IR 05000293/1987025
| ML20235V682 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 07/13/1987 |
| From: | Collins S, Howe A, Keller R, Lumb T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20235V665 | List: |
| References | |
| 50-293-87-25OL, NUDOCS 8707230273 | |
| Download: ML20235V682 (64) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT N (0L)
FACILITY DOCKET N FACILITY LICENSE N DPR-35
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LICENSEE: Boston Edison Company M/C Nuclear 800 Boylston Street Boston, Massachusetts I
FACILITY: Pilgrim Nuclear Poner Station EXAMINATION DATES: May 26 - 29, 1987 CHIEF EXAMINER: 8tkb >b, b 7 9. Q '
A. Howe, Reactor Engineer (Examiner) Date REVIEWED BY: 7-9-D l T. 4.umb, ReagjhF E611neer (Examiner) Date REVIEWED BY: )_ 7[/#[T7 R. M. Keller, Chief, Projects Section 1C Date APPROVED BY; , JH []ff)d ST J. Collins', Deputy Director }Dateh
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Division of Reactor Projects l
SUMMARY: Operator Licensing Examinations were administered to eight Reactor Operator (RO) candidates during the week of May 26, 198 All candidates passed the examination Because of the large number of significant plant changes made during the cur-rent outage, there is a concern about training in this are Therefore an inspec. tion of licensed operator training on modifications and procedure changes will be performed prior to plant startu $[k72hQ V
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REPORT. DETAILS
TYPE OF EXAMINATIONS: Replacement !
l EXAMINATION RESULTS: -l l
l RO l l Pass / Fail l I I I I I l Written l 8/0 l 1 I I I I I ,
l Operating l 8/0 l
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I i I I i I l Overall l 8/0 l l l l CHIEF EXAMINER AT SITE: S. Shankman, Acting Lead Reactor Engineer (Examiner) OTHER EXAMINERS: A. Howe, Reactor Engineer (Examiner)
G. Robinson, USNRC Consultant T. Lumb, Reactor Engineer (Examiner)
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3. Summary of generic strengths or deficiencies noted on operating tests:
This information is being provided to aid the licensee in upgrading license and requalification training program No licensee response is require STRENGTHS:
The candidates demonstrated a high level of familiarity with equipment i locations within the plan DEFICIENCIES:
1) The candidates were deficient in their knowledge of normal operations (e.g. control room indications, location of controls for instrument air, and the response of control rod drive (CRD) system parameters during normal rod motion).
2) The candidates hesitated in responding to questions about instrumen-tation since their training used information based on equipment which has been modified during the current outag . Summary of generic strengths or deficiencies noted from grading of written examinations:
This information is being provided to aid the licensee in upgrading license and requalification training program No licensee response is require The candidates had low class averages on the following questions:
Question Number Topic 1.09 Effects of a recirculation pump trip on the critical power ratio 2.06 Residual Heat Removal system response to a LOCA signal while in the Shutdown Cooling mode 2.07 Feedwater system controls 4.02 Procedure for shutdown outside the control room 4.09 Conditions that define a malfunction of an Emergency Core Cooling system component
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4 Personnel Present at Exit Interview:
NRC Personnel S. Shankman, Acting Lead Reactor Engineer (Examiner)
A. Howe, Reactor Engineer (Examiner)
i G. Robinson, USNRC Consultant l T. Lumb, Reactor Engineer (Examiner)
J. Lyash, Resident Inspector T. J. Kim, Resident Irspector Facility Personnel R. G. Bird, Senior Vice President, Nuclear P. E. Mastrangelo, Chief Operating Engineer S. Hudson, Operations Section Manager H. Balfour, Staff Assistant, Operator Licensing T. Sullivan, Nuclear Watch Engineer A. Shiever, Senior Nuclear Training Specialist L. A. Beckwith, Nuclear Engineering Department, Compliance Summary of NRC comments made at exit interview:
Dr. Susan Shankman, Chief Examiner, noted that she was detailed from Headquarters, NRR, to Region I, and will act as lead BWR examiner for the next several weeks. The specific results of the examinations would not be discussed at the exit meeting but they would be contained in the Examination Repor Every effort would be made to send the candidate's results in approximately 30 working day Commendations noted were:
1) smooth access to the plant, 2) good environment to conduct an operating test cr'ated through excel-lent cooperation from plant control room staff, 3) the examiners found content and organization of reference material sent in response to 90 day letter to be much improved; however, some inconsistencies were noted between the reference texts and the !
student guides, l l 4) the review of the written examination on May 26 went well; this !
review, however, was one hour longer than the standard two hour- !
review lengt '
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Concerns noted were: )
1) the untimeliness of information to the NRC about significant change in reactor vessel level instrumentatio ,
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The generic strengths and weaknesses given in paragraph three (3) of this report were summarize Overall: The Chief Examiner noted that generic weaknesses on normal opera-tions may be a reflection of the extended shutdown condition of the plan The experience and training received during the required one month at greater than 20% power and the completion of five significant control manipulations (as detailed in a May 7, 1987 letter to candidates) is particularly important for these candidates.
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Also concern was expressed that training on instrumentation was important.
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A follow-up on this training will be made as determined appropriate by regional managemen . Summary of licensee comments and commitments made at exit interview:
The licensee stated that the modifications to the reactor vessel level j instrumentation were modelled by the simulator. Also, that training on the modifications was scheduled to start on June 1, 198 The licensee stated that the additional required control room time will be structured training and closely supervise . Summary of Inspection Item (s) Opened:
As a result of the large number of significant plant modifications and procedure changes, an inspection of training given to licensed operators on major modifications and procedure changes affecting plant operations will be performed prior to restart. (50-293/87-25-01) Changes made to the written examination during the examination review and the resolution of facility comments are contained in Attachments 1 and Attachments: Written Examination and Answer Key (RO) Facility Comments on Written Examinations made after Examination Review and NRC Response to Facility Comments
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U.S. NUCLEAR REGULATORY COHISSION REACTOR OPERATOR LICENSE EXAMINATION Facility: n i s r. n i n Reactor Type: B UR- G E S Date Administered: S7/05/26 E.xaminer: n o B i tis 0N . Candicate: fA 4 5 7~F [ .
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 eacn cuestion are indicated in parentheses af ter the question. The passing grace 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 Cateoory
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7 c n3 Principles of Nuclear Power Plant Operation, Thermo-cynamics, Heat Transfer
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and Fluid Flow 7g .o 2;.p(3 Plant Design including Safety and Emergency '
Systems
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- n ' r . A.,; # Instruments and Controls -
-- ,, ,c g sw M Procedures - Normal, Abnormal, Emergency, and Radiological Control g o,?
wrn- 4 <,,,, TOTALS Final Grace %
All work done on this examination is my ow I have neitner given nor received ai Canoicate's Signature 0buw cA+.p/ g-ba. ;& fpf3 i
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l NRC RULES AND GUIDELINES FOR LICENSE EXAMINAT1DNS
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During the administrate on of thi s examination the f ollowing rules appl y: Cheating on the examination means an automatic. denial of'your application and could result in more severe penaltie l Restroom trips are to be l i mi t ed and only one candidate at a time may j leav You must avoid all contacts with anyone outside the examinati on l room to avoid even the appearance or possibility of cheatin . Use black ink or dark pencil only to f acilitate legible reproduction . Print your name in the blank pr ovi ded on the cover sheet of the examinatio l Fill in the date on the cover sheet of the examination (if necessary). ! Use only the paper pr ovi ded for answer . Print your name in the upper right-hand corner of the first.page of each section of the answer sheet.
I Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only og 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 onl y if they are commonl y used in facility literatur . The point value f or each cuestion is indicated in parentheses after the Question and can be used as a guide for the depth of answer require . Show all c al c ul at i ons , methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the Question or no . Par ti al credit may be gi ve Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAN . If parts of the examination are not clear as to intent, ask questions of the examiner onl . 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 examinatio This must be done after the examination has been completed.
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- 18. When you complete your examination, you shall: Assemble your ex ami nat i on as f oll ows:
(1) Exam questions on to (2) Exam aids - figures, tables, et (3) Answer pages including figures which are part of the answer, Turn in your copy of the examinati on and all pages used to answer the examination question Turn in all scrap paper and the balance cf the paper that you did not use f or answering the questions, Leave the examination area, as defined by the examiner. If after l eavi ng. you are found in this area while the examination is still in progress, your license may be denied or revoke l l
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. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
Ques t ion 1.01 (2 5)
- With no forced circulation available in the core, (1.5)
should the water level be increased or decreased (compared to the normal water level with forced circulation) to enhance natural circulation?
Briefly explain why this increase or decrease is recommende Gi ve three control room indications you would (1.0)
monitor to determine if natural ci rcul at ion exist Question 1.02 (3.0)
For each of th e events listed below, state whether the i
change will b ri ng the system CLOSER T0, FURTHER FROM or HAVE NO EFFECT ON the point at which the Reactor Recirculation Pumps will cavitat GIVE A BRIEF EXPLANATION FOR EACH ANSWE ! Increase in reactor water level (1.0) Loss of a fe e dwa t e r heater (1.0) Increase in Recirculation Pump Speed (1.0)
Question 1.03 (2.0)
The reactor has j us t s c ramme d after a long term full power ru (2.0)
All rods did not fully inser TEN hours after the scram, it has been determined that the Shutdown Margin is two percent. Assuming that no l
reactor coolant temperature changes occur er d all rod positions remain the same, explain how and_why the shutdown margin will change for the next 20 hocr (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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. PRINCIPLES OF NUCLEAR POWER PLANT O P t' R A T I O N ,
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THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW Question 1.04 (2 5)
During a reactor cooldown you are monitoring and (2.5)
recording the req ui re d temperatures and pressures according to Procedure 2.17, Vessel Heatup and .
Cooldow Over a hal f hour time period, the following '
is recorded REClR WIDE INLET TEMP RANGE LOOP A LOOP B RX PRESS 9: 00 AM 489'F 491*F 565 PSIG 9: 15 AM 475'F 476*F 465 PSIG j 9:30 AM 460*F 462*F 410 PSIG '
There has been some question concerning the :alib rat ion of the Wide Range RX Press i ns t r ume n t a t i o Using the steam tables provided, Indicate why you believe the pressure readings when compared with the recirculation loop readings would indicate that the wide range reactor p re s s u re instru-ine n t a t i o n is or is not properly calibrate Show all wor Ques t ion 1.05 (1.5)
Pumps should NOT be operated for significant p e r i o ds of
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lines close Briefly describe the consequences of such operation for Ce n t ri f uga l pumps ( two cons eq ue nc es ) (0. 8) Positive displacement pumps (one c o n s e q u e ri c e ) (0.7)
Question 1.06 (2.0)
For the following events, indicate whether the initial power response increases or decrease Briefly explain why, Single MSIV closure at 85% power (1.0) Isolation of extraction steam to H1gh Pressure Feedwate r Hea te rs at 90% power (1.0)
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.- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,
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THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW Question 1.07 (1.5)
State whether the fo' lowing would increase, decrease, or have no effect on Control Rod Worth, Moderator temperature decrease (0.5) Rod Density decrease (0. 5)
l Void Fraction increase (0.5)
Question 1.08 (2.5)
l During startup (power = 1 watt), a rod is pulled and a 60 second period is observed. With no further (2 5)
pulling of rods or changes in recirculation flow, could the operator maintain this reactor period for .
15 minutes? Show all work and i us t i f y your answe Question 1.09 (3.0)
Wl:h the reactor ope ra t i ng a t 100% steady stace condition, there is an i n a d ve r t e n t trip of one recirculation pum What is the initial response of INDICATED REACTOR ( 1. 5)
VESSEL LEVEL to this event and what is the reason f for this response? Does the Critical Power Ratio initially increase or decrease? Briefly jvstify your answe (1.5)
Question 1.10 (2.25)
The cancept of Subcritical Multiplication is used to describe the behavior of the reactor during refueling operations or startup I In a subcritical reactor, if the source s t re ng t h (0.75) I doubles, what wil' happen to the neutron level? Briefly j usti f y your answe In a subcritical reactor, if a reactivity of 0.003 delta K/k is added to the re ac to r , will it take (1.5)
longer to reach eq ui l i b ri um if the initial K-e f f e c t i ve is 0.95 or if K-effective is 0.9957 Briefly j us t i f y your choice.
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' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW Question 1.11 (2.25) !
Indicate whether the following s ta t emen ts are TRUE or FALS IF FALSE, CHANGE THE STATEMENT SO THAT IT IS If any part of the s t ri t e'm e n t is not
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CORREC NOTE:
true, mark the statement FALS i Beca us e the effective delayed neutron fraction (0.7$)
(B-e f f) is less at end of core life when compared to beginning of core life, the reactor perioa for a given positive reactivity insertion will be shorte r for the end of core cac : For a power decrease from 100 percent to 50 I (0. 75 )
percent, Xenon concentration increases to a peak 4-6 hours efter the powe r change Following a scram, reactor power as indicated (0.75)
by the neutronic instrumentation immediately ,
decreases on a negative eighty second perio l l
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' PLANT DFSIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS Question 2.01 (3.0)
Normal coeration of a Diesel Generator requires proper operatico of several auxiliary system For each of the ,
following component failures, explain what would happen 1 (Diesel would start, fall to start, stop, increase in speed, continue to operate with no change, etc.) and explain WHY the action you describe would occu The lubricating oil circulating pump fails while the (1.0)
diesel is operating during a surveillance tes ! The electric governor control signal is lost while (1.0)
operating after receiv:ng an emergency start signa A short occurs in the control circuitry for the (1.0)
AC powered pre-lube pump which supplies oil to the bearing An emergency start signal is receive Question 2.02 (2.5)
Consider the Automatic Depressurization System (ADS) The automatic blowdown has been Initiated by (1.0)
timing out of the 120 second time Indicate two cifferent actiorc or conditions that will inhibit the b l ov. iow n , snelude set points if I appropriate.
1 Explain the purpose of t h <: two minute time (0,5) Below what pressure will the safety relief (0.5) l valves reseat? Ars* *rc nos &o w do co a > TRUE or FALSE? The control switches for the relief (0 5)
l valves on the alternate shutdown panels are in the CLOSE position, the relief valves will not open in response to an ADS initiation signal but will respond to manual control from the control room (panel 903).
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O l PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
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Question 2.03 (2 5)
Consider the Reactor Water Cleanup System (RWCU) From what two locations does the RWCU system (0,5)
take a suction fo r cleanup of the reactor water? What two (di f fe rent) components of the RWCU (0.5)
system would the loss of RBCCW affect? In addition to High Space Temperature (1.5)
and Liquld Poison System Actuation, list '
three conditions that automatically isolate l the inboard (MO-2) and outboard (MO-5)
isolation valve Include setpoints.
l Question 2.04 (2.0) Match the following actions in Column A with (1.0)
the system pressures given in Column B. Pressures ;
may be used more than once or not Et al COLUMN A COLUMN B j motor driven fire pump starts 110 psi diesel d r i ve n fire pump starts 95 psi
3 jockey pump starts 85 psi If all fire pumps become inoperative, (1.0)
how is high pressure water obtained for the Fi re Protection water system?
Question 2.05 (2.5) Will indicated steam flow increase, de c re as e , (1.5)
or stay the s a me if one SRV opens when the
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reactor is at full power? Briefly explain your answe If the essential instrument air supply is lost, (1.0) ;
what two means are available to close the '
outboard MSIV's?
gy, , t, ye men -te close M 3W'J by Sit"'
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. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
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QUESTION 2.06 (3.0)
Consider the Residual Heat Removal System (RHR) What three functions does the Condensate Transfer (1,0)
System provide fo r RHR7 l The reactor is shutdown and operating with (2.0) l Shutdown Cooling. An actual LPC! initiation signal is receive What changes in valve and pump lineup occur automatically in the RHR and 1 Recirculation Systems? (Figure 1 is provided as an aid). Assume neither recirculation loop is broken and both shutdown cooling loops are operatin l l
QUESTION 2.07 (3.0) Each feedwater heater has three level switche (1.0)
Two switches send a high and low signal to the control room alar Briefly describe the function of the third level switch, Des c ri be the automatic actions which occur (2.0)
within the Condensate and Feedwater System if, while operating in the normal manner, at full power, one condensate pump trip Include time delays if appropriat QUESTION 2.08 (2.0)
Mechanica) seuls with seal purge wa;er are used to prevent excessive leEkage into the drywell from the Reactor Recirculation Pumps What sys tem s upplies the seal purge water? (0.5) If the number one seal fails, indicate ('.5)
two control room indications other than flow alarms that you woul d use to verify the failur (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
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' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY RYSTEMS l
Question 2.09 (2.5)
Consider the Primary Containment System l l Give two reasons why torus /drywell vacuum (1.0)
breakers are important to the proper operation and integrity of pri ma ry containmen In addition to s i gni fi can t drywell pressure (1.5) i and temperature changes, indicate three other i conditions that would indicate a leak inside the drywel Question 2.10 (2.0)
Consider the High Pressure Coolant injection System (9PCI) which is lined up for automctic initiation The booster pump normally takes a suction on (1.0)
the condensate s torage tan Indicate two separate conditions (i nc l ude set points)
which will automatically align the booste: pump for suction on t$e suppression poo Indicate whether the following are TRUE or FALS i Upon a HPCI system isolation, due to low (0.5)
steam pressure, the sys tem will not auto-matically restart after pressure is restored, even if the initiation signcis are still presen II If the HPCI turbine trips due to an over- (0.5)
speed condition it will restart when the speed coasts down to between 3000 and 4000 RPM.Gued 4w asha h y,7,,fg,r. raw +)
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3 INSTRUMENTS AND CONTROLS QUESTION 3.01 (3 0)
Consider the Average Power Range Monitors (APRM's) What are the two requirements regarding (1.0)
LPRM inputs to an APRM for the APRM to be considered operable? You cre operating at 92 percent of 'ull power (1.5)
and 100 percent co re flo Because of recirculation pump control p rob l e ms it is j desired to reduce core flew to 70 percent I while holding power constant at 92 percent I with control rod Would this be possible to do without the APRM's causing a s cram or a rod block? Justify your answe T RUE-ar-t# -5 E i ~
ioi, vie vi o ,, c ' ? m, c c, n , m , im, cc a c; M pe ! ' : .o p, k (,
QUESTION 3.02 (2.0) While operating at full power, one level transmitter in the ATWS/ARI System fails and indicates a low low water level and a spurious low low water level signal is received for two seconds by the other transmitter in that channe Does the reactor scram? Briefly explai (0.75)
il Do the recirculation pumps trip? Briefly explai (0.75) TRUE or FALSE 7 Alternate rod insertion duplicates the function of the RPS backup valves by venting the scram pilot valve air header pressure to cause a scra (0.5)
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3 INSTRUMEN15 AND CONTROLS
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QUESTION 3 03 (2.0) With the mode switch in STARTUP and the IRM's (1.0)
are indicating 20 on Range 4, an operator down ranged to Range 3 What trips, if any, would occur? Briefly justify your answe I With the mode switch in STARTUP, and IRM "C" (1.0)
reading 11 on Range 7, wha t t ri p(s), i f any, ;
would occur if IRM "C" was down ranged to Range 67 Briefly j us tify y;ur answe QUESTION 3.04 (2 5)
Consider the Refueling Flow Ventilation Exhaust Radiation Monitors (RFVE)
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' Indicate which of the conditions given below would ca us e the RFVE logic to trip during normai operatio Rad monitors "A" and "B' ere down scale (0,5)
and "C" senses high radiatio il Rad monitor "A" reads 20 mr/hr and (0.5)
"C" r6 ads 30 mr/hr lli All four rad monitors are downscale (0,5) What automatic action (s) occur when the t ri p (1.0) [
logic is tripped?
QUESTION 3.05 (3 0) The Recirculation Flow Control System uses two (1.5)
speed limi te rs (NO. I and NO.2) to limit the recirculation pump speed under certain condition For each speed limiter, give the condition that will initiate the limit and the percent of rated pump speed allowe The red scoop tube l o ck- up light indicates (1.5) .
that a lock-up is presen No Signal Failure {
Alarm occur Indicate three other possible !
causes of the lock-u (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
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3 INSTRUMENTS AND CONTROLS l
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QUESTION 3.06 (2.5) l With the reactor operating at full power under steady (2. 5)
state conditions in three element level control, an instrument technician mistakenly isolates and equalizes the pressure across one of the Main Steam Line flow transmitter Describe the response of the j feedwater control system until steady state conditions i are again establishe )
QUESTION 3.07 (3.0) ' Consider the Backup Scram Pilot Valves i Must the Backup Scram Pilot Valves be (0.5)
energized or deenergized to cause a l scram?
ii From what source are the Backup Scram (0.5) l Pilot Valves powered?
ili If one of the Backup Scram Valves s houl d (0.5)
fail to operate on receipt of the appropriate signal, what design feature assures that the other valve can perform j the necessary function? l The RPS busses provide 120 V power to what three (1.5)
radiation moritoring subsystems?
QUESTION 3 08 (2.0)
Briefly describe how the TIP System is isolated if it is in operation when a Group !! isolation is receive Indicate both automatic and manual operation For a normal TIP isolation (1.0) For the case where the detector cable cannot be (1.0)
withdrawn and a leak is suspected in the guide tub l
(****f CATEGORY 03 CONTINUED ON NEXT PAGE :****)
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3 INSTRUMENTS AND CONTROLS QUESTION 3.09 (2.C)
s/M List the conditions associated with the4" RETRACT (1.0)
PERMIT" 11pht that will cause a rod bloc I Give the three conditions that will cause the (1.0) !
rod block (in part a) to be bypasse QUESTION 3.10 (3.0) i Consider the A.C. Electrical Distribution a " Fast Transfer" automatically occurs when the (1.5) !
" Fast Transfer Switches" for the 4160 vac load centers are "0N" and an Auxiliary Transformer i failure or an opening of the Auxiliary Transformer j power supply breaker occur What th ree other i events will cause a fast transfer? j l Briefly explain why it is not a good practice to (1.0)
open a circuit breaker while it has a large current ;
flowing through it (unless an emergency condition j exis ts) . TRUE or FALSE? During operation, if an RPS MG set (0.5) .
fails, its associated RPS power panet will lose
{
power and will remain de energized until manually l re-energized from the B-10 backup power suppl l I
l (**f -* END OF CATEGORY 03 *****)
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- PROCEDURES - NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL CONTROL QUESTION 4.01 (3.0)
Consider a normal startup CRD coupling i nteg ri ty shall be verified when the rod (0.75)
's withdrawn the fi rs t time subsequent to each refueling outage or after maintenanc According to Startup From Shutdown (Proc. No. 2.1.1), how is coupling integrity verified? Under what conditions is primary containment required (0. 75)
to be maintained according to Proc. No. 2.1.17 In addition to the drywell and s upp res s i on ch ambe rs (1.5)
required to be intact and all blind flanges and man-ways closed, what are the th re e other req ui remen ts tha t must be satisfied for primary containment integrity to exist?
QUESTION 4.02 (3.0)
In accordance with procedure 2.4.143 - Shutdown From Outside Control Room Due to inhabitability of Control Room To what locations are the Ooerators and Operating (1.0)
Supervisor directed to go following assembly in the 23'
4 kv switchgear area? What is the preferred method to scram the reactor outside the control room 7 (1.0)
, When should the reactor feedwater pumps be (1.0)
tripped? (include all feedwater pumps)
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
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4 fROCEDURES -
NORMAL. ABNORMAL, EMERGENCY, AND RADIOLOGICAL
CONTROL QUESTION 4.03 (3 0)
The reactor is operating at 85 percent power when 125 VDC panel D 17 becomes and remains de-energize What three power supplies would you check in an (1.0)
effort to re-energize the panel? How will this malfunction affect reactor power? (1.0)
Briefly explai Which of the following components are NOT (1.0)
available as a result of this loss of power L4-d4 c s c i scociaivi p e /e k 2) RHR pump C 3) Core Spray Pump B 4) RCIC 5) p ri ma ry protective relays on b us A2 QUESTION 4.04 (2.5)
According to Procedure 2.4.147, Reset of Secondary Containment isolation List the four conditions that will cause a (1.6)
Secondary Containment Isolatio If the Secondary Containment isolation initiation (0.9)
signal has c l ea re d but the reset logic cannot be reset, what further verification should be performed?
l l (***fi CATEGORY 04 COLTINUED ON NEXT PAGE *****)
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. PROCEDURES -
NORMAL. ABNORMAL. EMERGENCY, AND RADIOLOGICAL
.
CONTROL QUESTION 4.05 (2.5)
in accordance with Procedure 5.5.1, General Fire Procedure and 5.5.2 Special Fire Procedure Who determines if the Plymouth Fi re Department (0.5)
is to be called in? (Give position title) Once the Plymouth Fire Department arrives on the (0.5)
fi re scene, who is responsible for control and l direction of fire-fighting ac ti vi t ies ? (Give {
position title)
I Who must authorize the activation of the Radiax (0. 5)
Emergency Backup Communications System? (Give position title) i I What.two types of fluids should NOT be used to (1.0) !
fight a fire in the vi ci ni ty of the new fuel j vault in the refueling floor?
QUESTION 4.06 (3.0)
in accordance with Procedures 6.1-024, Radiological Posting of Areas of the Station and 6.2-001, PNPS Radiation Expos ure Control Program What three access control conditions are required (1.0)
If a long-term radiation area is reading 1050 mrem /hr? In accordance with 10CFR20 whole body limits, would (1.0)
you be able to work in the area (in part a) for an hour? Justify your answe In accordance with PNPS administrative limits, (1.0)
under what conditions woul d you be allowed to work in the area (in part a) for an hour?
Assume you have worked at PNPS for several year (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
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. PROCEDURES -
NORMAL, ABNORMAL. EMERGENCY, AND RADIOLOGICAL
- 1 CONTROL l
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1 QUESTION 4.07 (3.0)- i
)
In accordance with Procedure E0P-02, RPV Control Power Indicate all the entry conditions for this procedur (1.5) ;
I By what th ree different methods can rod insertion (1.5)
failure be determined?
QUESTION 4.08 (2.5) .
On your shift while operating at 90 percent of full power, you notice that the suppression pool bulk temperature has risen to 88 degrees F and is increasing slowl A HPCI full flow test is in progress
} At what suppression pool temperature should (0.5)
suppression pool cooling be initiated? At what suppression pool temperature should HPCI (0. 5)
testing be secured? At what suppression pool temperature must the (0.5)
reactor be scrammed? l l What emergency operating p rocedure woul d you (1.0)
enter and when would you enter it?
QUESTION 4.09 (1.5) '
in Procedure E0P-01, RPV, Level and Pressure Caution (#10) indicates that it is permissible (1.5)
to secure ECCS or place it in the manuai mode if the system has malfunctioned. What are the two different conditions that define a malfunction of ECCS7 I i
QUESTION 4.10 (1.0)
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An increase in reactor power is occurring from (1.0)
unknown causet. In accordance with Procedure 2.4.13 Unexplained Rapid increase in Reactor Power, what
! parameter, when reached or exceeded, req ui res that a .
manual scram be initiated? (include setpoint if ,
appropriate) l i
i (***** END OF CATEGORY 04 *^***)
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TABLE II-3-1 PROPERTIES DE SATURATED STEAM AND SA*"URA*~ED (TEMPERA WATER ~UR.E) {
~
volume, tt /tc
Enthalpy. Btu /ic Temp Entropy. Blunt > a F Pres F psia water Evap Steam Water Evap Steam Water Evan Steam "
7 F 9 'rg V g hg h gg by Sg i sg gg 3't 0.0EE59 0.01602 3305 3305 -0 02 107 .5 0.0000 2.1873 2.1E 73 32 35 0.09991 0.01602 2945 294E 40 3.00 107 .1706 0.12163 0.01602 2u5 Pub 2.1767 35 45 8 C3 107 .0162 0.14 7c 0.01602 203 .B 13,04 2.1432 2.1594 40 50 0.17796 10681 10E .0262 2.1164 C.016 2 170 .E 18.05 2.142 E 45 60 0.25E l 105 E3 4 0.0361 2.0901 0.016;3 120 .6 28.06 2.1262 50 105 .7 0.0555 2.0391 70 2.0946 60 0.3E29 0 01605 EE B65 4 EO 3B05 10540 10921 0 0745 O5:EE O016:7 E 3 .9900 2.0645 70 90 48 04 *04! 4
,
1CYi 4 0.0932 06911 ;0;E10 4El 1 eE .942E 2.C3 5 9 , 60 EE 02 1042 7 110 l 100 0.9492 0 01E 12 250 4 35; 4 O.1115 1.E970 2.0066 90 1:0 6E 00 3C3 .2750 0.01617 2E E i 4 1.8530 1.9525 100 77.95 103 .3 0.1472 1.5105 1.9577 110 120 1.6927 0.01E20 203.25 203.26 E7 97 102 .2230 0.01625 1115 6 0.1645 1.7693 1.9339 157.32 157.33 97.96 101 .8592 0.01E29 111 .1817 1.7295 1.9112 122 95 122.00 107.95 101 .718 0.01E34 1122 0 0.1955 1.6910 1.6E95 97.05 97.07 117.95 140 160 4.741 100 .1 0.2150 1.6536 0.01640 77.27 77.29 127.96 1.8655 150 100 .2 0.2313 1.6174 170 1.64E7 160 1.993 0.01645 62 04 160 62.06 137.97 99 E .2 113 .511 0.01651 50.21 0.2473 1.5E22 1.E295 170 190 50.22 148 00 99 .2 9.3t 0 0.01 E 57 40.94 0.2E31 1.54E0 1.B111 160 200 40 96 155 04 96 .1 11.526 0.01664 33 62 0.2757 1.5145 1.7934 190 210 33.64 16E.09 97 .0 14.123 0 01E71 27 EO 0.2940 1AE24 1.7764 200 27.E2 17E.15 97 .3091 1.4509 1.7E00 210
. 212 14.E96 0.01E72 26 7E 26 50 '.50 17 220 17.156 97 .5 0.3121 1.4c 7 0.01676 23.13 23 15 ;SE.23 1.7568 212 230 20.779 0.01685 96 .3241 1.4201 19.364 19.381 195.33 95 .7 4 2 220 240 24.96E 0.01693 115 .3388 1.3902 16.304 16.321 20645 1.7290 230 250 29.E25 0.01701 95 .6 0.3533 1.3609 13.802 13.819 218.59 945 4 1.7142 240 116 .3677 1.3323 250 35 427 1 . '.~;9 1.7000 250 11 741 11 762 1 22E 7E 270 41.E 5i 00171E 935 E 11E7 4 0.3119 1.3041 280 10.042 10.060 23E.95 93 .E!E2 i 26 49.200 C01726 1170 E 0.3960 1.276i E.62 7 6.644 249.17 1.6729 270 290 57.550 92 .8 0.409 B 300 0 01736 7.u3 7 460 .25 .4 1.2501 1.fi99 280 E7.005 0 01745 6 44E 117 .223E 1.6473 6.466 26 .0 290 117 .1979 1.6351 310 77.67 300 0.01755 5.609 5.626 28 E9.64 0.01766 90 .5 0.4506 1.1726 4.896 4.914 2904 E9 .6232 310 340 117.90 0.01767 118 .1477 3.770 3.7BS 31 .8 1.6116 320 360 153.01 0.01811 119 .4902 1.0990 2.929 2 957 33 .5892 340 3B0 195.73 0.01535 86 .4 0.5161 1.0517 2.317 2.335 35 .5 1.5676 360 119 .5416 1.0057 1.5473 400 247.26 0.01864 380 1.E444 1.E630 3751 420 30E.7B 0.01294 E25 9 120 .56E7 09607 14505 14997 39 .5274 400 440 351.54 0.01925 B0 .1 0.5915 0.91 E5 460 1.1976 1.2169 41 .50SO 420 46 .0196 120 .6161 0.5729 480 0.9746 0 0942 u .2 24890 440 56 .0200 *- 0.7972 120 .8299 1.4704 0.E172 46 .6 460 120 .6648 0.7871 1.4518 500 630 9 0.0204 480
$20 0E545 0.6749 45 .3 l
81 .0209 1202 2 0.6E90 0.7 c 3 540 0535E 0.5596 51 .0 1.4333 500 96 .0215 0 u37 1199 0 0.7133 0.7013 1.4146 560 04651 53 .5 520 1133 4 0.0221 119 .7378 0.6577 1.3954 580 0.3651 0.3E71 562 4 62 .0228 118 .7E25 0.6132 1.3757 0.2094 0.3222 SE91 58 )
117 .7 E 76 0.5673 1.3550 GOD 154 i 0.0236 0.2435 0.2675 620 1756 9 61 .7 0.8134 0.0247 0.1962 0.2208 646 9 0.5196 1.3330 600 640 20593 50 .2 0.8403 0.0260 0.1543 0.1602 67 E9 1.3092 620 660 23E .0277 45 .7 0.8666 0 4134 680 0.1166 0.l u 3 71 .1 1.2621 640 270 .0304 0.0808 110 .3502 1.2498 660 l
0.1112 75E 5 31 .5 700 0.93E5 0.2720 1.20B6 680 s 309 .0366 0.0386 0.0752 E22 4 17 " 705.5 320 <
0.0505 0 99 .9901 0.1490 1.1390 700 0.050B 906 0 0 90 .0612 0 * 1.0612 70 {
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EDUATI0ti SHEET
.
f = ma v = s/t
. Cycle efficiency = (Ne: crx out)/(Energy in)
w= mg s=vt+ 1/2 c.:2 U
E = mc 2 KE = 1/2 mv a = (Vf - v9 )/t A = 2n A , g ge-11 PE = mgh V = V + at w = 6/t x = en2/t)pp = 0.692/: ,., i NPSH = P +
in sat ,
i
'l/2eff = [(+'1/2)(.'b) '
.
ma oAV [(t 1/2) * II b)3 AE = 931 am I = I e-Ex o
0 = mCpo 0 = UAch I = I e-WX o
Pwr = W ah f I=I o 10~*
TVL = 1.3/u-P = Pg 10sur(t)
HVL = -0. 693/u p = p et /,i o
SUR = 26.06/T SCR = 5/(1 - K,ff)
SUR = 26o / t = + (s - c)T CR x = S/(1 - Keffx)
CR)(1 . gdfi) = CR2 (1 - kdf2) !
T = ( t =/c ) + [(e . c )/[p ) M , jf()
geff) , gg,j;p c 7 = 1/( -8) \ j
~ = (E -c)/(ac) H = (1 - Keffo'/(1 . K eff1' -
l SDM = (1 - K l e-f)/Keff
.
.
o = (Kgf-I}/Kgf = t.K,ff/K d~ l' * IC' 58 0"d5
__
1 = 0.1 secones-I l o = [(I'/(T Kfff)) + [! g f (1 / + 17))
= I P = (:eV)/(2 x 10 0) I)d) 2 ,'I y]d3
'
I=eN 22 -
R/hr = (0.5 CE)/c#(r.eters )
NPSH = Static heac - h1-P 3n R/hr = 6 CE/d2 (feet';
Water Parameters Miscellaneous Conversions I 9E . = 8.345 1b I g a.} . = : . 76,t i t e r s I curie = 3.7 x 1CIC cps i ft# = I sc = 2.21 ID:
7.45 sa!. .
1.nb=2.itx 102 Density = 62.4 lom/ft- E af nr 1 n. = 3.41 x 106 g:gjn7 Density = 1 ;m/c J Heat of vapori:atier. = 970 B:u/'bt lin = 2.Et cr *F = 9/P; * 32
" eat of f usion = 14: E:c/lbe at = 5/9 <*r.32's IA m= 14.7;si= H..: in. H _ _ _ _ _ _ _ - - _ _ _ _ _ - _ _ - _ _ _ ~
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ftk5 rE (2_
', PRINCIPLES OF NUCLEAR POWER PLANT OPERATIO THERMODYNAMICS, HEAT TRANSFER AND FLUlD FLOW
, ANSWERS --
PILGRIM -87/5/26 - G. ROBINSON
]
ANSWER 1.01 (2.5)
e, increase (0.5)
With the reactor vessel level at normal (normal (1.0)
level during operation), natural circulation is retarded by the steam separator i Any three of the below answe rs (1.0)
i differential p res s u re across core plate Jet pump flow Ve rre / cha '" A"e Map I *
troda hr 1T=. C '*
6.e 4 PL+,
Vessel skin temperature g'u u / f,4 r core t empe ra t u re vessel pressure References Heat Transfer and Fluid Flow (S.G.) pgs. 8-55 6 56 objectives 10-4 and 5, pg 8-4 K/A Thermodynamics, K293008 Thermal Hydraulics K. l . 36 De s c . means op. can det. natural circ. (3.1)
K.1.37 Desc. means op. can enhance natural cire. (3.2)
ANSWER 1.02 (3 0) Further from cavitation (0 5)
As water level increases, the static head of water increases which increases NPSH (0. 5) Further from caviation (0.5)
Water entering the reactor is coole r which means the water is further from the saturation tem (0 5)
, closer to cavitation (0.5)
l As pump speed increases, the pressure in the eye of the impeller decreases, bringing it closer to cavitation (0.5)
(alternate answer further from cavitation, increase in recir flow, increases F.W. Flow, increases s ubcool i ng)
References Heat Transfer and Fl ui d Flow (S.G) 6-77a objectives 10-8 and 9, pg 6-3 K/A Thermodynamics 293006 Fluid Statics K 1.10 Define NPSH (2.7)
K 1.09 Define cavitation (2.8)
K/A Component 291004 Pumps Kl.01 Identification symptoms and consequences of cavi tat ion (3 2)
1
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- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, l
'
THERMODYNAhlCS. HEAT TRANSFER AND FLUID FLOW ANSWERS - - PILGRIM -87/5/26 - G. E. ROBINSON ANSWER 1.03 (2.0)
l The *ime at which the SDM was measured was near peak
,
l Xenon (Gr5)
l SDM will [IcNease over the next 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> (CCkf)
Xenon is past its peak thus it is decayin 'Since peak Xenon react,ivity is greater than two ,
pe/ N8c= 8 3et percent, the SDM will go to zero (The reactor will) N A6tuxdd ggo critical) (1.0)
I References Reactor Theory (S.G) pgs 1-36, 6-11 objectives 5-1 and 5 pg 1-3: 2-5 pg 6-2 K/A Reac*or Theory 292002 N-utron Life Cycle K 1.10 Define shutdown margin (3.2)
K/A Reactor Theory 292006 Fission Product Poisons K 1.107 Xenon "ollowing a Scram-State Effect on operation (3.2)
K/A APE 295015 incomplete Scram AK 1.01 Shutdown Margin (3.6)
AK 1.03 Reactivity Effects (3.8) l ANSWER 1.04 (2.5)
I VIDE RANGE j AVG REClRC EQUlVALENT SA RX PRESS LOOP TEMP PRESSURE READINGS t.490*F
,
9: 00 621 PSIA 580 PSIA {
9: 15 % 475'F 540 PSIA 480 PSIA
'
9:;0 N 460*F 467 PSIA 425 PSIA NOTE: Above calculation (1.0)
The wide range RX Press instrumentation is not calibrated properly (0.5) i The pressure r e a d i r.g s should be above the eq ui va l e n t saturated pressure readings for the recirc. loop (water in recir )
loop is subcooled) (1.0) i NOTE: same conclusion can be reached by converting wide range pressure readings to temperatur .
Keferences Heat Transfer end Fluid Flow (S.G) Chapter 3 objective 13 pg 3-2 K/A Thernedynamics 293003 Steam K 1.23 Use saturtted steam tables (2.8)
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- PRINCIPLES OF NUCLEAR POWER PLANT OPERATIO THERMODYNAMIC 5 HEAL TRANSFER AND FLUID FLDW
.
,
ANSWERS --
PILGRIM -87/5/26 - G. E. ROBINSON ANSWER I.05 (1.5) (pump will eventually add enough heat to the fluid)
to cause cavitation (0.4) and pump internal damage (caused by insufficient cooling) (0.4) pressure will build up (an d e ven t ua l l y rupture (0. 7)
the p i pe) (Not in Lesson Plan)
References Heat Transfer and Fluid Flow pg 6-108
.
K/A Component 291004 Pumps l K 1.04 Consequences of operating a centrifugal pump dead b aded (3 0)
K 1.18 Consequen:es of operating a positive
' displacement pump against a closed flow path (3 3)
ANSWER 1.06 (2.0) Increase (0.5)
Due to collapse of voids from the p res s ure spike (0.5) Increase (0.5)
Due to decrease in moderator temperature (or increase in density) (0.5)
References Reactor Theory (S.G) pgs 4-9 and 4-16 Objectives 1.5 and pgs 4-2 and 4-3 K/A System 239001 Main and Reheat Steam System A 1.10 Ability to predict changes in parameters associated with operating the Main Steam System controls including Reactor Power (3.8)
K/A System 259001 Reactor Feedwater System K3.12 Knowledge of the effect that a malfunction of the Reactor Feedwater System will have on Reactor Power ( 3. 8)
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- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUlD FLOW ANSWERS --
PILGRIM -87/5/26 - ROBINSON I ANSWER 1.07 (1.5) I Decrease (0. 5) Increase (0.5)
/ I- C ec re4 5 C (0.5)
l l References l
'
Reactor Theory (S.G) pgs 5-12a, 5-13a, 5-18a
,
Objectives 5.1 pg '
K/A System 201003 Control Rod and Drive Mechanism K 5.06 How control rod worth varies with moderator temperature and voids l K/A REACTOR THEORY 292005 Control Rods K 1.09 Explain di rec t ion of change in the magnitude of CRW for a change in control rod density (2.5)
ANSWER 1.08 (2 5)
P = P exp (t/ Tau) =
1 watt exp (900/60)
o = 3.27 MW (1.0)
l NO (0.5)
Moderator coefficient (or increase in Moderator temp)
would lengthen the period or heating range has been reached y : 5,,a PoM b g g ,, , ,,, y /. e - (1.0)
References 6~<'"" #'
Reactor Theory (S.G) pgs 7-10e and 3-17a Objectives 3-4, 3-6 pg 7-3 and 3-3 pg 3-4 K/A REACTOR THEORY 292008 Reactor Operational Physics K 1.11 Des c ri be reactor power and period response prior to reaching the POAH (3.6)
K 1.13 Explain characteristics to look for when the POAH is reached (3.8)
K/A REACTOR THEORY 292003 Reactor Kinetics and Neutron Sources i K1.08 Solve problems for power changes and period (2.7)
l l
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- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS --
PILGRIM -87/5/26 - G. E. ROBINSON ANSWER 1.09 (3.0) indicated water level increases (0.5)
Due to increased voiding in the core and decreased suction in the annulus (either one or both are acceptable) (1.0) decreases (0.5)
(The t i me constants of the fuel will cause) (1.0) l thermal power to lag behind the neutron flux and
, core flow decay and the mismatch between reactor I
power and recirculation flow results in a decrease in CPR y;- . , , , , ~ y s,ge w a L:; a r w --p a : ::- ! ;/ : ^ ^ l gg 7_ z: 5 , = - = jul' ::=d ; 'y :'~ h; .ss N References l Miscellaneous Training Ma te ri a l pg TAl-5 Objective 1.a og TAl-1 l K/A APE 295001 Partial Loss of Forced Core Flow Circulation l l
l AKl.03 Knowledge of the operational implications of partial loss of forced core flow circulation ;
to thermai limits (3.6) i AK2.03 Knowledge of the interrelations between I
partial loss of forced core flow circulation and reactor wa te r level (3.6)
K/A Thermodynamics 293009 Core Thermal Limi ts Kl.23 Describe the effects of mass flow on critical power (2.2) j
l l
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- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS --
PILGRIM -87/5/26 - G. ROBINSON l
ANSWER 1.10 (2.25) The count rate, thus the nectron level doubles (0.75)
since count rate is directly proportional to source strength. NOTE: use of CR = S/(1-k- !
effective) to jus t i fy answer is acceptabl The case when K-e f f e c t i ve = 0.995 will take longer to reach equilibrium ( 0 -. 5 ) I It takes longer to reach a new steady state beca us e of the increased n umbe r o f generations ,
which exist with the increased population. (1.0) i References '
Reactor Theory (S.G) pgs 3-8a, 3-9 a , and 7- 8 Objectives 1.3 pg 3-3 and 1.5 pg 7-2 K/A REACTOR THEORY: 292008 Ope ra t i ona l Physics )
Kl.04 Relate the concept of subc ri ti cal mul t i-plication to predi cted count rate and period response for Control Rod withdrawal during .
the approach to critica (3 3) )
Kl.05 Explain characteristics to be observed when i the reactor is very close to critica (4. 3) )
ANSWER 1.11 (2.25) TRUE (0.75) l TRUE (0. 7 5) FALSE (0.3)
Reactor power drops very fast (prompt drop)
and eventually follows a negative eighty second perio (0.45)
References Reactor Theory (S.G) pgs 3-36a, 6-10a and 7-22a Objectives 4.4 & 5.6 pg 3-4, 2.5.3 pg 6-2, 7.2 pg 7-4 K/A Reactor Theory 292003 Reactor Kinetics and Neutron Sources Kl.06 Explain the effect of delayed neutrons on re ac to r period (3.7)
K/A Reactor Theory 292006 Fission Product Poisons Kl.06 State the effect on reactor operations of maneuvering Xenon (2.6)
K/A Reactor Theory 292008 Reactor Operational Physics Kl.25 Explain the shape of a curve of reactor power ve rs us time after a scra (2.8)
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. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS ANSWERS -- PILGRlM -87/5/26 - G. E. R0BINSON l ANSWER 2.01 (3.0)
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' The diesel would shutdown (0.5)
l if pressure drops too low (less than 40 psi),
a pressure switch at the uppermost bearing (no. 13) sends a signal to shutdown the diese (0.5) The D.G. would speed up (0.5)
until the mechanical governor takes over speed control (0.5) Engine would start (0.5)
The A.C. powered pre-lube pump is not req ui red for en e rne r g e n c y star (0.5)
References System Reference Text, Book 2, Diesei Generator pgs 11-1, 5-2 and 28-1 Objectives -
System Student Guides and Objectives, Book 4, Diesel Generator System pg 4 ob , 9, 16, 22 and 25 K/A System 264000 Emergency Generators K6.03 Knowledge of the effect that a loss or malfunction will have on the D.G. of lube oil pumps l
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- , PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS ANSWERS --
PILGRIM -87/5/26 - G. RDBlNSON i
ANSWER 2.02 (2 5)
{
1 Any two of the below (0 5 each)
by resetting timer logic (both timers)
{
pull fuses on logic bus by securing RHR and CS pumps or less than 150 psi discharge press ure from ECCS pumps (NOTE: Student Guide does not agree with Lesson plan and Logic Diagram) to allow HPCI to restore water level before relief valves are actuated (0.5) psi (check if this is between vessel and containment) (0.5) ' FALSE (will not respond to manual control in the control room) (0.5)
References System Student Guides and Objectives, Book 3, ADS pgs L G 4, 5 and 6 Objectives 6, 11 and 14 pg L K/A System 218000 Automatic Depressurization System K1.05 Knowledge of cause-effect relationships between ADS and the remote shutdown systen (3 9)
K4.03 Knowledge of ADS design features of ADS logic control (3.9)
A1.04 Ability to predict changes in parameters associa ted wi th ope rat ing ADS including Reactor P res s ure (4.1)
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' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS ANSWERS -- PILGRIM -87/5/26 - G. E. ROBINSON ANSWER 2.03 (2.5) "A" Recirc loop (on the pump suction side) (0.25)
bottom of the reactor vessel (0. 2 5) Non-regenerative heat exchanger (0.25)
RWCU recirc. pump seals (0.25)
l l Low reactor water level (0 3); +9 inches (0.2)
NRHX high outlet temperature (0.3); 140*F (0.2)
i High Flow at System inlet (0. 3) ; 300 percent rated l flow (0.2)
References System Student G ui des and Objectives, RWCU System pgs LG 3,4 and 13 and Figure i Objectives 3, 10 and 22-8 pgs 4 and 5 K/A System 204000 RWCU System Kl.01 Knowledge of the physical connections between RWCU and the reactor vessel (3.1)
K6.01 Knowledge of the effect that a loss of component cooling water system will have on RWCU (3.1)
K4.04 Knowledge of RWCU system interlocks which provide for system isolaton (3 5)
ANSWER 2.04 (2.0) psi jockey pump starts (0.34)
95 psi elect ri c motor driven fi re pump starts (0.33)
85 psi diesel d ri ve n fire pump starts (0 33) System is supplied by the public water lead in line (0.5)
can be boosted to 125 psi by a fire truck (at PNPS site) (0.5)
References System Reference Texts, Fi re Protection Water System pgs 1-1 and 2-1 System Student Guides and Objectives, Book 4, Fire Protection System, objectives 7 and 9 pg 3 ( K/A Sys tem 286000 Fi re Protection System K4.01 Knowledge of Fire Protection System design features which provide adequate supply of water (3.4)
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' PLANT DESIGN I N C l. U D I N G SAFETY AND EMERGENCY SYSTEMS l l
ANSWERS --
PILGRIM -87/5/26 - G. ROBINSON l ANSWER 2.05 (2.5) indicated steam flow decreases (0.5)
Nozzles (Restrictors) which measure l steam flow are located downstream of SRV's (1.0) air accumulator (connected between air supply 4 check valve and the valve control unit) (0.5)
s p r i n g' force (0.5) ### A M#"'" ##""
h.p w de &p o p ht R References System Student Guides and Objectives, Main Steam System, pgs LG-7 and 15 and Figure 1
{
Objectives 18C, 19 and 20 pg 5 l
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K/A System 239001 Main and Reheat Steam System {
!
K6 Knowledge of the effect that a malfunction of the following will have on the Main Steam System K6.02 Plant Air Systems (3.2)
K6.04 Rellef Valve Operability (3.4)
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, PLANT DESIGN INCLUDING S A FE TY AND EMERGENCY SYSTEMS ANSWERS -- PILGRIM -87/5/26 - ROBINSON ANSWER 2.06 (3.0) Keeps LPCI system full (to te ,ure a vai l ab i li ty Immediately upon initiation) (0 34) I Flushes the RHR pumps and piping prior to lineup for shutdown cooling (0 33)
Provides torus makeup water (0.33) (Check at Facility)
Valves 50 and 47 close (0.4)
All pumps trip (0.4)
Valves 16A and B locked open (for 1 minute) (0.4) I Valves 29A and B (injection valves) close (0.4)
(LPCI Loop selection selects Loop B)
Loop # discharge valve closes and recir pump motor tripped if running (0.4)
References 4 System Student G ui des and Objectives, Book 3, RHR f System pgs LG 10, 14, 15, 21 and Figures 3 and 4 J
I Objectives 2, 3, 8, 10 and 14 pg 4 K/A System 203000 RHR/LPCI Mode J Kl.14 Knowledge of cause effect relationships between LPCI and Shutdown Cooling System (3.6)
K4.11 Knowledge of LDCI interlocks and loop selection logic (4.0)
Kl.01 Knowledge of physical connection LPCI l and Condensate Transfer System (2.8) l K1.04 Knowledge of physical connection LPCI and Keep Fill System (3 3)
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, PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS ANSWERS -- PILGRIM -87/5/26 - E. ROBINSON ANSWER 2.07 (3.0) The third switch interrupts instrument air to _ ,4 ,,,y,jl.0)
the bleeder trip valve"and spill valve whe heater water level reaches the very high level * CNd '
setpoin A4 4<- ILow NPSH shuts both coarse and fine reject valves (0.75)
- if low NPSH signal does not clear within (0.75) '
_ 60 sec, a feedwater pump is tripped
{Which feedwater pump tripped is determined by the RFP tripping ci rcui t
~ (0.5)
References 1( OP p f ,<. 4.p rek e k ss ,hd or oN W w g, %g pap ,,,, y 6.y b p ,v 4 4 "' M System Student Guide and Objectives, Book 2, Condensate and Feedwater Systems pgs LG 17, 18 and
Objectives 23, 27 and 55 K/A System 259001 Reactor Feedwater System K4 Knowledge of Reactor Fe6dwater System design and interlocks which provide for:
K4.01 Feedwater heating (2. 8) {
Kk.05 RFP protection (2.7) 1 A3.10 Ability to monitor automeric operations l of the reactor feedwater system including pump trips (3.4)
ANSWER 2.08 (2.0) control rod bydraulic system (0.5) increase in cavity number 2 pressure (to the same as number 1 about 1000 psi) (0.75)
increase in seal cavity temperature (0.75)
(either 1 or 2)
p, en. e va." ,,, h e k *r e N 'f"'fA#W References System Student Guides and Objectives, Book 2, Recirculation System pgs LG-4 and 27 Objectives 16, 17, 18, and 25e pgs 5 and 6 K/A System 202001 Recirculation System K4.04 Knowledge of Reci rc ul a t ion System l design and controlled seal flow (3.0)
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_ PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
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ANSUERS --
PILGRIM -87/5/26 - G. E. R0BINSON ANSWER 2.09 (2.5) They limit the dy nami c wa t e r level in the downcomers (to two feet or less) (0,5)
They prevent exceeding the drywell external design pressure (of 2 psig) (0,5)
Any H a excessive sump pump operation Dry c # * O lef4 d**. (0.5) '
(kw abrupt change in drywell h umi di ty ^ * * * d'1kil /% elfp (0.5)
higher than normal drywell radiation Eko,e pu p sur /e.hp. (0.5)
- '-^ ^ ; - '
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( /: ' n _; ggy
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References System Student Guides and Objectives, Book 4, Primary Containment System pg LG-14 and 57 Objectives 5 and 39 pgs 5 and 6 K/A System 223001 Primary Containment System K5.01 Knowledge of the operational application of the vaccum breaker / relief operation to the Primary Containment System (3.1)
A1.10 Ability to predict changes in parameters associated with operating the P ri ma ry Containment System including the d rvwe l l leak detection system (3.4)
ANSWER 2.10 (2.0)
/S CST level reaches J3rinches (0.5)
pool level rises to 5 inches (0 5) ,
1 FALSE (will restart when pressure is restored) (0.5)
TRUE (CAF) (0.5)
i References System Student Guides and Objectives, Book 3, HPCI pgs LG8 and 23 Objectives 12b and d K/A System 206000 HPCI System K4 Knowledge of HPCI design features and or interlocks which provide for the following K4.03 Resetting turbine trips (4.2)
K4.04 Resetting system isolation (4.0)
K6 Knowledge of the effect that a loss of or malfunction of the following will have or HPCI K6.04 Condensate storage tank level (3.5)
K6.05 Suppression pool level (3.5)
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3 INSTRUMENTS AND CONTROLS ANSWERS - PILGRIM -87/5/26 - G. E. R0BINSON
)
ANSWER 3 01 (3.0) At leas t two LPRM detectors per level (0. 5)
At least eleven detectors (total) (0,5) NO (at least a rod block depending on (0. 5)
value of FRP/MFLPD) i
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FRP Setpoint (RB) j (0.58 W + 50 percent) MF PD (1.0)
(Maximum FRP/MFLPD can be is onej 0.58 x 0 7 + 50 - 90.5 which is less than the 92 percent power Ab54 r- M-- ',,-)-- O '
References System Student Guides and Objectives, Book 3, Neutron Monitoring Systems pgs LG-13 and 19 Objectives 6 pg 4 (NO objectives found for parts b and c)
K/A System 215005 AFRM/LPRM Kl.04 Knowledge of the cause effect relationships between l
K4.01 Knowledge of APRM design features and interlocks which provide rod withdrawal blocks (3.7)
l A1.04 Ability to predict parameters associated with APRM's Scram and red block t ri p setpoint (4.1)
ANSWER 3 02 (2 0) i reactor scrams immediately on receiving two (0.75)
low-low water level signals in one channel il recirculation pump does not trip because (0.75)
there is a nine-second delay built into the trip circui TRUE (0.5)
References System Student Guides and Objectives, Book 3, RPS and ATWS Systems pgs LG 15 and 16 Objectives 21 and 22 K/A System 212000 Re a c to r Protection System K4. Knowledge of RPS design and interlocks which provide for system redundancy and reliability (3.4)
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3 INSTRUMENTS AND CONTROLS
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ANSWERS - PILGRIM -87/5/26 - G. R0BINSON ANSWER 3.03 (2.0) N o r, e (0,5)
l The IRM's would indicate 20 on Range 3; 0-40 scale (0.5)
l l An IRM upscale trip (rod block) would occur (0.5)
IRM would read 110 on the 0-125 scale (0. 5)
References System Student Guides and Objectives, Book 3, Neutron Monitoring System pg LG-22 and 32 Objectives 17 K/A Components, 291002 Sensors, Detectors K 1.20 Neutron Monitoring Units (3.2)
ANSWER 3.04 (2.5) i no t ri p (0. 5)
it ws. 49 (0.5)
lii trip (0.5) Standby Gas Treatment starts (0.5)
Reactor Building Ventilation isolates e'
(0. 5)
G ' c- L :t s e t J, s or h pp.ed
- References t
l System Guides and Objectives, Book 3 l Process Radiation Monitoring System pg LG-7 Objectives 2f, and 14c pgs 3 and 4 K/A System 272000 Radiation Monitoring System 4.02 Knowledge of Radiation Monitoring System de'ign s and i nte rlocks whi ch provide auto actions to contain the radioactive release in the event that the p re de te rmi ned release rates are exceeded (3.7)
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3 INSTRUMENT _5 AND CONTROLS
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ANSWERS --
PILGRIM -07/5/26 - ROBINSON l
ANSWER 3.05 (3.0) t
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l Speed Limi te r 1 Recire, discharge valve less than 90% open or (0.75)
Feedwater Flow is 20% of rated Maximum pump speed is 28% of rated Speed Limiter 2 $
All three RFP's are not operating and reactor (0.75) i water level is 19 inches or less Recirc. pump speed is limited to less than 65% (Any three) (0.5 pts. each)
M-G set drive motor bus undervoltage (80% of rated voltage) on bus A-3 or A-4 Lube oil high-high temp ( 210 * F)
Lube oil low pressure (6 sec. delay) (30 psig)
Generator Bus Lockout Relay trips Re f e re nces System Student Guides and Objectives, Book 2, Re ci rcul a t i on Eystem pgs LG-10 and 26 Objectives 13c, 26d, 31a and b pgs 4, 5 and 6 K/A System 202001 Re ci rcul a t ion System Kk.16 Knowledge of recirc. system interlocks (3.3)
which provide for recirc. pump runback K6.06 Knowledge of the effect that a loss of (3 1)
recirc. system M-G set has on the recir, system i
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3 INSTRUMENTS AND CONTROLS
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ANSWERS - PILGRIM -87/5/26 - G. ROBINSON ANSWEh 3.06 (2.5)
Total steam flow would indicate 75 percent while (0.5)
actual steam flow would remain at 100 percent The FWL control system would reduce feedwater (0. 5)
l according to the change in indicated steam flow Level will begin to decrease and a level error signal (0,5)
will be generated Feedwater flow will increase back to the 100 percent (0.5) !
level to bring level back near normal Final level will be established somewhat lower than (0. 5)
the original error (level error signal is equal but !
opposite to flow error signal)
Reference System Student Guides and Objectives, Book 2 I Condensate and Feedwater System pgs LG-26, 27 and 45 Objectives 42, 43, 44, 84 pgs 7 and 9 K/A Compenent 291002 Sensors / Detectors Kl.08 Level Modes of Fe ' l ure (3 3)
K/A System 216000 Nuclear Boiler Instrumentation l
Kl.12 Knowledge of cause effect relationships between Nuclear Boiler Instrumentation and the re a cto r water level control system (3.6)
ANSWER 3.07 (3.0) i energized (0.5)
il 125 v D.C. Station Battery (0.5)
111 check valve installed around one valve (0.5)
to prevent a valve f ail ure inhibiting the desired scram action g r v s a d rf u , s ( e l. uJ 4 , n ) re::::r i *' . : -: ':: : --- *4 (0. 5)
off gas system radiation monitors (0. 5)
main steam line radiation monitors (0.5)
References System Student Guides and Objectives, Book 3, RPS and ATWS Systems pgs LG-7 and 25 and Figure 1 Objectives 3,4, 10 and 19 pg 4 K/A System 212000 Reactor Protection System Kl.05 Knowledge of cause effect relationship l between RPS and Process Radiation Monitoring j Systems (3.3) l K3.06 Knowledge of effect that loss or malfunction 1
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of RPS has on scram air header solenoid operated valves (4.0)
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3 INSTRUMENTS AND CONTROLS ANSWERS - PILGRIM - 87/5/26 - ROBINSON ANSWER 3.08 (2.0) Any TIP detector not in its shield chamber is (0.5)
transferred automatically to manual re ve rs e When the detector is i r, its shleid chamber (as (0. 5)
indicated by the chamber switch) the associated valve closes A shear valve (outs i de the containment) is (1.0)
activated by a keylock switch on the valve control
. monitor References System Student Guides and Objectives, Book 3, Neutron Monitoring System pgs LG 18, 19 and 20 O b j e c t i v'e s 1, 27 and 28 pgs 4 and 5 K/A System 215001 TlP Kl.05 Knowledge of ca use effect relationship between TIP and PCIS (3 3)
A2.07 Ability to predict the impact of failure to retract during acc; dent conditions on T!P (3.4)
ANSWER 3 09 (2.0)
A rod block is initiated if SRM channel detects less than 100 CPS (0.5)
and its detector is not full inserted (o,3)
The rod block is bypassed if the mode switch is in RUN (0 34)-
The associated IRM range switches are of range 3 or above (0 33)
the channel is bypassed (0.33)
References System Student G ui des and Objectives, Book 3, Neutron Mon i to ri ng System, pg LG 5 Objectives 14 and 15 pg. 4 K/A System 215004 SRM System Kl.03 Kr.owledge of the ca us e effect ral a t i ons hi ps between SRMs and Reactor Manual Control (3.4)
K4.01 Knowledge of SRM system interlocks which provide rod withdrawal blocks (3.7)
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3 INSTRUMENTS AND CONTROLS ANSWERS - PILGRIM -87/5/26 - E. ROBINSON ANSWER 3.10 (3.0) generator lockout (0. 5) r,4.f- opei,,it of turbine lockout (0.5) Ace's fo+ e d " #
L160 > r-t e: ';;E2ut
.
(0.5)
ec .ew sc % opening a b re ak e r under these conditions will (1.0)
draw an arc between the breaker contacts causing pitting and other damage to the b reake r contacts TRUE (0. 5)
References System Student Guides and Objectives, Book 1, AC Electrical Di s t ri b ut ion pgs LG 6, 9 and 15 Objectives 7, 11. and 21 K/A System 262001 A.C. Electrical Distribution K/4 Knowledge of AC Electrical Distribution K4.03 Design features and interlocks between automatic bus transfer and breakers ( 3 .,1 )
K4.04 Redundant power sources to vital buses (3.6)
K/A Component 291008 Breakers, Relays and Disconnects Kl.08 Effects of closing breakers with too much load (3.4)
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. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND RADIOLOGICAL
. ENTROL ANSWERS - PILGRIM -87/5/26 - G. E. ROBINSON ANSWER 4.01 (3.0) observing discernible response of nuclear (0.35)
instrumentation when withdrawing control rods performing a coupling check when the rods is at (0.4)
full out position 48 p,LM ce co-ph et cire- when the reactor is critical or (0.4)
when the reactor water temperature is above 212*F (0.35) All manual co nt a i sime n t isolation valves on lines (0. 5)
connected to the RCS or containment not required to be open during accident conditions are close At least one door in each airlock is closed and sealed (0 5)
,
All automatic containment isolation valves are operable or de-activated in the isolated position (0.5)
References Proc. No. 2.1.1 Start up from shutdown pg 2.! l-4 System Stedent Guides and Objectives, Book 4, Primary Containment System pg LG-49 Objective 10 pg 5 K/A Systen 223001 Frimary Containment S y s t e:n and Auxiliaries System Gene ri c 10 Abil:ty to explain and apply
,
all system l imi ts and precautions (3.2)
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4. PROCEDURES - NORMAL, ABNORMAL. EMERGENCY AND RADIOLOGICAL CONTROL ANSWERS - PILGRIM - 87/5/26 - E. ROBINSON ANSWER 4.02 (3.0) One operator stays in the 23' elevation kv switch gear area (0.25)
One ope ra to r goes to the 37' switch gear a re a (0.25)
The CR0 goes to the RPS MG Set Room (0. 2 5)
Op. Supervisor goes to Inst. Rack 2205 or 2206 (0.25) Open the breakers to the APRM's at the RPS (1.0)
power panels two RFPs when the scram occurs (0.5)
one RFP when level i n d i, c a t i o n starts to increase (0.5)
" a i ff. do rec H ~ ~ F 'f-lo se pu su <
Reference Procedure 2.4.143 Shutdown from Outside Control Room Due to inhabitability of Control Room pg 2. 4.14 - 3- 7 a n d 8 K/A APE 295016 Control Room Abandonment System Gene ri c 10 Ability to perform aithout re f e re nce to procedures those actior5 * hat req ui re immediate operation of system cc mponer ts >
or controls (3.8) {
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4. PROLEDURES -
NORMAL. ABNORMAL. EMERGENCY AND RADIOLOGICAL ;
CONTROL ANSWERS - PILGRIM -87/5/26 - G. E. ROBINSON ANSWER 4.03 (3 0) l l MCC B-14 via Charger (0.34) l l MCC B-10 via Charger (0. 33) l Battery B (0. 3 3) l 1 Reactor power will decrease beca us e recir (1.0)
MG B has tripped ) '.' s i leet (0.2) d l* k 2) Not lost (0.29 3) Lost (0.2$
4) Not lost ( 0 . h)
5) Lost ( 0 . 25)
(an answer of 3 and 5 is expected)6 asja wort a e cm4 e< % cl--- L 6 *
References Procedure 5 3 12 Loss of Essential Bus D-5 pg 2 L* stem S t u d e r. t Guides and Objectives, Book 1, DO Electrical Distribution System pg LG-10 and Figure 1 Objectives 3 and 6 pg 3 K/A APE 295004 Partial or Complete Loss of D.C. Power l
AK2 Knowledge of the i n t e r r e 'l a t i o n between partial l loss of D.C. power and AK 2.01 Ba t te ry Charger (3.1)
AK 2.02 Batteries (3 0) ,
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AK 2.03 D.C. Bus loads (3 3)
System 263000 D.C. Electrical Di s t ri but i on ,
i K2.01 Knowledge of electrical power supplies to the major D.C. loads (31)
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. 4. PROCEDURES - NORMAL, ABNORMAL. EMERGENCY AND RADIOLOGICAL CONTROL ANSWERS -
PILGRIM -87/5/26 - G. ROBINSON ANSWER 4.04 (2.5) Low Reactor Water Level (0.4)
l High Drywell Pressure (0.4)
RFVE Rad Monitor High Radiation (0.4) )
RFVE Red Monitor downscale (0.4) Ve ri f y relays (on Panels 170/171) for closure (0.9)
of containment a tmos phe ri c control system isolation valves are energize e cr ea te av w ,C.j si.slu c i.c. a G por,J n a.drus4* k '
Reference Ar * * /r /, , ce hM Procedure 2.4.147 Reset of Secondary Containment Isolation pg 2.4.147-2 (
K/A EPE 295033 High Secondary Containment Area Radiation Levels System Generic 7 Ability to explain and apply all system limits and p re ca u t i ons (3.6)
System 290001 Secondary Containment I
A.301 Ability to monitor a utoma t i c ope ra t ions of the Secondary Containment incl udi ng s econdary containment isolation (3 9)
ANSWER 4.05 (2.5) F're Brigade Chief (0.5) ! Fire Brigade Chief (0.5) Watch Officer (0.5) Low Density fire fighting extinguisher foam (0. 5)
water fog or spray from a fire hose (0.5)
References Procedure 5. General Fi re Procedure pgs 5.5.1-2 and 3 Procedure 5. Special Fi re Procedure pg 5.5.2-3 K/A Plant Wide Ge ne ri cs 294001 Kl.16 Knowledge of facility protection requirements including fire brigade and portable fire-fighting equipment usage
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4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL ANSWERS -- PILGRIM -87/5/26 - E. ROBINSON ANSWER 4.06 (3 0) Posted with the words " CAUTION HIGH RADIATION AREA" (0.34)
Posted with the words "RWP REQUIRED FOR ENTRY" (0. 3 3)
l Lorked when not occupied (or when access is not controlled) (0.33) YES if for that quarter less than a dose ra te of (1.0)
200 mrem had been received (1250-1050)
or i f 5 (N-18) requirements f ul filled along wi th a NRC Form-4 available Mus t be authorized for an exposure upgrade (CAF) (0.5)
and must satisfy 5(N-18) and NRC Form-4 req ui re-ments previous exposure for the quarter exceeds (0.5)
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200 mrem References Procedure 6.1-024, Radiological Posting of Areas of the Station pg 6.1-024-4 Procedure 6.2-001, PNPS Radiation Exposure Control Program, pg 6.2-001 -4 and 5 K/A Plant-Wide Generics 294001 Kl.03 Knowledge of 10CFR20 and related facility radiation control req ui rements (3 3)
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4. PROCEDURES -
NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL
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CONTROL ANSWERS - PILGRIM -87/5/26 - ROBINSON ANSWER 4.07 (3.0) A condition e xi s t s wh i ch REQUIRES reactor scram and reactor power is cbove 3% (0,5)
or power cannot be determined (0.5)
, or all control rods a re rot inserted (0. 5)
past position 04 For selected rods the porition indicator will (0. 5)
provide (in digital di s play) the actual poaitio (A value greater than 04 Indicates incomplete insertion)
Indication lamps " Rods Fully inserted (Green) (0. 5)
I (for individual control rods)
Control rod drift alarm annunciate (0. 5)
Red " Rod Dri f t" indicating lamps will indicate the s peci fi c control rod 'x
%u Q pr: /su Ko o ~0 . gr he / M* 6 .%* k L f* *r' vsp W I't' 'U"* 4 References # M d 3'"' M '" *
Pom, ta n t n o + de es es s o m s Procedure E0P-02, RPV Control Power pgs E0P-2 and 3 K/A EPE 295037 Scram Condition Present and Reactor 4 I
Power above APRM Downscale or unknown EK 2.06 Knowledge of the interrelations between scram condition unknown and CRD mechanism (3.5)
System Generic 1 Ability to recognize abnormal indications for system operating parameters which .
I are entry-level conditions for emergency and abnormal operating conditions (4.4)
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e PROCEDURE - NORMAL, ABNORMAL. EMERGENCY AND RADIOLOGICAL
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PILGRIM -87/5/26 - G. E. ROBINSON ANSWER 4.08 (2.5) deg F (0. 5) If 90 deg F is exceeded, HPCI testing must be (0.5)
stopped deg F (0.5) E0P-4, Primary Containment Control Temperature (0.5)
Enter as soon as it was di s cove re d s uppres s i on (0.5) ;
pool temperature was above 80 deg l
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References l I
Procedure E0P-04 Primary Containment Control Temp pg 2 System Student Guides and Objectives, Book 4, Primary Containment System, pg LG-ll Tech Spec 3.7. K/A EPE 295026 Suppression Pool High Water Temperature System Generic 7 Ability to explain and apply all system limi ts and precautions (3.2)
System Generic 11 Ability to recognize a b r.o r ma l indications for system ope ra t i ng pa rame t e rs which are entry-level conditions for emergency and abnormal ope ra t i ng procedures (4.4)
K/A System 223001 Primary Containment System and Auxilaries System Generic 1 Knowledge of operator responsibilities during all modes of operation (3.8)
ANSWER 4.09 (1.5)
abberant initiations (0 75)
continued operation beyond automatic t ri p points (0.75)
References Miscellaneous Training Material, E0P Cautions pg B5-2 Procedure E0P-01, RPV Control, Level and Pressure K/A 295009 Low Reactor Vater Level System Gene ri c 12 Ability to utilize symptom based procedures (3.8)
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4. PROCEDURE - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL ANSWERS --
PILGRIM _- 87/5/26 - G. E. ROBINSON ANSWER 4.10 (1.0)
When any LPRM string has two or more LPRM's (1.0)
reading 110% or greater 1 References Procedure 2.4.13 Unexplained Rapid increase in Reactor Power pg 2.4.13-2 K/A APE 295006 SCRAM System Ge ne ri c 10 Ability to perform without reference to procedures those actions that req ui re immediate operation of system components or controls (4.1)
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ATTACHMENT 2 Facility Comments and NRC Resolutions of Comments on Written Examinations The following represents the facility comments and the NRC resolution to those comments made as a result of the current examination review polic Only those comments resulting .in significant changes to the master answer key, or those that were "not accepted" by the NRC, are listed and explained belo i Comments made that were insignificant in nature and resolved to the satisfac-tion of both the examiner and the licensee during the post examination review are not listed, i.e.: typo errors, relative acceptable terms, minor set point-change R0 EXAMINATION Question 1.01 FACILITY COMMENTS Request consideration for the following alternate answers:
1) Vessel drain line temperature 2) Moderator temperature 3) Recirculation pump suction temperature (it is the only direct indication available for moderator temperature.)
4) Reactor pressure 5) Vessel level REFERENCE PNPS Procedure 2.4.24, page 2, Procedure 2.4.25, page 2 and Procedure 2.1.7, page NRC RESOLUTION Comments Partially Accepte Recirculation pump suction temperature is not acceptable since pumps are not running. Therefore this temperature is not indicative of reactor water tem-peratur These alternate answers are not contained in the material referenced for development of this questio _ - _ _ _ - _ - _ - _ _ _ - _ - _ _ - _ _ - _ _ - _ - _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ - _ - _ - - _ - _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ . _ . _ - _ . _ _ - _
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. Attachment 2 2 l l
! Question 1.03 l'
FACILITY COMMENTS l
The answer key requires the candidate to state that the reduction in xenon con-centration is enough to cause shutdown margin (SDM) to go to zero without being given a graph of xenon concentration with respect to tim As stated on page 6-13 of the Reactor Theory Reference Text, xenon concentration changes over core life due to the difference in xenon fission yields for U-235 and Pu-239. Without a graph of xenon concentration verses time it would be dif-ficult for the candidate to determine that at " time equals 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />" 2 percent reactivity has been added by xenon decay resulting in a SDM of zer Consider- ,
ing this information, request consideration be given to allowing full credit for an answer stating that due to the reduction in xenon concentration, SDM will be reduce REFERENCE PNPS Reference Text on Reactor Theory, Page 6-1 I NRC RESOLUTION Comment Accepte Question 1.07 FACILITY COMMENTS Answer key states that Control Rod Worth would " increase" if core void fraction increases. It is our position that control rod worth will " decrease" as core void fraction increase The correct response for part "c" should be
" decrease".
REFERENCE PNPS Reference Text on Reactor Theory, page 5-1 NRC RESOLUTION Comment Accepte Question 1.08 FACILITY COMMENTS :
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Students are taught that the point of adding heat is reached somewhere between i 0.1% and 1.0% (depending upon moderator temperature at time of criticality). !
This corresponds to a thermal power level between 2.0 MW and 20 MW at Pilgri Since the power calculated in the first part falls between 2 and 20 MW, the i students may have concluded that the point of adding heat (P0AH), was or was not reached. Please consider the context of the students' answer when grading, i
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. Attachment 2 3 R !ERENCE PNPS Referance Text on Reactor Theory, page #7-1 NRC RESOLUTION Comment Accepte Question 2.03 FACILITY COMMENTS Recuest that " suction side of pump" be placed in parenthesis, and not required I
for full credi NRC RESOLUTION Comment Accepte Question 2;0S FACILITY COMMENTS Question may lead students to state other ways by which he may clo;a the MSIVs:
e.g. - control switch to close
- de-energize both RPS cabinets )
i Also, steam flow will assist in shutting the MSIV's '
REFERENCE PNPS Main Steam Student Guide LG-6 and NRC RESOLUTION Comment Accepte Question 2.06 FACII.ITY COMMENTS Last item should read " Loop B discharge valves closes. .." vice " Loop A dis-charge valve. ." Loop B is selected for injection if neither recirculation loop is brot' REFERENCE PNPS RHR Reference text page 13.
NRC RESOLUTIDf!
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- Attachment 2 4 Question 2.07 FACILITY COMMENTS Students may use the term " extraction valve" interchangeably with " bleeder trip valve". During normal full power operation the RFP SELECTIVE TRIP SELECTOR SWITCH ]
is in the ON position. Upon sensing a condensate pump trip, a correspond-ing feed pump will trip to prevent a low NPSH. (If the "RFP trip selector switch" is "0FF", the answer to part "b" would apply.)
REFERENCE PNPS Condensate and Feedwater Reference text, pages 16 and 17 l PNPS Procedure 2.2.96 page 1 NRC RESOLU110N Comments Accepte Terminology used in question development was based or, facility reference materia i Question 2.08 J FACILITY COMMENTS As alternate control room indications of a failed seal, please consider the following:
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Increase in leakage to floor or equipment sumps
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No. 2 seal pressure would approach No. 2 seal pressure
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Indication of steam leak in area of recirc pump (C-19 or C-85)
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Drywell pressure increasing
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Drywell humidity increasing (C-7)
- Drywell radiation levels increasing (C-19)
REFERENCE PNPS Recirc Student Guide, page LG-27; Procedure 2.4.22 and !
Recirc LG TP-9 (figure). l l
NRC RESOLUTION l Comments Partiaily Accepte _
Increase leakage to floor sump is not accepted because leakage would go to equipment sum _ _ _ _ _ _ _ _ _ _ _
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Attachment 2 5
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l The following were not accepted because they are indicative of both seals failin l
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Indication of steam leak in area of recirc. pump (C-19 or C-85) I
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Drywell pressure increasing
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Drywell humidity increasing (C-7)
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Drywell radiatior. levels increasing (C-19) i i
Question 2.09 I FACILITY COMMENTS In addition to " exceeding drywell external design pressure", please con-sider " preventing excessive drywell vacuum". For indication of a leak inside the drywell, request consideration of the following additional answers: j
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Drywell air cooler high drain flow (leakage) conditions
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Increase in drywell-to-torus d/p indication
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Recirc pump seal leakage condition REFERENCE PNPS Primary Containment Student Guide LG-14; and PNPS Proced-ure 2.4.1 NRC RESOLUTION Comment Accepte Part b. additional answers are not included in the facility l reference material used for question developmen i Question 2.10 l FACILITY COMMENTS Setpoint for low CST level should be 18" i
NRC RESOLUTION Commer.t Accepted.
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, Attachment 2 6 Question 3.0l FACILITY COMMENTS Answer could be either True or False. If a flow converter fails downscale a half scram will occur if reactor power is greater than 62%.
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REFERENCE PNPS APRM Reference Text, page 4 l
l NRC RESOLUTION
i Comment Accepte Part c. deleted since the question, as written, has no I
ability to determine knowledg Question 3.04 FACILITY COMMENTS . RFVE trip logic is "one out of two twice," with Channels A and C in one trip system and Channels B and D in the other trip syste A high radiation signal must be generated in both trip systems to get a tri The correct answer to a.ii . should be "no trip". Request consideration for the following alternate answer:
" Secondary Containment Isolation System is tripped (isolates reactor l building ventilation and Primary Containment Atmosphere Control ; stem),
Secondary Containment System is sometimes referred to as Reactor Building Isolation System (RBIS).
REFERENCE PNPS PRM Student Guide, page LG-7 PNPS Plant Ventilation Systems Student Guide, page LG-7 NRC RESOLUTION Comments Accepte Part b. alternate terminology is not provided in facility reference material used for question developmen Question 3.06 FACILITY COMMENTS For first part of answer, request that an explanation of " indicated steam flow" becoming less than " actual steam flow", be considered for full credit, instead of stating the 75% and 100% number NRC RESOLUTION Comment Accepted.
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Attachment 2 7
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Question 3.07b.
l I FACILITY COMMENTS RPS does not supply power to reactor building ventilation monitor RPS does supply power to refuel floor ventilation exhaust radiation monitors channels A and B. The correct answer would include refuel floor ventilation exhaust monitors channels A and REFERENCE PNPS PRM Reference Text, page 33 NRC RESOLUTION l Comment Accepte The facility reference material, from which this question was developed, is in erro Question 3.10 FACILITY COMMENTS l
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Procedure 2.1.6, Reactor Scram, page 3 states under the Operator Actions sec-tion that the operator should " VERIFY" or manually TRANSFER house locds to the Startup Transformer. This transfer is a result of the fast transfer featur The word verify implies that this is an automatic action that should occu Since the operators are responsible for the operator actions on a scram, and verifying transfer of house loads to the Startup Transformers is one of those actions, request that you consider as an additional answer to part "a."
" reactor scram."
Also, request that consideration be given to " simultaneous opening of ACB's 104 and 105" as an additional answer to part "a.". Tripping of these breakers
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simultaneously trips the Unit Auxiliary Transformer breaker Also, "4160 vac bus lockout" is an incorrect answer for part "a.". This was identified during a program audit. Students were retrained on this during the
"Needs Analysis Phase" of the program, and a Training Material Revision has been complete REFERENCE PNPS Procedure 2.1.6, page 3 PNPS AC Electrical Distribution Student Guide, page LG-9 and 10 NRC RESOLUTION Comment Accepte The facility reference material used for question develop-ment did not contain the above given alternate answer _ _ _ _ _ _ _ _ -
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- Attachment 2 8 question 4.01 FACILITY COMMENTS
Procedure 2.1.4 (Approach to Critical) is an integral part of rod pull on I startu Some students may interpret the question as asking the nLe_thod of performing a coupling check and answer per the directions of 2. Request this be considered as an alternate answer to 4.0 NRC RESOLUTION Comment Accepte Question 4.02 FACILITY COMMENTS The following are common usage terms at PNPS:
1) RPS MG Set Room is also referred to as the Vital MG Set Room, or just MG Sets Roo ) The 23' and 37' 4160V switchgear rooms are also referred to as upper and lower switchgear room, or 4160V breaker area .
l 3) The RPS Power Panels are referred to as C-511, or neutron monitoring !
panel ) Instrument Racks 2205 and/or 2206 are referred to as the 51' e i instrument rackt, pressure / level instrument racks, yarway racks, or RPS instrument rack ! Request consideration as an alternate answer to "one RFP when level indi-cation starts to increase," a statement which includes "at the direction of the supervisor" REFERENCE PNPS Procedure 2.4.143, page 8 NRC RESOLUTION Comments Accepte Part a. alternate terms are not contained in the facility reference material used for question developmen _______ -__ _____ _ _
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- Attachment 2 9 Question 4.03 ;
i FACILITY COMMENTS Consider use of terms Normal & Backup chargers for B-14 and B-1 b. and c. Common terminology at PNPS is to identify DC busses as either 125 VDC Bus A or Bus B, with the use of electrical prints to identify / locate specific leads (Objective 6 of DC Electrical Distribution Student Guide, page 3). Some students, therefore, may fail to identify D-17 as 125 VDC Bus If these stucients assumed D-17 was 125 VDC Bus A, then their answers should be as vcD ows: Recirc Pump A would be lost to cause the power decrease ) Diesel Generator A lost 2) Lost 3) Not Lost 4) Lost 5) Not Lost Please consider this as an alternate answe REFERENCE PNPS DC Electrical Distribution Student Guide, page LG-12 PNPS DC Electrical Distribution Student Guide, Objectives NRC RESOLUTION Comments Accepted. Grading will be based on which bus was assume ,
I NOTE: Part ) was deleted because it did not specify which diesel {
generato Point count was spread among other four parts of '
Question 4.04 FACILITY COMMENTS Students may have interpreted the question as asking for a list of items to be checked, as is discussed in the discussion section of 2.4.14 !
Please consider, as an alternate answer, statements to the effect of checking /
verifying switches in the safe (usually closed) positions and that initiation signals have been reset (the position - closed - of the isolation valves ;
energizes the relays in order for them to be reset).
REFERENCE PNPS Procedure 2.4.147, page 2 NRC RESOLUTION Comment Accepte l
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- - Attachment 2 10
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Question 4.05 l
FACILITY COMMENTS !
l b. and c. Consider including, as an alternate answer, a YES, based on the 10 CFR 20 limit of 3 Rem /qtr. with NRC Form 4 file and within cri-teria of 5 (N-18).
REFERENCE 10 CF". 20.101 (b)(1)(2)(3) l NRC RESOLUTION Comments not accepte The question is based on a specific dose rate and should be answered with that in mind, not answered in a generic (rote) manne Question 4.07b.
l FACILITY COMMENT Consider as additional answer, the following:
l 1) Computer printout of rod position (00-7)
2) Refuel Mode Select Permissive light illuminated with Mode Switch in Refuel Although these methods are not discussed in E0P-02, they are repeatedl stressed as valid methods at PNP '
REFERENCE PNPS Reactor Manual Control Reference text, page 7 PNPS Control Rod Drive Student Guide, TP 24 and 28 i NRC RESOLUTION Comment Accepte i l
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