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) | |
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r 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 '
REVIEWED BY: . eed j, /d[/yP4 Davit Lange, Le actor Edjineer (Examiner) Datd /
REVIEWED BY: /0//V /$4 Robert' M. Keller, Chief, Projects Section 1C Date'
APPROVED BY:
Harr/ B. Kistar, Chief
[b OM 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 examination OFFICIAL RECORD COPY OL NM 2 EXAM RPT - 0003. RA1029 PDR ADOCK 05000410 V PDR
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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 1 5/3 l l l l l 1 I I I l Oral Exam l 5/0 1 7/1 l l l 1 I I I I I l Simulator Examl 5/0 l 7/1 l l 1 I I I I I I l0verall I 5/0 l 5/4 l 1 1 I I CHIEF EXAMINER AT SITE: Frank Crescenzo 2. OTHER EXAMINERS: David Lange A. H. Howe M. O. Bishop (EG&G) Summary of generic strengths or deficiencies noted on oral exams: The candidates demonstrated good usage of the abnormal operating procedure 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 futur 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 session 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:
l The class, in general, performed only marginally well in sections '
4 (R.0. Procedures), 6 (S.R.0. Plant design and Instrumentation), 1 and 7 (S.R.0. Procedures). l The class performed poorly on the following specific questions:
Question N Topic Class Av !
4.01 Administrative definitions 66.0%
4.02 CSO 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% 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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during startup 0FFICIAL RECORD COPY OL NM 2 EXAM RPT
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4 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 Summary of NRC comments made at exit interview: Items 1.A and B were discussed with facility personnel presen The facility was informed that preliminary results of simu-lator/walkthrough exams were favorable, With a few exceptions, the simulator performed wel The training and operations staff were cooperative throughout the examination period. Summary of facility comments and commitments made at exit interview: The facility acknowledged the NRC comments noted abov The facility felt that the written examinations were .very comprehen-sive with Section 7 being " extremely difficult." The operating and simulator examinations were conducted in a competent and professional manner.
Attachments: Written Examination (s) and Answer Key (s) (SR0/R0) Significant Facility Comments and NRC Resolutions on written examinations made during Exam Review Facility Comments and NRC Resolutions on Written Examinations made after Exam Revie _
0FFICIAL RECORD COPY OL NM 2 EXAM RPT
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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
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FrankCfe$enzo Q Date REVIEWED BY: m e:
Davit Lange, Le M /d!/Y!P4 actor Edhineer (Examiner) Datv /
REVIEWED BY: ~
) /0//f//474 Robert' M. Keller, Chief, Projects Section 1C Date'
APPROVED BY:
Harry IF. Kist er, Chief [b OM 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 examination .
. REPORT DETAILS
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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 1 I I l l Oral Exam l 5/0 l 7/1 l 1 I I I I I l I l Simulator Examl 5/0 l 7/1 l l 1 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) Summary of generic strengths or deficiencies noted on oral exams: The candidates demonstrated good usage of the abnormal operating procedure The candidates did not routinely fill out surveillance or scram checklists during simulu+.or scenario 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 futur It was noted that procedure E0P-SCT was not trained on during simu-lator trainin 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:
' 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). The class performed poorly on the following specific questions:
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Question N Topic Class Av .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%
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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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- 4 j Personnel present at exit interview
NRC Personnel i
Frank Cre'scenzo, Reactor Engineer (Examiner)
David Lange, Lead Reactor Engineer (Examiner)
NRC Contractor Personnel M. Bishop (EG&G)
Facility Personnel
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K. Zollitsch, Superintendent of Training
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M. Dooley, Training Supervisor M. Jones, Operations Supervisor Summary of NRC comments made at exit interview: Items 1.A and B were discussed with facility personnel presen j
- The facility was informed that preliminary results of simu-
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lator/walkthrough exams were favorable, j With a few exceptions, the simulator performed wel The training and operations staff were cooperative throughout the i examination perio . Summary of facility comments and commitments made at. exit interview:
) The facility acknowledged the NRC comments noted above.
I The facility felt that the written examinations were very comprehen-sive with Section 7 being " extremely difficult." The operating and simulator examinations were conducted in a competent and professional manne : Attachments: Written Examination (s) and Answer Key (s) (SRO/RO) Significant Facility Comments and NRC Resolutions on written examinations made during Exam Review Facility Comments and NRC Resolutions on Written Examinations made after Exam Revie .
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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.
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Question N .0. 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 answe l 2.0. 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 gradin l 3. COMMENT: " Color coding is same for both AC and DC, which is implied and should not be required for full credit".
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q RESOLUTION: Comment accepted. Discussion of color coding for current '
type will be deleted from the answer ke . COMMENT: "First part of question is never answered in exa Key, should be "No".
{ 1) Pressure remaining high is fals !
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. 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 failur .1 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 initiate Recommend changing answer to " valve 94A opens".
RESOLUTION: Comment accepted. Answer key changed to reflect correct answe .0. COMMENT: " Question asked for basis of staying above minimum alter-nate flooding pressure, not necessarily the definitio 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 credi . COMMENT: Answer key has Remote Shutdown Panel L.C.0. of 7 day This specification was not supplied to examinees in their packet, so answer should be restricted to Spec 3.5.1 14-day LCO onl RESOLUTION: Comment accepte 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 N (SR0 6.08) 2.11 COMMENT: Should be either True or Fals 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 affecte Not Affected:
NMS - APRM ino RBM - All but downscale
, Ref: RMCS Ops Tech, Table 1 RESOLUTION: Comment accepte Part "b" deleted from the
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answer ke (R0 4.09) 6.11 COMMENT: 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 LPC ) LPCS test return returns to the suppression' pool via MOV30A which is the RHS loop A suppression pool retur 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 lineu If person answered using either of the cross connects above, part b makes no sens 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 in the answer key. No credit will be allowed for i) 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 answer i Others are:
1) Per E0P-C7, the normal control band is 15 ,
) 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 1 baron has been injected to allow operator i to return water level to normal ban l
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 ke l l
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2 MCl7/71C/7f/% Enshs u n o
. MASTER COPY U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: NINE MILE POINT 2 REACTOR TYPE: BWR-GES DATE ADMINISTERED: 86/07/21 EXAMINER: BISHO APPLICANT:
INSTRUCTIONS TO APPLICANT:
Use separate paper for the answers. Write answers on one side onl Staple question sheet on top of the answer sheet 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 start % OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 05. 3:M 25.00 -2 p# PRINCIPLES OF NUCLEAR POWER 25 50 PLANT OPERATION, THERMODYNAMICS, r* <* HEAT TRANSFER AND FLUID FLOW 24 50 M-25.00 { 5 . 00 *Q,b PLANT DESIGN INCLUDING SAFETY ce <* AND EMERGENCY SYSTEMS 24 hhp 2f5#
25.00 -2 5 . 00#' INSTRUMENTS AND CONTROLS-t+e?$ 9 z4 0cro-25.00 5 45.00 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 ai APPLICANT'S SIGNATURE
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS
. During the administration of this cxamination the following rules apply: Cheating on the examination means an automatic denial of your application and could result in more severe penaltie .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 cheatin ~ Use' black ink or dark pencil og to facilitate legible reproduction . Print your name in the blank provided on the cover sheet of the examinatio . Fill in the date on the cover sheet of the examination (if necessary). Use only the paper provided for answer . Print your name in the upper right-hand corner of the first page of each section of the answer shee . 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 shee . Number each answer as to category and nLmber, for example,1.4, . Skip at least three lines between each answe .
11. Separate answer sheets free pad and place finished answer sheets face down on your desk or tabl . Use abbreviations only if they are commonly used in facility literatur . 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 require . Show al) calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or no . Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND 00 NOT LEAVE ANY ANSWER BLAN . If parts of the examination are not clear as to intent, ask questions of the examiner onl . 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 complete . When you complete your examination, you shall: Assemble your examination as follows:
(1) Exam questions on to (2) Exam aids - figures, tables, et (3) Answer pages including figures which are a part of the answe 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 not use for answering the question 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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. PRINCIPLES OF NUCrJAR POWER PLANT OPERATIO PAGE 2
, 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 answe '
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 perio 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 rate a. Briefly explain the effect on power level. Include Void and Doppler effects and the effect on boiling boundr (1.5)
b. Briefly explain how the turbine EHC system reacts to this chang (0.5)
c. Briefly explain WHY the generator load increases as a result of this chang (1.0)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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. PRINCIPLES OF NUCLEAR POWER PLANT OPERATIO PAGE 3
. THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.05 (2.00) Give two reasons how feedwater heating improves power plant efficienc 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 actio QUESTION 1.06 (2.00)
In an emergency, a technique called " Breaking Vacuum" can be used to stop the turbine faster than norma BRIEFLY EXPLAIN how this helps stop the turbine and what hazard (s) exist for the turbine during this evolutio 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 shu B. Starting a centrifugal pump with the discharge valve full ope 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, Explain the sequence of events which lead to the pressure spike noted from time T=7 to T=10 sec (1.5) Explain the behavior of core inlet subcooling from time T=7 to T=20 sec (1.0)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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. PRINCIPLES OF NUCLEAR POWER PLANT OPERATIO PAGE 4
. 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 f Pump speed is then reduced so that pump head is 100 f WHAT is the new pump speed? SHOW ALL WOR 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 se Explain the reason for the differenc QUESTION 1.11 (2.50) WHY are installed neutron sources needed in the initial core /. 6 (
loading? (4-01" l 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 outag Rods are pulled to the 100% line and power is then increased to 100% with recirculation flo 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 actio [
a. What is the primary cause for this reduction in powe (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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. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 5
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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) In High Speed, WHAT is the MAJOR contributor to the NPSH required for the pumps? (0.5) 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 suctio (Setpoints required for full credit.)
QUESTION 2.03 (2.00)
Answer the following in regard to the ADS system: How many SRV's are dedicated to the ADS system? (0.5) What is the purpose of the 105 sec. time delay? (1.0) 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 receive (***** CATEGORY 02' CONTINUED ON NEXT PAGE *****)
. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 6
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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) 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? What is unique about the Group I isolation logic (MSIV only)
ARRANGEMENT in respect to the other group isolations? 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 RP What TWO conditions will automatically trip the RPS EPA's? 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 Pump In the event of gross failure of both seals with the RPV at operating pressure, what limits the leakage rate to the drywell? What indication would you have of a No. 1 seal failure? Why do the Recire Pumps ALWAYS start in Fast Speed? What system provides cooling to the pump seals?
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
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. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 7 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: The diesel generator is running in test, paralleled to the grid, when a LOCA signal occurs followed by a loss of offsite powe (1.5) The diesel generator is running in test, paralleled to the grid, when a loss of offsite power occur (1.5)
2.co QUESTION 2.11 (4,,40)#
Indicate whether the following statements regarding the RMCS are TRUE or FALSE: 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- _ pJ was The rod blocks imposed by the NMS (including the RBM) and the service and refuel platforms are' dependent on the Mode Switch positio _
_ A RWM rod block will occur if more than one control rod is withdrawn and the Rod Test pushbutton is depresse A double X, (XX), indication on the four rod display, indicates that the RPIS is receiving abnormal dat System hardware malfunctions in the RWM will not cause rod blocks when above the LPS (***** END OF CATEGORY 02 *****)
. INSTRUMENTS AND CONTROLS PAGE 8
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QUESTION 3.01 (2.00)
Briefly explain how RBCLCW temperature is controlle Include the COMPONENTS, HOW they function, and the Temperature maintaine 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 position QUESTION 3.03 (1.50)
Answer the following in regard to the Radiation Monitoring system (RMS). What TWO types of radiation detector are used in the ARM's? What type of radiation detector is used for the Main Steam Line Ra Monitors?
QUESTION 3.04 (2.00)
Plant emergency power is color coded for easy recognizatio 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 system? (1.5) At 100% power, what controlling circuit is actually positioning the control valves? (SYSTEMS AT NORMAL OPERATION) (0.5) 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 *****)
. INSTRUMENTS AND CONTROLS PAGE 9 QUESTION 3.06 (3.00)
State the calibration conditions for the following reactor water level rcnges: Shutdown Range (2 items) , Narrow Range (2 items) Wide Range (4 items) Fuel Zone Range (3 items)
QUESTION 3.07 (2.00) 1 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 pane 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 Wh loss of 1 steam flow signal input (3 element control) loss of 1 feed flow signal input (3 element control) loss of the selected NR water level input (downscale)
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. INSTRUMENTS AND CONTROLS PAGE 10
.
QUESTION 3.10 (1.00)
What are the feedback signals to each of the following RRFC controller outputs: Output of the master controller Output of the flux controller 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 occu 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 ># ,
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'***** END OF CATEGORY 03 *****)
. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 11 RADIOLOGICAL CONTROL QUESTION 4.01 (2.00)
Dafine the following terms: (IAW AP-1.0, Admin. Controls) Shall Should May 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% powe (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, Explain WHY this condition might occur and WHAT indications are available to the operator to assist in recognition of this conditio (1.0) Explain two methods by which an operator can prevent the above condition from occurin (2.0)
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
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. PROCEDURES - NORMAL. ABNORMAL EMERGENCY AND PAGE 12 RADIOLOGICAL CONTROL QUESTION 4.06 (2.50) Per N2-OP-21, Main Turbine, WHAT are FOUR indications of water induction to the Main Turbine? (2.0) 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 breake 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 )# Briefly explain HOW and WHY LPCS can be cross-connected with the RHS syste (1.5)
I While LPCS and RHS are cross-connected as described above, the operator is cautioned to:
1. Ensure LPCS suction valve MOV 112 is ope . 2. Ensure LPCS suction isolation valve 2CSL*V121 remains shu Explain the reasons for each of these caution (1.5)
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. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 13 RADIOLOGICAL CONTROL QUESTION 4.10 (2.00) Under WHAT conditions may a plant operator deviate from an approved procedure and depart from a license or Tech Spec condition?- Whose approval is required to do this?
QUESTION 4.11 (3.00)
Adequate core cooling is defined to be heat removal from the reactor I sufficient to maintain fuel clad temperature < 2200 deg. 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 mechanis .
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(************* END OF EXAMINATION ***************)
-. - _ - _
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- e
1 VE SSE L PFtE S ill$E (ppt d 1 PafuistorafLism 2 AVE 'stJHS AC( tel A l i l u g 2 Hf LIF F V ALVL F LOW 3 Cassat Irat i t I t s vi . 3 8vPAS$ V ALVF F LOW 4 UlHF int I T Silts 4 Der F USE H F LOW I (%)
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2 DOPPLE H HE ACTIVITY 3 TisFistiPJF S f t AMF t (?n y 3 SCH AM HE ACTivlTV 4 f E l L tW A T E H F t f F 4 10T AL HE ACTIVITY
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3 4 I I M l 2 l l l 1 l 0 to 20 30 40 SO O 10 20 30 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
l
.
EQUATICN SHEIT Cycle efficiency = (Net * -
f = ma v = s/t cut)/(Energy in)
, = 3g s = V ,t - 1/2 at
.
2 _,
E = mc A = AII A*A*s ,
4E = 1/2 mv a = (Vf - 13 )/t PE = agn
= e/t A= Ln2/tijg = 0.693/t1/2 y y . at t
1/2 8#f * b(*' )( ~*)] .,
- "
.
20 2 g(g /2) ,(; ))
A= 1
- 9' * $=V,yAo g , g0 , -h
. .
Q = mCast
[=Ie g h = UA pwr = w ,ah I = I,10"* O TYL = 1.3/u HVL = -0.593/u p = p lo sur(t)
o-
? = 7,e j '
SG = S/(1 - Kaff)
SUR = 25.06/T G x= S/(1 - K ,ffx)
G j(1 - K,ff1) = G 2(I ~ eff 2)
SUR = 25s/t* * (5 - o)T M = 1/(1 - K,ff) = CR)/G 3 T = ( t*/a ) + ((3 - s'/ Ta]
M = (1 - K,ff,)/(1 - K,ff;)
T = t/(s - s) f SCM = ( - K,ff)/K,ff T = (3 - o)/(Is) l 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)]
Id l1*Id2 Id Idj =2
2 7 = (:sv)/(3 x 1010) A/hr = (0.5 CE)/c2 (,,g,73)
- = 2N - R/hr = 6 CE/c2 (f,,g) .
Miscellaneous 0:nversiens Watar Partneters 10 egs I curie = 3.7 x 10 I gal. = 3.345 lo '4g no == 2.34 2.21 xlem 10 $ Stu/hr 1ga}.=3.78litars
= 7.48 ga mw = 3.41 x 100 5tu/hr Oensity = 62.4 lerg/ft3 lin = 2.54 cm Censity = 1 gm/c '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. H ft. H 2O = 0.4335 luf/i _ _
. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 14
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THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- NINE MILE POINT 2 -86/07/21-BISHOP, 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- 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) 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 ,
coolan Thus, doppler adds negative reactivity to turn the power excursio (1.0) Second: Moderator temperature coefficient begins adding negative reactivity as soon as sufficient heat is transferred to the coolant to raise coolant temperatur (1.0) , Third: Void coefficient will have little or no effect until saturation temperature is reache (1.0)
REFERENCE NMP-2, BWR Academic Series, Coefficients of Reactivity, PP. 4-1 thru 4-59 f
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. PRINCIPLES OF NUCLEAR POWER PLANT OPERATIO PAGE 15
. THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW i ANSWERS -- NINE MILE POINT 2 -86/07/21-BISHOP ANSWER 1.04 (3.00) 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 1 doppler broadenin As this new heat is transferred to the coolant, l more void formation occurs, thus more negative reactivity. Therefore, boiling boundary is back te-ite orginal 1 :: tion; approximatel (1.5) Power increasing causes pressure to increase?" The pressure increase is sensed by the EHC system which admits more steam to the turbin (0.5) 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) 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) 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)
l REFERENCE GE Reactor Fund. vol. 3 & GE Turbine Handout NMP-2, BWR Academic Series, Thermodynamic Cycles, PP. 5-48
__ PRINCIPLES OF NUCr.rAR POWER PLANT OPERATIO PAGE 16
. I_RERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- NINE MILE POINT 2 -86/07/21-BISHOP, 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 bladin 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 turbin (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;",,
." ;u .; . o r fg c.JM4 . * (1.0)
7*
B. Could cause excessively long starting currents or water hammer if the downstream piping was not fille (Will accept either answer.) (1.0) 1 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) Reactor level increases due to loss of recirc suction from annulus and increased voiding in core regio Increase in level causes turbine trip which in turn creates pressure spike to SRV opening. (1.5) As reactor pressure increases, core inlet temperature remains essentially constant thus inlet subcooling increases and decreases as a function of reactor pressur (1.0)
REFERENCE NMP2 FSAR Chap 15 Vol 27
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_ - - _ - , PRINCIPLES OF NUCLEAR POWER PLANT OPERATIO PAGE 17
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THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- NINE MILE POINT 2 -86/07/21-BISHOP, ANSWER 1.09 (1.50)
According to centrifugal pump laws:
2 ,
Head ~ (Speed) [0.5] l Therefore,
100 ft/160ft = (x/3600 RPM) [0.5]
2 l ix! = 12.96 x 10 E6 RPM (0.625)
x = 2846 RPM [0.5]
REFERENCE MNS Thermodynamics, p. 12- 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 I delayed neutron precursor.(0.5) (This decays with -80 sec. period.) (1.0)
se a'
REFERENCE Millstone Reactor Theory pg. 3.45 Pilgrim Reactor Theory pg 3.45 Nine Mile 2 Reactor Theory pg 3.45, L.O. ANSWER 1.11 (2.50) 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) 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 OPERATIO PAGE 18 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW
ANSWERS -- NINE MILE POINT 2 -86/07/21-BISHOP, l l
l 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 reache (1.0)
REFERENCE Dresden General Physics BWR RX Theory j
NMP-2 Academic Series, Poisons, PP. 6-6 thru 6-9 l
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_ _ _ _ PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE- 19
ANSWERS -- NINE MILE POINT 2 -86/07/21-BISHOP, ANSWER 2.01 (3.00) Maintaining the water level in RP .(0.5) Subcooling affect of the incoming feedwate (0.5) Auto High-to= Low Speed Transfer 1. Delta T between steam dome and recire, loop suction <10.7F for 15 se . FW flow <30% rated and FCV <19% open for 15 se . Vessel water level < level . 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 1 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) (0.5) Provides opportunity for CHS system [0.50] to recover RPV water level [0.50]. (1,0) Narrow Range (0.5)
REFERENCE NMP-2, Lesson Plan ECCS's, ADS, P , 13 and 14
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. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 20
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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 ge-<rst=~ d; c.s>>a + ,et,y /,e s ot o,. d,ff a.~4sL/ care n #
REFERENCE NMP-2, Lesson Plan, Standby diesel generator and auxiliaries, ANSWER -2,06 (1.50)
ed*ou~ ~4 g G*ool"rO g RBCLC% ., mm _ d , al?' isolation valves close [1.0] and the fens a trip
[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) 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) " 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
,me ausws,-) 4 REFERENCE NMP-2, Ops. Tech., Primary Containment Isolation System, P. 4 l
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- - . - .. - - - , ~ -. -.--..---.n , --,n.- , . -
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. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 21
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ANSWERS -- NINE MILE POINT 2 -86/07/21-BISHOP, 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]. (1.0)
1A - 2NHS-MCC008 [0.5] og us- S 7* ,,
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) Breakdown bushin No. 2 seal pressure will approach No. 1 seal pressur LFMG's will not supply breakway torqu RBCLCW [0.75 each] (3.0)
REFERENCE NMP-2, Ops. Tech., Reactor Recire. System, PP. 3 and 5 ANSWER 2.10 (3.00) DG bkr trips on LOCA 40.5) dicsel auc; tc crergency rede, (9. 5)"
when LOOP occurs, breaker closes and loads sequence as 7#
normal.(0.5)t- , (1.5) 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
. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 22
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ANSWERS -- NINE MILE POINT 2 -86/07/21-BISHOP, .00 ANSWER 2.11 (4rse)g True j True) D&M' #
4. True g c., True 4. c False REFERENCE NMP2 RWM LP pgs. 3,5 of 26 and table 2 E.O. 5
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. INSTRUMENTS AND CONTROLS PAGE 23 ANSWERS -- NINE MILE POINT 2 -86/07/21-BISHOP, 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 !
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NMP-2, Ops. Tech., RBCLCW, l i
ANSWER 3.02 (3.00)
Control switch selected to CIA Lead, C1B Lag, ClO Backu (0.5) .
'
l The station air compressor CIA stagts when the compressor control switch is placed in " START"V
'edmpressor control switch is starts STOP4(O.5)
automatically when th,e(0.5];,,;Jhe station air[0.5].
400rpsig compre
'(MFMThe 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 y s a s&h)
M (gggy n N REFERENCE NMP-2, Ops. Tech., Instrument, Service, and Breathing Air Systems, ANSWER 3.03 (1.50) G-M tubes and Ionization Chamber (1.0) Ionization Chamber (0.5)
REFERENCE NMP-2, Ops. Tech., Radiation Monitoring System, PP. 2 and 8 l
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. . INSTRUMENTS AND CONTROLS PAGE 24
ANSWERS -- NINE MILE POINT 2 -86/07/21-BISHOP, ANSWER 3.04 (2.00)
DIV. I - Green (2.0)
DIV. II - Yellow DI_V. III - Purple Co " 9 - - 0.50 each] (1.5)
-
olor codius is same for both AC and D (0.5)
(Will accept color designations for bother AC and DC.)
REFERENCE k 2h./y/,.
NMP, Ops. Tech., Plant DC ELECT. DIST. SYSTEM, P. 2' pg ANSWER 3.05 (3.00) Throttle (or Reactor) pressure Generator Load (MWe)
Turbine Speed [0.5 each] (1.5) The pressure control circuit (0.5) 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
. INSTRUMENTS AND CONTROLS PAGE- 25
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ANSWERS -- NINE MILE POINT 2 -86/07/21-BISHOP, ANSWER 3.06 (3.00) Shutdown Range in the vessel in the drywell Narronw Range 1. 1000 psig in the reactor 2. 135 in the drywell 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 Fuel Zone Range "0" psig in RPV e 2,2 'r /- t/v "0" psig in drywell ,,. zsz'M 4"% W No jet pump flow [0.27 each] (3.0)
REFERENCE NMP-2, Ops. Tech., Reactor Vessel Instrumentation System, Re ,
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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. INSTRUMENTS AND CONTROLS PAGE 26
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ANSWERS -- NINE MILE POINT 2 -86/07/21-BISHOP,M.
i ANSWER 3.09 (3.00) g fe lower than, since SF/FFCmismat (25%)theFWflowwillbe reduced to compensate for it. u (1.0) higher than, since FF/SFCmismatchM(50%)rd FF will increase to compensate for i (1.0) 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, of 6 ANSWER 3.10 (1.00) APRM feedback (C, normal; E, backup) (Also accept flux estimator feedbac o /. 6e /bar)/w Loop elbe"? flow . A<, c fled ye 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
_________ ________ High RPV pressure 128 psig Low RPV level, (level 3) 159.3 inches RHR area high temperature 135 F High Reactor Building Tem F High Reactor Building Pipe Chase Tem 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 l
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. INSTRUMENTS AND CONTROLS PAGE 27
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ANSWERS -- NINE MILE POINT 2 -86/07/21-BISHOP, ANSWER 3.12 (1.00)
to3 Y There is no load shed or sequencing for 2 ENS * SWG 40 (1,0)
( tu;// ah= ~~rt h-d sLJ nJ sy-~ A .?ns o swa lot as Aiks3d:
REFERENCE NMP2 Lesson Plan, Emergency AC Power Systems, page I. $MS funf "A sh Is (T> 1) z. a s p ,sf-fr (ra) M y. savu. w tn p.y ss,h (r.zs) .
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. PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE 28
, RADIOLOGICAL CONTROL ANSWERS -- NINE MILE POINT 2 -86/07/21-BISHOP, ANSWER 4.01 (2.00) Denotes a requiremen Denotes a recommendatio If not acted uyvu, ou maplanation is O Av-required justifying the oviivu Lokou. ^. pg Denotes permissio What is to be done or what is expected. Ncither a rcquir:rento pd ner rcccancndation.O. ,j [0.5 each] (2.0)
REFERENCE NMP-2, Procedure AP-1.0, Procedure for Admin. Controls, PP. 3 and 4 ANSWER 4.02 (2.00) Responsible for general operation of the control room (Subject to SSS and ASS). Responsible for direct supervision of operators on shif . Responsible for starting and stopping all major equipmen . Responsible for control of turbine.rrd rc::ter.2 pe Responsible for operation of major power board breaker . Responsible for operation of line breakers in switchyar . Responsible for determining if an ESF performs as required in the event of a LOCA or other abnormal inciden . ps?/a.4y JL cso (%b ) A*~) hf. u [any 4 @ 0.5 each] (2.0)
9 p,p.,v,;y. L cnbs / . /' r s* * be .,,s REFERENCE NMP-2, Lesson Plan, Composition and Responsibility of Unit Org., Item 1 NMP-2, Procedure AP-1.2, Composition and Responsibility of Unit Org., P. 5 ANSWER 4.03 (2.50) RPV water level < 159.3 inches RPV pressure > 1037 psig Drywell pressure > 1.68 psig A condition which requires an MSIV isolatio . A condition which requires a Rx scram, and Rx power is above 4%
or cannot be determine [0.5 each) (2.5)
REFERENCE ,
NMP-2, Emergency Operating Procedure N2-EOP-RQ, RP7 Reactivity '
Control, l t
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. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 29
, RADIOLOGICAL CONTROL ANSWERS -- NINE MILE POINT 2 -86/07/21-BISHOP, ANSWER 4.04 (1.50) W
, y .jk.A.cL~fush'- Dump H2 to atmospherea (to reduce H2 pressure) Trip the generator 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. ANSWER 4.05 (3.00) Thermal stratification as noted by axial skin temperature variation p>o4r**8 pd (1.0) . 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) . Rough starts Steam line water hammer High rotor eccentricity Large differential expansions Large vibration increase [any 4 @ 0.5 each] pe (2.0) 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 30
, RADIOLOGICAL CONTROL ANSWERS -- NINE MILE POINT 2 -8S/07/21-BISHOP, ANSWER 4.07 (2.00) (Verify annuncia:cr) and rod drift alarm on full core displa Select rod so its position can be monitore Check rod sequence sheet to determine proper position of the ro Place rod in its proper positio If rod continues to drift, check Tech. Spec Notify Reactor Analyst to correct possible flux problem 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. 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) 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. To prevent excessive D/P across the valv '-
(1.5)
(0.76) 1 '
2. To ensure the reactor does not drain to the suppression poo (0.75)
REFERENCE I
NMP2 OP 33 and LP LPCS; E.O. #'s I Y - .-
-
-
uh M
. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 31
, RADIOLOGICAL CONTROL ANSWERS -- NINE MILE POINT 2 -86/07/21-BISHOP, ANSWER 4.10 (2.00) To protect the public health and if no other approved method exist # y ss5 (1.0) Prior approval from a licensed SROjand if time permits SORC review and NRC notificatio (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 condition (1.00)
3. Steam Cooling proper pressure and steam flow through SRV' (1.00)
REFERENCE NMP2 EOP structure page 3- .
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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
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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
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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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U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION NINE MILE POINT 2 I FACILITY: _
REACTOR TYPE: BWR-GES ,
DATE ADMINISTERED: _ B6/07/21 _ _ _
EXAMINER: _CRESCENZO t _F.___________
rre-~
APPLICANT: __ gd[_f[3
_ -__ __
INSTRUCTIONS _Tg_ APPLICANT:
Write answers on one side onl Use separate paper for the answer Points for each Staple question sheet on top of the answer sheet question are indicated in parentheses after the question. The passing at grade requires at least 70% in each category and a final grade ofafter least 80%. Examination papers will be picked up six (6) hours ,
)
the examination start % OF CATEGORY % OF APPLICANT'S CATEGORY TOTAL SCORE VALUE CATEGORY VALUE--- ----- --- ----- __
_: 25. 00__ _ __25. 00 THEORY OF NUCLEAR POWER PLANT
_
___ ____
OPERATION, FLUIDS, AND THERMODYNAMICS 25.00 PLANT SYSTEMS DESIGN, CONTROL,
_25.g9__ ._
__
AND INSTRUMENTATION 25.00 PROCEDURES - NORMAL, ABNORMAL,
_25 99__ __
EMERGENCY AND RADIOLOGICAL CONTROL 25.00 _ ADMINISTRATIVE PROCEDURES, 25.00__ __ _ _ _
CONDITIONS, AND LIMITATIONS 199199__ 100 99 ___________ ________ TOTALS FINAL GRADE __
_%
All work done on this examination is my own. I have neither given nor received ai APPLICANT'S SIGNf.IURE
- _ - - ._ _ _
s PAGE 2 THEGRY_OF_ NUCLEAR POWER PLANT OPERATIONt FLUIDS t_AND
- *
THERMODYNAMICS
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,
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)
change 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 se (1.00)
Explain the reason for the differenc DUESTIDN 5.03 (2.50)
Why are installed neutron sources needed in the initial core I. b0 ' ; . 004 loading? 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 b'UC critical rod density change relative to a startup with ii.SGF l a source installed?
.
DUESTION 5.04 (3.00)
Regarding the xenon transient following a significant DECREASE in reactor power from high power operation: HOW will peripheral control rod worth be affected (INCREASE,
! DECREASE, REMAIN THE SAME) during the xenon peak? BRIEFLY (1.50)
EXPLAIN your answe 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 answe (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
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PAGE 3
THEORY OF NUCLEAR PDWER PLANT OPERATION, FLUIDS t AND
__ THERMODYNAMICS
.
QUESTION 5.05 (2.50)
The figure below represents six fuel assemblies within a critical rod reacto Why would an operator expect a higher than normal (2.50)
worth associated with control rod B-2?
X= control rod fully inserted O= control rod fully withdrawn X 0 X O X O X 0 X !
l QUESTION 5.06 (2.00)
Current plant conditions suggest that a severly degraded core condition exists. Several SRV's have been, or are currently, ope 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 answe l I
DUESTION 5.07 (3.00) List three parameters which contribute to AVAILABLE NPSH for a recirculation pump. Limit your answer to those parameters which (1.50)
are directly available in the control roo Consider two RPV conditions: low power and low flow (< 10 %) OR high power and high flow ( >B57.) . During which condition is REQUIRED NPSH for a recirculation (0.50)
pump greater? 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 4 THEdRY OF NUCLEAR POWER _PL. ANT OPERATIONt_ FLUIDSt_AND THERMODYNAMICS
.
QUESTION 5.08 (2.00)
(2.00)
Match the descriptions (a-d) to the corresponding 1/M plots (1-4).
(e) tDE AL (b) DETECTOR TO CLOSE TO SOURCE e C (c) DETECTOR TO FAR PRots MURCE (d) PREDICTON AFTER IDo BUNDLES 7'E l .0 - (L)
.9 -
.a -
.7 - (l)
.6 -
.5 -
4- s 3)
.s - \
\ CRITICAllTY 2_
~
(\,
ibo abo sbo abo soo sho
"
sammet or rutt swoLEs s narrn QUESTION S.09 (2.00)
The term " critical power" refers to that bundle power level corresponding to the onset of transition boiling (OTB) somewhere l
in that bundl State how critical power varies (ie. increases, decreases, or is l l
not affected) by each of the following:
(O.50) If coolant mass flow rate increases (0.50) If reactor pressure increases (0.50) If local power increases (0.50) If inlet subcooling increases
- )
(***** CATEGORY OS CONTINUED ON NEXT PAGE
. .
PAGE 5 THEO'RY_OF,NtJCLEAR POWER _ PLANT OPERATION g FLUIDS t_AND
' 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 motor Explain the sequence of events which lead to the pressure (1.50)
spike noted from time T=7 to T=10 sec Explain the behavior of core inlet subcooling from time (1.00)
T=7 to T=2O sec 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 powe = feedwater flow = 6400000 lb/hr MFW
= cleanup demin flow = 110000 lb/hr MCU
= steam flow = 6423000 lb/hr MG
= CRD flow to reactor = 23000 lb/hr MCRD HFW = enthalpy of feedwater = 345 Btu /lb HCU,IN = enthalpy cleanup flow in = 506 Btu /lb HCU,0UT= enthalpy cleanup flow out == 419 Btu /lb HG = enthalpy of steam 1194 Btu /lb
= enthalpy of CRD flow to reactor = 68 Btu /lb Hcrd Op = recir. pump energy input = 26500000 Btu /hr
= heat losses from nuclear boiler = 2040000 Btu /hr Of1 Qcore = reactor core thermal energy input = solve for (***** END OF CATEGORY 05 *****)
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I PAGE 6 6.__ PLANT _ SYSTEMS DESIGN t, CONTROL t AND INSTRUMENTATIDN
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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 Syste Include in your answer a discussion of how the FWCS will respond. Continue your discussion to a point where stability is reestablished. Assume FWCS is in three element control mode i end no operator action (1.00) Feedwater flow signal fails lo (1.00) Steam flow signal fails lo (1.00 RPV level signal fails lo 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 startu a. When the IRM is reading full scale on Range 10, the APRM's should be reading approximately 10*/. powe Shifting from Range 4, indicating 75, to Range 5, will result in an indication of 24, en Range Reactivity feedback, due to the moderator temperature coefficient should begin at approximately Range 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 i
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AND INSTRUMENTATION PAGE 7 6.__ PLAN'T SYSTEMS DES _I_GN, CONTROL 1
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l QUESTION 6.03 (3.00)
With the plant operating at 1007. power, recirc in Flux Manual, an operator inadvertently decreases the EHC pressure setpoint 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%. TCV positio BPV positio Reactor powe Reactor pressur )
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 referenc (0.50)
a. Back Up Hoist Limit (O.50)
b. Rod Block Interlock No. 1 (0.50)
c. Fuel Hoist Interlock (O.50) 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)
syste l
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l PAGE 8 6 3__ PLANT _gYgTEM@_DE@IGN t _CONTROLt_AND INSTRUMENTATION QUESTION 6.06 (2.00)
e. How does the reactor vessel water level instrumentation (1.00)
fail upon a loss of power? Include both meters and recorder 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- The rod blocks imposed by the NMS (including the RBM) and the service and refuel platforms are dependent on the Mode Switch (0.50)
positio c. A RWM rod block will occur if more than one control rod is (O.50)
withdrawn and the Rod Test puslinutton is depresse A double X, (XX), indication on the four rod display, indicates (0.50)
that the RPIS is receiving abnormal dat System hardware malfunctions in the RWM will not cause rod (O.50)
blocks when above the LPS ($$$$$ CATEGORY 06 CONTINUED ON NEXT PAGE *****)
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' PAGE 9 PLANT SYSTEMS DESIGN 1 CONTROL 1__AND INSTRUMENTATION l
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 powe The diesel generator is running in test, paralleled to the grid, (1.50)
when a loss of offsite power occur QUESTION 6.10 (2.00) Assume an EDG-2 starting circuit fuse failure has initiated 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 delay How would these same valves respond with a valid start signal 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)
delay QUESTION 6.11 (3.00) Briefly explain how and why LPCS can be cross-connected with (1.50)
the RHS syste While LPCS and RHS are cross-connected as described above, the operator is cautioned to:
1. Ensure LPCS suction valve MOV 112 is ope . Ensure LPCS suction isolation valve 2CSL*V121 remains shu (1.50)
Explain the reasons for each of these cautions
.
(***** END OF CATEGORY 06 *****)
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PAGE 10
- PROC'EDURES - NORMAL t ABNORMALt _ EMERGENCY ANE)
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? (0.50) In what specific operating condition is this applicable? What actions are required if this conditions exists? (0.50)
QUESTION 7.03 (2.50)
I 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 downwin Using the attached figures from EPP-8 and EPP-26, and a wind direction as shown on figure 14, determine which ERPA must (1.50)
seek shelte 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)
j dose rate limit 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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' PROCEDURES - NORMAL, ABNORMAL t EMERGENCY AND
'
RADIOLOGICAL CONTROL
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QUESTION 7.05 (3.00) During an ATWS condition if RPV water level should become undeterminable, EDP-C7, " LEVEL / POWER CONTROL," directs the injection operator to maintain power above 8% by controlling I
to the RPV. What is the significance of 8% and why must (2.00)
power be mintained above it? 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 pressur QUESTION 7.06 (3.00)
Adequate core cooling is defined to be heat removal from the reactor sufficient to maintain fuel clad temperature < 2200 deg. 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)
mechanis I QUESTION 7.07 (1.00)
i i
Define " Maximum Safe Operating Parameter" as it applies to (1.00)
the EOP' 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 until80RPV psig is pressure is 80 psig above suppression chamber pressur known as the Minimum Flooding Pressure. Discuss the basis for this (2.00)
pressur (See attached sheet 3 of 5, N2-EOP-C6)
f (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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~ PAGE 12 7._ PROCEDURES -_ NORMA k ABNORMALt _ EMERGENCY AND RADIOLOGICAL CONTROL QUESTION 7.09 (2.00)
a Explain the basis f or the f ollowing prerequisites to transf er of rscirculation pump from low to high spee "Feedwater Flow greater than 4.25 million pounds per hour (1.00)
(307. of rated)...." " Differential temperature between recirculation pump SUCTION F." (1.00)
and STEAM DOME TEMPERATURE is greater than 10.7 deg DUESTION 7.10 (2.00) 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 (1.00)
contaminated equipment. Explain the purpose of this cautio 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 operatio (1.00)
What is the purpose of this caution?
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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' PAGE 13 PROCEDURES - NORMAL 3_ ABNORMAL 1_ EMERGENCY _AND 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 stea indications Explain why this condition might occur and what are available to the operator to assist in recognition of (1.00)
this conditio Explain two methods by which an operator can (2,00)
prevent the above condition from occuring.
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(***** END OF CATEGORY 07 *****)
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' PAGE 14 ADMINISTRATIVE PROCEDURESt _ CONDITIONSt _AND_ LIMITATIONS QUESTION 8.01 (3.00)
The plant is operating at 30% power. The(it EDG-1 is currently has been inop for inoperative due to planned maintenance 10 hrs.). The I&C supervisor reports that instrument 25W8PSL95 (Service Water inlet pressure for EDG-2) has failed a channel functional tes Using the attached technical specifications, determine ALL (2.00)
applicable actions which must be take How would your actions be different if, rather than EDG-1 (1.00)
being inop for maintenance, LPCS was inop for maintenance?
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 cas 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 answe 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 setpoint (***** CATEGORY 08 CONTINUED ON NEXT PAGE $$$$$)
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PAGE 15
' ADMINISTRATIVE _ PROCEDURES CONDITIONSg_AND_ LIMITATIONS
_
QUESTION 8.05 (1.75) Indicate whether each of the following is considered a " Core (O.75)
Alteration" per the Nine Mile Point 2 Technical Specification . Withdrawal and insertion of an SRM detector to check the drive moto . Removal of an LPRM string for replacemen . Removal of an uncoupled control rod for replacemen While withdrawing a control rod to test a position indicator probe, it becomes necessary to suspend core alterations due to containment problems. Should the rod be reinserted or must it remain mid-positioned until conditions are such that core (1.00)
alterations are allowable?
QUESTION 8.06 (3.00)
a. State the 10CFR2O limits for penetrating radiation for the (2.00)
followings 1. Whole body. (without an NRC form 4)
2. Whole body. (with an NRC form 4)
3. Extremitie . Ski (1.00)
b. State the allowable emergency exposure limit QUESTION B.07 (1.50)
Where would you find the limits for the concentration of RAM (0.50)
released in liquid effluents to unrestricted areas?
(1.00) What are these limits based on?
QUESTION B.08 (2.50)
(2.50)
List 5 responsibilties the Emergency Director MAY NOT delegat *****)
(***** CATEGORY 08 CONTINUED ON NEXT PAGE
- _ __ _ . - .. ._
l
!
PAGE 16 ADMI'NISTRATIVE PROCEDURES t _CONDITIONSt__AND LIMITATIONS GUESTION 8.09 (2.00) Under what conditions may a plant operator deviate from an approved procedure and depart from a license or Tech Spec (1.00)
condition?
(1.00) Whose approval is required to do this?
QUESTION 8.10 (3.00)
Technical Specification 3/4.1.2 addresses " Reactivity Anomalies."
Briefly explain what a Reactivity Anomaly is; the method used to (3.00)
determine it; and why it is important to be aware of its existenc QUESTION 8.11 (1.25)
(1.00) List the Nine Mile Point 2 safety limit TRUE OR FALSE: Reactor vessel water level safety limits are (0.25)
applicable in operational condition (***** END OF CATEGORY 08 *****)
(************* END OF EXAMINATION ***************)
- .. _ _ _ , _ - .
__ -_ _ - - _ _ . -.
. ; a.QUA L 6 0ha j uA & A ShtL6
j t'
p - F, et/ T- *- * I/C1-h)
! I CL = 3.7 x 10105q N(c) - No a-U ap = - 1 x 10 5 gJ'T or= (Lf+L3 ) (cred)2 L (4*v3)
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 (2 )/lo x t=
. .-,.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') (EA )
5 = x5g + (1-x) Sg Q - E p at 1 in.- 2.54 cm ap - I L pv2 D 2sc
1 E*1* * 3 7E3 lit *f5 I - 64/Le 1 4 - 2.205 lb N pao /A p - k( e f f ) -1 Kle 1)
17.58 vat ts - 1 EIV/ min 1-K(e f f) 2 1 psi - 6.895 Pz
-"
CRI 1 psi = 2.936 - H 5(00}
"
M CT. I(eff)1 1pst - 27 68 - B 0 (G ';;
.
6 = .0071 0 - Nah ._
L =2x 10 , sec
,
Q - DAAT
.
m
%
.. .
i s
Table Saturated Steam: Temperature Table Enthalpy Entropy Ahs Pres Specific Volume Sa Sa Temp Sa Sa Sa Temp Lb per Sa Evap Vapor liquid Evap Vapor Fahr liqisid Ivap Vapor Li uid sg t Iahr Sq In g h ig h8 sg sig i p vg vig _ vg 0.0000 2.1873 2.1873 3 .5 107 .1802 3 .996 101 .9 0.0081 2 1651 2.1732 3 .008 107 .2 0 016020 28390 2839 0 0 0122 2.1541 2.1663 3 .018 107 .1 0 016019 2614 1 2634 2 18 8 018249 0.0162 2.1432 2.1594 4 .027 107 .1325 2.1527 42 0 48 8 112163 10 035 10698 107 / 221 .1459 4 I? 8 12.041 4 .6 108 .1111 2.1393 19657 19657 14 047 21327 4 .4 108 .0321 2.1006 1 Alp 0 1810 0 16 051 48 0 0 16514 0 016021 l 0 0361 2.0901 2.1262 5 .3 108 .0400 2.0798 2.1197 5 SO S 0 17796 20 057 106 .2 0 016024 15892 1589 2 0.0439 2.0695 2.1134 54 0 57 8 019165 22.058 106 .1 0 016026 1482 4 1482 4 0.0478 2.0593 2.1070 5 ).9 108 .6 0 0516 2.0491 2.1000 5 I O22183 26 060 106 .9 0 016011 17922 129 .0555 2.0391 2.0946 6 .060 10591 10871 0 25611 0016033 120 .0593 2.0291 2.0885 5 .059 105 .0192 2 0824 6 El 0 1056 5 32.058 105 .0670 2.0094 2.0764 6 I 34 056 105 .0708 1.9996 2.0704 68 8 68 0 36.054 105 .2 0 31889 0 016046 926 5 926 5 58 8 0.0745 1.9900 2.0645 7 .052 105 .1 0 36292 0016050 868 3 0.0783 1.9804 2.0587 7 .9 109 .9708 2.0529 1 .046 105 .8 0 016058 164 1 0.0858 1.%I4 1 0472 76.0
'
I4 I O41550 71 .9520 2.0415 7 .040 104 .6
ft016067 611 R 673 9 Ts e 04746I
- 8 .4 109 I.9426 2.0359 0 016072 633 3 633 3 48.037 8 .3 0 0969 1.9334 2.0303 0 016077 595 5 595 5 50 033 8 .1 10982 0.1006 1.9242 2.0248 0016082 560 3 560 3 52 029 0 MS 0$1102 104 .1043 1.9151 2.0193 88 0 0 61518 0 016087 227 5 52/5 54.026 0.1079 1.9060 2.0139 8 %8 496 8 56 022 104 .9 98 8 065551 0016091 0.1115 1.8970 2.0086 9 I 58 018 10423 1100 8 90 0 0 69813 0 01609"l 01152 1.8881 2.0033 9 .6 1101 6 92 8 074313 0 01610i 1102 5 01188 1.8792 1.9980 9 .5 94 8 0 19062 0 016111 0 1224 1.8704 1.9928 9 .3
> 98 I O84012 0 016111 0 1260 1.8617 1.9876 30 0 370 9 370 9 66 003 10382 1104 2 0 89356 0 016123 d, 90 8
-
e
. _ . . . _ _ ____ ___.____ _ ____ _ _ ______
- __
s
.
P Ahs Psess $pcolic Volume folhalpy Entropy
- Sal Sa Sa Sa Sa Sa Temp Temp lb per 1 alo Sq lo liquist ivap Vapos liquid Ivap Vapor liquid Evap Vapor Fahr vg hl h fg hg sg sig sg t
- I p vg vig 0 016130 350 4 350 4 61.999 103 .1295 1.8530 1.9825 IRO
'
ige 8 0 94924 10 I00789 0 016137 331 1 331 1 69 995 1035 9 110 .1331 1.8444 1.9775 i 182 I 1.8358 1.9725 1KS 0 016144 313 1 3131 11.992 103 .1366 104 8 1 06 % 5 1Es 0 016151 29616 296 18 73 99 1033 6 110 .1402 1.8273 1 % 75 l l$6 8 11347 0.1437 1.8188 1.9626 10 R0 28 280 30 7598 103 .5 i 100 0 26537 26539 77.98 103 .3 01472 1.8105 1.9577 lite lis t 1.2750 0 016165 251 38 79 98 103 I8021 1.9528 112.0
- 112I I3505 0 016173 251 37 238 21 238 22 81.97 1029I 111 .1542 1.7938 1.9480 11 .9 111 .1577 1.1856 1.9433 11 lig e I$133 102 .7 0.1611 1.7774 1.9386 11 lig e I6009 0 016196 214 20 f 214 21 85.97 203 26 87.97 1025 6 1113 6 0.1646 13693 1.9339 IMO its t . I6927 0 016204 203 25 0 016213 192 94 192 95 89 96 102 .1680 11613 1.9293 1 2 e 11891 0 1715 13533 1.9247 12 % 1023 3 1115 3
- 124e 18901 0 1749 13453 1.9202 1ES
'
0 016229 174 08 174 09 93 96 102 .1 178 8 1 9959 11374 1.9157 12 ; 95 96 102 .0 0 1783 128 0 2 1068 0 016247 157 32 15733 97.96 101 .8 01817 13295 1.9112 13 .9068 13 .3445 13140 1.9024 1R0 0 016265 14240 14241 101.95 101 .5 0.1884 1348 2 4717 1K0 0 016214 13555 135 57 103 95 1016 4 112 .1918 11063 1.8900 134 8 2 6047 16986 1.8937 13 .2 112 e I 2 1438 123 00 107.95 1014 0 112 .1985 1.6910 1.8895 1440 148 8 2 8892 0 016293 122 98 3 0411 0 016303 117 21 117 22 109 95 101 .8 02018 1.6534 1.8852 14 s 31997 0 016312 Ill 14 Ill 16 Ill 95 10113 1123 6 0 2051 1.6759 1.8810 14 .5 112 .6684 1.8769 14 les e 3 5181 0 016312 10168 10110 115 95 100 .3 02117 1.6610 1.8727 144.0
0 016343 9705 9707 117.95 1000 2 112 .2150 1.6536 1.0606 1R0 198 0 3 1184 3 9065 0016353 9266 92 68 119.95 100 .2183 1.6463 .l.8646 15 e 1540 4 1025 0 016363 8650 88 52 121 95 10058 11273 02216 1.6390 1.8606 i 1548 154 8 4 3068 0 016314 8456 84 57 123 95 100 I6318 1.8566 158 0 158 8 4 5197 0 016184 80 R2 R0 83 125 % 100 .6245 1.8526 15 .2 1130 2 0.2313 1.6174 1.8487 10 let I 4 F414 0 016395 17 29 113 .6103 1.8448 18 .0 0 016417 70 70 10 12 131.96 9998 113 .6032 1.8409 1RS 1640 52124 1Et 6767 6768 133 97 99 .6 0.2409 1.5961 1.8371 188 8 54623 0 016428 180 0 5 7223 0 016440 64 18 64 80 13597 99 .4 0.2441 1.5892 1.8333 10 .97 99 .2 0.2473 1.5822 1.8295 1R8 2 110 8
! 112 I 62136 0016463 5943 5945 139 98 99 .0 0.2505 1.5753 1.8258 17 t14 0 6 5656 0 016474 56 95 56 97 141.98 9938 113 .5604 1.8221 17 l 178 8 6 8690 0 016486 54 59 54 61 143 99 99 .5616 1.8184 IRO
j 170 0 71840 0 016498 5235 52 36 145 99 99 .4 0.2600 1.5548 1.8147 IMS
! .
/
-
, Q
- ' l p .
,
I a.
i
[nthalpy Entropy Abs P ess Specific Volume Sal Temp Sat Sal Sa Sal Temp tb pe Sal liquid Evap Vapor Fahr Ivap Vapo Liquid Evap Vapor Tahr Sq in ligmrt 5t I vg he h it ha 5 Sig i p v, vtg 0.2631 1.5480 1.8111 10 leg e 15110 0 016510 0.2662 I.5413 18075 14 IIII 1850 ' 0 016522 0.2694 15346 18040 1 8 .8 1139 8 184 8 8 203 0 016534 02125 1.5219 1.8004 10 .5 184 8 8 568 0016547 0.2756 1.5213 13 % 9 18 .3 tes s 8 947 0 016559
158 04 98 .1 0 2187 1.5144 11934 I f 40 957 19 I 160 05 98 .9 0 2818 1.5082 1.7900 9147 0016585 39 331 39 354 13865 19 II2 e 37 824 162 05 98 .5017 10 168 0 016598 37 808 0.2879 1.4952 1.7831 19 s 0016611 36 348 36 364 164 06 98 !!44 4 les e 10 605 166 08 97 .2910 1.4888 11798 198.0
' ige e 11 058 0 016624 34 954 34 970 1146 0 0 2940 1.4824 17764 20 .9 260 4 11 526 0 016637 03001 1.4697 17698 284.0 l
31135 31151 17211 975 4 1147.5
- 284 8 12 512 0 016664 0.3061 1.4571 13632 29 '
288 8 13 568 0016691 0 3121 1.4447 1.7568 21 .5 212 8 14 696 0 016719 0.3181 1.4323 17505 21 .0 216 0 15 901 23 148 188 23 965 2 115 I4201 13442' 22 s 17 186 0 016775 23 131 22 .9 0.3300 1.4081 1.7380 2248 18 556 0 016805 21 529 ?!545 228 I 20 056 20 073 196 31 960 0 115 .3%I 13320 228 8 20 015 0 016834 0 3411 13842 13260 23 .4 115 .3725 13201 23 .1 1160 6 0.3533 1.3609 11842 248 8 248 8 24 068 0 016926 16 304 24 l 15 260 212 50 949 5 116 .3494 17085 244e 26 826 0 016958 15 243 14 264 14 281 216 56 946 8 1163 4 0 3649 1.3379d 1.7028 248 0 248 8 28 196 0 016990 16912 25 .1 !!64 3 0 3706 1.3266 2520 30 883 0 011012 13 358 13 375 12 538 ??4 69 94 .1 0 3763 1.3154 16917 255 0 255 8 31091 0 017055 17 570 938 6 116 .3819 1.30f3 16862 26 .3876 1.2933 16808 26 I 37 894 0 3932 1.2823 I6755 26 I 1170 0 288 0 930 3 1171 3 0.3987 1.2715 1.6702 27 .s 0 1043 1.2607 1 6650 276 0 46 147 0 017228 9 162 9 180 245 08 9275 1172 5
!!8 0 24917 924 6 1113 8 0 4098 1.2501 1.6599 20 .0
041130 8 1780 8 1453 253 3 921 3 1175 0 284 I 52 414 0 4208 1.2290 16498 28 .2186 16449 29 .5 915 9 117 .2082 16400 29 R759 6 8433 265 6 9130 1118 6 1 2ts t
.
-- --
.
)
Os
[nthalpy Entropy j Abs Press Specific Volume Sa Temp Sa Sa Sa Sa Temp Lb per Sal Evap Vapor Fahr Ivap Vapor Li vid Evap Vapor Liquid Iahs Sqin liquid hg sg sg t I h ig sg e I p vg vig vs
910 0 11793 0 4372 1.1979 1.6351 300 0 61 005 0 01745 6 4483 6 4658 2693 3000 273 8 90 .1817 16303 304 0 304 0 71119 0 01149 6 0955 6 1130 3000 278 0 904 0 1182 0 0 4479 I.1776 1.6256 300 0 15 433 0 01753 57655 5 7830 1.6209 31 .0 11831 0 4533 1.1676 312 0 19 953 001757 1.1576 1.6162 316 0 5 1673 51849 286 3 89 .1 0 4506 310 0 84 688 0 01761 894 8 1185 2 0.4640 1.1477 1.6116 32 NG 89 643 0 01766 48%I 49138 290 4 32 .6 1186 2 04692 1.1378 1.6071 324 0 94 826 0 01770 1.6025 370 0 44030 4 4208 298 7 888 5 118 .1280 320 0 100 245 0 01774 f 0.4798 1.1183 1.5981 33 %6 30 .1 0.4850 1.1086 1.5936 3360 3 96AI 3 9859 30 .1 3M I . til 870 0 01183 878 8 11901 0 4902 1.0990 1.5892 34 !!7 992 875 5 1191 0 0 4954 1.0894 1.5849 34 .1 0 5006 1.0799 1.5806 340 0 0 01797 34018 3 4258 319 7 340 0 131.143 868 9 11923 0 5058 1.0705 1.5763 35 .5721 35 .3 86 l94 4 0 5161 1.0517 1.5678 360 0 300 0 153.010 0 01811 2 9392 001016 2 8002 2 8184 336 5 858 6 1895 2 0 5212 10424 1.5637 36 .0332 1.5595 360 0 0 01821 26691 2 6873 340 8 855 1 1195 9 0 5263 l 300 0 169113 1.0240 1.5554 372 0 0 01826 2 5451 2 5633 345 0 8516 11 % 3 0.5314
)
372 0 177 648 1.5513 37 I !!9 .5365 1.0148 353 6 844 5 1198 0 0.5416 1.0057 1.5473 300 0 300 0 195 729 0 01836 2 3110 2 3353 35 .5466 0.9966 1.5432 30 M I 205 294 001842 22120 2 2304 0 01847 21126 ?t311 36 .2 1199 3 0 5516 0.9876 1.5392 30 .5352 39 .4 1199 9 302 0 225 516 0.5617 0 % 96 I,5313 30s e 0 01858 I929I I9417 370 8 829 7 I200 4 MSI 236 193 0 01864 I8444 I8630 37 .9 120 .5667 0.9607 1.5274 400 0 400 0 247259 0 5717 0 9518 1.5234 40 .5 404 0 258725 0 5766 0.9429 1.5195 400 0 0 01875 16877 17064 383 8 818 2 120 .5816 09341 1.5157 41 I6340 3881 81 .4 412 0 282 894 0 5066 0.9253 1.5118 dis t 00lR87 1 5461 15651 392 5 810 2 120 % 9 8062 120 .9165 1.5000 420 0 420 0 308 180 0 01894 i
14184 14374 4013 802 2 120 .9077 1.5042 42 .5004 42 U1906 4320 130266 132119 4101 793 9 1204 0 0 6063 0 8903 1.4966 4320 35100 0 01913 414 6 1893 12042 0 6112 0 8816 1.4928 436 0 4MI 366 03 0 01919 124RR7 1 76806 0 01926 119761 121687 419 0 185 4 1204 4 0 6161 0 8729 1.4890 440 0 44 .1 1204 6 0 6210 0 8643 14853 4440 4440 397 56 0 01933 1.14874 1.16806 0 01940 1.10712 1 82152 428 0 176 7 12043 0 6259 0 8557 14815 440 0 440 0 414 09 43114 0 01947 1 05764 107711 ~12 5 172 3 1204 8 0 6308 0 8411 1.4778 ( cit (~~5 0 17 0 1678 1204 8 0 6356 0 8385 1 4741 GI el 44873 0 01954 1 01518 1 03472
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ .
- -
u
, Sperific Volume Enthalpy [ntropy Ahs l'ress Sat Sa Sa Temp lb per Sal Sal Sa temp liquid [vap Vapor Liquid [vap Vapor Fahr latu Sq lo l iquirl Ivan Vapor hg hg sig sg I I p vg v ig _ vg ht i sg 06405 0 8299 1.4704 48 .5 76 estI 466 87 0 01961 0 95557 44 .6 120 .4667 44 I 485 56 0 01969 0 93588 120 .6502 0 8127 1.4629 46 I 504 83 0 01916 0.6551 0 8042- 1.4592 47 .3 120 R4 0 6599 03956 1.4555 47 .3 475 8 54511 0H1997 0 6648 01871 1.4518 48 / 0 81711 '464 5 7398 120 .0 408 8 4691 73 % 03785 1.4481 404 I 58781 0 02009 016613 018622 44 .5 0.6745 03700 1.4444
'
610 10 0 01011 0 13641 1.4407 49 O 0 72820 478 5 724 6 120 I 613 03 0 07026 0 10194 1.4370 49 O f Ro .7 0 6842 01528 496 8 656 61 007014 71 .2 0 6890 01443 1.4333 58 .9 500 8 680 86 12013 0 6939 01357 1.42 % 50 Of>2938 0 64991 4923 70 .4258 50 .5 7033 120 .5 03036 07185 1.4221 51 .3 512 0 6921 119 .4183 51 % 997 058019 5071 5tl8 184 16 68 .4146 52 .2 0 7182 0 6926 1.4108 52 .3 524 8 07231 0 6839 1.4010 528 O 0 02112 0 49843 051955 521 8 67 .3 5288 870 31 0 7280 06152 1.4032 53 .3993 536 6 931.17 00?l34 0 46123 048757 5311 66 .4 536 8 65 .3 0 7378 06577 1.3954 54 I 65 .1 03427 0.6489 1.3915 54 /7 044834 541 8 5448 119 .3876 54 .6311 1.3837 55 .5 119 .6222 1.3797 55 f>0 5572 63 II871 0 7625 0.6132 1.3757 568 8 1133 38 0 02207 0 36507 0 38714 56 .3716 5640 0 02221 0 35099 0 31320 56 .5 118 .5950 1.3675 56 .5 118 I 1207 12 0 7775 0.5859 1.3634 57 .5 11823 572 8 0 7825 0.5766 1.3592 57 .2 118 I 1285 74 58 .5673 1.3550 500 0 1326 17 0 02279 0 29931 0 32216 58 Ses 8 0 7927 0 5580 1.3507 58 .4 117 Set t 1367 7 0 7978 0.5485 1.3464 58 .8 3 566 8 117 .5390 1.3420 59 .0 1453 3 0 02328 0 26499 0 28827 6051 e
55 .2 0 8082 0.5293 1.3315 598.8
" 588 8 14978 0 07345 0?S425 027770 61 . _
- _ _ _ - _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ . ..
_ _ _ _
O T
m
- . ~ . . - - . . . . . . . . . .
Temp th pe Sa Sa Sa Sa Sa Sa Temp liquid Ivap Vapor Li sid Evap Vapor Liquid Evap Vapor Fahr Iahr Sq in I I h Ig hg s, sg, s, g P 'J ' Ig_ 't 11673 0 8134 0.5196 1.3330 000A 026747 61 .6 see s 1543 2 0 02364 024384 54 .1 0.8147 05097 1.3284 48 .9 804 0 15897 53 .4 0 8240 0.4997 1.3238 60 .8 000 0 16313 5243 115 .4896 1.3190 61 .0 812 e 115 .8348 0.4794 1.3141 0 02444 020516 0 22960 6408 51 Sit e 1735 9 0.8403 0.4689 1.3092 820 820 I s !18 ! 022081 64 .3 115 .3041 82 .1 4? .8 824 8 1839 0 0 02489 0 18737 0 8514 0.4474 12988 520A 0 02514 017880 020394 65 .1 828 8 1892 4 0.8571 0.4364 12934 63 .9 47 .2 532 8 194 .8628 0 4251 1.2879 SES 0 18792 67 .7 113 .1 45 !!331 0.8686 0.4IM 1.2821 80 .8746 0 4015 1.2761 54 .1 112 g 2118 3 0.8806 0.3893 1.2699 44 .9 431.] 1124 0 848 0 21781 002657 013876 0 8868 03767 1.2634 852A 0 02691 0 13124 0.15816 700 0 4181 1118 1 052 0 2239 2 405 7 111 .3637 1.2567 S$6A 058 8 23017 0 02728 017387 015115 70 .1 110 .8995 0.3502 1.2498 00 .9 000 e 2365 7 0.9064 03361 1.2425 66 I1 0 10941 013757 72 est t 243 .1 199 .9137 0 3210 1.2347 08 .9212 0.3054 1.2266 57 .2179 47 .5 107 .5 31 .5 0.9365 0.2720 1.2006 00 .9447 0 2537', 1.1984 50 .2 105 I 003114 0 07349 -- 0 10463 0.9535 0.2337 l.1872 80 .2 104 . % 34 0.2110 1.1744 802A 0 03313 0 05197 0 09110 790 5 24 .6 802 8 2934 5 0.9749 0.1841 1.1591 99 A 101 I3 4 82 .2 0.9901 0.1490 1.1390 70 .3 0 03662 00385T 0 07519 0.1246 1.1252 702A 0 03824 0 03173 0 06957 835 0 1441 9791 1.0006 78 .5 '
1.0169 0.0076 1.1046 10 ?!92 0 06300 35 .0 95 .2 1.0329 0.0527 1.0856 705A 0 04427 0 01304 0 05730 87 .4 93 .0612 00000 1.0612 700s4F'
0 05078 0 00000 0 05078 90 .0 90 . 705 47' 3200 2
.
,
'
' Critical temperalute ] g
- -
.
Table 2: Saturated Steam: Pressure Table
.. .- .
Specific Volume Enthalpy Entropy Sa Sa Sa Sa . Abs Pres Abs Pers Temp Sal Sa liquid Evap Vapor Liquid Evap Vapor tb/Sq I th/Sg i Fahr liqunt Ivap Vapor sg p p i vi v,i vg hg hg g hg sg s gg 2.1872 2.1872 0.00085 3302 4 33024 0.0003 1075 5 107 .0425 2.0967 0.25 1235 5 1235 5 27 382 1060 1 108 .0370 8.58 64 .5 47.623 1048 6 109 .0925 1.9446 8 58 19 586 0 016071 1.8455 1.9781 I 110 .6094 1.8443 .1 114 .2836 1.5043 1.1879 1 .26 98 GIS
/26 199 180.17 970 3 1150 5 0.3121 1.4447 1.7568 14 598 21200 0 016719 26782 1 .21 %91 115 .3137 1.4415 1.7552 15 e , 21301 0016126 26 214 20 070 20 087 196 27 9601 115 .3358 1.3962 I.7320 2 .6995 3 .9 945 2 116 .3682 1.3313 30 e 250 34 0 017009 1.6765 4 I 933 6 116 .3921 1.2844 48 8 267 25 0 017151 1.6586 5 .1 0.4112 1.2474 58 8 281 02 0 017274 50 0 71562 7.1736 26 .6 0.4273 1.2167 1.6440 st 8 292?! 0 017383 1.1905 1.6316 7 .93 0017482 6 1875 6 2050 27 .8 118 g 1 .1 90 .1 0.4534 1.1675 1.6208 sta 312 04 0 017573 5 4536 2903 894 6 118 .4643 1.1470 1.6113 9 ge s 320 28 0 017659 4 8719 4 8953 29 .2 0.4743 1.1284 1.6027 10 e 3 32782 0 017740 4 4133 4 4310 305 8 88 I8 .4834 1.1115 1.5950 11 lit e 334 79 0 01782 4 0306 4 0484 31 .4 0.4919 1.0960 1.5879 12 I 27 0 01789 37097 3 7275 31 .7 0.4998 1.0815 1.5813 130 0 130 0 347.33 0 01796 3 4364 3 4544 325 0 86 .0 0.5071 1.0681 1.5752 14 .6 863 4 119 .0554 1.5695 15 eI 35843 001809 2 9958 30139 3361 859 0 119 .0435 1.5641 ISe O lie B 363 55 0 01815 28155 2 8336 34 .0322 1.5591 11 !s l96 9 0.5328 1.0215. # 1.5543 10 .6 0.5384 1.0113 1.5498 19 les e 37753 0otR33 23847 2 2873 355 5 84 .0016 1 5454 20 .9923 1.5413 tit t 218 s 385 91 0 01844 2 16373 364 2 835 4 1l99 6 0 5540 0.9834 1.5374 22 I 389 88 0 01850 2 06779 208629 199846 368 3 831 8 120 .5336 230 0 238 8 393 10 0 01855 I97991 39739 0 01860 189909 I91769 37 .6 0 5634 0.9665 I.5299 24 .84317 37 .1 0 5679 0 9585 1 5264 250 8 1.5230 26 .5 0.5722 0.9508 288 8 40780 0 01875 169137 I11013 383 6 818 3 120 .5197 218 8 210 8 20 I65049 38 .5805 0 9361 1.5166
- > 280 0 411 01 001880 163169 0DIRR5- 157597 I59482 39 .6 0 5844 0 9291 1.5135 298.8 6 20e g 414 25 394 0 80 .9 0 5882 0 9223 1.5105 30 s 8 417.35 0 01889 I52384 154274 409 8 19 .0 06059 0 8909 1.4968 35 .32554 1.16095 42 .6217 0.8630 1.4847 48 I 444 60 001934 1 14162
.
.. ..
. .
.
_______
___
.
-- - - - - -
-
. . . _ . _ __
Sal Sa Sal Sa Sa Sa Abs PvGss. . s Abs P ess Temp Vapor tb/Sg i Ivan Vapor liquid Evap Vapor Liquid Evap .
I" lb/Sq in Fahi lialuiil s, p I v,i vg hg hgg hg s, s ,,
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55628 0 02195 564 8 622 2 118 I3738 Ilse e 1150 8 56182 0 02214 0 35859 038073 I3683 1200 8 0 02232 0 34013 0 36245 57 l84 8 03714 05%9 itse 3 56119 578 8 603 8 118 I250 8 57238 0 02250 0 32306 0 34556 1758 8 1180 2 0 7843 05733 1.3577 1388 8 0 02269 0 30722 0 32991 5855 594 6 1398 I 577.42 13525 1350e 0 29250 031537 592 3 585 4 117 .3414 1400 8 0 2/811 0 301/8 598 8 516 5 1175 3 03966 0 5507 1400 0 58707 0 02301 117 es 591 70 0 02321 0 26584 028911 605 3 5674 1458 8 0 21119 611 3 558 4 11701 0 8085 *05288 13373 1500 8 1500 I $96 20 0 02346 0 25312 1 3324 15588 0 26601 618 0 549 4 116 .3274 Isle I 0 25545 624 2 540 3 1164 5 0 8199 0 5076 IE!G G 604 81 0 02381 0 23159 1 3225 1658 8 0 24551 630 4 531 3 116 .3116 1700 3 0 21178 0 23607 636 5 522 2 1158 6 0 8309 04867 lite s 613 13 0 02428 64 Ilse s 617.12 0 02450 0 20283 0 22713 13079 1800 O 0 21861 648 5 503 8 1852 3 0 8417 0 4662 1840 0 621 02 0 02412 0 19390 1.3030 1850 0 0 21052 654 5 494 6 1149 0 0 8470 0 4561 II58 0 624 83 0 02495 Ol8558 0 4459 I2981 1996 8 0 11161 0 20218 660 4 485 2 1145 6 0 8522 lies 0 628 56 0 02517 e 12931 19585 0lu999 0 19540 666 3 415 8 114 e 63222 0 02541 12881 2006 8 0 16266 0 18831 67 .3 0 8625 0 4256 feet a 63580 0 02565 12780 210e e 642 16 002615 0 14885 0 11501 683 8 4467 IJ 30 5 0 8727 04053 2l00 0 1122 2 0 8828 0 3848 I2616 2288 8 22*8 8 64945 0 02669 0 13603 0 16272 695 5 426 7 0 15133 107 2 4060 Ill3 2 0 8929 0 3640 12569 228 /121 0 12406 12460 2400 0 00/190 0II/H1 0 14016 719 0 384 8 11037 0.9031 0 3430 2499 8 662 Il 7313 3616 109 II O02859 0 10209 0 13068 0 9247 02977 11225 2544 8 613 91 0 02938 0 09112 0 12110 744 5 3316 1082 0 2808 e 0 2741 12097 27se p 679 53 0 03029 0 08165 011194 151 3 312 3 10693 0 9356 2168 I 0 2491 1.1958 2000 8 684 96 0 03134 0 07111 0 10305 110 7 2851 1055 8 0 9468 2880 0 1.1803 2984 8 0 06158 0 09420 7851 2547 1039 8 0 9588 0.2215 2900 8 690 22 0 03262 1.1619 3000 s 0 03428 0 05013 0 08500 8018 218 4 102 .1373 3180 9 0 031/1 0 01452 824 0 169 3 99 .1460 3100 8 700 28 0 016HI 1.0832 32se 8 0 01191 0 05663 815 5 561 93 .0351 0 0482 3294 I 105 08 0 04412 1.0612 3298.2*
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GEK-83288
- X T / / O MONO AUX TROLLEY' ROD BLOCK d% BACK UP HOIST FUEL HOIST HOIST AUX HOIST INTERLOO NO. I LIMIT INTERLOCK INTERLOG INTERLOCK
/ ~--@
e N g ROD BLOCK BRIDGE INTERLOCK REVERSE BRIDGE FAULT REVERSE LOCKOUT
/
NO. 2 STOP NO.1 STOP NO. 2
,
'
l C O l INTERLOCK STATUS DISPLAY Figure 3- Interlock Status Display Modale
- FIGURE 2 **
.
3-12
.
TYPICAL TURBULENCE Niagara Mohawk ASSOCIATED WITH OVERCAST-S10 Rov. 0R NOC1UR m SITUATIONS HAVING RELA-Cl8$$3 TlVELY STRONG WIND MECHANICALTURBULENCE
-
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Speed Direction
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. _ . __ _ - . - - - . _ - . . _ = _ ___-
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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 COMMENT S THE CENERAL POPUIATION_
RECOMMENDED ACTIONS (a)
g, Previously recommended protective Whole Body <1 No planned protective actions. (b) actions may be reconsidered or
State and County may issue an advisory to p ek terminate Thyroid <5 shelter and await further instruction (y (Child) Monitor environmental radiation level Shelter all ERPA's in affected sectors outi,b 11 constraints exist, special Whole Body 1 to <5 considerations should be given for toapointwherethedoseis<1remwhole7 evacuation of children and pregnant body and 5 rem thyroid. Consider evacuation wome of children and pregnant women in this areas Thyroid 5 to <25 Consider evacuation out to 2 miles radially (Child) from the station. Evacuate this area unless constraints male it impractica ,
Monitor environmental radiation levels. Control acces [
~ ~
conduct mandatory evacuation to 2 miles (.
Seeking shelter waald be an ,
Whole Body 5 and above alternative if evacuation were radially from the station and evacuate 4 not immediately possibl Thyroid 25 and above all ERPA's in affected sectors out to the @ "
(Child) point where dose is <5 rem whole body * Where possible, dose savinga from and <25 ren thyroi '
sheltering in lieu of or in combination with evacuation should be
,
Shelter all ERPA's in affected sectors considere 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 are Monitor environmental radiation levels and
,
" adjust area for mandatory evacuation based on *
.
these levels. Control access .
EPP-26 -6 June 1985
% "
.
. .
y . .
..
.,..
(
5. Plume Exposure Zone (0-10 miles) Protective Actions (Cont. ) (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 sector NOTE: Protective action recommendations will be made for entire ERPA's even though only a portion of that ERP may be affecte Repeat this procedure for each applicable ERFA (those ERPA's f4 affected by the plume). Inform State and County Emergency Operation Centers of protective action recommendations', ' dose projections and sample l4 results using EPP-20 and its associated Fact Shee . Ingestion Exposure Zone (0-50 miles) Protective Actim s 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 1 ( Compare the calculated deposition rate for each affected ERPA l' 4 with the appropriate preventive or emergency protective action guide levels listed on Figure " 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 fee !
" 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 commerc Based on the comparison performed above, use Eigure 7 to f4 determine recommended ingestion zone indicated protective BCtion j Repeat this procedure for each applicable ERPA or location l affected by the plum I EPP-26 -5 June 1985
.
t
. _ . -
. _ - -
.
Figure 3 (cont.)
1984 PERMANENT RESIDENT POPUIATION ESTIMATES EMERGENCY RESPONSE PLANNING AREAS EMERGENCY RESPONSE 1984 PERMANENT RESIDENT PIANNING AREA POPUIATION ESTIMATES 1 137 2 508 3 356 4 620 5 411 6 880 7 818 8 573 9 ^ 455 10 1,117 ?
11 1,423 12 9,145 13 11,238 l 14 122
'
> 15 1,028 )
16 1,692 !
17 577 18 1,048 19 1,003 l 20 1,553 21 2,157 22 6,488 TOTAL 43,349 i
.
a EPP 26-15 October 1984
_ _ _ . . . _ _ .
-- _ - - _
__ . _ _ . _ _ .
Figurn 3 (Cont. ) l TREE 2
ENWune rY EEEPONSE PLANNING AREES_
N m 'ZE SECTORS ET 2, 5 AMD 10 EZLES f f
5 MTr.ma l 10 uTT.ma
-_-_ vR 2 MILES l 28, 29 E l 27, 26 27 l l 29
'
EEE l 27 l l 29 EE i l 29 EEE l 27 l 14, 15, 29 27, 1, 2 l 4, 7 l E l 8, 15, 16, 17 1, 2 l 4, 7 l EEE I 17, 18 1, 2 l 4, 7*,.S*, 9 1 33 l 18, 19, 20 1, 2, 5* l 4, 9, 10 l SSR -l 19, 20 1,2,3,5 l 10 l S l 12, 19, 20, 21, 1, 3, 5 * I 6, 10, 11 1 asW l 1 22, 23, 24, 25 I i 6, 11, 12, 26 l 13, 21, 22, 23, SW l 1, 3 l I l 24 I 12, 22, 23, 28 1, 3, 26 l 6 1 EWW I l 26 l l 28 W l 28 EMW l 26 l
-
l 28 EW l 1 I 26 I l 28 ENN I i I
- I I Earginal for aissan i .
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.
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( EPP-26 -16 Jtme 1983 ,
.
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- %;^",::"L i.e., O" - - .-
Sector M-360 Degrees J.A. FitzPatrick/
,( 10-Mile Radius Nine Mile Point
., . ,,,ne ass Nuclear Power Stations
_-_------ _ _ _ _ _ _ .-.
-
'l
. Figure 4 ,
(*
) Of f-Site Protective Action Recommendation Work Shee t I
. ) Area of Concern (
location /ERPA l miles Distance miles (use Table 2 -
of Figure 3) Direction degrees ERPA
% Ingestion Zone Protective Action NOTE: If determining Recommendations only, proceed to Item 1 . hours Expected release duration
aph miles /hr Wind Speed = hours Plume travel time = (Item la)/(Item 3)
= ( _
)/( )
(a or b) hours Time until exposure begins ,
(a or b) If release has begun:
(
( Time = Item 4 - Time release has been in progress (
Time = ~
NOTE: If Item Sa is a negative number, enter zero hours for Item 5 If release will begin later: l
Time = Item 4 + Time until release j Time = + l Weather condition and season (circle onc for a & b ): Normal or Adverse l 4 Day or Night _ hours ERPA Evacuation time:
Use Figure 3 along with information recorded in Items 1 and 6 to determine the tim (Figure 3) are CAUTION: Times shown on Evacuation Time Estimates
. expressed as hours: minutes and aust be converted into hour EPP-26 -32 June 1985 ! '
i
. _ - - - - - .
.
.
.
' '
....- .......__...................... -_..........._.._...._..........
Titie: RPY Flooding ...__...______....
.__.______..____.___..__......... .___________________
.
ACTIONS INSTRUCTIONS C6-3./ THEN continue in this pro-E 6- WHEN all controls rods are cedur 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 W termined,-eO the foilowf ng l with
systers until at least 3 are open AND RPV SRVs j
^ pressure is not Eering AND is at least 80 psig
-
!
E5ve suppression chareer pressure:
1 HPCS, (.,
.
.
- Fee &ater pumps, i
I LPCS,
,
l
. LPCI, I
.;
- Condensate booster pumps,
. .6 Condensate pumps, CRD,
- Service water to RHR,
,
! . (0P-11, Section H},
Fire System, (0p.43, Section H),
keep full ,
l 4.10 ECCS (OF-32, 33, Section H),
SLC test tank , (OP- 36, 4.11
Section H),
boron tank ,
4.12 SLC (OP_36, Section H).
(
s AND ou au u .
- Fl
. _ ()
.cr
_
>
(
'
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____.._--______ _-___.. _______.-___._____..__...__-____............ _ (
. . ____ __ _ _______..._____.._________
______.______.____________
l C6 5 Paintain at least 3 SRVs open and RPV pressure at least 80 psig G
' above suppression chamber pressure by throttling injectio ACTIONS INSTRUCT IONS
.
,
determined,->O wi th the following sys-
!
teris until RPV watcr level is increasing:
i HPCS, O Fee &ater pumps, O f
~6. 3 LPCS,0
. LPCI,' O l Condensate pumps O (
b .6 Condensate booster pumps, O ,
k f l CRD,C RlR Service Water Tie (OP-11, Section H), C ,
Fire System (OP 43, Section H),D 6.10 ECCS keep full (OP- 32, 33, Section H), O (OP_36, 6.11 SLC test tank Section H),0 (O P-36, 6.12 SLC boron tank Section H). O p
A
.
( _
H2 ev rf Page ' cf 5 Rev. uu ry . l 19e-
-
. _ . .
.
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'
.
l 1 VESSEL PRES RESE toest 8 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 75
'
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-2s 30 40 go 0 10 70 O 20 30 40 so TIME isect O to flME (sect
' 8 VOIDREACTivtTV
~
t LEVEL tmch sel s*D thertl 2 DOPPLER RE ACTevlTY 2 VESSEL STE AMF tow 2 .I 3 SCRAM RE ACTivlTY 3 TURBINE STE AMF LOW
- 4 TOT AL RE ACTIVITY 4 F E EDW AT E H F t OW g im -
4 50 3
8
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- _2 30 40 M 0 40 50 0 10 20 0 10 70 30 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
_ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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PAGE 17 TH -----------OF
---_-_EORY -_
NUCLEAR POWER PLANT OPERATION,
-- ------
FLUIDS, AND
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IHERgggyNgglCg-86/07/21-CRESCENZO, ANSWERS -- NINE Mll.E POINT 2 " -
2 ,.s .
ANSWER 5.01 (2.50) As the reactor operates during the early part of the cycle, the burnable poison depletes more rapidly than the fuel, therefore, (1.00)
control rods must be inserted to hold the power constan 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(1.50) this
,
effect, the void coefficient becomes more negativ REFERENCE FJC 15 NMPC Operations Technology vol. I page 5-15, LO 2-5 Hope Creek LP RXPH19-01 pg.6 LO #4, RXPH28-01 pg.6 and trans. #1 LO #3 I 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. i 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)
scal Yes, criticality would be achieved with no change in critical (1.00)
rod densit REFERENCE NMPC Operations Technology pg. 2-3, Theory .1 FJC 68 PAGE 18 THEDRY_OF NUCLEAR POWER PLANT _ OPERATIONt FLUIDS 1_AND THERMODYNAMICS
- .---- -- - _
ANSWERS -- NINE MILE POINT 2 -86/07/21-CRESCENZO, g -
ANSWER 5.04 (3.00) _ Peripheral rod worth will increase [O.33 because the highest xenon concentration will be in the center of the core [O.33 This will where the highest flux existed previously [O.3 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.3 More than one half the value at 1007. power [O.5 The Xenon production rate is directly proportional to power i level, but removal rate is proportional to Xenon concentration i 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 pp , 6-1 FJC 78 l
Pilgrim Rx. Theory pgs. 6-7, 6-12 Nine Mile 2 Rx. Theory pg , 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 reacto Thus the size of the control rod is effectively much larger than its actual physical size would suggest,.and its ~
rod worth ig larg (2.50)
o i o o s v ., N ek s s s s M I ,3 c LL I1 tw k Sit % d c ~-
REFERENCE Millstone Rx Theory pg. 5-17 FJC 80 Pilgrim Rx theory pg. 5-17 Nine Mile 2 Rx Theory pg. 5-17, .6
I l
l l
l l
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AND PAGE 19 5.__THEggy_gF NUCLEAR POWER PLANT OPERATIgN, _ FLUIDS, _
IHEggggyNQUICS a,
-06/07/21-CRESCENZO, ,,
ANSWERS -- NINE MILE POINT 2
...
'
i (2.00) 2 p, ', m .. n , ; ., ,, e ,e _3 u-<
ANSWER , 5.06 IN
'
.l p: g -
.
.\ . . x ,<_ c . 3 y s. te 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 stea REFERENCE FJC264 Hope Creek M.O.C.D. LP 104 pg. 15 1.0. #1 Nine Mile 2 MOCD pg 7-8, S.L.O. 111.2,3 ANSWER 5.07 (3.00)
.- ~
+ <
'~
"r
'
'
'
'\0 ' ' ~ '\ k
'
' s* f}) u
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,
a. FW temp o -
...
<. >
,
FW flow ,,, e,, .'mn
~
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 FW flo (1.00)
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 Pilgrim reactor theory pg 3-14 FJC 293 Nine Mile 2 reactor theory pg. 3-14, S.L.O. 2.3,4 i
l l
- THEORY pf NUCLEAR POWER PLANT OPERAT ION t FLUIDS t_AND PAGE 20 THERMOBfRAMICS
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. . ,
ANSWERS -- NINE MILE POINT 2 -86/07/21-CRESCENZO, !
ANSWER 5.09 (2.00) l l
(0.5) Increases l
- (0.5) Decreases (0.5) Decreases (0.5) - Increases .
REFERENCE HC HT&T No. 11, Learning Objective 2, pages 8 and FJC 294 Pilgrim THT&FF pgs. 9-26 thru 9-30 i I
Nine Mile E THi&FF pgs. 9-26 thru 9-3 ~'
pov-c-
ANSWER 5.10 (2.50)
% 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) As reactor pressure increases, core inlet temperature remains essentially constant thus inlet subcooling increases and j
.
'
,
decrpases as a function of rx pressur cJ lj ,.4 ,q?uf i, (1.00)
e g_ n\ -
,, qsry q,,bw ,. ; ( o s g t .g j, - u . },t \ ,s c
\ ,s + , G cN $~ 4
- 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
- MCRD x HCRD (1.00)
Qcore = 6423000 x 1194 +110000(506-419) + 2040000 - 6400000 x 345 (1.00)
-26500000 - 23000 x 68 = 5444600000 Btu /hr = 1595 MWt REFERENCE Vermont Yankee Nuclear Power Corporation, SCRO-02-115, Reactor Heat Balance NMP 2 Thermo chap 8 FJC 347
._
_ . _ _ _ _ _ _ _ _ _ _ _ , . . . -- _ PLgNI_glgl{gg_Q{glGN, CONTROL, _AND_ INSTRUMENTATION PAGE 21
__
ANSWERS -- NINE MILE POINT 2 -86/07/21-CRESCENZO, l l
l ANSWER 6.01 (3.00) l The system sees low feed flow and will increase feedflow until ! level error balances the feedflow/steamflow mismatch at which (1.00)
l point level will stabilize at some new higher leve ; 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 leve (1.00) ; Feed flow will increase and continue to increase until the reactor I high level tri (i,e\,;ts t) (1.00) l
REFERENCE l NMPC Operations Technology FWCS pg 4 EO #6 FJC 35
\
. l ANSWER 6.02 (1.00) lj l
d g i
I REFERENCE l Hope Creek LP 14-01 pgs. 9, 10 .b FJC214 I Pilgrim LP IRM pg. 7 l Nine Mile 2 L.'P. IRM's 64 6 and 10 of 12.
ANSWER 6.03 (3.00) TGV remain at 100% due to load limit (.50) HPV open 5% due to max combined flow (.50) l Power decreases due to lower pressure (.50) Pressure decreases due to BPV (.50)
FINAL TCV at 100% position (.25) UPV shut (.25) power lower (.25) pressure slightly lower (.25)
REFERENCE Hope Creek LP 1.0. # 3,4,10 FJC 286 l
Nine Mile Point 2 EHC L.P. objective EO-6 l
I l
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PAGE 22 6 3__ PLANT _ SYSTEMS _DESIGNt__ CONTROL 1__AND INSTRUMENTATION ANSWERS -- NINE MILE POINT 2 -86/07/21-CRESCENZO, F.
ANSWER 6.04 (2.00)
+ 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%
1imit switc ( vi i
w m 'i t
? \im' " ( u . '. b bc ,5 13 d i .) 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 pover the vesse hs o r 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 loade t- 0I ' ' 'd O' ' ' ' ' 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 hoist ( r, .. wr r . _
REFERENCE GE Refueling Tools Familiarization Manual #2, pages 3-12, 3-13 FJC 312
'
NMP2 S.L.O. # E. I
~
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 (1.50)
separator ( cv gw s.f 4 .s y s h c. , d )
'
REFERENCE NMP2 Oper Tec, HPCS rev 2 pg 9 of 10, EO 5-3 FJC 320 l l
ANSWER 6.06 (2.00) Meters fail downscale. Recorders fail as is.' (1.00) ECCS will not initiate because logics are " energize to operate" (1.00)
'u a ' '
\T 't \, .t e \ v 4 &. vs .
en
.
- < (. 5
. -g .
(v s m.( , 3 . .m) x q
.
L'
s
..,-- - .. - -
- PLANT SYSTEMS _ DESIGN, CONTROL, AND INSTRUMENTATION PAGE 23 ANSWERS -- NINE MILE POINT 2 -86/07/21-CRESCENZO, REFERENCE NMP2 MOCD pg 75 of 80 E.O. #'s 1,2 FJC 348 ANSWER 6.07 (2.00)
see attached diagra REFERENCE NMP2 APRM LP Pg. 4 of 14 E.D. 2 FJC 349 1,k ANSWER 6.08 42rBO)
a. TRUE-M'-TRUE ----- ,
g r) .
}
<
((
\
-
c. TRUE d. TRUE e. FALSE REFERENCE NMP2 RWM LP pgs. 3,5 of 26 and table 2 E.O. 5 FJC 350
.s
-.4 ANSWER 6.09 (3.00) ,
'*
a. DG bkr trips on LOCA(diesel goes f to emergency mode,. when LOOP l occurs, breaker closes and(loads sequence as normai (1.50) l l
b. The offsite breaker will stay closed and diesel will attempt l to pick up offsite test loads. Directional current trip l 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 3__P(ANI_@y@IED@_@@@l@dt_@@dIB@67 AND INSTRUMENTATION PAGE 20 ANSWERS -- NINE MILE POINT 2 -86/07/21-CRESCENZO, . i ANSWER 6.10 (2.00) h Assuming 95A open, 94A closed in' auto. Valve 94A will samuL open due to EDG east running Valve 95A wi11 remain ope (1.00) Valve 94A will open as expected, once diesel is runnin after 1 minute time delay with service water pressure less than 20-psi valve 95A will clos (1.00)
- ---. ; . . ,
I ~
REFERENCE nmp2 FSAR figure 9.2-2 vol 18; service water LP EO 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 vesse (1.50)
0.95)- l
--trl .-to prevent-excessive-Dh' across tWe val ~v /~2.' To- ensure -the-reactoNoeh not draindadhe suppress 1 anr '
(Os-75) - l
- \
REFERENCE L)A y O P \
-
NMP2 op 33 and LP LPCS; E.O. #'s 2,9 FJC 353 l
l l
l
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i l
1 I
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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 25 RAD 19LggICAL CONIBQL-86/07/21-CRESCENZO, ANSWERS -- NINE MILE POINT 2 ANSWER 7.01 (1.50)
(0.75)
1. Health physics must sample the atmosphere in the containmen (O.75) The initiating condition must be correcte REFERENCE FJC 41 i N2-IOP-83 PCIS pg. 3 i l
i 4 ;
ANSWER 7.02 (1.50) 1 (O.50) High neutron flux alarm and/or scra At or near 1007. rod line, min recirc flo (0.50) '1.00; Insert control rods per reactor analyst or increase recirc flow .
( REFERENCE N2-OP-92 Neutron monitoring, precautions and off normal procedure FJC 47
.
ANSWER 7.03 (2.50)
a. Downwind direction will be NE. Table 2 of EPP-26- lists% C-(no ERPA within e m * M 1(1.50)
the 5 mile zone. ( mq +tei i3 MA a '3mmi d e a . 2- 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 mandator l ( Li (C )
l REFERENCE NMPC procedures EPP-8 and 26 FJC 302 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 baro .. -_ .-. . . . . - _
. . _ ,
PAGE 26 PROCEDURES _ _NORMAlt ABNORMALt_ EMERGENCY AND RAD _IOLOGICAL CONTROL
___ .
ANSWERS -- NINE MILE POINT 2 -86/07/21-CRESCENZO, I
!
REFERENCE N2-EOP-C7 LP pg 19 of 2 FJC 303 Mh , %
ANSWER 7.05 (3.00) BX is the power at which a reactor will stabilize during a full power ATWS and level is low enough to inhibit natural circulatio (flow stagnation power level). Further level reduction will result (2.00)
,in uncovering the cor 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% powe 'with no clad temperature _in excess of 1500 (1.00)
.. - - , i j, ,s d.>s_, .,
,
e mi TAF (1.00) Spray cooling- one core spray at or above design condition (1.00) Steam cooling- proper pressure and steam flow through SRV' (1.00)
REFERENCE NMP2 EOP structure page 3- FJC 305 ANSWER 7.07 (1.00)
Highest value of a parameter beyond which operation of '
3,3 p \
equipment impagtant to safety cannot be assume (;< 7c, (1.00)
oCt* 2 A is h, Ok 4- ,, o s- s , w m. s :- i(.(e j REFERENCE NMP2 EOP Structure page 3-5 FJC 306
7 PRgCEDURES __ NORMAL, ABNORMAL t EMERGENCY _AND
-
' p ~~
e
~
PAGE 27
ROpig6991Cgh CONTROL ANSWERS -- NINE MILE POINT 2 -06/07/21-CRESCENZO, 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 mi after scram from 1007. powe (2.00)
' REFERENCE N2-EOP-C6 LP pg 13 of 19 FJC 307 ANSWER 7.09 (2.00)
a. To prevent cavitation of the flow control valv (1.00)
b. To prevent cavitation in the the jet pump (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 shieldin (1.00)
b. Do not hold load at or near zero speed for more than 90 sec to-I prevent motor overheatin (1.00)
REFERENCE NMP2 procedure N2-OP-39 FJC 317 ANSWER 7.11 (1.00)
Both valves are powered from the same power sourc (1.00)
REFERENCE N2-IOP-62 FJC 319
_ _ _ . PROCEDURES _b_ NORMALt _.. ABNORMAL, EMERGENCY AND PAGE 28 g@DIOLOGICAL_ CONTROL ANSWERS -- NINE MILE POINT 2 -86/07/21-CRESCENZO, .
ANSWER 7.12 (3.00)
.ms Thermal stratification as noted by axial skin temperature variation (1.00)
b. Raise level to above the steam' separators (202") for natural
circulation or maintain SDC flow normal and throttle service wate (2.00) !
I'
REFERENCE NMP2 N2-IOP-101c pg. 6 FJC 354 l
l
,
-- ,c- - --.
. - - . , v Ei__gDglNigIggIlyg_ PROCEDURES, _ CONDITIONS, _AND_ LIMITATIONS PAGE 29 ANSWERS -- NINE MILE POINT 2 -86/07/?lI CRESCENZO, ANSWER B.01 (3.00) .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 hour8.333333e-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- (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. (1.00)
REFERENCE NMP2 T.S. 3.4.2, 3.5, (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. 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 MCPR to limits of figure 3.2.3.~1 when >257. power.(T.S. 3.0.4 is not appl,icablehtherefore can continue power increase [until limited
'
by MCPR.; s (2.50)
REFERENCE NMP2 .2.3, 3. FJC 310
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s PAGE 30 8t__ODMINigIggIlyE_PBggEgySEgz_ggdpillgNyt_@NQ_LIMil@IlgNy
- NINE MILE POINT 2 -86/07/21-CRESCENZO, ANSWERS
.
ANSWER 8.04 (2.50)
T=FRTP/CMFLPD if T is less than 1 then APRM setpoints are lowered by a factor of 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) no yes (0.25 ea) yes Suspension of core alterations shall not preclude completion of the(1.00)
movement of a component to a safe conservative positio REFERENCE Nine Mile 2 Technical Specifications, Definition Section FJC 327 ANSWER 8.06 (3.00) .25 REM /QTR REM /QTR NTE 5(n-18) .75 REM /QTR REM /QTR '.>
/ i, 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
PAGE 31 91__OEU191E10011YE_EO9EES90EE1_E9dE11190EE-AND LIMITATIONS ANSWERS -- NINE MILE POINT 2 -86/07/21-CRESCENZO, F.
, ANSWER O.07 (1.50)
B)
~ CFR2O appendix (0.50)
b. Provides assurance that the levels indicated will result in exposures within the limits of 10CFR to a member of the public and general populatio (1.00)
REFERENCE 10CFR2O and NMP2 Tech Specs 3/4.11.1.1 and base 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 exist hw ' ' s t .**\ '
\ *
- - C,.
' Dvd' , (1.00)
b. Prior approval from a licenged SRO and(if time permits SORC review and NRC notificatio (1.00)
REFERENCE NMP2 Procedure AP-4 FJC 331 14: .
@g__ADMINISTRATIVg_PRgCEDUREgt_CgNgITIgNgg_AND LIMITATIONS PAGE 32 ANSWERS -- NINE MILE POINT 2 -86/07/21-CRESCENZO, 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 necessar (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 FJC 333
.
..
.. ..
.. ._
. . .
.
TEST CROSS REFERENCE PAGE 1 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 FJCOOOO35 O6.10 2.00 FJCOOOO352 06.11 3.00 FJCOOOO353 25.00 07.01 1.50 FJCOOOOO41 __
%-
07.02 1.50 FJCOOOOO47 .z FJCOOOO302
~
07.03 2.50 ,. a=~
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
. , , . . , - _ .
. %u
_ -
.
..
. .. . .
TEST CROSS REFERENCE PAGE 2 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
^~
. - .
9 .y M"m x
_ _
. .
p______= ______________________= ________________________________q l FLOW CHANNEL i
' 7 I I COMPUTER I l 'N -
MODE x ISOL I
-
ISOLx SWITCH I j
l I ( A + D ONLY ) -
/
I
/ MODE N l L' P A l 'N MODE 8 '4 ,
I C
-
g N/ AND -
UPSCALE P lRC i s'x 1 SUMMER s 1 TEST -
N I W /N SQOARE, TEST - -
I F l / N T RIPS'N ROOT N-
-
' I
,- , SWITCH \ SWITCH '
-
X FR , ..N I
-
' ' -
'
l'
CONVERTER i l
l<
'h/, ' -
,ye MODULE MODE RELAYS l
l
/r l I
-
/
- INTERLOCK SW KI-K4 l , ' SQUARE
' '
l g
L PB R - - -
COMPARATOR , l R .lRC , _ CONVERTER -
TRIP q l I
TE , I W 'N F
X:> (R I
l S
/ ITC H'N 20V IN,0,P '.M --
-
p _ - _ _ _ _ _ _I g
- RBM I c
,
COM K2 r N FLOW REF OFF COMPUTER l
I N = -NORMAL ALARMS
-
ISO i =/ 8 ROD BLOCKS l
( A+ D ONLY) <~ ISOL = APRM TRIP g _ _ _ _ _ _ _J
,
bj REFERENCE l ,
' ISOL l
= TO FLOW COMP
,
UNIT (SEE TABLE 4)
!
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q ISOL l IOP LOOP O BYP~
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- tED) (BLACK) l BYP I TABLE 4) I- - - - - - - - - - - - - - - I FLOW UNIT BLOCK DIAGRAM RE PC FLOW REC P602
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ATTACHMENT 2 TO ENCLOSURE l
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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 key Question N .0. 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 answe .0. 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 gradin . COMMENT: " Color coding is same for both AC and DC, which is implied and should not be required for full credit".
j- RESOLUTION: Comment accepted. Discussion of color coding for current type will be deleted from the answer ke . COMMENT: "First part of question is never answered in exa Key, should be "No" 1) Pressure remaining high is fals 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. COMMENT: "Part a. Note, if transmitter power is lost, the recorder will fail downscale, .if lose recorder power, it fails as i 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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0FFICIAL RECORD COPY OL NM 2 EXAM RPT - 0010. /01/86
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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 failur .1 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
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initiate Recommend changing answer to " valve 94A opens".
RESOLUTION: Comment accepted. Answer key changed to reflect correct ~
answe .0. COMMENT: " Question asked for basis of staying above minimum alter-nate flooding pressure, not necessarily the definitio 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
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discussion than " adequate core cooling" must be included for full credit.
i 8. COMMENT: Answer key has Remote Shutdown Panel L.C.0. of 7 day .
This specification was not supplied to examinees in their 1 packet, so answer should be restricted to Spec 3.5.1 14-day l LC0 onl RESOLUTION: Comment accepte Reference to R.S.P. deleted from answe l l
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0FFICIAL RECORD COPY OL NM 2 EXAM RPT - 0010. /01/86 . _ . . _ _ _ . __ . _ . _ - _ . ._ _ -- - -- - ~ _---- - - - . - - ,
ATTACHMENT 3 TO ENCLOSURE i
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Facility Comments and NRC Resolutions on Written Examination made after Exam Review Question N_L.
(SRO 6.08) 2.11 COMMENT: Should be either True or Fals 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 affecte Not Affected:
NMS - APRM ino RBM - All but downscale Ref: RMCS Ops Tech, Table 1 RESOLUTION: Comment accepte Part "b" deleted from the answer ke (R0 4.09) 6.11 COMMENT: 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 LPC ) LPCS test return returns to the suppression pool via M0V30A which is the RHS loop A suppression pool retur 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 lineu If person answered using either of the cross connects above, part b makes no sens Ref: FSK 27-7 (RHS System)
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l 0FFICIAL RECORD COPY OL NM 2 EXAM RPT - 0012. l 10/01/86 l
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Question N .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 l discussion of the jockey pump as a system cross l connec Part "b" will be deleted since this '
mode is infrequently used and as such operators ,
need not maintain familiarity with these l caution l 17.04 COMMENT: The answer in key is one of three valid answer Others are:
1) Per E0P-C7, the normal control band is 15 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 ban ) RQ is being utilized simultaneously with C 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 ke OFFICIAL RECORD COPY OL NM 2 EXAM RPT - 0013. /01/86