IR 05000309/1986016
| ML20215E560 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 12/10/1986 |
| From: | Coe D, Collins S, Keller R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20215E558 | List: |
| References | |
| 50-309-86-16, NUDOCS 8612220412 | |
| Download: ML20215E560 (148) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT ,
EXAMINATION REPORT N /86-16 (OL)
FACILITY DOCKET NO. 50-309 FACILITY LICENSE NO. DPR-36 LICENSEE: Maine Yankee Atomic Power Company 83 Edison Drive Augusta, Maine 04336 FACILITY: Maine Yankee EXAMINATION DATES: October 21-23, 1986 CHIEF EXAMINER: I N / 0@
D. H. Coe, te'ad Reactor Engineer (Examiner) Date REVIEWED BY: vif M 12.//o /s c R. M. Keller, Chief, Project Section 1C Date APPROVED BY: IT0lkth4b)ll441) idl/ %
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Date S. J. Collins, Deputy Director, DRP SUMMARY: Written, Simu14 tor, and Oral Examinations were adiiiinistered to three Reactor Operator (RO), four Senior Reactor Operator (SRO) candidates, and one Instructor Certification candidate. All candidates passed all portions of the examination and were issued licenses.
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REPORT DETAILS TYPE OF EXAMS: Replacement EXAM RESULTS:
l RO l SRO I Inst. Cer l l Pass / Fail I Pass / Fail l Pass / Fail I l I l l I I I I l IWritten Exam l 3/0 l 4/0 l 1/0 l l l l l l l l l l l l Oral Exam l 3/0 l 4/0 l 1/0 l l l l l 1 I I I I I l Simulator Examl 3/0 l 4/0 l 1/0 l l 1 1 I l l l l l l l0verall l 3/0 1 4/0 l 1/0 l l 1 1 I I I I I I I CHIEF EXAMINER AT SITE: D. H. Coe OTHER EXAMINERS: J. Upton, PNL R. Clark, PNL The following generic deficiency was noted during oral exams. Both R0's and SRO's lacked full understanding of why upper and lower linear range nuclear instruments together totaled more than 100% power. Generic deficiencies noted from grading of written exams:
R0 examination Reason for caution on operation of only one RC CIS logic when in " BLOCK". Source of interlock signal for regulating withdrawal interlock SRO examination Changes to beta-bar and reactor period as the core age Primary indicator cf loss of natural circulatio How LHGR Safety Limit changes with core average burnu . Personnel Present at Exit Interview:
NRC Personnel D. Coe, Chief Examiner C. Holden, Senior Resident Inspector Facility Personnel J. Garrity, Plant Manager J. Frothingham, Operations Department Head A. Shean, Manager of Training M. Everingham, Operations Training Section Head J. Kirsch, Supervisor, Operations Training Group M. Swartz, Supervisor, Simulator Group B. O'Brien, Senior Instructor J. Sanoski, Senior Instructor Summary of NRC Comments made at exit interview:
Simulator Performance Due to an apparent simulator malfunction, the candidates lost control of the turbine control valves while at low power. This occurred twice; but, in both cases, the examination proceeded without interruption. Licensee corrective action will be verified during the next simulator examinatio Procedures The emergency classification chart which identifies the conditions under which a steam-line break will initiate an ALERT is imprecise in that the criteria are listed as "CSAS with containment pressare (greater than) 20
. psi." It does not specify if the pressure is in absolute or gage and is further complicated by intending to mean gage pressure when control room indications and E0P's use absolute (0 pen Item 86-16-01). The licensee has committed to correcting this using their normal change procedure Candidate Performance Although no specific preliminary results were given, the NRC commented that the license training programs appear to be providing a good knowledge and skill foundation for these operator . Facility comments on the written examinations:
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Comments were received from the facility and these comments and the NRC resolutions are provided in attachment 2 to this repor _ _
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Attachments: Written Examinations and Answer Keys (SR0/R0) Facility Comments on Written Examinations made after Exam Review and Resolutions
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U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: MAINE YANKEE REACTOR TYPE: PWR-CE DATE ADMINISTERED: 86/10/21 EXAMINER: CLARK, R. /UPTON, CANDIDATE: AMSHER KEY INSTRUCTIONS TO CANDIDATE:
Use separate paper for the answers. Write answers on one side onl Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination start % OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 25.00 25.00 PRINCIPLES OF NUCLEAR POWER c PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 25.00 25.00 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 25.00 25.00 INSTRUMENTS AND CONTROLS l 25.00 25.00 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL
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100.00 Totals
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All work done on this examination is my ow I have neither given l nor received aid.
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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 . 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 only o_n one side of the paper, and write "Last Page" on the last answer shee . Number each answer as to category and number, for example, 1.4, . Skip at least three lines between each answe . Separate answer sheets from pad and place finished answer sheets face down on your desk or tabl . Use abbreviations only if they are commonly used in facility literatur _
13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer require . Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
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15. Partial credit may be give Therefore, ANSWER ALL PARTS OF THE l QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
l 16. 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 after the examination has been completed.
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D 18. When you complete your examination, you shall:
a. Assemble your examination as follows:
(1) Exam questions on to (2) Exam aids - figures, tables, et (3) Answer pages including figures which are part of the answe b. Turn in your copy of the examination and all pages used to answer the examination question c. Turn in all scrap paper and the balance of the paper that you did not use for answering the question d. 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 revoked.
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PAGE 2 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION1 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.01 (2.00) Briefly EXPLAIN how the presence of delayed neutrons in the reactor core contribute to the control of the reactor power level (in response to a small reactivity change). (1.0) COMPARE the response of the reactor power level to a small positive step change in reactivity (reactivity addition) at BOC versus EOC. Answer by describing HOW and WHY the initial startup-rate (SUR) or the initial reactor period would be different at BOC compared to EOC (for a given reactivity change). (1.01 QUESTION 1.02 (1.00)
IDENTIFY two (2) features of the reactor system design that control the RADIAL neutron flux distributio (1.0)
QUESTION 1.03 (1.50)
LIST five (5) factors other than rod position or RCS boron concentration that affect SDM and are used in the SDM calculatio (1.5)
QUESTION 1.04 (2.00) DESCRIBE WHY the point of adding heat (P0AH) is important to a reactor operator who has been directed to take the reactor from just critical to a power level of 2%. (0.8) HOW is the P0AH recognized? (0.9)
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' At WHAT power level does it occur at Maine Yankee? (0.3)
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- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.05 (2.00) If the Maine Yankee power plant had been operating at 100%
power for 50 days and then was taken very quickly to 50%, the xenon (Xe) concentration in the core would change. SKETCH the Xe concentration from the time at which the power change was executed to 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> late (1.0) EXPLAIN why the initial change in Xe concentration is to increas (1.0)
QUESTION 1.06 (2.00)
Assume that the Maine Yankee power plant is operating at 70%
power level with all controllers in AUTO and controlling the plant parameters at their respective programmed values. Then, if the primary coolant flowrate decreased (insufficiently to cause any plant trips), WHAT would be the changes (if any) in the following plant parameters? (Assumenooperatoraction.) Tave (0.5) steam generator temperature and pressure (0.5) steam generator flowrate (0.5) pressurizer level (0.5)
QUESTION 1.07 (2.00)
The neutron multiplication factor, keff, has just changed from 0.920 to 1.004. ANSWER the following parts of this question by filling-
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i The reactor is now . (SUBCRITICAL, EXACTLY CRITICAL, SUPER-CRITICAL or PROMPT CRITICAL) (0.5) The reactivity of the core is now pc (0.75) The change in reactivity required in going from 0.920 to 1.004 was pc (0.75)
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' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 4 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.08 (2.00)
An estimated critical position (ECP) has been calculated for a reactor startup that is to be performed 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> after a trip. The trip was preceded by a 60-day run at 100% of full-power. For each of the parts to this question, SPECIFY whether the indicated actions would result in an actual critical rod position that is LOWER-THAN, HIGHER-THAN, or THE-SAME-AS estimated in the ECP. ASSUME a negative MT feeding the steam generators to increase their level by 15% (0.5) delaying the startup by 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> longer than the planned 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> (0.5) increasing the steam-dump pressure setpoint by 100 psig (0.5) increasing the Pressurizer level by using the dilute mode of boron-concentration control (0.5)
QUESTION 1.09 (2.50)
ANSWER TRUE or FALSE to each statement given below concerning subcritical operation following a 3-month run at 100% powe If the reactor had been shutdown for 2 months, the WR NI instruments would lose their ability to determine the level of the neutron flux because the flux level would be too lo (0.5)
, The neutron flux level after a 2-month shutdown is determined l primarily by the neutrons produced by the intrinsic sources, the largest of which is due to alpha-particle absorption in 0-1 (0.5) If the indicated countrate by the WR NI instrument doubled, (1-keff) has been reduced by one-hal (0.5) For each equal insertion (addition) of reactivity, it takes a longer amount of time for an equilibrium neutron-flux level to be reached as keff approaches unit (0.5) If 10 inches of CEA withdrawal increased the countrate by a WR NI instrument by 10 cps, then 20 inches of CEA with-drawal would have increased the countrate by 20 cp (0.5)
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- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 5 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.10 (2.00)
ANSWER TRUE or FALS The heat flux required to produce transition boiling and the heat flux required to produce DNB are the sam (0.5) Film boiling occurs at lower heat flux levels than nucleate boilin (0.5) Bulk boiling will occur ONLY in hot channel (0.5) WITHIN the nucleate boiling region heat removal is usually not enhanced as the heat flux increase (0.5)
QUESTION 1.11 (3.00)
GIVE three (3) parameters associated with the primary side of the plant that affect DNB and that can be observed or controlled by the reactor operator. Note whether an INCREASE in each parameter will RAISE or LOWER the DNBR. For example: reactor power increase will lower DNBR. Do not use the example as part of the answe (3.0)
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QUESTION 1.12 (2.00)
- WHAT is the function of the reheaters? (0.5)
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I i QUESTION 1.13 (1.00)
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If a tank contained 5 lbm of wet vapor of 96% quality, the mass of water in the tank would be (1.0)
(a.) 0.1 lbm
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(b.) 0.2 lbm (c.) 4.8 lbm (d.) 4.94 lbm i
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' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 6 QUESTION 2.01 (3.00)
ANSWER the following parts of this question which pertain to the Containment Spray System, Upon receipt of an RAS, valves are automatically repositioned
- to accomplish three (3) things. LIST these three (3) impacts on the Containment Spray System alignmen (1.5) WHY is sodium hydroxide injected into the Containment Spray System upon receipt of a containment spray initiation signal? (1.0) WHERE in the Containment Spray System is the sodium hydroxide injected? (0.5)
QUESTION 2.02 (2.00)
ANSWER the following parts of this question concerning the Neutron Shield Tan WHERE is the tank located with respect to the primary shield wall? (0.5) WHERE is the tank located with respect to the main coolant pipes? (0.5) WHERE are the excore nuclear flux detectors located with respect to the Neutron Shield Tank? (0.5) With WHAT material is the tank filled? (0.5)
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' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 7 QUESTION 2.03 (2.00)
SELECT the most nearly correct value, The temperature rise of the RCS charging flow through the regeneration heat exchangers at normal operation at 100%
power is (1.0)
( deg F (b. 200 deg F (c. 300 deg F (d. 500 deg F The MAXIMUM possible flow through the Letdown Flow Control Valves under operational conditions at 100% power is closest to being about (1.0)
( /4 times the normal letdown flo ( /2 times the normal letdown flo times the normal letdown flo times the normal letdown flo QUESTION 2.04 (1.00)
The four (4) motor-operated valves for the safety injection tanks (SITS) are (1.0)
(f.) closed during normal Mode 1 operation to isolate the SITS from the RCS but receive an open signal on an SIA (b.) open during normal Mode 1 operation but receive an open signal on an SIA (c.) interlocked with the Pressurizer pressure to automatically open if the RCS pressure decreases below 500 psig.
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(d.) interlocked with the Pressurizer pressure to ensure that they are closed if the RCS pressure increases above 500 psi Comment: This question needs to be rewritten for applicability to Maine Yankee which is a 3 loop plant with 3 SIT (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
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' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 8 QUESTION 2.05 (1.50)
Briefly DESCRIBE WHY a red light indicating OPEN, for a Pressurizer Power Operated Relief valve, may not give the true status of this valv (1.5)
QUESTION 2.06 (3.00)
IDENTIFY those Quench Tank parameters that are automatically maintained within limits during normal operatio For each parameter identified, DESCRIBE the automatic feature associated with i (3.0)
QUESTION 2.07 (2.00)
GIVE the reasons for the following guides on RCP operatio Avoid prolonged pump operation in the 1025 to 1250 ps. rang (1.0) Operation of only one RCP should be avoided whenever possibl (1.0)
QUESTION 2.08 (3.00)
COMPLETE the following statement Operation without seal injection (RCP) should be avoided because . (1.0) Three (3) reasons for burnable poison rods being installed in selected fuel assemblies are , , and . (1.5) The charging pumps are designed to produce a .naximum flow at runout of gp (0.5)
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' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 9 QUESTION 2.09 (2.00) SPECIFY the conditions in the RCS which require operability of EITHER ONE of the charging pumps as components of the ECCS train (1.0) SPECIFY the reactor power level which requires operability of two (2) charging pumps as components of the ECCS train (1.0)
QUESTION 2.10 (1.00)
WHAT is the purpose of the back-pressure regulating valve (LD-F-16)? (1.0)
QUESTION 2.11 (1.50)
GIVE the normal operating parameters for the VCT when at 100%
power, namely (1.5) temperature pressure spray flow
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' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 10 QUESTION 2.12 (1.50)
ANSWER the following questions TRUE or FALSE regarding the steam dump and turbine bypass control syste ~ The steam dump and the turbine bypass are controlled by completely separate and independent control system (0.5)
4 The Steam Dump Valve Open alarm in the control room responds whenever any one of the Steam Dump Turbine Bypass valves leave the fully shut positio (0.5) If the Reactor Regulating System fails As-Is and a turbine trip occurs while at full power, the steam generator safety valves will lif (0,5)
Comment: Question 2.12a needs to be rewritten or not used in future exams. Although the reference states that there are 2 independent control systems, the same signal is used for the quick-open mode for both system '
QUESTION 2.13 (1.50)
The Main Steam System in addition to transporting steam from steam generators to the turbine, supplies steam to the . LIST three (3). (1.5)
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3.- INSTRUMENTS AND CONTROLS PAGE 11 QUESTION 3.01 (1.00)
The electrical buses from which the reactor coolant pump's motors receive electrical power are energized during (1.0)
(a.) startup from the 115 kV switchyard through reserve station transformer X1 (b.) startup from the 115 kV switchyard through the tertiary windings of the reserve station transformer X1 (c.) full-power operation from the main generator through the station service transformer X2 (d.) full-power operation from the 345 kV switchyard through the main transformer 1 QUESTION 3.02 (2.50)
ANSWER the following parts of this question concerning the Containment Isolation Syste When the control switch is in AUT0, SPECIFY the contain-ment pressure switch logic required to generate a CIS "A". (0.5) When the control switch is in AUTO, SPECIFY the contain-ment pressure setpoint that would generate a CIS "A". (0.5) When the control switch is in BLOCK, SPECIFY the logic required to generate a CIS "A". (0.5) When the control switch is in MANUAL, SPECIFY the logic required to generate a CIS "A". (0.5) SPECIFY two (2) valves that are closed by CIS "A" and not
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by SIAS "A" or "B". (0.5)
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. INSTRUMENTS AND CONTROLS PAGE 12 QUESTION 3.03 (2.00)
ANSWER TRUE or FALS For the three (3) Subcooling Margin Monitors (SMM), one (1) RCS pressure signal, when selected, supplies this parameter to all three (3) SMMs (0.5) the temperature signals to the Reactor Core SMM and to the Reactor Head SMM at any one (1) time are the same (0.5) the temperature for a Steam Generator SMM can be selected by a three-position selector switch labelled S/G 1, 2, 3 (0.5) the range of the core region SMM is from 100 deg to-40 deg F (0.5)
Comment: Part "c." should be rewritten to make it clea QUESTION 3.04 (1.00) WHAT kind of detector is used for the calibration of the fixed incore detector (0.5) WHY is this calibration necessary? (0.5)
QUESTION 3.05 (1.50)
The computer uses the incore instrumentation signals to generate four (4) outputs. LIST three (3) of thes (1.5)
QUESTION 3.06 (2.50) BRIEFLY identify the signals which the Zero Power Mode Bypass Switch defeats or block (1.5) WHAT is the reason for these being blocked or defeated? (1.0)
Comment: The CEA design has been modified so that the reason given in the answer is no longer valid. The question should be modified for future us (***** CATEGORY 03 CONTINUED ON NEXT PAGE ""***)
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' INSTRUMENTS AND CONTROLS PAGE 13 QUESTION 3.07 (2.00)
A CWP signal for the CEAs is provided when any of three (3)
signals have been generated. LIST them and indicate any plant condition for which the signals would be disable (2.0)
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QUESTION 3.08 (3.00) IDENTIFY three (3) temperature-related output signals generated by the Reactor Regulatory Syste (1.5) To WHICH systems are these directed? (1.5)
i QUESTION 3.09 (2.00)
Several interlocks exist regulating CEA withdrawal. For each interlock named, IDENTIFY the source of the interlock signa (2.0)
INTERLOCK GENERATED BY EXAMPLE: 1. upper electrical limit REED SWITCH 2. shutdown insertion permissive 3. upper sequential permissive 4. upper group stop
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5. regulating withdrawal interlocks l
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, INSTRUMENTS AND CONTROLS PAGE 13a Points
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FIGU?.E 3.10 (00ESTIC'O
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' INSTRUMENTS AND CONTROLS PAGE 14 QUESTION 3.10 (2.00)
Refer to Figure 3.10 (QUESTION) which shows a typical level-measuring system for such closed tanks as the Pressurizer and the Steam j Generator. ANSWER the following parts of this question by choosing the correct response, by " filling-in-the-blanks", or by completing the sentenc The output of the D/P cell is P(R) - P(V). This output is equal to the water density times (hl,h2,h3,h4, orh5). (0.5) If the reference leg broke and some of the water in the reference leg drained out, the output of the D/P cell, (P(R) - P(V)), would (INCREASE, DECREASE,or STAY-THE-SAME) . (0.5)
_ If there were a power-plant transient which quickly reduced [
the pressure in the tank, the water level in the reference
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leg would decrease due to . (0.5) Increasing containment temperature from a line break (not from the tank or its level-measuring system) would cause the indicated tank level to (INCREASE, DECREASE, or STAY-THE-SAME) . (0.5)
QUESTION 3.11 (1.00)
WHAT is available to the operator to display xenon oscillations? (1.0)
QUESTION 3.12 (1.50)
ANSWER the following parts of this question that pertain to the excore nuclear instrumentation and the wide-range log channels specificall J In each channel of detection, there are HOW MANY fission chambers? (0.5) Each channel provides indicaticn of the neutron-flux level over HOW MANY decades? (0.5) The fission chambers are the most sensitive to WHAT type of neutrons (THERMAL or FAST NEUTRONS)? (0.5)
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- INSTRUMENTS AND CONTROLS PAGE 15 QUESTION 3.13 (2.00)
ANSWER TRUE or FALS When 1% power is indicated on the wide range channels, the input is coming from only one fission chambe (0.5) The output of the upper and lower VIC signals in the power channels is sent to the compute (0.5) The SUR trip is inhibited above 15% power by a bistable
, from the wide range channel (0.5) The high power rate of change channel trip bypassed is initiated after any two (2) power range safety channel level I bistables trip; i.e., at power greater than 15%. (0.5)
QUESTION 3.14 (1.00) WHAT is meant by a negative tilt in the core offset? (0.5) WHAT is the equation / calculation that provides the core offset? (0.5)
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' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 16 RADIOLOGICAL CONTROL QUESTION 4.01 (1.00)
The Short-Term Calormetric Procedure is used to perfom the short-form calorimetric adjustment (1.0)
(a.) whenever rate-of-change of power has been greater than 10%/hr for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or longe (b.) at least once per shift if the computer becomes inoperabl (c.) at least once per shift during power operation at 100% powe (d.) whenever the flux delta-T power comparator alarms during power operation at 100% powe QUESTION 4.02 (1.00)
According to the " Emergency Shutdown from Power or Safety Injection" Procedure, E-0, if two (2) or more rod bottom lights are not lit, then two (2) alternate steps are outlined
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to " ensure reactor trip." SPECIFY these two (2) alternate step (1.0)
QUESTION 4.03 (2.00)
! One of the IMMEDIATE ACTION steps of FR-S.1, " Nuclear i Power Generation /ATWS" states, " Ensure Reactor Sub-l cri ti cal . " STATE the criteria for this ste (1.0) This procedure is entered from F-0.1, "SUBCRITICALITY g
STATUS TREE" on WHAT conditions? (SPECIFY the color (s).) (1.0)
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QUESTION 4.04 (1.00)
HOW is " ensure proper ECCS valve alignment" determined according to the Emergency Operating Procedures, E-07 (1.0)
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' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 17 RADIOLOGICAL CONTROL QUESTION 4.05 (1.00)
When an operator takes a reading of the value of a parameter for the Primary Operators Log and the value is out of specifications /
limits, the operator should (1.0)
(a.) enter the parameter value on the log entry as all other such log entries, prepare a DR report for the PSS/ SOS, and make an entry in the comment remarks section of the lo (b.) circle the parameter value on the log entry, contact the Plant Engineering Department, and have an entry made into the rough log of the P0 in the control roo (c.) circle the parameter value on the log entry, report the situation to the PSS/ SOS, and make an entry in the comment remarks section of the lo (d.) underline the parameter value on the log entry, contact the Plant Engineering Department, and have an entry made into the +-
rough log of the P0 in the control roo QUESTION 4.06 (2.00)
According to Abnormal Operating Procedure, A0P-2-25, "High Radiation Levels," initial and subsequent actions include efforts to verify an alarm to verify that the radiation level has indeed increase LIST two (2) steps that should be considered for verification of any radiation alar (1.0) If the containment particulate monitor alarmed, WHAT
.
other installed monitors could be checked to verify the alarm? LIST two (2). (1.0)
QUESTION 4.07 (1.50)
When conducting the Daily RMS Operational Test, activating the high alarm on the Primary Component Cooling (PCC) process radiation monitor should cause two (2) control-function actions (like open or close a certain valve). LIST the two (2) control-i'untion actions and INDICATE which action should be manually reset immediately following the monitor tes (1.5)
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' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 18 RADIOLOGICAL CONTROL QUESTION 4.08 (1.00)
According to the "P-2A and 2B Feed Pump Operation" Procedure, the MDFP suction valve must be open prior to opening the warming lin WHY is this a precaution? (1.0)
QUESTION 4.09 (1.50)
The second IMMEDIATE ACTION step in the Emergency Operating Injection,"
Procedure, E-0 states, "" EnsureEmergency Shutdown Reactor fromthe Trip." LIST Power threeor Safety (3) parts of this step; i.e., WHAT three (3) items must be checked? (1.5)
QUESTION 4.10 (1.50)
Under the " Response Not Obtained" column for the second IMMEDIATE ACTION step (ensure reactor trip) for E-0, an alternate action step is specified with respect to the CEDM-bus under-voltage status light WHAT is the location of these status lights (i.e., on which panel)? (0.5) This alternate action step concerning these status lights is an "IF ... THEN ..." type of action step. STATE this action ste (1.0) ,
QUESTION 4.11 (1.50)
Emergency Operating Procedure, ES-0.0, " Event Diagnosis" provides a flow chart to allow the operator to determine or confirm the most appropriate past-accident recovery procedure. The flow chart may determine that either procedure E-1 (or ECA-1), E-2 or E-3 should be implemented. SPECIFY the type of event associated with these three emergency procedure (1.5)
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' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 19 * \i / /
RADIOLOGICAL CONTROL ; )b'
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QUESTION 4.12 (1.00)
According to " Reactor Trip Response," ES-0.1, the Pressurizer l'evel shall be monitored and actions taken depending on the observed leve To WHAT level should the Pressurizer be taken? (0.5) h If the Pressurizer level was less than 5%, WHAT action .
should be taken? (0'.5) ,
e t <
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QUESTION 4.13 (2.00)
According to the " Core Reloading" Procedure, core alterations or " '
movement of irradiated fuel within the containment shall cease .
immediately if any one of several conditions occur. SPECIFY . r those by correct number designation from the list below that s would cause this stop-action to occu (2.0)
' The containment equipment hatch is open and the reactor has been subcritical for more than 210 hour0.00243 days <br />0.0583 hours <br />3.472222e-4 weeks <br />7.9905e-5 months <br /> s Both personnel hatch doors are closed and the reactor has been suberitical for less than 210 hour0.00243 days <br />0.0583 hours <br />3.472222e-4 weeks <br />7.9905e-5 months <br /> . The containment venting and purge inlet and outlet trip valves are open and a refueling purge is not in progres . The containment low-range area monitor has been inoperable for greater than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and the containment ventilation system isolation is close ( , One wide-range log channel is operating and providing visual
indication in the control room and is generating an audible
countrate in the containment, the other wide-range log channel is inoperabl . One boric acid transfer pump is not availabl Comment: There are no conditions for stop-action in the lis A careful rewording of the conditions would make this a good questio (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
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i PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 20
-
RADIOLOGICAL CONTROL
.p f@g 4 : s Y h; ,
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( QUESTION ,4.14? (1.50)
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Maximum permissible exposure to external and internal radiation is set forth in the Code of Federal Regulations, " Standards for Protection Against Radiation" (10CFR20). PROVIDE those limits by completing the following table.
.
Occupational radiation exposure for individuals 18 years or older will be limited to the following per calendar quarter:
s
., ,
A REM
,_ i Yhol'e body, head and trunk, active blood-forming organs, lens of eyes, or gonads (0.5)
' Hands and forearms, feet and ankles (0.5) Skin of whole body (0.5)
QUESTION 4.15 (1.00)
An individual in a restricted area may receive a dose to the whole body greater than that permitted above provided that additional criteria are satisfie <
w h One such criteria is a limit on the dose to the whole bod SPECIFY this limit in rem / quarte (0.5)
,' Another such criteria is a limit on the accumulated occupational dose to the whole body. SPECIFY this limi (0.5)
'
1 _
I QUESTION !4.16 (1.00)
l LIST those emergency conditions (s)/ classification (s), after the declaration of which, the operations centers (EOF, TSC and OSC) are activate (1.0)
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- PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 21 RADIOLOGICAL CONTROL QUESTION 4.17 (1.50)
CHOOSE from the list below the required initial conditions for the Vacuum Priming Syste (1.5) diffuser vacuum priming system is in operation traveling water screens' wash system is operable condenser water box outlet valves are open condenser vacuum is greater than 22 inches raw water system is in operation instrument air system is in operation and supplying the condenser water box vacuum breakers
.
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. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 22 RADIOLOGICAL-CONTROL QUESTION 4.18 (2.00)'
Section 3.6 of the Technical Specifications, " Emergency Core Coolin and Containment Spray Systems" addresses the operability requirements for the ECCS. The statements below are taken from this Sectio COMPLETE the-statement (2.0)
SPECIFICATION:
The following equipment must be operable whenever the' reactor coolant system temperature and pressure exceed deg F and psig:
1.- Two safety injection tanks set for automatic initiatio Each tank shall contain 11,200 +/- 500 gallons of water borated to at least 1720 ppm and pressurized with nitrogen to 230 psig +
10 psi, - 25 ps . One operable ECCS train consisting of the following subsystems of the trai Each sub estem includes the manual valves that are aligned and locked in the position required for safeguards operation, the automatically operated valves set for automatic operation or aligned and locked in the position required for safeguards operation, the controls set for automatic initiation where appropriate, and a pump powered from an engineered safeguards bu One high pressure safety injection pump subsystem . Station service power in accordance with Technical Specification , 3.12.A supplying the same operable ECCS train as in 2. above.
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l PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 23 !
THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW l l
ANSWERS -- MAINE YANKEE -86/10/21-CLARK, ANSWER 1.01 (2.00) Delayed neutrons significantly lengthen the neutron generation time [+0.5], slowing the response time of the power level to a reactivity change [+0.5]. As the core ages, beta-bar-eff decreases [+0.5]. Hence, the reactor period would be shorter which means that the SUR is larger [+0.5].
REFERENCE Maine Yankee: Reactor Operator Reactor Theory and Lesson Plans, R0-L-1.9, " Prompt and Delayed Neutrons," p.1 . Maine Yankee: Reactor Operator Reactor Theory and Lesson Plans, " Period, SUR, In-Hour," R0-L-1.10, pp. 2 through ANSWER 1.02 (1.00) poison pins reflector distribution of fuel assemblies with different enrichments use of CEA's in a symmetrical pattern
[+0.5 each, +1.0 maximum]
REFERENCE Maine Yankee: Reactor Operator Reactor Theory and Lesson Plans, R0-L-1.8, " Flux Distribution", p. 5.
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' PRINCIPLES 0F NUCLEAR POWER PLANT OPERATION, PAGE 24 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- MAINE YANKEE - 86/10/21-CLARK, ANSWER 1.03 (1.50) RCS average temperature power defect (moderator, fuel, void, pressure defects = +0.2 f f listed separately) power level fuel burnup xenon concentration samarium concentration
[+0.3 each, +1.5 maximum]
'
REFERENCE Maine Yankee: Reactor Operator Reactor Theory and Lesson Plans, " Flux Distribution," R0-L-1.8, p.1.
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' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 25 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- MAINE YANKEE -86/10/21-CLARK, ANSWER 1.04 (2.00) At POAH, fuel and moderator temperatures begin to rise, automatically inserting negative reactivity [+0.5], which the operator must override with positive reactivity additions to reach desired power [+0.3].
At BOC, the reactor may have a positive MTC [+0.3]. If the MTC is positive, the P0AH is of concern because it would represent the point at which additional reactivity is added to the core; which, in turn, would increase the SUR [+0.5]. T-ave increases [+0.3]
SUR decreases [+0.3]
PZR level and pressure increases (with CVCS response) [+0.3]
Turbine bypass valves open further to keep S/G pressure from increasing [+0.3]
[+0.9 maximum] At about 10**-2% power [+0.3]
REFERENCE Maine Yankee: Reactor Operator Reactor Theory and Lesson Plans, R0-L-1.11, " Nuclear Coefficients / Defects," pp. 19 and 20, and Viewgraph 13.6-2.
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' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 26 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- MAINE YANKEE -86/10/21-CLARK, ,
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ANSWER 1.05 (2.00)
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< ,_ :. 4, The removal of Xe-135 from the core by neutron absorption decreases immediately. The major production term, the decay of I-135, does not decrease immediately. Hence, immediately after the power reduction Xe production > removal. [+1.0]
REFERENCE Maine Yankee: Reactor Operator Reactor Theory and Lesson Plans, R0-L-1.12, " Poisons," p. 1 ANSWER 1.06 (2.00) Tave would increase [+0.5] S/G temperature and pressure would increase [+0.5] S/G flowrate would increase (due to lower enthalpy steam) [+0.5] no change in PZR level if the setpoint is in MANUAL OR PZR level would increase if setpoint is auto-controlled by the RRS [+0.5]
- REFERENCE Maine Yankee
- Reactor Operator Reactor Theory and Lesson Plans, R0-L-1.12, " Poisons," pp. 18 to 20.
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.
' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 27 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- MAINE YANKEE -86/10/21-CLARK, ANSWER 1.07 (2.00) super-critical [+0.5] pcm or 0.398% delta rho ;+075;
. pcm or 9.09% delta rho _+ 0.75 REFERENCE Maine Yankee: Reactor Operator Reactor Theory and Lesson Plans, R0-L-1.10, " Period, SUR, In-Hour," pp. 1 and . Maine Yankee: Reactor Operator Reactor Theory and Lesson Plans, R0-L-1.7, " Neutron Multiplication," p. 3 and Viewgraph 1.3- ANSWER 1.08 (2.00) lower-than lower-than higher-than lower-than
[+0.5each]
REFERENCE Maine Yankee: Reactor Operator Reactor Theory and Lesson Plans, R0-L-1.13, "ECP 1/M."
ANSWER 1.09 (2.50) false false true true false
[+0.5 each]
REFERENCE Maine Yankee: Reactor Operator Reactor Theory and Lesson Plans, R0-L-1.1 .
e -, . . , , , -- - - y - - - - - , -
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- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 28 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- MAINE YANKEE -86/10/21-CLARK, ANSWER 1.10 (2.00) true false true false
[+0.5 each]
REFERENCE Maine Yankee: Reactor Operator Heat Transfer, Lesson Plans R0-L-7.3, pp. 12 and 1 .
ANSWER 1.11 (3.00) RCS coolant flow rate [+0.5] increase will RAISE DNBR [+0.5] coolant inlet temperature [+0.5] increase will LOWER DNBR [+0.5] Pressurizer pressure [+0.5] increase will RAISE DNBR [+0.5]
REFERENCE Main Yankee: Reactor Operator Heat Transfer, Lesson Plan R0-L-7.3, pp. 14 and 1 ANSWER 1.12 (2.00) to remove the moisture from the HP turbine exhaust [+0.5]
" The reheaters allow for an increase in efficiency. By removing the moisture from the HP turbine exhaust they allow the S/G to operate at a higher pressure, which means a larger delta (h) in the HP turbine. [+1.5]
i l REFERENCE Maine Yankee: Reactor Operator Heat Transfer and Lesson Plans, Lesson Plan R0-L-7.2, " Cycles and Efficiency" pp. 15, 16.
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' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 29 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- MAINE YANKEE -86/10/21-CLARK, ANSWER 1.13 (1.00)
(b.) [+1.0]
REFERENCE Maine Yankee: Reactor Operator Heat Transfer and Lesson Plans, Lesson Plan R0-LP-L-7.1, "HTFF/Thermo Fundamentals".
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2. ' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 30 ANSWERS -- MAINE YANKEE -86/10/21-CLARK, ANSWER 2.01 (3.00) . Shift the spray pumps' suction from the' Refueling Water Storage Tank to the containment Safeguards sum . Unisolate cooling flow through the Residual Heat Removal heat exchanger . Supply water from the discharge of the containment Spray pumps to the charging (HPSI) pumps' suction header . Vents for the spray pumps are re-aligned to containment sum [+0.5 each, +1.5 maximum] to remove radioactive iodine from the containment atmosphere [+1.0] to the mixing chamber of the Refueling Water Storage Tank [+0.5]
REFERENCE Maine Yankee: Systems Training Manual, Chapter 8, " Containment Spray System," NS-7, pp. 4, 6, and ANSWER 2.02 (2.00) inside the primary shield [+0.5] the top of the tank is located (at the level of the core and)
below the main coolant pipes [+0.5] the detector wells are mounted against the inner wall of the shield tank [+0.5] chromated water [+0.f]
REFERENCE Maine Yankee: Systems Training Manual, Chapter 9, " Containment Building," NS-8, p. 1 .
' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 31 ANSWERS -- MAINE YANKEE -86/10/21-CLARK, ANSWER 2.03 (2.00) (c.) '+1. 0'
' (b.) l+ REFERENCE Maine Yankee: Systems Training Manual, NS-4, Para. 5.0, pp. 99 and 10 . Maine Yankee: Systems Training Manual, NS-4, Para. 5.2, p.10 ANSWER 2.04 (1.00)
(b.) [+1.0]
REFERENCE Maine Yankee: Systems Training Manual, NS-5, "ECCS," Vol. 2, Para. 4.1.2, p. 10 ANSWER 2.05 (1.50)
The valve is operated normally by a OPERATES the red (OPEN) indication +0.5]. [ pilot Thus valve the redWHOSE (OPEN) CONTROL CIRCUIT light describes pilot valve operation [+0.5]. But if the PORV STICKS and does not open by actuation of its pilot valve, the RED light gives a false signal [+0.5].
REFERENCE Maine Yankee: Systems Training Manual, NS-3, Chapter 3, Para. 3.1.4, pp. 12 and 4 .
' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 32 ANSWERS -- MAINE YANKEE -86/10/21-CLARK, ANSWER 2.06 (3 '0)
LEVEL [+0.25] . Level control monitor [+0.75] turning two pumps en and off at pre-set level TEMPERATURE [+0.25]. High temperature will start the two pumps
[+0.75] to pump through Quench Tank heat exchanger (cooler) until temperature reaches preset control valu PRESSURE [+0.25] (The quench tank is pressurized as needed by the addition of N2 gas to maintain 2 psig.) The tank will automatically vent on increasing pressure (beyond 18 psig) to the hydrogenated vent header [+0.75] .
REFERENCE Maine Yankee: Systems Training Manual, NS-3, "PRZR Quench Tank,"
Para. 2.2.3, pp. 22 through 2 o ANSWER 2,07 (2.00) To avoid thrust shift reversal conditions causing excessive shaft movement and seal damage [+1.0]. During single pump operation the pump will tend to run out on its capacity curve to give high pump vibration. [+1.0]
REFERENCE Maine Yankee: Systems Training Manual, NS-2, "RCC," pp. 58 and 5 ..
- PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 33 ANSWERS.-- MAINE YANKEE -86/10/21-CLARK,- '
ANSWER 2.08 (3.00) crud from the primary system could get into the seal and
. eventually cause seal failure [+1.0] control excess reactivity at BOC [+0.5]
reduce initial reactor coolant boron concentration [+0. 5]
suppress flux peaks [+0.5] to 750 [+0.5]
REFERENCE Maine Yankee: Systems Training Manual, NS-1, Para. 2.1.2, p. . Maine Yankee: Systems Training Manual, NS-2, p. 5 . Maine Yankee: Systems Training Manual, NS-4, "CVCS," Par .1.13, p. 3 ANSWER 2.09 (2.00) When RCS temperature and pressure are greater than 210 deg F
[+0.5] and 400 psia [+0.5] .
. Reactor power > 2% [+1.0], boron concentration ( hot shutdown boron concentration [+0.5]
[+1.0 maximum]
REFERENCE Maine Yankee: Systems Training Manual, NS-4, Para. 5, p. 9 . Maine Yankee: Technical Specifications, Section 3.1 ANSWER 2.10 (1.00)
To maintain upstream system pressure at about 300 psig to prevent water from flashing to steam in letdown lin [+1.0]
REFERENCE Maine Yankee: Systems Training Manual, NS-4, Para. 1.2.1, p. . .
. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 34 ANSWERS -- MAINE YANKEE -86/10/21-CLARK, ANSWER 2.11 (1.50)
Any value between deg F psig gpm
[+0.5each]
REFERENCE Maine Yankee: Systems Training Manual, NS-4, Para. 5.2 and 5.3, pp. 98 through 10 . Maine Yankee: Surveillance Procedure, 3. ANSWER 2.12 (1.50)
' True False False
[+0.5 each]
REFERENCE Maine Yankee: Systems Training Manual, Chapter 25, " Steam Dump Turbine Bypass System," PGS-14, pp. 1 through ANSWER 2.13 (1.50)
' auxiliary steam system reheat steam system
. steam dump and turbine bypass system l turbine driven auxiliary feed pump turbine driven main feed pump
[+0.5 each, +1.5 maximum]
REFERENCE l
I Maine Yankee: Systems Training Manual, Chapter 22, " Reheat Steam," Para. 1.1.1, p. 1.
.
. INSTRUMENTS AND CONTROLS PAGE 35 ANSWERS -- MAINE YANKEE -86/10/21-CLARK, ANSWER 3.01 (1.00)
(c.) [+1.0]
REFERENCE Maine Yankee: Systems Training Manual, Volume 7, Chapter 32,
"High Voltage Electrical Systems," PGS-17, Rev. 3, p through 6 and Figure PGS-17- ANSWER 3.02 (2.50) /3 from 3 containment pressure sensors [+0.5] psig or 20 psia [+0.5] /3 from 3 containment pressure sensors AND Pressurizer pressure
> 1685 psig, 2/4 from 4 sensors [+0. 5] no other signals required [+0.5] . PCC return RCPs RCP seal return SL-M-29 RCP seal return SL-M-40 RCP seal return SL-M-51 PCC supply RCPs
[+0.25 each, +0.5 maximum]
REFERENCE Maine Yankee: Systems Training Manual, Chapter 6, " Emergency Core Cooling System," NS-5, pp. 43, 44, and 45; and Figure NS-5-12.
e
'S INSTRUMENTS AND CONTROLS PAGE 36 ANSWERS -- MAINE YANKEE -86/10/21-CLARK, ANSWER 3.03 (2.00) True False False False
[+0.5 each]
REFERENCE , Maine Yankee: Systems Training Manual, NS-3, Para. 2.2.4, p through 3 ANSWER 3.04 (1.00) a fission chamber [+0.5] the SPND signals (fixed incore detectors) change with exposure
'
to neutron flux due to burnup and this change must be measured for correction by recalibration [+0.5]
REFERENCE Maine Yankee: Systems Training Manual, Chapter 13, "Incore Instrument System," X/S-11, Para. 1.2, p. 3.
,
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- ANSWER 3.05 (1.50) Tq calculation INCA printout . corethermocouplemap)
various MPX (multiplex signals i
[+0.5 each, +1.5 maximum]
REFERENCE f Maine Yankee: Systems Training Manual, NS-11, Chapter 13, "Incore Instrumentation System," Para. 4.1, p. 16.
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3.- INSTRUMENTS AND CONTROLS PAGE 37 ANSWERS -- MAINE YANKEE- -86/10/21-CLARK, R.
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ANSWER 3.06 (2.50)
a.- Rx trip from low RCS coolant flow [+0.5] Rx trip from TM/LP trip [+0.5] dT input to the TM/LP calculator is blocked [+0.5]
l to allow CEAs to be pulled to five'(5) steps to prevent their binding during cooldown [+1.0]
REFERENCE Maine Yankee: Systems Training Manual, Chapter 11, " Reactor Protective System," NS-12, p. 4 ANSWER 3.07 (2.00)
. pre-trip from variable high power [+0.5] pre-trip from high rate of change of neutron flux [+0.5]
'
disabled below 10**-4% power [+0.25]
i pre-trip from TM/LP [+0.5]
disabledbelow10**-4% power [+0.25]
REFERENCE l Maine Yankee: Systems Training Manual, Chapter 12, "CEA Control System," NS-9, Para. 2.2.2.2, p. 15.
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. INSTRUMENTS AND CONTROLS PAGE 38
' ANSWERS -- MAINE YANKEE -86/10/21-CLARK, ANSWER 3.08 (3.00) . Tave - Tref alarm signals (HI and LO) PZR level setpoint Tave indication (MCB and 2 per recorder) Tref indication (2 per recorder) steam-dump demand signal quick-open override signal
[+0.5 each, +1.5 maximum] . steam dump system turbine bypass system annunciator system plant computer system PZR level control system
[+0.5 each, +1.5 maximum]
REFERENCE Maine Yankee: Systems Training Manual, Chapter 12, "CEA Control System," NS-9, Para. 2.1, p. 9 and Figure NS-9- ANSWER 3.09 (2.00) reed switch (given) reed switch [+0.5] computer (pulse counting) [+0.5] computer [+0.5]
, reed switch [+0.5]
REFERENCE Maine Yankee: Systems Training Manual, Chapter 12, "CEA Control System," NS-9, Para. 3.2, pp. 45 through 47.
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3. ' INSTRUMENTS AND CONTROLS PAGE 39 ANSWERS -- MAINE YANKEE -86/10/21-CLARK, ANSWER 3.10 (2.00) h2 [+0.5] decrease [+0.5] reference-leg water flashing to steam and flowing out of the condensing pot [+0.5] increase [+0. 5]
REFERENCE Generic: Combustion Engineering Document, " Controllers and Process Instrumentation," pp. 969 (82W3)/ds-29 -3 ANSWER 3.11 (1.00)
plot of symmetric-offset values [+0.5] from the plant computer [+0.5]
hand plot of symmetric-offset values [+0.5] with values from the plant computer or the RPS [+0.5]
[+1.0 maximum]
REFERENCE Maine Yankee: Systems Training Manual, Chapter 13, "Incore Instrumentation System," NS-11, Para. 5.2.1, p. 2 ANSWER 3.12 (1.50) or 9 thermal
[+0.5 each]
REFERENCE Maine Yankee: Systems Training Manual, Chapter 14, "Excore i
Nuclear Instrumentation System," NS-10, Para. 2.1.2, p. 5.
_
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. INSTRUMENTS AND CONTROLS PAGE 40 ANSWERS --' MAINE YANKEE -86/10/21-CLARK, ANSWER 3.13 (2.00) True True False False
[+0.5 each]
REFERENCE Maine Yankee: Systems Training Manual, Chapter 14, "Excore Nuclear Instruments System," NS-10, pp. 28 through 3 ANSWER 3.14 (1.00)
s A negative tilt means lower core power is greater than upper core power. [+0.5] It is determined from the ratio of the output signals of the UIC upper (U) and lower (L) power safety channels as U-L
[+0.5]
U+L REFERENCE Maine Yankee: Systems Training Manual, Chapter 11, "RPS,"
NS-12, Para. 2.1.10, p. 2 .
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. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 41 RADIOLOGICAL CONTROL ANSWERS -- MAINE YANKEE -86/10/21-CLARK, ANSWER 4.01 (1.00)
(d.) [+1.0]
REFERENCE Maine Yankee: Procedures, "Short Form Calorimetric," Proc. N , Rev. 6, p. !!
ANSWER 4.02 (1.00) manually trip CEDM ;enerator set output breakers manually open and reclose the 480 V supply breakers to buses 9 and 12 if all trippable CEAs are NOT inserted, then go to FR- [+0.5 each, +1.0 maximum]
REFERENCE Maine Yankee: Emergency Operating Procedures, " Emergency Shutdown from Power or Safety Injection," E-0, Rev. O, p. ANSWER 4.03 (2.00) . Check WR power channels - LESS THAN 2% Check start-up rate channels - ZERO or NEGATIVE
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[+0.5each] red or orange [+0.5each]
REFERENCE Maine Yankee: Function Restoration Procedures, " Nuclear Power Generation /ATWS," FR-S.1, pp. 1 and _ _ - _ _ - _ _ _ _ _ _
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. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 42 RADIOLOGICAL CONTROL
. ANSWERS -- MAINE YANKEE-86/10/21-CLARK, ANSWER 4.04 (1.00)
Verify on the SIAS lightbox indicators that all lights are green and not flashing. [+1.0]
REFERENCE Maine Yankee: Emergency Operating Procedures, " Emergency Shutdown from Power or Safety Injection," E-0, Rev. O, p. ANSWER 4.05 (1.00)
(c.) [+1.0]
REFERENCE Maine Yankee: Administrative Procedures, " Conduct of Operations,"
1-200-10, Maine Yankee Primary Operators Logs, MY-72-84, Rev. 9, p. . Maine Yankee: Administrative Procedures, " Conduct of Operations,"
1-200-10, Rev. O, p. 2 ANSWER 4.06 (2.00) . check alarmed channel for correct check area with portable meter take samples use other meters - portable or in-place
[+0.5 each, +1.0 maximum] . containment gas monitor 2, containment low area monitor crane area monitor
[+0.5each,+1.0 maximum]
REFERENCE Maine Yankee: Abnormal Operating Procedures, "High Radiation Levels," Proc. No. AOP-2-25, Rev. 12, pp. 2 and ~
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. PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE 43 RADIOLOGICAL CONTROL ANSWERS -- MAINE YANKEE -86/10/21-CLARK, ANSWER 4.07 (1.50) shut the PCC surge tank vent [+1.0] the PCC surge tank vent should be manually reset immediately
[+0.5]
REFERENCE Maine Yankee: Procedures, " Daily RMS Operational Test," Pro No. 3-6.2.2.17, Rev. 12, p. ANSWER 4.08 (1.00)
If the MDFP suction valve is not open, the full discharge pressure of the pump is put on the pump suction side, which would exceed the pressure rating of the suction-side piping. [+1.0]
REFERENCE Maine Yankee: Secondary Operating Procedures, "P-2A and 2B Feed Pump Operation," Proc. No. 1-104-1, Rev. 15, p. ANSWER 4.09 (1.50) ensure reactor trip breakers open ensure wide-range power levels decreasing ensure rod bottom lights on
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[+0.5 each]
REFERENCE Maine Yankee: Emergency Operating Procedures, " Emergency Shutdown from Power or Safety Injection," E-0, Rev. O, p. m a
4 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 44 RADIOLOGICAL CONTROL ANSWERS -- MAINE YANKEE -86/10/21-CLARK, ANSWER 4.10 (1.50) RPS trip-status panel [+0.5] IF two or more CEDM-bus under-voltage status lights are not lit [+0.5], THEN manually open and reclose the 480 volt supply breakers to bus 9 and bus 12 [+0.5].
REFERENCE Maine Yankee: Emergency Operating Procedures, " Emergency Shutdown from Power or Safety Injection," E-0, Rev. O, p. ANSWER 4.11 (1.50)
. E-1, loss of primary or secondary coolant E-2, steam-line break E-3, S/G tube rupture
[+0.5each]
REFERENCE Maine Yankee: Emergency Operating Procedures, " Event Diagnosis,"
ES-0.0, Rev. O, p. ANSWER 4.12 (1.00) % [+0.5] Ensure SIAS or manually initiate SIAS [+0.5]
REFERENCE Maine Yankee: Emergency Operating Procedures, " Reactor Trip
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Response," ES-0.1, Rev. O, pp. 5 and 6.
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.. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 45 RADIOLOGICAL CONTROL ANSWERS -- MAINE YANKEE -86/10/21-CLARK, ANSWER 4.13 (2.00)
1., 3., 5., and [+0.5 each for 1. , S., and 6 ; -0.5 each for 2. and 4.; + minimum, +2.0 maximum)
REFERENCE Maine Yankee: Procedures, " Core Reloading," Proc. No. 10-1, Rev. 9, pp. 4 and ANSWER 4.14 (1.50) .25 rem OR 3 rem with an updated NRC form-4 and not to exceed 5(N-18) .75 .50
[+0.5 each] ,
REFERENCE Maine Yankee: Radiation Protection Manual, Section 2.3, p. 2- ANSWER 4.15 (1.00) rem / quarter [+0.5]
f I (N-18) rem, N= age OR 75 rem and 25 rem for life saving and emergency actions [+0. 5]
REFERENCE Maine Yankee: Radiation Protection Manual, Section 2.3, pp. 2-2 and 2-3.
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~' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 46 RADIOLOGICAL CONTROL ANSWERS -- MAINE YANKEE -86/10/21-CLARK, .
ANSWER 4.16 (1.00) alert site emergency general emergency
[+0.33 each]
REFERENCE Maine Yankee: Emergency Plan, Section 6.0, " Emergency Measures,"
p. 6.2 and Figure 6- ANSWER 4.17 (1.50)
3., S., and [+0.5each]
REFERENCE Maine Yankee: Secondary Operating Procedures, " Vacuum Priming System" Proc. No. 1-100-1, Rev. 7, p. ANSWER 4.18 (2.00) one service water pump subsystem one component cooling pump subsystem one low pressure safety injection pump subsystem one containment spray pump and RHR heat exchanger subsystem
[+0.33 each]
REFERENCE Maine Yankee: Technical Specifications, Section 3.6, " Emergency Core Cooling and Containment Spray System," p. 3.6- '
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Class A
- F- Rev. N Rev. 0 -
,*' SUSCRITICALITY STATUS TREE Issue Date 7-1-86 '
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' Re.lew Date 7/88 '
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FR- / \ NO WR POWER () CHANNELS
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RANGE YES STARTUP WO RATE WORE NEGATIVE THAN
-C.Z DP M YE3 i . .
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STARTUP NO
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.___-___-_m.-_-.__.______.__,____.,_,.,___- . , _ _ . _ _ _ _ , _ _ _ _ _ _ _ _ _ _ _ . _ _ . . , _ _ _ . __ _
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_________...___._-_______..__-_-____-_________-_-___._______..____-_______
EQUATION FORMULA AND PARAMETER SHEET
____________ __._______ -_____-_..______________-___-__--___--__.---___ ..__
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, Where :a1 = m2
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(density)3(velocity)3(area)1 = (density)2(velocity)2(area)2
-! .-__.....__.___________.____-__..____________________________________.__ s t \u PE = mgh PEy + KE3+P3V3 = PE 2+KE2+P Y22 where V = specific Kh,=!} volume x P = pressure e
____________________________ .________ ____________________________... ___ Q = mcp(Tout -Tin) Q = UA (T,y,-Tstm) Q = m(hi-h2)
_..__.._____________________________________________________________________
(# - p)Em t
SUR = 26.06 T= = (d - p)
P = Po 10(SUR)(t) P = Po e /T I p p A,ff
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s ___________________________________ .__________..__.._________________..___
. x 1- t; delta K = (K,ff-1) CR1 (1-K,ff 3) = CR2 (1-Keff2) CR = S/(1-K,ff)
4 s 1 \
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(1-K,ff)) (1-X,ff) x 100% ,1
- ' .
SDM = A'If = 0.08 see K
M = (1-Keff2 I eff
...________________--_-_____________...__________________.-_________________
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decay constant = in (2) = 0.693 A 3 = Ag e (decay constant)x(t)
t 1/2 ;1/2
_________ . ..______________________-_____.-________________________________
Water Parameters fliscellaneous Conversions
1 gallon = 8.345 lbs 1 Curie = 3.7 x 10 dps j
1 gallon.= 3.78 liters 1 kg = 2.21 lbs 1 ft 3= 7.48 gallons 1 hp = 2.54 x 103 Btu /hr : Density =62.4lbg/ft 1 MW = 3.41 x 10 Btu /hr
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Density = 1 gm/cm 1 Btu = 778 ft-lbf l Heat of Vaporization = 970 Btu /lbm Degrees F = (1.8 x Degrees C) + 32 i Heat of Fusion = 144 Btu /lbm 1 inch = 2.54 centimeters 2 1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 ft-lbm/lbf-sec
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i ( 1 ft H 2O = 0.4335 lbf/i ___....._____..__-____..______ -__________..._______ ..____________________ ,
rTBob rne0Y /
U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: MAINE YANKEE REACTOR TYPE: PWR-CE DATE ADMINISTERED: 86/10/21 EXAMINER: UPTON, J. / CLARK, CANDIDATE: AUSUER KEY
_ INSTRUCTIONS TO CANDIDATE:
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Use separate paper for the answers. Write answers on one side onl Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination start % OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 25.00 25.00 THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 25.00 25.00 PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 25.00 25.00 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 25.00 25.00 ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 100.00 Totals Firal Grade
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All work done'on this examination is my ow I have neither*given
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nor received aid.
Cancidate's Signature .
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS tearing 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 . Print your name in the blank provided on the cover sheet of the exami .;atio . 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 only on o_ne side of the paper, and write "Last Page" on the last answer shee . Number each answer as to category and number,' for example,1.4, . Skip at least three lines between each answe . Separate answer sheets from 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 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
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the examiner onl . .
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17. Ycu 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 completed.
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18. When you complete your examination, you shall:
a. Assemble your examination as follows:
(1) Exam questions on to (2) Exam aids - figures, tables, et (3) Answer pages including figures which are part of the answe b. Turn in your copy of the examination and all pages used to answer the examination question c. Turn in all scrap paper and the balance of the paper that you did not use for answering the question d. 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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' THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 2 THERMODYNAMICS QUESTION 5.01 (1.00)
As a result of core burnup, beta-bar (eff) will change. This change in beta-bar (eff) would mean a change in the reactor period that would follow the insertion of a small amount of positive reactivit As the core ages from BOL to EOL, beta-bar (eff) will (1.0)
(a.) increase and the stable, positive reactor period will be large (b.) increase and the stable, positive reactor period will be smalle (c.) decrease and the stable, positive reactor period will be large (d.) decrease and the stable, positive reactor period will be smalle Comment: The question should address SUR instead of reactor pericd, seeing as the Maine Yankee instrumentation and procedures do not refer to reactor perio QUESTION 5.02 (2.00)
EXPLAIN the fact that tiie xenon poison (Xe-135) concentration in the reactor core increases after a reactor trip. Assume that the power plant had been operating at full power for 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> prior to the tri (2.0)
QUESTION 5.03 (1.00)
Listed below are six (6) radioactive elements that might be found in the RCS wate LIST those which would NOT indicate a break in the. fuel claddin (1.0)
l 1.- I-131 l l Xe-133 Co-60 Kr-85 -
l Fe-55 l
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- THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3 THERMODYNAMICS QUESTION 5.04 (3.00)
The reactor is to be taken critical after having been shut down due to a reactor trip from 100% full power. The estimated critical position (ECP) for the rods has been calculated. Consider each of the following situations separately and answer, "Will the actual ECP be HIGHER, THE SAME, or LOWER than the calculated ECP7" (HIGHER =
CEAs further out.) Briefly JUSTIFY your choice, The actual boron concentration is lower than that used in the calculation for the EC (1.0) Startup was delayed by four (4) hours beyond the time used in the calculation for the ECP. A shutdown time of sixteen (16)
hours was used in the calculation (1.0) The steam bypass / dump pressure setpoint is reduced 50 ps (1.0)
QUESTION 5.05 (2.00)
The data given below was taken during fuel loading into the cor WHAT is the predicted number of assemblies to obtain criticality?
(ASSUME that the assemblies are of equal worth.) SHOW your work and USE the graph provided if you desir (2.0)
Number of Assemblies Loaded Neutron Count Rate (cps)
10 75 15 105 e ,
1/M .
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10 15 20 25 30 35 40 No. of Assemblies (***** CATEGORY 05 CONTINUED ON NEXT PAGE *"***)
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' THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 4 THERMODYNAMICS QUESTION 5.06 (1.50)
Questions 5.06 and 5.07 assume that the Maine Yankee power plant has been operating at steady-state 80% of full power for 10 days. The fuel burnup status is that the core has reached 10 KMWD/MT. USE any of the provided figures and tables from the Technical Data Book to obtain your answer On your shift, you note that all CEAs were fully withdrawn, WHAT would be the critical boron concentration? (0.5) However, if the CEAs in Manual Sequential (MS) were at 520 steps withdrawn, WHAT would be the critical boron concentration? (1.0)
QUESTION 5.07 (3.00)
Again, assume that the Maine Yankee power plant has been operating at steady-state 80% of full power for 10 days. The fuel burnup status is that the core has reached 10 KMWD/MT. USE any of the
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provided figures and table Consider that on your next shift (1 day later) you observe that Tc is now higher than the previous (programmed) Tc by 2 deg F, the reactor coolant flowrate is the same, the turbine / generator output power is the same and the power plant is stable. EXPLAIN HOW and WHY each of the following parameters would have change Th (1.0) steam temperature and pressure ( steam flow rate ( ~
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QUESTION 5.08 (1.50)
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A rupture of a main steam line is a more severe incident (a more limiting condition) at E0C than at B0C. Briefly EXPLAI (1.5)
("**** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
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- THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 5 THERMODYNAMICS QUESTION 5.09 (1.00)
A variable-speed centrifugal pump (in a closed .subcooled system)
is rotating at 400 rpm, with a 200 gpm flowrate, at a discharge head of 30 psi drawing 60 kW of power. Pump speed is increased to 800 rpm. WHAT is the new flowrate and discharge head? (1.0)
QUESTION 5.10 (3.00)
GIVE three (3) parameters on the primary side affecting DNB that can be observed or controlled by the reactor operator, noting whether an INCREASE in each parameter will RAISE or LOWER the DNB For example: reactor power increase will lower DNBR. Do not use the example as part of the answe (3.0)
QUESTION 5.11 (1.50)
If the steam generator was operating at 900 psia, WHAT would be the moisture content of the main steam entering the MSRs?
STATE any assumption (1.5)
QUESTION 5.12 (2.00)
l The following refers to conditions during a natural circulation l
cooldown.
t WHAT is the primary indicator of a loss of natural circulation i flow? (0.5)
' If the feedwater flowrate to the Steam Generators was reduced, EXPLAIN the impact on natural circulation by explaining the l
effect on the primary coolant as it affects density, flowrate l and temperature (1.5)
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' THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 6 THERMODYNAMICS QUESTION 5.13 (1.00)
During a natural circulation cooldown assume that the steam generator pressure is 900 psia. WHAT RCS pressure would ensure that Tc is 40 deg F subcooled? (1.0)
QUESTION 5.14 (1.50) Heat transfer through a fuel-rod gas gap can be considered to be primarily conduction heat transfer. If the gas gap thick-ness is 4.25 x 10**-3 inches and Q-dot /A = 3.36 x 10**4 Btu /hr-ft**2, DETERMINE the gas gap delta Assume k(gas) =
0.125 Btu /hr-ft-deg (1.0) WHAT design feature of a fuel pin accounts for the heat-transfer coefficient used in this question? (0.5)
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(***** END OF CATEGORY 05 *****)
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' PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 7 QUESTION 6.01 (3.00)
The diagram below shows the four (4) 4160 VAC buses and the eight (8) 480 VAC buses. DRAW connecting lines between these buses showing the NC (normally closed) connections from the 4160 VAC buses to the 480 VAC buses and the NC (normally closed) bus ties among the 4160 VAC buse (It is not necessary to indicate transformers or breakers. The connections may have breakers and/or transformers in them, but they do not have to be indicated / drawn.) (3.0)
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4 l +t6 o 4i & c> 9i6ol R" ' Bus I ses 6 3,,s 5
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'? ks 'D ' Fus !3 Ba.c 7 Jes 7 fas if fas si ga,' sp, dro QUESTION 6.02 (2.00)
While operating at full power, DG-1A receives an emergency start signal,
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immediately tripped open automatically? (1.0) WHAT load breakers are immediately shut automatically? (1.0)
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PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 8 sTION 6.03 (1.50)
iWER TRUE or FALS r the three (3) Subcooling Margin Monitors (SMM),
On] (1) RCS pressure signal, when selected, supplies this pirameter to all three (3) SMM (0.5)
The temperature signals to the Reactor Core SMM and the Reactor Head SMM at any one (1) time are the sa (0.5)
The temperature monitor for a Steam Generator SMM can be selected by a three (3) position selector switch labelled S/G 1, 2, (0.5)
iTION 6.04 (1.50)
iWER TRUE or FALSE to the following statements about the itrols/ indications at the Primary Auxiliary Building Emergency iel (PABEP).
A controller for adjusting charging flow is availabl (0.5)
Pr:ssurizer pressure indication is available provided that a selector switch is switched to LOCA (0.5)
Main steam pressure indication is available without the need to place a selector switch in LOCA (0.5)
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-' PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 9 QUESTION 6.05 (3.50)
ANSWER the following parts of this question with respect to the -
Steam Dump and Turbine Bypass Syste SPECIFY the. number of ' valves that are temperature controlled and the number that are pressure controlle (1.0)
- Each temperature control valve has a maximum flow-capacity of WHAT percent of full-load steam flow? (0.5) SPECIFY the condition of the pressure control valves that would generate an alarm " STEAM DUMP VALVE OPEN." (0.5) SPECIFY the two (2) plant conditions that would provide an -
interlock preventing any dump valve from openin (1.0) With the reactor power greater than a certain level, a turbine trip would fully open all of the dump and bypass valves in approximately three (3) seconds. SPECIFY this power level. (Assume that all of the controllers are operating as programmed.) (0.5)
QUESTION 6.06 (1.50) ,
The decay-heat release valve, MS-A-162, allows removal of reactor decay hea WHAT is the capacity of this valve? SPECIFY your answer in percent of full-power steam flow or in lbm/h (0.5)
' WHAT signal (s) would automatically close this valve? (1.0) -
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j QUESTION 6.07 (1.50)
!. DESCRIBEthefunctionANDtheoperatingconsequencesofoperating
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the RAMP push button en the reheat steam system. In particular, SPECIFY the power level at which the RAMP push button should be
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, used and SPECIFY numerically the planned consequence (1.5)
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' PLANT StaTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 10 QUESTION 6.08 (3.00)
Liquid waste effluent flow is controlled by FC-3802 and monitored radiation monitor RI-380 DESCRIBE how the flow is controlled when the controller is in AUTOMATI INCLUDE in your answer the alignment of the flow control valves, the use of the flow control valve selector, and the flowrates attainable through the flow control valve (2.0) HOW does a high radiation signal from RI-3801 terminate the liquid waste release? (An incorrect answer would be that the signal removes the AC power from the test tank pumps.) (1.0)
QUESTION 6.09 (1.50)
Briefly STATE what condition (s) the following RPS trip functions are designed to preven Low Steam denerator Secondary Pressure (0.5) Loss of Coolant Flow (0.5) Axial Flux Offset (0.5)
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QUESTION 6.10 (1.00) WHAT is meant by a negative tilt ' the core offset? (0.5)
- WHAT is the equation / calculation that provides the core offset? (0.5)
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QUESTION 6.11 (3.00)
SPECIFY the trip or pretrip setpoints for the following RPS parameter RCS low coolant flow trip (0.5) Pressurizer pressure Hi pressure trip (0.5) high power rate of change trip (0.5) maximum variable overpower (0.5) Turbine trip (loss of load) (0.5) containment pressure trip (0.5)
Comment: 6.11b was deleted. There is no trip on L0 PZR pressur The question should be re-written to remove "or pretrip,"
seeing as there were no requests for pretrip setpoint In addition, the questt should specify that trip setpoints are required .ir the answer and not the Tech-Specs limits (if different).
a QUESTION 6.12 (1.00)
Fill-in-the-blanks with correct wordin The CEA Position Deviation High and High High occurs when any CEAs in a group deviate by steps and by step (1.0)
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QUESTION 6.13 (1.00) ,
GIVE four (4) components and/or svstems that the boric acid
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' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 12 !
RADIOLOGICAL CONTROL i
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I QUESTION 7.01 (1.00)
The "Short-Term Calormetric" Procedure is used to perform the short-form calorimetric adjustment (1.0) l l
(a.)wheneverrate-of-changeofpowerhasbeengreatertlian10%/hr for one hour or longe j (b.) at least once per shift if the computer becomes inoperabl ,
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(c.) at least once per shift during power operation at 100% powe (d.) whenever the flux delta-T power comparator alarms during power operation at 100% powe QUESTION 7.02 (1.00)
One of the IMMEDIATE ACTION steps of FR-S.1, " Nuclear Power Generation /ATWS" states, " Ensure Reactor Subcritical."
STATE the criteria for this ste (1.0)
QUESTION 7.03 (1.00)
According to the " Emergency Shutdown From Power or Safety Injection" Procedure, E-0, if two (2) or more rod bottom lights are not lit, then two (2) alternate steps are outlined to " ensure reactor trip." SPECIFY these two (2) alternate step (1.0)
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QUESTION 7.04 (.50) ,
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The' Daily RMS Operational Test Procedure shall be routed to for review and action on reported malfunction (0,5)
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- PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 13 RADIOLOGICAL CONTROL !
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QUESTION 7.05 (1.50)
ANSWER the following parts to this question which pertain to the Daily RMS Operational Tes When the Containment Vent and Purge System is in the ON-LINE purge mode of operation and the test is to be made on RI-3902Y (Primary Vent Gas Monitor), 'WHAT shall be accomplished with the Containment Vent and Purge System prior to testing this monitor? (0.5) When the containment Vent and Purge System is in the REFUELING mode of operation and the test is to be made on RI-6104 (Manipulator Crane Monitor), WHAT may be accomplished with the Containment Vent and Purge System prior to testing this monitor (to prevent cycling of the exhaust valves and purge supply fan)? (0.5) Under WHAT plants conditions can RI-6104 be successfully
" source checked?" (0.5)
QUESTION 7.06 (1.50)
CHOOSE from the list below the required initial conditions for the Vacuum Priming Syste (1.5) diffuser vacuum priming system is in operation traveling water screens' wash system is operable condenser water box outlet valves are open condenser vacuum is greater than 22 inches raw water system is in operation , instrument air system is in operation and supplying the condenser water box vacuum breakers
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- PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 14 RADIOLOGICAL CONTROL QUESTION 7.07 (2.00)
According to the " Turbine Driven Feed Pump (P-2C) Operation" procedure, certain conditions should be maintained whenever P-2C is shutdown, a vacuum exists in the Main Condenser, and the P-2C Turbine has NOT been isolate The P-2C (SHOULD/SHOULD NOT) be on the turning gea (0.5) LIST three (3) of the four (4) systems associated with P-2C that (due to this condition) are required to be in continuous operatio (1.5)
QUESTION 7.08 (1.00)
The procedure titled "P-2A and 28 Feed Pump Operation" outlines the action steps to depressurize and drain a MDFW pump. HOW is the actual depressurization of the MDFW pump accomplished? (Do not list steps - like " shut and white tag the pump suction.") (1.0)
QUESTION 7.09 (1.00)
HOW is " ensure proper ECCS valve alignment" determined according to the Emergency Operating Procedures, E-07 (1.0)
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- PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 15 RADIOLOGICAL CONTROL QUESTION 7.10 (2.00)
According to the " Core Reloading" Procedure, core alterations or movement of irradiated fuel within the containment shall cease immediately if any one of several conditions occur. SPECIFY those by correct number designation from the list below that would cause this stop-action to occu (2.0) The containment equipment hatch is open and the reactor has been subcritical for more than 210 hour0.00243 days <br />0.0583 hours <br />3.472222e-4 weeks <br />7.9905e-5 months <br /> . Both personnel hatch doors are closed and the reactor has been subcritical for less than 210 hour0.00243 days <br />0.0583 hours <br />3.472222e-4 weeks <br />7.9905e-5 months <br /> . The containment venting and purge inlet and outlet trip valves are open and a refueling purge is not in progres . The containment low-range area monitor has been inoperable for greater than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and the containment ventilation system isolation is close . One wide-range log channel is operating and providing visual indication in the control room and is generating an audible count rate in the containment, the other wide-range log channel is inoperabl . One boric acid transfer pump is not availabl Comment: There are no conditions for stop-action in the lis A careful rewording of the conditions would make this a good questio QUESTION 7.11 (1.50)
According to A0P 2-25, "High Radiation Levels," several radiation ,
monitors have automatic functions. 'For the following monitors, IDENTIFY their automatic functions, primary component cooling monitor (0.5)
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b .' spent fuel pool monitor (0.5) stack gas monitor (0.5)
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' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 16 RADIOLOGICAL CONTROL QUESTION 7.12 (2.00)
In the Emergency Operating Procedure, " Emergency Shutdown from Power or Safety -Injection," E-0, there is a sequence of IMMEDIATE ACTION steps. The third (3rd) of these steps is " Ensure turbine trip."
LIST the other four (4) of the first five (5) action steps. (Do not specify the details or subparts of each action step and do not specify the action steps required when a " Response Not Obtained" occur (2.0)
QUESTION 7.13 (1.00)
According to Emergency Operating Procedures, E-0, the details of the third (3rd) IMMEDIATE ACTION step (" Ensure turbine trip") is,
"all turbine stop, governor, intercept and reheat stop valves-CLOSED."
According to the " Response Not Obtained" column, WHAT alternate action step should be taken if the third (3rd) IMMEDIATE ACTION cannot be performed? (1.0)
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QUESTION 7.14 (1.00)
Attachment A of ES-0.2, " Natural Circulation Cooldown," provides a list of conditions that support an RCP restart. One of the listed conditions con ~siders the situation in which the RCS WR-T-Colds are less than the minimum pressurization temperature. Considering this condition for RCP restart, the criteria for the primary system is that the Pressurizer level be less than 80%. WHAT is the reason for this criteria? (1.0)
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. PROCEDURES - NORMAL, ABNORMAL, EMERGEACY AND PAGE 17
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RADIOLOGICAL CONTROL QUESTION 7.15 (2.00)
Assume that the Maine Yankee power plant has had a tube rupture in one steam generator and that the reactor has been tripped. The control-room operations staff is following the Emergency Operating Procedure, ES-0.1, " Reactor Trip Response." Under these assumptions, WHAT two (2) conditions in the power-plant would result in ES-0.1 directing you to re-enter E-07 (1.0) If you correctly used ES-0.0, " Event Rediagnosis" to determine the appropriate Emergency Operating Procedure to enter, WHICH procedure would you enter? SPECIFY the procedure numbe (0.5) After the affected steam generator has been isolated, WOULD the tube rupture procedure direct you to implement the procedure on loss of primary or secondary coolant? (0.5)
QUESTION 7.16 (2.00)
According to the " Emergency Shutdown from Power or Safety Injection" Procedure, E-0, an immediate action step is to
" check if RCPs can be run." LIST the steps to be carried out to accomplish thi (1.0) LIST the alternate action step (1.0)
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' ' PROCEDURES -~ NORMAL, ABNORMAL, EMERGENCY AND PAGE 18 RADIOLOGICAL CONTROL
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QUESTION 7.17- '(1.50)
Maximum permissible exposure to external and internal radiation is set forth in the Code of Federal Regulations, " Standards for Pro-tection Against Radiation," (10CFR20). PROVIDE those limits by completing the following tabl , Occupational radiation exposure for individuals 18 years and older will be limited to the following per calendar quarter:
REM Whole body, head and trunk, active blood-forming organs, lens of eyes, or gonads (0,5) Hands and forearms, feet and ankles (0.5) Skin of whole body (0.5)
QUESTION 7.18 (1.50)
An individual in a restricted area may receive a dose to the whole body greater than that permitted above provided that three (3)
criteria are satisfied. OUTLINE those criteri (1.5)
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- ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 19 QUESTION 8.01 (1.00)
SELECT the correct answe According to the Code of Federal Regulations, Part 10 CFR 55, (1.0)
(a.) individuals who direct the licensed activities of licensed operators should be certified (by an authorized representa-tive of the facility licensee) to have received the equiva-lent training of a Senior Reactor Operato (b.) each licensed individual should demonstrate his continued competence every four years in order for his/her license to be renewe (c.) the licensee is limited to performing the furetions of an operator or senior operator at those facilities of similar design as determined by the Commissio (d.) if a licensee has not been actively performing the functions of an operator or senior operator for a period of four months or longer, he/she shall, prior to resuming licensed activities, receive approval of the Commissio .
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' ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 20 QUESTION 8.02 (1.00)
In order to comply with the requirements of 10 CFR 50.36, the following actions shall be taken in the event a Safety Limit as described in Sections 2.2 and 2.3 of the Technical Specifications is violated: (1.0)
(a.) the facility shall be placed in at least HOT SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, operation shall not resume until authorized by the Commission, and a Safety Limit Violation Report shall be submitted within 14 days of the violatio (b.) the facility shall be placed in at least HOT SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, operation shall not resume until authorized by the NSAR Committee and the Manager of Operations, and a Safety Limit Violation Report shall be submitted within 14 days of the violatio (c.) the facility shall be placed in at least HOT SHUTDOWN within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, operation shall not resume until authorized by the Commission, and a Safety Limit Violation Report shall be submitted within 14 days of the violatio (d.) the facility shall be placed in at least HOT SHUTDOWN within I hour, operation shall not resume until authorized by the NSAR Committee and the Manager of Operations, and a Safety Limit Violation Report shall be submitted within 14 days of the violatio .
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. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 21 QUESTION 8.03 (1.00)
When an operator takes a reading of the value of a parameter for the Primary Operators Log and the value is out of specifications /
limits, the operator should (1.0)
(a.) enter the parameter value on the log entry as all other such log entries, prepare a DR report for the PSS/ SOS, and make an entry in the comment remarks section of the lo (b.) circle the parameter value on the log entry, contact the Plant Engineering Department, and have an entry made into the rough log of the PO in the control roo (c.) circle the parameter value on the log entry, report the situation to the PSS/ SOS, and make an entry in the comment remarks section of the lo (d.) underline the parameter value on the log entry, contact the Plant Engineering Department, and have an entry made into the rough log of the P0 in the control roo QUESTION 8.04 (1.00)
Following maintenance to safety classified equipment that cannot be functionally tested under existing plant conditions, the equipment (1.0)
(a.) can be considered provisionally operable and controls established for the Functional-Test Hold Binde (b.) cannot be considered operable until the functional tests are satisfactorily completed and the tags removed according to tagging rule (c.) may be determined functionally acceptable provided the PSS/ SOS is notified by the Responsible Department that conducted the <
' ' corrective action and the proper notation is made in the Deficiency Notification Log Boo ,
(d.) may be determined functionally acceptable provided that the equipment that has received the corrective action does not require a Reactor Coolant System Leak Test (Technical Specification 4.3).
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- ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE' 22 QUESTION 8.05 (1.00)
Consider the following situation as covered in procedure, " Drawing Control," 0-01-2. If a portion of a FE or FM drawing is found to be in error; but, due to the fact- that the power plant is at 100% '
of full power, the true as-built configuration cannot be verified, then the individual who noticed the needed changes shall: (1.0)
(a.) at the start of the next outage (and not before), originate a DCR and the Engineering Engineering Department PED) (Assistance shall be requested Group (EAG) in the to verify the Plant discrepancy and complete the DC (b.) originate a DCR and the EAG in the PED shall stamp all controlled" copies of the affected drawings as "NOT AS-BUILT, SEE DCR .
(c.) contact the PED to ensure that all controlled copies of the
, affected drawings are stamped, " DOCUMENT UNDER REVISION REFER
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TO EDRC " and at the start of the next outage (and not before), originate a DC ,
(d.) originate a DCR but leave the " Description of Change" portion blank and the EAG in the PED shall stamp all controlled copies of the affected drawings as," DOCUMENT UNDER REVISION REFER TO
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EDRC .
QUESTION 8.06 (1.00)
According to the Maine Yankee Emergency Plan the Plant Manager has
! the authority to declare an emergency condition and to initiate emergency measure In his absence, (1.0)
(a.) the Ort-Duty PSS and then the Technical Support. Department Head, IN THAT ORDER, would be designated as his alternat .
(b.) the Technical Support Department Head and the Operations -
Department Head, IN THAT ORDER, would be designated as his
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alternat (c.) the Assistant Plant Manager and then the Operations Department
- Head, IN THAT ORDER, would be designated as his alternat .
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- (d.) the On-Duty PSS and then the Operations Department Head, IN-
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THAT ORDER, would be designated as his alternate.
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' ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 23 QUESTION 8.07 (3.00)
The Maine Yankee Technical Specifications provides safety Limits for DNBR and for LHG DESCRIBE / DEFINE DNB (1.0) WHAT is the Safety Limit for DNBR7 (0,5) HOW does the LHGR Safety Limit change with Core Average Burnup (CAB)? (An example answer that is incorrect is, "The LHGR Safety Limit decreases exponentially with CAB becoming almost zero at E0C.") (1.0) WHAT is the " Basis" for the LHGR Safety Limit? (0.5)
QUESTION 8.08 (3.00)
Included at the end of the examination are several pages from Section 2.1, " Limiting Safety System Setting - Reactor Protection System,"
of Technical Specifications that apply to this question. Also included is the RPS Temperature Program curv Assume that the power plant is operating at 85% of full power and the axial flux tilt is -0.2, CALCULATE the Thermal Margin limit required by this part of Technical Specification (1.0) Under the conditions in part "a.", WHAT should be the limiting safety system setting for Thermal Margin / Low Pressure? (0.5) At this power level of 85%, WHAT is the acceptable range of'
the axial flux tilt according to this part of Technical Specifications? Assume that the power plant is operating in the three loop mode and that the incore detector system and computer are operationa (0.5)
- WHAT is the most limiting value for the axial flux tilt that would be within the limits for the symmetric offset trip function and would make the value of the Thermal Margin / Low Pressure limit the largest? (1.0)
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ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 24 0 QUESTION 8.09 (2.00)
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For each procedure number listed below, DESCRIBE the type of procedure that would correspond to the number
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QUESTION 8.10 (1.50)
. COMPLETE the following statement concerning temporary changes to procedures by filling-in-the-blanks. For blanks "a.", "b." and
"c." CHOOSE from the responses "1." through "10.". (1.5)
Technical Specifications 5.8.3 allows temporary changes to procedures provided that the temporary change does not alter the intent of the i procedure and receives the review and approval of within
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t 14 days of implementation. When a temporary change is needed to a class A or B procedure, the Department Manager, or other responsible person, writes a which is reviewed, approved and signed by and attached to the affected procedur . POTENTIAL RESPONSES
_ t'he PORC
\ the originator and the PORC Secretary the Plant Manager the PORC and the Plant Manager the Discrepancy Report, form no. 0-07-3-1 s a salaried employee and a person holding a SRO license both from the department responsible for the procedure s
- the Vendor Procedure Approval, form no. 0-06-2-5 tt .
5 'e the critjinator and the Plant Manager, provided that the originator holds a SRO license the Procedure Review Acknowledgement, form no. 0-06-2-1
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1 the Procedure Change Report, form no. 0-06-2-3 i
Comment: Item "6." should be rewritten to indicate that the person holding a SR0 license need not be from the responsible department.
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QUESTION 8.11 (1.00)
On the Maine Yankee Primary Operators Logs, it is nchessary to verify that the Spent Fuel Poob (SFP) Manipulator Crane is parked in the proper location, WHAT is the proper location? (0.5) WHY is this position specified in the administrative procedure? (0.5)
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QUESTION 8.12 (1.00)
From the list below, PROVIDE an answer to the two (2) parts of this question. For part "a." and for "b.", select one numbe As directed by the~ Maine Yankee Emergency Plan, WHO is responsible for coordinating offsite radiation monitoring? (0.5)
'- To WHOM should this individual relay the offsite dose monitoring data? (0.5)
'P NRC and DOE MY Executive Vice President Division of Health Engineering lesson officer Division of Health Engineering (DHE) Emergency Coordinator State E0C Offsite Dosimetry Assistants (ODAs)
QUESTION 8.13 (1.00)
LIST those emergency condition (s)/ classification (s), after the declaration of which, the operations centers (EOF, TSC and OSC) are activate (1.0)
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- ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 27 QUESTION 8.14 (2.0C)
Section 3.6 of the Technical Specifications, " Emergency Core Cooling and Containment Spray Systems" addresses the operability requirements for the ECCS. The statements below are taken from this Sectio COMPLETE the statement (2.0)
SPECIFICATION:
The following equipment must be operable whenever the reactor coolant system temperature and pressure exceed deg F and psig: Two safety injection tanks set for automatic initiatio Each tank shall contain 11,200 +/- 500 gallons of water borated to at least 1720 ppm and pressurized with nitrogen to 230 psig +
10 psi, - 25 ps . One operable ECCS train consisting of the following subsystems of the train. Each subsystem includes the manual valves that are aligned and locked in the position required for safeguards operation, the automatically operated valves set for automatic operation or aligned and locked in the position required for safeguards operation, the controls set for automatic initiation where appropriate, and a pump powered from an engineered safeguards bu One high pressure safety injection pump subsystem E, . Station service power in accordance with Technical Specification 3.12.A supplying the same operable ECCS train as in 2. above.
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' ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 28 QUESTION 8.15 (1.50)
For the following three (3) events classify them as an UNUSUAL EVENT, or as a GENERAL EMERGENC Effluent monitors detect levels corresponding to 1 R/hr whole body exposure at the site boundary under actual meteorological condition . (0.5) Primary vent stack monitor radiation level has an unexplained increase by a factor of 1 (0.5) LOCA and HPSI A and B trains inoperable in injection phas (0.5)
QUESTION 8.16 (1.00)
According to Technical Specification, Section 5.12, "High Radiation Area," any individual or group of individuals permitted to enter such areas shall be provided with two (2) types of radiation monitoring devices. DESCRIBE these two (2) device (1.0)
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- ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 29 QUESTION 8.17 (1.00)
Section 3.8 of Technical Specifications provides the specification covering the in-operation and/or operability of the following six (6) plant components for removal of reactor core energy-whenever there is fuel in the reacto . Residual Heat Removal System--Train A Residual Heat Removal System--Train B Steam Generator No. 1 Steam Generator No. 2 Steam Generator No. 3 a minimum of 23 feet of water above the top of the core with the reactor head removed WHAT is the in-operation / operability requirement? Examples of incorrect answers: (1) all components must be either in-operation or operable, (2) at least four (4) must be in-operation and a fifth (5th) must be operabl (1.0)
QUESTION 8.18 (1.00)
According to Technical Specifications, the reactor is critical when
. (1.0)
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- THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 30 THERMODYNAMICS ANSWERS -- MAINE YANKEE -86/10/21-UPTON, ANSWER 5.01 (1.00)
(c.) [+1.0]
REFERENCE Maine Yankee: Reactor Operator Reactor Theory Lesson Plans,
" Prompt and Delay N," R0-L-1.9, pp. S through . Maine Yankee: Reactor Operator Theory Lesson Plans, " Period, SUR, IN Hr.," R0-L-1.10, p. ANSWER 5.02 (2.00)
After the trip the neutron flux level decreases rapidly and this rapidly decreases the "Xe burnout" (neutron absorption in Xe-135).
The primary source of Xe-135 in the core is due to the decay of the fission product I-135. The I-135 concentration level will have been set by the 100% power level operation prior to the trip. After the trip the I-135 concentration will decrease at its radioactive decay rate (which is faster than the radioactive decay rate of Xe-135).
Hence, after the trip the rate of production of Xe-135 will not initially change while the rate of removal of Xe-135 by absorption will decrease immediately. [+2.0]
REFERENCE I
l Maine Yankee: Reactor Operator Reactor Theory Lesson Plans, l " Poisons", R0-L-1.12, pp. 8 - 1 ANSWER 5.03 (1.00)
4. and [+0.5each]
, REFERENCE
, Maine Yankee: Reactor Operator Reactor Theory Lesson Plans,
" Radioactive Decay," R0-L-1.3, pp. 7 and 8.
' Generic: C-E Training Department, " Chemistry," pp. TP-50, 128, and 12 .
' THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 31 THERMODYNAMICS ANSWERS -- MAINE YANKEE -86/10/21-UPTON, ANSWER 5.04 (3.00)
a, lower [+0.5]; CEAs must be lower to add negative reactivity not available from the boron [+0.5] lower [+0.5]; lower Xe concentration must be compensated for
[+0.5] lower [+0.5]; RCS temperature will go down, net positive reactivity must be accounted for [+0.5]
REFERENCE Maine Yankee: Reactor Operator Reactor Theory Lesson Plans,
"ECP/1/M," R0-LP-1.13, pp. 10, 13, 1 ;.
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' THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 32 THERMODYNAMICS ANSWERS -- MAINE YANKEE -86/10/21-UPTON, ANSWER 5.05 (2.00)
1/M(i = CR(o)/CR(1) -[+0.5]
1/M = 75/75 = 1.00 1/M = 75/105 = 0.714
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e [+1.0]
1/M
. s 28 [+0.5]
Ma s
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10 15 20 25 30 35 40 No. of Assemblies REFERENCE Maine Yankee: Reactor Operator Reactor Theory Lesson Plans,
"ECP 1/M," R0-LP-1.13, pp. 7 through ANSWER 5.06 (1.50) ppm [+0.5] ppm [+1.0]
REFERENCE Maine Yankee: Technical Data Book, Figures 1.1.1.1 and 2.6.1.
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~ THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 33 THERMODYNAMICS ANSWERS -- MAINE YANKEE -86/10/21-UPTON, ANSWER 6.07 (3.00) Th would be about 2 deg F hotter so that delta T-power in the primary is the same [+1.0]. This neglects any increase in thermal losse The increased primary temperatures would mean a higher steam and hence a higher steam pressure [+1.0]. The higher S/G temperature and pressure would mean a lower enthalpy steam. Hence the steam flowrate must be greater to generate the same output power [+1.')].
REFERENCE Maine Yankee: Reactor Operator Heat Transfer Lesson Plans, R0-L-7.2, pp. 7, 9, and 1 ANSWER 5.08 (1.50)
The MTC is more negative at EOC than at BOC [+0.5]. This difference increases the severity of the incident at E0C because the incident provides a sudden cooling of the RCS [+0.5] which because of the more negative MTC causes the reactivity of the core to increase more than at BOC [+0.5].
REFERENCE Maine Yankee: Reactor Operator Reactor Theory Lesson Plans, R0-L-1.11, " Nuclear Coefficients / Defects."
t ANSWER 5.09 (1.00)
400 gpm ;+0.5; 120 psi +0.5; REFERENCE Maine Yankee: Reactor Operator Heat Transfer Lesson Plans,
"Thermo Principles," R0-L-7.3.
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' THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 34
. THERMODYNAMICS ANSWERS -- MAINE YANKEE -86/10/21-UPTON, ANSWER 5.10 (3.00) RCS coolant flowrate [+0.5] increase will RAISE DNBR [+0.5] coolant inlet temperature [+0.5] increase will LOWER DNBR [+0.5] j Pressurizer pressure [+0.5] increase will RAISE DNBR ['0.5]
REFERENCE
' Main Yankee: Reactor Operator Heat Transfer Lesson Plans,
"Thermo Principles," R0-L-7.3, pp.14 and 1 ANSWER 5.11 (1.50)
14.5% if HP turbine is ideal if exhaust pressure is 200 psia
[+1.5]
REFERENCE Maine Yankee: Reactor Operator Heat Transfer Lesson Plans,
" Cycles and Efficiency," R0-L-7.2, p.1 ANSWER 5.12 (2.00) increase in incore thennocouple temperatures above T(hot) [+0.5] If the feedwater flowrate to the S/Gs was reduced, then the density difference between the primary coolant in the core and the primary coolant in the S/Gs would decrease and, hence, the primary coolant flowrate would decrease. The rate of decrease in Th would decrease - Th could increase. [+1.5]
REFERENCE Maine Yankee: Reactor Operator Heat Transfer Lesson Plans, R0-L-7,5, pp. 8 to 1 '
' THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 35 THERMODYNAMICS ANSWERS -- MAINE YANKEE -86/10/21-UPTON, ANSWER 5.13 (1.00)
1246 psia ['1.0]
REFERENCE Maine Yankee: Reactor Operator Heat Transfer Lesson Plans, R0-L- ANSWER 5.14 (1.50) delta T = (3.36 x 10**4)(4.25 x 10**-3)/(0.125)(12)
= 95.2 deg F [+1.0] The gas gap is filled with He gas [+0.5]
REFERENCE Maine Yankee: Reactor Operator Heat Transfer and Lesson Plans,
"Thermo Principles," R0-L-7.3, Classroom Problem No.16, p.1.
_ _ . . _ _ . _ _ _ _ _ , _ _ _ _ _ _ _ . ___ _ _ _ . _ _ _ _ _ _ . _ . . , _ . . _ _ , . _ _ _ _ _ , _ . _ . . _ _ _ _ , _ _ _ _ _ _ . _ _ _ _
.
' PLANT SYSTEMS DESIGN, CONTROL AND INSTRUMENTATION PAGE 36 ANSWERS -- MAINE YANKEE -86/10/21-UPTON, ANSWER 6.01 (3.00)
. .
,
'
At/49 4/60 4/fe CP
4/4 Ru.e .R
- -
But.[ gy, 6 gg, y i i . , ,
g
. . . .
,
1 -
. .
.... . . . . . . . . . _ . _ . . .... _ .
Je s 9 Jes /D '. Fas is Ba.c 7 Jas 7 '.Tas M fas n Bas's .rco
[+0.3 for each connecting line]
REFERENCE Maine Yankee: Systems Training Manual, Chapter 33, " Low Voltage Electrical Systems," PGS-18, Figure PGS-18- ANSWER 6.02 (2.00) T5 and SR [+0.5 each] . P-14A charging pump P-14S charging pump (if aligned as an "A") P-29A service water pump P-29C service water pump (one service water pump will trip back open)
[+0.33each,+1.0 maximum]
l REFERENCE
\
! Maine Yankee: Systems Training Manual, Chapter 34, " Diesel Generators," AS-12, pp. 56 and 57.
l
, . . . --- .. . . _ _ _ . - _ - - . - _ _ _ _ . _
r .,
s
" PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 37 ANSWERS -- MAINE YANKEE -86/10/21-UPTON, ANSWER 6.03 (1.50) True '+0. 5'
. False '+0. 5' False l+ REFERENCE Maine Yankee: Systems Training Manual, NS-3, Para. 2.2.4, p through 3 ANSWER 6.04 (1.50) true [+0. 5] false [+0.5] true [+0.5]
REFERENCE Maine Yankee: Systems Training Manual, NS-4, Para. 3.1.3, p. 8 ANSWER 6.05 (3.50) temperature control valves and 2 pressure control valves [+1.0]
OR 7 temperature controlled valves and 5 pressure controlled valves [+1.0] .5% [+0.5] when either one (1) reaches the fully open position [+0.5] condenser vacuum less than 20 in Hg steam header pressure less than 775 psig [i0.5:k+0.5] % to 84% [+0.5]
REFERENCE Mai:,e Yankee: Systems Training 5 Manual, Chapter 25, " Steam Dump and Turbine Bypass System," PGS-14, pp. 2, 4, and i
~
. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 38 ANSWERS -- MAINE YANKEE -86/10/21-UPTON, ANSWER 6.06 (1.50) /2% or 280,000 lbm/hr SIAS [+0.5] and CIS [+0.5][+0.5]
REFERENCE Maine Yankee: Systems Training Manual, Chapter 22, " Main Steam
& Reheat Steam System," PGS-22, Para. 2.1, p. 1 ANSWER 6.07 (1.50)
The RAMP pushbutton controls the timed opening of the reheater temperature control valves [+0.5]. The button is pressed when the turbine reaches 35% load [+0.5], starting the opening of reheater temperature control valves, which limits the maximum heatup rate to less than 100 deg/hr [+0.S].
REFERENCE Maine Yankee: Systems Training Manual, Chapter 22, " Steam and Reheat Steam," Para. 3.2.3.1, p. 2 ANSWER 6.08 (3.00) The controller output is directed to either of the two flow control valves. The two flow control valves are aligned in parallel [+0.5]; each with a flow orifice, 30 gpm and 120 gpm
[+0.5]. The control valve (or range) selector directs the control signal to the chosen valve [+0.5]. A setpoint is adjusted from 0 to 100% flow of either valve [+0.5]. A high radiation alarm trips the solenoid valves on the two flow control valves interrupting the air supply to the valves thereby closing them [+1.0].
REFERENCE Maine Yankee: Systems Training Manual, Chapter 38, " Liquid Waste Disposal System," AS-14, pp. 23 and 2 .. .
' PLANT SYSTEMS DESIGN, CONTROL,'AND INSTRUMENTATION PAGE 39'
ANSWERS -- MAINE YANKEE -86/10/21-UPTON, l ANSWER 6.09 (1.50) Protect against (prevent) excessive heat rencval caused by a steam line break [+0.5]. To prevent possible fuel / cladding damage OR to stay within DNBR Safety Limits - [+0.5] . To prevent power peaking leading to uneven core burnup and/or local fuel / cladding damage OR to stay within LHGR Safety Limits [+0.5] .
REFERENCE Maine Yankee: Systems Training Manual, Chapter 11, "RPS,"
NS-12, Para. 1.2.1, p. ANSWER 6.10 (1.00) A negative tilt means core power in the lower half is greater than in the upper half of the core. [+0.5] It is determined from the ratio of the output signals of the UIC upper (U) and lower (L) power safety channels as U-L
[+0.5]
U+L REFERENCE Maine Yankee: Systems Training Manual, Chapter 11, "RPS,"
NS-12, par 2.1.10, p. 24.
i I
i
!
,
!
l
,
.. _ .. _ _ _ _ _ _ _ _ _ . _ _ . _ . _ _ . . _ . . _ . _ l
.
'
. PLANT SYSTEMS DESIGN, CONTROL, AND IN3TRUMENTATION PAGE 40 ANSWERS -- MAINE YANKEE -86/10/21-UPTON, ANSWER 6.11 (3.00). % of full flow psig .6 dpm .6 (setpoint) OR 106.5% (Tech Specs) psig auto stop oil pressure OR 4 of 4 turbine stop valves closed .25 psig (setpoint) OR 5 psig (Tech Specs)
[+0.5 each]
REFERENCE Maine Yankee: Training Systems Manual, " Reactor Protection System," NS-12, p. 3 ANSWER 6.12 (1.00)
6 (reed switch deviation) [+0. 5]
5(pulsedeviation))OR[+0.5]
10 (pulse deviation REFERENCE Maine Yankee: Systems Training Manual, Chapter 12, "CEA Control System," NS-9, Para. 3.3, pp. 48 and 4 _-.
t
' PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION
_
PAGE 41 ANSWERS -- MAINE YANKEE -86/10/21-UPTON, ANSWER 6.13 (1.00) charging pump suction VCT RWST RHR spent fuel pool
[+0.25 each, +1.0 maximum]
REFERENCE Maine Yankee: Systems Training Manual, Chapter 4, "CVCS",
NS-4, Para. 2.1.18, p. 4 . . Maine Yankee: Systems Training Manual, Chapter 4, "CVCS,"
.
NS-4, Para. 1.2.4, p. 13.
!
!
l i
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- PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 42 RADIOLOGICAL CONTROL ANSWERS -- MAINE YANKEE -86/10/21-UPTON, ANSWER 7.01 .(1.00)
(d.) [+1.0]
REFERENCE Maine Yankee: Procedures, "Short-Form Calorimetric," Proc. N , Rev. 6, p. ANSWER 7.02 (1.00) Check WR power channels - LESS THAN 2% Check start-up rate channels - ZERO OR NEGATIVE
[+0.5 each]
REFERENCE Maine Yankee: Function Restoration Procedures, " Nuclear Power Generation /ATWS," FR-S.1, pp. I and ANSWER 7.03 (1.00) manually trip CEDM generator set output breakers manually open and reclose the 480 V supply breakers to buses 9 and 12 if all trippable CEAs are NOT inserted, then go to FR- [+0.5 each, +1.0 maximum]
REFERENCE Maine Yankee: Emergency Operating Procedures, " Emergency Shutdown from Power or Safety Injection," E-0, Rev. O, p. 3.
l
,
.
~
- PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 43 RADIOLOGICAL CONTROL ANSWERS -- MAINE YANKEE -86/10/21-UPTON, ANSWER 7.04 (.50)
I&C supervisor [+0.5]
REFERENCE Maine Yankee: Procedures, " Daily RMS Operational Test," Pro No. 3-6.2.2.17, Rev. 12, p. ANSWER 7.05 (1.50)
l shutdown and isolate the system [+0.5] the mode switches may be switched to ON-LINE (for up to 1/2 hour) [+0.5] only during refueling operation or refueling shutdown [+0.5]
REFERENCE Maine Yankee: Procedures, " Daily RMS Operational Test," Pro No. 3-6.2.2.17, Rev. 12, pp. 1, 2, and ANSWER 7.06 (1.50)
3., 5., and [+0.5 each]
REFERENCE Maine Yankee: Secondary Operating Procedures, " Vacuum Priming System," Proc. No. 1-100-1, Rev. 7, p. _ - . .
'
.
' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 44 RADIOLOGICAL CONTROL AN3WERS -- MAINE YANKEE -86/10/21-UPTON, ANSWER 7.07 (2.00) should [+0.5] . gland sealing steam gland exhaust vapor extractor [+0.5 each, +1.5 maximum]
REFERENCE Maine Yankee: Secondary Operating Procedures, " Turbine Driven Feed Pump (P-2C) Operation," Proc. No. 1-104-5, Rev. 8, p. ANSWER 7.08 (1.00)
take manual control of the recirc. valve and slowly throttle it open [+1.0]
REFERENCE Maine Yankee: Secondary Operating Procedures, "P-2A and P-2B Feed Pump Operation," Proc. No. 1-104-1, Rev. 15, pp. 10 and 1 ANSWER 7.09 (1.00)
Verify on the SIAS light box indicators that all lights are green and not flashing. [+1.0]
REFERENCE Maine Yankee: Emergency Operating Procedures, "Emer;ency Shutdown from Power or Safety Injection," E-0, Rev. O, p. 5.
I
&
.
- PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 45 RADIOLOGICAL CONTROL
,
ANSWERS -- MAINE. YANKEE -86/10/21-UPTON, ANSWER 7.10 (2.00)
1., 3., 5., and [+0.5 each for 1. , minimum,+2.0 maximum)S.,and6.;-0.5eachfor2.and4.;+ REFERENCE Maine Yankee: Procedures, " Core Reloading," Proc. No. 10-1, Rev. 9, pp. 4 and . ANSWER 7.11 (1.50) shuts the PCC surge tank vent [+0.5]
, stops upward movement of the new fuel elevator [+0.5] stops PAB exhaust fan or blocks it from starting [+0.25] and during an on-line purge shuts VP-A-1 through 5 and trips HV-9
[+0.25]
REFERENCE Maine Yankee: Abnormal Operating Procedures, "High Radiation Levels," Proc. No. A0P-2-25, Rev. 12, p. . Maine Yankee: Systems Training Manual, " Radiation Monitoring,"
AS-18, pp. 35 and 3 ANSWER 7.12 (2.00) manually trip the reactor and turbine ensure reactor trip ensure power to AC emergency buses 4 check if SIAS initiated
[+0.5each]
REFERENCE Maine Yankec: Emergency Operating Procedures, " Emergency Shutdown j from Power or Safety Injection," E-0, Rev. O, pp. 3 and 4 i
I
... . . _ _ _ _ _ - , _ _ _ - _ - _ _ , _ . , _ _ _ _ . - _ - - - - - _ _ , - - - . _ , . . _ _ _ . , . _ _ _ _ _ _ - - , _ _ _ _ - , -
, - _ . - . _ - . -
.
'
. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 46 RADIOLOGICAL CONTROL ANSWERS -- MAINE YANKEE -86/10/21-UPTON, '
ANSWER 7.13 (1.00)
manually stop the EHC HP fluid pumps (P-55A, P-55B) [+0.5]
if any turbine stop valve and its associated governor valve remains open, then close the EFCVs [+0.5]
REFERENCE Maine Yankee: Emergency Operating Procedures, " Emergency Shutdown from Power or Safety Injection," E-0, Rev. O, p. ANSWER 7.14 (1.00)
If the RCS temperature in the cold legs are less than required for NPSH, starting an RCP could produce a void in the RC The void would cause a swell in the primary. The criteria should prevent the Pressurizer from going solid upon starting an RCP, To prevent reverse heat transfer from the S/Gs to the RCS causing heatup and potential system overpressure. [+1.0]
REFERENCE Maine Yankee: Emergency Operating Procedures, " Natural Circulation Cooldown," ES-0.2, Rev. O, Attachment A, p. . Maine Yankee: Technical Specifications, Section . . - . _ . _ - . -
. _ ..
,
' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 47 RADIOLOGICAL CONTROL ANSWERS -- MAINE YANKEE -86/10/21-UPTON, ANSWER 7.15 (2.00) low PZR level ((5%)
-low PZR pressure ((1585 psig)
low core subcooling ((12 deg F)
,
[+0.33each] E-3 [+0.5]
'
, No [+0.5] . (This would recur after a steam-line rupture event and the Steam Generator had been isolated.)
i REFERENCE
, e Maine Yankee: Emergency Operating Procedures, " Emergency Shutdown from Power or Safety Injection," E-0, Rev. O, p. . Maine Yankee: Emergency Operating Procedures, " Event Rediagnosis," ES-0.0, Rev. O, p. . Maine Yankee: Emergency Operating Procedures, " Loss of Primary or Secondary Coolant," E-1, Rev. O, pp.1 and 2.
,
ANSWER 7.16 (2.00) .. Check core region subcooling - greater than or equal to 25 deg F Check CIS "A" and "B" trip signal annunciators - not lit I
[+0.5each] . If at least one charging /HPSI pump running, then stop all RCPs If CIS actuated, then s op all RCPs f
[+0.5each]
REFERENCE Maine Yankee: Secondary Operating Procedures, "P-2A and P-28 Feed Pump Operation," Proc. No. 1-104-1, Rev. 15, p. 5.
_ _ _ .- _ _ . . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ , . - . , - . . . _ _ - _ . . . . _ _ _ _ . . _ . - . _ . __. . - - . _ _ , _,
.
~ PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 48 RADIOLOGICAL CONTROL ANSWERS -- MAINE YANKEE -86/10/21-UPTON, ANSWER 7.17 (1.50) .25 rem OR 3 rem with an updated NRC. Form-4 and not to >
exceed 5(N-18) rem total .75 .50
[+0.5each]
REFERENCE Maine Yankee: Radiation Protection Manual, Section 2.3, p. 2- ANSWER 7.18 (1.50) The dose to the whole body shall not exceed 3 rem / quarte . The dose to the whole body, when added to the accumulated occupational dose to the whole body, shall not exceed 5(N-18). The individual's accumulated occupational dose to the whole body has been determined and recorded on Form NRC- . 75 rem and 25 rem for life saving and emergency actions
[+0.5 each, +1.5 maximum]
REFERENCE Maine Yankee: Radiation Protection Manual, Section 2.3, pp. 2-2 and 2- .
' ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 49 ANSWERS -- MAINE YANKEE' -86/10/21-UPTON, ANSWER 8.01 (1.00)
(d.) [+1.0]
REFERENCE _ Generic: Code of Federal Regulations, Part 10 CFR 55.31,
" Conditions of the Licenses," pp. 559 and 56 . Generic: Code of Federal Regulations, Part 10 CFR 55, Appendix A, p. 56 ANSWER 8.02 (1.00)
(c.) [+1.0]
-
REFERENCE Maine Yankee: Technical Specifications, Section 2.0, " Safety Limit Violation," p. 2.0- ANSWER 8.03 (1.00)
(c.) [+1.0]
REFERENCE Maine Yankee: Administrative Procedures, " Conduct of Operations,"
1-200-10, Maine Yankee Primary Operators Logs, MY-72-84, Rev. 9, p. . Maine Yankee: Administrative Procedures, " Conduct of Operations,"
1-200-10, Rev. O, p. 2 ANSWER 8.04 (1.00)
(a.) [+1.0]
REFERENCE Maine Yankee: Administrative Procedures, " Discrepancy Reporting Procedures," 0-07-3, Rev. 3, p. 9 through 1 .
' ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 50 ANSWERS -- MAINE YANKEE -86/10/21-UPTON, ANSWER 8.05 (1.00)
(b.) [+1.0]
REFERENCE Maine Yankee: Administrative Procedures, " Drawing Control,"
0-01-2, Rev. 4, pp. 2 through ANSWER 8.06 (1.00)
(c.) [+1.0]
REFERENCE Maine Yankee: Emergency Plan, Section 5.0, " Organizational Control of Emergencies," p. ANSWER 8.07 (3.00) DNBR is the acronym for Departure from Nucleate Boiling Rati It is the ratio of (1) the predicted heat flux (at a particular core location) that would correspond to the upper boundary of the nucleate boiling regime to (2) the actual heat flux (at thatlocation). [v1.0] .20 [+0.5] The LHGR Safety Limit decreases linearly with CAB with less than a 1% change for BOC to E0C. [+1.0]
" Operation with a peak linear heat rate below the Safety Limit on LHGR should insure that there is no fuel melting (which would occur initially along the fuel pellet centerline).
[+0.5]
REFERENCE Maine Yankee: Technical Specifications, Section 2.2, " Safety Limits - Reactor Core," pp. 2.2-1 and 2.2- ~
.
. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 51 ANSWERS -- MAINE YANKEE -86/10/21-UPTON, ANSWER 8.08 (3.00) with Y(1) = -0.2, A(1) = 1.06 with P = 85%, QR(1) = 0.94 and T(C) = 549 deg F Thermal Margin limit = 2025(1.06*0.94) + 17.9(549) - 1005 = 1782 [+1.0] versus 1835 psig means that the limit is 1835 psig [+0.5] using]
[+ Figure 2.1-2, the acceptable range is (-0.3 to +0.28) using Figure 2.1-la, Figure 2.1-2, and the Thermal Margin limit equation, an axial tilt of +0.28 is the answer [+1.0]
REFERENCE Maine Yankee: Technical Specifications, Section 2.1, " Limiting Safety System Setting - Reactor Protection System," p. 2.1-1, Figures 2.1-la, 2.1-lb, and 2.1- . Maine Yankee: Technical Data Book, "RCS Temperature Program,"
Figure 1. ANSWER 8.09 (2.00) temporary procedure, class A necial test procedure, class A (nuclear safety related) special test procedure, class B (nonnuclear safety related) equipment surveillance procedure, class D
[+0.5each]
REFERENCE Maine Yankee: Administrative Procedures, " Procedure Preparation, Classification and Format," 0-06-1, Rev. 7, pp. 2, 4 and ' Maine Yankee: Administrative Procedures, " Procedure Review, Approval, Distribution and Adherence," 0-06-2, Rev. 8, p. '
.
[i
' ' ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 52 3 ANSWERS -- MAINE YANKEE -86/10/21-UPTON, .
ANSWER 8.10 (1.50)
a. - b. - 1 c. - [+0.5 each] <
REFERENCE
- Maine Yankee: Administrative Procedures, " Procedure Review, Approval, Distribution, and Adherence," 0-06-2, Rev. 8, p. t Maine Yankee: Technical Specifications, Section 5. '
I
'
's -
'
,- , J?
,
ANSWER 8.11 (1.00) north end of pool, between the marks OR over the newly off 1 loaded spent fuel, between the marks [+0.5] ,
\ \ for earthquake protection (block-wall collapse) [+0.5]
REFERENCE Maine Yankee: Administrative Procedures, " Conduct of Operations,"
1-200-10. Maine Yankee Primary Operators Logs, MY-72-84, Rev. 9, pp. 7 and ANSWER 8.12 (1.00) 3
' . . . , (
[+0.5each] (
REFERENCE } Maine Yankee: Emergency Plan, Section 5.0, " Organizational i
'
',
Control of Emergencies," p. 5.1 i
.
%
d Y
_ _ . _ a _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _
. _ _ _ _ _ _ _ . . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ , . . . _ _ _ _ . _ _ . _ _ _ , _ _
, .
, ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 53
'
' ANSWERS -- MAINE YANKEE -86/10/21-UPTON, '
.
s
.
ANSWER 8.13 (1.00)
b alert
. site emergency general emergency
[+0.33 each]
REFERENCE g Maine Yankee: Emergency Plan, Section 6,.0, " Emergency Measures,"
\.
y p. 6.2 and Figure 6-5.
ANS ER 8.14 (2.00) one service water pump subsystem
\ one component cooling pump subsystem
, one lo r pressure safety injection pump subsystem one containment spray pump and RHR heat exchanger subsyster
'[+0.33 each}
'
REFERENCE'Is Maine Yankee: Tecnnical Specifications, Section 3.6, " Emergency Core Cooling and Containment Spray System," p. 3.6-1.
! ANSWER s8.15 (1.50)
'
'a . general emergency b. ' unusual event c'. ,deneral emergency
'
'
[+ 'each]
REFERENCE , Maine Yankee: Implementing Procedures to the Emergency Plan, Figure 2.50.0-1, " Emergency Condition Classification Table."
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_ _ _ _ . - . . . - . . . _ . . . - - - - _ _ _ .- __
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'
' ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 54 ANSWERS -- MAINE YANKEE -86/10/21-UPTON, ANSWER 8.16 (1.00) a radiation monitorin device which continuously indicates the radiation dose rate +0.5] a radiation monitoring device which continuously integrates the radiation dose rate and alarms at a present preset integrated dose [+0. 5] a radiation dose rate monitoring device in the hands of a radiological controls qualified individual [+0. 5]
[+1.0 maximum]
REFERENCE Maine Yankee: Technical Specifications, Section 5.12, "High Radiation Area," p. 5.12- ANSWER 8.17 (1.00)
at least one in-operation with a 2nd operable [+1.0]
REFERENCE Maine Yankee: Technical Specifications, Section 3.8, p. 3.8- ANSWER 8.18 (1.00)
wide range log-N channels read 10**-4 power [+1.0] -
REFERENCE Maine Yankee: Technical Specification Definitions, Procedure 1-2, p. . , _ - _ , _ . - - , _-__ - -_ . _ . . _ _ ..
- - _ _ _ - - _ _ - _ . ___-_-__ ___
p -
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CRillCAL DORON CONCEtHRA110N AT llFP VS. BURNUP CONDillONS: CORE 9, ARO, NORMAUZED TO MEASURED VALUES
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____________________________________________________________________________
EQUATION FORMULA AND PARAMETER SHEET
____________________________ .....________.. __ . __________________________
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Where mg = m2 (density)3(velocity)3(area)1 = (density)2(velocity)2(area)2
____________________________________________________________________________
__________________________________________________________________________ h=hAaT h=kAST/x
. . .
. .
Q = m:p(T out -Tin) Q = UA (T,y,-Tstm) Q = m(hi-h2)
____ ... ___________________....____________________________________________
P = P010(SUR)(t) P = P0 et /T SUR = 26.06 I T= Y - #) m = (3 _ o)
p p Ae,.
<,
_____________________________________________________________________. ____
delta K = (K,ff-1) CR y (1-K,ffy) = G2 (I-Keff2)
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M = (2-K ,,9)
.
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.. _________________________________________________________________________
decay con ant = In (2) = 0.693 -
- A1 = Ao e (decay constant)x(t)
'1/2 1/2
_____.. ____________________________________________________________________ Y Water Parameters Miscellaneous Conversions
.
"
I gallon = 8.345 lbs 1 Curie = 3.7 x 10 10 cps 1 gallon = 3.78 liters 1 kg = 2.21 lbs s
1 ft = 7.48 gallons 3 I hp = 2.54 x 10 Btu /hr Density = 62.4 lbm/ft 3 6 1 MW = 3.41 x 10 Btu /hr Density = 1 gm/cm# 1 Btu = 778 ft-lbf Heat of Vaporization = 970 Stu/lbm Degrees F = (1.8 x Degrees C) - 22 heat of Fusion = 144 Stu/lbm I inch = 2.54 centice ers, 1 Atm = 14.7 sia = 29.9 in H; ; = 32.174 f t-1:r 'i:f-s e:-
1 ft H 2O = 0.1325 1:f/i __--____-______________ __________.._______________________________________
____m
'tW L krTM6brn1rnf A FACILITY COMMENTS / RESOLUTIONS
^ Question 1.01 Facility Comment: Question asks for a response in terms of " reactor period." Although this unit is widely known and used in some plants, it is not measured or discussed in any Maine Yankee Operational Document An answer discussing response in terms of startup rate should be accepte It is again requested that the NRC refrain from use of this term in exams given at Maine Yanke Resolution: Maine Yankee Training Lesson, MT-T-65, Rev. 1, Lesson Number R0-L-1.10 is entitled, " Period, SUR, In-Hour." This lesson plan references NUS Manual Module 3 in which the title to Section 6.2 is
" Pe ri od . " In the answer key add to the reference list the following: Maine Yankee: Reactor Operator Reactor Theory and Lesson Plans,
" Period, SUR, In-Hour," R0-L-1.10, pp. 2 through Question 1.03 Facility Comment: This question refers to a calculation of SDM which is not performed at Maine Yankee. SDM is defined in Technical Specifications but no formula for calculation exists. All references to ensuring sufficient SDM exists refer the operation to approved curves in the TD The candidates will more than likely list many _of the factors acceptable by your answer key but may break down the power defect into its associated components (moderator, fuel, void, pressure defects) which should also be acceptable. It is again requested that the NRC refrain from reference to this calculation not used at Maine Yankee in future examinations. Your reference of R0-L-18 probably is incorrect as it does not discuss SD Resolution: Maine Yankee Training Lesson, MT-T-65, Rev. 1, Lesson Number R0-L-1.8, R0-L-1.9, R0-L-1.10, etc. all list on page 1 that "the trainee will be able to perform the following evolutions;--- b) Shutdown margin calculation." Change the answer key to read:
2. power defect (moderator, fuel, void, pressure defects =+0.2 f f listed separately)
Reference 1. Maine Yankee: Reactor Operator Reactor and Lesson Plans, " Flux Distribution," R0-L-1.8, p. Maine Yankee 1
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.
Question 1.04a Facility Comment: Maine Yankee may have a positive MTL at B0C. If the candidate assumes a positive MTC, the POAH is of concern because it will cause an addition of positive reactivity and therefore a larger SU See Lesson Plans R0-L-1.7 and R0-L-1.1 Resolution: In the question and in the answer key change P0H to POA Add to the answer key the following: At BOC, the reactor may have a positive MTC [+0.3]. If the MTC is positive, the PGAH is of concern because it would represent the point at which additional reactivity is added to the core; which, in turn, would increase the SUR [+0.5].
Change the reference to:
1. Maine Yankee: ......, pp. 19 and 20, and Viewgraph 13.6- Question 1.04b Facility Comment: Steam pressure will probably not increase since the turbine bypass system will maintain pressure constant if in AUTO (normal lineup) - Systems Training Manual, Vol. 5, Ch. 2 An increase in PZR level and pressure (with appropriate CVCS response, i.e., increase in letdown / decrease in changing) may be the operators best indications depending on rate of power rise to the P0A Systems Training Manual, Vol. 1, Ch. 4.
'
Resolution: The Reactor Operator Reactor Theory and Lesson Plans, R0-L-1.11 should be corrected. Change the answer key to read:
T-ave increases [+0.3]
SUR decreases [+0.3]
PZR level and pressure increases (with CVCS response) [+0.3]
Turbine bypass valves open further to keep S/G pressure from increasing
[+ 0.3]
- [+0.9 maximum]
l l
l l
l Maine Yankee 2 l
_
t. Question 1.06 Facility Comment: The question tells the candidate that the scenario begins with "all controllers in AUTO." This statement is confusing in that it could lead the candidate to assume conditions which are not normal for 70% powe I assume that the examiner means that system alignments are normal for this power leve If so, I believe the answer to be incorrec If the RCS flowrate is decreased while holding steam demand constant (which will be true with normal system lineups), then the RCS delta T (Th-Tc)willincrease(ThT,Tc}). T-ave as indicated to the operator will not change from steady state to steady state. If T-ave remains the same, there would be little or no change in steam generator temperature or pressure. If steam generator pressure does not change, then steam flow will not chang Regardless of what happens to T-ave, pressurizer level will remain the same since the level controller will be set to maintain 50% level at 70%
power. Letdown and charging flow rates will fluctuate to maintain this leve Systems Training Manual, Vol. 1, Ch. 3/ Other answers should be accepted depending on the candicates assumption of
"all controllers in AUTO."
Resolution: The answer to "a.", "b.", and "c." will be treated as a single question. If the candidate answers that T-ave would increase, then the answers given in the answer key would be correct and worth [+1.5]. If the candidate answers that T-ave does not change, an answer that the S/G conditions would not change would be correct and worth [+1.5]. The answer to "d." will be changed to "no change in PZR level if the setpoint is in MANUAL or PZR level would increase if setpoint is auto-controlled by the RRS [+0.5] . "
Maine Yankee 3
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Question 1.07b Facility Comment: Maine Yankee does not use the reactivity unit of pcm in any operational document. Assuming the students uses reactivity in terms of % delta rho:
K -K 2 1 1.004 - rho = ------- = ---------- = 0.4% delta rho KK 1.004
Resolution: During the examination it was announced that the answer could be given in any units. Hence, the answer key will read: b. 398 pcm or 0.398% delta rh Question 1.07c Facility Comment: Same comment as abov K -K 2 1 1.004 - 0.92 rho = ------- ------------- = 0.4% delta rho KK (1.004)(0.92)
It is requested that the NRC refrain from using reactivity terms other than those used in our operational documents in future examination See Lesson Plan R0-L-1.1 Resolution: During the examination it was announced that the answer could be giver in any units. Hence, the answer key will read: c. 9090 pcm or 9.09% delta rh Answer 2.02c Facility Comment: Answer Key has correct answer listed under (d.)
Answer should read "inside the neutron shield tank against the inner l wall."
Systems Training Manual, " Containment Building," Ch. PG-1 Pesolution: Interchange "c." and "d." in the answer ke Maine Yankee 4
._ .-__
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Question 2.04 Facility Comment: Maine Yankee has only 3 SIT's not The questicn refers to operational " MODES." Maine Yankee does not refer to operating conditions as MODES. We assume you mean " Power Operation Condition" or Condition This question has ne correct response since our SIT isolation MOV's are disabled OPEN .
It is again requested that the NRC refrain from referring to operational
" MODES" in examinations given at Maine Y?nke Resolution: Delete Question 2.0 Question 2.06 Facility Comment: N2 is normally isolated to containment and any overpressure will be relieved to the hydrogenated vent header via PR-A-45 (quench tank vent valve).
Systems Training Manual, "PZR and Press Relief," Ch. PG-2 Resolution: Change the wording in the answer key to read:
PRESSURE [+0.25] (The quench tank is pressurized as needed by the addition of N2 gas to maintain 2 psig.) The tank will automatically vent on increasing pressure (beyond 18 psig) to the hydrogenated vent header
[+0.75].
- Modify the reference to include pp. 22 through 26.
! Answer 2.07a Facility Comment: Correct answer should be to avoid thrust shift or thrust reversal conditions causing excessive shaft movement and seal damage (movement is vertical rather than radial).
OP-1-1 Step 7.0 and Systems Training Manual, "RCS," Ch. PG-6 Resolution: Change the answer to read: to avoid thrust shift reversal conditions causing excessive shaft movement and seal damage [+1.0].
Maine Yankee 5
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Question 2.09b Facility Comment: Two charging (HPSI) pumps are also required whenever boron concen!. ration is less than the Hot Shutdown Boron Concentration (Tech Specs 3.10). Even though technical specifications require 2 at power (72%), two may be required at lower power level Resolution: Change the answer to read: reactor power > 2% [+1.0]
boron concentration ( hot shutdown boron concentration [+0.5]
[+1.0 maximum]
Add a second reference: Maine Yankee: Technical Specifications, Section 3.1 Answer 2.11 Facility Comment: The answer key is incorrect. From reviewing the listed reference, it is apparent that the examiner has listed design valves and not " normal operating parameters."
Same reference Pg. 32 gives normal operating pressure as 10-20 psi Control room logs give normal band as 15-25 psig (Surveillance Pro .1.1).
Control room logs give normal band as90-110 deg (Surveillance Proc. 3.1.1).
"VCT Spray" is a term not used at Maine Yankee. It is assumed that the normal valve you request is that for letdown flow (Surveillance Pro .1.1) (logs) lists70-150 gpm as normal band.
i Resolution: Section 5.2, " Operator Monitoring Requirements," lists,
'
starting on page 98, nominal values as 100% of full power values. Letdown
_
flowrate is listed as 80 gpm, VCT temperature as 115 deg F, and VCT pressure as 13 psig. In the middle of page 99, Section 5.3, " Design Parameters" starts. On page 102 under VCT are listed normal operating pressure at 50 psig, normal operating temperature at 120 deg F, and normal spray flow at 80 gpm. As for pressure, neither 50 psig or 13 psig correspond to the surveillance range of 15 to 25 psig. As for temperature neither 120 or 115 deg F correspond to the surveillance range of 90 to 110 deg F. 80 gpm as the normal or nominal 100% power flowrate is consistent and consistent with the surveillance range of 70-150 gp The answer key will read as:
Any value between a. 90 to 110 deg F
,
b. 15 to 25 psig
! c. 70 to 150 gpm
[+0.5 each]
Maine Yankee 6 I
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.___a
. :.
The references will be modified to read:
1. Maine Yankee: Systems Training Manual, NS-4, Paragraph 5.2 and 5.3, p through 10 . Maine Yankee: Surveillance Procedure, 3. Answer 2.12a Facility Comment: This can be considered FALSE since the quick open override signal is common to bot Systems Training Manual, " Steam Dump and Turbine Bypass," p. Resolution: On page 1 of the reference, it is stated that each group of valves is operated by a separate control system. This statement requires TRUE for the answer. If the commonality of the quick open override signal can imply an answer of FALSE, then either answer would have to be considered correct. Question 2.12a is delete Answer 2.12c Facility Comment: The answer key is incorrect. The correct answer should be FALS If the RRS fails as is at 100% power, the signal to provide a quick open would be " locked in" therefore the system would operate upon a turbine trip to prevent safety valves from lifting. The initiation of a turbine trip which activates the system has nothing to do with RR Resolution: The answer key is amended to read, " FALSE".
!
Question 3.02c Facility Comment: This question has many potential answers depending upon initial plant conditions and subsequent pla ;t response The listed answer is true only if the switch was placed in block -
I pressure was then reduced and then raised to enable an auto unblock signal. Although this feature is available, the question does not clearly elicit the response required by the ke If the switch were placed in block as is normally done when allowed by l procedure and other normal pressure control procedures are used then no logic exists which will generate a CI Resolution: This part of the question does not ask for operational sequences, but what control logic could generate a CIS. Figure NS-5-12,
"CIS Block Diagram," supports the answer. No change to the question or to the answer key. Add Figure NS-5-12 to the reference page Maine Yankee 7 l
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Answer 3.03b Facility Comment: The answer key is incorrect. Each SMM (Rx Head and Core) have selectable temperature inputs which are normally selected to different input The Head from Th RTD - The Core from an In-Core Thermocoupl Resolution: The answer key is changed to FALS Answer 3.03c Facility Comment: The " temperature" input to the steam generator SMM is calculated from SG Press, the SG Press input is selectable from a switch on the MCB, it is therefore reasonable to answer TRUE to this questio If taken literally, it is reasonable to answer fals Question should be removed from gradin Resolution: Question 3.03c is delete Question ~:.05 Facility Co.' ment: Many more than 4 outputs from the plant computer exist which utilize "incore instrument signals." These include the failed detector log, rhodium alarms and others listed in the Systems Training
, Manual for Incore Instruments, pp.16 through 2 Resolution: Pages 16 through 21 does not refer to failed-detector logs or rhodium alarms. The question asks for 4 outputs instead of 4 printout Three (3) uniquely different outputs are acceptable response Answer 3.07 Facility Comment: 1) V0P input to CWP is not inhibited (10**-4%; 2) SUR input to CWP is inhibited (10**-4%.
Systems Training Manual, "CEDS," p. 15 and "RPS," p. 3 Resolution: The answer is corrected to move " disabled below 10**-4% power
[+0.25]" from "1. " to "2. "
Maine Yankee 8
!
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Answer 3.08a Facility Comment: The RRS generates the following " temperature related" output signals 1. Tave - Tref alarm signals (Hi and Lo)
2. PZR level RRS setpoint signal 3. Tave indication 4. Tref indication 5. variable steam dump demand signal (Tave - 547 def F)
6. quick open override signal (on if Tave >565 deg F) -
Resolution: The answer key is amended to read:
1. Tave - Tref alarm signals (HI and LO)
2. PZR level setpoint 3. Tave indication (MCB and 2 per recorder)
4. Tref indication (2 per recorder) Steam-durp demand signal Quick-open override si
[+0.5 each, +1.5 maximum] gnal The reference is amended to include Figure NS-9-2 as a pag Question 3.08b Facility Comment: Part b of this question involves " double jeopardy" in that if only 2 correct responses to A are listed than only 1.0 pts are possible for Overall - the following systems are available for RRS input from
" temperature related" output signal . steam dump system 2. turbine bypass system 3. annunciator system 4. plant computer system 5. PZR level control system Resolution: With six answers for 3.08a, there is no concern about double jeopardy. The answer key is amended to read:
1. steam-dump system 2. turbine bypass system 3. annunciator system 4. plant computer system 5. PZR level control system
[+0.5 each, +1.5 maximum]
Maine Yankee 9
. . _ . . _ . - . _ - .
r
-
Answer 3.11 \
Facility Comment: A hand plot of symmetric offset versus time is also maintained in the control roo t Resolution: The answer key is amended to read:
1. plot of symmetric-offset values [+0.5D from the plant computer [+0.5]
2. hand plot of symmetric-offset values [+0.5] with values from the plant computer or the RPS [+0.5]. '
[+1.0 maximum] "C
,'
,
Answer 3.12a s i Facility Comment: There are actually 9 fission chambers per channel,'there are 3 fission chamber detectors per channel, the student may answer either way and both should be acceptabl Systems Training Manual, "Excore Inst.," Ch. PG- R0-L-4.2, p. 5 (attached). ,
Resolation: The answer is amended to read: a. 3 or 9 Answer 3.13a Facility Comment: Above 2000 cps the WR NI channel cuts out 2 of the 3 fission chamber detectors and displays power in % rather than cps. This seems to indicate the answer should be true not false as indicated on the ke If the examiner is looking for false based on the fact that each -
fission chamber detector contains 3 fission chambers, I don't feel he is asking the question to test the candidates knowledge of the system. Tiue or False is correct depending on candidates understanding of questio Resolution: The answer key is amended to read: TRUE The reference is amended to include as pages, pp. 28 through 3 '
.
\
Maine Yankee 10 s
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'
4,
[ ,( Facility Comment: 3 RNO steps are listed for this condition ,
)* a) manually trip CEDM gen set output breakers; b) If two or more CEDM bus '
) under voltage status lights not lit then manually open and reclose 480 V
supply breakers to bus 9 and bus 12; c) If all trippable CEA's not .
inserted;thengog.oFR- . -
.
s Answer key only dists.a and ,
Answer b is de' pendent on CEDM bus under voltage'Ifghts
'
whose status is not listed in the questio s i
Grading of this question depends on the students assireption E-0, Rev. O, p. ,
Resolution: The answer key is amended.to add: ,
3. If all trippable CEAs NOT inserted, then go to FR- ,c
[+0.5 each, +1.0 maximum]
1 Answer 4.07
-
Fac'ility Comment: Answer #2 is not applicable since this trip function has been permanently disabled and the dump valve is continuously diverting to the PA3 sum OP 1-15-1, Step 2.1 System Training Manual, Vol. 4, Ch. 20,'pp.,14 and 3 \
Resolution: In the Precautions section of Procedure 1-15-1, " Primary Component Cooling System," there is the statemc r.t in 2.11.4 to maintain cdslection tank over' low aligned to PAB sump. However, Procedure 3-6.2.2.17 and Systems Training Manual, AS-1, have not been corrected (see Figure AS-19-I) to indicate that the dump valve to the PAB sump is locked open. The answer key is modified to read:
- shut the PCC surge tank vent [+1.0]
- the PCC surge tank vent should be manually reset immediately [+0.5]
,
, -
,,
f I '
%
Maine Yankee 11
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, _ _ _ . _ . , - . Question 4.13 Facility Comment: The statements listed in this question have been taken from Procedure 10-1 and reworded. The examiner has assumed that changing the words gives a Yes/No situation which may or may not be tru Example - Procedure 10-1, Precaution Step 3.1 reads - Core alterations or movement of irradiated fuel within the containment shall cease immediately.... If any of the following conditions occur:
3. . . A minimum of one boric acid transfer pump is not available....
The question states "One boric acid transfer pump is not available" and the answer key lists this as a condition which would cause a stop actio If one boric acid transfer pump was not available and since 3 pumps exist, a is easy and reasonable to assume 2 are available; therefore the answer key is incorrec This k re logic can be used for #5. Only 1 is inoperable and we have We only need and for #3 With the purge valves open and a purge not in progress refueling could continue as long as the valves were operable to auto close or if (120 hrs the system was lined up through the HEPA filters and for #1 The equipment hatch can be open as long as it is covered with plywoo Technical Specifications 3.13 and Procedure 10- There exists no statement which by itself would cause a stop action to occu Resolution: The question is delete Answer 4.14a Facility Comment: Could be answered as 3 rem with an updated NRC form-4 and not to exceed 5(N-18) (10CFR20.101)
Resolution: The answer key is modified to read:
a. 1.25 ren OR 3 rem with an updated NRC form-4 and not to exceed 5(N-18).
Maine Yankee 12
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FACILITY COMMENTS / RESOLUTIONS Question 5.01 Facility Comment: The use of the term Reactor Period is inappropriate for use in an exam given at Maine Yankee since it appears on no operational documents. See similar cc;nment for exam Section Resolution: Maine Yankee Training Lesson, MT-T-65, Rev. 1, Lesson Number R0-L-1.10 is entitled, " Period, SUR, In-Hour." This lesson plan references NUS Manual Module 3 in which the title to Section 6.2 is " Period." No change to the question and answer ke Answer 5.06a Facility Comment: The value listed in the answer key seems to be taken from TDB curve 1.1.1.1 which assumes Hot Full Powe At 80% power at 10 kMWD/MT the CB would be in the range of 240 to 270 pp If the candidate uses the curves provided and performs a linear interpolation:
HZP - 580 ppm HFP - 200 ppm 380 ppm (80%)(380) = 304 ppm 580 - 304 ~ 276 ppm If the candidate uses trained thumbrules and the curves provided HFP ~ 200 ppm Power Defect ~ 1.8% delta rho 20% (1.8) = 0.36% delta rho IBW ~ 100 ppm /% (0.36%) ~ 36 ppm 200 + 36 = 236 ppm Resolution: The candidates were informed to simply use the provided TDB curves for parts "a." and "b.", indicating what figures they were usin If the candidate calculates a boron concentration by either of the interpolation procedures suggested above, then those answers would be acceptabl Maine Yankee 1 l
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Question 5.06b Facility Comment: Question provides rod position at 520 steps which is impossibl Assuming a typo and you mean 120 steps on group 5:
Group 5 ARO ~ 1.43% delta rho Groupt 5 120 steps ~ 0.63% delta rho IBW ~ 100 ppm /%
-0.7% 100 ppm /% ~ -70 ppm 236 - 70 ppm ~ 166 ppm This question seems to involve double jeopardy in that if a is incorrect, b cannot be correct unless the grader takes the answer in a into accoun Both 5.06a and b also do not take into account the change in Xe not available from the curves provide Resolution: Figure 2.6.1 shows rod position in accumulated steps from 0 to 700; 520 corrects to 80 on group 5. The delta rho from ARO to 80 on group 5 is 1% delta k/k, which corresponds to 100 ppm boron. Hence, the answer accepted for part "b." is the answer in part "a." minus 100 pp Question 6.03 Facility Comment: This question is similar to R0 question 3.02, the same comments appl Resolution: The answer key is amended to read: FALSE [+0.5]
Maine Yankee 2
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l Answer 6.05a Facility Comment: The correct answer to this question is that 7 valves are temperature controlled (from the steam dump system) and that 5 valves are pressure controlled (from the turbine bypass system).
Systems Training Manual, " Steam Dump and Turbine Bypass," Ch. PG-1, PG-3, and PG- Resolution: On page 4 of PGS-14 it is stated that there are 10 temperature control valves and 2 pressure control valves and that 3 of the temperature control valves are used, in conjunction with the two pressure control valves, for the turbine bypass portio On page 1 it is stated the 7 valves comprise the Steam Dump portion of the system and the other five valves comprise the Turbine Bypass portion, and that each group of valves is operated by a separate control system. The Steam Dump control is based on Tavg and the Turbine Bypass control on steam header pressur The result of this confusion is that the answer key is amended to read:
a. 10 temperature control valves and 2 pressure control valves [+1.0] OR 7 temperature controlled valves and 5 pressure controlled valves [+1.0]
Answer 6.05e Facility Comment: Asking this question based on power level creates problems in that the signal is based on T-av Answer key should allow for a range of answer Resolution: The answer key is amended to provide a range of 70% to 84%
Answer 6.09b
'
Facility Comment: Candidates may respond in terms of Safety Limit being protected (i.e., DNBR) since training is conducted in that manner using the basis of Technical Specifications Resolution: The answer key is amended to read: b. to prevent possible fuel / cladding damage OR to stay within DNBR Safety Limits [+0. 5] .
Question 6.09c Facility Comment: Same as above except applicable Safety Limit would refer to kW/ft or LHG Resolution: The answer key is amended to read: c. to prevent power peaking '
leading to uneven core burnup and/or local fuel / cladding damage OR to stay within LHGR Safety Limits [+0.5].
Maine Yankee 3
. .--- -_ _ - - _ - _ .. . - - - - - = . -. ._ -. -. .-.
..- . . .
Question 6.11b Facility Comment: Maine Yankee does not have a pressurizer lo pressure tri The TM/LP trip which varies depending on factors including pressure has a minimum value of 1835 psi System Training Manual, "RPS," Section PG-2 Resolution: Question 6.11b is delete Answer 6.11d Facility Comment: 105.6 actual setpoint 106.5 by Technical Specifications System Training Manual, "RPS," p.1 Resolution: The answer key is amended to read: d.105.6 (setpoint) OR 106.5%
(Tech Specs).
Answer 6.11e Facility Comment: 45 psig ASO pressure or 4 of 4 turbine stop valves closed System Training Manual, "RPS," p. 22 and " Main Turbine," p.1 Resolution: The answer key is amended to read: e. 45 psig auto stop oil pressure OR 4 of 4 turbine stop valves close Answer 6.11f Facility Comment: 4.25 psig actual setpoint 5 psig Technical Specifications System Training Manual, "RPS," p. 2 Overall - This question asks for " Trip or pretrip setpoints," therefore the answer key should reflect correct pretrip valve System Training Manual, "RPS."
Resolution: The answer key is amended to read: f. 4.25 psig (setpoint) OR 5 psig (Tech Specs).
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l Answer 6.12 Facility Comment: The answer key is correct for deviation by puls A redundant deviation alarm is available from the reed switch syste "CEA position deviation reed switch" - 6 steps System Training Manual, "CEDS," pp. 48 and 4 Resolution: The answer key is amended to read:
10(pulse (pulsedeviation))OR 6 (reed switch deviation) [+0.5]
deviation [+0.5]
The reference is amended to include pp. 48 and 4 Question 7.03 Facility Comment: Same comments as R0 Question 4.0 Resolution: The answer key is amended to add: 3. If all trippable CEAs NOT Inserted, then go to FR-S.I. [+0.5 each, +1.0 maximum]
Answer 7.04 Facility Comment: The correct process for routing is SOS, PSS, then I&C superviso Resolution: The reference states that the Procedure shall be routed to the I&C Supervisor for review and action on reported malfunction No change to the answer ke Answer 7.08 Facility Comment: The proper response to this question is unclea The examiner should use wide discretion when gradin Resolution: Note Question 7.10 Facility Comment: Same comments as R0 Question 4.1 Resolution: Question 7.10 is delete Maine Yankee 5
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Answer 7.11c Facility Comment: Also shuts VP-A-1 through 5 and trips HV-9 on high radiation during an on-line purg System Training Manual, "AS-18," Ch. 20, p. 35/3 Resolution: The answer key is amended to read: c. stops PAB exhaust fan or blocks it from starting [+0.25] and during an on-line purge shuts VP-A-1 through 5 and trips HV-9 [+0.25].
The reference list is amended to include: 2. Maine Yankee: Systems Training Manual, " Radiation Monitoring," AS-18, pp. 35 and 3 Answer 7.13 Facility Comment: Additionally, it may be answered that "if any turbine step valve and its associated governor valve remains open, then close the EFCV Resolution: The answer key is amended to read:
-- manually stop the EHC fluid pumps (P-55A, P-558) [+0.5]
-- If any turbine stop valve and its associated governor valve remains open, then close the EFCVs [+0.5]
Answer 7.14 Facility Comment: The basis for this step also is due to the potential for reverse heat transfer from the S/G to the RCS causing heatup and potential system overpressur Technical Specifications Resolution: The answer is expanded to include: To prevent reverse heat transfer from the S/Gs to the RCS causing heatup and potential system overpressur The reference is expanded to include: 2. Maine Yankee: Technical Specifications, Section Maine Yankee 6
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Answer 7.15a Facility Comment: This question can also be answered from the fold-out pag core subcooling (12 deg F or 2 - PZR level Resolution: The answer key is amended to read: low PZR level ((5%)
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low PZR pressure ((1585 psig)
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low core subcooling ((12 deg F)
[+0.33 each]
Answer 7.15b Facility Comment: Answer is correct only if candidate assumes an SIAS is in progres Resolution: If a SIAS were NOT in progress, then ES-0.0 is not to be use No change to answer ke Question 7.17 Facility Comment: Same comments as RO Question 4.1 Resolution: The answer key is modified to read: a.1.25 rem OR 3 rem with an updated NRC Form-4 and not to exceed 5(N-18) rem total Question 7.18 is modified to include item . 75 rem and 25 rem for life saving and emergency actions [+0.5 each, + maximum]
Question 8.09 Facility Comment: Although these comments written here are not designed to address points of question relevancy or appropriateness, this question is totally inappropriat Since this exam sectin is suppose to have 25 points and 26.5 points exist it is requested that this question be totally removed from the exa Resolution: Section 8 has 25 point No change to question or answer key.
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Answer 8.10c Facility Comment: No correct answer exist The SRO licensed person does not I have to be from the same department responsible for the procedur Resolution: Question 8.10c is delete Question 8.16 Facility Comment: The statement from Tech Specs is incorrect in the questio One or more of 3 listed means of radiation detection must be provided not two (2) type The third method should be added to the answer ke Resolution: The answer key is modified to include item 3:
3. a radiation dose rate monitoring device in the hands of a radiological controls qualified individual [+0.5]
[+1.0 maximum]
.
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