Text

Requalification Program Exam Repts 50-317/85-32 & 50-318/85-27 on 851104-08.Exam Results:All Operators Passed Oral Exam,One Operator Failed Section of Written Exam & One Senior Operator Failed Written Exam
	ML20136F285
Person / Time
Site:	Calvert Cliffs
Issue date:	12/17/1985
From:	Dudley N, Keller R, Kister H; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20136F264	List: IR 05000317/1985032; ML20136F261;
References
	50-317-85-32, 50-318-85-27, NUDOCS 8601070333
	Download: ML20136F285 (64)
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U.S. NUCLEAR REGULATORY COMMISSION Region I 50-317 Docket Nos.

50-318 85-32 (0L)

DPR-53 Report Nos.

85-27 (OL)

License Nos. DPR-64 Licensee:

Baltimore Gas and Electric Company Post Office Box 1475-Baltimore, Maryland 21203 Facility:

Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Inspection.At: Lusby, Maryland Inspection Conducted:

No er 4-6, 1985 ( hd

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Inspector:

6W. tF. DudTey, L(a Reactor Engineer (Examiner)

date

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Reviewed By:

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R. Keller, Chief, Projects Section 1C

~date

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/2[/7/Ef Approved By:

Kister, Chief, Projects Branch No.1 date

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Summary:

An e aluation of the facility requalification program was made by substituting an NRC prepared written and oral examination for the facility annual requalification examination.

The -NRC prepared examination was admin-istered to 20% of the licensed operators who had not been examined by the NRC in the previous two years. All operators passed the oral examinations.

One reactor operator failed a section of the written examination, and one Senior Reactor Operator failed the written examination.

All other operators passed the written examination. As a result of this evaluation, the requalification program was found to be adequate, and no generic weaknesses were identified.

l 8601070333 851230 PDR ADOCK 05000317-

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a DETAILS A.

Scope The evaluation of the Requalification Program consisted of replacement of the facility annual requalification examination with NRC prepared written and oral examinations. These examinations were adeinistered to a selected sample of 20% of the licensed operators who had not been examined by the NRC in the previous two years. The evaluation criteria was dependent on the number of operators who passed the NRC administered examinations.

An NRC written examination was administered to six RO's and five SR0's.

Four RO's and two SR0's were on the operating shift which had its requal-ification examination scheduled for the week of November 4,1985. These operators were also administered NRC oral examinations.

The remaining licensed personnel, who were examined, were selected to ensure a sampling of staff licenses from different job classifications and different in{tial license issuance dates.

These operators were administered either a written or an oral examination. The written examinations were constructed to be 60% the length of licensing examinations and sampled a wide variety of areas. More detailed questions were asked in the areas covered in the previous year's requalification program.

NRC oral examinations were administered to six R0's and five SRO's. The oral examinations provided the same coverage as licensing examinations, and were conducted in accordance with NUREG-1021, Operator Licensing Examiner Standard.

The examinations lasted three to four hours.

The evaluation criteria used is detailed in NUREG-1021, Chapter ES-601, and states that for a program to be evaluated as satisfactory more than 80% of the evaluated operators must pass all portions of the NRC admin-istered examination.

B.

Findings The eleven operators who were administered oral examinations, by three NRC examiners, were evaluated as having a satisfactory level of knowledge. Of the eleven operators who were administered the written examination, one R0 failed Section 2, but passed overall, and one SRO failed Section 7, and failed overall.

All other operators passed all other sections of the written examination.

Individual weaknesses were identified on the oral and written examinations, however, no generic weaknesses were identified.

The requalification program was evaluated as satisfictor __

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C.

Exit Interview NRC Personnel N. Dudley, Lead Reactor Engineer (Examiner)

D. Trimble, Resident Inspector Facility Personnel L. Russell, Plant Superintendent R. Denton, General Supervisor, Training and Technical Services J. Hill, Supervisor of Operations Training J. Yoe, Principal Operator Instructor Summary of Comments Made at Exit Interview:

The NRC reviewed the number and type of examinations which had been admin-istered and noted that no generic weaknesses had been noted during the oral examinations.

.

A discussion of when examination results would be available was held.

Licensee stated that issuance of results after two months would impact the requalification program. The NRC stated that an attempt would be made to process examination results promptly but due to other commitments the examination results might be delayed.

The licensee commented on the NRC administered written examinations. The licensee stated that 50% of the material in the examinations had not been covered in the requalification program in the last two years, that 50% of the examinations was not operational oriented, and that 50% of the examin-ation required information that an operator would have available to him in the control room.

Also, the licensee noted that due to the shortened examination, an operator would lose more credit if he answered a single question incorrectly.

The licensee questioned when the NRC would begin using learning objectives to write examinations.

The NRC noted the difference in philosophy between the utility and NRC prepared examinations and lef t open the question of whether the annual examination should be a tool for evaluating operator's retention of the training provided by the previous year's requalification program or a tool for identifying weak areas which should be taught in future requalifica-tion programs.

The NRC stated that examinations were beginning to be written using NUREG-1122, Knowledges and Abilities Catalog for Nuclear Power Plant Operators, which was published in July 1985, but the requal-ification examinations had not been referenced to the Catalog.

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Changes to Written Examinations:

Utility supplied comments, made during the two hour examination review are contained in Attachment 3.

The comments were considered during grading of the examination, however, not all comments resulted in changes to the answer keys.

Answer No.

Change Reason 1.01 Change " shutoff head" to Corrects wording of answer and

" flow" and "1400 psta" to carrects pressure to the pressure

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"1100 psia". Add "(Design specified in question. Provides flow of HPSI pump is 345 design specifications not required gpm at 1075 psig)."

for full credit.

1,07 a & e Add " verify flow".

Allows more generalized statement of parameter being verified.

2.03 c Change "548F" to "557F".

Corrects temperature to Tave specified in reference.

2.03 Add "SD: Reactor Regulating Provides reference which specifies Reference System, p. 16".

value for quick open signal.

2.05 b Change " reduce flow" to Corrects logic for reducing RCS

' increase flow"; "less temperature while on shutdown water" to "more water"; and cooling.

" bypass".to " pass through".

3.03 Add "(4.75 psig)" and "(685 Allows technical specification psia)".

values to be used for setpoints.

3.03 Add "T.S. Table 3.3-4".

Provides reference for technical Reference specification setpoints.

3.04 Add " computer indication; Expands answer to allow for other metrascope digital".

means of verifying proper rod withdrawal.

3.05 b Add "and high power trip".

High power trip selects highest of NI power and delta T power.

4.01 d Change "240 F" to "280".

Changes temperature requirement for initiation of shutdown cooling due to technical specification requirement ~~

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Answer No.

Change Reason 4.02 b Add "or loss of all RCP".

Recognizes that E0P-200 is used for Natural Circulation.

4.03 d Add " bus indicating lights". Provides additional method of checking bus voltages.

6.06 a Change to read "What two Corrects question to match Question ESFAS or ASFAS actuation facility nomenclature of the signals..."

actuation signals.

6.06 Add " Preliminary Provides reference.

Reference Notification of Event or Unusual Occurrence PNO-I-85-55; DCS No.

50309/850808, date August 1985."

5.03 Add "or; yes if steam is Provides alternate correct-action removed from SG to drop for assumption that manual action Tsat below Tave".

is taken.

5.06 b Add "(Reactor would trip Provides automatic action which on high power)".

would be expected to terminate transient. Not required for full credit.

5.08 a Add "CEA withdrawal; E0C Provides additional conditions xenon transient; power which might cause flux tilt.

reduction using boron".

6.04 c Add "or; No. Th is averaged Recognizes that the second and TM/LP trip setpoint channel of TM/LP may not trip would not increase above since insufficient information was operating pressure".

provided in the question.

7.01 b Change to " pressure is Corresponds to facility controlled by saturation instructional methods, temperature of the pressurizer".

7.05 d Add " bus voltage lights".

Provides additional method of checking bus voltag l-

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Answer No.

Change Reason 8.07 b

. Add "or; make proper entries Requires less specific information in transient log".

for full credit.

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8.08 c Change to "No report, meets Refueling water tank level requirement for Mode 5".

requirement is less restrictive l

in Mode 5.

i Attachments:

1.

Written Examination and Answer Key:

R0 2.

Written Examination and Answer Key:

SR0 3.

. Facility Comments on Written Examination

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NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION l

FACILITY:

CALVERT CLIFFS

_________________________

REACTOR TYPE:

PWR-CE

_________________________

DATE ADMINISTERED: 85/11/06

_________________________

EXAMINER:

DUDLEY

_________-_______________

APPLICANT:

_________________________

INSTRUCTIONS TO APPLICANT:

__________________________

Use separate paper for the answers.

Write answers on one side only.

Staple question sheet en top of the answer sheets.

Points for each question are indicated in parentheses after the question. The passing 3rade requires at least 70% in each category and a final ersde of at least 80%.

Exaniination papers will be picked up six (6)

hours after the examination star ts.

% OF CATEGORY

% OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY

________ ______

___________

________ ___________________________________

14 0 23.33

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1.

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

___1_0___ ___1__3

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14 0 23 3 2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 17.00 28.33

________ ______

___________

________ 3.

INSTRUMENTS AND CONTROLS 15.00

___1_0

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PROCEDURES - NORMAL, ABNORMAL, 25 0

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EMERGENCY AND RADIOLOGICAL CONTROL 60.00 100.00 TOTALS

________ ______

___________

________

FINAL GRADE _________________%

All work done on this examination is my own. I have neither given not received aid.

EPPLEC5UII5~5YGU5YURE~~~~~~~~~~~~~~

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1.

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE

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QUESTION 1.01 (1.50)

What would be the approximate leak rate if a Loss of Coolant Accident occured and all automatic safety injection systems-functioned properly and RCS pressure stabilized at 1100 psia?

Justify your answer.

QUESTION 1.02 (1.00)

After operating for 1 week at 100% power the reactor is taken from 100% power to a just critical condition at 10E-4 % power.

What rod motion is necessary to maintain this power level for the next two hours?

Explain.

QUESTION 1.03 (2.00)

a.

Upon what THREE RCS parameters is the DNB Heat Flux (CHF)

dependent?

b. At what location in the core, top, bottom, or middker is the fuel the furthest from DNB? (i.e. Where is the DNB Ratio the largest?)

Justify your answer.

QUESTION 1.04 (1.00)

HOW and WHY would each of the following parameters compare 10 minutes and 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after confirming natural circulation flow resulting from a trip from extended 100% power operations accompainied by a loss of all Reactor Coolant Pumps.

a.

Delta T b.

Flow rate OUESTION 1.05 (2.50)

The ratio of the PU239 and Pu240 atoms to U235 atoms increases over core life. Explain the effect this ratio change has on

.

the following:

a.

Delayed neutron fraction (1.0)

b.

Reactor period (l'.0)

c.

Doppler Temperature Coefficient (1.0)

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

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PRINCIPLES OFm NUCLEAR POWER PLANT OPERATION, PAGE

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o QUESTION 1.06-

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You have just=corapleted a reactor startup and power level is at'the point of adding heat. For the following situations, INDICATE WHERE final power level will be in reference to

. initial power level (HIGHER, LOWER, OR THE SAME) and EXPLAIN your answer. (Assume the core is at mid-life, no operator action

and treat each situation separately).

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a.

Steam dump pressure setting is raised by 20 psis.

b.

A 1% steam leak develops outside of containment.

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c. An inadvertent 20 ppm boron addition is made.

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GUESTION 1.07 (3.00)

The five criteria: listed below are used to verify that natural circulation flow has been established.

E:< plain why each is important or what condition is.being checked.

l-a. 10 F to'50 F delta T b. Th constant or decreasing c.

Tc constant or decreasing

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d.

CET temperature consistant with Th

e.

Steamin3' rate'affects primary temperature

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(xxxxx END OF CATEGORY 01 xxxxx)

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PLANT' DESIGN INCLUDING SAFETY AND' EMERGENCY SYSTEMS PAGE

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-GUESTION 2.01 (2.00)

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Assume'that a gaseous radioisotope is dissolved in the reactor coolant system.' List the components in the flow path through

which this gaseous radioisotope could be. removed from the RCS, processed, and eventually released to the enviornment as part of a routine discharge.

QUESTION 2.02 (2.00)

For each operation. listed, state which mode of control is normally

?

used on.the Control Element Drive System.

a.

Exercising Shutdown Group CEA's for monthly surveillance.

b.

Withdrawing Shutdown Group CEA's during a reactor startup.

c.

Inserting Regulating Group CEA's during reactor shutdown.

d.. Recovering 4 dropped CEA.

QUESTION 2.03 (3.00)

A reactor trip-from full power has occurred causing the steam dumps and bypass valvesLto quick-open'

a.

At what' point will the dump valves so fully shut with the steam ^ bypass valves maintaining temperature?

(0.7)

.b.

After the dump valves are fully shut, the bypass valves fail shut.

How will the system function automatically to prevent the actuation of SG safety valves?- (Note any applicable setpoints.)

(0.8)

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c. What conditionscare necessry for the dump valves to quick-open AND how is.the system designed / constructed to accomp-lish ' Quick-Opening'?

(1.5)

'0UESTION 2.04 (2.00)

a.

Why are the CVCS letdown backpressure control valves needed during operations in hode 1.

(TWO reasons required).

b.

What operator. action is required if one of the letdown backpressure control valves failed open.

(xxxxx CATEGORY 02 CONTINUED ON NEXT PAGE xxxxx)

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2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE

00ESTION 2.05 (3.00)

a.

Using the attached diagram explain what the valve line up should be for normal shutdown cooling.

(1.5)

b.

Explain hou temperature would be reduced if the system was on shutdown cooling?

(0.75)

c.'What overpressure protection is provided for the primary just prior to establishing shutdown cooling?

(0.75)

QUESTION 2.06 (2.00)

a.

Why must the initiating SIAS signal be removed prior to stopping the diesel generator?

b.

How is the removal of the SIAS start signal to the diesel generator verified?

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

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INSTRUMENTS 6ND CONTROLS PAGE

____________________________

OUESTION 3.01 (2.00)

If a reactor trip signal was present. what effect would the simultaneous failure (to deenersize) of the Reactor Protection System (RPS) K-1 relay and K-2 relay have on the RPS?

What sould be done to correct the immediate problen?

A figure of the RPS breaker arrangement is provided.

QUESTION 3.02 (3.00)

Will the plant trip as a result of the following simultaneous instrument failures?

Explain your answers.

a.

SUR channels A and B fail high during a startup, when reactor is critical at 10 -6%.

b. SG-11 level channel A fails LOW and SG-12 level channel B fails HIGH while at 80% power.

c.

Loop 1 Tc channel A fails high and loop 2 Th channel B fails hi3h while at 80% power.

d.

The lower UIC detectors for safety channels D and D fail low at 50% power.

QUESTION 3.03 (3.00)

If during reactor plant operations at 95% power a feedline rupture were to occur inside the containment, what are the FOUR Engineering Safety Features (ESFs) that could possibly be actuated and what signals will cause these actuations? Include setpoints and logic.

QUESTION 3.04 (3.00)

The followins concern the control rod drive system.

a.

What effect would a lift coil failure have on rod withdrawal?

(1.5)

b. What means exist to determine whether a control rod in withdrawing properly? (Five required.)

(1.5)

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

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INSTRUMENTS AND CONTROLS PAGE

____________________________

GUESTION 3.05 (3.00)

a.

How man / Hot and Cold Les Temperature instruments are there in a single loop AND where are they located?

(0.8)

b.

How many of each type (T-hot and T-cold) are used for protec-tion AND what are they used for in the protection system?

(1.0)

c.

What specific SYSTEM (s) are controlled due to signals derived from loop t e ro p e r a t u r e s ?

(1.2)

GUESTION 3.06 (3.00)

What are FOUR different locations outside the main control room where it is possible to trip the Unit 2 turbine generator.

Explcin in terms of the EHC system how the turbine can be tripped from each location.

(*****

END OF CATEGORY 03

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4.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE G

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GUESTION 4.01 (2.00)

During cooldown to cold shutdown, prior to collapsing the bubble in the pressuriner, what should be the condition of each of the following components or indications?

a.

SIT's b.

RCP's c.

PORV's d.

RCS temperature e.

Pressurizer heaters and spray GUESTION 4.02 (3.00)

Under what conditions should each of the followins Emergency Operating Procedures be used?

a.

E0P-100, Reactor Trip b.

E0P-200, Loss of Off-Site Power / Natural Circulation c.

E0P-400, Excess Steam Demand d.

E0P-800, Fuctional Recovery Procedure OUESTION 4.03 (3.00)

Provide the two independent indications which would be used to complete each of the followins immediate actions contained in E0P 000.

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a.

Verify all CEA's are fully inserted.

b.

Verify pressuriner pressure stabilizes between 1850 and 2275 psia.

c.

Verify the reactor coolant system is subcooled greater than 30 F.

d.

Verify 4KV buses 11 or 14 energi=ed.

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

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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

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QUESTION 4.04-(3.00)

For each of the situations below,_ indicate whether the plant should

.be tripped immediately.

For situations which do not require an immediate trip explain at what point a reactor trip, if any, is required assimin3 conditions continue to deteriorate.

Assume plant has been operating for 1 week at 90% power.

Consider each situation separately.

.a.

The motor on the operating component cooling pump fails.

b.

It is discovered that containment integrity has been breached when a blind flange is found improperly secured.

c.

An unexplained dilution raises power by 5%.

d.

Instrument air pressure drops to 75 psis.

e.

The main journal bearing metal temperature is 230 F (5 F above the alarm set point) for the Unit 1 turbine.

f. The main journal bearing metal temperature is 225 F (5 F above the alarm set point) for the Unit 2 turbine.

QUESTION 4.05 (1 00)

State the two methods for restoring refueling pool level upon a cavity seal failure per AOP-6E recovery actions.

(Assuming a fuel assembly may be uncovered.)

OUESTION 4.06 (1.20)

List the four RMS alarms that are provided to help the operator identify a steam generator tube rupture event.

QUESTION

'4.07 (1.80)

A situation occurs where condenser water boxes need to be cleaned.

At one point it becomes necessary to remove two water boxes from cervice simultaneously.

What three. parameters should be observed prior to stopping the two associated circulating water pumps?

(*****

END OF. CATEGORY 04

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1.

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE

--~~isEER55is4RiCs-sEsi iEEssFEE Es5 FEUi5 FE5E

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ANSWERS -- CALVERT CLIFFS-85/11/06-DVDLEY ANSWER 1.01 (1 50)

Fr:d Leak rate would be som of charging pumps and shotofr-headrof the HPSI pumps at 14'00 p s i a. CO.73 120 3pm +

600 spm ~ 700 3pm. [0.63

( pr s u,v ra~ cr ms' omr* n p

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at IC'iFp;<p)

SD 7 and 8, SI systemst p 18, 90 ANSWER 1.02 (1.00)

Rod withdrawal CO.53 to maintain power due to nes. reactivity from xenon buildup. CO.53

.

REFERENCE CE-Nuclear Physics, Reactor Theory and Core Operating Characteristien.

p 204 ANSWER 1 03

'(2.00)

8.

Flow Temperature Pressure Power Cany 3 0 0 3 each]

b.

Dottom of CO.53, because this is where the temperature [0.4]

is the lowest and pressure the highest CO.13.

(1.0)

REFERENCE CE - Thermal Hydraulics, p 14 ANSWER 1 04 (1.00)

a. Delta T will be lower ofter 1 hr since there will be less decay heat to remove. CO.53 b. Flow rate will be lower after i hour sinco there will be less of a thermal driving head. CO.53 REFERENCE Roqualification Program 1905, Att. 3, p 10.

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1.

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE

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ANSWERS -- CALVERT CLIFFS-95/11/06-DUDLEY ANSWER 1.05 (2.50)

a.

Delayed neutron fraction decreases CO.53 because the beta

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is less for Pu229 as compared to U235. CO.03 (1re)

t b. Shorter reactor period CO.g3 because delayed neutron fraction

;

decreases. CO.93 ( 1 re)

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c. Doppler Coefficient is more negative CO.53 because Pu240 has

...;

a h' sher resonance eross section than U235. [0. 00 (1rO4

REFERENLE CE-Nucleur Physics, Reactor Theory and Core Operating Characteristics, p 153 - 156 ANSWER 1.06 (3.00)

a. Lower CO.2538 the steam dump pressure settins increase causes an RCS temperature increase CO.253. MTC CO.253 and FTC (Doppler) CO.253 both add negative reactivity to lower reactor power.

(1.0)

b.

Higher CO.3]l the increased flow will result in a lower RCS temperature CO.43. MTC will add positive reactivity and power will rise CO.33.

(1.0)

c. Lower CO.438 the negative reactivity inserted by the baron will cause power to decrease CO.63.

(1.0)

REFERENCE CF.- Nuclear Physics, Reactor Theory and Core Operating Characteristics, p

162-166, 170 ANOWER 1.07 (3.00)

Sufficient thermal driving head has been octablished; udir

'4 1'

a.

r b. Heat being removed from primary c. Heat being removed from primary d

Core is being cooled, saturation conditions have not been reached o. R C S a n d S G a r e c o u p l e d ;, w.ir r e n o w CO.6 each]

.

. -

-

. -.

.

..

- - - -

- -

-

.

- -

-4

-

__ - _ _ _ _ _ _ _ _ _ _ _ -_ _ - _ _ _ _ _ _ _

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _

_

-

,

.

1.

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE

~~~~isEEs559sARiEs-sEAi iEKssFEE An5 FEUi5 FE5E

ANSWERS -- CALVERT CLIFFS-85/11/06-DUDLEY REFERENCE Requalification program 1985, ATT.

2, E0P Objectives, p.

.

t

4

O S

A.m.

a

._m

.

2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE

_______________________________________________________

ANSWERS -- CALVERT CLIFFS-85/11/06-DVDLEY ANSWER 2.01 (2.00)

Lotdown to desassifier through the CVCS.

Tho desassifier removes the gas which is collected in the Waste Gas Surge Tank.

Cocpressors move gas to Waste Gas Decay Tanks.

I Vcnted through filters and RMS before reaching main vent.

l REFERENCE.

SD No. 14A Waste Gas System, p 2, Fig. A-1 ANSWER 2.02 (2 00)

a. Hanval Individual b. Hanval Group c. Manual Sequential d. Manual Individoal CO.5 each]

REFERENCE SD 60, CEDSI p 13, 14 ANSWER 2.03 (3.00)

a. The dump valves will 90 fully shut when error signal has oe-creased to 3 F.(Tave 535 F)

(0.7)

b. Dump valves will re-open on an 8 Degree error signal.

(Tave 540 F)

(0 8)

tr7 c.

Turbine tripCO.33 and 16 degree error signal (Tave 540*F)CO.33.

These conditions cause energizing of solenoidsCO.53, directing high pressure / volume air to actuators.CO.t3 (1.5)

Y REFERENCE SD 19, MS and MSIV systemi p 17 R cou. ca h'riatn rv. boir. s,

p/(

'

I E

_

, _ _

.

2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE

ANSWERS -- CALVERT CLIFFS-85/11/06-DUDLEY ANSWER 2.04 (2.00)

a.

Protect downstream purification equipment fron, overpressurization/ limit transients E0.53. Prevent upstrean letdown flow from flashing to steam CO.53.

b. Operator will isolate the failed valve and place other letdown letdown backpressure control valve in service.[1.03 REFERENCE SD 6, CVCSi p 15 ANSWER 2.05 (3.00)

a.

Letdown from hot les through 651 and 652 to suction LPSI C0.53 Common dircharge header splits to 306 and shutdown coolers [0.53 Out of discharge coolers through 657 and into LPSI injection header CO.53 winenSe tiaat b.

Throttle 657 to imduceeflow through SD coolers. E0.43 L-em4+ water will luft. p ass SD He a t E>:cha nge r i reducing RCS tenperature. [0.35]

Q PORV providb'uaprotection in MPT ENABLE mode. [0.753 c.

REFERENCE SD 7 and 8, SI and CS Systems; Fig.

A-2, Fig. A-3 OP 5, p8 ANSWER 2.06 (2.00)

c. The diesel generator would have a start failure, will not automatically restart, and is therefor out or service.C1.03 b. Observation of actuation modules in a 'non-tripped * condition before the DC are shutdown.[1.03 REFERENCE Encbling Objectives, GSO instructions 00-1

~.

.

..

.-

p a

,

.

,.,

>

,

>

,

...

3.

-INSTRUMENTS AND CONTROLS PAGE

.. ______________________ _____

,

' ANSWERS -- CALVERT CLIFFS-55/11/06-DUDLEY

,

!

ANSWER

3.~ 01 (2.00)

,

No.reacto'r trip. [0.73 Trip signal would nop be sent trip TCB's and CEDM would remain energized. [0.63 Manot11y trip plant.: E0.73 (2.0)

REFERENCE SD No. 59 RPSr. Fig A-2, p 39

-.

ANSWER 3.02 (3.00)

-

a.

No CO.353 not until power reaches.10 -4%.

[0.43

'b.-No E0.353 channels: auctioneer low fisnal therefore only

' channel A will trip. [0.43 c.

Yes CO.353 both TM/LP channels will trip CO.43 (One channel trips due to Qdnb driving set ppbbt high and the other channel

-

trips' due to Teal driving Pvar hish.)

~d.

Yes E0.353 both APD channels will trip E0.43

. REFERENCE SD'No.1 RPS, p 29, 31; Fig A-6, A-8

ANSWER 3.03.

(3.00)

'SIAS CO.43 and CIS CO.43 - High containment pressure E0.23, 14 77) 2. 8 psig E0.23, 2/4 E0.

4.25"43.

yj CSAS CO.43 - High containment pressure CO.2J psis E0.23 2/4 E0.13 SGIS'E0.43 - Low S/G-pressure E0.23 653 psia [0.13 2/4 E0.13

,

( id,

REFERENCE.

SD 63, ESFA System; p 88, 133 T. S.

TROL E - 3.3 ' V

.'

>

s y

.

>

r M

'

e

,

<

h

<bi r, w-wp

< -

.

3.

INSTRUMENTS AND CONTROLS PAGE

s.,

____________________________

ANSWERS -- CALVERT CLIFFS-85/11/06-DUDLEY ANSWER 3.04 (3.00)

a.

Rod will not move up CO.42 because the lift coil is used to raise the upper gripper CO.35], rod won't fall or insert CO.4]

due to action of lower gripper CO.353.

(1.5)

b.

1.

Metrascope RPI c c et r., r /J

.s,

+ :.,;

2.

Group deviation

_ g.

~' '

'~' "'

,

.,

3.

CEA motion-inhibit 4.

Pulse counter 5.

Rod bottom lights CO.3 each]-

(1.5)

REFERENCE SD 60, CEDS; p 9, 25-29 ANSWER 3.05 (3.00)

c. Five T-hot in each loop located between Rx. Vessel and Steam Generator. CO.43 Three T-cold per loop located between Coolant pump and R::.

Vessel. CO.43 (0.8)

b. Four in each hot les and two in each cold legCO.5]. They provide temperature (and Delta-t) signals to develop the TM/LP trip setpoint.C0/93 (1.0)

R a n.. a ec.. :.t sc.:v i<:;r a >1

~.

c. Cor trol rods, Pressurizer level, and Steam dump and by-pass s-cem.

CO.4 each]

(1.2)

REFERENCE SD 62, RCS Instrumentation; p3

.

b k

h;

.

3.

INSTRUMENTS AND CONTROLS PAGE

____________________________

ANSWERS -- CALVERT CLIFFS-85/11/06-DUDLEY

>

ANSWER 3.06 (3.00)

.

1.

From the front standard of the turbine. [0.33

,

Trip valve releases hydraulic fluid from auto stop head'er which allows 2-CV-8235 to open and dump oil from under control valve.EO.45]

2. Inside the front standard of the turbine. [0.33 Open 2-CV-8235 to dump oil from under control valves.EO.45]

any other two locations or actions which would cause trip signal to EHC system. For example *

3.

EHC pump control panel. [0.3J Securing Pumps which causes drop in auto stop header pressure and opening of 2-CV-8235.

[0.45]

4.

Shorting out solenoids on 2-SV-8235, 2-SV-8236, or 2-SV-8237. [0.33 Opening solenoids allows valves to open which will dump oil from the control valves. CO.45]

REFERENCE SD. 238, Fig. 238-5, Fis. 238-14

.

-..

. -

L

'4. ' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND.

PAGE 18.

~~~~ E555L55icat 55sTR5t R

-

~~~~---~~~--~~----

____________________

ANSWERS -- CALVERT CLIFFS-85/11/06-DUDLEY ANSWER-4.01 (2.00)

a. Outlet HOV's closed-and breakers tagged open

'b.

RCP's. stopped and tassed c.-PORV's in MP1 ENABLE; override switches in ' override shut" d. ' Le s s th a n ? rte F A

e. In manual CO.4 each]

REFERENCE

'OP-5, p 6-8

~ ANSWER 4.02 (3.00)

a. Following Reactor _ Trip with no conplications b.-Reactor. shutdown, feed and condensate system,'and all 4 RCP

' unavailable because of loss of of f-site power ca un ce 44-4Ct'

c.

Unisolable leak upstream of either MSIV d. One or more safety functions not met _and/or diagnosis is not possible

[0.75 each3 REFERENCE E0P 100,: p1 EOP 200, p 1-EOP-400,.p 1

'

E0P.800,fp 1 ANSWER 4.03-(3.00)

a.

Rod bottom lights (bottom reed switch)

- Hetroscope (reed switch)

~b. 2-of 4 safety channels

<c.

Subcooling monitor PZR pressure and Tc, The or Tave Ld. Breaker' indication Current meters

- Voltase meters

6n s c.m a rm-c. "IE8 0 0.375 each]

REFERENCE EOP 000,'p 5-7

-

%

--

a'z N2 - -

a

.

v

.

4.

PROCEDURES - NORMAL, ABNORMAL, EMLRGENCY AND PAGE

~~~~E5D ULb5EC5L E5sTs5L

-

--~~~~~~-~~~------------

____________________

ANSWERS -- CALVERT CLIFFS-85/11/06-DUDLEY ANSWER 4.04 (3.00)

a.

Trip if not restored in 10 min. [0.3] or alarm is received on RCP thrust bearing temperature. (>195 F) [0.23 b. No trip. CO.23 Trip if not in hot standby in-6 hours. [0.3]

c. No trip. [0.23 Trip if dilution raises power to RPS high power trip set point. [0.3J d. No trip. [0.23 Trip when pressure reaches 50 psis. E0.33 e.

No trip. CO.2] Trip at 240 F.

[0.3]

f.. Trip reactor. CO.5]

REFERENCE AOP 4, p 1 AOP 6, p 1-2 TS 3.6.1.1 AOP 7, p4 AOP 7D, p2 ACP 7E Unit 1,

p3 ADP 7E Unit 2, p 3 ANSWER'

4.05 (1.00)

Line UP a spent fuel pool pump taking a suction on alternate RWT. [0.53 Line up a LPSI Pump recirculating spilled RCS fluid from the containment floor through the core and out the leak. CO.53 REFERENCE LOR-320-1-85 ANSWER 4.06 (1.20)

Condenser Off Gas Blowdown Tank Blowdown Recovery Main Steam Line

[0.3 each]

REFERENCE LOR-300-6-85 m.

,

.

4.-

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

~~~~ d656L5EiEdt E5sTRUL R

-

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

____________________

ANSWERS -- CALVERT CLIFFS-85/11/06-DUDLEY

.AWSWER 4.07 (1.80)

Absolute backpressure (condenser vacuum)

Msnimum differential pressure between adjustment hoods Gross MW load EO.6 each]

REFERENCE 0I-14

.

__.,J

.

L./k V Ed'T cL1FF3 g0 gE Q WA L

-

lif/f CROSSREfERENCEr-t-18 TEST PAGE

QUESTION VALUE REFERENCE

,

__

_-------_-

01.01 1.50 DUD 0001120 01.02 1.00 DUD 0001121 01.03 2.00 DUD 0001122 01.04 1.00 0U00001123 01.05 2.50 DUD 0001124 01.06 3.00 DUD 0001125 01.07 3.00 DU00001126

_____-

14.00 02.01 2.00 DUD 0001127 02.02 2.00 DUD 0001128 02.03 3.00 0U00001129 02.04 2.00 DUD 0001130 02.05 3.00 DUD 0001131 02.06 2.00 DUD 0001132

-___-_

14.00

-

03.01 2.00 DVD0001133 03.02 3.00 DUD 0001134 03.03 3.00 DUD 0001135 03.04 3.00 DUD 0001136 03.05 3.00 DUD 0001137 03.06 3.00 DUD 0001138

______

17.00 04.01 2.00 DUD 0001139 04.02 3.00 DUD 0001140 04.03 3.00 0U00001141 04.04 3.00 DUD 0001142 04.05 1.00 DUD 0001165 04.06 1.20 DVD0001166 04.07 1.80 0000001167

______

15.00

--___-

__--_-

60.00

_u

w HASTER

-

..

,

lTTRC'/3 a7ettf b

-

U.

S.

NUCLEAR REGULATORY COMMISSION SENIDE REACTOR OPERATOR LICENSE EXAMINATION FACILITY:

CALVERT CLIFFS

REACTOR TYPE:

PWR-CE

--_-_--_--_____--____-_-_

DATE ADMINISTERED: 85/11/06

EXAMINER:

DUDLEY

-__-_---_-_--_-----~~----

APPLICANT:

_________________________

INSTRUCTIONS TO APPLICANT:

_--~~----------

Use separate Paper for the answers.

Write answers on one side only.

Staple question sheet on top of the answer sheets.

Points for each question are indicated in parentheses after the question. The passing grade requires at leest 70% in each category and a final grade of at least 80%.

Exanination papers will be picked up six (6)

hours after the examination starts.

% GF CATEGORY

% OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY

------

----------------------------------=.

15.00 25.00

-_______ ___---

_----------

5.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDSr AND THERMODYNAMICS 15.00

_'5.00

________

6.

PLANT SYSTEMS DESIGNr CONTROL, I.___

_

___________

AND INSTRUMENTATION 14.00

_['____

___________

________ 7.

PROCEDURES - NORMAL,. ABNORMAL, 3.33

________

EMERGENCY AND RADIOLOGICAL CONTROL

_

I ___ _ _I_

________ 8.

ADMINISTRATIVE PROCEDURES,

___________

CONDITIONS, AND LIMITATIONS 60.00 100.00 TOTALS

------

FINAL GRADE _________________%

All work done on this examination is my own. I have neither giv n not received aid.

5PPLIC5UTI5~555U5TURE~~~~~~~~~~~~~~

,

.

5.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE

---

_--_--_-------

GUESTION 5.01 (.50)

With the reactor critical at 10 E-4 %,

CEA group 5 is used to increase power to 10 E-3 %.

Select the statement that correctly describes the position of CEA group 5 after the power is stabilized at 10 E-3?..

'

A.

The group position will be higher than at 10 E-4% because mere fuel must be exposed to the available neutrons to maintain the higher power level.

B.

The group position will be higher than at 10 E-4% to overcome the power defect.

C.

The group position will be the same.

The outward rod motion needed to achieve a given startup rate equals the inward motion needed to reduce the startup rate to zero.

D.

The group position will be lower than at 10 E-4% due to the increased delayed neutron population associated with the higher power level.

QUESTION 5.02 (1.50)

Identify the secondary equipment associated with each labeled line on the attached Mollier Diagram.

(Processes are idealized.)

QUESTION 5.03 (1.50)

If following a LOCAr Tc is 530 Fr Th is 540 Fr RCS pressure is 1600 psia, Steam Generator (SG) pressures are both 995 psia and SG actual levels are both 0 inchese could the SG's be used as a heat sink?

Explain.

QUESTION 5.04 (1.50)

How does an increase in RCS temperature affect the relationship between indicated and actual core power as measured by the excore nuclear instruments?

Explain your answer.

.

'

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

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THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS, AND PAGE

- - -

GUESTION 5.05 (1.50)

During a reactor startup five hours after a trip from full power, power is leveled off at 10 E-4% to take critical data.

Enplain what rod motion, if any, is necessary to maintain this power level over the next hour.

QUESTION 5.06 (2.50)

a.

On the attached moderator to fuel ratio graph, indicate by the appropriate numbert where the reactor would be operating for each of the following cases.

Assume X indicates a core operating at 100% power middle of core lifer and consider each case individually.

1.

Moderator temperature increases by 10 degrees.

(0.4)

2.

A new core is loaded.

(0.4)

3.

Rods are inserted as boron concentration is reduced.

(0.4)

b.

What effect does a slightly positive MTC.have on a continuous rod withdrawal accident from low in the power range?

(1.3)

QUESTION 5.07 (3.00)

What effect would each of the following failures have on a natural circulation cooldown which is underway at 490 F.

Explain your answers and consider each failure independently.

a.

The steam dump valve which is being used to control cooldown rate fails open.

b.

Level is lost in the pressurizer.

c.

The Auxiliary feedwater valve to one of the SG fails shut.

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

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THEORY OF NUCLEAR POWEF PLANT OPERATION, FLUIDS, AND PAGE

---------------------------------------

___---------_-

GUESTION 5.08 (3.00)

While at 100% power the Axial Ghape Index (ASI) alara sounds acd the INCA printout shows:

ASI

- 0.1 Planar Radial Peaking Factor (Fxy)

Integrated Radial Peacking Factor (Fr)

1.4 Animuthal Power Tilt (Tq)

- 0.04 Value of N being used is 1.00 a.

What are two conditions which might cause these indications?

b.

What actions, if a..p r would improve this situation?

c.

What two Reactor Protection System trips would be affected by this situation?

(*****

END OF CATEGORY 05

- - - x)

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---

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6.

PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE

OUESTION 6.01 (2.00)

I s

If one level indicator on each steam senerator, 1-LT-ll14A and 1-LT-1124S, which feed the logic matrix for the Auxiliary Feedwater Acuation System (AFAS) failed as is, would the AFAS be able to provide its protective function?

Explain.

QUESTION 6.02 (2.00)

What actions should automatically occur in the Pressuriner Level and Pressure Control Systems if turbine generator ' cad dropped from

_

100% to 70%?

(2.0)

GUESTION 6.03 (2.00)

If a reactor trip signal was presente what effect would the simultaneous failure (to deenergine) of the Reactor Protection System (RPS) K-1 relay and K-2 relay have on the RPS?

What should be done to correct the immediate problem?

RPS diagram is attached.

QUESTION 6.04 (3.00)

Will the plant trip as a result of the following simu5taneous instrument failures?

Explain your answers.

a.

SUR channels A and B fail high during a startvo r when reactor is critical at 10 -6%.

b.

SG-11 level channel A fails LOW and SG-12 level channel B fails HIGH while at 80% power.

~

c.

Loop 1 Te channel A fails high and loop 2 Th channel d fails high while at 80% power.

d.

The lower UIC detectors for safety channels C and D fail low at 50% power.

QUESTION 6.05 (3.00)

Coepare the differences in detectors and signal processing between a linear power safety channel and a wide range los

'

channel when power is at 50%.

(3.0)

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

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PLANT SYSTEMS O2 SIGN, CONTROL, AND INSTRUMENTATION PAGE

-----------

GUESTION 6.06 (3.00)

If Unit 1 is operating at 100% power ard all root valves to the Steam Generator pressure safety channels are shut:

N ASFAS a. What two ESFAS" actuation signals would not function properly if a' major steam leak developed in the containment? Include what equipment would not receive expected signals.

b.

What automatic Reactor Protection System trips are available to mitagate consequences of a major steam leak in the containment?

c.

Would the plant return to power if the most reactive rod stuck out and all other systems functioned normally?

Justify your answer.

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

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-GUESTION' 7.01

'(1.50)

a..'Should shutdown. cooling be-secured before or after.formins a bubble in the:pressuriner?

b.

How'will~ pressure beLeontrciled during formation of a bubble?

0UESTION'-7.021 L(2.50)

A - 1200 e ppm dilution of the Reactor Coolant System (RCS)

-

has~been~ calculated to reach the critical boron concentration

.priorito a rod withdrawal startuP*

What actions should be

taken if af ter reducing RCS boron concentration by 600 ppm the source-range counts-changed from 10 cps to 20 eps?

' Explain why these actions should be taken.

'GUESTION 7.03

,(2.00)

~

Under what' conditions.~may an ESFAS initiated safe.ty feature tsystem be overridden?

Provide-two examples.

QUESTION

'7.04~

(2.00)

What initial actions should 4 9 taken if the immediate post trip actions have been completed and the-followins indications are

'present?

NO NOT include verification steps if they do not result infany acion.. EOP 800~is provided.

Loop 1 Loop 2 RCS Th 595 F 560 F RCS Tc-590 F 565 F SG Press 920 psia 900 psia-SG Level-100 inches

+ 20 inches Pressurizer-level 250 inches

,

Pressuriner temp.

615 F Pressurizer press.-

1800 psia-Containment' temp.

120 F Containment RMS~

2.5 rad /hr

Containment. pr ass'..

' ~0.5 psis (xxxxx CATEGORY 07 CONTINUED ON NEXT PAGE xxx*x)

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PROCEDURES - NORMA'., ABNORMAL, EMERGENCY AND PAGE

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QUESTION 7.05 (3.00)

Provide the two independent indications which would be used to conplete each of the following immediate actions contained in E0P 000, a.

Verify all CEA's are fully inserted.

b.

Verify pressuriner pressure stabilizes between 1850 and 2275 psia.

c.

Verify the reactor coolant system is subcooled greate; than 30 F.

d.

Verify 4KV buses 11 or 14 energized.

QUESTION 7.06 (3.00)

An entry into the containment is required while at 100% power and will result in an estimated whole body dose of 120 mrem.

Tha following four candidates are equally qualifieo to perfore.

tha task.

Which candidate may be allowed to perform the task in accordance with administrative procedures.

Explain your reasons for acceptins or rejecting each candidate.

No waivers can be obtained.

CANDIDATE

2

4 SEX male male female male AGE

38

20 WK/ EXPOSURE 100 mrem 30nrem Omrem 200 mrem QT/ EXPOSURE 1900 mrem 800 mrem 20 mrem

?

ACCUM LIFE EXPOSURE 5400 mrem 4000nrem 2200 mrem

?

REMARKS none None 3 months History pregnant unavai.lable (*****

END OF CATEGORY 07

- - - - )

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8.

ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIh1 TAT 10NS PAGE

GUESTION 8.01 (1.50)

The Calvert Cliffs Technical Specifications indicate. hat the maximum linear heat rate shall not exceed ;5.5 KW/ft. What TWO indications / conditions do the Tech. Specs. use to determine when this limit is exceeded?

DUESTION 8.02 (1.50)

a.

Why are the RCS Chenistry Transient limits different than the Steady State limits?

(0.9)

b.

Why must RCS pressure Le reduced below 500 psig if RCS chlor i he concentration exceeds the Steady State limit for greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in mode 5?

(0.6)

GUESTION 8.03 (2.00)

3.

Why was chapter 3/4.11r Radioactive Effluents, added to the Technical Specifications?

b.

What is the underlaying basis for all the Limiting Conditions a

for Operation contained in the Technical Specification Chapter 3/4.11, Radioacti'ce Effluents?

.

QUESTION 8.04 (1.00)

What determines the rate at which a plant shutdown should be made if an action statement for a limiting condition in the Technical Sp;cifications is entered?

OUESTION 8.05 (1.50)

a.

When may written procedures be departed from?

b.

What shall be done for any deviation from a procedure'

(***** CATEGORY 08 CONTIdUED ON NEXT PAGE '* * * :< * )

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8.

ADnINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE

QUESTION 8.06 (2.50)

a.

When should the Technical Support Center be activated?

(0.7)

6.

What is the function of the Technical Support Center?

(0.9)

c.

What is the function of the Emergency Operationt Facility?

(0.9)

DUESTION 8.07 (3.00)

The plant is-operating in Mode 5 with RCS temperature being maintained at 160 F.

Shutdown cooling is bein3 supplied by 11 LPSI Pump with 12 LPSI Pump tagged out.

A breaker control problem results in the loss of 11 LPSI Pump.

a. What actions are necessary by Technical Specifications?

b. When the shutdown ecoling loop is restored, what administrative requirements are necessary by Calver t Cliff Ins truction - 301?

00ESTION 8.08 (3.00)

~

For each of the following events explain why the NRC SHOULD or SHOULD NOT be notified within i hr.

a.

During instrument testing while in moce 3 three pressurizer pressure safety channels are momentarily bypassed.

b.

While at power, Tave momentarily dips to 510 F and then returns to normal.

c.

Refueling water tank level fells below 400,000 gallons and cannot be retored while in mode 5.

d.

During Surveillance testing an expected actuation of LPSI train A occurs.

.

(*****

END OF CATEGORY 03

- - - - )

(*************

END OF EXAMINATION

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THEORY OF~ NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE

-

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ANSWERS -- CALVERT CLIFFS-85/11/06-DUDLEY ANSWER 5.01 (.50)

C. Group position will be the same.

REFERENCE Nuclear Physics, p 170 ANSWER 5.02 (1.50)

A.

Steam Generator B. Control valves C. HP Turbine D.

Moisture seperator Reheaters E. LP Turbine F. Condenser E0.25 each]

ANSWER 5.03 (1.50)

No. CO.6]

There is insufficient subcooling margin between the SG and the primary to allow heat transfer across SG tubes. [0.9J C 4'

REFERENCE y cs rp 5 ; c,yg ;5 y,te s c<1 F.ec. < sit rc C.PCP T < <, r ac*a av Gaz Steam Tables ANSWER 5.04 (1.50)

As RCS temperature increases indicated power reads higher than cetual. CO.753 This is due to increased fast neutron leakage due to a decrease in coolant densitv. [0.75]

REFERENCE CE Nuclear Physics, p 166 SD No. 57:

NI, p3

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5.

THEORY OF NUCLEAR POWER FLANT OPERATION, FLUIDS, AND PAGE

--

ANSWERS -- CALVERT CLIFFS-85/11/06-DUDLEY ANSWER'

5.05 (1.50)

Rods must be withdrawn [0.63 to compensate for the buildup of Xenon resulting.from the reactror trip. CO.93 (Xe will continue to build three more hours.)

REFERENCE CE Nuclear Physics, p 206 ANSWER 5.06 (2.50)

a.

CO.4 each]

b. FTC must overcome the posit ve reactivity added by rods and MTC as the moderator temperature increases. [0.7]

FTC would turn ni her power level. [0. 63 (uncr: ' Jko itH r' 00 rin,1 l'c +- cR )

pouer but at a S

REFERENCE CE Reactor Theory, p 123-126 ANSWER 5.07 (3.00)

n. Increase cooldown rate CO.4] since more energy is >eing removed from the primary. [0.62 b.

May interupt natural cireviation CO.43 since hot legs maybe voided. CO.6]

c. Decrease cooldown rate [0.43 since SG tubes will become uncovered reducing heat removal. CO.62

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ANSWERS -- CALVERT CLIFFS-85/11/06-DUDLEY ANSWER 5.08 (3.00)

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b.

Reduce power by driving rods which will drive ASI positive.[1.0]

c.

Axial Power Distribution [0.5]

Thermal Margin / Low Pressure set point [0.53-REFERENCE SD No. 59: RPS, Fig A-6, A-8 TS 3/4 3.1

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6.

PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE

______________________________________________________

ANSWERS CALVERT CLIFFS-85/11/06-DUDLEY

--

ANSWER 6.01 (2.00)

Yes protection'would be provided. CO.8]

Three level detectors per SG are operable and would provide 2/3 cedundancy to Produce a trip signal. C1.2]

REFERENCE SD No. 34:

Auxiliary Feed System, p 45-47, A3-64 ANSWER 6.02 (2.00)

. Letdown valve ramps open. G L I Backup heaters turn off (+9 inches).LC.13 Backup heaters turn on (+12 inches). lex]

Spray valves open.(g4}

REFERENCE bD No. 5: RCS, Fig. A-20, p 65 ANSWER 6.03 (2.00)

No reactor trip. CO.7]

Tri P signal would not be sent trip TCB's and CEDM would remain enersi:ed. CO.63 Manually trip plant. [0.7]

(2.0)

REFERENCE SD No. 59: RPS, Fig A-2, p 39 ANSHER 6.04 (3.00)

a.

No CO.35] not until power reaches 10 -4%.

CO.4]

b.

No E0.35] channels auctioneer low signal therefore only channel A will trip. CO.4]

c.

Yes CO.35] both TH/LP channels will trip CO.42 (One c'hannel trips due to Odnb driving set piont high and the other channel trips due to real driving Pvar h i g h. ) ea //o ; G t.

na a,js /n a f.a r//ft,o rea r vr d.

Yes CO.353 both APD channels will trip E0.42 tu:.. c. m,,y c, f.,,, u a y c REFERENCE

'i/V.7 r,d u ff ruc- /W s 5 u //E SD No.

RPS, p 29, 31; Fig A-6, A-8

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s 6.

PLANT SY9TEMS DESIGNr CONTROL, AND INSTRUMENTATION PAGE

-----------------------------------------

ANSWERS -- CALVERT CLIFFS-85/11/06-DUDLEY ANSWER 6.05 (3.00)

Wide range los channels use a single group of fission chambers [0.4]

which supply a signal to an RMS (campbelling) circuit CO.4J and an LCR circuit E0.43.

LCR circuit provides constant output which is con.bined with the RMS output to produce power level. [0.3J (1.5)

Safety channels use two groups of UIC, U and L CO.53, which provide signals to two meters for Upper and Lower core

. power meters CO.52 and provides outputs of summes and differences of channel, power. CO.5]

(1.5)

REFERENCE SD No. 57: NI, p 4r 6 ANSWER 6.06 (3.00)

a.

AFAS BLOCK: AFW supply valves to faulted SG fr23 SGIS inhSIV lsl3 SGFP'sCCG Heater drain pumpsfCil Condensate booster pumps fell b. High power TM/LP Containment high pressure c.

No adequete shutdown margin would be maintained. Design of plant and safety systems.

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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

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ANSWERS -- CALVERT CLIFFS-85/11/06-DUDLEY ANSWER 7.01 (1.50)

a.

After formation of a bubble shutdown cooling should be secured. E0.6]

control-[w m oe i. e iwt o-~

b.

C VGFl ercourr--pemur

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REFERENCE OP -

1, p5 ANSWER 7.02 (2.50)

If dilution is completed the reactor would be critical. [0.8]

Stop dilution. [0.73 Recheck criticality calculations E0.53 and baron concentration E0.53 REFERENCE OP-2, p3 OI-28, p 6 ANSWER 7.03 (2.00)

May be overridden to support a threatened safety function. [0.83 EX: Override SIAS to prevent pressuri=ing plant.[0.63 Override MSIV to reestablish a heat sink. [0.63 (Other reasonaole examP les accepted.)

REFERENCE EOF - 400, p4 ANSWER 7.04 (2.00)

Energine pressuriner heaters. [0.6]

Manually position turbine bypass valves and atmospheric dump valves to maintain Tc less than 540 F.

[1.0]

Direct Chemisrty to place H2 analyzer in service and establish air mixing. [0.23 Start available iodine filters. [0.23 REFERENCE E0P 800, p 20, 30, 51, 52

,

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-7.

PROCEDURES - NORMAL, EBNORMAL, _EMEpBENCY AND PAGE

--- EE5i5E55iEEE E5siE5E--- :---------+---------

______at____________

ANSWERS -- CALVERT CLIFFS-85/11/06-DVDLEY ANSWER 7.05 (3.00)

,

a.

Rod bottom lights (bottom reed, switch)

Metroscope (reed switch)

b.

2 of 4 safety channels c. Subcoolins monitor

>

PZR pressure and Tc, Th, or Tave d.

Breaker indication Current meters Voltage meters 6 Ucc *;'"7'

E8 0 0.367 each3

)

t /#

REFERENCE

'

E0P 000, p 5-7 a

q, ANSWER 7.06 (3.00)

1. No CO.353 because he would exceed admin. limit'of 2000 mrem /qt CO.43 2. Yes CO.353 exposure would be less than 2000 mrem /qt CO.43 3. No CO.353 because she would exceed limit of 125 mrem /qt CO.43 4. No CO.353 because he would exceed 300 mrem /wk CO.43 (3.0)

RE.:ERENCE (k),

CCI-800Ar Att.

p 7

<

t

=

.

..

8.

ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE

__________________________________________________________

ANSWERS -- CALVERT CLIFFS-85/11/06-DUDLEY ANSWER, 8.01 (1.50)

a-Four.or more coincident incore channel alarms

[0.75]

b.

ASI outside of the Power dependent control limits [0.75]

REFERENCE Calvert Cliffs Tech Specs. pg. 3/4 2-1-ANSWER 8.02 (1.50)

s, Since (stress) corrosion is time and temperature dependent, time (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) is allowed to restore chemistry parameters prior to taking action.

(0.9)

6.

Reduce the effects of (stress) corrosion on RCS.

(0.6)

REFERENCE Tech. Specs 3/4 4-16 and B3/4 4-4 ANSWER 8.03 (2.00)

a. To incorporate items contained in the EnvyQgnmental Technical-Specifications, so that they could be deleted. [1.03 b. To prevent exposurer of a member of the public to radioactive isotopes, above limits set forth in 10CFR50 and 20. [1.03 REFERENCE TS, p B3/4 11-1

.

ANSWER 8.04 (1.00)

If the problem lends itself to correction a slower rate maybe used. [1.03 REFERENCE Administrative Procedures E.O.

80-12.1, 82-04

.

-

.

,

.

8.

ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE

ANSWERS -- CetVERT CLIFFS-85/11/06-DUDLEY ANSWER 8.05 (1.50)

a. In cases of emergency [0.353 where necessary to prevent injury to personnel or the public [0.23 or damage to equipment. E0.23 b. Los in SS 103 CO.353 Evaluate the need for a CCOM change report. [0.43 REFERENCE Administrative Procedure L.O.

CCI-300 ANSWER 8.06 (2.50)

.a.

On an Alert or higher. CO.73 b.

Provide plant related assessment and corrective actions. [0.93 c. Provide a communications link with Federals Stater and County energency organizations. CO.93 REFERENCE ERPIP Study Guide 3r 5,

ANSWER B.07 (3.00)

a. Suspend all operations which may increase decay heat load or a reduction in boron. [0.53 Deenergize Charging Pumps. [0.53 Close all containment penetrations within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . [0.53 b.

Record initial RCS temperature [0.53, pressure E0.53 and flou.

[0. 5 3 :, t'lth C PAGNA Er/ Trit s' M Tr.'

TRat/Sif u l L C&,

REFERENCE CCI-30lr p 5 Session 3 Handout

>

-

.

8.

ADMINISTRATIVE PROCEOURESr CONDITIONS, AND LIMITATI0rl5 PAGE

ANSWERS -- CALVERT CLIFFS-85/11/06-DUDLEY ANSWER 8.08 (3.00)

a.

Should report [0.353 since it prevented RPS fron fulfilling its safety function EO.43.

'

h.

No report E0.35] needed when an action statenent for an LCO is entered CO.43, Sh @ [Nreport [0.35]

st w rd o u w.,e n te in:bil1+" to -- e t L ctt a

c.

ecticn 2tstcment cy:irement: co,n2 rgens r?ccajic,9 y, s Fcd tiinF i d. No report CO.35] for ESF actuation during Surveillance testing

[0.43 (3.0)

REFERENCE 10CFR50.72(b)

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.

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,..

TEST CROSS-REFERENCE PAGE-

j-00ESTION'

VALUE-REFERENCE l

________

______

__________

R05.01

.50 DUD 0001088 05.02

~1.50 00D0001089 05.03-1.50.

DUD 0001090 05.04 1.50 DVD0001091

' 0 5.~ 0 5 1.50

~ DUD 0001092 05.06-2.50'

DUD 0001093

~

R05.07-3.00

. DUD 0001094 05.08 3.00

' DUD 0001095_

______

15.00 0'6.01 2.00 DUD 0001098'

~

06.02 2.00 0UD0001100

~

06.03.

2.00 DUD 0001101 06.04 3.00-

. DUD 0001102

"06.05'

3.00-DUD 0001103 o06.06 3.00-DUD 0001104

-

______

15.00

'07.01'

1.50 DU00001106-07.02 2.50-DUD 0001107 07.03'

2.00 DUD 0001108 07.04 2.00-DUD 0001109-07.05 3.00 DUD 0001110 071.06 3.00'

DUD 0001111-

______

14.00 08.01 1.50 DUD 0001112

~08.02 1.50 DUD 0001114

-

08.03-2.00 DVD0001119 08.04 1.00'

DUD 0001161

'08.05 1.50.

DUD 0001162-08.06-

.2.50 DUD 0001163.

08.07

, 3.00 DUD 0001164 08.08 3.00 DUD 0001116

______

- 16.00

______

. ______

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v t e Inspection Report - Calvert Cliffs - 1985032
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \| 25 \|
1985 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 9(0) 10(0) 11(0) 12(0) 13(0) 14(0) 15(0) 16(0) 17(0) 18(0) 19(0) 22(0) 23(0) 24(0) 25(0) 26(0) 27(0) 28(0) 29(0) 30(0) 31(0) 32(0) 33(0) 34(0) 35(0) 72(0) 31(0) Security Operator Exams Emergency Planning Other

IR 05000317/1985032

Text

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