ML20155F541

Exam Rept 50-315/OL-86-01 on 860311.Exam Results:Two Senior Reactor Operators Passed Written Exam

ML20155F541
Person / Time
Site:	Cook
Issue date:	04/15/1986
From:	Burdick T, Defferding L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II), NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20155F523	List: ML20155F519 ML20155F541
References
50-315-OL-86-01, 50-315-OL-86-1, NUDOCS 8604220254
Download: ML20155F541 (76)

Text

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U.S. NUCLFAR REGULATORY COMISSION l

REGION III I

Report No. 50-315/0L-86-01 Docket Nos. 50-315; 50-316 License No. DPR-58 Licensee:

Indiana and Michigan Electric Company P.O. Box 458 Bridgeman, MI 49106 Facility Name:

Donald C. Cook Nuclear Plants 1 and 2 Examination Administered At:

Donald C. Cook Nuclear Plants 1 and 2 Examination Conducted: Ma h 1, 1986 fk.

l Q

Examiner:

Leo J. Defferding Approvcd By:

T. M. Burdick, Chief Operating Licensing Section Date l.

i f

Examination Summary Examination administered on March 11, 1986 (Report No. 50-315/0L-86-01)

A written examination was administered to two SR0 candidates.

They had previously passed the plant walkthrough examination.

Results:

Both candidates passed the written examination.

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l 8604220254 860416 PDr ADOCK 05000315 V

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7-REPORT DETAILS

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1.

Examiners L. J. Defferding 4

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2.

Examination Review Meeting l

No longer conducted.

The specific facility comment, followed by the NRC response, appears below:

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QUESTION 5.02 a:

a.

To what level would the tank drop to as the pump continues to remove water from the tank?

FACILITY COM4ENT:

The keyed answer isdicated full. credit for 3.4 feet (four feet water - 0 256 psia) and 80% credit for l

four feet.

The equation used for NPSH found on Page 24 of the i

Thermodynamics Study Guide is:

i NPSH = PTANK + Z - hl SAT

-P if h is neglected (small) then L

l NPSH = PfANK + Z - PSAT Since the pressure in the tank will be saturation pressure for 60*F (.256 psia) t NPSH = Z (height of water)

[

r i

Therefore, if it is assumed that the tank level stops i

decreasing when minimum NPSH is reached, the full i

credit value should be four feet.

The tank level would i

continue to decrease, however, because the pump will still move water af ter it begins to cavitate.

i Eventually, the tank could empty.

Since the time from i

minimum NPSH to pump failure (and how much water will i

be removed from the tank) is not known, the final level in the tank cannot be realistically determined.

l We request that full credit be granted for four feet or an indication that the tank would empty and partial i

1 credit be assigned for identifying the elements r

(equation) for NPSH.

i NRC RESPONSE:

Accepted answer changed to read "four feet" (+0.5).

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f 4

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i QUESTION 5.04 c.:

During a loss of offsite power cooldown, the Natural Circulation Procedure instructs the operator to lower i

RCS pressure from 2235 psig to 200 psig in a time period in excess of 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br />.

c.

According to the procedure, RCS and S/G pressures are initially held at 2235 and 635 psig, respec-l tively.

Assuming natural circulation has been l

established, what is the approximate amount of RCS subcooling?

ANSWER 5.04 c.:

c.

110-130*F (+1.0)

(will except 131-160*F for +0.5).

t FACILITY C0P#4ENT:

The answer to this question makes a specific assumption that is critical to its correct calculation.

The assumption is that a very large 'T (approximately 38*F) exists between the hot and cold legs.

To obtain a ^T of this magnitude, two conditions must exist:

Nearly a maximum power history and a very short time period (minutes) following the reactor trip.

If these two assumptions are not made, the magnitude of the subcooling will be much greater than that allowed for full credit by the answer key.

Since the question does not state the amount of time elapsed, the amount of decay heat present, or suggest that the candidate assume a ^T for the calculation, we request that a calculation similar to below also be allowed for a full credit answer.

2235# (2250 psia) = 653*F 653*

635# (650 psia) = 495*F

- 495*

158"F l

(Assume T

=Tstm) avg This calculation assumes a low power history and three hours elapsed time following the trip to obtain a negligible 'T.

(See attached procedure OHP-4023.001.008, Step 5.3.2)

NRC RESPONSE:

Accepted.

Answer changed to read "110-160*F (+1.0) (exact amount depends on power history which was not given)."

S_t, ate w h each of the Rod Control System limits are QUESTION 5.12:

t provided.

a.

Withdrawal limit b.

Maneuvering band upper limit 3

c.

Maneuvering band lower limit d.

Insertion limit ANSWER 5.12:

a.

To help avoid dumping steam if a decrease in power is initiated.

(+0.75) (will accept "to provide suf ficient bite" for (+0.5)).

b.

To allow an "immediate" return to full power operation (without boron dilution).

(+0.75) c.

To maintain even axial flux depletion.

(+0.75) d.

To maintain adequate shutdown margin.

(0.75)

FACILITY COP 94ENT:

a.:

Since there are many factors that determine whether steam will be dumped on a power decrease, it is more appropriate to state the primary reason for the withdrawal limit as being control rod bite. We request that full credit be granted for "to provide sufficient bite."

b. and c.:

The term " Maneuvering Band" is not used at D. C. Cook.

We request that these parts be deleted from the exam for the following reasons:

1.

A maximum power escalation rate of six percent per hour precludes an immediate return to full power.

See Technical Data Book Figure 11.9.

2.

Rod position is dictated by maintaining Axial Flux Difference in the Target Band specified on Figure 13.1 of the Technical Data Book.

d.:

The Technical Specification Basis for the insertion limit (TS 3.1.3.5) identifies three reasons for each Unit.

Unit One specifies:

1.

Maintaining acceptable power distribution.

2.

Maintaining minimum shutdown margin.

3.

Minimize the effacts of an ejected rod.

We request that any of these reasons also be accepted without point loss.

NRC RESPONSE:

(a) Accepted.

Full credit (+0.75) given for "to provide sufficient bite."; (b) and (c) Comments noted and taken into consideration in grading; (d) Accepted.

Full credit given for (1) maintaining acceptable power distribution, (2) maintaining minimum shutdown margin, or (3) minimize the effects of an ejected rod.

4

QUESTION 5.13 b.:

Burnable poison rods are used in the first fuel cycle.

Why aren't they needed in subsequent fuel cycles?

b.

_h ANSWER 5.13 b.:

b.

The partially depleted fuel and the building of fission product poisons in the subsequent fuel cycles limit excess reactivity to a value controllable by soluble boron alone.

(+0.75)

FACILITY C0PetENT:

At D. C. Cook, burnable poison rods are used in subsequent fuel cycles for the same reason as in the first fuel cycle.

See Reactor Core Lesson Plan, LP-RTOP-NSI.

Since the candidate may attempt to answer the question as stated, we request the keyed answer or an explanation as to why we still use burnable poisons te accepted for full credit.

NRC RESPONSE:

Accepted.

Full credit (+0.75) given if candidate explains why burnable poison rods are used'in subsequent fuel cycles.

QUESTION 6.01:

a.

What automatic actions take place in the Component Cooling Water (CCW) system upon a Safety Injection signal? Assume standby pump is locked out.

b.

What automatic actions take place in the Component Cooling Water (CCW) system upon a Containment Spray signal?

ANSWER 6.01:

a.

Hx outlet open (+0.5).

RHR Hx outlet goes to INTERMEDIATE position giving approximately 3000 gpm flow (+0.5).

Isolation occurs on BA evaporator, spent fuel pit cooling, letdown Hx excess letdown Hx and Rx support coolers (+0.1 each).

b.

CCW to and from the RCPs, thermal barrier heat exchangers upper and lower bearing oil coolers, is isolated.

CCW to and from the CEQ fan motors is unisolated.

(+1.5)

FACILITY COMMENT:

a.

We request that isolation of CCW to either the BA evaporator or the South Rad. Waste evaporator (SRWE) should be accepted as correct.

The evaporators are unit specific.

The BA evaporator is supplied with CCW from Unit l's system; the SRWE is supplied with CCW from Unit 2's system.

Listing either evaporator indicates knowledge of this automatic action.

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b.

The isolation of CCW from the RCP thermal barriers and motor bearing oil coolers is summarized by the statement "CCW to RCP's isolates." Reference l

lesson plan R0-C-A501 0.1.b. (Page 11 of 13).

i-l l.

The question does not specifically elicit a L

response which lists discrece reactor coolant pump l

support systems supplied with CCW cooling.

As l

both RCP support systems supplied with CCW cooling I'

isolate on a Phase 8 isolation / containment spray signal, the above referenced summarized response l

should be accepted as a complete answer.

NRC RESPONSE:

(a) Accepted.

Credit (+0.1 each) given for isolation of CCW to either the BA evaporator or the South Rad. Waste evaporator (SRWE); (b) Accepted.

Full credit given for "CCW to RCP's isclates."

QUESTION 6.02:

List the associated components / pumps from the Emergency Core Cooling System needed for the Injection Phase and the RCS pressure at which they will begin to inject.

l ANSWER 6.02:

Pump Shutoff Head i

Centrifugal Charging Pumps 2540 psig (+0.5)

Safety Injection Pumps 1560 psig (+0.5) j i

Accumulators 620 psig (+0.5)

Residual Heat Removal 200 psig (+0.5)

FACILITY C0999ENT:

The shutoff head pressures listed are design numbers; pump operating characteristics and actual values will depend on specific operating conditions. We request I

that the answers within the following ranges of pressures be accepted for full credit.

Centrifugal Charging Pumps 2500-2550 psig Safety Injection Pumps 1550-1600 psig l

Residual Heat Removal 200-220 psig l

Accumulator nitrogen cover pressure is dictated by LCO 3.5.1.

We request that answers within the following pressure ranges be accepted for full credit.

Unit 1 Accumulators 585-658 psig i

t' Unit 2 Accumulators 599-622 psig i

I r

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NRC RESPONSE: Accepted.

Answer changed to read:

l l

Pump Shutoff Head Centrifugal Charging Pumps 2500-2550 psig (+0.5)

Safety Injection Pumps 1550-1600 psig (+0.5)

Accumulators or 580-660 psig (+0.5) l Units 1 Accumulators 585-658 psig (+0.25)

Units 2 Accumulators 599-622 psig (+0.25)

Residual Heat Removal 200-220 psig (+0.5)

(+2.0 maximum) l l

QUESTION 6.05:

List the components of the ECCS which receive an actuating signal on a safety injection and what action that component must take.

l ANSWER 6.05:

Centrifugal Charging Pumps - Start CCP Suction from RWST (IM0-910/911) - Open l

BIT Inlet and Outlet Valves - Open i

l SI Pumps - Start j

RHD Pumps - Start CCP Mini Flow Valves - Open Accumulator Isolation Valves - Open BIT Recirculation Valves - Close

(+0.15 for each component, +0.1 for each action)

FACILITY COMMENT:

The accumulator isolation valves are maintained open with their breakers racked out when pressurizer pressure is greater than 2000 psig (TS 3.5.1).

With no valve control power available, the safety injection valve opening signal performs no action. We request that the Accumulator Isolation Valve action not be required for full credit.

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NRC RESPONSE: Accepted.

" Accumulator Isolation Valves-Open" deleted l

from answer.

Point value changed (+0.2 for each component, +0.1 for each l

action) +2,0 maximum l

QUESTION 6.06:

List the four (4) ice condenser monitoring systems that l

are available to control room operators.

ANSWER 6.06:

1.

Ice bed temperature (+0.5) 2.

Ice condenser door position (+0.5) 3.

Ice condenser floor cooling temperature (+0.5) 4.

Equipment and personnel doors (+0.5)

FACILITY COMMENT:

This question is based on a poorly worded objective (No. 8),0f the lesson plan R0-C-NS14.

The only "systow' located in the control room is the Ice Bed Tempetiture Recorder.

The only indication in the contro!' room for the other " systems" are annunciators which indicate that an abnormal condition exists (see Paragraph VI.E. on Page 19 of R0-C-NS14).

The lesson plan clearly states in Section VI.B.2., VI.C.2, and VI.D.2. that these " systems" are operated from the Containment Auxiliary Subpanel (CAS).

The lesson does not indicate that this panel is not in the control room because its location in the Auxiliary Building is common knowledge for Operations personnel.

When the question was clarified to the candidates, they were told to consider " indications even if they were on other panels." This would lead the candidates to identify indications (status lights, valve positions, alarms, etc.) located in the control room, on the CAS Panel or Refrigeration Panel (also in the Auxiliary Building).

I Because of the confusion associated with this question, we request that it either be deleted or credit be awarded for indications associated with the " monitoring systems." These indications would include:

I Ice Bed Temperaturo Recorderc (SG-7, SG-17)

Control Room Annunciators (See lesson plan Paragraph VI.E.)

Indications located on the CAS Panel (See attached Drawing 1-5585-16) l 8

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Indications located on the Refrigeration Panel (See attached Drawings 12-5584 and 12-98297-0)

NRC RESPONSE: Accepted.

Other appropriate recorders, annunciators, and indicators accepted for credit.

QUESTION 6.07:

The reactor trip breaker shunt coils will:

(a.) energize on a manual trip signal.

(b.) energize on an automatic trip signal.

(c.) deenergize on a manual trip signal.

(d.) deenergize on an automatic trip signal.

ANSWER 6.07:

(a.) (+1.0)

FACILITY COMMENT:

RFC 12-2663 (completion notice attached) modified the Reactor Trip Breakers to activate both the shunt and undervoltage devices to trip the breaker on an automatic trip signal.

Unit I was completed October 30, 1985, and Unit 2 completed November 7, 1985. We request that the answer key be amended to allow full credit for "a and b."

NRC RESPONSE: Accepted with slight modification.

Credit given for (a.)

and (b.).

QUESTION 6.09:

This question asks the candidate to list the setpoints for VCT level control.

FACILITY COMMENT:

The keyed answer for Part b " Stops auto makeup" is incorrect.

The correct answer is 24% (see CVCS lesson plan R0-C-NS06, Page 5).

NRC RESPONSE:

Accepted.

Answer changed from 27 to 24.

QUESTION 6.13 c.:

Fill-in the blanks to make the following statements about reactor coolant pump operation correct.

c.

A pressure differential of at least psid should be maintained across the No. 1 seal.

ANSWER 6.13 c.:

c.

275 (+0.25)

FACILITY COMMENT:

Unit 1 RCP seal parameters changed prior to Cycle 9 restart.

Reference 1-0HP-4021.002.003, Precaution 4.5.

(See attached Page 3 of 1-0HP-4021.002.003 with Procedure Change Sheet 2 incorporated.)

Unit 2 minimum RCP No. I seal differential pressure remains at 275 psid.

Reference 2-OHP-4021.002.003, Precaution 4.5.

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We request that either.200 psid or 275 psid be accepted for full credit.

NRC RESPONSE: Accepted.

Credit given for 200 or 275.

QUESTION 6.14:

What activities does the failed fuel detector. system monitor when in service? (Note any differences between units.)

ANSWER 6.14:

Gross Gamma, I-135, Cs-137 Unit 1 (+0.5)

Gross Gamma, I-131 Unit 2 (+0.5)

FACILITY CCMt1ENT:

The keyed answer is not solicited by the question. The question requests " activities" while the key required specific isotopes. We request that full credit be awarded for Gross Gamma or fission product activity (see Page 2 of Primary Sample System lesson plan, R0-C-NS17).

NRC RESPONSE: Accepted.

Full credit given for Gross Gamma (+0.5) or fission prcduct activity (+0.5).

QUESTION 7.02:

Assume during reactor coolant or shutdown the RHR capability is lost.

List three (3) alternate cooling methods.

(State any assumptions.)

ANSWER 7.02:

Head on 1.

Condenser steam dump.

(+0.5) 2.

Atmospheric steam dump.

(+0.5)

Head off 3.

Spent fuel pit cooling.

(+0.5)

FACILITY COMMENT:

There are additional answers on the page referenced in the answer key that are not included in the exam answer key.

(OHP-4023.017.001, Page 3). We request that any of these be considered as correct answers in addition to those listed on the key.

Step 5.7.4:

" Additional cooling can be achieved by establishing maximum charging and letdown of the reactor coolant system."

(Utilizes NRHX as heat sink.)

Step 5.8.1:

"If not already done, flood the refueling cavity to the normal refueling level." (Provides a large volume of RWST water to the cavity as an immediate heat sink.)

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Step 5.8.5:

" Establish maximum refueling cavity filtering and purification." (This system does not provide cooling, but is included in the procedure because it does provide forced circulation of the refueling cavity water.)

Step 5.8.4:

" Operate sufficient containment ventilation units to maintain contain-i ment air temperature at or below about i

80 F."

(Increases driving force for j

l heat dissipation to ambient.)

l We would also request that the apparent assumptions (Head on and Head off) not be required for full credit.

The question did not specifically state Head on/ Head off configuration.

Also, the configuration would be implicit to the cooling method listed by the examinee.

For example, an operator would probably not think it necessary to state that the Rx vessel head would be on while dumping steam from the S/G's.

NRC RESPONSE:

Accepted.

" Head on" and " Head off" deleted from answer.

Credit given for following additional answers:

4.

Establish maximum letdown and charging (+0.5) 5.

Flood the refueling cavity (+0.5) 6.

Establish maximum refueling cavity filtering and purification (+0.5) 7.

Operate containment ventilation units (+0.5)

(+1.5 maximum)

QUESTION 7.05:

Answer the following questions about the emergency diesel engines, a.

What condition will cause the incomplete start sequence to occur?

b.

How can the operator in the control room tell this has occurred?

c.

List the two (2) conditions either of which will permit a restart.

ANSWER 7.05:

a.

During a EDG start < 95% rated speed > 10 second.

(+0.5) b.

A red failure light is energized in the control room.

(+0.5) 11

c.

Two minutes elapse oc push the EDG reset button.

(+0.5)

FACILITY COMMENT:

Though the lesson plan may be interpreted to imply that the "two minute closure of SV5" must be completed prior to restart of a diesel generator after an incomplete start sequence, this action is not required to be completed before restart.

If an automatic start signal is present, the diesel has the ability to start as soon as the reset pushbutton is depressed.

(A simplified control circuit diagram is included.) When the reset pushbutton is depressed, SV5 will receive an open signal.

The diesel will not restart unless the reset pushbutton is depressed.

We would request that the answer "two minutes elapse" be deleted from the answer key and that any additional

" reasonable" answers not be marked incorrect as the question did ask for two answers.

)

NRC RESPONSE:

Accepted.

Full credit given for SI and blackout start and "two minutes elapse or" was deleted from the answer key.

j QUESTION 7.10:

If system frequency drops to 58.8 Hz due to loss of generating capacity, what five (5) actions sre required after 30 minutes if the system frequency does not recover?

ANSWER 7.10:

1.

Open breakers K and K-1.

(+0.5) 2.

Verify steam dump controls have functioned to limit T-avg and Pressurizer pressure transients.

If not, trip the reactor.

(+0.5) 3.

Monitor reactor power to ensure the rods are moving inward to reduce power level.

If not, take manual control and reduce reactor power.

(Automatic Rod Withdrawal is blocked below 15%

C-5).

(+0.5) 4.

Adjust turbine speed to maintain frequency at 60 Hz.

(+0.5) 5.

Monitor T-avg to ensure the Steam Dump System is performing properly; take manual control if required.

(+0.5)

FACILITY COP 91ENT:

In order to correctly answer the question, all the right assumptions must be made.

If answers 1 and 2 are stated and a Rx trip is assumed, answers 3 and 4 would be quite impossible to carry out. We would request 12 2

that latitude in grading of this question be considered based on a possible lack of answers 3, 4 and/or 5 based on those given as 1 and 2.

NRC RESPONSE:

Accepted.

Comments noted and taken into consideration during grading.

QUESTION 8.01 c.:

Answer the following TRUE or FALSE.

c.

Areas outlined in magenta on the Radiation Information Board are EHR areas.

ANSWER 8.01 c.:

c.

FALSE (+0.5) (outlined and crosshatched in red)

FACILITY COMMENT:

This true or false question is technically correct.

We would, however, like to point out that all individuals entering the controlled area are required to check out the Radiation Information Board (RIB) prior to entry.

The RIB contains a color code /grahic legend to enable all individuals to ascertain conditions of a given area.

Additionally, EHR areas are specifically stated in PMP-6010. RAD.001 and are under RP key and lock control.

Individuals entering these areas require RP coverage.

It is not realistic for an individual to remember the color coding, which is not procedurally addressed except for EHR, where six (6) colors plus

" crosshatching" are used.

NRC RESPONSE:

Comments noted.

No change.

QUESTION 8.10:

If one (1) emergency diesel generator is taken out of service for maintenance and the reactor is in Mode 1, what action must be taken?

ANSWER 8.10:

Demonstrate operability of remaining AC source (+0.5)

(by performing Surveillance Requirements 4.8.1.1.1.a and 4.8.1.1.2.a.4) within one hour (+0.5) (and at least once every eight hours thereafter).

FACILITY COMMENT:

We request that the following be accepted as an additional correct answer.

" Restore to operable within one hour or conduct the Emergency Diesel Generator Operability Test (STP.027)".

STP.027 is performed to show compliance with TS 3.8.1.1 including breaker alignment (STP. 031).

STP.027 has been included for your information.

NRC RESPONSE:

Accepted.

Credit given for " restore to operable within one hour or conduct the Emergency Diesel Generator Operability Test" (STP.027).

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QUESTION 8.12

c. and d.:

For each of the following ECC Systems, specify whether the Technical Specification LCO is the SAME or DIFFERENT for Units 1 and 2 for Mode 1 operation.

c.

RWST boron concentration, Technical Specification 3.5.5.

d.

RWST minimum water temperature, Technical Specification 3.5.5.

l ANSWER 8.12

c. and d.:

c.

DIFFERENT (U : 1 1S50 ppm; U :

2000-2200 ppm) 2

(+0.5) d.

DIFFERENT (Ug: 70*F; U2 80*F) (+0.5)

FACILITY COMMENT:

c.

The answer key is correct except for the Unit i value for RWST boron concentration.

The value should be > 1950 ppm.

d.

The answer key is technically correct in accordance with technical specifications.

Attached is a copy of PMSO.074.

We feel that this question can be answered as either SAME or DIFFERENT.

Our plant procedure requires us to treat RWST minimum temperature as 80*F and should a violation occur, the Action Statement must be followed.

NRC RESPONSE:

c.

Accepted.

Answer changed to read DIFFERENT (U : > 1950 ppm; U : 2000-2200 ppm).

t 2

d.

Disagree.

Administrative limits are the same but technical specification limits are different.

SR0s should know technical specifications.

QUESTION 8.13 d.:

State whether each of the following, as specified in 10 CFR 72, requires NRC notification within one HOUR or four HOURS.

d.

Transfer of a contaminated individual to a local hospital.

ANSWER 8.13 d.:

d.

four hours (+0.5)

FACILITY COMMENT:

d.

The keyed answer is technically correct in accordance with 10 CFR 72 classifications.

However, PMP-2080.EPP.001 ECC-20 also classifies transfer of a contaminated individual to an offsite facility as on Unusual Event.

Unusual Events must be reported within one hour. We request that either one hour or four hours report be accepted as correct.

14

NRC RESPONSE:

Because the answer one hour or four hours is potentially correct, Question 8.13(d) was deleted from the exam.

QUESTION 8.14 b.:

A plant emergency occurs which requires activation of the D. C. Cook Emergency Plan.

b.

At what minimum emergency condition e.lassification is activation of each of the following required?

a.

Technical Support Center.

b.

Operations Staging Area.

c.

Emergency Operations Facility.

ANSWER 8.14.b.:

a.

Alert (+0.5) b.

Alert (+0.5) c.

Site Area Emergency (+0.5)

FACILITY COMMENT:

This keyed answer is technically correct by the current procedures.

At this time, all Emergency Plan procedures are undergoing a complete revision.

Annual Emergency Plan training, conducted during November and December of 1985, revealed several changes that were going to be implemented prior to June 10, 1986, NRC EVALUATED EXERCISE. One of these changes is the activation of the Emergency Operations Facility (EOF) (Part c) at the Alert level.

We request that Alert and Site Area Emergency be accepted for full credit based on the fact the activation of EOF is an administrative decision of plant management and does not violate current procedures.

NRC RESPONSE:

Accepted.

Full credit given for " Alert" or " Site Area Emergency."

(+0.5) 3.

Exit Meeting Not conducted because no plant walkthrough examinations were administered.

15

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4 Page 14 Cook 1 and 2 March 11, 1986 Points available OUESTION 5.10 With no fuel in the reactor, the count rate is 125 cps. After loading six (6) fuel assemblies, the count rate is 500 cps. A new detector, farther from the core, is now used and it has a count rate of 300 cps. Four (4) wre fuel assemblies are loaded, increasing the count rate to 750 cps on the new detector.

What is the value of k,ff after the ten (10) fuel assemblies have been loaded?

(1.0)

ANSWER 5.10 Moving the detector changes the base count rate:

Ci (new)

Co (new) = C

= (125 cps) [

S = 75 cps

[+0.5]

o Cj s

After loading four more fuel assemblies:

C (new) keff = 1 -

Cj "I~7

= 0.90

[+0.5]

s Alternately:

The value of keff with six fuel elements is:

keff = 1 -

=1-$

= 0.75 [+0.5]

After loading four more fuel assemblies:

C keff2 = 1 -

(1 - keffi)

=1-(1 - 0.75)

= 0.90

[+0.5]

Reference (s) 5.10 1.

DCC Reactor Theory, Neutron Sources and Subtritical Multiplication (I-4), pp. 13-24.

-Section 5.0 Continued on Next Page-

Page 15 Cook 1 and 2 March 11,1986 Points Available 00ESTION 5.11 Differential boron worth changes over core life (see Figure I-5.27below).

a.

Why does it initially decrease (become less hegative)?

(0.75) b.

Why does it subsequently increase (become more negative)? (0.75)

-10.0 I

i l

l E,

i.

-l

._.li. _

..i.

.. i--

..i L.. - : - r ---

- L--

i

?

.L..

. l...

. -i...

~

5 I

.l l

i l

C = 0 ppm

!'i l

B 2

' _. _ L_

5. -

.j 7

C

-9.0

. -.g + -

.i r-ii i

l i.

. _.l..

g-j..

t-.

f, E

i. '

= 1360 ppm,!-. -l..

I I

I i

-9.2 i

C o

../

i I

B m,

~},

2 9

j..j..

., _... j.

.;-..p-

.j.

-r f.

}

-8.8 ht

_ _._.[ _...!..

g*

-C changing -.

.. L..

B N

i

.!.!_ !.. p

_. L 1.

i 4

i a

i. \\.

r u_

I g

-8.4

._I._.._

- I- - I-l l

i

{ --

j l

I

._ l j

_. y i

l l

I

..l l

l.

j l

i

?

-8.0 i

I i

i O

2000 4000 6000 8000 10000 12000 14000 10000 CORE AVERAGE BUPfiUP (It',tD/MTU)

FIGURE I-5.27 Composite Graph of Changing Differential Boron Worth as a Functior, of Core Burnup

-Section 5.0 Continued on Next Page-

Page 16 Cook 1 and 2 March 11, 1986 Points Available ANSWER 5.11 a.

Fission product buildup (c'auses dbw to become less negative. This initiall outweights the decreasing boron concentrationeffect.)

+0.75]

b.

Decreasing boron concentration (causes dbw to become more negative. This subsequently outweights the fission product buildup.)

[+0.75]

Reference (s) 5.11 1.

DCC Reactor Theory, PWR Core Physics (I-5), pp. 31-35.

-Section 5.0 Continued on Next Page-

Page 17 Cook 1 and 2 March 11, 1986 Points Available OUESTION 5.12 State why each of the Rod Control System limits are provided, a.

Withdrawal limit (0.75)-

b.

Maneuvering band upper limit (0.75) c.

Maneuvering band lower limit (0.75) d.

Insertion limit (0.75)

ANSWER 5.12 a.

To help avoid dumping steam if a decrease in power is initiated.

[+0.75) (will accept "to provide sufficient bite" for

&6.753).

i b.

To allow an "immediate" return to full power operation (withoutborondilution).

[+0.75]

c.

To maintain even axial flux depletion.

[C.75]

d.

To maintain adequate shutdown margin. [+0.75] h M f N.

cQ4 :

74% W q% Aph, m W *"$

Mm M4

<W M Q Dw y &'lbv Reference (s) 5.12 1.

DCC Reactor Theory, Power Core Physics (I-5), pp. 50-51.

-Section 5.0 Continued on Next Page-

Page 18 Cook 1 and 2 March 11, 1986 Points Available OUESTION 5.13 Burnable poison rods are used in the first fuel cycle.

a.

Why are they necessary?

(0.75) b.

Why aren't they needed in subsequent fuel cycles?

(0.75)

ANSWER 5.13 a.

To maintain a negative moderator temperature coefficient (by providing partial control of the excess reactivity associated with the first fuel cycle).

[+0.75]

b.

The partially depleted fuel and the building of fission product poisons in the subsequent fuel cycles limit excess reactivity to a value controllable by soluble boron alone.

[+0.75] P ati.sotet! u g,,# a ey, g,a (

t,,,,,,.5

~6g t.m. t h po.w

,,,13 n,,, A ;s 9,, ;,,,,,,

4,,, g, j,,,

Reference (s) 5.13 1.

DCC Rea; tor Theory, PWR Core Physics (I-5), p. 57.

i

-Section 5.0 Continued on Ncxt Page-j

9 Page 19 Cook 1 and 2 March 11, 1986 Points Available 00ESTION 5.14 Select the correct statement regarding Xe behavior in the core?

(1.0)

(a.) Buildup and decay of Xe is similar to Sm.

(b.) The value of peak Xe, after a reactor trip, is independent of the pre-trip power level.

(c.) Xe peaking, after a reactor trip, occurs at a fixed time independent of the pre-trip power level.

(d.) Equilibrium Xe concentration, after a reactor trip, is independent of the pre-trip power level.

ANSWER 5.14 (d.)

[+1.0]

Reference (sl 5.14 1.

DCC Reactor Theory, PWR Core Physics (I-5.1), pp. 63-79.

-E,d of Section 5.0-1 i

Page 20 Cook 1 and 2 March 11, 1986 Points Available 6.0 PLANT SYSTEM DESIGN. CONTROL AND INSTRUMENTATION (25.0)

QUESTION 6.01 a.

What automatic actions take place in the Component Cooling Water (CCW) system upon a Safety Injection signal? Assume standby pump is locked out.

(1.5) b.

What automatic actions take place in the Component Cooling Water system (CCW) upon a Containment Spray signal?

(1.5)

ANSWER 6.01 a.

Hx outlet open [+0.5].

RHR hx outlet goes to INTERMEDIATE position giving approximately 3000 gpm flow [+0.5].

Isolation occurs on BA evaporator, spent fuel pit cooling, letdown Hx, excess letdown Hx and Rx support coolers

[+0.1 each] ALu M kN ** M itRVE).C C W to A *,04 #d@

n M k JAv & A.IguCu Jet 4 Ce v

• I A' CCW to and from the RCPs(il coolers,)is isolated.GCW to ~ ofg themal barrier heat exchangers _ r b.

upper and lower bearing o g

and from the CEQ fan motors is unisolated.

[+3PJ] (LLE MCAP cc W tc RCPb he4h E M 4 0,q $

-+ 0:15 Reference (s) 6.01 1.

DCC Lesson Plan, R0-C-A501, p. 6.

-Sect;on 6.0 Continued on Next Page-

Page 21 Cook 1 and 2 March 11, 1986 Points Available 00ESTION 6.02 List the associated components / pumps from the Emergency Core Cooling System needed for the Injection Phase and the RCS pressure at which they will begin to inject.

(2.0)

ANSWER 6.02 Pumn Shut-Off Head u.a.sne Centrifugal Charging Pumps 964trpsig

[+0.5]

irrc-suo Safety Injection Pumps M60*psig

[+0.5]

SWk o Accumulators 426 psig

[+0.5]

1c3=116%

o' Residual Heat Removal h psig

[+0.5]o.4 v

Wit I (tuwnaw 5 1 5-6 59 ysis

[s d a h

(%;L A Gwmak F5'I E'W.35 o

Reference (s) 6.02 44b '#4 g j,p, g,J 1.

DCC Lesson Plan, R0-C-NS12.

-Section 6.0 Continued on Next Page-

~ -

Page 22 Cook 1 and 2 March 11, 1986 Points Available OUESTION 6.03 In order to satisfy the interlock and open the recirculation s

sump suction valves, which of the following conditions are nol required?

,(1.0)

(a.) The RHR pump suctions from the RWST (IM0-310/320) must be closed.

(b.) The charging pump suctions from the RWST (IMO-910/911) must be closed,

e. - ts a ~.. ? u. ~r

~

(c.) The SI pump suctions from the RWST (IMO-215/225) must be closed.

(d.) Control power to the recirc sump suction valve must be on.

ANSWER 6.03 (b.)

[+1.0]

Reference (s) 6.03 1.

DCC Lesson Plan, R0-C-NS12.

4 1

-Section 6.0 Continued on Next Page-

Page 23 Cook 1 and 2 March 11, 1986 Points Available OUESTION 6.04 List the parameters that will initiate a Safety Injection signal along nith the actual setpoint for each unit. Do not include manual.

(2.5)

ANSWER 6.04 PZR low pressure: 1837 psig on Unit 1,1908 psig on Unit 2 [0.5]

Steam line delta-P:

100 psig [+0.5]

Containment high pressure:

1.1 psig [+0.5]

6 High steam line flow (1.42 to 3.88 x 10 pph,thisvaluenog required for full credit) on Unit I with 10-10 T-avg of 541 F or low steam line pressure of 600 psig

[+0.5]

Low steam line pressure: 600 psig on Unit 2 [+0.5]

Reference (s1 6.04 1.

DCC Lesson Plan, R0-C-NS12.

-Section 6.0 Continued on Next Page-

1 Page 24

, Cook 1 and 2 March 11, 1986 Points Avallabla-m OUESTION 6.05 List the components of the ECCS which receive an actuating signal on a safety injection and what action that component must take.

(2.0)

ANSWER 6.05

/

Centrifugal Charging Pumps - Start

/

CCP Suction from RWST (IM0-910/911) - Open BIT Inlet and Outlet Valves - Open /

/

SI Pumps - Start RHR Pumps - Start CCP Mini Flow Valves - Close /

tr!-"hter Echtf or, '.'aha3 - Oper >

BIT Recirc Valves - Close.

~

s

[+0.1Ffor each component, +0.1 for each action), r.0 th ~~

Reference (s) 6.05 1.

DCC Lesson Plan, R0-C-NS12.

-Section 6.0 Continued on Next Page-

~

i

Page 25 Cook I and 2 March 11, 1986 l

Points AyA1.1able QUESTION 6.06 Liit the four (4) ice condenser monitoring systems that are available to control room operators.

(2.0)

ANSWER 6.06 1.

Ice bed temperature

[+0.5]

2.

Ice condenser door position

[+0.5]

3.

Ice condenser floor cooling temperature

[+0.5]

4.

Equipment and personnel doors

[+0.5]

&, a ww, ara % Me@P Reference (s) 6.06 1.

DCC Lesson Plan, R0-C-NS14, p. 17.

OUESTION 6.07 The reactor trip breaker shunt coils will:

(1.0)

(a.) energize en a manual trip signal.

(b.) energize on an automatic trip signal.

(c.) deenergize on a manual trip signal.

(d.) deenergize on an automatic trip signal.

ANSWER 6.07 (a.)

[+1.0]

orb.

Reference (s) 6.07 1.

DCC Lesson Plan, R0-C-NS11.

-Section 6.0 Continued on Next Page-

l.

Page 26 Cook 1 and 2 March 11, 1986 Points Available OUESTION 6.08 a.

What automatic functions are performed when an Urgent Failure alarm is received?

(1.5) b.

What will cause a Non-Urgent Failure alarm?

(1.0)

ANSWER 6.08 a.

1)

Lift coil is deenergized.

[+0.5]

2)

Stationary and movable gripper coils are both energized at reduced current.

[+0.5]

3)

Stops manual and automatic rod motion.

[+0.5]

b.

Loss of any one of the two 120V AC power supplies.

[+1.0]

Reference (s) 6.08 1.

DCC Lesson Plan, R0-C-NSO4.

-Section 6.0 Continued on Next Page-

Page 27 Cook 1 and 2 March 11, 1986 Points Available OUESTION 6.09 Eillin the set points in the level control program for the Volume Control tank.

(1,0) a.

Hi Level alarm - full divert b.

Stops auto makeup c.

VCT low level alare d.

Shifts charging pump suction to RWST.

ANSWER 6.09 a.

87

[+0.25]

1+

b.

27*'

[+0.25]

c.

7

[+0.25]

d.

1

[+0.25]

Reference (si 6.09 1.

DCC Lesson Plan, R0-C-NS06.

i i

i i

l

-Section 6.0 Continued on Next Page-

Page 28 Cook 1 and 2 March 11, 1986 Points Available OUESTION 6.10 State the four (4) signals which will cause an automatic closure of the letdown orifice isolation valves (QRV 160, 161, 162).

Include setpoints if applicable.

(2.0)

ANSWER 6.10 1.

Containment Isolation

[+0.5]

2.

Loss of all charging pumps (breakers)

[+0.5]

3.

Closing of letdown isolation valves QRV-111,112 [+0.5]

4.

Low pressurizer water level of 17% [+0.5]

Reference (s) 6.10 1.

DCC Lesson Plan, R0-C-NS06.

-Section 6.0 Continued on Next Page-

Page 29 Cook 1 and 2 March 11, 1986 Points Available OUESTION 6.11 During a progranned poger increase from 0% power, the Auctioneer T-avg increases to 560 F.

The PZR levels should increase from (2.0)

U 1 to i

U

% to 2

Show calculations.

ANSWER 6.11 U 22% [+0.25] to 37% [+0.75]

t Ut Lvl = 22 + (46 - 22) 567 8 - 47 = 37 22% [+0.25] to 38%

[+0.75]

U2 U2 Lvl = 22 + (55 - 22) 5 8-7 = 38 Reference (s) 6.11 1.

DCC Lesson Plan, R0-C-NS03, p. 27.

2.

DCC Lesson Plan, R0-C-NSO4, Figure 2.1.1 and Figure 2.1 or R0-C-NS-06, page 26.

-Section 6.0 Continued on Next Page-

Page 30 Cook I and 2 March 11, 1986 Points Available OUESTION 6.12 The Pressurizer Pressure Relief System is designed to prevent a reactor trip under four (4) transient conditions. Litt them.

(2.0)

ANSWER 6.12 1.

Load at 5%/ min with auto rod control.

[+0.5]

2.

Unload at 5%/ min with auto rod control.

[+0.5]

3.

10% step change with auto rod control.

[+0.5]

4.

Step change from 100% to auxiliary load power with auto rod control and steam dumps.

[+0.5]

Reference (s) 6.12 1.

DCC Lesson Plan, R0-C-NS03.

-Section 6.0 Continued on Next Page-

Cook 1 and 2 Page 31 March 11, 1986 Points Available 00ESTION 6.13 Fill-in the blanks to make the following statements about reactor coolant pump operation correct.

a.

Minimum VCT pressure necessary for proper seal operation is psig.

(0.25)

U b.

Seal water injection temperature should not exceed F

as read at the VCT.

(0.25) c.

A pressure differential of at least psig should be maintained across the No. I seal.

(0.25) d.

Operation above F requires 4 RCPs to be in service.

(0.25)

ANSVER 6.13 a.

15

[+0.25]

b.

130

[+0.25]

• kbI5

[+0.25]

c.

d.

541

[+0.25]

Reference (s) 6.13 1.

DCC Lesson Plan, OHP-4021.002.003.

-Section 6.0 Continued on Next Page-

Page 32 Cook I and 2 March 11, 1986 Points Available OUESTION 6.14 What activities does the failed fuel detector system monitor when in service? (tiate any differences between units.)

(1.0)

ANSWER 6.14 Gross Gamma, I-135, Cs-137 Unit 1

[+0.5]

Gross Gamma, I-131 Unit 2 [+0.5]

CW V"~ f Om f%,w,n 59 cviast LUIsby I'#O C+C.53 Reference (s) 6.14 1.

DCC Lesson Plan, R0-C-NS17.

-End of Section 6.0-

O Page 33 Cook I and 2 March 11, 1986 Points Available

7.0 PROCEDURES

NORMAL. ABNORMAL. EMERGFNCY. AND RADIOLOGICAL CONTROL (25.0)

OUESTION 7.01 Operating Procedure OHP-4021.012.003 " Rod Control and Position Indication Operation", lists three (3) conditions that require insertion of the control banks during unit startup. List two (2)oftheseconditions.

(1.0)

ANSWER 7.01 1.

Critical with control bank height less than insertion limits for 0 power, Technical Specification 3.13.5.

2.

Critical at 1000 pcm below ECP.

3.

Rod withdrawal to 500 pcm past ECP and startup not guided by a 1/M plot.

Any two (2) [+0.5] each.

Reference (s) 7.01 1.

DCC Lesson Plan, R0-C-NSO4.

2, DCC Operating Procedure OHP-4021.012.003.

-Section 7.0 Continued on Next Page-

i Page 34 Cook 1 and 2 March 11, 1986 Points Available 00ESTION 7.02 Assume during reactor cooldown or shutdown the RHR capability is lost. List three (3) alternate cooling methods.

(State anyassumptions.)

(1.5)

ANSWER 7.02

Mi

r 1.

Condenser steam dump.

[+0.5]

2.

Atmospheric steam dump.

[+0.5]

M eff~

3.

Spent fuel pit cooling.

[+0.5]

Reference (s) 7.02 i

1.

DCC Emergency Operating Precedure OHP-4023.017.001, 3.

p aa-u.,.( c m p4 th. a as -r-z Of fsbe"- 4d ( IJ '7 (4[, $

4, E,., t, i j m e m., u.

7

s. v. ~,t m.e 4 u. j <.a,ca.g

f. Es+,tt:ss.

r.n ~~~ << L e t, t a V+y 4 6te.,

o, 4 p.., 4 ap,, [, g Q D c,$ }

Q cg....h c e -se.-

• s.' ve 4 n 4. s. ssn;t, (hl 5 UI -,,')

-Section 7.0 Continued on Next Page-

l Page 35 Cook I and 2 March 11, 19P5 Points Available QUESTION 7.03 Answer the following statements about Reactor Coolant Pump (RCP) operation TRUE or FALSE.

a.

The Shift Supervisor's approval is required before starting anRCPwithlessthan0.3gpmleakoffflowfromNo.Iseal.(0.5) b.

ihe maximum vibration allowed is greater on a ramp increase than during steady operation.

(0.5) 0 c.

If the RCS temperature is higher than 150 F, an emergency diesel must be operable when any RCP is running.

(0.5) d.

AnRCPcanbestartedwhiletheplantisgolid,aslong as the RCS temperature is higher than 175 F.

(0.5)

ANSWER 7.03 a.

TRUE [+0.5]

b.

FALSE [+0.5]

c.

TRUE [+0.5]

d.

FALSE [+0.5]

Reference (s) 7.03 1.

DCC Operating Procedure 2-OHP-4021.002.003, pp. 4-5.

i

-Section 7.0 Continued on Next Page-l

Page 36 Cook 1 and 2 March 11, 1986 Points Available 1

00ESTION 7.04 List the four (4) conditions that must be satisfied before a spurious Safety Injection can be reset.

(2.0)

ANSWER 7.04 1.

RCS pressure > 2000 psig and increasing.

[+0.5]

2.

PZR level above no-load value.

[+0.5]

U 3.

Subcool margin > 45 F for Thot RTD or 33 F for incore T/Cs. [+0.5 S/G wide range H O levels rising.

[+0.5]

4.

2 Reference (s) 7.04 1.

DCC Emergency Operating Procedure 1-OHP-4023.001.n02, p.5.

l j

-Section 7.0 Continued on Next Page-1

Page 37 Cook 1 and 2 March 11. 1986 1

Points Available OUESTION 7.05 i

Answer the following questions about the emergency diesel engines.

a.

h[ hat condition will cause the inconglete start sequence to occur?

(0.5) b.

tktw can the operator in the control room tell this has occurred?

(0.5) c.

List the two (2) conditions either of which will pennit a restart.

(0.5) 4 ANSWER 7.05 a.

During a EDG start ( 95% rated speed > 10 sec.

[+0.5]

b.

A red failure light is energized in the control room.

[+0.5]

s c.

? r'....

wi.p.

or push the EDG reset button.

[+0.5]

WM aiu p t (I E w n. t s6. c f.,

g,, q c,.,4,9 pea Reference (s) 7.05 1.

DCC Lest.on Plan, Vol. III, R0-C-ASIO, p. 25.

t

-Section 7.0 Cor.tinued on Next Page-

t Page 38 Cook 1 and 2 March 11, 1986 Points Available 00ESTION 7.06 According to Technical Specifications, ghich of the following does NOT have to be monitored and logged when the plant computer is inoperable.

(1.0)

(a.) rod position (b.) axial flux difference (c.) subcooled margin (d.) quadrant power tilt ratio.

ANSWER 7.06 (c.)

[+1.0)

Reference (s) 7.06 1.

DCC Lesson Plan, Volume III, R0-C-AS20, p. 28.

QUESTION 7.07 During cooldown using the natural circulation emergency procedure, what is the reason for the caution:

"When steaming single steam generator, (for cooldown) care must be taken not to get 100 psid between steam generators"?

(1.0)

ANSWER 7.07 If greater than 100 psid develops between S/Gs, SI will occur.

[+1.0]

Reference (s) 7.07 1.

OCC Emergency 0perating Procedure 1-0HP-4023.001.008, p. 3.

-Section 7.0 Continued on Next Page-

I 4

Page 39 Cook 1 and 2 March 11, 1986 Points Available OUESTION 7.09 a.

The Technical Specification operating temperature limits for the Pressurizer are O

Maximum heat up in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is F.

(0.5)

Maximum cooldown in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is F.

(0.5)

Maximum spray water temperature differential is

'F.

(0.5) b.

Answer TRUE or FALSE. The temperature limits in part a. are based on the pressurizer fatigue limit.

(0.5)

ANSWER 7.08 a.

1.

100"F [+0.5]

2.

200 F [+0.5]

3.

320*F [+0.5]

b.

TRUE

[+0.5]

Reference (s) 7.08 1.

DCC Technical Specifications, Units 1 and 2, 3.4.9.2.

2.

DCC Technical Specifications, Units 1 and 2, 3.4.9.2, (Bases).

-Section 7.0 Continued on Next Page-a

Page 40 Cook 1 and 2 March 11, 1986 Points Available OUESTION 7.09 Match the Technical Specification leakage categories in Column A with the item in Column B (some categories in A may have more than one (1) item B).

(3.0)

A B

a.

Identified leakage 1.

Leakage past reactor head o-ring b.

Unidentified leakage 2.

Leakage through atmospheric c.

Controlled leakage steam dumps d.

Boundary leakage 3.

S/G tube leakage 4.

Weld on loop 2 RTD manifold Tcold 5.

Leakage past No. 2 seal of the RCP 6.

Weld crack-wide range Thot thermowell 7.

PORV leakage ANSWER 7.09 a.

1.

F+0.5'

' 0.5' 3.

+

' 0.5' 7.

+

' 0. 5' 4.

+

(for ' item 4. in a. and/or b).

b.

4.

c.

5.

[+0.5]

d.

6.

[+0.5]

Reference (s) 7.09 1.

DCC Technical Specification 3.4.6.2.

2.

DCC Lesson Plan R0-C-NS2A, pp. 14-16.

-Section 7.0 Continued on Next Page-

Page 41 Cook 1 and 2 March 11, 1986 Points available OUESTION 7.10 If system frequency drops to 58.8 Hz due to loss of generating capacity, what five (5) actions are required after 30 minutes if the system frequency does not recover?

(2.5)

ANSWER 7.10 1.

Open breakers K and K-1.

[+0.5]

2.

Verify steam dump controls have functioned to limit T-avg and Pressurizer pressure transients. If not, trip the reactor.

[+0.5]

3.

Monitor reactor power to ensure the rods are moving inward to reduce power level.

If not, take manual control and

[+0.5](Automatic Rod Withdrawal is blocked reduce reactor power.

below15%C-5).

4.

Adjust turbine speed to maintain frequency at 60 Hz. [+0.5]

5.

Monitor T-avg to ensure the Steam Dump System is performing properly; take manual control if required.

[+0.5]

Reference (s) 7.10 1.

DCC Abnomal Operating Procedure 1-0HP-4022.001.002, p. 2.

I lP

-Section 7.0 Continued on Next Page-

Page 42 Cook 1 and 2 March 11, 1986 Points Available OUEST10N 7.11 It is acceptable to continue operation with a leaking PZR Safety Valve if what three (3) conditions are met?

(1.5)

ANSWER 7.11 1.

Charging pump can maintain PZR level.

[+0.5]

2.

Temperature, pressure, level 1.1 PRT can be maintained within limits.

[+0.5]

3.

Leakage must stay below Technical Specification leakage limit for identified leakage (( 10 gpm).

[+0.5)

Reference (sl 7.11 1.

DCC Abnormal Operating Procedure 1-OHP-4022.002.010, p.1.

2.

OCC Technical Specification 3.4.6.2.

-Section 7.0 Continued on Next Page-

Page 43 Cook 1 and 2 March 11, 1986 Points Av111able OUESTION 7.12 If PZR level and VCT level indicate a loss of letdown flow, whatfour(4)luunediateManualActionsarerequired?

(2.0)

ANSWER 7.12 1.

If PZR reaches high level limit, be sure Rx has tripped.

[+0.5]

2.

Reduce charging flow to maintain seal injection.

[+0.5]

3.

Stop charging through Regen. HX. [+0.5]

4.

Use excess letdown path.

[+0.5]

S.

hakeer to r c7 c, + p. ef c,.

G,,,.

e s ertst q t a r 6

b es+, o e m a.,,,, g, y g,,,,

Referencefs) 7.12 1.

DCC Abnormal Operating Procedure 1-OHP-4022.003.001, p. 4.

-Section 7.0 Continued on Next Page-i

Page 44 Cook 1 and 2 March 11, 1986 Points Available OUESTION 7.13 While manually withdrawing the control rod banks the rod movement stops. List the five (5) possible causes, including values where appropriate.

(2.5)

ANSWER 7.13 1.

Power range NI overpower (103%) (C-2).

[+0.5]

2.

Intermediate range nuclear overpower - current equivalent to 20% (C-1).

[+0.5]

3.

Overpower AT - 3% below trip setpoint (C-4).

[+0.5]

4.

Overtemperature AT - 31 below trip setpoint (C-4). [+0.5]

5.

Urgent failure alarm.

[+0.5]

Reference (s) 7.13 1.

DCC Lesson Plan R0-C-NSO4, p. 7.

-Section 7.0 Continued on Next Page-

Page 45 Cook I and 2 March 11, 1986 Points Available OUESTION 7.14 The reactor is operating

• 100% power when a Reactor Trip /

Safety Injection occurs. Reactor pressure has decreased to 1600 psig and containment pressure is 2.5 psig and is increasing.

a.

At what containment pressure should Phase B isolation occur?

(0.5) b.

When should reactor coolant pumps be tripped? State any assumptions.

(1.0)

ANSWER 7.14 a.

2.9 psig.

[+0.5]

b.

1450 psi [+0.5] or they should be tripped within 5 minutes after Phase B isolation occurs. [+0.5]

Reference (s) 7.14 1.

DCC Emergency Operating Procedure 1-OHP-4023.001.002, p.

4.

2.

DCC Abnormal Operating Procedure 1-OHP-4022.034.002, p.1.

-End of Section 7.0-t

Page 46 Cook 1 and 2 March 11, 1986 Points Available 8.0 ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS (Q

2s5 OUESTION 8.01 Answer the following TRUE or FALSE.

a.

The official version of a Radiation Work Permit is stored in the computer.

(0.5) b.

Separate keys are required to access hath the locked Extremely High Radiation (EHR) and locked High Radiation (HR) areas.

(0.5) c.

Areas outlined in magenta on the Radiation Information Board are EHR areas.

(0.5) d.

Areas where the major portion of the body could receive 0.1 R/hr are designated as HR areas.

(0.5) e.

The administrative limit for whole body radiation dose from external sources is 1.25 rem / quarter.

(0,5)

ANSWER 8.01 a.

TRUE [+0.5] (all others are copies) b.

TRUE [+0.5]

c.

FALSE [+0.5]

(outlined and crosshatched in red) d.

TRUE [+0.5]

(100 mR/hr) e.

FALSE [+0.5] (1.0 rem / quarter)

Referencefs) 8.01 1.

DCC PMP 6010. RAD.001, Radiation Protection Manual, Revision 6, p. 36, 122.

2.

DCC PHP 6010. RAD.002, Entry Into High Radiation Areas, Revision 2, pp. 2-4.

-Section 8.0 Continued on Next Page-

Page 47 Cook 1 and 2 March 11, 1986 Points Available 00ESTION 8.02 Answer the following TRUE or FALSE.

a.

A striped tag clearance issued to one shift supervisor is valid for all shift supervisors.

(0.5) b.

A clearance permit may include some equipment with red tags and some with striped tags.

(0.5) c.

A master clearance can be used to isolate a system where numerous unlisted activities are being performed.

(0.5) d.

A supervisor from the same department can sign off a clearance permit if the responsible individual is not available on site.

(0.5)

ANSWER 8.02 a.

TRUE [+0.5]

b.

TRUE [+0.5]

c.

FALSE (all activities must be listed)

[+0.5]

d.

TRUE [+0.5]

Reference (s) 8.02 1.

DCC PMI 2110, p. 3, 5, 9.

-Section 8.0 Continued on Next Page-

Page 48 Cook 1 and 2 March 11, 1986 Points Available QUESTION 8.03 A licensed operator shall not work more than hours in any 48-hour period.

(0.5)

ANSWER 8.03 24 [+0.5]

Reference (si 8.03 1.

Previously in PMI-1050, now cancelled. CAF.

i i

i

-Section 8.0 Continued on Next Page-i; I

t

Page 49 Cook 1 and 2 March 11, 1986 Points Available OUESTION 8.04 a.

Complete the table for minimum staffing, Modes 1, 2, 3 and 4, Technical Specification Table 6.2-1.

License Minimum Shift Cateaory Crew. One (1) Unit SOL (0.2,5)

OL (0.25)

Aux. operator (0.25)

(non-licensed)

STA (0.25) b.

List by licensing category, the personnel that may be shared between Units 1 and 2.

(0.5) c.

For how long a period may the crew be at less than minimum strength, provided steps are taken immediately to restore crew personnel?

(0.5)

MT/'ER 8.04

' 0.25' a.

1

+

' 0.25' 2

+

' 0.25' 2

+

' 0.25l 1

+

' 0.251 b.

SOL

+

' 0.25l:

STA

+

c.

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> [+0.5]

Reference (s) 8.04 1.

DCC Technical Specification Table 6.2-1 and footnotes.

-Section 8.0 Continued on Next Page-

Page 50 Cook 1 and 2 March 11, 1986 Points Available 00ESTION 8.05 a.

During ah emergency, it may be overly time consuming to process a clearance request. How does the Shift Supervisor ensure that control of isolation points, such as valves or circuit breakers, are protected from inadvertent operation?

(1.0) b.

Answer TRUE or FALSE. A Reg Tag clearance permit denotes that equipment is not to be operated except by request of the permit holder.

(0.5)

ANSWER 8.05 a.

The Shift Supervisor dispatches people to guard the points.

[+1.0) b.

FALSE (No manipulation of any kind allowed.)

[+0.5]

Reference (si 8.05 1.

DCC PMI-2110, Clearance Permit System, Revision 9, pp.1-3.

-Section 8.0 Continued on Next Page-1

Page 51 Cook 1 and 2 March 11, 1986 Points Available OUESTION 8.06 For each of the following, list the limiting condition for operation in Mode 6:

Shutdown k,ff (0.5) a.

b.

Shutdown time prior to fuel movement (0.5) c.

Water level above vessel flange (0.5) d.

Required nuclear instrumentation (1.0)

ANSWER 8.06 k,ff 5 0.95 [+0.5]

a.

b.

subcritical 2 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> [&0.5]

c.

2 23 ft over flange [+0.5]

d.

Two (2) SR monitors operable and operating with continuous visual indication in control room [+0.5] and audible in control room and containment.

[+0.5]

Reference (s) 8.06 1.

DCC Technical Specifications 3.9.1, 3.9.3, 3.9.10, 3.9.2, respectively.

/f

-Section 8.0 Continued on Next Page-

r v

Page 52 Cook 1 and 2 March 11, 1986 Points Arallable OUESTION 8.07 Identify har the RHR system is used to satisfy Technical Specification 3.4.9.3, " Overpressure Protection System",

when the RCS is cold and pressurized.

(1.0)

ANSWER 8.07 The RHR pump suction relief valve [+1.0] (in conjunction with two (2) operable PORVs) satisfies (paragraph b. of) the Technical Specification.

Reference (s) 8.07 1.

DCC Technical Specification 3.4.9.3.b.

2.

DCC RHR Lesson Plan, R0-C-NS08, Figure NS-8-5.

QUESTION 8.08 Describe the two (2) boron injection flow paths required to be operable by Technical Specification 3.1.2.2 while in Modes 1, 2, 3, and 4 (2.0)

ANSWER 8.08 1.

Boric acid tanks via a borte acid transfer pump and charging pump to RCS.

[+1.0]

2.

RWST via charging pump to RCS.

[+1.0]

Reference (s) 8.08 1.

DCC Technical Specification 3.1.2.2, paragraphs a. and b.

2.

DCC CVCS Lesson Plan, R0-C-NS06, Figure NS-6-9.

-Section 8.0 Continued on Next Page-

i f

Page 53 Cook 1 and 2 March 11, 1986 Points Available 00ESTION 8.09 Answer the follewing with regard to Technical Specification 3.2.1 on Axial Flux Difference limits, a.

When opere. ting above 90% power, how much time is the operator allowed to get AFD within the target band before he must reduce power?

(0.5) b.

When operating between 50% and 60% power level, whAt is the time limitation on the operator for operating outside the AFD target band?

(0.5)

ANSWER 8.09 a.

15 minutes

[+0.5]

b.

I hour in 24

[+0.5]

Reference (s) 8.09 1.

DCC Technical Specification 3.2.1.

-Section 8.0 Continued on Next Page-

Page 54 Cook 1 and 2 March 11, 1986 Points Available OUESTION 8.10 If one (1) emergency diesel generator is taken out of service for maintenance and the reactor is in Mode 1, what action must be taken?

(1.0)

ANSWER 8.10 Demonstrate operability of remaining AC sources [+0.5] (by perfoming Surveillance Requirements 4.8.1.1.1.a and 4.8.1.1.2.a.4) within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> [+0.5] (and at least once every 8 hh % ^ " "'r

• hours thereafter), 6 m,n e('TP **'?; _

G e, Q Lud-W % % W Reference (sl 8.10 1.

DCC Technical Specification 3.8.1.1, Action a.

-Section 8.0 Continued on Next Page-

Page 55 Cook 1 and 2 March 11, 1986 Points Available OUESTION 8.11 According to Technical Specification 3.4.11 " Relief Valves",

three (3) PORVs and their associated block valves shall be OPERABLE.

a.

If a single PORY is inoperable. What action (s) must be accomplished to allow plant operation at 100% power to continue?

(1.5) b.

What is the minimum number of OPERABLE PORVs required to allow at power operation to continue indefinitely according to this Technical Specification?

(0.5)

ANSWER 8.11 a.

1.

Restore PORY to OPERABLE status [+0.5] or 2.

Close the associated block valve [+0.5] and 3.

Remove power from the block valve [+0.5].

b.

0 (if "one or more" are inoperable, close block valves)

[+0.5]

Reference (si 8.11 1.

DCC Technical Specification 3.4.11.

-Section 8.0 Continued on Next Page-

Page 56 Cook 1 and 2 March 11, 1986 Points Anilable OUESTION 8.12 For each of the following ECC Systems, specify whether the Technical Specification LCO is the SAME or DIFFERENT for Units 1 and 2 for Mode 1 operation.

a.

Accumulator water volume, Technical Specification 3.5.1 (0.5) b.

Accumulator nitrogen cover-pressure, Technical Specift-cation 3.5.1 (0.5) c.

RWST boron concentration, Technical Specification 3.5.5 (0.5) d.

RWST minimum water temperature, Technical Specification 3.5.5 (0.5)

ANSVER 8.12 a.

SAME [+0.5]

b.

DIFFERENT (U : 585-658 psig; U : 599-644psig)

[+0.5]

1 2

c.

DIFFERENT (U :

950 ppm; U : 2000-2200 ppm)

[+0.5]

1 2

d.

DIFFERENT (0 : 70 F; U : 80 F)

[+0.5]

Win

,W e pt 5 4 5 4c-b h '" #

0 1

2 d

e. -.L + c 4 i. p i, i > w t s, t te, a.e m

9, c,

Reference (s) 8.12 1.

DCC Technical Specification 3.5.1.

2.

DCC Technical Specification 3.5.5.

-Section 8.0 Continued on Next Page-

Page 57 Cook 1 and 2 March 11, 1986 Points Available 00ESTION 8.13 State whether each of the following, as specified in 10 CFR 72, requires NRC notification within 1 HOUR or 4 HOURS.

a.

Shutdewn required by Technical Specifications.

(0.5) b.

Reactor trip.

(0.5) c.

Declaration of an unusual event.

(0.5) d.

h=fer af = cent: f sted Hi"td=10 : !cce!-h;;it:1.

(0.5).<-

ANSWER 8.13 a.

I hour [+0.5]

b.

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />

[+0.5]

c.

I hour [+0.5]

4.

k k-..

[4.sj ^~

Q, A cle ici,,(

,s e.+5e.

o<s-e<

A.5-r,

• s re cu,8ty j: c t c s t.'. I g c a r,,,t.

d e t :,

Reference (s) 8.13 c,,nt,, ms,. p, gw..( c n,0 y y,,,.4 s Ise., e c 4. S ' e

1. C '

1.

10 CFR 50.72.

l l

l

-Section 8.0 Continued on Next Page-

Page 58 Cook 1 and 2 March 11, 1986 Points Available OUESTION 8.14 A plant emergency occurs which requires activation of the D.C. Cook Emergency Plan.

A.

Wha is responsible for initiating PHP 2080 EPP.001, Emergency Plan Activation and Condition Classification?

(0.5)

B.

At what minimum emergency condition classification is activation of each of the following required?

a.

Technical Support Center.

(0.5) b.

Operations Staging Area.

(0.5) c.

Emergency Operations Facility.

(0.5)

C.

Liit four (4) individuals who can relieve the initial On-Site Emergency Coordinator.

(1.0)

ANSWER 8.14 A.

Shift Supervisor [+0.5]

' 0. 5' B.

a.

Alert

+

Alert l+0.5l rinup e<+.

a.- - e-- --- v c*

Q kM& Qp Lvo.s.1 C.

1.

Production Supervisor of Affected Unit 2.

Operations Superintendent 3.

Assistant Plant Manager - Operations 4.

Assistant Plant Manager - Maintenance 5.

D. C. Cook Plant Manager 6.

Vice President, Nuclear Operations Any four (4) [+0.25] each.

-Section 8.0 Continued on Next Page-

Page 59 Cook 1 and 2 March 11, 1986 Points Available Reference (s) 8.14 1.

DCC PMP 2080 EPP.001, Emergency Plan Activation Classification, pp.1-2.

2.

DCC PMP 2080 EPP.015, Responsibilities of the On-Site Emergency Coordinator, p. 2.

-End of Section 8.0-

-End of Exam-e

e t

EQUATION SHEET Where mi = m2 (density)1(velocity)1(area)1 = (density)2(velocity)2(area)2 2

PE +KE +P V 1 i = PE +KE +P Y22 where V = specific KE = mv PE = agh 2

2 i

i li volume P = Pressure Q = ic (Tout-Tin)

Q = UA (T

-T l

0

• 5Ih -h )

p ave stm l 2 P = P 10(SUR)(t) p, p e /T SUR = 26.06 T = (B-p)t t

o o

T-p delta K = (Kef f-1)

CR (1-Keff1) = CR Il-Keff2)

CR = S/(1-Keff) 1 2

~

M = (1-Keffi)

SDM = (1-Keff) x 100.

(1-Keff2)

Keff 1 = A e-(decay constant)x(t)

In (2) 0.693 A

decay constant

=

=

n t

t1/2 1/2 T

Water Parameters Miscellaneous Conversions 10 1 gallon = 8.345 lbs 1 Curie = 3.7 x 10 dps 1 gallon = 3.78 liters 1 kg = 2.21 lbs 3

1 ft3 = 7.48 gallons 1 hp = 2.54 x 10 Btu /hr 3

6 Density =62.4lbg/ft 1 MW = 3.41 x 10 Btu /hr Density = 1 gm/cm 1 Btu = 778 f t-lbf Heat of Vaporization = 970 Btu /lbm Degrees F = (1.8 x Degrees C) + 32 Heat of Fusion = 144 Btu /lbm I inch = 2.54 centimeters2 1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 f t-lbm/lbf-sec

l

.f EQUATION SHEET Where mi = m2 (density)1(velocity)1(area)1 = (density)2(velocity)2(area)2 2

PE +KE +P Y 1 i = PE +KE +P V22 where Y = specific KE = mv PE = agh 2

2 i

i li volume P = Pressure Q = sc (Tout-Tin)

Q = UA (T

-Tstm)

Q = m(h -h )

p ave i 2 P = P 10(SUR)(t) p p e /T SUR = 26.06 T = (B-p)t t

o o

T p

CR (1-Keffl) = CR Il-Keff2)

CR = S/(1-KeffI delta K = (Keff-1) 1 2

~

M = (1-Keff1)

SDM = (1-Keff) x 100%

(1-Keff2)

Keff 1 = A e-(decay constant)x(t) in (2) 0.693 A

decay constant

=

=

g t

t1/2 1/2 Water Parameters Miscellaneous Conversions 10 1 gallon = 8.345 lbs 1 Curie = 3.7 x 10 dps I gallon = 3.78 liters 1 kg = 2.21 lbs 3

t I hp = 2.54 x 10 Btu /hr l

1 ft3 = 7.48 gallons 6

3 1 MW = 3.41 x 10 Btu /hr l

Density = 62.4 lbg/f t Density = 1 gm/cm 1 Btu = 778 f t-lbf Heat of Vaporization = 970 Btu /lbm Degrees F = (1.8 x Degrees C) + 32 Heat of Fusion = 144 Btu /lbm 1 inch = 2.54 centimeters2 l

1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 f t-lbm/lbf-sec 1

l i

[

l

.