ML20214U326
| ML20214U326 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 06/04/1987 |
| From: | Burdick P, Defferding L, Gruel R, Sunderland P, Upton J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20214U301 | List: |
| References | |
| 50-255-OL-87-02, 50-255-OL-87-2, NUDOCS 8706110144 | |
| Download: ML20214U326 (81) | |
Text
{{#Wiki_filter:. _ _ _ . ._. U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-255/0L-87-02 Docket No. 50-255 License No. DPR-20 1 Licensee: Consumers Power Company Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043 Facility Name: Palisades Nuclear Plant Examination Administered At: Midland, MI and Palisades Nuclear Plant Examination Conducted: May 5, 6, 7 and 8, 1987 l Examiners: M Ncf[6 L. J. Tefferding u b [~#/~9 Date l j (.Ae M d 'N7 i R. L.-Gruel Q Date YlTlh&k 4 (nn S 4
- N')
J. W. Upton [/ ~ Date bS'D f.~/81bb%P R. Sunderlandtb/l U Date i Approved By: Nd b D < 'T. M. Burdick, Chief Date Operator Licensing Section Examination Summary i 4 Requalification examinations were caministered on May 5-8, 1937 (Report No. 50-255/0L-87-02) to five senior operators and four operators. Results: All examinees passed the examinations.
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REPORT DETAILS
- 1. Examiners L. J. Defferding, PNL R. L. Gruel, PNL
*J. W. Upton, PNL P. R. Sunderland, NRC, RIII
- Chief Examiner
- 2. Exit Meeting At the conclusion of the site visit the examiners met with representatives of the plant staff to discuss the results of the examinations. The results were combined with the results of the Replacement Examination (See Report No. 50-255/0L-87-01). Personnel present at the meeting included:
R. A. Fenech, CPCo, Operations Superintendent R. M. Rice, CPCo, Operations Manager J. G. Lewis, CPCo, Technical Director W. G. Merwin, CPCo, Training Administrator C. S. Kozup, CPCo, Technical Engineer E. A. Feury, CPCo, Supervisory Instructor R. B. Heimseth, CPCo, Supervisory Instructor L. A. Schmiedeknecht, CPCo, Supervisory Instructor E. R. Swanson, NRC, Senior Resident Inspector L. J. Defferding, Batelle Examiner J. W. Upton, Batelle Examiner R. L. Gruel, Bate 11e Examiner P. R. Sunderland, NRC, RIII Examiner The examiners made the following observations concerning your training program:
- a. Generic Weakresses
- The examinees were weak in understanding the theoretical background for system operation for components and component failures. ,
- The examinees were unsure as to personnel assignments when the control room was inaccessible.
- b. Generic Strengths The candidates exhibited (in the simulator) good training and knowledge in the handling of the plant in response to instrument failures and in response to accident conditions.
- The examinees displayed a positive attitude about the requirement to take a requalification examination.
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- c. Simulator The simulator exhibited the following problems:
Malfunction MS-11 (Turbine Bypass Malfunction did not work. Boration and Dilution modeling appeared rough.
- 3. Examinations Review Based on comments provided to the candidates during the examination and as a result of discussions with the facility staff, several modifications were made to the examinations and are noted in the enclosed answer keys.
Attachment A to this report is a copy of the facility comments and the resolutions, i l 3
ATTACHMENT A Question: 2.07 (RO) 7.05 (SR0) Comments: This question requires memorization of a resource assessment tree that is available for use during a complex casualty. It does not fall under the category of an immediate action and goes beyond what we expect an operator to recall from memory. For a question of this type, the reference material should have been furnished and the question worded to have the operator use the reference material to arrive at the proper resource tree and determine from plant conditions what the proper equipment available should be used. Recommendation: In this case, without proper reference material, allow additional answers in part b. for supplying borated water to the PCS. Safety injection tanks - approximately 200 1 50 psi Boric acid pumps - approximately 100 1 20 psi SIRW tank gravity feed - approximately 30 psi
Reference:
E0P 9.0, Functional Recovery. Resolution: This question does not require memorization of a resource assessment tree. The facility recommends allowing addition answers for part b. Part b. can be answered correctly as long as the candidate knows that the HPSI, LPSI, and containment spray pumps can be used to supply borated water to the PCS. We feel that this knowledge should be possessed by a candidate and we do not agree that this knowledge is only gained from memorization of a resource assessment tree. However, the facilities recommendation does identify an additional correct answer to Part b.; that being the use of the boric acid pumps. As question specifically requested flow paths using pumps, the safety injection tanks and SIRW tank gravity feed will not be added to the key. The answer key is modified to accept the boric acid pumps as an additional correct answer. 4
Facility: Palisades Nuclear Plant Examiner: Paul R. Sunderland i Dates of Evaluation: May 5, 6, 7 and 8, 1987 Areas Evaluated: X Written X Oral X Simulator l Examination Results: l R0 SR0 Total Evaluation Pass / Fail Pass / Fail Pass / Fail (S, M, or U) i Written Examination 4/0 5/0 9/0 S
, Operating Examination Oral 4/0 5/0 9/0 S Simulator 4/0 5/0 9/0 S Evaluation of facility written examination grading N/A Overall Program Evaluation s Satisfactory X Marginal Unsatisfactory (List major deficiency i
areas with brief l descriptive comments) 1 I Submitted: Forwarded: Appr ed: N & $'l 04s J P. R. Sunderland~ 4.~ M. Burdick,' Chief C. W. Hehl, Chief
- Operator Licensing Operations Branch Section 1
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a kf'id bu d \ %s %s# A - l U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR REQUALIFICATION EXAMINATION FACILITY: PALISADES REACTOR TYPE: PWR-CE DATE ADMINISTERED: 87/05/05 EXAMINER: UPTON/DEFFERDING/ GRUEL CANDIDATE: ANSHER KEY - N 1,%
/
INSTRUCTIONS TO CANDIDATE: Read the attached instruction page carefully. This examination replaces the current cycle factitty administered requalification examination. Retraining requirements for failure of this examination are the same as ' for failure of a requalification examination prepared and administered by your training staff. Points for each question are indicated in parentheses af ter the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up four (4) hours after the examination starts. X OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 15.00 25.00 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, NEAT TRANSFER AND FLUID FLOW 15.00 25.00 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 15.00 25.00 3. INSTRUMENTS AND CONTROLS
.00 25.00 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL '
CONTROL ao 6d.00 -
% Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.
Candidate's Signature ,
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
- 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
f 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 on one side of the paper, and write "Last Page" on tne last answer sneet.
- 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 i
down on your desk or table. .
- 12. Use abbreviations only if they are connonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the
! question and can be used as a guide for the depth of answer required. l 14. Show all calculations, methods, or assumptions used to obtain an answer 4 to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND 00 NOT LEAVE ANY ANSWER BLANK.
l l i j 16. If parts of the examination are not clear as to' intent, ask questions of the examiner only.
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- 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
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i been completed. i , i a . I 1 i
- 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 part of the answer.
- b. Turn in your copy of the examination and all pages used to answer the examination questions,
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions,
- d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
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1 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 2 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.01 (2.50) A radioactive spill had been surveyed at two (2) different times ' and from the two measurements a calculation was made for an effective decay constant.
- a. IF the effective decay constant was calculated to be 0.347 years **-1, WHAT would be the effective half life of the spill? (1.0)
- b. IF the spill was surveyed again at two (2) later times, WOULD the later-calculated decay constant be GREATER or LESS than the earlier one? EXPLAIN your answer. (1.5)
QUESTION 1.02 (1.00) Following a reactor trip from sustained operation at 100% power, the power (neutron flux) level drops immediately to about (1.0) (a.) 6% and then decreases with a decay rate that approaches an 80 sec**-1 decay constant. (b.) 6% and then decreases with a decay rate that approaches a
-1/3 dpm rate.
(c.) 99.6% and then decreases with a decay rate that approaches an 80 second period. (d.) 99.6% and then decreases with a decay rate that approaches an 80 sec**-1 decay constant. I t . (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3 THERMODYNAMICS, HEAL TRANSFER AND FLUID FLOW QUESTION 1.03 (1.00)
As core age increases during a cycle in the Palisades nuclear reactor, the integral rod worth at 100% full power of a regulating
. CEA located near the center of the core (1.0) 4 (a.) increases due to the fact that the CEA is operating in a region of higher neutron flux.
(b.) increases due to the fact that temperature of the CEA is higher. (c.) decreases due to the fact that the temperature of the CEA is higher. (d.) decreases due to the fact that the boric-acid concentration in the core is less. QUESTION 1.04 (2.00) Equilibrium xenon reactivity does not change as much for a 10% power change at a high power level as it does for a power change of 10% at a low power level. EXPLAIN. (2.0) QUESTION 1.05 (1.50) The Palisades power plant is operating at 100% of full power with all rods out (ARO). The age of the core is 8000 MWD /MTU. The boron concentration is equal to the " predicted boron concentration" for this reactor condition. It is necessary to insert the CEAs in the manual sequential (MS) mode until Group 4 is at 40 inches. A change in the boron concentration will be used to maintain 100% , power. Use the Palisades Technical Data Book as necessary.
- a. WHAT is the reactivity affect of inserting the CEAs? ,
(0.5)
- b. WHAT is the recipro:a1 boron worth for these conditions? (0.5)
I
- c. WHAT should be the new boron concentration? (0.5)
(***** CATEGORY 01CONTINUEDONNEXTPAGE*****)
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.. 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 4 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW l l l QUESTION 1.06 (1.00) During accident conditions, the following information is available: PZR pressure = 1000 psia Hottest CET temperature = 500 degrees F.
- a. WHAT is the temperature subcooled margin (SCM)? (0.5)
- b. WHAT is the pressure SCM7 (0.5)
QUESTION. 1.07 (2.00) A leak exists in the steam generator steam space while the plant is at 100% power. STATE the temperature AND the degree of subcooling/ quality / degree of superheat of the exit fluid. (2.0) QUESTION 1.08 (3.00) Included with this examination is the Characteristic Curve figure for the Primary Coolant Pump (PCP).
- a. DEFINE the term " shutoff head." For the Palisades PCP, WHAT is the shutoff head 7 (1.0)
- b. DEFINE the term " pump runout." WHERE on the figure does pump runout occur 7 (1.0)
- c. The figure shows the recommended NPSH. If the water l entered the PCP at 360 degrees F, WHAT would be the minimum PCP suction pressure in psia required to meet the NPSH recommendation? (1.0)
(***** CATEGORY 01CONTINUEDONNEXTPAGE*****)
1 1 r A 4 9 i c- , t;
- i. PAllSADES PLANT g
' m d PRIM ARY COOLANT PUMP CHAR ACTERISTIC CURVE =
T * : O -
-80 ,G :
s~ l2 w Total Head,~ Efficiency . ~ . . 70
. '.'t,'.
ft. '- -
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i ..d:
/
'. ~b0 l .- -
j . 4, . o ;
< 400 : -
Head ~
~M NPSHi ggge,,cy, R mended . .
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Min.~ NP s
- N % 5 Q o
.: o-300 -
- .100 ,
! 40 =
o . R i - --
., =
f, -... 1 200 - . + ., . Brake, {n j, a T * ~ > g2-
-10
[ 100 *
-20
! sf ' BHP @ SP. GR. = 1.0 [ ] IE. g N -
- 5,000
) 9: .
-10 4 o. .
i
%{~ . . . . . -
0 0 20 40 ; . .. 60 80 100 120 0 i
= !
1000 Gallons per Minute - t ir Og < , , *
.us . t 1
, t
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 5 f THERMODYNAMIC 5, HEAT TRANSFER AND FLUID FLOW t
QUESTION 1.09 (1.00) , The term " critical heat flux" refers to that heat flux between a fuel pin and the bulk of the primary coolant (1.0) (a.)atwhichtheheat-transfermechanismchangesbetweennatural convection and nucleate boiling. (b.) at which the heat-transfer mechanism changes between nucleate boiling and partial film boiling. (c.) at which DN8R would equal 1.3. (d.) which is 77% of that which would cause clad failure, i . l
?
- 4 I
I I i I (*"" END OF CATEGORY 01 '"")
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. 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 6 l l-
- QUESTION 2.01 (1.00) q j During a large break LOCA when the SIRW tank level reaches the :
( a RAS setpoint, WHAT design feature ensures an uninterrupted l suction flow path to the safeguard pumps? (1.0) l 5 i QUESTION 2.02 (3.00) '
- ' For each of the following condttions (a., b. and c.), LIST those actuation signals (1. through 6.) that should have triggered.
Consider each condition separately. Each condition may have for l Its answer no, one, or several actuation signal (s). (3.0) l
- 1. $1AS
- 2. CSAS I 3. Auto Containment Isolation i 4. MSIS vs. MSIV l
- 5. RAS
- 6. AFAS 1
i a. A steam-1tne break has occurred in containment and 3 l containment pressure = 7 psig containment radiation = background i S/G 1evels = 60% and 40% ! PZR pressure = 1800 psia i i SIRWT level = 80% ! S/G pressure = 680 and 580 psia - b.* A small-break LOCA has occurred and ! containment pressure = 2 psig 4 containment radiation = 6 R/hr ! S/G levels = 655 and 655 i PZR pressure = 1800 psia -
$1RWT level = 50% ,
. S/G pressure = 700 and 700 psia l
- c. A feedwater problem has occurred and j containment pressure = 1 psig i
containment radiation = background
! 5/G levels = 60% and 315 i
PZR pressure = 2200 psia SIRWT level = 805
! S/G pressure = 720 and 650 psia l
i ! i i ! I I i (***** CATEGORY 02 CONTINUE 0 ON NEXT PAGE ***") l _.--_.,-,___.-.--m--m____ . , , - -
l
- 2. PLANT DESIGN INCLUDING SAFETY AND ENERGENCY SYSTENS PAGE 7 QUESTION 2.03 (2.00)
STATE four (4) of the seven (7) automatic responses which occur . In the CVCS system when a Safety Injection Signal is generated wtth standby power available. (2.0) QUESTION 2.04 (1.50) LIST the three (3) loads that can be directly cooled from service water during a loss of CCW. (1.5) QUESTION 2.05 (2.50) The preferred AC bus Y20 supplies power to various loads.
- a. WHAT is the normal supply for the preferred AC bus Y207 (0.5)
- b. If the normal supply became unavailable, HOW would power be supplied to the preferred AC bus Y207 (1.0)
- c. Ifoffsitepowerwaslost,WHATtwo(2)Independentpower sources could be used to supply the preferred AC bus Y207 (1.0)
QUESTION 2.06 (2.00) The plant is operating at 2005 power when a reactor trip occurs
- a. If forced circulation cannot be maintained, HOW would the condition of natural circulation be verified per Palisades E0P 1.07 , (1.5)
- b. If containment isolation should have occurred and did not, HOW would it be initiated / established? (0,5)
(***** CATEGORY 02CONTINUEDONNEXTPAGE*****)
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 8 QUESTION 2.07 (3.00)
The Functional Recovery Procedure, E0P 9 Attachment A, provides-guidance for reactivity control. If CR0s cannot be inserted or driven in, two methods of boron addition are specified,
- a. LIST the three (3) flowpaths that are available to supply borated water from the charging pumps to the PCS. (1.5)
- b. LIST the three (3) flowpaths (pumas) for supplying borated water to the PCS ff the ciarging pumps are unavailable. (1~.5) 4 e
l I a 4
- 3. INSTRUNENTS AND CONTROLS PAGE 9 QUESTION 3.01 (1.50)
When operating in the Manual Indtvidual (NI) mode, LIST the thre'e (3) control rod interlocks which are NOT bypassed? (1.5) QUESTION 3.02 (1.50) With the Emergency Diesel Generator No. I controls in automatic, WHAT three (3) conditions would generate signals to automatically start'the diesel? (1.5) QUESTION 3.03 (3.00)
- a. During abnomally-high containment pressure conditions, WILL the indicated Steam Generator pressure and Pressurizer pressure be EQUAL T0, GREATER THAN or LESS THAN actual?
PROVIDE a rationale for your answer. (1.5)
- b. During abnormally-high containment temperature conditions, WILL the indicated Steam Generator level and Pressurizer level be EQUAL T0, GREATER THAN, or LESS THAN actual?
PROVIDE a rationale for your answer. (1.5) QUESTION 3.04 (2.00)
- a. If the "HIGH POWER RATE CHANNEL PRE-TRIP" alarm was activated by the RPS, WHAT two (2) other control signals / trips should have been activated? (1.0)
- b. If the " RATE TRIP CHANNEL BYPASSED" alarm was activated by the RPS, LIST the two (2) signals / trips, either of which should have been the cause. (1.0)
(***** CATEGORY 03CONTINUEDONNEXTPAGE*****)
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PAGE 10
- 3. INSTRUMENTS AND CONTROLS QUESTION 3.05 (1.00)
WHICH of the following is NOT a function of the excore NI start-up channels? To provide (1.0) (a.) core-power indications from one count per second to 10"5 counts per second (b.) core-power signal to rod control up to 10"-4% power (c.) high SUR alam at 1.5 dpm (d.) core-power signal to audio countrate amplifier for use during refueling QUESTION 3.06 (3.00) The PCS is provided with various temperature measurement and pressure taps,
- a. STATE the number of hot and cold leg RTDs AND the protec-tion or control signals developed from them. (1.5)
- b. Regarding PCS flowrate instrumentation:
- 1. STATE the number of PCS total flowrate measurement channels and the required coincidence .(logic) of these for a low-flow reacte trip. (0.5)
- 2. STATE the number of inputs to a representative PCS total flowrate channel and SPECIFY what flowrate is measured by each input. (1.0)
QUESTION 3.07 (3.00) The Reactor Protection System (RPS) generates e,leven automatic reactor trips.
)
! a. Two of the trips have automatic bypass features. STATE these two RPS trips AND the range over which they are
} (1.5) j bypassed.
!' b. STATE RPS trips including setpoints associated with the steam generators. (1.5) i (***** END OF CATEGORY 03 **"*)
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- 4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 11
, RADIOLOGICAL CONTROL QUESTION 4.01 (1.50)
If a fuel handling accident occurred in the auxiliary building fuel handling area, the control room should be notified and the fuel handling area should be evacuated. WHAT other three (3) immediate actions should be taken? (1.5) QUESTION 4.02 (1.50)
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ANSWER the following question related to Precautions and l Limitations regarding plant operation above 10% power. TRUE or FALSE. A turbine backpressure of 3" HgA is acceptable. (0.5) QUESTION 4.03 (3.00) The plant is operating at 100% power when a reactor trip occurs:
- a. WHAT actions should be taken in the event all full lingth control rods are not fully inserted? (1.75)
- b. What actions should be taken if Bus IA, IB, or IE has not transferred to startup power? (1.25)
QUESTION 4.04 (1.50) 4 STATE whether the start of each of the following operations would tend to INCREASE or DECREASE void size (if one exists in the I reactorvessel):
- a. Pressurizer spray operation .
- b. Pressurizer heater operation ,
- c. Reactor vessel head venting .
k I (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
.s. .
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s ,
, 4. PROCEDURES - NORMAL, ABNORMAL. EMERGENCY AND PAGE 12 RADIOLOGICAL CONTROL s 1
QUESTION 4.05 (.80) ' , '- If a visible sighting of a tornado was reported, WHAT immediate ' action should be taken? (0.8) QUESTION 4.06 (1.50) If power was lost to instrument AC bus YO1, WHAT immediate actions should be taken? (1.5) QUESTION 4.07 (1.20) As referred to in ONP 25.2, " Alternate Safe Shutdown Procedure," LIST the four (4) sets of items that the operators take with them when evacuating the control room? (1.2)
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QUESTION 4.08 (1.50) s Low Temperature Over-Pressure Protection (LTOP) must be operable under either of WHAT two (2) conditions when plant cooldown is proceeding from Hot Shutdown to Cold Shutdown? (1.5) r QUESTION 4.09 (2.50) You have received 1000 mrem of whole-body occupational dose in the first quarter of this year and 300 mrem in the second quarter of this year. Your cumulative dose to date is 22 rem. You are a healthy male who turned 24 on the first of the year. -
- a. According to the Palisades administrative dose limits, HOW i much additional whole-body occupational dose cah-yo's receive this second calendar quarter? SPECIFY any assumptions
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pertinent to your answer. - (1.0)_ s
- b. Based on the whole-body lifetime dose limits as set by 10CFR20, WHAT is the whole-body dose limit that you can receive this year? SPECIFY any assumptions pertinent to ycur answer. (1.5) i %.
(***** END OF CATEGORY 04 *****)
^ (*********** END OF EXAMINATION ***********)
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 13 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- PALISADES -87/05/05-UPTON, J.
ANSWER 1.01 (2.50)
- a. t-1/2 = 0.693/0.347
= 2 years [+1.0 for 1.5 to 2.5 years]
- b. Less [+0.5] . The short-lived radioactive elements decay away leaving the longer-lived isotopes (with the smaller decay constants) [+1.0].
REFERENCE
- 1. Palisades: Nuclear Operator Training, " Reactor. Theory (PQ),"
Theoretical Concepts (PQAO), Objective 4.2, pp. 8 through 11. ANSWER 1.02 (1.00) (b.) [+1.0] REFERENCE
- 1. Palisades: Nuclear Operator Training, " Reactor Theory (PQ),"
Theoretical Concepts (PQAO), Objective 4.6, p. 13. ANSWER 1.03 (1.00) (a.) [+1.0] REFERENCE
- 1. Palisades: Nuclear Operator Training, " Reactor Theory (PQ)," -
Theoretical Concepts (PQBO), Objectives 4.8, pp. 8 and 9. p
'.ti 4 i V';S .L.* .
B ,V t.- / . l 2 Y a2 %. %J m aL u %yL g I I i l
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 14
, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- PALISADES -87/05/05-UPTON, J.
ANSWER 1.04 (2.00) The production of Xe-135 is dependent on the flux or power level because the production is from fission fragments [+0.5]. The removal of Xe-135 is by burnout and by decay. The burnout of Xe-135 is power dependent while the decay term is not [+0.5]. As the power level increases, the burnout term becomes more dominant in Xe-135 removal and the condition is approached where a power-
' level change would equally affect production and removal producing no change in the equilibrium Xe-concentration [+1.0]
REFERENCE
- 1. Palisades: Nuclear Operator Training, " Reactor Theory (PQ),"
Control Poisons and Fission Poison Products (PQBO), Objective 4.27, 4.28, 4.29, 4.30 and 4.35, pp.19 through 23.
- 2. Palisades: Technical Data Book, " Equilibrium Xenon Versus Reactor Power," Figure 2.1, Rev. 2.
ANSWER 1.05 (1.50)
- a. Using Figure 5.1, the delta-rho is 0.54%. [+0.5]
~
- b. Using igure 4.1, the reciprocal boron worth 1s 97 pps/% delta rho. +0.5]
- c. delta-ppm = (97 pps/% delta rho) (0.54% delta rho)
= 52.4 ppm new concentration = 320 - 52.4 = 267.6 ppm
[+0.5] REFERENCE
- 1. Palisades: Technical Data Book, Dated 5/12/84, Figures 4.1 and 5.1.
192008K120 ...(KA'S)
= ,- -_.-...- - - . _ . - - _ _ - . _ _ . . _
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 15
. THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- PALISADES - 87/05/05-UPTON, J.
ANSWER 1.06 (1.00)
- a. 544.6 - 500 = 44.6 degrees F [+0.75]
- b. 1000 - 680.9 = 319.1 psia [+0.75]
REFERENCE
- 1. Palisades: Nuclear Operator Training, " Applied Reactor Heat Transfer ind Fluid Flow (PR)," SH-PRAO, pp. 35 and 36.
. 193008K115 . .(FA'S)
ANSWER 1.07 (2.00) Comment: The starting pressure for this calculation can range from 570 to 790 psia, p - 690 psia (saturated) h = 1202 BTU /lbm (saturated s*.eam) [+0.5] delta h = 0 [+0. 5] p-exit = 15 psi T-exit is approximately 320 degrees F [+ 0.5] 108 degrees F superheat [+0. 5] REFERENCE
- 1. Palisades: Nuclear Operator Training, " Applied Reactor Heat .
Transfer and Fluid Flow (PR)," SH-PRA0, Objective 4.35, pp. 45 and 46. I
,.,--__._-.,.-,,,3..m.
%7_m_f9..,_,_ 2 __,,.,_,_,_,_,_._,___my. ,_.y ,_,.___,,,,w _9y,, ,,- - - .,--- ,- -_.-.m. -,-.-, - -----*.yy
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 16 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- PALISADES -87/05/05-UPTON, J.
i ANSWER 1.08 (3.00)
- a. " Shutoff head" is the term used to give the pressure at which there is zero flow dischage pressure [(+0.5].except for recirculation)
The shutoff head for the PCP andis maximum 390 f t [+0.5] .
- b. " Pump runout" is the term used to describe the operation of a centrifugal pump when it is pumping at minimum head and maximum flowrate (and low NPSH and max amps) [+0.5]. It is indicated by the far-right in the figure [+0.5].
- c. NPSH = 100 ft H 0
- - 2 j = 43.4 psia sat. press. (360 degrees F) = 153.0 psia hence, min, press. = 196.4 psia [+1.0]
REFERENCE
- 1. Palisades: Nuclear Operator Training, " Applied Reactor Heat Transfer and Fluid Flow (PR)," SH-PRAO, pp.126 through 133 J
and Figure 10. ' 191004K112 191004K111 191004K105 ...(KA'S) ANSWER 1.09 (1.00) (b.) [+1.0] j REFERENCE i
- 1. Palisades: Nuclear Operator Training, " Applied Reactor Heat Transfer and Fluid Flow (PR)," SH-PRAO, pp. 50 through 52.
193008K106 ...(KA'S) l i I 4 i I ^
i e
- 2. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 17 ANSWERS -- PALISADES -87/05/05-UPTON, J.
ANSWER .2.01 (1.00) The containment sump valves start to open (one minute) before the SIRW outlet valves fully close. --OR-- The actions of these valves overlap. [+1.0] REFERENCE
- 1. Palisades: Nuclear Operator Trainin Engineered Safeguards Systems" Safety, (PNM)g, Advanced Injection (PNMA)",Systems Training, SH-PNMA, p. 25.
006000K409 ...(KA'S) ANSWER 2.02 (3.00) Comment: MSIS - such an actuation signal does not exist at Palisades. MSIVs do, of course, exist. Hence, the 2 alternate answers to "a.". The answer to "b." is changed to reflect the new setpoint of 10 R. ,
- a. SIAS, CSAS, Auto Containment Isolation, MSIS (1., 2., 3., 4.)
--OR-- (1., 2., 3.) [+1.00] l
- b. None [+0.5] ,
l
- c. None [+0.5]
REFERENCE
- 1. Palisades: Nuclear Operator Training, Advanced Systems Training, Engineered Safeguards Systems (PNM), " Safety Injection (PNMA)",
LP-PNMA, pp. 16 and 17; " Containment Spray (PNMB)", SH-PNMB,
- p. 23; " Containment Butiding (PNMF)", SH-PNMF, p.14,
- 2. Palisades: Nuclear Operator Trainin Secondary Plant Steam Systems ,
(PND)g,
" Main Advanced Systems Training, Steam (PNDB)",
SH-PN08, p. 11; " Auxiliary Feedwater (PNEB)", LP-PNEB, p. 8. 012000A301 ...(KA'S)
- 2. PLANT SYSTEMS DESIGN CONTROL AND INSTRUMENTATION PAGE 18 ANSWERS -- PALISADES -87/05/05-UPTON,J.
f ANSWER 2.03 (2.00)
- 1. Boric Acid Pumps start (P56A and B).
- 2. Coolant Charging Pumps start (P55A, B and C).
i
- 3. Boric Acid Gravity Feed Valves open (MOV-2169 and -2170).
- 4. Boric Acid Pumped Feed Stop Valve (MOV-2140) opens.
- 5. VCT Outlet Valve (MOV-2087) closes.
- 6. Concentrated Boric Acid Tanks' Recirc Valves (CV-2130 and
-2136) close.
- 7. Boric Acid Makeup Stop Valve (CV-2155) closes.
[+0.5each,+2.0 maximum] REFERENCE
- 1. . Palisades: Nuclear Operator Training, Advanced Systems Training, PCS Auxiliary and Containment (PNK), "CVCS (PNKA)",
Objective 4.24, SH-PNKA, p. 48. l ANSWER 2.04 (1.50) Comment: Due to the wording of the question in which " loads" was used instead of " pumps or types of pumps," three alter- ! nate answers are acceptable. An incorrect load or pump ! will result in less than full credit. I 1. HPSI 1. engineered safety pumps l t i 2. LPSI --OR-- and i 3. Containment Spray Pump 2. SDC Hx l l [+1.5 for either] --OR-- engineered safety pumps REFERENCE
- 1. Palisades: SH-PNAB, p. 40.
j 00026EA103 ...(KA'S) l I f I
PAGE 19
- 2. PLANT SYSTENS DESIGN. CONTROL. AND INSTRUNENTATION ANSWERS -- PALISADES -87/05/05-UPTON, J.
ANSWER 2.05 (2.50)
- a. 125 VDC Bus No. 2 --OR-- inverter No. 2 [+0.5 for either] l
- b. by (manual) transfer of the bypass regulator --OR--
by connecting Y01 AC instrument bust [+1.0 for either] l
- c. 1. 125 VDC battery
- 2. D/G-1 [+0.5 each, +1.0 maximum]
- 3. D/G-2 REFERENCE
- 1. Palisades: PNIC, Objectives 4.14 and 4.15, Figure PNIC-1.
ANSWER 2.06 (2.00) Comment: The answer to part "a." was crianged because E0P-1 lists 3 ftems while E0P-8 lists 5 items to be checked to verify natural circulation. ,
- a. Core delta T less than 50 degrees F CETs 25 degrees F subcooled Loop T-H within 15 degrees F of CETs Loop TC's constant or decreasing ,
4 CETs constant or decreasing [+0.5 each, +1.5 maximum]
- b.
ContainmentIsolationisinitiatedby]pushingeitherHIGH RADIATION INITIATE push button [+0.5 REFERENCE
- 1. Palisades: E0P 1.0, Standard Post Trip Actions, Section 3.0, Steps 000017EK10 11,13,(and14,pp.5and7of8.
... KA'S) Revision 0.
ANSWER 2.07 (3.00) l
- a. 1. Normal charging [+0.5]
- 2. HPSI[+0.5]
- 3. auxiliary spray [+0.5]
- b. 1. HPSI pumps [+0.5]
- 2. Containmentspraypumps[+0.5]
- 3. LPSI pumps [+0.5]
REFERENCE
- 1. Palisades: E0P 9.0, Functional Recovery, Attachment A.
001000K405 ...(KA'S)
.l
- 3. INSTRUMENTS AND CONTROLS PAGE 20 ANSWERS -- PALISADES -87/05/05-UPTON, J.
ANSWER 3.01 (1.50) upper electrical limit lower electrical limit Rod Withdrawal Prohibit (RWP) [+0.5each] REFERENCE
- 1. Palisades: Nuclear Operator Training, Advanced Systems Training, "RX Reg. and Rod Control (PNLB)", SH-PNLB, p. 28.
001000K408 ...(KA'S) ANSWER 3.02 (1.50)
- 1. turbine trip
- 2. bus IC undervoltage
- 3. S/U transformer undervoltage *
- 4. bus ID undervoltage l
[Any three (3) +0.5 each, +1.5 maximum] REFERENCE ,
- 1. . Palisades: Nuclear O
" Diesel GeneratorsPNHB)",(perator Training, Lesson Advanced Plan Outline, Systems Objectives Training, 4.13,
- p. 19.
ANSWER 3.03 (1.00) -
- a. The indicated pressures will be LESS THAN the actual pressure [+0.75]. PZR and S/G pressure instruments read ,
relative to containment pressure, therefore, as containment pressure increases, the indicated PZR and S/G pressures decrease [+0.75].
- b. The indicated levels will be GREATER THAN the actual levels
[+0.75]. The reference leg would be less dense resulting in less delta Ps and thus greater indicated levels [+0.75]. l REFERENCE
- 1. Palisades: SH-PNKE, p. 26 and 27.
191002K111 191002K108 ...(KA'S)
- 3. INSTRUMENTS AND CONTROLS PAGE 21 ANSWERS -- PALISADES -87/05/05-UPTON, J.
ANSWER 3.04 (2.00) Comment: The reference lists both RWP and AWP. The ARP lists only the RWP. The point count was adjusted as shown.
- a. RWP '+0.75' AWP ' 0.25l
;+1
- b. Power-Range Safety Channel > 15% ;+0.5; Wide-Range Log Channel ( 10**-4% + 0.5 REFERENCE
- 1. Palisades: Nuclear Operator Training, Advanced Systems Training, RX Monitorlag and Protection Systems (PNN), " Nuclear Instrumenta-tion (PNNA)",(Lesson 012000K603 ... KA'S) Plan Outline, Rev. 2, LP-PNNA, pp. 30 and 31.
ANSWER 3.05 (1.00) , (b.) [+1.0] REFERENCE
- 1. Palisades: Nuclear Operator Training, Advanc'ed Systems Training, RX Monitoring and Protection Systems (PNN), "liuclear Instrumenta-
' tion (PNNA)", Lesson Plan Outline, Rev. 2, SH-PNNA, Objective 4.7.
- ~ .
- 3. INSTRUMENTS AND CONTROLS PAGE 22
~ ' ANSWERS -- PALISADES -87/05/05-UPTON, J.
~
ANSWER 3.06 (3.00)
- a. Four channels per hot leg [+0.2] and two channels per cold leg [+0.2] feed the thermal margin / low-pressure trip point computers [+0.4]. One additional channel per leg [+0.2]
feeds the avera computer ;+0.4; ge. Onetemperature andfeeds cold leg channel differential the planttemperature computer _+0.1. .
- b. 1. Four channels (summers) [+0.2]; 2 of 4 required for trip
[+0.3].
- 2. Each PCS flow channel is fed by four differential pressure transmitters [+0.5]. Each of the four pressure transmitters provide a delta P across the S/Gs from hot leg to each suction leg [+0.5] .
REFERENCE
- 1. Palisades: Nuclear Operator Training, Advanced Systems Training, Primary Coolant System (PNJ), " Primary Coolant System* (PNJA),"
SH-PNJA, Section 4.16, pp. 28 through 30, Revision 2. 00200K606 ...(KA'S) ANSWER 3.07 (3.00)
- a. 1. loss of load trip [+0.5] is bypassed below'15% power [+0.25]
- 2. high rate of power change trip [+0.5 is not bypassed l between 10**-4% and 15% power [+0.25; .
- b. 1. low SG pressure [+0.5]; 500 or 510 psia [+0.25]
- 2. low SG water level [+0.5]; 25.7% (23-27 range) SG level l
[+0.25] REFERENCE
- 1. Palisades: SH-PNNB, p. 6, Revision 2.
012000A306 012000K603 ...(KA'S) 1
s* , a
- 2. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 23
. RADIOLOGICAL CONTROL ANSWERS -- PALISADES -87/05/05-UPTON,J. ANSWER 4.01 (1.50)
- 1. ensure fuel handling area charcoal filter is in operation (by placing HS-1894 to " Refuel" position)
- 2. trip the fuel handling area supply fan (V-7)
- 3. trip the fuel handling area fans V-69, 70A, 70B (by placing l HS-1893 to " Emergency Off")
[+0.5each] REFERENCE
- 1. Palisades: ONP 11.2, Fuel Handling Accident, Section 3.2,
- p. 1 of 2 Revision 0.
00036EK303 ...(KA'S) ANSWER 4.02 (1.50) Comment: Part "a." is deleted because the word acceptable is not sufficiently clear. Part "c." is deleted because the question does not state whether the condensate or the purification demineralizer is to be' considered. Part "e." was changed during the exam to reflect the changes in the procedure,
- a. False (below 7 mils is normal,14 alls requires trip) [+0.5]
- b. True (maximum allowed is 4.3" HgA) [+0.5]
- c. 562.8 degrees F [+0.5]
REFERENCE
- 1. Palisades: SOP 8, Main Turbine and Generating Systems, Sections 5.0.e and 5.0.v, pp. 2 through 4 of 61, Revision 16.
- 2. ' Palisades: 50P 11, Condensate System, Section 5.0.n, p. 4 of 27, Revision 4.
- 3. Palisades: GOP 6, Power Level Change, Sections 6.1.6 and 6.1.9, Revision 2.
045000A101 ...(KA'S)
l
- 4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 24 RADIOLOGICAL CONTROL ANSWERS -- PALISADES -87/05/05-UPTON,J.
ANSWER 4.03 (3.00)
- a. Push both [+0.25] reactor trip push buttons F+0.5]
Open CRD clutch power feeder breakers [+0.25'I , Set all CR0 clutch power toggle switches to CLUTCH OFF [+0.25] Commence emergency boration [+0.5]
- b. Attempt manual transfer [+0.5)
Close both MSIVs [+0.75] REFERENCE
- 1. Palisades: E0P 1.0, Standard Post Trip Actions, Section 3.0, Steps 4 and 5.d. pp. 2 and 3 of 8, Revision 0.
00029EK312 ...(KA'S) ANSWER 4.04 (1.50)
- a. increase
- b. decrease
- c. decrease
[+0.5each] . REFERENCE
- 1. Palisades: E0P 8.0, Loss of Forced Circulation Recovery, Steps 55 through 59, pp.11 and 12 of 15, Revison 0.
00074EK102 ...(KA'S) ANSWER 4.05 (.80) Start diesel generators [+0.8] REFERENCE
- 1. Palisades: ONP 12, Acts of Nature, Section 3.1, p.1 of 3, Revision 11.
062000SGA7 ...(KA'S)
- 2. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 25 RADIOLOGICAL CONTROL ANSWERS -- PALISADES -87/05/05-UPTON,J.
ANSWER 4.06 (1.50) Trip reactor [+0.5]. Isolate letdown (close CV-2001) :+0.5]. l Close PCP controlled bleedoff containment isolation va ves (CV-2083 and CV-2099) [+0.25' Open PCP controlled bleedoff reliefstopvalve(CV-2191)[+0.25]. REFERENCE
- 1. Palisades: ONP 24.5, Loss of Instrument AC Bus Y01, p. 1 of 3, Revision 12.
000057EK301 ...(KA'S) ANSWER 4.07 (1.20)
- 1. radios
- 2. Vital area door keys
- 3. locked valve keys ,
- 4. emergency flashlights
[+0.3each] REFERENCE .
- 1. Palisades: ONP 25.2, Alternate Safe Shutdown Procedure,
. 2 of 16, Revision 0, Section4.2,p(KA'S) 0000075GA6 ...
ANSWER 4.08 (1.50)
- 1. PCSlessthan300degreesF(250degreesFperTechnical Specifications) [40.75]
- 2. shutdowncoolingisolationvalves(No-3015andNO-3016)are open [+0.75]
REFERENCE
- 1. Palisades: 50P 1, PCS, Section 4.0.q and w, p. 4 of 45, Revision 16.
010000K403 ...(KA'S)
( .'
- 2. PROCEDURES - NORMAL. ABNORMAL ENERGENCY AND PAGE 26 RADIOLOGICAL CONTROL ANSWERS -- PALISADES -87/05/05-UPTON, J.
ANSWER 4.09 (2.50)
- a. If you had a completed NRC Form 4 lon file, the limit is 3000 mrem per quarter; so you can receive an additional 2700 mrem (full credit will be given if either the control or alert levels, 2500 and 2000 mres, respectively, are used) [+0.5]. l
- b. If you have a NRC Form 4, you can receive 3 rem per quarter l or 12 rem per year [+0.5]. However, you can only receive 5 (24-18) or 30 rem cumulative dose, so your dose limit would ,
be 8 rem this year [+0.5]. REFERENCE
- 1. Palisades: Administrative Procedure, " Radiation Dosimetry,"
Proc. No. 7.04, Revision 6, pp. 3 and 4. 00059EK102 ...(KA'S) a f
r - EQUATION FORMULA AND PARAMETER SHEET Where og = m2 - (density)g(velocity)g(area)1 = (density)2(velocity)2(area)2 KE=y"2 PE = mgh PEg + KEg+P Y 1 1 = PE2+KE2 +P Y22 where V = specific volume P = pressure 9 * "Cp(Tout-Tin) Q=UA(T,y,-Tstm) Q=m(h-h) l 2 P=P,10(SUR)(t) p,p t/T T= " = I# - #I SUR = Y delta K = (K,ff-1) CRg(1-K,ffg) = CR 2 (I"Keff2) CR=S/(1-K,ff) (1-K,ffg) (1-X,ff)x100% SDM = M = (1.Keff2I K eff A'II = 0.08 sec.3 decay constant = In (2) = 0.693 A 1
- Ao'-(decay constant)x(t) g1 /2 t 1/2 Water Parameters Miscellaneous Conversions 1 gallon = 8.345 lbs 1 Curie = 3.7 x 10 10 dps '
I gallon = 3.78 liters 1 kg = 2.21 lbs 1 ft3 = 7.48 gallons 1 hp = 2.54 x 10 3 8tu/hr 3 6 Density =62.4lbg/ft 1 MW = 3.41 x 10 Btu /hr Density = 1 gm/cm 1 Stu = 778 ft-lbf Heat of Vaporization = 970 Btu /lbm DegreesF=(1.8xDegreesC)+32 Heat of Fusion = 144 Btu /lbm 1 inch = 2.54 centimeters 1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 ft-lbm/lbf-sec 2 1 ft H 2O = 0.4335 lbf/in.2
r - PALISADES TECHNICAL DATA BOOK FIGURE 41, REV 2 CYCLE 7 HZP RECIPROCAL BORON WORTH VS BURNUP I10 - ' I : l . . t i i . . .
, . s . .
I l f 105- . s
. l N I -
Q. l ... g e e j .. .. ,
,s ___ . _ . . . _ . .
,oo .
@, N . _--,
f h m O . . . y g ,3 ! ___ . . . . . . . . . . _ . . . _ _ . . _ _ _ z I ' 8 l, a . . . . . . . . . . . . . . . _ 3 ' O g . 6 a u A. a go . - .. . . . _ . - . . . . M h
.........t .
REVEWED BY as 3d A A MYet - - - - - - - - - - - - - - REACTOR M N ER k . " L < NW v'k ~~ REACTOR ENCihllRiNG s
$UPERNTINOCNT . i I
I -- do , . . . - i .. . . 4 6 8 10 12 0 2 DURNUP - Gwo/WTd CONSUWERS 70*Dt C0WPANY REAciCR ENONCERee0/ PAuSAots GCPeekeeo alAct0R thoNttR GROUP Q2/27/86
r
'. PALISADES TECHNICAL DATA BOOK FIGURE 5.1, REV 2 CYCLE 7 INTEGRAL ROD WORTH i 2- -
! I . ,
a . s . . . . 1 . !
$ ,; .' l i.0- .
I I i l
. i
! ! REVEWED BY
- =
_ M. _1)al],__ -W4r, 06- - RocTOR EndEER g , i . f 4 ,0 7 M Han s b& __ --
* * ~
i l RDCf0R ENC 4NEERIM *
~ 4 s SUPERMTENDENT r: k i z . o.- . , ,
E 0*~ v
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o -
" l %5 i !
a i - 5 1, l a i g . . d. .. E 0.4-I hs \:.--.. -- . . - -- - l
......... m. . .....
. . . .l . t
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... ...... . . . . . ... . . 4 0 2 .
i i i t e l I l l
. I.
og. . . GR4 0 20 40 l 60 80 100 120 140 GR3 80 100 120 632 CONSUWERS POWER COWPAW ROD POSeTION - INCHES WiTHORAWN RUCTOR ENQNEGNC/ PAUSADES CCP a e it e e d RUCTOR ENONEER CROUP 02/27/86
. 4me..
g -e PALISADES TECHNICAL DATA BOOK FIGURE 6.1, REV 2 CYCLE 7 g,.-{d,h,.SREDICTEDBORONCONCENTRATION .
' y /.,,. BURNUP (HFP. ARO. EQUIL. XENON) 1300 . . . . . . . . ............ . . ~ . . . . .... .... . . . . . . . . . .. ... .. . . ..
. . . . . . 4. .. ... . . . . . . ...... ......... ....... . ..... . . . .
. . . . . r. , . . . . . . . . . . . . ... .. . . . . . . ... . . . . . . . . . ...... . . . .
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. . . . . . . . . , . . . ..... .. ... . ....... .. ...o....._.......... . . ..
1100-- - ---- - -
.y..
. .. 4........
. . . . . . . . . . ...... ~. ~..L~.r - -
.... 4.... .. .........
^
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1000 .-..... .... .... . . . . . . . . . . .....s ....... . .. .
- 3. . . . ..
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m z 800 ..... ..... ..... . .. . .. ... . ....... _ .- i ..
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700 - -.
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o ....... .. ...... ... .. . ... ..
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300 ~-----'-~~~~,--!!' i ... .... . . . . . . . . . . . . .
. . A h p L h ......... ..
E < L. .. . ... .. . ... ... . . . . . . [ , 200~,j g l , [ .g L ,3gp, ,, ,,,,,, ,, ,, ,,, , . , , , , , , , , ,
. ...... . _ nr.Acron EiMER .. ...... . . ... .
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~r- - - KM L c ]vU1 31/, .. . . . ........ . .. .. .-. . .... -4 . . t ..*. -
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1 i ii1 i t i . . i , a i i e i i i i i , l 4 5 6 7 8 9 10 11 12
-l 0 1 2 3 EXPOSURE GWO/WTU coeistAms powet coepAm ROCTom ENe M mpeV PAUS2 O GCPeckeed nOcton ENews enouP 02/27/#8 CONTROLLm mov
(
risura 8.1, R a 0
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s . . . _ , 9 ( . h a ,-- 3 ,t=.b- . i- 7.r w .gf 8, 9 . ' v -- ..,s w :w X. , " .- *
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tf-"y_g< {a L. h . ' u.s v W i U. S. NUCLEAR REGULATORY COMMIS$10N 1 SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION FACILITY: PALISADES REACTOR TYPE: PWR-CE DATE ADMINISTERED: 87/05/05 EXAMINER: UPTON/DEFFER0!!iG/ GRUEL CANDIDATE: ANSWER KEY -F is!!* , , ( INSTRUCTIONS TO CAN010 ATE: Read the attached instruction page carefully. This examination replaces the current cycle facility administered requalification examination. Retraining requirements for failure of this examination are the same as for failure of a requalification examination prepared 'and administered by your training staff. Points for each question are indicated in parentheses'after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up four (4) hours af ter the examination starts. t 0F CATEGORY % OF CAfl010 ATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 15.00 _25.00 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERHODYNAMICS 15.00 25.00 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 15.00 .25.00 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL
.15.00 _25.00 8. ADM!ft!STRATIVE PROCEDURES, CONDITIONS, AND LlHITATIONS 60.00 % Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.
Candidate's Signature
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
- 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination l
room to avoid even the appearance or possibility of cheating. l
- 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-nand corner of the first page of each section of the answer sheet, i
- 8. Consecutively number each answer sheet, write "End of Category ,,,,," as appropriate, start each category on a new page, write only on on,e side j of the paper, and write "Last Page" on the last answer sneet.
- 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 parentheses after the question and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND 00 NOT LEAVE ANY ANSWER BLANX.
- 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 part of the answer.
- b. Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions,
- d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
O
- 5. THEORY OF NUCLEAR POWER PLANT OPC'.Tf 0N. FLUIDS, AND PAGE 2
, THERMODYNAMIC 5 QUESTION 5.01 (1.50)
Plutonium-239 is produced in the core of the Palisades nuclear reactor.
- a. WHAT impact does the buildup of Pu-239 have on the effective beta of the core? (0.5)
- b. WHAT impact does the buildup of Pu-239 have on the fuel temperature coefficient (FTC) of the core? WHY7 (1.0)
QUESTION 5.02 (1.50) After a spent-fuel assembly had been in the spent-fuel pool for one year, it was surveyed and it was determined that the dose rate at one foot away was 50,000 R/hr. One year later (after two years inthepool),thedoserateatthesamelocationwasfoundtobe 5,000 R/hr. CALCULATE the effective half life, t-1/2, of the fuel assembly. (1.5) QUESTION 5.03 (1.00) . Following a reactor trip from sustained operation at 100% power, the power (neutron flux) level drops immediately to about (1.0) (a.)6%becauseabout94%oftheneutronsarepromptneutrons. (b.)6%becauseabout0.4to0.5%oftheneutronsaredelayed neutrons. (c.) 99.5 to 99.6% because about 0.4 to 0.5% of the neutrons are delayed neutrons. (d.)99.5to99.6%because99.5to99.6%oftheneutronsare
- prompt neutrons.
QUESTION 5.04 (1.50) Ecuilibrium xenon reactivity does not change as much for a 10% power change at a high power level as it does for a power change of 10% at a low power level. EXPLAIN. (1.5) (*"" CATEGORY 05 CONTINUE 0 ON NEXT PAGE "*")
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS. AND PAGE 3
,' THERMODYNAMICS QUESTION 5.05 (2.00)
The Palisades power plant is operating at 100% of full power with all rods out (ARO). The age of the core is 8000 MWD /MTU. The boron concentration is equal to the " predicted boron concentration" for this reactor condition. It is necessary to insert the CEAs in the manual sequential (MS) mode until Group 4 is at 40 inches. A change in the boron concentration will be used to maintain 100% power. Use the Palisades Technical Data Book as necessary. DETERMINE the number of PMW gallons required. Clearly OUTLINE your step:1. (2.0) QUESTION 5.06 (1.00) During accident conditions, the following information is available: PZR pressure = 1000 psia Hottest CET temperature = 500 degrees F
- a. WHAT is the temperature subcooled margin (SCM)? (0.5)
- b. WHAT is the pressure SCM7 (0.5)
QUESTION 5.07 (3.00) For each of the following, STATE the condition of the noted fluid.
- a. A leak exists in the steam generator steam space while the plant is at 100% power. STATE the temperature AND ths degree of subcooling/ quality / degree of superheat of the exit fluid. (1.5) .
- b. The regenerative heat exchanger inlet and outlet temperatures (letdown) are 538 degrees F and 240 degrees F, respectively.
The inlet temperature (charging) is 110 degrees F. STATE the outlet temperature (charging) of the regenerative heat exchanger. (1.5) l ("*" CATEGORY 05 CONTINUED ON NEXT PAGE "***)
i --
. 5. THEORYOFNUCLEARPOWERFLANTOPdRATION, FLUIDS,AND PAGE 4 THERMODYNAMICS ,
QUESTION 5.08 (3.50) Included with this examination is the Characteristic Curve figure-for the Primary Coolant Pump (PCP).
- a. DEFINE the term " pump runout." WHERE on the figure does pump runout. occur? (1.0)
- b. The figure shows the recommended NPSH.
- 1. If the water entered the PCP at 360 degrees F, l WHAT would be the minimum PCP suction pressure in
, psia required to meet the NPSH recommendation? (1.0)
- 2. According to SOP-1, " Primary Coolant System,"
can the PCP be placed in service with this PCS pressure? (0.5)
- c. The PCP is operating such that its parameters fall on the indicated head curve at the point corresponding to a water flowrate of 70,000 gpm. If the temperature of the water was raised, HOW would the operating point shift? EXPLAIN your answer. (1.0)
I 4
/
u (*****ENDOFCATEGORY05*****) >
.- * ^? .
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PALISADES PLANT E d PRIM ARY COOLANT PUMP CHAR ACTERISTIC CURVE 5 4 -
=
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- Head,
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-70 ft- > :- .' -
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- : -60
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Rebmmended M NPSH; ~yficiency, i,,s, Min. NP .: it .
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- a. .- -'
hp j 2= 100, - 10,000
-20 u
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- $$ O'
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0 20 40..; . 80 0 0 j 's n 8O Id0 120 L 1000 Gallons per Minute -
- 2 - -
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- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 5 QUESTION C 01 (1.00)
When operating in the Manual IndividQal (MI) mode, LIST the three (3) control rod interlocks which are NOT bypassed? (1.0) QUESTION 6.02 (1.50) With the Emergency Diesel Generator No. I controls in automatic, WHAT three (3) conditions would generate signals to automatically start the diesel? (1.5) (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
. 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 6 QUESTION 6.03 (2.00)
For each of the following conditions (a., b. and c.), LIST those actuation signals (1. through 6.) that should have triggered. Consider each condition separately. Each condition may have for its answer no, one, or several actuation signal (s). (2.0)
- 1. SIAS !
2 2. CSAS
- 3. Auto Containment Isolation
- 4. MSIS vs. MSIV l
- 5. RAS
, 6. AFAS
- a. A steam-line break has occurred in containment and containment pressure = 7 psig containment radiation = background S/G levels = 60% and 40%
PZR pressure = 1800 psia SIRWT level = 80% S/G pressure = 680 and 580 psia
- b. A small-break LOCA has occurred and containment pressure = 2 psig containment radiation = 6 R/hr S/G 1evels = 65% and 65%
PZR pressure = 1800 psia ~ SIRWT level = 50% S/G pressure = 700 and 700 psia
- c. A feedwater problem has occurred and containment pressure = 1 psig containment radiation = backgrcund S/G levels = 60% and 31% ~
PZR pressure = 2200 psia SIRWT level = 80% S/G pressure = 720 and 650 psia QUESTION 6.04 (3.00) STATE six (6) of the seven (7) automatic responses which occur in the CVCS system when a Safety Injection Signal is generated with standby power available. (3.0) (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
l 1
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 7 QUESTION 6.05 (2.00) i
- a. During abnormally-high containment pressure conditions, WILL the indicated Steam Generator pressure and Pressurizer pressure be EQUAL TO, GREATER THAN or LESS THAN actuali PROVIDE a rationale for your answer. (1.0)
- b. During abnomally-high containment temperature conditions, i WILL the indicated Steam Generator level and Pressurizer level be EQUAL TO, GREATER THAN, or LESS THAN actual?
PROVIDE a rationale for your answer. (1.0) QUESTION 6.06 (1.50) LIST the three (3) loads that can be directly cooled from service water during a loss of CCW. (1.5) QUESTION 6.07 (2.00)
- a. If the "HIGH POWER RATE CHANNEL PRE-TRIP" alarm was activated by the RPS, WHAT two (2) other control signals / trips should have been activated? (1.0)
- b. If the " RATE TRIP CHANNEL BYPASSED" alarm was activated by the RPS, LIST the two (2) signals / trips, either of which should have been the cause.
(1.0) QUESTION 6.08 (2.00)
- a. DESCRIBE the interlocks between the reactor-side tilt machine and the refueling machine. (1.0),
- b. DESCRIBE WHY both main transfer winches are not activated to manually transfer fuel and control rods between the containment and the fuel building when the transfer carriage is out of service. (1.0)
(***** END OF CATEGORY 06 *****)
. 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 8 RADIOLOGICAL CONTROL QUESTION 7.01 (3.50)
The plant is operating at 100% power when a reactor trip occurs: -
- a. WHAT actions should be taken in the event all full length control rods are not fully inserted even after both reactor trip pushbuttons are pushed? (1.0)
- b. If forced circulation cannot be maintained, HOW would the condition of natural circulation be verified? (1.5)
- c. If containment isolation and containment spray should have occurred and did not, HOW would they be initiated /
i established? (1.0) QUESTION 7.02 (1.50) STATE whether the start of each of the following operations would tend to INCREASE or DECREASE void size (if one exists in the reactor vessel):
- a. Pressurizer spray operation 0.5)
- b. Pressurizer heater operation 0.5)
- c. Reactor vessel head venting 0.5)
QUESTION 7.03 (1.00) During operation at 35% power, WHICH one of the following situations is NOT allowable beyond a two-hour period? (1.0)
- a. loss of one primary coolant pump
- b. an inoperable secondary system safety valve
- c. a closed safety injection tank isolation valve .
, d. two auxiliary feedwater pumps in manual for testing i
- QUESTION 7.04 (2.50) .
If power was lost to instrument AC bus Y01, WHAT immediate actions should be taken? Why? (2.5) i 't (***** CATEGORY 07CONTINUEDONNEXTPAGE*****) i
, _, -, . , . , . - _ , . - , - _ _ - _ , , - _ _ . _ _ . , _ _ . . . , , _ - , . , ~ . , , , _ , _ _ _ _ _ . _ , , , , _ , . , _ _ _ , _ .-m__m--_____.m._-.--.
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 9 RADIOLOGICAL CONTROL QUESTION 7.05 (3.00)
The Functional Recovery Procedure, E0P 9, Attachment A, provides guidance for reactivity control. If CRDs cannot be inserted or driven in, two methods of boron addition are specified.
- a. Regarding boron addition using the CVCS, LIST the three (3) flowpaths that are available to supply borated water from the charging pumps to the PCS. (1.0)
- b. LIST the three (3) flowpaths (pumps and PCS injection sites) for supplying borated water to the PCS if the charging pumps
. are unavailable. (2.0) QUESTION 7.06 (2.00) If a fuel handling accident occurred in the auxiliary building fuel handling area, WHAT four (4) immediate actions should be taken? (2.0) QUESTION 7.07 (1.50) Plant cooldown is proceeding from Hot Shutdown to Cold Shutdown,
- a. Low Temperature Over-Pressure Protection (LTOP) must be operable under either of WHAT two (2) conditions? (1.0)
- b. WHAT constraints are placed on the HPSI pumps when LTOP is in service? (0.5)
(*****ENDOFCATEGORY07*****)
', 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 10
)
QUESTION 8.01 (2.00) Included with this examination is a copy of Figure 2-3 from the Palisades Tech-Specs, Section 2.1, " Safety Limits-Reactor Core." -
- a. If the Primary Coolant System (PCS) pressure was 2100 psia, INDICATE on Figure 2-3 the region of acceptable operation. (0.5)
- b. WHAT is the objective or basis for this safety limit? (0.75)
- c. Section 3.1.1 of the Palisades Tech-Specs specifies that the nuclear-reactor vessel flowrate, with four (4) primary coolant pumps operating, shall be 126.9 x 10**6 lbm/h or greater, when corrected to 532 degrees F. If the vessel flowrate dropped below 126.9 x 10**6 lbm/h, EXPLAIN WHAT change should be made to the temperature limits of Figure 2-3. (0.75)
QUESTION 8.02 (2.00) Consider that at Palisades a condition existed that exceeded the Limiting Conditions for Operation as specified in the Palisades Tech. Specs. This condition was entered at 0930 (9:30 am) hours. The associated ACTION requirements called for one requirement to be completed in 12 hours and there were no special exceptions associated with noncompliance. It is known that the ACTION requirement would take 2 hours of work to complete. - (2.0)
- a. If, at 1530 (3:30 pm), work on the ACTION requirement had not been started and corrective measures allowed work on the ACTION requirement to start, WHAT additional actions (if any) would be required by Tech. Specs?
- b. If, at 2000 (8:00 pm), work on the ACTION requirement .
had not been started and corrective measures allowed work on the ACTION requirement to start, WHAT additional actions (if any) would be required by Tech. Specs.?
- c. If, at 2130 (9:30 pm), work on the ACTION requirement had not been started and corrective measures allowed work on the ACTION requirement to start, WHAT additional actions (if any) would be required by Tech. Specs.?
(***** CATEGORY 08 CONTINUED ON NEXT FAGE *****)
PALISADES QUESTION 8.0i. May 5, 1987 r'
. 1 l
ses , ! w s l W$k j '~ s&M i - h\ I - w\ N s s ' N
$15 g 30 100 110 REACTOR POWER-% OF 2530Mw
( Reactor Core Safety Limits Palisades figure 4 Pump Operaties Technical Specifications p3 has a 31 2 13 November 1,1977
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 11 QUESTION 8.03 (2.00) l l
Consider that at the Palisades nuclear power plant, a heatup is - progressing at 40 degrees F/h. The Primary Coolant System (PCS)
-pressure is 1400 psig.
- a. SPECIFY the PCS temperature range which is required by l Palisades Tech-Specs, Section 3.1.2, "Heatup and Cooldown
! Rates." (Section 3.1.2 of Tech-Specs refers to Figure 3-1 whichisprovided.) (0.5)
- b. Figure 3-1 refers to a fluence of 1.8 x 10**19 nyt. WHAT is
" fluence" in this situation and WHAT changes must be made before the stated fluence limit is reached? (1.5)
QUESTION 8.04 (2.50) According to the Palisades Tech-Specs, Section 3.1.4, " Maximum Primary Plant Coolant Radioactivity," the specific activity of the primary coolant shall be limited to
- 1. ( 1.0 micro Ci/gm dose equivalent I-131 AND
- 2. ( 100/E micro C1/gm.
- a. WHAT is the meaning of " micro Ci/gm dose equivalent I-131"? (0.5)
- b. Limit "2." is concerned with WHAT radionuclides? WHAT is "E"? (1.0)
- c. If limit "2." was exceeded, Tech-Specs requires that the plant should be shutdown with T-avg less than 500 degrees F within 6 hours. WHAT is the significance of 500 degrees F being used here as part of the specifications? (1.0)
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
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. .ui to
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$ Palisades Plant Pressure /TemperaturgLimits for Heatur g 1 . for Fluence to 1.8 x 10 avt F1- :e 3-1
\ \ )
i i
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS, AND LIMITATIONS PAGE 12 QUESTION 8.05 (2.50)
You have received 1000 mrem of whole-body occupational dose in the first quarter of this year and 300 mrem in the second - quarter of this year. Your cumulative dose to date is 22 rem. You are a healthy male who turned 24 on the first of the year.
- a. According to the Palisades administrative dose limits, HOW much additional whole-body occupational dose can you receive this second calendar quarter? SPECIFY any assumptions pertinent to your answer. (1.0)
- b. Based on the whole-body lifetime dose limits as set by 10CFR20, WHAT is the whole-body dose limit that you can receive this year? SPECIFY any assumptions pertinent to your answer. (1.5)
QUESTION 8.06 (2.00) According to Palisades Technical Specification 3.2.2(c), the combined content of the two Concentrated Boric Acid Storage Tanks (CBASTs) should be at least 118 inches of 6-1/4 to 10% by weight boric acid solution. Standing Order No. 28, Rev. 1, March 5, 1985 has placed a different operational limit on the CBASTs.
- a. WHAT is the different limit to which CPCo has committed itself (to the NRC)? Under WHAT power plant conditions does the different limit apply? (1.0)
- b. WHAT is the concern that has generated the modified limit? (1.0)
QUESTION 8.07 (2.00) , According to Standing Order No. 54 concerning the Technical Specification on Fire Rated and Fire Protection Assemblies, an " inoperable" cable tray enclosure requires one of two actions. LIST the two actions. (2.0) l l l I (***** END OF CATEGORY 08 *****) (************* END OF EXAMINATION *************)
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 13 THERMODYNAMICS ANSWERS -- PALISADES -87/05/05-UPTON, J.
/;. }Ax. /* "T" !"' "'
%.- 0
-t.
y m - p
.1 i 5. iM L.M h ;:m ii . ~
- # ly~
i \l!; ANSWER 5.01 (1.50)
- a. Pu-239 reduces beta-eff (because the beta for thermal fissioning of Pu-239 is about 0.002) [+0.5]
- b. Because Pu-239 has a neutron-absorption-fission peak just above thermal energy, it increases the magnitude (makes negative) of the FTC. --OR-- Because of Pu-240 buildup in the fuel due to n-absorption in Pu-239 and because Pu-240 also has an absorption peak, it increases the magnitude of the FTC. [+1.0]
REFERENCE
- 1. Palisades: Nuclear Operator Training, " Reactor Theory (PQ),"
Theoretical Concepts (PQAO), Lesson Plan Outline Objective 4.45,
- p. 32.
- 2. Palisades: Nuclear Operator Training, " Reactor Theory (PQ),"
Theoretical Concepts (PQAO), SH-PQAO, p. 36. ANSWER 5.02 (1.50) A = A(o) e**-0.693t/t-1/2 [+0.5] In A(o)/A = 0.693t/t-1/2 . (0.693)(1 yr) t-1/2 = --------------- in 50,000/5,000
= (0.693)/2.30
= 0.30 years [+1.0; full credit for 0.2 to 0.4 years]
REFERENCE
- 1. Palisades: Nuclear Operator Training, " Reactor Theory (PQ),"
Theoretical Concepts (PQA0), Objective 4.2, pp. 8 through 11. ANSWER 5.03 (1.00) (b.) [+1.0] REFERENCE
- 1. Palisades: Nuclear Operator Training, " Reactor Theory (PQ),"
Theoretical Concepts (PQAO), Objectives 4.12 and 4.13, pp. 13 through 20. 4
, _ _ . - _ - _ , ...m._. .-,_.__,_.___.,,,.m._,..__..._-__.. -,_ ,.,,...,,,-. -.- _ ,__ ___.-,__ ____,_.--_m
. 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 14 THERMODYNAMICS ANSWERS -- PALISADES -87/05/05-UPTON,J.
ANSWER 5.04 (1.50) The production of Xe-135 is dependent on the flux or power level because the production is from fission fragments [+0.5]. The removal of Xe-135 is by burnout and by decay. The burnout of Xe-135 is power dependent [+0.5]. As the power level increases, the burnout term becomes more dominant in Xe-135 removal and the condition is approached where a power-level change would equally affect production and removal equilibrium Xe-concentration [+0.5] producing no change in the REFERENCE
- 1. Palisades: Nuclear Operator Training, " Reactor Theory (PQ),"
Control Poisons and Fission Poison Products (PQBO), Objective 4.27, 4.28, 4.29, 4.30 and 4.35, pp. 19 through 23.
- 2. Palisades: Technical Data Book, " Equilibrium Xenon Versus Reactor Power," Figure 2.1, Rev. 2.
ANSWER 5.05 (2.00) Using Figure 6.1, the initial boron concentration is 320 ppm. Using Figura 5.1, the delta-rho is 0.54%. . Using Figure 4.1, the reciprocal boron worth is 97 ppm /% delta-rho. Therefore, delta-ppm = (97)(0.54) = 52.4 ppm. Using Figure 8.1, gallons / delta-ppm = 180. Therefore, answer = (180)(52.4) = 9430 gal. 4 (Using equation on Figure 8.2, answer = 10,300 gal) 5.77 X 10 in 1/f l [+0.33 for each step] REFERENCE
- 1. Palisades: Technical Data Book, Dated 5/12/84, Figures 4.1, 5.1, 6.1, 8.1, and 8.2.
- 2. Palisades: Nuclear Operator Training, " Reactor Theory (PQ),"
Control Poisons and Fission Product Poisons (PQBO), Objectives 4.17, 4.18, 4.20, and 4.22, pp. 12 through 15.
\
~ 50 THE0RY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 15 THERMODYNAMICS ANSWERS -- PALISADES -87/05/05-UPTON,J. ANSWER 5.06 (1.00)
- a. 544.6 - 500 = 44.6 degrees F [+0.5]
- b. 1000 - 680.9 = 319.1 psia [+0.5]
REFERENCE
- 1. Palisades: Nuclear Operator Training, " Applied Reactor Heat Transfer and Fluid Flow (PR)," SH-PRAO, pp. 35 and 36.
193008K115 ...(KA'S) ANSWER 5.07 (3.00) Comment: The starting pressure for this calculation can range from 570 to 790 psia. p = 690 psia (saturated) h = 1202 BTU /lba (saturated steam) [+0.5] delta h = 0- [+0.5] p-exit = 15 psi T-exit is approximately 320 degrees F [+0.5] 108 degrees F superheat [+0.5]
- b. EdeltaT =$deltaT [+0.5]
OR deltaTcharging=deltaTletdown(mletdown/Echarging)
$ letdown =40gpm [+0.25]
$ charging =44gpm [+0.25]
delta T charging = (538 - 240) (40/44) I
= 298 x 0.91
= 271 degrees F T = 110 + 271 = 381 degrees F l REFERENCE
- 1. Palisades: Nuclear Operator Training, " Applied Reactor Heat Transfer and Fluid Flow (PR)," SH-PRAO, Objective 4.35, 4.37, and 4.38, pp. 45 through 49 and 68.
- 5. THEORY OF NUCLFAR POWER PLANT OPERATION. FLUIDS. AND PAGE 16 THERMODYNAMICS
, ANSWERS -- PALISADES -87/05/05-UPTON, J.
ANSWER 5.08 (3.50)
- a. " Pump runout" is the term used to describe the operation i
ofacentrifugalpump(whenitispumpingatminimumhead and maximum flowrate and low NPSH and max amps) [+0.5].
- It is indicated by the far-right in the figure [+0.5].
. b. 1. NPSH = 100 ft H 0 2
= 43.4 psia sat. press. (360 degrees F) = 153.0 psia
- hence, min. press. = 196.4 psia [+1.0]
- 2. no, SOP-1 requires 250 psia [+0.5]
- c. When the PCS temperature is increased, the PCS water density is decreased. The decreased density results in the frictional losses being decreased. This means
, that the operating point would move to the right on the curve providing less head and more flowrate. [+1.0] REFERENCE ! 1. Palisades: Nuclear Operator Training, " Applied Reactor Heat Transfer and Fluid Flow (PR)," SH-PRAO, pp. 123 i through 133 and Figure 10. 191000K112 ...(KA'S) 1 I i i
. 6. PLANT SYSTEMS DESIGN. CONTROL, AND INSTRUMENTATION PAGE 17 ANSWERS -- PALISADES -87/05/05-UPTON, J.
ANSWER 6.01 (1.00) upper electrical limit lower electrical limit Rod Withdrawal Prohibit (RWP) [+0.33each] REFERENCE
- 1. Palisades: Nuclear Operator Training, Advanced Systems Training, "RX Reg. and Rod Control (PNLB)", SH-PNLB, p. 28.
001000K408 ...(KA'S) ANSWER 6.02 (1.50) 1. turbine trip
- 2. bus IC undervoltage
- 3. S/U transformer undervoltage
- 4. bus ID undervoltage l ,
[Any three (3) +0.5 each, +1.5 maximum] REFERENCE
- 1. Palisades: Nuclear Operator Training, Advanced Systems Training,
" Diesel Generators (PNHB)", Lesson Plan Outline, Objectives 4.13,
- p. 19.
1 l i
. 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 18
. ANSWERS -- PALISADES -87/05/05-UPTON, J. ANSWER 6.03 (2.00) Comment: MSIS - such an actuation signal does not exist at Palisades. MSIVs do, of course, exist. Hence, the 2 alternate answers to "a.". The answer to "b." is changed to reflect the new setpoint of 10 R.
- a. SIAS, CSAS, Auto Containment Isolation, MSIS (1., 2., 3., 4.)
--OR-- (1., 2., 3.) [+1.00] l
- b. None [+0. 5] l
- c. None [+0.5]
REFERENCE
- 1. Palisades: Nuclear Operator Training, Advanced Systems Training, Engineered Safeguards Systems (PNM), " Safety Injection (PNMA)",
LP-PNMA, pp.16 and 17; " Containment Spray (PNMB)", SH-PNMB,
- p. 23; " Containment Building (PNMF)", SH-PNMF, p.14.
- 2. Palisades: Nuclear Operator Training, Advanced Systems Training, Secondary Plant Steam Systems (PND), " Main Steam (PNDB)",
SH-PNDB, p. 11 " Auxiliary Feedwater (PNEB)", LP-PNEB, p. 8. 012000A301 ...(KA'S) ANSWER 6.04 (3.00)
- 1. Boric Acid Pumps start (P56A and B).
- 2. Coolant Charging Pumps start (P55A, B and C).
- 3. Boric Acid Gravity Feed Valves open (MOV-2169 and -2170). .
- 4. Boric Acid Pumped Feed Stop Valve (M0V-2140) opens.
- 5. VCT Outlet Valve (MOV-2087) closes.
- 6. Concentrated Boric Acid Tanks' Recirc Valves (CV-2130 and
-2136) close.
- 7. Boric Acid Makeup Stop Valve (CV-2155) closes.
[+0.5each,+3.0 maximum] REFERENCE
- 1. Palisades: Nuclear Operator Training, Advanced Systems Training, PCS Auxiliary and Containment (PNK), "CVCS (PNKA)",
Objective 4.24, SH-PNKA, p. 48.
l
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 19 ANSWERS -- PALISADES -87/05/05-UPTON, J.
ANSWER 6.05 (2.00)
~
- a. The indicated pressures will be LESS THAN the actual pressure [+0.5]. PZR and S/G pressure instruments read relative to containment pressure, therefo're, as containment pressure increases, the indicated PZR and S/G pressures decrease [+0.5].
- b. The indicated levels will be GREATER THAN the actual levels
[+0.5]. The reference leg would be less dense resulting in less delta Ps and thus greater indicated levels [+0.5]. l .. REFERENCE
- 1. Palisades: SH-PNKE, p. 26 and 27.
191002K111 191002K108 ...(KA'S) ANSWER 6.06 (1.50) Comment: Due to the wording of the question in which " loads" was used instead of " pumps or types of pumps," three alter-nate answers are acceptable. An incorrect load or pump j will result in less than full credit.
- 1. HPSI pumps 1. engineered safety pumps and
- 2. LPSI pumps --OR-- 2. SDC Hx
- 3. Containment spray pumps .
--OR-- engineered safety pumps l
[+1.5 for either] REFERENCE
- 1. Palisades: SH-PNAB, 00026EA103 ...(KA'S) p. 40.
ANSWER 6.07 (2.00) Comment: The reference lists both RWP and AWP. The ARP lists only the RWP. The point count was adjusted as shown.
- a. RWP '+0.75' AWP l+0.25l
- b. Power-Range Safety Channel > 15% ;+0.5; Wide-Range Log Channel ( 10**-4% + 0.5 REFERENCE
, 1. Palisades: Nuclear Operator Training, Advanced Systems Training, RX Monitoring and Protection Systems (PNN), " Nuclear Instrumenta-tion (PNNA)",(Le:: 012000K603 ... KA'S) son Plan Outline, Rev. 2, LP-PNNA, pp. 30 and 31.
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 20 ANSWERS -- PALISADES -87/05/05-UPTON, J.
1 ANSWER 6.08 (2.00)
- a. 1. The reactor-side tilt machine cannot be rotated from vertical to horizontal if the refueling machine is in the tilt machine zone. [+0.5]
- 2. The refueling machine cannot enter tilt zone unless tilt machine is vertical. [+0.5]
- b. One winch is driven to tow the carriage while the other winch freewheels to pay out cable. [+1.0]
REFERENCE
- 1. Palisades: PNEQ.
034000K402 ...(KA'S)
, , - . - - - , - - , , , , _ . - . .n .. a,- . , , . - - , . . - - - . , , - - - _ .
- 7. PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE 21 RADIOLOGICAL CONTROL ANSWERS -- PALISADES -87/05/05-UPTON, J.
ANSWER 7.01 (3.50) Comment: The answer to part "b." was changed because E0P-1 lists 3 items while E0P-8 lists 5 items to be checked to verify natural circulation.
- a. Open CRD clutch power feeder breakers [+0.25]
Set all CRD clutch power toggle switches to CLUTCH OFF [+0.25] Commence emergency boration [+0.5]
- b. Core delta T less than 50 degrees F CETs 25 degrees F subcooled Loop T-H within 15 degrees F of CETs Loop TC's constant or decreasing CETs constant or decreasing [+0.5 'each, +1.5 maximum]
l c. Containment Isolation is initiated by pushing either HIGH RADIATION INITIATE push button [+0.5] Containment spray is initiated by opening containment spray valves [+0.25] and starting containment spray pumps [+0.25] REFERENCE
- 1. Palisades: E0P 1.0, Standard Post Trip Actions, Section 3.0, Steps 4, 11, 13, and 14, pp. 2, 5, and 7 of 8, Revision 0.
00029EK312 ...(KA'S) ANSWER 7.02 (1.50)
- a. increase
- b. decrease
- c. decrease
[+0.5each] REFERENCE
- 1. Palisades: E0P 8.0, Loss of Forced Circulation Recovery, Steps 55 through 59, pp. 11 and 12 of 15, Revison 0.
00074EK102 ...(KA'S) ANSWER 7.03 (1.00) (a.) --OR-- (c.) [+1.0] l REFERENCE
- 1. Palisades: Technical Specifications, Sections 3.1.1.c (Table 2.3.1), 3.1.7.c, 3.3.2.a, and 3.5.2.b.
! 002020SGA8 ...(KA'S)
__- . - - . _ _ ~ - -_. . - _ ~ - _ - - . - - .. - ._-.- - - _ - . i
~
l
> - I
. 7. PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE 22 I RADIOLOGICAL CONTROL ANSWERS -- PALISADES - 87/05/05-UPTON,J.
- ANSWER 7.04 (2.50)
Actions - Trip reactor [+0.5]. Isolate letdown (close CV-2001) [+0.5]. l Close PCP controlled bleedoff containment isolation valves (CV-2083 and CV-2099) [+0.5 Open PCP controlled bleedoff relief stop valve (CV-2191)][+0.5]. Reason - To contain primary coolant discharge via the drain tank, due to loss of VCT level and other indications, due to loss of cooling-pondpumps,any]ofthereasonsusedforemergency boration [+0.5 for any REFERENCE l 1. Palisades: ONP 24.5, Loss of Instrument AC Bus YO1, p.1 of 3, Revision 12. 00057EK301 ...(KA'S) i ANSWER 7.05 (3.00)
- a. Normal charging [+0.4], HPSI [+0.3], or auxiliary spray [+0.3].
- b. 1.
I HPSI leg [+0.3 pump]s [+0.3] using either cold leg [+0.3] or hot injection. a
- 2. Containment spray pumps [+0.4] using LPSI cold leg injection [+0.3].
- 3. LPSI pumps using LPSI cold leg injection [+0.4].
REFERENCE
- 1. Palisades: E0P 9.0, Functional Recovery, Attachment A.
001000K405 ...(KA'S) i i 4 t- + ,- -
.__-,_._-__..-..-.-.,,.,__.y-,wm.-m.-y , #- _ . . - . ~, ,_.,_-- . __
-_m-%.,___---_._.,_._....-__m, _ . , . _ , , , . - . . _
- 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 23 RADIOLOGICAL CONTROL ANSWERS -- PALISADES -87/05/05-UPTON, J.
ANSWER 7.06 (2.00) .
- 1. notify the control room and evacuate fuel handling area
- 2. ensure fuel handling area charcoal filter is in operation (by placing HS-1894 to " Refuel" position)
- 3. trip the fuel handling area supply fan (V-7)
- 4. trip the fuel handling area fans V-60, 70A, 708 (by placing l HS-1893 to " Emergency Off")
[+0.5 each; item "1." can be split into two immediate actions] REFERENCE
- 1. Palisades: ONP 11.2, Fuel Handling Accident, Section 3.2,
- p. 1 of 2, Revision O.
00036EK303 ...(KA'S) ANSWER 7.07 (1.50)
- a. 1. PCS less than 300 degrees F (250 degrees F per Technical Specifications) [+0.5]
- 2. shutdown coolin are open [+0.5]g isolation valves (MO-3015 and MO-3016)
- b. A maximum of one HPSI pump is allowed to be operable [+0.5]
REFERENCE
- 1. Palisades: SOP 1, PCS, Section 4.0.q and w, p. 4 of 45, .
Revision 16.
- 2. Palisades: Technical Specifications, 3.1.8 and 3.3.2.g.
010000K403 ...(KA'S)
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 24 ANSWERS -- PALISADES -87/05/05-UPTON, J.
ANSWER 8.01 (2.00)
- a. below the 2100 psia parameter line [+0.5]
- b. to maintain the integrity of the fuel cladding --0R--
to prevent the release of significant amounts of fission products to the primary coolant --0R-- to maintain the DNBR greater than 1.3 [+0.75 for any version]
- c. The temperature limits should be reduced proportionately to the deficiency in the vessel flowrate (1 degree F per 1% in flowrate). [+0.75]
REFERENCE
- 1. Palisades: Technical Specifications, Section 2.1, " Safety Limits-Reactor Core," pp. 2-1 and 2-13.
- 2. Palisades: Technical Specifications, Section 3.1, " Primary Coolant System," p. 3-1b.
002020SGK5 ...(KA'S)
_ __ _ _ . _ _. _ . _ _ _ . . _ - __ _ _ _ _ _ . _ _ _ . . _ _ _ _ _ . m._
, 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 25 ANSWERS -- PALISADES -87/05/05-UPTON,J.
I ANSWER 8.02 (2.00) - i
] a. none [+0.5]
- b. Tech. Specs. Section 3.0.3 is going to become appitcable [+0.5 withoutthedetailsbelow).
4 By 2230 hours action shall have been initiated to place the unit in a condition in which the Specification does not apply [+0.25] by placing it, as applicable, in: 1
! 1. at least HOT STANDBY within the next 6 hours [+0.25]
i 2. at least HOT SHUTDOWN within the following 6 hours [+0.25], and
- 3. at least COLD SHUTD0WN within the subsequent 24 hours [+0.25]. !
l 4
- c. Tech. Specs. Section 3.0.3 applies with the actions listed in "b".
[+0.5; +1.0 for detailing the additional actions, if they are not j listed in part "b".] REFERENCE i Palisades: Technical Specifications, Section 3.0, " Limiting 1.
! Conditions for Operation", p. 3-1.
000000SGA8 ...(KA'S) 1 I ANSWER 8.03 (2.00)
- a. PCS temperature should be above 375 degrees F [+0.5]. l
! b. Fluence here is the fast neutron flux integrated over time
- to which the reactor vessel is exposed. --0R-- ,
I ! Fluence here is fast neutron dose or exposure to which l the reactor vessel is exposed. l [+1.0] i Before the fluence exceeds the limit in the figure, the l heatup and cooldown curves should be updated [+0.5]. I REFERENCE i I 1. Palisades: Technical Specifications, Section 3.1.2, "Heatup ! and Cooldown Rates," pp. 3-4 and 3-9. j 002020SGK5 ...(KA'S) l [ l i
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 26 ANSWERS -- PALISADES -87/05/05-UPTON, J.
ANSWER 8.04 (2.50) ,
- a. Dose equivalent I-131 shall be that concentration of I-131 which alone would produce the same thyroid dose as the quantity and isotopic mixture of the iodine isotopes actually present. [+0.5]
- b. Limit "2." is concerned with radioactive isotopes other than todines [+0.25] which have half-lives greater than 15 minutes [+0.25], E is an average of the disintegration energies [+0.5].
- c. 500 degrees F is used because the saturation pressure associated with 500 degrees F is below the lift pressure of the atmospheric steam-relief valves of the S/Gs.
--OR--
This would prevent a release to the atmosphere in the event of a S/G tube rupture. [+1.0 for either] REFERENCE
- 1. Palisades: Technical Specifications, Section 3.1.4, " Maximum Primary Plant Coolant Radioactivity," pp. 3-1,7 and 3-18.
- 2. Palisades: Technical Specifications, Section 1.4, " Miscellaneous Definitions, p. 1-4.
002020SGA8 .. (KA'S) l I
. 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND 1. IMITATIONS PAGE 27 ANSWERS -- PALISADES -87#3/05-UPTON,J.
ANSWER 8.05 (2.50) .
- a. If you had a completed NRC Form 4 on file, the limit is 3000 mrem per quarter; so, you can receive an additional 2700 mrem (full credit will be given if either the control or alert levels, 2500 and 2000 mrem, respectively, are used) [+0.5]. l
- b. If you have a NRC Form 4, you can receive 3 rem per quarter l or 12 rem per year [+0.5]. However, you can only receive 5 (24-18) or 30 rem cumulative dose, so your dose limit would be 8 rem this year [+0.5].
REFERENCE
! 1. Palisades: Administrative Procedure, " Radiation Dosimetry,"
Proc. No. 7.04, Revision 6, pp. 3 and 4. 00059EK102 ...(KA'S) ANSWER 8.06 (2.00)
- a. Both CBASTs will be maintained at or above 118 inches [+0.5]
whenever the reactor is critical [+0.5]. (If either level drops below 118 inches, then action step 3.2.3(b) would become applicable.) ,
- b. The concern is that if one D/G were inoperable concurrent with the loss of off-site power, the required amount of boric acid may not be available automatically unless the level of T-53A is always maintained above 118 inches. [+1.0]
REFERENCE
~
- 1. Palisades: Standing Orders, " Palisades Plant-Operability of Concentrated Boric Acid Tanks," Standing Order No. 28, Rev.1, March 5, 1985.
006000K601 ...(KA'S) t i 1
,-. ,, -----.,--,,...,-,m -
- 8. i ADMINISTRATIVE PROCEDURES. CONDITIONS AND LIMITATIONS PAGE 28 ANSWERS -- PALISADES -87/05/05-UPTON,J.
3 ANSWER 8.07 (2.00)
- 1. Establish a continuous fire watch on at least one side within
- one hour.
- 2. Verify the operability of fire detectors on at least one side and establish an hourly fire inspection.
[+1.0each] l REFERENCE I 1. Palisades: Standing Orders, " Palisades Plant-Supplemental Technical Specifications," Standing Order No. 54, Rev. 5 . January 13, 1987, p. 10. 086000K403 ...(KA'S) 4 9 i 4
~
EQUATION FORMULA AND PARAMETER SHEET Where og = m2
.(density)t(velocity)1(area)g = (density)2(velocity)2(area)2 ' ~
g - KE = y PE = mgh PEl + KE +PgV3 = PE +KE +P Y22 l 2 2 where V = specific '. _ volume . P = pressure / 'J Q = mcp (Tout-Tjn) Q = UA (T,y,-Tstm) Q=m(hg-h') 2 P=P10(SUR)(t) p ,p ,t/T 26.06 T= " ~ #) 0 0 SUR = T #
=
#Aeff .
i delta K = (K,ff-1) CRg(1-Keff1) = CR2 (1-Keff2) CR = S/(1.K,ff) , (1-Keyfg) (1-Keff) x 100% r SDM = M = (1-Keff2) K eff A'If = 0.08 sec_3 -
~
4
.e, decay constant'= (2) , 693 Ag ;
= Ag e-(decay constant)x(t) 1/2 1/2
- Water Parameters <
Miscellaneous Conversions I gallon = 8.345 lbs 1 Curie = 3.7 x 10 10
~
dps
,1 gallon = 3.78 liters 1 kg = 2.21 lbs 1 ft3 = 7.48 gallons I hp = 2.54 x 10 3Btu /hr 3 6 Density =62.4lbg/ft 1 MW = 3.41 x 10 Btu /hr l 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 I inch = 2.54 centimeters
, 1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 ft-lbm/lbf-sec 2 I ft H O 2 = 0.4335 lbf/in.2
. 1 I
/
/
/
~
m
. s .
. s -
s PALISADES TECHNICAL DATA BOOK FIGURE 41 REV 2 CYCLE 7 HZP RECIPROCAL BORON WORTH VS BURNUP -
-\
i10 -
% l -f . . .
i l e . j 6 , I i . . . t ** i . .,g IOS - A '! . I i
- s l e a . .
i . i
- c. ; I t l ..t
'T ,
e
. . . i
, e .
l 100 - g l . N ! ..
, i .
, ! t l t :
%, h z
- C3 y ,
g 95- _
---- - - - - - + - - - - - - - - - - -
z o 2 '
}' .
a . st . . . . . . _ . . . . . . . . . . _ - 0 l t o g . 4 1 ' o - I
'A 90- = ~ ~-- -
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. PALISADES TECHNICAL DATA BOOK FIGURE 5.1, REV 2 CYCLE 7 INTEGRAL ROD WORTH
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O PALISADES TECHNICAL DATA BOOK FIGURE 6.1, REV 2 CYCLE 7
* *n, WREDICTED BORON CONCENTRATION
' yf, .,. . BURNUP (HFP. ARO. EQUIL. XENON) 1300 ._ . . . . . . . . . ............ . ... . . . . . ... . . . . . . . . . . .. .. . . . .
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-1 0 1 2 EXPOSURE GWD/WTU coeisUWcR$ PowDt CouPAM REAcroR ENemtmose/ PAusaco GCPeekaed ROCTOR DeepG OROLF Q2/27/M CONTROLLm mov
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- Figuro 8.2, Pag 2 1 of 2. Rsv 0
, . FORMULA SMIET g - I. Boron Addition
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- l A. ROT V Gal. B.A. = '5 77 X 10 in (B.A.T.k. PPM-Pc(Initial)
(B.A.T.k. PPM-PC(Final) l
' B. COLD V Gal. B.A. = 8.h8 X 10 in (B.A.T.k. PPM-PC(Initial)
(B.A.T.k. PPM-PC(Final) 4 C. V Gal. B.A. for Gal. of Water Desired PPM
, desired PPM .
= to Berate X increase increase B.A.T.k. PPM i
II. Dilution %
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A. HOT V Gal. PMW = 5 77 X 10 in (PC Initial) (PC Final)
% i B. COLD V Gal. PMW = 8.h8 X 10 in (PC Initial)
- . (PC Final) i i
I 4 III. Blend Ratio B.A.T.k. PPM - 1 = #of Gal. PMW - PC PPM 1 Gal. B.A. REV!EWED BY i i h h 4 t's v .31, fletctor Engineer } 2
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. 4, e,. Figure 8.2, Page 2 of 2, R:v 0
. IV. Mixing Water and Concentrated B.A.
" " "* " *1 "*
o A. . Final Concentration = (Initial Gal. + Gal. H O2 Added) B.- o[7h0 ,(Initial Gal X Initial Conc. ) Find Cone. , Initial Gal. Desired Conc. V. Mixing 2 Tanks of Different Concentrations. ne. Of M "A" X Gal. W. "A")+f, Cone. Of 3."B" X Gal. n. A. Final Conc. = ,,A (Gal. Th. + Gal. Tk. B)
' red = Cal. OfPPM WaterOftoConc.
Borate X Desired B.A. PPM Increase
- PPM Final 73 ,
VI. To Predict Or Determine Evap. Bottoms Concentration Gal. Of Feed , Increase in Bottom Cone. X Evap. Concentrate To Evap. Cone. Of Feed To Evap. Volume ' -
'.V,-
VII. Pressure, Temperature Volume, Of A Gas > s PV PY ( 11 22 1 2 . P = Absolute pressure V = Volume T = Temp. Rankine ( F + h60)
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