ML20211J209
| ML20211J209 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 10/27/1986 |
| From: | Bill Dean, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20211J176 | List: |
| References | |
| 50-327-OL-86-02, 50-327-OL-86-2, NUDOCS 8611110037 | |
| Download: ML20211J209 (200) | |
Text
{{#Wiki_filter:_ - . _ . _ _ _ ENCLOSURE 1 EXAMINATION REPORT 327/0L-86-02 Facility Licensee: Tennessee Valley Authority ATTN: Mr. S'. A. White Manager of Nuclear Power 6N 38A Lookout Place 1101 Market Street Chattanooga, TN.. 37402-2801 Facility Name: Sequoyah Nuclear Plant
. Facility Docket No.: 50-327, 50-328
' Written, oral and simulator examinations wer administered at Sequoyah Nuclear Plant near Soddy-Da sy, Te psee, i- Chief Examiner:
WiMiain K Dean [ N ' _I /D Date Signed Approved by: / M 'oM7ds h4 F. Mu Agring Section Chief Date Signed Summary:
. Examinations: May 26-30, 1986 Written examinations were administered to 15 candidates, including .two-re-examinations; 12 passed, including one re-examination. Oral and simulator examinations were administered to 13 candidates, 11 of whom passed; one candidate failed the simulator exam and one candidate failed both the simulator and oral exams.
Based on these results, 4 of'5 R0s passed and 6 of 10 SR0s passed. Nk V ADO [
. - __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _]
REPORT DETAILS
- 1. Facility Employees Contacted:
*L. M. Nobles,. Superintendent of Operations
*J. M. Anthony, Operations Supervisor
*R. Joe Johnson, Director of Nuclear Training
*C. H. Noe, Supervisor Operator Training
*C. T. Benton, Simulator Training Supervisor
-* Clyde Brewer, Section Supervisor
*W. J. Glasser, Unit Supervisor, Audit, Surveillance, Training Unit
- Don Conner, Chief, E&TT Branch
*R. L. Merring, Engineering Training Section Supervisor
*B. E. Rodgers, TVA Corporate QA Branch
*R. C. Birchell, Compliance Engineer
*B. C. Lake, Training Shift Engineer
*Ed Keyser, Simulator Instructor
*W. G. Payne, Simulator Instructor
- Attended Exit Meeting
- 2. Examiners:
- Bill Dean Barry Norris Dave Nelson John Munro-Dave Graves
- Chief Examiner
- 3. Examination' Review Meeting At the conclusion of the written examinations, the exa:niners provided a copy.
of the written examination and answer key for review. The comments made.by the facility reviewers are included as Enclosure 3 to this report. The NRC Resolutions to these comments are given below. NRC Resolution of Facility Comments to Written Exam
- a. R0 Exam (1) Question 2.13 NRC Resolution: Agree. The answer key has been modified based on revised system description provided the NRC.
2 4' (2) ' Question 2.14 NRC Resolution: Agree. The answer key has been expanded to include reasonable radiation alarms indicative of a -leak. (3) Question 2.21
-NRC Rosolution: No action required. The question does not ask for specific mechanisms of instituting spray flow.
(4) _ Question 2.25
.NRC Resolution: Agree. The answer key has been modified to correct error in answer key. generation.
(5) Question 3.09 NRC Resolution: Agree. .. A dropped rod is a special case of a misaligned rod and may be substituted for.either response #4 or #5.
- (6) Question 3.11 l
l NRC Resolution: No action required. Candidates were instructed l by proctor that increasing severity of the same function would be acceptable provided that unique protective / control outputs occurred. (7). Question 3.14 NRC . Resolution: Agree. Based on revised procedure not held by the NRC, the answer key has been modified to allow answers in the range of 70% - 80% for full credi'. The previous answer of 85% is no longer an acceptable answer.
- b. SR0 Exam (1) Question 5.12 NRC Resolution: Agree. The answer key has been modified to reflect correct unit designation.
(2) Question 5.20b NRC Resolution: Disagree. Referring to the curves given with the exam, nuclear power is not at the C-2 rod stop setpoint at the time outward rod motion stops. Due to steam dumps being open, Tc decreases causing dT to increase. As-the dT increases, the OPdT setpoint decreases causing a runback and the rod stop. _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - ._ _]
E T. . - v. 3-l(3): _ Question 5.20c NRC Resolution: Agree. Th'e answer key has been modified to add an' additional correct answer. (4) Question 6.18 NRC-Resciution: Agree. The answer-key has been modified' based on
~a dditional material provided by the facility.
(5) Question 7.09 NRC Resolution: Agree. The answer key has been modified based on additional information not originally supplied by the facility and the answer key has been modified to accept Plant Manager (or authorized representative); however, Plant Superintendent is no longer an acceptable answer. (6) Question 7.15a NRC Resolution: The question instructs the candidates that a tube
-rupture has occurred. However, consideration will be given to candidates who made the assumption that a steam line' break occurred.
(7) Question 7.156 NRC Resolution: Disagree. The question asks the candidates to explain the bases for tripping the RCP's, not for the trip criteria as given in the facility comments. (8) Question 7.21 NRC Resolution: Agree. The answer key has been modified to reflect-the current CSF status trees. (9) ' Question 8.13 NRC Resolution: Disagree. Mode 6 is entered any time the vessel head bolts are less than fully tensioned. The facility comments support this position. (10) Question 8.14
~
NRC Resolution: Agree. Information in parenthesis is given for purposes of clarity and not required for full credit. (11) Question 8.15 NRC Resolution: Agree. The question has been deleted based on revised procedures and the sectional point value adjusted.
. . .. . va m
-4 (12): Question 8.16 NRC Resolution: Disagree. The question does:not ask for the DRPI
' operability requirements, ~ only for~ the TS requirements related to
~
Hot Channel Factors. No change to the answer key made. (13) Question 8.17
-NRC. Resolution: Disagree. The candidates should know TS 1.0 definitions; however verbatim response .is' not required for full credit. The point.value of the question is in accordance with the Examiner Standards.
(14) Question 8.20 1 NRC Resolution: Disagree. The question asks candidates how .to ensure the charging motor circuits will energize, not to determine
-if already energized. The charging ~ motor circuits will not energize if the toggle switch on the breaker cabinet is in the OFF-position. No change made to the answer key.
- c. R0/SRO Exam (1) Question 1.20/S.16 NRC Resolution: Agree. The answer key has been modified to accept <55F as the answer required for full credit.
(2) Question 2.12/6.11 NRC Resolution: Agree. The answer key has been . modified to accept either answer for full credit. (3) ' Question 2.11/6.12f NRC Resolution: Agree. This part has been deleted based on the possible confusion regarding which valve is being referred to, and the point value adjusted. (4)-Question 2.11/6.12j NRC Resolution: Agree. The answer key has been modified to accept CLOSE as the. full credit answer. (5) Question 2.16/6.14 NRC Resolution: Agree. The answer key has been modified to allow the inclusion of. the high current command to the stationary gripper coil based on the facility's comments. For full credit, however, the candidate must specifically state that the high current command is removed upon completion of rod motion.
.e .. 9 (6) Question 3.15/6.21 NRC Resolution: Disagree. The question asks for control or protective signals. RCS Subcooling meter and RVLIS are for indication only. COPS, however, does provide control / protection and the answer key has been modified to accept COPS. (7) Question 4.05/7.05 NRC Resolution: Agree that OPdT is more likely to cause a trip for the given situation. The answer key has been modified to accept OPdT as the full credit answer. (8) Question 4.18/7.19 NRC Resolution: Agree. This question has been deleted based on revised procedures and the sectional point value adjusted.
- 4. Exit Meeting At the conclusion of the site visit the examiners met with representatives of the plant staff to discuss the results of the examination.
There were _ no generic weaknesses noted during the oral and simulator examinations; however, it was noted that some Senior Reactor Operator candidates had a tendency to move too rapidly through Emergency Procedures, resulting in required steps not being accomplished. The cooperation given to the examiners and the effort to ensure an atmosphere in the control room conducive to oral examinations was also noted and appreciated. The licensee did not identify as proprietary any of the material provided to or reviewed by the examiners.
Enclorure 3 l SECTION 2 I QUESTIONS COMMENTS 2.13 RCP seal water return relief valves should be added to the answer key.
-See Attached 2.14 Condenser vacuum exhaust radiation monitoring should be added to the answer key.
2.21 There are two correct ways to initiate spray flow - See Attached t 2.25 Part 1 - Answer is reversed.
- a. Spring and Steam Pressure
- b. Air
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Page 36 of 40
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. SQNP
. GOI Units 1 & 2
. Page 8 of 25 . -
Rev. 40 I III. PREREQUISITES (Continued)
- 10. Select the highest reading source range channel and one inter-mediate range channel to be recorded on NR 45.
- 11. , Verify that the audio count rate channel is in operation and selected to the highest reading source range channel.
' 12. Notify chem lab that impending mode change from 3 to 2 re- -
quires sampling per SI-407 and SI-415 for startup conditional ,
.; requirements.
- 13. Shift Engineer verification that limiting conditions for op-3 eration are met without reliance on provisions in ACTION j statement unless otherwise excepted.
- 14. If a prolonged hot standby is evident.
t
- a. Borate the reactor coolant to the hot standby Xenon-free value according to SOI-62.2 boron
~
; concentration control.
CAUTION: The boron concentration differential between the pressurizer and reactor i coolant system shall not be greater than 50 ppm. When the difference is 1 50 ppm, turn on the pressurizer
, backup heaters to initiate automatic spray until concentrations are equalized.
- b. Insert control banks, if withdrawn, and establish hot standby conditions.
- c. Calculate S.D. margin and determine to be 1 1.6%.
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SQNP
, , GOI-5B - Units 1 & 2 {
Page 2 of 3 ' i Rev. 22 Unit Date III. PRECAUTIONS B. Anticipated changes in reactivity due to power variation or xenon changes will be compensated by programmed gradual changes in the loop boron concen-trations to minimize deviations from target flux maintaining an axial flux difference per TI-28, A.2. C. The baron concentration difference between the pressurizer and the reactor coolant systc= =ust not Exceed 50 ppm. D. If the loop boron concentration is changed by 10 ppm or greater, pressurizer sprays will be actuated by manual operation of sprays, causing B/U heaters to come on until the pressurizer concentration is within 10 ppm of the locp concentration. CAUTION: Monitor pressurizer pressure closely during manual spray operation to prevent dropping pressurizer pressure excessively. g E. In the event of a change in the rated thermal power level exceeding 15% 1 in one hour, notify the chem lab to initiate the conditional portions of
}{ SI-407 and SI-415 due to the thermal power change.
VI. INSTRUCTIONS Ub A. Drop turbine load from 100%~ to 80% power over one hour period so average load drop rate is about 1/3 %/ min. NOTE: This can be done by starting and stopping load drop on EHC system. B. Borate the amount according to Appendix A of this instruction dur-ing the load drop period to maintain control rods high in core. NOTE: The amount of boration may be varied to maintain flux within limits. - NOTE: Borate or dilute according to SOI-62.2. ib O t I
- - . . _. - .- . . . .- = -
4 . . . + Page 3 of 13 OPL271C014 02/11/86 Rev 0 Lesson Outline Instructor Notes
.A bistable is set to open the valve at
-1025 psig regardless of the controller setpoint. A pressure transmitter on^
each tee-header is used as the pressure reference , 4. Main Steam Isolation Valve and Check Valve Objective F
- a. Purpose - Isolates steam flow to the turbine in event of:
(1) Break upstream (2) Break downstream (3) Tube ' rupture ,
- b. Isolation signals for above accidents (1) High steam flow, 2/4 loops (2) Phase B containment isolation
- c. Closing time: 5 seconds or less (limits containment pressure rise)
- d. Type - 32 inch globe wye, air to open,
, spring to close with flow assist (angle). Air accumulator slows down valve to prevent its slamming shut.
- e. Check valve - prevents reverse flow of steam into faulted line from upstream break
- 5. Bypass Valve - Used to warm and pressurize
-e steam lines upon startup. When AP across the MSIV =25 psid, the HSIV may be opened
- 6. Piping
- a. 36"
- b. Supplies turbine via cross connection (36" header) which guarantees pressure (i 10 psid) equalization to all steam dumps. Also seals and HSR reheat tap off this line*
O s
SECTION 3 4 QUESTIONS CONMENTS 3.09 A dropped rod will bring in the alarm and should be considered as a correct answer. 3.11 The statement, " increasing severity of the same function like Lo and Lo-Lo level count as one response", is misleading. The Answer Key list Lo and Lo-Lo as separate responses. Lo T avg is 554*F and initiates feed water isolation. Lo-Lo T ayg is 540*F and allows block of steam line high flow SI. 3.14 The runback setpoint has been changed. The current revision of SOI-5.1
& 6.1 states the turbine will runback to ~75%. The candidates could give either answer. See Attached f
I
7...
.- . SQNP SOI-5.1 & 6.1 - Units 1 and 2 h0 Page 7 of 10 Rev. 10
," B. Continued
- 4. Open No. 3 heater drain pumps discharge header isolation valves to
, heater strings "A," "B," and "C" and place their control switches ; in the "P-auto" position.
1
- a. FCV 6-108 for "A" heater string
- b. FCV 6-109 for "B" heater string
, c. FCV 6-110 for "C" heater string I
L
- 5. At E40% power start 2nd #3 heater drain tank pump.
(, 6. At = 80 percent power start the 3rd #3 heater drain tank pump. l. CAUTION: If the . level in the No. 3 heater drain tank goes high ( enough to open the dump to condenser valves and turbine l^ load is > 80 percent, the turbine is run back to
- 75 percent power.
$ CAUTION: LCV 6-106B will close if the AP across the No. 3 heater I
drain pumps suction and discharge headers drops to 490 psi when one or more heater drain pumps are running. CAUTION: LCV 6-106B will have to be manually reset with local " Reset" pushbutton before the valve can be reopened. Close the air supply valve to LCV 6-106B before reseting the valve. After the PCV is reset open a'ir supply valve slowly while monitor-ing the No. 3 heater drain tank level to prevent low level trip. C. No. 7 heater drain pump operation *
- < 1. Check each No. 7 heater drain pump ready for operation by:
- a. Verify injection water flow to seals.
- b. Verify oil levels normal.
- c. Verify raw cooling water to oil coolers.
[, d. Place local " Test-Reset-Safe Stop" switch to " Reset" position. l 2. Place each No. 7 heater drain pump auxiliary oil pump local " Test-Reset-Safe Stop" switch to " Reset" position. I
- I e
g E e 6
Y , ' I SECTIONS 5 & 1 l QUESTIONS CONMENTS 5.12 The answer key addresses units 3 and 4; however, it should be:
- a. 2
- b. 1 5.16 Less than full load AT is not in our procedures. See 1.20 attachment.
5.20 b. Answer key states outward rod motion stops due to OPAT. The OPAT rod stop/ runback is at 3% below the OPAT trip setpoint
~109% - 3% = ~106% power.
Because the NIS power range rod step is at 103% power (lower than OPAT). The correct answer is NIS C-2 rod stop prevented outward rod motion. Answer key should allow NIS cod stop or the OPAT rod stop. 5.20 c. Answer key states the SF indication decreases due to P steam decrease. However, because the S/G press is decreasing, the steam flow should decrease because flow is proportional to the Y5 Answer key should be revised to accept statement that there is less delving force in S/G (lower S/G press); therefore SF decreases. J
.. = -. - - - --
r .
" ' SQNP AOI Units 1 & 2 Appendix A Page 1 of 1 Rev. 2 NATURAL CIRCULATION GUIDELINES A. Guidelines to determine if natural circulation is taking place in primary system under subcooled condition.
- 1. Relatively stable A T with < 55 F with gradual decrease.
- 2. Determine core A T as'follows:
(a). Wide range RTD's (hot and cold legs) OR (b). Wide range RTD's _(cold legs) and incore T/C's.
- 3. Incore T/C's temperature indicating below saturation temperature for the existing primary system pressure.
- 4. RCS heat being removed by secondary system:
(a). S/G's steaming and water being added to S/G's. (b). Steam pressure near saturation for RCS temperature B. Instructions to enhance natural circulation.
# 1. Keep S/G 1evels in narrow range (tubes covered), between 25% and 34 50% for post accident instrument error.
NOTE: (Unit 1) When cooling down during natural circulation and it becomes desirable to depressurize the RCS; use the 5 relocated incore T/C's to determine the upper head temperature. Then base the RCS pressure reduction on the highest operable T/C , of the 5 to prevent reaching saturation in the upper head area. NOTE: (Unit 2) At this time U-2 does not have the 5 relocated incore T/C's. Therefore use the existing 65 incore T/C's to determine the highest temperature in the Upper Head. Base reduction of RCS pressure on the highest operable of
- the 65 T/C's.
- 2. Keep RCS pressure above saturation pressure for the existing hot leg (W.R.) or incore T/C temperature.
- 3. Use S/G PORV's to steam of f and cool RCS.
- 4. Initiate SI if RCS pressure drops below 2000 psi unexpectedly then 9 evaluate condition and refer to emergency instructions.
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.s . SQNP g,, , ...-' ES-0.3 Unit 1 or 2 Pega 2 of 9 Rev. 0 j NATURAL CIRCULATION COOLDOWN STEP ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION: If SI actuation occurs, then E-0, Reactor Trip Or Safety Injection, should be used.
Note: If at any time an RCP can be restarted, then go to appropriate normal cooldown instruction 1 Try To Restart An RCP IF an RCP can NOT be started,
- a. Loop 2 preferred TIIEN verify natural circulation
- b. Refer to SOI-68.2
- RCS subcooling
- c. Start an itC'P an'd go to appropriate plant S/G press stable or instruction decreasing -
/.
T-hot stable or decreasing Core exit T/C stable or decreasing T-cold at saturation temp' for S/G press IF natural circulation
} TOT verified, TilEN increase dumping steam a
l? - ,
. +
SECTIONS 7 & 4 QUESTIONS COMMENTS 7.5 Answer key should accept OPAT as well at OTAT. OPAT should occur 4.5 first (~109%) and OTAT second (~114%) (The reason AOI-3 states OTAT as the trip function is because OPAT is not taken credit for in the FSAR accident analysis. Ref. TS bases for OPAT pg. B-2 attached) 4 4 7.9 b. Answer key should accept Site Emergency Director - See Attached ' IP-15 Answer Key states Plant Superintendent; however, title has been changed to Plant Manager, should accept either as correct answer. ! 1 i 7.15- a. The words " Faulted Steam Generator" is misleading and implies
- that a steam line break is present.
~
'b. The RCPs are not tripped on a SGTR unless the pressure decreases uncontrollable to less than 1250 PSI.
The question is based.on outdated procedures, i The answer key is confusing. i
- l 7.19 The question is misleading because it is based on outdated procedures.
4.18 ( Any answer that verifies safety injection should be accepted. i 7.21 The answer is based on outdated E0Is - It should reflect current status l trees. - See attached. l 1
V . LIMITING SAFETY ~ SYSTEM SETTINGS BASES Intermediate and Source Rance. Nuclear Flux (Continued) Range Channels will initiate a reactor trio at a current level proportional to approximately 25 percent of RATED THERMAL POWER unless manually blocked wnen F-10 becomes active. No credit was taken for operation of the trips associ-ated with either the Intermediate or Scurce Range Channels in the accident analyses; however, their functional capability at the sqecified trip settings is required oy tnis specification to enhance the overall reliability of the Reactor Protection System. Overtemoerature aT The Overtemperature delta T trip provides core protection to prevent DNS for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with' respect to piping transit delays frcm the core to the temcerature detectors (about 4 seconds), and pressure is within the range oetween the Hign and Low Pressure reactor trips. This setpoint includes corrections for axial power distribution, changes in density and heat capacity of water with temperature and dynamic compensation for piping delays from the core to the loop temperature detectors. With normal axial power distribution, this reactor trip limit is always below the core safety limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the di.fference c' etween top and bottom power range '- nuclear detectors, the reactor trip is automatically reduced according to the notations in Table 2.2-1. Operation with a reactor coolant loop out of service Delow the 4 loop P-8 setoolnt does not require reactor protection system set point modification because the P-8 setpoint and associated trip will prevent DN8 during 3 loop operation exclusive of the Ov'rtemperature e delta T setpoint. Three loop operation above the 4 loop P-8 setpoint is permissible after resetting the K1, K2, and K3 inputs to the Overtemperature delta T channels and raising the P-8 setpoint to its 3 loop value. In this mode of operation, the P-8 interlock and trip functions as a High Neutron Flux trip at the reduced power level. Overnower AT The Overpower delta T reactor trip provides assurance of fuel integrity, e.g., no melting, under all possible overpower conditions, limits the itu' sc range for Overtemperature delta T protection, and provides a backup ta tre Hign Neutron Flux trip. The setpoint includes corrections for changes in density and heat capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop temperature detectors. No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection Systee. SEQUOYAH - UNIT 2 8.2-4
K . SQNP
~'
REP-IPD SQN, IP-15 Page 1 of 2 Rev. 4 ENERGENCY EXPOSURE GUIDELINES .. 1.0 PURPOSE This procedure provides guidance as to the amount of radiation exposure that is acceptable for various types of activity. l 2.0 PROCEDURE 4 4
- p. 2.1 -Life Savina Actions (75 rems) 7 This applies to lifesaving actions for individuals or to prevent @
serious injuries to a large number of persons'. p 2.1.1 Y Rescue personnel must be aware of possible consequences of ff 2 such an exposure and selected on a voluntary basis unless r -
/F" they are members of an emergency team and have previously
- consented to receive this exposure. Following the exposure,
- these individuals must be removed from areas.where they- '
could receive another emergency dose. 2.1.2 Women capable of reproduction should not take part in these , actions. ,
, 2.1.3 Other things being equal, the oldest volunteer preferably should be selected.
2.1.4
~
p6- Planned dose to the whole body shall not exceed 75 rems. 1 2.1.5 Hands and forearms may receive an additional dose of up to
)F- 200 rems (i.e., a total of 275 rems).
2.1.6 Internal exposure should be, minimized by the use of respira-tory protection equipment. Respiratory protection factors are given in attachment 1 of this procedure. i 2.1.7 Contamination should be controlled by the use of available i protective clothing. 2.1.8, Normally, exposure under these conditions shall be limited to once in a lifetime. NOTE: The Site Emergency Director will determine the amount
, of exposure that will be permitted in order to perform i the emergency mission.
J 4 G-s' 1
. g.. .c:, - . g,
7 . SQNP RCI-l Page 9 of 20 n w, :n IV. RADIATION PROTECTION STANDARDS (continued) A. Maximum Permissible Exposure to Radiation (continued) Non-TVA personnel shall be limited to the following maximum whole body exposures: (1) 300 mrem / calendar quarter, or (2) 1,250 mrem / calendar quarter (up to 5 rem per year with lifetime occupational radiation exposure considered) if dose records are supplied for the indiviQ?al(s) for the present calendar-quarter. The exposure permitted shall be adjusted so that the total dose received shall not exceed the 1,250 mrem / calendar quarter limit., or (3) 3,000 mrem / calendar quarter (up to 12 rem per year with lifetime 6ccupational radiation exposure considered) if the requirements of item (2) above and 10CFR20.101(b) are met and the written authori-zation of the individual's employer is obtained.
- 2. Exposure Exceeding Limits of 10CFR20 Any individual who receives an exposure to radiation in excess of the limits shall be removed from further exposure for the~ remainder of the applicable period. A report shall be completed by the IIP Supervisor and an investigation with a written report of its findings and recommendations shall be conducted by the Plant Operations Review Committee (PORC).
~
Reports of the incident shall be made to the NRC per applicable parts of 10CFR20, Section 20.405, and AI-18, " Plant Reporting Requirements".
- 3. Emergency Exposure In emergency situations, normal exposure limits may be exceeded. Guidelines for limiting emergency exposures are set forth in Sequoyah Implementing Procedures, IP-15.
These guidelines shall not be used without the knowledge and consent of the Plant Manager, or his duly authorized representat'ive. 0008N/rmk
- Swn r FR-0 Unit 1 er 2 I '. -
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SECTIONS 6, 2 & 3 QUESTIONS COMMENTS 6.11 Air or Air Solenoid Valve should be accepted. 2.12 6.12 Part f. - RWST to SI Pump Suction - 2.11 Answer key states these valves receive an "Open" signal. If the question is referring to FCV-63-5, then the answer should be "No Change". If the interpretation for "SI Pump" is used for the CCP Suction [Sometimes called High-Head SI Pumps], then the answer would be open 4 (FCV-62-135/136) Request that both answers be accepted due to 2 different interpretations of SI Pump. Question could be referring to FCV-63-5 or FCV-62-135/136. Ref. 47W611-63 Part j. - Steam Supply Valve To Main Feed Pump Answer key states these valves receive "No Signal". However SIS 4MFW Isolation + Closes MFP steam. supply. Request answer key accept CLOSE 6.18 Liquid Space RID is used when a bubble is being drawn in the PZR. - See j Attachments I i 6.14 The stationary gripper will receive a high current signal first; 2.16 therefore, the correct answer is: b,c,f,e,b,d,a
, . o SECTIONS 6, 2 & 3 QUESTIONS COMMENTS
. J' h' 'ilt 6.21 The question asks what control or protection SIGNAL is generated by the 3.15 pressure instrument on T-hot leg.
In addition to the answer given on the key any one of the following should also be accepted: 1)-Cold over press input k d
- 2) RCS Subcooling meter input s frAidu p
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- 3) RVLIS Input
- 4) RHR interlock 380# and 700# #> W f y 6% j 4
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Initiate SI-127 "RCS and Pressurizer Temperature and u
' )(-'
g. y-- Pressure I.imits." i
- h. As the pressurizer temperature increases, water vapor and
- 2 ,
gas will be forced over to the PRT through the power relief valves. Monitor the PRT tank pressure and temperature as the pressurizer is heated. The pressurizer relief line j w temperature indicator should start increasing as soon as steaming is initiated in the pressurizer. i ,
- i. Cool the PRT by spray and drain operation as required to F J
~
maintain the temperature below the high temperature alarm. Vent the PRT if the gas pressure rises to near 3 psig through PCV-68-301. , l I j. When the water and vapor temperatures are within about 5 F of each other, close the power relief valves to c allow the pressurizer pressure to increase by action of j the pressurizer heaters.
-18. Maintain Lhe RCS temperature < 160 F by adjusting flow l ,,
' through the RHR heat exchangers (do not stup and start l RifA pumps for temperature control). ,
l Continue pressurizer heatup to 430 F. Allow RCS pressure 19. j j to increase to 325 psig. , x
! , 20. Gradually reduce pressurizer level to program (24.7%). fyl
! Place level control in AUTO when level is normal. This ^' is to limit temperature transients on loop 2 hotleg.
- 21. Start the reactor coolant pumps (SOI-68.2). After five minutes running, sample the RC for chemistry specifications (limits in SI-50 and 51). Partially open pressurizer sprays for mixing and turn on backup heater for pressure control.
- 22. Stop residual heat renoval system cooling operation by 3
closing HIC-74-16 and HIC-74-28 while maintaing flow l i thru HIC-74-32. Allow RC temperature to increase to 180'F. Continue ! 23. y SI--127. CAUTION: Do not exceed 180*F on RCS until water chemistry is within specifications. (TI-27, Table 48, or SI-50 and SI-51). i 1 -
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SECTIONS 6, 2 & 3 'I QUESTIONS COMMENTS 6.21 The question asks what control or protection SIGNAL is generated by the 3.15 pressure instrument on T-hot leg. In addition to the answer given on the key any one of the following should also be accepted:
- 1) Cold over press input
- 2) RCS Subcooling meter input
- 3) RVLIS Input
- 4) RHR interlock 380# and 700#
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;- PAGE I-b OF.lh._.-. ' l l !
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_- r B.14 SATURATION MONITORING
- INTRODUCTION The plant process computer is utilized to continuously monitor pressure and temperature margins-to-saturation of the primary coolant system, and to give early warning when any of these margins reaches preset limits. The following description is provided to familiarize the operator with the methods used for this monitoring, as an aid in the evaluation of saturation data to be used for trend or display, and in recognizing the type of abnormal indication that can result when certain computer inputs are unavailable or unreliable.
DESCRIPTION Saturation pressures, temperatures (UO901 thru UO905), and their associated margins-to-saturation (UO984 thru UO989), are calculated based on the following primary system inputs:
- 1. System pressure (lowest pressurizer or wide-range)
- 2. Hot leg temperature (hottest)
- 3. Average incore T/C temperature
- 4. Hottest incore T/C temperature The single pressure input used in the calculations is selected from the lowest of the four narrow-range pressurizer pressures when system pressure is above 1700
( psig. Wide-range pressure is used below 1700 psig and when there are no narrow-range inputs available and reliable. If the wide-range pressure also is unavail-able or unreliable, a minimum pressure value is used (unreliable zero).. D1is causes all margin values to be indicated as unreliable and removed from limit checking. Use of this minimum pressure in the calculations also causes the margin values to go low (usually negative), which results in a sudden zero scale pen indication for any margin being trended. Alarm printout and annunciation are suppressed under these conditions.
. The hottest wide-range loop temperature is selected from the four hot leg inputs on the basis of availability and reliability also. However, only those outputs associated with this temperature are affected (pen movement, alarm suppression) if none of the four inputs are usable. (700 degf unreliable is usad.)
Average incore temperature and hottest incore temperature are calculated by a standard-W program. This program uses temperature values for any thermocouple that is reliable and not in alarm; scan removed status is not checked. Therefore, during normal operation, if an incore thermocouple is removed from scan, a value of zero should be entered. This will ca'use it to be in low alarm and not be used in the average or selected as the hottest. ( B-30 2/81
/ . - - - ,
SUMMARY
OF INPUTS AND OUTPUTS A list of all program inputs and outputs can be obtained from the operator's { console using the DATA DUMP function. This listing should be requested when initially selecting a value for trend, and periodically thereafter to check data quality and verify pen recorder indication. Analog inputs that are scan removed or have been out of range will be marked with a letter 'S' or an as-terisk. Unreliable calculated values (program outputs) are marked with an asterisk and should not be used for trend. If there are no reliable outputs, a check of computer hardware and/or plant instrumentation should be requested. PEN RECORDER TREND Operator Procedure No. 1 (from W manual TPO44)
- 1. Push ANALOG TREND Function Button
- 2. Select Address on Alphanumeric Keyboard 3 Push ADDRESS Button 4., Select Pen Number (1 to 12) on Keyboard (see note)
- 5. Push Value 1 Button 6 Select Pen Position on Keyboard
- 7. Push Value 2 Button
- 8. Select Range on Keyboard
- 9. Push Value 3 Button
- 10. Push START Button Pen Position is the minimum scale value, Range is the maximum scale minus the I minimum scale. .
After the pen recorder is started, verify that the pen moves to the per cent of scale that corresponds to the value shown in the printout. Example: Pen Position = 0 (step 6) Range = 100 (step 8) TSAT Margin = 37.5 degf (from printout) Pen Indication = 37.5 per cent Note: It is suggested that Pen No. 1 be used as Pen No. 2 thru 12 are used for post accident monitoring, and are set up auto-matically by the PAM program. B-31 2/81
SATURATION MONITORING TABLE A PROGRAM INPUTS AP Sym Instrument Range Units PO480A PT-68-340 Pressurizer 1 press 1700-2500 psig PO481A PT-68-334 Pressurizer 2 press 1700-2500 psig PO482A PT-68-323 Pressurizer 3 press 1700-2500 psig PO483A PT-68-322 Pressurizer 4 press 1700-2500 psig PO499A PT-68-68A RCL System press 0-3000 psig T0419A TE-68-1 RCLA Hot temp 0-700 degf T0439A TE-68-24 RCLB Hot temp 0-700 degf T0459A TE-68-43 RCLC Hot temp 0-700 degf T0479A TE-68-65 RCLD Hot temp 0-700 degf UOO90 Inst Value of Hottest Incore T/C 0-2500 degf UOO91 Inst Value of Average Incore T/C 0-2500 degr K5502 PSAT Margin Annune Alarm Setpt 200 psi K5503 TSAT Margin Annune Alarm Setpt 15 deaf ( PROGRAM OUTPUTS UO900 Press Value Used for Sat Cale psig-UO901 Sat Press for Hottest RCL Temp psig UO902 Sat Press for Hottest T/C psig UO903 Sat Press for Avg Incore Te:np psig Hot Leg Temp Used for Sat Cale
~
UO904 degf UO905 Sat Temp for System Pressure degf UO984 PSAT Margin for Hottest RCL Temp psi Uo985 PSAT Margin for Hottest T/C psi UO986 PSAT Margin for Avg Incore Temp psi Uo987 TSAT Margin for (RCL T) for Sys Press degf UO988 TSAT Margin for (UOO90) for Sys Press degf UO989 TSAT Margin for (Uo091) for Sys Press degf ( I ! B-32 2/91 l
~ . .
~
. J 'e r- SATURATION MONITORING
. ,, . *ABLE B
(~- CALCULATIONS AND GRAPHICAL REPRESENTATION Pressure Margin to Sat. = System Pressure - Saturation Pressure (fig. 1) Temperature Margin to Sat. = Saturation Temp. -' Measured Temp. (fig. 2) i: Figure 1 2235 psig /A y M8 rg[g8 The saturation pressure indi-o cated is the pressure at which Pres , 'y - saturation will occur at the current temperature of 607 degf.
/ oE*b09Ngf f
Temp . Figure 2 3 for,b2$,hsig
/
/n sat margggmp The saturation temperature 607 degf ,,. ' # indicated is the temperature at which saturation will occur at Temp ,/ the current pressure of 2235 psig
/
[ l l Pres i t B-33 2/81 l l 1
o e. . SECTION 8 QUESTIONS CONNENTS l 8.13 Mode 5 should also be accepte'd because of RCS T>140*F. Definition of Mode 5: Ke rr <.99 0% pwr RCS T<200*F The Mode 5 condition does not footnote the " Fuel in the Rx vessel w/the vessel head closure bolts less than fully tensioned or head removed". Both Mode 5 & Mode 6 should be accepted. 8.14 Answer Key reference is wrong. It should be AI-8. AI-8 does not mention the storage area being lead shields. See Attached - 8.15 The question and answer is based on a old revision of OSLA-73.
~
The present revision does not designate a Stationary Fire Watch. 8.16 The Answer Key is based on bases for entholpy rise hot channel factor; however, LCO 3.1.3.2 states that rod position indication system shall , be capable of determining the rod position within il2 steps; j therefore 12 steps or i13 steps should be accepted. - See Attached l
, 8.17 This question's point value is too high.
Implies that the entire definition section of T.S. needs to t^ committed to memory. l 8.20 a. You cannot determine that the charging motor is energized prior to racking the breaker in.
/
/
/
o o.
- SQNP AI-8 Page 2 of 7 Rev. 15 2.0 PROCEDURE (continued) 2.2.3 Prior to closing containment and/or the annulus (and including fuel handling areas), after an outage / refueling or major maintenance, the SE in coordination with .
, the Public Safety Shift Supervisor, will require a search for explosives,
- incendaries and other devices. Two. teams a. se will conduct the search if necessary, one in upper containment and fuel handling -
areas, and one in lower containment and annulus. The teams will consist of a Public Safety Officer, an operator and HP representative. HP participation will be based on radiological conditions. Documentation of this requirement shall occur on the attached containment entry checklist data sheet. 7 2.3 The SE shall approve the number of personnel required for all initial entries to containment., 2.4 Prior to entry into lower containment or the annulus the incore flux detectors shall be verified
- to be in the storage position or inserted to within ten (10) feet of the core. The SE shall initiate a hold order clearance on the incore flux drive motors control power. This hold order shall remain in affect until the SE is assured all personnel have been cleared from containment and the -
personnel access is locked. Prior to issuing a ' radiation work permit (RWP) for lower containment or annulus, HP shall verify incore detec. tor 'ystem s is tagged with a hold order. The SE will issue incore detector system hold order clearance to HP Shift Supervisor by title. The incore' detector hold order clearance will remain issued to HP Shift Supervisor by title at all times except when running core maps or performing incore detector system maintenance that w; requires the detector system to be operated while , persons are in the incore instrument room.
'8' Fuel handling areas is defined as that area on the -
fuel floor that will be'within control zones delineated in AI-26. ' 0027A/mbs . 2 "p ' -
.J
Paga 1 O Hl. A'l3 02/20/86 I .. . SEQUOYAH NUCLEAR PLANT OPERATIONS GROUP OPERATIONS FIREWATCIIES
References:
- 1. SQO37, Fire Protection Manual N82FP-1, FP-3
- 2. Memorandum from C. R. Brimer to P. R. Wallace dated June 14, 1985 (S OI-850614-957)
- 3. CATS 85-409
- 4. Memorandum from C. R. Brimer to P. R. Wallace dated February 18, 1986 (S OI 860218 818)
Firewatch personnel are under the supervision of the Operations Supervisor and his subordinates whose responsibility is to ensure the watches are manned, tours are made according to the established routes and times, journals 'are kept, and to evaluate any problems reported by the firewatch personnel including initiating corrective action deemed necescary as a result of reports from them. There are two (2) designated firewatch patrol routes. Each route should be completed within forty-five minutes and shall be completed each hour. The designated routes are: A. Control Building - all elevations Auxiliary Building - elevations 759, 749, and 734 East Main Steam Valve Rooms - elevation 763 roof B. Auxiliary Building - elevations 714, 600, 660, and 653 Additional Equipment Buildings - elevation 706 West Main Steam Valve Rooms - elevation 714 Checklists to be completed by the firewatch each shift are provided in iI Appendix A. These checklists shall be turned in to the shift engineer for review then forwarded to the Operations Supervisor's office through the morning mail' pickup. Throughout the route, the firewatch shall observe and inspect the general area for substandard fire protection and i detection systems, open or breached firestops, seals, fire doors, or fire dampers and transient fire loads or any other fire safety threats, perform hourly checks or PFBBP in their area (Penetration Fire Barrier Breaching Permits), and shall promptly inform the SE of any detrimental conditions lI encountered in his duties. E *
- E l *
- Revised
- I '
I
s' ( - Paga 2 OSLA73
> o. . 02/20/86
=
m
- The Route B firewatch shall immediately notify Operations (shift engineer, assistant. shift engineer, unit operator) upon indication
" of a rapid rise in temperature or temperatures in excess of 130 F
- in the U1 or'U2 690 Pipe Chase or U1 side ERCW pipe tunnel.
- Firewatch personnel assigned to Appendix "R" Routes A and B e shall maintain a daily journal to record all abnormal conditions or actions observed by the firewatch. The logs shall be turned in to the shift engineer for review then forwarded through the morning m mail pickup to the Operations Supervisor's office.
The SE shall update the firewatch as necessary to meet the requirements of Tech Specs and Physi 13 by adding and/or e deleting firestop breaching and fire protection inoperability or other substandard equipment. o
- Stationary firewatches will relieve at their assigned watch stations, Do not leave firewatch stations (routes) unmanned. All relief shall be by direct face-to-face responsibility communication' with the person assuming for the watch.
- Firewatch patrol personnel shift relief stations are as follows:
Route A. Control Building at AUO sign in desk B Route B. Auxiliary Building el. 669 at AUO station Any problems encountered in supervising this program shall be
, brought to the attention of the Operations Supe isor~ for res on .'
, !i
.n ft/ff/h
*Supv Oper tions Group O'
- Itevised
'O
- rJ
!O l 10 i i L 1
I nao e POWER DISTRIBUTION LIMITS BASES Each of these is measurable but will normally only be determined periodically as .specified in Specifications 4.2.2 and 4.2.3. This periodic I surveillance is sufficient to insure that the limits are maintained provided:
- a. Control rods in a single group move together with no individual rod insertion-differing by more than + 13 steps from the group demand position.
- b. Control rod groups are sequenced with overlapping. groups as described in Specification 3.1.3.6.
- c. . The control rod insertion limits of Specifications 3.1.3.5 and i
3.1.3.6 are maintained.
- d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
F H will be maintained within its limits provided conditions a. through
- d. above are maintained. As noted on Figures 3.2-3 and 3.2-4, RCS flow. '
and F H may be " traded off" against one another to ensure that the calculated DNBR will not be below the design DNBR value. TherelaxationofFhasa function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits. When RCS flow rate and F" are measured, no additional allowances.are necessary prior to comparison with the limits of Figures 3.2-3 and 3.2-4. Measurement errors of 3.5 percent for RCS total flow rate and 4 percent for F > have been allowed for in determination of the design DNBR value. Rj , as calculated in Specification 3.2.3 and used in Figure 3.2-3, accounts for F H less than or equal to 1 49. This value is the value used in the various safety analyses where F AH influences parameters other than DNBR, e.g. peak clad temperature, and thus is the maximum "as measured" value alle. ed. R2 , as defined, allows for the inclusion of a penalty for Rod Bow on DNBR only. Thus, knowing the "as measured" values of F and RCS flow allow for H
" trade off" in excess of R equal to 1.0 for the purpose of offsetting the Rod Bow DNBR penalty.
SEP 2 91983 SEQUOYAH - UNIT 2 B 3/4 2-2 Amendment No. 21
y . REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEMS-OPERATING
. LIMITING CONDITION FOR OPERATION 3.1.3.2 The shutdown and control rod position indication system and the demand position indication system shall be OPERABLE and capable of determining the control rod positions within i 12 steps.
APPLICABILITY: MODES 1 and 2.
. ACTION:
- a. With a maximum of one rod position indicator per bank inoperable
~ ei ther:
- 1. Determine the position of the non-indicating rod (s) indirectly by the movable incore detectors at least once per 8 hours and immediate17 after any motion of the non-indicating rod which exceeds 24 steps in one direction since the last determination of the rod's position, or
- 2. Reduce THERMAL POWER TO less than 50% of RATED THERMAL POWER within 8 hours.
.b. With a maximum of one demand position indicator per bank inoperab.le either:
- 1. Verify that all rod position indicators for the affected bank are OPERA 8LE and that the most withdrawn rod and the least witndrawn red of the bank are within a maximum of 12 steps of each other at least once per 8 hours, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER .
within 8 hours. , SURVEILLANCE REQUIREMENT 3 4.1.3.2 Each rod position indicator shall be determined to be OPERABLE by verifying that the demand position indication system and the rod position indication system agree within 12 s.teps at least once per 12 hours exceot during time intervals wnen the Rod Position Deviation Monitor is incoerable, then compare the demand position indication system and the rod position indication system at least once per 4 hours. SEQUOYAH - UNIT 2 3/4 1-17 i
7-3
., r U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: SEQUOYAH 1&2 REACTOR TYPE: PWR-WEC4 DATE ADMINISTERED: 86/05/26 EXAMINER: D.J. NELSON APPLICANT: _ _ _ _ I _ _e__b______e_;d______
INSTRUCTIONS TO APPLICANT: Use separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the-question. The passing
~
grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 30 0 53
___1_0___ _'5 [_1__ ___________ ________ S. THEORY OF NUCLEAR POWER PLANT OPERATIONr FLUIDS, AND g THERMODYNAMICS _$[1 __ _ 1 ___________ ________ 6. PLANT SYSTEMS DESIGNr CONTROL, AND INSTRUMENTATION _ i __ _ 1 ___________ ________ 7. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND RADIOLOGICAL 3 ,, CONTROL
-2 7 , g 0- 23.40 l
- 8. ADMINISTRATIVE PROCEDURESr CONDITIONS, AND LIMITATIONS tl4 %
+17 .3 G- 100.00 TOTALS FINAL GRADE _________________%
All work done on this examination is my own. I have neither Siven nor received aid. 5PPL5CEUTI5~555U5TUf(E~~~~~~~~~~~~~~
]
r-S. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 2 QUESTION 5.01 (1.00) Which of the.following will cause the fuel temperature coefficient (pcm/ degree) to become less negative?
- a. fuel temperature increase
- b. baron concentration cecrease
- c. control rod insertion (at constant power) de increase in the ratio of Pu-240 to U-238 OUESTION 5.02 (1.00)
Select-the statement about single speed, motor driven, centrifugal pumps that is correct.
- a. Upon throttlin3 oPen the discharge valve to increase flow, discharge pressure decreases and therefore motor amps decreases.
- b. Upon throttling open the discharge valve to increase flow, net positive suction head required increases and differential pressure across the pump decreases.
- c. Upon throttling open the discharge valver flow increases, total developed head increases and net positive suction head available decreases.
- d. Pump cavitation can be reduced by throttling open the discharge valve thereby reducing total developed head.
QUESTION 5.03 (1.00) Of the following, which must the main condenser remove the most heat from to condense? (assume steam is of equal quality)
- a. one pound of steam ,t 0 psia,
- b. one pound of steam at 300 psia.
- c. two pounds of steam at 600 psia.
- d. two pounds of steam at 1200 psia.
l (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) l [
** K84 UNITED STATES #' 'g NUCLEAR REQULATORY COMMISSIUN "
-[ f, 'n AEGION il ' f a 101 MARIETTA STREET, N.W., SUITE 2000
# t o ATLANTA, GEORGIA 30323 %...../
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3 GUESTION 5.04 (1.00)
Which expression below describes the heat flux hot channel factor F9( )?
- a. Maximum fuel assemblybatheightZ/AvgbatheightZ
- b. Maximum fuel assemblybatheight/Avgbincore
- c. Average fuel assemblyb/ Maximum.batheightZ
- d. AveragebatheightZ/Avgbin core QUESTION 5.05 (1.00)
Which af the following would cause an inadvertant dilution accident? a) Overfilling a S/G while in hot standby. b) A Regenerative heat exchanger leak, c) Valving in a demineralizer that was not saturated. d) A VCT Lo-Lo level resulting in the RWST being used for charging. QUESTION 5.06 (1.00) Attached Figure 4 220 shows a power history and four possible samarium traces (reactivity vs time). Select (a, b, c, or d) the curve that correctly displays the expected samarium transient- for the given power history. QUESTION 5.07 (1.00) Attached Figure # 219 shows a power history and four possible xenon traces (reactivity ~vs time). Select (ar b, c, or d) the curve that correctly displays the expected xenon transient for the given power history. (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
f. 500-gn < . . . . . . . . . . .......... ee 30 3o 40 80 he *le g . . . . . . . . . . . 9- o -
- o. - .
to i. 4. iO i, 40
... . . _ . . . . - - - - - - - - - - - - - - - --- - - - - - - ~ ~ ~ ~ ~ ~ " ' ~ ' ~ ~ ~ ~ ~ ' ' * " ~
b, I O. lO t'a It de e'O [0 ip g - C- o O. .
, , , i . -
10 30 30 40 50 60 8 d. r O Q.
, , , i i i =
10 10 30 gg SO 44 70 TIME CORYS) F l GuRs ** 170
r i 6 WD' e . M t's O 60 fo de tio de g. q' d l A _ 1Le 90 60 90 les tse Its
- b. = - _ . - . - _ -
w to ** So leo ste age 1 C, i. . . _' . . . _ l 5 5 5 5 5 3 80 44 to 30 soo sa.e age l 1 d, g j _ . - - - - -
. . 7 ao 9e so 80 sco uo 140 TIME ( H o uRS) i l
l FicuRE*119
i
~5. -THEORY OF NUCLEAR POW'ER' PLANT OPERATION, FLUIDS, AND- PAGE 4 ~0UESTION 5.08 (3.00)
At 30% power a reactor coolant pump -trips. _With control rods.in manual ~and all other systems in automatic and no operator / protective actions.occurr
-indicate'the effects on the following at.the end of the transient:
(increases, decreasesi or remains the same) AFFECTED LOOP + UNAFFECTED LOOP
- a. steam generator level +
+
'b. steam flow +
+
c.-delta T- + QUESTION 5.09' (1.50) For the changes listed below (treat each one independently) indicate whether the moderator temperature coefficient will become MORE NEGATIVEr LESS' NEGATIVE or have N0'EFFECT. (Assume all other parameters.are constant) a) Neutron flux peak shifts radially outward to..the edge of the core. b) Boron concentration increases 100 ppm while core is at-HOL. c) All rods in instead of all rods out. QUESTION 5.10' (1.00)1 Indicate TRUE or FALSE for the following statements.concerning the effect
.that delayed neutrons..have on reactivity:
a.. Because delayed neutrons are born at-lower energies than prompt ~ neutrons, they are less likely to leak out of the core resulting in a positive effect.
- b. Delayed neutrons-are born at an average energy incapable of causing fast fission of U-238 creating a negative effect.
QUESTION 5.'11 (1.00) What are the two primary factors that provide the driving mechanism for Natural Circulation' flow?
-(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
R
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 5 QUESTION 5.12 (2.00)
Unit 1has just resterted following a refueling outage while Unit t is near EOL. Answer the following regarding the differences in plant response between the two units (explain your answers): A ssw-.e 7.h i- m% l a) At a steady power level of 10EE(-8) amps during a startup, equal reactivity additions are made ( a p p r o::i m a t e l y 100 pcm). Which Unit will have the higher steady state startup rate? b) At 50% power, a control rod (100 pcm) drops. Assuming NO RUNBACK or OPERATOR ACTION, which Unit will have the louer steady state Tavs? l GUESTION 5.13 (1.50) During the performance of an emergency baration while at power, how and why are the following parameters affected? (assume no control rod movement)
- a. subcooling
- b. over fewrA -- differential temperature setpoint
- c. control rod worth OUESTION 5.14 (1.00)
Explain how decreasing RCS flow (at constant power) will result in decreasing DNBR. QUESTION 5.15 ( .50) If the equilibrium count rate in a suberitical reactor TRIPLES due to a reactivity addition, what happens to the margin to criticality (direction and magnitude)? DUESTION 5.16 (3.00) What are THREE parameters AND their trends which are indications that Matural Circulation in the RCS is established? (numerical values not req'd) (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) 1
t
.5. THEORY OF NUCLEAR POWER PL. ANT OPERATIONr' FLUIDS, AND PAGE 6 QUESTION 5.17 (1.00)
Aux. Feed flow is more critical on a small LOCA than on a large LOCA. Why is this .true? QUESTION 5.10 (1.00) What are.the two reasons for shifting the SI mode from cold les recirculation to-hot le3 recirculation approximately 24 hours after a LOCA? QUESTION 5.19 (2.00) A roo drops and sticks at the core mid position from full power conditions with all rods'out. A Reactor Trip does not occur. If this condition were to persist for an extended period _of time (well beyond T/S limits), _ what will be the. effect on the Excore Axial-Offset of the Power _ Range NI for the quadrant in.which the dropped / stuck rod occurs. Include a discussion of xenon effects and a definition of axial offset. QUESTION ~ 5.20 (1.50) Use the attached curvesr- labeled S-24, to explain the following questions. Assume that'all systems are in automatic control and_that no operator action is taken. a). Why does S/G level rise at point 4? b) Why is there no outward rod motion after point 9e even though Tavs is less than Tref?
-c) -Why.does steam flow gradva'lly decrease at point 8?
QUESTION 5.21 (1.00) . Arrange the following types of radiation in order of penetrating power from LOW to HIGH:
- 1. Beta
- 2. Gamma
- 3. Neutron
- 4. Alpha
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
- . ,~ .. _ .-. ,. ,_ . _ . _ _
1 S-24 cf33 PRE 55UR12ER PRES 5URE (P5tG) S ANK + D" ROD 5 (57EP5) NUCLE AR POWE R (%) GENE R ATOR LOAD (WW) l l l Ii i i!!
!I -
I; < (137 3 l { . i a i
!l 3 -
! l !
e iI: i I N 3 2 ll-i ! l z ! ! 2 I
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i - i i i i i o 2235 3000 o 230 0 100 12o o 1200 1s00 TR E F - Tgyc - (*F) PRES $URIZER LEVEL (%) CH ARGING FLOW (GPM) STE AM DUMP (% DEM AND)
- 4 ll !
B l l
! k 3 54 \ /
- - , w L
i i Y 1 W l i 3 2 l 2 k 1 _
\ \ \
\ l \
l ' ' \ ; L ~-
\i i \
4 I 1
--(8A \ t ! !
e S.
, ii, l
-O, , ,,,,,,
3 , 53o Sn s3o o too o zoo o too t 1 ALL STE AM DUMP V ALVES F AILED OPEN - 100% POWER i S-24
S-24 06s3 Wg ( g /H a M t o') WF l s /H R X t o6 ) 5/G L EVE L (% N.R -) P STE AM (P$1C) I l i
,/ I 3 'h, - --s i I (v : i ,
- f. ,
; ' ~l i l l ij l
l us W l\ 'f ! e- 1i ! i , 3 2-~ ' z l t t l I 2 ' i l lI i
'! , l l 1 i
! i I i !
l! ! I i I J t(10)__
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e i, D,&,) i
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. ., o e o ion o icos uno I
( l l l l l i i 5 h l l l I l l l ALL STE AM DUMP VALVES F AILED OPEN - 100% POWER S-24 a
9 a E 5,- THEORY OFLNUCLEAR POWER PLANT OPERA 1IONr-FLUIDS, AND PAGE 7 GOESTION 5.22 .(2.00)
+
Given the following, calculate the required boron change to increase reactor power. from 75% to 100% while maintaining constant rod position. Moderator temp. coeff. -15 pcm/ degree F Do'ppler-only power coeff'. -12 pcm/% power Void reactivity change -25 pcm
' Xenon change -50 pcm '
Baron coeff. -9 pcm/ ppm - (xx*** END OF CATEGORY-05 *****) l r
n - 1 s O g. ;_r _A
;.l m,i; -. . -
/
c
- 6. PLANT' SYSTEMS DESIGNr CONTROL, AND INSTRUMENTATION PAGE .8 (1.00)
Qu'ESTIdh'{.01 Which statement below regarding the RCP shaft seals is.NOT correct? (1.0) a k. Only" tl anr' 12 seals are designed to withstand full system
. pressure Rb . Leakoff f rom 42 seal-is used to maintain the-level in the standpipe used to supply cooling.' water to 43~ seal
. Q.
An individual'il seal bypass line.cannot be isolated without isolbling the other.RCP seal bypass lines as well
- d. The 41 seal isla 'floatins' face seal vice a 'rubbins' face seal like the #2 and #3 seals.
QUESTION ~6.02 (1.00) According to 10CFR50.46, which of the followins is-NOT.~a design criteria of1the Emergency Core Coolins System subsystems.
- a. The calculat'ed peak centerline' temperature.shall not exceed
. . 2000.de,grees F.
- b. -The maximum claddins,oxidat.on shall not exceed 17% of the total cladding thickness.
~
c .- Thel calculated total amount of' hydrogen generated from the
~
cladding. reaction with water shall not exceed 1% of the amount that would be senerated if all the cladding around the fuel reacted..
- d. Calcola,ted changes in core geometry shall be such that the core remains in a coolable configuration.
=
(*****' CATEGORY 06 CONTINUED ON NEXT PAGE *****)
=
3: A
't 4
g i b% ) \- i I
V
'N
- 6. PLANT SYSTEMS DESIGN, CONTROL, INSTRUMENTATION' PAGE 9
__________________________________"AND _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
-GUESTION 6.03 f ( 1. 00 ) -
'The power ranse detector c'urrent comparator circuit compares which of the following?
- a. Each. individual. power ranse total power signal ~to the averas'etof-all power (ranse total power signals. . ',
s. g
- b. Each upper power ranse-detector: signal to its respective-lower power ranse signal.
- c. Each individual, upper (lower) power ranse detector' signal to the
,i averase of the upper (lower) power ranse detector signals.
- d. The average upper _ power ranse detector signal to the average lower power ranse detector signal.
j GUESTION 6.04 (1 00) i
- With three reactor ~ coolant pumps operatins indicate if the. flow in the
! S i ven loop sesciync will be in the NORMAL or REVERSED direction in_the loop with.the.non-operatins pump.
- a. T-h RTD manifold
- b. T-c RTD manifold OUESTION- 6.05 -(1.00)
.Besides the overspeed shutdownr which of the followig/ diesel-engine /
generator shutdowns-is enabled durins an emergency Sta'rt of the diesel? j a. Voltase restraint overcurrent relay, (51V).- 4
- b. Generator differential relayr (87).
- c. Phase balance relayr (46).
d.- Low lube oil pressure. l (*****. CATEGORY 06 CONTINUED ON NEXT PAGE *****) i I l
'l '
~~
pp -- 1 e 1'
. .4 e
' .}
- 15. ,PLANTSYSTEMSDESIGNrfCONTROLeANDINSTRUMENTATION' PAGE 10
-------------------------j---------------------- -----
, t e
.00ESTION 6.06 (1J00)
. .. . w . .
Which statement c~oncerji~ng th<eyRod Control System is CORRECT?-
... a v .
- a. The power,; cabinet p'rsvides AC power pulses t'o drive the control rod drive mechanisp.T):
b.= The reactor (control unit generates a rod speed andfdirection s"ignal in response to three ERROR' signals.
- c. T Nbi'ne' imp.ulse pressure provides si'anals to the rate.comparator, summing l unit.and the variable gain unit.
- d. Rod power is supplied bp two motor generator sets with a 260VDC .
nutput thf.ough.an isolation transformer. QUESTION a6.07 , .(1.00) Nhich of the below, features enhances.the operation.of. the ice condenser and containment spray for heat removal?
- a. Containment des,ign, such'that the delta P between upper and lower
, containment drides the air circulation.
s 0
- b. Ventilation coolers and recirculation fans are used to mix the air and pro, vide additional cooling.
- c. Air return fans. provide flow to return the air from the upper
- Icontainment L to the lower containment. l l
- d. Pressure oper<ated doces- open to allow upper containment air to l flow through to the lower containment.
i (***** CATEGORY 06 CONTINUED ON NEXT PAGE f
*****).
d'
- s. 4 4'
.1 t
5-
)
rY .L..
- 6. PLANT SYSTEMS DESIGN,. CONTROL, AND INSTRUMENTATION PAGE 11
~0UESTION -6.08 (2.50)
For the followins, how will-the indication respond (higher, lower, as'is)- to the given failure? a)-RTD open. circuit in detector b) Intermediate ranse compensation voltage fails hish withfreactor power at,100%.
~
c) Source ranse pulse height descriminator settin3 fails low with-reactor. power..in the source ranse. Ed) Thermocouple junction opens e) Steam 1 flow pressure compensation to steam flow detector fails high' QUESTION -6.09 - ( . 50) TRUE or FALSE?. After trippins a bistable in a 2/4 logic system, one'of three remaining-signals reaching the bit' able .setpoint .will cause a trip, even-though the logic SYSTEM remains as a'2/4 system. QUESTION 6.10' (3 00)
~ ~
Match the type (s) of' rod motion that is blocked'with the sisnal that causes the rod block! (More than one response mayfbe required for full credit)
-SIGNAL BLOCKED ROD. MOTION
- a. OP delta T 1.' Automatic Withdrawal
- b. OT delta T 2. Automatic Insertion
- c. Power Ranse at 103%' 3.-Manual Withdrawal
- d. Inter. Range at 20% equiv.. 4. Manual Insertion
- e. Control Bank D > 220 steps' 5. No Blocked Mo. tion f.-Urgent Failure in Power Cabinet j s. P imp < 15%
- h. Tave vs. Tref < 1.5 degrees F t
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) I J J 1
- ~ . - . . , . - , , ,,,--,-.,...n . -r., , . . . - . , . . . , , ..-,....v . , . . , . - - . . , , . . , ,
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 12 GUESTION 6.11 (1.00)
Fill in the blanks below to correctly complete the statement regarding the Motor Driven Auxiliary Feedwater Pump level control valves: These valves are ___________ operated and will fail _______ on a loss of air. They are normally set to maintain a S/G level of _____% and if pressure downstream drops to < ______ psis, the valves will close automatically. QUESTION 6.12 42.004
- 1. 9 For the following components, indicate whether they will receive an'0 PEN, CLOSE, or N0 signal as a result of a safety injection (with Phase 'A')
initiation signal.
- a. Control room supply ducts
- b. Main feed bypass valves
- c. SI accumulator discharge isolation valves
- d. Normal charging header isolation valves
- e. Main steam isolation valves r . -- e r to qI ,1 n. ;- n:tt. _1 _
pg.f .M 3 Seal water return isolation valve
- h. Component cooling isolation valve from RHR system
- i. Component cooling isolation from letdown heat exchanger
- j. Steam supply valves to +urbine-driven feed pum p Lla,Gh) A b Sed pnp.
QUESTION 6.13 (2.00) The following failures occur causing a subsequent reactor trip.
- What protection signal would cause the trip? Assume the reactor is initially at 100% power and steady state conditions, all systems in automatic and no operator action. Treat each case independently, a) CVCS flow rate drops to a minimum of 30 gpm.
b) A narrow range (controlling) cold leg RTD fails high. (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) 1
~. - _ . - - . - - ~
r . .i~ r
' 6 .s PLANT SYSTEMS DESIGN, . CONTROL,'AND. INSTRUMENTATION PAGE 13
,OUESTION 6.14 - (1.00)
~
i Arranse the followins in the correct sequence for rod-withdrawal'(one step)..
- a. Lift coil 0FF
- b. Stationary gripper coil DN-
- c. Hoveable gripper coil DN 5- d. Moveable gripper coil 0FF
- e. Lif.t coil ON
- f. Stationary gripper coil OFF
' 00ESTION 6.15 (2.50)
! List ALL the protection, alarm and control functions provided by the PZR. pressure instruments as pressure decreases from 2350 psis. (Include
-the applicable setpoints)
- GUESTION- 6.16 (1.50) l List the 5 Auto-start signals,for the Turbine driven AFW pump.
QUESTION 6 17 (2.00)- List four conditions that will senerate a " Computer Alarm Rod Dev and Seq HIS-.PWR. Range Tilts' alarm.
- ~0UESTION 6.18' (1.00)
' ~
'The pressurizer has a resistance temperature detector (RTD) in the. STEAM
; space that.is normally used to indicate the PZR saturation temperature. ,
Assumin3.this RTD'is operable, during what plant evolution is the RTO in i the PZR WATER space utilized, and why i sn't the PZR STEAM RTD used? 1 (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) a 0 e r,- ,,%, w-- -.,,-,.w y, - - . - -y ,- --r-.-,, r_, m n., , , e -y3 y,__ ,.c ,-sy_,,- , .,,yy,.y, ,,,,,--,,--,--1-_-.,.1m , , . , , . . . , .
6o PLANT SYSTEMS DESIGNr. CONTROL'r AND INSTRUMENTATION PAGE 14
~
QUESTION 6.19 (1.00) While operating at 92% power, the 13 hea'ter drain tank level Lgoes high enough to cause.the water in the tank to.begin dumping to the condenser. According to SDI's 5.1 &_6.1r what effect-will this have on the turbine? QUESTION 6.20 (1.00)
- a. When the RHR system is controlling RCS solid plant conditions, from where does the water leave the RCS?
b.RIf the control _ valve which separates RHR and CVCS fails snute what 3 relief valves would limit-RCS pressure? (redundant reliefs count as one response). QUESTION 6.21- (1.00) Most RCS pressure control / protection si3nals are generated by the PZR pressure instruments. What control or protection. signal is generated by
'the pressure instrument on a T-hot les?
OUESTION 6.22 (1.00) What is the reason for the interlocks on the CVCS letdown valves and orifice isolation valves? (***** END OF CATEGORY 06 *****) E i I L f a
- 7. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND .PAGE 15
~~~~R 5656EU55C5[~C6NTR5E-------- "-~~~~~--------
GUESTION E7.01 (1.00)- Which of the followins is NOT an immediate operator action for. a. Safety Injection as1 stated in E-0?
- a. Verify Containment-' Isolation.
b.1 Check Tavs.
- c. Verify AFW status.
d.' Verify Steam Dumps actuated. QUESTION 7.02 (1.00)
~
It is necessary to. reduce the critical boron, concentration by 200 ppm priorf t o pulling the control banks.. Prior to the dilution, the source range' instruments read 30 and-37 cps.- After.reducins. boron' concentration 100 ppm the same instruments read.62 and 75 cps.- Which of1the followins is the proper operator action'in accordance with GOE-2?
- a. Stop the dilution and borate _back t'o the original count rate.
^
b.-Stop the dilution and evaluate the'situati~on.
- c. Continue the dilution and continuously monitor the count rate.
- d. Continue ~the dilution and recalculate theDECC.
- e. Continue the dilution as nothing abnormal has occurred.
QUESTION 7.03 (1.00) During' normal CVCS. operation, whi~ch of the followins is an abnormal condition.and would require operator action to correct?
- a. .VCT pressure is 15 psis,
- b. The temperature of the fluid leavins the-letdown heat-exchangers is 127 F.
- c. The RCP seal injection water tempeature is 120 F and flow to the.
seals is.O spm/ pump.
- d. RCP seal differential pressure is 300 psid.
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
+
- 7. . PROCEDURES -l NORMAL, ABNORMAL, EMERGENCYLAND PAGE' 16:
~~~~ - -----~~~~~~~~~~~~~~~~~~-
RA656L665EAL E5NTR5t
/GUESTION' 7.04 ,(1.00)
Which1of the following statements concernins .the procedure for a
-dropped RCCA is correct?.
-a. Upon starting recovery.of the dropped RCCAr_an URGENT FAILURE alarm will occur because the lift coils for the other rods t in the group have been disconnected.
- b. The delta' flux target band.is not applicab'le during a dropped PCCA malfunction and recovery.
- c. If~two or more.RCCA's have dropped, manually trip.the reactor and proceed in_accordance with EP-1.00.
- d. Recovery from a dropped RCCA will be facilitated if Tavs is higher than. Tref prior to commencing withdrawal of the dropped RCCA.
.0UESTION 7.05 (1.00)
.During an inadvertent dilution accident while at 100% power, which of_the following will be the most probable cause of a reactor trip?
- a. Pressurizer low pressure.
b.-Over-temperature delta T.
- c. Over power delta T.
- d. Power ranse monitor positive-rate.
GUESTION 7.06 (1.00) If the reactor trip breakers are closed and the steam senerators_are under nitrosen pressorer the: nitrogen' ' pressure must be vented of f the steam i senerators prior to opening the MSIV's. Why'must this be done? (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) l l s
v v Y
- 7. . PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE' 17
~~~~~~~~~~~~~~~~~~~~~~~~ ~~~~ REUEULUUYUE[~EUNTRUL QUESTION ~7.07- (l'00) lAccording to a note in GOI-2't.what condition ~must be met prior-to exceedins 600- RPM on the main turbine?
Ja. ' Main Feedwat'er Regulating valves are--to be in automatic,
- b. Tavs is to be at the no-load'value..
- c. The low pressure turbine inlet metal-temperature must be steater t 400 degrees F.
- d. Steam dumps must be in Tavs mode.
QUESTION' 7.00 (1.50)
' Answer the folowing questions regarding EOPJusase TRUE or FALSE;.
a). If a: Function Restoration Procedure (FRP) is entered due to an ORANGE Critical Safety Function (CSF) condition, and a-HIGHER-priority ORANGE' condition is encountered, the original FRP must be completed prior to proceeding to the newly-identified FRP.
- b) Unless specified, a task need not be fully completed before proceeding to a subsequent step'as lons as that task is progressing-satisfactorily-c) If a procedure transition occurs, any tasks still in progress from the procedure which was in.effect need not be completed.
QUESTION 7.09 (1.50)
- a. Give the Sequoyah normal quarterly whole body dose limits for the
- following:
- 1. TVA personnel
- 2. Non-TVA personnel (without present quarterly records)
- b. Whose consent is' required before the emergency exposv e guidelines can be used?
(***** CATEGORY 07 CONTINUED-ON NEXT PAGE *****)
E7 . PROCEDURES - NORMAL, GNORMAL, EMERGENCY AND PAGE 18
- ------------~~----~~~~-- ~~~~R 5655L55fCAL C5sTR5t 00ESTION' 7.10 (1.00) a)' What l'evels are required in the containment sump and the RWST to transfer ECCS suction to the containment sump?
b) What actions are required if swapover has not been completed and RWST level reaches 0%? QUESTION 7.11 (1.00) What are the SI Re-initiation criteria of ES-0.2rz'SI Terminiation'? (Include pararneters- associated with adverse containment conditions) GUESTION 7.12 (1.00) Indicate what increase in count rate is required for the combination of nuclear instruments listed below, such that fuel shuffling operations would have t'o be immediately stopped! (Exclude Anticipated changes due to-detector or source movements) a) Increase on ANY Nuclear Channel b) Increase on ALL Nuclear Channels-QUESTION 7.13 (2.00) LIST the THIRTEEN immediate. actions to be taken for a Saftey Injection, in accordance with Emergency'Procedurer E-0. (Substeps are not required) 00ESTION 7.14 (1.50) 49 t?- O Accordir.s to "ImmediateActionsandDiagnostics'h . if containment pressure is greater than 2.81 psis, what THREE SPECIFIC conditions must be verified? k8 Ma eh Nc e.' chi D W-(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
h y
-7. PROCEDURES - NORMAL,~ADNORMAL, EMERGENCY-AND P' AGE':19
~~~~R5555L5GiB5t C5sTR5L QUESTION 7.15 (3.00) a) What are-FOUR methods that can_be used for identifying the faulted steam seneratorr during a steam senerator tube rupture accidente in accordance with EDI-3? (2.0) b) What.is the purpose of trippins the RCPs during a SGTR when the f
. appropriate. trip criteria are reached? (1.0)
GUESTION 7.16 (1.00) What are ALL the Immediate Operator Actions for a continuous insertion of a control rod bank? -c
~
00ESTION 7.17 (2.00)
- a. :According to GOI-3A, "Hydrazine must not be added to the coolant durins any phase of plant cooldown or shutdown, if
.the primary conlant system is to be opened.'" Explain WHY this precaution is necessary,
- b. Near the completion of plant cooldown (~140 F), Hydrogen Peroxide (H202) .is added to the RCS and circulated with Reactor Coolant Pumps. WHAT does,-this action accomplish and WHY is it necessary?
GUESTION 7.18 L(1.00) , A NOTE in GOI-2r " Plant Startup from Hot Standby to Minimum Load," states that if control rods were withdrawn 5 steps durin3 heatup, the control' rods must be fully inserted prior to withdrawing rods. a) Why are the rods withdrawn 5. steps durins heatup? b) Why must they be inserted prior to withdrawal? (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 20
- -----------~~-----~~~~~- ~~~~R A5i5L55iCAt C5UTR5L OUESTION 7.19 qVdh h (2,00)
A safety injection 1 signal is to be considered non-spurious unless Ca. specific cotiditions are exhibited. What are these conditions? [b. The 'SI Termination Criteria' for Loss of Primary Coolant and Loss of ( Secor dary Coolant have major differences. for these differences? What is the maaor REASON
)
GUESTION 7.20 (2.00) Sequoyah Procedure, GOI-2 states the following precautian!
'All shutdown banks must be at the fully withdrawn position whenever positive reactivity is being inserted by boron or Xenon concentration changes, reactor coolant temperature changes, or motion of control banks.'
State the TWO different plant conditions that are exceptions to this precaution. QUESTION 7.21 (1.00) GP S s What are the TWO guidelines from EDI-1 Appendix D hat indicate inadequete core cooling exists? QUESTION 7.22 (1.50) An irradiated fuel assembly is being moved from tne reactor vessel to the opender when it drops to the bottom of the refueling canal? A. What are the SRO's immediate actions if radiation monitors indicate increasir3 levels, in accordance with AOI-2?? B. What is the source (type) of the radiation activity released? (***** END OF CATEGORY 07 *****)
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- 8. AND LIMITATIONS PAGE 21
-___ ADMINISTRATIVE PROCEDURES,' CONDITIONS,______________________________________________________-
f00ESTION '8.01 (1.00) Fill inlthe blank with~one of'the following TS terms!
"A.________ shall be the. qualitative assessment of channel . behavior during operat' ion by observation. . This determination shall includer where possible, comparison of the channel indication and/or status with other indications;and/or -status derived from independent instrument channels measuring the same parameter".
- a. Channel Calibration b.' Channel. Check
- c. Channel Functional Test d.- Losic System Functional Test GUESTION 8.02 (1 00)
In accordance with 10 CFR 55, 'if a licensee has not been actively. Performins the functions of an operator or senior operator for a period of____(1) ___ months,-or longer, he sh~all, prior to resuming activities licensed pursuant to this parte demonstrate to the Commission that his knowledge and understandins of facility oper-ation and administration are satisfactory.' FILL IN THE BLANK WITH ONE OF THE FOLLOWING TIMEST
- a. 4
- b. 6
- c. 12
- d. 24
(***** CATEGORY 00 CONTINUED ON NEXT PAGE-*****)
a y, . 1
- 8 ~. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS ~ PAGE 22 QUESTION ?.03 (1.00)
Which of the followins_MAY proceed given that a Technical Specification Action Statement has been entered requiring that you1' suspend all CORE
. ALTERATIONS *?
- a. Removing a neutron source from the core,
- b. Using the bridge in the core is allowed,.Provided that the low load limit.is-jumpered out.
- c. Control, rods and burnable poison ~ rods may be shuffled as lons as K-effective is less than or equal to .95.
- d. Completion of the movement of a component to a safe conservative position within the ex pressure vessel.
QUESTION 8 .~ 0 4 (1.00) Per GOI-6,-Apparatus Operation, which of~the followins is the proper method for VERIFYING.the_ position of a locked (padlocked) valve?
- a. Attempt to move the valve handwheel or-operator in the OPEN direction.
b.- Attempt to move'the valve handwheel or operator in the CLOSED direction.
- c. Attempt ~to move the valve handwheel-or operator in the direction SPECIFIED as the correct position.
- d. DO NOT attempt:ta move'the valve handwheel or operator - Verify l proper valve position by direct observation of.the stem'and/or local position indicators.
(***** CATEGORY 00 CONTINUED ON NEXT PAGE'*****)
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONSr AND LIMITATIONS PAGE 23 QUESTION 8.05 (1.00)
Unit 1 is in COL'O SHUTDOWN with T' ave being maintained stable at 190 degrees F by RHR , The following equipment is INOP:
- Centr'ifugal Charging Pump 1D-B x (1 hr)
- Reciprocating Charging Pump 1B * (1 hr)
- DG 1A-A * (1 hr)
- There i's no estimate of repair time.
The Shift is directed to recommence the plant cooldown to a RCS T avs of 130 degrees F. SHUTDOWN MARGIN calculations indicate compliance with the TS LCD for " SHUTDOWN MARGIN - T-ave < ar equal to 200 degrees F' throughout the full range of the anticipated cooldown.
~
Which of the following actions most correctly detail the allowances and/or limitations imposed by the Technical Specifications in this instance? NOTE: APPLICABLE TSs ARE ENCLOSED FOR REFERENCE.
- a. Plant Cooldown may. recommence; OPERATIONAL MODE 6 may be entered with no restrictions on plant operations.
- b. Plant cooldown may recommence; OPERATIONAL MODE 6 may be entered BUT CORE ALTERATIONS are precluded.
- c. Plant cooldown may recommence; OPERATIONAL MODE 6 may not be entered.
- d. Plant cooldown is prohibited AND heatop to 200 degrees F is required.
- e. Plant cooldown is prohibited.
(mmr** CATEGORY 00 CONTINUED ON NEXT PAGE *****)
O. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 24 QUESTION 8.06 (1.00) Unit 1 is at 90% power with no INOP equipment. Ten minutes into the shift, two (2) level Instrument Channels associated with the "RWST Level - Low" function of ESFAS Instrumentation fail their CHANNEL FUMNIONAL TESTS. There is no estimate of repair time. Which of the following actions most correctly detail the allowances and/or limitations imposed by the Technical Specifications in this instance? NOTE: APPLICADLE TSs ARE ENCLOSED FOR REFERENCE.
- a. Operation may proceed provided the inoperable channels are restored to OPERADLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- b. Operation may proceed provided the inoperable channels are placed in the bypassed condition and the other Channels are demonstrated OPERADLE within i hour.
- c. Within one hour action shall be initiated to place the unit in at least HOT STANDBY within the next 6 hours; and at least HOT SHUTDOWN within the following 6 hours.
- d. Take Actions detailed by choice c. AND place the plant in COLD SiluTDOWN within the next 20 hours.
- e. Take Actions detailed by choice c. AND place the plant .n COLD SHUTDOWN within the subsequent 24 hours.
DUESTION 0.07 (1.50) Indicate whether alarms on the vertical boards listed below are the responsibility of the Unit 1 Dalance of Plant (BOP) operator, Unit 2 DOP or both DOPs. a) 0-M-27A (ERCW) b) 0-M-20A (Cooling Tower Pump Controls) c) 0-M-25 (Miteorological/ Environs Monitoring) (***** CATECORY 00 CONTINUED ON NEXT PAGE *****) l
I
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 25 GUESTION 8.00 (1.00)
Fill in the blanks to complete the followin3 statement regarding temporary alterations: The Shift Engineer. reviews requests to perform temporary alterations for completeness and coerectness, verifies it has been reviewed by _____ and approved by _____ and that a/an _____ is attached as required. He also performs _____ for emergency conditions requiring temporary alterations. QUESTION 0.09 (1.00) Fill in the blanks in the following statement regarding clearances: When tagging a breaker open that gives an alarm, if it is to be open for greater than _____, the electricians shall lift the wires to the annunciator and initial the clearance sheet. The ASE will the lggf3 with a _____. When the clearance is released the elec ~~ricians will connect the annunciator wires and _____ it on the return to normal section of the clearance sheet. QUESTION 0.10 (1.50) List the FIVE bases for the minimum temperature for criticality limit of the Technical Specifications,i.e. what does this limitation ensure? GUESTION B.11 (1.00) What ACTION (S) must IMMEDIATELY be initiated (per the TSs) if SHUTDOWN MARGIN decreases to less than 1.0% delta k/k in Mode 5? Be Specific. (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) l i
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONSr AND LIMITATIONS PAGE 26 GUESTION 8.12 (2.00)
STATE which Emergency Classification is appropriate for the following definitions.
- a. Events are in progress or have occurred which involve actual or potential substantial degradation of the level of safety of the plant.
- b. Events are in progress or have occurred which could develop intor or be indicative of, more serious conditions which are not yet fully realized, c .- Events are in progress or have occurred which involve actual or imminent substantial core failure with the poten-tial for loss of containment integrity.
- d. Events are in progress or have occurred which involve an actual or likely major failure of plant functions needed for protection of the public.
QUESTION 0.13 ( .50) Given the following plant conditions!
- K effective = 94
- % RATED THERMAL POWER = 0%
- AVERAGE COOLANT TEMPERATURE = 160 degrees F
- RPV head closure bolts less than fully tensioned
- Fuel in the Rx Vessel State the OPERATIONAL MODE of the plant as described above.
QUESTION 8.14 (1.00) Prior to entry into the lower Containment or the Annulus the position of the incore flux detectors shall be verified. What are the two (2) acceptable positions per the Access to Containment procedurer SONP AI-10? QUESTION B.15 ( .50) One stationary Firewatch position shall be maintained at ALL times per OSLA 73. What is the location of this firewatch? Provide D1dg. elevation and the Major equipment p r o::i m i t y . (***** CATEGORY 00 CONTINUED ON NEXT PAGE *****)
r 8.~ ADMINISTRATIVE PROCEDURES, CONDITIONSr AND LIMITATIONS- PAGE 27
-QUESTION 8.16 -(2.00)
What.are the.four conditions that Tech Specs cay must be met to ensure the-
-Nuclear Enthalpy Rise Hot Channel Factor is maintained within limits-during periods ~between in-core surveillances?
OUESTION 0.17 (2.50) Define IDENTIFIED LEAKAGE (as per Section 1.0 of the TSs).
. QUESTION 8.18 (1.50).
-Answer the following with regard to the Fuse' Control Procedurer SONP
- AI-16*
- a. Fill in the blank:
A blown control. circuit fuse may be replaced _____ time (s) with
'the correct fuse as identified adjacent to the fuse' block. ( 0.5)
- b. Assume the fuse replacement limit ~of.part a. has been met and the replacement _ fuse (s) have also blown. Who must be notified AND What must.be done~before further fuse replacement? (1.0)
QUESTION- 8.19 (1.00) Per GOI-6r Apparatus Operation:
- a. Why'are Manual-Operated Valves (excepting throttle valves) always backseated? (0.5)
- b. What specific valve damage could occur if excessive force is used during backseating? (0.5)
(***** CATEGORY 08 CONTINUED ON NEXT PAGE ***rx)
r o. 8.- ADMINIST'RATIVE:PROCEDURESr CONDITIONS, AND LIMITATIONS PAGE ' 2Er QUESTION 8.20 -(1.00) Concider'the proper procedure for racking in a 6.9 kV breaker-per ! GOI-6,_ Apparatus. Operation: l
.a. -Before besinning racking in, how do ensure that'the charging l motor circuits d energize?
- b. buring racking in, how do ensure that the charging motor enargizes when the closing fuses are installed?
QUESTION 0.21 '(1.00) The TS ACTION Statement for SPECIFIC ACTIVITY requires that the plant be in at least HOT STANDBY with T avg less than-500 degrees F within 6 hours should the specific activity exceed the LCO Limit. Explain the TS Basis for reducing T avs to less than 500 degrees F. QUESTION 8.22. (2.00) a) List the 4 rooms which are required by Tech Specs to have operable low pressure'CO2 systems. .(1.0) b) What actions are required.within i hour if one of these systems were-to become INOPERABLE? (1.0) GUESTION 8.23 ( .50) How long may the. quarterly surveillance requirement (0 - 92 days) be extended without declaring the component INOP due to the-surveillance testing not being performed? ! I i (***** END OF CATEGORY 08 *****) (************* ENC OF EXAMINATION ***************) l i i I i l l
, , - . . . - , . - m.-_-. , , . _ _ , _ _ - _ . _ _ _ . - . - . . _ . . . . . . . _ . _ . . - - _ . _ . . _ _ . . . - . . , . ~ . . . . . . _ . _ _ . _ . . , . . . - _ . _ _ - - - - _ - _ - - ,
3/4 LIM! TING CONDITIONS F0E ODE:tATION AND SURVE*LLANCE REOUIREMENTS 3/4.0 ADPLICABILITY B LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Doeration contained in the sue:eeding Specifications is required during the ODERATIONAL EDES or otner conditions spe:ified therein; except that upon failure to mee the Limi*ing Conditions for Operation, the associated ACTION requirements shall be met, t 3.0.2 Noncompliance with a Specification shall exist when the requirements of i the Limiting Condition for Operation and associated ACTION requirements are not met within the spe:ified time intervals. If the Limi-ing Condition for . Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required. 3.0.3 When a Limiting Condition for Operation is not met,. except as provided in the associated ACTION requirements, within one hour action shall be initiated j to place the unit in a MODE in which the Specification does not apply by placing it, as applicable, in: [
- 1. At least HOT STANDBY within the next 6 hours,
- 2. At least HOT SHUTDOW within the following 6 hours, and
- 3. At least COLD SHUTDOW within the subsequent 24 hours.
; Where corre:tive measures a a completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with tne specified time limits o as meas ; red from the time of failure to meet the Limiting Cor.dition fer Operation. "
Exceptions to these requirements are stated in the individual Specifications. 3.0.4 Ent.y into an OPERATIONAL MODE or other specified condition sM11 not be made unless the conditions for the Limiting Condition for Operatica are met I without reliance on provisions contained in the ACTION requirements. This provision shall not prevent passage through OPERATIONAL MODES as required to comply with ACTION requirements. Exgeptions to these requirements are stated in the individual Specifications. i 3.0.5 When a system, subsystem, train, component or cevice is determined to be . inoperable solely because its emergency power source is inoperable, or solely I because its normal power source is incperable, it may be considered OPERABLE l for .the purpose of satisfying the requirements of its applicable Limiting , Condition for Operation, provided: (1) its corresponding normal or emergency I power source is OPERABLE; and (2) all of its recundant system (s), subsystem (s),
- l. train (s), ccaponent(s) and device (s) are OPERARLE, or likewise satisfy the requirements of this Specificatien. Unless poth conditions (1) and (2) are
, g satisified, within 2 hours action shall be initiated to place the unit in a
.L MODE in which the applicable Limiting C:ndition for Operation does not apply by pla:ing it as applicable in:
l 1. At least HOT STANDBY within the next 6 .'rpurs,
! 2. At least HOT SHUTDOW within the following 6 hours, and
- 3. At least COLD SHUTDOW within the subsequent 24 hours.
This Specification is not applicable in MODES S er 6. SEQU3YAH - UNIT 1 3/4 0-1 gpy7 9 1
...w- - - - - . - -w , _ _ - - ,, . . ,-m ,,y e. _ _m,,_.,,y_,
-_ , . _ _ , _,_.w-m,_,.,_m___ _ ,,_.,,--m.. _ - - . . . _ _ _ _ -
j l MCTIV*TV CO'iTR3L SYSTEMS 3 /4.1. 2 BDRAT10h 3YSTEMS FLOW PATHS - SH'JiDOWN LIMITING CONDITION FOR OPERATION 3.1.2.1 As a minimum, one of the following beror. injection flow paths sna11 be OPERABLE:
~
- a. A flow path from the boric acid tank via a boric acid transfer pump and charging pump to the Reactor Ccolant System if only the coric acid storage tank in Specification 3.1.2.Sa is OPERABLE, or
- b. The flow path from the refueling water storage tank via a charging pump to the Reactor Coolant System if only the refueling water storage tank in Specification 3.1.2.5b is OPERABLE.
I APPLICA8ILITY: MODES $ and 6. ACTION: With none of the above flow paths CPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. I SURVEILLANCE REOUIREMENTS I 4.1.2.1 At least one of the above required flow paths shall be cemonstrated OPERA 8LE:
- a. At least once per 7 days by verifying that the tenperature of the heat tracea portion of the flow path is greater than or equal to 145'F when a flow path from the beric acid tanks is used.
l b. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position. . l
REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR 00ERATION 3.1.2.2 At least two of the following three boron injection flow paths shall be OPERABLE:
- a. The flow path from the boric acid tanks via a boric acid transfer pump and a charging pump to the Reactor Coolant System.
- b. Two flow paths from the refueling water storage tank via charging pumps to the Reactor Coolant System.
APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to DPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% delta k/k - at 200*F within the next 6 hours; restore at least two flow paths to OPERA 8LE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REOUIREMENTS i' 4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE: I ! a. At least once per 7 days by verifying that.the temperature of the j heat traced portion of the flow path from the boric acid tanks is greater than or equal to 145'F when it is a required water source,
- b. At least once per 31 days by verifying that each valve (manual, power I operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in i,ts correct position.
- c. At least once per 18 months during shutdown by verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal.
- d. At least once per 18 months by verifying that the flow path recuired l by Specification 3.1.2.2a delivers at least 10 g:m to the Reactor
! Coolant System. htAR 2 51952 3/4 1-8 Amene. tent No. 12 SEgr0YAH - L' NIT 1
j REACT!V!'Y l0fiTE0L SYSTEMS
. CHAR 3!NG PUM8 - 5HUTDOWN LIMITING CONDITION FOR ODERATION 3.1.2.3 One charging pump in the beron injection flow path required by Soe:ification 3.1.2.1 sna11 be OPERASLE and capable cf being powered f ree an OPERABLE shutcown boa F..
An*LICA5ILITY: M00E3 5 and 6. ACTION: With no charging pump OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. s SURVEILLANCE REOUIREMENTS h 4.1.2.3 The aoove required charging puso shall be dem nstrated OPERABLE by verifying, that on recir:vlation flow, the pump eevelops a discharge pressure of greater than or equal to 2400 psig when testad pursuant to Specification 4.0.5. 5
'- SEQUOYAH - UNIT 1 3/4 1-0
)
EEACTIVITY CONTROL SY5TEMS C'AROING 8"975 - 08 ERAT:N; LIMITINO CONDITION FOR OPERATION 3.1.2.4 At least two chargir.g pumps shall be 07 ERA 5LE. APPLIIAf f LITY: MODES 1, 2, 3 and 4. ACTION: With only one cr.arging pump OPERA 5LE, restere at least two charging pumps to OPERABLE status within 72 hours or be in at least Hot STANDEY and carated to a SHUIDOWN MAROIN equivalent to at least 1% delta k/k at 200'T wi'hin the next 6 hours; restore at least two charging pumps to CPERABLE status within the , next 7 cays or be in COLD SHJTDCWN within the next 30 hours. SURVEILLANCE REOU!RE.MENTS
~
4.1.2.4 At least two charging puncs shall be cesor.strated CFERABLE by
- verifying, that on recirculation flow, each pump cevelops a cischar;e ;tessure cf greater than or equal ts 2400 psig when testec pursuant to Spccification 4.0.5.
i I . i ' l I i sE:t'cfAs - en:T : 2/4 1-10 Amer.crer.t xc. i
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'l EMERGENCY CORE COOLING SYSTEMS (ECCS) I 3/4.5.2 ECCS SUBSYSTEMS - T, Greater Than or Equal to 350*F LIMITING CONDITION FOR OPERATION _ . , _ 3.5.2 Two' independent ECCS subsystems shall be OPERABLE with each subsystem comprised of:
- a. One OPERABLE centrifugal charging pump,
- b. One OPERABLE safety injection pump,
- c. .One OPERABLE residual-heat removal heat exchanger,
- d. One OPERABLE residual heat removal pump, and
- e. An OPERABLE flow path capable of taking suction from the refueling water storage tank on a safety injection signal and automatically transferring suction to the containment sump during the recirculation phase of operation.
APPLICABILITY: MODES 1, 2 and 3.
- ACTION:
- a. With one ECCS subsystem inoperable, restore the inoperable subsystem 1
to OPERABLE status within 72 hours or be in at least HOT STAN08Y within the next 6 hours and in HOT SHUTDOWN within the following 6 j hours.
- b. In the event the ECCS is actuated and injects water-into the Reactor
- Coolant System, a REPORTABLE EVENT shall be prepared and submitted to the Commission pursuant to Specification 6.6.1. This report shall include a description of the circumstances of the actuation and the R40 total accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be provided in this report whenever its value exceeds 0.70.
i l SURVEILLANCE REQUIREMENTS i 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:
- a. At least once per 12 hours by verifying that the following valves are in the indicated positions with power to the valve operators i removed:
, November 23, 1984
- SEQUOYAH - UNIT 1 3/4 5-5 Amendment No. 36 l
I EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.3 ECCS SUBSYSTEMS - T,yg Less Than 350*F
._ ~
LIMITING CONDITION FOR OPERATION _ 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:
- a. One OPERABLE centrifugal charging pump,
- b. One OPERABLE residual heat removal heat exchanger,
- c. One OPERABLE residual heat removal pump, and
- d. An OPERABLE flow path capable of taking suction from the refueling-water storage tank upon being manually realigned and automatically transferring suction to the containment sump during the recirculation phase of operation.
APPLICABILITY: MODE 4. ACTION:
- a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour or be in COLD SHUTDOWN within the next 20 hours.
- b. With no ECCS subsystem OPERABLE because of the inoperability of either the residual heat removal heat exchanger or residual heat removal pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reactor Coolant System T**9 less than 350*F by use of alternate heat removal methods.
- c. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a REPORTABLE EVENT shall be prepared and submitted to the Commission pursuant to Specification 6.6.1. This report shall l include a description of the circumstances of the actuation and the R40 total accumulated actuation cycles to date. The current value of-I the usage factor for each affected safety injection nozzle shall be provided in this report whenever its value exceeds 0.70.
l Nove$ber 23, 984 SEQUOYAH - UNIT 1 3/4 5-9 Amendment No. 36 h -
s . -- i W'[ 3/4.! ELECTT.ICAL POW'ER SY!iEMS 3 /4. 2.1 A.Cr SOURCES
- OPERATING 4
s LIMITING CONDITION FOR OPERATION
.:~-
, - 3.8.1.1 As a misimum, the following A.C. electrical power sources snail be i
x OPERABLE:
- a. Two plysically independent circuits between the offsite transmission
- network and the onsite Class 1E distribution system, and
', b. Four separate and independent diesel generator sets each with:
- 1. 'Two diesels driving a common generator t
- 2. Two engine-mounted fuel tanks containing a minimum volume of
'" 250 gallons of fuel, per tank
' A separate fuel storage system containing a minimum volume of 3.
62,000 gallons of fuel. I 4. A separate fuel transfer pump, and N .
- 5. A separate 125-volt D.C. distribution panel, 125-volt D.C.
~ battery bank and associated charger.
APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: 4-l ',
. a. With either an offsite circuit or diesel generator set of the above l
required A.C. electrical power sources inoperable, demonstrate the j s OPERABILITY of the remaining A.C. sources by performing Surveillance i
- Requirtn=nts 4.8.1.1.1.a and 4.8.1.l.2.a.4 within ene hour and at
' least.ence per 8 hours thereafter; restore at least two offsite I circu ns and four diesel generator sets to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hcurs and in
. COLD SHUTDOWN within the following 30 hours.
- b. With one offsite circuit and one diesel generator set of the' above recuired A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.1.a and 4.8.1.1.2.a.4 within one hour and at least once per 8 hours thareafter; restore at least one of the inoperable sources to OPERABLE status within 12 hours or be in at least NOT STANDBY within the next 6 hours and in COLD SHUTDOWN within
' the following 30 hours. Restore at least two offsite circuits and b four diesel generator sets to OPERABLE status within 72 hours from the, time of initial loss or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
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SEQUDYAH - UNIT 1 3/4 8-1 o D
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L. I : e' ELEC8!*2t *WE: SYS?!",!,
- !cN (*e-tinU?dI /
- c. With two cf the above recuired offsite A.C. circuits incoerable, -
de-onstrate the OPERAEILITY cf a diesel generator se:I by perfori::ing Surveillance Pecuirement 4.8.1.1.2.a.: within one hour and at least once per ! hours thereafter, unless tas diesel generator sets are already operating; restore at least one of tne inoperable offsite sources to OPERAELE status within 24 hours or be in at least HOT STMDBY within the next 6 hours. Witn only one offsite source ' restored, restore at least two offsite circuits to OPERA::LE status within 72 hours frten tirse of initial loss or be in at least MCT STANDSY within the next 6 hours and in COLD SHUTDCWN within tne following 30 hours. (. . With either diesel oenerator tats 1A-A and/cr 2 A-A incoerable sinultaneous with 13-B and/cr 23-B, de enstrate the OPEPARILITY of two offsite A.C. circuits by perfomine Surveillance Pecuirenent 4.8.1.1.1.a within ene hour and at least once per 8 hours thereafter; restore at least 1) 1A-A and 2.t-A er 2) 15-F and 25-2 to OPEPARLE status within 2 hours or be in at least FOT STANDBY within the next 6 h:urs and in CULD SHUTNVN within the followino 30 bcurs. 8estore at least fcur diesel penerator sets to OPERABLE status withfr 72 hcurs from time of initial less or be in least HOT $ TANSY within the nex*. 6 hours and in COLD $NUTDOWN within the followino 30 hours. SUPVEILLANCE FEDU!PE*ENTS 4.8.1.1.1 Each of the above reevired indeoendent circuits between the cffsite trans::rission networic and the onsite Class 1E distributien systs: shall be:
- a. Detemined CPERAELE at least once per 7 days by verifying c:rrect breaker %nrents, indicated ;cwr availability, and
- b. Demonstrated OPEFAELE at least once per 18 months durino shutdown by transferrvig (ranua11y and aut:ratically) unit power sucoly fr:rn the nomal circuit to the alternate circuit.
4.2.1.1.2 Each diesel generator set shall be de-enstrated OPE 8ABLE:
- a. In accordance with the frecuency soecifiec in Table 4.!-1 en a ST E E?ED TEST BASIS by:
( Verifying the #uel level in the encinenunted day tanks. 1.
- 2. Verifying the *uel level in the 7 day tank.
- 3. Verifying the fuel trans'er ;urt can be started and transfers fuel fr:m the s:crace syste, to the encine reunted 'uel tanks. -
SECUOYau - tm!T.1 3/4 ! 2 .
. . . , _ . . , ,, ~ . , . . . - - - , , - - _ , , _ _ , _- , _ . _ _ ,
l ELETRICAL POW'ER SYSTEMS SHUTDOWN LIMITING CONDITION FOR ODERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:
- a. One circuit Detween the offsite transmission network and the onsite Class 1E distribution system, and
- b. Diesel generator sets 1A-A and 2A-A or 18-8 and 28-B each with:
- 1. Two diesels driving a common generator, .
- 2. Two engine-mounted fuel tanks containing a minimum voline of 250 gallons of fuel per tank,
- 3. A fuel storage system containing a minimum volume of 62.000 gallons of fuel,
,. 4. A fuel transfer pump, and
- 5. A separate 125-volt D.C. distribution panel, 125-volt D.C.
battery bank and associated charger. APPLICA8ILITY: MODES 5 and 6. ACTION: With less than the above minimum required A.C. electrical power sources OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. SURVEILLANCE REOUIREMELT5 4.8.1.2 The above required A.C. electrical power sources shall be demon-strated OPERABLE by the performance of each of the Surveillance Requirements of 4.8.1.1.1 and 4.8.1.1.2 (except for requirement 4.8.1.1.2.a.5), 4.8.1.1.3, t and 4.8.1.1.4. l t ( l l \ -- SEQUOYAH - UNIT 1 3/4 S-S l l
I INSTRL'*.ENT ATION 3 /a. 3. 2 ENGINEERED SAFETY FEATURE A*TUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION [ l 3.3.2.1 The Engineered Safety Feature Actuation Systee (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall belCPERABLE with their trip i setpoints set consistent with the values shown in tho' Trip Setpoint column of. Table 3.3-4 and with AESPONSE TIMES as shown in Table'3.3-5. AoPLICABILITY: As shown in Table 3.3-3. j ACTION: j
- a. With an ESFAS instrumentation channel or-interlock trip setpoint less conservative than the value shown in the A11bwable Values column of Table 3.3-4, declare the channel inoperable'and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERA 3LE status with the trip setpoint adjusted consistent with the Trip Setpoint value.
- b. With an ESFAS instrumentation channal or interlock inoperable, take the ACTION shown in Table 3.3-3.
SURVEILLANCE REQUIREMENTS a.3.2.1.1 Each ESFAS instrumentation channel shall be demonstrated CPERABLE by the performance of the CNANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Taole 4.3-2. 4.3.2.1.2 The . logic for the interlocks shall be demonstrated OPERABLE during the automatic actuation logic test. The total interlock function shall be demonstrated CPERABLE at least once per la months during CHANNEL CALIBRATION testing of each channel. affected by interlock operation. 4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function
- shall be demonstrated to be within the limit at least once per 18 months.
Enth test shall include at least one logic train such that both logic trains are tested at least once per 36 months and cne channel per function such that all channels are tested at least once per N tiees 18 months where N is the total number of reduncant channels in a specific ESFAS fune-ion as shown in
. the " Total No. of Channels" Column of Table 3.3-3.
{~ SEQUOYAH - UNIT 1 3/a 1-14
.( .
y ,- g -
. t TABLE 3.3-3 (Continued)
M jj ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION - S! E MINIMUH TOTAL NO. CllANNELS CilAllHELS APPLICABLE { FUllCTIONAL UNIT OF CilANNELS TO TRIP OPERABLE MODES ACTION P 8. EllGINEERED SAFETY _ FEATURE ACTUATION SYSTEM INTERLOCKS
- a. Pressurizer Pressure - 3 2 2 1,2,3 22a Hot P-11
- b. T,yg - P-12 4 2 3 1,2,3 22b
- c. Steam Generator 3/ loop 2/ loop 3/ loop 1, 2 22c
- Level P-14 any loop us 30 9. AUTOMATIC SWITCHOVER TO ,
us CONTAINMENT SUMP u ll A. RWST Level - Low 4 2 3 1, 2, 3, 4 18 COINCIDENT WITil Containment Sump Level - High 4 2 , 3 1,2,3,4 18 AND Safety Injection (See 1 above for Safety Injection Requirements)
- g J;
1" R 5 mg . s' . %--
- 4
s _ 7 lt$TD.'. HEN ATION TAELE 3.3-3 (Oe.tireedi TAELE NOTAT*0N
#Trip function may be bycassed in this MODE below P-11 (Pressuri er Pressure p Blo:k of Safety Inje: tion) setooin .
IP Tric fun: tion may De cypassed in this MDDE oelow P-12 (T,y9 51::t of Safety Inje: tion) setpoint.
"The channel (s) ssociatec with the protective fun:tions derive: from the
.out of service Reactor. Coolant Loop shall be pla:ed in the tripoed moce.
The provisions of Specification 3.0.a are not a:plicaole. ACTION STATE'MENTS ACTION 15 - With the number of OPERABLE Channels one less than the Total Number of Channels, be in at least HOT STANDBY within 6 hours and in COLD. SHUTDOWN within the following 30 hours; however, one channel may be bypassed for up to 1 hour for surveillan:e g testing per Specification 4.3.2.1.1 provided the other channel is GP:RABLE.
-ACTION 16 - With the numoer of OPERABLE Channels one less than the Total
- Number of Channels, operation may proceed until performance of the next required CHANNEL FUNCTIONAL TEST, provided the incoersole channel is placed in the tripped condition within i hour.
I ACTION 17 - .With a channel associated.with an operating loop inoperable, restore the inoperable channel to OPERABLE status within 2 hours or be in at least HOT SHUTDOWN within the following 12 hours; however, one channel associateo with an operating loop may be bypassen for ~up to 2 hours for surveillance testing per U Specification 4.3.2.1.1. ACTION 18 - With the number of OPERABLE Channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is demonstrated within 1 hour; one additional channel may be bypassed for up to 2 hours for-surveillance testing per Specification 4.3.2.1.1. ACTION 19 - With less than the Minimum Channels OPERABLE, operation may continue provided the containment ventilation isolation valves j are maintained closed. l ACTION 20 - With the number of OPERABLE Channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT. STANCBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. l . i SEQUOYAH ,0 NIT 1 3/4 3-22 g 7g j
T a INSTR'JMENTATION TAELE 3.3-3 (Continued) ACTION 21 - With the number of OPERABLE Channels one less than the lotal humber of Channels, STARTUP anc/or PD*ER OPERATION may proceec provided the following conditions are satisfiec:
- a. The inoperable channel is placed in the tripped condition within I hour,
- b. The Minimum Cnannels OPERASLE requirements is met; however, one additional channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1.1.
ACTION 22 - With less than the Minimum Number of Channels OPERABLE, declare the interlock inoperable and verify that all affected channels of R16 the functions listed below are OPERABLE or apply the appropriate ACTION statement (s) for those functions. Functions to be evaluated are:
- a. Safety Injection Pressurizer Pressure
- b. Safety Injection High Steam Line Flow Steam Line Isolation High Steam Line Flow Steam Dump Turbine Trip
- c. ,
Steam Generator Level High-High Feedwater Isolation Steam Generator Level High-High ACTION 23 - With the number of OPERABLE channels one less than the Total Number of Channels, be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6 hours; however, one channel may be bypassed for up to 1 hour for surveillance testing per Specification 4.3.2.1. l ACTION 24 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within 6 hours and in at.least HOT SHUTDOWN within the following 6 hours. l ACTION 25 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE , status within 48 hours or declare the associated valve incperable < K16 ! f and take the ACTION required by Specification 1.7.1.5. I I MAR 251982 SEQUQVAH - UNIT 1 3/4 3-23 Amencment No.12 l l
j,, . _ , , . , _ _ . _ . _ , _ . _ , - .
. 3. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDSr AND PAGE 29 ANSWERS -- SEQUOYAH 182 -86/05/26-D.J. NELSON ANSWER 5.01 (1.00) a REFERENCE SON /HDN Nuclear theory ANSWER 5.l02 (1.00) p b.
i REFERENCE SON /WBN HTFF Chap. 2E ANSWER 5 03 (1 00) c. REFERENCE
, steam tables
! ANSWER 5.04 -(1.00) b 4 REFERENCE TS 3/4.2.2 ANSWER 5.05 (1.00)' C , REFERENCE-TPT Requal Lesson Plan, Cycle IIr-Day 1-1985 ' TPT SD13, 'CVCS', pp 23 SON 'CVCS's pp 24
-004/020; A2.13(3.4/3.9)
- - - _ _ . . . . . . - ~_- . . .-- . - . . .._
_ _ __ ..7_--_....- p# K84 UNITED STATES
/ 9[o,, NUCLEAR RESULATCRY COMMISSION ,
{ w g REG 40N il
- t 101 MARIETTA STREET, N.W., SulTE 2000
- 2 o, ATLANTA, GEORGIA 30323 s,....../
- , - - - . , - , a, ,, - , , - - , . , - , . , - , - - , - - - - - - - - - - . - ---,-..-n..- . - . - - - - - - -- . - - . , - - - --------.------ - -
7
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 30 ANSWERS -- SEQUOY AH 1&2 -86/05/26-D.J. NELSON ANSWER 5.06 (1.00) d REFERENCE EIH: L-RO-606r pp 4, 5; Fig. 4 OSEP: 02-2/3-A, pp 177 - 100; 02-0G-A, pp 60 - 61 DFNPt Xenon' and Samarium LPr pp 5, 6; R0 84/03/05 Westinghouse Nuclear Training Operations, pp. I-5.77 - 79 Turkey Point, Reactor Core Control, pp. 4 34 001/000-K5.13 (3.7/4.0)
ANSWER 5.07 (1.00) C REFERENCE EIH: GPNT,Vol VII, Chapter 10.1-83-86 BSEP: L/P 02-2/3-A, pp 172 - 176; 02-0G-A, pp 57 - 60 Westinghouse Nuclear Reactor Theory, pp. I-5.77 - 79 Turkey Point, Reactor Core Control, pp. 4 28 001/000-K5.13 (3.7/4.0) ANSWER 5.08 (3.00) AFFECTED LOOP + UNAFFECTED LOOP [ OL esiv fa p
- a. decreases + increases E0.53 each
- b. decreases + increases CO.53 each
- c. decreases + increases CO 5] each REFERENCE SON /WBN HTFF/ Nuclear theory L
g 7 _ _ - _ . . . . . _ _ . ~ . _ - - _
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 31 ANSWERS -- SEGUOYAH 1&2- -86/05/26-0.J. NELSON ANSWER 5.09 (1.50) a) More Negative (+0._5 ea) ,
b) Less Negative c) More Negative REFERENCE Westinghouse Nuclear Trainins Operations, pp. I-5.6 - 16 CN10, " Reactor Core Control", pp 3-16/28 001/000; K5.26(3.3/3.6) ANSWER 5.10 (1.00) a .- true CO.5]
- b. true EO.53 REFERENCE Nuclear theory,' Inst. Notes VI ANSWER 5.11 (1.00)
- 1) Density difference'between cold and hot les (+.5 ea)
(or Heat sink and Heat source with a Delta T)
~2) Height d.fference between hot and cold legs (or S/G and Core)
REFERENCE ~ CNTO, " Thermal / Hydraulic Principles.and Applications II", pp 14-16/17 ~ Westinghouse, " Mitigating Core Damage", CH 1, pp 11/12 002/000; K5.10(3.5/3.9) ANSWER 5.12 -(2.00) I a) Unit 4 * (+.5) -due to a lower Beta coefficient at EOL (+.5) b) Unit & I (+.5) due to MTC beins less negative, so Tavs must decrease come to add + reactivity) -(+.5) 4EFERENCE CNTO ' Reactor Core Control *, pp 3-21 & " Fundamentals of Nuclear Reactor Physics", pp 7-31
a
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 32 ANSWERS -- SEQUOYAH 1&2 -86/05/26-D.J. NELSON 001/000; K5.49(2.9/3.4) & KS.10(3.9/4.1)
ANSWER 5.13 (1.50)
- a. increases due to decreasing Tave E.53
- b. increases ove to decreasing Tave E.53
- c. decreases due to decreasing Tave and increasing beron concentration E.5J REFERENCE SON /WE:N Nuclear theory ANSWER 5.14 (1.00)
Lower f1'ow at the same power level results in a larger delta T; CHF decreases toward the top of the core. Lower coolant velocities result in less stripping action which removes nucleate bubbles, a steam film can form at louer heat flux. REFERENCE SON /WBN HTFF ANSWER 5.15 ( .50) Margin to criticality decreases E0.1] by 2/3 E0.4]. REFERENCE SON /WBN Neutron Sources and Suberit. Mult. ANSWER 5.16 (3.00) (three req'd)-
- 1. Delta T across the core E0.5] constant / decreasing E0.4] and less than ET F h '-
-d -'._1
- 2. Core outlet temp.CO.5]
L u - . E -: , i r constant /decreasingEO.5] [0./]
- 3. Teold = Tsat for P(S/G)[0.5}]* constant /decreasingEO.53
- 4. SG pressure E0.53' decreasing or stableEO.53
- 5. .Thot E0.5] stable or decreasing E0.5]
REFERENCE WTS; Ch. 14, p. 27; ADI-35; SON ES-0.3 l l
)
w .n. .n,
, n . - - - - ~ ; v^' ^~: Tr^27- ~~ ; =-~"~ -
i.
- 5. ~ THEORY OF NUCLEAR POWER PLANT OPERATION, CLUIDS, AND PAGE '33 ANSWERS -- SEQUOYAH 1&2 -86/05/26-D.J. NELSON
. ANSWER 5.17 (1.00)
In a small LOCA core heat is not beins removed sufficiently by the break an'd'little'ECCS flow is bein3 delivered due to elevated _RCS pressure. REFERENCE-SON /WBN S0;' Aux Feed Sysi p. 8.f8 ANSWER 5.18 -(1.00) remove boric acid that is precipita'ted on' upper core surfaces (+.5) terminate any boilins or steam formation in upper head region (+.5)
~ REFERENCE Westinshouse PWR Systems Manual, pp 4.2 ~TPT SD-21,.'ECCS", pp 26 .
EPE-Olli EK3.13 (3.8/4.2) j
' ANSWER 5 ~.19 ( 2. 0 0')
EA0 = C(Pt.- Pb)/(Pt + Pb)3 x 100% (+.5) (or DELTA I/RX POWER = EAO)
~
Initially, Grester power senerated in lower seqment- _of quadrant and EAD will be more ne3ative. This condition will be~ accentuated as xenon burns out in the lower and~ builds in the. upper segments ~. (+.75) As.. xenon builds into the' lower segment while depleting in the upper section
-due to the neutron flux shift, a hisher percentase of-power will be_senera-ted in the upper ses' ment and EAD will. shift towards a positive value.(+.75) l REFERENCE TPT.0P-12304.8; CNTO, " Reactor Core Control"7 PP 4-28/29 and Section 8
-001/010;~K5.34(3.2/4.1) l 1
3 1 P l I p L t~
1
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 34 ANSWERS --'SEGUOYAH 1&2 -86/05/26-D.J. NELSON ANSWER 5.20 (1.50) a) Swell due to increase in steam flow (+.5 ea) b) OP Delta T Rod stop is actuated c) Pstm decreases affects the density compensation on Stm flow indication Ahe, lowsr P,r- n a. - , t ~<r eQrca., sa . pa,. s + -- pt, ~ .
REFERENCE SON Lesson plans on " Steam Dumps", " Rod control'r " Main Steam'r 'RPS' ANSWER 5.21 (1.00) 4, 1, 3, 2 ( .25 for each switch required to put in correct order) REFERENCE TPT GET Radeon Training Lesson Plan, pp 3 068/000; K5.04(3.2/3.5) ANSWER 5.22 (2.00) Tave: 30.4 X 0.25 X -15 = -114 pcm Power: 25 X -12 = -300 pcm Void: -25 pcm Xenon: -50 pcm total; -489 Pcm Doron: -409 / -9 = 54.3 ppm dilution (accept 52 to 56) REFERENCE SON /WBN Nuclear theory
- - - - ~ - - - - - - - - -- - - ---- ~-
- 7---
. y : : = - -- ;'. - -
1 t i
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 35 ANSWERS - -SEGUOYAH 182. -86/05/26-D.J. NELSON ANSWER '6 01 (1.00) b.
REFERENCE i SON /WBN SD; RCSi CVCS West. PWR.Sys~. Manual ANSWER- 6.02 (1.00).
; a a
2 REFERENCE , 10CFR50.~46 ANSWER. 6.03 (1.00) c. 4
-ANSWER 6.04 (1.00)
- a. reversed b.~ normal REFERENCE West. PWR Sys. Manual / SON /WBN HTFF ,
ANSWER 6.05 (1.00) . b. l REFERENCE SONP Diesels handout, p. 6. e 4 + f 4 Y b 1 t
,.~.-.----..,..,m._.-._ m_..--. ._r.m- . . . . - . - , . , - - - _ . . . , - , . - . ~ - . . - _ _ . ~ . _ . _ . . , - . . . - - ~ , _ _ - , , -
-~
.. . . . ,. . , . ~ 3 4 - r *, s 6 .. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 36 ANSWERS -- SEQUOYAH 1&2 -86/05/26-D.J. NELSON ANSWER 6.06 (1.00) c REFERENCE SONP Rod Control Lesson, pp 5 & 6 of 11.
- Systems-Manual, Chapter 11.1, p. 11.1-63.
ANSWER 6.07- (1.00) c. 1 REFERENCE i System Manual, Chapter 4, p. 4.0-2 i ANSWER 6.00 '(2.50) I
- a. higher
- b. as is
- c. higher
- d. lower I
e, hisher
- ANSWER 6.09 ( .50)
True. 4 REFERENCE Reactor Protection Lesson, .p. 8 of 13, item d. i ANSWER 6.10 (3.00) , s. 1,3 e. 1 j b. 1,3 f. 1,2,3,4
- c. 1,3 s. 1
- d. 1,3 h. 5 E0.23 each of 15 responses REFERENCE West. PWR Sys. Manual, 11.1 i
i I
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 37 ANSWERS -- SEGUDYAH 1&2 -86/05/26-D.J. NELSON ANSWER 6.11 (1.00)
Solenoid ,Airi open; 33%; 500 [+.25 ea3 aw o n, REFERENCE Aux Feed Lesson Plan, p. 5 of 8. ANSWER 6.12 IE.uv,-(d.h
- a. CLOSE
- b. CLOSE
- c. OPEN
- d. CLOSE
- e. N0
- f. 'OP""- '0-
, S. CLOSE
- h. NO
- i. NO
- j. -MO- if CE [0.2 ea.]
REFERENCE SONP System Description, ECCSr CVCSr MNSTMr CCW ANSWER 6.13 (2.00)
- a. high pressuriner level (letdown isolates)
- b. low pressurizer pressure (rods drive Tave and PZR level down)
[1.03 each REFERENCE Channel Failure Handout; TAB A&C: 3,4-11 ANSWER 6.14 (1.00)
~
crfrerbrdra cet l> % o t < N Ar
- i REFERENCE West. PWR Sys. Manual L
g-- _ ._ 3
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 38 ANSWERS -- SEQUDYAH 182 -86/05/26-D.J. NELSON ANSWER 6.15 (2.50) decreasing
- 2335 PORV closes (4.05 for setpoint, +.2 function) 2310 sprays start closing 2260 sprays closed 2250 variable heaters start to come on 2220 variable heaters full on 2210 low pressure alarm, backup heaters on 2210 Backup heaters on 1970 Low pressure SI block enabled 1970 Low pres.sure reactor trip 1870 Low ~ Pressure SI REFERENCE SON SD; RCS,RPS ANSHER 6.16 (1.50)
Lo-Lo. level in 2/4 S/Gs (+.3 ea) SIS Loss of both MFP Loss of one MFP > 00% power LOSP
-REFERENCE Aux.-Feed Lesson Plan, pp 4-6. .
ANSWER 6.17 (2.00)
~
- 1. two percent radial flux tilt (^.5 ca for any 4)
- 2. improper rod sequence
- 3. shutdown bank ods less than 220 steps
- 4. rods within a bank greater than 12 steps from the bank demand
- 5. rods greater than 12 steps froin each other within a bank REFERENCE'
. SON, SOI-55-1M4, XA-55-48, p 25 ANRHER 6.18 (.1.00)
Used during cooldown because the steam space detector response is poor due to poor heat transfer.
. e ,- , = - - - , - - ,
y_ _ a
.6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 39 ANSWERS -- SEGUOYAH 1&2 -86/05/26-D.J. NELSON REFERENCE SON /WDN Inst. Notesi RCS ANSWER 6.19 (1.00)
Turbine runbackt0.53 to eG440.53.
~
" 7 0, REFERENCE w/ no./ g'o2 00I-5.1 & 6.1; p7.
AMSWER 6.20 (1.00)
- a. loop 4 hot leg (?T)
- b. either the relief valve downstream of the letdown orifices if it is .
unisolatede PORV on the PZR, or RHR suction reliefs. REFERENCE SGN Inst. Notesi RHR ANSWER 6.21 (1.00) Interlock to prevent openning of RHR when RCS pressure > 380_psis. . REFERENCE " b M M058 ) . SON /WBN Inst. Notes; RCS os, sh.f ,S def. w p my w / N M ANSWER 6.22 (1.00)~ To ensure.the regenerative heat exchanger always has RCS system pressure in it to prevent flashing of high temperature water.- REFERENCE SON /WBN SDi CVCS
l
- 7. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 40
- --------~~~~~~~~~~~~~~~~ ~~~~R 55i5tBEiEEt E5sTE5t ANSWERS -- SEQUOYAH 1a2 -86/05/26-D.J. NELSON ANSWER 7.01' (1.00) d.
REFERENCE SONP E-0 pp. 2-5 ANSWER 7.02 (1.00) b. REFERENCE SONP GOI-2, pg. 1 ANSWER 7.03 (1.00) a. REFERENCE SONP S01-62.18, pp. 8, 9 ANSWER 7.04 (1.00) a. REFERENCE SONP AOI-20, pp. 10 - 12 ANSWER 7.05 (1.00) H (* REFERENCE SONP AOI-3D, p. 1 of 2 J
f
).
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 41
~~~~R565UL5EICAt E5sTR5L-~~~~~~~~~---~~~~~~~~~~~
ANSWERS -- SEQUOYAH 182 -86/05/26-D.J. NELSON ANSWER 7.06 (1.00) To prevent EST actuation (+.7) on lo-lo S/G level (+.3) REFERENCE SONP GOI-1, p. 4; precaution T. ANSWER 7.07 (1.00) b REFERENCE SUNP GOI-2, p. 16 ANSWER 7.08 (1.50) a) False (+.5 ea) b) True c) False REFERENCE Westinghouse User's Guide for TPT EOPs, pp 5-12 ANSWER 7.09 (1.50)
- a. 1. 3 rem C+.5 ea]
- 2. 300 mrem
- b. Plant E- -i,i_. ' '-
(or authorized representative) E0.5] REFERENCE E - SON, RCI-ir p. 7-8 ANSWER 7.10 (1.00) a) RWST: 29% SUMP:10% (+.25 ea) b) Stop all pumps taking a suction on RWST until swapever is complete (+.5)
- 7. PROCEDURES - NORMAL, ACNORMAL, EMERGENCY AND PAGE 42
~~~~R ED UL55iE5t E5sTR6L-------~~~-~~---~~~~~~--
ANSWERS -- SEQUOYAH 1&2 -86/05/26-D.J. NELSON REFERENCE SONP ES-1.3 p. 1 of 4; ES-1.2 p. 1 of 3, App. A ANSWER 7.11 (1.00) PZR level (+.35) < 20 % (+.1)
< 50 % adverse containment (+.1)
RCS Subcooling (4.35) < 40 degrees F (+.1) REFERENCE SONP ES-0.2, pp 3, 4 ANSWER 7.12 (1.00) a) 5 (+.5 ea) b) 2 REFERENCE SONP, FHI-7, p 4. ANSWER 7.13 (2.00)
- 1. Verify Reactor Trip.
2'. Verify Turbine Trip.
- 3. Verify Shutdown Boards Enersized.
- 4. Check if SI Actuated.
- 5. Verify ECCS status.
- 6. Verify Cntmt. Isolation. 6 E45
- 7. Verify MFW Isolation.
G. Verify AFW status.
- 9. Verify CCS Pumps Running.
- 10. Verify ERCW Pumps Running.
- 11. Verify EGTS and ABGTS Running.
- 12. Check Cntmt. press less than 2.01 psis.
- 13. Check Tavg.
REFERENCE SQNP E-0, pp 2-5
r 1 l
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE' 43
~~~~R E5i5L55iCAL 55sTR5L-~~~~~~~~----~~~--------
ANSWERS -- SEQUOYAH 1&2 -86/05/26-D.J. NELSON ANSWER 7.14 (1.50)
- 1. Containment spray pumps running.
- 2. MSIVs closed.
- 3. Phase "B' Isolation (status monitor panel 6E and 6F lights). (1.5)
REFERENCE SONP E0I-Or p5 ANSWER 7.15 (3.00) Q g, p 44 a) 1. U n e >:p e c t e d rise in on S/G level with feedwater flow (+.5 ea) reduced or stopped,
- 2. High radiation from any S/G B/D line by rad monitor
- 3. High radiation from any one S/G B/D line by analysis or rad monitor
- 4. High radiation from monitoring of steam lines b) SDLOCA considerations If the pumps are not tripped, an inadvertant loss of RCPs later (+.3) could result in uncovering the core due to e:< c e s s i v e loss of mass (+.7)
L. 1. (At least or_ f the four CCP/SI p u n. ,s running) If '451 pressure d. crees 7s below 1250 rsig ( 2 credit givei for Uncontrol ed Dept .ssurizationi 2 Phase d entmt, isol atior [1.03 REFERENCE E01-37 pp 2-3. Westinghouse ERG Manual on E-3
d"~#,2:%#:-- :z . .
- t. . . .
2 e 1
.7. . PROCEDURES'- NORMAL, ABNORMAL, EMERGENCY AND PAGE 44
---~~---------------~~~~
~~~~ R5656L65 65E~C6HTR5L .
. ANSWERS -- SEQUOYAH 182 -06/05/26-D.J. NELSON.
i 4 ANSWER 7.16 (1.00)
- 1. Rod control. in manual (or individual bank select) and restore Tave.
- 2. If unable to stop rod insertion, trip reactor.'
E0.53 each l REFERENCE SON,AOI-28, . p '1 l iANSWER 7 .17._ (2.00) -
- s. Addition oflhydrazine will result in the additional production F, of gases that must be removed prior to opening the RCS. Also accepted,.
the gas presents personnel hazards if opened to atmosphere when-
- it exists innthe RCS. (1.0)
.b . -(H202) will cause activated corrosion products (Cof58, Co 60,
~
and others') to be put-into solution in the RCS. E0.53 This will result in decreased radiation levels (and corresponding radiation exposures). [0.53 (1.0) REFERENCE - GOI-3A, p3
- GOI-3C, P 30 S01-62.30,p 6.
i . ANSWER 7.1S (1.00)
- a. to prevent thermal _ lock up.
l b. to prevent bank overlap malfunction. i REFERENCE . .-S O N ,-_ G O I - 1 , - p --3 , 4 .t - 1
- _ , _ _ , . . . _._- _ _ . . , _ . . . - _ - . _ _ . - . _._ .. f.....__ _ , _ _ - _ - . .-
f-
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 45
- -~~~~--------------~~---
~~~~R565ELUU565L UUNTRUL ANSWERS -- SEQUOYAH 1&2 -86/05/26-D.J. NELSON
. k ANSWER 7.19 (2.00)
- a. Normal readings on containment temperaturer pressurer radiation, and sump level. [0.53 Normal readings on Aux building radiation and ventilation monitoring.
CO.53 Normal readings on Steam Generator blowdown and vacuum pump exhaust radiation. E0.53
! . Due to the low RCS temperatures and pressures following a loss of secondary coolant (with the RCS intact), SI repressurization could lead
\ to RCS overpressurization a d da age. (PTS) E1.03
& t REFERENCE SON EDP-2r p.5; EOP-1 p.4; EOP-0, p.7 ANSWER 7.20 (2.00)
- a. The RCS temperature'and baron concentration are beins maintained at the hot shutdown, Xenon free condition. (1.0)
- b. The RCS has-been borated to the cold shutdown concentraton AND the plant is beins cooled down. (1.0)
REFERENCE SONP GOI-2, p2 ANSWER 7.21 (1.00)
- a. avi c th'n fnur innnen T 'C ' giuots thsn er m ,ml Lu 12:0 .
IkJ. 4v 3 ii i U ' > yeuued i u u ii' i / C 0 " M " g (1.0) b. p Ob REFERENCE t u r : csp s e. i , 6
--1
- b. sd=4 < "o. # 7 ". *g
[pt=:@7:;-4: C :-ll
]
k : ,- t a 7 '. PROCEDURES ' NORMAL, ABNORMAL, EMERGENCY-AND PAGE 46
+
~~~~RA5iBE55iEEE E5nTR5E------~~------~~~~~~----
ANSWERS - SEQUDYAH 1.82 -86/05/26-D.J. NELSON cANSHER 7.22 (1.50) A. Announce radiation abnormal in the reactor building and.over the ~ PA for-all personnel to evacuate the co'ntainment_ building. (1.0) j~ B. Airborne gases'(I, Kr, etc.) (0.5) REFERENCE AOI-29. 4 1 l' + }
^
(- (
- i. '
l - I' f
~
I I 1 I a i I': ( i .
~
g -
+n- vm -s r m-,y- - , , - ,c -,-+-.-g---- -- ,--,------+---,y , wm, - . , -
E. . .
- 8. ' ADMINISTRATIVE PROCEDURES, CONDITIONSr AND LIMITATIONS PAGE 47 ANSWERS -- SEQUOYAH 182 -86/05/26-D.J. NELSON ANSWER 8.01 (1.00) b REFERENCE GGNS TSs Definitions
; SON TSs Definitions Section 1.0.
4 ANSWER 8.02- (1.00) a
- REFERENC
10 CFR 55.31.e ANSWER 8.03 (1.00) d. REFERENCE SON TS Section 1.0 ANSWER 8.04 (1.00) b. REFERENCE SONP GOI-6, p.-4 ANSWER 8.05 (1.00) e. REFERENCE SON TSs 3.0.5, 3.1.2.1, 3.1.2.2, 3.1.2.3, 3.1.2.4r 3.8.1.1, 3.8.1.2 f.
-~
v . . ,
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 48 ANSWERS -- SEQUOYAH 1&2 -86/05/26-D.J. NELSON ANSWER 8.06 (1.00) e.
REFERENCE SON TSs 3.0.3, 3.3.2.1, 3.5.2, 3.5.3 ANSWER 8.07 (1.50) a) Unit 2 ) b) Unit 2
-c) Unit 2 REFERENCE SONP AI-27 pp 10 PWG-23: Station Directiv'es related to staffing / activities (2.8/3.5)
ANSWER 8.08 (1.00) PORC; Plant Manager; US001 safety review (+.25 ea) REFERENCE SONP Al-9, pp 2 PWG-23: Use of. procedures / station directives (2.8/3.5)
. ANSWER 8.09 (1.00) one 8 hour shift; tas the disconnected leads; H.N.; double verify (+.20ea)
REFERENCE SONP AI-3, pp 4 PWG-14: Tagging / Clearance Procedures (3.6/4.0). L.
r s O. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 49 ANSWERS -- SEGUOYAH 182 -86/05/26-D.J. NELSON ANSWER 0.10 (1.50) 1 MTC within analyzed range (0.3)
- 2. Trip instrumentation within operating range (0.3)
- 3. Above P-12 setroint (0.3)
- 4. Pr capable of being. operable (0.15) with a steam bubble (0.15) (0.3)'
- 5. Rx vessel above its RT (NDT) (0.3)
REFERENCE Cat, TSr p. 3/4 1-17 FNP/SGNP TS B3/4.1.1.4 001/050; PWG-5 (2.9/4.3) ANSWER 8.11 (1.00) Immediately initiate and continue boration (0.3) at greater than or equal to 10 gpm (0.3) of a solution containing greater than or equal to 20,000 ppm or equivalent (0.3) until the required SHUTDOWN MARGIN is r e s t o r e d l+.1) REFERENCE SON TS 3.1.1.2 ANSWER 8.12 (2.00)
- a. Alert
- b. (Notification of) Unusual Event
- c. General Emergency
- d. Site Area Emergency (0.5 each)
REFERENCE , EIH: GET Handbook, pp 57, 58, 60, 61 HNP-x-4420, HNP-x-4520, HNP-x-46207 HNP-x-4720 BFNP: BFN-IPD, IP-1, p 1; RQ E5/04/01 SON : SON IPD RIP-1; NUREG-0654 s
- k. .
f
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 50 ANSWERS -- SEQUDYAH 182 -86/05/26-D.J. NELSON ANSWER 8.13 ( .50)
NODE 6 - REFUELING' REFERENCE C0i! ~~ TAbtE 1.1 ANSWER 0.14 (1.00)
- Storage position, (the lead shielded storage area in the seal table room)
- Inserted to within 10 feet of the core REFERENCE '
SONP AI-\0, p. 2 ANSWER 8.15 ( .50) Auxiliary Bldg.(.17) Elevation'669 (.17) near U-1 Auxiliary Turbine Feed Pump Room (.16) REFERENCE SON OSLA 73, p. 1 ANSWER 8.16 (2.00)
- 1. Control rods in a single group move together with no individual rod insertion differing by more than + or - 13 steps from the group demand position.
- 2. Control rod gro.ips are sequenced with overlapping groups (as described in TS 3.1.3.6) per procedure.
-3. The control rod insertion limits of TSs (3.1.3.5 and 3.1.3.6) are maintained.
- 4. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
REFERENCE SON TS B 3/4 2.2 and'3/4.2.3 TPT TS B3.2.5
.i
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 51 ANSWERS -- SEQUOY All 1&2 -86/05/26-D.J.-NELSON 001/000; K5.46(2.3/3.6)
ANSWER 8.17 (2.50) 1, Leakese (except CONTROLLED LEAKAGE) into closed systemsr(0.5) (such as pump seal or valve p~acking leaks) that are captured and conducted to a sump or collecting tankr(0.5) or
~
- 2. Leakage into the containment atmosphere' from sources that are both specifically located and known (.34) .either not to interfere with the operation of leakaSe detection systems (.33) or not to be PRESSURE BOUNDARY LEAKAGE, (.33) OR
- 3. Reactor coolant leakage through'a steam generator to the ,
secondary system.(0,5) REFERENCE SON TSs Section 1.0 ANSWER 8.18 (1.50)
~
- a. onc (0.5)
- b. Maintenance section must be notified (0.5) to check circuit before further fuse replacement (0,5).
REFERENCE SON AI-16, p. 1 ANSWER 0.19 (1.00)
- a. Always backseat valves to isolate packing from line pressure.
- b. Excessive force could cause the separaton of stem and disc.
REFERENCE SONP G01-6, p. 3
- 8. ADMINISTRATIVE PROCEDURES,' CONDITIONS, AND' LIMITATIONS PAGE 52 ANSWERS -- SEQUOYAH-1&2 -86/05/26-D.J. NELSON ANSWER 1 8.20 (1.00)
- a. Check the.tossle switch on for the chargins motor circuit.
- b. Listen for the closins spring charsing up when.the closins fuses are' installed.
REFERENCE SONP GOI-6' p. 6 . ANSWER 2.21 (1.00) Reducin.3'T avs to < 500 degrees F prevents the release of activity should a SG tube rupture-(0.5) since P sat of the primary ~ coolant - c~ is'below the lift pressure of the atmospheric steam relief valves.(0.5) REFERENCE SUN TS B.3/4.4.8
' ANSWER 8! . 22 : (2.00) a) Computer Roomi Aux Instrument Roomi EDG Rooms; Fuel Oil Pump Room
~(+.25 ea) b) . Establish.a continuous fire watch (+.25) with backup fire. suppression (s.25)-for those areas in which redundant sys'tems/ components could-be damaged.(+.25)
Establish hourly fire watch patrol for other areas (+.251 REFERENCE 4 SONP TS 3.7.11.3 096/000; K4.06 (3.0/3.3) & PWG-36 (2.8/3.7) i.- ANSWER' 8.23 ( .50) 23 days (25% x 92 days) REFERENCE
' SON TS 4.0.2 i
l l I
r-w=
' o U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: SEGUOYAH 1&2 REACTOR TYPE: PNR-WEC4 DATE ADMINISTERED: 86/05/26 EXAMINER: D.J. NELSON APPLICANT: ____ _I _
jf---_-__ INSTRUCTIONS TO APPLICANT: Use separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination star ts.
% OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 30 0 25 0
___1_0___ ___1_0_ ___________ ________
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICSr HEAT TRANSFER AND FLUID FLOW
^ ^^ 25
_~_' _ _ _ _'_ _ _ _ _ I _0 0 _ ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 30.00 25.00
- __---. ----_----_- -------- 3 INSTRUMENTS AND CONTROLS 5__
^
25.00 PROCEDURES - NORMAL, ADNORMAL, _----___ _-____ _ _--_--___ ________ 4. CONTROL ll?* 8^ ^ - t .' ^ 100.00 TOTALS FINAL GRADE _________________% All work done on this examination is my own. I have neither given not received aid. 5PPLIC5Ui~5~5 5U5TURE~~~~~~~~~~~~~~ J
f y
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 2
--- isEER557sERICs- GEEi iEEssFEE EA6 FEUi5 FE5s QUESTION 1.01 (1.00)
Which of the following will cause the fuel temperature coefficient (pcm/ degree) to become less negative?
- a. fuel temperature increase
- b. boron concentration decrease
- c. control rod insertion (at constant power)
- d. increase in the Pu-240 to U-238 ratio GUESTION 1.02 (1.00; Select the statement about single speed, motor driven, centrifugal pumps that is correct.
- a. 'Upon throttling open the discharge valve to increase flow, discharge pressure decreases and therefore motor amps decreases.
- b. Upon throttling open the discharge valve to increase flow, net positive suction head required increases and differential pressure across the pump decreases,
- c. Upon. throttling shut the discharge valve, flow decreases, total developed head decreases and. net positive suction heard available increases.
- d. Pump cavitation can be reduced by throttling open the discharge valve thereby reducing total developed head.
QUESTION 1.03 (1.00) The most serious problem wit h reaching the critical heat flux is caused by
- a. the poor thermal conductivity of steam,
- b. the heat stress on the fuel cladding when nucleate boiling ceases and cladding temperrture increases.
- c. the displacement af baron from the core as steam bubble formation becomes significant.
- d. the pressure fluctuations in the RCS caused by steam bubble formation and subsequent collapse.
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
f
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3
---~isEEs557sEsiEs- REEi iEEssFEE Es5 FEUi5 FE5s QUESTION 1.04 (1.00)
Which of the following describes the changes to steam that occurs.between the inlet a'nd outlet of a REAL (not ideal) turbine?
- a. enthalpy decreases, entropy decreases, quality decreases.
- b. enthalpy decreases, entropy increases, quality decreases.
- c. enthalpy decreases, entropy constant, quality constant.
- d. enthalpy constante entropy increases, quality decreases.
QUESTION 1.05 (1.00) Of the following, which must.the main condenser remove the most heat from to condense? (Assume steam is of equal quality) a.'one pound of steam at 0 psia.
- b. one pound of steam at 300 psia.
- c. two pounds of steam at 400 psia.
- d. two pounds of steam at 1200 psia.
QUESTION 1.06 (1.00) Attached Figure 4 219 shows.a power history and four possible xenon traces (reactivity vs' time). Select (a, be er or d) the curve that correctly displays the expected xenon transient for the given power history. l (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****) i I l. i i
P e . 39' - e . s'e O to s see do de g. q, i _ _ . . . . . . A ~ to .W 60 90 les age pts 5
- b. = = - - - . . - - -
m w a 80 too eso ige 1 C. i. . _ _ M 46 6e 30 too sta 410 l E d, g j d--J . . . _ _ _ so 90 ao 80 soo no sgo TIME ( H o uRS) F l C,URE Uli
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 4
~~~~TREER55YsERICs?' SEAT TREU5FER U6~ FLU 56~FLUU GUESTION 1.07- (3.00)
At 30% power a reactor coolant pump trips. With control rods in manual and all other systems in automatic and no operator / protective actions occur, indicate the effects on the following at the end of the trtosient (increases,-decreases, or remains the same) AFFECTED LUDP + UNAFFECTED LOOP
- a. steam senerator level +
+
- b. steam flow +
+
- c. de3ta T +
OUESTION 1.08 (1.50) A reactor is taken critical with xenon concentration at zero. Power is raised to 50% at 5%/ min. A trip occurs as power reaches 50%. What is the xenon concentration trend (increasins, decreasing, or near equilibrium):
- a. one hour after. the trip?
- b. 12 hours after the trip?
- c. 4 hours after the trip and the reactor has been restarted and power is at 25%?
JUESTION 1.09 (1.50) Indicate wh'ther e the following changes cause the differential baron worth to become MORE NEGATIVE, LESS' NEGATIVE, or REMAIN THE SAME. Consider each separately.
- a. Baron concentration increases (0.5)
- b. Moderator temperature increases (0.5)
- c. Core age increases'(at a constant boron concentration) (0.5)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 5
- ~ ~~~~TU5RR567sARICs? SEAT TRAssFER 5UU~FLUiU FL6E QUESTION 1.10 (1.00)
Indicate TRUE or FALSE for the following statements concerning the effect that deleyed neutrons have on reactivity:
- a. Because delayed neutrons are born at lower energies than prompt neutrons, they are.less likely to leak out of the core resulting in a positive effect,
- b. Delayed neutrons are born at an average energy incapable of causing fast fission of U-238 creating a negative effect.
QUESTION 1.11 (1.00)
- a. TRUE or FALSE: During cold plant conditionse you would expect the COLD calibrated PZR level instrument to indicate HIGHER than the HOT calibrated level instrument.
- b. Give two different conditions involving the reference leg which could result in a false high level on the PZR level instrument.
QUESTION 1.12 (2.00) p.93,s v4,) BothPu-239andPu-240toncentrations[increaseovercore life. Indicate whether this will cause the MAGNITUDE of the following parameters to INCREASE, DECREASE or HAVE NO EFFECT: a) Average Delayed Neutron fraction b) Core Reproduction Factor c) MTC (assume it is negative) d) FTC QUESTION 1.13 ( .50) Answer TRUE or FALSE for the following statement concerning Xenon-135
- production and remevait
- a. At full power equilibrium conditions about half of the xenon is
-produced by iodine decay and the other half is produced as a direct fission product.
(***** CATEGORY 01 CONTINLPD ON NEXT PAGE *****)
- 1. PRINCIPLES OF NUCLEAR POWER PLANY OPERATION, PAGE 6
--- iREER557 Easies- sEAi isAs5 FEE As5 FEUi5 FE6s -QUESTIra 1.14 (1.00)
List three significant heat transfer advantages of a counter flow heat exchanger. over a parallel flow heat exchanger. QUESTION 1.15 -(1.50) During the performance of an emergency boration while at power, how and why are the following par ameters affected? (assume no control rod movement)
- a. r.ubcooling PowE A
- b. .over ;2 c :::. c differential temperature setpoint
- c. control rod worth OUESTION 1.16 (1.00)
If bank D control rods were positioned at core midplane long enough to establish equilibrium conditions, and then withdrawn, describe the effect on Delta I. QUESTION 1.17 (1.00) Explain how decreasing RCS flow. (at constant power) will result in decreasing DNBR. QUESTION 1.18 ( .50) If the equilibrium count rate in a suberitical reactor TRIPLES due to a reactivity addition, what happens to the margin to criticality (direction and magnitude)? 00ESTION 1.19 (1.00) Enplain why the equilibrium (at power) value of samarium reactivity is independent of power level. (***** ' A1 EGORY 01 CONTINUED ON NEXT PAGE *****)
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 7
- ~ ~~~~T sERR55isAsics? SEAT TREU5FER 5U5~FLUi5 FL5s . QUESTION 1.20 (3.00)
What are THREE parameters AND their trends which are indications that Natural Circulation in the RCS is established? (numerical values not req'd) QUESTION 1.21 (1.00). Aux. Feed flow is more critical on a sinall LOCA than on a large LOCA. Why is this true? QUESTION 1.22 (1.50) From 80% power, 100% flow (assume delta T = 40 degrees F) a station blackout occurs. Natural circulation is established and core delta T stabilizes at 40 degrees F. If decay heat is 2.0% of full power, what is the mass flow rate (% of full flow)? Show all calculations. QUESTION 1.23 (2.00) Given the following, calculate the required baron change to increase reactor power from 75% to 100% while maintaining constant rod position. Moderator temp, coeff. -15 pcm/ degree F Doppler-only power coeff. -12 Pem/% power Void reactivity change -25 pcm Xenon change -50 pcm Baron coeff. -9 pcm/ ppm (***** END OF CATEGORY 01 *****)
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 8 00ESTION 2.01 (1.00)
Which statement below regarding the RCP shaft seals is NOT correct? (1.0)
- a. Only ti and t2 seals are designed to withstand full system pressure
- b. Leakoff from 12 seal is used to maintain the level in the standpipe used to supply-coolins water to 13 seal
- c. When an individual ti seal bypass line is isolated, the other RCP seal bypass lines are isolated as well
- d. The #1 seal is a " floating' face seal vice a " rubbing" face seal like the 12 and 13 seals.
QUESTION 2.02 (1.00) According to 10CFR50.46, which of the following.is NOT a design criteria of the Emergency Core Cocling System subsystems,
- s. The calculated peak centerline temperature shall not exceed 2000 degrees F.
- b. The maximum cladding oxidation shall not exceed 17% of the total cladding thickness.
- c. The calculated total amount of hydrogen generated from the cladding reaction with water shall not exceed 1% of the amount that would be generated if all cladding around the fuel reacted.
- d. Calculated chanages in core geometry shall be such that the core remains in a coolable configuration.
QUESTION 2.03 (1.00) Which of the following is NOT a purpose of.the Reactor Protection Syst.em?
- a. Prevent maximum DNBR from . increasing above 1.3.
- b. Prevent the maximum power density (kw/ft.) from exceeding limits.
- c. Prevent RCS pressure from exceeding 110% of design pressure.
- d. Prevent secondary system pressure from exceeding 110% of design pressure.
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
o ,
- 2. PLANT DESIGN INCLUDING SAFETY'AND EMERGENCY SYSTEMS PAGE 9 QUESTION 2.04 (1.00)
With three reactor coolant pumps operating indicate if the flow in the given' loop segment will be in the NORMAL or REVERSED direction in the loop with the non-operating pump. a) T-h RTD manifold b) T-c.RTD manifold GUESTION 2.05 (1.00) The purpose.of the CVCS demineralizers is to:
- a. Remove all chemicals from the RCS fluids.
- b. -Remove solvable and insolvable material from the RCS.
- c. Replace insolvable material with solvable ions.
- d. Provide a method for boron control during reactor operations.
QUESTION 2.06 (1.00) Besides the overspeed shutdown,.which of the following diesel engine / seneratne Thutdowns is enabled.durins an emersency start of the diesel?
- a. Voltage restraint overcurrent relay, (51V).
- b. Generator differential t .7 1 a y , (87).
- c. Phase balance relay, (46).
- d. Low lobe oil pressure.
QUESTION 2.07 ( .50) With normal power ~ unavailable and ONE vital battery out of service, how long will the remaining THREE batteries be capable of supplying all loads
' required for safe shutdown of BOTH units?.
i i (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****) a
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYST. EMS PACE 10 GUESTION 2.08 (1.00)
Which of the below features enhances the operation of the ice condenser and containment spray for heat removal?
- a. Containment design, such that th'e delta P between upper and lower containment drives the air circulation.
- b. Ventilation coolers and recirculation fans are used to mix the air and provide additional cooling.
- c. Air return fans provide flow to return the air from the upper containment to the lower containment.
d .- Pressure operated doors open to allow upper containment air to flow through to the lower containment. QUESTION 2.09 ( .50) TRUE or FALSE? After tripping a' bistable in a 2/4 logic system, one of three remaining signals reaching the bistable setpoint will cause a trier even though the logic SYSTEM remains as a 2/4 system. QUESTION 2.10 -(1.00) Name the plant temperature (HOT or COLD) and the RCP status (starting FIRST , or LAST reactor coolant pump) for which you would expect the highest pump starting current. (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****) L
7
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE ?.1 OUESTION 2.11 ' 5 . 0 "-
- i. t o For the following components, indicate whether they will receive an OPEN, CLOSE, or NO signal as a result of a -safety injection (with Phase 'A') >
initiation signal.
- a. Control room supply ducts
- b. Main feed bypass valves
- c. SI accumulator discharge isolation valves
- d. Normal charging header isolation valves
- e. Main steam isolation valves f.
g. F.W C T Lu S I c . ..+ ~ a c ' 10 r. Seal water return isolation valve
'! v e - Ad d McM
- h. Component cooling isolation valve from RHR system
- i. Component cooling isolation from letdown heat exchanger
- j. Steam supply valves to turbine-driven feed pump Cl*rbY Aly &h p&
QUESTION 2.12 (1.00) Fill in the blanks below to correctly comp l ete the statement regarding the Motor Driven Auxiliary Feedwater Pump level control valves: These valves are ___________ operated and will fail _______ on a loss of air. They are normally set to maintain a S/G level of _____% and if pressure downstream drops to < ______ psig, the valves will close automatically. QUESTION 2.13 (1.00) List four relief valve discharges accepted by the PRT from inside containment. (other than PZR PORV and safties) DUESTION 2.14 (2.00) Aside from alarms / annunciators, list eight indications that are monitored which would indicate a coolant leak from the primary system. QUESTION 2.15 (2.00) List four different ways to emergency borate. (Valve Numbers not required) (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 12
'0UESTION- 2.16 (1.00)
Arranse the followins in the correct sequence for rod withdrawal (one step).
- a. Lift coil 0FF
- b. Stationary gripper coil DN
- c. ' Moveable gripper coil-ON
- d. Moveable gripper coil 0FF
- e. Lift coil DN
- f. Stationary grfpper coil 0FF QUESTION 2.17 (1.50)
Wh,t segments of the Rod Control System are provided with signals from the Turbine Impulse Pressure Detectors? DUESTION 2.10 (1.50) List the 5 Auto-start signals for the Turbine driven AFW pump. QUESTION 2.19 (1.00) Why is a seal b / pass necessary for the reactor coolant pump 41 seal? 00ESTION 2.20 (1.00) What_is compared in the power ranse detector current comparator? 00ESTION 2.21 (1.00) Followins a RCS boron dilution at power, how is the water in the pressurizer brought to the same concentration? (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
i 2. PLANT DESIGN INCLUDING SAFETY AND1 EMERGENCY SYSTEMS PAGE 13' :
'0UESTION 2.22 (1.00)
- a. When the RHR system is. controlling RCS pressure during solid plant-
- ' conditions, where does the water leave the RCS? i b. If-the control-valve which separates RHR and CVCS fails shut, what 3 relief valves could limit RCS pressure?'-(redundant system reliefs count as~one response). OUESTION 2.23 (1.00) The pressurizer has a resistance temperature detector.(RTD).in the STEAM
' space that is normally used to ~ indicate the PZR saturation temperature.
! Assuming-this.RTD is operable, during what.P l ant evolution is the RTD in
- the PZR WATER space utilized, and why isn't the PZR STEAM RTD used?
QUESTION 2.24 (1.50) Which chemical is used to accomplish the following. methods of RCS l corrosion control? i a. . Control pH during startup
- b. Scavenge oxygen during a startup from cold conditions l
c.-Control oxygen during normal' at power operations QUESTION 2.25- (2.50) s . . a) What force (s) is/-are used for Main Steam ~ Isolation Valve (MSIV)) l a. closing ? i b. opening ? ! b) What. component prevents MSIV damage on fast closure? E l c) What are ALL the MSIV automatic closure signals? I. i i
- (***** END OF CATEGORY 02 *****)
3' f t l' i
, , , , . - ,-,,.,,.,-.,--Lr,e,,,-,,._,n ,n-n ,m,. ~.m.-, - + - . , , . . . , ,...-,,,e ,-.,---,_,--,,n,_,.n,.-,-,,,.- ,n,
o.
- 3. INSTRUMENTS AND CONTROLS PAGE 14 OUESTION 3,01 (1.00)
-What set of signals below are sent to the Reactor Protection System to indicate a Turbine Trip?
- a. Throttle valves closed & Auto Stop Oil pressure l'o w
- b. Throttle valves closed.& EHC pressure low
- c. Governor valves closed & Auto Stop Oil pressure low
- d. Governor valves closed & EHC pressure low OUESTION 3.02 (2.50)
For the following, how will the indication respond (higher, lower, as is) to the given failure?
- a. RTD open circuit in detector
- b. Intermediate range compensation voltage fails high with reactor power at 100%.
- c. Source range pulse height descriminator setting fails low with reactor power in the source range.
- d. Thermocouple junction opens
- e. Steam flow Pre 55Ure compensation to the eteam flow detector fails high QUESTION 3,03 (1.50)
Indicate whether the OT Delta-T AND OP Delta-T GETPOINT will INCREASE, DECREASE or NOT CHANGE if the following operating parameter changes occur. CONSIDER EACH CHANGE INDEPENDENTLY.
- 1. Pressurizer pressure decreases 100 psig.
- 2. The N-41 lower detector fails low.
- 3. Overdilution of the RCS, which causes rods to insert slowly to maintain constant load and Tave.
7] g4 in ScrY. (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
- 3. INSTRUMENTS AND CONTROLS PAGE 15
' QUESTION 3.04 (3.00)
Hatch t'he typ.e(s) of. rod motion thatl is blocked with the signal that causes the rod blockt ~
' SIGNAL BL'OCKED ROD MOTION
-a. OP delta T 1. Automatic ~ Withdrawal l
- b. OT delta T 2. Automatic Insertion
- c. Power Ranse at.103% 3. hanual Withdrawal
- d. Inter. Range at 20% equiv.. 4. Manual Insertion
- e. Control Bank D > 220 steps
_ 5. No Blocked Motion
- f. Ursent Failure in-the Power Cabinet
- s. P imp < 15%
-h. Tave vs. Tref < 1.5 degrees.F QUESTION 3.05- (1.00) a) If the turbine trips from an initial power of LESSJthan_50%,
which. steam dump controller.will contrul the steam dumps? b )- In addition to beins an input to the steam. dump controllersr what function does Teve provide in the Steam Dump Control circuitry? GUESTION 3.06 (2.00) The followins failures occur causins a subsequent automatic reactor trip. What' protection signal would;cause the trip? Assume the reactor is initially at 100% power and steady state conditions, all systems in automatic and no operator action. Treat each independently. a).CVCS flow rate drops to a minimum of 30 spm. b) A narrow range.(controlling) cold les RTD fails hish. (****r CATEGORY 03 CONTINUED ON NEXT PAGE *****)
1 .,
- 3. INSTRUMENTS ~AND CONTROLS PAGE 16-100ESTION 3.07 . (3.00)
- a. List THREE Main Feedwater isolation signals.
- b. List ALL automatic actions that occur onLa feedwater isolation signal.
- c. What automatically happens as' main feed pump suction pressure i approaches saturation (low NPSH)?
- d. What is the1setpoint-for the ' Low NPSH st MFP's" alarm AND what are the sensin9 Point locations for this signal?
) QUESTION '3.08 (1.75) List the SEVEN RPS Permissives by nomenclature (eg P-2), that. block safeguards actions. Indicate the permissives that must be done manually. QUESTION. -3.09 (2.00) List four-conditions that will-generate a ' Computer Alarm Rod Dev and Seq HIS PWR Range Tilts' alarm. QUESTION 3.10 (3.00) What are three functions of the Overspeed Protection Controller (OPC)? In your answer include which valves are actuated for each function. QUESTION 3.11 (1.75) List the protective / control outputs for the following: (Increasing severity of.the-same function like lo and lo-lo level count as one response) a) LOOP Tave Instrument, (FOUR REQUIRED) b) Auctioneered HIGH.Tave Instrument, (THREE REQUIRED) (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
- -- 4
- 3. INSTRUMLNTS AND CONTROLS PAGE 17-1 QUESTION _.3 12 (2.50)
List ALL the protection,' alarm and control' functions provided by the PZR; pressure instruments as pressure decreases from 2350 psis. (Include the applicable setpoints) QUESTION 3.13' (1.00) ( What is the reason for the interlocks on'the CVCS letdown valves and orifice isolation valves? QUESTION 3.14 (1.'00) While operating at 92% power, 'the 13-heater drain tank level . goes high enough to'cause the water in the tank to begin ( dumping to the condenser. . According to SOI's.5 1 & 6.1, what effect j will'this have on.the turbine? QUESTION 3.15 (1.00) Most.RCS pressure control / protection signals-are generated by the PZR preccure. instruments. What control or protection signsl-is generated by the pressure instrument on a-T-hot les? QUESTION' 3.16 (2.00)
~
List'ALL the automatic actions that occur upon detection of high radiation in the'following radiation. monitors
- a) . Component Cooling water liquid effluent monitor.-
b) Steam. Generator blowdown liquid effluent monitor. c) Containment purge air exhaust monitor. d) Fuel Pool radiation monitor. (***** END OF CATEGORY 03 *****)
-4. PROCEDURES - NORM AL , ABNORM At r': EMERGENCY ~AND PAGE 18
'-~~~EA5i5E55i5EE 55 ATE 5E------~~-----------~~~--
J QUESTION 4.01 (1.00) Which of the following is.NOT an immediate' operator action for a F Safety Injection as stated.in-E-0?
- a. Verify Containment Isolation.
- b. Check.Tavs.
- c. Verify.AFW status.
- d. Verify Steam Dumps actuated.
OUESTION 4.02 . ( 1. 00 ) . It-is necessary to reduce the critical baron concentration by 200 ppm i prior to pullin3Lthe control-banks. Prior. to the dilution, the' source i range-instruments read CO and 37 cps. 'After reducing the boron concentra-tion by 100 ppm the same instruments read 62 and 75 cps. Which of the j following is the proper operator _ action in.accordance with'GOI-2? 7 a. Stop the dilution and borat back to the original ~ count rate. I b. Stop the dilution and evaluate the situation.
- c. Continue the dilution and continuously monitor the count rate.
- d. Continue the dilution and recalculate the ECC.
QUESTION 4.03 .(1.00) During normal CVCS operation, which of the following is an abnormal condition and would require operator action-to. correct? .. .a. VCT pressure is 15 psis. I b. The temperature of the fluid leaving the letdown heat exchangers l is 127 F. I- c. The RCP seal injection water tempeature is 120 F and flow to the I seuls is 8 spm/ pump.
- d. RCP seal' differential pressure is 300 psid.
i. I L (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****) 1 l' ; i ii._...__.. . . _ _ , , . _,_..__._..__,__.__.,,_,4-__._____ _ . , _ _ , _ , , . . _ . , . , _ . . . _ , . .
n...
- 4. PROCEDURES - NORMAL,-ABNORMAL, EMERGENCY AND PAGE' 19
- ------------------~~----
~~~~R5656LUUEC5L cUsTRUL QUESTION 4.04 (1.00)
'Which of the following statements.concerning the procedure.for a-
_ dropped RCCAcis correct?
.a. Upon starting. recovery of the dropped RCCAr an URGENT FAILURE alarm will occur because.the lift toils for the other rods in the group have been' disconnected. -
- b. The delta flux target band is not applicable during a dropped
'RCCA malfunction and recovery.
.c . If two or more RCCA's1have dropped, manually trip-the reactor and-proceed in accordance with EP-1.00.
- d. Recovery from a dropped RCCA will be facilitated if Tavs !
is higher than Tref prior to commencing withdrawal of the dropped RCCA.
. 00ESTION 4.05 (1.00)
During an inadvertent dilution accident while at 100% power, with controls in automatic and no operator action, which of the following will be the most probable cause of a reactor trip?
- a. Pressurizer-low pressure.
- b. Over-temperature delta T.
- c. Over power delta'T.
d.. Power range monitor positive rate. QUESTION 4.06 (1.00) According to a note in-GOI-2, what condition must be met prior to exceeding 600 RPM on the main turbine?
- a. Main Feedwater Regulating valves are to be in automatic,
- b. Tavs is to be at the no-load value.
- c. The low pressure turbine inlet metal temperature must be greater t 400 de3rees F.
d.' Steam dumps must be in Tave mode. (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
- s
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 20
- ---------------~~~~~~~~~
~~~~R A5i5L55iEAL 56sTs5L
-GUESTION 4.07' (1.50)
List SIX alarms that-may actuate'which would be symptoms of rods failing to insert following a decrease in turbine load. QUESTION 4.00 (1.00) If. the reactor trip breakers are closed and the steam senerators are under nitrogen pressure, the nitrogen pressure must be vented off the steam senerators prior to opening the MSIV's. Why must this be done? 00ESTION 4.09 (1.00) TRUE or FALSE?
- a. The transfer of ECCS suction to the Containment Sump is accomplished RWST level is < 29% and Containment Sump Level is > 10%.
- b. When RWST level reaches 0%r the Containment Spray Pumps are shifted to the Containment Sump.
QUESTION 4.10 (2.00) TRUE or FALSE?
-a. When the axial flux difference monitor is inoperabler the AFD must be lossed once a shift by performing SI-44.
- b. Any off-frequency turbine operation is to be reported to the results section for record keepins.
- c. If the " Rod Control Banks Limit Low" alarm comes in when criticalt commence boration to clear'the alarm.
- d. When the quadrant power tilt ratio alarm is inoperabler the GPTR must be calculated every 12 hours by performing SI-133.
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****) l
* : e
.4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 21
- -~~~~~-- "-----~~~~~~~~~
~~~~R A5i5L6EiEAL E5sTR5L 1
QUESTION -4.11 (2.50) Match the evolutions performed during a power increase in column A with the
'ower at which it.is normally performed in column B.' Column-B answers may p
be used more than once. COLUMN A COLUMN B
'a . verify. chemistry within limits 1. 35%
- b. place-MSR in service 2. 90% .
- c. load second MFPT 3. 50%
- d. Perform calorimetric. calibration 4. 30%
- e. verify P-8 light goes out 5. 40%
GUESTION 4'.12 '(1.00) What are the.SI Re-initiation criteria of ES-0.2, 'SI'Terminiation'? (Include'. parameters associated with adverse containment ~ conditions) QUESTION 4.13 (1.00) List the quarter'ly. exposure limits for the following a),-TVA employeer Whole Body. b) TVA employeer extremities c) Non-TVA employee without exposure record, Whole Body QUESTION 4.14 (2.00) LIST the THIRTEEN immediate actions to be taken for a Saftey Injectiont ~in accordance with Emersoney Procedurer E-0. (Substeps are NOT required) QUESTION 4.15 (1.00) What are ALL the Immediate-Operator Actions for a continuous insertion of a control rod bank? (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****) t
r-
- 4. PROCEDURES - NORMALr ABNORMALr EMERGENCY AND PAGE 22
~~~~R5656LUU5CEL C5ATR5t-----~~~~-------~~~-~~--
00ESTION 4.16 (1.00) A NOTE in GOI-2r " Plant Startup from Hot Standby to Minimum Loadr* states that if control rods were withdrawn 5 steps during heatupr the control rods must be fully inserted prior to withdrawing rods. a) Why are the rods withdrawn 5 steps during heatup? b) Why must they be inserted prior to withdrawal? QUESTION 4.17 (2.00) a) If the plant was operating in Mode 1 when the RCS pressure exceeds 2735 psis, what action must you taker in accordance with the Technical Specifications? INCLUDE applicable time limit. b) If the plant was in Mode 3 when the RCS pressure exceeds 2735 psis, what action must you taker in accordance with the T. S.? INC'.UDE applicable time limit. QUESTION 4.18 7pdSY(2.00) p r r a s'. W. t Er i n io s?
- b. The 'SI Termination Criteria' for Loss of Primary Coolant and Loss of
. Secondary Coolant have maj r diff rences. What is the major REASON for these differences?
..[
QUESTION 4.19 (3.50) According to A01-4D, ' Nuclear Instrumentation Malfunction, Power Range Failure,' the immediate operator action requires that the rod control be switched to MANUAL if in AUTOMATIC when a PR Instrument fails HIGH. a) Identify the particular component of the rod control system that makes this immediate action necessary and explain its basic operation. (1.5) b) Identify four bypasses or bistables which must be tripped if this situation were ta occur. (2.0) (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
. 1 p2"4 UNITED STATES
( y,
~9 o
NUCLEAR RECULATORY COMMISSION REGION I! a
$ g 101 MARIETTA STREET, N.W., SUITE 2000 s
- 2 ATLANTA, GEORGIA 30323
\...../
4 f 1 1 l 4
l
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 23
~~~~~~~~~~~~~~~~~~~~~~~~
~~~~R5056LUU 55L'UUUTRUL QUESTION 4.20 (1.00) g,j,,~ g h C h > %, }
What are the TWO guidelines from EDI-1 Appendix D that indicate inadequate core cooling exists? QUESTION 4.21 (1.50) Make a rough sketch of the Curves for Reactor Core Safety Limits for four loop operation (Exact Numbers are NOT required). Ensure that you indicate the parameters that are being measured and the region of acceptable operation. (x*r** END OF CATEGORY 04 *****) (xxxxxx******* END OF EXAMINATION ***************) i
i 8 UNITED 8TATES c# "%,,$ , .. NUCLEAR f.E20LATORY COMMISSICN , .
; g REGION il 4
g E 101 MARIETTA STREET, N.W., SUITE 2000 o, ATLANTA, GEORGIA 30323
%, p*[
4 n 4 i I 1 i l l l l 4 I l l
f~o m3 - * , -
@yCDe ei91Ciency'o (tet tork out)/(Energy in) 2
, o cg s e.vo t - 1/2 at
[ = mC'
- KE = 1/2 mv a = (Vf - 73 )/t A = AN A=Ae' g
PE = mgn yf = y, + at w = e/t A= an2/t1/2 = 0.693/t1/2 W=v P
~
A= nD 2 1/2'N
- bl / S Y 4 [(t1/2)
- Itb))
AE = 931 am
. m=VavAo -Ix Q .= i.nah
- I=Ieo
. t Q = mCpat Q = UAL T I = I n
e'"* , { Pwr = Wfah I = I,10'*/ M TVL = 1.3/v 5 P = P 10 "'II) HVL = -0.693/u p = p et/T o = SUR = 26.06/T SCR = 5/(1 - K,ff) CR, = S/(1 - K,ff,) SUR = 26o/t* + (s - o)T CR j (1 - K,ff)) = CR2 II ~ keff2) T = ( **/s ) + [(8 - 8 Y Io l N
- II(I ~ Keff) = CR)/CR 3 7 = s/(o - s) M * (I ~ Keffo)/II ~ Keffl)
T = (8 - e)/(y,) SOM = ( - K,ff)/K ,ff a = (Keff-l)/Keff * 'Keff/Keff t' = 10 second . A =0.1 seconds] o = [(t*/(T K,ff)] + [a,ff /(1 + IT)] Idjj=Id P = (IeV)/(3 x 1010) I jd) 2 ,2gd 2 22 2 I = oN R/hr = (0.5 CE)/d (meters) R/hr = 6 CE/d2 (f,,g) , Water Parameters Miscellaneous Conversions I gal. = 8.345 lem. I curie = 3.7 x 1010 gps I ga;. = 3.78 liters 1 kg = 2.21 lbm 1 ft* = 7.48 gal I hp = 2.54 x 10 Btu /hr . Density = 62.4 1 /ft3 1 mw = 3.41 x 10 Btu /hr Density = 1 gm/c lin = 2.54 cm Heat of vaporization = 970 Btu /lom 'F = 9/5'C + 32 Heat of fusion = 144 Btu /lbm ,
'C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. Hg. 1 BTU = 778 ft-lbf I ft. H O 2
= 0.4335 lbf/in.
e = 2.718 l l ~
= - --
Volume, ft'/lb [ritMpp. 8tupb Enteopy. 8tur.t , I
', Water Evap Steam Water tvap Steam Water Evep Ste e m T At A A 8t 8 4 s,
't 'e 's t t
.-C.02 1075.5 1075.5 0.0000 2.1873 2.1873 32 32 0.08859 0.01602 3305 3305 3.00 1073.8 1076.8 0.0061 2.1706 2.1767 35 35 049991 Obl602 2948 2948 l
8.03 1071.0 1079.0 0.0162 2.1432 2.1594 40 40 0.12163 0 01602 2446 2446 13.04 1068.1 1081.2 0 0262 2.1164 2.1426 45 45 0.14744 0.01602 2037.7 2037.8 18.05 1065.3 1083.4 0.0361 2.0901 2.1262 to 90 0.17796 0 41602 1704.8 1704.8 28.06 1059.7 1067.7 0.0M5 2.0391 2A946 60 60 0.2561 0.01603 1207.6 1207.6 l 868.4 38.05 1054.0 1092.1 0.0745 1.9900 2 0645 70 70 0.3629 Ohl605 868.3 30 4 633.3 48.04 1048.4 1096 4 0.0932 1.9426 2.0359 to 0.5068 0.01607 633.3 468.1 58.02 1042.7 1100.8 01115 1A970 2.0086 to 90 0.6981 0.01610 468.1 350.4 350.4 68.00 1037.1 1105.1 0.1295 12530 1.9825 100 300 0.9492 0.01613 0.1472 1A105 1.9577 110 0.01617 265.4 265.4 77.98 1031.4 1109.3 110 1.2750 87.97 1025.6 1113.6 0.1646 1.7693 1.9339 130 130 1A927 OD1620 201.25 203.26 . l 157.33 97.96 1019 3 1117A 0.1817 1.7295 1.9112 130 130 2.2230 Obl625 157.32 0.1985 1.6910 1A895 140 2 0.01629 122.98 123.00 107.95 1014.0 1122.0 140 2.8892 0.2150 1.6536 1.8686 150 4 0.01634 97.05 97.07 117.95 1008.2 1126.1 4 150 3.718 0.2313 1.6174 1A487 160 0.01640 77.27 77.29 127.96 1002.2 1130.2 160 4.741 62.04 62.06 137.97 996.2 1134.2 0.2473 1.5822 1A295 170 170 5.993 021645 0.2631 1.5480 1A111 130 041651 50.21 50.22 148.00 990.2 1138.2 130 7.511 0.2787 1.5145 1.7934 ISO 0A1657 40.94 40.96 158.04 984.1 1142.1 l 190 9J40 977.9 1146.0 0.2940 1.4824 1.7764 300 300 11.526 0.01664 33.62 33.64 168D9 27A2 178.15 971.6 1149.7 0.3091 1.4509 1.7600 210 210 14.123 0.01671 27A0 180.17 970.3 1150.5 0.3121 1.4447 1.7568 212 212 14.696 0.01672 26.78 26.80 23.15 188.23 965.2 1153.4 0.3241 1.4201 1.7442 230 l 220 17.186 0A1678 23.13 230 19.381 198.33 958.7 1157.1 0.3388 1.3902 1.7290 i 230 20.779 0.01685 19.364 340 16.321 208.45 952.1 1160.6 0.3533 1.3609 1.7142 240 24.968 041693 16.304 250 13.819 .218.59 945.4 1164.0 0.3677 1.3323 1.7000 i 250 29.825 0.01701 13.802 1 228.76 938.6 1167.4 0.3819 1.3043 1.6862 360 240 35.427 0.01709 11.745 11.762 l 238.95 931.7 1170.6 0.3960 1.2769 1.6729 270 270 41.856 0 01718 10.042 10.060 l 924.6 1173A 0.4098 1.2501 1.6599 3RD 380 49.200 0.01726 8.627 8.644 249.17 259.4 917.4 1176.8 0.4236 1.2238 1.6473 250 390 57.550 0 41736 7.443 7.460 269.7 910.0 1179.7 0.4372 1.1979 1.6351 300 300 67.005 0.01745 6.448 6.466 l 280.0 902.5 1182.5 0.4506 1.1726 1.6232 310 310 77.67 0.01755 5.609 5.626 4.914 290.4 394.8 1185.2 0.4640 1.1477 1.6116 330 320 39.64 0.01766 4A96 340 3.770 3.788 311.3 878.8 1190.1 0.4902 1A990 1.5892 Sto 117.99 0.01787 1.5678 350 2.939 2.957 332.3 862.1 1194.4 0.5161 1.0517 360 153.01 0.01811 1.5473 Sep 2.317 2.335 353.6 844.5 1198.0 0.5416 1A057 380 195.73 0.01836 375.1 825.9 1201.0 0.5667 0.9607 1.5274 400 400 247.26 0.01864 1.8444 12630 420 1.4997 396.9 806.2 1203.1 0.5915 0.9165 1.5080 420 305.78 0.01894 1.4808 440 1.2169 419.0 785.4 1204.4 0.6161 0.8729 1.4890 440 381.54 0.01926 1.1976 460 0.9942 441.5 763.2 1204.8 0.6405 0.8299 1.4704 460 466.9 0.0196 0.9746 480 0.8172 464.5 739.6 1204.1 0.6648 0.7871 1.4516 480 566.2 0.0200 0.7972 0.6749 487.9 714.3 1202.2 0.6890 0.7443 1.4333 500-500 680.9 0.0204 0.6545 1.4146 520 0.55 % 512.0 687.0 1199.0 0.7133 0.7013 520 812.5 0.0209 0.5386 1.3954 540 0.4437 04651 536.8 657.5 1194.3 0.7378 0.6577 540 962.8 0.0215 0.6132 1.3757 560 0.3651 0.3871 562.4 625 3 1187.7 0.7625 5(0 1133.4 0.0221 0.5673 1.3550 550 0.2994 0.3222 589.1 589.9 1179.0 0.7876 580 1326.2 0.0228 0.2675 617.1 550.6 1167.7 0.8134 0.5196 1.3330 900 600 1543.2 0.0236 0.2438 1.3092 820 0.2208 646.9 506.3 1153.2 0.8403 0.46S9 620 1786.9 0A247 0.1962 640 0.1802 679.1 454.6 1133.7 0.8666 0.4134 1.2821 640 2059 9 0.0260 0.1543 1.2458 660 0.1443 714.9 392.1 1107.0 0.8995 0.3502 660 2365.7 0 0277 0.1166 1.2086 680 0.1112 758.5 310.1 1068.5 0.9365 0.2720 640 2708.6 0.0304 0.0808 0.0386 0.0752 822.4' 172.7 995.2 0.9901 0.1490 1.1390 700 700 3034.3 0 0366 1.0612 0 1.0612 705.5 0.0508 0 0.0508 906.0 0 906.0 705.5 3203 2 TABLE A.2 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (TEMPERATURE) A.3
votome, it'/it, E nthalpy. Stubb Entrapy. Sto/sb a f Energy. 9tw/lb
- P'e n 1
- r"P Evep Steam Water Evep Steam Crater trop Ste:m W:ter Steam P" p**
Wreter psie F hg h, so sq 8, er g er, Vs *q *e he 0.01602 3302.4 3302.4 0.00 1075.5 1075.5 0 2.1872 2.8872 0 1021.3 OA8e4 ep.ese6 32.018 0.01602 2945.5 29455 3 03 1073 8 10763 0 0061 2.1705 2.1766 343 1022.3 S.10 0.10 35 023 0 0271 2.1140 2.1411 13.50 1025 7 0.15 0.15 4Edb3 0 01602 2004 7 20047 13.50 1067.9 1081.4 1526 3 21.22 1063 5 10841 0 0422 2 07?8 2.1160 21.22 1028.3 DJo 0.20 L3.160 0 01603 1526.3 32.54 1032 0 64484 0 01604 1039 7 1039.7 32.54 10b7.1 1089.7 0.0641 2.0169 2.0809 0.30 0.30 0.0799 1.9762 2.0562 40.92 1034.7 OA0 0.40 72 869 0.01606 792.0 792.1 40.92 1052.4 1093.3 0 01607 641.5 641.5 47.62 1048 6 1096 3 0 0925 1.9446 2.0370 47.62 1036 9 0.5 0.5 79.586 5324 1038.7 9.6 85.236 0 01609 540.0 540.1 53 25 1045.5 10933 0.1028 1.9186 2.0215 0.6 e . '. s *. gy .e .go og . D01%10. t 46633% ' 46694 5610 -10487 4 1008 94..%.. 68966e.2.0083 58.ne .1049.3 . 0.7 , 94.38 0.01611 413.67 411.69 62.39 1040.3 1102.6 0.1117 1.8775 1.9970 6239 1041.7 OA 0.8 66.24 1042.9 0.9 0.9 98.24 0.01612 368 41 36843 66.24 10381 1104.3 0 1264 12606 1.9870 0.01614 333.59 333 60 69.I3 1036.1 11058 0.1326 12455 13781 69J3 1044.1 i 3J 1.0 101.74 94A3 1051A RA 2.0 126.07 0.01623 373.74 17336 94.03 1022.1 1116.2 0.1750 13450 1.9200 118.73 109.42 1013.2 1122.6 0.2009 1.6854 14864 109 41 1056.7 S.0 3.0 141 47 0.01630 118.71 120.90 1060.2 4.0 4.0 152.96 0.01636 90 63 90 64 120.92 1006 4 1127.3 0.2199 1.6428 13626 73.53 130 20 1000.9 1131.1 0.2349 1.6094 13443 130.18 1063.1 5.0 S.0 16224 0.01641 73.515 61.98 138 03 996.2 1134.2 0.2474 1.5820 1A294 138.01 1065.4 6.0 6.0 170.05 0.01645 61.967 0.01649 53.634 53.65 144 83 992.1 1136 9 0.2581 1.5587 1A168 14431 1067.4 7A 7.0 176 84 S.O 182 86 0.01653 47.328 47J5 150 87 988.5 1139.3 0 2676 1.5364 13060 15034 1069.2 S.0 9.0 189.27 0 01656 42.385 42.40 156.30 985.1 1141.4 0.2760 1.5234 1J964 15628 10704 9.0 38.404 38 42 161.26 982.1 1143.3 0.2836 1.5043 1J879 161.23 1072.3 30 10 193.21 0.01659 14.696 212.00 0.01672 26382 26 80 180.17 970.3 1150.5 0.3121 1A447 IJ568 180.12 1077.6 14.696 15 213.03 0.01673 26.274 26.29 181.21 969.7 1150.9 0.3137 1.4415 1.7552 181.16 1077.9 35 20 227.96 0.01683 20 070 20 087 19627 9601 1156.3 0.3358 1.3962 1J320 196.21 1082.0 to 30 250.34 0.01701 13 7266 13.744 218 9 945.2 1164.1 0.36B2 1.3313 1.6995 218 5 1087.9 30 40 267.25 0 01715 10 4794 10 497 236.1 933.6 1169.8 0.3921 1.2844 1.6765 236 0 1092.1 40 50 261.02 0.01727 8 4967 8 514 250.2 923 9 1174.1 0 4112 1.2474 ).6585 250.1 1095.3 50 60 292.71 0.01738 7.1562 7.174 262.2 915 4 1177.6 0.4273 1.2167 1.6440 262.0 1098.0 60 70 302.93 0.01748 6.1875 6205 2723 907A 1180 6 0 4411 1.1905 1.6316 272.5 1103.2 70 j 80 312.04 0.01757 54536 5 471 232.1 900.9 1183.1 0.4534 1.1675 1.6208 281.9 1102.1 80 90 320.28 0.01766 4.8777 4.095 2903 894.6 1185.3 0.4643 1.1470 1.6113 293.4 1103.7 90 0.4743 1.1284 1.6027 298.2 1105.2 100 4 100 327.82 0 01774 4.4133 4.431 298.5 808.6 1187.2 i 120 34127 0.01789 3.7097 3728 312 6 877A 11934 0 4919 1.0960 1.5879 312.2 1107.6 120 340 353 04 0 01803 3.2010 3 219 325.0 868.0 1193.0 0.5071 1.0681 1.5752 324.5 1109.6 140 360 363 55 0.0;815 2.8155 2A34 336.1 859.0 1195.1 0.5205 1.0435 1.5641 335.5 1111.2 160 380 373 08 0 01827 2.5129 2.531 346.2 850.7 1196.9 0 5328 1.0215 1.5543 345.6 1112.5 180 200 351.80 0 01839 2.2689 2.287 355.5 842 2 1198.3 05438 1.0016 1.5454 3542 1113J 300 i 250 40097 0 01865 1.8245 1A432 376.1 825 0 1201.1 0.5679 0 9585 1.5264 3753 1115.8 250 ) 808.9 1202.9 0.5682 fl.9223 1.5105 392.9 1117.2 300 J 300 41735 0 01859 1.523B 1.5427 394 0 43133 0 01913 1.3064 1.3255 409 8 7942 1204 0 0 60M 0 8909 1.4968 406.6 1118 1 350 350 422.7 111E 7 400 400 444 60 0.0193 1.14162 1.1610 424.2 760 4 1204.6 0 6217 0 8630 1.4647 450 456 78 0 0195 1.01224 .l.0318 437.3 767.5 1204.8 0 6360 0.8378 1.4738 4353 1118.9 450 0 90787 0 9276 449.5 755.1 1204 7 0 6490 0 8149 1.4639 447.7 1118 8 500 500 4(7 01 0 0199 550 47694 0 0199 0 82183 0 8412 4609 743.3 1204 3 0 6611 0 7936 1.4547 456.9 1118 6 550 J 403 48520 0 0201 074962 0.7693 471.7 732.0 1203 7 0.6723 07738 1.4461 469.5 li1E. 2 500 063505 06556 491.6 710.2 1201.8 0 692R 07377 1.4304 488.9 1116 9 700 l 703 .503 08 0 0205 833 518 21 0 0209 0.54809 0.5690 509.8 689 6 11994 07111 03051 1.4163 506 7 1115.2 803 ! 900 132 93 0 0212 04796S 05009 526 7 669 7 1196 4 07279 06753 1.4032 5232 1113 0 900 0 4460 542.6 f 50 4 1192.9 03434 06476 1.3910 53";6 1110 4 1003 3000 544.% 0.0216 0 42436 0 4005 557.5 631.5 1169.1 0 7578 0.6216 1.3794 553.1 1107.5 1100 3100 $t! 2/ 0 0220 0 37863 556 9 1104.3 1200 i 3230 267.19 0 0223 0 34013 0.3625 571.9 613 0 1184 8 0.7714 0.59 9 1.3653 0.3299 585.6 594.6 1180 2 03843 0 5733 1.3577 580.1 1100 9 1300 1903 57742 0 0227 0 30722 1 0 3018 59B B $76 b i175.3 0.7966 05507 1.3474 592.9 1097.1 1400 14CD 557 07 0 0731 0 27871 3500 59020 0 0235 025372 0 2712 611 3 558 4 1170 1 0.8035 0 5253 1.3373 6052 1093.1 1500 2000 635 b0 0 02's? O16766 0ISS3 672 3 4657 1133 3 086M 04256 17881 662 6 10GS 6 2000 { 716 5 1032.9 2500 2500 (is l1 0 02rl 0 10204 0 1307 731 3 3616 1093 3 C 9139 0 3206 1.2345 3003 0 050/3 0 0850 801 8 216 4 10703 0 9728 C IEDI 1.1619 1822 973.1 3003 695 33 0 0343 3298 2 70; 47 0OWB 0 0 050d 906 0 0 936 0 10512 0 1.0612 675.9 875 9 37081 l _= , TABLE A.3 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (PRESSURE) A.4 j
32mpeestm, F
, Abe pcs.
b/agla. 900 400 800 600 700 000 000 ID: 1100 1200 1300 1400 3500 (set, temp) 100 200 h e 00361' 392 5 452.3 511.9 S71.5 631.1 690 7 3 6 68 00 1150 2 11957 1241 A 1288 6 3336 1 1984 5 (101.74) s 01295 2.0b09 2.1152 21722 2.2237 2.2708 2J144 o 00161 76 34 90 24 102.24 314.?! 126 15 138 08 150 01 161.94 173 86 18578 197.70 209 62 221.53 233 45 6 a 68 On i14C 6 1144 8 1241.3 1286 2 1335 9 1384 3 1433 6 1863 7 1534 7 1586 7 1639 6 1693 3 17480 18035 (167.24) s 01795 1.8716 1.9369 3.9943 2.0460 2.0932 23369 2.1776 2 2159 2 2521 2.28 % 2.3194 2.3509 2.3811 24101 e 0 0161 38 B4 44 93 SI 03 67.04 63 03 69 00 74 9B 80 94 86 91 9287 98 84 104 80 110 76 116 72 30 6 68 02 1146 6 11937 1240 6 1207.8 1335.5 13840 1433 4 1483 5 1534 6 ISM 6 16395 1693.3 1747.9 18;34 (192.21) s 0 1295 1.7926 1.8593 1.9173 1.9692 2 0166 2.0603 2.1011 2 1394 2 1757 2.2101 22430 2.2744 2.3046 2.3337 e 0 0161 0 0166 29899 33 963 37.985 41.9F6 45.978 49 % 4 53 946 57.926 61 905 65882 69358 73 833 77207 36 6 68 04 168 09 31925 1239 9 1287.3 13352 1383 8 1433 2 1483 4 1534 5 1586 5 1639 4 1693.2 1747A 1803 4 (213.03) s 01795 0.2940 1.8134 1.8720 1.9242 1.9717 ,2.0155 2.0563 2JD946 2 1309 2.1653 2.1982 22297 2.2599 22890 e 0 0161 0 0166 22.356 25428 28 457 31.466 34 465 37.458 40 447 43 435 46420 49 405 S2 20 6 68 05 16811 1191.4 1239.2 1286.9 1334.9 1383 5 1432 9 1483 2 1534.3 1586.3 1639.3 1693.1 17472 1803.3 (227.96) s 01295 0.2940 1.7805 1A397 1A921 1.9397 3.9836 2.0244 2.0628 2.0991 2.1336 2.1665 2.1979 2.2282 2 2572 e 0.0161 0 0166 31 035 12.624 14.165 15 685 17.195 18 699 20 199 21.697 23 194 24689 26 183 27.676 29.168 i 40 6 68 10 168 15 1186 6 1236 4 1285 0 1333 6 1382.5 1432.1 1482.5 1533.7 1585 8 1636 8 1992 7 1747.5 1803 0 (267.25) s 0.1295 0 7940 1.6992 1.7608 1A143 13624 1.9065 1.9476 1.9660 2.0224 2.0569 2.0899 2.1224 2.1516 2.1837 e 0.0161 0 0156 7.257 8354 9.400 10 425 11.438 12 446 13.450 14 452 15.452 16.450 17A48 18.445 19 441 60 6 68 15 16920 1181 6 1233.5 1283.2 1332.3 1381.5 1431.3 1481.8 1533.2 1555.3 1638 4 1692.4 1747.1 18022 (292.71) s 0.1295 0.2939 1.6492 1.7134 1.7681 1A168 13612 1.9024 1.9410 1.9774 2.0120 2.0450 2.0765 2.1068 2.1359 e 0.0l61 0 0166 0.0175 6.238 7.038 7.794 8.560 9 319 10.075 10.829 11 581 12.331 13.081 13.829 14 577 80 6 68 21 168 24 269 74 1230.5 12bl 3 1330.9 13805 1430.5 1481.1 1532 6 1584.9 1638 0 1692.0 1746.8 1832.5 (312.04) s 0.1295 02939 0.4371 1 6790 1.7349 1.7842 1A289 1 8702 1.9089 1.9454 1.9800 2.0131 2.0446 2.0750 2.1041 e 0 0161 0.0166 0 0175 4 935 5 588 6.236 6 833 7.443 8050 8655 9258 9 860 10 460 11.060 11A59 100 6 6E 2C 168 29 269 77 1227.4 1279.3 1329 6 1379 5 1429 7 14B0 4 1532 0 1564 4 1637.6 1691.6 1746.5 1802.2 (327.62) s 0.1295 0.2939 0 4371 1.6516 1.70ES 1.7586 1.8036 13451 1A839 1.9205 1.9552 1.9883 2 0199 2.0502 2.0794 e 0 0161 0 0165 0 0175 4 0786 4.6341 S.1637 56831 6.1929 6 7006 7.2060 7.7096 8.2119 8.7130 9.2134 9.7130 120 A 48 31 168 31 269 81 1224.1 1277.4 1328.1 1378 4 1428.8 1479.8 1531.4 1583 9 1637.1 1691.3 17462 1602D (341.27) s 0.1295 0 2939 04371 1.6286 1A872 1.7376 1.7829 12246 1A635 1.9001 1.9349 1.9680 1.9996 2.0300 2&592 e 00161 0 0166 0 0175 3 4651 3 9526 4 4119 4A585 S 2995 S.7364 6 1709 66036 7J0349 7A652 7A946 83233 148 6 68 37 168 38 26S85 1220 8 1275.3 1326 8 1377.4 1428 0 1479 1 1530 8 1583 4 1636 7 1890.9 1745.9 1801.7 l (353 04) : 0 1295 0 2939 0 4370 14055 1.6teb 1.7196 1.7652 1.8071 12451 12323 1.9175 1.9509 1.*825 2.0124 7 647) e 0 0161 0 0166 0 0175 3 0060 3 4413 3.8480 4.2420 4 6295 S.0132 S 3945 S.7741 6 1522 6 5293 '6.9055 7A811 360 6 68 42 168 42 269.89 1217.4 1273 3 1325 4 1376 4 1427.2 1478 4 1530.3 1582.9 1636.3 1990.5 1745.6 1801.4 (363 55) s 0.1294 0.2938 0 4370 1.5906 1.6522 1.7039 1.7499 1.7919 1A310 1A678 1.9027 1.9359 1 A676 - 1.9980 2.0273 e 0 016! 0 0166 00!?4 26474 3 0433 3 4093 3.7621 4.1064 4.4505 4.7907 S.1289 S4657 62014 6.1363 6.4704 180 6 6847 16647 260 9/ 1213 8 1271.2 1324 0 1375.3 1426 3 1477.7 1529 7 1582 4 1635 9 1640 2 1785.3 1801.2 (373.C81 s C 1294 0.2938 0 4370 1 5743 1.6376 1.6900 1 7362 1.7784 1A176 1.8345 1.8894 1.9227 1 9545 1.9649 2.0142 e 00!(I 00!66 00!?4 2 3598 2.7247 3.0583 3.3783 3 6915 4.0008 4 3077 4.6128 4.9165 52191 S.5209 SA219 200 > 68.52 168 51 269 96 12101 1269 0 13221 1374.3 1425 5 1477.0 15291 1581.9 1635 4 1689 8 1745.0 1800 9 i (3Bi&O) s 01294 0 2935 0 4359 1.5593 1.6242 1.6776 1.7239 1.7663 13057 1.6426 1.8776 1.9109 13427 1.9732 2A025 e 0 0161 0 01E6 0 0174 0 0166 2.1504 24662 24872 2.9410 3 1909 3 4382 3 6837 3 9278 4 1709 4 4131 4 6546 68 65 166 63 270 05 3/510 1263 5 1319 0 1371.6 1423 4 1475 3 1527.6 1580 6 1634 4 1688 9 1744 2 1800.2 ! 250 e (400 97) : C 1294 02937 0 4366 0 5567 1.5951 11502 1 6976 1.7405 1.7601 1.8173 1.8524 1.8d58 1 9177 1.9482 1.9776 e 0 0161 001ES C 0174 0 0186 1.7655 2.0044 2.2263 2.4407 2.6509 2 6585 3 0643 3.2688 3 4721 3.6746 3 8764 300 6 6679 !$6 74 ??u ld 375.15 1237.7 1315 2 1368 9 1421.3 1473 6 1526 2 1579 4 1633 3 1688 0 1743 4 1799.6 (417.35) s 01294 0 2937 0 4.M7 C5%5 1.5703 1.6214 1.6758 1.7192 1.7591 1.7964 12317 1.8652 13972 1.9278 13572 e 0 0161 0 0166 0 0174 0 0186 1.4913 3.7028 13970 2 0332 2.2652 2 4445 2.6219 2.7980 2.9730 3.1471 3 3205 350 a 68 92 16985 27024 37521 1251.5 1311 4 13662 1419 2 1471 6 1524 7 1578.2 1632.3 1667.1 1742 6 1798 9 0 2935 0 4357 0 5644 1.5483 1.6077 1.6571 1.7009 1.7411 1.7187 1.8141 13477 1379S 1.9105 1.9400 (431.73) n 01293 l v 0 0161 0 0166 0 0174 0 0162 1 2841 1.4763 1 6499 1.8151 1 9759 2 1339 2 2901 2.4450 2.5997 2.7515 2 9037 400 m (9 05 16E 47 270 33 375 27 124S 1 1307.4 130 4 le 17.0 1470 1 1523 3 1576 9 1631.2 16562 17419 1793 2 (44440) s 51293 02935 0 4365 0 % 63 1.5282 1 5901 1 6406 14850 1 7255 1.7632 1.79BB l.8325 13647 1.8955 1.9250
! e 0 0161 0 0166 0 0174 0 0lE6 0 9919 1.1584 13037 1.4397 1.570B 1 6992 187tt 1.9507 20746 2 1977 2.3200 S00 > 69 37 1 % 14 270 51 3tL 3R 12312 12991 13573 le17 7 1466 6 1520 1 15744 16291 1(A4 4 1740 3 179E9 ; (467.011 s 0I?97 6 2934 04E4 0 M60 149?! 15595 1 U23 1(5/8 1(990 1 7371 17733 180f 9 11393 18702 11993 TABLE A.4 PROPERTIES OF SUPERHEATED STEAM AND COMPRESSED WATER (TEMPERATURE AND PRESSURE)
A.L
.~ _- _ . - - - - - _ - - _ _ - - . - = - . _ - - - - - . . _ - - .-
h
- w,,.m. w ... ,
! . .C/94 b (eAl. tsmp) 300 200 800 400 600 600 700 800 900 1000 1100 12 % 3300 1400 1500 o 00161 0 01(4 0 0174 00186 07944 0 94 % 10726 11892- 1.3006 14093 1.5160 16711 1.77b? I8764 3.9309 680 6 69.58 169 42 270 70 375 49 3215 9 1290 3 1851 8 1408 3 1463 0 1517.4 1571 9 3677.0 16826 3738 8 1795 6 ge8620) s 01292 0.2933 04362 0 % 57 14590 1.63?9 1.M44 3.63bl 16769 17155 8.7517 17859 13184 laos 1 A792 e 00161 0 0166 0 0174 0 0186 0 0704 0 7928 0 9072 1.0102 1.1078 17023 1 7948 1.3858 1.4757 1 % 47 1 6530 ! 780 6 6904 169 65 270 89 375 61 487 93 1281 0 1345 6 14037 1459 4 1514 4 1%94 16748 M60 7 17372 1794 3 (503C8)s 0 1291 02932 04MO 0 % 55 0 6609 1.5090 3.% 73 1 6154 3.6580 16970 173h 17676 18003 18318 18617 e 0 0161 0 0166 0 0174 0 0186 0 0704 0 6774 0 7879 0 8759 0 9631 I N70 1 1289 12093 1.2875 1 3669 1.4446 300 6 70 11 169 86 271 07 375 73 487M 1271 1 13392 13991 1455 R 1511 4 1 % 6 9 .167? 7 167E 9 17h 0 1792 9 0182.) . 0.1290 0 2930 0 4358 0.E52 0 6885 14E69 1 5484 1.% 50 16413 I M07 1.7375 17522 1.76bl 1 8164 14464 e 0 0161 0.0166 00174 0 0186 0C204 05869 0 6E% 07713 0 8504 092(2 0 99M 10720 1.1430 1 2131 1.2625 ! 300 6 70 37 170 10 271.26 37524 487A3 1260 6 1332 7 1394 4 1452.2 1506 5 1%44 3620 6 16771 17341 1791 6 i (531.95) s 0 1290 0.2929 04357 0 % 49 0 6881 1.4659 1.5311 1.5822 16263 1 M62 1.7033 3.7382 3.7713 14028 33329 1
e 00161 00166 0.0174 0 0186 0 020* O$137 0 6080 0 6875 0 7603 0 8295 0 89 % 09622 1.07 % 1.0901 1.1529
$800 6 70 63 170 33 271.44 375 96 487.79 1249.3 .3325.9 1389.6 1448 5 1504.4 1 % 1.9 1618 4 1675 3 1732 5 1790 3 (544.58) s 0.1269 0.2928 0.4355 0.5647 04876 1.4457 1.5149 1.5677 14126 36530 14905 1.7256 3.7589 1.7905 14707 i
e 00161 0 01 % 00174 0.0185 0.0203 04531 0 5440 0 6188 06865 0 7505 0 8121 0 8723 0 9313 0 9894 1.0468 3500 4 70 90 170.56 27143 376 08 467.75 1237.3 1318 8 1384 7 1444 7 1502 4 1559 4 1616 3 1673.5 1731.0 1789.0 SM23) s 0.1269 0.2927 0 4353 05M4 0 6872 1.4259 1.4996 1.5542 16000 1.6410 14787 1.7141 1.7475 1.7793 13097 e 00161 0 01 % 0.0174 0 0185 0 0203 0 4016 0 4905 0.% 15 0 6250 0 6845 0 7418 0.7974 08519 0.9055 0 9584 8200 4 71.16 170.7 A 27122 376.20 437.72 1224 2 1311.5 13797 1440 9 3449 4 15%9 1614.2 16714 1729 4 1787.6 (567.19) s 0.1288 0.2926 0.4351 0.5642 0 6868 1.4061 1.4851 13415 1.5883 14298 1 % 79 1.7035 1.7371 1.769) 3.7996 4 e 0 01(.I O0166 00174 0 0185 0 0203 0 3176 0 4059 0 4712 0 5282 0 5809 06311 0 6794 0 7272 0.7737 0 8195 ? 1400 4 71 68 17124 272.19 376 44 487.65 1194.1 12961 1369 3 14332 1493 2 15518 1609 9 1%80 1726 3 17850 ! 08?A7) s 0.1287 02923 04MB 0.5636 0 6659 IJ652 1.4575 1.5182 1. % 70 140M 14484 14645 1.7185 1.7508 1.7815 1 i e 00161 0 01 % 0.0173 0 01A5 0 0202 0 0236 0.3415 0 4032 0 4555 0.5031 0 5482 0 5915 0 6336 06748 0.7153 l 1600 6 72.21 171.69 272.57 376 69 487 60 61( 77 1279 4 13565 1425.2 14EC 9 1546 6 1605 6 1%43 1723 2 17E2.3 (% 4 87) s 0 1256 0 2921 04M4 0 M31 0 6651 0.6129 14312 1.4963 1.5476 1.5936 1 6312 1.6676 1.7022 1.7344 1.7657
. e 0 0160 0.0165 0 0173 0.0185 0 0202 0 0235 0 2906 0 3500 0.3988 0 4426 0 4836 0 5229 0 % 09 0 5980 0 6?43 3800 a 12.73 172.15 272.95 376 93 487.56 615.58 1261.1 1347.2 3417.1 1480 6 1541.1 1601.2 1%07 1720.1 1779.7 4621/3F)a 0.1284 0.2918 0.4341 0.5626 0 68'3 0.8109 1.4054 1.4768 1.5302 1.5753 1.61 % 14528 14876 1.7204 1.7516 e 0 0160 0.016% 0.0173 0.0184 0.0201 0 0233 02488 0.3072 0.3534 0.3942 0 4320 0 4680 0.5027 0.5365 0 5695 3000 t 7326 172 60 273.32 377.19 487.53 614 48 1240.9 1353 4 1408 7 1447.1 1536 2 1596.9 1657.0 1717.0 1777.1 l (635 60) s 0.143 v.2916 0 4337 0 % 21 OssM 0.8091 1.3794 1.4578 1.5138 1.% 03 1.6014 14391 1.6743 1.704 1.7369 e 0 0160 0.0165 0.0173 0.0184 0.0200 0 0230 Ol681 0.2293 02712 0.3068 0.3390 0 3692 0.3980 0 4259 04529 i 9800 6 74.57 173 74 274.27 377.87 487.50 612.08 1176 7 1303 4 3386.7 1457.5 1522.9 1585 9 1647A 1709.2 3770 4 I
(688t!)s 0.1280 0.2910 0 4329 0.W0' 3 6815 0 8048 IJ076 1.4129 1.47 % 1.5269 1.5703' 1A094 14456 1.6796 1.7136 e Q 0160 0 0165 0 0172 0 0183 0 0200 0 0228 0 0982 0 1751 0.2161 02484 0.2770 0 3033 0 3282 0 3522 0.3753 3000 A 7583 17#88 275.22 378 47 467.52 610 06 1060 5 1267.0 33632 3440.2 15014 1574.8 1635 5 1701 4 17f 1.8 (69L13) s 0.127? 0.29a4 0.4320 0 5597 0 6796 0 8009 1.1966 1.3692 1.4429 1.4976 1.5434 1.%41 1.621/ 14 61 1 M88 1 e 0.0160 0 0165 0 0172 0 0183 0 0199 0 0227 0.0335 0.1588 01987 0.2301 0 2576 0 2827 0.306% 0.3291 0.3510
- - 3200 h 76 4 1753 27b 6 3787 487.5 609 4 800 8 1250 9 3353 4 1433.1 1503 8 1570.3 16343 16983 1761.2 0 05.08) s 03276 0 2902 0.4317 0 5592 06768 0.7994 0 9708 1.3515 1.4300 1.48 % 1.5335 1.5/49 1Al?6 1.6477 14806 e 0 0160 0 0164 00172 0 01E3 0 0199 0 0225 0.0307 0.1364 01764 0 2066 02326 0 2563 0.2784 02995 0.319?
I 9600 4 77.2 176 0 276 2 3791 487.6 6064 779 4 1224 6 13382 1422 2 14955 IM33 16292 1693 6 17b7.2 j e 0.1274 0.2899 0 4312 0 5585 0 6777 0 7973 09508 1.3242 1.4112 34709 1.5194 3.% 18 1.6002 3.6355 1.6691 ! e 0 0159 0 0164 00172 0.0182 0 0198 0 0223 0 0287 01052 0.1463 0.1752 0 1994 02210 0.2411 0.2601 02783
. 4800 A 7e.5 177.2 277.1 3798 487.7 606 5 763 0 1174.3 1311.6 1403 6 1481.3 1552.2 1619 8 1665 7 1750.6 l
e D.1273 02993 0.4304 0M73 06760 0 7940 09343 1.2754 IJ807 3.4461 1.497G 1.5417 1.5812 14177 3.6516
- e 0 0159 0.0164 00171 0 0181 0.0196 0 0219 0 0268 0.0591 0 1038 0.1312 0 1529 0 1718 0 18*0 0 7050 0.2203 5000 a 81 3 179 5 2791 381.2 4881 604 6 746 0 1042 9 1252.9 33(4 6 14521 15291 16009 1670 0 1727.4 l s 0.1765 026b1 0 4267 0.5550 0 6726 0.7880 0 9153 1.1593 1.3207 3.4001 1.4582 1.5061 15481 1.5E63 1A216
- 0 0159 0.0163 0 0170 0 0lbu 0 0195 0 0216 0 0256 0 0397 0 0757 0 1020 0 1221 0.1391 0.1544 01684 0 1817
! 4000 4 63 7 181.7 281.0 362 7 AP8 6 602 9 7361 9451 llE8 8 1323( 1422 3 1505 9 1%20 16542 1724? e 0 1258 0.2670 0 4271 0 5528 06693 0 7826 0 9026 1.0176 1.2615 1.3 5 N 1.4229 1.4745 1.5194 1.5593 15962 o 00158 0.0163 0 0170 0OlF0 0 0193 0.0?!3 0 0248 0.0334 0 0573 0 0316 0 1004 01160 0 !?St 01424 0.1542 , 7CCD A 862 IP4 4 283 0 384 2 489 3 601 7 729 3 901.8 3124 9 12P) 7 1392 2 1492 6 1%31 1639 6 1711 1 j ,_ s 01252 0Pf59 04256 05'07 0 6563 0 7/77 0 8976 10350 12055 1.21ii 1 1934 144e6 14935 153-5 157JL TABLE A.4 PROPERTIES OF SUPERHEATED STEAM AND COMPRESSED l WATER (TEMPERATURE AND PRESSURE) (CONTINUED) A.0
amepr.swrc.r a, .., ,, ... . ,, ,, ,, ,, ,, . ,, ,, ,,
,. < rTI,kl.h U) M.
, n . ~ .. -
4/M,r, /A / N // N '* A.r Fi I nobo b bbY / I
/ F
/ N/
s / IAL I.
//% I nno ;
/// p /N / 1
,., M IA) % I7s /N/ Ti///W r I anb'
~
/)MT [ / ,
,. /II /n2N /
ki M f/I% i > <1 i
)im
/JM'ff V
f
. , . HHCl H R) K /
N//IW r M I J I hr
> IE
/$EIl1W'I N /MJ
/ "
IfEGKh% y' /75, ~ 1 Mhfs75MMW 7 7 s wwwh;%27M ygnyj 79 I._ .
'~
1 ' l _ WBMK I !! / ff f . V // W8 / / /
/4' N 7 SM h/ /
dW ~ i seso _ MMN7#4 ~
. MNNMM N ~
UMXMM47Y
~
A y ~
%kV6M.7/ ~
.JW M W M 30 1.1 1.7 3.3 34 3.$ 34 3.y
,t atropy. Stu/h. F 18 13 2.0 2.5 22 FIGURE A.5 MOLLIER ENTWJ.,PY-ENTROPY DIAGRAM A.7
1 i l PROPENTIES OF WATER 1 Density c (Ibsitt') PSIA Temp Saturated _ 2300 240D 2500 8000 Liquid 1000 2000 2100 3200 (* F) '
- 62.909 62.93 62.951 63.056 62.637 62.846 62.667 62.888 32 62.414 62.846 62.87 62.99 i 62.75 62.774 62.798 62.822 60 62.36 62.55 62.465 62.559 62.409 62.427 62.446 61.989 62.185 62.371 62.390 100 60.568 60.587 60.606 ' 60.702 00.314 60.511 60.53 60.549 200 60.118 67.859 67.882 67A98 67.767 67.79 67.813 57.836 300 67.310 57.537 64.373 64.529 64.249 M.28 64.311 M.342 400 63.651 63.903 64.218 64.11 l 63.86 63.89 63.925 63.95 63.248 63.475 63.79 63.825 410 63.46 63.50 63.63 63A9 63.025 63.36 63.40 63.425 420 62.798 63.065 63.09 63.265 62.675 62.925 62.95 62.99 63.02 430 62.356 62.64 62.56 62.275 62.42 62.45 62.475 62.51 440 61.921 62.125 62.21 62.41 62.065 62.10 62.14 62.175 450 ' 61.546 61.66 62.025 61.96 61.64 61.66 61.725 61.76 61.020 61.175 61.56 61.61 450 61.22 61.25 61.30 61.50 60.70 61.1 61.14 61.175
' 470 60.505 60.76 60.625 51.035 60.62 60.66 60.7 60.74 480 60.00 50.20 60.35 60 575 60.175 60.22 60.265 60.31 4DO 49.505 49.685 60.13 60.098 49.714 49.762 49.81 49.858 48.943 49.097 49.618 49.666 500 49.203 49.254 49.305 49.56 48.51 49.05 49.101 49.152 610 48.31 48.68 48.735 49.01 48.46 48.515 48.57 48.625 620 47.85 47.91 48.155 48.45 47.919 47.970 48.037 46.006 630 47.17 47.29 47.56 47A9 47.362 47.428 47.494 47.56 l 46.51 47.23 47.296 640 46.794 46 862 46.93 47.27 i 46.59 46.658 46.726 650 45.87 46.216 46.29 46.66 45.92 45.994 46.068 46.142 660 45.25 45.64 45.62 46.02 45.22 45.30 45.38 45.46 670 44.64 44.844 44.93 45.36 44.50 44.586 44.672 44.758 680 43 66 44.11 44.205 44.66 43.73 43.825 43.92 44.015 650 43.10 43.33 43.434 43.956 42.913 43.017 43.122 43.226 600 42.321 42.432 42.55 43.14 41.96 42.08 42.196 42.314 610 41.49 41.483 41.616 42.283 40.950 41.083 41.217 41.35 620 40.552 41.44 630 39.53 40.388 640 38491 39.26 650 37.31 38.000 660 36.01 36.52 670 34.48 34A98 680 32.744 . 32.144 690 30.516 TABLE A.6 PROPERTIES OF WATER, DENSITY
- A.8
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 24
~
~~~~Th5E566 N555657~555T TR5U5E5E~5U6~ELU56~ELUU ANSWERS ---SEQUOYAH 1&2 ~86/05/26-D.J. NELSON J
AMSWER 1 01 (1.00) a 4 REFERENCE SON /HDN Nuclear thacry ANSWER 1.02 (1.00) b. REFERENCE SON /WBN HTFF Chap. 2E ANSWER 1.03 (1.00) a. REFERENCE SON /WBN HTFF ANSWER 1.04 (1.00) b. REFERENCE SON /WBN thermo ANSWER 1.05 (1.00) i c. REFERENCE
- r. t e a m tabler
F
- 1. PRINCIPLES OF NUCLEAR POWER PLANT UPERATION, PAGE 25
--~~isEER557sARIEs- sEAi isAs5FEs Es5 FE5i5 FE5s ANSWERS -- SEQUOYAH 1&2 -86/05/26-D.J. NELSON ANSWER 1.06 (1.00) c REFERENCE Elli: GPNT,Vol VII, Chapter 10.1-03-86 BSEP: L/P 02-2/3-A, pp 172 - 176; 02-0G-A, pp 57 - 60
' Westinghouse Nuclear Reactor Theory, pp. I-5.77 - 79 Turkey Point, Reactor Core Control, pp. 4 20 001/000-K5.13 (3.7/4.0)
ANSWER 1.07 (3.00) AFFECTED LOOP + UNAFFECTED LOOP c f (ce*'*i N T'4 4
- a. decreases + increases 0.53 each
- b. decreases + increases to.53 each
- c. decreates + increases E0.53 each REFERENCE SON /WDN HTFF/ Nuclear theory ANRWFR 1.00 (1.50)
- a. increasing CO.5]
- b. decreasing CO.5]
- c. decreasing [0.53 REFERENCE SON /WBN Nuclear theory ANSWER 1.09 (1.50)
- a. Less negative (0.5)
- b. Less negative (0.5)
- c. Less negative (0.5)
REFERENCE Turkey Point, Reactor Core Control, Chapter 5, Fig. SNP-RF-9 004/0003 K5.06(3.0/3.3) I l
9 .
- 1. PRINCIPLES OF NUCLEAR POWER PLANT =0PERATION, PAGE 26
'~~~~ ~ ~
TEER5667 555657~5EST TR5U5FER 5U67FL656~FL6U ANSWERS -- SEQUOYAH 182 -86/05/26-D.J.' NELSON ANSWER 1.10 (1.00) a, true CO.53
- b. true CO.53 REFERENCE Nuclear' theory, Inst. Notes VI ANSWER 1.11 (1.00) 3
- a. FALSE CO.53
- b. reference les draining voids in reference les elevated reference les temperat'ure (two required) CO.253 each REFERENCE SON /WBN SD AMSWER 1.12 (2.00)
- a. Decrease (+.5 on)
- b. . Increase
- c. Increase
- d. Increase REFERENCE NUS, Nuclear Energy Trng.- Reactor Op., pp. 9.2-5, 11.3-2, 11.4-3 ANSWER 1.13 ( .50)
- 8. FALSE REFERENCE SON / WON Nuclear Theory Inst. Notes
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 27-
--~~isERR55isARICs- REAi iEAssFEE AR6 FCUi5 FC50 ANSWERS -- SEGUOYAH 182 -86/05/26-D.J. NEL50N ANSWER 1.14 (1.00)
-Hinimizes thermal stress due to more uniform temp difference of fluids
-The outlet temp of the colder fluid approaches the inlet temp of the hotter fluid
-A'more uniform heat transfer rate is achieved throughout the heat exchanser (+.33 ea)
(more efficient is an acceptable response also) REFERENCE CNTO, " Thermal / Hydraulic Principles and Applications', pp 5-10 004/020; K5.02(2.5/2.9) 1 ANSWER 1 15 (1.50)
- a. increases due to decreasing Tave E.53
- b. increases due to decreasins Tave E.53
- c. decreases due to decreasing Tave and increasing baron concentration E.53
' REFERENCE SON / WON Nuclear theory ANSWER 1.16 (1.00)
Delta I becomes more positive due to the increase of neutron flux in the
. top of the core relative to the bottom.
REFERENCE SGN/WDN Nuclear theory ANSWER 1.17 (1.00) Lower flow at the same power level results in a larger delta T8 CHF-decreases toward the top of the core. Lower coolant velocities result in less stripping action which removes nucleate bubbles, a steam film can form at lower heat flux. REFERENCE SON / WON HTFF
o .
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 20
~
~~~~YUEE566EU555C5I~5E5T ~TE5U5E5R 506~ELUI6~FL50 ANSWERS -- SEQUDYAH 182 -06/05/26-D.J. NELSON ANSWER 1.10 ( .50)
Margin to criticality decreases E0.1] by 2/3 E0.4]. REFERENCE SUN /WDN Neutron Sources and Subcrit. Mult. ANSWER 1.19 (1.00) Since samarium is a stable isotope both the production rate and removal rate are proportional to power level. REFERENCE West. Nuc, Trna. Ops., p.I-5.77 ANSWER 1.20 (3.00) (three req'd) ,
- 1. Delta T across the core E0.5]; constant / decreasing E0.4] and less than SS,P Jul' im -f -4. l t ? . [ 0 .13-
- 2. Core outlet temp.CO.5]; constant /decreasingEO.53 (o.]
- 3. Teoid = Tsat for S/G PressureEO.5]; constant or decreasing E0.5]
- 4. SC pressure E0.5]; decreasing E0.53
- 5. That [0.5]; constant or decreasing E0.53 REFERENCE WTSi Ch. 14, p. 27; ADI 35 ANSWER 1.21 (1.00)
In a small LOCA core heat is not being removed sufficiently by the break cnd little ECCS flow is being delivered due to elevated RCS pressure. REFERENCE SON /WDN SD; Aux Feed Sys; p. 0.f0
+ ,
+
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 29
--- isEsR557sAsiEs? REsi isXssFEE AR5 FEGi5 FE55 ANSWERS -- SEQUOYAH 1&2 -86/05/26-D.J. NELSON ANSWER 1.22 (1.50) 0=mcDelta T(+.5) .02/.0=(m/100)(40/40) (+.5) >>> m= 6/200 x 100%= 3%(+.5)
REFERENCE SON /HDN HTFF ANSWER 1.23 (2.'00) Tave! 30.4 X 0.25 X -15 = -114 pcm Powert 25 X -12 = -300 pcm Void! -25 pcm Xenon! -50 pcm totalt -409-pcm . Baron! -409 / -9 = 54.3 ppm dilution (accept 52 to 56) REFERENCE SON /HBN Nuclear theory 4 f i l-i i {
o .. i2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 30 ANSWERS -- SEQUOYAH 1a2 -86/05/26-D.J. NELSON ANSWER 2.01 (1.00) b. REFERENCE SON /WBN SD; RCS; CVCS West.-PWR Sys. Manual ANSWER 2.02 (1.00) B REFERENCE 10CFR50.46 ANSWER 2.03 (1.00) a. REFERENCE SON SDI RPS ANSWER 2.04 (1.00)
- a. reversed
'b. normal REFERENCE West. PNR Sys. Manual / SON /WDN HTFF ANSWER 2.05 (1.00) b.
REFERENCE System Manual, Chapter 3r pp 4, 10 09NP System Descriptions, CVCS
l l
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 31 l ANSWERS -- SEQUOYAH 1&2 -86/05/26-D.J. NELSON ANSWER 2.06 (1.00) b.
REFERENCE SONP Diesels handout, p. 6. 1 ANSWER 2.07 ( .50) 30 minutes REFERENCE Review of Elec. Distribution Lesson, p 10. I ANSWER 2.08 (1.00) c. REFERENCE , System Manual, Chapter 4, p. 4.0-2 l ANSWER 2.09 ( .50) True. l REFERENCE l Reactor Protection Lesson, p. O of 13, item d. l 1 ANSWER 2.10 (1 00) coldt last pump l
- REFERENCE SON / WON HTFF 1 l i
i i 1
=
- 2. PLANT DESIGN INCLU3ING SAFETY AND EMERGENCY SYSTEMS PAGE 32 ANSWERS -- SEQUOYAH if2 -06/05/26-0.J. NELSON ANSWER 2.11 2 ^ '; r
- a. CLOSE
- b. CLOSE
- c. OPEN
- d. CLOSE
;: "L p.+ +.a u
- g. CLOSE
- h. NO
- i. NO
- j. M d.(O S e# : to.2 ea.]
REFERENCE GUNP System Description, ECCS, CVCS, M N G'I M , CCW ANSWER 2.12 (1.00) Solenoid Airi open; 33%i 500 L4.25 ea3 A h REFERENCE A t.m F e e d L e u c o n Plan, p. S of G. (4NSWER 2.13 (1.00)
- 1. RHR suction (+.25 en up to 1.0)
- 2. RilR discharge
- 3. SI pump suction
- 4. 91 pump discharge
- 5. letdown ralinf valve also nrificloQ.c P p. wram s d,$
RL,FERENCE SON /WUN Innt. Notesi RCS
j 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 33 ANSWERS -- SEQUOYAH 1&2 -86/05/26-D.J. NELSON ANSWER 2.14 (2.00) Containment sump level and flow monitoring. Containment upper and lower radiation monitoring. I Containment temperature. Containment humidity. .ls , c,,)**wr LAa 'te ey4aat8 Containment pressure. Pressurizer level and pressure. 09 g/ 91a.l Volume control tank makeup. (eight required-CO.25 each]) . REFERENCE SON /WDN SDI RCS,CVCS, Cont. ANSWER 2.15 (2.00)
- 1. Manual emergency borate valve to blender. N d
- 2. Emergency borate valve in main control room. < w M"p .
- 3. Normal boration path to charging pump suction.
, 4. Divert charging pump suction from VCT to RWST.
- 5. Align the charging pumps to inject the BIT by openning the motor operated inlet and outlet valves.
REFERENCE SQh/WDN SDI CVCS i ANSWER 2 16 (1.00) crfrerbrd,a lm vesA hrsY ;<[50l Nwt b W*A 55Yr REFERENCE "1 I*
- b ** b ' / ' " * ' ' "/
West. PWR Sys. Manual ANSWER 2 17 (1.50) Summing Unite Rate Comparator and Variable Gain Unit (4.5 ea) REFERENCE SONP Rod Control Lessono pp 5 & 6 of 11. Systems Manual, Chapter 11.1r p. 11 1-63. 4 J
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 34
' ANSWERS -- SEQUOYAH 1&2 -06/05/26-D.J. NELSON ANSWER 2.10 (1.50)
Lo-Lo level in 2/4 S/Gs (+.3 ea) SIS Loss of both MFP , Loss of one MFP > 80% power LOSP REFERENCE Aux. Feed Lesson Plan, pp 4-6. ANSWER 2.19 (1.00) When there is low flow due to low RCS pressure (+.3), this will increase the flow to improve the cooling for the pump lower bearing (+.7) REFERENCE West. PWR Sys. Manual SON Lesson Plan 'RCS*, pp 22 ANSWER 2.20 (1.00) Each individual detector (both upper and lower) (+.5) is compared to the average of the som of the upper (or lower) detectors (+.5) ANSWER 2 21 (1.00) PZR continuous spray flow will equilize the boron concentration. REFERENCE SON /HDN SDI RCS ANSWER 2.22 (1.00) a, loop 4 hot leg
- b. either the relief volve downstream of the letdown orifices if it is unisolated, PORV on the PZR, or RHR suction reliefs.
REFERENCE SON Inst. Notest RHR L
O
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 35 ANSWERS -- SEQUOYAH 1&2 -86/05/26-D.J. NELSON ANSWER 2.23 (1.00)
Used during couldown because the steam space detector response is poor due to poor heat transfer. REFERENCE SON /WDN Inst. Notes; RCS ANSWER 2.24 (1.50)
- a. lithium
- b. hvdrazine
- c. hydrogen REFERENCE SON /WBN Inst. Notesi RCS, CVCS ANSWER 2.25 (2.50) 1.g a . air CO.5].
ab. spring and steam pressure [0.5]
- 2. Air accumulator E0.5]
- 3. High steam flow with low steam pressure or low low Tave. CO.5]
Phase B isolation ' 4 4 2. p Nr. [0.5] f(Lt Lt(LNCE SON /WBN SD; MS J
.. , v e
- 3. : INSTRUMENTSAND CONTROLS PAGE 36 i ____________________________
ANSWERS'-- SEQUOYAH 182- -86/05/26-D.J. NELSON ANSWER 3. 0 f- (1.00) a. i REFERENCE. SONP System Descrip. "RPS", pp 10 & RPS~Hechanical Logic Drawing 012/0003 K6.03 (3.1/3.5) ANSWER 3.02 (2.50)
- a. higher
- b. as is
- c. higher
- d. lower-
- e. higher ANSWER 3.03 (1.50)
- a. OP. Delta-T OT Delta-T
- 1. no change decreases 2.. no change decreases
- 3. no change- decreases E0.25 each3 (1.5)
'REFERENEE Sequoyah Technical Specifications P 2 2-10, B2-4,5.
- ANSWER 3.04 (3.00) a '. 1,3 e.
l' 1
- b. 1,3 f, 1,2,3,4
) c. 1,3 3 1 { d. 1,3 h. 5 E0.2] each of'15 responses ! REFERENCE .
- West. PWR Sys. Manual, 11.1 l
f ( l
.]
e .
- 3. INSTRUMENTS AND CONTROLS PAGE 37 ANSWERS -- SEQUOYAH-1&2- -86/05/26-D.J. NELSON ANSWER 3.05- (1.00) a) . Load rejection controller'(C-7; rate).- E0.53
.b ) Provides a blocking signal for-9 valves 1during cooldown.below the low-low Tave setpoint (i.e. uncontrolled cooldown). E0.53 REFERENCE Instructor Notesi Steam Dump Control System.
ANSWER 3.06 (2.00)
- a. high pressurizer level (letdown isolates):
b.. lou pressurizer pressure (rods drive Tave and-PZR level down) E1. 03~each. REFERENCE Channel Failure Handouti TAD ARCi 3,4-11 ANSWER 3 07- (3.00)
- a. 1. High-high level-(75%) in any S/G (P-14).
- 2. SIS.
.3. Rx trip with low Tavs-(554-F). CO.33*each] (1.0) i b.-(Gutti iiFF's Leip) liFRV's eiid bypess valves clossi MFW isolsticr.
I ~ valves close.'(condensate system recirc's to condenser) ~(1.0) ! Condensate oooster pump (selected for.P-auto) starts. (0.5)
- c. . (Because the suction valve opens)
- d. 100-psid decreasins; between No. 2 heater shell and MFP suction. (0.5) i i REFERENCE ,
Sequoyah System Description, Section 8, p 6,105 Annunciator Response Vol I,. tab 5, p 29. l L i i. i l l i t
m o , f , 30 INSTRUMENTS AND CONTROLS PAGE 38 ANSWERS -- SEQUOYAH 1&2 -86/05/26-D.J. NELSON ANSWER 3.08 (1.75) Permissives- 6, 10, 11, 12 are manually initiated (+.25 EA) Permissives 7, 8, 9 are automatic REFERENCE PLSi pp 7-8. ANSWER 3.09 (2.00)
- 1. two percent radial flux tilt
- 2. improper rod sequence
- 3. shutdown bank rorJs less than 220 steps
- 4. rods within a bank greater than 12 steps from the bank demand
- 5. rods greater than 12 steps from e ch other within a bank REFERENCE SON, SOI-55-1M4, XA-55-40, p 25 ANSWER 3.10 (3.00)
- 1. anticipated overspeed E0.5]; governor and intercept valves E0.5]
- 2. overspeed E0.5]; governor and intercept valves CO.5]
- 3. partial unloading protection E0.5]; intercept valves E0.5]
REFERENCE SON Inst. Notes; Turbine. Control, p.5 ANSWER 3.11 (1.73) A. 1. Over Temp delta T
- 2. Over Pwr delta T
- 3. Lou Tave intlk.
- 4. Low-low Tave intlk.
B. 1. Rod Control
- 2. Steam Dump Control
- 3. P:r Level Control E0.25] each REFERENCE SNP Exam Ganki 3-6.
1
7 O a a
- 3. INSTRUMENTS AND. CONTROLS PAGE 39 ANSWERS -- SEQUOYAH 1&2 -86/05/26-D.J. NELSON ANSWER 3.12 (2.50) decreasing: 2335 PORV closes (+.05 for setpoint, +.2 function) 2310 sprays start closing 2260 sprays closed 2250 variable heatert start to come on 2220 variable heaters full on 2210 low pressure alarm, backup heaters on 2210 Backup heaters on 1970 Low pressure SI block enabled 1970 Low pressure reactor trip .
1970 Low pressure SI REFERENCE SON SD; RCSrRPS ANSWER 3.13 (1.00) To ensure the regenerative heat exchanger always has RCS system-pressure in it to prevent flashing of high temperature water. REFERENCE SON /WBN SDi CVCS ANC!!ER 3.1^ (1.00) Turbine runback [0.53 to REFERENCE 1757.ne4EO o,c c . 53 .M 70
- F0%
SOI-5.1 & 6.1; p7. ANSWER 3.15 (1.00) Interlock to prevent openning of RHR when RCS pressure - 300 psig. REFERENCE tt. 0/, a.Jo st.d A > 7<m /sd - SON /WBN Inst. Notes; RCS accepQ also : COIR massu. pekeL i ps1. e
o . .
- 3. INSTRUMENTS AND CONTROLS .PAGE 40 ANSWERS -- SEQUDYAH 1&2 -86/05/26-D.J. NELSON ANSWER 3.16 (2.00) a) Surge tank vent valve shut 6 (+.5 ea) b) B/D diverted to Con DI c) Containment Ventilation Isolation d) Aux Blds isolation and Aux Blds emerg gas treatment starts REFERENCE Systems Manual; Radiation Monitoring, pp 2-25.
t.
F 1
- 4. PROCEDURES - NORMAL, ADNORMAL, EMERuENC( AND PAGE 41
~
~~~~R 5656L66iU5L UUUTRUL~~~~~~~~~~~~~~~~~~~~~~~~
ANGWERO -- OEQUDY Ali la2 -86/05/26-0.J. NELSON ANSWER 4~.01 (1.00) d. REFERENCE SONP E-0 pp. 2 -5 ANSWER 4.02 (1.00) b. REFERENCE SONP GOI-2r pg. 1 ANSWER 4.03 (1.00) a. REFERENCE SONP S0I-62.1Br pp. Or 9 ANSWER 4.04 (1.00) a. REFERENCE SONP AOI-20, pp. 10 .1:2 ANSWER 4.05 (1.00)
% C.*
REFERENCE SONP AOI-3D, p. 1 of 2
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 42
~~~~R5656LUU5EAL C5 sir 5t----~~~~~~~-----~~~~~~~~
ANSHERS -- SEQUOYAH 1&2 -86/05/26-0.d. NELSON ANSWER 4.06 (1.00) b REFERENCE SONP GOI-2, p. 16 ANSWER 4.07 (1.50) Reactor Coolant Loop Tref-Tauct Hi-Lo (+.25 ea up to six) PZR Level High, Backup Heaters On Reactor Coolant Loop Tavg/Tauct deviation High-Low Rod Withdrawal Stop due to OPDeltaT OPDeltaP High Flux at 103% C-5(15% impulse power) IR Instrument (20%) REFERENCE o c~ J *V Y" ADI-2A, p2 ANSWER 4.08 (1.00) To prevent ES'E actua, tion (+.7) on lo-lo S/G level (+.3) REFERENCE A N SONP GOI-ir p. 4; precaution T. ANSWER 4.09 (1.00)
. TRUE 14 %
- b. FALSE.
REFERENCE SONP ES-1.3 p. 1 of 41 ES-1.2 p. 1 of 3, App. A
. .. o
- 4. ' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 43
- --~~~~~~----------~~~~~~ ~~~~REDi5L5EiCEL E5UTR5L ANSWERS -- SEGUOYAH 1&2 -S6/05/26-0.J. NELSON ANSWER 4.10 (2.00)
- a. False. (+.5 ea)
- b. True.
- c. True.
- d. True.
REFERENCE SONP GOI-5A, pp 2 & 3 ANSWER 4.11 (2.50)
.a. 4
- b. 1
- c. 5
- d. 2
- e. 1 E0.53 each REFERENCE SON, GOI-5A, p 5-9 ANSWER 4.12 (1'.00)
PZR level (+.35) < 20 % (+.1)
< 50 % adverse containment (+.1)
RCS Subcooling (+.35) < 40 degrees F (+.1) HEFERENCE
'SONP ES-0.2, pp 3, 4 ANSWER 4.1.3 (1.00) a) 3 Rem (+.33 ea) b) 18.75 Rem c) 300 Mrem REFERENCE SON RCI-1, pp 7 J
f
c-
' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND
~
- 4. PAGE 44
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- - AHSHERS;-- SE000YAH 182 -R6/05/26-D.J.' NELSON 1.
ANSWER 4.14 (2.00)
- 1. Verify' Reactor Trip.
- 2. Verify Turbine Trip.
- 3. Verify Shutdown Boards Energized.
4' 4. Check if SI Actuated.
- 5. Verify.ECCS status.
- 6. Verify Cntmt._ Isolation.
.7. Verify MFW Isolation.
- 8. Verify AFW status.
9 -Verify CCS Pumps Running 10 . - Verify ERCH' Pumps Running.
- 11. Verify EGTS and ADGTS-Running.
,12. Check-Cntmt. press less than'2.81_psig.
~ 13 .' Check Tavs.
LREFERENCE i SONP E-0, pp'2-5 I ANSWER 4.15 ~(1.00) i- -1.' Rod control in manual (or individual-bank select) -and restore Tave.
- 2. If unable to.stop rod insertion, trip reactor.
CO.53 each REFERENCE , SGN.AOI-2Br'p 1 ANSWER- 4.16 (1.00) - a, to prevent thermal lock up.
- b. to prevent bank overlap _ malfunction.
! REFERENCE SON, GOI-1, p 3,4 I f t l i-j 1 J.
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- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 45
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ANSWERS -- SEGUOYAH 1R2 -86/OS/26-D.J. NELSON ANSWER 4.17 (2.00) A. Be in Hot Standby within one hour. (1.0) l l B. Reduce RCS pressure to less than 2735 psis within 5 minutes. (1.0) REFERENCE TS, p 2-1. , he+ Q ANSWER 4.18 ins (2.00)
- a. Normal readings on containment temperature, pressure, radiation, and sump level. [0.5]
Normal readings on Aux building radiation and ventilation monitoring. [0.25] Normal readings on Steam Generator blowdown and vacuum pump exhaust radiation. [0.25] q
- b. Due to the low RCS temreratures and pressures following a loss of secondary coolant (with the RCS intact), SI repressurization could lead to RCS overpressurization and pa ag . (PTS) E1.0]
~
REFERENCE SON E0P-2, p.5; E0P-1 p.4; E0P-0, p.7
. ANSWER 4.19 (3.50)
- a. power mismatch circuit E0.5]; auct. nuclear power goes high causing a large P-ref/ nuclear power mismatch A ,-Ja _ 'ac -- ,+ + ' ' - -i-"~.
*- C1.03
- b. 1. overpower rod stop
- 2. power mismatch (for rod control)
- 3. upper detector current comparator
-4. lower detector current comparator
- 5. comparator channel defeat
- 6. nuclear flux bistables (remove control power fuses) 4 req'd [0.53 each REFERENCE SON ADI-4D l
_ _ _ _ _ _ _ . . _ _ --__.___.J
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- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 46
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' ANSWERS -- SEQUDYAH 182. '-86/05/26-D.J. NELSON ANSWER 4.20 (1.00)
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