Text

Exam Repts 50-282/OL-88-01 & 50-306/OL-88-01 on 880711-14. Exam Results:Three Senior Reactor Operators & Three Reactor Operators Passed Exams
	ML20151U319
Person / Time
Site:	Prairie Island
Issue date:	08/08/1988
From:	Hare S, Parkinson K, Shepard D; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:	;
Shared Package
ML20151U316	List: ML20151U312; ML20151U319;
References
	50-282-OL-88-01, 50-282-OL-88-1, 50-306-OL-88-01, 50-306-OL-88-1, NUDOCS 8808190036
	Download: ML20151U319 (158)
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s U. S. NUCLEAR REGULATORY COMMISSION REGION III Reports No. 50-282/0L-88-01; 50-306/0L-88-01 Docket Nos. 50-282; 50-306 Licenses No. DPR-42; DPR-60 -

Licensee: Northern States Power Company 414 Nicollet Mall Minneapolis, MN 55401 Facility Name: Prairie Island Examination Administered At: Prairie Island Examinations Conducted: Senior Reactor Operator and Reactor Operator h)nti b1 l Examiners: 9'/7/f8 S. M. Hafe /' Date

. Aff _ . . Wil77

0. L. Shepard Date Adf ls t . lTh' K. L. Rarkinton e Approved By: [Thomas M. Burdick, Chief kkff Dath '

Operator Licensing Section Examination Summary l Examination administered on July 11-14, 1988 (Reports No. 50-282/0L-88-01; No. 50-306/0L-88-01)

Areas Inspected: Written and/or operating exams were administered to five reactor operators and five senior reactor operators.

Results:.. Three senior reactor operators and three reactor operators passed

, the examinations.

gjS19003688081o y ADOCK 050002 2 PD

f DETAILS 1 Examiners S. M. Hare, NRC D. L. Shepard. NRC X. L. Parkinson, Sonalysts

2. Examiner Observations During the administration of the operating exams the examine s did not observe generic strengths or weaknesses. The examiners did note isolated weaknesses in the area of control board manipulations. These weaknesses may be addressed by the facility training staff on a case by case basis.

M ter administering the written examinations the examiners became aware of the "G Procedures" which discuss the cdministrativa controls on surveillance testing. These procedures were not provided to the examiners prior to the exam. These procedures should be included in the reference material package provided to the NRC prior to the next scheduled examination (Replacement or Requalification).

3. Exit Meeting An exit meeting was held following the examinations on July 14, 1988 with facility management and training staff representatives. ,

NRC representativec in attendance were:

S. M. Hare, Chief Examiner D. L. Shepard, Examiner K. L. Parkinson, Sonalyst Contractor Examiner l J. E. Hard, Senior Resident Inspector !

M. M. Moser, Resident Inspector l

Facility representatives in attendance were:

T. Amundson, Superintendent of Training ,

L. Waldinger, Manager, Production Training i D. Reynolds, Operations Training Supervisor i D. Mendele, Superintendent Engineering and Radiation Protection l W. Bell, Instructor R. Holthe, Shift Manager M. Waldley, Shift Manager The facility management acknowledged the examiners observations discussed in Section 2 of this report, l

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At the conclusion of the~ written examination the facility is given a copy of-the examination and the answer key for both the SR0 and R0 examinations. The facility then has five working days in.which to provide written comments concerning the examinations'to the NRC.

The fol16 wing paragraphs contain the facility comments concerning the examinations followed by the NRC comments.

REACTOR 0PERATOR'S EXAMINATION QUESTION 1.22 (1.00)

Which of the following steam _ generator parameters is nearly linear (directly proportional) to reactor power as turbine load is ramped from 25% to 75%?

(Rod control is in automatic) (SELECT THE CORRECT PARAMETER). (1.0)

A. Delta-H (steam anthalpy - feed enthalpy)

8. Delta-T (steam temperature - feed temperature)

C. Steam flow

0. Steam generator level ANSWER '. 22 (1.00)

C. Steam flow COMMENT 1.22 (1.00)

The first possible answer (A) is delta-H. The use of Capital H (total enthalpy) rather than lower case h (specific enthalpy) is misleading. In reality Q = AH.

The use of Enthalpy (H) rather than specific enthalpy (h) should make either case "A" or "C" acceptable resporses since Q = mAh (Ah decreases as power )

i.icreases) and Q = AH (AH is directly proportional to power). '

NRC Response:

l The answer key was revised to accept answer A. [ Delta-H (steam enthalpy - I feed enthalpy)] or answer C. [ Steam flow] for full credit.

QUESTION 1.24 (1.00) ;

Which of the following indications DOES NOT verify that natural circulation is !

established? (SELECT THE INCORRECT INDICATION) (1.00) l l

A. Core Outlet temperatures are constant or decreasing i

B. Decrease in steam pressure during constant decay heat generation C. Constant or decreasing Delta T across the reactor core less than the full load Delta T D. Constant steam generator level.with a' constant auxiliary feedrate and a constant or decreasing steam pressure.

ANSWER 1.24 (1.00)

B. (Decrease in steam pressure during constant decay heat generation) (1.0)

COMMENT 1.24:

There is no incorrect response. Steam pressure stable or decreasing is used as an indication of natural circulation.- This question should be deleted.

NRC Response:

The reference, IES-1.1 Revision 4, Page 1 of 1, Natural Circulation Conditions, Attachment A, provided by the facility states "SG pressure - STABLE OR DECP. EASING is a condition that supports or indicates natural circulation flow". The reference, Westinghouse, Thermal - Hydraulic Principles and Applications to the Pressurized Water Reactor,1982, Page 14-27,. provided as

'the reference for the correct answer states an interruption of natural pre-existing change circulation is indicated by a decrease in steam pressere during constant decay heat generation. The question directed the candidates to select the indication that DOES NOT verify that natural circulation is established. Of the answers provided, answer B; decrease in steam pressure during constant decay heat generation, is an indication that occurs when aatural circulation is interrupted and is therefore the correct answer. The question is retained. )

QUESTION 3.03 (2.10)

A. The "Low Power" rod stop blocks outward rod motion in automatic.

1. WHAT is the "Low Power" rod stop SETPOINT? (0.3)

2. WHY is the "Low Power" rod stop REQUIRED? (0.3)

B. The "High Bank D Rod Position" rod stop blocks outward rod motion in automatic. l l

Other than blocking outward rod r..otion, WHAT does the "High Bank D Rc,d Position" rod stop PREVENT? (0.3)

C. Other than blocking automatic rod withdrawal, what are the other functions of the "overtemperature delta-T" rod stop? (LISTTWO)(0.6)

D. What is the SETPOINT and COINCIDENCE for the "overpower delta-T rod stop"? (0.6) 2 i i

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ANSWER 3.03 (2.10)

A. 1. Turbine Impulse pressure < 15 power (0.3)

2. Because of instabilities in the rod control system durinq Icw power conditions (0.3)

B. Misalignment of tne system step counters (0.3)

C. 1. Blocks manual rod withdrawal (0.3)

2. Actuates a turbine runback (0.3)

D. Overpower delta-T trip setpoint -5% (0.3) 2/4 coincidence (0.3)

COMMENT 3.03:

3.03A The "Low Power" rod stop could be interpreted as the Intermediate Range High Flux (Low) rod stop at 20% current equivalent on 1/2 IR channels. It functions to limit the consequences of a startup accident (X-HIAW-1-243). This should be an alternate acceptable answer.

The Rod Control System is unstable at < 15% power due to the non-linearity of turbine first stage pressure at that low power (Ref. RE-12). This should be an alternate acceptable answer.

3.038 The question insinuates another function for the "High Bank D rod .

position" rod stcp signal. The Bank D rod withdrawal limit rod stop only prevents outward automatic rod motion of Bank 0

, (Ref. X-HIAW-1-239). This question should be deleted.

3.03C The question insinuates two other functions exist for the overtemperature delta-T rod stop. Exceeding 0 TAT runback limit provides a signal to turbine control and to rod control. The signal to the rod control system (the overtemperature delta-T rod stop) ;

stops outward rod motion in auto and manual. The only acceptable ;

answer is "blocks manual rod withdrawal". Answer 2, "Actuates a i turbine runback" should be deleted. l l

NRC Response:

3.03A The answer key was revised to accept the recommended alternate l answer.

3.03B The reference, System Description B-5, Revision 0, Paragraph 4.2, Pages 23 and 24, provided with the examination answer key, states in part "The 'High Bank D Rod Position' rod stop blocks outward rod motion in automatic only due to all rods being fully withdrawn from 3

the core. The interlock pravents misalignment of the system step counters." The facility's comments referenced X-HIAW-1-239 an incorrect reference for the coment. The referenced drawing, X-HIAW-1-239, is titled Logic Diagrams Primary Coolant System Signals. The correct reference is X-HIAW-1-243, which is titled Diagram Rod Control and Rod Blocks. Since a conflict existed with the facilities reference material, the question was deleted from the examination as requested by the facility.

3.03C The reference, System Description.B-5, Revision 0, Paragraph 4.2, Pagus 23 and 24, provided with the examination answer key, states in part "The 'Overtemperature delta-T' rod stop blocks bcth automatic and manual rod withdrawal. The setpoint is the OT delta T trip setpoint - 5% with a 2/4. Drawing X-HIAW-1-239, Revision A, states pre-existing modification that Overtempcrature delta T blocks auto-matic and manual rod withdrawal and auto turbine runback. Since a conflict exists with the facility reference material, the answer "actuates a turbine runback" was deleted from the answer key as requested by the facility.

QUESTION 3.06 (2.50)

During power operation a load rejection occurs that causes a 5F delta T (auctioneered high Tavg - Tref) error signal. If the steam dumps fail to automatically respond to the. load rejection, what could cause the steam dump valves to fail to respond automatically? (LIST FIVE [5] CONTROLS OR CONDITIONS you would check or verify, INCLUDE COINCIDENCE AND SETPOINTS where appropriate) (Do not consider cou out solenoids, blown fuses, etc.)ponent failures such as binding valves, burnt (2.5)

ANSWER 3.06 (2.50) ,

1. The steam dump "Off/ Reset-On-Bypass Intlk" switches (0.25) in the ON position (0.25)

2. The steam dump "Mode Selector Control" switch (0.25) in the :

Ta!g CONTROL position (0.25)

3. Reactor coolant loop temperatures (0.25) _540F (0.25)

4. No turbfee trip (0.25) 2/2 stop valves (0.25)

5. Air pressure available to the valves i

6. Circulating water pumps operating (breaker closed) (0.25) 1/2 (0.25) i

7. Condenser vacuum _15" Hg (0.25) in both shells (0.25)

(any 5 0 0.5 each) (2.5) i l

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COMMENT 3.06:

The question initially asks for things which could cause steam dump to respond, then parenthetically asks for controls or conditions to check or verify.

Another condition required for steam dump to actuate is that steam dump is armed. Therefore, another alternative answer is:

Turbine impulse pressure rate of change insufficient to activate loss of load interlock, i.e. . steam dump did not arm (setpoint is variable based on 120 sec time constant).

Also condenser steam dump permissive reset is 15" Hg and trip (permit for dump) is > 16" Hg according to 47014-0205. Setpoint listing indicates 15" Hg stops steam dump and 16.5" Hg permits steam dump. A range of values should be permitted for this response.

Also, Item number 4 should include 2/3 auto stop oil pressures less than 50 psi (Ref. NF 40777-2) or 2/2 stop valves.

NRC Response:

The answer key was modified to accept the recommended alternative answers and the recommended range of values.

QUESTION 3.12 (3.00)

The Engineered Safeguards System is capable.of initiating a number of 1 different signals. Each signal is a response to a different set of abnormal plant conditions and initiates a number of different actions. For example: A safety injection (S) signal results in a reactor trip and provides emergency makeup water to the reactor coolant system. LIST SIX (6) additional signals initiated by the Engineered Safeguards System (INCLUDE WHAT THE SIGNAL PREVENTS, MINIMIZES, REDUCES OR ENSURES) (3.0) l 1

ANSWER 3.12 (3.00) l

1. Containment isolation (CI) signal (0.25) which prevents the release of !

fission products to the environment (0.25)

2. Containment Ventilation isolation signal (0.25) which prevent the release of fission products to the environment (0.25)

3. Control room ventilation isolation signal (0.25) which ensures the habitability of the control room (0.25)

4. Steamline isolation signal (0.25) which prevents a continuous uncontrolled blowdown of more than one steam generator (0.25) ,

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5. Feedwater isciation signal (0.25) which minimizes the feedwater effects on a cooldown (0.25) a l j 5

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6. Containment spray (P) signal-(0.25) which reduces containment pressure, (temperature and airborne fission product level) (0.25)

7. Auxiliary feedwater pump start signal (0.25) which en tres a feedwater supply is available (0.25) any 6 @ 0.5 each) (3.0)

COMMENT 3.12:

The wording of the question is such that it ic not clear that the expected response was those listed in the answer key since an "S" signal generates many of the signals listed. If the question was interpreted to mean what other engineered safety features are initiated by the "S" signal, then additional answers are:

Diesel generators start and load rejection restoration rcquencing start.

Control Room chillers start.

Shield Building Ventilation System starts.

The Auxiliary Building Special Ventilation starts.

The condensate and feedwater pumps trip.

The containment fan co 's shift to slow speed and dampers shift to the dome position.

Component cooling water pumps start.

Diesel cooling water pumps start.

121 cooling water pump discharge aligns to the affected unit.

NRC Response:

4 The answer key was modified to accept the additional answers.

QUESTION 4.12 (1.70)

Plant Operations Manual, Section F2, Radiation Safety, specifies eight (8) radiation protection responsibilities individual's shall have. For the following radiation protection responsibilities for individuals, FILL IN THE MISSING WOR 0(S).

A. Read carefully and observe all requirements as spelled out on applicable

[THREE (3) WORDS] (0.5)

B. Take special care to assure he is free of detectable contamination before leaving the (1) [0NE (1) WORD] area and the (2) [0NE (1) WORD] area. (0.4)

C. Eliminate unnecessary exposure by refraining from lingering in (1) [0NE (1) WORD) fields, by maximizing (2) [0NE (1) WORL; from radiation source, and by utilizing existing (3) [0NE (1) WORD). (0.6)

D. Attempt to keep exposure [0NE (1) ACRONYM] by reviewing work procedures and conducting as much preliminary work outside radir'. ion areas as feasible. (0 1

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ANSWER 4.12 (1.70)

A. Radiation Work Permits (RWPs) (0.5)

8. (1) Controlled (Protected) (0.2)

(2) Protected (Controlled) (0.2)

C. (1) radiation (0.2)

(2) distance (0.2)

(3) shielding (0.2)

D. ALARA Comment 4.12:

Part C is not written according to NUREG1BR-0122 Guidance 4.5.2. Three blanks are used and the guidance is no more than two:

4.5.2 Completion (' fill in the blank") items Completion items require the candidate to supply a word, symbol, or short phrase in response to a question or to complete a statement. The items are most appropriate for testing formulas, numerical solutions, or the recall of simple facts.

Completion items should be constructed according to the following guidelines:

(1) The statement should be complete enough so that (a) there can be no doubt as to its meaning; and (b) there is only one correct answer.

Poor: Saturation conditions are being approached in the reactor coolant system as indicated by .

Better: One indication from the control room that saturation conditions are being approached in the reactor coolant system i is .

1 (2) Avoid the use of too many blanks. Use no more than two in any question.

NRC Response:

1 Comment noted.

QUESTION 4.15 (2.00)

Prairie Island Units 1 and 2 Technical S ? (cifications specifies two (2) safety limits:

1. Safety Limit, Reactor Core, and

2. Safety Limit, Reactor Coolant System Pressure.

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A. What is the objective of the Reactor Core Safety Limit? (STATE THE TS OBJECTIVE / PURPOSE of the Reactor Core Safety Limit) (0.b)

B. What are the parameters that apply to the Reactor Core Safety Limit?

(LIST THREL PARAMETERS SPECIFIED IN T5 as applying to the Reactor Core Safety Limit)

(0.75)

C. What is the objective of the Reactor Coolant System Pressure Safety Limit? (STATE THE TS OBJECTIVE /PURPdSE of the Reactor Coolant System Pressure Safety Liinit)

(0.5) l D. What is the TS specification for the Reactor Coolant System Pressure Safety Limit? (STATE THE TS PRESSURE LIMIT) (0.25)

ANSWER 4.15 (2.00)

A. To maintain the integrity of the fuel cladding (0.5)

B. 1. Thermal power (0.25)

2. Reactor Coolant System pressure (0.25)

3. Reactor Coolant System temperature (0.25)

C. To maintain the integrity of the Reactor Coolant System (0.5)

D. RCS pressure < 2735 psig (0.25)

COMMENT 4.15:

Answer for Part B should also accept Flow and AT as possible answers.

The safety limit is based upon a pre-assumed flow and AT as stated on Figure TS 2.1-1.

NRC Response:

The answer key was modified to accept flow and delta T as possible answers.

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SENIOR REACTOR 09ERATOR'S EXAMINATION QUESTION 5.03 (1.00)

Which of the following stateinents concerning fission product poisoning effects. <

is TRUE? (SELECT THE CORRECT STATEMENT)

14. Xenon is produced directly from fission and neutron absorption but .

Samarium produced from fission only.

.B. Xenon is removed by decay and neutron absorption but Samarium is removed by neutron 3bsorption only.

C. Equilibriua xenon is indepel. dent of power level but equilibrium Samarium is dependent on power level. -

D. Xenon equilibrium is reached in about 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months following a power transient but samarium equilibrium is reached in about 200 hours0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months following a power transient. ,

ANSWER 5.03 (1.00)

B. (Verion is removed by decay and neutron absorption but Samarium is removed 4 by neutron absorption only)

COMMENT 5.03:

Answer (D) is false due to a technicality. This is not in accordance with the l Examiner's Handbook Section 4.6.2(2), and could mislead the candidate. In fact, there is only a difference of - 50 pcm reactivity between 200 hours0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months and 400 hours0.00463 days 0.111 hours 6.613757e-4 weeks 1.522e-4 months .

Answer (0) should also be accepted.

[ NRC Response:

1 As indicated in the reference material provided, it is clear that Samarium equilibrium is reached only after 400 hours0.00463 days 0.111 hours 6.613757e-4 weeks 1.522e-4 months . The answer key will not be ,

modified. '

QUESTION 5.17 (2.25) (Clarified Part b to all candidates)

a. At what axial location in an o[erating PWR core is the Departure from Nucleate Boiling Ration at its lowest? (Limit your answer to TOP, MIDDLE, or BOTTOM) (0.75)

b. How does the MINIMUM critical neat flux va'ue change as the following 7

parameters are INCREASE 0? (Limit your answer to INCREASE, DECREASE, or NO CHANGE. Consider each case independently.) (1.5)

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1. Tave

2. RCS pressure

3. RCS flow ANSWER 5.17 (2.25)

a. Top (0.75)

b. 1. Decrease

2. Increase

3. Increase (0.5 pts each)

COMMENT 5.17:

The term "minimum critical heat flux" is not used at Prairie Island or in the -

section of Westinghouse text cited as a reference. The term "critical heat flux" is defined, and then compared to actual local heat flux. Finally, DN8R is developed and defined. The effects of varying Tave, RCS pressure, and RCS flow on DNDR are then described.

If a candidate correctly described the effects on DNBR, then the candidate has demonstrated understanding of DNB and the effect on the plant. Full credit should be given. If the candidate was not able to correlate DNRB to minimum critical heat flux, no credit should be taken off.

NRC Response:

The examiners clarified the question to the candidates during the exam.

However, if the candidate damonstrates understanding of DNB and its relationship to critical heat flux, full redit will be awarded.

QUESTION 7.12 (2.00)

As a result of a large break loss of coolant accident, adverse containment conditions have been declared.

a. Explain briefly why adverse containment conditions require the use of alternate instrument setpoints while implementing Prairie Island Emergency Procedures? (1.0)

b. Assuming adverse containment was declared initially due to containment pressure, what conditions must be satisfied prior to using normal containment instrument setpoints? (1.0)

ANSWER 7.12 (2.00)

a. Accuracy of instrumentation can be adversely affected (0.5) by temperature (0.25) and/or pressure (0.25).

b. Engineering evaluation should be performed. (0.33) 10

Containment pressure less than. 5 psig. (0.33)

Containment radiation less than 10 E4 R/hr (0.33)

COMMENT 7.12:

Question and answer b:

The question clearly states that high containment pressure was the reason for declaring adverse containment conditions, and therefore full credit should be ,

given if the only response was to say " . . . containment pressure less than 5 psig . . ."

Containment radiation levels need not be discussed to receive full credit.

Containment radiation would only need to be addressedfif the setpoint has been exceeded, which would not necessarily occur during the LOCA.

NRC Response:

The question clectly states that adverse conditions were declared due to a large break LOCA (not a steam line or feed line break). Therefore, it is reasonable to assume that the containment radiation levels would increase. In addition, the question also states that adverse containment was declared initially due to pressure (implying containment radiation could be exceeded).

Therefore, containment radiation is a required answer as it is reasonable that it would oe exceeded shortly after a 1&rge break LOCA and the setpoint is likely to be exceeded prior to pressure decreasing below its limit.

With regards to requirina an "Engineering evaluation should be performed," ,

an acceptable alternative would be a discussion of integrated raafation dose.

QUESTION 7.14 (2.25)

Answer the following regarding the Cooling water system (suction)

a. What are two situations in which the emergency cooling water line ;

is used? (1,0) ,

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b. What action (s) are required to use the emergency cooling water line? (0.5)

c. What precautions should be used prior to using the emergency ,

cooling water line? (0.75) l ANSWER 7.14 (2.25)

a. High river water temperature. (0.5)

Loss of normal cooling water suction. (0.5)

b. Manually close the emergency cooling water gates. (0.5)

c. The emergency cooling water line should be backflushed. (0.75) 11 l

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COMMENT 7.14 Question and answer a:

Question is requiring two situations when only one is addressed in the procedure (loss of normal suction). Although it has been used during high river temperature conditions, it is not addressed in any p.roredure, and therefore should not be required as part of the answer. The use during high temperature was an attempt to solve a high temperature situatton in the cooling water system.

Questien and answer c:

The question should be eliminated because there is no written procedure that statc3 this as a precaution prior to using the emergency suction lines.

Attached is a ccpy of C35 which dictates operation of the cooling water system. The NRC inspection report ti.at if raferenced as the source of the answer is incorrect in stating "Procedures . . . have been clarified to require prior backflushing". A surveillance procedure was clarified that permits backflushing if the pipe is to be used.

NRC Response:

a. The use of the emergency cooling water line by Prairie Island to solve high temperature problems in the cooling water system in 1987 resulted in significant problems in safety systems that relied on the cooling system. The significance of the 1987 event and the fact that this mode of operation is not described in the procedure makes it more important that the candidate know about it,

b. Failure to backflush the energency cooling water line as indicated above ,

resulted in significant problems in the system for cooling. Additionally, communications with the Senior Resident inspector who discussed this issue with the Plant Manager indicated that the referenced inspection report was correct. For these reasons the question is valid and will remain on the exam.

QUESTION 8.09 (1.50)

Smoke from MCC 2AC activated the firc alarm in Fire Detection Zone #35.

EXPLAIN why the "Special Instructions" on F-5 Appendix A (Attachment 8.1),

state that temporary ventilation may be required.

ANSWER 8.09 (1.50) l l

If/When the fire dampers close, hydrogen (0.75) (which is produced by the batteries) may accumulate and create an explosive hazard. (0.75)

COMMENT B.09:

The need for "temporary ventilation" is not needed solely for the purposes of l removing Hydrogen gas.

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9 A;though the answer provided has the potential of being correct, additional reasons for using temporary ventilation still exist in every fire situation in the battery roomt.

The question solicits a response as to why tempora.ry ventilation may be required. The correct answer to this question is simply that "The fire dampers may have closed (actuated) causing a loss in the normal ventilation to the room."

There is no reference in the c;uestion which asks why ventilation is required.

Only why "temporary" ventilation is required. If a basis for ventilation is required for full credit, then other plausible answers shou?d also be accepted for full credit in addition to the Hydrogen removal answer provided in the answer key. Other acceptable answers would be:

Removal of heat Removal of smoke Removal of fire gases NRC Response:

Do not concur. The question asks the candidate to EXPLAIN WHY temporary ventilation may be required. This elicits a more detailed response than the facility suggested answer "the fire dampers may have closed," which is simply a reiteration of the Special Instructions on the exam handout. The other alternate answers (removal of heat, smoke, and fire gases) suggested in the facility comment are plausible, however, they are not significant with respect to the candidates knowing the batteries are a source of hydrogen.

However, partial credit will be awarded for these responses to a maximum of 0.2 points, f

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

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_EBalBIE_ISLeUQ_182______

I "*;s ,. ,4r 'g REACTOR TYPE: _EWB-WEQ2________________

S ($

s, 's;. s .: g

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DATE ADMINISTERED: _SQlDZ/11________________

' '"*" EXAMINER: _BaBEt_St___________ ____

CANDIDATE: _________________________

IUSIBUCIIQUS_IQ.CeUDIQeIE1 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 lacst 80%. Examination papers will be picked up six (6) hours after the examination starts.

% OF CATEGORY % OF CANDIDATE'S CATEGORY

__VeLUE_ _IDIeL ___SQQBE___ YaLUE__ ______________CaIEQQBl_____________

_25tZ5__ _25tk5 ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS

_25 2D__ _25t20 ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION

_2e1k5 _ _2fthh ___________ ________ 7. PROCEDURES - N . MAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL

_2e ZQ__ _2dtkQ ___________ ________ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS l 1001$0-- ___________ ________% Totals I Final Grade l l

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

Candidate's Signature

/ tau atCTCnCups, n r .

- NnC 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 '

end could r e s u l.t in more severe penalties.

2. Rostroom trips are to be limited and only one candid-te at a time may loave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance >r possibility of cheating.

3. Use black ink or dar k pencil QDly to facilitata 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 gggb section of the answer sheet.

B. Consecutively number each answer shee+, write "End of Category __ as appropriate, stort each category on a ogy tage, write QD1Y QD QDR Eidt of the paper, and write "Last page" on the last answer cheet.

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

10. Skip at least ibres lines between each answer.

11. Soperate 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 liitt31yte.

13. Thw point value for each question is indicated in parentheses aftor 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. Ther6 fore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANX.

16. If parts of 't h e examination are not clear as to intent, ask questions of the gE301Dgt 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 dona efter tha avsmination 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.

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

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

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

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

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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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PAGE 2 5t__IBEQBl_QE_NUGLEeB_EQWEB_EL8HI_QEEB6110Ni_ELVIQ1t_&NQ IBEBdQQ1NedIGH QUESTION 5.01 (1.00)

What is the expected count rate response resulting from a.brief rod withdrawal during the approach to criticality? (SELECT THE CORRECT COUNT RATE RESPOI4SE) (1.0)

A. An immediate rapid rise continuing to criticality 8, An immediate rapid rise followed by a gradual increase to a higher steady state value ;

C. A gradual increase followed by a rapid decrease when rod withdrawal is stopped D. A gradual increase continuing to criticality QUESTION 5.02 (1.00)

The DOMINATING EFFECTS on K eff by control rods are the changes in' [ SELECT THE CORRECT CHANGES)

(1.0) .

A. Resonance escape probability (p) and non-leakage probabilities (Pf and Pth)

B. Thermal utilization (f) and non-leakage probabilities (Pf and ,

1 Pth)

C. Resonance escape probability (p) and thermal utilization (f)

D. Resonance escape p-abability (p), thermal utilization (f) and !

I non-leakage probabilities (Pf and Pth)

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, IMEBdQQ1NedIGS QUESTION 5.03 (1.00)

Which of the following statements concerning fission product poisoning offects is TRUE? [ SELECT THE CORRECT STATEMENT)

[1.0]

A. Xenon is produced directly from fission and neutron absorption but Samarium produced from fission only B. Xenon is removed by decay and neutron absorption but Samarium is removed by neutron absorption only C. Equilibrium xenon is independent of power level but equilibrium Samarium is dependent on power level -

D. Xenon equilibrium is reached i n about 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months following a power transient but samarium equilibrium is reached in about 200 hours0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months ;

following a power transient i

QUESTION 5.04 (1.003 If reactor power is increased at a constant 1.0 dom ctertup rate up to the point of adding heet, which one of the following responses will be the INITIAL indication that the point of adding heat has been reached? [ SELECT THE CORRECT INITIAL INDICATION OF REACHING THC POINT OF ADDING HEAT) (1.0) ,

I A. RCS temperature increases

8. Pressurizer level increases C. Reactor power stabilizes D. Startup rate decreenes

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. IBEBdQQ1NedIQ1 QUESTION 5.05 (1.00)

How will pressurizer level indication compare to actual pressurizer level if containment temperature is inc; eased (steam leak) from 100 degrees F to 180 degrees F? (Consider only rC'erence leg heating effects) (SELECT THE

. CORRECT COMPARISON) (1.0)

A. Indicated pressurizer level will be higher than actual pressurizer level B. Indicated pressurizer level will be lower than actual pressurizer level C. Indicated pressurizer level will be equal to actual pressurizer level D. Indicated pressurizer level will fall at is QUESTION 5.06 (2.00)

A reactor at BOC is critical at 10E(-9) amps. Rods are withdrawn at 48 stops per minute for 10 seconds in bank sequenet. :

a. Calculate the SUR immediately AFTER rod motion stops. Assume differential rod worth is 10 pcm/ step, effective delayed neutron fraction is 0.006 and the effective precursor decay constant is 0.1 see-l. (1.25)

b. Compared to part (a) above, HOW and WHY would the $UR change if the initial conditions were at EGC? No alculations are necessary. (0.75) l l

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IBEBdQQ1 God 101 QUESTION 5.07 (2.50)

a. What is the subcooling margin of the plant if the fol10 wing conditions exist: (Show all CJ1culations) (1.00)

Thot 2 568 F Tavg = 553 F Teold = 539 F Ppar = 2235 psig Psg = 810 psig

b. If power is raised from 50% to 100%, how vill the subcooling margin change (increase, decrease, or remain the same)? Justify. ( Assume all control systems in auto) (0.75)

c. Which one of the following conditions would result in the smailest subcooling margin? Assume RCS pressure is th6 same for all cases.

Justify. (0.75)

1. Controlled natural circulation cooldown immedittely following a reactor trip from loss of RCS flow.

2. Continued operation at 5% power.

3. All RCPs era operating at normal no-load temperature after on extended shutdown.

QUESTION 5.08 (1.00)

When the flow rate through a centrifugal pump decreases, ovailable NPSH (increases or decreases) and required NPSH (increases or decreases). Choose correct responses. (1.0)

QUESTION 5.09 (1.00)

The Reactor Protection System would become unreliable for LNB protection from the OT delta T trip if voids were allowed tu form in tti Reactor 1 Coolant System because: (Choose correct answer) l

a. The thermocouples respond much slower in voids as compared to a subcooled fluid.

b. The voids in the downcomer region will cause the NI detectors to j indicate in error. l

c. The heat transfer coefficient of the cladding is reduced !

significantly. l

d. Reactor power is no longer proportional to delta T as measured by the loop RTDs.

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r-PAGE 6 542_IBEQBl_QE_NUGLE88.EQWEB_EL8NI_QEEB8I1QNt ELU101t_8NQ-IBEBdQQ1NedlGS QUESTION 5.10 (1.50)

The plant is starting up with pressurizer pressure at 985.psig and RCS Teve at 420 F. A pressurizer PORV begins leaking to the pressurizer relief tank which is at 5 psig.

o What is the downstream tail pipe temperature? C0.5)

b. What is tne enthalpy of the fluid entering the PRT? (0,5)

c. The fluid entering the PRT is a: (SELECT ONE) (0.5)

1. Subcooled liquid

2. Wet vapor

3. Saturated vapor 4, Suparheated vapor QUESTION 5.11 (1.50)

Tho reactor is producing 65% rated thermal pow.tr at a core delta T of 36 degrees and a mass flow rate of 100% when a blackout occurs.

Notural circulation is established and core delta T goes to 20 degrees. If decay heat is 2%, what is the core mass flow rate (in %)?

QUEST!'ON 5.12 (1.50)

During power operation, control rod insertion limits are in force.

Explain the three possible consequences of NOT maintaining rod insertion limits.

QUESTION 5.13 (1.50)

During refueling operations, you, as the SRO, are given the following information:

Boron sample on the refueling canal was 2027 ppm and the total reactivity in the reactor vessel is -5500 pcm after the last fuel bundle was loaded.

Give a brief expl3 nation and reasons as to why the fuel movement should or chould not continue. (Show all work.)

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QUESTION 5.14 (1.00)

Which one of the following occurs over core life, i.e, from BOC to EOC7

o. Axial power distribution flattens

b. Effective fuel temperature increases

c. MTC becomes less negative

d. Reactor response time increases QUESTION 5.15 (1.00)

As moderator temperature decreases, control rod worth for the controlling bank ________

(INCREASES, DECREASES, or REMAINS THE SAME)

QUESTION 5.16 (2.00)

Answer the following concerning xenon:

a. List two methods of xenon removel. (0.5)

b. Describe how and why xenon reactivity varies after a 2 5's powe r increase to 100% power (equilibrium conditions). -(1.0)

c. At 100% power, the reactor trips due to a mistake by an ISC l technician. Criticality is planned 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after the trip. If 1 the startup is delayed for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , how and why, is the Estimated Critical Position affected? (0.5) 1 l

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PAGE 8 5t$_IHEQBX_QE_UUCLE68EQWEB.EL8HIQEEBeI10Nt_ELW1 dst _88Q IBEBdQQ188d101 QUESTION 5.17 (2.25)

o. At what axial location in an operating PWR core is the Departure From Nucleate Boiling Ratio at its Icwowt? (timit your answer to TOP, MIDDLE, or BOTTOM) (0.75)

. How does the MINIMUM critical heat flux value change as the following paremeters are INCREASED? (Limit your answer to INCREASE, DECREASE, or NO CHANGE. Consider each case independently.) (1.5)

1. Teve

2. RCS pressure

3. RCS flow QUESTION 5.18 (1.00)

(True or False)

The Delayed Neutron Importance Factor at Prairie Island is less than one (1) because delayed neutrons are less likely to leak from the ,

core.

QUESTION 5.19 (1.00)

During normal plant operations, WHEN does the reactor vessel experience tho highest stresses and WHAT TWO (2) primary parameters can ba controlled to limit these stresses.

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QUESTION 6.01 (3.25)

Anawer the follow 4ng regarding the Chemical cnd Volume Control System.

1. What system design feature reduces pressure oscilletions on a Charging Pump start? (0.53

2. What are the interlocks associated with the letdown orifice isolatibn valves and what are these interlocks designed to protect / prevent against? (1.0)

3. What are the advantages (chemically) of maintaining a pressurized Hydrogen cover gas in the VCT? (1.0)

4. What is the minimum Hydrogen pressure in the VCT and why is there a minimum Hydrogen pressure? (0.75)

QUESTION 6.02 (1.80) 1 Tho plant is operating at 100% power with Feedwater Regulating valve FCV-466 in AUTO. For each of the following conditions, indicate the initial direction of travel (answering either OPEN or CLOSE) for FCV-466.

s. Controlling S/G level transmitter fails low

b. Controlling G/G feed flow transmitter fails high

c. Safaty Injection Actuation

d. Controlling S/G pressure transmitter fails low

c. Controlling S/G steam flow transmitter fails high

f. S/G program level fails high concurrent with a Mi Hi S/G 1evel signal

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QUESTION 6.03 (2.50)

Answer the following concerning the nuclear instrumentation system:

e. Explain the response of the reactor protection system upon a s'multaneous failure of both IR channels (where both channels ov.t puts go to zero) for each of the following two cases.

Briefly explain the differences, if any.

1. The recetor is at 5% power during a startup.

2. The reactor is at 20% steady state power. ( 1. T)

b. Assume that while at 100% power one IR channol fails HIGH.

If a recctor trip occurs while in this condit ion, what additional action must be taken during the emergency procedures to ensure the proper operation of the N1 system? (0.9)

QUESTION 6.04 (2.50)

The plant is in Hot Standby:

e. Under what c o n dit io ns , if any, could e Reactor 04olant Pump (RCP) be operated without seal injection? (1.0)

b. What effect, if any, would a containment isolation signal have on the flow of water past the Number 2 seal on a RCP7 (0.5)

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c. What TWO conditions might result in a low level alarm on a RCP

. standpipe? (1.0)

QUESTION 6.05 (2.25)

a. List All. Permissives/ conditions required to arm the Steam Dump System (CDS) with the Mode Select Switch in Tavg Control with the plant operating at 100% power. Inciude setpoints. (1.0)

6. Assume the SOS has been armed as described in part a. due to prior testing r.f related i nst r ument s t io.T . The plant is operating at 50 percent power with all cUntrol systems in automatic. How will the SDS respond to a Channel II inve instrument failure to 569 degrees ;

Fahrenheit? Assume no reactor trip. ,

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QUESTICN 6.06 (1.00),

TRUE or FALSE.

Tho component cooling water heat exchanges outlet temperature should not normally exceed 105 degrees F. However, a maximum temperature of 120 degrees F for a two hour period is acceptable when the RH system is operating for o unit cooldown.

QUESTION 6 97 12.50)

A. How are the RHP hot leg supply isolation valves, MV-32164 and MV-32165, interlocked? COESCRtDE TWO INTERLOCKS / AUTO FEATURES &

INCLUDE SETPOINTS)

(1.0) ,

B. Why is the motor-operator breaker for the RHR hot leg supply isolation valve, MV-32164, opened during normal sower operation? [ EXPLAIN WHAT IS PREi!ENTED AND WHY IT MUST 3E FREyhWT ED) (0.5)

C. How are the RHR heat exchanger to SI pump suction volves interlocked?

(DESCRIBE TWO INTERLOCKS 8 INCLUDE SETPOINTS) (1.0)

QUESTION 6.08 (1.00)

If suctioneered T-avg was high due to e failed temperature instrument, how is pressurizer level affected?

QUESTION 6.09 (1.50)

Answer the following concerning the Overtemperature d e lt a-1 function;

a. What TWO control ections are provided by the OT delta-T function?

(1.0)

b. What is the setoo\r.1 et which these occur? (0.5) 4

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hi _EL8HI 1IIIEdi_DE110Nt_QQNIB9Lt eUQ Id1IBWMENI8IlQN PAGE 12 QUESTION 6.10 (2.50)

Answer EACH of the following questions concerning the Rod Position Indication System:

a. Is it possible to have a Control Bank D rod stuck on the bottom without the "ROD BOTTOM / ROD OROP" alarm energized with the Control Bank 0 bank demand position at 31 steps? (Explain) (1.5)

b. Could you use the rod bottom indicating light? to detect the stuck rod in part a? (Explain) (1.0)

B QUESTION 6.11 (2.00)

An0wer the following cuestions concerning the Radiation Monitoring System.

o. What are six of the seven channels in the Radiation Monitoring system which are designed to remain functional following a LOCA. (1.50)

b. What two automatic functions occur when a high radiation alarm is received on Unit 2 Auxiliary Building Vent Gas Monitor (AR-37). (0.50) )

QUESTION 6.12 (1.00)

Initial plant conditions are: Plant at 100% power all control systems are in automatic. Pressurizer level selector switch is in its normal l position (2/3). EXPLAIN what AUTOMATIC actions would occur if a leak occurred in the reference leg capillary of LY-428 (Blue).

1 4

QUESTION 6.13 (1.50) I i

On May 28, 1987, Unit 1 was in its power ascension program following the Cycle 11-12 refueling.

1

a. Explain why at an indicated NIS power level of 34.6% that calorimetric data showed that actual pcwer was 45.7% ?

l b. Why was this event significant?

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- PAGE 13 88DIQLQQ198L GP91BQL QUESTION 7.01 (2.25)

Ancwer the following regarding procedure C3, Reactor Coolant Pump.

1. Why are only two successive RCP pump start attempts allowed? (0.50)

2. RCS pressure is at 325 psig with RCS temperwture at 250 degrees F and Steam Generator temperature at 180 degrees F. Explain why it is/is not advisable to start'the RCPs. (Assume Pressurizer steam bubble has been established) (1.0)

3. The reactor is critical in the intermediate range (prior to the point of adding heat) and an RCP trips. One hoar later troubleshooting reveals the RCP trip was caused inadvertently due to an instrument technicians mistake. Explain why the RCP may or may not be restarted.

(0.75)

QUESTION 7.02 (2.00)

Anower the following in accordance with E-2 "Faulted Steam Generator Isolation":

a. How is a faulted Steam Generator identified? (2 symptoms) (0.75)

b. List five actions taken to isolate a faulted steam generator. (1.25)

QUESTION 7.03 (1.00)

A caution statement in Procedure ECA-3.1 "SGTR with loss of reactor coolant - subcooled recovery desired" states that voiding may occur in the RCS during RCS deprescurization.

a. Why does this voiding occur? (0.50)

b. What would be the indication (s) that voiding occurred in the RCS?

(0.50)

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-QUESTION 7.04 (1.50)

Tho Ecergency Plan states that in the event of a major radiation emergency, PI administtative dose limits may be exceeded if authorized by the Emergency Director. 'Jhat are the PI normal administrative dose limits for at

a. Quarter (whole body)

b. Year-(ahole body)

If these administrative dose limits were suspended, what are the emergency dose limits when

e. Emergency action is required. (not lifesaving)

d. You volunteered in a lifesaving situation.

QUESTION 7.05 (1.00)

(TRUE or FALSE)

When the Residual Heat Removal System is in operation, the component cooling water supply to the residual removal heat exchangers should be throttled to regulate temperatures in the RCS.

3 QUESTION 7.06 (1.00)

, (TRUE or FALSE)

Functional Restoration Procedures may not be entered while in ECA-0.0 even if a Red Peth on Suberiticallity exists.

QUESTION 7 07 (2.40)

In accordance with the Optimal Recovery Procedures, state FOUR conditions when the Reactor Coolant Pump trip criteria is not applicable. ,

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. B8Q10LQQ1G6L_CQUIBQL QUESTION 7.08 ( .50)

A. Units 1 & 2 are opersting at 99% power. A Unit i reactor trip and cafety inj ection occurs. As shift supervisor you direct the immediate actione and subsequent actions of your reactor operator. The reactor operator informs you of the following indications:

a. Power range < 5%

b. C o nt a i nrae nt pressure = 18 psig

c. Pressurizer level < 15%

d. Core exit TCs = 895F

e. RVLIS full range = 31%

f. RCPs off

g. Both CG wide range levels indicating 52%

h. RCS subcooling based on core exit TCs = 15F Which Critical Safety Function is in a RED CONDITION? [ STATE the CRITICAL '

SAFETY FUNCTION)

Part B deleted B. Unit 1 is operating at 100% power. A unit 2 startup is in progress with power atabilized at approximately 10 E-8 emps. Indications on both units are normal for the existing conditions. What is the Critical Safety Function Status Tree indications or conditions that exist for these oporating conditions? (SELECT THE CORRECT ANSWER)

1. Both units subcriticellity - RED

2. Unit i subertticallity - RED

3. Unit 2 suberiticellity - RED

4. Both units suberiticallity - ORANGE l l

QUESTION 7.09 (2.00)

a. Withdrawing the shutdown banks is administrative 1y controlled in the startup procedure (C1.2). State the two plant conditions that may exempt the shutdown banks from being withdrawn? (1.0)

b. What is the minimum shutdown margin that must be maintained ,

with all shutdown and control banks inserted? (0,5)

)

c. What is the minimum shutdown margin that must be maintained with all shutdown and control banks withdrawn? (0,5)

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PAGE 16 Zu _EBQQEQUBE1_ _NQBdeL&_eEUQBdekt_EdEBQEUQ1_8SQ BeQIQLQElG6L_QQUIBQL QUESTION 7.10 (2.00)

Units 1 and 2 are both operating at 100% power. A relay room fire occurs that causes the Unit i Shift Supervisor to make the decision to evacuate the control room.

A. What is the INITIAL evacuation assignment of the UNIT 1 Lead Plant Equipment and Reactor Operator (LPERO)? (STATE THE PLANT LOCATION OR EQUIPMENT TO PROCEED TO) (0.5)

B. What is the INITIAL evacuation assignment of the UNIT 2 Lead Plant Equipment and Reactor Operator (LPERO)? (STATE THE PLANT LOCATION OR EQUIPMENT TO PROCEED TO) [0.5)

C. What is the INITIAL evacuation assignment of the UNIT 1 Plant Equipment and Reactor Operator (PE&R0]? (STATE THE PLANT LOCATION OR EQUIPMENT TO PROCEED TO) (0.5)

D. What is the INITIAL evacuation assignment of the UNIT 2 Plant Equipment and Reactor Operator (PE&R0]? (STATE THE PLANT LOCATION OR EQUIPMENT TO PROCEED TO) [0.5)

QUESTION 7.11 (2.00)

According to procedure 1ES-0.4, "Natural Circulation Cooldown with Steam Void in Vessel : ,

s. If RVLIS is less than 95%, pressurizer level must oe greater than 90 %. Explain the reason for this high level requirement.

b. If a RCP cannot be restarted, pressurizer level should be between 19% and 23%. Explain the reason for this low level requirement.

QUESTION 7.12 (2.00)

As a result of a large break loss of coolant accident, adverse containment conditions have been declared.

o. Explain briefly why adverse containment conditions require the use of alternate instrument setpoints while implementing Prairie Island Emergency Procedures? [1.0)

b. Assuming adverse containment was declared initially due to containment pressure, what conditions must be satisfied prior to using normal l containment instrument setpoints? [ .'. 0 ) l l

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QUESTION 7.13 (1.50)

Spocial Order PINGP 301 (attachment 7.1) was generated in rasponse to Prcirte Island's investigation regarding NRC Information Notice 87-63.

WHAT is the concern addressed by this Special Order? (BASIS) ,

QUESTION 7.14 (2.25)

Answer the following regarding the Cooling Water System.

a. What are two situations in which the emergency cooling water line is used? [1.0)

b. What a c t i o r. ( s ) are required to use the emergency cooling water line?

[0.5)

c. What precaution (s) should be used prior to using the emergency cooling wat6r line? [0.5)

QUESTION 7.15 (1.25)

A leak has developed in a CVCS Letdown piping component located outside the containment building which must be manually isolated.

Tho general area radiation level in the area where the leak is to be icolated is 600 millirem per hour. The one available person to porform the work informs you that his present quarterly exposure and lifetime exposure levels are 2.90 Rem and 54.75 Rem respectively.

a. Using only 10 CFR 20 whole body exposure limits as a guide, how long may this person work in the area before the quarterly exposure limit 10 exceeded? (SHOW YOUR WORK) (0.75)

b. What is the minimum age that this person may be to perform the work?

(SHOW YOUR CALCULATIONS) . (0.50) l (***** END OF CATEGORY 07 *****)

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PAGE 18 IzS_8QdINISIB6IIVE EBQQEQWBEft_QQuQ1I1QNSt_860_LidIIeI1QNS QUESTION 8.01 (2.25)

In accordance with F-2, "Radiation Safety"

a. What is the administrative requirement which determines if a radiation area needs to be locked? (.75)

b. Who has custody of the key to a locked radiation area? (0,5)

c. What are four local (4) areas in the plant that have the potential for being Very High Radiation Areas? (0.5)

QUESTION 8.02 (2.50)

What are the following reactor trips specifically designed to protect against? (T/S Basis)

a. Power range'high flux-high level t rip. (0.5)

b. Low pressurizer pressure. (0.5)

c. Low RCS flow rate. (0.5)

d. High pressurizer level. (0.5)

o. Low-low steam generator water level. (0.5)

QUESTION 8.03 ( .75)

TRUE or FALSE An individual shall not work or have worked more than a total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, excluding shift turnover.

QUESTION 8.04 (1.50)

On July 3 mid shift (0000 - 0800) pressurizer PORV #1 was declared inoperable. On July 11 at 0800, with the first PORV already inoperable, it ecs established that the second pressurizer PORV could not be operated from the control room. What oo Technical Specifications require? (Include all options)

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PAGE 19 A[._eQUINISIBeIlhE_EBQQEQUBEnt_QQND1IlQNi&_eSQ_L1011&IlQN1 QUESTION 8.05 (1.50)

c. For power operation at 70% power, you exceed i hour of time outside the target exial flux difference band in the last 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months . What actions are required? (1.0) (Assume you remained inside the "doghouse / envelope" throughout this period)

b. If the reactor is at 35% power and it is planned to go to 100% power, what axial flux difference requirements must be met? (0.5)

QUESTION 8.06 (2.50)

During a walkdown of the main control boards prior to relieving the Reactor Oporator, you note that reactor makeup pump #11 "Start / Normal /Stop" control switch is in the "Pullout" position with a Hold Card attached (#11 motor meintenance is in progress). Yesterday you noted that reactor makeup pump

22 "Start / Normal /Stop" control switch was in the "Stop" position with a Secure Card atteched (#21 pump was being tested and oper7 tion of #22 pump could invalidate the #21 pump test).

A. Who is responsible for maintaining the Hold and Secure Card Record Logs? (STATE POSITION) (0.5)

6. What is a Hold Card used fort (STATE THE PURPOSE of a Hold Card)

(0.5)

C. What is a Secure Card used for? (STATE THE PURPOSE of a Secure Card)

(0.5)

D. Is the Hold Card being used prope.rly? (ANSWER YES OR NO and ,

JUSTIFY / EXPLAIN your choice) (1 0)

I QUESTION 8.07 (1.50) l Answer the following regarding SWI-0-3, "Safeguard Hold Cards and Component Locking or Blocking"*

a. Who is responsible for control of the blocking and locking devices? l (0.5) l

b. DEFINE a locking device and EXPLAIN the difference between a locking i device and a blocking device. (1.0) l

(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)

1

PAGE 20 I Ai _eQdI'11IB6IlyE_EBQQEQUBEf d t_GQUQ1Il0Ni&_6UQ_L151I6I1QN1 i

QUESTION 8.08 (1.70) j Answer the following questions per procedure SWI-0-4, "Operations Section i Work Management"

a. (TRUE or FALSF.)

ALL data sheets, including periodic equipment logs SH...L be checked by the 2300 - 0700 Shift Supervisor for completeness and routed to the General Superintendent. (0.5)

b. Six hand written logs are maintained by Operations personnel. What are these six logs? (LIST) (1.2) ;

QUESTION 8.09 (1.50)

Smoke from MCC 2AC activated the fire alarm in Fire Detection Zone #35.

EXPLAIN why the "Special Instructions" on F-5 Appendix A (attachment 8.1),

otote that temporary ventilation may be required.

QUESTION 8.10 (1.25)

With the RCS temperature less than the MPT (Technical Specification 3.1-G), t under WHAT TWO CONDITIONS may both SI pumps control switches be in auto?

QUESTION 8.11 (1.50)

Technical Specification 3.8.B.5 limits the number of recently discharged fuel assemblies in Fuel Pool Number One to 45.

1. WFy doesn't a similar Technical Speci.fication apply to Fuel Poo!

Number Two? (0.5)

2. What in the basis for this T.S. 11mit? (1.0)

QUESTION 8.12 (1.00)

! Soction 5 of procedure F-2 "Radiation Safety" deals with equipment control requirements. Explain briefly WHAT "White Tags" and "Green Tags" are used for.

(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)

Bt.'_6Db18111BSIIVE_EBQQEQUBEft_Q0W0111081t_8N0 L10118I1Q61 PAGE 21 QUESTION 8.13 ( .50)

Tho RCS has been drained per procedure 0-2 "Draining the Reactor Coolant Syotem" and is being maintained at the centerline of the RCS hot leg.

a. An RCP low level standpipo level alarm occurs. Explain why you should be concerned? :

Part B deleted from exam

b. What operational concerns during this mode of operation governs the flow of the RHR pumps?

QUESTION 8.14 ( .50)

(TRUE or FALSE)

Whon purging the Generator, Hydrogen is purged directly with Air for the purposes of personnel safety QUESTION 8.15 (1.75)

Procedure SAWI 3.10.1 requires that 4.16 KV breakers which are being racked into the connect position shall be verified either by operation or visibly dith prior approval.

a. Who must approve visual verification of breaker position? [0.5)

b. Wh&t are the five steps used to visually verify breaker position?

[1.25) .

b QUESTION 8.16 (1.00) l To exit a vital ar6a, card readers are installed to unlock the door. Under chat three situations is it allowable to turn the knob or use a key to exit a vital ares? ,

l i

QUESTION 8.17 (1.50) i It is discovered today, July 11 day shift (0800 - 1600), that a monthly j curve 111ance due lect July 3 mid shift (0000 - 0800), was not performed.

This surveillance has been performed on time for the past six months. Is this item considered operable? Explain your answer.

i

(***** END OF CATEGORY 08 *****)

(************* END OF 6XAMINATION ***************)

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ATTACHMENT 7.1 i ;

231 PINGP 301 SPECIAL ORDER No.

Rev. 7 Page 1 of

SUBJECT:

Use of Containment spray on Long Tera Aecirculation 1 l EXPIRES: UPon Update of ERC's ,

I i

OPERATIONS COMMITTEE REVIEW REQUIRED: HTG#/DATE 2/18/88 i i

Yes 111 No l[]

DISTRIBLTr10N:

Admin. Services Supv. OC Secretary

, Control Room (1) Supt. Quality Engineering .

Shift Supervisor Supt. Training - PI NRC Resident Inspector CAUTION: This Special Order is only applicable when the ECCS is in the Containment Sump Recirculation mode af ter a large break LOCA.

1) The MR pur:ps shall not be operated in a mode where they are injecting to the vessel of cold leg and supplying a containment >

spray pump.

2) If two RRR pumps are available when.svitching to recirculation, one f pump shall be injecting to the reactor vessel and the other pump shall i be supplying the associated CS pump. The RHR pump supplying the

~

containment spray pump sh 11 have its discharge valve to the reactor vessel closed.

3) If only one RHR pump is available for long term recirculation it shall be injecting into the reactor vessel.

4) It is acceptable to run an RER pump supplying both a CS pump and a SI pump provided a) the RHR discharge valve to the reactor vessel is closed or b) RCS pressure is above the shut of f head ci the RHR pump (140 psig).

- - - - er Prepared By - Dcte M'/8*88 Approved for Issuance N Date J- /b d M lant Man per Approved for Termination -

Date Plant Manager 1RM

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. IBEBdQQ16801Q1 ANSWERS -- PRAIRIE ISLAND 112 -88/07/11-HARE, S.

i ANSWER 5.01 (1.00)

B. (An immediate rapid rise followed by a gradual incre?se to a higher steady state value) [1.0)

REFERENCE Westinghouse, Fundamentals of Nuclear Reactor Physics, 1983, page 8-54 192308K103 ...(KA'S)

ANSWER 5.02 (1.00)

C. (Resonance escape probability (p) and thermal utilization (f)) (1.0)

REFERENCE Westinghouse, Reactor Core Control for Large Pressurized Water Reactors, 1983, page 6-10 192005K107 ...(KA'S)

ANSWER 5.03 (1.00)

8. [ Xenon is removed by decay and neutron absorption but Samarium is removed by neutron absorption only)

REFERENCE Uestinghouse, Reactor Core Control for Large Pressurized Water Reactors, 1983, pages 4-11 through 4-33 192006K103 192006K104 192006K116 192006K117 ...(KA'$)

ANSWER 5.04 (1.00)

D. (Startup rate decreases) (1.0)

REFERENCE Wootinghouse, Reactor Core Control for Large Pressurized Water Reactors, 1983, page 9-17 ;

192008K117 ...(KA's) l l

l 1

1 i

'12*_IBEQBY_QE.NWQLE8E.EQWEB_EL8NI_QEEB8IIQNt_ELWIQ1t_6HD PAGE 23

. IBEBdQQYNed1G1 A'NSWERS -- PRAIRIE ISLAND 142 -88/07/11-HARE, S.

ANSWER 5.05 (1.00)

A. (Indicated pr'ssurizer level will be higher than ac tual pressuriter level] (1.0)

REFERENCE Westinghouse, Thermal-Hydraulic Principles and Applications to she Prcssurized Water Reactor !!, 1982, pages 11-27 &ad 11-28 191002K106 ...(KA'S)

ANSWER 5.06 12 001

c. Lambda = 0.1 enc-1 (0.25)

Beta eff = 0.006 (0.25)

Rho = (lo sec)(1 min /60 sec)(10 pcm/ step)(48 steps / min)

= 80 pcm (0.25)

After rod motion:

SUR = 26(lambda (rho))/(beta eff - rho) (0.75)

= 26(0.1(8E-4))/(0.006-0.0008)

= 0.40 dpm (0.25)

b. At E0C, beta effective is less (0.50) which gives a higher GUR (0.25)

REFERENCE Wootinghouse F.N.R.P. Chapter 7 192003K107 ...(KA'$)

ANSWER 5.07 (2.50)

a. Tsat for 2250 psia (L235 psig) (0.20)

= 652.67 F (0.40)

SCM = Tsat - Thot = 652.67 - 566 = 84.67 F (+/- 2 F) (0.40)

b. SCM will decraese (0.25)

Thot will increase as power increases (0.50)

c. 1 - Natural Circulation Cooldown (0.25)

Core delta-T during NC cooldown will approach full load delta-t (0.50)

12;.JBEQB1_QE_NUGLE68_EQWEB_ELeNI_QEE86I1QNt_ELVIQ1&_8NQ PAGE 24 ISE809Q16851Q1 NSWERS -- PRAIRIE ISLAND 142 -88/07/11-HARE, S.

REFERENCE P.I. plant info summary 002000K509 191004K106 ...(KA'S)

ANSWER 5.08 (1.00)

(0.5 each) increase, decrease REFERENCE Westinghouse T.H. 10-56 >

004000K604 191004K106 ...(KA'S)

ANSWER 5.09 (1.00) i

d. l REFERENCE Wootinghouse Thermal-Hydraulic Principles and Applications to PWR II Chapter 13, pp 53-56 0120000K50 ...(KA'S)

ANSWER 5.10 (1.50) '

(0.5 each)

o. 300 degrees F (295-305 F seceptable)

b. 1193 BTU /LBM (1190-1195 BTU /LBM seceptable)

C. 4 REFERENCE i Wootinghouse Reactor and Pressurizer Thermodynamics Steam Tables 000008E101 0100000A10 ...(KA'S) l, l

n l

1

li _IBEQBl_QE_NWQLE88_EQWEB EL8HI_QEEB8110NA ELUIQ1t_86Q PAGE 25 IBEBdQQ1Ned1G1 ANSWERS -- PRAIRIE ISLAND 112 - d' 0 / L ; ; HARE, S.

ANSWER 5.11 (1.50)

To determine flow in NC:

Q = m cp deltaT => 65 = 100

cp

36 => ep = 65/C100

36) cp = .0181 THEREFORE 2% = m * .0181

20 => m= 2%/(.0181

20) = 5.5%

(1.0 for right method)

REFERENCE Weotinghouse Thermal-Hydraulic Principles and Applications to PWR II Chapter 14, pp 15-29 002000A105 ...(KA'S)

ANSWER 5.12 (1.50)

(.33 each)

1. Potentially inadequate trip reactivity '

2. No assurance that power distribution limits are met

3. Increase the consequences of a rod ejection accident REFERENCE T/S 15.3.10 e 15 001050G006 ...(KA*S)

J ANSWER 5.13 (1.50)

Tho fuel movement may continue (0.5), because Technical Spocification (3.8.A 4) on refueling requires > 2000 ppm boron concentration or Keff < .95 (whichever is most restrictive) during fuel movement. (0.5)

If Total reactivity (p) = -5500 pcm then K= 1 / (1-p) = 1 / (1-C- 0.05500)) = .948 (0.5) ]

ALSO ACCEPT 0.948 is too close to 0.95 to continue to load fuel so fuel handling should stop (0.50) e l

1

?; - 1 iz__IBEQBY_QE_UVQLE88 EQWEB.EL8HI_QEEB8110Nt_ELW1Qft_88Q PAGE 26 IHEBdQQ108dIQS ANSWERS -- PRAIRIE ISLAND 182 -8b/07/11-HARE, S.

REFERENCE Prairie Island Technical Specifications 3.8.A.4 l 192002K112 ...(KA'S)

ANSWER 5.14 (t.00)

a. (1.0)

REFERENCE Westinghouse Reactor Core Control for Large PWR's, pp. 8-21, 2-45, 3-19, 2-73 192004K106 ...(KA'S)

ANSWER 5.15 (1.00)

Docreases.

REFERENCE !

W Reactor Core Control, P. 6-22 001000K520 192005K107 ...(KA'S)

ANSWER 5.16 (2.00) ,

s. (0.25 pts. each)

1. Neutron absorption (burnout)

2. Decay

b. Xenon reactivity initially decreases (0.253 because of increased burnout. (0.25) Production of Xe and I increases due to a higher fission rate. (0.25) This causes Xe reactivity to increase as these isotopes build up. The result is an equilibrium reached after about 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months at a higher xenon concentration. (0.25)

c. Xenon concentration is increasing at a high rate, adding negative reactivity (0.25) so the rods are ut a higher position to achieve criticality. (0.25)

REFERENCE W Reactor Core Control, p. 4-11 through 4-22 192006K102 192006K105 192006K106 ...(KA'S)

- 52_ _IH E Q Bl_ Q E _NW Q L E 88_ E QW E B _ EL 8NI ..Q E E B &Il0N t _ELW lQ 1 t _8NQ PAGE 27 IHEBUQQ188dlG1 ANSWERS -- PRAIRIE !$ LAND 142 -88/07/11-HARE, S.

f ANSWER 5.17 (2.25)

a. Top (0.75)

b. 1. Decrease

2. Increase

3. Increase (0.5 pts each)

REFERENCE Wootinghouse Thermal-Hydraulic Principles, pgs. 13-18, 13-24, 13-34 19300CK106 ...(KA'S)

ANSWER 5.18 (1.00)

False (1.0)

(Less than one because delayed neutrons do not cause fast fission of U-238)

REFERENCE Wostinghouse Fundamentals of Nuclear Reactor Physics, pg. 7-34, and PI Plant Information Summary 192003K107 ...(KA'S) .

F ANSWER 5.19 (1.00)

During plant cooldown (0.5)

Tonperature or cooldown rate (0.25)

Prossure (0.25]

REFERENCE i PI-P8157L-004 K/A 193010K1.07 4.1 193010K107 ...(KA'S) l r

i t

L

6A..EL8NI 1111Edl QEllGNA GQd180LA 8NQ lygIgygg61&I196 PAGE 28 ANSWERS -- PRAIRIE !$ LAND 142 -88/07/11-HARE, S.

ANSWER 6.01 (3.25) ,

1. Desurger (0.5) (on pump discharge piping)

2. Interlocked with letdown isolation valves (0.20) and charging pump breakers (0.20) and "N0 containmennt isolation" (0.20). These interlocks limit the loss of RCS coolant (0.20) if the letdown line ruptures downstream. (0.20) ,

also accept pressurizer level > 14.8% for (0.10)

3. Reduces Oxygen concentration in the RCS (0.5) which reduces corrosion (0.25) and limits formation of Nitric Acid. [0.25]

4. 15 psig (0.375) Provides backpressurc for proper operation of the RCP -

seals (0.375)

REFERENCE P1872L-001A pp 10,11,12,15 C-12 0 6 004000K104 004000K504 004020K403 ...(KA'S) i ANSWER 6.02 (1.80)

A

a. OPEN

b. CLOSE

c. CLOSE

d. CLOSE

e. OPEN

f. CLOSE ,

(0.3 each)

REFERENCE i 1

PI $ystem description B-7, p 38 059000A211 ...(KA'S) l 1

l l

l l !

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ti__EL68I_SISIENS_ DES 19Ni_QQUIGQLi_eWQ.INGIBWHENI&IIQN PAGE 29 ,

ANSWERS -- PRAIRIE ISLAND 1A2 -88/07/11_ HARE, S.

ANSWER 6.03 (2.50)

a. 1. The rtsetor will trip on SR high flux (0.4]

due to P-6 dropping out and reenergizing SR nuclear instruments.(0.4)

2. The reactor will not trip (0.41

. due to P-10 which prevents reenergization of the SR N! even if P-6 drops out. (0.4)

b. The SOURCE RANGE RESET switches must be used (0.5) in order to allow the SR NI to reenergize. (0.43 also accept "Manually unblock source ranges for full credit REFERENCE P.I. Plant information summary 015000A403 015000K406 ...(KA'S) ;

ANSWER 6.04 (2.50) !

a. 1. RCP seal leakage rate is 5 gpm or less (0.50) and at least 195 1

gom of CCW is flowing through the thermal barrier heet exchanger. (0.50) ,

b. Increase flow (since return valve would shut and relief valve to PRT would increase dp across #2 seal) [0.5)

c. (any two at 0.5 each)

Lack of leakage through #2 seal Excessive leakage through #3 seal Leak in standpipe Loss of seal injection RFTERENCE 6-3, Reactor Coolant Pumps, P 25 C-3 RCP Procedure, P2 ASME 84-PVP-80, The Westinghouse RCP Shaft Seal System Evolving i to Improve Availability 003000K108 003000K602 0033000K60 103000A203 ...(KA'S) i

)

l l

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1 t

62_ tLeNI_SYSIJd1_QESIGNt_QQNIBQLt_6NQ_INSIBudENI8IlgN PAGE 30

" WERS -- PRAIRIE ISLAND 112 -88/07/11-HARE, S.

ANSWER 6.05 (2.25)

a. 1. 10 percent loeu decrease (J.26) in < 120 seconds (0.25) also accept equivelent ramp rate to 10 % in 120 seconds

2. At least one circulating water pump running (0.25) with condenser vacuum > 16" Hg (0.25)

3. Teve > 540 F (0.25)

b. Condenser valve (CV-3100) wil) trip open (0.33). Atmospheric valves (CV-31085 & CV-31090) will fuily open (0.33) and second set of atmospheries will modulate open as necessary (0.33)

REFERENCE P.I. P8174L-02, D.I. B-7, P 19 041020G007 041020K105 .. (KA'S)

ANSWER 6.06 (1.00)

Trua.

REFERENCE Prairie Island System Description B-14, p. 26 008030G010 ..,(KA'S)

o . ... . . ..... .. . - - . -

61.LEL881_111IEd1_QE11GNt_QQUIBQL&_88Q_INSIBydENI611QN PAGE 31 ANSWERS -- PRAIRIE ISLAND 1&2 -88/07/11-HARE, S.

ANSWER 6.07 (2.50)

A. 1. Valves can not be opened [0.25] unless RCS pressure < 425 psig (0.25] [0.5]

2. ,en the control switch is in the AUTO position (0.1], the valves close automatically (0.2] if RCS pressure increases to 600 psig (0.2]. [0.5)

B. Prevent an inadvertent valve opening (0.25) which could overpressurize the RHR system (0.25]. [0.5]

[ Full credit {0.5) will be given for "Prevents an incorrect flowpath during emergency core cooling operation"]

C. 1. The valves can not be opened [0.1] unless the associnted RHR pump discharge pressure (0.2] is < 210 psig [0.2].

2. The valves can not be opened (0.1) unless the associated SI pump suction isolation valve (0.2] from the RWST is shut (0.2]. [0.5 REFERENCE System Description B-15, Rev. 2, Para 3.4.A, pages 15 a 16 System Description B-15, Rev. 2, Para 3.4.0, pago 17 005000K407 006000K409 ...(KA'S)

ANSWER 6.08 (1.00)

The level program controller limits reference level to 50% [0.50] so actual level wou1J seek 50%. [0.50] or- levels off at a higher level for (0.50]

REFERENCE PI B-7 Figure B7-19 011000A104 (3.1/3.3) 000028A202 (3.4/3.8) 000028A202 011000A104 ...(KA'S)

ANSWER 6.09 (1.50)

e. Turbine Runback (0.5)

Automatic and Manual Rod withdrawal inhibition (0.5)

(-- 0.2 if both automatic and manual not included)

b. 54 below trip setpoint. (0.5)

EI EL6NI_SISIEdk_QESIGNt_QQUIBQLt_oNQ INSIBudENIallQN PAGE 32 ANSWERS -- PRAIRIE ISLAND 1&2 -88/07/11-HARE, S.

I REFERENCE l PI Plant info Summary P 2 l

I ANSWER 6.10 (2.50)

a. Yes (0.5), the rod bottom bypass bistables will block the Rod l Bottom / Rod Orop alarm (0.5] until the Control Bank D bank demand position reaches 35 stops [0.5]. '

b. Yes (G.5), rod bottom lights function at all times (0.25) and will be on when the rod is below 20 steps (0.25).

REFERENCE PI Rod Control and Rod Position Indication Lesson Plan, 3 66 ,

001000K401 ...(KA'S) i ANSWER 6.11 (2.00)

( any six at 0.25 EACH )

n. 1. R-16 Containment fan-coil

2. R-18 Containment fan-coil

3. H-26 RHR Cubical 11 and 21

4. R-27 RMR Cubical 12 and 22

5. Containment Red. Monitor R-48

6. Containment Accident Area Monitor R-49

7. Shield Buil ding High Stack Monitor R-50

( 0.25 each )

b. Starts 121 Aux Building Spoeial Exhaunt Fan Closes Gas decay tank vent isolation valve REFERENCE PI B-11 pg 20, 38, Table B-11-1 KA 073000K1.01 3.9 072000A3.01 3.1 072000K4.03 3.6 072000A301 072000K403 073000 kid 1 ...(KA'S) l

Et_'_EL88I_111IEdi_QESIGNt_CQUIBQLt_aND_INSIBudENI8IIQN PAGE 33 ANSWERS -_ PRAIRIE ISLAND 182 -88/07/11-HARE, S.

ANSWER 6.12 (1.00)

Auto actions: (0.5 each)

Turn on Back Up heaters Charging pump speed decrease signal REFERENCE KA 011000K1.01 3.9 011000K1.04 3.9 011000K3.01 3.4 011000K5.13 3.4 011000A2.10 3.6 PI B7 pp 15,20 011000A210 011000K101 011000K104 011000K301 011000K513

...(KA'S)

AMSWER 6.13 (1.50)

a. Less than predicted radial neutron leakage. (0.5)

b. The power range high flux low setpoint (0.5] would have tripped the reactor at a power level above the 25 % Technical Specification limit. (0.5)

REFERENCE 015000K402 ...(KA'S) ,

I i

I l

I i

Zt_'_EBQQEQWBEl_ _NQBd8Lt_eBNQBd6Lt_EdEBQENQ1_68Q PAGE 34

, 86010LQQ198L_QQNIBQL ANSWERS -- PRAIRIE ISLAND 1 r,2 -88/07/11-HARE, S.

ANSWER 7.01 (2.25)

1. Prevent meter winding damage (0.25) from excessive heat (0.25)

2. Should not start the RCPs (0.25) until the RCS and S/G temperature are styroximately equal (0.375] to minimize RCS pressure transient (b.375)

3. Cannot be restarted (0.375) bectase of rositive reactivity insertion (0.375)

REFERENCE Procedure C.3 pp 3,4 0030000001 (3.7/3.8) 003000A102 (2.9/2.9) 003000K501 (3.3/3.9) 003000A102 003000G001 003000K501 ...(KA'S)

ANSWER 7.02 (2.00)

c. 1. 56 pressure decreasing in an uncontrolled manner (0.375 each)

2. SG completely depressurized

b. 1. Isolate MFJ flow (0.25 each, max of 5 correct answers)

2. Isolate AFW flow

3. Close steam supply valves to TDAFW pump

4. Verify SG PORVS closed

5. Verify SG B isolation valves closed

6. Check closed or close MSIVS on effected SG REFERENCE PI E-2 000040EA201 (4.7) 000040EK304 (4.7) 00C040A201 000040'<304 ...(KA'S)

ANSWER 7.03 (1.00)

a. The high temperature coolant in the upper head which has a slight amount of cooling flow will flash to steam creating voids in the upper head. (0.50)

b. Rapidly increasing pressurizer level. (0.50) l REFERENCE l 000038K306 (4.2/4.5)

ECA 3.1 p 9 00003BK306 ...(KA'S) l l

L Zz_*_E80GEQU8El_ _NQBd8Lt_8BNQBM8Li_EdESQENQY_8NQ PAGE 35

- 88DIQLQQIQ8L_QQUIBQL AN'SWERS -_ PRAIRIE ISLAND 182 -88/07/11-HARC, S.

ANSWER 7.04 (1.5LJ

a. 1250 mrem

b. 5 Rem

c. 25 Rem

d. 75 Rem (0.375 each)

REFERENCE PI E-Plan PI F-2 o 14 194001K103 (2.8/3.4) 194001K103 ...(KA'S)

ANSWER 7.05 (1.00) 4 l

False.

REFERENCE PI C-15 p 6 005000K402 (3.2/3.5) l 005000K402 ...(KA'S)

I I

ANSWER 7.06 (1.00) I l

True REFERENCE PI Emergency Peacedures, ECA-0.0, p. 2 000055G012 .(KA'S)

. ~, . .

Z [_'_ E B Q Q E RV B E S _: _ N Q B d 6L c _6d dQ B d 6 L i _ E d E B Q E N Q 1_8N Q PAGE 36 BSD10LQQ196L_Q2NIBQL ANSWERS -- PRAIRIE ISLAND,182 -88/07/11-HA3E, S.

1 AN*WER 7.07 (2.40)

Following SI initiated f-om cold SD, HSB, or startup conditions (0.6]

During recovery operatAont following initiation of operator controlled cooldown. (0.6]

Following RCP restart specified in ERG recovery inst r uct ions. (0.6]

Uhen EWG specific criteria do not apply. [0.63 If no SI pumps are running and RCS pressure > 12C0 ps ig (1500 psig) (0.6)

REFERENCE Background Information for Westinghouse Owners group ERG, Generic Issue RCP trip / restart, HP/LP-Rev. 1, Sep. 1383, p 58.

P8197L-012, E-1/E-2 Review; EO-7 Lesson Plan P8197L-012, Question 3.c, p 64 000011K314 ( /4.2) 000074K308 ( /4.2) 000011K314 000074K308 ...(KA'S)

ANSWER 7.08 ( .50)

A. Core Cooling Part B deleted B. 2. (Unit i Suberiticallity - RED)

REFERENCE F-0.2, Core Cooling, Rev.2, Page 1 F-0.1, Suberiticellity, Rev. 1, page 1 000074A207' 194001A113 ...(KA'S) e

-- --. , , . , - , , - - , - - . - , . . . . _ ~ ,,- y .p. , w - - --

Zt__EBDCEQUEE1_=_NQBdakt_ABNQBd&Lt EDEBGENGL.AdQ PAGE 37 B6DIQLQQ1 CAL 99.UIBQL i ANSWERS -- PRAIRIE ISLAND 182 -08/07/11-HARE, .S.

ANSWER 7.09 (2.00)

e. RCS borated to at le ast the cold shutdown concentrati?n (or greater)

[0.50] or borated to the hot xenon free concentration (0.25) and is being maintained at no-load average temperature. [0.25)

6. 3% (0,5)

c. 1%

also accept i to 2 percent for b or e only if technical specification limit Ourve is discussed or referenced ,

REFERENCE P.I. Procedure C1.2 pp 5,6,7 C1A para 5, p 11 001010K506 ...(KA'S)

ANSWER 7.10 (2.00)

A. Hot shutdown panels (0.5)

8. Aux Bldg Elev 695 iMSIVs] [0.5)

C. D1 Diesel Room (0,5)

D. Turbine Front Standard (0.5) also acceppt for C or 7 (not for both) trip reactor (0.1) close msivs (0.1) close pressurizer block valves (0.1) assure turbine trip (0.1)

REFERENCE Plant Operations Manual, Section F5 Appendix B, Control Room Evacuation (Fire) Safe shutdown Procedure, Rev. 4, Para 3.8. and 3.C., page 4 000068G006 ...(KA'S) l l

I l

Z1__RBQQERUBEl_:_NQBbelt_eBNQBdekt_EUEBGENQX_eUQ PAGE 38 BeQ10LQQ1 gel _QQUIBQL A'NSWERS _- FRAIRIE ISLAWD 182 -88/0~/11-HARE, S.

ANSWER 7.11 (2.39)

a. As a RCP is strrtede reflood or the vessel upp3r head with cool water will collapse ibe atame 9ubble (0.5) and pressurizer level will decrease rapidly as pressur?xer wett fills the void.(0.5)

5. If n RCP cannot be starteu etj e rapid cooldown is initiated, the void in the upse: head croms leraer10.5) which causes an insurge into the p r er s u r i;~ :. r .

The lower level prevents taking the pressurizer 3o11J.to.5) 52Pk%EUaE P.I, fr9cedurs 1E*,0.4 Backgfound Informatien ppl 3000274201 ,s ,;rA*S)

ANSWER 7 li (2.08:

e. A c c u r a c ,- of ti,atrumentation con he adversely affected (0,5) by t emp e r at 'we envior presoved , 0. 't ,' ] and radiation (0.25).

b. Engineering evaluation should ae performed. [0.333 J

( also accept discusion of integrated _ radiation dose)

Containment pressure less than 5 psig. [0.33]

Containment radiation less than 10 E4 R/hr (0.33)

REFERENCE 2-0 foldout page Woetinghouse generic emergency procedure bases 000011K312 000069G011 191002X111 ...(KA'S)

ANSWER 7.13 (1.50)

The net positive suction head requirements (also accept RHR wump runnout)

(0.75) of the RHR pumps were not met when the RHR pumps were in the rocirculation mode with the simultaneous use of; low head upper plenum vegsel inj ect ion, high head cold leg vessel injection, or containment spray. [0.75) (also accept with the simultaneous use of vessel inj ection/ containment spray)

REFERENCE PINGP 301 LER 88-001 080025M101 . ..(KA'S)

g.

Zs'__EBQQEQVBES_:_NQBdakt_eBNQBd6Lt_EdEBGENC1_6NQ PAGE 39 B8DIQLQQ1CGL_QQNIBQL i ANSWERS -- PRAIRIE ISLAND 182 -88/07/11-HARE, S.

l 1

ANSWER 7.14 (2.25) l

s. High river water temperature (0.5) i Loss of normal cooling water suction (0,5) l

b. Manually close the emergency cooling water safeguards gates.-(0.5)

c. The emergency cooling water line should be backflushed. (0.75)

REFERENCE !

B-35 p 31,32 C-35 p 13 NRC inspection report 282/87014 ,

076000G007 076000K121 ...(KA'S) i ANSWER 7.15 (1.25)

I

a. 3000 - 2900 mrem = 100 mrem dose remaining (0.25) 600 mR/Hr X 1 Hr/60 minutes = 10 mrem per minute (0.25)

(100 mR)/(10 mR/ minute) = 10 minutes (0.25)

b. 5(N-18) = 54.75 + 0.10 (0.25) i N- 18 = 54.85/5 l

N- 18 = 11 N= 11 + 18 = 29 (0.25) l l

REFERENCE !

Wostinghause Radiation, Chemistry and Corrosion, p 4-15 PI F-2 p 14 l 194001K103 ..-(KA'S) l l

i I

I i

l i

l

Sz_*_8DdINISIBoIIVE_PBQQEQUBE$t_QQUQlIIQUlt_8NQ_LidlI811QUS PAGE 40 ANSWERS -- PRAIRIE ISLAND 1&2 -88/07/11-HARE, S.

ANSWER 8.01 (2.25)

e. An arca where the dose is in excess of 1000 mrem (0.5) in one hour. (0.25)

b. Shift Supervisor (0,5)

c. Incore Thimble Chase Area (sump c)

Reactor Head Reactor Coolant Loops Spent Resin Tank Room Primary System Demineralizers VCT room Resin liner in fuel recieving area

( 4 at 0.25 each)

REFERENCE PI F-2 pp 11, 12 K/A 194001K1.03 (2.8/3.4) 194001K101 ...(KA'S)

ANSWER 8.02 (2.50)

e. Prevents core power from reaching o value. (0.25) at which fuel pellct centerline melting woulo occur. (0.25) ( It-acts during power excursions that are too rapid 40 be protected by

' temperature er 1 pressure protective circuitry.)

b. Loss of coolant accident or DNB (0.5)

c. Cladding f ailure f r om high temperature or DNB. (0.5)

d. It protects the safeties against water damage. (0.5)

e. Loss of heat sink (during a loss of main feedwater accident due to auxiliary feedwater pump starting delays.) (0.5)

PEFddENCE Tochnical Specification 2.3-4 K/A 012000 GD06 ()

a[_'_6DdINISIB6IIVE_EBQQEQUBi1&_GQNRIIION1t_8ND_LIdII8IIQN1 PAGE- 41 ANSWERS -- PRAIRIE ISLAND 1&2 -88/07/11-HARE, S.

ANSWER 8.03 ( .75)

True REFERENCE 5ACD 3.15, pg. 4 K/A 194001A1.09 (2.9/3.9) 194001A109 ...(KA'S)

ANSWER 8.04 (1.50)

The second PORV must be restored to an operable condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (0.5), or the associated block valve shall be closed. (0.5) Otherwise place the reactor in cold shutdown (0,5) within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

REFERENCE T/S 3.1-2 011000G011 ...(KA'S)

ANSWER 8.05 (1.50)

a. 1. Reduce power below 50%. (0.5)

2. High neutron flux setpoint reduced to less than or equal to 55% of rated power. (0.5)

b. Axial Flux Difference must be in the target band for at least 22

(.25) of the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . (.25) (less than 60 penalty minutes in the last 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months )

REFERENCE PI Technical Specifications, p. TS 3.10-3 and 4 K/A 001010G005 (3.7/4.1)

S[_'6DdINISIB6IIVE_E20GEQUBEft_QQUQlIIQUS&_6UQ_LidllallQNg PAGE 42 ANSWERS -- PRAIRIE ISLAND 182 -88/07/11-HARE,- S.

ANSWER 8.06 (2.50)

A. Lead Plant Equipment and Reactor Operator (LPERO) (0.5)

B. To safeguard human life (0.5)

C. To safeguard equipment or service or for other operational reasons (0.53 (full credit will be given for equivalent wording)

D. No [0.5)

Hold Cards shall not be attached on switchboard or benchboard remote control devices for clearance to work (0.5]

(full credit will be given for equivalent wording)

REFERENCE Administrative Work Instruction SAWI 3.10.3, Rev. 1, Para 5.2, page 5, Para 4.1.1, page 2, para 6.2.2.e., page 8 K/A 194001K1.02 ()

194001K102 ...(KA'S) i 1

ANSWER 8.07 (1.50)

I O. Shift Supervisor (0,5) ;

l

b. A locking device consists of a padlock (0.25) and chain or cable l (0.25). This device when applied to a component prevents its status from being changed without the removal of the lock (0.25). A blocking device is simply a snap hook which would be applied as a lock (above)

(0.25)

REFERENCE SWi-0-3 pg 2 l REFERENCE 194001K102 ...(KA'S)

Se_'_eQdINISIBoIIVE_EBQQEQUBESt_CQUQlIl0NSt_eNQ_LidlI&I1QNS PAGE 43

~

ANSWERS -- PRAIRIE ISLAND 182 -88/07/11-HARE, S.

ANSWER 8.08 (1.70)

e. True

b. Plant Operations Log Electrical Log Unit i Reactor Log Unit 2 Reactor Log Turbine Building Log (water treatment)

Weste Treatment Log (Aux operators log)

(0.2 each = .2)

REFERENCE SWI-0-4 pp 3, 4 K/A 194001A1.06 (3.4/3.4)

ANSWER 8.09 (1.50)

If/When the fire dampers close, hydrogen (0.75) (which is produced by the botteries) may accumulate and create en explosive hazard. (0.75) also accept: Heat (0.1], Fire gases (0.1] or smoke (0.1) for a maximum of [0.2)

REFERENCE F-5 Appendix A K/A 000067 K3.02 (2.5/3.3), K3.04 (3.3/4.1), 194001 K1.15 (3.4/3.8), Kl.16 (3.3/3.4) 000067K302 00067K304 19400K116 194001K115 ...(KA'S)

ANSWER 8.10 (1.25)

Stoam or gas bubble in pressurizer (0.375) and the SI pump discharge valves shut (0.375) (BOTH ARE REQUIRED FOR FULL CREDIT) also accept (for purposes of the integrated SI test) (0.375)

Reactor vessel head is removed [0.5)

REFERENCE T/S 3.1.G p 20 010000K403 (3.8/4.1) 010000G006 (2.6/3.6) 010000G006 010000K403 ...(KA'S)

E 1 at_'_6DdIN.I SIB 6IIVE_EBQQEQUBEf t_QQNQ111081&_68Q_ Lid 1I6IlQNS PAGE 44 -

ANSWERS -- PRAIRIE ISLAND 182 -88/07/11-HARE, S.

1 ANSWER 8.11 (1.50)

1. Number One fuel pool is smaller. l l

2. Provides assurance that in the event of a loss of pool cooling [0.33) !

at least.0 hours0 days 0 hours 0 weeks 0 months are evailable (0.33] until the onset of boiling. [0.33]

REFERENCE I l

T/S 3.8.B.5 000036G003 000036G004 ...(KA'S)

ANSWER 8.12 (1.00)

White Tags - Conditional release of tools / equipment to a clean area. (0.5) l l

Green Tags - Unconditional release of tools / equipment to the clean area.

[0.5) ,

l REFERENCE l

F-2 Page 20 194001KiO3 . . . ( Y. A ' S )

l ANSWER 8.13 ( .50)

a. Radioactive gas release out of the RCP seals. (0.503 Part B deleted from exam

b. Vortexing/eir binding of the pumps (0.50]

RCS temperature less than 140 F [0.50)

REFERENCE D-2 p 3 000025K101 003000A110 ...(KA'S)

ANSWER 8.14 ( .50)

Felse REFERENCE C 22.2 p 3 194001K115 ...(KA'S)

A/__eQUINISIB8IIVE_EBQQEQUBEft_CQNQIIIQNSt_8NQ_LIdII8IIQNS PAGE 45

- 1 ANSWERS -- PRAIRIE ISLAND 1&2 -88/07/11-HARE, S. f l

I I

I ANSWER 8.15 (1.75) l

a. The General Superintendent Plant Operations or his designee [0.5)

b. [0.25 each)

Visually check the breaker is in the connect position Check springs charged Breaker manual trip button is not depressed Red or green light on cubical door Visual or physical check to ensure latching is complete REFERENCE l 5AWI 3.10.1 062000A401 194001K107 ...(KA'S)

ANSWER 8.16 (1.00)

[0.33 each)

An emerger.cy If the badge is lost A door or reader does not function REFERENCE SAWI 5.1.1 p 4 194001K105 ...(KA'S)

ANSWER 8.17 (1.50)

The item is NOT operable. (0.50)

Surveillance intervals for a monch are not to exceed 7 days (0.50)

The surveillance is 8 days overdue (0.50)

REFERENCE PI Tech Specs 4.1.1 4

- , , 5a - ,,_ ,- - ~ - --

' .1 TEST CP.0SS REFERENCE PAGE l

.. 1 VALUE REFERENCE l QUES'r10N

' 05.01 1.00 HMS0001585 ,

05.02 1.00 HMS0001586 05.03 1.00 HMS0001587 05.04- 1.00 HMS0001588 .

05.05 1.00 HMS0001589 05.06 2.00 HMS0001590 '

05.07 2.50 HMS0001591 )

05.08 1.00 HMS0001592 !

05.09 1.00 HMS0001593 05.10 1.50 HMS0001594 1 05.11 1.50 HMS0001595 I 05.12 1.50 HMS0001596 05.13 1.50 HMS0001597 05.14 1.00 HMS0001598 l 05.15 1.00 HMS0001599 l 05.16 2.00 HMS0001600 l 05.17 2.25 HMS0001601 05.18 1.00 HMS0001602 05.19 1.00 HMS0001603 25.75 06.01 3.25 HMS0001604 l 06.02 1.80 HMS0001605 06.03 2.50 HMS0001606 06.04 2.50 HMS0001607 06.05 2.25 HMS0001608 1 06.06 1.00 HMS0001609 )

06.07 2.50 HMS0001610 06.08 1.00 HMS0001611 06.09 1.50 HMS0001612 06.10 2.50 HMS0001613 l 06.11 2.00 HMS0001614 06.12 1.00 HMS0001615 06.13 1.50 HMS0001626 ,

l 25.30 07.01 2.25 HMS0001617 l 07.02 2.00 HMS'001618 l 07.03 1.00 HMSL001619 07.04 1.50 HMS0001620 I 07.05 1.00 HMS0001621 l

07.06 1.00 HMS0001622 07.07 2.40 HMS0001623 07.08 .50 HMS0001624 07.09 2.00 HMS0001625 07.10 2.00 HMS0001626 07.11 2.00 HMS0001627 07.12 2.^0 HMS0001628 1

/ . _ - . - , , _ .

PAGE g' , TEST CROSS REFERENCE 2 l

\

dVEsi10N VALUE REFERENCE l

07.13 1.50 HMS0001629

' 07.14 2.25 HMS0001630 ,

07.15 1.25 HMS0001647 l 24.65 1

08.01 2.25 HMS0001631 08.02 2.50 HMS0001632 08.03 .75 HMS0001633 1 08.04 1.50 HMS0001634 .

0C.05 1.50 HMS0001635 )

08.06 2.50 HMS0001636 08.07 1.50 HMS0001637 08.08 1.70 HMS0001638 08.09 1.50 HMS0001639 08.10 1.25 HMS0001640 08.11 1.50 HMS0001641 08.12 1.00 HMS0001642 08.13 .50 HMS0001643 i 08.14 .50 HMS0001644 08.15 1.75 HMS0001645 l 98.16 1.00 .HMS0001646 08.17 1.50 HMS0001648 <

24.70 MWemWW 100.40 DOCKET NO 282 l

1 i

l o

1 9

>* l 1

1

~

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

l

' FACILITY: _EB8181g_1gL88Q_1h2______ l i

A -

REACTOR TYPE: _EWB-WEC2________________ 1 S^ 'q l z.-

DATE ADMINISTERED _SQLQZ411________________ l

( EN##"# ' EXAMINER: _E6BEINSQNa_Es __________

CANDIDATE: _________________________

INSTRUCTION 3 TO CANDIDATE: I l

Uno separate paper for the answers. Write answers on une side orly.

Stcple question sheet on top of the answer sheats, Points for each l qu stion are indicated in parentheses after the question. The passing ;

grcde requires at least 70% in each category and a final grade of at j lonst 80%. Examination papers will be picked up six (6) hours after i tho sxamination starts. i l

% OF CATEGORY % OF CANDIDATE'S CATEGORY

__296U5_ _IDIe6 ___HROB5___ _286Ug__ ______________G625098Y_________m___ i l

_2E 99__ _2Et2R _____,_____ ________ 1. PRINCIPLES CC NUCLEAR POWER )

PLANT OPERATION, THERMODYNAMIOS, l HEAT TRANSFER AND FLUID FLOW ;

_2E199__ _2Ez2E ___________ ________ 2. PLANT DESIGN INCLUDING SAFF.TY AND EMERGENCY SYSTEMS l 23:39__ _2dA6Z ___________ ________ 3. INSTRUMdNTS AND CONTROLS

_23:E9__ _23 22 ___________ ________ 4. PROCEDURES - NORMAL, ABNORMAL, EMERCENCY AND RADIOLOGICAL CONTROL

_29129__ ___________ _______..% Totals Final Grade All work done on this examination is my own. I have r.either given nor received aid.

Candiuate's Signature

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

1. Cheating on the examination means an automatic denial of your application cnd 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 gely to f acilitate legible reproductions.

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

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 gach nection of the answer sheet.

B. Consecutively number each answer sheet, write "End of Category __" as sopropriate, start each category on a Ogg page, write gely gg goe 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 Icast th t ee 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 f acility litstatutg.

13. The point value f or 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 mathumatical problems whether indic&ted in the question or noto

15. Partial credit may be given. Tne-efore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK. ,

16. If parts of the examination are not clear as to intent, ask questions of the gxamingt 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.

19. When you complete your ex amination , you shalli

c. Assemble your examination as follows:

(1) Exam questions on top.

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

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

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

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

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

t e

J

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

, . - . - - ,--,-.---,-nn- - -n,----e, - . ~---w - .- , - ---

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 2

. THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.01 (1.00)

The primary reason delayed neutrons make the reactor more controllable is d31cyed neutrons haves CSELECT THE CORRECT REASON 3 C1.03 A. Higher eftergy B. A longer life time C. A shorter life time D. Lower energy QUESTION 1.02 (1.00)

Shutdown margin is the actual amount of reactivity: CSELECT THE CORRECT STATEMENT 3 E1.03 A. iJhich would be inserted by shutdown bank rods B. Inserted by but wable poisons at BOL C. Inserted by dissolved beron in the RCS D. By which the reactor is suberitical l

I l

(***** CATEGORY 01 CENTINUED ON NEX7 PAGE *****)

. . . _ . - _ - - .- - - n.-,-, , , - , - , , , - - . , ,._..,,n - , _ - . .,.,4 -. , , , _ , _ . , _ , , , - - , - . - ----.en., . , , - ~ . . , , -

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 3

,. THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.03 (1.00) !

Wh3ther the moderator temperature coef ficient is positive or negative d:psnds on [ SELECT THE CORRECT STATEMENT 3 C1.03 A. Whether the given water-boron mixture absarbs more neutrons than it moderates or moderates more neutrons than it absorbs B. Only the boren concentration because raising the boron !

concentration makes the moderator temperature coefficient more l negative i

C. Only the moderator temperature because the magnitude of the I moderator temperature coef ficient is greater at higher temperatures )

D. Both the boron concentration and moderator temperature because although both the moderator pressure and void coefficients are positive they are neglected QUESTION 1.04 (1.00) j i

Why does the moderator temperature coefficiert become more negative from BOL to EOL7 [ SELECT THE PRIMARY REASON 3 E1.03 A. A decrease in the fuel to clad gap over cora age results in a de.rease in the fuel temperature B. A decrease in the boron concentration during core life due to fuel burnup and the production of fissiori f ragment poisons C, Plutonium building over core age results in more fissionable materi al being available to compete with boron atoms for neutrons D. Plutonium building over core age causes hardening of the nautron flux which results in more fast neutrons available for fast fission and an increase in the fast fission factor 1

l i

4 l

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

1. PRINCIPLES OF NUCLEAR PUWER PLANT OPERATION. PAGE 4

. THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.05 (1.00)

Which one ce the following factors is the predominant effect on making the Doppler coefficient less negative at EOL7 CSELECT TME CORRECT ANSWER 3 C1.03 A. The buildup of gases in the fuel rod gap B. Fuel densification C. Clad creep D. Plutonium 240 buildup

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

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 5

. THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.06 (2.00)

A rccctor has been shut down from 100 percent power and cooled down to 140 d:grces F over 3 days. During the cooldown, boron concentration was incrcased by 100 ppm. Given the f ollowing absolute values of reactivity chcnge over the 3 days:

Xenon = C 3 2575 pcm Rods = C 3 6938 pcm Temperature = C3 550 pcm Boron = C 3 1140 pcm Power Defect = C 3 1525 pcm A. What was the shutdown margin at the time of shutdown? CSELECT THE CORRECT VALUE3 C1.03

a. - 5413 pcm

b. - 5963 pcm

c. - 6938 pcm

d. - 7913 pcm B. What is the shutdown margin after all the reactivities were added?

CSELECT THE CORRECT VALUE3 C1.03 i

a. - 3428 pcm

b. - 4863 pcm

c. - 6478 pcm

d. - 9053 pcm

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

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. FOGE 6

, THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW I

,- 1 QUESTION 1.07 (1.00)

Aftcr operating at 50% power for several days, reactor power is increased I to 100%. How will reactivity change in the next one to two hours? [ SELECT THE CORRECT ANSWER 3 C1.03 A. Positive due to the faster burnup of Zenon at higher neutron flux levels B. Positive due to the increased rate of decay of Xenon to Cesium C. Negative due to the increase production of Xenon at the higher fission rate D. Constant initially due to the 6.2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months half-life of lodine 135 QUESTION 1.08 (1.00)

Why are rod insertion limits specified to maintain a minimum rod height I during reactor operations? CSELECT THE CORRECT ANSWER 3 C1.03 1 A. Maintaining the rods below the rod inserir. ion limit minimizes the consequences of an ejected rod B. Maintaining the rods below the rod insertion limit guarantees a sufficient shutdown margin from a given power level j C. Maintaining the rods above the rod insertion limit minimizes the !

effects of a dropped rod D. Maintaining the rods above the red insertion limit produces an axial flux distribution which prevents high local peak power l 1evels l

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

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 7

, THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.09 (1.00)

Which of the fellowing statements concerning fission product poisoning effoe.ts is TRUE7 CSELECT THE CORRECT STATEMENT 3 C1.03 A. Xenon is produced directly from fission and neutrcn absorption but Samarium produced from fission only B. Xenon is removed by decay and neutron absorption but Samarium is

. removed by neutron absorption only C. Equilibrium xenon is independent of power level but equilibrium Samarium is dependent on power level l

D. Xenon equilibrium is reached in about 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months following a power I transient but samarium equilibrium is reached in about 200 hours0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months !

following a power transient OUESTION 1.10 (1.00) '

I If the control rods are maintained above the rod insertion limits during routine reactor operations at power, which of the following parameters i+

most likely within limits? CSELECT THE CORRECT PARAMETER 3 C1.01 A. Axial Flux Difference CAFD3 1

B. Quadrant Power Tilt Ratio EQPTR3 C. Local Power Density CKW/ft3 D. Departure from Nucleate Boiling Ratio CDNBR3 l

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 8

. THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW

.~

OUESTION 1.11 (1.00)

Whct is the expected count rate response resulting from a brief rod withdrawal during the approach to criticality? CSELECT THE CORRECT COUNT RATE RESPONSE 3 C f O's A. An immediate rapid rise continuing to criticality B. An immediate rapid rise followed by a gradual increase to a higher steady state value C. A gradual increase followed by a rapid decrease when rod withdrawal is stopped D. A gradual increase continuing to criticality i

OUESTION 1.12 (1.00) l During a reactor startup, control rods are withdrawn such that the new I ctacdy-state count rate is 800 cps. If the count r ate bef ore rod i withdrawal was 400 cps and Keff was 0.96, what is Keff following rod withdrawal? [ SELECT THE CORRECT Keff3 [1.03 A. 0.97 B. 0.975 C. 0.98 I

D. 0.985 l

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 9

. THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.13 (1.00) ;

If roactor power is increased at a constant 1.0 dpm startup rate up to the point of adding heat, whicn one of the following responses will be the ;

INITIAL indication that the point of adding heat has been reac.1ed? CSELECT THE CORRECT INITIAL INDICATION OF REACHING THE POINT OF ADDING HEAT 3 C1.03 A. RCS temperature increases B. Pressurizer level increasaw C. Reactor power stabilizes D. Startup rata decreases QUESTION 1.14 (1.00) i During the initial startup following refueling, which of the following :

combinations of actions is most likely to occur when increasing nuclear pow;r from 15% to 100% power? CUELECT THE CORRECT COMBINATION 3 C1.03 A. Incr6csing turbine first-stage pressure, decreasing RCS boron concentration and increasing rod height

5 B. Decreasing turbine first-stage pressure, decreasing RCS baron i concentration and increasing rod height i I

C. Increasing turbina first-stage pressure, increasing RCS boron ;

concentration and decreasing rod height l D. Decreasing turbine first-stage pressure, increasing RCS horon concentration and decreasing rod height l

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1. PRINCIPLES OF NUCLEAR POWER PLANT DPERATION. PAGE 10

. THERMODYNAMICS. HEAT TE8MEFER AND FLUID FLOW Luo'.!ON 1.15 (1.00)

How will pressurizer level indication compare to actual pressurizer level if containment temperature is increased Esteam leak 3 from 100 degress F tc 180 degrees F7 EConsider on1v ref erence leg heating ef f ects3 ESELECT THE CORRECT COMPARISON] C1.03 A. Indicated pressurizer level will be higher than actual pressurizer level B. Indicated pressurizer level will be lower the.n actual pressurizer level C. Indicated pressurizer levn1 will be equal to actual pressuriser level D. Indicated pressurizer level will fail as is QUESTION 1.16 (1.00)

Unit 1 is operating at 30% load following a 2 day outage to repair a steam lock. A ramp increase in load is commenced with the steam flow density componsation failed at the 30% load value. How will the 100% load steam flow indication be af f ected by the steam flow density compensation f ailurw?

CSELECT THE CORRECT COMPARISON 3 E1.03 A. The indicated steam flow will be greater than actual steam flow B. The indicated steam flow will be less than actual steam flow C. The indicated steam flow will be equal to actual steam flow D. The indicated steam flow will fail at the 30% load value

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1. PRINCIPLES OF. NUCLEAR POWER PLANT OPERATION. PAGE 11

,"THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.17 (1.00)

How does the convective heat transf er coef ficient vary as reactor coolant flows up along a fuel rod which experiences departure from nucleate boiling CDNB3 near the top of the fuel rod? CSELECT THE CORRECT ANSWER 3 E1.03 A. Increases continuously B. Decreases continuously C. Increases, then decreases D. Decreases, then increases I

QUESTION 1.18 (1.00)

Which of the following core conditions will DECREASE the departure from nuc1cate boiling ratio CDNBR37 (Consider each sepsrately3 CSELECT THE J CORRECT CONDITION 3 ti.03 A. Reactor power decreases ;

B. Reactor coolant pressure increases C. Reactor coolant temperature decreases D. Reactor coolant flow decreases QUESTION 1.19 (1.00) i During a Unit 2 plant cooldown and depressurization with forced circulation, RCS loop flow indication becomes erratic. What is the most likoly cause of the erratic RCS loop flow indication? CSELECT THE MOST l

LIKELY CAUSE3 C1.03 j

1 A.

RCP runout l l'

B. RCP cavitation C. RCS hot leg saturation D. RCS loop water hammer l l

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 12

THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.20 (1.00)

Which of the following conditions is required for brittle fracture of a ))

rocctor pressure vessel to occur? ESELECT THE CORRECT ANSWER 3 C1.03 i

A. Neutron embrittlement of the vessel wall, temperature below Nil Ductility Temperature ENDTJ, tensile stress applied to tne vessel wall B. Material defect in the vessel wall, temperature below Nil Ductility Temperature CNDT3, tensile stress applied to the vessel wall C. Neutron embrittlement of the vessel wall, temperature below Nil Ductility Temperature CNDT3, material def ect in the vessel wall D. Material defect in the vessel wall, neutron embrittlement of the l vessel wall, tensile stress applied to the vessel wall QUESTION 1.21 (1.00)

How cre letdown system upstream pressure and flow rate af fected by manually closing the Letdown Pressure Control Valve ECV-312033 an additional 10%? .

CSELECT THE CORRECT AK7elERJ C1.03 )

l A. Pressure decreases and flow rate increases I B. Pressure ducreases and tiow rate decreases C. Pressure incruases and flow rate decreases D. Pressure increases and flow rate increases ;

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Lu PRiUCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 13

, THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.22 (1.00)

Which of the following steam generator parameters is nearly linear Edircctly proportional] to reactor power as turbine load is ramped from 25%

to 75% load? CRod control is in automatic 3 CSELECT THE CORRECT PARAMETERJ C1.C3 A. Delta-H Cateam enthalpy - feed enthalpy]

B. Delta-T Csteam temperature - feed tamperature3 C. Steam flow D. Steam generator level QUESTION 1.23 (1.00)

Which of the f ollowing ef f ects will occur as a result of increased f ouling of the main condenser tube bundles? CAssume cooling water flow and sheam flow rates remain constant] CSELECT THE EFEZCT THAT WILL OCCUR 3 C1.03 A. Cooling water outlet temperature will decrease B. Condenser hotwell temperature will decrease C. Condensate depression will increase D. Condenser Heat rejection will increase QUESTION 1.24 (1.00)

Which of the f ollowing indications DOES NOT verif y that natural circulation 10 octablished? [ SELECT THE INCORPECT INDICATION 3 C1.03 A. Core Outlet temperatures are constant or decreasing B. Decrease in steam pressure during constant decay heat generation C. Constant or decreasing Delta T across the reactor core less than the full load Delta T D. Constant steam generator level Nith a constant auxiliary feedrate and a constant or decreasing steam pressure

(***** END OF CATEGORY 01 *****)

2 FLANT DESIGN INCLUDING SAFETY AND_ EMERGENCY SYSTEMS PAGE 14 QUESTION 2.01 (2.50)

A. Why is a small continuous bypass spray flow provided around the pressurizer spray valve? CLIST TWO REASONS 3 CO.53 B. Why is each pressurizer saf ety valve provided with a water-filled loop seal between the saf ety valve and the pressurizer? CSTATE PURPOSE OF SEAL 3 CO.53 C. What are the air supplies provided for operating the pressurizer PORVs? CLIST TWO AIR SUPPLIES 3 CO.53 D. Which relief valves discharge to the Pressurizer Relief Tank (PRT)?

CLIST FOUR RELIEF VALVES 3 CPressurizer PORV and safety valve discharge not accepted for credit 3 C1.03 QUESTION 2.02 (3.00)

A. What are the functions of the VCT7 CLIST FOUR FUNCTIONS 3 C2.03 E. Charging flow normally passes through the loop "B" return iso 1& tion valve (CV-31328) and into the RCS. CV-31328 is equipped with a bypass line containing a check valve.

1. W'.at is the PURPOSE of the bypams line? CO.53

2. How does the check valve operate to perform the bypass line purpose / function? CINCLUDE CHECX VALVE DESIGN AND SETPOINT3 Co.53

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4 i

2.

PkeNT DESIGN INCLUDING WBEETY ANR, EMERGENCY SYSTEMS PAGE 15 l

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QUESTION 2.03 (2.00) .

A. What are the lodds or components supplied by the Unit 1 Component l Cooling (CC) system? [ LIST TEN & INCLUDE COMPONENTS SHARED WITH UNIT '

23 CDifferent components in a system will be accepted but component parts will be accepted only once as the component 3 C1.03 i 4

B. 1. What are the conditions / signals that will doenergize the 21 CC pump yellow indicating (ready) light on the CC pump control board l switch? CLIST TWO CONDITIONS / SIGNALS 3 to.53 !

l

2. If the 21 CC pump yellow indicating (ready) light goes out, can the 21 CC pump be started [DO NOT CONSIDER THE BULB TO BE DEFECTIVE 3

a. Manually? CO.253

b. Automatically? Co.253 QUESTION 2.04 (2.50)

A. How are the RHR hot leg supply isolation valves, MV32164 and MV-32165, interlocked? CDESCRIBE TWO INTERLOCKS 6 INCLUDE SETPOINTS3 ti.03 4

B. Why is the motor-operator breaker for the RHR hot leg supply isolation valve, MV-32164, opened during normal power operation? CEXPLAIN WHAT IS PREVENTED AND WHY IT MUST BE PREVENTED 3 CO. 5 3 C. How are the RHR heat exchanger to SI pump suction valves interlocked?

CDESCRIBE TWO INTERuOCKS & INCLUDE SETPOINTS3 ti 03 i

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2. PLANT DESIGM INCLUDING SAFETL AND EMERBENCY SYSTEMS PAGE 16 QUESTION 2.05 (2.50)

A. What are the fuel transfer system interlock conditions that must be met or satisfied before the conv9yor car can be moved? CDESCRIDE TWO INTERLOCK CONDITIONS 3 [1.03

9. How is the f uel transf er system conveyor car interlocked with the ilitjng frames to prevent lifting frame operation? [ DESCRIBE TWO INTERLGrK CONDITIONS 3 to.753 C. How is the fuel transfer system lifting frame in containment interlocked with the manipulator crane? CDESCRIBE THE OPERATION (S)

PREVEN 1ED3 CO.753 QUESTION 2.06 (2.00)

Tho SI test line to R%ST inalation valves are interlocked with the containment sump "B" isolation valves.

A. How are the valves interlocked? CDESCRIBE INTERLOCK 3 C1.03 B. What are the conditions /mvents that this interlock prevents? CLIST TWO3 ti.03

'. 1 OUESTION 2.07 (2.00) 1 What actions are initiated in the ECCS by the safety injection "S" signal? I

ELIGT THE ECCB COMPONENTS ACTUATED AND THE TYPE OF SIGNAL RECEIVED BY THE I COMPONENTS 3 CFor example _____ _ pump (s) - atxrt/stop or _______ valve (s) l

opcn/close3 CInclude normally open valves] C2.03 QUESTION 2.08 (2.00) l What are the renditions that will cause a saf ety injection "S" signal to be g: ncrated? [ INCLUDE COINCIDENCE AND SETPOINTS3 C2.03 l

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2. ,P_LANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 17 QUESTION 2.09 (2.00)

A. What are the conditions that will cause a containment spray "P" signal to be generated? CINCLUDE COINCIDENCE AND SETPOINTS3 C1.03 B. How does the containment spray system respond following generation of a "P" signal? CLIST FIVE ACTIONS THAT ARE INITIATED 3 C1.03 j l

QUESTION 2.10 (1.50)

Whct are the 4.16 KV power supplies fo? the following safeguard buses?

A. 15 Bus? CLIST COUR3 CO.43 l

, B. 16 Bus? CLIST FOUR3 IO.,4 3 C. 25 Bus? CLIST THREE3 CO,33 D. 26 Bus? CLIST FOUR3 Coe4 ,

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l QUESTION 2.11 (1.00) {

SELECT THE ANSWER that correctly describes the DC distribution panel's powsr supplies. C1.03 A. DC Panel 15 Normal supply in DC Panel 11 and DC Panel 17 Standby I supply is DC Panel 12. l 1

B. DC Panel 17 Normal supply is DC Panel 21 and DC Panel 15 Standby supply is DC Panel 12.

C. DC Panel 15 Normal supply is DC Panel 12 and DC Panel 17 Standby i supply is DC Panel 21. l l

D. DC Panel 17 Normal supply is DC Penel 11 and DC Panel 15 Standby supply is DC Panel 21.

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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 18 QUESTION 2.12 (1.00)

A. What are the tanks that are vented to the cover gas / vent header? CLIST SIX3 CWhere both units rave similar tanks, credit will be given only for the unit i or unit 2 tank, not both3 to.63 B. What are the equipments that drain to the RCDT7 CLIST FOUR3 C0.43 AUESTION 2.13 (1.00?

Whst aFe the sources of auxiliary feedwater for the following AFW pumps?

A. AF pump #117 CTHE SUCTION HEAGER (C-41-2, CROSS TIE) MANUAL ISOLATION IS SHUT 3 CO.253 B. AF pump #127 CTHE SUCTION HEADER (C-41-2, CROSS TIE) MANUAL ISOLATION IS SHUT 3 to.253 C. AF pump #217 CBACKUP SUPPLY WHEN THE NDRMAL SUPPLY IE DEPLETED 3 to. 2^ t D. AF pump #227 CBACKUP SUPPLY WHEN THE NORMAL SUPPLY IS DEPLETED 3 to.253 l

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

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

3. INSTRUMENTS AND CONTROLS PAGE 19 1

QUESTION 3.01 (1.80) l j

A. Why are the reactor vessel level instrumentation mystem (RVLIS) DP cells located outside of containment? CSTATE THE FRIMARY REASON 3 CO.63 B. Since the RVLIS DP cells are installed outside of cCntainmant, how does the RVLIS design reduce the hazard of piping breaks outside of ;

containment? CDESCRIBE TWO COMPONENTS THAT WILL LIMIT LEAKAGE OUTSIDE l OF CONTAINMENT] [1.23 1

QUESTION 3.02 (1.75)

During operation at power C93%3, you are manually withdrawing the controlling bank of rods and the "Logic Cabinet Urgent Failure" annunciator 10 cctuated in the control room.

A. How ie rod motion af f ected? C ANSHER FOR BOTH MANUAL AND AUT03 CO.63 B. What are the logic cabinet conditions /f ailures that cause a "Logic Cabinet Urgent Failure"? ELIST THREE CONDITIONS / FAILURES 3 [0.93 C. When the malf unction has been corrected, how is the alarm circuit reset? CSTATE THE CONTROL USED3 CO.253 I

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3.

INSTRUMENTB AND CONTROLS PAGE 20 QUESTION 3.03 (1.50)

A. The "Low Power" rod stop blocks outward rod motion in automatic. ,

1. WHAT in the "Low Power" rod stop SETPOINT7 E0.33 2 WHY is the "Low Power" rod stop REQUIRED 7 CO.33 B. The "High Bank D Rod Position" rod stop blocks outward rod motion in automatic.

Other than blocking outward rod motion, WHAT doea the "High Bank D Rod Position" rod stop PREVENT 7 CO.33 QUESTION DELETED C. Other than blocking automatic rod withdrawal, what are the other functions of the "overtemperature delta-T" rod stop? [ LIST TWO3 CO.63 D. What is the SETPOINT and COINCIDENCE f or the "overpower delta-T rod 9 top? CO.63 QUESTION 3.04 (1.20)

Tho Rod Position Indication System consists of two redundant position indication systems. How does each system measure the rod position?

[ STATE THE NAME AND DESCRIBE HOW EACH SYSTEM DETECTS ROD POSITION 3 [1.23 QUESTION 3.05 (2.00) l Tha Tavg signal is used as an input to alares and information light (s).

What are the alarms and informetion light (s) that receive a Tavg input cigno17 CLIST FOUR, INCLUDE SETPOINTS and COINCIDENCE if coincidence is rcquired3 [2.03 l

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Iz__ INSTRUMENTS AND_ CONTROLS PAGE 21 QUESTION 3.06 (2.50)

During power operation a load rejection occurs that causes a 5F delta T tcuctioneered high Tavg - Tref 3 error signal. If the steam dumps fail to cutenatically respond to the load rejection, what could cause the steam dump valves to fail to respond automatically? CLIST FIVE E53 CONTROLS OR CONDITIONS you would check or verify, INCLUDE COINCIDENCE AND SETPOINTS wh ro appropriate 3 CDo not consider component failures such as binding volvcs, burnt out solenoids, blown fuses, etc.3 C2.53 QUESTION 3.07 (2.40) l 1

Whilo operating at 75% power, the Pressurizer Pressure Control System j rofcrence pressure, Pref, CProportional-Integral-Derivative (PID) i Controller setting] fails to zero. If NO OPERATOR ACTION is taken, what i ALARM SIGNALS will be provided by the Pressurizer Pressure Control System? ;

CLIST FOURI INCLUDE SETPOINTS AND COINCIDENCE if any3 [2.43

)

QUESTION 3.08 (2.90)

A. What condition (s) are required for the P-4 permissive actions to occur? CLIST TWO COMPONENTS AND THEIR REQUIRED POSITION 3 CO.63 B. What are the actions that the P-4 permissive initiates? CLIST FOUR ACTIONS and ASSOCIATED CONDITIONS CSETPOINTS & COINCIDENCES] as appropriate 3 C2.03 i

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l

- 3. INSTRUMENTS AND CONTROLS PAGE 22 i

I

.' ,.i,

@UESTION 3.09 (2.00) i LABEL ITEMS A. through J. on the following source range block diagram (2.0]

/

BF3 A.

B. Power supply

I c.

I/A D.

pulse shaper I

pulse driver I

E.

i l

F.

G.

I/A computer )

4 H. !

I

, I.

containment evacuation s

J.

$ /

s I

I I

i I 1

2 ..

'(aeu.m CATEGORY 03 CONTINUED ON NEXT PAGE meeae) i

, , m- - . ,..,,,.7- .- , . ----,e- --v - w w- .---r--,,- - - - - , - - -, y.- w , -- ,-~%- . --

3.

INSTRUMENTS AND CONTROLS PAGE 23 J

QUESTION 3.10 (1.75)

A. What are the primary functions of the incore instrumentation system?

CLIST TWO3 C1.03 B. What are the secondary functions of the incore instrumentation system?

CLIST THREE3 CO.753 QUESTION 3.11 (1.60) 4 high level alarm on the spent fuel pool (SFP) air monitors (R-25 and R-31) initiates five C53 automatic actions in the plant ventilation system.

If a high level alarm occurs on the spent fuel pool air monitor, what are the cutomatic actions initiated in the plant ventilation system 7CLIST FOUR3 C1.63

! OUESTION 3.12 (3.00)

Tho Engineered Saf eguards System is capable of initiating a number of diffcrent signals. Each signal i s a response to a different set of cbnormal plant conditions and initiates a number of dif f erent actions.

For example: A safety injection (S) signal results in a reactor trip and provides emergency makeup water to the reactor coolant system. LIST SIX C63 additional signals initiated by the Engineered Saf eguards System

[ INCLUDE WHAT THE SIGNAL PREVENTS, MINIMIZES, REDUCES OR ENSURESJ [3.03 1

1

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4. PROCEDURES - NORMAL. ARMQRMAL . EMERGENCY AND PAGE 24

'E8DIOLOGICAL CONTROL QUESTION 4.01 (2.00)

Whilo operating at 93% load on Unit 2, you, as the reactor operator, cbecrve that pressurizer level and pressure are slowly decreasing. As you otcrt to announce your observation pressurizer level and pressure rapidly d:crcase and a reactor trip and safety injection occurs. What immediate cctions are performed in accordance with Emergency Operating Procedure 2E-0, Reactor Trip Or Safety Injection? CLIST SEVEN IMMEDIATE ACTIONS 3t2.83 QUESTION 4.02 (1.00)

Em:rgency Operating Procedure 1E-3, Steam Generator Tube Rupture, contains tho following:

STEP ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 10 Establish Instrument Air To Containment 11 Verify Safeguard Buses Attempt to restore

- ENERGIZED BY OFFSITE offsite power.

POWER CAUTION Isolation of the ruptured SG SHALL be complete before i continuing to Step 12 unless the ruptured SG is needed for RCS cooldown l

12 Check Ruptured SG(s) Go to 1ECA-3.1, SGTR WITH l Pressure - GREATER LOSS OF REACTOR COOLANT: ;

THAN 250 PSIG SUBCOOLED RECOVERY, l Step 1.

i Whilo performing Emergency Operating Procedure 1E-3, Steam Generator Tube Rupture, the ruptured steam generator was identified and isolated. Steps 1 through 10 have been completed satisfactorily. If offsite power cannot be roctored, what action should be taken? CWHAT IS THE GENERAL RULE FOR USAGE OF EMERGENCY PROCEDURES IF THE RESPONSE NOT OBTAINED ACTION CANNOT BE C1.03 PERFORMED OR IS NOT SUCCESSFUL 3 I

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, 4. PROCEDURES - NORMAL. ABNORMAL. EMERG.ENCY AND PAGE 25

. MADIOLOGICA.1, CONTRQL l

c i

Ql'ESTION -

4.03 (1.00) <

A. Emergeniy Plan Implementing Procedure F3-2, Classification Of Emergencies, defines four graded levels of emergency classifications. What cro the four C43 Emorgency Action Levels that indicate that an emergency cxicts? [NAME FOUR3 C1.03 i

QUESTION 4.Q4 (1.00)

Sc o of the diesel generator fuel oil day tank level valves require ,

indtpendent verification. How can you determine if a diesel generator fuel oil day tank level column isolat3on valve requires independent verification f ollowing repositioning? [BTATE/ EXPLAIN HOW YOU ACCESS THE INDEPENDENT '

VERIFICATION LIST CIVL33 C1.03 I

QUESTION 4.05 (2,50)

During a walkdown of the esin control boards prior to relieving the Reactor Op:rctor, you note that reactor makeup pump #11 "Start / Normal /Stop" control !

cwitch is in the "Pullout" position with a Hold Card attached C#11 motor i maintenance is in progress 3. Yesterday you noted that reactor makeup pumn i 022 "Start / Normal /Stop" control switch ?$as in the "Stop" position with a S: cure Card attached C#21 pump was being tested and operation of #22 pump l would invalidate the #21 pump test 3.

A. Who is responsible f or maintaining the Hold and Secure Card Record Logs? ISTATE POSITION 3 E0.53 B. What is a Hold Card used for? CSTATE THE PURPOSE of a Hold Card 3 to.53 C. What is a Secure Card used for? CSTATE THE PURPOSE of a Secure Card 3 ;

[0.53 )

D. Is the Hold Card being used properly? CANSWER YES OR NO and JUSTIFY / EXPLAIN your choice 3 [1.03 4

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4.

PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 26

, E8D10 LOGICAL CONTROL l l

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GUESTION 4.06 (1.20)

A ref ueling outage is nearing completion on Unit 1. You are performing vcrious checklisto on Unit 1 outside of the control room.

A. A valve on checklist C1.1.15, Residual Heat Removal System, which is i specified to be in the open position by the checklist is found in the shut position. You should open the valve, initi al the checklist and inform the control room operator. CANSWER TRUE OR FALSE 3 CO.43 B. Whe;1 you have completed checklist C1.1.15, Residual Heat Removal System, it is acceptable for you to call the control room and transfer the C.1.15 checklist working copy information to the master copy of the checklist. CANSWER TRUE OR FALSE 3 CO.43 C. While performing checklist C1.1.4, Reactor coolant system, you discover that your working copy of the checklist is dirty / smudged and you are unsure of two [23 items on the checklist. To minimize your exposure, you should call the control room and have the control room !

operator direct you Cover a communication system 3 in performing the

'cwo items you are unsure of. CANGWER TRUE OR FALSE 3 CO.43 i

QUESTION 4.07 (2.70) I Function Restoration Procedure 1FR-S.1, Response To Nucinar Power G:ncration/ATWS, specifies three C33 immediate action steps including: ;

1. Verify Reactor Trip, and

2. Verify Turbine Trip. I l

A. How oo you verify Reactor Trip in accordance with 1FR-B.17 CLIST FOUR l C43 ACTIONS / EXPECTED RESPONSES] [1.23 B. When you verify Turbine Trip, you fail to get the expected response.

What action will you take in accordance with 1FR-S.17 CLIST FOUR 243 RESPONSE NOT OBTAINED STEPS IN THE SPECIFIED ORDER 3 C1.23 C. What is the third Z3ro3 1FR-C.1 immediate action step? CSTATE l ACTION / EXPECTED RESPONSE 3 t o. 33 !

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4 PROCEDUAES - NORMAL. ABNORMAL. EMERBENCY AND PAGE 27

. RADIDLOGICAL CONTROL QUESTION 4.08 (0.50)

A. Units 1 & 2 are operating at 99X power. A Unit i reactor trip and '

ccfoty injection occurs. As reactor operator you perform your isoediate cctit1s and subsequent actions an directed by the shift supervisor. You obecrve the f ollowing indications:

a. Power range < 5%

b. Containment pressure = 18 psig

c. Pressurizer level < 15%

d. Core exit TCs = 895F

e. RVLIS full rarge = 31%

f. RCFs off

g. Both SG wide range levels indicating 57%

h. RCS subcooling based on core exit TCs = 15F l Which Critical Saf ety Function is in a RED CONDITIO!17 CSTATE the CRITICAL ;

l SAFETY FUNCT7nN3 CO.53 i

)

B. Unit 1 is operating at 100% power, A unit 2 startup is in progress i

with power stabilized at approximately 10-8 amps. Indicattor,s on both 1 unito are normal for the existing conditiens. What is the Critical Safety Function Status Tree indications or conditions that exist for these opcrcting conditions? CSELECT THE CORRECT ANSWER 3 Co.53

1. Both units subtriticality - RED

2. Unit i subcriticality - RED 1

3. Ur.it 2 subcraticality - RED I 1

4. Both units subcriticality - ORANGE PART B. OF THE QUESTION WAS DELETED l

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4. PROCEDURES _- NORMAL. ABNORMAL. EMERGENCY AND PAGE 28 R&DIOLOGICAL CONTROL QUESTION 4.09 (1.50)

Units 1 and II are both operating at 100% load. A Unit II reactor trip cccurs. As the Unit II reactor operator you perform the 1E-0, Reactor Trip Or Stfety Injection, immediate actions and have exited f rom 1E-0 to 1ES-0.1, Reactor Trip Recovery. 1ES-0.1 steps have been performed octiofactorily through Step 12, Maintain Stable Plant Conditions. Given that the Inf ormation Pagn For E-0 Series Procedures provides the following RCP Trip Criteria Trip both RCPn if all conditions listed below occurs

a. SI Pumps - AT LEAST ONE RUNNING

b. RCS Pressura - less than 1250 PSIG C1550 PSIG3

c. An operator controlled cooldown hSs NOT been initiated Undar what other conditions would you trip the RCP(s)? CSTATE THREE C33 CONDITIONS DR LIMITS] [1.53 QUESTION 4.10 (2.10)

If e reactor trip with safety injection occurs, how is the 1E-0, Reactor Trip Or Safety Injection, immediate action "VENIFY SAFEGUARD COMPOttENT ALIGNMENT" performed? CLIST THE SEVEN INDICATIONS, COMPONENTS OR ITMMS THAT MUST BE CHECKED TO COMPLETE THE VERIFICATION 3 C2.13 QUESTION 4.11 (1.50)

Section Work Instruction SWI-0-2, Section Work Instructions Operations S ction Shift Organization, Operation & Turnover, establishes areas of rocponsibilities for control room operators. Which CUNIT i or UNIT II3 Plcnt Equipment and Reactor Operator is responsible for the following chcrcd areas:

A. butside Areas? [ ANSWER UNIT 1 OR UNIT II3 CO.53 B. Auxiliary Building? CANSWER UNIT 1 OR UNIT 113 Co.53 C. Service Building? CANSWER UNIT 1 OR UNIT 113 Co.53 i

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(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)

i, PRQCEQQBES - NORMAL. ABNORMAL, EMERGENCY AND PAGE 29

.BBDIOLOGICAL CONIBQL QUESTION 4.12 (1.70)

Plcnt Operations Manual, Section F2, Radiation Safety, specifies eight C83 rcdiction protection responsibilities individual's shall have. For the following radiation protection responsibilities for individuals, FILL IN THE MISSING WORD (S).

A. Read carefully and observe all requirements as spelled out on applicable __________________________________. CTHREE [3] WORDS3 CO.53 B. Take special care to assure he is free of detectable contamination before leaving the _______(1)_______ CONE [13 WORD 3 area and the

_______(2)________ CONE [13 WORD 3 area. [0.43 C. Eliminate unnecessary exposure by refraining from lingering in

_______(1)________ CONE [1] WORD 3 fields, by maximizing

_______(2)________ CONE [13 WORD 3 from radiation source, and by utilizing existing _______(3)________ CONE [13 WORD 3. CO.63 D. Attempt to keep exposure _________________ [ONE [13 ACRONYM 3 by reviewing work procedures and conducting as much preliminary work outside radiation areas as feasible. [0.23

/

QUESTlON 4.13 (2.00)

Unito 1 and 2 are both operating et 100% power. A relay room fire occurs thct causes the Unit 1 Shif t Supervisor to make the decision to evacuate tho control room.

A. What is your INITIAL evacuation assignment if you are the UNIT 1 Lead Plant Equipment and Reactor Operator CLPERO37 [ STATE THE FLANT LOCATION OR EQUIPMENT YOU PROCEED TO3 [0.53 B. What is your INITIAL evacuation assignment if you are the UNIT 2 Lead Plant Equipment and Reactor Operator [LPERO3? [ STATE THE PLANT LOCATION OR EQUIPMENT YOU PROCEED TO3 [0.33 C. What is your INITIAL evacuation assfgnment if you are the UNIT 1 Plant Equipment and Reactor Operator EPE&RO37 CSTATE THE PLANT LOCATION OR EQUIPMENT YOU PROCEED TO3 [0.53 4

D. What is your INITI AL evacuation assignment if you are the UNIT 2 Plant Equipment and Reactor Operator [PE&ROI? [ STATE THE PLANT LOCATION OR ECUIPMENT YOU PROCEED TO) [0.53

(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)

4. PROCEDUREG m NORMAL. ABNORMAL. EMERGENCY AND PAGE 30 KADIOLOGlCAL CONTROL ,

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1 QUEDTION 4.14 (1.00) j i

Whct is the type of radiation that is primarily measured by a pelf-reading ,

pockct dosimeter CSRPD37 [ SELECT THE CORRECT ANSWER 3 tj.03

]

A. Alpha B. Beta 1

0. Gamma D. Neutron QUESTION 4.15 (2.00) :

Prcirie Island Units i&2 Technical Specifications specifies two [2] safety lirifts:

1. Safety Limit, Reactor Core, and

2. Safuty Limit, Reactor Coolant System Pressure.

A. What is the objective of the Reactor Core Safety Limit? [ STATE THE TS OBJECTIVE / PURPOSE of the Reactor Core Safety Limit 3 to.53 B. What are the parameters that apply to the Reactor Core Saf ety Limit?

CLIST THREE PARAMETERS SPECIFIED IN TS as applying to the Reactor Core Safety Liioit3 CO.753 C. What is the objective of the Reactor Coolant System Pressure Safety Limit? CSTATE T!iE TS OBJECTIVE / PURPOSE of the Reactor Coolant System l Pressure Safety Limit 3 CO.53 l D. What is thm TS specification for thi Reactor Coolant System Pressure Safety Limit? CSTATE THE TS PRESSURE LIMIT 3 CO.253 ,

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END OF CATEGORY 04 *****)

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

i

ATTACHMENTS PRAIRIE ISLAND 182 RO EXAH PARKINSON, K.

88/07/11 i

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i 4

EQUATION SHEET f = ma y = s/t Cycle efficiency = (Net work out)/(Energy in) I 2

w = eg s = V,t + 1/2 at g w? . .

KE = 1/2 mv 2

, , (yf , y g)fg 4 , 13 g , g ,4 %

_ o Pe = agn Vf = V, + at w = e/t A = an2/t.1/2 = 0.693/t1/2 '

y , y gp, -

n0 2 1/2 "

A= ,

[(t1/2)

IIb))

al = "31 am

l m = V,yAo ,

-Ex Q = mCpat Q = UAa T' I " I,e'#

Pwe = W ah f I = I,10"*/ M TYL = 1.3/u .

J P = P,10 sur(t) HYL = -0.683/u ,

t P = Po e /T SUR = 26.06/T SCR = S/(1 - K,ff)

CR x = S/(1 - K,ffx)

SUR = ?.6o/ t* + ($ - o )T CR j (1 - K,ffj) = CR2 (1 - keff2) I l

T = ( t*/o ) + ((8 - s y To ] M = 1/(1 - K,ff) = CR j/CR, T = 1/(o - 8) M = (1 - X ,ff,)/(1 - K,ffj)

T=(8-o)/(Is) SOM =.( - K,ff)/K,ff o=(K,ff-1)/K,ff=AK,ff/K,ff t' = 10 seconds ]

I = 0.1 seconds *I I

o = ((t=/(T K,ff)] + (a,ff /(1 + IT)] , I Ij dj = I d P = (taV)/(3 x 1010) Id 2 ,2 2 d jj g'2 2 i

I = oN R/hr = (0.5 CE)/d 2(meters) l R/hr = 6 CE/d2 (feet) 1 Water Parw.eters Misem11aneous Conversions 1 gal. = 8.345 10m. 1 curio = 3.7 x 1010 dps 1 ga]. = 3.78 liters 1 kg

2,21 lbm 1 ft4 = 7.48 gal. 1 np = 2.54 x 103 Stv/hr 1 Density = 62.4 lbrIi/ft3 l Oensity = 1 gm/c9 1 on = 3.41 x 106 5tu/hr '

lin = 2.54 cm Heat of vacorn ation = 970 5tu/lom *F = 9/5'C + 32 Helt of fusion = 144 Stu/lem *C = 5/9 (=F-32) i 1 At.i = 14.7 psi = 29.9 in. Hg. 1 BTU = 778 ft-lbf I ft. H O 2

= 0.4335 Inf.in. !

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l Table 1. Saturated Steam: Temperature Table Abs Press. Specific Volume Enthalpy Entropy Temp tb per Sat. Sat. Sat. Sat. Sat. Sat. Temp Fahr SqIn. Liquid Evap Vapor Liquid Evap 'Japor Liquid Evap Vapor Fahr t p vg vgg vg hf h fg hg sg sig Og i 32 8 0 08859 0 016022 33041 3304 7 0 0179 1075 5 1075.5 0 0000 2.1873 21873 32 8 34 2 0 09600 0 016021 3061.9 3061.9 I996 1074 4 10764 0 0041 2.1762 2 1832 34 8 36 8 010395 0 016020 28390 2839 0 4 003 1013 2 1077.2 0 0081 2.1651 21732 36 8 38 0 0 11249 0 016019 26341 2634 2 6 018 10723 10781 0 0122 2.1541 2X63 38 8 48 8 1.12163 0 016019 2445 8 24458 8 027 10710 Z790 0 0162 2.1432 2 1594 48 8 42 8 013143 0 016019 2272.4 2272.4 10 035 1069 8 1079.9 0 0202 2.1325 2.1527 42.8 as s 0 14192 0016019 2112 8 2112.8 12041 1068 7 10801 0 0242 2.1217 2.1459 44 8 45 8 015314 0 016020 1965 7 1965 7 14 047 1067.6 1081 6 002C2 2.1111 2.1393 43 8 41 8 0 16514 0 016021 1830 0 1830 0 16 051 10664 10825 0.0321 2.1006 2.1327 48.8 50 0 0 17796 0 016023 1704 8 17048 18 054 10653 1083.4 0 0361 2.0901 2.1262 58.8 52 8 019165 0 016024 1589 2 1589 2 20 057 10642 1084 2 0 0400 2.0798 2.1197 52 5 58 8 020625 0 016026 1482.4 14P1.4 22 058 10631 10851 0043 2 0695 2 1134 54.8 SE S 022183 0 016028 1333 6 1383 6 24 059 1061.9 1086 0 0 0478 2 0593 2.1010 56 8 5;8 023843 0 016031 1292.2 1292.2 26 060 1060 8 1086 9 0 0516 2 0491 2.1008 58 8 58 8 025611 0 016033 1207.6 1207.6 28 060 10591 10871 G0555 2.0391 2 0946 88 8 52 8 027494 0016036 1129.2 !!29 2 30D59 1058 5 1088 6 0 0593 2.02S1 2.0885 E2.8 54 8 0 29497 0 016039 1056.5 1056 5 32.058 10574 1089 5 0 0632 2.01c2 2 0824 54 8 El 8 0 31626 0 016043 989 0 9891 34 056 10563 10904 0 0670 2 00M 2 0764 EE 8 El D 0338739 0 016046 926.5 926 5 36 054 1055 2 1091.2 0 0708 1.999G 2 0704 58 8 78 8 0 36292 0 016050 8683 868.4 38 052 1054 0 1092.1 0.0745 1.9900 2 0645 78 8 72 8 0'8844 0 016054 814 3 814 3 40 049 1052.9 1093 0 0 0783 1.9804 2 0587 12.8 74 8 0 41550 0 016058 764.1 7641 42 046 1051.8 1093 8 0 0821 1.9708 2.0529 24 8 76 8 0 44420 0 016063 717.4 717.4 44.043 10507 10941 0 0858 1.%I4 2.0472 7E S 78 8 0 47461 0 016067 673.8 673 9 46 040 1049 5 1095.6 0 0893 1.9520 2.0415 78 8 88 8 050683 0.016072 6333 633 3 48 037 1988.4 1096.4 0 0932 1.9426 2 0359 IS 8 57 8 054093 0 016077 595.5 595 5 50 033 .1047.3 10973 0 0969 1.9334 2 0303 52 8 Ss 3 0 5.~102 0 016082 560 3 560 3 52 029 10461 1098 2 0.I006 13242 2 0248 54 8 86 8 0 61518 0 016087 227.5 527.5 54 026 1045 0 1099 0 0.1043 1.9151 2 0193 IE S 88 8 065551 0 016093 4%8 496 8 56 022 1043 9 1099 9 0.1079 1.9060 2.0139 IS 8 98 8 0 63813 0 0160'39 468 1 4681 58 018 10421 1100 8 01115 1 8970 200F6 38 8 37 8 074313 0 016105 4413 441 3 60 014 1041 6 1101 6 01152 I8881 2 0011 32 8

. 34 8 0 79062 0 016111 4163 41E 3 62 010 1040 5 1102 5 O!!88 18702 19M0 33 8 35 0 0 34012 0 016117 392 8 392 9 64 006 1039 3 11033 0 1224 1 8704 1 9928 96 8 SII O89356 0 016123 370 9 370 9 66 003 1938 2 1I04 2 0 1260 -! 8517 19876 38 8

Abs Prus. Specific Volume Enthafpy intropy Temp Lt per Sal. Sal. Sat. Sal. Sat. Sal. Temp Fahr Sq In. tiquid Evar vapor tiquid Evap Vapor liquid Evap Vapor Fahr 1 p eg vgg vg hg h rg hg sg sgg sg t leg g 0 94924 0 016130 350 4 350 4 67.999 1037.1 1105.1 0.1295 IJ530 13825 181 8 102 5 1 00789 0 016137 331.1 3311 69 995 1035 9 1105 9 01331 1.8444 13775 182.0 10t 8 106965 0 016144 313 1 313 I 71.992 1034 8 1106 8 01366 I2358 1.9725 led s IDE 8 11347 0 016151 29616 29618 7399 1033 6 !!O75 0 1402 1.8273 19675 ISE a 7598 1032.5 1108.5 0 1437 1 8188 1.9626 Ige s itB O 12030 0 016158 28028 28030 12750 0 016165 16537 265 39 77 98 1031.4 !!09 3 0.1472 1.3105 13577 lis B 118 8 117 8 1 3505 0 016173 25137 25138 79 98 1030 2 1110 2 01507 1 8021 1.9528 1172 Iis 8 I4299 0 016180 23821 238 22 81.97 10291 till 0 0.1542 13938 13480 114 3 -

I!E 8 1 5133 0 016185 22584 225 85 83 97 1027.9 t i ll.9 0.1577 13856 1.9433 115 e 214 21 85 97 IC268 11123 01611 13774 1.9386 Ils e til 5 16009 0 0161 % 214 20 16927 0 016204 203 25 20326 87.97 1025.6 1113 6 0 1646 13693 1.9339 120 8 173 0 13891 0 016213 192.94 19295 89 96 1024 5 1114 4 0 1680 13613 1.9293 122 8

.72 8 01715 13533 1.9247 124.0 I8901 0 016221 18323 183 24 91 96 1023 3 1115 3 124 8 12ill 1 9959 0 016229 17408 174 09 93 % 10222 11161 0 1749 13453 1.9202 17E D 178 9 2 !068 0 016238 16545 16547 95 96 1021.0 1117.0 (11783 13374 13157 178 8 22230 0 016247 15732 15733 97.56 1019.8 1117.8 01817 13295 13112 130.0 13e 5 23445 0 016256 149 64 149 66 9995 10183 1118 6 01851 13211 13068 132 e IIIa

> 134 8 24717 0 016265 14240 14241 101.95 1017.5 111*

0 1854 13140 13024 1348 2 6047 0 016274. 13555 13557 103 95 1016 4 T,: 01918 13063 88980 13E 8

& 135 8 0 016284' 129 09 129.11 105 95 1015 2 l' M 01951 1.6986 18937 13e 8 133 3 2 7438 2 8892 0 016293 122.98 123 00 If!7.95 1014S L- .3985 1.6910 13895 14ee 148 3 3 0411 0 016303 117.21 117.22 If,> 95 Ir*2F lh _ <t2018 1.6534 18852 1428 1s23 16759 13810 144.8 3 1997 0 016312 11134 Ill 76 111.95 16 i 3 1I23 6 12051 144 3 14G 10658 106 59 113 95 1010 ' 1824 5 02084 1.6684 1J769 lif 8 3 3653 0 016322 100> ' M.' 02117 1.6610 1.7727 1488 148 8 3 5381 0 016332 10168 10130 115 95 158 3 3 7184 0 016343 97.05 9707 117.95 10081 IIM.I 03150 1.6536 12605 150.8 3 9065 0 016353 92.66 92.63 - 119 95 1001.0 1126.9 0.2183 1.6463 1J646 1521 152 3 4 1025 0 016363 88 50 8852 121.95 10058 11273 02216 1.6390 1J606 1548 1543 0 2248 1.6318 IJ566 ISE 3 4 3068 0016314 84 56 8451 123.95 10046 1128 6 15C8 4 5197 0 016384 8082 8083 125 96 -1003.4 1129.4 02281 1.6245 12526 158.8 153 0 4 7414 0 016395 7717 7729 127.96 1002.2 1130 2 0 2313 1.6174 13487 168.0 150 3 4 9722 0 016406 73 90 7392 129 % 1001.0 1131.0 0 2345 1.6103 13443 162 8 182 3 5 2124 0 016417 70 70 70 72 131.96 9998 1131.8 02377 1.6c32 12409 164.8 164 8 54623 0 016428 6767 6768 13397 598 6 1132.6 0 2403 15961 12371 155 8 IEE 8 5 1223 0 016440 64 78 64 80 -

135.97 997.4 1133 4 0 2441 15892 18333 168 I 163 3 0 016451 62 04 6206 13737 9962 1134.2 02473 1582Z 1 8255 178.8 tre s 5 9926 1135 0 0 2505 15753 I8258 172 s FI 3 (' 736 0016463 5943 5945 13998 995 0 56 95 5697 14198 , 9938 1135 8 0 2537 1.5684 18221 174 8 174 8 i?656 0 016474 II36 6 0 2568 15616 18184 IIE F 6s690 0 016486 5459 54 61 143 99 992 6 UE8 a% 99 991 4 1137.4 02600 15548 12147 USS ais s 7IMO 0 016498 52 35 5236

m ^

Abs Press Specific Volume Enthalpy Entropy Temp tb per Sat. Sat. Sat. Sat. Sst. Sat. Temp fahr Sq In. Liquid Evap Vapar Liquid Evap Vapor Liquid Evap Vapor Fahr i p v, vfg vg hf h fg hg 5f Sig S g I 1800 7.5110 0 016510 50.21 5022 148 00 990.2 1138.2 0.2631 1.5480 1.8111 188 3 182 0 7.850 0 016522 48.172 18 189 150 t.1 989 0 1139.0 0.2662 1.5413 1.8075 182.8 184 0 8 203 0 016534 46 232 46.249 152 01 987.8 1139 8 0.2694 1.5346 1.8040 184 0 186 0 8568 0 016547 44.383 44 400 154 02 986 5 114n 5 0 2725 1.5279 I8004 186 9 188 0 8 947 0 016559 42 621 42.638 156 03 985.3 1141.3 0 2756 1.5213 13959 188 0 190 0 L.0 0 016572 40.941 40.957 158.04 984.1 1142.1 02787 1.5148 17934 ISO B 192 0 9347 001E585 39.337 39.354 160 05 982.8 1142.9 0 2818 1.5082 17900 192.G 19:G 10 168 0 016598 37.808 37.824 162.05 981.6 11433 0 2C48 1.5017 13865 1940 1960 10 605 0 016611 36.348 36.364 164 06 980.4 1144 4 0 2879 1.4952 1.7831 1960 1980 11.058 0 016624 34.954 34.970 166 08 979.1 1145 2 0 2910 1.4888 13798 198.0 200 0 11 526 0016637 33.622 33.639 168 09 977.9 11460 02940 1.4824 13764 200.8 i 12 512 0 0!6664 204 0 31.135 31151 17211 "15 4 1147.5 03001 1.4697 13698 204 0 208 0 13 568 0 016691 28.862 28878 176.14 :72.8 1149 0 0 3061 1.4571 1.7632 2080 3 212 0 14 696 0 016719 26382 26199 180.17 970.3 1150.5 0.3121 1.1447 1 7568 2120 a 216 0 15.901 0 016747 24 878 24 894 184.20 567.8 1152.0 0 3181 1.4323 13505 216.0 I 2to o 17.186 0.016775 , 23.131 23.148 188 23 965.2 1153.4 03241 1.4201 13442 2200 2240 18.556 0.016805 21.529 21.545 192.27 962.6 1154.9 0 3300 1.4081 17380 224 8 2280 20.015 0 016834 20.056 20 073 19631 960 0 1156.3 0 3359 1 3961 13320 228 8 2320 21.567 0 016854 18.701 18318 200 35 957.4 1157.8 0.3417 13842 13260 2320 2360 23.216 0 016895 17.454 17.471 7 G3 954.8 11592 0 3476 1.3725 13201 2360 240 0 24.968 0.016926 16??4 16.321 20845 952.1 1160 6 0.3533 13609 13142 24s e 2440 26826 0 016958 Is % 15.260 212.50 349.5 1162.0 0 '591 1.3494 11085 2440 2480 28 796 0 016990 14.264 14.281 216.56 946 8 1163.4 0.1649 1.3379 13028 2488 252 0 30 883 0 017022 13.353 13 375 220 62 944.1 11643 0 3706 1.3266 16972 2520 256 0 33 091 0 017055 12.520 12.538 224 69 941.4 1166.1 0 3163 1 3154 1.6917 2560 260 b 35 427 0 017089 11.745 11.762 228 76 938 6 1167.4 03819 1.3043 16862 250 0 264 0 37.894 0 017123 11.025 11.042 232 83 935 9 11683 03816 12933 16808 2640 268 0 40300 0 017157 10.358 10.370 23691 9331 1170 0 0 3932 13823 1 6755 2680 212 0 43 249 0 017193 9 738 9 755 240 99 930 3 1171 3 03987 I2715 I5702 272 0 276 0 46147 0 017228 9.162 9180 24508 927.5 1172.5 04043 I26C7 1.3550 276 0

~

49 200 280 0 0 017264 8 627 8 644 24917 924 6 1173 8 04098 12501 16599 280 0 284 0 52 414 0 01730 8 1280 8 1453 253 3 9213 1175 0 0 4154 12395 'l6548 281 0 2330 55 795 0 01734 7.6634 7.6807 257.4 918 8 11762 0 4?08 1 2290 16498 2880 292 0 59 350 0 01738 73301 72475 261.5 915 9 1177.4 04263 1 2186 1 6449 232 0 296 0 63 034 0 01741 6 8259 6 8433 265 6 913 0 1178 6 04317 12092 16400 295 0

Abs Press. Specific Yolume Enthalpy Temp tb per Entropy Sat. Sal. Sat. Sat. Sat.

Fahr Sq In. Liquid Sal. Temp Evap Vapor Liquid Evap Vapor Liquid Evap Vapor Fahr t p vg vig vg hr h ig hg sg sfg . S E I 300 8 67 005 001745 6 4483 64658 269.7 910 0 1179 7 0.4372 1.1979 354 8 71.119 0 08749 6 0955 1.6351 3000 6 1130 273 8 907.0 1180 9 04426 398 O 75 433 1.1877 1.6303 334 0 0 01753 51655 5 7830 278 0 904 0 1182.0 3125 79 953 0 01757 0 4479 1.1176 1.6256 308O 54566 5.4742 282.1 9010 1183.1 -

316 4 84 688 04533 1.1676 16209 312 0 0.01761 51673 51849 286 3 897.9 1184.1 0.4586 1.1576 1.6162 316 0 320 a 80643 001766 48%1 4 9138 290 4 894 8 1:85 2 3248 i 326 0 01770 4.6418 4.6595 294 6 891.6 1186 2 O.464G 1.1477 1.6116 320 8 3288 100 245 0 4692 1.1378 1.6071 3240 0 01774 4 40M 4 4208 298 7 888.5 1187.2 332 8 105 907 04745 1.1280 1.6025 3280 0 01779 4.17N 4 1966 302.9 885 3 1188.2 336 8 111 820 0.4798 1.1183 1.5981 J32.0 0 01783 3.9681 3 9859 307.1 882.1 1189 1 0.4850 1.1086 1.5936 3360 340 0 117 992 0 01787 3.7699 3 7878 311.3 878.8 11901 04902 344O 124 430 0.01792 3.5834 3 6013 315 5 10990 1.5892 340 8 343 0 815.5 1191 0 0 4954 10894 1.5849 131.142 001797 34018 3 4258 3440 319 7 8 72.2 1191.1 0 5006 352 0 138138 0 01801 3 2423 3 2603 323 9 10799 1.5806 3480 868 9 1192.7 0 5058 356 0 145 424 0 01806 30863 3 1044 1.0705 1.5763 352.0 3281 865.5 1193 6 05110 10611 1 1.5721 356.0 i > 360s 153 010 0 01811 2.9392 2.9573 332 3 862.1 1194 4 364 0 160 903 0 5161 10517 1.5678 JE0 0

0 01816 2.8002 2.8184 336.5 858 6 3680 1195 2 0 5212 1 0424 1.5637 169 113 0 01821 2 6691 2.6873 340 8 3648 855.1 1195.9 0 5263 312 O I77648 0 01826 2.5451 2.5633 345 0 1 0332 1.5595 3680 851 6 1196 7 0 5314 10240 1.5554 376 O IS6517 0 01831 2.4219 2.4462 349 3 8481 372 8 1197.4 0 5365 1.0148 1.5513 316.8 380 0 195 729 0 01836 2.3170 2.3353 353 6 844 5 !!98 0 0.5416 10057 384 0 205 294 0 01842 2.212n 1.5473 360 0 2 2304 3579 E40 8 11981 3I80 215 220 0 01847, 2.1126 05456 0 9966 1.5432 3840 2.1311 362.2 837.2 1199 3 0 5516 392.0 225 516 001853 2.0184 09876 1.5392 3888 2.0369 366 5 833 4 1199.9 0 5567 0 9786 396 0 236 193 001858 1.9291 1.5352 3928 1.9477 3708 829 7 1200 4 0 5617 0 9696 1.5313 396 s 400 0 247.259 0 01864 1.8444 1.8630 3751 E25.9 1201.0 0.5667 09607 404 0 258 725 001870- 13640 13827 1.5274 400.8 379 4 822 0 1201.5 0 5717 09518 438 8 270 600 0'01875 1.6877 13064 1.5234 404 e 383 3 818 2 1201.9 0 5766 0 9429 412 0 282 894 001881 1.6152 16340 1.5195 408 0 3881 814 2 1202.4 0 5816 0 9341 416 0 295 617 0 0i887 1.5463 1.5157 412.0 1.5651 392 5 810 2 1202.8 0 5866 09253 1.5118 416.0 420 0 308 780 0 01894 1.4808 1.4997 396 9 806 2 12031 0 5915 0 9165 4240 322 391 0 01900 1.4184 1.5080 428 0 1.4374 401.3 802 2 1203.5 0 5964 0 9077 428 0 336 463 001906 1.3591 1.5042 424.0 13782 405 7 7980 12031 0 6014 0 8990 432 0 351 00 0 01913 1.30266 1.5004 428 0 1.32179 410 1 793 9 1204 0 0 6063 0 8903

435O 366 03 001919 1.24887 1 4966 4320 I26806 414 6 789 7 1204 2 0 6112 0 8816 1.4928 436 8 440 0 381 54 0 01926 1.19761 1.21687 419 0 7854 1204 4 061GI 0 8729 NIO 397$6 0 01933 1.14874 1.4890 448 8 1.16806 d23 5 781.1 1204 6 0 6210 0 8643 ist D 414 09 0 01940 1.10212 erg 0 14853 4aa e 1.12152 776 7 12047 06259 4W 1520 431 14 0 01947 1 05764 1 01711 ezS 0 8557 ~1.4815 772 3 1204 8 0 6308 4%a 113 73 0 01054 I0151R I01412 4170 7678 1204 8 05356 0 8471 14Dd +C#

0 8385 14741 456 o

Abs Press. Specific Volume Enihalpy Entropy Temp lb per Sat. Sat. Sat. Sat. Sat. Sat. Temp fahr Sq In. Liquid Evap Vapor Liquid Evap Vapor Liquid Evap Vapor Fahr i p V, veg vg he h fg hg se sig S g I 460 0 466 8 f 0.01961 0.97463 0 99424 441.5 763 2 .204 8 0.6405 0 8299 1.4704 460.0 464.0 485 56 0 01969 0 93588 0 95557 4461 758 6 1/04 7 0 M54 0 8213 1.4667 4649 4680 504 83 0 01976 0 89885 0.91862 4333 754 0 1204 6 0.6502 0 8127 1.4629 468I 4720 524 67 0 01984 0 86345 0 88329 4A2 749 3 12R5 0.6551 0 3042 1.4592 472.5 476 0 54511 001992 0 82958 0 84950 459 9 744.5 12043 0 6599 - 07956 1.4555 476 >

480 0 566.15 0 02000 039716 0 81717 464 5 739.6 1204.1 0.6648 07871 1.4518 480.0 484 0 587.81 0 02009 036613 0 78622 4691 7343 1203 8 06696 0.7785 1.4481 4848 4880 610.10 0 02017 033641 035658 473 6 7293 1203.5 0.6745 0 7700 1.4444 4888 497 0 633 03 0 02026 0 70794 0 72820 478 5 724 6 1203.1 0 6793 0 7614 1.4407 492.5 4960 656 61 0 02034 0 68065 010100 483 2 719 5 12023 0 6842 0 7528 1.4370 496.8 5000 680 86 0.020t3 0.65448 067492 487.9 7143 1202.2 0 6890 0.7443 1.4333 588.0 5040 70538 0 020i3 0.62938 0 64991 492.7 709 0 12013 0 6939 07357 1.42 % 504.!!

5080 731.40 0 02062 0 60530 0 62592 497.5 7033 1201.1 0.6987 03271 1.4258 50t.8 y 512 0 75732 0 02072 0 58218 0 60289 502.3 698.2 1200.5 0.7036 03185 1.4221 512.0 y

. 515 0 784 76 0 02001 ,,0 55997 0.580/9 507.1 G92.7 1199.8 0.7085 03099 1.4183 516 8 520 0 812.53 0 02091 0.53864 0 55956 . 2.0 687.0 1199.0 0.7133 01013 1.4146 520.0 524 0 841 04 0 02102 0 31814 0.53916 516.9 6813 1198.2 01182 0 6926 1.4108 524.0 528 0 870 31 0.02112 0 49843 051955 521.8 6755 11973 01231 0 E839 1.4070 528 0 5320 900 34 0 02123 0.47947 0.50070 526 8 669.6 1196 4 03280 0 6752 1.4032 532.8 5360 931.17 0 02134 0 46123 0.48257 531.7 663.6 1195.4 01329 0 6665 13993 536.5 540.0 962.79 0 02146 0.44367 0.46513 536.8 657.5 1194.3 03378 0.6577 1.3954 546.5 5440 995 22 0 02157 0 42677 0 44834 541.8 6513 1193.1 0 7427 06489 13915 544.5 5480 1028 49 0 02169 0 41048 0 43217 5469 645.0 1191.9 03476 0 6400 13876 5488 552 0 1062.59 0 02182 039479 0 41660 552 0 638 5 1190.6 03525 0 6311 13837 552 8 556 0 1097.55 0.02194 037966 340160 557.2 632.0 1189.2 03575. 06222 13797 556.8 560 0 1133 38 0 02207 036507 0 38714 562.4 6253 1187.7 03625 0 6132 1.3757 560 8 5640 1170 10 0 02221 0 35099 0 37320 567.6 618.5 1186.1 03674 0 6041 1.3716 564 I 568 0 120732 0 02235 033741 0 35975 572.9 611.5 1134.5 03725 0 5950 13675 568 0 5720 1246.26 0 02249 0 32429 0 34678 578 3 604.5 11823 03775 0 5859 1.3634 572.8 5760 1285.74 0 02264 0 31162 033426 583.7 597.2 1180.9 03825 0 5766 13592 576 8 5800 1326 17 0 02279 039937 0 32216 589.1 589.9 '1179.0 03876 0 5673 1.3550 580.8 584 0 13673 0 02295 0 28753 031048 594.6 582 4 1176.9 0 7927 05580 13507 584 5

$88 0 1410 0 0 02311 0 27608 0 29919 600 1 5743 1174 8 0.7978 0 5485 1.3464 588 8 5920 1453 3 0 0?328 0.26499 0 28527 605 7 566 8 1172.6 0.8030 0 5390 1.3420 592d 556 0 14978 0 02345 0.25425 0 27770 611.4 558 8 1170 2 0 8082 0 5293 1 3375 596 0

8%

~

Abs Press. Specific Volume Enthalpy Entropy Temp 1.b per Sal. Sat. Sat. . Sat. Sat. Sat. Temp fahr SqIn. Liquid Evap Vapor Liquid Evap Vapor Liquid Evap Vapor Fahr i p vg vfg vg hg h gg hg se sgg sg t 600 0 15432 0 02364 0 24384 0 26747 C17.1 550.6 1167.7 0.8134 0.5196 1.3330 E08.8 E04 0 1589 7 0 02382 0.23374 0.25757 622.9 542.2 1165.1 0 8187 0.5097 1.3284 504 0 608 0 1637.3 0 02402 0.22394 0.24796 628 8 533.6 1162.4 0 8240 0.4997 1.3238 EOS.O E12.0 1686.1 0 02422 0 21442 0 23865 634 8 5243 1159.5 0.8294 04896 1.3190 E12.E 640 8 515.6 1156 4 0.8348 0.4794 1.3141 515.0 E16 6 1735.9 0 02444 0 20516 0 22960 506.3 1153.2 0 8403 0.4689 1.3092 E20.0 E20 0 1786.9 0 02466 0.1%I5 0 22081 646.9 1839 0 0 02489 0.18737 0 21226 653 1 406 6 1149.8 0 8458 0.4583 1.304i E24.5 624 0 628 0 1892 4 0 02514 0.17880 0 203 % 659 5 486 7 1146.1 0.8514 04474 1.2988 628 8 0 19583 665 9 476.4 1142.2 0 8571 0.4364 1.2934 532.8 532 0 19470 0 02539 0.17044 672.4 4653 11381 0 8628 0 4251 1.2879 635.8 536 0 2002 8 0G2566 0.16226 0.18792 E40 0 20599 0.02595 0.15427 0.13021 679.1 454.6 1133.7 0.8686 0.4134 1.2821 548.8 644 0 2118 3 0 02625 0.14644 017269 685.9 443.1 1129 0 0.8746 0.4015 1.2761 E44.0 648 0 21781 0 02657 0.13876 0 15534 692.9 431.1 1124.0 0.8806 0.3893 1.2699 E48 I E52.0 22392 0 02691 0 13124 0.15816 700 0 4181 1118 7 0.8868 0.3767 1.2634 ES2 O 0 02728 0.12387 0.15115 707.4 4051 1113.1 0.8931 0.3637 1.2567 E55 0 Y

656 0 23013 650 0 2365.7 0 02768 0.11663 0.14431 714.9 392.2 1107.0 0.8995 0.3502 1.2498 560.0 EE4 0 2431.1 0 02811 0.10947 0.13757 722.9 377.7 1100 6 0.9064 0.3361 1.2425 664.0

' EE8 0 24981 0 02858 0.10229 0.13087 731.5 362.1 1093.5 0.9137 0.3210 1.2347 EE8.8 612 0 2566 6 0 02911 0 09514 0.12424 7402 3451 1085.9 0.9212 0.3054 1.2266 E72.9 EI6 0 26368 0.02970 0.08799 0.11769 74 9.2 328.5 1077.6 0.9287. 02892 1.2179 E75.0 E80 0 2708 6 0 03037 0.08080 0.11117 758 5 310.1 1068.5 0.9365 0.2720 1.2086 588.8 E84 0 2782.1 0 03114 0.07349 0.10463 768.2 290 2 1058.4 0.9447 0.2537 1.1984 584.0 688 0 28574 0.03204 0 06595 0 09799 778 8 268.2 1047.0 0.9535 0.2337 1.1872 E88.I E92.0 2934 5 0 03313 0.05797 0 09110 790.5 243.1 1033.6 0.9634 0 2110 1.1744 592.8 596 0 3013.4 0.03455 0.04916 0 08311 804.4 212.8 1017.2 0.9749 0.1841 1.1591 596.0 700 0 3094.3 0.03662 0 03857 0 07519 822.4 172.7 995.2 0.9901 0 1490 1.1390 - 100.0 102 0 3135.5 0 03824 0 03173 0 06997 835 0 144.7 9791 1.0006 0.1246 1.1252 702.5 3177.2 0 04108 0 02192 0 06300 854 2 102.0 956.2 - 10169 0 0876 . 1.1046 704.0 704 0 105 0 3198 3 0.04427 0.01304 0 05730 873.0 61.4 934.4 1.0329 0 0527 1.0856 705 O 105 47* ?208 2 0 05078 0 00000 0 05078 906 0 0.0 906 0 1.0612 0 0000 1.0612 705.47*

Critical temocrature

e.

~

Table 2: Saturated Steam: Pressure Table Specific Volume Enthalpy Entropy Abs Press. Temp Sat. %t. Sat. Sat. Sat. Sat. Abs Press.

lb/Sg in. Fahr liquid Evap ,apor Liquid Evap Vapor Liquid Evap Vapor LblSq In.

p t vg *g hg h fg h g sg s fg S g E vfg 3 0003 1075.5 1075.5 0 0000 2.1872 2.1872 C08865 008865 32 018 0 016022 3302.4 3302.4 1087.4 0 0542 2.0425 2.0967 G 25 0 25 59 323 0 016032 1235 5 1235 5 27.382 1060.1 641.5 47.623 1048 6 1096 3 0 0925 1.9446 2.0370 0 50 0 50 79 586 0 016071 641.5 to 101.74 333 59 333 60 69 73 10361 11058 0.1326 1.8455 1.9781 '10 t 0 016136 50 162 24 0016407 73.515 73 532 13020 1000 9 121.1 0 2349 1.6094 1.8443 50 38420 161.26 9821 !!43 3 0 2836 1.5043 1.7879 10 0 10 0 193 21 0 016592 38404 14 695 212 00 0 016719 26 782 26 799 180 17 970 3 1150 5 0 3121 I4447 1.7568 14 696 15 0 213 03 0 016726 26.274 26.290 181.21 969.7 1150.9 0.3137 1.4415 1.7552 15 0 20 0 227.96 0 016834 20 070 20.087 19627 960 1 11563 0 3358 1.3962 1.7320 20 0 30 0 250 34 0 017009 137266 13 7436 218 9 945 2 11641 0 3682 1.3313 1.6995 30 0 40 4 267.25 0 017151 80.4794 10.4965 236.i 933 6 1169 8 0.3921 1.2844 1 6765 ao s

! 50 0 28102 0 017274 8.4967 8 5140 250 2 923 9 1174.1 0.4112 1.2474 1 6586 50 0 60 0 25'2 71 0 017383 7.1562 7.1736 262.2 315 4 1177.6 0.4273 1.2167 1.6440 60 0 70 0 302 93 0 017482 6.1875 6 2050 272.7 907.8 1180 6 0 4411 1.1905 1 6316 70 0

> 80 0 312 04 0 017573 ~5.4536 5.4711 282.1 900 9 1183! O4534 1.1675 1.6208 80 0 I> 90 0 320 28 0 017659 .48779 4.E953 290.7 894 6 11855 0.4643- 1.1470 1.6113 50 0 100 0 32782 0 017740 4.4133 4.4310 298.5 88 5 1187.2 0.4743 1.1284 1.6027 100.0 1103 334 79 0 01782 4 0306 4.0484 305 8 883.1 1188.9 0 4834 1.1115 1.5950 11GO 111L O 341.27 0 01789 3.7097 3 7275 312 6 877.8 m34 0.4919 1.0 % 0 1.5879 120 0 130 0 347.33 0 01796 3.4364 3.4544 319 0 872 8 1191., 0 4998 1.0815 1.5813 130 0 i

140 0 353 04 0 01803 3.2010 3.2190 325c 868 0 1193 0 0.5071 1.0681- 1.5752 140 0 150 0 35843 0 01809 2.9958 3 0139 330 6 863.4 1194.1 u:'41 1.0554 1.5695 1500 153 8 363 55 0 01815 2.8155 2.8336 3361 859 0 1195.1 0 5206 1.0435 1.5641 160 0 1700 368 42 0 01821 2.65'S 2.6738 341.2 854 8 1196 0 0 5269 a.;M 1.5591 170 W 180 0 373 08 0 31827 2.5129 2.531.- 346.2 850 7 1196.9 0.5328 1.0215 1.5543 1800 1900 377.53 0 01833 2.3847 2.403b 350 9 846.7 1397.6 0.5384 1.0113 1.5498 190 0 3a1.80 0 01839 2.2689 2.2873 355.5 842.8 1198 3 0 5438 1.0016 1.5454 2000 2000 210 0 38591 0 01844 2.16373 2.18217 359 9 839 1 1199 0 0 5090 0.9923 1.5413 210 0 2200 389 88 0 01850 2 06779 2.08629 364 2 835 4 1199 6 0 5540 0.9834 1.5374 220 0 230 0 39370 0 0185; 1.97991 139846 368 3 831.8 12001 0.5588 09748 15336 2300 240 0 397.39 0 01860 1.89909 1.91769 372 3 828 4 1200 6 0.5634 0 9665 1 5299 240 0 250 0 40G 97 0 01835 1 82452 1.84317 376 1 825.0 1201.1 0.5679 0 9585 1.5264 250 0 260 0 40444 0 01870 175548 177418 379 9 821 6 1201.5 0 5722 0 9508 1.5230 260 0 270 0 40780 0 01875 169137 1.71013 383 6 818.3 1201.9 0 5764 0 9433 15197 270 0 280 0 411 07 0 01880 1 63169 1 65049 387.1 815.1 1202.3 0 5805 0 9331 1 5166 250 0 233 0 414 25 0 01885 1.57597 1.59482 390 6 812 0 1202 6 0 5844 0 9291 1 5135 290 0 300 0 41735 0 01889 1.52384 1.54274 3920 808 9 1202.9 05882 0 9223 15105 300 e no0 411 11 001912 130647 1 37 % 1 4aq 8 7942 1704 0 0 f>0VI O8909 I 4'M 3500

Specific Volume Enthalpy Entropy Abs Press. Temp Sat. Sat. Sat. Sat. Sat. Sat. Abs Press. .

lb/Sq In. fahr liquid Evap Vapor Liquid Evap Vapor Liquid Evap Vapor. tb/Sq In.

p t vg hg h hg 3 3g vig 't is 1 S

ig P 450 0 456 28 0 01954 1.01224 1.03179 437.3 767.5 1204.8 0 6360 0.8378 1.4738 450 0 5000 46701 001975 0 90787 0 92762 449.5 755.1 1204.7 06490 0 8148 1.4639 500 8 550 0 47694 0 01994 0 82183 0.84177 460.9 74!.3 1204.3 0 6611 0 7936 1.4547 5500 600 0 486 20 0 02013 0 74962 0 76975 471.7 732 0 1203.7 0 6723 03738 1 4461 6000 650 0 494 89 0 02032 0 68811 0 70843 431.9 720.9 1202.8 0.6828 07552 1.4381 650e 700 0 503 08 0 02050 0 63505 0 65556 491.6 710 2 1201.8 0.6928 0 7377 1.4304 700s 750 0 510 84 0 02069 0.58880 0 60949 500.9 699 8 :200 7 03022 03210 1.4232 1540 100 0 518 21 0 02087 0 54809 0 56896 509 8 689 6 1199.4 01111 03051 1.4163 8000 850 0 525 24 0 02105 051197 053302 518.4 679.5 1198 0 03197 0 6899 1.4096 350.5 900 0 531.95 0 02123 0 47968 050091 5263 66S 7 1196 4 0 7279 0 6753 1.4032 30s e 550 0 53839 0 02141 0 45064 0.47205 534 7 660 0 l'?43 03358 0 6612 1.3970 950 8 10L00 544 58 0 02159 0 42436 0.44596 542 6 650.4 1192.9 03434 0 6476 1.3910 10008 1050 0 550 53 0 02177 040047 0 42224 550.1 6409 1191.0 0 7507 0 6344 1.3851 1050 0 1100 0 556 28 0 02195 0 37863 0 40058 557.5 631.5 11891 03578 0 6216 1.3794 1100 8 1150 0 561 82 0 02214 0.35859 038073 564.8 622.2 1187.0 03647 0.6091 1.3738 1150 0 12000 56719 0 02232 0.34013 0 36245 571.9 613.0 1184 8 01714 0.5969 1.3683 12000 1250 0 572.38 0 02250 .132306 0.34556 578 8 603.8 1182.6 07780 0.5850 1.3630 1258 8 13000 577.42 0 02269 b W 22 0.32991 585 6 594.6 1180 2 0 7843 0 5733 1.3577 1300 8 1350 0 582.32 0 02288 0 29250 0 31537 592.3 585 4 1177 8 07906 0 5620 1.3525 135s0 4 14000 58707 0 02307 0 27871 0 30178 5988 576 5 1175.3 03966 0.5507 1.3474 1400 8

>' 1450 0 59130 0 02327 0.26584 028911 605 3 567.4 1172.8 0 8026 05397 1.3423 14508 1500 0 596 20 0 02346 0 25372 0 27719 6113 558 4 1170.1 0.8085 0 5288 1.3373 15000 6 1550 0 60059 0 02366 0 24235 0 26601 618 0 549 4 1167.4 0 8142 0 5182 1.3324 1550 0 16000 604 87 0 02387 0 23159 0 25545 624 2 54n 3 1164.5 0 8199 0 5076 1.3274 16008 16508 609 05 0 02407 0 22143 0 24551 630 4 531.3 1161.6 0 8254 0 4971 1.3225 16500 17000 613 13 0 02428 021178 0 23607 636.5 522.2 1158 6 0.8309 0.4867 1.3176 17000 1750 0 617.12 0.02450 0.20263 0.22713 642.5 513.1 1155 6 0 8363 0 4765 1.3128 1750 8 1800 0 621 02 0 02472 0.19390 0.21861 648 5 503 8 1152.3 0 8417 0.4662 1.3079 18000

. 1850 0 624 83 0 n2495 0.18558 0 21052 654 5 494 6 1149 0 0 8470 0.4561 1.3030 1850 e 19000 628 56 0 02517 0.11761 0.20278 660 4 485.2 1145 6 0 8522 0 4459

1.2981 1900 0 1950 0 632 22 0 02541 0.16999 0.19540 E66.3 475 8 1142.0 0 8574 0.4358 1.2931 1950 0 20000 635 80 0 02565 0.16266 0.18831 672.1 466 2 1138 3 0 8625 0 4256 1.2881 20000 2100 0 64236 0.0?615 0.14885 0 17501 683 8 4463 1430.5 0 8727 04053 1.2780 21000 22000 64945 0 02669 0 13603 0 16272 695 5 4263 1122.2 0 8828 0.3848 1.2676 2200.0 2300 0 655 89 0 02727 012406 0 15133 707.2 4060 1113 2 0 8929 0.3640 1.2569 23000 2400 0 662.!! 0 02790 0.11287 0.14076 719 0 384.8 11033 0.9031 0.3430 1.2460 24000 2500 0 66811 0.02859 0.10209 0.13068 731.7 361.6 1093.3 0 9139 0.3206 1.2345 2500.5 2600 0 673 91 0 02933 0 09172 0.12110 744.5 3376 1082.0 0 9247 0 2977 13225 2600 0 21000 679 53 0 03029 0 08165 0.11194 -757.3 312.3 1069 7 0 9356 0 2741 1.2097 21000 2800 0 634 36 0 03134 0 07171 0 10305 770 7 285 1 1055.8 0 9468 0 2491 1.1958 2800 0 7900 0 690 22 0 03262 0 06158 0 09420 785 1 254 7 1039 8 0 9588 0 2215 1.1803 2900 d 30000 695 33 0 03428 0 05073 0 08500 801 8 218 4 1020 3 0 9728 0 1891 1.1619 3000 0 3I00 0 700 28 0 03681 0 03771 0 07452 824 0 169 3 993 3 0 9914 0 1460 11373 31000 3700 0 705 08 0 04472 0 01191 0 05663 875 5 561 931.6 1 0351 0 0482 1 0832 3200 0 N082- 70547 0 05078 0 00000 0 05078 00& O 00 906 0 1 0612 00300 1 0612 3208 2-

e g e' o a

/ l I

w I l

Table 3. Supelhe.iled Ste.u.e

.Aln i'ren LtnSo la 5al 541 lemoetals'e-Delins f aiweWil (Sal tem 0) weiei Steam 200 210 348 354 899 850 lit 600 Itt 840 999 litt litt f283 g Sn

, 0 01614 3336 H25 is 75422 lol264IM452?6 24I ?$ 29876 34876 398 26 49876 $78 ?6 698 26 790 ?6 896 ?6 993 tt 109 76 3 4871 O I'I till 544 7 578 ) 431 3 690 7 758 4 8et t 861 4 979 8 618 ,

h 69 73 1105 8 11507 1112 9 1895 7 tilB 7 1748 8 1265 1 1288 6 13M i IJ84 5 1433 7 14s3 8 1534 9 anal 8 1619 7 s $1326 1 9781 2 05C9 20H1 21152 2144S 2 3722 21985 2 2237 72108 2 3144 t hil 2 M34 2 4196 2 4648 2 4569 sa 37;6 0776 13773 18776 23776 18776 337 76 43i76 $33 !6 447 76 73776 81776 93714 303776 I

1362 241

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= 9095 0'84 $ 0 9996 1 0473 8 0953 4 1476 l is77 1 7781 1 1678 1 30?! 131M 170 4 0 0787 0 0314 00pg 0 0100 0 0418 O NS8 0 0496 0 0534 0 0573 0 0t 46 0 6714 i

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a t0 0778 est0aal 6747t 00?$8 0 07FI OO42 I?6f0t 13196 037913778 0 016014136 0 0393 16te0t 0479 0 N64 0 764 9 t

t>46 g47 6 loll t 1375 7 line 17003 15704

$h = su M' hell.I 0 $477 01466 1 0340 1 0797 1 I?47 1 1674 1 7073 1 2457 i F8ll I 3131 1 3 t = spes.fic volume. til !! per Ib h = enthalpy. Sfg per Ib 5 = Gril10py. Stu per R per Ib B-10 -

k

1 A

i l'

ANSWER KEY PRAIRIE ISLAND 1&2 RO EXAM PARKINSON, K.

88/07/11 Q

1. PRINCIPLES OF NUCLEAR POWER PLANT Q' .ATION. PAGE 31 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- PRAIRIE ISLAND 1&2 -BB/07/11-PARKINSON, K.

1 1

ANSWER 1.01 (1.00) l B. C A longer lif e time] [1.03 REFERENCE W3ctinghouse, Fundamentals of Nuclear Reactor Physics, 1983, page 7-30 l 192003K107 ...(KA'S)

ANSWER 1.02 (1.00) l D. CBy which the reactor is subcritica13 C1.03 REFERENCE Wantinghouse, Reactor Core Control for Large Pressurized Water Reactors, 1983, page 7-17 ,

192002K110 ...(KA'S)

J ANSWER 1.03 (1.00)

A. CWhether the given water-boron mixture absorbs more neutrons than it moderates or moderates more neutrons than it absorbs 3 C1.03 REFERENCE

)

W30tinghouse, Reactor Core Control for Large Pressurized Water Reactors, I pcg3 3-4 192004K101 192004K103 192004K110 192004K106 ...(KA'S)

ANSWER 1.04 (1.00)

B. CA decrease in the baron concentration during core life due to fuel burnup and the production of fission fragment poisons] C1.03 1

REFERENCE l

Wastinghouse, Reactor Core Control for Large Pressurized Water Reactors, 1983, page 3-28 192004K106 ...(KA*S) l l

I

l

1. PRINCIPLES OF NUCLEAR POWER-PLANT OPERATION. PAGE 32 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- PRAIRIE ISLAND 1&2 -BS/07/11-PARKINSON, K.

l ANSWER 1.05 (1.00)

C. [ Clad creep 3 [1.03 REFERENCE Wastinghouse, Reactor Core Control for Large Pressurized Water Rear. tors, 1583, page 2-45 192004K107 ...(KA*S)

ANSWER 1.06 (2.00)

A. a. C- 5413 pcm3 C1.03 B. a. C- 3428 pcm3 01.03 REFERENCE W3ctinghouse, Reactor Core Control For Large Pressurized Water Reactors, 1983, pages 17-21 through 17-23 192002K114 ...(KA'S)

ANSWER ).07 (1.00)

A. [ Positive due to the faster burnup of Zenon at higher neutron flux levels 3 E1.03 REFERENCE Wactinghouse, Reactor Core Control For Large Pressurized Water Reactors, 1983, page 4-24 192006K111 ...(KA'S)

- l

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATIDNi PAGE 33 1

, THERMODYNAMIC U EAT TRANSFER AND FLQLD fLQM I ANSWERS -- PRAIRIE ISLAND 1&2 -88/07/11-PARKINSON, K.

ANSWER 1.08 (1.00)

D. CMaintaining the rods above the rod insertion limit produces an axial flux distribution which prevents high local peak power levels 3 C1.03 REFERENCE I I

Wr=tinghouse, Reactor Core Control for Large Pressurized Water Reactors, 1983, page 6-30 192OO5K115 ...(KA'S)

ANSWER 1.09 (1.00)

B. CXenon is removed by decay and neutron absorption but Samarium is removed by neutron absorption only3 C1.03 l REFERENCE l W3ctinghouse, Reactor Core Control for Large Pressurized Water Reactors, 1983, pages 4-11 through 4-33 192006K103 192006K104 192006Kt16 192OO6K117 . . . (KA*S)

ANSWER 1.10 (1.00)

A. CAxial Flux Difference] CAFD3 C1.03 or C. CLocal Power Density [KW/ft33 REFERENCE W20tinghouse, Reactor Core Control for Large Pressurized Water Reactors, 1983, page 6-32 192005K115 ...(KA*S) 1 i

e w--- ,-- w . , - --r-----,.--,v.

y-.w, - , , , , , , ,,,,_.y., , ._.m, ..-_,__y ..%, ., _ ,_ g m. -ps -,,e---g.,---,,y,y 3----1

1. PRINCIPLES OF NypLEAR PbWER PLANT OPERATION. PAGE 34

,IBERMODYNAMICSr HEAT TRANSFER AND FLUID FLOR ANSWERS -- PRAIRIE ISLAND 1&2 -88/07/11-PARKINSON, K.

. l i

i I

r ANSWER 1.11 (1.00) i B. CAn immediate rapid rise followed by a gradual increase to a higher steady state value3 C 1. 0 3 REFERENCE '

W2ctinghouse, Fundamentals of Nuclear Reactor Physics, 1983, page B-54 192000K103 ...(KA'S)

ANSWER 1.12 (1.00)

C. CO.983 C1.03 REFERENCE W;stinghouse, Reactor Core Control for Large Pressurized Water Reactors, 1983, page 9-10 192000K104 ...(KA*S)

ANSWER 1.13 ti.00)

D. CStartup rate decreases 3 C1.03 REFERENCE W stinghouse. Reactor Core Control for Large Pressurized Water Reactors, 1983, page 9-17 192008K117 ...(KA'S) 1 1

ANSWER 1.14 (1.00) l l

A. Cincreasing turbine first-stage pressure, decreasing RCS baron .

enncentration and increasing rod height 3 C1.03 l

REFERENCE Wantinghouse, Reactor Core Control for Large Pressurized Water Reactors, 1983, page 9-22 192008K119 ...(KA'S) l 4

1

__ _ _ _ __ . _ - . . _ . _ _ _ _J

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 35 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- PRAIRIE ISLAND 1&2 -88/07/11-PARKINSON, K. I ANSWER 1.15 (1.00) ,

I A. CIndicated pressurizer level will be higher than actual pressurizer leve13 ti 03 i

REFERENCE W:stinghouse, Thermal-Hydraulic Principles and Applications to the i Proccurized Water Reactor II, 1982, pages 11-27 and 11-28 l 191002K109 ...(KA*S) 1 l

l i

ANSWER 1.16 (1.00)

A. CThe indicated steam flow will be greater than actual steam flow 3C1.03 I REFERENCE Wactinghouse, Thermal-Mydraulic Principles and Applications to the Proccurized Water Reactor II, 1982, pages 11-19 and 11-20 191002KiO2 ...tKA'S) l ANSWER 1.17 (1.00)

C. [ Increases, then decreases 3 C1.03 l

REFERENCE Wantinghouse, Thermal-Hydraulic Principles and Applications to the Prcoturized Water Reactor I, 1982, page 3-77 193008K102 ...(KA'S) l j

ANSWER 1.18 (1.00)

]

D. CReactor coolant flow decreassal C1.03

)

REFF3ENCE i W2stinghouse, Thermal-Hydratatic Principles and Applications to the l Pr occurized Water Reactor II, 1982, page 13-24 193008K105 ...(KA*S) 1 1

1. PRINCIPLEE OF NUCLEAR POWER PLANT OPERATION. PAGE 36

. , THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- PRAIRIE ISLAND 1&2 -BS/07/11-PARKINSON, K.

1 ANSWER 1.19 (1.00)

B. CRCP cavitation 3 C1.03 l REFERENCE Wactinghouse, Thermal-Hydraulic Principles and Applications to the PrCocurized Water Reactor II, 1982, page 10-54 ,

191004K106 ...(KA'SV ANSWER 1.20 (1.00)

B. CMaterial defect in the vessel wall, temperature below Nil Ductility Temperature CNDT3, tensile stress applied to the vessel wall] C1.03 REFERENCE Weatinghcuse, Thermal-Hydraulic Principles and Applications to the Proccurized Water Reactor II, 1982, pages 13-60 and 13-61 ;

193010K101 ...(KA*S)

ANSWER 1.21 (1.00)

C. CPressure increases and flow rate decreases 3 C1.03 ;

REFERENCE W:ctinghousa, Thermal and Hydraulic Principles and Applications to the I Proccurized Nater Reactor II, 1982, page 10-73 191001K103 ...(KA'S)

I I

c la PRINCIPLES OF NUCLEAR POWER PLANT OPERATIOMs .PAGE 37 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW j ANSWERS -- PRAIRIE ISLAND 1&2 -BG/07/11-PARKINSON, K. 1 4

ANSWER 1.22 (1.00) l A. CDelta-H Csteam enthalpy - feed enthalpy33 i or C. CSteam flow 3 C1.03 REFERENCE f Wactinghouse, Thermal-Hydraulic Principles and Applications to the Prcccurized Water Reactor II, 1982, pages 12-12 and 12-13 l 19200BK121 ...(KA'S) l ANSWER 1.23 (1.00) l A. CCooling water outlet temperature will decrease 3 C1.03 REFERENCE W2stinghouse, Thermal - Hydraulic Principles and Applications to the Preocurized Water Reactor, 1982, pages 9-8 through 9-20 191006K112 ...(KA'S)

ANSWER 1.24 (1.00) 1 B. CDecrease in steam pressure during constant decay heat generation 3 C1.03 REFERENCE W3stinghouse, Thermal - Hydraulic Principles and Applications to the Proccurized Water Reactor, 1982, page 14-27 19300BK122 ...(KA'S) l l

l

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

2. PLANT DESIGN INCLUDINE_ SAFETY AND EMERGENCY SYSTEMS PAGE 38 ANSWERS -- PRAIRIE ISLAND 1&2 -88/07/11-PARKINSON, K.

ANSWER 2.01 (2.50) i A. 1. To ensure that the PZR boron concentration CO.1253 is equal to the reactor coolant boron concentration CO.1253. [0.253

2. To prevent thermal shock CO.13 to the piping CO.0753 and the spray nozzle Co.0753 CO.253 B. Prevant the leakage of any thydrogen3 gas CO.253 which has accumulated in the top of the PZR CO.253. [0.53 Chalf credit CO.253 given for "prevent steam leakage past the safety valve"3 C. 1. Instrument air header. CO.253

2. Local accumulator. [0.253 D. 1. Combined RHR pump suction header relief valve,

2. RHR to reactor vessel injection headcr relief valve,

3. RCP seal leakoff return header relief valve, and

4. Letdown orifice outlet header relief valve.

CO.25 each, no credit allowed for PZR safety and PORV3 C1.03 REFERENCE Syctcm Description B-4A, Rev. 2, Para 3.5.A., page 24 Systcm Description B-4A, Rev. 2, Para 3.6.A., page 26 Syotcm Description B-4A, Rev. 2, Para 3.7.A., page 20 Syntcm Description B-4A, Rev. 2, Para 3.8.A., pages 29 & 30 010000K401 010000K403 010000G007 ...(KA*S)

O I

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2. PLANT DESIGN INCLUDING SAFETY AND Ei1ERGENCY SYSTEMS PAGE 39 l

ANSWERS -- PRAIRIE ISLAND 1&2 -88/07/11-PARKINSON, K.

I ANSWER 2.02 (3.00)

A. 1. Provide a backpressure for the #1 seal on the RCPs to.253 to ensure flow through the #2 coal to.253,

2. Provide net positive suction head (NPSH) to.253 for the charging pumps [0.253,

3. Aid in oxygen control of the RCS CO.253 by providing a hydrogen atmosphere into which the letdown water is sprayed during normal plant operations [0.253 tequivalent wording accepted]

4. Provide a means of adding makeup water to the RCS, and

5. Aid in degasification of the RCS CO.253 when necessary prior to a plant cooldown CO.253.

tany 4 0 0.5 each] C2.03 B. 1. Protect the tube side CO.13 of the regenerative heat exchanger CO.13 from overpressure CO.13 if the charging path is isolated to.13 and letdown flow continues CO.13 [ equivalent wording accepted 3 CO.5J l

2. Spring-loaded check valve [0.23 opens to relieve the overpressure 4 condition [0.21 when the dp across the valve reaches I approximately 200 psid CO.13. Cmquivalent wording accepted 3 to.53 REFERENCE Systcm Description B12A, Rev. 2, Para 3.2.A, pages is & 19 Syotsm Description B12A, Rev. 2, Para 3.4.E, page 33 004000K405 004000G007 ...(KA's) j l

G

, . - - - - - -. - , ,,,...__~--.----m., , . ~ , - - , - - ._-,-,-v-- -,

l l 2 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 40

[ .

ANSWERS -- PRAIRIE ISLAND 1&2 -88/07/11-PARKINSON, K.

l l

l l f.NSWER 2.03 (2,00)

A. 1. RHR heat exchangers 2. RHR pumpe

3. SFP heat exchangers 4. Letdown heat exchangers

5. Excess !ktdown heat exchanger 6. Seal water heat exchanger

7. Reactor coolant pumps G. S/G blowdown sample analysis panel Csample coolerz3

9. Waste gas compressors 10. Boric acid evaporator package

11. Waste evaporator package 12. Safety injection pumps

13. Containment spray pump 14. Hydrogen recombiners heat exchangers CNUTE: for RCPs will accept uppor oil cooler, lower oil cooler, or thermal barrier heat exchanger for full credit but will accept only 1 for RCP credit 3 CNOTE: .for BA evaporator will accept vent condenser or evaporator condenser for full credit but will accept only 1 for l BA evaporator credit)

Cany 10 0 0.1 each3 C1.03 B. 1. Local /Remots switch is in the LOCAL position CO.253 Class of P-250 process computer CO.253 for that unit 3 Pump start cignals CO.13 reached the Large Motor Monitor System i limit CO.053 within the specified time period CO.053 without a soaking period CO.053 Cequivalent wording accepted 3 CO.253 ,

2. a. yes CO.253 1

b. yes CO.253 i REFERENCE System Description B-14, Rev. 1, Para 2.1, page 5 Systcm Description B-14, Rev. 1, Para 3.1.C, page 8 OOS000K102 OOSOOOK401 ...(KA*S) l l

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2 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 41 ANSWERS -- PRAIRIE ISLAND 1&2 -SS/07/11-PARKINSON, K.

l ANSWER 2.04 (2.50)

A. 1. Valves can not be opened [0.253 unless RCS pressure < 425 psi EO.253 CO.53

2. When the control switch is in the AUTO position CO.13, the valves close automatically CO.23 if RCS pressure increases to 600 psig CO.23. [0.53 .

B. Provent an inadvertent valve opening CO.253 which could overpressurize tis: RHR system CO. 253. [0,53 CHalf credit (0.25) will be given for "Prevents an incorrect flowpath during emergency core cooling operation"3 J C. 1. The valves can not be opened EO.13 unless the associated RHR pump discharge pressure CO.23 is < 210 psig CO.23. CO.53 l

2. The valves can not be opened CO.13 unless the associated SI pump i suction isolation valve to.23 from the RWST is shut CO.23. [0.53 REFERENCE Syntom Description B-15, Rev. 2, Para 3.4.A, pages 15 & 16 ]4 Syctcm Description B-15, Rev. 2, Para 3.4.C, page 17 I OO5000K407 ...(KA*S)

I 1

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, 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE d2 i - .

ANSWERS -- PRAIRIE ISLAND 1&2 -88/07/11-PARKINSON; K.

l 1

I ANSWER 2.05 (2.50) i 1

I A. Both lifting frames CO.253 must be in the down position CO.253 and the transfer tube valve EO.253 must be open [0.253. E1.03 Enquivalent wording accepted for full credit] .

l B. The conveyer car CO.253 must be at the end of its travel [0.253 under j the lifting frame EO.253. [0.753 Coquivalent wording accepted for full credit]

l C. To prevent the frame from being moved CO.253 while the crane is placing a fuel element into or removing a fuel assembly [0.253 from the upender CO.253. Co.753 tequivalent wording accepted for full credit 3 REFERENCE ;

Systcm Description B-17, Rev. O, Para 3.3.4., pages 21 & 22 l 034000K402 ...(KA'S) l 4

ANSWER 2.06 (2.00) )

A. The sump "B" isolation valves cannot be oper.ed CO.53 unless one of tha l SI test line to RWST isolation valves is shut CO.53 E1.03 i B. 1. Losa of cooling water back to RWST,

2. Airborne radioactive release through the RWST vent, and

3. Contamination of the RWST by water from zump "B".

Cany 2 9 0.5 each3 ti.03 REFERENCE Syctcm Description B-18A, Rev. 1, Para 3.3.F, page 19 006000K409 ...(KA*S)

J

_ _ _ _ - . _ _ . - . - , , _ , . , _ _ __ , _ . - , ., _ _ , ,y- _ _ -. .- - , --

o ..

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 43 ANSWERS -- PRAIRIE ISLAND ik2 -88/07/11-PARKINSON, K.

l 1

l ANSWER 2.07 (2.00) !

RHR CO.153 and SI pumps CO.153 - start CO.13 CO.43 SI pump BA supply isolation valves CMV-32001, MV-32002 and MV-320833 CO.33 - open LO.13 CO.43 High head SI to RCS cold leg isolation valves CMV-32068 and MV-320703 CO.33 - open CO.13 CO.43 RHR to reactor vessel injection valves CMV-32064 and MV-320653 CO.33 - .

open CO.13 CO.43 )

Accumulator to cold leg isolation valves CMV-32071 and MV-320723 CO.33

- open CO.13 CO.43 Ceither valve descriptions or numbers accepted for full credit 2 REFERENCE I Syntcm Description B-18B, Rev. 1, Pare 3.5.A, pages 12 and 13 006030K404 ...(KA*S) i l

ANSWER 2.08 (2.00) 2/3 CO.13 PZR pressu;*e instruments CO.23 < 1815 psig CO.23 Co.53 j 2/3 Co.13 containment pressure instruments CO.23 > 4 psig CO.23 CO.53 2/3 CO.13 steamline prassure detectors CO.13 on 1/2 CO.13 main steam headers Co.13 < 500 psig Co.13 CO.53 1/2 Co.253 manual safety injection switches CO.253 CO.53 REFERENCE I Syntcm Description B-188, Rev. 1, Para 3,5.A, page 12 006030A201 ...(KA*S) 1 I

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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 44 ANSWERS - PRAIRIE ISLAND 1&2 -88/07/11-PARKINSON, K.

ANSWER 2.09 (2.00)

, A. 1/2 Co.13 c3ntainment High-High pressure signals to.13 at 23 psig CO.13 on 3/3 CO.13 sets of containment pressure detectors CO.13 CO.53 2/2 CO.153 manual [0.153 "containment spray actuation " switches CO.13 placed in ACTUATE CO.13 CO.53 B. 1. Caustic addition valves CCV-31941 and CV-319383 open,

2. CS pump discharge valves CMV-32103 an6 MV-32105] open,

3. CS pumps start,

4. CS pump suction valves from the RWST EMV-32093 ane MV-320993 open,

5. CS pump discharge valves are blocked open, and

6. CS pump unit coolers are started.

Cany 5 0 0.2 erch3 C1.03 REFERENCE Syntcm Description B-1BD, Rev. 1, Para 3.1.C., pages 7 & 8 025000A301 ...(KA'S) l 4

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

2 PLANT DESIGN INCLUDING __ SAFETYJND EMERGENCY SYSTEMS PA!E 45 AN!WERS -- PRAIRIE ISLAND 1&2 -99/07/11-PARKINSON, K. i l

l ANSWER 2.10 (1.50) i A. 1M tr isformer to.13 !

1R t.ensformer CO.13 But z6 C15-26 tie breaker 3 CO.13 l D1 emergency diesel generator CO.13 B. 1R transformer CO.13 CT11 Bus C0.13 i Bus 25 C16-25 tie breaker 3 CO.13 l D2 emergency diesel generator to.13 C. 2RY transformer CO.13 Bus 16 C16-25 tie breaker 3 E0.13 D2 emergency diesel generator to.13 D. 2RY transformer to.13 CT12 Bus CO.13 Bus 15 C15-26 tie breaker 3 to.13 D1 emergency diesel generator [0.13 l i

REFERENCE l Syottm Description B-20.5, Rev. O, Para 2.0, pages 3 & 4 ;

062000K407 ...(KA'S)

ANSWER 2.11 (1.00)

D. [1.03 REFERENCE F,'otcm Description B-20.3, Have 0, Para 3.4, page 5 063000K201 ...(KA*S) e 9

m -

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

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 46 ANSWERS -- PRAIRIE ISLAND 1&2 -98/07/11-PARKINSON, K.

ANSWER 2.12 (1.00)

A. 1. CVCS holdup tanks,

2. BA evaporator condensers,

3. RCDTs,

4. PRTs,

5. Non-aerated drains monitor tanks,

, 6. Non-aerated dr_ ,a sump tanks, and -

7. VCTs.

tany 6 0 0.1.each3 CO.63 B. 1. Excess letdown heat exchanger letdown flow,

2. Reactor coolant pump seal #2/#3 laakof f CO.13

3. Reactor vessel flange leakof f ,

4. Safety injection accumulators drains,

5. Reactor coolant loops drains,

6. PRT, and

7. Refueling canal drains.

tany 4 0 0.1 each3 CO.43 REFERENCE System Description B-21A, Rev. O, Para 2.1, page 4 System Description B-218, Rev. O, Para 3.4.B, page 28 071000 GOO 4 06SOOOGOO4 ...(KA*S)

2. PLANT DESIGN _ INCLUDING SAFETY AND EMERGENCY SYSTEME PAGE 47 ANSWERS -- PRAIRIE ISLAND 1&2 -S7/07/11-PARKINSON, K.

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ANSWER 2.13 (1.00)

A. #11 condensate storage tank [0.253 B. #21 and #22 condensate storage tank Co.253 Chalf credit if only #21 or #22 provided3 C. Loop A cooling water header CO.253 D. Loop B cooling water header [0.253 REFERENCE CyctCm Description B-20B, Rev. O, Para 2.1, page 4 061000K401 ...(KA*S) l 1

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INSTRUMENTS AND CONTROLS PAGE 48 ANIWERS -- PRAIRIE ICLAND 1&2 -88/07/11-PARKINSON, K.

ANSWER 3.01 (1.80)

A. To reduce any inaccuracy CO.33 which could be introduced by the containment post-accident environment [0.33. CO.63 Chalf credit to.33 will be given f or "easier access f or calibration and maintenance during normal and post-accident conditions"3 B. 1. High volume sensor CO.33 check valve feature EO.33 CO.63

2. Hydraulic isolators CO.33 provide automatic containment isolation in the event of a break in a line downstream of the isolators CO.33 CO.63 REFERENCE Sycttm Description B-49, Rev. O, Para 3.1.A, pages 4 & 5 016000K101 ...(KA'S)

ANSWER 3.02 (1.75)

A. Manual rod motion stops CO.33 Auto rod motion stops CO.33 B. 1. The pulser fails, CO.33 1

2. The slave cycler fails, and CO.33 1

3. A printed circuit card is loose or removed. [0.33 1

C. Alarm reset pushbutton CO.253

)

REFERENCE Syotcm Description B-5, Rev. O, Para 3.3.D, page 16 OO1050K401 ...(KA'S) 1 1

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3. INSTRUMENTS AND CONTROLS PAGE 49 AN WERS -- PRAIRIE ISLAND 1&2 -88/07/11-PARKINSON, K. ,

ANSWER 3.03 (1.50)

A. 1. Turbine Impulse pressure < 15% power CO.33 CIntermediate Range High Flux 20% current equivalent e'a '/2 y f e nnels-accepted ior-full-credit 3

2. Beceuse of instisilities in the rod control system during low power conditions CO.33 CLimit the conssquences of a startup accident accepted for full credit 3 CThe red control systemn is unstable at <15% power due to the non-lirearity of turbine first stage pressure at that low power accepts:d f or f ull credit 3 B. Misalignment'uf the system step counters CO.03 QUESTION DELETED C. 1. Blocks manual rod withdrawal CO.33

2. Actuates a turbine runback Co.03 ANSWER 2. ACTUATE 5 A TURBINE RUNBACK DELETED ,

D. Overpower delta-T trip setpoint -5% CO.33 2/4 coincidence LO.33 REFERENCE l Syctcm Description B-5, Rev. O, Para 4.2, pages 23 & 24 l 001010K410 ...(KA*S)

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3. INSTRUMENTS AND CONTROLS PAGE 50 1

ANSWERS -- PRAIRIE ZSLAND 1&2 -88/07/11-PARKINBON, K. J l

ANSWER 3.04 (1.20)

Tho individual red position indication system CO.23 measures actual rot) poaition by using coils mounted on the outside of the rod drive pressure j housing CO.43 CO.63 Ccquivalent wording accepted for full credit 3 Tho bank demand position indication system CO.23 counts the pulses 90nerated by the rod control sywtem to move each rod grottp CO.43 CO.63 Ccquivalent wording accepted for full credit 3 REFERENCE Syctcm Description B-6, Rev. 1, Para 3.1, Page 4, and Para 3.2, Page 5 014000K502 ...(KA'S)

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3. INSTRUMENTS AND CONTROLS PAGE 51 ANSWERS -- PRAIRIE ISLAND 1&2 -08/07/11-PARKINSON, K.

AN2WER 3.05 (2.00)

1. "Reactor Coolant System High Tavg" alarm CO.23 567F CO.23 on 1/4

[ 0.1 *J

2. "Reactor Coolant Loops Lo-Lo Tavg" alarm to.23 540F CO.23 on 1/4 C0.13

3. "Reactor Coolant Loops Tave Deviation" alarm Co.23 1/4 CO.13 + or -

2F from auctioneered high Tavg CO.23 4 "Auctioneered Tavg-Tref Deviation" alarm CO.23 auctioneered high Tave CO.153 + or - 5F from Tref Co.153

5. "Lo-Lo Tavg Steam Dump Interlock" light CO.23 < 540F Co.23 on 2/4 CO.13

6. Bistable light RC LP LO Tavg at 554F

7. Bistable light RC LP LO Tavg at 540F Cany 4 0 0.5 each3 C2.03 REFERENCE Gystcm Description B-7, Rev. O, Para 3.1.B., pages 11 & 12 016000G008 ...(KA'S)

3. ' INSTRUMENTS AND CONTRDLS PAGE 52 ANS'WERS -- PRAIRIE ICLAND 1&2 -88/07/11-PARKINSON, K.

ANSWER 3.06 (2.50)

1. The steam dump "Off/ Reset-On-Bypass Intlk" switches CO.253 in the UN position CO.253

2. The steam dump "Mode Selector Control" switch CO.253 in the Tavg ,

CONTROL position Co.25]

3. Reactor coolant loop temperatures CO.253 > 540F CO.253 4 No turbine trip C0.253 2/2 stop valves CO.253 C2/3 auto stop oil pressure to.253 <50 psi CO.2533

5. Air pressure available to the valves 6.

Circulating water pumps operating Ebreaker closed] to.253 1/2 CO.253

7. Condenser vacuum > 15" [15-16.5"3 Hg C0.253 in both shells to.253 C. Turbine imoulse pressure rate of change insuf ficient to activate loss of load interlock, i.e., steam dump did not arm Esotpoint is variable based on 120 sec time constant] ,

Cany 5 e 0.5 each3 C2.53 i REFERENCE

, System Description B-7, Rev. O, Para 3.2.B. page 20 041000C015 ...(KA*S) i .

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3. INSTRUMENTS AND CONTROLS PAGE 53

)

AN!WERS -- PRAIRIE ISLAND 1&2 -BB/07/11-PARKINSON, K.

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ANSWER 3.07 (2.40)

1. "Pressurizar High/ Low Pressure" C0.23 2185 psig C0.23 1/3 C0.23

2. "Low Pressurizer Pressure Block SI" to.23 2000 psig CO.23 2/3 [0.23

3. "Pressurizer Low Pressure Reactor Trip" C0.23 1900 psig CO.23 2/4 with turbine or reactor power > 10% (P7) [0.23

4. "Pressurizer Low Pressure SI Channel Alert" to.23 1810 psig to.23 1 1/3 to.23 )

5. "Pressurizer Low Pressure Safety-Inj" to.23 1815 psig C0.23 2/3 to.23 l

6. "Pressurizer Control High Press Or Backup Htra On" C0.33 2310 psig l C0.33 i l

7. Pressure Alert High/Lew Pressure Co.23 1900 psig C0.23 1/4 CO.23 l l

Cany 4 9 0.6 each3 C2.43 REFERENCE l Syatcm Description B-7, Rev. O, Para 3.3, pages 25 - 34 j 010000K103 ...(KA'S) 1 i

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3. INSTRUMENTS AND__ CONTROLS PAGE 54 AN WERS -- PRAIRIE ISLAND 1&2 -88/07/11-PARKINSON, K.

ANSWER 3.08 (2.90)

A. Reactor trip breaker CO.23 and it's bypass breaker CO.23 are both open CO.23 CO.63 B. 1. Trips the main turbine CO.53

2. Closes the main feed regulating valves CO.23 and ieals in the signal if 2/4 Tavg channels CO.23 are < 554F Co.".J CO.63

3. Seals in the main f eed regulating valve closure signal CO.23 if the valves were closed by a SI CO.23 or high steam generator water level CO.23 CO.63

4. Permits an SI to be reset CO.33 after 90 second delay CO.33 CO.63

5. P-4 ir. puts to SAS for activation of subtriticality CSF is an alternate acceptab1w answer REFERENCE Syctem Description B-8, Rev. 1, Para 3.2, page 14 Syctcm Description B-1GC, Rev. O, Para 4.2.B, page 29 012000K610 ...(KA*S)

3. INSTRUMENTS AND CONTROLE PAGE 55 ANSWERS - PRAIRIE ISLAND 1&2 -98/07/11-PARKINSON, K.

ANSWER 3.09 (2.00)

/

A. LOSS OF DETECTOR BF3 VOLTAGE LOSS OF -

DETECTOR VOLTAGE -

B. PREAMP PREAMP power supply l

C. PULSE AMP PULSE AMP / DISCRIMINATOR D'.SCRIMINATOR I/A AUDXO COUNT RATE D. AUDIO COUNT RATE pulse shaper I

pulse driver E. LOG PULSE LOG PULSE INTEGRATOR INTEGRATOR

. I F. LEVEL AMP LEVEL AMP SUR CIRCUIT G. SUR CIRCUIT I/A computer MCB METER &

H. MCB METER & RECORDER RECORDER HI FLUX AT SHUTDOWN I. HI FLUX AT SHUTDOWN containment evacuation J. HI LEVEL TRIP HI LEVEL TRIP

/

[0.2 each] CO.Q)

REFERENCE Syctem Description B-9, Rev. O. Figure B9-6 '

015000 GOO 7 ...(MA'S) i

)

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4

3. INSTRUMENTS AND CONTROLS PAGE 56 ANSWERS -- PNAIRIE ISLAND 1&2 -88/07/11-PARKINSON, K.

ANSWER 3.10 (1.75)

A. 1. Provide information on neutron flux distribution Efuel assembly power 3 CO.53

2. Provide information on fuel assembly outlet temperature Edelta T3 Csubcooling3 CCSF status tree on subcooling] [0.53 Cfull credit will be given for confirm reactor core design parameters (0.5) and calculate hot channel factors (0.5)3 B. 1. Calibrate and verify the response of the excere nuclear instrumentation system

2. Dutect misplaced fuel assemblies

3. Detect or verify rods out of position

4. Determine post accident core conditions Cany 3 0 0.25 each3 CO.753 REFERENCE Syctcm Description B-10, Rev. 1, Para 1.1 & 1.2, page 3 015020G004 017020G004 ...(KA*S) 4 1

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. i 1 INSTRUMENTS AND CONTROLS PAGE 57 !

ANSWERS -- PRAIRIE ISLAND 162 -99/07/11-PARKINSON, K. l

. i ANSWER 3.11 (1.60)

1. SFP normal supply fan stops ;

i

2. SFP normal exhaust fan stops ;

3. Unit 1 and 2 containment in-service purge supply and exhaust valves shut 4 Starts the containment in-service purge and SFP exhaust fan !

l l 5. Opens the SFP special exhaust damper Cany 4 8 0.4 each] [1.63 I

REFERENCE !

Syctcm Description B-11, Rev. O, Para 3.2.A., page 15 l 073000K401 ...(KA*S) 1 l

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3. INSTRUMENTS AND CONTROLS PAGE 58 '

AN WERS -- PRAIRIE ISLAND 162 -88/07/11-PARKIN80N, K.

ANSWER 3.12 (3.00)

1. Containment isolation (CI) signal CO.253 which prevents the release of ,

fission products to the environment [0.253

2. Containment Ventilation isolation signal [0.253 which prevent the release of fission products to the environment [0.253

3. Control room ventilation isolation signal [0.253 which ensures the habitability of the control room [0.253

4. Steamline isolation signal CO.253 which prevents a continuous uncontrolled blowdown of more than one steam Generator [0.253 l

5. Feedwater isolation signal [0.253 which ininimizes the f eedwater effects on a cooldown EO.253

6. Containment spray (P) signal CO.253 which reducus containment pressure, Etemperature and airborne fission product leve13 Co.253 l l

7. Auxiliary feedwater pump start signal CO.253 which ensures a feedwater supply is available [0.253 ,

i Cany 6 4 0.5 each] [3.03 CNOTE: the following additional answers are accepted for full credit provided reasonable explanation of what the signal prevents, minimizes reduces or ensures was also provided 2

Diesel generatorn start and load rejecton restoration sequencing start

Control room chillers start

Shield building ventilation system starts

Auxiliary building special ventilation starts

Condensate and feedwater pumps trip

Containment fan coils shift to slow speed and dampers shift to the dome position

Component cooling water pumps start

Diesel cooling water pumps start

121 cooling water pump discharge aligna to the affected unit 3 REFERENCE Systcm Description B-18C, Rev. O, Para 2.3, page 6 013000G004 ...(KA'S) i l

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4. ' PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 59 RADIDLOGICAL CONTROL

. l ANSWERS -- PRAIRIE ICLAND 1&2 -89/07/11-PARKINSON, K.

i l

ANSWER 4.01 (2.80) ,

1. Verify Reactor Trip

2. Verify Turbine Trip ;

1

3. Verify Safeguard Buses Energized 4 Check If SI Is Actuated

5. Verify Safeguard Component Alignment

6. Check If Main steamlines Should Be Isolated l i

7. Check Containment Pressure Call seven C73 S 0.4 each3 C2.83 REFERENCE Em:rgency Operating Procedure 1E-0, Reactor Trip Or Safety Injection, Rev.

4, pcges 3 through 6 000011G010 ...(KA*S) n I

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4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 60

, , RADIOLOGICAL CONTROL ANSWERS -- PRAIRIE ISLAND 1&2 -88/07/11-PARKINSON K. l 1

ANSWER 4.02 (1.00) l prccced to the next step or substep in the ACTION / EXPECTED RESPONSE column

[preceed to step 123 Eperform step 123 CCheck Ruptured SG(s) Pressure -

Groctor than 250 psig3 C1.03 REFERENCE SWI-0-10, Rev 10, Para 3.1.3 (3) (c) , page 3 194001A102 ...(KA*S)

ANSWER 4.03 (1.00)

A. 1. Notification of Unusual Event CUnusual Event 3 CUE 3

2. Alert

3. Site Area Emergency CSite Emergency 3 CSAE3

4. General Emergency CGE3 Call 4 8 0.25 each3 C1.03 REFERENCE l Em:rgency Plan Implementing Procedure 73-2, Classifications Of Emergencies, R v. 6, pages 1, 2&3 194001A116 ...(KA'S) l ANSWER 4.04 (1.00) l Dicplay of the IVL is available thru the Plant Information Computer "O" ,

Run. Chard copy if the IVL in the control room 3 [1.03 j CHalf credit CO.53 will be given for "plant CSperry3 computer"3 I

REFCRENCE Adminirtrative Work Instruction 5AWI 3.10.2, Independent Verification Of Comp 6nents List, Rev. 3, page 9 194001K101 ...(KA*S)

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4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY ANQ PAGE 61

- RADIOLOGICAL CONTROL ANSWERS -- PRAIRIE ISLAND 1&2 -GB/07/11-PARKINSON, K.

ANEWER 4.05 (2.50)

A. Lead Plant Equipment and Reactor Operator CLPER03 to.53 B. To safeguard human life [0.53 C. To safeguard equipment or service or for other operational reasons

[0.53

~Cfull credit will be given for equivalent wording 3 D. No [0.53 ,

use of safety tags AWI specifies that Hold Cards shall not be attached on switchboard or benchboard remote control devices for clearance to work [0.53

[ full credit.will be given for equivalent wording 3 REFERENCE Administrative Work Instruction 5AWI 3.10.3, Rev. 1, Para 5.2, page 5, Para 4.1.1, page 2, para 6.2.2.e., page B ,

194001K102 ...(KA'S) ,

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4. ' PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 62 ;

RADIDLOGICAL CONTROL ANS'WERS -- PRAIRIE ISLANC 1&2 -89/07/11-PARKINGCN, K. -

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l AN WER 4.06 (1.20) ,

i A. Falso to.43 '

. B. True CO.43 C. Falso to.43 I REFERENCE SWI-0-10, Rev 10, Para 3.3.2 and 3.3.3, pages 4 and 5 l 194001K101 ...(KA'S) l l

ANSWER 4.07 (2.70) ,

A. 1. Rod bottom lights CO.153 - lit (0.153 CO.33 1

2. Reactor trip breaker and bypass breaker to.153 - open CO.153CO.33

3. Rod position indicators to.153 - zero CO.153 to.33

4. Neutron flux to.153 - decreasing CO.153 CO.33 B. 1. Manually trip turbine to.33 l

2. Verify all turbine control valves closed CO.33 '

3. Manually close control valves to.33

4. Close MSIVs and bypass valves CO.33 ;

C. Check AFW pumps [MD or TD3 to.153 running CO.153 to.33

' l REFERENCE 4 FRP 1FR-S.1, Rev. 4, page 3 000029G010 ...(KA*S)

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4. PROCEDURES - NORMAL. ABNORMALe EMERGENCY 8MR PAGE 63 -

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- i ANSWERS -- PRAIRIE ISLAND 1&2 -88/07/11-PARKINSON, K. l l

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ANSWER 4.08 (0.50)

A. Core Cooling CO.53 B. 2. Cunit 1 subcriticality - RED 3 PART B. OF THE QUESTION WAS DELETED Eboth units display green when in this cendition]

REFERENCE F-0.2, Core Cooling, Rev.2, Page i 1

F-0.1, Subcriticality, Rev. 1, page 1 000074A207 194001A113 ...(KA'S) i i

AN WER 4.09 (1.30) 4 2 ain CO.253 without CCW CO.253 CO.53 .

l 200F Co.25) at radial bearing CO.53 ,

200F CO.253 at motor bearing CO.53 CCOo1 leakage CO.253 greater than 6.5 gpm Co.2533 tocol leakage CO.253 (0.5 gpm Co.2533 CRCP vibration Co.253 >10 mils CO.2533

REFERENCE Optrcting Procedure C14, Component Cooling System, Rev. 6, Para 6.8.1, page 13 000015K209 j 000015K200 ...(KA*S) l I

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4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 64 RADIOLOGICAL CONTROL i ANSWERS -- PRAIRIE ISLAND 1h2 -88/07/11-PARKINSON, K.

ANSWER 4.10 (2.10)

SI nat ready lights CO.153 - not lit Co.153 91 cctive lights CO.153 - lit with exceptions ,

Crntcinment isolation lights CO.153 - lit with exceptions Stcp spent fuel pool CO.153 normal ventit ation Co.153 Step rad waste building CO.153 ventilation system Co.153 Cctcgory I doors CO.153 - closed CO.153 Ch;ck operation log CO.13 for any ventilation openings CO.13 that must be eleccd within 6 minutes Co.13 Call seven in any order G 0.3 each3 C2.13 REFERENCE Emergency Operating Procedure 1E-0, Reactor Trip Or Safety Injection, Rev.

4, Step 5, page 5 000007K202 ...(KA's) 4

, ANSWER 4.11 (1.50) 1 A. Unit II Co.53 B. Unit ! CO.53 ;

i C. Unit II CO.53 REFERENCE

, SWI-0-2, Rev. 16, Para 3.3.J, page 6 194001A110 194001A112 ...(KA*S) 1 t

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4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 65 RADIOL,0GICAL. CONTROL ANSWERS -- PRAIRIE ISLAND 1&2 -88/07/11-PARKINSON, K.

ANSWER 4.12 (1.70)

A. Radiation Work Permits CRWPs3 C0.53 B. (1) Controlled CProtected3 CRestrictedJ CPosted3 to.23 (2) Protected EContro11ed3 ERestricted3 CPosted3 so.23 i

C. (1) radiation CO.23 (2) distance CO.23 (3) shielding CO.23 D. ALARA to.23 i REFERENCE PlcNt Operations Manual, Section F2, Rev. 5, Para 2.2.1g Page 8 194001K104 194001K103 ...(KA'S) i 1

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4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 66 RADIOLOGICAL CONTROL ANSWERS -- PRAIRIE ISLAND 1&2 -88/07/11-PARKINSON, K.

. l ANSWER 4.13 (2.00) i A. Hot shutdown panels CO.53 l B. Aux Bldg Elev 695 [MSIVs3 E0.53 j C. D1 Diesel Room CO.53 1

. D. Turbine Front Standard to.53 CNOTE: for either C. or D. but not both will also accept trip reactor C0.13 cloco MSIVs (0.13 close pressurizer block valves to.13 assure turbine trip 2 C0.133

, REFERENCE Plcnt Operations Manual, Section F5 Appendix B, Control Room Evacuation (Firo) Safe shutdown Procedure, Rev. 4, Para 3.B. and 3.C., page 4 000018G006 ...(KA'S) j AN2WER 4.14 (1.00)

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C. [GammaJ ti.03 REFERENCE Weatinghouse, Radiation, Chemistry and Corrosion Considerations f or Nuclear Pow;r Plant Application, 1983, page 5-32 191002K119 ...(KA'S) '

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4. PROCEDURES - NORMAL..ABNDRMAL._ EMERGENCY AND PACE 67

. RADIDLOGICAL CONTROL

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ANSWERS -- PRAIRIE ISLAND 1&2 -98/07/11-PARKINSON, K.

I ANSWER 4.15 (2.00)

O. To maintain the integrity of the fuel cladding CO.53 B. 1. Thermal power j 2. Reactor Coolant System pressure

3. Reactor Coolant System temperature

4. Flow 0

5. Delta T t

Cany 3 0 0.25 each3 CO.753 :

C. To maintain the integrity of the Reactor Coolant System CO.53 [

D. RCS pressure < 2735 psig C0.253 i REFERENCE Tcchnical Specification TS-2.1 and TS-2.2 010000G005 ...(KA'S) l e

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'PRk!RIEISLAND1h2 98/07/11 PARKINSON, K.

SECTION 1 1.01 1.00 1.02 1.00 1.03 1.00 1.04 1.00 1.00 1.00 1.06 2.00 1.07 1.00 1.00 1.00 .

1.09 1.00 1.10 1.00 '

1.11 1.00 1.12 1.00 1.13 1.00 1.14 1.00 1.15 1.00 1.16 1.00 1.17 1.00 1.18 1.00

, 1.19 1.00 1.20 1.00 1.21 1.00 1.22 1.00' 1.23 1.00 1.24 1.00 25.00 SECTION 2 2.01 2.50 2.02 3.00 2.03 2.00 2.04 2.50 2.05 2.50 l 2.06 2.00 l 2.07 2.00 l 2.08 2.00 1 2.09 2.00 !

2.10 1.50 2.11 1.00 ,

2.12 1.00 l 2.13 1.00 -

25.00 i

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, PRAIRIE ISLAND 1&2 88/07/11 PARKINSON, K.

SECTION 3 3.01 1.80 3.02 1.75 3.03 1.50 3.04 1.20 3.05 2.00 3.06 2.50 3.07 2.40 3.08 2.90 3.09 2.00 4

3.10 1.75 3.11 1.60 3.12 3.00 24.40 SECTION 4 4.01 2.90 4.02 1.00 4.03 1.00 4.04 1.00 4.05 2.50, 4.06 1.20 4.07 2.70 4.08 0.50 4.09 1.50 4.10 2.10 J 4.11 1.50 4.12 1.70 1

4.13 2.00 4.14 1.00 4.15 2.00

_______ 3 24.50 i i

1 Tctol 98.90 ,

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ML20151U319

Text

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