ML20210C325
| ML20210C325 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 01/22/1987 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20210C309 | List: |
| References | |
| 50-458-OL-86-01, 50-458-OL-86-1, NUDOCS 8702090367 | |
| Download: ML20210C325 (151) | |
Text
2 DETAILS
- 1. Examination Results All 17 license candidates were administered written, simulator, and oral operating examinations. One R0 candidate failed the simulator examination and one R0 candidate failed one section of the written examination.
All nine operators examined during the Requalification Program audit were administered written and simulator examinations. One R0 failed the written examination.
- 2. Examiners D. N. Graves, NRC Chief Examiner J. L. Pellet, NRC T. L. Morgan, EG&G Idaho, Inc.
K. M. Spencer, EG&G Idaho, Inc.
J. B. Sherman, EG&G Idaho, Inc.
W. C. Cliff, Battelle G. S. Sly, Battelle
- 3. Examination Report Performance results for individual candidates are not included in this report. This report is comprised of the sections listed below:
- a. Written Examination Review Coment Resolution In general, editorial comments or changes made during the examination or subsequent grading reviews are not addressed in this section. A copy of the facility's review coments is attached. If a coment is not addressed in this section, it has been accepted and incorporated into the examination answer key. Coments and resolutions are listed by section question number.
Licensing exam: 1.10 Requalification exam: 1.07 5.10 5.07 Resolution: Question retained. The question objective is for the candidate to understand the local / average flux relationship and core flux conditions at 40 percent rod density vs. 90 percent rod density.
Licensing exam: 3.06 Requalification exam: 3.04 6.08 esolution: Question retained. The operating procedures do not 8702 67 870122 PDR JCK 05000458 V PDR
-*)
3 T
prohibit Flux Auto operation. The control system '
response to the event is considered pertinent information by the chief examiner.
Licensing exam: 3.12a Requalification exam: 6.07a g 6.10a 3.07a x
-Resolution: Division III answer will lose credit. The referenced procedure specifically states that emphasis should be placed on using Division I equipment..
1 b. Exit Meeting Summary i An exit meeting was held on November 21, 1986,'to conclude the site visit. The following personnel were present at the meeting: -
NRC Licensee (Gulf States Utilities)
J. L.. Pellet, Examiner D. L. Andrews I D. D. Chamberlain, SRI D. J. Ashley W. B. Jones, RI W. H. Odell D. W. Williamson
) The following observations were made to the licensee:
(1) Escorted access to the facility and vital areas inside the facility was inconvenient and time consuming during the first week of examinations. The situation was much improved during the second week and posed no problem.
\
(2) Candidates, in general, were weak on details of fire protection (types of extinguishers, use of extinguishers, details of Halon, etc.).
(3) Candidates had difficulty using prints, especially electrical prints.
(4) Candidates found key control confusing and cumbersome.
i (5) .Several procedure errors were found. These were pointed out to the licensee at the exit meeting.
(6) Several simulator malfunction discussions in the malfunction book referenced items that were not available to the examiners.
~
(7) Results would be forthcoming as soon as possible but, per NUREG-1021, no preliminary results would be given now, i
l (Items 8-10 were noted to the training staff at the end of the first l week of examinations.)
4 (8) Surveillance procedures in the simulator were out of date or did not reflect how the simulator performs.
(9) The simulator did not respond as expected on several occasions.
The simulator operator had a list of what was noted as simulator i*
problems noted during the examinations.
(10) The Licensed Operator Training Manuals are very simplistic in the description of systems. Determining how a system is going to respond to various input conditions or malfunctions is very difficult, if not impossible, due to the lack of detail. ,
- c. Examination Master Copies Copies of the R0, SRO, requalifications R0, and requalification SRO examinations plus answer keys are attached.
i
- d. Requalification Program Evaluation Report ,
A copy of the requalification program evaluation report is attached.
i 5
4 4
0 i
l
EXAMINATION REVIEW COMMENTS Lie: 1.09 b Req: 1.06 b An equally correct answer is:
"In approximately 20 days, when equilibrium Samarium is reached."
Reference:
LOTM-TH-4.21, pg. 4 Lie: 1.10 Req: 1.07 5.10 5.07 Recommend deleting question for the following reasons:
The question asks to compare rod worths for withdrawing a central rod at 90%
vs. 40% rod density.
The reference in the answer key is LOTM-TH-4.20-1, pg. 4 The reference material compares the worth of withdrawing a central control rod with all other rods inserted vs. inserting a central rod with all other rods withdrawn or mostly withdrawn.
The two conditions do not match up. Fourty-percent rod density is not having all control rods withdrawn or mostly withdrawn and inserting vs. withdrawing with l one condition being 40% rod density and the other at 0% rod density makes the situation difficult to analyze, Withdrawing a central control rod worth all rods inserted can be viewed as an out-of-sequence withdrawal.
r i
i l
l
Likewise, withdrawing a center control rod at 40% rod density could be considered as an out of sequence withdrawal.
However, the insertion of a center control rod when all control rods are withdrawn is not an out of sequence maneuver.
According to the GE Station Nuclear Engineer's Manual:
"Thus, control rod withdrawal will increase pitch, increase flux, and increase thermal diffusion length. From the equation (H-1), these three factors will be competing and control rod worth will either increase or decrease depending on the relative magnitudes of the variables involved."
See Figure H-7 in the SNE Manual which illustrates the drastic effect on rod i worth due to sequence errors.
I The analysis is further complicated by the change in rod worth due to withdrawal of adjacent control rods.
Having adjacent control rods withdrawn tends to increase the local flux.
i This is one of the reasons a control rod drop accident is worst at 50% rod density.
t
References:
NEDO-24810B Vol.1, pg. H-2, H-17/H-18 River Bend FSAR, Chapter 15 i
I l
l
.. Lie: ' 2.03 Reg: 2.03 6.01 6.01 Another acceptable answer for the flow control valve position response is:
" Flow control valve opens (0.5) to maintain total system flow constant (0.5)."
l The second sentence correctly states drive water differential pressure will increase. Since reactor pressure is essentially constant, this means the' pressure i upstream of the pressure control valve increased. This is the pressure downstream of the flow control valve. Therefore, the differential pressure 4 across the flow control valve has decreased. Since the flow controller will position the flow control valve as necessary to maintain the set flowrate, the valve will have to open, i.e. lower it's head loss, to allow a constant flow rate with a lower differential pressure across it.' This can be demonstrated in either the simulator or the plant by throttling the pressure control valve shut, i
However, the answer provided in the key is also acceptable in terms of perceptible indication of CRD FCV position in the Control Room.
References:
LOTM-TH-1.1 and 1.2, Fluid Flow Concepts i
)
i i
t k
i l
i I
i i
i
--- -.n-,. - ,,--n.,n-- , - . . . , . . . - - , . - - - . - ~ , . - . - - - - - - - , ,,--,-,,-..,s. - - - -.., --- - - . - - , - - - .
Lle: 2.06 e Req: - 2.05 e )
l 6.02 e 6.02 e ,
l Answer should include as acceptable for full credit:
"All modes are still functionalif only one of the valve's two logic circuits is placed in test."
A separate test switch exists for each of the two logic circuits for an SRV.
Placing one of the switches in its test position disables its associated logic, but leaves the SRV fully functional through its other logic channet Placing both switches in test is not allowed nor covered by test procedures at River Bend.
Recommend question be reworded taking into account the dual channel configuration of the SRV logic.
l
Reference:
LOTM-7-1, pg. 5 and Figure 10 (Attached with applicable portions highlighted) i l
l i
Lie: 2.10 a Req: 2.08 a 6.05 a 6.04 a Reference material, LOTM Ch 20-0 pg. 9, discusses the F048 valve only in reference to other valves associated with suppression pool cooling. As referenced, it is the only RIIR valve that cannot be overridden. Ilowever, the question refers to the entire RIIR System.
Therefore any other valve that can't be immediately overridden with a initiation signal present should be acceptable. The following valves will not stroke with a LOCA signal present:
E12-F02G A RIIR Hx flow to RCIC E12-F011 A/B RIIR IIx flush E12-F021 RIIR C pump test return E12-F052 A/B RIIR Stm Supply Shutoff E12-F051 A/B Steamline press reducing valve E12-F087 A/B RIIR IIx Stm Supply BP valve The following valves maybe repositioned, but return to their original position when the C/S is released:
E12-F048 A/B Hx Bypass E12-F027 A/B Inlet Shutoff E12-F003 A/B Hx Outlet Recommend the answer key be revised to accept any one of the above listed valves.
l
Reference:
LOTM-20-0, Valve logic tables, pg. 54-56 l
~
l l
l
.=_ - ._ _-
Lie: 2.11 a The associated emergency diesel switchgears should also be acceptable as specific power supplies to each Standby Service Water pump following a loss of offsite.
Pump Power Supply 2A ENS *SWG1 A 2B ENS *SWG1B 2C E22*S004 2D ENS *SWG1B
References:
LOTM-22b, pg. 26, and LOTM-43, pg.1 i
Lie: 2.12 a Level 1,- 143 inches should also be acceptable as a LOCA signal. The question doesn't specify a particular LOCA signal
References:
LOTM-IC-6, pg.19, and LOTM-18, pg.1 Lie: 3.02 An additional control function provided by the Turbine Stop Valve position is part of the low condenser vacuum bypass logic contained in the MSIV isolation logic.
Reference:
LOTM-lC-12, pg.15 i
m - -, __ ____.______ __ _ . _ . .
__. . _ _ , ~ , _ . . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . -
Lie: 3.03 Req: 3.02 An additional correct initiation signalis " Low system flow at 465 gpm." LOTM-11 has not yet been updated to include this feature of the plant's generator runback circuit.
Reference:
GEK 75521 (Attached with applicable sections highlighted)
Lie: 3.06 Req: 3.04 6.08 Recommend deleting required response for Recirculation Flow Control System because plant does not allow this operating configuration, operator aid posted on P680 prohibits operation in " flux auto" mode of recire flow control
Reference:
Operator Aid for Recire Flow Controls 1
Lie: 3.08 Req: 3.06 Recommend this question be deleted.
i i The question states that there is one exception to the requirement for deliberate 1
i operate actio.n to reopen isolation valves after a reset. This statement is incorrect as there are no exceptions.
The MSIV's will not automatically reopen upon reset. A deliberate action must be taken to place each MSIV control switch to the CLOSE position before the logic can be reset and the valves reopened.
Referenced LOTM discussion is incorrect. The logic drawing in LOTM-12 (Figure 6, Page 76) indicates that the MSIVs must be closed. The logic drawing is correct.
Reference:
AOP-0003, Rev. 2, pg. 3 (Attached with applicable section highlighted)
Lic: 3.09 Reference is incorrect and will be revised. RPS logic for the Main Steam Line (MSL) Radiation Monitors is the same as most other parameters inputted to RPS. MSL monitors A, B, C, and D cause trips on high-high (three times normal full power background) in RPS channel A, B, C, and D respectively.
Included are Applicable Process Radiation Monitoring and Reactor Protection System GE elementary drawings establishing the relationship above.
Therefore correct answers are:
"a, c, d, and f"
References:
GE DWG 828E243AA, sheets 5 and 16 GE DWG 828E531 AA, sheets 5 and 9 (Attached with applicable areas highlighted)
LOTM-11, pg.1 Lic: 3.10 b Request " ADS Timer Reset pushbuttons," or " Reset ADS Timer", or " ADS Reset" be acceptable as alternative answers for " ADS Timer / Level 3 Seal-in Reset pushbuttons."
Reference:
Same as exam question l
I
! Lic: 3.12 a Reg: 6.07 a l 6.10 a 3.07 a Request credit not be lost for including Division III along with Division I. AOP-0031 states Div. III is assumed operable and in procedure steps prescribes setup and use.
i
Reference:
AOP-0031, Rev. 3, pg. 3,12, and 15 (Attached with applicable steps highlighted) i l
L
Lie: 4.06 a Req: 4.04 a Another correct answer, (if RO-2A is assumed to be the instrument used) would be four times the difference in meter readouts. This is due to required beta correction factor applicable to this instrument. RPP-0089 on operation and calibration of this instrument states the correction factor.
Answer in key is also correct for other detectors. Applicable RPP sections provided describing how to obtain beta readings from E-520 and the 6112B teletector. Both are consistent with key.
References:
RPP-0082, 0089, 0091 (Attached with applicable steps highlighted)
Lic: 5.09 e Request "Never samarium free after power operation" as an acceptable correct response fully justifying the incorrectncss of the statement given in the question.
Reference:
LOTM-21 Lic: 6.09 a Answer should include " Containment Ilydrogen" as a parameter monitored by Post Accident Monitoring System Hydrogen analyzers monitor both containment and drywell separately.
Reference:
River Bend Tech Specs, Table 3.3.7.5-1, Item 8, pg. 3/4 3-82
Lie: 6.11 b For second part of answer, may receive response according to SOP-0053, which states, "The reinstatement of all trips following a manual emergency start requires the diesel to be shutdown and returned to standby." Recommend this also be acceptable for full credit as second part of the answer.
Reference:
SOP-0053, Rev. 4, pg. 4 and 21 (Attached with applicable sections highlighted)
Lie: 8.05 e ADM-0031 has a Temporary Change Notice (TCN 86-1283) currently in effect which provides other additional correct information on maximum time allowed for a temporary alteration. A discussion of these with or without the key answer should be satisfactory for full credit.
Reference:
TCN 86-1283 (Attached with applicable sections highlighted)
Lie: 8.08 a Req: 8.04 a Request allowing partial credit for responses which state "upon declaration of ALERT or greater", as this includes SITE AREA and GENERAL EMERGENCY.
Reference:
Same as exam question i
U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _B1VEB_ BEND _1____________
REACTOR TYPE: _RWB-QEh_________________
DATE ADMINISTERED: _akl112Qd________________
EXAMINER: _QB6 VEST _Qt______________
CANDIDATE: _________________________
INSIBURIl0NH_IQ_R8NQlRaIE1 Use separate paper for the answers. Write answers on one side only.
Stcple 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 locst 80%. , Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY
__V6LVE_ _IQIeL ___EQQBE___ _Y6LME__ ______________Q6IEQQBX_____________
_2EtDQ__ _j$.QQ ___________ ________ 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 2EzQQ__ _23200_ ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
_2dzQQ__ _J$ Q1 ___________ ________ 3. INSTRUMENTS AND CONTROLS
_25t0Q__ _15 Q1 ___________ ________ 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL
< _QQtRQ__ ___________ ________% Totals Final Grade i
All work done on this examination is my own. I have neither given nor received aid.
Candidate's Signature
. . _ - - . . - . - ~ - - _ -. . . . .
t i
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS.
During the administration of this examination the following rules apply:
- 1. Cheating on'the examination means an automatic denial of your application and could result in more severe penalties.
4
'2. Restroom trips-are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination
- room to avoid even the appearance or possibility of cheating.
3.- Use black ink or dark pencil 2DlY to facilitate legible reproductions.
4
- 4. Print your name in the blank provided on the cover sheet of the examination.
I
- 5. Fill in the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers.
- 7. Print your name in_the upper right-hand corner of the first page of Raah section of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a ngw page, write 2DlY 2D 2D2 11d2 of the paper, and write "Last Page" on the last answer sheet.
i.
- 9. Number each answer as to category and number, for example, 1.4, 6.3.
- 10. Skip at least 1bEgg lines between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
, 12. Use abbreviations only if they are commonly used in-facility litgtstyng.
- 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 enswer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND 00 NOT LEAVE ANY ANSWER BLANK.
3
- 16. If parts of the examination are not clear as to intent, ask questions of the gggm1Det only.
i
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.
~ , _ - . - . - - - . . -
- 18. When you' complete your examination, you shall:
- c. Assemble your examination as follows: '
(1) Exam questions on top.
(2) Exam aids - figures, tables, etc.
(3) Answer pages including figures which are part of the answer.
- b. Turn in your copy of the examination and all-pages used to answer the examination questions.
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.
- d. Leave the examination-area, as defined by the examiner. If after
.i leaving, you are found in this area wnile the examination is still in progress, your license may be denied or revoked.
4 l
l t
l l
i
(
lt__EBINCIELE1_9E_NUGLEaB_EQWEB_EL6SI_9EEB611981 PAGE 2 IBEBdQQIU801G1t_HE61_IBoNSEEB_eUQ_ELUIQ_ELQW QUESTION 1.01 (2.00)
M0tch each of the four lettered items with one of the numbered items. A letter-number sequence is suf f icient. (2.0)
- 1. MAPRAT 5. PCIOMR
- 3. CPR 7. PEAKING FACTOR
- 4. FLPD 8. LHGR
_____a. Parameter by which plastic strain and deformation are limited to less than 1%.
_____b. Ratio of bundle power required to produce onset of transition boiling somewhere in the bundle to actual bundle power.
_____c. Parameter by which peak clad temperature is maintained less than 2200 degrees F during postulated design basis accident.
_____d. Accounts for the non-uniformity of power distribution in the Core.
QUESTION 1.02 (1.50)
Indicate HOW each of the following changes will affect reactor CRITICAL POWER (INCREASE or DECREASE). If CRITICAL POWER will not be affected, state this. (1.5)
- e. Loss of extraction steam to a feedwater heater
- b. Mass flow rate through the core is increased
- c. Reactor pressure is increased QUESTION 1.03 (2.00)
Using the steam tables, indicate whether water at each of the following is SUBC00 LED, SATURATED, or SUPERHEATED. (2.0)
- c. 200 psig, 387.7 F
- b. 1000 psig, 544.6 F
- c. 1200 psig, 603.9 F
- d. 900 psig, 531.1 F
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
- _. .- ..- . ~ . . .. . .. . . . --
b li__EBINQ1 ELE 1_QE_NUGLEeB_EQWEB_EL6HI_QEEB8IlQNt PAGE 3 4
IBEBdQQ1N851 Git _BE8I_IB8NSEEB_8ND_ELUID_ELQW 4
QUESTION 1.04 (2.00) ;
Answer the following with regard to centrifugal pump operation.
- a. How does increasing the pump's speed affect the REQUIRED Net Positive l Suction Head (NPSH) for the pump? Answer INCREASE, DECREASE, or l UNCHANGED. (0.5)
- b. If the discharge valve for the pump is moved in the close direction, how will each of the following change? Answer INCREASE, DECREASE, or UNCHANGED. (1.5) i i 1. System flow
- 2. motor amps
- 3. pump discharge head i .r
- QUESTION 1.05 (2.50) i ' Answer the following with regard to the Control Cell Core (CCC) operating
! strategy *
}~
- a. What constitutes a CONTROL CELL? (0.5)
- b. While operating at power, what is the position of the non-control 2 cell centrol rods? (0,5)
- c. Why do control rod movements result in a lower kw/ft. change per notch in a CCC'than in a conventional core? (1.0)
! d. TRUE or FALSE. Using the CCC operating strategy eliminates
! the need for control rod pattern exchanges. (0.5) r-i l -
I r
- l i
- (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****) ,
i
- , ~ . , n-n-,-, - . ,_,,. _ _ ,- -,, - ,,- ,, - c ,._, -.,,,,.,.,,,.-,,.-,...,n,-_,,---,-,.._,-----,,,,-_--.
l
~ 11__EBINQ1 ELE 1_QE_NUQLE88_EQWEB_EL8MI_DEEB811QNi PAGE 4
-IBEBdQDIN8510Ht_BE8I_IB8NSEEB_8ND_ELVIQ_ELQW QUESTION 1.06 (2.00)
For each of the following, choose the correct word in parentheses: (2.0)
- o. The moderator temperature coef ficient. is more positive at (HIGH/ LOW) moderator temperatures.
- b. The moderator temperature coefficient is (LESS/MORE) negative at'EOL.
- c. The Doppler coefficient tends to be (LESS/MORE) negative at high fuel temperatures, and (LESS/MORE) negative at high moderator temperatures.
- d. The Doppler coefficient is more negative at (BOL/EOL).
! o. The void coefficient is (LESS/MORE) negative at high void fractions.
- f. The void coefficient is (LESS/MORE) negative at high fuel temperatures.
- g. The void coefficient is most negative near (BOL/E0L).
QUESTION 1.07 (1.50)
Describe (INCREASE, DECREASE, OR UNCHANGED) the effect on void fraction, doppler reactivity and feedwater enthalpy due to a recirculation flow i rate increase at power. (1.5) i t
4
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
-li__EBINQIELE1_QE_NVQLE88_EQWEB_EL8NI_QEEB8I1QNi PAGE' 5 IBEBdQQ1Ned10ft_BE8I_IB8NSEEB_8ND_ELulD_ELQW QUESTION 1.08 (1.00)
The reactor trips from full power, equilibrium xenon conditions.
Four (4) hours later the reactor is brought critical and power level in-maintained on range 5 of the IRMs for several hours. Which of the following statements is CORRECT concerning control rod motion during this period? (1.0)
- c. Rods will have to be rapidly inserted since the critical reactor will cause a high rate of xenon. burnout.
- b. Rods will have to be withdrawn due to xenon build-in.
- c. Rods will have to be inserted since xenon will closely follow
' its normal decay rate.
d.. Rods will.approximately remain as is as the xenon establishes its equilibrium value for this power level.
QUESTION 1.09 (2.00)
'The reactor is started up after a refueling outage. Rods are pulled to the 100% line and power is then increased to 100% with recirculation flow.
After approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, reactor power has decreased to about 98%.
Aasume no operator action.
i L a. What is the primary cause for this reduction in power? (1.0)
- b. When would you expect the power decrease to stop and WHY does
- it stop? (1.0)
I-t
. QUESTION 1.10 (1.50)
. _Concerning control rod worth, compare withdrawing a center control rod at 90% rod density to withdrawing a center control rod at 40% rod density. In which situation is control rod worth greater for the withdrawn control rod? Explain your answer. (1.5) f i
l
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
It__EBINQ1 ELE 1_QE_NVQLE68_EQWEB_EL6NI_QEE8611QNt PAGE 6 IBEBdQQ1N8dIQ1t_BE61_IB8NSEEB_8NQ_ELVIQ_ELQW QUESTION 1.11 (2.00)
'HOW and -WHY will the Shutdown Margin j ust prior to a refueling outage compare with the Shutdown Margin following the refueling? Two (2) t rossons required. (2.0)
QUESTION 1.12 (2.S0)
- a. Briefly explain each of the following terms relating to neutrons. (1.S)
- 1. Beta (B) 2.- Beta Core
- 3. -Importance Factor (I)
- b. Write a mathematical expression to show the relationship of the
- above items to BETA EFFECTIVE. (1.0)
QUESTION- 1.13 ( . 50)
Which of the following neutron sources is NOT considered an INTRINSIC neutron source: (0.S) i
- n. curium 242
- b. antimony-beryllium
- c. gamma-deuterium (photo-neutron)
- d. alpha-oxygen l
4 t
t l
l l
I
)1
(***** CATEGORY C1 CONTINUED ON NEXT PAGE *****)
- -, ,n .- g.v.,,., ,-,.+-,c ,- - , . - , - - - - - - , -, ,-w .-----,.----..>-,,r .--.-------e- , - - , - - - - , - - - ~
'It__EBINGIELES_QE_NVQLE88_EQWEB EL8NI_QEEBellDNA PAGE ~ 7.
~IBEBdQQXNedICSt_BE6I_IB8NSEEB_8NQ_ELUID_ELQW
, QUESTION- 1.14. (2.00)
Une a'1/M' plot and PREDICT the number of control r o'ds required to'be withdrawn to achieve criticality.
NOTES: 1. 'CR = Count Rate
- 2. USE.THE FIGURE AT THE END OF THE SECTION AND INSERT IN YOUR ANSWER PACKAGE CR0 = 40 cps CR4 = 191' cps CR1 = 50 cps. CRS = 333 cps CR2 =.89 cps CR6 = 800 cps CR3 = 129 cps Each CR reading is recorded following a 5 rod withdrawal with CR0 being 100% rod density.
5 10 15 20 25 30 35 40 45 50 55 1.0-----l----l----l----l----l----l----l----l----l----l----l----1.0 l 0.9- -0.9 0.8- -0.8 I 0.7- -0.7 3
1/M 0.6- -0.6 0.5- -0.5 0.4- -0.4 0.3- -0.3
- 0 .' 2 - -0.2 0.1- -0.1 0.01----l----l----l----l----l----l----l----l----l----l----l----0.0 0 5 10 15 20 25 30 35 40 45 50 55 Control Rods Withdrawn i
5
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
It__EBINQIELES_QE_NUQLE68_EQWEB_EL6NI_QEE8611QN t PAGE 8 IBEBdQQIU6dIQSt_ME61_IB6NSEEB_6ND_ELUIQ_ELQW QUESTION 1.15 C .00)
Une a 1/M plot and_ PREDICT the number of control rods required to be withdrawn to achieve criticality.
NOTES: 1. CR = Count Rate
- 2. USE THIS FIGURE AT THE END OF THE SECTION AND INSERT IN YOUR ANSWER PACKAGE CR0 = 40 cps CR4 = 191 cps CR1 = 50 cps CRS = 333 cps CR2 = 89 cps CR6 = 800 cps CR3 = 129 cps
-Ecch CR reading is recorded following a 5 rod withdrawal with CR0 being 100% rod density.
5 10 15 20 25 30 35 40 45 50 55 1.0-----l----l----l----l----i----l----l----l----l----l----l----1.0 0.9- -0.9 0.8- -0.8 0.7- -0.7 1/M 0.6- -0.6 u0.5- -0.5 0.4- -0.4 0.3- -0.3 0.2- -0.2 0.1- -0.1 0.01----l----l----l----l----l----l----l----l----l----l----l----0.0 0 5 10 15 20 25 30 35 40 45 50 55 Control Rods Withdrawn
(***** END OF CATEGORY 01 *****)
n 2t__EL8NI_QE11GN_INGLVQ1NQ_16EEII_6NQ_EMEBGENGl_111IEd1 PAGE g
-QUESTION 2.01 (1.50)
The CRD Pressure Equalizing valves (C11-PEV-F150A and B) operate to
-maintain or minimize the dp between the CRD exhaust header and the cooling water' header. DESCRIBE the sequence of conditions / events in the CRD system that requires these valves to actuate / operate. (1.5)
QUESTION 2.02 (2.00)
- a. Immediately following a r*.7ctor scram, the CRDH flow indicator reads offscale high (>100 gpm). Is this a normal indication for this condition? If so, explain why flow indicates high. If not, give TWO (2) possible reasons for the flow to be high. (1.0)
- b. Not all of the CRD pump discharge flow is directed through the
. flow-element. What three (3) pump discharge paths are NOT directed
-through the flow element? (1.0)
- -QUESTION 2.03 (2.00)
A single control rod'is inserted one notch. The insert stabilizing valve (s) f ail' to reposition following the rod movement. How does this af f ect DRIVE WATER DIFFERENTI AL PRESGURE and FLOW CONTROL VALVE position? EXPLAIN YOUR ANSWER. (2.0)
QUESTION 2.04 (2.00)
! -List three_(3) interlocks specifically designed to limit or preclude cavitation of the recirculation pumps, jet pumps, or flow control valves.
Include initiating parameters, setpoints, and any action that may occur cm a result of that interlock. (2.0)-
QUESTION 2.05 (1.50)
- a. What provides the normal and backup sources of air to the SRV's? (1.0)
- b. Is the changeover from normal to backup source of air a manual or automatic evolution? (0.5)
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
2t__EL6HI_ DES 10N_ INCLUDING _HaEEII_6UQ_EdEBGENGl_11SIEd1 PAGE 10 QUESTION 2.06 (2.00)
For each SRV switch position / condition below, state which mode of SRV operation will still function (use SAFETY, PRESSURE RELIEF, ADS, and LOW LOW SET as the modes). Assume all modes of valve operation apply to the valve in question. (2.0)
- c. OFF
- b. AUTO
- c. The valve's logic circuit is in TEST QUESTION 2.07 (1.50)
The reactor is operating at 95% power. Condensate pump A is out of cervice. (pumps B and C operating). The Condensate Recirculation Velve (Short Cycle Cleanup to Condenser FCV114) fails OPEN. EXPLAIN the effect this will have on reactor feed flow. (1.5)
QUESTION 2.08 (1.50)
Describe how the failure of a feedwater heater extraction steam non-return valve to shut on a turbine trip can result in over-speeding the turbine. (1.5)
QUESTION 2.09 (2.50)
Indicate whether each of the following valves in the HPCS system receive cn OPEN signal, a CLOSE signal, or N0 signal upon automatic start of the cystem. (2.5)
- o. F001 Condensate Storage Tank (CST) Suction Valve
- b. F015 Suppression Pool Suction Valve
- c. F010 Test Valve to CST
- d. F023 Test Valve to Suopression Pool
- e. F004 Injection Valve
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
2t__EL8dI_QESIQU_IUCLVQ1NQ_16EEIY_8UQ_EUEBQEUQ1_SYSIEUS PAGE 11 QUESTION 2.10 (2.50)
- e. Following a LOCA, what is the only valve in the RHR system that cannot be manually overridden IMMEDIATELY following LPCI initiation? (0.5)
- b. What are two (2) indications that a RHR Injection Valve (F042A/B/C) has been manually overridden? (1.0)
- c. What conditions / actions must be met /taken to clear the manual override? (1.0)
QUESTION 2.11 (3.00)
- c. What is(are) the power supply (Les) to the four SSW pumps (A - 0) following a loss of offsite power? Be specific for each pump. (1.0)
- b. For each of the following SSW loads, indicate whether it is supplied by LOOP A ONLY, LOOP B ONLY, OR EITHER LOOPS of SSW. (2.0)
- 1. Emergency _ supply to RHR System
- 2. Backup supply to fuel building fire protection
- 3. RHR Heat Exchanger 10
- 4. Drywell unit coolers QUESTION 2.12 (3.00)
- c. What is a LOCA signal (include setpoints)? (1.0)
- b. Following a LOCA signal, which loads on RPCCW are isolated, AND which loads are maintained? (2.0)
(***** END OF CATEGORY 02 *****)
2&__INSIBudENIS_aug_CgNIBgLS PAGE 12 QUESTION 3.01 (2.00)
There are four (4) arm and depress isolation pushbuttons (A - 0) on P680 for the MSIV's. Which pushbutton combinations below will effect a full MSIV isolation? (2.0)
- c. AB
- b. AC
- c. AD
- d. BC
- o. 80
- f. CD QUESTION 3.02 (1.00)
What two (2) control / protective functions utilize Turbine Stop Valve position as an input? (1.0)
QUESTION 3.03 (2.50)
Describe the response of the turbine generator to a degradation in the etator water cooling supply to the main generator. INCLUDE ANY APPLICABLE SETPOINTS in your discussion. (2.5)
QUESTION 3.04 (2.00)
- c. The RCIC flow transmitter has failed low (indicates 0 gpm). Select
, the answer below which best describes how the RCIC turbine will j respond upon receipt of an initiation signal. (1.0)
- 1. The turbine will trip on overspeed
- 2. The turbine will run continuously at minimum (idle) speed
- 3. The turbine will run continuously at approximately 4500 rpm
- 4. The system will not start (F045 Steam Supply remains closed)
- b. Which RCIC Isolation Division (1 or 2) is provided with a manual
- isolation feature? (0.5) l c. When is the manual isolation in part "b" above functional? (0.5) l
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
k
2i__INSIBudENI1_eND_QQNIBQL1 PAGE 13 1
QUESTION 3.05 (3.00) a.. What rod movement restraints are in force as a result of exceeding the LPSP and HPSP during power escalation? (1.0)
'b. At what power levels do the LPSP and HPSP take effect? (1.0)-
- c. Why is control rod withdrawal undesirable if power is above the LPSP and a Main Turbine Bypass Valve is open? (1.0) 1 QUESTION 3.06 (2.50)
The reactor is operating at 100% power.- Recirculation flow control
. io in MASTER MANUAL. The Reactor Operator mistakenly depresses the l
[ DECREASE load button on the EHC panel and holds it. Describe the response of the turbine control and recirculation flow control systems to this event. Terminate the event with a reactor scram, including the cause. (2.5)
QUESTION 3.07 (3.00)
., State whether each of the following parameters directly initiate SCRAMS, i ROD BLOCKS, BOTH, or NEITHER. Setpoints are NOT required. (3.0)
- a. Scram discharge volume level
- b. MSIV position 4
- c. Main steam line high flow j d. Reactor vessel high level
- e. Recirculation flow
- f. Reactor pressure ,
r
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
'2i_i1NSIBudENI1_eNQ_QQNIBQL1 PAGE -14 QUESTION 3.08 (2.00) a Which combination (s) of. Main Steam Line Radiation Monitors / Trip Channels bolow will initiate a REACTOR SCRAM if tripped? -(2.0)
- e. AB
- b. AC
- o. BD
- f. CD 1
QUESTION 3.09 (2.00)
- a. Which-condition (s) necessary for ADS automatic initiation must be present1to manually initiate ADS (per the initiation logic)? (1.0)
- b. How can the ADS logic be reset once it has completely actuated? (1.0)
QUESTION 3.10 (2.00)
List ten (10) parameters that are monitored as part of the Post Accident Monitoring System. (2.5)
QUESTION 3.11 (2.00)
- c. During a control room fire, efforts should be focused on using Division (I, II, III) equipment to put the station in a safe shut-down condition. (0.5)
- b. How is control of a component transferred to the Remote Shutdown Panel from the control room? (0.5)
- c. What are two (2) indications in the control room that would indicate that transfer of a valve's control to the Remote Shutdown Panel had been performed? (1.0)
(***** END OF CATEGORY 03 *****)
di__EBQQEQUBES_:_NQBd8Lt_8BNQBd8Lt_EdEBGENQ1_6NQ PAGE 15 88010LQQ196L_QQNIBQL QUESTION <4. 01 (1.50)
Answer the following'with regard to GOP-0001, Plant Startup:
- a. When is the reactor determined to be critical? (1.0)
- b. Control rods are withdrawn until the reactor is critical and a steady, reasonable period is achieved. Per the referenced procedure, what is a steady, reasonable period? (0.5)
QUESTION 4.02 (3.00)
Mctch the following events (a-f) with the approximate pressure at which they should be performed per GOP-0001, Plant Startup to Low Power Alarm Point. The pressures may be used more than once or not at all. (3.0)
- a. Place steam seals in service 1. 20 psig
- b. Startup the Offgas System _ 2. 100 psig
- c. Line up RCIC CStandby mode) 3. 150 psig
- d. Startup the first reactor feed pump 4. 250 psig
- o. Start turbine shell warming 5. 400 psig
- f. Startup the SJAE 6. 450 psig
- 7. 600 psig
- 8. 850 psig 1
N l
t w
' % 4
's
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
i 4t__EBQQEQMBEQ_:_UQBUalt_8HUQBd8Lt_EUEBQEUQ1_6UQ PAGE 16 bod 19LQQ1 gel _QQUIBQL QUESTION 4.03 (2.50)
Given the following cooldown data, DETERMINE if the RPV cooldown rote limit has been exceeded. STATE the RPV cooldown limit and HOW the determination was made. IF the limit was exceeded, STATE the period (s) of time the limit was exceeded. (2.5)
TIME TEMPERATURE (Deg F)
START 0600 520 s 0815 490 0830 460 0845 445 0900 430 0915 400
~'
0930 370 /
0945 340
, 1000 325 1015 -
310 1030 275 1045 ~
250 1100 225 1115 205 r
QUESTION 4.04 (2.50) _
'Briefly state the reason for each of the following statements pOrtaining to a reactor / plant shutdown from power:
- c. The turbine generator must remain on the line until' Reactor Thermal Power is below 10%. (0.5)
- b. Maintain steam packing exhauster and air removal pumps in operation for up to four (4) hours after breaking. vacuum. (0.5)
- c. Small control ro"d movements are essential when operating with the reactor critical and the MSIV's closed. (0.5)
- d. sThe main turbine must be on the turning gear whenever steam seals are on or c. u r b i n e metal tempe,ratures are above 150 deg F. (0,5)
- o. Do not allow RPV pressure to decrease to O psig with the MSIV's open and condenser vacuum > or = 20 Hg. (0.5)
A k s 4
(***** CATEGORY 04 CONTINUED ON NEXT,PAGE *****)
'at__EBQQEQVBEl_=_NQBU6Li_6BNQBd6Lt_EdEBEENQ1_6ND PAGE 17 86DIQLQQlC6L_CQNIBQL QUESTION 4.05 (1.50). l What do-the words "shall", "should", and "may" denote when they are uced in River Bend procedures? (1.5)
QUESTION 4.06 (1.50)
An irradiated component is surveyed with a-portable instrument. The open window reading is 2.65 Rem /hr and the closed window reading is 1.75 Rem /hr. SHOW ALL WORK!
- e. What is the beta dose rate at the instrument? ( 0. 5 ) ~
- b. How long could you remain at the survey distance without exceeding administrative and 10CFR20 quarterly exposure limits? (1.0)
QUESTION 4.07 (2.50)
- e. What two (2) general limitations or limiting plant parameters may restrict or inhibit removal of feed water heaters from service? (0.5)
- b. Explain how a reactivity transient may result from fluctuating 5th or'6th point heater levels. (1.0)
- c. Why must the operator exercise caution when shutting down the Heater Orain' Pumps ? (1.0)
QUESTION 4.08 (1.50)
Instrument Air header pressure is decreasing. All systems are re-sponding properly to the decreasing air pressure. Besides keeping the Shift Supervisor informed of plant conditions, what actions should be taken by the' control room operator (s) and at what point (if applicable) por AOP - 0008, Loss of Instrument Air? (1.5)
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
i-8t__EEQQEQM8El_ _NQ858kt_8ENQBd8kt_EME8QENQY_8NQ :PAGE 18 88010LQQlD8L_QQUIBQL-
. QUESTION 4.09 (1.00)
What two (2) conditions must be met so that an IMMEDIATE reactor ceram is NOT required during a loss of Normal Service Water? (1. 0 )'
QUESTION 4'.10 (2.00)
- c. The loss of one RPS bus. requires system isolations to be reset.
Which system isolation must be reset within a very short period of time to prec'.ude requiring further operator action? C0.5)
- b. What length of time is a "very short period of time" per A0P-0010, I
-Loss of One RPS Bus? (0.5)
- c. What action must be taken if the isolation mentioned in part a above is not1 reset within the specified time? (1.0)
QUESTION _ 4.11 (2.50)
- c. With the facility at power, what actions should be taken prior to evacuating the control room? (2.0)
- b. Where should the control room personnel assemble following the evacuation?. (0.5)
QUESTION 4.12 '(3.00)
For each of the following conditions, state which E0PCs), if any, should be applicable. (3.0)
[
- e. Reactor-water level 12.5 "
- b. Drywell pressure 1.87 psig
- c. Suppression pool temperature 97 deg F c
- d. Drywell to containment dp 27.3 psid
- e. Containment temperature 88 deg F
(***** END OF CATEGORY 04 *****)
(*************
END OF EXAMINATION
- )
EQUATI0tt SHEET f = ma v = s/t Cycle efficiency = (Net work out)/(Energy in) 2 w = og s = V,t + 1/2 at I
E=K -
2 KE = 1/2 av a = (Vf - V,)/t A = AM A=A,e*
PE = agn Vf = V, + at -
w = e/t a = an2/t1/2 = 0.693/t1/2 y , 3p
,o 2 tify - msg A= , [(g1/2) * (*bI3 aE = 931 an m = V,yA , ,
-Ex Q = aCpat 6 = UAaT I=I,e*
- Pwr = Wfah I = I ,10-* N L TVL = 1.3/u P = P 10sur(t) HVL = -0.693/v p = p et /T o
SUR = 26.06/T SCR = S/(1 - K,ff) l CR, = S/(1 - K,ff,)
l SUR = 26s/s* + (s - p)T CRj (1 - K ,ffj) = G2 II - Ieff2)
T = ( t*/s ) + [(B - 8 Y I83 N
- IIII ~ Keff) = CRj /3 ,
T = s/(p - s) M * (I - Keffo)/II-Eeff1)
T = (B - 8)/(I8) SDM = (1 - K,ff)/Keff a = (K ,ff-1)/K ,ff = AKeff M eff t* = second 1 = 0.1 seconds _
p = ((1*/(T K,ff)] + [s,ff /(1 + IT)]
! Ij dj = I d Id 2 ,g gd 2 P = (14V)/(3 x 1010) jj 22 2 I = oN R/hr = (0.5 CE)/d (meters) i R/hr = 6 CE/d2 (feet) ,
Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lbm. I curie = 3.7 x 1010dps 1 ga;. = 3.78 liters 1 kg = 2.21 lbm 1 ft* = 7.48 gal. 1 np = 2.54 x 103 Stu/nr Density = 62.4 lbm/ft3 1 nw = 3.41 x 10o Stu/hr i Density = 1 gm/c:n3 lin = 2.54 cm Heat of vaporization = 970 Stu/lDm 'F = 9/5'C + 32 Heat of fusion = 144 Bru/lbm 'C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. Hg. 1 8TU = 778 ft-lbf 1 ft. H 2O = 0.4335 lbf/in.
l 1
11__EBINGIELE1_QE_NuGLEoB_EQWEB_ELeNI_9EEB6119Nt PAGE 19 IBEBdQQ1Ned191t_BEoI_IBoNSEEB_8ND_ELulD_EL9W ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 1.01 (2.00)
- a. 8
- b. 3
- c. 2
- d. 7 (0.5 each)
REFERENCE LOTM-TH-3.8-0, Critical Power, pg 3 LOTM-TH-3.9-0, LHGR, pg 2, 5 LOTM-TH-3.10-0, APLHGR, pg 1 ANSWER 1.02 (1.50)
- c. CRITICAL POWER increases (0.50).
- b. CRITICAL POWER increases (0.5).
- c. CRITICAL POWER decreases (0.5).
REFERENCE LOTM-TH-3.7-0, Transition Boiling, pg 3 ANSWER 1.03 (2.00)
- c. Saturated
- b. Subcooled
- c. Superheated
- d. Subcooled a (0.5 each)
REFERENCE Steam Tables
11__EBINGIELER_QE_N9GLEaB_EQWEB_ELaNI_QEEBaIIQUt PAGF 20 IBEBdQQ1NadIGEt_BEaI_IBaNSEEB_aND_ELVIQ_ELQW ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 1.04 (2.00)
- c. increase (0.5)
- b. 1. decrease (0,5)
- 2. decrease (0,5)
- 3. increase (0.5)
REFERENCE LOTM-TH-1.4-0, Pumps, pg 3, 17 ANSWER 1.05 (2.50)
- c. Four fuel bundles and a control rod (0.5)
- b. Fully withdrawn (0,5)
- c. All control rod movements are associated with low reactivity cells (1.0).
- d. True (0.5)
REFERENCE LOTM-2-1, Nuclear Fuel, pg 20-23 ANSWER 1.06 (2.00)
- c. Low
- b. Less
- c. Less, More
- d. EOL
- o. More
- f. More 9 BOL (0.25 each)
REFERENCE LOTM-TH-4.17 through 4.19 l
It__EBINCIELER_9E_NUGLEeB_E9 WEB _EL8HI_QEEBoII9Nt PAGE 21 IBEBdQQINed1GSt _BEeI_IBeNSEEB_eND_ELUIQ_EL9W ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 1.07 (1.50)
Void fraction decreases (0,5)
Doppler reactivity increases (0.5)
Enthalpy increases (0.5)
REFERENCE LOTM-TH-4.18-1, Void Coefficient LOTM-TH-4.19-1, Doppler Effect LOTM-14b-0, Feedwater Heaters and Heater Drains ANSWER 1.08 (1.00) b (1.0)
REFERENCE LOTM-TH-4.22-0, Xenon ANSWER 1.09 (2.00)
- c. As the reactor operates at power, Xenon builds in to equilibrium (0.5),
adding negative reactivity, causing power to decrease (0.5).
- b. In 40-50 hours (0.5) when equilibrium Xenon is reached (0.5).
Also accepted approximately 20 days (0.5) when equilibrium Samarium is reached (0,5). One point total.
REFERENCE LOTM-TH-4.22-1, Xenon o
"It__EBINQIELE1_QE_NUQLE88_EQWEB_EL6NI_QEEB6119Nt .P AGE- 22
-IHEBdQQINadIQ1t_HE8I_IB6NSEEB_6NQ_ELVID_ELQW LANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
1 ANSWER '.1.10 (1.50)'
Withdrawal'of a center control rod at 90% density'has greater worth.(0.5)-
The control rod worth'is proportionel-to the (local neutron flux / the core average neutron flux) squared.CO.25)
With 90% rod density.the core average-neutron flux is very small.
Withdrawing a central control rod, increases the local flux in the area
, of withdrawn rod substantially. Because the rod causes the value of the
. term'Clocal neutron flux / core average neutron flux) squared to be large
! :its worth for this condition is quite high.
~
Higher than withdrawing the rod at 40% rod density, when core average flux will be higher. (0.75)
REFERENCE LOTM-TH-4.20-1, Control Rod Worth, pg 4 i
I ANSWER 1.11 (2.00)
I
[ SDM prior to the outage will be larger (0.5) due to fission product
- poisoning (0.75) and fuel depletion (0.75).
. REFERENCE i LOTM-TH-4.11-0, Shutdown Margin, pg 5 ANSWER 1.12 (2.50) 9
- c. 1. BETA: fraction of neutrons born from a particular isotope that are born delayed (0.5) 4
- 2. BETA CORE: Weighted average of the betas of all the different fuels in the core (0,5)
- 3. IMPORTANCE FACTOR: Indicates the relative probability of delayed neutrons causing fission to prompt
! neutrons causing fission (0.5) t b. BETA EFFECTIVE = (B CORE) X (I) = BETA EFFECTIVE (1.0)
REFERENCE LOTM-TH-4.14-1, Prompt and Delayed Neutrons
it__EBINQ1ELES_QE_NuGLE88 EQWEB_ELeNI_QEEBoI1QNA- PAGE 23
.IBEBdQQ1NedIQ1t_ME6I_IBeNSEEB eNQ_ELulD_ELQW ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
l ANSWER 1.13 ( .50) b (0.5)
REFERENCE LOTM-TH-4.12-1, Neutron Sources-ANSWER 1.14 (2.00) l Drawing a straight line between the last two *'s predicts 34-35 control rods must be withdrawn.
(0.1 for each point plotted, 1.3 for line and prediction) 5 10 15 20 25 30 35 40 45 50 55 1.0*----I----I----I----1----1----I----1----I----1----I----I----1.0 0.9- -0.9
! 0.8- * -0.8 0.7- -0.7 4 _ -
1/M 0.6- -0.6 I
0.5- -0.5 l - * -
O.4- -0.4 t 0.3- * -0.3 0.2- * -0.2 l -
i 0.1- * -0.1 0.01----l----l----l----l----l----l----l----l----l----l----l----0.0 0 5 10 15 20 25 30 35 40 45 50 55 Control Rods REFERENCE LOTM-TH-4.13.1, Subcritical Multiplication and Count Rate Comparison ANSWER 1.15 ( .00)
NOT APPLICABLE
-2i__ELeNI_QE11EN_INCLVQINQ_18EEI1_8NQ_EMEBGENGl_HISIEMS PAGE 24
. ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 2.01 .(1.50)
On a scram, the' exhaust water header. will be at a greater pressure than the scram discharge header (0,5). This causes the exhaust water hoader to depressurize to the scram discharge header (0.5) via the HCOs' 121 valves (0.5). The Pressure Equalizing Valves quickly repressurize the exhaust header from the cooling water header.
' REFERENCE
.LOTM-4-1, CRDH System, pg 8-9 ANSWER 2.02 (2.00) e.- yes (0.5). CRD system flow increases as the scram accumuletors recharge (0.5). If examinee answers NO, allow 0.25 for each of two possible causes of high flow.
- b. CR0 pump miniflow (0.4)
Recirc pump seal flow (0.4)
Sample flow (0.2)
REFERENCE LOTM-4-1, CRDH System, pg 4 ANSWER 2.03 (2.00)
J Flow control valve position does not change (0.5) because total system flow has not changed (0.5). Drive water differential pressure will increase (0.5) because flow previously bypassing the pressure control
. valve is now-passing through it (0.5).
Also accepted that the FCV opens slightly to maintain flow due to slightly ;
[
increased backpressure.
REFERENCE LOTM-4-1, CRDH System, pg 16-17 r
f I'
t 2t__EL8NI_DE11GN_INCLUDIN9_18EEI1_6NQ_EdEBGENQ1_111IEd1 PAGE 25 ANSWERS - RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 2.04 (2.00)
-<30% (0.2) total feedwater flow (0.2) for 15 seconds (0.2)
- dt <8 deg F (0.2) for 15 seconds (0.2) between the main steam line (0.2) and the recirculation pump suction (0.2) 1
- Reactor level (0.2) < or = level 3 (0.2)
- Any of the above will initiate a transfer to low speed (0.2)
- REFERENCE
_LOTM-5-1, Recirculation System, pg 12, 20
- - ANSWER 2.05 (1.50)
- a. normal: SRV air compressors (0.5) backup: Penetration Valve Leakage Control System Air Compressors (0.5)
- b. automatic (0.5)
REFERENCE LOTM-7-1, Main Steam, pg 4, 41 i
ANSWER 2.06 (2.00)
- a. Safety, ADS
- b. Safety, Pressure Relief, ADS, Low Low Set
- c. Safety (accept ALL if noted that only 1 logic circuit is in test)
(0.285 each, 2.0 total)
REFERENCE
- LOTM-7-1, Main Steam, pg 5-7, 16 ANSWER- 2.07 (1.50) i Suction pressure to the RFP will decrease (0.5). Reactor feed flow will decrease (0.5) due to RFP trip on low suction pressure or the reduction in RFP discharge pressure (0.5 accept either).
REFERENCE LOTM-12-1, Condensate System, pg 5, 41 Simulator Malfunction 85b i
r-, ~ . , - . . , ~ . , - - . , . _ , - . . . ,,_,-_,,__...<--_-,__.-,m.-._, , _ , _ - . _ , , . . - _ . . . -.._,__m.. . . _ , , , . . = , _ _ . _ , , , . . - - - - ,
2t__EL6NI_QE110N_INGLUDING_18EEIl_6NQ_EMEBGENQ1_11SIEd3 PAGE 26 ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 2.08 (1.50)
When the turbine trips, pressure in the turbine decreases to condenser pressure (0.5). The condensed extraction steam flashes to steam (0.5),
and backflows up the extraction steam line through the turbine (0.5), which con cause it to overspeed.
REFERENCE LOTM-14b-1, Feedwater Heaters, Vents, and Drains, pg 17 ANSWER 2.09 (2.50)
- c. Open
- b. No
- c. Close
- d. Close
- o. Open (0.5 each)
REFERENCE LOTM-18-0, HPCS System, pg 30-31 ANSWER 2.10 (2.50)
- c. RHR HX Bypass valve (0.5). Other valves were acceptable as alternate answers. See the valve logic tables in the stated reference.
- b. - amber light above the valve switch (0.5)
- annunciator (RHR Inj ection Valve in Man Override) (0.5)
- c. - LOCA signal clear (0.5)
- Initiation Reset pushbutton(s) depressed (0.5)
REFERENCE LOTM-20-0, RHR System, pg 9
2i_iEL6NI_QE11EN_INCLUQ1NG_HeEEI1_8NQ_EMEBGEN91_1111Ed1 PAGE 27
-ANSWERS -_ RIVER BEND 1 -86/11/04-GRAVES' , D.
i ANSWER. 2.11 .C3.00)
B - D/G B or ENS *SWG1B-(either acceptable) 4 C -~HPCS D/G or E22*S004 (either acceptable)
D - D/G B or ENS *SWG18-(either acceptable)
, (3.25 each)
- b. 1. loop 8 "
i 2. either loop
- 3. loop A
- 4. either loop
'(0.5 each) 3
-REFERENCE-
- LOTM-22b-0, SSW system, pg 2-3, 24 1
i ANSWER 2.12 (3.00)
- a. 1.68 psig (0.25) drywell pressure (0.25) j reactor vessel-low water level (0.25) of level 2 (0.25)
Also accepted level 1 low water level.
.! Recirculation pumps l DW equipment drain sump cooler i
Maintained: RHR pumps A and B Fuel pool coolers RWCU pumps ,
CRD pumps (8 at 0.25 each) i-REFERENCE LOTM-23-0, RPCCW, pg 6 l
o r
i i
l l
[
i
li__IN11ByMENI1_6NQ_QQNIBQL1 PAGE 28 ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 3.01 (2.00) c,:c, d ,f (0.5 each)
REFERENCE LOTM-7-1, Main Steam, pg 17 ANSWER 3.02 (1.00)
E00 RPT (0.5)
Also accept low condenser vacuum bypass logic.
REFERENCE LOTM-5-1, Recirculation System, pg 33 LOTM-IC-11-0,-Reactor-Protection System, pg 12 LOTM-IC-12-0, Containment and Reactor Vessel Isolation Control System, ,
pg 15 ANSWER 3.03 (2.50)
A turbine runback will occur (0.5) at 13 psig stator inlet pressure (0.25),
81 deg C stator outlet-temperature (0.25), or 465 gpm low flow. If generator load is >80% after 2 minutes (0.25) or >24% (0.25) after 3.5 minutes-(0.25), .the turbine trips (0.5). Low flow was accepted, but l not required, due to recent modification.
i.
REFERENCE
- LOTM-11-1, Main Generator Auxiliaries, pg 3, 10-11
! GEK 75521 i
i ANSWER 3.04 (2.00) l l- e. 3 (1.0) i
- b. 1 (0.5)
- c. initiation signal must be present (0.5)
[ REFERENCE LOTM-21-0, RCIC System, pg 7, 13, 37, 38 i
i I
SA__IN11BudENIS 6NQ_QQNIBQLH PAGE 29
' ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER- 3.05 (3.00)
- c. LPSP: Rod withdrawal is limited to four notches (0.5) before having to deselect and reselect a rod.
HPSP:: Rod withdrawal is limited to two notches (0.5) before having to deselect and reselect a rod.
- b. LPSP: 27.5% power or 157.6 ps ig first stage turbine pressure (0.5)
HPSP: 62.5% power or 410.7 psig first stage turbine pressure (0,5)
Accept + or - 1%-power and + or - 5 psig pressure
- c. Reactor power-would be higher than that indicated by turbine first stage pressure. Rod movement could be performed that could conceivably exceed a fuel safety limit (1.0).
REFERENCE LOTM-IC-7-0, RC & IS, pg 31, 47 River Bend Technical Specifications Bases, 3.1.4.1 ANSWER 3.06 (2.50)
As the load set motor runs down, the EHC system will be calling for the control valves to close to reduce load, which they will do (0.5).
As the CV close, the turbine bypass valves will begin to open to meintain pressure (0.5). Once the BPV are fully open and the CV are still shutting, reactor pressure and power will begin to increase (0.5) cousing the recirculation flow control valves to close down in an
- ettempt to keep reactor power at 100% (0.5). This continues until the reactor scrams on high pressure or power (0.5)
REFERENCE l LOTM-IC-8-0, Recirculation Flow Control, pg 15, 23
!. LOTM-IC-9-0, EHC, pg 85-86, 114 i
k k
2t__INSIBUDEN11_6NQ_QQUIBQLS PAGE 30 ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 3.07 (3.00)
- c. both
- n. scram
- c. neither
- d. scram
- o. rod block
- f. scram (0.5 each)
REFERENCE LOTM-IC-11-0, RPS, pg 16 LOTM-IC-4-0, APRM, pg 8 River Bend Technical Specifications, Table 3.3.6-2 ANSWER 3.08 (2.00) c, c, d, f (0.5 each)
REFERENCE LOTM-IC-12-0, Containment and Reactor Vessel Isolation Control System, pg 9 ANSWER 3.09 (2.00)
- c. Low pressure ECCS pump running (1.0)
- b. Depressing the ADS TIMER / LEVEL 3 SEAL-IN RESET pushbuttons (1.0)
Other names acceptable for the pushbuttons, i.e. timer reset.
REFERENCE LOTM-IC-14-0, ADS, pg 4, 5 l
l
- - _ . . - _ . . - . - - _ ~ - __ . _ . . . . , _ _ ._ - . - - . . .-
2t__INSIBMUENIH_86Q_QQNIBQLS PAGE 31 ANSWERS -- RIVER BEND 1 -86/ll/04-GRAVES, D.
ANSWER 3.10 (2.00)
Containment /Drywell dp Conteinment pressure Containment radiation Containment temperature Containment hydrogen Drywell Hydrogen Drywell pressure Drywell temperature Drywell radiation Suppression pool temperature Suppression pool level Roactor water level Roactor pressure (10 required at 0.2 each)
REFERENCE LOTM-IC-23-0, Post Accident Monitoring, pg 2-4 ANSWER 3.11 (2.00)
- c. I (0.5)
- b. By taking the Remote Shutdown Transfer Switches to the EMERGENCY position (0.5).
- c. Annunciator (s) in the control room (0.5)
Loss of position indication for that valve in the control room (0.5).
REFERENCE LOTM-IC-16-0, Remote Shutdown, pg 7-8
st__880CEQUBE3_:_NQ808kt_8BUQ806Lt_EUE8GENQ1_6NQ PAGE 32 86DIQLQQlCoL_CQUIBQL ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 4.01 (1.50)
- c. The reactor is critical when the neutron count rate increases (0.5) at an exponential rate without further control rod motion (0.5).
- b. Accept 50 - 150 seconds (0.5)
REFERENCE GOP-0001, Plant Startup, Rev 6, pg 20 ANSWER 4.02 (3.00)
- c. 3 or 150 psig
- b. 4 or 250 psig
- c. 2 or 100 psig or 3 or 150 psig
- d. 5 or 400 psig
- o. 2 or 100 psig
- f. 4 or 250 psig (0.5 each)
REFERENCE GOP-0001, Plant Startup, Rev 6, pg 22-24 ANSWER 4.03 (2.50)
The cooldown rate limit is < or = 100 deg F (0.25) in any one hour portod (0.25). The limit was exceeded (0.5) for three one hour poriods: 0845 - 0945 (105 deg F), 0900 - 1000 (105 deg F), and 1015 - 1115 (105 deg F). (3 periods of time at 0.5 each)
REFERENCE River Bend Technical Specification 3.4.6.1 i
i t-
31__EBQQEQWBEl_ _NQBd6Lt_6HNQBd6Lt_EMEBQESQX_6NQ -PAGE 33 B6DIQLQQ196L_QQNIBQL
-ANSWERS'-- RIVER BEND 1 -86/11/04-GRAVES, D.
' ANSWER 4.04 (2.50).
- e. ' Turbine BPV.are only rated at 10% total steam flow. Will also accept to prevent a high pressure scram. (0.5)
- b. High airborne activity in the turbine building can' result (0.5) if vacuum is broken within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from operation at full power.
- c. There is no automatic pressure control (0.5).
- d. To prevent warping the shaft (0.5).
- o. Prevent flashing the reference legs (0.5) on the RPV instruments.
REFERENCE River Bend GOP 0002, Power Decrease / Plant Shutdown, Rev 3,.pg 8-10 ANSWER 4.05 (1.50) chall - denotes a requirement (0.5)
.chould - denotes a recommendation (0.5) may - denotes permission (0.5)
- REFERENCE l River' Bend ADM 0003, Development, Use, and Control of Procedures, Rev 10, pg 5 i
, ANSWER 4.06 (1.50)
- a. Beta dose rate = open - closed window = 0.9 Rem /hr -(0.5)
Also accept 3.6 if beta factor of 4 used, i
( b. Stay time: 10CFR20: 1.25 RC1mts)/1.75 R/hr = .71 hr = 43 minutes (0.5)
[ Admin.: 1.00 RC1mts)/1.75 R/hr = .57 hr = 34 minutes (0.5) i REFERENCE l GET II, p. 4
}
i r
l l
dt__E8QQEQUBEl_:_NQBd8(t_8BNQBd8(t_EdEBGEUQ1_6NQ PAGE 34 88Q19LQQ196L_GQUIBQL ANSWERS -- RIVER BEND 1 -86/ll/04-GRAVES, D.
ANSWER 4.07 (2.50)
- c. MCPR or Turbine limits (0.5)
- b. Extreme high levels in the heaters will auto bypass the condensate stream around the heaters (0.5) causing feed water temperature to decrease and change reactor power (0.5).
- c. Flow through the demineralizers must increase as pumped forward flow is reduced (0.5). Subsequent pressure oscillations may cause feed pump trip (0.5).
REFERENCE GOP-0002, Power Decrease / Plant Shutdown, Rev 3, pg 8-9 ANSWER 4.08 (1.50)
When instrument air header pressure has decreased to 65 psig (0.5) or it bocomes apparent that instrument air will not be restored (0.5), scram the roactor (0.5).
REFERENCE River Bend A0P - 0008, Loss of Instrument Air, Rev 2, pg 3 ANSWER 4.09 (1.00)
- The main turbine is NOT on the line (0.5).
- Turbine BPV are NOT being used to control reactor pressure (0.5).
REFERENCE River Bend A0P 0009, Loss of Normal Service Water, Rev 2, pg 4 ANSWER 4.10 (2.00)
- o. RPCCW (0.5)
- b. One minute (0.5)
- c. Trip both recirculation pumps (1.0)
)
it _iE B Q Q E Q W B E H _:._N Q Bd e k t _8 HN Q B d a k t _ E d E B Q E N QY _6N Q PAGE 35 B8010LQQIQaL_QQNIBQL ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
REFERENCE River Bend AOP 0010, Loss of One RPS Bus, Rev 2, pg 2,4 ANSWER 4.11 (2.50)
- c. 1. Arm and depress all four manual scram buttons.
- 2. Mode Switch to SHUTDOWN.
- 3. Verify control rods fully inserted.
- 5. Close the-MSIV's (5 at 0.4 each)
- b. . Division I Remote Shutdown Panel (0.5)
REFERENCE A0P-0031,-Shutdown from Outside the Control Room, Rev 3, pg 6 ANSWER 4.12 (3.00) j c. none
- E0P-0002, Primary Containment Control
- c. E0P-0002, Primary Containment Control
- o. none
, (6 at 0.5 each)
REFERENCE River Bend E0P's 0001, Rev 6; 0002, Rev 5; 0005, Rev 3 i
U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _BlVEB_QENQ_1____________
REACTOR TYPE: _QMB-QE6_________________
DATE ADMINISTERED: _QhtlllQd________________
EXAMINER: _QB6MEkt_Qt______________
CANDIDATE: _________________________
IN5189CIl0NE_IQ_GaNQ1QoIE1 Uso separate paper for the answers. Write answers on one side only.
Stcple 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 locst 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY
__V8LUE_ _IQI6L ___hQQBE___ _VaLUE__ ______________GoIEQQBl_____________
_25tDQ__ _2510Q ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 25tQQ__ _251QQ ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION
_25tQQ__ _25tQQ ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL
_25tDQ__ _25tDQ ___________ ________ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 1QQtQQ__ ___________ ________% Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.
Candidate's Signature
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
- 1. Cheating on the examination means an au t on.a t ic denial of your application and could result in more severe penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil 90lY to facilitate legible reproductions.
- 4. Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of gggb section of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a ogw page, write QDlY QD 2DR Ridt of the paper, and write "Last Page" on the last answer sheet.
- 9. Number each answer as to category and number, for example, 1.4, 6.3.
- 10. Skip at least ibtgg lines between eacn answer.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
i
- 12. Use abbreviations only if they are commonly used in facility 111gtgigtg.
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND 00 NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the SERT 1Det only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.
L _ _ _ _ _ _ _ _ _ _ . _, . _ _ _ , _
- 18. When you complete your examination, you shall:
- c. Assemble your examination as follows:
(1) Exam questions on top.
(2) Exam aids - figures, tables, etc.
(3) Answer pages including figures which are part of the answer.
- b. Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions,
- d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revcked.
5t__IBEQBl_QE_NVQLE68_EQWEB_ELaNI_QEEBoIIQNt_ELu1 dst _8NQ PAGE 2 IBEBb001N8b1C1 QUESTION 5.01 (2.00)
Match each of the four lettered items with one of the numbered items.
A letter-number sequence is sufficient. (2.0)
- 1. MAPRAT 5. PCIOMR
- 3. CPR 7. PEAKING FACTOR
- 4. FLPD 8. LHGR
_____a. Parameter by which plastic strain and deformation are limited to less than 1%.
_____b. Accounts for non-uniformity in core power generation
_____d. LHGR over LHGR limit QUESTION 5.02 (3.00)
Describe HOW and WHY each of the following changes will affect reactor CRITICAL POWER. If CRITICAL POWER will not be affected, otate this.
- a. Loss of extraction steam to a feedwater heater (1.0)
- b. Mass flow rate through the core is increased (1.0)
- c. Reactor pressure is increased (1.0)
(***** CATEGORY 05 CONTINUE 0 ON NEXT PAGE *****)
l l
l 1
l Si__IBEQBI_QE_UUQLE88_EQUEB_EL6HI_QEEB8IIQNt_ELUIDSt_6NQ PAGE 3 j IBEBdQQ188d101 l
l I
l QUESTION 5.03 (2.00) j A reactor water sample indicates the following isotopes present:
Tritium (1H3) Kr 87
'Mn 56 I 131 Ni 57 Xe 133
, Fe 59 Cs 137 Co 60 Ba 140
- c. Which of the above are fission products and which are corrosion products? (1.0)
- b. Does the presence of fission products in the coolant mean a i fuel element failure or defect exists? If not, account for l their presence in the reactor coolant. (1.0) J QUESTION 5.04 (2.00)
Answer the following with regard to centrifugal pump operation. j
- o. How does increasing the pump's speed affect the REQUIRED Net Positive Suction Head (NPSH) for the pump? Answer INCREASE, DECREASE, or UNCHANGED. (0.5)
- b. If the discharge valve for the pump is moved in the close direction, how will each of the following change? Answer INCREASE, DECREASE, I or UNCHANGED. (1.5) !
1
- 1. System flow -
- 2. motor amps l l
- 3. pump discharge head l
l 4
l 1
l l
t l
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) l l
it__INEQBX_QE_NWQLE68_EQWEB_EL8HI_QEE86IIQNt_ELVIQ1t_68Q PAGE 4
-IBEBdQDINed1G1 QUESTION 5.05 (1.00)
ARRANGE the following in order of INCREASING heat transfer coef-ficient (Lowest Coefficient to Highest Coefficient). (1.0)
- o. Free Convection, Air
- b. Boiling Water (Free Convection)
- c. Boiling Water (Forced Convection)
- d. Forced Convection, Water
- o. Forced Convection, Air QUESTION 5.06 (1.00)
A reactor heat balance was performed (by hand') during the midnight chift due to the Process Comouter being 000. The OAF's were computed, but the APRM OAIN ADJUSTMENTS HAVE NOT BEEN MADE.
Which of the following statements is TRUE concerning reactor power? (1.0)
SELECT ONLY ONE ANSWER Conly one is truel)
- c. If the feedwater temperature used in the heat balance calcu-lation was LOWER than the actual feedwater temperature, then the actual power is HIGHER than the currently calculated power.
- b. If the reactor recirculation pump heat input used in the heat balance calculation was OMITTED, then the actual power is LOWER than the currently calculated power.
- c. If the steam flow used in the heat balance calculation was LOWER than the actual steam flow, then the actual power is LOWER than the currently calculated power.
- d. If the RWCU return temperature used in the heat balance cal-culation was HIGHER than the actual RWCU return temperature, then the actual power is LOWER than the currently calculated power.
(***** CATEGORY 05 CONTINUE 0 ON NEXT PAGF *****)
'\
k 5t__IBEQB1_QE_NVQLE88_EQWEB_EL8BI_QEEB8I1QNt_ELVIQSt_86Q ,
PAGE 5 IHEBdQQ1NedIQS QUESTION 5.07 (2.50)
Answer the following with regard to the Control Cell Core (CCC) operating otrategy:
- o. What constitutes a CONTROL CELL 7 (0.5)
- b. While operating at power, what is the position of the non control cell control rods? . . (0.5)
- c. Why do control rod movements result in a lower kw/ft chande; per notch in a CCC than in a conventional core? (1.0)
- d. TRUE or FALSE. Using the CCC operating strategy eliminates the need for control rod pattern exchanges. ' (0.5)
QUESTION 5.08 (3.00)
Ooscribe (INCREASE, DECREASE, OR UNCHANGED) the effect on void fraction, doppler reactivity and feedwater enthalpy due to a recirculation flow rote increase at power. EXPLAIN WHY for each change. (3.0)
QUESTION 5.09 (2.50)
Indicate whether each of the following is TRUE or FALSE. If FALSE, explain WHY it in FALSE. (2.5)
- o. Xenon and Samarium concentrations increase following a scram from high power operation (within the first five hours). ,
- b. Even though Samarium has a higher microscopic absorption cross section than Xenon, Samarium is not as significant an operating concern as Xenon. ,
- c. A reactor startup several days after a scram from extended high power operation is considered to be Xenon and Samarium free.
- d. The equilibrium concentration of Samarium at 50% power is approximately the some as at 100% power. ,
- c. The equilibrium concentration of Xenon at 50% pow *r is approximately one-half the equilibrium concentration at 100% power.
(***** CATEGORY 05 CONTINUE 0 ON NEXT PAGE *****)
7 . .#
q, X . r. . ,
4 2_IHEQBl_QE_NuGLE88_EQWEB_EL8NI_QEE86I1QNt_ELW1Qh_8NQ 'PAGE 6 IBEBtl00186t!101 .,
s s
~
Qd5STION 5.10 / C 1. 5 0.)
s Concerning control rod worth, compare withdrawing a center control rod at 90% rod density to withdrawing a center control rod'at-40% rod density. In which situation is control rod worth greater for the-withdraw'n control rod? Explain your answer. .(1.5) j '
. e. -
'fA . m QUESTION 5.11 (2.50)
- o. Define SHU[DOWN MARGIN per the Technical Specifications. (1.0).
- b. The required minimum value for-Shutdown Margin per the Technical Specifications.is one of two values. What are these two values
, and.what determines which value to use? .(1.5) w ,
s
'# ss Y
/ g O',
f..
5 i
4 L
- 7. - ;,
M' l ,
- s. .,
l
= 4
- 4 h .4
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
-11__IHEQE1_QE_ NUCLE 88_EQWEB_EL8HI_QEEB8Il0Nt_ELUIDSt_8NQ PAGE. 7
.IBEBdQQ188d1GS 1
l
?
v JQp QUESTION 5.12 (2.00)
I
'Une a 1/M plot and PREDICT the number of control rods required to be i withdrawn to achieve criticality.
7 j
NOTES: 1. 'CR'= Count Rate 1
- 2. USE THE FIGURE AT THE END OF THE SECTION AND INSERT IN YOUR l
- ANSWER PACKAGE j i
CR0 = 40 cps ~CR4 = 191 cps i CR1 = 50 cps CR5 = 333 cps CR2 = 89 cps CR6 = 800 cos CR3 = 129 cps ?,
Eoch CR reading'is recorded following a 5 rod withdrawal with CR0 being 100% rod density.
5 10- 15 20 25 30~ 35 40 45 5 0 -' 55 l 1.0-----l----l----l----l----l----l----l----l----l----l----l----1.0 0.9- -0.9
- i
'I i 0.8- -0.8 0.7- -0.7 1/M 0.6- -0.6 0.5- ,
-0.5 0.4- ,
, -0.4 0.3- -0.3 0.2- -0.2 L v -
O.1- -0.1 0.01----l----l----l----l----l----l----l----I ---l----l----l--- 0.0 0- 5 10 15 20 25 30 -35 40 45 50 55 3
Control Rods Withdrawn l
i
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
u St__IBEQBl_QE_NUQLE88_EQWEB_EL8NI_QEEB8I1QNt_ELV10St_8NQ PAGE 8
-IBEBdQQ1N86101 QUESTION 5.13' ( .00)
Une a 1/M plot and PREDICT the number-of control. rods required to be withdrawn to achieve criticality.
NOTES: .1. CR = Count Rate
- 2. USE THIS FIGURE AT THE END OF THE SECTION AND INSERT.IN YOUR
!- ANSWER PACKAGE CR0 = 40 cps CR4 = 191 cos
- CR1 = 50 cps CRS =~333 cps CR2 = 89 cps -CR6 = 800 cos CR3 = 129 cps Ecch CR reading is recorded following a 5 rod withdrawal with CR0 being 100%_ rod density.
5 10 15- 20 25 30 35 40' 45 50 55 1.0-----l----l----l----l----l----l----l----l----l----l----l----1.0
)
0.9- -0.9 0.8-- -0.8 0.7- -0.7 1/M 0.6- -0.6 0.5- -0.5 0.4- -0.4 i 0. 3-- -0.3 0.2- -0.2 0.1- -0.1 0.01----l----l----l----l----l----l----l----l----l----l----l----0.0 0 5 10 15 20 25 30 35 40 45 50 55 I Control Rods Withdrawn e
(***** END OF CATEGORY 05 *****)
_. _ _ . _ _ .. . _ . _ _ , _ _ . , . . _ . _ _ _ _ ~ - . - . _ .._,.-._ _.-
Ex__ELeUI_SISIEUS_DESIQut_QQUIBQLt_eUQ_INSIB90ENI6IIQU PAGE 9 QUESTION 6.01 (2.00)
A single control rod is inserted one notch. The insert stabilizing volve(s) fail to reposition following the rod movement. How does this affect DRIVE WATER DIFFERENTIAL PRESSURE and FLOW CONTROL VALVE position? EXPLAIN YOUR ANSWER. (2.0)
QUESTION 6.02 (2.00)
For each SRV switch position / condition below, state which mode of SRV operation will still function (use SAFETY, PRESSURE RELIEF, ADS, and LOW LOW SET as the modes). Assume all modes of valve operation apply to the valve in question. (2.0)
- c. OFF
- b. AUTO
- c. The valve's logic circuit is in TEST QUESTION 6.03 (1.50)
The reactor is operating at 95% power. Condensate pump A is out of service. (pumps B and C operating). The Condensate Recirculation Velve (Short Cycle Cleanup to Condenser FCVil4) fails OPEN. EXPLAIN the effect this will have on reactor feed flow. (1.5)
QUESTION 6.04 (1.50)
Dascribe how the failure of a feedwater heater extraction steam non-return valve to shut on a turbine trip can result in over-speeding the turbine. (1.5)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
.ht__PL6NI_1111EN1_QEllGNt_CQNIBQLt_6HQ_IU11BVMENI6IlQU PAGE 10
. QUESTION 6.05 -(2.50)
- a. Following a LOCA, what'is the only valve in~the RHR system that cannot be manually overridden IMMEDIATELY following LPCI initiation? (0.5)
- b. What are two (2) indications that a RHR Inj ection Valve (F042A/B/C) has been manually overridden? (1.0)
- c. What conditions / actions must be met /taken to clear the manual override? (1.0).
QUESTION 6.06 (2.00)
- c. The RCIC flow transmitter has failed low (indicates O gpm)'. Select the answer below which.best describes how the RCIC turbine will respond upon receipt of an initiation signal. (1.0)
- 1. - The turbine will trip on overspeed
- 2. The turbine will run continuously at minimum Cidle) speed
- 3. The turbine will run continuously at approximately 4500 rpm
- 4. The system will not start (F045 Steam Supply remains closed)
- b. Which RCIC Isolation Division (1 or 2) is provided.with a manual isolation feature? (0.5)
- c. When is the manual isolation in part "b" above functional? (0.5)
QUESTION 6.07 (3.00)
- a. What rod movement restraints are in force as a result of exceeding
.the LPSP and HPSP during power escalation? (1.0)
- b. At what power levels do the LPSP and HPSP take effect? (1.0)
- c. Why-is control rod withdrawal undesirable if power is above the LPSP and a Main Turbine Bypass Valve is open? (1.0)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
6t__EL8HI_H11IEd1_QESIGNt_QQNIBQLt_8ND_INSIBUMENI8IIQU PAGE 11
. QUESTION' 6.00 (2.50)
The reactor is operating at 100%_ power. Recirculation flow control io in MASTER MANUAL. The Reactor Operator mistakenly depresses the OECREASE load button on the EHC panel and. holds it. Describe the response of the turbine control and recirculation flow control systems to this event.- Terminate the event with a reactor scram, including the cause. (2.5)
QUESTION 6.09 (2.00)
List ten (10) parameters that are monitored as part of the Post Accident Monitoring System. (2.5)
QUESTION 6.10 (2.00)
- a. During a control room fire, efforts should be focused on using Division CI, II, III) equipment to put the station in a safe shut-down condition. (0.5)
- b. How is control of a component transferred to the Remote Shutdown Panel from the control room? (0.5)
- c. What are two (2) indications in the control room that would indicate that transfer of a valve's control to the Remote Shutdown Panel had been performed? ( 1. 0 )'
QUESTION 6.11 (2.50)
- c. A Diesel Generator (Div I) has been started manually using the EMERGENCY START pushbutton in the control room. What are the only three (3) diesel generator trips that are armed? (1.5)
- b. With no LOCA condition or bus undervoltage signals present, but the diesel running from the above start, how are the normal protective trips reinstated? (1.0)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
hi__EL6NI_SYSIEdi_DESIGNt_QQNIBQLt_6NQ_INSIBudENI611QN PAGE 12 w.
QUESTION 6.12 .(1.50)
With reg'ard to the'Standoy Gas Treatment System (GTS), match each~of-the U signals (a- d) with the automatic action / mode (1 - 4) that applies.
' Items a d may have.one or more answers, and items 1 - 4 may be used more than once or-not at all. o(1.5)
ACTION / MODE-
[ e. Hi-H1 Radiation in the 'l. . Alignment of the GTS with the Auxiliary Building exhaust Auxiliary Building exhaust
- b. - Hi-Hi Radiation in the 2. Alarm only Annulus exhaust 7'
- 3. Alignment of the GTS with
- d. Low Flow in the Annulus 4. Alignment of the GTS with Pressure Control System (APCS) the Drywell/ Containment Purge System-t l
t 4
i i
(***** END OF CATEGORY 06 *****)
t.
. + , . - - -
w---' ..-,,v -,<.e.,,,--,-,~,- , , ~ , - - - - = , - . + , - - m.,~,,,----*,,,~.-,4- #-+
Y "Zi__EBQQEQVBES_:_NQBuelt_eHNQBd6Lt_EMEBQENQ1.8NQ PAGE 13 88Q10LQQ198L_QQNIBQL QUESTION 7.01. (3.00)
~
Match the following events (a-f) with the approximate pressure at which they should be performed per_GOP-0001,-Plant Startup to Low Power Alarm Point. The pressures:may be used more than once'or not at all. (3.0)
- c. Place steam seals in service 1. 20 psig
- b. Startup the Offges System 2. 100 psig
- c. -Line.up RCIC (Standby mode) 3. 150 psig
- d. Startup.the first reactor feed pump 4. 250 psig
-o. Start turbine shell warming 5. 400 psig
- f. Startup the SJAE 6. 450 psig
- 7. 600 psig
- 8. 850 psig QUESTION 7.02 (2.50)
Given the following cooldown data, DETERMINE if the RPV cooldown rate limit has been exceeded. STATE the RPV cooldown limit ~and HOW the determination was made. 'IF the limit was exceeded, STATE tne period (s) of time the limit was exceeded. (2.5)
TIME TEMPERATURE (Deg F)
-START- 0800 520 0815 490 0830 460 0845 445 0900 430 0915 400 0930 370 i 0945 340 1000 325 1015 310 1030 275 1045 250 1100 225 1115 205 l
(***** CATEGORY 07 CONTINUEC ON NEXT PAGE *****)
Zz__EBQQEQUBES_:_NQBd6Lt_6HNQBd6(t_EMEBQENQ1_6NQ PAGE 14
.88DIQLQQIQ8L_QQNIBQL QUESTION 7.03 (2.50)
Briefly state the reason for each of the following statements portaining to a reactor / plant shutdown from power:
- a. The turbine generator must remain on the line until Reactor Thermal Power is below 10%. (0.5)
-b. Maintain steam packing exhauster and air removal pumps in operation for up to~four (4) hours after-breaking vacuum. (0.5)
- c. Small control rod movements are essential when operating
.with the reactor critical and the MSIV's closed. (0.5)
- d. The main turbine must be on the turning gear whenever steam seals are on or turbine metal temperatures are above 150 deg.F. (0.5)
- e. Do not allow RPV pressure to_ decrease to O psig with the MSIV's-open and condenser vacuum > or = 20" Hg. (0.5)
QUESTION 7.04 (1.50)
What do the words "shall", "should", and "may" denote when they are used in River Bend procedures? (1.5)
, QUESTION 7.05 (2.50) i
.o. What two (2) general limitations or limiting plant parameters may I
restrict or inhibit removal of feed water heaters from service? (0.5) l t b. Explain how a reactivity transient may result from fluctuating 5th or 6th point heater levels. (1.0)
- c. Why must the operator exercise caution when shutting down the Heater Drain Iumps ? (1.0) i 1
l l
l l
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
r
._ . . , . , , ,_ . . - . - , . _ . . . _ , _ . , , ., , ,_m- . _ . . .
Zz__E80CEQUBES_:_UQBdekt_eBNQBualt_EUEBQENCY_8ND PAGE 15 86010LQQ1G6L_GQUIBQL QUESTION 7.06 (1.50)
Instrument Air header pressure is decreasing. All systems are re-cponding properly to the decreasing air pressure. Besides keeping the Shift Supervisor informed of plant conditions, what actions should bG taken by the control room operator (s) and at what point (if applicable) por A0D - 0008, Loss of Instrument Air? (1.5)
QUESTION 7.07 (1.00)
What two (2) conditions must be met so that an IMMEDIATE reactor scram is NOT required during a loss of Normal Service Water? (1.0)
QUESTION 7.08 (2.50) c.- With the facility at power, what actions should be taken prior to evacuating the control room? (2.0)
- b. Where should the control room personnel assemble following the evacuation? (0.5)
QUESTION 7.09 (3.00)
For each of the following conditions, state which E0PCs), if any, should be applicable. (3.0)
- o. Reactor water level 12.5 "
- b. .Drywell pressure 1.87 psig j
- c. Suppression pool temperature 97 deg F l
- d. Drywell to containment dp 27.3 psid
- o. Containment temperature 88 deg F
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
L
-Zi__EBQQEQUBE1_=_NQBd8Lt_8ENQBd8Lt_EdEBQEUQ1_6NQ PAGE- 16 88Q10LQQ108L_CQNIBQL QUESTION 7.10 (1.00)
-When an emergency condition requires implementation and use of the E0P's, what positions / roles are assumed by the Control Operating Foreman end the' Shift Supervisor? (1.0) 1 QUESTION 7.11 (2.50)
.A 24 year old individual, with a lifetime exposure of 23 REM through the 1 cst quarter, will be working in a 200 MREM /hr gamma field. In addition to the lifetime dose, the individual has received 850 MREM this quarter.
~
-a. How long may this individual work in this area before he/she reaches a radiation exposure limit? (1.0)
- b. Whose recommendation (s) and approval (s) are needed to allow the individual to exceed this limit? (1.0)
- c. If the radiation field had been beta instead of gamma, how would the limit in part "a" have changed (state the limit)? (0.5)
, QUESTION 7.12 (1.50)
The Operator "At-The-Controls" will manually initiate a reactor scram whenever one of two general conditions or situations occur. What are these_two general conditions or situations? (1.5) 4 a
i
(***** END OF CATEGORY 07 *****)
- At__6Dd1NISIB6I1YE_EBQQEQUBESt_QQNQ1110NSt_8NQ_LidIISIIQNg PAGE 17
-QUESTION 8.01 (3.00)
Fill in the blanks regarding the River' Bend Safety Limits: (3.0)
THERMAL POWER shall not exceed ___(a)___ % of RATED THERMAL POWER with
.the reactor vessel steam dome pressure less than ___(b)___ psig or core flow less than ___(c)___ % of rated flow.
The. Minimum Critical Power Ratio (MCPR) shall not be less than ___(d)___
with the reactor vessel steam dome pressure greater than or equal to
___(b)___ psig and core flow greater than or equal to ___(c)___ % of rated flow.
The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed ___Ce)___ psig.
-The reactor vessel water level shall be above ___(f)___.
QUESTION 8.02 (2.50)
Indicate whether each of the following is considered a " core alteration" per the River Bend Technical Specifications: (2.5)
- a. Withdrawal and insertion of an SRM detector to check the drive motor
- b. Removal of an LPRM string for replacement
- c. Removal of an uncoupled control rod for replacement #
- d. Removal of a control rod's position indicator probe for repair
- a. Control rod withdrawal and insertion to test the position indicator probe
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
at__8QUINISIBoIIME_EBgggguBEft_CQUDIIIQUlt_eUQ_LIMII8119US PAGE 18 QUESTION 8.03 (2.50)
Indicate whether the following statements are TRUE or FALSE. (2.5)
- 9. A Shift Supervisor may concurrently fill the position of the STA while on shift.
- c. All core alterations must be directly supervised by a licensed Senior Operator (or Senior Operator Limited to Fuel Handling).
- d. An Operator license is required for an operator to perform a core alteration.
o During Operational Condition 4 or 5, an individual with a valid
. Operator license may be designated to assume the control room command function during an absence of the Shift Supervisor from the control room.
QUESTION 8.04 (1.50)
Answer the following with regard to temporary procedure changes per the Technical Specifications.
- a. A temporary change to a procedure must be reviewed by whom (include any qualifications required of the individual (s))? (1.0)
- b. With the exception of the above review, what other condition must the temporary change meet? (0.5)
QUESTION 8.05 (1.50)
- c. When does removal of a temporary alteration require double verification ? (0,5)
- b. Under what condition (s) may a required double verification be waived? (0.5)
- c. What is the maximum time a temporary alteration is allowed to remain in effect? (0.5)
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
8t__6Dd1NISIBoIIVE_EB9CEQUBESt_C9NQ1110Nat_eUQ_L10116I1QN3 PAGE 19 QUESTION 8.06 (1.00)-
List five (5) agencies that must be notified within 15 minutes after e General Emergency is declared at River Bend. (1.0)
QUESTION 8.07 (2.50)
- c. Persons authorized as " requestors", individuals that request Tagging Orders, must be designated in writing by WHOM7 (0.5)
- b. Where is the list of authorized requestors maintained? (0.5)
- c. What two (2) employee classifications are " Designated Operators "
for Tagging Order purposes? (0.5)
- d. What are four (4) smployee classifications designated as " Tagging Officials"? (1.0)
QUESTION 8.08 (2.50)
Answer the following with regard to the River Bend Emergency Plan Implementing Procedures:
- c. When shall offsite radiation dose calculations be initiated? (1.0)
- b. Who is responsible for the initial and follow-up of f site dose calculations (at least until relieved by personnel in the TSC)? (0.5) i c. A determination has been made that offsite dose calculations are necessary.
- 1. When should the initial assessment be completed? (0.5)
- 2. When should follow-up calculations be completed? (0.5) l l
l l
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
Rz__8051NISIB6I1YE_EBQQEQUBESt_QQUDII1ONSt_6NQ_LIU1I8IIQUS PAGE 20
-QUESTION 8.09 (1.S0)
During preparation for a reactor plant startup from Operational Condition 3, the Shift Supervisor (SS) is informed that LPCI pump.A hcs failed a surveillance test, but it is probable that the problem can be corrected within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Based upon this in-formation, the SS determines that even if repairs take longer he still
-hcs 7 days before.he exceeds the time requirement in the Action statement. He also decides to continue the reactor startup. . Is this decision correct? Justify your position. (1.S)
QUESTION- 8.10 (2.50)
- c. If an individual performing an activity cannot or believes he/she should not follow the procedure governing that activity, what two (2) actions should be taken? (1.0)
- b. .What two (2) methods of resolving the procedural discrepancy are available? (0.S)
- c. .A safety-related valve position checklist is being performed with the initial performer and independent verifier performing their functions simultaneously. Is this satisfactory? EXPLAIN. (1.0)
QUESTION 8.11 (2.00)
- a. Where would an operator find the interpretation on a Technical Specification requirement that had required previous interpretation? (1.0)
- b. When is the operator allowed to enter an operating condition where an LCO cannot be met without relying on provisions of an action
- statement? (1.0) i 8.12
^
QUESTION (1.00)
In addition to utilization of Time, Distance, and Shielding, a concept called Source Reduction can be utilized to lower personnel exposure.
List three examples or methods that can be utilized for reducing SOURCE STRENGTH. (1.0)
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
,,At__8DdIN11IB8IIVE_EBQGEQWBEft_CQNQ1IIQNit_8NQ_LIMII8110NS PAGE. 21 1'
1~
I l- QUESTION ~8.13 (1.00)
W'hy is the Turbine Generator overspeed protection system addressed in 1 River Bend ~ Technical Specifications when the. Turbine Generator is not.
, cafety-related? (1.0)
I 4
4 4
1 1
1, j
i J
I t
i j
i i
[
h t
f- i l (***** END OF CATEGORY 08 *****) .
L
(*************
END OF EXAMINATION
- )
EQUATICri 5HEET f a ma y= s/t Cycle efficiency = (Net work out)/(Energy in) 2 w = ag s = V,t + 1/2 at 2
E = ac ,
~AI KE = 1/2 av a = (Vf - V,)/t A = AN A = A,e PE = ogn Vf = V, + at w = e/t A = an2/t1/2 = 0.693/t1/2 w = ,ap -
2 tjfg = UtmW]
A= ID,
[(t1/2)*I*b)3 aE = 931 an -
""Y avA ' g.g. -h o
Q = aCpat I = I ,e~"*
6 = UAAT Pwr = Wfah I = I ,10-* E I TVL = 1.3/u 8
P = P 10 "'II) HVL = -0.693/u t <
P = Po e /T l
l SUR = 26.06/T S G = S/(1 - K,ff)
CR,= S/(1 - K,ff ,)
SUR = 26e/a* + (s - e)T Gj (1 - K,ffj) = G2II - "eff2)
T = ( a*/s) + [(s - e y Te] M = 1/(1 - K,ff) = CR /CRj ,
y = s/(, - s) M = (1 - K ,ff,)/(1 - K,ffj)
T = (8 - e)/(I,) SDM = ( - K,ff)/Keff a = (X,ff-l)/Keff
- 8Keff/K,ff s' = 10 seconds A = 0.1 seconds _
p = (( a*/(T Kgff)] + [3 gff /(1 + IT)]
Idjj=Id Id 2 ,g gd 2
P = (teV)/(3 x 1010) jj 22 2 I = oN R/hr = (0.5 CE)/d (meters)
R/hr = 6 CE/d2 (f,,g) .
l Water parameters Miscellaneous Conversions I gal. = 8.345 lbm. 1 curie = 3.7 x 1010dps 1 ga:. = 3.78 liters 1 kg = 2.21 lbm I hp = 2.54 x 103 Stu/hr 1 ftd = 7.48 gal.
Density = 62.4 lbm/ft3 1 ,= 3.41 x 100 Stu/hr Density = 1 gm/c:n3 lin = 2.54 cm l
Heat of vaporization = 970 Stu/lom *F = 9/5'C + 32
'C = 5/9 (*F-32)
Heat of fusion = 144 Stu/lbm l 1 Atm = 14.7 psi = 29.9 in. Hg. 1 BTU = 778 ft-lbf i i ft. H O 2
= 0.4335 lbf/in.
l
Sz__IBEQBX_QE_UUQLEoB_EQWEB_EL6UI_QEEBoIIQUt_EL91 dst _eUQ PAGE 22 IBEBdQQXUedIQS ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 5.01 (2.00)
- e. 8
- 6. 7
- c. 1
- d. 4 (0.5 each)
REFERENCE LOTM-TH-3.9-0, LHGR LOTM-TH-3.10-0, APLHGR ANSWER 5.02 (3.00)
- c. CRITICAL POWER increases (0.25). A greater enthalpy rise is required to bring the coolant to saturation conditions resulting in less steam generation for a given power level (0.75) or similar.
- b. CRITICAL POWER increases (0.25). Same reason as above (0.75).
- c. CRITICAL POWER decreases (0.25). A lower enthalpy rise is required to change a given mass of coolant from liquid to vapor at a higher pressure (0.75).
REFERENCE LOTM-TH-3.7-0, Transition Boiling, pg 3 ANSWER 5.03 (2.00)
- c. Fission products: Tritium, I, Kr, Xe, Cs, Ba (0.1 each)
Corrosion products: Fe, Ni, Co, Mn (0.1 each)
- b. No (0.25). The activity is due to tramp uranium fission (0.75).
REFERENCE Mitigating Core Damage, pg 5-8 Chart of the Nuclides
Si__IBEQBY_QE_NUQLE88_EQWEB_EL8MI_QEEBoIIQNz_ELu1DSt_6NQ PAGE 23
-IBEBdQQ1Ned103
. ANSWERS ---RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 5.04 (2.00) a.- increase (0.5)
- b. 1. decrease (0.5)
- 2. decrease (0,5)
- 3. increase (0.5)
REFERENCE LOTM-TH-1.4-0, Pumps, pg 3, 17 ANSWER. 5.05 (1.00)
- 1. a
- 2. e 3.' d
- 4. b f 5. c (0.25 for each manipulation necessary to put the items in the correct i
order)
REFERENCE
-LOTM-TH-3.2-0, Modes of Heat Transfer LOTM-TH-3.5-0, Boiling Heat Transfer ANSWER 5.06 (1.00)
}: b (1.0) i REFERENCE LOTM-TH-2.5-0, 1st Law of Thermodynamics LOTM-TH-2.13-0, Reactor Heat Balance
'51__IBEQB1_QE_NuGLE68_EQWEB_EL8HI_QEEBoI1QNt_ELVIDS _8ND PAGE 24 IHEBdQQXNedIQH-ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 5.07 (2.50)
- e. Four fuel bundles and a control rod (0,5)
- b. Fully withdrawn (0.5).
-c. All control rod movements are associated with low reactivity cells (1.0).
- d. True (0.5)
-REFERENCE LOTM-2-1~, Nuclear Fuel, pg 20-23 ANSWER 5.08 (3.00)
-Void fraction decreases (0.5). The void fraction is slightly. smaller because the negative effects of doppler are compensated for by the void fraction decrease (0.5).
Doppler reactivity increases (0.5). As power increases, the fuel temperature increases. This temperature increases cause doppler to cdd more negative reactivity (0.5).
Enthalpy increases ( 0. 5 ) . - The increased reactor power allows more
. extraction steam to be drawn off thus increaseing inlet enthalpy (0.5).
REFERENCE LOTM-TH-4.18-1, Void Coefficient LOTM-TH-4.19-1, Doppler Effect LOTM-14b-0, Feedwater Heaters and Heater Drains r
4 e
i Ez__IBEQB1_QE_NUGLE68 EQWEB_EL6NI_QEEB6110Nt_ELUIDSt_6NQ PAGE_ 25 IBEBdQQ1NadIQ1 ANSWERS -- RIVER BEND 1 -86/ll/04-GRAVES, D.
ANSWER 5.09 (2.50)
- c. True (0.5) b~ . False (0.25). Xenon has a higher microscopic absorption cross section (0.25).
- c. False (0.25). It may be Xenon free, but' Samarium will increase following the scram (0.25). Also accept "never samarium free following power operation."
- d. True (0.5).
- o. False (0.25). 50% equilibrium Xenon ~is > one-half the 100%
equilibrium value (0.25).
REFERENCE LOTM-TH-4.21-1, Samarium and Reactor Poisons LOTM-TH-4.22-1, Xenon ANSWER 5.10 (1.50)
Withdrawal of a center control rod at 90% density has. greater worth.(0.53-The control rod worth is proportional to the (local neutron flux / the core average neutron flux) squared.(0.25)
With 90% rod-density the core average neutron flux is very small.
-Withdrawing a central control rod, increases the local flux in the area of withdrawn rod substantially. Because the rod causes the value of the term (local neutron flux / core average neutron flux) squared to be large
~
its worth for this condition is quite high. Higher than withdrawing the rod at 40% rod density, when core average flux will be higher. (0.75)
REFERENCE LOTM-TH-4.20-1, Control Rod Worth, pg 4 i
l I
l r
1
'St__IBEQBl_QE_NVQLEeB_EQWEB_EL8NI_QEEBel10Nt_ELVIQ1t_eND PAGE 26 IHEBdQQ1Ned102 1
ANSWERS.-- RIVER BEND 1. -86/ll/04-GRAVES, D.
. ANSWER 5.11 (2.50) s.- ~The amoun't-of reactivity :by which the ~ reactor is subcritical (0.1)'
or would be suberitical (0.1) assuming all control rods.are fully inserted (0.2)'except'for the single control rod of highest reactivity worth is assumed to be fully withdrawn (0.2) and the reactor is in the shutdown' condition, cold'(0.2) and xenon free (0.2),
b
. b. 0.38% dk/k (0.5) 0.28% dk/k (0.5)
- ' Which value is limiting is determined by whether the highest worth
' control rod was determined by test (0.28%)'or analytically ~(0.38%)(1.0)
REFERENCE
. River Bend Technical Specifications Definitions Technical-Specification 3.1.1 s
i l
i
, I I-l L
w i
i 5t__IBEQBl_QE_NUQLE88 EQWEE EL6HI_QEEB6110Nz_ELU101t_8NQ PAGE 27 IBEBdQQXNadIQS ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER .5.12 (2.00)
Drawing a straight line between the last two *'s predicts 34-35 control rods must be withdrawn.
(0.1 for each point plotted, 1.3 for line and prediction) 5 10 15 20 25 30 35 40 '45 50 55 1.0*----l----l----l----l----l----l----l----l----l----l----l----1.~0 0.9- -0.9
'0.8- * -0.8 0.7 . -0.7
'1/M 0.6-- -0.6' 0.5- -0.5 0.4- -0.4 0.3- * -0.3 0.2- * -0.2 0.1- * -0.1 0.01----l----l----l----l----l----l----l----l----l----l----l----0.0
'O .5 10 15 20 25 30 35 40 45 50 55 Control Rods REFERENCE LOTM-TH-4.13.1, Suberitical Multiplication and Count Rate Comparison ANSWER 5.13 ( .00)
NOT APPLICABLE
n 6t__ELeNI_11SIEMS_QE11GNt_QQNIBQLt_6NQ_IN11BudENI8IIQN PAGE 28 ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 6.01 (2.00)
Flow control valve position does not change (0.5) because total system flow has not changed (0.5). Drive water differential pressure will increase (0.5) because flow previously bypassing the pressure control velve is now passing through it (0.5).
Also accept that the FCV opens slightly to maintain flow due to slightly increased backpressure.
REFERENCE LOTM-4-1, CRDH System, pg 16-17 ANSWER 6.02 (2.00)
- s. Safety, ADS
- b. Safety, Pressure Relief, ADS, Low Low Set
- c. Safety (accept ALL if noted that only 1 logic circuit is in test)
(0.285 each, 2.0 total)
REFERENCE LOTM-7-1, Main Steam, pg 5-7, 16 ANSWER 6.03 (1.50)
Suction pressure to the RFP will decrease (0.5). Reactor feed flow will docrease (0.5) due "to RFP trip on low suction pressure or the reduction in RFP discharge pressure (0.5 accept either).
REFERENCE LOTM-12-1, Condensate System, pg 5, 41 Simulator Malfunction 85b ANSWER 6.04 (1.50)
When the turbine trips, pressure in the turbine decreases to condenser pressure (0.5). The condensed extraction steam flashes to steam (0.5),
and backflows up the extraction steam line through the turbine (0.5), which con cause it to overspeed.
REFERENCE LOTM-14b-1, Feedwater Heaters, Vents, and Drains, pg 17
It__EL8HI_111IEd1_QE11GNt_GQNIBQLt_8NQ_INSIBudENI611QN PAGE. 29
- AN S WE R S ---- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 6.05 (2.50)
- a. RHR HX Bypass valve (0.5). Other valves were accepted as alternate answers. See.the valve logic tables in the stated reference.
- b. - amber light above the valve switch (0.5)
- annunciator-(RHR Inj ection Valve in Man Override) (0.5)
- c. - LOCA signal clear (0.5)
- Initiation Reset pushbutton(s) depressed (0.5)
REFERENCE LOTM-20-0, RHR System, pg 9 ANSWER 6.06 (2.00)
- a. 3 (1.0)
- b. 1 (0.5)
- c. initiation signal must be present (0.5)
REFERENCE LOTM-21-0, RCIC System, pg 7, 13, 37, 38 ANSWER 6.07 (3.00)
- o. LPSP: Rod withdrawai'is limited to four notches (0.5) before having to deselect and reselect a rod.
HPSP: Rod withdrawal is limited to two notches (0.5) before having to deselect and reselect a rod.
- b. LPSP:. 27.5% power or 157.6 psig first stage turbine pressure (0.5)
'HPSP: 62.5% power or 410.7 psig first stage turbine pressure (0.5)
Accept + or - 1% power and + or - 5 psig pressure
- c. Reactor power would be higher than that indicated by turbine first stage pressure. Rod movement could be performed that could conceivably exceed a fuel safety limit (1.0).
REFERENCE LOTM-IC-7-0, RC & IS, pg 31, 47 River Bend Technical Specifications Bases, 3.1.4.1
ht__EL6NI_111IEH1_QE11GNt_QQNIBQLt_8NQ_IN11BudENI6IIQN PAGE 30 ANSWERS -- RIVER BEND 1 -86/ll/04-GRAVES, D.
~ ANSWER 6.08 (2.50)
An the load set motor runs down, the EHC system will be calling for the control valves to.close to reduce load, which they will do.(0.5).
-As the CV close, the turbine bypass valves will begin to open to
-maintain pressure (0.5). Once the BPV are fully open and the CV are still' shutting, reactor pressure and power will begin to increase (0. 5) .
cousing the recirculation flow' control valves to close down in an ettempt to keep reactor power at 100% C0.5). This continues until the reactor scrams on high pressure or power (0.5)
' REFERENCE LOTM-IC-8-0, . Recirculation Flow Control, pg 15, 23 LOTM-IC-9-0, EHC, pg 85-86, 114 ANSWER 6.09 (2.00)
Containment /Drywell dp Containment pressure Containment radiation Containment temperature Containment hydrogen Drywell Hydrogen Drywell pressure Drywell temperature Drywell radiation Suppression pool temperature Suppression pool level Reactor water level Reactor pressure (10 required at 0.2 each)
REFERENCE LOTM-IC-23-0, Post Accident Monitoring, pg 2-4 I
i I
1 l
i
.6t__EL6NI_111IEdH_DESIGNt_QQNIBQLt_eSQ_INSIBudENI6IIQN PAGE 31 ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 6.10 (2.00)
- c. I (0.5)
- b. By taking the Remote Shutdown Transfer Switches to the EMERGENCY position (0.5).
- c. Annunciator (s) in the contrcl room (0.5)
Loss of position indication for that valve in the control room (0.5).
REFERENCE LOTM-IC-16-0, Remote Shutdown, pg 7-8 ANSWER 6.11 (2.50)
- e. .overspeed generator differential manual (0.5 each) b .' Press the control room EMERGENCY START RESET pushbutton (0.5) and the NORMAL START pushbutton at the local control panel (0.5).
Also accepted requirement to shutdown EDG and return to STBY per SOP-0053, Rev 4, pg 4 and 21.
REFERENCE LOTM-35-0, Standby Diesel Generator, pg 12, 20 ANSWER 6.12 (1.50)
- e. 2
- b. 3
- c. 1, 3
- d. 3 (5 at 0.3 each)
REFERENCE LOTM-32-0, Standby Gas Treatment 3ystem, pg 2 LOTM-31a-0, Reactor Building HVAC, pg 5-7 LOTM-31b-0, Auxiliary Building HVAC, pg 6
4 Z1__EBQGERVBES_:_UQBd8Lt 6BUQBd8L&_EdEBQENGX_68Q PAGE 32 B6DIQLQQ1 col _GQUIBQL ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
{
~ _
ANSWER 7.01 (3.00)
^
- c. 3 or 150 psig g
- b. 4 or 250 psig
- c. 2 or 100 psig or 3 or 150 psig *
- d. 5 or 400 psig
- o. 2 or 100 psig _
- f. 4 or 250 psig (0.5 each)
REFERENCE GOP-0001, Plant Startup, Rev 6, pg 22-24 ANSWER 7.02 (2.50)
The cooldown rate limit is < or = 100 deg F (0.25) in any one hour period (0.25). The limit was exceeded (0.5) for three one hour periods: 0845 - 0945 (105 des F), 0900 - 1000 (105 deg F), and 1015 - 1115 (105 deg F). (3 periods of time at 0.5 each)
REFERENCE -
River Bend Technical Specification 3.4.6.1 ANSWER 7.03 (2.50)
- c. Turbine BPV are only rated at 10% total steam flow. Will also accept to prevent a high pressure scram. (0.5)
- b. High airborne activity in the turbine building can result (0.5) if vacuum is broken within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from operation at full power.
- c. There is no automatic pressure control (0.5).
- d. To prevent warping the shaft (0.5).
- o. Prevent flashing the reference legs (0.5) on the RPV instruments.
REFERENCE River Bend GOP 0002, Power Decrease / Plant Shutdown, Rev 3, pg 8-10 -
9
= ^
17 . f
, . ti
- i; .
- 3
'N? . . .. . .
.. ?? ;
ed Zi__EBQGEQWBES._ _NQB58Li_8ENQB58Lt_EMEBGENQ1_8ND- :PAGE .33-1*7: + -B8DIQL921G8L_GQNIBQL:
s> . .
s4_
(/ ANSWERS,-- RIVER BEND 1. -86/ll/04-GRAVES, D.
{fQ T
-Qye m;.
y
$[ ANSWER 7.04 (1.50) chall - denotes a requirement (0.5) y chould:- denotes a: recommendation (0.5)
A may - denotes permission-(0.5) 1 ,.
> REFERENCE
!u/; River Bend ADM 0003, Development, Use, and Control of Procedures, E' Rov 10, pg 5
. /-
. ANSWER: .p 7.05 (2.50)
~ .. ,
ce 1e. MCPR or Turbine limits'(0.5) b'. -Eitreme high levels in the heaters will auto bypass the condensate stream around.the heaters (0,5) causing-feed water-temperature to decrease and change reactor power (0.5).
' c. ' Flow through the demineralizers must increase as pumped forward flowJis reduced (0.5). Subsequent pressure oscillations may cause
. feed p' ump trip (0.5).
L
- =~ r . . ,
" j? r # ; R E F_E R E N C E' LG0P-0002,. Power Decrease / Plant. Shutdown, Rev 3, pg 8-9 y
t_vs ANSWER f 7.06' (1.50) f Wh6n i strument air header pressure has decreased.to 65 psig (0.5) or it becomes apparent that instrument air will not be restored (0.5), scram the I
J_ r'r e a c"t o r' ( 0 . 5 ) '.
. REFERENCE.
River Bend AOP - 0008, Loss of Instrument Air, Rev 2, pg 3 ANSWER 7.07 (1.00)
--The main turbine is NOT on the line (0.5).
. Turbine BPV are NOT being used to control reactor pressure (0.5).
, REFERENCE h River BendoA0P 0009, Loss of Normal Service Water, Rev 2, pg 4 pur ::wc F[ .,v.
1Zz__tBQQEQUBE1_=_NQBM8Lt_8HNQBd6Lt_EMEEGENQX_6NQ PAGE 34 B8DIQLQQIQaL_GQNIBQL
' ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 7.08 (2.50)
- c. 1. Arm ~and depress allifour manual scram' buttons.
2.. Mode Switch to SHUTDOWN.
- 3. Verify control rods 1 fully inserted.
- 5. Close.the MSIV's-(5 at 0.4 each)
- b. ' Division I Remote Shutdown Panel (0.5)
~ REFERENCE _ _ .
AOP-0031, Shutdown from Outside the~ControlzRoom,-Rev~3, pg 6 ANSWER- 7.09 (3.00)
- c. none-b._ E0P-0001, RPV Control E0P-0002,-Primary Containment Control c .- .EOP-0002,--Primary Containment Control d.- E0P-0005, RPV Flooding
- e. none (6 at 0.5 each)
REFERENCE' River Bend.EOP's 0001,.Rev 6; 0002, Rev 5; 0005, Rev 3 d
ANSWER .7 .~ 10 (1.00).
tThe Control Operating Foreman will assume the role of E0P Director /
Coordinator (0.5) and the Shift Supervisor will assume the role of
-advisor"to the C0F (0.5).
REFERENCE 0SP-000g, Author's Guide / Control and Use of Emergency Procedures, Rev 1, pg 18 o
S , s ,, , , ,,.~ n , - ,~-e- r - -
.- e+n.a , - - - - -
-+w - - - , , - - - - - , . . - - .
Zi__EBQQEQUBER_:_NQBd6Li_8BNQBd6Lt_EMEBQENQ1_8NQ PAGE 35 88DIQLQQIQaL_QQNIBQL ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES,.D.
ANSWER- .7.11' (2.50) a.- 1000 MREM - 850 MREM = 150 MREM 150 MREM / (200 MREM /hr) = 0.75 hr or 45 minutes -1 l
-(0.5 pt for the limit, 0.5~for the stay time)
- b. Assistant Plant Manager recommendation (0.5)
Supervisor-Radiological Programs approval (0.5)
REFERENCE GET II-1, pg 4
~ ANSWER 7.12 (1.50)
- 1. A scram setpoint is exceeded and an automatic scram did not occur-or 2.. Plant conditions are approaching an. unsafe: condition-and a. manual scram will_ mitigate the consequences of this condition (concept).
(2 at 0.75 each)
REFERENCE A0P-0001,-Reactor Scram, Rev 3, pg 2 I
t l.
i
at__6DdIN1118611YE_EBQQEQVBEft_QQNDIIIQUEt_oND_LIMII6IIQN3 PAGE 36 ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 8.01 (3.00)
- c. 25
- b. 785
- c. 10
- d. 1.06
- c. 1325
- f. the top of the active irradiated fuel (6 at 0.5 each)
REFERENCE River Bend Technical Specifications, Safety Limits ANSWER 8.02 (2.50)
- c. no
- b. yes
- c. yes
- d. no
- o. yes (5 at 0.5 each)
REFERENCE River Bend Technical Specifications, Definition Section ANSWER 8.03 (2.50)
- n. True
- b. False
- c. True
- d. False
- c. True (5 at 0.5 each)
REFERENCE River Bend Technical Specifications, Section 6.2.2
.8t __6Dd1NISIB8IIVE_PBQQgQUBggt_GQUQlllQU$t_6NQ_LidlI611983 PAGE 37 ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
-ANSWER 8.04 (1.50)
- c. Two members of plant management staff (0.5), one of which holds a Senior Operator license (0.5). Also accept section supervisor, discipline supervisor or foreman AND C0F or SS as the two members of plant management staff, if candidates' answers are that specific.
- b. The intent of the procedure is not changed OR to continue work already in progress (either one at 0.5).
REFERENCE River Bend Technical Specifications, Section 6.8.3 ANSWER 8.05 (1.50)
- c. Double verification is required if the temporary alteration is on a safety related component or system (0.5).
- b. A double verification may be waived only to prevent significant radiation exposure (0.5).
- c. 90 days (0.5). A discussion of the current policy per Temporary Change Notice 86-123 was accepted with or without the 90 day answer.
REFERENCE ADM - 0031, Temporary Alterations, Rev 0, pg 3,4,6 ANSWER 8.06 (1.00)
- Louisiana Nuclear Energy Division Louisiana Office of Emergency Preparedness East Feliciana Parish West Feliciana Parish
- Pointe Coupee Parish
- East Baton Rouge Parish
- West Baton Rouge Parish
- Miss iss ipp i Highway Safety Patrol (5 required at 0.2 each)
REFERENCE EIP-2-005, General Emergency, Rev 3, pg 5 EIP-2-006, Notifications, Rev 6, pg 3 1
1 J Az__8DMIN11IB811YE;EBQQEQUBElt_QQNDIIIQN1t_aND_LIMIIeI1QN1 PAGE 3;8
" ANSWERS -- RIVER BENO l -86/ll/04-GRAVES, D..
f ,
ANSWER 8.07 -(2.50)
, s. Operations' Supervisor (0.5)
- b. The list of authorized requestors is kept in front of the
-Tagging-Log-Notebook (0.5).
- c. Nuclear Equipment Operator (0.25) 4 Nuclear Control Operator (0.25)
. d. 0perations Supervisor Assistant Operations Supervisor Shift Supervisor-
- -Cont ro 1~ Ope r at ing Foreman l Nuclear Control Operator l
(4 required at 0.25 each)
REFERENCE River Bend AOM 0027, Protective. Tagging, Rev 3, pg'2,3 i
^
ANSWER H8.08 (2.50) la. Of fsite dose calculations shall be perf ormed anytime a Site Area (0.25). t or General' Emergency (0.25) is declared due to a radiological i accident (0.25), or any other time the Emergency Director or
- Recovery Manager deems necessary (0.25).
i .
!. b. . Shift Foreman (0.5)
, c. 1. Initial assessment should be completed within approximately
! 15 minutes of'the declaration (0.5)
I
- 2. Follow-up calculations should be performed approximately every 30 minutes following initial-projections (0.25),
or sooner if release rates change significantly (0.25).
REFERENCE River Bend EIP-2-025, Offsite Dose Calculations - Computer Method, Rev 2, pg 2 i
ANSWER 8.09 (1.50)
No (0.5). Entry into an Operational Condition or other specified condition shall not be made unless the conditions for the LCO are met without reliance on provisions contained in the ACTION require-mants (1.0).
~ At__8Dd1NISIB8IIVE_EBQQEDMBESt_QQNQ1I1QNSz_8ND_ Lid 1I8I1QN3 PAGE 39
. ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
REFERENCE River Bend Technical Specifications, 3.0.4 and 3.5.1, ANSWER 8.10 (2.50)
- s. - Place the system / component in a stable.and safe condition (0.5)
- Inform the responsible supervisor (0.5)
+
- b. - Determining the method by which the activity can be performed using-the procedure as written (0.25)
- Submitting a procedure change (0.25)
- c. ' No (0.25). The-independent verification will.be performed separately without visual or audible contact with the first performer (0.75).
REFERENCE River Bend ADM 0022, Conduct of Operations, Rev 7, pg 24, 26
~ ANSWER 8.11 (2.00)
- a. There are interpretation letters included in front of the Technical Spec if ic at ions (1.0).
, b. When it is required to change modes to meet an action statement, or TS 3.04 is specified as N/A (1.0).
REFERENCE River Bend Technical Specifications i
t ANSWER 8.12 (1.00) l
- 1. Decontamination f
- 2. Draining & Purging of equipment
- 3. Storing for decay-
- 4. Reducing reactor power level.
(3 at 0.33 each)
REFERENCE GET II-1, pg42 L
i g ..- -+ , --. ,. e ,.~~-m -,- e,--,.~.w., -..w- - , - , , .,-,-,.~-~,,--v -, , - - - , - ,. ,
At__8DMIN11188IIVE_EBQQEQUBElt_QQNDIIIQNit_8NQ_L15116I1QN3 PAGE 40
! ANSWERS -- RIVER BEND 1 -86/ll/04-GRAVES, D.
ANSWER 8.13 - (1. 0 0 ) -
Failure:to~ trip on overspeed could result in the turbine becoming a missile-hazard (which could impact safety related equipment, systems or barriers).
' REFERENCE RBS_ Technical Specification 3.3.8
U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR REQUALIFICATION EXAMINATION FACILITY: _BlyEB_gENQ_1____________
REACTOR TYPE: _RWB-QEh_________________
DATE ADMINISTERED: _Qk411LQd________________
EXAMINER: _QBAVEgt_Qt______________
CANDIDATE: _________________________
INSIBuQ110NE_IQ_QaNQ1QaIE1 Road the attached instruction page carefully. This examination replaces tho current cycle facility administered requalification examination.
Ratraining requirements for feilure of this examination are the same as for' failure of a requalification examination prepared and administered by your training staff. 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 four (4) hours after the examination starts.
% OF
- CATEGORY % OF CANDIDATE'S CATEGORY
__VaLUE_ _IQIaL ___100BE___ _ VALVE __ ______________QaIEQQBl_____________
_15tDQ__ _J$ 21 ___________ ________ l. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, 4
HEAT TRANSFER AND FLUID FLOW
_15tDQ__ _J3.QQ ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
_14z00__ _l! @3_ ___________ ________ 3. INSTRUMENTS AND CONTROLS
_15zQQ__ _25.Q1 ___________ ________ 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL
_EQtDQ__ ___________ ________% Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.
I Candidate's Signature
, - - _ , _ - . ~ . _ - _ _ . _ - . . .- ,. . _ . _ _ . _ - --- _ . _ _ ____. _ ._ _._
4 NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS
.During1the administration of this examination the-following rules-apply:
- 1. Cheating on the-examination means an automatic denial of your application end.could result in more severe penalties.
.2. 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 i room to avoid even the appearance or possibility of cheating.
s
- 3. Use black ink or dark pencil 2Dly to facilitate legible reproductions.
- 4. Print your name in the blank provided on the cover sheet of the
. examination. '
5.- Fill in the date on the cover sheet of the examination (if necessary).
l
- 6. Use only the paper provided for answers, r
- 7. Print your name in the. upper right-hand corner of the first page of RS2h
- section of the answer sheet.
- 8. - Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a Dgg page, write 2DlY 2D 2D1 Eidt of the paper, and write "Last Page" on the last answer sheet.
- 9. Number each answer as to category and number, for example, 1.4, 6.3.
10.-Skip at least ihrgg lines between each answer.
- 11. Separate answer sheets from pad and place finished-answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly used in facility 111gtglyng.
- 13. The point value for each question is indicated in parentheses after the l
Question and can be used as a guide for the depth of answer required.
I 14. Show all calculations, methods, or assumptions used to obtain an answer.
. to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE i
QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of
, the RER51DRt only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.
4
- - - - - . ,..c ,,v._.--,-m ,,- - ,,....m1, ,. -.,-.,,,..w. . . , . . . . - - _ , , , - , _ . , , . ~ - - _ _ _ . _ . - _ , , , . . . -
.. ~
~18. When you complete'your examination, you shall:
y ~ a. Assemble your examination as follows:
Cl) . Exam questions on top.
(2) Exam aids --figures, tables, etc.
I
.(3) -Answer pages including figures which are part of~the answer.
~
- b. Turn in your copy of the examination and all pages used to answer the examination questions,
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering.the questions.
- d. Leave the examination area, _as defined by the examiner. If after
. leaving, you are found=in this area while the examination is stil'1 in progress, your license may be denied or revoked.
I b
I,,
i i
a w - a y e,--.~n,-m- , ., c --n,.--e,- - , , . , - , - , - - - .,
- - , --n---,e,~ v.,, - . . - ,,m n ,
- iz__EBINQlELE1_QE_NuGLE88_EQWEB_EL8MI_QEEB811QN t PAGE 2 IHEBdQQ1N8dlDit_HE8I_IB8NSEEB_8ND_ELulD_ELQW QUESTION 1.01. (1.50)
Indicate HOW each of the following changes will affect reactor
' CRITICAL POWER (INCREASE or DECREASE). If CRITICAL POWER will not bo affected, state this. (1.5)
- c. Loss of extraction steam to a feedwater heater
- b. Mass flow rate through the core is increased c .' Reactor pressure is increased QUESTION 1.02 (2.00)
Using the steam tables, indicate whether water at each of the following ie SU8C00 LED, SATURATED, or SUPERHEATED. (2.0)
- a. 200 psig, 387.7 F
- b. 1000 psig, 544.6 F
- c. 1200 ps ig, 603.9 F
- d. 900 ps ig, 531.1 F QUESTION 1.03 (2.00)
Answer the following with regard to centrifugal pump operation.
- c. How does increasing the pump's speed affect the REQUIRED Net Positive Suction Head (NPSH) for the pump? Answer INCREASE, DECREASE, or i UNCHANGED. (0.5)
- b. If the discharge valve for the pump is moved in the close direction,
! how will each of the following change? Answer INCREASE, DECREASE, or UNCHANGED. (1.5)
- 1. System flow l 2. motor amps
- 3. pump discharge head j
i I
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
c11__EBINQ1 ELE 1_QE_NuQLE68 EQWEB_EL6NI_QEEB6I1QNt PAGE 3 IHEBdQQ1Ned10ft_HE81_IB8NSEEB_8ND_ELu1Q_ELQW QUESTION 1.04 (2.50)
Answer the following with regard to the Control Cell Core (CCC) operating strategy:
- e. What constitutes a CONTROL CELL? C0.5)
- b. While operating at power, what is the position of the non-control cell' control rods? (0.5)
- c. Why do control rod movements result in a lower kw/ft change per~ notch in a CCC than in a conventional core? (1.0) d .- TRUE or FALSE. Using the CCC operating strategy eliminates the need-for control rod pattern exchanges. (0.5)
L 1.05 QUESTION (1.50) i Dsscribe (INCREASE, DECREASE, OR UNCHANGED) the effect on void fraction, doppler reactivity and feedwater enthalpy due to a recir culation flow rate increase at power. (1.5)
QUESTION 1.06 (2.00)
The reactor is started up after a refueling outage. Rods are pulled to the 100% line and power is then increased to 100% with recirculation flow.
! After approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, reactor power has decreased to about 98%.
Assume no operator action.
- c. What is the primary cause for this reduction in power? (1.0)
- b. -When would you expect the power decrease to stop and WHY does it stop? (1.0)
QUESTION 1.07 (1.50)
Concerning control rod worth, compare withdrawing a center control rod at 90% rod density to withdrawing a center control rod at 40% rod density. In which situation is control rod worth greater for the withdrawn control rod? Explain your answer. (1.5)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
'1x__EBINQlELER_QE NVQLE88_EQWEB_EL8NI_QEEB8IIQNr. PAGE 4
-IBEBdQQ1N8510ft_BE81_IB8NSEEB 8NQ_ELVIQ_ELQW 4
QUESTION. 1.08 (2.00) t use a 1/M plot and PREDICT.the-number'of control rods required to be j withdrawn to achieve criticality.
NOTES: 1. CR = Count Rate
- 2. USE THE FIGURE AT THE END OF THE SECTION AND INSERT IN YOUR ANSWER PACKAGE CR0 = 40 cps CR4 = 191 cps CR1 = 50 cps CR5 = 333 cps CR2 = 89 cps CR6 = 800 cps CR3 = 129 cps Esch CR reading is recorded following a 5 rod withdrawal with CR0 being .
100% rod density.
l 5 10 15 20 25 30 35 40 45 50 55 1.0-----l----l----l----l----l----l----l----l----l----l----l----1.0 0.9- -0.9 l - -
0.8- -0.8 0.7- -0.7 l .1/M 0.6- -0.6 0.5- -0.5 l 0.4- -0.4 i 0.3- -0.3
- 0. 2-1 -0.2 t
0.1 . -0.1
! 0.01----l----l----l----l----l----l----l----l----l----l----l----0.0 i :D 5 10 15 20 25 30 35 40 45 50 55 l Control Rods Withdrawn I,
1
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
l
11__EBING1 ELE 1_QE_URCLEeB_EQWEB EL6UI_QEEBoI1QNi PAGE 5 IBEBdQQlded1Gli_BEeI_IBeBSEgB_aug_ELu1R_EL9W QUESTION 1.09 ( .00)
Une a 1/M plot and PREDICT the number of control rods required to be t1ithdrawn to achieve criticality.
NOTES: 1. CR = Count Rate
- 2. USE THIS FIGURE AT THE END OF THE SECTION AND INSERT IN YOUR ANSWER PACKAGE CR0 = 40 cps CR4 = 191 cps CR1 = 50 cps CR5 = 333 cps CR2 = 89 cps CR6 = 800 cps CR3 = 129 cps Ecch CR reading is recorded following a 5 rod withdrawal with CR0 being 100% rod density.
5 10 15 20 25 30 35 40 45 50 55 1.0-----l----l----l----l----l----l----l----l----l----l----l----1.0 0.9- -0.9 0.0- -0.8 0.7- -0.7 1/M 0.6- -0.6 0.5- -0.5 0.4- -0.4 0.3- -0.3 0.2- -0.2 0.1- -0.1 0.01----l----l----l----l----l----l----l----l----l----l----l----0.0 0 5 10 15 20 25 30 35 40 45 50 55 Control Rods Withdrawn
(***** END OF CATEGORY 01 *****)
t 2i__EL6NI_QE119N_INGLUQ1NG_S8EEIl_8NQ_EdEBGENC1_SX1IEd3 PAGE 6 J
QUESTION 2.01 (1.50)
The CRD Pressure Equalizing valves (C11-PEV-F150A and B) operate to meintain or minimize the dp between the CRD exhaust header and the
- cooling water header. DESCRIBE the sequence of conditions / events in 4
the CRD system that requires these valves to actuate / operate. (1.5) i 4
QUESTION 2.02 (2.00) [
- o. Immediately following a reactor scram, the CROH flow indicator reads offscale high (>100 gpm). Is this a normal indication for '
1 this condition? If so, explain why flow indicates high. If not, j give TWO (2) possible reasons for the flow to be high. (1.0) j
- b. Not all of the CRD pump discharge flow is directed through the flow element. What three (3) pump discharge oaths are NOT directed t
{ through the flow element? (1.0) l l
l QUESTION 2.03 (2.00)
A single control rod is inserted one notch. The insert stabilizing
- valve (s) fail to reposition following the rod movement. How does j this' affect ORIVE WATER DIFFERENTIAL PRESSURE and FLOW CONTROL VALVE position? EXPLAIN YOUR ANSWER. (2.0) l f QUESTION 2.04 (1.50)
I a. What provides the normal and backup sources of air to the SRV's? (1.0)
- b. Is the changeover from normal to backup source of air a manual or automatic evolution? (0.5) t l
i i
t
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
2&__EL8NI_QE11GN_INQLUDINQ_S8EEIl_88Q_EMEBQENQ1_11SIEd3 PAGE 7 QUESTION 2.0S (2.00) i For each SRV switch position / condition below, state which mode of SRV operation will still function (use SAFETY, PRESSURE RELIEF, ADS, and g LOW LOW SET as the modes). Assume all modes of valve operation apply I to the v.sive in' question. (2.0)
- s. ' 0FF i.
- b. AUTO
- c. The valve's logic circuit is in TEST QUESTION 2.06 (1.S0)
Describe how the failure of a feedwater heater extraction steam non-return valve to shut on a turbine trip can result in over-
]
opeeding the turbine. (1.S) l I
, QUESTION 2.07 (2.00)
Indicate whether each of the following valves in the HPCS system receive on.0 PEN signal, a CLOSE signal, or N0 signal upon automatic start of the cystem. (2.0)
- e. F001 Condensate Storage Tank (CST) Suction Valve
- b. F01S Suppression Pool Suction Valve
{
j c. F023 Test Valve to Suppression Pool I d. F004 Injection Valve t
- QUESTION 2.00 (2.50)
- c. Following a LOCA, what is the only valve in the RHR system that cannot be manually overridden IMMEDIATELY following LPCI initiation? (0.S)
- b. What are two (2) indications that a RHR Inj ection Valve (F042A/B/C) has been manually overridden? (1.0) i c. What conditions / actions must be mat /taken to clear the manual override? (1.0) 4
(***** END OF CATEGORY 02 *****)
-L2t__INSIBudENI1_eNQ_CQNIBQL1 PAGE 8
. QUESTION .3.01 (2.00)
There>are four (4) arm.and depress isolation pushbuttons'(A - D) on P680 for the'MSIV's. Which-pushbutton combinations below will effect a full MSIV isolation? ( 2. 0 )'
- c. AB-
- b. AC
- o. BD
- f. CD QUESTION 3.02 (2.50)
Doscribe the response of the turbine generator to a degradation in the stator water cooling supply to the main generator. INCLUDE ANY APPLICABLE SETPOINTS in your discussion. (2.5)
QUESTION 3.03 (2.00)
- a. The RCIC flow transmitter has failed low (indicates 0 gpm). Select the answer below which best describes how the RCIC. turbine will
-respond upon receipt of an; initiation signal. (1.0)
- 1. The turbine will trip on overspeed
- 2. The turbine will run continuously at minimum (idle) speed
- 3. The turbine will run continuously at approximately 4500 rpm
- 4. The system will not start (F045 Steam Supply remains closed)
- b. Which RCIC Isolation Division (1 or 2) is provided with a manual isolation feature? (0.5)
- c. When is the manual isolation in part "b" above functional? (0.5)
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
2t__INSIBudENIS_eUQ_GQUIBQL1 PAGE 9 QUESTION 3.04 (2.50)
The reactor is operating at 100% power. Recirculation flow control to in MASTER MANUAL. The Reactor Operator mistakenly depresses the DECREASE load button on the EHC panel and holds it. Describe the response of the turbine control and recirculation flow control systems to this event. Terminate the event with a reactor scram, including the cause. (2.5)
QUESTION 3.05 (3.00)
State whether each of the following parameters directly initiate SCRAMS, ROD BLOCKS, BOTH, or NEITHER. Setpoints are NOT required. (3.0)
- 2. Scram discharge volume level
- b. MSIV position
- c. Main steam line high flow
- d. Reactor vessel high level
- o. Recirculation flow
- f. Reactor pressure QUESTION 3.06 (2.00)
- c. During a control room fire, efforts should be focused on using Divis ion (I, II, III) equipment to put the station in a safe shut-down condition. (0.5)
- b. How is control of a component transferred to the Remote Shutdown Panel from the control room? (0.5)
- c. What are two (2) indications in the control room that would indicate that transfer of a valve's control to the Remote Shutdown Penel had been performed? (1.0)
(***** END OF CATEGORY 03 *****)
4 A__EBQQEQ9 bel _=_UQBd6Lt_6BNQBd6Lt_EUEBGEUQ1_6NQ PAGE 10 B6010LQQ1Q6L_QQUIBQL QUESTION 4.01 (1.50)
Answer the following with regard to GOP-0001, Plant Startup:
O. When is the reactor determined to be critical? (1.0)
- b. Control rods are withdrawn until the reactor is critical and a steady, reasonable period is achieved. Per the referenced procedure, what is a steady, reasonable period? (0.5)
QUESTION 4.02 (2.50)
Briefly state the reason for each of the following statements portaining to a reactor / plant shutdown from power:
- o. The turbine generator must remain on the line until Reactor Thermal Power is below 10%. (0,5)
- b. Maintain steam packing exhauster and air removal pumps in operation for up to four (4) hours after breaking vacuum. (0.5)
- c. Small control rod movements are essential when operating with the reactor critical and the MSIV's closed. (0.5)
- d. The main turbine must be on the turning gear whenever steam seals are on or turbine metal temperatures are above 150 deg F. (0.5)
- o. Do not allow RPV pressure to decrease to O psig with the MSIV's open and condenser vacuum > or = 20 Hg. (0.5)
QUESTION 4.03 (1.50)
What do the words "shall", "should", and "may" denote when they are ueed in River Bend procedures? (1.5)
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
at__EB9CEDVBE1_:_UQBdeLi_6BUQBdakt_EUEBGENGl_6NQ PAGE 11 860196901G86_00UI896 QUESTION 4.04 (1.50)
An irradiated component is surveyed with a portable instrument. The epen window reading is 2.65 Rem /hr and the closed window reading is 1.75 Rem /hr. SHOW ALL WORKl
- c. What is the beta dose rate at the instrument? (0.5)
- b. How long could you remain at the survey distance without exceeding administrative and 100FR20 quarterly exposure limits? (1.0)
QUESTION 4.05 (1.50)
Instrument Air header pressure is decreasing. All systems are re-cponding properly to the decreasing air pressure. Besides keeping the Shift Supervisor informed of plant conditions, what actions should bo taken by the control room operator (s) and at what point (if applicable) por AOP - 0008, Loss of Instrument Air? (1.5)
QUESTION 4.06 (1.00)
What two (2) conditions must be met so that an IMMEDIATE reactor ecram is NOT required during a loss of Normal Service Water? (1.0)
QUESTION 4.07 (2.50)
- 9. With the facility at power, what actions should be taken prior to evacuating the control room? (2.0)
- b. Where should the control room personnel assemble following the evacuation? (0.5)
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
I
- dz__EBQQEQUBE1_ _NQBd8Lt_8BNQBd8Lt_EMEBQENQ1_8NQ PAGE 12 88DIQLQQ198L_CQNIBQL QUESTION 4.08 (3.00)
For each of the following conditions, state'which E0PCs), if any, should-bo applicable. (3.0) c.- Reactor water level 12.5 "
- b. Drywell pressure 1.87 psig c.- Suppression pool temperature 97 deg F
- d. Drywell to containment dp 27.3 psid
- o. Containment temperature 88 deg F i
l l
(***** END OF CATEGORY 04 *****)
(************* END OF EXAMINATION ***************)
EQuATI0tt SHEET f = ma v = s/t Cycle efficiency = (.4et work out)/(Energy in) 2 w = ag s = V ,t + 1/2 at 2
E=m ,
KE = 1/2 av 2
, , gyf ,y g)fg g , g, g , gg,-At PE = agn Vf = V, + at w = e/t A = an2/t1/2 = 0.693/t1/2 y . ,3p A"
,o 2 ti/r = WJW
! 4 [(t1/2I
- II DI3 aE = 931 am
- m = V,,Ao ,
-Ex Q = aCpat I = 1 ,e~"*
6 = UA AT Pwr = Wfah I = I ,10-x m L TVL = 1.3/u 8
' P = P*10 "'III HVL = -0.693/u f ,p o,t/T SUR = 26.06/T SCR = S/(1 - K,ff)
CR,=S/(1-K,ff,)
SUR = 26e/t* + (s - e)T CR(1-K,ffj)=CRII~Ieff2) j 2 i
T = (t*/s) + [(s - sy Ie] M = 1/(1 - K,ff) = CR j/CR, l T = s/(e - s) M = (1 - K ,ff,)/(1 - K,ffj) j T = (s - o)/(Is) SDM = ( - K,ff)/K,ff a = (K ,ff-1)/K,ff = AK,ff/K,ff t= = 10 seconds
- I = 0.1 seconds ~I l e = [(t*/(T K,ff)] + [a,ff /(1 + IT)]
l l id; = I d Id 2 ,g gd2
! P = (seV)/(3 x 1010) jj 22 I 1 = eN R/hr = (0.5 CE)/d (meters)
R/hr = 6 CE/d2 (f,,g) i Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lbm. 1 curie = 3.7 x 1010dps 1 ga:. = 3.78 liters I kg = 2.21 lbm 1 fta = 7.48 gal. I hp = 2.54 x 103 Stu/hr Density = 62.4 lbrp/ft3 1 m = 3.41 x 100 Stu/hr Density = 1 gm/c9 lin = 2.54 cm Heat of vaporization = 970 Stu/lem *F = 9/5'C + 32 Heat of fusion = 144 Bru/lbm 'C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. Hg. I BTU = 778 ft-Ibf I ft. H O 2
= 0.4335 lbf/in.
it__EBINQ1 ELE 1_QE_NVQLE88_EQWEB_EL8NI_QEEB8110Nt PAGE 13 IHEBdQQINad101t_HE8I_IB8NSEEB_8NQ ELVIQ_ELQW ANSWERS -- RIVER BEND 1- -86/11/04-GRAVES, D.
ANSWER 1.01 (1.50)
- o. CRITICAL POWER increases (0.50).
- b. CRITICAL POWER increases (0.5).
- c. CRITICAL POWER decreases (0.5).
REFERENCE LOTM-TH-3.7-0, Transition Boiling, pg 3 ANSWER 1.02 (2.00)
- c. Saturated
- b. Subcooled
- c. Superheated
- d. Subcooled (0.5 each)
REFERENCE Steam Tables ANSWER 1.03 (2.00)
- c. increase (0.5)
! b. 1. decrease (0.5)
- 2. decrease (0.5)
- 3. increase (0.5)
REFERENCE LOTM-TH-1.4-0, Pumps, pg 3, 17 l
\
I i
I
l l
It__EBINDIELER_QE_NUQLE68_EQWEB_EL8NI_QEEBoIIQNt PAGE 14 IBEBdQQ1N80101t_UE81_IB8NSEEB_eUQ_ELUID_ELQW ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 1.04 (2.50)
- o. Four fuel bundles and a control rod (0.5)
- b. Fully withdrawn (0.5)
- c. All control rod movements are associated with low reactivity !
cells (1.0). j
- d. True (0.5)
REFERENCE i LOTM-2-1, Nuclear Fuel, pg 20-23 1 l
ANSWER 1.05 (1.50) l Void fraction decreases (0.5) l Doppler reactivity increases r0.5)
Enthalpy increases (0.5)
REFERENCE l LOTM-TH-4.18-1, Void Coefficient !
LOTM-TH-4.19-1, Doppler Effect LOTM-14b-0, Feedwater Heaters and Hester Orains l
1 ANSWER 1.06 (2.00)
- o. As the reactor operates at power, Xenon builds in to equilibrium (0.5), I adding negative reactivity, causing power to decrease (0.5). l
- b. In 40-50 hours (0.5) when equilibrium Xenon is reached (0.5).
Also accept approximately 20 days, when equilibrium Samarium is reached.
REFERENCE LOTM-TH-4.22-1, Xenon l
l l
I i
l
li__EBINGIELE1_QE_UUCLE68_EQWEB_EL6HI_QEEB6IIQUt PAGE 15 IBEBdQQ188 digit BE81_IB6USEEB_eUQ_ELul0_ELQW ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 1.07 (1.50)
Withdrawal of a center control rod at 90% density has greater worth.(0.5)
The control rod worth is proportional to the (local neutron flux / the core Cverage neutron flux) squared.(0.25)
With g0% rod density the core average neutron flux is very small.
Withdrawing a central control rod, increases the local flux in the area of withdrawn rod substantially. Because the rod causes the value of the torm (local neutron flux / core average neutron flux) squared to be large its worth for this condition is quite high. Higher than withdrawing the rod at 40% rod density, when core average flux will be higher. (0.75)
REFERENCE LOTM-TH-4.20-1, Control Rod Worth, pg 4 i
I i
l l
l l
1
.li-_EBINQlELEl-QE_NWQLE88 EQWEB EL8NI_QEEB8IIQNi PAGE 16 i IBEBdQQ1NedIQ1t_UE8I_IB8NSEEB 880_ELU1D_ELQW ANSWERS - . RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 1.08 (2.00)
Drawing a straight line between the last two *'s predicts 34-35 control rods must be withdrawn.
(0.1 for each point plotted, 1.3 for line and prediction) 5 10 15 20 25 30 35 40 45 50 55 1.0*----l----l----l----l----l----l----l----l----l----l----l----1.0 0.9- -0.9 0.8- *' -0.8 0.7-- -0.7 1/M 0.6- -0.6 0.5- -0.5 0.4- -0.4 0.3- * -0.3 0.2- * -0.2 0.1- * -0.1
- w -
0.01----l----l----l----l----l----l----l----l----l----l----l----0.0
, 0 5 10 15 20 25 30 35 40 45 50 55 i Control Rods i
! REFERENCE
! LOTM-TH-4.13.1, Subcritical Multiplication and Count Rate Comparison 1
i i ANSWER 1.00 ( .00)
- NOT APPLICABLE e
i
(
i l
t
f' t
.. f e . ua ** '.
2i__EL6NI_ DESIGN _INGLUDINQ_SeEEII_8NQ EdEBGENGl_1111Ed3 - PAGE $7
_ (-
ANSWERS -- RIVER BEND.1 -86/11/04-GRAVES, D. s;
{j,j u A i < (h v? ya m [ I' .
r wp\
1 .'-
ANSWER 2.01 (1.50)
,s d,<9, Y.
On a scram, the exhaust water header will be at a greater preds. ure
.than the scram discharge header (0.5). This causes the exhauV,t water header to depressurize to the scram discharge header (0.5) via the ,
t +3/i HCUs' 121 valves (0.5). The Pressure Equalizing Valves quickly ropressurize the exhaust header from the cooling wateg header. *,
) '
t 3
REFERENCE z,4 LOTM-4-1, CRDH System, pg 8-9 q;!
A.
- 4. ..
y ANSWER 2.02 (2.00) , il. ' '
O. yes (0.5). CRD system flow increases as the scram accumulators ,6 recharge (0.5). If examinee answers NO, allow 0.75 for each of ,
two possible causes of high flow. ,,
- b. CR0 pump miniflow (0.4) h '
Recirc pump seal flow (0.4) ' "
,' f Sample flow (0.2) 4 REFERENCE LOTM-4-1, CRDH System, pg 4 4 . i
- I, .
j ANSWER 2.03 (2.00) ,
, { ,.
? l "
Flow control valve position does not change (0.5) tecaus) total systen '
v flow has not changed (0.5). Drive water dif f erential pressure will
- increase (0.5) because flow previously bypassing the pressura cor3r.%1 volve is now passing through it (0.5). Also accept that the FCV yp~ ens clightly to maintain flow due to slightly increased backpressure.
i REFERENCE , //
LOTM-4-1, CRDH System, pg 16-17 . , ,,
o ,,s.
(
i
, m .
1 v ;
ANSWER 2.04 (1.50) M' '
- ! a. normal SRV air compressors (0.5) .
backup! Penetration Valve Leakage Control SysiUJFAir Compressp,do ;
(0.5) .
i %
- b. automatic (0.5)
%r, e
] _, w - w n ., --.~-~,,gw, .r ,w,.----,~-,
aix '
,l
'21__ELANI_DE119N_INCLUDINQ_18EEI1_860_EMEBQEUGl_SYSIEd3 PAGE 18
(~
AbsWERS';q>'RhVERBEND1 hf -86/11/04-GRAVES, D.
L'- X REFERENdE LOTM-7-1) Mpin Steam, pg 4, 41
?>?
AN3WER 2.05 (2.00)
- 6. Safety, ADS b >/ 1'arety, Pressure Relief, ADS, Low Low Set ej J.Saf ety' (accept ALL if noted that only 1 logic circuit is in test)
,J4.285 each,'2.0 total)
.i ,.
REFERENCE '
LOTM-7-1, MUin Steam, pg 5-7, 16 (J
ANSWER 2.06 (1.50)
When the turbine trips, pressure in the turbine decreases to condenser pressure (0.5). The condensed extraction steam flashes to steam (0.5),
cnd backflows up the extraction steam line through the turbine (0.5), which ccn cause it to overspeed.
, i REFESENCE L,0 T M-14 b-[, i feedwater Heaters, Vents, and Drains, pg 17 s
ANSWER LI . 07 (2.00) r ,
o./ Open; .
,, bi r P!o 6.' Close
- d. ' Opes 3, (0.5, inch) ,-
REFERENCR LOTM-18-0, HPCS System, pg 30-31 sa
<r,
\s 4,* ( / <-
,' s F, ~, J,
,h f k ae l *S
2t__ELsNI_QESIGN_INGLUQ1NQ_18EEIY_8NQ_EMEBGENC1_111IEd3 PAGE .19 ANSWERS -- RIVER BEND 1 -86/ll/04-GRAVES, D.
ANSWER' 2.08' (2.50)
- See the valve logic tables in the. stated reference.
b.' - amber light.above the valve switch (0.5)
- annunciator (RHR Injection Valve in Man-Override) (0.5)
~
- c. LOCA signal clear (0.5)
- Initiation Reset pushbutton(s) depressed (0.5) 5' REFERENCE LOTM-20-0,. RHR System, pg.9 J:
I 3 .
l t
t.
4 4
1 1
<2i__INSIBudENI1_8ND_GQNIBQL1 PAGE- 20 ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER ~3.01 (2.00) e, c, d ,f C0.5 each)
' REFERENCE .
LOTM-7-1, Main Steam, pg 17 ANSWER 3.02 (2.50)
A turbine runback will occur (0.5) at 13 psig stator inlet pressure (0.25),
Si deg C stator outlet temperature (0.25), or low flow of 465 gpm. If load .
->80% (0.25] after 2. minutes (0.25) or.>24.5% (0.25)after 3.5 minutes (0.25) the turbine trips (0.5). Low flow is accepted but1not required due to a rocent plant modification.
REFERENCE'
'LOTM-11-1, Main Generator Auxiliaries, pg 3, 10-11 4
ANSWER 3.03 (2.00) e.- 3 (1.0)
- b. 1 (0.5) lc. initiation signal must beLpresent (0.5)
REFERENCE LOTM-21-0,-RCIC System, pg 7, 13, 37, 38 ANSWER 3.04 (2.50)
As'the load set motor runs down, the EHC system will be calling for the control valves to close to reduce laad, which they will do (0.5).
l An the CV close, the turbine bypass valves will begin to open to maintain pressure (0.5). Once the BPV are fully open and the CV sre still shutting, reactor pressure and power will begin to increase (0.5) causing the recirculation flow control valves to close down in an
, attempt to keep reactor power at 100% (0.5). This continuce until the reactor scrams on high pressure or power (0.5)
REFERENCE LOTM-IC-8-0, Recirculation Flow Control, pg 15, 23 LOTM-IC-9-0, EHC, pg 85-86, 114 t
9 ysy ..-m-,- ,, r- . gm-v.,,-m-- -y--y- - , --- ,y -~.s .,-
y%, , . %m ,.,,re
- 4 -- w,,--m .--, , ,,-m _
~
J2i__INSIBudENI1_8NQ_QQNIBQL1 PAGE 21 ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES,-D.
'l ANSWER 3.05 (3.00) c.
~
both
- b. scram
- c. neither.
- d. scram
- o. ~ rod block
- f. scram (0.5 each)
REFERENCE 4
LOTM-IC-11-0, RPS, pg'16 ,
lLOTM-IC-4-0, APRM, pg 8 River Bend Technical. Specifications, Table 3.3.6-2 ANSWER 3.06 (2.00)
- e. I~(0.5)
- b. ~By taking the Remote Shutdown Transfer Switches to the EMERGENCY position (0.5).
- c. Annunciator (s) in the control room (0.5)
. Loss of position indication for that valve in the control room (0.5).
- i. REFERENCE LOTM-IC-16-0,-Remote Shutdown, pg 7-8 4
f 1
i
- - s- , y e e ,,w- ,,m
,m.w, -- ---.-ee,-,m.4 y - e .--.e ,
'dz__EBQQEQMBE1_:_NQBd8Lt_6BNQ858Lt_EME8QEUQX_68Q PAGE 22 88QIQLQQIQeL_QQNIBQL ANSWERS 1-- RIVER BEND'l -86/11/04-GRAVES, D.
ANSWER 4.01 (1.50)
- a. The reactor is critical when the neutron count rate increases (0.5) at an exponential' rate without further control rod motion (0.5),
- b. Accept 50 - 150 seconds (0.5)
REFERENCE GOP-0001, Plant Startup, Rev 6, pg 20 ANSWER 4.02 (2.50)
- a. Turbine BPV are only rated at 10% total steam flow. Will also accept to prevent a high pressure scram. (0.5)
- b. High airborne _ activity in the turbine building can result (0,5) if vacuum is' broken within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from operation at full power.
'c. There is no automatic pressure control (0.5).
- d. To prevent warping the shaft (0.5).
- e. Prevent flashing the reference legs (0,5) on-the RPV instruments.
REFERENCE River Bend GOP 0002, Power Decrease / Plant Shutdown, Rev 3, pg 8-10 ANSWER 4.03 (1.50) chall - denotes a requirement (0.5) chould - denotes a recommendation (0.5) may - denotes permission (0.5)
REFERENCE River Bend ADM 0003, Development, Use, and Control of Procedures, Rev 10, pg 5
st__EBQGEQUBE1_=_NQBdekt_6BNQBdekt_EdEBGENQ1_8ND PAGE 23
~B8010LQQ198L_QQNIBQL ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 4.04 (1.50)
- o. Beta dose rate = open - closed window = 0.9 Rem /hr (0.5)
Also accepted 3.6 Rem /hr -if Beta correction factor of 4 used.
- b. ' Stay time: 10CFR20: '1.25 R(1mts)/1.75 R/hr = .71 hr = 43 minutes (0.5)
Admin.: 1.00 RC1mts)/1.75 R/hr = .57 hr = 34 minutes (0.5)
REFERENCE GET II, p. 4 ANSWER 4.05 (1.50)
When instrument-air header pressure has decreased to 65 psig (0.5) or it becomes apparent that instrument air will not be restored (0.5), scram the reactor (0.5).
REFERENCE River Bend AOP - 0008, Loss of Instrument Air, Rev 2, pg 3
-ANSWER 4.06 (1.00)
- The main turbine.is NOT on the line (0,5).
- Turbine BPV are NOT being used to control reactor pressure (0.5).
. REFERENCE River Bend AOP 0009, Loss of Normal Service Water, Rev 2, pg 4 ANSWER 4.07 (2.50) l e. 1. Arm and depress all four manual scram buttons.
- 2. Mode Switch to SHUT 00WN.
.3. Verify control rods fully inserted.
. 5. Close the MSIV's (5'at 0.4 each)
- b. Division I Remote Shutdown Panel (0.5)
4 t__EB9CEQUBE2_:_UQBd6Lt_6HUQBdebt_EUEB9ENGl_aNQ PAGE 24 88DIQLQQ1G6L_G90IB9L ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
REFERENCE -
A0P-0031, Shutdown from Outside the Control Room, Rev 3, pg 6 ANSWER 4.08 (3.00)
- c. none
- b. E0P-0001, RPV Control E0P-0002, Primary Containment Control
- c. E0P-0002, Primary Containment Control
- e. none (6 at 0.5 each)
. REFERENCE River Bend E0P's 0001, Rev 6; 0002, Rev 5; 0005, Rev 3
U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION FACILITY: _BlyEB_BENQ_1____________
REACTOR TYPE: _RWB-QEq_________________
DATE ADMINISTERED: _ghtlllq4________________
EXAMINER: _QB6ME5t_Qz______________
CANDIDATE: _________________________
INSIBuQIl0N5_IQ_QaNQ1061El Road the attached instruction page carefully. This examination replaces the current cycle facility administered requalification examination.
Rotraining requirements for failure of this examination are the same as for failure of a requalification examination prepared and administered by your training staff. 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 four (4) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY
__V8LME_ _IQI6L ___500BE___ _V8LUE__ ______________QoIEQQBY
_15tDQ__ _251Q0 ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS
_15zQQ__ _25tQQ ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION
_15tQQ__ _251QQ ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL
_15tDQ__ _25tQQ ___________ ________ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS
_kQtDQ__ ___________ ________% Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.
Candidate's Signature
NRC RULES AND GUIDELINES FOR' LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
- 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
- 2. 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 ap p e a.* a n c e or possibility of cheating.
- 3. Use black ink or dark pencil 201y to facilitate legible reproductions.
- 4. Print your.name in the blank provided on the cover sheet of the examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of ggeb
-section of the answer sheet.
'8.
Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a Dgw page, write SDAY 2D 2DR 11dt of the paper, and write "Last Page" on the last answer sheet.
- 9. . Number each answer as to category and number, for example,-1. 4, 6.3.
- 10. Skip et least 1btgg lines between each answer. ,
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use abbreviations-only if they are commonly used in facility litstglyng.
l 13. The point value for each question is indicated in parentheses after the i question and can be used as a guide for the depth of answer required.
- l. 14. Show all calculations, methods, or assumptions used to obtain an answer l to mathematical problems whether indicated in the question or not.
l L 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND 00 NOT LEAVE ANY ANSWER BLANK.
l 16.-If parts of the examination are not clear as to intent, ask questions of the REAmiD2n only, i
j 17. You must sign the statement on the cover sheet that indicates that the l 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 I been completed.
1 i
- , ..-n- , . . , , - . . - . , , , . , , , . , . - -
18.;When:you complete:your examination, you shall:
- s.' - Assemble your examination as follows:
(1) Exam questions on top.
(2) Exam--aids - figures, tables, etc.
(3) . Answer pages including figures which are part.of the answer.
b.- Turn in your copy of the examination and all pages used to answer the examination-questions, c .. Turn in all scrap paper and the balance of the paper that.you did not use for answering the questions.
- d. Leave the examination area, as defined by the examiner.. If after leaving,.you are found in this area while the examination is still in progress, your license may be denied or revoked.
I l-i I
i l
l
-. _, . - . . - . . . . - . . _ . . , . - , ,.. ... , ,.m .. . . . - . . - . . _ _
51__IMEQBl_QE_NUGLE88_EQWEB_EleNI_QEEB8110Nt_ELVIDS t_6HD
'PAGE 2 IBEBdQQ1N851GS QUESTION 5.01. (2.00)
Dcscribe HOW and WHY each of the following changes will affect roactor CRITICAL POWER. If CRITICAL POWER will not be affected, etate this,
- c. Loss of extraction steam to a feedwater heater (1.03'
- b. Mass flow rate through the core is increased (1.0)
QUESTION 5.02 (2.00)
A reactor water sample indicates the following isotopes present:
Tritium (1H3) Kr 87 Mn 56 I 131 Ni 57 Xe 133-Fe-59 Cs 137 Co 60 Ba 140
- a. Which.of the above are fission products and which are corrosion products? .(1.0)
- b. Does the presence of fission products in the coolant mean a fuel element failure or defect exists? If not, account for their presence in the reactor coolant. (1.0)
QUESTION 5.03 (1.00)
ARRANGE the following in order of INCREASING heat transfer coef-ficient (Lowest Coefficient to Highest Coef f icient) . (1.0)
.a. Free Convection, Air l
- b. Boiling Water (Free Convection)
- c. Boiling Water (Forced Convection)
- d. Forced Convection, Water
- e. Forced Convection, Air i
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
St_aIBEQBl_QE_UUQLE68_EQWEB_EL6NI_QEEB6Il0Nt_ELu1Dat_8ND PAGE 3 IBEBdQQ1Ned101 QUESTION 5.04 (1.00)
A reactor heat balance was performed (by hand) during the midnight chift due to the Process Computer being 000. The GAF's were computed, but the APRM GAIN ADJUSTMENTS HAVE NOT BEEN MADE.
Which of the following statements is TRUE concerning reactor power? (1.0)
SELECT ONLY ONE ANSWER COnly one is truel)
- a. If the feedwater temperature used in the heat balance calcu-lation was LOWER than the actual feedwater temperature, then the actual power is HIGHER than the currently calculated power.
- b. If the reactor recirculation pump heat input used in the heat balance calculation was OMITTED, then the actual power is LOWER than the currently calculated power.
- c. If the steam flow used in the heat balance calculation was LOWER than the actual steam flow, then the actual power is LOWER than the currently calculated power.
- d. If the RWCU return temperature used in the heat balance cal-culation was HIGHER than the actual RWCU return temperature, then the actual power is LOWER than the currently calculated power.
QUESTION 5.05 (2.50)
Answer the following with regard to the Control Cell Core (CCC) operating strategy:
- c. What constitutes a CONTROL CELL? (0.5) b .' While operating at power, what is the position of.the non-control cell control rods? (0.5)
- c. Why do control rod movements result in a lower kw/ft change per notch in a CCC than in a conventional core? (1.0)
-d. TRUE or FALSE. Using the CCC operating strategy eliminates the need for control rod pattern exchanges. (0.5)
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
. .~. - - _ - - - , _ - _ - _ . . - - - _ , - -
5t__IBEQBl_QE_UUQLEoB_EQWEB_ELeUI_QEEBeI10Nt_ELUIQSt_6UQ PAGE 4 IBEBdQQ1N8dIQ1 QUESTION 5.06 (3.00)
Dascribe (INCREASE, DECREASE, OR UNCHANGED) the effect on void fraction, doppler reactivity and feedwater enthalpy due to a recirculation flow rcte increase at power. EXPLAIN WHY for each change. (3.0)
QUESTION 5.07 (1.50)
Concerning control rod worth, compare withdrawing a center control rod at 90% rod density to withdrawing a center control rod at 40% rod density. In which situation is control rod worth 9reater for the withdrawn control rod? Explain your answer. (1.5)
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
Ei__IBEQBl_QE_N90LE88 EQWEB_EL8NI_QEEBoIIQNt_ELu1Q1t_8NQ PAGE 5 IBEBdQQ1N85101
, QUESTION 5.08 '(2.00)
Use;a 1/M plot and PREDICT the number of control rods required to be withdrawn to_ achieve criticality.
NOTES: 1. CR = Count Rate ~
2.- USE THE FIGURE AT THE END OF THE SECTION AND INSERT IN YOUR-ANSWER PACKAGE
'CR0 = 40 cps CR4 = 191-cps CR1 = 50 cps CR5 = 333 cps CR2 = 89 cps CR6 = 800 cps CR3 = 129 cps Each CR reading is recorded following a 5 rod withdrawal with CR0 being 100% rod density.
5 10 15 20 25 30 35 40 45 50 55 1.0-----l----l----l----l----l----l----l----l----l----l----l----1.0 0.9- -0.9 0.8- -0.8 0.7- -0.7 1/M 0.6- -0.6 0.5- -0.5 0.4- -0.4 0.3- -0.3 0.2- -0.2 0.1- -0.1 0.01----l----l----l----l----l----l----l----l----l----l----l----0.0 -
, 0 5 10 15 20 25 30 35 40 45 50 55 Control Rods Withdrawn L
i l
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
s
'St_ IBEQBY-QE NVQLE68 EQWEB_EL6HI-QEEBoI19Nt ELu1 dst eNQ PAGE 6 IBEBdQQ1Ned1GS
? QUESTION 5.09' ( .00)
Une a.1/M~p1'ot and PREDICT the number of control rods required to be (11thdrawn-to achieve eriticality.
NOTES: l '. CR = Count Rate
- 2. USE THIS FIGURE'AT THE END OF THE SECTION AND INSERT I'N.YOUR F^ ANSWER PACKAGE
'CR0 = 40 cps CR4 = 191 cps CR1 = 50 cps CR5 = 333 cpe CR2 = 89-cps CR6 = 800 cps CR3 = 129 cps -
Each CR reading is recorded following a 5 rod withdrawal with CR0 being .
100% rod density.
5 10 15 25 30 35 40 45 50 55 1.0-----l----l----l----l----l----l----l----l----l----l----l----1.0 0.9- -0.9 0.8- -0.8 0.7- -0.7 4 - -
.1/M 0.6- -0.6 O.5- -0.5 0.4- -0.4 0.3- -0.3 0.2- -0.2 t- - -
l-0.1- -0.1 0.01----l----l----l----l----l----l----l----l----l----l----l----D.0 0 5 10 15 20 25 30 35- 40 45 50 55 .
Control Rods Withdrawn i
r l
I l
l l
(***** END OF CATEGORY 05 *****)
t-
, _ , . . , , . _ . . . , , . . . . . , - . _ , - . .,c ,.
...a_.____,_--_.__.___...,__., - - _ . . . , , _ _ _ _ _ _ _ _ , . . _ _ _ . . , _ _ _ _ _ _ _ _ _ , . . . . _ . , _ _ _ _ , . _ , , . . . . _ - - - _ ,
4t__EL8HI_111IEd1_QE110Nt_QDNIBQLt_8ND_1NSIBudENI611QN PAGE 7 QUESTION 6.01 (2.00)
A. single control rod 19 in=ar+ad ona notch. The insert stabilizing o vetvels) fail to reposition following the' rod movement. How does this' affect DRIVE WATER DIFFERENTIAL PRESSURE and FLOW CONTROL VALVE position? EXPLAIN YOUR ANSWER. (2.0)
QUESTION 6.02 (2.00)
For each SRV switch position / condition below, state which mode of SRV operation will still function (use SAFETY, PRESSURE RELIEF, ADS, and LOW LOW SET as the modes). Assume all modes of valve operation apply to the valve in ouestion. -(2.0)
- c. OFF
- b. AUTO c.- The valve's logic circuit is in TEST QUESTION 6.03 (1.50)
Describe how the failure of a feedwater heater extraction steam non-return valve to shut on a turbine trip can result in over-
. speeding the turbine. (1.5)
QUESTION 6.04 (2.50)
- c. Following a LOCA, what is the only valve in the RHR system that cannot be manually overridden IMMEDIATELY following LPCI initiation? (0.5)
- b. What are two (2) indications that a RHR Inj ection Valve (F042A/B/C) has been manually overridden? (1.0)
- c. What conditions / actions must be met /taken to clear the manual override? (1.0)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
- ht__EL8HI_SYSIEUS_DESIGNi_C9NIBQLt_8NQ_INSIBudENI811QN PAGE 8 QUESTION 6.05 (2.00)
- a. : The RCIC flow transmitter has failed low (indicates-0_gpm). Select
-the answer below which best describes-how the RCIC turbine will respond upon receipt of an initiation signal. (1.0)
- 1. The turbine will trip on overspeed
- 2. The turbine will run continuously at minimum (idle) speed
- 3. The turbine will run continuously at approximately 4500 rpm
- 4. The system will not start (F045 Steam Supply remains closed)
-b. Which RCIC Isolation Division (1 or 2) is provided with a manual isolation feature? (0.5)
- c. When is the-manual isolation in part'"b" above functional? (0.5)
QUESTION 6.06 (3.00)
- s. What ' rod movement. restraints are in force as a result of exceeding the LPSP and HPSP during power escalation? (1.0)
- b. At what power levels do the LPSP and HPSP take effect? (1,0)
- c. Why is control rod withdrawal undesirable if power is above the LPSP and a Main Turbine Bypass Valve-is open? (1.0)
QUESTION 6.07 (2.00)
- a. During a control room fire, efforts should be focused on using Division (I, II, III) equipment to put the station in a safe shut-down condition. (0.5)
- b. How is control of a component transferred to the Remote Shutdown Panel from the control room? (0.5)
- c. What are two (2) indications in the control room that would indicate that transfer of a valve's control to the Remote Shutdown Panel had been performed? (1.0)
(***** END OF CATEGORY 06 *****)
Zi__EBQQEQUBES_:_UQBd8Lt_oBUQBdekt_EUEBGEUQ1_60Q PAGE 9 BoQ19LQQlCoL_QQUIBQL QUESTION 7.01 (2.50)
Given the following cooldown data, DETERMINE if the RPV cooldown rete limit has been exceeded. STATE the RPV cooldown limit and HOW the determination was made. IF the limit was exceeded, STATE the period (s) of time the limit was exceeded. (2.5)
TIME TEMPERATURE (Deg F)
START 0800 520 0815 490 0830 460 0845 445 0900 430 0915 400 0930 370 0945 340 1000 325 1015 310 1030 275 1045 250 1100 225 1115 205 QUESTION 7.02 (2.50)
Briefly state the reason for each of the following statements portaining to_a reactor / plant shutdown from power:
- a. The turbine generator must remain on the line until Reactor Thermal Power is below 10%. (0.5)
- b. Maintain steam packing exhauster and air removal pumps in operation for up to four (4) hours after breaking vacuum. (0.5)
- c. Small control rod movements are essential when operating with the reactor critical and the MSIV's closed. (0.5)
- d. The main turbine must be on the turning gear whenever steam seals are on or turbine metal temperatures are above 150 deg F. (0.5)
- c. Do not allow RPV pressure to decrease to O psig with the MSIV's open and condenser vacuum > or = 20" Hg. (0,5) l
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
e
s.
/
3 Zi__EBQQEQUBER_:_NQBd8Lt_8BNQBd8Lt_EMEBGENQX_8NQ. PAGE 10 88Q10LQQ1Q8L_GQNIBQL-QUESTION 7.03 (l'.50)~
Instrumont Air header pressure is decreasing. All systems are re-
- cponding' properly to the decreasing air-pressure. Besides keeping
-the. Shift-Supervisor informed of' plant conditions, what actions should be taken by the control room operator (s) and at what point (if applicable) per AOP - 0008, Loss of Instrument Air?
~
(1.5) 4 QUESTION 7.04 (1.00)
'What two-(2) conditions must be met so that an IMMEDIATE reactor scram is NOT required during a loss of Normal Service Water? ( 1. 0 3 -
- QUESTION 7.05 (2.50)
. -e. With the facility at power, what actions should be taken prior to ,
P ovacuating the-control: room? (2.0)
- b. Where'should the control room personnel assemble following the evacuation? (0. 5)'
- _ QUESTION 7.06 (3.00)
!= For-each of the following conditions, state which E0PCs), if any, should bo; applicable. (3.0)
- c. Reactor water level 12.5 "
- b. -Orywell. pressure 1.87 psig 1
- c. Suppression pool temperature 97 deg F d.~ Drywell to containment dp 27.3 psid e :e.. Containment temperature 88 deg F i
I
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
y g-e, y- e---y ,.-----+,-----we,yy --.y--,--.-ve- e, +s-e..-=,w--w-, ,y--, -..v--..- -.,er, ,wy- . . . . , - - ,-em - - , - , - - . . ,.-wi- * - , ----<=ww--e,
I .
LZi__EBQQEDMBER_:_NQBdekt_6BNQBdakt_EMEBQENQ1_eND PAGE 11 B8DIQLQQIQaL_QQNIBQL QUESTION 7.07 (2.00)-
A 24 year old individual, with a lifetime exposure of 23 REM.through the' last quarter, will be working in a 200 MREM /hr gamma field. In addition to the lifetime dose, the individual has received 850' MREM this quarter.
- m. How long may this individual work in this area before he/she reaches a radiation exposure limit? (1.0)
- b. 'Whose recommendation (s) and approval (s) are needed to allow the individual to exceed this limit? (1.0) 1 4
a e
3
(***** END OF CATEGORY 07 *****)
1 At__8Dd1N11188I1YE_EBQGEQUEEli_QQNQ1Il0NSt_8ND_LidlleI1QNS PAGE 12 )
1 1
QUESTION 8.01 (2.00) l Indicate whether each of the following is considered a " core alteration" por the River Bend Technical Specifications: (2.0) i
- c. Withdrawal and insertion of an SRM detector to check the drive motor I
- b. Removal of an LPRM string for replacement
- c. Removal of an uncoupled control rod for replacement
- d. Removal of a control rod's position indicator probe for repair l
QUESTION 8.02 (2.50)
Indicate whether the following statements are TRUE or FALSE. (2.5) l
- a. A Shift Supervisor may concurrently fill the position of the STA while on shift.
- c. All core alterations must be directly supervised by a licensed Senior Operator (or Senior Operator Limited to Fuel Handling). '
- d. An Operator license is required for an operator to perform a core alteration.
- e. During Operational Condition 4 or 5, an individual with a valid i Operator license may be designated to assume the control room command function during an absence of the Shift Supervisor from the control room.
QUESTION 8.03 (1.00)
List five (5) agencies that must be notified within 15 minutes after o General Emergency is declared at River Bend. (1.0)
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
at__eQUINISIBeIIVE_EBQQEDUBEnt_QQUDIIIQUnt_eUQ_LIMIIeIIQN3 PAGE 13 QUESTION 8.04 (2.50)
Answer the following with regard to the River Bend Emergency Plan Implementing Procedures:
- e. When shall of fsite radiation dose calculations be initiated? (1.0)
- b. Who is responsible for the initial and follow-up offsite dose calculations (at least until relieved by personnel in the TSC)? (0.5)
- c. A determination has been made that offsite dose calculations are necessary.
- 1. When should the initial assessment be completed? (0.5)
- 2. When should follow-up calculations be completed? (0.5)
QUESTION 8.05 (2.50)
- a. If an individual performing an activity cannot or believes he/she should not follow the procedure governing that activity, what two (2) actions should be taken? (1.0)
- b. What two (2) methods of resolving the procedural discrepancy are available? (0.5)
- c. A safety-related valve position checklist is being performed with the initial performer and independent verifier performing their functions simultaneously. Is this satisfactory? EXPLAIN. (1.0)
QUESTION 8.06 (2.00)
- c. Where would an operator find the interpretation on a Technical Specif ication requirement that had required previous interpretation? (1.0)
- b. When is the operator allowed to enter an operating condition where an LCO cannot be met without relying on provisions of an action statement? (1.0)
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
- Az__8Dd1NISIB8IIVE_EBQQEQUBEft_QQNQ1Il0Nat_oNQ_L1011611QU2 PAGE 1 44
, QUESTION 8.07 (1.50)
Answer the following with regard to temporary-procedure changes per the Tcchnical Specifications.
- c. -A temporary change to a procedure must be reviewed by whom (include any qualifications required of the individual (s))? (1.0)
- b. With the exception of the above review, what other condition must the temporary change meet? (0.5)
QUESTION 8.08 (1.00)
Why is the Turbine Generator overspeed protection system addressed in River Bend Technical Specifications when the Turbine Generator is not safety-related? (1.0)
I i
i l
l (***** END OF CATEGORY 08 *****)
( (************* END OF EXAMINATION ***************)
EQUATICri 5HEET f a ma y= s/t Cycle efficiency = (Net work out)/(Energy in) 2 w = og s = V,t + 1/2 at E
E=K -
-At KE = 1/2 av a=(Vf - V,)/t A = AN A=Aeg PE = ogn Vf = V, + at w = e/t 1 = an2/t1/2 = 0.693/t1/2 y = , 3p
,n 2 tiff = UtuM]
A" 4 [(t1/2) * (*b)3 aE = 931 am m = V,yAo ,
-h Q = aCpat 6 = UAa T I = I ,e-"* -
Pwr = Wfah I = I ,10-* E TVL = 1.3/u 5
P = P 10 "III) HYL = -0.693/u
- p = p et /T O
l SUR = 26.06/T SCR = S/(1 - K,ff)
CR, = S/(1 - K,ff,)
SUR = 26s/s* + (s - e)T CR j (1 - K,ffj) = G2 II ~ Ieff2)
T = (t*/s) + ((s - sy To) M = 1/(1 - K,ff) = CRj /G ,
T = s/(o - 8) M = (1 - K ,ff,)/(1 - K,ffj)
T = (s - o)/(Io) SDM = ( -K,ff)/K,ff a = (K,ff-1)/K,ff = AK,ff/K eff s' = 10 seconds I = 0.1 seconds" o = ((t*/(T Kgf9))+(Egff/(I + 5I)) i Idlj=Id Id 2 ,z gd 2 P = (seV)/(3 x 1010) jj 22 2 i I = eN R/hr = (0.5 CE)/d (meters)
R/hr = 6 CE/d2 (f,,g) ,
Water Parameters Miscellaneous Conversions I gal. = 8.345 lem. 1 curie = 3.7 x 1010dps Iga;.=3.78 liters 1 kg = 2.21 lbm 1 ft* = 7.48 gal. I hp = 2.54 x 103 Stu/hr Density = 62.4 lbm/ft3 1 aw = 3.41 x 10 Stu/hr Density = 19m/cm3 lin = 2.54 cm Heat of vaporization = 970 Stu/lbm *F = 9/5'C + 32 Heat of fusion = 144 Bru/lbm 'C = 5/9 (*F-32) l Atm = 14.7 psi = 29.9 in. Hg. 1 8TU = 778 ft-lbf I ft. H O 2
= 0.4335 lbf/in.
'Sz__IBEQBI_QE_UVGLEaB_EQWEB_EL6NI_QEEBoI1QUt_ELVIQHt_6NQ PAGE 15 IEEBdQQ1Ned1GH ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 5.01 (2.00)
- c. CRITICAL POWER increases (0.25). A greater enthalpy rise is required to bring the coolant to saturation conditions resulting in less steam generation for a given power level (0.75) or similar.
- b. CRITICAL POWER increases (0.25). Same reason as above (0.75).
REFERENCE LOTM-TH-3.7-0, Transition Boiling, pg 3 ANSWER 5.02 (2.00)
- a. Fission products: Tritium, I, Kr, Xe, Cs, Ba (0.1 each)
Corrosion products: Fe, Ni, Co, Mn (0.1 each)
- b. No (0.25). The activity is due to tramp uranium f is s io n (0.75).
REFERENCE Mitigating Core Damage, pg 5-8 Chart of the Nuclides ANSWER 5.03 (1.00)
- 1. a
- 2. e
- 3. d
- 4. b
- 5. c (0.25 for each manipulation necessary to put the items in the correct order)
REFERENCE LOTM-TH-3.2-0, Modes of Heat Transfer LOTM-TH-3.5-0, Boiling Heat Transfer
St__IBEQBl_QE_UUGLE8B_EQWEB_EL6NI_QEE86I196t_EL91 dst _eUR PAGE 16 IBEBdQQ1Ned191 ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 5.04 (1.00) b (1.0)
REFERENCE LOTM-TH-2.5-0, let Law of Thermodynamics LOTM-TH-2.13-0, Reactor Heat Balance ANSWER 5.05 (2.50)
- c. Four fuel bundles and a control rod (0.5)
- b. Fully withdrawn (0.5)
- c. All control rod movements are associated with low reactivity cells (1.0).
- d. True (0.5)
REFERENCE LOTM-2-1, Nuclear Fuel, pg 20-23 ANSWER 5.06 (3.00)
Void fraction decreases (0.5). The void fraction is slightly smaller bocause the negative effects of doppler are compensated for by the void fraction decrease (0.5).
Doppler reactivity increases (0.5). As power increases, the fuel temperature increases. This temperature increases cause doppler to cdd more negative reactivity (0.5).
Enthalpy increases (0.5). The increased reactor power allows more extraction steam to be drawn off thus increaseing inlet enthalpy (0.5).
REFERENCE LOTM-TH-4.18-1, Void Coefficient LOTM-TH-4.19-1, Doppler Effect LOTM-14b-0, Feedwater Heaters and Heater Drains
Ez__IBEQBl_QE_NUCLEeB_EQWEB_ELeNI_9EEBoIIQNt_ELUIDSt_eND PAGE 17 IBEBdQD1Ned1GS 1
ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 5.07 (1.50)
Withdrawal of a center control rod at 90% density has greater worth.(0.5)
The control rod worth is proportional to the (local neutron flux / the core cverage neutron flux) squared.(0.25)
With 90% rod density the core average neutron flux is very small. j Withdrawing a central control rod, increases the local flux in the area of withdrawn rod substantially. Because the rod causes the value of the torm (local neutron flux / core average neutron flux) squared to be large its worth for this condition is quite high. Higher than withdrawing the 4 rod at 40% rod density, when core average flux will be higher. (0.75) l REFERENCE LOTM-TH-4.20-1, Control Rod Worth, pg 4 l
l l
l l
l l
l l
l
L __IBEQBl_QE NUQLE88_EQWEB_EL8HI_QEEB8110Ni_ELu1Rh 8NQ PAGE 18 IBEBdQQ1Ned191 ANSWERS -- RIVER' BEND'1 -86/11/04-GRAVES, D. w ANSWER 5.08 (2.00) 6
't'
'Orawing a straight _line between the last two *'s predicis'34-35 control' ., /
rods must be withdrawn.- ' >##
(0.1 for each point plotted, 1.3 for line and predictio,n) 5 10 15 20 25 30 35 40 45 50 55 :
1.0*----'l----1----I----1----I----1----I----I----I----I----I----1.O qp 0.9- -0.9 , 7 0.8- -* -0.8 .
^
.0.7- -0.7 1/M 0.6- -
s -0.6 -i 0.5- .,
-0.5
- s -
0 .' 4 - -0.4 0.3- * -0.3 0.2- * -0.2 0.1- * -0.1 O.01----l----l----l----l----l----l----l----l----l----l----l----0.0' .
0 5 10 15 20 25 30 35 40 .45 50 55' '
REFERENCE LOTM-TH-4.13.1, Suberitical Multiplication and Count Rate Comparison ym ANSWE'R 5.09 ( .00)
NOT APPLICABLE
m ,
gs nht_-)C6HI_11SIEdS_DESIGNt_QQUIBQLt_6NQ_INSIBudENI6IIQN PAGE 19
^
ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
~
ANSWER 6.01 (2.00)
Flow control valve position-does not change (0.5) because total system flow has Sot changed (0.5). Drive water differential pressure will increase (D.5) because flow previously bypassing tSt pressure control
.vs1ve cis now passing through it (0.5). Also accept that the FCV opens slightly to maintain flow due to slightly increased backpressure.
REFERENCE LOTM-4-1, CRDH. System, pg 16-17
- h. ,
'4
ANSWER 6.02 .(2.00)
- c. gSafety (accept ALL if noted that only 1 logic circuit'is in test)
(0.285 each, 2.0 total) r REFERENCE LOTM-7-1, Main Steam, pg 5-7, 16 ANSWER 6.03 (1.507 A d ~~ W h e n t h e turbine trips, pressure in the turbine decreases to condenser r' . pressure (0.5). The condensed extraction steam flashes to steam (0.5),
ond backflows up the extraction steam line through the turbine (0.5), which
$' can cause it to overspeed.
REFERENCE
..- LOTM-14b-1, Feedwater Heaters, Vents, and Drains, pg 17
'D ANSWER 6.04 (2.50) c.- RHR HX Bypass valve (0.5) Other valves are acceptable answers. See the valve logic tables in the stated reference.
- b. - amber light above the valve switch (0.5)
- annunciator (RHR Inj ection Valve in Man Override) (0.5)
- c. - LOCA signal clear (0.5)
- Initiation Reset pushbutton(s) depressed (0.5) l l
l i
. . - - - - - - - . . . - . ~ . -- ,
61__EL8HI_111IEd1_DE11GN _QQNIBQLt_8NQ_IN11BMMENI611QN t PAGE 20
-ANSWERS -- RIVER' BEND 1- -86/11/04-GRAVES, D.
REFERENCE LOTM-20-0,.RHR System, pg 9
' ANSWER 6.05 (2.00)
- e. -3 (1.0)
- b. 1 (0.5)
- c. initiation signal must be present (0.5).
REFERENCE LOTM-21-0, RCIC. System, pg 7, 13, 37, 38 ANSWER 6.06 (3.00) a.' LPSP:- Rod withdrawal is limited to four notches (0.5) before having to deselect and reselect a rod.
HPSP: Rod withdrawal is limited to two notches (0.5) before having to deselect and'reselect a rod.
- b. LPSP:' 2'7. 5% power or.157.6 psig first stage turbine pressure (0.5)
HPSP: 62.5% power or 410.7 psig first' stage turbine pressure (0.5)
Accept +'or - 1% power and + or - 5 psig pressure
,- c. Reactor power would be higher than that indicated by turbine first stage-pressure. Rod movement could.be performed that could conceivably exceed a -fuel safety limit (1.0).
REFERENCE
-LOTM-IC-7-0, RC & IS, pg 31, 47 River Bend Technical Specific 1ti: ns Bases, 3.1.4.1 l \
l ANSWER 6.07 (2.00)
- a. I (0.5) b .- By taking the Remote Shutdown Transfer Switches to the EMERGENCY position (0.5).
- c. Annunciator (s) in the control room (0.5)
. Loss of position indication for that valve in the control room (0.5).
L l --
l r
L b) >
)J
Ei__EL8HI_111IEd1_QE11QNi_QQNIBQLt_680_INSIBudENIallQN PAGE 21.
, ANSWERS'-- RIVER-BEND l' -86/11/04-GRAVES, D.
REFERENCE LOTM-IC-16-0, Remote Shutdown, pg 7 ,
1..
4' 1
4 f
f
- > 4 e
I i
E
\
i l
l l
i.
l l
Zi__EBQGEQUBE1_=_NQBd8Lt_8BNQBU6Lt_EMEBEENGX_6ND PAGE 22 88DIQLQQ1G8L_CQUIBQL ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
ANSWER 7.01 (2.50)
-The cooldown rate limit is < or = 100 deg F (0.25)'in any one hour poriod.(0.25). The limit was exceeded (0.5) for three one hour poriods: 0845 - 0945 (105 deg F), 0900 - 1000 (105 deg F), and 1015 - 1115 (105 deg F). (3. periods of time at 0.5 each)
' REFERENCE
.' River Bend Technical Specification 3.4.6.1 ANSWER 7.02 (2.50)
- a. Turbine BPV are only rated at 10%. total steam flow. Will also accept to prevent a high pressure scram. (0.5) b '. High airborne activity in the turbine building can result (0.5) if vacuum is broken within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from operation at full power.
c.- There is no automatic pressure control (0.5).
- d. To prevent warping the shaft (0.5),
- e. Prevent flashing the reference legs (0.5) on the RPV instruments.
REFERENCE River Bend GOP 0002, Power Decrease / Plant Shutdown, Rev 3, pg 8-10 l ANSWER 7.03 -(1.50) i When instrument air header pressure has decreased to 65 psig (0.5) or it
, becomes apparent that instrument air will not be restored (0.5), scram the l reactor (0.5).
r REFERENCE River Bend AOP - 0008, Loss of Instrument Air, Rev 2, pg 3 ANSWER 7.04 (1.00)
- The main turbine is NOT on the line (0.5).
- Turbine BPV are NOT being used to control reactor pressure (0.5).
Zz__EBQQEQUBE1_:_NQBd8Lt_eBNQBdekt_EdEBGENQ1_6NQ PAGE 23 86DIQLQQIQaL_GQNIBQL-
' ANSWERS -- RIVER' BEND -86/11/04-GRAVES, D.
REFERENCE River Bend AOP 0009, Loss of Normal Service Water, Rev 2, pg 4 ANSWER 7.05 (2.50) s.. 1. Arm and depress all four manual scram buttons.
- 2. Mode Switch to SHUTDOWN.
- 3. Verify control rods fully inserted.
- 5. Close the MSIV's (5.at 0.4 each)
- b. Division I Remote Shutdown Panel'(0.5)
REFERENCE A0P-0031, Shutdown from Outside the Control Room, Rev 3, pg 6 ANSWER 7.06 (3.00)
- e. none
- b. -EOP-0001, RPV Control E0P-0002, Primary Containment Control c.. E0P-0002, Primary Containment Control
- o. none (6 at 0.5 each)
REFERENCE.
U . River Bend E0P's 0001, Rev 6; 0002, Rev 5; 0005, Rev 3 l
ANSWER 7.07 (2.00)
- a. 1000 MREM - 850 MREM = 150 MREM i
150-MREM / (200 MREM /hr) = 0.75 hr or 45 minutes
- (0.5 pt for the limit, 0.5 for the stay time)
- b. Assistant Plant Manager recommendation (0.5)
Supervisor-Radiological Programs approval (0.5)
REFERENCE GET II-1, pg 4 i
w.
.Al__8DMIN11188IIVE_P89QEQMBESt_QQUQlI1QNit_8NQ_ lid 116IIQNS .PAGE 24 ANSWERS ---- RIVER BEND L 1 -86/ll/04-GRAVES,-th ANSWER 8'.01 '(2.00)
-o. no
- b. yes
- c. 'yes
- d. no (4 at 0.5 each)
REFERENCE River. Bend Technical Specifications, Definition Section I
. ANSWER 8.02 (2.50) 2
.o. True
- b. False
- c. ~True
- d. False.
e .- True
-(5 at 0.5 each)-
REFERENCE River Bend Technical Specifications, Section 6.2.2 ANSWER 8.03 (1.00)
- Louisiana Nuclear Energy Division Louisiana Of fice of Emergency Preparedness East Feliciana Parish West Feliciana Parish
- Pointe Coupee Parish
- East Baton Rouge Parish West. Baton Rouge Parish
- Mississippi Highway Safety Patrol (5 required at 0.2 each)
REFERENCE EIP-2-005, General Emergency, Rev 3, pg 5 EIP-2-006, Notifications, Rev 6, pg 3
at__6QdlUISIB611ME_EBQGEQMBElt_GQUQlllQUlt_6UQ_LldlI611QU$ PAGE 25 ANSWERS -- RIVER BEND 1 -86/ll/04-GRAVES, D.
ANSWER 8.04 (2.50)
- c. Offsite dose calculations shall be performed anytime a Site Area (0.25) or General Emergency (0.25) is declared due to a radiological accident (0.25), or any other time the Emergency Director or Recovery Manager deems necessary (0.25).
- b. Shift Foreman (0.5)
- c. 1. Initial assessment should be completed within approximately 15 minutes of the declaration (0.5)
- 2. Follow-up calculations should be performed approximately every 30 minutes following initial projections (0.25),
or sooner if release rates change significantly (0.25).
REFERENCE River Bend EIP-2-025, Offsite Dose Calculations - Computer Method, Rev 2, pg 2 ANSWER 8.05 (2.50)
- c. - Place the system / component in a stable and safe condition (0.5)
- Inform the responsible supervisor (0.5)
- b. - Determining the method by which the activity can be performed using the procedure as written (0.25)
- Submitting a procedure change (0.25)
- c. No (0.25). The independent verification will be performed separately without visual or audible contact with the first performer (0.75).
REFERENCE River Bend ADM 0022, Conduct of Operations, Rev 7, pg 24, 26 ANSWER 8.06 (2.00)
- a. There are interpretation letters included i- front of the Technical Specifications (1.0).
- b. When it is required to change modes to meet an action statement, or TS 3.04 is specified as N/A (1.0).
at__aQdIN11186IIVE_EBQQEQUBEst_QQUQ1Il0 Nit _aND_LidIIaI1983 'PAGE 26 ANSWERS -- RIVER BEND 1 -86/11/04-GRAVES, D.
REFERENCE River Bend Technical Specifications ANSWER 8.07 (1.50)
- c. Two members of plant management staff (0.5), one of which holds a Senior Operator license (0.5). Also accept section supervisor, discipline supervisor or foreman AND C0F or SS~as the two members of plant management staff, if candidates' answers are that specific.
- b. The intent of the procedure is not changed OR to continue work already in progress (either one at 0.5).
REFERENCE River Bend Technical Specifications, Section 6.8.3 ANSWER 8.08 (1.00)
Foilure to trip on overspeed could result'in the turbine becoming a missile hczard (which could impact safety related equipment, systems or barriers).
REFERENCE RBS Technical Specification 3.3.8
REQUALIFICATION PROGRAM EVALUATION REPORT Facility: RTVFR RFNn STATTnN Examiner: n. N_ GRAVFS Date(s) of Evaluation: NnVFPRFR 4 6. 19R6 Areas Evaluated: yy Written Oral yy Simulator Examination Results:
R0 SRO Total Evaluation Pass / Fail Pass / Fail Pass / Fail (S, M or U)
Written Examination 2/1 6/0 8/1 s
. Operating Examination i
Oral Simulator 1/n s/n 9/n s
- Evaluation of facility written examination grading s*
Overall Program Evaluation Satisfactory _1g_ Marginal Unsatisfactory (List major defi-ciency areas with brief descriptive j comments) 1 i
- The only facility examination grading reviewed was the facility retake examination administered to the one RO failing the initial requalification written examination.
Submitted: Forwarded: Approved:
U 7$
Examiner f 8Branch a bslA Chief mb\
.ectionChief[