ML20138R888

Exam Rept 50-298/OL-85-04 on 850923-26.Exam Results:Four Senior Reactor Operators Passed Exam & One Failed & Two Reactor Operators Passed Exam & One Failed

ML20138R888
Person / Time
Site:	Cooper
Issue date:	11/08/1985
From:	Cooley R, Graves D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20138R885	List: ML20138R884 ML20138R888
References
50-298-OL-85-04, 50-298-OL-85-4, NUDOCS 8511190204
Download: ML20138R888 (62)

Text

.( .A APPENDIX U.S.' NUCLEAR REGULATORY COMMISSION REGION IV

.NRC Examination Report: 50-298/0L-85-04 License: DPR-46 Docket: 50-298

-' Licensee: Nebraska Public Power District (NPPD)

P. O. Box 499 s

Columbus, Nebraska 68601 Facility Name: CooperNuclearStation(CNS)

~Requalification examinations administered at Cooper Nuclear Station-(CNS)

Examinations Conducted: September 23-26, 1985 T

' Chief Examiner: Y 72 AM//r David N. Graves . Date Approved: .h-R. A. Cool 6y, SectFon Chief

///f/[f Date' Inspection Summary An audit of the Cooper Nuclear Station Requalification Program was conducted by Region IV personnel. This audit found the requalification program to be SATISFACTORY.

8511190204 851108 PDR ADOCK 05000298 G PDR

_ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ __ - _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _______._._________.-_________._____.__.____O

c 1

DETAILS

1. Facility-Generated Written Examination Facility examinations to be administered to Reactor Operators (RO) and

~

Senior Reactor Operators (SRO) were reviewed for compliance with 10 CFR 55, Appendix A. This review showed the examinations to be appropriately structured as to content. The examinations were also satisfactory when compared to the guidelines of NUREG 1021, Operator Licensing Examiner Standards, Section ES-601, Administration of NRC Requalification Program Evaluation.

2. Persons Examined Five SR0s and three R0s were administered NRC-generated written examinations and oral operating examinations.

3. Results of NRC Administered Examination One R0 failed the written examination and the oral operating examination. One SR0 failed the written examination.

4. Examiners D. N. Graves, Chief Examiner J. E. Whittemore

5. This section pertains to the site visit conducted September 24-26, 1985, and consists of the following subsections:

a. Examination Review Comments and Resolutions.

Facility comments are listed by section question number and followed by the Chief Examiner's resolution.

1) Question 1.05 (Also 5.01).

Comments: Value of 1*.USAR uses 50 microseconds; GE Reactor Physics Review uses 100 microseconds. NRC has previously accepted values for 1* between 50 and 150 microseconds. The value of 1* used will significantly change the answers _in several parts and this should be taken into account in grading.

Resolution: Agree. Accepted 1* values from 50 to 150 microseconds.

)

me-4.

2) Question 1.08 (Also 5.08).

Coments: Assuming question is referring to available NPSH, Rx dome. pressure should also be an acceptable answer. We teach:

NPSH,y43)=(Pdome+Pheight -P loss)-P sat Resolution: Accepted Rx dome pressure as subcooling.

3). Question 1.11.

Comments: Student may also answer question in terms of ensuring MCPR-Safety Limit is not exceeded. This should also be an acceptable answer.

Resolution: Agree.

-)

4 Question 2.06.

Comments: REC lemperature Control is manual or automatic.

Manual by varyirg SW flow with RCV-451 A or B in open position.

Automatic Contrel by TCV-451 A and B is Auto Valve position controlled by Rif Temperature out of Rx. Temperature adjustable

, at controller. 3

^

Resolution: Agiee.

. c.

'5) Question 3.01c (Also 6.02c).

Comments: Question could be misinterpreted. Student may assume

question implies that SRV did not open. If student assumes SRV

~m

~-

did not open, answer would be:

Red' Green Blue Off On On This answer should also be accepted.

Resolution: Accepted if candidate states assumption.

6) Question 3.05(Also6.07).

Comments: Plant Terminology / Common Usage LI-91-A and B: Fuel Zone, Rosemount, Wide Range, Wide Range Yarways LI-85 A and B: Narrow Range Yarways LI-94 A, B, C: GEMAC, Narrow Range GEMAC Resolution: Accepted.

1 s:

7) Question 3.07.

Comment: Plant Terminology. Student may refer to Log Rad Monitor as Off Gas Monitor or Off Gas RAD Monitor. Either of these terms should be accepted.

Resolution: Agree.

~

8) Question 4.01(Also8.0i).

Comment: Due to the fact that as a normal course of action, an announcement regarding the nature of the alarm and immediate response actions required is made over the "public address"

. system, we believe that this question is inappropriate for a licensed operator examination. -Further, we expect that all would recognize that a " ringing gong" would not be correct for any of the emergency signals; hence, one incorrect response results in at least two incorrect answers, a " double jeopardy" situation. On this basis.we believe the question should be deleted or, if not, its value be readjusted downward.

Resolution: Disagree. Question and answer stand as written.

9) Question 4.03.

Comment: May receive the answer, " Main Steam Channel A and B low pressure alarms on panel 9-5 clear." May also receive an answer of "APRM gain adjusted to 1.0" as it is in the same paragraph as the other.three answers and should be considered a correct answer.

Resolution: Agree.

10) Question 4.04 (Also 7.02).

Comment: Undesirable effects. CRD supplies seal purge H9 0 to the RR Pumps and RWCU Pumps. If this supply is lost, seat life .

could be shortened. The loss of the pumps will eliminate the possibility of moving the control rods with the RMC system as hydraulic pressure is lost. Additional information regarding l- possible adverse effects if the CRD pumps are lost is provided in the referenced procedure.

Resolution: Agree.

L 11) Question 4.09d (Also 7.08d).

Comment: We believe that there may be other ways to answer this question. Procedure 2.1.10 states, " Suppress the oscillations by inserting control rods and/or increasing core flow. It is

_ preferred to reverse the actions that caused the flux oscillations."

(

i Resolution: Abnormal Procedure 2.4.2.2.1, Trip of Reactor Recirculation Pumps, covers this particular situation and makes no provision for increasing flow with the remaining pump.

Answer stands as written.

12) Question 5.02a.

Comment: Boiling boundary. Student may answer question in terms of " sweeping away voids" and void content. This should also be an acceptable answer.

Resolution: Agree.

13) Question 5.06.

Comment: Question could possibly be misinterpreted. (Student could answer in terms of resonance capture, and the effect of resonance escape probability on Keff.) FOR FUTURE

REFERENCE:

Question may be reworded as such: State two (2) effects that cause resonance capture to increase with an increase in fuel I temperature.

Resolution: Noted for future reference.

14) Question 6.04.

Comment: The referenced procedure assumes Standby Mode while the question does not address EDG condition. It is felt that some answers that are addressed in attaining Standby Mode may be included as correct.

(a) DC Control Power and Maintenance Lockout Keylock is in the "On" position. Lack of this Control Power will prevent Breaker operation.

(b) DG Breaker Selector Switch in Auto. This switch being out of Auto position will prevent Breaker operation other than manually, with synch scope interlock satisfied.

In addition, to fully answer the question, i.e., "and re-energize its associated bus?", it is felt that, "IFE (GE) closed," is also a correct answer.

Resolution: Agree.

15) Question 7.01.

Comment: While responses to the second part of this question may have been provided, it is no longer an applicable question for CNS since maintenance of DW/ Torus dp is no longer required.

Resolution: Agree. Points redistributed.

16) Question 8.02 bac.

Coment: Designation of a shift communicator should also be

cons.'dered a correct response.

Resolution: Agree.

f 17) Questicn 8.04.

. Comment:

! (a) The basis of Tech Specs (pg.102) states "A limiting l Control Rod Pattern is a pattern which results in the core being on a Thermal Hydraulic L1mit (i.e. , MCPR = 1.07, and l LHGR - as defined in 1.0.A.4)" Ref: Tech Specs; Definition l Section, pg. I and Basis Section, pg. 102. It seems that 4 the procedure has shortened the Tech Spec explanation /

definition for ease of use. The above answer may be given

, and should be considered correct.

(b) We believe that this question should be deleted. The answer is based upon information provided in the Discussion section of the referenced procedure. Administrative Procedure 0.4 " Preparation, Review and Approval of Procedure" states that the discussion section of a procedure is to " provide a short discussion of the ,,

procedure and its objectives." The discussion section is intended; therefore, as no more than an aid; consequently, we would not expect an individual to commit to memory information contained in that section. While we recognize that licensed operations personnel need to " commit to memory" immediate action steps in an emergency procedure, (Note that Procedure 10.10 does not fit that classification), subsequent actions required need not be, let alone information contained in a Discussion section.

In that respect, CNS places the responsibility for determination of limiting control rod patterns on Reactor Engineering personnel. Such a responsibility is defined in both procedures and Technical Specifications. The following references are cited:

The basis of Tech Specs (pg. 102) states "It is the responsibility of the Reactor Engineer to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to the occurrence of inoperable control rods in other than limiting patterns."

o In addition, Procedure 10.10 II.B states "G.E. Studies in

'this field have indicated that Engineers who specialize in Control Rod pattern development and analysis learn to tell by visual examination whether a limiting configuration exists, but there has not been any simplified method to date; general curves, tables, or overlays which would guarantee that such an assessment could be made on an absolute basis by someone else. (i.e. Operators) The combination of odd circumstances to be covered to guarantee absolute certainty would be almost infinite in extent."

Also, Procedure 10.10.C pg 1 states "As such, each of the above abnormal withdrawal sequences shall be considered a potential limiting Control Rod pattern, unless, specific analyses are perfonned which demonstrates a limiting pattern does not exist. Also, Procedure 2.4.1.4; Step I.A states "The Reactor Engineer will notify the Shift Supervisor and have it entered in the Operations Log when a limiting Control Rod Pattern exists."

Further, the answer requires that three (3) responses be provided. This comprises all of the potential responses identified in the Discussion section. Hence, to obtain full credit, all of the elements within the Discussion section need to be addressed. We believe that, in itself, is inappropriate. Finally, the point va 4

assigned amounts to 1/6 of the total point valuelue in that section. This is an excessive value assigned to a question which we feel to be inappropriate due to the fact that Reactor Engineering Personnel are specified in the Technical Specifications as being responsible for identifying control rod patterns.

Resolution: (a) Agree (b) Disagree. The question does not ask to identify a specific limiting control rod pattern. The question states the general way to obtain a limiting pattern is by an abnormal withdrawal sequence. It is reasonable to expect an SR0 to be aware of conditions under which abnormal patterns (not specific patterns) could be encountered.

Question and answer stand.

b. Exit Meeting Summary At the conclusion of the site-visit, the examiners met with utility representatives to discuss the results of the examinations. The i

k

[; )

p3 9.

following' personnel were present for the exit meeting: "

NRC Facility D. L. Dubois, Senior Resident Inspe'ctor R. D. Black D. N. Graves ' '

D. L. Reeves, Jr.

. T. Sandner

- 'J. Sayer

., P. V. Thomason l

'The meeting was started by Mr. Graves announcing preliminary results of the oral operating examination with one R0 not being a " clear pass" as of.the exit meeting. The facility was informed that this did not mean the candidate failed, but that a more detailed review of his performance must be made.

l Observations made during the' examination were discussed as follows:

1) Several SR0s were indecisive and hesitant concerning actions required by the Emergency Plan. Given a set of-conditions,

.several were hesitant to make decisions or commit to a specific course of action.

2). In at least one instance, the Emergency Action Level classification chart identifies one event under two classifications. Specifically, a main steam line break with loss of primary containment integrity is addressed specifically as a Site Emergency yet also satisfies the criteria for a loss of two of three fission product barriers, which is a General Emergency.

The meeting concluded with the examiners thanking the facility staff

• for their cooperation and informing them that final results would be f forthcoming as soon as possible.

t Requalification Program Evaluation c.

The Requalification Program Evaluation Report is presented on the next page.

d. Master Copies of the R0 and SR0 Written Requalification Examination Copies of R0 and SR0 examinations follow the Requalification Program Evaluation Report.

a. . . .. .

4 REQUALIFICATION PROGRAM EVALUATION REPORT Facility: COOPER NUCLEAR STATION Examiner: D. N. r, raves Date(s) of Evaluation: Seotember 24-26, 1985 Areas Evaluated: XX Written XX Oral Simulator Written Examination

1. Overall evaluation of examination: (NRC written) 6 of 8 candidates passed

2. Evaluation of facility examination grading: N/A Oral Examination

1. Overall evaluation 7 of 8 candidates passed NRC administered oral exams,

2. Number conducted 8 Simulator Examination

1. Overall evaluation N/A

2. Number conducted N/A Overall Program Evaluation Satisfactory XX Marginal Unsatisfactory (List major deft-ciency areas with brief descriptive comments)

Submitted: Forwarded: Approved:

// k.d. M, O NW,v.

Examiner SectionChieg gBranchChief

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

REACTOR TYPE: _BWB-QE4_________________

DATE ADMINISTERED: _Q5(QQL2h________________

EXAMINER: _QB8VElt_Qt______________

APPLICANT: _________________________

18118UQ11081_IQ_aEEL108HIl 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 locst 80%. Examination papers will be picked up six (6) hours after the examination starts.

% OF CATEGORY  % OiF APPLICANT'S CATEGORY

__V8LUE_ _IQI6L ___SQQBE___ _VaLUE__ ______________GoIEQQBY

.15tQQ__ _25tQQ ___________ ________ 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

_15tQQ__ _25100 ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS

_15tDQ__ _25tDa ___________ ________ 3. INSTRUMENTS AND CONTROLS

_15tDQ__ _25tQQ ___________ ________ 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL

_kQt0D__ 10Q10Q ___________ ________ TOTALS FINAL GRADE _________________%

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

~~~~~~~~~~~~~~

PPL55 UT~5~55GU TURE

lt__EBIUCIELE1_QE_NMQLE68_EQWE8_8L6Bl_QEE8611QN t PAGE 2 IBEBdQQ1Ned1Q1&_BE61_IBaNSEEB_6ND_ELulD_ELQW QUESTION 1.01 (2.00)

c. What combination of recirculation flow and control rod position will yield the MAXIMUM PERCENT VOIDS in the core? (Specific values not required) (1.0)

b. Why does the void percentage'OECREASE as power is raised from the level in part "a" to 100% power? (1.0)

QUESTION 1.02 (2.00)

What are five (5) mechanisms or methods by which reactivity additions are made in the power range? (2.0)

QUESTION 1.03 (2.00)

Consider a turbine trip from approximately 25% power. Briefly explain what would be expected to occur to the plant in the next 15 minutes with no operator action. Include parameters such as power, pressure, temperature, etc. Provide reasons for all changes. (2.0)

QUESTION 1.04 (2.00)

Concerning control rod worths during a reactor startup from 100% PEAK XENON versus a startup under XENON-FREE conditions, which statement is correct? JUSTIFY YOUR CHOICE. (2.0)

o. PERIPHERAL control rod worth will be LOWER during the PEAK XENON startup than during the XENON-FREE startup.

b. CENTRAL control rod worth will be HIGHER during the PEAK XENON startup than during the XENON-FREE startup.

c. BOTH control rod worths will be the SAME regardless of core Xenon conditions.

d. PERIPHERAL control rod worth will be HIGHER during the PEAK XENON startup than during the XENON-FREE startup.

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

it__EBINGIELE1_QE_UUGLE68_EQWEB_EL6NI_QEE8611QN t PAGE 3 IBEBdQQ10adIGS&_BE61_IBauSEEB_600 ELUIQ_ELQW QUESTION 1.05 (1.00)

How much (by what factor) would power increase in one second in a prompt critical reactor at E0L? (1.0)

QUESTION 1.06 (1.00)

Using the Steam Tables, calculate a reactor cooldown rate (F/hr) for a reactor pressure decrease from 1000 psig to 250 psig in one hour and forty-five minutes (105 minutes total). Show all work for full credit. (1.0)

QUESTION 1.07 (1.00)

Select the answer below which best describes Net Positive Suction Head:

a. The difference between the pump suction pressure and the pump (1.0) discharge pressure.

b. The pressure above which the pump can no longer provide flow

c. The difference between Psat at the pump discharge and the actual pump discharge pressure.

d. The difference between Psat at the pump suction and the actual pump suction pressure.

QUESTION 1.08 (1.00)

What are the three (3) maj or parameters affecting the NPSH of the recirculation pumps? (1.0)

QUESTION 1.09 (1.00)

During a LOCA, large amounts of hydrogen may be produced. Why will odequate core cooling prevent a great deal of this hydrogen from boing produced? (1.0)

(***** CATEGORY 01 CONTINUE 0 ON NEXT PAGE *****)

lt__EBIUQ1ELsS_QE_UVQLE68_EQWEB_ELaNI_QEEB811QUt PAGE 4

-IBEBdQQ18edIQSt_BE61_IB6USEEB_8UQ_ELUID_ELQW l

QUESTION 1.10 (1.00)

Which type of fuel, new fuel or exposed fuel, would have the longer thermal time constant? EXPLAIN. (1.0)

QUESTION 1.11 (1.00)

How is it assured that 99.9% of the fuel pins in the core do not experience transition boiling during a transient? (1.0)

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

2t__ELeNI_DE1108 INGLVQ1NQ SeEEIl_680_EdE89EUQ1_211IEUS PAGE 5 QUESTION 2.01 (2.00)

e. What are six (6) areas served by the Low Pressure CO2 System?

Indicate whether the area is covered AUTOMATICALLY or requires MANUAL actions. (1.5)

b. After the initial 50 second discharge of CO2, how are subsequent discharges. initiated (2 methods or locations required)? (0.5)

QUESTION 2.02 (2.00)

For each of the following diesel generator controls, describe its offect(s) on diesel generator performance / operation.

Include how the listed controls affect automatic and manual starts and the relationship (s) between the controls as applicable.

s. Control Mode Selector Switch (when in REMOTE) (0.5)

b. Diesel Control Isolation Switch (when in LOCAL) (0,5)

c. Diesel Generator Stop Pushbutton (on D/G engine panel) (0,5)

d. Emergency Stop Pushbutton (on D/G engine panel) (0,5)

(***** CATEGORY 02 CONTINUED ON NEX1 PAGE *****)

2i__ELaNI_QESIGU_INGLUQ1NQ_SaEEIl_800_EMEBGEUQ1_SISIEd1 PAGE 6 QUESTION 2.03 .(2.00)

For each of the following items (a- d), indicate whether it describes

= proper system response or not. If it does not, describe how the discussed event should occur.

c. With the RCIC system operating, a low level occurs in an ECST.

.The pump suction from the ECST (RCIC-MO-18) closes and the pump suction from the suppression chamber (RCIC-MO-41) then opens. (0.5)

b. The RCIC system is operating with reactor level increasing.

When the high level setpoint is reached, the steam supply inboard and outboard isolation valves (RCIC-MO-15 and 16) close. (0.5)

c. The RCIC system is operating in the TEST mode, discharging to the ECST. A valid low reactor level initiation signal is received. The test circuitry is automatically bypassed, the test bypass to ECST closes, and the flow controller controls RCIC flow automatically. (0,5)

d. With the RCIC system operating, a high steam line space temperature isolation signal is generated. The following valves close as a result of the isolation: Steam supply inboard and outboard isolation valves (RCIC-MO-15 and 16), the minimum flow valve (RCIC-MO-27), the RCIC pump discharge valve (RCIC-MO-20), and the injection valve (RCIC-MO-21). NOTE: The above listed valves are not the only valves that are affected by the isolation.

EVALUATE ONLY THE VALVES LISTED. 00 NOT ATTEMPT TO COMPLETE THE LIST. (0.5)

QUESTION 2.04 (1.50)

c. The ORYWELL COOLING SYSTEM is operating with three (3) units running and one in STAND 0Y. How does the system respond to a

~LOCA signal? (0.5) i l

b. What system supplies cooling water to the ORYWELL COOLING SYSTEM? (0.5)

c. What will cause a ORYWELL COOLING SYSTEM ur.it to automatically start? (0.5)

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

2t__ELaul_QE11GN_18CLUDIUQ_S6EE11_600_EUEBGENC1_11SIEd1 PAGE 7 QUESTION 2.05 (1.50)

c. What two (2) mechanisms or methods are used in the Offgas System to maintain or reduce hydrogen concentration of the offges? (1.0)

b. Why is there a maximum power limit imposed on the use of the mechanical vacuum pumps? (0,5)

QUESTION 2.06 (1.50)

c. How is the temperature of the REC system controlled? Include whether control is automatic or manual, and how the temperature is adj usted. NO VALUES are required. (1.0)

b. What will cause the REC critical loops to automatically be placed into service? (0.5)

QUESTION 2,07 (1.50)

c. Where does the Standby Gas Treatment System line up to take a suction on an automatic initiation due to a refueling accident? (0.5)

b. Other than the normal automatic initiation supply and dilution air, what are two (2) additional areas or components that can provide a supply to the Standby Gas Treatment System? (1.0)

QUESTION 2.08 (2.00)

a. What provides the NORMAL and BACKUP cooling supplies to the three (3) plant air compressors? BE SPECIFIC. (1.5)

b. Is the normal to backup cooling water relationship a manual or automatic feature? If automatic, include what parameter (s) cause(es) the transfer to occur. Setpoints not required. (0.5)

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

2t__EL6HI_DE11GH_IUQLuQ1ND_18EEIX_600_EDEBGEUGl_1111EUS PAGE 8 QUESTION 2.09 (1.00)

Control Rod Drive Hydraulic System flow is automatically maintained ot approximately 45 gpm. Describe the flow path for this water, with no CRD motion, from the CRD Flow control Valve to where it nixes with the maj ority of the reactor coolant. (1.0) i

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

2t__INSIBudENIS 8NQ_QQNIBQLS PAGE 9 QUESTION 3.01 (3.00)

Describe the indication (ON or OFF) of the three (3) indicating lights for a Safety / Relief Valve under each of the following conditions: (3.0)

c. No actuation signals present, valve shut

b. The valve-is leaking significantly

c. 'The valve handswitch is in OPEN

d. The valve opens due to high reactor pressure

'o.

ADS logic actuated QUESTION 3.02 (1.50)

Following a reactor scrom, the four rod display position goes blank, but the green full-in light on the full core display for that control rod is lighted. Is this normal? If so, explain why it occurs. If not, describe the probable cause. (1.5)

QUESTION 3.03 (2.00)

e. Which two (2) SRM rod blocks are bypassed when the IRM's are on range 5? Setpoints not required. (1.0)

b. Which two (2) SRM rod blocks are bypassed (in addition to the 2 above) when the IRM's are on range 8 or above? Setpoints not required. (1.0)

QUESTION 3.04 (1.50)

Ocacribe three (3) rod blocks associated with refueling equipment.

Include required Mode Switch position (s), as applicable. (1.5) l (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)

l-

2t__INSI8udENIS_6NQ_GQUIBQL$ PAGE 10 QUESTION 3.05 (2.50)

o. List eight (8) indications of reactor vessel water level in the Control Room. Include on which panel the indication is located. (2.0)

b. How far is INSTRUMENT ZERO (0 on most level ranges) above the top of the active fuel (TAF)? (0.5)

QUESTION 3.06 (2.00)

Indicate whether each of the following statements about the operation of the DEH system in Mode 4 (Turbine Follow - Reactor Manual) are TRUE or FALSE. For those that are FALSF., BRIEFLY EXPLAIN WHY.

c. Raising the pressure control signal above the load reference will cause the bypass valves to open. (0.5)

b. Reducing the valve position limiter setting to below the pressure control signal will result in an increase in reactor pressure. (0,5)

c. Reactor power is controlled by the operator adj usting the load reference signal. (0.5)

d. Governor valve control will remain in AUTOMATIC if a loss of the speed loop (speed signal) occurs. (0.5)

QUESTION 3.07 (2.00)

c. What combination (s) of Off Gas Radiation Monitoring System trips will start the Off Gas Valve Timer? (1.0)

b. What four (4) valves are isolated (shut) at the completion of the Off Gas Valve Timer cycle? Valves may be identified by number or description. (1.0)

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

34__INSIBudENIS_6NQ_CQUIBQLS PAGE 11 QUESTION 3.08 ( .50)

Which of the following provides the signal for a Turbine Control Volve (TCV) Fast Closure scram? (0.5)

1. TCV position limit switches

2. Rate of TCV position change

3. Power to the TCV fast acting solenoids

4. Turbine control fluid pressure

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

st__EBQQEQUBES_:_NQBd8(t_8ENQBd8(t_EUEBQEUQ1_88Q PAGE 12 88Q10LQQ1Q8L_QQUIBQL i

1 QUESTION 4.01 L1.50)

Match the emergency signal (a - c) with its tone description. (1.5)

_____a. Emergency alarm 1. distinct steady tone

_____b. Fire alarm 2. one steady up and down tone

_____c. All clear 3. ringing gong

4. distinct pulse tone QUESTION 4.02 (2.00)

A control rod coupling test is to be performed. Briefly describe how the check is performed AND provide four (4) indications that the operator would see if the rod was uncoupled. (2.0)

QUESTION 4.03 (1.50)

! L ist three (3) verifications that should be made prior to placing l the Mode Switch to RUN during a plant startup. (1.5)

QUESTION 4.04 (2.00) l With the reactor operating at power, a loss of both CRD pumps f occurs. What are two (2) undesirable effects of this loss and l what makes these effects undesirable? (2.0) l l

l I

QUESTION 4.05 (2.00)

c. The control room is filling with a noxious vapor from en undetermined source. The Shift Supervisor decides to implement I the " Toxic Gas in Control Room" procedure. What IMMEDIATE ACTIONS are required of operating personnel by this procedure? (1.0) l L

b. The above actions are determined to be inadequate and the control room is to be evacuated. What actions should be taken, j if possible, prior to leaving the control room, by the Control l

Room Operators? (1.0) l l

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

i

-di__EBQQEQUBES_:_UQBdokt 6BUQBdoLt_EMEBQENQ1.60Q PAGE 13 B6010LQQ1QaL_QQUIBQL QUESTION 4.06 (1.S0)

c. A gaitronics' announcement is made concerning a Class Bravo fire.

What is meant by a Class Bravo fire? '

(1.0)

b. If the-fire involves radioactive material (s), an individual should not remain in the area for longer than _________ without respiratory protection. '

(0.5)

QUESTION 4.07 (1.00)

Boron inj ec t ion into the reactor is required per E0P-1, RPV Control, and the Standby Liquid Control S,ystem is incapable of inj ect ing into the RPV. Describe basically how boron inj ection to the RPV is cccomplished under the above circumstances. Specific procedural steps ARE NOT required. (1.0)

QUESTION 4.08 (1.S0)

What are the Emergency Dose Exposure Limits per EPIP 5.7.12, Emergency Radiation Exposure Control, for:

a. Sampling under accident conditions (0.5)

b. Corrective or protective actions (0.5)

c. Life-saving actions (0.5)

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

At__EBQQEQUBES_:_NQBd8Lt_ABNQBdeLt_EMEBQEUQ1_60Q PAGE 14 BSDIQLQQ106L_QQUIBQL

, QUESTION 4.09 (2.00)

The plant is operating at power with both recirculation pumps operating at minimum speed. ONE pump trips and backflow is NOT catsblished in the idle jet pumps, i

i 0. In this condition, how will indicated core flow compare with actual core flow? (0.5)

! b. In the condition above, in what region of the power / flow map will the plant appear to be operating? (0.5)

c. What action must the operator take to ensure that a correct core flow signal is used by the process computer for core parameter calculations? (0.5)

d. What action should be taken to avoid or control abnormal neutron flux oscillations (not actions to observe flux oscillations, but to control them)? (0,5)

(***** END OF CATEGORY 04 *****)

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

NRC LICENSE EXAMINATION HANDOUT EQUATIONS, CONSTANTS, AND CONVERSIONS 6 = rii*Cp *deltaT 6=U*A*deltaT P = Po*10sur*(t) P = P *et /T SUR = 26/T T = 1*/p + (p-p)/X p T=1/(p-p) T = (p-p)/Xp P = (Keff-1)/Keff = deltaKeff/Keff p=1*/Reff+jeff/(1+1T) .

A = In2/tg = 0.693/tg K = 0.1 seconds-1 I = Io*e-"*

CR = S/(1-Keff) '

2 R/hr = 6*CE/d feet Water Parameters 1 gallon = 8.345 lbm = 3.87 liters 1 ft3 = 7.48 gallons Density 9 STP = 62.4 lb Heat of vaporization = m/ft3 = 1 gm/cm3 970 Btu /lbm Heat of fusion = 144 Btu /lbm 1 atmosphere = 14.7 psia = 29.9 inches Hg.

Miscellaneous Conversions I curie = 3.7 x 101U disintegrations per second 1 kilogram = 2.21 lbm I horsepower = 2 54 x 103 Btu /hr 1 mw = 3.41 x 105 Btu /hr 1 inch = 2.54 centimeters degrees F = 9/5 degrees C + 32 degrees C = 5/9 (degrees F - 32) 1 Stu = 778 ft-lbf

l It__EBINQlELER_QE_URGLEaB_EQWEB_LLoNI_QEEBoI1QUt PAGE 15 IBEBdQQ1NadlGat_BEBI_IB861EEB_oNQ_EL91D_ELQW ANSWERS -- COOPER -85/09/26-GRAVES, D.

ANSWER 1.01 (2.00)

o. At the 100% rod line (0.5) with the recirc pumps at minimum speed (0.5).

b. Void percentage decreases to counter the increased negative reactivity from the fuel temperature coefficient (1.0) or similar explanation.

REFERENCE +

GE Reactor Physics Review, pg 52 ANSWER 1.02 (2.00)

- control rod movement

- recirculation flow changes

- fuel temperature changes

- fission product poison chenges

. void content change (pressure changes)

(5 at 0.4 each)

REFERENCE GE Reactor Theory Review, og 51 ANSWER 1.03 , (2.00)

BPV's open to pa'ss'the steam previously going to the turbine (0.5).

FGedwater temperature will decrease due to a loss of extraction oteam (0.5). Reactor power will increase due to the decrease in j feedwater temperature (0,5). Reactor pressure will increase j due to the increase in reactor power (0.5). l REFERENCE '

e BWR-4 Transients ANSWER 1.04 (2.00)

"d" is the correct answer (0.5). The highest xenon concentration will be in the center of the core (0.5), the high flux region from the previous operating period (0.5). This will increase the flux in tije area of the peripneral rods (0.5) thus increasing their worth. ,

Iz__EBINDIELE1_QE_UMGLEaB_EQWEB_ELaNI_QEEBoI1QUt PAGE 16 IBEBbQQ10ad10St_BEaI_IBaNSEEB_auR_EL91R_ELQW ANSWERS -- COOPER -85/09/26-GRAVES, D.

' REFERENCE GE Reactor Physics Review, pg 36-37 ANSWER 1.05 (1.00)

T=1-star /p + (B-p)/tp (0.1) no for prompt critical neglect delayed term & T=1-star /p (0.1) 1* = 10EE-04 seconds (accept 50 to 150 microseconds) (0.1) p~ .005 (0.1)

T= 10EE-4/.005 = 0.02 seconds (0.25)

P/Po=eEEtime/T (0.1)

P/Po=eEE50 = 5.18 X 10EE21 (0.25)

Numerical values do not have to be exact for full credit. Reasonable casumptions accepted.

REFERENCE GE Reactor theory Review, pg 20 ANSWER 1.06 '(1.00)

Cbtain corresponding temperatures from steam tables by interpolation:

1000 peig = 546.3 deg F (0.25) 250 psig = 406.0 deg F (0.25)

Ostermine the temperature change: 546.3 - 406.0 = 140.3 deg F (0.25)

Determine the rate of cooldown: 140.3/1.75 hr = 80.2 deg/hr (0.25)

-REFERENCE Steam Tables ANSWER 1.07 (1.00) d (1.0)

REFERENCE GE Thermodynamics, Heat Transfer and Fluid Flow, pg 7-91

Iz__EBIUCIELE1_QE_NVQLEaB_EQWEB_ELaNI_QEEBallQNt PAGE 17 IMEBdQQ1Nad101t_BEaI_IBaNSEEB_aND_ELu1R_ELQW ANSWERS -- COOPER -85/09/26-GRAVES, D.

ANSWER 1.08 (1.00)

- height of the column of water above the eye of the pump

- amount of subcooling of the above water or reactor dome pressure

- irreversible flow losses in the suction line (3 at 0.33 each)

REFERENCE GE Thermodynamics, Heat Transfer and Fluid Flow, pg 7-93 ANSWER 1.09 (1.00)

The hydrogen generation becomes measurable when the cladding temperature becomes elevated (> 2200 deg F). Adequate core cooling prevents the required temperature from being reached. (or similar explanation) (1.0)

REFERENCE CNS Mitigating Core Damage, Gas Generation Section ANSWER 1.10 (1.00)

Now fuel would have the longer thermal time constant (0.5) due to the increase in fuel to clad contact in exposed fuel (0.5).

REFERENCE General Electric Heat Thermodynamics, Heat Transfer, and Fluid Flow pg 9-131, question 34 Problem Solutions, pg 9-5 ANSWER 1.11 (1.00)

Transition boiling is avoided by maintaining MCPR above the operating or safety limit (1.0).

REFERENCE GE Thermodynamics, Heat Transfer and Fluid Flow, pg 9-93

- 21__EL8MI_DE11GN_INCLUQ1NG_18EEll_6NQ_EMEBGENC1_111IEd1 PAGE 18 ANSWERS -- COOPER -85/09/26-GRAVES, D.

ANSWER 2.01 (2.00)

c. Reactor building MG set room - Manual Control. building cable spreading room - manual Control room entrance - manual Turbine building switchgear room - manual Main generator - manual Turbine bearing #1 area - automatic Turbine. bearing #2 area - automatic Turbine bearing #3 area - automatic (6 areas at C.15 each, 6 initiations at 0.1 each)

b. - Reset button on the sprinkler control and fire alarm panel in the. Control Room (0.25)

- Manual pu'shbuttons in the NW corner and on the North wall of the turbine generator operating floor shield wall (0.25)

REFERENCE

~0P 2.2.2, Carbon Dioxide System, Rev 14, pg 2,5.

ANSWER 2.02 (2.00)

n. Allows the 0/G to be manually started and stopped from the

. Control Room-(0.5).

b '. When in LOCAL, disconnects all remote control and automatic start signals to the diesel (0.25) as well as Control Room indication for the D/G (0.25). u

c. Allows ~ local shutdown of the 0/G (0.25) if the Control Mode Selector Switch is in LOCAL (0.25),

d. Allows shutting down the D/G under any conditions (0.5)

REFERENCE LSOP 2.2.20, Standby AC Power System (D/G), Rev 19, pg 14 s

J t

e 2t__EL8Sl_QEllGU_lUDL90lNQ_18EEIl_8UQ_EdE8QEUQ1_1111Ed1 PAGE 19 ANSWERS -- COOPER -85/09/26-GRAVES, D.

ANSWER 2.03 (2.00)

a. No, the torus suction valve opens first, then the ECST suction valve shuts (0.5).

b. No, the steam supply block valve (RCIC-M0-131) shuts (0.5),

c. Yes (0.5)

d. No, the discharge and inj ection valves (RCIC-MO-20 and 21) do not close on an isolation (0.5).

REFERENCE SOP 2.2.67, RCIC, Rev 24, pg 2-4 RCIC System Description, pg 5 ANSWER 2.04 (1.50)

a. All units that are running will stop automatically (0.5).

b. REC C0.5)

c. High area temperature near CRD hydraulic piping (0.5)

REFERENCE SOP 2.2.40, HVAC Drywell Cooling, Rev 7, pg 2, 4 ANSWER 2.05 (1.50)

c. - dilution (0.5)

- recombination (0.5)

b. Due to the possibility of combustion within the vacuum pump (0.5)

REFERENCE Offgas and Augmented Offges System Description, pg 8, 10 i

l I

2t__EL6NI_QESIGN_INCLUQ1NG_18EEI1_8NQ_EUEBQENQ1_1111Ed1 PAGE 20 ANSWERS -- COOPER -85/09/26-GRAVES, D.

ANSWER 2.06 (1.50)

c. Temperature control is manual or automatic (0.5). REC temperature is adj usted by varying the service water flow through the REC heat exchangers (0,5).

b. When any of the Core Standby Cooling Systems (Core Spray, RHR, HPCI, or RCIC) are placed into service (0.5). Also accept Group VI isolation.

REFERENCE REC System Description, pg 16 ANSWER '2.07 (1.50)

n. Reactor building exhaust plenum (0.5)

b. - HPCI gland steam condenser exhauster (0.5)

- Drywell and torus purge exhaust (0.5)

REFERENCE S8GTS System Description, pg 6, Figure 1 ANSWER 2.08 (2.00)

c. NORMAL BACKUP Compressor A: REC TEC Compressor 8: TEC REC Compressor C: TEC REC (6 at 0.25 each)

b. Manual (0.5)

REFERENCE Plant Air System Description, pg 6 ANSWER 2.09 (1.00)

The coolant flows from the flow control valve into the cooling water header (0.33) and into the insert line of each individual CRD (0.33). From there it flows through the CRD into the reactor

_ vessel.(0.33).

l 1

2t__EL6MI_DE11GN_INGL90lN9_S6EEIY_680_EdF.BQEN91_SYSIEd3 PAGE 21 ANSWERS -- COOPER -85/09/26-GRAVES, D.

REFERENCE CRD System Description, pg 21-22, Figure 18 4

J t

t

SA__lN11BQMEUI1_60Q_QQUIBQL1 PAGE 22 ANSWERS -- COOPER -85/09/26-GRAVES, D.

ANSWER 3.01 (3.00)

RED GREEN BLUE

a. off on on

b. off on off

c. on off off

d. off on off

o. on off off (15 at 0.2 each)

REFERENCE Nuclear System Pressure Relief System Description, pg 4,6,8,9 ANSWER 3.02 (1.50)

Ycs, it is normal (0,5). The drive piston moves the RPIS magnet past the "00" reed switch and actuates only the green full-in " overtravel" reed switch (1.0).

REFERENCE CRD System Description, pg 18 RMC System Text ANSWER 3.03 (2.00)

c. - SRM downscale (0.5)

- Retract Not Permitted (0,5)

b. - SRM Inop (0.5)

- SRM Upscale (0.5)

REFERENCE Instrumentation Operating Procedure 4.1.1, SRM's, Rev 6, pg 2

x 9.

. 21__IN1189dENI1_680_GQNIBQL1 PAGE 23:

ANSWERS -- COOPER -85/09/26-GRAVES, D.

r ANSWER 3.04 (1.50)

ROD BLOCK MODE SWITCH POSITION

- Service platform jib crane. loaded Startup/ Hot Standby or Refuel Refueling , platform.over core Startup/ Hot Standby

- Refueling platfrom over core with any of-its 3 hoists loaded Refuel

- Refueling platform over core with fuel grapple not fully up Refuel (3 blocks at 0.25 each, 3 positions at 0.25 each)

REFERENCE Instrumentation Operating Proce' dure 4.3, Reactor Manual Control System, Rev-10,_ pg ,6, 7

' ANSWER 3.05 (2.50)'

O. -

JPost-accident monitor A or LI-91A. Located on panel 9-3 Post-accident monitor B or LI-918. Located on panel 9-3

- Wide range for vessel flooding or LI-86. Located on panel 9-4

- ECCS Yarway A or LI-85A. Located on panel.9-5

- ECCS Yarway B'or.LI-85B. Located-on panel 9-5 GEMAC feedwater control.A or LI-94A. Located on panel 9-5

- -GEMAC feedwater control-B or LI-948. Located on panel 9-5 GEMAC-feedwater control C or LI-94C. Located on panel 9-5 Narrow Range Recorder or LR-6-97. Located on panel 9 -

Wide range during shutdown for core coverage or LR-6-98. Located.

on panel 9-5 (8 indications required at-.125 each, locations at .125 each)

NOTE:. LI-91A'a-B: Fuel Zone, Rosemount, Wide Range, Wide Range Yarways LI-85A a.B: Narrow Range Yarways LI-94A, B,& C: GEMAC, Narrow Range GEMAC These are alternate names for the above-instruments.

b. 164.19 inches.. Accept 160 - 170 inches (0.5)

REFERENCE Instrumentation Operating Procedure 4.6.1, Reactor Vessel Water Level Indication, Rev 9, pg 2, LAttachment "C" to the above procedure

21__INSIBRUENIS_8NQ_GQUIBQL3 PAGE 24 ANSWERS -- COOPER -85/09/26-GRAVES, D.

ANSWER 3.06 (2.00)

c. True (0.5)

b. False (0.25). Bypass valves will open to maintain pressure at its present value (0.25).

c. False (0.25). Reactor power is controlled with control rods and recirculation flow (0.25).

d. True (0.5).

REFERENCE DEH System Description ANSWER 3.07 (2.00)

e. Log Red Monitor High - High (0.25) Log Red Monitor = Off Gas Red Mon.

Log Rad Monitor Downscale (0.25)

Log Rad Monitor Inop (0.25)

Any combination of one of the above in each channel will activate the Off Gas Valve Timer , i.e. a High - High in channel A and a Downscale in 9h annel B (0.25).

b. OG - 254AV or ERP inlet valve (0.25)

OG - 902AV or A0G outlet valve to the ERP (0.25)

OG - A0-12 or holdup pipe drain valve (0.25)

OG - A0-13 or Offges filter drain valve (0.25)

REFERENCE Instrumentation Operating Procedure 4.7.2, Air Ej ector Of f Gas Radiation Monitoring System, Rev 9, pg 1, 2 Off Gas System Description, figure 2 ANSWER 3.08 ( .50) 4 or turbine control fluid pressure (0.5)

REFERENCE RPS Systen Description, pg 9

4 t__EBQQEQMBES_:_UQBd6Lt_6HUQBd6Lt_EMEBQEUQ1_600 PAGE 25

-86DIQLQGlQ6(_QQUIBQL ANSWERS -- COOPER -85/09/26-GRAVES, D.

ANSWER 4.01 (1.50)

c. I

b. 4

c. 2 (0.5 each)

REFERENCE SOP 2.2.4, Communications, Rev 13, pg 6 ANSWER 4.02 (2.'00)

With the control rod at position 48 (0,5), initiate a rod out signal (0.5). If the control rod is uncoupled: 1) The indicating number 48 will go out, 2) the background will go blank, 3) and the rod full indicating light will go out. 4) ROD OVERTRAVEL and 5) ROD ORIFT annunciators will alarm when the timer cycle is complete.

(4 of 5 indications required at 0.25 each)

REFERENCE Instrumentation Operating Procedure 4.3, Reactor Manual Control System, Rev 10, pg 12 ANSWER 4.03 (1.50)

- APRMs are not downscale

- Vacuum is established

- Reactor pressure is greater than 825 psig or Main Steam Channel A and B Low Pressure alarms on 9-5 clear

- APRM gain adj usted to 1.0 (3 required at 0.5 each)

REFERENCE Gsneral Operating Procedure 2.1.1, Cold Startup Procedure, Rev 39, pg 10 o

SA__EBQQEQQBEH_:_NQBdakt_8HNQBd8Lt_EdEBQENQ1_8NQ PAGE 26 B6DIQLQQ196L_QQNIBQL ANSWERS -- COOPER -85/09/26-GRAVES, D.

ANSWER 4.04 (2.00)

The loss of the CRD pumps causes a loss of cooling water flow C0.5),

which shortens CRD seal life (0.5), and also allows the CRD hydraulic cccumulators to slowly depressurize (0.5) which reduces the scram ccpability of the reactor (0.5), especially at low pressures.

!Also accept loss of seal purge to RWCU and RR pumps which can shorten coal life and loss of ability to move control rods with RMCS as hydraulic pressure is lost.

REFERENCE Abnormal Procedure 2.4.1.1.4, Loss of CR0 Pump, Rev 5, pg 2 System Procedure 2.2.8, Control Rod Drive System, pg 2 ANSWER 4.05 (2.00)

c. - Start the Control Room Ventilation Booster Fan BF-C-1A (0.33)

- Secure the Control Room Ventilation Supply Fans SF-C-1A & B (0.33)

- Essential Control Room personnel obtain self contained breathing apparatus and use as necessary (0.33)

b. - Scram the reactor (0.25)

- Verify all control rods inserted (0.25)

- Trip-the main turbine (0.25)

- Ensure the reactor feed pump turning gear control switches are in Auto (0.25) ,

REFERENCE Abnormal Procedure 2.4.8.5, Toxic Gas in Control Room, Rev 2, pg 1 Emergency Procedure 5.2.1, Shutdown from Outside the Control Room, Rev 13, pg 1 ANSWER 4.06 (1.50)

o. Flammable or combustible liquids, flammable gases, greases, and similar materials (1.0).

b. 2 minutes (0.5)

REFERENCE Emergency Procedure 5.4.1, General Fire Procedure, Rev 17, pg 1, 5

F. .

_.4

-st__88QQEQMBEl_:_NQ858kt_8HNQBd8kt_EdE89ENQ1_8NQ PAGE 27 I- '88Q19LQQlQ8L_QQU189L

!, ANSWERS -- COOPER -85/09/26-GRAVES, D.

I ANSWER 4.'07 (1.00)

Filling the RWCU demineralizers with borated water and -inj ecting this -

via the RWCU-system (1.0).

REFERENCE Emergency Procedure 5.2.14, Alternate Means to Inj ect Boron to RPV, Rev 0 ANSWER 4.08 (1.50)

a. 5 REM (0.5)

b. 25 REM (0.5).

c. 75 REM C0.5)

REFERENCE Emergency Plan Implementing Procedure 5.7.12, Emergency Radiation Exposure Control, Rev 5, Attachment "A" ANSWER 4.09 (2.00)

c. Indicated core flow will be lower than actual (0.5) b.- To the left of the natural-circulation line (0.5)

c. Th'e operator must enter a substitute value for core flow equal to the natural circulation flow for the existing power level (0.5).

d. Insert' control rods until power is~< or = 75% load line (0.5).

REFERENCE

-General Operating Procedure 2.1.10, Station Power Changes, Rev 9, Attachment "A" '

General Operating Procedure 2.1.15, Reactor Recirculation Pump Startup and Shutdown, Rev 13, pg 3, 4 Abnormal Procedure 2.4.2.2.1, Trip of Reactor Recirculation Pumps, Rev 12,-

pg 1-3

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

REACTOR TYPE: _RWB-qed _________________

DATE ADMINISTERED: _QELQQl2h________________

EXAMINER: _QB6ME3z_Qu______________

1 APPLICANT: _________________________

INSIBUCIl0NH_IQ_8EELIQ6N11 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 lecst 80%. Examination papers will be picked up six (6) hours after

, the examination starts.

% OF CATEGORY  % OF APPLICANT'S CATEGORY

__MaLUE_ _IQIaL ___1QQBE___ _VaLME__ ______________CaIEQQBY_____________

_15tDQ__ _2519Q ___________ ________ S. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS

_lEzQQ__ _25zQQ ___________ ________ 6. PLANT SYSTEMS DESIGN, C O NT R01. ,

AND INSTRUMENTATION

_15tHQ__ _25zQQ ___________ ________ 7. PROCEDURES _ NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 4 _15tQQ__._25tDQ ___________ ________ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS

_hDzQQ__ 1DQ1Dg ___________ ________ TOTALS FINAL GRADE _________________%

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

i

~

AEEE55 5YI5~555U 5Uk5~~~~~~~~~~~~~~

-Sz__IBEQBl_QE_UUCLEaB_EQWEB_EL6NI_QEE86IlQNt_E(ylQSt_6NQ PAGE 2 IBEBdQQ1865102 QUESTION 5.01 (1.00)

~How much (by what factor) would power increase in one second in a prompt critical reactor at EOL? (1.0)

QUESTION 5.02 (2.00)

The reactor is operating at 75% power. Recirculation flow is subsequently increased to provide 100% power and 100% flow.

Describe the effect that the above increase will have on each of the below items. Continue your description until steady state conditions are reached.

a. Core Void Content (1.0)

b. Core Net Reactivity (1.0)

QUESTION 5.03 (2.00)

Concerning control rod worths during a reactor startup from 10G% PEAK XENON versus a startup under XENON-FREE conditions, which statement is correct? JUSTIFY YOUR CHOICE. (2.0)

a. PERIPHERAL control rod worth will be LOWER during the PEAK XENON startup than during the XENON-FREE startup.

b. CENTRAL control rod worth will be HIGHER during the PEAK XENON startup than during the XENON-FREE startup.

c. BOTH control rod worths will be the SAME regardless of core Xenon conditions.

d. PERIPHERAL control rod worth will be HIGHER during the PEAK XENON startup than during the XENON-FREE startup.

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

St__IBEQBl_QE_NUQLE68_EQWEB_EL6NI_QEEB6IlQUt_ELUlQ1t_60Q PAGE 3 IEEBUQQ1NadlQ1 QUESTION 5.04 (3.00)

.The reactor is operating at 100% power when HPCI inadvertantly initiates. Describe the response of the following parameters during the transient, including why the parameter changes as it does.

Assume NO SCRAM occurs. Continue your description until steady state conditions are reached.

c. Reactor Pressure (1.0)

b. Reactor. Water Level (1.0)

c. Feedwater Flow (1.0)

QUESTION 5.05 (1.50)

List the three (3) primary sources of hydrogen during an accident. (1.5)

QUESTION 5.06 (1.00)

Give two (2) reasons or factors that cause the fuel temperature coefficient to be negative. (1.0)

QUESTION 5.07 (2.00)

Using the steam tables, indicate whether water at each of the following is SUBC00 LEO, SATURATED, or SUPERHEATED. (2.0)

a. 200 psig, 387.7 F

b. 1000 psig, 544.6 F

c. 1200 psig, 603.9 F

.d. 900 psig, 531.1 F

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

'5z__IBEQBX_QE_NVQLEaB_EQWEB_EL8SI_QEEBal10Nt_ELU10st_8NQ PAGE 4 IBEBdQQXNad103 QUESTION 5.08 (1.00)

What are the three (3) maj or parameters affecting the NPSH of

-the recirculation pumps? (1.0)

QUESTION 5.09 (1.00)

Pellet / cladding interaction, a stress corrosion type cracking that can occur during rapid power increases in irradiated fuel, occurs due to two (2) factors. List both factors. (1.0)

QUESTION 5.10 ( .50)

a. TRUE or FALSE? APLHGR is a function of fuel burnup. (0.25)

b. -FILL IN THE BLANKS. The APLHGR limit is based on keeping __Ci)__

below __(11)__ degrees F. (0.25)

(***** END OF CATEGORY 05 *****)

ht__EL8HI_1111gd1_QgSigNt_qquIggLt_aup_lugIgudgNIaligU PAGE 5 QUESTION 6.01 -(2.00)

a. What are six (6) areas served by the Low Pressure CO2 System?

Indicate whether the area is covered AUTOMATICALLY or requires MANUAL actions. (1.5)

b. After the initial 50 second discharge of CO2, how are subsequent discharges initiated (2 methods or locations required)? (0.5)

QUESTION 6.02 (3.00)

Ooscribe the indication (ON or 0FF) of the three (3) indicating lights for a Safety / Relief Valve under each of the following conditions: (3.0)

a. No actuation signals present, valve shut

b. The valve is leaking s ig n if ic ant ly

c. The valve handswitch is in OPEN

d. The valve opens due to high reactor pressure

e. ADS logic actuated QUESTION 6.03 (1.50)

Following a reactor scram, the four rod display position goes blank, but the green full-in light on the full core display for that control rod is lighted. Is this normal? If so, explain why it occurs. If not, describe the probable cause. (1.5)

QUESTION 6.04 (1.50)

What are six (6) of the eight (8) permissives that must be met in order for'the emergency diesel generator breaker to close and re-energize its associated bus? (1.5)

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

~

Ez__ELaNI_SYSIEUS_RES1Qut_QQNIBQLt_6NQ_lN118UdENI611QN PAGE 6 i QUESTION 6.05 (2.00)

The HPCI system was in a normal standby lineup when an initiation signal was received. For each of the following situations, state whether it is indicative of a malfunction or not, and explain your choice. All parameters were within their respective normal ranges when the initiation signal was received.

I

a. During pump startup, discharge pressure is 75 psig, flow is O gpm and the minimum flow valve (HPCI-MO-25) is SHUT. (0.5)

b. The GLAND SEAL CONDENSER CONDENSATE PUMP indicates NOT running. (0.5)

c. The GLAND SEAL CONDENSER BLOWER indicates NOT running. (0,5)

d. The HPCI pump suction torus valve (HPCI-MO-58) strokes open. (0.5)

QUESTION 6.06 (2.00)

Describe four (4) rod blocks associated with refueling equipment.

Include required Mode Switch position (s), as applicable. (2.0)

QUESTION 6.07 (2.00)

List eight (8) indications of reactor vessel water level in the Control Room. Include on which panel each indication is located. (2.0)

, QUESTION 6.08 (1.00)

o. LPRM accuracy may be effected significantly by depletion of the fission chamber. How does this depletion affect the LPRM readings (higher or lower than actual power)? (0.5)

b. A' Core Thermal Power and APRM Calibration program (00-3) is performed and shows APRM A with a Gain Adj ustment Factor of 1.03. What does this tell the operator about the relation-ship between actual and indicated power on APRM channel A? (0.5)

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

Zz__EBQGgDUBgS_:_uQadeLt_agsgBuatt_gegBGgggy_aGQ PAGE 7 BaQ19LQQ1 gal _GQNIBQL t

QUESTION 7.01 (2.50)

During a reactor and plant startup, primary containment oxygen content must be ___(a)___ within ___(b)___ hours after ___(c)___. (2.5)

QUESTION 7.02 (2.00)

With the reactor operating at power, a loss of both CRD pumps occurs. What are two (2) undesirable effects of this loss and what makes these effects undesirable? (2.0)

QUESTION 7.03 (1.00)

With the reactor operating at power, an accumulator trouble alarm sounds. Investigation reveals a nitrogen leak on the accumulator.

The control rod is FULLY INSERTED. Is that control rod OPERABLE?

If not, how can it be made operable? (1.0)

QUESTION 7.04 (2.50)

c. The control room is filling with a noxious vapor from an undetermined source. The Shift Supervisor decides to implement the " Toxic Gas in Control Room" procedure. What IMMEDIATE ACTIONS are required of operating personnel by this procedure? (1.0)

b. Tne above actions are determined to be inadequate and the control room is to be evacuated. What actions should be taken, if possible, prior to leaving the control room, by the Shift Supervisor and the Control Room Operators? (1.5)

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

l

Zz__EBQQEQUBES_:_UQBdakt_eBNQBd8L&_EMEBGENQ1.6NQ PAGE 8 BoQ10LQQ1 gal _QQUIBQL QUESTION 7.05 (1.50)

c. A . g a i t r o n i c r. announcement is made concerning a Class Bravo fire.

What is meent by a Class Bravo fire? (1.0)

b. If the fire involves radioactive material (s), an individual should nat remain in the area for longer t h a n _________ w it ho ut respiratory protection. (0.5)

-QUESTION 7.06 (1.00)

Boron injection into the reactor is required per E0P-1, RPV Control, and the Standby Liquid Control System is incapable of inj ect ing into the RPV. Describe basically how boron inj ection to the RPV is accomplished under the above circumstances. Specific procedural steps ARE NOT required. (1.0)

QUESTION 7.07 (2.50)

Emergency Dose Exposure Limits are defined for three (3) categories in EPIP 5.7.12, Emergency Radiation Exposure Control. List the three categories AND their associated exposure limits. (2.5)

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

1

Z1__EBQQEQUBEH_:_NQBd8Lt_6BNQBd8(t_EMEBQENQ1_6NQ PAGE 9 88DIQLQQlG8L_QQNIBQL

-QUESTION 7.08 (2.00)

The plant is operating at power with both recirculation pumps operating at minimum speed. ONE pump trips and backflow is NOT established in the idle j et pumps.

a. In this condition, how will indicated core flow compare with actual core flow? (0.5)

b. In the condition above, in what region of the power / flow map will the plant appear to be operating? (0,5)

c. What action must the operator take to ensure that a correct core flow signal is used by the process computer for core parameter calculations? (0,5)

! d. What action should be taken to avoid or control abnormal neutron flux oscillations (not actions to observe flux oscillations, but to control them)? (0.5) l l

I i

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

Hz__6DMIN11IBoIIYg_PBQQEQQBESt_QQNQlIlQN1t_6NQ_Lidll611QNS PAGE 10 QUESTION 8.01 (1.50)

Match the emergency signal (a- c) with its tone description. (1.5)

_____a. Emergency alarm 1. distinct steady tone

_____b. Fire alarm 2. one steady up and down tone

_____c. All clear 3. ringing gong

4. distinct pulse tone QUESTION 8.02 (1.50)

For each of the following three Government Communications Systems, state the color of the telephones, the person (by title or organization) who should establish the communication, and whether or not the system is in the Control Room.

-c. Health Physics Network (HPN) (0,5)

b. Emergency Notification System (ENS) (0.5)

c. Nebraska State Patrol (NSP) Hotline (0,5)

QUESTION 8.03 (1.50)

Briefly describe how a TEMPORARY setpoint change to a Technical Specification setpoint is performed administratively. Include the appropriate documentation and approvals / concurrences required. (1.5)

QUESTION 8.04 (2.50)

a. Define a LIMITING CONTROL ROD PATTERN. (1.0)

b. It has been concluded that such an occurrence could result only from en abnormal. withdrawal sequence at a high power level.

List or describe three (3) operations or conditions that could result in an abnormal withdrawal sequence. (1.5)

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

St__6DdlNISIB8IIVE_PBQQEQUBEft_QQNQlllQN1t_6NQ_LidlI8IlQN1 PAGE 11

. QUESTION 8.05 (2.00)

Pertaining to CNS Procedure 0.9, Equipment Clearance and Release Orders:

.a. List the three (3) methods given that may be used for verifying the position of manual valves in the main flow path of safety related equipment when they are returned to service. (1.5)

b. Who may sign the CLEARANCE RELEASED BY blank when the person who signed the CLEARANCE ISSUED TO blank is not on site and the tags need to be picked.up? (0.5)

QUESTION 8.06 (2.00)

-The. Shift Supersivor shall immediately (or as soon as possible) notify the Operations Supervisor and Division Manager of Nuclear Operations if any of.FOUR (4) GENERAL SITUATIONS exist. What are these four gen-eral situations? (2.0)

QUESTION 8.07 (1.50)

Identify.three (3) situations or conditions that warrant using a small magnetic base red arrow on a white background as a highlighting device. (1.5)

. QUESTION 8.08 (1.50)

The reactor is. operating at 100% power when a turbine trip occurs.

A review-of the alarm printer shows the reactor scrammed on APRM high flux. Does this violate Technical Specifications? If so, JUSTIFY YOUR ANSWER. (1.5)

QUESTION 8.09 (1.00)

Once emergency conditions have been detected and classified, initial state and local notifications must be completad within ___(a)___ and initial NRC notification must be made within ___(b)___. (1.0)

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

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

NRC LICENSE EXAMINATION HANDOUT EQUATID,'iS, CONSTANTS AND CONVERSIONS 6=m*C*deltaT p d=U*A*deltaT P = Po*10sur*(t) P = P *et /T SUR = 26/T T=1*/p+(p-p)/Xp T=1/(p-p) T = ($-p)/Xp P = (Keff-1)/Keff = deltaKeff/Keff p = 1*/TKeff + feff/(1+ X.T) .

X = In2/tg = 0.693/tg K = 0.1 seconds-1 I = Io*e'"*

CR = S/(1-Keff) 2 R/hr = 6*CE/d feet Water Parameters 1 gallon = 8.345 lbm = 3.87 liters 1 ft3 = 7.48 gallons Density 9 STP = 62.4 lbm/ft3 = 1 gm/cm3 Heat of vaporization = 970 Stu/lbm Heat of fusion = 144 Btu /lbm '

1 atmosphere = 14.7 psia = 29.9 inches Hg.

Miscellaneous Conversions 1 curie = 3.7 x 101U disintegrations per second 1 kilogram = 2.21 lbm I horsepower = 2 54 x 103 Btu /hr 1 mw = 3.41 x 105 Btu /hr 1 inch = 2.54 centimeters degrees F = 9/5 degrees C + 32 degrees C = 5/9 (degrees F - 32) 1 Stu = 778 ft-lbf s

-5t__IBEQBl_DE_NVCLE6B EQWEB EL8MI_QEEB8Il0Nt_ELU10S t_6NQ~ PAGE 12-IBEBdQQ1Nad191 ANSWERS -- COOPER -85/09/26-GRAVES, D.

ANSWER. 5.01 (1.00)

'T=1-star /p + '(B-p)/tp- (0.1) so for prompt. critical neglect delayed term & T=1-ster /p (0.1) 1* = 10EE-04 seconds (accept 50 to 150 microseconds) (0.1) p~ .005 (0.1)

T= 10EE-4/.005 = 0.02 seconds (0.~25)

P/Po=eEEtime/T (0.1)

P/Po=eEE50 = 5.18 X 10EE21 (0.25)

Numerical valu~es do not have to be exact for full credit. Reasonable casumptions accepted.

REFERENCE GE Reactor. Theory Review, pg 20 ANSWER 5.02 (2.00)

e. -VoidsLin'itially decrease (0.25) as the increased flow moves l the boiling boundary farther into the core (0.25). As power l increases,- the rate of boiling increases (0.25), and the boiling

-boundary returns to near its original level (0.25).

b. The decrease in void content initially causes a positive reactivity addition (0.25). As power increases, negative reactivity is added due to increased void formation (0.25) and the increased

. fuel temperature (0.25). Net reactivity returns to 0 at steady l' ' state conditions-(0.25).

REFERENCE GE Reactor Physics Review, Figure 66 ANSWER 5.03 (2.00)

! "d" is the correct answer (0.5). The highest xenon concentration will tue in the center of the coro(0.5), the high flux region from the previous operating periodCO.5). This will increase the flux ,

in the area of the peripheral rods (0.5) thus increasing their worth.

REFERENCE

~GE Reactor Physics Review, pg 36-37 i

I

^,ee.m q .. ge:g -ee m- e mm,%-,,,q,, .,ap-w ay-e,. mqis-.--.yy i e-- ----y,-yggws- ,m-pr g,y.mm- ,-.-&,-----?------t p- sv'7-p ,--w- - , - - gre..-. -

5t__IB5981_QE_UQQLEaB_EQWEB_EL6NI_QEEB61100t_ELQ1QSt_68Q PAGE 13 IBEBdQQ188dICS ANSWERS -- C00PL' . -85/09/26-GRAVES, D.

ANSWER 5.04 (3.003-

c. Reactor pressure would increase (0.5) due to the increase in reactor power caused by the increased subcooling (0,5).

b. Reactor level will increase (0.5). A level error must be generated to reestablish steady state conditions in the FWLCS (0.5).

c. Feedwater flow will decrease (0.5). The HPCI inj ect ion is providing a portion of the required feed for the reactor and this is not sensed by the Feed Flow detectors (0.5).

REFERENCE BWR-4 Transients ANSWER 5.05 (1.50) -

Zirconium-water reaction

-- Steel-water reaction

- Radiolysis of water (3 at 0.5 each)

REFERENCE Mitigating Core Damage, Gas Generation S .w. t i o n ANSWER 5.06 (1.00)

- Doppler broadening (0.5)

- Self-shiciding (0.5)

REFERENCE GE Reactor Physics Review, pg 30 ANSWER 5.07 (2.00)

c. Saturated

b. Subcooled

c. Superheated

d. Subcooled (0.5 each)

- 5t__IBEQB1-QE_NUGLE88_EQWEB_ELeNI_QEEBallQNt_ELQ1QSt _6NQ PAGE 14 IBEBdQQ1NadIQ1 ANSWERS --- COOPER -85/09/26-GRAVES, D.

REFERENCE Steam-Tables

> ANSWER 5.08 (1.00)

-: height.of the column of water above the eye of the pump ramount of.subcooling of the above : water .or reactor dome pressure

- irreversible flow losses in the suction line (3 at 0.33 each)

REFERENCE GE Thermodynamics, Heat Transfer and Fluid-Flow, pg 7-93 1

- ANSWER 5.09 (1.00)

High: localized stresses caused by differential pellet / clad. expansion-(0.5)

Gnd'the presence of_1embrittling f iss ion p ro,1uct species (0.5) such as Iodine and Cadmium.

' REFERENCE GE. Thermodynamics,' Heat. Transfer and. Fluid Flow, pg 9-124 s.

ANSWER 5.10 ( .50)

c. TRUE.- (0.25)

b. i.- clad temperature- (0.125)'

11. 2200 (0.125)

REFERENCE GE Thermodynamics, Heat Transfer and Fluid Flow, pgs 9-71'and 9-72 t

6%

. ;(

l Ez__EL8NI_SISIEdS_QESIGNt_QQUIBQLt_6NQ_INSIBQUENI611QN PAGE 15 ANSWERS -- COOPER -85/09/26-GRAVES, D.

ANSWER 6.01 (2.00) 4

e. Reactor building MG set room - Manual Control building cable spreading room - manual Control room en'.r ance - manual Turbine building switchgear room - manual Main generator - manual Turbine bearing #1 area - automatic Turbine bearing #2 area - automatic Turbine bearing #3 area - automatic (6 areas at 0.15 each, 6 initiations at 0.1 each)

b. - Reset button on the sprinkler control and fire alarm panel in the Control Room (0.25)

- Manual pushbuttons in the NW corner and on the North wall of the turbine generator operating floor shield well (0.25)

REFERENCE OP 2.2.2, Carbon Dioxide System, Rev 14, pg 2,5 ANSWER 6.02 (3.00)

RED GREEN BLUE

c. off on en

b. off on off

c. on off off

d. off on 'off

c. on off off (15 at 0.2 each)

REFERENCE Nuclear System Pressure Relief System Description, pg 4,6,8,9 ANSWER 6.03 (1.50)

Yes, it is normal (0,5). The drive piston moves the RPIS magnet past the "00" reed switch and actuates only the green full-in " overtravel" reed switch (1.0).

REFERENCE CRD System Description, pg 18 RMC System Text

.e-ht__EL8NI_11SIEdi_DE11ENt_GQUIBQLt_6ND_INEIBudENIal1QN PAGE 16

- ANSWERS -- COOPER -85/09/26-GRAVES, D.

ANSWER 6~.04 (1.50)

-- Emergency transformer 4.16 KV ACB open (1FS/1GS)

- Diesel generator lockout relay ~ reset

- Diesel generator tie breaker lockout relay reset 1

- Bus 1F (1G) tie breaker 1AF (IBG) OR Bus 1A (18) tie breaker 1FA (1GB) open

- Diesel-generator at rated speed Diesel generator at rated voltage (94 %)

-7 Sustained' loss of critical' bus voltage > 5 seconds

--Diesel Control Isolation Switches in REMOTE

.00 Control Power and Maintenance Lockout Keylock in "0N" D/G bkr Selector Switch in AUTO (6 required at 0.25 each)

REFERENCE .

')

SOP 2.2.20, Standby AC Power System (D/G),.Rev 19, pg 25 l ANSWER _ 6'.' 0 5 (2.00)

n. . Malfunction.' The minimum flow valve should open when the initiation signal is received. Any explanation that demonstrates this fact is acceptable (0.5).

b.' This-is not . indicative of a problem. The condensate pump cycles on level in the hotwell and is independent of the HPCI initiation signal (0.5).

c. . Malfunction. The blower should start on the initiation signal and run continuously (0.5).

- d .- Malfunction. The torus suction valve should not receive an open signal due to the HPCI initiation (0,5).

- REFERENCE' SOP :2.2.33,'HPCI, Rev 26, pg 4-7 1

kt__PL6dI_SISIEd1_QESlQUt_QQUIBQLt_6NQ_lN11BQUENI611QU PAGE 17 ANSWERS -- COOPER -85/09/26-GRAVES, D.

ANSWER 6.06 (2.00)

R0D BLOCK MODE SWITCH POSITION

- Service platform jib crane loaded Startup/ Hot Standby or Refuel

- Refueling platform over core Startup/ Hot Standby

- Refueling ~ platfrom over core with any of its 3 hoists loaded Refuel

- Refueling platform over core with fuel grapple not fully up Refuel (0.25 for each block, 0.25 for each position)

REFERENCE Instrumentation Operating Procedure 4.3, Reactor Manual Control System, Rev 10, pg 6, 7 ANSWER 6.07 (2.00)

Post-accident monitor A or LI-91A. Located on panel 9-3 Post-accident monitor 8 or LI-918. Located on panel 9-3 Wide range for vessel flooding or LI-86. Located on panel 9-4 ECCS Yarway A or LI-85A. Located on panel 9-5 ECCS Yarway B or LI-858. Located on panel 9-5 GEMAC feedwater control A or LI-94A. Located on panel 9-5 GEMAC feedwater control 8 or LI-948. Located on panel 9-5 GEMAC feedwater control C or LI-94C. Located on panel 9-5 Narrow Range Recorder or LR-6-97. Located on panel 9-5 Wide range during shutdown for core coverage or LR-6-98. Located on panel 9-5 (8 indications required at .125 each, locations at .125 each)

NOTE: LI-91A & B: Fuel Zone, Rosemount, Wide Range, Wide Range Yarways LI-85A & B: Narrow Range Yarways LI-94A, B, & C: GEMAC, Narrow Range GEMAC These are alternate names for the above level indicators.

REFERENCE Instrumentation Operating Procedure 4.6.1, Reactor Vessel Water Level Indication, Rev 9, pg 2

ht__ELaNI_11SIEUS_QE1100t_QQUIBQLt_aND_INSIBUUENI8l1QU PAGE 18 ANSWERS -- COOPER -85/09/26-GRAVES, D.

ANSWER 6.08 (1.00)

c. Readings will decrease as the detector ages (0.5).

b. The APRM reading is lower than actual power (0.5).

REFERENCE Nuclear Performance Procedure.10.5, LPRM Calibration, Rev 17, pg i Nuclear Performance Procedure 10.1, APRM Calibration, Rev 14, pg 4

LZz__EBQQEQUBE1_:_NQB56Lt_6HNQB56Lt_EMEBGENQ1_6NQ PAGE 19 88QIQLQGlQ6L_CQNIBQL ANSWERS -- COOPER' -85/09/26-GRAVES, D.

ANSWER 7.01- (2.50)

c. < 4%

24 b.

c. ~ going to the:RUN Mode (0.833~each)

' REFERENCE

-General Operating Procedure 2.1.1, Cold Startup Procedure, Rev ~39, pg 11 ANSWER 7.02 (2.00)

The' loss of the CRD pumps causes a loss of cooling water flow-(0.5),

.which shortens CRD seal life (0.5), and also allows the CRD hydraulic-

,cccumulators to slowly depressurize (0.5) which-reduces the scram cepability of .the reactor (0.5), especially at low pressures.

Also accept loss:of-seal purge to the RWCU and RR pumps which can

! shorten seal life, and loss of ability to move control rods with RMCS as hydraulic pressure ~is lost.

REFERENCE-Abnormal Procedure 2.4.1.1.4, Loss of CRD Pump, Rev 5, pg 2 System Procedure 2.2.8, Control Rod Drive System, pg 2 t

ANSWER 7.03 (1.00) lThe control rod is INOPERABLE (0.5). To make it operable, the-leak must be repaired and the accumulator trouble cleared (0.5)

REFERENCE CNS Technical Specification 3.3.A.2.e

Z1__EBQQEQuBEH_ _UQBdebt_8BNQBdebt_EdEBQENQ1_aNQ PAGE 20 BaQ10LQQIQaL_QQNIBQL ANSWERS -- COOPER -85/09/26-GRAVES, D.

ANSWER- 7.04 (2.50)

a. - Start the Control Room Ventilation Booster Fan BF-C-1A (0.33)

- Secure the Control Room Ventilation Supply Fans SF-C-1A & B (0.33)

- Essential Control Room personnel obtain self contained breathing apparatus and use as necessary (0.33) b'. - Announce the event over the station paging system (0.25)

- Direct operations personnel to assemble in some designated area area (0.25)

- Scram the reactor (0.25)

- Verify all control rods inserted (0.25)

- Trip the main turbine (0.25)

- Ensure the reactor feed pump turning gear control switches are in Auto (0.25)

REFERENCE Abnormal Procedure 2.4.8.5, Toxic Gas in Control Room, Rev 2, pg i Emergency Procedure 5.2.1, Shutdown from Outside the Control Room, Rev 13, pg 1 ANSWER 7.05 (1.50)

c. Flammable or combustible liquids, flammable gases, gresses, and similar materials (1.0).

b. 2 minutes (0.5)

REFERENCE Emergency Procedure 5.4.1, General Fire Procedure, Rev 17, pg 1, 5 ANSWER 7.06 (1.00)

Filling the RWCU demineralizers with borated water and inj ecting this via the RWCU system (1.0).

REFERENCE Emergency Procedure 5.2.14, Alternate Means to Inj ect Boron to RPV, Rev 0

Z1__EBQQEQUBE1_ _NQBdakt_aRUQBdaL&_EMEBQENQ1_8NQ PAGE 21 BaQ10LQQIQaL_QQUIBQL

-ANSWERS -- COOPER -85/09/26-GRAVES, D.

ANSWER 7.07 (2.50)

- Sampling under accident conditions (0.33): 5 REM (0.5)

- Corrective or protective actions (0.33): 25 REM (0.5)

- Life saving actions (0.33): 75 REM (0.5)

REFERENCE Emergency Plan Implementing Procedure 5.7.12, Emergency Radiation Exposure Control, Rev 5, pg 1, Attachment "A" ANSWER 7.08 (2.00)

a. Indicated core flow will be lower than actual (0.5)

b. To the left of the natural circulation line (0.5)

c. The operator must enter a substitute value for core flow equal to the natural circulation flow for the existing power level (0.5).

d. Insert control rods until power is < or = 75% load line (0.5).

REFERENCE General Operating Procedure 2.1.10, Station Power Changes, Rev 9, Attachment "A" General Operating Procedure 2.1.15, Reactor Recirculation Pump Startup cnd Shutdown, Rev 13, pg 3, 4 Abnormal Procedure 2.4.2.2.1, Trip of Reactor Recirculation Pumps, Rev 12, pg 1-3

at__aQUINISIB811VE_EBQQEDMBESt_QQURIIIQUnt_oND_L1011811QUE PAGE 22 ANSWERS.-- COOPER -85/09/26-GRAVES, D.

ANSWER 8.01 (1.50)

c. 1

b. 4

c. 2 (0.5 each)

REFERENCE S0P 2.2.4, Communications, Rev 13, pg 6 ANSWER 8.02 (1.50)

c. - beige (0.1)

- NRC (0.2)

NOT in Control Room (0.2)

b. - red (0,1)

- DMNO, STA, SS, communicator (any of the four acceptable) (0.2)

- IN Control Room (0.2)

c. - green (0.1)

. Emergency Director (or communicator) (0.2)

- IN Control Room (0.2)

REFERENCE S0P 2.2.4, Communications,-Rev 14, pg 3, 11-13 Procedure 2.0.5, Steps II.B.1.d, pg 1 and Step III.E, pg 5

~ ANSWER 8.03 (1.50)

A Setpoint Change Request form (0.25) marked TEMPORARY (0.25) will be filled out and signed by 2 SR0's (1.0).

REFERENCE Instrumentation Operating Procedure 4.0.1, Instrument Setpoint Control, Rev 0, pg 3

p .o .

o 'Ri__8DdlN1118811VE_EBQGEQVBEli_GQUQIIIQuit_8NQ_L1011811QN1 PAGE 23

. ANSWERS -- CdOPER -85/09/26-GRAVES, D.

ANSWER ~ 8.04. (2.50)

-e. One which contains a rod which, if completely withdrawn (0.25),

could result-in a MCPR of less than 1.07 (0.75).

b. - Interchange of normal control rod patterns (0,5)

_ Establishment of special control rod patterns as-an aid for identifying core regions'having failed fuel assemblies (0.5)

- Establishment of special control rod patterns resulting from control-rod drive system malfunctions-(0.5)

' REFERENCE Nuclear Performance Procedure 10.13,. Limiting Control Rod Pattern Determination, Rev 7, pg 1 ANSWER 8.05 (2.00)

a. l .- Position light indication in the control room if the-valve is so equipped (0.5)

2. LocallyLby a second operatpr (0.5)

3. Satisfactory performance of a flow operability surveillance-(0.5)

b. One of the individual's supervisors (0.5)

REFERENCE CNS. Procedure 0.9, Equipment Clearance and Release Orders, Rev 2, pg 2, 3 ANSWER 8.06 (2.00)

- Entry into a limiting condition for operation as required by Technical Specifications (0.5)

- Any plant condition which requires any of the Emergency, Abnormal, or Emergency Operating Procedure to be implemented (0.5)

- Any continuing of f-normal condition which limits plant power capability or could limit power production if corrective action was not taken (0.5)

- 'A 1 condition for entry into the Emergency Plan Implementing Procedures (0.5)

REFERENCE Conductaof Operations Procedure 2.0.1, Operations Department Policy, Rev 2, pg 2

at__6DdlN1118611VE_EBQGEQUBEft_QQNQlllQU$t_8NQ_Lidll611QU$ PAGE 24 ANSWERS -- COOPER -85/09/26-GRAVES, D.

ANSWER 8.07 (1.50)

- Any time a control is placed in an out of normal position

- Annunciators and other indications that may be out of service

- Annunciators or other indications that may require particular attention

- Nuisance alarms (3 required at 0.5 each)

REFERENCE Conduct of Operations Procedure 2.0.3, Control Room Conduct and Manning, Rev 0, pg 2 ANSWER 8.08 (1.50)

Yes (0.5). A safety limit shall be assumed to be exceeded when a acram is accomplished by a menas other than the expected scram signal (1.0). The scram should have occurred due to the Turbine =

Stop Valve Closure.

REFERENCE CNS Technical Specifications 1.1.0, pg 6, 12, 13 ANSWER 8.09 (1.00)

a. 15 minutes (0.5)

b. I hour (0.5)

REFERENCE EPIP 5.7.6, Notification, Rev 5, pg 1

>