ML20206R448

Exam Rept 50-275/OL-86-01 on 860715-23 for Units 1 & 2.Exam Results:One Reactor Operator Candidate Passed Written Exam & Four Senior Reactor Operator Candidates & Nine Operator Candidates Passed Oral & Written Exams

ML20206R448
Person / Time
Site:	Diablo Canyon
Issue date:	09/02/1986
From:	Defferding L, Hannon J, Morrill P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20206R431	List: ML20206R427 ML20206R448
References
50-275-OL-86-01, 50-275-OL-86-1, NUDOCS 8609050430
Download: ML20206R448 (126)

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U. S. NUCLEAR REGULATORY COMMISSION REGION V Report No. 50-275/0L-86-01 Docket Nos. 50-275 and 50-323 Licensee: Pacific Gas and Electric Company 77 Beale Street San Francisco, California 94106 Facility Name: Diablo Canyon Units 1 and 2 Examinations at: Avila Beach, California Examination cond ted: July 15-23, 19 6 Examiners: '

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9-?-8' Date Signed L. Def rding V ff/&n~d)UVw 9-a-es Date Signed M )"b T.Ja/ gar . (/ Date Signed Approved by: -

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J.Eyn, Chief,Op6rathnsSection- Date Signed Summary:

Examinations were cond'cted u July 15 through 23, 1986. The written examination was administered on July 15, 1986 to four senior reactor operator candidates (SRO) and ten reactor operator candidates (RO). The oral and simulator examinations were administered to the candidates during the period July 17-23, 1986. All candidates passed the oral and written examinations with the exception of one RO candidate who failed the oral portion of the examination.

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REPORT DETAILS

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1. . Examiners:'

P. Morrill, Chief Examiner, Region V' J. Hannon L. Defferding,..PNL lz ,

F. Jagger, INEL ,

Obse rvers ' '{, r i , ' J. Zwetsig, RV, .

K. Parkinson, Sonalyst,>Inc.

. B. Haagensen, Sonalyst Inc. *

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2. Persons Attending the Fxit Meeting: ,

July 24',1986 I i

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i NRC ,

.s P. Morrill, Region V ,

Pacific Gas and Electric Company' ,

R.'Thornberry,' Plant. Manager J. Sexton, Plant Superintendent J. Townsend iAssistant Plant Manager R. Exner, Training Supervisor B. Terrell, Senior Training Instructor J. Welsch, Senior Training Instructor C. Leach, Senior Training Instructor J..Molden, Operations. Training Supervisor R. Jett, Simulator Supervisor

3. Facility Review of the Written Examination After the conclusion of the written examinations, the examiners met with I J. Tinlin, J. Leach, J. Welsch, and J. Molden of the Training Department to review the written examinations and answer keys. Their comments were incorporated in to the master examination keys prior to grading the candidates' responses.

4.- . Exit Meetings Operating Examinations t

Simulator and operating examinations were conducted on July 16, 17, 22 and 23, 1986. No generic weaknesses were observed. ,

'The facility comments and resolutions are described in Attachment A to ,

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this report.

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< DIABLO CANYON FACILITY COMMENTS, RESOLUTIONS, AND EXAM KEY REVISIONS Question 1.03(a) ,

, Key Revision NPSH is absolute pressure above saturation: therefore, Answer is 5 ft - 1 ft

='4 ft'for full credit.

Question 1.12' Facility' Comment Request that you accept answers (a) 2800 pcm or (b) 3000 pcm because BOL is

-2852 pcm and EOL is -3069 pcm. (Vol. 9 XENON Worth Chart attached as reference,' Attachment 1).

Resolution DCNP teaches 2850 PCM. Will accept (a) or (b) as correct.

Reference:

Operator Information Manual, Page R-7-1.

Question 2.06 Facility Comment Part a l

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Answer key for valve 8701 has an auto close at 700 psig. (Like 8702) This is missing-from the key. (Attachment 2) j Resolution l Comment accepted.

Question 2.06 b.

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Facility Comment Should be deleted because it does not apply anymore. Lesson plan has been revised. (See Attachment 3 OP B-2:V page 2).

Resolution ,

Delete - New instructions do not: call for breakers open.

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2 Question 2.09 c.

~t Key Revision <

Either answer ~is correct.. If high pressure side of the double dam seal fails,-

standpipe level'could increase and if the low pressure side fails,;the.,1 vel9 ,

.will decrease. ,

1 Question 2.13 --

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-' Key Revision '. ,

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Add "or prevent brittle fracture". , ,

Question 3.07(b)

Key Revision-Question did not solicit the first part of-the answer. Full credit for the reactor trip.

J Question 3.08 Key Revision -

(a)(1) Add " containment gas".

Reference:

01MB-6-9.

(a)(3) Answer was left_off the key. Add " containment.Hi Hi press (22 psig)".

(b)(1) Answer in key not complete. Add "CCW supply to RCP oil coolers and thermal' barriers (FCU. 749) ."

Question 3.10 Key Revision

'(a)(2) Question _jgesnotrequestthefirstpartoftheanswer(1outof2 irs > 10 amp). Full credit for last half of answer.

(b), Add "or manually energize".

Question 3.11 Key Revision (3) Answer "SI (with no under-voltage on 4 KV buses)".

3 Question 4.08'.

Key Revision (b)~ Diablo Canyon uses magenta instead of orange.

• i Question 4.10 Key Revision (b)(2) 'Since there'has to be an SRO.in the control room, an added licensed operator either SRO or R0 is OK. -

Question 5.01 _

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E Facility Comment: , ,

Answer of 315*F for part(b) .si inaccurate and reflects misinterpretation'of '

steam tables' Mollier diagram. 305*F is a more accurate answer.'

Resolution ' '

Comment incorrect. Key will be graded to 315* 10*F.

Question 5.02 Facility Comment Question does not state whether core life is BOL or EOL, although the Key reflects EOL.

Note that B for shouldbe7Ifkpcm.ourcoreis.0070,E0L,thereforethecandidates' answer If the candidate assumed BOL, the' answer should be 52 pcm (part (c)).(B,fg BOL

'is .0050).

Resolution s

4 Commentfaccepted. Answer will'be based on candidates assumptions regarding B,gg per operator information manual.

Question 5.05 1 Facility Comment Candidates' answer will likely state "...due to buildup of fission product poisons." Which'would include samarium, xenon, and other less significant

. fission product poisons.

Resolution Comment accepted to be consistent with listed reference.

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4 Question 5.08

' Facility Comment The Key answer (b) is a typo. It.should read "2 (linear increase)" vice "3 (linear increase)". Also note thatcthis is a " linear" increase as indicated.

on semi-log chartspaper. .

Resolution:

Comment' accepted. -Typo corrected.

Question 6.02:

Facility Comments For both parts a &,b, chare are other acceptable answers, namely:

1. SPDS ,

2. P-250 Plant Computer "

3. Dedicated Shutdown Panel Resolution Comment accepted. Components added to key per PG&E Dwg. 102035 sheet 6D.

Question 6.03 Facility Comment

- Choice (b) is the correct answer. The reasons:

1. , Choice (a) is not correct. Since the IR logic to reinstate SR instruments is 2/2, the students will assume (correctly) that if one channelisreadinglowghereisnotproblemsincetheOPERABLEchannel g

must also be below 10 amps in order for.the SR NIS to reinstate.

2. "Startup level" is not standard terminology for DCPP. Essentially everyone presented with this term interprets it to mean "Startup rate" vice "IR level", thus they choose choice (b) as their answer (which is an accurate description of SUR response).

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. Resolution:

. Comment accepted. Typo corrected, i

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Facility Comment:-

Candidates will likely include the " Steam Flow Chart Recorders" on' Panel CC3 for Part.(C).

Also, there is some potential for problems here since the question asks'for

" instruments or-control systems" and does not specifically solicit protection features.

Resolution:

Steam flow chart recorder added as a correct answer.

Question 6.07 Facility Comment Many candidates wil1~ describe the final S/G level-is being slightly higher

'than programmed level,-sincesif level returned exactly to_the " original i setpoint", as stated, then there would be no level error, so level would start to rise again.

Resolution:

Comment incorrect. The level error will eventually dominate a feed-steam flow mismatch.

Question 6.09 Facility Comment The stated answer is the basis for OTdeltaT protection. _For OPdeltaT-protection, the basis is,"to limit rod heat flux (KW/ft) to ensure fuel integrity under al1 possible overpower conditions...."

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Resolution:

. Comment accepted. ,

Reference:

. Technical Specification page B2-4.

Question 6.11 Facility Comments The question specifically states " numerical values and coincidence...not

-desired." In the answer Key, then, bus UV and UF coincidence seem to be required. Inconsistent. -

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the bus UV trip.

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, Answer key modified to include-2/2 logic.

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' Question'7.03 -

Fac'ility Comment- g

! Part (a) solicits-indications of " uncontrolled or.itnexplained reactivity '-

-increases" which will require' emergency boration. The.. answer Key tests the ,

sub-symptoms of symptom #4'as the answer, however, symptom #3 is also an uncontrolled reactivity increase.

Part'(c) Answer Key has incorrect valve; number. Emergency boration valve j _ number is 8104,~not 8471.

The manual valve is 8471, as. stated.

Resolution J

Added " uncontrolled RCS cooldown following a reactor trip with no ESF

, actuator" to part (a) per OP AP-6, page 1. Corrected valve number in part (C).

Question 7.04(b) l' Facility Comment i

OP'A-2:II is a complex and infrequent procedure required to be "in-hand".

Candidates are not required to memorize NOTE statements from such procedures.

Resolution l Candidates should be aware of general procedure for removing the RPV head.

l Comment not accepted.

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Question 7.05 Facility Comment I' The answer key does not assume 2 things (1. and 2. below) which are vital in proper compliance with this'LCO.

1. At time 0900, the Rx is not only outside of the target band, but is outside of the Acceptable Operation Limits _ curve. Therefore, thermal power must be less than-50% within 30 minutes (0930 hours0.0108 days <br />0.258 hours <br />0.00154 weeks <br />3.53865e-4 months <br />). We were not at 50% till 1035.

i h 2. 50% is not acceptable. The ACTION ~ statement clearly states that to j

comply, you must be below 50% (and not equal to). .

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3. .Even if you assume that 50% is acceptable, (i.e. 50% means _50%) then the answer to part (d) is 09371 based on the time that the LCO requirements and the logged data: I hour of cumulative penalty-time in the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4. Also, reduced trip setpoint was done late.

-Resolution:

1. ' Violation added to key.

2.. The question and answer are adequate. No change needed.

3. Comment accepted.

4. Comment accepted.-

Question 7.07 ,

s Facility Comment General comment.' It11s inapprophiate to' expect SRO's to be able to memorize various accident classifications given that the classification procedure is approximately-65 pages long. A more appropriate question is to give them the procedure and event in-timed sequence and have them classify it.

(5) Could be "(A)"Jor "(B)";if 'five' is in an area that has vital equipment.

Resolution -

"(B)" will be accepted,as correct if candidate states appropriate assumption.

Question 7.08 Facility Comment

'Part (b) Although the question says "following a reactor trip with an inadvertent SI", some candidates may list the conservative termination criteria which include adverse containment criteria of (20%] as shown in the relevant procedures.

Resolution: l Commen't accepted. Key changed to read " greater than 4[20]%" vice " greater than 4% per EP E-1.

Question 8.03 Facility Comment The Key answer is correct in that the startup may not proceed due to the requirements of Technical Specification 3.0.4. The loss of a diesel fuel oil transfer pump no longer makes a diesel generator inoperable, as indicated in the answer Key. New ACTION f. covers loss of a diesel fuel oil train.

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Resolution: s ,

c, ' r .2 E Word " Gene $t'or" citarged to " fuel oi1> transfer tr'ain".

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Question'8.04- ,

Facility Comment- -

'The candidate-may be inclined to answer "yes" to part 5 if he interprets the high wind as a "..'. natural. phenomenon that posesan actual threat to the i

safety of the nuclear plant or significantly hampers site personnel in.the performance of duties necessary for safe operation of the plant.", per -

10 CFR 50.72(b)(iv).

Resolution:-

A "yes" answer for part 5 will be~ accepted if the candidate qualifies his answer as described.

Question 8.05 Facility Comments Specification definitely does not apply in part (b),-therefore, the answer should be "Does not-apply".

1. 'The LCO_page provided in the handout has applicability in Modes 1 through

4. The problem posed isifor Mode 5.

i' 2. Our Tech Specs have no LCO for ASW in Mode 5.

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l Comment accepted.

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gyi cyg e p <A r L hu r U. S. NUCLEAR REGULATORY COMMISSION Y B S",6,T REACTOR OPERATOR LICENSE EXAMINATION A h Sl0 D b Y FACILITY: DIABLO CANYON 1 REACTOR TYPE: PWR-WEC4 DATE ADMINISTERED: 86/07/15 EXAMINER: DEFFERDING, L APPLICANT: ANSWER KEY INSTRUCTIONS TO APPLICANT:

Use separate paper for the answers. Write answers on one side only.

Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.

% OF .

CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 25.00 25. 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 4

25.00 25.9d 3. INSTRUMENTS AND CONTROLS 25.00 25.h 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL-sf e

'I b0 100.00

. TOTALS FINAL GRADE  %

All work done on this examination is my own. I have neither given nor received aid.

APPLICANT'S SIGNATURE

_1_ ____  : _

NRC RULES AND GUIDELINES FOR LICENSE EvAMINATIONS During the administration of this examination the following rules apply:

1. ,, Cheating on the examination raeans an automatic dental of your application and could result in more severe penalties.

2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.

3. Use black ink or dark pencil only to facilitate legible reproductions.

4. Print your name in the blank provided on the cover sheet of the

<_ examination.

5. Fill .in the date on the cover sheet of the examination (if necessary).

6. Use only the paper provided for answers.

7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.

8. " as Consecutively appropriate, start each category on a new page, write only one si of number each answer sheet, write "End of Category _de

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

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down on your desk or table.

12. Use abbreviations only if they are commonly used in facility literature.

13. The point value for each question is indicated in parenthesis after the question and can be used as a guide for the depth of answer required.

14. Show all calculations, methods, or assumptions used to obtain'an answer to mathematical problems whether indicated in the question or not.

• 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND 00 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.

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18. When you complete your examination, you shall:

a. Assemble your examination as follows:

(1) Exam questions on top.

J (2) Exam aids - figures, tables, etc.

(3) Answer pages including figures which are a part of the answer. ~

b. Turn in your copy of the examination and all pages used to answer the examination questions,

c. Turn in all scrap paper and the balance of paper that you did not use for answering the questions.

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d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 2 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.01 (1.00)

EXPLAIN WHY departure from nucleate boiling is allowed in a S/G but not in the core. (1.0) .

QUESTION 1.02 (1.50)

Letdown water (75 gpm) at 538 deg F enters the regenerative heat exchanger and exits at 290 deg F. ASSUMING normal, steady-state, 4

at power operation (Total Charging Flow of 87 gpm at 115 deg F),

WHAT is the temperature of the charging water entering the RCS from the regenerative heat exchanger? SHOW all work and STATE all assumptions. (1.5)

QUESTION 1.03 (2.50) l l A large vr.nted water tank 60-ft high has a small capacity centrifugal pump taking a suction from its base. The pump is located at a vertical elevation corresponding to 1 ft below the bottom of the tank and it requires 5 ft of net positive suction head (NPSH) to prevent cavitation. The tank is initially entirely full of water and is maintained at 60 deg F by heaters. ASSUME that the vent becomes totally clogged (fully closed) while the pump is in operation.

ANSWER the following questions.

a. To WHAT level would the tank drop to as the pump continues to remove water from the tank? (1.0)

b. WHAT two (2) types of equipment failures could result from continued pump operation with a clogged vent? (1.0)

c. To WHAT level could the pump continue to pump water to if the vent was then opened? ,

(0.5) l 1

QUESTION 1.04 (1.50)

The steam generators are producing 14.5 million Ibm /hr of steam i

flow at 520 deg F and 100% quality with a feedwater temperature of '

I 440 deg F. Use steam table values to CALCULATE the total RCS flow

• rate if T(H) = 608 deg F and T(C) = 544 deg F. SHOW all work. (1.5)

(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)

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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3 1.

THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW s

QUESTION 1.05 (1.00)

a. WHAT change occurs in beta effective from BOL to E0L? (GIVE values.) (0.5) ,

b. WHAT causes this change? (0.5)

B QUESTION 1.06 (2.00)

The count rate increased from 12 cps to 36 cps during withdrawal of the shutdown banks. If the initial value of Keff, KO, was 0.94, CALCULATE the reactivity worth, in pcm, of the shutdown banks, rho S/D? SHOW all work. (2.0)

QUESTION 1.07 (1.50)

Power increases from 10**-8 amps to 5 x 10**-7 amps in two minutes.

DETERMINE the stable SUR. (1.5)

QUESTION 1.08 (2.00)

ANSWER the following statements, relating to Doppler only Power Coefficient, TRUE or FALSE.

a. Clad creep causes Doppler Only Power Coefficient to become less negative as core age increases. (0.5)

b. Fuel growth causes Doppler Only Power Coefficient to become less negative is core age incre,ases. (0.5)

c. The Doppler Only Power Coefficient is more negative at high power levels than at low power levels. (0.5) l

c. The Doppler Only Power Coefficient is less negative at high boron concentrations than at low boron concentrations. (0.5) 1

(***** CATEGORY 01 CONTINUED ON NEXT PA3E *****)

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PRINCIPLES OF NUCLEAR POWER PLANT OPERATIOil, PAGE 4 1.

o THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW B

QUESTION 1.09 (1.00)

WHICH one of the following conditions is NOT necessary to cause brittle fracture? (1.0) .

a. pre-existing defects

b. load stress greater than yield stress I

c. temperature below the nil ductility transition temperature

d. residual stresses QUESTION 1.10 (1.50)

NAME three (3) situations which could cause a Quadrant Power Tilt. (1.5)

QUESTION 1.11 (2.00)

The hot channel factor limits will be met for normal operation provided four (4) conditions are observed. NAME these four (4)

conditions. (2.0)

QUESTION 1.12 (1.00)

CHOOSE the CORRECT response. The approximate value for 100% power equilibrium xenon reactivity at Diablo Canyon is: (1.0) l

a. 2800 pcm

b. 3000 pcm

c. 3800 pcm

d. 4600 pcm l

l QUESTION 1.13 (1.50) i A PORV on the pressurizer is leaking. Primary pressure is 2250 psia and the PRT pressure is 20 psig.

a. WHAT is the temperature in the tailpipe? (1.0) -

WHAT is the condition of the fluid in the tailpipe (degrees l b.

superheated, degrees subcooled, or % moisture)? (0.5) l j (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 5

. THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

QUESTION 1.14 (3.00)

Would power range indications tend to indicate an INCREASE, DECREASE, or NO CHANGE for the following events? Assume actual power was

• unchanged at 75% (before and after the event). Briefly EXPLAIN your answer. (Considereachseperately.)

a. Heatup of the reactor coolant (1.0)

b. Core aging (1.0)

c. Increase in boron concentration (1.0)

QUESTION 1.15 (1.00)

WHY are burnable poison rods used instead of increasing the boric acid concentration in the coolant? (1.0) l QUESTION 1.16 (1.00) i WHICH will cause the greatest change in the temperature of the affected fluid? Adding one BTU of energy to one (1) lbm of: (1.0)

, . . saturated water at 180 deg F l . superheated steam at 400 deg F

. subcooled water at 180 deg F

. subcooled water at 570 deg F ,

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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 6 QUESTION 2.01 (1.00)

WHAT are the two (2) major components the Vibration and Loose Parts Monitoring system monitors? (1.0)

QUESTION 2.02 (1.00)

Fluid escaping from a leaking PZR PORV valve is cooled in the PRT by: (1.0)

(a.) Water sprays in the PRT that are fed from primary make up water.

(b.) Venting and pressurizing the N(2) cover blanket.

(c.) Mixing with PRT water through perforated pipe.

(d.) "unning it through the RCDT heat exchanger before introducing it into the PRT.

QUESTION 2.03 (2.00) -

Downstream of the non-regenerative heat exchanger in the CVCS letdown line there are two valves providing protective functions. DESCRIBE these functions for:

a. Letdown temperature divert valve, TCV-149 (1.0)

b. Low pressure letdown control valve, PCV-135. (1.0)

QUESTION 2.04 (1.50) .

DESCRIBE HOW the 125 VDC Vital loads' Panel 11 receives its power.

l INCLUDE the three (3) sources from the 480 VAC source (s) through to the panel. (1.5) 1

(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 7 QUESTION 2.05 (1.50)

You are conducting a recovery from a reactor trip during which all reactor coolant pumps tripped. Conditions have been restored to the point where reactor coolant pumps can be started. WHICH pump -

should you start first and WHY? (1.5)

QUESTION 2.06 (3.00)

a. DESCRIBE four (4) interlocks or safety devices which prevent overpressurization of the RHR system when placing it in service or conducting a plant cooldown? (2.0) b WHAT measure is taken to prevent the inadvertent isolation of j)9 the RHR suction from the RCS? (1.0)

Y QUESTION 2.07 (1.00)

The Moveable Incore Detector System (MIDS) is used to monitor the neutron flux level at various positions in the core.

a. HOW MANY detectors are used in this system? (0.5)

b. WHATtypeofdetector(s)isusedbytheMIDS? (0.5)

QUESTION 2.08 (2.00)

For a DBA LOCA, LIST the ECCS components that will inject water

into the reactor coolant system. INCLUDE the operating pressures l of each component. *

(2.0) l

(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 8 t

QUESTION 2.09 (3.50)

a. WHAT is the purpose of the No. I seal bypass valve associated with the Reactor Coolant Pump and WHAT are the RCS minimum and maximum pressure requirements imposed for opening the valve? ~ '

(Specific values required.) (1.0)

b. The RCP No.1 seal water return flow to the CVCS is isolated I

upon containment isolation. WHAT provision is made for maintaining seal leakoff after the isolation valves close? (1.0)

If the leakage through No. 3 seal is greater than normal, will

/. the No. J[ seal standpipe level INCREASE or DECREASE? (0.5)

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d. WHAT establishes the DP across No. 2 seal during normal operation? (1.0)

E QUESTION 2.10 (1.50)

- The feed pump speed control is designed to maintain a programmed pressure differential between the feedwater pump (s) discharge and steam header by controlling the speed of the turbine driver (steam governor speed changer). WHAT are the three (3) reasons for this design? (1.5)

I QUESTION 2.11 (2.00)

a. WHAT method is used to seal the reactor vessel closure head to the reactor vessel and HOW are these seals activated? (1.0)

b. HOW would a failure of the seal be detected? (0.5)

c. WHAT is the Technical Specification category for the type of leakage from the reactor vesseT seal? (0,5) l QUESTION 2.12 (1.00)

WHAT are the two (2) functions of the PZR sprays manual bypass *

! valves? (1.0) j (***** CATEGORY 02CONTINUEDONNEXTPAGE*****)

PAGE 9

. 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS QUESTION 2.13 (1.00)

WHY is it necessary to have only one (1) centrifugal charging pump operable (i.e., two pumps are not allowed to be operable) when the RCS is below 323 degrees F7 (1.0)

QUESTION 2.14 (1.00)

A reactor shut down is in progress at 10**-8 amps. The compensating voltage for one of the intermediate channels fails. WHAT change in the intermediate range indication occurs? g*,; ) u,4,3 4 u.7 (1.0) s QUESTION 2.15 (2.00)

Unit 1 is in Mode 1 at 100% power. Reactor trip initiate alarms are received in the control room on PK-04. However, a Reactor Trip does not occur. By procedure, the operator is directed to attempt a manual trip first, and then to de-energize two 480-Volt load centers.

WHAT two (2) load centers are de-energized? (1.0) a.

b. WHY are these two (2) particular load centers required to be de-energized? (1.0) i i

I I

l l

i

(***** END OF CATEGORY 02 *****)

3. INSTRUMENTS AND CONTROLS PAGE 10 QUESTION 3.01 (2.50)

PROVIDE an RCS temperature value for each item listed.

a. Lo Tavg setpoint (0.5)

b. Lo-Lo Tavg setpoint (0.5)

c. No-load Tavg (0.5)

d. Minimum temperature for criticality (0.5)

e. "Tavg DEVIATION FROM REF" annunciation (0.5)

QUESTION 3.02 (2.00)

For the following list of circuits and list of circuit uses from the Reactor Coolant System temperature instrumentation, MATCH only one (1) circuit to each circuit use. A circuit can be used only once. (2.0)

Circuit Use Circuit

a. C-4 rod stop 1. OP delta T

b. Feedwater Isolation 2. T-cold wide-range

c. RVLIS 3. OT delta T I

d. Steam Dump System demand 4. Tavg control

5. Tavg protection

6. T-hot wide-range i

i r

i

(

i

(***** CATEGORY 03CONTINUEDONNEXTPAGE*****)

3. INSTRUMENTS AND CONTROLS PAGE 11 QUESTION 3.03 (2.00)

For each of the following process radiation monitors, DESCRIBE the automatic function, if any, that occurs when an alarm is received.

a. Containment Air Particulate Monitor RE-11. (0.5)

b. Component cooling Liquid Monitor RE-17A, 17-B. (0.5)

c. Gas Decay Tank Discharge Gas Monitor RE-22. (0.5)

d. Control Room Air Particulate Monitor RE-21. (0.5)

QUESTION 3.04 (3.00)

Unit 1 is operating at 45% power with all systems in automatic control. For each of the following conditions, GIVE the direction of initial rod motion and EXPLAIN WHY there is rod motion.

a. A steam generator Atmospheric Power Relief Valve fails open. (1,0)

b. A feedwater heater string becomes isolated. (1.0)

c. The lower detector of the power range channel N-44 fails low. (1.0)

QUESTION 3.05 (1.50)

ANSWER the following 1 RUE or FALSE regarding the construction and operation of the POWER RANGE NUCLEAR INSTRUMENTATION detector.

l a. Each upper and lower section provides inputs to a delta flux l ueu.r, delta flux recorder, and, current comparator. (0.5)

b. BF3 gas is used in the power range detectors for neutron detection. (0.5)

c. The detector uses no compensation circuitry to remove gamma current. (0.5)

(***** CATEGORY 03CONTINUEDONNEXTPAGE*****)

3. INSTRUMENTS AND CONTROLS PAGE 12 l

QUESTION 3.06 (2.00)

WHAT conditions actuate steamline isolation? (INCLUDEsetpoints andcoincidence.) (2.0)

QUESTION 3.07 (2.00)

a. If Unit 1 is at 100% power and one reactor trip bypass breaker is racked in and shut, WHAT indications does the control room operator have of this condition? (1.0)

b. If the remaining bypass breaker is also racked in and shut, WHAT effect will this have on the plant? (1.0) yi N 1 to 6 7sy r4 <& M S QUESTION 3.08 (3.00)

a. WHAT signals, other than manual, directly activate the following:

1. Containment ventilation isolation -

(1.0)

2. Phase A isolation 0.

3. Phase B isolation 0.

b. If a Phase B isolation occurs after a Phase A isolation has occurred, WHAT additional valves will close? (Name or number.) (1.0)

QUESTION 3.09 (2.00)

On a sketch of the steam generator, SHOW the location of the taps for the wide range and narrow range level instruments relative to the tube sheet. INDICATE height covered by narrow range. (2.0)

(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)

=

3. INSTRUMENTS AND CONTROLS PAGE 13 QUESTION 3.10 (2.00)

a. During a reactor startup with the source range detector energized: -.

1. At WHAT count rate level will the reactor trip? (0.5)

2. WHAT operator actions can prevent this trip? (1.0) 06 c 4

• r-

b. HOW are the source range detectors energized during a reactor shutdown? (0.5)

QUESTION 3.11 (2.00)

LIST the three (3) Auto Start signals for the component cooling water system. INCLUDE in your description which auto start is independent of a minimum CCW pump oil pressure. (2.0)

QUESTION 3.12 (1.00)

LIST the three (3) conditions that will initiate an automatic

'FEEDWATER ISOLATION' signal. (1.0) i l

l l

(***** END OF CATEGORY 03 *****)

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 14 RADIOLOGICAL CONTROL QUESTION 4.01 (3.00)

Following a reactor trip with an inadvertent SI, four (4) conditions must be met to terminate SI. -

a. LIST the four (4) (one item has 2 parts). (2.0)

b. After terminating SI either of two (2) conditions will require SI reinitiation. LIST the two (2) conditions. (1.0)

QUESTION 4.02 (2.50)

ANSWER the following questions related to the Subcooled Margin Monitor,

a. TRUE or FALSE 7 The OPERABILITY of the Subcooled Margin Monitor is required by Technical Specifications. (0.5)

b. LIST all inputs to the Subcooled Margin Monitor. (1.5)

c. TRUE or FALSE? There are protection and/or control functions associated with the Subcooled Margin Monitor. (0.5) i l

QUESTION 4.03 (2.00)

STATE four (4) of the five (5) conditions in which core loading must be suspended, in accordance with the operating procedure OP B-8D. (2.0) l l

(***** CATEGORY 04CONTINUEDONNEXTPAGE*****)

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 15

. RADIOLOGICAL CONTROL I

QUESTION 4.04 (2.50) . l I

a. OP L-2, HOT STANDBY TO MINIMUM LOAD, instructs that with the main unit at approximately 15% power, " Transfer the governor valves from single valve operation to sequential valve operation."

1. HOW does the operator accomplish this? (0.5)

2. WHAT is the difference between single valve operation and sequential valve operation? (1.0)

b. OP L-2 also instructs the operator to press IMP IN at approximately 15% power. WHAT change does this cause in the control system? (1.0)

QUESTION 4.05 (1.00)

One of the programs utilized by the P-250 plant computer is the Post Trip Review program. EXPLAIN HOW this program operates to assist the operator in evaluating plant operations after a trip. (1.0)

QUESTION 4.06 (2.00)

You are the Control Operator on Unit 2 at 20% power when Instrument Inverter IY-24 fails, resulting in a loss of power to Vital Instrument AC Distribution Panel PY-24. LIST two (2) symptoms of this electrical casualty, other than possible alarms. (2.0) l QUESTION 4.07 (1.00)

a. During an onsite emergency, WHO

• becomes the Interim Site Emergency Coordinator? (0.5)

b. As a minimum, the Interim Site Emergency Coordinator shall

appoint two (2) people to support his efforts. LIST the two t (2) positions. (0.5) l l

l

(***** CATEGORY 04CONTINUEDONNEXTPAGE*****)

. - - . - - _ 1___

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 16 RADIOLOGICAL CONTROL QUESTION 4.08 (3.50)

There are six (6) critical safety function status trees which are an emergency transient. The status trees contain monitored during(4) different colors.

paths with four

a. LIST the six (6) status trees in order of priority (one (1) being the highest). (1.5)

b. LIST the color of the paths in order of priority, and EXPLAIN WHAT general operator action is required for each color. (2.0)

QUESTION 4.09 (1.50)

After working for 45 minutes in a radiation zone that you were told was 20 mr/hr, you discover that your pocket dosimeter reads offscale >200 mr. A subsequent survey found a " hot" valve that used 2 r/hr gamma at 1 ft. You were working 2 ft from the valve.

a. HOW much exposure did you receive? (1.0)

b. If your previous exposure for the quarter was 900 mrem did you exceed the 10 CFR 20 quarterly limits. (0.5)

QUESTION 4.10 (2.00)

a. After an unscheduled reactor trip, WHO (position) must give approval to start the reactor? (0.5)

b. LIST the two (2) additional persons that must be in the centrol room during the start up from an unscheduled reactor trip. (1.C)

c. You are the control room operat'or when you detect a reactor e

trip setpoint has been exceeded. WHOSE approval do you need to shut the reactor down? (0.5) l l

l QUESTION 4.11 (1.50) ,

WHAT are the three (3) Technical Specification requirements for the Refueling Water Storage Tank in modes 1-4, for it to be considered OPERABLE? (1.5)

(***** CATEGORY 04CONTINUEDONNEXTPAGE*****)

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 17 RADIOLOGICAL CONTROL E

QUESTION 4.12 (1.50)

LIST three (3) of the five (5) conditions that must be met to ensure Containment Integrity during power operation? (According to Technical .

Specifications.) (1.5)

QUESTION 4.13 (1.00)

ASSUME you are part of a crew that made an entry into containment to identify a leak from the RCS. On return to exit containment you can not get the containment side door closed. To reach the emergency hatch would require excessive exposure to radioactive steam from the RCS leak. HOW can you get out of containment without shutting down the reactor and using the equipment hatch? (Theleakcanbe repaired when you get the proper equipment.) (1.0) i l

1

(*****ENDOFCATEGORY04*****)

(************* END OF EXAMINATION ***************)

. 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 18 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- DIABLO CANYON 1 -86/07/15-DEFFERDING, L ANSWER 1.01 (1.00)

In a S/G, the heat transfer surface temperature cannot exceed the hottest RCS coolant temperature; while in the reactor, the cladding temperature can increase until the fuel rods destroy themselves due to the accumulation of decay heat. [+1.0]

REFERENCE

1. Westinghouse: Thermal-Hydraulic Principles and Applications to the PWR.

ANSWER 1.02 (1.50) h =h [+0.5]

letdown charging E delta T = b delta T 1 1 c c delta T = delta T (E/m) c 1 1 c h = 55 gpa (charging less 32 gpm for RCP seals) [+0.5]

c delta T = 248 deg F (75/55) = 338 deg F c

Outlet temperature is therefore 115 + 338 = 453 deg F [+0.5] ,

REFERENCE

1. Westinghouse: Thermal-Hydraulic Principles and Applications to the PWR.

l

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 19 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- DIABLO CANYON 1 -86/07/15-DEFFERDING, L ANSWER 1.03 (2.50)

a. 3.4 feet (4 ft water - 0.256 psia) [+1.0] (will accept 4 ft up s # , An =e * #

for [+0.4]) . saf

• 4i' #6' 6e v'T

b. 1. Pump damage due to cavitation. [+0.5]

2. Tank collapse due to low internal pressure. [+0.5]

c. O ft (all the water). [+0.5] (Theaddedpressureof14.7 psia (34 ft of water) would allow all of the water to be removed.)

REFERENCE

1. Westinghouse: Thermal-Hydraulic Principles and Applications to the PWR.

I e

i 0

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,, PAGE 20 THERMODYNAMICS., blat TRANSFER AND FLUID FLOW ANSWERS -- DIABLO CANYON 1 -86/07/15-DEFFERDING, L ANSWER 1.04 (1.50)

Q =Q [+0.5]

RCS SEC

$ delta h -E delta h RCS RCS SEC SEC m delta h m (h -h )

SEC SEC SEC stm fw m = =

RCS delta h h -h RCS T(C) T(H) 6 (14.5x10 lbm/hr)(1199.0 BTU /lbm - 419.0 BTU /lbm) m =

RCS 628.8 BTU /lbm - 541.8 BTU /lbm 8

= 1.3 x+10i lbm/hr or 130 million Ibm /hr [+1.0]

REFERENCE

1. Westinghouse: Thermal-Hydraulic Principles and Applications to PWP.

ANSWER 1.05 (1.00)

a. BOL 0.007 [+0.25] E0L 0.005 l b. Decrease in U-235 and increase in PU-239 [+0.25] [+0.5]

REFERENCE

1. Westinghouse: Fundamentals of Nuclear Reactor Physics. ,

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 21 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- DIABLO CANYON 1 -86/07/15-DEFFERDING, L ANSWER 1.06 (2.00)

CR1/CR0 = 1-K0/1-K1 [+0.5]

K1 = 1-(1-K0)* CR0/CR1

= 1 - (1-0.94)* 12/36 = 1 - (0.06)* 1/3 = 0.98 [+0.5]

rho S/D = delta rho = (K1 - KO)/(K1 x KO) [+0.5]

rho S/D = (0.98 - 0.94)/(0.98 x 0.94) = 0.04/0.921 = 0.0434 rho S/D = 4342 pcm [+0.5] (will accept 4300-4400 pcm for full credit)

REFERENCE

1. Westinghouse: Fundamentals of Nuclear Reactor Physics.

ANSWER 1.07 (1.50)

P = P(0) 10**(SUR*t(min)) [+0.5]

SUR = log (P/P(0))/t(min)

= log (5 x 10**-7/10**-8)/2 min

= 0.85 DPM [+1.0]

REFERENCE

1. Westinghouse: Fundamentals of Nuclear Reactor Physics.

ANSWER 1.08 (2.00)

a. TRUE '+0. 5' '

b. TRUE l+0.5:

c. FALSE '+0. 5'

d. FALSE ' 0.5l

.+ ,

REFERENCE

1. Westinghouse: Reactor Core Control For Large PWRs, Chapter 2.

, 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 22 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- DIABLO CANYON 1 -86/07/15-DEFFERDING, L ANSWER 1.09 (1.00)

(b.) [+1.0]

REFERENCE

1. Westinghouse: Thermal-Hydraulic Principles and Applications to PWR II, Chapter 13.

ANSWER 1.10 (1.50)

1. rod misalignment or dropped rod

2. x-y xenon transient

3. improper fuel load

4. improper burnable poison pattern

5. uneven fuel depletion

6. difference in moderator density (uneven loop temps)

Any three (3) [+0.5] each, +1.5 maximum.

REFERENCE l 1. Westinghouse: Reactor Core Control For Large PWRs, Chapter 8.

ANSWER 1.11 (2.00)

1. Control rods in a single group move together with no individual rod insertion differing by more than +- 12 steps, indicated, from the group demand position. [+0.5]

2.

described in Technical Specification (3.1.3.6). Control +0.5] rod groups are sequenced with

3. The control rod ins'ertion limits of Technical Specifications (3.1.3.5 and 3.1.3.6) are maintained. [+0.5]

4. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits. [+0.5]

REFERENCE

1. Diablo Can. yon: Technical Specifications, 3/4.2.2.

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 23 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- DIABLO CANYON 1 -86/07/15-DEFFERDING, L ANSWER 1.12 (1.00)

[+1.0] T h y #

• 4 " ** Ub W Nb REFERENCE

1. Westinghouse: Reactor Core Control For Large PWRs, Chapter 4.

ANSWER 1.13 (1.50)

a. From Mollier chart using constant enthalpy expansion to the 20 psig (35 psia) line, ;+0.5] temperature in pipe = T(sat) for 35 psia = 259 deg F + 0.5]

b. hg 2250 = 1117.7 constant enthalpy process hg 35 psia = 1166.5 hf 35 psia = 227.5 x+y = 1 x227.5x + 1166.5y]=

= 0.052 [+0.5 1117.7 OR from Hollier chart fluid contains 5-6% moisture [+0.5]

REFERENCE

1. Generic: Steam tables and Mollier chart.

ANSWER 1.14 (3.00) ,

a. Increase [+0.5] - Temperature increases, density decreases, leakage increases. [+0.5]

b. Increase [+0.5:1 - As center fuel burns up, flux shifts to core extremities. I:+0.5]

c. Decrease [+0.5] - Boron concentration increases, absorption in downcomer increases, leakage decreases. [+0.5]

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 24

. THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- DIABLO CANYON 1 -86/07/15-DEFFERDING, L REFERENCE

1. Westinghouse: Fundamentals of Nuclear Reactor Physics, Chapter 6.

ANSWER 1.15 (1.00)

To ensure that the moderator temperature coefficient is negative for power operations. [+1.0]

REFERENCE

1. Westinghouse: Reactor Core Control For Large PWRs, Chapter 2.

ANSWER 1.16 (1.00)

(b.) (superheated steam at 400 deg F; the Cp of superheated steam is only about 0.5 BTU /lbm-deg F) [+1.0]

REFERENCE f

1. Westinghouse: Thermal-Hydraulic Principles and Applications to the PWR I.

e

, . , . , . . --..e.-.., . , ,,-e.,. . , - - . , - - , - - - , - - , . - , , , - - , .- , , , , . - , , , . . . , . , - , . . ,

. 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 25 ANSWERS -- DIABLO CANYON 1 -86 / 07 /15-DEFFERDING, L ANSWER 2.01 (1.00)

Steam generators [+0.5] and reactor pressure vessel [+0.5].

REFERENCE

1. Diablo Canyon: Lesson Plan LA-2e.

ANSWER 2.02 (1.00)

(c.) [+1.0]

REFERENCE

~

1. Diablo Canyon: Lesson Plan LA-4b.

ANSWER 2.03 (2.00)

c. TCV-149 protects the demineralizers downstream by diverting flow around the demineralizers if temperature goes too high (136 degrees F). [+1.0]

b. PCV-135 maintains a 350 psig backpressure on the letdown flow through the ortftces to prevent the letdown flow from flashing (and eroding the orifices). [+1.0]

REFERENCE I

1. Diablo Canyon: Lesson Plan B-la, " Chemical and Volume Control System."

l l

b o j*

_ . _ - . . . . .,_. _ , _ _ . . , , , _ _ . _ _ _ _ , _ , , . . , _ ..__.-,_._,.._..m_____.___ _ _ _ _ . - _..,..m. _ _ . . _ _ _ _ . . . _ _ _ - _ _ -

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 26 ANSWERS -- DIABLO CANYON 1 -86/07/15-DEFFERDING, L I

l l

ANSWER 2.04 (1.50) i

1. 480 VAC Vital Bus "F" 0.5 through battery charger 11 to the panel [+0.5]

2. 480 VAC Vital Bus "H" 0.5 through battery charger 121 to the panel [+0.5]

3. Battery 11 [+0.5]

REFERENCE

1. Diablo Canyon: Operator Information Manual, Lesson Plan-DC System J-9.

ANSWER 2.05 (1.50)

RCP #2 or RCP #1. [+0.5]

(Other RCPs will not provide adequate spray flow.)This [+1.0 will restore normal spray ] flow.

If applicant explains that RCP #1 can be started resulting in reduced, but adequate, spray flow available; then full credit should be given.

REFERENCE

1. Diablo Canyon: Lesson Plan A-4a.

2. Diablo Canyon: EP OP-1.1, page 12 of 20.

4 I

c. l y c du'r s~ve Ra f L *" $ '

p

. 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 27 ANSWERS -- DIABLO CANYON 1 -86/07/15-DEFFERDING, L ANSWER 2.06 (3.00)

a. 1. The RHR suction from loop four isolation valve (8701) can only open if RCS pressure is below 390 psig and pressurizer vapor space temperature is below 475 degrees F.

2. The other RHR suction from loop 4 isolation valve (8702) can only be opened if RCS pressure is below 390 psig and will automatically close if RCS pressure exceeds 700 psig.

3. Relief valves are located on the RHR pump suction, the cold leg discharge header, and the hot leg discharge header.

4. Two PORVs are reset to less than 450 psig.

[+0.5] each

. The breakers for the suction valves (MOV 8701 and 8702) are maintained in the open position (except when it is necessary to strokethevalves). [+1.0]

REFERENCE

1. Diablo Canyon: Lesson Plan B-2.

2. Diablo Canyon: RHR System Operating Procedure OP B-2:V, RHR Place in Service During Plant Cooldown.

ANSWER 2.07 (1.00)

a. six detectors [+0.5]

b. fission chamber detectors [+0.5]

REFERENCE ,

1. Diablo Canyon: Lesson Plan LB-5.

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 28 ANSWERS -- DIABLO CANYON 1 -86/07/15-DEFFERDING, L ANSWER 2.08 (2.00) g:

Centrifugal charging pump (s) 2500 psig, safety injection pump (s) 1520 psig, accumulator (s) 650 psig, and RHR pump (s) 175 psig.

[+0.3] for each component, [+0.2] for each pressure REFERENCE

1. Diablo Canyon: Lesson Plan B-3.

ANSWER 2.09 (3.50)

a. Allows for additional seal injection water flow through the pump bearing for cooling purposes. [+0.6] Primary system pressure must be >100 psig +-10 [+0.2] and (1000 psig +-100

[+0.2].

b. Relief valve directed to the PRT. [+1.0]

Decrease. [+0.5]

d. VCT pressure and RCDT pressure. [+1.0]

REFERENCE 1

1. Diablo Canyon: Lesson Plan A-6.

ANSWER 2.10 (1.50)

I 1. Maintain feedwater control valves in a linear range for better throttling characteristics. [+0.5]

2. Reduce pump power requirements at part load. [+0.5]

3. Reduce the possibility of valve plug erosion due to excessive closure at part load. [+0.5]

I REFERENCE ,

Diablo Canyon: Lesson Plan C-8b.

1.

i l

- -- - - , - , . ,-,,-,----,-n-,, -,.,-,,,c- - . - - - . , - - - - - , . , - . - . , ,

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 29 ANSWERS -- DIABLO CANYON 1 -86/07/15-DEFFERDING, L ANSWER 2.11 (2.00)

a. Two concentric 0-ring seals [+0.5] are used. The inner diameter ~

of rings is slotted to allow RCS pressure to inflate or expand ring and thus sealing closure head. [+0.5]

b. Any leakage past the inner 0-rtrg would pass through a leak detection flow path [+0.25], that includes a temperature element and associated alarm. [+0.25] (Temperature can be read on VB-2.)

c. Since this leakage is collected and known it would be classified as identified leakage. [+0.5]

REFERENCE

1. Diablo Canyon: Lesson Plan LA-2a.

ANSWER 2.12 (1.00)

1. To preheat the lines downstream of the spray valves in order to minimize thermal shock when the valve opens. [+0.5]

2. Maintain uniform water chemistry in the PZR and RCS (particularly boron concentration). [+0.5]

REFERENCE l 1. Diablo Canyon: Lesson Plan A-4a.

l l

ANSWER 2.13 (1.00)

This criteria (along with verifying both safety injection pumps inoperable) ensures that a mass addition pressure transient will be adequately relieved by a single PORV. [+1.0] g 8p/8/c fru/uq ,

REFERENCE

1. Diablo Canyon: Lesson Plan B-la.

I l_ _ _ _ _ . - _ . _ _ _ _ _ . . . - . _-.

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 30 ANSWERS -- DIABLO CANYON 1 -86/07/15-DEFFERDING, L ANSWER 2.14 (1.00)

a. Intermediate range indication for that channel will increase slightly. [+1.0 REFERENCE

1. Diablo Canyon: Lesson Plan LB-4.

ANSWER 2.15 (2.00) r

a. De-energize 13D and 13E. [+1.0]

b. 13D and 13E are the power supplies to the two Control Rod Drive Motor Generators. De-energizing the power supply to the CRDMGs 1 will cause a loss of power to the Control Rod Drive Assemblies i and cause the Control Rods to drop into the core even though the Reactor Trip breakers remain closed. [+1.0]

REFERENCE i 1. Diablo Canyon: Lesson Plan LJ-7.

2. Diablo Canyon: EP FR-S.1, Re/. O.

l l

e l

l l .

3. INSTRUMENTS AND CONTROLS PAGE 31 ANSWERS -- DIABLO CANYON 1 -86/07/15-DEFFERDING, L ANSWER 3.01 (2.50)

' ^

a. 554 deg F

b. 543 deg F

c. 547 deg F

d. 541 deg F

e. plus or minus 3 deg F

[+0.5] each REFERENCE

1. Diablo Canyon: Lesson Plan LA-2c.

ANSWER 3.02 (2.00)

a. I

b. 5

c. 6

d. 4

[+0.5] each REFERENCE

1. Diablo Canyon: Lesson Plan LA-2c.

ANSWER 3.03 (2.00)

a. high radiation alarm initiates containment ventilation isolation

[+0.5]

b. high activity initiates closurd of surge tank vent (RCV-16)

[+0.5]

c.

high activity) initiates tanks (RCV-17 [+0.5] closure of discharge from gas decay

d. no automatic actions [+0.5] ,

REFERENCE

1. Diablo Canyon: Lesson Plan G-3, " Area Radiation Monitor System."

I i

_ -.~ . ,.. - . . , _ _ . _ . _ _ _ _ _ . _ . _ _.___. , _ _ _ _ . . . _ . _ _ _ _ . . . ~ . - _ _ , , _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ , _ . _ -

3. INSTRUMENTS AND CONTROLS PAGE 32 ANSWERS -- DIABLO CANYON 1 -86/07/15-DEFFERDING, L

2. Diablo Canyon: Lesson Plan G-4, " Process Radiation Monitors."

ANSWER 3,04 (3.00) a.

Steam the RCS, flow increases reducing Tave.causing [+0.5 ]Tave increased

- Tref removal deviationof heat from causes rod control circuit to withdraw rods to restore Tave. [+0.5]

b. (This causes reduced efficiency in the secondary plant cycle for the same turbine load output.) Tave will decrease because of greater heat removal. [+0.5] Tave - Tref deviation causes rod control circuit to withdraw rods to restore Tave. [+0.5]

c. This causes decreased N-44 output. Rod control sees auctioneered high nuclear power, [+0.5] so no rod movement. [+0.5]

REFERENCE

1. Diablo Canyon: Lesson Plan A-3.a.

ANSWER 3.05 (1.50)

a. TRUE [+0.5]

b. FALSE 0.5

c. TRUE [+[+0.5))

REFERENCE i

1. Diablo Canyon: Lesson Plan B4.

1 9'

, - - - - - - . - _ _ ~ . . . - ._ . . . . - . _ . . . - - . - -.

3. INSTRUMENTS AND CONTROLS PAGE 33 ANSWERS -- DIABLO CANYON 1 -86/07/15-DEFFERDING, L ANSWER 3.06 (2.00)

1. High steam flow [+0.5]

Variable, 1/2 sensors on 2/4 loops [+0.1]

AND

a. Low steam [+0.5]

600 psig [ pressure +0.1],2/4 loops [+0.1]

OR

b. Low-low Tave [+0.5]

543 deg F [+0.1], 2/4 loops [+0.1] gy ,

REFERENCE

~

1. Diablo Canyon: Lesson Plan B-6a.

ANSWER 3.07 (2.00) .

a. General Warning Annunciator [+0.5]

Rx trip bypass breaker shut irdication on VB2 [+0.5]

b. (Generalwarningonbothtrains 0.5] causing a reactor trip t i"/*"0""

[+0. 5] . e sk <d f* * ','

LR s ecep t REFERENCE g r r, /p /'-/'

credal

1. Diablo Canyon: Lesson Plan B-6b.

I O

3. INSTRUMENTS AND CONTROLS PAGE 34 ANSWERS -- DIABLO CANYON 1 -86/07/15-DEFFERDING, L ANSWER 3.08 (3.00)

a. 1. SI [+0.5] or high radiation [+0.5] (plant vent of particulate or gas or containment particulate).-pes), 0 /M O - M

.' fee / 1 Hi- >Vi t**t T2 ( 2 2 P s ip )

g I 749) [+0.33] Mc t % rply b *I' M 74 ' ~ I 6= " 't-RCP thermal barrier CCW return isolation (FCV 357a 750) 11 .3 CCW supply header isolation (FCV 355 and 356) [+0.33]

1. Diablo Canyon: Lesson Plan B-6c. dw e4 /06 7 'V 54"'

ANSWER 3.09 (2.00)

(587") wide range upper and narrow range upper 12' or 144" (443") narrow range lower (12") wide range lower tube siteet

[+2.0]

l REFERENCE

1. Diablo Canyon: Lesson Plan A-5, Figure SG.3.

I

4 .

, 3. INSTRUMENTS AND CONTROLS PAGE 35 ANSWERS -- DIABLO CANYON 1 -86/07/15-DEFFERDING, L ANSWER 3.10 (2.00)

a. 1. 10**5 cps [+0. 5] g j,,3 o e -

2.(when1of2IR'sexceed10**-10ampsfP-6)sourcerangetrip N*I can be manually blocked. [+1.0]

b. Auto reset when both IR's read less than 10**-10 amps. [+0.5] / #' *" #

1 e f z. in > 'o'"

REFERENCE o ccast o,a s sa ly

1. Diablo Canyon: Lesson Plan B.4. 4 4a -

ANSWER 3.11 (2.00)

1. Low vital header pressure with no SI and no transfer to EDG

2. Bus transfer to diesel generator

3. SI4tithnoundervoltageon4KVbuses*2 rf I v.- N 6 M

• CCW pump starts regardless of oil pressure 1 . m i f 8 t s ca s, u u, pp u ;r ,nt

[+0.5] each 5" REFERENCE ,

1. Diablo Canyon: Lesson Plan LF-2.

ANSWER 3.12 (1.00)

The necessary conditions are:

1. hi-hi level on 1 of 4 S/Gs (at 67% (P-14)) [+0.33]

2. Safety Injection Signal, (train A or train B) [+0.33]

3. Reactor tri (P-4 with low Tavg ((less than 554) on 2 of 4 channels) +0.33 REFERENCE

1. Diablo Canyon: Lesson Plan LC-8a.

., n. - - , - - ._.--..n-- , - - - , < _ _ , . , _ , . - ~ , - - - - - - . .-.,-,---_--,,n.---,,----n- - - - - . _ _ . _ _ - . _ - ____.

__ ~.

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 36 RADIOLOGICAL CONTROL ANSWERS -- DIABLO CANYON 1 -86/07/15-DEFFERDING, L ANSWER 4.01 (3.00)

a. 1. RCS subcooling greater than 20 deg F

2. Secondary heat sink greater than 460 gpm aux feed flow OR greater than 4% NR on 1 S/G.

3. RCS pressure stable or increasing

4. PZR level greater than 4%

b. 1. RCS subcooling less than 20 deg F

2. PZR level cannot be maintained above 4%

[+0.5] each REFERENCE

1. Diablo Canyon: EP-0 and EP 1.1.

t ANSWER 4.02 (2.50)

a. TRUE(AccidentMonitoringInstrumentation) [+0.5]

b. Six Incore Thermocouples [+0.5]

Four Wide Range Hot Leg Temperatures [+0.5]

Two Wide-Range RCS Pressure [+0.5]

l c. FALSE [+0.5 l

l REFERENCE

1. Diablo Canyon: Lesson Plan LA-2c.

2. Diablo Canyon: Technical Specifications, p. 3/4.3.3.6.

ANSWER 4.03 (2.00) ,

1. Any Nuclear Channel having an unexpected increase by a factor i of 5.

2. All Nuclear Channels having an unexpected increase by a factor of 2.

3. Any unexpected RCS temperature change by greater than 10 deg F.

4. Any RCS boron change by greater than 20 ppm.

5. Communications lost between the Control Room and Containment. .

l Any four (4) [+0.5] each, +2.0 maximum.

i 1

. _ - . . . _ _ - _ _ _ _ _ ._ ~ . _ __ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ . _ _ . . . . _ _ _ _ - _ _ . , _

^;, 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 37 RADIOLOGICAL CONTROL ANSWERS -- DIABLO CANYON 1 -86/07/15-DEFFERDING, L REFERENCE

1. Diablo Canyon: Lesson' Plan LB-8. -

ANSWER 4.04 (2.50)

a. 1. By depressing the SINGLE VALVE / SEQ VALVE push button on the DEH control panel. [+0.5]

2. In single valve, all four of the governor valves respond as one valve. [+0.5] If selected to sequential valve, the valves open in the predetermined order (presently 1, 2, and 3 as a bank first, then 4). [+0.5]

b. In IMP IN, the generator load reference signal is compared to a signal representative of the turbine first stage impulse pressure (PT-8). [+1.0]

REFERENCE l 1. Diablo Canyon: Lesson Plan LC-3c.

2. Diablo Canyon: Lesson Plan L-2, p. 17.

ANSWER 4.05 (1.00)

Various parameters are pre-programmed to be trended (such as S/G level, pressurizer pressure and level, RCs temperatures, etc.) on an EIGHT-SECOND CYCLE. [+0.5] A reactor trip will cause the computer to store all information trended 2 minutes prior to the trip to 3 minutes after the trip. The P-250 will print out this trend and allow the operator to evaluate the trip based on the trended conditions. [+0.5]

REFERENCE

1. Diablo Canyon: Lesson Plan LJ-11.

. - - . _ _ - - - . - - . - ,, - - . , . , - - - - - - - - - , -,.._,...,..--,-..-.,--,,,---,------,n --.,-_..n. -

. 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 38 RADIOLOGICAL CONTROL ANSWERS -- DIABLO CANYON 1 -86/07/15-DEFFERDING, L ANSWER 4.06 (2.00)

1. Multiple indicators fail to their deenergized position (the yellow color dot on the meter and recorder faces indicate that thisisalossofPY-24).

2. Multiple Hagan controllers will be inoperable, fail to MANUAL or AUTO HOLD. (These also are identified by a yellow color dot.)

3. All postage stamp status lights for reactor protection channel 4 will be off.

Any two (2) [+1.0] each, +2.0 maximum.

REFERENCE

1. Diablo Canyon: Lesson Plan LJ-10.

2. Diablo Canyon: Lesson Plan AP-4.

ANSWER 4.07 (1.00) i a. Shift foreman. [+0.5]

b. 1. Liaison coordinator (for offsite notification). [+0.25]

2. Operationscoordinator(whohasatleastanR0 license).

[+0.25]

REFERENCE

1. Diablo Canyon: Administrative Procedures, NPAP 5-5/NPG-2.3.

I 1

t

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 39 RADIOLOGICAL CONTROL ANSWERS -- DIABLO CANYON 1 -86/07/15-DEFFERDING, L ANSWER 4.08 (3.50)

a. 1. subcriticality

2. core cooling

3. heat sink

4. Integrity

5. containment

6. Inventory

, [+0.25] each

b. RED [+0.25] - stop optimal recovery in progress and immediately initiate indicated functional restoration unless there is a higher red [+0.25]

M erreh M [+0.25] - finish status tree check then initiate functional restoration of highest ranking orange (assume no reds) [+0.25]

YELLOW [+0.25] - monitor trees. Operator prerogative on continuing optimal recovery or functional recovery. [+0.25]

GREEN [+0.25] - no action required [+0.25]

REFERENCE j 1. Westinghouse: E0P Guidelines.

l l

ANSWER 4.09 (1.50)

a. 2 r/hr x 1**2/2**2 = 500 mr/hr at work location 500 x 3/4 = 375 mr [+1.0]

b. (10 CFR 20 limits 1250 mrem.) 'Yes, you exceeded the limit.

l [+0.5]

REFERENCE j

1. Generic: 10 CFR 20.

l' 2. Westinghouse: Radiation, Chemistry and Corrosion.

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 40

. RADIOLOGICAL CONTROL ANSWERS -- DIABLO CANYON 1 -86/07/15-DEFFERDING, L ANSWER 4.10 (2.00)

a. assistant plant manager, plant superintendent or his delegate.

[+0.5]

b. 1. an engineer (assistant plant manager-technical services, engineering manager, power production engineer or the STA) [+0.5] ,

2. a licensed SRO [+0.5] A te# 540 * /ad= 4 < :e co - / e /%

(o r efe

• V %f Add / mem f oi e s te .

c. no one, the licensed operator has authority and responsibility

[+0.5]

REFERENCE

1. Diablo Canyon: Administrative Procedure, NPAP A100.

ANSWER 4.11 (1.50)

1. a minimum contained borated water volume of 400,000 gallons

2. a boron concentration of between 2000 and 2200 ppm

3. a minimum solution temperature of 35 degrees F

[+0.5] each REFERENCE

1. Diablo Canyon: Lesson Plan LB-3.

a s

O

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 41

. RADIOLOGICAL CONTROL ANSWERS -- DIABLO CANYON 1 -86/07/15-DEFFERDING, L ANSWER 4.12 (1.50)

1. All penetrations required to be closed during accident conditions are automatic either isolation a) capable system of being [+0.25], closed or bby)an closedOPERABLE to manual containment valves, blind flanges or deactivated automatic valves secured in their closed position. [+0.25]

2. All equipment hatches are closed and sealed. [+0.5]

3. Each air lock is operable. [+0.5]

4. Containment leak rates are within limits. [+0.5]

5.

i Thesealing[mechanismassociatedwitheachpenetrationis OPERABLE. +0.5]

Any three (3), +1.5 maximum.

REFERENCE

1. Diablo Canyon: Lesson Plan LB-6c.

ANSWER 4.13 (1.00)

There is an interlock mechanism inside the personnel airlock which can be defeated. This allows the outer door to be opened even if the containment side door is not closed. [+1.0]

REFERENCE

1. Diablo Canyon: EP M-8.

I e

EQUATION SHEET Where,mg = m2 (density)i(velocity)1(area)i = (density)2(velocity)2(area)2 where V = specific KE = -}2PE = agh PE + KEg+P Y l 1 1 = PE 2+KE 2+P Y22 volume P = Pressure O " "Cp (Tout-Tj ,) Q = UA (T,y,-Tstm) Q = m(hi-h2) 26.06 T = (B-p)t P = P,10(SUR)(t) P = Po et /T SUR = 7 p

{ _

,e delta K = (K,ff-1) CRg(1-K,ff3) = CR2 (1-Keff2) CR = S/(1-K,ff) 1 (1_Keffi) SDM =

(1-Xeff) x 100%

M = (1-Keff2I E eff decay constant = In (2) = 0.693 A g=A*e-(decayconstant)x(t) t 1/2 *1/2 Water Parameters Miscellaneous Conversions 1 gallon - 8.345 lbs 1 Curie = 3.7 x 10 10 dps .

1 gallon = 3.78 liters 1 kg = 2.21 lbs 1 ft3 = 7.48 gallons I hp = 2.54 x 103 Btu /hr -

3 6

Density =62.4lbg/ft 1 W = 3.41 x 10 Btu /hr .

Density = 1 gn/cm 1 Btu - 778 ft-lbf Heat of Vaporization = 970 Btu /lbe DegreesF=(1.8xDegreesC)+32

, Heat of Fus. ion = 144 Btu /lba 1 inch = 2.54 centimeters 1 1 Ata = 14.7 psia = 29.9 in Hg g = 32.174 ft-lbs/lbf-sec 2 '

)

1 l

i

- . - - , - - - - , , . --~ v - ., . . . , - . _ , . _ , . , , , , __-- ---. n,_ ___ - - . . . ~ . - - , - - - -

W a.

U.S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION

~

. Facility: ninhio canven Units 1 and 2 Reactor Type: Unacinohnuse PWR Date Administered: aniv ts. 1986 Examiner: p.31 untr411 Candidate:

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 indi-cated 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 25 5. Theory of Nuclear

• Power Plant Operation, Fluids, and Thermo-dynamics 25 25 6. Plant Systems Design, Control, and

Instrumentation 25 25 7. Procedures - Normal, Abnormal, Emergency, j

' and Radiological Control i 25 25 8. Administrative Pro-

! cedures, Conditions, and Limitations

.- 100 Totals ~

Final Grade All work done on this examination is my own, I have neither given nor received aid.

( ,

l Candidate's Signature l

ES-201-2 ATTACHMENT 2

REQUIREMENTS FOR ADMINISTRATION OF WRITTEN EXAMINATIONS -

I

! 1. A single room shall be provided for completing the written examina-4 tion. The location of this room and supporting restroom facilities shall be such as to prevent contact with all other facility and/or i contractor pert.onnel during the duration of the written examination.

If necessary, the facility should make arrangements for the use of i a suitable room at a local school, motel, or other building. Ob-i taining this room is the responsibility of the licensee.

i

2. Minimum spacing is required to ensure. examination integrity as determined by the chief examiner. Minimum spacing should be one candidate per table, with a 3-ft space between tables. 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 candi-dates are to have lunch, coffee, or other refreshments. These arrangements shall comply with Item 1 above. These arrangements shall be reviewed by the examiner and/or proctor.

4. The facility staff shall be provided a copy of the written examination and answer key after the last candidate has completed and handed in J, C his written examination. The facility staff shall then have five working days to provide formal written comments with supporting documentation on

! the examination and answer key to the chief examiner or to the regional i office section chief.

5. The licensee shall provide pads of 8-1/2 by 11 in, lined paper in unopened packages for each candidate's use in completing the exam-

ination. The examiner shall distribute these pads to the candidates.

i All reference material needed to complete the examination shall be furnished by the examiner. Candidates can bring pens, pencils, calculators, or slide rules into the examination room, and no other equipment or reference material shall be allowed.

i 6. Only black ink or dark pencils should be used for writing answers I

to questions.

i l

lL Examiner Standards 11 of 18

ES-201-2 NRC RULES AND GUIDELINES'FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:

1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.

2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.

3. Use black ink or dark pencil only to facilitate legible reproductions.

4. Print your name in the blank provided on the cover sheet of the examination.

5. Fill in the date on the cover sheet of the examination (if necessary).

6. Use only the paper provided for answers.

7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.

8. Consecutively number each answer sheet, write "End of Category " as appropriate, start each category on a new page, write only one side of the paper, and write "Last Page" on the last answer sheet.

9. Number each answer as to category and number, for example, 1.4, 6.3.

10. Skip at least three lines between each answer.

11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.

12. Use abbreviations only if they are commonly used in facility literature.

13. The point value for each question is indicated in 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 AFD 00 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. -

Examiner Standards 12 of 18

ES-201-2 i

F 18. When you complete your examination, you shall:

a. Assemble your examination as follows:

1 .

(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. I

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.

1 i

l i

i i

i Examiner Standards 13 of 18

\

_ -_ _ _ ~ _ _ - _ _ - . . . , , , _ . _ ,

l J'

EQUATION SilEET f = ma v = s/t

"*E w = mg s=vt+ ls at 2 Cycle efficiency =

f]t)

E = mC a = (vg - vg )/t KE = %

2 v g=v + at A = AN A=Ae g -t PE = m8h m = 0/t A = in 2/tg = 0.693/tg W = v4P AE = 931Am t q(eff) = (t )(ts) i (tg+t) b '

Q=[nCAT

, P I.Ieo -EX Q = UAAT I.Ie -UX Pwr = W g It I.I o lo -XM P=P 10 SUR(t) TVL = 1.3/u yp t o

e /T HVL = 0.693/u SUR = 26.06/T T = 1.44 DT SCR = S/(1 - K,gg) fA

• o\

SUR = 26 1 g I

CR = S/(1 - K,gg )

1(

~

eff}1 " 2(1 *K,gg)2 T = (1*/p ) + [(f 'p)/x ff] o T = 1*/ (p - F) M = 1/(1 - K,ff) = CR /CR g 0

"( ~ P)! eff P M = (1 - K,gg)0! ( ~

eff)1 o = (K,ff-1)/K,gg = AK,gg/Keff SDM = (1 - K,gg)/K,gg

~

p= [ 1*/TK,'gg .] + [B/(1 + A,ggT )] ,

g* = 1 x 10 seconds

-1 P = I(V/(3 x 10 0) A,gg = 0.1 seconds I = No Idg1=Id22 WATER PARAMETERS Idg =Id2 1 gal. = 8.345 lbm R/hr = (0.5 CE)/d (meters)

I gal. = 3.78 liters R/hr = 6 CE/d (feet)

I ft = 7.48 gal. MISCELLANEOUS CONVERSIONS .

10 Density = 62.4 lbm/ft 1 Curie = 3.7 x 10 dps Density = 1 gm/cm 1 kg = 2.21 lbm Heat of vaiorizationi = 970 Etu/lbm I hp = 2.54 x 10 3BTU /hr -

6

, Heat of fusica = 144 Btu /lbm 1 Mw = 3.41 x 10 Btu /hr k 1 Atm = 14.7 psi = 29.9 in. I'g- 1 Btu = 778 ft-lbf 1 f t.110 = 0,4333 lbf/in 1 inch = 2.54 cm 2

F = 9/5 C + 32 C = 5/9 ( F - 32)

E

.. ., i 1

Properties of Saturated Steam and Saturated Water

• Absolute Pressure Vacuum Temper. Heat of Latent Heat Total Heat Specific Volume inches Inches the of Lbs.N.r Sq. of Hg of Hg ature Liquid Evaporation of Steam y P' o...

g

. p. storm. aio ris.

b Water Steam neo ris. co.r. ,,,is co.re. ,.,is.

0.0087 0.02 29.90 32.018 0.0003 1075.5 1075.5 0.10 U.016022 3302.4 0.20 29.72 35.023 3.026 1073.8 1076.8 0.016020 0.15 0.31 2945.5 29.61 45.453 13.498 1067.9 1081.4 0.016020 0.20 0.41 2004.7 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 0.30 0.61 0.016032 1235.5 29.31 64.484 32.541 1057.1 1089.7 0.016040 0.35 0.71 1039.7 29.21 68.939 36.992 1054.6 1091.6 0.016048 0.40 0.81 898.6 29.11 72.869 40.917 1052.4 1093.3 0.016056 0.45 0.92 29.00 792.1

, 76.387 44.430 1050.5 1094.9 0.016063 708.8 0.50 1.02 28.90 79.586 47.623 1048.6 10 % .3 0.60 1.22 0.016071 641.5 28.70 85.218 53.245 1045.5 1048.7 0.016085 0.70 1.43 540.1 28.49 90.09 58.10 1042.7 1100.8 0.016099 O.80 1.63 28.29 466.94 94.38 62.39 1040.3 1102.6 0.016112 411.69 0.90 1.83 28.09 98.24 66.24 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

. 1.6 3.26 26.66 117.98 85.95 1026.8 1112.7 1.8 3.66 0.0161 % 214.33 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 2.2 1116.2 0.016230 173.76 4.48 25.44 129.61 97.57 1020.1 1117.6 2.4 4.89 0.016245 158.87 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 2.8 7.70 24.22 138.78 106.73 1014.7 1121.5 0.016287 126.67 3.0 6.11 23.81 141.47 109.42 1013.2 1122.6 3.5 0.016300 118.73 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 90.44 4.5 9.16 20.76 157.82 125.77 1003.5 1129.3 0.016384 83.03 5.0 10.18 19.74 162.24 130.20 1000.9 1131.1 0.016407 73.532 5.5 11.20 18.72 166.29 134.26 998.5 1132.7 6.0 0.016430 67.249 12.22 17.70 170.05 138.03 996.2 1134.2 6.5 0.016451 61.984 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 7.5 15.27 0.016441 53.650 14.65 179.93 147.93 990.2 1138.2 0.016510 50.294 8.0 16.29 13.63 182.86 150.87 982.5  !!39.3 8.5 0.ul6527 47.345 17.31 12.61 185.63 153265 986.8 1140.4 0.016545 9.0 44.733 18.32 11.60 188.27 156.30 985.1 1141.4 0.016561 9.5 19.34 42.402 -

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 33.420 11.0 22.40 7.52 197.75 165.82 979.3 1145.1 12.0 24.43 0.01 % 22 35.142 5.49 201.% 170.05 976.6 1146.7 0.016650 32.394 13.0 26.47 3.45 205.88 174.00 974.2 1148.2 0.016676 30.057 14.0 38.50 1.42. 209.56 177.71 971.9 1149.6 0.016702 28.043 Pressure Temper- Heat of Latent Heat Total Heat Specific Volume Lbs. per Sq. In. ature the of of Steam Absolute Gage Liquid Evaporation y P' P t

o.m.. v.

k Water 6 team siuns. seo ne. s orm. cu.ri. m. co. re. ,,, is.

14.696 0.0 212.00 180.17 970.3 1150.5 0.016719 15.0 0.3 213.03 26.799 181.21 969.7 1150.9 0.016726 26.290 16.0 1.3 216.32 184.52 %7.6 1152.1 17.0 0.016749 24.750 2.3 219.44 187.66 %5.6 1153.2 0.016771 18.0 3.3 23.385 222.41 190.66 %3.7 - 1154.3 0.016793 19.0 4.3 225.24 193.52 22.168

%I .8 1155.3 0.016814 21.074 20.0 5.3 227.%. 1%.27 %0.1 1156.3 21.0 0 016834 20.087 6.3 230.57 198.90 958.4 1157.3 0.016854 22.0 7.3 233.07 19.190 201.44 956.7 1158.1 0.016873 23.0 8.3 235.40 18.373 203.88 955.1 1159.0 0.016891 24.0 9.3 237.82 206.24 17.624 953.6 1159.8 0.016909 16.936 25.0 10.3 240.07 204.52 952.1  !!60.6 26.0 11.3 0.016927 16.301 242.25 210.7 950.6 1161.4 0.016944 27.0 12.3 244.36 15.7138 212.9 94.9.2 1162.1 0.016 % I 15.1684 28.0 13.3 246.41 214.9 947.9 1162.8 0.016977 14.6607 29.0 14.3 248.40 217.0 946.5 1I63.5

• 0.016993 14.1869 30.0 15.3 250.34 218.9 945.2 1864.1 31.0 16.3 0.017009 13.7436 252.22 220.8 943.9 1164.8 32.0 17.3 0.017024 13.3280 254.05 222.7 942.7 1165.4 33.0 18.3 0.017039 12.9376 255.84 124.5 941.5 1166.0 0.017054 34.0 19.3 257.58 12.5700 226.3 940.3 11 % .6 0.017069 12.2234

Properties of Saturated Steam and Saturated Water-continued Pressure Temper. Heat of Lbs. pe? Sq. In. Latent Heat Total Heat Specific Volume sture the of Absol,ute Gage i Liquid Evaporation of Steam t

p P n o..c "8 Water

. F. seuris. neutid. neuria.

i Steam

~

cu. re. c., is. I co. re. c., in.

35.0 20.3 259.29 228.0 36.0 939.1 1167.1 0.017083 21.3 260.95 229.7 938.0 - 11.8959 37.0 22.3 262.58 1167.7 0.017097 11.5860 231.4 936.9 1168.2 38.0 23.3 264.17 233.0 0.017111 11.2923 39.0 24.3 935.8 1168.8 0.017124 265.72 234.6 934.7 11.0136 40.0 1169.3 0.017138 10.7487 25.3 267.25 136.1 41.0 26.3 933.6 1169.8 0.017151 268.74 237.7 932.6 10.4965 42.0 27.3 270.21 1170.2 0.017164 10.2563 239.2 931.5 1170.7 43.0 28.3 271.65 240.6 0.017177 10.0272 44 ft 19.3 930.5 1171.2 0.017189 273.06 242.1 929.5 9.8083 45.0 30.3 1171.6 0.017202 9.5991 274.44 143.5 928.6 1172,0 46.0 J1.3 275.80 244.9 0 017214 9.3988 47.0 32.3 927.6 1172.5 0.017226 277.14 246.2 926.6 9.2070 48.0 33.3 278.45 1172.9 0.017238 9.0231 247.6 925.7 1173.3 49.0 34.3 279.74 248.9 0.017250 8.8465 924.8 1173.7 0.017262 50.0 35.3 281.02 250.2 8.6770 51.0 36.3 923.9 1174.1 U.017274 282.27 251.5 923.0 8.5140 52.0 37.3 283.50 1174.5 0.017285 8.3571 - -

252.8 922.1 1174.9 53.0 38.3 284.71 254.0 0.0172 % 8.2061 54.0 39.3 921.2 1175.2 0.017307 285.90 255.2 920.4 8.0606 55.0 1175.6 0.017319 7.9203 40.3 187.08 256.4 56.0 919.5 1175.9 0.017329 41.3 288.24 257.6 918.7 7.7850 57.0 42.3 289.38 1176.3 0.017340 7.6543 258.8 917.8 1176:6 58.0 43.3 290.50 259.9 0.017351 7.5280 59.0 917.0 1177.0 =

0.017362 44.3 291.62 261.1 7.4059 916.2 1I77.3 0.017372 60.0 45.3 192.71 262.2 7.2879 61.0 915.4 1177.6 0.017383 46.3 293.79 263.3 7.1736 62.0 47.3 914.6 1177.9 0.017393 294.86 264.4 913.8 7.0630 63.0 48.3 295.91 1178.2 0.017403 6.9558 265.5 913.0 1178.6 64.0 49.3 2 % .95 266.6 0.017413 6.8519 912.3 1178.9 0.017423 65.0 50.3 197.98 267.6 6.7511 66.0 911.5 1879.1 0.017433 51.3 298.99 268.7 6.6533 67.0 52.3 910.8 1179.4 0.017443 299.99 269.7 910.0 6.5584 68.0 53.3 300.99 1179.7 0.017453 6.4662 -

270.7 909.3 1180.0 69.0 54.3 301.% 271.7 0.017463 6.3767 908.5 1180.3 0.0D 472 70.0 55.3 302.93 272.7 6.2896 71.0 56.3 907.8 1160.6 0.0174 4 103.89 173.7 407.1 1180.8

'.2050 72.0 57.3 304.83 274.7 0.017491 6.1226 73.0 58.3 906.4 1181.1 0.017501 305.77 275.7 , 905.7 6.0425 74.0 59.3 306 69 1181.4 0.017510 5.9645 276.6 905.0 1181.6 75.0 60.3 307.61 0.017519 5.8885 277.6 904.3 1881.9 76.0 61.3 308.51 0.017529 5.8144

- 77.0 278.5 903.6 1182.1 62.3 309.41 279.4 0.017538 5.7423 78.0 63.3 902.9 1182.4 0.017547 310.29 280.3 902.3 5.6720 79.0 64.3 311.17 1182.6 0.017556 281.3 901.6 1182.8 5.6034 80.0 65.3 J12.04 0.017565 5.5304 282.1 900.9 1183.1 81.0 66.3 312.90 283.0 0.017573 5.4711 82.0 67.3 900.3 1183.3 0.017582 313.75 283.9 899.6 5.4074 83.0 68.3 314.60 1I83.5 0.017591 284.8 899.0 1183.8 5.3451 84.0 69.3 315.43 285.7 0.017600 5.2843 898.3 1184.0 0.017608 85.0 70.3 316.26 286.5 5.2249 86 0 897.7 1884.2 0.017617 71.3 317.08 287.4 897.0 5.1669 87.0 72.J 317.89 1184.4 0.017625 288.2 8%.4 1184.6 5.1101 88.0 73.3 318.69 289.0 0.017634 5.0546 89.0 895.8 1184.8 0.017642 74.3 319.49 289.9 895.2 5.0004 90.0 1185.0 0.017651 75.3 320.25 290.7 894.6 4.9473 91.0 76.3 321.06 1885.3 0.017659 291.5 893.9 1185.5 4.6953 92.0 77.3 321.84 292.3 0.017667 4.8445 93.0 893.3 1185.7 78.3 322.61 193.1 892.7 0.017675 4.7947 44.0 79.3 323.37 1185.9 0.017684 293.9 892.I 1186.0 4.7459 95.0 80.3 324.13 0.017692 4.6982 294.7 891.5 lih6.2

%.0 81.3 324.MM 0.017700 4.6514 97.0 295.5 891.0 1186.4 N 2.3 325.63 2%.3 890.4 0.017708 4.6055 94.0 M3.3 326.36 11 6 .6 0.017716 99 0 297.0 8H9.8 1186.8 4.5606 M4.3 327.10 297.8 0.017724 4.U 66 M74.2 i187.0 400.0 n5.3 J27.M2 0.017732 4.4734 29M.5 8BMA 1l87.2 101.0 M6.3 328.54 0.017740 101.0 299.3 888 I 1187.3 4.4310 M7.3 329.2t. 300.0 887.5 0.01775 4.3895 103.0 4R.3 1887.5 0.01776 329.97 300.8 886.9 4.3487 104.0 M9.3 330.67 1I87.7 0.01776 301.5 R86.4 1187.9 4.3087 105.0 90..I 331.37 0.01777 4.2695 106 0 302.2 8M5.8 1184.0 98.3 332.06 303.0 BMS.2 0.01778 4.2309 107.0 92.3 332.75 1184.2 0.01779 Into 303.7 884.7 1184.4 4.1931 93.3 333.44 304.4 BM4.1 0.01779 4.1560 104 0 94 3 1183.5 0.017A0 334.11 305.1 s53.o 4.1195 i188.7 0.01781 4.0837 L

Properties of Saturated Steam and Saturated Water-continued Pressure Temper- Heat of jLatent Heat ; Total Heat Specific Volume )

Lbs. per Sq. In. ature the i of of Steam p Absolute Gage I Liquid ' Evaporation li 1 j W ater

-st ea m I; ;

P' P our... F. neurm. niu ris. . n.uiis. cu rsc., is. co. ri. c., in.

110.0 95.3 334.79 305.8 883.1 1185.9 UE82 4.04s4 111.0  %.3 335.44 306.5 882.5 1889.0 0.01782 4.0138 112.0 97.3 336.12 3C7.2 882.0 1189.2 0 017x3 3.9794 .

113.0 98.3 336.78 307.9 551.4 1189.3 0.01784 3.9464 114.0 99.3 337.43 308.6 880.9 1889.5 0.01785 3.9136

!!5.0 100.3 338.08 309.3 880.4 1189.6 0.01785 3.8513 116.0 101.3 338.73 309.9 879.9 1189.8 0.01786 3.8495 117.0 102.3 339.37 310.6 879.3 1189.9 0.01787 3.8183 e 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 1190.2 0.01788 3.7573 120.0 105.3 341.27 312.6 877.8  !!90.4 0.01789 3.7275 121.0 106.3 341.89 313.2 877.3 1190.5 0.01790 3.6983 g 122.0 107.3 242.51 313.9 876.8 1190.7 0.01790 3.6695 123.0 108.3 343.13 314.5 876.3 1190.8 0.01791 3.6411 124.0 109.3 343.74 315.2 875.8 1190.9 0.01792 3.6132 125.0  !!0.3 344.35 315.8 875.3 1191.1 0.01792 3.5857 126.0 111.3 344.95 316.4 874.8 1193.2 0.01793 3.5586 *

• 127.0 112.3 345.55 317.1 874.3 1191.3 0.01794 3.5320 128.0 113.3 346.15 317.7 873.8 1191.5 0.01794 3.5057 129.0 114.3 346.74 318.3 873.3 1191.6 0.01795 3.4799

• 130.0 115.3 347.33 319.0 872.8 1891.7 0.017 % 3.4544 -

131.0 116.3 347.92 319.6 872.3 1191.9 0.01797 3.4293 132.0 117.3 348.50 320.2 871.8 1192.0 0.01797 3.4046 133.0 118.3 349.08 320.8 871.3 1192.1 0.01798 3.3802 134.0 119.3 349.65 321.4 870.8 1192.2 0.01799 3.3562 135.0 120.3 350.23 322.0 870.4 1192.4 0.01799 3.3325 136.0 121.3 350.79 322.6 869.9 1192.5 0.01800 3.3091 137.0 122.3 351.36 323.2 869.4 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 140.0 125.3 353.04 325.0 868.0 1193.0 0.01803 3.2190 141.0 126.3 353.59 325.5 867.5  !!93.1 0.01803 3.1972 142.0 127.3 354.14 326.1 867.1 1193.2 0.01804 3.1757 143.0 128.3 354.69 326.7 866.6 1193.3 0.01805 3.1546 144.0 129.3 355.23 327.3 866.2 I193.4 0.01805 3.1337 145.0 130.3 J55.77 327.8 865.7 1193.5 0.01806 3.1130 146.0 131.3 356.31 328.4 865.2 1193.6 0.01806 3.0927 147.0 132.3 356.84 329.0 864.8 1193.8 0.01807 3.0726 148.0 133.3 357.38 329.5 864.3 1193.9 0.01805 3.0528 149.0 134.3 357.91 330.I' 863.9 1194.0 0.01808 3.0332 150.0 135.3 358.43 330.6 863.4 1194.1 0.01809 3.0139 152.0 137.3 359.48 331.8 862.5 1194.3 0.01810 2.9760 154.0 139.3 360.51 332.8 861.6 1194.5 0.01812 2.9391 156.0 141.3 361.53 333.9 860.8 1194.7 0.01813 2.9031 158.0 143.3 362.55 335.0 859.9 1194.9 0.01814 2.8579 160.0 145.3 363.55 336.1 859.0 1195.1 0.01815 2.8336 162.0 147.3 364.54 337.1 858.2 1195.3 0.01817 2.8001 164.0 149.3 365.53 338.2 857.3 '1195.5 0.01818 2.7674 166.0 151.3 366.50 339.2 856.5 1195.7 0.01819 2.7355 168.0 153.3 367.47 340.2 855.6 1195.8 0.01820 2.7043 170.0  !$5.3 363.42 341.2 854.8 11 % .0 0.01821 2.6738 172.0 157.3 369.37 342.2 853.9 11 % .2 0.01823 2.6440 174.0 159.3 370.31 343.2 853.1 11 % .4 0.01824 2.6149 176.0 161.3 371.24 344.2 852.3 11 % .5 0 01825 2.5864 178.0 163.3 372.16 345.2 851.5 11 % .7 0.01826 2.5585 180.0 165.3 373.08 346.2 850.7 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 9.01831 2.4527 188.0 173.3 376.65 350.0 847.5 1197.5 0.01832 2.4276 190.0 175.3 377.53 350.9 846.7  !!97.6 0.01833 2.4030 192.0 177.3 378.40 35t.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 I198.1 0.01836 2.3322 198.0 183.3 380.% 354.6 843.0 1198.2 0.01838 2.3095 200.0 185.3 381.80 355.5 842.8 1898.3 0.01539 2.28728 205.0 190.3 383.88 357.7 840.9 1898.7 0.01841 2.23349 210.0 195.3 385.91 359.9 839.1 1199.0 0.01844 2.18217 -

215.0 200.3 387.91 362.1 837.2 1899.3 0.01847 2.13315 220.0 205.3 389.88 364.2 835.4 1199.6 ' O.01850 2.08629 225.0 210.J 398.80 366.2 833.6 1899.9 0.01852 2.04:43 230.0 215.3 393.70 36s.3 831.8 1200.1 0.01855 1.99446 235.0 220.3 395.56 370.3 830.1 1200.4 0.01857 1.95725 240.0 225.3 397.39 372.3 828.4 1200.6 0.0lMho 1.91769 245.0 230.3 399.19 374.2 826.6 1200.9 0.01863 1.87970

Properties of Saturated Steam and Saturated Water-concluded Pressure Temper. Heat of Latent Heat Total Heat Specific Volume Lbs. per Sq. In. - ature the of of Steam Absolute Gage Liquid Evaporation p I t

h' Wa F P o,em, r. nions. siuns. seu ris. co.e. ,t,er

, is.

5 team

. co. ri. c., is.

250.0 2J5.3 400.97 376.1 825.0 1201.1 0.01865 1.84317

~

255.0 240.3 402.72 378.0 823.3 1201.3 0.01868 1.80802 260.0 245.3 404.44 379.9 821.6 1201.5 0.01870 1.77418 265.0 250.3 406.13 381.7 820.0 1201.7 0.01873 1.74157 270.0 255.3 407.80 ~ 383.6 818.3 1201.9 0.01875 1.71013 275.0 260.3 409.45 385.4 816.7 1202.1 0.01878 280.0 1.67978 265.3 411.07 387.1 815.1 1202.3 0.01880 1.65049 '

285.0 270.3 412.67 388.9 813.6 1202.4 0.01882 290.0 1.62218 275.3 414.25 390.6 812.0 1202.6 0.01885 295.0 1.59482 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 0.01889 1.54274 320.0 305.3 423.31 400.5 802.9 1203.4 1

0.01899 1.44801 340.0 325.3 428.99 406.8 797.0 1203.8 360.0 0.01908 1.36405 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 0.01925 1.22177 400.0 385.3 444.60 424.2 780.4 1204.6 420.0 0.01934 1.16095 -

405.3 449.40 429.6 775.2 1204.7 0.01942 440.0 425.3 1.10573 454.03 434.8 770.0 1204.8 0.01950 1.05535 460.0 445.3 458.50 439.8 765.0 1204.8 480.0 0.01959 1.00921 465.3 462.82 - 444.7 760.0 1204.8 0.01967 0.96677 500.0 485.3 467.01 449.5 755.1 1204.7 520.0 0.01975 0.92762 505.3 471.07 454.2 750.4 1204.5 0.01982 0.89137 540.0 525.3 475.01 458.7 745.7 1204A 560.0 0.01990 0.85771 545.3 478.84 463.1 741.0 1204.2 0.01998 0.82637 580.0 565.3 482.57 467.5 736.5 1203.9 , 0.02006 0.79712 600.0 585.3 486.20 471.7 732.0 - 1203.7 620.0 0.02013 0.76975 605.3 489.74 475.8 727.5 1203.4 0.02021 0.74408 640.0 625.3 493.19 479.9 723.1 1203.0 660.0 0.02028 0.71995 645.3 4 %.57 483.9 718.8 1202.7

• 0.02036 0.69724 680.0 665.3 499.86 487.8 714.5 1202.3 0.02043 0.675S1 700.0 685.3 503.08 491.6 710.2 1201.8 720.0 0.02050 0.65556 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 0.02065 760.0 0.61822 745.3 512.34 502.7 C 780.0 800.0 820.0 765.3 785.3 805.3 515.30 518.21 521.06 506.3 509.8 513.3 697.7 693.6 689.6 685.5 1200.4 1199.9 1899.4 1198.8 0.02072 0.02090 0.02087 0.02094 0.60097 0.58457 0.56896 0.55408 840.0 825.3 523.86 516.7 681.5 1198.2 860.0 0.02101 0.53938 845.3 526.60 520.1 677.6 1197.7 0.02109 0.52631 880.0 865.3 529.30 523.4 673.6 1197.0 0.02116 0.51333 900.0 885.3 531.95 526.7 669.7 11 % .4 920.0 0.02123 0.50091 905.3 534.56 530.0 665.8 1195.7 0.02130 0.48901 940.0 925.3 537.13 533.2 661.9 1195.1 960.0 0.02137 0.47759 945.3 539.65 536.3 658.0 1194.4 0.02145 0.46662 980.0 %5.3 542.14 539.5 654.2 1193.7 0.02152 0.45609 1000.0 985.3 544.58 542.6 650.4 1892.9 1050.0 0.02159 0.44590 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 1135.3 561.82 564.8 622.2 1187.0 0.02214 0.38073 1200.0 1185.3 567.19 571.9 613.0 1184.8 0.02232 0.36245 1250.0 1235.3 572.J8 578.8 603.8 1882.6 0.02250 0.34550 1300.0 1285.3 577.42 585.6 594.6 1180.2 0.02269 0.32991 1350.0 1335.3 582.32 592.2 585.6 1177.8 0.02238 0.31536 1400.0 1385.3 587.07 598.8 1450.0 567.5 1175.3 0.02307 0.3017a 1435.3 591.70 605.3 567.6 1172.9 0.02327 0.28909 1500.0 14M5.3 5 % .20 bil.7 558.4 1170.1 0.02346 0.27719 1600.0 1585.3 604.87 624.2 540.3 1864.5 0.023S7 0.25545 1700.0 1685.3 6l3.13 636.5 522.2 1158.6 0.02428 0.23607 1800.0 1785.3 621.02 648.5 503 3 1152.3 0.02472 0.21861 1900.0 1885.3 62N.56 660.4 4M5.2 1145.6 0.02517 0.20278 2000.0 19M5.3 6J5.no 672.1 466.2 2100.0 1138.3 0.02565 0.18331 2045.3 642.76 683.8 446.7 1130.5 2200.0 2185.3 649 45 0.02615 0.17501 695.5 426.7 1122.2 0.02669 2300.0 2285.3 655.89 707.2 0.16272 406.0 1113.2 0.02727 0.15133 l 2400.0 23MS.3 662.11 719.0 384.8 1103.7 0.02790 0.14076 i 2500.0 24M5.3 664.11 7Jf.7 361.6 1093.3 0.02859 U.lJ065 2MNI.0 25M5.3 673.91 744.5 337.6 1032.0 0.02938 0.12110' 2700.0 26MS.3 679.53 757.3 312.3 2800.0 1069.7 0.03029 0.11194 2785.3 6M4.96 770.7 285.1 1035.8 29tN1.0 2 MMS.3 690.22 0.03134 0.10305 7AS.I 254.7 1039.8 0.03262 0.09420 l Jutm.u 24M5.3 695.33 808.8 3100.0 2 8 M.4 1020.3 0.03428 0.us500 30MS.3 700.28 824.0 169.3 993.3 J200.0 345.3 705.n4 0.03651 0.07452 M75.5 56.1 931.6 320M.2 3191.5 705.47 0.04472 0.056n3 ho 0.0 906 0 0.05078 0.05078 i

i

l Properties of Superheated Steam i

V= specific volvme, cubic feet per pou.J h,-i.e.i h.., of .. seu p.r p d Pressure Sat.

Lbe. per Temp.

Sq.In. Total Temperature-Degrees Fahrenheit (t)

Abs. Gage 350* 400' 500* 600* 700* 800* 900' P' P r 100F 1100* 1300* 1500*

15.0 0.3 213.03 V 31.939 33.963 37.985 41.9M 45.978 49.964 53.946 57.926 61.905 69.858 77.807 A, 1216.2 1239.9 1247.3 1335.2 1383.8 1433.2 1483.4 l 1534.5 1586.5 1693.2 1803.4 20.0 5.3 217.% V 23.900 25.428 "28.457 31.466 34.#5 37.458 40.447 43.435 #.420 52.388 58.352 A, 1215.4 1239.2 1286.9 1334.9 1383.5 1432.9 1483.2 1534.3 15M.3 1693.1 1803.3 30.0 15.3 250.34 7 15.859

{

16.892 18.929 10.945 22.951 24.952 26.949 A, 1213.6 28.943 30.936 34.918 38.8 %

1237.8 1286.0 1334.2 1383.0 1432.5 1482.8 1534.0 1586.1 1692.9 1803.2 40.0 25.3 267.25 7 11.838 12.624 14.165 15.685 17.195 18.699 20.199 21 697 A, 1211.7 1236.4 23.194 26.183 29.168 1285.0 1333.6 1382.5 1432.1 1482.5 15M.7 1585.8 1692.7 1803.0 50.0 35.3 281.02 P 9.424 10.062 11.3M 12.529 13.741 14.947 16.150 A, 1209.9 17.350 18.549 20.942 23.332 1234.9 1284.1 1332.9 1382.0 1431.7 1482.2 1533.4 ! 1585.6 1692.5 1802.9 60.0 45.3 292.71 F 7.815 8.354 9.400 10.425 11.438 12.4 % 13.450 A, 1208.0 14.452 15.452 17.448 19.441 1233.5 1283.2 1332.3 1381.5 1431.3 1441.8 1533.2 1585.3 1692.4 1802.8 70.0 55.3 302.93 7 6.664 7.133 8.039 8.922 9.793 10.659 11.522 12.382 A, 1206.0 1232.0 13.240 14.952 16.661

] 1282.2 1331.6 1381.0 1430.9 1481.5 1532.9

• 1585.1 1692.2 1802.6 80.0 65.3 312.04 7 5.801 6.218 7.018 7.794 8.560 9.319 10.075 A, 10.829 11.581 13.081 14.577 1204.0 1230.5 1181.3 1310.9 13803 1(30.5 1481.1 1532.6 1584.9 1692.0 1807.5 90.0 75.3 320.28 E 5.128 5.505 6.223 6.917 7.600 A, 8.277. 8.950 9.621 10.290 11.625 12.956 1202.0 1228.9 1280.3 1330.2 1380.0 1430.1i 1480.8 1532.3 1584.6 1691.8 1802.4 100.0 85.3 327.82 F 4.590 4.935 5.568 6.216 6.833 7.443 8.050 A, 1199.9 8.655 9.258 10. # 0 11.659 1227.4 1279.3 1329.6 1379.5 1429.7 1480.4 1532.0 1584.4 1691.6 1802.2 120.0 105.3 341.27 V 3.7815 4.0786 4.6341 5.1637 5.6813 6.1928 6.7006 7.2060 7.7096 8.7130 9.7130 A, 1195.6 1224.1 1277.4 1328.2 1378.4 1428.8 1479.8 1531.4 1583.9 1691.3 1802.0 140.0 125.3 353.04 y 3.4661 3.9526 4.4119 4.8588 5.2995 5.7364 A,

1220.8 1275.3 6.1709 6.6036 7.452 8.3233 1326.8 1377.4 1428.0 1479.1 1530.8 1583.4 1690.9 1801.7 160.0 145.3 363.55 V 3.0060 3.4413 3.8480 4.2420 4.6295 5.0132 5.3945 5.7741 6.5293 A, 1217.4 1273.3 7.2811 1325.4 1376.4 1427.2 1478.4 1530.3 15f2.9 1690.5 1801.4 180.0 165.3 373.08 7 2.6474 3.0433 3.4093 3.7621 4.1084 4.4508 4.7907 5.1289 5.8014 6.4704 A, 1213.8 1271.2 1324.0 1375.3 1426.3 1477.7 1529.7 1582.4 1690.2 1801.2 200.0 185.3 381.80 V 2.3598 2.7247 3.0583 3.3783 3.6915 4.0006 4.3077 4.6128 5.2191 A, 1210.1 1269.0 5.8219 1322.6 1374.3 1425.5 1477.0 1529.1 1581.9 1689.8 1800.9 220.0 205.3 389.88 7 2.1240 2. # 38 2.771C 3.0642 3.3504 3.6327 3.9125 4.1905 4.7426 5.2913 A, 1206.3 12 % .9 1321.2 1373.2 1424.7 1476.3 1528.5 1541.4 1689.4 1800.6 240.0 225.3 397.39 P 1.9268 2.2462 2.5316 2.8024 3.0661 3.3259 3.5831 3.8385 A, 1102.4 4.3456 4.8492 1264.6 1319.7 1372.1 1423.8 1475.6 1527.9 1580.9 1689.1 1800.4 2'0.0 6 245.3 404.44 7 2.0619 2.3289 2.5808 2.8256 3.0M3 3.3044 3.5408 4.0097 4.4750 A, 1262.4 1318.2 1371.1 1423.0 1474.9 1527.3 1580.4 1688.7 1800.1 280.0 265.3 411.07 V I.9037 2.1551 2.3909 2.6194 2.4437 3.0655 3.2855 A, 3.7217 4.1543 1260.0 1316.8 1370.0 1422.1 1474.2 1526.8 1579.9 1688.4 1799.8 300.0 285.3 417.35 V 1.7%5 2.0044 2.2263 2.4407 2.6509 2.8585 3.0643 A, 3.4721 3.8764 1257J 1315.2 1368.9 1421.3 1473 4 1526.2 1579.4 1M8.0 1799.6 320.0 305.3 413.31 F 1.642 1.8725 2.0823 2.2843 2.4821 2.6774 2.8700 3.2534 3.6332 A, 1255.2 1313.7 1367.8 1420.5 1472.9 1525.6 1578.9 1687.6 1799.3 340.0 325.3 428.99 7 1.5399 1.7561 1.9552 2.1%3 2.3M3 2.5175 2.7000 3.0611 3.4186 A, 1252.8 1312.2 1366.7 1419.6 1472.2 1525.0 1578.4 1687.3 1799.0 ,

360.0 345.3 434.41 F l.4454 1.6525 1.8421 2.0237 2.2009 2.3755 2.5482 2.8898 A, 3.2279 1150.3 1310.6 1365.6 1418.7 1471.5 1542.4 1577.9 1686.9 1798.8

l Properties of ivperheated Steam continued V-,.cac e.sv.., cu6ic t ,

A,-woi h , or n..., seu p.,p.cppound ond Premure Set.

Lbs. per Temp.

Sq. In.

Total Temperature-Desrees Fahrenheit (4

• N 500*

P' P " 400* 700

• 900* 900* 1000* 1100* 1200' ' 1500'

• 300.0 365.3 439.61 y 1.3606 ' t.5598 1300'!1400'!l A, 1247.7 1309.0 ' 1364.5 1J4101417.91.9139 2.0825 2.2484 1470.8 1523.8 2.4124' 2.5750! 2J366 2.8973 ' 3.0572 400.0 385.3 444.60 y 1.2841 1.4763 1577.4 1631.6 1686.5 1742.2

• 1798.5 1245.1 1.6499 1.8151 1.9759 2.1339 2.2901 A, 1307.4 1363.4 1417.0 1470.1 1513.3 1576.9 2.4450 2.5987 2.7515 2.9037 420.0 405.3 449.40 V 1.2148 1.4007 1.%76 IJ258 1633.2j 1646.2.1741.9 1798.2 A, 1242.4 1.8795 2.0304 2.1795 1305.8 1362.3 1416.2 I#9.4 1822.7 1576.4 2.3273I 2.4739 2.6!% 2.7647 440.0 425.3 454.03 1630.8 1685.8 1741.6 1798.0 7 1.1517 A, 1239J 1.3319 1.4926 1.6445 1.7918 1.9363 2.0790 2.2203 1304.2 1361.1 1415.3 !#8J 1512.1 1575.9 2.3605 ' 2.4998 2.6384 460.0 445.3 458.50 1630.4 F 1.0939 1.2691 1.4242 1.5703 IJ117 1.8504 1685.5 1741.2 1797.7 A,

1234.9 1302.5 1360.0 1434.4 I#8.0 1521.5 1.9872 2.1226 2.2569 2.3903 2.5230 1875.4 1629.9 480.0 #5.3 # 2.82 [ 1685.1 1740.9 1797.4 A, 1.0409 1.2115 1.3615 1.5023 1.6384 IJ716 1.9030 1234.1 1300.8 1358.8 1413.6 IM7.3 1520.9 1574.9 2.0330 2.1619 ' 2.2900 2.4173 500.0 485.3 1629.5 1 % 4.7, 1740.6 1797.2 467.01 P 0.9919 1.1584 1.3037 1.4397 1.5708 1.6992'I.8256 A, 1.9507 1231.2 1299.1 1357.7 1412.7 14 % .6 1520.3 1574.4 2.0746 2.1977 2.3200 520.0 505.3 1629.1 471.07 F 0.9466 1.1044 1.2504 1.3819 1.5085 1.6323 IJ542 1684.4 1740.3 17 %.9 A,

1228.3 1297.4 1356.5 1431.8 1465.9 1819.7 1573.9 1.87 # 1.9940 2.1125 2.2302 540.0 16281 ' 1684.0, 1740.0 17%J 525.3 (75.01 F 0.9045 1.0640 1.2010 1.3284 1.4508 l A,

1225.3 1295J 1355.3 1410.9 1#5.1 1.5704 1.6800 1.8042 1.9193 2.0335 2.1471 M0.0 545.3 1819.1 1573.4 M28.2 1683.6 1739.7 , 17 %.4 478.84 P 0.8653 1.0217 1.1552 A, 1.2787 1.3972 1222.2 1293.9 1354.2 1.5129 1.62 % 1.7388 1410.0 1#4.4 1518.6 1572.9 1627.8 1.8500 1.9603 2.0699 580.0 565.3 402.57 F 1683.3 , 1739.4 17 %.1 0.8287 0.9824 1.1125 I A, 1219.1 1.2324 1.3473 1.4593 1.M 93 I.6780 1292.1 1353.0 1409.2 1#3.7 1518.0 1572.4 . 1627.4 1.7855 : 1.8921 1.9980 400.0 585.3 486.20 V 1682.9g 1739.1 1795.9 01944 0.9456 1.0726 A, 1215.9 1290.3 1351.8 1.1892 1.3008 1.4093 1.5160 1.62tI I 1408.3 ' 1 # 3.0 1.9309 680.0 635.3 494.09 F 0.7173 1517.4 1571.9 1627.0 1.J252 02.6, 1738.8 ' l.8284 1795.6 0.8634 0.9835 A, 1207.6 1285.7 1.0929 1.1969 1.2979 1.3969 1.4944 1.5909 3.6864 1.7813 1348.7 1406.0 1#1.2 700.0 1515.9 1570J 1625.9 1681.6 685.3 503.08 F 17J8.0 1794.9 0.7928 0.9072 A, 1281.0 1.0102 1.1078 1.2023 1.2948 1.3858 1.4757 ! 1.5647 1.6530 1345.6 1403.7 1459.4 1514.4 750.0 1M9.4 1624.8 1%0.7 1737.2 735.3 510.84 V 0.7313 0.8409 0.93M 1794.3 A, 1.0306 1.1195 1.2063

.., 1276.1 1342.5 1401.5 1.2916 1457.6 1512.9 1%8.2 1.3759 ~ 1.4592 1.5419

- 800.0 1623.8 785.3 518.21 F 0.6774 0.7828 0.8759 1679.8 . 1736.4 1793.6 A, 0.9631 1.0470 1.1289 1271.1 1339.3 1399.1 1.2093 1.2885 ' 1.3 % 9 1.4446 1455.8 1511.4 15 %.9 850.0 1622.7 835.3 525.24 V C.62% 1678.9 . 1735.7 1792.9 A, 0.7315 0.8205 0.9034 0.9830 1265.9 1336.0 13 % .8 1454.0 1810.0 1.0606 1.13 % . 1.2115 1.2855 1.3588 900.0 1h5.7 885.3 531.95 F 0.5M9 0.6858 0 7713 1621.6 1678.0 1734.9 1792.3 A, 1260.6 1332.7 1394.4 0.8504 0.9262 0.9998 1.0720 1.1430 1.2131 ' t.2825 980.0 1452.2 1508.5 1564.4 935.3 538.39 F 1620.6 1677.1 ' 1734.1 ;1791.6 A, 0.5485 0.6449 0J272 0.0030 0.8753 0.9455 t 1255.1 1329.3 1392.0 1450.3 1507.0 1h3.2 1.0142 1.0817 ! 1.1484 'l.2143 3000.0 985.3 544.58 C 1619.5 1676.2 1733.3 1791.0 A, 0.5137 0.6080 0.6875 0.7603' O.8295 0.89 %

1249.3 1325.9 1389.6 1448.5 1505.4 0. % 22 1.02 % 1.0901 .1.1529 1MI.9 1618.4 1675.3 1732.5 1950.0 1035J 556.33 V .

1790.3 -

A, 0.4821 0.5745 0.6515 0.7216 0J881 1243.4 1322.4 1387.2 1446.6 1503.9 0.8524 0.9151 0.9767 1.0373 1.0973 15401 1617.4 1674.4 17J1.8 1100.0 1985.3 556.28 7 1789.6 'l A, 0.4531 0.5440 0.6188 1237.3 1318.8 1384.7 0.6865 0.7505 0.8121 0.8723 0.9313 0.9894 1.0#8 1444.7 1502.4 1559.4 M16.3 1673.5 1731.0 1789.0 1180.8 1135.3 561.82 7 ...

A, . 0.4M3 0.5162 0.5889 0.6M4 0J161 0.7754 0.8332 0.8899 0.9456 1.0007 1230.9 1315.2 1382.2 1442.8 1500.9 1558.1 MIS.2 1672.6 1730.2 1788.3 g

Properties of Supdrheated steam concluded V-s,.a volu , cubic r.. p.c pound h,= total hoot of steam,8tv per pound -

Pressure Set .

Temp.

Lbs.

sq. fn. Total Tempersture-Degrees Fahrenheic 0)

Abs. Gase P' 65r 700

• 75r 800 P r 900* 1000*  !!00* 1200

• 1300*

_ 1400 1509 1200.0 !!85.3 567.19 E 0.4497 0.4905 0.5273 0.5615 0.6250 0.6845 01418 0.7974 A,

1300.0 1285.3 577.42 1271.8 1311.5 13%.S 13791 I440.9 1499.4 1556.9 1614.2 1671.

F 0.4052 A,

0.4451 0.4804 0.5129 0.5729 0.6287 0.6822 0.7341 0

!!61.9 1303.9 1340.8 1374.6 1437.1 14%.3 1554.3 1612.0 1%9.8 1 1400.0 1385.3 587.07 7 A, 0.3667 0.#59 0.4400 0.4712 0.5282 0.5809 0.6311 0.6798 0.7272 0

!$00.0 1485.3 5 % .20 7 1251.4 12%.! 1334.5 1369.3 1433.2 1493.2 1551.8 1609.9 I%8.0 1 A, 0.3328 0.3717 0.4049 0.4350 0.4894 0.5394 0.5%9 0.6327 0.6773 1600.0 1585.3 604.87 C 1240.2 1287.9 1328.0 1364.0 1429.2 1490.1 1549.2 1M7J 1%6.2 A, 0.3026 1228.3 1279.4 0.3415 1321.4 0.3741 1358.50.4032 1425.2 1486.9 0.4555 15 % 0.5031

.6 0.5482 0.5915 0.6336 1700.0 1685.3 613.13 C 1605.6 1%4.3 1723.2 1782.3 A,

0.2754 0.3147 0.3468 0.3751 0.4255 0.4711 0.5140 1215.3 1270.5 1314.5 1352.9 1421.2 1483.8 1544.0 0.5552 0.5951 0.6341 0.6724 1800.0 1785.3 621.02 T 0.2505 0.2906 0.3223 0.3500 1603.4 1%2.5 1721.7 1781.0 A, 1201.2 1261.1 0.3988 0.4426 0.4836 1307.4 1347.2 1417.1 1480.6 1541.4 0.5229 0.5609 0.5980 0.6343 1601.2 1%0.7 1720.1 1779.7 1900.0 1885.3 628.56 i 0.2274 0.2687 0.3004 0.3275 0.3749 A,  !!85J I251.3 0.4171 0.4565 0.4940 0.5303 1300.2 1341.4 1412.9 0.5656 0.6002 1477.4 1535.8 1599.1 1658.8 1718.6 2000.0 1985.3 635.80 F 1778.4 0.2056 0.2488 0.2805 0.3072 0.3534 0.3942 0.4320 A, II68.3 1240.9 1292.6 1335.4 0.4680 0.5027 0.5365 0.5695 1408.7 1474.1 1536.2 15 % .9 2!00.0 1657.0 1717.0 1777.1 2085.3 642.76 C 0.1847 0.2304 0.2624 0.2888 A, 1148.5 0.3339 0.3734 0.4099 0.4445 0.4778 1229.8 1284.9 1329.J 0.5101 0.5418 1404.4 1470.9 1533.6 1594J 1655.2 1715.4 1775.7 2200.0 2185.3 649.45 V 0.1636 0.2134 0.2458 0.2720 A, 1123.9 0.3161 0.3545 0.3897 0.4231 0.4551 1218.0 1276.8 1323.1 1400.0 C.4862 0.5165 1467.6 1530.9 1592.5 1653.3 1713.9 2300.0 2285.3 655.89 F 1774.4 0.1975 0.2305 0.25% 0.2999 0.3372 Ag 1205.3 1268.4 0.3714 0.4035 0.4344 0.4643 1316.7 1395.7 l#4.2 1528.3 0.4935 1590.3 1651.5 1712.3 2400.0 2385.3 M2.11 7 0.1824 0.2164 0.2424 177J.1 A, 0.2850 0.3214 0.3545 0.3856 1191.6 1259.7 1310.1 1391.2 0.4155 0.4443 0.4724 1460.9 1525.6 1588.1 1649.6 1710.8 2300.0 2485.3 M8.11 0 0.1681 0.2032 0.2293 1771.8 A, 0.2712 0.3068 0.3390 0.3692 1176 1 1250.6 1303.4 1386.7 0.3980 0.4259 0.4529 1457.5 1522.9 1585.9 1647.8 2600.0 2585.3 F 1709.2 1770.4 673.91 0.1544 0.1909 0.2171 0.2585 0.2933 0.3247 A, 1160.2 0.3540 0.3819 0.4088 1241.1 12 %.5 1382.1 1454.1 0.4350 1520.2 1583.7 16 % .0 1707.7 2700.0 2685.3 679.53 I 1769.1

%1411 0.1794 0.2058 0.2468 0.2809 A, 1142.0 0.3114 0.3399 0.3670 0.3931 1231.1 1289.5 1377.5 1450J 0.4184 1517.5 1581.5 1644.1 1706.1 2800.0 2785.3 664.% U 1767.8 0.1278 0.1685 0.1952 0.2355 A, 1121.2 c.2693 0.2991 0.3268 0.3532 1220.6 1282.2 1372.8 1447.2 0.3785 0.4030 1514.8 1579.3 1642.2 1704.5 2900.0 2885.3 690.22 I 0.1138 0.1581 0.1853 17 % .5 A, 0.2256 0.2585 0.2877 0.3147 0.3403 0.3649 0.3887 1995.3 1209.6 1274.7 1368.0 1443.7 1512.1 1577.0 1640.4 170J.0 1765.2 3000.0 2985.3 695.33 E 0.0982 0.1483 A, 1060.5  !!97.9 0.1759 0.2161 1267.0 1363.2 1440.20.2484 1509.40.2770 1574.8 0.3033 th38.5 0.3282 0.3522 0.3753 J100.0 3085.3 700.28 9 1701.4 1763.8 0.1389 A, i185.4 0.1671 0.2071 0.2390 0.2670 0.2927 0.3170 0.3403 0.3628 3200.0 3185.3 705.08 9 1259.1 1358.4 1436.7 1506.6 1572.6 1636J 1699.8 1762.5 A, ~

0.1300 1172.3 0.1588 0.1987 0.2301 0.2576 0.2827 0.3065 0.3291 0.351 l 3300.0 3285.3 . P 1250.9 1353.4 1433.1 1503.8 1570.3 1634.8 1698.3 1761.2 A,

0.1213 0.1510 0.1908 0.2218 0.2488 0.2734 0.2966 0.J187 0.340 3400.0 3385.3 . 7 s..

1158.2 1242.5 1348.4 1429.5 1541.0 1568.1 1623.9 16%J 1759.9 A, ...

0.1129 0.1435 0.1834 0.2140 0.2#5 0.26% 0.2872 0.3088 0.32%

l 1I43.2 1233J 1343.4 1425.9 1498.3 1565.8 163I.1 1695.1 1758.5

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SECTION 5 THEORY OF NUCLEAR POWER PLANT OPERATION.

FLUIDS, AND THERMODYNAMICS .

5.01 QUESTION (3.0)

The plant is in hot standby with one of the steam generators at 805 PSIG. Use the attached steam tables to answer the following questions.

(a) What is the steam temperature in the steam generator ? (1.0)

(b) If an atmospheric relief valve is opened what will be the temperature of the discharged steam ? (assume discharge pressure i s atmospheric) (1.0) 4 (c) Will the discharned steam be superheated, saturated, or a mixture of saturated steam and water ? Briefly explain how you reached your conclusion. (1.0) 5.01 ANSWER (1.0)

(a) 521 F

(*lo") (1.0)

(b) 315 3 F (c) superheated. Steam temperature is above saturation temperature (212) at atmospheric pressure. (1.0)

REFERENCE:

Thermal-Hydraulic Principles and Applications to the PWR, Part II, Chapter 7 9

I c - - - - - - -, , , , . , , , - - - , - - - - -r <

5.02 OUESTION (3.0)

Refer to Figure 5.1 which shows an instantaneous, positive -

reactivity insertion into an already critical reactor core at time T = 0, followed by a removal of this positive reactivity after a stable reactor period is reached at time T = 10 minutes. During this ten minute time period power increases from 10-8 amps to 10-5 amps. (Show your work and state assumptions needed for any calculations.)

(a) On Figure 5.1 sketch the resulting reactor startup rate as a function of time for these reactivity changes. (numerical values are not required) (1.0)

(b) On Figure 5.1 sketch the reactor power level as a function of time for these reactivity changes. (1.0)

(c) Based on the stated power increase and time, what was the initial positive reactivity insertion in percent millicho (pcm) 7 (1.0) 5.02 ANSWER CAMbl0VF MM STATE (a) & (b) See attached drawing.  : 0 007 Sol (t) (10) g , 0.cos got (c). P=Pon 10**(SUR x t) 10-5 = 10-8 n 10**SUR x 10 SUR = 0.3 DPM T = 26/SUR = 86 sec-1 T= (b - p)/(ap) 86 = (0.0072 - p)/(0.1 x p) 8.6 x p = 0.0072 - p 9.6 x p = 0.0072 p = 0.00075 75 pcm 73 pun p .007 (1.0)

SJ fYM p = .007

REFERENCE:

Fundamentals of Nuclear Reactor Physics, Chapter 7 l

Figure 5.1 i.

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5.~O3 QUESTION (2.0)

The Unit 2 reactor plant is operatina normally with the plant parameters indicated below.

Plant power- 50 %

Total RCS flow 130 million pounds per hour RCP PSID 90 RCP supply voltage 11.5 KV RCP motor power- 4400 KW Grid frequency 60 H' ,

Due to an off-site problem the grid frequency decreases to 55 Hz and the RCPs continue to operate.(Show your work and state assumptions needed for any calculations.)

-(a) After the frequency decrease, what is the Total RCS flow ?

(in million pounds per hour) (0.5)

(b) After the frequency decrease, what is the RCP PSID ? (0.5)

(c) Assuming constant motor efficiency, what is the RCP motor power at the lower frequency 7 (in KW) (1.0) 5.03 ANSWER (a) .55/60 x 130 = 119 million pounds per hour (flow is proportional to pump speed) (0.5)

(b) (55/60)**2 x 90 = 76 PSID (pressure drop is proportional to speed squared)

.(0. 5)

(c) 3390 Kw (Power is proportional to speed to the third power (55/60)**3 x 4400 = 3390) (1.0)

REFERENCE:

T-H Principles and Applications-to the PWR, Vol II, Chapter 10

FIGURE 5.3 m

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CRITICAL BORON CONCENTRATION CURVE I

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4 5.05 OUESTION (1.0)

Refer to Figure 5.3. Early in core life, why does critical. baron concentration drop quickly f rom 1200 to 850 ppm ? (1.0) 5.05 ANSWER

~

Critical baron concentration drops quickly early in core life due

- to the build-up ;f c: rri r. fi5stoo P"doct Fosseus (Aemeueruv (1.0)

X6900 AWD SAMAAIVM)

REFERENCE:

Reactor Core Control for Large PWRs, page 3-26 I

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5.04 OUESTION -(2.0)

To answer questions (a) and (b) below, refer to Figure 5.2 which depicts four xenon transients.

(a) Which curve best depicts the xenon transient you would expect starting from equilibrum conditions at 50 % power followed by a rapid increase to 100 % power ? (0.5)

(b) Which curve best depicts the xenon transient starting from equilibrum conditions at 100 % power followed by a rapid decrease to 50 % power 7 (0.5)

(c) Xenon instability is a slow oscillation in power distribution due to changing xenon concentrations. What i s the principal problem resulting from xenon instability ? (1.0) 5.04 ANSWER

! (a) B (0.5)

(b) C (0.5) 1 (c) Local increase in neutron flux causes greater local heat generation than expected or provided for in accident analyses.

(1.0) i

REFERENCE:

Reactor Core Control for Large PWRs, 4-20 through 4-29 f r i

e 7 - a e-- - -e- --g--- - , - - - - - - ,

5.06 OUESTION (3.0) .

Following a 60 day run at 100 % power _the reactor has been in hot-shutdown for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Using STP R-17 an ECP of 850 ppm baron has been calculated for 100 steps on control bank D and 552 F.

Assume the following reactivity coefficients, defects, and values. (Show your work and state assumptions needed for any calculations.)

Moderator temperature coefficient - 15 pcm/F Differential baron worth - 10 pcm/pom Average rod worth 6 pcm/ step Initial source range count rate (with all rods in) 20 CPM (a) After the ECP is computed you find that RCS temperature is actually 545 F and reactor engineering tells you that the ECC should have used a critical rod height of 120 steps. What is the new critical baron concentration 7 (1.0)

(b) While conducting the start-up you observe that the count rate has increased to 50 CPM when all the shutdown banks have been with-drawn. Based on this observation, what is the approximate reactivity worth of the shutdown banks ? (The reactor engineer informed you that with all rods in, 545 F and borated to be critical at 120 steps on control bank D Keff should be 0.931) (1.0)

(c) If the start-up is delayed one day how will samarium affect the ECC ? (boron higher, same or lower) (0.5)

(d) If the start-up is delayed one day how will xenon affect the ECC 7 (boron higher, same or lower) (0.5) 5.06 ANSWER (a) (545 - 552) x (-15) + (120 - 100) x 6= 105 + 120 = 225 pcm 255/10 pcm/ ppm = 23 ppm additional baron (873 ppm) (1.0)

(b) CR1(1 - Keff1) = CR2(1 - Keff2) 2O(0.069) = 50(X): 1.38/50 = 0.0276 (negative reactivity with shutdown banks out): p= (K - 1)/K = 0.0276/O.9724 =

0.02838; 6900 - 2838 = 4062 pcm .

Alternate for 75 % credit: CR1 : p1 = CR2 x p2; so 50 X =

20 x 6900, X = 2760 pcm, 6900 - 2760 = 4140 pcm (1.0)

(c) Lowers baron required (samarium build-up adds negative '

reactivity) (0.5)

(d) Same boron required (xenon han decayed away) (0.5)

REFERENCE:

Reactor Core Control for Large PWRs, Chapter 9 and Operator Information Manual R-4-1 through R-6-1

'.t

5.07 QUESTION (3.0) .

Emergency Diesel Generator 1-1 is being paralleled with bus H per Operating Procedure OP J-6b to load test the diesel.

(a) Why is the diesel speed adjusted to obtain the synchroscope rotating slowly in the fast direction just before shutting the generator breaker ? (1.0)

(b) How are the following electrical meter indications affected if the diesel speed control switch is momentairly switched to " RAISE" while the diesel generator is paralleled with off-site power '

(increase, decrease, stay the same)

Generator KW (0.5)

Generator VARs-(assume initial 100 KVAR lag) (0.5)

(c) How are the following electrical meter indications affected if the diesel voltage adjuster control switch is momentairly switched to " RAISE" while the diesel generator is paralleled with off-site power ? (increase, decrease, stay the same)

Generator KW (0.5)

Generator VARs (assume initial 100 KVAR lag) (0.5) 5.07 ANSWER (a) To prevent reverse current trip and/or to cause the diesel generatcr to pick up some load when it is paralleled. (1.0)

(b) Generator KW increases (picks up real load) (0.5)

Generator VARs stays the same(reactive load unchanged) (0.5)

(c) Generator KW stays the same (real load" unaffected) (0.5)

Generator VARs increase (reactive load increased) (0.5)

REFERENCE:

OP J-6B:IV, DG 1-1 Manual Operations, and Lesson Plan

! J-6B f

a E

p y --. - e ,-m .n---- ,.y.-, - - - , ,,-~,e,, .m. , , , , . - , - - . - - - . - - . - - u o

5.08 QUESTION (2.0)

For each of the questions below select the best answer.

(a) what reactivity addition is required to double the count rate if Keff is 0.98 7 (1.0)

1. 1000 pcm

2. 5000 pcm

3. 100 pcm

4. 500 pcm (b) If a reactor core with a neutron source is exactly critical at 1000 CPS in the source range, over'the next few minutes the count rate should: (1.0)

1. Increase exponentially with time

2. Increase linearly with time

3. Remain constant at 1000 CPS

4. Increase geometrically with time 5.08 ANSWER (a) 1 (1000 pcm) (1.0)

(1.0)

(b) SL (linear increase)

REFERENCE:

Fundamentals of nucler Reactor Physics, Chapter 8

5.09 OUESTION (2.0)

A heat balance by manual calculation has been completed in -

accordance with STP R-2B, Operator Heat Balance.

(a) If blowdown flow is omitted from the calculations, briefly e:tplain how calculated thermal power is affected ? (1.0)

-(b) If erosion has enlarged the feedwater venturis, briefly explain how calculated power is affected ? (1.0) 5.09 ANSWER (a) Calculated power is higher .than actual due to less energy removed by blowdown or more energy removed by steam. (1.0)

(b) Calculated power is lower than actual because feedwater flow calculated is lower-due to decreased delta pressure across venturis. (1.0)

REFERENCE:

STP R-2B 5

9 I

h. _

P l

5.10 DUESTION (2.0)

While operating at 100 % power, T(cold) is 545 F, T(hot) i s -610 F ,

and RCS flow rate is 134 million pounds per hour.

Following a unit trip and loss of all off-site power, a natural circulation couldown is established with T(cold) of 544 F and core delta T of 60 F. It is estimated that decay heat is 1.6 % of full power.

Assume Cp of water between 550 and 650 F is 1.3 Btu /Lbm-F, condensate storage.tenk is at 100 F, and use the attached steam tables, if appropriate, to answer the following questions. (Show your work and state assumptions needed f or any calculations. )

(a) What is the natural circulation flow rate in Lbm/Hr ? (1.0)

(b) What is the auxiliary feedwater flow rate (Lbm/Hr) to the steam generators required to maintain a constant RCS temperature ?

(1.0) 5.10 ANSWER (a) O = M Cp (delta T); 134 MM/Hr :: 1.3 x (610 - 545) = 11.32+9 Btu /hr; 1.6 % of 11.32+9 = 1.81+8 Btu /Hr; 1.81+8 = M x 1. 3 ::

60 so M = 2.3 Million pounds per hour (1.0)

(b) O=M (delta h); 1.81+8 = M (1192 - 68) so M = 1.61+5 Lbm/Hr (partial credit given for correct calculation with heat rate from part (a)) (1.0)

REFERENCE:

T-H Principles and Applications to the PWR, Chapter 2, 7, 12, and 14 1

1

T 5.11 QUESTION (1.0)

During normal operation what are the three principal chemicals added to the RCS ? (1.0) 5.11 ANSWER (0.33 f or each of the following)

Hydrogen, boric acid, and lithium hydroxide.

REFERENCE:

Radiation, Chemistry, and Corrosion Considerations for NPP Application, 7-15 t< 16

5.12 QUESTION (1.0)

Which of the following actions in most likely to cause water.

hammer 7 (a) Starting a centrifugal pump with the discharge valve shut (b) Feeding a steam generator with water colder than saturation (c) Placing vacuum breakers on the high points of water systems (d) Draining a steam generator when it is above 212 F 5.12 ANSWER (B)

REFERENCE:

CAF END OF SECTION 5 9

SECTION 6 PLANT SYSTEMS, DESIGN, CONTROL AND INSTRUMENTATION .

6.01 QUESTION (3.0)

Regarding the auxiliary feedwater system.

(a) How are the motor driven auxiliary feedwater pumps protected from "run-out" conditions (such as a feedwater line break) ?

(1.0)

(b) List five of the six signals which will start the motor driven auxiliary feedwater pumps. (1.0)

(c) If maintenance personnel switch the hot shutdown pannel auxi-liary feedpump 1-2 transfer switch from the " Control Room" to the

" Local" position, what two indications occur in the control room ?

(1.0) 6.01 ANSWER 4

> (a) Pump discharge low pressure signals are used as signals to close the appropriate steam generator level control val ve (s) (in automatic, no run-out protection in manual or bypass). (1.0)

(b) (0.2 each for any five of the following)

Control room start HSDP start ,

SI Signal i S/G low-low level (15% on 2/3 channels in 1/4 S/Gs)

Trip of both MFPs 4KV bus transfer to the D/G 4

(c) " Hot Shutdown Panel" annunciator (on PKO9) (0.5)

All lights but the white status light above the 1-2 pump control switch will go out (red, green, and blue lights out;

! with the pump not running and not tripped the green light on VB-3 will go out) (0.5)

REFERENCE:

Lesson Plan D-1, Auxiliary Feedwater System, pages 14 -

21

~

i l

6.02 QUESTION (2.0)

In addition to narrow range hot and cold leg temperature instru-mentation the RCS is monitored with wide range hot leg and cold leg temperature instruments.

(a) List the three control and/or monitoring devices which receive .

signals from the wide range hot leg temperature instruments.

(1.0)

(b) List the three control and/or monitoring devices which receive signals f rom the wide range cold leg temperature instruments.

(1.0) 6.02 ANSWER (a) (0.33 ecch) -

Control board recorder SPDS RVLIG P-pso Subcooled Margin Monitor ,D EDKw'Isb S*>TDOW U PWNEL, T 1 (b) (0.33 cach)

Control board recorder Water Solid Alarms System Low temperature Overpressure Protection System V

REFERENCE:

Lesson Plan, A-2c, RCS Temperature Instrumentation, page 17; 'Pir.D ioaos5, % co, hi p, 4

6.03 OUESTION (1.0)

One of the intermediate range detectors has had its' compensating voltage set too high (Over-compensated). Which of the following statements is most correct ?

(a) On a normal reactor shutdown the source range detectors may be reinstated too early resulting in a reactor trip.

(b) During a start-up a. false high start-up level will be indicated by the affected channel.

(c) During a shutdown the soure range detectors will not automatically be reinstated if the affected channel stays above 5 x 10-11 Amps.

(d) During a shutdown if the affected channel stays above 10-10 Amps the source range detectors can be blocked.

6.03 ANSWER (h) is correct

REFERENCE:

Lesson Plan B-4, Excore NI System, page 33 i

l I

i t

I t

r-6.04 OUESTION (3.0)

Haaen controllers are used to send demand signals to valves and other equipment which regulate their position or function. According to OP O-2 " Operation of Hagen Controllers" there are two types of power to these controllers, " automatic" and " manual" power.

(a) If a Hagen controller is operating in AUTO and automatic power is lost momentarily, how will the controller respond 7 (0.5)

(b) If a Hagen controller is operating in AUTO and manual power is lost momentarily, what is the immediata response of the controller ?

(0.5)

(c) If a Hagen controller is operating in AUTO and manual power is lost momentarily, what is the final response of the controller once manual power is restored 7 (0.5)

(d) What two controller indications would you expect if both automatic and manual power are lost to a Hagen ccatroller ? (1.0)

(e) Do Hagen controllers show actual valve or device position 7 (0.5) 6.04 ANSWER (a) The controller will go to MANUAL (and stay there). (0.5)

(b) The controller will go to AUTO-HOLD. (0.5)

(c) The controller will go to MANUAL. (0.5)

(d) AUTO demand meter goes to zero and (0.5)

• the lights on the controller go out. (0.5)

(e) No (they show demand, with the exception of the Au:.iliary Feedwater Level Control System). (0.5)

REFERENCE:

OP O-2, Operation of Hagen Controllers e

e-,, - - . - - - ,

- - . . - - , , - - - - , , - - , - - - - - - - - - ,yc- - , -,,--,,

6.05 QUESTION (2.0) -

Regarding the Fuel Handling Building Ventilation System in " Auto":

'(a) Not counting manual-selection from the control room or FHB, list the two ccnditions or-signals which will shift the FHBV System from

" Normal" to " Iodine Removal" mode. (1.0)

(b) When starting supply f an S-1 f rom Ventilation Control Panel #1 in the cable spreading room what two verifications must an operator conduct and why are'they necessary ? (1.0) t

- 6.05 ANSWER

-(a) (0.5 for each of the following)

High radiation in the spent fuel pool area.

Failure of fan E-4 (normal exhaust)

(b) The operator must verify proper damper alignment exists and that an exhaust fan is running because the "On" position of the control switch will start the fan without regard to damper position or exhaust fan status. (This is done to prevent placing positive pressure on the FHB and excessive positive or negative pressures on the ducts and plenums.) (1.0)

REFERENCE:

Lesson Plan H-7, FHB Ventilation Systen, pages 11 & 21

-~ q ,c- , , , , - , - - . , , , , - - , -

' 6.06 QUESTION (3.0)

Regarding the Main Steam System: -

(a)_ What are the two automatic closure signals for the Main Steam (1.0)

' Isolation valves ?

(b) What is the purpose of the check valve portion of the Main (1.0)

Steam Isolation valves ?. '

c

- (c) List four of the five instrumnets or control systems which receive main steam flow signals. (1.0)

-6.06 ANSWER Hi-Hi containment pressure (22 PSIG) and (0.5)

(a)

High Steamline Flow coincident with Low-Low Tave or Low steam.

line pressure ([>. 5)

(b) In the event of a steamline break upstream of a MSIV the' check valve prevents backflow from the other three steam generators.

t (1.0)

(c) (0.25 each for any four of the following)

Steam flow indicators .

High steamline flow (coincidence with low steamline prcesure  ;

or low-low Tave SI and MSL isolation) ,

Low f eed flow trip (Low S/G anticipatory trip G 25% SF/FF mismatch)

S/G Water Level Control System Main feedwater pump speed control sTuMW TANJ CHAftT- (tscoRbEf25.

REFERENCE:

Lesson Plans C-2a. Main Steam Piping Sysem, pace 26 &

27; C-8b. Main Feedwater Control System, page 48 8

e L

i

6.07 OUESTION (3.0) j Regarding the Steam Generator Level Control S'ystem (a)- Describe how the steam generator level _will change if the controlling steam flow signal had a small step increase (due to instrument malfunction) with no change in actual. steam flow ?

' E:<pl ai n your answer.

(2.0)

(b) Why.is the no-load S/G 1evel limited to'33 % ? (1.0) 6.07 ANSWER (a) S/G 1evel will increase above the programmed setpoint then slowly return to original setpoint. The difference'between' feed-flow and steam flow will. generate an error signal which will'open the feedwater regulating valve, which will initially raise'S/G 1evel..The level contro11er' integrates the difference between programmed level and actual level which makes the level error eventually dominate-over the flow error and shut the regulating (2.0)'

valve to decrease S/G level.

(b) This limits S/G mass for a steamline break inside containment accident (which limits RCS cooldown positive reactivity insertion and containment pressurization). (1.0)

REFERENCE:

Lesson Plan, C-Eb, Main FW Control System, pages 15 &18 l

i s

L

2 T

6.08'OUESTION (2.0)

The steam dump system uses several pressure transmitters for-

]

< control and interlock signals. CPT-505 (Tref), PT-506(turbine load rejection), and PT-507(main steam pressure)3 (a) With the plant initially operating normally at full power, how will PT-506-failing low af f ect the steam dumps ? Briefly. explain (1.0) your answer.

(b) With the plant initially at 5 % power during start-up (main turbine not operating), how will PT-507 failing low affect the(1.0) j steam dumps ? Briefly e:< plain your answer.

)

6.08 ANSWER (a)'The steam dumps will stay shut. (0.5) PT-506 failing-low will arm all the steam dumps (C-7A & B interlocks), but there would bo

no steam -dump demand signal . (0.5)

(b)-The steam dumps will shut if they are open or stay shut in any

? ' case. ( 0. 5) ~ In the steam pressure mode (used during start-up) the dumps are actuated by a steam pressure signal. With sensed steam pressure sianal failing low no opening signal will'be generated.

', (0.5)

REFERENCE:

Lesson Plan, C-26, Steam Dump System 4

i 4

3 e

t t

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~

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(I 6.09 QUESTION (2.0)

The Technical Specification bases describe the reactor tripts) generated by the-reactor protection system which protect the reactor from various events. List cach event for which the following trips are designed to protect the reactor.

(a) Positive rate trip (0.5)

(b) OPdelta T trip (0.5)

(c) S/G Low-Low water level (0.5)

(d) Pressurizer high water level (0.5) 6.07 ANSWER (a) Rod ejection accident (0.5)

F0EL tuTEG.RITY OM Alt.

(b) T. Em7 RCD E"crrdin; DNOHSAT Flux4A(M/pt)(Ostious)

OVE#.Po con (0.5)

(c) Loss of heat sink (0.5)

(d) Prevents water relief through the pressuriner S/Vs (0.5)

REFERENCE:

Technical Specifications B 2-5, 6, - & _7 i-1 I

k

6.10 QUESTION (2.0)

Unit 1 is-at-80 % steady state power with all controls in automatic.

How will each of the following events affect control rod motion ?

(assume one event at a time) Briefly explain your answer.

(a) One reactor coolant pump trips (0.5)

(b) One atmospheric dump valve is inadvertantly opened (0.5)

(c) One narrow range Th instrument f ails high (0.5)

(d) A power range upper detector f ails high (0.5) 6.10 ANSWER (a) Reactor-SCRAM. Loss of flow (one RCP above P-8) trip (0.5)

(b) Rods will-move out. Greater steam demand will lower Tave.

Since Tref didn't change rods will step out to eliminat Tave/ Tref

-mismatch. (0.5)

(c) Rods will step in. Tave signal is auctioneered high Tave so one channel high will generate a Tave/ Tref mismatch. (0.5)

(d) Rods will~ step in then stop. The power mismatch signal is a rate signal which initailly reacts as though reactor power exceeds turbine power. When the rate circuit output'becomes zero'Tave will be below Tref, however outward rod motion Will be blocked (by the high range nuclear power block C-2). (0.5)

REFERENCE:

Lesson Plan, A-3a, Full Length Rod Control System-a e

I

ab.11 QUESTION (2.0)

Lict tne reactor coolant pump reactor trips. Include associated reactor. power levels (0-10%, 10-35% and above 35%). (numerical

' values and coincidence logic such as 1/2 or 2/3 not desired).

6.11 ANSWER Below 10 % (P-7) no trips (0.4)

Between 10 and 35 % (between P-7 and P-8) loss of flow in two loops will cause a trip. 7,o (0.8)

(a) Undervaltage on swun buses'l;;; c: t.;; 1_;p;; will cause a reactor trip. g o , y coop 3)

(b) Underfrecuency on one bus will cause a reactor trip.

(c) Two reactor coolant pump breakers tripped will cause a reactor trip.

(d) Low flow in two loops will cause a reactor trip Above 35 % loss of flow in one loop will cause a trip. (0.8)

(a) Undervoltage on IEd. busewill cause a reactor trip.

(b) Underfrequency on one bus will cause a reactor trip.

(c) One reactor coolant pump breaker tripped will cause a reactor trip.

(d) Low flow in one loop will cause a reactor trip i, -

REFERENCE:

Lesson Plan, B-6a, Reactor Protection System END OF SECTION 6 4

4 9

)

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,-,---,,-,---,-,-n, ,n--- . - . , . - , , , , - - , , , , , . . - -

SECTION 7 PROCEDURES.--NORMAL, ABNORMAL EMERGENCY, AND RADIOLOGICAL CONTROL -

7.01 QUESTION (2.0)

During refueling operations:

(a) Why must fuel movement not occur until the reactor has been

' subcritical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> ? (1.0)

(b) Why should the Dillon Load Cell weight indicator on the

- manipulator crane control board be kept on the "A" (0 - 3000 Lb) scale except for emergency operations ? (1.0) 7.01 ANSWER (a) To allow for decay of short-lived fission products (f or -

consistency with accident analyses). (1.0)

(b) The interlocks depend upon weight indicator needle position.

Using a different scale will offset all the interlock (weight) trip setpoints. (1.0)

REFERENCE:

Technical Specifications 3.9.1 and 3.9.3 (Bases), Lesson Plan B-8, Fuel Handling System, FHRS-10 and page 30 I

-- .- -. . . - - - , , , - , - .--.n--n,-,,--- . - - - , , , , ,a.

7.02 QUESTION'(2.0)

Based on'10 CFR 20, " Standards for Protection against Radiation";.

(a) What are the' quarterly occupational radiation dose standards for individuals'in restricted areas ? (Whole body, skin, and (1.5) extremities)

(b)-How are these standards modified for a person less than-18 years ,

I (0.5) old ?

7.02 ANSWER (0.5)

(a) Whole body 1.25 rem /Qtr (0.5)

Skin 7.5 rem /Qtr-18.75 rem /Qtr. (0.5)

Extremities 10 */. Of the adult quarterly ~ limits (0.5)

(b)

REFERENCE:

10 CFR 20.101 and 20.104

-7.03 OUESTION (3.0)

Regarding Emergency Boration with the plant in modes 1 through 6:

(a) What are three indications of uncontrolled or unexplained reactivity increases which will require emergency boration ?

(1.0)

(b) What are the three normal operator actions required to initiate.

emergency boration ? (1.0)

(c) If the VCT make-up system is not available, what are three i alte.rnate means of emergency boration ? (1.0) 7.03 ANSWER (a) (0.33 each for any three of the following)

Unexplained control rod insertion (in auto)

Increased Tave with no increased load demand Increased nuclear power with no increased load demand Unexplained increasing count rate with-the plant shutdown uncurft+U.s0 R.c$ co0LDOWN Pou.oww0. A N TW.tP wm 40 e5F Acropin0M.

(b) (0.33 each)

Place VCT make-up control in borate position Set baron flow controller pot to maximum Place M/U controller in start position (verify 10 gpm flow)

(c) (0.33 for any three of the following)

-810 Emergency boration valve (CVCS-e+,4) +>

RWST BIT Manual emergency borate valve (CVCS-8471) LAST

REFERENCE:

OP AP-6, Emergency Boration i

l 1

(

l

{

- 7.04 OUESTION (2.0)

Draining the Reactor Coolant Vessel is a prerequisite to -

refueling. ,

(a) Why is it important to keep reactor vessel level above the centerline of the RCS loops 7 (El. 107') (1.0)

(b) While draining the RPV, what response would you expect to see in the Vessel Fill and Drain Level Indication System when the vessel head vent is opened 7 Briefly explain your answer. (1.0) 7.04 ANSWER' (a) A lower level may-cause suction to the RHR pump (s) to be lost.

(1.0)

(b) The indicated level will decrease (by about 14") due to venting the (0.5 PSIG) nitrogen supply which pressurized the

~

vessel head during draining. (The reference leg of the indication system is open to the atmosphere.) (1.0)

REFERENCE:

OP A-2:II. Reactor Vessel - Draining the Reactor Coolant System i

l 9

4 s i

7.05 OUESTION (3.0)

Due to major problems at other power plants Diablo Canyon has been required to make several rapid up power and down power changes during the last f ew hours. As a consequence the axial flux difference has been outside the target band during the last shift.

Use the attached Technical Specifications, AFD Curve, and the log entries listed below to answer the following questions.

0845 Unit 2 reached 100 % power 0847 Unit 2 AFD -7%

0850 Unit 2 AFD -9% (outside target band) 0852 Unit 2 AFD -10%

0855 Began power decrease 0900 Unit 2 at 90 % power, AFD -15%

0905 Unit 2 steady at 80 % power, AFD -12%

1000 begin power reduction to 50%, AFD at -11%

1035 Power steady at 50%, AFD -7%

1040 AFD -6% and tending towards cero (inside target band) 1340 power neutron flux trip set to 55%

(a) How long can you operate Unit 2 above 90% power with -10% AFD without violating the Technical Specifications ? (0.5)

(b) How long can you operate Unit _1 above 90% power with -10% AFD without violating the Technical Specifications ? (0,5)

(c) Describe the Technical Specification violations which occured ?

Briefly explain your answer (s). (1.0)

(d) Assuming that subsequent operations do not violate Technical Specifications or trigger any limiting conditions for operations, when (what time) may reactor power be raised above 50% power ?

Briefly explain your answer. (1.0) 7.05 ANSWER (a) 15 minutes above 90% power maximum (0.5)

(b) No time limit (0.5)

S. A+ tme 0900 Rn 00TsbE TMS Acca0M6lE WGt4Mn/ untr$ cv44 (0,5)

(c) 1. The decrease to450% should have started by 0950.

2. Power should have been below 50% g 30 minutes. (0.5) 09376. i'D NO (d) 4+dMrthe following day. The plant can go above 50% power when the penalty deviation time is less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. (1.0)

REFERENCE:

Lesson Plan B-4, Excore NI System, pace 35, Technical Specifications

l 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXfAL FLUX DIFFERENCE LIMITING CONDITIONS FOR OPERATION 3.2.1.1 For Unit I the indicated AXIAL FLUX DIFFERENCE (AFD) shall be main-tained within the allowed operational space defined by Unit 1 Figure 3.2-la. _.

APPLICABILITY: For Unit 1 MDDE 1 AB0VE 50 PERCENT RATED THERMAL POWER *.

ACTION:

i af With the indicated AXIAL FLUX DIFFERENCE outside of the Unit 1

.- Figure 3.2-la limits,

1. Either restore the indicated AFD to within the Unit 1 Fig- -

ure 3.2-la limits within 15 minutes, or

! 2. Reduce THERMAL POWER to less than 50K of RATED THERMAL POWER -

', within 30 minutes and reduce the Power Range Neutron Flux - High Trip setpoints to less than or equal to 55 percent of RATED' THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b.

- THERMAL POWER shall not be increased above 50K of RATED THERMAL POWER unless the indicated AFD is within the Unit 1 Figure 3.2-la limits.

SURVEILLANCE REQUIREMENTS 4.2.1.1.1 For Unit 1 the indicated AXIAL FLUX DIFFERENCE shall be determined to be within its lief ts during POWER OPERATION above 50 percent of RATED THERMAL

POWER by

-

a. Monitoring the indicated AFD for each OPERABLE excore channel:

1. At least once per 7 days when the AFD Monitor Alars is OPERABLE, and

2. At least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after restoring the AFD Nonitor Alara to OPERABLE status.

b. Monitoring and logging the indicated AXIAL FLUX DIFFERENCE for each OPERA 8LE excore channel at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least once per 30 minutes thereafter, when the AXIAL FLUX DIFFERENCE Monitor Alam is inoperable. The logged values of the indicated AXIAL FLUX DIFFERNECE shall be asstaned to exist during the interval preceding each logging.

4.2.1.1.2 For Unit 1 the indicated AFD shall be considered outside of its limits when at least 2 OPERABLE excore channels are indicating the AFD to be outside the limits.

L "See Special Test Exceptions Specification 3.10.2 DIABLO CANYON - UNITS 1 & 2 3/4 2-1 AMENDMENTS NOS. 3 & 1

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FIGURE 3.2-2a l

UNIT 1 AXIAL FLUX DIFFERENCE LIMITS l ~ -

l AS A FUNCTION OF RATED THERMAL POWER e

i l DIABLO CANYON - UNITS 1 & 2 3/4 2-14 M D M S NOS. 3 & 1 l

I l

- - - - . - . . . . . . . . . . - --------.,- , _ --,, . ,e---<--- mm-- ,-,a- - - - , , -

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1' AXIAL FLUX DIFFERENCE LIMITING CONDITION FOR OPERATION 3.2.1.1 For Unit 2 the indicated AXIAL FLUX DIFFERENCE (AFD) shall be main- [

tained within a 25% target band (flux difference units) about the target flux difference when THERMAL POWER is above 50K of RATED THERMAL POWER. The indi-

cated AFD any deviate outside the above required target based at greater than l or equal to 505 but less than 905 of RATED THERMAL POWER provided the indicated AFD is within the Acceptable Operation Limits of Unit 2 Figure 3.2-lb and the l cumulative penalty deviation time does not exceed I hour during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l The indicated AFD may deviate outside the above required target band at greater than 155 but less than 505 of RATED THERMAL POWER provided the cumulative penalty I

deviation time does not exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. ..

APPLICA8ILITY: For Unit 2 MDDE I above 155 of RATED THERMAL POWER *.

[

l ACTION:

l a. With the indicated AFD outside of the 25% target band and with THERMAL

! f POWER greater than or equal to 905 of RATED THERMAL POWER, within 15 minutes, either:

(

1. Restore the indicated AFD to within the target band Itaits, or

2. Reduce THERMAL POWER to less than 905 of RATED THERMAL POWER.

b. . With the indicated AFD outside of the above requi. red target band for more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of cumulative penalty deviation time during the

, previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or outside the Acceptable Operation Limits of Unit 2 Figure 3.2-Ib and with THERMAL POWER less than 905 but equal l to or greater than 50K of RATED THERMAL POWER, reduce:

1. THERMAL POWER to less than 505 of RATED THERMAL POWER within 30 minutes, and

2. The Power Range Neutron Flux - Hight Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

• See Special Test Exceptions Specification 3.10.2

1. Surveillance testing of the Power Range Neutron Flux channel any be performed pursuant to Specification 4.3.1.1 provided the indicated AFD is maintained within the Acceptable Operation Units of Unit 2 Figure 3.2-lb. A total of l' 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> operation say be accumulated with the AFD outside of the above required target band during testing without penalty deviation.

DIABLO CANYON - UNITS 1 & 2 3/4 2-lb AMENDMENTS NOS. 3 & 1

~

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued) c.

With the indicated AFD outside of the above required target band for more than I hour of cumulative penalty deviation time during the pre-vious 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and with THERMAL POWER 1ess than 505 but greater than 155 of RATED THERMAL POWER, the THERMAL POWER shall not be increased equal to or greater than 50K of RATED THERMAL POWER until the indicated AFD is within the above required target band.

• SURVEILLANCE REQUIREMENTS j 4.2.1.2.1 For Unit 2 the indicated AFD shall be determined to be within its l limits during POWER OPERATION above 15% of RATED THERMAL POWER by: .-

! a. Monitoring the indicated AFD for each OPERA 8LE encore channel:

1) At aMleast once per 7 days when the AFD Monitor Alarm is OPERABLE, i

2) At least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after restoring the AFD Monitor Alarm to OPERABLE status.

b. Monitoring and logging the indicated AFD for each OPERA 8LE excore

channel at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least i

once per 30 minutes thereafter, when the AFD Monitor Alarm is in-operable. The logged values of the indicated AFD shall be assumed to i

' i exist during the interval preceding each logging.

t 4.2.1.2.2 For Unit 2 the indicated AFD shall be considered outside of its + 5%

target band when two or more OPERABLE excore channels are indicating the AFD to be outside the target band. Penalty deviation outside of the + 5X target band shall be accumulated on a time basis of:

~

a. One minutai penalty deviation for each I minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above [

505 of RATED THERMAL POWER, and

b. One-half minute penalty deviation for each I minute of POWER ,

OPERATION outside of the target band at THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER.

l 1

DIABLO CANYDN - UNITS 1 & 2 3/4 2-2 AMENDF5iTS NOS. 3 & 1

C, POWER DISTRIBUTION LIMITS l

SURVEIn. LANCE REQUIREMENTS (Continued) 4.2.1.2.3 For Unit 2 the target flux difference of each OPERA 8LE excore channel shall be determined by measurement at least once per 92 Effective Full Power l Days. The provisions of Specification 4.0.4 are not appitcable.

4.2.1.2.4 For Unit 2 the target flux difference shall be updated at least once per 31 Effective Full Power Days by either determining the target flux dif- l '

forence pursuant to Specification 4.2.1.3 above or by linear interpolation I between the most recently measured value and W at the end of the cycle life.

The provisions of Specification 4.0.4 are.not applicable.

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G DIABLO CANYON - UNITS 1 & 2 8 3/4 2-3

7.06 OUESTION (2.5)

A major fire has broken out in reference paperwork stored in the control room. It will be necessary to evacuate the control room and implement Operating Procedure AP-8 " Control Room Inaccessability"e According to the procedure, what are the five steps which.are to be taken (for each reactor) prior to evacuation ?

7.06 ANSWER (0.5 for each of the following)

Manually trip the reactor Verify reactor trip Verify turbine trip Manually close MSIVs and bypass valves Establish Auto make-up control

REFERENCE:

OP AP-8, Control Room Inaccessibility, page 4

7.07 QUE3 TION (2.5)

Match the potential indicated condition with the appropriate Emergency Action Level (EAL).

1) Loss of offsite power and loss of onsite power for more than 15 minutes. - (0.5)

2) Major damage to spent fuel in containment or fuel handling building. (0.5)

3) Effluent monitors detect levels corresponding to 1 Rem /Hr. whole body or 5 Rem /Hr thyroid at the site boundary under actual meterological conditions. (0.5)

4) Loss of-both residual heat removal trains. (0.5)

5) Fire within the plant lasting more than 10 minutes after initiating fire fighting efforts. (0.5)

EALs A. Unusual Event D. Alert C. Site Area Emergency D. General Emergency 7.07 ANSWER

1) C (0.5)

2) C (0,5)

3) D (0.5)

',g og 6 (,w FAG auDMMitS d'ThL SGusMGNI~ W '0"##I#

y a nMomenou) e

REFERENCE:

EP-G1. Accident Classification and Emergency Activation, Table 1

,.e , - , . . . - - - . ,- .. , , , , _ , , _ . . . , ,y--.n--, . - , , , , - , , - _ , ,,_.,m,--- a- , , , , , ,,,,,_,,,_---,n,_.--.n,,--,w,.,----.-a,,,a,, .. . - --

7.08 QUESTION (3.0) j Following a reactor trip with an inadvertent SI, (a) List eight of the ten immediate verifications performed by the operator. (2.0) 4 (b) List the four conditions that must be met to terminate SI. (One 1 item has two parts) (1.0) 7.08 ANSWER (a) (0.25 for any eight of the following)

Verify Reactor trip Turbine trip Vital 4 KV bus status Check if SI Actuated Verify Phase A Isolation

Containment Vent Isolation ESF pump and valve status FW isolation

Main Steamline isolation not required Containment Spray and Phaso B not required (b) (0.25 for each of the following) l RCS subcooling greater than 20 F Secondary Heat sink greater than 460 GPM Aux. feed flow or greater than 4 % NR on one steam generator RCS pressure stable or' increasing PZR level greater than 4%

i dAO%)

REFERENCE:

EP E-0, Reactor Trip or Safety Injection pages 3, 4, 5, and 11, EP E-l, LoS5 om REACTDR, oR SecWo4Ry coomuT, PA(rE 7 i

i i I

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7.09 QUESTION (2.5)

There are six critical safety function status trees which are monitored during an emergency transient. The status trees contain paths with four different colors.

(a) List the six status trees in order of priority (the first being the highest). (1.0)

-(b) List the color of each othe four paths in order of priority and explain what general operator action is required for each color.

(1.5) 7.09 ANSWER (a) (0.17 each)

1. subcriticality

2. core cooling

3. heat sink

4. RCS integrity S. containment

6. inventory (b)

1. Red (0.2) - stop optimal recovery in progress and immediatly initiate indicated functional restoration unless there is a higher priority red path. (0.2)

2. @k re

_ W (0.2) - finish statuu tree check then initiate functional restoration of highest priority eramee (assuming no red paths exist). (0. 2) MWh

3. Yellow (0.2) - Monitor trees. Operator perogative on continuing optimal recovery or functional recovery. (0.2)

4. Green (0.2) - No action required. (0.1)

REFERENCE:

Westinghouse EOP Guidelines

=

7.10 QUESTION (2.5)

You are supervising a routine monthly surveillance test of Diesel Generator 1-3.

(a) Which of the following protective relays are cut out for normal service but can be cut in for the duration of the test ? (1.5)

1. Overcurrent

2. Differential

3. Loss of Field

4. 4 KV bus differential

5. Directional ~ power (b) According to procedure it is preferable to conduct the test with the 4 KV bus paralleled with the unit Aux. transformer. What is the problem with conducting the test with the 4 KV bus paralleled with the Start-up bank ? (1.0) 7.10 ANSWER (a) (0.5 each)

1. Overcurrent

3. Loss of field

5. Directional power (b) With the 4 KV bus paralleled with the Aux. transformer a main unit trip will automatically trip the bus feeder breaker. If the start-up bank is paralleled with the 4 KV bus and the start-up power were lost the start-up feeder breaker will not trip. In the second case the diesel will try to pick up the entire 4 KV and start-up buses. (1.0)

REFERENCE:

STP M-9A-3, Diesel Engine Generator 1-3, Routine Surveillance Test, paaes 8 and 11 END OF SECTION 7

SECTION 8 ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS -

8.01 QUESTION (3.0)

In accordance with the Technical Specifications for Administrative controls, Section 6.2.2 " Plant Staff":

)

1 (a) What is the minimum crew composition required with both units in mode 17 (1.0)

(b) What is the minimum crew composition required with one unit in

[

mode 1 and the other unit in mode 6 7 (1.0)

(c) What are the manning requirements for the site Fire Brigade 7 (1.0) i i 8.01 ANSWER (a) (0.25 for each) 2 - SROs (SS and SOL) 3 - ROs (at least one per unit plus one) 3 - AOs (at least one per unit plus one)

! 1 - STA funless one of the SROs meets the requirements for STA)

(b) (0.25 for each) 2 - SROs (SS and SOL) 3 - ROs (at least one per unit plus one) 3 - ads (at least one per unit plus one) 1 - STA (unless one of the SROs meets the requirements for STA)

(c) At least five members on site at all times, not including the Shift Supervisor or the two other members of the minimum shift crew a necessary.to shutdown the unit or any personnel required for other essential functions during a fire emergency. (1.0)

REFERENCE:

Technical Specifications, Table 6.2-1 and paragraph 6.2.2 i e.

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8.02 OUESTION (3.0)

The CODE OF FEDERAL REGULATIONS (10 CFR 55) defines the general provisions for Operator's Licenses. In accordance with these regulations:

(a) What are the " controls" of a nuclear facility ? (1.0)

(b) When is an individual deemed to be " operating" the controls of a nuclear facility ? (1.0)

(c) Who may operate the " controls" of.a nuclear facility without an-Operator License ? (1.0) 8.02 ANSWER (a) " Controls" are apparatus and mechanisms the manipulation of which directly affects the reactivity or power level of the reactor.

(1.0)

(b) An individual in deemed to operate the controls of a facility if he (1) directly manipulates the controls or (2) directs another to manipulate the controls. (1.0)

(c) An individual may manipulate the controls as part of his training to qualify for an operator license under the direction and in the presence of a licensed operator or senior operator. (1.0)

REFERENCE:

10 CFR 55

?

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A 9

8.03 QUESTION (2.0)

While the plant is in mode 3 a routine start-up is progressi.ng.

During surveillance testing on one of the Diesel Generators a diesel fuel oil' transfer pump fails to operate. The Maintenance Supervisor states that the pump motor will have to be replaced and that this can be accomplished in no later than 49 hours5.671296e-4 days <br />0.0136 hours <br />8.101852e-5 weeks <br />1.86445e-5 months <br />. Explain why the startup can or cannot proceed. (A cooy of the Technical Specifications for emergency power sources is provided.)

0.03 ANEWER F0EL oil TA M PER TILA* = one fuel oil A diesel4 ;rnr 2tr' is out of transfer pump. With SEDS DO*$Nfyrviceduetothelossof of service you are in an action statement and the startup cannot proceed until the LCD is met. Entry into an operating mode is not allowed unless all LCOs are met for that operating mode without reliance on action statements.. (2.0)

REFERENCE:

Technical Specification 3.0.4

SEE PSRC INTERPRETATION

~~

3/4.8 ELECTRICAL POWER SYSTEMS 3/4 .1 A.C. SOURCES '

y .NG O

LIMITING CONDITION FOR OPERATION

3. 8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

a. Two independent circuits (one with delayed access) between the offsite transmission System, and network and the Onsite Class 1E Distribution

b. Three separate and independent diesel generators,* each with:

1. A separate engine-mounted fuel tank containing a minimum volume of 200 gallons of fuel, and

2. Two supply trains of the Diesel Fuel Oil Storage and Transfer System with a cotoined storage of 31,023 gallons of fuel for one unit operation and $2,046 gallons of fuel for two unit operaticn.

APPLICAEILITY: MODES 1, 2, 3, and 4. '

ACTION:

a. With either an offsite circuit or diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Specifica-tions 4.8.1.1.la. and 4.8.1.1.2a.2) within I hour and at le'ast once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter; restore at least two offsite circuits and three diesel generators to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOW',

within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With one offsite circuit and one diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Specifica-tions 4.8.1.1.la. and 4.8.1.1.2a.2) within I hour and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter; restore at least one of the inoperable sources to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or be in at least H T STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Restore at least two offsite circuits anc tnree diesel generators to OPERA 3LE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from the time cf

~

initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. '

L'

'CDERAE:LITY of the third (common) diesel generator shall incluce the caoatility cf functioning as a power source for the required unit upon autcmatic cems,c fro. that unit.

CAE.' : CANYON - UNITS 1 & 2 3/4 8-1

ELECTRICAL POWER SYSTEMS LIFITING CONDI ON FOR OPERATION ACTION (Continued) c.

With one diesel generator inoperable in addition to ACTION a. or b.

above verify that:

1.

All required systems, s'ubsystems, trains, components and devices that depend on the remaining OPERABLE diesel generators as a source of emergency power are also OPERABLE, and

2. When in MODE 1, 2, or 3 that at least two auxiliary feedwater pumps are OPERABLE.

If these conditions are not satisfied within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> be in at least:

HOT following STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUT 004N within tre 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

d.

With two of the above required offsite A.C. circuits inoperable, demonsteate the OPERABILITY of thre:- di e :al ger.erators by performing the requirements of Specification 4.8.1.1.2a.2) within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter, unless the diesel generators are already operating; restore at least one of the inoperable offsite sources to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT r STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. With only one offsite source restored, restore at least two offsite circuits to OPERABLE status (

within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from time of initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

e. With two or more o+ the above required diesel generators incperable, demonstrate the OPERABILITY of two offsite A.C. circuits bT perforning the requirements of Specification 4.8.1.1.la. within I hour and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter; restore at least two of the inoperable diesel generators to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or te in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHJTOCWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Restore at least three diesel gene a-

- tors to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from time of initial loss c-be in least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDCdN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

f. With one supply train of the Diesel Fuel Oil Storage and Transfer System inoperable, restore the inoperable system to OPERABLE statu within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 ho.rs and be in HOT SHUTCOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

g. With both supply trains of the Diesel Fuel Oil Storage and Transfe-System inoperable, restore at least one supply train, incluoing the -

common storage system, to OPERABLE status within I hour or be in at least HOT STANDBY witnin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

CIAB.C CANYDN - UNITS 1 & 2 3/4 8-2

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMEN1 4.8.1.1.1 Each of the above .equired independent circuits between th'e offsite transmission network and the Onsite Class IE Distribution System shall be:

a. Determined OPERABLE at least once per 7 days by verifying correct breaker alignments, indicated power availability, and

b. Demonstrated OPERABLE at least once per 18 months during shutdown by:

1) Transferring 4 kV vital bus power supply from the normal circuit to the alternate circuit (manually and automatically) and to the delayed access circuit (manually), and

2) Verifying that on a Safety Injection test signal, without loss of offsite power, the preferred, immediate access offsite pc-er source energizes the emergency busses with permanently connected loads and energizes the auto-connected emergency (accident) loads through sequencing timers.

4.8.1.1.2 Each diesel generator

• shall be demonstrated OPERABLE:

a. In accordance with the frequency spec'.fied in Table 4.8-1 on a STAGGERED TEST BASIS by:""

! 1) Verifying the fuel level in the engine-mounted fuel tank,

( 2) Verifying the diesel starts from ambient condition and accelerates to at least 900 rpm in less.than or equal to 10 seconds. The generator voltage and frequency shall be 4160 1 420 volts and 60 2 1.2 Hz within 13 seconds after the start signal. The

diesel generator shall be started for this test by using one of the following signals

a) Manual, or r b) Simulated loss of offsite power by itself (Startup bus undervoltage), or c) A Safety Injection actuation test signal by its11f.

" Tests of Diesel Generator 3 to satisfy the frequency specifiec in Table 4.5-1 and in Surveillance Requirement 4.8.1.1.2.b for one unit may be countec in determining whether the frequency specified in Table 4.8-1 and in Surveil-lance Requirement 4.8.1.1.2b for the other unit is satisfied. Unit-specific portions of this Surveillance Pequirement for Diesel Generator 3 shall be performed on an alternating schedule with signals from Units 1 and 2.

• • All diesel generator starts for the purpose of this surveillance test may be prececed by an engine prelube period. Further, all surveillance tests, with int exception of once per 184 days, may also be preceded by warmup procedL-es -

(e.g., gradual acceleration and/or gradual loading > 150 sec) as recem-encec ty the manufacturer so that the mechanical stress and weee on the diesel engine is minimized.

CIAELC CANYON - UNITS 1 & 2 3/4 8-3

ELECTRICAL POWER SYSTEMS

^

SURVEILLANCE REOUIREMENTS (Continus )

3)

Verifying the generator is synchronized, loaded to greater than or equal to 2484 kW in less than or equal to 60 seconds, anc operates for greater than or equal to 60 minutes,

4) -

Verifying the diesel generator is aligned to provide standby power to the associated emergency busses," and

5) Verifying the diesel engine protective relay trip cutout switen is returned to the cutout position following each diesel generator test.

b.

At least once per 18 months, during shutdown, by:

1) Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for this class of standby service;

2) Verifying that the load sequence timers are OPERABLE with ea-h load sequence timer within the limits specified in Table 4.8-2; 3)

Verifying the generator capability to raject a load of greater than or equal to 508 kW while maintaining voltage at 4160 + 420 volts and frequency at 60 + 3 Hz; -

4) ~ Verifying the generator capability to reject a load of greater than or equal to 2484 kW without tripping. The generator voltage shall not exceed 4580 volts during and following the load rejecticr;

5) Simulating a loss of offsite power by itself, and:

a) Verifying de-energization of the emergency busset and load shedding from the emergency busses, and l

b) Verifying the diesel starts on the auto-start signal, energizes the emergency busses with permanently connecte:

loads within 10 seconds, energizes the required auto-connected loads through sequencing timers and operates for greater than or equal to 5 minutes while its generator is loaded with the permanent and auto-connected loads. After

! energization of these loads, the steady state voltage a t i frequency of the emergency busses shall be maintained at 4160 2 420 volts and 60 2 1.2 Hz during this test.

"May or 4. be the assqciated bus in the other unit if that unit is in MODE 1, 2, 3 DIAB.'.0 CANYON - UNITS 1 & 2 3/4 8-4 l

B.04 OUESTION (3.0)

The CODE OF FEDERAL REGULATIONS 10 CFR 50.72 and 10 CFR 20.403 requires immediate notification (within a period of one hour) to the NRC Operations Center via the Emergency Notification System f or various events. Which of the events listed below reauire immediate notification 7 (0.5 each)

1. A worker recieven an exposure of 250 Rem to his right hand.

2. Declaration of an Unusual Event at Unit 2.

3. Damage to property on the site in e:: cess of $250,000.

4. A reactor trip followed by depressurication to 1700 PSIG without safety injection caused by a stuck open PORV.

5. Due to high winds one meterological tower is lost.

6. The facility receives warnino of a Tsunami wave that could affect the intake structure.

l 8.04 ANSWER

1. no (0.5)

2. yes (0.5) l 3. yes (0.5)

4. yes (0.5)

5. no (0.5) p NES , n F C ANDtDATli! >$6uMG5 wts Posss AM ActvAL THA8M

6. yes (0.5) 70 TIM 98 MT. (0,5)

REFERENCE:

10 CFR 50.72 and 10 CFR 20.403 i

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8.05 QUESTION (2.0)

Technical Specification 3.0.3 (attached) describes actions to be taken when LCOs are not met.

OP (1.0)

(a) What is the bases Technical Specification 3.0.3 ?

3 (b) In which of the following cases does Technical Specification 3.0.3 apply ? (Driefly explain why.) (1.0)

1. In mode 1. charging pumps 1-1 and 1-3 are inoperable.

2. In mode 5, Aux. SW pumps 1-1 and 1-2 are inoperable.

8.05 ANSWER

-(a) This specification delineates the measures to be taken f or those circumstances not directly provided for.in the action statements.and whose occurance would violate the intent of a specification. (1.0)

(b) 1. Does not apply. T.S. cover situation with one charging pump operable. (0.5)

Does not et PPW*

2. ^;;1 ire. T.S. de mrt cover case with no ASW pumps availablegd.

M005 5 (0.5)

REFERENCE:

Technical Specification 3.0.3 and Bases l

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3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY '

LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.

3.0.2 Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.

3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within I hour action shall be initiated to it, place the unit in in:

as applicable, a MODE in which the specification does not apply by placing

a. At least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,

b. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and

c. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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.

3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the conditions for the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION requirements. This pre-vision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION statements. Exceptions to these requirements,are stated in the individual specifications.

t 3.0.5 Limiting Conditions for Operation including the associated ACTION re-quirements shall apply to each unit individually unless otherwise indicated as follows:

a. Whenever the Limiting Conditions for Operatien refers to systems or components which are shared by both units, the ACTION requirements will apply to both units simultaneously. This will be indicated in the ACTION section;

b. Whenever the Limiting Conditions for Operation applies to only one unit, this will be identified in the APPLICABILITY section of the specification; and

c. Whenever certain portions of a specification contain operating para--

eters, Setpoints, etc., which are different for each unit, this will be identified in parentheses, footnotes or body of the reacirement.

L DIABLC CANYON - UNITS 1 & 2 3/4 0-1

SEE PSRC INTERPRETATION APPLICABILITY SURVEILLANCE REOUIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless othemise stated in an individual Surveillance Requirement.

4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with:

a.

A maximum interval, butallowable extension not to exceed 25% of the surveillance b.

The combined time interval for any 3 consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval.

4.0.3 Failure to perform a Surveillance Requirement within the specified time interval shall constitute a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications. Surveillance Requirements do not have to be performed on inoperable equipment.

4.0.4 Entry into an OPERATIONAL MODE or other specified condition s' hall not be m;'e er.;ess the Surveillance Requirement (s) associated with the Limit ng Condition for Operation has been performed within the stated surveillance interval or as otherwise specified.

4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2 and 3 components shall be applicable as follows:

a. Inservice inspection of ASME Code Class 1, 2 and 3 components and inservice testing of ASME Code Class 1, 2 and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler anc Pres-sure Vessel Code and applicable Addenda as required by 10 CFR 50, Sec-tion 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g)('6)(i);

b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vesse~1 Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:

ASME BOILER AND PRESSURE VESSEL REQUIRED FREQUENCIES FOR CODE AND APPLICABLE ADDENDA PERFORMING INSERVICE TERMINOLOGY FOR INSERVICE INSPECTION AND TESTING INSPECTION AND TESTING ACTIVITIES ACTIVITIES Weekly Monthly At,least once per 7 days I Quarterly or every 3 months At least once per 31 days At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days - s Yearly or annually At least once per 366 days CIABLO CANYON - UNITS 1 & 2 3/4 0-2 i

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PLANT SYSTEMS -

3/4.7.4 AUXILIARY SALTWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4.1 At least two auxiliary saltwater trains shall be OPERABLE.

APPLICABILITY: HODES 1, 2, 3, and 4.

ACTION:

With only one auxiliary saltwater train OPERABLE, restore at least two trains to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE RE0VIREMENTS 4.7.4.1 At least two auxiliary saltwater trains shall be demonstrated OPERABLE at least once per 31 days by verifying that each valve (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position, is in its correct position.

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DIAE.0 CANYON - UNITS 1 & 2 3/4 7-12

e 8.06 QUESTION (3.0)

Regarding the use and administration of plant procedures:

(a) Describe the three Technical Specification Requirements which must be met when making a temporary change to required procedures.

(1.5) ~

(b) List three of the four situations wnich require a procedure to be "in hand" during the performance of the associated task.

(1.5) 8.06 ANSWER (a) (0.5 for each of the following)

1. The intent of the original procedure is not altered.

2. The change is approved by two members of the plant management staff, at least on of whom hold a SRO license on the unit affected.

3. The change is documented, reviewed by the PSRC and approved by the Plant Manager within 14 days of implementation.

(b) so.5 each for any three of the following)

1. Procedure is complex or extensive in nature

2. Procedure is infrequently used.

3. The procedure involves tasks which must be performed in a specified sequence.

4. The procedure is of a check list nature or incorporates the required data directly into the body of the procedure.

REFERENCE:

Technical Specification 6.8.3, and NPAP E-4, Procedures, page 11 3

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8.07 OUESTION (3.0)

Regarding the Technical Speci fication Equipment Operability -Status Sheet:

(a) When are these sheets required to be filled out ? (0.5)

(b) Who is allowed to fill out this form ? (0.5)

(c) List four of the five times (or situations) when the Shift Foreman must review all outstanding Technical Specification Equipment Operability Status Sheets. (2.0) 0.07 ANSWER (a) Whenever Technical Specification related equipment is rendered or declared inoperable. (0.5)

(b) The Shift Foreman or his delegate can fill out the form.

(0.5)

(c) (0.5 each for any four of the following)

1. At the beginning of each shift

2. Prior to removing any T.S. equipment from service which is required in the current plant mode

3. Immediatly after declaring any equipment, recuired by T.S.,

inoperable

4. Prior to any planned plant mode changes

5. As soon as possible following a reactor trip or forced shutdown

REFERENCE:

AP C-6S4, Contro of Equipment Required by the Plant Technical Specifications 8.08 OUESTION (2.0)

Regarding sealed valves and the associated checklists which-are controlled in accordance with'AP C-951, Supplement i to Nuclear Plant. Administrative Procedure Sealed Valve, are the following statements TRUE or FALSE 7 (a) The Reactor. Operator on shift has authority to give permission to operate a sealed valve. (0.5)  !

(b) A sealed valve which appears on two checklists can have its position / seal checks transposed from one checklist to the other.

(0.5) -l (c) To check a sealed valve OPEN, one should attempt to move the valve closed to verify valve movement, then restore the valve to fully open position. . (0.5)

(d) To check a sealed valve CLOSED, one one should attempt to move the valve open to verify movement, then restore the valve to the fully closed position. (0.5)

'O.08 ANSWER (a) False (Shift Foreman or his designated alternate) (0.5)

(b) True (must be from valid checklist at SFs direction) (0.5)

(c) True (0. 5)

(d) False (0.5)

REFERENCE:

AP C-9S1, Supplement 1 to Nuclear Plant Administrative Procedure Sealed Valve

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8.09 OUESTION (3.0)

At Diablo Canyon MAN-ON-LINE, CAUTION, and INFORMATION. tags are used.

-(a) Briefly describe the use of each of these tags. (1.5)

(b) When testing a device which has been removed from service.who must " report-off" prior to the test ? (0.5)

(c) Can a CAUTION tag be used in place of a MAN-ON-LINE tag for uncompleted work following testing 7 (0.5)

(d) If'two people will-be working cn1 a motor to be replaced, what MAN-ON-LINE tags must be on the circuit breaker while the work is in-

.progess ? (0.5) 0.09 ANSWER (a) (0.5 for each of the following)

1. A MAN-ON-LINE taa is used in conjunction with a. clearance to mark equipment for non-operation.

2. A CAUTION tag is used to mark any equipment which is not cleared, but for safety reasons should not be operated except under specific. instructions.from the station or the named individual.

3. An INFORMATION tag is used to provide general information regarding status of plant equipment.

?

(b) Everyone who " reported-on" must " report-off" prior to the test.

(0,5)

(c) No.(MAN-ON-LINE tags must be used) (0.5)

(d) One tag for each worker, and the Shift Foreman's tag. (0.5)

REFERENCE:

NPAP C-7,-Tagging Requirements

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9.10 QUESTION (1.0)

There are Technical Specification requirements for both off-site and on-site electrical power sources. What problem would be created if the Motor Operated Disconnect Switch was inperable and in the closed position ?

8.10 ANSWER The SOOKV switch-yard would not be available as one of the "off-site transmission network" sources. In other words there would be only one of'the two independent circuits between the off-site trans-mission network and the on-site class IE distribution system re-quired by Technical Specifications. (1.0)

REFERENCE:

Technical Specification 3/4.8.1 END OF SECTION EIGHT 1

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