Text

Exam Rept 50-302/OL-88-02 on 880112-14.Exam Results:Three Senior Reactor Operators & Six Reactor Officers Passed Written & Operating Exams
	ML20196K407
Person / Time
Site:	Crystal River
Issue date:	02/18/1988
From:	Brockman K, Bill Dean; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML20150A985	List: ML20196K385; ML20196K407; RA88-001, Forwards Comments on 880112 Senior Reactor & Reactor Operator Exams.Question 1.12 Will Be Deleted from Exam & Answer Key for Questions 1.17 & 6.03 Will Be Revised;
References
	50-302-OL-88-02, 50-302-OL-88-2, NUDOCS 8803150304
	Download: ML20196K407 (312)
	v • d • e

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~l ENCLOSURE 1 EXAMINATION REPORT 302/0L-88-02 Facility Licensee: Florida Power Corporation Facility Name: Crystal River Nuclear Plant Unit 3 Facility Docket No.: 50-302 Written examinations and operating tests were administered at Crystal River Unit 3 near Crystal River, Florida.

Chief Examiner: [ d/g- *y ,)// 2 /g Wilrihm M. Dearf ' Date Signed Approved by: [ "- 8 /66' Ken)dth E. Bro'cTanan, Chief Operator Licensing Section 2 Date Signed Summary:

Examinations on January 12-14, 1988.

Written and operating tests were administered to three Senior Reactor Operator (SRO) and six Reactor Operator (RO) candidates. Three of three SR0s and six of six R0s passed the written and operating exam.

Based on the results described above, six R0's passed and three SR0's passed overall.

Seven of the 16 (44%) changes made to the answer key were a result of inadequate or incomplete reference material being provided to the NRC for exam ger.eration.

For several of these changes, time did not permit correction of the information provided the NRC, since the modifications in question were made during the recently completed outage period.

8803150304 880301 PDR ADOCK 05000302 V PDR

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REPORT DETAILS

1. ~ Facility Employees Contacted:

Paul .McKee, Director Plant Operations

Bruce Hickle, Manager, Nuclear Plant Operations

Bill. Marshall, Acting Operations Superintendent

Larry Kelly, Manager, Nuclear Operations Training

Johnnie Smith FPC Training Supervisor

Malcolm Holmes, Instructor

James Owen Instructor

'* David Harper, Licensing Assistant

Attended Exit Meeting

2. Examiners:

William Dean Charles Casto Robert Picker, INEL (JeffTedrow,ResidentInspector,attendedexit)

Chief Examiner

3. Examination ~ Review Meeting At the conclusion of the written examinations, the examiners provided your training staff with a copy of the written examination and answer l key for review. The coments made by the facility reviewers are included l as Enclosure 3 to this report. The NRC resolutions to these coments are listed below.

a. R0 Examination (Applicable SR0 questions are in parentheses)

1. Question 1.12: Coment accepted. Due to lack of specificity in the question, it will be deleted.

2. Question 1.17: Coment noted. The answer key encompasses (5.17) two of the responses referenced by the facility in one answer (fuel cladding degradation). Facility recomended answer is equivalent to answer key and no changes are required.

3. Question 2.02: Coment accepted. Due to a recent (6.03) modification which was not reflected in reference material provided to the NRC, question will be delated.

4. Question 2.05: Coment accepted. Due to the similarity (6.06) between responses "a" and "c", they will both be accepted, t i

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5. Question 2.11: Coment accepted.- Since the question was not specific-as to which MDEFW Pump component was in question, the recomended answer will also be accepted.

6. Question 2.13: Coment noted. Due to a recent modification (6.15) that was not reflected in the reference material provided, the question will be deleted.

7. Question 2.14: Coment not accepted. Knowledge of the basic electrical characteristics of the 120 and 480 VAC systems is considered appropriate.

Note that the weighting for this portion of the answer is minimal.

8. Question 2.16: Coment noted. Based on clarifications made (6.16) to this question during the exam, the candidates are expected to answer as if any ES actuation had occurred affecting that component. The recommended modification regarding the RB fan assemblies will be made to the answer key, but there will be no change to the answer for

- the Sea Water pump as the question asked for the "response of" the component to an ES initiation.

9. Question 2.20: Comment accepted. Since the information (6.22) utilized for question development was from j an outdated lesson plan (that should be l

updated to concur with recent modifications),

the second part of the answer will not be j required for full credit.

10. Question 2.26: Comment not accepted. The question was (6.25) specific enough to relate the IE Notice to the particular application of the d/p

! transmitters in question. No reference material was provided by the facility to show how other IE notice information presented in classes may have confused the candidates, nor how this problem has been proven to be a non-issue at Crystal River.

No change to the answer key.

11. Question 3.02: Coment accepted. Typographical error will (6.01) be corrected as recomended.

12. Qt.estion 3.11: Coment accepted. Due to the possibility that EDGs could be considered one component, credit will not be deducted if a candidate lists an incorrect second component.

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13. Question 3.14:~ Content accepted. The answer key reflects what will happen to the "air-lock" valve for the pneumatic operator, vice the valve itself.

Answer key will be modified as recommended.

14. Question 3.15: Comment noted. Recommended answer is equiva-(6.12) lent to answer key. No change required.

15. Question 3.17: Comment accepted. Based on additional (6.13) information provided that was not reflected in the reference material-utilized, the recommended answer will also be accepted.

16. Question 3.18: Comment accepted. The additional recommended answer will also be accepted.

17. Question 3.19: Comment accepted. Typographical error in the (6.17) answer key will be corrected as recommended.

18. Question 3.22: Content noted. The answer key still reflects the desired information. If interim responses are given, credit will be deducted only if the information provided is incorrect.

19. Question 3.25: Comment not accepted. Candidates should have a basic understanding of logic diagrams and symbology. The information desired of the candidate requires that he understand how the interlock works and be able to translate that to a simple logic diagram. Credit will be given for a verbal explanation of how the interlock works.

20. Question 4.04: Comment accepted. Due to the inability of (7.03) the question to elicit the desired response, it will be deleted.

21. Question 4.12: Comment accepted. Typographical error will be corrected as recommended.

22. Question 4.16: Comment accepted. The answer key will be (7.14) modified to coincide with the way the question was phrased,

b. SR0 Examination

1. Question 5.13: Comment not accepted. The question clearly states that the size of the leak is small.

In such an instance the cooldown will not be nearly as dramatic as the situations presented in the Technical Bases Document.

Also, the answer listed in the key is emphasized in the facility trainino material.

No change to answer key.

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2. Question 5.16: Comment accepted. The additional recommendea answer will also be accepted.

3. . Question 7.10: Comment noted. No Remedial Actions were contained in the answer key. Question will be modified to improve its quality.

4. Qcastion 7.16: Comment accepted. The information provided was not available to the examiners prior to the exam. The recommeded answers will be added to the answer key as required information.

5. Question 7.19: Comment accepted. Due to additional information provided, the recommended answer will also be accepted.

6. Question 8.06: Comment noted. No change to answer key is required. Typographical error in question will be corrected.

7. Question 8.15: Comment accepted. Based on a more recent revision of material than was originally supplied to the examiners, the answer key will be changed as recommended.

8. Question 8.19: Comment accepted. Typographical error in the answer key will be corrected.

9. Question 8.23: Comment accepted. Based on information in OP-412B that is not reflected in the' referenced lesson plan..the answer key will be expanded to include requirements contained in this OP.

10. Question 8.24: Comment accepted. As the appropriate curve was not provided, the question will be deleted.

4. Exit Meeting At the conclusion of the site visit the examiners met with representatives of the plant staff to discuss the results of the examination.

There were some generic weaknesses noted during the oral examination. The areas of below normal performance were in the candidate's knowledge of the Non-Nuclear Instrumentation System and R0 level understanding of Technical Specifications. It was also noted that since the plant had just emerged from a refueling cutage, the candidates were not familiar with many of the recent modifications, of which restructuring of annunciator panels was most notable. The facility gave assurances that all the candidates would receive training on all the modifications prior to assuming any licensed duties.

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5 The. examiners also expressed their continued concern with the structure of the emergency operating procedures. It is possible to have the operators performing as many as four different p:ocedures simultaneously just for a relatively uncomplicated plant casualty. Until the facility has a simulator with which the operators can be examined in a real-time manner utilizing these procedures, these doubts will still exist.

The cooperation given to the examiners and the effort to ensure an atmosphere in the control room conducive to oral examinations was also noted and appreciated.

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

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. 0 U. S. NUCLEAR REGULATORY COMMISSION -

SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: CRYSTAL RIVER

REACTOR TYPE: PWR-B&Wi77 DATE ADMINISTERED: 88/01/12 EXAMINER: DEAN. WM CANDIDATE:

INSTRUCTIONS TO CANDIDATE:

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

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

% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 2%6r /

J 30.00 _1 W 5. THEORY OF NUCLEAR POWER PLANT HERMOD MC 40.00- 25.00- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION atq.o 2 r. 2 00.00 =25.00 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL g 7f.T 2 </ p do.ea 2Lve0 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS p

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

Candidate's Signature I

NBC 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 andfcould result in more severe penalties.

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

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

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

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

6. Use only the paper provided for answers.

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

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

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

10. Skip at least three lines between each answer.

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

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

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

14. Show all calculations, methods, or aesumptiens 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, bd

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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 includis,T figures which are part of the answer.

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

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

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

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 2 THERMODYNAMICS QUEST'ID,N 5.01 (1.00)

Which one of the following is NOT a potential adverse condition that could result from successfully establishing "HPI/PORV Cooling"?

a) Exceeding NDT/ Brittle Fracture Limits, b) Increasing the potential for an uncontrollable LOCA via a failed Code Safety, c) Creating voids in the reactor vescel upper head and "candy cane" regions.

d) Developing unreliable instrumentation indications inside containment.

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. 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE S THERMODYNAMICS QUESIION 5.02 (1.00)

Which'one of the following represents the maximum linear power density which would be expected in the core during full power operations?

a. Local Power Density multiplied by Nuclear Heat Flux Hot Channel Factor.

b. Radial Peaking Factor multiplied by the Local Peaking Factor.

c. Average Kw/ft for the core multiplied by the Nuclear Heat Flux Hot Channel Factor.

d. Nuclear Heat Flux Hot Channel Factor multiplied by the Maximum Local Power Density.

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. 5. THEORY OF NUCLEAR POWEB PLANT OPERATION. FLUIDS, AND PAGE 4 THERMODYNAMICS QUESTION, 5.03 (1.00)

The reactor is critical at 10-8 amps early in core life. These conditions were achieved about 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months prior to peak xenon occurring following a reactor trip from 75% power. Assuming that feedwater and steam pressure are controlled automatically and no operator actions nor reactor trip occurs, which one of the following correctly describes the behavior of the reactor over the next several hours?

a) The reactor stays critical due to the effect of the Moderator Temperature Coefficient.

b) The reactor will go suberitical in about one hour, and will remain suberitical.

c) The reactor will go subcritical in about one hour, will return to criticality in another hour and remain at approximately 10-8 amps.

d) The reactor will go suberitical in about one hour, will return to criticality in another hour and then reach the POAH in another hour.

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

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QUEST 4.0N 5.04 (1.00)

Which'one of the following reactivity coefficients would be the first to turn power back down following a steam break accident?

a) Moderator Temperature Coefficient b) Pressure Coefficient c) Doppler Coefficient d) Void Coefficient t

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS,-AND PAGE 6 THERMODYNAMICS QUESTION 5.05 (1.00)

Which one of the following describes where the maximum fuel centerline temperature occurs in a core with a symmetric axial neutron flux about the core midplane?

a) Top of the core b) Between the top and midplane of the core c) Core midplane d) Between midplane and bottom of the core e) Bottom of the core

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 7 THERMODYNAMICS QUESTION 5.06 (1.00)

Referring to the attached page showing temperature vs distance from the fuel centerline, which curve correctly represents this relationship for a typical coolant channel while at power?

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.N (1.00)

Which one of the following statements correctly describes the effect of adding Emergency Feedwater (EFW) during a Natural Circulation condition?

a) It LOWERS the OTSG thermal center while INCREASING the strength of the heat sink.

b) It LOWERS the OTSG thermal center while DECREASING the strength of the heat sink.

c) It RAISES the OTSG thermal center while INCREASING the strength of the heat sink.

d) It RAISES the OTSG thermal center while DECREASING the strength of

the heat sink.

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5. THEORY OF NUCSEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 9 THERMCDYNAMICS QUESTI.ON 5.08 (1.00)

Which'one of the curves on the attached page correctly shows the affect of recently installed "GREY" APSRs on the axial imbalance compared to the affect the old "BLACK" APSRs had on axial imbalance?

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 10 THERMODYNAMICE QUESTIpH 5.09 (1.50) t.

With'4he Unit operating at 100% power with all control systems in automatic, a Turbine Bypass Valve fails full open. Indicate how the following parameters will change relative to their initial values when plant conditions stabilize: (INCREASE, DECREASE, REMAIN THE SAME) a) Tavg b) MWe c) Reactor power -

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5. ' THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 11 THERMODYNAMICS QUESTION 5.10 (1.50)

For e~dch of the following, indicate whether they will cause the power range instrument to indicate HIGHER, LOWER or the SAME as actual power, if the instrument was adjusted to 100% based on a calculated heat balance:

a) The feedwater temperature used in the heat balance was HIGHER than actual feedwater.

b) If the reactor coolant pump heat input used in the heat balance is OMITTED.

c) If the steam flow uced in the heat balance was HIGHER than actual.

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'5. ' THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 12 l THERMODYNAMICS !

l QUESTION 5.11 (2.00)

Indicate whether each of the following will cause the differential rod worth to INCREAGE, DECREASE or have NO EFFECT.

a) An adjacent rod is withdrawn.

b) Moderator temperature is DECREASED.

c) Boron concentration is INCREASED.

d) A Burnable Poison Rod depletes.

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QUESTION 5.12 (1.50) i

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Assuming that OTSG pressure is stable, what are three indications that the i operator can utilize to confirm the existence of natural circulation? Give numerical values where appropriate.

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'3 "qQUESTI.pN 5.13 ;

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What' parameter will be the greatest aid to an operas,r in helping x.

%:to determine whether a small steam leak or a small LOCA is present k

'inside containment?

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5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS AND PAGE 16 THERMODYNAMICS QUESTIpH 5.15 (1.00)

>i What characteristic shape should an inverse multiplication plot (1/M plot)

-have during fuel loading when the fuel is loaded so that the distance between the detector and the fuel steadily decreases? A sketch is sufficient to answer the question.

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5. THEOPJ OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AE2 PAGE 17 j

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Q U E S T @ ,N 5.16 (1.00)

There are two effects that cause differential boron worth to change over core life. List these two effects and indicate their relative impact on differential boron worth, t

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS. AND PAGE 18 TEEBt10 DYNAMICS QUESTION. 5.17 (1.50)

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List 6 potential sources of gas intrusion into the RCS if a significant

-LOCA were to occur. '

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. 5. THEORY OF NUCLEAR POWER P_LANT OPERATION, FLUIDS. AND PAGE 19 THERMODYNAMICS QUESTI N 5.18 (2.00)

How does each of the following parameter changes affect the DNBR (INCREASE, DECREASE or NO EFFECT)? Briefly explain your ansder and DO NOT consider the transient effects or any control system or operator actions.

a) Pressurizer temperature increases 5 degrees.

b) Maec flow rate in the core increases 10%.

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. 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 20 THERMODYNAMICS 1

QUESTI,0,N 5.19 (1.50)

If the OTSG low level limit were decreased, indicate-whether reactor power would have to be HIGHER, LOWER or the SAME to achieve a TAVG of 579 degrees. Explain your answer.

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5; THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS AND PAGE 21 THERMODYNAMICS QUEST. ION 5.20 (2.00) i While' operating at 60% power, it is recognized that a control rod has been

! oignificantly misaligned BELOW its group average for several days. The rod l is realigned, and a positive quadrant tilt develops in the quadrant of the aisaligned rod.

l l a) Why does a positive tilt exist, even though all rods are correctly aligned?

l l b) Assuming no operator action, how would this tilt change over the next ten hours? Explain your answer.

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS. AND PAGE 22 THERMODYNAMICS 1 QUESTION 5.21 (2.00)

Refer to the attached figure 11-27, "Error Adjusted Rod Index Alarm Setpoint" to answer the following:

a) Assume that you are at the indicated point outside of the permissible area of operations for "power level vs. rod index" and "axial power imbalance". List the three methods (indicated by Paths A, B and C on both curves) which are utilized to restore plant conditions within the region of permissible operations. (1.5) b) Path C for "axial power imbalance" was not successful in vacating the restricted region. What action is taken as indicated by Path D, which will restore axial imbalance to the permissible operating region? (0.5) i e

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FIGURE 11-27: ERROR-ADJUSTED ROD INDEX ALARt1 SETPOINTS

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Power-Imbalance Envelope for Operation Error-Ad}usted RodIndex Alerm Setpoints Power Level Power (%)

(%)

100 -

Shutdown

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Restricted C Restricted

-40 Region Region 40 -

-30 Permissible

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--10 Rodindex (%)

0 50 Ido 150 2$0 250 300 40-30 13 0 10 20 30 40 50 [

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. 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 23 THERMODYNAMICS l

1 l %

QUEST.IgN 5.22 (1.00)

What are the two reasons that there is a 15 minute time period that must be observed between RCP "bumps" (until Natural Circulation redevelops) when using RCP "bumps" to try and restore Natural Circulation, assuming it was lost during a LOCA condition?

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

, 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 24 THERMODYNAMICS QUESTI,0N 5.23 (1.50)

Attached is a curve showing Keff vs. moderator-to-fuel ratio for 500 ppm Boron concentration. For the following, indicate where on the curve the applicable location would be located: (e.g. above and to the left) a) Overmoderated region b) Negative HTC region c) Direction optimum point shifts as Boron concentration decreases

(***** END OF CATEG0FY 05 *****)

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. 6. PLANT SYSTEMS DESIGN, CONTROL. AND INSTRUMENTATION PAGE 25 QUESTION 6.01 (1.00)

Which'ione of the statements below correctly describes the operation of the Load Control Valve (LLCV) and the Main Feedwater Block Valve (MBV) during a power decrease from 100% to 15%, assuming both valves are in Automatic?

a) The MBV starts to close as Leop FW Demand drops below 80% and the LLCV starts to close when the MBV reaches the fully closed position.

b) The MBV starts to close as Loop FW Demand drops below 50% and the LLCV starts to close when the MBV reaches the 80% open position.

c) The MBV starts to close as Loop FW Demand drops below 45% and the LLCV starts to close when the MBV reaches the 50% open position.

d) The MBV starts to close as Loop FW Demand drops below 45% and the LLCV starts to close when the MBV reaches the fully closed position.

I

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

6.' PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 26

.g QUESTION 6.02 (1.00)

Which'etatement below correctly describes how the ICS determines the RCS flow s,ignal for determining whether a runback is required?

a) ICS calculates a flow signal based on RCP breaker status, b) ICS calculates a flow signal based on input from RCS loop flow D/P transmitters.

c) ICS calculates a flow signal based on input from RCS loop flow tubes.

t d) ICS calculates a flow signal based on a selectable input using a patch cord from the RPS cabinet.

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(***** CATEGORY 06 CONTINUED ON NEXT PAGE 4****)

6. PLANT SYSTEMS DESIGN. CONTROL AND INSTRUMENTATION PAGE 27 QUESTION 6.03 (1.00)

Whil~e' . performing -354-B (B-EGDG surveillance) a loss of off-site power occ rs along with an exciter field overcurrent. Whic one of the following correctly describes the response of the Diesel Generator engine and output breaker?

a) The output b eaker will trip open and lockout, the engine shuts own, b) The output bre her will remain closed, the engine continues to o erate.

c) The output brea r will trip open, the engine shuts down and then re tarts after the overcurrent condition clears. The output breaker will then Close.

d) The output breaker ill trip open, the engine will continue to run, th output breaker will close after the overcurren conditions clears.

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

6. PLANT SYSTEMS DESIGN. CONTROL, AND INSTRUMENTATION PAGE 28 QUESTION 6.04 (1.00)

Whic *one of the following sets of conditions would directly cause the EFIC system to initiate Emergency Feedwater?

(assume all controls are in auto) a) A trip of one Main Feedwater pump at 90% power.

b) A low level (<6") in "A" OTSG with HPI actuation on both channels, c) While at power an operator depresses the "Trip 1" button at the local panel.

d) An operator has placed the SG BYPASS / RESET switches for Channels A - D in the BYPASS (up) position with main steam pressure at 700 psig, and a OTSG low pressure (<600 psig) signal is subsequently received.

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

[

6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 29 QUESTION 6.05 (1.00)

Which'ione of the following indications and/or automatic cetions of the NNI X system power supply monitor would indicate a loss of one +24 VDC and one -24 VDC power supply? [

a) Upper light in upper section would be out...no automatic actions, b) Both lights in upper section would be out...no automatic actions, c) Upper light in both sections will be out...after a time delay (.5 secs) the Si and S2 breakers will open.

d) Both lights in upper section would be out...after a time delay (.5 secs) the Si and S2 breakers will open.

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

6. PLANT SYSTEMS DESIGN. CONTROL, AND INSTRUMENTATION PAGE 30

'i QUESTION 6.06 (1.00)

Whic5/one of the following is the most important reason why core flood tank pressure is carefully controlled?

a) During a large LOCA, CFT injection will occur immediately after "blow-out" of the water from the lower part of the core occurs, b) Pressure must be maintained to promote adequate "Nitrogen Sparging" (mixing) during CFT depres-surization, c) Pressure must be low enough to minimize CFT injection until immediately after "blow-out" of the water from the lower part of the core occurs.

d) During CFT injection the decrease in CFT liquid j volume will cause nitrogen injection into the core.

i

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

6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTPUMENTATION PAGE 31 QUESTION 6.07 (1.00)

Whickone of the following is correct concerning the limits and precautions of the CA system?

a) Of system components that contain a boric acid solution of 5% wt. or greater, only tanks need to be maintained at greater than (105 deg F).

b) While draining the BAMT ensure the heaters automatically de-energize at the appropriate tank level, c) Hydrazine addition to the RCS during operation of makeup demineralizers prevents releaae of chlorides, d) Sufficient water shall be available to dilute a spill to a (100:1) ratio when handling NaOH, LiOH, or hydrazine.

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

6. PLANT SYSTEMS DESIGN, CONTROL. AND INSTRUMENTATION PAGE 32 QUESTION 6.08 '

.c.)

$ Indikstewhetherthe <ng statement is TRUE or FALSE:

If the HPI Safegua .uation d.s bypassed prior to pressure falling to 1500 psig, then a suvo,quent LPI actuation takes place, the HPI system would actuate.

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

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'6. ' PLANT SYSTEMS DESIGN. CONTROL AND INSTRUMENTATION PAGE 33 QUESTION 6.09 (1.50)

For e OTSG Level Instruments listed below, indicate the number of channels available per OTSG, and indicate whether the instrument is temperature compensated or not.

c) Operating Range b) EFIC High Range c) Full Range 1

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

\ _ _ _ _ ___

6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 34 QUESTION 6.10 (1.00)

For th'e operating conditions listed below, ' indicate what error signal would '

be the controlling input into the Megawatt Calibrating Integral. Assume that all other controls are in Automatic, with the unit at power.

a) SGRX in Manual b)

Turbine not in ICS Auto 4

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

6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 3S QUESTION 6.11 (1.00) ;

J Indichte for each of the following situations whether the indicated breaker can be closed or if cross-tie blocking will prevent closure. Refer to the i attached drawing showing the 4160ES Distribution.

a) Breakers 3207, 3208 and 3209 are closed. An attempt to close 3210 is i made.

b) Breakers 3205, 3208 and 3210 are closed. An attempt to close 3209 is made.

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

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6. PLANT SYSTEMS DESIGN CONTROL, AND INSTHUMENTATION PAGE 36

'. F i

QUESTION 6.12 (2.00) I LisE#the four methods by which the Low Pressure CARDOX system to the Feedwater Pump Area may be actuated. Provide locat'.onc where applicable. t I

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(***** CATEGOBY 06 C0!iTINUED Oli NEXT PAGE *****)

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6." PLANT SYSTEMS DESIGN, CONTROL. AND INSTRUMENTATION PAGE 37 QUESTION 6.13 (1.50)

List-,t.he 2 valves monitored by the Decay Heat Removal System's Automatic Closure Initiation (ACI), the 2 automatic actions that occur if the actuation setpoint is reached (assuming these valves are open) and how this interlock may be bypassed.

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

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, 6. ' PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUt1ENTATION PAGE 38-

\

I I QUESTION 6.14 (1.00)

The'iq11owing actions associated with the EHC and OPC systems occur:

The test solenoids on the reheat intercept valves are Cnergized, causing the intercept valves to'c},$ae. After a time delay, the valves reopen and then recloevi (repeating this process several times).

State the most likely cause of these events.

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

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' \._\ gb(j,f ' h , PLANT SYSTEMS DESIGN. CONTRob. AhD INSJRUMENTATIONPAGE 39 V

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/jt QUESTION 6.15 (1.00)

The 2perator ha placed the Diamond Control Station in automatic.in ace rdance with the operating procedure. A control rod drive programmer senses rod motion without a corresponding comm si nignal.< State the response of the Diamond Control Std !.on, with respect to its ability to respond to command c gnals, as a rer, ult of this condition.

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

6. ' PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 40 QUEST,IOFI 6.16 (2.00)

Conce[ningtheDCsystem, state the response of the following equipment to an automatic initiation signal from ES:

1. RB fan assemblies.

2. Nuclear Services Emergency Sea Water Pumps.

3. Closed Cycle Cooling Pumps.

4. Normal Sea Water Pump.

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

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 41 QUESTION 6.17 (1.50)

WhiiN at 75% power, with all control systems in AUTO, the SGFX Master is placed in HAND. Describe the effect on the ICS from increasing its output to 80% Discuss only the systems affected by the control manipulation and how the ICS compensates for the increased demand.

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

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 42 j l

l QUESTION 6.18 (1.00)

An I'&'C technician wants to remove the High Pressure Trip module from RPS Channel "A". He states that by placing the channel in bypass, the other channels will be prevented from receiving a trip signal from the "A" Channel. Explain why you AGREE or DISAGREE with the I & C tech.

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

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6 '. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 43 l

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QUESTION 6.19 (1.00)

. 's 1 Why are the DH System Drop Line Valves DHV-3, 4, and 41 Opened with their

breakers in the "Lock Reset" position when the reactor vet sel head is removed?

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

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 44 QUEST, ION 6.20 (1.50)

Whi1Eat 100% power, there is a failure HIGH of the lower detector of the Power Range NI feeding the A Channel of the RPS. Describe the effect of this failure on the PPS In your answer, indicate what inputs to the RPS are affected and the final state of the RPS.

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

6. PLANT SYSTEMS DESIGN CONTROL. AND INSTRUMENTATION PAGE 45 QUESTION 6.21 (2.00)

ForibneEmergencyFeedwaterControlValvesEFV-55,EFV-56, EFV-57, and EFV-58 provide the following information:

1. EXPLAIN the motive force used to operate the valves.

(0.5)

2. What system provides the controlling signal (s) for the valves? (0.5)

3. What is the basis for the "fill rate" limiting signal for valve operation? (1.0) l l

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

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 46

\

QUESTION 6.22 -( 1. 00 )- [06)

With't'he Makeup System in service, a loss of either ICS or NNI power occurs. The operator attempts to close MUV-49 Letdown Isolation Valve and MUV-40 Letdown Cooler Outlet Valve. Is it possible to close these valves with the loss of ICS or NNI power.7-AND ic thic.:n ppropriate evtion to--

t-ahe given + hee: conditiens?

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

6. PLANT SYSTEMS DESIGN. CONTROL.-AND INSTRUMENTATION PAGE 47

~

QUESTION 6.23 (1.00)

'i Assume' a loss of-offsite power occurs coincident with a HPI

. actuation. Explain the feature of the HPI pump start celector switches which will prevent two HPI pumps from being loaded onto one diesel generator.

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

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6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 48 e

QUESTI,0N 6.24 (1.00)

Assumd the following conditions exist:

"A" channel RCPPM .- "C" and "D" RCPs MW transducers have failed The reactor is at power, no other abnormal conditions exist.

Explain why this condition would cause an automatic reactor trip.

(***** CATEGORY 06 C0t1 tit 10ED Ott tiEXT PAGE *****)

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6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 49 QUESTI,0N 6.25 (1.50)

NRC'51f Information Notice 85-100 concerns the application of Rosemount dp transmitters used for measurement of RCS flow.

EXPLAIN the abnormality that may occur, including the conditions that may precipitate this, and the effect on the indication of flow.

I

(***** END OF CATEGORY 06 *****) ,

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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND l PAGE 50 i RADIOLOGICAL CONTROL !

i QUESTION 7.01 (1.00)

Which one of the following is NOT a RCS Leakage Detection System required by Technical Specifications?

a. Containment atmospheric Iodine radioactivity monitoring

b. Nuclear Services Closed Cooling Water monitoring

c. Containment atmospheric gaseous radioactivity monitoring

d. Containment sump level monitoring

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

7. PROCEDURES - NORMAL, ABNORMAL. EMERGENCY AND PAGE 51 RADIOLOGICAL CONTROL QUESTJON 7.02 (1.00)

Consider a situation where a Large LOCA has occurred, and you are operating the plant under AP-380, "Safeguards Actuation" and EP-290, "Inadequate Core Cooling". While performing a needed step in EP-290, you find you are unable to complete the action as called for in either the "ACTIONS" or the "DETAILS" column of the procedure. Which statement below correctly describes the appropriate action the operator should take, assuming additional methods NOT LISTED in EP-290 could be utilized to complete the required action?

a. The operator at the controls should take the appropriate mitigating actions, and then inform the NSS or ANSS.

b. The operator at the controls should await direction from the NSS or ANSS, and if none is forthcoming, continue with the procedure.

c. The operator at the controls should inform the NSS or ANSS of his alternative action for concurrence, prior to performance,

d. The operator should continue attempts to take the action as stated-in the EP, while continuing on with the procedure.

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7. PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE 52 RADIOLOGICAL CONTROL QUESTION 7.03 (1.00)

Assuming a unisol ble leak in the RCS, with saturated conditions existing and the RCPs secur d,.which one of the following statements correctly describes how chang s in Excore NI behavior can provide an indication of the status of the co e and downcomer water levels?

a. SR counts wil SLOWLY DECREASE as the core begins to uncover, assuming downc mer level remains relatively stable,

b. SR cour.ts will PIDLY DECREASE as the core uncovers, assuming downcomer level emains relatively stable.

c. SR counts will SLO LY INCREASE as the downcomer region refille, assuming the level n the core remains relatively stable.

d. SR counts will SLOWL INCREASE as the downcomer region depletes, assuming the level in the core remains relatively stable.

e. SR counts will RAPIDLY ECREASE as the downcomer region refills, assuming the level in t core remains relatively stable.

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

7. PROCEDURES - NORMAL ABNORMAL, EMERGENCY AND PAGE 53 RADIOLOGICAL CONTROL QUESTI,0,N 7.04 (1.00)

In accordance with OP-204, "Power Operations", which one of the following is considered the best method for dampening a Xenon Oscillation?

a. Determine the period of the oscillation ar. soon as possible, and then utilize t!.3 APSRs and Boration/ dilution as necessary several hours BEFORE the peak deviation to achieve an average axial power imbalance,

b. Determine the period of the oscillation within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , then use boration/ dilution, as appropriate, several hours AFTER the peak deviation to achieve an average axial power imbalance.

c. Determine the period of the oscillation over several days, and once peak deviation occurs so that a POSITIVE axial imbalance exists, drive control rods inward to reduce power 10-15%.

d. Determine the period of the oscillation over several days, then make the appropriate rod position correction several hours BEFORE the peak deviation to achieve an average axial power imbalance.

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

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 54 RADIOLOGICAL CONTROL QUESTI.ON 7.05 (1.50)

For each of the following, indicate whether that condition alone would be sufficient grounds to initiate a reactor trip, as listed in the entry conditions of AP-580, "Reactor Trip":

a) 1 MSIV on each loop (MSV-411 and MSV-414) have been inadvertently shut.

b) Both Main Feed Pumps are lost with reactor power at 15%.

c) A Turbine Trip from 40% power occurs.

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

. 7. PEOCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 55 RADIOLOGICAL CONTROL QUESTIO.N 7.06 (1.00)

With respect to removal of the equipment hatch to the reactor building, match the responsibility in column A witn the appropriate individual in column B.

COLUMN A COLUMN B a) Determines need for hatch removal 1) SSOD

2) Nuclear Ops Superintendent b) Approves the hatch removal 3) Health Physics Supervisor

4) Chem / Rad Protection c) Concurrence required prior to removal Superintendent

5) Outage Shift Manager

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

I

7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 56 RADIOLOGICAL CONTROL QUESTION 7.07 (1.00)

Fill ~in the blanks:

a) Thermoluminescent dosimeters should be re-zerced prior to reaching a maximum of % full scale.

b) The background reading on a fricker used for whole body monitoring should be no more than CPM.

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(***** CATEGORY 07 CONTItiUED ON 11 EXT PAGE *****)

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7. PROCEDURES - NORMAL ABNORMAL, EMERGENCY AND PAGE 57 RADIOLOGICAL CONTROL a

QUESTy0N ,

7.08 (1.00) a) When performing an RCP "bump" in an effort to restore Natural Circulatii;n, why is OTSG Pressure reduced so that OTSG Tsat is 40-60 degrees below Incore Temperatures, prior to starting an RCP?

b) What two methods / indications can be utilized by the operator to determine if RCS pressure is high enough to allow a RCP "bump"?

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

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 58 RADIOLOGICAL CONTROL QUESTION 7.09 (1.50)

While performing a shutdown in order to conduct a refueling, is it permissible to begin detensioning the reactor vessel head bolts without having audible counts of the source range in the control room, as long as this audible Explain your indication answer. is actuated before removal of the head begins?

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

7. PROCEDURES - NORMAL, ABNORMAL. EMERGENCY AND PAGE 59 RADIOLOGICAL CONTROL

i ;

QUESTJON 7.10 (1.50)

,7 List the Immediate Actions for AP-460, "Steam Generator Isolation Actuation". The remedial actions are NOT required.

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

7. PROCEDURES - NORMAL ABNORMAL, EMERGENCY AND PAGE 60 RADIOLOGICAL CONTROL i,

QUESTI.ON r ,.

7.11 (1.00)

What"two individuals, by title, can authorize an Inplant Clearance?-

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7. PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE 61 RADIOLOGICAL CONTROL

^;;

QUESTION 7.12 (1.50)

'f:/

In accordance with AP-580, "Reactor Trip", what are the 3 actions required to initiate emergency boration?

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'7.

?ROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 62 '

RADIOLOGICAL CONTROL QUEST 30N 7.13 (1.00)

List the four Critical Safety Functions in the order of priority, as monitored in Verification Procedure VP-580. ,

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7. ' PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 63

. RADIOLOGICAL CONTROL -

QUESTION 7.14 (1.00)

fi What'three conditions must be met in order to secure Nuclear Service Water supply to the CRDs?

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7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 64 RADIOLOGICAL CONTROL QUESTI,0N 7.15 (1.75) a) Assuming reactor power is 35%, what are the 3 parameters, which if exceeded require entry into AP-660, "Turbine Trip"? (.75) b) While performing the IMMEDIATE ACTIONS of AP-660, the Governor Valves fail to close and the generator output breaker fails to open. What are the appropriate REMEDIAL ACTIONS? (1.0) l l

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

7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 65 RADIOLOGICAL CONTROL

/

QUESTI,,0,H 7.16 (1.00)

A continuing problem with the EFIC system is LED failures causing 1/2 trip signals. If such a condition were to occur, necessitating maintenance on the EFIC system while at power, what system manipulation is done before work is commenced, and what administrative action is required while maintenance is ongoing?

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

- . -m -,

,- 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 66 RADIOLOGICAL CONTROL e,

QUESTIO,N 7.17 (1.75)

In accordance with AP-380, "Engineered Safeguards Actuation", what are the HPI throttling criteria?

(***** CATEGOBY 07 CO!1 tit 40ED Ott t1 EXT PAGE *****)

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7. PROCEDURES - NORMAL ABNORMAL EMERGENGL AND PAGE 67 RADIOLOGICAL CONTROL '

.1 s

RUESTJ,0,N 7.18 (1.50)

Indicate what the status of the following components / systems should be in order to perform OP-208, "Plant Shutdown" (20% to 0%):

a) OTSG Level b) Feedwater System Pump Status c) RCS Pressure and Temperature t

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7. FROCEDURES - NORMfL. AB!!ORMAL. EMERGENCY AND PAGE 68 RADIOLOG1 CAL CONTdQL QUESTION 7.19 (1.50)

Assume that the Main Condenser is NOT available. What are two bases for allowing an Emergency Cooldown Rate of 240 degrees / hour to a Thot of 500 degrees while performing EP-390, "Steam Generator Tube Rupture"?

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7. PROCEDURES - NORMAL ABNORMAL. EMERGENCY AND PAGE 69 RADIOLOGICAL CONTROL s

QUESTION 7.20 (1.50)

Answer the following questions regarding fuel handling operations:

a) While handling new fuel assemblies utilizing the overhead crane, why must the 1.5 ton auxiliary hook be utilized when loading into R0W 1 of the new fuel ctorage racks, vice the main hock?

b) What indication can be utiitzed to verify that the Fuel Transfer Carriage is located ir its ncrmal storage area against the "hardstops" in the "A" Fuel Storage Pool?

c) After a fuel assembly has been lifted, as indicated by the "Grapple Tube Up" light, what must be check to verify grapple position?

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7. PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE 70 RADIOLOGICAL CONTROL

?

QUESTgN 7.21 (1.00)

According to the ATOG Guidelines, what is the major concern of operating the-HPI system excessively during a Steam Generator Tube Rupture?

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7. PROCEDURES - NORMAL ABNORMAL EMERGENCY AND PAGE 71 RADIOLOGICAL CONTROL QUESTLON >

7.22 (2.00) l a) What is the basis for establishing a OTSG level of 65% while performing AP-530, "Natural Circulation"?

b) What is the basis behind raising OTSG level to 95% if subcooling margin degenerates while performing AP-5307 1

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

7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 72 RADIOLOGICAL CONTROL QUESTJON 7.23 (1.00)

While performing OP-209, "Plant Cooldown", step 6.2.13 (see attached) has the operator refer to TS 3.7.1.2, EFW Syoten. What is the reason for referring to this TS7 l

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. Initials

6.2.12 When secondary steam pressure has decreased to about

'; ; 250 psig, stop MTVP per OP-605, Feedwater system. Use the TV booster pump for TV supply. _,

NOTS: If coming to refueling mode, perform SP-418, -

Main Teodwater Pump Trip Test.

., ,6.2.13 When secondary steam pressure is ( 200 psig, enter STS action of 3.7.1.2. lc 6.2.14 Between 900 and 600 psig RCS pressure, bypass LPI ES channels ' A' and 'S' RC-4, RC-5, and RC-6, 6.2.14.1 If the plant is expected to be in COLD SHUTDOWN for 72 firs. or more, contact the Inservice Inspection (ISI)

Specialist to ascertain whether SP-405, Core Flooding

Systen Check Valve Operat.'on Demonstration and Leak i

Testing, Part '8*, and SP-603, Decay Heat Check Valve Leak Testing, need to be performed. -

6.2.14.2 If coming to refueling mode, perforn $P-603, Decay l Heat Check Valve Leak Testing; SP-402, Core Flooding

Systen Isolation Valves Alaras Actuation; and SP-405, l

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CP-209 Rev.60 Pave 12

7. PROCEDURES - NORMAL. ABNORMAL EMERGENCY AND PAGE 73 !

RADIOLOGICAL CONTROL QUEST.IO,N 7.24 (1.00)

What are the two reasons for establishing an OTSG level of 97-99% operate range, when performing a plant heatup? (See attached precaution from OP-202) i l

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

4.1.10 Prior to placing a sakeup desinerclizer in service after having '

.used hydrazine for oxygen scavenging, assure that chemistry .

analysis of the RC showed hydrazine concentration as 'less than

.j ,

j detectable *,

4.2 AZACTOP CORE LIMITS AND PRECAUTIONS FOR REATUP Control Rod Safety Groups 1 thru 4 will be fully withdrawn i

during deboration.

4.3 STT_am GENDt1 TOR LIMITS AND PRECADTIONS FOR unTUP 4

4.3.1 The secondary side of the steam generator (s) shall not be pres-surized above 237 psig if the temperature of the steam genera- -

tot vessel shell is belcw 110*T (STS 3.7.2.1).

j 4.3.2 The minimum steam generator operating level is 18 inches (STS 3.4.5).

4.3.3 (deleted) 4.3.4 The naminua steam generator heatup rate is 100*F/hr.

4.3.5 Cycle cleanup shall be complete before feeding once-through steam generator (OTSG).

I 4.3.6 The 0750 sain feedwater nozzles should be flooded by maintain- #

ing level at 97-99% on the operate range until RC systen tes-

. perature is greater that 190'F. The OTSG 1evel should be

lowered to .( 350 in. prior to entering Mode 4.

- - ~ _

)

CP-202 # '

Rev. L 7 5

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 74

/, ,

a;f QUESTION 8.01 ( .50)

Whibioneofthefollowingcombinationsoftagshungona single component is permissible?

a) 2 blue tags b) 1 red and 1 blue c) 3 red tags 4

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 75 I ..

QUESTION 8.02 (1.00)

Whic5 one of the following prevents a fuel assembly from being'raiced above the minimum shield depth?

a) By ebserving the ZZ tape indication and stopping at the required height.

b) A microswitch/ interlock.

c) A hardstop.

d) Refueling water level limits, i

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8.

ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 76 QUESTION 8.03 (1.00)

Which one of the following is limited to ' ensure that tn3 "powdy peaking" limits are maintained?

a) control rod speed.

b) axial power imbalance.

c) RCS pressure, d) RCS flow.

,a

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

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8: ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIOl1 PAGE 77 QUESTION 8.04 (1.00)

Acc9rding to the OSIM, which one of the following must be

, used .for proper tracking of "special valve line-ups"?

a) Placing an entry in the operator's log.

b) Listing on the shift relief check list.

c) Placing an entry in the shift supervisor's log.

d) Making a temporary procedure change.

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8.

ADMINISTRATIVE PROCEDURES,_ CONDITIONS, AND LIMITATIONS PAGE 78 QUESTION 8.05 (1.00)

Which one of the following may proceed, given that a Tech Spec qction statment has been entered requiring you to "suspend all coro alterations"?

a) Removing a neutron source form the core or positioning the auxiliary neutron detector.

b) Using the bridge to remove a fuel assembly from the Core.

c) Shuffling control rods as long as Keff = or < .95.

d) Placing a fuel assembly which is in transient in a secure position, i

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8.* ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 79 4

QUESTION 8.06 (1.00)

Using attached TS 3.2.2, determine which one of the following statements is correct concerning the Quadrant Power Tilt (QPT).

a) If QPT exceeds the maximum limit of 20.0, the reactor must be immediately shutdown.

b) If misalignment of a control rod causes the QPT to exceed the transient limit, thermal power must be reduced within 30 minutes.

c) No action is required within one hour regardless of the QPT limit exceeded (steady state, transient or maximum).

d) If QPT exceeds the steady state limit, but is less than the transient limit, operation may continue indefinitely only up to 60% allowable for the RCP combination.

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

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8.' ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 80 QUESTION 8.07 (1.00)

Unitj3. is operating in Mode 2. After performance of the surveillance test for the RCS High Pressure Trip, it is found that the response time for one channel exceeds 0.44 seconds. Which one of the following actions would be Permissible in accordance with TS 3.3-1 attached?

a) apply action statement 3.a. only.

b) apply action statement 3.b. only.

c) apply action statement 3.0.3.

d) apply action statement 3 and continue power aseencion into Mode 1.

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 81 4

t QUESTION 8.08 (1,00)

Ifad,unexplainedreactivityadditionoccurredduringMODE4 operations, what three (3) individuals (by position) would be notified in accordance with AI-500 Conduct of Operations?

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 82 QUESTION 8.09 (1.00)

StaV the two minimum protective actions that are required when 5 General Emergency is declared.

(***** CATEG0FY 08 C0t1TINUED ON NEXT PAGE *****)

8.'

ADMINISTRATIVE FROCEDURES, CONDITIONS. AND LIMITATIONS PAGE 83 ;

f QUESTION 8.10 (1.50) t What/three conditions aust be met to allow the operator to !

leave l a fuel handling bridge unattended? - I i

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS. AND LIMITATIONS PAGE 84 QUESTION 8.11 (1.00)

PldNt, conditions are as follows: Mode 1, 97% RTP, B-EDG is 00$ for-routine maintenance. Twenty hours into the work on the B-EGDG the control power fuses blov for the A-DHP.

Refer to the attached Tech Spec excerpts and determine which LCO/ Action statement covers this situation.

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

8. ADMINISTRATIVE PROCEDUREE, CONDITIONS, AND LIMITATIONS' 'PAGE 85 4

QUESTION 8.12 (1.50)

List'five (5) areas which require authorization by the Nuclear Fire Protection Specialist to weld, cut, or perform spark-producing work while in Modes 1, 2 or 3.

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

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 86 o

QUESTION 8.13 (1.00)

ThehiantisinMODE3andareactorstartupisinprogress. t The power supply for NI-5 fails. The Chief I&C technician on duty informe you that it will take two days to replace the power supply. Using the attached Tech Spec excerpts, determine which LCO/ Action statement applies, and briefly EXPLAIN what action (s) must be taken. Also, address whether or not the startup may continue.

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i

, 8. ' ADMINISTRATIVE PROCEDURES,' CONDITIONS, AND LIMITATIONS PAGE 87-QUESTION 8.14 (1.00)

As thle SSOL, what are your responsibilities when you recieve

an NCOR7

(***** CATEGORY 08 CONTIFUED 08 NEXT PAGE *****)

8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 88 i.

QUESTION 8.15 (1.00)

Who-[byposition) is required to authorize exceeding 1.25R/QTR7 What admininstrative limit would then be enforced?

i

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i

8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 89-i QUESTION 8.16 ( .50)

The' plant status is as follows: Mode 1, 984 RTP, 2155 pai, i 579 deg. F Tavg, 2 F Przr level. SP-333 Control Rod i Exercises is in progress. It has been determined that l Group-6 Rod-3 will not move. Explain how the Shutdown !

Margin requirement is changed.

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

8. ADMINISTRATIVE PROCEDURES CONDITIONS, AND LIMITATIONS PAGE 90 i

QUESTION 8.17 (1.00)

Sta Nftwo (2) requirements which must be satief. Led to allow additional individual (s) to work under another individual's

, clearance.

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

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..,.'8.

ADMINISTRATIVE PROCEDURES CONDITIONS AND LIMITATIONS PAGE 91 ,

i:

'i QUESTION 8.18- ( .50)

i.

Exp1'ain the relationsh'ip-between an RCA and a Radiation-Area, .

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(***** CATEGORY 08 CONTI60En ON t' EXT PAGE *****) [

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8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 92 i

QUESTION 8.19 (2.00)

During a plant startup, in Mode 3, the Chemistry Technician on duty informs you that CAV-126 (RC Letdown Isolation valve) will not close. Briefly EXPLAIN what LC0(S)/ Action statement (s) is/are affected by this failure AND whether or not startup may continue.

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

O^

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 93 i

QUESTION 8.20 (1.00)

Corr #edtly apply the Emergency Classifications to the following summary statements:

a) An event that has or is causing major failure of systems required to protect the public. It is likely that some exposure to the public may occur. r b) Actual or potential substantial degradation of the '

level of safety of the plant, i

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 94 (

i QUESTION 8.21 (1.50)-

Wha Da're the Federal Quarterly dose limits for a radiation I' worker with AND without a Form-47 Note: include whole body, okin and extremity limits.

b I

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, 8. ADMINISTRATIVE PROCEDURES CONDITIONS. AND LIMITATIONS PAGE 95 '

QUESTION 8.22 (2.00)

Ansde/eachofthefollowing in accordance with the requirements of AI-401 "... Revisions to P0QAM Proceduree": ,

t

1. When can interim changes be used?

2. What changes can be made by an interim change?

3. Compare the type of review that is required for an IC with that required for a permanent revision.

4. How many interim changes can be in effect on any one procedure?

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(***** CATEGORY 08 CONTINUED ON HEXT PAGE *****) ;

8 '. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 96 e

-QUESTION 8.23 (1.00)

Statd four requirements that must be met prior to and during the release of radioactive gases froe the waste gas decay tanks.

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

8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 97 QUESTION 8.24 (1.50)

Whi1I#operatin in Mode 1,la reactor trip occurs. In reviewing the p rameters following the trip, it is determined that he reactor tripped on High Temperature (618 deg F) and at th time of the trip pressure was 2000 psig cnd had decreased to 1750 before recovering. (the variable pressure /temperatu e trip failed to trip the reactor)

STATE all of the re uired actions, in accordance with Tech Specs, relating to t is event.

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8. ADMINISTRATIVE PROCEDURES CONDITIONS. AND LIMITATIONS PAGE 98 QUESTION 8.25 (1.50)

a. Explain the "procedure adherence policy" according to AI-400 P0 GAM. (0.5)

b. Give a brief explanation of three (3) exceptions to the rule that "the written ' Controlled */' Working Copy

procedure shall be present and followed step-by-step" as outlined in AI-400 P0QAM. (1.0) l (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)

8. ADMINISTRATIVE PROCEDURES. CONDITIONS, AND LIMITATIONS' PAGE 99 QUESTION 8.26 (2.00)

SummE/ize four (4) incidents which require immediate l notification in accordance with HPP-100 Radiological .

Protection Plan.

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

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

. g .

'9 3/4 LIMITING CONDITIONS FOR OPERATION AND SURYEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Limiting Conditions for Operation and ACTION requirements shall be applicable during the OPERATIONAL MODES or other conditinns specified for each specification.

3.0.2 Adherence to the requirements of the Limiting Condition for Oper-ation and/or asscciated ACTION within the specified time interval shall constitute compliance with the specification. In the event the Limiting Condition for Operation is mstored prior to expiration of the specified time interval, completion of the ACTION statement is not required.

3.0.3 k' hen a Limiting Condition for Operation is not met, except as provid-d in the associated ACTION requirements, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action shall be initi,tc4 ' f' to place it, placing theasunit in a MODE applicable, in:in which the Specification does not apply to

1. At least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , 2.

At least HOT SHUTOOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

3. At least COLD SHUT 00hN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Where corrective measures are complete.1 that permit operation under the ACTION ,

requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting condition for Operation. ~

Specifications. Exceptions to these requirements are stated in the individual 3.0.4 Entry into an OPERATIONAL MODE or other specified applicability ;

condition shall not be made unless the conditions of the Limiting Con-dition for Operation are met without m11ance on provisions contained in '

the ACTION statements unless otherwise excepted. This provision shall not prevent passage through OPERATIONAL MDES as required to comply. with ACTION statements.

i 3.0.5 When a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable !

Limiting Condition for Operation, provided: (1) its corresponding normal or ,

emergency power source is OPERABLE; and (2) all of its redundant system (s),

sybsystem(s), train (s), component (s) and device (s) are OPERABLE, or likewise t satisfy the requirements of this specification. Unless both conditions (1) and (2) are satisfied, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months action shall be initiated to place the ,

unit in a MODE in which the applicable Limiting Condition for Operation does not apply by placing it as applicable in: i

1. At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , ;

2. At least HOT SHUTOOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and [

3. At least COLO SHUTOOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . ;

This Specification is not applicable in MODES 5 or 6. [

CRYSTAL RIVER - UNIT 3 3/4 0-1 Amendment No. 40 [

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y .I" T:NG CON:* :Of; =0E ODER ~ ION _ _ _

3.2.4 THE OUCRANT P0' DER TILT sna11 not ex:ee: the Stenev State Limit

of Table 3.2-2. i APPLICAEILITY: MODI 1 above 15% of RATED THER%L POWER.-

l ACTION:

a. Witn the OUCRANT POWER TIL~ eetermined to ex:ee: the Steaev S dte Limit but less : nan or ecual to the irknsien: Limit of Iaole 3.2-2.

1. Within 2 hou-s:

a) Ei:ne recu:e :ne OUCRANT PonR TILT :: witnin its Steacy State Limit, or b) Recu:e THERMAL POWER so as no :: ex:ee: THER%L

, DOWER, in:1uding power level cutoff, allowable for l ne rea::or to:iant pump combinatier less at leas:

l i 25 for ea:h 11 :f OUCRANT POWER TILT in excess of

ne Steacy S a:e Limit and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , reduce

ne hu: lea Overoower Trio Setooin and :ne Nu: lear Ove-cowe Base: on RCS Flo.: ane AY.It DOWER IMBALANCE

-1: Se:o:in; at leas: 21 for ea:n it of OUCRAC ,

POGE TILT in ex:ess of the Steacy State Limit.

2. Veri'y tha: :ne OUCRANT POWER TILT is within its Steady Sta e Limi wi nin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after ex:ee:ing the Stency State Limit er recu:e THERMC POWER to less tnan 60% of THER%u POWER allowable for the rea: tor :ociant one combination within the nex: 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and recu:e :ne hu: lear Overpower Trio Seto:in: to < 65.5% of THERME POWER allowable for :ne rea: or coolant pum: :ombination viinin the next 4 heues.

2. Icen-if v an: :o-re:: :ne cause of ne out f limi: :on-ettien 'o-le :: in:reasin; THERMt POWER: subsecuen:

POWER ODERATIOh above 60% of THERKR POWE allowable for the -ea::o :o:lant pum: com ination m!y 0"o:ee:

orovicee :ne: :ne OUCRAC DOWER TILT is verifie:

witnin its 5:ea:y State Lirit at least on:e oe- hou fo-

, 12 nou-s or until verifie: a::ectacle at eg; er greate*

RATED THERMAL DOWER.

'See Spe: Tai Tes Ex:eo:1on 3.1C.1.

CRYSTC RIVER 'JNI 3 2/4 2-5

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4

// L*v!T NG CCNDITICN 2CR CPE?ATION I"entinueC

. With :ne OUACRANT ?CWER 7!L7 :etar?.ined :: exceec :ne Telnsien; Limi: Out less : nan :ne Maximum Limi: Of Table 3.2-2, :ue ::

misalignmen cf ei:ner a safety, regulating Or axia :cwer shaming rec: i

1. Recuca THERFAL PCWER at leas; 25 for each li Of incica ad OUACRANT PCWER TIL7 in axcess :f the 5:aady Sta:a *.imi-wi nin 30 minutes. ,

f

2. Verify :na: :ne CUACRANT PCWER 7'L7 is witnin ( s 7 insten:

Limi: niinin 2 heur$ a' tar ex aecing :ne 7rtnsien Limi

Or recuce THERMAL PCWER: : less : nan 50% :( MERMAL :CWER aliewa 1e der :n.e renc::r c:cian: :um ::moinatien ai nin

ne next Z neurs anc reduca :ne Nucitar Cver:cwer 7-i:

I Set:cin: :: < 55.5 Of THERMAL :C'aER alicwaole ' r :ne .

reac :r ::cian; :um: c:mciaati:n aithin :ne nex 1 9eurs.

[

I 3. I:entify and c:r te; ne :ause O' :ne cu: ef limi ::n-l :1:i:n :ri:r :: inertasing 7'-ERFAL :CWER; su:secuen:

, :CkER CPERAT:CN a:cve 50% :# TWERMAL :CWER nit:wa:1e ':r r

. :ne react:r :: clan: :ume ::::: nation ?.ay :recaec :r:v :ec d :na: :ne ;UACRANT PCkE3 7:L7 's serifiec ai hin its :ta:y M 5:2:a Limi at leas: Onca :er 9:ur f:r 12 curs :r 2n- '

i sert'iec ac:a :a:ie a: 35 :r ;rta:ar RAT!0 79ER."AL

C'a ER .

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., :. Wi:n :.9e UACRANT :C'WER TIL7 ce:armi ec :: excato :ne -$ns er.-

L'9*: :u: less : nan :.9e Saximum .:r Of Tacie 3.I.2, :ue ::

, :auses ::ner : nan :ne misaligeren: :' ei:ner i sa't:y, .-tquit:.

j ing er axial :cwer sna:ing r:c:

'. Recuce THERFAL :C'WER :: less nan 5C% of 79ERFAL :C'aER

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al'. cwa:!e f:r :ne reac r ::c'ar :um: ::ecina:i:r. 4i:nie 2 neurs arc reduca :ne Nuclear 'eer:caer Tri Se::ci9-I. :: < 55.5: ef 7'-ERMAL :CkER ti':waele 3:e :ne renc :e

!! c:cian* Oum :*m:1"ati:n wi:ni" :ne .*eX: . "curs.

i l' 2. * *en:' #y inc ::ratc* re :ause :#

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':'On Or* P :: 'acetas'*g ~ ERMAL ChER . s'. s te'.ea -

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se....w ...

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Ou-* ******a**:P "ay Or :19e :* './'",Sc

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v sp. ;r.i r.aa.s ... .g ,...e.

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. A2T:0ti: (Continued) :

C. k'itn the OVADRAN* POWER TILT determined to exceed tne Maximum !

Limit of Table 2.2-2, recu:e THERMAL POWER te < l5t cf RATED

~

THER%AL POWEE within 2 neurs.

SUE'!EI'.L

. ANCE REQUIREMEfC5 4.2.4 Tne OUADRANT POWER T LT sna11 de octeminec te be within :ne 11rits at least on:e every 7 etys curing ope-ation aoove 151 ef RATE:

THERMAL POWER exce:: wner. :ne OUADRANT POWER TILT monito .s inopertole, ,

nen :ne OUADRANT POWER . T:'T snail be calculate: a: least on:e pe- 12 nours. '

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TABLE 3.2-2 QUADRANT POWER TILT LIMITS STEADY STATE TRANSIENT MAXIMUM LIMIT LIMIT LIMIT QUADRANT POWER TILT as Measured by:

Symmetrical Incore Detector System 3.20 9.08 20.0 Power Range Channels 1.61 6.96 20.0 Minimum incore Detector System 1.73 4.40 20.0 CRYSTAL RIVER - UNIT 3 3/42-!! Amendments Nos. Jf J), 72,77

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3/4.3 INSTRUMENTATION

> 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPEPATION

3. 3.1.1 As a minimum, the Reactor Protection System instrumentation chant.?ls ano bypasses of Table 3.31 shat' ue OPERABLE with RESPONSE TIMFS as shown in Table 3.3-2.

APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As sheen in Table 3.3-1.

SURVE!LLANCE REQUIREMENTS 4.3.1.1.1 Each Reactor Protection System instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK CHANNEL CAllBRATION And CHANNEL FUNCTIONAL TEST

operations during the PODES and at

, the frequencies shown in Table 4.3-1.

4.3.1.1.2 The total bypass function shall be demonstrated OPEPABLE at least once per 18 renths during CHANNEL CALIBRATION testing of each channel affected by bypass operation.

4.3.1.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIPE of each reactor trip function shall be demonstrated to be within its limit at least once per IP months. Each test shall include at least one channel per function such that all charnels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the "Total Fe. of Channels" column of Table 3.31.

i l'

N r

l J

0

!RYSTALRIVEP-UNIT C 3 3/4 3-1

TAeLE 3.3-1 .

REACTOR PROTECTION SYSTEM INSTRtNENTATION .; , ,-

Q .

g TOTAL NO.

MINIMUM y FUNCTIONAL UNIT OF CHAINELS CHANNELS CHANNELS APPLICABLE TO TRIP OPERABLE IWDES ACTION E 1. Manual Reactor Trip 1 1 1 1, 2 and' 8 -

2. Nuclear Overpower 4 2 3 1, 2 2#

5 3. RCS Outlet leperature - Hf gh 4 2 ,

3 1, 2 3#

4 Nuclear Overpower Based on RCS 4 2(a) 3 1, 2 ' 2#

Flow and AXIAL POWER IPEALANCE

5. RCS Pressure - Low 4 2(a) 3 1, 2 3#

6. RCS Pressure - High 4 2 3 1, 2 3#

7. Variable Low RCS Pressure 4 2(a) 3 1, 2 3#

8. Reactor Containment Pressure - High 4 2 3 1, 2 3#

w 9. Interinedtate Range, Neutron Flux 2 1 and Rate 0 2 1, 2 and* 4

10. Source Range Neutron Flus and Rate A. Startup 2 0 2 2 # # and* 5

8. Shutdown 2 0 1 3, 4 and 5 6

11. Control Rod Drive Trip Breakers 2 per trip 1 per trip 2 per 1, 2 and* 7#

a system system trip system k 12. Reactor Trip Module 2 per trip 1 per trip 2 per g system system 1, 2 and* 7#

trip system '

& 13. Shutdown Bypass RCS Pressure - Hfgh 4

- 2 3 2**, 3**,

w 6#

4**, 5**

7 14 Reactor Coolant Pump Power Monttors 2 per pump 1 from 2 2

or more pumps (a,b) per pump 1,2 25

. 15. AntIcfpatory Reactor Trip 4

. - Main Turbine 2(c) 3 1 3#

w 16 Anticipatory Reactor Trip 4 p r pump 2 per pump (d) 3 per pump

." - Both Main Feedwater Pumps 1

, 3r - j 3

l TABLE 3.3-1 (Continued) .,

l TABLE NOTATION 7i With the control rod drive trip breakers in the closed position and the control rod drive systra capable of rod withdrawal.

When Shutdown Bypass is actuated.

The provisions of Specification 3.0.4 are not applicable.

H High voltage to detector may be de-energized above 1010 amps on both Intern.ediate Range channels.

(a) Trip may be manually bypassed when RCS pressure less than or equal to 1720 psig by actuating Shutdown Bypass provided that:

(1) The Nuclear Overpower Trip Setpoint is less than or equal to 5% of RATED THERMAL POWER, ,

(2) The Shutdown Bypass RCS Pressure--High Trip Setpoint of less than or equal to 1720 psig is imposed, and l (3) The Shutdown Bypass is removed when RCS pressure greater than 1800 psig.

(b) Trip may be manually bypassed when reactor power is less than or equal to ,

2475 MWT and four reactor coolant pumps are operating.

I l

1 (c) Trip automatically bypassed below 45 percent of RATED THERMAL POWER. f (d) Trip automatically bypassed below 20 percent of RATED THERMAL POWER.

ACTION STATDIENTS ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next six hours and/or open

the control rod drive trip breakers.

ACTION 2 With the number of OPERABLE channels one less than the Total

- Number of Channels, STAATUP and/or POWER OPERATION may proceed provided all of the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within one hour.

b. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to two hours for surveillance testing per specification 4.3.1.1.

CRYSTAL RIVER - UNIT 3 3/433 Amendment No. hl, JJ, J). If.

TA3LE 3.31 (Continued)

C ACTION $1ATEm(NT3 (Contf aved) and the inoperable channel above any be by.

Passed for up to 30 sinwtes in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Mried when necessary to test the trip breater associated with the logic of the channel being tasted per Specif tsation 4.3.1.1, and

c. Either. THERML POWER is rt:tricted ta a 75 of RATID AATID TNCRML and the Nuclear Iver.

power Trip 5etMint is reduced to 1 85% of RATED TWCWL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months er the QUADUNT POWER TILT is monitored at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTICN 3 - With tne nweer of CPERA4LE channels one less than the Total Nuncer of Channels START 1JP and POWIR OPERATION may proceed provided both of the following consttions are satisfiec: '

a. The inoperstle channel is placed in the tripped l

condition witnin one howr.

b. The Minirwm Channels OPERA 8LE retwirenunt is met; he=cver, one additional cf.annel any be typassed for up to 2 howrs for surveillance testing per Specification 4.3.1.1, and the inexranle cunnel above may be bypassed for up to 30 sinwtes in any 24 howr period =cen necessary to test the trip treater associated wita the logic of the caannel being testad per Specification 4.3.1.1.

Acti n a . Wita the avseer of cKennels CPERA3LI one less than '

retwirte by the Miniew Channels CPERAILI recwirywnt and with the THERML Power level:

s. < 55 of RATED THERML POWER restore the inoperable channel to UPERABLE status prior to increasing THERML POWER above 5% of RATED THE M L POWER.

b. >51 of RATED THERMAL POWER. POWER OPERATION may continue.

3/434 A:end:ent No. 55 CRYSTAL t*VER L*N:73

ACTION 5 -

, With the ntater cf channels 0PDA8LE .cne bs: than required '

by the Minisnm Channels OPE?AELI requirw. ant and with the "

Tii!RrAL POWER level: , ,

a. /10*I9 ames en the Intar nediate Range (IR) in.

/j T:rv::en ation. restert the inocerable channel :s

OPEPABLE,gatus prior: s iner, easing TiiEAK4L POWER above 10 amps en the IR instru:.antation, b., > 10 10 amps en the !R instrumentation operation s

may continue.

ACTION 6 * - With the ntr.her of channels CPERABLI one less than re-quired by the Miniers Channels OPEPAit.! requirement. .

verify complianca with the SHUTDOWN PARGIN requirt ants of Specificatica 3.1.1.1 within ene hour and at least onca per 12 hcurs thereafter.

ACTION 7 -

With the number of CPERAELE channels one 1ess' than the-ic:a1 t umter of Channels START 11P and/or POWER OPERAi!ON ety pr:cted provided all cf the following conditions are ratisfied;

a. *'i:

. Min 1 heur: .

1. P' lace the incpert:1e char.nel in the ri;;ed c:ndition, or

2. Aes:ve gewer sucpif ed to the cen:rel r:c trip device asscciated wita the inecerative enannel,

b. One additional channel :.ay te typassed for v: to 2 hcurs for surveillance testing per Scecification 4.3.1.1, and the inc;erable channel above e.ay te typassed for ,u:! to 20 minutes in anv 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months geriod wnen necessary t: test the trio brtaker associa:ac with the logic of the channel being tastet car

Specifica:1cn a.3.1.1. The inecerable channel ateve city not te tycassad to tas: tae logic of a channel of the trip sys;arn asscciatad with the inceersale chancel.

ACT!on 8 -

. With the nu=ter of channels OPGABLE less than re:uirtd by the Minicum Channels CPE?AELE requirer.ent. te in 4:

least NOT STAN06Y witain 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 25 .

With the ntre'er of channels CPERASLE one less : nan the ettutree itinteun Channels CPERASLE requirement, plant oceration r.ay continue until the next required Channel Functional fest pec-vided the inoperable enaneel is placed in the trisced conct: ton within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

CRYSTAL R!yER-UNIT 3 J/4 3-5 Amendment Ib. Si

TAALE 3.3-2 '

REACTER PROTECTISIt SYSTDI IIISTRWEIITATION Resp 0NSE TIK5 '

..i n

-o l

y functlenal tinit Response Times E

h 1. Manual Reactor Trip Not App 1fcable E 2. Nuclear Overpower

1 0.266 seconds w 3.

l RCS Outlet Temperature - High Not Applicable ,

4 Nuclear Overpower Based on RCS Flow and AXIAL POWER INBALANCE

1 1.842 seconds

5. RCS Pressure - Low 1 0.44 seconds 6 RCS Pressure - Hfgh 1 0.44 seconds

7. Variable Low RCS Pressure Not Applicable

8. Pimip Status Based on RCPPMs**

1 1.44 seconds

9. Reactor Contatsument Pressure - High t Not Applicable

10. Antfcipatory Reactor Trfp - Not Applicable Main Turbine II. Anticipatory Reactor Trip - Not Applicable Both Main Feedwater Pumps 4

y Neutron detectors are exempt from response time testing. Response time of the neutron flux

=

signal portion of the chan wl shall be measured from detector output or input of first electronic component in channel.

e **

P Time response testing of the RCPf% may exc3ede testing of the current and voltage sensors I and the watt transducer.

TO N

,LE 4.3-1 REAcitut PROTECTICII SYSTEM IIISTRtNENTATICII SINtVEILLANCE REQUIROENTS

. ~.

G CMAINYEL MODES IN INitCM CMAINFEL g ,FUIICTIDIIAL 1851T CKCK CHAIBIEL FUNCTIOIIAL SURVEILLAsICE r- CALI3 RAT 1011 TEST M QtfIRED

1. N nual Reactor Trip M.A. M.A.

% S/U(1) N.A.

~

2. Muclear Overpower S E D(2) and Q(7) M 1,2 Q 3. RCS Outlet Temperature--High 5 R M 1,2

4. Nuclear Overpower Based on RCS 9 Flow and AIIAL POWER IMBALANCE S(4) M(3) and Q(1,8) '

M 1,2

5. RCS Pressure--Low $ R M 1,2

6. RCS Pressure--High S R M 1,2 7 Variable Low RCS Pressure 5 R M 1,2 w 8. Reactor Contalanent Pressure--High 5 1 R M 1,2

9. Intermediate Range, Neutron Flux and Rate -

S R(7) S/U(1)(5) 1, 2 and*

10. Source Range Heutron Flus and Rate 5 R(7) S/U(1)(5) 2,3,4 aind 5 11 Control Rod Drive Trip Breaker N.A.

Y N.A. M and S/U(1) 1, 2 glwf*

g 12. Reactor Trip Module M.A. N.A. M 1, 2 and

R g 13. Shutdown Bypass RCS Pressure--High 5 R

,+ M 2**,3**,4** 5**

14 Reactor Coolant Pump Power Monttors S R(9) M 1,2 3 15. Anticipatory Reactor Trip - Main Turbine 5 R M 1 16 Anticipatory Reactor Trip - Both Main

. Feedwater Pumps S w R M 1

e l ';

/ TABLE 4.3-1 (Continued' NOTATION

- With any control rod drive trip breaker closed.

- - When Shutdown Eypass is actuated.

(1) - !! not performed in previous 7 days. .

(2) - Heat balance only, above 13% of RATED THERMAL POWER.

(3) - When THERMAL POWER (TP) is above 30% of RATED THERMAL POWER (RTP), compare out-of-core measured AXIAL POWER IMBALANCE (APlo) to incore measurec AXIAL POWER IMBALANCE (APig) as follows: .

h (APlo AP!g) e imbalance Error Recalibrate if the absolu;e value of the imbalance Error is equal to or greater than 3.3%.

(4) -

AXIAL POWER IMSALANCE a,nd loop flow indications only.

(3) - Verify at least one decade overlap if not verified in previous 7 days.

(6) - Each train tested every other month.

(7) - Neutron detectors may be excluded from CH ANNEL CAllBR ATION.

(8) - Flow rate measurement sensors may be excluded from CHANNEL CALIBRATION. However, each flow measurement sensor shall be calibrated at least once per 18 months.

(9) - Current and voltage sensors may be excluced from CHANNEL CAllB R ATION.

envtras oivro.,,ew i tio s.e e . .. d. . o u 41 77

t I

{

s l

.. j EMERGENCY CORE COOLING SYSTEMS _

i ECCS SUS $YSTEMS, . T.v > 280'F LIMITING CONDITION FOR OPERATION 1 3.5.2 Two independent ECCS subsystems shall be OPERA 8LE with each subsystem comprised of: ,

a. One OPERA 8LE high pressure injection (HP!) pump,

b. One OPERA 8LE low pressure injection (LPI) pump.

One OPERA 8LE decay heat cooler and c.

d. An OPERABLE flow path capable of taking suction from the borated water storage tank (BWST) on a safety injection signal and manually transferring suction to the containment sump during the recirculation phase of operation.

APPLICA8ILITY: M00ii 1. 2 and 3.

ACTION:

a. With one ECCS subsystem inoperable, restore the inoperable subsystem to OPCRABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT ,

SRUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . l

b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared are

submitted to the Commission pursuant to Specification 6.9.2 '

i within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. ,

l l l i

i i

i CRYSTAL RIVER - UNIT 3 3/4 5 3

t . ,

I

.t EMERGENCY CORETCOOLING SYSTEMS G .

S'URVEILLANCE REQUIREMENTS , ,

4.3.2 Each ECCS subsystem shall be demoastrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power opdrated or automatic)in the flow path that is tvit locked, sealed or otherwise secured in position, is in its correct position.

b. By a visual inspection wh!ch verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suction during LOCA conditions. This visualinspection shall be performed:

1. For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and 2, Of the areas affected within containment at the completion of each containment entry when CONTAINMENT INTEGRITY is established.

c. By verifying the correct position of each mechanical position stop for*the following HPl stop check valves prior to restoring the HPl system to OPERABLE status following periodic valve stroking vr maintenance on the valves.

l

l. MUV-2 ,

2. MUV-6

3. MUV 10

d. By verifying that the flow controllers for the following LPI throttle valves l operate properly prior to restoring the LPl system to OPERABLE status following periodic valve stroking or maintenance on the valves.

1. DHV.!!0

2. DHV-!!!

e. At least once per 18 months by:

1. Verifying automatic isolation and interlock actiors of th! DHR system from the Reactor Coolant System when the Reactor Coolant System pressure is greater than or equal to 280 psig. \

CRYSTAL RIVER - UNIT 3 3/434 Amendment No.17,77

t ""==~ m m .r. . r n .y. ~ ~ ,-

.. .- . . . . 1

- - - t, ; o :.. .

DE3tCDtCT CORI C00LiNC* ffTT80 I -

..... ... ....'.,)*',,,

.,,_- p]

SURVI!LLANCE RIO wt f Continued) E

M.'i 6MI 5 -

w ~.,m_ . _ _ i

-...,_.,,w W 2. Verif ying the correct position of each mechanical position stop

' e' for each of the stop check valves listed in Spectiteation 4.5.2.c.

3. Verifying that the flow controllers for the throttle valves listed in frecification 4.5.2.d operate proper 1;.

4. A visual inspection of *he containment emergency susp which verifies that the subsysten suction inlets are not restricted by debris .and that the susp components (trash 'r'acks, screens, etc.) show 'no evidence of structural 'istress or corroston,

5. Verifying a total leak rate less than or equal to 6 gallons per hour for the LP! systen att a) Normal operating pressure or a hhrostatic test p of greater than or equal to 150 psig for those part ressure s of the systes downstress of the pump suction isolation valve, and b) Creater than or equal to 55 psig for the piping from the containment energency sump isolation valve to the pua; suction isolation valve.

f. At least once per '18 months, in MODE 6, by

1. Verif ying that each autcoatic valve in the flow path actuates to its correct position on a high pressure or low pressure j

safety injection test signal, as appropriate.3

2. Verifying that each NpI and Lp! pump starts automatteally upon receipt of a high pressure or low pressure safety injection test signal, as appropriate.3 !

g. Following completion of NPI or LPI systen modifications that could have altered systes flow characteristics , by l performance of a !!ow talance test during shutdown to confira the following in)ection

. flow rates into the Reactor Coolant Systes:

MPI systep - Sinele Pus 52 LPI syster sinle Pu-e Single pump flow rate greater than 1. Injection Les A - 2500 or equal to 500 gpa at 600 psig, to 3100 gps.

While injecting through 4 Injection 2. Injection Let B

2800 Legs, the flow rate for all to 3100 gps.

combinations of 3 Injection Legs greater than or equal to 350 gps at 600 pstg.

I Flos talance tests performed prior to complete ins:allatt:t 0:

zodtfacattens are valtd t! performed <tta the syste: :na .;e sna:

could alter flow character;stics in effect 2

Ine Hp! T1ow Bala,nce Test shall te perf ormed in MODE 3-3 Tor Cycle VI, the surveillance f requendy shall te at least once pe:

fuel cycle in MODE 6.

Sym em: . m- 1 1 : (. m:c* 22. '?. 77.;

l Cp1 Al*(NT $YiTJS 3/4.6.3 CONTAINM(NT ISOLATION VALV[$

i i LIMITING CONDITION FOR OPERATION __

3.6.3.1 The containment isolation valves specified in Table 3.41 shall be OPELA8LE with isolation times as shown in Table 3.61.

APPLICABILITY: MODES 1. 2, 3 and 4 ACTION:

- With one or more of the isolation valve (s) specified in Table 3.61 inoperable, either:

Restore the inoperable valve (s) to OPERA 8LE status within 4 l

4.

hours, or

b. Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one deactivated automatic valve secured in the isolation position, or

c. Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by une of at least one closed e.anual valve or blind flange; or

d. Be in at least NOT STAN;lY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD 5=UTL'. =t tnin the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

1

$URVEILLANCE REQUIREMENTS 1

4.6.3.1.1 The isolation valves specified in Table 3.61 shall be demonstrated OPERABLE prior to returning the valve to service af ter painten-ance, repair or replacement work is peri'ormed on the valve or its assectated actuator, control or power circuit by performance of a cycling tef t and '

verification of isolation time. .

l1,RT$ Tat elves , t!Nf T 1 A-e nd en t No. )*,5P,6 3 3/4 6 14

O CONTAINMENT SYSTEMI .

l

$URVElLLANCE RE@tREM(WT5 (Continued) 4.6.3.1.2 Each isolation valve specified in table 3.61 shall be demonstrated OPERA 8LE during the COLD 5WT00WW or AIFU(LIM MODE at least once per 18 months by:

a. Verifying that on a containment isolation test signal,' each automatic isolation valve actuates to its isolation position.

c. Verifying that on a containment radiation-high test If enal, each purge and exhaust automatic valve actuates to its '

isolation position. ,

~

i

. I CRYSTAL RIVER - Uti!T 3 3/4 6 16

TABLE 3.6-1 CONTAINf4E!d ISOLATION VALVES O

V YALVE NUMBER. FUNCTION.,

ISOLATION TIME (seconds)

A. CONTAINMENT ISOLATION

1. BSV-27 check i closed dure nor, operation NA and open dur. RB spray BSV-3 i "

'60 BSV-26 check # NA BSV-4 f- 60

2. CAV-126(A)*+ iso. CA sys. fr. RC.letdn. 60 CAY-1 (A)*+ iso. CA "sys. fr. pzr. 60 CAV-3 (A)*+ 60 CAV-2 (B)*+ iso. CA sys. fr. R8 60 CAV-4 # (A)*

isolate liquid sampling 60 system C.W-6 f * "

60 CAV-5 i * "

60 CAY-7 i * "

60

3. CFV-20 check iso. N2 supply fr. CFT-1A CFV-28(A/B)*+

NA 60 l C'FV-18 check . iso. MU system fr.' CFT-1B NA ,

CFV-26 (A/B)*+ 60 l CFV-19 check iso. MU system fr. CFT-1A NA CFV-25 (A/8)*+

60 l CFV-42(B)*.+ iso. liquid sampling fr. 60 }

'CF system i CFY-15

iso. WD sys. fr. CF tanks 60 CFV-16 * "

60 CFY-29 (B

60 l

' CFY-11 (A)*+ iso. CF tanks fr. liquid 60 l sampling,system

.C FV-12 ( A) * + 6'O l CFY-17 check iso. N2 supply fr. CFT-1B NA CFV-27 ( A/8 ) * + 60 l CRYSTAL RIVER - UNIT 3 3/4 6-17 Amendment flo. X#,9A 91 1

l

, TABLE 3.6-1 (Continued)

(^T CONTAINHE'NT ISOLATION VALVES V

VALVE tUi3ER FUNCTION ISOLATION TIME (seconos)

Blind Flange 34B iso. fdel transfer tube from NA' Trans fer' Canal BTind Flange 436 NA Equipttqnt Hatch iso. RB HA Personnel Hatch iso. RB RA i Not subject to Type C Leakage Test

The provisions of Specification 3.0.4 are not applicable.

+ May be opened on an intermittent basis under administrative control consisting of a dedicated operator stationed at the valve in continuous communication with the control room with necessary parameters to rapidly isolate the penetration on a valid indication.

(A)' Isolates on Diverse Isolation Actuation Signal A O (B)

(A/B)

Isolates on Diverse Isolation Actuation Signal B Isola'tas on Oiverse Isolation Actuation Sigrial A or B O

CRYSTAL RIVER - UNIT 3 3/4 6-21a MendmentNo.#M9I

e I

CLECTRICAL POWER SYSTEMS

'/j SHUTDOWN LIMITING CONDIT!ON FOR OPERATION l

i

".3.1.2 As a minimum, tne follcwing A.C. electrical power sources snall

i. OPERABLE: .

a. One circuit between the offsite transmission network and the l!

onsite Class IE distribution system, and i ;

b. One diesel generator with:

,j 1. Day fuel tank containing a minimum volume of 400 gallons q of fue',

4. A fuel storage system cer.tciaing t m.inimum vo'.are of 20,300 galions cf fuel, anc h-

3. A fuel transfer pumo.

APo'ICABillTY: MODES 5 anc 6.

I

' " ~ Ct. :

'aith less than the above minimum recuired A.C. electrical power sources j OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity enanges until the minimum required A.C. electrical power sources are restored to OPERABLE status.

f

SURVEIL AN: E TECUIREMENT5 l

,j4.6.1.2 ine soove requirea A.C. electrical, power sources snall be h cemonstrated OPERABLE by performance of each of the Surveillance Re-

ouirements of 4.8.1.1.1 and 4.8.1.1.2, except recuirement 4.8.1.1.2.a.5.

I CRYSTAL RIVER - UNIT 3 3/4 8-6

e g .

l . .

! ..L j l i .

l l ELECTRICAL POWER $YSTEMS 5URvEtLL ANCE REOUIREMENTS (Continued) _

) 2. Verifying the generator capability to reject a load of b 315 kw without tripping. :

I d . 3. Simulating a loss of of filte ,mwer in conjunction with Reactor Building ,

high pressure ud Reactor building high high pressure tests signals, andi a) Verifying de-energization of the emergency buses and loac shedding from the emergency busses, b) Verifying that the 4160 v. emergency bus tie breakers open, c) Verifying the diesel starts from ambient condition on the auto-start signal, energizes the emergency busses with permanently connected loads, energizes the auto-connected emergency loses through the load sequencer, and operates for k3 minutes while its generator is loaded with the emergency loads.

4. Verifying the diesel generator operates for h 60 minutes wh!Je loaded to > 3000 kw, f

3. Verifying that the auto-connected loads to each diesel generator do not exceed the 2000 hour0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months rating of 3000 kw, and

6. Verifying that the automatic load sequence timers are OPERABLE with each load sequence time interval within + 10A

This test shall be performed in MODI 3 8 The specified 18 month frequency may be waived for Cycle VI startup CRYSTAL RIVER - UNIT 3 3/4 8 3 Amen dm e n t s .*10 s . ,8, 24'. 7 9

ELECTRICAL POWER SYSTEMS ..

SURVEILLANCE REQUIREMENTS (Continued) 4 At least once per 18 months, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actua energency loads for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months when the battery is subjected to a battery service test.

5. At least once per 60 months, by verifying that the battery capacity i-at least 80% of the manufacturer's rating when subjec ted '.o performance discharge test. This performance discharge test shall Me performed subsequent to the sa tis fac tory completion of the requiren battery service test.

4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE.

a. At least once per 31 days on a STAGGERED TEST BASIS by:

1. Verifying the fuel level in the day fuel tank,

2. Verifying the fuel level in the fuel storage tank,

3. Verifying the fuel transfer pump can be started and transfers fuel from the storage system to the day tank, 4 Verifying the diesel starts from ambient condition and can :.

accelerated to at least 900 rpm,

5. Verifying the generator is synchronized, loaded to greater than or ecual to 1500 kw, and operates for greater than or equal to 60 minutes, and

6. Verifying the diesel generator is aligned to provide standby power to the associated emergency busses,

b. At least once each 92 days by verifying that a sample of diesel fuel from t9 fuel storage tank is within the acceptable limits specified in Table 1 of ASTM D975-68 when checked for viscosity, water and sediment.

c. At least once per 184 days in lieu of surveillance 4.8.1.1.2.a.4 by verifying the diesel starts from ambient condition and accelerates to at least 900 rpm in less than or equal to 10 seconds,

d. At least once per 18 months, by:

1. Subj ec ting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's reconnendations for this class of standby service, CRYSTAL RIVER - UNIT 3 3/4 S.4 Amendment No, p, y, 7).96

1 i

i. , ELECTRICAL POWER SYSTEMS SURVEILLANCE REOUIREMENTS

c. Demonstrated OPERABLE by de:emining that each battery supplying DC control power to the 230kv switchyard breakers is CPERA5LE;

-1

1. At least once per 7 days by verifying that:

a) The electrolyte level of each pilot cell is between the minimum and maximum level indication marks, b) The pilot cell specific gravity, corrected to 77'F, and full electrolyte level is > l. F c) The pilot cell voltage is > 2.15 volts, and d) Tne overall battery voltage is > 120 volts.

2. At least once per 92 days by verifying that:

a) The voltage of each connected cell is > 2.15 volts under float charge and has not decreased more tnan 0.10 volts from the value observed during the base-line tests, and b)

The specific gravity, corrected to 77'F and full electrolyte level of each connected cell is > 1.20 and has not decreased more than 0.01 from tai value observed during the previous tests, and c) The electrolyte level of each connected cell is between the minime and maxim a level indication

. marks.

3. At least once per 18 months by verifying that: .

a) The cells,. cell olates, and battery racks shew no visual indication of phy.sical damage or acnormal deterioration.

b) The cell-to cell and tenninal connections are

! clean, tight and coated with anti-corrosion .rateria'is, l

c) The battery charger will supoly at least 95 amoeres f

' at 125 volts for at least 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

l

ELECTRICAL POWER SYSTEMS ACTION feentinued)

d. With two of the above required of f site A.C. circuits inoperable, demonstrate the OPERABILITY of two diesel generators by performing Surveillance Requirement 4. 8.1.1. 2. a . 4 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , unless the diesel generators are already operating; restore at least one of the inoperable offsite sources te OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . With only one offsite source restored, restore at least two of f-site circuits to OpIRABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from time of initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

e. With two of the above required diesel generators inoperable, demonstrate the OPERABILITY of two offsite A.C. circuits by performing Surveillance Requirement 4.8.1.1.1.4 within one hour and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore at least one of the inoperable diesel ger.e ra to r s to CPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Restore at least two diesel generators to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months f rom time of Ir.itial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COL:

SHUTDOWN within the f ollowing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.8.1.1.1 Each independent circuit between the of f site transmission network and the onsite Class 1E distribution systes shall be:

a. Determined CPERABLE at least once per 7 days by verifying correct breaker alignments and indicated power availability.

b. Demonstrated OPERABLE at least once per 18 months during shutdown by transferring unit power supply from the normal circuit to the alternate l circuit.

l -

l l CRYSTAL RIVER - UNIT 3 3/4 3.; Amendment No. 4*' 98

~

3/4.s ritCTRICAI. POVD SYSTEMS

^ ; 314.8.1 A. C. 500Rcts l OPERATING LIMITING CONDITION FOR OPEPATION 1 -

i 3.8.1.1 As a siniaua, the following A.C. electrical power sources shall be OPERABLI 1

a. Two physically independent circuits between the of fsite transmission networ 4

and the onsite Class 1E distribution systes, and l b. Two separate and independent diesel generators each with:

1

1. A separate day fuel tank containing a siniaua volume of 400 gallons c :

fuel,

2. A separate fuel storage systes containing a sinimus volume of 20,30u gallons of fuel, and

3. A separate fuel transfer pump.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTICN: ,

a. With one of the above offsite circuits inoperacle, demonstrate th-OPERABILITY of the remaining A.C. sources by performing Surveillance Requirement 4.8.1.1.1.4 within one hour and at least once per 8 hour:

thereaf ter; and 4.8.1.1.2.a.4 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , unless the diesel generator:

are already operating. Restore at least two of fsite circuits to OpERABLI status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least NOT STANDBY within the next 6 hourt and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With one diesel generator inoperable, d escristrate the operability of the remaining A.C. sources by performing Surveillance Requirement 4.8.1.1.1.a within one hour and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereaf ter; and 4.8.1.1.2.a.4 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Restore two diesel generators to OPERABLE status within 7I hours or be in at least ROT STAND 8Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLI SHUTDOWN withia the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With one offsite circuit and one diesel generator of the above required A.C.

electrical power sources inoperable, demonstrate the OPEAABILITY of the remaining A.C. sourcss by performing Surveillance Requirements 4.8.1.1.1.4 within one hour and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereaf ter; and 4.8.1.1.2.a.4 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , unless the diesel generator is already operating. Restore at least one of the inoperable sources to OPERABLE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months er be in at least HOT STAND 8Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDCVN withir, the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Restore at least two offsite circuits and two diesel generators to optRABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months f rom the time of initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

CRYSTAL RIVER - UNIT 3 3/4 8-1

, . ji,, -THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS. AND PAGE 100 THERMODYNAMICS ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M ANSWEN' 5.01 (1.00) c REFERENCE CR ROT 3-4, obj 3 K/A (4.5/4.9) 000074K103 ...(KA'S)

ANSWER 5.02 (1.00) c.

REFERENCE Oconee OP-0C-SPS-THF-PD pp. 10 obj. 2d 2.9/3.3 CR ROT 2-9, obj 3 193009K107 ...(KA'S)

ANSWER 5.03 (1.00) d REFERENCE CR ROT 5-10, obj 9 K/A (3.4/3.4) 192006K107 ...(KA'S)

ANSWER 5.04 (1.00) e REFERENCE CR ROT 1-8, obj 3 K/A (3.6/3.8) 192008K121 ...(KA'S)

, ' E_;. THEORY OF NUCLEAR POWER PLANT OPERAYION. FLUIDS. AND PAGE 101 THERMODYNAMICS .

ANSWERS -- CRYSTAL RIVER- -88/01/12-DEAN, W M 9

ANSWER' 5.05 (1.00) b REFERENCE CR ROT-2-9, obj 8 K/A (2.2/2.4) 193008K127 ...(KA'S)

ANSWER 5.06 (1.00) b REFERENCE CR ROT 3-2, obj 7 K/A (2.4/2.6) 193008K116 ...(KA'S)

ANSWER 5.07 (1.00) c REFERENCE CR ROT 3-3, obj 6, 7 K/A (3.9/4.1) 193008K123 ...(KA'S)

ANSWER 5.08 (1.00) d REFERENCE CR ROT 2-10, obj 5 K/A (3.2/3.5) 192005K114 ...(KA'S)

. ji. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 102 THERMODYNAMICS ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W H ANSWER 5.09 (1.50) e) Decrease (+.5 ea) b) Decrease c) Increase REFERENCE OP-00-TA-NT pp 7/8; LO lb CR ROT 3-3, obj 2 (3.6/3.9) 041020A202 ...(KA'S)

ANSWER 5.10 (1.50) a) Lower (+.5 ea) b) Higher c) Higher REFERENCE CR ROT 5-7, obj 8 K/A (3.5/3.8) 015000A101 ...(KA'S)

ANSWER .5.11 (2.00) e) Increase (+.5 ea) b)- Decrease c) Decrease d) Increase REFERENCE CR ROT 1-7, obj 15 K/A (2.5/2.8) 192005K107 ...(KA'S)

I 1

l l

.. 5 .- THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 103 THERMODYNAMICS .

ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M ANSWER' 5.12 (1.50)

1) Tc=Tsat for the OTSG (+.5 ea)

2) Delta T develops / stabilizes at 45-55 degrees

3) Average of 5 highest thermocouples follows Th within 10 degrees REFERENCE CR ROT 3-3, obj 10 K/A (4.2/4.2) 193008K122 ...(KA'S)

ANSWER 5.13 (1.00) a) Containment Radiation (+.5 ea) b) High Flow on the Broken Line REFERENCE CR ROT 3-8, fig i K/A (4.2/4.6; 3.5/3.8) 000009A202 000009A236 ...(KA'S)

ANSWER 5.14 (1.00)

1) Pu 240 buildup-- More Negative (+.5 ea)

2) Accumulation of fission products-- More Negative REFERENCE CR ROT 1-9, obj 2 K/A (3.4/3.7) 192004K103 ...(KA'S)

ANSWER 5.15 (1.00) concave down REFERENCE CR ROT 1-12, obj 7 K/A (2.9/3.5) 001010K516 .

.(KA'S)

i 5 .c THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 104 THEBM0 DYNAMICS

-ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M ANSWER 5.16 (1.00) 1). Less boron required due to fuel burnout, this increases the boron worth due to less competition. (or flux hardening is decreased, so higher offective boron absorption cross-section) ( + . 5 ea FLrder L)

2) Fission products poisons build up, decreasing the boron worth.

&cna_o flu x caha.ner oVer cbcc /4 -9 mcmje s harm c.> der R FERENCE CR ROT 1-7, obj 19 K/A (2.8/2.9) 192004K109 ...(KA'S)

ANSWER 5.17 (1.50)

1) Radiolytic decomposition (+.25 ea for any 6)

2) CFTs

3) Decreased ability to retain dissolved gases

4) Fuel Cladding degradation

5) Steam formation

6) BWST via ECCS/ Makeup systems

7) Pressurizer REFERENCE CR ROT 3-5, obj 1 K/A (4.6/4.8; 4.3/4.4) 000074K102 002000A201 ...(KA'S)

ANSWER 5.18 (2.00) a) Increases (+.5) Par temperature increase will cause a pressure increase, increasing margin to saturation (+.5) b) Increases (+.5) Delta t across the core will be lower to produce the same power. Th will decrease and the coolant in the upper regions will be farther from saturation.

(Also higher flow removec bubbles from rod surface)(+.5)

REFERENCE DPC Thermodynamics / Fluid Flow pp 196-198 CR ROT 2-9, OBJ 11 (3.4/3.6) 193008K105 ...(KA'S)

' 5 .' THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS. AND PAGE.105 THERMODYNAMICS ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M ANSWER' 5.19 (1.50)

Lower (+.5) due to the heat transfer area decreasing, the Delta T across the OTSG must increase to achieve the same power. (+1.0) (Q:uA DeltaT)

REFERENCE CR ROT 2-5, obj 10/11 K/A (3.8/4.0) 035010K109 ...(KA'S)

ANSWER 5.20 (2.00)

Q) The lower level of xenon in that quadrant results in a higher power production (+.5) b) Tilt will increase (+.5)due to rapid burnout of the existing xenon in that quadrant (+.5), and then decrease as xenon concentration increases

(+.5)

REFERENCE CR ROT 3-7, obj 5 ; ROT 1-10, obj 9 K/A (3.2/3.6) 000005K103 ...(KA'S)

ANSWER 5.21 (2.00) a) A: Borate w/ ICS in AUTO (+.5 ea)

B: Borate w/ rods in Manual C: Reduce ULD (ICS in AUT0) b) D: Insert Group 8 REFERENCE CR ROT 2-11, ob) 6 K/A (3.6/4.0) l 001050A206 ...(KA'S)

ANSWER 5.22 (1.00)

1) Limit Liquid flow out the break (+.5 ea)

2) Gives operator time to observe Natl Cire develop i

I

. 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 108 THERMODYNAMICS .

ANSWERS -- CRYSTAL' RIVER -88/01/12-DEAN, W M REFERENCE CR ROT 3-5, pp 21

-K/A (3.4/3.8) 000017A211 ...(KA'S)

ANSWER 5.23 (1.50) e) Right of the optimum point (+.5 ea) b) Left of the optimum point c) Above and to the right REFERENCE CR ROT 1-8, obj 10; CR ROT 1-9, obj 4 K/A (2.9/2.9) 192004K107 ...(KA'S)

r .6.- PLAN 1 SYSTEMS DESIGN. CONTROL, AND~ INSTRUMENTATION PAGE 107 ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M ANSWE{; _ 6.01 (1.00) ed REFERENCE CR NA0 96, obj 6 K/A (2.4/2.6) 059000K401 ...(KA'S)

ANSWER 6.02 (1.00) e REFERENCE CR-ROT 4-9, obj 2 K/A (3.2/3.2) 059000K107 ...(KA'S)

ANSW 6.03 (1.00) ,

d) e REFEREN CR ROT 4- obj 9 (4.4/4.6; 2.6/3.0; 3.3/3.7) 000055A214 064000A216 064000K409 ...(KA'S)

ANSWER 6.04 (1.00) b)

REFERENCE CR ROT-4-15 obj 9 (3.3/3.7) 061000K402 ...(KA*S)

ANSWER 6.05 (1.00) b)

REFERENCE CR ROT 4-9 obj 6 (2.9/3.2) 016000A202 ...(KA'S)

.- 6 ., PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 108 ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M ANSWERf 6.06 (1.00)

.c) ch6l. C)

REFERENCE CR ROT 4-01 (3.4/3.9) 006000K602 ...(KA*S)

AN3hER 6.07 (1.00) d)

REFERENCE CR ROT 4-2 obj 4 (3.1/3.4) 004000G10 ...(KA'S)

ANSWER 6.08 ( .50)

TRUE (+.5)

REFERENCE CR ROT 4-13, pp 39, LO 11 K/A (3.7/3.9) 013000K412 ...(KA*S)

ANSWER 6.09 (1.50) a) 2 channels and temperature compensated (+.5 ea) b) 4 channels and temperature compensated c) 1 channel and not temperature compensated REFERENCE CR ROT 4-32, obj 6 K/A (3.6/3.8) 035010K401 ...(KA'S)

. >6.. PLANT SYSTEMS DESIGN.' CONTROL. AND INSTRUMENTATION PAGE'109 ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M i

ANSWERj 6.10 (1.00) c) Input / Output across SG/RX (+.5 ea) b) Header Pressure' error REFERENCE CR ROT 4-14, obj 3 K/A (2.7/2.9) 045000K401 ...(KA'S)

ANSWER 6.11 (1.00) o) Cross tie blocking in effect (+.5 ea) b) Breaker will close REFERENCE CR ROT 4-3, obj 5,6 K/A (2.8/3.1) 062000K403 ...(KA'S)

ANSWER 6.12 (2.00)

1) Heat Senoor in the area (+.25 for method, +.25 for location)

2) Manual Actuator in Control Room

3) Manual Actuator on wall outside Control Complex on turbine deck

4) Manual Spurt button near the FW pumps REFERENCE CR ROT 4-7, pp 32 K/A (3.0/3.3) 086000K406 ...(KA'S)

ANSWER 6.13 (1.50)

DHV-3, 4 (+.5 ea)

DHV-3and4willclose(<wuMAW and are electrically disabled from opening Bypassed by key interlocks in the ES Test Cabinets REFERENCE CR ROT 4-1, pp 131/139, LO 3-3 K/A (3.2/3.5) 005000K407 ...(KA'S)

a '

6s PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 110 ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M ANSWER / 6.14 (1.00)

A partial load rejection has occurred.

REFERENCE CR ROT 4-22 obj 8 (2.4/2.5; 3.1/2.9) 045000A401 045000K407 ...(KA'S)

ANSWER .15 (1.00)

The Diamond ontrol Station will automatically revert to canual contro REFERENCE CR ROT 4-28 objs & 12 (3.7/3.9; 3.4/3.8) 001050A201 001 OK401 ...(KA'S) l ANSWER 6.16 (2.00)

1. Transfers the fans from the industrial cooler to the SW system, o 9- (GA fH<rr N W XPsED

! 2. Starts the pumps.

l 3. Starts the pumps.

j 4. Trips the pump.

l (3.5 ea.)

REFERENCE l CR ROT-4-2 obj. 7 (4.1/4.4)

! 013000K103 ...(KA'S) l ANSWER 6.17 (1.50) l Feed flow and reactor power will increase due to the increased demand.(+.5) l This will cause OTSG pressure ,o r[se above the turbine header pressure TheTurbineggy setpoint (+.5). , pas; Valves will open to restore the header pressure to setpoint, accounting for the extra 5% heat generation (+.5)

REFERENCE CR ROT 4-14, obj 4, 5, 8 K/A (2.9/3.3) 041020K401 ...(KA'S)

. 6.- PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 111 ANSWERS -- CRYSTAh RIVER -88/01/12-DEAN, W H ANSWER; 6.18 (1.00)

Technician is correct. (+.5) When in bypass, the trip relay is supplied power via a separate path not affected by module removal (+.5)

REFERENCE CR ROT 4-12, obj 15 K/A (3.3/3.6) 012000K604 ...(KA'S)

ANSWER 6.19 (1.00)

This prevents isolation of the DH system (+.5) on a spurious ACI signal

(+.5)

REFERENCE CR ROT 4-1, pp 137 K/A (3.3/3.5) 005000G010 ...(KA'S)

ANSWER 6.20 (1.50)

The High Flux Trip and the Power / Imbalance / Flow tripe for Channel A will actuate (contacts open)(+0.75). This will cause the channel A trip relay to deenergize and open contacts (KA1, 2, 3 and 4) in each trip module, making 1 side of the 2/4 required logic (+.75).

REFERENCE CB ROT 4-12, pp 13-14,33,38, LO 3 K/A (4.1/4.2) 015000K101 ...(KA*S)

ANSWER 6.21 (2.00)

1. The valves are modulating solenoid valves that utilize an increasing DC current opposing spring pressure to modulate the valve (0.5)

2. EFIC (0.5)

3. This is done to minimize overcooling of the reactor coolant system when EFW is initiated.(1.0)

' - 6: PLANT SYSTEMS DESIGN. CONTROL AND INSTRUMENTATION PAGE 112 ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M REFERENCE CR ROT 4-15 obj.6 (2.7/2.7)

C61009K411 ...(KA'S)

ANSWER 6.22 p00[ {0 '

The manual operation of-these valves will not be affected by the loss of instrument power. 3;; thi= 4a a propae-ecLion.

(0.5 ea.)

REFERENCE CR ANO-82 obj. 11 ANSWER 6.23 (1.00)

The start selector switch prevents a selected pump from tripping on an undervoltage condition and ensures that the non-selected pump does trip when an undervoltage condition exists on the 4160 ES bus. (1.0)

-or-When a pump is selected, a contact in seriea with the contact that is closed by the undervoltage relay is opened.

This prevents the trip signal from reaching the breaker tripping device. Conversely, when a pump is not selected, the contact is closed and the pump breaker will be tripped if an undervoltage condition occurs on the bus. (1.0) l REFERENCE CR 4-1 p. 90 obj chap. 2-2a. (4.3/4.4) 006000K405 ...(KA'S)

ANSWER 6.24 (1.00)

All of the RPS channels would sense "C" and "D" RCPs tripped via the contact monitors, resulting in a reactor trip.

REFERENCE CR ROT 4-12 obj 3 (3.1/3.5) 012000K603 ...(KA*S)

S

' 6: PLANT SYSTBilS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 113

' ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M ANSWER 6.25 (1.50)

The Rosemount dp transmitters have been observed to experience a zero reference shift (0.5) at a pressure different from that at which they were calibrated (0.5).

This may cause the indicator to read some low value when actual flow does not exist (0.5).

REFERENCE CR ROT 4-9 (2.3/2.5) 016000K601 ...(KA'S)

L i

i I

o i

'7: PROCEDURES - NORMAL, ABNORMAL. EMERGENCY AND PAGE 114 RADIOLOGICAL CONTROL ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M ANSWEi 7.01 (1.00) b REFERENCE CR TS 3.4.6.1, CR ROT 5-1, LO 11 K/A (3.6/3.8); (3.6/4.1) 002020G005 002020K401 ...(KA'S)

ANSWER 7.02 (1.00) c ,

REFERENCE CR ROT 5-14, LO 5 ,

K/A (4.0/4.1) 000011G012 ...(KA'S) i ANSWER 7.0 (1.00) o REFER CE CR R 3-3, LO 6,7 K/ (4-1/4.7) 74A207 ...(KA*S)

ANSWER 7.04 (1.00) d REFERENCE CR OP-204, pp 17; CR ROT 5-2, LO 5

'K/A (3.6/4.0) 001050A206 ...(KA'S) 2 >

> 7 .* PROCEDURES - NORMAL. ABNORMAL EMERGENCY AND PAGE 115 RADIOLOGICAL CONTROL ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M ANSWER 7.05 (1.50) a) TRIP (+.5 ea) b) TRIP c) DON'T TRIP REFERENCE CR ROT 5-28, LO 1; AP-580 K/A (4.2/4.1) 000007G010 ...(KA'S)

ANSWER 7.06 (1.00) c) 5 (+.33 ea) b) 1 c) 3 REFERENCE CR OP-417, pp 4 K/A (3.5/3.8) 103000G001 ...(KA'S)

ANSWER 7.07 (1.00) a) 75% (+.5 ea) b) 100 cpm REFERENCE CR ROT 5-43, LO 5; RSP-101 K/A (2.8/3.4) 194001K103 ...(KA*S)

ANSWER 7.08 (1.00) o) So that the OTSG will provide a good heat sink (+.5 ea) b) SPDS or RCP NPSH curve with plant wide range pressure indication REFERENCE CR ROT 3-5, pp 26; AP-530; ROT 5-25, LO 4 K/A (4.0/4.4) 000074K307 ...(KA*S)

. .7 PROCEDURES - NORMAL. ABNORMAL ~. EMERGENCY AND PAGE 116 RADIOLOGICAL CONTROL ,

ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M

'i ANSWER' 7.09 (1.50)

NO (+.5) TS (3.9.2) require audible counts prior to entering Mode 6 (+0.5) cnd detensioning of a head bolt constitutes entrance into Mode 6 (+0.5)

REFERENCE CR TS 3.9.2, TS 1.4; CR LER 87-023; ROT 5-1, LO 5 K/A (3.3/3.8) 015000G005 ...(KA'S)

ANSWER 7.10 (1.50)

1) Ensure valves on affected OTSG are closed: (+.25) ,

MSIV (411/412 or 413/414)

MBV (FWV-30 or 29)

LLBV (FWV-31 or 32)

SUBV (FWV-36 or 33)

Cross-tie (FWV-28)

MFP Suction (FWV-14 or 15)

(+.75 for valves)

2) Ensure MFPs on affected OTSG Tripped (+.5)

REFERENCE

CR-ROT 5-23, LO; AP-460 K/A (4.1/4.2) 000040G010 ...(KA'S) ,

ANSWER 7.11 (1.00)

1) SSOD (+.5 ea)

2) ANSS REFERENCE CR ROT 5-40, LO 3; CP-115, pp 3 K/A (3.7/4.1) 194001K102 ...(KA'S)

'~7 PROCEDURES _ NORMAL. ABNORMAL. EMERGENCY AND PAGE 117 RADIOLOGICAL CONTROL ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M ,

t e

ANSWER 7.12 (1.50)

1) Start CAP-1A or 1B (+.5 ea)

2) Open CisV-60 (Emergency Boration Isolation)

3) Establish Maximum Letdown REFERENCE CR ROT 5-28, LO 3; AP-580 K/A (4.0/4.0) 000024G010 ...(KA'S)

ANSWER 7.13 (1.00)

1) Reactivity Control (+.2 for CSF, +.05 for position)

2) Thermal Control

3) Radioactive Inventory Control

4) Equipment Availability REFERENCE CR ROT 5-14, pp 9/10; LO 10 K/A (3.8/3.9) 000007G012 ...(KA'S)

ANSWER 7.14 (1.00)

1) J ha CRD Stator alarm ( tt160 degrees) (+.33 ea) 2)pjLay CRD energized

3) RCS Temp )[ 200 degrees 4.,

REFERENCE CR OP-502, pp 5; ROT 4-28, LO 3 K/A (3.3/3.5) 001050G010 ...(KA'S)

'7' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 118 RADIOLOGICAL CONTROL ANSWERS - -CRYSTAL RIVER -88/01/12-DEAN, W M ANSWER 7.15 (1.75) c) 1) > 50 degrees delta T between condensers (+.25 ea)

2) > 8 Degrees delta T between highest / lowest hot gas temperatures

3) > 10" Hg absolute condenser vacuum b) Close MSIVs (+.5 ea)

Notify Load Dispatcher REFERENCE CR ROT 5-29, LO 1 & 3;.AP-660 K/A (2.8/3.9), (2.9/3.2) 045000G014 045000G015 ...(KA'S)

ANSWER 7.16 (1.00) 1)

(tM et.)

Both actuation output breakers are opened (+Ja<Y ea)

2) A dedicated operator must be stationed at the EFIC controls to manually actuate the system if required.

3 Tf C kev retswr wcia; xswww t./) MCx daua nea v.unc,ptaym g,mys-R FERENCE CR LERs87-001, 87-002 K/A (3.7/3.9) 061000G001 ...(KA*S)

ANSWER 7.17 (1.75)

1) Admquate subcooling margin exists (+.25)

2) Maintain RCS Press / Temp < NDT (+.5)

3) Maintain HPI Flow < 540 gpm/ pump (+.5)

4) Maintain subcooling < 100 degrees when no RCPs are operating (+.5)

REFERENCE AP-380; CR ROT 5-22, LO 4 K/A (3.6/4.2) 000009A234 ...(KA'S)

' 7) PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 119 ,

RADIOLOG1 CAL CONTROL ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M ,

o.

ANSWER 7.18 (1.50) a) Controlling on Low Level Limits (+.5 ea) b) one MFP and one MFBP c) 2155 psig/579 degrees

REFERENCE CR OP-208; CR ROT 5-3, LO 4a K/A (3.6/3.5), (2.7/2.9), (4.0/3.9) 010000A302 035010A361 059000A103 ...(KA*S)

ANSWER 7.19 (1. 50) [or am L)

1) Minimize the release of radiation to the environment (+.75 ea) 1

2) Minimize level. increase in the affected OT G (Carryover) >

3) mma Nek+ o 4 non- Ncounc Ve M (* us REFERENCE B & W Technical Bases Document, CH III-E, pp 28/29 CR ROT 5-20, LO 4; K/A (4.2/4.5) 000038K306 ...(KA'S)

ANSWER 7.20 (1.50) c) The Main hook will not reach ROW 1 (+.5 ea) b) POOL CARRIAGE OVERTRAVEL light c) "Z-Z" Tape full-up position must be checked ;

REFERENCE t CR FP-302, pp 5; FP-601, pp 5, 8 ;

K/A (2.7/2.9) 034000G010 ...(KA'S) f ANSWER 7.21 (1.00)

Lifting of secondary safeties due to overfilling the OTSG (liquid release) or overpressurizing OTSG. (+1.0)

REFERENCE ATOG Guidelines, III-E, pp 55 -

K/A (4.1/4.5) 000038K306 ...(KA'S)

q ., , .

7*.

PROCEDURES - NOJ11ML, ABNORMAL. EMERGENCY AND PAGE 120 RADIOLOGICAL CONTROL ?-

1 .

ANSWERS--CRYSTALRIVERj N -88/01/12-DEAN, W M

.\

\q .s' -'

4 \

os. 'f, hA ,, -

e 3 i g i

" D. \ < '. s y v .h ,. q q~ (.,NbbP ANSWER M_ .2Q '-

s n 4 .

1!

.h is above s c) . With levah of 65%, tt $rmal chter of the 0TSG s the thermal

,h')3 Since

. center loss of the corhs (+1.0) of subcooling 7

is an "?ndict. tion of void formation being

'N possible,j Reflux soiling may nee,d to'be established, which requires a greater heat transfer area. (+1.0) s

)

4 REFERENCE CR ROT:3-3, pp20/21; CR ROT 5-25, LO 4 i

is

' 'c' g

K/A ( 6.1.,'4 . 2,\ .l1. . . ( K A ' S ) \

'010001/KPS{

- \ , s a ) ',

, i 3

1. % 2 ,

C\

+ ANSWER ,7.23 (1.0d) , s. ,

3,3 Below 2.Nf psittisecondary' press t 4 N cannot supply sufficient flow with the .

TDEFW Pump, so the pump is ulsettially .f.noperable (+1x01,in iode'3.

4

~

\

. REFERENCE .

3 Pa. LO 5 2 Ch'.L3R K/A (3.5/3.6), 87-17 ( 3; OP-203;

. 3 / -1. A;r 5- 3, N;3 X k- -

061000G005 0610006010 ... ^Kn(S) g ,{ .

~

'b . k g {

ANSWER ' sl'J 2 4 ! a(}.00)

Allow the nc.c$les Po heat up with the,*eedwater (+.5) and aid !c.

transferring he,2t'go as much of the O UGg shell as possible ( + . 's )

.. 1 REFERENCE; s.

s CR ROT 4-32, LO 3; CB' ROT 5-2, LO3a\

? .K/A '3.i/G.4) '

035010G010 ...(KA~S) h I , ,

s s

[ [ O s

4 g

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% -- - _% _' -_\--

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' ' 8 .' ADMINISTRATIVE PROCEDURES, CONDITIONS. AND LIMITATIONS PAGE 121 A!iSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M ANSWER'i 8.01 ( .50) c)

REFERENCE CR TOR-5-40 objs 7 & 8 (3.7/4.1) 194001K102 ...(KA'S)

ANSWER 8.02 (1.00) c)

REFERENCE CR ROT-4-26 obj 6 (2.5/3.0) 034000G007 ...(KA'S)

ANSWER 8.03 (1.00) b)

REFERENCE CR TS p B2-2 Cycle 2 84 (2.9/3.8) 001006G006 ...(KA'S)

ANSWER 8.04 (1.00) d)

REFERENCE CR OSIM V-2 (3.6/3.7) 194001K101 ...(KA'S)

ANSWER 8.05 (1.00) d)

REFERENCE CR FP-203 p.18 (2.4/3.5) 034000G011 ...(KA'S)

1 8'. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 122 ANSWERS -- CRYSTAL RIVER -89/01/12-DEAN, W M ANSWER >* 8.06 (1.00) b)

REFERENCE CR TS 3.2.4, p 3/4.2-8 (3.7/4.1) 001050G005 ...(KA'S)

ANSWER 8.07 (1.00) d)

REFERENCE

CR TS 3.3.1 (3.4/4.3) 012000K005 ...(KA'S)

ANSWER 8.08 (1.00) f SOTA, MOC, NOS (.33 ea)

REFERENCE CR ROT-5-38 obj 3. (3.7/3.8) 001000G001 ...(KA*S)

ANSWER 8.09 (1.00) l Evacuate to the two mile radius all sectors.

Shelter to the five mile radius in affected sectors.

(0.5 ea.)

REFERENCE CR ROT-5-51 obj 7 (3.1/4.4) 193001A116 ...(KA*S)

I ANSWER 8.10 (1,50)

1. grapple disengaged

2. grapple full-up

3. equipment de-energized '

b o

. c a> 8 r ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE-123 ANSWERS - CRYSTAL RIVER -88/01/12-DEAN, W M REFERENCE CR ROT-4-26 obj 6 (2.8/3.3) 034000G014 ...(KA'S)

ANSWER 8.11 (1.00)

STS 3.0.5 REFERENCE CR ROT-5-1 obj 10

NOWER 8.12 (1.50)

1. Cable Spreading Room

2. ES Switchgear Room

3. Station Battery Room

4. CRD Equipment Room

5. Control Center (5 at .3 ea)

REFERENCE CR ROT 5-40 obj CP-118 1. (3.1/3.4) 086000G001 ...(KA'S)

ANSWER 8.13 (1.00)

STS 3.3.1.1(28)... That channel (RPS Channel A) must be placed in the tripped condition within one hour.(0.5)

Startup may proceed since 3.0.4 is not applicable. (0.5)

REFERENCE CR ROT 5-1 obj. 10 (3.3/3.5) 015000G010 ...(KA'S)

ANSWER 8.14 (1.00)

Perform an initial evaluation to determine is any STS or restrictions are applicable (0.33)

Determine notification requirements (0.33)

Provide reportability (0,33)

r_ _______ _ . _ _ _ _

8'. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 124

' ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M ;

I REFERENCE l CR ROT-5-40 obj 1 ANSWER 8.15 (1.00)

Plant Manager DNSO (0.5) t JMNMD mrem not to exceed 5(N-18) (0.5) afoo '

REFERENCE CR ROT-5-43 obj 5 y(MPWY 3.3/3.5) 194001K104 ... KA*S)

ANSWER 8.16 ( .50)

STS 3.1.1.1.a increase the SDM by an amount at least equal to the worth of that rod.

REFERENCE CR ROT-5_1 obj 11 (3.5/4.2) 001000A203 ...(KA'S)

ANSWER 8.17 (1.00)

1. Who ever holds the clearance is agreeable

2. Permission is obtained from the clearance authority (0.5 ea)

REFERENCE CR ROT-5-50 obj 4 (3.7/4.1) 194001K102 ...(KA*S)

ANSWER 8.18 ( .50)

An RCA is used to control access to Radiation Arear-REFERENCE CR ROT-5-43 obj 1 (3.3/3.5) 194001K104 ...(KA'S)

" 8I ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 125 ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M ANSWER / 8.19 (2.00) ,

4 3.6.3.1 Action (s) a, b, e or d...there are two pesibilities:

(1.0)

1. as long as actions a, b, er c can be complied with then startup may continue.(0.5)

2. If neither action a, b, or e can be met then the plant will be forced to enter Mode Ar in the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .(0,5)

J' 30 REFERENCE CR ROT 5-1 obj 10 & 11 (3.3/4.1) 103100G005 ...(KA*S)

ANSWER 8.20 (1,00) a) Site Area Emergency b) Alert (0.5 ea)

REFERENCE CR ROT 5051 obj 2 (3.1/4.4) 194001A116 ...(KA'S)

ANSWER 8.21 (1.50)

With: 3 rem /qtr whole body...not to exceed 5(N-18)

Without: 1.25 rem /qtr With or Without: 7 5 rem /qtr skin 18.75 rem /qtr extremities (0.3 ea)

REFERENCE CR ROT 5-43 obj 2 & 3 (2.8/3.4) 194001K102 ...(KA'S)

* (, ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 126 ANSWERS -- CRYSTAL. RIVER -88/01/12-DEAN, W H ANSWER / 8.22 (2.00)

1. When time prevents implementation by other procedure change methods.

2. They may be used for the addition and deletion of steps not normally addressed in the procedure for a specific period of time, or event.

3. IC's must be reviewed and approved by the same review and opproval cycle as required for new and permenanent revisions. Handwritten IC's may be used.

4. One (0.5 ea)

REFERENCE CR ROT-5-39 obj 2 ANSWER 8.23 (1.00)

1. Approval of analysis shortly before their release.

2. Only through paths that require positive manual operation in order to effect the release.

3, Only through charcoal and HEPA filters.

4.Throughapathjnwhichthegasismonitoredtwice.

f' *ap' q$ f k aaVg< corred rtJNefl.

REFERENCE CR ROT-5-48 obj 5 [(2.7/2.9) 071000G001 ...(KA*S)

/-0)

ANSW3R 8.24 ,

-STS Ga-fety Limit 2.1.

he& b _ > avceedad_4A54. Be 4n-WM Standby-d ho . . Requirements of STS 6.7.1 must be cet (reports) (04k5). 'nce violation was caused by a failure of the RPS varia e pressure / temperature trip, a prompt notification is re ired per STS 6.9.1.8.a.(0,5)

REFERENCE CR ROT 5-1 obj 7 000007G004 ...(KA*S)

" " 8i ADMINISTRATIVE PROCEDURES. CONDITIONS, AND LIMITATIONS PAGE 127 ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M ANSWER / 8.25 (1.50)

c. "Controlled"/"Working" copies of written procedures chall be stricly adhered to in all matters relating to nuclear safety. (0,5)

b. 1. If otherwise specifically authorized in the individual procedure, or as authorized in POQAM.

2. While performing routine actions that are frequently repeated.

3. If the conditions of the QC hold point cannot be catisfied in a timely manner, parallel work may proceed (with the concurrence of the Shop Supervisor).

(3 at 0.33 ea)

REFERENCE CR ROT 5-38 obj 14 ANSWER 8.26 (2.00)

1. Exposure of the whole body of any individual to 25 rems or more of radiation; skin 150 rems or more; feet, ankles, hands or forearms of any individual to 375 rems or more or radiation.

2. The release of radioactive material in concentration which, if averaged over a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , would exceed 5,000 times the limits specified in App. B, Table II.

3. A loss of one working week or more of the operation of any facilities affected.

4. Damage to property in excess of $200,000.

, (4 at .5 ea)

REFERENCE CR ROT 5-43 obj 4.

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oa

. g- .

\ ,

U. S. NUCLEAR REGULAT'ORY'C011 MISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: CRYSTAL RIVER j REACTOR TYPE: PWR-B&W177 DATE ADMINISTERED: 88/01/12 EXAMINER: DEAN. WM CANDIDATE: _

INSTRUCTIONS TO CANDIDATE:

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

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

% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY d '3 5

^ 3d _2iJilbi2 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. THERMODYNAMICS.

p HEAT TRANSFER AND FLUID FLOW

_2 ? E2 25.00 2. PLANT DESIGN INCLUDING SAFETY +

AND EMERGENCY SYSTEMS 30.00 _2LjiLQ 3. INSTRUMENTS AND CONTROLS

^M a

_rG.20 25.00 4. PROCEDURES - NOFMAL ABNORMAL, t EMERGENCY AND RADIOLOGICAL g, CONTROL >

107.0c % Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.

Candidate's Signature

4 l 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.

r

2. Resiroom trips are to be limited and only one candidate at a time may '

leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating. '

3. Use black ink or dark pencil only to facilitate legible reproduct' ions. ,

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 neces5aiy). '

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 agw page, write only an 2na pid9 of the paper, and write "Last Page" on the last answer sheet.

9. Number each answer as to categor/ and number, for example, 1.4, 6.S.

10. Skip at least three lines between each answer,

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

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

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

84. 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, ANSWEB 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.

27. 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 l completing the examination. This must be done after the examination has been completed.

l

b 4

18. When you complete your examination, you shall:

a. Assemble your examination as follows:

(1) Exam questions on top.

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

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

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

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

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

0 l

t i

l i

n e - -

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 2

,' THERMODYNAMICS. FEAT TRANSFER AND FLUID FLOW QUESTI N 1.01 (1.00)

Whici one of the following correctly describes the observed reactor response for the same small addition of reactivity, one positive and one negative?

a) The response will be faster for the negative addition at all times in core life.

b) The response will be faster for the nega?,1ve addition at BOC but faster for the positive addition at EOC.

c) The response will be faster for the positive addition at all times in core life, d) The response will be faster for the positive addition at BOC but faster for the negative addition at EOC.

e) The response will be the same for both the positive and negative addition.

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

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3

', THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.02 (1.00)

Which'one of the following would be the major problem associated with conducting a forced circulation cooldown with one OTSG dry and depressurized?

a) OTSG Tube-to-shell differential temperature b) Delta Tc c) Excessive subcooling Margin d) Pressurized thermal shock

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

1 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 4 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTIO,N 1.03 (1.00)

Referring to the attached page showing temperature va distance from the fuel centerline, which curve correctly represents this relationship for a typical coolant channel while at power?

t

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

I

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Le PRINCIPLES OF NUCLEAR _EQMER PLANT OPERATION. -

PAGE 5 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW i

QUESTION 1.04 (1.00)

Which one of the following statements correctly describes the effect of cdding Emergency Feedwater (EFW) during a Natural Circulation condition?

a) It LOWERS the OTSG thermal center while INCREASING the strength of the heat sink.

b) It LOWERS the OTSG thermal center while DECREASING the strength of the heat sink.

c) It RAISES the OTSG thermal center while INCREASING the strength of the heat sink.

d) It RAISES the OTSG thermal center while DECREASING the strength of the heat sink.

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

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 6

', THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 4

i QUESTION 1.05 (1.00)

Which'one of the curves on the attached page correctly shows the affect of recently installed "GREY" APSRs on the axial imbalance compared to the offect the old "BLACK" APSRs had on axial imbalance?

1 i

e

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

i

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A, - g3 .. --- -- _ _B p..._.-_._ .

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1 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 7 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTIO,N 1.06 (1.50)

Indicate whether the following will INCREASE, DECREASE or REMAIN THE SAME:

a) Available NPSH for a MFP as volumetric flow rate increases.

b) Minimum required RCP NPSH as volumetric flow rate increases, c) Available NPSH to condensate (hotwell) pumps as condenser subcooling increases, i

5

+

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

1. PhINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 8 THERMODYNAMICS. MEAT TRANSFER AND FLUID FLOW QUESTIO,H 1.07 (1.50)

With the Unit operating at 1004 power with all control cystems in automatic, a Turbine Bypass Valve fails full open. Indicate how the following parameters will change relative to their initial values when plant conditions stabilize: (INCREASE, DECREASE, REMAIN THE SAME) c) Tavg b) MWe c) Reactor power i

l J

4

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

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.

4 PAGE 9

, THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.08 (1.50) -

Indicate whether each of the following INCREASE, DECREASE or REMAIN THE SAME as the core ages: (answer in terms of change in magnitude) a) Fuel Temperature Coefficient of Reactivity <

b) Equilibrium Xenon Worth c) Initial Power Drop following a Reactor Trip at Power I

l l

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i 1

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

1 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE le THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW

.g.

QUESTION 1.09 (1.00)

Indicate if Beff is LARGER, SMALLER or the SAME as Beore. Explain your answer.

(***** CATEGORY 01 CONTItiUED ON t1 EXT PAGE *****)

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

. THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.10 (1.50) !

Indic' ate whether the following will cause the power range instrument to i indicate HIGHER, LOWER or the SAME aa actual power, if the instrument has been adjusted to 1004 based on a calculated heat balance: ,

r o) The feedwater temperature used in the heat balance was HIGHER than .

actual feedwater. l b) If the reactor coolant pump heat input used in the heat balance is OMITTED. t c) If the steam flow used in the heat balance was HIGHER than actual.

1 t

i

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

1. , PRINCIPLES'0F NUCLEAR POWER PLANT OPERATION, PAGE 12 THERMODYNAMICJ. HEA.T TRANSFER AND FLUID FLOW '

t t

QUESTIO,N 1.11 (2.00) :

Indicate whether each of the following will cause the differential rod ;

worth to INCE1ASE, DECREASE or have NO EFFECT. f c) An adjacent rod is withdrawn. !

b) Moderator' temperature is DECREASED.

t c) Boron concentration is INCREASED. !

d) A Burnable Poison Rod depletes. l t

i 1

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

i i

1. PRINCIPLES OF NUCi.2fR POWER PLART OPERATION. PAGE 13

'. THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.12 ( .50)

TRUE'OR FALSE: In e OTSG heat is transferred from a higher enthalpy fluid to a lower en alpy fluid.

b

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

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 14 '

THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW 4

r i

QUESTION 1.13 (1.50) i i

State three factors that contribute to the production of high boron concentrations in the core post LOCA.

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

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 15 !

', THEkMODYNAMICS, HEAT TRANSFER AND FLUID FLOW !

.'. i l

s QUESTION 1,14 (1,50) !

i Assuming that OTSG pressure is stable, what are three indications that the I operator can utilize to confirm the existence of naturs1 circulation? Give ;

numerical values where appropriate, i i

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

1

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION < PAGE 16 s ,< -

THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW l

.:>- t t

4 i QUESTION 1.15 (1.00) l a'

Attached is a figure showing the current plant heatup curves. ;

r c) In what' direction will the maximum limit (segment A-F) shift over the !

li'le of the plant? i b) Why does the curve shift?

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

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPER.\ TION, PAGE 17 -{

      =
           . THERMODYNAMICS. HEAT TRANSFER AND FLUIL FL0d                                    I 1

I I QUESTION 1.16 (1.00)

                 ~
           -There are two effects that cause differential boron worth to change over
          . core life. List these two effects and. indicate their relative impact on differential boron worth.

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

1 ~. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 18

      . THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION ,

1.17 (1.50) List 6 potential' sources of gas intrusion into the RCS if a significant LOCA were to occur. (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)

4 J

        - 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,                           PAGE               19
            ',  THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION     1.18         (1.50)

If the OTSG low level limit were decreased, indicate whether reactor power would have to be HIGHER, LOWER or the SAME to achieve a TAVG of 579 degrees. Explain your answer. f I (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)

            =                        - - - - ,                        - - , ,     - - - - - - . - , . _ ,
                    =  -

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 20

         . THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION   1.19            (2.00)

While operating at 60% power, it is recognized that a control rod has been eignificantly misaligned from its group average for several days.- The rod is realigned, and a positive quadrant tilt develops in the quadrant of the misaligned rod, a) Why does a positive tilt exist, even though all rods are correctly aligned? b) Assuming no operator action, how would this tilt change over the next ten hours? Explain your answer. 'l l l l

l l l l l

                        /.***** CATEGORY 01 CONTINUED ON NEXT PAGE  *****)

l

_1 . PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 21 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.20 (2.00) Refer to the attached figure 11-27, "Error Adjusted Rod Index Alarm Setpoint" to answer the following: a) Assume that you are at the indicated point outside of the permissible area of operations for "power level vs. rod index" and "axial power imbalance". List the three methods (indicated by Paths A, B and C on both curves) which are utilized to restore plant conditions within the region of permissible operations. (1.5) b) Path C for "axial power imbalance" was not successful in vacating the restricted region. What action is taken as indicated by Path D, which will restore axial imbalance to the permissible operating region? (0.5) i (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)

                                                                                   ~
 !                              FIGURE 11-27: ERROR-ADJUSTED ROD INDEX ALARM SETPOINTS 2:

i Power-imbalance Envelope for Operation I Error-Mjusted Rodindex Alarm Setpoints I Power Level Power (%) l (%) 100 - Shuttbwn - - {- -100 l A or . 90 t1ergin & Limit D

                                                                                                      'g      -                                                                                          C 80   -

Restricted > p Operation -

                                                                                                                  -70 Permissible 3y                         Operating           >      --60 60    -       Operatton                             g i                                                                                   Region Not Allowed                                                                                -
                                                                                                                  -50

Restricted i C Restrfcted Region

'l Region

                                                                                                                  -40 l      40    -

i

                                                                                                               -  -30 l

Permissible Operation I

                                                                                                                   -10 Rodindex (%)                             i     i        i              i    i     e i 0                          ,         i 150 200       250 i

300 40-30 10 0 10 20 30 40 50 0 50 100 Axia1 Power imbalance (%) Bank 5 , i i Bank 6 , Bant 7 , 11 - 41 ,

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 22

    . THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION  ,

1.21 (1.00) What are the two reasons that there is a 15 minute time period that must ba observed between RCP "bumps" (until Natural Circulation redevelops) when using RCP "bumps" to try and restore Natural Circulation, assuming it was lost during a LOCA condition? l I (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****) ;

  -1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,                  PAGE 23
   '. THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION     1.22        (1.00)

The pressurizer PORV is leaking by during operation at 85% power. Assuming a Quench Tank pressure of 20 psia and i aaturation-conditions in the pressurizer corresponding to 2243 psia, determine the quality of the steam on the downstream cide of the PORV utilizing STEAM TABLES. Show all work. i l l. l l l (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)

.-a

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 24

   . THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW 4

QUESTION 1.23 (1.50) Attached is a curve showing Keff vs. moderator-to-fuel ratio for 500 ppm Boron concentration. For the following, indicate where on the curve the applicable location would be located: (e.g. above and to the left) a) Overmoderated region b) Negative MTC region c) Direction optimum point shifts as Boron concentration decreases (***** END OF CATEGORY 01 *****)

l

1.6 OPTINUM YALUE

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g ...................... ...... .......... g 1

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                                                                              .                        .i 0.4                                                      i.

O.2 I I 2 3 4 5 T 8 7 8 T 1 10 2 Moderstar-To-Fuel Ratio ll 1 l l l l 1

2. PLANT' DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 25 4

    . QUESTION           2.01        '( 1. 00 )

Which one of the following is NOT a purpose for startin.g air during the start of an EGDG7 a) Provides air as a motive force to roll the engine, b) Suppliee air to the oil charged bearing booster. c) Supplies air to the governor booster. d) Provides air to the blower side of the turbocharger. t (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)

p. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 26 QUESTION 2. (1.00)

Whilejperform ng SP-354-B (B-EGDG surveillance) a loss of off-site power occurs along with an exciter field overcurrent. ich one of the following correctly describes the response of he Diesel Generator engine and output breaker? a) The outpu breaker will trip open and lockout, the ' engine shu down. b) The output b eaker will remain closed, the engine continues to perate. c) The output bre er will trip open, the engine shuts down and then r starts after the overcurrent condition clears The output breaker will then close. d) The output breaker ill trip open, the engine will continue to run, th output breaker will close after the overcurren conditions clears. g@6D ( * * * *

CATEGORY 02 CONTINUED ON NEXT l' AGE * * * * * )

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 27 0'

4 QUESTION 2.03 (1.00) WhicK/one of the following sets of conditions would directly

      -cause the EFIC system to initiate Emergency Feedwater?
        -(assume all controls are in auto) a)  A trip of one Main Feedwater pump at 90% power.

b) A low level (<6") in "A" OTSG with HPI actuation on both channels. c) While at power an operator depresses the "Trip 1" button at the local panel. d) An operator has placed the SG BYPASS / RESET switches for Channels A - D in the BYPASS (up) position with main steam pressure at 700 psig, and a OTSG low pressure (<600 psig) signal is subsequently received. (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 28 QUESTION 2.04 (1.00)

WhicWe'one of the following indications and/or automatic cetions of the NNI X system power supply monitor would indicate a loss of one +24 VDC and one -24 VDC power supply? a) Upper light in upper section would be out...no automatic actions. b) Both lights in upper section would be out...no automatic actions. c) Upper light in both sections will be out...after a time delay (.5 secs) the Si and S2 breakers will open, d) Both lights in upper section would be out...after a time delay (.5 secs) the Si and S2 breakers will open. - (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)

           ,2 . PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS                           PAGE 29 QUESTION            2.05          (1.00)

Which'/cne of the following is the most important reason why core"flood tank pressure is carefully controlled? a) During a large LOCA, CFT injection will occur immediately after "blow-out" of the water from the lower part of the core occurs, b) Pressure must be maintained to promote adequate "Nitrogen Sparging" (mixing) during CFT depres-surization. c) Pressure must be low enough to minimize CFT injection until immediately after "blow-out" of the water from the lower part of the core occurs. d) During CFT injection the decrease in CFT liquid volume will cause nitrogen injection into the ccre. (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)

 ,2 . PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS         PAGE 30 QUESTION     2.06        (1.00)

Which'ione of the following oil pumps has as a purpose to "supply oil to the speed increaser gears and bearings when the Makeup and Purification Pump is STARTED"? a) Main Gear Oil Pump b) Shaft Driven Gear Oil Pump c) Main Lube Oil Pump d) Shaft Driven Speed Increaser Oil Pump (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)

_2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 31 QUESTION 2.07 (1.00) With 'vegard to the plant ventilation systems, which one of the following systems is required for emergency operation? a) Reactor Cavity Fans b) Reactor Building Purge Supply System c) Decay Heat Closed Cycle Cooling Pumps Air Handling Units d) Reactor Building Operating Floor Fans (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)

l

   -2.   ' PLANT DESIGN INCLUDING SAFETY AND EMERGRNCY SYSTEMS          PAGE 32 n       -
                                                                              ^;

QUESTION .2.08 (1.00) Whic(one of the following is correct concerning the limits and precautions of the CA system? a) Of system components that contain a boric acid solution of 5% wt. or greater, only tanks need to be naintained at greater than (105 deg F). b) While draining the BAMT ensure the heaters automatically de-energize at the appropriate tank level. c) Hydrazine addition to the RCS during operation of makeup demineralizers prevents release of chlorides. d) Sufficient water shall be available to dilute a spill to a (100:1) ratio when handling NaOH, LiOH, or hydrazine. l i r l l l (***** CATEGORY 02 CONTINUED ON HEXT PAGE *****)

           ..                              =                _

f

2. PLANT DESIGN' INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 33 QUESTION -2.09 (2.00)

List' ell the interlocks which must be satisfied to start a Reactor Coolant Pump. i P l (***** CATEGORY 02 C0t1TINUED 011 NEXT PAGE *****)

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 34 QUESTION 2.10 ( .50)

Whic pcomponent of the Incore Instrumentation System has as ' its purpose 'to provide gamma compensation'? r, ' 1 i 4 i ; P I I i t l l 1 t 1 (***** CATEGORY 02 C0tiTIt10ED Oli 11 EXT PAGE *****) f i t

2. PLANT DESIGN INCLUDING SAbJTY AND EMERGENCY SYSTEMS PAGE 35 QUESTION 2.11 (1.50)

For each component listed below, state which system supplies its normal cooling water. a) Spent Fuel Coolers b) Decay Heat Removal _ Heat Exchangers c) RB Fan Assembly Cooling Coils d) Motor Driven Emergency Feedpump e) Makeup Pump 1B (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 36 QUESTION 2.12 (1.00)

The 5'd11owing actions associated with the EHC and OPC systems occur: The test solenoids on the reheat intercept vcives are energized, causing the intercept valves to close. After a time delay, the valves reopen and then reclose (repeating this process several times). State the most likely cause of these events. i (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)

2 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 37 QUESTION 2.13 (1.00) The operator has pl ced the Diamond Control Station in cutomatic in accorda ce with the operating procedure. A

control rod drive pr rammer senses rod motion without a corresponding command signal. State the response of the Diamond Control Static , with respect to its ability to rcspond to command sign 18, as a result of this condition.

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

2,f PLANT DESIGN INCLUDING SAFETY AND EMERGENCY-SYSTEMS PAGE 38 QUESTION 2.14 (1.00) Briefly describe the power supply ARRANGEMENT for the AC powered Radiation Monitoring System Panel, and the sampler < pump assemblies (exclude RMG 21 - 24 Waste Processing Area). Note: include voltage, number of phases, source (s) of power, and any backup supplies provided. r I ' 4 4 (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****) 1

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 39 7

4 QUESTION 2.15 (1.00) e) WWa't chemical is added to the Building Spray System? b) What is the basis for the addition of this chemical? i l (***** CATEGORY 02 ColiTINUED ON NEXT PAGE *****)

s - y n

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 40

                                         !,                                             .                                i
                                                                          /           $

4 /.

                                                                                   .z QUESTIOR                     2.16                        (2.00)
                                 ~

ConceTy$1ng the DC system, state the response of the ' following equipment to an automatic initiation signal from ES:

.7

1. RB fan assemblies.

2. Nuclear Services Emergency Sea Water Pumps.

3. Closed Cycle Cooling Pumps.

4. Normal Sea Water Pump,.

                                                          'l P

i/ i l l 1 t j t

i l (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****) !

n- ,

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE- 41

,e .' t ! , QUESTION- 2.17 _ ( .50) , l Which'/ level of Nuclear Instrumentation System provides an - 1 input"to the Reactor Diagnostic system? y. i L I L [ t r l t l 1 N l l J i r i (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****) ! i 1

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L ,_1$ ANT DEUff1NOLUDINC'JSAFETY AND JLH189JNCY SYSTEki, PAGE 42 m

                                                                             ,                                                                                                                                      ,                                  t s

1 s. QUESTION 2.18 (1.00) , '

                                     , . ,                                                                              ~ . .

Desce.tbe afl of the conditions-and/or steps (including cpplicable setpoint(s)F that are required to 'open. 'f4b-30 (crosstie.-oetween IA ancfSA) following a loss c f I Arl '

')

pressure. t > 3 t

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CATEGOR) 02 COtlTINUED ON NEXT PAGE * * *4 *; }

t

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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 43 0

QUESTION 2.19 (2.00) For the Emergency Feedwater Control Valves EFV-55, EFV-56, EFV-57, and EFV-58 provide the following information:

1. EXPLAIN the motive force used to operate the valves.

(0.5)

2. What system provides the controlling signal (s) for the valves? (0.5)

3. What is the basis for the "fill rate" limiting signal for valve operation? (1,0) l 1

l i l l (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****) u ~ ,- --

2_. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 44 QUESTION 2.20 'i.00)-(O.i) Wit' ihe Makeup System in service, a loss of either ICS or NNI power occurs. The operator attempts to close MUV-49 Letdown Isolation Valve and MUV-40 Letdown Cooler Outlet Valve. Is it possi le to close these valves with the loss ofICSorNNIpowergANOlo. Lhie eu avvive iste action to tekc given th aa vuudiidensS-(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 45 QUESTION 2.21 (1.00)

Assume a loss cf offsite power occurs coincident with a HPI Ectuation. Explain the feature of the HPI pump start celector switches which will prevent two HPI pumps from being loaded onto one diesel generator. a 'l 'I k l i (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****) 1 P

i

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 46 4

     ' QUESTION    2.22        (1.00)

What'two sets of conditions will result in an ES/UV Block (86 lockout) on an ES 4160VAC Bus? (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 47 -

          .'                                                                                                                                                                              I QUESTION                                    2.23                                            (1.00)

Assume the following conditions exist:

            "A" channel RCPPM                                                                     -
                                                                                                          "C" and "D" RCPs MW transducers have failed                                                        -

The reactor is at power, no other abnormal conditions exist. ' Explain why this condition would cause an automatic reactor trip. ] ! r t L e r J : i ) (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****) P

             - - - . ,       .--- -                                     ,o-            . . - - , .-    ,,,e      -%,         ,     ,,.                                      - e

   -2.                      PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS                                   PAGE 48-QUESTION                                                2.24        (1.00)

Statsythe purpose of the 120V AC Vital Bus Static Transfer Switches (VBXS). . (***** CATEGORY 02 CCNTINUED ON NEXT PAGE *****)

1 __FLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 49 QUESTION 2.25 (2.00) For the Absolute Position Indication instrumentation, answer the f611owing:

o. What ir he source of indication? (0.5)

b. List f (5) applications of the system output. (1.5)

Pote: ider multiple uses of the same indica .4/ interlock / control etc. as one application, i l l (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 50 QUESTION 2.26 (1.50) ;

        -NRC IE Information Notice 85-100 concerns the application of Rosemount dp transmitters used for measurement of RCS flow.

EXPLAIN the abnormality that may occur, including the conditions that may precipitate this, and the effect on the indication of flow. W l i t t t i (***** END OF CA"fGORY 02 *****) !

f

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

3. INSTRUMENTS AND CONTROLS PAGE 51 QUESTION 3.01 (1.00)

Which,one of the following is NOT automatically isolated / turned off if RMA-2 "Aux Building Ventilation Exhaust Duct" were to reach its alarm cotpoint? a) WDV-436, 437 and 438 (WGDT Outlet Isolation) b) AHV-1A, 1B, 10 and ID (Containment Purge supply / exhaust valves) c) AHF-30 (Chem Lab Supply Fan) d) AHF-10 (Fuel Handling Area Fan) e) AHF-9A and 9B (Penetration Cooling Fans) 1 (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)

3. INSTRUMENTS AND CONTROLS PAGE 52 QUESTION 3.02 (1.00)

Whic6/one of the statements below correctly describes the operation of the Load Control Valve (LLCV) and the Main Feedwater Block Valve (MBV) during a power decrease from 100% to 15%, assuming both valves are in Automatic? a) The MBV starts to close as Loop FW Demand drops below 80% and the LLCV starts to close when the MBV reaches the fully closed position. b) The MBV starts to close as Loop FW Demand drops below 50% and the LLCV starts to close when the MBV reaches the 80% open position. c) The MBV starts to close as Loop FW Demand drops below 45% and the LLCV starts to close when the MBV reaches the 50% open position. d) The MBV starts to close as Loop FW Demand drops below 45% and the LLCV starts to close when the MBV reaches the fully closed position. l (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)

3. INSTRUMENTS AND CONTROLS PAGE 53 QUESTION 3.03 (1.00)

Which/one of the following combinations of OTSG pressures will permit feedwater flow to BOTH OTSGa? A OTSG B OTSG a) 650 psig 550 psig b) 550 psig 400 psig c) 550 psig 450 psig d) 400 psig 575 psig (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)

3. INSTRUMENTS AND CONTROLS PAGE 54 l

QUESTION 3.04 (1.00) Which statement below correctly describes how the ICS determines the RCS flow signal for determining whether a runback is required? a) ICS~ calculates a flow signal based on RCP breaker status. b) ICS calculates a flow signal based on input from RCS loop flow D/P transmitters. c) ICS calculates a flow signal based on input from RCS loop flow tubes. d) ICS calculates a flow signal based on a selectable input using a patch cord from the RPS cabinet. '. i l i 1 't i 'f i (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****) l

                                       %eg.+

3. INSTRUMENTS AND CONTROLS PAGE 55 QUESTION 3.05 (1.00)

Which',one of the following is the set of conditions required to place the RCS pilot-actuated relief valve (RCV-10) into a low pressure overpressure code of operation? a) RCV-10 in MANUAL and the PORV setpoint key switch in NORM. b) RCV-10 in AUTO and the PORV setpoint keyswitch in NORM. c) RCV-10 in AUTO and the PORV setpoint keyewitch in LO. d) RCV-10 in AUTO, the PORV setpoint keyswitch in NORM and the LPI 500 psig trip bistables RESET. e) RCV-10 in MANUAL, the PORV setpoint keyewitch in LO, and the LPI 500 psig trip bistables RESET. i (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)

INSTRUMENTS AND CONTROLS PAGE 56 QUESTION 3.06 (1.00) Which'ione of the following is correct regarding the Main Turbine Overspeed Protection devices? a) There is a mechanical overspeed trip at 103% rated speed and an backup electrical trip at 111% rated speed. b) With the overspeed protection control switch in the "test" position, the electric overspeed device is bypassed, c) At approximately 103% rated speed, ONLY the governor and intercept valves will close. d) In the Overspeed "test" position, ONLY the reheat and intercept valves close. (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****) .t t

i l 3. INSTRUMENTS AND CONTROLS PAGE 57 l

1. -

t'- , j ! l QUESTION 3.07 ( .50) Indicate whether the following statement is TRUE or FALSE: l l If the HPI Safeguards Actuation is bypassed prior to pressure falling to 1500 psig, then if a subsequent LPI actuation were to take place, the HPI cystem would actuate. r f i I (+**** CATEGORY 03 CONTINUED ON NEXT PAGE *****) <

3. INSTRUMENTS AND CONTROLS PAGE 58

                                                                                                                                                                  *i QUESTION                 3.08                                 (1.00)                                                     j Fill'.'in the blanks to complete the following regarding degraded voltage                                              "

protection of the 4160 Volt ES System: The 4160 Volt ES System'a degraded voltage relays will actuate if a low voltage condition (< 3800 volts) Persists for greater than seconds. ' This must be sensed by of 3 degraded voltage relays. The : loss-of-voltage auxiliary relays will then actuate if the degraded voltage condition persists for a total of seconds. The ES bus will then be Otripped of all loads with the exception of . 4

I l

a i i i

,                                                                                                                                                                    l
!                                                                                                                                                                    r i

t-l (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****) '

3. INSTRUMENTS AND CONTROLS PAGE 59 .

QUESTION 3.09 (1.50) : For the OTSG Level Instruments listed below, indicate the number of channels available per OTSG, and indicate whether the instrument is temperature compensated or not. ; c) Operating Range b) EFIC High Range c) Full Range r l t i I (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****) l i : 1 J I i

i 1 INSTRUMENTS AND CONTROLS PAGE 60 a ,. QUESTION 3.10 (1.00) For th'e operating conditions listed below, indicate what error signal would be the controlling input into the Megawatt Calibrating Integral. Assume that all other controls are in Automatic, with the unit at power. a) SGRX in Manual b) Turbine not in ICS Auto I l l 1 I i F l i L I I 1 f

)'

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

l l

l t

'~

3. INSTRUMENTS AND CONTROLS 9 PAGE 61 QUESTION 3.11 (1.50) a) What two components will still actuate upon an ES signal even if i their controls are selected to the Remote Shutdown Panels (RSP)? (0.5) .

, b) If the relay status light on the "A" Side of the RSP is illuminated, and the one on the "B" Side is NOT illuminated, what is the status of the RSP? (1.0)

1 P

1 t i f r r 4 J 4 s a I v I 4 : I

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

e i I

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

3. INSTRUMENTS AND CONTROLS PAGE 62

                                              '^
                                                                                                                                      % i I

QUESTION 3.12 (1.00)

Indicate for each of the following situations whether the indicated breaker can be closed or if cross-tie blocking will prevent closure. Refer to the t
ettached drawing showing the 4160ES Distribution. '

4 a)- Breakers 3207, 3208 and 3209 are closed. An attempt to close 3210 is made. I b) Breakers 3205, 3208 and 3210 are closed. An attempt to close 3209 is made. (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)

E -

                   ?                       t.ujJ             ow2r s                                          t UJJ wmT 3                    .                p u-2r : -e t                 '
              ;                                              svaar-oe                                                   umzaanY          .                           swar- ve
 !            w gm                 ma<5+m92                                        (0P        Y***" 5** *                     -

n 1.m,,amaa 5 {

                                                                                        \                                                                               '

N < - i D3205 - 07 it 06 D3208 O D212 4160V ES BUS 3A 4160V ES BUS 3B 3209 J221 3210 p2O p222

] U O _J _J

;;                                                                                   LL JL3                                                     UL W ECOG 3A                                              C                                          EGOG30                      &m i                                                                                                                                                              O l                                                                       4BOV ES BUS Al                                                   l480V ES BUS B j PLANT AUX OUS3] ,,,

l 4160V ES DISTRIBUTION l E' ~ l l x 1- -__ _ ._ _ __ _. _ - - - _ __ - . . - _ _ . __

3. INSTRUMENTS AND CONTROLS PAGE 63 QUESTION 3.13 (1,50)

Assuming that Makeup Pump MUP A is running and MUP C is in STANDBY, what should be the status of the following MUP related controls to ensure that oil three MUPs are running following an ES Actuation? a) MUP B Power Supply b) Position of A-B & B-C Selector Switches c) Status of MUP B before the ES Actuation l 1 (***** CATEGOBY 03 CONTINUED ON NEXT PAGE *****)

).

INSTRUMENTS AND CONTROLS PAGE 64 QUESTION 3.14 (1.00) a) When the "Air Fail Reset" puenbutton for Pressurizer Level Control Valve MUV-31 is illuminated, in what position is the valve? b) If an operator wants to attempt to manipulate MUV-31 on a loss of air pressure to the valve, what must be done with the "Air Fail Reset" pushbutton? (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)

(

    - @.                                        INSTRUMENTS AND CONTROLS                                                                                                                                                                           PAGE 65 4
                                                                                                                                                                                                                                                         "{

QUESTION 3.15 (2.00) List'the four methods by which the Low Pressure CARDOX system to the Feedwater Pump Area may be actuated. Provide locations where applicable. .) J l i 1 1 1

i (***** CATEGORY 03 CONTIt3UED ON NEXT PAGE *****)

3. INSTRUMENTS AND CONTROLS -

PAGE 66 i QUESTION 3.16 (1.50) Asidn'ifrom the CRDMs, list 6 different loads which are isolated from the a Nucle'ar Services Closed Cycle Cooling System upon an ES Actuation. l l (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)

3. INSTRUMENTS'AND CONTROLS PAGE 67 *

                                                                                                              'i    ,

r J , QUESTION 3.17 (1.50) s 1 List the 2 valves monitored by the Decay Heat Removal System's Automatic f Closure Initiation (ACI), the 2 automatic acticne that occur if the ractuation setpoint is reached (assuming these valves are open) and how

this interlock may be' bypassed. '

1 i l l J ) J tn, l - i 1 i i l l i t (***** CATEGORY 03 CONTINUED ON HEXT PAGE *****)

c________-______________ ( ' e ', .

             ).* ?lNSTPUMENTS AND CONTROLS PAGE    68 l
                ?- (                                                                         ,e

_7_l QUESTION 3.18 (1.00) What't'wo Safeguards signals cause the Reactor Building Main Fans to shift to slow speed and what is the purpose of slowing the fans down in these i instances? -

                       .ir r
                                                                                                  ?

b h i I i I k i t i l t 1

       'l                                                                                          ;

i if

 'i '                                                                                               ;

1 (***** CATEGOBY 03 CONTINUED ON NEXT PAGE *****) ! I i

                                                                                     . .-       ,      . 1. ,          .

s,

3. INSTRUMENTS AND CONTfpId PAGE 69 3.19 QUESTION [1.50)

        . While7st 75% 1c.wer,.with all control systems in AUTO, the SGRX Master is .             .

placed in HASD. Describe the effect on the IC4 from increasing its output - to.80% Discuss only the systems affected by the control manipulation and. how the ICS compensates for the increased den.snd. i

i t 3

A s 1 (***** CATEGORY 03 CONTIliUED ON t4 EXT PAGE *4***) ,

b. ..

3. INSTRUMENTS AND CONTROLS PAGE 70 O

                                                                                 ; 1 QUESTION   3.20        (1.00)

An I'& C technician wants to remove the High Pressure Trip module from RPS Channel "A". He states that by placing the channel in bypass, the other channels will be prevented from receiving a trip signal from the "A" Channel. Explain why you AGREE or DISAGREE with the I & C tech. i l (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****) l

3. INSTRUMENTS AND CONTROLS PAGE 71 QUESTION 3.21. (2.00) a) Between what two setpoints should the EFIC systems "Low-Steam Pressure Stiutdown" be bypassed? (0.5) b) What two methods may be utilized to perform this bypass operation, and what are the indications for each method that the bypass took effect?

(1.5) l l f (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)

3, . INSTRUMENTS AND CONTROLS PAGE 72' t QUESTION 3.22 (1.00) What*Will happen to OTSG level control if the following situation took placi in the EFIC system?

1) An EFW actuation took place, with all RCPs secured.

2) The OTSG level setpoint was manually raised to 95% on the PSA/EFIC panel.

3) A RCP was then started.

(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****) 1

                       . ~ .     .            .   -.-   . - . -

3,. ' INSTRUMENTS AND CONTROLS PAGE 73

     -QUESTION      3.23            (1.00)

Describe the interlock associated with the EFW suction valves from the Main' Condenser. (***** CATEGORY 03 CONTItiUED ON NEXT PAGE *****)

3. INSTRUMENTS AND CONTROLS PAGE 74 QUESTION 3.24 (1.50)

WhilCiat 100% power, there is a failure HIGH of the lower detector of the Power Range NI feeding the A Channel of the RPS. Describe the effect of this failure on the RPS. In your answer, indicate what inputs to the RPS are affected and the final state of the RPS. (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)

(.

3. INSTRUMENTS AND CONTROLS PAGE 75

        . QUESTION                        3.25            (1.00)

Refer ~;to the attached logic drawing showing the Rod Withdrawal Inhibit Logic and sketch the missing portion of the drawing, showing the correct logic symbols. t

i l l 4 i (***** END OF CATEGORY 03 *****) ; E i

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

FIGUAE il RCD WITHORAWAL INHIBIT LOGIC sa sa in sa al. 21 3 3:0% 3 10 % N1 3 al.4 wt ny pa pe 2 2 0 Pts orp 2 t Dru 0F# pa pa 1 2 3 0PW 230M8 41 8 al.4 n! F NI 4 V V u - 1 1 I I

                                                                 .L _L 0
                                                                                 .~.

I

O ROS I M$

1 l l l I I

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 76 RADIOLOGICAL CONTROL QUESTION 4.01 (1.00)

Which one of the following is NOT a RCS Leakage Detection System required by Technical Specifications?

a. Containment atmospheric Iodine radioactivity monitoring

b. Nuclear Services Closed Cooling Water monitoring

c. Containment atmospheric gaseous radioactivity monitoring

d. Containment sump level monitoring l

l [ (***** CATEGORY 04 CONTINUED ON NEXT PAGS *****) '

d,, PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE 77 RADIOLOGICAL CONTROL QUESTION 4.02 (1.00) Consi er a situation where a Large LOCA has occurred, and you are operating the plant under AP-380, "Safeguards Actuation" and EP-290, "Inadequate Core Cooling". While performing a needed etep in EP-290, you find you are unable to complete-the action as called for in either the "ACTIONS" or the "DETAILS" column of the procedure. Which statement below correctly describes the appropriate action the operator should take, assuming additional methods NOT LISTED in EP-290 could be utilized to complete the required action?

a. The operator at the controls should take the appropriate mitigating actions, and then inform the NSS or ANSS.

b. The operator at the controls should await direction from the NSS or ANSS, and if none is forthcoming, continue with the procedure.

c. The operator at the controls should inform the NSS or ANSS of his alternative action for concurrence, prior to performance,

d. The operator should continue attempts to take the action as stated in the EP, while continuing on with the procedure.

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

O A. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 78 RADIOLOGICAL CONTROL QUESTION 4.03 (1.00) UtiliIingtheattachedexcerptfromtechspecs,determinewhichoneofthe following statements is correct concerning the Quadrant Power Tilt (QPT)?

a. If QPT exceeds the maximum limit of 20.0, the reactor must be shutdown immediately.

b. If misalignment of a control rod causes the QPT to exceed the transient limit, thermal power must be reduced in 30 minutes,

c. Action is NOT required for an hour, regardless of the QPT limit exceeded (transient, steady state or maximum).

d. If QPT exceeds the steady state limit, but remains less than the transient limit, operation may continue indefinitely as long as power does not exceed 60% allowable for the existing RCP combination.

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

4

        ; DOWER C:STRIBL' TION LIM:TE
          'DUADRAfC DOWER TILT l LIM!TINCCON::TI0t;FORODERATION 3.2.4 THE QUADRANT POWER TILT snali no; exceed the Steady State Limit of Table 3.2-2.

ADPLICAEILITY: MODE 1 above 15% of RATED THERMAL POWER.* ACTION:

a. Witn the OUADRANT POWER TILT determined to ex:eed the Steacy State Limit but less than or equal to the Transient Limit of Iable 3.2-2.

1 Within 2 hours: a) Either redu:e ne OUADRANT POWER TILT to within its Steacy State Limit, or b) Redu:e THERKAL POWIR so as not to ei.ceed THERKAL POJER, in:iuding power level cutoff, allowable for

ne rea: tor coolant Dumo combination less at least 25 for each it of OUADRANT POWER TILT in excess of

ne Steady State Limit and within 4 hours, reduce the hu: lear Overoowe- Trip Setooint and the Nu: lear Ove*oower Base: on RCS Flow and AXIAL DOWER IMBALANCE Trio Se:ocin: at least 2L for ea:n it of OUADRANT ,

POWER TILT in excess of tne Steady State Limit.

2. Veri #y :na: the OUADRANT POWER TILT is within its Steady S ate Limi: within 24 nours after ex:ee:ing the Steaoy State Limit or reouce THERMAL POWER to less than 60% of THERKAL POWER allowable for the reactor co0lant pump combination within the next 2 hours and redu e the Nuclear Overpower Trio Seto int to < 65.Ei of THERMAL POWER allowable for the reactor coolant pum: combination within

ne next 4 hours.

2. Icentify and corre: the cause of :ne out of limit con-dition orter :: increasing THERMAL POWER; subseouent DOWER ODERATION above 60% of THERKAL POWER allowable for the rea: tor coolant Dumo combination may pro:eed previoed :na the OUADRANT DOWER TILT is verified within its Steacy State Limit at least on:e per hour for 12 hou-s or until verifie: a::eo;atie at 9Et or greater RATE 0 THERMAL DOWiR.

        'See Soe: Tai les: Ex:eo:1on 3.10.1.

CRYSTA1 RIVER - UNIT 2 2/4 2-5 __

j

                                             -                                                                                                                        < \

q :n,geg s .r.c R I3U?. '.#.N '. .'M f. . 5

                 .,.                    LIMITING CONDITICN .:CR CPERAT!CN (Continuec'

b. With tne QUACRANT PCWER T!LT detarmined :o exceed the Transien:

Limi but less : nan :ne Maximum Limi: Of Table 3.2-2, due :: misalignmen; of eitner a safety, regulating or axial cwer shacing red:

1. Reduce THERMAL P0h'ER 4 : leas: 2% for each li e' indica:ad CUACRAN7 POWER TILT in excess of :ne 5:eady State Limit withir. 30 minutes.

2. Verify na: the CL'ACRANT PCWER TILT is witnin 13 Transien-Limit witnin 2 hetrs after exceeding the Transien: Limi-Or reduce THERMAL PCWER :: less : nan 50% of TMERMAL PCWER allewaole for :ne react:r c:elant :uma c:moination wi nin

ne next 2 hcurs anc reduce ne Nuclear Over:cwer Tri:

Se :oint t: < 55.5% of THERMAL PCWER allowaole for :ne i I i reac:Or c: clan: :umo e:mcinatien wi-hin ne next 1 icurs. l 3. Identify and c rrec the cause of :ne cu :f ifmit c:n - t l di:icn :rier :: increasing THERMAL CWER; su:secuen: I t PCWER CPERATICN acove 5C% :f THERMAL PCWER allewacle for l,

ne react:r c:clant :umo c:m:ina:icn ?.ay :recaec Or:vi:ec na: :ne OUACRANT ?CWER 7:LT is verified within i:s :eacy l N Sta:a Limit a: leas onca :er neur f:r 12 Meurs :r un : 1 veri

                                                                           #iec ac:a :aole a- 35% Or grea:ar RATID THERFAL
                                                                . e.ne R .

l dll :. Wi:n :ne ;UACRANT #CWER TILT ce:armired :: excaec :ne Transien-U Limi :u- less : nan :ne 'daximum imi Of Tacie 3.2-2, :ue :. causes ::ner : nan.:ne misalignmen: :# ei ner a sa#e y, regu'a - jl ing er axial Pcwer snacing rec: 1 Recuce THERMAL OCWER :: less nan 5C% of TFERMAL C%ER alicwacie f:r :ne reacter :: clan : cec ::m:inati:n 41:ni-l 2 neurs anc reduce tne Nuclear CVGFOcwer Iri: 3e*:Ci iI : <- 55.5% of :-ERMAL PCWER allewaai e ' r :. e reac :e il 4 c:Cian *um0 :Om:1na*icn 11*nin :ne nex - 9eurs.

                           ,' j

2. I:en*i#y and ::r sc: :*e causa *#

                                                                                                                               *9e
                                                                                                                                .      u* Of ' f 9i *. : n-l 0'f    :n
                                                                . r :.:-

r*0r- : iacreasdnc ~2ERMAL 'ChEA. su:sacuent :C'a !.:

                                                                         . s . . . .y 3 e .t . : v.

2 reac;:P :: clan

                                                                                             . . e         v .f.: v.a t :r. a .:.:.   ..-,nac;
                                                                                                                                         .     . e.r ...

Lc: ::.?:i a-*:n ay :r :sec :rev :ec :na-

                          ,l                                      ne UACRAN               .:C%ER 7:LT is eer'diec *i nin d s 5:aacy
                          ';                                    .im'; a- less: :nca :er acur ':r 12                                 curs r .n- ' fe 5:a          ed e

ec

                         ,                                      .. ..:::. c ,.. r a. a ..a r :.: ~. T..' . ' .:.:.9.a l :c.'a :.:.
                         ,f i
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                                      . .:.v. e. ; L :. . j :..:. . . . - .-

I

4

        , DOWEP. 0:5TE B'J TION LIMITS
             'i l.IM:T!NGCONDITION0R ODERATION (Continued)

ACTION: { Continued)

c. Wi:n the OUADRAKT POWER TILT determined to exceed ne Maximum Limit of Table 3.2-2, recu e THERMAL POWER to <~ 15% of RATED THERMAL POWER within 2 hours. '

SURVEILLANCE REOUIREMENTS 4.2.4. Tne OURDRAKT POWER TILT shall de oetermined to be witnin :ne limits at least on:e every 7 days ourine operation aoove 15% of RATEC THERMAL POWER exce:: wnen the OUADRANT POWER TILT monitor is inoperad3e,

nen :ne OUADRANT POWER TILT shall be calculate: at leas on:e per 12 nours.

l l l i t CP.YSTA'. RIVER - US 73 3/4 2-10 -

4 a 1 TABLE 3.2-2

           /                                     QUADRANT POWER TILT ',IMITS
             /

i STEADY STATE TRANSIENT MAXIMUM ! LIMIT LIMIT LIMIT QUADRANT POWER TILT as Measured by: Symmetrical Incore Detector System 3.20 9.08 20.0 Power Range Channels 1.61 6.96 20.0 Minimum incore Detector System 1.73 4.40 20.0 CRYSTAL RIVER - UNIT 3 3/4 2-11 Amendments Nos J$, J), 32,77

     . .                                                                              PAGE   79 9

ABNORMAL. EMERGENCY AND

4. PROCEDURES - NQEMAL.

RADIOLOGICAL CONTROL , 4.04 (1.00) ith saturated conditions existing QUESTION Assuming a unis lable leak in the f the following RCS, w statements correctlya and the RCPs sec red, which one water odescribes uncover, levels? how cha g the status of the core and downcomerill l tively stable.SLOWLY DE

a. SR counts assuming d wncomer level remains re a uncovers, assuming ble.

b. SR counts wi 1 RAPIDLY downcomer le el remains relatively sta DECREASE as the c fills, wil SLOWLY INCREASE as the downcomer i s relatively stable. regio

c. SR counts assuming the 1 vel in the core rema n depletes,

d. SR counts will LOWLY INCREASE as the downc assuming the lev downcomer region refills, SR counts will RA IDLY DECREASE as theins relatively sta

e. assuming the level in the core rema l

D@ l I l

                                                                                 +****)

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

4. PROCEDURES - NORMAL, ABNORMAL. EMERGENCY AND PAGE 80

       . RADIOLOGICAL CONTROL v

QUESTION 4.05 (1.00) In ac~cordance with OP-204, "Power Operations", which one of the following is considered the best method for dampening a Xenon Oscillation?

a. Determine the period of the oscillation as soon as possible, and then utilize the APSRs and Boration/ dilution as necessary several hours BEFORE the peak deviation to achieve an average axial power imbalance.

b. Determine the period of the oscillation within 12 hours, then use boration/ dilution, es appropriate, several hours AFTER the peak deviation to achieve an average axial power imbalance.

c. Determine the period of the oscillation over several days, and once peak deviation occurs so that a POSITIVE axial imbalance exists, drive control rods inward to reduce power 10-15%.

d. Determine the period of the oscillation over several days, then make the appropriate rod position correction several hours BEFORE the peak deviation to achieve an average axial power imbalance.

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

4. ~ PROCEDURES -' NORMAL. ABNORMAL. EMERGENCY AND PAGE 81-RADIOLOGICAL CONTROL

 -QUESTION    4.06        (1.50)

For e ch of the following, indicate whether that condition alone would be cufficient grounds to initiate a reactor trip, as listed in the entry conditions of AP-580, "Reactor Trip": a) 1 MSIV on each loop (MSV-411 and MSV-414) have been inadvertently shut. b) Both Main Feed Pumps are lost with reactor powe" at 15%. c) A Turbine Trip from 40% power occurs. (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)

4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND

         *.                                                                          PAGE. 82
     .       RADIOLOGICAL CONTROL QUESTION 4.07        (1.00)

Fill"in the blanks: a) Thermoluminescent dosimeters should be re-seroed prior to reaching a maximum of % full scale, b) The background reading on a frisker used for whole body monitoring

s should be no more than CPM.

i

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

                                                                       .-- , ,     -       r-,
                            -                    . _ . . -      e

4. '

PROCEDURES - NORMAL ABNORMAL. EMERGENCY AND PAGE 83 RADIOLOGICAL CONTROL QUESTION 4.08 (1.00) c) WdenperforminganRCP"bump" in an effort to restore Natural Circulation, why is OTSG Pressure reduced so that OTSG Tsat is 40-60 degrees below Incore Temperatures, prior to starting an RCP? b) What two methods / indications can be utilized by the operator to determine if RCS pressure is high enough to allow a RCP "bump"? f L [

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

     . . . _ _ _    _ , _         =_  _.,       . _ _ _ .
                                                                                                     -_ __ - ,_ ---, J

4.

   "     PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND

PAGE 84 RADIOLOGICAL CONTROL QUESTION 4.09 (1.50)

          *i While' performing a shutdown in order to conduct a refueling, is it permissible to begin detensioning the reactor vessel head bolts without having audible counts of the source range in the control room, as long as this audible indication is actuated before removal of the head begins?

Explain your answer. (***** CATEGORY 04 ColiTINUED ON NEXT PAGE *****)

4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 85

         **  RADIOLOGICAL CONTROL          ,

l . l.

        -QUESTION  4.10         (1.50)

List the Immediate Actions for AP-460, "Steam Generator Isolation Actuation". The remedial actions are NOT required. l (*****' CATEGORY 04 CONTINUED ON NEXT PAGE *****)

        ~L
        ~ '.

PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE 86-RADIOLOGICAL CONTROL i QUESTION 4.11 (1.00) What'two individuals, by title, can~ authorize an Inplant Clearance? i i

i

i -

.i ,l . )' . l 1 i-(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****) .

4. FROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND

    '                                                                   PAGE 87
..     .~ RADIOLOGICAL CONTROL d

QUESTION 4.-12 (1.00) List four areas-which require authorization from the Nuclear Fire Protection Specialist on a Fire Work Permit. (***** CATEGORY 04 C0t1TItiUED ON NEXT PAGE *****)

   , 4. PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND                     PAGE  88 RADIOLOGICAL CONTROL QUESTION 4.13        (1.50) c)  In accordance with AI-500, "Conduct of Operations", what are the 4 areas which if affected by a procedure, require that the procedure be done in a step-by-step manner?                                     (1.0) b)  What is meant by conducting a procedure in a "step-by-step" manner?

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

4. PROCEDURES'--NORMAL. ABNORMAL. EMERGENCY AND PAGE 89 RADIOLOGICAL CONTROL QUESTION 4.14 (1.50)

l In accordance with AP-580, "Reactor Trip", what are the 3 actions required ; to initiate emergency boration? 1 l r t [ f i 1 f I J h

i ;

(***** CATEGORY 04 CotiTItiUED CritiEXT PAGE *****) r

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

l

                  .4.    '

PROCEDURES - NORMAL. ABNORMAL. EMEFGENCY AND "1E 90

                           . RADIOLOGICAL CONTEQL                                           ,
                                                                         .,                  ?

J' 4.15 (1.00) QUESTION , List the four Critical Safety Functions in the order of-priority, as conitored in Verification Procedure VP-580. . .

                                                                                                                                         .y
                                                                                 .)
                                                                       >    .d                      <-
                                                                     '                 , ,r I                                                      [ .  s 3

4 6

(***** CATEGORY 04 CONTINUED Oil NEXT PAGE +4*++)

                                                                                                       /
 - ~ . . . . . . . . . .                                                                                                  . .. ...

                    ' 4 '. PROCEDURES -' NORMAL.' ABNORMAL, EMERGENCY AND
    'f                '*                                                                                                                                                                                     PAGE' 91 RADIOLOGICAL CONTROL

[* i.

'y QUESTION                                           4.16          (1.00)                                                                            ,

What supplyihree to the cenditions CRDs? must be. net in order to secure Nuclear Service Water 4 s 1 e L t i 1 . I

                                                                                                                     )          /                                              >                                               h f                                                                                                          \           .

I s k i t i. 5 i, I

r

i t

i (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****) I l i

                                ,y-                       .-                        . .-                                        _

p -

                                 .o 4 .,       h50dN00BES-NORMAL, ABNORMAL,EMERGENCYAND PkGE   92
      ..                       RA1}IULOGICAL COtiIEQL
                                                       ,,                        .s Wg%

QUESTION , 4.17 (1.75) .g,

                     'o )_      Assuming reactor power is U h , 0 hat are the 3 parameterr,, which if                                         '

y. exceeded require entry into AP-060, "

Turbine Trip"? (.75) b) While peYforming the IMMEDIATE ACTIONS of AP-660, the Governor Valvea fail to close and the generator output breaker fails to open. What are the appropriate REMEDIAL ACTIONS? (1.0) T

                                            '.         a.

I t 4 71,. L (***** CATEGORY 04 CONTINUED ON NEXT PAGC *****) l-

4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND '

                                                                                                                            PAGE  93 RADIOLOGICAL CONTROL t

QUESTION 4.18 (1.00) A continuing. problem with the EFIC system is LED failures causing 1/2 trip [ oignals. If such a condition were to occur, necessitating maintenance on , the EFIC system while at power, what system manipulation is done before- , work is commenced, and what administrative action is required while naintenance is ongoing? t (***** CATEGONY 04 CONTINUED ON NEXT PAGE *****)

4.

PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE -94 RADIOLOGICAL CONTROL QUESTION. 4.19 (1.75)

In accordance with AP-380, "Engineered Safeguards Actuation", what are the HPI throttling criteria? L l

h i . L

l i (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****) i. l [ i l l

4. PROCEDURES - NORMAL.~ ABNORMAL. EMERGENCY AND PAGE 95 RADIOLOGICAL CONTROL i-e.

               -QUESTION                       4.20                            (1.50)

Indic$tewhatthestatusofthefollowingcomponents/systemsshouldbein order.to perform OP-208, "Plant Shutdown" (20% to 0%): a) OTSG Level b) Feedwater System Pump Status c) RCS Pressure and Temperature b t i { i I r i e l l (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****) i i L l 4 +'---7---T---v-e r-- -e-se'--iv--------eg-- a+-- -

                                                                                                       ----m---+m--

1 j 9

4. PROCEDURES - NORMAL, ABNORMAL. EMERGENCY AND PAGE 96 RADIOLOGICAL CONTROL 1

QUESTION _4.21 (1.00) i

               '. . I

, According to the ATOG Guidelines, what is the major concern of operating 1 the HPI system excessively during a Steam Generator Tube Rupture? t i 4 J i (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)

l

4. PROCEDURES - NORMAL. ABNORhAL, EMERGENCY AND PAGE 97 RADIOLOGICAL CONTROL t

QUESTION 4.22 (2.00) a) What is the basis for establishing a OTSG level of 65% while performing' AP-530, "Natural Circulation"? b) What is the basis behind raising OTSG 1evel to 954 if subcooling margin degenerates while performing AP-5307 i 4 i 5 j t

,                                                                                                                                                     i i                                                                                                                                                      i I

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

4 4_. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND- PAGE 98 RADIOLOGICAL CONTROL QUESTION 4.23 ( .50) Explain the relationship between an RCA and a Radiation Area. (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)

4. PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND

       *.                                                                                                              PAGE 99 RADIOLOGICAL CONTROL
      -QUESTION  4.24         (1.00)
              ^

While performing OP-209, "Plant Cooldown", step 6.2.13 (see attached) has the operator refer to TS 3.7.1.2, EFW System. What is the reason for referring to this TS7 i (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)

                                                                                                                .                                         Initials           !

6.2.12 when secondary steam pressure has decreased to about 250 psig, stop MFWP per OP-605, Feedwater System. Use

~ ~

the FW booster pump for FW supply. ; NOTE: If coming to refueling mode, perform SP-418, Main Feedwater Pump Trip Test.

                .,                                                        ,6.2.13   When secondary steam pressure is < 200 psig, enter STS action of 3.7.1.2.                                                               l53 ;

6.2.14 Between 900 and 600 psig RCS pressure, bypass LPI ES channels 'A' and 'B' RC-4, RC-5, and RC-6. 6.2.14.1 If the plant is expected to be in COLD SHUTDOWN for 72

                                                                        %-          hrs or more, contact the Inservice Inspection (ISI)

Specialist to ascertain whether SP-405, Core Flooding

Systen Check Valve Operation Demonstration and Leak Testing, Part 'B', and SP-603, Decay Heat Check Valve Leak Testing, need to be performed. -

6.2.14.2 If coming to refueling mode, perform SP-603, Decay Heat Check Valve Leak Testing; SP-402, Core Flooding System Isolation Valves ' Alarms Actuation; and SP-405, l 1 I I l i 0P-209 Rev.60 Pase 12 n - . r.- z ,- ,m..., _,. m . , , . , . , 6 .. . .- , , - . . < . ,.- 4. - L

4. PROCEDURES - NORMAL ABNORMAL, EMERGENCY AND

    --                                                                   PAGE 100 RADIOLOGICAL CONTROL         ,

a QUESTION 4.25 (1.00)

           *;                                                                       t What'are the two reasons for establishing an OTSG level of 97-99% operate range, when performing a plant heatup? (See attached precaution from          1 OP-202) t i

r i I i (***** END OF CATEGORY 04 *****) (************* END OF EXAMINATION ***************)

O . 4.1.10 Prior to placing a aakeup desineralizer in service after having used hydrazine for oxygen scavenging, assure that chemistry - analysis of the RC showed hydrazine concentratio. as 'less than

        ...                detectable'.

4.2 enctoR cent LIMIYs inn PRtcAtTTIONs FOR RFATUP Control Rod Safety Groups 1 thru 4 will be fully withdrawn during deboration. 4.3 STI1N GENERATOR LIMITS AND PRICAUTIONS FOR HEATUP

                                                                                           ' 4 .,

4.3.1 The secondary side of the steam generator (s) shall not be pres-surized above 237 psig if the cesperature of the steaa genera- - ter vessel shell is below 110'T (STS 3.7.2.1). 4.3.2 The ainlaus steam generator operating level is 18 inches (STS 3.4.5). 4.3.3 (deleted) 4.3.4 The nazimum steam generator heatup rate is 100*T/hr. 4.3.5 Cycle cleanup shall be cesplete before feeding once-through

steam generator (OTSG).

     }       4.3.6       The 0750 main feedwater nozzles should be flooded by maintain-ing level at 97-99\ on the operate range until RC systen tem-
                      , perature is greater than 190*T.         The OTSG 1evel should be lowered to 1 350 in, prior to entering Mode 4.

1 - O OP-202 Rev.l75'

                                                                                      "       4

{

1. PRINCIPLES OF-NUCLEAR POWER PLANT OPERATION. -PAGE 101

[- THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW y ' ,_

       . ANSWERS -- CRYSTAL RIVER             -88/01/12-DEAN, W M ANSWER       1.01        (1.00) c REFERENCE CR ROT 1-6, obj 8, 9 K/A (3.0/3.0) 192003K107      ...(KA'S)

ANSWER 1.02 (1.00) o REFERENCE CR ROT 3-3, obj 12 K/A (3.7/3.8) 000040K106 ...(KA'S) ANSWER 1.03 (1.00) b REFERENCE CR ROT 3-2, obj 7 K/A (2.4/2.6) 193008K116 ...(KA'S) ANSWER 1.04 (1.00) c REFERENCE CR ROT 3-3, obj 6, 7 K/A (3.9/4.1) 193008K123 ...(KA'S)

. , . . 1.

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 102 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M ANSWER 1.05 (1.00) d REFERENCE CR ROT 2-10, obj 5 K/A (3.2/3.5) 192005K114 ...(KA'S) ANSWER 1.06 (1.50) i c) Decreases (+.5 ea) b) Increases I c) Increases REFERENCE DPC Thermodynamics / Fluid Flow, CH 2E i CR ROT 2-8, OBJ 17 ' (2.6/2.8) 191004K115 ...(KA'S) ANSWER 1.07 (1.50) c) Decrease (+.5 ea) b) Decrease c) Increase REFERENCE OP-0C-TA-NT pp 7/8; LO lb CR ROT 3-3, obj 2 (3.6/3.9) 041020A202 (KA'S) ANSWER 1.08 (1.50) c) Increase (+.5 ea) b) Increase c) Increase REFERENCE CR ROT 1-8, obj 3; ROT 1-10, obj 10; ROT 1-5, obj 3, 9, 12; K/A (2.9/2.9; 2.6/2.8; 2.9/3.1)

1. ' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 103 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W H 192004K107 192008K122 192008K123 ...(KA*S)

                 *i ANSWER        -1.09          (1.00)

Smaller (+.5) Since delayed neutrons are born at lower energies, they cause less fast fissioning (+.5) REFERENCE CR ROT 1-5, obj 11 K/A (2.4/2.4) 192003K104 ...(KA*S) ANSWER 1.10 (1.50) a) Lower (+.5 ea) b) Higher c) Higher REFERENCE CR ROT 5-7, obj 8 K/A (3.5/3.8) 015000A101 ...(KA'S) ANSWER 1.11 (2.00) e) Increase (+.5 ea) b) Decrease c) Decrease d) Increase REFERENCE CR ROT 1-7, obj 15 K/A (2.5/2.8) 192005K107 ...(KA*S) ANS R 1.12 ( .50) false 5) REFERENCE CR ROT 2-2, obj 8 K/A (3.3/3.4) 193003K125 ...(KA'S) I

1. PRINCIPLES OF NUCLEAR POWER-PLANT OPERATION. PAGE 104 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M i

ANSWER 1.13 (1.50) i

1. High injection water boron concentration

2. boiling off of RCS coolant in the core due to decay heat

3. Low thru core flow rate REFERENCE Oconee OP-0C-SPS-PTR-AM-1 pp. 37 obj 19 CR ROT 3-6, obj 1 l ANSWER 1.14 (1.50)

1) Tc:Tsat for the OTSG (+.5 ea)

2) Delta T develops / stabilizes at 45-55 degrees

3) Average of 5 highest thermocouples follows Th within 10 degrees REFERENCE CR ROT 3-3, obj 10 K/A (4.2/4.2) 193008K122 ...(KA'S) l ANSWER 1.15 (1.00) l a) To the right (+.5 ea) b) Neutron embrittlement causes the NDTT to increase REFSRENCE CR ROT 4-34, obj 6 K/A (2.9/3.0) 193010K105 ...(KA'S)

ANSWER 1.16 o v y' 7 3 t(t <-1.00)

                                   /' e Lessboron[requiredduetofuelburnout,

1) this increasee the boron worth due to less competition. (or flux hardening is decreased, so higher effective boron absorption cross-section) (+.5 ea)

2) Fission pr9ducts poisons build ups,^(decreasing the boron worth.

      $       <%J Hg 3c,              -5e 5-      <*'s-y          40 m    U 'W 4 REFE     CE                                                               -

CR ROT 1-7, obj 19 ' K/A (2.8/2.9) 192004K109 ...(KA*S)

if__ PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 105. THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M ANSWER' 1.17 (1.50)

         '1)                   Radiolytic decomposition (+.25 ea for any 6)

2) CFTs-

3) Decreased ability to retain dissolved gases

4) Fuel Cladding degradation

5) Steam formation

6) BWST via ECCS/ Makeup aystems

7) Pressurizer REFERENCE CR ROT 3-5, obj 1 K/A (4 6/4.8; 4.3/4.4) 000074K102 002000A201 ...(KA'S)

ANSWER 1.18 (1.50) Lower (+ 5) due to the heat transfer area decreasing, the' Delta T across th OTSG must increase to achieve the same power. (+1.0). REFERENCE i CR ROT 2-5, obj 10/11 K/A (3.8/4.0) 035010K109 ...(KA*S) ANSWER 1.19 (2.00) a) The lower level of xenon in that quadrant results in a higher power production (+.5) b) Tilt will increase (+.5)due to rapid burnout of the existing xenon in that quadrant (+.5), and then decrease as xenon concentration increases (+.5) REFERENCE CR ROT 3-7, obj 5; ROT 1-10, obj 9 K/A (3.2/3.6) 000005K103 ...(KA'S)

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION2 PAGE 106 l THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW f
ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M i i ;

l r ANSWER 1.20 (2.00) , c) A: Borate w/ ICS in AUTO (+.5 ea) B: Borate w/ rods in Manual C: Reduce ULD (ICS in AUT0) - b) D: Insert Group 8 REFERENCE , CR ROT 2-11, obj 6 K/A (3.6/4.0) i 001050A206 ...(KA*S) ANSWER 1.21 (1.00)

1) Limit Liquid flow out the break (+.5 ea)

2) Gives operator time to observe Natl Cire develop REFERENCE CR ROT 3-5, pp 21 K/A (3.4/3.8) 000017A211 ...(KA*S)

ANSWER 1.22 (1.00) at 2240 psia, hg 1115 BTU /lb (+.5) et 20 psia, at saturation conditions, hg : 1156 BTU /lb and hf : 196 BTU /lb calculate: (1156-1115)/(1156-196) : .043 >>> 95.7% quality REFERENCE OP-GA-SPS-THF-STM pp 20/21; LO 2e CR ROT 2-2. obj 9 (3.3/3.4) 193003K125 ...(KA'S) l 4

1. ' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 107 i THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M i t

t

                  */                                                                                     ;

ANSWER' -1.23 (1.50) l c) Right of the optimum point (+.5 ea) ! b) Left of the optimum point : c) Above and to the right REFERENCE CR ROT 1-8, obj 10; CR ROT 1-9, obj 4 K/A'(2.9/2.9) 192004K107 ...(KA*S) ' i I t I I i I I i F t i I i 4 i i l r I

t i

I L

              -                                                                                                                                       [
      ..                                                                                                                                              r

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 108 l l'

ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M ANSWERi 2.01 (1.00) d) REFERENCE CR ROT 4-6 obj 1 (3.4/3.9) 064000K105 ...(KA*S) < ANSW 2.02 (1.00) ' d) REFERENC ; CR ROT 4- obj 9 (4.4/4.6; 2.6/3.0; 3.3/3.7) 000055A214 064000A216 064000K409 ...(KA*S) i ANSWER 2.03 (1.00) b) REFERENCE I CR ROT-4-15 obj 9 (3.3/3.7) < 061000K402 ...(KA*S) ! l ! ANSWER 2.04 (1.00) b) REFERENCE CR ROT 4-9 obj 6 l (2.9/3.2) 016000A202 ...(KA*S) l ANSWER 2.05 (1.00) i f a) o g_ C[) REFERENCE CR ROT 4-01 (3.4/3.9) ( 006000K602 ...(KA'S) , t

  !                                                                                                                                                   I f

t _ _ _ _ _ . ._ __-I

F .

l. .

2.' PLANT DESIGN INCLUDING SAFETY ~AND EMERGENCY SYSTEMS PAGE 1C9 ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M ANSWER;- 2.06- (1.00) a)

, REFERENCE CR ANO-82 (2.5/2.6) 004020K618 ...(KA'S) ANSWER 2.07 (1.00) c) REFERENCE CR ROT 4-8 obj 3 (3.3/3.4) 005G000007 ...(KA'S) ANSWER 2.08 (1.00) d) REFERENCE CR ROT 4-2 obj 4 (3.1/3.4) 004000G10 ...(KA*S) l i ANSWER 2.09 (2.00) l l 1. Lift oil pressure 5. Seal flow

2. Seal cooling water flow (SW) 6. Power level < 30%

3. Upper oil pot level 7. RCS temp > 500 F

4. Lower oil pot level or t

                                                                    < 3 BCPs running control bleed off open

8. Motor cooling water flow (0.25 ea.)

REFERENCE CR ROT 4-1 obj 4 (2.6/2.9) 003000K614 ...(KA*S) l

12. PLANT DESIGN INCLUDING 6AFETY AND EMERGENCY SYSTEMS PAGE 110 ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W H ANSWER,- 2.10 ( .50)

The background detector REFERENCE CR ROT 4-11 obj 2 (2.6/2.9) 015000K602 ...(KA*S) ANSWER 2.11 (1.50) o) SW b) DC c) CI d) SW 09 setr-CM( D o) SW (0.3 ea) REFERENCE CR ROT-4-2 chap 1 obj 2; chap 2 obj 2: (2.1/2.2; 3.4/3.4) 076000K118 076000K120 ...(KA*S) ANSWER 2.12 (1.00) A partial load rejection has occurred. REFERENCE CR ROT 4-22 obj 8 (2.4/2.5; 3.1/2.9) 045000A401 045000K407 ..(KA'S)

ANSWER 2.13 (1.00) The Diamon Control Station will automatically revert to manual contr 1. 3 REFERENCE CR ROT 4-28 obj 8 & 12 (3.7/3.9; 3.4/3.8) 001050A201 0 (050K401 ...(KA*S)

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 111 ANSWERS -- CRYSTAL-RIVER -88/01/12-DEAN, W H ANSWER,. 2.14 (1.00)

Components - 120 v AC, 1 phase,-from the battery backed inverter fed vital buses. The sampler pump assemblies receive power from 480 v AC, 3 phase ESF MCCs. [0.5 ea.] REFERENCE CR ROT 4-25 obj 3 (2.3/2.7) 073000K201 ...(KA*S) ANSWER 2.15 (1.00) o) NaOH b) Minimize the fission product Iodine in the RB atmosphere Raising the Ph level of the spray water (0.5 ea.) REFERENCE CR ROT-4-1 obj 1 (2.0/2.7; 2,8/3.2) 026020KK50 026020K401 ...(KA*S) ANSWER 2.16 (2.00)

1. Transfers the fans from the industrial cooler to the SW system. o(L (/W L 5Hvfr W W SfT 4D

2. Starts the pumps.

3. Starts the pumps.

4. Tripe the pump.

(0.5 ea.) REFERENCE CB ROT-4-2 obj. 7 (4.1/4.4) 013000K103 ...(KA'S) ANSWER 2.17 ( .50) Power Range

 . ,                                                                                              )

2.' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 112 ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W H REFERENCE CR ROT 4-10 obj 3 015000K100 ...(KA*S) ,

ANSWER 2.18 (1.00)'

With IA header pressure greater than 80 psi (0.25), turn the local selector switch to the closed position and then return to the auto position (0.75). REFERENCE CR ROT-4-5 obj 3 (IA) (3.2/3.5) ! 078000K402 ...(KA*S) ' t ANSWER 2.19 (2.00) *

1. The valves are modulating solenoid valves that utilize an increasing DC current opposing epring pressure to modulate i the valve (0,5)

2. EFIC (0.5) '

3. This is done to minimize overcooling of the' reactor coolant system when EFW is initiated.(1.0) y REFERENCE CR ROT 4-15 obj.6 (2.7/2.7) 061000K411 ...(KA*S)

ANSWER 2.20 pM-h f The manual operation of these valves will not be affected by [ the loss of instrument power. (0.5 ea.) Yes thic le a proper-eotiern-- hg . REFERENCE CR ANO-82 obj. 11 i a f i I , i 5

            ----      -     -,-,v ---,- ,                  v-  - , - - - - ,   ,n    --- --n-    - - - , - - , - - -

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 113 !

ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M i i i 8 ANSWER f 2.21 (1.00) The start selector switch prevents a selected pump from 4 tripping on an undervoltage condition and ensures that the non-selected pump does trip when an undervoltage condition oxists on the 4160 ES bus. (1.0)

         -or-When a pump is selected, a contact in series with the contact that is closed by the undervoltage relay is opened.

This prevents the trip signal from reaching the breaker tripping device. Conversely, when a pump is not selected, the contact is closed and the pump breaker will be tripped if an undervoltage condition occurs on the bus. (1.0) REFERENCE CR 4-1 p. 90 obj chap. 2-2a. (4.3/4.4) 006000K405 ...(KA*S) ANSWER 2.22 (1.00) an undervoltage followed by the respective EGDG output breaker closing or an EGDG breaker closed with an ES signal present (0.5 ea.) REFERENCE CR ROT 4-13 obj 14 (2.8/3,2) 062000K401 ...(KA'S) ANSWER 2.23 (1.00) All of the RPS channels would sense "C" and "D" RCPs tripped via the contact monitors. resulting in a reactor trip. REFERENCE CR ROT 4-12 obj 3 (3,1/3.5) 012000K603 ...(KA*S)

2. PLANT DESIGN INCLUDING SAFEYY AND EMERGENCY SYSTEMS PAGE 114 ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M ANSWERi 2.24 (1.00)

The VBXS are (solid state deviceis) designed to monitor their input and repidly transfer to their alternate source if the inverter should fail. (1.0) REFERENCE CR ROT 4-3 obj. 9. (2.8/3.1) 062000403 ...(KA'S) ANSWER 2.25 (2.00)

a. There are 45 reed switches strapped to the motor tube,

b. 1 Position indication in the contol room

2. Individual rod position indication is used to determine group average position.

3. Asymmetric Rod

4. Group In and Out limits

5. Provides input to the sequence enable

6. auto inhibit and out inhibit circuits

7. Bleed and feed permits (5 at 0.3 ea)

REFERENCE CR ROT 4-28 obj 4 (3.5/3.8) 001000K401 ...(KA*S) ANSWER 2.26 (1.50) The Rosemount dp transmitters have been ooserved to experience a zero reference shift (0.5) at a pressure different from that at which they were calibrated (0.5). This may cause the indicator to road some low value when actual flow does not exist (0.5). REFERENCE CR ROT 4-9 (2.3/2.5) 016000K601 ...(KA*S)

3. INSTRUMENTS AND CONTROLS PAGE 115 i ANSWERS -- CRYSTAL RIVER 88/01/12-DEAN, W M ANSWER; 3.01 (1.00) b REFERENCE CR ROT 4-25, obj .4 K/A (3.6/3.9) 073000K101 ...(KA'S)

ANSWER 3.02 (1.00) ed REFERENCE CR NA0 96, obj 6 , K/A (2.4/2.6) 059000K401 ...(KA'S) ( l ANSWER 3.03 (1.00) l-l C I REFERENCE CR ROT 4-15, obj 17 K/A (3.5/3.7) 061000K414 ...(KA*S) . ANSWER 3.04 (1.00) a REFERENCE CR-ROT 4-9, obj 2 K/A (3.2/3.2) 059000K107 ...(KA*S)

.: . .4

3. INSTRUMENTS AND CONTROLS 2? AGE 116 ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M a '

y, .

                                                                                                   -  a ANSWER /      3.05        (1.00)                                                                     I-t              ;     ,

e '

                                                                                           .. j
                                                                                               /,; ,r

REFERENCE , CR ROT 4-1, obj 6 K/A (3.8/4.1) ' 010000K403 ...(KA'S) ANSWER 3.06 (1.00) e REFERENCE CR NAO-112, obj 2, 3, 5 , K/A (2.6/2.8) 045000K413 ...(KA*S) t ANSWER 3.07 ( .50) TRUE (+.5) REFERENCE CR ROT 4-13, pp 39, LO 11 K/A (3.7/3.9) 013000K412 ...(KA'S) ANSWER 3.08 (1.00) , 4 7; 3;20;RunningESblockkloads(+.25ea) REFERENCE CR ROT 4-3, obj 7, pp 54 K/A (4.1/4.4) 062000K102 ...(KA'S) s

                                                                       .;y.
 ;;U6 3 _.lESTRUMENTS AND CONTROLS                                                    PAGE 117 f"       ANSWERS -- CRYSTAL RIVER                     -88/01/12-DEAN, W M     ,

oc0 . f 3 ANSWER,. 3.09 ( 1.50) o) 2 channels and temperature compensated (+.5 ea) b) 4 channels and temperature compensated

n c) 1 channel and not temperature compensated REFERENCE CR ROT 4-32, obj 6 K/A (3.6/3.8) 035010K401 ...(KA*S)

ANSWER 3.10 (1.00) o) Input / Output across SG/RX (+.5 ea) b) Header Pressure error REFERENCE CR ROT 4-14, obj 3 K/A (2.7/2.9) 045000K401 ...(KA'S)

. +

ANSWER 3.11 (1.50)

c) AandBEDGs(+.5)[tF4TPft.d'%diCMCoNWd D26munrea'MA) s # U#N'#

'b)

A RSP is energi=ed and B is not (1.0) REFERENCE CR ROT 4-16, obj 3 K/A (3.9/4.1; 3.3/3.4) 000068A121 000066K207 ..(KA*S) y u I ANSWER 3.12 (1.00) j' , e) Cross tie blocking in effect (+.5 ea) b) Breaker will close t REFERENCE CR ROT 4-3, obj S,6 K/A (2.8/3.1) 062000K403 ...(KA'S) t

7 - 7 g .;

                                                                                                                    ..                       ,- xx

e. .. . , ,/

m, - . [' ;[L.!5 3 . INSTRUMENTS AND CONTROLS - PAGE 118

            . u-
       +

f'LANSWERS -- CRYSTAL RIVER -88/01/12-DEAH, W M

  .qyl'c 1; / >                                                                                                                                                      .
     */
      ,,                6
                      -ANSWER.            3.13                     ( 1,00, )

a) ESLPue"A (+ 5 ea) A b) ~ B.end C' ' c) Studdby

                                                                         .:. ~                                                        '
                                                                                   ' ~                                                  ,s REFERENCE CR ANO.82, obj 4, 12                                                   ~

i K/A.(4.3/4.4)

  .i 006000K405          ,

1 . . (KA'I) ,

                                    \.       )                                                                                                        -

9 ANSWER ~ 3.14 ( 1.0Ui d i s i JWT5 - a) ~1end

.\

( + . 'd e a ) m

                      ) b)      Hold      the-buttoi            depres.3erY               ..

REFERENCE s CR ANO 82, pp 44 w- { K/A (3.6/4.2)

         ',             004010A204                   . . .     (KA'S)
                                                                                                                           '                  ~

l' \

8. .

             <                                                                                                      i ANSWER              3.15                    (2.00)                                                   t 1

'. 1) Heat Sensor in the ar.?t (+.25 for method, +.25 for location)

   'E ' 2) Manual Actuator in Cen+Jol Room
              \                                                                                                                            <

T 3) -Manual Actuator on va m outside Cont.rol Complex on turbiteideck

4) Panual Spurt button ae.s; 'the FW pumps f (

1 REFERENCE CR ROT 4-7, pp 32 K/A (3.0/0.3) !. 086000K406 . . (KA'S) l-

6 s

s N-a l' i

                                                                                                                                   / k 3

m, . . - - -

3. ' INSTRUMENTS AND CONTROLS PAGE 119 ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M LANSWER,, -3.16 (1.50)

1) Letdown Coolers-(+.25 ea for any 6)

2) RCDT Cooler.

3) RCPs 4)' Seal Return Cooler

5) RC Evaporator

6) Waste Evaporator 7). Waste Gas Compressor REFERENCE CR ANO 84, obj 6 K/A (3.6/3.9) ~

000026K302 ...(KA'S) ANSWER 3.17 (1.50)

[ DHV-3, 4 (+.5 ea) gudc#""yalamoccu DHV-3 and 4 will close and are electrically disabled from opening Bypassed by key interlocks 11. the ES Test Cabinets REFERENCE

             'CR ROT 4-1, pp 131/139, LO 3-3 K/A (3.2/3.5) 005000K407          ...(KA'S) 3.18         (1.00)
          . ANSWER' OL E

1500 PSIG HPI initiation or 4 PSIG RB signals (+.5[6cc48 1-) Protect against motor overload while operating in a denue atmosphere (+.5) REFERENCE ic CR ROT 4-8, pp 43, LO 6 K/A (3.1/3.4) 022000K402 ...(KA'S) b

       '3. INSTRUMENTS ~AND CONTROLS                                         PAGE 120 ANSWERS -- CRYSTAL RIVER                -88/01/12-DEAN, W M ANSWER /       3.19       (1.50)

Feed flow and reactor power will increase due to the increased demand.(+.5) This will cause OTSG pressure to rise above the turbine header pressure setpoint (+.5). The Turbine PfyE33' Valves will open to restore the header

         . pressure to setpoint, accounting for the extra 5% heat generation (+.5)

REFERENCE CR ROT 4-14, obj 4, 5, 8 K/A (2.9/3.3) 041020K401 ...(KA'S) ANSWER 3.20 (1.00) Technician is correct. (+.5) When in bypass, the trip relay is supplied power via a separate path not affected by module removal (+.5) REFERENCE CR ROT 4-12, obj 15

         -K/A (3.3/3.6) 012000K604        ...(KA'S)

ANSWER 3.21 (2.00) a) Between 600 and 750 psig (+.5) b) 1) Depress Channel A, B, C and D bypass pushbuttons on PSA panel (+.5) the buttons will backlight when bypass in effect (+.25)

2) Depress the SG Bypass / Reset toggle in all 4 EFIC cabinet Initiate modules to BYPASS (+.5) The SG LOW PRESSURE bypass light on the cabinet alarm panel will blink (+.25)

REFERENCE , CR ROT 4-15, pp 28/29, LO 18 K/A (4.0/4.2) 061000K406 ...(KA'S) ANSWER 3.22 (1.00) The level setpoint will be adjusted to 24 inches automatically (+1.0)

3. ' INSTRUMENTS'AND CONTROLS PAGE 121

                                  ' ANSWERS -- CRYSTALLRIVER                             -88/01/12-DEAN, W H REFERENCE oCR ROT 4-15, pp 16, LO 14 K/A (3 7/2.9)-                               .

061000K411 ...(KA'S) ANSWER 3.23 (1.00)- At least one vacuum breaker must be open (+.5) to electrically open EFV-1 or EFV-2 (+.5) REFERENCE CR ROT 4-15, LO 4 K/A (3.9/4.2) 061000K401 ...(KA'S) ANSWER 3.24 (1.50)

                                  -The High Flux Trip and the Power / Imbalance / Flow trips for Channel A will actuate-(contacts open)(+0.75). This will cause the channel A trip relay to deenergize and open contacts (KA1, 2, 3 and 4) in each trip module, making 1 side of the 2/4 required logic (+.75).

REFERENCE CR ROT 4-12, pp 13-14,33,38, LO 3 K/A (4.1/4.2) 015000K101 ...(KA'S) ANSWER 3.25 (1.00) see attached REFERENCE CR ROT 4-10, fig 11, LO 3 K/A (3.7/3.9) 015000K402 ...(KA'S)

                                                                                                       ,.ey-m         - - - - .n _

i

    ,O l                                        -
                                                                               F10URE    ll ROD WITH0RAWAL INHIBIT LO0lc
                                                 ==                        ==           in I

N!=1 N!=2 M13 i. Nt.4 ai.s a .s 2: = M 1 2:= M 23 = := 4!. 4?. 4?r 4?. 1 l

  !                                                                 I

(+.ia cE.zc) I I i l () LJ (b. r) ,. O 2 ,, .. ,,, ,,,, _=_.

                                                                                               'aa'
                                         =!".'i?.                              .f". ??.      4.", 4*.

I I 1 Q -Q I I I I I I

;                                                             = = .

m. mu.e, .

O SOURCE RANGE HIGH VOLTAGE OFF LO0lc O 3

~

4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 122.

RADIOLOGICAL CONTROL ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M ANSWER' 4.01 _(1.00) b REFERENCE CR TS 3.4.6.1, CR ROT 5-1, LO 11 K/A (3.6/3.8); (3.6/4.1) 002020G005 002020K401 ...(KA'S) ANSWER 4.02 (1.00) e REFERENCE CR ROT 5-14, LO 5 K/A (4.0/4.1) 000011G012 ...(KA'S)

   ; ANSWER        4.03        (1.00) b REFERENCE CR TS 3.2.4; CR ROT 5-1, LO 10 K/A (3.7/4.1) 001050G005          ..(KA'S)

ANSW R 4.04 (1.00) REFER (CE CR ROT 3-3, LO 6,7 K/A (4. V4.7) 000074A207 ...(KA'S)

le 4; PROCEDURES ~- NORMAL. ABNORMAL. EMERGENCY AND PAGE 123 : RADIOLOGICAL CONTROL ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M ANSWER' 4.05 (1.00) d REFERENCE CR OP-204, pp 17; CR ROT 5-2, LO 5 K/A (3.6/4.0) 001050A206 ...(KA'S) ANSWEB 4.06 (1.50) c)- TRIP (+.5 ea) b) TRIP c) DON'T TRIP REFERENCE . CR ROT 5-28, LO 1; AP-580 K/A (4.2/4.1) 000037G010 ...(KA'S) ANSWER 4.07 (1.00) L c) 75% (+.5 ea) b) 100 cpm REFERENCE CR ROT 5-43, LO 5; RSP-101 K/A (2.8/3.4) 194001K103 ...(KA'S) ANSWER 4.08 (1.00) a) So that the OTSG will provide a good heat sink (+.5 ea) b) SPDS or RCP NPSH curve with plant wide range pressure indication REFERENCE CR ROT 3-5, pp 26; AP-530; ROT 5-25, LO 4 K/A (4.0/4.4) l 000074K307 ...(KA'S) i y .- . ,, , . , - -._ _ -.v,_ , __. r,m. __ , ,- , r

4. PROCEDURES - NORMAL. ABNORMAL'. EMERCENCY AND RADIOLOGICAL CONTRQL PAGE 124 ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN,-W M ANSWER 4.09 (1.50) NO (+.5) TS (3.9.2) and detensioning of a head bolt constitutes entrance into Mode REFERENCE CR TS 3.9.2, TS 1.4; CR LER 87-023; ROT 5-1, LO 5 K/A (3.3/3.8) 0150000005 ...(KA'S) ANSWER 4.10 (1.50) 1) Ensure valves on affected OTSG are closed: (+.25) MSIV MBV (411/412 or 413/414) (FWV-30 or 29) LLBV (FWV-31 or 32) SUBV (FWV-36 or 33) Cross-tie (FWV-28) MFP Suction (FWV-14 or 15) 2) (+.75 for valves) Ensure MFPs on affected OTSG Tripped (+.5) REFERENCE CR-ROT 5-23, LO; AP-460 K/A-(4.1/4.2) 000040G010 .

                                    .(KA'S)

ANSWER 4.11 (1.00)

1) SSOD (+.5 ea)

2) ANSS REFERENCE CR ROT 5-40, LO 3; CP-115, pp 3 K/A (3.7/4.1) 194001K102 ...(KA'S)

 +,         -w-        -    -                          -   .- ,,--r-,w.- s ,---
                                                                                --r- r-,,          v

 ' t .,    .. -                                                                            .

4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 125-RADIOLOGICAL CONTROL ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M ANSWER' 4.12 (1.00)

1) Cable Spreading Room (+.25 ea for any 4)

2) ES Switchgear Room

3) Station Battery Rooms

4) CRD Equipment Rooms

5) Control Coaslea REFERENCE CeNer/W CR ROT 5-40, CP-118 LO 1; CP-118, pp 2 K/A (3.7/4.1) 194001K102 ...(KA'S)

ANSWER 4.13 (1.50) c) 1) Nuclear Fuel (+.25 ea)

2) Primary Coolant Pressure Boundary

3) Containment Integrity

4) ES Systems b) As each step is performed, the performer must sign / initial the step

(+.5) REFERENCE AI-500, pp 42; CR ROT 5-38, LO 13 K/A (4.1/3.9) 194001A102 ...(KA'S) l ANSWER 4.14 (1.50) ! 1) Start CAP-1A or 1B (+.5 ea)

2) Open CAV-60 (Emergency Boration Isolation)

3) Establish Maximum Letdown REFERENCE CR ROT 5-28, LO 3; AP-580 K/A (4.0/4.0) 000024G010 ...(KA'S)

m

    ' 4' . PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND'                      PAGE 126 RADIOLOGICAL CONTROL ANSWERS -- CRYSTAL RIVER                  -88/01/12-DEAN, W H ANSWER'        4.15          (1.00)

1) Reactivity Control (+.2 for CSF, +.05 for position)

2) Thermal Control

3) Radioactive Inventory Control

4) Equipment Availability REFERENCE CR ROT 5-14, pp 9/10; LO 10 K/A (3.8/3.9) 000007G012 ...(KA'S)

ANSWER 4.16 (1.00)

1) w P CRD Stator alarm (/4160 degrees) (+.33 ea)

2) p>Afr7 CRD energized -

3) RCS Temp j

                            / 200 degrees.

9EFERENCE JR OP-502, pp 5; ROT 4-28, LO 3 K/A (3.3/3.5) 001050G010 ...(KA'S) 4 ANSWER 4.17 (1.75) a) 1) > 50 degrees delta T between condensers (+.25 ea)

2) > 8 Degrees delta T between highest / lowest hot gas temperatures

3) > 10" Hg absolute condenser vacuum b) Close MSIVs (+.5 ea)

Notify Load Dispatcher REFERENCE CR ROT 5-29, LO 1 & 3; AP-660 K/A (2.8/2.9), (2.9/3.2) 045000G014 045000G015 ...(KA'S) l l l L'

e

4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 127 RADIOLOGICAL CONTROL

     -ANSWERS -- CRYSTAL RIVER                          -88/01/12-DEAN, W M ANSWER                4.18       (1.00)                    , ('  -
                                                                    ,h m

1) Both actuation output breakers are opened 44avfr ea)

2) AdedicatedoperatormustbestationedattheEF/Cpontrolstomanually actuate the system _if required ,_ Eu - / c c h'-

RE C f #7"" ' CR LERs 87-001, 87-002 K/A (3.7/3.9) 061000G001 ...(KA'S) ANSWER 4.19 (1.75)

1) Adequate subcooling margin exists (+.25)

2) Maintain RCS Press / Temp < NDT (+.5)

3) Maintain HPI Flow < 540 gpm/ pump (+.5)

4) Maintain subcooling < 100 degrees when no RCPs are operating (+.5)

REFERENCE AP-380; CR ROT 5-22, LO 4 ! K/A (3.6/4.2) ( 000009A234 . .(KA'S) ANSWER 4.20 (1.50) a) Controlling on Low Level Limits (+.5 ea) b) one MFP and one MFBP c) 2155 psig/579 degrees j REFERENCE l CR OP-208; CR ROT 5-3, LO 4a l K/A (3.6/3.5), (2.7/2.9), (4.0/3.9) j 010000A302 035010A301 059000A103 . .(KA'S) l l l

   - ANSWER                 4.21       (1.00)

Lifting of secondary safeties due to overfilling the OTSG (liquid release) or overpressurizing OTSG. (+1.0) REFERENCE ATOG Guidelines, III-E, pp 55 K/A (4.1/4.5)

m p

 ,'    .4 ?

PRbCEDURES-' NORMAL.ABNORMdL.EMERGENCYAND RADIOLOGICAL CONTROL PAGE 128 ANSWERS -- CRYSTAL RIVER -88/01/12-DEAN, W M-000038K306 ...(KA'S) ANSWER 4.22 (2.00) a) With level center of 65%, of the core thermal (+1.0) center of the OTSG is above the thermal b) Since lossReflux possible, of subcooling boiling ma is an indication of void formation being greater heat transfer area.y (+1.0) need to be established, which requires a REFERENCE CR ROT 3-3, pp20/21; CR ROT 5-25, LO 4 K/A (4.1/4.2) 000017K307 ...(KA'S) ANSWER 4.23 ( .50) An RCA is used to control access to a Radiation Area (+.5) REFERENCE CR~ ROT 5-43, LO 1 K/A (3.3/3.5) 194001K104 ...(KA'S) ANSWER 4.24 (1.00) BelowPump, TDEFW 200 psig secondary pressure, cannot supply sufficient flow with the so the pump is essentially inoperable (+1.0) in Mode 3. REFERENCE CR LER 87-17; OP-209; ROT 5-3, LO 5 K/A (3.5/3.6), (3.3/4.0) 061000G005 061000G010 ...(KA'S) ANSWER 4.25 (1.00)

    -Allow the nozzles to heat up with the feedwater (+.5) and aid in transferring heat to as much of the OTSG shell as possible (+.5)

REFERENCE CR ROT 4-32, LO 3; CR ROT 5-2, LO 3a K/A (3.2/S.4) 035010G010 ...(KA'S) i I i}}

ML20196K407

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