ML20196D615
| ML20196D615 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 01/27/1988 |
| From: | Elin J, Meadows T, Morrill P, Royack M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20196D612 | List: |
| References | |
| 50-275-OL-87-02, 50-275-OL-87-2, NUDOCS 8802170304 | |
| Download: ML20196D615 (221) | |
Text
{{#Wiki_filter:,-r s , U. S. NUCLEAR REGULATORY COMMISSION REGION V Report No. 50-275/0L-87-02 .,. Docket Nos. 50-275 and 50-323 Licensee:- Pacific Gas and Electric Company 77 Beale Street San Francisco, California'94106 i Facility Name: Diablo Canyon Units 1 and 2 Examinations at: Avila Beach, California Examination conducted: December 8-17, 1987 ; Examiners: h Date Signed I P. Morrill , XY - A FF- w </r e /sr Date Signed T. Meadows
^* _
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< IfmlW M. Royack Date Sfgned Approved by: / A_ // 7 ,
./Elin, Chitf, Operations Section Ddte 5'igned Sumary: 1 Examinations were conducted December 8 through 17, 1987. The' written ,
examination was administered on December 8, 1986 to three senior reactor operator candicates (SRO) and to twelve reactor operator candidates (RO). The oral and simulator examinations were administered to the candidates from December 9 through 17, 1987. All candidates passed the operating and ; written examinations. l I l' 8802170304 880201 PDR V ADOCK 05000275 PDR s
REPORT DETAILS
- 1. Examiners:
P. Morrill, Chief Examiner, Region V T. Meadows M. Royack ,
- 2. Persons Attending the Exit Meeting December 18, 1986 NRC P. Morrill, Region V T. Meadows, Region V M. Royack, Region V J. Elin, Region V Pacific Gas and Electric Company J. Townsend, Acting Plant Manager T. Martin, Training Manager J. Becerra, Senior Training Instructor B. Terrell, Senior Training Instructor J. Welsch, Senior Training Instructor C. Leach, Senior Training Instructor J. Molden, Operations Training Supervisor
- 3. Written Examination Review The written examination was administered on December 8, 1987. At the conclusion of the examination a copy of the examination key was provided to Mr. J. Molden, of the licensee's Training Department, for review. The written examination key was reviewed by the licensee's Messrs. Terrell, Welsch, Leach, and Molden.
The licensee's review of the written examination resulted in coments which are included in Attachment A of this report. Licensee examination coments were given to the NRC examiners on December 11, 1987 in draft form and were subsequently sent to the NRC on January 12, 1988 by PG&E i Letter No. DCL-88-002 (Shiffer to Martin). The written examination master key was revised, as described in Attachment A to this report, prior to grading the candidates responses.
- 4. Operating Examinations 1 Simulator and oral examinations were conducted from December 9 through December 17. 1987. During the simulator examination, no generic !
problems were identified. l 1
- 5. Exit Meeting l 1
At the conclusion of the site visit the examiners met with ; representatives of the plant staff to discuss the examination.
i i ATTACHMENT A ! DIABLO CANYON UNITS 1 AND 2 NRC REACTOR OPERATOR EXAMINATION REVIEW ! Examination review conducted by, Molden, Welsch, Terrell, and Leach. i QUESTION: 1.02 r COMMENT: l We request that the examiners accept answers "c" or "d" as correct. j Using the provided Steam Tables yields an Enthalpy value of 1190 btu /lbm.
- When this value is cross-referenced on the provided Mollier diagram, the results are approximately 300 'F. The possible choices are 296 and 305 *F.
We therefore feel that both choices are correct. EVALUATION: j 3 The facility coment is correct, the use of the Mollier diagram will result <
- in approximately 300 *F.
RESOLUTION: ; Facility coment accepted. The answer key is revised to accept answer "c or d". , i QUESTION: 1.18 i COMMENT: , 4 We request that the examiners accept answers "c" or "d" as correct. - l, Depending on the assumed starting point for VARS, either answer would be ' I correct: ; l If VARS were some level "OUT", then lowering voltage will cause VARS to I decrease, answer "c". 4 i If you continue to lower voltage until a Unity Power Factor is reached, or l ! initially assumed that we were operating there, VARS will begin to increase i ! in the "IN", out, direction, answer "d". l Y 1 EVALUATION: I ! i The facility coment is correct. Depending on the assumed starting point of l l
"VARS" in, out, or unity answer "c or d" would be correct. "
j
- I j RESOLUTION
i Facility coment accepted. The answer key is revised to accept answer "d or c". 4 i
.--- , _ _ _ _ . . - _. . .. - , . _ _ = _
QUEST 10N: 1.19 COMMENT: We request that the examiners accept answers "a" or "b" as correct. Our reasons are similar to the above question, the possible choices are too close together. Using the provided tables, all 3 reviewers came up with an answer of approximately 130 gallons, halfway between the two possible choices. Also, at DCPP we teach a thumbrule of 3 gallons of boric acid to increase the ppm by 1. Applying this yields 126 gallons. Therefore we feel both answers are correct. EVALUATION: The facility coment is correct. The tolerances in the use of the graphs and nomographs could lead to a response that falls in between answers "a and b". RESOLUTION: Facility comment accepted. The answer key is revised to accept answer "a or b". i QUESTION: 2.01 COMMENT: We request that the examiners delete part "c" from this question. And increase the point value of the parts of this answer. The problem is in the reference material provided to you. The Unit 1 drawing labels this correctly as the Core Barrel, while the Unit 2 drawing labels the same item as the Core Plate. The true answer was not among the choices. A TIP will be initiated to correct our System Descriptions. Attached are the highlighted unit drawings from our system descriptions. EVALUATION: The reference material provided for the examination preparation was incorrectly labeled. RESOLUTION: Facility coment accepted. The answer key is revised to delete item number ]
"c" from the response and the point value of the question has been reduced ,
to 2.5 points. l 1 I
QUESTION: 2.02 COMMENT: r We request that the following be added to the answer key as possible correct answers to part "b": 1 Verification of value positions on the Monitor Box. STP done every shift to verify valve positions. Verification that the 8980 valve closed alarm is not in. Use of a locking device on the manual valve SI-1. The answer given in the key is correct, however, we feel that any of the above : answers are also just as correct. j EVALUATION: Verification of valve position by use of the monitor lights for valve position indication on the Monitor box is required by STP I-1A and is done every shift to verify valve position. This is an action which is taken to insure that the valves from the normal source of water to the RHR pumps are or remain open. (Reference STP I-la.) ., The verification of SI-1 being locked open is part of the valve line up checks and as stated is not an acceptable answer to the question. , The use of alarms requires an action after the valve has left the open , i position. This does not insure that the valve remains or is open. RESOLUTION: Facility comment partially accepted. ; The answer key is revised to accept "Verification of valve position on the monitor box or that STP I-1A is done every shift to verify valve position." I i i l i l l t 4 i 2 ) 1 i
l l j QUESTION: 2.03
- COMMENT: ,
- We request that the answer key be modified to include brackets in parts "c" '
- and "d" as follows
3 ) c. Reactor coolant system (loop 4) pressure at 700 psig. },
- d. Reactor coolant system pressure (PT-403 or 405).
The permissive and auto closure signals for 8701 come from PT-405, while the 8702 signals come from PT-403. The required knowledge should not include the . loop origin for the signal, and in part d, the' loop origin depends upon which valve is being referenced. EVALUATION: The noun name or equipment numbers of the components are equivalent. <
- RESOLUTION
Facility coment accepted. The answer key is revised as follows: j c. Reactor coolant system loop 4 (PT-403) pressure, at 700 psig, j j d. Reactor coolant loop 3 or 4 (PT-405 or PT-403) (Hot leg) pressure, j i h i l l l
- ?
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- t I i s >
i t [ i I d '
I . QUESTION: 2.04 I COMMENT: \ l We request that the examiners accept 12 kv busses D and E as the correct answer to part "a" without reference to which pump is supplied by which bus. We do not train the operators to memorize which pump is supplied by which bus, instead we train to know which busses supply the pumps and what actions to do on a loss of that bus. We also feel that the KSA catalog does not support this detail. I We also request that the answer to part "b" be modified to include as a- ; j possible answer the mechanical operation of the anti-rotation device (by the i pawls on the flywheel engaging the frame mounted ratchet as the rotor comes ; to a stop thereby preventing reverse rotation). We feel that the question, as written, could be interpreted as solicting this answer also. EVALUATION: For part "a" of the question, the knowledge of power supplies for major loads j is important to the safe operation of the plant. The question is in , accordance with Examiner Standard ES-202, paragraph B.3 and NUREG 1122 KSA
- catalog knowledge requirements.
For part "b" of the question, it could be interpreted to solicit the actual l mechanical operation of the anti-reverse rotation device. l
- RESOLUTION
- r Facility coment partially accepted. For part "b" of the question the answer i key is revised to accept "by the pawls on the flywheel engaging the frame i mounted ratchet as the rotor comes to a stop thereby preventing reverse !
- rotation." for 0.S points. ,
1 6 1 i 4 ! I i l [ 5 l . ____ ,
e, QUESTION: 2.05 COMMENT: We request that the examiners accept "Reactor Vessel Head Vent" as a possible answer to part "a". This is a more common name for the same device. We also request that the examiners accept "PT-403 and PT-405" as possible connection names to part "b" of this question. Again the answer key is correct, but these are also methods of referring to the requested loop connections. EVALUATION: In part "a", the names appear to be interchangeable and are for the same component. Inpart "b", the equipment numbers are equiva'.ent to the equipment noun names. RESOLUTION: Facility coment accepted. The answer key is revised as follows:
- a. "Reactor vessel head vent" is added as an alternative Start-up head vent,
- b. "PT-405 and PT-403" are added as alternatives to loop 3 and 4 respectively.
QUESTION: 2.06 COMMENT: We request that the examiners accept "releasing into the Sparger ring or Sparged" as possible answers to part "c" of this question. EVALUATION: The sparger is physically located below the water line in the PRT, therefore, "releasing into the sparger ring or sparged" are correct alternative answers. RESOLUTION: Facility coment accepted. The answer key is revised to accept "releasing into the sparger ring or sparged". l l l
t ] l QUESTION: 2.07 COMMENT. We request that the word "instrumert" be bracketed in part "c" of the l answar. The question asks for the actuating fluid, not system. ; EVALVATION: i The "actuating fluid" for the letdown orifice stop valves is "air". 2 therefere, the word "instrument" is not required for_a complete answer, t i RESOLUTION: l Facility coment accepted. The answer key is revised to put "instrument" i in parentheses. 1 QUESTION: 2.09 COMMENT: We again request that the word "instrument" be bracketed in part "a" of the I answer. The question calls for fluid not system. ! EVALUATION: ' The "fluid" used to open the MSIVs is "air", therefore, the word "instrument" is not required for a complete answer. - 1 j RESOLUTION: I 1 Facility coment accepted. The answer key is revised to put "instrument" in ' j parentheses. I ! l QUESTION: 2.10 i i i
- COMMENT
- i 3
l We request that part "c" of the answer also include as possible answers a "failed closed by design". As this is also a design feature to prevent { inadvertent RCS cooldown. j EVALUATION: l , j The facility coment is correct. The 10% steam dump valves are designed to ' i fail closed upon a loss of air and/or electrical control power. Reference j System Description C-2b. l RESOLUTION: l Facility cement accepted. The answer key is revised to accept "Fail Closed" as an acceptable answer, ) l ;
- t I !
QUESTION: 2.11 i COMMENT: We request that the examiners accept "higher rod worth" as a possible answer to the question. The stated references give both as the reasons. It is attached and highlighted. EVALUATION: The rod worth for Unit 2 is 1% less than for Unit 1. (ReferenceSystem Description A-3a, Unit Differences). RESOLUTION: Facility comment accepted. The answer key is revised to accept "Higher rod , worth" as an acceptable answer. .i QUESTION: 3.01 , 4 d COMMENT: We request that the examiners accept "PT-505" as a possible answer to part "a". It is the P impulse signal that produces T ref. EVALUATION: The "P impulse" signal is produced by PT-505, first stage turbine pressure. RESOLUTION: Facility comment accepted. The swer key is revised to accept "PT-505" as an alternative answer. 1 4 I
t r QUESTION: 3.03 COMMENT: We request that the examiners delete part "a" to this question, and increase the other portions of the question accordingly. As stated above in the comments to question 2.04, we do not feel this kind of memorization is required nor supported by the KSA catalog. We request that part "b" of the answer key be modified to accept as a possible answer "available for auto", as this is the practical use of the indication. EVALUATION: 1 For part "a" of the question, knowledge of safety related power supplies to major loads is an important knowledge. The question is in accordance with Examiner Standard ES-202, paragraph B.3 and NUREG 1122 KSA catalog knowledge requirements. For part "b" of the question, "available for auto" is synonymous with "auto-after-off". RESOLUTION: , Facility comment partially accepted. The answer key is revised to accept "available for auto" as an acceptable answer for part "b" of the question. j a a
QUESTION: 3.07 C0!HENT: We request that the examiners accept the high pressure trip of 2385 psig es a , possible answer, , The tem "increase button" can be interpreted two ways, one which strictly looks at the increase arrow button, and the other that looks at the application of increasing cressure by using the decrease button. It should : be clear in the candidates response how he interpreted your question, and the 1 answer key should reflect the acceptance of either answer as the concept is i understood in either case. EVALUATION: In order to increase pressurizer pressure the master pressure controller output signal is decreased by pressing the "increase" or "down" pointed arrow on the controller. In order to decrease pressurizer pressure the output from the master pressure controller is increased by pressing the "decrease" or i "up" pointed arrow on the centroller. Interpretation of the question could mean that increasing the output signal of the controller was taking place or that a decreasing output signal was increasing the actual pressurizer pressure. Reference System Description A-4a. RESOLUTION: Facility comment partially accepted. The answer key is revised to accept "2385 psig" as an acceptable response if the response is provided with an explanation. QUESTION: 3.08 C0tPENT: We request that the examiners accept "P-14" as a possible answer to part "
"a.1" of the answer key.
EVALUATION: Steam generator high-high level is the noun name for P-14, which causes a . feedwater isolation. ; RESOLUTION: l Facility coment accepted. The answer key is revised to accept "P-14" as an acceptable response to part a.1. l i I I
/
QUESTION: 3.10 COMMENT: We request.that.the examiners accept "busses F and.H" as the correct answer. Reasons stated in response to questions 2.04 and 3.03 above. EVALUATION: Knowledge of safety related power supplies to major loads is an important knowledge. The question is in accordance with Examiner Standard ES-202, paragraph 8.3 and NUREG 1122 KSA catalog knowledge requirements. RESOLUTION: Facility comment not accepted, s
^
QUESTION: 4.03 s COMMENT: ?- . WerequestthatthN'exaMinersaccept"Ma.nuallyde-energizeloadcenters130
& E" as a possible answer to.-part."a_" of this question.
The operators are taught on a Unit- 1 simulator, and are, therefore, likely to respond with unit i load center numbers. The concept of the desired answer is correct, even with unit i load center. numbers. EVALUATION: , , The control switches for load centers 13 D and E and 23 D and E are located in identical positions for both Units 1 and 2. Ths question clearly states that Unit 2 rector has tripped, therefore, the response requires a response for. Unit 2. RESOLUTION: '
+ -
+, , ,
Facility'cShmentpartia'llyaccepted;,Thea'nswerkeyisrevisedtoaccept t
"Manually de-energize load centers 13.'D'and E" as acceptable answer for half ..
credit (0.5 points). s' l 4 Msf
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QUESTION: 4.07 COMMENT: We request that the examiners reconsider the point allocating of this question. We feel that the value given to part "b" is too high. EVALUATION: The point allocation for part "b" of the question is justified since the information requested falls within the one hour time frame for operator action as required by Technical Specifications and required operator knowledge. RESOLUTION: Facility comment not accepted. QUES 110N: 4.09 COMMENT: We request that the examiners delete part "b" to this question. The question solicits a response for the actions for Hot Shutdown, the answer given is for Hot Standby. If the answer were to be modified for Hot Shutdown, it would be a confusing answer involving 3.3.3 implications with assumptions of the actions taken or not taken to get into Hot Standby in the correct amount of time. This is obviously not the examiners intent in this question. EVALUATION: The facility comment is correct. The term "Hot Standby" should have been used in part "b" of the question. RESOLUTION: Facility comment accepted. The answer key is revised to delete part "b" of the question.
.. . ~. - . . .-- .
I QUESTION: 4.11 COMMENT: We request that the examiners accept parts "a.1, a.3 and a.4" as the correct-answer, without given'"a.2". The question solicits the answer of the SI termination criteria. .The actual. criteria is: Subcooling 20 RCS pressure stable or increasing Pzr level 4% Heat sink available: SG NR level 4% in 1 SG or AFW flow 460 gpm Therefore, if the operator listed the first 3 above, and assumed the heat sink criteria satisfied from the question (due to SG level), we feel that this should be considered for full credit. EVALUATION: The facility comment is correct. For part "a" of the question the stem could imply that the "heat sink availability" criteria is met. RESOLUTION: Facility comment accepted. The answer key is revised to accept a.1, a.3, and a.4 as an acceptable response for full credit at 0.667 points each. I l l
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ATTACHMENT-A DIABLO CANYON UNITS 1 AND 2 NRC SR0 EXAMINATION REVIEW
' Examination review conducted by: Molden, Welsch, Terrell, Leach.
QUESTION: 5-4 COMMENT: Answer "a" is also correct. EVALUATION: The question did not state whether the tank or the reference leg varied from calibration temperature. Therefore, the level instrument could indicate correctly. RESOLUTION: Comment accepted, either answer (a) or (c) is correct. QUESTION: 5-7 COMMENT: The figure given, 5-4, is very simplistic and does not give adequate selection points to pick off the various processes. We request that the examiners accept either answer for part (c) and part (f). r (c) 1-2 or 1-3-3 (f) 2-3-4 or 3-4 EVALUATION: The figure does not clearly show the endpoint of feedwater heating. Therefore the facility comment is correct. However, if an individual chose 1-2-3 for feedwatcr heating then 3-4 would be the only correct answer for heat added by steam gererators. , RESOLUTIIN: Connent partially accepted. Add 1-2-3 to answer (c) and 3-4 to answer (f) with the provision described above.
l QUESTION: 5-10 COMMENT: In part (b) it is asked if blowdown. flow was "actually" 400 gpm. There are 2 ways to interpret this.
- 1. If you recalculate power while blowing down 400 gpm, you must remove that mass from the total mass flow rate. This will result in a lower "calculated" power.
This is the interpretation used in the Key.
- 2. If, however, you assume in the question that it was 400 gpm and you didn't know it (i.e., used 0) then your "calculated" power would be the same.
Either answer should be acceptable. EVALUATION: The facility expressed the concern that part (b) of the question was ambiguous. The candidates' assumption may change the answer to that indicated by the facility. RESOLUTION: The grader will consider the candidates' responses in light of any assumptions stated. If no assumptions are stated, the question will be graded to the existing key. The examiner will also revise the point distribution of part (a) to 0.4, 0.4, 0.4, and 0.3 as discussed with the facility reviewers. QUESTION: 5-13 COMMENT: The "best" answer is, of course, answer (b) high gamma flux. We request that the examiners also accept answer (d) high radiation flux. EVALUATION: The "best" answer is clearly (b), therefore the answer should be "(b)". RESOLUTION: Comment not accepted, no change will be made to the key.
QUESTION: 5-24 COMMENT: Equilibrium Samarium does not vary with Power level, answer. (d) is correct, notanswer(c). , Reference enclosed. EVALUATION: The facility is correct. During. editing the order or responses was changed
~w ithout changing the answer. Page 4-31 of the cited reference applies to this situation.
RESOLUTION: Comment accepted, the key will be changed to indicate that (d) is the correct-answer. QUESTION: 5-26 COMMENT: The Test did not give the Student the Curve for Boron Worth that.they use in i calculating an ECC. The Power Change Worksheet (attached) simply uses , 10 pcm/ ppm. They were given the Inverse Boron Worth Curve which they do not use. They should be expected to calculate the change in Power Defect of -235 pcm. Their answer for the corresponding Boron' Change required will vary depending on their assumed Boron Worth. We request that the answer key reflect this latitude. EVALUATION: The inverse boron worth curve supplied by the facility was used due to the fact that the boron worth curve described above was not supplied. It is acceptable to use 10 pcm/ ppm with an answer of - 22.5 +/- 3. ppm. RESOLUTION: Comment accepted, the key will be changed as described above. In the future the facility personnel were requested to send the curves the operators actually used to the NRC examiners. c -s er~ "-m - r r- w ~ - - w
\. - 1 1 (t 1/2 = 53.1 hrs). Therefore, Pm-149 is assumed to be a direct result of 1.07 percent of all fissions. ANp , . I At *TPm f*- Pm m At equilibrium, Pm-149 production Pm-149 removal: AN g at =0=y,If p $-A,N,p p y Ag N,=yIf+ p and Sm-149 production = Sm-149 removal: AN g At = 0 = Ap , Np , - o, N3 , + 1 APm "Pm " #a "Sm
- Substituting and solving for N Sm' Y
PmI f * * *a N,4 3 TPmIf d N,= 3 Sa d
)
TPmI f "Sm " o a Samarium-149 has an equilibrium concentration independent of the flux level. Figure FND-RF-251 is a graph of the reactivity due to Sm-149 versus time af ter startup for a clean reactor core. The Sm-149 concentration reaches its equilibrium value in about 400 hours. 0468C - . , ,
7/86 Page 2 of 2 TITLE: POWER CHANGE WORKSHEET
- 3. Rod Worth
- a. Refer to Volume 9. Figure R4-1F-4 (R4-2F-4)
- b. Initial Rod Position ___ steps. Initial Worth
- pcm
- c. Final Rod Position steps, Final Worth pcm
- d. Net Rod Worth Change (C.3.b - C 3.a) pcm
- 4. Het Reactivity Change (C.2.d + C.3.d) ocm
- 5. Het Boron Change (C.4 + 10 gem) ppm ,
ppm
'- 6. FinalBoronConcentration(C.1+C.5) ppm
- 7. Boration
- a. Refer to nomagraph, Vol. 9 Sect. IA Figure IA-2. Pg. I-2
- b. Gallons 12% Boric Acid to Add callons
- 8. Dilution
- a. Refer to nomograph, Vol. 9. Sect. IA, Fig. IA-4, Pg IA-4 !
- b. Gallons Primary Water to Add gallons. 1 D. Fxy:
- 1. Power level at which Fxy was last measured: % ;
Consult with plant technical department for above j value.
- 2. Refer' to Precaution and Limitations, item D.
Work sheet data and calculations performed by Reviewed by - Comments l 1 l l l 00P18115.064 14XV
4 QUESTION: 6-3' COMMENT: Request that brackets.be placed in part "a" of the answer key as.follows:
"Upper - during natural circulation (or when the RCP in the loop with the hot leg connection is not operating)"
EVALUATION: The subject phrase was intended for clarification. The examiner agrees. RESOLUTION: Comment accepted, the subject phrase will be placed in brackets for clarification. QUESTION: 6-4 COMMENT: In part (c) they should also accept 328-330'F as in part (b) EVALUATION: The facility is correct. The same tolerance as used in part (b) was intended to be used in part (c). . r RESOLUTION: Comment accepted, 328F will be changed to 328 - 330 F. QUESTION: 6-5 COMMENT: Request that the answer key for part "c" have the "Pzr level falls...." deleted, the question doesn't solicit this. EVALUATION: The referenced phrase was intended to be supporting material. It was'not i solicited by the question. RESOLUTION: Comment accepted, the subject phrase will be placed in brackets. I
QUESTION: 6-6 COMMENT: Answer for part (a) lists 0 tot twice, should be 0 PAP on one of them. Answer for part (c). lists "above 15% power" which has nothing to do with the Steam Dumps. Please delete this from the key. EVALUATION: The fifth line of the answer should be "Overpressure delta T channel activated". The comment in brackets in answer (c) is left over from editing and is not applicable. RESOLUTION: Commentaccepted,theanswertoparts(a)and(c)willbechangedasdescribed above. QUESTION: 6-7 COMMENT: ! In part (a) it is not clear which switch at the Hot Shutdown Panel is referred to. If one assumes the ectual Control Switch, rather than the Control Transfer Switch, then it will not matter which position it is in. Accept "Remote" as synonymous with "Control Room". Answer for part (b) should be: (any2)
- 1. Less than 40 psig discharge pressure
- 2. Low voltage on opposite bus 4
- 3. Bus Transfer to Startup
- 4. Bus Transfer to Diesel Ref: Page 13 of System Description E-5 EVALUATION:
In answer (a) "Control Room" or "Remote" express the same switch position from the remote shutdown panel, consequently either answer is acceptable. In answer (b) based on PG&E drawing 437594, Change 14, the facility is correct. Any two of the four start signals would be correct. ; RESOLUTION: Comment accepted. The answers to part (a) and (b) will be modified as described above.
QUESTION: 6-8 COMMENT: Request that the examiners accept the simplified answers given in the Operator Information Manual. Reference is attached. EVALUATION: The question states that the bases for the trips are described in the Technical Specifications. The descriptions in the Operator Information Manual are not the same as the Technical Specification Bases and contains less detail. See Operator Information Manual, pages B-6-4a and B-6-4b. RESOLUTION: Comment not accepted, the examiner will grade based on the existing key. Partial credit may be given for partial answers, i.e. some of the Operator Information Manual data as long as it is consistent with the Technical Specification Bases. QUESTION: 6-9 COMMENT: Answer for part (e) should say "Alarm Only" Answer for part (f) should also say "Transfer Blowdown Tank Outlet from Outfall to EDR" (FCV-498/499 swap to EDR) EVALUATION: In part (e) Unit i discharge isolation valve is jumpered and does not isolate, while in Unit 2 the discharge isolation valve is not jumpered. With proper detector operation Unit 2 would isolate, however, the detector cannot be calibrated at this time. Facility personnel stated that the entire matter is under review with the hope of purchasing a sufficiently i sensitive instrument which can be calibrated to the proper setpoint. In part ' (f) based on PGSE drawing 102931 the tank outlet is switched from the outfall to the EDR and valves FCV-498 and 499 are shut. This is a more precise answer than the one originally solicited. . RESOLUTION: Comments accepted. In part (e) due to the on going facility work on this device, the only correct answer would be a lengthy explanation well beyond the purpose of the question. Part (e) is therefore deleted. In part (f) the answer will be modified to include that the tank outlet is switched to the EDR.
-NUCLEARINSTRUMENTATIONSIGNALSS TRIP SETPOINT** C0lNCIDENCE INTERLOCKS PROTECTION AFFORDED Source Range 10s cps 1/2 P-6, P-10 Start-up Accident I Hi Flux Intermediate 25% power 1/2 P-10 start-up accident Range Hi flux (current equivalent)
Power Range 25% power 2/4 P-10
- Start-up accident Hi Flux (low)
Power Range 109% power 2/4
- Over power (kw/ft)
Hi Flux (high)
+5% power /2sec Ejected rod Power Range 2/4 Rate Trip
-5% power /2sec DNBR >1.30 during single / multiple rod drops
- REACTOR COOLANT SYSTEM SIGNALS -
i OTAT 117.4% i 2/4 *DNBR >1.30
/17.y 2 17 4 : k -A:
OPAT 107.9% - 2/4 *KW/ft penalties
-ws-7 7 44 Loop low flow 90% 2/3 sensors on: P-7, P-8 *DNBR >1.30 2/4 LOOPS (<P-8) 1/4 loops (>P-8)
RCP breaker Open 2/4 loops P-7 DNBR >1.30 open RCP bus under 8050 volts 1/2 sensors P-7 *DNBR >1.30 voltage 2/2 busses RCP bus under 54 Hz 2/3 sensors P-7 DN8R >1.30 frequency 1/2 busses
- Protection assumed in FSAR analyzed accidents 0*Resetvaluesapprox.1%differentthansetpointvalue(seeappropriatesetpoint documentation for exact values).
B-6-4a REV.2 Y00066.TM 101
- PRESSURIZER SYSTEM SIGNALS -
i TRIP SETPOINT** COINCIDENCE INTERLOCKS PROTECTION AFFORDED Low Pressure 1950 psig 2/4 P-7
- ONB High Pressure 2385 psig 2/4
- RCS integrity High Level 92% 2/3 P-7 Prevent water relief, RCS integrity
- SECONDARY SYSTEMS SIGNALS -
Steam Generator 25% NR Lv1-and- 1/2 sensors on any S/G Loss of heat sink low level and SF>FF by - and - flow mismatch 1/2 sensors on same S/G 1.45X10lf/hrU1 1.49X10 f/hr U2 Steam Generator 15% 2/3 sensors on
- Loss of heat sink ,
low low level 1/4 loops I Turbine trip Auto Stop 2/3 P-7
- Limit temperature /
011<50 psig pressure transients ( -or- stop on the RCS i l valves closed 4/4 I
- MISCELLANEOUS SIGNALS -
1 Manual 1/2 Operator judgement Safety Any "S" signal
- Limit consequences injection of accidents Seismic .35g 2/3 sensors Trip reactor in the (in the same event of a double direction) design earthquake
- Protection assumed in FSAR analyzed accidents
** Reset values approx. 1% different than setpoint value (see appropriate setpoint documentationforexactvalues).
( B-6-4b REV.2 Y00066.TM 111
T QUESTION: 6-10 COMMENT: Request that the examiner accept "temperature" as well as "power" in the answers for parts "b and c". This is the actual function of the rods. EVALUATION: Considering that power defect and NTC will stabilize power at a higher or-lower temperature (assuming no rod motion) either power or temperature could be viewed as controlling the rods and evaluation of the effects of the transient. RESOLUTION: Comment accepted. Appropriate changes to the key will be made. QUESTION: 6-15 COMMENT: Request that the examiners accept a description of the Mechanical interlock for the correct answer. EVALUATION: A description of the mechanical interlock demonstrates that the candidates know what prevents paralleling power supplies, consequently this would also be a correct answer. RESOLUTION: Comment accepted. However, no change to the key needs to be made. QUESTION: 7-5 COMMENT: Answer in part (b) should also list as an action for Criticality below the RIL: - Emergency Borate 100 ppm EVALUATION: Unit 2 has a higher rod insertion limit than Unit 1 (for which the question I was originally written), consequently the facillity is correct. < RESOLUTION: l Comment accepted. The key will be changed appropriately. ,
)
QUESTION 7-12 COMMENT:- The answer to part d. should accept partial credit for indicating that there is required access control per 10 CFR 20 for the pump room, tank room, and valve room. Only the tank room is required to be locked or access control per 10 CFR 20. After reading the answer it would be obvious that the "or access controlled per 10 CFR 20" is meaning other options if the tank room can not be locked. The examinee could interprete the questi;n ad identify that all the rooms require access control per 10 CFR 20. EVALUATION: The question asks which area (s) must be locked or access controlled per 10 CFR 20. The facility uses Access Control to mean the "Restricted Areas" which are the radiation areas of 10 CFR 20. The facility comment has some merit. In the future it would be beneficial to avoid the phrase "or access controlled". The ambiguity introduced by this phrase defeats the purpose of the question. RESOLUTION: Question 7-12(d) will be deleted. 2 I l y . - -- , - - . -
..~-~w - - 4 w, , -,, . -,., -,--._i., - -.-- . ,. - ~-- -y, b w ---
QUESTION: 8-01 COMMENT: The Abnormal Procedures for Steam Generator Tube Leaks are as follows: ' OP AP-3A "Steam Generator Tube Leak" OP AP-3B "Steam Generator Tube Failure" In these procedures the actions are to find the leak and reduce power as l necessary for OP AP-3A "SG Tube Leak", and to find the leak and commence shutdown for OP AP-3B "SG Tube Failure". The question denotes AP-3 as the Minor SG Tube Failure, and AP-3A as the SG Tube Leak. But in either instance, the answer key is incorrect. We request that the examiner accept answers "a" or "d" as correct. Dependent upon which I the candidate assumed to be 3A or 3B, one of the two would be the correct l answer. Excerpts of OP AP-3A and 3B are enclosed. EVALUATION: The reference material sent to the NRC by the facility included AP-3 "Minor Steam Generator Tube Failure" dated 7/20/84, and AP-3A "Steam Generator Tube Leak" dated 6/22/87. The existence of AP-3B "Steam Generator Tube Failure" dated 6/22/87 was identified to the NRC personnel during the examination review. Neither AP-3 or AP-38 were listed in the list of material forwarded-to the.NRC. Aside from the administrative problems described above, the purpose of the question was to determine if the candidates knew the different major actions required by procedure for a "tube leak" vs. a "minor tube failure". Considering that procedure AP-3B appears to replace AP-3 and the question reference a procedure no longer in use, the question appears to be no longer valid. RESOLUTION: Comment accepted, however question 8-1 will be deleted.
- v. + y r.- , - -- - - , - = + - --
+s . --e -
p--
PACIFIC GAS AND ELECTRIC COMPANY NUM ER OP AP-3A REVISION 1 DEPART 1ENT OF NUCLEAR POWER GENERATION PAGE 1 0F 5 l DIABLO CANYON POWER PLANT UNITS . AND ABNORMAL OPERATING PROCEDURE TITLE: STEAM GENERATOR TUBE LEAK APPROVED: N'N [odMN PLANT MANAGER ,/ DATE EFFECTIVE DATE SCOPE This procedure provides instructions in the event steam generator tube leakage
~
is indicated by secondary radiation alarms or increased secondary activity. This procedure is applicable for leak rates that might exceed Technical Specification limits but are too small to observe increased charging flow or pressurizer level fluctuations. This procedure and changes thereto require PSRC review. SYMPTORS
- 1. Secondary radiation alarms or up-scale readings
- a. Air ejector off-gas high radiation (RE-15)
- b. S/G blowdown high radiation (RE-19, 23, 27)
'c. Mainsteamlinehighradiation(RE-71,72,73,74)
- 2. Increase in sampled secondary coolant activity
- 3. VCT level trending down or increased automatic rakeup
- 4. Possible Main Annunciator Alarms:
- a. S/GBLOWDOWNHIRAD(PK11-17)
- b. HIGH RADIATION (PK11-21)
- c. MAIN STEAM LINE HI RAD (PK11-18)
- d. PLANT VENT RADIATION (PK11-25)
- e. STM JET AIR EJECT HI RAD (PK11-06) UNIT 2 ONLY 00641201.022 11 J
I DIABLO CMYON POWER PLANT NUM ER OP AP-3A (- REVISION PAGE 1 3 0F 5 TITLE: STEAM GENERATOR TUBE LEAX UNITS 1 MD.2 ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 4. CHECK SG Specific Activity:
- a. Notify Chem and Rad to perform specific activity analysis on leaking steam generator,
- b. Refer to Tech Spec 3.7.1.4 for IF secondary activity limit is compliance to SG specific exceeded activity limits. THEN GO to Cold Shutdown per Tech Spec action statement.
- 5. CHECK SG Leak Rate:
l
- a. Perf orm STP R-10 Part C. l
- b. Notify CARP to perfom CAP D-15'
'(' to determine specific SG 1eak I rate.
- c. Refer to Tech Spec 3.4.6.2 for compliance to SG 1eakage limits. .
- 6. DETERMINE Response To SG Tube Leak:
o If continued operation is desired go to OP L-4 NORML OPERATION AT POWER. o If leak rate reduction is desired continue with this procedure.
- 7. REDUCE Power level as needed to attempt to reduce leakrate.
B. REVERIFY leak Rate LESS Than Return to Step 7 if leakrate Technical Specification Limit _: reduction is desired. o Perform STP R-10 Part C o Notify CARP to perfom CAP D-15 o Verify compliance with Tech Spec 3.4.6.2 , 00641201.022 31
O PACIFIC GAS AND ELECTRIC CCMPANY Mt#EER iOPAP-38 ! REVISION 2 DEPARTENT OF NUCLEAR POWER GENERATION PAGE 1 0F 13 DIABLO CANYON POWER PLANT UNITS , ABNORMAL OPERATING PROCEDURE AND TITLE: STEAM GENERATOR TUBE FAILURE APPROVED: E b'27iY NM'fd PLANT M NAGtk f-) DATE EFFECTIVE DATE SCOPE This procedure provides instructions and general guidelines for operator actions in the event of a steam generator tube failure. The conditions asstaned for entry into this procedure are primary to secondary leak rate clearly in excess of Technical Specification limits and secondary radiation monitors in alaru - condition. This procedure and changes thereto require PSRC review. OBJECTIVE l The objective of this procedure is to rapidly shutdown the reactor, isolate the affected steam generator, cooldown and depressurize RCS to the point where primary to secondary break flow is stopped, without actuation of reactor protection or safeguards systems. SYMPTONS
- 1. Indication of RCS leakage:
- a. Increase in charging flow
- b. Increased VCT makeup frequency
- 2. Secondary systen radiation monitor alarms:
- a. Air ejector off-gas (RE-15) I 1
- b. Steam generator blowdown (RE-19, 23, 27)
- c. Mainsteamline(RE-71,72,73,74) 1
- 3. Increase in sempled ser.ondary coolant activity.
l 00641302.02z 11 1
DIABLO CANYON F0WER PLANT NUEER OP AP-3B REVISION 2 PAGE 5 0F 13
. TITLE: ABNORMAL OPERATING PROCEDURE STEAM GENERATOR TUBE FAILURE UNITS 1 AND 2 ACTICW/ EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 4. DETERMINE Affected Steam Generator:
- a. Contact Chemistry and Radiation Protection - C0lHENCE SAMPLING SGs.
- b. Check main steam line radiation monitors. -
- c. Use portable radiation detectors to survey main steam leads or SG blowdown lines.
- d. Continue with Step 5 concurrent -
. with Step 4.
- 5. COMMENCE Unit Shutdown:
- a. Notify system dispatcher that unit is coming off the grid.
- b. Begin load reduction in accordance with OP L-4 up to 50 MW/ min ramp down rate.
- 6. PLACE Reactor in HOT STAN08Y in accordance with OP L-5 00641302.02z 51
_____._2._- - _ a-________-_-__._.__-_. -
I 1 QUESTION: 8-3 . l COMMENTS: l See attached drawings. No key was provided. Copy of NPAP enclosed to check against examiners' key. EVALUATION: The NPAP was not previously requested to avoid possible question compromise. The Chief Examiner requested this document from the licensee after the examination was completed. RESOLUTION: Supplement I to NPAP C-101, Figure 1 and 2 will be used for the examination key. - QUF.STION: 8-4 COMMENT: I Administrative Procedure C-150 was rescinded in May of 1987, its' scope is - now covered by the switching procedures. Refer to the atteched AP C-7S1 for l description of caution tag usage. Therefore, request that the examiners ! accept the following variation to the answer key: ' Part C: Caution tag. This is actually a more correct answer to the type of tag used. The switching tag is the' document used as a procedure, unlike the- ,- name implies. EVALUATION: 1 - Based on the esitnination of AP C-150 the switching procedures now cover the. same scope. The caution tag would be appropriate for the grounding switch based on C-751, Revision 2, page 3. RESOLUTION: The coment is accepted, the correct answer is the caution tag.
. 6 PACIFIC GAS AND ELECTRIC COMPANY DEPARTMENT OF NUCLEAR PLANT OPERATIONS
- DIABLO CANYON POWER PLANT UNIT NOS.1 AND 2 SUPPLEMENT 1 TO NUCLEAR PLANT ADMINI,STRATIVE PROCEDURE C-101 TITLE: CONFINES OF CONTROL ROOM AT DIABLO CANYON -
SCOPE This procedure identifies the boundaries of the Control Room at Diablo Canyon. PROCEDURE
- 1. The nonnal boundaries of the control room for the Control Operator (operator at the controls) shall be as shown on Figure 1.
- 2. The emergency boundaries of the control room for the Control Operator shall be as shown on Figure 1. ,
- 3. The boundaries of the control room for the senior Ifeensed operator shall
- include aret.s 1 and 2 above, the Shift Foreman's office and other appropriate adjacent areas. This area is shown on Figure 2.
FIGURE 1
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- I h EMattrNcV CdWDW Raum*r 4xre PAGE I 0F 2 2 REVISION DATE 3/28/80 APPROVAL [ M F S/4" #O PLANT SUPERINTE @ / DATE l .
DIABLO CANYON POWER PLANT UNIT NOS0 -1AS2 .' SUPPLEMENT 1 TO NUCLEAR PLANT ADMINISTRATIVE PROCEDURE C-101 . TITLE: CONFINES OF CONTROL ROOM AT DIABLO CANYON F ( i t , . FIGURE 2 ! i ! v >. x. w..v,y,..a: ,:.. ....;p:.;.9,:.!.<: Nrh 9*_:jj.gr.g;.; g.;;..;.;
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i 1 PAGE_2 0F 2 REVISION 2 DATE 3/28/80 i
- . - _ _ _ , _ . _ . , - , , _ y - _ - , - _ _ . _ . _ _ , _ _ -- , . - , , _v.-- - _ -
DIABLO CANYON POWER PLANT Nt#EER AP C-7S1 REVISION 2 PAGE 3 0F 6 TITLE: PLANT TAGGING REQUIREMENTS UNITS 1 AND 2 1 3.3.3 Equipment that is tagged with Man-On-Line Tags shall not be operated for any reason, j 3.3.4 A Man-On-Line Tag shall be filled out with the following information, as a minimum. ,
- a. The description or number of the device to which the tag is to be attached.
- b. The name of the person who is taking the Clearance,
- c. The Clearance Request number.
.. 3.3.5 Man-On-Line Tags shall be removed and destroyed when no longer needed. ' l 3.4 <aution Tags 1 3.4.1 Caution Tags shall be used to identify any plant equipment that should not be operated for reasons other than those requiring use of a Man-On-Lina Tag. Examples of these i applications are: Tagging open vents and drains in l conjunction with a clearance, tcgging closed an electrical I ground switch, and for Administrative Tag Outs.
3.4.2 Caution Tags shall not be hung in the plant unless their installation and removal is documented by an approved procedure, Clearance Request, Work Order Activity or , Administrative Tag Out. 3.4.3 Caution Tags shall be completely filled out with the l following information. l
- a. The description or number of the device to which the l Tag is to be attached.
- b. The persons name who initiated the Tag,
- c. The Clearance, Procedure number, or Administrative Tag Out number.
3.4.4 Caution Tags should be hung by personnel who are
. responsible for carrying out the procedure requiring the tag. (Operations personnel hang tags involved with clearances and Administrative Tag Outs, I&C personnel hang tags involved with instrumentation work, etc.)
00064202.012 31
QUESTION: 8-7 COMMENT: Request the following modifications be accepted as part of the key for part b of this question: AP C-104S1 page 2 also states that if a component has been cleared on a previous clearance which was independently verified, verification of the subsequent clearances is not necessary. Therefore, this answer should be~ accepted for full credit also. ' Admin procedure excerpts included. EVALUATION: The facility comment and reference to C-104S1, page 2, are correct. The ' alternate answer would demonstrate a high level of knowledge in the subject area of the question. RESOLUTION: The comment is accpeted. QUESTION: 8-11 COMMENT: Request that the examiner change the key to part "d" of this question to "ALERT". The question is an ATWS Without apparent core damage. No mention of an SI or other transient in the question. Our procedures on this are G-1, and the Appendix Z of FR-S1, Both are enclosed and highlighted. , l Appendix Z designates this as an ALERT, only upgraded to a SITE AREA EMERGENCY if an SI was initiated before the rods were inserted into the core. In EP G-1 the ATWS concern is called out in the ALERT ar.d the SITE AREA EMERGENCY areas. Under the NUREG Column the difference is in the Transient condition needed for a SAE category, the DCPP column is not as clear. It in - fact uses the same criteria for both. This problem is being forwarded to the Emergency Planning group for correction in future revisions to EP G-1. EVALUATION: j Based on a review of EP G-1, pages 25 and 33 and FR S-1. Appendix Z, the facility comment is correct. The facility identified the ambiguity in procedure EP G-1 and appears to be taking appropriate corrective action to ; clarify that the correct classification is "ALERT". l RESOLUTION-The coment is. accepted and the key will..be.cbanged. J
. . . .. .. . . .. . . .. .. . . . . . . . . . . - . . .. . . w. . c . c . . . . . . . . . . .
PACIFIC GAS AND ELECTRIC COMPANY NUM ER AP C-104S1 ( REVISION 5 ( DEPARTMNT OF NUCLEAR POWER GENERATION PAGE 1 0F 8 DIABLO CANYON POWER PLANT UNITS l AND ADMINISTRATIVE PROCEDURE TITLE: INDEPENDENT VERIFICATION OF OPERATING ACTIVITIES SUPPLEENT 1 APPROVED W. 8 PLANT MNAGER ,_ / MlM DATE EFFECTIVE DATE SCOPE This procedure describes the method of-implementing the requirements of independent verification of operating activities at Diablo Canyon as defined in NPAP C-104 Independent Verification of Operating Activities. This procedure and changes thereto require PSRC review.
. DISCUSSION ,,
Independant verification of configuration changes to plant systems important to safety shall be performed during routino system aligment changes, removal from ( and return to service for plant maintenance or testing, installation and removal of jumpers, lif ted circuits, or mechanical bypasses, and alignment to initiate liquid or gaseous radwaste or chemical discharges. This provides additional assurance that these tasks are properly performed to comply with the plant procedures ated technical specifications and that the health and safety of the public and environmental quality is protected. INSTRUCTIONS
. A. Removal from Service / Clearances When equipment which is important to safety is removed from service for I maintenance or testing, independent verification will be performed to assure that other plant systems are not affected unknowingly by the clearance. This verification shall proceed as follows.
- 1. A second licensed operator, other than the Shift Foreman shall review the clearance prior to the clearance being initiated to verify that the proper citarance points are addressed and all l items documented on the clearance (Tech Spec Applicability, Post l Maintenance Test, etc,) have been properly addressed. This I verification will be documented by that person's signature in the ,
spot designated as "Reviewed By" under the preparer's signature in l Section III of the Clearance Request and Job Assignment Sheet. 1 X000907a.01 11 l
i* . DIABLO CANYON POWER PLANT IENGER AP C-104S1
^
REVISION 5 PAGE 2 0F 8 TITLE: INDEPENDENT VERIFICATION OF OPERATING ACTIVITIES SUPPLEMENT 1 UNITS 1 AND 2 ,
- 2. Following the clearance being approved and after all tags are hung, a second persen shall independently verify the clearance by:
a) Verifying that all clearance points are in their proper position (valves closed, breakers open, etc.). b) Verifying that all tags are on their respective clearan;e points. J c) Verifying that any necessary additional measures called ' out on the clearance have been satisfied. (Tanks drained. . piping vented, etc.) This verification shall be l documented in the independent verification signoff on the ' clearance fom. (d) In the case where multiple clearances are issued on the same piece of equipment, independent verification is not necessary after the first clearance is issued providing the additional clearances do not increase the scope of the original clearance. If independent verification of a clearance is waived due to a previous clearance on the equipment, it should be documented in the Remarks section of that clearance, including the clearance number that was independently verified. e) This independent verification may be perfomed by eitner , an operator or the journeyman taking the clearance. B.
- Return to Service / Clearances When equipment which is important to safety is returned to service following maintenance or testing, independent verification will be performed .to assure the equipment is This verification will proceed as follows: properly returned to service.
- 1. If an approved surveillance test is performed as a post-maintenance test which demonstrates that all alignments are correct on that piece of equipment, then the test will satisfy the independent verification requirements. If this is the case, the performance of the test shall be entered in the independent verification signoff on the clearance request fom, d
X000907a.01 21 i
NUMBER EP FR-S.1
, . . - h w. ._._:0 Pacific Gas and Electric Company REVISION 0 DEPARTMENT OF NUCLEAR PLANT OPERATIONS
^ !
!I 0iAetO CANYON POWER PLANT UNIT NO(S) 1 AND 2 PAGE 1 0F . 9
[I'j TITLE EMERGENCY PROCEDURE RESPONSE TO NUCLEAR POWER GENERATION /ATWS IMPORl ANT APPROVED
- ' M 5- 5 TO PLANT MANAGgil l DATE SAFETY SCOPE This procedure provides actions to add negative reactivity to the core if it is observed to be critical when expected to be shut down. This procedure and changes thereto requires PSRC review.
SYMPTOMS OR ENTRY CONDITIONS
- 1. E-0, REACTOR TRIP OR SAFETY INJECTION, Step 1, when reactor trip
, is not verified and manual trip is not effective.
- 2. F-0.1, SUBCRITICALITY Critical Safety Function Status 1ree on either a RED or ORANGE condition.
DCO273 IVI
_ , c . , , _ . _ _ - - _ _ _ _ _ _ - _ _ _ _ _ _ . _ _ _ . a DIABLO CANYON POWER Pt. ANT UNIT NO(S) 1 AND 2 "' ON DATE 3/7/25 l b TITLE RESPONSE TO NUCLEAR POWER GENERATION /ATWS APPENDIX 2 . ENERGCNCY PROCEDURE NO,TIFICATION INSTRUCTIONS
- 1. When the emergency pdIcedure has been activated and upon direction from the Shift foreman proceed as follows:
- a. Designate this event an Alert. Notify plant staff and response organizations required for this classification by i Emergency Procedure G-2 "Establishment of On-Site
- Organization" and Emergency Procedure G-3 "Notification of
'Off-Site Organization" in accordance with Emergency Procedure G-1 "Accident Classification and Emergency Plan Activation."
- b. Designate this, event a Site Area Emergency if safety injection was initiated before rods were inserted into the
-core but no core damage is evident (no abnomal increase in RCS i:oolant activity and no abnomal increcse in gross F failes fuel inditstion), Notify plant staff and response organizatiers required by EP G-2 and EP G-3 ir. accordance
( with EP G-1.
- c. Designate this event a General _Ereroency if one of the-following conditions exist:
- 1) Core damage is evident by:
a) keactor coolant activity greater than 300 uti/cc equivalent I-131, pr b) Radiation levels indicate greater than 2001 gap release (Refer to Appendix H of EP OP-1),
- 2) Complete loss of safe shutdown system simultaneous with rods not inserted in the core.
- 3) Loss of CVCS capability to increase boric acid concentration in the RCS simultaneous with rods not inserted into the core.
- 2. Notify plar:t staf f and response organizations required by EP G-2 l and EP G-3 in accordance with EP G-1.
I i i DC0273 9VI
i PACIFIC GAS MD ELECTRIC COMPANY NUIEER EP G-1 REVISION 8 DEPARTENT OF NUCLEAR POWER GENERATION PAGE 1 0F 51 ' UNITS lIABLOCANYONPOWERPLANT nw . . EMERGENCY PROCEDURE AND
, TITLE: ACCIDENT CLASSIFICATION AND EMERGENCY PLM ACTIVATION M2$87 APPROVED: , e F 47-92
// PLANT MANAGER DATE EFFECTIVE DATE
(/ SCOPE This procedure describes the guidelines for Accident Classification and responsibilities for Activation of the Emergency Plan. Implementation of this procedure constitutes declaration of an emergency condition. This procedure and revisions thereto require PSRC review. GENERAL This procedure provides guidance on activating the emergency plan and classifying an accident. The steps required by this procedure are in addition to the steps required to maintain or restore the plant to a safe condition, k Prompt notification of off-site authorities should be given within about 15 minutes for the Unusual Event class and sooner (consistent with the need for other emergency actions) for other classes. The time is measured from tne time which the Shif t Foreman recognizes that events have occurred which make declaration of an emergency class appropriate. This procedure is organized as follows: ACTIVATION OF EMERGENCY PLAN The initial steps to be taken for each of the established accident l classifications are listed below under l
- 1. Notification of an Unusual Event
- 2. Alert
- 3. Site Area Emergency
- 4. General Emergency
)
00223608.3Az 11 1
'DIABLO CANYON POWER PLANT NUPEER EP G-1 REVISION 8 PAGE 25 0F 51 TITLE: ACCIDENT CLASSIFICATION AND EMERGENCY PLAN ACTIVATION UNITS 1 AND 2
~~ .
I ALERT - (cont.) NUREG-0654, APPENDIX 1 DIABLO CANYON POTENTIAL CONDITIONS INDICATED CONDITIONS 3.4-1) or 100 pCi/gm specific activity (Technical Specification 3.4.8) (Unusual Event l Condition No. 3). !
- 10. Complete loss of any function 10. Loss of both residual heat
, needed for plant cold removal trains. ) shutdown.
- 11. Failure of the reactor 11. Plant conditions indicate the protection system to initiate required conditions for and complete a scram which Reactor Trip has occurred or brings the reactor the required coincidence of I subcritical. bistables have tripped, or l
'k trip is manually activated, I and Nuclear Instrumentation l indicates reactor not ;
subtritical (non-negative ) start-uprate).
- 12. Fuel damage accident with 12. a. High Contairner ', Radiogas release of radioactivity to and/or particulate alarms containment or fuel handling or Containment Ventilation building. Isolation caused by high containment activity while in the refueling mode or
- b. High fuel Handling Building Area Radiation l Alarm or Fuel Handling !
Building Ventilation automatic change to the l Iodine Removal Mode, while l irradiated fuel is in the l building. 00223608.3Az 25I I
DIABLO CANYON POWER PLANT NUEER EP G-1 REVISION 8 PAGE 33 0F 51 TITLE: ACCIDENT CLASSIFICATION MD EMERGENCY i PLM ACTIVATION UNITS 1 MD 2 SITE AREA EMERGENCY (cont.) NUREG-0654, APPENDIX 1 DIABLO CANYON POTENTIAL CONDITIONS INDICATED CONDITIONS
- g. Complete loss of Instrumentation or Controls required for any of the systems capabilities in items 7.a-f. above.
- 8. Transient requiring operation 8. Plant Conditions indicate the i of shutdown systems with required conditions for Reactor
~
failure to scram (continued Trip has occurred or the power generation but no core required coincidence of damage issnediately evident), bistables have tripped, or trip is manually activated (ALERT Condition number 11), and power generation indicated on power range channels, and no gross fuel failure evident (absence of ALERT Condition No. 1). ;
- 9. Major damage to spent fuel in 9. a. High Containment Radiogas ,
l containment or fuel handling and/or particulate alarms building (e.g., large object or Containment Ventilation l damages fuel or water loss Isolation caused by high !
. below fuel level). containment activity while in the refueling mode or
- b. High fuel Handling Building Area Radiation Alars or fuel Handling Building Ventilation automatic I change to the lodine I Removal Mode, while i irradiated fuel is in the building. (Alert Condition l number 12) and Confirmed j gross fuel damage or loss i of water level to below fuel level.
00223608.3A2 331 l L
1 I QUESTION: 8-13 ,
- COMMENT
- ,
Request the examiner change the key for part "b" of this question to include all 4 of the conditions. Centrifugal Charging Pump 21 is activated by Train A of SSPS, while RHR pump 21 is activated by Train B of SSPS, therefore, it would be a 3.0.3 condition. At DCPP it is not a safe assumption to make that all pumps "1" are Train A, we have many variations to this rule. Charging pump 21 inoperable and Charging pump 22 inoperable would constitute the same conditions as listed above. As well as the inability to satisfy TS 3.5.2 on ECCS Subsystems, or TS 3.1.2.4 on Charging pumps. The examiner might have assumed CCP 21 to be the Reciprocating pump, this is pump 23. EVALUATION: Based on facility drawing 445651, sheet 2, change 3, which tabulates the ESF equipment with power supplies and logic trains, the facility comments are correct. The intent of the question was to evaluate the candidates' knowledge of 3.0.3 not of this particular drawing or of the unique ordering of components, power supplies, and logic channels. It appears that the original purpose of the question is no longer satisfied and therefore the question is no longer valid. RESOLUTION: Comment accepted, however, the question will be dropped, i i 4 i
)
2 1
CoABEC726 2y/3M NG U.S. Nuclear Regulatory Commission Reactor Operator License Examination Facility: DIABLO CANYON UNITS 18 2 Reactor Type: WESTINGHOUSE PWR 4 LOOP , Date Administered: DECEMBER 8, 1987 l Examiner: MICHAEL J. ROYACK l Candidate: " * "
- KEY " * " " l INSTRUCTIONS TO CANDIDATE:
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 ' l 1 east 705 in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts, j
% of Category X of Candidate's Category Value Total j Score Value Cateoory 25 1. Principles of Nuclear Power d'25OM 2+ "
Plant Operation, Thermodynamics, 2 44 Heat Transfer and Fluid Flow
.2 yJ2N 8
- 2. Plant Design Including Safety and sm Emergency Systems 25 , I(J51Y 3. Instruments and Controls
[ 24 4. Procedures - Normal, Abnormal, g tv " Emergency, and Radiological control j d - TOTALS Final Grade KEY X All work done on this examination is my osn. I have neither given nor received aid.
. . . . . . K EY * * * * "
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 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 covsr 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 parentheses after the question and can be used at 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 BLANK. l
- 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.
- l
(
- 18. When you complete your examination, you shall: ,
- a. Assemble your examination as followi:
(1) Exam questions on top. (2) Exam aids - figures, tables, etc. (3) Answer pages including figures which are a part of the answer,
- b. Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.
- 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.
l 4 l
..s
~
s' EQUATION SHEET f = ma y = s/t Cycle efficiency = N Work out) w = mg s = v,t + \at 2 E = aC a = (vg - v,)/t KE = my 2 v g=v A = AN A = A,e' *
+ at PE = agh w - e/t A = in 2/tg = 0.693/tg W = vaP AE = 931am tg(eff) = (t )(ts) 4 = m oh (g 4g) b h
Q=[nC4Tp I = I 4X Q = UAAT I-Ieg -ux Pwr = Vf In ~ I = I, 10 *
. P=P 10 SUR(t) TVL = 1.3/u P=P o et/T HVI = 0.693/u SUR = 26.06/T T = 1.44 DT SCR = S/(1 - K,gg)
SUR = 26 fAeff ) 0 g CR, = S/(1 - K,gg ) T = '(t*/p )+ {(f /p)/x pg CR y (1 - Keff)g = CR (1 *Keff)2 2 7 = t*/ (p p) M = 1/(1 - K,gg) = CR /CR g 0
~ 8)! eff' M = (1 - K,gg) /(1 - K,gf)
# " ( eff" }! eff " OKeff/Keff SDM = (1 - K,gg)/K,gg o= [1*/TK,gg .] + [H/(1 + 1,ggT )]
~
t* = 1 x 10 seconds
-I P = I$V/(3 x 1010) A,gg = 0.1 seconds E = No Idgg=1d22 WATER PARAMETERS Id =1d22 g
1 gal. = 8.345 lba 2 R/hr = (0.5 CE)/d (meters) 1 gal. = 3.78 liters R/hr = 6 CE/d (feet) I ft = 7.48 gal. HISCELI.ANEOUS CONVERSIONS . 3 Density = 62.4 lbm/ft 1 Curie = 3.7 x 10 dps 10 Density = 1 gm/cm 1 kg = 2.21 lba Heat of varorization = 970 Etu/lbm 3 I hp = 2.54 x 10 BTU /hr , Heat of fusica = 144 Btu /lba 6 1 Hw = 3.41 x 10 Btu /hr 1 Atm = 14.7 psi = 29.9 in. Ig. 1 Stu = 778 ft-lbf I ft. H oy = 0.4333 lbf/in 1 inch = 2.54 cm F = 9/5 C + 32
*C = 5/9 ( F - 32)
196b .. - 1335 1650 1000 ~~
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1.0 1.2 1.'4 l'6 1.'8 ~ ~ ~IO~ 2 ~~I2
, # [
/
' 1700 T700 - ,
. cuar ,
1500 1600 [ ,
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1500 L 1500
' MOD MOD ' / netwust i/ / . .
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k / . f pmpg,7-- 7 T100 es 'n / /. 4 y 110
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900 j ///
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900 B00
~800' 1.0 / / U.2 / / 1.41 / .6 1 -- 1.8 2.0 1965 2.2-2 300 1335 1650 J 1000 ENTROPY,s,8tv/4 =R
W b Table 1. Saturated Steam: Temperature Table Specific Volume Enthalpy Entropy Abs Press. Sat. Temp Sd. Sat. Sat. Sat. Sat. Temp tb per Vapor Fahr Liquid Evap Vapor Liquid Evap Vapor Liquid Evap Fahr SqIn. i vtg vg he h tg h5 5 Sfg sg t p vg 0 016022 3304 7 33X7 0 0179 1075.5 1075.5 0.0000 2.1873 2.1873 32.0 32.3 0 08859 0 0041 2.1762 2.1502 34.8 0 016021 3061.9 3061.9 1.996 1074.4 1076 4 34 8 0 09600 0 0081 2 1651 36 I 0 016020 28390 2839 0 4 008 1073.2 1077.2 2.1732 36 8 0.10395 0 0122 2.1541 38.8 0 016019 26341 2634.2 6 018 1072.1 1078.1 2.1663 38.8 0 11249 0 016019 2445.3 2445 8 8.027 1071.0 1079.0 0 0162 2.1432 2.1594 48.8 40 8 1.12163 0 0202 2.1325 2.1527 42 8 0 016019 22/2.4 22//.4 10 :35 1069 8 1079.9 42.8 0.13143 0 016019 2112.8 2112.8 12.041 1068 7 10801 0 0242 2.1217 2.1459 44.5 44 0 0 14192 2.1393 46 8 0 016020 1%57 1965 7 14.047 1067.6 1081.6 0 0282 2.1111 45 8 015314 0 0321 2.1006 2 1327 48.5 0 016021 1830 0 1830 0 16.051 1066.4 1082.5 48 5 0.16514 1704 8 lii054 C m*3 4 0.0361 2.0901 2.1262 50 0 58 I 0.17796 0 016023 1704 8 0.19165 0 016024 1589.2 1589 2 20 057 10642 1084.2 0.0400 2.0798 2.1197 52.0 52.8 020625 0 016026 1482.4 1482.4 22 054 1063.1 1085.1 0.0439 2.0695 2.1134 54 8 54 5 0 22183 0.016028 1383.6 1383.6 24.059 1061.9 1086.0 0 0478 2.0593 2.1070 56 8 55 8 0.23843 0 016031 1292.2 1292.2 26.060 1060.8 1086.9 0 0516 2.0491 2.1008 58.8 58.8 1207.6 28.060 1059.7 1087.7 0 0555 2.0391 2.0946 50.0 88.8 025611 0 016033 1207.6 027494 0 016036 1129 2 1129.2 30 059 1058.5 1088 6 0.0593 2.0291 2.0885 52.8 E2.8 0.29497 0.016039 1056.5 1056.5 32.058 10574 1089.5 0.0632 2.0192 2.0824 E4.0 54.8 56 8 0.31626 0 016043 989 0 989.1 34 056 1056.3 1090.4 0.0670 2.0094 2.0764 EE.8 0 33889 0 016046 926.5 926.5 36 054 10552 1091.2 0.0708 1.9996 2.0704 58 8 58.8 0.36292 0.016050 868.3 868.4 38 052 1054.0 1092.1 0 0745 1.9900 2.0645 78 8 78.8 72 8 0.38844 0 016054 814.3 814.3 40 049 1052.9 1093 0 0.0783 1.9804 2.0587 72.8 74.8 0 41550 0 016058 764.1 764.1 42.046 1051.8 1093.8 0 0821 1.9708 2.0529 74 8 78.8 0.44420 0016063 717.4 717.4 44 043 1050 7 10947 0.0858 1.9614 2.0472 76 8 78.5 0.47461 0 016067 673 8 673.9 46 040 1049.5 1095 6 0 0895 19520 2.0415 18 I 88.8 0 50683 0 016072 633.3 633.3 48 037 1048.4 1096 4 0 0932 1.9426 2.0359 88.8 82.8 0.54093 0 016077 595 5 595.5 50 033 1047.3 1097.3 0 0969 1.9334 2 0303 82.8 0 016082 560 3 560 3 52 029 1046.1 1098 2 0.1006 1.9242 2.0248 HI 84.8 057702 0 1043 1.9151 2.0193 86 3 061518 0 016087 227.5 527.5 54 026 1045 0 1099 0 85.8 065551 0 016093 4%8 496 8 56 022 1043 9 1099.9 0.1079 1.9060 2.0139 88.8 30.8 0 69813 0 016099 468.1 4681 58 018 1042 7 1100 8 01115 1.8970 2.0086 90 0 88.5 2 0033 92 8 0 016105 441.3 4413 60 014 1041.6 1101 6 0 1152 1.8881 82 8 0 74313 34 s 4163 416 3 62.010 1040.5 1102 5 0 1188 1.8792 1.9980 94.0 0 19062 0 016111 0 016117 392.8 392 9 64 006 1039 3 1103 3 0 1224 1.8704 1.9928 SE S 2 SE.8 0 84072 30 0 0 89356 0 016123 370 9 370 9 66 003 10382 11042 0 1260 1.8617 1.9876 Se 8 [, ) .- -. -
Y Abs Press. Specific Volume Enthalpy [ntropy Temp tb per Sat. Sat. Sat. Sat. Sal. Sat. Temp Fahr SqIn. Liquid Evap Vapor Liquid Evap Vapor Liquid Evap Vapor Fahr t p vg vtg vg hg h eg hg s, sig 5g i 100 g 094924 0 016130 350 4 350 4 67.999 1037.1 11051 0.1295 18530 1.9825 180.0 182.s 1 00789 0 016137 331.1 331.1 69 995 1035 9 1105 9 0.1331 1.8444 1.9775 182.8 1M.I 1.06 % 5 0 016144 3131 313.1 71 992 1034 8 1106 8 01366 1.8358 1.9725 IM.3 106 I 1.1347 0 016151 29616 29618 13 99 1033 6 1107 6 0 1402 18273 19675 188.8 Igg I 1.2030 0 016158 280 28 28030 75 98 10323 1108.5 0.1437 18188 1.9626 10s.0 118.3 1 2750 0 016165 265 37 265 39 77.98 1031.4 11093 01472 18105 1.9577 113.8 112.8 13505 0016173 25137 25138 79 98 1030 2 1110 2 01507 I8021 1.9528 112.0 lie s 1 4299 0 016180 238 21 238 22 81.97 1029 1 1111 0 0.1542 11938 1.9480 114.5 118.8 1.5133 0 016188 225 84 225 85 83 97 1027.9 1111.9 0.1577 11856 13433 115.0 118.8 1 6009 0 016196 214 20 21421 85 97 1026 8 11h23 0 1611 13774 1.9386 118.0 120.0 16927 0 016204 20325 203 26 87.97 1025 6 1113.6 0.1646 1.7693 1.9339 120.0 122.s 11891 0 016213 192.94 192.95 89 % 1024 5 1114 4 01680 13613 1.9293 122.8 124.s 1 8901 0 016221 183 23 183 24 91 % 1023 3 1115 3 01715 13533 1.9247 124.0 126 8 1.9959 0 016229 174 08 174 09 93 % 1022.2 1116.1 0 1749 1.7453 I.9202 125.0 123.8 2.1068 0 016238 16545 16547 95 96 1021.0 1117.0 0 1783 13374 1.9157 128.5 130_I 2 2230 0 016247 15732 15733 97.% 1019 8 1117.8 01817 13295 1.9112 130.0 132.5 23445 0 016256 149 64 149 66 99 95 1018 7 1118.6 0.1851 13217 1.9068 132.8 134.0 2.4717 0 016265 14240 142.41 101 95 10173 1119.5 01884 13140 1.9024 134.0 136.8 2.6047 0 016274 135 55 13557 103.95 1016 4 1120 3 0 1918 13063 1.8980 138.0 138 8 2.7438 0 016284 129 09 129 11 105.95 1015 2 1121.1 0.1951 16986 1.8937 138.8 148.8 2.8892 0 016293 122.98 123 00 107.95 1014 0 1122.0 0 1985 1.6910 1.8895 140.8 142.8 3 0411 0 016303 11721 11722 109 95 1012.9 1122.8 0 2018 1 6534 1.8852 142.3 144.8 3.1997 0 016312 11134 11116 111.95 10113 1123 6 02051 1.6759 1.8810 144.5 146.3 3 3653 0 016322 106 58 106 59 113 95 10105 11243 0 2084 1.6684 18769 146.8 148I 3 5381 0 016332 101 68 10110 115.95 1009 3 1125 3 02117 1.6610 1.8727 lesI 158.5 33184 0 016343 9705 9707 117.95 1008.2 1126.1 02150 1.6536 1.8686 150.0 152.5 3 9065 0 016353 92 66 92.68 119.95 1007.0 1126.9 02183 1.6463 1.8646 1528 154.8 4.1025 0 016363 8850 88 52 121.95 1005 8 11273 02216 1.6390 1.8606 1548 156 8 4 3068 0 016374 84 56 84 57 123 95 1004.6 1128.6 02248 1.6318 1.8566 158.R 158 8 4 5197 0 016384 80 82 80 83 125 % 1003.4 1129.4 02281 1.6245 1.8526 158.5 lle O 43414 0 016395 77.27 7729 127.% 1002.2 1130.2 0 2313 1.6174 1.8487 100.0 152.8 4 9722 0 016406 73 90 7392 129 96 1001.0 1131.0 0 2345 1.6103 1.8448 182.8 154.8 5 2124 0 016417 7010 70 72 131.96 999.8 1131.8 0 2377 1.6032 1.8409 154 3 165 5 54623 0016428 6767 6768 13397 998 6 1132.6 0.2409 13 %1 1.8371 188.3 168 I 5 7223 0 016440 64.78 6480 13597 997.4 1133.4 0 2441 13892 1.8333 158.0 178.8 5 9926 0 016451 62 04 62.06 137.97 9% 2 1134.2 0.2473 15822 1.8295 170.8 172.8 6 2736 0 016463 59 43 5945 139.98 9950 1135 0 02505 13753 1.8258 172.8 174.0 6 5656 0016474 56 95 5697 141.98 993.8 1135 8 0 2537 15684 1.8221 174.3 175.8 6 8690 0 016486 54 59 54.61 14399 992.6 1136 6 0 2568 13616 1.8184 175 0 tilI 1.1840 0 016498 5235 52.36 14599 991.4 1137.4 0 2600 13548 1.8147 173.8 G G w a
Enthalpy Entropy Abs Press. Specific Volume Sat. Temp Sat. Sat. Sat. tb per Sat. Sat. Fahr Temp Vapor liquid Evap Vapor liquid Evap Vapor SqIn. Liquid Evap I Fahr h eg he St 5fg Sg v, vig va he t p 14800 990.2 1138.2 02631 1.5480 18111 130.0 0 016510 50.21 50 22 1.5413 1.8075 182.0 IIe 8 7.5110 150 01 989 0 1139 0 02662 1.850 0.016522 48.172 18.189 1.5346 18040 184 8 182.0 152.01 98T8 1139 8 0 2694 0 016534 46 232 46 249 134 g 8.203 44.400 154 02 986 5 1140.5 0 2725 1 5279 1.8004 196 8 8 568 0.016547 44383 0.2756 1 5213 13%9 188 8 18E 8 42.621 42.638 156 03 9853 11413 188.8 8.947 0.016559 1142.1 02787 1.5148 13934 1988 40.941 40.957 158 04 984.1 198 8 9 340 0.0!6572 982.8 1142.9 0 2818 1.5082 11900 192.5 39337 39354 160 05 192.8 9347 0.016585 981.6 11433 0 2848 1.5017 13865 194 8 0 016598 3T808 37.824 162.05 194 g 10.168 980.4 1144 4 0 2879 1.4952 13831 196 I 36 348 36364 164 06 10.605 0.016611 979.1 1145 2 0 2910 1.4888 11798 190.3
~195 8 0016624 34.954 34.970 166 08 198 8 11.058 11460 0 2940 1.4824 11764 2005 33.622 33.639 168 09 977.9 200 3 11.526 0 016637 11475 0 3001 1 4697 11698 284 8 31.135 31.151 172.!! 975 4 284.5 12.512 0 016664 03061 1.4571 11632 298.8 28878 176.14 972.8 1149 0 13.568 0 016691 28 862 0 3121 1 4447 11568 2128 200.8 18017 970 3 1150.5 0 016719 26 782 26 799 212.8 14 696 24 894 18420 967 8 1152.0 03181 1.4323 11505 216.8 15.901 0.016747 24 878 215.0 188 23 %5.2 1153.4 0 3241 14201 11442 229 e 17186 0 016775 23 131 23.148 13380 224.8 220.0 19227 %2.6 1154.9 0 3300 1.4081 0.016805 21.529 21.545 11320 2245 18.556 20 073 1 % 31 9600 1156 3 0 3359 13%1 228 8 20 015 0016834 20 056 0 3417 13842 17260 232 8 228.8 20035 9574 11578 21.567 0 016864 18301 18318 13725 11201 235 8 232.0 20440 954 8 1159.2 0 3476 23 216 0016895 1T454 17471 236 8 16321 20845 952.1 1160 6 0 3533 13609 17142 248 8 240.9 24.% 8 0 016926 16304 0 3591 13494 11085 2448 15 243 15 260 212.50 949 5 1162.0 2440 26.826 0016958 03649 13379 13028 2488 14 264 I4 281 21656 946 8 1163 4 240.0 283 % 0.016990 0 3706 13266 16972 252 0 13358 13 375 220 62 944.1 11643 252.0 30.883 0 017022 03763 13154 1 6917 256 5 12.520 12.538 224 69 941.4 1166.1 236 8 33 091 0 017055 0 3819 13093 1.6862 2EE.8 11345 11362 228 76 938 6 1167.4 203.0- 35.427 0.017089 03816 12933 16808 254 8 11.025 11.042 232 83 935 9 11683 264 0 3T894 0 017123 0 3932 1 2823 1 6755 258 8 0.017157 10 358 10375 236 91 9331 1170 0 268.8 40.500 9355 240 99 930 3 1171 3 0 3987 12715 16702 272.5 43.249 0 017193 9 738 1 2607 16650 276 8 272.0 245 08 9275 1172.5 0 4043 46.147 0.017228 9.162 9 180 275.0 04098 12501 16599 280.8 8 627 8.644 24917 924 6 1173 8 280.0 49.200 0 017264 0 4154 1 2395 1.6548 284.8 8 1280 8.1453 253 3 921 7 1175 0 284.9 52 414 901730 0 4208 12290 16498 288 8 0.01734 T6634 76807 257.4 918 8 11762 55795 0 4263 12186 1 6449 2920 280 0 ))77 4 7.2301 7 2475 261.5 915 9 292.9 59.350 0 01738 04317 12082 16400 296 8 0.01741 6 8259 6 8433 265 6 9130 1178 6 63 064:
P 295 8
? cn Abs Press. Specific volurr,e Enthalpy Entropy Temp tb per Sat. Sat. Sat. Sat. Sat. Sat. Temp Fahr SqIn. Liquid Evap Vapor liquid Evap Vapor Liquid Evap Vapor Fahr t p vg vig vg hg h ig hr 5f Sig 5g I 67.005 0 01745 6 4483 64658 269.7 910 0 1179.7 0 4372 11979 I.6351 388 g 388.0 304.0 71.119 0 01749 6 0955 6 1130 273.8 907.0 1100.9 0 4426 1.1877 1.6303 384 8 388.0 75.433 0.01753 57655 51830 278 0 904 0 1182.0 0 4479 1.1776 1.6256 388 I 312.0 79 953 0 01757 54566 54742 282 1 901.0 1183.1 04533 1.1676 1.6209 312.8 318.8 84 688 0.01761 5 1673 51849 286 3 897.9 1184.1 0 4586 1.1576 16162 315 I 320.8 89 643 0 01766 48%1 4.9138 290 4 894 8 !!85 2 04640 1.1477 1 6116 328.8 324.s 94 826 0 01770 4 6418 4 6595 294 6 891.6 1186 2 04692 1.1378 16071 324 8 328.I 100 245 0 01774 44030 4 4208 2987 888 5 1187.2 0 4745 1.1280 1.6025 328 9 332.0 109 907 0 01779 4.1788 41%6 302.9 885 3 1188 2 0.4798 1.1183 1 5981 3320 338 5 111.820 0 01783 3 9Hil 3.9859 307.1 882.1 1189 1 04850 1.1086 1.5936 336 8 348.3 117.992 0.01787 31699 31878 311.3 878 8 11901 0.4902 10990 I5892 348.9 344.5 124.430 0 01792 3 5834 3 6013 315 5 875 5 1191 0 04954 1.0894 1.5849 344.s 348.8 131.142, 0 01797 34018 3 4258 319 7 8 72.2 1191.1 0 5006 1 0799 1 5806 348 8 352.3 138.138 0 01801 3 2423 3 2603 323 9 868.9 11921 0 5058 1.0705 1.5763 3528 356.0 145 424 0 01806 3 0863 3 1044 328.1 865 5 1193 6 0 5110 1 0611 1.5721 356 8 30s.9 153.010 0 01811 2.9392 2 9573 332.3 862.1 1194 4 0.5161 1 0517 1 5678 358.3 384.0 160.903 0 01816 2 8002 2 8184 336.5 858 6 1195 2 0 5212 10424 1.5637 354 s 388.5 169.113 0 01821 2 6691 2 6873 340 8 8551 1195 9 0.5763 1 0332 1.5595 3Eg g 372.0 177.648 0 01826 2.5451 2.5633 345 0 8516 11961 0.5314 10240 1.5554 312 3 37s.g 186 517 0 01831 2.4279 2.4462 349 3 848.1 1197.4 0 5365 1.0148 1.5513 376.8 388.I 195.729 001836 2.3170 2.3353 353 6 844.5 1198 0 05516 10057 1.5473 30s e 384.0 205 294 001842 2.2120 2.2304 357.9 840 8 1198 7 05466 0.9%6 1.5432 384.8 30s.3 215220 0 01847 2.1126 2.1311 362.2 8372 1199 3 0 5516 09376 1.5392 389 8 392.0 225 516 0 01853 2.0184 2.0369 366.5 833.4 1199 9 0 5567 09786 1.5352 332.8 398.9 236.193 0 01858 1.9291 1.9477 370 8 8291 1200 4 0.5617 0 9696 1.5313 39Es 488.8 247.259 001864 1.8444 1.8630 375.1 825.9 1201.0 0.5667 09607 1.5274 Age 8 404.0 258725 001870 1.7640 11827 3794 822 0 1201.5 0 5717 09518 1.5234 484 0 488.8 270 600 0 01675 1.6877 11064 383 8 818 2 1201.9 05766 0 9429 1.5195 488 8 412.8 282.894 001881 1.6152 16340 388.1 814.2 1202.4 0 5816 09341 1.5157 412.8 418.8 295 617 0 01887 1.5463 1.5651 392.5 810 2 1202.8 0 5866 09253 1.5118 416 8 420.0 308 780 0 01894 1.4808 1.4997 396 9 806 2 12031 0 5915 0.9165 1.5080 428 8 424g 322 351 0 01900 1.4184 1.4374 4013 802.2 1203.5 0 5964 0 9077 1.5042 424.9 428.8 336 463 001906 1.3591 13782 405 7 798 0 12031 0 6014 0 8990 1.5004 428 0 432.0 351 00 0 01913 130266 1.32179 410 1 793 9 1204 0 0 6063 0 8903 1.4966 4320 436.8 366 03 0 01919 1.24887 126806 414 6 789 7 1204 2 0 6112 0 8816 1.4928 43E 8 448.8 381.54 0 01926 1.19761 121687 419 0 785 4 1204 4 0 6161 0 8729 1 4890 448 8 444.s 397 56 0 01933 1.14874 1.16806 423 5 781.1 1204 6 0 6210 0 8643 14853 4448 448 I 414 09 0 01940 1.10212 1.12152 4280 776 7 1204 7 06259 0 8557 14815 44n a PI 431 14 0 01947 1.05764 1 07711 '5 772 3 1204 8 0 6308 0 8411 1.4778 Q3 448 73 001954 101518 103472 0 7678 1204 8 06356 0 8385 14741 au4
O h ' Abs Press. Specific Volume Enthalpy Entropy Temp tb per Sat. Sat. Sat. Sat. Sat. Sat Temp Fahr SqIn. Ltquid Evap Vapor Liquid Evap Vapci Liquid Evap Vapor Fahr t p vg Vfg Vg hg h ft hg s, sig sg t 450.8 466 87 0.01 % I 0.97463 0 99424 441.5 763 2 1204 8 0.6405 0 8299 1.4704 468.8 454 S 485.56 001969 0 93588 0 95557 446.1 758 6 1204.7 0 6454 0 8213 1.4667 454.0 450 0 504 83 0.01976 0 89885 0 91862 450 7 754.0 1204.6 0.6502 0 8127 1.4629 458.8 472.8 524 67 0 01984 0 86345 0 88329 455 2 749 3 1204.5 0.6551 0 8042 1.4592 472.5 475 8 545.11 0 01992 0 82958 0.84950 459 9 744.5 12043 0 6599 0 7956 14555 475 8 480 8 566.15 0 02000 0 79716 0 81717 464 5 739 6 1204.1 0 6648 01871 1.4518 480 0 484.5 58781 0 02009 016613 018622 469.1 7341 1203 8 0 66 % 01785 1.4481 484.0 488.8 610 10 0 02017 0 73641 0 75658 473 8 7291 1203.5 0 6745 01700 1.4444 488.8 492.8 633 03 0 02026 010794 0 72820 478.5 724 6 1203.1 0 6793 0 7614 1.4407 492.5 496 8 656 61 002034 0.68065 010100 483 2 719.5 1202.7 0.6842 01528 1.4370 496 8 500.8 680 86 0 02043 0 65448 067492 487.9 714 3 1202.2 0 6890 0 7443 1.4333 500.0 504.8 70518 0 02053 0S2938 064991 4921 709 0 1201.7 06939 0.7357 1.42 % S04.5 588.5 131.40 0 02062 0 60530 0 62592 497.5 7031 1201.1 0 6987 01271 1.4258 50s.s 512.8 757.72 0 02072 0.58218 0 60289 5023 698 2 1200.5 0 7036 0 7185 1.4221 512.8 515.8 784.76 0 02081 0.55997 0.58079 507.1 6921 :199.8 0 7085 . 01099 1.4183 516.8 520 8 812.53 0 02091 0.53864 0 55956 512.0 687.0 1199.0 01133 0 7013 1.4146 528.I 524.0 841 04 0.02102 0 51814 0.53916 516.9 6813 1198.2 01182 0 6926 1.4108 524.8 528.8 87031 0.02112 0 49843 0 51955 521 8 675.5 11973 0 7231 0 6839 1.4070 5285 532.0 900 34 0 02123 047947 0.50070 526 8 669 6 1196 4 01280 0 6752 1.4032 532.0 535.0 931.17 0 02134 0 46123 0 48257 5311 663.6 11954 01329 0 6665 13993 536.0 540.8 % 2 79 0 02146 0 44367 0 45513 536 8 657.5 11943 01378 06577 13954 548.8 544.3 995.22 0 02157 0 42677 0 44834 541.8 6513 1193.1 01427 0 6439 13915 544.5 548.5 1028.49 0 02169 0 41048 0.43217 546 9 645 0 1191.9 0 7476 0 6400 13876 548.8 552.3 1062.59 0 02182 039479 0 41660 552 0 638.5 1190 6 01525 0 6311 13837 552.8 555.8 1097.55 0 02194 03 7966 0.40160 557.2 632.0 1189.2 01575 0 6222 1.3797 556.8 500.8 1133 38 0 02207 036507 038714 $62.4 625 3 1187.7 0.7625 0 6132 13757 580 0 554.0 1170.10 0 02221 035099 0 37320 567.6 618.5 1186.1 0.7674 0 6041 13716 554 8 550.0 1207.72 0 02235 0 33741 0 35975 572.9 611.5 1184 5 01725 0 5950 13675 558 5 572.5 1246 26 0 02249 032429 034678 5133 604.5 11821 0 7775 0 5859 13634 572.s 576.8 1285.74 0.02264 0 31162 033426 5831 *972
, 1180.9 0.7825 0 5766 13592 575 8 50s.8 1326 17 0 02279 0.29937 0 32216 589.! 589.9 21790 03876 0.5673 13550 500 0 504.0 13671 0.02295 0.28753 031048 594 6 58?.4 1176 9 019?7 0 5580 13507 5848 508.8 1410 0 0 02311 0 27608 029919 600.1 5741 1174.3 01978 0.5485 1.3464 588 I P
592.0 14533 0 02328 026499 0 28827 6051 5668 !!72.6 08030 0 5390 13420 592.5 596.8 1497.3 0.02345 0 25425 017770 611.4 558.8 1170.2 0 8082 0 5293 13375 596 8
Y m Abs Press. Specific Volume . Enthalpy Entropy Temp lb per , at. S Sat. Sal. Sat. Sal. Sal. Temp fahr SqIn. Liquid Evap Vapor Liquid Evap Vapor liquid Evap Vapor Fahr 1 p vi vrg vg hg hgi hg sg sig sg t 0.02364 0 24384 026747 617.1 550.6 11673 0.8134 0.5196 1.3330 50s.e 800.0 15432 0.8187 05097 1.3284 584 8 15891 0 02382 0 23374 0.25757 622.9 5422 1165.1 584.8 1162.4 0 8240 0.4997 13238 50s.8 0 02402 0 22394 0247 % 628.8 533.6 500.0 16373 0.8294 0.48 % 1.3190 512.8 1686.1 0.02422 021442 0.23S65 634.8 5241 1159.5 512.8 1156.4 0.8348 0.4794 1.3141 515.8 0.02444 0.20516 0.22960 640.8 515.6 515.5 1735.9 e02466 0.1 % 15 022081 646.9 506 3 1153.2 0.8403 0.4689 1.3092 528.8 525 I 1786.9 0.3458 0.4583 13041 524.8 0.02489 0.18737 0 21226 653.1 406.6 1149.8 524.8 1839 0 0 02514 017880 020394 659.5 4861 1146.1 0 8514 04474 1.2988 528.8 528.0 1892.4 0.8571 0.4364 1.2934 532.0 0.02539 0.17044 0.19583 665.9 476 4 1142.2 532.8 1947.0 0 02566 0.16226 0.18792 672.4 4653 1138.1 0 % 28 0 4251 1.2879 535.8 535.0 2002.8 0 02595 0.15427 0.18021 679.1 454.6 11333 0 8686 0.4134 12821 540.0 540.0 2059.9 0.8746 0.4015 1.2761 544.0 0.02625 0.14644 0.17269 685.9 443.1 1129.0 544.8 21183 0.8806 03893 1.2699 648.8 0.02657 0.13876 0.16534 692.9 431.1 1124 0 548.8 2178 1 0.8868 03767 1.2634 552.0 22392 0 02691 0.13124 0.15816 700 0 4181 11181 552.8 0.8931 03637 12567 555.8 0 02728 012387 0 15115 707.4 405.7 1113.1 555.0 23011 gas.g 23651 0.02768 0 11663 0.14431 714.9 392.1 1107.0 0 8995 03502 12498 800.5 2431.1 0.02811 0.10947 0.13757 722.9 377.7 1100.6 0.9064 03361 12475 554.5 554.B 0.9137 03210 1.2347 558.8 0 02858 0.10229 0.13087 731.5 362.1 1093.5 558.8 2498.1 0 02911 0 09514 0.12424 7402 345.7 1085.9 0 9212 03054 1.2266 572.5 572.8 2566.6 2636.8 0.02970 0 08799 0.11769 7492 328.5 1077.6 0 9287 02892 1.2179 575.8 575.0 1068.5 0.9365 02720 1.2006 ges.3 ges.O 2708.6 0 03037 0.08080 0.11117 758 5 310 1 0.03114 0.07349 010463 7682 290.2 1058 4 0.9447 02537 1.1984 584.0 584.8 2782.1 0.03204 0.06595 0 09799 778 8 2682 1047.0 0.9535 02337 1.1872 586.9 500.5 2857.4 0 03313 0 05797 0 09110 790.5 243.1 1033 6 0. % 34 0.2110 1.1744 592.8 092.8 2934.5 0 03455 0 04916 0 08371 804.4 212.8 1017.2 0.9749 0.1841 1.1591 595.8 095.0 3013.4 30943 0.03662 0 03857 0 07519 822.4 1723 995.2 0.9901 0.1490 1.1390 798.0 798.0 0 03824 0 03173 0.06997 835 0 1441 979.7 1.0006 0.1246 1.1252 732.0 782.8 3I355 1.0169 0 0816 1.1046 794.8 0 04108 0 02192 0 06300 854 2 102.0 956.2 704.8 3171.2 0.04427 0 01304 0 05730 873 0 61.4 934.4 1.0329 00527 1.0856 795.8 795.8 31983 1.0612 0 0000 1.0612 795.47* 32082 0 05078 0.00000 0 05078 906 0 0.0 906.0 795.47'
- Critical temperature
a O Table 2: Saturated Steam: Pressiure Table Specific Volume Enthalpy Entropy Abs Press. Temp Sat. Sat. Sat. Sat. Sat. Sat. Abs Press. Lb/Sg in. Fahr liquid Evap Vaper Liquid Evap Vapor Liquid Evap Vapor Lb/Sq In. p t vg v ,g vg hg hrg h 8 St 5f8 5 8 P I.00085 32.018 0 016022 3302 4 3302 4 0 0003 10753 10753 0.0000 2.1872 2.1872 3.00085 325 59 323 0 016032 1235 5 12355 27382 1060.1 1087.4 0 0542 2 0425 2 0967 3 25 930 79 586 0 016071 6413 6413 47.623 1048 6 10963 0 0925 1.9446 2.0370 8 50 1.8 10134 0 016136 333 59 333 60 6913 1036.1 1105 8 0.1326 1.8455 1.9781 1.8 5.s 16224 0.016407 73 515 73 532 13020 1000.9 1131.1 0 2349 1.6094 1.8443 53 10 0 193 21 0 016592 38404 38.420 16126 982.1 t143 3 0 2836 13043 13879 13.8 14595 212.00 0 016719 26382 26399 180.17 970 3 11503 0 3121 1.4447 13568 14.595 15.s 113 03 0 016726 26 274 26 290 181.21 %9.7 1150.9 03137 1.4415 13552 15.0 23 3 227.96 0 016834 20070 20.087 1 % 27 9601 11563 0 3358 13 %2 13320 28.g 30.g 250 34 0.017009 133266 133436 218.9 945 2 1164.1 0 3632 13313 16995 30 s og g 26725 0.017151 10 4794 10 4%$ 236.1 933.6 1169 8 0 3921 12844 1.6765 40 8 5e e 281.02 0017274 8 4967 8 5140 250 2 923 9 1174.1 0.4112 12474 1.6586 5e O Eg.g 29211 0 017383 7.1562 7.1736 262.2 9154 1177.6 04273 1.2167 16440 50 0 Je g 302.93 0 017432 6 1875 6 2050 2723 907.8 1180 6 0 4411 1.1005 1.6316 Ta g 30.g 312.04 0 017573 5.4536 54711 282.1 900.9 1183.1 0 4534 1.1675 1 6208 Og g 30.3 320 28 0 017659 4.8779 4.8953 2903 894 6 11853 0.4643 1.1470 1.6113 Se e 133.0 327.82 0.017740 4.4133 4.4310 298 5 888 6 1187.2 0 4743 1.1284 1.6027 100.8 118.8 33439 0 01782 4 0306 4.0484 305.8 883I 1188.9 0 4834 1.1115 13950 lig I 120 g 34127 0 01789 33097 33275 312 6 877.8 1190.4 0 4919 1.0960 13879 120 0 13d e 34733 0 017 % 3.4364 3 4544 319 0 872.8 II91.7 0 4998 10815 13813 138 8 14e B 353 04 0.01803 32010 3.2190 325 0 868.0 1193 0 0 5071 1.0681 13752 140.8 150.3 358 43 0.01809 2.9958 3 0139 330 6 863.4 1194.1 0$141 10554 1.5695 15e g 180.0 363 55 0 01815 2.8155 2.8336 336.1 8590 1195 1 0 $206 10435 I5641 15e e 170.3 36842 0 01321 2.6556 2.6738 341.2 854 8 11 % 0 0 5269 1 0322 1 5591 178 0 130.3 373 08 0 01827 23129 2 5312 346 2 8501 1196 9 0 5328 1.0215 13543 Igg g 190.3 37733 001833 23847 2.4030 350.9 8463 1197.6 0 5384 10113 13498 190 0 293.s 381.80 0 01839 2.2689 23873 355 5 842.8 1198 3 0 5438 1.0016 1 5454 298.g 210.3 385.91 0 01844 2.16373 2 18217 359 9 839.1 1199 0 0 5490 0 9923 13413 213 0 223 8 339.88 0 01850 2.06779 2.08629 364 2 835 4 1199 6 0 5540 0 9834 15314 223 8 230.3 39310 0 01855 1.97991 1.99846 368 3 831 8 12001 0 5588 09748 15336 23e 8 240.8 39139 0 01860 1.89909 1.91769 3723 828 4 1200.6 0 5634 0 9665 1 5299 24e g 250.3 40097 0 01865 1.82452 1.84317 376.1 825 0 1201.1 0 5679 0 9585 15264 25g e 25g g 404 44 0 01870 135548 117418 3 79.9 821 6 12013 0 5722 0 9508 13230 260 e 27s 3 407.80 0 01875 169137 131013 383 6 818 3 1201.9 05764 0 9433 15197 278 e >= 20s.s 411.07 001880 163169 I65049 387.1 8151 1202 3 0 5805 0 9361 I5166 283 s 6 290.8 41425 0 01885 137597 1.59482 390 6 812.0 1202.6 0 5844 0 9291 1 5135 2M s 303 3 41735 0 01889 152384 154274 394 0 808 9 1202.9 0 5882 0 9223 1.5105 380 8 35g 8 43133 001912 130642 132554 409 8 794 2 1204 0 0 6059 0 8909 I4968 35g 8 age.g 444 60 0 01934 1.14162 1.16095 424 2 780 4 1204 6 0 6217 0 8630 14847 4M e
Specific Volume Ecthalpy Entropy Abs Press. Temp Sat. Sat. Sat. Sat. Sat. Sat. Abs Press. P tb/Sg in. Fahr liquid Evap Vapor liquid Evap Vapor liquid Evap Vapor tb/Sg in. $ p t v, v ,, vg ht h ig hg s, s ,g s, p 450 0 45628 0 01954 1.01224 1.03179 437.3 767.5 1204.8 0 6360 0 8378 14738 450 8 SeeB 46701 0 01975 0 90787 0 92762 449.5 7551 1204 7 06490 0 8148 14639 500.0 558 8 47694- 0 01994 0 82183 0 84177 460 9 743 3 1204 3 0 6611 0 7936 14547 550 8 500 0 486 20 0 07013 0 74962 0 76975 4711 732 0 12031 0 6723 0 7738 14461 See s Ele 8 494 89 00?032 0 68811 010843 481.9 720 9 1202.8 06828 0 7552 14381 558 s 7ss e 503 08 0 02050 0 63505 0 65556 491.6 710 2 1201.8 0 6928 0 7377 14304 Ige s 750 0 51034 0 02069 0.58880 0 60949 500.9 699 8 12001 0 7022 0 7210 1.4232 758 I 800.8 518 21 002087 054809 0 568 % 509 8 689 6 1l99 4 01111 0 7051 14163 see e ISO I 52524 0 02105 0 51197 053302 518.4 679.5 1198 0 01197 06899 I4096 35e s 90s O 53195 0 02123 0.47% 8 0.50091 526 7 6697 11 % 4 0 7279 06753 1.4032 set t 950 s $38 39 0 02141 0.45064 0 47205 5341 660 0 11941 01358 0 6612 1.3970 350 0 less 8 544.58 0 02159 0 42436 0 445 % 542 6 650 4 1192.9 0 7434 0 6476 I3910 1900 3 1850 s 550 53 0 02177 0 40047 0 42224 5501 6409 1191.0 0 7507 0 6344 1 3851 1950 s 1100 s 5 % 28 0 02195 0.37863 0 40058 557.5 631.5 1189 1 01578 0 6216 1.3794 tiss e 1150.9 561.82 0 02214 0.35859 0.38073 564 8 622 2 1187 0 01647 0 6091 1 3738 1 58 s 1200.s 567.19 0 02232 0 34013 0.36245 571.9 613 0 1184 8 01714 05%9 1.3683 1200 s 1250.8 572.38 0 02250 0 32306 034556 578 8 603 8 1182 6 0.7780 0 5850 1.3630 1758 8 1308 8 577.42 0 02269 030722 0 32991 585 6 594 6 1130 2 0 7843 0 5733 1.3577 1398 8 1350.3 582.32 0 02288 0 29250 0.31537 592.3 585 4 1177 8 01906 0 5620 1.3525 13588 1400 e 587.07 0 02307 027871 0.30178 598 8 576.5 1175.3 01966 0.5507 13474 lage 8 1458.8 59110 0 02327 0 26584 0 28911 605.3 567.4 1172.8 0 8026 0 5397 1.3423 1450s 1588 8 5%20 0 02346 0 25372 0 27719 6113 558_4 1170 1 0 8085 0 5288 1.3373 15000 155ee 600 59 0 02366 0 24235 0 26601 618 0 549.4 1167.4 0 8142 0 5182 1.3324 1558s 1688 8 604 87 0 02387 0 23159 0 25545 624 2 540.3 1164.5 0 8199 05076 13274 1580 0 1650 s 609 05 0 02407 0 22143 024551 630 4 531.3 1161 6 0 8254 04971 1 3225 165s t 1788 8 613 13 0 02428 021178 0 23607 636.5 522.2 1158 6 0 8309 04867 1.3176 tree s 1750 617.12 0 02450 020263 0 22713 642.5 513.1 1155 6 0 8363 04765 1.3128 II50 s less.8 s 621.02 0 02472 019390 0 21861 648 5 503 8 1152.3 0 8417 0 4662 1.3079 1800 e 1858.3 624 83 0 02495 018558 021052 654.5 494b 1149 0 0 8470 0 4561 1.3030 te m 190s.s 628 56 0 02517 0.17761 010278 660 4 4851 1145 6 0 8522 0 4459 17981 1900 e 1958s 63212 0 02541 0 16999 0.19540 666.3 475 8 1142.0 0 8574 0 4358 1.2931 1958 e 2800.0 635 80 0 02565 0 16266 018831 6721 466 2 1138 3 0 8625 04256 12881 2sse e fles 8 64216 0 02615 0 14885 0I7501 683 8 446 7 IJ305 0 8777 04053 12780 210e e 2280.0 649.45 0 02669 0 13603 0 16272 695 5 426.7 1122.2 0 8828 0 3848 1.2676 2280 0 2308.8 655 89 0 02727 012406 0 15133 707 2 406 0 1113.2 0 8929 0 3640 12569 2300 8 2400.0 662.11 0 02190 0.11287 0 14076 719 0 384.8 11031 0 9031 0 3430 1 2460 2408 8 2500 0 66811 0 02859 0 10209 0.13068 731 1 3616 1093 3 0 9139 0 3206 1.2345 2500 e 2508 8 673 91 0 02938 0 09172 0.12110 744.5 3376 1082.0 0 9247 0 2977 1 2225 2600 0 2 Fee 8 679 53 0 03029 0 08165 0 11194 757.3 3123 1069 7 0 9356 0 274I I2097 2788 8 2000e 684 % 0 03134 0 07171 0 10305 770 7 2851 1055 8 0 9468 0 2491 11958 2808 e 2900 8 690 22 0 03262 0 06158 0 09420 7851 254 7 1039 8 0 9588 0 2215 11803 2900 e 3000 8 69533 0 03428 0 05073 0 08500 8018 218 4 1020 3 0 9728 01891 11619 3000 0 1100 0 700 28 0 03681 0 03771 0 07452 874 0 169 3 993 3 0 9914 0 1460 1 1373 31Be e ' 3288 s 705 08 0 04472 0 01191 0 05663 875.5 56I 931.6 1.0351 0 0482 10332 3280 e 3200.2' 70547 0 05078 0 00000 0 05078 906 0 00 906 0 1Obl2 0 0000 10612 1288.2* N 'Crbi pressure
APPENDIX B SUPEREATED STEAM TABLES B-1
(Intentionally Blank) 3
)
B-2
s Table 3 Superheated Sleam ADS Pttil lemsmave-Deg'Ms f aWM 10/$4 M Sat let ist 001 lest H00 use Sol 1**01 It ater 1:eam Pte 264 306 364 440 die He tot 800 Sa u 76 148 M iW M 248 76 F$tlit M 344 tt les M eos 631 t wli p M les 26 1 nu 6:0,7 een Pw2na74eWeg 26 os 76es toe 9?6 is6sute i M 412 4 as/ at: 1 nn, n te n.3.. t i6 s t, gg; pg,
.e 4 01614n n.3 6H. nd2 pin im g nii, im . no i n il6 nMi i.4a l a t !)M 1 9741 29M # 00d l I ll%2 Ilub 21722 i 1986 # f?J? 21700 2 3l64 23%I IMM let 18W f otes 93? P6 103f ?6 la 37 76 4? ?6 137 M 187 M 23? ?6 19? ?4 33f M 431 76 63'
!?61% I M?6 D663776 1600173? 161MWB37 1966 '6 les 79 19? ?0 I
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t 4 Med . M35 , W12 1 l?M i 6044 1631< 16?tt i 2232 1 7630 140c3 1 43 % 1 M90 I 9c10 1 9316 i M10 j la 12 ll 82ll 3265 182 6l Pt? ll M2 66 442ll 64265 642 65 782ll 88266 942 65 1042 65 W e t SitH I M27 16fM I?MS 4483 2 0044 2tM) 2 4407 3 6609 I tsu 3 M43 !?st 3 4?28 3 6746 3 8'64 4417 J$1 6 M3 99 !?S29 82M 7 !?%7 7 .20' l }316 2 IMat 1421 3 14736 15?6 2 15734 lu)3 16 40 1743 4 17999 e 4 5482 14105 1 1361 I S?03 . 60C3 3 6274 ll?S8 t ill2 I Mt! ! ?tb4 18317 IM62 l et?2 1 9278 1 9677 Sn 29 64 Pt M 71 W 79 64 f ?1 W 3't M e ?t W $79 W 679 W ??l W 8't 64 979 W 1979 W l'8 e 8 41SW l 48M 1 5763 , 7044 8233 .SM3 ) Ii?0 2 3400 2 M38 2 76M itus 3 162) 3HW 35%) 3 Mot d20 M, 6 Je ?Jo 1203 i !!24 i ,7% 6 .706 ) .314 % the4 14?0 9 14732 I5M P. li'l 2 1633 I IW78 !?O 3 !?M 4 6 k H20 1 6076 1 5311 l MS) SMO i L233 16719 ! ?!k) I?M3 1 7927 1 8280 IMin l#1M 1 9241 19tM th 7669 76 69 ' M 69 ,?t H !?llt 374 69 4?l 69 $4 H 6'649 ??llt $4 69 97649 llM H EUI e 9 01999 1 4440 1 6207 I H62 PU) 8'?$ J ot23 12W) taci 16774 14100 30c8 SM38 3 4438 3 6332 H23331 4 400 S3 iM34 1722 l !?%$ 2 78$ 3 .313 7 IM78 1420 1 14729 lim 4 1678 9 1632 9 4647 6 1743 1 4 79% 3 4 4S9M LSM8 1 5261 13412 1 H18 8 6I92 1 ten I 7116 1 ?$16 1 1990 1 4243 8 86?1 1 4091 1 5206 I DSOC la 'il4W4 42 F3 82IIf?OM 312 73 82 273 42 373 t? 473 82 173 12 67342 773 62 4?) 12 til 82 10?!12 MI e 0 91903 4048 I Hil 81M t elW 22132 tom 4 t itM 37 28 IM u 3154 33Mt 3 S227 i M?618 e 403 ?0 .M36 ??0 9 !?wl !?W e 313 4 IJ67 3 leM O 14725 lim 3 15'8 7 1632 7 16876 17429 1791 7 ' e 83991 3021 ,1213 i$%8 i M'8 6113 1 M43 170?$ 1 7440 t ?tu 4 8204 14%4 I seu ! 0171 4kW L:s till T161 121 81 171 91 fil et 371 91 47191 $7101 671 01 77141 971 01 tilat 197101 . WB e 1 81900 1 3680 14191 1 999 1 911 !?MI 19%f 2140 23333 25175 ! ?oJO IMit 3 Mll a fect 3 4!M I l 1428 996 6 SE 80 IM38 U19 2 iM28 !?tl e 1312 2 iM67 1419 4 14722 lim t 15'84 1632 $ 164' 8 170 8 17990 t 6 6026 l ette WH 13h3 IS&M 16ll4 I M06 ! ?0ad IPut ! ?t?0 1 8174 1 8610 l at3! [nst 1 9432 an 19 27 64 f? 118 77 64 ?? 768 t? Ma t? 464 27 M827 06827 764 t? 068 ?? 96427 10W ff thi e 9 01912 1 12 % 137?$ 1 4913 I WC2 . 7029 1 9970 2 0612 !?H? 3 4445 2 6?ll !?tto It7M 3 14)I 3 3706 43173) a 409 63 !?os 0 12176 It%I l lit 4 311 4 13e4 2 14It i 14714 It?4 7 litt 2 1u24 IW7i It42 6 iPM l 6 6 6059 44M8 1 4119 1 5443 1 5717 6077 1 6671 i ?tM 1 7411 17787 13141 IW77 l a'98 1 9106 I WOO la 19 59 6559 ' 15 59 ' 6S H 26S H 36559 4H lt M5it 665 59 76569 Mill 965 H IMlM tal e 4 91917 tell 13?tt l eave M21 H?) I k21 3 023? 1 7001 2J7% 7 982 7 7tM 2 00t4 30%2 3 ?? ?) M)d ell n 412 81 704 1 !!!n l IMc3 791 6 110 6 IM56 14:07 1471 5 1174 4 1571 6 16 # 3 [686 9 1742 6 l798 8 8 4 6062 ,4943 4 M73 I heI . S?St .6040 4 8SM i M76 3 7329 The i AIO6 iSeen 48'u i 9673 4the Sn 19 ft to M 18 M ,MH 60 M 260 19 5039 4H39 let M ese 79 760 M Me M 968 M 1000 M tot e 30145 ??ll 12472 13E4 406 1 7410 1 9139 2 0th !)eed 74124 ilFM 27M4 l et?) 3 0S F2 MJS 61) t 418 H 204 4 1212 4 1247 7 .2 79 l .m 0 1%4l 1417 9 1470 8 1123 I in?? 4 1631 i !W65 !?a2 i 17t4 6 4 4 61M , 44t4 4 Ht2 1SMO i ht3 i SMS I 6470 t illa 1 F316 !?H2 4 0047 1 &3W 1 4 ?06 8 9011 1 U07 Sh a superheat F h = entupy Btv per tb i t = SptCifst 90kme, tu fl per it) 5 = (Mropy,6ty per R pef Ib B-5
Table 3. Superheated Steam-ConNaved b on Sat Set Tenperature - Cet' eel f ahrenhe<1 (1st fe+pl matee $ team ett $N let 480 450 3W M M 1N0 11W llW 13W 1400 1600 b S 40 $s 40 05 40 15s 40 Ms 40 ts540 35540 4H 40 66540 6H 40 PH 40 pH 40 9H 40 185 40 480 e 90llM lillt 31734 1 7944 .3834 14?63 8 % 46 1 6491 t illl 1959 18331 2 ?tcl 2 44W 21987 2 Mll itCJ 7 4444 601 4 42417 I?04 6 !?c8 8 124$ 1 .2776 IMP 4 13M t IM34 1417 0 14708 1573 3 1674 9 1631 2 16M P 1741 9 !?ts ? 0 6267 l4MP l aste i ntt? Mll I hcl 1 6163 1HM 1 6450 1756 1 1632 !?M8 1 43?) I M47 1 9966 19?% b 60 to 60 00 60 50 60 70C 60 750 W JSC 60 4W 60 SW W 00 W ?$c to SW W 950le 10$0 60 434 e 0 0lW2 11057 1 1071 1 2448 3113 . 4JC 7 1 44 % I M76 1 7758 1 8795 28%4 t i?th 2 3??) 24739 # 41M ? ?647 tut 401 6 4?t M I ?08 7 1705? 1742 4 7764 30$ 8 13M S 1M23 1416 7 1469 4 IS?? 7 l$76 4 1008 164$ 8 1741 6 1794 0 0 G W76 1 4402 1 4408 1 5206 . lW2 , hth 1 6100 16M6 16791 1 7197 1 7673 I ?t32 1 069 18HI 1 8499 l l;tt
> 4517 tit? 4197 95 97 74597 Mll? 44597 $46 97 64547 74$ l7 Nlt? Wit? 164l97 441 e 9 01950 10%4 11517 1 2454 : His 4138 14 t?6 1 H45 I ?t:6 ItM) 2 0'90 22203 iMOS 2 at98 2064 44W C31 6 4M 77 I?04 8 12M 7 12734 Ms? 333 7 1341 1 1415 3 1464 7 lit? l 1P59 16304 16as t 1741 2 17177 6 6 6332 34159 1 6132 1 S474 ,6772 6040 4 6?66 167H I flat 1 7621 1 18?8 8 8216 i 8438 1SMF i 9143 Sh 41 50 91 % 41 50 91 50 241 50 MI SO 441 50 641 50 641 50 741 50 kl 54 941 50 1041 50 488 e 4 019H 1 0092 l et39 118$2 tall M82 14142 1 D03 1 7117 10%4 I H72 21726 27%9 !)tC3 2 U30 84W 101 6 0983 !?04 8 12M t 127 3 M26 .331 8 IM00 1414 4 1480 1916 16 '$ 4 16?tt 164$ 1 174c t 17974 8 46M7 44748 130W ! $409 , Ull ,3982 16D0 l uto I ?0tt 1 7469 178M 1814 1 8448 18'97 iKH de Sk if18 8718 3718 87iB 237 Ii 33718 43711 53711 437il 737I8 83718 93718 1037 II e 0 0lM7 O M64 l edet 11300 2115 1984 I Mil ! $02) 1644 1 7716 l tC30 2 0330 11619 2 7W 2 417) 942sh t 444 75 !?Qs 8 UMi 1769 I 300 0 DOS 1%Il lill t 1867 3 IU09 1674 9 16?t t 164d ? 1740 6 17977 8 6649 L 477 1 4990 LAM 6 , Ml? I lth 4 6176 i M28 1 1938 1 7410 1 7777 1 8116 iMM I 8748 1904$
b 32 99 82 H 32 99 82 H 232 99 332 99 432 99 132 ft 632 99 732 99 432 H 4? 99 1037 M tu e t elPS 09??6 0 9')19 19791 lud Ju? 1 3037 14797 1 908 16992 18?% 1927 20746 J it?7 2 3?00 1667 013 6 449 S? !?os 7 1731 2 IM70 ?tt I !?t i 1367 7 le12 7 leu 6 IS?0 3 1674 4 16?9 1 16ua f eo 3 l FM t t 06490 146M i 4921 1 pts 9 96 5871 1 6123 1 6678 16990 1 7378 1 ??)0 8 8069 ISM 3 842 1 9998 D 7893 78 13 'ftH 7813 728 93 3?l H 4?B 93 t?t 93 628 H ??l H E813 SM 93 1878 93 170 e 0 01982 6 8914 0 9464 ISMI 10W 1816 12%4 1 3819 l bat $ 1 632) 17WP 1 8746 1 9440 glin 2 Dc7 It ?! 8?) t ebe 18 I?Os 6 !??S 3 !?W 8 PS 7 4 327 ? 13 % 6 1411 0 1465 9 llit 7 1U39 16?t 7 1644 0 1740 0 !?M ? 06M6 1 4401 14M3 I U23 Sul .68;8 8 6072 1690 16943 173?$ 1724 i M24 ISM 8 i M$ 7 ithe h to 99 74 H ' 24 95 74 91 724 H 324 99 424 H 524 H 624 95 724 H 824 99 U4 99 10?4 99 680 e 4 01990 6897 0 9045 0 9064 06 4 IM2 I ?ol0 3?64 1828 1 904 1 6440 1 042 lelu 283M 2 1471 e476 011 6 458 71 I?M 4 12M 3 !?u l 7tg 7 D63 Ins) 410 9 lea l Ints l I$f 3 s 1U12 1683 6 I?M 7 17964 s $ 658 r 14WS 14766 1 6164 6445 U67 1 60:3 1 983 4640 ! ?260 ! ?640 I ?tti 1 &Jo$ 1 8415 1 9918
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> 1743 6743 life) 6743 21743 317 4) 41743 617 0 61743 71743 5174) 91743 1917 43 tm : 0 0?006 4 M71 e a?lf 0 9100 0 6424 Otu a ll?6 12nd i 34 7) 14%) I us) 1 4'50 i tan i es?! I Mt:
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$h 46 32 M 12 144 (? IM 92 796 if )N12 4M t? M6 92 SM t? ?M 97 $W 92 9% t?
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> 1805 68 05 118 05 1680% 768 0% M8 ti MtM M4M 068 M M4 05 064 0% M4 06 W e 0 82173 O $009 $1263 0 6469 06M8 6 64s4 87?'l 0824 0 5762 4 9tus 8 0'20 1 8430 8 2131 12t?$
6 31 951 6 W6 70 1IM 4 till 6 !?to 6 !?tt 6 13J? 7 1394 6 Hs: ? its i stes 4 M>l 6 16774 17M i !?ti 6 L7279 8 4432 14223 iout 1 W10 1331l t 6427 1 06) &tM/ 1 7933 1 7382 4 7713 1 80?8 8 E3?t
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8 6440 9 latt0 e7414 llM S0 1614 7174 2 6 1671 4519 6 ??t e j ft? 4 64711 t 57186 IlW S FH1 : ?t96 5 67h4 1 Mal I4al iett 148$1 111W l Sell i M54 1 5443 1 6714 Iwt I tt35 i F371
> !? S4 72 44 122 54 172 44 722 58 772 $8 322 54 472 54 %?? 54 622 54 772 $4 122 S4 t?? W 1380 e 8 0?Mt 03?tt O MPO 0 40t2 tust 04 04 Olitt 0 5434 0 5 ??t 04?I7 0 64 2 07Ml 07M7 04Mn e as4 an 4t
- Set 58 IIE t !?ot t !?61 9 13C3 9 IMO9 8374 4 14164 14371 14963 inM 3 8817 0 lutt $7?? t 1766 3 8 $?pl 13S7 F 43E0 14M0 14hl I W:7 1 5216 llM4 16773 141H I H70 tilD 1 7279 IMM 1 7902 ,
412 9 117 93 Sn 17 93 6293 117 93 16293 fif t) 762 93 312 t3 41793 512 93 617 93 712 93 0 77U)08?M less e 0 6 E7 0 30lt $ 3176 OM47 t est 0 4430 0 471t CMM 0 6?tt 0 $409 06hl 04798 S Pt?{ 6870?l 6 lett) 1176 3 IlW I l?S14 429t l 13M S IMt3 1432 0 14H 2 lau f IHil IWet 16u s {7?63 17110 t t h=4 i M?4 1302 14181 isb?) 18900 1 6182 1 6436 i M!O 1 60 % l H44 16NS 171st s?We I ?llt
> 3 80 %3 60 103 to ill 60 703 M TS3 to 303 to 403 80 543 K 6C3 l' 703 90 Elte 903 K that e 6 00 M4 027M 01176 2120 OlVI tirl) 04Mt 049 satts Satu etna twH $6Mf $477) 0 410 0 76M 3 !?o t us't p?t 9 }%40 }M74 14:17 14 % I 1992 16c? ? 1646 r [440 17837 0% 20t t 611 64 1170 )
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$h = luperheet i ti = enthaty Btw pet tb 8 = entrepy. Stu ptf R pef lb 9 = Specific volume, tv tt pet Ib l
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SECTIONbbE Principles of Nuclear Power Plant Operation Thermodynamics, Heat Transfer, and Fluid Flow
- Question 1.01 (1.0)
Multiple Choice (Sc]ect the correct response.) The Required Net Positive Suction Head (NPSHg ) of a pump is the minimum suction head required to prevent cavitation in a pump. Which of the following conditions will cause HPSH R f r the main feedwater pump to increase?
- a. Decreasing main feedwater pump turbine speed.
- b. Increasing condenw;.te system pressure.
- c. Increasing main feedwater pump discharge flow.
- d. Decreasing condensate system temperature.
- ANSWER 1.01 (1.0) l c.
- Thermal-Hydraulic Principles and applications Chapter 10, Pages 10-55 to 10-61.
191004K115 1 I l
- QUESTION 1.02 (1.0)
Multiple Choice (Select the correct response.) One of the steam generator safety valves has lifted at a steam generator pressure of 1040 psig. Use the steam tables provided. What idill be the temperature of the steam when it reaches atmospheric pressure?
- a. 551 F
- b. 549 F
- c. 305 F l
- d. 296 F
- ANSWER 1.02 (1.0) c or d
- REFERENCE Thermal-Hydraulic Principles and Applications, Chapter 2.
KSA 193OO4K115
- QUESTION 1.03 (1.0)
Multiple Choice (Gelect the correct response.) A centrifugal pump is being driven by a two (2) speed motor. The pumps speeds are 1200 and 1800 RPM. By what factor will the pump head change if the pump speed is increased from 1200 to 1800 RPM?
- a. Pump head will increase by a factor of 1.5.
- b. Pump head will increase by a factor of 0.67.
- c. Pump head will increase by a factor of 0.45.
- d. Pump head will increase by a factor of 2.25.
- ANSWER 1.03 (1.0) d.
- REFERENCE Thermal-Hydraulic Principles and Applications Chapter 10, Pages 10-37, -38, -39.
191004K115 i 1
- QUESTION 1.04 (1.0)
Multiple Choice (Sc1cet the correct response.) A centrifugal pump is being driven by two (2) speed motor. The pump spcods are 1200 and 1800 RPM. By what factor will power change if the speed of a two (2) speed motor is reduced from 1800 RPM to 1200 RPM?
- a. Power will decrease by a factor of 0.30.
- b. Power will increase by a factor of 0.30.
- c. Power will decrease by a factor of 0.44.
- d. Power will increase by a factor of 0.44.
- ANSWER 1.04 (1.0) a.
- REFERENCE Thermal-Hydraulic Principles and Applications Chapter 10, Pages 10-37, -38, -39, 191004K115 i
I l
- QUESTION 1.05 (1.0)
Multiple Choice (Sclect the correct response.) Which one of the following conditions will cause the Available Net Positive Suction Head (NPSH ) of a centrifugal 3 pump to increase?
- a. Decreasing the level of the storage tank being filled.
- b. Decreasing the level of the storage tank being pumped from.
- c. Increasing the flow rate by opening the discharge valve.
- d. Decreasing the temperature of the fluid being pumped.
- ANSWER 1.05 (1.0) d.
- REFERENCE Thermal-Hydraulic Principles and Applications Chapter 10,
~ Pages 10-54 through 10-61. 191004K115 l
- QUESTION 1.06 (2.0)
Unit 1 is operating with all steam generator pressures at 785 pais, a feedwater temperature 7of 432 F and a total feedwater flow rate of 1.45 x 10 lbm/hr. Rated thermal power for unit 1 is 3338 Mwt. Using the steam tables and the Mollier diagram answer the following questions. Without considering blowdown: What is the actual percent power of Unit 1 with the secondary plant conditions stated above? (2.0) uANSWER 1.06 (2.0)
- a. Q = A Zih, [Q = $ (h steam -hfeedwater)] (0.5) h steam (800 psia) = 1199.4 BTU /LBM, (0.25) h feedwater (432 F) = 410.1 BTU /LBM (0.25) 7 Q = 1.45x10 LBM/IIR (1199.4 BTU /LBM - 410.1 BTU /LBM) (0.25)
Q = 1.45x10 LBM/IIR (789.3 BTU /LBM) Q = 1.144x10 10 BTU /IIR (0.15) 10 0 Mw = 1.144x10 BTU /IIR (1Mw / 3. 41x10 BTU /IIR) (0.25) Mw = 335G.3 Mw (0.1) 335G.3 Mw / 3338 Mw = 1.005 or 101% 4 /- 2% (0.25)
- REFERENCE Thermal-llydraulic Principles and Applications, Chapters 2 and 12.
103007K108
- QUESTION 1.07 (2.0)
Matching The attached figure 1.07 is a simplified temperature - entropy diagram of the Diablo Canyon Power Plant steam cycle. The numbered points on the figure are parts of the stcom cycle which identify a change in phase or the condition of the process fluid, steam or water. Match each System / Component action (a through d) with its correct number range Choice, given below, that corresponds with its associated cycle process identified on figure 1.07. (0.5 points for each correct choice.) System / Component Action:l Matching Response l Choices l l l l
- a. Work out LP Turbine l l 1 to-2 l l 2 to 3 l l 3 to 4
- b. Reheater Superheating l l 4 to 5 l l 5 to G l l G to 7
- c. Vaporization in Steam l l 7 to 8 Generator l l l l l l
- d. Condensing l l l l
- ANSWER 1.07 (2.0) (0.5 points each maximum 2.0 points)
- a. 7 to 8
- b. 6 to 7
- c. 3 to 4
- d. 8 to 1
- Reference Thermal-Hydraulic Principles and Applications, Chapter 7, Page 7-91.
193008V,101
FIGURE 1,07 : O 1 SL 3 e / 5 i 7
/
lk
,--------______1\ \i 1 ,
(@] '
/
h ENTROPY 1 S 9 e-
I l I
- QUESTION 1.08 (1.5)
Multiple Choice (Sclect the correct responso.) Delayed neutrons are less than one percent (1%) of all fission neutrons, but are significant in operating and controlling a reactor. Which of the following statements correctly describes the average delayed neutron fraction cause and effect on reactor operations with respect to core life?
- a. The average delayed neutron fraction decreases over core life causing reactor transients to be faster at the end of core life.
- h. The average delayed neutron fraction decreases over core life causing reactor transients to be slower at the end of core life.
- c. The averago delayed neutron fraction increases over core life causing reactor transients to be faster at the end of core life.
- d. The average delayed neutron fraction increases over core life causing reactor transients to be slower at the end of core life.
- ANSWER 1.08 (1.5) a.
- Reference Fundamentals of Nuclear Reactor Physics, Chapter 7, Pgs. 7-5 through 7-38.
192003K107
- QUESTION 1.09 (1.0)
Multiple Choice (Scicct the correct response.) The count rate of a reactor that has no fuel in it is 150 counts per minuto. After adding 10 fuel assemblics the count rate increases to 600 counts per minute. What is the multiplication factor (Keff) for the count rate increase?
- a. 0.25
- b. 0.30
- c. 0.65
- d. 0.75
- ANSWER 1.09 (1.0)
- d. (Keff = 1 - C, / C(10), 1- 150/000 = 0.75)
- REFERENCE Fundamentals of Nuclear Reactor Physics, Chapter 8, Pgs. 8-18 through 8-28, 192003K102
- QUESTION 1.10 (1.0)
Multiple Choice (Select the correct response.) The secondary neutron source is used as the only installed neutron source after the initial fuel cycle. What statement below DOES NOT correctly state the purpose of installed neutron source assemblies?
- a. To provide a base neutron level to insure an orderly and controlled approach to criticality.
- b. Used to check radiation monitors on the refueling floor during refueling operations.
- c. To provide a means to monitor reactivity changes in the core when the reactor is shutdown.
- d. To provide for verification of the proper operation of nuclear instrumentation.
- ANSWER 1.10 (1.0) b.
- REFERENCE Fundamentals of Nuclear Reactor Physics, Chapter 8, Pgs. 8-3 through 8-12, 192003K111
- QUESTION 1.11 (1.0)
Multiple Choice (select the correct response.) The Doppler temperaturo coefficient is the rato at which reactivity changes with fuel temperature. Which of the following factors would be tho predominate reason for the Doppler Temperature Coefficient becoming LESS NEGATIVE over core life?
- a. The half life of samarium and its buildup in the core.
- b. The decrease in boron concentration over the life of the Core.
- c. The increasing strength of the installed neutron sources under power conditions.
- d. The offect of fuel cladding creep which causes the fuel clad gap to decrease.
- ANSWER 1.11 (1.0) d.
- REFERENCE Reactor Core Control, Chapter 2, Pgs. 2-40 through 2-40.
i i
- QUESTION 1.12 (1.5)
Multiple Choice (Select the correct resp'onse.) How much reactivity (in pcm) must be_gdded to the Unit I reactor if it is just critical at 10 amps and you are to take it to 3% power at a startup rate of 0.75 DPM. Detagff(/S) = 0.007, Lamda (h) = 0.1 sec'I , j# = 2x10 -5 sec.
- a. 0.156 pcm
- b. 1.56 pcm
- c. 15.6 pcm
- d. 15G pcm
- ANSWER 1.12 (1.5) d.
(T = 2G.0G/SUR = 2G.06/0.75 = 34.75, p=,(*/T4 Beta gff / 14 Lamda x T p = 2x10 -5 /34.75 4 .007/ 1 4 0.1 x 34.75 p = 0.0015G ZiK/K p= 156 pcm)
- Reference Reactor Core Control, Chapter 6, Pgs. G-3 throuGh G-10.
001000K547
l
- QUESTION 1.13 (1.5)
Multiple Choice (Sc3 cot the correct response.) The Unit 2 reactor is shutdown (mode 4) by a calculated
-4.75% AsK/K. All rod control cluster assemblies (RCCA's) are fully inserted.
What is Keff for the Unit 2 reactor when it is shutdown by
-4.75% 21 K/K?
- a. 0.83
- b. 0.95
- c. 0.97
- d. 0.98
- ANSWER 1.13 (1.5)
~ b. (2h K/K = -4.75 = Keff - 1 / Keff 1 = Keff + 0.0475 Keff Keff = 0.95)
- Reference Fundamentals of Nuclear Reactor Physics, Chapter 7.
001000K508 i
- QUESTION 1.14 (1.0)
Multiple Choice (Sclect the correct response.) The unit i reactor is shutdown by -4% /1 K/K, with all RCCA's fully inserted. The most reactive RCCA is worth 2% dsK/K. What is the shutdown margin of the reactor in % /1 K/K?
- a. 1.6% /1 K/K
- b. 2.0% dsK/K
- c. 2.4% AsK/K
- d. 4.0% dsK/K
- ANSWER 1.14 (1.0) b.
(4% /hK/K - 2% esK/K = 2% Zi K/K)
- Reference Technical Specification 3/4.1.1, Pg. 3/4 1-1. Reactor Core Control, Chapter 7, Pan. 7 7-23.
192002K113
- QUESTION 1.15 (1.0)
Multiple Choice (Sc1cet the correct response.) Which of the following statements concerning Axial Flux Difference (AFD) is correct?
- a. When power is distributed equally throughout the core the AFD is equal to one (1).
- b. When AFD.is negative more power is Deing produced in the top of the core.
- c. As power level increases cooler water enters the bottom of the core causing the AFD to be more negative.
- d. At beginning of core life (BOL) axial power distribution peaks in the top of the core making AFD more negative.
- ANSWER 1.15 (1.0) c.
- REFERENCE Reactor Core Control, Chapter 8, Pga. 8-25 through 8-29.
001000K553 1 l l 1 l l
- QUESTION 1.16 (0.5)
True or False At the end of core life the Doppler coefficient becomcc lens negative due to the' increased absorption of neutrons by Plutonium 240.
- ANSWER 1.16 (0.5)
False
- Reference Reactor Core Control, Chapter 2, Pga. 23 through 30.
001000K549 s l
- l 1
l l
- QUESTION 1.17 (1.5)
Multiple Choice (Select the correct response.) Figure 1.17 is an illustration of the behavior of Xenon as a result of step power changes in a reactor. Which of the following statements correctly describes the reason the Xenon concentration decreases at point A on figure 1.17?
- a. The xenon poison build-in is from increased neutron absorbtion rate at increased higher powcr levels and a decrease in the production of iodine,
- b. The xenon poison burn-out is from the decreased neutron absorbtion rate at increased higher power levels and a decrease in the production of iodine,
- c. The xenon poison build-in is from the slow decrease in the production of iodine and the build-in of xenon directly from increasing neutron flux.
- d. The xenon poison burn-out is from the slow increase in the production of iodine and the burn-out rate of xenon directly from neutron flux.
- ANSWER 1.17 (1.5) d.
- Reference Reactor Core Control, Chapter 4, Pgs. 4-20 through 4-27.
192000K106
FIGURE 1.17 100 - - POWER 75 .. (%) 50 .. 8 I C e I I I 8 25 . . l , 8 i i l l ' I l 0 - a 0 20 30 60 8d l'00 s i 8 . I e l l l l l 1 ' i s ! e i
, s 0
)
-6000 . . g i I 8
l 3 I I I l
' I
-5000 - - l s l l XENON : I,
-4000 . .
g g REACT m TY , g (PCM) i A i
-3000 -- l ;
i
,l l 1
-2000 . .
l l l 1 I i i i l
-1000 .
l ,, i e i I a 1 0 ! ! I. . 0 20 40 60 80 100 TIME (HOURS)
- QUESTION 1.18 (1.5)
Diesel generator 1-1 is operating in manual and is paralleled to the grid on Bus H for testing. Which of the following statements correctly describes the response of control board indications if you lower the voltage control switch?
- a. Line voltage increases, VARS increase, and out-put frequency decreases,
- b. Line voltage decreases, VARS decrease, and out-put frequency increases.
- c. Line voltage remains constant, VARS decrease, and out put frequency remains constant.
- d. Line voltage remains constant, VARS increase, and out put frequency remains constant.
- ANSWER 1.18 (1.5) c or d
- Reference DCPP System Description J-6B, Diesel Generators, Pg. 52 and DCPP OP-J6B: IV. ;
191005K107 l l l i I
- QUESTION 1.19 (2.0)
Multiple Choice (Select the correct response.) The Unit 2 reactor is operatt-g at 80% power, BOL, boron concentration of 1500 ppm. Assume no Xenon effects and 10 pcm/ppe worth for boron. What is the volume of 12% boric acid required to reduce power from 80% to 40% without changing control rod positions? Use attached figures 1.19-1, and 1.10-2.
- a. 140 gallons
- b. 120 gallons
- c. 100 gallons
- d. 75 gallons
- ANSWER 1.19 (2.0) a or b
- Reference
. DCPP OP-L4 Power Change Worksheet Pgs. 1 and 2 of 2. DCPP Volume 9, Curves and Miscellaneous Data. 004000K506 l l 1
FIGURE 1.19-1 TOTAL POWER DEFECT AS A FUNCTION OF POWER LEVEL AT BOL UNIT : CYCLE 2 FOR BURNUP 0-5000 MWD /MTU 1800 i I' ii ! 'i i !! ! i ; I i!!! ..
. I i i!lI IIl' il'! ll.e illi I i llll '
{ i I:ii ill, lill llll il i til- I!;l llll i i 1400
: iI ' !'I I I3I '
II '! i ' 600 1: li'i ill! ti;i l'li i!il tii it , iill t ./ ii- ii: lili tisi lili !lil !!! 11: 1 till /. :
,i ! i i !! ii, ill 1 il i:l; lli! 'i/i / 900
'I' ' I !
1200 i i
I' ' II I !
ilii i:I i'Ii : !i Iii! !!i . IIei /!;s/: ,/ 1200 lli lill liI; t ill! Illi l!!l Ilil lll/ ll/l 1/
- i: ll Illi lil! !!ll lill 11: lill l yf l fli'f;!lp 1500 :
'!' I' ' ! ' '
1000 ! II! I TOTAL *iii i!** ' 1300
!! ! !Ii' iiI! 'i l i 3/ !/i 8
/i p ', /
POWER : !l4 llil l: ! llll l lt llll Ifj lflt/f gi 1 DEFECT .{ji l;li
!/ l f (PCM) ,,,, ;;;i i;; i ;i ; ;. ;lll Af yjy;jy j, ,
l i;g. : ,, ;i, g,g j ,7 , g. , ii ii!! ,Ii. ' I i.i .. i fi/ j/i f; e_/ ;I! *
'Ill Ili lill lill ilil lIly@ j/lyfi f;li Ill! ! .' i Ill! li ! li' i l/'JMi l/f lill ll! (!
ML li I! '.i 600 ' M i! li i li i ' i;;' II: 1 11 1
- :i M/fMN IN 600:PF ult' '
'I '
II'l i' I i :///AMi N TN 900 PF W
!i iiI' ////fi iWNIN1200 PF W 400 i i;I///// '
i ' i !'Nh1500.PF W'
' iiii i ://///l i !! i! N800 PF W -
I!I Ill ///# ti' Illi li ill' liit !! i i': l l)V///1 Illi l': ! I: ll 11ll tili i' l' 200
! II ' '
!' I iI I'I' '
il ##M i iiti .! i ! '!, tit .. il11Mi l l l llii I::' I . iil. it'i .
'[ l! '
l 1 i: ii'! ,;,; ,I. iil 0 7 '! ' i - 0 10 20 30 40 50 60 70 80 90 100 l l
~
POWER LEVEL (%)
i FIGURE 1.19-2 _ BORATION ll0M0 GRAPHS BOR0!1 ADDITION l 5000 .4 4500 -. i 4000 - l 3000 - 3500 - 1000 l
~
000 3000 - 800 l 2500 - _ 700 l l 2000 - -- 600 ) _ 1750 - - 500 2500 - # 1500 - l 400 e.- 1250 - E 1000 - - 300 l o 900 - l 7- E 800 - - 250 8 E 700 - ~ 200 ^ g 2000 - M 600 - J ! E $ 500 -- g 450 - - 150 ~ l 5 l 5 y 400 - E l g - 350 - ;- o . - 100 -
" 1500 - g 300 - -
90 2* 5 d 250 - - 80 , E
=
3 - 70~ 5 ! 200 - - 60 d 175' - "
- . 150 -
50 a. t 5 1000 - . 125 - - 40
- E
" 100 -
30 75 - - 25 500 - - 20 15 100 - 10 - .- 10 l l l l
- QUESTION 1.20 (1.0)
Multiple Choice (Select the correct response.) Which of the following statements correctly-describes the results of placing the cation domineralizer in the Chemical and Volume Control System (CVCS) in service?
- a. The cation domineralizer adds lithium lithium to the reactor coolant to increase pH.
- b. The cation domineralizer removes lithium from the reactor coolant to increase pH.
- c. The cation domineralizer adds lithium and reduces pH in the reactor coolant.
- d. The cation domineralizer removes lithium and reduces the pH of the reactor coolant.
- ANSWER 1.20 (1.0) d.
- Reference DCPP System Description B-la, Chemical and Volume Control System, Pgs. 23 and 24.
004020KS01 END OF SECTION ONE CONTINUE ONTO SECTION TWO i l l 1 1 l l
SECTION TWO PLANT DESIGN, INCLUDING SAFETY AND EMERGENCY SYSTEMS
*DUESTION 2.01 (2.5)
Matching Item "C" DELETED. For the parts labeled A through F on Figure 2.01 correctly match the letter (A through F) with the list of component names given below. (0.5 pts. each maximum 2.5 points.) Matching Part Name Letter
- 1. Instrumentation thimble guides (neutron detector).______
- 2. Bottom Support Forging. ______
- 3. Lower Core Plate. ______
- 4. Upper Core Plate. ______
- 5. Outlet Nozzle. ______
- 6. Inlet Nozzle. ______
- 7. Rod Cluster Control Guide Tube. ______
- 8. Thermal Sleeve. ______
l
- 9. Control Rod Drive Mechanism. ______ l
- 10. Closure Head Assambly. I
- 11. Upper Support Plate. ______
l
- 12. Core Plate. ______
)
- 13. Fuel Assemblies. ______ l l
- 14. Baffle. ______
- 15. Lower Core Support Plate. ______
- ANSWER 2.01 (2.5) 0.5 pts. each max 2.5 points.
9-A, 11-B, (12-C DELETED), B-D, 7-E, 1-F
- Reference DCPP System Description A-1 Figure RCS-36.
OO2OOOK613
- FIGURE 2,01 CROSS-SECTIONAL CUTAWAY OF REACTOR UNIT 2
- nd b p $ a)j:
., y
, +. U: .
'7 ,,
- 0 4 'n t) iun g #
g
- f. mm 5
' l ll
/.
s b.ht l l F.I r I N g ys
~
e . x ~ b!! If]t i
! FIGURE 2.01 ocoaggyooco ( CROSS-SECTIONAL CUTAWAY OF REACTOR UNIT 2
!$k , i l kI f,,. [. . ,$.5 .i!!f, "[,
3
,a. f 3
[ ,, Control Rod 'gv Drive Mechanism N - pk . Thermal Sleeve Rod Cluster Control Closure Head Assembly ,' ti . i Upper Support Plate % - .
! o-( '3 re Plate g
3 }lf 1
; *-Inlet Nozzle s l lg e h' i
# Outlet Nozzle h
% Upper Core Plate Fuel Assemblies--- n
'l
;t
(!-:: m Baffle 4 3 i . ..l
,:' 9 l Lower Core Support Plate M f ower Core Plate l U l Bottom Support Forging
' ?) Instrumentation Thimble Guides -
Neutron Detector is 1001 1143/T23006.024 RV.20 Rev. 1
- QUESTION 2.02 (1.5)
During normal plant operations the Residual Beat Removal System (RHR) is aligned to act as part of the Emergency Core Cooling System (ECCS). A single line from the source of water supplies both trains of RHR.
- a. What is the normal source of water to the RHR pumps when it is aligned to act as part of the ECCS system? (0.5)
- b. What action, besides valve line up checks, is taken to insure that the valve from the normal source of water to the RHR pumps during ECCS alignment remains open? (1.0)
- ANSWER 2.02 (1.5)
- a. Refueling Water T ak (RWST) (0.5)
I
- b. Power Removed (Motor De-energized, supply beaker opened) or (Verification of valve position on the monitor box or STP done every shift verify position). (1.0)
- Reference l DCPP System Description B-2, Pg. 15. DCPP P&ID 102010 & 1 102009.
005000K106 l 1 I 1
- QUESTION 2.03 (2.5)
The Residual Heat Removal System (RilR)- is placed in service to remove decay heat during phaso II cooldown by opening RHR loop isolation suction Valva's 8701 and 8702. Before openinqc valve g701, for phase II ccoldown, the RCS system must below 475 F and less than 390 psig.
- a. From what reactor coolant syr, tem loop does the RHR system take suction from during phase II cooldown? (0.5)
- b. What is the source of the 475 F signal? ( 0. 5)^
~ . Wnat signal will automatically clcse valve 8702 (source and set point)? (1.0)
- d. What is the source of the 390 psig pressure signal? (0.5)
- ANSWER 2.03 (2.5)
- a. Loop 4 (0.5)
- b. Pressurizer vapor spage temperature (TE-454). (0.5)
- c. Reactor coolant system- loop 4 (PT-403) pressure, (0.5) at 700 psig. (0.5)
- d. Reactor coolant loop 3 or 4 (PT-405 or PT-403) (hot leg) pressure. s (0.5)
- Reference-DCPP system Description B-2, Pgs. 17 and 18.
OO5000K106 1 1 j l I i a
- QUESTION 2.04 (2.0)
The reactor coolant pumps move large volumes of reactor coolant to remove heat from the reactor core.
- a. What are the two (2) 12 Kv busses that supplies power to reactor coolant pumps 1-1 and 1-2, respectively? (0.5)
- b. How does the anti-reverse rotation device used on the reactor coolant pump protect the pump motor? (1.0)
- c. What component of the reactor coolant pump ensures short term continuation of forced reactor coolant flow if all AC power is lost to the reactor coolant pumps si mul taneousl y? (0.5)
- ANSWER 2.04 (2.0)
- a. RCP 1-1, BUS E (0.25)
RCP 1-2, BUS D (0.25)
- b. Over heating of reactor coolant pump motor if the pump is started while rotating in the reverse direction.
(Starting of a reactor coolant pump while it is rotating in the reverse rotation would result in overheating of the motor, due to excessive starting curronts). (1.0) (By the pawls on the flywheel engaging the frame mounted ratchet as the rotor comes to a stop thereby preventing reverse rotation. CO.5 points])
- c. RCP Flywheel. (0.5)
- Reference DCPP System Description A-5, Pg. 20 & 21 and System Description J-5, figure 12Kv.3.
OO3OOOK201
I
- QUESTION 2.05 (2.5)
The Reactor Vessel Level Indication System (RVLIS) is designed to provide the operator with reactor vessel level indication during abnormal and emergency operating events.
- a. What connection does the upper tap for the RVLIS come from? (0,5)
- b. From what two loop connections does the RVLIS loop connections tap off of? (Provide loops (0.25 each) and connection name (0.5), connection numbers not required.)
- c. How are the RVLIS transmitters protected frca the harsh LOCA environment? (1.0)
- ANSWER 2.05 (2.5)
- a. Start-up head vent (Reactor vessel head vent). (0.5) s
- b. Loop (wide range) pressure connection, (0.5)
Loop 3 (PT-405) (0.25) Loop 4 *PT-403) (0.25)
- c. The transmitters are located outside of the reactor containment. (1,0)
- Reference DCPP System Description A-2d, Pgs. 5 through 9.
002000K603 i l l l l
^
- QUESTION 2.06 (2.0)
The Pressurizer Relief Tank (gRT) is normally3 Perated at 81 to 87% full of water, 1600 ft of the 1800 ft
- a. What component protects the PRT from exceeding 100 psig?
(0,5)
- b. What gas is added to the PRT to prevent the formation of an explosive mixture of hydrogen and oxygen in the PRT during normal operation? (0,5)
- c. Where is the steam from a PORV released into the PRT that insures that the large volume of steam entering the tank is able to be handled by the tank? (1.0)
- ANSWER 2.06 (2.0)
- a. Rupture Disc. (0,5)
- b. Nitrogen. (0,5)
- c. Steam is released below the water, (Where the steam is condensed.) (Released into sparger ring or sparged.)(1.0)
- Reference DCPP System Description A-4b, Pgs 9 through 13.
010000K604 l I i
- QUESTION 2.07 (2.0)
The Chemical and Volume Control System (CVCS) provides a path for make-up and let down to the reactor coolant system,
- a. What are the normal and alternate reactor coolant loops the CVCS charges directly into during normal tower operations? (7.5)
- b. What reactor coolant system loop does normal letdown come from? 'O.5)
- c. What is the normal actuating fluid for the CVCS .1 1tdown orifice stop valves? (0.5)
- d. What is the back-up actuating flufd used to operate the letdown orifice stop valves? (0.5)
- ANSWER 2.07 (2.0)
- a. Loop 4 (RCS cold Leg, normal) (0.25)
Loop 3 (RCS cold leg, alternate) (0.25)
- b. Loop 2 (RCS cold leg, crossover) (0,5)
- c. (Instrument) air. (0.5)
- d. Nitrogen. (0.5)
- Reference DCPP System Description B-la, Pgs. 15 through 41. DCPP P&ID 102008.
004000K101 l I l l l
- QUESTION 2.08 (2.0)
The main steam line flow restrictor Jimits the main steam velocity and flow rate during a main steam line break to reduce the magnitude of pipe whip and contributes to the pressure drop for steam flow measurements. What are the two (2) other reasons for the main steam line flow restrictor during a main steam line break? (2.0)
- ANSWER 2.08 (2.0)
- 1. Limits the mass flow rate (OR differential pressure Delta P) the main steam isolation valve must close against (during a main steam line break). (1.0)
- 2. Limits the cooldown rate of the reactor coolant system.
(1.0)
- Reference DCPP System Description C-2a, Pgs. 9 through 11.
000040EK202
- QUESTION l 2.09 (3.0) )
There is a main steam isolation valve (MSIV) and a main steam check valve in each of the main steam lines from the steam generators.
- a. What fluid is used to open the HSIV? (0,5)
- b. What two (2), component (s) or fluid (s), close the MSIV when a MSIV control switch is turned to the closed position while steam is flowing through the pipe? (1.0)
- c. What design accident (0.5) and subsequent reactor transient (0,5) is he MSIV designed to minimize?
- d. What main steam line accident is the main steam line check valve designed for? (0.5)
- ANSWER 2.09 (3.0)
- a. (Instrument) air. (0.5)
- b. 1. Spring, (0,5)
- 2. steam flow. (0.5)
- c. Excess reactor coolant system cooldown, (0,5) from main steam line breaks downstream of the main steam isolation valves. (0.5)
- d. Main steam line break upstream of the main steam isolation valves. (0,5)
- Reference DCPP System Description C-2a, Pgs. 16 and 17.
000040EK301 1 l l 1
- QUESTION 2.10 (3.0)
The steam dump system is designed to remove excess heat from to reactor coolant system. The 10% power operated relief valves can operate independently or as a group.
- a. How many atmospheric dump valves are available to dump main steam down stream of the main steam isolation valves? (0,5)
- b. in what operating modedo the 10% atmospheric dump valves act independently, other than manual? (1.0)
- c. How are the 10% atmospheric dump valves designed to operate if there is a loss of control air and electrical power to the valve? (1.0)
- d. Which group of steam dump valves (10%, 35%, and/or 40) can be operated from the hot shutdown panel? (0.5)
- ANSWER 2.10 (3.0) a nine (9) (0,5)
- b. When in the overpressure protection mode. (1.0)
- c. Valves have local manual control (direct linkage handwheel) or fail closed. (1.0)
- d. 10%. (0.5)
- Reference DCPP System Description C-2b, Pgs. 17, 21, 22, SDS.1 and P&ID 102004.
041020K603
- QUESTION 2.11 (1.5)
Why are the rod insertion limits higher for unit 2 than they are for unit 17 (1.5)
- ANSWER 2.11 (1.5)
The power defect for unit 2 is greater (because of the higher thermal rating) or higher rod worth.
- Reference DCPP System Description A-3a, Pgs. 85 and 86.
001000K504 ! l I l
END OF SECTION TWO CONTINUE ONTO SECTION THREE i l l l 1
SECTION THREE INSTRUMENTS AND CONTROLS
*OUESTION 3.01 (2.5)
Tave, Tref, and nuclear power instrumentation are used to control rod speed and direction,
- a. What signal is used to produce Tref? (0.5)
- b. What is the control rod speed for the following conditions? (1.0)
- 1. Tave = 571 F
- 2. Tref = 565 F
- 3. Nuclear power = 100X?
- c. Why does the out-motiog bistable and relay de-energize at (1.0) a Tref minus Tave of 1 F7
- ANSWER 3.01 (2.5)
- a. P impulse (PT-505 First stage turbins pressure). (0.5)
- b. 72 spm (Steps per Minute) (1.0)
., c. Prevent system oscillations. (1.0)
- Reference DCPP System Description A-3a, Pgs. 30 through 36, 77 and
- figures RC.B and 11.
! OO1000K403 i
I l t i i 4
- QUESTION 3.02 (3.0)
Control Rod stops C-5, turbino power less than 15%, and C-11, control banh D withdrawal limit greater than 220 ' stops, stops only AUTOMATIC outward rod motion. What are the four (4) control rod stops (0.5 points cach)and their set points (0.25 points each) that will stop all outward rod motion? (3.0)
- ANSWER 3.02 (3.0)
- 1. Intermediate rango nuclear overpower (C-1) (0.5),
20% (0.25).
- 2. Power range (high rango) nuclear overpower (C-2) (0.5),
103% (0.25).
- 3. Overtemperature delta T (C-3)(0.5), 3% (0.25).
- 4. Overpower delta T (C-4)(0.5), 3% (0.25).
- Reference .
DCPP System Description A-3a, table RC-2. 001000K407 I 1 l 2 1 )
- QUESTION 3.03 (3.5)
The pressurizer heaters are powered from 480 volt vital (emergency) and non vital power supplies. Back-up heater control switches for groups 1-2, 1-3, and 1-4, have five (5) breaker status lights, green, white, blue, red, and amber. When back-up heater groups 1-2 and 1-3 are aligned to their , vital (emergency) power supply there is no power available for any of the associated breaker status lights.
- a. What are the two (2) 480 volt vital busses that supply back-up power to back-up heater groups 1-2 and 1-3 respectively? (0.5)
- b. What does the AMBER breaker status Ifght for the back-up heater control switches for heater groups 1-2 and 1-3 indicate? (0.5)
- c. What instrument indication is available to the operator to indicate the operation of back-up heater groups 1-2 and/or 1-3 are being supplied from their vital pawer source (besides no breaker status lights)? (1.0)
- d. What two (2) back-up heater groups have both an individual back-up heater control transfer switch and an individual back-up heater local control switch on the Hot Shutdown Panel (HSP)? (1.5)
- ANSWER 3.03 (3.5)
- a. 1-2 (vital bus) 1G (0.25) 1-3 (vital bus) 1H (0.25)
- b. Back-up heater control switch (on CC-1) is in the Auto-after-off position (available for auto). (0.5)
- c. Wattmeter (is provided to indicate power drawn by these j back-up heater groups). (1.0) l l
- d. Back-up heater groups 1-2 (0.75) )
l and 1-4. (0.75) j l
- Reference l DCPP System Description A-4a, Pgs. 51 through 55.
010000K201 l i
)
h i i
- QUESTION 3.04 (1.0) !
l Figure 3.04 is a diagram of the gas amplification curve for gas filled detectors.
- a. In what region of the curve dose the DF 3 source rance detectors operato? -(0,5) ,
- b. In what region of the curve do the power range, operate?
(0.5) J-4
- ANSWER 3,04 (1.0)
- a. III (Proportional) (0,5)
- b. II (Ionization Chamber Region) (0.5) .
]
- Reference ,
i - DCPP System Description B-4, Pgs. 19 through 32. l 015000KbO1 l 1 l l l t e B i I l l 1 l 1 il 1 ! I i
)
FIGlRE.3,04 Gas Amplification Curve . - I I II I III l IV l V l VI I I I l l I I I I I I I I Log of l 1 . Ion Pairs j g Collected .l l Per Event l l l g 1 1 I Alpha / I . g g g l Voltage Gradient, volts /cm I: Recombination Region II: Ionization Chamber Region III: Propbrtional Region IV: Limited Proportional Region V: Geiger-Mueller Region VI: Continuous Discharge Region 1
# e I ,
l
- QUESTION 3.05 (1.5)
The pressurizer levo] control system has four (4) pressurizer differential pressure detectors for level control and indication. Three of tho level sensors are hot calibrated and the fourth is cold calibrated.
- a. Ilow will the cold calibrated level sensor road during normal 100% power operating temperatures and pressurcs, (high, low, came as hot calibrated)? (0.5)
- b. Ilow will the charging flow rato change if the reference leg for the selected pressurizer level control channo]
leaked and drained dry? (1.0)
- ANSWER 3.05 (1.5)
- a. Low (0.5) ,
- b. Decrease (flow to pressurizer). (1,0)
- Reference DCPP System Description A-4a, Pgs. G: through GG, figure Pc.
4 and 23. 011000K301 P f
- QUESTION 3.06 (1,0) ,
The output of the master Icvel controller (MLC) for the pressurizer Icyc1 control system can be thought of as the charging flow demand signal. Reference level (Lref) is one signal input to the MLC,
- a. What other signal is required to produce the correct output signal from the MLC (besides Lref)? (0,5)
- b. What signal is used to generate Lref? (0.b)
*ANGWER 3.06 (1.0)
- a. Actual (pressurizer) level. (0.5)
- b. Auctioneered high Tave. (0.5)
- Reference DCPP Gystem Description A-4a, Pgs. G8 through 72.
011000K404 I J i i I a l { l
- QUESTION 3.07 (1.25)
The reactor is operating at 100% power with pressurizer level at 60% and pressurizer pressure at 2200 psig. An operator places the master pressure controller in manual to increase pressure to 2235 psig (normal operating pressure) by pressing the increase button. At what pressurizer pressure will the reactor trip at if the operator maintains the increase signal to the master pressure controller? (1.25)
- ANSWER 3.07 (1.25) 1950 psig (pressurizer low pressure.)
(2385 psig, to increase the pressurizer pressure the signal from the master pressure controller is decreased.)
- Reference DCPP System Description A-4a, Pgs. 43 and 77.
010000K302
I
- QUESTION 3.08 (2.5)
There are three plant signals which can produce a feedwater isolation signal. One of the signals is P-4, reactor trip permissive, in coincidence with low Tave on two of four l loops.
- a. What are the other two (2) plant signals that can cause a main feedwater isolation? (1.0)
- b. Why are the main feedwater regulating bypass valves closed on a main feedwater isolation signal initiated by a low Tave? (1.0)
- c. What is the set point for the two out of four Taves' that will initiate a main feedwater isolation? (0,5)
- ANSWER 3.08 (2.5)
- a. 1. Steam generator high-high level (67%) (on two of three level detectors on one steam generator) (P-14). (0.5)
- 2. Safety Injection. (0,5)
- b. Prevents excess RCS cooldown (which could cause excess positive reactivity to be added to the core). (1.0)
- c. 554 F (0.5)
- Reference DCPP System Description B-6a, Pgs. 51 and 52 and Systeu Description C-8b, Pgs. 23 and 31.
059000K419 l l l
i
- QUESTION a 3.09 (0.75) ;
The unit is_boing started up, a steady positive startup rato : is indicated, when t.in discovered that the intermediate ranco nuclear instrumentation is undercompensated. . W111 the actual Startup rato be higher or lower than tho intermediate range instruments aro showing? (0.75) ;
- ANSWER ,
3.09 (0.75) F
- 111ghe r
*Referenco ;
DCPP Gystem Description B-4, Pga. 19 through 32. 015000K502 4 i ll t i t a ) i . i i 8 ] a j ! 4
- l !
a i l 1
)
- QUESTION j 3.10 (2.0)
The motor driven auxiliary feedwater pumpn are provided with control switchen and four status lightn above each control switch. 4
- a. What does the white status light above the control switch indicate? (0,5)
- b. What does the blue status light indicate? (0.5)
- c. What 4160 volt busnes are motor driven auxiliary feedwater pump motors 12 and 13 powered from? (1.0)
- ANSWER 3.10 (2.0)
- a. Power is available (for the pump and control circuit).
(0.5)
- b. The breaker han tripped on overcurrent. (0.5) ,
- c. 12 Du n 11 ( 0. 5 ) , 13 Bus F (0.5) ,
i
- Reference DCPP System Description D-1, Pc, 14 and 43 a 001000K202 I
f i i i l a i
.
- QUESTION l 3.11 (2.5)
The motor driven and steam driven auxiliary feedwater pumps automatically start on a two of three low-low water level in
- one steam generator. The turbino driven pump also starts on 1 a loss reactor coolant pump bus power.
- a. What are three (3) other signals or conditions that will automatically start the motor driven auxiliary feedwater pumps? (2.5)
- b. What two (2) valve groups are closed by the same signal that starts the turbino driven auxiliary feedwater pump, (valves are not aux FW pump related)? (1.0)
L
- ANSWER 3.11 (2.5)
- a. 1. Both main feedwater pumps tripped, (0,5)
I 2. Gafety injection signal, (0.5)
- 3. Transfer to dicsol generator (Loss of power). (0.5)
' b. Gteam generator blowdown valves outsido containment,(0,5) Gtcam generator sample valves. (0.6)
- Reference i DCPP System Description D-1, Pgs. 17 and AFW.4 & S, and 4
Gyntem Doncription B-Ga, Pgn. 52 and 53. 4 j 061000K402 i l
\ l i
1 l i
)
- QUESTION 3.12 (2.0)
A safety injection signal on two of two trains will initiate a containment vontilation isolation (CVI) signal. High radioactivity readings on RE-11 and RE-12 will also init.into a CVI. What are the two other radiation detectors that will d a. initiate a CVI signal? (Name or number (0.25 points cach) and type of activity, gaseous or particulate (0.25) points each), required; A/B counts as one detector.)(1.0)
- b. What type of detectors are RE-11 and RE-12 (0.5 points each)?
(Geiger-Mueller, Ion Chamber detector, or scintillation.) (1.0)~
- ANSWER 3.12 (2.0)
, a. 1. RE-28 A/B (Plant vent air) (0.25) particulate, (0.25)
- 2. RE-14 (Plant vent air) (0.25), Radiogas (0.25).
- b. RE-11, Scintillation. (0.5)
RE-12, Geiger-Mueller. (0,5)
- Reference DCPP System Description D-Go, Pgs. 7, 8, and CIG.1 and DCPP System Description G-4, Pgs. 26 through 20, 103000K40G f
i 1 J { e
l i
- QUESTION 1 3.13 (1.5) i Liquid radioactive wantos arc monitored by radiation element RE-18 prior to being dincharged.
- n. What type of detector is radiation c]cment RE- 107 (0.5)
(Goicer-Muellor, Ion chamber, Scintillation.)
- b. What two (2) automatic valves change position (0.5 pointa each), if RE-10 reaches ita Hi radiation alarm set point?
(Valve numbers or name required) (1.0)
- ANSWER 3.13 (1.5)
- a. Scintillation (0,5)
- b. RCV-18 (shut off to circulating water discharge) (0,5), ,
FCV-477 (liquid radwanto equipment drain rocciver dump
; valvo) (0,5).
- Reference DCPP System Duncription G-1, Pga. 34 and 35, and System Doncription G-4, Pg. 33. ,
3 068000KG10 0 1 4 i i I i
l i END OF GECTION TIIREE CONTINUE 011T0 SECTION FOUR .i
PROCEDURES NORMAL, ADNORMAL, EMERGENCY, and RADIOLOGICAL CONTROL
- QUESTION-4.01 (2.0)
The Technical Specifications state that if a proscuriner power operated relief valve (PORV) block valve is inoperabic, in modes 1, 2, and 3, the PORY block valve must be restored to operability within one hour. What two actions must be taken-if the PORY block can not be returned to the operable status within one (1) hour?
- ANSWER 4.01 (2.0)
- 1. Close the PORY block valvo, (1.0)
- 2. Remove power from the block valve. (1.0) i -
- Reference DCPP Technical Specifications 3/4.4.4, PC. 3/4 4-10 010000KGG5 a
i a d i 4 i i
E 1 t
- QUESTION 4.02 (2.0) 4 Unit 1 in in mode G with diesel generator 1-2 out of [
- se rvice , diesel generator 1-3 cupplying p'ower to unit 2 and ;
' auxiliary transformer 1-2 in out of servico. [ What four (4) immediate actiona , other than restore cource ! to operable, must be taken to be in comp]iance with Technical Specification 3.8.1.2 if the engine mounted diesel - fuel tank for diesel generator 1-1 han drained and in not : j refilling? l
- ANSWER 4.02 (2.0)
- 1. Suspend all core alterations, (0.5) l
- 2. Suspend positive reactivity changos, (0.5) ,.
- 3. Suspend irradiated fuel movement in the spent-fuel pool, (0.5)
- 4. Suspend any crane operations with loads over fuel storage pool. (0.5) 1
- Reference !
DCPP Technical Specification 3/4.8.1, Pg. 3/4 8-11. OG4050KSG8 l
- i 1 .
I i i i l ! i 1 4 l < 1
- QUESTION 4.03 (3.5)
The unit 2 reactor trip actuated alarm has come in and the unit has not tripped. You have entered FR-S.1, Response to Nuclear Power Generation /ATWS, from EP E-0, Reactor Trip or Safety Injection. The first immediate response is to manually trip the reactor,
- a. What immediate action will you take if after manually initiating a reactor trip the reactor does not indicate that a trip has taken place? (1.0)
- b. What immediate action will you take if the turbine generator stop valves are not closed af ter a manual trip and run back of the turbine has failed? (1.0) t'
- c. What are 'ehe three ways of verifying that the reactor has tripped in accordance with EP E-07 (1.5)
- ANSWER 4.03 (3.5)
- a. Manually de-energize 480 volt busses (23 D & E) or (13 D -
& E CO.5 points 3) feeding rod drive motor generators.
(1.0)
- b. Close the MSIVs (0.5)
- and bypass valves. (0.5)
- c. 1. Verify reatter trip, (0.25) and bypass breakers are open. (0.25)
- 2. Rod bottom lights lit, (0.5)
- 3. Decreasing neutron flux. (0.5)
- Reference DCPP EP E-O and FR-G.1.
000029EK301 l 1 I i 1 l i 1 1
- QUESTION 4.04 (1.0)
A reactor trip has occurred and the EP E-0.1, Reactor Trip Response has been entered. Duririg the recovery the schcooling monit.or iails. What two'(2) indications wil l yco use, by procedure,to determino subcooling? 10.5 peirit.a cach)
- ANSWER '
4.04 ('1.0)
~
Wide rango RCS rpessure, (0.5) Core exit thermocouples. (0.5)
- Reference DCPP EP E-0.1 '
003000K556
\
e e y l I l l J
> +
- - -* w e- 1n--- - -- - " - -
---toe =-y w -- e+ e sw <-
- QUESTION 4.05 (2.5)
During a loss of all AC power EP ECA-0.0 is entered. One of the immediate operator actions is to verify that the reactor coolant system (RCG) has been isolated and to verify that auxiliary feedwater is in operation. i
- a. What are the four (4) things (not specific valve numbera) that must be checked to verify that the RCG has been isolated? (2.0)
- b. What flow rate (GPM) must auxiliary feedwater flow be greater than to meet its safety function? (0.5)
- ANSWER 4.05 (2.5)
- a. 1. PZR PORVs closed, (0.5)
- 2. Letdown isolated, (0.5)
- 3. Excess letdown isolated, (0.5)
- 4. RCP Seal return valves isolated. (0.5) 4
- b. 460 GPM 4/- 10 (0.5)
- Reference
.DCPP EP ECA<0.0 000055EK302
l l
- QUESTION 4.06 (2.0)
EP ECA-0.0, Loss of All AC Power, notes that the steam generators should be depressurized at the maximum rate to minimize the loss of reactor coolant inventory due to a component failure but EP ECA-0.0 Cautiona the operator not to allow steam generator pressure to go below 160 psig. ' During the depressuriz.. tion of the steam generators upper reactor vessel head voiding may occur.
- a. What component failure is being anticipated when reducing the steam generator pressure as rapidly as possibic?(1.0)
- b. What cor.dition is prevented by not reducing the steam generator pressure below 160 psig? (1.0)
- ANSWER 4.06 (2.0)
- a. Reactor coolant pump seal failure. (1.0)
- b. Injection of nitrogen into the RCS from the SI accumulators. (1.0)
- Reference DCPP EP ECA-0.0.
00005GEK302
- QUESTION 4.07 (2.5)
The Unit 2 reactor is operating at 100% power with all rods out (ARO). One of the control rods dropa into the core and is declared inoperable but trippable. A shutdown margin calculation is completed and shutdown margin in 1.1% delta K/K. After shut down margin is restored continued operation is allowed at a reduced thermal power of level in accordance with Technical Specification 3.1.3.1.
- a. What minimum flow rate and boron solution concentration are you required to borate at according to Technical Specification 3.1.1.1 to restore required shutdown margin? (1,0)
- b. What percent of rated thermal power is the reactor allowed to operate at with the inoperabic control rod?
(1.0)
- c. How many steps is a control rod allowed to out of indicated position before it is considered misaligned?
(0.5)
- ANSWER 4.07 (2.5)
- a. 10 gpm (0.5) 20,000 ppm (0.5)
- b. 75% (of rated thermal power). (1.0)
- c. 4/- 12 steps. (0.5)
- Reference DCPP Technical Specification 3.1.1.1 and ., .1. 3 .1.
000005EK306 i l
- QUESTION 4.08 (1.5)
Reactor core safety limits an specified by Technical Specification 2.1 state that the combination of thermal power , pressuriner pressure, and the highest operating loop coolant Tave shall not exceed the limits shown in Figure 4.08.
- a. What action must be taken within one (1) hour of exceeding the limits shown on figure 4.087 (1.0)
- b. Who must be notified within one (1) or less if the limits shown on figure 4.08 are exceeded? (0.5)
- ANSWER 4.08 (1.5)
- a. De in Hot Standby (within one hour). (1.0)
- b. NRC (By phone). (0.5)
- Reference DCPP Technical Specification 2.1, Pg. 2-1, and Specification 6.7, Pg. 6-13.
010000SGKS
. FIGURE I4.08 -
670 _ i I i 660 N UNACCEPTABi M I N2400 PSIA OPE RAT lON I 650 N
- I ,
' ~
N I I 640 ~ 2_2 5 0 P S s ! -
.m N \ i .
- u. N N -
i I 630 N ' 620 * \
< 610 1954.7 PSI A \
X\ : 600 ACCEPTABLE \1 OPERATION K 590 l' N 1 580 i il 570 0 20 40 60 80 100 12C PERCENT OF RATED THERMAL POWER
~
l DIdBLOCANYON-UNITS 1&2 I' .- .
l 1 l 1
- QUESTION 4.09 (1.0)
The unit 1 reactor announced as critical at 1:00 am. At 1:30 am he loop 3 Tave has been confirmed as being less than 541 ...
- a. How long do you have to restore Tave to within its limits? (1.0)
- b. At what time must you be in Hot Shutdown if Tave can not be restored to within its limits? (DELETED)
- ANSWER 4.09 (2.0)
- a. 15 minutes. (1.0)
- b. ** DELETED **
- Reference DCPP Technical Specification 3.1.1.4, Pg. 3/4 1-6.
002020SGKS 1 1 i
- QUESTION 4.10 (3.0)
An elbow in the CVCS system is emitting a dose rate of 1500 mrom per hour-at one (1) meter. You have just started a calendar year and have zero accumulated done.
- a. Ilow long can you work at a position 2 meters from the elbow before you meet your 10 CFR 20 whole body quarterly dose limit? (1.5)
- b. What are the two (2) criteria that allows your 10 CFR 20 quarterly limit to be increased up to 3 rem per quarter?
(1.0)
- c. What is your DCPP emergency lifesaving action whole body limit? (0.5)
- ANSWER 4.10 (3.0)
- a. 3.33 Hours (0,5) 1500 mrom = 1.5 rem 1.5 rem /hr x (1)2 meter : D2 x (2)2 meters (0,5)
(1.5 x 1) / 4 = 0.375 rem /hr . 1.25 rem / quarter (0.5) (1.25 rem) / 0.375 rem /hr 3.33 hours
- b. 1. Form NRC-4 is current, (0,5)
- 2. Life time dose dose not exceed 5(N-18). (0,5)
- c. 75 rem (0.5)
- Reference DCPP Operator Information Manual, and 10 CFR 20.
194001K103
- QUESTION 4.11 (3.0)
Emergency Operating Procedure EP E-0, Reactor Trip or Safety Injection, and EP E-1.1, Safety Injection Termination, provide termination criteria for safety injection after a reactor trip and safety injection. One of the criteria for termination of safety injection is at least one (1) narrow range steam generator level indication greater than 4%.
- a. What are the other four criteria that need to be met before safety injection can be terminated or reduced?(2.0)
- b. What are the two conditions that require that safety injection be reinitiated after it has been terminated?
(1.0)
- ANSWER 4.11 (3.0) a.(2.0 points)
{a.1, a.3, a.4 required for full credit 0.667 points each)
- 1. RCS subcooling greater than 20 F (based on core exit thermal couples).
- 2. Total auxiliary feedwater flow to steam generators greater than 460 GPM.(Not required)
- 3. RCS pressure - stable or increasing.
- 4. Pressurizer level greater than 4%.
- b. (0.5 points each maximum 1.0 points)
- 1. RCS subcooling less than 20 F.
- 2. Pressurizer level can not be maintained greater than 4%.
- Reference DCPP EP E-0 and EP E-1.1.
r - 4 END OF SECTION FOUlf END OF EXAMINATION i I
co/2Ec7E D EXW G U.S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION Facility: DIABLO CANYON Reactor Type: WESTINGHOUSE FOUR LOOP Date Administered: DECEMBER 8, 1987 Examiner: P.J. MORRILL Candidate: N/A INSTRUCTIONS TO CANDIDATE 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 Candidate's Category Value Total Score Value Category 25 00 25.78 5. Theory of Nuclear Power Plant Operation,
' Fluids, and Thermodynamics 94 67 2539 X f 6. Plant Systems Design, Control and Instrumentation 24.S0 QS).I 28 F5 7. Procedures - Normal, Abnormal. Emergency, and Radiological Control
.23.w 2347
# Ef 8. Administrative Procedures, Conditions, and Limitations 19tf TOTALS Final Grade All work done on this examination is my own, I have neither given nor received aid.
Candidate's Signature 1 l
i ES-201-1 Enclosure 2 REQUIREMENTS FOR ADMINISTRATION OF WRITTEN EXAMINATIONS j. 1. A single room shall be provided for completing the written examination. The location of this room and supporting restroom facilities shall be such as to prevent contact with all other facility and/or contractor personnel during the duration of the written examination. If necessary, the facility should make arrangements for the use of a suitable room at a local school, motel, or other building. Obtaining this room is the responsibility of < the licensee. I
- 2. !
Minimum spacing is required to ensure examination integrity as detemined by the with a 3-ftchiefspace examiner. Minimum between tables.spacing should bc one candidate per table, No wall charts, models, and/or other training materials shall be present in the examination room. 3. Suitable arrangements shall be made by the facility if the candidates are to have lunch, coffee, or other refreshments. These arrangements shall comply with Itse 1 above. These arrangements shall be reviewed by the examiner and/or proctor. 4. The facllity staff sha'il be provided a copy of the written examination and answer key after the last candidate has completed and handed in his written examination. The facility staff shall then have five working days to pro-vide formal written comments with supporting documentation on the exanina-tion and answer key to the chief examiner or to the regional office section chief. 5. The facility licensee shall provide pads of 8-1/2 by 11 in. lined paper in l unopened packages for each candidate's use in completing the examination. The examiner shall distribute these pads to the candidates. All reference natuial needed to complete the examination shall be furnished by the examiner. Candidates can bring pens, pencils, calculators, or slide rules i into the examination room, and no other equipment or reference material shall be allowed. 6. Only black ink or dark pencils should be used for writing answers to questions. ' 6 Examiner Standards ..
ES-201-1 Enclosure 2 i 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 applicat' ion 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).
EEE-
- 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 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 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.
l 1 1 Examiner Standards - l
ES-201-1 Enclosure 2 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 a part of the answer.
- b. Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn in all scrap paper and the balance of the paper that you did not use fot answering s 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. i i 1 l l I I 4 l Examiner Standards . e.
---- -.- - . , , -. - . . . - . - _,,_, . , , , .,y. -,,, -.r-
I KQUATION SHEET f = ma v = s/t w = ag 2 Cycle efficiency = Net Work (out) a = v,t + at Energy (in) E = aC a = (v' - y*)/t EE = b2 vg = v, + at A = AN A = A,e
-At Pz = agh a = 6/t A = In 2/t = 0.693/t g W = v&P -
l iz . ,314m V ) - ((t, )(t ) t g + g)
, k = h pAT y . I p,-IX l , .'-
6=UAAT y . g -ux
'. Pwr = Ug 'E . .
i.g go-x/ m g ; P=P 10 5UR(t) O . gyt . g,3/9 ! P=P o et/T EVL ='O.693/u i SUR = 26.06/T - ' T = 1.44 DT SCR = S/(1 - E,fg) fA*re cT SUR = 26 g,, CR, = S/(1 - K,ggg)
~
T = D*/o ) + [(i ' p)/A,,fo ] 1( eff)1 " 2 (I ~~ Keff)2 y .1*/ (, _ y; - M = 1/(1 - K,gg) = CR 3/CR0
~8 eff'
#" ~
M = (1 - K,gg) /(1 - K,,f)}
/K eff .. aff " #eff aff .
,..,, SDM = (1 - K,gg)/K,gg p= ~
[1*/TKygg] + [5/(1 + 1,ggT )] ,
'1* = 1 x 10 seconds P = Z$V/(3 x 1010) ~I A,gg = 0.1 seconds ..
E = Ho - Idgg=Id22 WATER PARAMETERS Id =1d2 g
,1 gal. = 8.345 lba 2 R/hr = (0.5 CE)/d (meters) ,
I'tal. = 3.78 liters R/hr = 6 CE/d (feet) I ft3 = 7.48 gal. MISCELLANEOUS CONVERSIONS , Density =.62 4 lbm/ft 3 1 Curie = 3.7 x 1010dps Density = 1 gn/cm 3 1 kg = 2.21 lba Heat of valorization = 970 Etu/lbm I hp = 2.54 x 103 BTU /hr Heat of fusica = 144 Btu /lbm 0 1 N = 3.41 x 10 5tu/hr 1 Ata = 14.7 psi = 29.9 in. I g. I BLu = 778 ft-lbf 1 ft. H 2O = 0.4333 lbf/in 2 1 inch = 2.54 cm F = 9/5'c + 32
'C = 5/9 ('T - 32) '
. . .n ' -
Properties of Saturated Steam and Saturofed Water * , Absolute Pressure Vacuum Temper- Heat of Latent Heat Total Hea t Specific Volume 1.bs, Inches Inches ature the of of Steam y Sq. E of Hg of Hg Liquid Evaporation t be Water Steam P' o.em. v. m ns. u ns u ns cw r.. w n cu re. ,,e n. 0.0087 0.02 29.90 32.018 0.0003 1075.5 1075.5 0.016022 3302.4 0.10 0.20 29.72 35.023 3.026 1073.8 1076.8 0.016020 2945.5 0.15 0.31 29.61 45.453 13.498 1067.9 1081.4 0.016020 2004.7 0.20 0.41 29.51 53.160 21.217 1053.5 1084.7 0.016025 1526.3 0.25 0.51 29.41 59.323 27.382 1060.1 1087.4 0.016032 1235.5 0.30 0.61 29.31 64.4fA 32.541 1057.1 1089.7 0.016040 1039.7 0.35 0.71 29.21 68.939 36.992 1054.6 1091.6 0.016048 898.6 ! 0.40 0.81 29.11 72.869 40.917 1052.4 1093.3 0.016056 792.1 1 l 0.45 0.92 29.00 76.387 44.430 1050.5 1094.9 0.016063 708.8 1 j 0.50 1.02 28.% 79.586 47.623 1048.6 10 %.3 0.016071 641.5 1 O.60 1.22 28.70 85.218 53.245 , 1045.5 1098.7 0.016085 540.1 0.70 1.43 24.49 90,09 58.10 1042.7 1100.8 0.016099 4 % .94 0.80 1.63 24.29 94.38 62.39 1040.3 1102.6 0.016112 411.69 0.90 1.83 28.09 98.24 %.24 1038.1 1104.3 0.016124 368.43 1.0 2.04 27.88 101.74 69.73 1036.1 1105.8 0.016136 333.60 1.2 2.44 27.48 107.91 75.90 1032.6 1108.5 0.016158 280.% 1.4 2.85 27.07 113.26 81.23 1029.5 1110.7 0.016178 243.02 0 , 1.6 3.26 26.% 117.98 85.95 1026.8 1112.7 0.0161 % 214.33 j 1.8 3.% 26.26 122.22 90.18 1024.3 1114.5 0.016213 191.85 2.0 4.07 25.85 126.07 94.03 1022.1 1116.2 0.016230 173.76 j 2.2 4.48 25.44 129.6! 97.57 1020.1 1117.6 0.016245 158.87 2.4 4.89 l 25.03 132.88 100.84 1018.2 1119.0 , 0.016260 146.40 2.6 5.29 24.63 135.93 103.88 1016.4 1120.3 0.016274 135.80 i 2.8 7.70 24.22 138.78 106.73 1014.7 1121.5 0.016287 126.67 i 3.0 6.11 23.81 141.47 109.42 1013.2 1122.6 0.016300 118.73 3.5 7.13 22.79 147.56 115.51 1009.6 1125.1 0.016331 102.74 4.0 8.14 21.78 152.% 120.92 1006.4 1127.3 0.016358 90.64 4.5 9.16 20.76 157.82 125.77 1003.5 1129.3 0.016384 83.03 i 8.0 10.18 19.74 162.24 130.20 1000.9 1131.1 0.016407 73.532 5.5 11.20 l 18.72 1%.29 134.26 998.5 1132.7 0.016430 67.249 i 6.0 12.22 17.70 170.05 138.03 996.2 1134.2 0.016451 61.984 i 6.5 13.23 16.69 173.56 141.54 994.1 1135.6 0.016472 57.506 7.0 14.25 15.67 176.84 144.83 992.1 1136.9 0.016401 53.650 7.5 15.17 14.65 179.93 147.93 990.2 1138.2 0.016510 50.294 8.0 16.29 13.63 { 182.86 150.87 988.5 1139.3 0.016527 47.345 8.5 17.31 12.61 185.63 153,65 986.8 1140.4 0.016545 44.733 1 9.0 18.32 11.60 188.27 156.30 985.1 1141.4 0.016561 42.402 9.5 19.34 10.58 190.80 158.84 983.6 1142.4 0.016577 40.310 ! 10.0 20.36 9.56 193.21 161.26 982.1 1143.3 0.016592 38.420 11.0 22.40 7.52 197.75 165.82 979.3 1145.1 0.01 % 22 35.142 12.0 24.43 5.49 201.% 170.05 976.6 1146.7 0.01 % 50 32.394 13.0 26.47 3.45 205.88 174.00 974.2 1144.2 0.01 % 76 30.057 14.0 38.50 1.42. 209.56 177.71 971.9 1149.6 0.016702 28.043 I Pressure Temper- Heat of Latent Heat Total Heat Specific Volume ! Lbs. per Sq. In. the of of Steam Abeolute GaSe ature Liquid Evaporation y . i t he Water 5 team P' P o r. em ns. mns. s ns. cu en. w w cu. rs. ww 14.6 % 0.0 212.00 180.17 970.3 1150.5 0.016719 26.799 15.0 0.3 213.03 181.21 % 9.7 1150.9 0.016726 26.290 16.0 1.3 216.32 184.52 %7.6 1152.1 0.016749 24.750 17.0 2.3 219.44 187.% 965.6 1153.2 0.016771 23.385 18.0 3.3 222.41 190.% %3.7 - 1154.3 0.016793 22.168 19.0 4.3 225.24 193.52 % 1.8 1155.3 0.016814 21.074 20.0 5.3 227.96 1%.27 960.1 1156.3 0.016834 20.087 21.0 6.3 230.57 198.90 %8.4 1157.3 0.016854 19.190 22.0 7.3 233.07 201.44 956.7 1158.1 0.016873 18.373 8,3 l 23.0 235.49 203.88 955.1 1159.0 0.016891 17.624 ' 24.0 9.3 237.82 206.24 953.6 1159.8 0.016909 16.936 25.0 10.3 240.07 208.52 952.1 1860.6 0.016927 16.301 26.0 11.3 242.25 210.7 950.6 1161.4 0.016944 15.7138 27.0 12.3 244.36 212.9 949.2 1162.1 0.016 % 1 15.1 % 4 28.0 13.3 246.41 214.9 947.9 1162.8 0.016977 14. % 07 29.0 14.3 248.40 217.0 946.5 1163.5 0.016993 14.1869 36.0 15.3 250.34 218.9 945.2 1164.1 0.017009 13 7436 31.0 16.3 252.22 220.8 943.9 1164.8 0.017024 '13.3280 32.0 17.3 254.05 222.7 942.7 1165.4 0.017039 12.9376 33.0 18.3 255.84 224.5 941.5 11 % .0 0.017054 12.5700 34.0 19.3 257.58 226.3 940.3 11 % .6 0.017069 12.2234
~, 1 Properths cf S tursted Stocm cnd S:turated Wct;r-entinund Pressure Temper. Heat of Latent Heat Total Heat Specific Volume i Lbs. per Sq. In. sture the of of Steam Liquid Evaporation y l Absol,ute Gase !
P P o F. neu nd. siu ns. sio ns. co.ri. ,,,is. co.ri.p.,is. 35.0 20.3 259.29 228.0 939.1 1167.1 0.017083 11.8959 36.0 21.3 260.95 229.7 938.0 1167.7 0.017097 11.5860 37.0 22.3 262.58 231.4 936.9 1168.2 0.017111 11.2923 - 38.0 23.3 264.17 233.0 935.8 1168.8 0.017124 11.0136 39.0 24.3 265.72 234.6 934.7 1169.3 0.017138 10.7487 l j 40.0 25.3 267.25 236.1 933.6 1169.8 0.017151 10.4 % 5 41.0 26.3 268.74 237.7 932.6 1170.2 0.017164 10.2563 42.0 27.3 270.21 239.2 931.5 1170.7 0.017177 10.0272 l 43.0 28.3 271.65 240.6 930.5 1171.2 0.017189 9.8083 1 44.0 29.3 273.06 242.1 929.5 1171.6 0.017202 9.5991 45.0 30.3 274.44 243.5 928.6 1172.0 0.017214 9.3%6 46,0 31.3 275.80 244.9 927.6 1172.5 0.017226 9.2070 47.0 32.3 277.14 246.2 926.6 1172.9 0.017238 9.0231 48.0 33.3 278.45 247.6 925.7 1173.3 0.017250 8.8465 49.0 34.3 279.74 248 9 924.8 1173.7 0.017262 8.6770 i 50.0 35.3 281.02 250.2 423.9 1174.1 0.017274 8.5140 51.0 l 36.3 282.27 251.5 923.0 1174.5 0.017285 8.3571 1 52.0 37.3 283.50 252.8 , 927 [ 1174.9 0.0172 % 8.2061 53.0 38.3 284.71 l 254.0 921.4 1175.2 0.017307 8.0606 54.0 39.3 285.90 255.2 ', 920.4 1175.6 0.017319 7.9203 1 T,. 55.0 40.3 287.08 256.4 919.5 1175.9 0.017329 7.7850 56.0 41.3 288.24 257.6 918.7 1176,3 0.017340 7.6543 57.0 42.3 289.38 258.8 917.8 1176.6 0.017351 7.5280 58.0 43.3 290.50 259.9 917.0 II T7.0
- 0.017362 7.4059 59.0 44.3 291.62 261.1 916.2 1177.3 7.2879 Q.0173 1 60.0 45.3 292.71 262.2 915.4 1877.6 0.017383 7.1736 61.0 46.3 293.79 263.3 914.6 1177.9 0.017393 7.0630 l 62.0 47.3 294.86 264.4 913.8 1178.2 0.017403 6.9558 63.0 48.3 295.91 l
265.5 913.0 1178.6 0.017413 6.8519 ' 64.0 49.3 2 % .95 266.6 912.3 1178.9 0.017423 6.7511 65.0 50.3 297.98 267.6 911.5 1179.1 0.017433 6.6533 M.0 51.3 298.99 268.7 910.8 1179.4 0.017443 6.5584 67.0 52.3 299.99 269.7 910.0 1179.7 0.017453 6.4%2 - 1 68.0 $3.3 300.99 270.7 909.3 1180.0 0.017463 6.3767 69 0 54.3 301. % 271.7 908.5 1180.3 0.017472 i 6.28 % 70.0 55.3 302.93 272.7 907.8 1180.6 0.017482 71.0 6.2050 56.3 303.89 273.7 907.1 1180.8 0.017491 6.1226 72.0 57.3 304.83 274.7 906.4 1181.1 73.0 0.017501 6.0425 58.3 305.77 275.7 , 905.7 1181.4 0.017510 5.9645 74.0 59.3 306.69 276.6 905.0 1181.6 0.017519 5.8885 75.0 60.3 307.61 277.6 904.3 1181.9 0.017529 76.0 5.8144 61.3 308.51 278.5 903.6 1182.1 0.017538 5.7423 77.0 62.3 309.41 279.4 902.9 1182.4 0.017547 78.0 5.6720 63.3 310.29 280.3 902.3 1182.6 0.017556 5.6034 79.0 64.3 311.17 281.3 C01.6 1182.8 0.0175M 5.5364 i ' 80.0 65.3 312.04 282.1 900.9 1183.1 0.017573 81.0 5.4711
%.3 312.90 283.0 900.3 1183.3 0.017582 5.4074 82.0 67.3 313.75 283.9 899.6 1183.5 0.017591 5.3451 83.0 68.3 314.60 284.8 899.0 1183.8 0.017600 84.0 69.3 5.2843 315.43 2C5.7 898.3 1184.0 0.017608 5.2249 85.0 70.3 316.26 286.5 897.7 !!84.2 86.0 0.017617 5.1%9 71.3 317.08 287.4 897.0 1184.4 0.017625 5.1101 87.0 72.3 317.89 288.2 8%.4 1184.6 88.0 0.017634 5.0546 73.3 318.69 289.0 895.8 1184.8 0.017642 5.0004 89.0 74.3 319.49 289.9 895.2 1885.0 0.017651 4.9473 90.0 75.3 320.28 290.7 894.6 1885.3 91.0 0.017659 4.6953 76.3 321.06 291.5 893.9 1185.5 0.017667 4.8445 92.0 77.3 321.84 292.3 893.3 1185.7 93.0 0.017675 4.7947 78.3 322.61 293.1 892.7 1885.9 0.017684 4.7459 94.0 79.3 323.37 293.9 892.1 1186.0 0.017692 4.6982 95.0 40.3 324.13 294.7 891.5 1886.2
%.0 0.017700 4.6514 l 81.3 324.88 295.5 891.0 1186.4 0.017708 4.6055 97.0 82.3 '325.63 2%.3 890.4 1186.6 98.0 0.017716 4.5606 i 83.3 326.36 297.0 889.8 1186.8 0.017724 4.5166 99.0 84.3 327.30 297.8 889.2 1187.0 0.017732 4.4734 100.0 65.3 327.83 298.' 888.6 1187.2 101.0 0.017740 4.4310 86.3 328.54 299.3 888.1 1187.3 0.01775 4.3895 J
102.0 87.3 329.26 300.0 887.5 1187.5 103.0 '0.01776 4.3487 < 88.3 329.97 300.8 886.9 1187.7 0.01776 4.3087 104.0 R9.3 330.67 301.5 886.4 1187.9 0.01177
)
4.2695 l 105.0 90.3 3J 1.37 302.2 885.8 1888.0 106.0 0.01778 4.2309 91.3 332.06 303.0 885.2 1188.2 0.01779 4.1931 l 107.0 92.3 332.75 303.7 8 84.7 1188.4 108.0 0.01779 4.1560 93.3 333.44 304.4 884.1 1188.5 0.01780 4.1195 109.0 94.3 . 334.11 305.1 883.6 1188.7 0.01781 4.0837 I
- . _. _ _ . _ -. _ _ . w- - - - - - . _ _ _ _ _ _ _ -- -
~
Eropertb8 cf Snturatsd Staum cmd 5sturat:d Watsr-continu:d Pressure Temper- Heat of Latent Heat ; Total Heat Specific Volume Lbs. per Sq. In. sture the Liquid of of Steam p i Absolute Gage ! Everwration g Wate 5 team P, P o.a r. s.u nd. miu ns. m o nt . c ry.,r is. t u. ri. e., is s lo.0 95.3 334.79 303.8 483.1 1868.9 0.01762 4.04s4 111.0 %.3 335. # 306.5 882.5 1189.0 0.01782 4.0138 112.0 97.3 3M.12 307.2 882.0 1189.2 0.01783
- 3.9798 113.0 98.3 336.78 307.9 881.4 1189.3 0.01784 3.9464 114.0 99.3 337.43 308.6 880.9 1189.5 0.01785 3.9136 115.0 100.3 334.08 309.3 .480.4 1189.6 0.01785 3.8413 316.0 101.3 338.73 309.9 879.9 1189.8 0.01786 3.8495 117.0 102.3 339.37 310.6 479.3
- 1189.9 0.01787 3.8183 118.0 103.3 340.01 311.3 878.8 1190.1 0.01787 3.7875 119.0 104.3 340.64 311.9 878.3 11 % .2 0.01788 3.7573 120.0 105.3 M t.27 312.6 877.8 1890.4 0.01789 3.7275 221.0 106.3 341.89 313.2 877.3 1190.5 0.01790 3.6983 122.0 . 107.3 M2.51 313.9 876.8 1190.7 0.01790 3.% 95 123.0 100.3 343.13 314.5 876.3 1190.8 0.01791 3.6411 .
124.0 109.3 343.74 315.2 ~ 875.8 1190.9 0.01792 3.6132 125.0 110.3 M4.35 315.8 875.3 1191.1 0.01792 3.5857 126.0 111.3 344.95 316.4 874.8 1191.2 0.01793 3.55M 127.0 112.3 345.55 317.1 874.3 1191.3 0.01794 3.5320 128.0 113.3 3 # .15 317.7 873.8 1191.5 0.01794 3.5057 129.0 114.3 3#.74 318.3 873.3 1191.6 0.01795 3.4799
*
- 130.0 115.3 347.33 219.0 872.8 1191.7 0.017 % 3.4544 I 131.0 116.3 347.92 319.6 872.3 1191.9 0.01797 3.4293 320.2 871.8 1192.0 0.01797 3.40 # l 132.0 117.3 348.50 133.0 114.3 M 9.08 320.8 871.3 !!92.1 0.01798 3.3802 1 134.0 119.3 349.65 321.4 870.8 1192.2 0.01799 3.3M2 l 135.0 120.3 350.23 322.0 870.4 1192.4 0.01799 3.3325 IM.0 121.3 350.79 322.6 #9.9 1192.5 0.01800 3.3091 137.0 121.3 351.36 313.2 869.4 1192.6 0.01801 3.2861 138.0 123.3 351.92 323.8 868.9 1192.7 0.01801 3.2634 139.0 124.3 352.48 324.4 868.5 1192.8 0.01802 3.2411 3.2190 '
140.0 125.3 353.04 315.0 868.0 1193.0 0.01803 141.0 126.3 353.59 325.5 867.5 1193.1 0.01803 3.1972 142.0 127.3 354.14 326.1 867.1 1193.2 0.01804 3.1757 \ 143.0 128.3 354.69 326.7 8%.6 1193.3 0.01805 3.1546 I 144.0 129.3 355.23 327.3 8%.2 1193.4 0.01805 3.1337 !
!45.0 130.3 355.77 327.8 E5.7 1193.5 0.01806 3.I130 le.0 131.3 356.31 328.4 865.2 1193.6 0.01806 3.0927 147.0 132.3 3 M.84 329.0 #4.8 1193.8 0.01807 3.0726 148.0 133.3 357.38 329.5 864.3 1193.9 0.01808 3.0528 149.0 134.3 357.91 330.l" E3.9 1194.0 0.01808 3.0332 ;
150.0 135.3 358.43 330.6 863.4 1194.1 0.01809 3.0139 i 152.0 137.3 359.48 331.8 862.5 1194.3 0.01810 2.9760 154.0 139.3 365.51 332.8 861.6 1194.5 0.01812 2.9391 156.0 141.3 Mt.53 333.9 860,8 1194.7 0.01813 2.9031 154.0 143.3 M 2.55 335.0 859.9 1894.9 0.01814 2.8679 160.0 145.3 363.55 336.1 859.0 !!95.1 0.01815 2.8336
- 162.0 147.3 364.54 337.1 458.2 1195.3 0.01817 2.8001 164.0 149.3 %5.53 338.2 857.3 '1195.5 0.01818 2.7674 IM.0 151.3 an.50 339.2 SM.5 1195.7 0.01819 2.7355 168.0 153.3 M7.47 M0.2 S&5.6 1195.8 0.01820 2.7043 170.0 155.3 MS.42 M1.1 854.8 11 % .0 0.01821 2.6738 172.0 157.3 369.37 342.2 853.9 1196.2 0.01823 2.6440 174.0 159.3 370.31 343.2 453.1 1196.4 0.01824 2.6149 176.0 161.3 371.24 344.2 852.3 11 % .5 0.01825 2.5h4 178.0 163.3 372.16 345.2 851.5 11 % .7 0.01826 2.5585 180.0 165.3 373.08 3#.2 850.7 11 % .9 0.01827 2.5312 182.0 167.3 373.98 347.2 849.9 1197.0 0.01828 2.5045 184.0 169.3 374.88 348.1 849.1 1197.2 0.01830 2.4783 186.0 171.3 375.77 349.1 848.3 1197.3 0.01831 2.4527 188.0 173.3 376.65 350.0 847.5 1197.5 0.01832 2.4276 190.0 175.3 377.53 350.9 8#.7 1197.6 0.01833 2.4030 192.0 177.3 378.40 351.9 845.9 1197.8 0.01834 2.3790 194.0 179.3 379.26 352.8 845.1 1197.9 0.01835 2.3554 1% 0 181.3 380.12 353.7 844.4 1198.1 0.01834 2.3322 198.0 183.3 380.% . 354.6 843.6 1198.2 0.01838 2.3095 200.0 145.3 381.80 355.5 842.8 !!98.3 0.01839 2.28728 205.0 190.3 343.84 357.7 840.9 1198.7 0.01841 2.23349 210.0 195.3 385.91 359.9 439.1 1199.0 0.01844 2.13217 215.0 200.3 387.91 M2.1 837.2 1199.3 0.01847 2.13315 220.0 205.3 389.88 364.2 835.4 1199.6
- 6.01850 2.08629 215.0 210.3 391.80 3%.2 833.6 1199.9 0.01852 2.04143 230.0 215.3 393.70 368.3 831.8 1200.1 0.01855 1.99846 235.0 120.3 395.56 370.3 830.1 1200.4 0.01857 1.95725 240.0 225.3 M7.39 372.3 828.4 1200.6 0.01860 1.91769 145.0 230.3 399.19 374.2 826.6 1200.9 0.01863 1.87970 1
6
-w.- -,eww e- , .,.. ,1,---,,,e --w-----wwww w w w --y w-----. ww w s w ,. y-------# w e- w,vy---, . - -w,,w ~,.--<,.w..e- s - - , ,,-,m.w--w--,.ev--ww n . .,v-,ve.m-.re-r
Properths of Saturated Steam and Saturated Water--concluded Pressure Temper- Heat of Latent Heat Total Heat Specific Volume Lbs. per Sq. In. the of Abeolute Gage sture Liquid Evaporation of Steam p I . F F o p. siu ns, seuns. siu nd. Wa cu. ri. ,t,er
, is.
Stea cu. ri. ,m,, is 250.0 235.3 400.97 376.1 825.0 7201.1 0.01865 255.0 1.84317 240.3 402.72 378.0 823.3 1201.3 0.01868 1.80802 260.0 245.3 404.44 379.9 821.6 1201.5 0.01870 265.0 1.77418 250.3 406.13 381.7 820.0 1201.7 0.31873 1.74157 270.0 255.3 407.80 383.6 818.3 1201.9 0.C1875 ' e 1.71013 275.0 260.3 409.45 385.4 816.7 1202.1 0.01878 1.67978 240.0 265.3 411.07 387.1 815.1 1202.3 0.01880 1.65049 I 285.0 270.3 412.67 388.9 813.6 1202.4 0.01882 1.62218 l 290.0 275.3 414.25 390.6 812.0 1202.6 0.01885 1.59482 295.0 280.3 415.81 392.3 810.4 1202.7 0.01887 1.56835 300.0 285.3 417.35 394.0 808.9 1202.9 0.01869 1.54274 320.0 305.3 423.31 400.5 802.9 1203.4 0.01899 1.44801 340.0 325.3 428.99 406 8 797.0 1203.8 0.01908 1.36405 g 3M.0 345.3 434.41 412.8 791.3 1204.1 0.01917 1.28910 380.0 365.3 439.61 418.6 785.8 1204.4 C.01925 1.22177 j 400.0 385.3 444.60 424.2 780.4 1204.6 0.01934 1.16095 ' 420.0 405.3 449.40 429.6 775.2 1204.7 0.01942 1.10573 440.0 425.3 454.03 434.8 770.0 1204.8 0.01950 1.05535 460.0 445.3 458.50 439.8 765.0 1204.8 0.01959 1.00921 -e eaJ.0 465.3 462.82 444.7 760.0 1204.8 0.01 % 7 0.96v77 500.0 485.3 # 7.01 449.5 755.1 1204.7 0.01975 520.0 0.92762 505.3 471.07 454.2 750.4 1204.5 0.01982 0.89137 543.0 525.3 475.01 458.7 745.7 1204.4 0.01990 560.0 0.85771 545.3 478.84 #3.1 741.0 1204.2 0.01998 0.82637 540.0 565.3 482.57 467.5 736.5 1203.9 , 0.02006 0.79712 600.0 E3 486.20 471.7 732.0 1203.7 0.02013 O.76975 620.0 M 5.3 489.74 475.8 127.5 1203.4 H0.0 0.02021 0.74408 625.3 493.19 479.9 723.1 1203.0 0.02028 0.71995 M0.0 645.3 496.57 483.9 718.8 1202.7 = 0.02036 0.69724 680.0 us.3 499.86 437.8 714.5 1202.3 0.02043 0.67581 700.0 685.3 503.08 491.6 710.2 1201.8 720.0 0.02050 0.655 % 705.3 506.23 495.4 706.0 1201.4 0.02058 0.63639 740.0 725.3 509.32 499.1 701.9 1200.9 760.0 0.02065 0.61822 745.3 512.34 502.7 697.7 1200.4 0.02072 0.60097 780.0 765.3 515.30 506.3 693.6 1199.9 0.02080 0.58457 800.0 785,3 518.21 509.8 689.6 1199.4 0.02087 0.%8% 820.0 805.3 521.06 513.3 685.5 840.0 1898.8 0.02094 0.55408 825.3 523.86 516.7 681.5 1198.2 0.02101 0.53988 860.0 845.3 526.60 520.1 677.6 1197.7 880.0 0.02109 0.52631 865.3 529.30 523.4 673.6 1197.0 0.02116 0.51333 900.0 885.3 531.95 526.7 M 9.7 11 % .4 0.02123 0.50091 920.0 905.3 534.56 530.0 us.8 1195.7 940.0 0.02130 0.48901 925.3 537.13 533.2 M t.9 1195.1 0.02137 0.47759 960.0 945.3 539.65 536.3 658.0 1194.4 980.0 0.02145 0.46662 965.3 842.14 539.5 654.2 1193.7 0.02152 0.45609 1000.0 985.3 M4.58 542.6 650.4 1050.0 1192.9 0.02159 0.445 % 1035.3 550.53 550.1 640.9 1191.0 0.02177 1100.0 1085.3 0.42224 556.28 557.5 631.5 1189.1 0.02195 0.40058 1150.0 I135.3 M1.82 564.8 622.2 1200.0 1187.0 0.02214 0.38073 1185.3 M7.19 571.9 613.0 1884.8 0.02232 0.36245 1250.0 1235.3 572.38 578.8 # 1.8 1182.6 0.02250 0.34556 1300.0 1285.3 577.42 585.6 548.6 1180.2 0.02269 1350.0 0.32991 l 1335.3 582.32 592.2 585.6 1177.8 0.02288 1400.0 1385.3 0.31536 587.07 598.8 567.5 1175.3 0.02307 0.30178 1450.0 1435.3 591.70 605.3 567.6 1172.9 0.02327 0.28909 1500.0 1485.J 5%.20 611.7 558.4
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1170.1 0.02346 0.27719 1600.0 1585.3 M4.87 624.2 540.3 1164.5 1700.0
. 0.02387 0.25545 1 2 5.3 613.13 636.5 522.2 1158.6 0.02428 1800.0 1785.3 0.23607 621.02 644.5 503.8 1152.3 0.02472 0.21861 1900.0 1885.3 628.56 660.4 485.2 1845.6 0.02517 0.20278 2000.0 1945.3 635.80 672.1 466.2 Ii38.3 2100.0 0.02565 0.18831 2085.3 642.76 683.8 446.7 1830.5 0.02615 2200.0 0.17501 2185.3 649.45 695.5 426.7 1122.2 0.02 % 9 2300.0 2285.3 0.16272 655.09 707.2 406.0 1113.2 0.02727 0.15133 2400.0 1345.3 % 2.11 719.0 384.8 1103.7 0.02790 0.14076 2500.0 24k5.3 % 8.11 7Jl.7 361.6 1093.3 0.02859 0.13068 2600.0 2545.3 673.91 744.5 337 6 1082.0 2700.0 0.02938 0.12110' 2685.3 679.53 757.3 311.3 1069.7 0.03029 28000 2785.3 0.11194 684.% 770.7 285.1 1055.8 0.03134 0.10305 2900.0 2845.3 690.22 785.1 254.7 1039.8 0.03262 0.09420 Jono.O 2945.3 695.J3 801.5 218.4 1020.3 0.03428 0.08500 3100.0 3085.3 700.28 824.0 169.3 993.3 3200.0 0.03641 0.07452 3185.3 705.08 875.5 56.1 931.6 '0.04472 0.05 % 3 3208.2 3193.5 705.47 906.0 0.0 906.0 0.05078 0.05078
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CATEGORY 5 THEORY OF NUCLEAR POWER PLANT OPERATION FLUIDS ANO THERMODYNAMICS
- QUESTION 5-1 (0.5)
MULTIPLE CHOICE - SELECT THE BEST ANSWER Refer to Figure 5-1, which shows two test manometers connected to a liquid waste hold-up tank for testing the tank level instrumentation. The liquid level in the tank ist (a) 5 feet (b) 7 feet (c) 12 f eet (d) 19 feet
- ANSWER (c)
- REFERENCE Thermo-Hydraulic Principles, 2-17 to 26 and 11-27
- KW
- QUESTION 5-2 (0.5)
MULTIPLE CHOICE - SELECT THE BEST ANSWER Refer to Figure 5-1, which shows two test manometers connected to a liquid waste hold-up tank for testing the tank level instrumentation. The pressure in the tank is equivalent to (a) 5 feet of water gauge pressure. (b) 7 feet of water gauge pressure. (c) 27 feet of water absolute pressure. (d) 29 feet of water absolute pressure.
- ANSWER )
(a)
- REFERENCE Thermo-Hydraulic Principles, 2-17 to 26 and 11-27
*KW i 1
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- QUESTION 5-3 (0.5)
MULTIPLE CHOICE - SELECT THE BEST ANSWER Refer to Figure 5-1, which shows two test manometers connected to a liquid waste hold-up tank tor testing the tank level instrumentation. The level instrument is a differential pressure cell with a ref erence leg which has been filled wi th air. As tank level is increased the indication f rom the D/P cell: (a) will not indicate correct initial level but will indicate correct changes in level. (b) will indicate correct initial level but will indicate changes smaller than actual. (c) will not indicate correct initial level and will indicate changes smaller than actual. (d) will indicate correct initial level but will indicate changes in level larger than actual.
- ANSWER ta)
- REFERENCE Thermo-Hydraulic Principles, 2-17 to 26 and 11-27 okW
- QUESTION 5-4 (0.5)
MULTIPLE CHOICE - SELECT THE DEST ANSWER Refer to Figure 5-1, which shows two test manometers connected to a liquid waste hold-up tank for testing the tank level instrumentation. The level instrument is a differential pressure cell with a reference leg which is filled with of water. When the reference leg is colder than the tank the indication from the D/P cell (a) may not indicate correct initial level but will indicate correct changes in level. (b) will indicate correct initial level but will indicate changes smaller than actual. (c) may not indicate correct initial level and will indicate changes smaller than actual. (d) will indicate correct initial level but will indicate changes in level larger than actual.
- ANSWER (a) or (c)
- REFERENCE Thermo-Hydraulic Principles, 2-17 to 26 and 11-27
- KW 2
^
\
- QUESTION 5-5 (0.5) l MULTIPLE CHOICE - SELECT THE BEST ANSWER j i
Refer to Figure 5-2 "Enthalpy vs Entropy for Water" to answer the 1 following question. A line of constant temperature is depicted by lines (a) line A. l (b) line B. (c) line C. (d) line D.
- ANSWER (a)
- REFERENCE Thermal-Hydraulic Principles, 2-69 and 7-8 to 7-16
- KW
- QUESTION 5-6 (0.5)
MULTIPLE CHOICE - SELECT THE BEST ANSWER Refer to Figure 5-3 "Temperature vs Entropy for Water" to answer the following question. A region of compressed liquid is indicated by: (a) region A. (b) region B. (c) region C. (d) region D.
- ANSWER (a)
- REFERENCE Thermal-Hydraulic Principles, 2-69 and 7-8 to 7-16
- KW i
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- QUESTION 5-7 (2.0)
MATCHING - SELECT THE BEST CHOICE Refer to the attached Figure 5-4, which is a simplified temperature - entropy (T-s) diagram of the Diablo Canyon Power Plant steam cycle. The numbered points on the figure are parts of the steam cycle which identify beginning or end points of various processes. Match each process listed below (a through f) with its' corresponding numbers on the T-s diagram. Allowable choices are at least two numbers and may include three numbers. (examples:
"1-2" or "1-2-3" are possible answers while "1" or "1-2-3-4" are not acceptable.) (0.33 each)
(a) Work out of the HP turbine (b) Vaporization in steam generator (c) Feedwater heating (d) Reheater superheating (e) Condensing in the condenser Heat addition by the steam generators I (f)
- ANSWER ,
(a) 4-5 ) (b) 3-4 I (c) 1-2 or 1-2-3 Ewith 3-4 selected for (f ) ] l (d) 6-7 (e) 8-1 (f) 2-3-4 or 3-4 Cwith 1-2-3 selected for (c)]
- REFERENCE Thermal-Hydraulic Principles, 7-80 through 7-91
*KW
- QUESTION 5-8 (1.5)
Refer to Figure 5-5 which is a sketch of a closed cooling water system, the system characteristic curve, and pump P-1 characteristic curve. Both pumps are identical. (a) On Figure 5-5 sketch and label the new pump curve when both pumps P-1 and P-2 are running. Indicate the new operating point. (0.75) (b) On Figure 5-5 sketch and label the new system curve when the throttle valve is partially closed. Indicate the new operating point with pump P-1 running. (0.75)
- ANSWER See attached drawing
- REFERENCE Thermal-Hydraulic Principles, 10-41 THROUGH 10-48
*KW 4
4
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- QUESTION 5-9 (1.5)
Refer to Figure 5-5 which is a sketch of an idealized closed cooling water system, the system characteristic curve, and one pumps
- chare.cteristic curve. Pump P-1 is initially operating at 36 PSID, 1000 GPM, 24 amps, and 1800 RPM. Pump P-2 is off.
(a) If pump speed is changed to 1200 RPM, what is the new flow rate? (0.75) (b) If pump speed is changed to 1200 RPM, what is the new current? (0.75)
- ANSWER (a) N1/N2=Q1/Q2, 12OO/1800=X/1000, X= 667 GPM (0.75)
(B) (N1/N2)**3=P1/P2, 0.296=X/24(power proportional to current), X= 7 AMPS (0.75)
- REFERENCE Thermal-Hydraulic Principles, 10-34 through 10-41
*KW 4
5
l
- QUESTION 5-10 (2.0)
A calorimetric is being conducted to calibrate nuclear instrumentation. The following data has been recorded or is known. Indicated power NIs 99.5% 99.4% 99.5% 99.4% , Feedwater temperature 440 F i Loop----------> 1 2 3 4 Feedwater flows 3.75 3.78 3.77 3.77 (x 10+6 lbm/hr) Steam pressure 994 989 990 990 (PSIG) Reactor coolant pumping power 20 MW Losses to ambient 7 MW Steam generator blowdown O gpm (a) What is the total reactor power in Megawatts? (1.5) (b) If blowdown flow was actually 400 gpm how would the calculated power be affected? (higher, lower, or stay the same) (0.5)
- ANSWER (a) QRx + QRCPs = M(Hst-Hfw) + Oamb (0.4)
QR = 1.507x10+7 * (1193-419) - 20MW + 7MW (0.4) QR = (1.1664x10+10 Btu /hr/3413OOO Btu /hr-MW) - 13MW (0.4) QR = 3418MW - 13MW = 3405MW (0.3) (b) The calculated power would be lower (since the blowdown ; would have a lower enthalpy than steam) (0.5)
- REFERENCE ,
Thermal-Hydraulic Principles, 13-41 through 13-44 l
- KW l I
l I i 6
- QUESTION 5-11 (0.5)
MULTIPLE CHOICE - SELECT THE BEST ANSWER The primary coolant system is maintained at a pH between 4.2 and 10.5. PH ist (a) a measure of the oxidation-reduction potential of a water solution. (b) a measure of the hydrogen ion concentration on a logarithmic scale. (c) a measure of disassociation of water into hydrogen and oxygen as temperature rises. (d) a measure of the chemical activity of dissolved solids in a water solution.
- ANSWER (b)
- REFERENCE Radiation, Chemistry, and Corrosion Considerations, page 6-5
*KW'
- QUESTION 5-12 (0.5)
MULTIPLE CHOICE - SELECT THE BEST ANSWER Corrosion rates of most metals in a water environment will decrease whens (a) flow velocity increases and pH is neutral or slightly higher. (b) flow velocity increases and pH is neutral or slightly lower. (c) flow velocity decreases and pH is neutral or slightly higher. (d) flow velocity decreases and pH is neutral or slightly lower.
- ANSWER (c)
- REFERENCE Radiation, Chemistry, and Corrosion Considerations, page 6-13
*KW 7
1
- QUESTION 5-13 (0.5)
MULTIPLE CHOICE - SELECT THE BEST ANSWER The recombination of oxygen and hydrogen in a nuclear reactor is promoted by: (a) a high temperature. (b) a high gamma flux. (c) a high neutron flux. (d) a high radiation flux.
- ANSWER (b)
- REFERENCE Radiation, Chemistry, and Corrosion Considerations, page 7-5
*KW
- QUESTION 5-14 (0.5)
MULTIPLE CHOICE - SELECT THE BEST ANSWER Lithium hydroxide is added to the primary coolant system to (a) prevent the formation of nitric acid. (b) reduce the disassociation of water to oxygen. (c) control the pH of the coolant water. (d) strip metal surfaces of deposits removed by purification.
- ANSWER (c)
- REFERENCE Radiation, Chemistry, and Corrosion Considerations, page 7-17
*KW
- QUESTION ,
5-15 (0.5) MULTIPLE CHOICE - SELECT THE BEST ANSWER Most of the tritium in the reactor coolant comes frome (a) activation of hydrogen in water. (b) activation of deuterium in water. (c) fission product leaking through the cladding.
- (d) neutron reactions with beren in the r.colant.
- ANSWER (d)
- REFERENCE l Radiation, Chemistry, and Corrosion Considerations, page 7-14 l *KW i
8 i i
- QUESTION 5-16 (0.5)
MULTIPLE CHOICE - SELECT THE BEST ANSWER When reactor power is rapidly increasing (a) reactor period is a large positive value and startup rate is a large positive value. (b) reactor period is a large positive value and startup rate is a small positive value. (c) reactor period is a small positive value and startup rate is a large positive value. (d) reactor period is a small positive value and startup rate is a small positive value.
- ANSWER (c)
- REFERENCE Fundamentals of Nuclear Reactor Physics, page 7-17
- KW
- QUESTION 5-17 (0.75)
MULTIPLE CHOICE - SELECT THE BEST ANSWER A reactor is at 10E-10 amps power and has a start-up rate of 0.5 DPM. How long will it take to reach 10E-6 amps? (a) 2 minutes (b) 4 minutes (c) 8 minutes (d) 20 minutes
- ANSWER (c)
- REFERENCE Fundamentals of Nuclear Reactor Physics, page 7-19
*KW 9
- QUESTION 5-18 (1.0)
MULTIPLE CHOICE - SELECT THE BEST ANSWER A reactor is at 10E-10 amps power and has a start-up rate of + 1.3 DPM. What amount of reactivity must be added to cause the reactor to be just critical? (Note: b= 0.005 and L=0.1) (a) -250 pcm (b) -210 pcm (c) -167 pcm (d) -114 pcm
- ANSWER 26/1.3=20 20=(b-p)/Lp 20= (0. OO5-p ) /O.1 (p ) 2p=0.OO5-p 3p=0.OO5 p=0.OO167 p=+167 pcm with +1.3 DPM (c)
- REFERENCE Fundamentals of Nuclear Reactor Physics. page 7-21 and 7-42
- KW
- QUESTION 5-19 (1.0)
MULTIPLE CHOICE - SELECT THE BEST ANSWER A reactor is subtritical with 10 counts per second indicated, Keff=0.93, hot, no xenon, and all rods in. The shutdown banks, worth 4000 pcm are withdrawn. What is the new count rate? (a) 21 CPS (b) 41 CPS (d) 64 CPS (d) 90 CPS
- ANSWER O. 04= (x-0. 93) /O. 93 (x ) 0. 0372 (x ) =x-0. 93 0. 9626 (x ) =0. 93 x=0.9661 i CRO/CR1=(1-K1)/(1-KO) 10/CR=(0.0339/O.07) CR=21 CPS (a) 21 CPS
- REFERENCE Fundamentals of Nuclear Reactor Physics, Chapter 8 l
- KW l
10
- QUESTION 5-20 (1.0)
Refer to Figure 5-6 which shows negative reactivity being inserted into an initially supercritical core Sketch the resulting (log scale) fission rate as a function of time. Specific numbers are not desired, however the shape and slope of ; the curve describing the fission rate should be shown.
- ANSWER See attached sheet
- REFERENCE Fundamentals of Nuclear Reactor Physics, page 7-67
- KW
- QUESTION 5-21 (1.0)
Refer to Figure 5-7 which shows positive reactivity being inserted into an initially subcritical core. Sketch the resulting fission rate as a function of time. The shape of the curve describing the fission rate and steady state values (if any) should be shown.
- ANSWER See attached sheet
- REFERENCE Fundamentals of Nuclear Reactor Physics, page 8-55
- KW +
- QUESTION 5-22 (1.0)
MULTIPLE CHOICE - SELECT THE BEST ANSWER The difference between delta I and Axial Offset ist (a) delta I =P(lower)-P(upper) while axial offset =P(upper)- P(lower) devided by P(upper)+P(lower) (b) delta I =P (upper)-P (lower) while axial offset =P(upper)- P (lower) devided by P(upper)+P(lower) (c) delta I =P(lower)-P(upper) while axial of f set =P (l ower )- P (upper) devided by P(lower)+P(upper) (d) delta I =P(upper)-P(lower) while axial offset =P(lower)- P(upper) devided by P(lower)+P(upper)
- ANSWER (b)
- REFERENCE Reactor Core Control for Large PWRs, Chapter 8
*KW 11
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- QUESTION 5-23 (0.5)
MULTIPLE CHOICE - SELECT THE BEST ANSWER Unit i has been operating at 50% power for six months. Power is then increased to 100%. Equilibrum xenon wills (a) increase by more than 50%. (b) increase by 50%. (c) increase by less than 50% (d) stay the same.
- ANSWER (c)
- REFERENCE Reactor Core Control for Large PWRs, Chapter 4
*KW
- QUESTION 5-24 (0.5)
MULTIPLE CHOICE - SELECT THE BEST ANSWER Unit 1 has been operating at 50% power for six months. Power is then increased to 100%. Equilibrum samarium will (a) increase by more than 50%. (b) increase by 50%. (c) increase by less than 50% (d) stay the same.
- ANSWER (d)
- REFERENCE Reactor Core Control for Large PWRs, Chapter 4 (4-31)
*KW
- QUESTION 5-25 (0.75)
MULTIPLE CHOICE - SELECT THE BEST ANSWER Xenon burn-out will cause the fastest reactivity change when: 1 (a) power is increased rapidly after a long shutdown. ' (b) power is increased rapidly after a short shutdown. (c) power is increased slowly after a long shutdown. (d) power is increased slowly af ter a short shutdown. ;
- ANSWER i
(b) '
- REFERENCE Reactor Core Control for Large PWRs, Chapter 4
*KW 12 l
- QUESTION 5-26 (1.5)
Diab}o Canyon Unit 2 is initially operating at beginning of life, 2 GWD/MTU, 70% power, 180 steps on control bank D, 1200 ppm baron, Tave is programed normally with the RCS in auto. Use the attached Figures 5-8, 5-9, 5-10, 5-11, 5-12,and 5-13 to answer the following. With no rod motion what change in boron concentration would be necessary to increase power to 90% over a period of 30 minutes?
- ANSWER 70% to 90% G12OOppm = 850 to .'.075 pcm = 225 pcm
-123 ppm /Xdp x 0.225 = [-28 +/-3 ppm 3 or 225 pcm/10 pcm/ ppm = E-22.5 +/-3 ppm 3
- REFERENCE Reactor Core Control for Large PWRs, Chapter 5 and 9
*KW
- QUESTION 5-27 (1.5)
Diablo Canyon Unit 2 is initially operating at beginning of life, 2 GWD/MTU, 70% power, 180 steps on control bank D, 1200 ppm boron, Tave is programed normally with the RCS in auto. Use the attached Figures 5-8, 5-9, 5-10, 5-11, 5-12,and 5-13 to answer the following. What new rod position would be necessary to decrease power to 50% in 30 minutes?
- ANSWER 70% to 50% power is 850 to 615 pcm = -235 pcm
-260 pcm 9180 steps to -495 pcm CG137 steps +/-5 steps bank D3
- REFERENCE Reactor Core Control for Large PWRs, Chapter 6 and 9
*KW 13
DIABLO CANYON POWER PLANT OPERATION DATA FIGURE 5 - 9 DIFFERENTIAL AND INTEGRAL ROD WQRTH VS. STEPS WITHDRAWN, BANKS D, C. B. UN AND A MOVING WITH 100 STEP OVERLAP, AT BOL, HFP, EQUILIBRIUM XENON CYCLE 2 FOR BURNUP < 7500 MWD /T
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. I jis pn SOURCE: WCAP - 11450, Rev. O. Figure 6.1 Revision 1 Date: 06/08/87 e>
1 I DIABLO CANYON POWER PLANT OPERATION DATA 1 TABLE 5-9 UNIT HZP INTEGRAL WORTH AS A FUNCTION OF STEPS WITHDRAWN FOR BANKS D AND C WITH 100 STEP OVERLAP (Cycle 2) Steps Withdrawn 0-7500 MWD /T BOL. HZP 6500-EOL MWD /T Predicted Predicted BANK C BANK D Integral Worth Integral Worth 228 228 -0.0 .-0.0
. 228 220 -10.0 -50.0 228 210 -50.0 -160.0 228 200 -115.0 -320.0 228 190 -190.0 -500.0 228 180 -260.0 -630.0 228 170 -320.0 -760.0 228 160 -380.0 -870,0 228 150 -430.0 -940.0 228 140 -480.0 -1020.0 228 130 -530.0 -1080.0 228 120 -595.0 -1120.0 228 110 -650.0 -1160.0 228 100 -705.0 -1190.0 218 90 -780.0 -1250.0 208 80 -860.0 -1370.0 198 70 -960.0 -1500.0 188 60 -1060.0 -1610.0 178 50 -1170.0 -1730.0 168 40 -1300.0 -1840.0 158 30 -1420.0 -1910.0 148 20 -1515.0 -1990.0 138 10 -1600.0 -2040.0 128 0 -1660.0 -2080.0 118 0 -1705.0 -2100.0 108 0 -1740.0 -2120.0 98 Rod Insertion Limit 0 -1790.0 -2150.0 SOURCE: WCAP - 11450 Rev. O Figure A.4. Figure A.5.
Revision 2 Date: 06/08/87
1 I . 1 DIABLO CANYON POWER PLANT OPERATION DATA j FIGURE 5-/o i TOTAL POWER DEFECT AS A FUNCTION OF POWER LEVEL AT BOL UNIT l CYCLE 2 FOR BURNUP 0-5000 MWD /HTV sm - ' iii ; , . . i . III i !' .. I i . i!l. Ii;i ;ii i!.. ii: i i i.. !!!i i
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. ii;; ii i i'i i+ii :i ! Iiit i;i ieii /: ;t/t / 1200
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0 14 M M 64 M M M M M im P0ftit LEYEL (PDCEMI) l SOURCE: WCAP - 11466. Rev. O. Figure 4.1 1 1 Revision 1 Date: 06/08/87 ' l l l
i I DIABLO. CANYON POWER PLANT OPERATION DATA FIGURE $ - /l INVERSE BORON WORTH AT HFP, ARO, EQUILIBRIUM XENON VS. BURNUP UNIT
-. CYCLE 2
-* , i . i ... ... . ,i i ,i . .
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$ 0 3 J 4 8 4 I $ $ F) Il il 13 14 t$
RM7 (sv0/viv) SOURCE: WCAP - 11450 Rev. O. Figure 5.14
. Revision 1 Date: 06/08/87 D
DIABLO CANYON POWER PLANT - OPERATION DATA FIG uR E 5-/2 CYCLE 2 GOOD FOR 0-5000 MWD /T 90L. DORON FREE IENON WORTH (PQt) YS. 71NE FOLLINING PLANT TRIP AFTER STEADY 57 ATE OPERATION POWER LtyEL TIME AFTER PLANT TRIP (> CURS) 0 2 4 8 8 10 12 14 18 18 20 25 30 35
~
100 -3379 -4753 -5584 -9013 -8151 -9083 -8872 -5508 -5208 -4817 -4417 -3451 2914 -1937 55 -3345 -4467 -5443 -8871 -5998 -5920 -5717 -5418 9008 -4485 -4294 -3353 -2539 -1881 50 -3313 -4582 -8343 -5729 -5445 -5789 -5H2 -5209 -4924 -4552 -4171 -32H 2464 -1828 85 -3281 -4494 -8222 -5587 -M92 -M13 -5407 -5110 -4782 -4419 -4049 -3158 -2300 -1770 50 -3249 -4419 -5101 -5445 -5539 -54H -5252 -4970 -4640 -4287 -3929 -3001 -2315 -1714 73 -3197 -4288 -4933 5250 -8331 -5244 -5043 -4708 -4440 -4104 -2781 -2930 -2215 -1940 70 -3148 -4164 -47M -5058 ~5122 -5031 -4834 -4He -4258 -2930 -M 98 -2799 *2115 - 1 HS 95 -3093 -4044 -4591 -4881 -4914 -4819 -4424 -4M5 -4087 -3758 -3431 -2949 -2015 -1490
. DO -3040 -3922 -4431 -4 ate -4705 -4007 -4415 -4153 -3870 -3873 -3296 2538 -1915 -1418 55 -2949 -3749 -4208 -4410 -4435 -4334 -4147 -3908 -M34 -3347 -3058 -2374 -1790 -1322 50 -2857 -M75 -2978 ~4154 -4105 -4001 -3879 -M49 -3382 +3122 -2850 -3209 -tH4 -1229 45 -2796 -3402 +3754 -3898 -3894 -3787 -3812 3393 -3149 *28M *1942 -2045 -1530 -1135 40 -2675 -3229 -MIS -3642 -3824 -M t 4 -3344 -S t M -2007 -1670 -2434 -1880 -1414 -1042
!S -2494 -2994 -3215 -3302 -3375 -3158 -3000 -2818 -9000 -2364 -2180 -1982 -1203 -930 30 -2314 -2703 -2002 -2M3 -2929 -2822 -2874 -2499 -2211 2118 -1927 -1484 -1113 -819 25 -2133 -2441 -2589 -2923 -2577 -2475 -2339 -2181 -2013 *1842 -1973 -1285 -H2 -707 30 -1953 2178 -2276 -2383 -2228 -2129 -1003 -1842 -1714 - t MS *1420 -1087 812 Ste it -1815 -1770 -1831 -1824 -1771 -1986 -1582 -1487 -1348 -1320 -1913 -850 -834 -484 10 -1277 -1362 -1248 -1385 -1314 -1243 -1181 -1072 -982 -893 -807 -813 -454 -333 5 -838 -881 -993 -882 -957 -022 580 -534 -401 -448 -403 -307 228 -187 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 l
l l SOURCE: WCAP - 11466 Rev. O. TABLE 2.1 Revision 2 __ Date: 06/08/87
, _]
1
- DIABLO CANYON POWER PLANT OPERATION DATA
. B0 RATION N0tiOGRAPHS Figure 6-/3 BORON DILUTION 10 ,
10 y . M in - 8.33 Gr, 20 20 50 50 S - 100
- = ,
- g - .
200 3 . W 100 e w
- u * -
- E .
I E, 500
~ .. 200 5 , b 8
m a 1000
,3
- a =
y - g 10 5 a
, 2000 3
= - -
E 500 - - -
- E a.
5
.a 5000
- =
a ._. ,
- 100 ;
10,000
- m -
]
o 1000 - g -_._ l
- " 20,000
- -- y l
= -
2000 - 1000 50,000 2000 --- 100.000 I 3000 3000 REFER TO TABLE 1 FOR CORRECTION FACTORS
%w e e REVISION O DATE 6/3/77
- QUESTION 5-28 (1.0) -
Diablo Canyon Unit 2 is initially operating at beginning of life, 2 GWD/MTU, 70% power, 100 steps on control bank D, 1200 ppm baron, Tave is programed normally with the RCS in auto. Use the attached Figures 5-8, 5-9, 5-10, 5-11, 5-12,and 5-13 to answer the following. With power decreased to 50%, what in the expected change in equilibrum xenon?
- ANSWER G70% Eq. Xe = -3145 pcm G50% Eq. Xe = -2857 pcm CChange = +288 +/-10 pcm3
- REFERENCE Reactor Core Control for Large PWRs, Chapter 4 and 9
*KW
- QUESTION 5-29 (0.5)
Diablo Canyon Unit 2 is initially operating at beginning of life, 2 GWD/MTU, 70% power, 180 steps on control bank D, 1200 ppm boron, Tave is programed normally with the RCS in auto. Use the attached Figures 5-8, 5-9, 5-10, 5-11, 5-12,and 5-13 to answer the following. How much dilution water is required to decrease boron concentration by 40 ppm?
- ANSWER See Nomograph 2000 gallons
- REFERENCE Reactor Core Control for Large PWRw, Chapter 4 and 9
*KW i
l l < END OF CATAGORY 5 I GO ON TO CATEGORY 6 14
1 i CATAGORY 6 - PLANT SYSTEMS j DESIGN, CONTROL, AND INSTRUMENTATION ' i
- QUESTION 6-1 (1.5)
The Reactor Trip Breakers are located physically alongside the Reactor Trip Bypass Breakers. (a) What is the purpose of the bypass breakers? (0.75) (b) If bypass "A" and bypass "B" breakers were racked in at the same time, what would happen to the reactor trip breakers? (0.75)
- ANSWER (a) The bypass breakers allow testing the SSPS without tripping the reactor. (0.75)
(b) The reactor trip breakes would trip. (0.75)
- REFERENCE B-6b, Pg. 23
- KW
- QUESTION 6-2 (1.0)
Operating Procedure A-61 limits VCT pressure to a minimum value prior to starting an RCP and during RCP operation. (a) What is thin minimum pressure? (0.5) (b) What is the purpose of this minimum pressure? (0.5)
- ANSWER (a) 15 PSIG (0.5)
(b) This ensures sufficient backpressure on the
#1 seal to force adequate flow through the
#2 seal to meet its' cooling and lubrication 1 requirements. (0.5) l
- REFERENCE '
A-6 Pg. 33 and A-6I
- KW 1
i
- QUESTION 6-3 (2.0)
The Reactor Vessel Level Indication System has three indicating ranges; Upper Range, Full Range, and Dynamic Range. (a) When is each range used? (1.0) (b) What is the indication range in the reactor vessel for each range? (1.0)
- ANSWER (a) (0.33 mach)
Upper - during natural circulation (or when the RCP in the loop with the hot leg connection is not operating) Full - during natural circulation Dynamic - when any combination of P. cps are running. (b) (0.33 each) Upper - hot leg to top of vessel Full - bottom to top of vessel Dynamic - bottom to top of vessel
- REFERENCE A-2d, Pg. 16 - 17
*KW *DUESTION 6-4 (2.5)
Low Temperature Overprescure Protection (LTOP) is provided for the reactor coolant system using PORVs 455C and 456. (a) What is the purpose of LTOP7 (0.5) (b) What are the two plant inputs to the LTOP ' system? (include setpoints) (1.0) (c) What two requirements must be met to arm a PORV for LTOP protection? (1.0)
- ANSWER (a) Low pressure overpressure protection prevents brittle fracture in the RCS. (0.5)
(b) Cold leg temperature 320-330 F (0.5) i Wide range RCS pressure 435-450 PSIG (0.5) l (c) Low setpoint protection cutout switch is in CUT IN and RCS temoerature at or below 328-330 F. (1.0) ,
- REFERENCE A-4a, Pg. 49 - 51 & A-1, Pg. 28 - 29
*KW 2
\
Il
- QUESTION 6-5 (2.0)
The Unit 1 reactor is operating at 50% power and all systems are operating normally. The pressurizer pressure controller is in AUTO and the PDP is operating. Unknown to the licensed operators, a trainee turns the potentiomenter on the pressurizer hand controller (HC-455K) from 8 turns to 10 turns (ies to 100%). (a) How does this action affect the controlling ' channel setpoint? (0,5) (b) With no operator action, what will initially happen? (0.5) (c) With no operator action, what will cause the RCS inventory to decrease? (1.0)
- ANSWER (a) Setpoint is increased to 2500 PSIG (0.5)
(b) Backup and proportional heaters will go on (0.5) (c) Increased pressure causes PORVs to open (PZR level f alls due to charging inability to keep up) (1.0)
- REFERENCE A-4a, Pg. 30 - 39
*KW a
l i i 3
I
- QUESTION 6-6 (3.0)
Reactor unit 1 is operating at 100% power and all systems are in automatic and/or normal line up. The loop 1-1 T(cold) fails high. (a) State four of the seven immediate alarms that would occur? (1.0) (b) How will the Rod Control System be affected? (0.5) (c) Why will the steam dump system NOT actuate? (0.5) (d) How will the operator determine that the T(c) RTD failed high? (1.0)
- ANSWER (a) (0.25 each for any four)
T(ref) deviation from auctioneered high T(ave) Delta T deviation High T(ave) alarm Overtemperature delta T channel activated Overpressure delta T channel activated T(ave) deviation from auctioneered high T(ave) Protection channel activated (b) The Control Rod Drive System will drive rods in to counteract the apparent high T(ave). (0.5) (c) The steam dump system will have a demand, but will not be armed. (0.5) (d) Compare delta-T and T (ave) indications for each loop. High T(ave) with low delta-T indicates a failed high T(c). (1.0)
- REFERENCE Lesson Plans A-2c, A-3a, C-2b and AP-5, PK-401 402 403 etc.
*KW l
l 4 1
- QUESTION 6-7 (2.0) ,
l The Auxiliary Salt Water System of unit 1 is operating normally i l with pump 1-1 operating. , (a) What is the position of the following switches to allow , auto-start of pump 1-27 (1.0) l Control room control switch ' l l Control room standby selector switch Hot shutdown pannel switch l (b) With no SI, what two automatic signals will start ASW pump 1-27 (1.0)
- ANSWER (a) (0.33 each)
Control room control switch - neutral Control room standby selector switch - Auto Hot shutdown pannel switch - Control Room or Remote (b) (0.5 each for any two of the following) Less than 40 PSIG discharge pressure Low voltage on the opposite bus. Bus transfer to startup Bus transfer to diesel
- REFERENCE E-5, Pg. 12 - 14
*KW l
1 1 5
- QUESTION ,
6-8 (2.5) The Reactor Protection System is designed to protect the reac*.or , from specific events by tripping the reactor and the turbine. The bases for the trips are described in the Technical Specifications. For each of the following reactor trips; what is the event the trip is designed to mitigate? (0.5 each) (a) Intermediate range hign neutron flux trip. , (b) Reactor trip initiating a turbine trip. (c) Undervoltage and underfrequency RCP bus trips. Overpower delta-T trip \ (d) (e) Overtemperature delta-T trip
- ANSWER (0.5 each)
(a) Intermediate range high neutron flux trip. Core protection during startup (subtritical) from a continuous rod (cluster) withdrawl event (b) Reactor trip initiating a turbine trip. Prevents the reactivity insertion that would otherwise result from cooldown and avoids unnecessary Safety Injections, i (c) Undervoltage and underfrequency RCP bus trips. Core protection against END as a result of complete loss of forced coolant flow (d) Overpower delta-T trip Provides assurance of fuel integrity, no fuel pellet cracking or melting for ovde power cc:iditions such as steam line breaks. (e) Overtemperature delta-T trip Core protection against DNB for all combinations of pressure, power, temperature, and axial power I distribution. (for slow transients)
- REFERENCE Technical Specifications Bases, pages B 2 B 2-7
*KW
/ 6 4
- QUESTION 6-9 (2.0)
The Process Radiation Monitoring System provides both alarms and and automatic operation of some components. For each of the monitors listed below; what automatic action (if any) occurs on a high radiation condition? (0.33 each) (a) Plant vent Particulate monitor RE-28A&B (b) Plant vent Iodine monitor RE-24 (c) Component cooling water monitor RE-17A&B (d) RHR Heat exchanger compartment exhaust monitor RE-13 (e) Condensate demineralizer monitor RE-16 (f) Steam generator blowdown monitor RE-23
- ANSWER (0.33 each)
P (a) Plant vent Particulate moniter - containment ventilation isolation (b) Plant vent Iodine monitor - alarm only (c) Component cooling water monitor - Surge tank vent clsoes (d) RHR Heat exchanger compartment exhaust particulate monitor -
) alarm only 1
(e) DELETED - UNIT ONE HAS ALARM ONLY UNIT TWO SHUTS FVC-161 INSTRUMENTATION NOT FULLY FUNCTIONAL AT THIS TIME (f) Stuam generator blowdown monitor - blowdown and sample ' isolation valves shut and blowdown tank outlet transf ers from outfall to EDR (FCV-498 & 399 swap to EDR)
- REFERENCE G-4, Pg. 28 - 37
*KW i
l l i I' 7 1 i
, l
s
-n
.e Y
~
- QUESTION 6-10 (2.5) _
Unit 2 is initially at 75% power, with all controls in automatic and/or nornmal. With no operator action, how will each of the following events affect the plant? (Only one item at a time., Describe the plant response and control rod motion until the reactor trips or the unit stabilizes.) (a) One atmospheric dump valve opens. (1.0) o f
..s s
, (b) Turbine load is decreased 5%. (0.75) ,
. N- ,
W) One roci drops into the core. (0.75) -
*ANSWFP -
;- a.
(a) drie' atdospheri . dump valve ope w - steam demand incru:as$s, - T (ave) /T (ref ) mismatch p ages.g#Pode move out to increase power at new steam demand.T[^, .. - .
, J1.0)
(b) : Turbine load is decreas'e'd5 t - steam demand decreases, T (ave) /T (ref ) missmatch, r o s,r - fn o v e in to decIease power (or
~', temperature). ,, (0.75)
(c) One rod drops into the core power decreases, T(h) and T(ave) decrease, T(ave)/T(ref) mismatch, rods move out to ! re-establish T(ave) (or inscrease power). (0.75)
- REFERENCE -
Lesson Plan A-3a, Pg. 15 - 19 & 3Q -- 35
*KW
,-mg ,
/
} 4
'G 7
,, /
LT 1 4 8 i
- QUESTION 6-11 (0.5)
MULTIPLE CHDICE - SELECT THE BEST ANSWER When the F.CP seal water return stop valves are shutt (a) RCP leakoff and excess letdown flow to the reactor coolant drain tank via the excess letdown heat exchanger relief valve. (b) RCP leakoff and excess letdown flow to the pressurizer relief tank via the excess letdown heat exchanger relief valve. (c) RCP leakoff and excess letoown flow to the reactor coolant ! I drain tank via the seal leakoff containment relief valve. (d) RCP leakoff and excess letdown flow to the pressurizer relief tank via the seal leakoff containment relief valve.
- ANSWER (d) I
- REFERENCE j Lesson Plan B-la, Pg. 44 '
*KW
- QUESTION 6-12 (0.5)
MULTIPLE CHOICE - SELECT THE BEST ANSWER From inside containment excess letdown can be diverted to: (a) The seal water heat exchanger or the reactor coolant drain tank. (b) The seal water heat exchanger or the pressurizer relief tank. , (c) The volume control tank or the reactor coolant drain tank. (d) The volume control tank or the pressurizer relief tank.
- ANSWER (a)
- REFERENCE B-lb, Pg. 45
*KW 9
- QUESTION 6-13 (1.0)
How is pump runout protection provided for the auxiliary feedwater pump 1-37
- ANSWER The auxiliary feedwater level controlers receive signals from the SGWLC transmitter AND a pressure transmitter in the discharge piping. As the pressure in the discharge piping decreases the signal to the level control valve is biased to close the LCV.
- REFERENCE D-1, ;. 20
*KW
- QUESTION 6-14 (1.0)
What are the two automatic start signals for auxiliary feedwater pump 1-17 (include logic)
- ANSWER Undervoltage on 1 of 2 channels both 12KV buses Lo-Lo level on 2 of 3 channels in 2 Of 4 steam generators
- REFERENCE D-1, Pg. 24
*KW
- QUESTION 6-15 (1.0)
What prevents paralleling an instrument AC inverter and a back-up power supply voltage regulator through a vital instrument AC distribution pannel?
- ANSWER The two supply breakers are mechanically interlocked to prevent both from being closed at one time.
- REFERENCE J-10, Pg. 9
*KW END OF CATAGORY 6 GO ON TO CATEGORY 10
CATAGORY 7 PROCEDURES: NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL CONTROL l
- QUESTION 7-1 (1.0)
MULTIPLE CH01CE - SELECT THE BEST f.NSWER l l In the event emergency boration is necessary and the VCT make-up I system is unavailable, the prefered method of baration, in order of preference in Abnormal Operating Procedure DP AP-6, Emergency Boration, is: 1 (a) (11 the manual emergency borate valve (8471), (2) refueling ! water storage tank, (3) boron injection tank, and (4) emergency ( boration valve (8104). (b) (1) emergency boration valve (8104)-, (2) refueling water storage tank, (3) boron injection tank, and (4) the manual emergency borate valve (8471). (c) (1) emergency boration valve (8104), (2) the manual l emergency borate valve (8471), (3) boron injection tank, and (4) refueling water storage tank. (d) (1) boron injection tank, (2) refueling water storage tank, (3) emergency boration valve (8104), and (4) the manual emergency borate valve (8471).
- ANSWER (b)
- REFERENCE OP AP-6
- KW l
1 l 1
)
- QUESTION 7-2 (2.0)
Based on Abnormal Operating Procedure DP AP-6, Emergency Boration, for each of the situations listed below, when can emergency boration be stopped? l (0.5 each) l (a) Two stuck control rods following a SCRAM (b) Control rods below rod insertion limit at power (c) Shutdown margin less than that required by Technical Specifications when shutdown (d) Criticality achieved inadvertently below the rod insertion limits
- ANSWER (a) 200 ppm (100 per stuck rod) (0,5)
(b) Borate until rods above RIL (0.5) (c) Borate until shutdown margin restored (0.5) (d) 100 ppm (0.5)
- REFERENCE OP AP-6, Pg. 4
- KW
\
l 1 1 1 2
i i
- QUESTION 7-3 (3.0)
Unit 1 is being heated up from cold shutdown to hot stand-by per Operating Procedure OP L-1, Plant Heatup from Cold Shutdow to Hot Stanoby. The plant is at 325 PSIG and 100 F. (a) Based on Technical Specifications, what three plant conditions identify the hot shutdown mode? (1.0) (b) What three steps (operations) are used in L-1 to establish a steam bubble in the pressurizer? (1.0)
'c) . What indications would you look for to verify that a steam bubble is forming in the pressurizer? (1.0)
- ANSWER (a) K(eff)<O.99, 0% power, and 200 F <T(ave)< 350 F (1.0)
(b) Energize pressurizer heaters, place PCV-135 in AUTO, and reduce charging (to 35 GPM). (1.0) (c) Letdown flow is greater than charging flow and pressure is stable or increasing. (1.0)
- REFERENCE OP L-1
*KW 3
- QUESTION 7-4 (3.0) j
' I Unit 1 is being heated up from hot shutdown to hot stand-by per Operating Procedure OP L-1, Plant Heatup from Cold Shutdown to ,
Hot Standby. The plant is at 950 PSIG and 400 F. fa) Based on Operating Procedure OP A-6:I, Reactor Coolant Pumps - Place in Service, what two conditions would require opening the No. 1 seal bypass valve (8142)7 (1.0) (b) What four conditions must be met before opening the No. 1 seal bypass valve? (2.0)
- ANSWER (a) The pump radial bearing outlet temperature or the No. 1 seal leakoff temperature approach their alarm level. (1.0)
(b) (0.5 each) RCS pressure is greater than 100 and less than 1000 PSIG The No. 1 seal leakoff valves (8141a,b,c,d) are all oper No.1 seal leakoff flow is less than 1 GPM Seal injection to each seal is greater than 6 GPM
- REFERENCE OP A-6:I, Pg. 2
*KW l 4 l .. - -
l l
- QUESTION 7-5 (2.5)
Unit 2 is being started up per Operating Procedure OP L-2, Hot Standby to Minimum Load. The plant is at 2250 PSIG and 547 F. Refer to Figure 7-1, which shows rod integral worth as a function of steps withdrawn. The ECP is 80 steps on bank D. The Control Operator has just declared the reactor critical at 60 steps on bank C. (a) What two problems will cause you (as the SRO) to direct that all control bank rods be inserted? (1.0) (b) Besides inserting all contr - banks, what are three of the other four activos required by OP L-2? (1.0) (c) Whose permission must be obtained prior to continuing the start-up? (0.5)
- ANSWER (a) Criticality below rod insertion limits and criticality below ECP-120 (1.0)
(b) (0.33 each for any three) Unblock the high flux et shutdown alarms Determine the RCS baron concentration by analysis Recalculate the ECC Notify plant management Emergency borate 100 ppm (c) The Plant Superintendent (0.5)
- REFERENCE OP L-2, Pg. 9
- KW
)
5 j
DIABLO CANYON POWER PLANT OPERATION DATA
};'I G L)(?Ei 7~l L)ffl1f HZP INTEGRAL WORTH AS A FUNCTION OF STEPS WITHDRAWN FOR BANKS D AND C WITH 100 STEP OVERLAP (Cycle 2)
Steps Withdrawn 0-7500 MWD /T BOL. HZP 6500-E0L MWD /T Predicted Predicted DANK C BANK D Integral Worth Integral Worth 228 228 -0.0 .-0.0 ~ 228 220 -10.0 -50.0 228 210 -50.0 -160.0 228 200 -115.0 -320.0 228 190 -190.0 -500.0 228 180 -260.0 -630.0 228 170 -320.0 -760.0 228 160 -380.0 -870.0 228 150 -430.0 -940.0 228 140 -480.0 -1020.0 228 130 -530.0 -1080.0
. 228 120 -595.0 -1120.0 228 110 -650.0 -1160.0 228 100 -705.0 -1190.0 218 90 -780.0 -1250.0 208 80 -860.0 -1370.0 198 70 -960.0 -1500.0 188 60 -1060.0 -1610.0 178 50 -1170.0 -1730.0 168 40 -1300.0 -1840.0 158 30 -1420.0 -1910.0 148 20 -1515.0 -1990.0 138 10 -1600.0 -2040.0 128 0 -1660.0 -2080.0 118 0 -1705.0 -2100.0 108 0 -1740.0 -2120.0 98 Rod Insertion Limit 0 -1790.0 -2150.0 SOURCE: WCAP - 11450 Rev. O Figure A.4, Figure A.S. ~^70 - -
Date: 06/08/87
- QUESTION 7-6 (2.0)
Valves and other equipment are positioned or regulated by Hagen controllers. Based on Operating Order D-2, Operation of Hagen Controllers, each controller has two power supplies; AUTO and MANUAL. (a) With a Hagen controller in MANUAL and manual power is lost, how will the controller respond? (0.5) (b) With a Hagen controller in AUTO and manual power is lost, how will the controller respond? (0.5) (c) With a Hagen controller in AUTO and manual power is lost, how will the controller respond after manual power is restored? (0.5) (d) How can you identify a loss of both manual and auto power to the controller? (0.5)
- ANSWER (a) Controller goes to AUTO-HOLD (0.5)
(b) Controller goes to AUTO-HOLD (0.5) (c) Controller goes to MANUAL (0.5) (d) Controller output goes to zero and lights on the controller go out. (0.5)
- REFERENCE OP O-2, Pg. 2
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- QUESTION 7-7 (2.0) )
While starting up Unit 2 the Control Operator selects two nuclear instruments channels to be recorded on recored NR-45. Operating Procedure OP L-2, Hot Standby to Minimum Load, provides guidance as to what channels are to be selected. For each of the following conditions, what channels should the CD select to record on NR-45? (a) Withdrawing rods for criticality (1.0) (b) Reactor power stable at 10E-8 AMPS to take data (1.0)
- ANSWER (a) Select the highest source channel and one intermediate channel. (1.0)
(b) Select one intermediate range and one power range. (1.0)
- REFERENCE OP L-2, Pg. 6 and 11
- KW
- QUESTION 7-8 (2.0)
Technical Specification Limiting Condition for Operation 3.4.8 establishes the maximum reactor coolant activity limits. (a) What are the two specific activity limits for the Reactor Coolant System 7 (1.0) (b) What are the Technical Specification Bases for these limits? (1.0)
- ANSWER (a) 1 microcurie / gram dose equavaient I-131 and 100/E microcuries/ gram gross activity (1.0)
(b) The site boundary 2 hour dose will not exceed a small fraction of 10CFR100 dose guideline values following a steam generator tube rupture with an assumed previous 1 GPM S/G tube leak. (1.0)
- REFERENCE T.S. B3/4 4-5 and 3/4 4-25 !
*KW !
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- QUESTION 7-9 (0.5)
MULTIPLE CHOICE - SELECT THE BEST ANSWER Fo11cwing a control room fire, the control room is being evacuated in accordance with OP A-8, Control Room Inaccessibility. The Unit 2 control operator is responsible for: (a) Turbine building watch Unit 1 (b) Auxiliary building watch Unit 2 (c) Unit 2 hot shutdown pannel operator (d) Superviscr of Unit 1 shutdown
- ANSWER (c)
- REFERENCE OP A-B
- KW
- QUESTION 7-10 (0.5)
MULTIPLE CHOICE - SELECT THE BEST ANSWER Following a control room fire, the control room is being evacuated in accordance with OP A-8, Control Room Inaccessibility. The Unit 2 assistant control operator is responsible for (a) Turbine building watch Unit 1 (b) Auxiliary building watch Unit 2 (c) Unit 2 hot shutdown pannel operator j (d) Supervisor of Unit i shutdown
- ANSWER (a)
- REFERENCE OPA-8
- KW I 1
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- QUESTION 7-11 (0.5)
MULTIPLE CHOICE - SELECT THE BEST ANSWER Following a control room fire, the control room is being evacuated in accordance with OP A-8, Control Room Inaccessibility. The Shift Seoervisor is responsible fort (a) Turbine building watch Unit 1 (b) Auxiliary L'uilding watch Unit 2 (c) Unit 2 hot shutdown pannel operator (o) Supervisor of Unit 1 shutdown
- ANSWER (d)
- REFERENCE OP A-B
- KW l
- QUESTION l
7-12 (1.5) A survey of radiation and contamination levels has been made by Health physics personnel. Figure 7-2 is a map which documents the results of this survey. Figure 7-3 is an extract from 10CFR2O ! Appendix B, Concentrations in Air and Water Above Natural Background. Answer the following questions based on 10CFR20. (a) What posting is required for the pump room? (0.5) l 1 (b) What posting is required f or the tank room? (0.5) I (c) What posting is required for the valve room? (0.5) (d) Which area (s) must be locked or access controlled per 10 CFR 207
- ANSWER (a) Caution - Radiation Area (0.5)
(b) Caution - High Radiation area (0.5) (c) Caution - Airborne Radioactivity Area (0.5) (d) DELETED - POSSIBLE AMBIGUITY
- REFERENCE 10CFR2O
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- QUESTION 7-13 (2.0)
Following a reactor trip and safety injection, Emergency Procedure E-0, Reactor Trip or Safety Injection and the associated Emergency Operating Procedures have been implemented. (a) At what two times are the Critical Safety Function Status Trees required to be monitored? (1.0) (b) What are the two reactor coolant pump trip criteria per the "foldout page"? (1.0)
- ANSWER (a) When directed by E-0, and after transition out of E-0. (1.0)
(b) Trip all RCPS within 5 minutes of a Phase B actuation, and trip all RCPs when either one CCP or one SI pump is running and WR RCS pressure is less than 1275 PSIG. (1.0)
- REFERENCE EP E-0, Pg. 13 & 19
- KW
- QUESTION 7-14 (2.0)
Following a reactor trip and safety injection, Emergency Procedure E-0, >1 actor Trip or Safety Injection and the associated Emergency Operating Procedures have bEen implemented. Subsequently, E-1.1, SI Termination, is entered. (a) What are the two SI re-initiation criteria? (1.0) (b) What are the three adverse contaimnment l criteria? (1.0) i
- ANSWER (a) RCS subcooling less than 20 F or pressurizer level cannot be maintained above 4%C2OJX (1.0)
(b) Containment pressure above 3 PSIG Greater than 10E5 R/Hr dose rate Greater than 10E6 R dose (1.0)
- REFERENCE EP E-1.1, Foldout
- KW END OF CATEGORY 7 GO ON TO CATEGORY 8 10
CATAGORY 8 ADMINISTRATIVE PROCEDURES PRECAUTIONS AND LIMITATIONS
- QUESTION 0-1 (0.0)
MULTIPLE CHOICE - SELECT THE BEST ANSWER Operating procedures AP-3, Minor Steam Generator Tube Failure, and AP-3A, Steam Generator Tube Leak, both describe actions to be taken when steam generator leakage is discovered. Regarding the actions to be taken when implementing these procedures: (a) AP-3 requires reactor shutdown while AP-3A does not require shutdown. (b) Both AP-3 and AP-3A require shutdown, but with different time requirements. (c) Neither AP-3 nor AP-3A require that the reactor be shutdown. (d) AP-3 does not require reactor shutdown while AP-3A requires reactor shutdown.
- ANSWER DELETED - REFERENCES NO LONGER VALID l l
- REFERENCE OP AP-3 and AP-3A, Pg. 1
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- QUESTION 8-2 (3.0)
Technical Specification Limiting Condition for Operation 3.4.6.2 states the limits f or Reactor Coolant System leakage. After returning from a vacation you are informed of the following RCS leakages. The unit is being shutdown and is in mode 2 at 2250 PSIG. 0.8 GPM through RCS pressure isolation valves (past dist(S)) 3.1 GPM through a valve stem packing 5.0 GPM through a pressurizer relief 0.3 GPM tube leaks in steam generator 1-3 (f or last 48 hours) 0.9 GPM tube leaks in steam generator 1-1 (f or last 24 hours) 0.1 GPM RCS leakage through a socket weld on the pressurizer O.8 GPM unidentified leakage 9.1 GPM RCP 1-1 seal injection flow 11.1 GPM RCP 1-2 seal injection flow 15.7 GPM RCP 1-3 seal injection flow 9.2 GPM RCP 1-4 seal injection flow What leakage (s) exceed Technical Specification LCO limits? (Include the leakage specification for each out of LCO condition)
- ANSWER (1) Greater than 40 GPM controlled leakage (45.1 GPM) (0.60)
(2) No pressure boundary leakage allowed (0.1 GPM) (0.60) (3) Greater than 1 GPM through all steam generators (1.2 GPM) (0.60) (4) Greater than 500 GPD through one steam generator (0.9 GPM) (0.60) (5) Greater than 10 GPM identified leakage (10.1 GPM) (0.60)
- REFERENCE T.S. 3/4 4-19 & Definitions
*KW 2
- QUESTION 8-3 (1.5)
Administrative Procedure NPAP C-101, Requirements to Remain in the Confines of the Control Room, describes the control room areas to be occupied by the Control Operator (CO) and Senior Control Operator (SRO). Assume Unit 1 at 70% power and Unit 2 shutdown (mode 5). Use the attached Figure 8-1 to draw in the boundaries as requested below. (a) What are the Unit 1 CD's NORMAL boundaries? (0.5) (b) What are the Unit 1 CD's EMERGENCY boundaries? (0.5) (c) What are the SRO's boundaries? (0.5)
- ANSWER See attached drawings
- REFERENCE NPAP C101 Figures 1 and 2
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o-6 PACIFIC GAS AND ELECTRIC COMPANY DEPARTMENT OF NUCl. EAR PLANT OPERATIONS DIABLO CANYON POWER PLANT UNIT NOS.1 AND 2 g SUPPLEMENT 1 TO NUCLEAR PLANT ADMINISTRATIVE PROCEDURE C-101 TITLE: CONFINES OF C0hTROL ROOM AT DIABLO CANYON - SCOPE This procedure identifies the boundaries of the Control Room at Diablo Canyon. PROCEDURE
- 1. The nonnal boundaries of the control room for the Control Operator (operator at the controls) shall be as shown on Figure 1.
- 2. The emergency boundaries of the control room for the Control Operator shall be as shown on Figure 1. ,
- 3. The boundaries of the control room for the senior licensed operator shall
~ include areas adjacent areas.1 and 2 above, the Shift Foreman's office and other appropriate This area is shown on Figure 2.
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l t DIABLO CANYON POWER PLANT UNIT NOS.1 AN,9 2 SUPPLEMENT 1 TO NUCLEAR PLANT ADMINISTRATIVE PROCEDURE C-101 TITLE: CONFINES OF CONTROL ROOM AT DIABLO CANYON 9 FIGURE 2 !
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- QUESTION 0-4 (3.0) 1 While heating up in preparation for startup (mode 4), Electrical Maintenance has just found that diesel-generator 1-3 requires electrical maintenance. The Electrical Foreman has just submitted a Request for Clearance for your immediate approval. Answer the following based on Administrative Procedures NFAP C-7, Tagging Requirements, C-7S1, Plant Tagging Requirements, and C-150, Instruction for Manual Electrical Switiching.
(a) What is the hierarchy of Tags, from the most powerful to the least? (five required) (0.5) l (b) With the generator power leads to be disconnected, ; what two tags should be hung on the generator i output breaker? (1.0) 1 (c) What tag should be used f or a grouriding switch? (0.5) (d) In an emergency, if diesel generator 1-3 were l urgently required, whose permission is required to remove the highest priority tags? (2 people) (1.0)
- ANSWER l (a) (0.1 each) j 1
Maintenance red tag l l Man on line tag Caution tag ! Information tag Action request tag (b) Maintenance end MOL tags (1.0) (c) Caution tag (0.5) (d) The shift Foreman with the concurrence of the requestor or his supervisor. (1.0)
- REFERENCE NPAP C-7, AP C-7S1, and C-150
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- QUESTION B-5 (2.0)
When disabling the safety injection pumps to comply with Technical Specification requirements in mode 4 an Administrative Tag Out is used. Based on AP C-6S1, Clearance Request / Job Assignment, what are four of the five rules governing Administrative Tag Outs?
- ANSWER (0.5 each for any four of the following)
No work can be done No work order is required No MOL tags can be used 1 All tags will be caution tags I No red (or maintenance) tags may be used !
- REFERENCE AP C-6S1
- KW 5
- QUESTION 0-6 (3.0)
Procedure AP C-SS4, Control of Equipment Required by The Plant Technical Specifications, implements methods for control and ; tracking of equipment required to be operable by the Technical i Specifications. l (a) When is the Unit Shift Foreman required to review l all outstanding Technical Specification Equipment i Operability Status Sheets? (four out of five required) (2.0) (b) Who is responsible for the Technical Specification Equipment Operability Status for equipment common to both units? (0.5) (c) When would both a Train A and Train B Technical Specification Equipment Operability Status Sheet be filled out for one piece of equipment? (0.5)
- ANSWER ,
(a) (O.5 for any four of the following) At the beginning of each shift Prior to removing TS equipment from service required by current mode Immediatly after declaining any TS required equipment inoperable Prior to any planned unit mode changes As soon as possible after a unit Rx trip or forced outage l (b) The Unit i Shift Foreman (0.5) (c) When the equipment is actuated by both trains. (0.5)
- REFERENCE AP C-6S4
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- QUESTION j 8-7 (2.5) 1 While operating at 100% power, safety injection pump 2-1 has been removed from service on a clearance due to a large pump seal leak. An independent verification in accordance with NPAP C-lO4, Independent Verification of Operating Activities, and AP C-104S1, Independent Verification of Operating Activities, is in progress.
(a) How much time is allowed to complete the independent verification after removing the SI pump from service? (0.5) (b) What situation (other that a plant emergency) ! would allow the Shift Foreman to waive part of l the clearance verification? (1.0) ') (c) In what situation would it be appropriate for the independent verifier to accompany the individual actually removing equipment from , service? (1.0) l
- ANSWER i l
(a) 4 hours or as soon as possible (0.5) l I (b) Independent verification may be waived - j (1.0 for either of the following) if it involves entry into a high radiation area that would not otherwise be made or if a component has been cleared on a previous clearance which was independently verified. (c) The clearance person should accompany the verifier when incorrect operation could result in a reactor trip of ESF actuation. (1.0)
- REFERENCE NPAP C-104 and AP C-104S1
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- QUESTION .
8-B (1.0) Following repairs to a SI pump it is being returned to service and it is necessary install valve seals in accordance with AP C- 1 951, Suppliment i to Nuclear Plant Administrative Procedure I Sealed Valve. l (a) Prior to installation of a valve seal, how I does the operator verify a CLOSED valve? (0. 5) ' (b) Prior to installation of a valve seal, how does_the operator verify an OPEN valve? , (0.5)
- ANSWER (a) Attempt to move the handwheel in the close direction. If the valve is in the correct position no motion will occur. (0.5) l s'
I (b) Attempt to move the handwheel in the close l direction only enough to verify valve movement. , Return the valve to full open. (0.5)
- REFERENCE AP C-9S1
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- QUESTION B-9 (2.0)
Due to suspected fuel problems, both reactors have been shutdown for inspection of fuel el ecients . Unit 1 is in mode 5 and unit 2 is in mode 6 with fuel being removed from the core for underwater examination. Based on the Minimum Shift Crew Composition required by Technical Specification 6.2.2, what is the minimum shif t crew composition? (Include licensed personnel, non-licensed operation personnel, plant staff required to be on site, and their locations if restricted. Do not include Fire Brigade)
- ANSWER
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(0.25 for each of tbh fol'1owing) SRO Shift Foreman SRO Directly supervising fuel movements in Unit 2 l RO Control Operator in Un'at 1 control room RO Control Operator in Unit 2 control room NL Auxiliary Operator for Unit i NL Auxiliary Operator for Unit 2 NL Auxiliary Operator NL Health Physics Technician l
- REFERENCE TS 6.2.2
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- QUESTION 8-10 (1.0)
During abnormal or emergency conditions the primary function o# the Shift Foreman is command and control. NPAP A-102, General Authorities and Responsibilities of the Shift Foremcn, describes situations when other personnel can assumne the control room command and control function. To answer the following questions assuma that it is 2:00 AM and only the minimum shift crew is on site. (a) If the Shif t Foreman is incapacitated and there is no other concurrent emergency, who may assume the control room command and control function? (0.5) (b) If the Shift Foreman is incapacitated and there is another emergency situation, who should assume the control room command and control function? (0.5)
- ANSWER (a) A Senior Control Operator (0.5)
(b) The Shift Technical Advisor / Shift Engineer (0.5)
- REFERENCE NP4P A-102
*KW
- QUESTION 8-11 (2.0)
Using the Emergency Plan clasifications used in EP G-1, Accident Classification and Emergency Plan Activation, (pages 1 throu0h 10 are attached) what are the correct classifications f or each of the following events? (0.5 each) (a) Tsunami causes a 5 minute shutdown of both ASW pumps. (b) A LOCA with RCS activity over 300 uCi/cc and leaking containment hatch. (c) Loss of the annunciator system during steady state operation. (d) An ATWS uvent occurs without apparent core damage. l
- ANSWER l (a) Alert
, (b) General Emergency N g (c) Al ert 3* A. (d) Alert g
- REFERENCE EP G-1 pages 25 and 33, FR-S1, Appendix Z
*KW 10
1 1 l I l l I l 3.0 3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within 1 bour action shall be initiated to place the unit in a MODE in which the specification does not apply by placing it, as applicable, in:
- a. At least HOT STANDBY within the next 6 hours,
- b. At least HOT SHUTDOWN within the following 6 hours, and
- c. At least COLD SHUTDOWN within the subsequent 24 hours.
Where corrective measures are completed that permit operation under the ACTION requirements, the action may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions of these requirements are stated in the individual specifications. This specification is not applicable in MODE 5 or 6. M cr u R;= 8-2 ,
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NUMBER EP G-1 7 S'2G Pacific Gas and Electric Company REV!SION 6 DATE 1/6/86
, DEPARTMENT OF NUCLEAR PLANT OPERATCNS3 ggg g PAGE 1 0F 60
[e [ DIABLO CAggPg ggt NO(S) ACCIDENT CLASS]FICATION AND EMERGENCY 7i7tg PLAN ACTIVATION NPORTANT TO APPROVEo SAFETY VLANI MANAULK / UAIL O SCOPE This procedure describes the guidelines for Accident Classification , and responsibilities for Activation of the Emergency Plan.
- Implementation of this procedure constitutes declaration of an l
emergency condition. This procedure and revisions thereto require PSRC review. GENERAL This pro:edure provides gv, dance on activating the emergency plan and classifying an accident. The steps required by this procedure are in addition to the steps required to maintain or restore the plant to a safe condition. Prompt notification of offsite authorities should be given within about 15 minutes for the Unusual Event class and sooner (consistent with the need for other emergency actions) for other classes. The time is measured from the time which the Shift Foreman recognizes that events have occurred which make declaration of an emergency class appropriate. j ( This procedure is organized as follows: AClIVATION OF EMERGENCY PLAN The initial steps to be taken for each of the established accident l classifications are listed below under:
- 1. Notification of an Unusual Event
- 2. Alert
- 3. Site Area Emergency
- 4. General Dnergency Figure 1 may be used for guidance in assignment of shif t personnel to activate the Emergency Organization and implement the Emergency Plan, t
20C0161 1XIII
DiAsLO CANYON POWER PLANT UNIT NO(S) } AND 2 5 ON 6 DATE 1/6/86 PAGE 2 0F 60 yi7 m ACCIDENT CLASSIFICATION AND , EMERGENCY PLAN ACTIVATION ACCIDENT CLASSIFICATION Table 1 provides guidance and criterla for determining if an event meets the emergency action levels requiring declaration of one of the four accident classifications. The left column lists events, as provided in the reference NRC criteria for emergency classification, the right coluan provides conditions which are sufficient for declaration of that emergency condition at Diablo Canyon. Other indications could also appropriately indicate certain of the emergency conditions, depending on the situation. Nomally the classification guidance contained in Appendix Z of the appropriate E, ECA, FR, R, or 1 5 M series emergency procedures will be used to detennine the initial classification. In the event none of the E. ECA, FR, R or M series l procedures is appropriate to the immediate situation, Table 1 and judgement should be used for the initial classification (and future events as necessary). POTE: If multiple emergency situations are occurring simultaneously such that the probability of a release of radioactive materials is increased over what it would be for the single occurrence, classify the emergency one level higher than it would otherwise have been, based on th? most severe single occurrence. ! Table 2 sumarizes the emergency classification guidance in the E, ECA, FR, R, and M series procedures. , In addition, procedures included in the ewrgency procedures which meet the NRC requirenerts for innediate notification (10 CFR 50.72) but do not meet the criteria for implementing the emergency plan are included in Table 3 for reference. Refer to Administrative Procedure C-11 "Non-Routine Notification and Reporting to the Nuclear Regulatory j Comission (NRC) and Other Government Agencies" and Supplements to AP C-11 for appropriate reporting for these events. _ PROCEDURE
- 1. NOTIFICATION OF (P.JSUAL EVENT.
- a. Description Unusual events, generally characteri:e off-normal plant conditions that are in process or have occurred which indicate > potential degradation of the level of safety of the plant i# proper action is not taken or if circumstances beyond the usntrol of the operating staff render the situation more serious from a safety standpoint. No releases of radioactive material requiring offsite response i
DC0161 2XIII
O'S DiAsto CANYON POWER PLANT UNIT NO(S) 1 Ab'D 7 , DATE 1/6/86 PAGE 3 CF 60 ACCIDENT CLASSIFICATION AND TIT a EMERGENCY PLAN ACTIVATION or monitoring are expected for this classification unless further degradation of the level of safety of the plant occurs,
- b. Actions
- 1) Assign on-shift personnel to perfonn the functions required for implementation of the Emergency Plar..
Assignments may vary at the discretion of the interim Site Emergency Coordinator, however, e typical organ;zation and assignments are given in Figure 1. Duties and responsibilities are listed in EP G-2, "Establishment of the Onsite Emergency Organization." a) If organizational requirements are given N , the appropriate E, ECA, FR, R, or M l l Procedure, they should be followed. ) b' The minimum functions which must be assigned are: (1) Operational control of the plant by on-shift personnel (Emergency Operations Coordinator). (2) Notification of offsite organizations and off-shift staff (Energency Liaison Coordinator).
- 2) Notify off-shift plant staff of the emergency situation per EP G-2 "Establishment of the Onsite Emergency Organization."
- 3) Promptly notify and inform the county, state, NRC and on-call Recovery Manager of the nature of the Unusual Event situation per EP G-3, "Notification of Offsite Organizations."
- 4) Escalate to a more severe class, if appropriate.
OR
- 5) Close out with a verbal sumary of corrective actions or termination of th? event to offsite authorities.
(The nature of the event will detennine when this should be done), i DC0161 !XIII
DIABLO CANYON POWEQ PLANT UNIT NO(S) 1 AND 2 Sf0N6 DATE 1/6/86 ' I PAGE 4 0F 60 mu ACCIDENT CLASSIFICATION AND EMERGENCY Pl.AN ACTIVATION 1 Retain all notification records and other documentation l 6) of the event for use in preparation of a written l sumary of the event within 24 hours of closeout (or on l the next normal working day). 2- ME I
- a. Description Events are in progress, or have occurred, which involve an actual or potential substantial degradation of the level of I
= safety of the plant. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. It is the lowest level of l classificatiun where emergency near-site or offsite response j may be anticipated. For most Alert events, the plant would ! be brought to a safe condition, and radioactive releases, if i any, would be minimal.
- b. Actions l
- 1) Assign on-shift personnel to perform the functions required for implementation of the Emergency Plan. ;
Assignments may vary at the discretion of the interim ' Site Emergency Coordinator, however, a typical organization and assignments are given in Figure 1. Duties and responsibiiities are listed in EP G-?,
"Establishment of the Onsite Emergency Organization." l l
a) If organizational requirements are given in the ! appropriate E, ECA, FR, M, er R procedure, they l should be followed. b) The minimum functions which must be assigned are: (1) Operational control of the plant (Emergency Opert,tions Coordinator). (?) Notification of offsite orgar.izations and off-shift staff (Emergency Lfaison Coordinator).
)
i l DC0161 4XIII -
DIABLO CANYON POWER PLANT UNil NO(S) 1 AND 2 Sf0N DATE I/6/86 PAGE 5 0F 60 i ACCIDENT CLASSIFICATION AND W EMERGENCY PLAN ACTIVATION (3) Evaluation of plant conditions and radiological assessment (Emergency l Evaluations and Recovery Coordinator). l
- 2) Sound the site emergency signal (if appropriate) and i initiate an all-call on group call 400 and 411, using 4 the Health Physics local radio, to inform plant I personnel of the emergency and to initiate site assembly and accountability per EP G-4 "Personnel Assembly and Accountability". The site emergwcy signal should be followed-up with an announcmant over the plant wide paging system, g: The nature of an emergency may make site assembly unneccessary, impractical or hazardous to personnel, for example a breech in security at the Security Building. In such cases the Shift Foreman may decide not to sound the site emergency signal.
- 3) hotify off-shift plant staff of the emergency situati.on
,( and their assignments in the long-term emergency organization per EP G-2, "Establishrent of the Onsite Emergency Organization."
- 4) Promptly notify and inform the county, state, NRC, and on-call Recuery Manager of the Alert status and their l anticipated response per EP G-3, "Notification of Offsite Organizations." l l
- 5) Initiate onsite monitoring and associated comunications per EP RB-7, "Emergency Onsite Radiological Monitoring Program," if a release in excess cf 100X Technical Specificatiun Limits is occurring or anticipated. (See EP R-2, "Release of ;
Airborne Radioactive Material").
- 6) Determine the need for evacuating nonessential site personnel per EP G-5, "Evacuation of Nonessential Site Personnel". (See EP R-2).
- 7) Provide periodic (approximately every 30 minutes) plant status updates per EP G-3.
i DC0161 5XIII ,
l l DIABLO CANYON POWER PLANT UNIT NO(S) 1 AND 7 ON DATE I/6/86 PAGE 6 Or 60 7gg ACCIDENT CLASSIFICATION AND EMERGENCY PLAN ACTIVATION
- 8) Escalate to a more severe class, if appropriate.
OR
- 9) Closecut or reconnend reduction in emergency class by 4 verbal connunication to offsite authorities. I
- 10) Retain all notification records and other documentation of the event for use in preparation of a written sunnary of the event within 24 hours of closecut (or the next nonnal working day).
- 3. SITE AREA EM5RGENCY
- a. Wicription
, Events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public. The Site Area Emergency classificatien reflects conditions where there is a clear potential for significant releases of radioactive material, such releases are likely, or they are occurring, but in all i cases a core meltdown situation is not indicated based on - current information. Any releases are not expected to exceed EPA Protective Action Guideline Exposure levels except near the site boundary.
- b. Actions
- 1) Assign on-shift personnel to perform the functions required for implementation of the Emergency Plan.
Assignments may vary at the discretion of the interim Site Emergency Coordinator; however, a typical organization and assignments are give in Figure 1. Duties and responsibilities are listed in EP G-2, "Establishment of the Onsite Emergency Organization." a) If organizational requirements are given in the appropriate E. ECA, FR, N. or R procedure, they l should be followed. bl The minimum functions which must be assigned are: (1) Operations 1 control of the plant (Emergency Operations Coordinator). DC0161 6XIII
DLASLO CANYON POWER PLANT UNIT NO(Sj 1AND$ ON DATE 1/6/6C PAGE 7 0F 60 ACCIDENT CLASSIFICATION AND i W EMERGENCY PLAN ACTIVATION (2) Notification of offsite organizations and off-shift staff (Emergency Liaison Coordinator). (3) Evaluation of plant conditions and radiological assessment (Emergency Evaluations and Recovery Coordinator).
- 2) Sound the site emergency signal (if appropriate) and initiate an all-call on group call 400 and 411, using the Health Physics local radio, to inform plant <
U personnel of the emergen:y and to initiate site l assembly and accountability per EP G-4 "Personnel Assembly and Accountability". The site emergency signal should be followed-up with an announcement over the plant wide paging system. g: The nature of an emergency may make site assembly unneccessary, impractical or hazardous l to personnel, for example a breech in security at the Security Building. In such cases the
-(- Shift Foreman may decide not to sound the site emergency signal.
- 3) Notify off-shift plant staff of the emergency situation and their assignments in the long-term emergency organization per EP G-2, "Establishment of the Onsite Emergency Organization."
- 4) Promptly inform the county, state, NRC and on-cell Recovery Manager of the site area emergency situation, per EP G-3 "Notification of Offsite Organizations."
- 5) Determine the need for evacuating non-essential site
>ersonnel per EP G-5 "Evacuation of Non-essential Site
'ersonnel."
- 6) Initiate on-site and off-site monitoring per EP RB-7, "Emergency On-site Radiological Monitoring Program,"
and EP RB-8, "Emergency Off-site Radiological Monitoring Program."
- 7) Provide periodic (approximately every 30 minutes) plant status updates per EP G-3, "Notification of Offsite Authorities."
i DC0161 7XIII
i - otAato CANYON POWER PLANT UNIT NCHS) 1 AND 2 Sf0N6 DATE 1/6/86 PAGE 8 0F 60 yggg.' ACCIDENT CLASSIFICATION AND EMERGENCY PLAN ACTIVATION
- 8) Escalate to General Emergency class, if appropriate.
OR
- 9) Closecut or recommend reduction in emergency class by verbal consnunication to offsite authorities.
- 10) Retain all notification records and other documentation of the event for use in preparation of a written summary of the event within 24 hours of closeout.
4-. GENERAL EMERGENCY
- a. Description The General Emergency action level reflects accident situations involving actual or imminent substantial core cegradation or melting with the potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.
I
- c. Actions
- 1) Assign on-shift personnel to perform the functions required for implementation of the Emergency Plan.
Assignments may vary at the discretion of the interim Site Emergency Coordinator; however, a typical organization and assignments are given in Figure 1. Duties and responsibilities are listed in EP G-2, "Establisheent of the Onsite Energency Organization." a) If organizational requirements are given in the appropriate E ECA, FR, M, or R procedure, they l should be followed. b) The minimum functions which must be assigned are: (1) Operational control of the plant ( Emergency Operations Coordinator). (2) Notification of offsite organizations and off-shift staff (Emergency Liaison Coordinator). DC0161 8XIII
DLABLO CANYON POWER PLANT UNIT NO(S) 1 AND 2 f0N DATE 1/6/86 PAGE 9 0F 60 / T'T'E~ ACCIDENT CLASSIFICATION AND EMERGENCY PLAN ACTIVATION (3) Evaluation of plant conditions and radiological assessment (Emergency Evaluations and Recovery Coordinator). (4) Evacuation of nonessential site p(ersonnel See EP G-5) (SiteEvacuationCoordinator).
- 2) Sound the site emergency signal and initiate an all-call on group call 400 and 411, using the Health Physics local radio, to inform plant personnel of the emergency and to initiate site assembly and accountability per EP G-4 "Personnel Assembly and Accountability". The site emergency signal should be followed-up with an announcement over the plant wide paging system.
- 3) Notify off-shift plant staff of the emergency situation and their assignments in the long-term emergency organization per EP G-2, "Activation and Notification of Onsite Emergency Organization."
( 4) Pr..mptly notify and inform the county authorities of the emergency situation, its classification and their anticipated response per EP G-3, "Notification of Off-Site Organizations," Recomend evacuation out to the LPZ and alerting of the general public in the Basic Emergency Planning Zone to County Authorities. Recommend sheltering ia affected sectors of the Baste Emergency Planning Zone where a release is iminent or actually occurring, and further from the plant, if a dose of >500 mR (whole body) is projected at that distance.
- 5) Promptly inform the state, NRC and on-call Recovery Manager per EP G-3, "Notification of Offsite Organizations."
- 6) Order evacuation of nonessential site 1ersonnel per EP G-5, "Evacuation of Nonessential Site 'ersonnel." on completion of assembly and accountability, i
1 0C0161 9XIII
l OLASLO CANYON POWEQ PLANT UNIT NO(S) 1 AND 2 Sf0N 6 DATE 1/6/86 PAGE 10 0F 60 ACCIDENT CLASSIFICATION AND TITLE EMERGENCY PLAN ACTIVATION
- 7) Dispatch onsite and offsite monitoring teams, per EP RB-7 and EP AB-8.
- 8) Provide periodic (approximately every 30 minutes) plant status updates per EP G-3, "Notification of Offsite Authorities."
- 9) Closeout or recommend reduction of emergency class when appropriate by briefing of offsite authorities. ,
l
- 10) Retain all notification records and other documentation
- of the event for use in preparation of a written sumary of the event within 24 hours of closecut. TABLES )
- 1. "Emergency Action Levels"
- 2. "Emergency Operating Procedures - Accident Classifications"
- 3. "Emergency Operating Procedures - NRL Inwdiate Notification" i
- 4. Technical Specifications Applicable to Unusual Event Condition ,
he. 9. . FIGURES . l
- 1. "Typical On-Shift Emergency Organization and Assignments" l SUPPORTING PROCEDUkES EP G-2 "Establishment of the Onsite Emergency Organizatir.,n" EP G-3 "Notification of Offsite Organizations" EP G-4 "Personnel Assembly and Accountability" EP G-5 "Evacuation of Nonessential Site Personnel" EP R-2 "Release of Airborne Radioact; e Material" i
DC0161 10XIII
l
- QUESTION 8-12 (1.0)
The Code of Federal Regulations, 10CFR55, defines the general provisions for Operators' licenses. Due to the needs of PG&E you j have been working off-shift for 10 weeks of the calendar quarter. What on shift activity is necessary to maintain your license in an active status? (per quarter)
- ANSWER One must perform the functions of a SRO on a minimum of seven B- i hour or five 12-hour shifts per calendar quarter.
- REFERENCE 10CFR55.53
*KW ,
l l
*GUESTION '
B-13 (1.0) Refer to Figure 8-2, which is a copy of the Technical Specifications containing Limiting Condition for Operation 3.0.3. l (a) What are the bases for LCO 3.0.37 (1.0) (b) For which of the following does 3.0.3 apply? (0. 0 ) Charging pump 2-1 and RHR pump 2-1 inoperable Charging pump 2-1 and 2-2 inoperable RHR pump 2-1 and RHR pump 2-2 inoperable Accumulator 2-2 and 2-4 inoperable
- ANSWER l (a) The specification delineates the measures to
) be taken f or those circumstances not f ound in 1 the action statements and whose occurance would i violate the intent of the specification. (1.0) l 4 (b) DELETED - DIABLO CANYON DIVISIONS, CHANNELS, AND EQUIPMENT NUMBERING ARE NOT THE SAME SEE DWG 445651
- REFERENCE TS 3.0.3 and Bases
*KW END OF CATEGORY 8 )
END OF EXAMINATION i l 11 !
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