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Exam Rept 50-302/OL-87-02 on 870126.Exam Results:One Reactor Operator Passed.Exam Master Copy & Util Comments Encl
	ML20211M947
Person / Time
Site:	Crystal River
Issue date:	02/13/1987
From:	Lawyer S, Munro J; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML20211M941	List: ML20211M934; ML20211M947; RA86-023, Forwards Comments on 861217 Senior Reactor Operator Exams. Related Documentation Encl;
References
	50-302-OL-87-02, 50-302-OL-87-2, NUDOCS 8702270398
	Download: ML20211M947 (129)
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! - ENCLOSURE 1

, EXAMINATION REPORT 302/0L-87-02 3 <

s , Facility Licensee: Florida Power Corporation s s 11,

, Facility \Name: Crystal River Unit 3 Facility Docket No.: 50-302 I Written examination w admi iste in he ~bffices in Atlanta, G orgi .

Chief Examiner: + /> 37 ~

Sandf Lawyerf ( 'Date 51gned Approved'by: arv A//S!F7 ohg . pnW/Sec)rfon Chief Date Signed Summary:

Examinations on January 26, 1987' One candidate was administered a written R0 re-examination. That one candidate passed.-

Based on the results described above, one of one R0 passed.

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REPORT DETAILS

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1. Facility Employees Contacted: 1 James Owen, Licensed Operator Instructor l'

2. Examiners:

S. Lawyer *

Chief Examiner

3. Examination Review Meeting At the! conclusion of the written examination, the examiner provided Mr. Owens~ with a, copy of the written examination and answer key for review.

The comments m ade' by the facility reviewers are included as Enclosure 3 to this report, and the NRC Resolutions to these comments are listed below.

R0 Exam (1)iQuestion 1.02 - Agreed. C is the correct answer. The answer key was changed accordingly.

(2) Question 1.21(c) - Agreed. The answer key was changed as recommended.

(3) Question 4.02 - Agreed. " Exit cooling water temperature /180 F" was addedtojheanswerkey.

(4) Question 4.33 - Agreed. The question was modified accordingly.

(5)- Question 4.10 - Agreed. Answer key was changed as recommended.

(6) Question 4.14 - Agreed. The question was modified for future use.

4. Exit Meeting Since the examination was conducted in the RII Offices and no oral examina-tions were given, no exit meeting was necessary.

The licensee did not identify as proprietary any of the material provided to or reviewed by the examiners.

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6 o MASTER U.S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION Facility: CRYSTAL RIVER Reactor Type: PWR-B&W177 Date Administered: 86/12/17 Examiner: HUENEFELD, J.

Candidate: ' ANSWER KEY INSTRUCTIONS TO CANDIDATE:

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

Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination

papers will be picked up six (6) hours after the examination starts.

Category % of Candidate's % of Value Total Score Cat. Value Category

'7 34 27.f4 5. Theory of Nuclear Power Plant Operation, Fluids and Thermodynamics J

31 25,/ 6. Plant System Design, Control and Instrumentation

.ff.C 3 28f 23.M 7. Procedures - Normal, Abnormal, Emergency, and Radiological Control 7

29 23.M 8. Administrative Procedures, Conditions, and Limitations g

XI TOTALS Final Grade %

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

Candidate's Signature

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

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

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

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

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

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

6. Use only the paper provided for answers.

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

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

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

10. Skip at least three lines between each answer.

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

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

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

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

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

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

17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.

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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 a part of the answer.

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

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

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

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 1 AND THERMODYNAMICS QUESTION 5.01- .(2.00)

-DESCRIBE the~ response of neutron power for each of the following two (2) situations: (2.0)

1. !The reactor is critical at 10**-8 amps and control rods are inserted by 1% RI.

2. The reactor is at.100% FP with the ICS in " Track," and control rods are inserted by 1% Rod Index. (Note: a detailed description of-ICS ops is NOT required.)

QUESTION 5.02 (2.00)

MAKE a sketch of a graph of neutron countrate versus time (linear scale) assuming a short duration rod pull is made with the reactor critical below the point of adding heat. NEGLECT feedback effects.

EMPHASIZE on your graph the effects of prompt and delayed neutrons. (2.0)

QUESTION 5.03 (0.5)

Inserted rod worth with increasing temperature during a plant heatup. (0.5)

. increases

. decreases

. remains constant t

-QUESTION 5.04 (0.5)

The negative moderator coefficient of reactivity in j magnitude as rods are withdrawn. (0.5)

a. increases

b. decreases

c. _ remains constant

! (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 2 AND THERMODYNAMICS QUESTION 5.05 (0.5)

The negative moderator coefficient of reactivity in magnitude as boron is increased. (0.5)

a. increases

b. decreases

c. remains constant QUESTION 5.06 (1.50)

AFW can have a proportionately larger cooling effect on the RCS for the same flow rate than MFW. STATE three (3) reasons for this.

(Assume forced convection in the RCS.) (1.5)

QUESTION 5.07 (1.00)

WHICH one (1) of the following would NOT be an acceptable means for elimination of a void that has formed in an idle loop during a' single loop natural circulation cooldown? (1.0)

. venting through the hot leg vents

. bumping an RCP

. establishing full HPI flow untti the void collapses

. allowing the void to condense through ambient heat losses QUESTION 5.08 (2.00)

The reactor is shut down by 6% delta k/k with a source neutron countrate indication of 50 cps. Rods are withdrawn to raise the source range indication to 285 cps. WHAT is the value of

- reactivity when counts are 285 cps? (SHOW all work) (2.0) i QUESTION 5.09 (1.00)

STATE the two (2) most positive indications in a subcooled RCS that natural circulation has been lost. (1.0)

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. v Si -THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 3 AND THERMODYNAMICS-

' QUESTION 5.10' (1.50)

With the RCS saturated,- WHAT is the best indication of a loss of natural circulation flow or interruption of boiler condenser cooling -(reflux boiling)? (1.5)

QUESTION 5.11 (1.00)

During an overcooling event, the RCS will remain subcooled unless:

(SELECT _one.) (1.0)

a. RCPs are lost.

b. emergency feedwater actuates.

c. the overcooling is not rapidly terminated.

d. the pressurizer empties.

QUESTION 5.12 (1.50)

During a plant heatup, OP-202, a vacuum is drawn in the Steam Generators. This action promotes even warming of the Steam Generator shell. EXPLAIN why this occurs. (1.5)

QUESTION 5.13 (1.00)

The water occurs duringhammer problem WHAT range duringtemperatures?

of feedwater heatup at Crystal River typically)

(SELECTone. (1.0)

a. below 180 deg F

b. 210 to 230 deg F

c. 230 to 300 deg F

d. above 300 deg F

(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)

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5. THEORY OF NUCLEAR POWER PLANT OPERATION.' FLUIDS. -PAGE 4 AND THERMODYNAMICS QUESTION 5.14 (1.00)

- There are three (3)- regions 'of heat transfer in the steam generator.

a. NAME the region that:provides the greatest heat flux. (0.5)

-b. NAME the region that expands the largest amount due.to a power level increase. (0.5)

. QUESTION 5.15 (2.00)

a. . Approximately HOW MUCH of the RCS flow will bypass the core as

. .it flows through the vessel? (SELECTone.) (1.0) ,

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a. 1%

b. 6%

c. 15%

d. 19%

b. Even though the above amount of flow bypasses the core, it

, still picks up heat energy. DESCRIBE the mechanism by which this occurs.

(1.0)

QUESTION 5.16 .(3.00)

a. WHAT special complication is associated with conducting a '

natural circulation cooldown with only one -(1) OTSG and with l OTSG full of water? (1.5)

b. WHAT special complication is associated with conducting a forced cooldown on one (1) OTSG and with the other OTSG dry and-depressurized? (1.5)

QUESTION 5.17 (2.00)

-IS the boiler-condenser or reflux boiling method of RCS circulation susceptible to blockage by buildup of non-condensables? EXPLAIN. (2.0)

(***** CATEGORY 05CONTINUEDONNEXTPAGE*****)

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 5 AND THERMODYNAMICS QUESTION 5.18 (0.50)

ANSWER TRUE or' FALSE.

After a reactor trip early in life, emptying a full makeup tank of i deborated water into the reactor coolant system will result in a reactivity addition of less than 1.0% delta k/k. (0.5)

QUESTION 5.19 (1.00)

WHAT does the term " gray" mean in gray APSRs? (1.0)

QUESTION 5.20 (0.50)

ANSWER TRUE or FALSE.

A one (1) decade per minute (DPM) startup rate implies that reactor

-power will change by one-half (1/2) of one decade (a factor of 5) in thirty (30) seconds. (0.5)

QUESTION 5.21 .(2.00)

Reactor power drops to about 7.5% instantly after a reactor trip.

Given that the delayed neutron fraction is less than 1%, WHY doesn't reactor power drop instantly to less than 1%? Address the importance of delayed neutrons and wuberitical reactivity in your answer. (2.0)

QUESTION 5.22 (1.50)

. STATE three (3) RCS symptoms that, when taken separately, are indeterminate but when observed coincidently, indicate that pressurizer level is no longer a good indicator of RCS inventory. (1.5) 1

(***** CATEGORY 05CONTINUEDONNEXTPAGE*****)

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS. PAGE 6 AND THERMODYNAMICS QUESTION 5.23- (1.50)

During power operations HOW can you determine, at a glance, that the amount of positive reactivity that will be inserted by dIppler and moderator temperature during a reactor trip will NOT exceed the amount of negative reactivity that will be inserted by control rods? (1.5)

QUESTION 5.24 (3.00)

WHICH contains the greater enthalpy, a cubic foot of saturated water at 70 degrees, or a cubic foot of saturated water vapor.

at 70 degrees?- (Use the steam tables provided to prove your answer.) (3.0) k 4

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6. PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 7 INSTRUMENTATION

~ QUESTION 6.01- (3.00)

WHY is it NOT necessary for the ICS to program OTSG water level

-versus power in order to achieve equilibrium conditions? (3.0)

QUESTION 6.02 (1.50)

STATE six (6) of the seven (7) RCP start permissives as stated in OP 302. Setpoints are NOT required. (1,5)

QUESTION 6.03 (0.50)

ANSWER TRUE or FALSE.

The self-powered neutron detectors are susceptible to thermionic-effects causing them to read very high when exposed to very high temperatures. (0.5)

QUESTION 6.04 (0.50)

ANSWER TRUE or FALSE.

The incore instrumentation is compensated for gamma radiation. (0.5)

QUESTION 6.05 (0.50)

ANSWER TRUE or FALSE.

The SPDS displays the average of the five (5) highest of the 52 incore thermocouples. (0.5)

(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

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6. PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 8 INSTRUMENTATION QUESTION 6.06 (0.50)

ANSWER TRUE or FALSE.

Essentially complete gamma compensation is achieved in the Intermediate Range nuclear instrumentation by surrounding the detectors by four (4) inches of lead. (0.5)

QUESTION 6.07 (0.50)

ANSWER TRUE or FALSE.

Voiding in the downcomer region of the reactor vessel can cause a significant increase in nuclear instrument indication. (0.5)

QUESTION 6.08 (1.50)

Given a loss of forced RCS flow, the T(c) indication may decrease even though natural circulation of the RC does NOT exist. WHY? (1.5)

QUESTION 6.09 (2.00)

a. WHAT effect does elevated reactor building temperature have on the indicated level of a level indicator that has a wet reference le EXPLAIN. (One or two sentences should be

-sufficient.)g? (1.5)

b. IS this effect observable on control room instrumentation?

(YES'orN0) (0.5)

(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

6. PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 9 YNSTRUMENTATION QUESTION 6.10 (1.00)

WHICH one (1) of the following phases best describes the behavior of a geiger-mueller detector when exposed to extremely high INCREASING radiation fields? The meter will: (SELECT one.) (1.0)

a. saturate at full scale and stay pegged.

b. saturate at some maximum value and remain at that value.

c. saturate at soue maximum value and then slowly drop off.

d. saturate at some maximum value and drop abruptly to zero.

QUESTION 6.11 (2.00)

STATE four (4) of the six (6) loads serviced by the Decay Heat Closed Cycl.e Cooling system. (Multiple pumps from a single system are to be counted as a single load. Pumps and the motors that drive them are to be considered as a single load.) (2.0)

' QUESTION 6.12 (3.00)

a. STATE the three (3) types of emergency diesel auto-start signals. (1.5)

b. STATE the five (5) permissives that must be satisfied in order for an auto-start to occur on signal. (1.5)

QUESTION 6.13 (1.00)

WHERE do the DC system and the SW system surge tanks vent and relieve to? (1.0)

QUESTION 6.14 (2.50)

a. WHY is it more important to check for leakage from the SW system into the RB than into other locations? (1.0)

b. STATE three (3) different SW supplied components that are instrumented to monitor for leakage from the SW system to the RB. (1.5)

(***** CATEGORY 06CONTINUEDONNEXTPAGE*****)

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. 6. ' PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 10-INSTRUMENTATION QUESTION 6.15 (1.50)

WHAT is the safety significance of maintaining the SW system pressurized? (1.5)

QUESTION 6.16 (1.00)

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WHAT are the two (2) sources of water to the Industrial Cooler (CI) system? (1.0)

QUESTION 6.17 (1.50)

EXPLAIN WHY the Nuclear Instrumentation tends to become less

- conservativeaspowerincreasesfrom/FPto100%FP. (1.5)

IC8 QUESTION 6.18 (1.50)

At WHAT Intermediate Range (IR) level is the Source Range instrument deenergized? DOES it take one or both of the IR instruments to cause this to happen? (1.5)

QUESTION 6.19 (1.50)

WHY is it important that the NNI X or Y +/- 24 VDC system be

totally deenergized by the power supply monitor should a degraded voltage condition develop? (1.5)

QUESTION 6.20 (1.50)

HOW is the RCS flow signal to the ICS derived? (1.5)

(***** CATEGORY 06CONTINUEDONNEXTPAGE*****)

6. PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 11 INSTRUMENTATION

. QUESTION 6.21 (1.50)

GIVE a general, one (1) or two (2) sentence, description of the major difference between a fixed water spray system, a wet pipe sprinkler system, and a pre-action sprinkler system. Also, GIVE cne (1) example application of each type. (1.5)

QUESTION 6.22 (1.00)

WHAT is the minimum volume of fuel oil that a day tank must have in order for the associated DG to be considered OPERABLE?

(SELECTone.) (1.0)

a. 100 gallons

b. 200 gallons

c. 300 gallons

d. 400 gallons

(*****ENDOFCATEGORY06*****)

7. PROCEDURES - NORMAL, ABNORMAL. EMERGENCY PAGE 12 AND RADIOLOGICAL CONTROL QUESTION 7.01 (1.50)

WHY is it necessary that all of the hydrazine be removed from the RCS prior to placing a makeup demineralizer in service? (1.5)

QUESTION 7.02 (0.50)

ANSWER TRUE or FALSE. l During a plant heatup (OP-202), pressurizer level is to be controlled such that there is an out-surge from the surge line n zzle. (0.5)

QUESTION 7.03 (1.50)

According to AT0G there are three (3) specific instances during a less of forced RCS flow when EFW should be " turned on full." STATE

one (1) of these three (3). (1.5)

N QUESTION 4J4 (0.50)

ANSWER TRUE or FALS Upon loss of all feedwater, t e s of service water as a source of emergency feedwater is preferable to e ishing HPI cooling. (0.5) r l

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7.- PROCEDURES - NORMAL, ABNORMAL, ENERGENCY PAGE~13 .

AND RADIOLOGICAL CONTROL QUESTION 7.05 (1.00)'

WICH one (1) of the following statements best describes why two . I (2) HPI pumps should be_used for HPI/PORV cooling? (SELECTone.) (1.0)

(a.) One HPI pump is NOT adequate for core heat removal.

(b'.).OneHPI-pumpCANNOTkeepupwiththeleakoutofthePORV.

.(c.) The time to match core decay heat is much shorter for two pumps than for one pump.

-(d.)-Two pumps result-in balanced heat removal.

QUESTION 7.06 -(0.50)

ANSWER TRUE or FALSE.

Following inadequate core cooling AND hydrogen production, only the PZR vent may be used for RCS depressurization. (0.5)

QUESTION 7.07 (1.00)

a. WHAT is the maximum allowable flow for one (1) HPI pump? (0.5)

b. At WHAT RCS pressure will this runout condition occur? (0.5)

QUESTION 7.08 (1.00)

According to ATOG, under WHICH one (1) of the following conditions should a reactor coolant pump NOT be started or bumped? (SELECT' one.) (1.0)

. RCS subcooled with natural circulation existing.

. RCS subcooled with no natural circulation.

. RCS saturated with natural circulation existing (HPI on).

. RCS saturated with no natural circulation existing (HPI on).

(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY PAGE 14 AND RADIOLOGICAL CONTROL QUESTION 7.09 (2.00)

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a LOCA, action must be taken to preclude the possibility of boron precipitation in the reactor vessel. GIVE a one (1) sentence description of each.cf the t7e(2) methods that are to be used at Crystal River. (2.0)

QUESTION 7.10 (0.50)

WHAT parameter are you procedurally directed to observe to independently verify that generator output voltage is, in fact, 22 kV7 (0.5)

QUESTION 7.11 (1.00)

WHICH one of the following conditions does NOT constitute grounds for an immediate manual reactor trip: (SELECT one) (1.0)

(a.) Two (2) main steam isolation valves on different loops have been inadvertently shut.

(b.) Reactor power is 10% and main feedwater is lost.

(c.) Pressurizer level is 295 inches and decreasing.

(d.) All four safety groups drop into the core while at 100% power.

' QUESTION 7.12 (1.50)

Subcooling margin is lost as a result of a small break LOCA.

STATE the three (3) major actions, automatic or manual, that must be initiated. (1.5)

(***** CATEGORY 07CONTINUEDONNEXTPAGE*****)

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'7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY PAGE 15 AND RADIOLOGICAL CONTROL

. QUESTION 7.13 -(1.50)

Various procedures (APs/EPs) require that you verify proper OTSG level following actuation of the EFIC system. LIST the correct EFIC level for each of the following conditions. (1.5)

a. RCPs running... reactor tripped

b. no RCPs running... adequate subcooling margin

c. no RCPs running... inadequate subcooling margin OR less than two (2) HPI pumps QUESTION 7.14 (1.50)

LIST the Immediate Actions for AP-555 (Continuous Control Rod Withdrawal). (1.5)

-QUESTION 7.15 (1.00)

While 1A must operating at 95%

be removed from power, service.you (astwo WHAT SS00)(are informed

2) actions are you that RCP required to ensure have been taken PRIOR to securing the pump per the limits and precautions of OP-204 (Power Operation)? (1.0)

QUESTION 7.16 (2.00)

During full power operation you discover that the PORV block valve is open and will not respond to the control switch on the Main Control Board. Per Technical Specifications WHAT actions must be taken? (2.0)

QUESTION 7.17 (3.00)

WHAT are the nine (9) Immediate Actions for AP-380 (Engineered SafeguardsSystemActuation)? (3.0) 4 j (***** CATEGORY 07 CONTINUE 0ONNEXTPAGE*****)

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY PAGE 16 AND RADIOLOGICAL CONTROL QUESTION 7.18 (2.00) a.- During an approach to criticality, you note that RCS Tave has decreased to 521 deg F. WHAT actions are required per Technical Specifications? (1.0)

b. STATE two (2) of the four (4) bases for the minimum temperature for criticality. (1.0)

QUESTION 7.19 (3.00)

LIST the nine (9) "Immediate" actions (remedial actions not required) of the Reactor Protection System Actuation

~ procedure (AP-580). (3.0)

QUESTION 7.20 (2.50)

a. DEFINE shutdown margin. INCLUDE any assumptions about axial power shaping rod or control rod positioning. (1.5)

b. During a plant heatup, modes 5, 4, and 3, the reactor must be shutdown by >/= 1% delta k/k. When group 1 rods are withdrawn for the heatup, does their inserted worth count as part of that 1% delta k/k? WHY? (1.0) l

(***** END OF CATEGORY 07 *****)

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~8. ADMINISTRATIVE PROCEDURES. CONDITIONS, PAGE 17 AND LIMITATIONS QUESTION 8.01 (0.50)

~ ANSWER TRUE or FALSE.

Compliance with Normal Operating Procedures is required by the Code of Federal Regulations. (0.5)

QUESTION 8.02 (0.50)

ANSWER TRUE or FALSE.

The Code of Federal Regulations allows a licensed senior operator to take reasonable action that departs from a Technical Specification in an emergency. (0.5)

. QUESTION 8.03 (2.50)

STATE five (5) reportable events that according to the Code of Federal Regulations must be reported to the NRC either immediately t or within one (1) hour. (2.5)

QUESTION 8.04 (1.50)

WHAT are the Nuclear Instrumentation requirements in Modes 3-5 if the CRDM breakers are closed and the rods are capable of being withdrawn (i.e., how many power range, intermediate range, and source range channels are required)? (1.5)

QUESTION 8.05 (1.00)

Assume that a MVP breaker fails to shut. An initial investigation is undertaken via work request to determine the cause of failure.

No cause for the failure is evident. WHAT are the minimum testing requirements for reestablishing the operability of this breaker? (1.0)

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

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, PAGE 18 AND LIMITATIONS QUESTION 8.06 (1.50)

WHICH departments must supply personnel for the Fire 8rigade',

HOW many individuals must they supply, and WHICH department must supply the team leader? .(1.5)

QUESTION 8.07 (0.50)

ANSWER TRUE or FALSE.

The generic action statement, TS 3.0.3 (i.e., requires one (1) hour S/D if certain LCOs are not met), may be entered for planned maintenance purposes. (0.5) t)UESTION 8.08 (1.00)

The OSIM states that it is permissible to use the PORV (RCV-10) to avoid high pressure reactor trips given two (2) conditions. STATE those two (2) conditions. (1.0)

QUESTION 8.09 (1.00)

It is determined that the audible function of a portion of the main control board annunciators is out of service. STATE two (2) actions that are required by the OSIM. (1.0) i I

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, PAGE 19 AND LIMITATIONS QUESTION 8.10 (1.00)

The authority to startup and return to power operation rests with WHAT two (2) positions? (1.0) 1 QUESTION 8.11 (2.00)

AI-500 requires the Nuclear Shift Supervisor to perform five (5) functions in the event of a reactor trip or plant shutdown. LIST the five (5) functions. (2.0)

. QUESTION.8.12 (2.00)

According to the emergency plan implementing procedure, there are seven (7) Emergency Teams that may be activated by the Emergency Coordinator during an emergency. LIST four (4) of these Emergency Teams. (2.0)

QUESTION 8.13 (1.50)

WHAT are the three (3) categories of decisions that the Emergency Coordinator cannot delegate the responsibility for? (1.5)

QUESTION 8.14 (0.50)

If generating complex personnel are to evacuate the site, HOW many suitable evacuation routes are available from the site? (0.5)

QUESTION 8.15 (1.50)

In any case where a " General Emergency" has been declared, WHAT is the minimum protective action recommendation that should be made r:garding the welfare of the surrounding population? (1.5)

(***** CATEGORY 08CONTINUEDONNEXTPAGE*****)

% g -

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, PAGE 20 AND LIMITATIONS

- QUESTION 8.16 (1.00)

The on-shift Nuclear Shift Supervisor has the authority of the Emergency Coordinator until relieved by the designated Emergency Coordinator. WHO is the designated Emergency Coordinator? (1.0)

QUESTION 8.17 (1.00)

According to EM-201, Duties of an Individual Who Discovers an Emergency, WHAT is the first action that an individual who discovers an emergency condition locally out in the plant should take? (1.0)

' QUESTION 8.18 (2.00)

STATE'the five (5) conditions required for containment integrity to exist. (2.0)

QUESTION 8.19 (1.50)

CP-115 (In Plant Equipment Clearance and Switching Orders) requires PRC approval prior to issuance of a clearance which cannot meet the double valve protection guidelines.

a. WHAT are the double valve protection guidelines? (1.0)

b. WHEN may the Clearance Authority approve clearances which do not meet these guidelines without PRC approval? (0.5)

QUESTION 8.20 (1.00)

All steam systems at CR-3 have been administratively established as RCAs. GIVE two (2) other examples of systems that are normally not contaminated, yet have been administratively established as RCAs. (1.0)

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

.,w,

8. " ADMINISTRATIVE PROCEDURES, CONDITIONS, PAGE 21 AND LIMITATIONS-

' QUESTION 8.21. '(1.00)

Using one (1) or two (2) sentences, STATE the main difference b2 tween an RWP and an SRWP. (1.0)

-QUESTION 8.22 (1.00)

.WHAT minimum radiation field, in mr/hr, constitutes a "High

-Radiation Area"? (1.0)

QUESTION 8.23. (1.00)

WHAT significant personnel safety hazard, other than radiological, is associated with using tools in the spent fuel pool area, and WHAT must be done to eliminate it? (1.0)

QUESTION 8.24 (1.00)

WHAT restrictions apply to raising an irradiated fuel asssembly with the new fuel elevator?~ (1.0)

- - END OF CATEGORY 08 *****)

(********(*****ENDOFEXAMINATION*************)

l l '

5. THEORY OF-NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 22 AND THERMODYNAMICS ANSWERS -- CRYSTAL RIVER -86/12/15-HUENEFELD, J.

ANSWER 5.01 (2.00)

The reactor will go, and remain, subcritical and power will 1.

decrease to the subcritical multiplication level corresponding to the amount of negative reactivity inserted. [+1.0]

4

2. Reactor power will decrease and level off.at a new critical power level, with the amount of the decrease being determined

by the power deficit. [+1.0]

i REFERENCE

1. Crystal River: Nuclear Energy Training, Module 3, " Reactor Operation," pp. 13.5-2 and 13.5-4.

4 I

i l

s i

i

?

s

..- s a g *

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 23 F- AND THERMODYNAMICS ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J. ,l L %

..i ANSWER 5.02 (2.00) s Countrate

-[+1.0] *

\

j Prompt response to rod withdrawal Constant period dominated h by delayed neutrons

, }

[+1.0]

Time s

REFERENCE

1. Oconee: Fundamentals of Nuclear Reactor Engineering, Duke Power Co., pp. 101 through 104.

2. Crystal River: Nuclear Energy Training, Module 3, Unit 6 (general). ,

t ANSWER 5.03 (0.5)

(a.) [+0.5]

REFERENCE

1. Crystal River: Nuclear Energy Training, Module 3, Figure 16-6. ,

'% 4

5. ' THEORY OF NUCLEAR POWER' PLANT OPERATION, FLUIDS, PAGE 24

-AND THERMODYNAMICS.

.. ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.

x ANSWER 5.04 (0.5)

, '(b.) [+0.5]

J'

, REFERENCE i

[1 [l+- Crystal River: Nuclear Energy Training, Module 3, Figure 16-3.

j c ANSWER'5.05 (0,5) s p

(N)/ [+0.5] See a h ed d a M C-k k

REFERENCE

1. Crys'tal River: Nuclear Energy Training, Module 3, Figure 16-3.

l~'

ANSWER 5.06 (1.50)- i 1.- AFW enters near 'the top of the OTSG. This results in an effective increase' in heat transfer area.

4- ,

2. AFW is not pre-heated, and is therefore colder than MFW.

~

N '" 3. 'AFW has,a steam pressure reduction effect that MFW does not '

have because it is injected into the steam space of the OTSG.

[n'. ,;p m, ,

I'

[+0.5] each g; '

l ,

s REFERENCE ,,

o

1. B&W Technical Document, Emergency Procedures Technical Bases.

4 $

i

)

1 M

8sntcc Oca C<nwol fa< Laqe. PW L

.a 6Ievtviv C, wpm # cra La Ad(% u t; p~6 d A S * /6 e l so t .. ..

= ..

l

. . _ _. = . . _ _ . ._. . '

. . ~.:_

.-~..

= . . _ ... . .- :_- ~. . ..

9 04.

- '%L~ - 00* t g%

-t=:g;; ... ..=-. .% ,.

g

% m...- .

s N ,. _

y

.. j N_. mN. _. . - . .

. _.N. N.

e _ . _ .

8 I...

=..

l

____._ _ _ _ . A., . , .

. _ . . .__. . . _ . . . _ . . . . . _ . w h.

s.

\

n,,o.u

*30

\.,

g __

x. . ..

1

-30 - - - -

.- c _ - - _ __ _. _.

Q< _4e soo o soo soo soo .oo soo W 94R4704 ftGBPERATURE (*F)

FIGURE SNP-RF-6t; 900ERATOR TEMPERATURE COEFFICIENT CURVE

( AT BOL AND EOL, CYCLE 1) ROOS IN (REV.1) .

increasing the moderator temperature increases the probability of neutron leakage into the control rod and loss to the fission chain j

reaction. ".arefore for a given moderator temperature change, more l .

negative reactivity is inserted when control rods are in the core.

Consider the equation for aT 1 .

2 2 dL f dL 1 df I dl 2 dT th)

+ - B +

T f dT p dT (dT 3 23 W-l l

0469C l

l i

.- , ,, - ,----,-+

5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. PAGE 25 AND THERMODYNAMICS ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD,J. ,

1 ANSWER 5.07 (1.00)'

(c,) [+1.0]

REFERENCE

1. B&W Technical Document, Emergency Procedures Technical Bases.

l

. ANSWER 5.08 (2.00)

Given p = -6% CR(1) = 50 cps CR(2) = 285 cps 1 1

=.K = ---- = 0.943 OK if estimated p = k - 1 [+0.5]

1-p 1 1.06 CR 1-K 1-K 1 2 50 2 i --- = ------ = --- = ---------

[+0.5]

CR- 1-K 285 1 - 0.943 2 1 50(0.057)

= 1 - K therefore K = 0.99 [+0.5]

. 285 2 2 i

K-1 0.99 - 1 p = ----- = -------- =~-0.01 [+0. 5]

K 0.99 or 1% shutdown REFERENCE

1. Basic Reactor Theory, Formula Sheet.

. . N -. e

-e e.. -- -- , ---,- -,n-,, - . , - - - - - . ,- - . . , , - - . , - , - - , . r - - - . - - , - -- - -,+--

5. -THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 26 AND THERMODYNAMICS ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.

. ANSWER 5.09 (1.00)

1. A divergence developing between the incore T/C and Thot.

2. A "decoupling" between Tcold and SG pressure.

[+0.5] each REFERENCE

1. B&W Technical Document, Emergency Procedures Technical Bases,

p. II.B.9.

ANSWER 5.10 (1.50)

A trend of incore T/C temperature versus RCS pressure increasing away from the SG Tsat along the saturation curve. [+1.5]

REFERENCE

1. B&W Technical Document, Emergency Procedures Technical Bases,

p. B-9.

ANSWER 5.11 (1.00)

(d.) [+1.0]

REFERENCE

1. B&W Technical Document, Emergency Procedures Technical Bases,

p. II.B-9.

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 27 AND THERMODYNAMICS ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD,J.

ANSWER 5.12 (1.50)

The warming of the shell is promoted because when a vacuum is drawn, the water in the OTSG flashes to steam at the saturation temperature. The steam helps to evenly warm the OTSG shell.

[+1.5]

REFERENCE

1. Crystal River: Basic Thermodynamics, OP-202, p. 23.

ANSWER 5.13 (1.00)

(b.) [+1.0]

REFERENCE

1. Crystal River: OP-202.

ANSWER 5.14 (1.00)

a. nucleate boiling region ;+0.5;

b. nucleate boiling region +0.5 REFERENCE

1. Crystal River: ROT-3-2, pp. 30 to 37.

b 4-

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 28 AND THERMODYNAMICS ANSWERS - . CRYSTAL RIVER -86/12/17-HUENEFELD, J.

' ANSWER 5.15 (2.00)

.a. (b.) ' [+1.0]

b. By conduction [+0.5] from the vessel and other components that are heated by conduction and gamma heating [+0.5].

REFERENCE

1. Crystal River: ROT-3-2, pp. 47 through 48.

r ' ANSWER-5.16 (3.00)

a. The idle OTSG will lag significantly behind the operating OTSG; a loss of subcooling margin in the idle loop could result if RCS pressure is reduced too fast. [+1.5]

b. There is very little coupling between the tubes and the shell of the idle OTSG. Because the tubes are already under extra tension due,to the depressurized secondary side, it is important that cooldown rate be limited to minimize tube-to-shell delta.T. [+1.5]

REFERENCE

1. Crystal River: ROT-3-3, p. 11.

- ANSWER 5.17. (2.00)

Yes [+0.5] . The potential exist thst non-condensable gases could collect on the heat transfer surface of the primary side OTSG l

tubes.- If this were to occur, non-condensable gases of sufficient quantity could shut off the boiler-condenser process due to the increase in resistance to heat transfer. [+1.5]

REFERENCE

1. Crystal River: ROT-3-5, p. 6.

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

.. 1

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 29 !

AND THERMODYNAMICS j ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.

ANSWER 5.18 (0.50)

True [+0.5]

REFERENCE

1. Crystal River: ROT-3-8, Section 10, p. 2.

ANSWER 5.19 -(1.00)

The " gray" APSRs are made of inconel and they are not as strong as the previously used black APSRs. [+1.0]

REFERENCE ~

1. Crystal River: ROT-2-10, p. 10-8.

ANSWER 5.20 (0.50)

False [+0.5]

REFERENCE

1. Crystal River: Basic Reactor Theory

2. Crystal River: NUS, Module 3.

l

~. : ..

5. -THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 30 AND THERMODYNAMICS

. ANSWERS -- CRYSTAL RIVER -

86/12/17-HUENEFELD, J.

ANSWER 5.21: (2.00).

l The delayed neutrons behave as a source. Keff after a reactor trip will be well above 0.9; therefore, significant subcritical multiplication of the delayed neutrons take place. [+2.0]

REFERENCE-

1. Crystal River: Basic Reactor Theory.

'2. Crystal River: ROT-1-33, p. 20.

ANSWER 5.22 (1.50) o< t,Ol O e

1. increasing 21L]evel ./

2. decreasing)RCS pressure

3. relatively stable RCS temperature

[+0.5] sach

1. d 1

REFERENCE

1. Crystal River: ROT-3-11, p. 23.

ANSWER 5.23 (1.50)

By observing that all control rods are operable and aligned with their groups within the limits of the Technical Specification rod withdrawal curves. [+1.5]

REFERENCE

1. Oconee: OP-0C-SPS-RT-RC.

.2. Crystal River: Basic Reactor Theory.

6%- ,, .- - . - - . - - - - , - . . - -- , . - , .

4

'4 6 .

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, PAGE 31 AND THERMODYNAMICS

. ANSWERS -- CRYSTAL RIVER - 86/12/17-HUENEFELD,J.

ANSWER 5.24 (3.00)

At 70 degrees v(f) = 0.016050 ft**3/lbm rho = 62.305 lbm/ft**3 [+0.5]

m(f) - rho y = 62.305 lbm [+0.5]

H(f) = (38.052 BTU /lbm)(62.305) = 2370.841 BTU [+0. 5]

- v(g) = 868.4 ft**3/lbm rho = 0.00115 lbm/ft**3 [+0.5]

m(g) = ev = 0.00115 lbm [+0.5]

H(g) = (1092.1)(0.00115) = 1.26 BTU [+0.5]

The 70 degrees liquid water contains the most ethalpy per ft**3.

I REFERENCE

1. Oconee: Thermodynamics, Fluid Flow and Heat Transfer for Nuclear Power Plants.

I e

- , , - , i, . - - - , --

.' \

'. . j

6. PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 32 INSTRUMENTATION ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.

ANSWER 6.01 (3.00)

There are two frames of reference from which the candidate may answer this question. The first frame of reference is to discuss the heat transfer across the steam generator from the primary to the secondary. The second frame of reference is to discuss the primary heat balance and the secondary heat balance. Either frame of reference is equally correct. Representative full credit answers from each frame of reference are given below:

FRAME OF REFERENCE 1: The ICS controls selected parameters to ensure a balance between the heat being generated in the reactor, and the heat being transferred at the steam generator. There are four variables that affect this energy balance: the total energy produced (i.e., generated megawatts), the primary to secondary heat transfer area-(1.e., OTSG level), the primary side temperature (i.e., Tave), and the secondary side saturation pressure (i.e.,

steamheaderpressure). By fixing any three of these four variables, the remaining variable must by necessity assume a specific value in order for equilibrium to exist. If level is not at this specific value, then an upset condition exists.

The ICS was designed to fix the following three variables:

generated megawatts, Tave, and steam header pressure. The ICS will a control upsets in these three variables so as to drive them to their selected setpoints. Feedwater flowrate is adjusted in concert to ensure an equilibrium steam header pressure is achieved and that any upset condition is eliminated. OTSG level is, by virtue of this control of feedwater, indirectly driven to the correct level. Any drifting from the correct level will cause an upset in steam header pressure along with a resultant correcting adjustment in feedwater flow.

..u_

. ~ . .

6. -PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 33 INSTRUMENTATION ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.

FRAME OF REFERENCE 2: In order for equilibrium heat transfer.to exist, 'the primary heat balance must equal the secondary heat balance. In the primary, the only variables to be controlled are Thot and Tcold. .The ICS does not control them specifically, but rather controls their mid-point by fixing a constant Tave. 1[n the secondary, the variables of concern are: feedwater flow, inlet enthalpy, and outlet enthalpy (i.e., steam header pressure). The inlet enthalpy, because it coincides with -the liquid phase, is very nearly a constant. The two remaining variables of control are

'directly controlled by the ICS. 0TSG level is not one of the variables affecting a primary heat balance, nor a secondary heat balance.

i REFERENCE l

1. Crystal River: ROT-3-2, pp. 31 and 32.

1 ANSWER 6.02 (1.50)

1. reactor power ( 30% (administrative)

2. oil lift pressure > 200 psig 3.- NSCCCW return flow > 260 gpm/ pump

4. upper and lower oil reservoirs above low alarm

5. seal injection flow > 3 gpm/ pump

6. controlled bleedoff' valves MUV-258, 259, 260, and 261 open

7. Tc > 500 degrees F to start the fourth RCP Sept.b, ors awt es keh Ser hit wilit .

. Any six (6) [+0.25], t+1.5~ maximum REFERENCE

1. Crystal River: ROT-3-5, p. 25 and OP-302.

1

,,,,w-- -y - - - ---,,- --- - - , - -

' 6. PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 3A INSTRUMENTATION ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.

ANSWER 6.03 (0.50)

True [+0.5]

REFERENCE

' 1.

Crystal River: ROT-3-9, pp. 3 and 4.

-ANSWER 6.04 (0.50)

True [+0.5]

REFERENCE

1. Crystal River: ROT-3-9, p. 7.

ANSWER 6.05 (0.50)

False [+0.5]

REFERENCE

1. Crystal River: ROT-3-9, p. 15.

ANSWER 6.06' (0.50)

False [+0.5]

REFERENCE

1. Crystal River: ROT-3-10.

l - t i i

6. PLANT SYSTEMS' DESIGN, CONTROL, AND PAGE 35 INSTRUMENTATION

-86/12/17-HUENEFELD,J.

ANSWERS -- CRYSTAL RIVER ANSWER 6.07 (0.50)

True [+0.5]

REFERENCE

1. Crystal River: ROT-3-10, p. 16.

ANSWER 6.08 (1.50)

RCP seal injection can flow down the cold leg stratifying in the low point, at the location of the Tc RTD. [+1.5]

REFERENCE

1. Crystal River: ROT-3-11, p. 20.

ANSWER 6.09 (2.00)

a. The high ambient temperature decreases tne density of the wet reference leg.- This causes the detector to sense a larger relative pressure of the fluid being measured, therefore, increasing the indicated level. [+1.5]

b. Yes [+0.5]

REFERENCE-

1. Crystal River: ROT-3-11, pp. 23 and 24.

w - ,- ,,--e,

6. ' PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 36 INSTRUMENTATION ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.

ANSWER 6.10 -(1.00)

(c.) - [+1.0]

REFERENCE

1. Crystal River: ROT-3-12.

' ANSWER 6.11 '(2.00)

1. decay heat removal heat exchangers

2. decay heat services seawater pump motors

3. DC pump motor handling units

4. decay heat pumps and motors

5. reactor building spray pumps and motors 61 makeup and purification (MU) pumps and motors 1A and IC Any four (4)' [+0.5] each, +2.0 maximum.

REFERENCE

1. Crystal River: ROT-4-2, p. 5.

i l

l

6. PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 37 INSTRUMENTATION ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.

ANSWER 6.12 (3.00)

a. 1. ES actuation (HPI)

2. ES bus degraded voltage

3. ES bus undervoltage (loss of voltage)

[+0.5] each

b. 1.. diesel start mode select switch (43) on main control board (one for each diesel) selected to auto

2. air shutoff valves EGV-35(A), EGV-39(B) open

3. control at engine - normal switch (on the engine gauge panel) selected to normal

4. 86 lockout relay (generator control cabinet) reset

5. SDR seal-in reset (accomplished by pressing the Reset pushbutton CPB4) located on the engine gauge panel

[+0.3] each REFERENCE

1. Crystal River: ROT-4-6, p. 55.

ANSWEF. 6.13 (1.00) the waste gas system [+1.0]

ce the M A M .. vewt; Q,m Spb.

1. Crystal River: ROT-4-2, p. 47.

I

_ _ _ _ - l

l

~

6. PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 38 4- ' INSTRUMENTATION

. ANSWERS -- CRYSTAL RIVER. -86/12/17-HUENEFELD, J.

ANSWER 6.14 (2.50)

a. Because the SW system is a source of unborated water that could potentially dilute water in the RB sump. [+1.0]

b.- 1. RB AHUs

~2. ' letdown coolers

3. RCDT ,

4. CRD coolers

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

REFERENCE

1. Crystal River: ROT-4-2, p. 33.

ANSWER 6.15 (1.50)

To ensure that_a minimum pressure (60 psig). exists for all components inside the RB thereby eliminating the SW system as a flowpath for release from the RB. [+1.5]

REFERENCE

1. Crystal River: ROT-4-2, p. 9.

l d

ANSWER 6.16 (1.00)

1. demineralized water [+0.5]

2. condensate system [+0.5]

o REFERENCE

1. Crystal River: ROT-4-2, p. 12.

i

!~

i

I?

6.- PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 39

(. INSTRUMENTATION ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.

ANSWER 6;17 (1.50)'

J As power is increased, the inlet water temperature decreases causing the

-water between the thermal shield and the reactor vessel wall to become more idense and therefore to shield more neutrons from the detectors. [+1.5]

REFERENCE

1. . Crystal River: ROT-4-10, p. 34.

ANSWER 6.18 (1.50)_

1 x 10**-9 amps [+0.5] . It takes both IR instruments greater than setpoint.

to cause -the SR to deenergize [+1.0].

REFERENCE

-1. Crystal River: ROT-4-10, Figure 11.

ANSWER 6.19 (1.50)

If-the NNI system were allowed to operate with a degraded voltage, instrument' readings would be affected. Actions taken by either the operator or by automatic system response would be in response to inaccurate readings. [+1.5]

REFERENCE

1. Crystal River: ROT-4-9, p. 15.

l

I

6. PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 40 INSTRUMENTATION ANSWERS .-- CRYSTAL RIVER -86/12/17-HUENEFELD, J.

ANSWER 6.20 (1.50)

A simulated flow signal is derived from signal generators which produce a l signal equivalent to full flow from one pump anytime the breaker to that pump.is closed. [+1.5]

REFERENCE

1. Crystal River: ROT-4-9, p. 11.

ANSWER 6.21 (1.50) 1.. Fixed Water Spray System - A fixed system designed for specific discharge patterns generally over one specific fire hazard.

2. Wet Pipe Sprinkler System - A general area system that is fully pressurized. Water flow is actuated when individual sprinkler heads get hot enough to soften fusible solder links.

3. Pre-action Sprinkler System - Similar to the wet pipe system, except the system is dry. A main flow control valve must be actuated separately by a heat or smoke detector.

[+0.5] each For examples see ROT-4-7, Tables 1, 2, and 3 on pp. 16, 20, and 24, respectively.

REFERENCE

1. Crystal River: ROT-4-7.

i i

--- e w - - m ,- - - - - - , - - - e.- -- - , - , - - , , , , . , -n,- - -,,,-----,-wr---w a--nr,-,.e--- - - - -. w r - - ----- --- - -- - , . 7 y

W WL ' CRYSTAL RIVER PAGE 40a r j TABLE 1 -- FIXED WATER SPRAY SYSTEMS Protected Area l Detector l Monitor l Actuation l Remarks Type l System l Req'ments l I Aux. Bldg. Charcoal l heat l Fire Srvce l 2 detec. l Filter Banks (5) l l Panel l l l l l l Control Complex Filter l heat l Fire Srvce l 2 detec. l Banks (2) l l Panel l l l l l l Hydrogen Seal Oil l heat l Fire Srvce l 1 detec. l Unit l l Panel l l l l l l Turbine Lube 011 l heat l Fire Srvce l 1 detec. l Storage Tank l l Panel l l l l 1 I Feedwater Pump l heat l Fire Srvce l 1 detec. l Consoles l l Panel l l l 1 1 I Unit Auxiliary l heat l Fire Srvce l 1 detec. l Also Trips Transformer l l Panel l l Water Wall l I i i Start-Up Transformer l heat l Fire Srvce l 1 detec. l Also Trips

, l l Panel l l Water Wall l I I I 3, Unit Transformers l heat l Fire Srvce l 1 detec. l Any 1 Detec Q (3-1,3-2,3-3) l l Panel l l Trips Other l l l l 2 Units &

. l l l l Water Wall l I I I TSC Air Cleanup Units l smoke l PYR-A-LARMl 1 detec. l Alarms on l l l l Fire Service l l l l Panel 9'

i ROT-4-7 16 Rev. 0 l

l

f V l

CRYSTAL RIVER PAGE 40b IABLE 2 - WET PIPE SPRINKLER SYSTEMS l Protected Area l Detector l Monitor l Type l System l Actuation l Remarks l Reg'ments l 1 1 Fire Pump House I I l press. sw.l Fire Srvce l1 sprinkler l l l Panel l head l I

958 El. Turbine Bldg l press. sw.Il Fire Srvce l1 sprinkler l2I headers I l l Panel l head l south / north l

I 119' El. Turbine Bldg l press. sw.lI Fire Srvce l1 sprinkler l2 headers I i l l Panel l head l south / north I

95' El. Aux. Building l press sw.IlPyrotronic l1 sprinkler lSeeI I note 5 l l Module 5 l head l 1 l l 119' l El. Aux. Building l press, sw.JPyrotronic l1 sprinkler lSee note 6 l l Module 5 l head l l l 958 1 El. Interm. Bldg l press. sw.llPyrotronic l1 sprinkler l l l Module 5 l head l 1 I I 119' i El. Interm. Bldg l press. sw.lPyrotronic l1 sprinkler l l l Module 5 l head l 1 1 I

^-

958 El. Control Comp. l heat lPYR-A-LARM l I

) (Ch,em/ Rad) l l 1 detec. lSee note 1 l l l 1 I I 124' El. Control Comp.l press. sw.lPyrotronic l1 sprinkler lSee note 4 i l l Module 5 l head l 1 1 l I I office Building R cords Storage Vault ll press. sw.ll Fire Srvce l1 sprinkler l Panel l head l I I ,

Technical Support l smoke I

lPYR-A-LARM l 1 detec. [See note 2 I i Center l" I go- l l i

! 1 I '

Environmental Whse i l press. sw.l Local only l1 sprinkler lSee note 3 l lat present j head l l l I Flammable Liquids I Storage Area ,)l press. sw.llat Local only l1 sprinkler lSee note 3 present l head l l l l l l l l l ROT-4-7 20 Rev. 0 l

7 .

CRYSTAL RIVER PAGE 40c TABLE 3 - PRE-ACTION SPRINKLER SYSTEM g.

Protected Area l Detector l >Ionitor l Actuation l Remarks Type l System l Reg'ments i I ,

Emergency Diese,1 Gen. l heat l,71re Srvce l1 sprinkler lafter pre-

& Control Rooms l (20) l Panel i head laction i I I Ialarm

. I I I I I I I I I

l l

l

\

1 i

l l

l 1

i i

l l

1 i

Y,.

i i

..*==

e 1

k, h

ROT-4-7 24 Rev. 0

l L ,; :

6. PLANT SYSTEMS DESIGN, CONTROL, AND PAGE 41 INSTRUMENTATION ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.

ANSWER ~6.22 (1.00)

(d.) [+1.0]

REFERENCE

1. Crystal River: Technical Specifications, p. 3/4 8-1.

2. Crystal River: ROT-4-6.

.c :

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY PAGE 42 AND RADIOLOGICAL CONTROL

. ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.

. ANSWER 7.01 (1.5b)

Because the hydrazine has a greater affinity for the exchange resin than do other ions like chlorine. Sustained use of the demineralizer with hydrazine in the RCS could lead to a release of undesirable ions from the demineralizer. [+1.5]

REFERENCE-

1. Crystal River: OP-202, p. 4.

4 ANSWER 7.02 (0.50)-

1 True [+0.5]

REFERENCE

1. Crystal River: OP-202, p. 29.

ANSWER 7.03- (1.50)

1. (EFW must be turned on full) if natural circulation stops and 4

the steam generator level is below the set oint. (It can be

, throttled when natural circulation starts.

2. (EFW must be turned on full) if its actuation was delayed.

(It can be throttled when natural circulation starts.)

3. (EFW must be. turned on full) if it is injecting into only one

. generator. -(It can be throttled when natural circulation starts.).

[?0.5] cach- #cte.: Only o.>e. ci h do.'e N v5guirv/ bv kIl uelif-t i

REFERENCE

1. Crystal River: ROT-3-3, p. 7.

7. PROCEDURES - NORMAL', ABNORMAL, EMERGENCY PAGE 43 AND RADIOLOGICAL CONTROL ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.

ANSWER 7.04 (0 k True~ [+0.5]'

Oufp4d f4 bodibyregt#9(

1. : Crystal River: ROT-3-4, p. 2.

' ANSWER 7.05 (1.00)

(c.) [+1.0]

REFERENCE 1.- Crystal River: ROT-3-4, p. 4.

ANSWER 7.06 (0.50)

True [+0.5]

REFERENCE-

1. Crystal River: AP-530, p. 10.

-ANSWER 7.07 (1.00)

a. 545 gpm +0 -45 gpm ;+0.5;

b. 600 psi +/- 50 pst + 0.5 REFERENCE

1. Crystal River: ROT-3-4, p. 14.

l l

7 .~ PROCEDURES - NORMAL, ABNORMAL, EMERGENCY PAGE 44 AND RADIOLOGICAL CONTROL ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.

ANSWER 7.08' (1.00)-

(c.) . [+1.0]

REFERENCE

1. Crystal River: ROT-3-5, p. 23.

ANSWER 7.09 (2.00)

1. Gravity drain-through the DH drop line to the RB sump. [+1.0]

2. Feed via the auxiliary spray flow path to the pressurizer.

[+1.0]

REFERENCE

1. Crystal River: ROT-3-6, p. 7.

ANSWER 7.10 '(0.50) exciter current .[+0.5]

REFERENCE

1. Crystal River: OP-203, p. 20.

l

7 .- PROCEDURES - NORMAL, ABNORMAL, EMERGENCY PAGE 45 AND RADIOLOGICAL CONTROL ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD,J.

ANSWER 7.11 (1.00)

(b.) [+1.0]

REFERENCE

1. Crystal River: OP-203, p. 3.

2. Crystal River: OP-502, p. 4.

ANSWER 7.12 (1.50)

1. Initiate full HPI 2 .~ stop all reactor coolant pumps

3. feed up OTSGs to 95% on the OR

[+0.5] each REFERENCE

1. B&W Technical Bases Document.

ANSWER 7.13 (1.50)

a. W TH

b. 65% EFIC Hi range

c. 95% EFIC Hi range

,._ [+0.5] each REFERENCE

1. Crystal River: ROT-3-3, Rev. 4, Objective 11.

i

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY PAGE 46 AND RADIOLOGICAL CONTROL ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.

ANSWER 7.14 (1.50)

1. IF asymmetric rod condition exist THEN go to AP-542

2. Select "J0G" on " SELECTOR" switch

3. Sto supply) p rod withdrawal (transfer rod (s) to alternate power REFERENCE

1. Crystal River: ROT-5-27.

2. Crystal River: AP-555, Rev. O.

ANSWER 7.15 (1.00)

1. power reduced to (60%

2. T(c selected to the unaffected leg 3, 6sle)ef NM%M M ost. R c P REFERENCE I%

1. Crystal River: ROT-5-2, Rev. 1, Objective 3C.

2. Crystal River: OP-204, Rev. 43, p. 5.

ANSWER 7.16 (2.00)

_Rentore the _ block va]Se_to operability _WLtMn oneJ1our_[t0_.7] or close the block valve and remove power from it [+0.7], or close the, P0RY and_ rempower from the_ solenoid [+0.6].

REFERENCE

1. Crystal River: STS - 3.4.3.2.

l L

4 ;- l

-.. l g.;

I 7. -PROCEDURES - NORMAL, ABNORMAL, EMERGENCY P5GE 47'. - ._

AND RADIOLOGICAL CONTROL ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J. ,

-j ANSWER 7.17 (3.00).

1. verify a valid actuation 4

2. depress "HPI actuation" pushbuttons -

3. ensure HPI trains start [+0.4]

4. ensure BWST suction valves open [+0.4] ,,

5. ensure HPI' valves open [+0.4] ^

6. ensure LPI trains start ,

7. ensure EDGs start

8. ensure diverse containment isolation actuation -

9. verify adequate subcooling margin

[+0.3] each with exceptions noted

-REFERENCE ,

/. l

1. Crystal River: AP-380, Rev. 6.

2. Crystal River: ROT-5-22.

ANSWER 7.18 (2.00)

a. Restore Tave to within its limit within 15 minutes or be in -

HOT STANDBY within the next 15 minutes. [+1.0]

b. 1. ensures that MTC is within its analyzed range

2. . ensures that the protective instrumentation is within its . A normal operating range

3. ensures that the pressurizer.is capable of being operable i l

with a steam bubble A '

(M i 4. ensures that the pressure vessel is above its minimum k '.

"T RT(NOT)

Any two (2) [+0.5] each, +1.0 maximum REFERENCE i.g;

1. Crystal River: STS 3/4.1.1.4. -

i I

l i

i

,. - ~. . ..

_.. ~

.., j:-.

," "( l

m. .

,.s w, ,

T- - )

1 i.

c.- a.. m 3.

7; PROC 0RESI-NORMAL,'ABNORMALhEMERGENCY PAGE 48 AND RADIOLOGICAL CONTROL -< '

l 7 ,

ANSWERS -- CRYSTAL. RIVER -86/12/17-lEENEFELD,J.

t i

.u. '

~.f i ! ' Wj' Cf. ,

W 4N$lER 7.19 (3.00) ? I

%% .5

= 6' [i ^ Q f"gl.' ensure GRP 1-7 rods ful7y[;inserte/f . >!

Zi ensure fitt2 decreasing " +0.4] W ; V R 3

' 3 ." Snsure mairi turbine TVs and GVs clbsed

%y ' I i .

4. a p si.tre main block valves closed . B0.41 5.%emure low load block valves closed- [+0.4] N

.. 6 / nil)ntain PZR level >/= 50' inches I a

'" 17s 7 0sure steam header press'u're at 1010 psig

(),

8. ensure output breakers open ' 1" 9.- close the block orifice bypass valve .

{^0.3]ea'chwithexceptionsno_ted .g

u o 4 .

x;

\

REFERE,NCE .;

+

3 -

r a\ :

N t

1.f '

Crystal River: AP-580, Rev. 6, pp. I through 3. t c

2. Crystal River: ROT-62-82, Generic Objective $b 'i s F s ' '

)Q s ,

e- i I t S.

P.,3m i A?SWERf.20 (2 7'50)

I 2 ? 3 .: k3 ai 5hutdown margin \shall be the instantanehds~ amount of . N .

p Trea'ctivity by which the reactor: is subc'ritical or would be' w 3Dberitical- fro'n its present c6ndition assuming: 1) no cMnge it in 4 3,tal powerishapingi assemlies(safetyand.ftdposition,and2)allcontrol;dd regupting) are fully inserted except for the single rod assembly of highest reactivity worth which is ais6med tos be fully withdrawn. [+1.5]

F a . L i; 5 s. ..

V

b. No. Even if group 1 rods are to be withdrawn, the shutdown -

value must beQ/= 1% delta k/k (i.e., Keff must be (0.99).

[+1.0] p 4 , ,

s j q, ,

, q , -

4

~

l REFERENCE- -

m

~)t m

1. Crystal Rivefi' Technical Specifications, pp.1-3 and 1-9.' 9 l

c. :

2.- Crystal River: Plantfatup,OP-202,p.37.

c n ' -

k^1 *

[,

5 7

1.

P i

_ -. -- . . - . . ..: . 4(

8. ~ ADMINISTRATIVE PROCEDURES, CONDITIONS, PAGE 49 AND LIMITATIONS ANSWERS -- CRYSTAL RIVER -

86/12/17-HUENEFELD,J.

ANSWER 8.01 (0.50)

True [+0.5]

REFERENCE

1. 10 CFR 50, Appendix B, V.

ANS O U.02 (0.50)

True [40.5)

REFERENCE'

1. 10CFR50.54,(x).

g - w-- -,+- - p- y- t vw--, ---e+- - - - - --- -w--- y ,-r -,,ey*wc-,

5

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, PAGE 50 AND LIMITATIONS ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.

ANSWER 8.03 (2.50)

1. any emergency of the classes in the Approved Emergency Plan

2. receipt of a package with > 0.01 uc/100 cm**2 on surface

3. Tech Spec safety limit exceeded 4.. automatic safety system does not function as required

5. limiting control setting exceeded

6. limiting conditions for operation exceeded

7. results of a trace investigation of radioactive shipment

8. . theft or diversion attempt of licensed special nuclear material

9. event that threatens effectiveness of physical security system

10. plant shutdown required by Tech Specs I 11. departure from Tech Specs authorized to protect public health and safety

12. degradation of plant and/or principal safety barriers

13. natural phenomenon or external condition threatens safety or hampers personnel duties for safe operation

14. -any event that causes (or should have caused) ECCS injection

15. major loss of emergency assessment capability, offsite response capability, or communications capability

16. plant event that threatens safety or hampers personnel duties for safe operation; includes fires, toxic gas release and g, ; .

ui4+e 4 % ;o e re 2o.4o3 av_ d M4, ;

Any. five (5) f+0.5] e +2.5 maximu REFERENCE

1. Crystal River: ROT-3-15, pp. I and 2. '

s:

~i ..

8. ADMINISTRATIVE PROCEDURES. CONDITIONS. PAGE 51

-AND LIMITATIONS-ANSWERS -- CRYSTAL RIVER -

86/12/17-HUENEFELD,J.

ANSWER 8.04 (1.50)

.no power range channels two IR channels two SR channels

[+0.5] each -

REFERLiCE

-1. Crystal River: ROT-4-10, p. 40.

ANSWER 8.05 (1.00)

~

The breaker'must successfully pass two (2) consecutive retests.

[+1.0]~

REFERENCE

1. Crystal River: AI-500, p. 8a.

ANSWER 8.06 (1.50)

Operations - three individuals, one is leader Maintenance - two individuals

[+1.5]

f REFERENCE

1. Crystal River: OSIM, p. V-13.

E w ,.w. - - - , , - ,- n ,- ,- - e..,-a,. , , ,. - , , . --, --,,,>s - , . , - . - - . . ,., , .-.. - . - - - , , ,v-- . . . -- , - ~ - ,

o! t ua p w m a. . m......- ~ ~ . . . . . . .

the reg- (3 )(1) A revision 01 tne ongmal unoer s SU.2hos or t av compromises FSAR containing those original pages k shall notify the NRC Operations significantly Center via the Emergency Notification safety; plant operate a that are still applicable plus new re- (B) In a condition that is outside the gnt to the placement pages shall be filed within 2 of this 24 months of either July 22,1980, or f'

System of:

(1) The declaration of any of the design basis of the plant; or Emergency Classes specified in the li- (C) In a condition not covered by the y, as pro- the date of issuance of the operating censee's approved Emergency Plan." plant's operating and emergency pro-

-nd (4) of license, whichever is later, and shall analysis bring the FSAR up to date as of a

[

6 or cedures.

mitted as maximum of 6 months prior to the 1 (ii) Of those non-Emergency events (iii) Any natural phenomenon or he operat- date of filing the revision. specified in paragraph (b) of this sec- other external condition that poses an y' tion. actual threat to the safety of the nu-

> inf orma- (ii) Not less tnan 15 days before (2) If the Emergency Notification clear power plant or significantly ham-

} contains i 50.71(e) becomes effective, the Direc. ,

" System is inoperative, the licensee pers site personnel in the performance ped. This tor of the Office of Nuclear Reactor shall make the required notifications of duties necessary for the safe oper-

)e changes Regulation shall notify by letter the via commercial telephone service, ation of the plant.

ation and licensees of those nuclear power plants other dedicated telephone system, or (iv) Any event that results or should P** " initially subject to the NRC's system- y any other method wWh wM msure have resulted in Emergency Core N atic evaluation program that they

[r that a report is made as som as p Cooling System (ECCS) discharge into pssim & need not comply with the provisions of ca the reactor coolant system as e result riate h this section while the program is being {[ 3) The ce ee s a no f the of a valid signal.

' updated conducted at their plant. The Director iI NRC immediately after notification of (v) Any event that results in a major of the Office of Nuclear Reactor Reg- the appropriate State or local agencies loss of emergency assessment capabil-iclude the ,

in the fa- ulation will notify by letter the licens- and not later than one hour af ter the ity, offsite response capability, or com-ee of each nuclear power plant being ,. time the licensee declares one of the munications capability ( e.g., signifi-bed in the evaluated when the systematic evalua. Emergency Classes.

(4) When making a report under cant portion of control room indica-

icns per. "

,er in sup. tior' program has been completed. tion. Emergency Notification System.

tendments Within 24 months after receipt of this - paragraph (a)(3) of this section, the 11 or offsite notification system).

. tons that notification, the licensee sha!! file a censee shallidentify: (vi) Any event that poses an actual Lnreviewed complete PSAR which is up to date as iI (i) The Emergency Class declared; or threat to the safety of the nuclear rialyses of of a maximum of 6 months prior to (ii) Either paragraph (b)(1), "One-I by or on the date of filing the revision. Hour Report.*' or paragraph (b)(2), power plant or significantly hampers (4) Subsequent revisions shall be "Four-Hour Report," as the paragraph site personnel in the performance of ommission ,

of this section requiring notification of duties necessary for the safe operation if c rmation filed no less frequently than annually of the nuclear power plant including

eil within and shall reflect all changes up to a f' , the Non Emergency Event. ~

maximum of 6 months prior to the j" (b) Non-cmergency et'ents ;(IIOnc. - fires, toxic gas releases, or radioactive pdated in- date of filing.

r

. jhout reporti. If not reported as a dec- releases. (2) Four-hour reports. If not report-

'1aration 'ofan Emergency Class under

'd on a re- (5, Each repluement page shall in.

paragraph (a) of this section, the 11 ed under paragraphs (a) or (b)(1) of Tall be ac- clude both a change indicator for the censee shall notify the NRC as soon as this section, the licensee shall nctify iientifies area changed. e.g., a bold line vertical. practical and in all cases within one the NRC as soon as practical and in all AR follow- ly drawn in the margin adjacent to the hour of the occurrence of any of the cases, within four hours of the occur-ne signed portion actually changed, and a page following:

rence of any of the following; pies of the change identification (date of change .

(iXA) The initiation of any nuclear (1) Any event, found while the reac-be filed or change number or both). plant shutdown required by the tor is shut down, that, had it been tr Reactor plant's Technical Specifications. found while the reactor was in oper-h gulatory (33 FR 9704. July 4.1968, as arnended at il

,C. 20555. FR 16446. Apr.19.1976; 41 FR 18303. May ABT Anj. deviatiori. froni. thiplant*S ation, would have resulted in the nu-3.1976; 45 FR 30615. May 9.1980) (Technical Specifications authorized clear power plant, including its princi-clude (f) a pal safety barriers, being seriously de-

>rized offi- dursuant to i 50.54(x) of this part .

"M2 I " '""E " "" # C"U"" "4 "I"' fli) Any event or condition during graded or being in an unanalyzed con-

~

ner the in. ments for operating nuclear power re- operation that results in the condition dition that significantly compromises ts changes of the nuclear powerplant, including plant safety.

submittal. its principal safety barriers, being seri- (11) Any event or condition that re-tation and (a) General requirements.2 (1) Each ommission nuclear power reactor licensee licensed ously degraded; or results in the nucle- sults in manual or automatic actuation ar power plant being: of any Engineered Safety Feature ommission " , ng We Mador NM h changes 'Other requirements for immediate notifi- tion System ( RPS). However, actu-fication of cation of the NRC by licensed operating nu- ,

'These Emergency Classes are addressed ation of an ESF, including the RPS, avisions of clear pow er react ors are contained else- In Appendix E of this part. that results from and is part of the bmitted to w here in t his chapter, m particular. 8 Commercial telephone number of the NRC Operations Center is (202) 951-0550. preplanned sequence during testing or i120.205. OtL403. 50.36. and 73.71.

457 456

4p '

iL U% $ 20.403 10 CFR Ch. I (1-1-85 Editieskanlacr tsgulstsry Ctmmissisn

I E tai m 7500. May 9,1969 as amended at 38 (d) Reports made by licensees in th ' - (D Estimates of each individus

{ $y m 1271.Jan.11.1973; 48 FR 33859. July 26.

19831 sponse to the requirements of this soo"9 tion must be made as follows:

pesure as required by paragraph a

31s section-f 5 20.403 Notifications of incidents. (1) Licensees that have an irstalled ' - (t!) Levels of radiation and ce

?: p Emergency Notification System sha8v g (a) Immediate noff/icoffon. Each 11- make the reports required by pareT Irations of radioactive materi fk censee shall immediately report any graphs (a) and (b) of this section te.

events involving byproduct, source, or ;

~ Nud-tim The cause of the special nuclear material possessed by the NRC with I Operations Center in accor$;h Wels or concentradons; and b the licensee that may have caused or ance 50.72 of this chapter.

(iv) Corrective steps take E

p (2) All other licensees shall make theg threatens to cause: reports required by paragraphs (a) and planned to prevent a recurrence.

t c-]I (1) Exposure of the whole body of (b) of this section by telephone and by[i . (b) Any report filed with the mission pursuant to paragraph h any individual to 25 rems or more of telegram, mailgrnm, or facsimile to the y

% l. I radiation; exposure of the skin of the Administrator of the appropriate NRCJ this section shall include for eac I, f '

whole body of any individual of 150 Regional Office listed in Appendix D; rems or more or radiation; or exposure of this part.

of the feet, ankles. hands or forearms tidual exposed the name. social -

tF number, and date of birth, r p estimate of the individuals ext W b of any individual to 375 rems or more (27 m 5905. Jane 22,1962, as amended at . The report shall be prepared s y 28 m 6823, July 3.1963; 42 FR 43965. Sees.% '

of radiation; or 1.1977; 43 FR 2719. Jan.19.1978; 48 FR. this information is stated in a se 9 I. (2)The release of radioactive materi- 33859, July 26,19831 e part of the report.

" al in concentmtions which, if averaged (c)(1)In addition to any notif-N y[ over a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , would 5 20.404 (Reserved]

tequired by 6 20.403 of this part pi 5" exceed 5,000 times the limits specified '

licensee shall make a report in $

for such materials in Appendix D, 5 20.405 Reports of overexposures and esg,, of levels of radiation or releaser k&hL !?

Table II of this part; or cesalve levels and concentrations.

3' dioactive material in excess of speelfled by 40 CFR Part 190, (3) A loss of one working week or (aX1)In addition to any notificatlos, ql W C

more of the operation of any facilities required by 5 20.403 of this part, eacM fonmental Radiation Pro' ;

affected; or licensee shall make a report in writing g Randards for Nuclear Power

@l P (4) Damage to property in excess of concerning any one of 30 thedays following;L

stlons," or in excess of license Lt% $200.000, types of incidents within of hs. tions related to compliance u (b) Twenty-four hour notf/featfon. occurrence: U CFR Part 190.

J1 M Each !!censee shall within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of (1) Each exposure of an individual tec' (2) Each report submitted

( -

L discovery of the event, report any radiation in excess of the applicable; event involving licensed material pos- limits in ll 20.101 or 20.104(a) of this paragraph (c)(1) of this sectiot describe:

sessed by the licensee that may have part, or the license; t (1)The extent of exposure of I T caused or threatens to cause: .

(11) Each exposure of an Individual l , sals to radiation or to radioacti

(1) Exposure of the whole body of to radioactive materialin excess of the ' terial;

.f any individual to 5 rems or more of ra- applicable limits in il 20.103(a)(Ik t (11) Levels of radiation and <

r e diation: exposure of the skin of the 20.10*Hs)(2), or l 20.104(b) of this part. ' l trations of radioactive mates PT whole body of s.ny individual to 30 or in the license; , . ,olved;

} i W rems or more of radiation; or exposure (!!!) Levels of radiation or concentrs. ( 111 ) The cause of the exi e p of the feet, ankles, hands, or forearms tions of radioactive material in a tv levels, or concentrations; and

, f to 75 rems or more of radiation; or stricted area in excess of any other ap. (iv) Corrective steps tak-

) (2)The release of radioactive materi- plicable limit in the license; ,

planned to assure against a s s alin concentrations which,if averaged (iv) Any incident for which notiftes. rence, including the schedu

, over a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , would tion is required by 120A03 of this ' echieving conformance with 4

p/ exceed 500 times the limits specified part; or s T: > for such materials in Appendix B, (v) Levels of radiation or concentrs i - Part 190 and with associated

-}U Table II of this part; or tions of radioactive material (whether -

(3) A loss of one day or more of tM or not involving excessive exposure of

" tenditions' (d) For holders of an opera pi K operation of any facilities affected; or any individual) in an unrestricted ares {eyden 3

V ( (4) Damage to property in excess of in excess of ten times any applicable: include n r grap!

i $2.000. limit set forth in this part or in the 11 (c) of this section must be repe

'f'p b (c) Any report filed with the Com- cense. accordance with the procedm G mission pursuant to this section shall (2) Each report required under pars i scribed in 1501 (b), (c), (d), (

4 i f be prepared so that names of individ- graph (a)(1) of this section must de- ts) of this chapter and must :

F' unis who have received exposure to ra- scribe the extent of exposure of Indl. , clude the information regul r '

p diation will be stated in a separate viduals to radiation or to radionctive , paragraphs (a) and (c) of this .l g part of the report. material, including-Incidents reported in accordan*

n-150.73 of this chapter need no' I' " 256 a ~

lii[h Mh hb fy w.

%NdMWb yy m

%l' Qc d[

n:l I l ?&,,%;N.,

v w

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, PAGE 52 AND LIMITATIONS

ANSWERS -- CRYSTAL RIVER ~ -86/12/17-HUENEFELD, J.

! ANSWER 8.07 (0.50) .

False . [+0.5]

REFERENCE ,

1. Crystal River: OSIM, p. V-9.

ANSWER 8.08 (1.00)

1. The-PORV block valve (RCV-11) is operable (capable of

. closing).

. 2. The o)erator on the switch does not leave the PORV (RCV-10) switc1 unattended (dedicated operator) until he assures that the PORY RCV-10 is closed and there is no flow thru through

<-- the PORV RCV-10 .

4

[+0.5] each .

REFERENCE-

'1. Crystal' River: OSIM, p. V-14.

ANSWER 8.09 (1.00)

]

~1. corrective maintenance is to be performed around the clock

2. a dedicated operator is to be assigned to monitor the

defective annunciators

! [+0.5] each REFERENCE

1. Crystal River: OSIM, p. V-21.

1

r 1)1 E "E

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, PAGE 53 AND LIMITATIONS ANSWERS -- CRYSTAL. RIVER -86/12/17-HUENEFELD, J.

ANSWER 8.10 (1.00)

1. . Nuclear Operations Superintendent

2. Nuclear Plant Manager _ or % c.s uit

[+0.5] each ,

. REFERENCE 1.. Crystal River: AI-500, p. 3.

ANSWER 8.11 (2.00)

1. ensure the plant is under control with existing procedures and requirements

2. call the man-on-call, NUC Ops Superintendent, SOTA, Resident NRC Rep, and the NRC

3. determine subsequent actions (i.e., cooldown) start back up

4. fill out the RX trip / shutdown report and assign shutdown number

5. ensure the information is entered in the N0 logs and the SSOD logs

[+0.4] each REFERENCE

1. Crystal River: AI-500, Rev. 55.

2. Crystal River: Section 2.4, p. 17.

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

s

8. ADMINISTRATIVE PROCEDURES. CONDITIONS. PAGE 54 AND LIMITATIONS ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD,J.

ANSWER 8.12 (2.00) 1.. Emergency Medical Team

2. Radiation Emergency Team

3. Plant Fire Brigade

4. Environmental Survey Team

5. Sampling Team

6. Emergency Repair Team

7. Dose Assessment Team Any four (4) [+0.5] each, +2.0 maximum.

REFERENCE

1. Crystal River: EM-202, p. 12.

ANSWER 8.13 (1.50)

1. Emergency Classification

2. Notifications

3. Protective Action recommendations

[+0.5] each REFERENCE

1. Crystal River: EM-202, p. 6.

a

[ ANSWER 8.14 (0.50) only one [+0.5]

REFERENCE

1. Crystal River: EM-202, p. 5.

7-

=

.8. ADMINISTRATIVE PROCEDURES, CONDITIONS, PAGE'55 AND LIMITATIONS

. ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.

ANSWER 8.15 (1.50).

Evacuate all people within a 2-mile radius and shelter all people 5

=niles in the potentially affected sectors. [+1.5]

REFERENCE

1. Crystal River: EM-202, p. 5.

ANSWER 8.16 (1.00)

Director, Nuclear Plant Operations or his designated alternate, the Man-On-Call. [+1.0]

REFERENCE 1._ Crystal River: EM-202, p. 4.

. ANSWER 8.17 (1.00) contact and notify the control room [+1.0]

REFERENCE

1. Crystal River: EM-201, p. 3.

~3

-v s

s

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, PAGE 56 AND LIMITATIONS ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.

ANSWER 8.18 '(2.00)

1. all penetrations required to be closed off during an accident are able to be closed off by an operable automatic closure system OR closed by a manual valve, blind flange, or closed and de-energized auto valve

2. containment hir locks are operable

3. equipment hatch is closed and sealed

4. containment leakage rate is within limits

5. all penetration seal mechanisms (i.e., 0-rings, gaskets) are operable

[+0.4] each REFERENCE

1. Crystal River: Technical Specifications, p.1-2.

ANSWER 8.19 (1.50) a=-

a. 500 psig and/or 200 deg F and ohne 0.5 in. diameter opening

[+1.0]

b. with written justificatic AND approval of the MOC [+0.5]

REFERENCE

1. Crystal River: ROT-5-40, Rev. O, Objectives 13 and 14.

2. Crystal River: CP-115, Rev. 56.

p .

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, PAGE 57 AND LIMITATIONS ANSWERS -- CRYSTAL RIVER -86/12/17-HUENEFELD, J.

ANSWER 8.20 (1.00)

1. portions of the condensate system

2. turbine building sump 3.- nitrogen system Any two (2) [+0.5] each, +1.0 maximum REFERENCE

1. Crystal River: RSP-101, p. 22.

\

ANSWER 8.21 (1.00)

An SRWP authorizes groups of individuals to conduct routine tasks, and RWP is more detailed and specific.

REFERENCE

1. Crystal River: RSP-101, p. 8 ANSWER 8.22 (1.00) 100 mr/hr [+1.0]

REFERENCE

1. Crystal River: RSP-101.

--m -

s

..+

c

. a ..

- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, PAGE 58 AND LIMITATIONS A"WERS -- CRYSTAL RIVER -86 / 12 /17-HUENEFELD, J.

i'

- ANSWER 8.23 - (1.00)

Electrical. The crane power bus n:ust be deenergized to prevent accidental contact and electrocution. [+1.0]

REFERENCE

1. : Crystal River: FP-601, p. 5.

ANSWER 8.24 (1.00)

It must NOT be done. (An irradiated fuel assembly must never be raised, no matter witat the circumstances.) [+1.0]

REFERENCE

1. Crystal River: FP-601, p. 4.

i i

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a EQUATION SHEET Where al ".m2 (density)1(s.!ocity)1(area)1 = (density)2(velocity)2(area)2 KE="[2.-PE = mgh PE + KE +P V l l 1 1 = PE +KE 2 +P 2 Y22 where.V = specific volume P = Pressure

-Q ='mcp(Tout -Tin). Q = UA (T,y,-Tstm) Q = m(ht-h2)

P = Po10(SUR)(t)-- P = Po et /T SUR = 26.06 T = (B-p)t T p delta K = (K,ff-1) CRi (1-Keff1) = CR 2 (1-Keff2) CR = S/(1-K,ff)

(1-Keffi) SDM =

(1-Keff) x 100%

M = (1-Keff2) K,ff decay constant = in (2) = 0.693 A1 = A e-(decay constant)x(t) t t 1/2 1/2 Water Parameters Miscellaneous Conversions 1 gallon = 8.345 lbs 1 Curie = 3.7 x 10 10 dps I i gallon = 3.78 liters 1 kg = 2.21 lbs 1 ft 3= 7.48 gallons 1 hp = 2.54 x 103 Btu /hr 3 6 Density = 62.4 lbg/ft 1 MW = 3.41 x 10 Btu /hr Density = 1 gm/cm 1 Btu = 778 ft-lbf Heat of Vaporization = 970 Btu /lbm Degrees F = (1.8 x Degrees C) + 32 Heat of Fusion = 144 Btu /lba 1 inch = 2.54 centimeters 2 1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 ft-lbm/lbf-sec

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U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINAT ION FACILITY: _CRYSIAl,_ RIVER.___,._____,,___

REACT OR T YPE: _PWR;BsM1ZZ.______________

D(ITE ADMINISTERED: _9Zl91/26__,________,,,__,__

EXAMINER: _MORGON 1 _1.______________

CANDIDATE: __________________________

INSTB WTIONS__lO_C3NDID81E: .

Uue sc:p ar at e paper ior the answers. Writo answers on one s2de only.

Staplo question st.eet on top of the answer sheets. Points for each question are indicated in parenthebes af ter the qucstion. The passing grade requires at least 70% in each category and a final grade of at least 00%. I'xamination papers will be picked up sir (6) hourn after the ex cutii nati on starts.

/. OF CATEGORY % OF CANDIDATE'S CATEGORY

_ _V_OL_L.lE _ ,_19 LOL _ ._ _S C O BE. ._ . _ _ _V(LUE__ , _ ._ _, _ _ . _ _ . __ __C GI E O 9 8 Y __ , _ _ , _ _ _, _. _ _ _

22 99._ 24112 _ . _ _ _ . , _ _ _ , , _ _ . _ _ _ _ . _ _ _ _ _ _

1. PRINCIPLES OF NUCLEf4R POWER PLANT OPERATION, THERMODYNAMICS, llEAT TRANSFER AND FLUID FLOW

_$9299_._ _2Dz99 _ _ _ _ _ _ _ _ _ _ . _ _ _______2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS

_21199__ _25123 .___________ ______._3. INSTRUMENTS AND CONTROLS

_39z99__ _25199 _ __________ _________

4. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 129299 _ ____.____ __ ________"/. Tota 1s Final Grade All work done on this e>tamination is my own. I have neither given nor received aid.

Candidate's Si.gnature

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

1. Cheating on the examination means an automatic deni al of your app 12 cation and could result in more severe penalties.

2. Rectroom tript 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 goly to facilitate legible reproductions.

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

t,. Fi11 in the date on the covet sheet of the examination (ii necessery).

6. Use only the paper ;;r ovi d ed for answers.

7. Print your name i n the upper right-hand corner of the first page ci eac t.)

section of the answer sheet.

G. Consecutively number each answer sheet, write "End of Category , _ _ " as appropriate, start each category on a new page, write pnly gn one sidg of the paper, and write "Last Page" on the last answer sheet.

Y. Numbcr each answer at to rat eciory and number , for cxample, 1. dl . 6.3.

10. Skip at least three lines between each answer.

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

12. Use abbreviations only if they are commonly used in f acilit y Li terature.

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

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

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

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

10..When you complete your examination, you shall:

a. Assemble your e:: ami n at i on as follows:

(1) Exam questions on top.

(2) Exam aids - iIgures, 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 scr ap paper and the balance of the paper that you did not use f or- answering the questions.

d. Lenve the e::ami nat i on area, as def ined by the e::aminer. If after leaving, you are found in this area while the examination is still in progrens s your 12 cense may be denied or revoked.

i l

l

"12_ EBINC I PL ES,,pE _NUC LE OB_ EOWE B_ EL.ONI_9EEBOllgN, PAGE 2 IUEBDQDYNODICS3 _UEOl_lBONSEEB_OLJD_ELUID_ELgW QUESTION 1.01 (1.00 Which one of the f ollowing is NOT a TRUE statement.

a. During Natural Circulation, the operator can increase the heat removal rate from the RCS by reducing steam pressure, or increasing OTSG 1evel.

b. A LOCA with no RCP's running can result in more inventory loss than the same LOCA with RCP's running.

c. A total and prolonged loss of OTSG fecd can lead to a loss of RCS liquid inventory. (Assume no steam generator tube leak exists.)

d. The primary concern when fuel clad t emperature reachec 1400 degrees F is the production of hydrogen.

QUESTION 1.02 (1.00)

Concerning the Schsvlor of s a m a r i u m-- 149, in the reactor, which one of'the following statements is CORRECT?

a. Most of the Sm produced comes directly from fission.

b. Most of the removal of Sm is by radioactive decay.

c. Sm reactivity is independent of flux once it has reached equi 1ibrium.

d. Equilibrium SM is reached about 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months after the inital startup of the reactor.

DUESTION 1.03 (1.00)

a. Does the change in Beta-effective over core life affect the STABLE reactor period after a reactor trip? (YES or NO)

b. HOW does the change in Beta-ef f ective over core life affect the TRANSIENT reactor period after a reactor trip (before reaching a stable period)?

(**444 CATEGORY 01 CONTINUED 014 NEXT PAGC *****)

'$i 1 f1r ' PRINCIPLES'DF' NUCLEAR POWER PLANT OPERATIONi PAGE' 13.

~

?INEB59DYN8dlCS 3_UE81_IB8NSEEB_8ND_ELUID_ELOW m -

1 TOUdSTION! 1.04 :(1.50)

~

a.. i HOW does equi 1ibrium. Xenon (Xe-eq)_ reactivity at hot ful1fpower change-as a function of core. age'(EFPD)? -00.53

b. .WHY'does Xe-eq : change as a ' f unction of core age ~(EFPD)?-E1.03 QUESTION 1.05 ( .50)

LTRUE or# FALSE? (NO expl anation required. )-

a. As condensor -vacuum is increased (pressure decreased), more energy can be extracted from the steam.

~

b .'- Increasing condensate depression will increase overall plant efficiency.

QUESTION 1.06 (1.00)-

'Wfith the reactor critical at 10 E-7 amps, rod withdrawal is used to

increase' power to 10 E-6 amps.

Gelect.the statement that correctly describes the position of rods Lafter-the power is stabilized at 10 E-6 amps.

a.. The rod position will be higher than at 10 E-7 amps because more fuel must-be exposed to the available. neutrons-to maintain the.

higher power level.

.tr. The rod position will be higher than at 10 E-7 amps to overcome the power defect.

c. The rod position will be the same. The outward rod motion needed

-to achieve a'given startup rate equals the inward motion needed to reduce the.startup rate to zero.

d.- ;The. rod position will be lower than at 10 E-7 amps due to the increased delayed neutron' population associated with the higher power level.

e

(***** CM EGORY 01 CONTINt1ED ON NEXT F wGE. *****)

+, , ,

'la__EBINQIPLES'OF NUCLEAR POWER E6 I ANT;QPERATIQN 1 PAGE. 4

. ISEEDQQyN6dlCS i_UEST'TRANSEER_ANQlFLUID_FLQW

OUESTION '1.07 _(3.00)

Assume that your. plant has experienced a degraded'electricallpower.

condition and.that youfare monitoring the. plant's cooldown on natural.

1 circulation. Indicate whether.you AGREE or DISAGREE with-each of the-following' statements. and WHY.

' ~

.a. A slow downward trend in indicated Tave is a good indication of a

' swell-established. natural circulation flow.

b. . Natural ci r cul ati on flow ratercan be increased by rapidly increasing the steam flow rate.-

.c. . A 'di f f erence between wide-r ange Th and wide-range Tc of 65'F'and

~

slowly increasing indicates developing natural circulation flow.

QUESTION 1.08 (2.00)

HOW nould the actual critical rod position vary (HIGHER, LOWER, OR NO ;

CHANGE) from the estimated critical rod posit. ion (ECP) for EACH.of the following situations.

' Include a BRIEF explanation WHY.

a.- After aLtrip from 100% power, an ECP is-calculated using zero

'. xenon' reactivity for a-startup 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after the trip.

~b. 20 EFPD is used.in the ECP instead of the actual.200 EFPD.

c. The' actual boron concentration is 100 ppm lower than the value 3

-used for the ECP.

, - d. The source count rate has decreased from 20 cps-to 10 cpn between the time the ECP was' calculated and the beginning of the stertup.

o I

(***** CATEGORY 01 CONTINUED ON NEXT PAGE 44***) ;

...t.- :

t ils.__EBINcIELES_9E_ NUCLE 88_EQWEB_EL@NI_QEgBOIlg& PAGE- 5 IUEgtiggyN9MICS, _HE81_IB9NSEEB ,9Np_ELUlp_ELgN QUEST 1UN 1.99 (1.00)

Critical Heat' Flux (CHF) is defined as-the heat' flux at which Departure ifrom Nucleate Boiling (DND) occurs. For an INCREASE in'each<of'the

-parameters below, tell how the CHF will change.

(Consider eachl parameter separately.)

U . Limit your-anser to INCREASE, DECREASE, or REMAINS THE SAME a . -- Reactor Cool ant Flow Rate.

b. Reactor - Cool ant Temperature QUESTION 1.10 (1.00)

TRUE or FALSE? (No enplanation required.)

a. :For a centrifugal pump, maximum power is required at shutoff head.

=b. For an axi al . f low pump, maximum power is required at shutoff head.

I c. The point.where NPSH(min required) ano NPSH(avail) curves meet is pump' runout.

d. NPSH required increases as pump flow decreases.

' QUESTION 1. '1 I (1.00)

The-ratio of both Pu-239 and Pu-240 atoms to U-235 atoms changes over the core life. Which one of the pairs of parameters below are most affected.by this change?

a.. Moderator temperature coefficient and doppler coefficient

b. . Doppler coefficient and beta ..

.c. " Beta and thermal neutron diffusion 1ength. l

d. Thermal neutron diffision length and moderator temperature coefficient.

( *

4 4 *- CAT EGORY 01 CONTINUED ON NEXT PAGE ***44)

p,... ..

J 'hi_iEBINCleLEg_pE_NUCLEBB_EQWEB_EL991_QPEB811gNz. 'PAGEL 6-

,11HEBM99YNOMigS,_MEel_lB8NSEEB_9ND_ELVID_ELQW EQUESTl'N' O 1.12 -(1.00)

. Following a trip f rom f ull power with the reactor shutdown and 4.RCP operating, the bias is suddenly removed from the turbine bypass

. valves. Which.one of-the following statements best describes plant-response?

a. The OTSG' saturation temperature drops causing a decrease in RCS'Tc and'a rapid drop in pressurizer level.

b '. OTSG pressure drops and 1cvels rise. The increased OTSG

-levels cause an overcooling of the RCS.

c. Since OTSG. pressures drop by several pounds, BTU limit alarms'will be' received on both generators and feedwater will cut back.

1

, 'd. Thelresulting-cooldown of the RCS will decrease the shutdown

(. margin to less than Tech Spec limits, i

OUESTION 1.13 (1.00)

OP-210, " Reactor Startup", requires that the critical rod position be taken at 10-8 amps on the intermediate range. If, during a xenon free reactor startup at MOL, the operator " overshots" 10-8 amps and instead leveled off at 10-7 amps, which one of the following statements is

. CORRECT?

a. At 10-7 amps, ther e are little or no effects from nuclear heat but since the roactor is a decade higher in power, the critical rod position would be higher.

b. At 10-7 amps, there are little or no effects from nuclear heat; therefore, the critical rod position should be the same at 10-8 amps.

c. At 10-7 amps there are substantial effects from nuclear heat; therefore, the critical rod positions will be higher than at 10-8 amps.

d. At 10-7 amps, nuclear heat xenon and the decade higher in power level will result in a higher critical rod positon.

(**444 CATEGURY 01 CONTINUED ON NEXT PAGE *****)

x e

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l7

'PAGE Jz__EBINg1PLES OF;Nyg(EAR POWER' PLANT _ OPERATION, IMEBMQQyN9dJgSi _ME91_IB9NSEEB_9NQ_ELylD_ELQW

~

QUESTION 1.14 (1.00)

A general rule'is often stated " doubling the count rate halves the margin to cr2ticality." This is mathematical 1y stated by the- _

' equation.

CR1/CR2'= 1-keff 2/1-keff 1 Which one of - the f ollowing . statements is CORRECT concerning the above statement'and equation?

a. Both keff1 and keff2 have to be less than 1.0.

b. Equel changes in keff resuit in equal changen in subcritical multiplication level,

c. - Thr .equati on only approximat es the instantaneous change in count

. rat'; once equ.ilibrium value is reached, the count rate will be higher,

d. A second doubling of the count rate will result in the reactor bccoming critical or supercritical.

QUESTION 1.15- (1.00)

.. Which one of the f ollowing statements about Net Positive Suction Head

'(NPSH) is INCORRECT?

a. NPSH is the amount by which the suction pressure is greater than the saturation pressure for the water being pumped.

b. -NPSH can be calculated by subtracting the saturation pressure from the' actual suction pressure.

c. NPSH is essential'for operation of centriiugal pumps but not for positive displacement pumps.

d.- When a pump is sr.arted, the NPSH will decrease by the amount of the pressure drop in the suction piping.

(***** CATEGORY 01 CONTINUED ON NEXT PAGE **4*4)

my a _.*-

Pi__EBING1ELgs_gE_NgCLgeB_egNgB_e(8NIdQEEB811gN,. PAGE B l

~

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.OUESTION~ 1.16~ '(1.60)

During a reactor startup,. power is!being' raised =above the point of adding heat ( PO AH ) . --

Which-one.of tho'following statements'is CORRECT?

(Assume a linear reactor power increase to about 3 percent power)

~

a.~. LSince hea' der pressure 'i s 885 psig, Tave 'will not-rise above the; corresponding satuation-temperature of 532 degrees Farenheit.

- b. 'Since the OTSGs are low level limited and header pressure i s being maintained at 885 psig,-Tave will rise and1the steam temperature

.wil1 - tend to follow Th.

c. With the. header-pressure beinq maintained'at 885 psig,~the OTSGs will remain at naturated condit4ons and no superheat will be-

-added.

d.. SinceLthe OTSGs are low level limited, the steam is superheated at zero power conditions and rises proportionally with power.

'OUESTION 1.17 (1.00)

During fuel loading, which one of the following will have NO effect on the shape of)a 1/M plot?

a. The location of the neutron sources in the core.

. b. 'Ttua strength of the neutron sources in the core.

c. The location of the neutron detectors around'the core.

, d. The orcer- of placement of fuel assemblies provided the proper enrichments are placed in their proper location.

1 N

i i

l- .(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)

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[13;;EB1Ng1ELES OF NUCLEAR POWER PLANT OPERATIONx- PAGE 9 lINEBdQDYN@ digs t_ME@l_l'y@NSEgB_@ND_ELUlp_ELQW n

LOUESTION 1.18 (1.00)-

.The. reactor trips from full' power, equilibrium xenon. conditions. .Six (6) hours later the reactor is brought critical at 10 E-8 amps on the intermediate range. -If power level is maintained at 10 E-8 amps which one hof the following statements is CORRECT concerning control rod' mot i on?--

a . -- ' Rods will have to be withdrawn since xenon'will closely follow -

its normal buildup rate.

L 'b. -Rods wil1 have to be inserted due to ::enon decay.

c. Rods will have to be rapidly inserted since the critical reactor will cause a high rate of xenon-burnoct.

d. Rods will approximately remain as is as the xenon establishes its equilibrium value for'this power level.

QUESTION 1.19 (1.00)

'A moderator is necessary to slow neutrons down to thermal energies.

Which one of.the following is the CORRECT reason for operating with thermal instead of fast neutrons?

a.. Increased neutron efficiency since thermal neutrons are less likely to leak out of the core than fast' neutrons.

b. Reactors operating primarily on fast neutrons are inherently unctable and have a higher risk of going prompt critical.

c. The fission cross section of the U-235 f uel is much higher for thermal energy neutrons than fast neutrons.

d. Doppler and moderator temperature coefficients becomo positive as neutron energy increases.

'(***** CATEGORY 01 CONTINUED ON NEXT PAGE ***4*)

I

- - _ - _ ________ __ _ _ ________ ____ __ _ n

y J ~.1 ! PRINCIPLES OF NUCLEAR POWER _PLBNT_QEERATIQN, 'PAGE 10

-IMEBU9DYN001CS,_MEBI_IB@N@EgB_9ND_E691D_EL9W

.OUESTION. 1.20 -(1.00)

A negative MTC is great to'have for safe reactor. control,'but creates a problem when-it-comes to_ cold 1 water accidents or a steam /feedline l break. Which one of the below inherent features of'the CR3 low enriched core acts intially to limit the severity of these transients?

a. Beta Effettive

b. --Doppler Coefficient

c. Pressure CoeHicient d.- Burnable Poison l

OUESTION 1.21 (3.00)-

WhatJwill-be the trend of the imbalance f or each of the f ollowing conditions described? Consider each condition separately.

Select an answer from the. choices listed for each condition.

' CONDITIONS

a. Assume reactor power-is increased from 20% to 100% steady state power operation, without any rod motion (control or APSR).

b. Assume reactor power is decreased from.100% to 50% steady state power operation, with rod motion only.

c. Reactor power-operation from Beginning of Life to-End of Life.

CHOICES-

-1. Shift in the positive direction

2. Shift in the negative direction

3. . Remain the same I.

e

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

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G iz__EBINg1ELES_QE_NQg(E88_EQWEB_ELONI_QEgBBIlgN i PAGEJ 11 INEBdQQyN9dJgS3_BE@l_IB9NSEEB_8ND_ELUlp_ELQW 4

,' QUESTION 1.22 -(1.00)

Which one of.the below: sets of parameters indicates a water system that'is subcooled by greater than 30 F7 TEMP.s (F) PRESS. (psla)'

.a. 540 1000

'b. 560 1500

c. 665= 2000

d. 640 2400 .

~ QUESTION 1123 (1.00)

TRUE or FALSE

a. Equiligrium Xenon worth increases over core li f e.

b .- Equilibrium Xenon-concentration increases over core life.

~

c. Equilibrium Samarium worth increases over core life,

d. Equilibrium Samarium concentration increases over core life.

=OUESTION- 1.24 (1.00)

Choose which one of the following indicates what the new flow and

~ discharge head would be if a veriable speed centrifugal pump, that is trunning at 1800 rpm, with a discharge head of 15 psig, flow rate of 60

.gpm, and.a motor power draw of 15 Kw, is increased in speed until the power draw is equal to 120 Kw.

a.- -60 psig 180 gpm

b. 60 psig 120 gpm

c. 120 psig 180 gpm

d. 120 psig 240 gpm

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

hz__ELONI_DEgigN_INCLUDINQ_S8EEIy_8ND_EdEBQENCy_SySIEdg PAGE 12 OUESTION 2.01 (2.00)

With respect to the 125 volt DC system:

a. On a sustained loss of AC power, the batteries have a sufficient capacity to carry their recpective loads for ___?___ hours.

Chocce the correct length of time.

1. 0.5 hours

2. 2.0 hours

3. 4.5 hours

4. 6.0 hours

b. Can a spare battery charger be used on two buses sinsultaneously?

LYES cr NO3 Explain HOW or WHY NO1.

OUESTION 2.02 (2.00)

From the following list, identify ALL of the components that can be controlled from the Remote Shutdown Panel when in the Remote Shutdown '

Panel mode of operation.

a. HPI Injection Valves

b. RCP SW Supply and Return Valves

c. Seal Injection Flow Control Valve

d. Seal Injection Isolation Valve

e. Makeup Flow Control Valve (MUV-31)

f. Decay Heat Pumps

g. SW Emergency Duty Pumps (SWP-1A & 1B)

h. SW CRDM Booster Pumps (SWP-2A & 2B)

i. CRDM SW Supply and Return Valves

j. PORV (RCV-10)

k. PORV Dlock Valve (RCV-11)

1. - Block Orifice Bypass Valve (MUV-51)

m. Industral Cooling Water Pumps (CIP-1A & 18)

n. Makeup Pumps Recirc Valve (MUV-257)

o. Emergency FW Isolation Valves

(***** CATEGORY O2 CONTINLIED UN NEXT PAGE *****)

L

T i ,

-'at__ELONI_DE@lGN_INCLyp1NQ_@@ Eely _8ND_EdEBGENCy_SYSIEMS .PAGE 13 OUESTION. 2.03 (1.00).

> Which onelof the f ollowing statements _ is'. accurate concerning the OTSG:

I --

a.- " Primary and'_ secondary sido blowdown (during plant heatup)is-accomplished by-means of drain connections near the lower.

l tubesheet..

~

b. The1startup range . instruments wi11 provide indication of flooding

of the aspirating-ports.

c. JThe auxiliary feedwater header penetrates riear the top of the OT5G shell and sprays,the feedwater on the upper cylindrical baffle.

~

d.- ' Orifice plates, located in the 1ower<downtomer section may be adjusted to balance out~the internal recirculation system.

i

' OUESTION 2.04 -( 1. 00 )

Which one of the following correctly describes the trip system of the main turbine?

4

a. $When the auto-stop (turbine control) oil pressure decreases, the interface trip valve will open allowing the EHC Control oil to dump to drain.

Lb. When the EHC Oil pressure decreases, the interface trip valve

! - will open, allowing the auto-stop (turbine-control) oil to dump to drain.

c. The interface trip valve is solenoid actuated and when open, will dump both auto-stop (turbine control) oil and EHC control s

oil to drain.

I.

d. A full turbine trip requires the servo valves for all four sets of turbine valves (throttle, governor, reheat and

! interceptor) to open.

l l

'(***** CATEGORY O2 CONTINUED ON NEXT PAGE *****)

'Bi__ELONI_DEg1GN_INCLyp1NQ_ S8Egly_9Np_EMEBQENCy_SYSIENS ) PAGEL 14-DUESTION' 2.05 (1.00)

' ~

' Which ' one of - the f ollowi ng statements is CORRECT regardi ng .the. desi gn

.of the internals vent. valves?

a.- The vent-valves are designed to open in.the event of a' HOT. leg

-break when the pressure differential reaches at least 15 psi.

b. The vent valves are designed to open in the' event of a' COLD leg break when the pressure ~ differential reaches at_least 15 psi.

.. c . In the. event of a HOT. leg break, the. valves should begin to open with a delta p of about O.3 psid and be fully open at 1.5 psid.

d. In the event of COLD leg break, the valves should begin to open with a delta P of about 0.3 psid and be fully open at 1.5 psi d.

QUESTION 2.06 (2.00)

The Nuclear Services Closed Cooling System supplies four separate

. coolers on each Reactor Coolant' Pump and Motor.- List these f our

. coolers.

I

(***** CATEGORY O2 CONTINUED ON NEXT PAGE *****)

^

, + ,

[3z__ELONI_pgg1QN_INgbyDIN9_g6 Eely _9ND_EMEBgENCy_@ySIgdg . .PAGE' 15J n

DUESTION 2.07- (3.00)

To_ prevent both BSP-1A and EFP-1.from starting on'the

~

'A' ES bus:

together, as block'4 lcads, their start sequence is controlled.

Indicate which pump and on which block it whould start for each of the following conditions.

a. If HPI1is actuated or has been actuated and then bypassed on.two out'of three channels and RD pressure is greater tnan 30 psig on

'twolout'of 3 channels, ___?___ pump will start as a block.__?__

load. EO.53

b. If-HPI is actuated or has been actuated and then bypassed on two out of three' channels and an emergency f eedwater actuation occurs,

___?___ pump will start as a block __?__ load. [0.53

c. If HP1 and RD spray and EFW are all actuated simultaneously, and the -RD . spray actuation occurred less than 5 seconds af ter the EFW actuation, the ___?___ pump will start as c block __?__ load and the ___?___ pump will start as'a block __?__ load. El.03

d. If HPI and RD spray and EFW are all actuated simultaneously, and the RD spray actuation occurred-greater than 5 seconds after the EFW actuation, .ttue ___7___ pump will start as a block __?__ load and the ___7___ pump will start as a block __?__ load. El.03

(*4*** CATEGORY O2 CONTINUED ON NEXT PAGE *****)

si__EL8HIingSIGN_lNGLUDIN@_SBEETY_AND_EMERGENGY_SYSIEMS PAGE 16'

. QUESTION 2.08 '(3.00)

.Regarding the. Emergency Feedwater Initiation and Control System (EFIC), answer the.following:

a. -List by name the valvesLthat will get an isolation. signal when the t1ain. Steam.Line Isolation (MSLI) is initiated. E1.83 (Provide a~ list for both a MANUAL.and an AUTOMATIC initiation of' the system.)

b. When the Vector Logic is enabled and no EFW overfill signal is present, the. valve open/close commands from the Feed Only Good Generator (FOGG) Logic are determined by the relative. values of the OTSG pressure. 01.23 Indicate WHAT the command signal would be'for each of the Pressure States listed ~below. (OPEN/CLOSE)

PRESS STATES SG-A Valves SG-B Valves Command Command

1. SG A & B > 600 PSIG ___A___ ___B___

2. SG A 125 PSIG > SG B ___A___ ___B___

3. - SG~A ( 600 PSIG & SG D > 600 PSIG ___A___ ___B___

4. SO A & B < 600 PSIG AND.

SG'A & B within 125 PSIG ___A___ ___B___

E8 @ 0.15 ea3 QUESTION ~ 2.09 (1.00)

Which one of the f ollowing statements identifies the pressure cavities in the RCP seals that are monitored for pressure?

a. The first and second seal cavities

b. The second and third seal cavities

c. The first and third seal cavities

d.

First, second and third seal cavities (4**** CATEGORY O2 CONTINUED ON NEXT PAGE *****)

[3i__ELONI_DESIDN_INcLyo1No_geEgIy_eyp_gDEBQENCY_SYSIEDS 'PAGE 17.

QUESTION 2.10 (1.00)

Which one of the followng actions will cause the'"A" CRDM breaker to opent

a. Placing more than one channel "A" module test switch in

" test".

b. Racking out both the channel "A" low pressure bistable and the channel'"A" high flux bistable.

c. Placing the "A" channel output trip test switch on the "A" reactor trip module in " test".

d. Placing the "A" channel in " channel bypass" and then deenergining the "A" RPS cabinet.

QUESTION 2.11 (1.00)

Which one of the following, WILL NOT result in the Main Feedwater block valves auto shutting?

a. Reactor trip

b. Feedwater demand reaches 50% decreasing c.- Main Feedwater pump trip

d. Feedwater pump' discharge crosstie is not shut.

(**t** CATEGORY O2 CONTINUED ON NEXT PAGE *****)

I8t__ELON1_DESIEN_INGLUQ1N@_S9 Eely _@NQ_EdEB@ENQy_SYSIEdg. PAGE' 18-DUESTION 2.12 (1.00)

When conducting a plant cooldown, reveral operations are required to prevent' inadvertent ES actuation. Which one of the f ollowing statements is TRUE'during a plant cooldown?

a. When RC pressure is reduced to 1800 psi, the HPI white bypass

-permit lights WILL come on.

b. If HPI was properly bypassed, the 1509 psi bistable tripped lights WILL NOT come on when pressure is reduced below this value,

c. Utuan RC ' pressure reaches 900 psi , the LPl white bypass permit lights WILL come on allowing the operator to bypass LPI and RD

. spray,

d. When each channel was bypassed,its respective amber channel bypassed light would have como ON, and the green channel function enabled Itghts and the green bypass / reset lights would have gone OUT.

QUESTION 2.13 (1.00)

Which one of the following statements about Emergency Diesel System is true?

a. The air start reservoirs f or either diesel are designed to allow at least G starts without recharging.

4 b. The fuel oil day tank for each diesel must be refilled onto per l' day during operation at rated power.

c. During startup, oil is supplied to the main drive end bearing by an auxiliary gear-driven pump.

d. Operator override valves parallel the air start solenoid valves, i

allowing manual starting of each engine, if necessary.

j I

i h

(***44 CATEGORY O2 CONTINUED ON NEXT PAGE $64**)

hz__P68NI_ DESIGN _INCLUp1NQ_S6 Eely _8ND_EMgBGENCy_Sy@lEMS PAGE 19 DUESTION 2.14 (1.00)

Choose the one correct statement concerning the operation of the Emergency Diesel Generators.

a. The RUN light on the Main Control Board when lighted indicates that the associ ated diesel is up to 900 RPM and is ready to close in on the bus if necessary.

b. 1he 'A' L "B' Start Circuit Lights f or the 'A' EDG when lite only indicate that 125 VDC power is available to the air-start control relays and that the Mode select switch is in AUTO for that diesel,

c. The STOP push button on the Main Control Doord must be held depr t e sed inor der to shut down the associated Emergency Diesel if an ES signal is present. (Assiuming a shutdown of the diesel was necessary.)

d. In order to use t he AUTO Vol+ age Regulator Control in the Control Room the Excitation Voltage Adjust Select Switch in the Emergency Diesel Gener ator Room must be selected to the Control Room position.

DUESTION 2.15 (1.00)

Which one of the falltwing conditions in an indication of improper operation for the Main Feedwater pump and turbine oil supply system 7

a. Lube oil illter delta p less than 20 psi,

b. Lighted red light on oil test panel associated with the DC oil pump.

c. Lighted red light on oil test panel associated with either AC cil pump.

d. Control oil pressure equals 110 psig.

<$$tt* CATEGORY O? CONIINUED ON NEX1 PAGE 4$$64)

'at__ELON1_QE@lGN_lNGEVQ1NQ_SOEEly_@NQ_EMEBGENGy_Sy@lEMS PAGE 20 DUESTION 2.16 (2.00)

Identify ALL of the following condition (s) that will put the Integrated Control System (ICS) into the tracking mode?

a. Cross Limits

b. Steam Generator Reactor Demand Hand / Auto Station in ' Manual'

c. A Feedwater Loop Master Hand / Auto Station in ' Manual'

d. Doth the Diamond Control Station in ' Manual' AND the Reactor Demand Hand / Auto Station in " Hand'

e. Turbine E.H.C. not in oper ator ICS mode of contr el

f. A generator output breaker tripped

g. The Reactor tripped OUEST10N 2.17 (1.00)

Ascuming the ICS is in i t s norrnal automatic lineup and power output is at 750 MWE. Which one of the following statements most accurate 1!

describes the response the ICS would take if one of the bypass valves on the 'A' side failed open?

a. The increased steam flow would stert to decrease loop 'A' Tc.

The delta Tc controller would reratio feedwater, cutting back on the 'A' side and increasing 'B' side feed. EHC will decrease turbine throttle setting to return header presuure to setting.

b. The increased steam flow would start to decrease loop 'A' Tc.

The delta Tc controller would reratic feedwater reducing 'A' feed and increasing 'B' side feed to balance delta Tc. Reactor demand would pull rods to recover Tave.

c. The increased steam flow would start to decrease steam header pressure which would then cause an error signal between header pressure and set pressure. This error signal would then be given to the control valves to close to compensate for tho increased steam f l ow.

d. The increased steam flow would caute a decrease in Tove thereby causing the reactor demand to pull rods to compensate for the decrease. With the correction being grooter t h a n S*/. . feedwater would be cross limited and incroaced by 27. to maleup for the -

increased 5, t e a ra flow.

(*4**$ CATEGORY O2 EONTINUED UN NExT PAGE *****;

c c

Nz__ELONI_DEgION_INGEUDJNg_S8EEIy_8ND_EDERGENGy_Sy@IEd5 PAGE 21 OUESTION 2.10 (2.00)

Match the following lights and switches (a-d) on the Diamond Panel with the brief explanations (1-5) of their purpose:

LIGHTS / SWITCHES PURPOSES

o. Latch 1. Clears conditions such as Asynmetric rods, out inhibits, programmer lamp malfunction, etc.

b. Motor Fault 2. Will reclose tripped breakers only when CRD groups are at in-limit. .

3. Shown power supply programmer running

c. Clamp / Clamp improperly.

Rel ease 4 Used to cross connect Auxiliary power supply and DC hold Bus.

d. Fault Reset 5. Allows CRD motors to be driven in insert direction past the electrical in-limit.

QUESTION 2.19 (1.00)

Which one of the following is a TRUC statement regarding the Remote Shutdown Panel (RSP)?

a. The pressurizer level indications on the RSP are both temperature compensated by the pressurizer temperature indications.

b. The operator has both control and flow indication of the RCP seal injection on the RSP.

c. The Emergency Duty Nuclear Services Closed Cycle Cooling pumps (SWP-1A and ID) can be started from the RSP.

d. If power is lost to the 'A' transfer switch for the RSP while in normal operation, the components controlled by that transfer switch will transfer to the RSDP f or control.

(*44$6 CGTEGORY O2 CONIINUED ON hEXT PAbE *4***)

bu__EL8ul_DESIGU_INELUDIUS_E8 Eely _8UQ_gdESQEUgy_Sy@lEUS PAGE 22 DUESTION 2.20 (2.00)

The MU & P pumps are aligned as follows:

1. 'A' pump is running supplying normal makeup and seal injection requirements.

2. 'B' and 'C' pumps are to be in ES standby For this situation, answer the following:

. How should the pump start selector switches on the 'A' and 'C' pump breaker cubicles be aligned?

b. How would you as the Control Room Operator know from the Control Room indication that the selector switches are in the right positions?

c. After an HPI initiation. e loss of offsite power occurs. Which HP1 pumps should be running after the diesels re-energize the buses?

d. What prevents an operator from selecting the "D' pump on both

'A' and *C' breaker cubicles start selector switches?

(444$4 END OF CAIEGOf<Y O2 **444)

Ih__INSIBUMNIS._BND,,CQNIBgLS PAGE 23 UUESTION 3.01 (2.00)

List the f our DIFFERENT parameter inputs into the saturation monitors.

(e.g. pressuriner level was an input, then pressurizer level A&B would be one answer not two.)

OUESTION 3.02 (2.00)

There are two cross limits associated with the ICS. DRIEFLY describe each of these limits. Include in your discussion the conditions under which the limi t will be in effect and the demand signal (s) which will be modified by the limit (including direction of change).

QUESTION 3.03 (1.50)

THere is a circutt in the CRDr1 logic which will cause the programmer to move by one stop (to a two phase energized position) whenever the programmer is stopped in a position with three phases energized.

a. What is the purpone of this circuit? E1.01

b. In which direction will the programmer turn (Insert or Withdrawal)? EO.53 OUESTION 3.04 (2.00)

Concerning the recent modification associated with RPS 'A' cabinet patch cord and ICS's Reactor Coolant Flow Signal

c. What was this modification to the ICS f1ow Signal?

b. What doc;. thin mean in terms of identifying a failed input to the ICS?

t46444 CAILbukY 03 CONIIt!ULD UM fil i l PAbE **44*)

h _INSIBUMENID_0ND_G9 NIB 9LS PAGE 24 DUESTION 3.05 (l.00)

The reactor is being shutdown. A stable neutron decay rate of -1/3 DPM SUR has been established from the point at which both intermediate rcnge channels read 5 x 10 -8 amps. One intermediate detector is properly compensated, the other is undercompensated.

Sielect the CORRECT statement concerning reset of the SRM high voltage bistablo.

a. The bistable will reset in about six minutes regardless of the undercompensation.

b. The bistable will reset in greater than six minutes due to the undercompensated detector,

c. The bistable will reset in less than sir minutes due to the undercompensated detector,

d. The backup from the power range channels will reset the bistable in about six minutes due to the undercompensated IRM not resetting the bistable.

QUESTION 3.06 (1.00)

TRUE or FALSE (NO explaination is required)

The System of concern is the Makeup and Purification Systemt

a. The block orifice has two bypasses, both of which are remotely operated from the control room.

b. The letdown line connections to the Decay Heat Removal System aru prior to the profilters and after the letdown filters.

c. A temperature element (TE-5) on the letdown line alarms at 130 degrees F and closes the letdown cooler outlet valves (MUV-40 and MUV-41) at 135 degrees F to protect the letdown coolers.

d. The makoup dominerali :ers may be operated in parallel or uurien.

(44844 Cu llbOhY 'C. CUf Ji INUlli Oil NEX1 l'nli 44444)

b.e.__INSIBUDENIS,,@ND_GQNISQLD PAGE 25 OUESTION 3.07 (1.00)

Which one of the f ollowing RPS trips is NOT bypassed when the RPS is in " Shutdown Dypass"?

a. Low Pressure (1000 psig)

b. High Flun

c. Flux / Delta Flux / Flow

d. Variable Pressure Temperature OUESTION 3.08 (2.00)

Assume that a small leak has developed on the reference leg of the pressuriner lovel transmitter selected for control of pressuri ::er lovel (assume LT-1). Describe the response you would expect to seu on EACH of the three pressuri;:or level transmitters. Explain why this responte occurs. Assume the condensing pot cannot maintain level in the reference leg.

DUESTION 3.09 (2.00)

The turbine bypass valve may be biased by a O psig, a 50 ps10 or a 125 puig signal. Indicate which one of thouc biases apply to each of the four situations below.

1. The reactor and turbine are not tripped. the turbine bypass valvos are closed, and header pressure deviation is less than 10 psig.

2. The reactor is tripped as indicated by a TRIP CONF light on the Diamond Panel.

3. The reactor and turbine are not tripped, all turbine bypass valves are NOT closed and ULD is greater than 15 percent.

4. The reactor is not tripped, and the main turbine in tripped.

(44444 C M EbOhY 0 5 CON T INUt'.D ON ND 1 F Obt. $4444i

b.__INSIBut}gNI@_QNQ_CQNIBQLQ PAGE 26 QUESTION 3.10 (3.00)

Rcgarding the existing RD spray actuation circuitry answer the following questions.

c. List the two sets of conditions, either of which will result in an RD spray PERMIT. [1.03

b. In addition to the RD Spray PERMIT, what ather signal is required for an RD Spray ACTUATION 7 [0.53

c. The RD Spray PERMIT, once set, can be reset in any one of three different ways. List each of theco three ways. [1.53 DUEST10N 3.t1 (3.00)

Regarding the Reactor Coolant Pump starting interlocks:

a. List (4) of the seven (7) interlocks. [2.03 Sotpoints are NOT required.

b. How does an Operator know those interlocks are satisfied prior to attempting a pump start. [1.0)

OUESTION 3.12 (1.00)

WHICH one of the f ollowing condi tions will cause a r eactor trip?

c. One RPS channul trips, another is in channel bypass

b. One RPS cabinet is dennergi:ed

c. One RPS channel tripo on high RCS prensure, another trips on high RCS temperaturo

d. MFP "A" and MFP "b" tost uwitctes are both placed in test ior RPS channul "A" OUESTION 3.13 (3.00)

STATE the normal and Alternato cooling water uupplion to each of the throo (3) makeup pumpa.

r (444*6 C(iTEGORY 03 CONIINULD ON tlE*1 IME **644)

9 L.__INDIBWtigNI@_@ND_GQN1896@ PAGE 27 DUESTION 3.14 (1.50)

The Nuclear Services Closed Cycle Cooling System supplies numerous components through out the plant, of these loads eight will be icolated during an emergency. ,

Lict six of those loads that are isolated during an emergency isolation.

QUESTION 3.15 (1.50)

With the Reactor Control and/or Diamond Station in Manual. Tave control is transferrred to the feedwater system. In order for the feedwater system to accept control, three additional conditions must be mot.

List these tbree condition.

QUESTION 3.16 (1.50)

Identify each of the three modes of operation of the 15ui l di ng Spray flow control valves (DSV-3 and 4), and onplain how flow is controlled in each of these modes of operation.

OllEST10N 3. l '7 (2.00)

The NN1 Controlling Temperature Selector Switch has six inputs and two outpuc.

List t h u s s' inputu and outputs.

($$$$$ LND OF Chill 40kY 03 *****)

'!i__P89CEDWBES_:_NQBd % _8pNQBNGL,_gdgBQENGf,8ND PAGE 28 BOD 196991GOL_G9 NIB 96 DUESTION 4.01 (2.50)

For-EACH of the below listed parameters, indicato at WHAT value'a MANUAL reactor trip would be required, if an automatic trip did not occur?

c. Hot Leg Temperature Increasing

b. Pressurizer Lovel Increasing

c. RCS Pressure Decreasing

d. RCS Pressure Increasing

o. Reactor Power Increasing [3 RCPJ QUESTION 4.02 (2.00)

Answer the following with regard to Reactor Coolant Pump operation:

M What erwS control room indications of abnormal RCP operation, W

a. q require the punip to be shutdown immediately?

b. What control room indication of abnormal RCP operation, requiron power lovel to be reduced to 727., at 307. min, fallowed by tripping the affected RCP7 (Includo numerical valuo.)

l

{ OUESTION 4.03 (1.00)

All Reactor Coolant Pumps have boon lost because of a loss of off-sito power. Plant control is being maintained in accordanco with the

! Natural Circulation proceduro, AP-530. According to AP-030, which one i of the follc.ing in CORRECT regarding OTGG lovel control?

a. If loss than 2 HPI pumps are availablo, than OTSG lovels should M"M bo ontablished at /. c a.M S 0%If

b. If Pressurizer level in Icss than SO" then DO NOT mtcued an OT5G lovel o f M*/. .

W

! c. If cubcooling margin is 25 degroos F and RCS pressuro is >1500 j l

psig. Then maintain OTSO levol a t M*/. . J if

d. If operatinn range level indication is lost on ono OTSG. then maintain lovel in the other OTSG at 9137. unt t i RCG flow.is reontablichod.

i (44446 CAIEGORY 04 CONT INUE.D ON NE X 1 F'6bL **4*4)

. . . r 5x__EB9GEDWBED_ _N985b_ODNDBMb_EMEB9ENGLOND .PAGE 29 69D196991GOL_G9 NIB 9L' ;

r i

' OUESTION 4.04 (1.00) i TRUE or FALSE (No Explaination required) 1 i According to OP-412 " Waste Gas Disposal Disposal Ssytem":

A

a. An increase on a portable radiation detector (such as an

! E-102) is used to indicate that all water has buen drained from the waste gas surOe tank drain pot.

l

b. Tho "Operatur at the Controls" is responsibic for verification of the radiation monitor setpointo as they are specified on the release permit.

c. If meterological conditions show a delta temperaturo of zero or positivo, you are not allowed to proceed with a gaseous rolcano.

h d. _1f_the flowrote monitor is inoperablo, Technical Specifications allow continuing the gaseous roloaso (assuming the action i statement IS met).

I' >

UUESTION 4.05 (1.00)

According to the Plant Heatup Procedure OP-202, a vacuum is drawn on the Steam Gonorators during tho heatup to 200 F.

-( WHAT is the purpose of this stop? i t

a

! OUESTION 4.06 (2.50) i During a Steam Generator tube rupture casulty:

1 a. When is 1t required to establish Emorgency Cooldown limits?

i (3 conditions required for full credit.) [1.53 !

b. What are tha TWO Emergency Cooldown limits? C1.03 i p

! OUESTION 4.07 (3.00) i A Ruactor Trip has junt occurrud. In accordanco with AP-500 [

j " Reactor Protection Actuation". WHAT are the reactor opurators j

required immodtato action steps?

d j t j ($4*** CAIEbDRY 04 COHl!NULD ON NEX1 t ' AGE *****)

i i ww-Y w -PN-*NJN- _ _ . _ _ _ _ - _ _ - . _ _ _ _ _ - _ _ . . _ _ _ - _ _ _ - _ _ _ _ . _ _ _ _ _ _ - _ _ _

PAGE 30 h z _ _ EB9G E DUB E D_:_ N9 Bd 8L a _0 k U9 Bd861_ EDEB9 EN G L @N D BOD 196991G06_G9 NIB 96 UUESTION 4.00 (1.50)

a. What is the operating area for the " Operator at the Controls" according to Al-500 " Conduct of Operations? (De specific for Non-Emergency contitions.) [0.53

b. In the event of an emergency affecting the safety of operations, the Operator at the Controls may momenterily be absent frcm the operating area provided WHAT conditions are met? [1.03 OUESTION 4.09 (2.50)

Shi f t Operating per connel, when on duty, are to remain on duty with full responsibilities of their pocition until proporly relieved.

According to Al-500 " Conduct of Operationc" HOW in " Properly Relieved" de f i ned'l (i.e. State the five things that constitute the definition.)

DUESTION 4.10 (1.00)

Suloct the CORRECT ctatement concerning transfer of Non-Nuclear Instrumentation utgnals to the ICS (as pur OP-501).

a. Disconnecting the RC flow signal source from the RPG cabinctc has no affoct on the ICS.

b. If operating signal source mal f unctionn mal:e cigr.a1 source tranufer necennary, the trancfor to another nource should be done i mmedi ately r egardl et.n of ICO operating mode.

c. When changing narrow rango RC pressure nignals, the PORV (RCV-11) should be open, with the heatern and spray valven in manual.

d. Duffer carde on buf f er card modulen may be ruptaccd whi1e leaving the affacted controlloru in auto.

QUESTION 4.11 (3.00)

WHAT immediate actions will bo tal en. according to AP-990 " Shutdown f rom Out.ut do Control Room". prior to ovacuating the control room.

( Accumn time permi t t t no f or al1 act1(;no)

(*4444 Cn1 LOURY 04 COU1INilLD ON NL: Ai PAbt 44444i

at__EBQGEDUBES_:_NOBdOLt_0BNQBdOLt_EdgBGENQY_OND PAGE 31 BOD 10L90lGOL_QQNIBQL QUESTION 4.12 (3.00)

Answer the following questions according to Emergency Plan Implementing procedure EM-201 " Duties of an Individual Who Discovers cn Emergency".

o. How is an emergency defined? [1.03

b. What are the required actions of the individual who discovers an emergency condition? [2.03 OUESTION 4.13 (2.00)

Answer the f ollowi ng questionn regarding radiation limits and guidelines,

a. Without a NRC form 4 on file what are the Whole Dody and Extremity limits per 10 CFR 207

b. With a NRC form 4 on file what are the Whole Itody and Skin limits l per 10 CFR 207

c. What ere the limitu for the fallowing per FPC aministrative exposure limits HPP-314?

l 1. Weet.ly limit when no current quarter extimate for inception quarter is on file.

2. Duarterly limit when on incomplete NRC form 4 is on file.

3. Quarterly and Yearly Whole Dody limits when all recorde. are completed. NRC form 4 on file.

l l

l I

1 l

(*4444 UniLbORY 04 CON 1IfllW.D ON NtXi fML 44444)

I, h__EBQGEDUBES_:_NQBdBLa._00NGBdeks._EdEBEENGY_eNQ PAGE 32 889196Q01G06_GQNIBQ6

('

OUESTION 4.14 (2.00)

Below is a 1ist of four diffcrent oporating modes and six different d2finitions for operating modes. Match the operating modes with its d2ffinition.

MODES DEFINITIONS Keff %'PWR Tave

a. Startup 1. >/= 0.99 </= 5% >/= 280 F

, 2. >/= 0.99 > 5% >/= 280 F

b. Hot Standby 3. ( 0.99 0 '>/= 280 F

4. ( 0.99 0 </= 200 F

c. Hot Shutdown 5. </= 0.95 0 </= 140 F

6. ( 0.99 0 200 F > Tavo > 200F

d. Cold Shutdown OUESTION 4.15 (1.00)

During an approach to criticelity. you note that RCS Tave has decreased to 522 F.

What ections tro required por Technical Specifications?

(include time limits and setpointu if epplicable.)

(44444 CAlEGOftY 04 CONTINULD ON NEX1 PAGl!. *****)

c '

r

. A__EB9GEDWBED_r_N9Bbeba_ BEN 9Bdebi_EMEB9ENGY_eND PAGE 33 BOD 196991GOL_G9 NIB 96 s

4 h

QUESTION 4.16 (1.00)

Indicate which of the f ollowing definitions accurately defines each of-

-the list Radiological areas.

' AREA

a. Radiation Area

b. Radiation Control Area

c. . Control Point Area

d. Restricted Area DEFINITIONS

1. Any area to which access is controlled for purposes of prot'ection of individuals from exposure to radiation and radioactive materials.

2. Accessible area where a major portion of the whole body could receive in excess of 100 mrem in any 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

3. Accessible area where a major portion of the whole body could receive a dose in any one hour in excess of 5 mrem, or in any 5 consecutive days a dose in excess of 100 mrem.

4. An area established by HP personnel specifically for entry / exit control of personnel, tools, and/or equipment across an RCA boundary.

5. Any area, designated by HP section, which may contain elevated levels of Radiation / Radioactive Material.

6. Where there is loose, removable contamination in oncess of 30 dpm/100 -cquare cm alphe and/or 300 dpm/100 square cm beta-gamma '

activity, or whero there are fixed contamination levels > 0.25 mram/hr at 1 inch distance.

(***** END OF CATECsORY 04 *****)

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

' 11__EBING1ELES_QE_NyGLE8B_EQWEB_ELONI_QEEBOllgN1 PAGE 34 ISEBMggyN8MlGEt_UEGI_IB6NSEEB_6ND_E(ylD_E(QW ANSWERS --- CRYSTAL. RIVER -87/01/26-MORGAN, T.

ANSWER 1.01 (1.00)

b. .

REFERENCE AND Sample Questions and Answer Keys, 01.47 (OCAN068314)

Ut,.W ATOG Guidelines. Part II, Vol. 1 CR3 ROT 3-3 pg 4 EP-290 pg 12, ROT 2-9 pg 54-57 ANSWER 1.O2 (1.00)

A c.

REFERENCE Westinghounc NTO, p. 1-5.77 Duke Power Company, FNRE; p. 169 CR3 NET, Module 3 Rn Op 10. 5 pg 2 t< 3 ANSWER 1.03 (1.00)

a. NO (0.5)

b. Transient period shorter (due to faster response w/ cmaller Beta).

(0,5)

REFERENCE AN31 PLANT SPECIFIC REf)CTOR THEORY. Core Characteristics CR3 NET Module 3 Rx Op it.4 pg 4 ANSWER 1.04 (1.50)

a. Xe-eq gets larger as a function of core age. (0.5)

b. Xe-eq is a f unction of flux (not power) (0.5) and flux increases as a function of core age (0,5).

REFERENCE ANO2 PLANT SPECIFIC REACTOR THEORY. Fission Product Poisons CH3 NE.T Mcdule 3 Rx Op 10.2 po 1, 2, t< 3

1s__EBINCIELES_OF_NQCLE88_EQWEB_ELONI_QEE68110N t PAGE 35 IHEBMQDYNOMICSx_UkOl_168NSEE8_8ND_E6QlD_ELOW ANSWERS -- CRYSTAL RIVER -07/01/26-MORGAN, T.

ANSWER 1.05 ( .50)

a. True. .

b. False. -

REFERENCE ANO1 HEAT TRANSFER THERMODYNAMICS AND FLUIDS HANDBOOK CR3 ROT 2-3 Sec 3.15 pg 31 - 35 ANSWER 1.06 (1.00) c.

REFERENCE Basic Nuclear Physics CR3 NET Module 3 Rn Op 13.5 pg 3 ANSWER 1.07 (3.00)

a. Disagree LO.25] - Tave is a calculated indication and one parameter decreasing will cause Tave to decrease giving a f alse indication. [0.75]

(Agree - If other indications are used in conjunction with Tave.)

b. Agree [0.25] - lowering steam pressure will lower saturation temp which will increase heat' transfer across the tubes. [0.75]

(If examinee disagrees on the basis that voids may form in the hot leg which will interupt N.C. flow, he must state this will occur only ofter plant cooldown has caused sufficient contraction of the RCS to empty the pressuriner thus shifting the bubble to the vessel head or the hot legs)

c. Di sagr ee CO.25] - Natural Circulation is indicated by Th stabili:ing then tends to decrease. Tc tends to track OTSG sat.

temp. [0.75]

REFERENCE DOW17'7 EOB CR3 ROT 3-3 pg 6 5 7

It__EBINGlELE@_QE_NUGLE88_EQWEB_ELON1_QEEB811QNs PAGE 36 IUEBMQQYN8 MIGS t _UE81_IB8NSEEB_8NQ_ELylp_ELQW ANSWERS -- CRYSTAL RIVER -07/01/26-MORGAN, T.

ANSWER 1.00 (2.00)

a. Higher [0.253 due to Xenon peak after the trip. EO.253 ,

b. Higher 00.25] due to less excess reactivity remaining in the core -

at 200 EFPD.EO.253

c. Lower [0.253 with less boron in the RCS the Control Rods will need to be inserted to compensate for the lower reactivity of the boron.EO.253

d. No Change EO.253 since initial count rate has no effect on ECP (only pcWor at which ECP occurs) [0.253 (Accept HIGHER if examinee states an assumption that negative reactivity has been added from some source to cause count rate to decreaue.)

REFERENCE Nuclear Funs.amental s CR3 NET Module 3 R:' Op 12.5 pg 1 ANSWER 1.09 (1.00)

7. Increase

b. Decrease (2 @ O.5 ca)

REFERENCE Westinghouse Thermal Science; Ch. 13, pg 33-52 CR3 ROi 3-2 pg 17 6 18, ROT 2-9 pg 39 6 40 ANSWER 1.10 (1.00) .

a. False,

b. True.

c. True

d. False. (4 answers @ 0.5 ea.)

REFERENCE ANO1 THERMODYNAMICS HEAT TRANSFER AND FLUIDS HANDBOOK CR3 ROT 2-8 pg 19, 20, 33, 34 6 37 ANSWER 1.11 (1.00) b.

=;

W:

lt__EB1NG1ELES_QF NUCLEAR'POWEB_EL8NI_QPEBOligNt - PAGE' 37 IBEBd90YNediQgx_UE81,IB@NSEE8_8NQ,ELQ1p_E6QW ANSWERS -- CRYSTAL RIVER -87/01/26-MORGAN, T.

> REFERENCE-

- NUS, NETRO, p. 11.4-3.

CR3-NET Module-3 Px Op'-11.4 pg 3 ANSWER _ 1.12 ( 1. 00)'

.a.

REFERENCE

-STM-504 CR3 ROT 3-0 SLB ,

ANSWER - 1.13 (1.00) b '.

. . . .c

-REFERENCE OP-210, Rev. 16, p. O CR3 NET Modulo 3 Rx Op 13.5 pg 2 & 3 ANSWER 1.14 (1.00) a.

REFERENCE Duke Power Company,FNRE; pp. 115-120 CR3 NET, Modulo 3 Rx Op 12.1 pg 4 ANSWER 1.15 (1.00) c.

REFERENCE NUS Plant Performanco,-pp 6.5-1 to 6.b-3 CR3 ROT 2-8 pg 34 ANSWER 1.16 (1.00)

b. .

1s__EBINglELgg_QE_NQgLE@B_EQWEB_EL8NI_QEEB811gN i PAGE- 38

. y; -IBESd90XN001GEi_UE01_IBONSEEB_8ND_ELUID_ELOW gja ANSWERS -- CRYSTAL RIVER -87/01/26-MORGAN, T.

REFERENCE

CR3-ROT 3-7 pg 3 and Fig 2 ANSWER 1.17 '( 1. 00) . .

b.

REFERENCE Duke Power Company, FNRE;.pages 126-128

CR3 NET Module 3 Rx Op 12.3-pg 1 &2 LANSWER 1.'18 (1.00) a.

REFERENCE CR3 NET Module 3 Rx Op : 10.3. pg 4 t< Fig 10.3-2

' ANSWER _ 1.19 (1.00)

C . -~

REFERENCE-CR3 NET Module 3 .Fb: Op 1.4 pg 1 & 2-,

' ANSWER 1.20 (1.00) b, .

REFERENCE-CR3_ NET Module 3 Rx Op 8.2 pg 2

, ANSWER' 1.21 -(3.00)

.i

a. 2.

' b .' 2.

c. 34 l

1t__EBINQ1PLES_QF NUCLEAR POWER PLANT-QEgB011QUt PAGE~ 39

--IMEBOODYN@dlC@t_Ug81_IB6NSEgB_6NQ_E(UlQ_ELQW

ANSWERS - ' CRYSTAL R10ER -87/01/26-MORGAN, T.

REFERENCE -

CR3. NET Module. 3 Reactor Operation 7.3 t< 13.7 ANSWER 1.22 (1.00) -

b.

REFERENCE

. Thermodynamics, Fluid-' flow, and Heat. Transfer for Nuclear Power Plants Steam: Tables ANSWER' 1.23 (1.00)

^a' . 'True

b. ~ False

c. True r

-d. False REFERENCE ",

CR3 NET. Module 3 Rx Ops 10.2 and 10.5.1 and-CR3 Question Bank pg 2 ANSWER 1.24 (1.00)

!- b.

REFERENCE CR3 ROT 2-8 PG 15 and CR3 Question Bank J

g g

9 cui

- s -w ve.n.- - - . - - - - , n - . - - , , , - - - , , - .-r,m-w--e w wg e m.,e,w w e-4- g v-e-r,,

2. PLANT DESIGN INCLUDING _SAEETy_AND_ EMERGENCY _ SYSTEMS PAGE 40 ANSWERS -- CRYSTAL RIVER -87/01/26-MORGAN, T.

ANSWER 2.01 (2.00)

a. 2. (1.0)

b. No, CO.53 A mechanical interlock prevents the charger from being ,

aligned to more than one bank of batteries at a time. LO.5]

REFERENCE CR3 ANO-111 pg 8 and ANO-110 pg 7 ANSWER 2.02 (2.00) a, b, e. g, i, j, k, 1, n, t< o (10 5) 0.2 ea)

REFERENCE CR3 ROT-4-16 ?, Question Bank ANSWER 2.03 (1.00) d.

REFERENCE STM-2-34, 27, 54 snd 16 CR3 NAO-113, pg 3, 5, 7, 9 t< 10 ANSWER 2.04 (1.00) a.

REFERENCE STM-28-5.

CR3 NAO-112 pg 5 ANSWER 2.05 (1.00) d.

REFERENCE STM-1--17 to 20 CR3 AND-77, ROT-4-34, Obj 10 CR3 Question Bank

2t__PLONI_DE@lGU_1NCLUDINQ_@@EEJy_6ND_EMEGGENCY_SYSIEMS PAGE 41 ANSWERS -- CRYSTAL RIVER -87/01/26-MORGAN, T.

1 ANSWER 2.06 (2.00)

1. Upper Thrust Oil Bearing Cooler

2. Lower Guide Bearing Cooler .

3. Motor air cooler

4. Seal Area Cooler (4 & O.5 ea)

REFERENCE STM-23-7 and 8 CR3 ANO-78 pg 11 ANSWER 2.07 (3.00)

PUMP BLOCK

a. BSP-1A 4

b. EFP-1 5 - " * -

c. BSP-1A 4 EFP-1 5 or EFP-1 5 B5P-1A 4

d. EFP-1 5 BSP-1A 6 or BSP-14 6 EFP-1 5 (6 & O.5 ea)

REFERENCE CR3 ROT-4-13 pg 51 and 52

~

32__ELONI_ DESIGN _lNCLUDIN@_S@EEIy_GND_ EMERGENCY _SYSIEMS PAGE 42 ANSWERS -- CRYSTAL RIVER ~87/01/26-MORGAN, T.

ANSWER 2.08 (3.00)

a. Auto and Manual MSIV's ,

Automatic Only .

Main Feedwater Block Valve Lo-Load Block Valve Startup Block Valve MFWP Suction Valve Feedwater Crosstie Valve (6 valves e 0.25 ea)

(0.3 for correct Auto & Manual Seperation)

b. 1. A open B open

2. A open B close

3. A close B open

4. A open B open (8 e 0.15 ea)

REFERENCE CR SLRM handout CR3 ROT-4-15 pg 18 & 22 ANSWER 2.09 (1.00) b.

REFERENCE CR3 ANO-78 pg 7 and CR3 Question Bank ANSWER 2.10 (1.00) d.

REFERENCE CR-3 ROT-4-12 pg 3, 4, Fig 2 6 9 & CR3 Question Bank ANSWER 2.11 (1.00) b.

REFERENCE CR3 OP-504, 5.2.27 5.2.28 Page 9,

I 3.__E60Ul_DESIGU_INQQUQ1NG_g8EEIY_8NQ_gdERQENGy_Qy@lEdS' PAGE 43 ANSWERS -- CRYSTAL RIVER -87/01/26-MORGAN, T.

ANSWER 2.12 (1.00) d.

REFERENCE -

CR3 ROT-4-13 pg 62 t< 63 ANSWER 2.13 (1.00) d.

REFERENCE CR3 ROT-4-6 pg 7 ANSWER 2.14 (1.00) b.

REFERENCE CR3 ROT-4-6 and CR3 Ouestion Bank 3/7/86 ANSWER 2.15 (1.00) b.

REFERENCE CR3 NGO- 982, 23 ANSWER 2.16 (2.00) a., b., d., e. t< g . (5 & O.40 ea)

REFERENCE CR3 ROT-4-14 pg 12 ANSWER 2.17 (1.00)

C. .

2E__ELONI_QESIGN_INCLUQLNQ_@@ Eely _@NQ_ EMERGENCY _SYSIEMS PAGE 44 t ANSWERS -- ' CRYSTAL' RIVER ' -87/01/26-MORbAN,.-T.

EREFERENCE'

-CR3 ROT-4-141 pg.115-119'

? ANSWER .2.18 ( 2.' 00 ) -

a. 5.

b. 3.

c . - 4. .

d. 1. (4 G O.5 ea)

~ REFERENCE CR3 RO-93 Lesson plan f or CRD Electric CR3 Question Bank requal. exam

.2/3/86

' ANSWER '2.19' (1.00)

- - ~

c.

REFERENCE

-CR3 ROT-4-16 pg 97,.1, 53 CR3 Question Bank Requal Exam 2/3/86 ANSWER 2.20 (2.00)

a. The switch on the 'A' cubicle selected to 'B'. 00.253 The switch on the 'C'. cubicle selected to "C'. EO.253

~b. The operator should see'the white ES selected light for the 'B' MUP on the control switch-for the 'B' MUP on the 'A' side of the-ES panel is illuminated [0.253, and the ES selected light for the

'C' MUP is illuminated. CO.253

, c. 'B' and 'C' E2 G O.25 ea3

d. There is only one control handle for operating both selector EO.253' switches and the handle cannot be removed from the switch if it is selected to the 'B' MUP position on either circuit breaker. 00.25]

REFERENCE CR3 AND-82 pg 27 & 28 and CR3 Question Bank 3/7/86 s -

b

c .

};- " INSTRUMENTS AND CONTROLS: PAGE 4 5 ~-

ANSWERS --- CRYSTAL RIVER -87/01/26-MORGAN, T.

ANSWER- H3.01 (2.00)

1. Incore Thermocouples:

2. - Hot Leg RTD's A.& B ,

3. Cold Leg RTD's A & B,

4. RCS WR Pressure A & B REFERENCE CR3 ROT 4-9 pg 11'& CR3 Question Bank ANSWER 3.02 (2.00)

1. FEEDWATER CROSS LIMIT (0.25)

If. f eedwater flow ' deviates f rom f eedwater demand ~ by >/= 5% (0.25),

reactor demand will be modified by the amount of error in' excess

'of 5% (0.25). Reactor demand will only be deceased by.the actions ~

of?this limiter (0.25)..

2. REACTOR CROSS L'IMIT (0.25)-

If reactor power deviates f rom reactor demand by >/= 5% (0.25),

then feedwater demand will be modified by the amount of-errorrin excess of 5%-(0.25). Feedwater demand may be increased or. decreased to match reactor power (0.25).

-REFERENCE; CR3 ROT-4-14, CR3 Question Dank

-ANSWER 3.03 (1.50)

a. This circuit ensures that the heat generation in the stator is within (0.5) the capability of the cooling water flow supplied (0.5).

b. Insert (0,5) 8.

REFERENCE CR3 ROT-4-28 (CRDM ELECTRICAL) & CR3 Question Bank i-i I

+

h lz__INDIBUMENIS_8NQ_CQNIBQLS PAGE' 46 TANSWERS -- CRYSTAL RIVER ' -87/01/26-MORGAN, T..

1

- ANSWER 3.04 (2.00)

a. Modification removed.the ICS input from the actual ~ flow signal (0.5) and' replaced this signal with a. simulated signal. (0.5) . .'

b. The only method available.to identify a failed input to;the -

ICS is by identification of the ICS response. (1.0)

' REFERENCE

[

CR3 ROT-409 pg 10 & 11 L

I-ANSWER 3.05 (1.00) 1

a.-

. REFERENCE CR3 ROT-4-10 pg 16 1

ANSWER 3.06 (1.00)

a. False i- b. .True l c. False

d. True (4 9 0.5 ea)

REFERENCE '

! ~STM 17-4, 4, 5, 7.

.CR3 ANO-82 pg 9, 10, 13, 29 & Fig 1 ANSWER 3.07 (1.00) b.

! REFERENCE

'CR-3 ROT-4-12 pg 8 fig 2 l

L l

? .

r I

l--

It__IUSIEQMENIS_8NQ_CQNIBQLS PAGE 47 ANSWERS -- CRYSTAL RIVER -87/01/26-MORGAN, T.

ANSWER 3.08 (2.00)

A loss of fluid from the reference leg would cause a decrease in the high side pressure to the delta pressure transmitter (0.5). This .

would cause the indicated level to approach full scale (0.5). -

However, since this transmitter is selected for control, makeup will be reduced to maintain level at setpoint (0.5). This will result in a steady decrease in the indications fro LT-2 and LT-3 while the indication for LT-1 will be steady (0.5).

REFERENCE CR3 ROT-3-11 b CR3 Ouestion leank ANSWER 3.09 (2.00)

1. 50 psig

2. 125 psig

3. 50 psig

4. O psig (4 S 0.5 ea)

REFERENCE CR3 ROT-4-14 pg 137, 138 L 139 ANSWER 3.10 (3.00)

a. 1. HPI Block 3 and Blo'ck 4 actuated on 2 out of 3 channels (RC-1, RC-2, RC-3) or (0.5)

2. HPI bypassed after an actuation on 2 out of 3 channels (RC-1, RC-2, RC-3) (0.5)

b. >30 psig signal (2 of 3 channels RB-4, RB-5, RB-6) (0.5)

c. 1. 2 reset pushbuttons on ESF-ALB sections of main control board. (0.5)

2. If HPI is bypassed and the associated ES 4160V bus UV relays actuate, the permit will be automatically reset. (0.5)

3. If HPI is bypassed and is then reactuated by either LPI or RB ISO and COOLING, the permit will automatically be reset.

(0.5)

REFERENCE CR3 ANO-90 pg 13 h 14

f IIi__1NEIBudENIE_eNQ_GQNIBQ65 PAGE -48 fANSWERS -- CRYSTAL-RIVER -87/01/26-MORGAN, T.:

.. ANSWER? 3.11 .( 3.00) a.- 1. Lift oil pressure

2. SW flow .

3. Reservoir oil level -

4. Seal injection flow

5. Reactor power

6. Fourth pump only

7. ' Controlled bleedoff valve open (4 9 0.5 ea)

b. The status of the starting interlocks is displayed by white lights on the pump control switch back plate (0.5), when the lights are energized the interlock is satisfied (0.5).

REFERENCE-CR3 AND-78 pg 9

' ANSWER 3.12 (1.00)-

C.

REFERENCE CR3 ROT-4-12 figure 2 ANSWER 3.'13 (3.00)

NORMAL ALTERNATE MUP 3A NSCCCS- -DHCCCS 'A' MUP 3B NSCCCS NONE MUP 3C- DHCCCS 'B' NSCCCS E6 9 0.5'ea3

. REFERENCE

- CR3 ANO-82 pg 23 a

1t__INEIBudENIEiBNQ_CQNIBQLE -PAGE 49 ANSWERS -- CRYSTAL RIVER -87/01/26-MORGAN, T. ,

i 4

ANSWER 3.14' (1.50)

1. Letdown Coolers 2 .- :RC Drain Tank Cooler. , ,

3. CRDM-coils -

4. Reactor Coolant Pumps

5. . Seal Return Coolers

6. Waste Evaporator

7. RC Evaporator

8. Waste Gas Compressors (6 0 0.25 ea)

REFERENCE

'CR3 ANO-84 pg i i

t- -ANSWER- 3.15 (1.50)

1. Neither OTSG can be.on-BTU limits

' One OTSG must be off low level limits

2.

3. One FW Loop master station must be in auto (3 @ 0.5 ca) ,

1

REFERENCE I

CR3 ROT-4-14.pg 68 b ANSWER 3.16 (1.50) i l '1. Local Manual EO.253 - Co'ntrol of the valves is done manually by use of the toggle bar on the controller. E0.25]

Local Automatic E0.253 - Automatic control at setpoint selected-by 2.

operator. CO.253

3. Remote EO.253 - Valve controlled in automatic to maintain 1550 j gpm. CO.25]

, REFERENCE i CR3 ANO-90 pg 9

! i

)..

.e 5

I

l Iz__INQlBQdENIS_6NQ_CQNIBQLS PAGE 50 ANSWERS -- CRYSTAL RIVER -07/01/26-MORGAN, T.

ANSWER 3.17 (2.00)

INPUTS 1. Loop A Tc OUTPUTS 1. Selected Th

2. Loop B Tc 2. Selected Tc ,

3. Loop A & B Tc avo -

4. Loop A Th

5. Loop B Th

6. Loop A & B Th avg (8 @ 0.25 ea)

REFERENCE CR3 ROT-4-9 pg 8

. . ~ .

14___EB9GEDUBES..:_NQBd@Lt_6BNQBd86x_gdEBGENQY_8NQ PAGE 51 BOQ196901G66_GQNIBQL ANSWERS -- CRYSTAL-RIVER -87/01/26-MORGAN, T.

ANSWER 4.01 (2.50)

a. 610 F (+/- 5 F) ,

b. 290 in (+/- 5 in)

c. 1800 psig (+/- 20 psig)

d. 2300 psig (+/- 20 psig)

e. 80, % (+/- 5 %) (5 3 0.5 ca)

REFERENCE CR3 ROT-4-12 fig 2, ROT-4-1 L CR3 Question Bank ANSWER 4.02 (2.00)

a. Controlled bleed off temp. >170 degrees F (0.5) (Verified) High seal stage pressure drop > 2/3 RCS pressure (0.5) -***-

. Total seal outflow exceeds 2.5 gpm (0.5) and is rapidly increasino (O. E 44 % 4 &psh t / 90'h t s h)

REFERENCE v CR3 OP-302, Rev. 25 pg. 6 ANSWER 4.03 (1.00) c.

REFERENCE CR3 AP-530 pg 3 of 12 ANSWER 4.04 (1.00)

a. True

b. False

c. False

d. True REFERENCE ,

CR3 OP-412, Page 8, 11, 13 & 16 l

L 4 t__BBQGEDQBES_ _NQBdebt_8ENOBdebt_EDEBGENQY_@NQ PAGE 52-60D19b091GOL_G9 NIB 96

-ANSWERS -- CRYSTAL RIVER -87/01/26-MORGAN, T.

ANSWER 4.05- (1.00) 14 vacuum is drawn on the S/G's to ensure even heating. .

REFERENCE CR3 OP-202 pg 23 ANSWER 4.06 (2.50)

a. 1. If main condenser is not available

2. HPI is required to maintain P2R level

3. RCP's are not operating E3 G O.5 ea3

b. 1. When RC temp > 500 F then cocidown at </= 240 F/HR

2. When RC temp </= 500 F then cooldown at </= 100 F/HR E2 G O.5 ea3 REFERENCE CR3 EP-390 pg.6 Of 12 ANSWER 4.07 (3.00)

1. Ensure GRP 1-7 rods f ully inserted.

2. Ensure f lu>: decreasing

3. Ensure main turbine TV's and GV's closed

4. Ensure main block valves closed

5. Ensure low load block valves closed

6. Maintain PZR level >/= 50 "

7. Ensure steam header pressure at 1010 psig 8.. Ensure output breakers open, (BKR 1661 and 1662)

9. Close block orifice bypass valve, (MUV-51)

E9 G O'.33 ca]

REFERENCE CR3 AP-580 pg 2 & 3 ANSWER 4.08 (1.50) .

a. The Red-Carpeted general area in the Control Center. (0.5)

b. The Operator remains within the confines of the Control Center (0.5) and maintains an unobstructed view of the operational control panels (0.5).

it__EBQGEDUBEE_:_N9Bd66t_8hNQBd8(t_Edg8@gNQY_8NQ PAGE 53-88D196001GOL_QQN18QL-

? ANSWERS -- CRYSTAL RIVER- -87/01/26-MORGAN, T.

REFERENCE CR3 AI-500 pg.6 ANSWER- 4.09 (2.50)

Reli ef Licensed Operators must:

1. Walk down Control Board to determine plant status.

'2. Review Appropriate Operators Log and sign off as required.

3. Discuss operations and plant status with on-duty personnel.

4. Read short term inntructions.

5.- Perform ~a walk down of the area of responsibility.

[5 & O.5 ea]

REFERENCE CR3 AI-500 pg 11 ANSWER 4.10 (1.00)

REFERENCE CR3 OP-501, Rev. 8, pgs. 2, 3 and 6 ANSWER 4.11 (3.00)

1. Announce over the PA system that the control room is being evacuated.

2. Depress " Reactor Trip" push button.

3. Ensure the turbine trips.

4. Actuate Emergency Feed Flow and Assure OTSG 1evels are being controlled.

5. Actuate Main Feed Isolation.

6. Actuate Main Steam Isolation.

.7. Close MUW-49

8. Ensure power is available to all ES buses.

9. Close RCV-11 (PORV block valve).

(9 3 0.33 ea)

REFERENCE CR3 AP-990 pg 2 84 3

St__EBQGEQQBES_ _NQBd86t_@ENQBd8Lt_EMEBQENCy_8NQ PAGE 54 E8910600lG86_GQNIBQL ANSWERS -- CRYSTAL RIVER -87/01/26-MORGAN, T.

ANSWER 4.12 (3.00)

a. An emergency is an incident or condition requiring immediate ,

attention [0.53, which could result in personnel injury 00.25] -

and/or damage to plant components [0.253.

b. 1. Notify the Control Room giving the following information; (0.2)

A. Type of Emergency (0.2)

D. Location of Emergency (0.2)

C. Whether or not there are injured personnel (0.2)

D. Visible damage to the plant (0.2)

2. Take any immediate actions qualified to perform (0.25)

3. Withdraw to saf e area (0.25)

4. If possibility of personnel contamination exists, remain in safe area until monitored. (0.25)

5. Follow the instructions issued by the Control Room. (0.25)

REFERENCE CR3.EM-201 pg 1, 3 8< 4; CR3 Question Bank 7.12, 3/7/86 pg 20 ANSWER 4.13 (2.00)

a. Whole Body is 1.25 Rem /Otr EO.253 and Extremity is 18.75 Rem /Otr EO.253

b. Whole Body is 3.0 Rem /Otr not to exceed 5(n-18) CO.253 and Skin is 7.5 Rem /Otr 00.253

c. 1. O.125 Rem /Wk [0.25I

2. 0.90 Rom /Otr CO.253

3. 1.25 Rem /Otr [0.253 and 4.0 Rem /Yr E0.25]

REFERENCE CR3 ANO-49 Table i pg 1 ANSWER 4.14 (2.00)

a. 1.

b. 3.

.a

c. 6.

d. 4. (4 @ 0.5 ea)

REFERENCE CR3 ROT-5-1 STS sec 1.0 table 1.1 pg 1-9 W.

. -q Jaz__EB9GEDUBES_ _N9BdeLa_0DN9Bbebr_EDEBEENgy_@ND PAGE 55.

BOD 196991906_cggIggL-

= ANSWERS -- CRYST AL RIt 'ER. -87/01/26-MORGAN, T.

ANSWER 4.15 -(1.00)

Restore'Tave to within limits of >525 F within 15 minutes CO.53 or be ..

lLn Hot Standby within the next 15 minutes CO.53. .

REFERENCE CR3 STS 3.1.1.4 PG 3/4-1-5 and CR3 Question' Dank 2/3/86 ANSWER 4.16 (1.00) a.' 3.

b. 5.

c. 4.

d. 1.

i-

=-

REFERENCE

! CR3 ANO-49 Tabic 2 pg 3 I

1

)

1 4

I 1.

3 i

f, i

i

7

~

TEST CROSS REFERENCE PAGE 1

DUESTID'N VALUE REFERENCEL

-01.01 1.00 TLMOOOO359 01.02- 1.00 TLMOOOO360 01.03 1.00 TLMOOOO361' 01.04; 1.50 .TLMOOOO362 01.05 .50- TLMOOOO363 01.06 1.00 TLMOOOO364 01.07 3.00 TLMOOOO365 01.08' 2.00 'TLMOOOO366

. 01.09 1.00 TLMOOOO367 i 01.10 1.'00 TLMOOOO368 i 01.11 1.00~ TLMOOOO369 J01.12 1.00 TLMOOOO370 101.13 1.00' TLMOOOO371 l' ~O1.14 1.00 TLMOOOO372 l 01.'15 1.00 TLMOOOO373 ~

! 01.16 1.00 TLMOOOO374 01.17 1.00 TLMOOOO375 01.18. 1.00 TLMOOOO377 01.19 1.00 TLMOOOO378 01.20 1.00 TLMOOOO379-01.21 3.00 TLMOOOO415

a-

01.22 1.00 TLMOOOO416

} 01.23 1.00 TLMOOOO417 01.24 1.00 TLMOOOO418

~~

)

! 29.00 l '02.01 2.00 TLMOOOO356 02.02 2.00 TLMOOOO350 02.03 1.00 TLMOOOO380 i 02.04 1.00 TLMOOOO381 02.05 1.00 TLMOOOO382 '

[ 02.06 2.00 TLMOOOO383 i 02.07 3.00 TLMOOOO384

! 02.08 3.00 TLMOOOO385

!. 02.09 1.00 TLMOOOO386 02.10 1.00 TLMOOOO389 02.11 1.00 TLMOOOO392 l 02.12 1.00 TLMOOOO393 l 02.13. 1.00 TLMOOOO394 i 02.14 1.00 TLMOOOO395

! 02.15- 1.00 TLMOOOO396 l 0

'2.16 2.00 TLMOOOO397 l 02.17 1.00 TLMOOOO398 02.18 2.00 TLMOOOO419

. 02.19 1.00 TLMOOOO420 ,

02.20 2.00 TLMOOOO421

! 30.00 i

j 03.01 2.00 TLMOOOO352 .

l 1-

TEST CROSS REFERENCE F' AGE 2 OUESTION VALUE REFERENCE 03.02 2.00 TLMOOOO353 03.03 1.50 TLMOOOO354 03.04 2.00 TLMOOOO357 03.05 1.00 TLMOOOO376 03.06 1.00 TLMOOOO387 03.07 1.00 TLMOOOO390 03.08 2.00 TLMOOOO391 03.09 2.00 TLMOOOO399 .

03,10 3.00 TLMOOOO400 -

03.11 3.00 TLMOOOO401 03.12 1.00 TLMOOv0402 03.13 3.00 TLMOOOO403 03.14 1.50 TLMOOOO422 03.15 1.50 TLMOOOO423 03.16 1.50 TLMOOOO424 03.17 2.00 1LMOOOv425 31.00 04.01 2.50 TLMOOOO355 04.02 2.00 TLMOOv0388 04.03 1.00 TLMOOOO404 04.04 1.00 TLMOOOO405 04.05 1.00 TLMOOOO406 04.06 2.50 TLMOOOO407 04.07 3.00 TLMOOOO408 04.00 1.50 TLMOOOO409 04.09 2.50 TLMOOOO410 04.10 1.00 TLMOOOO411 04.11 3.00 TLMOOOO412 04.12 3.00 TLMOOOO413 04.13 2.00 TLMOOOO414 04.14 2.00 TLMOOOO426 04.15 1.00 TLMOOOO427 04.16 1.00 TLMOOOO428 30.00 120.00

ML20211M947

Text

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