ML20236T026
| ML20236T026 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 11/19/1987 |
| From: | Burdick T, Damon D, Sunderland P, Vinnola A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20236S973 | List: |
| References | |
| 50-454-OL-87-02, 50-454-OL-87-2, NUDOCS 8711300242 | |
| Download: ML20236T026 (92) | |
Text
{{#Wiki_filter:_ _ _ _ _ _ _ _ _ _ 4 U.S. NUCLEAR REGULATORY COMMISSION y REGION III Report No. 50-454/0L-87-01 Docket No. 50-454/455 Licenses No. NPF-23; CPPR 131 Licensee: Commonwealth Edison Company 4: Byron Nuclear Station 4450 N. German Church Road Byron, IL 61010 Facility Name: Byron Station l Examination Administered At: Byron Station and Production Training Center Examination Conducted: November 2-4, 1987 i Examiners: - D. J.,/Damon Il[lkf7 Date A. J. Vinnola M b k fet. Date I1 6 7 G?udlRSMAd P. R. Sunderland Date n-m -67 Approved By: T. M. Burdick e bh-O Date Examination Summary i Examinations were administered on November 2-5, 1985 (Report No. 50-454/0L-87-02) j to five operator and one senior o)erator candidates J hesults: All candidates passed t1e examination, i 8711300242 B $54 PDR ADOCK O PDR V L__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
4 4 REPORT DETAILS 1.- Examiners D. J. Damon
- P. R. Sunderland A. J. Vinnola
- Chief Examiner
- 2. Exit Meeting On November 5, 1987, the chief examiner met with members of the plant staff to discuss generic findings made during the course of the examinations. Attendance included:
R. C. Ward, CECO, Services Superintendent R. Pleniewicz, CECO, Production Superintendent L. A. Sues, CECO, Assistant Operations Superintendent A. J. Chernick, CECO. Training Supervisor
'L. D. Bunner, Ceco, Training Instructor J. K. Heaton, CECO, Training Instructor P. R. Sunderland, NRC, RIII, Chief Examiner The following areas were discussed:
- a. Generic Strengths (1) The candidates were well trained. They took positive action in all situations, displaying good self confidence.
(2) The candidates displayed a positive approach toward the examinations and towards Byron Station,
- b. Generic Weaknesses None were noted.
- c. Simulator Operation The simulator performed all of our scenarios without any problems.
- 3. Examination Review Following are facility comments and their respective NRC resolutions:
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. . _ _ . . . - - . ~ - . -~
se REACTOR OPERATOR EXAM QUf.STION 1.14 Comment: Use of.the Mollier diagram to calculate the temperature and condition of the steam should be acceptable for answering the question. The answer key appears to require the enthalpy of the steam to receive full credit; whereas, the question only requires temperature and condition of the steam. Also, regardless of the method used to answer the question, inaccuracies from reading the Mollier diagram or interpolating the steam tables do not appear to be taken into account. Byron proposes that a correct answer be considered as "Superheated. 300'F ( 10"F)." NRC Response: Comment accepted. The answer key was modified to reflect the comment. QUESTION 2.01 Comment: Letdown rate to CVCS is controlled by PCV-CV131. Byron propotes that PCV-CV131 also be accepted as a correct answer. NRC Response: Comment accepted. PCV-CV131 will be accepted as an alternate answer. QUESTION.2.02 Coninent: No automatic action will occur upon a LO-2 RWST level alarm actuation. An SI signal must be present in conjunction with the LO-2 RWST level signal before the SIB 811A and B will open. In addition, a modification to the , CV system automatically closes CV8810 and CV8111 (CVPp recirc) on LO-2 RWST l level with an SI signal present. Byron proposes that this question be deleted due to the omission of the SI i signal in the question. ) NRC Response: Comment accepted. The question was deleted.
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QUESTION 2.05 Comment: The answer key is correct. Requiring the addition of the manual valve in the answer to part b. is excessive detail. Byron proposes that the reference to the manual valve be deleted from the answer key. NRC Response: Comment accepted. The manual valve was removed from the answer. QUESTION 2.06 Comment: This question is.not a unit specific question, yet answer number 5 requires the Unit One HI-2 level setpoint. The supporting reference material for the NRC answer does not list setpoints for the HI-2 S/G water level trip of the Turbine Driven Main FW Pumps and thus should not be required. The permissive signal (P-14) is synonymous with HI-2 S/G water level, and should also be accepted as a correct answer. NRC Response: Connent accepted. The setpoint in question was removed from the answer key and P-14 is accepted as an alternate answer. QUESTION 2.10 Comment: Question 2.10 specifically asks what components actuate on a Safety Injection. Valves CV8110 and CV8111 close on a lo-lo RWST level in conjunction with the SI signal and valves CV8114 and CV8116 cycle on wide range RCS pressure in ! conjunction with an SI signal. The question does not state that the 10-10 RWST level or the wide range RCS pressure interlocks are present. i Byron proposes that modulate not be required for valves CV8110, 8111, 8114, and 1 8116, in order to receive full credit. NRC Response: . i Conment not accepted. The candidate should know that these valves are ) modulated under certain conditions. \ _ ___ _______________-______a
i _ QUESTION 2.11 Comment: Part a. As stated in " System Description, CVCS," Chapter 15a, page 15, CV-8149A, B, and C, are also interlocked with the letdown isolation valves CV-459 and CV-460. If CV-459 or CV-460 closes, then CV-8149A, B, and C will automatically close. Byron proposes that the closure of CV-459 or CV-460 be accepted as a correct answer. Part b. .Another purpose of PCV-CV131 is to control RCS pressure when solid and on RHR, as referenced by " System Description, CVCS," Chapter 15a, page
- 65. Byron requests that this purpose not be considered an incorrect answer.
NRC Response: Comment accepted. The answer key was modified to reflect the comments. However, note that a reference to loss of solid plant pressure control is required to get full credit in part b. QUESTION 2.12 Comment: , An alternative answer to " Return to power after blackout" would be " auto start following an undervoltage on bus 142." Byron proposes that this also be accepted as a correct answer. NRC Response: Comment accepted. The undervoltage on bus 142 is accepted as an alternate for answer b for full credit. QUESTION 2.13 Comment: For RCS loops B and C, the accumulator and high head injection points are j incorrect. These inject into the cold leg, not the hot leg. For loop B, the normal charging line connects to the cold leg, not the hot leg. For loop C, the letdown line and spray line connect to the cold leg, not the hot leg. Byron proposes that the attached figure 2-4 be used as the correct answer. NRC Response: Comment accepted. The Byron figure will be used as the answer key answer. Loops C and B had their pumps and S/Gs reversed on the drawing and the connections were not changed. L. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
QUESTION 2.15 Comment: c '. . FALSE - Speed and voltage may be manually controlled in two modes after an emergency start: Manual Emergency Mode Manual Test Mode Byron. requests that FALSE be accepted as the correct answer. NRC Response: Comment accepted. The answer is changed to FALSE. QUESTION 3.03 Comment: Byron proposes'that " iconic" be accepted as an alternative to SPDS, as they are synonymous. NRC Response: Connent accepted. Iconic is accepted as an alternate answer for SPDS for full credit. QUESTION 3.05 Comment: System Description Chapter 43b, pages 33 and 34, describe the response of
. control room HVAC System and the Control Room Offices HVAC System if high radiation is detected in the outside air intake.
Byron station proposes that the following actions not be considered as incorrect answers for part b of question 3.05:
- 1. Recirculation Charcoal absorber inlet and outlet isolation valves open.
- 2. Recirculation Charcoal absorber bypass damper closes.
- 3. Control Room Offices HVAC fans trip.
NRC Response: Comment accepted. In fact, the answers in the comment were added to the answer key and are required for full credit. QUESTION 3.11
Comment: The purpose of the OT Delta-T circuit (i.e., Rx trip and runback) is to provide DNB protection and the purpose of the OP Delta-T circuit (i.e., Rx trip and runback) is to provide KW/ft protection. Byron proposes that DNB and KW/ft not be considered as incorrect answers for part b. NRC Response: Comment accepted. DNB and KW/ft protection is accepted as an alternate answer to part b. for full credit. QUESTION 3.12 Comment: In question 3.12, part a., " actuated" implies that the cold overpressure protection system has responded to a cold over-pressure condition. There are two methods by which it can be " actuated" from the MCB:
- 1. Manually place the PORV's in the OPEN position, or
- 2. Automatically - Actual WR RCS pressure greater than the PORV lift setpoint (as described in answer b).
In addition, the NRC answer in part a. is attempting to describe how the Cold Overpressure Protection System (COP) is placed in service, not actuated. To place the C0P in service, the PORV control switch must be in the ARM LOW TEMP position. Due to the confusing nature of this question, Byron proposes that part a. be deleted from the exam. In regard to Question 3.12, part c., " deviation alarm" should be accepted as the correct answer. The question does not ask for the alarm setpoint, and therefore, should not be required in the answer. Another alarm on the MCB, " LOW TEMP ARMED AND PORV ISOL VLV CLOSED," also actively warns the operator of a possible cold overpressurization. This alarm is generated when the Cold Overpressure Protection System is in service (armed) and the PORV Isolation Valve (block valve) is closed. Cold Overpressurization is possible in this condition because the PORV's are now incapable of relieving an RCS pressure transient. Byron requests that the " LOW TEMP ARMED AND PORV ISOL VLV CLOSED" alarm not be considered an incorrect answer. NRC Response:
)
Comments noted. Part a. was clarified during the exam with questions. Consideration on a case-by-case basis will be given to other answers similar to the ones in the comment. In part c., the setpoint was removed from the ! answer key. The " LOW TEMP ARMED AND PORV ISOL VLV CLOSED" alarm will be given only 1 credit. The deviation alarm should also alarm if you have a cold overpressurization ir this condition.
QUESTION 3.14 Comment:
.The question in part a. asks for SI actuations only. Therefore, the answer should be:
- 1. PZR LOW PRESSURE, 1830 psig
- 2. STM LINE LOW PRESSURE, 640 psig.
Byron proposes that Steam Pressure Rate 50 psig in 100 seconds not be required for full credit. NRC Response: Comment accepted. The answer key is modified to reflect the comment. QUESTION 4.05 Comment: Byron requests that for part b., Reactor-to-secondary leakage, or steam generator tube leakage be accepted as correct answers, as well as "' identified" leakage. NRC Response: Comment not accepted. The candidates need to know their Technical Specification definitions. QUESTION 4.07 Comment: Question 4.07, part b. answers should be 350 F and 360 gsig (not pisd). Question 4.07, part c. answers should be _325 psig and _200 psid. Question called for RCS temperature. However, BGP 100-5 does not list a 325 F temperature requirement to maintain the RCP's running. Byron proposes deleting the first answer in part c., as there is no correct answer as the question is worded. NRC Response: i Comment accepted. The first part of part c. is deleted. It was corrected
" incorrectly" during the exam. Also during the exam, psid was changed to psig in part b.
QUESTION 4.15 l l l l l
Conunent: Step Sa (blank #) on page 2 of 2BGP 100-7T2 requires the # from RRD step 1.c. to be entered. The answer key lists the # from RRD step 1.g. This is incorrect. This also affects step Sb on page 3 of-2BGP 100-7T2 on page 3. NRC Response: Comment accepted. Also be advised that Figure 10A of BCB-2 was misread. The answer key was changed to reflect these changes. e i l l 4 I i _._._______._.______U
SENIOR REACTOR OPERATOR EXAM QUESTION 6.01 1 Comment: A Phase A isolation signal will not provide an auto start signal to any pumps. The CC Pps auto start on low discharge pressure, SI, and return to power after a blackout only. Byron proposes that answer number 1 for part a. not be required for full credit. NRC Response: The question stated that a phase A signal and an SI signal are present. As indicated in the original reference and in the facility comment, the SI signal causes a non-running pump to receive a start signal. For this reason, answer No. 1 for part a. is a correct answer. The answer key remains unchanged. QUESTION 6.04 Comment: There are two additional logic conditions required for breaker 2423 to automatically close:
- 1. Breakers 2411, 2412, 2414 open
- 2. Control switch for 2413 in after trip, or
- 3. Control switch for surveillance test in Surveillance test position (Note:
this is a recent modification to both units that allows the DG to be in test and will still close onto the bus in the event of a blackout).
' NRC Response:
Comment accepted. Answer key modified to include additional interlocks. llowever, the additional response of part b. is modified to read "'.untrol switch for 2423 in after trip" and part a. modified to read " breakers 2421, 2422, 2424 open." j i QllESTION 6.05 Comment: The lo-l'o Tave setpoint at Byron Station is 550 F. Byron proposes that Tave greater than 550 F be accepted as the correct l answer for Question 6.05, part a.1 and b.1. NRC Response: Comment accepted. Answer key modified to correct the value for 10-10 Tave.
,y
. QUESTION 6.06-
. Comment:
Candidats may also respond to Answer No. '1; by writing: On a trip, an immediate: generator trip coincident with a failure of ABT's would cause a loss of flow which may' result in exceeding DNB limits. Byron proposes that this response also be accepted as correct. NRC Response: The facility requested answer is viewed simply as a rewording of the original. answer key. As such, it will be accepted as a correct answer. Answer key remains unchanged. Alternate equivalent wording is always acceptable. QUESTION 6.07 Comment: Question 6.0/ is misleading. .The-Unit. Aux 1ransformer output voltageJ11sted
- does not specify whether.it is feeding the 6.9kv or 4.16kv bus. Therefore,-
5200 volts could be interpreted as a low voltage on a 6.9kv bus, of as high-voltage on'a 4.16kv bus. The second interpretation would make answer no. 1
. (RCP undervoltage) incorrect.
Byron proposes that answer number 1 not be required for full credit. NRC Response:
- The question states that the candidate should assume that all channels read the same. Thus, the voltage specified.should be considered as the same for both feeds to the service busses. 'Thus, the voltage on the 6.9kv bus is low, and the voltage on the 4.16kv bus is high. The high voltage causes no
' trip, while the low voltage causes a trip as stated in the original answer key.
The answer key-remains unchanged. QUESTION 6.08 Comment: Part a. of the answer key is incorrect. BIS 3.1.1-213. " turbine impulse pressure surveillance," only allows the test to be performed in Mode 1 >10% power or Modes 3, 4, 5, or 6 provided the Rx trip breakers are open. t Byron proposes that part a. of the answer key not be required for full credit. 4 U__._.__m. _ _ _ . _ - - _ _ _ . . _ _ . . . _ _ _ . . _ _ _ . . _ _ _ _ - . _ . _ . _ m . _ _ _ _m _..-_.__ . _ .. __ _ __ -_ ._ _ _ -. _ _ _ . _ _ . - - . _ . _ . , _ _ . _ _ _ _ _ _ _ _ _ . _ . . - _ _ _ . _ _ . _ _ _ _ _ . _ _ _ . - _ _ _ _ _ _ _ _ _
NRC Response: The original answer key, which is different from the comment supplied by the facility, states that either answer a. or answer b. will be accepted for full credit along with an explanation that the concern is making up logic for P-7 with a turbine trip. While it is recognized that the answer for part a. is contrary to approved procedures, it is a viable alternative from a strictly systems viewpoint. Since this question is in the systems section and not in the procedures section, this answer will be considered adequate for this section of the test. As an additional note, the reference given in the facility response was not provided to the NRC prior to the examination. For the above stated reason, as well as the fact that part a. of the original answer is not required to receive full credit, the answer key remains unchanged. QUESTION 6.13 Comment:
- 1. Candidate may have included the following in his answer:
- a. Manually block BDPS prior to startup
- b. Manually block HIGH FLUX AT SHUTDOWN alarm prior to startup
- 2. Action for P-10 should also include:
- a. Automatically blocks SR Rx trip and deenergizes the SR high voltage,
- b. Manually block Intermediate Range High Flux Rod Stop Byron proposes that the above statement should not be considered as incorrect answers.
NRC Response: Comment partially accepted. The answers as stated in the facility comment will not be considered incorrect. However, the automatic actions as stated for P-10 were not elicited by the question, which asked only for manual actions. Answer key remains unchanged. QUESTION 7.02 Comment: The answer for part h. of question 7.02 should be Yes. Byron Station no-load Tave is 557 F. The steam dump failure has dropped temperature to 540 F. This is a 17 F drop in temperature that should have been terminated by the P-12 LO-2 Tave interlock at 550 F or by operator action. Tave dropping below 550 F implies that an uncontrolled cooldown was in progress and per B0A-PRI-1, " Emergency Boration," an emergency boration is required.
a Byron proposes that Yes be accepted as the correct answer to part h. of the question 7.02. j NRC Response: l Comment accepted. Answer key modified to reflect-the comment. QUESTION 7.09 4 Comment: Per the User's Guide, page 5, if the contingency action in the right hand column cannot be performed, then go to the next contingency action. If further contingency actions aren't provided, then return to the next step or substep in the left hand column. Byron proposes that " perform the next step or substep" be acceptable, and that "in the left hand column" not be required for full credit. NRC Response: Partially concur. Answer key modified as follows: "If further contingency actions are provided, perform next contingency action. (.5). If further contingency actions are not provided, perform the next step or substep in the left-hand column. (.5)." QUESTION 8.03 Comment: The NRC answer is correct. However, it appears that "resulting uncontrolled RCS cooldown" is required for full credit in the " Accident" part of the i answer. A steam line break at no load operating temperature is understood to result in an uncontrolled cooldown. ; i Byron proposes that uncontrolled cooldown should not be required for full I c.' edit in this portion of the answer. ! NRC Response: Do not concur. Contrary to the facility contention, a steam line break does not always result in an uncontrolled cooldown. If a steam line break is , downstream of the MSIV, the operator can take action to control the cooldown. the answer key remains unchanged. However, full credit will be awarded if an "unisolatable steam line break" is specified in the answer. QUESTION 8.04 Comment: Question 8.04 b. should be answered " Alert." At Byron Station, a ramp in power to 1050 MWe is considered a controlled evolution, and not a plant transient, u-__-- - - . _ - - - _ - - - - - - _ - - - - _ _ - - - - - _ _ _ _
1-l , Byron? proposes that '" Alert" be accepted as the correct answer for part b. of
- question 8.04.
Question 8.04 d. should be answered " Unusual Event.'. A demonstration outside the_ main gate that prevents the oncoming crews from entering the plant for two hours is considered a civil disturbance. A civil disturbance is a spontaneous collective group gathering (demonstration) which disrupts normal operations (prevented oncoming crews from entering plant' for
'two hours). To classify the event as an." Alert," it would have to be an ongoing ' security. threat of increasing severity that persists for more than 60 minutes.
We agree that.'it does persist for more than 60 minutes, but it is not an ongoing security threat of increasing severity. Byron proposes that " Unusual Event" be accepted as the correct answer for part
'd. of question 8.04.
NRC Response: Comment' accepted. Answer key modified to reflect the comment. QUESTION 8.10 Comment: The containment purge isolation system (i.e., containment purge isolation valves) are referenced in two technical specification LC0's; 3.9.9 and 3.6.3.
'LC0 3.9.9 is not applicable in Mode 1. However, LC0 3.6.3 is applicable in Mode 1.
Part d. of question 8.10 states that the containment purge isolation system is inoperabic and that the purge isolation valves are shut. Technical Specification LC0 3.6.3 action statement states that with one or more isolation valves inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and within four hours:
- 1. Restore the inoperable valve (s) to OPERABLE status, or
- 2. Isolate each affected penetration by use of at least one deactivated automatic valve secured in the isolation position, or 3.. Isolate each affected penetration by use of at least one closed manual valve or blind flenge.
Byron proposes that "Yes" be accepted as the correct answer for part d. of question 8.10. 1 NRC Response: i Comment accepted. Answer key modified to reflect the comment. L_ __ _ ____ _ _ ____________ _ _
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L s .! [ {tMGU ATORY COMMISSION 2 REACTOR OPERATOR LICENSE EXAMINAT]ON FAC1L1TY: _B_YKON REACTOR TYPE: PWR-WEC4 l l DATE ADMINISTERED: 81211Z02 EXAMINER: S_UNDEBL A_ND . P. CANDIDATE j INSTRUCTIONS TO CANDIDATE 1 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 i 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 CANDIDATE'S CATEGORY V b.L U E _ ._.T D T A L SCQRE VALUE_ CATEgQ.RY 26.00 35.32 ._
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID ELOW 24._00 24.30 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 25.00 26.32 3. INSTRUtiENTS AND CONTROLS
_24.75 26.06 ,_ _ 4 . PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL . 98.75 % Totals Final Grade All work done on this examination is my own. I have neither given nor received aid. Candidate's Signature
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS D6 ring-the. administration of this examination the following rules apply:
- 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties. i
- 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating. l
- 3. Use black ink or dark pencil only to facilitate legible reproductions.
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- 4. Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary). .
- 6. Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of each ,
section of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only on one side ,
of the paper, and write "Last Page" on the last answer sheet.
- 9. Number each answer as to category and number, for example, 1.4, 6.3. l
- 10. Ship at least three lines between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use abbreviations only if they.are commonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer
- to mathematical problems whether indicated in the question or not. J l
- 15. Partia) credit may be given. Therefore, ANSWER ALL PARTS OF THE l QUESTION AND DO NOT LEAVE ANY ANSWER BLANK. l l
- 16. If parts of the examination are not clear as to intent, ask questions of j the examiner only.
]
- 17. You must sign the statement on the cover sheet that indicates that the l
j work is your own and you have not received or been given assistance in j ! completing the examination. This must be done after the examination has j been completed. l u - - .-. - - - - - - - - _ _ - . _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
f-18,,. . 5 hen you complete your examination, you shall:
- a. Assemble your examination as follows:
(1) Exam questions on top. (2) Exam aids - figures, tables, etc. (3) Answer pages including figures which are part of the answer.
- b. Turn in your copy of the examination and all pages used to answer the examination. quest 1ons.
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.
'd . Leave the examination area, as defined by the examiner. If after leaving, you are round in this area while the examination is still in progress, your license may be denied or revoked.
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1*? PRINCIPLES OF NUCLEAR POWER _ PLANT OPERATION 1 Page. 4 fliERMODYNAMICS, j HEAT TRANSFER AND_ FLUID FLQH ; QUESTION 1.01 (1.00) Fill in the blank with the most correct choice given. During a Xenon-free reactor startup, critical data was inadvertently taken two decades below the required Intermediate Range Level (IR). Assuming-that RCS temperatures and boron concentrations were the same, the critical rod position taken at the proper IR level the critical rod position taken two decades below the proper IR level,
- a. is less than
- b. is the same as -
- c. is greater than I
- d. cannot be compared to QUESTION. 1.02 (1.50)
Indicate whether the following are TRUE or FALSE when adjusting the power range channels to 100% based on a calculated calorimetric. (Consider each case separately)
- a. If feedwater temperature used in the calorimetric was lower than actual feedwater temperature, actual power will be higher than calculated power.
- b. If the feedwater flow used in the calorimetric was lower than actual feed flow, actual power will be less than calculated power.
- c. If RCP heat input used in the calorimetric was neglected, actual power will be less than calculated power.
QUESTION 1.03 (1.00) True or False?
- a. One of the pump laws for centrifugal pumps states that the volumetric ,
flow rate is proportional to the speed of the pump. (0.6)
- b. As VCT level increases, volumetric flow rate from the centrifugal charging pump increases. (0.5) i 4
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *4444) l _. _______-__-O
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'1' ; EglNCIELES OF_ NUCLEAR-POWEB_ELANT OPERATIQN 1 Page 5
., THERMODYNAMICS. HEdT: TRANSFER _AND FLUID FLQW
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L QUESTION 1.04 (1.00) Prior to reaching criticality during a reactor startup, an-initial
-reactivity addition causes count rate to increase from 20 cps to 40 cps.
A second reactivity addition causes count rate to increase-from 40 to 80 cps. Which one of the following statements is correct?
- a. The first reactivity addition was larger.
- b. The second reactivity addition was larger,
- c. The reactivity additions were equal.
- d. There is not enough data given to determine the relationship of the reactivity additions.
i QUESTION 1.05 (1.00) Choose the correct phrase to complete the following sentence: l As the core ages from BOL to EOL, the ratio of Pu-239 atoms to U-235 atoms
-increases. This. changing ratio causes the _
- a. Reactor period to decrease.
- b. Void coefficient to become less negative.
- c. Moderator temperature coefficient to become less negative.
- d. Delayed neutron fraction to increase.
l (4**** CATEGORY 1 CONTINUED ON NEXT PAGE *****) l l j
- j. PB1HCIPLES OF NUCLEAR POWER PLANT OPERATIQdi Page 6
.THEBMODYNhb1CS, BEAT _IBhBSEEB_AND ELM 1D FLQW ~
1 i QUESTION 1.06 ('1.00) Which ONE of the following statements concerning the production and removal of Xe-135 is correct?
- a. At full power, steady state, about half of the Xenon is produced j by Iodine decay and the other half is produced as a direct fission i product. l l .
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- b. Following a reactor trip from equilibrium conditions, Xenon peaks )
because delayed neutron precursors continue to decay to Xenon while ] neutron absorption (burnout) has ceased- , I
- c. Xenon production and removal increases linearly as power level l increases, i.e., the value of 100% equilibrium Xenon is twice that j of 50% equilibrium Xenon. 1 3
- d. At low power levels., Xenon decay is the major removal methcd. At j high power levels, burnout is the major removal method.
QUEST 10N 1.07 (1.00) The - (1/3) DPM SUR following a reactor trip is caused by which one of the i following?
- a. The decay constant of the longest-lived group of delayed neutrons.
- b. The decay constant of the shortest-lived group of delayed neutrons.
- c. Doppler adding positive reactivity following a trip due to cooldown.
- d. The amount of negative reactivity added being greater than the
- shutdown margin.
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1 l I l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) ] l l l l
Ef. EBINCIELES OF NUCLEAB_POWEB_EL. ANT 0EEB61 IONz Page 7
. THERMODYNAMICS _HE&I_IBANDFER AND ELUID FLOW QUESTION 1.08 (1.00) ,
I T' 4 2200 degrees-F maximum peak cladding temperature is based on l (Choose the most correct answer) _)
- a. It is 600 degrees F below the fuel cladding melting point. !
- b. Any clad temperature higher than this correlates to a fuel centerline temperature.
- c. A zircalloy-water reaction is accelerated'at temperatures above 2200: degrees F. <
- d. The thermal conductivity of aircalloy decreases above 2200 degrees F causing an unacceptably sharp rise in the fuel centerline t temperature.
1.09 (1c50) j QUESTION Answer the following concerning shutdown margin (SDM) during power operations,
- a. What condition during normal operation insures adequate SDM? (0.5) ,
- b. Describe if the SDM INCREASES, DECREASES, or REMAINS THE SAME for the following conditions. (Assume control systems are in automatic)
- 1. Xenon concentration increases (0.5)
- 2. RCS boron dilution (0,5) 1 l
QUESTION 1.10 (1.25) I
- a. What three indications / conditions are used by the reactor operator to l declare the reactor critical? (0.75) j l
- b. What condition is the reactor in at the point of declaring )
criticality? (Limit your answer to SUBCRITICAL, CRITICAL, or j SUPERCRITICAL) (0.5) l (***** CATEGO'RY 1 CONTINUED ON NEXT PAGE *****) i ______--_-_a
[ T. P81dQ1thES-UF NUMkM68_EQWE8_.Eh&NT OPE 86TIOU1 Page 8
,THEBOODYNAUICS, HEAT TR60eFER AND FLUID __ FLOW i
s -- r; QUESTION 1.11 (1.00) Knowing that the heat transferred from the primary system equals the heat transferred to the secondary system, why do the primary and secondary mass flow rates differ? (1.0) QUESTION 1.12 (2.00) The plant is operating at 30% power when one reactor coolant pump trips due to an electrical fault. The control systems are all in AUTOMATIC. Bank D rods are initially at 163 steps. Assume the plant does not trip. Indicate whether the following parameters INCREASE, DECREASE, or REMAIN THE SAME.
- a. Turbine power
- b. Reactor power (final)
- c. Core Delta-T
- d. T-avg (affected loop)
- e. T-avg (unaffected loop)
- f. Final rod height
- g. Delta-T (affected loop)
- h. Delta-T (unaffected loop)
QUESTION 1.13 (1.00) Refer to Figure 1-1 and describe the dominant processes that are occurring in parts a and b of the critical boron concentration versus burnur graph.
. QUESTION 1.14 (1.50)
Steam passing through a relief valve undergoes a throttling process. , Calculate the temperature and condition (saturated, subcooled, ; superheated) of the steam downstream of a steam generator atmospheric relief valve. (Assume the pressure drops from 1100 to 14.7 psia across the valve) (***** CATEGORY I CONTINUED ON NEXT PAGE *****) l
~
l -1V EBIUC11'LES OF NUCLE 68_EOWER PLANT OEERATIONz Page 9 j
.IBERMQDYNAMICSt _ BEAT TRANSEEB_AND FLUID FLOW I
QUESTION 1.15 (2.00)
.Since DNB is not an observable condition, operators must monitor certain parameters to ensure that DNB does not occur. LIST SlX (6) ]
of those parameters. j 1 ( i j QUESTION 1.16 (3.00) -1 I During a startup, the reactor is suberitical at 3000 CPS on the Source hange Instruments when a steam du.np valve f ails open.
- a. EXPLAIN what happens to reactor power and T-avg and why. Your answer should be continued until stable conditions are reached with NO operator action. (Assumptions include that the reactor is !
undermoderated, at BOL, no reactor trip occurs, and steam dump valve stays open.) (2.0)
- b. How will the transient and final conditions differ if the transient in part "a" happened at EOL instead of BOL? (1.0) i
-QUESTION 1.17 (2.00)
An incident at ANO-1 resulted in fuel damage when a control rod was found to be 90 inches further into the core than the remaining rods in its group for a period of 12 days. The rod was withdrawn to align it with the rest of the group within one hour while the plant continued to operate at full power. Why is fuel damage likely to occur in such a situation? QUESTION 1.18 (1.25) An operator is taking data for a 1/M plot. The value of K-eff it, about 0.97. The operator withdraws the control rods to insert +1000 pcm and immediately records the count rates.
- a. Describe how the 1/M plot is affected. (.75)
- b. Does this result in over or under predicting the critical rod height?
(0.b) . (****d END OF CATEGORY 1 *****) t
e . ... g j -s ~M' $ hie &__bh6HI_kgSIGN'INCLUDINGQAFEIX_ANDEMERGENCY 'Page 10-SY .EYSTEde-q. QUESTION 2.Olf (1.00) LThe RCS is~ cooled down,and. solid, on RHR. How'is pressure control g.l maintained?
. QUESTION'- 2 . 0'2 ~ (0.00) 1 Deleted'
, n
- l ' <
'Ji QUESTION 2.03 (1.00)
Describe the interlock and its purpose for opening CS001A(B) (Containment Spray Pump RWST Suction Isolation Valves). (1.0)
-QUESTION 2.04 (1.50)-
Answer the following concerning the Main Steam Atmospheric Relief Valves (S/G PORVs):
- a. If a S/G PORV' accidentally failed full open during normal operations at 100% power, what would you expect the new steady state reactor power to be? (0,5)
- b. What are the setpoints for-opening the S/G PORV in automatic?- (0.5)-
- c. The S/G PORV for "A" S/G loses power from its 480V ESF source. How may it be operated? (0.5) ,
QUESTION 2.05 (2.00) Answer the following concerning the Reactor Makeup Control System (RMCS):
- a. .While performing a BORATION, the Boric Acid Totali=er counts out.
What response is seen in the system? (1.0)
- b. Figure 2-1 shows the RMCS in the DILUTE mode. Modify the figure to add the piping and components included in the ALTERNATE DILUTE mode. (1.0)
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) - _ = - _ _ - _ _ -
n L't. Page 11 ELAUI_DEElGM_lECLV.ElEC_EAEETY AHU_ EMERGENCY l .SYSTESS.
.s L l QUESTION 2.06 (1.'50)
List six conditions (other than manual) that trip the Turbine Driven Main Feedwater Pumps. Include setpcints, where applicable.
. QUESTION 2.07 (1.50)
Indicate whether the following valves fail open or' closed upon a loss of instrument air.
- a. Pressurizer PORV (RY-45bA)
- b. Auxiliary Feed flow Control Valve (AFOOS-C)
- c. Letdown Isolation Valve (CV-459)
- d. Letdown Pressure Control Valve (CV-131)
- e. RCP Seal Flow Control Valve (CV-182)
- f. Startup Feed Pump Recirc Valve (FW-076)
/
JyESTION 2.08 (2.10) Answer the following concerning the Auxiliary Feedwater System (AFW):
- a. Refer to Figure 2-2 and choose (a through e) the correct point at which AFW flow enters'the main feedwater system. (0.5)
- b. What four signals will cause the 1A AFW punp (motor) to start.
Include coincidences, if applicable. (1.0)
- c. What signals will open the emergency supply water valves to the suction of the AFW pumps? (0.6)
QUESTION 2.09 (1.50) List six conditions that cauce a Reactor Coolant Pump trip. (Do not include procedural requirements) I (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ .___-_______O
4 Page 12 j d$ J
-PLANT _tESIGN ltCLUDING SAFETY AND_ EMERGENCY !
SIEIEMS
~
i i l i QUESTION 2.10 -(2.40) Indicate on Figure 2-3 (High Head SI System) what components actuate automatically on a Safety Injection. For valves, indicate whether they open (O), close (C), or modulate (M). . I i QUESTION .2.11 (1.60) Answer the following concerning the letdown portion of the Chemical and Volume Control System:
- a. What three conditions (other than manual operation) will cause the orifice isolation valves to close? (.75)
Lb. What is the purpose of PCV-131 (Low Pressure Letdown Valve) Include in your answer, any potentially damaging effects to the letdown system. (.75) QUESTION 2.12 (2.00) Answer the following concerning the Component Cooling Water (CCW) System:
- a. Which of the following loads are cooled by CCW? (1.0)
- 1. Containment Penetration Cooling
- 2. Waste Gas Compressor
- 3. CVCS Regenerative Heat Exchanger ,
- 4. Excess Letdown Heat Exchanger
- b. During normal operations, what is the load that receiver the most CCW flow? (0.25) (Do not limit your answer to the loads in part a.)
- c. Under what three conditions will the "O" pump start in " standby" if it is operated on bus 142 with.the 1B pump in PULL TO LOCK?
(0.75) I (**+44 CATEGORY 2 CONTINUED ON NEXT PAGE **4*4)
4 El EbbMI_RE&lGN IHgkUDING SAFETY AND EMEBGEllg1 Page 13
- EXSIEUE s
QUESTION 2.13 (2.00) Use Figure 2-4 (Reactor Coolant System) to indicate where the following systems / components connect to the Reactor Coolant Loops:
- a. Pressurizer surge line
- b. Pressurizer spray lines (2)
- c. Letdown line
- d. Normal charging line
- e. Excess letdown lines (4)
- f. Accumulator injection lines (4)
- g. Alternate charging line
- h. Hi-Head Charging lines (4)
QUESTION 2.14 (2.00) Answer the following concerning DC power distribution: I
- a. On loss of DC Bus 113 (non-vital 125 VDC Bus), how is operation of the Permanent Magnet Generator (PMG) affected? (1.0)
- b. The loss of either vital 125 VDC bus at power requires a Manual Reactor Trip. Why is this necessary? (1.0) l i
l i (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) j i l
2 PLANT DESIGN lHCLUDTHg_ SAFETY AND EMERGENCY Page 14
.SYSIEUS QUESTION 2.16 (2.00)
Answer the following concerning the Emergency Diesel Generator (EDG):
- a. Where does the "1A" EDG get field flash power from? (0.5)
- b. For the EDG in the AUTO MODE, which of the following trips are I disabled in an emergency start? (1.0) ,
- 1. Reverse Power
- 2. Loss of Field
- 3. Engine Vibration 4, Overspeed
- c. TRUE or FALSE?
After an emergency start, the speed and voltage may only be manually controlled in the MANUAL EMERGENCY MODE. (0.5) { 1 l l i (***** END OF CATEGORY 2 *****)
- 3. INSTRUMENTS AND CONIROLS Page 15 QUESTION 3.01 (1.00)
Fill in the blanks with the correct response concerning the Rod Control System. (Note: A blank may contain more than one word)
- a. A failure in the Power Cabinet requires replacement of a printed circuit card. To avoid dropping rods, the provides power to the Power Cabinets to keep the control rods in. position,
- b. Normal system parameters place speed of the rods in MANUAL at spm and the SHUTDOWN banks at spm.
QUESTION 3.02 (1.50)
~ Fill in the blank for the following concerning the Rx Trip and Bypass Breakers:
- a. The Reactor Trip Breakers have two trip mechanisms. In the event that- the UV coil deenergizing does not give a reactor trip, the device still activates to open the breakers.
- b. Train A's bypass breaker is racked in and closed for testing. Train B's bypass breaker is then inadvertently closed. The general warning.
circuit will generate a _.
- c. Opening of the Rx Trip and Bypass breakers generates permissive p
QUESTION 3.03 (0.75) There are two display hardware subsystems for the Reactor Vessel Level
- Indication System (RVLIS). List where the readouts are for each system and indicate which one is safety grade l
l (***** CATEGORY 3 CONTINUED ON NEXT PAGE * **** ) 1 li
2 THSIBUMENTS__AND_CDUIBOLS Page 16 QUESTION 3.04 (1.00) You are pulling control rods on a reactor startup. Indicate whether the DRFI rod bottom lights for the following groups are ON or OFF when the ROD AT BOTTOM alarm clears-.
- a. Control Group A
- b. Control Group B
- c. Control Group C
- d. Control Group D QUESTION 3.05 (2.00)
Describe the interlocks for the following radiation detectors if a high radiation signal is detected:
- a. RE-PR008 (Steam Generator Blowdown Monitor)
- b. ORE-PR031 and ORE-PR032 (r trol Room Outside Air Intake A Monitors)
- c. RE-ARO11 (Containment Fuel Handling Incident)
- d. RE-PR009 (C'i Heat Exchanger Water Outlet Monitor)(Unit 1)
QUESTION 3.06 (1.00) Concerning MFW isolation, refer to Figure 3-1 and label the inputs for ' items a through d. (1.0) QUESTION 3.07 (2.75) List ALL (11) signals, other than indications and alarms, that are derived
-from Power Range Nuclear Instrumentation. Include control signals, protection signals, permissives, and interlocks. Indicate setpoints and if signal is auctioneered, if applicable.
l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) l 1
S' Page 17 INSTRUd? MIS AND CQNTROLS
#e QUEST]ON 3.08 (2.50)
Answer the following questions concerning S/G Water Level Control: n .- Refer to Figure.3-2 and list the inputs to the SGWLC system at points a through e. (1.25)
- b. Does MFP speed initially INCREASE, DECREASE, or REMAIN THE SAME if PT-BOB (MFW Header Pressure) fails high? Explain your answer.
(1.25) QUESTION 3.09 ( 2. 00 )' Concerning P-7:
- a. How is F-7 derived? Include setpoints and coincidences. (1.0)
- b. What is P-7 used for? (1.0)
QUESTION 3.10 (1.60) List all control, protection, and alarm functions as PZR level increases from 0 to 100% level. Include all coincidences, setpoints, and applicable conditions as port of your answer. QUESTION 3 11 (2.00) Answer the following concerning Main Turbine Control: .
- a. Describe what occurs during a Turbine Runback. (1.0)
- b. What is the purpose of a Turbine Runback? (0.b)
- c. List the TWO (2) conditions that cause a Turbine Runback. Include setpoints. (0.b)
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) 1
v- .
]
Page IU. dm__lBSIEUMEUId_AUL CONIBOLS 9' c
.I M M,UESTION. >
3.12 (2.00) h Answer the following concerning PZR Cold Overpressure Protection. ]
~
b $f 'lA . How is Cold Overpressure Protection actuated from the MCB? (0.b) hf f i
- b. - How does Cold Overpressure Protection work? Include any inputs l
[f[cc-. and comparisons. (1.0) ! Q) 5 How is the operator actively warned of a possible cold Ui," c .. overpressurization? (0.5)
]3
,d ( 3 w(iQUE3TIbl. 3.13 (2.50) Inswer ti.eJfollowing concerning Steam Dump Control:
- a. 1%for to Figure 3-3 (Arming Circuit) and indicate what signals-a+ s1 . operate the contacts designated a through d. (Note that a
M" ;
^
through e are interchangeable.) (1.0) s
- b. - In the STEAM PRESSURE mode of control, what is compared in the N ' signar;comparator? (0,5) u>
.. a If the tbrbine trips, how is a HIGH setpoint generated? (0.5) lto4p or c. How dous'a HIGH setpoint (c. above) affect the Steam Dump System, 1 L 'd
. ,/.. assuming that all in.terlock conditions are satisfied and the r
r, . steam dumps are armed? (0.5) a h ! q(jQUESTION 3.14 (i.50) ' Permissive P-11 (PZR Low Pressure) allows for an orderly reduction of RCS Z!p pressure without causing a Safety In.iection. i n. What SI signals can be blocked with P-11 present? Include
" setpoints. (1,0) q
' i.;j
'L
- b. How are the blucks in part a. automatically removed? (0.b)
{. D ' j > a 6 (**.*+4' CATEGORY 3 CONTINUED ON NEXT PAGE **++4~) ? b_L U m :x n
- _ - -- _ _-. _ - _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ -
i
Page 19 A___INSTRl!LEL4TS_AND CONTBQLD QUESTION 3.15' (1.00) Which one of the following wi.11 not cause a ROD CONTROL URGENT FAILURE alarm to actuate? a, A Slave Cycler failure
- b. DRI'l data A. and B failure
- c. Loose circuit card.in the logic cabinet
- d. A rod drive motor-generator set trip i
(*S444 END OF CATEGORY 3 *****)
4 4' . EEggEDURES --NORM &L. ABNORM &L, EMERGENCY Page 20 l AUD RADlQLQGICAL_gpNTROL QUESTION 4.01' (1.00) Which of the following is a RCP trip criteria per the BEP-O fold-out page.
- a. CC water to RCP bearings is lost
- b. -Containment Phase A 1 solation
- c. RCS Pressure 1250 psig with SI pump flow at 75 gpm without a cooldown in progress
- d. .The conditions in c. above with a cooldown in progress QUESTION 4.02 (1.50)
Answer the following questions TRUE or FALSE concerning a plant heatup (BGP 100-3). 1
- a. Starting a RHR Pump while using RHR letdown with the RCS solid causes an inadvertent RCS pressure reduction if PCV-CV-131 is left in AUTO,
- b. If the control rod drive system is capable of rod withdrawal, then at least one manual reactor trip channel must be operable,
- c. When directed by the Shift Engineer, the MSIV bypass lines may be opened to heat up the main steam lines. The bypasses should be opened as quickly as possible to avoid water hammer.
QUESTION 4.03 (1.00) . Choose the most correct answer for the following: 1 An area accessible to personnel where radiation exists at levels that a j major portion of the body could receive a dose of greater than b mrem in j any one hour is considered a(n)
- a. Radiation Area
- b. High Radiation Area
- c. Airborne Radioactivity Area
- d. Radioactive Materials Area
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
.______________-__a
4 Page 21 EBMGEDURES - HQ306kt_6BBQBUAbu_EMEEGENCY AND RAD]OLOGICAL CQHTROL QUEST 10N' 4'04
. (2.00)
Refer to Figure 4-1 (BGP 100-5 Flowchart) when answering the following questions.
- a. What is the significance of the large circle? (0.5)
- b. What does the small circle in Step 12 mean? (0.5)
- c. What does the dotted line around Step 13 permit you to do? (0.6)
- d. What does the box in Step 3 mean? (0.5)
QUESTION 4.05 (2.00). FILL IN THE BLANK for the following concerning Plant Technical Specification:
- a. Reactor Coolant System pressure shall not exceed psig.
- b. Reactor Coolant System leakage through a steam generator to the secondary coolant is known as leakage.
- c. If a FZR code safety valve (Mode 1) is deemed INOPERABLE, it must be fixed within or else be in HOT STANDBY within 6 hours.
- d. RCS temperature (Tavg) in Mode 1 must be greater than degrees F.
QUESTION 4.06 (1.00)' Answer the following TRUE or FALSE concerning controlling external dose I within Commonwealth Edison: ]
- a. Only the Station Superintendent may allow quarterly dose equivalents l above 1250 mrem. ]
- b. Supervisory approval for daily dose equivalents is not needed until '
you plan to exceed 100 mrom. (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
Lim __EBRREMMREE - EREdakt_ABNOBd&L, Et1EBGEligl Page 22
.AND_B6DlOh9Glgeh_COHTROL QUESTION 4.07 (1.75)
Fill in the blanks for the following concerning a Plant Shutdown and Cooldown per BGP 100-b:
- a. The administrative cooldown limits are ' degrees F/hr for (0,5) the RCS and degrees F/hr for the Pressurizer.
- b. RCS temperature must be below degrees F and RCS Pressure must be below psid before RHR is lined up to the RCS.
(0.5)
- c. To maintain RCPs running, #1 seal dp must be greater than psid. (0.25)
- d. Emergency Boration must be performed if or more control rods are not fully inserted. The amount of boration that must take place is at least ppm for each control rod. (0.5)
QUESTION 4.08 (1.50) In the following procedurec, state the conditions that would require a manual reactor trip.
- a. Rods moving outward (automatic) with Tave and Tref matched (BOA ROD-1). (0.75)
- b. Excessive Primary Leakage (BOA PRI-1) (0.75)
QUESTION 4.09 (1.50) . Byron Unit I has experienced a Reactor Trip. Step 2 of E-O has you verify that the turbine tripped.
- a. How do you verify that the turbine tripped? (0.6)
- b. What four methods in order (assuming that the previous one was unsuccessful) does E-O have you utilize to shut down the turbine?
(1.0) 1 (44*** CATEGORY 4 CONTINUED ON NEXT PAGE *****) l
Page.23 L__EhQCEMBES - NOBM&kt_aMQBM6kJ EMEBGENCY i_
- 6ND E6DI_QLQfilCak_ ggt!TEQL QUESTION 4.10 (1 50)
You are performing BEP ES-0.2, NATURAL CIRCULATION COOLDOWN, following a loss of offsite power, when you notice that PZR level is varying by 3arge amounts inconsistent with your cooldown rate. What is the problem and what would you do to counteract it assuming no situation existed that would require immediate cooldown and depressurization? (1.5) QUESTION 4.11 (2.00) Answer the following concerning BCA-0.0 (Loss of All AC Power);
- a. Which one of the following is not an IMMEDIATE ACTION step to the procedure? (0,5) ,
- 1. Verify turbine trip
- 2. Attempt to restore offsite power
- 3. Verify AFW flow
- 4. Verify RCS isolated The S/G's are to be depressurized AT MAXIMUM RATE to 260 psig. Why b.
maximum rate? (1.0)
- c. How is the S/G depressurized? (0,5) l l
QUESTION 4.12 (2.00) Answer the following concerning the Response to Nuclear Power Generation /ATWS (BFR-S.1) procedure when the reactor will not trip;
- a. Why is manual Safety injection unwise under these conditions? (1.0)
- b. Rods are driven inward in Automatic initially. When do you switch to manual? (0.5)
- c. What is the preferred path for immediate boration? (0.5)
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
- - _ _ _ _ - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _____ _ __ _J
- l. L fB99EDUBES._ __M9306L2_3DBORMAkt._EMEEGENQY Page.24 AMD_BbDigLoGICAL COB.IBDL-1 l
I l QUESTION 4.13 (1.00) The plant'is operating at 100% power with rod control in automatic, control bank.C at 225 steps, Tavg matched with Tref, PZR pressure at 2245 psig, and PZR level one percent from programmed band.
- a. What problem exists? (0,5)
- b. What must be done to correct it? (0.5)
QUESTION 4.14 (1.00) Byron 2 is recoverinc from a Steam Breah per 2BEP-1. Per foldout page for the 2BEP-1 series procedures, LIST the two independent. conditions that would cause you tc reinitiate Safety Injection. Assume that containment conditions are normal. QUESTION 4.15 (4.00) Calculate the Estimated Critical Position using the procedures and graphs contained in Attachment 4-1 for a startup ten hours after a reactor trip. Plant temperature is at no-load value (557 degrees F) with boron concentration 600 ppm. Desired boron concentration is 500 ppm. i l l (***** END OF CATEGORY 4 4****) i (*******44'* END OF EXAMINATION **********) l
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DURINP (WWD/NTU) L i i l CRITICAL BORON CONCENTRATION VS. BURNUP FOR HFP, I l ARO EQUILIBRIUM XENON CONDITIONS I i
- , ,dPROVED J _)
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dC1 M6n ~ L - [' 2BGP 100-7 Revision 51 l L REFERENCE REACTIVITY DATA CALCULATION AND ESTIMATED CRITICAL CONDITION CALCULATION A. STATEMENT OF APPLICABILITY: This procedure describes the steps required for recording Reference Reactivity Data (RRD) just prior to or immediately following a Reactor Shutdown, and provides instructions for calculating the Estimated Critical. Condition (ECC). B.
REFERENCES:
- 1. 2BGP 100- 7T1, Reference Reactivity Data Worksheet.
- 2. 2bCP 100- 7T2, Calculation of Estimated Critical Condition based on a known Boron Concentration.
- 3. 2BGP 100-7T3, Calculation of Estimated Critical Condition based on a known Rod Height, i
C. PREREQUISITES: l None D. _ PRECAUTIONS: None E. LIMITATIONS AND ACTIONS:
- 1. In order to calculate the condition of the core to ensure Shutdown Margin or to calculate the Estimated Critical Rod Position, use the data available from last time K pp e equaled one. (Data from a recent known previous, stable critical condition may be used. Note in remarks if this was done.)
- 2. 2BGP 100-7T1, 7T2 and 7T3 are to be retained as Plant Documentation.
DO NOT DISCARD. Forward completed tables to the Operating Staff when they are no longer required on shift for reference. ! F. MAIN BODY: 4
- 1. Reference Reactivity Data Calculation:
- a. RECORD the date and time that Keff equaled one, immediately prior to a planned shutdown or as soon as possible following a reactor trip.
- b. RECORD the controlling bank and its height prior to shutdown from the step counter on the Main Control Board.
- 7. -
f APPROVED OCT 03 $86
)
. . . R.
(7523P/0153P)
3BGP 100-7
, Revision 51
- c. Using the HZP Integral Rod Worth Curve (BCB-2, Figure #2A),
RECORD the remaining negative reactivity that could be withdrawn I from the core based on the recorded bank height. d, RECORD the critical boron concentration from either the most i recent sample, the calculated boron concentration if dilution or { boration operations are in progress during the shutdown, or from samples taken following the shutdown. If boron concentration is ] 1 from a se.mple, ENSURE that sufficient time has been allowed for i mixing (approximately 15-30 niinutes). (On the RRD Data ' Worksheet, indicate which process is used.)
- e. RECORD the power level the last time Kepp equaled one from the NR45 or Excore Nuclear Instrumentation Meters NI418, NI428, NI43B and NI44B. (Use the highest power level when recording power level).
*%MMMMn M M% nM MM M MnMMMnMMM MMM M M MM MMM N R MM uMMM M Mk M M M MM MMMMMM M
- NOTE *
.
- In the following step, if power level was not measurable in *
- the Power Range-last time Keff = .1, then the Power *
- Defect will be O.
- umun u n un nun nu nn um umn* u *nmu p*nunn
- f. RECORD the total Power Defect for the associated power level using the Total Power Defect Curve ( BCB-2, Figure #17A or BCB-2, Table 2-1).
, g. ADD the reactivity values determined in Step 1.c. and 1.f. This value will be. the total negative reactivity in the core due to rods and Power Defect the last time when K epp equaled one.
- h. OBTAIN the equivalent power history for Xenon Calculations.
This portion of Table need only be performed if the Process
- Computer is not available and use of the NIX Code Xenon Tables,
, BCB-2, Table 1-3, is not desired.
l 2. For ECC Calculation based on a known Boron Concentration:
+
- a. EXECUTE 2BGP 100-7T2.
- 3. For ECC Calculation based on a known Rod Height:
- a. EXECUTE 2BGP 100-7T3.
)
I-4 APPROVED OCT 031986 (Final)
, s.o.S.R.
1 (7523P/0153P) I
2BGP 200-7T1 Revision 51 2BGP 100-7T1 REFERENCE REACTIVITY DATA WORKSHEET l e * * * *
- e e
- n e m o n e.ne.. * * * * * ** * * * * *
- n e *
- e n e m a n n e e e * * *
- e e n e * *
- e e n e .
NOTE
- I This worksheet will be retained as plant documentation. *
- DO NOT DISCARD. Forward completed form to the Operating
- Staff when no longer required on shift for reference. Data
- l may be taken from point history as required (NOTE in *
- remarks which coniputer points were used). *
*
- e e n e* e n e* *
- ne ee n *
- e n e * * * * * * * * * *
- e n ee e n *
- e ne.... * *
- e n e e * * *
- 1.a. Shutdown Date: 1i-2 -87 Shutdown Time: 0600 1.b. Control Bank O at 67 steps.
1.c. Inserted Control Bank Worth (From BCB-~2, Figure #2A): (-) ocm 1.d. Critical Boron Concentration from (CIRCLE mode of analysis used): (1 most recent sample (2) calculation, or (3) sample after shutdown CB = 600 ppm 1.e. Power level prior to shutdown: lO percent. 1.f. Total Powar Defect (From BCB-2, Figure #17A or BCB-2, Table 2-1): (-) pcm l 1.g. TOIAL REACTIVITY (Step 1.c. + 1.f.): (-) pcm APPROVED l OCT 03 $86 i (7530P/0153P) 1 B.O.S.R. 1
2BGP 100-7T1
'
- Revision 51 2BGP 100-7T1 (Continued)
REFERENCE REACTIVITY DATA WORKSHEET 1.h }: 'uivalent Power For Xenon Calculation:
~
Equivalent Power.for Xenon Calculation: Hours Average Prior to Power. Multiplier Product Xe Power = Total = 6370 = shutdown (t) 91 91 70 _t j 0 to 1 10 x6 = 60 1 to 2 20 x5 = 100 2 to 3 30_ x5 = 150 g ' 3 to 4 50 x5 = 250 - 4 to 5 _ 60. . x4 = 2 __
~ VO 5 to 6 60 x4 = 240_l 6 to 7 60 x4 = 2'/O <
7 to 8 60 x4 = 240 / 8 to 9 60 x4 = 240 I 9 to 10 . x3 =' 210 10 to 11 ' ___70 90 x3 = 240 __ 11 to 12 (iD x3 = 270 12 to 13 97 x3 = 291 13 to 14 97 x3 = 29I 14 to 15 97 x3 = 291 15 to 16 ('1 x3 = 2 91 16 to 17 V7 x2 = 194 17 to 18 97 x2 = 144 18 to 19 97 x2 = 14 4 19 to 20 97 x2 = lH 20 to 21 CD x2 = ISO 21 to 22 (;D x2 = IPe 1
- 22 to 23 QD x2 = lRD 23 to 24 (D x2 = 19b 24 to 25 fD x2 = 180 25 to 26 9D x1 = 90 26 to 27 40 x1 = 90 27 to 28 90 x1 = 90 28 to 29 90 x1 = TD 29 to 30 90 x1 n 9D 30 to 31 100 x1 = 10 0 31 to 32 10 0 x1 = J00 ' 32 to 33 100 x1 = 100 f 33 to 34 10 0 x1 = 10 0 34 to 35 l00 x1 = 10 0 35 to 36 10 0 x1 = loo I 1
TorAt. = 6370 Remarks: Et 4500ed Od 10 lo Dower dMe do feed oschla b s W Qotho drwm 4b' 4Yt a StAnm' IEdk tm an (OSE. j I
-m o i
62S /A2-07/ OBQ( SRO Date Time (Final) APPROVED OCT 031986
! (7530P/0153P)
B.O.S.R. f
---- - - - _ _ n
>. 'a T -2BGP 100-7T3 Revision 51 2BGP'100-7T2' CALCULATION OF ESTIMATED CRITICAL CONDITION BASED ON A KNOWN BORON CONCENTRATION..
'l i'
Unit 2 Startup Number 87 - 027 Startup Date 11 / 02 / 87
'i Startup Time 1600 Time Interval since Shutdown 10 Hours I
l MMMMMMMMWMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMKMMMMMMMMMMMMM M NOTE M.
- 2BGP 100-7T2 will be used to calculate Critical Rod
- Position based on a known boron concentration. 2BGP 100-7T3**
- will be used to calculate Critical Boron Concentration
- based on a known rod position. 2BGP 100-7T2 will be *
- retained as plant documentation. DO NOT DISCARD. Forward
- M completed form to the Operating Staff for retention *
- when no longer required on shift for reference.
- MM M M M MM K MMMMM MM M M MMMMM MM M M MM M M MM M MM M MM M MM M M MM M M M M MMM MM M M M M MM M M A. REACTIVITY CHANGES ,
- 1. Reactivity Change Due to Power Reduction:
1 0 pcm - (-) ocm Power defect Power defect at = (+) pcm l at startup shutdown (RRD Step 1.f)
- 2. Reactivity Change Due to Temperature Deviation from Programmed Tave:
., ( "F - 557'F ) x (-) pcm/*F = () pcm j Tave expected Program Tave MTC from
- at startup at startup BCB-2, Figure #5
- 3. Reactivity change due to Fission Product Poison:
- a. Reactivity Change Due to Samarium l Average power over 5 days prior to shutdown from 280S NR-1, i
i Power History Hourly Surveillance IhD % Pav
-4 (-) ocm -
(-) pcm I Sm worth at Sm worth at = (-) pcm F startup time shutdown (time =0) j for Pav from for Pav from BCB-2, BCB-2, Tb1 1-4 Table 1-4
)
i
+ i I )
APPROVED l
' OCT 031986 i
E3.().53,Fg,
, (7516P/0153P) i x _ - _ _ _ _ _ _ _ _
2 2BGP 100-713 Revision 51 j 2BGP 100-7T2 (Continued) CALCULATION OF ESTIMATED CRITICAL CONDITION BASED ON A KNOWN BORON CONCENTRATION ,, ,
- b. Reactivity Change Due to Xenon [In order of preference, obtain i l
from (1): Process Computer, (2); NIX tables, BCB-2, Table 1-3, or (3): RRD worksheet step 1.h] DCm - DCm
'Xe worth at Xe worth at =() pcm startup shutdown (Zero if BCB-2, Figure :
#8A is used) j l
- c. Total Reactivity Change Due to Fission Product Poisons. q i
e (-) pcm + () ocm
- Samarium. Xenon =() pcm from A.3.a from A 3 b l
- 4. Reactivity Change Due to Adjustments in Boron Concentration: I
- a. Desired critical boron concentration at startup 500 ppm.
- b. Reactivity Change Due to Boron Adjustments. )
l SOO ppm - 600 ppm x (-> pcm/ ppm j
- Critical boron Boron conc. prior Differential Boron !
at startup to shutdown (RRD Worth from BCB-2,
~
(A.4.a) Step 1.d) Figure #10A , i
~
=() pcm
- c. DETERMINE the amount of reactivity for which Control Banks must correct.
(+).__pcm + ( ) pcm + ( ) 'pcm + ( ) pcm = { ) pcm
! A.3.c A.4.b y A.1 A.2 C T , @Qg 5. Calculation of Estimated Critical Rod Position.
O o 2 " " " " " " " " " " " " " " * " " " * " " " " " " " " " " " " "* O i
$u 2
m
- NOTE
- If the value for A.S a is NEGATIVE, it indicates that
- O
- criticality cannot be achieved at the desired boron l t
- concentration in A.4.a. RCS Boron Concentration must be *
-
- reduced and the ECC must be recalculated.
n un* n un n o n mu n n* n n n un u n*
- n u n n u m u n* n i' Calculation of Inserted Rc,d Worth Required for Criticality.
a.
\ pcm i () pcm - (-) ocm = ( )
! Control bank reactiv- Inserted control bank Inserted rod worth ity correction for worth prior to shutdown required for j criticality I criticality (A.4.c) (RRD Step 1.c) i l (7516P/0153P)
~R
~ ~
J-l
. \
~
2BGP'100-7T2 l Revision 51 2BGP 100-7T2 (Continued) CALCULATION OF ESTIMATED CRITICAL CONDITION BASED ON A KNOWN BORON CONCENTRATION..
- b. Using BCB-2, Figure 7A, DETERMINE the control bank position '!
corresponding to tb.s inserted worth calculated in A.S.a above. j Control Bank at ' steps j B. ESTIMATED CRITICAL CONDITION
SUMMARY
Present Boron Concentration 600 ppm Estimated Critical Boron Concentration 600 ppm (A.4.a) f Estimated Critical Bank Position CB _,,at steps.(A.S.b) 1 ECC performed by i NAME DATE ECC reviewed by q
/ / /
C, ESTIMATED CR CAL POSITION FROM CO TS INCREASING EIGHTF0
* *xxxx*x**xxxxxx***xx* ********xx**xxxx*xxxxx M***xx**xxxxx {
NOTE *
- A 1/M plot may be intained during star if desired. *
****xxx**x*x*****x *xxx**xxxxx*******xxx x x x***** x** **x xx* x* x x ,/
- 1. Base Count Ra With Shutdown Banks t cps
- 2. Counts In ase Eightfold at Con ol Bank at steps.
l
- 3. Predic d Critical Position om BCB-2, Figure #9.
y ontrol Bank at steps CDH $T
- 4. NSURE the predicted ritical Control Bank Po tion in C.3 is AB0VE the Rod Insertion imits (i.e. CBC is grea than 47 steps
~
9O" 2 withdrawn). If t, DO NOT PROCEED WITH PROACH TO CRITICALITY. Th RCS Boron Co entration must be incre .,ed and the ECC recalculated.
. SRO Date fme , O
- 5. ENe RE the predicted Criti . Control Bank Position in .3 is BELOW i e Withdrawal Limit for D of 223 steps. If not, v NOT PROCEED WITH APPROACH TO CRIT LITY, The RCS Boron Con ntration must be reduced and the EC recalculated. / /
SRO Date Time D. REMARKS:
/ / /
/
/ /
/ / /
[(Final) j/ (7516P/0153P)
BCB-1 rigure 2 ! Revision O ' 8 48 06 120 188 208 248 298 526 568 490 446 5500 14 i,1 n i i
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! SANK 0 .-
i BnNK C 6 113 228 l ,
' 228 BANK 9 e III Control Bank Position (steps withdrawn) i i
! DIFFERENTIAL AND INTEGRAL ROD WORTH VS. STEPS WITHDRAWN. !
BANKS D, C, B, MOVING WITH 113 STEP OVERLAP, BOL, HFP, EQUILIBRIUM XENON CONDITIONS
! t 4
APPROVED OCT 101986 l ! , B.O.S.R. I i .,
Bcs-2 j Figure 5 j Revision i APPROVED s I 0.CT 10 $86 1 B. O. S. R. ie , 1
. . . . . .. i . .
i
. g
. , . . . . . . \ . s I t . J i .; : . . .. i . i ;, . . ;_ .1 4 J
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_.i_ i. s..:.__ \
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i 1500 ppm,_ i 9
.. . _ . . : -r . . .: ; . . . .: . . ;
( I ' t , . . g _ t- ' l t t R 1200 )p'in K r- l {, 1J i , e __ _ _
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g ... ! . ._ . . . _ _ _ _ __1. _ g L__4 q_._1 ; N g L. L._.L _ ;; . I I 4 i N .
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.. . . . . _ . . _ . . . . _A I' .
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' l i JiD 400 500 600 i 0 100 .' 0 0 F00
.! MODERATOP TEMPEPH1UFE IOf :
5 1 i
! l 1
1
; MODERATOR TEMPERATURE COEFFICIENTS l L VS. MODERATOR TEMPERATURE F
AT BOL AND EOL CYCLE 1, CZP TO HZP, ARO 1 l
. w y, =gg= e. .a.aw--'*.-.*=e .e ]
--_--____.__-,c, gm y ___gpm
4 i APPROVED scB-2 OCT 10 886 Fipg BA Revision 0 B.O.S.R. REACTIVITY INSERTION DUE TO XENON ; l 4 L VS TIME FOLLOWING PLANT TRIP 1 l
- 4000 I-~
^
- 3500 / \ -
. -- - 100% REACTOR i' J 5.- 1-( }'~'s \
rr x ,
.,, 9 0% PORER
- 3000 ') \ L -- '
J lf
!/
_\ ) ' L T L. - - " " _-
- 80% BEFORE ,___,
l
- 2500 if 7 -(s M\
n ; < t i - -- - _ 70% TRIP [ g . LkV i i x \ i l 60% )
^ ' <l ,' 'i -" ' "
- 2000 1 s _--- ')
J I 1JR"T\ .
- Ir \ l i \ L ,
e I; \ 't i \
- i500 'l i \i n 1 1 1 1 \ L r i L tit
\ \ \ ,\
~1000 \
\ \
\\',n I
\ \\' \
I 1 1 \\1
- 500 t,
\
i t\( o : l L <)il i l <\t 1 1LU\ J 0 t,.u\ j u, v l 1 \Gi I L, \ A I l500 u,u,uu T'. \'A 1'illi
, Link 1000 E
e, , I "l 1500 l m u 2000 %'t s w -% 2500 Y,a eh i
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- r. PBINCIELES OF NU.QLEAB_f0WER PLANT OEERATION 1 Page 25 '
l'UEBMODYNAMICS _UEAT TRBNSFEB_AND_ELMID FLOW ANSWER 1.01 (1.00)
- b. ( 1. 0 )-
j REFERENCE 1 Reactor Theory, Chapter 9, pgs. 13-14 Terminal Performance Objective 4 192008K112 ..(KA's) ANSWER 1.02 (1.50) (0.5 pts each)
- a. False
- b. False
- c. True REFERENCE Thermodynamics, Chapter 2, pg. 167 Terminal Performance Objective 6c 015000A101 015000K504 ..(KA's)
ANSWER 1.03 (1.00) (0.5 pts each)
- a. True *
- b. False REFERENCE Fluid Flow, Chapter 2, pgs. 27, 39 Terminal Performance Objectives 4, 5 191004K112 191004K115 191004K105 . . (KA's)
ANSWER 1.04 (1.00)
- a. (1.0) 1
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
e L . . I 1* PR1HCIPLEG_9E_ NUCLEAR POWER ELANT OPERATION t Page 26 THEBUQDYNAMICS, HEAT TBAESEER AND FLUID FLOW l , REFERENCE Westinghouse. Reactor Core Control for Large PWRs, pg. 9-10 192008K103 ..(KA's) AN5WER 1.05- (1.00)
- a. (1.0)
REFERENCE Nuclear Theory, Chapter 6, pgs. 11-17 Termina1' Performance Objective 5-d 192003K106 ..(KA's) ANSWER 1.06 (1.00)
- d. (1.0)
. REFERENCE Westinghouse Reactor Core Control for Large PWha, Chapter 4, pgs. 11-18' 192006K104 192006K103 001000Kb33 . .(KA's)
ANSWER 1.07 (1.00)
- a. (1.0)
REFERENCE , Nuclear Theory, Chapter 7, pg. 76 192008K123 . (KA's) ANSWER 1.08 (1.00)
- c. .(1.0) 1
(**44* CATEGORY 1 CONTINUED ON NEXT PAGE *t44*) I l 1
- 2. :EBIMgithES OF NUCLEAR POWEB_EhANT OEERATION 1 Page 27 l
THERMODYNAMICS t _ HEAT TRANSFER AND FLUID FLOW
-l
)
REFERENCE ! j Heat Transfer, Chapter 8, pgs. 23-24 ;
' Thermal Performance Objective 2 j 017020K501- 193009K105 (KA's) k ANSWER 1.09 (1.50)
- a. Maintaining control banks above the rod insertion limit. (0.5)
- b. (0.5 pts each)
- 1. Increases-
- 2. Decreases REFERENCE Reactor Theory, Chapter 7-, pgs. 7-9 Terminal Performance Objective 8 192002K114 ..(KA's)
ANSWER 1.10 (1.25)
- a. (0.25 pts each)
- 1. Constantly positive SUR
- 2. Steadily increasing count rate
- 3. No control rod motion (also accept no positive reactivity addition) -
- b. Supercritical. (0.5)
REFEFENCE Reactor Theory, Chapter 9, pg. 33 Terminal Performance Objective 3 Heat Transfer, Chapter 7, pg. 39 015000K505 292008K109 ..(KA's) L 1 (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) L______-_-__------- ,
Page 28
- 2. .' EBINC1ELES OF NUGLEAR_POWEB PLANT OEEBATION.
THERMODYNAMICS, HE6T TR&NSFER_dND FLUID FLOW { ANSWER 1.11 (1.00) The secondary system involves a phase change. (1.0) (A larger delta T in secondary is an acceptable answer for 1/2 credit) REFERENCE' Heat Transfer, Chapter 7, pg. 95 Terminal Performance Objective 4 002000K511 002000K501 .(KA's) ANSWER 1.1 'i (2.00) (0.25 pts each)
- a. Remains the same
- b. Remains the same
- c. Increases
- d. Decreases
- e. Remains the same
- f. Remains the same
- g. Decreases
- h. Increases REFERENCE Heat Transfer, Chapter 7, pg. 65-68 Terminal Performance Objective 9a !
000017K104 ..(KA's) ANSWER 1.13 (1.00) (0.5 pts each)
- a. Initial buildup of fission products in a clean core.
- b. Fuel depletion and fission products are only partially compensated by burnable poison rods.
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
1- -d81NCIELES OF NUCLEAR EQWER_EL&NI OPERAT_lgRm Page 29 ISE8dODYNAMigEt HEAT TRANSEEB_AND FLUID FLOW REFERENCE BCB, Figure 11 l Westinghouse Reactor Core Control for Large PWRs, pg. 5-10 192007K104 ..(KA's) ANSWER 1.14 (1.50) Superheated (0.75), 300 degrees-F (+/- 10)(0.75) REFERENCE Steam Tables (and Mollier Diagram) Thermodynamics, Chapter 3, pgs. 165-166 Terminal Performance Objective 9 193004K115 193003K125 .(KA's) ANSWER 1.15 (2.00) ( Any six e ;.33 pts each)
- 1. Pressure
- 2. Power ^
- 3. Flow
- 4. Temperature
- 5. Delta-I
- 6. Rod Insertion Limits
- 7. Rod Position (within 12 steps)
- 8. Rod Bank Overlap REFERENCE 9 Heat Transfer, Chapter 9, pgs. 66-66 19300BK105 012000K501 ..(KA's)
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
._ _-____________-__A
'PR1RQlELKE_OF NUCLEAR' POWER PLANT OPERATION, -Page 30:
~Q.
.IJijERMgDYN&MICS, HEAT TR&NSEER__AND FLUID ELOl ANSWER 1.16- (3.00)-
- a. -(The excess steam flow causes) T-avg to decrease [0.253-which inserts-c positive reactivity'[0.25], and' power increases [0.25). '(At the a POAH) increased power will increase temperature which inserts . .
' negative reactivity via doppler and MTC [0.25]. Power will stabilize
~ higher than POAH [0.5] and T-avg willEbe lower than the.no-load value
-(minus:the number of degrees.needed to overcome doppler) [0~5]. .
- - ' b '.
' Power ~ increase rate-is higher at EOi.,because.of: changes'in' beta-bar.
(MTC) [0.5]. Final. power is the :same [0.25] but T-ave will'be higher (closer to no-load temperature) because of a larger'MTC.(more
- negative)-[0.25]. (Note: The answer in b. depends on part a. ' Grade partlb.'accordingly')
REFERENCE Reactor Theory, Chapter 9, pg. 14
. Heat, Transfer,: Chapter 7, pgs. 42, 38
'041020K507 '192008K114 ..(KA's) 4
. ANSWER 1.17 (2.00)-
Fuel in the vicinity of the inserted rod experiences lower Xe (and Iodine) concentrations due to flux depression [0.5]. When the rod is withdrawn to group position, flux in the region increases markedly [0.5]. Xenon burns out due to higher flux-[0.3]. This all results in severe power peaking in-the region [0.7]. .(Partial credit of up to 1.0 pts may be given for an answer that talke .5out the rest of the core being at higher power because flux is supprest d in'the region with the misaligned rod.) REFERENCE Nuclear Theory, Chapter 6, pg. 26 192006K108 192005K110 000005K103 001000K507 001000A203
..(KA's)
' ANSWER 1.18 (1.25)
- a. The data taken will result in incorrect 1/M plot because time was not allowed for suberitical multiplication effects to increase the count rate. (.75)
- b. Overpredicting. (0.5)
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
EBINCJELES OF NUCLEAR _ POWER PLANI _QEEB&IlOMz Page 31 J. THERM 0pYNAMICS uSE61_I8&NSEEB_&BD_ELUID FLOW REFERENCE Nuclear Theory, Chapter 8, pgs. 53, 58 Terminal Performance Objective 4 192008K106 ..(KA's) l 1 J (***** END OF CATEGORY 1 *****)
- g. -
? f -Page 32
%__ELalfLl!ESIGN AMD12!#21MG SAEEI12&ND EMERGENCY
~ SleIEde:
E ANSWER 2.01- -(1.00) 1 Regulating letdown rate.to CVCS-(from downstream of the RHR heat. 1exchangers). (Also-accept controlled by PCV-131 as. alternate answer)
. REFERENCE System Description, RHR, Chapter'18, pgs. 38-39 004010K101 ..(KA's) i ANSWER 2.02 (0.00)
Deleted-REFERENCE Deleted ANSWER 2.03 (1.00)
..To'open CS001A(B),.its respective trains recirc sump suction isolation valve (CS009) must.be fully closed;[0.5]. This prevents inadvertently supplying a' drain flowpath-from the RWST to the containment recirculation sumps.[0.5).
REFERENCE System Description, Containment Spray System, Chapter 59, pgs. 26, 27 . I Terminal Performance Objective 8b
'026020K404 ..(KA's) .
ANSWER- 2.04 (1.50) j
- a. 102.75% (+/- .75%) (0.5) '
- b. Starts to open at 1115 psig (0.25) j Full open at 1175 psig (0.25) l
- c. "A local hand pump [0.1] is provided for manual operation [0.4]. l l
i 4 (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) ) 1 ____--______._________ _ _ _ - .h
- 1. ELAUI_DESlGN_ INCLUDING SAEEIX_&ND EMERGENCY Page 33 l
L SYSTEMS ] ( l 1 REFERENCE 1 System Description, Main Steam, Chapter 23, pg. 13 . Terminal Performance Objectives 3b, 3c 035010K602 ..(KA's) i I ANSWER 2.05 (2.00)
- a. (0.33 pts each) :
l
- 1. CV-110A shuts
- 2. CV-110B shuts
- 3. Boric Acid Transfer Pump Stops
- b. A line should be connected to a point between FCV-111B and the flow instrument [0.25] to a point between the VCT and CV-112B [0.25).
The line should include a flow control valve (FCV-110B) [0.5]. REFERENCE System Description, RMCS, Appendix A and Figure 15b-5 004020A401 (KA's) ANSWER 2.06 (1.50) (Any six @ 0.25 pts each, 0.05 pts are deducted for an incorrect setpoint, where applicable)
- 1. Overspeed, 5720 rpm *
- 2. Low bearing oil pressure, 10 poig
- 3. Low auto-stop oil pressure, 35 psig
- 4. Low Vacuum in Main Condenser, Zone C at 14" HgV ;
- 5. Hi-2 S/G water level (also accept P-14)
- 6. SI
- 7. Thrust bearing wear, 10 mils REFERENCE {
i System Description, Cond and MFW, Chapter 25, pg. 80 { Terminal Performance Objective 7a Ob9000K416 . (KA's) (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) i
.- __ __________J
1
. g. P.k&HLpKSIGN INRLUDING SAFETY AND _ EMERGENCY Page 34 n -
SYSTEfjS 1 i 1 ANSWER 2.07 (1.50) (0.25 pts each)
- a. Closed
- b. .Open
- c. Closed
- d. Open
- e. Open
- f. Open REFERENCE System Description, Serv. and Inst. Air, Chapter 53, pgs. 46-48 Terminal Performance Objective 8 078000K302 078000K301 ..(KA's)
ANSWER 2.08 (2.10)
- a. b. (0.5)
- b. (0.25 pts each) i
- 1. SI
- 2. 2/4 Low-Low S/G 1evel in any S/G -
- 3. Undervoltage on 2/4 RCP Busses
- 4. Undervoltage on Bus 141 (Sequenced)
Low AFW pump suction [0.4] coincident with any of bl, b2, c. or b3, above [0.'2]. REFERENCE System Description, Cond and MFW, Chapter 25, Figure 25-1B System Description, AFW, Chapter 26, pg. 41 Terminal Performance Objective 2, 3 061000K401 061000K402 061000K102 . (KA's) (***** CAT $ GORY 2 CONTINUED ON NEXT PAGE *****)
Page 35
. sm__ PLANT DESIGtJ __ INCLUDING SAFETY AND EMETfGEtJCY EXEIEde ANSWER 2.09 (1.50)
(Any six @ 0.25 pts each)
- 1. Manually from MCB
- 2. UF on RCP Bus l 3. UV on RCP Bus
- 4. Phase Overcurrent
- 5. Ground Overcurrent
- 6. Loop Iso Valve interlock not satisfied
- 7. Manually from RSP REFERENCE System Description, RCP, Chapter 13, Appendix A 002000K602 ..(KA's)
ANSWER 2.10 (2.40) See Figure 2-3A REFERENCE System Description, ECCS, Chapter 58, Figure 58-2 Terminal Performance Objective 9a, b
- 006020A301 00602CK404 .(KA's)
(***** CATEGORY 2 CONTINUED ON NEXT PAGE ****4)
- 5. PLANT _ DESIGN INCLUDIBG SAFETY AND_ EMERGENCY Page 36 SYSTEMS ANSWER 2.'11 (1,.50)
- a. .(any three at 0.25 pts each)
- 1. Phase A containment isolation
- 2. PZR Low Level (17%)
- 3. Loss of-air
- 4. Closure of CV-459 or CV-460
- b. Controls pressure upstream of PCV-131 high enough to prevent two-phase flow (saturation or flashing) [0.5] in the letdown flow orifices which could cause erosion [0.25]. (Also accept; Controls RCS pressure when solid, however a loss of solid pressure control must be refered to for full credit.)
REFERENCE System Description, CVCS, Chapter iba, pgs. 15-19 004010K403 004020K403 ..(KA's) l ANSWER 2.12 (2.00)
- a. 1, 2, 4 (1.0)
- b. Spent Fuel Pit Heat Exchanger (0.25)
- c. (0.25 pts each)
- 1. Low Pressure (85 psig)
-2. Return to power after blackout (also accept; UV on Bus 142)
- 3. SI REFERENCE System Description, CCW, Chapter 19, pgs. 12, 31 006000K401 008000K102 ..(KA's)
*****)
(***** CATEGORY 2 CONTINUED ON NEXT PAGE
1:E Page 37 EkaQI_ DESIGN INCLUDING SAFEIY AND EMERGENCY SYSTEME ANSWER 2.13 (2.00) See Figure 2-4A
. REFERENCE System Description, RCS, Chapter 12, Figure 12-1A Terminal Performance Objectives 2h System Description, CVCS, Chapter 15a, pg. 43 002000K109 002000K108 002000K106 ..(KA's) i ANSWER 2.14 (2.00)
- a. PMG output breaker must be tripped locally [1.0) (because control power is lost),
- b. Feed reg valves and bypasses fail closed [1.0]. (S/G level would rapidly drop to the Low S/G Level setpoint).
REFERENCE System Description, 48 and 125 VDC Power Systems, Chapter Ba, pg. 22 063000K302 063000K201 ..(KA's) ANSWER 2.15 (2.00)
- a. 125 VDC (ESF) Bus 111 (0.5)
- b. 1, 2, 3 (1.0) ,
- c. False (0.5)
REFERENCE System Description, EDG and Aux. Systems, Chapter 9, pgs. 47-52 Terminal Performance Objective 5 064000A401 064000K402 064000K104 .(KA's) (***** END OF CATEGORY 2 *****)
B. 1NSTRUMSUIS_AND CONTBQLS Page 38 q
' ANSWER 3.01 (1.00) l 1
(0.5 pts each) i
- a. DC hold cabinets
- b. 48, 64 (0.25' pts each) i REFERENCE System Description, Rod Control, Chapter 28, pgs. 27, 36,: 47, 60 Terminal Performance Objective 4b, be, 9 001000K402 001000K403 ..(KA's)
ANSWER 3.02 (1.50) (0.5 pts each)
- a. shunt trip
'b. reactor trip I
- c. 4 i
REFERENCE System Description, SSPS, Chapter 60a, pg. 20 System Description, RPS, Chapter 60b, pg. 21 012000K610 001000K603 ..(KA's) ANSWER 3.03 (0.75) l
- 1. SPDS (0.25) (also accept; iconics)
- 2. MCB (0.25) - safety grade (0.25) {
i REFERENCE l System Description, ICCS, Chapter 34b, pg. 13 002000K402 002000K107 ..(KA's) (v4*** CATEGORY 3 CONTINUED ON NEXT PAGE *****) I
N' T. Page 39
' lHSIRUMEHIE AND CONTROLS
-l 2
ANSWER 3.04 (1.00) [ a. Off 1 b. On j- c. On
- d. On REFERENCE 1 , .
System Description, DRPI., Chapter 29, pg. 21 Terminal Performance Objective 4e 001000K401 014000K403 ..(KA's) ANSWER 3.05 (2.00) (0.5 pts each)
- a. .S/G blowdown sample valves (PS 179A-D) close
- b. The outside air intake A dampers close (0.1), the makeup area unit fan (OVCO3CA) starts (0.1), the main control room turbine building air intake A dampers open (0.1), the recire charcoal absorber is placed in service (0.1), and the control room offices HVAC fans trip (0.1)
- c. Purge dampers (VQ001-5) close (Also accept Containment Vent Isolation)
- d. Sur6e tank vent valve (CCRCV017) shuts REFERENCE System Description, Control Room HVAC, Chapter 43b, res. 33-34 System Description, Process RMS, Chapter 49, pgs. 43-46 Terminal Performance Objectives 4a2, 4b2, 4b3, 4b5 008000K103 029000K403 072000K204 073000K101 (KA's) 1
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) L_ . - _ _ . . - _ _ - . _ _ _ - _ - _ _ - - - _ - _ - - - - . _ _ _ - _ _ _ _ _ _ . _ _ - _ _ _
L 8; Page 40 IUEIEMMENTS AHE_9QNISQLS i
-ANSWER 3,06 (1.00)
(0 25 pts each)
- a. P-14 or (S/G Hi-H1 Level) (Note: a and b may be interchanged)
- b. SI
- c. P-4 or (Reactor Trip)
- d. Low T-ave REFERENCE System Description, Cond z.nd MFW, Chapter 25, Figure 25-6 Terminal Performance Objective 6b Ob9000K419 ..(KA's)
' ANSWER 3.07 (2.75)
(0.25. pts each, setpoint or auctioneered counts 0.05 pts, if applicable)
- 1. Input to delta-flux for OP delta-T 2 .. Input to delta-flux for OT delta-T
- 3. Pwr Range Low Range Flux Level High Rx Trip, 25% power
- 4. Pwr Range Hi Range Flux Level High Rx Trip, 109% power
- 5. P-10 permissive, 10% power
- 6. Pwr Range Hi Flux Rate (+) Rx Trip, + 5% in 2 sec
- 7. Pwr Range Hi Flux Rate (-) Rx Trip, - 5% in 2 sec
- 8. C-2 Rod Stop, 103% power
- 9. Rod Control (Auct)
- 10. Feed-Reg Dypass Valve (auct) 11.. P-8 permissive, 30% power
(***** CATEGORY 3 CONTINUED ON NEXT PAGE 44***)
2; Page 41 IEEIERMEHIS AND~COMIBOLS . REFERENCE System Description, Pwr Rng NIS,. Chapter 33, Figure 33-1
. Terminal Performance Objectives S, 7 Olb000K103 015000K407 015000K405 015000K402 ..(KA's)
ANSWER 3.08 (2.b0)
- a. (0.26 pts each)
- a. Impulse Pressure
- b. Steam Flow
- c. Feed Flow (b and e are interchangeable)
- d. S/G Level.
- e. Auctioneered Nuclear Power
- b. Decreases [0.5) The delta-P between steam and feed increases beyond program. To reduce the delta-P MFP speed decreases [0.75).
REFERENCE System Description, SGWLC, Chapter 29, Figure 27-9, pgs. 29-30 Terminal Performance Objective i Terminal Performance Objective 15c 035010A203 035010K101 035010K401 ..(KA's) ANSWER 3.09 (2.00)
- a. P-7 is activated when power is below 10% [0.50) as indicated by 2/2 let stage turbine pressure [0.25) and 3/4 power ranges [0.25).
Alternate answer: Trips (in part b) are reactivated when indicating above 10% power by 1/2'P-13 inputs or 2/4 P-10' l inputs) l
- b. P-7 bypasses the following reactor trips:
- 1. High PZR Level (0.25) l l
- 2. Low PZR Pressure (0.25)
- 3. Turbine Trip (0.25)
- 4. Reactor Coolant Low Flow (0.25)
(Give 1/3 credit for each of RCP Bus, UF and UV, Rx Coo] ant l Pump Bhr Trip, as they make up the RCS Low Flow Trips) 1 (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****;
Page 42 O lueIWldEMIE_aHD CONIBOLS REFERENCE System Description, RFS, Chapter 00b, pg. 22 012000K610 012000K406. ..(KA's) ANSWER 3.10 (1.50) (0.3 pts for each of the five items)
- 1. 17% . letdown isolation, all heaters off, alarm
- 2. 5% below program - alarm
- 3. 5% above program B/U heaters on, alarm i
4, 70% - alarm
- 5. 92% - Rx Trip (if power > 10%)(2/3), alert (1/3), alarms REFERENCE System Descriptions, PZR, Chapter 14, Figure 14-2 Terminal Performance Objectives, 7d, 22, 23 011000A101 011000K406 011000K405 011000K401 011000K101
..(KA's)
ANSWER 3.11 (2.00)
- a. Power is lowered [0.5) at 200%/ min [0.1) for 1.5 sec [0.1) then waits for 28.5 secs [0.1] and repeats [0.1] until conditions are cleared [0.1].
- b. Clear conditions which could cause a Rx Trip. (0.5)
(Accept as alternate answer; Provide DNB and KW/ft protection)
- c. OF Delta T - 3% below setpoint (0.25) -
OT Delta T - 3% below setpoint (0.25) REFERENCE System Description, DEHC, Chapter 37a, pg. 152 Terminal Performance Objective 11-12 045000K412 .(KA's) 1 i i 1 (***** CATEGORY 3 CONTINUED ON NEXT PAGE ****d) l I __-_ - _ a
'- Page 43 l J. INSTRUMENTS _AND CONTROLS 1 ANSWER- 3.12 (2.00)
- a. By placing the PORV control mode switch selector switch in
" COLD OF ARM" -(0.5)
- b. A variable PORV lift setpoint is generated by input from auctioneered RCS wide range temperature [0.25]. The setpoint is compared to RCS wide range pressure [0.25] (one channel per PORV) and if RCS P is greater than the setpoint (per Tech Specs), the PORV lifts
[0 5].
- c. A deviation alarm (0.5) (For 1/2 credit accept; LO TEMP ARMED AND PURV ISOL VLV CLOSED alarm)
REFERENCE System Description, PZR, Chapter 14, pg. 42, Figure 14-13a, 13c Terminal Performance Objective 24 010000K403 ..(KA's) ANSWER 3.13 (2.50)
- a. (0.25 pts each, a-c are interchangeable)
- a. Steam Pressure Mode
- b. Loss of Load (Load Rejection) (C-7)
- c. Turbine Trip (C-8)
- d. C-9 (Condenser Pressure)
- b. Pressure setpoint controller [0.25] and Steam Header Pressure (PT-bO7) [0.25].
- c. If the error between Auctioneered T-ave [0.15] and No Load T-ave
[0.15] is greater than 15 degrees F [0.2]. d~. Six (of the twelve) dump valves snap fully open. (0.5) REFERENCE System Description, Steam Dumps, Chapter 24, pg. 20, Figures 24-1 and 24-10 Terminal Performance Objective 2g, 3a, bb, 8a 041020A405 041020A404 041020A408 041020K418 041020A102 041020K603 ..(KA's) (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
I l Page 44 L 3. INSTRUMENTS-AND CONTROLS ] l
- j i
i J F . ( ANSWER 3.14 (1.50) l I
- a. (0.5 pts each)
- 1. Low PZR Pressure, 1829 psig
- 2. Low Steamline Pressure, 640 psig
- b. .When 2/3 PZR Pressure signals are greater than 1930 psig. (0.5) _l REFERENCE System Description,.ESF, Chapter 61, pgs. 33-34 Terminal Performance Objectives 5 and 6 012000K604 012000K603 012000K610 ..(KA's)
ANSWER 3.15 (1.00)
- d. (1.0)
REFERENCE System Description, Rod Control, Chapter 28, pgs. 64, 65 Terminal Performance Objective 13 001000K403 001050K401 ..(KA's) l (***** END OF CATEGORY 3 *****) !
, 4. PBOCEDUEES - NgBM3L2_3DNQBMAL, EMEE9EBQY Page 45 L' - AHD._ESDlgkggigAL couTEgL. I . I
~ ANSWER 4.01 (1.00) a.
REFERENCE BEP-F.O 000007A104 ..(KA's) LANSWER- 4.02 (1.60) (0.5 pts each)
- a. True
- b. False
- c. False
' REFERENCE BGP 100-1, pgs. 3, 10, 24 193006K104 039000G013 005000G013 002000G013 001000G013
..(KA's) i ANSWER 4.03 (1.00)
- a. (1.0)
REFERENCE BRP 1000-A1, pg. 5 194001K103 ..(KA's) ANSWER 4.04 (2.00)
- a. The large circle signifies a major step in the procedure. (0.5)
- b. A step to be completed and initialled by the NSO. (0.5)
- c. A dotted line permits you to bypass that step. (0.5)
- d. A box requires prior approval before starting and SRO initials after completing the step. (0,5)
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
1
~
'4. EROCEDUBES -' NORMAL 2_ABE9EMAL1_EMEB9EUGY Page 46 l-
-4_ AUD_BADIOLoglDAL_CONThgL l
REFERENCE BAP 340-1, pg. 12 j 194001A102 ..(KA's)
' ANSWER- 4.05 (2.00)
(0.5 pts each) i l
- a. 2735
- b. Identified
- c. 15 minutes
- d. -550 REFERENCE Byron Technical Specifications, pgs. 1-3, 2-1, 3/4 1-6, 3/4 4-10 002000G005 010000G005 . (KA's)
ANSWER 4.06 (1.00)
- a. False (0.5)
- b. False (0.5)
REFERENCE BRP 1000-A1, pg. 23 194001K104 ..(KA's) ANSWER 4.07 (1.75) (0.25 pts each blank)
- a. 50, 100
- b. 350, 360
- c. 200
- d. 2, 100 REFERENCE i
BGP 100-5, pgs. 4, 5, 8 , 000024A205 000024K301 002000A103 002000A101 003000A404 i 003000A107 ..(KA's) (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
'4- ESQQEDUBES'- NOBM&Lx_aEHQBMALx_ EMERGENCY Page 47 e AND RaQlOLOG1 CAL COrlTSQL i ANSWER 4.08 (1.50) ,
- a. If after the rods are placed in MANUAL [0.25] and they are still moving [0.5].
- b. If unable to maintain PZR level > 4%. '( 0. 7 5.)
REFERENCE BOALROD-1, pg. 1 BOA PR1-1, pg. 4 012000A401 000001A105 ..(KA's) ANSWER 4.09 (1.50)
- a. (0.25 pts each)
- 1. Verify all turbine throttle stop valves closed.
l
- 2. Verify all turbine governor valves closed.
- b. (0.25 pts each)
- 1. Manual Trip
- 2. Runback
- 3. EH Pumps to Pull-to-lock
- 4. Steamline Isolation REFERENCE BEP-0, pg. 4 '
000007A107 000007A101 ..(KA's) ANSWER 4.10 (1.50) You probably have voiding in the reactor vessel or core [0.75] and you need to: (any 1 of 3 at 0.75 pts) !
- a. repressuriae to collapse them.
- b. hold pressure and cooldown.
- c. slow down the cooldown and start CRDM fans
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
r , r, . a v, q
.t y . y ,
"A. _PB99EDUBES - NORMAk. ABYQEM&m EMERGENCY,, .
Padg 48
- s ABI2_Bbp19 log 1ph CON'IROL, 4) /< 1,
-9 J,> , y y p.
3 REFERENCE d -
.. 1 BEP-ES-0.2, pg. 10
002000A403 000055K302 000055K101 ..(KA's) l ( n ANSWER 4.11 (2.00) ,- (
- a. 2 (0.5)
/
- b. To minimize RCS inyentory loss through seals (1.0)
- c. Manually operating. S/G PORVs (0.5) -
REFERENCE f {\ 1 BCA 0.0, pgs. 3, 4, 19, 20 t. , 000055A201 000055K302 ~000055G010
..(KA's) fr e
r is .g
\
ANSWER 4.12 (2.00) # ', #
< v ,
- a. Loss of Heat Sink due'to.Feedwater Isolation. (1.0)
- b. When automatic speed reach ((; 48 spm. {0.5) (Also accept until manual speed is greater than autoj
- c. RWST supplies charging pump suct' ion. (0.5) /d Y
? }p ,
REFERENCE f/ .g W j D 0 bK31 b00029K311 000029K310 000029K309 ..(KA's) j
~,
ANSWER 4.13 (1.00) ' 7 . g (0.5 pts each) e
- a. Rods are below the rod insertion limit (Lo-Lo RIL alarm).
- b. Emergency borate.
, p (Note: Answer b. is a result of answbr a., grader dise' ret 3pn is allowed if b. is correct for a wrong answer!a.) 3
- ! -, I; 1
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) l
\ '
_g
4 PERGEDUSES - EQ3Mah&_6FBQBMak&_EMERGEERY Page 49 L s AED RADlQLQ21g&L_gpjlIRQL L ( 1 o
/ rI ,
REFERENCE ,
- 1 ; /
Byron Technical Specifications, pg. 3/4 1-22 BOA TR1-2, pg. 1 00002.G011 001000G011 001000K504 000024K301 . . (KA's) t.) ,
, L ANSWER 4.14 (1.00) ,,
?? .c (0.5 pts each) ,'
l 4:
- 1. RCS subcooling unacceptable .r; L 2. PZR level low < 4% ,
h REFERENCE BEP-1, Foldout Page ' 000040A204 ..(KA's) < f ANSWER 4.15 (4.00) See (Ansver Attachment 4-1 A REFEREt4C(s f p 7 j Rescoor Theory, Chapter 7, Attachment ^1'
. Terminal Performance Objective 16 192008K107 .(KA's)
- ' /
s ( , e s. e i Vi 1 l' , (***** END OF CATEGORY 4 *****) (********** END OF EXAMINATION **********) ' ir .,
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[ N8C q wtRV) k (-] 2Bo 1x-n1 Revision 31 2BGP 100-7T1 REFERENCE REACTIVITY DATA WORKSHEET e**eneenene****en......****enee..e******e.....****e......... NOTE
- This worksheet will be retained as plant documentation.
- DO NOT DISCARD. Forward completed fom to the Operating
- j Staff when no longer required on shift for reference. Data *
{ may be taken from point history as required (NOTE in * ! remarks which corrputer *
* * * ... e **
- e ne n e n n e n e n e.* points were used) .
enenemene**eneseeen.......e****enemene 1.a. Shutdown Date: I I - 7. - N Shutdown Time: 0600 1.b. Control Bank b at b steps. 1.c. Inserted Control Bank Worth (From BCB-2, Figure #2A): (-) I500 pcm (2.5)@ 1.d. Critical Boron Concentration from (CIRCLE mode of analysis used): h most recent sample (2) calculation, or (3) sample after shutdown CB = 600 ppm 1.e. Power level prior to shutdown: 10 percent. l.f. Total Power Defect (From BCB-2, Figure #17A or BCB-2, Table 2-1): (-) 150 pcm (to) Q l.g. TorAL REACTIVITY (Step 1.c. + 3.f.): (-) 1650 pcm (cro)]
@ Poiai vaku. aceea<s in << box .
@ Wambew in (. g a m % k ) are a 4 h tl @
$Ok8fiMC8S .
@) it em appeacs in (pacenheA , A ',9 h YiumhPA $s cor(cci of wbcd AP P R O V E D ,'g g
< m opfo m p % % OmdidcA , tk fud OCT 031986 c,wEf b
@h B.O.S.R.
2BGP 100-7T2 Revision 51 2BGP 100-7T2 CALCULATION OF ESTIMATED CRITICAL CONDITION BASED ON A KNOWN BORON CONCENTRATION., Unit 2 Startup Number 87- 0$37 Startup Date il / 0 Z / Of7 Startup Time 16 0/) Time Interval Since Shutdown lC) Hours MMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMM M NOTE M M 2BGP 100-7T2 will be used to calculate Critical Rod *
- Position based on a known boron concentration. 2BGP 100-7T3M
- will be used to calculate Critical Boror. Concentration *
- based on a known rod position. 2BGP 100-7T2 will be *
- retained as plant documentation. DO NOT DISCARD. Forward *
- completed form to the Operating Staff for retention *
- when no longer required on shift for reference.
- MMMMMMMMMMMMMMMMMMMMMMMMMMMMHMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMM A. REACTIVITY CHANGES
- 1. Reactivity Change Due to Power Reduction:
O pcm - (-) 15 0 ocm(crd) Power defect Power defect at = (+) l$5Cl pcm (d45) ,1 at startup shutdown (RRD Step 1.f)
- 2. Reactivity Change Due to Temperature Deviation from Programmed Tave:
( 551 r- 557 F ) x (--) NA ocm/ r = ( >_ O pcm .I Tave expected Program Tave MTC from at startup at startup BCB-2, figure #5 6
- 3. Reactivity change due to Fission Product Poison:
- a. Reactivity Change Due to Sanarium Average power over 5 days prior to shutdown from 2BOS NR-1, Power History Hourly Surveillance 80 x Pav
(-) 666 pcm '2 - (-) 637 - ocm
; Sm worth at Sm worth at = (-) d5l pcm (cFD) b]_
startup time shutdown (time =0) for Pav from for Pav from BCB-2, BCB-2, Tb1 1-4 Table 1-4 , APFROVED OCT 031986 ' (7516P/0153P) E3. C). 55. Ft.
2BGP 100-712' Revision 51 2BGP 100-7T2 (Continued). CALCULATION OF ESTIMATED CRITICAL CONDITION L BASED ON A KNOWN BORON CONCENTRATION ,, ,
- b. Reactivity Change Due to Xenon [In order of preference, obtain from (1): Process Computer, (2): NIX tables, BCB-2, Table 1-3, or (3): RRD worksheet step 1.h]
- 2500' pcm -
O pcm Xe worth at - Xe worth at *g = (-) 2500 pcm (CFD) ,1 i
-s startup (503 shutdown (Zero
*g if BCB-2, Figure
#8A is used) ]
- c. Total Reactivity Change Due to fission Product Poisons.
.]
(-) $l pcm + (-) 2500 pcm 1 Samarium Xenon (CFD) = ( -) 2551 pcm (CFD) ,1 from A.3.a '4 (cro) from A.3.b ,g
- 4. Reactivity Change Due to Adjustments in Boron Concentration: ]
a
- a. Desired critical boron concentration at startup 500 ppm.
- b. Reactivity Change Due to Boron Adjustments.
50 0 ppm - 600 ppm x (-> 12,2 pcm/ ppm (. 2) Critical boron Boron conc. prior Differential Boron at startup to shutdown (RRD Worth from BCB-2, (A.4.a) Step 1.d) Figure #10A
~
= (+) 1220 pcm (cFD) h
- c. DETERMINE the amount of reactivity for which Control Banks must correct.
(+)150_pcm + ( ).Q_pcm + (-)?ssi pcm + (+) 12 20 pcm = (-) 1l81 pcm j A.1 ,g A.2 A.3.c A 4,b (AL} y 3 <l
- i o 5 PQg 5. Calculation of Estimated Critical Rod Position.
.O o 2O * *** " ""*"* " " * " **** " " "* * * " " * " " " " " " " *
$u <
- NOTE
- If the value for A.S.P is NEGATIVE, it indicates that 2 p1 O
- criticality cannot be achieved at the desired boron M
- concentration in A.4.a. RCS Boron Concentration must be N
- reduced and the ECC must be recalculated.
- MMMMMMVMMMMMMMMXMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMMM*pN
- a. Calculation of In;erted Rod Worth Required for Criticality.
(-) ll91 oc n - (-) 1500 pcm - (4) Slct ocm
. Control bank reactiv- Inserted control bank Inserted rod worth worth prior to shutdown required for 'j ity correction for criticality (A.4.c) (RRD Step 1.c) (CFD)-l criticality (CFD) .I (cFD) ,1 l i
l j' (7516P/0153P)
2BGP 100-7T2 Revision 51 2BGP 100-7T2 (Continued) CALCULATION OF ESTIMATED CRITICAL CONDITION BASED ON A KNOWN BORON CONCENTRATION..
- b. Using BCB-2, Figure 2A, DETERMINE the control bank position I corresponding to the Inserted worth calculated in A.S.L above.
Control Bank "D at 11 3 steps (C F D) .5 B. ESTIMATED CRITICAL CONDITION
SUMMARY
Present Boron Concentration 600 ppm Estimated Critical Boron Concentration 50 0#13 . ppm ( A.4.a) Estimated Critical Bank Position CB 2 at _ steps ( A.S.b) (20sf95)fj 1 ECC performed by NAME DATE ECC reviewed by NAME DATE 4 C. ESTIMATED CRITICAL P 'ITION FROM COUNTS INCREASINJ IGHTFOLD. I
*****x*xxx x x x x x x* **
- x x***** x x* * ** x xx*d* x x x* x x*
- x
- x x x x* x x x* x x
- NOTE *
- A plot may be maintained du,ri g startup if desired.
- g
**x xM*x x* ** **x ***** x xx
- wM* x* M*- Nx x****xx x x x**x x*x xx* x x x x x* x* x* f-
- 1. ase Count Rate With Shutdo Banks Out cps
/
Counts Increase Eightfp at Control Bank at steps. i 3. Predicted Critical osition From BCB-2, Figure #9.
!) Control Bank at steps Q)H $1
- 4. ENSURE the redicted critical Control Bank Positi the Rod sertion Limits (i.e. CBC is greater in C.3 is ABOVE n 47 steps 9O 2 withdr n). If not, DO NOT PROCEED WITH APP,R0ACH TO CRITICALITY. The RCS oron Concentration must be increased 'nd the ECC recalculated. -
i p m "$ m
/
SRO Date Time
/
O ENSURE the predicted Critical ntrol Bank Position in C.3 i / ELOW the Withdrawal Limit for CP of 223 steps. If not, DO NO ROCEED WITH APPROACH TO CRITIC ATY. The RCS Boron Concentra 'on must be reduced and the ECC r alculated. / / 0 Date Time D. REMARKS: . / l s / / / l (
/ / / l J !
3 j (Final) 1 { (7516P/0153P)}}