactor Service Equipment," a stainless steel jamming button must be installed on the auxiliary hoist cable if this hoist is used to handle contaminated equipment. Explain the purpose of this caution.
(1.00)
b. What limitation exists in using the " load float switch" on the (1.00)
main hoist while it is loaded?
QUESTION 7.11 (1.00)
Procedure N2-OP-62, "DBA Hydrogen Recombiner System," cautions the operator to ensure both the inboard and outboard isolation valves for the recombiner are shut during standby modes of operation.
(1.00)
What is the purpose of this caution?
(*****
CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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PROCEDURES - NORMAL _ ABNORMAL _ EMERGENCY _AND
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1 7.
RADIOLOGICAL CONTROL QUESTION 7.12 (3.00)
Procedure N2-IOP-101C, " Plant Shutdown," notes that while in it ehutdown cooling with both reactor recirculation pumps of f, the vessel or venting is possible to observe pressurization of off of steam.
indications Explain why this condition might occur and what a. are available to the operator to assist in recognition of (1.00)
this condition.
b. Explain two methods by which an operator can (2,00)
prevent the above condition from occuring.
l
.
(*****
END OF CATEGORY 07
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8.
ADMINISTRATIVE PROCEDURES _ CONDITIONS _AND_ LIMITATIONS t
t QUESTION 8.01 (3.00)
The plant is operating at 30% power. The EDG-1 is currently (it has been inop for inoperative due to planned maintenance 10 hrs.). The I&C supervisor reports that instrument 25W8PSL95B.
(Service Water inlet pressure for EDG-2) has failed a channel functional test.
Using the attached technical specifications, determine ALL a.
(2.00)
applicable actions which must be taken.
b. How would your actions be different if, rather than EDG-1 being inop for maintenance, LPCS was inop for maintenance?
(1.00)
QUESTION 8.02 (2.00)
The plant is operating at 75% power when it is determined that SRV 121 must have its control circuits for all three solenoids de-energized simultaneously for maintenance. Use the attached technical specifications to discuss ALL actions which could be (2.00)
Epplicable for this case.
QUESTION B.03 (2.50)
With the mode switch in RUN, and reactor power approximately 16%,
it is determined that one of the main turbine bypass valves is inoperative. Using the attached technical specifications, determine what actions must be taken if it is desired to continue
<
the startup and proceed to RATED conditions (if possible).
Reference all technical specifications used in developing (2.50)
your answer.
i QUESTION 8.04 (2.50)
Explain how and why the CNFLPD and the FRACTION OF RATED THERMAL (2.50)
POWER are used to adjust APRM scram setpoints.
(*****
CATEGORY 08 CONTINUED ON NEXT PAGE $$$$$)
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8.
ADMINISTRATIVE _ PROCEDURES CONDITIONS _AND_ LIMITATIONS
'
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QUESTION 8.05 (1.75)
Indicate whether each of the following is considered a " Core Alteration" per the Nine Mile Point 2 Technical Specifications.
(O.75)
a.
1.
Withdrawal and insertion of an SRM detector to check the drive motor.
2.
Removal of an LPRM string for replacement.
3.
Removal of an uncoupled control rod for replacement.
While withdrawing a control rod to test a position indicator b.
it becomes necessary to suspend core alterations due to probe, Should the rod be reinserted or must it containment problems.
remain mid-positioned until conditions are such that core (1.00)
alterations are allowable?
QUESTION 8.06 (3.00)
10CFR2O limits for penetrating radiation for the a. State the (2.00)
followings 1. Whole body. (without an NRC form 4)
2. Whole body. (with an NRC form 4)
3. Extremities.
4.
Skin.
(1.00)
b. State the allowable emergency exposure limits.
QUESTION B.07 (1.50)
RAM Where would you find the limits for the concentration of (0.50)
a. released in liquid effluents to unrestricted areas?
(1.00)
b. What are these limits based on?
QUESTION B.08 (2.50)
(2.50)
List 5 responsibilties the Emergency Director MAY NOT delegate.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
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8.
ADMI'NISTRATIVE PROCEDURES _CONDITIONSt__AND LIMITATIONS PAGE
t GUESTION 8.09 (2.00)
Under what conditions may a plant operator deviate from an a. approved procedure and depart from a license or Tech Spec (1.00)
condition?
(1.00)
b. Whose approval is required to do this?
QUESTION 8.10 (3.00)
Technical Specification 3/4.1.2 addresses " Reactivity Anomalies."
the method used to Briefly explain what a Reactivity Anomaly is; determine it; and why it is important to be aware of its existence.
(3.00)
QUESTION 8.11 (1.25)
(1.00)
List the Nine Mile Point 2 safety limits.
a.
b. TRUE OR FALSE: Reactor vessel water level safety limits are (0.25)
applicable in operational condition 1.
(*****
END OF CATEGORY 08 *****)
(************* END OF EXAMINATION ***************)
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_
-.
a.QUA L 6 0ha j uA & A ShtL6
.
j t'
I/C1-h)
p - F, et/ T-
- -
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!
I CL = 3.7 x 10 5q o = (L +L ) (cred)2 ap = - 1 x 10 5 gJ'T r
f
(4*v3)
L y. - 1 x 10 3 tJJ: volds n - v/(1 + d)
,
E P - I e v/(3.7 x 1010)
g,. - t 5 x 10 ' crd:"I x
t=
(2 )/lo
.
.-,.3 x 1o
,is:,..e, t - 1/ + (ei g)/xg
.
K t-1/(c-s)
I(c) = Io e-1 vvf + xvgg T1/' - in( 2) /1 E = zhg + (1 z) bg Cp = (CFbase) (K') (E )
A 5 = x5g + (1-x) Sg Q - E p at 1 in.- 2.54 cm ap - I L pv2 1 E*1* * 3 7E3 lit *f5
D 2sc I - 64/Le 1 4 - 2.205 lb p - k( e f f ) -1 N
pao /A Kle 1)
17.58 vat ts - 1 EIV/ min
CRI 1-K(e f f) 2 1 psi - 6.895 Pz 5(00}
1 psi = 2.936 - H
-"
"
B 0 (G ';;
M CT. 2.
1-I(eff)1 1pst - 27 68 -
6 =.0071
.
0 - Nah L =2x 10, sec
._
,
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.
m
%
..
.
i s
Table 1.
Saturated Steam: Temperature Table Ahs Press.
Specific Volume Enthalpy Entropy Temp Lb per Sat.
Sat.
Sat.
Sat.
Sat.
Sat.
Temp Iahr Sq In liqisid Ivap Vapor Li uid Evap Vapor liquid Evap Vapor Fahr i
p vg vig _
vg g
h ig h
sg sig s
t
g 37 8 0 08859 0 016022 33047 3304 7 0 0179 1075.5 1075.5 0.0000 2.1873 2.1873 32.8 34 0 0 09600 0016021 3061 9 3061.9 1.996 1014.4 10764 0 0041 21762 2.1802 34.0 36 0 0 10395 0 016020 28390 2839 0 4.008 1073.2 1077.2 0.0081 2 1651 2.1732 36.0 18 8 018249 0 016019 2614 1 2634 2 6.018 1072.1 1018.1 0 0122 2.1541 2.1663 30.0 48 8 112163 0 016019 2445 8 2445 8 8.027 1071.0 1079 0 0.0162 2.1432 2.1594 40.0 42 8 0 13143 0 016019 2212 4 / 2212.4 10 035 10698 1079.9 0 0202 2.1325 2.1527 42 0 0 14192 0 016019 2112 8 21I? 8 12.041 10681 10807 0 0242 2 1217 2.1459 44.s 46 8 015314 0016020 19657 19657 14 047 1067.6 1081.6 0 0282 2.1111 2.1393 46.0 44 0
-
l 48 0 0 16514 0 016021 1 Alp 0 1810 0 16 051 1066.4 1082.5 0.0321 2.1006 21327 48.0 SO S 0 17796 0 016023 1704 8 1704 8 18 054 1065.3 1083.4 0 0361 2.0901 2.1262 50.0 57 8 019165 0 016024 15892 1589 2 20 057 1064.2 1084.2 0.0400 2.0798 2.1197 52.0 54 8 0 20625 0 016026 1482 4 1482 4 22.058 1063.1 1085.1 0.0439 2.0695 2.1134 54 0 58 I O22183 0 016028 1383 6 1383.6 24 059 106).9 1086.0 0.0478 2.0593 2.1070 56.0 58 8 0 23843 0 016011 17922 1292.2 26 060 1060.8 1086.9 0 0516 2.0491 2.1000 58.8 88 8 0 25611 0016033 1207.6 1707 6 28.060 10591 10871 0.0555 2.0391 2.0946 60.8 El 0 0 27494 0 016036 1129 2 1129 2 30.059 1058.5 1088 6 0.0593 2.0291 2.0885 52.0 54 8 0 29497 0 016039 1056 5 1056 5 32.058 1057.4 1089 5 0 0632 2.0192 2 0824 64.0 68 0 0 31626 0 016043 989 0 989 I 34 056 1056.3 1090 4 0.0670 2.0094 2.0764 64.0 58 8 0 31889 0 016046 926 5 926 5 36.054 1055.2 1091.2 0.0708 1.9996 2.0704 68 8 70 0 0 36292 0016050 868 3 868 4 38.052 1054.0 1092.1 0.0745 1.9900 2.0645 70.8 12 0 0 38844 0 016054 814 3 814 3 40 049 1052.9 1093.0 0.0783 1.9804 2.0587 72.8 I4 I O41550 0 016058 164 1 764 1 42.046 1051.8 1093.8 0 0821 1.9708 2.0529 14.8 76 8 0 44420 0 016063 117 4 717.4 44 043 10501 10941 0.0858 1.%I4 1 0472 76.0
'
Ts e 04746I ft016067 611 R 673 9 46.040 1049.5 1095.6 0 0895 1.9520 2.0415 78.0
88 8 050683 0 016072 633 3 633 3 48.037 1048.4 1096.4 0 0932 I.9426 2.0359 88.8 82 8 0 54093 0 016077 595 5 595 5 50 033 10473 1097.3 0 0969 1.9334 2.0303 87.8 MS 0$1102 0016082 560 3 560 3 52 029 1046.1 10982 0.1006 1.9242 2.0248 84.0 88 0 0 61518 0 016087 227 5 52/5 54.026 1045.0 1099 0 0.1043 1.9151 2.0193 06.8 98 8 065551 0016091 4%8 496 8 56 022 1043.9 1099.9 0.1079 1.9060 2.0139 88.9 90 0 0 69813 0 01609"l 4681 468I 58 018 10423 1100 8 0.1115 1.8970 2.0086 98.0 92 8 074313 0 01610i 441 3 4413 60 014 1041.6 1101 6 01152 1.8881 2.0033 92.8 94 8 0 19062 0 016111 416 3 416 3 62 010 1040.5 1102 5 01188 1.8792 1.9980 94.0
>
98 I O84012 0 016111 392 8 392 9 64 006 1039 3 1103.3 0 1224 1.8704 1.9928 95.0 d,
90 8 0 89356 0 016123 370 9 370 9 66 003 10382 1104 2 0 1260 1.8617 1.9876 30 0
-
e
.
...
.
-
__
s
.
P Ahs Psess
$pcolic Volume folhalpy Entropy Temp lb per Sal Sal.
Sal.
Sal.
Sal.
Sat.
Temp
1 alo Sq lo liquist ivap Vapos liquid Ivap Vapor liquid Evap Vapor Fahr I
p vg vig vg h l h fg h
sg sig s
t g
g
ige 8 0 94924 0 016130 350 4 350 4 61.999 1037.1 11051 0.1295 1.8530 1.9825 IRO
'
i 182 I I00789 0 016137 331 1 331 1 69 995 1035 9 1105.9 0.1331 1.8444 1.9775 102.8 104 8 1 06 % 5 0 016144 313 1 3131 11.992 1034.8 1106 8 0.1366 1.8358 1.9725 1KS l
l$6 8 11347 0 016151 29616 296 18 73 99 1033 6 1101.6 0.1402 1.8273 1 % 75 1Es i
100 0 12030 0 016158 2R0 28 280 30 7598 1032.5 1108.5 0.1437 1.8188 1.9626 100.0 lis t 1.2750 0 016165 26537 26539 77.98 1031.4 1109.3 01472 1.8105 1.9577 lite
112I I3505 0 016173 251 37 251 38 79 98 1030.2 1110 2 01507 I8021 1.9528 112.0 114 5 14299 0 016180 238 21 238 22 81.97 1029I 1111.0 0.1542 1.7938 1.9480 114.0 lig e I$133 0 016188 225 84 22585 83 97 1027.9 1111.9 0.1577 1.1856 1.9433 116.0 214 21 85.97 1026.8 1112.7 0.1611 1.7774 1.9386 110.0 lig e I6009 0 016196 214 20 f its t.
I6927 0 016204 203 25 203 26 87.97 1025 6 1113 6 0.1646 13693 1.9339 IMO 122 e 11891 0 016213 192 94 192 95 89 96 1024.5 1114 4 0.1680 11613 1.9293 1 22.0 124e 18901 0 016221 18323 18324 91 %
1023 3 1115 3 0 1715 13533 1.9247 124.0 178 8 1 9959 0 016229 174 08 174 09 93 96 1022.2 1116.1 0 1749 13453 1.9202 1ES
-
'
128 0 2 1068 0 016238 16545 165 4; 95 96 1021.0 1117.0 0 1783 11374 1.9157 128.0 130 0 2 2230 0 016247 157 32 15733 97.96 1019.8 1117.8 01817 13295 1.9112 130.0 132 8 2.3445 0 016256 149 64 14966 99 95 1018 7 1118 6 0 1851 13217 1.9068 132.0 1348 2 4717 0 016265 14240 14241 101.95 1017.5 1119.5 0.1884 13140 1.9024 1R0 134 8 2 6047 0 016214 13555 135 57 103 95 1016 4 1120.3 0.1918 11063 1.8900 1K0 13e I 2 1438 0 016784 129 09 129 11 105 95 1015.2 1121.1 0 1951 16986 1.8937 130.0 148 8 2 8892 0 016293 122 98 123 00 107.95 1014 0 1122.0 0.1985 1.6910 1.8895 1440 142 8 3 0411 0 016303 117 21 117 22 109 95 1012.9 1122.8 02018 1.6534 1.8852 142.0 144 s 31997 0 016312 Ill 14 Ill 16 Ill 95 10113 1123 6 0 2051 1.6759 1.8810 144.0 148 8 3 3653 0 016322 10658 106 59 113 95 1010.5 1124.5 02084 1.6684 1.8769 140.0 les e 3 5181 0 016312 10168 10110 115 95 1009.3 1125.3 02117 1.6610 1.8727 144.0
198 0 3 1184 0 016343 9705 9707 117.95 1000 2 1126.1 0.2150 1.6536 1.0606 1R0 152 e 3 9065 0016353 9266 92 68 119.95 1007.0 1126 9 0.2183 1.6463.l.8646 152.0 i
1548 4 1025 0 016363 8650 88 52 121 95 10058 11273 02216 1.6390 1.8606 1540 154 8 4 3068 0 016314 8456 84 57 123 95 1004.6 1128 6 0 2248 I6318 1.8566 158 0 158 8 4 5197 0 016184 80 R2 R0 83 125 %
1003.4 1829 4 0 2281 1.6245 1.8526 150.0 let I 4 F414 0 016395 1727 17 29 127 96 1002.2 1130 2 0.2313 1.6174 1.8487 100.0 182 8 49722 0 016406 73 90 1392 129 96 1001.0 1131.0 0 2345 1.6103 1.8448 182.0 1640 52124 0 016417 70 70 10 12 131.96 9998 1131.8 0 2377 1.6032 1.8409 1RS 188 8 54623 0 016428 6767 6768 133 97 998.6 1132.6 0.2409 1.5961 1.8371 1Et 180 0 5 7223 0 016440 64 18 64 80 13597 997.4 1133.4 0.2441 1.5892 1.8333 100.0
110 8 5 9926 0 016451 62 04 62 06 137.97 996.2 1134.2 0.2473 1.5822 1.8295 1R8
!
112 I 62136 0016463 5943 5945 139 98 995.0 1135.0 0.2505 1.5753 1.8258 172.0 l
t14 0 6 5656 0 016474 56 95 56 97 141.98 9938 1135.8 02537 1.5604 1.8221 174.0
178 8 6 8690 0 016486 54 59 54 61 143 99 992.6 1136 6 02568 1.5616 1.8184 IRO j
170 0 71840 0 016498 5235 52 36 145 99 991.4 1137.4 0.2600 1.5548 1.8147 IMS
/
,
Q
!
.
-
l p
'
.
,
I a.
i Abs P ess Specific Volume
[nthalpy Entropy Temp tb pe Sal Sat Sal Sal.
Sal Sal Temp Tahr Sq in ligmrt Ivap Vapo Liquid Evap Vapor liquid Evap Vapor Fahr i
p v,
vtg vg he h it h
Sig
I a
t leg e 15110 0 016510 50 21 50 22 148 00 990 2 1138 2 0.2631 1.5480 1.8111 100.4 IIII 1850
' 0 016522 48 172 18 189 150 01 989 0 1139 0 0.2662 I.5413 18075 142.0 184 8 8 203 0 016534 46 232 46 249 152 01 987.8 1139 8 0.2694 15346 18040 1 84.0 184 8 8 568 0016547 44 383 44 400 154 02 906 5 1140.5 02125 1.5219 1.8004 106.8 tes s 8 947 0 016559 42 671 42 638 156 03 985 3 1141.3 0.2756 1.5213 13 % 9 188.0 198 I 9 340 0 016512 40 94f 40 957 158 04 984.1 1142.1 0 2187 1.5144 11934 IN.8
II2 e 9147 0016585 39 331 39 354 160 05 982.8 1142.9 0 2818 1.5082 1.7900 192.5 194 s 10 168 0 016598 37 808 37 824 162 05 981.6 1143 7 02848 1.5017 13865 194.0 les e 10 605 0016611 36 348 36 364 164 06 980.4
!!44 4 0.2879 1.4952 1.7831 196.8 ige e 11 058 0 016624 34 954 34 970 166 08 979.1 11452 0.2910 1.4888 11798 198.0
'
260 4 11 526 0 016637 33 622 33639 16809 977.9 1146 0 0 2940 1.4824 17764 200.0
284 8 12 512 0 016664 31135 31151 17211 975 4 1147.5 03001 1.4697 17698 284.0 l
'
288 8 13 568 0016691 28 862 28 878 176 14 972 8 1149 0 0.3061 1.4571 13632 290.0 212 8 14 696 0 016719 26 782 26799 18037 970 3 1150.5 0 3121 1.4447 1.7568 212.0 216 0 15 901 0016747 74 818 24 894 184 20 9678 1152.0 0.3181 1.4323 17505 216.0 220 s 17 186 0 016775 23 131 23 148 188 23 965 2 1153.4 03241 I4201 13442'
220.8 2248 18 556 0 016805 21 529
?!545 19227 962 6 1154.9 0.3300 1.4081 1.7380 224.0 228 8 20 015 0 016834 20 056 20 073 196 31 960 0 1156.3 0 3359 1.3%I 13320 228 I 232 0 21 567 0 016864 18 101 18718 200 35 957.4 1157.8 0 3411 13842 13260 232.4 236 8 23216 0 016895 17 454 17471 204 40 954 8 1159 2 0 3476 1.3725 13201 236.0 l
248 8 24 068 0 016926 16 304 16 321 208 45 952.1 1160 6 0.3533 1.3609 11842 248 8 244e 26 826 0 016958 15 243 15 260 212 50 949 5 1162.0 0 3591 1.3494 17085 244.0 248 8 28 196 0 016990 14 264 14 281 216 56 946 8 1163 4 0 3649 1.3379d 1.7028 248 0 2520 30 883 0 011012 13 358 13 375 220 62 944.1
!!64 3 0 3706 1.3266 16912 257.8 255 8 31091 0 017055 17 570 12 538
??4 69 941.4 1166.1 0 3763 1.3154 16917 255 0 290 0 35 427 0 017089 11 145 11 762 22876 938 6 1167.4 0.3819 1.30f3 16862 264.8 284 I 37 894 0 017123 11 025 11 042 232 83 935 9 1168 7 0.3876 1.2933 16808 264.8 288 0 40 500 0 017157 10 358 10 375 236 91 933I 1170 0 0 3932 1.2823 I6755 268.8 212.s 43 249 0 017193 9138 9 155 240 99 930 3 1171 3 0.3987 1.2715 1.6702 272.0
!!8 0 46 147 0 017228 9 162 9 180 245 08 9275 1172 5 0 1043 1.2607 1 6650 276 0 200 0 43200 0 017264 8 627 8 644 24917 924 6 1113 8 0 4098 1.2501 1.6599 200.9 284 I 52 414 041130 8 1780 8 1453 253 3 921 3 1175 0 0 4154 12395 16548 284.0
288 8 55 795 0 01734 76634 16807 2574 918 8 1176 2 0 4208 1.2290 16498 284.0 292 8 59350 0 01738 72301 72475 261.5 915 9 1177.4 0 4263 1.2186 16449 292.0
2ts t 63 084-0 01741 6 R759 6 8433 265 6 9130 1118 6 04317 1.2082 16400 295.0
.
--
--
.
)
Os j
Abs Press Specific Volume
[nthalpy Entropy Temp Lb per Sal Sal.
Sal.
Sal.
Sat.
Sat.
Temp Iahs Sqin liquid Ivap Vapor Li vid Evap Vapor Liquid Evap Vapor Fahr sg s
t vs I
h ig h
sg e
g g
I p
vg vig
3000 61 005 0 01745 6 4483 6 4658 2693 910 0 11793 0 4372 1.1979 1.6351 300 0 304 0 71119 0 01149 6 0955 6 1130 273 8 907.0 1180 9 0 4426 1.1817 16303 304 0 300 0 15 433 0 01753 57655 5 7830 278 0 904 0 1182 0 0 4479 I.1776 1.6256 3000 312 0 19 953 001757 5 4566 54142 2821 901.0 11831 0 4533 1.1676 1.6209 312.0 310 0 84 688 0 01761 5 1673 51849 286 3 897.9 1184.1 0 4506 1.1576 1.6162 316 0 3NG 89 643 0 01766 48%I 49138 290 4 894 8 1185 2 0.4640 1.1477 1.6116 320.0 324 0 94 826 0 01770 4 6418 4 6595 294 6 891.6 1186 2 04692 1.1378 1.6071 324.0 320 0 100 245 0 01774 44030 4 4208 298 7 888 5 1187.2 04745 1.1280 1.6025 370 0 f
312 0 105 907 0 01779 4 1788 41%6 302.9 885 3 1188 2 0.4798 1.1183 1.5981 332.0 til 870 0 01183 3 96AI 3 9859 307.1 882.1 1189.1 0.4850 1.1086 1.5936 3360 3M I
.
340 0
!!7 992 0 01181 3 1699 3 7878 311 3 878 8 11901 0 4902 1.0990 1.5892 340.0 344 0 124 430 0 01792 3 5834 3 6013 315 5 875 5 1191 0 0 4954 1.0894 1.5849 344.0 340 0 131.143 0 01797 34018 3 4258 319 7 872 2 1191.1 0 5006 1.0799 1.5806 340 0 3520 138 138 001801 3 2423 3 2603 323 9 868 9 11923 0 5058 1.0705 1.5763 352.0 350 0 145 424 0 01806 3 0863 3 1044 329 1 865 5 1193 6 0 5110 1 0611 1.5721 356.0 300 0 153.010 0 01811 2 9392 2 9573 332.3 862.1 1l94 4 0 5161 1.0517 1.5678 360 0 304 0 160 903 001016 2 8002 2 8184 336 5 858 6 1895 2 0 5212 10424 1.5637 364.0 l
300 0 169113 0 01821 26691 2 6873 340 8 855 1 1195 9 0 5263 1.0332 1.5595 360 0 372 0 177 648 0 01826 2 5451 2 5633 345 0 8516 11 % 3 0.5314 1.0240 1.5554 372 0
)
370 0 106 517 0 01831 24219 24462 349 3 848I
!!97.4 0.5365 1.0148 1.5513 375.0 300 0 195 729 0 01836 2 3110 2 3353 353 6 844 5 1198 0 0.5416 1.0057 1.5473 300 0 3M I 205 294 001842 22120 2 2304 357.9 840 8 1198 7 0.5466 0.9966 1.5432 304.0 300 0 215 220 0 01847 21126
?t311 362.2 837.2 1199 3 0 5516 0.9876 1.5392 300.0 302 0 225 516 001853 20184 2 0369 366 5 833.4 1199 9 05567 0 9786 1.5352 392.0 MSI 236 193 0 01858 I929I I9417 370 8 829 7 I200 4 0.5617 0 % 96 I,5313 30s e 400 0 247259 0 01864 I8444 I8630 375.1 825.9 1201.0 0.5667 0.9607 1.5274 400 0 404 0 258725 0 01870 11640 17827 379 4 822 0 1201.5 0 5717 0 9518 1.5234 404.0 400 0 270 600 0 01875 16877 17064 383 8 818 2 1201.9 0 5766 0.9429 1.5195 400 0 412 0 282 894 0 01881 16152 I6340 3881 814.2 1202.4 0.5816 09341 1.5157 412.0 410 0 295 617 00lR87 1 5461 15651 392 5 810 2 1202.8 0 5066 0.9253 1.5118 dis t i
420 0 308 180 0 01894 14808 14997 3% 9 8062 1203.1 0 5915 0.9165 1.5000 420 0 424 0 322 391 0 01900 14184 14374 4013 802 2 1203.5 0 5964 0.9077 1.5042 424.0 420 0 336 463 0U1906 13591 13782 405 7 198 0 1203 7 0 6014 0 8990 1.5004 420.0 4320 35100 0 01913 130266 132119 4101 793 9 1204 0 0 6063 0 8903 1.4966 4320 4MI 366 03 0 01919 124RR7 1 76806 414 6 1893 12042 0 6112 0 8816 1.4928 436 0 440.0 381 54 0 01926 119761 121687 419 0 185 4 1204 4 0 6161 0 8729 1.4890 440 0 4440 397 56 0 01933 1.14874 1.16806 423 5 181.1 1204 6 0 6210 0 8643 14853 4440 440 0 414 09 0 01940 1.10712 1 82152 428 0 176 7 12043 0 6259 0 8557 14815 440 0 (~~5 0 43114 0 01947 1 05764 107711
~12 5 172 3 1204 8 0 6308 0 8411 1.4778 ( cit el 44873 0 01954 1 01518 1 03472 17 0 1678 1204 8 0 6356 0 8385 1 4741 GI
-
.
u
-
-
,
s.
Ahs l'ress Sperific Volume Enthalpy
[ntropy temp lb per Sal Sal Sal.
Sat Sat.
Sat.
Temp latu Sq lo l iquirl Ivan Vapor liquid
[vap Vapor Liquid
[vap Vapor Fahr I
p vg v ig _
vg ht hg h
sg sig sg I
i g
estI 466 87 0 01961 097463 0 99424 441.5 763.2 1204 8 06405 0 8299 1.4704 488.9 464 I 485 56 0 01969 0 93588 0 95557 446.1 758.6 1204.7 0 6454 0 8213 1.4667 444.0 468 I 504 83 0 01916 0 89885 0 91862 450 7 754 0 1204.6 0.6502 0 8127 1.4629 468.0 472 8 524 67 0019R4 0 86315 0 88329 455 2 749.3 1204.5 0.6551 0 8042-1.4592 472.0 475 8 54511 0H1997 0 87958 0 84950 459 9 744 5 1204.3 0 6599 03956 1.4555 476.0 408 8 566 15 0 02000 0 79716 / 0 81711
'464 5 7398 1204.1 0 6648 01871 1.4518 480.0 404 I 58781 0 02009 016613 018622 4691 734.7 1203 8 0 66 %
03785 1.4481 484.0 488 O 610 10 0 01011 0 13641 0 15658 413 8 7293 1203.5 0.6745 03700 1.4444 448.9 487 I 613 03 0 07026 0 10194 0 72820 478 5 724 6 1203.1 0 6793 0 7614 1.4407 492.0
'
496 8 656 61 007014 O f Ror.5 0 70100 4832 719 5 1202.7 0 6842 01528 1.4370 498.0 500 8 680 86 0 02043 0 65448 0 67492 487.9 714.3 1202.2 0 6890 01443 1.4333 588.8 5848 705 78 0 07053 Of>2938 0 64991 4923 709.0 12013 0 6939 01357 1.42 %
504.0 588 8 73140 0 02062 0 60530 0 62592 497.5 7033 1201.1 0 6987 03271 1.4258 500.0 512 0 15712 0 02072 0 58218 060289 502.3 698 2 1200.5 03036 07185 1.4221 512.0 5tl8 184 16 0 07081 0 % 997 058019 5071 6921 1199.8 07085 03099 1.4183 518.0 528 8 812 53 0 02091 0 53864 0 55956 512 0 687.0 1199 0 07133 01013 1.4146 529.9 524 8 841 04 0 02102 051814 053916 516 9 681.3 1198.2 0 7182 0 6926 1.4108 524.5 5288 870 31 0 02112 0 49843 051955 521 8 675.5 1197.3 07231 0 6839 1.4010 528 O 532 8 900 34 0 02123 047947 0 50070 5268 669 6 1196 4 0 7280 06152 1.4032 532.0 536 8 931.17 00?l34 0 46123 048757 5311 663.6 1195.4 0 7329 06665 1.3993 536 6 548 I 96279 0 02146 644367 0 46513 536 8 657.5 1194.3 0 7378 06577 1.3954 540.0 5448 99522 0 02157 0 476/7 044834 541 8 651.3 1893.1 03427 0.6489 1.3915 544.0 5488 1028 49 0 02169 0 41048 0 43217 546 9 645 0 1191.9 0 7476 0 6400 3.3876 548.0 5528 1062 59 0 02182 0 39479 0 41660 552 0 638.5 1190.6 0 7525 0.6311 1.3837 552.8 558 8 1097 55 0 07194 0 11966 0 401f>0 5572 632.0 1189 2 03575 0.6222 1.3797 556.9 548 8 1133 38 0 02207 0 36507 0 38714 562.4 625 3 II871 0 7625 0.6132 1.3757 568 8 584 8 1170 10 0 02221 0 35099 0 31320 567.6 618.5 1186.1 01674 0 6041 1.3716 5640 548 I 1207 12 0 02235 0 33141 0 35915 572 9 611.5 1184.5 0 7725 0.5950 1.3675 568.0 572 8 1246 26 0 07249 0 37429 0 34618 578 3 604.5 11823 0 7775 0.5859 1.3634 572.s 578 I 1285 74 002764 011167 0 33426 5837 597.2 1180.9 0 7825 0.5766 1.3592 576.8 Ses 8 1326 17 0 02279 0 29931 0 32216 589.1 589.9 1179 0 0 7876 0.5673 1.3550 500 0 Set t 1367 7 0 02295 0 28753 0 31048 594 6 582.4 1176.9 0 7927 0 5580 1.3507 584.0 588 8 1410 0 0 02311 0 27608 0 29919 6001 5143 1174.8 0 7978 0.5485 1.3464 588.0 e
592.0 1453 3 0 02328 0 26499 0 28827 6051 566 8 1172.6 0 8030 0.5390 1.3420 592.0
"
588 8 14978 0 07345 0?S425 027770 611.4 558.8 1170.2 0 8082 0.5293 1.3315 598.8
-
-
-
-
.
..
. _
O T
m
-..........
-. ~.. -
Temp th pe Sal.
Sal.
Sal.
Sal.
Sal.
Sat.
Temp Iahr Sq in liquid Ivap Vapor Li sid Evap Vapor Liquid Evap Vapor Fahr I
P
'J
' Ig_
't I
h Ig h
s, sg, s,
g g
see s 1543 2 0 02364 024384 026747 617.1 550.6 11673 0 8134 0.5196 1.3330 000A 804 0 15897 0 02382 0 23374 025757 622.9 542.2 1165.1 0.8147 05097 1.3284 484.0 000 0 16313 0 02402 0 22194 0 24196 628.8 533.6 1162.4 0 8240 0.4997 1.3238 608.0 812 e 16861 0 02422 0 21442 0 23865 6348 5243 1159.5 0 8294 0.4896 1.3190 612.0 Sit e 1735 9 0 02444 020516 0 22960 6408 515.6 1156.4 0.8348 0.4794 1.3141 888.0 01961! 022081 646.9 506.3 1153.2 0.8403 0.4689 1.3092 820 820 I s
!186.9 0 02466 824 8 1839 0 0 02489 0 18737 021226 653.1 4?6.6 1149.8 08458 0 4583 1.3041 824.0 828 8 1892 4 0 02514 017880 020394 659.5 4863 1146.1 0 8514 0.4474 12988 520A 532 8 1947.0 0 02539 0 17044 0 19583 665.9 476.4 1142.2 0.8571 0.4364 12934 632.0 838 0 2007 8 0 07566 0 16126 0 18792 672.4 465.7 1138.1 0.8628 0 4251 1.2879 SES 840 0 2059 9 0 02595 0 15427 0 18021 679.1 454.6
!!331 0.8686 0.4IM 1.2821 800.0 844 g 2118 3 0 02625 0 14644 017269 685 9 443.1 1129.0 0.8746 0 4015 1.2761 544.8 848 0 21781 002657 013876 0 16534 692.9 431.]
1124 0 0.8806 0.3893 1.2699 448.0 052 0 2239 2 0 02691 0 13124 0.15816 700 0 4181 1118 1 0 8868 03767 1.2634 852A 058 8 23017 0 02728 017387 015115 707.4 405 7 1113.1 00931 0.3637 1.2567 S$6A 000 e 2365 7 0 02768 011663 0 14431 714.9 392.1 1107.0 0.8995 0.3502 1.2498 000.0 est t 2431.1 0 028I1 0 10941 013757 722.9 3773 1100 6 0.9064 03361 1.2425 664.0 860 0 24981 0 02858 010229 0 13087 7315 362.1 1993.5 0.9137 0 3210 1.2347 080.0 812 8 2566 6 0 02911 0 09514 0 17424 740 2 3451 1085 9 0.9212 0.3054 1.2266 572.0 476 8 2635 3 0 02970 0 08199 011769 749 2 328.5 1071.6 0 9287 02892 1.2179 476.0 880 3 27086 0 03037 0 08080 0 11117 758.5 310.1 1068.5 0.9365 0.2720 1.2006 008.0 884 0 2782I 003114 0 07349 -- 0 10463 7682 290.2 1058.4 0.9447 0 2537', 1.1984 504.0 800 0 28574 0 03204 0 06595 0 09799 778 8 268.2 1041.0 0.9535 0.2337 l.1872 808.0 802 8 2934 5 0 03313 0 05197 0 09110 790 5 243.1 1033.6 0. % 34 0.2110 1.1744 802A 800 0 30I3 4 0 03455 0 04916 0 08371 804 4 212A 1017.2 0.9749 0.1841 1.1591 998.0 700 0 3094.3 0 03662 00385T 0 07519 822.4 1723 995.2 0.9901 0.1490 1.1390 700.8 782.8 3135.5 0 03824 0 03173 0 06957 835 0 1441 9791 1.0006 0.1246 1.1252 702A 704 0 3171.2 0 04100 0 0?!92 0 06300 354.2 102.0 956.2 1.0169 0.0076 1.1046 104.0
'
705 8 3198 3 0 04427 0 01304 0 05730 873.0 41.4 934.4 1.0329 0.0527 1.0856 705A 705 47'
3200 2 0 05078 0 00000 0 05078 906.0 0.0 906.0 1.0612 00000 1.0612 700s4F'
.
.
,
' Critical temperalute
]
g
'
- -
.
Table 2:
Saturated Steam: Pressure Table
..
.-.
Specific Volume Enthalpy Entropy Abs Perss.
Temp Sal Sat.
Sat.
Sat.
Sat.
Sat.
. Abs Press.
th/Sg in.
Fahr liqunt Ivap Vapor liquid Evap Vapor Liquid Evap Vapor tb/Sq In.
s p
v hg hg h
sg s gg g
v, g
p i
vi g
g i
880845 32 018 0 016022 3302 4 33024 0.0003 1075 5 1075.5 0 0000 2.1872 2.1872 0.00085 8 25 59 323 0 016032 1235 5 1235 5 27 382 1060 1 1087.4 0 0542 2.0425 2.0967 0.25 8 58 19 586 0 016071 641.5 641.5 47.623 1048 6 1096.3 0.0925 1.9446 2.0370 8.58
10314 0016136 33159 333 60 6913 1036I 1105.8 0 1326 1.8455 1.9781 1.8
16224 0 016407 13 515 13 532 130 20 1000 9 1131.1 0 2349 1.6094 1.8443 5.8 10 0 193 21 0 016592 38404 38420 161.26 982.1 1143.3 0.2836 1.5043 1.1879 10.0 14 598 21200 0 016719 26782
/26 199 180.17 970 3 1150 5 0.3121 1.4447 1.7568 14 GIS 15 e,
21301 0016126 26 214 26290 181.21
%91 1150.9 0.3137 1.4415 1.7552 15.8 20 0 22796 0 016834 20 070 20 087 196 27 9601 1156.3 0.3358 1.3962 I.7320 28.8 30 e 250 34 0 017009 13 7266 13 7436 218.9 945 2 1164.1 0.3682 1.3313 1.6995 30.0 48 8 267 25 0 017151 10 4794 10 4965 236I 933 6 1169.8 0.3921 1.2844 1.6765 40.0 58 8 281 02 0 017274 8 4967 8 5140 2502 923 9 1174.1 0.4112 1.2474 1.6586 50.0 st 8 292?!
0 017383 71562 7.1736 262.2 915 4 1177.6 0.4273 1.2167 1.6440 50 0 18 g 302.93 0017482 6 1875 6 2050 272.7 907.8 1180.6 0 4411 1.1905 1.6316 70.8 sta 312 04 0 017573 5 4536 5 4111 282.1 900.9 1183.1 0.4534 1.1675 1.6208 10.8 ge s 320 28 0 017659 4 8719 4 8953 2903 894 6 1185.3 0.4643 1.1470 1.6113 90.0 10e 3 32782 0 017740 4 4133 4 4310 298.5 888 6 1187.2 0.4743 1.1284 1.6027 108.8 lit e 334 79 0 01782 4 0306 4 0484 305 8 883.1 1I88.9 0.4834 1.1115 1.5950 110.0 128 8 34I 27 0 01789 37097 3 7275 312.6 877 8 1190.4 0.4919 1.0960 1.5879 120.0 130 0 347.33 0 01796 3 4364 3 4544 319.0 872 8 1191.7 0.4998 1.0815 1.5813 130 0 148 8 35304 0 01803 3 2010 3 2190 325 0 868.0 1193.0 0.5071 1.0681 1.5752 140.0 15eI 35843 001809 2 9958 30139 330.6 863 4 1194.1 0 5141 1.0554 1.5695 158.8 lie B 363 55 0 01815 28155 2 8336 3361 859 0 1195.1 0 5206 1.0435 1.5641 ISe O 178 8 368 42 0 01821 2 6556 26738 341.2 854 8 1196 0 0 5269 1.0322 1.5591 119.8
!ss.g 373 08 0 01827 2 5129 25312 346 2 8507 1l96 9 0.5328 1.0215. # 1.5543 100.0 les e 37753 0otR33 23847 2 4030 350 9 8467 1197.6 0.5384 1.0113 1.5498 190.0 298 8 381 80 0 01839 22689 2 2873 355 5 842.8 1198 3 0 5438 1.0016 1 5454 200.8 218 s 385 91 0 01844 2 16373 218217 359 9 839 1 1199 0 0 5490 0.9923 1.5413 tit t 228 I 389 88 0 01850 2 06779 208629 364 2 835 4 1l99 6 0 5540 0.9834 1.5374 228.8 238 8 393 10 0 01855 I97991 199846 368 3 831 8 1200.1 0 5588 0 9748 1.5336 230 0 240 8 39739 0 01860 189909 I91769 372.3 828 4 1200.6 0 5634 0.9665 I.5299 248.8 250 8 40097 0 01865 1 82452 1.84317 376.1 825 0 1201.1 0 5679 0 9585 1 5264 254 9 288 8 404 44 001870 115548 117418 379 9 821 6 1201.5 0.5722 0.9508 1.5230 268.0 210 8 40780 0 01875 169137 I11013 383 6 818 3 1201.9 05764 09433 1.5197 218 8
- >
280 0 411 01 001880 163169 I65049 387.1 815 1 1202 3 0.5805 0 9361 1.5166 200.0
20e g 414 25 0DIRR5-157597 I59482 390.6 812 0 1202.6 0 5844 0 9291 1.5135 298.8 30s 8 417.35 0 01889 I52384 154274 394 0 808.9 1202.9 0 5882 0 9223 1.5105 300.0 350 0 43113 0 01912 1 30642 1.32554 409 8 194.2 1204.0 06059 0 8909 1.4968 358.0 448 I 444 60 001934 1 14162 1.16095 424.2 780 4 1204 6 0.6217 0.8630 1.4847 480.8
.
..
..
.
.
.
_______
___
.
... _. _
__
-
--
- - - - -
Abs P ess Temp Sal Sal.
Sal Sal.
Sal.
Sal.
Abs PvGss.. s I"
lb/Sq in Fahi lialuiil Ivan Vapor liquid Evap Vapor Liquid Evap Vapor tb/Sg in.
.
v hg hgg h
s, s,,
s, p
v, g
g 2;
p I
vi i
45s 0 456 28 0 01954 101224 103119 437 3 767 5 1204 8 06360 0 8378 14138 458 8 500 0 46101 0 01915 0 90181 0 92162 449 5 7551 12043 0 6490 0 8148 14639 500 0 5500 416 94 001994 0 8/183 0 84111 460 9 143 3 1204.3 0 6611 0 1936 14547 554 8 Ef;8 3 486 20 0 02013 0 14962 0 16915 471 7 132 0 12033 0 6723 0 7738 1 4461 see t 850 8 494 89 0 02032 0 68811 0 10843 4819 720 9 1202.8 06828 07552 14381 558 8 III e 50308 0 07050 0 63505 06 556 4916 110 2 1201.8 06928 03377 1.4304 70s a 158 I 510 84 0 02069 0 58880 0 60949 500 9 699 8 1200 7 03022 07210 1.4232 758 8 808 I 518 21 0 01081 0 54809 0 56896 509 8 6896 1l99 4 01111 07051 14163 see a 850 8 525 24 0 02105 051191 0 53302 518 4 619 5
!!98 0 0 7197 06899 I4005 35e 8 998 8 53195 002123 0 41968 050091 526 7 669 7 1196 4 01279 06753 1.4032 See t 958 e 538 39 0 02141 0 45064 0 41205 534 7 660 0 11941 03358 06612 13970 354 8 1948 0 544 58 0 02159 042436 0/ 4596 542 6 650 4 1192.9 0 7434 0 6416 1.3910 leet s 1858 I 550 53 0 02111 0 40041 0 42224 5501 6409 1191 0 01507 0 6344 13851 less e IIIIO 55628 0 02195 0 31863 040058 557.5 631 5 1189.1 0 7578 0.6216 1.3794 1100 8
'
1150 8 56182 0 02214 0 35859 038073 564 8 622 2 1187.0 03647 0 6091 I3738 Ilse e itse 3 56119 0 02232 0 34013 0 36245 571.9 613 0 1l84 8 03714 05%9 I3683 1200 8 1758 8 57238 0 02250 0 32306 0 34556 578 8 603 8 1182.6 0 1780 0 5850 13630 I250 8 1398 I 577.42 0 02269 0 30722 0 32991 5855 594 6 1180 2 0 7843 05733 1.3577 1388 8 13540 582 32 002288 0 29250 031537 592 3 585 4 1177.8 07906 0 5620 13525 1350e 1400 0 58707 0 02301 0 2/811 0 301/8 598 8 516 5 1175 3 03966 0 5507 1.3414 1400 8 1458 8 591 70 0 02321 0 26584 028911 605 3 5674 1172.8 0 8026 0 5397 13423 145es 1500 I
$96 20 0 02346 0 25312 0 21119 611 3 558 4 11701 0 8085
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TYPICAL TURBULENCE Niagara Mohawk ASSOCIATED WITH OVERCAST-S10 Rov. 0R NOC1UR m SITUATIONS HAVING RELA-Cl8$$3 TlVELY STRONG WINDS.
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FIGURE 1 i
RECOMMENDED PROTECTIVE ACTIONS FOR GENERAL POPULATION AND EMERGENCY WORKERS
l l'
l l
General Population
__
l PROJECTED DOSE (REM) TO RECOMMENDED ACTIONS (a)
COMMENT S THE CENERAL POPUIATION_
No planned protective actions. (b)
g, Previously recommended protective
Whole Body
<1 State and County may issue an advisory to p ek actions may be reconsidered or terminated.
Thyroid
<5 shelter and await further instructions.
(y Monitor environmental radiation levels.
(Child)
Shelter all ERPA's in affected sectors outi,b 11 constraints exist, special Whole Body 1 to <5 toapointwherethedoseis<1remwhole7 considerations should be given for body and 5 rem thyroid. Consider evacuation evacuation of children and pregnant of children and pregnant women in this areas women.
Thyroid 5 to <25 Consider evacuation out to 2 miles radially (Child)
from the station. Evacuate this area unless constraints male it impractical.
,
Monitor environmental radiation levels. Control access.
[
~
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Seeking shelter waald be an
,
alternative if evacuation were radially from the station and evacuate
Thyroid 25 and above all ERPA's in affected sectors out to the @
not immediately possible.
"
point where dose is <5 rem whole body Where possible, dose savinga from (Child)
and <25 ren thyroid.
sheltering in lieu of or in
'
combination with evacuation should be
,
Shelter all ERPA's in affected sectors considered.
out to a point where the dose is <1 rem whole body and 5 ren thyroid. Consider evacuation of children and pregnant women in this area.
Monitor environmental radiation levels and
,
adjust area for mandatory evacuation based on
"
.
these levels. Control access.
EPP-26-6 June 1985
%
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.
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5.3.2 Plume Exposure Zone (0-10 miles) Protective Actions (Cont. )
b.
(Cont.)
(4 For very turbulent and constantly changing wind conditions, cmsider recommending protective actims for all IRPA's out to a distance radially from the plant where the dos,e is 1 rem whole body and 5 rem thyroid in affected sectors.
NOTE:
Protective action recommendations will be made for entire ERPA's even though only a portion of that ERPA.
may be affected.
f4 c.
Repeat this procedure for each applicable ERFA (those ERPA's affected by the plume).
d.
Inform State and County Emergency Operation Centers of protective action recommendations', ' dose projections and sample l4 results using EPP-20 and its associated Fact Sheet.
5.3.3 Ingestion Exposure Zone (0-50 miles) Protective Actim s a.
Calculate the deposition rate for the area of concern using l4
]
EPP-8 (projected) and/or EPP-7 (environmentally determined via sampling) and record in Figure 4, Item 18 or 19.
(
b.
Compare the calculated deposition rate for each affected ERPA l 4 with the appropriate preventive or emergency protective action
'
guide levels listed on Figure 7.
" Preventive PAG's"- Projected dose commitment values at which recommendations-should be made to responsible officials. These actions should have minimal impact to prevent or reduce the
,
radioactive contamination of human food or antmal feed.
" Emergency PAC's"-
Projected dose commitment values at which recommendations should be made to responsible officials to isolate food containing radioactivity and thus prevent it's introduction into commerce.
f4 c.
Based on the comparison performed above, use Eigure 7 to determine recommended ingestion zone indicated protective BCtions.
l
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Repeat this procedure for each applicable ERPA or location affected by the plume.
EPP-26-5 June 1985
.
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.
Figure 3 (cont.)
1984 PERMANENT RESIDENT POPUIATION ESTIMATES EMERGENCY RESPONSE PLANNING AREAS EMERGENCY RESPONSE 1984 PERMANENT RESIDENT PIANNING AREA POPUIATION ESTIMATES
137
508
356
620
411
880
818 573
^ 455
10 1,117
?
1,423
9,145
11,238
122
1,028
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1,692
577
1,048 l
1,003
1,553
2,157
6,488 TOTAL 43,349 i
.
a EPP 26-15 October 1984
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Figurn 3 (Cont. )
TREE 2
ENWune rY EEEPONSE PLANNING AREES_
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10 uTT.ma l
28, 29 E
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10-Mile Radius Nine Mile Point Nuclear Power Stations
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1.
Area of Concern (
location /ERPA la.
miles miles (use Table 2
-
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Distance of Figure 3)
b.
Direction degrees
%
c.
ERPA NOTE:
If determining Ingestion Zone Protective Action Recommendations only, proceed to Item 18.
2.
hours 2.
Expected release duration
3.
miles /hr 3.
Wind Speed =
aph 4.
Plume travel time =
(Item la)/(Item 3)
4.
hours (
)/(
)
=
_
5.
Time until exposure begins (a or b)
5.
hours
,
(a or b)
If release has begun:
a.
(
(
Time = Item 4 - Time release has been in progress (
Time =
~
If Item Sa is a negative number, enter zero hours for Item 5 NOTE:
b.
If release will begin later:
l
Time = Item 4 + Time until release j
Time =
+
6.
Weather condition and season (circle onc for a & b ):
a.
Normal or Adverse l
b.
Day or Night 7.
_ hours 7.
ERPA Evacuation time:
Use Figure 3 along with information recorded in Items 1 and 6 to determine the time.
shown on Evacuation Time Estimates (Figure 3) are Times CAUTION:
expressed as hours: minutes and aust be converted into hours.
.
EPP-26 -32 June 1985
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...__...______....
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.
ACTIONS INSTRUCTIONS THEN continue in this pro-C6-3./
WHEN all controls rods are cedure.
E 6-3.1 THserted to at least position 02,+O E
AN6no'
The Rx is shutdown, Tdd into boron nas been injec
'
2e RW. +O l
I THEN connence and raise l C6 4 IF RPV water level cannot be C6-4 T6Tection into the RPV l
W termined,-eO with the foilowf ng systers until at least 3
SRVs are open AND RPV pressure is not Eering j
!
AND is at least 80 psig
-
^
E5ve suppression chareer
pressure:
4.1 HPCS, (.,
.
4.2 Fee &ater pumps,
.
-
i I
4.3 LPCS,
,
4.4 LPCI, I
.
.;
l 4.5 Condensate booster pumps,
.
.
4.6 Condensate pumps, 4.7 CRD,
4.8 Service water to RHR,
,
(0P-11, Section H},
!
.
4.9 Fire System, (0p.43, Section H),
l 4.10 ECCS keep full,
(OF-32, 33, Section H),
4.11 SLC test tank,
(OP-36, Section H),
(
(OP_36, Section H).
s AND ou au u.
- Fl
- - - - -
.
()
.cr
_
>
(
'
,
____.._--______ _-___.. _______.-___._____..__...__-____............ __.
(
_ _______..._____.._________
.. ____ __
______.______.____________
l Paintain at least 3 SRVs open and RPV pressure at least 80 psig G
above suppression chamber pressure by throttling injection.
C6 5
'
ACTIONS
INSTRUCT IONS C6-6.
THEN commence and increase
.
j C6 6.
IF RPV water level can be THIiction into the RPV determined,->O wi th the following sys-
,
teris until RPV watcr
!
level is increasing:
6.1 HPCS, O
i f
6.2 Fee &ater pumps, O
~6. 3 LPCS,0
.
l 6.4 LPCI,' O 6.5 Condensate pumps O (
Condensate booster pumps, O b.6
,
k f
6.7 CRD,C 6.8 RlR Service Water Tie (OP-11, Section H), C
,
6.9 Fire System (OP 43, Section H),D 6.10 ECCS keep full (OP-32, 33, Section H), O 6.11 SLC test tank (OP_36, Section H),0 6.12 SLC boron tank (O P-36, Section H). O p
A
.
(
_
H2 ev rf Page ' cf 5 Rev. uu ry. l 19e-
-
. _..
.
i l
'
.
l 1 VESSEL PRES RESE toest
9 NEUTRO4 F LUM 2 RELIEF W ALVE FLOW 2 AVE SUHF ACE HE AT F lux 3 SVPASS VALVE FLOW 3 CORE INLE T F LOW 4 OlF FUSER FLOW t t%9 4 COHE INLE T Sus s Dif FU$ER FLOW 7 (%f
'
O
4 l
100
s l
'
O B
9-
-
y
~
rs
-
"
'
w
l g s0 N
,
Q
_
.
w
\\
l
!
!
!
-2s O
to
30
so
10
30
go O
TIME isect flME (sect
'
8 VOIDREACTivtTV
~
t LEVEL tmch sel s*D thertl 2 DOPPLER RE ACTevlTY
.I 3 SCRAM RE ACTivlTY 2 VESSEL STE AMF tow 3 TURBINE STE AMF LOW
4 TOT AL RE ACTIVITY 4 F E EDW AT E H F t OW g
3 im
-
3
,
-
,
s0 -
,
.
'
E
4
2
!
!
!
l l
_2
10
30
50
10
30
M
TIME feect TIME (sec)
FIGURE 15.3-2 TRIP OF BOTH RECIRCULATION PUMP MOTORS
.
NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
-
."
---_-_EORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
TH
-- ------
----- -----
3.
-_
IHERgggyNgglCg ANSWERS -- NINE Mll.E POINT 2-86/07/21-CRESCENZO, F.
"
,.s.
-
ANSWER 5.01 (2.50)
a. As the reactor operates during the early part of the cycle, the burnable poison depletes more rapidly than the fuel, therefore, control rods must be inserted to hold the power constant.
(1.00)
bo As the control rod density increases, the power producing regions of the core become more undermoderated; the moderator to fuel ratio is decreasing. In effect, as total power production has remained the constant but the power producing volume has become smaller, operating volume of the core has become undermoderated. Because of this (1.50)
,
effect, the void coefficient becomes more negative.
,
REFERENCE NMPC Operations Technology vol. I page 5-15, LO 2-5 FJC 15 Hope Creek LP RXPH19-01 pg.6 LO #4, RXPH28-01 pg.6 and trans. #1 LO #3 RXPH16-01 trans. #2
_
ANSWER 5.02 (1.00)
(Theperiod during power increasqs is governed by how quickly the neutron population can increase.) The same holds true on a power decrease however, the neutron population is dominated by the longest. lived delayed neutron prec6Fsor.(This decays with -80 sec. period.)
(1.00)
REFERENCE Millstone Reactor Theory pg. 3-45 FJC 67 Pilgrim Reactor Theory pg. 3-45 Nine Mile 2 Reactor Theory pg. 3-45, L.O.
5.6 ANSWER 5.03 (2.50)
ao Installed sources are used to raise flux levels in the core to a point where it is on scale for the nuclear instrumentation. Initial intrinsic source levels are not high enough to bring the instrumentation on (1.50)
scale.
b. Yes, criticality would be achieved with no change in critical (1.00)
rod density.
REFERENCE NMPC Operations Technology pg.
2-3, Theory L.O.
1.1 FJC 68
_.
FLUIDS _AND PAGE
5.
THEDRY_OF NUCLEAR POWER PLANT _ OPERATION
t THERMODYNAMICS
_
-.---- --
-
ANSWERS -- NINE MILE POINT 2-86/07/21-CRESCENZO, F.
-
g
-
ANSWER 5.04 (3.00)
_
a.
Peripheral rod worth will increase [O.33 because the highest xenon concentration will be in the center of the core [O.33 where the highest flux existed previously [O.33.
This will suppress the flux in the center of the core [O.33 and increase the flux in the area of the peripheral rods, thereby, increasing their worth [O.33.
b.
More than one half the value at 1007. power [O.53.
The Xenon production rate is directly proportional to power i
i level, but removal rate is proportional to Xenon concentration and it contains a power dependant term, thermal neutron flux.[O.50]
Since flux is directly proportional to power level the burnout term becomes less significant. This results in an equilibrium Xenon value which is lower than the original equilibrium value but greater than one half the original concentration.[O.503 REFERENCE
Millstone Rx Theory ppg.
6-7, 6-12.
FJC 78 Pilgrim Rx. Theory pgs.
6-7, 6-12 Nine Mile 2 Rx. Theory pgs.
6-7, 6-12 L.O.
2.5.1,4
.
ANSWER 5.05 (2.50)
i With control rods A-1, B-1, B-3, and C-2 withdrawn, and B-2 still fully inserted, the " effective" core consists of four small four-bundle reactors. Each is essentially uncoupled because neutrons from one can't pass through the inserted control rod. By withdrawing the control
.
rod, 16 bundles are immediately coupled into one 20 bundle reactor.
Thus the size of the control rod is effectively much larger than (2.50)
its actual physical size would suggest,.and its rod worth ig large.
I1 tw k Sit % d c ~-
N ek s s s s M I,3
~
LL o i s.
o o s v.,
c REFERENCE Millstone Rx Theory pg. 5-17 FJC 80 Pilgrim Rx theory pg. 5-17 Nine Mile 2 Rx Theory pg. 5-17, L.D.
1.6 I
l
.
AND PAGE
5.__THEggy_gF NUCLEAR POWER PLANT OPERATIgN, _ FLUIDS, _
IHEggggyNQUICS a,
ANSWERS -- NINE MILE POINT 2-06/07/21-CRESCENZO, F.
,,
...
'
i
_3 u-<
- .,
,,
e
,e ANSWER,
5.06 (2.00)
p, ', m.. n,
.l IN
'
g
-
e p:
.\\.. x
, < _
c. 3 y s. t
.Under severe degraded core conditions with vessed pressure remaining high, and the core uncovered, temperatures significant1y' higher'th,ag,,
those calculated from the constant enthalpy line indicate the (2.00)
presence of superheated steam.
REFERENCE Hope Creek M.O.C.D.
LP 104 pg. 15 1.0.
- 1 FJC264 Nine Mile 2 MOCD pg 7-8, S.L.O.
111.2,3 ANSWER 5.07 (3.00)
'\\0 ' ' ~ '\\ k s*
f}) u,
'~
"r
'
.- ~
'
'
'
-~
'
a.
FW temp o
...
<.
>
<
-
+
,.'mn FW flow
,,,
e,,
RPV pressure
~
RPV level (3 reqd @ O.5ea)
(O.5)
b.
High flow high power c. High flow high power, due to increased inlet subcooling from increased (1.00)
FW flow.
REFERENCE Pilgrim HTFF pg. 6-81 FJC 292 Nine Mile 2 HTFF pgs. 6-77, 81 L.D.
10.10 ANSWER 5.08 (2.00)
a-1, b-2, c-3, d-4 0.5 each REFERENCE FJC 293 Pilgrim reactor theory pg 3-14 Nine Mile 2 reactor theory pg. 3-14, S.L.O.
2.3,4 i
5.
THEORY pf NUCLEAR POWER PLANT OPERAT ION FLUIDS _AND PAGE
t t
THERMOBfRAMICS
---
..,
ANSWERS -- NINE MILE POINT 2-86/07/21-CRESCENZO, F.
ANSWER 5.09 (2.00)
l l
(0.5)
a.
Increases
(0.5)
b.
Decreases (0.5)
c.
Decreases (0.5) -
d.
Increases
.
REFERENCE HC HT&T No. 11, Learning Objective 2, pages 8 and 9.
FJC 294 Pilgrim THT&FF pgs. 9-26 thru 9-30 i
I Nine Mile E THi&FF pgs. 9-26 thru 9-30.
~'
pov-c-
ANSWER 5.10 (2.50)
%
a. Reactor level increase due to loss of recirc suction from annulus and increased voiding in core region. Increase in level causes turbine trip which in turn creates pressure spike to SRV opening. (1.50)
b.
As reactor pressure increases, core inlet temperature remains essentially constant thus inlet subcooling increases and
.
j
,
'
qsry q,,bw
,. ; ( o s g t.g j, -
,q?uf decrpases as a function of rx pressure.
cJ lj
,.4 (1.00)
i, u. },t \\,s e g_ n\\
-
,,
, G cN $~ 4 c
\\,s
+
- s
REFERENCE NMP2 FSAR Chap 15 vol 27 FJC 346 s
ANSWER 5.11 (2.00)
Ocore = MG x HG + MCU(HCU,IN - HCU,0UT) + Of1 - MFW x HFW - Op (1.00)
- MCRD x HCRD Qcore = 6423000 x 1194 +110000(506-419) + 2040000 - 6400000 x 345-26500000 - 23000 x 68 = 5444600000 Btu /hr = 1595 MWt (1.00)
REFERENCE Vermont Yankee Nuclear Power Corporation, SCRO-02-115, Reactor Heat Balance FJC 347 NMP 2 Thermo chap 8
.
,..
.
--
_
._
6.
PLgNI_glgl{gg_Q{glGN, CONTROL, _AND_ INSTRUMENTATION PAGE
__
ANSWERS -- NINE MILE POINT 2-86/07/21-CRESCENZO, F.
l l
l ANSWER 6.01 (3.00)
l
!
a. The system sees low feed flow and will increase feedflow until level error balances the feedflow/steamflow mismatch at which l
point level will stabilize at some new higher level.
(1.00)
b. The system sees low steam flow and will decrease feedflow until I
level error balances the feedflow/steamflow mismatch at which point l
level will stabilize at some lower level.
(1.00)
c. Feed flow will increase and continue to increase until the reactor I
l high level trip.
(i,e\\,;ts t)
(1.00)
l REFERENCE NMPC Operations Technology FWCS pg 4 EO #6 FJC 35
\\
.
l lj ANSWER 6.02 (1.00)
l d
i g
I REFERENCE l
Hope Creek LP 14-01 pgs.
9,
I.O.
4.b FJC214 I
l Pilgrim LP IRM pg. 7 Nine Mile 2 L.'P.
IRM's 64 6 and 10 of 12.
ANSWER 6.03 (3.00)
a.
TGV remain at 100% due to load limit (.50)
l b.
HPV open 5% due to max combined flow (.50)
c.
Power decreases due to lower pressure (.50)
d.
Pressure decreases due to BPV (.50)
FINAL a.
TCV at 100% position (.25)
b.
UPV shut (.25)
c.
power lower (.25)
d.
pressure slightly lower (.25)
REFERENCE l
Hope Creek LP 1.0.
Nine Mile Point 2 EHC L.P.
objective EO-6 l
I l
-
-.
6 __ PLANT _ SYSTEMS _DESIGNt__ CONTROL 1__AND INSTRUMENTATION PAGE
3 ANSWERS -- NINE MILE POINT 2-86/07/21-CRESCENZO, F.
ANSWER 6.04 (2.00)
+
a.
Back Up Hoist Limit:
This lamp lights only if the normal maximum (0.5)
up limit limit fails and the hoist is stopped by the backup haiq%
?
\\im' " ( u. '. b bc,5 13 d i.)
1imit switch.
( vi i
w m 'i t
b.
Rod Block Interlock No. 1: Occurs when a fuel assembly load is on (0.5)
any hoist and refuel switch #1 is activated when the refueling platform is over the vessel.
p hs o r c.
Fuel Hoist Interlock: Indicates a condition when the platform is (0.5)
over the reactor, a control rod is withdr; awn, and the grapple is loaded.
t-0 I
'
'd O'
' ' ' '
'
d. Bridge Reverse Stop No. 1:f Prohibits bt idge travel toward t e (0.5)
reactor when a signal from the control room indicates that a control rod is withdrawn, the platform is on a switch indica ing that the platform is about to move over the reactor, and a load is on any of the hoists.
(
_
r,.. wr r.
REFERENCE GE Refueling Tools Familiarization Manual
- 2, pages 3-12, 3-13 FJC 312 NMP2 L.P.
S.L.O.
- E.O.4
'
~
l ANSWER 6.05 (1.50)
j A differential pressure cell compares the pressure as sensed in the HPCS header and from the above core plate pressure tap. If a break were to occur in the vessel, but outside the shroud, a differential pressure would exist because of the p(essure drop across the steam s h c., d )
(1.50)
separators.
( cv gw s.f 4.s y
'
REFERENCE NMP2 Oper Tec, HPCS rev 2 pg 9 of 10, EO 5-3 FJC 320 ANSWER 6.06 (2.00)
a.
Meters fail downscale. Recorders fail as is.'
(1.00)
b.
ECCS will not initiate because logics are " energize to operate" (1.00)
' '
\\T
'u
't
\\,
.t e
a
\\ v 4 &. vs.
L
.
.
-g (v s m.(, 3..m) x q
'
en
- < (. 5 m..
.
.
s
..,--
-
..
-
-
-
6.
PLANT SYSTEMS _ DESIGN, CONTROL, AND INSTRUMENTATION PAGE
ANSWERS -- NINE MILE POINT 2-86/07/21-CRESCENZO, F.
REFERENCE NMP2 MOCD pg 75 of 80 E.O.
- 's 1,2 FJC 348 ANSWER 6.07 (2.00)
see attached diagram.
REFERENCE NMP2 APRM LP Pg. 4 of 14 E.D.
FJC 349 1,k ANSWER 6.08 42rBO)
g
} ( (
a.
TRUE
\\
-M'-TRUE -----,
r).
<
-
c.
TRUE d.
TRUE e.
FALSE REFERENCE NMP2 RWM LP pgs. 3,5 of 26 and table 2 E.O.
FJC 350
.s
-.4 ANSWER 6.09 (3.00)
,
'*
a. DG bkr trips on LOCA(diesel goes to emergency mode,. when LOOP f
occurs, breaker closes and(loads sequence as normai (1.50)
b.
The offsite breaker will stay closed and diesel will attempt to pick up offsite test loads. Directional current trip will open offsite breaker and isolate bus with EDG (1.50)
REFERENCE NMP2 FSAR pg. 8.3-18c LP E.O.
- 's 3,4 FJC 351
lh 6 __P(ANI_@y@IED@_@@@l@dt_@@dIB@67 AND INSTRUMENTATION PAGE
3 ANSWERS -- NINE MILE POINT 2-86/07/21-CRESCENZO, F.
.
i h
ANSWER 6.10 (2.00)
a. Assuming 95A open, 94A closed in' auto. Valve 94A will samuL open due to EDG east running Valve 95A wi11 remain open.
(1.00)
b.
Valve 94A will open as expected, once diesel is running.
after 1 minute time delay with service water pressure less than 20-psi valve 95A will close.
(1.00)
- ---. ;..
,
I ~
REFERENCE nmp2 FSAR figure 9.2-2 vol 18; service water LP EO 4.
FJC 352 l'
ANSWER 6.I1
'(3.~dDI
'
I ao By cross connected the LPCS suction and RHS SD cooling suction l
via isolation valves and a removeable spoolpiece. This is done to allow testing of the LPCS system with a suction from the reactor vessel.
(1.50)
l
--trl.-to prevent-excessive-Dh' across tWe val ~ve.
0.95)-
\\
'
l
/~2.' To-ensure -the-reactoNoeh not draindadhe suppress 1 anr (Os-75) -
L)A y O P \\
-
REFERENCE l
- 's 2,9 FJC 353 l
l l
. l-
I I
i l
I
.
_
7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
RAD 19LggICAL CONIBQL ANSWERS -- NINE MILE POINT 2-86/07/21-CRESCENZO, F.
ANSWER 7.01 (1.50)
(0.75)
Health physics must sample the atmosphere in the containment.
1.
2.
The initiating condition must be corrected.
(O.75)
REFERENCE FJC 41 i
N2-IOP-83 PCIS pg. 3 i
l i
ANSWER 7.02 (1.50)
(O.50)
a.
High neutron flux alarm and/or scram.
b.
At or near 1007. rod line, min recirc flow.
(0.50)
c.
Insert control rods per reactor analyst or increase recirc flow
'1.00;
.
(0.5 REFERENCE N2-OP-92 Neutron monitoring, precautions and off normal procedures.
FJC 47
.
ANSWER 7.03 (2.50)
a. Downwind direction will be NE. Table 2 of EPP-26 lists no ERPA within the 5 mile zone. ( mq +tei i3 MA a '3mmi d e a. 2- - % C-( e m * M 1(1.50)
b.
The' lower number is the point at which the P.A.
should be made especially for sensitive populations. The higher number indicates 44,,*iO)-
the P.A.
is mandatory.
( Li (C )
REFERENCE FJC 302 NMPC procedures EPP-8 and 26 ANSWER 7.04 (2.50)
l While controlling RPV power with water level, very little baron mixing
'
occurs. When the hot shutdown baron weight is injected, water level is increased to promote natural circulation which will mix and distribute (2.50)
the stratified baron.
.. -_
.-.._.
.
._.
. - _
.. _
,
ABNORMAL _ EMERGENCY AND PAGE
7.
PROCEDURES _ _NORMAlt t
RAD _IOLOGICAL CONTROL
___.
ANSWERS -- NINE MILE POINT 2-86/07/21-CRESCENZO, F.
I
!
REFERENCE N2-EOP-C7 LP pg 19 of 23.
FJC 303 Mh,
%
ANSWER 7.05 (3.00)
a. BX is the power at which a reactor will stabilize during a full power ATWS and level is low enough to inhibit natural circulation.
(flow stagnation power level). Further level reduction will result (2.00)
,in uncovering the core.
b.' Minimum RPV pressure at which steah flow through open SRV's is sufficient-to remove decay heat from a completely uncovered core by steam" heat transfer alone after a scram from 100% power.
'with no clad temperature _in excess of 1500 F.
(1.00)
mis_,
e t
r.
"'n
.. -
-, i j,
,s
.,
,
-
.
...
.
,
.
REFERdNCE N2-EOP-C7 LP pg. 5 of 23 and C6 LP pg 7 of 19.
FJC 304
?"
.
ANSWER 7.06 (3.00)
.
..
1.
Submergence-water level > TAF (1.00)
2. Spray cooling-one core spray at or above design conditions.
(1.00)
3. Steam cooling-proper pressure and steam flow through SRV's.
(1.00)
REFERENCE FJC 305 NMP2 EOP structure page 3-1.
ANSWER 7.07 (1.00)
Highest value of a parameter beyond which operation of 3,3 p \\
'
(1.00)
equipment impagtant to safety cannot be assumed.
(;<
7c, i(.(e j A
is h,
Ok 4-
,, o -
s g.
, w m. s :-
s oCt* 2 REFERENCE FJC 306 NMP2 EOP Structure page 3-5
' p ~~
PAGE
e
~
PRgCEDURES __ NORMAL, ABNORMAL EMERGENCY _AND
-
t
ROpig6991Cgh CONTROL ANSWERS -- NINE MILE POINT 2-06/07/21-CRESCENZO, F.
ANSWER 7.08 (2.00)
Lowest differential pressure at which steam flow through the minimum number of SRV's required for emergency depressurization (3) is sufficient to remove decay heat via boiling heat transfer 10 min.
(2.00)
after scram from 1007. power.
' REFERENCE N2-EOP-C6 LP pg 13 of 19 FJC 307 ANSWER 7.09 (2.00)
a.
To prevent cavitation of the flow control valve.
(1.00)
b.
To prevent cavitation in the the jet pumps.
(1.00)
REFERENCE NMP2 HTFF pg. 6-81 FJC 311
-
NMP2 10P-29
,
'
-
,..
ANSWER 7.10 (2.00)
a.
To provide a backup to the normal limit switch to ensure sufficient water shielding.
(1.00)
b.
Do not hold load at or near zero speed for more than 90 sec to-I prevent motor overheating.
(1.00)
REFERENCE NMP2 procedure N2-OP-39 FJC 317 ANSWER 7.11 (1.00)
Both valves are powered from the same power source.
(1.00)
REFERENCE N2-IOP-62 FJC 319
_ _ _.
7.
PROCEDURES _b_ NORMAL _.. ABNORMAL, EMERGENCY AND PAGE
t g@DIOLOGICAL_ CONTROL ANSWERS -- NINE MILE POINT 2-86/07/21-CRESCENZO, F.
.
ANSWER 7.12 (3.00)
.ms a.
Thermal stratification as noted by axial skin temperature variations..
(1.00)
b.
Raise level to above the steam' separators (202") for natural
circulation or maintain SDC flow normal and throttle service water.
(2.00)
!
I'
REFERENCE NMP2 N2-IOP-101c pg. 6 FJC 354
,
. - -.
,
_._
-
--
,c-
-
--.
v Ei__gDglNigIggIlyg_ PROCEDURES, _ CONDITIONS, _AND_ LIMITATIONS PAGE
ANSWERS -- NINE MILE POINT 2-86/07/?lI CRESCENZO, F.
ANSWER B.01 (3.00)
a.
3.3.9.b Plant Systems directs operator to table 3.3.9.1.
action 145 to lock closed the service water valves to the HPCS diesel and take action for inop EDG-2.
T.S.
3.8.1.1.d.
allows 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> till declare HPCS inop. Since EDG-1 is inop, take action of 3.8.1.1.i which directs operator to take actions of 3.8.1.
1.b,d,e which allows 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Must perform surveillance of offsite power sources and of EDG-3.
(2.00)
b.
Once HPCS declared inop (72 hrs), then operator would attempt to apply 3.5.1.
which in this case would not apply since division 1 (LPCS)
is inop. Apply t.s.
3.0.3.
(1.00)
REFERENCE NMP2 T.S.
3.4.2, 3.5, 3.0.
(FJC 308)
.
~.
ANSWER 8.02 (2.00)
3.5.1.
14 day LCO. 373 r7.- 4 r-R7S. P -7-d ay - LCO (2.00)
(3.4.2.
for safety valve does NOT apply)
,
REFERENCE NMP2 T.S 3.4.2, 3.5.1, 3.3.7.4 FJC 309
.
ANSWER 8.03 (2.50)
T.S 3.7.9 to take action of 3.2.3.
when > 257..
T.S.,3.2.3.
limits
>257. power.(T.S.
MCPR to limits of figure 3.2.3.~1 when 3.0.4 is not appl,icablehtherefore can continue power increase [until limited
'
by MCPR.;
s (2.50)
REFERENCE NMP2 T.S.
3.2.3, 3.7.9 FJC 310
\\
.
.
.
.
s PAGE
8 __ODMINigIggIlyE_PBggEgySEgz_ggdpillgNyt_@NQ_LIMil@IlgNy t
ANSWERS
- NINE MILE POINT 2-86/07/21-CRESCENZO, F.
.
ANSWER 8.04 (2.50)
T=FRTP/CMFLPD if T is less than 1 then APRM setpoints are lowered by a factor of T. The setpoints are lowered when the combination of thermal power and CMFLPD indicates a higher peaked power distribution to ensure that an LHGR transient would not be increased in the (2.50)
degraded condition.+
REFERENCE FJC 324 NMP2 Tec Specs 3/4.2.2 and bases ANSWER 8.05 (1.75)
a.1.
no 2.
yes (0.25 ea)
3.
yes b. Suspension of core alterations shall not preclude completion of the movement of a component to a safe conservative position.
(1.00)
REFERENCE Nine Mile 2 Technical Specifications, Definition Section FJC 327 ANSWER 8.06 (3.00)
a.1.
1.25 REM /QTR 2. 3 REM /QTR NTE 5(n-18)
3.
18.75 REM /QTR
'.>
i,
/
4.
7725 REM /QTR b.
75 REM life saving; 25 REM non life saving cr r s v.7 e-4 6 (O.50 each)
REFERENCE FJC 328 10CFR2O
-
r w
l I
91__OEU191E10011YE_EO9EES90EE1_E9dE11190EE-AND LIMITATIONS PAGE
ANSWERS -- NINE MILE POINT 2-86/07/21-CRESCENZO, F.
ANSWER O.07 (1.50)
,
B)
~
10CFR2O appendix (0.50)
a.
b.
Provides assurance that the levels indicated will result in exposures within the limits of 10CFR to a member of the public and general population.
(1.00)
REFERENCE 10CFR2O and NMP2 Tech Specs 3/4.11.1.1 and bases.
FJC 329
ANSWER 8.09 (2.50)
Decision to notify off site agencies Making PAR Classification of the event Determining necessity for site evacuation
-
Authorizing emergency exposures (0.5 each)
REFERENCE SEP 5-4 FJC 330
,
ANSWER 8.09 (2.00)
!
a. To protect the,public health and ifpoother approved method exists.J hw ' '
.**\\
\\
'
Dvd'
(1.00)
'
- - C,.
s t
,
b.
Prior approval from a licenged SRO and(if time permits SORC review and NRC notification.
(1.00)
REFERENCE NMP2 Procedure AP-4 FJC 331 14:
_.
.
@g__ADMINISTRATIVg_PRgCEDUREgt_CgNgITIgNgg_AND LIMITATIONS PAGE
ANSWERS -- NINE MILE POINT 2-86/07/21-CRESCENZO, F.
ANSWER 8.10 (3.00)
A reactivity anomaly is a deviation from the predicted or calculated reactivity of the core. It is determined by comparing actual rod
,
'
density to predicted rod density. Since the SDM requirement for the reactor is so small, a careful check on actual conditions to the predicted conditions is necessary.
(3.00)
REFERENCE NMP2 Tech Specs 3/4.1.2 and bases FJC 332 ANSWER 8.11 (1.25)
a.
Thermal power; low pressure or low flow Thermal power; high pressure and high flow Coolant system pressure Vessel level
.
b.
FALSE
~
~
(0.25 each)
REFERENCE NMP2 Tech Specs 2.1 FJC 333
.
..
..
..
..._
..
.
.
TEST CROSS REFERENCE PAGE
QUESTION VALUE REFERENCE
-=
_
05.01 2.50 FJCOOOOO15 05.02 1.00 FJCOOOOO67 05.03 2.50 FJCOOOOO68 05.04 3.00 FJCOOOOO78 05.05 2.50 FJCOOOOOOO 05.06 2.00 FJCOOOO264 05.07 3.00 FJCOOOO292 05.08 2.00 FJCOOOO293 05.09 2.00 FJCOOOO294 05.10 2.50 FJCOOOO346 05.11 2.00 FJCOOOO347 25.00 06.01 3.00 FJCOOOOO35 06.02 1.00 FJCOOOO214 06.03 3.00 FJCOOOO286 06.04 2.00 FJCOOOO312 06.05 1.50 FJCOOOO320 06.06 2.00 FJCOOOO348 06.07 2.00 FJCOOOO349 06.08 2.50 FJCOOOO350 06.09 3.00 FJCOOOO351.
O6.10 2.00 FJCOOOO352 06.11 3.00 FJCOOOO353 25.00 07.01 1.50 FJCOOOOO41
%-
__
07.02 1.50 FJCOOOOO47
.z 07.03 2.50 FJCOOOO302
~
a=~
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07.04 2.50 FJCOOOO303 07.05 3.00 FJCOOOO304 07.06 3.00 FJCOOOO305 07.07 1.00 FJCOOOO306 07.08 2.00 FJCOOOO307 07.09 2.00 FJCOOOO311 07.10 2.00 FJCOOOO317 07.11 1.00 FJCOOOO319 07.12 3.00 FJCOOOO354
.
25.00 08.01 3.00 FJCOOOO3OB 08.02 2.00 FJCOOOO309 08.03 2.50 FJCOOOO310 08.04 2.50 FJCOOOO324 08.05 1.75 FJCOOOO327 08.06 3.00 FJCOOOO328 08.07 1.50 FJCOOOO329
.,,.., - _.
.
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.
TEST CROSS REFERENCE PAGE
QUESTION VALUE REFERENCE
__
_ _ -
08.08 2.50 FJCOOOO330 08.09 2.OO FJCOOOO331 08.10 3.00 FJCOOOO332 08.11 1.25 FJCOOOO333 25.OO 100.OO
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_
ATTACHMENT 2 TO ENCLOSURE
l Facility Comments and NRC Resolutions on Written Examination made during Exam Review NOTE: The following represent facility comments made during the examination review which resulted in significant changes to the examination answer keys.
Question No.
2.0.7.a.
COMMENT:
"This also allows a plant cooldown using the bypass valves to depressurize per OP-101-C, page 5 and 6."
RESOLUTION:
Comment accepted. Answer key changed to. reflect additional correct answer.
2.0.7.c.
COMMENT:
"This is not the only unique feature of an MSIV isolation (i.e. valves are hydraulically operated, a reactor scram is a direct result when in run on mode switch, only group at level one, only group affected by mode switch).
RESOLUTION:
Comment accepted. Alternate answers will be considered during grading.
-
3.0.4.
COMMENT:
" Color coding is same for both AC and DC, which is implied and should not be required for full credit".
RESOLUTION:
Comment accepted. Discussion of color coding for current j-type will be deleted from the answer key.
5.0.6.
COMMENT:
"First part of question is never answered in exam.
Key, should be "No" 1) Pressure remaining high is false.
RESOLUTION:
"No" is implied by text of answer; however, the key has been changed to reflect this explicitly. Comment regarding pressure is accepted, key changed to reflect this."
6.0.6 COMMENT:
"Part a. Note, if transmitter power is lost, the recorder will fail downscale,.if lose recorder power, it fails as is.
Part b. Since part (a) of the question talked about
" instrumentation", the responses to part (b) may indicate that if power still exists to the transmitter, ECCS will initiate if level drops to the initiation setpoint. This should be acceptable as' opposed to key answers."
,
0FFICIAL RECORD COPY OL NM 2 EXAM RPT - 0010.0.0 10/01/86
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__
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_
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- -
.
..
.
.
.
. _ _ _ _ _ _.
-
l
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.-
i l
RESOLUTION:
Comment will be considered during grading.
Full credit will be awarded based on candidate's assumption of exact nature of power failure.
6.10.a.
COMMENT:
" Answer is misleading, the question stated that " circuit fuse failure has. initiated logic for service water to the l
CSH Diesel". This implies that valve 94A has opened since
'
this is the only occurrence that takes place when SW is initiated.
Recommend changing answer to " valve 94A opens".
RESOLUTION: Comment accepted. Answer key changed to reflect correct
~
answer.
7.0.5.b.
COMMENT:
" Question asked for basis of staying above minimum alter-nate flooding pressure, not necessarily the definition.
Per C7 bases p.14 of 23 '11.a. "As long as. RPV pressure remains above the Min. Alternate Flooding Pressure, the core is adequately cooled irrespective of whether any water is being injected into the RPV".
To ensure adequate core cooling should be sufficient answer."
RESOLUTION:
Comment accepted with exception that a more detailed f
discussion than " adequate core cooling" must be included
'
for full credit.
i 8.0.2 COMMENT: Answer key has Remote Shutdown Panel L.C.0. of 7 days.
.
This specification was not supplied to examinees in their
packet, so answer should be restricted to Spec 3.5.1 14-day LC0 only.
RESOLUTION: Comment accepted.
Reference to R.S.P. deleted from answer.
l
.
)
i i
l
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I
!
0FFICIAL RECORD COPY OL NM 2 EXAM RPT - 0010.1.0 10/01/86
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-- - ~ _---- - - -. - -,
ATTACHMENT 3 TO ENCLOSURE i
Facility Comments and NRC Resolutions on Written Examination made after Exam Review Question N_L.
(SRO 6.08) 2.11b.
COMMENT:
Should be either True or False.
Not all rod blocks in these groups are affected by mode switch position but some are.
Statement was not clear as to whether all or some were affected.
Not Affected:
NMS - APRM inop.
RBM - All but downscale Ref:
RMCS Ops Tech, Table 1 RESOLUTION: Comment accepted.
Part "b" deleted from the answer key.
-
(R0 4.09) 6.11 COMMENT:
a.
RHS system and LPCS are cross connected in several ways that the key does not reflect:
1)
Jockey Pump, a single pressure holding pump supplies RHS loop A and LPCS.
2)
LPCS test return returns to the suppression pool via M0V30A which is the RHS loop A suppression pool return.
Note:
there are no procedural steps for the line up discussed in this question, only cautions. So operators are not required to have an in depth knowledge of this non-routine lineup.
If person answered using either of the cross connects above, part b makes no sense.
Ref: FSK 27-7 (RHS System)
0FFICIAL RECORD COPY OL NM 2 EXAM RPT - 0012.0.0 10/01/86
-
_
_
.,
Question No.
6.11 RESOLUTION:
Facility comment-to part "a" will be considered during grading, however, full credit must include a discussion of the cross connect'as described in the answer key. No credit will be allowed for discussion of the jockey pump as a system cross connect.
Part "b" will be deleted since this
'
mode is infrequently used and as such operators
,
need not maintain familiarity with these cautions.
17.04 COMMENT:
The answer in key is one of three valid answers.
Others are:
1)
Per E0P-C7, the normal control band is 159.3 to 202.3.
Per page 19 of 23, the definition of Hot' Shutdown boron weight includes level assumed to be at "high RPV water level trip (202.3")" so answer could be sufficient boron has been injected to allow operator to return water level to normal band.
2)
RQ is being utilized simultaneously with C7.
Action steps in RQ taken to insert rods include " resetting Rx scram" (Step 15.1)
which is the basis for setting the lower band limit of 159.3" (p. 13 of 23, C7 bases, 8.c and RL bases, p. 11 of 15 6c).
So valid i
answer would be " sufficient boron injected to raise water level to above 159.3" in order to reset the alarm".
RESOLUTION:
Comments will be considered during grading and partial credit awarded as is appropriate.
Full credit answer must include discussion of boron stagnation as described in the answer key.