ML20215C523

Exam Rept 50-255/OLS-86-01 on 860624,0722-24 & 0827.Exam Results:Both Reactor Candidates & Seven Senior Reactor Operator Candidates Passed

ML20215C523
Person / Time
Site:	Palisades
Issue date:	09/30/1986
From:	Clark R, Higgins R, James Smith, Upton J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20215C516	List: ML20215C513 ML20215C523
References
50-255-OLS-86, 50-255-OLS-86-0, NUDOCS 8610100320
Download: ML20215C523 (139)

Text

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U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-255/0LS-86-01 Docket No. 50-255 License No. DPR-20 Licensee: Consumers Power Company 212 West Michigan Avenue Jackson, MI 49201 Facility Name:

Palisades Nuclear Plant Examination Administat ed At:

Palisades Nuclear Plant, Covert, MI and Midland Training Center, Midland, MI Examination Conducted: June 24, July 22, 23 and 24, and August 27, 1986 Examiner (s):

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?[3d[f6 Date JV R. G. Clark

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Date-h J. D. Smit 3ddl Date

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4f&k Approved By:

hom)as M. Burdick, ChiefV b

o Operating Licensing Section Da'te '

Examination Summary Examination administered on June 24, July 22, 23 and 24, and August 27, 1986 QeportNo. 50-255/0LS-86-01)

Examinations were administered to two reactor operator candidates and eight senior reactor operator candidates.

Results: Both reactor operator candidates and seven senior reactor operator candic'ates passed.

8610100320 860930

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PDR ADOCK 05000255 V

PDR

REPORT DETAILS 1.

Examiners R. L. Higgins, RIII R. G. Clark, PNL J. D. Smith, PNL J. W. Upton, PNL 2.

Examination Review Meeting An examination review meeting is no longer conducted. A copy of the examinations and answer keys are left with the facility after the last candidate completes his exam. The facility has five working days in which to make comments to the NRC concerning the examinations and answer keys. The following paragraphs contain the facility comments concerning the reactor operator examination followed by the NRC response.

Question 1.11 Facility Comment:

The examination answer key indicates a value of lambda equal to.08.

Our student handout assumes a value of.1 for lambda, which would change the answer to the question of rod height equal to 92 inches (+ or - 2 inches).

Recommendation:

An alternate answer for rod height should be acceptable based on the value of lambda assumed.

Reference:

Palisades student handout SH-PQD0 Reactor Startup, Objective No. 4.24 (PQD00K5.20).

NRC Response:

Agree. The answer key was expanded to also award full credit if the candidate used a value of lambda equal to.1.

Question 2.01 4

Facility Comment:

The examination answer key indicates a minimum of two gross radioactivity monitors for a liquid batch release.

The reference cited on the key indicates only one monitor is required, which agrees with Palisades Technical Specifications.

Recommendation:

Change the answer key to reflect "B. One" as the correct answer for the minimum number of gross radioactivity monitors for a liquid batch release.

Reference:

Palisades Technical Specification Table No. 3.24-1, Student handout SH-PNPD Radwaste, Objective No. 4.7 (PNPD0G5.03).

NRC Response:

Agree. The answer key was change to "B."

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Question 2.02a Facility Comment:

The examination answer key indicates that the diesel generator K-6A jacket water cooler is supplied by critical header A.

The reference cited indicates critical header B is the cooling water supply for K-6A.

Recommendation:

Change the examination answer key to critical header B.

References:

Palisades P&ID M-208 SH.1A, Student handout SH-PNAA Service Water System, Objective No. 4.4 (PNAA0K1.01).

NRC Response:

Agree. The answer key was changed to " critical header B."

Question 2.04c Facility Comment:

The examination answer key indicates the motive force for filling a SIT is a HPSI pump.

Depending upon the mode of operation assumed in answering the question, the containment spray pumps may also be used.

Recommendation:

Accept containment spray pumps as an alternate answer.

Reference:

S0P 4 Containment Spray and Iodine Removal System Procedure Section 7.1.1.

NRC Response:

Agree. The answer key was expanded to also grant full credit for the response " containment spray pumps."

Question 2.07 Facility Comment:

The examination answer key describes two flow paths by which decay heat is removed from the core to the " ultimate heat sink" immediately following a post LOCA Recirculation Actuation Signal.

The answer key is not in accordance with the reference cited.

Recommendation:

Revise answer key to reflect the reference cited.

Reference:

Student Handouts:

SH-PNMB Containment Spray System, Objective No. 4.14 (PNMB0G4.05), and SH-PNMC Containment Air Cooling System, Objective No. 4.4 (PNMC0K1.01).

NRC Response:

Agree. The following changes to the answer key were made:

(1) Accepted "LPSI pumps" as a correct alternative to "minirecirc pumps"; (2) Replaced " cooling towers" with

" lake"; (3) Accepted "CCW" as a correct alternative to "CNTM air coolers."

3

Question 2.10c Facility Comment:

The examination answer key only lists one source of water to routinely fill the spent fuel pool. PMW or UW may also be used to fill the spent fuel pool on a routine bases.

Recommendation:

Accept PMW via P-90 or UW via P-91 as alternate answers.

Reference:

SFP0-1 Addition of Water to the Spent Fuel Pool.

NRC Response:

Agree. The answer key was modified to also grant full credit for any of the following responses:

" utility water";

" primary makeup water"; or " gravity feed from the SIRWT."

Question 2.11b Facility Comment:

The examination answer key lists "none" as the effect of a turbine trip signal on the feedwater regulating valves.

This answer is based on the reference cited (PNGAOK3.01),

which is a learning objective looking at a very specific interrelationships between systems that had been previously presented. There is an additional reference that was not cited that describes the interrelationship between a turbine trip and the feedwater regulating valves. As described in that reference the valves " remain in their as is position."

Recommendation:

Change the examination answer key from "None" to "as is."

Reference:

Student handouts:

SH-PNGA HP and LP Turbines, Objective No. 4.15 (PNGAOK3.01), SH-PNFB Steam Generator Water Level Control, Objective No. 4.5 (PNFBOK5.03).

NRC Response:

Agree. The answer was changed to "as is."

Question 2.12b,c,d,e,i Facility Comment:

The answers listed on the examination answer key are based on the reference cited (EOP-1, Attachment 3, Revision 16) which provides Reset SIS repositioning criteria; however, the question deals with SIS Actuation response.

Recommendation:

Change the examination answer key to:

b.

trip c.

all 3 start with standby power available d.

closed e.

closed i.

no effect

References:

Palisades Logic Diagram Engineered Safeguards E-17 SH.3; E0P-1, Attachment 3, Revision 16; XE-17 SH.11.

NRC Response:

Agree. The answer key was changed to b. " trip"; c. "all 3 start with standby power available"; d. " closed"; e.

" closed"; and 1. "no effect."

4

Question 3.01a Facility Comment:

The examination answer key states "Both valves shut" which is not correct, only the main feedwater regulating valve on the affected steam generator will shut.

Recommendation:

Change the examination answer key to "the affected steam generator feedwater regulating valve would shut."

Reference:

Palisades P&ID M-207 SH. 1, Palisades System Information Manual Chapter 6b, Page 26, Section c.2.

NRC Response:

Agree. The answer key was changed to "the feedwater regulating valve for the affected steam generator would shut."

Question 3.02 Facility Comment:

The examination answer key lists the synchro-transmitters as providing the input signal to the nixie tubes; actually, the synchro-transmitters provide the input signal to the Position Indication Primary (PIP) which provides the input signal to the nixie tubes. The key lists the CRDM limit switches as the input devices to the matrix board display; however, the Secondary Position Indication (SPI) provides input to the matrix board display as well.

Recommendation:

Change the answer key to accept, for full credit, either the synchro-transmitters provide the input signal or the Position Indication Primary (PIP) provides the input signal to the nixie tubes.

Change the answer key to also accept SPI as an input to the matrix board display.

I

Reference:

Palisades System Information Manual Chapter 10b, Pages 10 and 11.

NRC Response:

Agree. The following changes were made to the answer key:

(1) Accept either " PIP" or " synchro-transmitters" as providing the input signal to the nixie tubes; (2) Accept either " SPI" or " limit switches" as providing the input signal to the matrix board display.

Question 3.03 Facility Comment:

The examination answer key is not all inclusive of the automatic actions associated with loss of the instrument AC bus.

Recommendation:

Change the examination answer key to also accept che items listed in the references attached.

5

4

References:

ONP-24.5 Loss of Instrument AC Bus Y01, Revision 11; ARP-24, Revision 36, annunciator number 30; and P&ID M-653 SH.3.

NRC Response:

Expanded the answer key to also grant credit for the following responses:

(1) SPI is lost; (2) PIP is lost; (3) Rod matrix display is lost; (4) Startup, intermediate and power range recorders are lost; (5) Cooling tower pumps trip; (6) Pressurizer PORV temperature indication is lost; (7) All pressurizer temperature indications are lost; (8) Containment temperature and humidity indications are lost; (9) Letdown and charging temperature and flow indications are lost; (10) Boric acid storage tank levels are lost; (11) AFW discharge pressure indication is lost; and (12) CCW surge tank level indication is lost.

Question 3.05d Facility Comment:

The examination answer key lists the suction damper (P0-8001) closing as a result of an automatic action from the fuel handling building radiation monitor (RE-5712).

The damper only closes automatically on the fan motor being de-energized.

Recommendation:

Change the examination answer key to eliminate the closing of the damper for full credit as an automatic action of the associated radiation monitor.

Reference:

Palisades P&ID M-658, Revision 12.

NRC Response:

Agree.

The answer key was modified to no longer require

" closes suction damper (PO-8001)" for full credit.

Question 3.07 Facility Comment:

The examination answer key does not include all of the Engineering Safety Features that could possibly be actuated on a feedline rupture inside the containment at 95% of full power.

Recommendation:

Add to the examination answer key as alternate answers to the following:

Main feedwater regulating and bypass valves closing on the affected steam generator low pressure of 500 psi.

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~ --

Auxiliary feedwater initiation on two out of four low level in either steam generator of 28.7%.

Control Room ventilation actuation on Containment High Pressure (CHP).

Reactor trip on CHP or Low S/G Pressure.

Reference:

Palisades ARP-21 Annunciators:

Page 4, Number 8, Page 5, Number 2., Page 9, Number 8.; Student Handout SH-PNPC Control Room Ventilation System, Objective No. 4.32 (PNPCG12.04).

NRC Response:

Agree. The answer key was expanded to also grant credit for the following:

(1) " main feedwater regulating and bypass valves closing on the affected steam generator if pressure drops below 500 psi;" (2) " auxiliary feedwater initiation due to a level indication of 28.7% or less on two out of four steam generator level indicators;" (3) " control room ventilation actuation on containment high pressure;" and (4) " reactor trip on containment high pressure or low steam generator pressure."

Question 3.08a Facility Comment:

The examination answer key does not reflect the same answer as cited in the reference.

Recommendation:

Include in the examination answer key an alternate answer that reflects the cited reference.

References:

Student Handout SH-PNEB Auxiliary Feedwater System, Objective No. 4.10 (PNEBG12.02).

NRC Response:

Modified the answer key to also accept " Feed Only Good Generator (FOGG) is unreliable."

Question 3.09a Facility Comment:

The reference cited in the examination answer key had nothing to do with the question that was asked.

In order to answer the question, formulas and values for the constants and conversion factors that the I&C Department is given and uses while calibrating the instruments should have been provided to the operators during the test.

Furthermore, our calculations do not support a calculated trip set point of 2300 psia which would result in a reactor trip per the answer key for Channel A.

We assumed a normal Th reading of 585 degrees F for non affected channels, a Th reading of 600 degrees F for the affected channel, Tc readings 535.5 degrees F and a PCS pressure of 2010 psia. Using the attached I&C Department Monthly Technical Specification Test (MI-2A), we calculated the less than full power value (PY 1) 7

at 2.46 volts and the greater than full power ~value (PY 2) at 3.012 volts. Using the standard I&C volt to psia conversion formula. {[(volt-1)/4]*1000}+1500 we'obtained a value of 1865 psia for a calculated trip set point for the less than full power condition and a value of 2003 psia calculated trip set point-for the greater than full power condition. These calculated trip set points result in a

-trip free condition on Channel A with a PSC pressure of 2010 psia.

Recommendations:

Because the examinees were unable to calculate the absolute value for the set point, the examination answer key should be changed to give full credit to answers that state the assumption of a Channel A TM/LP trip or trip free condition and then demonstrate the ability to determine if a reactor trip occurred in accordance with the reference cited.

Reference:

Student Handout SH-PNNB Reactor Protective System, Objective No. 4.41 (PNNBG27.01); Palisades Technical Specification Test Reactor Protective Trip Units Procedure No. MI-2A, Revision 2.

NRC_ Response:

Agree.

Since the exact trip setpoint for the TMLP trip can not be calculated with the information available in the facility literature, it is uncertain whether a trip would occur. The answer key was modified to award full credit if mention was made that the TMLP setpoint was raised.

Question 3.09c Facility Comment:

Based on the interpretation of the question "PCS flow indication to" the examinee could assume failure of the same signal that provides input to the flow indication as to the trip devices.

Recommendation:

Include in examination answer key an alternate answer that allows for an assumption of a trip condition on Channel C and therefore concludes with a reactor trip.

References:

Palisades P&ID M-204 SH. 1, Revision 33.

NRC Response:

Agree.

If the candidate mentions that the flow indication to Channel C is a differential pressure signal, the candidate will be awarded full credit for stating that the reactor would trip.

Question 3.10 Facility Comment:

The reference cited states, "using all available indications."

Recommendation:

Accept alternate listings of "all available indications,"

especially with respect to Parts "c" and "d."

8

Reference:

Palisades E0P-1, Attachment 3, Revision 16, Page 1.

NRC Response:

Partially agree. The answer key will be modified as follows:

(1) the response " core exit thermocouples" will also be accepted as a correct response for Part "a;" (2) the statement "on two channels" will not be required in part "d" for the " steam flow" response.

Question 4.04 Facility Comment:

The objectives in our licensed operator training program do not require trainees to memorize General Operating Procedures.

However, we do include objectives similar to the following:

Given a simulator scenario with only one start up detector available, using GOP-3, perform an approach to criticality.

NRC Response:

The candidates were not required to memorize the procedure, but they should be familiar enough with the procedure to know the major steps.

Question 4.09a Facility Comment:

The examination answer key indicates that the basis for establishing and maintaining PCS temperature at 525 degrees F to 530 degrees F is to maintain subcooling margin before initiating cooldown, which is not reflected in the reference cited.

Recommendation:

Revise the examination answer key to indicate that the basis for establishing and maintaining PCS temperature at 525 degrees F. to 530 degrees F. is to maintain shutdown margin before initiating cooldown.

Reference:

Palisades E0P8.2, Revision 15, Pages 1-7.

NRC Response:

Agree. The answer key was changed to " Maintain shutdown margin..."

Question 4.10 Facility Comment:

4.10a.

Based on the guidance given in our DNP-3 Loss of Feedwater, the value for either tripping the plant or reducing power is given as an approximation due to the necessary averaging of eight channels of indicators disagreeing within their expected allowed tolerances.

4.10c. and g.

Neither of these states a specific value for the amount of PCS leakage, which would allow assumptions to be made by the examinees.

4.10f. The examination answer key does not reflect the reference cited.

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1

. Recommendation:

4.10a. Revise examination answer key to include as an alternate answer " Trip."

4.10c. and g.

Revise examination answer key to include reasonable assumptions that could be made about the amount of PCS leakage, such as, " continue operations" due to PCS leakage within Technical Specification Limits.

4.10f. Change the examination answer key to include as an alternate answer " reduce power" based on guidance given in the reference cited.

Reference:

4.10a. ONP-3 LOSS OF FEEDWATER, Revision 11.

4.10c. and g.

Palisades Technical Specifications, Section 3.1.5.

4.10f.

E0P-5 LOSS OF INSTRUMENT AIR, Revision 13.

NRC Response:

Partially agree. The following modifications were made to the answer key:

(1) for part "a," the NRC disagrees with the facility, so the required answer remains " reduce power;"

(2) for parts "c" and "g," the answer " continue operations if leakage is within Technical Specification Limits" will also be awarded full credit; and (3) for part "f," the answer " reduce power if the leak rate is small enough to allow time for component isolation and additional supply from the Feedwater Purity Building" will be awarded full credit.

The following paragraphs contain the facility comments concerning the senior reactor operator examination followed by the NRC Responses.

Question 5.02 Facility Comment:

If the candidate uses only the Mollier diagram provided and follows constant enthalpy from interpolated points for 1715 psia /715 psia to 35 psia, he has to again interpolate between hundred degree lines which are one-half inch apart.

Therefore, the allowable temperature band for an acceptable answer should be expanded.

Recommendation:

The acceptance band should be expanded to +/- 15 degrees F.

instead of +/- 5 degrees F.

The concept of isenthalpic expansion is what is important, not the exact final temperature.

Reference:

Mollier diagram from test, Attachment 5.02.

NRC Response:

Disagree.

The answer key was not changed.

10

i Question 5.08 Facility Comment:

The question infers exclusion of DNB as a possible answer.

Recommendation:

Specific reference to DNB should not be a criterion to award full credit.

NRC Response:

Disagree. The answer key was not changed.

Question 5.13 Facility Comment:

There are several factors which change over core life to influence flux which thereby changes control rod worth.

The answer key concentrates on only one of those factors.

Additional acceptable factors are suggested.

Recommendation:

Allow the following as an additional acceptable answer:

" Neutron flux increases over the life of the core due to several factors (reduction in boron concentration, depletion of burnable poisons, fuel depletion necessitating an increase in flux to maintain a constant fission rate, etc.).

This increase in flux results in a larger control rod worth since rod worth is proportional to neutron flux (thermal and epithermal)."

Reference:

Palisades student handout SH-PQBO, Attachment 5.13.

NRC Response:

Agree. The answer key was modified to also grant credit for the following responses:

(1) " increase thermal neutron flux at EOL;" and (2) " depletion of burnable poisons and the reduction in boron reduces the competition for neutrons, causing the rods to be exposed to a greater neutron flux."

Question 6.01.b Facility Comment:

The question does not require a listing of the components supplied by CCW in containment.

Recommendation:

Allow the following as an additional acceptable answer:

" Prevent a loss of instrument air from causing a loss of CCW to the components in containment (cooled by CCW)."

NRC Response:

Disagree. The candidate must state the specific CCW components in containment to receive full credit.

Question 6.04 Facility Comment:

The description of the trip devices listed as answers 2 and 3 on the answer key may be described differently by the applicants.

The following descriptions are provided to the examiner for clarification.

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Recommendation:

Allow the following descriptions of the main feed pump trip devices as additional acceptable answers:

1.

The manual trip button (electrical) at the main feed water pump (answer 2) is actually located on the local control console.

2.

The manual trip ' lever (mechanical) at the main feed water pump (answer 3) is actually a button or plunger type trip device.

Reference:

Control Oil System Drawing for Utility Boiler Feed Pump Drive, Attachment 6.04.

NRC Response:

Agree. The answer key was modified as follows:

(1) for answer 2, both phrases "at the main feedwater pump" and "on the local control console" were awarded full credit; and (2) for answer 3, full credit will be given for " button or plunger type trip device" as well as " manual trip lever."

The original answers were taken verbatim from facility literature.

The facility is admonished to revise its literature to reflect actual plant conditions.

Question 6.06 Facility Comment:

The backup sources of seal oil vary depending upon the operating mode of the main turbine (operating or on the turning gear). Therefore, additional acceptable sources are recommended.

Recommendation:

If normal seal oil pressure is decreasing, the seal oil pressure regulator automatically opens to supply backup oil from a variety of sources. These sources are:

• Air Side High Pressure Seal Oil Backup Pump

• Turning Gear Oil Pump

• Emergency Bearing Oil Pump

• Shaft Driven Main Lube Oil Pump Emergency Air Side Seal Oil Backup Pump

• These pumps will supply seal oil via the seal oil pressure regulator. The operating condition of each of these pumps is dependent upon bearing oil pressure and turbine conditions which will determine if each pump is available, not seal oil pressure.

Therefore, the order of these four (4) pumps is irrelevant when considering decreasing seal oil pressure.

It is recommended that the acceptable answers be evaleated accordingly.

12

Reference:

Seal Oil Diagram, Attachment 6.08.

NRC Response:

Agree. The candidates will be awarded full credit regardless of the order in which their answers appear.

Question 6.07 Facility Comment:

One of the automatic trips of the diesel generator output breaker is the diesel engine trip. Therefore, any condition which causes an automatic trip of the diesel engine (thereby causing the breaker to trip) should be an additional acceptable answer.

Recommendation:

Allow the following additional acceptable answers:

-Low lube oil pressure

-Overcrank

-Overspeed

Reference:

Palisades Student Handout SH-PNHB, Attachment 6.07.

NRC Response:

Agree. The answer key was modified to accept " Low lube oil pressure," "overcrank," and "overspeed" as acceptable responses in lieu of the response " engine trip."

Question 6.10.c Facility Comment:

The allowable band for the approxima'te value of Primary Coolant Pump vapor seal pressure is too restrictive based upon operational experience. The PCP vapor seal pressure is manually controlled by throttling a manual valve (MV-2194) which controls PCP seal bleed off flow to the Volume Control Tank. Therefore, if VCT pressure is altered the vapor seal pressures will be effected also. The maximum allowable VCT pressure is 50 psig (High pressure alarm). At 50 psig in the VCT the vapor seal pressure of the PCPs will be approximately 100 psia (See control room traces from SH-PNJD, Objective No. 4.25 supporting information).

Recommendation:

Revise the allowable band for approximate values of PCP vapor seal pressure to 20 to 100 psi.

Reference:

.10.c, excerpts from SH-PHJD and ARP 4.

Palisades P&ID M202, Sheet 1, Revision 32.

NRC Response:

Partially agree.

The answer key was modified to accept answers between 20 and 50 psia. The facility is admonished to revise its literature to correctly reflect actual plant conditions.

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l

l Question 6.10.d Facility Comment:

The allowable band for the approximate value of primary coolant pump amperage is too restrictive based upon plant experience.

Recommendation:

Revise the allowable band for approximate values of Primary Coolant Pump amperage to 600 to 650 amps.

Reference:

Control Room Log Sheet No. 1, May 10, 1986,.10.d.

NRC Response:

Partially agree. The answer key was expanded to "600-630 amps." The facility is admonished to revise its literature to reflect actual plant conditions.

Question 6.11 Facility Comment:

The noun name of the valve downstream of the blender (CV-2155) is commonly described as the " blender outlet valve" vs. the " boric acid makeup valve."

Recommendation:

Allow an additional acceptable answer of " blender outlet valve" as a description of CV-2155.

Reference:

.11, Palisades P&ID M-202, Sheet 1A, Revision 3.

NRC Response:

Agree.

The answer key was modified to also grant credit for the response " blender outlet valve" in lieu of " boric acid makeup stop valve."

Question 6.12.a Facility Comment:

An alternate acceptable answer is to describe only the configuration of the pressurizer pressure " controller" (due to the wording of the question).

("60 to 70 psi lower than the normal pressure setpoint of 2010 psia") is too restric-tive based upon operational experience.

Ideally, the pressure controller should be set 75 gsi lower than the desired PZR pressure. This setpoint will allow the PZR spray valves to start to open thereby maintaining the desired setpoint (2010 psia); however, this is an ideal setpoint which assumes the calibration of the pressure controller and spray valves are exact.

In reality, the operators adjust the setpoint of the controller to whatever pressure is necessary to maintain the desired pressure. Therefore, the acceptable band for the controller pressure should be expanded to "approximately 50 to 75 psi lower than the normal pressure setpoint of 2010 psia."

Recommendations:

1.

Allow an alternate acceptable answer to be restricted only to the configuration of the pressurizer pressure

" controller."

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h, 2.

Expand the allowable range of the pressurizer pressure controller setpoint to "approximately 50 to 75 psi lower than the normal pressure setpoint of 2010 psia."

References:

.12.a, excerpts from:

Palisades Student Handout SH-PNKD (Objectives No. 4.14 and 4.13), Palisades SOP 1, Revision 16.

NRC Response:

Disagree. The answer key was not changed.

Question 6.12.b Facility Comment:

Due to the configuration of the pressurizer pressure control system, the spray valves will be partially open when pressure is being maintained at the desired pressure (as described in the comment to part "a" of the question).

Upon an increasing pressure transient the spray valves will continue to open thereby limiting a pressure increase.

According to the original design of the pressure control system these valves would not open until 2085 psia.

Therefore, in accordance with our present configuration, during a transient which results in increasing pressure, the spray valves would respond more quickly than if the control system was configured according to design.

During normal operation the pressurizer pressure will fluctuate if operated according to design. The controllers for the proportional heaters will not provide the fine control necessary to maintain the desired pressure. The result is a "saw toothed" trace on the pressure recorder.

With our present configuration of the control system the spray valves provide a more smooth pressure control during normal operation.

Recommendation:

Allow an additional acceptable answer of "provides for smoother pressure control."

Reference:

.12.b, excerpts from Palisades Student Handout Sh-PNKD.

NRC Response:

Agree.

The answer key was expanded to also grant credit for the response " smoother pressure control."

Question 6.14 Facility Comment:

The answer to this question (and the plant training material) is incomplete. While it is true that the water is less dense during the recirculation phase this does not fully explain why the flowrate through the containment spray pumps increase post-RIAS.

15

Prior to RIAS the suction pressure of the spray pumps is j

determined by the level, pressure and temperature of the water in the SIRW Tank.

Post-RIAS the suction pressure of I

the pumps is determined by the level, pressure and l

temperature of the water in the containment sump.

The net i

effect of the change in these parameters is to cause the head of the containment spray pumps to decrease upon shifting of the suction to the containment sump, primarily because the suction pressure has increased dramatically.

The net effect is an increase in total spray flow.

Recommendation:

Allow "because the suction pressure increases post-RIAS" as an acceptable answer.

Reference:

.14, FSAR, Table 6-7.

NRC Response:

Agree. The answer key was modified to also grant full crecit for the following response:

"The pump must add greater head during the injection phase than during the recircu'ition phase.

During the injection phase the pump's suction is from the SIRWT and it is pumping against high containment pressure. During the recirculation phase the pump's suction is from the containment sump, so the high containment pressure adds to the pump's suction pressure."

Question 6.18 Facility Comment:

The answer key only lists valve numbers as acceptable answers. Tht. noun names of these valves are provided for clarification.

Recommendation:

Allow the following valve names in addition to valve numbers:

CV-0738, S/G Surface Blowdown Valve CV-0739, S/G Surface Blowdown Valve CV-0770, S/G Bottom Blowdown Valve l

CV-0771, S/G Bottom Blowdown Valve

)

CV-0704, Blowdown Tank to Mixing Basin Valve

Reference:

.19, SOP-7, Revision 9.

NRC Response:

Full credit was awarded if the candidate used the correct valve name instead of the valve number.

Question 7.03.b Facility Comment:

The answer to this question is in E0P 1, Step No. 4.10, which is a subsequent action. Our objectives in the lesson plan for E0P 1 state that students must be able to state symptoms, automatic actions, and immediate actions from memory, without the aid of references.

In accordance with NOREG 1021, ES-402, students were given the clear expectation 16

I that they did not have to memorize specific information in subsequent actions unless there was a specific objective over a particular section of the subsequent actions, but only had to " describe generally the objectives and methods."

We reviewed each of the subsequent actions of each E0P and ONP and identified several such objectives, but did not determine that memorizing the steps in Section 4.10 of E0P 1 was necessary for successful task performance, since the operator would be required to use the procedure in executing the task.

Recommendation:

Allow tne following as an acceptable answer:

" Isolate the steam generators and emergency borate."

NRC Response:

Disagree. Operators must know what to do during circumstances which require prompt action, whether that prompt action is called "immediate action" or not. The candidate must state how to isolate the steam generator and how much to emergency borate in order to get full credit.

Question 7.04 Facility Comment:

In accordance with ES-402 the candidate is expected to

" describe generally the objectives and methods" of subsec,uent actions. Therefore, exact setpoints and datails should not be required for full credit.

Recommendation:

Accept the following answer as the minimum information required for full credit:

1.

Unable to maintain adequate PCS subcooling.

2.

Unable to maintain adequate PZR level.

3.

No steam generators are available for removing heat.

NRC Response:

Disagree. Operators must know what to do during circuinstances which required prompt operator action, whether that action is called an "immediate action" or not. The answer key was not changed.

Question 7.05 Facility Comment:

In accordance with ES-402 the candidate is expected to

" describe generally the objectives and methods" of subsequent actions.

The basis for this question is Step No. 4.9 of E0P 2.1, therefore, exact setpoints and details should not be required for full credit.

)

17 L

Recommendation:

Accept the following answer as the minimum information required for full credit:

"The two conditions are: battery discharge amps must be minimized quickly and maintained at the reduced discharge rate."

NRC Response:

Partially agree. The answer key was modified to grant full credit for the response " battery current is reduced to less than 150 amps within 30 minutes of the onset of the loss of AC Power."

Question 7.06 Facility Comment:

We do not have an objective addressing memorizing Section 4.1 of E0P-4, but have decided to add an objective requiring students to know these general conditions from memory.

Since the basis for the question is a subsequent action (allowing time for reference to procedures) the following answer is suggested.

Recommendation:

Accept the following answer as the minimum information required for full credit:

1.

Primary Coolant Pump seal bleed off temperature is alarming (setpoint of alarm is 180*F.).

2.

Primary Coolant Pump bearing temperature is alarming (setpoint of alarm is 175 F.).

3.

All or most CRDM seal Bleed-off temperatures are alarming (setpoint of alarm is 200 F.).

4.

CCW flow is interrupted and cannot be restored (greater than 10 minutes).

NRC Response:

Disagree. The operator must know the conditions and actions to take when a prompt operator action is required, whether or not that action is called "immediate action." The answer key was not changed.

Question 7.07 Facility Comment:

The wording of this question does not exclude the candidate from providing conditions listed in the immediate actions (see Step No. 3.4) which necessitate tripping the Reactor. The following answer is therefore suggested.

18

Recommendation:

Allow the following as the minimum required information for awarding full credit:

"Any two (2) of the following three (3):

-air pressure less than 50 psig

-instrument air pressure dropping rapidly

-indication of erratic equipment behavior" NRC Response:

Disagree. The operator must know the conditions and actions to take when prompt operator action is required, whether or not that action is called "immediate action."

The answer key was not changed.

Question 7.10 Facility Comment:

In accordance with ES-402 the candidate is expected to

" describe generally the objectives and methods" of subsequent actions.

The basis for this question is subsequent action Step No. 4.24.3.6 of E0P 8.1, therefore, details should not be required for full credit. Our analysis indicated that students do not need to state this information from memory. On-the-job performance would be done with the use of the procedure as an aid.

Recommendation:

Allow the following as the minimum information required for full credit:

" Establish HPSI flow to the PCS, through the PORV's (feed and Bleed)"

Reference:

.10, E0P 8.1, Revision 18 NRC Response:

Disagree.

The operator must know the conditions and actions to take when prompt operator action is required, whether or not that action is called "immediate action."

The answer key was not changed.

Question 7.11 Facility Comment:

The answer key omitted one of the four immediate actions of E0P 9 (Activate Site Emergency Plan) for the " fuel handling accident on the reactor refueling side" event.

Recommendation:

Allow any three (3) of the following four (4) actions for full credit:

1.

Dispatch personnel immediately to secure the equipment access door and the personnel air lock doors.

2.

Notify the Control Room and Evacuate Containment.

l i

19

3.

Evaluate status of containment penetrations and initiate actions necessary to prevent air leakage from containment.

4.

Activate the Site Emergency Plan.

Reference:

.11, E0P-9, Revision *2.

NRC Response:

Agree. The answer key was expanded to also grant credit for the response " activate the Site Emergency Plan."

Question 8.03 Facility Co:nment:

This question does not match the condition statement in our objective which reads: "Given the Plant Administrative Procedures and an employee's work history for the last 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />, determine if the employee would exceed restrictions if assigned overtime." Our objective does not require the student to state this information from memory.

Recommendation:

For comment only. No recommended action for the purposes of this examination.

NRC Response:

Disagree. A senior reactor operator must know how much time operators are allowed by administrative procedure to work during a week.

Question 8.05 Facility Comment:

The answer key does not state that the Shift Engineer must activate the Site Emergency Implementing Procedure during an accident.

Recommendation:

Allow the following as an additional acceptable answer in lieu of " Function as site emergency director until relieved."

" Implement the Site Emergency Implementing Procedures" (thereby establishing the Shift Engineer as the SED)

Reference:

.05, EI-1.

Revision 11 NRC Response:

Agree. The answer key was expanded to also grant full credit for the response " implement the site emergency implementing procedures."

Question 8.06 Facility Comment:

This question does not match the condition statement in our objective. We do require students to know from memory that a refamiliarization requirement exists, but we do not require memorization of the details of that section of the procedures.

We expect the supervisor to look up the procedure when making the determination that a refamiliarization shift is required for somebody.

20

Recommendation:

For comment only. No recommended action for the purposes of this examination.

NRC Response:

Disagree. A senior reactor operator must know whether or not the operators under his supervision require refamiliari-zation training in order to satisfy the currency requirements of the plant administrative procedures.

Question 8.16.a Facility Comment:

Plant Administrative Procedure 4.02 allows a valve / breaker found out of position to be repositioned with the authoriza-tion of the Shift Supervisor.

In such a situation, it would be unnecessary for the person performing the system checklist to note the condition on the " Record of Exceptions" sheet.

Recommendation:

Allow the following additional answer for part "a" of the question:

" Notify the Shift Supervisor and reposition the valve / breaker as authorized."

Reference:

.16.a, Administration Procedure No. 4.02 NRC Response:

Agree. The answer key was expanded to also grant full credit for the response " Notify the shift supervisor and reposition the valve / breaker as directed."

Question 8.19.b Facility Comment:

The plant terminology used for part "b" of the answer may not be familiar to the examiner, therefore, clarification is provided. The floor plugs are identified by the room under the floor plug (i.e., the West Safeguards room floor plug is actually located in the CCW room floor and the Safeguards room ceiling).

NRC Response:

Noted.

Question 8.28.a Facility Comment:

The question does not sufficiently limit the scope of possible correct answers.

For example, one possible correct answer could be "1250 mrem" in accordance with 10 CFR 20 criteria.

Recommendation:

Evaluate the candidates answers in accordance with the assumptions stated.

NRC Response:

Disagree. The question refers directly to a specific limitation stated in the facility's administrative procedure. The answer key was not changed.

21

3.

Exit Meeting On July 24, 1986, the PNL examiners met with representatives of the plant staff to discuss the results of the simulator and oral examinations. The following personnel attended the Exit Meeting:

PNL:

R. G. Clark J. D. Smith J. W. Upton Palisades:

W. G. Merwin R. B. Heimsath S. Ghidotti R. J. Frigo NRC Senior Resident Inspector:

E. R. Swanson The examiners made the following observations concerning the training program:

a.

The candidates had been well trained in the knowledge and use of plant P& ids. They indicated competency in the use of electrical drawings as well as piping diagrams.

b.

The candidates were responsive to the questions of the examiners and demonstrated their intent to convince the examiners that they could operate the power plant in a safe manner.

c.

The overall rating of the simulator teams of candidates for communication would be barely satisfactory. Numerous examples were provided in the meeting.

d.

The candidates did not appear to be well trained in the implementation of the Emergency Plan. This deficiency was noted in both the simulator portion of the examination and in the operating portion.

In the simulator examinations the candidates (as the Shift Supervisor) were reluctant to make a classification because that would be the Shift Engineer's responsibility. Upon followup at the plant, the candidates seemed to have difficulty in classifying an event in which the severity of the conditions would determine the classification.

22

P fs U.

S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY:

_P6 LIQ 8QES _ __

(

REACTOR TYPE:

_PWB-GE________________

DATE ADMINISTERED: _@6/9@Z12____,,____ _ _

EXAMINER:

DUDLEY

_N.__

2 CANDIDATE:

INSIByCIlgNS_IQ_CBND198IEi Write answers on one side only.

Use separate paper for the answers.

on top of the answer sheets.

Points for each Staple question sheet The passing question are indicated in parentheses after the question.

and a final grade of at Crade requires at least 70% in each category least 80%.

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

% OF CATEGORY

% OF CANDIDATE'S CATEGORY

__YOLUE_ _IQIOL

___SGQBE___

_y@(UE__ _ _

COIEGQBy___

-_ 1.

PRINCIPLES OF NUCLEAR POWER

_2Ez99__ _2Ez99 PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

________ 2.

PLANT DESIGN INCLUDING SAFETY

_2Ez99__ _2Ez99 AND EMERGENCY SYSTEMS

=_

-- - 3.

INSTRUMENTS AND CONTROLS

_25z99__ _25z99 4.

PROCEDURES - NORMAL, ABNORMAL,

_2Ez99__ _2Ez99 EMERGENCY AND RADIOLOGICAL r

CONTROL q,-

Totals 199299__

Final Grade I have neither given All work done on this examination is my own.

nor received aid.

_ _ _ Candidate's Signature L

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

1.

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

Restroom trips are to be limited and only one candidate at a time may 2.

leave.

You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.

3.

Use black ink or dark pencil gely to facilitate legible reproductions.

name in the blank provided on the cover sheet of the 4.

Print your examination.

5.

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

6.

Use only the paper provided for answers.

Print your name in the upper right-hand corner of the first page of gach

~/.

section of the answer sheet.

~

Consecutively number each answer sheet, write "End of Category __" as 8.

start each category on a Ogw page, write gnly gn gng sidg appropriate, and write "Last Page" on the last answer sheet.

of the paper, 9.

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

10. Skip at least thrge lines between each answer.

Separate answer sheets from pad and place finished answer sheets face 11.

down on your desk or table.

Use abbreviations only if they are commonly used in f acility litgtatutg.

12.

The point value for each question is indicated in parentheses after the 13.

answer required.

question and can be used as a guide for the depth of

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

15. Partial credit may be given.

Therefore, ANSWER ALL PARTS OF THE DUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

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

You must sign the statement on the cover sheet that indicates that the 17.

is your own and you have not received or been given assistance in work I

completing the examination.

This must be done after the examination has been completed.

l l

l

10. When you complete your examint; ion, you shall:

a.

Assemble your examination as follows:

(1)

Exam questions on top.

(2)

Exam aidc - figures, tables, etc.

(3)

Answer pages including figures which are part of the answer.

b.

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

c.

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

d.

Leave the examination area, as defined by the examiner.

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

D k

PAGE 12_ _ EBI NQ1 EL E S_ DE_ NgGL E 88_ EQWEB_ EL 9NI_ DEE8GIlg N 2 JUEBDQDXNBulGE4_MEBI_IB8NEEEB_eNp_ELylp_ELQW QUESTION 1.01

(.50)

Which one of the f ollowing descriptions best supports the reason why X non reactivity increases sharply after a trip from 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> at 100% power?

A. Xenon decays less rapidly due to a reduction in the neutron flu >:.

B. Iodine half-lif e is meseh shorter than Xenon half-lif e.

C.

Iodine production is greatly reduced and Xenon production is greatly increased due to the reduction in neutron flux.

D.

Due to reduced neutron absorption, Iodine concentration increases, and Xenon decays directly from Iodine, thus Xenon increases.

QUESTION 1.02

(.60)

Select the most correct statement from the following.

A.

I'f two centrifical pumps are in parallel then the combined pump

~

head will be approximately the sum of the individual pump heads.

B.

If two centrifical pumps are in series then the combined power requirements will be approximately equal to the cube of the individual pump power.

C.

If two centrifical pumps are in parallel then the combined flow will be approximately equal to the sum of the individual pump iIows.

D.

If two centrifical pumps are in series the flow of each pump will be approximately equal to the square of the individual pump speed.

QUESTION 1.03 (2.50)

A nearly instantaneous load reduction fron 100% power to 50% power is made. The plant is now at a steady state power level of 50% with Tave at its proper value and no rod motion.

Ecw much makeup water must be added during the next six hours to maintain Tave at its proper value with no rod motion? Assume the initial PCS boron concentration at 50% power is 500 pp:

cnd the plant burnup is 2.6 GWD/MTU. Portions of the Technical Data Book have been provided.

Show all calculations, reference any figures used, and state all assumptions.

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

L

PAGE 3

Iz__EBINQlELE5_DE_NUgkEBB_EQWEB_EL9NI_DEEB911QB2 IBEBDDDYBedIGE,_HE91_IB9NEEEB_9Np_ELylp_ELQW QUESTION 1.04 (1.50)

c. For an operator taking data for a 1/M plot, how will the value of Keff affect the time elapsed before a stable count _ rate can be obtained after withdrawing rods ?

b. How will the initial count rate affect the count rate at criticality?

DUESTIDN 1.05

( 1.00)

With all systems in manual and no operator action, what effect (increase, decrease, or no change) will decreasing the circulation water temperature have on the following?

a.

Condenser vacuum l

b.

Condensate Pump NPSH 1

QUESTION 1.06 (2.00) 1

c. List two plant evolutions which could result in a waterhammer.

(0.8) b.

Give two examples of how waterhammer can be minimized.

(1.2)

QUESTION 1.07 (1.50 )

Indicate whether each of the f ollowing will INCREASE, DECREASE, or REMAIN UNCHANGED as the discharge valve of a running, motor operated, centrifugal pump is throttled (valve moved in the shut direction):

a.

Pump motor amps i

f b.

Pump discharge flow

.c.

Pump discharge pressure

)

l i

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

t

PAGE 4

Iz__EBINCIELE5_QE_MQQLE68_EQWEB_EL9NI_QEEB9IlpN2

'IVEB599YN951QS,_UE91_IBBNSEEB_6NQ_ELQ1p_ELQW QUESTION 1.08 (2.40)

Why does nucleate boiling heat transfer remove more heat than o.

non-boiling heat transfer?

b. Why does film boiling heat transf er remove less heat than non-boiling heat transfer?

QUESTION 1.09 (2.50) c.

In Figure 5.2 in the Technical Data Book, explain why rod worth at 80 inches increases over core life.

(1.0)

b. Explain qualitatively what effect a dropped rod would have on reactor power and Tave if the plant was initially at 100% power with all rods withdrawn.

Assume all systems in manual and no reactor (1.5)

~

trip occurs.

DUESTION 1.10 (2.50)

a. The overall power coefficient is a combination of three coefficients.

List those coef ficients in order from the largest to the smallest contributor to the overall power coefficient.

(1.5) b.-What are two reasons for the Total Power Defect increasing with (1.0) power?

DUESTION 1.11 (2.00)

If the reactor is critical at 10 E -4 % power and rod group 4 is at 80 inches at EOC, how far should group 4 be withdrawn to establish a O.3 DPM startup rate?

Show all calculations, reference any figures, and state any assumptions.

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

w

PAGE 5

lt__EBINclegg5_DE_N99LE98_EDNEB_EL9NI_DEEBBIIQN, INEB5DDYN951R52_NESI_IB9NSEEB_9ND_ELUID_ELDN DUESTION 1.12 (3.00)

What effect, if any, will each of the following plant upsets have on subcooling margin?

Assume all systems are in manual.

Justify your answers and consider each upset separately.

a. A pressurizer PORV begins leaking while at 80% power.

b.

One PCP is secured f or physics testing while operating at 10% power.

c. Steam Generator level is being controlled 4% above setpoint while at 30% power.

QUESTION 1.13 (3.00)

Determine and explain what effect, if any, each of the following changes in plant parameters will have on shutdown margin.

Consider each parameter separately.

During a reactor startup the levels in both steam generators are a.

rapidly increased 5%.

b. After raising power from 50% to 100% conditions are stabilized and no further operator actions are taken for six hours.

c. Rods are borated to all out position from some intermediate position.

I i

i

(***** END OF CATEGORY 01

• • • • • )

t

PAGE 6

'2i__EL6NI_p5 Sign _INCLUplNg_g8EEIy_8ND_EDEBQgNCy_gySIEdg QUESTION 2.01

(.60)

What is the minimum number of gross radioactivity monitors required for a liquid batch release through a three inch discharge line?

Choose the most correct answer.

A. None B. One C.

Two D.

Three OUESTION 2.02 (2.00)

Indicate whether the f ollowing service water loads are supplied by the non-critical header, critical header A, critical header B, or critical header A and B.

~

a." Diesel generator K-6A Jacket water cooler b.

Component cooling water heat exchanger E-54B

c. Engineered safeguards pumps seal cooling

d. Auxiliary building air conditioning condensers e.

Containment air coolers QUESTION 2.03 (1.50)

a. What design feature of the reactor vessel prevents complete draining of the vessel?

b. What is the design purpose of having the reactor vessel supported by steel pads resting on sliding plates?

l DUESTION 2.04 (2.10)

a. What would be the probable cause of the pressure instrument (PT 0338) downstream of the 12 inch check valve to the Safety Injection Tank (SIT) reading 2000 psig?

See attached figure.

b. How would the pressure be reduced?

c. What is the source and motive force used to fill an SIT?

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

L

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2t__P(@NI_DE@lGN_lNCLUQ1NG_@@EEIy_6ND_EdEBGENCY_SYSIE55 PAGE 7

QUESTION 2.05

( 2. 51 )

0 For each of the following valves indicate whether the v41ve will fail open, fail shut, f ail as is, or will not be affected by a loss of instrument air.

a. Pressurizer spray valve

b. Condensate sump recirculation valve (CV-0730A) c.

Main feedwater bypass valve (CV-0734)

d. Main steam isolation valve e.

Atmospheric steam dump QUESTION 2.06

( 2.00 )

A double ended load center is a 480 volt bus which can be supplied by two different power sources.

a.

How is paralleling of double ended load centers prevented?

( 0.8 )

b.

What are three reasons for not paralleling double ended load centers?

(1.2)

QUESTION 2.07 (2.00)

Describe the two paths by which decay heat would be removed immediately following a post LOCA Recirculation Actuation Signal.

Start at the fuel cladding and continue to the ultimate heat sink.

Include major components or systems.

QUESTION 2.08 (2.50)

c. What two design features would prevent control rod damage during a rod drop?

( 1. :2) b.

Would any control rods drop if power was lost to MCC1?

Assume the selector switch in the spreading room was selected to MCC1.

Justify (1.3) your answer.

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

• • • • • )

u

PAGE E

,2t__ELON1_QE@l@N_lNQ(UDINQ_S@ EELY _@ND_EMEBGENQY_EYSIEd@

QUESTION 2.09 (2.40)

I 1

What is the reason for the limitation allowing operation with one Primary Coolant Puinp (PCP) seal f ailed but requiring immediate shutdown if more than one PCP seal fails?

See item (i) on the precautions and limitations provided.

QUESTION 2.10 (2.40)

State the components which are routinely used to:

Maintain the clarity and chemistry of the spent fuel pool.

a.

b.

Reduce the temperature of the spent fuel pool.

c. Raise level in the spent fuel pool.

QUESTION 2.11 (2.00)

What effect, if any, does a turbine trip signal (45 psig auto oil pressure) have on:

a. Reactor Protection System b.

Feedwater regulating valves c.

Cooling tower fans d.

Emergency diesel generator QUESTION 2.12 (3.00) should be observed for each of the following What response, if any, components af ter a Saf ety Injection Actuation Signal?

a. Service water outlet valve (VHX-4)

b. Containment air coolers (V-1B,2B,3B, and 43)

c. Component Cooling Water (CCW) pumps (P-524, B, ind C)

d. CCW to spent fuel pool and evaporators (CV-0944H)

e. VCT outlet valve (MD-2OB7)

f. Gravity feed valves (MO-2169 and 2170)

g. CCW to containment (CV-0910)

h. 1E bus feeder breaker (152-303)

1. Emergency diesel generators

j. Instrument air compressors

(***** END OF CATEGORY O2 *****)

PALISADES NUCLEAR PLANT b e d on 4.,g Ne SOP 1 SYSTEM OPERATING PROCEDURE Rsvisien 16 Page 3 of 44 TITLE:

PRIMARY COOLANT SYSTEM g.

During heat-up or cooldown of the primary coolant system, at least one primary coolant pump must be operation if shutdown cooling is not operating.

h.

Primary coolant pumps shall not be started if rolling in the reverse direction.

i. Failure of more than one of the three main pressure seals on a Primary Coolant Pump requires immediate shutdown and cooldown in accordance with GOP 8 and GOP 9.

Operation with one main seal and low pressure seal failed is possible if additional leakage is within limitation of the radwaste system.

Maximum expected leakage with all main seals failed should not exceed approximately 30 GPM.

(See ARP 5, " Primary Coolant Pump Seal Pressure Off Normal," for failed seal criteria.)

j. Primary coolant pump seals shall be vented prior to starting of the pump following any operation which could introduce air into the system including " Steam Generator Air Sweeps."

k.

Primary coolant pump seals shall be provided with a source of cool, clean, borated water (flush flow) during any filling of the primary coolant system f rom a level below the upper seal or af ter reactor head replacement, during venting of the seals, during primary coolant pump run for " Steam Generator Air Sweeps" and during seal run-in operation of the primary coolant pump (s).

1.

Primary coolant pump seals shall be run in for a period of two hours subsequent to any repair or replacement of the seals, or if the primary coolant level has been below the elevation of the upper seal for a week or more.

If pressurizer sprays are operated when differential temperature m.

between spray water and pressurizer water is greater than 200*F, log time, differential temperature and pressurizer pressure in Reactor Logbook.

A cumulative total of spray cycles with greater than 200*F differential temperature will be kept inside the cover of each Reactor Logbook. The abnormal operating limit is 350'F, J

but can be exceeded in an emergency situation.

A reactor coolant pump shall not be started with one or more of the n.

PCS cold leg temperatures less than or equal to 250*F unless; 1) i the pressurizer water volume is less than 700 cubic feet; or 2) the secondary water temperature of each steam generator is less than i

70*F above each of the PCS cold leg temperatures.

(Reference c) l I

No more than two (2) primary coolant pumps shall be operated when o.

the primary coolant temperature is less than 250*F.

No more than three (3) primary coolant pumps shall be operated when p.

primary coolant temperature is less than 400 F.

l l

pa0679-0284a-89 e

PAGE 9

Its_JNSIBUDENIS_AND_C9 NIB 9bE QUESTION 3.01 (2.40)

What effect, if any, would a failure high of steam generator level instrument, LIA-0702A, have on:

(Figure PNFB-6 is provided)

Main feedwater regulating valves to steam generators E-50A and B.

a.

b. Bypass valves to steam generators E-50A and B.

c.

Main feedwater pumps to steam generators E-50A and B.

d.

Overall plant operations if no operator action is taken.

GUESTION 3.02 (2.40)

List.four control room indications, not including alarms, which provide control rod position indication, and explain HOW the indication signals are produced.

QUESTION 3.03 (2.50)

What five automatic actions would occur if the instrument AC bus was deenergized during power operations?

QUESTION 3.04 (1.40)

List FOUR automatic diesel generator trips and indicate whether each trip is overridden by a Safety Injection Actuation Signal.

DUESTION 3.05 (2.40) 1 Explain what automatic actuations, if any, are associated with each of the following radiation monitors.

Do not include alarm indications.

a.

Stean generator blowdown monitor (RE-0707) j b.

Stack monitors (RIA-2318) c.

Engineered safeguards room west (RE-1811)

)

i

d. Fuel handling building (RE-5712)

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

f

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PAGE 10 21__JNSIBydENIg_AND_CDMIBQL5 QUESTION 3.06 (2.40)

What signals, including power levels, type of channels, and number of channels, produce each of the following?

a. Bypass of the reactor trip on "High Startup Rate" (1.2)

, 0.6)

(

b.

Removal of voltage from the startup channels

c. Bypass of the reactor trip on " Loss of Load" (0.6)

QUESTION 3.07 (2.00)

If during reactor plant operations at 95% power a feedline rupture were to occur inside the containment, what are the FIVE Engineering Safety Features (ESFs) that could possibly be actuated and what signals will cause these actuations? Include setpoints and 1,ogic.

QUESTION 3.08 (3.00)

Why are the eight motor operated valves associated with the

a. auxiliary f eedwater system normally positioned to the open position rather than the auto position?

b. What actions, if any, should an operator take if one minute after the receipt of an Auxiliary Feedwater Actuation Signal the auxiliary feedwater pumps P-BA and P-8B have not started and P-8C is running?

Feed flow to SG E-50A is 100 gpm and to SG E-50B is 50 gpm.

Pressure in SG E-50A is 930 psia and in SG E-50B is 900 psia.

Justify your answer and assume all switches are in automatic.

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

PAGE 11 Ic__IN51BydEN15_@ND_CQNIBQLS DUESTION 3.09 (3.00)

For each of the following pairs of simultaneous instrumentation failures indicate whether the reactor should or should not have tripped.

JUSTIFY your answer and consider each pair of failures separately.

Figure PNNB-7 is supplied.

a. Th indication to channel A fails to 600 F and RCS pressure indication to channel B fails to 1700 psia.

b. SG A level indication to channel A fails to 20% and SG B level indication to channel A fails to 20%.

c. PCS flow indication to channel C fails to 0% and loop 2A differenbial pressure indication to channel D f ails low.

d. Power supply to channel A drawer fails and the upper detector on nuclear instrument power range safety channel B fails high.

DUESTION 3.10 (3.50)

What indications must be checked to verify compliance with each of the following Safety Injection Actuation Signal reset criteria.

Include the number of channels that must be checked.

a. PCS is greater or equal to 25 F subcooled with Primary Coolant Pumps (PCP) running.

(0.9) 1

b. PCS is greater than or equal to 25 F subcooled with PCP's not (0.8) running.

c.

Pressurizer l evel is greater than 20% and constant or increasing.

i (0.6) d.

At least one steam generator is available f or removing heat.

(1.2) l I

i l

l l

(***** END OF CATEGORY 03

• • • • • )

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4 __PBQCEgyBES_;_NgBb963_9pNQBMBL3_gBgBQENGy_9ND PAGE 12 t

BSD196991996_G9NIBQL QUESTION 4.01 (2.00)

a. What is a licensed operator 's responsiblility concerning the RPS and RPCIC panels when there is fuel in the reactor?

b.

Under what circumstances, if any, can an operator depart from an operating procedure?

c.

What actions should an operator take if, while perf orming a checklist, he finds a valve in the Safety Injection System shut which should be open?

d.

Who is responsible f or maintaining the fuel status board?

QUESTION 4.02

( 1.50 )

Answer the f ollowing questions concerning nonmal reactor startups.

a.

What is the MINIMUM temperature for criticality?

e b.

What is the MAXIMUM startup rate permitted under normal conditions?

What is the MINIMUM number of PCPs required to be running prior to c.

startup utilizing control rods?

QUESTION 4.03

( 1.00 )

From the Emergency Operating Procedure (EOP) Index provided, list the EDP which should be used to:

a. Determine power supplies and room locations f or motor operated valves.

b.

Determine the reset criteria for SIAS and SIS equipment.

t

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

h.__P8QQEQUBE5_,_MQBd6(a_6SNQ856(,_[UEB@ENQY_6NQ PAGE 17

'6001DLQQ1G96_GQNIBQL QUESTION 4.04 (2.50)

Explain how an approach to criticality should be conducted with only one operable startup detector.

Limit your explaination to the order cnd levels of positive reactivity additions made by dilution or rod withdrawal.. State your initial assumptions concerning boron i

concentration and rod position.

Actual values are not required.

QUESTION 4.05 (2.50)

a. What actions, if any, are required if there is indication that FOUR full length control rods have not inserted following a reactor trip?

(1.0)

I

b. Why are two Primary Coolant Pumps (PCP) tripped when PCS pressure drops below 1300 psia and all PCP's are tripped at a lower PCS i

pressure following initiation of Safety Injection after a reactor (1.5) trip?

QUESTION 4.06

( 3.0- )

0

c. Explain why each of the f ollowing procedural steps is an indication of natural circulation.

1.

Delta T is less than 50 F

2. SG 1evel > -84% wide range 3.

Loop Tc constant or decreasing (1.5) b.

During depressurization while on Natural Circulation PCS pressure is 650 psia, core exit thermo-couples are at 500 F, Th is 500 F and Tc is 480 F.

Has a void formed in the PCS?

A page of ONP-21

( 0.5) is provided.

c. What are the maximum and minimum PCS pressures allowed in the PCS if PCS temperature is 290 F and f orced circulation has just been reestablished by starting a PCP?

SDP 1 figure 1-2a is provided.

( 1.0 )

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

PALISA1st'S NUCLEAR PLAfff EEEEhCY OPERATING PNOCEDURES INDEX 81ENhlAL o

REVISI0ff REVIEW IRDBER PERPORDED DOCl2 eff SPONSOk f

PROCEDURE TITLE PROCEDURE NUPSER n

16 86/03/17 RJFriso REACIOR TRIP EOP 1 as 14 84/08/23 ILIFriso LOSS OF AC POWER 5

EOP 2 1

86/01/24 RJFriso IDSS OF SERVIG WATER EOP 3 2

83/12/05 RJFrigo LOSS OF CDl900 err CDOLINC l

EOP 4 13 83/12/05 RJFriso IDSS OF INSTRIDeft AIR EOP 5 16 86/03/17 ILIFriso MAIN STEAM LI E BREAK / MAIN FEEDWATER LINE BREAK INSIDE EOP 6 CDerIAllteft 14 84/03/17 RJFrigo MAIN STEAM LINE BREAK /NAIN FEEDWA1ER LINE hREAK OUTSIDE E0P 7 000fIAlleeft 18 86/03/17 RJFriso LOSS OF CODIANI ACCIDEffT EOP 8.1 15 86/03/17 RJFriso S1TAH UENERATOR TU8E RUF2URE EOP 8.2 12 86/01/24 RJFriso RJEL HANDl.ING ACCIDENT EuP 9 Canceled (DNTRCL ROOM EVACUATION E0P 10 1

85/04/25 RJFriso FIRE WilOf THitEATEMS SAFETY RELATED EQUIPteff EOP 10.1 15 85/05/28 RJFrigo ALTERNATE SAFE SHUTDOWI PROCEDURE EOP 10.2 l

13 83/12/05 RJFrigo LOSS OF CDef!AllGENT INTEGRITY EOP 11 12 86/01/24 RJFriso ABNORMAL RELEASE OF RADIOACTIVITY EOP 12 l

l l

p*Se 1' 1NM:X-19101 5/21/86

PAGE 14 4 __PBQCgpuBES_ _NDBd662_OpdQBd@L2_EMEBGENCy_9ND 2

'B9D196DD1996_GDNIBg6

" QUESTION 4.07 (2.50)

During a serious emergency, operators may be called upon to assist in search and rescue or recovery operations in the plant.

a.

In such cases, what dose could an operator receive 1)

To bring an injured worker to safety?

(0.5) 2)

To eliminate the further escape of radioactive effluents ?

(0.5) b.

What are the possible effects of receiving radiation exposures of the levels of 50 rem ?

Include short and long term effects.

(1.0)

Who must authorize this voluntary radiation exposure up to the c.

(0.5) emergency limits ?

QUESTION 4.08 (3.00)

What immediate actions should be taken if the Shift Supervisor determines that the plant cannot be maintained in hot standby due to a fire in the cable spreading room?

QUESTION 4.09 (3.00)

Explain the basis or reasons for the following subsequent actions in the Steam Generator Tube Rupture procedure, EOP-8.2.

Establish and maintain PCS temperature at 525 F to 530 F a.

(Listed in order of preferrence).

b.

Reduce and maintain PCS pressure as low as possible in the operating region.

Maintain PCS pressure approximately 0 - 100 psi higher than the c.

isolated Steam Generator.

d.

If affected Steam Generator level approaches full scale, drain to the Miscellaneous Waste System.

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

PA!,ISADES WUCt.EAP PLA'IT Proc No ONP 21 I

OFF-Pf0RMAL PROCEDURE Rsvisicn 12 P:33 7 cf 9 l

TITLE: NATURAL CIRCULATION Question 4.06 1

4.8.9-During PCS depressuration, monitor for void formation.

If substantial void formations are indicated by one or more of the following, go to Section 4.9 and 4.10.

l During steady state conditions, charging a known voluene of a.

l water / boric acid into the PCS does not result in a corresponding increase in pressurizer level.

b.

Pressurizer level increases significantly greater than expected while operating auxiliary spray.

If pressurizer level control system is in automatic, an c.

unanticipated letdown flow greater then charging flow.

NOTE:

During periods of abnormal pressurizer level behavior, charging and letdown _centrols should be placed in manual since automatic actions may be opposite to those which should be teken.

d.

Qualified core exit thermocouples read higher than saturation conditions or are erratic.

l Increased core AT abcve full power (50*F).

e.

f.

Startupneutrondeteck:orsshowingerraticindication.

Loop T s increasing or erratic.

h g.

h.

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1

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PAGE 15 dz__EBOGEggggS_ _NQBb8L,_9pNQBbg63_gbEggENgy_8ND BBD196991Q86_GQNIBQL QUESTION 4.10 (4.00)

For each of the f ollowing situations indicate whether the. operator should continue operations, reduce power, shutdown the reactor, or trip the reactor.

Consider each situation seperately, Reactor power is 70% and one of the two condensate pumps' trip.

a.

b. Reactor power is 20% and condenser vacuum has dropped to 17" of Hg.

c.

Reactor power is 50% and a steam generator tube leak is identified.

d.

Reactor power is 80% and component cooling water has been lost for two minutes.

There are no temperature alarms.

e.

Reactor power is 80% and non-critical service water has been lost for two minutes.

There are no temperature alarms.

f.

Reactor power is 100% and instrument air header pressure has decreased from 90 to 80 psig over the last 10 minutes and is continuing to decrease.

g. Reactor power is 75%, total leakage is 3 gpm and a CRDM seal has been confirmed to be leaking.

h.' Reactor power is 100% and a fuel element failure has been identi fi ed by coolant activity as being above Technical Specification limits.

i. Reactor power is 90% and a single rod drops into the core.

J. Reactor power is 100% and a major fire is reported in the HPSI pump room.

(*****

END OF CATEGORY 04

• • • • • )

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

~

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t

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( 153.4 0 016C70 1965 7 IMS7 14 087 10676 5021 6 0 0262 2 1111 2 1393 at 48 8 0165 e4 0 C3C21 1830 0 183e 0 16 051 leb64 80125 0 L321 2 1006 21327 as e De t 0 1779!

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PALISADES TECHN2 CAL DATA BOOK 4/2/85 LIST OF EFFECTIVE PAGES 1 Figure Title g 6 I, l 1l I 1 i 2.0, Xenon Worth 2.1 Equalibrium Zenon vs. Reactor Power 1 2.2 Xenon Worth vs. Time. Trips fron'Various Powers 1 2.3 Xer.on Worth vs. Time, Step Decreases from 100% Power 1 2.4 Xenon Worth vs. Time, Step Increases to 100% 1 from various Power Levels ~ .41 0, Power Defect 3.1 Power Defect vs. Power 2 3.2 100% Power Defect vs. Burnup 2 ~ 3.3 Programmed PCS Temperature vs. Reactor Power 1 -,.-e .,m- .m ---_,--,-,.-g --.-,-.,,,.,,.,.n..-., ,,,,--,,w,__,,,--,.,,., --g_,,-,- ,,+,-, 7v

1 PALISADES TECHNICAL DATA BOOK 10/26/84 LIST OF EFFECTIVE PAGES 3 Fipure Title Rev. 4.0, Reciprocal Boron Worth 4.1 HZP Reciprocal Boron Worth vs. Burnup 1 5.0, control Rod Worth 5.1 Integral Rod Worth 1 5.2 Group 4 Worth at 80 Inches vs. Burnup 1 6.0, Boron Rundown 6.1 Predicted Boron Concentration vs. Burnup 1 7.0, Boration I 7.1 Boric Acid Batching Graph 0 7.2 Boraticn Rate Addition 0 7.3 Boration Volume Addition 0 8.0, Dilution 8.1 Gallons PMW per ppm Boron 0 i 8.2 Formula Sheet (Pages 1 & 2) 0 8.3 Boron Dilution Rate (120*F) 0 8.4 Baron Dilution Rate (532*F) 0 8.5 Dilution Make-up Volume (120*F) 0 8.6 Dilution Make-up Volume (532*F) 0 l l ( EP0584-0189A-TC03-DM01 l

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. i. :. = =.. I_ .\\.. .s .i,! 1,. l. l, g _....L_._.. r a _.9 C {... p_._. _. _ o y y ._. v:_ _... _ [.._........ q g i g,,, _4_ \\_ ._N. g,e. _- j .. I. _. .'O _ _.. l. _.. _ _ g \\. l p l . _ _ ___..I...._ I. S i s y ll -=

= =M N

y...

.,.. - N.. _.A i i N _.__i N.: 1 N i-. =i :- :t =. = .. W. CN " =M - =:\\ l - l-K I -N -N

• - t E:

.E= Hd yi c_wN-

= e#. \\i; 1'NJ d N S K :-- 4-3.. '.::t Mil z:(@

+ f. 3 ?:- =- y h* ~ Q. " lY pk r$. __ 1=f- -5 _7,55.ML%.f E E55:.1 4 W-N _ _;p__u _-A.. _ _.. _ ._n : p__. 2.,7.... Q...; _._.,;.._ __ ... l... ..= _ - - _ w g. .i _. e.._.,,.I.... g, i. _. _..;. :. =

==_. l _= ___ p. _.,, ,p- -.ln .. l a. t l :..1..- - i = [ E: ?) 5: L=;- Eli= ;-Q: 55$ _ _- Er 9 M"-l 3:-l: k i N5,, I

• I.

I .. N - .~_ N ".* _. l __ _.. L _. .__ \\. _ _ _ _ ...4 i... ._.._. (.,. _..._._~ N A _e-. _A

a.. _

..fr \\ 7.. _.. s._ _ .. l g_ m. . v._ 1 {.. A. U.~t...i.. i_. .i _ _i. . I.. 4 E O b 8 T SS b N O T IFI N [. ! # -liJ f 1 bdG M .9 DN 8 f 8

i, ,.. uczen PAGE la ic_ E61NCIELEE DE WWCLEBB EQWEB EL6MI-QEEBBIlgNa 'e IUEBOODYN951gS,_MEDI_189MSEE8_6NQ ELU10_ELQW -86/08/12-DUDLEY, N. ANSWERS -- PALISADES ANSWER 1.01 (.50) B [O.53 REFERENCE PDBOOK5.28 KA: p. 3.1-17 K5.20 3.6 ANSWER 1.02 (.60) C. [0.63 REFERENCE PRADOK6.09 Appendix pg A-9 2.9 ANSWER 1.03 (2.50) CO.83 3.33 - 2.65 = 0.68 % p Figure 2.3 90 ppm / Xp X O.68 %p = 61.2 ppm Figure 4.1 rev 1 [1.03 7,600 gal (+/- 300 gal) Figure 8.6 rev 0 [0.73 REFERENCE PDBOG28.02 OG35.01 OKOS.29 KA: p 3.1-45 EK1.02 3.6 ANSWER 1.04 (1.50)

a. The closer to criticality, (large Kef f) the longer time required to reach a stable count rate.

[0.753

b. A higher initial count rate will result in a higher count rate at criticality.

[0.753 REFERENCE 1 l n.

PAGC 17 h~_EBING1ELEE DE_NWGLE9B_EDHEB_EL9N1 DEEB611QL IHEBUDDYW951GEo UEBI IBBNSEEB 6HD ELVID EL9W -86/OB/12-DUDLEY, N. ANSWERS -- PALISADES KA: p 3.1-7 K5.OB 2.9 ANSWER 1.05 (1.00)

• a.

Increase [0.5 each3 \\ b.* Increase REFERENCE PRA04.50 KA: p 3.5-3 A1.05 3.2 p 3.2-2 K5.11 4.0 ANSWER 1.06 (2.00) Valve operation, opening or closing

a. Pump starting or stopping

[any 2 @ O.4 each] Oscillation of auto control valves Rcasonable ansvers vill be cecepted. control during velvc operationt b. Insure adequate pressure insulation and steam drains to prevent f ormation of use thermal condensate Slowly opening of valves between voided and f ull systems Proper venting of components Adequate level on tanks in systems where the tanks provide supply or surge function Proper use of steam traps and vents (Two required) (1.2) Proper sequencing of valves in pressurized systems Reascnable answers vill be accepted. REFERENCE PRA04.101 PRA04.102 ANSWER 1.07 (1.50) a. decrease b. decrease c. increase [0.5 each3 REFERENCE PRA04.91

PAGE 10 1_ EBING1ELEE-QE WGLE98_EQUEB ELON1-QEEB011QN3 IUEBUQDYN951 gen UE01 IB0NSEEB_eND ELU10_ELON ANSWERS -- PALISADES -86/OB/12-DUDLEY, N. s ANSWER 1.08 (2.40) Nucleate boiling creates turbulent flow which promotes mor'e mixing (1.2)c a. The coolant picks up latent heat of vaporization and carries OR it to cooler parts of the channel (1.2). b. In film boiling, a film of steam coats the clad surface and forms an insulating layer ( 1.2 ). REFERENCE PRADOK5.11 KA: p 3,4-23 EK1.03 4.5 ANSWER 1.09 (2.50) a. The fission product inventory increases at EOL. [0.3]so chc neutren Tissien preducts arc f ood therral neut ron abserbers, energy spectruc will shift towards the epitherts1 region. [0.4] Rods are good absorbers of epithermal neutrons. [0.3] OR Thermal neutron flux increases at EOL due to fuel burnup and boron removal. [0.5] Rod worth is directly proportional to thermal neutron flux. [0.5] [0.53 b. Reactor power will decrease and return to equilibrium power. Tave will decrease. [O.53 The negative reactivity inserted by the dropped rod would be (0.5) countered by positive reactivity inserted by MTC. REFERENCE PQDBOOK5.06 J =-------------------_ KA: p 3.1-2 K 5.05 3.5 K 5.10 3.9

PAGE 19 1s-EBINGLELEE_QE_NUGLE0B_EONEB_ELONI-QEEB8110Na INEBdQQYN9NIGS _NEGI_I69NSEEB 9ND_E(UlQ_ELQN i -86/OB/12-DUDLEY, N. ANSWERS -- PALISADES ANSWER 1.10 (2.50)

a. FTC [0.43 MTC [0.43 Void defect [O.43

[0.33 f or proper order Program Tave increases as power increases which adds negative b. reactivity through MTC [0.53 temperature increases as power increases which adds negative Fuel reactivity through FTC [O.53 REFERENCE PQCOG37.14 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - _ - = _ - _ _ KA: p 3.1-3 K 5.49 3.4 ANSWER 1.11 (2.00) . l pu [0.43 SUR = 26 (lam da) (p) / (B - p) [0.73 0.3 = 26 ( ) (p) / (0.0053 - p) [0.23 (26 X O.08 + 0.3) p = 0.3 X O.0053 [0.13 p = 0.00067 or 6.7 E -4 From figure 5.1 rev 1 curve 3; rod height = 94 in (+/- 2 in) CO.63 REFERENCE PODOG35.1

---

_-__=-

= =------------- _=. _ _--=__ -=----- KA: p 3.1-3 K5.47 2.9

PAGE 29 1 :. EBING1ELEE DE_NWGLEeB EQWE8_EleM1_QEEBBI1QL ISER5ppyN951GSg_ME91_IBONSEEB_6MQ_E(Q1Q_E(QW -86/OS/12-DUDLEY, N. ANSWERS -- PALISADES ANSWER 1.12 (3.00)

a. Subcooling margin would Secrease

[0.5) due to decreased PCS pressure. [0.53 cy[.< M g /ec.~wa - $ $+#<e

b. Subcooling margin would not change (0.5]

[0.53 since Tave would remain the same. ( Ita T across the SG would increase. ). ' & 22 y w /,an - w p -+f/dm <$ a a

c. Subcool ng margin wpuid not change (0.5]

[0.53 since heat transf er would not be af f ected. REFERENCE PRADOI:6.01 KA: p 3.4-22 EA2.01 4.6 ANSWER 1.13 (3.00) Shutdown margin decreases. [0.5) [0.53 PCS temperature decreases adding positive reactivity to the core. a. Shutdown margin remains the same. [0.5] Reactivity effects of xenon and temperature will compensate f or b. each other. [O.53 [0.5] Shutdown margin increases. [0.53 More rod worth will be available for insertion on a trip. c. REFERENCE PQFOOA1.03 KA: p 3.1-17 K5.19 3.5

PAGE 21

21. PLeNI-Q(ElGN ING(yQLNG SBEEIY_eNQ_EDE6QENQY EYSlgdg

-86/OB/12-DUDLEY, N. ANSWERS -- FALISADES ANSWER 2.01 (.60) C. Two [0.63 g, Ou REFERENCE PNPDDG5.03 KA: p 3.11-14 K6.10 2.5 ANSWER 2.02 (2.00) a.)(3

b. A

c. A and B d.

non-critical e. B [0.4 each] REFERENCE 1.01 PN A ADI: KA: p3.5-45 K1.05 3.8 K1.01 3.4 K1.16 3.6 ANSWER 2.03 (1.50)

a. Elevation of the lowest penetration (hot leg) is (6 ft) above the top of the f uel assemblies. [O.753

[O.753

b. Allows f or thermal expansion of the reactor vessel.

REFERENCE PNJBDK4.03 Q1 K4.02 Q1 KA: P 3.2-1 K4.01 2.7 e

i PAGE 22 2 &,_P($N1_ DES IGN _ LNGLUQlWQ _@@EEIY_6NQ _ENER@gNQY _EYSIENS c -86/09/12-DUOLEY, N. ANGWCR5 -- PALISADES ANSWER 2.04 (2.10) the primary check valve. [0.7J

c. Leak by of Bleed pressure through CV-3038 to radioactive waste treatment system.

[0.73 b.

c. SIRWT [O.43 by using HPSI pump [O.33 CS ev Co 3]

n v v REFERENCE PNMAOK1.02 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ = _ KA: p 3.2-9 K1.06 2.2 3.2-12 K2.03 3.3 ANSWER 2.05 (2.50) a, closed b. closed

c. open d.

as is e. cl osed [O.5 each] REFERENCE EOP-5 Rev 13, p 2, 3 _ = _ _ = _ _ _- .-= _____-__----- KA: p 3.8-10 EA2.OB 2.9 ANSWER 2.06 (2.00)

a. Prevented by mechanical interlocks. [0.83

b. Cross connect a safeguard bus with a non-safeguards bus Elimination of saf eguard bus redundancy Cross connecting out of phase

[any 3 e 0.4 each3 Exceeding, bus ratings <~ f Mle bak fdfi-M MM' I#, REFERENCE y l PNIBG37.01 _-_____________ _=_ KA: p 3.7-3 A2.06 3.4 A2.15 2.8

J PAGE 23 Zu-ELON1_QEgigN_lyg(yQLNG_$@ EELY _6NQ_EUEBGENgy_EY@ LEU @ -86/OS/!2-DUDLEY, N. ANSWERS -- FALISADES K4.05 2.7 ANSWER 2.07 (2.00) Dut the break to the sump [0.13 Out break to the sump [0.13 LI5hT+w=qanew Minretir pumps [0.13 Through CS system [O.23 CCW eT CNTH Air Coolers [0.43 SDC HX [O.33 SW [0.23 CCW [O.23 -Ceclin; trarre [0.13 SW [0.23 4kACE r~ cling tc= crc [O.13 2#-4. REFERENCE PNMBOG4.05 PNMCOK1.01 KA: p 3.2-15 K4.03 3.4 ANSWER 2.08 (2.50)

a. Damping of buffer postion on a wet scram [O.63 Energy a b s or ber prevente damage on a dry scram [0.63

[O.33 No rod drop. b. provides power to rod drive motor and brake which is engaged MCC1 when deenergized. [O.63 [O.43 The clutch which provides union is not effected by loss of MCC1. REFERENCE PNLAOK4.01 K2.01 G4.03 _= _. - = - -. KA: p 3.1-2 K4.07 3.7 3.1-1 K2.03 2.7

e PAGE 24 2.c__ELeNI_DEgigN_LNQwg1NQ_geEEI!_eNQ_EDESQENQY_stglEdg -86/OB/12-DUDLEY, N. ANSWERS -- PALISADES ANSWER 2.09 (2.40) Pumps can be operated with one failed seal because the other seals will prevent gross leakage. [1.2] If more than one seal has f ailed,JSir e R -;r ' - '

_A bleesi of f fiv.

-y 7. ^ 13 _=- @ additional rapid seal failure resulting in a LOCA. [1.2] /EL$thYf -ft n os 'm* J REFERENCE PNJDOG7.01 KA: p 3.4-2 K6.02 2.7 K4.07 3.2 ANSWER 2.10 (2.40) a. Spent fuel pool filter [0.23 Spent fuel pool demineralizer [0.33 SI:2 n.rier LO.33

b. Spent fuel pool heat exchanger [0.4 3 P-51 A or B [0.4J

c. SIRWT CO.~4 3 trkj EN[kj P-51B

[0.43 S I f[ y f REFERENCE PNOADG4.02 G9.01 SDP-27, p3 . _ _ = _ KA: p 3.11-5 K4.02 2.5 K3.03 3.0 K4.01 2.9

PAGE 25 2,.__ELON1_QE@lGN _lNGLypihg _E@Egly_6ND _EMEB@ENQY _EYSIED E ANSWERS -- PALISADES -86/OS/12-DUDLEY, N. ANSWER 2.11 (2.00)

a. Loss of load, signal if greater than 15'/. power b. 4desser 0-4 M c.

Trips cooling tower fans d. Start signal for diesel generator [0.5 each3 REFERENCE PNGAOK3.01 KA$ 3 5-11 A3.bk 3.4 ANSWER 2.12 (3.00) a. open b. ' cta -t ir alun => p c d #2$. [ wM[m] c. 2-C' htert-t/A 3 M

d. w a0s w J-A e.

f. open

g. open

h. open

i. St:-t but-da nat !cc.d c.,r. Bus Ar

[0.3 each]

j. Not effected REFERENCE EDP-1, Att.

3, rev. 16; p 3, 4 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ = - - - KA: p 3.2-27 A3,02 4.1 1 l

PAGE 26 ac-_INSIBydENIg_@NQ_GONIBQLS -86/08/12-DUDLEY, N. ANbwERG - FALISADES ANSWER 3.01 (2.40) MdLig1S 4.of W valves shtsts CO.63 a.

b. No effect [0.63 c.

No effect [0.63 [0.63

d. Rx trips on low steam generator water level REFERENCE PNFBOK6.O2

- _ - - _ - = _ KA: p 3.5-36 K4.05 2.5 p 3.5-56 EA2.05 3.5 ANSWER 3.02 (2.40) Position Indication Primary CO.33 synchro-transmitters [O.33 Secondary Position Indication [O.33 reed switches [O.33 Nixie tubes [0.33 synchro-transmitters [0.33 et f1f CC,3] Matrir board display [0.33 limit switches CO.33 o sg REFERENCE PNLADA4.04 A4.03 A4.02 A4.01 $ g* gQ KA: 3.1-31 K4.03 3.2 y '# j"] 9 K4.06 3.4 b Y. 5% t Pf /N+ Est K1.02 3.2 PIP && g ANSWER 3.03 (2.50) g O[kM M Charging pump suction shifts from VCT to SIRWT g[7g [f All charging pumps start Intermediate and letdown bypass valves close Feedpumps go to minimum speed gM +/ Backup pressurizer heaters energize 4, 4 + d;A Jtg* 6.+ [O.5 each] BA5T#eahd J REFERENCE A F u/ Lwy PNICOA2.01

---

g33g--

f KA: p 3.9-1 K1.01 3.4

PAGE 27 h__INSIByt!ENI5_8ND_QDNIB%S ANSWERS -- PALISADES -86/OB/12-DUDLEY, N. ANSWER 3.04 (1.40) Low lube oil pressure (40 psig) Overcrank ( 35 sec jacket water pressure < 10 psig, speed < 120 rpm) Overspeed ( 990 - 1G35 rpm). CO.3 each3 Generator differential relay Not overridden by SIAS signal [0.23 REFERENCE PNHBOK4.04 KA: p 3.7-11 K4.04 3.9 ANSWER 3.05 (2.40) Isolates blowdown tank (5 valves) [0.63 a. b. None [O.63 Shuts ventilation dampers [0.63 3 c. d. Trips fuel area supply fan (V-69) [O.E3 E'----- -___t:7 u--_., y. (V-70A or B) [O.E3 Stops exhaust fan non-selected for standby 3 REFERENCE PNPAOK1.O1 KA: p 3.9-23 K4.01 4.0 ANSWER 3.06 (2.40) a. Bel ow 10E -4% power [O.3 o wide range channels [O.33 Above 15% power [0.33 on 4 power range safety channels [0.33 LJ

b. Above 10E-5% power [0.33 on f2 wide range channels [O.33 ad Below 15% power [O.33 on14 power range saf ety channels [O.33 c.

REFERENCE PNNAOG9.01 K1.06 K4.03 KA: p 3.9-5 K4.01 3.1 K4.06 3.9

PAGE 28 ag__INSIBWENig_68Q_QQNISAg ( ANSWERS -- PALTS4 DES -86/08/12-DUDLEY, N. ANSWER 3.07 (2.00) SIAS [0.33 CIS [O.33 CSAS [0.33 (CNTM Air Cooler DNBR' mode) all ori HI Containment press. [0.23 of 3.7 - 4.4 psig CO.13 on 2/4 logic [O.13 Steam Line Isolation [0.33 on t_ow g press. [0.23 F/O of.306 psia [O.13 on 2/4 logic [O.13 REFERENCE AFAS PNMA4.19 cR// yf p PNMB4.09

pfy, PNMFA2.01 br TS Table 3.16.1 ANSWER 3.08 (3.00)

F0(.2Cr k.wx1&.-m;M s u~)5usb w ;~h' 54,_ U 'M On a massive SG tube f ailure the combination 6cf ramping turbine

a. load and filling the SG could result in isolation of the incorrect SG. [1.53 OR Attempt to start P-8A.

[0.5] Take no action.

b. The steam-driven AFW pump, P-8B, will not automatically start until 80 seconds after the AFAS.

[1.0) REFERENCE PNEBG12.02 PNEBOK4.05 - - - - = _ - - - - - - - - = _ - - = - - - - - _ KA: p 3.5-42 K4.02 4.5 K5.01 3.6

PAGE 29 L.__INSIBWt!ENIE_8NQ_GQNIBQLS -86/OS/17-DUDLEY, N. ANSWERS -- PALISADES ANSWER 3.09 (3.00) w rctit/b L 0* "' tuh4 a. Trip [O.250 TM/LP set point raises to 2300 pria on channel A Pressure is below TM/LP floor on channel B. [O.53 [0.53 No trip [0.253 Both trips on same channel. b. [O.253 Flow indication does no provide trip signal.[0.53 No trip d. Trip [O.253 Trip signal produced when trip drawer is deenergized. Second trip from high power. [O.53 s g i[ay-M [ L< w: 4 4 4_ g_f v ^^' a w "2~j dNed L muu RE ERENCE j _f_a %g<d.. __t e _P_N_N B G 27_._01__ __ _ _ _ KA: p 3.9-1 K4.02 3.9 ANSWER 3.10 (3.50)

a. Hot leg temperature on 4 channels

[0.33 c-t' f ce# Cold leg temperature on 4 channels CO.33 PCS pressure on 4 channels [O.33 b. Qualified core exit thermocouples, 4 required [0.43 PCS pressure on 4 channels CO.43

c. PZR level

[0.63

d. Steam flow [on two channels) [0.43 Main feed oY Aux. f eed flow on two channels

[O.43 Steam dump or bypass position on two channels [O.43 REFERENCE EOP 1, ATT. 3, rev. 16, p 1 KA: p 3.4-24 EA2.01 4.6 EA2.05 3.4 EA2.OB 3.8

PAGE 30 L.- EBQGEDWEE_: NQBueLa_8ESQBdeLi EMEgggggy_eMQ 8991060 GIG 66_CONIBDL ANSWERS -- PALISADES -86/OB/12-DUDLEY, N. ANSWER 4.01 (2.00) Maintain panels in view at all time. [0.53 a. In an emergency situation needed to protect the public health

b. and safety. [O.33 when authorized by an SRO CO.23

[0.53 Notify SRO and request permission to open valve. c.

d. Control operator [0.53 REFERENCE PSOOG23.03 G23.06 G13.06 G26.05 KA:

System Wide Generics 23 2.8 22 4.3 13 3.7 ANSWER 4.02 (1.50) a. 525 F b. 1 DPM

c. 4 PCP

[0.50 each] REFERENCE GOP-3, Att 1, rev 5, p 1, 4 - - - - - - - - - _ - - _ - - = - - - - - _ - KA: p 2-1 Plant Wide Generics 12c 3.5

PAGE 31 At-EBQGEDWBES :_NQBdBLa BENQBdebo EUE6GENQY_@ND ~ BBQ1960 GIG 06 GQNIBQL -86/08/12-DUDLEY, N. ANSWERS -- PALISADES ANSWER 4.03 (1. 00)

a. EOP-10.1

b. EOP-1

[0.5 each3 REFERENCE EOP-10.1, Att 1, rev 2 EOP-1, Att 3, rev 16 KA: Plant Wide Generics 28 2.9 ANSWER 4.04 (2.50) Establi ' shutdown boron or critical boron concentrati on (plus 50) wh i t.. ever is greater. LO.63 Pull rods until Group 4 is at 130 inches. [0.63 Determine critical boron from 1/M plot (critical approach sheet) CO.7J [O.63 Dilute to criticality (stop 2 minutes prior to 1/M plot prediction) REFERENCE GOP-3, Att 1, rev 5; p 5, 6 =_ _ =_ KA: p 3.1-7 K5.06 3.3 KS.13 3.1 K5.16 2.9

PAGE 32 h_ EBQGEDWBEE_ _UQBUBLa_6ENQ856(g_(d[8Q[NGY_0Np BBDIOLOGICet_CQNIgpl -86/OB/12-DUDLEY, N. ANSWERS -- PALISADES ANSWER 4.05 (2.50) Depress backup manual trip button (RPS panel CD6) [O.53 Emergency borate 225 per rod (900 ppm) or cold shutdown [O.53 a. With LOCA in hot leg PCP's would cause deeper core uncovery [0.53 b. [O.53 Better heat removal under f orced circulation. Protect pumps f rom damage. [O.53 REFERENCE PTBHOK6.02 EDP-1, rev 16, p 1 DNP-7, rev 10, p1 - = - - _ - KA: p 3.1-50 Generic 11 4.5 p 3.3-8 EK3.23 4.2 ANSWER 4.06 (3.00) a.

1. Indicates flow has not stagnated. [0.53

[O.53 Indicates SG is available for heat removal. 2.

3. Indicates flow has not stagnated. CO.53 5/6c = w we.?wg/M in

b. YES [0.53

c. 250 to 375 psia [1.03 REFERENCE PTBNDA2.02 A2.07 G28.02 KA:

p 3.7-21 EK1.01 3.7 EK1.04 3.1 l

PAGE 33 4 __PBQGEQQ8ES_ _NQ85@(,_6BNQBd@(,_EUEBQENGY_6NQ 3 8901960 GIG 96_QQNIBQL i -86/08/12-DUDLEY, N. ANSWERS -- PALISADES ANSWER 4.07 (2.50) a. 1) 75 REM - 2) 25 REM [0.5 each3 Short tern. Increased likelihood of cancer, particularly leukemia. b. somatic effects include blood changes. [1.03 c. SED CO.53 REFERENCE Pr or.. No EI-2.1 p 2, 3 -= ----- __--_

---

_=.

-- --- ------==_ KA: 3.11-27 EK1.01 3.5 3.11-29 EK1.02 2.6 ANSWER 4.08 (3.00) Trip the reactor Verify turbine trip Trip main feedpump Trip all but one PCP in each loop Activate Site Emergency plan [O.6 each] REFERENCE PTBEG11.OS - -_== - - - - - - = - - - - - - - - - - - - - - - - - - - - - - - - - - -=-- KA: p 3.8-13 System Generic 11 4.5

l PAGE 34 la__EBQGEDUBEE_:_NQBuela_0BNQBd@(,_EUE8GENGL6NQ BODIOLQGige;_QONISOL ANSWERS -- PALISADES -86/08/12-DUDLEY, N. ANSWER 4.09 (3.00) 0l --rt a. Maintain - ^ --!i..; margin before initiating cooldown. [0.75] OR ) Prevent lifting secondary safety valve. [0.75]

b. Prevent lif ting secondary saf ety valve. [0.75] OR Reduce flow out tube leak.

[0.75]

c. Prevent PCS dilution. [O.753

d. Prevent release of contaimination to the environment. [0.75] OR Prevent overstressing Main Steam piping.

[0.75] REFERENCE EOP-8.2, rev 15, p 4 KA: p 3.3-17 EK3.05 3.7 ANSWER 4.10 (4.00)

a. Reduce power (below 60%)

b. Trip M ^^bIhc$ ^./ f) de /24 e c. Shutdown et d. Remain at power

e. Trip A

cp;bef a a-r-kNHcd t ifj Shutdowne m.2<-ra eb ' d #A: w e/e [J M j

f. Shutdown et g.

h. Shutdown

i. Reduce power

.j. Trip [0.4 each] REFERENCE EDP-3, rev 1, p 1 DNP-5, rev 12, p 1 EDP-4, rev 2, p 1 DNP-11, rev 10, p 1 EDP-5, rev 13, p 1 DNP-23.1, rev 12, p 2 EDP-10.1, rev 2, p 1 ONP-23.2, rev O, p 1 _ - = - , KA: Plant Wide Generic 10 4.1 -v- .. - - - +. - -, -,---y ,..y-a w w w

o ? U. S. NLCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _P@LIg@DEg REACTOR TYPE: _PWR-gE__________________ DATE ADMINISTERED: _g6f96/24________________ EXAMINER: _HIggINg3_R._ APPLICANT: INgIBUgIlgNg_Ig_8EELig8 nil U = ceparate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each qu:estion are indicated in parentheses after the question. The passing grcdo requires at least 70% in each category and a final grade of at 1 cast 80%. Examination papers will be picked up six (6) hours after -'ths cxamination starts. % OF FATEGORY % OF APPLICANT'S CATEGORY __ve6UE_ _Igl@L ___@ggBE___ _V@6UE__ ______________g@IEgDRY .2Es99__ _2Ez99 ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS .2E299__ _2E 99 ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 2E 99__ _2E 99 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 43Ez99__ _25z99 ________ 8. ADMINISTRATIVE PROCEDURES, I CONDITIONS, AND LIMITATIONS [99299__ 199t99 ________ TOTALS FINAL GRADE _________________% 11 work dorie on this examination is my own. I have neither ivcn nor received aid.

ES-201-2 NRC RULES AND GUIDELINES _FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties. 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating. 3. Use black ink or dark pencil only to facilitate legible reproductions. 4. Print your name in the blank provided on the cover sheet of the examination. 5. Fill in the date on the cover sheet of the examination (if necessary). 6. Use only the paper provided for answers. 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet. 8. Consecutively number each answer sheet, write "End of Category " as appropriate, start each category on a new page, write on only one side of the paper, and write "Last Page" on the last answer sheet. 9. Number each answer as to category and number, for example,1.4, 6.3.

10. Skip at least three lines between each answer.

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

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

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

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

15. P3rtial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

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

NRC Rules and Guidelines for 2 ES-201-2 License Examinations

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

18. When you complete your examination, you shall:

a. Assemble your examination as follows: (1) Examination questions on top. (2) Examination aids < figures, tables, etc. (3) Answer pages including figures which are a part of the answer. b. Turn in your copy of the examination and all pages used to answer --~ the examination questions. Turn in all scrap paper and the balance of the paper that you did c. not use for answering the questions. d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked. l T 2

5 __IMEg8f_gE_ NUCLE 68_EgWEB_EL@NI_QEEB911gN _ELUIDS _6ND PAGE 2 1 1 ISEBdQDyN8 digs DUESTION 5.01 (1.50)

o. TRUE or FALSE.

Excessive motor amperage is one indication that a ccntrifugal pump is at shutoff head. (.5) b. What are the amperage limitations for the motor-driven AFW pumps? (1.0) i DUESTION 5.02 (1.00) What will be the temperature of the PORV discharge line if a PORV opsne when there is a steam bubble in the pressurizer, quench tank prs = cure is 20 psig, and pressurizer pressure is: c. 1700 psig? (.5) b. 700 psig? (.5) -QUESTION 5.03 (1.50) State six indications of void formation. QUESTION 5.04 (1.00) Why does the reference transition nil ductility temperature of the roector vessel increase with reactor operation? QUESTION 5.05 (1.00) a. TRUE or FALSE. Assuming all other conditions are identical, the flow obtainable through two 2 inch diameter pipes will be greater than the flow obtainable through one 3 inch diameter pipe. (.5)

b. TRUE or FALSE.

Assuming all other conditions are identical, the flow through a hole with a dif f erential pressure of 200 psi will be more than twice the flow through the same hole if the differential pressure is only 100 psi. (.5) (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)

5. THEORY OF NUCLEAR POWER PLANT DPERATION,,FLUIps,_@Np PAGE 3 2 IHEBdppyN8dICS QUESTION 5.06 (1.00) What is meant by the term " minimum required NPSH"? I QUESTION 5.07 (1.00) l' If a step change increase in turbine load occurs, what is the initial roeponse of:

a. steam generator level?

(.5) b. pressurizer level? (.5) QUESTION 5.08 (1.00) For 3-pump operation the limiting condition is void f raction rather than DNBR. Why is excessive void fraction a threat to fuel cladding integrity? QUESTION 5.09 (1.00) Tha speed of a positive displacement pump must be increased in ordsr to increase the flow rate. Why is a higher pump head developed at the increased flow rate, even though the pump head at each point along the pump characteristic curve remains the same as pump speed is increased? Rafer to Figure FND-FF-44. t QUESTION 5.10 (2.00)

c. What type of startup requires that the shift supervisor use an inverse multiplication plot?

(.5) b. Assume that the initial equilibrium count rate during a startup is 8 counts per second. A certain amount of reactivity is then added, raising the equilibrium count rate to 14 counts per second. If this same amount of reactivity is added again, what will be the new equilibrium count rate? Show your work! (1.5) (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)

1 ~ N2=2Ng g w V2 = 2V; Ng N2 I .J N R VOLUMETRC FLOW RATE = 9 FIGURE FNO-FF-44: TYPICAL CHARACTERISTIC CURVES FOR POSITIVE DISPLACEMENT PUMP AT VARIOUS SPEEDS (REV. 4) [

3___IHgggy_gE_NyC6g83_EgWgB_E69NI_gEgggIlgN _E69]pS _8Ng PAGE 4 2 3 _TH_E_RM_O_D_YN_AM_IC_S QUESTION 5.11 (2.00)

c. The reactor shall not be made critical, even during low power physics toetc, if primary coolant temperature is less than _____.

(.5)

b. What is the basis for the minimum temperature for criticality of 525 F in c11 instances except low power physics tests?

(1.5) DUESTION 5.12 (3.00) Calculate how long it will take to raise power at the maximum permissible Otertup rate from the power level at which the reactor is considered critical for the purposes of adminiistrative centrol to the highest pSrmitted power level in the hot standby condition. Show your worh! QUESTION 5.13 (1.50) Explain why control rod worth increases from BOL to EOL. Refer to Figure 5.1. QUESTION 5.14 (2.00) Explain why the maximum xenon concentration after a trip varies almost linscrly with the power level just prior to the trip. Refer to Figure 2.2. QUESTION 5.15 (1.00) Explain why the 100% power defect increases with fuel burnup. Refer to Figure 3.2. QUESTION 5.16 (1.50) State the three purposes of the power dependent insertion limits. (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)

PALISADES TECHNICAL DATA BOOK FIGURE 5.1, REV 2 CYCLE 7 INTEGRAL ROD WORTH 3.2 l i i I, i c 3 i i i i i i.0-i i t . i i l i t, REVEWED BY q I u I _3V!.. O 2,. 3A/ir. REACTOR EMGddEER 08- } 4. /'t y Wpen.,/[ & .' l Q o REAc' tor ENCatEmc ~ i SUPERWTDtDENT 4 s 5: ) x i i i. o e t D e gs. c i i o5

i

+4 i p. J r a$. s s 4 .' i mo t

d..

y l k \\ s E 04-y 5 5 e7 .e .e f I*...... a a.d. g I 4 0 2 i l 1 l I . { a l l l 0.0- g..... ...g .....g.... ...g........ 4 'e''''4''''i l 60 80 100 120 140 GRd 0 20 40 GR3 80 100 120 132 ROD POSIT 10N - INCHES W l TrtDR AWN CONSUMERS POWER COWPANY REAC10R ENQlNEDtMC/ PAUSADES GCPaekeed REACTOR ENQNEER CROUP 02/27/86

l l PALISADES TECHNICAL DATA BOOK FIGURE 2.2, REV 2 CYCLE 7 XENON WORTH VS tlME TRIPS FROM VARIOUS DOWERS I i i l b.0-3 100R + l f i r 4.b-i - - - =-- - I l een ~ \\g i

REVEWED BY

,TL LA #4: 70p 8 REAc.Toftsac4NEER fy KA* % /t M : ~ f react' E. ENotNEERINGI Jb-o '~ Q-I SUPERNTENDENT . a l g g 3,a. 8 D z i O e Il ( t 5 + 2.5-l l i t 2.0-t .) au \\ L i.3 7 -~ o _..a u x g 0.5-s i ~%i 0.0 -................................................................................ 0 10 20 30 40 50 60 70 80 T4WE (HOURS) ( 0045UWERS POWER COMPANY REACTOR DeQlNEERMC/ P== GCPaekeed REacio,t ENQldst CROUP 02/27/86

PALISADES TECHNICAL DATA BOOK FIGURE 3.2, REV 3 CYCLE 7 100% POWER DEFECT VS BURNUP i.8 I i t I i h l t l t t i i i 1.6-l l t l i a 4 i ~ i I i i j j g .I ~~ pc i i E-. 1.4-j i g W I C i a: w I it 4 i e On. I i de e i C3 g 3 1.2- ~ f I 4 I ~ ~ REVEWED BY i.o ?Y /T :'l bbl. REACTOR MEER 8 f.r% A19 NMr .ucT0e tNcwEERao P SUPERWTENDENT t 0.8 - 0 2 4 6 8 10 12 BURNUP - GWD/MTU CONSUMERS POWER COMPANY REACTOR ENGLNEERMC/ PAuSADEs GCPaekeed O2/27/86 REACTOR ENGINEIR CROUP

E __IMEQBy_QE_Nyg(E@B_EQBEB_E(@NI_QEEB@llgN _E(Q1pg3_@NQ PAGE 5 2 ISEBdggyN@ dig @ QUESTION 5.17 (1.00) RI A 2326 is the normal range noble gas activity monitor. What is a " noble gno"7 OUESTION 5.18 (1.00) Th3 nitrogen-16 equilibrium radioactivity value in the primary coolant is 121 u c/ml. Explain how nitrogen-16 is formed. Equations may be used, but j 'cro not required. (***** END OF CATEGORY 05 *****) L

PAGE 6 6___P69NI_gygIEdg_DEgJgN _CQNIBg62_9ND_INgIBUdEUI@IlgN 3 QUESTION 6.01 (1.50)

a. What is the f ailed position of the CCW containment isolation valves

-(CV-0910, 0911 and 0940) if instrument air is lost? (.5)

b. What is the basis for having the CCW containment isolation valves fail to this position upon loss of instrument air?

(1.0) I DUESTION 6.02 (2.00) State all the interrelationships between the fire protection system and tha following systems:

a. osrvice water system.

-- b. epent fuel pool system. QUESTION 6.03 (1.00) Whct are the four automatic trips for the motor driven AFW pumps? QUESTION 6.04 (1.50) What are the three manual methods of tripping a main feedwater pump? QUESTION 6.05 (1.00) Explain how the f ollowing signals will affect the main feed water regulating valve and the main feed water regulating bypass valve. Consider each case separately.

c. Steam generator level of 90%.

b. Steam generator pressure of 460 psig.

QUESTION 6.06 (1.00) i l State the f our backup supply sources f or the main generator seal oil in i order cf decreasing pressure. (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

6:__P69NI_gyglEDg_ggglgN _CgNIBgb3_gNg_INgIBUDEUI@llgN PAGE 7 2 QUESTION 6.07 (2.00) What are the five automatic trips for the diesel generator output breaker? DUESTION 6.08 (1.00) State all the signals which will initiate the DBA sequencers. QUESTION 6.09 (1.00) State the two interlocks, including uetpoints, which prevent starting a PCP. QUESTION 6.10 (1.00) State the approximate valucc for the following PCP parameters during plcnt operation at power:

a. Lower seal temperature.

b. Middle seal pressure. c. Vapor seal pressure. d. Motor amperes. QUESTION 6.11 (2.00) State all seven automatic responses which occur in the CVCS system when en SIS is generated with standby power available. QUESTION 6.12 (2.00)

a. How is the pressurizer pressure controller normally configured f or powsr operation?

b. Give two reasons f or using this configuration f or pressurizer pressure control. (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

6 __PL@NI_gygIgDg_DgglGN _CgNIBg(3_@ND_INgIBUdgNI@IlgN PAGE B 3 k -QUESTION 6.13 (2.00)

c. What interlock is associated with the hot leg saf ety injection volves?

b. What is the basis f or this interlock? QUESTION 6.14 (1.00) Why does the containment spray pump have a higher capacity during the racirculation phase than it does during the injection phase? QUESTION 6.15 (1.00) Exp1&in how the containment hydrogen sample isolation valves can be rGopened when a CHP signal is still present. . QUESTION 6.16 (1.00) What will initiate a " dropped rod" alarm on a power range safety channel? QUESTION 6.17 (1.00) What is the function of the suppression coil network systems which are incorporated in the four "M" relay control power circuits? DUESTION 6.18 (1.00) i What automatic actions, if any, are associated with the steam generator blowdown radiation montitor? QUESTION 6.19 (1.00) What two methods are used for bridge and trolley indexing? (***** END OF CATEGORY 06 *****)

Z. _PEOGEQUBE@_ _NOBd@61_@@NQBd@61_EdEBGENQY_@NQ PAGE 9 6891060G1986_99NIBg6 QUESTION 7.01 (1.00) What two indications are used to verify a reactor trip? QUESTION 7.02 (2.00) EOP 1 requires additional immediate action (s) to be taken if a safety inicction occurs in conjunction with a reactor trip. What are these adJ:. t i on al immediate action (s), and what indication (s) are used to detsrmine whether these additional immediate actions must be performed? (Include setpoints as applicable). ' DUESTION 7.03 (2.50) If, following a safety injection, the pressure in either steam generator falls below a certain value and continues to drop, four actions must be taken.

c. What is this steam generator pressure?

b. What are these four actions? 4 QUESTION 7.04 (1.50) Whct three conditions require the reinitiation of HPSI after SIAS has basn reset? QUESTION 7.05 (1.00) During a loss of all immediately available AC power, battery capacity will not be exceeded during the expected four to six hour discharge period provided two conditions are met. What are these two conditions? QUESTION 7.06 (2.00) A manual trip of the reactor and turbine generator is required if any of four conditions occur during a loss of component cooling. What are these four conditions? i ) (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) J

PROCEDURES - NORMAL _ABNgRMAL _ EMERGENCY _AND PAGE 10 7. 3 3 BSDigLgg1C8L_CgNIBgL QUESTION 7.07 (1.OC) What two conditions asscciated with a loss of instrument air require a rcactor trip? QUESTION 7.08 (1.00) How is containment isolation manually initiated? QUESTION 7.09 (1.00) What is the reason f or the precaution against simultaneously starting -- two or more PCPs? OUESTION 7.10 (2.00) What five actions must be taken to facilitate PCS heat removal if shutdown cooling cannot be established and heat removal by steam gene.ators can no longer be maintained? QUESTION 7.11 (1.50) What three immediate actions must be taken if a fuel handling accident occurs on the reactor refueling side? QUESTION 7.12 (1.00) What four types of items should the operators take with them when ovacuating the control room? QUESTION 7.13 (1.00) How can the reactor be tripped if it can not be automatically or manually tripped from the control room? (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

PROCEDURES - NORMAL _ABNgRMAL _EMgRggNgY_ANQ PAGE 11 '7. 3 2 889196991986_GQNIBQL QUESTION 7.14 (1.00) What action must be taken if lake water rises above the 590 foot level cnd enters the main plant building? QUESTION 7.15 (1.00) What is basis for the requirement that PORV breakers 52-196 and 224 normally be open? QUESTION 7.16 (1.00) -- What action must be taken if f uel pool or reactor cavity water l evel bcgins to decrease rapidly during irradiated fuel movement? QUESTION 7.17 (1.00)

c. What is the maximum permissible containment building air temperature?

b. What action must be taken if this temperature is exceeded?

. QUESTION 7.18 (1.00) ~ What is the basis for the precaution against changing diesel generator voltage control from automatic to manual while the diesel is loaded onto cn engineered bus? QUESTION 7.19 (1.50) If one of the two startup range channels fails during a startup, is it pcrmissible to continue with the startup? Explain. (***** END OF CATEGORY 07 *****)

9:__8Ddit!1SIBBIIVE_P89CEpuBES3_CgNp1IIQN@3_@Np_LiblI@I19N@ PAGE 12 QUESTION 8.01 (.50) PORV breakers 52-196 and 52-224 shall normally be open when PCS tcmperature is above.325 F. Which individual on shift, if any, can. direct that the breaker be closed? QUESTION 8.02 (.50) TRUE or FALSE. The tagging procedures specified in section 8 of the SOPS cre mandatory and must be followed whenever work is to be performed on thm respective component. QUESTION 8.03 (1.00)

c. What is the maximum amount of time which an auxiliary operator is allowed to work during any 168 hour period 7 (.5) b.

Who can authori:e deviations f rom the working hour restrictions? (.5) . QUESTION 8.04 (1.00) Whan, if ever, is it permissible to take actions which intentionally d: pert from Technical Specifications? QUESTION B.05 (1.00) What are the two priorities of the shift. W.r.,cierr, in order of importance, during an accident? QUESTION D.06 (.50) A-refamiliarization shift is required of any licensed RO who has not pcrformed as a licensed RO within the past _____. QUESTION 8.07 (.50) Which individual is normally responsible f or maintaining the control room logbook? (***** CATEGORY 08 CONTINUED ON NEXT PAGE

• • • • • )

L

9:__BDdINJgIg811yE_PBgCEpUBEg2_CgNplIJgNg3_8Np_LJDJIBIJgNg PAGE 13 QUESTION 8.08 (.50) TRUE or FALSE. The radwaste logbook is to be treated as a legal document subject to being entered in a court record. QUESTION 8.09 (1.00)

a. At least _____ auxiliary operators shall be on shift at all times.

(.5) b. How many licensed personnel (specify RO and SO) must be in the control room at all times when the plant is in a mode other than cold or refueling shutdown? (.5) " DUESTION 8.10 (1.00) What acti on must be done to perf orm an evolution f or which no procedure cxicts? QUESTION 8.11 (.50) Pcrmanent plaques listing operational information may be affixed to control panels provided the _____ concurs. QUESTION 8.12 (1.00) What two conditions must be satisfied prior to removing one diesel gcnsrator from service when the plant is less than 325 F7 QUESTION 8.13 (1.00) How is the repositioning of locked breaker documented if the locked br@cker is repositioned without using a checklist, Technical Specifi-cation surveillance or a switching and tagging order? QUESTION 8.14 (.50) TRUE or FALSE. Switch positions designated by switching and tagging ordsrs supersede system checklists. (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)

O___8Dd1NigIB@JJyE_PBgCEDUBgg2_CgNDillgNg2_8ND_ Lid 1I@IlgN@ PAGE 14 QUESTION 8.15 (.50) During startup evolutions, portions of a system's checklist may be weived by the _____. QUESTION B.16 (1.50)

c. What action should an operator take if, while perf orming a checklist, o valve which is required to be open by the " Place in Position" column is f ound to be closed? (.5) b.

Which individual can authorize opening a normally-open valve which 10 found to be closed? (.5) c. Which individual can authorize a " locked open" valve to be " locked - clocad"? (.5) QUESTION B.17 (.50) TRUE or FALSE. All caution tags issued, including those controlled within a Technical Specification surveillance procedure or switching cnd *.agging order, shall be entered in and controlled by the caution tcg log. QUESTION 8.18 (1.00)

c. Under what circumstance can the requirement f or quarterly verification of workmen's protective tags be waived for specific tags? (.5)

~ b. Who has the authority to grant this waiver? (.5) QUESTION 8.19 (2.00) Tha removal or opening of three floor plugs / hatches require additional control beyond simple Shift Supervisor approval whenever PCS temperature excoads 325 F.

a. What additional control is required?

(.5)

6. Name these three floor plugs / hatches. (1.5)

(***** CATEGDRY 08 CONTINUED ON NEXT PAGE *****) t

O___8Pd1NJpIgg}}yg_PBgggpuBggy_ggNp]I]QN!3_8Ng_LJd]I@IJgNg PAGE 15 QUESTION 8.20 (1.00) How is a control valve tagged to maintain it in the shut position if the valve f ails to the open position upon loss of air? DUESTION 8.21 (.50) TRUE or FALSE. The responsibility f or making the decision tc startup the rosctor after a reactor trip is vested in the Operations Superintendent. QUESTION B.22 (1.00) -- Maintenance immediately necessitated by f our situations is defined as cccrgencpfmaintenance. What are these four situations? QUESTION 8.23 (.50) What priority is assigned to urgent maintenance? QUESTION B.24 (.50) TRUE or FALSE. Access into high radiation areas under general RWPs is allowed with radiation safety technician coverage and without dedicated covcrage by qualified operations personnel. QUESTION 8.25 (.50) Which individual, if any, may extend a standard RWP past its expiration dcto? QUESTION B.26 (.50) What additional requirement is imposed upon personnel who enter a rcdiologically controlled area under a general RWP when work is being psrformed in the area under a standard RWP7 (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) L

0 __0DdlNiglB@llyE_88QCEQUBES _CQNDillgNS _@ND_Lidll@IlgNE PAGE 16 x 2 QUESTION B.27 (1.00) How must the primary dosimetr device be worn to insure proper monitoring of

c. beta dose? (.5) 1:

b. neutron dose? (.5) I' OUESTION B.28 (2.00)

c. What is the whole body dose limit f or newly employed f emales with unknown prior exposure histories? (.5) b.

Which three individuals must approve non-emergency whole body -- cxposures in excess of 5 rems per year? (1.5) ' QUESTION B.29 (1.50) Which three decisions may not be delegated by the SED during an cmnrgency? (***** END OF CATEGORY OB *****) (************* END OF EXAMINATION ***************) L

F l S. Fl!llATIONS REACTOR THEORY RADIATION Fl.UIDS/TilERHO/ HEAT TRANSFEM ~ T e l P = P,e* * = P,10 "** N = N,e" E 8 = A192V = Aap2V2 A = AN Q = A1Va = A2V2 l 0~ ~# t= 8J + or T = ) p Ap 10 g,g,-px -x/ E,= E,,L + AE I 10 3 stored = ~ ~ p= ' as o g ATis = 0.693 E = KE + PE + U + PV+Q+W 2 R/hr 9 d feet = g 2 k<1 - ka Iadga = 12d 2 Point s m ca 2 b"( g Igda = 12d2 - line source -=1-k N R/hr x time = R reduced for - turbine, SC pump. nozzle. 3 _ cps

• Rad x QF = Rem orifice condenser. PI e, Rx P

l M

cys, h

xT flow a (dp j Tis,gg = Bio Rad y2 g l

• act = A(pdoppler + Omod + Ovoid bBio +

IS h** I'"*

• • "
• Y Rad D 2g c

head loss a Ap +pXe

• 8Se
• PPu +

P=h + p dient =k Y a E F = PA c

1. oron + Drod + # uel +

8 f Ap2 AP1 = phase phase x K o,g,, ) k = f(quality & Pressure) p Pump laws speed a flow j ka = kg + 5k (speed)2a pressure (8 Peed)3a Power 6k=k-1 MATil j Q = kAAT = hAAT = UAAT l SUR = 1 ' y" = b 0 *

• h0 N " *A log b=a teV y

Q = cot ,, 3.1 x I0ne log x, = c lug x All - m c. AT P E = No 8 E*~ EI

• * 'y

~ 1 log xy = log x + log y H=U+PV AS = 8 Defect = Coeff x A Parameter T pV = nRT El b. = E1 12 T1 T2 4 CtVi + C2V2 = Cs(Vi + V2) l'

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f 5 __IMEQBy_QE_NgCLE@B_EgWEB_EL@NI_gEEB811gN _ElglDS _8ND PAGE 17 1 2 IHEBugDyN8dlCS ANSWERS -- PALISADES -86/06/24-HIGGINS, R. 1 ANSWER 5.01 (1.50)

o. FALSE.

(.5) b. P-BA 101 amps (.5) P-BC 93 amps (.5) REFERENCE EDP-1, step 3.6 4 ANSWER 5.02 (1.00)

a. 258 F (250 - 260)

(.5) - -13. 325 F (320 - 330) (.5) REFERENCE i Storm Tables i ANSWER 5.03 (1.50) Any six of the following (.25 each): 1. Core outlet temperature reading above saturation. e2. Core differential temperature above full power (50 F). 3. Erratic indication on the startup neutron detectors. _4. Erratic indication on qualified core exit thermocouples.

75. Hot leg temperature increasing or erratic.

4. Cold leg temperature erratic.

7. Pressurizer level increases significantly greater than expected when operating auxiliary spray. O. Unanticipated letdown flow greater than charging flow when pressurizer level control is in automatic. 9. During steady state conditions, charging a known volume of water or boric acid into the PCS does not result in a corresponding increase in pressurizer level. REFERENCE EOP 8.1, step 4.8.a ONP 21, step 4.8.9 ANSWER 5.04 (1.00) Fact neutron irradiation of the reactor vessel.

9 5 __IHEQBy_QE_NyC6E86_EQWEB_E66NI_QEEB611gN _E6MIDB _6ND PAGE 18 2 2 THERMODYNAMICS ANSWERS -- PALISADES -86/06/24-HIGGINS, R. REFERENCE Tech Spec 3-6 ANSWER 5.05 (1.00)

c. FALSE

(.5) b. FALSE (.5) REFERENCE Th;rmal-Hydraulic Principles and Applications to the PWR, B-20 and 10-5 ANSWER 5.06 (1.00) Smallest amount of net positive suction head a pump must have to prevent 4 cavitation. REFERENCE Th rmal-Hydraulic Principleti and Applications to the PWR, 10-56 ANSWER 5.07 (1.00) +

c. increase

(.5) b. decrease (.5) REFERENCE Thurmal-Hydraulic Principles and Applications to the PWR, 12-52 and 55 ANSWER 5.08 (1.00) Excessive void fraction causes flow instability (.5) which could cause tha premature onset of DNB (.5). REFERENCE Tsch Spec 2-2 and 2-7 ANSWER 5.09 (1.00) Greater system resistance to flow at the increased fluid velocity.

5. THEORY OF NUCLEAR POWER PLA'NT OPERATION,_FLUIpS _gNp PAGE 19 2 IHEBdgDyN901CS ANSWERS -- PALISADES -86/06/24-HIGGINS, R. REFERENCE Thsrmal-Hydraulic Principles and Applications to the PWR, 10-52 i ANSWER 5.10 (2.00)

c. Whenever only one source range instrument is operable.

(.5) (1 - Keff2)/(1 - Keffi) (.1) b. (CR1)/(CR2) = Kaff2 = Keffi + AK (.1) (1 - Keffl - AK)/(1 - Keff1) (.1) (CRI)/(CR2) = 1 - [ (AK) / (1 - Keff1)] (.1) (CR1)/(CR2) = 1 - [(CR1)/(CR2)3 (.1) (LK)/(1 - Koff1) = 6/14 = 3/7 (.1) = 1 - (8/14) Keff1) (4K)/(1 = 4K = 3(1 - Keff1)/7 (.1) (7/3) ( K) (.1) Keff1) (1 = (1 - Keff3)/(1 - Keff2) (.1) (CR2)/(CR3) = Kef f 3 = Kef f 2 + LK = Kef f 1 + 24K (.1) (1 - Keffi - 22K)/(1 - Kef f i - AK) (.1) (CR2)/(CR3) = (2AK) J / [ ( (7/3) (AK) ) - (AK)3 (.1) = [ ( (7/3) (AK) ) (CR2)/(CR3) (1/3)/(4/3) = 1/4 (.1) C(7/3) - 23 / E (7/3) - 1] (CR2)/(CR3) = = CR2 = (1/4) (CR3) (.1) (4)(14) = 56 (.1) CR3 = (4) (CR2) = REFERENCE NUS Reactor Operation 12.4-2 GOP 3, step 5.0 + hNSWER 5.11 (2.00)

c. 352 F

(.5)

b. Restrict the amount of positive reactivity which could result from PCS d: pressurization (1.0), since MTC will be more positive at lower temperatures

(.5). REFERENCE Tsch Spec 3-13

5 __IHEgBY_gE_NUC6E@B_EgWEB_E68NI_g[EB@llgN _ELUlDS _@ND PAGE 20 1 1 1HEBMQDYN8MICS ANSWERS -- PALISADES -86/06/24-HIGGINS, R. ANSWER 5.12 (3.00) Maximum permissible startup rate is 1.0 DPM. (.5) Pownr level at which the reactor is considered critical for the purposes of cdministrative control is 10**(-4)%. (.5) Highest permitted power level in hot standby is 2%. (.5) P= (Pc) 10**[ (SUR) (ti me) 3 (.5) P/P, 10n*C(SUR)(time)3 = (SUR) (ti me) = log (P/Pc) timm = E1/(SUR)31og(P/Pe) (.5) timm = (1/1)1og[2/10**(-4)3 = log (2*10**4) = 4.3 minutes (.5) REFERENCE __GLC 3, step 2.3.2 Tcch Spec 1-1 ANSWER 5.13 (1.50) Fiesion product inventory increases with fuel burnup (.5). Fission products are strong thermal neutron absorbers, so as fission product inventory increases the neutron flux spectrum shifts to the epithermal rcnge (.5). Control rods are strong epithermal neutron absorbers, so the relative strength of the control rods will increase as fission product " inventory increases (.5). [f NCE w Ra r Cor

Control, 6-N

+ M 67 0 N Oi7 kN [ -m. ANSWER 5.14 (2.00) Most of the xenon after a trip is due to iodine decay, rather than to the xanon which was present at the time of the trip and has not yet decayed (1.0). Since iodine concentration is linear with reactor power, the maximum xenon concentration after a trip will be linear with reactor l power since the only removal mechanism for xenon after a trip is xenon dscay (1.0). REFERENCE PWR Core Physics B-4, p 23-25 i

5 __IMEggy_QE_Ngg6 EBB _EQHgB_E6 ANT QPERATIgN _FLUIgS _ANg PAGE 21 1 1 IUE60QQyN@ digs ANSWERS -- PALISADES -86/06/24-HIGGINS, R. ANSWER 5.15 (1.00) Baccuse MTC, which is part of the power defect, becomes much more n gstive with fuel burnup (.5) due to boron removal by dilution (.5). REFERENCE PWR Core Physics B-3, p 63 ANSWER 5.16 (1.50)

1. Limit the worth of the control rods.

(.5)

2. Ensure adequate shutdown margin.

(.5) ~~^ 3. Maintain hot channel factors within limits of transient analysis. (.5) REFERENCE SH PNLB 4.19 ANSWER 5.17 (1.00) A gss which is chemically inert. (Does not react chemically.) (A gas such as helium, argon, krypton, xenon or radon.) + REFERENCE -W;bster's New Collegiate Dictionary SH P A PA 4.10 ANSWER 5.18 (1.00) I R2 action of oxygen-16 with a high energy neutron. REFERENCE Nuclear Engineering Handbook, p 7-20 l SH PNPA 4.10 l t I

t. 6___E60NI_Sy@IEME_QEgl@N _QQNIBQL3_@NQ_lN@lRQMENI@IlgN PAGE 22 1 ANSWERS -- PALISADES -86/06/24-HIGGINS, R. ANSWER 6.01 (1.50)

c. Open.

(.5) b. Prevent a loss of instrument air from causing a loss of CCW to the PCPs.(.25), CRDMs (.25), letdown heat exchanger (.25) or shield cooling coi l s (.25). REFERENCE SH PNAB 4.11 ANSWER 6.02 (2.00) c. 1. The fire protection system can provide partial backup supply to, both critical service water headers. (.5) 2. The fire protection system can provide cooling water to the service / instrument air compressors. (.5) 3. The service water boos.ter pumps can be aligned via temporary connections as a backup for the fire system jockey pump. (.5)

b. Fire water can be supplied as emergency makeup to the spent f uel pool via a swing-elbow connection.

(.5) REFERENCE SH-PNAD 4.9 ,w CNSWER 6.03 (1.00) Ecch of the following is worth.25.

1. Bus undervoltage.

2. Low pump suction pressure.

3. Overcurrent.

4. Load shed. REFERENCE 7 SH PNEP 4.16 I ! ANSWER 6.04 (1.50) i

1. Manual trip button in the main control room.

(.5) l

2. Manual trip button at the main feed water pump.

(.5)

3. Manual trip lever at the main feed water pump.

(.5) ( 0 S YVc I

6___E68NI_gy@IEd@_DE@lGN _CgNIRg61_@ND_lNQIRydENI@IIQN PAGE 23 1 .' ANSWERS -- PALISADES -86/06/24-HIGGINS, R. REFERENCE SH PNFA 4.8 ANSWER 6.05 (1.00)

c. Feed water regulating valve will shut.

(.25) Feed water regulating bypass will be unaffected. (.25) b. Feed water regulating valve will shut. (.25) Feed water regulating bypass valve will shut. (.25) REFERENCE SH PNFB 6.6 RNSWER 6.06 (1.00) i (.25 each) 1. Main oil pump if the turbine is rotating f aster than 2/3 speed. 2. Air side HP seal oil backup pump. 3. Emergency air side seal oil backup pump.

4. Turning gear oil pump.

REFERENCE SH PNGD 4.9 8NSWER 6.07 (2.00) (.4 each) 1. Engine trip. 2. Loss of excitation. 3. Overcurrent. 4. Differential relay. 5. Bus transfer. REFERENCE SH PNHB 4.17 ANSWER 6.08 (1.00) 1. Bus 1C or 1D undervoltage and SIS (.5) OR 2. Reactor trip or turbine trip and undervoltage on the startup transformer and SIS (.5).

6 __PL@NI_@lgIEMg_DEgl@N _CgNIBg(2_@jD_lN@lBgMENI@IlgN PAGE 24 1 ANSWERS -- PALISADES -86/06/24-HIGGINS, R. REFERENCE SH PNHD 4.4 ANSWER 6.09 (1.00) (.5 each) 1. Lift oil pressure below 2000 psig. 2. CCW flow less than 80 gpm. REFERENCE SH PNJC 4.9 _AN,SWER 6.10 (1.00) (.25 each)f g a. 100 - W: F b. 1200 - 1400 psia 20-jfIpsia c. d. 600 - 620 psia 630 REFERENCE SH PNJD 4.15 w ANSWER 6.11 (2.00) (.3 each except where noted) 1. Boric acid pumps start (P 56 A and B). 2. Coolant charging pumps start (P 55 A, B and C). 3. Boric acid gravity feed valves open (MOV-2169 and 2170). 4. Boric acid pumped feed stop valve (MOV-2140) opens. 5. VCT outlet valve (MOV-2087) closes. 6. Concentrated boric acid tanks recire valves (CV-2130 and 2136) close. 7. Boric acid makeup stop valve (CV-2155) closes. (.2) REFERENCE SH PNKA 4.24 t

6 __P6@NI_SySIEMg_pE@I@N _CQNIRQ61_@Np INSTRUMENTATION PAGE 25 2

• ANSWERS -- PALISADES

-86/06/24-HIGGINS, R. ANSWER 6.12 (2.00)

c. Backup heaters are all in the "DN" position (.5) with the pressure controller auto setpoint set for 60 to 70 psi lower than the normal prGeeure setpoint of 2010 psia

(.5), b. (Any two of the following at .5 tsach) 1. Quicker spray response. 2. Better baron mixing between the PCS and PZR.

3. Less thermal shock on the spray no::les.

REFERENCE SH PNKD 4.14 N SWER 6.13 (2.00)

n. A hot leg safety injection valve can not be opened until the cold leg cafoty injection valve in its train is shut.

(1.0) b. Insure HPSI pumps do not experience pump run out. (1.0) REFERENCE SH PNMA 4.23 and 4.24 ANSWER 6.14 (1.00)

Tip water during the recirculation phase is less dense.

a Y~1 W yemuAAe <fwk.04A W n REFERENCE M. Am a #.r.,_ t.,3

1. ,a Lg, g a

SH PNMB 4.11

u. b d Erg T',_ff,u g,,, %1;fg24 g

, g,g,,,'3 FS4R(A bjg%<, A#4 Md o ,, 7j g, ,f JM7 IWA CM rew f ANSWER 6.15 (1.00) AMMd /

• J s-Turn key switchen HS-2418 and 2419 to the " Accident ' position

(.5), then turn hand switches HS-2416 and 2417 to the "close" and then to the "opun" position (.5). hf gj REFERENCE SH PNMD 4.10 yJM MY M2

6 __ PLANI _SYSIEdS_DEg1GN _ CON 16QL _@ND_lN@l6QdENI@ll@N PAGE 26 1 3 " ANSWERS -- PALISADES -86/06/24-HIGGINS, R. ANSWER 6.16 (1.00) Any one of the four power range safety channels detecting a drop in pownr (.4) of more than 8% (.3) in less than 8 seconds (.3). REFERENCE CH PNNA 4.46 ANSWER 6.17 (1.00) Provent arcing across the matrix relay contacts when a reactor trip cignal deenergizes a matrix relay, causing its associated matrix contact to open. REFERENCE SH PNNB 4.39 . ANSWER 6.18 (1.00) (Either of the f ollowing f or full credit) 1. Isolates the steam generator blowdown tank.

2. Shuts CV-0704, 0738, 0739, 0770 and 0771.

+ REFERENCE _CH PNPA 4.10 ' ANSWER 6.19 (1.00) (.5 each) 1. Digital readout on the console.

2. Mechanical pointer system along the trolley rail and the bridge.

REFERENCE SH PNOC 4.15 l l I

Z __P6QCEDQ6ES_ _NOBdBL _@@NgBd8L _EMEBGENgy_8ND PAGE 27 1 1 RADIOLOGICAL CONTROL j ANSWERS -- PALISADES -86/06/24-HIGGINS, R. l AN5WER 7.01 - (1.00) (.5 each) 1. All the full length control rods indicate fully inserted.

2. Reactor power is decreasing (negative startup rate).

REFERENCE EOP 1, step 3.1 ANSWEP 7.02 (2.00) If pressurizer pressure is less than 1300 psia, trip two PCPs (one in --ecch loop, preferably P-50A and 50D). (1.0) If pressurizer pressure drops below the minimum pressure for PCP opsration, then trip the remaining two PCPs. (1.0) REFERENCE EOP 1, step 3.6 ANSWER 7.03 (2.50) wo. COO psia. (.5) b. (.5 each) 1. Close both MSIVs (CV-0510 and 0501). 2 2. Close both feedwater regulating valves (CV-0701 and 0703). 3. Close both feedwater regulating bypass valves (CV-0734 and 0735). 4. Emergency borate to establish 3.75% shutdown margin or cold shutdown boron concentration, whichever is greater. REFERENCE EDP 1, step 4.10 ANSWER 7.04 (1.50) (.5 each) 1. PCS subcooling is less than 25 F. 2. Pressurizer level is below 20% or decreasing. 3. No steam generators are available f or removing heat. l REFERENCE EOP 1, step 4.15 I

Z___EBQCEDQBEQ_ _NQBd@(1_@@NQ6d@61_EMEBGENCy_@ND PAGE 28 88D1969GIC86_CQNIBQL ANSWERS -- PALISADES -86/06/24-HIGGINS, R. ANSWER 7.05 (1.00) Battery current is reduced to less than 150 amps within 30 minutes of the loss of AC power (hp) and maintained below 150 amps isrestoredtothebatte[rychargers WMy). tha anset of until AC power REFERENCE EDP 2.1, step 4.9 ANSWER 7.06 (2.00) (.5 each) --d. PCP seal bleed off temperature exceeds 185 F.

2. PCP bearing temperature exceeds 175 F.

3. All (or most) control rod drive seal leak off temperatures exceed 200 F. 4. CCW flow to the PCPs is interrupted for more than ten minutes. REFERENCE EDP 4, step 4.1 ANSWER 7.07 (1.00) (.5 each) 1 1. Instrument air pressure drops below 50 psig. 2. First indication of erratic equipment behavior. REFERENCE EOP 5, step 4.1

ANSWER 7.08 (1.00)

Puthing left and right HIGH RADIATION INITIATE pushbuttons. l REFERENCE EOP 6 l l I

Z___P899EgyEES_ _NQEd@(2_QBNQBd@62_EdEEGENQy_@NQ PAGE 29 6891969 GIG 86_99N18QL ANSWERS -- PALISADES -86/06/24-HIGGINS, R. ' ANSWER 7.09 (1.00) Provent causing excessive starting currents on Bus 1A or 1B. REFERENCE EOP B.1, step 4.19 ANSWER 7.10 (2.00) (.4 each) 1. Maximize HPSI cold leg injection flow. 2. Close PORV breakers. --3. Open PORV isolation valves.

4. Check PORV handswitches in auto.

5. Pull out any two pressurizer high pressure bistable trip modules on the RPS to open the PORVs.

REFERENCE EDP 8.1, step 4.2.4.3.b ANSWER 7.11 (1.50) (.5 each) ,1. Dispatch personnel to secure the equipment access door and the personnel air lock doors if they are open.

2. Notify the control room and evacuate containment.

3. Evaluate the status of containment penetrations and initiate any cctions necessary to prevent air leakage from containment.

REFERENCE EOP 9, step 3.1 ANSWER 7.12 (1.00) (.25 each)

1. Radios.

2. Vital area door keys.

a

3. Locked valve keys.

4. Emergency flashlights.

Z __P69gEgggg@_ _NgBd@(1_@@NgBd@(1_Ed[8@ENgy_9NQ PAGE 30 8091969E1C66_CQNIBQL ANSWERS -- PALISADES -86/06/24-HIGGINS, R. REFERENCE EDP 10.2, step 4.1 ANSWER 7.13 (1.00) Opsn breakers RPS 42-1 and 42-2 at the CRDM clutch power supply trcnsformers in the cable spreading area. REFERENCE ONP 7, step 4.1 ANSWER 7.14 (1.00) Trip the reactor and follow up with EOP 1. REFERENCE ONP 12, Seich, step 4.2 ANSWER 7.15 (1.00) Th9 basis is a fire in the cable spreading room or control room could

• ccute " hot shorts" simultaneously in the control circuits for a PORV cnd its block valve

(.5), resulting in the equivalent of a LOCA during a afiro (.5). REFERENCE SDP 1, step 4.0.u ANSWER 7.16 (1.00) Imm diately place the bundle back in the core, tilt machine or spent fual rack, whichever is most convenient. REFERENCE SDP 28, step 5.26

Zi__EB9EEE96EE_2_d98U6hA_6Ed98d661_Edg69 Edgy _6dD PAGE 31 689196991G66_QQNI696 ANSWERS -- PALISADES -86/06/24-HIGGINS, R. ANSWER 7.17 (1.00)

c. 150 F.

(.5) b. S ut own the plant in accordance with Tech Spec 3.0.3. (.5) A SS) US&b S&j NY SDP 24, step 4.0.1 ANSWER 7.18 (1.00) Th re will be a voltage perturbation because auto and manual output

__ voltage will not be the same unless they are adjusted prior to loading.

REFERENCE CDP 22, step 7.5.3.e ANSWER 7.19 (1.50) It is permissible to continue the stYtup (.5) if both WRL channels and tho other startup range channel remain operable (.5), and a 1/M plot is encintained until criticality is achieved (.5). REFERENCE = COP 35, step 4.0.b y. _. _ ~ ._.~

8 __0Dd1NISI6611YE_66QCEDQBE@i_CQND111QNg2_6ND_LIMlI611gNS PAGE 32 t .- ANSWERS -- PALISADES -86/06/24-HIGGINS, R. ANSWER 8.01 (.50) Shift Supervisor. REFERENCE SDP 1, step 4.0.u ANSWER B.02 (.50) Fal ce. REFERENCE Any SOP, section 8 ANSWER 8.03 (1.00)

c. 72 hours.

(.5)

b. Plant General Manager.

(.5) REFERENCE Admin Proc 1.00, steps 26.1 and 26.3 + 8NSWER 8.04 (1.00) In cn emergency (.5) when abiding by Technical Specifications would not provide adequate or equivalent protection to the public (.5). REFERENCE 10CFR50.54(X) ANSWER 8.05 (1.00) 1. Function as site emergency director until relieved. (.5)

2. Perform the accident assessment function. (.5)

REFERENCE Admin Proc 4.00, step 4.10.1.1

8. ADMINISTRATIVE PROCEDURES _CgNDITIQN@1_AND_LIMITATIgN@ PAGE 33 2 , ANSWERS -- PALISADES -86/06/24-HIGGINS, R. ANZWER 8.06 (.50) 35 days. REFERENCE Admin Proc 4.01, step 5.1 i ANSWER 8.07 (.50) Control operator 1. REFERENCE Admin Proc 4.01, step 5.7.2.b.2 ANSWER 8.08 (.50) Trua REFERENCE Admin Proc 4.01, step 5.7.2 w ANSWER B.09 (1.00) 10. 4. (.5)

b. Two licensed individuals (.25), at least one of whom holds an 50 liccnse (.25).

REFERENCE Admin Proc 4.01, step 5.1.1 ANSWER 8.10 (1.00) A procedure shall be developed and approved prior to commencing the cvolution. REFERENCE Admin Proc 4.01, step 5.3.9

8___8pd1NIS160llyE_66QCEQQ6E@s_CQNQlligNS _@NQ_Lidl18IlgN@ PAGE 34 1 .. ANSWERS -- PALISADES -86/06/24-HIGGINS, R. ANSWER 8.11 (.50) Chift Supervisor. REFERENCE Admin Proc 4.01, step 5.8 ANSWER 8.12 (1.00) Ono service water pump (.4) and one component cooling water pump (.4) cro operable on the operable diesel generator (.2). REFERENCE

--Admin Proc 4.01, step 5.11.a

) AN'WER 8.13 (1.00) Cy logging it in the control room logbook (.5), along with the names cf the repositioner and verifier (.5). REFERENCE Admin Proc 4.02, step 5.3.4 w -kNSWER 8.14 (.50) Trua. REFERENCE Admin Proc 4.02, step 7.2 ANSWER 8.15 (.50) Operations Superintendent. REFERENCE Admin Proc 4.02, step 7.2.1 f a

4 0 __e901NIEIBOI1YE_PBQgEpyBE@1_QgNQlIlgN@2_@NQ_LidlI@llgNE PAGE 35 ANSWERS -- PALISADES -86/06/24-HIGGINS, R. ANSWER B.16 (1.50)

c. Note the condition on the " Record of Exceptions" sheet. (.5) b.

Shift Supervisor. (.5) c. Operations Superintendent. (.5) REFERENCE Admin Proc 4.02, step 7.2.2 and 7.2.3 ANSWER 8.17 (.50) Folce --REFERENCE Admin Proc 4.03, step 14.1 AN!WER 8.18 (1.00) c. If the verification would result in significant radiation cxposure. (.5) b. Operations Superintendent. (.5) wREFERENCE Admin Proc 4.03, step 10.2 ANSWER 8.19 (2.00)

c. A jumper link and bypass must by used. (.5) b.

1. Engineering safeguards room floor plugs. (.5) 2. West engineering safeguards room hatch. (.5) 3. AFW room floor plug. (.5) REFERENCE Admin Proc 4.03, step 19.0 . ANSWER 8.20 (1.00) i A m2chanical block is installed on the valve to keep it shut (.5), then j c tcg is placed on the mechanical block (.5).

H __8Dd1NISIE@IlyE_PEQCEDUBES _CQNQlIlQNS _8ND_(ldlI@IlgN@ PAGE 36 1 1 ANSWERS -- PALISADES -86/06/24-HIGGINS, R. REFERENCE Admin Proc 4.03, attachment 1, step 2.7.b ANSWER B.21 (.50) Felse REFERENCE Admin Proc 4.08, step 4.1 ANSWER B.22 (1.00) (.25 each) --1. Allow safe shutdown of the plant.

2. Reduce imminent possibility of personnel injury.

3, Reduce imminent possibility of damage to major equipment.

4. Reduce imminent possibility of hazard to public safety.

REFERENCE Admin Proc 5.01, step 5.12 ANSWER 8.23 (.50) + Priority 8.

1. REFERENCE Admin Proc 5.01, attachment 1, block 19 ANSWER B.24

(.50) True REFERENCE Admin Proc 7.03, step 5.1.c ANSWER 8.25 (.50) Rcdiation Saftety Supervisor a

a___0901Ni@I6611VE_P6QCEQQ6E@1_CQNQlligN@i_@NQ_(ldlI@llgN@ PAGE 37 ,,AN2WERS -- PALISADES -86/06/24-HIGGINS, R. REFERENCE Admin Proc 7.03, step 5.2.a ANSWER 8.26 (.50) Road, sign and comply with the standard RWP. REFERENCE Admin Proc 7.03, step 7.2.c AN:WER O.27 (1.00)

c. With the window on the front of the badge facing away from the

.--b dy. (.5) f b. Close to the body. (.5) REFERENCE Adain Proc 7.04, steps 5.3.1.c and 5.4.d ANSWER 8.28 (2.00)

c. 300 mrem. (.5) b.

1. Radiological Services Manager (.5) y

2. Plant General Manager (.5)

3. Vice President of Nuclear Operations (.5)

REFERENCE Admin Proc 7.04, step 5.2.b; table 1 ANSWER 8.29 (1.50)

1. Recommend protective actions to off-site organizations (.5)

2. Evacuate the site.

(.5) 3. Authorize exposures in excess of 10CFR2O limits. (.5) REFERENCE Sito Emergency Plan 5.4.1 >}}