ML20214J580
| ML20214J580 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 08/05/1986 |
| From: | Cooley R, Mccrory S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20214J575 | List: |
| References | |
| 50-368-OL-86-03, 50-368-OL-86-3, NUDOCS 8608150084 | |
| Download: ML20214J580 (48) | |
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- 5. p ARKANSAS NUCLEAR ONE, UNIT 2 EXAMINATION REPORT 50-368/0L-86-03 Docket No: 50-368 Lice.nse No.: NPF-6 Licensee: Arkanasa Power & Light Company P.O. Box 551 Little Rock, AR 72203 Examinations administered at Arkansas Nuclear One Unit 2 Chief Examiner: . [/ /
Date/ N 5.'t. 04cCrory, Lead Examiner ~ Approved by: Da'te
/b R.A. Cooley,SecponChief Summary Examinations conducted on July 15, 1986.
Written and operating examinations were administered to seven (7) Senior Reactor Operator Upgrade candidates. All candidates passed these examinations. h 8608150084 86dG08' > PDR ADOCK 05000368 V PDR ___)
2 Report Details I. Examination Results SR0 candidates R0 Candidates Total Pass- Fail % Total Pass Fail % 7 7 0 100 0 II. Examiners S. L. McCrory, Chief Examiner D. Graves i III. Examination Report This Examination Report is composed of the sections listed below. Examination Review Comment Resolution Exit Meeting Minutes 1 Generic Comments ANO-2 Examination Key (SR0 Questions and Answers) l l Performance results for individual examinees are not included in this l report because examination reports are placed in the NRC Public Document ! Room as a matter of course. Individual results may be retained in the NRC l Region IV office during the period that the facility is evaluated as L unsatisfactory but are not subject to public disclosure. A. Examination Review Comment Resolution In general, editorial comments or changes made during the ! . examination, the examination review, or subsequent grading reviews are not addressed by this resolution section. This section reflects resolution of substantive comments made.during the examination l ! review. The modifications discussed below are included in the master l examination key which is provided elsewhere in this report as are all other changes mentioned above but not discussed herein. Attachment 1 l
~
is th'e_ facility comments on the examination. Unless otherwise , indicated in this section, the facility comments were incorporated ( into the answer key. r .
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3 ! COMENTS
- 1. 5.02 See attachment.1 for comment.
- Resp. In general, expanded ranges for acceptable answers were used. Since the problern stated core age as MOL, no .
credit was given for using BOL or E0L curves if MOL. data was available.
- 2. 5.04 Should accept 185 + 10.
Resp. Reject. Since the-data provided plots as an ideal straight line, it is not necessary to " fair in" a curve to fit the data points.
- 3. 5.09' If " normal" operating ' temperature is 545 to 580 deg.F. t "C" is correct. If " normal" operating temperature is- 80 to 580 deg.F, "B" is correct.
Resp. The question was graded.using 545 to 580 deg.F as normal operating temperature. B. Exit Meeting Summary At the conclusion of the examination period, the examiners met with the members of the facility staff to discuss the results of the examinations. The following personnel were present for the exit meeting: NRC UTILITY E McCrory J. Vandergrift D. Graves W. Perks C. Harbuck S. Quennoz E.-Ewing P. Crossland C. Anderson NRC informed the facility staff that no generic weaknesses were - identified during the course of these examinations. It was noted, however, that some candidates indicated to the examiners during the course of the operating examination that the facility walk-through examinations did not focus or. the Emergency Plan to the extent that the NRC examination did. Additionally, several of the candidates expressed to the examiners that they were not clear on what to use as a basis for determining offsite exposure for making evacuation recomendations to local authorities. I
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7 4 C. General Comments NRC was pleased with candidrte performance on both the written and operating examinations. The Unit 2 training staff has done well in increasing the performance excellence of the Unit 2 operators and senior. operators. D. ANO-2 Examination Key Date Administered: 7/15/86 Type of, Examination: Senior Reactor Operator Upgrade Comments: During the grading of the examination, some questions were found to have ambiguous point values assigned. Where this occurred, the
+
point value appearing on the question number line was used. 3
g .. U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _6BK6N161_NQQLE68_QNE_2__ REACTOR TYPE: _EWB-QE__________________ DATE ADMINISTERED:_ghlDZZ1E________________ EXAMINER: _dQQBQB1t_St_____________ APPLICANT: _________________________ INSIBUQIl0NS_IQ_8EELIQ8 nil Uno separate paper for 'the answers. Write answers on one side only. Stcple question sheet on top of the answer sheets. Points for each qu0stion are indicated in parentheses after the question. The passing grcde 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 tho examination starts.
% OF CATEGORY % OF APPLICANT'S CATEGORY
__Y8LUE_ _IQI8L ___SQQBE___ _V8LUE__ ______________Q6IEQQBY_____________ _25tDQ__ _25121 ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS _22tQQ__ _221B2 ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION _22tEQ-_ _24125 ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL _25tDQ__ _25t21 ___________ ________ S. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS _2htEQ__ 1QQtDQ ___________ ________ TOTALS FINAL GRADE _________________% All work done on this examination is my own. I have neither givon nor received aid. PPL5C UiI5~55GU TUR5~~~~~~~~~~~~~~
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 appearance or possibility of cheating. Use black ink or dark pencil only to facilitate legible reproductions. 3. I 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).
t
- 6. Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category " as appropriate, start each category on a new page, write on only olie side of the L 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 three lines between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
L 12. Use abbreviations only if they are commonly used in facility literature. E- 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 calcualtions, 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 examiner 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.
- 18. When you complete your examination, you shall:
- a. Assemble your examination as follows:
(1) Exam questions on top. (2) Exam aids - figures, tables, etc. (3) Answer pages including figures which are a 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.
t
- d. Leave the examination area, as defined by the examiner. If after leaving, L-you are found in this area while the examination is still in progress, your E
license may be denied or revoked.
- e. Do not dicuss the examination with other licensee staff personnel until the formal examination review is complete.
51__IBEQBl_QE_NuGLE88_EQWEB_EL8NI_QEEB6110Nt_ELu1QSt_6NQ PAGE 2
-IHEBdQQ1NedIGS QUESTION 5.01 (2.00) -The relative worth of a CEA is dependent on the neutron flux reaching it.
LIST and EXPLAIN four (4) factors or conditions which affect the flux roaching a particular CEA. (2.0) QUESTION 5.02 (3.00) The reactor is at 2250 psig, 545 deg.F (Mode 3), and Keff = 0.997 with rods et 10" on group 6. Core depletion is 255 EFPD and boron concentration is 790 ppm. Ignore xenon and samarium concentrations unless otherwise stated. Graphs have been provided to assist with calculations but it may not be n0cessary to use each graph. (3.0)
- c. How much rod motion will be required to bring the reactor j ust critical? Show all calculations. (1.0)
- b. How much additional rod motion will be required to establish a 1 DPM startup rate after the reactor is made critical? Show all calculations. (1.0)
- c. If the 1 DPM startup rate is allowed to continue until power enters the power range, what will be the final values of RCS temperature and pressure and reactor power? Assume no operator action with rod control in manual and all other systems in normal startup condition.
(1.0) l QUESTION 5.03 (1.50) l l An RCS heatup with an MTC of +.3 dK/K/deg. F and an RCS cooldown with an l MTC of .3 dK/K/deg. F will each produce a reactor power increase. If the l tcmperature change is the same in both cases, is the change in reactor l pcwer the same? Why or why not? (1.5) l l I (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
Sz__IHEQBl_QE_NUGLE88_EQWEB_EL8NI_QEEB8IIQNz_ELu1D$t_88Q PAGE 3 IBEBdQQ1NedICS
. QUESTION 5.04 (1.50)
- c. With the data provided below, construct a "1/M" plot. Use the blank graph sheet attached. (1.0)
Neutron count rate (n/sec) fuel assemblies loaded 100 0 105 10 116 27 127 41 217 100 263 115 500 148
- b. How many fuel assemblies are needed to make the reactor critical? (0.5)
QUESTION 5.05 (1.00) Consider the operating characteristics of a centrifugal pump and explain ehy the HPSI pumps were arranged in parallel rather than in series. (1.0) QUESTION 5.06 (2.50)
- e. Where in the reactor core is nucleate boiling most likely to begin?
Why does it start there rather than elsewhere? (1.5)
- b. What is the impact of going to nucleate boiling from the non boiling condition? (1.0) i I
QUESTION 5.07 (2.00) 00 scribe HOW and WHY increasing condensate depression in the condenser will offect the following: (2.0)
- o. REQUIRED net positive suction head of the condensate booster pump.
- b. AVAILABLE net positive suction head of the condensate booster pump.
- c. Overall plant efficiency.
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
5t__IBEQBl_QE_NUQLE88_EQWEB_EL8NI_QEEB8Il0Nt_ELMIQat_6NQ PAGE 4 IBEBdQQ1N8MIQS QUESTION 5.08 -(3.00)
- o. What TWO parameters can the operator control to prevent pressurized thermal shock? (1.0)
- b. What is the sequence of generic events that could lead to pressurized thermal shock conditions? (1.0)
- c. Why does the concern about brittle fracture of the reactor pressure vessel increase as the ANO unit 2 plant ages? Include in your answer the specific material property the.t is affected. (1.0) l l
l I (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
5t__IBEQBl_QE_NUQLE88_EQWEB_EL8NI_QEEB8IIQNt_ELu1Dit_6HD PAGE 5 IBEBdQQ1NedIQ1 QUESTION 5.09 (1.50) Choose the' correct response to each of the following.
- 1. Moderator temperature coefficient becomes more negative from BOL to EOL primarily because of: (0.5)
- a. The larger change in resonant escape probability per degree change in moderator temperature.
- b. The larger change in core leakage per degree change in moderator temperature.
- c. More thermal neutrons are available for absorption in the moderator.
- d. The smaller change in thermal utilization factor per degree change in moderator temperature.
- 2. Doppler coefficient (pcm/ degree F fuel) becomes more negative from BOL to EOL because: (0.5)
- a. An increase in effective fuel temperature.
- b. Clad creep and fuel pellet swell.
- c. The production of plutonium 240.
- d. The overlapping of resonant peaks.
- 3. Which SINGLE condition below results in the GREATEST rod worth within the normal operating range of temperature and boron? (0.5)
- a. Higher boron concentrations
- b. Higher moderator temperatures
- c. Lower boron concentrations
- d. Lower moderator temperatures
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
5t__IHEQBl_QE_NUGLEAB_EQWEB_EL8NI_QEEB6IIQNt_ELVIQ1t_8NQ PAGE 6 IUEBdQQ1N8dICS QUESTION 5.10 (2.00)
- o. If a trip from 100% power occurs with xenon at equilibrium, what is the approximate time interval after trip that xenon will again be at the 100% equilibrium value? (0.5)
- b. How would this approximate time interval compare if the trip occurred from 50% equilibrium conditions? (0.5)
- c. When a reactor is returned to_100% power from peak xenon conditions, why does xenon reactivity undershoot the 100% equilibrium value?(1.0)
QUESTION 5.11 (2.00) For each of the following conditions (a-d), which of the two choices results in the greater INDIVIDUAL (differential or integral as indicated) red worth? Rod Worth Condition Choice 1 Choice 2
- e. Integral Tavg 150-F 500-F (0.5)
- b. Integral Core life BOL EOL. (0.5)
- c. Differential Rod position 40% inserted 70% inserted (0.5)
- d. Differential Rod in Group 6 an inserted the rod (0.5) which is next cod withdrawn to a module with QUESTION 5.12 (1.00)
Explain HOW and WHY ASI is expected to change as power is increased during a normal power increase 20% to 70%. Assume ARO and power increase by boron dilution. (1.0) (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
;51__IBEQBY_QE_NUQLE8B_EQWEB_EL8NI_QEEB6110Ni_ELU1011_6NQ PAGE 7 IBEBdQQ1N8M101 QUESTION 5.13 (2.00)
TRUE or FALSE? No explanation required. (2.0)
- c. The operator can increase the heat removal rate from the RCS by reducing steam pressure.
- b. A LOCA with NO RCPs running can result in more inventory loss than a LOCA WITH RCPs running.
- c. A total and prolonged loss of feedwater can lead to a loss of RCS liquid inventory.
- d. The primary concern when fuel clad temperature reaches 1400 deg. F is f
the production of hydrogen. (***** END OF CATEGORY 05 *****)
6t__ELANI_SYSIEMS_DESIGNt_QQNIBQLt_aNQ_INSIBudENIaIIQN PAGE 8 QUESTION 6.01 (2.00) Match the voltages with the appropriate letter designation. (2.0)
- c. 6900 VAC 1. A
- b. 4160 VAC 2. AB
- c. 480 VAC- 3. B
- d. 120 VAC 4. D
- o. 120 INST AC 5. EH
- f. 125 VDC 6. H
- g. 120 VAC 7. LA RPS/ESF 8. RA
- h. 125 VDC 9. RS RPS/ESF 10. SA
- 11. Y QUESTION 6.02 (1.50)
- o. List six (6) individual instrumentation indications available to the
, operator for monitoring the condition of the RCP seals. (Excluding alarms) (0.6)
- b. What indications would you have if a coolant pump lower seal failed and the other seal stages remained intact? (0.9)
QUESTION 6.03 (2.00) l What are two (2) undesirable reactions to simultaneous SIAS and RAS cctuation? Explain WHY the reactions are undesirable. (2.0) QUESTION 6.04 (2.50) l ( o. Name three (3) ways of tripping a 4160 V or 6900 V breaker using DC control power. (1.0)
- b. If a 4160 or 6900 V breaker is shut and DC control power is lost, how will the breaker respond? (0.5)
- c. If a 4160 or 6900 V breaker is open when DC control power is lost, what condition (of the breaker) must be satisfied to shut the braaker? (1.0) l l
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
Ez__EL8NI_111IEd1_QE11GNt_QQNIBQLt_8ND_INSIBudENI8IlQN PAGE 9 QUESTION 6.05 (1.50) Give the normal position for each of the following valves when operating in mode 1. (1.5)
- o. RWT outlet
- b. Spray header isolation
- c. NaOH tank outlet
- d. Containment sump isolation inside containment
- o. Minimum flow isolation QUESTION 6.06 (2.00)
During normal power operation, the two temperature sensing elements on the outlet of the letdown heat exchanger fail high. What control functions-occur? (2.0) QUESTION 6.07 ( .00)
*** EMPTY QUESTION ***
QUESTION 6.08 (3.00) The reactor is operating at 100% power with all systems in normal. For the following failures explain the plant response AND indicate what reactor protective signal will cause the roactor to trip. Assume N0 operator action and consider each foilure independently.
- c. Controlling Pressurizer level channel fails high. (1.5)
- b. Controlling Steam Generator level channel (LT-1111) fails low. (1.5)
QUESTION 6.09 (1.50) During a plant startup, while bypassing steam to the condenser, a slow, oteady increase in condenser pressure occurs. Describe how this pressure increase affects the signals going to the bypass and dump valves from the SDBCS as pressure slowly goes to atmospheric. (1.5) (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) f
ht__EleNI_1YSIEMS_DESIGNt_QQNIBQLt_6NQ_INSIBudENIeIIQN- PAGE 10 QUESTION 6.10 (1.50) What effect does a CIAS have on the Penetration Room Ventilation System? (1.5) QUESTION 6.11 (2.00) With the plant at 15% power during a plant startup, the operator is maintaining steam generator levels with all FWCS Hand / Auto stations in
"=anual".
- c. What will the response of the FWCS be if a reactor trip occurs? (1.0) -
- b. What two problems could this create if there were no operator action to correct? (1.0)
QUESTION 6.12 (1.50) Explain how you can determine from instruments available in the coritrol room whether high activity in the RCS is due to a crud burst or from failed fuel. (1.5) QUESTION 6.13 (2.00) The excore nuclear instrumentation inputs a " raw" signal to the CPCs. i Briefly describe two of the corrections to this " raw" signal that the CPCs porform to obtain " calibrated neutron power". (2.0) 4 .I (***** END OF CATEGORY 06 *****)
Zi__EBQQEQUBE1_=_NQBd8Lt_8BNQBd8Lt_EMEBGENQ1_8NQ PAGE 11 88DIQLQQ1Q8L_QQNIBQL QUESTION 7.01 ( .00)
*** EMPTY QUESTION ***
QUESTION 7.02 (2.00)
- c. Why are we concerned about " Quality Factors" with regard to radiation? (1.0)
- b. What are the " Quality Facters" for the various kinds of radiation that may be present at AN027 (1.0)
QUESTION 7.03 (2.00) What is the MAXIMUM dose you could expect to receive in each of the following radiologically controlled areas in one hour? (2.0)
- c. Radiation area posted inside controlled access.
- b. High. radiation area which is not locked.
l QUESTION 7.04 (1.50) ! List three of the symptoms for " degraded power" in the Reactor Trip L Rocovery Tab of E0P 2202.01 (1.5) l l t j QUESTION 7.05 (2.00) What two conditions during a SIAS require securing RCPs? (2.0) l QUESTION 7.06 (1.50) One of the immediate actions for reactor trip is to verify that pressurizer lovel is >29% and (.=82%. What are the bases for these limits? (1.5) l l l (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
e Zi__EBQGEQUBER_=_NQBd6Lt_8BNQBd6Lt_EMEBGENQ1_8NQ PAGE 12 88DIQLQQ108L_GQNIBQL QUESTION 7.07 (1.50) , List two separate conditions for entry into inadequate core cooling recover ' cction. (1.5) QUESTION- 7.08 (2.00) What are two criteria for securing RCPs after MSIS occurs due to a steam line break inside containment? (2.0) L QUESTION 7.09 (3.00) The immediate actions of E0P 2202.01 list seven safety functions which rcquire verification of various parameters and automatic actions. For each of.the safety functions given below, list the verifications to be performed. The number in parenthesis after the safety function indicates the number of verifications required for that safety function. (3.0)
- c. RCS INVENTORY CONTROL C5)
- b. RCS HEAT REMOVAL (3)
- c. CORE HEAT REMOVAL (2)
- d. CONTAINMENT INTEGRITY (2) l QUESTION 7.10 (2.00) l If any EDG auto sturts but does not energize its respective bus during a blackout, why is it important to attempt to shut the EDG output breaker as quickly as possible? (2.0) l QUESTION 7.11 (1.50)
Why should outward rod motion be in small frequent steps rather large steps et one time when above 50% reactor power? (1.5) l (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
.Zz__EBQQEQUBER_ _NQBd8Lt_8BNQBd6Lt_EdEBQENQ1_8NQ PAGE 13 88Q10LQQIQ8L_QQNIBQL QUESTION 7;12 (1.00)
How shoul'di NORMAL power decreases be initiated from steady state high power icve ls ? ' # (1.0) QUESTION 7.13 (2.00) The plant shutdown and cooldown procedure, OP 2102.10, lists alternative nothods for RCS cooldown and depressurization if normal processes are not cvailable to get to the SDC mode. What is to be employed as the LAST RESORT for both cooldown and depressurization of the RCS? (2.0) QUESTION 7.14 (1.50) A precaution in the CVCS operating procedure states' " Placing a freshly charged ion exchanger-in service may alter RCS boron concentration." HOW cnd WHY vill RCS boron concentration be affected? (1.5) i l l I r i l l l l (***** END OF CATEGORY 07 *****) l l
Az__6DdINISIB8IIVE_EBQQEQUBgst_QQNDIl10NSt_8ND_LIMII611081 PAGE 14 QUESTION 8.01 (1.00) When the shutdown margin is determined to be less than the Limiting Condition for Operation for the operational mode the plant is in, what is the operator Tech Spec action that must be taken? (1.0) QUESTION 8.02 (1.50) For operations with Tave => 300 deg F. Technical Specifications require 2 independent ECCS subsystems operable. What constitutes an operable ECCS subsystem? (1.5) QUESTION 8.03 (2.50)
- a. Why is a leak rate done for the RCS? (0.5)
- b. What are the maximum allowable leak rates? (1.5)
- c. How often is a leak rate required in mode l? (0.5)
QUESTION 8.04 (2.50) Give the emergency classification for the following Emergency Action levels (EALs). (2.5)
- a. Evacuation of the control room required, control of systems can be established.
- b. Projected summed releases from either unit exceed 1 MPC for one hour at the site boundary assuming annual average X/Q.
- c. Instrumentation indicates the RCS is in a saturated condition.
- d. Measured dose rates at the site boundary is 300 MR/HR whole body.
- e. Ongoing security threat within plant buildings, but not within the control room or vital areas.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
At__8051N1118811ME_PBQQEQMBEft_QQNQlllQN$t_6NQ_LidlI6IlQN1 PAGE 15 QUESTION 8.05 (1.50) The minimum temperature for criticality limitation.(525 deg. F) is required to ensure four criteria are met. List three of the criteria. (1.5) QUESTION 8.06 (1.00) What is the bases for requiring that the reactor be subcritical for cpecified time periods before moving spent fuel from the core to the spent fuel pool? (1.0) QUESTION 8.07 (1.00) During refueling operations the water height above the core must be 23 feet. This is based on ___ . (Chose one answer.) (1.0)
- a. maintaining adequate heat sink
- b. absorbing iodine gas activity
- c. providing adequate shielding
- d. limiting visual distortions QUESTION 8.08 (1.00)
Technical Specifications require that fire hose stations be OPERABLE at i specified locations. What must be done if one of these stations is l INOPERABLE? (1.0) l l QUESTION 0.09 (1.00) l ! List two of the RC3 leakage detection systems that shall be operable during Modes 1-4. (1.0) I 1 l l i l l l (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) l L _ ._ __.__ .___ .__ _ _ . - . . . _ _ _ _ - - _ - _
Hz__8DdINISIB8IIVE_EBQQEQUBEst_QQNDIIIQNit_8NQ_L1dII8IIQN1 PAGE 16' QUESTION 8.10 (1.00)
- e. Why will maintaining a Hydrogen overpressure in the VCT ensure that Oxygen levels are controlled in the RCS during power operations?(0.5)
- b. How is Oxygen controlled in the RCS during extended outages? (0.5)
QUESTION '8.11 (1.00) During refueling operations, an unexpected increase in neutron count rate occurs. Whose approval (s) is(are) required to continue refueling operations once the cause is corrected? (1.0) QUESTION 8.12 (1.50) Explain the reporting time frame for reportable occurrences requiring;(1.5)
- a. Immediate notification
- b. Prompt notification i
QUESTION 8.13 (1.00) What two people (positions) may authorize a tag out that puts the plant in G Technical Specification action statement? (1.0) QUESTION 8.14 (2.00) I OP 1622.009 (Reactor Building Entries) lists entry requirements for entry when the reactor is at power (>10 E-1%). List four of these requirements. l l l I (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
Az__8DdIN11188I1YE_EBQQEQUBEst_QQNQ1I1QNit_8ND_L151I8I1QN1 PAGE 17 QUESTION 8.15 (3.00)
- o. List four categories or types of temporary modifications which are governed-by the Jumper and Lifted Lead Control Procedure (OP 1000.28). (1.0)
- b. What temporary modifications are exempt from this procedure? (1.0)
- c. When should control of temporary modifications be governed by OP 1000.27 (Hold and Caution Card Control)? (1.0)
QUESTION 8.16 (2.50) Give five examples of SERIOUS INJURY which by definition envoke the Personnel Emergency Procedure EP 1903.93. (2.5) l l (***** END OF CATEGORY 08 *****) (************* END OF EXAMINATION ***************)
MC LICENSE EXANINATION HM00UT EQUATIONS. CONSTANTS. AfW CONVERSIONS 6=m*C*deltaT p 6=U*A*deltaT P = Po*10sur*(t) p , po.,t/T SUR = 26/T T = 1*/p + (p-p)/K p T=1/(p-5) T = ($-p)/Xy p = (Keff-1)/Keff = deltaKeff/Keff p=1*/1Keff+hff/(1+X.T) ' A = In2/tg = 0.693/tg K = 0.1 seconds-1 I = Io*e'"* CR = S/(1-Keff) 2 R/hr = 6*CE/d feet r Water Parameters 1gajlon=8.345lbm=3.87 1 ft = 7.48 gallons liters Density 9 STP = 62.4 lbm/ft3 = 1 gm/cm3 Heat of vaporization = 970 Btu /lbm Heat of fusion = 144 Btu /lba 1 atmosphere = 14.7 psia = 29.9 inches Hg. Miscellaneous Conversions 1 curie = 3.7 x 10W disintegrations per second . 1 kilogram = 2.21 lba 1 horsepower = 2 54 x 103 Btu /hr 1 mw = 3.41 x 10E Btu /hr 1 inch = 2.54 centimeters degrees F = 9/5 degrees C + 32 degrees C = 5/9 (degrees F - 32) 1 Btu = 778 ft-lbf
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5t__IHEQB1_QE_NUGLE88 EQWEB_EL8NI_QEEB8110Nt_ELUIDSt_8NQ PAGE 18 IHEBdQQ1N8dICS . ANSWERS -- ARKANSAS NUCLEAR ONE-2 -86/07/15-MCCRORY, S. ANSWER 5.01 (2.00) (ANY 4)
- 1. Temperature - Neutrons travel longer distance at higher temperature, therefore, higher probability of capture.
- 2. Boron concentration - As concentration goes up, fewer neutrons reach the rods.
- 3. Fission product poisons - As poison concentration increased fewer neutrons reach the rods.
- 4. Rod shadowing - When adj acent rods are inserted, the existing rod sees a lower flux..
- 5. Radial position - Rods near the center of the core see higher flux than peripheral rods.
- 6. Axial position - The farther a rod is inserted into the core, the greater the flux to which it is exposed.
- 7. Core age - As the core ages, boron is reduced which causes rod worth to increase.
- 8. Anti-shadowing - An inserted rod suppresses the nearby flux but causes a flux peak further out. A rod inserted into this peak will have a higher worth.
(0.15 each for factor / condition, and 0.35 each for explanation) l REFERENCE CE EQB t l l l I
St__IBEQBl_QE_NuGLE88_EQWE8_EL8NI_QEEB8IIQNz_ELVIQ1t_8NQ PAGE 19 IHEBdQQ1NedIQ1 ANSWERS -- ARKANSAS NUCLEAR ONE-2 -86/07/15-MCCRORY, S. ANSWER 5.02 (3.00)
- c. rho = (Keff-1)/Keff (0.25)
= (0.997-1)/0.997
= -0.003/0.997
= -0.003009 Cor -0.003)
= -0.3% (0.25)
From figure B-4.B rod insertion worth at 10" on group 6 ~ -0.56% dp. Therefore rod insertion worth when j ust critical = -0.26% dp. From figure B-4.B, this is about 70" on group six or about 60" of rod motion. (0.5)
- b. rho = beta /(1+1amda T) T= 26/SUR = 26/1 DPM =~26 (0.25)
= 0.0062/(1+0.1X26)
= 0.0062/3.6
= 0.00172
= 0.17% (ACCEPT 0.15 - 0.2%) (0.25)
From figure B-4.B rod position based on 'inal insertion worth of 0.09 (0.06 -0.11%) dp, rod height will be about 127" (125-130"). Therefore the additional rod motion will be 57" (55-60"). (0.5)
- c. (At this point in mode 3, the SDBCS would be set to control S/G pressure at about 1000 psia (545 deg. F). As reactor power passes the point of adding heat (1%), the RCS will begin to heatup which will raise S/G pressure causing the bypass valves to modulate open thereby reducing temperature and pressure. Equilibrium will be achieved when reactivity added by the power defect equals the excess inserted by the rods.) Final conditions will be*
RCS temperature - 545 deg. F (controlled by SDBCS) (0.4) RCS pressure - 2250 psia (pzr press control system) (0.2) Reactor Power 18% (based on power defect) (0.4) REFERENCE AN02 OP 2103.15 figures, OP 2104.02, Plant Specific Reactor Theory Manual ANSWER 5.03 (1.50) No. (0.5) Power will increase less during the heatup condition because of doppler feedback to turn power. (1.0)
It__IBEQBl_QE_NUQLE88_EQWEB_EL8NI_QEEB8IIQNi_ELu1Q1t_6NQ PAGE 20 IBEBdQDINed191 ANSWERS -- ARKANSAS NUCLEAR ONE-2 -86/07/15-MCCRORY, S. REFERENCE ANO2 RTTM, CH 17 ANSWER 5.04 (1.50)
- o. See attached graph.
- b. 185 +/- 5 REFERENCE ANO2 RTTM CH 15 ANSWER 5.05 (1.00)
When centrifugal pumps are arranged in parallel, you get an increase in the volumetric flowrate in the system for each pump that is started. (This cbility to increase flowrote is critical in an emergency situation. The pcrallel arrangement helps ensure reliability such that the seizure of a i pump will not impede flow from the other pumps but this consideration is not based on centrifugal pump characteristics.) REFERENCE PUMP LAWS and SAR considerations l l ANSWER 5.06 (2.50) l
- o. Nucleate boiling is most likely to start at the top of the core or fuel assembly (0.5) because this is the hottest part of the core and therefore nearer saturation conditions. (1.0)
- b. At the onset of nucleate boiling the heat transfer rate increases (1.0) (due to the removal of the heat of vaporization and the break up of the laminar layer at the clad surface).
REFERENCE bcsic thermo and core characteristics i
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Ez__IBEQBl_QE_NWQLE88_EQWEB_EL8NI_QEEB8IIQNt_ELu1Dit_8NQ PAGE 21 IBEBdQQ188dIQ1 ANSWERS -- ARKANSAS NUCLEAR ONE-2 -86/07/15-MCCRORY, S. ANSWER 5.07 (2.00)
- c. No change since required NPSH is a design criteria of the pump.(0.66)
- b. Actual NPSH will increase since temperature of the fluid is decreased. (0.67)
- c. Plant efficiency is decreased due to the additional energy removed in the condenser. (0.67)
REFERENCE CE EQB ANSWER 5.08 (3.00)
- c. Coolant temperature (0.5)
Coolant pressure (0.5)
- b. Rapid cooldown and depressurization (0.5) followed by rapid repressurization (0,5).
- c. Neutron exposure (integrated) (0.5) makes the. material more brittle (raises NOTT) (0.5).
REFERENCE CE EQB ANSWER 5.09 (1.50)
- 1. d
- 2. c
- 3. c 0.5 each l
REFERENCE l CE Reactor Theory l l t
Il i
'li__IHEQBI_QE_ NUCLE 88_EQWEB_EL8HI_QEEB8Il0Ni_ELU1911_8NQ PAGE 22 IHEBBQQIN851G1 ANSWERS -- ARKANSAS NUCLEAR ONE-2 -86/07/15-MCCRORY, S.
ANSWER 5.10 (2.00)
- 0. ~ 24 Hrs. Ceccept 20 - 30 hrs.) (0.5)
- b. Time would be shorter. (< 20 hrs.) (0.5)
- c. Due to the time delay in Xenon production from the decay of Iodine. (1.0)
REFERENCE CE EQB . ANSWER 5.11 (2.00)
- c. 2 (500 deg. F)
- b. 2 (EOL) c 1 (40%)
- d. 2 (Rod withdrawn) 0.5 each REFERENCE CE EQB ANSWER 5.12 (1.00)
Maderator density comes less at the top of the core causing the flux peak to move down in the core. (0.5) ASI will become more positive as the power to increased (since it is a function of 1-u/1+u). (0,5) REFERENCE CE EQB ANSWER 5.13 (2.00)
- c. true
- b. false
- c. true
- d. false 0.5 EACll I
li__IHEQBI_QE_NWQLE83_EQWER_EL8HI_QEE86IIQNi_ELWIQ1t_8NQ PAGE 23 IHEBdQQ1NedIQ1 ANSWERS -- ARKANSAS NUCLEAR ONE-2 -86/07/15-MCCRORY, S. REFERENCE BASIC REACTOR THEORY. h ) t l l
6t__EL8HI_1IIIE51_QE11SNt_GQNIBQLt_880_INSIBuBENI8110N PAGE 24 ANSWERS -- ARKANSAS NUCLEAR ONE-2 -86/07/15-MCCRORY, S. ANSWER 6.01 (2.00)
- o. 6.
- b. 1.
- c. 3.
- d. 7.
- o. 11.
- f. 4.
- g. 9.
- h. 8.
0.25 EACH REFERENCE AN02 STM-2-32 ANSWER 6.02 (1.50)
- c. 1. Lower seal cavity temperature
- 2. CB0 temperature
- 3. Middle seal cavity pressure
- 4. Upper seal cavity pressure
- 5. Vapor seal cavity pressure
- 6. CB0 flow 0.1 each
- b. Mid-stage seal cavity pressure would go to RCS pressure (0.3), upper seal cavity pressure would increase to approximately one half of RCS pressure (0.3). Potential increase in CB0 flow and temperature.(0.3) i REFERENCE ANO2 STM-2-03 CAA-52002-001-38)
; ANSWER 6.03 (2.00)
- 1. HPSI pump recirc valves shut, HPSI pumps (centrifugal pump) is running with a minimal flow -> overheating -> possible pump damage.
- 2. With containment sump valves opening and RWT valves open but closing, RWT is gravity draining to the sump.
l 1.0 each REFERENCE AN02 AOP 2203.18 n
6t__EL8HI_111IEdi_DE110Nt_GQNIBQLt_8NQ_INSIBudENIAIIQN PAGE 25 ANSWERS -- ARKANSAS NUCLEAR ONE-2 -86/07/15-MCCRORY, S. ANSWER 6.04 (2.50)
- o. 1. Remotely via interlocks
- 2. Automatic control circuits
- 3. Local control switch on breaker
- 4. Remote control switch (usually in control room)
- 5. Contacts in control circuit from protective circuits
- 6. Contacts in control circuit from ESFAS cny 3 at 0.33 EACH
- b. The breaker will remain shut. (0.5) l
- o. The breaker spring must be charged. (1.0)
REFERENCE AN02 OP 2107.01 ANSWER 6.05 (1.50)
- o. open
- b. closed
- c. closed
- d. open
, o. open 0.3 each REFERENCE i AN02 OP 2102.01 1 ANSWER 6.06 (2.00) i
- 1. One temperature element failing high causes the CCW flow controller valve from letdown heat exchanger to go full open. (1.0)
- 2. The other temperature element failing high will isolate the letdown flow to the CVCS red monitors and boronometer and also bypass letdown
, flow around the domineralizers to the VCT. (This feature protects i these components from high temperature conditions which they are not designed to handle.) (1.0) i i REFERENCE AN02 STM-2-4 (AA-52002-003-40)
1 l 6t__EL6HI_111IEdi_DE11ENt_GQNIBQLt_6NQ_IN11BudENI611QN PAGE 26 I ANSWERS -- ARKANSAS NUCLEAR ONE-2 -86/07/15-MCCRORY, S. ANSWER 6.07 ( .00) REFERENCE ANO2 LESSON PLAN AA-52002-016, T.S. 3.8, OP 1104.23 (bulk tank capacity) OP 2104.36 ANSWER 6.08 (3.00)
- c. Letdown valve opens (0,5)
Backup charging pump stops (0.5) PZR level decreases eventually trips on DNBP or LPPS (0.5]
- b. FWRV opens (0.6] causing S/G water level to increase Reactor trips on high S/G water level C+50 in) (0.9)
OR If switch is selected to "both", no trip occurs, only level deviation alarm. REFERENCE CE EQB ANSWER 6.09 (1.50) The increase in pressure decreases the modulate signal to the bypass valves Cnd increases the modulate signal to the atmospheric dump valves proportionally so that less steam is dumped to the condenser and more to Ctmosphere. (1.0) If pressure does not recover in the condenser (vacuum) ct 6.1" Hg, the bypass valves are interlocked closed. (0.5) REFERENCE AN02 STM-2-23 ANSWER 6.10 (1.50) Signals from CIS will initiate closure of all penetration room normal cupply and exhaust duct isolation dampers, At the same time, the CIS cignals will start both penetration room exhaust fans (2VEF-38A 18). REFERENCE AN02 STM-2-70 (AA-52002-013-2A)
)
6t__EL8HI_111IEdi_DE11GNt_CQNIBQLt_8NQ_INSIBudENI6IIQN PAGE 27 ANSWERS -- ARKANSAS NUCLEAR ONE-2 -86/07/15-NCCRORY, S. ANSWER 6.11 (2.00)
- c. Feed water flow will remain roughly constant. (1.0)
- b. 1. Resulting in an overfeeding of steam generator and (0.5)
- 2. Overcooling of the RCS. (0.5)
REFERENCE AN02 STM-2-69 (AA-52002-015-6A) ANSWER 6.12 (1.50) The radiation monitor on the letdown line provides for monitoring of gross cotivity and activity of a specific nuclide (I-135). 1 Increased gross activity with NO increase in Iodine activity would indicate o crud burst. Increased Iodine activity AND increased gross activity would indicate foiled fuel. (0.5 each) REFERENCE AN02 STM-2-62 (AA-52002-018-4C) ANSWER 6.13 (2.00) (ANY 2 AT 1 PT EACH)
- 1. Temperature shadowing correction: Corrects for changes in Tc which will affect the amount of neutron leakage from the vessel.
- 2. Rod shadowing correction: Corrects for neutrons absorbed by control rods inserted near the detectors.
- 3. Shape annealing factor: Compensates for stacked detectors seeing neutrons f rom areas of the core other than the adj acent area.
REFERENCE AN02 STM-2-65
4 Zz__EBQGEQWBE1_ _NQBU8L&_8BNQBM8Lt_EMEBEENGl_8NQ PAGE 28 88DIQLQQIG8L_GQNIBQL ANSWERS -- ARKANSAS NUCLEAR ONE-2 -86/07/15-MCCRORY, S. 1 ANSWER 7.01 ( .00) REFERENCE AN02 STM-2-70 ANSWER 7.02 (2.00)
- o. Quality Factor - Different types of radiation have different ionizing ability and therefore different potential for causing biological damage. (1.0)
- b. Gamma or X rays - 1 Beta - 1 Internal Alpha - 20 Thermal neutrons - 3 Fast neutrons - 10 (0.2 EACH)
REFERENCE AN02 AA-52009-001 ANSWER 7.03 (2.00)
- c. 100 mrem or j ust slightly less (based on next posting level beginning
', at 100 mR/hr).
- b. 1000 mrem or just slightly less (high rad areas are REQUIRED to be
( locked at 1000 mR/hr). REFERENCE AN02 AA-52009-001 ANSWER 7.04 (1.50) l 1. Temporary loss of normal control room lighting.
- 2. DG(s) auto started AND supplying 4160 V ESF bused (2A3/2A4).
j 3. No voltage indicated on non-vital 6900V AND 4160V buses. 1 (0.5 each) REFERENCE < AN02 E0P 2202.01 i
Zi__EBQGEQUBE1_ _NQBd8Lt_8BNQBd8Lt_EMEBGENGl_8NQ PAGE 29 88DIQLQGIG8L_GQNIBQL ANSWERS -- ARKANSAS NUCLEAR ONE-2 -86/07/15-MCCRORY, S. ANSWER 7.05 (2.00) 1.. IF PZR pressure decreases to less than or equal to 1400 psia (0.5) secure one RCP PER LOOP (0.5).
- 2. IF RCS margin to saturation decreases to less than or equal to 30 deg. F (0.5), secure all RCPs (0.5).
REFERENCE ANO2 OP 2202.01 ANSWER 7.06 (1.50)
- 1. >29% - based on pressurizer low level heater cutout setpoint (0.5)
- 2. <=82% - T.S. 3.4.4 limit, ensures a steam buoble is maintained in pressurizer to accommodate pressure surges during operation (prevent hydraulic solid system, protects reliefs against water relief).(1.0)
REFERENCE AN02 E0P ANSWER 7.07 (1.50)
- 1. CETs indicate that a superheat condition exists for the indicated pressure of the RCS.
- 2. Complete loss of feed to the S/Gs.
REFERENCE AN02 E0P ANSWER 7.08 (2.00)
- 1. If all RCPs are running and RCS temperature is < 500 deg. F, secure one RCP.
- 2. If pressurizer pressure decreases to <= 1400 psia, secure two RCPs.
i
. _ _ _ _ _ _ . _ . . _ _ _ _ . . _ _ _ _ _ . _ _ _ . _ . . _ _ _ . _ ~ _ __ _ _ _ . _ _ _ _ _ . . ~ . _ . . _ _ _ _ _ . . _ _ . . . _ , _ . . _ , . .
Zi__EBQGEQWBE1_=_NQBd8Li_8tNQBd8Li_EMEBGENGl_8NQ PAGE 30 86DIQLQEIG8L_GQNIBQL ANSWERS -- ARKANSAS NUCLEAR ONE-2 -86/07/15-MCCRORY, S. REFERENCE
~ AN02 E0P 4
. ANSWER 7.09 (3.00)
- o. 1. Pzr level control system CONTROLLING pzr level >29% but <82%.
- 2. MONITOR VCT level and pressure
- 3. NO S/G sample coolers, condenser off gas OR main steam radiation monitors showing abnormal trends.
- 4. NO CAMS OR containment area monitors in ALARM.
- 5. Containment sump level NOT INCREASING.
i
- b. 1. Teve > 540 deg. F but < 555 deg. F.
- 2. SOBCS CONTROLLING S/G pressure >950 psia but <1100 psia.
- 3. S/G 1evel >10% but (90% with MFW OR EFW systems RESTORING 1evel.
- c. 1. All RCPs RUNNING
> 2. RCS margin to saturation >30 deg. F.
- d. 1. Containment building pressure >13 psis but <16 psia.
- 2. Containment temperature <146 deg. F.
(0.25 each) REFERENCE I AN02 E0P 2202.01 ANSWER 7.10 (2.00) The EDG may over heat (1.0) due to lack of service water since the pump is , pcwered from the vital bus (1.0). i REFERENCE ANO2 E0P 2202.01 ANSWER 7.11 (1.50) f This will reduce the likely hood of fuel failures (0.5) due to large local power density changes (0.5) in the vicinity of the CEA finger tips (0.5). REFERENCE j ANO2 OP 2102.04
%* --..-r-,. ..-----.---------y_.---..,-.-s-,--,.~,,,,y. , . , - ,-...r--- .----~_,-.m_..--.---.-.,,.w.--.--r...,.--,,.-.-w-,,e--.--nv- + .---c.--+ r- ---+we.+o---y--
Zi__EBQGEQUBE1_=_NQBdeLa_8BNQBdeLt_EMEBEENGl_8NQ PAGE 31 88DIQLQEIG8L_GQUIBQL
= ANSWERS -- ARKANSAS NUCLEAR 0NE-2
-86/07/15-MCCRORY, S.
ANSWER 7.12 (1.00) By=boration (followed by rod motion to control ASI). REFERENCE AN02 OP 2102.04 ANSWER 7.13 (2.00) Cooldown - open pzr ECCS valve and initiate HPI (1.0) D pressurization - open pzr ECCS valve (1.0) REFERENCE AN02 OP 2102.10 ANSWER 7.14 (1.50) The RCS boron concentration will decrease (0.75) because a new resin bed uill absorb boron until it reaches equilibrium with the RCS concentration (0.75). f REFERENCE , l AN02 CVC SYSTEM MATERIAL l l l l l
At__8Qd1NISIBel1VE_EBQGEDMBESt_GQNQ1Il0 Nit _aND_L1dII6IIQN1 PAGE 32 ANSWERS -- ARKANSAC NUCLEAR ONE-2 -86/07/15-MCCRORY, S. ANSWER 8.01 (1.00) Immediately initiate and continue boration at => 40 gpm of 1731 ppm boric caid solution or equivalent until the required SDM is restored (or ccorgency borate). REFERENCE ANO2 T.S. 3.1.1.1 ANSWER 8.02 (1.50)
- 1. One OPERABLE high pressure safety inj ection pump (0.3)
- 2. One OPERABLE low pressure safety injection pump (0.3)
< 3. An independent OPERABLE flow path capable of taking suction from the refueling water tank on a SIAS and automatically transferring suction to the containment sump on a RAS. (0.9)
REFERENCE AN02 T.S. 3.5.2 ANSWER 8.03 (2.50)
- c. A leak rate is performed in order to identify potential problems of leakage from the primary system. (0.5) l b. Leakage shall be limited to:
- 1. No pressure boundary leakage
- 2. 1 gom unidentified leakage
- 3. 1 gpm total primary to secondary leakage through both steam generators or 0.5 gpm through a single steam generator
- 4. 10 gpm identified leakage from the RCS
- 5. Leakage specified (in Table 3.4.6-1) for RCS pressure isolation valves l (0.3 each)
- c. Leak rates are required every 72 hours during steady state operation
- except when in the shutdown cooling mode. (0.5)
REFERENCE l ANO2 T.S. 3.4.6 l l i l I l _ ___ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _
At__8Dd1N11IB8IIVE_ESQQEQUBEft_QQNDIIIQN1t_8ND_ Lid 1I8IIQNS PAGE 33 ANSWERS -- ARKANSAS NUCLEAR ONE-2 -86/07/15-McCRORY, S. ANSWER 8.04 (2.50) 2
- c. Alert
- b. NOUE
- c. SAE
- d. 'GE
- o. SAE (0.5 each)
REFERENCE AN02 OP1903.10 ANSWER 8.05 (1.50) (ANY 3)
- 1. Moderator. temperature coefficient is within its analyzed range.
- 2. Protective instrumentation is within its normal operating range.
- 3. The pressurizer is capable of being in an OPERABLE status with a steam bubble.
- 4. The reactor pressure vessel is above its minimum RT-NOT temperature.
(0.5 each REFERENCE AN02 T.S. BASES 4.1.1.5 ANSWER 8.06 (1.00) The time limits ensure inat the decay heat generated by the spent fuel will not exceed the cooling capacity of the spent fuel ptol cooling system. REFERENCE ANO2 T.S. 3/4.9.3 ANSWER 8.07 (1.00) B REFERENCE AN32 T.S. 3/4.9.9
At__6Dd1N11188I1YE_EBQGEQUBEnt_GQNDIIl0 Nit _8NQ_ Lid 1I8Il0NS PAGE 34 ANSWERS -- ARKANSAS NUCLEAR ONE-2 -86/07/15-MCCRORY, S. ANSWER 8.08 (1.00) Route an additional hose from an OPERABLE hose station. (Establish a fire watch.)
, REFERENCE AN02 T.S. 3.7.10.3 ANSWER 8.09 (1.00)
(ANY 2)
- 1. Containment Atmosphere porticulate radioactive monitoring system
- 2. Containment sump level monitoring system
- 3. . Containment atmosphere gaseous radioactive monitoring system (0.5 each)
REFERENCE AN02 T.S. 3.4.6.1 ANSWER 8.10 (1.00)
- e. The overpressure in the VCT ensures that a certain amount of Hydrogen is dissolved in the reactor coolant. In the presence of a neutron flux dissolved Hydrogen combines with dissolved Oxygen to form water,
- b. Specific chemicals (usually hydrazine) are used to scavenge Oxygen during extended shutdown periods where the residual neutron flux is inadequate for recombination.
(0.5 each) REFERENCE AN02 ops s ANSWER, 8.11 (1.00) Operaitons Superintendent (0.5) and Nuclear Engineering (Reactor Engineering) (0.5) i- REFERENCE ANO2 OP 2502.01
At__8DdIN11IB8IIVE_EBQQEDVBEft_GQNQ1IIQNit_8ND_ lid 1I611QN1 PAGE 35 ANSWERS -- ARKANSAS NUCLEAR ONE-2 -86/07/15-MCCRORY, S. ANSWER 8.12 (1.50) Icmediate - report within one or four hours (0.5) after determining reportability (0.25). Prompt - reports made expeditiously as possible (0.25) but within 24 hours (0.5). REFERENCE AN02 OP 1000.08 ANSWER 8.13 (1.00)
- 1. Operations Superintendent
- 2. Operations Manager REFERENCE AN02 OP 1000.27 ANSWER 8.14 (2.00)
(ANY 4)
- 1. RWP required
- 2. Must use buddy system
- 3. Neutron badges and dosimeters
- 4. Air sample l 5. Notify S.S. prior to entry
- 6. HP spproved
! (0.5 each) REFERENCE l l AN02 OP 1622.009 i i l l l l l I I l L_
At._8Dd1NISIB8I1YE_EBQQEQUBEst_QQNQ1110N1t_6ND_ lid 1I8I1QNd PAGE 36 ANSWERS -- ARKANSAS NUCLEAR ONE-2 -86/07/15-MCCRORY, S. ANSWER 8.15 (3.00)
- o. (any 4)
- 1. Electrical j umpers
- 2. Lifted leads
- 3. Mechanical bypasses or spool pieces
- 4. Mechanical configuration alterations
- 5. Blind or blank flar.ges
- 6. Rotate spectacle flanges (0.25 each)
- b. 1. Temporary modifications covered by approved procedures or work plans. (0.5)
- 2. Install / remove hand held j umpers (0.5)
- c. Temporary modifications (lifted leads, temporary grounds,'cr blank flanges) made soley for personnel or equipment protection. (1.0)
REFERENCE AN02 OP 1000.28 ANSWER 8.16 (2.50) (ANY 5)
- 1. Loss of consciousness (more than momentary)
- 2. Fracture (actual or suspected)
- 3. Head inj ury
- 4. Inj ury that may have damaged internal organs
- 5. Serious burn
- 6. Hemorrhaging (heavy bleeding)
- 7. Large dose of radiation (>50R).
l (0.5 each REFERENCE AN02 EP 1903.23 I l l l l l [ t l ,}}