ML20212D850
| ML20212D850 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 07/28/1986 |
| From: | Brockman K, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20212D722 | List: |
| References | |
| 50-259-OL-86-01, 50-259-OL-86-1, NUDOCS 8608120577 | |
| Download: ML20212D850 (175) | |
Text
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ENCLOSURE 1 EXAMINATION REPORT 259/0L-86-01 Facility Licensee: Tennessee Valley Authority 1101 Market Street Chattanooga, TN 37402-2801 Facility Name: Browns Ferry Nuclear Plant Facility Docket Nos.: 50-259, 50-260, and 50-296 Written and oral examinations were administered at the Browns Ferry Nuclear Plant near, Decatur, Alabama. Simulator' examinations were administered at the Power Training Operations C nter near, Soddy Daisy, Tennessee. Chief Examiner: w{bwk s Wl24 Date Signed KefE.Brodk / Approved by: John F. Muhro, Actin 6$ection Chief b~h u Nkk Date Signed Summary: Examinations on June 16-19, 1986 Written examinations were administered to 6 candidates, 5 of whom passed. Oral examinations were administered to 6 candidates, 5 of whom passed. Simulator examinations were administered to 6 candidates, 5 of whom passed. Based on the results described above, 5 of 5 R0s passed and 0 of 1 SR0 passed. 060012o577 e60001 PDR ADOCK 00000259 V PDR
REPORT DETAILS ,
- 1. Facility Employees Contacted:
*C. H. Noe, Division of Nuclear Training
*A. R. Champion, BFNP' Operations Training ;
*E. S. Howard, BFNP Operations Training :>
*N. S. Catron, BFNP Simulator Training'- , .
*W. J. Glasser, Division of Nuclear Training
*R. J. Moll, BFNP Operations Training
*L. H. Sain, Division of Nuclear Training
*R. Dawson, BFNP Simulator Training
- Attended Exit Meeting
- 2. Examiners: 7, ,
*K. E. Brockman, RII '
n W. C. Cliff, PNL - G. A. Sly, PNL '
- Chief Examiner
- 3. Examination Review Meeting e
At the conclusion of the written' examinations, the examiners provided . A. R. Champion, with a copy of the written examination and answer key for review. Enclosure 3 includes. the comments made by the facility reviewers; NRC resolutions to these comments are listed below,
- a. R0 Exam and SRO Exam !
Question 1.01/5.01 NRC Resolution: Comment acknowledged. Sufficient guidance is provided in the existing answer key to encompass the facility concern. Answer key was modified to include the alternate wording for future exams. Question 1.02/5.02 NRC Resolution: Comment acknowledged. The facility comment is a more elaborate explanation of the effect of delayec neutrons and was not required for full credit. The existing answer key contains a more general descrip-tion of the transient effects and was not changed to require the more specific explanation provided by the facility. No change to answer key. L
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, e Enclosure 1 2 August'1, 1986 Question 1.03/5.07b NRC Resolution: Comment not accepted. The facility comment. states that "the void coefficient may increase, decrease, or remain the same" and provides Figure 4-10 as a reference which supports the facility position.
But when the situation of question 1.03b is considered '(100% power), the percent voids will remain relatively constant from BOL to EOL and, according
.to Figure 4-10, would become less negative. Therefore, the comment is denied and the answer key remains as is.
Question 1.04-NRC Resolution: Comment accepted. The facility comment is already inferred ' l in the existing answer key. Answer key was modified to not require the radial distribution response. Question 1.05 NRC Resolution: Comment not accepted. G0I-100-1 step 5, states to " observe the period meter when pulling rods and govern withdrawal rate to avoid having a period shorter than 6,0 seconds." This step directs the operator to maintain a period greater than 60 seconds and is the administratively
" minimum permissible -sustained positive period permitted." Reference to any periods less than 60 seconds (e.g., 30 seconds) is not permitted by GOI-100-1 and requires it to be immediately corrected to greater than 60 seconds.
Question 1.08d NRC Resolution: Comment acknowledged.~ Answer key was modified to read that the temperature must correspond to the saturation temperature for the ' pressure given, i Question 1.09a , NRC . Resolution: Comment acknowledged. The facility comment is already inferred in the existing answer key. Answer key was not modified. Question 1.09b c NRC Resolution: Comment accepted. Answer key modified to read: i j (b) (2) increase (3) decrease 1 1 7 l 4
s . i- ' Enclosure 1 3 August 1, 1986 Question 1.14/5.13 NRC Resolution: Comment n'ot accepted. While "several hours" is not totally explicit, to address a period of time lasting for over five hours would require discussing both rod withdrawal and insertion. Only answer "a" is possible from time = 5 hours after shutdown. No change to exam or answer key. Question 2.01 NRC Resolution: Comment accepted. Answer key was modified to reflect the facility position. Component 1, 3, 4, and 6, 2 could occur in any consec-utive order due to being parallel paths. , Question 2.03 s NRC Resolution: Comment acknowledged. If the candidate indicates that he misinterpreted the wording of the question due to a typographical error, the response will be evaluated accordingly. Answer key not modified. Question 2.05a NRC Resolution: Comment not accepted. Facility resolution would not be a
. proper response to the question without referring to the explosive potential
~ of excessive hydrogen concentration. Inerting alone does not preclude the
, generation of hydrogen following a LOCA. Answer key was not modified.
Question 2.05b NRC Resolution: Comment accepted. Typographical errors corrected to change i . "condensible" to "non-condensible." Question 2.06b NRC Resolution: Comment acknowledged. The answer key already contains the facility comment. No changes made. Question 2.06c NRC Resolution: Comment accepted. Concentrations 660/1075 will be changed to 600/750 respectively to reflect proper Browns Ferry values. Also the reference to "RHR dilution" was deleted and full credit assigned to
" imperfect mixing."
Question 2.09b
+
NRC Resolution: Comment accepted. Answer key was expanded to accept any combination of Hi-Hi-Hi and Dewnscale/Inop occurring simultaneously in both trip channels; also, Loss of Power to (the solenoid) FCV 66-28 was added to the key. 4 August 1, 1986 Question 2.15 NRC Resolution: Comment accepted. Due ' to wording of the question, two answers ("a" and "d") are correct. Answer key modified to accept either response. Question 3.01/6.01 NRC Resolution: Comment accepted. Due to the technical clarification given to candidates by the proctor, the answer key was required to be modified to read: 3.01(a) used for trip functions 3.01(b) not used for trip functions Question 3.05/6.05a NRC Resolution: Comment acknowledged. The facility concern is already inferred in the existing answer key. Answer key was not modified. Question 3.07a/6.12a NRC Resolution: Comment accepted. Answer key was modified to accept alternative wording. Question 3.12 NRC Resolution: Utility comment acknowledged. The operator should know that the ball valve will not close until the in-shield limit switch is closed (i.e., the TIP is fully withdrawn). This is not detailed circuitry knowledge, and is considered appropriate for both the R0 and SRO. Answer key, however, was modified to reflect more accurate partial credit. Question 3.13b NRC Resolution: Comment accepted. Answer key was modified to "YES [+0.5], because there are greater than 13 operable LPRM inputs in APRM C [+0.5]." Question: 4.01a/7.07a NRC Resolution: Comment not accepted. Deenergizing of the 250 VDC breaker is an important safety function in locking the valve in the open position, and is necessary for full credit. Answer key was not modified. Question 4.01b/7.07b NRC Resolution: Comment accepted. Part "b" of question deleted. Question 4.04c/7.01c NRC Resolution: Comment accepted. Answer key was expanded to include "but do not exceed 100 mrem /hr".
~
Enclosure 1 5 August 1, 1986 Question 4.05/8.01 NRC Resolution: Comment accepted. Item "e" changed from "not required" to
" required". -
Question 4.07/7.14 NRC Resolution: Comment not accepted. " Check RFPT and main turbine on turning gear" is a NOTE and not an Immediate Action step, which means it is not required prior to exiting the control room. No credit will be given or taken off for inclusion. Answer key was not modified. Question 4.08/7.03 NRC Resolution: Comment accepted. The following changes were made to the entry condition setpoints: l' to 1", 6.25' to 6.25", and 165 F to 160 F. Question 4.09 NRC Resolution: Comment accepted. Question deleted. Question 4.10/7.11 NRC Resolution: Comment not accepted. Question and Figure are referenced to Page 2 of 5, PC/P Unit 1, dated 4/16/86 which is the latest revision received. There was a typo in question, " torus" versus " suppression chamber", but this shouldn't effect the proper response. Facility comment is without merit and proper justification for requested answer key modifi-cation. The vacuum breakers are one of the design limitations in the development of the "Drywell Spray Initiation Pressure Limit" curve. The upper portions of the curve are still based on the "old" limits and the spray flow rates are still based on maximum rated (9,000 to 10,000 gpm) flow. Credit will be given if candidate can provide alternate reasons for negative pressure, but they must be plausible for full credit. Answer key was modified to reflect the above alternate response. Question 4.12/7.13 NRC Resolution: Utility comment accepted. Due to the utility's changes to the E0I Entry Conditions, the nur setpoints will be placed in the answer key. Note that the setpoints ar still required for full credit. Question 4.15/7.04 NRC Resolution: Utility comment accepted. The response provided is more direct under the bases of C-5, vice RC/L, of the BWROG Guidelines. The reason for ADS inhibition is direct and specifically addressed. Answer key modified to reflect new correct response. i
Enclosure 1 6 August 1, 1986 Question 4.16b NRC Resolution: Utility comment is essentially a repeat of the answer key. The key words are " directed to take action." No change to answer key required.
- b. SR0 Exam only Question 5.05 NRC Resolution: Comment accepted. Answer key modified rated core flow from "108.5" to "102.5".
Question 5.07c NRC Resolution: Comment not accepted. Page 4-17, Figure 4-5 (G.E. Reactor Theory and page 3 of 17 (BFN-HLT) both state that the void coefficient becomes less negative with decreased void fraction or decreased fuel temperature. Due to the wording of the question the candidate would have had to assume that all parameters except power level remained constant, or specifically state his/her assumptions (as directed to by proctors). Because of the contradicting positions between the material, the facility comments, and the wording of the question, the answer key has been changed to accept the response for Moderator Temperature Coefficient as the only valid item in part "c". This reduces the points for part "c" to 0.25 and the question to 2.25 from 1.0 and 3.0, respectively. Part "c" of the answer key is now: (c) alpha (mod) becomes less negative (+0.25) Question 5.08 NRC Resolution: Comment not accepted. While it is admitted that the referenced points A, B, & C were not on the diagram, if a candidate did not understand a question he was directed on a number of occasions to ask the proctor for clarification. The question was also written so that the reference to specific paint location would not hamper from the understanding or answering of the question. Answer key was not modified. Question 5.09a NRC Resolution: Comment not accepted. No formulas are necessary to answer the question. If candidate did not make the assumption that all flow from the tank stops at 5 feet of head, he should state when he did assume flow to stop and respond accordingly. Answer key was not modified. Question 5.10 NRC Resolution: Comment acknowledged. Facility concern was already inferred in existing answer key. Answer key was not modified. l l
._ . . _ _ _ __ .. 7 August 1, 1986 Question 6.02a NRC Resolution: Comment accepted. Answer key was modified so as not to require the inclusion of < 130 psid across the valve "for full credit".
Question 6.02b NRC Resolution: Comment accepted. Answer key was changed to accept the facility comment, which provided a more direct wording than the answer key, as the new correct response. Question 6.03e NRC Resolution: Comment not accepted. The reason why the reactor water is on the tube side of the heat exchanger is not only a general design characteristics, but more importantly, a plant specific characteristic. The fact that the low pressure heater tubes pass through the condenser (shell side) is a major system design criteria and relevant to reactor operation. Answer key was modified to accept the above alternate response for the reason "WHY". Question 6.04 a/c NRC Resolution: Comment not accepted. Questions are within the scope of current 10 CFR and Examiner Standard practices. Answer key was not modified. Question 6.10a NRC Resolution: Comment acknowledged. Answer key was modified to allow alternate name of " Master Controller" in addition to " Speed Demand Limiter". Question 7.06 NRC Resolution: Comment accepted. Answer key was changed to reflect proper time period. Question 7.10b NRC Resolution: Comment accepted. Reference to " routine RWP" was deleted from the answer key; the other two responses will be worth 0.5 points each instead of 0.33 points each. Question 8.02 NRC Resolution: Comment acknowledged. Answer does not require specifying which S. I. is to be performed.
Enclosure 1 8 August 1, 1986 Question 8.03 NRC Resolution: Comment not accepted. The candidate was told (as the utility states) that the "a" portion of the reference material for question 8.03 had been removed from the SR0 handout because it provided an answer to another questior. (8.01). This was done prior to the start of the examination and the cindidate was also told that "if he had a problem with answering the question without the material he should inform the proctor and something would be worked out." The time factor referred to by the facility is of little consequence in that the candidate completed the examination prior to the required time limit. The content of the entire question is within the scope of the current licensing standards. Answer key was not modified. Question 8.06 NRC Resolution: Comment acknowledged. The candidate had enough information to respond to the question correctly. The fact that Table 3.7.A (pages 250-255) of the TS was missing is moot, in that the question stated that the valve was an isolation valve (and the candidate would have wasted valuable time verifying the accuracy of this statement). The deletion of page 243, which contained a memorizable action (i.e., close the other isolation valve) was an oversite by the examiner, but shouldn't have effected a valid response. The candidate is responsible to be knowledgeable of the PCIS Technical Specification LCOs and the consequences of exceeding these LCOs. This was the emphasis of the question and the candidate had sufficient
-information at his disposal to answer the question properly. Question stands as is but the answer key was modified to allow for partial credit if candidate demonstrated TS knowledge by addressing only the HPCI system.
Question 8.07
- NRC Resolution
- Comment not accepted. The question is within the scope of l the required memorization knowledge found in 10 CFR 50.72, 10 CFR 50.73, and l NUREG-1021 (i.e., one hour reportable items). Answer key was not modified.
1 Question 8.08 f NRC Resolution: Comment not accepted. Upon review of the referenced ! material, facility comment as well as answer key were found to be incorrect. The correct response is: l r (a) TRUE (per TS) FALSE (per BF 12.24, 3SRO, SRO) (b) FALSE (per TS, ISRO, 2RO) FALSE (per BF 12.24, 2SRO, 4RO) i i
- c. 3 i
l I Enclosure 1 9 August 1, 1986 l Part "b" of the answer key has been corrected to read FALSE. Since part "a" of the question places the candidate in a position where the SP and the TS's ^ provide conflicting guidance, the question will be deleted unless the candidate states which reference he is referring to, thereby reducing the total point value of the question to 0.5 from 1.0. Question 8.09 a/d NRC Resolution: Comment not accepted. The examiner feels that sufficient information is provided in the wording to elicit a proper response for all parts of the question. The fact that the exact terminology " maintenance tag" is, or is not, used at Browns Ferry does not distract from the concept or intent of the question. Answer key was not modified. Question 8.10 NRC Resolution: Comment Accepted. Answer key was modified to refer to any of the items on Attachment 1 of OSIL 34 as acceptable responses for credit. Question 8.13 NRC Resolution: Comment acknowledged. Alternate wording for part "b" was added to the answer key. Question 8.14 NRC Resolution: Utility comment accepted. The question's wording does allow for the interpretation described by the utility. The alternative responses requested will be allowed. Answer key modified to reflect additional responses. Question 8.16 NRC Resolution: Utility comment acknowledged. The answer key is adequate to grade the question - it provides sufficient guidance to cover both situations described by the utility. Also, no automatic functions are discussed in the key. The utility cautions that " care must be taken to accept an answer based upon the question and not word for word phrases ...
. Likewise, the utility should not look at the answer key as a " bean count", but as a guide to ensure equitable and consistent evalua~ ion of examinee's demonstrated knowledge. No change to answer key required.
Question 8.05 NRC Resolution: During the grading of the examination the answer to this question was reevaluated. If all (three) pressure taps on one calibrated jet pump were to fail, that jet pump would fail the "10% delta pressure for lower plenum to diffuser pressure versus the average for all jet pumps" surveillance. It would not pass this surveillance as stated in the answer key. For this reason the answer to question 8.05 has been changed to read "No (+0.5), due to the failure of the single jet pump to average jet pump pressure surveillance requirement (+1.0)." I L
n nclosure 1 10 August 1, 1986
- 4. Exit Meeting At the conclusion of the site visit the examiners met with representatives of the plant staff to discuss the results of the examination.
There were three generic weaknesses noted during the oral examination. The areas of below normal performance were:
- a. Fire Protection / Equipment
- b. Escort Duties
- c. Refueling floor equipment, duties, and operations The cooperation given to the examiners and the effort to ensure an atmosphere in the control room conducive to oral examinations was also noted and appreciated.
The licensee did not identify as proprietary any of the material provided to or reviewed by the examiners. l l 1 t (
c
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U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: BROWNS FERRY 1, 2&3 REACTOR TYPE: BWR-GE4 DATE ADMINISTERED: 86/06/16 EXAMINER: CLIFF, W. APPLICANT: 8as he /S/v'
/ /
INSTRUCTIONS TO APPLICANT: Use separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY
.W:.U- 25.4
. 25dtf 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 25.00 25d6 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS -
- r.e e AU E520 2P:tf5 3. INSTRUMENTS AND CONTROLS es. . as.r 25de 2540 4. PROCEDURES - NORMAL, ABNORMAL,
! EMERGENCY AND RADIOLOGICAL CONTROL GVf 100coff 100.00 TOTALS FINAL GRADE % l All work done on this examination is my own. I have neither given nor received aid. APPLICANT'S SIGNATURE l
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
- 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil only to facilitate legible reproductions.
- 4. Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers. _
- 7. Print your name in the upper right-hand corner of the first page of 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 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.
- 10. Skip at least three lines between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly used in facility literature.
- 13. The point value for each question is indicated in parenthesis after the question and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must-be done after the examination has been completed.
- 18. When you complete your examination, you shall:
- a. Assemble your examination as follows:
(1) Exam questions on top. (2) Exam aids - figures, tables, etc. (3) Answer pages including figures which are a part of the answer.
- b. Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn in all scrap paper and the balance of paper that you did not use for answering the questions.
- d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
d l
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 2 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.01 (2.00)
Regarding the xenon transient following a significant decrease in reactor power from high power operation:
- a. If the decrease in reactor power was from 100% to 50%, WHY is the equilibrium xenon reactivity more than one-half the 100% equilibrium value? (1.0)
- b. WOULD the resultant peak from a 100% to 50% power maneuver occur in the same time as the result peak from a 100% to 0% power maneuver? EXPLAIN your answer. (1.0)
NOTE: Consider all production and removal mechanisms in your answer. _, QUESTION 1.02 (1.00) The reactor is critical at 10E+6 cps. A stable period of 60 seconds is achieved. If rods are inserted continuously until the period meter decreases to infinity and then the rod insertion is immediately stopped, WILL the reactor be (CRITICAL, SUPERCRITICAL, or SUBCRITICAL) in the time following the rod stoppage? EXPLAIN your response. (1.0) l l 1
- i l
l \ . l (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.03 (2.50)
You are currently operating at 100% power BOL when you lose partial feedwater heating.
- a. If the STA tells you that reactor coolant temperature i decreased by 10 deg. F, voids decreased by 2%, WHAT I WOULD BE the corresponding temperature change to the fuel temperature? (ASSUME no rod movement, no recirculation flow changes and the reactor reactivity returns to zero.) (1.0)
- 1. Increase by 30 deg. F l
- 2. decrease by 30 deg. F
- 3. increase by 300 deg. F
- 4. 1 m. . _ by 300 deg. F -
1 nec nut.
- b. If the same situation were to occur at E0L, WHAT would be the corresponding reactivity changes (MORE NEGATIVE, LESS NEGATIVE, NO CHANGE) to each of the coefficients
[i.e., delta (T) mod; delta (% voids); delta (T) fuel]? (1.5) QUESTION 1.04 (2.00) - DEFINE what is meant by each of the following control rod descriptions, and EXPLAIN how each of these two types of control rods affect the core axial flux distribution and/or power when they are moved.
- a. shallow rods (1.0)
- b. deep rods (1.0)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****) h
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 4 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.05 (2.00)
The reactor is critical at "50" on Range 2 of the IRMs. A control rod is withdrawn three notches, resulting in a power increase with a stable reactor period equal to the minimum permissible sustained positive period permitted in GOI 100-1, Plant Start-Up. Heating power (POAH)is estimated to be at "30" on Range 7. SHOW all your assumptions and work for the following calculations. (NOTE: Assume you range up on the IRMs according to procedures.)
- a. WHAT is the doubling time? (1.0)
- b. HOW long will it take to reach heating power? (1.0)
QUESTION 1.06 (2.00) STATE whether the following failures would INCREASE, DECREASE, or NOT CHANGE overall plant efficiency? ASSUME plant is at 100% rated condition.
- a. loss of one (1) high pressure feedwater heater (0.5) b.10% mass flow tube leak in the regenerative heat exchanger of the RWCU (0.5)
- c. loss of all control rod drive cooling water flow (0.5) .
- d. SRV lift. (0.5)
NOTE: Consider thermodynamic effects only. QUESTION 1.07 (1.50) Increasing recirculation pump speed will cause WHAT change (INCREASE, DECREASE, or REMAIN THE SAME) in each of the following parameters? (Assume normal operating conditions.)
- a. actual bundle power (0.5)
- b. critical power (0.5)
- c. critical power ratio (0.5)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 5 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.08 (2.50)
During your Shift, an SRV inadvertently opens from 100% power and 1000 psia. Use a Mollier Diagram or the Steam Tables to answer the following: NOTE: ASSUME A SATURATED SYSTEM AND INSTANTANE0US HEAT TRANSFER
- a. STATE the tailpipe temperature, assuming atmospheric pressure in the Suppression Pool and No Reactor Depressurization. (0.5)
- b. If the Suppression Pool Pressure were to start INCREASING, STATE whether the Tailpipe Temperature would INCREASE, DECREASE, or REMAIN THE SAME as compared to the temperature -
you calculated in (a.) above. (0.5)
- c. If the reactor starts depressurizing when the SRV is opened, STATE whether the Tailpipe Temperature will INITIALLY INCREASE, DECREASE, or REMAIN THE SAME, as compared to the temperature you calculated in (a.) above. (0.5)
- d. STATE the Reactor Pressure and Temperature at which the Tailpipe Temperature would be at its MAXIMUM value (during thedepressurization.) (1.0)
QUESTION 1.09 (2.00) A minimum Net Positive Suction Head (NPSH) is required for assuring that the Recirculation System pumps do not cavitate.
- a. PROVIDE a brief definition of NPSH. (0.5)
- b. STATE how the AVAILABLE NPSH changes (INCREASES, DECREASES, or REMAINS THE SAME) for each of the following:
- 1. Reactor water level decreases from normal level to just above the low level scram setpoint. (0.5)
Feedwater heating is lost.
- 2. (0.5)
- 3. Recirc. pump speed is changed from 40% flow to l 80% flow. (0.5)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****) ; l
x
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 6 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.10 (2.00) i The attached figure represents a transient that could ocur at a BWR, ico Te Given: An inadvertent initiation of HPCI at % power No other operator actions are taken.
Recorder speed: 1 division = 1 minute EXPLAIN the cause(s) of the following recorder indications: (2.0)
$ %\ \ Mwr
- a. Gece flow 88cfMR9E at point
- b. NO ~ w ike=ai. peint J15Ed55 T the p;;;;a,c :pik: :t
; & t - ww . - at pc.h F i r s e, .. _. - u t 49 a)t
- c. Reactor vessel level INCREASE at point #13
- d. Total feedwater flow-DeeiEEE at point HtMb 9 3 -
pct .s ,._. QUESTION 1.11 (1.50) STATE for the following conditions, whether pump amperage would INCREASE, DECREASE, or STAY THE SAME.
- a. the pump discharge valve is slowly throttled closed (0.5)
- b. increase in inlet subcooling (0.5)
- c. rotor lock-up (0.5)
QUESTION 1.12 (1.00) Which of the following radioactive isotopes found in the reactor coolant WOULD NOT indicate a leak through the fuel cladding. . (1.0)
- a. Co - 60
- b. Xe - 133
- c. I - 131
- d. Kr - 87
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
0
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 7 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.13 (1.00)
Which of the following correctly describes the Maximun Fraction of Limiting Power Density (MFLPD)? (1.0)
, a. LHGR-actual / LHGR-limit ; must be maintained ( 1
- b. LHGR-limit / LHGR-actual ; must be maintained > 1
- c. LHGR-limit / LHGR-actual ; must be maintained ( 1
- d. LHGR-actual / LHGR-limit ; must be maintained > 1 QUESTION 1.14 (1.00)
The reactor trips from full power, equilibrium xenon conditions. Four (4) hours later the reactor is brought critical and power level is maintained on range 5 of the IRMs for several hours. WHICH of the following statements is CORRECT concerning control rod motion during this period? (1.0)
- a. Rods will have to be withdrawn due to xenon build-in.
- b. Rods will have to be rapidly inserted since the critical reactor will cause a high rate of xenon burnout,
- c. Rods will have to be inserted since xenon will closely follow its normal decay rate.
- d. Rods will approximately remain as is as the xenon establishes its equilibrium value for this power level.
, QUESTION 1.15 (1.00) As part of the scram procedure, the operator is directed to i insert the SRM's and IRM's. EXPLAIN how these systems could be used to provide a crude l indication of water level if level could not be confirmed by normal instrumentation. 1 1 1 (*****ENDOFCATEGORY01*****)
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 8 QUESTION 2.01 (3.00)
With regard to the Condensate and Feedwater System:
- a. Other than normal feedwater, NAME four (4) major systems that are provided a flow path to the reactor through the feedwater system. (1.0)
- b. PUT the following components in the order in which they occur going from the main condenser hotwell to feedwater spargers. (2.0)
- 1. steam packing exhauster condenser
- 2. startup bypass valve
- 3. off gas condenser
- 4. steam jet air ejector condensers
- 5. high pressure feedwater heater strings -
- 6. reactor feed pumps
- 7. condensate booster pumps
- 8. low pressure feedwater heater strings QUESTION 2.02 (2.00)
An automatic HPCI initiation has occurred. Subsequently HPCI injection was automatically terminated due to high reactor water level (+54 inches).
- a. WHAT valve (s) auto close on this HPCI turbine trip? (1.0) -
- b. Assuming no operator action, HOW WILL HPCI respond to a subsequent decreasing water level (to low low level, -51.5)? (0.5)
- c. If the HPCI system had switched sources from the CST to the torus due to low CST level and the CST level subsequently recovers, WILL the HPCI system automatically switch back to the CST suction? (0.5) l l
(***** CATEGORY 02CONTINUEDONNEXTPAGE*****)
- 2. PLANT DESIGN INCLUDING SAFETY k~DN EMERGENCY SYSTEMS PAGE 9 QUESTION 2.03 (1.00)
The following concerns the Browns Ferry Raw Cooling Water System.
- a. LIST the Raw Cooling Water Suction head interconnections for Units 1, 2, and 3. (0.5)
- b. WHAT is the purpose of the sodium hypochlorite injected into the raw cooling water by the sodium hypochlorite system? (0.5)
QUESTION 2.04 (2.25) - - - LIST the three (3) modes of Condenser Circulating Water (CCW) System operation and DESCRIBE the basic flow paths. (2.25) QUESTION 2.05 (1.50) Concerning Primary and Secondary Containment Systems:
- a. WHY are the Unit 1 torus and drywell inerted? (0.5)
~
- b. WHY is there a minimum AND maximum suppression pool level limit? (1.0)
(***** CATEGORY 02CONTINUEDONNEXTPAGE*****)
, - - - - - - - . _ , - . , _ _ . . . - . - - - , - - - - . - - . . . . . - - - - . - - - - . - - - - ..-----,n.,---- - , , - . - . - .n-- .
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 10 QUESTION 2.06 (2.75)
With regard to the Standby Liquid Control System (SLC)
- a. During the filling of the Standby Liquid Control Storage Tank, some of the liquid is spilled and goes into the SLC system drains. WHERE would the spilled fluid discharge to AND WHY? (0.5)
- b. If SLC were manually injected, LIST five (5) indications that the operator could observe in the control room which would indicate that the SLC system was injecting. (Do not include system initiation control switch or manual shutoffvalvepositions.) (1.25)
- c. If the maximum amount of boron concentration necessary to shutdown the reactor from a full power condition is 660 --
ppm, WHY is the total design concentration 1075 ppm? (1.0) QUESTION 2.07 (3.50) Regarding the ADS System
- a. WHAT is the power supply for the ADS valves? (0.5)
- b. CAN you or CAN you NOT still actuate the ADS valves following a failure of the air supply to the ADS valves AND WHY? (1.0) .
- c. DESCRIBE the effect of the following signals clearing, (by themselves), prior to expiration to the ADS timer (i.e., will ADS initiate or not initiate)?
- 1. High drywell pressure (0.5)
- 2. Low water level (0.5)
- d. WHAT is the purpose of the vacuum breaker in each relief-valve discharge line? (1.0)
I j (***** CATEGORY 02CONTINUEDONNEXTPAGE*****)
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 11 QUESTION 2.08 (1.00)
Concerning the SRM Detection System
- a. WOULD an increase in the gas fill pressure (INCREASE, DECREASE, or NOT CHANGE) the sensitivity of the detector? (0.5)
- b. WOULD elimination of the discriminator (INCREASE, DECREASE, or NOT CHANGE) the control room SRM count rate? (0.5)
QUESTION 2.09 (1.50) With regard to the Off Gas System: _
- a. WHICH of the following gases is in the greatest abundance (most SCFM) at the exit of the steam jet air ejectors going to the Off Gas System. (0.5)
- 1. N2
- 2. H2
- 3. 02
- 4. normal air
- 5. water vapor
- 6. xenon -
- b. LIST two (2) conditions which will cause the off gas isolation valve (66-28) to close. (1.0)
QUESTION 2.10 (1.00) WHICH the designone (1) of the
/ operation following of the ECCS Kee? statements correctly Fill System PSC)?(describes (1.0)
- a. It can take its water from the pump suction of either Core Spray System I OR System II.
- b. The PSC pumps run continuously to maintain > 48 psig on system piping.
- c. CST static head automatically supplies a backup pressure source in the event of PSC failure,
- d. The pump suction valves close on a PCIS Group 2 isolation signal.
(***** CATEGORY 02CONTINUEDONNEXTPAGE*****)
+
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 12 1
I QUESTION 2.11 (.50) A DG automatic initiation signal is present and the operator depresses OG A's Control Room Emergency Stop PB. STATE the action which the operator must take from the Control Room to reset the automatic start lock-out. (0.5) QUESTION 2.12 (2.00) Listed below are four (4) parameters which can indicate a Failed Jet Pump. STATE whether these parameters will INCREASE, DECREASE, or REMAIN THE SAME for a Jet Pump Failure. NOTE: THIS IS NOT A RISER BREAK ___
- a. Core Flow, as calculated from Core Plate dP. (0.5)
- b. Failed Jet Pump Flow. (0.5)
- c. Companion Jet Pump Flow (Other Jet Pump on the common riser). (0.5)
- d. Companion Jet Pump Loop RECIRCULATION FLOW (Other JP Loop). (0.5)
QUESTION 2.13 (1.00) STATE the interlocks associated with the Normal and Alternate Feeder Breakers to a RPS BUS. Include in your response the purpose of each interlock. (1.0) i l (***** CATEGORY 02CONTINUEDONNEXTPAGE*****)
tr
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 13 QUESTION 2.14 (1.00)
The Recirculation MG Set Oil System is in its normal lineup for power operation when the running AC oil pump trips. The DC oil pump auto starts when the standby AC pump fails to restore oil pressure above 20#. Which of the following correctly reflects the current equipment status? (1.0)
- a. The AC oil pump (s) AND the MG set drive motor have tripped.
- b. The AC oil pump (s) continue to run; the MG set drive motor has tripped.
- c. The AC oil pump (s) have tripped; the MG set drive motor continues to run.
- d. The AC oil pump (s) and the MG set drive motor continue to run.
QUESTION 2.15 (1.00) The plant is operating normally at power when you receive a
" Pump A Seal Staging Flow High/ Low" alarm and note an INCREASE in No.2 Recirc Pump seal pressure with NO CHANGE in No. I seal pressure. Which of the following failures would cause these indications? -
(1.0)
- a. Failure of No. I seal
- b. Failure of No. 2 seal
- c. Plugging of the No.1 internal restricting / breakdown orifice
- d. Plugging of the No. 2 internal restricting / breakdown ;
orifice NOTE: NO OTHER ALARMS ARE PRESENT l l l . (***** END OF CATEGORY 02 *****)
l
- 3. INSTRUMENTS AND CONTROLS PAGE 14 i
l QUESTION 3.01 (3.00) Given below are the four (4) ranges of level indication. For each range GIVE: to WHAT location their value is referenced, WHETHER they are temperature compensated or not, and WHETHER they are used for RPS trip functions or not. (3.0)
- a. Normal Control Range
- b. Emergency Range
- c. Shutdown Vessel Flooding Range
- d. Post Accident Flooding Range a h w - O $ var. ble bg net Mt6IL wilui.~h QUESTION 3.02 (2.50)
In regards to the Browns Ferry Main Turbine, --
- a. STATE WHAT occurs at each of the following exhaust hood temperatures. (1.5)
- 1. 135 deg. F
- 2. 175 deg. F
- 3. 225 deg. F
- b. GIVE the position for each of the following valves when the turbine is in chest warming. (1.0)
- 1. main stop valves
- 2. control valves
- 3. CIVS stop
- 4. intercept QUESTION 3.03 (1.00)
Assume that Browns Ferry is operating in 3 element control at 32% power.
- a. WILL either the RWM or RSCS be controlling the rod withdrawal sequence? (Ifso,LISTthesystems.) (0.5)
- b. WILL the auto light on the RWM operator's panel be on or off? (0.5) l
- i
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
U
- 3. INSTRUMENTS AND CONTROLS PAGE 15 QUESTION 3.04 (4,00)
The following concern the Source Range Monitors.
~
- a. STATE the difference between the retract permissive indicating light on Panel 9-5 and Panel 9-12. (0.5)
- b. Except for irop, LIST all the conditions that will generate a rod block in the SRM circuit (include setpoints and when when the block is not bypassed). (3.0)
- c. DESCRIBE how the SRMs can be used to produce a scram. (0,5)
QUESTION 3.05 (2.75) -- The following concerns the Brown Ferry Reactor Water Cleanup System (RWCU).
- a. LIST four (4) of the RWCU isolations (INCLUDE setpoints). (2.0)
- b. WHICH RWCU valves auto close on a RWCU isolation?
(Actual valve numbers not required but may be used in lieu of descriptive description.) (0.75) QUESTION 3.06 (2.25)
- a. WHAT are the five (5) start signals for the Emergency Diesel Generators (setpoints not required)? (1.25)
- b. In the event of an emergency start (accident conditions)
WHICH of the diesel generator protection auto trips are still functional? (1.0) i (***** CATEGORY 03CONTINUEDONNEXTPAGE*****) {
tr
- 3. INSTRUMENTS AND CONTROLS PAGE 16 QUESTION 3.07 (2.00)
Regarding the Reactor Building Closed Cooling Water System (RBCCW).
- a. STATE all of the conditions which will cause automatic closure of the nonessential loop isolation valve (MOV-48). (1.0)
- b. If the condition that caused the nonessential loop isolation valve (M0V-48) to auts close clears, WILL MOV-48 auto open or is operator action required to reopen MOV-48? (0.5)
- c. MOV-47 controls the RBCCW supply to ESSENTIAL equipment.
WILL MOV-47 auto close? If so, LIST one (1) signal which will auto close MOV-47. (0.5) ___ (***** CATEGORY 03CONTINUEDONNEXTPAGE*****)
? .f a
- 3. INSTRUMENTS AND CONTROLS ( PAGE 17 QUESTION' 3.08 (1.00)
Unit 2 is operating at 100% RTP, with recirc in Master Manual._. An operator inadvertently INCREASES the " Pressure Set" on EHC ' by 5 psig. ASSUME: 1. No Further Operator Actions ! , f
- 2. All other EHC control settings are normal
- 3. Starting Parameters / /-
o TCV's - 100% Steam Flow Positten' o BPV's -
' 0% Steam Flow! Position o Power - 100% Rated Thermal Power o Pressure - 1005 psig ,/
NOTE: FIGURE # 257 IS ATTACHED FOR REFERENCE /
,n Which of the following most accurately describes both the .
INITIAL RESPONSE and FINAL STATUS of the different pa'rameters and components. (1.0) a b c d INITIAL RESPONSE o TCV's CLOSE (~83%) CLOSE(~83%) CLOSE(~83%) NO CHANGE o BPV's NO CHANGE OPEN (~17%) , NO CHANGE OPEN (~2ET.) o Power INCREASE NO CHANGE ' INCREASE DECREASE o Pressure INCREASE ' NO . CR.ANGE INCREASE . DECREASE FINAL STATUS 7,
~
o TCV's ~100 % ~ 83 % ~ 83 % ~ 100 % o BPV's 0% ~ 17 % ~ 17 % 0% o Power > 100 % 100~ % > 100 % ( 100 % o Pressure >1005psig 1005 psig >1005 psig (1005 psig i l l l l l (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
o
- 3. INSTRUMENTS AND CONTROLS PAGE 18 QUESTION 3.09 (1.00)
The reactor is being started up. Assume the following: o Reactor Power is below the LPSP o Rod Withdrawal Sequence B is in effect o No RWM errors or blocks exist o The RWM Normal-Bypass Switch was positioned to Bypass following complete withdrawal of all RWM Group 6 rods to Position 36 The operator incorrectly attempts to continue withdrawing Group 6 rods to position 48. WHICH one of the following most correctly details the outcome of this operator error? (1.0) NOTE: FIGURES # 467 A - 0 ARE PROVIDED FOR REFERENCE __
- a. Rod withdrawal will occur with NO Rod Position Restrictions.
- b. Rod withdrawal will occur, provided all Group 6 rods are maintained within 1 notch of each other.
- c. Rod withdrawal will NOT occur beyond the RWM Withdrawal Limit for Group 6.
- d. Rod withdrawal will NOT occur beyond Position 36, due to RSCS immediately imposing a Rod Block.
QUESTION 3.10 (1.00) Attached Figure # 457 illustrates the " Core Spray System Pipe Break Detection Instrumentation". STATE 7he pressure relationships between Points 6 and 7 while 0;erating normally at power? JUSTIFY your answer. (1.0) i I (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
u-
- 3. INSTRUMENTS AND CONTROLS PAGE 19 QUESTION 3.11 (2.00)
'When paralleling a diesel generator with the 4KV bus:
- a. WHAT do you use the governor control for
- 1. before the breaker is closed? (0.5)
- 2. after the breaker is closed?' (0.5)
- b. WHAT do you use the voltage regulator adjust for
- 1. before the breaker is closed? (0.5)
- 2. after the breaker is closed? (0.5)
QUESTION 3.12 (1.00) A TIP trace is being run in the automatic mode: the SCAN light is lit. A Group 2 PCIS isolation signal is recieved. DESCRIBE how the TIP system will respond. (1.0) QUESTION, 3.13 (1.50) Given the following data for APRM Channel C: LPRM Level: A B C D Number of LPRMs assigned: 6 5 5 5 Number of LPRMs bypassed: 3 4 0 0
- a. If APRM Channel C selector switch on the local (back) panel was placed to the COUNT position, WHAT would be the expected meter reading? (0.5)
- b. Based on the above data, is APRM Channel C operable:
ANSWER YES or N0'and EXPLAIN WHY. (1.0) (*=***ENDOFCATEGORY03*****)
O
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 20 RADIOLOGICAL CONTROL QUESTION 4.01 (2.00)
According to 0I-71 " Reactor Core Injection Cooling System"
- a. WHAT must be done if the 250 VDC is removed from FCV-71-39 (RCIC pump discharge valve) if RCIC is required to be operable? (1,0)
- b. If the turbine is tripped by the electrical overspeed trip, DOES the trip throttle valve have to be manually reset? (0.5)
- c. During testing of RCIC, the suppression pool temperature must not be allowed to exceed WHAT temperature? (0.5)
QUESTION 4.02 (1.50) According to 0I-92 " Neutron Monitoring System", during normal startup:
- a. You must verify that all operable SRM channels have a count rate greater than WHAT value? (0.5)
- b. All unbypassed IRM range switches are set to position
- 1. WHY? (0.5)
- c. WHAT values should all APRM channels be between before transferring to run? (0.5)
QUESTION 4.03 (.50) Concerning a main turbine startup OI-47, " Turbo-Generating System", you are cautioned NOT to operate the turbine for prolonged times between 900 to 1500 rpm. WHY? (0.5) (***** CATEGORY 04CONTINUEDONNEXTPAGE*****)
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 21 RADIOLOGICAL CONTROL QUESTION 4.04 (2.00)
According to Browns Ferry Radiation Protection Procedures
- a. WHAT are the administrative whole body dose limit for 1 year? (0.5)
- b. Based upon 10CFR20, WHAT is the maximum allowable whole body accumulated dose for a 30 year old person? (0.5)
- c. WHAT is the definition of radiation area? (1.0)
QUESTION 4.05 (2.00) STATE whether each of the following scrams need to be operable OR need NOT to be operable, in the COLD SHUTDOWN CONDITION. (2.0)
- a. main steam line high radiation scram
- b. IRM high flux scram
- c. scram discharge volume high level scram
- d. low reactor water level scram
- e. APRM 15% flux scram
- f. drywell high pressure scram
- g. manual scram
- h. SRM high flux scram
(***** CATEGORY 04CONTINUEDONNEXTPAGE*****)
, 4. PROCEDURES - NORMAL, ABNORMAL. EMERGENCY AND PAGE 22 RADIOLOGICAL CONTROL QUESTION 4.06 (1.00) You are executing E0I-1, " Reactor Control," in response to a reactor scram and subsequent low water level. Drywell cooling has been lost due to an electrical malfunction to the fans. After 10 minutes, you receive High Drywell Pressure at 2.5 psig and slowly increasing. Which of the following is the correct action to take according to the generic guidance provided in using the E0I's.
- a. Re-enter E0I-1 at the beginning (Only required action)
- b. Continue in E0I-1 from where you are - it will direct you to the correct action.
- c. Exit E01-1 and Enter E0I-2, " Primary Containment Control"
- d. Re-enter E01-1 at the beginning AND enter E0I-2.
Qb'ESTION 4.07 (4.00) According to Browns Ferry Unit 1 EPM-Control Room Abandonment Procedure
- a. WHAT are six (6) actions the operator should perform '
if sufficient time is available? (3.0)
- b. WHERE would the Unit i unit operator go? (0.5)
- c. WHERE would the Unit 3 unit operator go? (0.5) l QUESTION 4.08 (2.50)
LIST all of the entry conditions for E0I-2, " Primary 1 Containment Control." (2.5) l (***** CATEGORY 04CONTINUEDONNEXTPAGE*****)
e
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 23 RADIOLOGICAL CONTROL QUESTION 4.09 (1.00)
You are cautioned in E0I-2, " Containment Control" not to secure or place an ECCS in manual mode unless you are assured by at least two (2) independent indications, that one (1) of two (2) conditions exist. WHAT are these conditions? (1.0) QUESTION 4.10 (1.00) Attached Figure f 437 , "Drywell Spray Initiation Pressure Limit", is used in conjunction with the CAUTION provided in l E0I-2, " Containment Control". ___
"If' torus, pressure exceeds 14.5 psig BUT ONLY IF l dryweR temperature and pressure are within the '
Drywell Spray Initiation Limit, THEN ....
... initiate drywell sprays at rated flow."
STATE the basis for this Drywell Spray Initiation Pressure Limit. (1.0) QUESTION 4.11 (1.00) . GIVEN: The D/G CO-2 System has initiated in its Vital Area - Per BF-0I-39, " CO-2 Storage, Generator Purging, and Fire Protection": STATE what action (s) must be taken AND where it/they must be performed to enable the D/G C0-2 System to reinitiate (upon receipt of a valid signal). (1.0) NOTE: ASSUME THAT AN ADEQUATE CO-2 SUPPLY STILL EXISTS. QUESTION 4.12 (2.50) LIST ALL the entry conditions for Procedure E01-1,
" Reactor Control." (2.5) ,
(***** CATEGORY 04CONTINUEDONNEXTPAGE*****)
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 24 RADIOLOGICAL CONTROL QUESTION 4.13 (1.00)
You have been operating at 60% power when one (1) recirculation pump trips. You have been requested to restart the idle loop. According to 0I-68, Recirculation Procedure, WHAT are the two (2) Technical Specifications temperature limits that apply to the restart of an idle loop? (1.0) QUESTION 4.14 (1.00) Per the E0I's, DEFINE " Adequate Core Cooling." (1.0) QUESTION 4.15 (1.00) E0I I-1, RC/L states: If any control rod is NOT inserted to or beyond position
- disable ADS auto blowdown function (reference Appendix 3)(02),
and enter CS (LEVEL / POWER CONTROL). STATE the basis for disabling the ADS auto blowdown function in this plant condition. (1.0) QUESTION 4.16 (1.00) Concerning the generic useage of the E0I's:
- a. DIFFERENTIATE between the terms "WHEN" and "BEFORE" (0.5)
- b. EXPLAIN the significance of using all CAPITAL LETTERS to specify operator actions.
(e.g., " EMERGENCY RX DEPRESSURIZATION IS REQUIRED") (0.5)
*****)
(***** END OF CATEGORY 04 (************* END OF EXAMINATION ***************)
EQUATION SHEET Where mi = m2 (density)i(velocity)i(area)g = (density)2(velocity)2(area)2 KE="{2 PE = mgh PE + KE +P V i i 1 1 = PE 2+KE 2+P Y22 where V = specific volume P = Pressure Q = mcp (Tout -Tin) Q=UA(T,y,-Tstm) Q = m(ht-h2 ) P = Po10(SUR)(t) P = Po et /T SUR = 26.06 T = (B-p)t T p delta K = (K,ff-1) CRi (1-Keff1) = CR 2 (1-Keff2) CR = S/(1-K,ff) (1-K,ffi) (1-K,ff) x 100% SDM = M = (1-Keff2) K,ff decay constant = in (2) = 0.693 Ag = A e-(decay constant)x(t) t t 1/2 1/2 Water Parameters Miscellaneous Conversions 1 gallon = 8.345 lbs 1 Curie = 3,7 x 10 10 dps 1 gallon = 3.78 liters 1 kg = 2.21 lbs 1 ft = 7.48 gall'ons 1 hp = 2.54 x 103 Btu /hr 3 6 Density =62.4lbg/ft 1 MW = 3.41 x 10 Btu /hr Density = 1 gm/cm 1 Btu = 778 ft-lbf Heat of Vaporization = 970 Btu /lbm Degrees F = (1.8 x Degrees C) + 32 Heat of Fusion = 144 Btu /lbm 1 inch = 2.54 centimeters 2 1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 ft-lbm/lbf-sec a
- . _ - - . - - - - . - - . . . . _ . - - . - - - - , - --m-, -,
20A EVENT INA*iVELTENT HP31 STAR'T POWER 10M 5 :- 1 4 --- . 3
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2 ' , 0 20 40 60 80 100 120 0 a, 8 12 16 CORE FLOW TOT AL STE AM FLOW (X 1 MILLION lbs/hr) (X 1 MILLION lbs/hr) i 8 :- h 11 i t 7
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6 "'~ ' '..- 9 950 970 990 1010 1030 1050 0 4 8 12 16 RE ACTOR PRESSURE TURBINE STE AM FLOW (PSIG) ( X 1 MILLION lbs/hr) e- , - ---- -_,. - , ,e ,--mn------~---m-- - - - - - e--- e-a- e-wn -------,----~wn--ww----v e w --w-----~mw~'v----w------~-----w---*
20B EVENT INADVERTENT HPCI START POWER 100%
=
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( 0 25 50 75 100 125 0 4 8 12 16 APRM TOT AL FW FLOW (% POWER) ( X 1 MILLION lbs/hr) j
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- Pag 2 _32 of 37 08/03/84
~
Lesson Plan 25 59 4 3 4 3 q s 2 1 2 1 2 7
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51 4 3 4 3 4 p3 N
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03 3 4 3 4 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58 FIGURE 2 SEQUENCE B, RWM R0D GROUPS 1-4; RSCS R0D GROUPS B12 AND B34 -
Pcgu 36 of 37 08/03/84 Liston Pirn 25 [ ] - E 7 6 6 7 7 N / 47 p 8 11 10 10 11 8
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03 1 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58 1 FIGURE 6 NOTCH GROUPS 5-13, SEQUENCE,B
._ g P'ge a 37 of 37 08/03 F Lesson Plan ,25 -
l 59 14 15 14 q
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51 17 18 19 18 17 r N / , 43 16 20 21 22 21 20 16 35 N / _ 15 19 -22 21 22 19 15
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, 19 22 21 22 19 15 19 16 20 21 22 21 20 16 L
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J 17 18 19 18 17 07 \ 03 L ,4 15 ,4 J 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58 FIGURE 7 NOTCII GROUPS 14-22, SEQUENCE B
-- - . , . . - - , - - - . . , _ _ _ , _ , , . . . , _ _ _ . _ . _ _ _ . _ _ - - _ _ . , _ _ _ _ - , - - - - - - . . . - _ ~ _ - , - - , - - - - - , . . . - - , . . - - - . - , , . _ . -_,
Pag 2 21 of 22 08/14/85 Lesson Plan 45 Rev. O C *
,o J L STEAM DRYER DRYWELL WATER LEVEL
- . . f.- - - --
~ ~ aa - - ) 7" AO MO MO f %
- r. % V, M ><H N
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o____ ___ __ W kir I(r Is0TE: DETA P esBTIquhsses? IS COMsECTEO 70 efwe A P0efffv6 ftEAesse Af flATED Cosemesse se OfeGR g 70 Ofv6 A FAE, SAFS sesTimassur. ABOVE CORE
;; PLATE poAH
;; PRESSURE 80
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- STANOBY LIQUID CONTROL i
I FIGURE 4 CORE SPRAY SYSTEM PIPE BREAK DETECTION INSTRUMENTATION
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figure C. DRYWELL SPRAY INITIATION LIMIT -- E 280 ' ' ' t i i , , 7 O- y , , l ! o m 205- l z R 1ao- II l-nos
-180 94 .,
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- i hhj, 130-I
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M',,, l 8 , CV REG l
.i eu, g a SPEED OPEN WifEN GEN. OUTPUT g y _7 ERROR LOAD REJECT ' '
IV R[Eg 0 - r--I W BREAKER IS CLOSED
> - , a,,,
i r- , , M s ,igen i START UP '* s > CV REGULATION'N/ *r
+ .L RATE ,
,f yto,y
! o g 4 Z,4 J's, q b ,6 > LOAD SELECTOR g 4 C C - n -
- RUNBACK ON LOAO REJECT n STEAM ~ t----eb- RUN8ACK SYNCSPEED ,
NOT SELECTED 0 4 4
. THROTTLE I e PRESSURE
- o -
o - t-* A y ON e- s:AS PRESS tun 8tNE TRfPPED w >> MOT )y, t-8 contact m PftESSUftE SET , r VA(VE 4i % FLW DEM4880 t > Z +1 Fei PRESS. LUAD j B 88AS Z F ALVES Foo.R STEAM + REG. - THROTTLE PRESSURE I. ~ STOP V TCtOSEo , s' sy, Ass
~
_A
~ ^ VALVE DEMAND gg , o .,
SMALL gypg g AO gyE ancg e go)$ On* t.d 8, "
~
s~ O l \ .
- -y v p. . ~
.r -.L: l Q .:,~. T ~ .1 - .
T *
=-
\ . - . . .-
a 7
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 25 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W.
ANSWER 1.01 (2.00)
- a. { Xe(eq.) PProduction: Removal. The Production term is a flux dependent term; the Removal term is dependent on burnup and decay with only burnup being flux dependent. }
While all the production term would reduce by one-half, (+0.5) only the burnup portion of the removal term would reduce by one-half; thereby higher,e utlibrium leaving level 0.5).(+the xenon concentration at aSA cb ,qwM A& p .LQ n % A, . o/t- yM u ffta d w ,e u m, p (+1.0 Total) # . M e c.ox) _
- b. No (+0.25), because xenon peak following a scram is approximately the square root of the power change (10 h,rsh and xenon peak on a 100% to 50% power maneuver is approximately 5 hrs (+0.95).0c.t:)
(+1.0TotAi) DdM f' ' " #' REFERENCE
- 1. BFNP, Xenon and Samarium, pp. 2, 3, 4, Student Learn-ing Objectives 1, 3, 4.
ANSWER 1.02 (1.00) Supercritical (+0.5) (When the period reaches infinity, the reactor is exactly critical). After the rod insertion stops, the delayed neutron precursors that were formed in previous generations and at a higher power level tend to pull power back up (+0.5). (Therefore, the reactor is still supercritical due to thelatenteffectofdelayedneutrons). (+1.0 Total) REFERENCE
- 1. BFNP, Neutron Slowing Down and Diffusion, pp.1, 8, 9 of 24, Student Learning Objectives 1, 6. '
- 2. BFNP, Reactor Power and Reactor Period, pp. 4, 5, 6 of 11, Student Learning Objectives 3, 4.
tTI l i
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 26 I THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW l ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W.
ANSWER 1.03 (2.50)
- a. 3 (+1.0)
- b. delta (ro) = alpha (x)
- delta (X) delta (ro)dop = more negative (+0.5) delta (ro) void = less negative (+0.5) delta (ro) mod = less negative (+0.5)
(+1.5 Total) REFERENCE
- 1. BFNP, Reactivity Coefficients, pp. 2, 3, 4, 5 of 17, Student Learning Objectives 1, 2, 3.
ANSWER 1.04 - (2.00)
- a. Shallow rods are those rods that are notch positions 32 - 48 (+0.5). ( They are also called shaping rods. )
They affect the local axial flux individually l (whiletakentogether,theyaffecttheradialppw(er tistribution depending on pattern and symmetry) +0.5). (+1.0 Total)
- b. Deep rods are those rods that are notch positions 00 -
16 (+0.5). (Theyarealsocalledpowerrods.) Deep rods affect the gross power production in the core (+0.5). (+1.0 Total) REFERENCE
- 1. BFNP, Control Rod Worth, pp. 7, 8 of 25, Student Learning Objectives 1, 4.
e
.. _ . _ _ - . _ , . . ~ . _ - .__ _ . _ _ . _ _ _ _ _ _ _ _ . . _ _ _ _ ._.__,__ . _ . _ , _ . . . _ _ _ _ _ _ . _ _ . _ .
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 27 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W.
ANSNER 1.05 (2.00)
- a. Minimum permissible stable period = 60 seconds (+0.5)
DT = period /1.445 or 1.443 (+0.25)
= 60/1.445 = 41.6 seconds (+0.25)
(+1.0 Total) t/T
- b. P = P e (+0.25) P/Po = 3000/5 = 600 (+0.25) in 600 = t/T t = 60
- In 600 = 384 seconds = 6.4 min. (+0.5)
(+1.0 Total) REFERENCE
- 1. BFNP, Reactor Power and Reactor Period, pp.1, 2, 4, Student Learning Objectives 1, 2.
- 2. BFNP, GOI 100-1, p. 13
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 28 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W.
ANSWER 1.06 (2.00)
- a. decrease (+0.5)
- b. increase (+0.5)
- c. increase (+0.5)
- d. decrease (+0.5)
REFERENCE
-~
- 1. BFNP, Rankine Cycle pp. 7, 8 of 12, Student Learning Objective 2.
- 2. BFNP, Plant Efficiency, pp. 9,10 of 11, Student Learning Objective 3.
- 3. BFNP, Reactor Heat Balance, pp. 5, 6, 7 of 9, Student Learning Objective 1.
ANSWER 1.07 (1.50)
- a. Increase (+0.5)
- b. increase (+0.5)
- c. decrease (+0.5)
REFERENCE
- 1. General Electric Thermohydraulic, Heat Transfer, and Fluid Flow Text, pp. 9-85, 9-86, 9-92.
0 . - . . . - , . - - _ _ _ _ _ , ___ _ _ _ _ . - _ _ _ __ _ _ - - . . - . _ _ _ - . . . _ _ _ _ _ , . , .____..-~..
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 29 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W.
ANSWER 1.08 (2.50)
- a. 295 deg. F (+- 15 deg F) (+0.5)
- b. increase (+0.5)
- c. increase (+0.5) ,
(4, gze,. u . W !L SM Ufv/w,
- d. 450 psia (+- 50 psia) (+0.5) - 455 deg. F (+0.5)
REFERENCE
- 1. BFNP, Hollier Diagram, pp. 4, 5 of 7, Student Learning --
Objectives 3, 4.
- 2. BFNP, Steam Tables, pp. 6, 7 of 23, Student Learning Objectives 2, 4.
- 3. BFNP, Lesson Plan 43, p. 11 of 17
- 4. General Electric Heat Transfer and Fluid Flow, Chapter 3.
ANSWER 1.09 (2.00)
- a. The difference between static pressure at the eye of the pump and saturation pressure. (+0.5)
- b. 1. decreases (+0.5)
- 2. M easee (+0.5) ,e ,, :, e m.
- 3. ,
a2:_t W same (+0.5) dec.resu-REFERENCE
- 1. General Electric Thermohydraulic, Heat Transfer, and Fluid Flow Text, pp. 7 7-96.
a
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 30 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W.
ANSWER 1.10 (2.00)
- a. Due-to-the-recire pn=p trin at 1120-pefg. UN Yd C'# k' F# f~*^'" '
b.-Due to-the-anticipatory-MSIVJosure serem. / > tN d V 5"W '"g
- c. Due-to'MSRV~opintn r eausi Ny-a levaL-epth, n u r d u e . A t:. W ll C '"'
- d. Due tc, tb Emng bachic!:: 5wearfiew. //r? I gj <d.. c / m.:Je-REFERENCE
- 1. BFNP, Transients, Inadvertent HPCI Start, p. 1, Student Learnitg Objective A. _
ANSWER 1.11 (1.50)
- a. decrease (+0.5)
- b. increase (+0.5)
- c. increase (+0.5)
REFERENCE
- 1. General Pump Dynamics.
ANSWER 1.12 (1.00)
- a. (+1.0 TOTAL)
REFERENCE
- 1. BFNP, BFN Hitigating Rx Core Damage, pp 17 - 18; RQ 85/02/01 ANSWER 1.13 (1.00)
- a. (+1.0)
REFERENCE
- 1. BFNP LHGR and BASES LP, pp. 8,9
- 2. GGNS HCD, THERHAL LIMITS, p. 74
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 31 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W.
ANSWER 1.14 (1.00)
- a. (+1.0)
REFERENCE
- 1. 8FNP: XENON & SAMARIUM LP, pp. 4, 12
- 2. GGNS: OP-NP-514 ANSWER 1.15 (1.00)
By observing the neutron level while moving the nuclear _. Instrumentation. A significantly HIGHER (approximately 300 times) count rate would be seen for the UNVOIDED areas of the core as opposed to the VOIDED. (1.0) REFERENCE General Electric Heat Transfer and Fluid Flow, Chapter 9 General Electric Reactor Theory, Chapter 4 EIH: L-RQ-540 (M.8) (MCD) BFNP HTFF, Chapter 9; Rx Theory, Chapter 4, L.O.'s 3.3 & 3.6 l
v
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 32 ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W.
l ANSWER 2.01 (3.00)
- a. RCIC, HPCI, RWCU, CRD ,
(+0.25 for each correct answer)
- b. P4 SJAEcondensers) s steam packing exhauster condenser) gM J 1
off gas condenser) 7 condensate booster pumps) 8 low pressure feedwater heater strings) g @ startup bypass valve) op l.3 reactorfeedpumps) __ 5 high pressure feedwater heater strings) (+0.25 for each answer) REFERENCE
- 1. BFNP, Lesson Plan 11, Objective F, pp. 1, 3, 4.
ANSWER 2.02 (2.00)
- a. HCPI turbine stop valve (73-18) (+0.5) and minimum flow '
bypass valve to suppression pool (73-30) (+0.5) (+1.0 Total)
- b. Auto restart on the low low level (+0.5)
- c. No (+0.5)
REFERENCE
- 1. BFNP, Lesson Plan 42, pp. 12, 26.
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 33 ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W.
ANSWER 2.03 (1.00)
- a. Suction headers for Unit 1 and Unit 2 are interconnected.
(Unit 3isseparate.) (+0.5)
- b. Used to inject sodium hypochlorite into the raw cooling water system to control marine growth (+0.5)
REFERENCE
- 1. BFNP, Lesson Plan 48, Objective B, pp. 3, 5.
ANSWER 2.04 (2.25)
- a. 1. Open mode (+0.25) water flows from intake structure to main condenser and discharged back to the reservoir (+0.5)
- 2. Helper mode (+0.25) water flows from the intake structure to the main condenser to the cooling towers and is discharged back to the reservoir (+0.5)
- 3. Closed mode (+0.25) water flows from the intake structure to the main condenser to the coolin .
and then back to the intake structure (+0.5) g towers REFERENCE
- 1. BFNP, Lesson Plan 50, Objective A, pp. 1, 3, 4.
- 2. BFNP, Lesson Plan 50, Objective C, pp. 1, 9.
l l
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 34 ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W.
ANSWER 2.05 (1.50)
- a. Reduce potential explosive concentrations of oxygen /
hydrogen (+0.5)
- b. Provide sufficient water for steam condensation (+0.5) and enough air space for the primary containment n on-condensibles (+0.5)
REFERENCE
- 1. BFNP, lesson Plan 16, Objectives C, D, H, I, pp. 1, 14, 26, 32. _.
ANSWER 2.06 (2.75) a, collected in 55 gallon drums (+0.25) because the sodium pentaborate can precipitate out and plug the lines (to prevent plugging up rad waste lines) (+0.25)
- b. SLC storage tank level decreasing reactor power decreasing pump discharge pressure squib continuity monitors (indicating lights)
(indicatinglights pump systemrunning flow (indicating lights)) squib current flow (back of 9-5 panel) (+0.25 for each correct answer, +1.25 maximum)
- c. dth to allow tt e ( 4for imperfect
- 5) (+1.0 Total mixinf (N) and RHR REFERENCE
- 1. BFNP, Lesson Plan 39, Objective B, pp. 1, 5, 7, 14.
h
'2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 35 ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W.
ANSWER 2.07 (3.50)
- a. 250 VDC (+0.5)
- b. Can (+0.5), accumulators still functional (+0.5)
(+1.0 Total) c.1. initiate (+0.5)
- 2. not initiate (+0.5)
(+1.0 Total) _.
- d. to prevent drawing water up into the line due to steam
, condensation following termination of relief valve operation (+1.0) REFERENCE
- 1. BFNP, Lesson Plan 43, pp. 4, 7, 8.
ANSWER 2.08 (1.00)
- a. increase (+0.5)
- b. increase (+0.5) l REFERENCE
- 1. BFNP, Lesson Plan 19, p. 19.
i O l l }
. - - . . - - . . . ~ . - . - - - . - . - . _ _ . . . - , . . - . _ - . - . . -
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 36 ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W.
ANSWER 2.09 (1.50)
- a. H2 (+0.5) [2)
- b. 1. manual on control room panel 9-53
- 2. loss of air
- 3. Hi-Hi-Hi radiation in the off gas post treatment system 2 (G % combindion of - ) N/ /,,.of !e c)
(d.fys(c,c. +0,5'each, +1.0 maximum) tl. ja h . p v ie ( ss%c& FCV- W
- f --
ANSWER 2.10 (1.00) d (+1.0 TOTAL) REFERENCE
- 1. BFNP Lesson Plan No. 45, p.7-8 ANSWER 2.11 (.50)
Take the' Control Switch to RESET REFERENCE
- 1. BFNP, 0I-82 i
ANSWER 2.12 (2.00) j a. Decrease
- b. Increase
- c. Decrease l d. Increase (+0.5each)
! REFERENCE
- 1. BFNP: OI-68
~n
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 37 ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W.
I ANSWER 2.13 (1.00)
- 1. Both Alternate breakers can't be closed at the same time
(+0.25) tc prevent powering both RPS buses from the same power supply. (+0.25)
- 2. The Normal and Alternate breakers for the same RPS bus can't be closed simultaneously (+0.25) to prevent paralleling an RPS MG Set with the Alternate power source. (+0.25)
REFERENCE
- 1. BFNP, RPS L/P (# 28), L.0. C.
ANSWER 2.14 (1.00)
- a. (+1.0 TOTAL)
REFERENCE
- 1. BFNP Lesson Plan No. 7, p.16 ANSWER 2.15 (1.00)
G ed d. (+1.0 TOTAL) - REFERENCE
- 1. BFNP, Lesson Flan No. 7, p. 28
- 2. EIH: L-RQ-714, Figure 714-6; HNP-2-2447
- 3. GGNS SD B33-1, pp 5, 6; OP-833-1-501, p 5; ARI B33-FAL-L603A e
- 3. INSTRUMENTS AND CONTROLS PAGE 38 ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W.
ANSWER 3.01 (3.00) Normal Control Range Referenced to instrument zero Temperature compensated (by a pressure signal) ihut used for trip functions Emergency Range Referenced to instrument zero Temperature compensated (via heat clamps between reference and variable leg) jv37 Used for trip functions (because a Yarway does not require electrical power for indication) , Shutdown Vessel Flooding Range Referenced to instrument zero No temperature compensation Not used for trips Post Accident Flooding Range Referenced to 360 inches above vessel zero (which is top of active fuel) No temperature compensation Not used for trip functions (+0.25 for each correct answer) REFERENCE
- 1. BFNP, Lesson Plan 3, pp. 5-8.
h i
- 3. INSTRUMENTS AND CONTROLS PAGE 39 ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W.
ANSWER 3.02 (2.50)
- a. 1. Automatic temperature control valve (ATCV) begins to open. (+0.5)
- 2. Alarms in control room (+0.25) and ATCV is full open (+0.25).
- 3. turbine trip (+0.5)
(+1.5 Total) , t'0 0 m.,5
- b. 1. No. 2 bypass open: 1, 3, 4 closed (+G.25) __
- 2. closed (+0.25)
- 3. open (+0.25)
- 4. closed (+0.25)
(+1.0 Total) REFERENCE
- 1. BFNP, Lesson Plan 10, Objective F, pp. 1, 18.
- 2. BFNP, lesson Plan 10, p. 42, Objective D, p. 1.
ANSWER 3.03 (1.00)
- a. No (30% of rated is maximum RWM/RSCS power) (+0.5)
- b. Off (must be above 35%) (+0.5)
REFERENCE
- 1. BFNP, Lesson Plan 24, pp. 10, 28.
- 2. BFNP, Lesson Plan 12, p. 37.
- 3. BFNP, Lesson Plan 25, p. 11. '
- 3. INSTRUMENTS AND CONTROLS PAGE 40 ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W.
/
ANSWER 3.04 (4.00)
- a. 9-5 Panel = light on greater than or equal to 100 cps 9-12 Panel = light off greater than or equal to 100 cps
(+0.5)
- b. SRM high (+0.25), greater than 1 x 10E+5 (+0.25), rod block below range 8 on IRM (+0.25) and not in run (+0.25)
SRM downscale (+0.25), less than 3 cps (+0.25), rod block below range 3 on IRM (+0.25) and not in run (+0.25). SRM retract not permitted less than 100 cps (+0.25), rod block if detector not full in (+0.25) and below range -- 3 on IRM (+0.25) and not in run (+0.25). (+3.0 Total)
- c. By removing the shorting links (+0.35), in the RPS (+0.15).
(+0.5 Total) REFERENCE
- 1. BFNP, Lesson Plan 19, Objective G, pp. 1, 18.
- 2. BFNP, Lesson Plan 19, Objective E, p. 1.
- 3. BFNP, Lesson Plan 19, Objective I, p. 1.
O,
- 3. INSTRUMENTS AND CONTROLS PAGE 41 l
ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W. ' ANSWER 3.05 (2.75)
- a. 1. Low reactor water level (+0.25), +11 inches (+0.25)
- 2. High tem deg. F (perature +0.25) outlet of non-reg Hx (+0.25),140
- 3. Standby liquid control initiated (+0.5)
- 4. High RWCU e utpment area temperature (+0.25),160 to 180 deg. F +0.25)
- 5. High temperature in floor drains in RWCU area (+0.5), __
160 to 180 deg. F (+0.25) (Any four at +0.5 each, maximum +2.0)
- b. inboard and outboard inlet isolation valves (FCV 69-1 and FVC 69-2) and the return isolation valve (FCV 69-12)
(+0.75) REFERENCE
- 1. BFNP, Lesson Plan 13, Objective E, pp. 1, 10.
l
- 3. INSTRUMENTS AND CONTROLS PAGE 42 ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W.
ANSWER 3.06 (2.25)
- a. 1. High drywell pressure, (2.45 psig)
- 2. Low low low level, (-114.5 inches)
- 3. Loss of voltage to 4160 shutdown board
- 4. Degraded voltage to shutdown board
- 5. Manual
(+0.25 for each correct answer) (+1.25 Total)
- b. High differential current (+0.5) __
Overspeed (1035 rpm = 115%) (+0.5) (Note: Overcurrent, loss of field and reverse power will not trip.) (+1.0 Total) (+1.0 Total) REFERENCE -
- 1. BFNP, Lesson Plan 38, pp. 10, 11.
ANSWER 3.07 (2.00) a, loss of normal AC power in conjunction with an accident (Jr.b:' e# do @ ^" or gud in.a-ec- We AccodW ea,rd.hk o sci pja y,,en,,'5 g w;gw g,,,,;} low RBCCW pump discharge header pressure (both answers at +0.5 each)
- b. operator action is required (+0.5)
- c. There are no auto close signals for MOV-47. (+0.5)
REFERENCE
- 1. BFNP: Lesson Plan 47, pg. 8. L.0. D 1
. - . - . , . -. - . . _ . . . , - - . - - . _ , - - - - _ , - - ._--,----_---c.- , . - , -
- 3. INSTRUMENTS AND CONTROLS PAGE 43 ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W.
ANSWER 3.08 (1.00)
- a. (+1.0 TOTAL)
REFERENCE
- 1. BSEP, RTN-033, 012; SD 26.2
- 2. EIH: L-RQ-705, pp 18, 19; GPNT, Vol. VII, Chapter 9.4 ANSWER 3.09 (1.00)
- b. (+1.0 TOTAL) _ . , _
REFERENCE
- 1. BFNP, RSCS LP; RWM LP; RQ 85/02/02 & 03 ANSWER 3.10 (1.00) p-7 = p-3 + the height of cooler water in the sensing line.
Therefore p-7 ) p-6 even though p-3 ( p-2. REFERENCE
- 1. BFNP, LPf45, p 13, 14; RQ 85/01/05 ANSWER 3.11 (2.00)
- a. 1. diesel speed +0.5
- 2. load control +0.5
- b. 1. voltage control (+0.5)
- 2. VAR control (+0.5) l REFERENCE
- 1. BFNP: DG SYSTEM DESCRIPTION, OPL171.038 l
- 2. LaSalle: System Description, Chapter 47. j l
I e
m
- 3. INSTRUMENTS AND CONTROLS PAGE 44 ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W.
ANSWER 3.12 (1.00)
- 1. The TIP will auto trnsfer to the manual reverse mode. (+0.5)
- 2. The ball valve will close (+0.25), when the in-shield limit switch is closed. (+0.25)
REFERENCE
- 1. BFNP: LPf23, pp. 21, 22 ANSWER 3.13 (1.50)
- a. 70% (+0.5) _
96 g 4e G
- b. Je (+0.5), there are Lass than two operable inputs on Level b A pdm t. . (+0.5)
REFERENCE
- 1. BFNP: APRM LPf OPL171.022, p. 37 l
l l 0
i
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 45 RADIOLOGICAL CONTROL ANSWERS -- BROWNS FERRY 1, 2&3 -
86/06/16-CLIFF, W.
/. ) J ANSWER 4.01 (3.06)
- a. the valve must be opened (+0.5) and its breaker placed in the inoperable position (+0.5)
(+1.0 Total) L. No-(only orrmechanicaf overspfFtrip) (;0.5). ddck
- c. 105 deg. F (+0.5)
REFERENCE
- 1. BFNP OI-71, p. 3.
ANSWER 4.02 (1.50)
- a. 3 cps (+0.5)
- b. because rod withdrawal blocks due to IRM downscale trip are bypassed on range 1 (+0.5)
- c. greater than or equal to 5% and less than or equal to .
12% (+0.5) REFERENCE
- 1. BFNP 01-92, pp. 4, 5.
ANSWER 4.03 (.50) This is the critical speed range of the turbine (large vibrations could occur damaging the turbine). (+0.5) REFERENCE
- 1. BFNP 0I-47, pp. 20, 23.
e v.v r --- - - - - - - - - , , , , , _ _ , y, , _m-- -_-. - - , . _ _ , - - _ _ , - _ , , - . - - - - - , . _ , , _ . . . - - _ -
l
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 46 RADIOLOGICAL CONTROL
)
ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W. l l ANSWER 4.04 (2.00) 1
- a. 4 rem / year (+0.5) l
- b. 5(N-18) = 5(30-18) = 60 rem (+0.5)
- c. Any area accessible to personnel in which there exists :
or is likely to exist radiation such that a major portion of the body could receive an exposure rate in excess of 5 mrem /hr or in any 5 consecutive days a dose in excess of 100 millirems) (the) b + ad 4 a<ae6 100 me/g- (40) -~ REFERENCE
- 1. BFNP: RCI-9, p. 16, 10CFR20.101 and ICCFR20.202.
ANSWER 4.05 (2.00) a,d,/,f,h are not required b, c, g,e are required (+0.25each,2.0 maximum) REFERENCE
- 1. BFNP Tech. Specs. pp. 33-36a, Table 3.1.A ANSWER 4.06 (1.00)
- d. (+1.0)
REFERENCE
- 1. BFNP, E0I-1, Sections 2.1 and 3.3; E0I-2, Section 2.1; E0I L/P, L.0. I
- 2. BWROG E0I Users Guide, Section 2.1. '
-. , . - , , - - - -- e - - , , , - - - ,-
m
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 47 RADIOLOGICAL CONTROL ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W.
1 ANSWER 4.07 (4.00) l l
- a. 1. manually scram the reactor
- 2. check that all rods are fully in
- 3. continue to operate with normal feedwater system and bypass valve operation as long as possible
- 4. close main steam isolation valves
- 5. trip feedwater pumps 1
- 6. trip main turbine .
- 7. trip condensate pump
- 8. trip condensate booster pump
- 9. start diesel _
(+0.5 for each correct answer up to a maximum of six (6) correst answers, +3.0 Total)
- b. Unit 1 panel 25-32 or shutdown board room A. (+0.5)
- c. Unit 3 panel 25-32 or MOV board A. (+0.5)
REFERENCE
- 1. Browns Ferry Control Room Abandonment Procedure, p. 2 and Figure 2.
ANSWER 4.08 (2.50)
- 1. Suppression pool above 95 deg. F (+0.5)
- 3. Suppression pool level above - 1" (+0.5)
- 4. Supperssion pool level below - 6.25" (+0.5)
/LO
- 4. drywell temperature above '165r deg. F (+0.5)
- 5. drywell pressure above 2.45 psig (+0.5)
REFERENCE
- 1. BFNP E0I-2, pp. 3, 4. ,
- - - - -+ . _-.-e. - - - - - - , - - . . - - . , _ . _ . . -
. , , . . . , , - - - . - - - - - , - , - , - - , - - - - - - + ,-
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 48 RADIOLOGICAL CONTROL ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W.
o.0 ANSWER 4.09 (M) 1 nntroperation-itt -automat f e-mode-irconf f rmed, (+0.5) eradequatTcorEiioTing is~aisored-f+0di) SOSC C ANSWER 4.10 (1.00) ' l "I f* Spray initiation above this limit may result in containment depressurization rate which exceeds the relief ca city of the drywell and reactor building vacuum breakers.(0.3 This could - - - result in the negative design pressure of the drywell being exceeded. (0.7) REFERENCE
- 1. BFNP, E0I-2; PC/P-3, p. 2 AN5WER 4.11 (1.00)
Reset the System (0.5) at the local relay cabinet (0.5) REFERENCE
- 1. BFNP, BF-0I-39, Sections III.E & IV ANSWER 4.12 (2.50) i 1. Reactor water level below +11" (+0.5)
- 2. Drywell pressure > 2.45 psig (+0.5) l 3. A condition which requires a MSIV isolation (+0.5)
- 4. Reactor pressure above 1043 psig (+0.5)
- 5. A condition which requires a scram and Reactor power i
1 is > 3% or connot be determined (+0.5) REFERENCE
- 1. BFNP: E0I-1, pp.1
-._,__m...
v
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 49 RADIOLOGICAL CONTROL ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W.
ANSWER 4.13 (1.00)
- 1. Steam dome space to bottom drain less than or equal to 145 deg. F (+0.5)
- 2. Idle loop to operating loop less than or equal to 50 deg.
F (+0.5) REFERENCE
- 1. BFNP: Technical Specifications 3/4.3.6A, p. 175
- 2. BFNP: OI-68 ANSWER 4.14 (1.00)
Heat removal from the reactor sufficient to restore and maintain the peak fuel clad temperature (PCT) LTE 2200 deg F. (Heat up is allowed as long as PCT will nowhere exceed 2200 deg F.) (+1.0) REFERENCE
- 1. BFNP, E0I L/P, p. 5, L.O. D.
- 2. BWROG E0I Users Guide.
ANSWER 4.15 (1.00) , Preclude excessive heat up of the Suppression Po~ol to p,reclude l'""*^* '^ " # the loss of ECCS pump NPSH (and the subsequent inabiliAy to # cts (esd sx4d vi c . maintaig a,dequate core cool,ing, potential unstable steam ra.9 ,n j e ,;, 1 f e,c .o r condensarTon, and containment failure). (+1.0) f ew 2.c a-us., hud.. 3 4 x;, g#,1 c u rs cl-+ e. - REFERENCE
- 1. BFNP, E0I L/P, C-5.
- 2. BWROG Users Guide, E0I RC/L, p. 3.
t
v
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 50 RADIOLOGICAL CONTROL ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-CLIFF, W.
ANSWER 4.16 (1.00)
- a. BEFORE means that you are to take an action before a condition is met (there is no specific margin to allow before your action) (0.25);
WHEN means that you are to wait until the described conditions exist and then you can take the required action (0.25). (0.5)
- b. You are being directed to take that action (at some other location
-{withintheE0I's). (0.5)
REFERENCE BFNP: E0I-1, Section 4.0 e 6
* \
U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: BROWNS FERRY 1, 2&3 REACTOR TYPE: BWR-GE4 DATE ADMINISTERED: 86/06/16 EXAMINER: SLY, G. APPLICANT: //brslc'r ft. V
/ /
INSTRUCTIONS TO APPLICANT: Use separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after -- the examination starts.
% OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY
.sv. >r 2&r06 25.00 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 25.00 25.00 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSlRUMENTATION , ,
24.s: 25-06 25.00 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL l CONTROL 25.00 25.00 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 100.00 TOTALS l FINAL GRADE % All work done on this examination is my own. I have neither given nor received aid. ' APPLICANT'S SIGNATURE
.o
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
- 1. Cheating on the examination means an automatic dental of your application and could result in more severe penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil only to facilitate legible reproductions.
- 4. Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of 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 one sTde of the paper, and write "Last Page" on the last answer sheet.
~~9. Number each answer as to category and number, for example,1.4, 6.3.
- 10. Skip at least 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 parenthesis after the question and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in -
completing the examination. This must be done after the examination has been completed. em
- 18. When you complete your examination, you shall:
- a. Assemble your examination as follows:
(1) Exam questions on top. (2) Exam aids - figures, tables, etc. (3) Answer pages including figures which are a part of the answer.
- b. Turn in your copy of the examination and all pages used to answer the examination questions,
- c. Turn in all scrap paper and the balance of paper that you did not use for answering the questions.
- d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
4 e e h
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 2 THERMODYNAMICS QUESTION 5.01 (2.00)
Regarding the xenon transient following a significant decrease in reactor power from high power operation:
- a. If the decrease in reactor power was from 100% to 50%,
WHY is the equilibrium xenon reactivity more than one-half the 100% equilibrium value? (1.0)
- b. WOULD the resultant peak from a 100% co 50% power maneuver be of the same magnitude as the result peak from a 50% to 0% power maneuver? EXPLAIN your answer. (1.0)
NOTE: Consider all production and removal mechanisms in your answer. QUESTION 5.02 (1.00) The reactor is critical at 10E+6 cps. A stable period of 60 seconds is achieved. If rods are inserted continuously until the period decreases to infinity and then the rod insertion is immediately stopped, WILL the reactor be (CRITICAL, SUPERCRITICAL, or SUBCRITICAL) in the time following the rod stoppage? EXPLAIN. (1.0) 1 (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3 THERMODYNAMICS QUESTION 5.03 (1.00)
The reactor is critical at 10E+6 cps. WHICH of the following best describes the behavior of neutron power following a prompt insertion of negative reactivity (e.g., single rod scram)? (1.0)
- a. Neutron power drops immediately to " beta" (delayed neutron fraction) times the neutron power prior to the prompt insertion of negative reactivity.
- b. Neutron power decreases linearly with time after the initial prompt drop.
- c. After the initial prompt drop, neutron power decreases on a constant negative period; the magnitude of the period determined by the amount of negative reactivity inserted. --
- d. Because only delayed neutrons are left immediately after a negative reactivity insertion, neutron power decreases on an 80-second period regardless of the size of the negative reactivity insertion.
QUESTION 5.04 (3.00) With the plant at rated conditions, the EHC pressure setpoint (on the controlling pressure regulator) is lowered to its minimum value with the DECREASE pushbutton on the 9-7 Panel. - Assuming N0 further operator action, ANSWER the following using attached FIGURE 4.
- a. WHY does APRM power gradually decrease at Point 147 (0.5)
- b. WHAT is causing total steam flow to be greater than 100% rated flow between Point 5 and Point 67 (0.5)
- c. WHY did total feed flow increase to full scale at Point 177 (0.5)
- d. WHAT caused total feed flow to go to zero at Point 187 (0.5)
- e. WHAT is indicated by the oscillations in the wide range reactor pressure trace? (0.5)
- f. WHY do the peaks in the pressure oscillations occurring '
in wide range reactor pressure trace become farther apart with time? (0.5) (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 4 THERMODYNAMICS QUESTION 5.05 (2.00)
Given the following two (2) conditions and using the supplied information, DETERMINE which condition is operating MORE CLOSELY to its MCPR limit. (SHOW all work and STATE any assumptions.) K-f graph is provided. (2.0) Condition 1 Condition 2 Core flow = 54.25 Mlb/hr Core flow = 81 Mlb/hr Rx power = 1660 MW Rx power = 2490 MW P-1 MCPR = 1.57 P-1 MCPR = 1.37 Tau = 0 Tau = 0 Core flow limit = 117.0 Core flow limit = 117.0
-~
QUESTION 5.06 (1.50) STATE whether the following situations would (INCREASE, DECREASE, or NOT CHANGE) control rod worth,
- a. Restart 10 hr following a scram from 100% power condition (peripheral rod only). (0.5)
- b. Second rod in a rod group following the withdrawal of the first rod in that group. (0.5)-
- c. Localized voiding of region not previously voided. (0,5)
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
y
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 5 THERMODYNAMICS 21'>
QUESTION 5.07 (1500) You are currently operating at 100% power BOL when you lose partial feedwater heating.
- a. If the STA tells you that reactor coolant temperature decreased by 10 deg. F, voids decreased by 2%, WHAT WOULD BE the corresponding temperature change to the fuel temperature?
(ASSUME no rod movement, no recirculation flow changes and the reactor reactivity returns to zero.) (1.0)
- 1. increase by 30 deg. F
- 2. decrease by 30 deg. F
- 3. increase by 300 deg. F
- 4. decrease by 300 deg. F _ _ .
- b. If the same situation were to occur at EOL, WOULD the fuel temperature difference be (LARGER, SMALLER, or NOT CHANGE)? WHY? (INCLUDE the effect for each coefficient.) (1.0)
- c. If the same situation was to occur at 50% power, WOULD
.. the fuel temperature difference be (LARGER THAN, SMALLER THAN, or NOT CHANGE) at 100% power? WHY? (INCLUDE the u l')
effect for each coefficient.) J1<07 l l l l l l l l 1 (***** CATEGORY 05CONTINUEDONNEXTPAGE*****)
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 6 THERMODYNAMICS QUESTION 5.08 (2.00)
Figure 8 shows plots of the coolant temperature and coole.it enthalpy as a function of the height in the core or flow path length,
- a. EXPLAIN WHY the coolant temperature remains constant over much of the flow path while the enthalpy continues to rise. (1,0)
- b. EXPLAIN WHY the slope of the coolant enthalpy curve increases from Pt. A to Pt. 8 then decreases from Pt. 8 to Pt. C. (1.0)
~
.4 6
N __
\
COOLANT TEMPERATURE
\ =\ \ -
FUEL ROD SURFACE
\ TEMPERATURE I \
Z \ ERA E E \
- I 5
l sutusoLao COoLAuf EufnALPy I .-- 67 g l I cTl FUEL g
/ FLM ROC g
/ l NUCLEATE
[ g BOLNO g
/ sue. ! sus- I
/
/
We'?fi"l t f.L"l
*f CO% SOTTOM TEMPERATURE, ENTHALPY lWAT FLUE l
Figure 8. Plot of Coolant Temperature Enthalpy. l l l (***** CATEGORY 05CONTINUEDONNEXTPAGE*****)
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 7 THERMODYNAMICS QUESTION 5.09 (2.00)
Given a large vented condensate tank 30 ft in diameter and 60-ft high with a small capacity centrifugal pump taking a suction from its base. The pump is located at a vertical elevation corresponding to the bottom of the tank and it requires 5 ft of net positive suction head (NPSH) to prevent cavitation. The tank is entirely full of water (with no make-up available) and is maintained at 60 deg. F by heaters. ASSUME a frictionless system and that the vent becomes totally clogged (fully closed) while the pump is in operation. ANSWER the following questions.
- a. WHAT is the lowest pressure (at the bottom of the tank) that the tank will drop to as the pump continues to __
remove water from the tank? (0.5)
- b. WILL the pump lose NPSH and begin to cavitate prior to reaching a level of 5 ft in the tank? (STATE any assumptions.) (0.5)
- c. COULD the pump continue to pump water at a level below 5 ft without cavitation if the vent were open? EXPLAIN. (1.0)
QUESTION 5.10 (2.50) Your reactor operator informs you that MAPRAT is 1.02:
- a. IS the MAPRAT, as stated, conservative? EXPLAIN your answer and your actions, if any? (1.0)
- b. In regards to MAPRAT, which of the following statements are TRUE and which are FALSE. (1.5)
- 1. Maintaining MAPRAT within limits ensures that transition boiling will not occur in 99% of the fuel bundles.
- 2. Maintaining MAPRAT limits ensures that the APLHGR limits are met.
- 3. Maintaining MAPRAT limits ensures that peak clad '
temperatures will not reach 2200 deg. F during a LOCA. (***** CATEGORY 05CONTINUEDONNEXTPAGE*****)
e
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 8 THERMODYNAMICS QUESTION 5.11 (1.00)
Consider that Browns Ferry is operating at 100% power when a leak results at the Recirc A Reactor Water Cleanup Suction location. As the effluent discharges from the leak into the drywell, it WILL: (1.0)
- a. all flash to superheated steam,
- b. all flash to saturated steam.
- c. part flash to steam and part stay liquid.
- d. remain as a liquid.
QUESTION 5.12 (1.00) During heat up, WHICH of the following best describes the process that contributes the most significant amount of energy (provides the most thermal power)? (1.0)
- a. The fission process yields large fission fragments with high kinetic energy, which is converted into heat,
- b. The fission process yields neutrons with high kinetic energy, which is converted into heat. '
- c. The fission process yields gamma radiation with high energy, which is converted into heat.
- d. The fission process yields beta particles with high energy, which are converted into heat.
(***** CATEGORY 05CONTINUEDONNEXTPAGE*****)
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 9 THERMODYNAMICS QUESTION 5.13 (1.00)
The reactor trips from full power, equilibrium xenon conditions. Four (4) hours later the reactor is brought critical and power level is main-tained on range 5 of the IRMs for several hours. Which of the following statements is CORRECT concerning control rod motion during this period?
- a. Rods will have to be withdrawn due to xenon build-in.
- b. Rods will have to be rapidly inserted since the critical reactor will cause a high rate of xenon burnout.
- c. Rods will have to be inserted since xenon will closely follow its normal decay rate.
- d. Rods will approximately remain as is as the xenon establishes its equilibrium value for this power level.
iku QUESTION 5.14 (2.00) J The attached figure represents a transient that could ocur at a BWR. Given: An inadvertent initiation of HPCI at 10% power No other operator actions are taken. ( Recorder speed: 1 division = 1 minute EXPLAIN the cause(s) of the following recorder indications: (2.0) S 4v,, 2?n e ao
- a. Eere flow DEC4fM9E at point JF #3
- b. NO-power Optbt point #15 due tu use point #5 / hee- <<" *A M / & # 5rc::ure 3 pike et
- c. Reactor vessel level INCREASE at point #13
- d. Total feedwater flow DECREASE at point #28 /7
~
(***** END OF CATEGORY 05 *****)
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 10 QUESTION 6.01 (1.00)
WHICH level indication range (s) is (are) used for RPS trip functions? EXPLAIN the reason (s) for using these (this) range (s) for trip functions. (1.0) QUESTION 6.02 (3.00) With respect to the Browns Ferry main turbine system:
- a. STATE and EXPLAIN the two (2) reasons for the stop valve internal bypass valve. (2.0) kW _.
- b. WHAT is the purpose of the feedwater' extraction non-return valves valves and EXPLAIN how this purpose is accomplished? (1.0)
QUESTION 6.03 (3.00) With regard to the Condensate and Feedwater System
- a. HOW MANY demineralizers are required for 100% flow? (0.5) i
- b. WHAT is the capacity of the condensate demineralizer bypass valve? (in % flow) (0.5)
~
- c. At WHAT delta P across the filter demineralizers does l the bypass auto open? (0.5)
- d. Upon loss of air to the bypass valve, DOES the bypass valve: (0.5)
- 1. fail open
- 2. fail closed
- 3. stay as is
- e. DOES the main flow path of the condensate /feedflow go through the tube side or shell side of the low pressure feedwater heaters AND EXPLAIN WHY, (1.0)
(***** CATEGORY 06CONTINUEDONNEXTPAGE**"**)
1
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 11 QUESTION 6.04 (2.50)
Both the SRM and IRM compensate their detector signals with a unique type of discrimination process,
- a. Briefly DESCRIBE WHY discrimination is needed in the SRM and IRM range and DESCRIBE WHY it is not needed in the APRM range. (1.0)
- b. STATE the methods by which each system SRM/IRM accomplishes the discrimination process. (0.5)
- c. EXPLAIN WHY there is a difference between the SRM and IRM discrimination process? (1.0) 6+.
QUESTION 6.05 (3.00) The following concerns the Browns Ferry Reactor Water Cleanup System. .
- a. LIST the five (5) RWCU isolations (include setpoints). (1.5)
- b. STATE the two (2) conditions (setpoints not required) which will cause automatic closure of the blowdown valve (FCV 69-15) and EXPLAIN WHY they are needed? (1.0)
- c. WHAT system design component is used to mitigate the consequences of FCV 69-15 (flow control valve to rad waste /
condenser) failing open? (0.5) QUESTION 6.06 (2.00) An automatic HPCI initiation has occurred. Subsequently HPCI injection was automatically terminated due to high reactor water level (+54 inches).
- a. WHAT valve (s) auto close on this HPCI turbine trip? (1.0)
- b. Assuming no operator action, HOW WILL HPCI respond to a subsequent decreasing water level (to low low level, -51.5)? (0.5)
- c. If HPCI system had switched sources from the CST to the torus due to low CST level and the CST level subsequently
; recovers, WILL the HPCI system automatically switch back to the CST suction? (0.5)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE **"*)
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 12 QUESTION 6.07 (1.00)
For the control room alarm below STATE the ABNORMAL CONDITION being indicated AND briefly EXPLAIN HOW the abnormal condition could prevent the system from accomplishing its designed function when required. Core Spray valve leakage test (with a setpoint of 450 psig and increa' sing). (1.0) QUESTION 6.08 (1.00) One (1) of the required initiation signals for the ADS (Automatic Depressurization System) is a -114.5 in. water level signal. STATE the design bases for the choice of
-114.5 in, over another level for logic initiation. (1.0)
QUESTION 6.09 (2.00) For each of the following situations (i and 11) select the correct Feed-water Control System / plant response from the list (a through e) which follows. An answer may be used more than once, and N0 operator actions are taken.
- a. Reactor water level decreases and stabilizes at' a lower level.
- b. Reactor water level decreases and initiates a reactor scram.
- c. Reactor water level increases and stabilizes at a higher level.
- d. Reactor water level increases and initiates a turbine trip.
- e. None of the above.
- 1. The plant is operating at 90% power in 3-element control '
when the HPCI system inadvertently initiates and injects. (1.0)
- 11. The plant is operating at 70% power, 3-element control, when one MSIV fails closed. (1.0)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION , PAGE 13 bgW(
# t9 QUESTION 6.10 (2.00) 1 The plant is operating at 23% power and both Recirc Pump M/A Transfer Stations are in MANUAL and' set for 50% speed.
The "Recirc Flow B Limit" annunciator is CLEAR. For each of the following instances, STATE how the speed of Recirc Pump "B" will change (i.e., INCREASE, DECREASE, REMAIN THE SAME) and WHICH COMPONENT (S) of the control / positioning system is/are LIMITING. NOTE: FIGURE # 474 IS PROVIDED FOR REFERENCE
- a. Recirc Pump "B" M/A Transfer Station placed in " BALANCE". (1.0)
- b. Recirc Pump "B" M/A Transfer Station manual potentiometer is turned fully in the counter-clockwise direction.
(1.0) QUESTION 6.11 (1.00) The reactor is being started up. Assume the following: o Reactor Power is below the LPSP o Rod Withdrawal Sequence B is in effect o No RWM errors or blocks exist o The RWM Normal-Bypass Switch was positioned to Bypass following complete withdrawal of - all RWM Group 6 rods to Position 36 The operator incorrectly attempts to continue withdrawing Group 6 rods to position 48. Which one of the following most correctly details , the outcome of this operator error? NOTE: FIGURES # 467 A - C ARE PROVIDED FOR REFERENCE
- a. Rod withdrawal will occur with NO Rod Position Restrictions.
- b. Rod withdrawal will occur, provided all Group 6 rods are maintained within 1 notch of each other.
- c. Rod withdrawal will NOT occur beyond the RWM Withdrawal Limit for Group 6.
j d. Rod withdrawal will NOT occur beyond Position 36, due to RSCS immediately imposing a Rod Block. (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION ,
PAGE 14 l 1 l QUESTION 6.12 (1.00) STATE ALL of the automatic signals which will CLOSE the RBCCW Sectionalizing Valve (FCV-70-48). QUESTION 6.13 (1.00) For each of the following RHR systems, STATE the RHR system (s), (unit and system), IF ANY, with which it can be DIRECTLY cross-connected. Consider only Unit-to-Unit cross-tie capabilities.
- a. Unit 1, System I
-~
- b. Unit 2, System 1 QUESTION 6.14 (.50)
A DG automatic initiation signal is present and the operator depresses DG A's Control Room Emergency Stop PB. STATE the action which the operator must take from the Control Room to reset the automatic start lock-out. (0.5) QUESTION 6.15 (1.00) The reactor is critical at approximately 10 psig and the "HEATUP AND PRESSURIZATION" phase of G0I-100-1 is being performed. The NR GEMAC LIs in the control room read the following (approximate values). A - 37 inches B - 38 inches C - 37 inches The two (2) emergency system / accident "Yarway" control room indicators should read approximately: (CHOOSE ONE) (1.0)
- a. O inches
- b. 15 inches
- c. 38 inches
- d. +60 inches
*****)
(***** END OF CATEGORY 06
w
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 15 RADIOLOGICAL CONTROL
]
I i QUESTION 7.01 (2.00) According to Browns Ferry Radiation Protection Procedures
- a. WHAT are the administrative whole body dose limit for 1 year? (0.5)
- b. Based upon 10CFR20, WHAT is the maximum allowable whole body accumulated dose for a 30 year old person? (0.5)
- c. WHAT is the definition of radiation area? (1.0)
~~
QUESTION 7.02 (1.00) You are executing E0I-1, " Reactor Control," in response to a reactor scram and subsequent low water level. Drywell cooling has been lost due to an electrical malfunction to the fans. After 10 minutes, you receive High Drywell Pressure at 2.5 psig and slowly increasing. Which of the following is the correct action to take according to the generic guidance provided in using the E0I's.
- a. Re-enter E0I-1 at the beginning (Only required action)
- b. Continue in E0I-1 from where you are - it will direct '
you to the correct action.
- c. Exit E0I-1 and Enter E0I-2, " Primary Containment Control"
- d. Re-enter E0I-1 at the beginning AND enter E0I-2.
QUESTION 7.03 (2.50) LIST all of the entry conditions for E0I-2, " Primary Containment Control." (2.5) (***** CATEGORY 07CONTINUEDONNEXTPAGE*****)
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 16 RADIOLOGICAL CONTROL l
QUESTION 7.04 (1.00) E0I-1 RC/L-1 states: If any control rod is NOT inserted to or beyond position (02), disable ADS auto blowdown function (reference Appendix 3) and enter C5 (LEVEL / POWER CONTROL) STATE the basis for disabling the ADS auto blowdown function in this plant condition. (1.0)
~
QUESTION 7.05 (1.50) Operations Section Instruction Letter No.19 " Liquid Effluent Monitoring For RHRSW, RBCCW and RCW" states " anytime a RHRSW heat exchanger is placed in service the chemical laboratory takes a sample every 4 hours to verify compliance to 10CFR20 release rate limits".
- a. WHY is it necessary for the chemical laboratory to conduct a routine sampling for verification of compliance to 10CFR207 (1.0)
- b. When a release rate limit as specified in IP 1 of the REP is violated, as determined by the chemical laboratory analysis, WHAT action is required by the SE? (0.5)'
QUESTION 7.06 (2.00) ANSWER the following question concerning the main generator and load changes. USE the attached Power Factor Chart. You are operating at max. real load, with a 0.95 lagging power factor and H2 at 75 psig, when the load dispatcher orders you to drop the power factor to a 0.9 lagging power factor and maintain maximum MWE output. In general, HOW would you change your operating condition? (INCLUDE in your answer): (2.0)
- a. The initial state (MWe, KVAR)
- b. The Final state (MWE, KVAR)
- c. State how you would get from the Initial state to the Final state.
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 17 RADIOLOGICAL CONTROL QUESTION 7.07 (2.00)
According to 0I-71 " Reactor Core Injection Cooling System"
- a. WHAT must be done if the 250 VDC is removed from FCV-71-39 (RCIC pump discharge valve) if RCIC is required to be operable? (1.0)
- b. If the turbine is tripped by the electrical overspeed trip, DOES the trip throttle valve have to be manually reset? (0.5)
- c. During testing of RCIC, the suppression pool temperature must not be allowed to exceed WHAT temperature? (0.5)
QUESTION 7.08 (1.50) According to 01-92 " Neutron Monitoring System", during normal startup
- a. You.must. verify that all operable SRM channels have
< a count rate greater than WHAT value? (0.5)
- b. All unbypassed IRM range switches are set to position 1.
WHY? (0.5) ,
- c. WHAT values should all APRM channels be between before transferring to run? (0.5)
't
- QUESTION 7.09 (.50)
Concerning a main turbine startup per 0I-47, " Turbo-Generating
- System, WHY are you cautioned NOT to operate the turbine for prolonged times between 900 to 1500 rpm? (0.5) i I
(***** CATEGORY 07CONTINUEDONNEXTPAGE*****)
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND , PAGE 18 RADIOLOGICAL CONTROL QUESTION 7.10 (1.50)
According to BF RCI-9, Radiation Work Permits (RWPs):
- a. WHAT is the normal lifetime limitation on an RWP? (0.5)
- b. Radiation Work Permits are numbered as follows: XX-XXXX.
WHAT three (3) items of information can be identified by this numbering scheme? (1.0) QUESTION 7.11 (1.00) _. Attached Figure # 437 , "Drywell Spray Initiation Pressure Limit", is used in conjunction with the CAUTION provided in E0I-2, " Containment Control",
"If torus pressure exceeds 14.5 psig BUT ONLY IF '
drywell temperature and pressure are within the Drywall Spray Initiation Limit, THEN ....
... initiate drywell sprays at rated flow."
STATE the basis for this Drywell Spray Initiation Pressure Limit. (1.0) QUESTION 7.12 (1.00) GIVEN: The D/G CO-2 System has initiated in its Vital Area - Per BF-0I-39, " CO-2 Storage, Generator Purging, and Fire Protection": l l STATE what action (s) must be taken AND where it/they must be performed to enable the D/G C0-2 System to reinitiate (upon receipt of a valid signal). (1.0) NOTE: ASSUME THAT AN ADEQUATE CO-2 SUPPLY STILL EXISTS. QUESTION 7.13 (2.50) LIST ALL the entry conditions for Procedure E0I-1,
" Reactor Control." (2.5)
I (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 19 RADIOLOGICAL CONTROL ;
i I J
- QUESTION 7.14 (4.00)
According to Browns Ferry UNIT 1 EPM-Control Room Abandonment Procedure
- a. WHAT are six (6) actions the operator should perform if sufficient time is available? (3.0)
- b. WHERE would the Unit 1 unit operator go? (0.5)
- c. WHERE would the Unit 3 unit operator go? (0.5)
~~
C'JESTION 7.15 (1.00) Per the E0I's, DEFINE " Adequate Core Cooling." (1.0) - (*****ENDOFCATEGORY07*****)
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 20 QUESTION 8.01 (2.00)
STATE whether each of the following scrams need to be operable OR need NOT to be operable in the COLD SHUTDOWN CONDITION 7 (2.0)
- a. main steam line high radiation scram
- b. IRM high flux scram
- c. scram discharge volume high level scram
- d. low reactor water level scram
- e. APRM 15% flux . scram
- f. drywell high pressure scram
- g. manual scram
- h. SRM high flux scram QUESTION 8.02 (1.00)
The Division 1 diesel is operating and is 30 minutes into a surveillance test when the air starting system fails. The maintenance repair team estimates a 2-day minimum repair time. (USE the attached Tech. Spec. to EXPLAIN your answers to the following questions.) , IS the diesel generator inoperable according to Tech. Spec.? EXPLAIN. (1.0) 1 (***** CATEGORY 08CONTINUEDONNEXTPAGE*****)
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 21 QUESTION 8.03 (3.00)
ANSWER the following questions with action (s) that most accurately detail the allowances and/or limitations imposed by the Technical Specifications.
- a. IS IT PERMISSIBLE to go from start up to run if IRMs A, B, and C are inoperable? EXPLAIN. (1.0)
- b. If the same IRMs were found inoperable while in run, WOULD you violate any Technical Specifications by:
- 1. Staying in run? EXPLAIN. (1.0)
- 2. Placing the mode switch in Startup? EXPLAIN. (1.0)
~
NOTE: APPLICABLE TS's ARE ENCLOSED FOR REFERENCE QUESTION 8.04 (2.50) A weekly surveillance, normally performed on Wednesday, was performed on the following days due to manpower limitations over the Christmas Holiday.
~
Wednesday - December 18 Monday - December 23 5 days from last survey) Tuesday - December 31 8 days from last survey) Wednesday - January 8 8 days from last survey) - l
- a. HAVE'the surveillance requirements been exceeded for this set of dates (YES/N0)? EXPLAIN your answer. (1.5)
- b. WHEN is the maximum allowable date that the next surveillance can legally be performed? (INCLUDE HOW you determined this date.) (1.0) r
( (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 22 QUESTION 8.05 (1.50)
The reactor is operating at 100% reactor power and 100% recirculation flow with the two (2) recirculating loop flows matched. During a routine surveillance, it is found that the calibrated jet pump diffuser to lower plenum, as well as the diffuser to throat differential pressure readings (on the same jet pump), are both downscale. STATE whether the LC0 for jet pumps has been met. (YES/N0) EXPLAIN the bases for your decision. (1.5) NOTE: APPLICABLE TS's ARE ENCLOSED FOR REFERENCE QUESTION 8.06 (1.50) __ The HPCI outboard steamline isolation valve (FCV-73-3) is found to be failed in the stuck open position. Maintenance is currently attending to the problem. By using the attached Technical Specifications STATE whether HPCI is OPERABLE or IN0PERABLE and GIVE ANY necessary action statement (s) required. (1.5) NOTE: APPLICABLE TS's ARE ENCLOSED FOR REFERENCE QUESTION 8.07 (2.50) STATE whether the following events WOULD or WOULD NOT be a 1-hour reportable event to the NRC. (2.5) mb .
- a. While performing rounds it was observed that the 99 pet pumps mechanical overspeed trip had not been locally reset following system testing.
- b. A maintenance worker loses 1 week of work after receiving a strained back while repairing a feedwater turbine.
- c. While operating at 80% power with a 108% rod pattern, a MSIV rapidly closes. No scram occurs. (POWER to FLOW MAP provided).
- d. You experience a loss of control power to the control ,
room. (Loss-of.- one-125VOCW
- e. HPCI fails to start following a reactor scram due to a faulty ramp generator circuit.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE ***W*)
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 23 QUESTION 8.08 (1.00)
STATE whether the following questions on minimum shift staffing are TRUE or FALSE.
- a. If three (3) units are in operation, you are required to have at least three (3) SR0s and four (4) R0s. (0.5)
- b. If one (1) unit is in operation, the minimum number of licensed operators is four (4). (0.5)
QUESTION 8.09 (2.00) A reactor operator candidate is performing a valve walkdown and line up of the HPCI system and discovers that the minimum flow valve had been tagged out for maintenance. Upon checking the valve identification tag he also discovers that it does not correspond to the procedural valve number. For this situation, ANSWER the following questions TRUE or FALSE. -
- a. If the reactor operator contacts the control room and
-discovers that the valve had been cleared 1 week prior, he may remove the tag. (Valve in normal position.) (0,5)
- b. The reactor operator, once he verifies that the valve identification tag was mismarked, can retag the valve without a second person's approval. (0.5) c..If the procedure was in error, the mistake is classified as a non-intent change and can be corrected on the spot by the R0 with only concurrence from one (1) senior reactor operator (i.e., temporary procedure change). (0.5)
- d. You must reshow operability of the minimum flow valve, prior to removal of the maintenance tag, even though the valve had been cleared I week earlier. (All the forms had been properly approved and listed the correct valve numbers.) (0.5) 4 l
o (***** CATEGORY 08CONTINUEDONNEXTPAGE**"**)
s
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 24 QUESTION 8.10 (1.50)
The Shift Engineer is required by Operator Section Instruction Letter No. 34 " Report - Daily Plant Operations Report by Phone" to provide a report by phone to the Plant Operations Section. LIST five (5) items that must be included in this report. (1.5) QUESTION 8.11 (.50) With regard to the Shift Engineer's Key Cabinet (Operations _ Section Instructions Letter 16, " Shift Engineer Key Cabinet Audit") WHICH BOP shift engineer is responsible for auditing the Shift Engineer's Key Cabinet? (0.5) QUESTION 8.12 (1.00) . WHICH of the following scenarios requires application of the Power Transient Fuel Cladding Safety Limit of the Unit 1 Technical Specifications? (1.0)
- a. Reactor power is at 70% RTP; a steam leak to the drywell occurs and drywell pressure rises; the reactor SCRAMS at approximately 2 psig; Diesel Generator auto-initiation does not occur, but manual start is successful; the reactor is brought to a cold shutdown condition.
- b. Reactor is in Start-Up, at 7% RTP; power is increased by rod pull; the reactor SCRAMS at 10.5% power, by APRMs; level and pressure are maintained by normal systems for the plant status.
- c. Reactor power is at 42% RTP; the main turbine trips due to an EHC malfunction; the reactor SCRAMS based on the Turbine Trip; the BPVs control pressure thereafter,
- d. The reactor is at 18% RTP; 1-1/2 BPVs are open in preparation for turbine warmup; controller failure reduces pressure to 875 psig; MSIVs close; the operator then manually SCRAMS the reactor; level and pressure are maintained by normal systems for the plant status.
l (***** CATEGORY 08CONTINUEDONNEXTPAGE*****)
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 25 QUESTION 8.13 (1.50)
Concerning the administrative controls on sign-offs in G0I-100-1,
" Integrated Plant Operations":
- a. DIFFERENTIATE between the sequencing requirements for completing and signing-off STEPS and SECTIONS in this procedure (Specifically address the differences in the requirements and in order which they must be completed.) (1.0)
- b. STATE the methods which you should use to identify and sign-off Non-Applicable portions of the procedure. (0.5)
QUESTION 8.14 (1.50) _ Unit 1 Technical Specifications specify for REACTIVITY CONTROL ...
"A sufficient number of control rods shall be oper-able so that the core could be made subcritical in ..."
LIST the three conditions / assumptions which must be met to verify this " Reactivity Margin - Core Loading." (1.5) QUESTION 8.15 (1.00) - Per OSIL-43, which. of the following is the proper method for CONFIRMING the position of a locked valve. (1.0)
- a. Turn the valve hand wheel in the OPEN direction; confirm the locking device integrity and proper installation.
- b. Turn the valve hand wheel in the CLOSED direction; confirm the locking device integrity and proper installation.
- c. Turn the valve hand wheel in the DESIRED POSITION direction; confirm the locking device integrity and proper installation.
- d. DO NOT turn the valve hand wheel - confirm valve position by observing the stem position; confirm locking device integ-rity and porper installation.
(***** CATEGORY 08CONTINUEDONNEXTPAGE*****) I
m i
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 26 QUESTION 8.16 (1.00)
As per Standard Practice 14.15, " Removal of Fire Protection Equipment From Service", STATE the requirements which must be met if a fire protection system is to be removed from service for testing or servicing for a period of time exceeding one (1) hour. (1.0) (***** END OF CATEGORY 08 *****) (*************ENDOFEXAMINATION***************)
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 27 THERMODYNAMICS ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-SLY, G.
ANSWER 5.01 (2.00)
- a. ( Xe(eq.) = Production / Removal. The Production term is a flux dependent term and the Removal term is dependent on burnup and decay with only burnup being flux dependent.}
While all the production term would reduce by one-half, (+0.5) fs., gc, j3/ only the burnup portion of the removal term would reduce by 4jy' .,
",'"f ;
one-half; higher equilibrium thereby leaving level (+the 0.5). xenon concentration h go /. 4 . faat/4 a L -"> " e / ' h', (+1.0 Total) @,p,k~# , n , , , r , r g r g.-.
- b. No (+0.25), because the xenon peak following a scram is larger due to the lose of the burnout decay term and does not affect the production term. (+0.75).
(+1.0 Total) REFERENCE
- 1. BFNP, Xenon and Samarium, pp. 2, 3, 4, Student Learn-ing Objectives 1, 3, 4.
ANSWER 5.02 (1.00) Supercritical (+0.5). ( When the period reaches infinity, the reactor is exactly critical.) After the rod insertion stops, the delayed neutron precursors that were formed in previous generations and at a higher power level tend to pull power back up (+0.5). ( Therefore, the reactor is still supercritical due to the latent effect of delayed neutrons.) (+1.0 Total) REFERENCE
- 1. BFNP, Neutron Slowing Down and Diffusion, pp. 1, 8, 9 of 24, Student Learning Objectives 1, 6.
- 2. BFNP, Reactor Power and Reactor Period, pp. 4, 5, 6 of 11, Student Learning Objectives 3, 4. ,
.o
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 28 THERMODYNAMICS ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-SLY, G.
ANSWER 5.03 (1.00)
- c. (+1.0 Total)
REFERENCE
- 1. BFNP, Reactor Power and Reactor Period, pp. 5, 6 of 11, Student Learning Objectives 1, 2, 4.
ANSWER 5.04 (3.00) _.
- a. The decreasing reactor pressure is causing an increase in core voids. (+0.5)
- b. Steam flow through the turbine bypass valves. (+0.5) ,
- c. The FWCS responding to the rapid decrease in reactor water level. (+0.5)
- d. The RFPs ran out of steam following the HSIV closure.
(+0.5)
- e. SRVs lifting to control reactor pressure. (+0.5)
- f. Less core decay heat. (+0.5)
REFERENCE i 1. BFNP Transient Pressure Setpoint Decrease, No. 17A and 178. Provide Transient Figures I e e.
a
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND ,
PAGE 29 THERMODYNAMICS ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-SLY, G. ANSWER 5.05 (2.00) sv .'d Assuming 100% core flow is 10er5 Mlb/hr (+0.25), minimum MCPR (limit) = 1.26 (+0.5) For Condition 1:
/ cts $bD % core flow - 54.25/108,4 = 50%
from Figure 3.5.2, Kf = h175 Idi (+0.25) therefore, the MPCR (limit) = 1.26(1.175) = h457 / 432 7 _ I . u v: delta (MCPR) = 1.57 - 4 A57 = 0.113 C 0 6 For Condition 2:
- 74. 0L
% core flow = 81/108.5 = 74r6%
t . ;' Figure 3.5.2 Kf = h65 (+0.25) t.07 therefore, the MCPR (limit) = 1.26(hC05) = t-25 l 5W
.w y delta (MCPR) = 1.37 - 4d5 = 0.08tr 0 011 Condition 2 is closer to limits (+0.5)
(+0.25 for math) REFERENCE
- 1. General Electric Thermodynamics, Heat Transfer, and Fluid Flow, MTC, March 1983, pp. 9-96 to 9-99.
- 2. BFNP Tech. Specifications 3/4 3.5K, Minimum Critical Power Ratio, Figure 3.5.2.
Provide Kf figure for Technical Specifications
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 30 THERMODYNAMICS ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-SLY, G.
ANSWER 5.06 (1.50)
- a. increase (+0.5)
- b. decrease (+0.5)
- c. decrease (+0.5)
REFERENCE
- 1. BFNP, Control Rod Worth, pp. 2, 3, 4, 5, 8 of 25, Student Learning Objectives 2, 5. __
ANSWER 5.07 (t06) A W
- a. 3 (+1.0)
- b. smaller (+0.25) alpha (mod) becomes less negative (+0.25) alpha (void) becomes less negative (+0.25) alpha (fuel) becomes more negative (+0.25)
(+1.0 Total)
- c. smallcr @0c25-) du "U alpha (mod) becomes less negative (+0.25)<
-alpha (void) bccemes f$5I r.eg&tive (+0.25)
. l< \< \ l alpha (fuel) tecoacs acrc ncgattve (+0.25) 4 \ LI do es ,;zi cha.,*a '
(+1.0 Total) REFERENCE
- 1. BFNP-HLT Reactivity Coefficients, pp. 2, 3, 4, 5 of 17, Student Learning Objectives 1, 2, 3.
.o
_ , . , . . _ . . . . _ - , _ _ . _ _ _._ - . , _ , . _ _ . . _ .- .____.____...,y _ _ _ _ _ - _ , _ . . . - - - . - _ - - - _ _ _ _ - . - , , - , , .
]
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 31 THERMODYNAMICS ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-SLY, G.
ANSWER 5.08 (2.00)
- a. The coolant temperature rises continuously as heat
-(energy) is absorbed until saturation conditions are reached. After this, the temperature remains constant, but the quality increases. (+1.0)
- b. Enthalpy changes continuously over the flow path because the water continues to absorb energy from the fuel. The rate of energy absorption is proportional to the heat flux (neutron flux). Since the flux is highest near the center of the core, the slope is greatest at this point.
(+1.0) -- REFERENCE
- 1. General Electric Thermohydraulics, Heat Transfer and Fluid Flow Text, pp. 9 9-33.
ANSWER 5.09 (2.00)
- a. 0.256 psig vapor pressure plus height of water -OR- 2.17 psia at the bottom (+0.5) J
- b. No (+0.5), (assuming a frictionless system. Cavitation would not occur above 5 ft)
- c. Yes (+0.5), the added pressure of 14.7 psia would allow all of the water to be removed (+0.5).
(+1.0 Total) REFERENCE
- 1. BFNP, Fluid Flow and Pumps, p. 5
- 2. G.E. T, TH, AND FF., pp. 7-91 to 7-96.
4 h em
.J
u
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS. AND ,
PAGE 32 THERMODYNAMICS ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-SLY, G. ANSWER 5.10 (2.50)
- a. The MAPRAT of 1.02 is NOT conservative (+0.25). With a MAPRAT of greater than 1, it means that the MAPLHGR has been exceeded (+0.25). Restore within limits within 15 minutes (+0.5).
MAPRAT=MAPLHGR(actual)/MAPLHGRLC0 (+1.0 Total)
~~
- b. 1. False (+0.5)
- 2. True (+0.5)
- 3. True (+0.5)
(+1.5 Total) REFERENCE
- 1. General Electric Thermohydraulics, Heat Transfer, and Fluid Flow, p. 9-74.
- 2. BFNP Tech. Specs. 3/4 S.I, p. 159.
ANSWER 5.11 (1.00)
- c. (+1.0 Total)
REFERENCE
- 1. Steam Tables.
ANSWER 5.12 (1.00)
- a. (+1.0 Total)
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 33 THERMODYNAMICS ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-SLY, G.
ANSWER 5.13 (1.00)
- a. (+1.0 Total)
REFERENCE BFNP: XENON & SAMARIUM LP, pp 4, 12 GGNS: OP-NP-514 ANSWER 5.14 (2.00)
- a. One -ta t% r;;i. u pui,,p 4rtp at 11?n netg. g g/. ef v,t clue /c [*tf-//W df _
l'"'2 C
- b. Dua ta th anticipetery "S!V claenra ecxam. p'< a ud 5 4c A r
- c. Gws t0 MSRV 0;;;r.ing cousins ; icvci spi',a. fim^' el mw o,,
i tyu {/w d e,wal d.-Dwe te th CS cett!n; h;k iv fui'e ete - <tay. g(t f 2,tj < <. I. c, . g ,_ , te r REFERENCE
- 1. BFNP, Transients, Inadvertent HPCI Start, p. 1, Student Learning Objective A.
l
^
l l e em n ., --n., - a. ,n - - - - - ---
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 34 ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-SLY, G.
ANSWER 6.01 (1.00) Only the emergency. range (-1$5 to +60' inch) range is [lem d Rur t, b 60') used.',(+0.5) This is..because this range uses Yarway instruments which do not require electricalhwer for indicatio'A (+0.5) (+1.0 Total) REFERENCE
- 1. BFNP, Lesson Plan 3, p. 6.
ANSWER 6.02 (3.00) a.(Themainstopvalvesaredssignedtoopenatlessthan 130 psid across the valveJ The internal bypass valve allows for equalizing pressure across the valves prior to opening. (+1.0) The Number 2 main stop valve internal bypass valve is also used to admit steam for chest and high-pressure shell warming. (+1.0) (+2.0 Total) .
- b. The purpose is to protect the turbine from overspeed j"'" y,j,' 3,3 ,, ,.,c on a turbine trip. (+0.5) They. protect -the- tVrbine , ",7 blades. . . (A subsequent-lowering-of pressure -in t'e
- A c G e, " J IJ " g I '
the turbine-and-heatersrdue-to-vacuum in the condenser), dde 4 4;. f.,6b. m. ,e will-cause hot-water _ from-the-heater to flash to- steam, pa gr , .n c alm s, m This wet steam _would cause7xtensive turbine blading ; Ls , p r, ,,z (w m,,, ,, de me as-f t -flows-to -the-condenser _(+0J)- e c c <. . ; . , - (+1.0 Total) REFERENCE
- 1. BFNP, Lesson Plan 10, p. 7.
- 2. BFNP, Lesson Plan 10, p. 10.
(Both lesson Objectives, BFNP, Lesson Plan 10, p. 1) '
.o e
- 6. PLANT SYSTEMS DESIGN, CONTROL. AND INSTRUMENTATION PAGE 35 ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-SLY, G.
ANSWER 6.03 (3.00)
- a. 8 (+0.5)
- b. 33% (+0.5)
- c. greater than or equal to 45 psid (+0.5)
- d. fails open (+0.5)
- e. tube side - tube side is designed to take the higher pressure (+1.0) - o R. Me. idt - Iow. ptwee. hear bl>u pus 4. ngl REFERENCE co. deaur (.skelQ 4e ret 6% exWJ Lg+ (% ga g, g,;c g,,,gy , Y-a
- 1. BFNP, Lesson 11, p. 4.
O O a
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i
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 36 ANSWERS -- BROWNS FERRY 1, 2&3 -
86/06/16-SLY, G. l ANSWER 6.04 (2.50)
- a. In the SRM and IRM range the gamma radiation is not proportional to power and must be eliminated from the power reading. (+0.5) In the APRM range the gamma radiation is eithiinated. (proportional +0.5) to power and thus need not be
(+1.0 Total)
- b. SRM = pulse height discrimination,(neutron pulse is greater than gamma) (+0.25)
IRM = squaring circuit or cambelling,(square of neutron -~ signal is greater than gamma) (+0.25) (+0.5 Total) c.SRMhasgreatersensitivity(+0.25),dealswithindividual counts (+0.25). (Large difference in detector amplitude of gamma event and neutron event.) IRM deals with average. (+0.5) (Gammaeventandneutron event difference not as easily separated. Squaring of signal creates larger difference.) (+1.0 Total) - REFERENCE
- 1. BFNP, Lesson Plan 19, pp. 6, 7.
- 2. BFNP, Lesson Plan 20, pp. 9, 10.
e ee ____.-_m....-. - . _ _ , _ _ . . _ - _ _ _ - _ _ , __ . , _ _ . _ _ . _ _ . . - .
~
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 37 ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-SLY, G.
ANSWER 6.05 (3.00)
- a. 1. Low reactor water level (+0.2), +11 inches (+0.1)
- 2. High temperature outlet of non-reg Hx (+0.2), 140 deg. F (+0.1)
- 3. SBCL initiated (+0.3)
- 4. High RWCU equipment area temperature (+0.2), 160 to 180 deg. F (+0.1)
- 5. High temperature in floor drains in RWCU area (+0.2),
160 to 180 deg. F (+0.1) -~ (+1.5 Total)
- b. Upstream low pressure (+0.25) to prevent draining the entire RWCU system piping in a siphon action to the main condenser or rad waste. (+0.25)
Downstream pressure high (+0.25) to protect downstream low pressure piping. (+0.25) (+1.0 Total)
- c. Restricting orifice (upstream) (+0.5)
REFERENCE
- 1. BFNP, Lesson Plan 13, Objectives C, D, E.
ANSWER 6.06 (2.00)
- a. HCPI turbine stop valve (73-18) (+0.5) and minimum flow bypass valve to suppression pool (73-30) (+0.5)
(+1.0 Total)
- b. Auto restart on the low-low level (+0.5) ,
- c. No (+0.5) i
.o
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 38 ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-SLY, G.
-REFERENCE ! 1. 8FNP, Lesson Plan 42, pp. 12, 26.
ANSWER 6.07 (1.00) 1 Indicates leakage back from the reactor through the testable check valve and the injection valve (+0.5) which could cause overpressurization -OR- possible rupture of the low pressure portion of the CS discharge.(+0.5) (+1.0 Total) REFERENCE ~~
- 1. BFNP, Lesson Plan No. 45, CS, p. 8 1
ANSWER 6.08 (1.00) Depressurizes the reactor vessel in time to allow fuel cooling - i i by systems CS and LPCI CRDH, (feedwater, systems (+0.5) RCIC, and HPCI (following)a fall to maintain LOCA) if other makeup [ vessel level. (+0.5) l (+1.0 TOTAL) REFERENCE - 4
- 1. BFNP, Lesson Plan No. 43, ADS, pp. 4,5 ANSWER 6.09 (2.00)
- 1. c
- 11. e (+2.0 Total,+1.0 pts,each)
REFERENCE i BFNP LPf12,P.24; TRANSIENT #20;0I-57,P.53 l i
.I 4
U
- 6. PLANT SYSTEMS DESIGN. CONTROL, AND INSTRUMENTATION , PAGE 39 ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-SLY, G.
ANSWER 6.10 (2.00) Muke3 04. ll tP
- a. Increase (50%) (0,5); Sp :d 0;a.end '.4atter (0,5)
- b. Decrease (20%) (0.5); Limit Switches on Bailey Positioner (Low Speed Mechanical Stop) (0.5)
REFERENCE EIH: L-RQ-714, Figure 4.1(8); GPNT, Vol V, Chapter 4.1 BSEP: SSM 10-2/3-A, Section 3.2.1.1, pp 25 - 30 BFNP: RECIRC FLOW CONTROL LP; RQ 84/04/02 ANSWER 6.11 (1.00)
- b. (+1.0 Total)
REFERENCE BFNP: RSCS LP; RWM LP; RQ 85/02/02 & 03 ANSWER 6.12 (1.00) Loss of Normal AC Power (0.25) in conjunction with an Accident Signal (0.25) .ca._ .14. f. U.'s., d 4 d v l uu ./ Jbd zogel -oh Adde 1 s y-/
$ ^6 0 ,'3 (w- . ~. 1 s us a <in k w d.
Low pump discharge' header pressure (0.50) REFERENCE BFNP: LP 47; 01-70 ANSWER 6.13 (1.00)
- a. None
- b. Unit 1, System 2 (0.5each)
REFERENCE BFNP: RHR LP, p 39; RQ 85/01/04
- o-
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 40 ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-SLY, G.
REFERENCE BFNP: RHR LP, p 39; RQ 85/01/04 ANSWER 6.14 (.50) Take the Control Switch to RESET REFERENCE BFNP: 0I-82 ANSWER 6.15 (1.00)
~ ~
- d. (+1.0 Total)
REFERENCE
- 1. BFNP: Hot License Lesson Plan #3
- - - . . , .,,..--...--,,_.-----,----,--,--,----.----------..e-,~. . - - + - , - - - - - - ~ - - - - - -----------r-, - - - -----v-~ -- v ~ -- -~w~----'*w '
u
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 41 RADIOLOGICAL CONTROL
, ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-SLY, G. ANSWER 7.01 (2.00) l
- a. 4 rem / year (+0.5)
- b. 5(N-18) = 5(30-18) = 60 rem (+0.5) l
- c. Any area accessible to personnel in which there exists or is likely to exist radiation such that a major portion of the body could receive an exposure rate in excess of 5 mrem /hr or in any 5 consecutive days a dose in excess of 100 milliremsy bd+t-0) n + +o 4.w c te d ese m e/ne. (t/ Cs _~
REFERENCE
- 1. BFNP: RCI-9, p. 16, 10CFR20.101 and 10CFR20.202.
ANSWER 7.02 (1.00)
- d. (+1.0)
REFERENCE
- 1. BFNP, E0I-1, Sections 2.1 and 3.3; E0I-2, Section 2.1; E0I L/P, L.0. I
- 2. BWROG E0I Users Guide, Section 2.1.
ANSWER 7.03 (2.50)
- 1. Suppression pool above 95 deg. F (+0.5)
- 3. Suppression pool level above - 1" (+0.5)
- 4. Supperssion pool level below - 6.25" (+0.5)
- 4. drywell temperature above 165 deg. F (+0.5) ;
- 5. drywell pressure above 2.45 psig (+0.5)
REFERENCE I
- 1. BFNP E0I-2, pp. 3, 4. 1
a
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 42 RADIOLOGICAL CONTROL ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-SLY, G.
ANSWER 7.04 (1.00)
~
Preclude, excessive heatzup of the Suppretsfor. Pool to precidde the A /" e " # '. " loss of'ECCS pump NPSH '(and th(subsequent inability to maintain c. gees (c od r, it
' containmejtt failur,e)g',adequa'te
(+1W) y h s t cope euchecoolin (La e, ach,0., potential u0 stable steam cond
/esdin ho e ore c u a:.t .
REFERENCE
- 1. BWROG Users Guide, EDI RC/L, P 3
- 2. BFNP: E0I L/P, C-5 ANSWER 7.05 (1.50)
- a. Because the RHRSW on-line process monitors lack required sensitivity to detect a small leak that could be in excess of 10CFR20 release rate limits. (+1.0)
- b. Initiate a IP 2. (Unusual event) (+0.5)
REFERENCE
- 1. BFNP Operations Section Instruction Letter No. 19. .
ANSWER 7.06 (2.00)
- co l InitialState-$h;75MWeand425KVAR(+0.25) i
; iso 4 l You should reduce generator load by recirc. or rods to 1r2M MWe (+Gr26), then raise reactive load (VAR) by adjusting the l l
AC8oltageregulator(+0r26). (+0.5 for order of steps) , \ h o.s ) w. , l Final State - lyeit MWe and 600 KVAR (+0.25) l /, / 5 V GO l REFERENCE
- 1. BFNP, GOI-100-1 p. 18
- 2. BFNP, 01-47, Section E.
l { l l l
U
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 43 RADIOLOGICAL CONTROL ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-SLY, G.
l ANSWER 7.07 (2.00) i a, the valve must be opened (+0.5) and its breaker placed in the inoperable rMsition (+0.5) (+1.0 Total) 44(only-on-mechanical-overspeed-trip)-(+0rE) dddE
- c. 105 deg. F (+0.5)
REFERENCE _ _ _
- 1. BFNP OI-71, p. 3.
ANSWER 7.08 (1.50)
- a. 3 cps (+0.5)
- b. because rod withdrawal blocks due to IRM downscale trip 1's bypassed on range 1 (+0.5)
- c. greater than or equal to 5% and less than or equal to 12% (+0.5)
REFERENCE
- 1. BFNP OI-92, pp. 4, 5. , G er- / t c rg.
ANSWER 7.09 (.50) This is the critical speed range of the turbine (large vibrations could occur damaging the turbine). (+0.5) REFERENCE
- 1. BFNP 01-47, pp. 20, 23. ,
i e em
- 7. PROCEDURES - NORMAL, AB' NORMAL, EMERGENCY AND ,
PAGE 44 RADIOLOGICAL CONTROL ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-SLY, G. ANSWER 7.10 (1.50) h* h A Dd of
- M 'O'l{'I-
- a. One j;;;- (+0.5)
- b. 1. Year RWP was issued
- 2. RWP sequential number 3.-If-the-RWP-is-a-rotttiiis ;""'
r O.4 (+0.33 pts each, +1.0 Total) REFERENCE
- 1. BFNP, RCI-9, pp. 2, 10. _
ANSWER 7.11 (1.00) Spray initiation above this limit may result in a containment depressurization rate which exceeds the relief capacity of the drywell and reactor building vacuum breakers.(0.3) This could result in the ne exceeded. (0.7) gative design pressure of the drywell being REFERENCE
- 1. BFNP, E01-2; PC/P-3, p. 2 ANSWER 7.12 (1.00)
Reset the System (0.5) at the local relay cabinet (0.5) REFERENCE
- 1. BFNP, BF-0I-39, Sections III.E & IV l
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 45 RADIOLOGICAL CONTROL ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-SLY, G.
ANSWER 7.13 (2.50)
- 1. Reactor water level below +11" (+0.5)
- 2. Drywell pressure > 2.45 psig (+0.5)
- 3. A condition which requires a MSIV isolation (+0.5)
- 4. Reactor pressure above 1043 psig (+0.5)
- 5. A condition which requires a scram and Reactor power is > 3% or connot be determined (+0.5)
~~~
REFERENCE
- 1. BFNP: E0I-1, pp.1 ANSWER 7.14 (4.00)
- a. 1. manually scram the reactor
- 2. check that all rods are fully in
- 3. continue to operate with normal feedwater system and bypass valve operation as long as possible
- 4. close main steam isolation valves
- 5. trip feedwater pumps '
- 6. trip main turbine
- 7. trip condensate pump 8.- trip condensate booster pump
- 9. start diesel
(+0.5 for each correct answer up to a maximum of six (6) correst answers, +3.0 Total)
- b. Unit 1 panel 25-32 or shutdown board room A. (+0.5)
- c. Unit 3 panel 25-32 or MOV board A. (+0.5)
REFERENCE
- 1. Browns Ferry Control Room Abandonment Procedure, p. 2 and Figure 2. ,
\
~
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 46 RADIOLOGICAL CONTROL ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-SLY, G.
ANSWER 7.15 (1.00) Heat removal from the reactor sufficient to restore and maintain the peak fuel clad temperature (PCT) LTE 2200 deg F. (Heat up is allowed as long as PCT will nowhere exceed 2200 deg F.) (+1.0) REFERENCE
- 1. BFNP, E0I L/P, p. 5, L.O. D.
- 2. BWROG E0I Users Guide.
=6 W
6 eo
, - , . - . ~
. , - , , - . , - . , . - - - - - . , . , - . , - - . , , - - - - - - - , , n . , ,
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS, AND LIMITATIONS PAGE 47 ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-SLY, G.
ANSWER 8.01 (2.00) a,d,s4f,h are not required b, c, g,c are required (+0.25each,2.0 maximum) REFERENCE
- 1. BFNP Tech. Specs., pp. 33-36a, Table 3.1.A ANSWER 8.02 (1.00) __
Yes (+0.5), due to failure of surveillance (4.9.A.I.a) (+0.5) (+1.0 Total) . REFERENCE
- 1. BFNP Tech. Spec., 3.9A, 3.9B, pp. 292-298.
ANSWER 8.03 (3.00) .
- a. Yes (+0.5), following putting RPS trip System A in the tripped position (+0.5)(as per action 1.A for Table 3.1.A (IRM))
(+1.0 Total)
- b. 1. No (+0.5), IRMs are not required in Condition 1 and you may stay there (+0.5)
(+1.0 Total)
- 2. Yes (+0.5), unless you had the RPS trip System A in the tripped position (+0.5).
(+1.0 Total) REFERENCE
- 1. BFNP Tech. Specs., 3/4 3.1, pp. 33, 34, 35, 36.
em J
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 48 ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-SLY, G.
Provide Tech. Specs. ANSWER 8.04 (2.50)
- a. No (+0.5), allowed to exceed weekly by 25% (+0.5) (no restriction doing them early). Also did not exceed 3.25 times interval for three (3) consecutive surveillances
(+0.5).
- b. Next surveillance would be Tuesday, January 14 (+0.5),
because you are limited by the three (3) consecutive interval limit (22 days) from December 23 (+0.5). REFERENCE - - -
- 1. BFNP Tech. Spec., 1.0 AA., p. 7.
ANSWER 8.05 (1.50) { m + ^ * " C' " > '" #" C*
- O' A.Le e.
N (+0.5), due to the ' assing p of Jx>th the_operabiltty c 4 as wcll as the# ( .A desurveillance requirements a A f, I- (t-/dL..- ett u (+1.0). M it& f M ""? ' (+1.5 Total) REFERENCE
- 1. BFNP Technical Specifications 3.6.E (Jet Pumps), pp.181, 182.
ANSWER 8.06 (1.50) Operable (+0.5), apply action for 3.7.D.2 (i.e., close the valve) (+0.5) . Valve closure makes @ CI inop then action 3.5.E.2 applies (+0.5). gg g REFERENCE b f " $'"3 ' $ '~
- i. BFNP Tech. Spec. 3/4.5.E (HPCI) and 3/4.7.D (Primary Containment),pp. 154, 242.
Provide 3/4.5.E (HPCI) and 3/4.7.D (Primary Containment), Tabhr%J,A., pp.15g, 154, 242, 243, 250-255@ tte p -i,' em
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 49 ANSWERS '-- BROWNS FERRY 1, 2&3 -86/06/16-SLY, G.
ANSWER 8.07 (2.50)
- a. no (+0.5)
- b. no (+0.5)
- c. yes (+0.5)
- d. yes (+0.5)
- e. yes (+0.5)
(+2.5 Total) __ REFERENCE
- 1. BFNP Incident Report and Reportable Occurrences Program, 31t=AP.ZZ-000, Attachment 1. r6 7- / s 2 ' .h - ,. - 4 .
6/-/5. L-ANSWER 8.08 (1.00) I#" ' 7,., p., 7.s . W Q v M -'2 N 3W U '
)
- a. f4lse (+0.5)
- b. '(+0.5) iv f 7 52- fah<- ((s' i - ' ;W) 2 V'
- d d" ')
1 ytv; a Ks
& false -{+h5) \
g,_ g \t u v. s.- A A 4-c / REFERENCE ,,
- u. , , , , . t .g ,_ ( A c . , \, c. , ,
- 1. BFNP Tech. Spec. 6.8, pp. 358, 360 66-/2.M pg. (.,
em
-.mse.
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 50 ANSWERS -- BROWNS FERRY 1, 2&3 -
86/06/16-SLY, G. ANSWER 8.09 (2.00)
- a. false (+0.5)
- b. false (+0.5)
- c. true (+0.5)
- d. true (+0.5)
REFERENCE
- 1. BFNP Standard Practice 3.11, p. 3; 14.25, p. 23, and 2.3, p. 2. -
ANSWER 8.10 (1.50)
- 1. load status of each unit
- 2. load or status changes expected to occur during next shift and reason for change
- 3. major equipment or operational problems
- 4. training drills or exercises held .
- 5. significant safety events
- 6. violation of tech. specs.
- 7. predicted delta T maximum cooling tower mode - river flow and temperature and actual maximum during last 24 hours l 8. other pertinent information requested by Division duty officer gsuw on A g,_d I of 0s 34-q,(+0.3'each,+1.5 maximum)
REFERENCE
- 1. BFNP Operations Section Instruction Letter No. 34. '
em
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 51 ANSWERS -- BROWNS FERRY 1, 2&3 -86/06/16-SLY, G.
ANSWER 8.11 (.50) Group 3 B0P Shift Engineer (+0.5) REFERENCE
- 1. BFNP Operations Section Instruction Letter 16, p. C.
ANSWER 8.12 (1.00) d (+1.0) - - - PEFERENCE
- 1. EIH, Unit 1 Tech. Spec., 1.1.C.
- 2. BFNP, Unit 1 Tech. Spec., 1.1.B.
ANSWER 8.13 (1.50)
- a. All applicable steps must be completed and signed off in each section prier to proceeding to the next section. .
(0.5) None of the actions outlined in each of the procedure sections are required to be performed in sequence. (0.5)
- b. Non-Applicable portions are NOT PRECEDED by an (R).-c'E (0.25) pr ee y . ., grces:, s s~Lf They must be initialed by the SE. (0.25) l REFERENCE l BFNP: G0I-100-1, p 1 l
l l e o ** i
)
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 52 l ANSWERS -- BROWNS FERRY 1, 2&3 -
86/06/16-SLY, G. ANSWER 8.14 (1.50) Highest worth rod (0.25) fully withdrawn (0.25) Xenon free Cold (68 deg F) (0.5each) enade sut Ja.'uJ . A b ; :.a b - to.sd +1 a ci 3 t s ' ' t- W +" 1 I '8-REFERENCE "
- W ^** h General Electric Reactor Theory, Chapter 1 EIH: U2 TS, 1.0 "SDM" BFNP: U1 TS, 3.3/4.3.A; RQ 84/03/05 & 85/02/04 Rx Theory, Chapter 1, L.0. 5.1 ANSWER 8.15 (1.00) b REFERENCE BFNP: OSIL-43, pp 6, 7; OSIL L/P, L.0. X ANSWER 8.16 (1.00)
- 1) Establish an appropriate Fire Watch. (0.5)
- 2) Implement plant procedures for the removal of fire protection equipment from service. (PM, or designee approval; Posting of manual operating instructions near the manual actuation station -
or stationing of a properly instructed fire watch at the station) (0.5) rk %. Luc 6 ; .~ , REFERENCE BFNP: SP-14.15, p 1: SP-14.15 L/P, L.0. B i l l l em
--__m _-, _ . _ _ _ , - . _ _ _ _ . - ,
4 EQUATION SHEET Where mi = m2 (density)1(velocity)i(area)g = (density)2(velocity)2(area)2 KE="{2 PE = mgh PE + KE +PgV1 = PE +KE +P Y22 l 1 2 2 where V = specific volume P = Pressure Q = mcp(Tout -Tin) Q = UA (T,y,-Tstm) Q = m(ht-h2 ) P = Po10(SUR)(t) P = Po et /T SUR = 26.06 T = (B-p)t T p delta K = (K,ff-1) CR1 (1-K,ff1) = CR2 (1-Keff2) CR = S/(1-K,ff) (1-Keff1) (1-K,ff) x 100% SDM = M = (1-Keff2 K eff decay constant = in (2) " 0.693 A1
= Ag e (decay constant)x(t) t t 1/2 1/2 Water Parameters Miscellaneous Conversions 1 gallon = 8.345 lbs 1 Curie = 3.7 x 10 10 dps 1 gallon = 3.78 liters 1 kg = 2.21 lbs
- 1 ft 3= 7.48 gallons I hp = 2.54 x 103' Btu /hr 3 6 Density = 62.4 lbg/ft 1 MW = 3.41 x 10 Btu /hr Density = 1 gm/cm 1 Btu = 778 ft-lbf Heat of Vaporization = 970 Btu /lbm Degrees F = (1.8 x Degrees C) + 32 Heat of Fusion = 144 Btu /lbm 1 inch = 2.54 centimeters 2 1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 ft-lbm/lbf-sec .
t en h
+m,,,.y-- - --- -
.. M&'sy ?
BROWNS FERRY SRO WRITTEN HANDOUT
- 1. EQUATION SHEET
- 2. EHC TRANSIENT FIGURES 17A/178 EXAM FIGURE 4
- 3. Kf CURVE FIGURE 3.5.2
- 4. INADVERTENT HPCI INITIATION TRANSIENT FIGURE 20A/B
- 5. RECIRCULATION LOGIC CONTROL EXAM FIGURE 474 ,
1
- 6. RSCS ROD SEQUENCE 'B' EXAM FIGURE 467 A-C I
- 7. GENERATOR POWER FACTOR CURVE
- 8. DRYWELL SPRAY INITIATION CURVE
- 9. TECHNICAL SPECIFICATIONS 1.0 DEFINITIONS pp. 2-2a 3/4 3.1 REACTOR PROTECTION SYSTEM pp. 31-34 -
3/4 3.5.E HPCI pp. 154-155 3/4 3.6.E JET PUMP pp. 181-182 3/4 3.7.D PRIMARY CONT. pp. 242-243 3/4 3.9 AUX. ELECTRICAL SYSTEM pp. 292-298 w
%~## \
tr EQUATION SHEET Where mi = m2 (density)i(velocity)i(area)1 = (density)2(velocity)2(area)2 KE="{2 PE = mgh PEl + KE +P Y l 1 1 = PE 2+KE2+P Y22 where V = specific volume P = Pressure Q = mcp(Tout -Tin) Q = UA (T,y,-Tstm) Q = m(ht-h2) P = Po10(SUR)(t) P = Po et /T SUR = 26.06 T = (B-p)t T p delta K = (K,ff-1) CR (1-Keff1) = CR 2 (1-Keff2) 1 CR=S/(1-Keff)
~
(1-Keff1) (l'Eeff) x 100% SDM = M = (1-Keff2) K eff decay constant = In (2) = 0.693 A1 = A0 e-(decay constant)x(t) t t 1/2 1/2 Water Parameters Miscellaneous Conversions 1 gallon = 8.345 lbs 1 Curie = 3.7 x 10 10 dps 1 gallon = 3.78 liters 1 kg = 2.21 lbs 1 ft 3= 7.48 gallons ' hp = 2.54 x 10 3Btu /hr 3 6 Density = 62.4 lbg/ft 1 MW = 3.41 x 10 Btu /hr Density = 1 gm/cm 1 Btu = 778 ft-lbf Heat of Vaporization = 970 Btu /lbm Degrees F = (1.8 x Degrees C) + 32 Heat of Fusion = 144 Btu /lbm 1 inch = 2.54 centimeters 2 1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 ft-lbm/lbf-sec . 1 l l 0
. , - - - - . - - . - - . - - , _- . - . - , _ . . - - - . _ _ _ . . _ - - - - . .______________s
17A EVENT PRES 8URE SETPOINT GECREA82 POWER 1C0% e
\
\ _
k 2 - Ns s,
's /
s~
~-- / ,
0 20 40 60 80 100 120 0 4 8 12 1' 8 CORE FLOW TOT AL STE AM FLOW ( - (X1 MILLION lbsthr) (X 1 MILLION lbsthr) l
' s e
~~
__; ___.... \ 950 970 990 1010 1030 1050 0 4 0 12 go RE ACTOR PRESSURE TURBINE STE AM FLOW (PSIG) ( X 1 MiLLIOP lbs/hr)
- A_ -,,--r--- -_,--.y,-
- e 17B EVENT PRES 8URE SETPOINT DECREASE POWER 1005
{ l l 1
~- l ,
P l l . I I q,--,
\
f '
---- k s 4 h )_
0 300 600 900 1200 1500 0 10 20 30 40 50 60 RE ACTOR PRESSURE RE ACTOR VESSE'. LEVEL C ~~ (PSIG) (IN C H E S) l J [ 18 y""
--, .t t
0 25 50 75 100 125 0 4 8 12 1e APRM (% POWER) ( X 1 MILLION lbs/hr)
b O, b ! I i I BROWNS FER*:.Y, NUCLE AR PL ANT i-FIGURE 3.5.2 ' . 4 i KgFACTORj. .. l.4 r-i . l .- , 1 l f 1.3 - , l i l. C LJ AUTOMATIC FLOW CONTROL l 1.2 - i Et u ) W i i
'- n Kg 1.1 -
h. w l MANUAL FLOW CONTROL Tube Scooption Set-Point such that Calibration posi Flowmax = 102.5 % 107.0 */ f.0 - . 112.0 */.d I
- 117.0 %
l ~ l 1 I I f f I f f .f 50 60 70 80 90 10 0 30 40
!l CORE Ft.OW,%
*K g =
1.0 for core flow 3 100% l i -.s l
20A EVENT !N ADVERTENT HPCI ST AR'T POWER 100% 5 : 1 4 - -- .. 3 :- / __ __,,-- N
~~~
,,,,, ---~
O 20 40 60 80 100 120 0 o 8 12 1' 6 CORE FLOW TOT AL STE AM FLOW (X 1 MILLION Ibs/hr) ( X 1 MILLION lbs/hr) 8 h 11 i (
)
- . / 10 :-
[
... eb ^,.. - -' '
- ' ~ .
e _ , ,, , - 9 950 970 990 1010 1030 1050 0 4 8 1' 2 16 RE ACTOR PRESSURE TURBINE STE AM FLOW (PSIO) (X 1 MILLION Ibs/ht)
20B EVENT INADVERTENT HPCI START POWER 100%
- 14 P
4 - 1 l
>_- l --.
-7 12 :- (
l j_. -- 13 O 300 600 900 1200 1500 0 10 20 30 40 50 60 - RE ACTOR PRESSURE RE ACTOR VESSEL LEVEL (PSIG) (IN C H E S) 21 l l i 20 l
)
j 16 )' * --
--- -> g 17 \ , , -- 19 l
t- 9 ab -,, ~ O15 18 0 25 50 75 100 125 o 4 8 12 16 APRM TOTAL FW FLOW (% POWER) ( X 1 MILLION lbs/hr) l
P ga 38 of 44 08/13/84 Lesson Plan 8 ELECTRO C, HYDRAULIC CONTROL SYSTEM I E2=:2 . MASTER A , At CONTROLLER o l Hl.102?. OF R ATED O OUAL RECIRC PUMP SPEED O LO 30% SPEED TO "B" LOOP (OPTIONAL SETTING) SPEED CONTROL , '
==::
M M/A A M TRANSFER DUAL FUNCTION o STATION UNIT l OPTIONAL SETTING p - - I- - - - --- q CLOSED WHEN DISCHARC!-~
' GENERATOR d?.
O tiMITER I 4 vAtvE Futt OPEN l SPEED (UNLOADED) Q 28?. l _ CLOSED WHEN F.W. FLOW >209
-- --~~- -_
SPEED =- CLOSED WHEN REACTOR
, INDICATION j _ 4. WATER LEVEL ABOVE 2T-LIMITE R g OR WHEN ALL 3 RFP C ..
g
..y ,.
75?.
..y ..
FLOWS ARE >204 T" GENERATOR SPEED ff Il I r ~
' L ERROR SIGNAL SPEED 4. I LIMITING NETWORK y y ITACHOMETERI -
SPEED DEMAND Y SPEED INDICATOR A CONTROLLER a DEMAND "Y" CONTACTS FUNCTION SHOWN FOR FIELD GENERATOR BREAKER CLOSED l FEEDBACK l l [~-~~l L A POSITIONER F LURE CAM L DETECTOR AMPLIFIE R 4 + BRAKE . POSIT SCOCP
' MOTOR TUBE
, FIGURE 2 RECIRCULATION SYSTEM FLOW CONTROL NETWORK (Shown for A Loop, Typ. for B) '
! o ! = u POWER / FLOW
,j l *
- OPERATING MAP POINT 7 ioo* ioo%
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Rsvistd 2-6-81 r
/ 1.0 DEFINITIONS The succeeding frequently used terms are explicitly defined so that a uniform interpretation of the specifications nay be achieved.
A. Safety Limit - The safety limits are limits below which the reason-able maintenance of the cladding and primary systems are assured. Exceeding such a limit requires unit shutdown and review by the Atomic Energy Commission before resumption of unit operation. Operation beyond such a limit may not in itself result in serious
, consequences but it indicates an operational deficiency subject to regulatory review.
B. Limiting Safety System Setting (LSSS) - The limiting safety system setting are settings on instrumentation which initiate the automatic protective action at a level such that the safety limits will not be exceeded. The region between the safety limit and there settings represent margin with normal operation lying below these settings. The margin has been established so that with proper operation of the instrumentation the safety limits will never be exceeded. -- C. Limiting Conditions _for Operation (LCO) - The limiting conditions for operation specify the minimum acceptable levels of system performance necessary to assure safe startup and operation of the facility. When these conditions are met, the plant can be operated safely and abnormal situations can be safely controlled. I 1. In the event a Limiting Condition for Operation and/or associated requirements cannot be satisifed because of circumstances in excess of those addressed in the specifi-cation, the unit shall be placed in at least Hot Standby within 6 hours and in Cold Shutdown within the following 30 hours unless corrective measures are completed that permit operation under the permissible discovery or until the reactor is placed in an operational condition in which the specification is not applicable. Exceptions to these requirements shall be stated in the individual specifications. This provides actions to be taken for circumstances not directly provided for in the specifications and where occurrence would violate the intent of the specification. For exampla, if a specification calls for two systems (or subsystems) to be opersole and provides for explicit requirements if one system (or subsystem) is inoperable, then if both systems (or subsystems) are inoperable the unit is to be in at least Hot Standby in 6 hours and in Cold Shutdown within the following 30 hours if the inoperable condition is not corrected. l
, , _ _ , _ _ _ , _ _ - , _ _ _ _ . , _ _ _ _ , _ . . - _ . , . . , m , .,,,_-,-.-.-v_ -
I
. l R2 vised 2-6-81 l
I 1.0 DEFINITIONS (continued)
- 2. When a system, subsystem, train, component or device is determined to be inoperable solely because its onsite power source is inoperable, or solelybecause its offsite power source is inoperable, it may be considered operable for the purpose of satisfying the requirements of its applicable Limiting Condition For Operation, ;
provided:
)
(1) its corresponding offsite or diesel power source is operable; { I and (2) all of its redundant system (s), subsystes(s), train (s), i component (s) and device (s) are operable, or likewise satisfy these requirements. Unless both conditions (1) and (2) are satisfied, the unit shall be placed in at least Hot Standby within 6 hours, and in at least Cold Shutdown within the following , 30 hours. This is not applicable if the unit is already in 41d i Shutdown or Refueling. This provision describes what additional j conditions' met be satisfied to permit operation to continue consistent with the specifications for power sources, when an offsite or onsite power source is not operable. It specifically prohibits operation when one division is inoperable because _ its offsite or diesel power source is inoperable and a system, subsysten, train, component or device in another division is inoperable for another reason. This provision permits the requirements associated with individual systems, subsystems, trains, components or devices to be consistent with the requirements of the associated electrical power source. It allows operation to be governed by the time limit of the sequirements associated with the Limiting Conditica For Operation for the offsite or diesel power source, not the individual requirements for each systei, subsystem, train, component- or device that is determined
'to be inoperable solely because of the inoperability of its offsite or -diesel power source.
D. DELETED 9 0 W 2a , l I l
x SEP 191985 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.1 REACTOR PROTECTION SYSTEM 4.1 REACTOR PROTECTION SYSTDI Apolicability Applicability Applies to the instrumentation Applies to the surveillance of and associated devices which the instrumentation and asso-initiate a reactor scram. ciated devices which initiate reactor scram. Objective Objective To assure the operability of the To specify the type and frequen-reactor protection system cy of surveillance to be applied to the protectinr. instrumenta-tion. Specification Specification A. When there is fuel in the ves- A. Instrumentation systems shall be sel, the setpoints, minimum functionally tested and calibrat-number of trip systems, and , ed as indicated in Tables 4.1.A minimum number of instrument and 4.1.B respectively, channels that must be operable for each position of the mode switch shall be as given in J t Table 3.1.A. B. Two RPS power monitoring chan~ nels for each inservice RPS 110 i sets or alternate source shall be operable.
- 1. With one RPS electric power l monitoring channel for in-service RPS MG set or alt-ernate power supply inoper-able, restore the inoper-able channel to opercble status within 72 hours or remove the associated RPS I MG set or alternate power supply from service. ~~
t 31
, Amendment Nos. ,1 ,123
LIMITING CONDITIONS FOR OPERATION DEC 121983 SURVEILLANCE REQUIREMENTS 3 .1 REACTOR PROTECTION SYSTEM 4 .1 REA_CTOR PROTECTION SYSTEM . B.2 Vith both RPS electric power B. The RPS power monitoring \ monitoring channels for an system instrumentation inservice RPS MG set or alter- shall be determined operable: nate power supply inoperable, restore at least one to 1. At least once per 6 months operable status within 30 by performance of channel minutes or remove the associated RPS MG set or f unctional te st s. . - . . alternate power supply from service. 9 e l O w a0 32 e h e _ --,--s -
" ' ' * ' #~ ~ ' ' '
x LIMITINC CarSTTIONS FOR OP21ATION SURVEILLANCZ r.IOUIREMENTS 3.5.D Equin ent Area Coolers 4.5.D Equioment Aree Caolers e'
- 1. The equipment ares cooler 1. Each equipcent ares coolers associated with esch RHE la opersted in conjunctica pump and the equipment ares with the equipment served by cooler associated with ecch that patticular coolcr; set of core spray pumps (.e. therefore, the equipcent area and C or 5 and D) must be coolers are tested at the operable at all times when same frequency as the pumps the pump or purpo served by which they serve, that specific cooler is considered to be operable.
- 2. When an equipoent area cooler is not operable, the puzp(s) served by that cooler =ust be considered inoperable for Technical Specification pur-poses.
E. Hieh Pressure Coolant Injection E. System (n?CIS) Hirh Fredsure Coolant Inicetica - System (HPCIS)
- 1. The HPCI system shall be 1. HPCI Subsystem testing chall operable: be performad as follows:
(1) prior to startup from a a. Simulated Cold Condition; or once/ Autocacic operating .- Actuation (2) vhenever there is irra- Test cycle ( diated fuel in the rese-tor vessel and the reactor b. Pump opers- Onec/ vessel pressure is greater bility conth than 122 paig, except no speciffed in specifica- c. Motor Operated tion 3.3. E. 2. Once/. Valve Opern- conth bility
- d. Flow Rate at once/3 nor=sl reactor conths l vessel opera-ting precoure i
- e. Flow Rate at Once/
150 peig op era t ir.g i cy-le The HPCI pump shall delirer
~
at leant 5000 gp= durin; ~ each flow rate test. 154 -
1.1MIT!!!C. C0!!GTTIO!:S FOP. OPE'IATION SUrsVT. ILLA?!CC RFQUIRP.MO.TS 3.5.E Hirh l'rqsqure Coolant Intcetion System (llPCIS) 4.5.E liith Preneure Cuolant InJcetion ., Systen (HPCIS)
- 2. If the HPCI system is inopera- 2. When it is determined that s, ble, the reactor may remain in the IIPCIS is inoperable the operation for a period not to ADS actuation logic, the exceed 7 days, provided the RCICS. the RNRS (LPCI), and ADS, CSS, RHRS (LPCI). and the CSS shall be demonstrated RCICS are operable. to be operable icsediately:
The RCICS and ADS logic shall be demonstrated to be operable daily thereafter.
- 3. If specifications 3.5.E.1 or 3.5.E.2 are not met, an orderly shutdown shall bc initiated and the reactor vessel pressure shall be reduced to l22 psig or less within 24 hours.
~
F. Reactor Core Isolation Cooling F. Reactor Core Isolation Coolin4 System (RCICS) System.(RCICS)
- 1. The RCICS,shall be operabic: 1. RCIC Subsysten testing shall be performed as follows:
(1) prior to startup f rom a Cold Condition; or
~ a. Simulated Auto- Onec/
matic Actuation operating (2) whenever there is irra- Test cycle '-- dinted fuel in the reac-- , tor vessel and the reac- b. Pump Operability Onec/ tor vessel pressure is month abovel22 psig, except as specified in 3.5.F.2. c. Motor Operatej Cnce/ Valve Operability :enth
- d. Flow Rata at Once/3 normal reactor months vessel operating pressure
- e. Flow Rate at 150 Onec/
psig opera ti.. ; cycle
~
The RCIC pump shall deliver at least 600 sp= during e2.:- flew test. 156 l s-v I
- t.t'ITIND M CONDtTIONS TOR OFF.R_ATION SimVEII.IANOF. *1QUIRDtENTS' L 25 a0 -
3.5.F Reactor Core Isolation tuolina 4.5.F Reactor Core Isolation Cooline
- 2. If the RCICS is inoperable, 2. When it is deter::ined that the the reactor may remain in cperation for a period not RCICS is inoperable, the RPCIS shall be demonstrated to be
{(b to execed 7 days if the operable immediately.
. HTCIS in operable during such time. .
- 3. If specifications 3.5.T.1 or 3.3.T.2 are not met, an '
' orderly ahur.down shall be initiated and the reactor shall be depressurizecd to less than 122 psig within 24 hours.
.C. Aistematic Depressurization
.C. Aut oma t i,c_De pres s uriza t ion system t,ws ) , System (ADS)
- 1. Tcur of the six valves of 1. During each operating cycle the Automatic Deprensuri- ', the following tests shall be zation System shall be performed on the ADS:
operabic: ,
- a. A siniulated aute:tatic *
(1) prior to a startup . actuation test shall bu
- frna a Cold Condition, performed prior to stars.nis or, af ter each refueling out-age. Manual surveillance (2) whenever there is irra- of the relief valves is diar.ed fuel in ttei reactor envered. in 4.6.D.2.
- CN vessel and the tesctor vessel pressure is greater than 105 p:;ig, er. cept as spectfled in 3.5.C.2 and 3.5.C.3 below. 2. When it is determined that ri..o r than two of the ADS valves sie
- 2. If thice lif the six Ans valves incapable of autoca:ic o;,: ration.
are known to be in:apabic of thi If1'C15 shall be detter.stratert autcmitic operation, the to b 2 operable imediatc;y '.uid reactor may reiuin in opera- daily thereafter as len as tton for .' prvind not to Specification 3.5.C.2 applies. er :eed 7 d.sys, providt.l the lirC1 aystem ta operable. Olote that the tu es.<nre retter fun.ti.in of thene , valves is ast.urcel bv section 3.6.D nf thuso specifleatinns and that this , hrec t ' f rat. ion nnly tit p t ien to the AIM f unes ie.n.) If more than thse: nf tne six ADr. vaj ve:; ni c kn. un te be incap-abic o f ems om it i.- npe: a t t un , ._ an Jeme il.it e nriles ly r.hutdown ' shall he in tt].it it. it h the 157 . reserar in a het sh".it...,i cosi-dit tem lu f,hourr md f ri a cold-shutdom con <lition in the ,
- i following 18 hours. ,
!s - - t Amendment No. 59 e
SEP 191984 l LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMEMTS 3.6.C Coolant Leakage 4.6.C Coolant Leaksee '
- 2. Both the sump and air sampling (
systems shall be operable during D. Relief Valves reactor power operation. From 1. Approximately one-half of ill and after the date that one of relief valves shall be bench-these systams is made or found checked or reolaced with a to be inoperable for any reason, bench-checked valve each reactor power operation is operating cycle. All 13 permissible only during the valves will have been checked i succeeding 72 hours. or reolaced upon the comole-n every second cycle. The air sampling system may be removed from service for a 2. Once during each operating period of 4 hours for cali- cycle, each relief valve bration, function testing, and shall be manually opened maintenance without providing until thermocouples and a temporary monitor. acoustic monitors downstream of the valve indicate steam
- 3. If the condition in 1 or 2 is flowing from the valve.
above cannot be met, an orderly
~ '
shutdown shall be initiated 3, The integrity of the relief and the reactor shall be shut- valve bellows shall be continously down in the Cold Condition monitored when. valves incorporatin;; within 24 hours. the hellows design are installed.
- 4. At least one relief valve shall D. Relief Valves be disassembled and inspected
- 1. When more than one relief valves each operating cycle, are known to be failed, an
.C ,
orderly shutdown shall be initiat-ed and the reactor depressurized E. Jet Pumos to less than 105 psig within 1. Whenever there is recirculation 24 hours. flow with the reactor in the i startup or run modes with both E. Jet Pumos recirculation pumps running.
- 1. Whenever the reactor is in the jet pump operability shall be startup or run modes, all jet checked daily by verifying that pumps shall be operable. If the following conditions do not it is determined that a jet pump occur simultaneously:
is inoperable, or if two or more jet pump flow instrument a. The two recirculation loop failures occur and cannot be have a flow imbalance of corrected within 12 hours, an 157. or more when the pumps orderly shutdown shall be initi- are operated at the same ated and the reactor shall be speed. shutdown in the Cold Condition within 24 hours. 0
~
L w , ,w- w----- -
, - , , w ,- - - -
6 1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 0CT 111984 4.6.E Jet pnmos ' 3.6.F Recirculation Pumn Oneration - b. The indicated value of
- 1. The reactor shall not be core flow rate varies operated with one '
from the value derived from loop flow recirculation l'o'o p out of measurements by more service for more than 24 than 10%. hours. With the reactor operating, if one c. recirculation loop is out The diffuser to lower of servico, the plant shall Plenum differential pressure reading on an be placed in a hot shutdown individual jet pump condition within 24 hours unless the loop is sooner varies from the mean of all j e t pump returned to service, differential pressures
- 2. Following one pump by more than 10%.
operation, the discharge 2. Whenever there is _. valve of the low speed pump recirculation flow with the may not be opened unless reactor in the Startup or t..e speed of the faster Run Mode and one pump is less than 50% of recirculation pump is its rated speed. operating with the
- 3. Steady state operation with equalizer valve closed, the diffuser to lower plenum both recirculation pumps differential pressure shall out of service for up to 12 hours is permitted. During be checked daily and the l differential pressure of an such interval restart of individual jet pump in a the recirculation pumps is 100p shall not vary from permitted, provided the the mean of all jet pump loop discharge temperature differentini pressures in ,
is within 75*F of the that loop by more than saturation temperature of 10%. the reactor vessel water as , de te rmine d by done pressure. The total 4.6.F _ Recirculation Pueo Operation e1apsed tine in natura1 1. circulation and one pump Recirculation pump speeds shall be operation must be no . checked and logged at least once per day. grcater than 24 hours.
- 2. No additional curveillance required.
- 3. Before starting either recirculation pump durine steady state operation, check and lov,the loop dincharce temperature and dome saturat ion ter.perature. -
t' K - 182
p1 ,Ny CONDITIONS FOR OPERATION SURVEILLANCE RECUIK AiENTS
.7.. Secondary containment 4.7.C Secondary Containeant f~ 4. If refueling zone secondary I \* == ' contaimment cannot be maintained the following conditions shall be met:
- a. Handlinr. of spent fuel and all operations over spent fuel pools and open reac-tor wells containing fuel shall be prohibited.
- b. The standby gas creatuent system suction to the re-
-fualing zone vill be blocked except for a con-trolled leakage area sized to assure the achieving of a vacuum of at least 1/4-inch of water and not over _
3 inches of water in all three reactor zones. a.. Primary Containment Isolaeion Valves D. Primary Containment Isolation Valves
- 1. Durinc. reactor power operation, 1. The primary containment isola-all isolation valvea listed in tion valves surveillance shall
. ,/
Table 3.7.A and all reactor be performed as follove: coolant system instrument line flow check valves shall be a. At least once per operating operable except as specified cycle the operable isola-in 3. 7.D. 2. tien valves that are power operated and auto-matica11y initiated shall be tested for simulated automatic initiation and closure times.
- b. At least once per quarter:
(1) All normally open power operated isolation valves '(except for tha main steam line power-operated isolation !
~ j valves) shall be fully closed and reopened.
242 O e
MAY 2 1985 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.9 AUXILIARY ELECTRICAL SYSTEM C Applicability 4.9 AUXILIARY ELECTRICAL SYSTEM {[k Applicability Applies to all the auxiliary electrical power system. Applies to the periodic testing requirements of the auxiliary Objective electrical system. Objective To assure an adequate supply of electrical power for operation of those systems required for safety. Verify the operability of the auxiliary electrical system. Specification Specification A. Auxiliary Electrical Equipment A. Auxiliary Electrical System
- 1. The reactor shall not be .
- 1. Diesel Generators started up (made critical) from the cold condition a.
unless the following are Each dicsci generator satisfied: shall be manually' started and loaded
- a. Diesel generators A, once each month to B, C, and D operable. demonstrate opera-tional readiness.
- b. Requirements 3.9.A.3 The test shall C through 3.9.A.6 are ,
continue for at le met. a 1-hour period at < 757. of rated load or
- c. At least two of the greater.
following offsite power sources are During the monthly available: generator test, the diesel generator (1) The 500-kV starting air com-system is pressor shall be available to checked for operation the units 1 and and its ability to 2 shutdown recharge air boards through receivers. The the unit I operation of the station-service diesel fuel oil transformer TUSS transfer pumps shall IB with no be demonstrated, and credit taken the diesel starting
- for the two time to reach rated _
' 500-kV Trinity voltage and speed lines. If the shall be loge.,4.
i unit 2 station- b. service trans- Once per operating former is the cycle, a test will second source, be conducted simu-I a minimum of lating a loss of -' ( l two 500-kV l lines must be available. 292
MAY 2 1985 - LI!!ITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS UNIT 1 f 3.9 AUXILIARY ELECTRICAL SYSTE31 4.9 i( AUXILIARY ELECTRICAL SYSTEli (2) The 500-kV system is offsite power and . available to the similar conditions units 1 and 2 that would exist shutdown boards with the presence through the unit of an actual safety-2 station- injection signal to service trans- demonstrate the former TUSS 2B following: with no credit (1) taken for the Deenergization two 500-kV of the emer-Trinity lines, gency buses If the unit 1 and load station-service shedding from transformer is the emergency the second buses. source, a minimum of two (2) The diesel _ 500-kV lines starts from must be ambient con-available. dition on the , auto-start (3) The Trinity signal, ener-161-kV line is gizes the emer-gency buses l available to the ' with perma- [
' units 1 and 2 nently
( i shutdown boards through both connected l common station-loads, ener-
! gizes the auto-service t rans fo rmers. connected emer-gency loads NOTES FOR (3): through load sequencing, (a) If unit 3 and operates
! is claiming for greater the Trinity than or equal line as an to five minutes offsite while its gene-source, rator is loaded see unit 3 with the emer-technical gency loads.
[, specifi-i , cations, (3) On diesel gene- ,_ i section rator breaker trip, the loads 3.9.A.1.c.2. are shed from the emergency buses and the . diesel restarts on the auto-start signal, the emergency buses are { energized with permanently 292a
U LIMITING CONDITIONS'FOR OPERATION SURVEILLANCE REQUIRE.'!ENT Y 3.9 AUXILIARY ELECTRICAL SYSTEM 4.9 AUXIL\RY ELECTRICAL SYSTEM (b) If unit 2 is in cold connected - shutdown, loads, the only one auto-connected Common emergency loads station- are energized service through load trans former sequencing, is and the diesel required. operates for greater than or (4) The Athens equal to five 161-kV line is minutes while available to the its generator units 1 and 2 is loaded with shutdown boards the emergency through a common loads. station-service c. transformer when Once a month the~' unit 2 is in quantity of diesel cold shutdown fuel available shall and unit 3 is be logged. nat claiming the d. Athens line as Each diesel generator l an offsite shall be given an i annual inspection
.C ..
source. NOTE FOR (3) AND (4): accordance with With no cooling instructions based on tower pumps or fans the manulacturer's running, a cooling recommendations, tower transformer e. may be substituted Once a month a. sample for a common of diesel fuel shall station-service be checked for transformer. quality. The quality shall be within 2. The reactor shall not be acceptable limits started up (made critical) specified in Table I from the hot standby of the latest condition unless all of revision to ASTM D975 and logged. the following conditions are satisfied: 2. DC Power System - Unit
- s. At least one offsite Batteries (250-V),
, power source is Diesel-Generator _
available as speci- Batteries (125-V) and fled in 3.9. A. I .c. Shutdown Board Battertes (250-V)
- b. Three untts I and 2 diesel generators -
shall be operable. ~ L 2') 3 k
LIMITING CONDITIONS FOR OP A ION
~
21985SuavEII.tANet nEcutaEnENTS 9 AUXILIARY ELECTRICAL SYSTEM 4.9
- c. AUXILIARY ELECTRICAL SYSTEM An additional source a.
of power consisting Every week the - of one of the specific gravity. following: voltage and temperature of the pilot cell and (1) A second offsite power source available as overall battery specified in voltage shall be l 3.9.A.I.c. measured and logged. (2) A fourth b. Every three months operable units the measurement shall 1 and 2 diesel be made of voltage of generator. each cell to nearest d. 0.1 volt, specific - Requirements 3.9.A.3 gravity of each cell, through 3.9.A.6 are and temperature of met. every fifth cell. -
- 3. These measurements Buses and Boards shall be logged.
Available
- c. A battery rated
- a. The respective start discharge (capacity) bus is energized for test shall be each common station- ,
performed and the service transformer voltage, *.ime, and designated as an output current offsite power source. measurements shall be logged at inter-
- b. The 4-kV bus tie vals not to exceed -
board is energized 24 months. and capable of supplying power to 3. Logic Systems the units 1 and 2 shutdown boards if a. Both divisions of a cooling tower the common accident transformer is signal .ogic system designated as an shall be tested every offsite power 6 months to demon-source. strate that it will j function on actuation
- c. The units 1 and 2 of the core spray 4-kV shutdown system of each
, boards are reactor to provide an energized. automatic start -
signal to all 4 units
- d. 1 and 2 diesel gene-The 480-V shutdown boards IA and IB rators.
are energized. ,
! e.
The units 1 and 2 diesel auxiliary boards are energized. 293a .
LIrlITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRE!!ESTMAY 2 1985 3.9 AUXILIARY ELECTRICAL SYSTE!! 4.9 AUXILIARY ELECTRICAL SYSTEll l f. Loss of voltage and degraded voltage b. Once every 6 months,
- relays operable on the condition under 4-kV shutdown boards which the 480-V A, B, C, and D. load shedding logic system is required l g. Shutdown busses 1 and shall be simulated 2 energized. using pendant test switches and/or push-
- h. The 480-V reactor button test switches motor-operated valve to demonstrate that (Rt10V) boards ID & IE the load shedding are energized w.i h t
logic system would motor generator /mg) initiate load sets IDN, IDA, IEN, shedding signals on and IEA in service. the diesel auxiliary boards, rmov boards,
- 4. The three 250-V unit and the 480-V shut-batteries, the four shut- down boards. _
down board batteries, a 4. battery charger for each Undervoltage Relays battery, and r.ssociated a. battery boards are (Delete'd) operable.
- b. Once every 6 months,
- 5. L'ogic Systems ,
the conditions under C .- which the loss of
- a. Common accident signal voltage and degraded logic system is voltage relays are ope rabic , required shall be simulated with an
- b. 480-V load shedding undervo!* age on each logic system is shutdowr ocard to..
operable, demonstrate that the associated diesel
- 6. There shall be a minimum generator will start.
of 103,100 gallons or
- c. The loss of voltage diesel fuel in the standby diesel generator and degraded voltage fael tanks, relays which start the diesel generators from the 4-kV shutdown boards shall be calibrated
, annually for trip .
and reset and the measurements logged. These-relays shall be calibrated as - specitied in Table 4.9.A.4.c. C t' 294
LI:tITING CONDITIONS FOR OPERATION N St CE REQUIRE!!ENTS ( 3.9 AUXILIARY ELECTRICAL SYSTE!! 4.9 AUXILIARY ELECTRICAL SYSTE!! ( d. 4-kV shutdown board
- voltages shall be recorded once every 12 hours.
- 5. '
480-V Rt!0V Boards ID and IE i
- a. Once per operating cycle the automatic transfer feature for 480-V Rt!0V boards ID and IE shall be functionally tested to verify a uto-t rans fer capability.
-{e-L- (s-O e 294a -
x LI!!! TING CONDITIONS FOR Ol'ERATION MAY 2 1985 ' SURVEILLANCE REQUIRE!!ENTS 3.9 AUXILIARY ELECTRICAL SYSTE!! 4.9 AUXILIARY ELECTRICAL SYSTEM B. Operation with Inoperable Equipment B. Operation with Inocerable - Equiement k'henever the reactor is in 1. k' hen only one offsite Startup mode or Run mode and not in a cold condition, the power source is operable, availability of electric all units 1 and 2 diesel power shall be as specified generators and associated in 3.9.A except as specified boards must be demon-herein. strated to be operable immediately and daily
- 1. t he rea t *.c r .
From and after the date that only one offsite 2. power source is available, hten a required offsite power source is reactor operation is per-missible for 7 days. unavailable to unit I because the 4-kV bus tie
- 2. From and af ter the date board or a start bus is that the 4-kV bus tie inoperable, all unit 1~
board becomes inoperable, and 2 diesel generators reactor operation is and associated boards permissible indefinitely shall be demonstrated pr,ovided one of the operable immediately and required offsite power datly thereafter. The sources is not supplied rematning offsite source
, and associated buses
- fiom the 161-kV system through the bus tie shall be checked to be board. -
energized daily.
- 3. 3. kten one of the units 1 k' hen one of the units 1 and 2 diesel generators and 2 diesel generator is inoperable, continued is found to be reactor operation is inoperable, all of the CS, RHR (LPCI and permissible during the succeeding 7 days, containment cooling) provided that 2 offsite systems and the remaining power sources are diesel generators and available as specified in associated boards shall 3.9. A. I.c and all of the be demonstrated to be CS, RHR (LPCI and operable immediately and containment cooling) daily thereafter.
systems, and the rematning 4. three units I and 2 When one 4-kV shutdown diesel generators are board is found to be qperable. If this inoperable, all remaining - requirement cannot be 4-kV shutdown boards and met, an orderly shutdown associated diesel gene-shall be initiated and rators, CS, and RIC4 (LPCI the reactor shall be shut and containment cooling) down and in the cold systems supplied by the condition within 24 remaining 4-kV shutdown . hours. boards shall be demon-( strated to be operable immediately and daily thereafter. e 295
- 1 l
LIMITING CONDITIONS FOR2OPERATIONMAY 1985 3.9 SURVEILLANCE REQUIREMENTS AUXILIARY ELECTRICAL SYSTEM 4.9 4 AUXILIARY ELECTRICAL SYSTE:1 When one units I and 2 5. 4-kV shutdown board is When a shutdown bus is - inoperable, continued found to be inoperable, reactor operation is all 1 and 2 diesel gene-permissible for a period rators shall be proven of 5 days provided that operable immediately and 2 offsite power sources daily thereafter.
- are available as specified in 3.9.A.I.c 6.
When one units 1 and 2 and the remaining 4-kV diesel auxiliary board shutdown boards and is found to be associated diesel inoperable, the remaining generators, CS, RIIR diesel auxiliary board (LPCI and containment and each unit I and 2 cooling) systems and all diesel generator shall be 480-V emergency p,ower proven operable boards are operable. If immediately and daily this requirement cannot be thereafter. met, an orderly shutdown ' shall be initiated and the reactor shall be shut down and in the cold condition within 24 hours. 5. {(( When one of the shutdown buses is inoperable, reactor operation is ( permissible for a period of 7 days. 6. When one of the 460-V diesel auxiliary boards becomes inoperable, reactor operation is permissible for a period of 5 days. 7. From and af ter the date that one of the three 250-V unit batteries and/or its associated battery board is found to be inoperable for any r,cason, continued reactor ._ operation is permissible during the succeeding 7 days. Except for routine surveillance testtag, !;RC .shaii be - notitied wtthin 24 295a '
MAY 2 1985 LI!!ITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRE!!ENTS , 3.9 AUXILIARY ELECTRICAL SYSTE!! 4.9 AUXILIARY ELECTRICAL SYSTE.'! hours of the situation, the precautions to be - taken during this perioil, and the plans to return the failed component to an operable state. 8. From and after the date that one of the 250-V shutdown board batteries and/or its associated battery board is found to be inoperable for any reason, continued reactor operation is permissible during the succeeding five days in accordance with 3.9.B.7. _ 9. When one division of the logic system is inoperable, continued reactor operation is permissible under this ( - condition for seven days, provided the CSCS requirements listed in j specification 3.9.B.3 are satisfied. The NRC shall be notified within 24 hours of the situation, - the precautions to be taken during this period, and the plans to return the failed component to an operable state.
- 10. (deleted)
- 11. The following limtting conditions for operation exist for the undervoltage relays which start the diesel generators on the 4-kV shutdown boards.
- a. The loss of voltage relay channel which '
starts the diesel generator f or a - complete loss of e 296
\
lit!ITING CONDITIONS FOR 2 1985 OPERATIONMAY 3.9 SURVEILLANCE REQUIRE!!ENTS AUXILIARY ELECTRICAL SYSTE11 4.9 voltage on a 4-kV AUXILIARY ELECTRICAL SYSTE.'I shutdown board may be inoperable for 10 days provided the degraded voltage relay channel on that shutdown board is operable (within the surveillance schedule of 4.9.A.4.b). b. The degraded voltage relay channel which starts the diesel generator for degrader voltage on a 4-kV shutdown board may be _. inoperable for 10 ' days provided the loss of voltage relay channel on that shutdown board is operable (within C. the surveillance schedule of ( N. 4.9.A.'.b). c. One of the three phase-to phase degraded toltage relays provided to detect a degraded voltage on a 4-k'/ shutdown board may be inoperable for 15 days provided both of the following conditions are satisfied.
- 1. The other two phase-to phase degraded -
voltage relays on that 4-kV
- hut town board are operable (within the survei1 lance schedule of 4.9.A.4.b).
l 296a
MAY 2 1985 LIMITING CONDITIONS FOR OPERATION SURVETT.MCE REQUIREMENTS 3.9 AUXILIARY ELECTRICAL SYSTEM 4.9 AUXILIARY ELECTRICAL SYSTEM
- 2. The loss of voltage telay channel on that shutdown board is ope.able (within the surveillance schedule of 4.9.A.4.b).
- d. The degraded voltage relay channel and the loss of voltage relay channel on a 4-kV shutdown board may be inoperable for 5 days provided the other shutdown _
boards and under-voltage relays are ope rable . (Within the surveillance schedule of 4.9.A.4.b). ~C **
- 12. Winen one 480-V shutdown board is found to be ineperable, the reactor will be placed in hot standby within 12 hours and cold shutdown within 24 hours.
- 13. If one 4S0-V RMOV board mg set is inoperable, the reactor may remain in operation for a period not to exceed seven days, provided the remaining 480-V RMOV board mg sets and their associated loads remain operable.
- 14. If any two 480-V RMOV board mg sets become -
inoperable, the reactor shall be placed in the cold :;hutdown condition w' thin 24 hours.
~
297
MAY 2 1985 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.9 AUXILIARY ELECTRICAL SYSTEM ~ 4.9 AUXILIARY ELECTRICAL SYSTEM
- 15. If the requirements for {
operating in the conditions specified by 3.9.B.1 through 3.9.B.I4 cannot be met, an orderly shutdown shall be initiated and the reactor shall be shut down and in the cold condition within 24 hours. r - i( . C 0 _
.o 297a '
a MAY 2 1985 LDfITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRE!ESTS 3.9 AUXILIARY ELECTRICAL SYSTE!! 4.9 C. Operation in Cold Shutdown AUXILIARY ELECTRICAL SYSTE}] Whenever the reactor is in cold shutdown condition with irradiated fuel in the reactor, the availability of electric power shalt be as specified in Section 3.9.A except as specified herein. 1. At least two units 1 and 2 diesel generators and their associated 4-kV shutdown boards shall be operable. 2. An additional source of power energized and capable of supplying power to the units 1 and 2 shutdown boards consisting of at least . one of the following: a. ( - One of the of fsite power sources specified in (((, 3.9.A.I.c. . l b. A third operable diesel generator.
- 3. At least one 480-V shutdown board for each unit must be operable.
- 4. One 480-V RMOV board mg set is required for each l RMOV board (ID or IE) required to support operation of the RHR system in accordance with 3.5.B.9.
O - e (s. e 298
B EQUATION SHEET Where mi = m2 (density)t(velocity)i(area)1 = (density)2(velocity)2(area)2 2 KE = "V PE = mgh PE + KE +P Y where V = specific l l 1 1 = PE 2+KE 2+P Y22 volume P = Pressure Q = mcp(Tout -Tin) Q = UA (T,y,-Tstm) Q = n(hl-h2) P = Po10(SUR)(t) P = P et /T SUR = 26.06 T = (B_p)t T p delta K = (K,ff-1) CR(1-K,ffi)=CR(1-Keff2) i 2 CR=S/(1-X,ff) (1-K,ffi) (1-K,ff) x 100% SCM = M = (1-Keff2) K eff decay constant = in (2) " 0.693 A1 = Ag e-(decay constant)x(t) t t 1/2 1/2 Water Parameters Miscellaneous Conversions 1 gallon = 8.345 lbs 1 Curie = 3.7 x 10 10 dps 1 gallon = 3.78 liters 1 kg = 2.21 lbs 1 ft 3= 7.48 gallons I hp = 2.54 x 103 Btu /hr 3 6 Density =62.4lbg/ft 1 MW = 3.41 x 10 Btu /hr Density = 1 gm/cm 1 Btu = 778 ft-lbf Heat of Vaporization = 970 Btu /lbm Degrees F = (1.8 x Degrees C) + 32 Heat of Fusion = 144 Btu /lbm 1 inch = 2.54 centimeters 2 1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 ft-lbm/lbf-sec 0 e a
20A EVENT IN ADVERTENT HPCI STAR'T POWER 100% l e 5 n 1
* %mg me 3 :- /
__,. ..-- M O 20 40 60 80 100 120 0 d, 8 12 1' 6 CORE FLOW TOTAL STE AM FLOW ( X 1 MILLION lbs/hr) ( X 1 MILLION lbs/hr) 8 . k 11 1 L
)
7 l } g 10 :_ ( a* sr -
/ ,
6 - , ,, - 9 950 970 990 1010 1030 1050 0 d , 8 1' 2 16 e RE ACTOR PRESSURE TURBINE STE AM FLOW (PSIG) ( X 1 MILLION lbs/hr) 1
20B EVENT INADVERTENT HPCI START P O W E R _100 %
=
8
. h -
P l l l 7 012 [ I - e---..._ ~ 13 1 0 300 600 900 1200 1500 0 10 20 30 40 50 60 - RE ACTOR PRESSURE RE ACTOR VESSEL LEVEL (PSIG) (IN C H E S) I l l l l
)
/ I
--- s )
I *
-~~----....._____ , . - ---
- - n
( 0 25 50 75 100 125 8 12 0 4 16 APRM TOT AL FW FLOW (% POWER) ( X 1 MILLION lbs/hr)
Paga 32 of _37 08/03/84 L:ss:n Plen 25 59 - 4 3 4 3 q s 2 1 2 1 2 7 51 4 3 4 3 4 j3 N / 47 p 2 1 2 1 2 p 1 2
]
43 4 3 4 3 4 4 p3 3 N / 39 1 2 1 2 2 p1 1 N / 35 .3.. 4 3 4 3 4 3 4 j
~
31 [ 2 1 2 j1 2 1 2 i ~ 27
~
4 3 4
/ \
p3 4 3 4 3 23 1 2 2 2 p1 1 1 . 19 3 4 4 j3 3 4 3 4 L h 15 11 2i 3
/
,1 '
4 2 3 1 4
-2 3
i s
\
4 2 f 7 07 2 1 2 1 2 \ 03 3 4 3 4 l 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58 L _ FIGURE 2 SEQUENCE B, RWM ROD GROUPS 1-4; RSCS ROD GROUPS B12 AND B34 - l
r Pag 2 36 of 37 08/03/84 Lzss:n Plan 25 s
- 7 6 6 7 7
51 47 p 8 11 10 10 11 8
]
N / _ 39 5 9 12 13 13 12 9 5
~ ~ '
35 ( 10 13 31 12 12 13 10 6
.T ?'
27 .
- 23 5 9 12 13 13 12 9 5 19 -
i 15 8 11 10
. 10 11 8 u
07
/ .
N\ 7 6 6 7 03 02 06 10 14 18 22 26 30 34 38 42 46 54 50 58 fL . FIGURE 6 NOTCH GROUPS 5-13, SEQUENCE ,B
P~ age 37 of 37 08/037 W Lesson Plan ,25 l 59 14 15 14
]
N
/
51 17 18 19 18 17 r N '
/ ,
43 16 20 21 22 21 20 16 35 N / 15 19 22 21 22 19 15
"=
3, X ~ 27
} , '15 19 22 21 ' 22 19 15 =>
19 16
/ -
N 20 21 22 21 20 16 L 15 11
/ \ J 17 18 19 18 17 07 03 L 24 15 14 J 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58 s
FIGURE 7 NOTCH GROUPS 14-22, SEQUENCE B e
. - - - - + - . . _ - - , _ , , . . - , , , . - . . - . , - - - - - - - - - . - - - , , - . -, - - - -
Page 21 of 22 08/14/85 Lesson Plan 45 Rev. 0
..'* ~
J L STEAM DRYER DRYWELL WATER LEVEL
,/ x/ ) ,..
aw W l V O h 01 E _ _0_ _ _ __ W ANr r I( ..
.s.T.: TA, .,T.
04HeseCTED TO Ofv5 A PoemWE RSAOese AT RAff0 C0esomosse es OfWER g 70 Ofwe A PAIL SAFC r. ABOVE CORE
;; PLATE pdAH
;; PRESSURE si
- g. I O1 g HI LOW [Pdl]3
! N h STANDBY LIQUID CONTROL l l - l t e ! FIGURE 4 CORE SPRAY SYSTEM PIPE BREAK DETECTION INSThUMENTATION
-o e
r I
\
figure C. DRYWELL SPRAY TNITIATICN LIMIT - E 280 ' ' ' ' ' ' ' ' ' ' '
' 2 80 E 255- -255
<
- 230- UNSAFE !g -230
- Z tu -
o cz: 205-
~_
3 # ' q e -205 2 180- <!,.~ ,h 10) . 0>-< ..emsygyd 3
- -_ .J 180 in .j
-! $4 % 30
;, ' . y,, __ _- .
I
!= "+
- m 80 80 L1 0 to 20 30 40 50 60 DRYWELL PRESSURE (PSIG)
I a O
. . ~;1 ; ----.:-.-
T~
,'k i i i .- I , START UP , RATE * -,4 m:
e
- dies I * - *
'ri if J
'M TURBINE A OPEN .
E SPEED "I
- h sesgo ER R '
m EMr
, stitcf '
y sostEncart IV REGULATRON + c" N I ' I % ',,, SPEED _8 ,
'CV REG i
spe, I a l m - , - f SPEED ERROR LOAO REJECT ' ' IVR[EG ggy OPEN WI#EN GEN. OUTPUT BREAKER IS CLOSED t- g y _l 9 Q s 3,,,ye,, START UP
, y ?
CV REGULATION N/ 'r
.,L
.% FLost 7 g, g
[' l ". i dO LOAD SELECTOR '
.h.
~
~ 0" n -
' - RUN8ACK ON LOAD REJECT n STEAM
~ --- 49 --= RUN8ACK, SYNC SPEED NOT SE LECTED 4 4
.-s
. THROTTLE I ~
- o PRESSURE -
o - t* A g O Pd t~ gggs PRESS ' F> MOT t j , WR8tNE P-8 MMD i n y _ . cominot PRESSURE SET r Agyt 44 FLovt E F 8 I +1 P'i BIAS PRESS. LOAD .l { g STEAM REG. Fto.e THROTTLE I i FSTOP ALVES V PRESSURE - TCtoSED < , s' ggI gy,,33
-c h, DEMAND VALVE
.g o ,
SMALL gypggg - CLOSE . BIAS JACK toD1 m, w o u
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. . _ _ _ _ _ __ _ ___ _ _ _ _ _ _}}