IR 05000245/1988016
| ML20246P392 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 03/17/1989 |
| From: | Conte R, Walker T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20246P386 | List: |
| References | |
| 50-245-88-16OL, NUDOCS 8903280213 | |
| Download: ML20246P392 (150) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT N (OL)
FACILITY DOCKET N FACILITY LICENSE N DPR-21 LICENSEE: Northeast Nuclear Energy Company P. O. Box 270 Hartford, Connecticut 06141-0270 FACILITY: Millstone Unit 1 EXAMINATION DATES: November 7 - 10, 1988 CHIEF EXAMINER: M bb T. Walker? Tenior Operations Engineer
_ 3/& /P9 Date APPROVED BY: MU C" foA Rf'c hard J. Co'rit~e, Chief, BWR Section 2//7[f7 Dats '
Operations Branch, Division of Reactor Safety SUMMARY: Written and operating examinations were administered to two (2)
reactor operator (RO) candidates and seven (7) senior reactor operator (SRO) candidates. All RO candidates passed the examinations. Six (6) SR0 candidates passed the written examination. One (1) SRO candidate failed the written examinatio Six (6) SRO candidates passed the operating test. One (1) SRO candidate failed the operating tes PDR V ADDCK 05000245 pg
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DETAILS TYPE OF EXAMINATIONS: Replacement EXAMINATION RESULTS:
E RO E SRO E E Pass / Fail E Pass / Fail E I E E E E E E EWritten E 2 / 0 5 6 / 1 E E E E E E E E E E0perating 5 2 / 0 5 6 / 1 E l 5 I E I I I E E E0verall 5 2 / 0 E 5 / 2 E I
E E E E CHIEF EXAMINER AT SITE: T. Walker, Senior Operations Engineer OTHER EXAMINERS: D. Florek, Senior Operations Engineer J. Hanek, Examiner (EG&G)
M. Spencer, Examiner (EG&G) The following is a summary of generic strengths and weaknesses noted on the operating tests. This information is being provided to aid the licensee in upgrading license and requalification training program No licensee response is require STRENGTHS
- Familiarity with in plant equipment and operations
- Knowledge of administrative procedures l WEAKNESSES l
- Informal communications
- Verification of step completion when using the Emergency Operating Procedures (EOPs)
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3 The following is a summary of generic strengths and deficiencies noted from the grading of the written examinations. This information is-being provided to aid the licensee in upgrading-license and requalification training programs. No licensee response is require STRENGTHS a. Understanding of plant parameter effect's on Shutdown Margin -
Questions 1.03 and 5.03 b. Understanding of level instrument calibration --Questions 1.05 and 5.05 l c. Understanding of Xenon effects during a startup - Questions 1.07 and 5.07 d. Ability to use Steam Tables - Question 5.11 e. Knowledge of Standby Gas Treatment System automatic start signals - )
Questions 3.01 and 6.01 f. Understanding of APR System logic - Questions 3.03 and 6.03 g. Understanding of equipment response to a FWCI initiation signal -
Questions 2.04 and 6.04
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h. Understanding of Diesel Generator voltage response - Question 6.08 i. Understanding of conditions which require Primary and Secondary Containment integrity - Question 2.11 j. Understanding of emergency generator start and trip signals - .
Question 3.08 1 k. Knowledge of the local indications of a tripped breaker - Question 3.11 1. Knowledge of the immediate actions for a loss of station AC power - 'l Question 4.04 m. Understanding of the use of tags - Question 4.05 n. Understanding of radiation exposure control - Question 4.10
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o. Ability to recognize a safety limit violation - Question 8.05 p. Understanding of Gas Turbine Generator Technical Specifications -
Question 8.10'
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WEAKNESSES a. Understanding of the effects of a Load Reject and Turbine Generator Runback - Questions 3.02 and 6.02 b. Understanding of overpressure protection and NPSH considerations for the Core Spray System - Questions 2.06 and 6.06 c. Understa 'ing of RPS and PCIS signals - Questions 2.07 and 6.07 d. Understanding of LPCI loop selection logic - Question 6.10 e. Understanding of Reactor Manual Control System and Rod Worth Minimizer functions - Questions 3.04 and 3.07
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f. Knowledge of the indications and immediate actions for uncontrolled power oscillations - Questions 4.01 and 7.01 g. Understanding of the purpose of steps in EC? 575, RPV Level /Rx Power
- Questions 4.03 and 7.03 h. Knowledge of Shift Supervisor and Supervisory Control Operator functions in the event of a control room evacuation - Question 7.07 1. Understanding of emergency classifications - Question 4.09 j. Ability to determine deportability time requirements - Question 8.13 5. The following personnel were present at the exit meeting of November 10, 198 NRC Personnel D. Lange, Chief, BWR Section, DRS T. Walker, Senior Operations Engineer D. Florek, Senior Operations Engineer W. Raymond, Senior Resident Inspector L. Kolonauski, Resident Inspector Facility Personnel J. M. Black, Director, Nuclear Training R. Lueneberg, Supervisor, Operator Training M. C. Jensen, Assistant Training Supervisor D. Meekoff, Senior Instructor R. Palmieri, Unit 1 Operations Supervisor
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l 5- i A review'of the written examination was held with facility representatives following administration of the written examinatio The facility was reminded to send their written comments to the NRC and j EG&G within five' working day There were some problems wi... occess to the plant during the plant !
tours. The problems were resolved and no problems were encountered during administration of the operating tests. Operations and training personnel were cooperative during the examination proces i l The generic strengths and weaknesses noted on the operating tests (see Section 3 of this report) were presented. Due to the identification of ,
the weakness in verification of step completion when using the E0Ps, a i meeting was held with the licensee to discuss what was expected of the operators with respect to use of the E0Ps during plant transients. The j specific observations on use of the E0Ps in the simulator during the ;
examination were attributed to the examination process. Shift crews are J expected to verify that all required actions have been performed.-
(Subsequent to the exit meeting several shift crews were observed during {
requalification examinations and it was verified that the licensed i operators performed as expected with respect to use of the E0Ps.).
During the examination the examiners noted several cincerns related to the E0P Wide range level indication is used to determine reactor vessel water level during an ATWS situation when water level is lowered to control l reactor powe Because the wide range level instrument,is calibrated I
for cold conditions, actual water level is approximately five feet higher than indicated water level at high pressures. This could result in non-conservative action with respect to lowering vessel water level during level / power control. It could also cause premature entry into emergency depressurization during a loss of high pressure feed with the reactor at high' pressur The flowpath for alternate injection of boron using the control rod drive system appears to pass through a 50 micron filter and thus will not result in adequate flo The torus water level indication chart paper scales do not match the scale on the meter indicatio Based on a telephone conversation between Mr. D. Florek, NRC, and Mr. Palmieri, Northeast Nuclear, on March 14, 1989, the licensee plans short term and long term corrective actions. The short term actions are to make facility procedural changes to the E0P's to account for the error created by reactor vessel water level cold calculation coincident with implementation of Revision 4 to the E0P's. Implementation of Revision 4 shall also address the other concerns noted above (tentatively scheduled forSeptember,1985). In the long term the licensee is considering hardware modifications.
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l Licensee actions to address the concerns will .be. reviewed in a subsequent inspections on the-implementation of Revision 4 to the E0P' j i
The results of the examination would not be discussed at the exit meeting
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but would be contained in the Examination Report. Every effort would be made to send the candidates' results in approximately thirty day . -Management Meeting The following personnel were present at the management meeting on January 18, 1989:
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NRC Personnel l
R. Gallo, Chief, Operations Branch, DRS 0. Lange, Chief, BWR Section, DRS T. Walker, Senior Operations Engineer D. Florek, Senior. Operations Engineer L. Bettenhausen, Chief, Projects Branch 1, DRP Facility Personnel R. Lueneberg, Supervisor, Operator Training R. Palmieri, Unit 1 Operations Supervisor The meeting was held at the NRC Region I office to' discuss the facility comments on the written examinations. The NRC questioned the facility I comments since many of the comments generated were not identified by th facility during the review held immediately following the administration of the examinations. The facility stated that the additional comments were the result of the facility debrief of the candidates after administration of the written examination subsequent to the post examination review. The facility indicated that their comments were ;
intended as constructive criticism, not as an attack on the validity of 1 the examination The results of the examinations were discussed at the meeting and th NRC asked about the training that was provided to the candidates. The i facility indicated that all candidates had received identical training j and had met the standards for participation in the NRC examinations. The watchstanding capabilities of the upgrade candidate that did not pass
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the SR0 examination would be assessed prior to his resuming duties as a ,
reactor operato ,
Attachments:
Reactor Operator Written Examination and Answer Key , Senior Reactor Operator Written Examination and Answer Key l Facility. Comments on Written Exa~minations NRC Response to Facility Comments Simulation Facility Fidelity Report L_-_--_-_____=________-_________-_-_-_____-____-
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U. NUCLEAR REGULATORY COMMISSION
, REACTOR OPERATOR LICENSE EXAMINATION FACILITY: Millstone 1 q REACTOR TYPE: BWR-GE3
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. - 6 j j d DATE ADMINISTERED: 88/11/07
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I EXAMINER: NRC REGION I _
CANDIDATE LMSIRILQTIONS TO CANDID &TEL
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Use separate paper for the answers. Write answers on one side onl Staple question sheet on top of the answer sheet Points for each The passing- ;
question are indicated in parentheses after the question. grade of at !
grade requires at least 70% in each category and a final least 80%. Examination papers will be picked up six (6) hours after I the examination start ,
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% OF CATEGORY % OF CANDIDATE'S CATEGORY
.__YAld!E_ TOTAL SCORE VALUE CATEGOBY p$A% l
._25.00 N PRINCIPLES OF NUCLEAR POWER
- " PLANT OPERATION, THERMODYNAMICS,
,p r.to HEAT TRANSFER AND FLUID FLOW 2 y, so :2th & - n J - 25.005 M PLANT DESIGN INCLUDING SAFETY 20.$l AND EMERGENCY SYSTEMS 2v. 2r 26,64- ru-2 5 d S4 e5~;$$* INSTRUMENTS AND CONTROLS
$ ;24,43+
PROCEDURES - NORMAL, ABNORMAL, E .GG s$*u edC . s EMERGENCY AND RADIOLOGICAL M'M CONTROL SW %
I _140 -7 " Totals l Final Grade All work done on this examination is my ow I have neither given nor received ai Candidate's Signature 2 O% oua 2 01 c. m MJ'IpS Sf f(y
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NRC RULES AND GUIDELINES
_ FOR LICENSE EXAMINATIONS During the administration of this examination, the following rules apply: Cheating on the examination means an automatic denial of your application and could result in more severe penaltie . 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 only to facilitate legible reproduction l Print your name in the blank provided on the cover sheet of the examination.
l Fill in the time you START and STOP on the cover sheet of the examination.
l Use only the paper provided for answer . Print your name in the upper right-hand corner of each page of the exa . The exam has one question per page. Write the answer beneath the question (start just below ***** CATEGORY . . ) . Write only on_ one_
side of the exam and any extra answer sheet . Number each answer continued on additional paper as to category and number, for example,1.4, 6.3.
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l 1 Attach continued answers to back of question to which it applie l 1 Place finished answer sheets face down on your desk or tabl i 1 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 all 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 DO NOT LEAVE ANY ANSWER BLAN . If parts of the examination are not clear as to intent, ask questions of the examiner onl _ - _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in i completing the examination. This must be done after the examination ,
has been complete ]
18. When you complete your examination, you shall: Assemble your examination as follows:
(1) Exam questions with answers on to )
f (2) Exam aids - figures, tables, et I (3) Scratch paper used during the examinatio Turn in your copy of the examination and all pages used to answer the examination question Turn in all scrap paper and the balance of the paper that you did not use for answering the question ,
l Leave the examination area, as defined by the examiner. If after l leaving, you are found in this area while the examination is still in progress, your license may be denied or revoke ;
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'.1 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION t Page 2 j IBEEMQDYEAMES. HEAT _IBANJEER_ANE_ELUID._ELOl
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l QUESTION 1.01 (2.00)
Will control rod worth INCREASE, DECREASE, or REMAIN THE SAME for each of the following? Increasing moderator temperature Increasing the percent voids Increasing the fuel temperature i Increase in Xenon concentration following a power change l
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QUESTION 1.02 (3.00)
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Concerning Prompt and Delayed Neutrons STATE whether EACH of the ,
following is TRUE or FALSE: l The percentage of delayed neutrons produced from fission increases as the age of the core increase (0.5) The energy level at which delayed neutrons are produced categorizes them as thermal neutron (0.5) Neutrons produced from the moment of fission until 10 E-14 seconds are considered promp (0.5; i Delayed neutrons will not cause fast fission of U-23 (0.5) Delayed neutrons are. the major f actor in- determining the rate of cla9' of
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reactor power 10 seconds after a scram o
REGION __I CANDIDATE LEEIBEllONS TO_ GANDIDAIF Use separate i
' Staple question sheetpaper for the answer Write answers on one side onl on top of the answer sheets.
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l grade least 80%.
requires at least 70% in each categoryquestion .
The passin are indicate grade of at-the examination starts. Examination s papers will be picked up six and (6) a hours final after 1 i
CATEGORY % OF % OF CANDIDATE'S CATEGORY {
VALUE_ __ TOTAL SCORE
__VALUE__ CATEGORY
_21,10__
4_ " _
_ THEORY OF NUCLEAR POWER PLANT g2 o y, q3 OPERATION, FLUIDS,AND o, % THERMODYNAMICS i
_25 C C+ __ N _ Eco .pc,w _
PLANT SYSTEMS DESIGN, CONTROL, 2 - ;,. ; a 2c.-Nf-H AND INSTRUMENTATION
_4dAO _Mr&lFA _ _ _ ./ PROCEDURES - NORMAL, ABNORMAL,
,7rr.st ew* qq;M * EMERGENCY AND RADIOLOGICAL CONTROL
_25cO&f _W__ _
__ ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS M" ,
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All work done on this examination is my ow .i nor received ai I have neither given I
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NRC ltULES AND GUIDELkNES-FOR LICENSE EXAMINATIONS During the adminstration'of this examination, the-following rules apply: Cheating on the examination means an automatic denial of your application and could result in more severe penaltie . 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 only to facilitate legible reproduction . Print your name in the blank provided on the cover sheet of the examinatio . Fill in the time you START and STOP on the cover sheet of the examinatio . Use only the paper provided for answer . Print your name in the upper right-hand corner of each page of the l exa . The exam has one question per pag Write the answer beneath the question (start just below **"* CATEGORY . . . ) . - Write only on one side of the exam and any extra answer sheet . Number each answer continued on additional paper as to category and :
number, for example,1.4, . Attach continued answers to back of question to which it applie !
1 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 all calculations, methods, or assumptions used to obtain an answer to mathematical problems' whether indicated in the-question or no . Partial credit may be give Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAN ~
1 If parts of the examination are not clear as to intent, ask questions of the examiner onl !
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17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been complete . When you complete your examination, you shall: Assemble your examination as follows:
(1) Exam questions with answers on to (2) Exam aids - figures, tables, et (3) Scratch paper used during the examinatio Turn in your copy of the examination and all pages used to answer the examination question 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 examine If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoke ,
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5. TLIKQRY OF NUChEAR__EQEER PLANT OEEBATIO _
Page 2 EL.MIDLallD THE8tiRRYtMtilGE
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QUESTION 5.01 (2.00)
Will control rod worth INCREASE, DECREASE, or REMAIN THE SAME for each of the following? Increasing moderator temperature Increasing the percent voids Increasing the fuel temperature
, Increase in Xenon concentration following a power change
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QUESTION 5.02 (3.00) .
- Concerning Prompt and. Delayed Neutrons; STATE whether EACH of the following is TRUE or FALSE: .The percentage of delayed neutrons produced'from fission increases as the age of the core increase '(0.5) The energy level at which delayed' neutrons are produced categorizes them as thermal neutron (0.5) Neutrons produced from.the moment of fission until 10 E-14 seconds.are considered promp (0.5) Delayed neutrons will not cause fast fission of U-23 .(0.5) Delayed neutrtus are the major-factor in determining the rate ofdda-fe #
reactor power 10 seconds after a scram occur (0.5) Delayed neutrons increase the average' neutron generation' (0.5)
tim C b~yr m e.d' c/wy edam
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. l QUESTION 5.03 (3.00) .
i Will actual Shutdown Margin (SDM) INCREASE or DECREASE for each of the following? For EACH item provide a brief discussion justifying your choic Poison concentration decreases (0.75) A number of control rods are inserted into the core (0.75) j Moderator Temperature increases (0.75) Plutonium 240 concentration increase (0.75)
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QUESTION 5.04 (1.50) .
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MATCH each of the lettered items a-c with one of the numbered items 1- Numbered items may be used more than once or not at all as appropriat I FRP 5. PCIOMR APLHGR 6. APF CPR 7. TPF GEXL 8. LHGR Parameter by which plastic strain and deformation are limited to less than 1%. Ratio of bundle power required to produce onset of transition I boiling somewhere in the bundle to actual bundle powe Parameter by which peak clad temperature is maintained less l than 2200 degrees F during postulated design basis acciden I l
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QUESTION 5.05 (2.00) .
.72 M 6CAldA.It- 80L The narrow and wide range yarway level instruments are calibrated for the reactor at hot pressurized conditions with a specified drywell temperatur When the reactor is cold and depressurized will indicated level be HIGHER or LOWER than actual leve (1.0) l WHICH instrument range will have the greater error?
EXPLAIN YOUR ANSWE (1.0)
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QUESTION 5.06 (3.00) *
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During operation at 100% power a feedwater train automatically l isolates due to high water level in a heate l I
STATE whether the following coefficients of reactivity will j MITIGATE or INCREASE the severity of this transien INCLUDE a j brief reason in your answer. Consider the entire transient UNTIL l the scram occurs and assume no operator action is taken i a. Doppler Coefficient l l
b. Moderator Temperature Coefficient l l
c. Void Coefficient I
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.l QUESTION 5.07 (1.00) . I You are performing a reactor startup 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> following a scram which occurred after 30 days of full power operatio WHICH statement below DESCRIBES the expected effects of Xenon concentration when performing the startup? Thermal neutron flux will be highest in the same areas where ;
the flux was. highest during the previous operational perio ,
i Thermal neutron flux will be higher in areas of high Xeno concentration than during the previous operational. phase to maintain the.same reactor' power leve l Thermal flux will be pushed to the periphery of the core, making the periphery rods have a high rod wort . Xenon burnup during the reactor'startup will-make the reactor go critical earlier in the rod sequence than is norma .
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QUESTION 5.08 (2.00) ,
i The reactor is brought critical at 50/125s on range 2 of the IRM with a stable positive period of 45 seconds. The point of adding heat has l
been determined to be 50/125s on range 8 of the IR '
a. WHAT is the expected doubling time if period remains constant? j b. HOW LONG from criticality will it take for power to reach the f point of adding heat if period remains constant? I
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I QUESTION 5.09 (2.50) *
Will critical power INCREASE, DECREASE, or REMAIN THE SAME for the '
following changes? Consider each independentl Increase in core flow Decrease in subcooling Axial power peak shift from bottom to top of channel Increase in local peaking factor Decrease in pressure l
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QUESTION 5.10 (2.00) ..
Answer each of the following TRUE or FALSE: During equilibrium power conditions,-the production rate:of indirect. Xenon from Iodine is faster than the decay rate of Xenon to Ceniu Slowing the rate'of a power decrease, lowers 1the height of the.
! resultant Xenon' pea .The resultant Xenon peak due to a scram from 50% power is I l.
( larger than the peak resulting from a scram at 100% powe During an increase in powerLfrom equilibrium Xenon conditions, Xenon concentration initiallyidecrease I n
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QUESTION Sill -(2.50) ' ;
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During your' Shift,-an SRV inadvertently opens at 100% power and 1000 l psia. .Use a Mollier. Diagram or the Steam Tables to answer EACH of the l following: (ASSUME A SATURATED SYSTEM AND INSTANTANEOUS HEAT TRANSFER) STATE the tailpipe temperature, assuming' atmospheric pressure-in the Suppression Pool and No Reactor Depressurizatio (1.0) If the Suppression Pool. Pressure were.to INCREASE, STATE whether the Tailpipe Tempera'ture would INCREASE, DECREASE, or REMAIN THE
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SAM (0.5)
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! If the reactor starts to 'depressurize when the SRV opens, STATE whether the Tailpine Temperature will initially INCREASE; DECREASE, or REMAIN THE SAME. Assume suppression chamber pressure-remains constan (0.5) What would be the Thermodynamic State (Superheated Steam, Saturate Steam, or Subcooled Liquid) of the the fluid at the discharge point of the SRV7 (0.5)
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"62 PLANT SYSTEMS DESIGN, CONTROL &_AND INSTRUMENTATION Page 13
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QUESTION 6.01 (2.50) .
LIST FIVE (5) conditions which will initiate an automatic start of the Standby Gas Treatment Syste (SET POINTS REQUIRED) ,
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QUESTION 6.02 (3.00) . What effect does the initiation of a " LOAD REJECT" signal have on the following during reactor operation at 95% power? The Reactor Protection Syste (1.0) The APRM Syste (0.5) Using the attached figure (next page) as an aid, BRIEFLY DESCRIBE the operation of the turbine pressure control system and the turbine generator during a " RUNBACK" from 90% power when the ,
turbine throttle valves fail to clos (1.5) i l
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1.50 QUESTION 6.03- ( 2-00-) =
Using the attached Figure (Next page of the exam) answer the following YES or NO:
Assume Reactor water level is -40 i ches and drywell pressure is 3 psi ,q s sm/;s~ ji paGJ Pr pul A sky a. If-the "LPCI or CS_ pump _pr_essurs " onntact_.in_serles-wi-th-rel-ay-145 OPENS 7 rim Lv L h e-120, e c on d-time-d el-a y-com pletrin g-th e-d ele y ,
will the APR system-actuate? If the relay "106"Af ailed such that its associated contacts OPEN after the time delay completed its cycle, will the APR system actuate?
l If the pressure in the ECCS systems decrease to 20 psig after the APR system has actuated, will the APR valves remain OPEN? ,
i If the drywell pressure decrease to 0.5 psig after the APR system '
has actuated, will the APR valves remain OPEN?
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QUESTION 6.04 (2.00) .
The unit is at 55% power, the B condensate pump is out of service for maintenance (4160 supply breaker racked out), A and C feedwater pumps are in service B feedwater pump is in standb The FWCI Selector switch on CRP 926 is in the "A-AUT0". A SPURIOUS initiation signal is received by the FWCI syste DESCRIBE the response of the below listed equipment to the above signa o The FWCI condensate transfer pump The condensate transfer discharge valve (1-MW-96A) FRV FW-5A J % FRV FW-5C l
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QUESTION 6.05 (2.00) .
J For each condition given below STATE whether the affected APRM channel l channel will have an INOP trip? (YES or NO?) l 1 APRM channel 4 APRM mode switch in POWER, channel not bypassed b. APRM channel 5 has 4 LPRM inputs in "A" level; 3 LPRM inputs i in "B" level; 3 LPRM inputs in "C" level and 1 LPRM input in j
"D" level, channel not bypassed !
I APRM channel 3 (with function switch in COUNT) meter I indicates 50*4, channel not bypassed
d. APRM channel 1 APRM mode switch in ZERO, channel bypassed j
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d QUESTION 6.06 (-BAO-) . STATE HOW the. discharge piping of the core spray pumps is protected from overpressurization? (0.5) .
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b, -I-f-a-los swf-normal-power 4LNP ) occurs-concurrent ui+k or
_ subsequent _to the ' initiation--si-gnal-for-oore-spray,--which i
rore spray systeur wi-id stari, .immediately &fter the 4100 "AG
--bus-i-s-ene rgi-zed ? -( 0. 51- l
/ p'. WHAT are TWO (2) actions / evolutions the control room operator ca take during accident conditions to improve the NPSH of..the core
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sso l QUESTION 6.07 (2.004 * '
The Unit is operating at 93% of rated power and all systems are norma Given the following events, SELECT the F4 REP action which occurs (1, 2,3, or 4) from the list belo l EVENTS Steam tunnel high temperature at 200 degrees Main Steam line radiation level of seven times normal full power backgroun c . Mai-n-Steam linc f-low-4n-excese-of-1-1-6%.
6 Loss of AC and DC power to the isolation valve solenoids i while the MODE SWITCH is in RU )
ACTIONS 1. Reactor protection system tri J 2. Group One Isolatio I
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3. Neither 1 or )
4. Both 1 and '
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d QUESTION 6.08 (1.50) .
While operating the Diesel Generator with the voltage regulator in manual contrcl for a test, the A core spray pump is inadvertently started. (Anu - /b dMel (,taval'c h /b edy S*-ett of purr s w by th da) '
WHAT effect would this have on the Diesel Generator output voltag (INCREASE, DECREASE, or REMAIN THE SAME) (0,5) If another A-C motor were operating at this time off the output of )
the Diesel Generator WHAT would be the effect (INCREASE, DECREASE, l or REMAIN THE SAME) on the following parameter (1.0) Motor Current Draw Motor Winding Temperature I
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,kf* laY QUESTION 6.09 fC2-563 .
EXPLAIN the reason the Automatic Pressure Relief (APR) system is i required to prevent core damage during small break loss of coolant l transient (-2-5+
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1 10 QUESTION 6.10 63-003 .
The reactor is operating at 45% power with Recirculation pump A running and Recirculation pump B tripped (B loop idle) when a design basis LOCA occurs in recirculation loop B. During the LOCA, Recirculation pump A fails to trip on low leve Answer the following questions concerning the LPCI (Low Pressure Coolant Injection) LOOP SELECTION LOGI STATE the action initiated by the loop select logic PRIOR to proceeding with loop selectio (1.0) LIST trhe-ei-gne-1-which-prev 1<ies-the LPCI LOOP SELE &T-IM-PERI'ISSIVE . (IliefMDEM NY SETPOIliTS) (0.5}
b HOW does the loop select logic determine that loop B of the i
recirculation system has broken / ruptured.
l (BE SPECIFIC, SETPOINT NOT REQUIRED) (1.0)
4. d . LIST the PERMISSIVE SIGNAL for the LPCI Loop A injection valve to OPEN. (INCLUDE ANY SETPOINTS) (0.5)
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QUESTION 6.11 (2.50) .
During accident conditions operation of SBGT with suction on the drywell is prohibited by the EOPs, if drywell temperature is above 212 degrees What condition is indicative if drywell temperature is above 212 degrees F? (0.5) DESCRIBE how continued venting above 212 Degrees F could lead to l containment failure if drywell sprays were initiate (2.0)
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7 EBQCJDQBES - NQBMalo___ABSORMak uEMERGEEC1 Page 24 1 AND_. BAD 10LQGlGAL CONTRQL )
l 3.1r QUESTION 7.01 (0.75)
Answer the following questions concerning.ONP 526, Unit.1. Uncontrolle Power Oscillation !
a. DESCRIBE the INDICATIONS of an uncontrolled oscillation event. (0.75)
b.- LIST the required IMMEDIATE ACTIONS for each of the following uncontrolled power oscillations listed below: If the uncontrolled power oscillation resulted from withdrawal of a single control ro (0.5) If the uncontrolled power oscillation resulted from a single. recirculation pump tri (O.5)
) STATE what events (conditions) in ONP 526 direct the operator to IMMEDIATELY scram the reactor if indications of.a uncontrolled power-oscillation exis (-1-5-)
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' This event is most likely to occur under which of the'following .
conditions? (CHOOSE ONE) (0.5) High power, high flow High power, low flow Low power, high flow Low power, low flow Cbyr mA /Ub N"* GN'NY l
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/, 3v QUESTION 7.02 (erOO3 .
STATE (YES or NO) whether each of the following sets of plant conditions j could cause thermal stratification in accordance with (OP 206 -Plant 1 cooldown to cold shutdown). Consider each set of conditions separatel I 1 The reactor recirculation pumps are secured, reactor water j level is +55 inches, and the shutdown cooling system is in ;
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operation with one pump at maximum flo The-reactor-reeinu+atrion-pump "B" is-i-n-ser-v4oe-and-th s h u td ow n-o oo l-in g-sys te m-i-s-i-n-ope re t4:o n-w i4rh-tw o-p um ps-at the-maximum-f4e ,
4 p The reactor recirculation pumps are in service, reactor water level is +60 inches, the shutdown cooling system is in I
operation with two pumps at maximum flow, and the SDC heat I exchanger outlet temperature is within 40 degrees F of the "B" ;
I recirculation loop temperature and 45 degrees F of the "A" j recirculation loop temperatur l l
c (. Only the "A" recirculation loop is in operation, reactor water level l I
is +58 inches, two shutdown cooling pumps are in operation at I maximum flow, and the suction temperature at the suction of the j shutdown cooling pumps is approaching 150 degrees !
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~ QUESTION 7.03 (2.50)- a
'EOP 575, RPV' Level /Rx Power, directs the operator to. terminate and prevent'all injection into the RPV except from Boron Injection and CRD until' reactor power drops to less than 3%,.or RPV water level reaches -127", or all SRVs remain shu j l
a. Why is CRD injection not prevented by this. procedure?. (0.5)
b. How does lowering RPV' water level decrease reactor power? (1.0)
c. When 585 lbs. of sodium pentaborate have been injected (16%'SLC storage tank level),.the operator is directed.to restore and maintain water level between +10" and +50". WHAT are TWO (2)
reasons for raising water level to-the indicated-control band at this time? (1.0)
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i QUESTION 7.04 (Or2&) .
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Millstone Unit i has experienced of station blackout (loss of off-site'
and on-site AC power). The unit'was operating at 100% power when the power failure occurred. Answer.the following' questions concerning operator. actions per ONP-503B, Unit 1 Loss of Off-site and On-site AC-Power (Station Blackout).
-)' LIST the reason (s) for each of the TWO "Immediate Actions" stated below during the performance of ONP 503B:
~ "Close all MSIV's from CRP 903 and; place the Isolation i Condenser in servic .(0.75)
i After automatic start of Diesel Generator and/or Gas Turbine Generator, start-one RBCCW pump and one CRD pum (0.5) LIST the reason (s) for the " Subsequent. Action"' stated below during the performance of ONP.503B (Loss of All Station AC Power):
Verify drywell radiation levels.are normal by selecting
" Radioactivity Control" from SFDS Displays, or check drywell radiation on CRP 918. Then open the drywell nitrogen compressor suction valves and restart the compresso (0.75)
c. If the Gas Turbine fails to start, the procedure has the operator place the Gas Turbine mode switch in OFF. EXPLAIN'the purpose for this actio (0.5)
d-I-f-drywel-1-bul-k-temperature is approachi-ng-384-degrees-F-and-normal-drywel-1-cool-ing-eannot- be restoredT-how-can-drywel-1 P \-
- cool-ing-be-accompl-1-shed4--4nel-ude-the eyetems--and interconnections requi-re (0.75)
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QUESTION 7.05 (2.00) .
EOP-574, RPV Steam Cooling, has been entered and the Isolation Condenser is unavailable, therefore, steam cooling must be performed by opening one safety / relief valv The procedure directs the operator to wait until RPV water level reaches -220" to open the SR WHAT would be the negative consequence of opening the SRV prematurely? (1.0) If RPV pressure drops below 700 psig during steam cooling with one SRV, the procedure directs the operator to emergency depressurize using EOP 577. WHAT is significant about 700 psig and WHY is RPV depressurization required below 700 psig? (1.0)
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QUESTION 7.06 (3.00) -
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l The Millstone Emergency Operating Procedures direct the prevention I of the APR depressurization during two-transients 4 EOP 570, RPV Level Control and EOP 572 Reactor Power Contro WHAT is the bases for preventing the APR depressurization in each instanc fl f erpmw r of I
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QUESTION 7.07 (2.50)
~~~Th'U~~Utif tTC^on tro l-Room-i s r e qui r ed d o-b e-eva cua t ed-because_o f a Lire._ Answer-- the Loll owi.ng _ questions-concer-ni-ng-ONP-54-1,r-Uni-t--I Plant.-. Shutdown. Erom Dutside- the Contr-ol Roo _
a s- H OW -i-s-th e-+e a c-t or-sc r a mm e d -6r-om-o u t sid e the cnnten1 rnnm i n .-
-accor-dance-wi-th-ONP-51-1-f or--each-of the-.i oLLowi ng r nnrii + i nnC (1.O)
l -1,--The-Nai-n--Gondenser--and-F.eedwat-er= E y c t c m ^oE avai1ab M 2r-Th e-M a i n-C-on d en s ee-o e--the-F-eedwa ter Sy:t i c NOT-ava4-1-abLe7 b .-HOW-i-s-the-pl-ent--cool-down-r-at-e-mon 4-t-or-ed ie t-h e m a i - condenses l In o m nri +n = chi ava c eld-shutdown-c-ond4-t-ions c- (1.0)
l i,.. WHERE-does-the-Shi+t-supemor-r-epcr t te " hen the Centen1 Annm 4-s-aban d on ed ? (0.25)
dr-WHC i s r esponsi b1 c fcr th e--4-i-r-e-4-i-g ht-i-ng-d ut-ie cf thc pe-eeaae1
--w+vol vcd ia thc Unit Chtitdown c- (O.25)
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'5 lush \ Lunm( Sccenx 'is repirec1 b be_. ecacuM becca c 04 c~ C a . A osu <s A_ 0000 3 pwd.%
conte.rneg owP szs A, begraA,_J m in G onha ha er-cch oaa .
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L.j H G 9 f Ict' each ok k w b \\.o pKc/c 6k3c h MC pe; son m-ph % 4dca.g eac-ken c a b & scia his FixucTiotJ c4 b 4 IoccJ.on Swp em a,~g don %\ operch e (. o . ?c)
" .Sh C\- bpevis o r . G O lo c akens) C/5)
b, W. Proteck c\, cec _b A 6.h ( 6S k not.Eg Rs t io ass % cornmuna ca.on tv: pens.b.1. 6 Lt i t e iA 6919 4 1/2 . b) RAT N spe n s 'w i \', I',< s i.s - ce6c.g b? Go. ar)
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QUESTION 7.08 (2.00) .
An individual permitted to enter a high radiation area shall be provided with or accompanied by a combination of items per Millstone Technical Specification LIST TWO (2) of these items / requirements other than a radiation work permi (2.0)
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.I QUESTION 7.09 (3.50) '
Concerning ONP 525E, Degraded' Fire in: Shutdown. Coo' ling Pump are q DESCRIBE the basic flow path' utilized tofachieve' Cold' Shutdown
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after the isolation condenser is no longer effective (< 250-280 ;
degrees F) and the main condenser is not available as a heatsin I Include all systems and' major components utilized in this procedure and specific A, B etc. designation for components described in the flowpat (2.25) j LIST TWO reasons Reactor water level is raised to +100 inches after the isolation condenser is secure '(0.50) Why is it necessary to maintain the Reactor Coolant temperature out of the' heat exchanger greater than 100 degrees F.? (0.50)
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d Why will it be necessary to eventually run the A or C ESW pump?
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QUESTION 7.10 (1.50) .
While performing a core offload a fuel bundle is dropped in the Reactor Cavity area resulting in a Refuel Floor High Radiation Alar STATE THREE (3) immediate actions required by ONP 519, Dropped Fuel Bundl (1.5)
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I QUESTION- 8.01 (2.00)
DETERMINE if the following statements are TRUE or FALSE per 1 U
ACP-QA-2.06A, Station Taggin a. If a red tagged breaker is removed from the breaker cubicle, the tag remains on the breake b. Additional work orders may be released under a clearance
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I utilizing. Blue Tags only if it 1s-agreeable to the perso '
requesting the clearance, c. If a Red Tag is to be attached to a plant component, any Yellow Tag already attached to the' component is to be lifted for possible reattachment after the Red Tag is removed.
I d. An air operated valve that fails open may be utilized as.a system isolation boundary point for clearance if.the valve is gagged in-the closed position and its solenoid is deenergized .
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QUESTION 8.02 (1.50) ,
Jumper devices installed on OPERABLE plant equipment and positively identified / controlled by plant approved procedures can be exempted from the requirements of ACP-QA-2.06B, Station Bypass and Jumper Control, provided three criteria are met in the controlling procedur LIST the THREE requirements the controlling procedure must satisf l l
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QUESTION 8.03 (2.50) .
Answer the following questions concerning ACP-QA-9.02, Station Surveillance Progra a. HOW is a "high risk" surveillance defined by ACP-QA-9.0 (0.5)
b. Test data taken while performing a monthly surveillance does not satisfy the procedure's " Acceptance Criteria." List FOUR l
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actions to be performed by the Shift Supervisor-before returning the test data sheet to the cognizant Department Hea (2.0)
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QUESTION 8.04 (2.50) .
The Shift Supervisor is authorized by ACP-QA-2.20, Independent i Verification, to waive the independent verification requirements if excessive radiation exposures would resul What exposure is provided as a guideline by this procedure to indicate an excessive exposure? (0,5) l
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b. Alternative means of independent verification are recommended in circumstances where excessive radiation exposure may resul LIST FOUR alternative verification techniques recommended by ACP-QA-2,2 (2.0)
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QUESTION 8.05 (1.50) .
Answer the following questions regarding the Rod Worth Minimizer administrative requirement a. Technical Specifications require a second independent ope:ator or engineer when the Rod Worth Minimizer (RWM) is inoperable in the run mode below 20% thermal power. WHAT does this independent operator specifically verify and WHEN is the verification performed? (1.0)
b. EXPLAIN why the Shift Supervisor is NOT allowed to perform the independent operator function? (0.5)
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QUESTION 8.06 (2.50) .
While operating at 100% power, a Group I isolation and a reactor scram occu Data collected from the plant computer and plant operators indicates the following occurred:
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The group I isolation was caused by a Technician erro The reactor scram was caused by high reactor pressur The Isolation Condenser initiate All operating feed pumps tripped and FWCI Inhibit alarm was receive Reactor water level decreased and FWCI auto started and injected to the reactor vesse An operator reset Feedwater Flow Control and shifted to level control to maintain reactor vessel level at 37".
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Continuing to follow Millstone procedures, the plant is placed in a normal shutdown conditio SELECT the correct statement below and JUSTIFY your choice:
, Power operation cannot resume because FWCI auto-initiated ,
I and injected to the reactor vesse '
l Power operation cannot resume because a safety limit may have been violate Power operation cannot resume because the feed pumps should
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not have trippe Power operation cannot resume until permission is received from State authorities and the NRC because the Emergency Plan was implemented.
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Technical Specification 3.3.A.2 requires that a rod be taken out of service if'it cannot be moved with drive pressur a. WHY is. disarming the control rod electrically. preferred to valving-out the drive (hydraulically isolating the drive)? (1.0)
b. WHEN the rod,is required to be electrically disarmed, how'is-this physically accomplished? ' ( 1. 0 ) .
l If the control' rod cannot be fully inserted before it is disarmed, WHAT additional Technical Specification requirement.must be verified? - ( 0. 5 ) L
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QUESTION 8.08 (2.00) *
The Millstone Technical Specifications require containment integrity to be established any time the reactor is critical or when reactor water temperature is above 212 degrees F and fuel is in the reacto LIST FOUR conditions specified by Tech Specs to establish Primary Containment integrity (in addition to having the drywell and suppression chamber intact).
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QUESTION 8.09 (2.00) ,
What are FOUR major concerns (identified in ACP-QA-2.02C, WORK ORDERS)
of the Shift Supervisor / Supervisory Control Operator when implementing a work order that will place a Category 1 system out of service?
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QUESTION 8.10 (2.00) .
Millstone Unit 1 is operating at 100% power. The Gas Turbine Generator was determined to be inoperable six (6) hours ag Using the attached Technical Specifications answer the following:
J a. WHAT conditions does Technical Specifications place on continued reactor operations? (1.0) l b. WHAT actions would be required per the Technical Specifications l if Core Spray pump "A" was then found to be inoperable in addition to the Gas Turbine Generator? (1.0)
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QUESTION 8.11 (2.00)
The reactor is operating at 75% power with all neutron monitoring systems operabl An Instrument Technician shorts out the power supply to APRM 3 requiring you to declare it inoperable and have it bypasse Using the attached Technical Specifications answer the following:
a . EXPLAIN why Technical Specification action is required with the Neutron Monitoring Instrumentation in this configuratio (1.0) ] WHAT action is required by Technical Specifications? (1 0)
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QUESTION 8.12 (1.00) *
LIST TWO reasons the Technical Specifications require the SRM system to be operable with the Reactor mode switch in RU :
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QUESTION 8.13 F1--009 Using the attached copy of EPIP Form 4701-4, DETERMINE the NRC deportability time requirements for each of the following events: Hurricane approaching the immediate vicinity of the plan l l Inadvertent ~ initiation of the Isolation Condenser while performing calibration of pressure instrumentatio Transportation of a potentially contaminated injured man to ,
off-site medical facilities l I Engineering determination of a potential for ECCS pump suction i strainers to foul with insulation debris following a LOC J G$uye ha d frk k dm d is h /k m
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. THEORY OF NUCLEAR POWER PLANT OPERATION t Page 47 ELulDS&AND_IllERMQDYliAMIGS -
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ANSWER :5.01 (2.00) Increase Decrease Remains the same Decrease (4 @ 0.5 ea'.] (2.0)
REFERENCE Millstone 1, Operator Training Reactor Theory Chapter 5, . Lesson Objective #1 K109(2.5-2.6)
292005K109 ..(KA's)
ANSWER 5.02 (3.00)- False
"- '- True MASTER COPY ' False True True [6 @ 0.5 ea.] .(3.0)
REFERENCE MP1 Operator Training Reactor Theory Chapter 3 Lesson Objectives 11 14 &
1 K102(3.2)
i 292001K102 ..(KA's)
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. THEORYLOF NUCLEAR POWER PLANT OPERATION, Page 48 ELUIDE&AND_TM BdQDXHAMIGE-
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ANSWER 5.03 (3.00)
Given SDM = 1 - Keff (units are delta K)
i SDM decreases [0.25] A decrease in poison concentration adds- l positive reactivity which increases Keff. [0.75] ! SDM Increases [0.25] Control rod insertion adds negative' l
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reactivity.which decreases Kef [0.75]
l SDM Increases [0.25] An increase in moderator' temperature adds-
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negative reactivity which decreases Keff. [0.75]
i SDM Increases [0.25] Plutonium 240 is a strong resonance absorber.- 1 As PU 240 increases the total resonance absorption increases, . 1 decreasing Kef [4 @ 0.75.ea.]:(3 0) i
- REFERENCE MP1 Operator Training Reactor Theory Chapter 1 Lesson Objectives 11 &
1 K114(2.6) .
. 29002K114 ..(KA's)
ANSWER 5.04 (1.50) l [3 @ 0.5 ea.] (1.50)
REFERENCE L Millstone 1, Operator Training Thermodynamics, Chapter 11, Lesson Objectives #12, #2, and # K108(3.0-3.4) 293009K106(3.4-3.8) :
293009K110(3.3-3.7) 293009K118(3.2-3.7)
293009K118 293009K110 293009K106 293009K108 ..(KA's)
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ANSWER 5.05- (2.00)? Indicated level willLread higher that actual leve (1.0). The error.in the wide range will be greater than the narrow range [0.5] due to its lower variable leg.vesselinozzle and thus- l a greater variable. leg water column. [0. 5]- . .(1.0) 4
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b&e pr g 9is o r; -pre.e,/
f ec"o ifyhaa <~ 4,pr . d r eror /-i 44< -, . ffesfelutn'6,fy, u /4, ~
Arp REFERENCE- ge 4 6.rj Millstone 1, Operator Training Thermodynamics,-Chapter'7;.
Operator Training Systems Vol 1, Chapter 1.
l 216000K507(3.6-3.8) 216000K511(3.2-3.3) 291002K108(2.8-2.9) ~
291002K108 216000K511 216000K507 ..(KA's)
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ANSWER 5.06 (3.00)
a. Doppler will mitigate the transient [0.5) because of the increase in fuel temperature. [0.5] (1.0)
b. Maderator Temperature Coefficient will increase the severity, [0.5)
due to increase in core inlet subcooling. [0.5] (1.0) ,
' Void coefficient will increase the severity of the transient [0.5]
because of the increase in core inlet subcooling. [0.5] -(1,0)
REFERENCE r
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Millstone 1,- Operator Training Reactor Theory, Chapter 4, Lesson Objective #2 K203(3.3-3.4) 295014K204(3.2-3.3) 295014K206(3.4-3.5)
295014K206 295014K204 295014K203- ..(KA's).
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l ANSWER 5.07 (1.00)
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3 (1.0)
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= THEORY OF NUCLEAR POWER PLANT OPERATION, Page 50 FLUIDS AND_JBEBdQDJEAMICS
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REFERENCE-Millstone 1 Operator Training Reactor Theory, Chapter 6,. Lesson Objective #9 K108(2.8-3.2)
l 292006K108 ..(KA's)
l ANSWE .08 (2.00) ;
I a. P = Po e**(t/T) (0.25 pts) .for doubling time, P = 2 Po (0.~25 pts). i T = 45 sec o,'
'P : 67 '7i Mh'/ - t = T.In(P/Po) (0.25 pts) l
= T-in 2 ~
i
= 45 (0.69) = 31.2 sec (0.~25 pts) (1.0)
-i b. 50% on range 2 = 0.05% on range 8 (0.25 pts) I Po = 0.05 Pt = 50 T = 45 see Pt = Po e**(t/T) (0.25. pts)
t: T In(P/Po) (0.25 pts) l
= 45 see in(50/0.05) ~" ^
= 310.8 sec = 5.2 min (0.25 pts) -(1.0) !
REFERENCE Millstone 1, Operator Training Reactor Thgory, Chapter 3, Lesso Objective 5 K108(2.7-2.8)
of 101 .N,9 ns4 h c.rk.a.C{y /4y/f p p j' .f& 'N ,
292003K108 ..(KA's)
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Page 51 ELU.lDS_dED_IHEBMQDINAMIGE
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ANSWER' 5.09 (2.50) Increase Decreas Decrease Decrease Increase (5 at 0.5 each)
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REFERENCE
Millstonet i, Operator Training Thermodynamics, Chapter'11,_page 54, 293009K124(2.7-3.2) 293009K123(2.8-3.2) 293009K122(2.9-3.3)
293009K122 293009K123 293009K124 ..(KA's)
i 5.10 (2.00)
-
ANSWER
' True True False True [4 @ 0.5 each]' (2.0)
REFERENCE
, Millstone 1 Operator Training Reactor Theory Chapter 6, Lesson L Objectives #6, 7, & K103(2.9-2.9) 292006K104(2.9.2.9) 292006K105(2.9-2.9)
292006K106(2.7-2.7) 292006K107(3.2-3.2)
292006K107 292006K106 292006K105 292006K104- 292006K103
..(KA's)
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~ THEORY OF NUCLEAR POWER PLANT OPERATION, Page 52 I ELUIDS2.AND_IHEBdQDINAMIGS
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I ANSWER 5.11 (2.50)
i deg F (+/- 15 deg F) (1.0) Increase (0.5)' Increase (0.5) Superheated stea (0.5)
REFERENCE
Millstone 1, Operator Training Thermodynamics, Chapter 06, Lesson Objective #7 l 218000A101(3.4 3.6) l l Steam Tables /Mollier Diagram l 218000A101 ..(KA's)
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. PLANT SYSTEMS DESIGN, CONTROL,_AND INSTRUMENTATION Page 53
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ANSWER 6.01 (2.50) . Reactor building exhaust plenum high radiation (0.25) at 11 millrem per hou (0.25) Reactor building refueling floor high radiation (0.25) at 100 millrem per hou (0.25) Steam tunnel high radiation ( 0.25) at 50 millrem per hour. (0.25) Drywell high pressure (0.25) at 2 psi (0.25)
l +1
' Reactor vessel low water level (0.25) at && inche (0.25)
[ SET POINTS REQUIRED] (10 at 0.25 each)
REFERENCE i
l Millstone 1, Operator Training Systems Vol 4, SBGT Text, Lesson l Objective #2 and page 15.
l 261000K401(3.7-3.8) 261000K301(3.2-3.3) 261000K302(3.2-3.1)
261000K302 261000K301 261000K401 ..(KA's)
ANSWER 6.02 (3.00) . The RPS initiates a control valve fast closure scram if bypass valves 1-6 are not open within approx. 1/4 of a i second after acceleration relay operatio (This scram I is automatically " bypassed" if the valves are open within the time limit.) (1.0) The APRM scram setpoint is reduced to 90% powe (0.5) . The load from the generator will not be reduced to the 25% requirement in the 3 minute time limi . The turbine will trip thus closing the stop valve (Reactor powisr being greater than 45% will initiate a reactor scram) The bypass valves will open to control reactor pressur CANDIDATE'S ANSWER MUST INCLUDE THE ABOVE THREE 1TEMS FOR FULL CREDI (1.5)
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- ELANT SYSTEUS_ RESIGN 1 CONTBOL- _6MD INSTRUMENI&TlQM Page 54
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REFERENCE Millstone 1, Operator Training Systems Vol 3, Turbine Generator Text, {
Lesson Objectives 29bw and 35. Figure 2 K301(4.1-4.1) 241000K302(4.2-4.2) 241000K304(3.8-3.8) )
241000K306(4.1-4.1)
241000K306 241000K304 241000K302 241000K301 ..(KA's)
/, >$ I ANSWER 6.03 (-2-00-) Y-EG l
l c )f, YES l l )
g o'. NO 3' /T i c YES [4 @ 0.5 each] (3rO-) l I
REFERENCE Millstone 1, Operator Training Systems Vol 5, APR Text, Figure 6.
i Lesson Objectives 6 & 1 j l
218000K501(3.8-3.8)
218000K501 . (KA's)
ANSWER 6.04 (2.00) starts opens continues to operate as normal to maintain vessel water level shuts or receives a shut signal [4 @ 0.5 ea.] (2.0)
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REFERENCE l
Millstone 1, Operator Training Systems Vol 4, FWCI Text, Lesson l Objective K108(3.6-3.7) 25'001K302(3.8-3.8) 259001K607(3.8-3.8)
259001K607 25'001K302 259001K108 ..(KA's)
ANSWER 6.05 (2.00)
a. YES b. NO l YES d. NO [4 @ 0.5 ea.] (2.0)
REFERENCE
Millstone 1, Operator Training Systems Vol 6, APRM Text, Lesson **
Objective 4 l 215005K104(3.-3.6) 215005A203(3.6-3.8)
215005A203 215005K104 ..(KA's) ,
. t.ro ANSWER 6.06 ( 9-00-) relief valve (set @ 375 psig)(o,.4-) .fn(trld g,.g ,,deggy, $' W"' [PY
/Je,ug,,2r) (0.5) Core spray system "B" (0.5)
f Decreasing torus water temperature [0.5] and decreasing core spray flow, [0.5] are means to enhance the NPSH of the core spray pump (1.0)
REFERENCE l
Millstone 1, Operator Training Systems Volume 5, Core Spray, Lesson Objectives 7, 9, 12(a); pages 2, 8, 11./10perator Training Procedures, Emergency Operating Procedures, page 1 K401(3.2-3.4) 209001K602(3.8-3.9)
209001K105(3.7-3.7) 209001K102(3.4-3.4)
209001K102 209001K105 209001K602 209001K401 ..(KA's)
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/. P ANSWER 6.07 ( 2-00-)
a. : -2- 9
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b. =4-c. : 3 J /. J D 6 4. =1 [+ @ 0.5 ea.] (-e-O-)
l REFERENCE Millstone 1, Operator Training Systems Volume 3, Main Steam: Lesso Objectives #25, 17k,'27, and 28; Pages 26 - 2 K127(4.0-4.1) 239001K401(3.8-3.8)
239001K401 239001K127 ..(KA's)
l l ANSWER 6.08 (1.50) Output voltage would decreas (0.5) , . Increased current flow to the operating motor would result Increased operating temperature of the motor would resul *
[2 @ 0.5 ea.] (1.0)
REFERENCE MSP1: Operator Training Systems Manual Volume 5,-Diesel Generator Syste Lesson Objectives #15 & 16 I
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264000A103(2.8-2.9)
-264000A103 ..(KA's)
W ANSWER 6.09 (2.50)
hvo Breaka nf thie siz e_ result- in-s ystem-lo s s es- wh ich-ex eeed-the-make-up capaci Ly v f-the-Fee dwate r-Goo lant-End e c bien-s ys tem , [-0 r53-or-fo14 ewing-a .
failure of-the-FWOI system-[S-53-may-not-be-lerge-enough-to - l depressurite-the reactor-49-St-to-the-di-scharge pressure of the spray i-ey stem-[-O r53-before-unacceptable-e-fuel-temperatures-are-reached . [0- 5]
.
04 @ 0.5 ea.] (-2-5-) !
] b /4 In the event of a f ailure of. the FWCI system ETdM or a break '
larger than'the capacity of the FWCI E0.53 the APR opens selected SRVS to rapidly depressurize the reactor CO.53 to within the injection pressure of the core spray and LPCI systems. [0.53 l
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REFERENCE Millstone 1, Operator Training Systems Volume 5,'APR System, Lesson Objective 1 218000K301(4.4-4.4)- 2 /J'do # 6 o# 9 ( F '/ d 1
218000K301 ..(KA's) l
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'2 50 _l ANSWER 6.10 (0.00) Trips signals are sent to both reactor recirculation pumps [0.5] 1 A close signal is sent to Recire loop cross-tie and cross-tie bypass valves (to insure they remain closed) [0.5] (1.0)
br-Low-RPV-pressurc [0. 263 000-psi 1g [0.25] +0-6-) ;
6 Loop select logic monitors d/p between recire. riser'A and B
/ [0.50] and identifies the pressure in riser A is greater than the pressure in riser B [0.50]. (1,0)
6, 51'. (Low)RPVpressure [0. 2G ]-- 350 psig -[-0-eSt ;( O '. 5 ) I
REFERENCE
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MP1: Operator Training, LPCI/ Containment Cooling, TX-1335, PP.-25-26, K/A 203000K401(4.2*/4.2), 203000K410(3.9/4.1), 203000K411(4.0/4.0)-
ANSWER 6.11 (2.50) It is likely that steam is being admitted to the drywell (0.5)
, If the drywell air temperature were greater than 212F steam would be l replacing the noncondensible [0.5] If sprays.were.then initiated, the containment pressure would decrease rapidly, !
[0.5] at a rate exceeding vacuum breaker capacity. [0.5] This'would ;
create a negative pressure in the drywell exceeding the design limi '
[0.5] ( 2. 0 ).
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REFERENCE Millstone 1, Operator Training Procedures, EOP text, page 136; Operator Training Systems, Vol 2, Lesson Objectives 32 and-3 K109(3.4-3.6) 223001K501(3.1-3.3) 223001G001(3.8-4.0)
223001G001 223001K501 223001K109 ..(KA's)
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7 ~. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Page.59 AND_BADLQLQG.lGAL_GQNIBQL
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.7,27 ANSWER 7.01- (-3@ l
! APRM peak-to-peak oscillations greater than 10% power [0.25] ,
with no indication of level.[0.25].or pressure [0.25]-control system malfunction (0.75) j
' . If uncontrolled power oscillation resulted from withdrawal o single control rod, then insert the rod to its previous positio (0.5)
2 .' If uncontrolled power oscillation resulted from a single recirculation pump trip, initiate a select rod inser (0.5)
[ H L . 1 [0 %) vu L. 2 [0 M] obvve do avl LuualuaLu Lhe vsv111aLiva, l l i=cdictcly ccr;;; the rc;cto (-1 -fH l
[
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If both recirculation pumps have tripped, immediately~ scram . !
the reacto (e-5-)~ (high power, low flow) (0.5)
REFEF.ENCE MP1: ONP 526, Unit 1 Uncontrolled Power Oscillations, Rev. O, PP. 1-2 295001K104(2.5-3.3) 295001K305(3.2-3.6) 295001A202(3.1-3.2)
295001AG011(3.9-4.2)
295001G011 295001A202 295001K305 295001K104 ..(KA's)
/..sa ANSWER 7.02 60-) ew 20 Yec f g'. . No i
) d ro
- J i'. -Ves NB [,,4' @ 0 . 5 e a . ] 4-B-0-) i
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REFERENCE Millstone 1, Operator Training Procedures, General Operating i Procedures, Lesson Objectives 63b, 63a, 63 I 202001K507(3.3-3.4) 202001K122(3.5-3.6) J 202001K122 202001K507 ..(KA's)
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ANSWER 7.03 (2.50)
a. The CRD system is being used concurrently with this procedure 1 (in EOP 572) to attempt to insert control rod (0.5)
b. Reducing the natural circulation driving head due to height
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differential reactor power )r ]- resultshinpincreased
[0.5]. h,.g p,0core voiding (and decreased (1.0)
c. The water level must be raised to reestablish natural circulation vsf6-63 and distribute the boron throughout the core. [0.5] (1.0)
re s.eh sc - ASa WA c - c a>t g z @ 0, fed rue. k) nod ed.o\ b4 -Loperrar- AkW d l REFERENCE l MP1: EOP 575, RPV Level /Rx Power, Rev. 2 (Chng. 1), PP. 2&7
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.. Operator Procedure Training, Emergency Procedures, TX-1500B, j PP. 54,55,66, & 67, 295015K101(3.6-3.9) 295015K103(3.8-3.9) 295015K201(3.8-3.9)
295037K102(4.1-4.3) 295037K104(3.4-3.6) 295037K107(3.4-3.8)
295037K205(4.0-4.1) 295037K209(4.0-4.2) 295037K303(4.1-4,5)
295037K304(3.2-3.7)
295037K304 295037K303 295037K209 295037K205 295037K107 295037K104 295037K102 295015K201 295015K103 295015K101
..(KA's)
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ANSWER 7.04 (-3-259 , . Avoids vessel inventory. losses [0.25] while providing a method to remove decay heat [0.25] since no normal method of providing significant amounts of water to the vessel is available [0.25].
(0.75) Prevents ECCS initiation on low vessel level.or high drywell l pressure (0.25]. Also, avoids temperature damage to the- 1 recirculation pump seals and the control-rod drive mechanisms [0.25] . (0.5)
o.575' Ensures no radioactive release to the reactor building. [-0 . 2 5 ] - 1 Restores nitrogen.provides operational control for i
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reactor pressure control equipment [0.25] and the drywell cooler dampers [0.125] . (0.75)
f Secure the DC auxiliary pumps [-0,-25] and conserve the Gas Turbine l batter [0-M]
o.ro
_0.5)
(
' Utd+tze-Di .v ., ell Opray cooling-EO . 2 5] by-connectri-ng-a-f-i-re-hose-
[4.25] to the"A" LI'CI heat exchanger ( s hel-1--e-id e-deal-n+-[-OrM3-(4r 74-)
i I
.. REFERENCE i
MP1: ONP 503C, Rev. 2, P. 1&5 .
Operator Procedure Training, Off-Normal Procedures, TX-1500A, LO 30, 32, 34 & 37, PP. 52 & 53 295003K106(3.8-4.0) 295003K201(3.2-3.3) 295003K205(3.8-4.0)
295003K302(2.9-3.1) 262001A203(3.9-4.3) 262001A204(3.8-4.2)
295003K302 295003K205 295003K201 295003K106 ..(KA's)
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ANSWER 7.05 (2.00)
a, insuf-f-icient-thermal-gradient-weuad-be-developed--[-0. 3] rcsul-ting-i-n
_insuff4<-iant steam f1.ow [0. 3] to-remove-deoay-heat-f-rom-the-core
[0 4] wl-th-etea ti-TM b. 700 psig is the minimum RPV pressure that will produce adequate steam i flow through one SRV to limit peak clad temperature to 2200^F (or to assure adequate core cooling). (0.5) The surge of steam flow resulting !
from rapid depressurization will lower fuel temperatures providing additional time to establish a source of coolant injection. (0.5)
g* t:?tueen2el 5ltu m c.' o* lly e(feslEs/e neJS [o.s) a-1/ cltcfersetA (1.0)
&e a ya,%$l/ :4 8si ey a lle%/lsh .2Jedu .fy.g/ws[o.r) (/,c)
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.. PROCEDURES - NORMAL. ABBORMAL1 EMERGENCY Page 62-AND_BADIQLOGlGAL CONTROL
i REFERENCE MP1 Operator Training - Procedures, TX-1500B - Emergency Operating Procedures, pgs 49 & 50, Obdect4ve-2+
295031EK101(4.6-4.7) 295031G07(3.7-4.0) 2_9.r(oJ/(k D V ((P -94,l 295031K101- 295031G007 ..(KA's)
ANSWER 7.06 (3.00) ,
i During the reactor level control process (EOP 570) while level is maintained between +10 and -127 inches and the APR timer has initiated, the operator is instructed to prevent blowdow )
BASES FWCI may be controlling Reactor water. level [0.25] at a level below the APR setpoin [0.25] If the blowdown is allowed to proceed, l kjfv\0 there is no assurance the low pressure ECCS'will actually injec [0.25] (0.75)
APR initiation imposes a severe transient on the vessel [0.25] and is not necessarily conducive to level contro [0.25] The operator ycd) can progress toward stable conditions in'a more controlled
[0.25] (0.75)
~~
fashio . During reactor power control (EOP 572), when Standby Liquid j Control initiation is required, the APR blowdown is to be ;
prevente i y BASIS
} g -, The injection of all ECCS would result in reactivity bv7fF' . additions [0.25] from boron dilution, void collapse, and r#,* temperature effects [0.75] which produce power excursions rP f o [0.25] large enough to damage the core. [0.25] (1.5)
Q REFERENCE-MP1: EOPs 570 & 572. Operator Training Procedures, EOP Text, PP. 27&41 295031K301(3.9-4.2) 205031K305(4.2-4-.3) l 295031K305 295031K301 ..(KA's)
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a i ac -3 mm co.26 ro mag o % g ,m s _
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' PROCEDURES'- NORMAL, ABNORMAL &_ EMERGENCY Page 63 AND_ BAD 10LD919AL_CDETEDL
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ANSWER 7.07 (2.50)
+.'l. Venti-ng-the-soram-a-i-r-header . (0.5)
-2. Tripping-both-RPG-MG-set-output-breaker (-04 bv--M a i-n-s te am-pre s s u re-g a g e-by-th e-# 8G6-s witch es--[4-53--a n d-the
- loca14y-mounted-pressure /-temperature-ohar-t . [0,5] ( 1,-04-c. Shutdown-communications-area-4by--the-makenp-
-tiemimeral-iter-area) (41.J.54 Unit 2 ( fi ra b ri ga ria ) personnel (0.251 l
i REFERENCE-diPi+-ONP 511. " nit i Plant Shutdown From Outsi-de the Control noore, Rev 1, PP. 1 -4, 10 295016K202(4.0-4.1) 295014A14844.0" .0) 295016A20S(3.3-3.0)
G960-14G006t4 .1 -4 .1-)
295014G006 295044A306 295044A408 295014K202 ..(KA' )
ANSWER 7.08 (2.00) A radiation monitoring device which continuously indicates the radiation dose rate in the are . A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is receive . An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring devic '
[any 2 @ 1.0 ea.] (2.0) i
,
REFERENCE Millstone 1 Technical Specifications 294001K103(3.3-3.8)
294001K103 ..(KA's)
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I lI ANSWER 7.09 (3.50) From the A Recire Loop suction [0.25] to the shutdown cooling system suction, [0.25] through either of the shutdown cooling heat exchangers-[0.25] and gravity drain to the Torus [0.25] via the B i side LPCI test lin [0.25]
From the Torus [0.25] to either A or C LPCI pump [0.25] back to the ;
vessel [0.25] via the A Recirc loo [0.25] [9 @ 0.25 ea.] (2.25) To provide increased head pressure for flow to Torus [0.25] and assure natural circulation if Recire pumps are secure [0.25] I (0.50)
l To ensure Reactor Coolant temperature remains above the Tech Spec I limit of 86 degrees F [0.25] with the head tensioned. [0.25] (0.50) To maintain Torus Temperatur (0.25)
REFERENCE
MP1: ONP 525E ,
205000K503(2.8-3.1) &&5000A203(3.2 3.2' 205000K303(3.9-3.9) l 295021K104(3.6-3.7) 295021K301(3.3-3.4) 295021K302(3.3-3.4)
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295021K302 295021K301 295021K104 205000K303 205000A203 i 205000K503 ..(KA's) 295e2 IA toy (J 4 3,t) 29Jw/gb;-[J.y_y.j')
295oJ/ OelDo/-34) .293 b)/gos 7Q 9 -3,1) 2%myf;ocfG27,y .
ANSWER 7.10 (1.50) Evacuate all personnel from the Refuel Floo . Conduct an evacuation of the drywel . Conduct an evacuation of the Reactor Building. [3 @ 0.5 ea.] (1.5)
REFERENCE MP-1 Off Normal Procedure 519, Dropped Fuel Bundle, Step A101(3.6-4.1) 295023A203(3.4-3.6) 295023A301(3.6-4.3)
295023A301 295023A203 295023A101 ..(KA's)
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__ _ ~ ADMINISTRATIVE PROCEDURES, CONDITIONS, Page 65 AND_LldIIAIl0NS-
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' ANSWER 8.01 (2.00)
a. False b. False-c. True d. True [4 @ 0.5 ea.] (2.0);
REFERENCE MP1: ACP-QA-2.06A, Station Tagging, Rev.-13, PP. 9-13 Administrative Procedure Training, TX-1100, Rev 1, LO 13,14,15,17-294001K102(3.9-4.5)
294001K102 ..(KA's)
ANSWER 8.02 -(1.50) - de_L2 kc\
'
1. SS/SCRO notification and signoff when jumper' devices are installed or removed.
g 2. Documentation of jumper device installation and removal is
- require . Independent verification of jumper device installation and removal is required on Quality system [3 @ 0.5 ea.] (1.5)
REFERENCE MP1: ACP-QA-2.06B, Station Bypass and-Jumper Control, Rev. 8, P. 7 294001K102(3.9-4.5) /d c.fM sy, y,7 294001K102 ..(KA's)
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ANSWER 8.03 (2.50) A surveillance test which could result in a " loss of generation". J (if a system or component failed during the test)= : ( 0. 5 )-
If the' candidate only mentions RESULTS IN A REACTOR SCRAM -0.25 b. 1. Determine system / component operabilit . Document system operability on. test data shee . Verify technical Specification. requirements are satisfie )
4. Complete / submit a Plant Incident Repor [4 @ 0.5 ea.~] (2.0) j REFERENCE MP1: ACP-QA-9.02, Station Surveillance Program, Rev. 16. P. 6,12,13 294001A103(2.7-3.7)
294001A103 ..(KA's)
, ANSWER 8.04 (2.50)
l greater than 10 mrem (0,5)
b. Any 4 of the following @ 0.5 each; 1. Remote position indicator . System process parameters (flow, pressure, etc.)
3. Observation of valve stem which has been marked by some metho . Authorized scribe marks on throttled equipment 5. Functional mechanical indicator (2.0)
REFERENCE MP1: ACP-QA-2.20, Independent Verification, Rev. 1, PP. 2.5,&7'
294001K101(3.7-3.7) /0 c f4 gup/J .
294001K101 ..(KA's)
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AMD._ Lid 1IAIION l
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ANSWER 8.05 (1.50)
a. The second operator checks that all control rod positions are correct [0.5] prior to commencing each new rod step in.the sequence, [0.5] (or rot ye-p)
-
(1.0)
b. The independent operator or engineer can have no other concurrent duties (0,5)
REFERENCE MP1: Operator Training (Systems), RWM, TX-1302B, Rev. 2, PP.65-67 LO 24,25,26
,
Technical Specif.ications Bases, 3/4.3.A.2, P. B 3/4 3-2 l
201006K301(3.2-3.5) 201006A207(2.5-2.8) 201006G001(3.5-3.8)
201006G001 201006A207 201006K301 ..(KA's)
i ANSWER 8.06 (2.50)
b. [1.0]
~
EXPLANATION: On a group I isolation the scram signal should come from MSIV closure and NOT steam line pressur [0.%] Failure of the primary source scram signal (steamline isolation valve closure scram)
may have allowed a safety limit to be violate [0.%) Technical Specifications prescribe that exceeding a safety-limit in nause for a-u nitr-s hut d own-and-review--by--NRG-before-rec ump t ion-of-u ni-t-ope rat i-on ,
j [4 -53 reyWes /de w tl je /L oA L //o7-5 fM A$ (2.5) )
REFERENCE MP1 Tech. Spec 2.12. and 6. I bOS4-Techn4"al Apa"4 f.14at-ions PP. B 2-E TO B-2-6- l Technical Speci&icatriens-64-.-1-395005K40H4-0-4-1-)-295005 KdO2t3-2 - 0 . 0 ) 235005Kf01(0.C-0.S)
895005KGOH-3r6-3-63-295035G005-( 2. 7 - 0. 7 )
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295005G005 295005K30-1 295005K201 295005KdO2 235005K40-1
. .(KA'S) 2 9.ro J oklol (). / - 2 p} 293 o.1oA?od (3, y .7.L)
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. ADMINISIBATIVE PROCEDURES t CORDITIOHet Page 68 ANQ_LIMIIAIl0MS
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ANSWER 8.07 (2.50)
a. CRD cooling water is still available to cool the drive and prevent crud accumulation if the CRDM is not hydraulically disarme (1.0)
b. The four amphenols type plus connectors are disconnected from the drive insert and withdraw solenoid (1.0) Shutdown Margin (0.5)
REFERENCE MP1: Technical Specifications Bases 3.3.A.2, P. B3/4 3-1 and 3-2 201003K303(3.2-3.8) 201003G006(2.8-3.7)
l po i coJNsof'(3,/- JQ 20/006 soJ~(71-),1)
201003G008 201003K303 ..(KA's)
ANSWER 8.08 (2.00)
1. All manual isolation valves on lines connecting to the reactor l coolant system or containment isolation valves not required to be open during accident conditions are close ., At least one door in the airlock is closed and neale . All automatic containment isolation valves are operable are I deactivated in the isolation positio . All blind flanges and manways are close [4 @ 0.5 ea.] (2.0)
REFERENCE MP1: Technical Specifications Definitions, P. 1-2 & 1-3 223001K101(3.7-3.8) 223001G005(3.3-4.1)
223001G005 223001K101 ..(KA's)
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ANSMER' 8.09 (2.00) . Tagging requirements . Technical Specifications (identify any LCO requirements and perform any necessary surveillance testing of redundant safet systems) Retest requirements (approve or' amend retest ~ recommendations) Radiation control-s A m p Fh't add .s [4 @ 0.5 each]-(2.0)
cJ rey nlrofi pudonk wk REFERENCE MP1: ACP-QA-2.020, Work Orders, Rev 13, PP. 11.-12 &'2 iKt02( 0. 0 4W 294001A103 ( 2. 7-3. 7 ) . /o - < /X #273 204001K10E 294001A103 ..(KA's)
l ANSWER 8.10 (2.00) . . .
a. (T.S. 3.5.F.3) allows 4 days if the diesel generator, all APR components, all' low pressure core cooling and containment cooling subsystems are operable 6Gr&+
40 ,
b. ( T . rJ . 3.5.A.6 and T.S. 3.5.F.4) specify an_ orderly shutdown of the
.
reactor shall be initiated and the reactor shall be in the COLD SHUTDOWN or REFUEL cond.ition within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (1.0)-
REFERENCE MP1: Technical Specifications, PP. 3/4 5-1 & 3/4 5-8 209001G005(3.3-4.2) 264000G005(3.4-4.1)
L 264000G005 209001G005 ..(KA's)
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ANSWER 8.11 (2.00) Less . San three operable IRM High-High/ Companion APRM Downscale trips are operable [0.5) (APRM 3 is a companion signal to channel 13 and 14 IRM High-High trips in RPS Channel A.) with the Reactor Mode Switch in the RUN mod [J.5] (1.0) Place the affected trip system (RPS trip system A) in a trip conditio (Half Scram) (1.0)
REFERENCE MP1: Technical Specification 3.1.A, Table 3. Operator Systems Training, Reactor Protection System, TX-1408, Rev. 1, Fig. 8 215005K101(4.0-4.0) 215005K102(3." :).7) 215005G005(3.3-4.2)
215005G005 215005K102 215005K101 ..(KA's)
ANSWER 8.12 (1.00) Available for a planned shutdown or scra (0.5) Provide for continuous indication of reactor perio .5)
RFyERENCE MSP1: Operator Training Systems Manual Volume 6, SRM Syste Lesson Objective #8 215004G006(2.8-3.7)
215004G006 ..(KA's) ;
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ANSWER 8.13- (t-001 5 cc locl ou ' Immediate (A.II u d' b " , g h '^ d ## /
D
I* '
Imme t to 'A.1) FA-73
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/ %L'*q;i <gf,s f/' "&w,9 Vf //T,p ,wxzm ,
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4~ ,r zy
. &y b y y & for-?,N9 %,_zg- e-ry " <
Immedia e (A.1) y/ A 7 fp.74 oc g,
. /g2jc';,7
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= y stow <r --
am ej(,(/(.ri 7 REFERENCE
' '
MP-1 EPIP Form 4701-4, 10 CFR 50.72 & 50.7 LER 88-004 294001A116(2.9-4.7)
294001K116 ..(KA's) !
i meacbc& GA.D or i he repo4 or- lo CFP 6032. (bM QuiQ [c.zQ SNb 3D b 3 LE R c>,-
1o CFR SO. n QL2.3Ui ih C o. L'[l q a
b- I mW.cb cle- ( cr* 9 he- reph or t o cJT4 50,72. (.tStQd d Coafl iWb D bq LG A oc So , E (4.'y.t3C h6 Eo.2G Icm auSe_ LA .d oe 4 he reped or to c.rR Sd.72 teczxtii') cos3 ;
d 4 he re porb oc /O c,FA Sc.72.(tht2.XI,1 ) h.2b
,.
A pb or 57).73 (dz-M U [o.2d l 20 bc.3 lea :
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EQUATION SHEET
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2 Cycle efficiency = Net Work (out) -
, , ,g ,.y g
e+ gg ag Energy (in)
E = aC -
a = (vg - v,)/t KE = laiv vg A = AN A = A,e
= v, + at PE = mgh w = 9/t A = in 2/tq = 0.693/tg W = v4P- .
AE = 931Am eg(eff) = (t,.)(ts)
.
(g + )
Q=$ CAT ~
IX P , I = I
, o Q = UAAT I=I
-
ux
' ~
' 'Pwr = W'g m" I=I 10-x h
,
g SUR(t), TyL . 1.3/u
-
to y.p .t/T HVI. a 0.693/u o
~
'SUR = 26.06/T T = 1.44 DT SCR = S/(1 - K,gg)
fA*gr oT SUR = 26 g,,l CR, = S/(1 - K,gg,)
? CR 1 (1 - K*ff)I = CR 2 (l ~ X*ff)2
'=~
T = '(t /o ) + [(6 ' p)/1,ggp ]
T ,= 1*/ (p D M " I/II - Keff) = CR /CR0 g
"
~# #
vff M = (1 - K,ff)0/II ~ Eeff)1 8 " IEeff -1)IKaff = AK,fg/K,gg g , ,
~
p= [t*/TKygg-] + [I/(1 + A,ggT )] t* = 1 x 10 seconds
,
P = I(V/(3 x 10 0) A,fg A= 0.1 seconds Ia No
.
Idgg=Id22 UATER PARAMETERS Idg =Id22 1 gal. = 8.345 lba R/hr = (0.5 CE)/d (meters)
1 gal. = 3.78 liters R/hr = 6 CE/d (feet) -
1 ft = 7.48 ga MISCEI.L\NEOUS CONVERSIONS .
Density = 62.4 lbm/ft 1 Curia = 3.7 x 10 dps O Density - L gm/cm i kg = 2.21 1ha liest of vag orizations - 970 Etu/lba 1 hp = 2.54 x 10 3BTU /hr Heat of fusica = 144 Stu/lbm 6 1 Mw = 3.41 x 10 Btu /hr 1 Atm = 14.7 psi - 29.9 in, l' Btu = 778 ft-lbf I ft. H O 2
= 0.4333 lbf/in 1' inch = 2.54 cm F = 9/5 C + 32 C = 5/9 (*F - 32) ____________________
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QUESTION -VALUE BEEEBENCE
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5 '.' 01 .2.00 ZZZ0000002 l
'5.02 3 00 ZZZ0000003 5.03 3.00 ZZZ000000 i 5.04 1.50- .ZZZ0000006 d 5.05- 2.00- ZZZ0000007 'd 5.06- 3.00~ ZZZ0000008 5.07 1.00 ZZZ0000009 5.08 2.00 ZZZ0000010 5.09 2.50 ZZZ0000011 )
5.10 2.00 ZZZ0000001 J 5.11 2.50 ZZZ0000005 i
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l 24.50 6.01 2.50 ZZZ0000013 6.02 3.00' ZZZ0000014 6.03 /<Y Br&& ZZZ0000015 6.04 2.00 ZZZ0000016 6.05 2.00 ZZZ0000017 6.06 # e-06 ZZZ0000020 6.07 /J'.3rOO ZZZ0000022 6.08 1.50 ZZZ0000012 G.09 ' E" ev50/JDZZZ0000021
{
6.10 2.r $T00 ZZZ0000018' q 6.11 2.50 ZZZ0000019
______
m 114 l 7. 01 3M 8-Hi ZZZ0000026 l 7.02 /.5- 2 M ZZZ0000029 j L 7.03 2. ), 2. 50 ZZZ0000025 '
l 7.04 Sve6 ZZZ0000023 7.05 2.00 ZZZ0000024 7.06 3.00 ZZZ0000027 7.07 2.50 ZZZ0000028 !
7.08 2.00 ZZZ0000030 7.09 3.50 ZZZ0000031 7.10 1.50 ZZZ0000032 d Elf ______
2&r&O 8.01 2.00 ZZZ0000033 ;
8.02 1.50 ZZZ0000034 1 8.03 2.50 ZZZ0000035 8.04 2.50 ZZZ0000036 !
8.05 1.50 ZZZ0000037 l 8.06 2.50 ZZZ0000038 8.07 2.50 ZZZ0000039 -
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I 8.08 2.00 ZZZ0000040 .l 8.09 2.00 ZZZ0000041 8.10 2.00 ZZZ0000042 8.11 2.00 ZZZ0000043 8.-12 1.00 ZZZ0000044 8.13 Fr00 ZZZ0000045 l
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dupWDWWW 1-0,DMT .
$ DOCKET NO 245
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