ML101270171

2010-03 Draft Written Exam

ML101270171
Person / Time
Site:	Arkansas Nuclear
Issue date:	03/16/2010
From:	NRC Region 4
To:	Entergy Operations
References
50-313/10-03, 50-368/10-03
Download: ML101270171 (321)

Text

Form ES-401 -2 PWR Examination Outline ES4OI 31512010 j Date of Exam:

Unit I Facility: Arkansas Nuclear One SRO-Only Points A2 G* Total Group AG Tier KKKKKKAAA

• Total 6 1 2 3 4 1 2 3 4 5 6

.1. 18

1. 1 1. .1. ..L 4 Emergency & 1 2 N/A L 9 2 2 2 1 N/A Abnormal 10 4 5 4 27 Tier Totals 5 5 4 Evolutions 5 233 28 1 33233222 3

2. 111 1 10 2 0111111 Plant 8 3 3 3 4 4 38 Systems Tier Totals 3 4 3 4 4 3 1 2 3 4 7 2 3 4 10 ities 1

3. Generic Knowledge and Abil Categories 3 2 2 3 sampled within each tier of the RO every applicable K/A category are Note: 1. Ensure that at least two topics from for one category in Tier 3 of the SRO-only outline, the Tier Totals and SRO-only outlines (i.e., except than two).

in each K/A category shall not be less match that specified in the table.

tier in the proposed outline must revisions.

2. The point total for each group and tier may devi ate by +/-1 from that specified in the table based on NRC each grou p and The final point total for t total 25 points.

points and the SRO-.only exam mus The final RO exam must total 75 systems or evolutions that do not appl y

ution s with in each grou p are identified on the associated outline; syste ms/e volu tions that are not Systems/ evol site-specific

3. justified; operationally important, rding the elimination at the facility should be deleted and Refe r to Sect ion D. I .b of ES-40I for guidance rega ld be adde d.

included on the outline shou of inappropriate K/A statements. m or evolution in the group befo re as man y syste ms and evol ution s as possible; sample every syste

4. Select topics from m or evolution.

selecting a second topic for any syste g (lR) of 2.5 or higher shall be selec ted.

plan t-spe cific prior ity, only thos e K/As having an importance ratin Abse nt a ctive ly.

5. the RO and SRO-only portions, respe Use the RO and SRO ratings for categ ories.

from the shaded syste ms and K/A

6. Select SRO topics for Tiers I and 2 2 of the K/A Catalog, but the topics 2 shall be selected from Sect ion s.

The gene ric (G) K/A s in Tiers I and ion D.1. b of ES-401 for the applicable K/A 7

relev ant to the appl icabl e evol ution or system. Refer to Sect gs (IRs )

must be s importance ratin enter the K/A num bers, a brief description of each topic, the topic the grou p and tier total s On the following page s, and category. Ente r

8. the point totals (#) for each system or G* on the for the applicable license level, and led in othe r than Cate gory A2 e; if fuel handling equipment is samp 2 (Note #1 does not apply). Use duplicate for each category in the table abov p enter it on the left side of Column A2 for Tier 2, Grou SRO-only exam, exam s.

pages for RO and SRO-only the K/A numbers, descriptions, IRs, ion 2 of the K/A catalog, and enter

9. For Tier 3, select topics from Sect K/A s that are linked to 10 CFR 55.43.

poin t totals (#) on Form ES-4 01-3. Limit SRO selections to and Form ES-401-2 ES-401

RO Written Exam Tier I Group I

CHANGE 031 PAGE 2 of 25 1202.001 REACTOR TRIP INSTRUCTIONS CONTINGENCY ACTIONS Manually Trip Rx.

A. Perform the following:

A. Verify all rods inserted

1) lFRxfailstotrip, AND THEN depress CRD Power Supply Breaker Trip PBs on C03 Reactor power dropping.

(A-501 and B-631).

a) IEA-501 or B-631 fails to trip, THEN manually insert rods at C03.

AND Dispatch an operator to open CRD AC Power Supply breakers.

2) IF more than one rod fails to fully insert OR Rx power is dropping, THEN perform Emergency Boration (RT 12).

3) DO NOT continue until the reactor is shutdown.

CHANGE 1202.001 REACTOR TRIP 031 PAGE 3 of 25 INSTRUCTIONS CONTINGENCY ACTIONS Manually trip Turbine.

A. Verify Turbine throttle and governor valves A. Perform the following:

closed.

1) IF 125V DC Bus DOl is de-energized as indicated by both of the following, THEN perform Loss of 125V DC (1203.036) Loss Of Bus DOl section in conjunction with this procedure.

• Turbine Trip Solenoid Power Available light off.

• Breaker position indications on left side of Cl 0 off.

2) jf SG press is < 900 psig, THEN perform the following:

a) Actuate MSLI for affected SG(s)

AND actuate EFW AND verify proper actuation and control (RT 6).

b) Advise Shift Manager to implement Emergency Action Level Classification (1903.010).

c) GO TO 1202.003, OVERCOOLING procedure.

I CHANGE 1202.001 ] REACTOR TRIP 031 PAGE 4 of 25 INSTRUCTIONS CONTINGENCY ACTIONS Check adequate SCM. 3. Check elapsed time since loss of adequate SCM AND perform the following:

A. IE 2 minutes have elapsed, THEN trip all RCPs.

B. IF >2 minutes have elapsed, THEN leave currently wnning RCPs on.

C. Advise Shift Manager to implement Emergency Action Level Classification (1903.010).

D. GO TO 1202.002, LOSS OF SUBCOOLING MARGIN procedure.

4. Advise Shift Manager to implement Emergency Action Level Classification (1903.01 0).

5. Reduce Letdown by closing Orifice Bypass (CV-1 223).

6. Open BWST Outlet to OP HPI pump (CV-1 407 or 1408).

7. j Emergency Boration is QI in progress, THEN adjust Pressurizer Level Control setpoint to 100.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

Rev: 0 Rev Date: 9/03/09 Source: New Originator: S.PuIlin QID: 0771 Objective: 2 Point Value: 1 TUOI: AILP-RO-AOP Section: 4.2 Type: Generic APEs System Number: 008 System

Title:

Pressurizer (PZR) Vapor Space Accident ility safety related equipment.

==

Description:==

Ability to determine operability and/or availab KIA Number: 2.2.37 CFR

Reference:

41 .7/43.5/45.13 3.6 RO Select: Yes Difficulty: 3 Tier: 1 RO Imp:

SRO Imp: 4.6 SRO Select: Yes Taxonomy: C Group: I Question: RO: 2 SRO:E Given:

close the Spray valve and stopped the

- Pressurizer Spray fails open and the ATC operator was able to Reactor Coolant system pressure decrease.

in.

- Annunciator alarm PZR HEATER GROUND FAULT (K09-E3) comes energized.

- RCS pressure response abnormally slow with Pressurizer heaters ed Pressurizer Heater Checkout (1307.009) to

- Maintenance is requested to perform Unit I Emergency Power determine operability of vital powered pressurizer heaters which KW output of the vital powered Which heaters groups are the vital powered pressurizer heaters, and ication 3.4.9?

heaters will satisfy the operability requirements of Technical Specif

, Group 4 heaters, 120 KW output.

A. Group I proportional heaters, Group 2 proportional heaters

, Group 5 heaters, 135 KW output B. Group I proportional heaters, Group 2 proportional heaters

, 124 KW output C. Group 1 proportional heaters, Group 3 heaters, Group 5 heaters

, 128 KW output D. Group 2 proportional heaters, Group 4 heaters, Group 5 heaters Answer:

, Group 5 heaters, 135 KW output B. Group I proportional heaters, Group 2 proportional heaters Notes:

A. is incorrect wrong groups of heaters and KW output to low ility requirements of TS 3.4.9 B. is the correct answer correct groups of heaters and KW meets operab C. is incorrect wrong groups of heaters and KW output to low ments of TS 3.4.9 D. is incorrect wrong groups of heaters and KW meets operability require

References:

1203.015 change 016 T.S. 3.4.9 amendment # 215 History:

New for the RO/SRO 2010 exam

CHANGE 016 PAGE 7 of 24 1203.015 PRESSURIZER SYSTEMS FAILURE SECTION 3--INOPERATIVE PRESSURIZER HEATER(S)

ENTRY CONDITIONS One or more of the following:

setpoint:

• Pressurizer heaters do not energize in AUTO at proper

- Banks I and 2 full on: 2135 psig

- Bank3on: 2l35psig

- Bank4on: 2l2Opsig

- Bank 5 on: 2105 psig energized

• RC pressure response abnormally slow with Pressurizer heaters 3)

• Annunciator alarm PZR HEATER GROUND FAULT (K09-E

CHANGE 016 PAGE 9 of 24 1203.015 PRESSURIZER SYSTEMS FAILURE SECTION 3-- INOPERATIVE PRESSURIZER HEATER(S)

NOTE

• In order to satisfy the require ments of TS 3.4.9, Group I Proportional heaters, Group 2 be operable. This ensures Proportional heaters and Group 5 vital powered heaters shall offsite power concurrent with a that 126 KW (nominal) is available in the event of a loss of single failure of one EDG.

l transfer via B55/56. In the event

• Group 5 vital powered heaters shall be capable of manua rizer Heaters is B55/56 can not be manually transferred, then one train of Pressu considered inoperable.

both trains of Pressurizer

• If Group 5 vital powered heaters are declared inoperable, then is NOT applicable. Entry into Heaters are considered inoperable. Per Licensing, TS 3.0.3 of Pressurizer Heaters.

TS 3.4.9 Condition C is required for inoperability of both trains (1307.009) is used to determine

• Unit I Emergency-Powered Pressurizer Heater Checkout operability of vital powered Pressurizer Heaters.

5. j any Pressurizer heater is declared inoperable, THEN perform the following:

• Initiate action to repair heater.

• Initiate a Condition Report.

• Refer to TS 3.4.9 and TRM 3.4.9.

Limits of the ANOI COLR

• Refer to RCS Pressure, Temperature and Flow DNB Surveillance (TS 3.4.1).

• Notify Ops Manager.

Pressurizer 3.4.9 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 Pressurizer LCO 3.4.9 The pressurizer shall be OPERABLE with:

a. Pressurizer water level 45 inches and 320 inches; and

b. A minimum of 126 kW of Engineered Safeguards (ES) bus powered pressurizer heaters OPERABLE.

OPERABILITY requirements on pressurizer heaters do not apply in MODE 4.

APPLICABILITY: MODES 1, 2, and 3, MODE 4 with RCS temperature> 262°F.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressurizer water level not A. I Iestore level to within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> within limits, limits.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 4 with RCS 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> temperature 262°F.

C. Capacity of ES bus C.1 Restore pressurizer heater 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> powered pressurizer capacity.

heaters less than limit.

D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition C not met.

D.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ANO-1 3.4.9-1 Amendment No. 215

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I Rev: 0 Rev Date: 9/3/09 Source: New Originator: S.Pullin QD: 0772 Objective: 5 Point Value: I TUOI: A1LP-RO-AOP Section: 4.1 Type: Generic EPEs System Number: 009 System

Title:

Small Break LOCA apply to a small break LOCA: RB sump level

==

Description:==

Ability to operate and monitor the following as they KIA Number: EM .02 CFR

Reference:

41.7/45.5/45.6 3.8 RO Select: Yes Difficulty: 4 Tier: 1 RO Imp:

1 SRO Imp: 3.8 SRO Select: Yes Taxonomy: Ap Group:

Question: RO:I 3 SRO:1 3 Given:

Small break LOCA has occurred.

The Reactor building sump is filling at a rate of 2%/minute.

Reactor Building sump level is 44%

how long can the Reactor building sump be What is the RCS leak rate and with the leak size remaining steady used for an accurate leak rate calculation?

can be used for 3 minutes.

A. RCS leak rate approximately 91 gpm, and the RB sump level can be used for 16 minutes.

B. RCS leak rate approximately 80 gpm, and the RB sump level can be used for 28 minutes.

C. RCS leak rate approximately 72 gpm, and the RB sump level can be used for 3 minutes.

D. RCS leak rate approximately 45 gpm, and the RB sump level Answer:

can be used for 3 minutes.

A. RCS leak rate approximately 91 gpm, and the RB sump level Notes:

the RB sump can only be used for leak rate A. is the correct answer due to sump is 45.4 gallons per/ % and te leak rate due to volume determination up to 50% level after that level you can not get an accura uncertainties B. is incorrect due to the wrong leakrate and time.

C. is incorrect due to the wrong leakrate and time.

D. is incorrect due to the wrong lea krate.

References:

STM 1-08 Rev. 14 History:

New for the RO/SRO 2010 exam

Reactor Building Spray & Containment Building STM 1-08 Rev. 14 2.9.2.1 RB Sump Isolation Each RB sump ECCS suction line is equipped with two Valves isolation valves. The four ECCS isolation valves are 14-inch motor operated gate valves. The inside RB isolation valves are manufactured by Anchor Darling. They are designated as CV-1414 and CV-1415 with a design pressure and temperature of 100 psig and 150°F. Both inside isolation valves are located in the RB sump. The outside isolation valves are manufactured by Aloyco. They are designated as CV-1405 and CV-1406 with a design pressure and temperature of 75 psig and 300°F.

Isolation valves CV-1414 and CV-1405 provide the suction line from the RB Sump to P-34A/P-35A. CV-1405 is located in the A DH vault. They are controlled by hand-switches located on panel C-18.

Isolation valves CV-1415 and CV-1406 provide the suction line from the RB Sump to P-34B1P-35B. CV-1406 is located in the B DH vault. They are controlled by hand-switches located on panel C-16.

The inside containment isolation valves (CV- 1414 & CV- 1415) are normally left open. The outside isolation valves (CV-1405 &

CV-1406) are normally closed. All four RB sump ECCS isolation valves are required to be operable whenever RB integrity is required.

They can be either manually or remote-manually operable. The four RB sump isolation valves do not receive an ESAS signal to open or close since their safety function supports ECCS.

RB Sump isolation valve power supply and associated handswitch for each are provided in the following table.

RB Sump supply to P-34A & P-35A CV-1414(RBlnsidelsol) HS-1414/C-l8 B-51112 CV-1405 (RB Outside Isol) HS-1405 /C-18 B-5l113 RB Sump supply to P-34B & P-35B CV-1415 (RB Inside Isol) HS-1415 / C-16 B-6163 CV-1406 (RB Outside Isol) HS-1406 / C-16 B-6166 (Refer to Table 8.01) 2.9.2.2 RB Sump Level Instrumentation RB sump level is independently measured by level transmitter LT-1405 and level sensor LE-1405B. Each level instrument is mounted in the RB sump and has a measuring range of 0 to 56 inches, which correlates to 0 to 100%. The instruments measure sump level from 6 inches (0%) above the sump bottom to the sump maximum level of 46 inches (71.4%). Each percent from 0% to 50%

of level indication equals a maximum o 45.4 gallons. S leve indication fo determining CS eak rat is limited to 50% ue to volume uncertainty above that level. Each instruments range extends 16 inches above the RB floor.

LT-1405s signal is used for indiction and an annunciator alarm function. RB sump level can be read on panel C-14 from LI-1405 or SPDS (L1405). LT-1405 has a sensitivity of 1/4 inch or 0.45% of 17

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

Rev: 1 Rev Date: 8/9/05 Source: Direct Originator: J. Haynes QID: 0198 Objective: 6 Point Value: 1 TUOI: Al LP-RO-RBS Section: 4.1 Type: Generic EPEs System Number: 011 System

Title:

Large Break LOCA Break LOCA and the following: Pumps.

==

Description:==

Knowledge of the interrelations between the Large K/A Number: EK2.02 CFR

Reference:

41.7/45.7 2.6 RO Select: Yes Difficulty: 2 Tier: 1 RO Imp:

SRO Imp: 2.7 SRO Select: Yes Taxonomy: K Group: 1 Question: RO: 4 SRO:r Given:

- A large break LOCA has occurred.

- Offsite power has been lost.

200 gpm prior to transferring to Reactor Building Why must Reactor Building Spray flow be throffled to 1050-1 sump suction?

A. To ensure adequate NPSH for ECCS pumps.

B. To prevent pump runout on the Spray pumps.

C. To lower load on the EDGs.

D. To limit corrosion of reactor building equipment.

Answer:

A. To ensure adequate NPSH for ECCS pumps.

Notes:

(a.) is correct.

is achieved during ES conditions.

(b.) is incorrect. The spray pumps are designed for the full flow that spray pumps at full flow.

(c.) is incorrect. The EDGs are designed to handle the load of the for corrosion due to RB spray.

(d.) is incorrect. The design of RB equipment includes allowances

References:

1202.012, Chg. 008 History:

Developed for use in 98 RO Re-exam.

Selected for use in 2005 RO exam, replacement question.

Selected for the RO/SRO 2010 exam.

CHANGE I 1202.012 REPETITIVE TASKS 008 PAGE 37 of 50 Pagelof6 r WARNING in areas damaged, THEN initia tion of sump recirculation may cause high radiation i.E core is significantly near HPI, LPI, and RB Spray system piping.

CAUTION on press.

p recirc may result in inadequate pump sucti

• Failure to throttle RB Spray before initiating sum tes.

y can remove 6 of water from BWST in 5 minu

• Full flow from both trains of HPI, LPI, and RB Spra NOTE be overridden as necessary to perform this task.

If ES has actuated, individual component sicinals may

15. Shift to RB sump suction:

A. Verify LPI pumps running (P34A and B).

associated HPI pump.

1) i.E either LPI pump is unavailable, THEN stop

2) Verify LPI Room Coolers running:

I P-34A Room P-34B Room I 0 and 1401).

3) Verify both LPI Block valves fully open (CV-140 B. Verify Letdown isolated by either:

Letdown Coolers Outlet (CV-1 221)

OR Letdown Cooler Outlets (CV-1214 and 1216).

(15. CONTINUED ON NEXT PAGE) f 1202.012 I RTI 5 U Rev 3-17-08

CHANGE 1202.012 REPETITIVE TASKS 008 PAGE 38 of 50 Page 2 of 6

15. (Continued).

C. IF HPI is in service, THEN perform the following:

1) IF either of the following sets of conditions is satisfied, THEN terminate HPI as follows:

All of these conditions satisfied: OR Both of these conditions satisfied:

• CET SCM is adequate

• CETs do not indicate superheated

• y LPI flow exists

• LPI flow meets the following criteria:

• HPI throttled to 110 gpmlpump 2 LPI pumps I LPI pump

• RCS press and temp are ! rising 2800 gpm/pump 3050 gpm a) Start AUX Lube Oil pumps for running HPI pumps.

b) Stop running HPI pumps.

c) Close all HPI Block valves.

d) Close RCP Seal INJ Block (CV-1206).

2) IF HPI termination criteria are not met, THEN verify both Decay Heat Supply to Makeup Pump Suctions open (CV-1 276 and 1277).

a) IF CV-1 276 or 1277 fails to open, THEN stop associated HPI pump.

D. IF RB Spray has actuated, THEN perform the following:

1) Verify RB Spray flow throttled to maintain 1050 to 1200 gpm per train.

2) IF NaOH Tank T10 level is 16, THEN close RB Spray NaOH Addition T-10 Outlets (CV-1616 and 1617).

E. Verify RB Sump Outlets open (CV-1414 and 1415).

F. Align LPI to take suction from RB sump as follows:

1) Open RB Sump Outlets (CV-1405 and 1406).

a) IF CV-1 405 or 1406 fails to open, THEN stop associated LPI, HPI and RB Spray pumps.

(15. CONTINUED ON NEXT PAGE) 1202.012 RTi5 Rev 3-17-08

NSAS INITIAL ROISRO EXAM BANK QUESTION DATA ARKA NUCLEAR ONE UNIT I -

Rev: 0 Rev Date: 8/9/05 Source: Direct Originator: CorklPullin QID: 0609 Objective: 19 Point Value: I TUOI: Al LP-RO-ARCP Section: 4.2 Type: Generic APEs System Number: 015 System

Title:

Reactor Coolant Pump Malfunctions the Reactor Coolant Pump Malfunctions (Loss of RC

==

Description:==

Knowledge of the interrelations between Flow) and the following: RCP indicators and controls.

K/A Number: AK2.10 CFR

Reference:

41 .7/45.7 2.8 RO Select: Yes Difficulty: 3 Tier: I RO Imp:

SRO Imp: 2.8 SRO Select: Yes Taxonomy: K Group: 1 Question: RO:J 5 SRO:I Which of the following indications would require stoppin g a Reactor Coolant Pump?

A. Seal cavity pressures oscillating at 600 psi peak to peak B. Seal bleedoff temperature 160°F temperature C. Seal beedoff temperature 60°F above 1st stage seal D. Failure of one stage as indicated by zero stage DP Answer:

ature C. Seal beedoff temperature 60°F above 1st stage seal temper Notes:

section 1 of 1203.031.

Answer C is correct, this exceeds 40°F delta-T specified in monitoring frequency of an RCP.

Answers A, B and D just indicate a need for increased

References:

1203.031, Chg. 018 History:

New for 2005 RO exam, replacement question.

Selected for the RO/SRO 2010 exam.

CHANGE Y 018 PAGE 4 of 38 1203.031 REACTOR COOLANT PUMP AND MOTOR EMERGENC SECTION 1 SEAL DEGRADATION NOTE

• RCP seal stage zP is determined as follows:

1st stage zP = system pressure lower seal cavity press.

press.

2nd stage P = lower seal cavity pressure upper seal cavity heric press.

3rd stage \P = upper seal cavity pressure RB atmosp leakage in excess of design will

• Third stage seal leakage by design is 0 to 0.08 gpm. Third stage affect upper seal cavity pressure and seal bleed off flow.

4. Determine if any of the following conditions exist:

-peak

• RCP seal cavity pressure oscillations exceed 800 psi peak-to

• AP across any stage exceeds 2/3 of system pressure

• A loss of seal injection AND 2.5 gpm total seal oufflow, including seal bleedoff (excluding shaft sleeve leakage)

• Seal bleed off temp >40°F above 1st stage seal temp

• RCP seal bleed off or seal stage temp reaches 180°F ICW flow.

fj no interruption of seal injection A. IE any of the above conditions exist, the capacity of the unaffected RCP combination, using THEN reduce reactor power to within Rapid Plant Shutdown (1203.045)

B. WHEN power reduction is complete, Operation (1103.006).

THEN stop the affected RCP(s) per Reactor Coolant Pump

1) IE only 1 RCP in operation per loop, ion A (18-hour time clock).

THEN enter Tech Spec 3.4.4 Condit (continued)

CHANGE 1203.031 REACTOR COOLANT PUMP AND MOTOR EMERGENCY 018 PAGE 38 of 38 ATTACHMENT A Page 1 of I RCP PARAMETERS Seal DeciradationlSeal Failure 1 Seal

1.  !:4! of the following are criteria to SECURE the affected RCP per Section Degradation

• RCP seal cavity pressure oscillations exceed 800 psi peak-to-peak

• A? across any stage exceeds 2/3 of system pressure on a running RCP exceeds 80% of system pressure on an idle RCP.

• 2.5 gpm total seal outflow, including seal bleedoff (excluding shaft sleeve leakage),

AND a loss of seal injection

• Seal bleed off temp >40°F above 1st stage seal temp

• RCP seal bleed off or seal stage temp reaches 180°F, Q no interruption of seal injection OR ICW flow.

2. fff of the following are criteria to TRIP the affected RCP per Section 2 Seal Failure

• 10 gpm rise in RCS leak AND a change in seal cavity pressure behavior.

• RCP seal bleed off or seal stage temp reaches 200°F no change in seal injection ICW flow.

• A? across a single stage equal to RCS press, with seal bleed off flow established.

Loss of Cooling Water to RCP Motors or MotorlBearing Trouble IF Motor Bearing Temperature >190°F (1 67°F for P-32B) continues to rise, THEN SECURE the affected RCP per section 4 andlor section 5 of this procedure.

2. ANY of the following are criteria to SECURE the RCP per section 5 of this procedure:

• P32B, P32C or P32D PUMP SHAFT vibration; more than one channel 25 mils, after startup stabilization

• P32A PUMP SHAFT vibration; more than one channel 28 mils, after startup stabilization

3. ANY of the following are criteria to TRIP the affected RCP per section 4 andlor section 5 of this procedure:

• Motor current exceeds 800 amps

• Winding temperature exceeds 300°F

• Bearing temperature exceeds 225°F (176°F for P32B)

• P-32B or D MOTOR vibration; more than one channel >20 mils after startup stabilization

• P-32A orC MOTOR vibration; more than one channel >0.8 in/sec after startup stabilization

• ANY RC PUMP SHAFT vibration 29 mils after startup stabilization

DATA ARKANSAS INITIAL ROISRO EXAM BANK QUESTION NUCLEAR ONE - UNIT I Source: New Originator: S. Pullin QID: 0773 Rev: 0 Rev Date: 9/3/2009 Objective: 23 Point Value: I TUOI: ANO-1-LP-RO-DHR Section: 4.2 Type: Generic APE oval System System Number: 025 System

Title:

Loss of Residual Heat Rem apply to a Loss of implications of the following concepts as they

==

Description:==

Knowledge of the operational S during all modes of oper ation.

Residual Heat Removal System: Loss of RHR K/A Number: AKI.01 CFR

Reference:

41.8/41.10/45.3 Difficulty: 2 I RO Imp: 3.9 RO Select: Yes Tier:

SRO Select: Yes Taxonomy: a Group: I SRO Imp: 4.3 Question: RO:J 6 SRO:j6 Given:

cement.

- The RCS is drained to 374 feet for seal repla

- RCS Temperature 140 F.

- RCS pressure is 5 psig.

RCS leakage measured at 50 gpm. closed CV-I 050 Decay Heat Suction Valve has been A Decay Heat Pump has been stopped and AOP.

per 1203.028, Loss of Decay Heat Removal for these AOP, what is the preferred makeup flow path Per 1203.028, Loss of Decay Heat Removal conditions?

A. Gravity feed from the BWST.

B. Low Pressure Injection.

C. Spent Fuel Cooling Pump P-40A.

D. High Pressure Injection.

Answer:

B. Low Pressure Injection.

Notes:

use the RCS is pressurized A. Gravity feed from the BWST is incorrect beca B. Low Pressure Injection is correct. ed.

because it is the least preferred method allow C. Spent Fuel Cooling Pump P-40A is incorrect HPI could overpressurize the RCS.

D. High Pressure Injection is incorrect because

References:

1203.028 Change 21 History:

New for the RO/SRO 2010 exam.

CHANGE

[ 021 PAGE 82 of 82 I 1203.028 LOSS OF DECAY HEAT REMOVAL ATTACHMENT H

{1 and 3)

Page 1 of 1 RCS MAKEUP METHODS

1. Consider the existing plant conditions listed below:

• RCS press

• RCS level

• RCSopenorintact

• Leak rate

• DH Removal system isolated or unisolated

• Available MU flow rate

• Available time

• Available equipment NOTE

• The six RCS makeup methods are listed in order of preference.

• Each method is effective only as long as the limitations listed are met.

2. Select a makeup method below based upon existing plant conditions fl perform the applicable attachment:

REQUIRED APPLICABLE METHOD RCS OTHER LIMITATIONS ATTACHMENT PRESS Gravity Feed from 0 psig Requires BWST level >21 ft A BWST LPI <200 psig Requires operable LPI pump B RB Spray Pump <220 psig Available flow rate 1500 gpm C Borated Water Recirc <100 psig Available flow rate 180 gpm D Pump_(P-66)

Spent Fuel Cooling <45 psig Available flow rate 1000 gpm E Pump_(P-40A)

HPI N/A Can over-pressurize RCS F

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I Rev: 0 Rev Date: 11/21/00 Source: Direct Originator: D.Slusher QID: 0395 Objective: 14 Point Value: 1 TUOI: AILP-RO-NNI Section: 4.2 Type: Generic APEs System Number: 027 System

Title:

Pressurizer Pressure Control Malfunction Pressurizer Pressure Control Malfunction and the

==

Description:==

Knowledge of the interrelations between the following: Controllers and positioners.

K/A Number: AK2.03 CFR

Reference:

41.7/45.7 2.6 RO Select: Yes Difficulty: 3 Tier: I RO Imp:

SRO Imp: 2.8 SRO Select: Yes Taxonomy: C Group: 1 Question: RO:1 SRO:1 The plant is shutdown and cooled down.

RCS pressure is 220 psig.

l&C is performing calibration checks on A RPS channel.

38, Why will l&C request the Pzr Control Pressure Selector, HS-10 be placed in the Y positio n?

A. To allow remote indications to be checked during calibration.

ion of the RCS.

B. To prevent the ERV opening, causing a rapid depressurizat C. To maintain pressurizer heaters off during testing.

D. To allow the ERV low setpoint to be checked.

Answer:

ion of the RCS.

B. To prevent the ERV opening, causing a rapid depressurizat Notes:

t is in effect.

Answer [b] is correct, testing will cause ERV to open since the LTOP setpoin betwee n local and remote indications.

Answer [a] is incorrect, the selector switch does not select is incorrect, PZR heaters are in manual contro l for cooldo wn.

Answer [c] ERV at this point.

Answer [d] is incorrect, l&C verifies the setpoint, it is undesirable to operate

References:

1105.006, Chg. 010 STM 1-69, Rev. 13 History:

Exam.

Direct from regular exambank QID#5545 for 2001 RO/SRO Selected for 2005 RO exam, replacement question.

Selected for the RO/SRO 2010 exam.

PAGE: 4 of 15 PROC.IWORI( PLAN NO. PROCEDUREIWORK PLAN TITLE:

REACTOR COOLANT SYSTEM NNI CHANGE: 010-00-0 1105.006 AUTO pushbutton must be pressed to 3.13 After a SASS trip has occurred, the AUTO . Tran sfer to AUTO is inhibited if a return the channel to mismatch exists.

to bypass a channels input 3.14 The Mismatch Alarm Bypass Switch is used to SASS MISMATCH (K07-B4).

C04 selects either of two 3.15 Pressurizer Level Transmitter HS on or LTl002) as input to the compensated level signals (LT-lOOl following:

(CV-1235) H/A station

• Pressurizer Level Control Valve

• Pressurizer Lo-Lo Heater Cutoff (LS-lOOl)

• Pressurizer Hi/Hi-Lo/Lo Alarm

• Dasey Panel PZR LVL (LI-bOO) recorder and indicator on C04 are The compensated Pressurizer Level s and the Pressurizer Level totally independent of the NNI X/Y system Transmitter HS on C04.

HS on C04 selects either of two 3.16 Pressurizer Temperature Transmitter 01A or TE-10 02A) to feed the Pressurizer temperature elements (TE-10 plant Temp indicator on CO4. The signal not selected is sent to the computer.

r level signals is accomplished Temperature compensation of pressurize ms. Each level signal is compensated independent of the NNI X/Y syste re sign al at EFIC Sign al Conditioning Cabinet by a specific temperatu (C539 or C540) a SASS selector switch which 3.17 RC Pressure RPS A RPS C HS on C04 is RPS C (PT-b038) for control of selects input from RPS A (PT-bO2l) or the following systems:

• Pressurizer Heater Control.

• Pressurizer Spray Valve Control Electromatic Relief Valve Control (high pressure setpoint)

RPS A (PT-102l) is selected as the preferred In SASS ENABLE position, input.

C03 selects T-hot of loop A, 3.18 The three-position Cntrlg T-Hot HS on age of loop s A and B (marked UNIT, T-hot of ioop B, or the aver 23 in C47). The selected signal from RC Loop A/B Hot Leg T-ave TY-10 This sign al is also used by Reactor is used by the ICS for control.

C13 and the reco rder s HI alarm contact, RC Coolant T-hot (TR-1023) on Loop A/B Hot Leg (TS-1023).

STMI -69 Rev. 13 Non-Nuclear Instrumentation System 3.3.7 RCS Pressure Ten pressure transmitters monitor RCS pressure. The pressure Instruments transmitters are located on instrument racks 1 and 2 inside the reactor building. The pressure taps for the pressure transmitters are located The on the RCS hot leg piping on the vertical piping to the OTSGs.

pressure transmitters supply input to the Engine ered Safegu ards Actuation System (ESAS), Reactor Protection System (RPS),

and EFIC instrument cabinets C-539 and C-540 (suppl ies inputs to SPDS).

39 Pressure transmitters PT-1021, PT-1023, PT-1038 and PT-10 supply inputs to A, B, C, and D RPS channels, respectively. The ntial pressure transmitters that supply RPS are Rosemount differe capacitance detectors. A and C RPS channels supply pressure psig.

recorders on C04. The range of indication is 1700 psig to 2500 A and C RPS channels also supply inputs to NNIX for pressu re control.

e Pressure transmitters PT-1020, PT-1022, and PT-1040 provid input to A, B, and C ESAS analog channels, respectively. The pressure transmitters that supply ESAS are Rosemount differential ion on capacitance detectors. ESAS analog channel A supplies indicat 2500 C-I 66 (Dasey Panel). The range of the indication is 0 psig to l A also inputs to NNIX for pressu re psig. ESAS analog channe for control (ERV low setpoint at 400 psig). PT-1020 is also used Heat Remov al System . CV-over pressure protection of the Decay ck 1050 will close if RCS pressure exceeds 320 psig. The interlo than 290 psig.

allows opening CV-1050 when RCS pressure is less Pressure transmitters PT-1041 and PT-1042 provide input to EFIC instrument cabinets C-540 and C-539, respectively. The pressure transmitters that supply C-539 and C-540 are Rosemount differential capacitance detectors. These transmitters satisfS REG.

Guide 1.97 environmental qualification and Appendix R fire requirements (C-540). All outputs from C-539 and C-540 are ent buffered so that an output device failure will not affect the instrum wn), ICCM DS string. C-540 supplies outputs to SPDS (Safe Shutdo channel B, DROPS channel 2 and P1-1041 (located on C04). C-539 l A, supplies outputs to SPDS (Alternate Shutdown), ICCMDS channe DROPS channel 1, and PR 1042 (located on C04). The range of indication is 0 psig to 3000 psig. C-540 also supplies an input to ESAS analog channel 2. The input is used for over pressure if protection of the Decay Heat Removal System. CV-1 410 will close interlo ck allows openin g CV-RCS pressure exceeds 385 psig. The 1410 when RCS pressure is less than 290 psig.

RPS channels A and C supply outputs from PT-1021 and PT-3.3.8 NNIX pressure 1038 to the NNIX instrument cabinets for RCS pressure control. A control transfer relay selects which signal inputs to the NNIX pressure control channel. The relay is powered from the NNIX 120-volt AC bus. A three-position switch located on C04 controls the transfer relay. The switch positions are A, Auto, and C.

14

Non-Nuclear Instrumentation System STMI -69 Rev. 13 In the Auto position SASS controls which signal inputs into NNIX. Normally, RPS channel A is selected for input. If RPS channel A signal fails, SASS would de-energize the transfer selecting the RPS channel C input. The A and C switch positions allow the operator to select RPS channel A or C independent of SASS (signal is hard selected and SASS cannot change it). The input scheme is shown below:

125 PSI BIAS TRANSFER I RELAY ERV NNIX The SASS selected pressure signal inputs into an isolation amplifier. A 125 psi bias is input into the isolation amplifier when contact A closes. The bias is applied when either MFWP trips and reactor power is greater than 80%. This immediately opens the pressurizer spray valve to control RCS pressure. The output of the isolation amplifier is input to a difference amplifier and the ERV signal monitor.

The ERV signal monitor opens and closes the ERV in response to the input from the isolation amplifier. The signal monitor has two adjustable setpoints (a high and a low setpoint). The signal monitor opens the ERV when RCS pressure reaches 2450 psig (high) and closes the ERV when RCS pressure reaches 2395 psig (low). ESAS analog channel 1 supplies wide range pressure input to a signal monitor. The ESAS input and associated signal monitor opens the ERV when RCS pressure is 400 psig and closes the ERV when RCS pressure reaches 350 psig.

Three switches are associated with the ERV, the ERV setpoint selector switch, HS-1013, and two auto/open switches, HS-1012 and HS-1-14. HS-1013 (located on C-04) allows selecting either the high ERV setpoint (2450 psig) or the low ERV setpoint (400 psig): Hand switches HS-1012 (located in NNI cabinet C-47-2) and HS-1014 (located on C-04) allow manual opening of the ERV. Each handswitch has two positions; AUTO, and OPEN.

15

ANSAS INITIAL ROISRO EXAM BANK QUESTION DATA ARK NUCLEAR ONE UNIT I -

Rev Date: 9/3/2009 Source: New Originator: S. Pullin QID: 0582 Rev: 0 Objective: 26 Point Value: 1 TUOI: Al LP-RO-EFIC Section: 4.1 Type: Generic EPEs (ATWS)

System Number: 029 System

Title:

Anticipated Transient Without SCRAM ing as they apply to the ATWS: Reactor trip alarm.

==

Description:==

Ability to determine or interpret the follow KIA Number: EA2.02 CFR

Reference:

43.5 /45.13 RO Select: Yes Difficulty: 4 Tier: 1 RO Imp: 4.2 SRO Imp: 4.4 SRO Select: Yes Taxonomy: An Group: I Question: RO: 8 SRO:[

Given:

- Plant startup is in progress.

- Reactor power is 20%.

- Total Main FW flow is 1.6 x e 6 Ibm/hr.

- Generator load is 180 Mwe.

Subsequently the following indications are observed:

- Reactor power dropping rapidly,

- Regulating groups In-Limit lights ON,

- Safety groups Out-Limit lights ON.

- Turbine Generator Lockout alarm is in,

- EFW actuated on both trains.

annunciator, could cause the above indications?

Which of the following annunciators, and reasons for the MFW pump has tripped causing a reactor trip with A. K08-A3 REACTOR TRIP because the in-service power >9%.

loss of transformer X8 has tripped the Regulating B. K08-F2 CRD MOTOR POWER FAILURE because a Groups.

s Nl-501 and NI-502 were not calibrated within 3% of C. K08-A5 AMSAC TRIP because both Gamma Metric heat balance as required.

atory trip for Turbine has not been reset.

D. K08-A3 REACTOR TRIP because the RPS anticip Answer:

s Nl-501 and NI-502 were not calibrated within 3% of C. K08-A5 AMSAC TRIP because both Gamma Metric heat balance as required.

Notes:

all of the control rods to insert not just the regulating A. Is incorrect because a reactor trip would have caused groups.

of the AC power supplies to the rods and no rods B. Is incorrect because a loss of X8 would only lose one would trip.

given feedwater flow, an AMSAC Trip would be C. Is correct, if gamma metrics indicated >45% with the initiated.

all of the control rods to insert not just the regulating D. Is incorrect because a reactor trip would have caused groups.

References:

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

11 02.002 Change 082 STM 1-59 Rev. 1 History:

New for the RO/SRO 2010 exam.

Stage 1Inch Warmup 18.20 Open moisture separatorreheater 2nd Valves on Cli.

• CV6808

• CV6837

• CV6814

• CV6842 (127 MWe) 18.21 Adjust icload limit setpoint to 12.7%

and NI502 (SPDS 18.22 Verify that NI501 (SPDS point NI1LP)

Line ar Power Instruments are point NI2LP) Gamma etrics within +3 of hsat balan ce power.

able 18.22.1 IF instruments are outside the allow diffe renc es, nts per THEN have I&C Dept calibrate instrume nels Test (130 4.05 5).

Source Range Chan 18.23 IF available, n, PLANT MODE THEN using Plant Computer UTILITY scree selec t Powe r Ops mode to enable ASSIGNMENT (PMA),

computer alarms for the power ops mode.

18.24 Verify plant at -25% power.

Adjust Low Load Limit setpoint to 20% (200 MWe).

18.24.1 18.24.2 WHEN gland steam condenser zP is >2.8 (per PDIC2905 on C02),

THEN place CV-2906 handswitch in AUTO

Diverse Reactor Overpressure Prevention System STM 1-59 Rev I Foxboro isolators NY-501G and NY-502G. NI-501 inputs reactor power to DROPS channel 1 and NI-502 inputs to channel 2.

The Loop A and B Main Feedwater Flow signals from the flow transmitters are provided to DROPS through non-IE Bailey voltage buffers in NNI cabinets C47-4 and C48-7.

Refer to table below for MFW flow transmitters associated with each DROPS channels.

I Channel 1 Channel 2 Loop A MFW flow PDT-2628 Loop A MFW flow PDT-2627 Loop B MFW flow PDT-2677 Loop B MFW flow PDT-2678 he AMSAC turbine trip and EFW initiation signals are generated when MFW flow is less than 5% of 6.0 x 106 LB/hn rated flow in both loops and when reactor power is greater thin 45070.

(Refer to Figure 59.04)

The turbine trip signals are summed in the existing turbine trip circuitry and upon receipt of both DROPS channels AMSAC signals, the auto-stop oil trip solenoid and the auto-stop back-up oil trip solenoid will be energized. Energizing either of the auto-stop oil trip solenoids will trip the turbine.

The DROPS AMSAC signal to trip the main turbine is accomplished by two relays in the turbine trip circuitry. The relay contacts are wired in series to form the 2 out of 2 coincidence logic to actuate the Auto-Stop oil trip and backup trip solenoids which trip the main turbine. The power for the coil and contacts on these relays is supplied from the 125 vdc bus in the turbine trip circuitry.

DROPS Turbine trip confirmation is provided by the two trip contacts wired in series which provide a 125 vdc trip confirm signal back to DROPS. For additional information on the turbine trip circuitry refer to STM 1-24 Main Turb & Controls.

The DROPS AMSAC subsystem provides an energize to trip signal to initiate Emergency Feedwater. EFW actuation signals from DROPS inputs to EFIC channels A and D Initiate modules.

DROPS channel 1 trip signal inputs to EFIC channel A and channel 2 trip signal inputs to EFIC channel D. Initiation of EFIC Channels A and D will result in full EFW actuation.. Since EFIC trips actuate on loss of input signal or loss of EFIC power to the initiate modules the signals from DROPS are inverted by a Anticipatory Trip Initiation relay in the EFIC cabinets. The 1 E relay coil and contacts are powered by the associated EFIC cabinet 28 vdc power supply. This relay is normally dc-energized and its associated contacts closed. The AMSAC trip signal will energize the anticipatory trip initiation relay causing it contacts to open actuating EFW utilizing the normal initiation process. The non-1E AMSAC signal interfaces with the 1E portion of EFIC through photo-optic Ii

Diverse Reactor Overpressure Prevention System STM 1-59 Rev I 2.4.1 DROPS Testing Periodic testing of DROPS is required and will occur during normal plant operation. Due to DROPS interaction with other systems it is important to recognize equipment failures that can potentially cause adverse affects during testing. When performing a DROPS channel DSS or AMSAC test the trip function will cause the associated actuation features relays or contacts to change state.

For example: en performing DROPS channel I DSS subsystem est the A or main power to CRDgroups , , 7 an the Aux. bus gate drives contacts will o en along with energizing one of the two relays the turbine trip circuitly. When problems exist in either the CRD, EFIC, Gamma Metrics or the turbine trip circuitry DROPS testing should not be performed.

DROPS testing is performed by utilizing the common test jacks and test enable push-buttons located on the control module. The system test will remove the process inputs and internally simulate the inputs. The setpoints and the trip signals can be monitored at the front panel for verification during testing. Since the DROPS is a 2 out of 2 logic system the channel not being tested and the trip feature not being tested will be placed bypass. Placing the channel in bypass requires the DSS and AMSAC bypass switches to be placed in the bypass position. This will disable the associated trip function and prevent system actuation during testing.

When the associated DROPS channel being tested is placed in test the DSS I AMSAC in test annunciator will alarm alerting the operator of this condition.

For additional information refer back to section 2.1.5 Control Module and section covering the DSS and AMSAC test enable buttons.

2.5 Annunciators This section will cover the annunciators associated with the DROPS system. For additional information refer to 1203.0 13G Annunciator Corrective Actions.

The annunciators associated with DROPS are systematically arranged and grouped in panel K08 located on panel C-13 in the control room.

The following alarms are provided for the DROPS system:

DSS Trip: K08-A5 will alarm when RCS pressure sensed by PT-1041 and PT- 1042 indicates> 2430 psig and both DROPS channels have received an DSS trip confirm signal.

AMSAC Trip: K08-B5 will alarm when Reactor power is greater than 45% and MFW flow indicates less than 15% of total flow (.9 x 106 lbmlhr) and both DROPS channels have received an AMSAC trip confirm signal.

DSS I AMSAC Trouble: K08-C5 will alarm when an abnormal condition exist associated with the subsystems of DROPS. The conditions that can cause the DSS or AMSAC trouble alarm are listed below.

13

Diver lactor Overpressure Prevention System STM 1-59 Rev I PRESSURE (PT-1042)

AMSAC-1 AMSAC-2 DSS-1 DSS-2 (AMSAC RELAYS C-20) (DSS RELAYS C-62)

TURBINE TRIP REACTOR TRIP TURBINE TRIP (TRIP 5,6,7 & AUX BUS)

FIGURE 59.01: BLOCK DIAGRAM OF DROPS 18

INITIAL RO!SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I Rev: 0 Rev Date: 11/8/00 Source: Direct Originator: J.Cork QID: 0364 TUOI: Al LP-RO-EOPO6 Objective: I Point Value: I Section: 4.1 Type: Generic EPEs System Number: 038 System

Title:

Steam Generator Tube Rupture

==

Description:==

Knowledge of EOP mitigation strategies K!A Number: 2.4.6 CFR

Reference:

41.10/43.5/45.13 1 RO Imp: 3.7 RO Select: Yes Difficulty: 4 Tier:

Group: I SRO Imp: 4.7 SRO Select: Yes Taxonomy: An Question: ROI 9 SRO:

After a reactor trip, the following indications are observed:

- Makeup Tank level has lost 5 inches in the last 5 minutes

- RB and Aux. Bldg. Sump levels are stable

- A EFIC level is 35 and rising

- B EFIC level is 31 and stable

- A MFW Flow is 0.1 mlb/hr

- B MEW Flow is 0.3 mlb/hr Which of the following actions would be required to minimize the threat of a potential radioactive release to the public?

A. Initiate HPI per RT-2 B. Cooldown and isolate the B SG C. Cooldown and isolate the A SG D. Commence a rapid RCS cooldown at 240 °F/hr Answer:

C. Cooldown and isolate the A SG Notes:

cooldown should be Answer [C] is correct, the SG level parameters indicate a rupture on the A SG and a a secondary safety on commenced to reduce RCS temperature to <500 F to minimize the possibility of lifting the A SG.

within the capacity of

[a] is incorrect, the leak size is about 30 gpm (30.86 gaIIin. x 5 in./5 mm.). This is normal makeup.

[b] is incorrect, a cooldown and isolation is required but not on this SG.

SG is imminent.

[d] is incorrect, a rapid cooldown at this rate is not required until overfilling of ruptured

References:

1202.006, Chg. 11 History:

Created for 2001 RO/SRO Exam.

Selected for 2002 ROISRO exam.

Selected for 2005 Jon Gray RO re-exam.

Selected for 2010 RO/SRO Exam

CHANGE 011 PAGE 3 of 42 1202.006 TUBE RUPTURE INSTRUCTIONS CONTINGENCY ACTIONS

5. Begin controlled plant shutdown at 5% per minute.

6. Determine bad SG using one or more of the following:

• OTSG N-16 monitors:

SGA SGB RI-2691 Rl-2692

• SGTR display on SPDS.

• Plant Monitoring System Alarms.

• Steam Line High Range RAD Monitors (may be inconclusive due to insufficient shielding between MS lines):

SGA SGB RI-2682 RI-2681

• Local steam line radiation survey.

• Nuclear Chemistry sample.

• At low FW flow rates:

Higher than expected SG level Lower than expected FW flow rate Lower than expected MFW pump speed

7. Perform Control of Secondary System Contamination (1203.014) in conjunction with this procedure.

8. WHEN bad SG is known, THEN place bad SG EFW pump Turbine (K3)

Steam Supply valve in MANUAL !iQ. close:

SGA SGB CV-2667 CV-2617

CHANGE 1202.006 TUBE RUPTURE 011 PAGE 14 of 42 INSTRUCTIONS CONTINGENCY ACTIONS

18. E emergency cooldown rate is required OR RCS T-hot is 500°F, THEN establish RCS cooldown rate of 100°FIhr as follows:

A. For good SG, place TURB BYP valves in A. IF TURB BYP valves are not available, HAND THEN operate ATM Dump Control System AND for good SG in HAND to maintain adjust to maintain cooldown rate 100°FIhr. cooldown rate 100°F/hr.

SGA SOB ATM CV-2676 DUMP ISOL CV-2619 ATM CV-2668 DUMP CNTRL CV-2618

1) j both SOs are bad, THEN steam both SOs.

B. When RCS press is <1700 psig, THEN bypass ESAS.

C. jf only one SG is bad, C. IF both SGs are bad, THEN steam bad SG only as necessary to THEN steam both SGs.

maintain:

• MSSVs closed

• SG press:

990 psig if using TURB BYP valves 1 040 psig if using ATM Dump Control system

• SO level 410.

• SC Tube-to Shell L\T 100°F (tubes colder).

• Desired cooldown rate if good SO TBV or ADV is full open.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0551 Rev: 0 Rev Date: 3-30-05 Source: Direct Originator: J.Cork TUOI: A1LP-RO-EOPO3 Objective: 10 Point Value: I Section: 4.2 Type: Generic APEs System Number: 040 System

Title:

Steam Line Rupture Steam

==

Description:==

Knowledge of the operational implications of the following concepts as they apply to the Line Rupture: Consequence of PTS.

K/A Number: AKI .01 CFR

Reference:

41.8 /41.10/45.3 Tier: I RO Imp: 4.1 RO Select: Yes Difficulty: 3 Group: I SRO Imp: 4.4 SRO Select: Yes Taxonomy: C Question: RO:1 10 SRO:j 0 Which of the following would invoke Pressurized Thermal Shock (PTS) limits during a Steam Line Rupture?

A. HPI on with all RCPs off

5. RCS cool down rate 105°F/hr with Tcold 360°F C. RCS cool down rate 55°F/hr with Tcold 31 0°F D. SG Tube to shell DT 150°F tubes colder Answer:

A. HPI on with all RCPs off Notes:

Answer A is correct per RT-1 4.

Answer 8 is incorrect, cooldown rate is >100°F/hr but Tcold >355°F.

Answer C is incorrect, cooldown rate is >50°F/hr but Tcold >300°F.

Answer D is incorrect, this is a limit but not a PTS limit.

References:

1202.012, Chg. 8 History:

New for 2005 RO exam.

Selected for 2010 RO/SRO exam

T CHANGE 1202.012 REPETITIVE TASKS 008 PAGE 34 of 50 Page 1 of 3 NOTE

. PTS limits apply if y of the following has occurred:

HPI on with all RCPs off RCS C/D rate> 1 00°F/hr with Tcold < 355° F RCS C/D rate> 50°F/hr with Tcold < 300°F

. Once invoked, PTS limits apply until an evaluation is performed to allow normal press control.

. When PTS limits are invoked OR SGTR is in progress, PZR cooldown rate limits apply.

14. Control RCS press within limits of Figure 3.

A. !E PTS limits apply or RCS leak exists, THEN maintain RCS press j within limits of Figure 3.

B. IF RCS press is controlled AND will be reduced below 1650 psig, THEN bypass ESAS as RCS press drops below 1700 psig.

C. IF PZR steam space leak exists, THEN limit RCS press as PZR goes solid by one or more of the following:

1) Throttle makeup flow.

2) IF SCM is adequate, THEN throttle HPI flow by performing the following:

a.) Verify both HPI RECIRC valves (CV-1 300 and 1301) open.

b.) Throttle HPI.

3) Raise Letdown flow.

a) I ESAS has actuated, THEN unless fuel damage or RCS to ICW leak is suspected, restore Letdown flow (RT 13).

4) Verify ERV Isolation open (CV-1 000) fQ cycle ERV (PSV-1 000).

(14. CONTINUED ON NEXT PAGE) 1202.012 RT14 Rev 3-17-08

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0774 Rev: 0 Rev Date: 9/4/2009 Source: Modified Originator: S Pullin TUOI: Al LP-RO-EOPO2 Objective: 8 Point Value: 1 Section: 4.2 Type: Generic APEs System Number: 054 System

Title:

Loss of Main Feedwater (MFVV)

Loss of Main Feeciwater

==

Description:==

Ability to determine and interpret the following as they apply to the (MFV: AFW adjustments needed to maintain proper T-ave and S/G level.

K/A Number: AA2.06 CFR

Reference:

43.5/45.13 I RO Imp: 4.0 RO Select: Yes Difficulty: 3 Tier:

Group: 1 SRO Imp: 4.3 SRO Select: Yes Taxonomy: C Question: RO:j I SRO:j A reactor trip has occurr ed from 100% power due to a loss of both MEW Pumps.

The following conditions have existed for three minutes:

- CET temperature = 580 degrees F.

- RCS pressure = 1600 psig.

Which of the following operator actions will be performed?

A. Trip all running RCPs.

B. Verify EFW flow to each Steam Generator is -320 gpm.

C. Verify Reflux Boiling setpoint is selected on both EFIC trains.

D. Verify EFW in hand and flow to each Steam Generator is -570 gpm.

Answer:

C. Verify Reflux Boiling setpoint is selected on both EFIC trains.

Notes:

s expired without A. incorrect, this would be done for loss of subcooling margin but only if <2 minute had tripping the RCPs.

adequate and the B. Incorrect this is done for loss of subcooling margin but only if EFW flow is less than value given is similar but less than the minimum flow rate of greater than or equal to 340 gpm.

be selected in this C. Correct since subcooling margin is lost and the Reflux Boiling setpoint is required to situation.

D. Incorrect, this would be done if only one SG was available.

References:

1202.012 Change 008, RT-5 History:

Modified from QID 368.

Selected for the RO/SRO 2010 exam.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0368 Rev: 1 Rev Date: 8/8/05 Source: Direct Originator: J.Cork TUOI: Al LP-RO-EOPO2 Objective: 8 Point Value: I Section: 4.1 Type: Generic EPEs System Number: 009 System

Title:

Small Break LOCA

==

Description:==

Knowledge of the operational implications of the following concepts as they apply to the small break LOCA: Natural circulation and cooling, including reflux boiling.

K/A Number: EKI .01 CFR

Reference:

41.8 /41.10 / 45.3 Tier: 1 RO Imp: 4.2 RO Select: No Difficulty: 3 Group: 1 SRO imp: 4.7 SRO Select: No Taxonomy: An Question: RO:1 SRO: I A reactor trip has occurred from 100% power.

The following conditions have existed for three minutes: , e 1 4

- RCS temperature = 590 degrees F.

- RCS pressure = 1700 psig.

Which of the following operator actions will be performed?

A. Trip all running RCPs.

B. Verify EFW flow to each Steam Generator is 320 gpm.

C. Verify Reflux Boiling setpoint is selected on both EFIC trains.

D. Go to Overheating EOP, 1202.004.

Answer:

C. Venfy Reflux Boiling setpoint is selected on both EFIC trains.

Notes:

Answer [c] is correct since subcooling margin is lost and the Reflux Boiling setpoint is required to be selected in this situation.

Answer [a] is incorrect, this would be done for loss of subcooling margin but only if <2 minutes had expired without tripping the RCP5.

Answer [b] is incorrect, this is done for loss of subcooling margin but only if EFW flow is less than adequate and the value given is similar but less than the minimum flow rate of greater than or equal to 340 gpm.

Answer [d] is incorrect, this would not be done since the entry conditions for Overheating have not been met and loss of subcooling margin

References:

1202.012, Chg. 004-03-0, RT-5 History:

Direct from regular exambank QID 3030.

Selected for use in 2002 SRO exam.

Modified for use in 2005 RO exam, replacement question.

CHANGE REPETITIVE TASKS 008 PAGE 11 of 50 1202.012 Page 1 of 3

5. Verify proper EFW actuation and control:

A. Verify EFW actuation indicated on Bus I and 2 of Trains A and B on C09.

h applicable B. Verify at least one EFW pump (P7A or B) running with flow to SG(s) throug EFW CNTRL valve(s).

SGA SGB CV-2645 P7A CV-2647 CV-2646 P7B CV-2648 Table I EFIC Automatic Level Control Setpoints Condition Level Band Automatic Fill Rate Any RCP running 20 to 40 No fill rate limit All RCPs off AND Natural Circ selected 300 to 340 2 to 8/min All RCPs offf Reflux Boiling selected 370 to 410 2 to 8/min C. if SCM is adequate, THEN perform the following:

1) Select Reflux Boiling setpoint.

NOTE Table 2 contains examples of less than adequate/excessive EFW flow.

2) Verify EFW CNTRL valves operate to establish and maintain SG levels 370 to 410.

a) j SGs are available, THEN verify SG level rising and tracking EFIC setpoint until 370 to 410 is established.

(1) IF EFW flow is less than adequate, THEN control EFW to applicable SG in HAND to maintain 340 gpm to applicable SC until level is 370 to 410.

(2) j[ EFW flow is excessive AND

> 340 gpm to either SG, THEN throttle EFW to applicable SG in HAND to limit SG depressurization.

Do not throttle below 340 gpm on either SG until SG level is 370 to 410.

b) IE only one SC is available, THEN feed available SG in HAND at 570 gpm until SG level is 370 to 410.

(5. CONTINUED ON NEXT PAGE) 1202.012 RT-5 U Rev 3-16-06

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I Rev: 0 Rev Date: 12/8/2003 Source: Direct Originator: NRC QID: 0496 Objective: 29 Point Value: 1 TUOI: ELP-NLO-ELECI Section: 4.1 Type: Generic EPEs System Number: 055 System

Title:

Station Blackout they apply to a Station Blackout: Battery, when

==

Description:==

Ability to operate and monitor the following as approaching fully discharged.

K/A Number: EAI .05 CFR

Reference:

41.7 / 45.5 /45.6 3.3 RO Select: Yes Difficulty: 3 Tier: I RO Imp:

I SRO Imp: 3.6 SRO Select: Yes Taxonomy: C Group:

Question: RO:J 12 SRO:112 Unit 1 has been in a station black-out for 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> with battery bank D06 supplying bus D02 with power without a battery charger online for this entire time.

following statements describes the batterys If the equipment on bus D02 does NOT change, which one of the ed?

discharge rate (expressed as amperage) as the battery is expend battery capacity A. The battery amperage will be fairly constant until the design is exhausted.

capacity is exhausted.

B. The battery amperage will drop steadily until the design battery capacity is exhausted.

C. The battery amperage will rise steadily until the design battery battery capacity is D. The battery amperage will be fairly constant until the design exhausted and then will rapidly drop.

Answer:

capacity is exhausted.

C. The battery amperage will rise steadily until the design battery Notes:

e will drop and current (battery P=lE; As the battery discharges under a constant load, battery voltag amperage) will rise.

References:

ELP-NLO-ELEC I History:

Developed by NRC.

Used on 2004 RO/SRO Exam.

Selected for 2005 Jon Gray RO re-exam.

Selected for the RO/SRO 2010 exam.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0366 Rev: 0 Rev Date: 1/8/00 Source: Direct Originator: J.Cork TUOI: AILP-RO-ESAS Objective: 5 Point Value: 1 Section: 4.2 Type: Generic APEs System Number: 056 System

Title:

Loss of Offsite Power apply to the Loss of Offsite

==

Description:==

Knowledge of the reasons for the following responses as they Power: Order and time to initiation of power for the load sequen cer.

K/A Number: AK3.01 CFR

Reference:

41.5, 41.10 / 45.6 / 45.13 I RO Imp: 3.5 RO Select: Yes Difficulty: 2 Tier:

Group: 1 SRO Imp: 3.9 SRO Select: Yes Taxonomy: K Question: RO: 3 SRO:j actuation of the even channels.

An electrical storm has caused a Degraded Power situation with a spurious ES In which order will the following ES components be started automatically?

A. SW pump, HPI pump, LPI pump, RB Spray pump B. HPI pump, SW pump, LPI pump, RB Spray pump C. SW pump, HPI pump, RB Spray pump, LPI pump D. HPI pump, LPI pump, SW pump, RB Spray pump Answer:

D. HPI pump, LPI pump, SW pump, RB Spray pump Notes:

actuation.

Answer [d] lists the correct order of load sequence with loss of offsite power and ES The others are incorrect sequences of the correct components.

References:

1305.006, Chg. 030 History:

Created for 2001 RO/SRO Exam.

Selected for 2005 Jon Gray RO re-exam.

Selected for the 2010 RO/SRO exam

i PROC.IWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: I PAGE: 169 of 170 1305.006 INTEGRATED ES SYSTEM TEST CHANGE: 030 SUPPLEMENT 1 Page 72 of 73 LIMITING IS DATA WITHIN TEST MEASURED NORMAL RANGE FOR LIMITING RANGE?

QUANTITY INSTRUMENT VALUES RANGE OPERABILITY (CIRCLE YES OR NO)

Loop II SW Control Attachment 4 logic test N/A N/A N/A satisfactory YES NO ES Even Channels Control Attachment 6 logic test N/A N/A N/A satisfactory YES NO l hour

@2600-2750 KW AND DG2 temperatures loaded Clock Mm N/A stable YES NO DG1 At rated (CH 2) DAS Data Sec. N/A speed and YES NO DG2 from ESAS voltage in (CH 2) Actuation Sec. N/A l5 sec. YES NO Shed on loss N/A N/A of power YES NO Resequence N/A N/A on buses YES NO HPI pump Even DAS Data Sec. N/A 4.7-5.3 sec YES NO channels from Loss LPI pump ES load of Power Sec. N/A 9.6-10.4 sec YES NO sequencing SW pump Sec. N/A 14.4-15.6 sec YES NO RBS pump Sec. N/A 33.6-36.4 sec YES NO VSF lC Sec. N/A 48-52 sec YES NO VSF-lD -

Sec. N/A 48-52 sec YES NO 5.2 IF No is circled in any space above, THEN declare the affected component inoperable, immediately notify the Shift Manager, write a Condition Report and initiate corrective action.

PERFORMED BY OPERATOR DATE/TIME OPERATOR DATE/TIME OPERATOR DATE/TIME OPERATOR DATE/TIME

INITIAL ROISRO EXAM BANK QUEST ION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0624 Rev: 0 Rev Date: 11/2/05 Source: Direct Originator: J.Cork TUOI: AILP-RO-NNI Objective: 7 Point Value: I Section: 4.2 Type: Generic APEs System Number: 057 System

Title:

Loss of Vital AC Electrical Instr ument Bus

Description:

Ability to determine and interpret the following as they apply to the Loss of Vita Bus: S/G pressure and level meters. l AC Instrument K/A Number: AA2.05 CFR

Reference:

43.5 / 45.13 Tier: 1 RO Imp: 3.5 RO Select: Yes Difficulty: 3 Group: 1 SRO Imp: 3.8 SRO Select: Yes Taxonomy: C Question:

RO:114 SRO:I 14 What would be the effect on the SG pressure and level instruments on C03, if a loss of the RS-1 bus occurred?

A. Instrument power would automatically be transferred to YO-2 by the ABT, SG pressure instruments would not be effected. and level B. The NNI-X Si and S2 switches would open and SASS would transfer to NNI-Y, SG press instruments would fail to mid scale. ure and level C. Both NNI-X SG pressure indicators would fail so ICS could not generate a BTU limit alarm.

D. Instrument power would automatically be transferred to YO-1 by the ABT, SG pressure instruments would fail low. and level Answer:

A. Instrument power would automatically be transferred to YO-2 by the ABT, SG pressure instruments would not be effected. and level Notes:

A is correct, a loss of RS-i would simply caus e NNI-X to be powered from YO-2, -24vDC auctioneered and instrument power would logic power is transfer by the ABT within 0.5 seconds no effec B is incorrect, it would take a loss of both RS-1 t on instruments.

and YO-2 to cause the SI and S2 switches to C is incorrect, although SG pressure does open.

input to the BTU limit alarm, the NNI-X SG would not be failed due to the power transfer pressure indicators to YO-2.

D is incorrect, the alternate power to NNI

-Y is from YO-I.

References:

STM 1-69, Rev. 13 History:

New for 2005 RO re-exam.

Selected for the 2010 RO/SRO exam

Non-Nuclear Instrumentation System STMI -69 Rev. 13 3.1.3 Controls and The following controls and indications are associated with the indications SASS system.

SASS module front panel controls and indications:

• Reset switch Switch is located under the front cover and may be used to reset (initialize the computer and start the data gathering).

• Auto push-button When depressed, the associated SASS channel will return to automatic if the NNIX and NNIY signals are within the mismatch setpoint.

• Test toggle switch The test toggle switch inserts a +5 VDC signal into the signal conditioning board of the associated channel. The SASS computer will see the signal step change and generate a mismatch and thp indication. The SASS transfer function is blocked when the toggle switch is taken to either the X or Y position.

• Auto Indicator Green LED indicator shows the SASS system is capable of initiating a signal transfer. Indicator should normally be on.

• Mismatch Indicator Amber LED indicator lights when the computer detects a mismatch (NNIX and NNIY signals exceeds the mismatch setpoint).

• Trip X Indicator Red LED indicator lights when the NNIX signal has failed. The SASS channel should have initiated a transfer to the NNIY channel.

• Trip Y Indicator Red LED indicator lights when the NNIY signal has failed. The SASS channel should have initiated a transfer to the NNIX channel (for Tc only).

3.2 NNI Power Supplies The NNI power supplies provide the power necessary for the NNI system operation. A 120-volt AC bus supplies power to relays, transmitters, and generally components not inside the NNI cabinets.

+/-24 volt DC buses supply power to the internal electronic circuits which process the signals. NNIX is supplied from RS-l and Y-02.

NNTY is supplied from RS-4 and Y-01.

The vital (RS) and instrument (Y) buses each supply one positive 24 volt DC and one negative 24 volt DC power supply through circuit 7

Non-Nuclear Instrumentation System STMI -69 Rev. 13 breakers S-i and S-2. S-i and S-2 are located in the NNI cabinets.

Diodes auctioneer the outputs of the two positive 24 volt DC and two negative 24 volt DC power supplies. Therefore, a loss of either of the power sources A (RS or instrument power) will not cause a loss of the

+/-24 volt DC power.

The power supply monitor monitors the output of the +/-24-volt DC power supplies. The power supply monitor will initiate an annunciator alarm when power supply voltage is abnormal. The power supply monitor will also cause the S-i and S-2 switches to open when either the positive or negative 24 volt DC power is lost (loss > >.5 seconds). The S-i and S-2 switches also provide overcurrent protection for the power supply and NNI components.

An automatic bus transfer switch (ABT) is fed from the vital and instrument buses (vital is the normal source). The ABT will transfer to the instrument AC when the vital power source is lost. The ABT will shift back to the vital source 10 minutes after vital power is restored to the ABT.

3.3 Reactor Coolant System Instrumentation 3.3.1 RCS Hot Leg (Th) Three dual element RTDs are located on each RCS hot leg on the vertical piping at the outlet of the reactor vessel. Hot leg RTD locations are as follows:

RCS A Loop RCS B Loop TE-lOli TE-1039 TE-1012 TE-1040 TE-1013 TE-1041 TE-1014 TE-1042 TE-ilil TE-1139 TE-1112 TE-1140 TE-i 139 and TE-1 112 provides an input into C-540B. TE-i lii inputs into C-539B. The temperature elements input into an RTD bridge that converts the resistance of the RTD to a corresponding output voltage. The output then goes to isolation amplifiers. The isolation amplifiers supply outputs to the SPDS computer and hot leg temperature indication on C03. The range of temperature indication is 50 °F to 700°F.

Non-Nuclear Instrumentation System STMI-69 Rev. 13 NNIXBUS RS -I C-47 YO-2 I)

POWER SUPPLY MONITOR )

MONITORS OUTPUT OF + LAJ...4J SOLA TRANSFORMER AND -24 VDC POWER SUPPLIES. HELPS SMOOTH OUT UNREGUlATED AC INPUT A LOSS OF EITHER VOLTAGE FOR .5 SECONDS CAUSES BOTH ALTERNATE S-i AND S-2 TO TRIP.

NORMAL TRIP S-2 TRIP S-i BUS TRANSFER SWITCH NORMAL SEEKING WITH A 10 MINUTE TIME DELAY. POWER SEEKING AT ALL TIMES. TRANSFER TIME .5 SECONDS. CAN BE MANUALLY TRANSFERRED INSIDE CABINET.

Y02 AG I BKR37 110 VAC FAN INSTRUMENT POWER POWER I AUC11ONEER I

+24VDC -24VDC jQJ: S-i AND S-2 ALSO ACTS AS CIRCUIT BREAKERS IN THE EVENT OF AN OVERLOAD.

jQ]: S-i AND S-2 MUST BE MANUALLY RESET FOLLOWING A TRIP.

FIGURE 69.01: NNI X BUS 56

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0187 Rev: I Rev Date: 4/25/2002 Source: Direct Originator: S.PuIIin TUOI: Al LP-RO-AOP Objective: 4.5 Point Value: 1 Section: 4.2 Type: Generic APE System Number: 058 System

Title:

Loss of DC Power

==

Description:==

Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power: Battery charger equipment and instrumentation.

K/A Number: AKI.0l CFR

Reference:

41.8/41.10/45.3 Tier: I RO Imp: 2.8 RO Select: Yes Difficulty: 3 Group: 1 SRO Imp: 3.1 SRO Select: Yes Taxonomy: C Question: RO:1 SRO:I 5 Given the following indications at 100% power:

- Annunciator D02 UNDERVOLTAGE (K0l-A8) in alarm.

- Annunciator D02 TROUBLE (K01-D8) in alarm.

- Annunciator D02 CHARGER TROUBLE (K01-E8) in alarm.

- The reactor has tripped.

- The turbine trip solenoid light is on.

- Breaker position lights on the RIGHT side of ClO are off.

What are the actions required of the CBOT?

A. Trip the main generator output breakers.

B. Transfer DII to emergency supply DOl.

C. Trip all RCPs.

D. Transfer D21 to emergency supply DOl.

Answer:

D. Transfer D21 to emergency supply DOl.

Notes:

[d] is correct per 1203.036 as the conditions are indicative of a loss of D02.

[a] and [b] are incorrect due to this a loss of D02 not DOl these are actions for the loss of DOl.

[c] is incorrect due to we have not loss seal injection and seal cooling, this is an action in this procedure section if both of the before mentioned system functions are lost

References:

1203.036, Chg. 08 History:

Developed for use in 98 RO Re-exam Selected for use in 2002 RO/SRO exam, revised slightly.

Selected for 2005 Jon Gray RO re-exam.

Selected for the 2010 RO/SRO exam

I PROCJWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: I PAGE: 13 of 44 1203.036 LOSS OF 125V DC CHANGE: 008 SECTION 2 -- LOSS OF D02 1.0 SYMPTOMS 1.1 Low DC Voltage Alarms:

• D02 UNDERVOLTAGE (IC0l-A8)

• D21 LOSS OF VOLTAGE (K0l-B8)

• RA2 LOSS OF VOLTAGE (K0l-C8)

• D02 TROUBLE (K0l-D8)

• H2 DC CONTROL POWER OFF (K02-B5)

• A2 DC CONTROL POWER OFF (K02-C7)

• A4 DC CONTROL POWER OFF (K02-D7)

• EOS SYSTEM TROUBLE (K04-C5) 1.2 Loss of breaker indicator lights for plant buses on right side of dO and switchyard mimic on dO.

2.0 IMMEDIATE ACTION NONE 3.0 FOLLOW-UP ACTIONS 3.1 IF reactor trips, THEN perform Emergency Operating Procedures (1202.XXX) in conjunction with this procedure.

3.2 IF RCP seal injection ND seal cooling are BOTH lost, THEN trip all running RCPs.

3.2.1 Isolate RCP Seal Bleedoff (Normal) by closing the following valves:

• CV-l270

• CV-127l

• CV-l272

• CV-1273 3.2.2 Place RCP Seal Bleedoff (Alternate Path to Quench Tank) controls in CLOSE:

• SV-l270

• SV-l27l

• SV-1272

• SV-1273

PROC JWORK PLAN NO. PROCEDUREIWORK PLAN TITLE:

I PAGE: 14 of 44 1203.036 LOSS OF 125V DC CHANGE: 008 SECTION 2 -- LOSS OF 002 (continued) 3.3 Isolate letdown by closing Letdown Cooler E-29A/B Outlets:

• CV-l214

• CV-l216 3.4 At ClO, transfer 021 to EMERG SUPPLY 001.

3.4.1 IF transfer of 021 is NOT successful, THEN attempt local transfer of 021 to 001, while continuing.

3.5 Notify SM to implement Emergency Action Level Classification (1903.010) 3.6 IF reactor is NOT tripped, THEN GO TO step 6.0.

3.7 IF transfer of 021 is successful, THEN GO TO step 4 .0.

3.8 IF transfer of 021 is NOT successful, THEN perform the following:

3.8.1 Dispatch an operator to perform Attachment 2, while continuing.

3.8.2 GO TO step 5.0.

4.0 IF transfer of 021 is successful, THEN perform the following:

4.1 Verify Condenser Vacuum Pump (C-5A OR C-SB) running.

4.2 IF OP HPI pump is tripped, THEN restart as follows:

4.2.1 Place the following in HAND AND close:

• RC Pump Seals Total INJ Flow (CV-1207)

• PZR Level Control (CV1235) 4.2.2 Verify RCP Seal INJ Block (CV-1206) closed.

4.2.3 Start Aux Lube Oil pump for OP HPI pump.

4.2.4 Start the OP HPI pump.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0281 Rev: 0 Rev Date: 9-3-99 Source: Direct Originator: D. Slusher TUOI: ANO-1-LP-RO-MSSS Objective: 3 Point Value: 1 Section: 4.2 Type: Generic AOP System Number: 062 System

Title:

Loss of Nuclear Service Water

Description:

Knowledge of the reasons for the following responses as they apply to the Loss of Nuclear Service Water: The automatic actions (alignments) within the nuclear service water resulting from the actuation of the ESFAS.

K/A Number: AK3.02 CFR

Reference:

41.4, 41.8 / 45.7 Tier: 1 RO Imp: 3.6 RO Select: Yes Difficulty: 3 Group: 1 SRO Imp: 3.9 SRO Select: Yes Taxonomy: C Question: 16 RO:i SRO:1 16 Service Water Pumps P-4A, P-4B (supplied from A-4), and P-4C are running.

An ES actuation coincident with a loss of off-site power occurs.

Which service water pumps will autostart when A-3 and A-4 are re-energized?

A. P-4A, P-4B and P-4C B. P-4A and P-4B C. P-4B and P-4C D. P-4A and P-4C Answer:

D. P-4AandP-4C Notes:

When ESAS actuates and the buses are re-energized the P-4A and P-4C handswitch position will interlock P 4B and keep P-4B from starting. Therefore, a, b, and c responses are incorrect.

References:

STM 1-42, Rev. 18, Service and Auxiliary Cooling Water, page 13, 14, 15 History:

Developed for 1999 exam.

Used in 2001 RO/SRO Exam.

Selected for the 2010 RO/SRO exam.

Service & Auxiliary Cooling Water STM 1-42 Rev. 18 Each vacuum breaker returns flow back to its respective service water bay.

Each vacuum breaker is provided with a manual isolation valve and a bypass. The isolation valves, SW-li 8A, B & C are Category E valves normally locked open.

The service water pumps are driven by a 350 HP, 4160 Volt AC induction motors. The motors are located on the second floor of the Intake Structure. This location ensures pump operability in the event of a flood.

Additional information on SW pump design is contained in the table below.

Line Shaft Diameter 2-3/16 Discharge Column CS, Flanged Impeller Diameter 18-1/2 Power supplies for the motors are as follows:

• P-4A is powered from Bus A3 (4.16KV) through breaker A-302. If offsite power is unavailable and the #1 emergency diesel generator is running, A3 will be powered from DG #1 (K4A) through generator output breaker A-308.

• P-4C is powered from Bus A4 (4.16KV) through breaker A-402. If offsite power is unavailable and the #2 emergency diesel generator is running, A4 will be powered from DO #2 (K4B) through generator output breaker A-408.

• Service water pump P-4B is a swing pump. It can be powered from either A3 or A4 through motor operated disconnect (A6). P4B power can be electrically swapped using HS-3 608 or by manually swapping A6 to the opposite bus. HS-3 608 is located on panel C- 18. To ensure system redundancy, it must be selected to the associated bus for the pump that it is in standby for. If P-4B is backup to P-4A then HS-3608 will be in the A-3 (breaker A-303) position and A-4 (breaker A-403) for P-4C backup. The MOD for P4B is located in the upper level of the Intake Structure in the electric fire pump room.

Note: Logic for auto-start is not determined by selector switch position but by breaker aIinment.

13

Service & Auxiliary Cooling Water STM 1-42 Rev. 18 The following table contains SW pump handswitch location and its associated positions.

SW Component HS # HS Location Remarks Pump P4A SWLoopl 3611 C-18

• P4B Swing SW Pump (A3) 3609 C-18

• Swing SW Pump (A4) 3600 C-16

• Bus selector switch 3608 C-18 Selects power to either A3 or A4 P4C SWLoopll 3610 C-16

• Handswitch; start/stop/normal/pull-to-lock; spring return too normal.

2.3.4.1 SW Pump Start During normal operation the A3 and A4 buses are powered from Logic non-vital 4160-volt buses Al and A2 respectively through bus tie breakers. A3 is fed from Al through tiebreaker A309 and A4 is fed from A2 through tiebreaker A409. Al or A2 can be supplied power from one of the three power supplies available. During turbine generator operation, Al and A2 are powered from the Unit Aux transformer, which provides power to all in house loads. Following a turbine trip, electrical power is automatically transferred to the SU 1 transformer. If SU 1 becomes inoperable then power can be manually aligned to SU 2. SU2, which can provide power to either Unit 1 or Unit 2 or both is provided with a load shed feature to limit load placed on SU 2. For additional information on SU 2 load shed refer to 1107.001 Electrical System Operation.

Each SW pump is provided with 15-second time delays, which will time out prior to restarting the SW pumps previously running when power is restored. If no offsite power is available then an under voltage condition on either A3/A4 or B51B6 will cause the bus tiebreaker to open, associated EDG to start and tie onto the bus.

When A3 or A4 are re-energized, the SW pump(s) will restart after their associated time delay times out.

To prevent from exceeding EDG loading during an ESAS actuation with a loss of offsite power and three SW pumps in service, modifications to the SW pump start circuitry were incorporated.

Prior to these modifications the potential existed for two SW pumps to be placed on a single EDG due to time delays for each pump are set at 15 seconds. This condition would occur if the time delay for the swing pump timed out before the lead pump. This event would start the swing SW pump and the lead SW pump overloading the EDO.

DCP-92-1016 modified the SW pump start logic to prevent this event from occurring. Modifications to the system included replacing P4A and P4C handswitches and P-4B start permissive circuitry. The new handswitches, HS-3610 for P-4C and HS-361 1 for P-4A provided additional contacts that tie into the swing SW pump start logic. A contact in each handswitch is wired into the auto-start 14

Service & Auxiliary Cooling Water STM 1-42 Rev. 18 circuitry for service water pump P-4B that allows pump to auto-st art when a specific condition exists.

For ease of discussion the logic explanation will cover P-4B auto-start when selected to the A3 bus. If any of the following conditions exists, then P-4B will auto-start when an ESAS actuation occurs along with or without a loss of offsite power.

HS-361 1 (P-4A) in normal after stop (green flagged).

HS-3611 in Pull to Lock.

HS-361 1 placed in stop position.

Feeder breaker for P-4A trips open with HS in normal after start (red flagged).

Service water cross-connect isolation valves will automatically align to provide flow from P-4B to the affected loop. Additional information on Service Water crosstie valve logic will be discus sed in section 2.3.8.

The following are automatic starting and stopping interlocks associated with the service water pumps:

• Pump motor will stop when turned to off or P-T-L.

• Pump motor will stop on a loss of voltage.

• Pump motor will stop on an electrical fault.

• Pump motor for P-4A and P-4C will restart after a loss of voltage if its handswitch is in normal after start.

• Pump motor for P-4B will restart after a loss of voltage only if the handswitch for the primary pump is in the Stop, Normal-After-Stop or Pull-To-Lock position.

• Pump motor will start when hand-switch is placed in start.

2.3.4.2 SW Pump (Refer to Figure 42.01 & Table 42.02)

Instrumentation P-4A, B, and C motor winding temperature are continuously and Alarms monitored by temperature elements. These temperature elements send a signal for their respective pump motor windings to trend record er TR-2808 located on panel C-19 in the control room. The SW Pump Motor winding temperatures can also be read on the plant computer.

When motor winding temperature reaches 2 50°F, annunciator K10-C4 SW Pump Mtr Wdg Temp Hi will alarm alertin g the operator of this condition. TEs associated with each SW pump are listed below.

• P-4A Mtr Wdg temp (TE-3 650)

• P-4B Mtr Wdg temp (TE-3613)

• P-4C Mtr Wdg temp (TE-3610) 15

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0335 Rev: 0 Rev Date: 9-7-99 Source: Direct Originator: D. Slusher TUOI: ANO-l-LP-RO-EOPO4 Objective: 6 Point Value: 1 Section: 4.3 Type: B&W EPE/APE System Number: E04 System

Title:

Excessive Heat Transfer

Description:

Ability to operate and / or monitor the following as they apply to the (Inadequate Heat Transfer):

Desired operating results during abnormal and emergency situations.

K/A Number: EAI .3 CFR

Reference:

CFR: 41.7/45.5 / 45.6 Tier: I RO Imp: 3.6 RO Select: Yes Difficulty: 2.5 Group: I SRO Imp: 3.8 SRO Select: Yes Taxonomy: C Question:

RO:J 17 SRO: I 7 Given:

- Loss of all Feedwater

- HPI core cooling started What indicates adequate HPI core cooling?

A. CET temperatures stable after 100 minutes.

B. T-cold tracking associated SG T-sat.

C. T-hot tracking CET temperatures.

D. T-hotiT-cold differential temperature dropping.

Answer:

A. CET temperatures stable after 100 minutes.

Notes:

A is correct since the only criteria for evaluation of adequacy of core cooling via HPI is decrease a in CET tern ps.

B, C, and D are individual indications of adequate primary to secondary heat transfer.

References:

1202.004 Change 6 History:

Developed for 1999 exam.

Used on 2004 RO/SRO Exam.

Selected for the 2010 RO/SRO exam

CHANGE 1202.004 OVERHEATING 006 PAGE 7 of 17 INSTRUCTIONS CONTINGENCY ACTIONS

6. (Continued).

F. Isolate Pressurizer Spray Line as follows:

1) Place Pressurizer Spray Control in HAND AND verify closed (CV-1 008).

2) Close Pressurizer Spray Isolation (CV-1 009).

7. jf MU Tank level drops below 18, THEN close Makeup Tank Outlet (CV-1 275).

8. Check Letdown in service. 8. jf CET SCM is adequate, THEN unless fuel damage or RCS to ICW leak is suspected, restore Letdown flow (RT 13).

9. Control RCS press within limits of Figure 3 (RT 14).

10. Check CET temps stable or dropping. 10. Perform one of the following:

A. IF HPI flow is <full flow from one HPI pump, THEN GO TO step 18.

B. Hold at this point until one of the following conditions is met:

1) IF EFW becomes available, THEN GO TO step 13.

2) E MFW or AFW pump becomes available, THEN GO TO step 12.

3) IF CET temps begin to drop, THEN GO TO step 11.

4) IF 120 minutes on HPI cooling elapse AND CET temps are still rising, THEN GO TO step 18.

5) jf CET temps are superheated moving away from the saturation line, THEN GO TO 1202.005, INADEQUATE CORE COOLING procedure.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0775 Rev: 0 Rev Date: 9/8/2009 Source: New Originator: S. Pullin TUOI: AILP-RO-GEN Objective: 7 Point Value: 1 Section: 4.2 Type: Generic APEs System Number: 077 System

Title:

Generator Voltage and Electrical Grid Disturbances

==

Description:==

Ability to interpret reference materials, such as graphs, curves, tables, etc.

K/A Number: 2.1.25 CFR

Reference:

41.10/43.5/45.12 Tier: 1 RO Imp: 3.9 RO Select: Yes Difficulty: 3 Group: I SRO Imp: 4.2 SRO Select: Yes Taxonomy: C Question: RO: 8 SRO:r 8 REFERENCE PROVIDED Given:

Plant 100% power Electrical storm caused an grid disturbance The Dispatcher calls Control Room and requests Unit I Generator power factor With the Unit I Generator operating at 880 MWe gross out, what reactive load must it carry to be at a 0.98 power factor?

A. 140 MVAR B. 180 MVAR C. 200 MVAR 260 MVAR Answer:

B. 180 MVAR Notes:

Using Attachment N of Op-I 102.004 B. is correct A, C and D are associated with different power factors or generator loads.

References:

1102.004 Change 048 History:

Developed for the 2010 RO/SRO exam.

0 C)

(Under-excited) MEGAVARS (Over-excited)

- C..) .

C..) N) - 4 Co o o 0 C 0 0 0 o o 0 0 0 0 _Q o 0 C C C C 0 0 C 0 0 0

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(31

RO Written Exam Tier I Group 2

Form ES-401-2 ne PWR Exammation Outli ES-401 Form ES-40l-2 PWR Examination Outline Tier 1/Grou 2 RO ormal Plant Evolutions

-401 Emer enc and Abn # QID Typ IR K/A Topic(s) e K K K A A G Function E/APE # / Name / Safety 12312 n/a Not selected Withdrawal /1 n/a 000001 Continuous Rod Not selected trol Rod /1 n/a 000003 Dro ed Con Not selected Control Rod / 1 n/a 000005 Inoperable/Stuck Not selected 2.8* 19 776 N ation / I leak 000024 Emer enc Bor AKI .01 PZR reference el Malfunction /2 x

abnormalities.

000028 Pressurizer Lev 3.1 20 777 N rce-X AA2.04 Satisfactory sou overla Range NI /7 ran e / intermediate-range 000032 Loss of Source n/a Not selected diate Range NI /7 000033 Loss of Interme domly n/a AK2.1 Changed to ran m 068 AK 2.07 l Handling Accident /8 selected S ste 000036 (BW/A08) Fue tank AAI.10 CVCS makeup 2.9 21 778 N level indicator or Tube Leak /3 000037 Steam Generat n/a Not selected ser Vacuum /4 n/a 000051 Loss of Conden Not selected uid RadWaste Rel. /9 randomly n/a 000059 Accidental Li AKI .04 Changed to I .01 seous Radwaste Rel. /9 selected S tern 028 AK 000060 Accidental Ga 3.4 22 634 D tained in x AK3.02 Guidance con s stem.

rms / 7 alarm res onse for ARM 000061 ARM System Ala 3.1 23 695 DR AKI .02 Fire Fi hting x

3.3 24 779 D 0067 Plant Fire On-site /8 AK2.07 ED/G Room Evac. /8 x -

n/a J0068 BW/A06) Control Not selected CTMT Inte ri /5 n/a 000069 W/E14) Loss of Not selected lin / 4 n/a 000074 /E06&E07 Inad. Core Coo Not selected Coolant Activi /9 n/a 000076 Hi h Reactor Not selected

& SI Termination /3 n/a W/EO1 & E02 Rediagnosis Not selected Over-pressure / 4 n/a W/E13 Steam Generator Not selected odin 15 n/a W/E15 Containment Flo Not selected t Radiation /9 n/a W/E16 Hi h Containmen Not selected 1 n/a 8W/AOl Plant Runback / Not selected I-X IY /7 n/a BW/A02&A03 Loss of NN Not selected BW/A04 Turbine Tn / 4 AK2.1 Components, and x

trol and safety sel Actuation /6 functio ns of con BW/A05 Emergency Die udi ng systems, incl 4.0 25 349 D instrumentation, signals, and interlocks, failure modes, ures.

automatic and manual feat ropriate AA2.2 Adherence to app 3.3 26 780 N within procedures and operation lity s BW/A07 Flooding / 8 the limitation s in the faci license and amendments n/a Not selected coolin Mar in /4 n/a BW/E03 Inade uate Sub Not selected ldown Depress. /4 n/a BW/E08; W/E03 LOCA Coo Not selected E10 Natural Circ. 14 limiting W!E09; CE/A13; W/E09& 2.2.22- Knowledge of 4.0 27 595 N ope rati ons and Enclosures conditions for

/E13&E14 EOP Rules and safe limits.

n/a Not selected PTS / 4 CE/Al 1; W/E08 RCS Ove rcoolin -

Form ES-401-2 I

ES-401

Form ES-401-2 PWR Examination Outline W-4O1 n/a Not selected

/2 CE/Al 6 Excess RCS Leaka e Not selected n/a CE/E09 Functional Recove 9 2 2 1 1 2 1 Grou Point Total:

A CategO Point Totals:

Form ES-4014 ES-401

ARKANSAS IT IA L RO IS RO EX AM BANK QUESTION DATA IN I

NUCLEAR ONE UNIT -

e: New Originator: S. Pullin Rev: 0 Rev Date: 9/8/2009 Sourc QID: 0776 Point Value: 1 2 Objective: 9a TUOI: ASLP-RO-CMPO Section: 4.2 Type: Generic APEs Level Control Malfunction Sy ste m

Title:

Pressurizer (PZR)

System Number: 028 to pressurizer pli cat ion s of the follow ing concepts as they apply of the operational im

Description:

Knowledge PZR reference leg abnorm alities.

level contro malfunctions:

l 10/45.3 CFR

Reference:

41 .8/41.

KIA Number: AK1.01 Difficulty: 2 RO Im p: 2.8 RO Select: Yes Tier: 1 Taxonomy: C SR O Im p: 3.1 SRO Select: Yes Group: 2 RO: 19 SRO:1 Question:

Given:

Plant at 100% power rizer reference leg Leak develops on the pressu ve, CV-1235?

e on lev el ind ica tion and pressurizer level control val What effect does this hav 35, opens to control level.

rea ses and pre ssu riz er level control valve, CV-12 A. Indicated level dec trol level.

ssu riz er lev el con tro l val ve, CV-1235, closes to con ses and pre B. Indicated level decrea to control level.

pre ssu riz er lev el con trol valve, CV-1 235, opens ses and C. Indicated level increa ses to control level.

and pre ssu riz er lev el control valve, CV-1235, clo ses D. Indicated level increa Answer: control level.

pre ssu riz er lev el con tro l valve, CV-1235, closes to and

0. Indicated level increases Notes: of the level rise uld cau se ind ica ted lev el to increase. As a result reference leg wo D. is correct, a leak in the nt.

ore der to maintain level at setpoi CV-1235 will close in fer ent pos sib le combinations.

, using the dif A, B, and C are incorrect

References:

ASLP-RO-CMPO2 Rev 2 History:

/SRO exam.

New selected for 2010 RO

UIDE KEY POINTS, A IDS, INSTRUCTOR G SWERS QUESTIONS/AN bient temperature

a. An increase in am wet reference will cause density of leg to decrease in lower DIP

b. This will result d indicated sensed by DIP cell, an actual than level will be greater level ct produces lower

c. The opposite effe ambient indicated level when s

temperature decrease scussed, radiation

4. As previously di affect detector levels near DIP cell integrity environment can

a. A high radiation le detector cell, permanently embritt its elasticity causing cell to lose acteristics, as and altering its char itive well as degrade sens electronics K. Failure Indications s with wet reference

1. For level detector igh pressure side leg connected to h g failure modes of DIP cell, followin exist Objective 9b ble leg of D/P cell

a. A break in varia ing sensed by creates higher D/P be level D/P cell, resulting in tin g low level instrument indica Objective 9a rence leg brea

b. Conversely, refe nsed across creates lower DIP se indicated DIP cell, esulting in al level level higher an actu PHYSICS

© 1999 GENERAL TS I CHAPTER 7 54 of 136 CORPORATION PWR / COMPONEN REV 2 RS SENSORS/DETECTO

AS ISR O EX AM BA NK QU ESTION DATA ARKANS INITIAL RO NUCLEAR ONE - UNIT I Source: New Originator: S. Pullin Rev: 0 Rev Date: 9/8/2009 QID: 0777 Point Value: I Objective: 4 TUOI: AILP-RO-NOP Section: 4.2 Type: Generic APEs tion Range Nuclear Instrumenta er: 032 Sy ste m

Title:

Loss of Source System Numb to the Loss of Source Range int erp ret the following as they apply

Description:

Ability to det erm ine and ate-range overlap Instru me nta tion: Sat isfa ctory source-range intermedi Nucle ar 5.13 CFR

Reference:

43.5/4 K/A Number: AA2.04 Difficulty: 3 RO Imp: 3.1 RO Select: Yes Tier: I Taxonomy: C SRO Imp: 3.5 SRO Select: Yes Group: 2 RO: 20 SRO: J Question:

Given:

nts Source Range 5 X 10 4 cou Ra nge I X 10- 9 amps Intermediate nts per second.

sta rtu p, the sou rce ran ge instruments fail to 3 cou During the dition?

or action for the given con What is the required operat reactivity changes..

rations involving positive A. Immediately suspend ope D trip breakers open.

B. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> verify CR C. Continue the startup..

trol rods.

shutdown and insert all con D. Immediately initiate a Answer:

C. Continue the startup..

Notes: icate >10 -10 amps.

allo ws con tin uin g wit h sta rtup if intermediate range ind C. is correct, procedure nts fail and both the act ion s to tak e wh en both source range instrume to these are A, B and D are incorrect due cha nne ls indicate <10 -10 amps.

intermediate range

References:

1203.021 Change 10 History:

exam.

New for the RO/SRO 2010

PROCEDUREIWORK PLAN TITLE: I PAGE: 7 of 8 PROC JWORK PLAN NO.

I 1203.021 LOSS OF NEUTRON FLU X INDICATION CHANGE: 010 SECTION 3 ROUGH 5 NELS IN MODES 2 TH ON E OR MOR E SO URCE RANGE NI CHAN LOSS OF

1. 0 SYMPTOMS ectly.

tion reading incorr 1.1 Source range indica alarm.

IBITED (K08-A2) 1.2 CRD WITHDRAWAL INH 2.0 IMMEDIATE ACTION NONE 3.0 FOLLOW-UP ACTIONS NO TE

-scale indication co nd iti on s ap ply , there is no on lowing If all 4 of the fol of neutron flux:

are 5% power, er range instruments

• Three of four pow >10-10 amps, ge instrument is

• No intermediate ran s,

trument is <l0 cp

• No source range ins perable.

(NR-502) is ino de Range Recorder

• Reactor Power Wi flux is available, no on -sc ale ind ication of neutron 3.1 IF r s THEN trip reacto njunction with thi rm Re ac tor Tr ip (1202.001) in co AND perfo procedure.

operable, e range channel is >10-10 amps, 3.2 IF only one sourc ne ls indicates 1 of 2 int erm ediate range chan OR 3.3.9).

THEN continue pla nt operations (TS il, ge instruments fa 3.3 IF both source ran ls indicate 10° am ps, bo th int erm ed iate range channe AND following:

THEN perform the NOTE are tivity additions s wh ich re su lt in positive reac Sh utd ow n cha nge in the Plant temperature is accounted for temperature change allowed provided the s.

Marqin calculation Condition A.

3.3.1 Refer to TS 3.3.9 olving positive d operations inv 3.3.2 Immediately suspen reactivity changes.

ol insert all contr me dia tely in iti ate a shutdown and 3.3.3 Im rods.

akers open.

rify CRD trip bre 3.3.4 Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ve

STM 1-67 Rev. 11 Nuclear Instrumentation STM 1-67 n Nuclear Instrumentatio 1.0 Introduction Core) tion on the Excore (Out of This STM contains informa I. It incl udes s) for ANO Unit Nuclear Instrument System (NI the plan t and component locations in operational theory of detectors, ons . The ons and equipment conditi normal and abnormal operati the effect effect nuclear instruments have on plant operation, and discussed.

the Nuclear Instruments is plant operations have on is found in theory of detector operation Additional information on STM 1-62, Radiation Monitoring sure (NI) System is designed to mea The Nuclear Instrumentation 1.1 System Function flux using ten channe ls of out of core over twelve decades of neutron 67. 01) The entation. (Refer to Figure neutron detectors and instrum rato r and are layed to the Reactor Ope full range of indications are disp sys tem s.

ion and Integrated Control supplied to the Reactor Protect plet e and ed to overlap to provide com Measurement ranges are design tor.

continuous information of the full operating range of the reac ES UMENTAflON FLUX RANG FIGURE 67.01: NUCLEAR 1NSTR o

DLO.CI0)RNEI7IRON FUJX, 0 0 0

• 0 (0 0 10 --

0 0 II 0 10 bF-- -4 H- (0 1 0 10 .1 10 0

01 10 III 0 0 0 I(OAI 1(10 O1WLR. 3 COUNTS PER SFC000 01.1 V RZ COMMA I0 I( 11S7 0 6(11)11CC M Ii 6 IL, I)

RANOP 1.1 IOU POWER (NI NR.)1L2 IC II 6 WII)1I 6.00(1) .1 0 0 10 0 10 -.

(Al ION CUI(RI-NI ACIPI)RCS 00_ER- -j MEDIAII) +.-. -

II (0 10 -

LI/sOUL 11)0 lO III (0 II I0 01.5.6.? A 0

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0778 Rev: 0 Rev Date: 9/8/2009 Source: New Originator: S. Pullin TUOI: Al LP-RO-ALEAK Objective: 3 Point Value: 1 Section: 4.2 Type: Generic APEs System Number: 037 System

Title:

Steam Generator (S/G) Tube Leak

==

Description:==

Ability to operate and / or monitor the following as they apply to the Steam Generator Tube Leak: CVCS makeup tank level indicator.

KIA Number: AAI.10 CFR

Reference:

41 .7/45.5/45.6 Tier: 1 RO Imp: 2.9 RO Select: Yes Difficulty: 2 Group: 2 SRO Imp: 3.1 SRO Select: Yes Taxonomy: K Question: RO:] 21 SRO:1 21 Given:

Plant at 100% power Makeup Tank level dropping at 1 inch every 2 minutes.

A OTSG N-16 TROUBLE (K07-A5)

PROC MONITOR RADIATION HI (K10-B2)

What is the A OTSG Tube Leak rate?

A. logpm B. l5gpm C. 2ogpm D. 25gpm Answer:

B. lsgpm Notes:

B. 15 gpm is correct based on makeup tank level is 30 gallons per inch, at a rate of change of 1 inch per 2 minutes equals 15 gpm leak.

A, C and D are incorrect.

References:

1203.039 Change 011 History:

New for the RO/SRO 2010 exam.

CHANGE 1203.039 EXCESS RCS LEAKAGE 011 PAGE 11 of 14 ATTACHMENT I Estimate of RCS Leakrate NOTE

. The RB Sump contains 45.4 gal/percent.

. Dirty Waste Drain Tanks (T-20s) contain 52.5 gal/percent.

. Auxiliary Building Sump contains 8.98 gal/percent.

. ICW Surge Tank T-37B Level (PDIS-2229) 0.5 to 27 PS id (1 psid 333 gallons).

. Estimated MU Flow During RCS Cooldown is contained in Attachment 2 of this procedure.

1. Estimate RCS Ieakrate using the following formula:

. Use the following table to perform mass balance estimate.

NOTE

. When the BWST is aligned to the Makeup Tank, Makeup Tank Level changes should generally .QI be used for leak rate estimation.

. Pressurizer and Makeup Tank level changes can either be added subtracted to estimate leak rate.

• jf applicable, record current cooldown rate for leak estimation:

• Calculate Seal bleedoff flow for RCPs + + +

Makeup Flow F1238/C04 Gpm Plus Seal Injection Flow F1239/C04 Gpm Plus HPI Flow SPDS/C16 and C18 Gpm Plus Pressurizer Level Gpm Minus (IF rising)

X 12.4 gal/in Change Plus (IF lowering)

Makeup Tank Level X 30.86 gal/in Gpm Plus (IF rising)-

Change (N/A IF Minus (IF lowering)

BWST Outlet Open)

Letdown Flow F1236/C04 Gpm Minus Seal Bleedoff flow F1270-31C13 Gpm Minus Makeup Flow Due to Gpm Minus Attachment 2 Cooldown TOTAL Gpm

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0634 Rev: 0 Rev Date: 11/8/05 Source: Direct Originator: J.Cork TUOI: AILP-RO-RMS Objective: 7 Point Value: 1 Section: 4.2 Type: Generic APEs System Number: 061 System

Title:

Area Radiation Monitoring (ARM) System Alarms

==

Description:==

Knowledge of the reasons for the following responses as they apply to the Area Radiation Monitoring (ARM) System Alarms: Guidance contained in alarm response for ARM system.

KIA Number: AK3.02 CFR

Reference:

41.5, 41.10/45.6/45.13 Tier: 1 RO Imp: 3.4 RO Select: Yes Difficulty: 4 Group: 2 SRO Imp: 3.6 SRO Select: Yes Taxonomy: C Question: RO:j 22 SRO:1 Given:

- AREA MONITOR RADIATION HI (K10-B1) in alarm

- RADIATION MONITOR TROUBLE (Kb-Cl) in alarm In accordance with the alarm response procedure, the area monitors on C25 Bay 3 must be inspected.

What indication(s) would you expect to find on the alarming monitor drawer with both of the above annunciators in alarm?

A. WARNING and POWER ON lights on B. POWER ON light off C. HIGH ALARM light on and POWER ON light off D. FAILURE light on Answer:

B. POWER ON light off Notes:

B is correct, a loss of power will cause both the Hi Radiation and Trouble annunciators to come in.

A is incorrect, this would cause the Hi Radiation but not the Trouble annunciator.

C is incorrect, the POWER ON light off will cause both annunicators but the HIGH ALARM light will not be on with a loss of power.

D is incorrect, this will cause the Trouble annunciator but not the Hi Radiation annunciator.

References:

1203.0121, Chg. 046 STM 1-62, Rev. 11 History:

New for 2005 RO re-exam.

Selected for 2010 RO/SRO exam.

Radiation Monitoring STM 1-62 Rev. 11 (Refer to Figure 62.09) In addition to an analog meter, station indicating units have:

Four status lights to indicate Power On (green), Failure (white), High Alarm (red) and Warning (amber). The failure alarm occurs when the signal drops below a preset value.

• One three position switch allows for checking the warning and high alarm setpoints. Operation of the Alarm Setting switch does not bring in high alarms or initiate any automatic actuation.

• Another three position switch is provided for alarm reset and check source operation.

• When either of the three position switches is removed from the normal position of operate, a Rad Monitor Test in Progress Alarm will annunciate in the control room on K-b.

• Each drawer can be slid away from the panel face to gain access to potentiometers for setpoint adjustment.

High alarm of all the monitors is interlocked to give audible and visual remote alarms at the location of each monitor. The ailure alarm or a loss of power to the unit will actuate the Radiation Monitor Trouble annunciator (Kb-Cl) in the Control Room. A High alarm or a loss of pow to the unit will actuate the Area Monitor Radiation Hi annunciator (Kb0-B1).

The ARMs provide inputs to the plant computer. A listing of the ARMs and their current value can be displayed on the plant computer by going to Group Display (GD), then ARMS.

2.1.2 Control Room The control room envelope (Unit 1 and Unit 2) is monitored for Radiation Monitor excessive radiation by five detectors. These radiation detectors are RE-8001, 2RE-8001A, 2RE-8001B, 2RE-8750-1A, and 2RE-8750-1B. The Unit 1 Control Room Area Monitor (RE-8001) is located on the east wall of the control room. In addition, radiation monitors 2RE-8001A and 2RE-8001B are mounted in the air supply and operating area ductwork for the Unit 1 Control Room. High radiation on any one of these monitors will cause Control Room isolation for both Control Rooms. The Control Room Supply Duct Radiation Hi annunciator (K16-D1) is the associated alarm. Refer to STM 1-12 for Control Room Ventilation.

The warning alarm on Control Room Area Monitor (RI-8001) provides the actuating signal for control room isolation Power to this unit is from RS-4 through C-25 Bay 3.

The actuation level for high radiation is sufficiently below hazardous radiation levels to minimize operator dose during an accident and is sufficiently above normally experienced background levels to minimize spurious actuations.

PROCJWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 3 of 68 1203.0121 ANNUNCIATOR K10 CORRECTIVE ACTION 046 CHANGE:

Page lof2 Location: C16 Device and Setpoint:

See Radiation Monitoring System Check and Test (1305.001) AREA MONITOR Supplement 6, Area Radiation Monitor Weekly Alarm Check. RADIATION HI Alarm: Kl0B1 1.0 OPERATOR ACTIONS

1. Inspect C25 Bay 3 and determine alarming monitor.

A. Determine if alarm is due to high radiation or loss of power.

2. IF alarm is due to momentary spike, THEN reset alarm.

3. IF loss of power, THEN GO TO RADIATION MONITOR TROUBLE (Kb-Cl).

4. IF confirmed high radiation within reactor building AND personnel are inside RB, THEN sound reactor building evacuation alarm.

A. IF high radiation outside RB AND within a Radiologically Controlled Area, THEN GO TO step 6.

5. IF confirmed high radiation outside reactor building AND outside Radiologically Controlled Areas, THEN announce high radiation warning on plant public address system.

6. IF Control Room (RI800l) in alarm, THEN refer to ACTUATION -- CONTROL ROOM ISOLATION (K16-B2).

7. Initiate action to have high radiation area surveyed.

8. IF SF Pool (RI-8009) in alarm AND Spent Fuel Pool is the radiation source, THEN maximize SF Pool purification flow per Spent Fuel Pool Purification section of Spent Fuel Cooling System (1104.006)

A. IF radiation levels inside a Radiologically Controlled Area are determined to be > limits of EVACUATION (1903.030),

THEN GO TO 1903.030.

9. IF radiation rises to 2.5 mrem/hour outside a Radiologically Controlled Area, THEN GO TO EVACUATION (1903.030).

10. IF projected summed releases exceed NUE criteria for one hour at site boundary, THEN notify SM to review EAL5 (1903.010).

PROC.IWORK PLAN NO. PROCEDUREIWORK PLAN TITLE:

I PAGE: 4 of 68 1203.0121 ANNUNCIATOR K10 CORRECTIVE ACTION CHANGE: 046 KlOBl Page 2 of 2

11. IF it is desired to raise alarm setpoint, THEN perform applicable sections of Area Radiation Monitor Monthly Alarm Check (1305.001 Supplement 6) 2.0 PROBABLE CAUSES NOTE This annunciator has multiple input without reflash.

1. Any area monitor in C25 Bay 3 senses radiation above alarm setpoint

2. Any area monitor in C25 Bay 3 de-energized

3. Any area monitor in C25 Bay 3 alarm lamp removed or burned out

3.0 REFERENCES

1. Schematic Diagran Annunciator Kb (E460, sheets 1 3)

PROC.IWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: I PAGE: 5 of 68 1203.01 21 ANNUNCIATOR K10 CORRECTIVE ACTION CHANGE: 046 Page 1 of 3 Location: C16 Device and Setpoint:

De-energization of or FAILURE ALARM on any radiation RADIAT ION monitor in Radiation Monitoring System Panel (C25 Bays 13 MONITOR and Bay 4 of C24). Monitors are listed on next page. TROUBLE Alarm: Kb-Cl 1.0 OPERATOR ACTIONS

1. Observe monitors at C24 and C25 for FAILURE ALARM light(s) on or POWER ON light(s) off.

2. IF power is off to all monitors in a bay, THEN verify supply breaker closed:

A. Rad Monitor Panel C24, Rad Monitor Panel C25, Bay 1 (RS1, bkr 8)

B. Rad Monitor Panel C24, Rad Monitor Panel C25, Bay 2 (RS2, bkr 8)

C. R&d Monitor Panel C25, Bay 3 (RS4, bkr 8)

3. IF either of the following monitors is inoperable (FAILURE ALARM or power loss):

• Spent Fuel Pool (RI8009)

• Fuel Handling Area (RI-80l7)

AND fuel handling in progress, THEN stop fuel handling until radiation monitoring requirement is satisfied per Control of Unit 1 Refueling (1502.004) OR Control of Fuel and Control Rod Movement in the Ub Spent Fuel Area (1502.010).

(TEN 3.9.1 and TEN 3.9.2)

4. IF Control Room (RI8001) is inoperable (FAILURE ALARM or power loss),

THEN verify control room emergency ventilation actuation. (TS 3.7.9)

5. IF Liquid Radwaste (RI-4642) is de-energized, THEN verify CZ Disch to Flume Flow (CV4642) is closed or auto closes.

(ODCM App.l, L2.l.l)

PROC.IWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 6 of 68 1203.0121 ANNUNCIATOR K10 CORRECTIVE ACTION CHANGE: 046 Kb-Cl Page 2 of 3 NOTE The following alignment stops gaseous release and diverts flow to Waste Gas Surge Tank (T17)

6. IF Gaseous Radwaste (RI4830) is de-energized, THEN verify the following: (ODCM App.l, L2.2.l)

• T18s Discharge to Gaseous Radwaste Discharge Header Flow Control (CV4820) closed

• Gaseous Radwaste Discharge Isol (CV-4830) closed

• ABVH Diversion to T-17 (CV-4806) open

7. IF RB Atmos Gaseous Monitor is inoperable, THEN refer to Reactor Building Ventilation (1104.033). (TS 3.4.15)

S. Initiate steps to survey areas for which radiation monitors are inoperable.

9. Initiate steps to have failed monitor(s) checked and repaired.

10. IF alarm was caused by FAILURE ALARM on monitors, THEN all monitors that are failed, must be reset using ALARM RESET switch on front of monitor to clear KbCl.

2.0 PROBABLE CAUSES NOTE

• This annunciator has reflash capability. If the alarm window is lit solid due to one cause and another cause actuates, the alarm will go to fast flash with an audible alarm.

• FAILURE ALARM light on monitor indicates that the monitor has had no input from the detector for one minute; detector failure.

1. Any radiation monitor FAILURE ALARM in C25 or Bay 4 of C24

2. De-energization of any radiation monitor in C25 or Bay 4 of C24

3. Any radiation monitor in C25 or Bay 4 of C24 alarm lamp removed or burned out

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0695 Rev: I Rev Date: 4/1/2008 Source: Repeat Originator: Steve Pullin TUOI: ASLP-RO-FRHAZ Objective: 4B Point Value: I Section: 4.2 Type: Generic APEs System Number: 067 System

Title:

Plant Fire on Site

==

Description:==

Knowledge of the Operational implications of the following concepts as they apply to plant fire on site: fire fighting.

K/A Number: AKI.02 CFR

Reference:

41 .8/41.10/45.3 Tier: 1 RO Imp: 3.1 RO Select: Yes Difficulty: 2 Group: 2 SRO Imp: 3.9 SRO Select: Yes Taxonomy: K Question: RO:1 SRO:1 23 Per 1015.007, Fire Brigade Organization and Responsibilities, which of the following describes the Ops Manning composition of the Fire Brigade for the initial response to a fire on Unit 1?

A. Unit I supplies the Fire Brigade Leader, Unit 2 supplies 3 Fire Brigade members, Security supplies one support member.

B. Unit I supplies the Fire Brigade Leader and 2 Fire Brigade members, Unit 2 supplies 1 Fire Brigade member, Security supplies one support member.

C. Unit 2 supplies the Fire Brigade Leader, Unit I supplies 3 Fire Brigade members, Security supplies one support member.

D. Unit 2 supplies the Fire Brigade Leader and I Fire Brigade member, Unit I supplies 2 Fire Brigade members, Security supplies one support member.

Answer:

A. Unit 1 supplies the Fire Brigade Leader, Unit 2 supplies 3 Fire Brigade members, Security supplies one support member Notes:

A is correct per the requirements of 101 5.007 B is incorrect. This answer was previously correct for a fire on Unit I prior to the latest revision.

C is incorrect. This is correct for a fire on Unit 2 D is incorrect This answer was previously correct for a fire on Unit 2 prior to the latest revision.

References:

101 5.007, Fire Brigade Organization and Responsibility Chg. 019 History:

Selected for 2008 RD Exam Selected repeat for the 2010 RO/SRO exam

I I PROCEDURE/WORK PLAN TITLE: II I PROC IWORK PLAN NO. PAGE: 4oflO I 101 5.007 FIRE BRIGADE ORGANIZATION AND RESPONSIBILITIES I I CHANGE: 019 5.3 Fire Brigade Members 5.3.1 Under the direct supervision of the Fire Brigade Leader, Fire Brigade Members are responsible for primary extinguishment efforts (extinguishers, hoses, etc.).

5.3.2 The Fire Brigade Members of the unaffected unit shall respond to a fire in the affected unit.

5.3.3 Restore fire equipment after use in accordance with Fire Equipment Restoration section of this procedure.

5.4 Security Force 5.4.1 Shall assign a support person to the Fire Brigade Support Team to respond to a fire.

5.4.2 The support person is responsible for providing support activities under direct supervision of the Fire Brigade Leader or the three fully trained Fire Brigade Members.

These activities will normally include, but are not limited to, supplying additional equipment, supplying SCBA5, hose laying, etc. Under normal circumstances the support person should not perform extinguishing activities unless directly instructed by the Fire Brigade Leader.

5.4.3 Assist with the restoration of fire equipment after use in accordance with Fire Equipment Restoration section of this procedure.

6.0 INSTRUCTIONS 6.1 Assignment of Fire Brigade Personnel 6.1.1 The Unit 1 Fire Brigade consists of the following:

A. Unit 1 Fire Brigade Leader B. Three Fire Brigade Members from Unit 2 C. Fire Brigade Support Member from Security Force 6.1.2 The Unit 2 Fire Brigade consists of the following:

A. Unit 2 Fire Brigade Leader B. Three Fire Brigade Members from Unit 1 C. Fire Brigade Support Member from Security Force 6.2 The fire is reported to the Control Room of the affected unit.

6.2.1 The SM/CRS of the affected unit dispatches the Fire Brigade to the scene of the fire. The SM/CRS of the unaffected Unit will dispatch the Fire Brigade for zones identified in 1203.049/2203.049, Fires In Areas Affecting Safe Shutdown.

6.2.2 The Fire Brigade Leader of the affected unit responds and assumes command of the fire fighting activities.

6.2.3 The Fire Brigade Members from the unaffected unit respond along with the Security Fire Brigade Support Member. This comprises the initial fire fighting force.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0779 Rev: 0 Rev Date: 9/8/2009 Source: Direct Originator: S. Pullin TUOI: ANO-1-LP-RO-EDG Objective: 26 Point Value: I Section: 4.2 Type: Generic APEs System Number: 068 System

Title:

Control Room Evacuation

==

Description:==

Knowledge of the interrelations between the Control Room Evacuation and the following: ED/G KIA Number: AK2.07 CFR

Reference:

41.7/45.7 Tier: 1 RO Imp: 3.3 RO Select: Yes Difficulty: 2 Group: 2 SRO Imp: 3.4 SRO Select: Yes Taxonomy: K Question: RO: 24 SRO:

Given:

Fire has occurred in the Cable Spread Room Performing 1203.002 Alternate Shutdown CRS follow-up actions are to place

1. 1 EDG and #2 EDG are running and have been placed in a No DC start condition.

What protection is operable to the Emergency Diesel Generators?

A. Positive crankcase pressure trip B. Low lube oil pressure trip C. Mechanical over speed trip D. High jacket water temperature trip Answer:

C. Mechanical over speed trip Notes:

C will mechanically trip the fuel rack.

A and B require DC power to the emergency trip relay.

D does not exist.

References:

1104.036, Emergency Diesel Generator Operation, Change 049 History:

Direct Selected for 2010 RO/SRO exam

i PROC.IWORK PLAN NO.

I PROCEDUREIWORK PLAN TITLE: I PAGE: 38 of 271 1104.036 EMERGENCY DIESEL GENERATOR OPERATION CHANGE: 049 13.0 DG1 START WITHOUT DC CONTROL POWER CAUTION If fault condition that caused loss of DC is NOT removed, be aware that a fault may still be present and will have to be dealt with when presented.

NOTE Following sequence assumes no AC or DC is available.

13.1 IF known, THEN remove fault condition that caused loss of DC.

13.2 Place DG1 Engine Control Selector switch (HS5234) on C107 in MAINT.

CAUTION With loss of control power, the only functional DG protection is the mechanical overspeed device.

13.3 Open the following local breakers to prevent shutdown when DC power is restored:

• DG1 Local Field Flashing Power (Dlll6A)

(inside voltage regulator cabinet E-ll)

• DG1 Engine Control Power (D-lll4A).

(inside engine control panel Cl07)

NOTE

• Refer to ES Electrical System Operations (1107.002), Breaker Local Operation Without DC Control Power section, for manual operation of 4160 and 480 volt load center breakers.

• This is a serious condition and even if ESAS is required, ES signal must be overridden and de-energized.

13.4 To prevent full ES actuation upon restoration of power, deenergize ESAS digitals by opening following breakers:

• ESAS Panel C86 and C87 Breaker (RS14)

• ESAS Panel C9l and C92 Breaker (RS24)

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0349 Rev: 0 Rev Date: 9-7-99 Source: Direct Originator: D. Slusher TUOI: ANO-1-LP-RO-ELEC Objective: I IJ Point Value: I Section: 4.3 Type: B&W EOP/AOP System Number: A05 System

Title:

Emergency Diesel Actuation.

==

Description:==

Knowledge of the interrelations between the (Emergency Diesel Actuation) and the following:

Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

K/A Number: AK2.1 CFR

Reference:

CFR: 41.7/45.7 Tier: 1 RO Imp: 4.0 RO Select: Yes Difficulty: 3 Group: 3 SRO Imp: 3.8 SRO Select: Yes Taxonomy: C Question: RO:j 25 SRO:I 25 Diesel Generator #1 is running for a surveillance test.

Low reactor coolant system pressure causes a reactor trip and ESAS actuation.

What will the ES Electrical response be?

A. A-3 and A-4 powered from SU #1, both diesel generators running unloaded.

B. A-3 and A-4 powered from SU #1, Diesel Generator # I tripped, Diesel Generator # 2 running unloaded.

C. A-3 powered from Diesel Generator #1, A-4 powered from SU #1, Diesel Generator # 2 running unloaded.

D. A-3 powered from Diesel Generator #1, and A-4 powered from Diesel Generator #2.

Answer:

A. A-3 and A-4 powered from SU #1, both diesel generators running unloaded.

Notes:

A is correct, electrical response should be the normal response for an ESAS.

B is incorrect, nothing should trip #1 EDG.

C is incorrect, the #1 EDG output breaker should open on an ES signal.

D is incorrect, both busses should be powered from SU #1.

References:

STM 1-32, Rev. 33 History:

Used in 1999 exam.

Modified from ExamBank, QID# 453.

Selected for 2010 RO!SRO exam

Electrical Distribution STMI-32 Rev. 33 LOSS OF TRIPS DURING LOCAL MANUAL AUTOMATIC MANUAL BUS i3 DG VOLTAGE PARAL L EL C I OSE CLOSE CLOSE TRIP LOCK OUT LOCK OL T REGUI AT[ON OPERATION CLOSED

/ IN REMOTE 152-308 LOCAL CON IROL 186-A3 L CLOSED a (CLOSED WHEN SWITCH / I3USA3 IN REMOTE 7I OCK OUT OUTPUT BKR 152-308 CLOSED)

SWIICH 4 a

-L (OPEN / / _(CLOSED WHEN SWITCH PULL-TO-LOCK) 7 OUTPUT BK R Acy:.

T j

CLOS El))

I SYNCH 127 I)GI(K-19)

CONTROL SELECT 1 (CLOSED WHEN SWI1CH SWITCH DG OUTPUT LOCK OLJT+/-

LOCAL OLTAGE K2 1 R IS NORMAL) 286-GI 186-STI )86-ST2 (OPEN ON AUTO START)

I IA-112 IA-113 1 A-1l1 l25S-DGI _.._. a 152-410 a a 186-A3 186-DGI SYNCH

._-0 b C HEC K (CLOSED WHEN A3/A4 I 27-DO 1

• TIE BKR OPEN) 4 (K19)

(CLOSE ON 152-309 7 UNDERVOLTAGE) b L CLOSEI)

(CLOSED WHEN 152-308 3(19 IS OPEN) IN REMOTE a CLOS El) 127-A3 (ClOSED WHEN IN REMOTE x2 BKR 309 JLOSE.D WHEN Cl OsED)

A3 IS l)EAI)

CLOSE I1\ <-2450 VOLTSs)

TRIP EIGURE 32 72 DIESEL (IENERAT()R OUTPUT BREAKER A308 A408 193

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0780 Rev: 0 Rev Date: 9/09/2009 Source: New Originator: S.PuIIin TUOI: AILP-RO-AOP Objective: 4 Point Value: I Section: 4.3 Type: B&W EPEs/APEs System Number: A07 System

Title:

Flooding

==

Description:==

Ability to determine and interpret the following as they apply to the (flooding): adherence to appropriate procedures and operation within the limitations in the facilities license and amendments.

K/A Number: AA.2.2 CFR

Reference:

43.5/45.13 Tier: 1 RO Imp: 3.3 RO Select: Yes Difficulty: 3 Group: 2 SRO Imp: 3.7 SRO Select: Yes Taxonomy: K Question: RO:] 26 SRO:1 26 Given:

Plant power 100%

A Decay Heat pump OOS Dardanelle Lake Level 350 feet rising 1 ft/hr due to heavy rains Corps of Engineers predicts peak flood levels will reach 355 feet What action is required per Natural Emergencies procedure 1203.026 section 4 Flood?

A. Perform rapid plant shutdown per 1203.045 and align B Decay Heat pump for Decay Heat B. Perform rapid plant shutdown per 1203.045 and transfer plant auxiliaries to SU 2 transformer C. Trip Reactor and refer to 1202.001 and perform a Forced flow Cool down 1203.040 D. Trip Reactor and refer to 1202.001 and perform a Natural Circulation cool down 1203.013 Answer:

B. Perform rapid plant shutdown per 1203.045 and transfer plant auxiliaries to SU 2 transformer Notes:

B. is correct due to 1203. 025 directs you to perform a shutdown per 1203.045, and SU2 transformer is designed for flooding and should be used during a flood A. is incorrect 1203.025 directs you to perform a shutdown per 1203.045, and align a LPI pump for DH if both pumps are operable in this case A DH pump is OOS C. is incorrect the procedure does not call for a reactor trip but you should use rapid plant shut down and forced flow cool down D. is incorrect the procedure does not call for a reactor trip but you should use rapid plant shut down and forced flow cool down not Natural Circulation CD

References:

Natural Emergencies 1203.025 change 028 History:

New for 2010 RO/SRO exam

I I 0 o D)

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CHANGE 1203.025 NATURAL EMERGENCIES 028 PAGE 34 of 40 SECTION 4 FLOOD INSTRUCTIONS

1. Notify Unit 2 Control Room.

NOTE Information may be obtained from the Corps of Engineers throughout the implementation of this procedure at the following numbers:

• Dardanelle Lock and Dam Project Office 479-968-5008 ext. 241

• Dardanelle Lock and Dam Powerhouse 479-229-1863 (Fri-Sun use ext. 0)

• Little Rock District Engineer 501-324-5697

2. Establish contact with Corps of Engineers for peak flood condition forecasts and updates.

3. Notify Little Rock TOG Dispatcher.

4. Initiate lake level monitoring using one of the following methods:

NOTE SPDS displays level in feet. PMS/PDS displays level in inches above 324 reference level.

Instructions that follow give level in feet and corresponding level from PMS/PDS in brackets, e.g., 340 [PMS 192 in.]. At flood levels >349 [PMS 300 in.], SPDS and PMS/PDS are off-scale above sensor span.

• SPDS monitor SW pump aligned to lake (P-4A, P-4B1, P-4B2, P-4C)

• PMS/PDS monitor SW or Circ bay aligned to lake (SW bays L3664, L3666, L3668, and B & C Circ Bays L3601, L3602)

• WHEN 349 [PMS 300 in.],

THEN monitor lake level locally on an hourly basis.

5. WHEN directed by plant management, THEN begin plant shutdown per Rapid Plant Shutdown (f 203.045).

A. IF directed by plant management, THEN begin plant cooldown per Plant Shutdown and Cooldown (1102.010) or Forced Flow Cooldown (1203.040).

6. Notify Shift Manager to implement Emergency Action Level Classification (1903.01 0).

7. Initiate evaluation of plant risk in accordance with COPD-024, Risk Assessment Guidelines.

(continued)

CHANGE 1203.025 NATURAL EMERGENCIES 028 PAGE 35 of 40 SECTION 4 FLOOD

8. WHEN Lake Dardanelle level greater than 345 ft. (PMS 252 in.),

THEN perform Local Flooding Actions Attachment B of this procedure.

NOTE The Little Rock TOC Dispatcher will notify and call out personnel to install jumpers for breakers, switches and other equipment necessary for maintaining off-site power for shutdown and emergency operation.

9. Coordinate with Little Rock TOC Dispatcher and Unit 2 Control Room to initiate the following tasks:

NOTE Jumpers and associated hardware are located at Air Break Tower (B1217).

A. Installation of jumpers from the primary side of Startup Transformer (SU-2) directly to the 161KV transmission line.

B. Issuance of switching orders to allow work on Startup Transformer SU-2.

C. De-energize and bypass SU-2 Voltage Regulator.

10. IF both decay heat removal ioops are available, THEN align one loop for decay heat removal as follows:

A. Ensure that one decay heat ioop is aligned for ES standby (LPI) per Decay Heat Removal Operating Procedure (1104.004), Attachment A.

B. Align the opposite decay heat ioop for DH removal per 1104.004, Decay Heat Removal During Cooldown section.

1) Align DH system AUX spray per 1104.004, Depressurizing RCS Using DH System AUX Spray section.

11. if only one decay heat removal loop is available, THEN verify loop is aligned for ES standby (LPI) per 1104.004, Attachment A.

A. WHEN plant cooldown is at the point of switching to decay heat, THEN align the available ioop for DH removal per 1104.004, Decay heat Removal During Cooldown section.

B. IF DH system is NOT accessible, THEN continue RCS cooldown utilizing steam generators ensure RC pressure is maintained as per NPSH curve for RC pump operation.

(continued)

CHANGE 1203.025 NATURAL EMERGENCIES 028 PAGE 36 of 40 SECTION 4-- FLOOD (Continued)

12. Remove equipment from service AND de-energize power supplies to below-grade equipment prior to flooding.

NOTE At flood levels >349 [PMS 300 in.], SPDS and PMS/PDS are off-scale above sensor span.

Level must be observed locally.

13. Prior to flood waters exceeding elevation 354, perform the following:

A. Secure nonessential electrical loads.

B. Verify all necessary work is completed on SU 2.

C. Coordinate with Unit 2 Control Room to transfer plant auxiliaries to SU 2 using Electrical System Operation (1107.001), Startup Transformer Operations section.

14. For each component verified in position Attachment B, install a Caution Tag stating, This component is positioned for Unit I flooding concerns. Contact the Unit I Control Room prior to repositioning.

15. Annotate on the Shift Turnover Sheet that verification of Attachment B of 1203.025 is required daily while Lake Dardanelle is greater than 345 ft.

16. Conduct further operations as directed by plant management.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0595 Rev: 0 Rev Date: 9/09/2009 Source: New Originator: S.PuIIin TUOI: A1LP-RO-RCS Objective: 26 Point Value: I Section: 4.3 Type: B&W EPEs/APEs System Number: E13 System

Title:

EOP Rules and Enclosures

==

Description:==

Knowledge of limiting conditions for operation and safety limits.

K/A Number: 2.2.22 CFR

Reference:

41.5/43.2/45.2 Tier: 1 RO Imp: 4.0 RO Select: Yes Difficulty: 3 Group: 2 SRO Imp: 4.7 SRO Select: Yes Taxonomy: C Question: RO:I 27 SRO:

In accordance with Technical Specification bases, what is the purpose of the Code Safeties and what is the design bases accident that defines their minimum capacity?

A. The Code Safeties prevent exceeding the safety limit of 2500 psig during a 100% load rejection without a reactor trip.

B. The Code Safeties prevent exceeding the safety limit of 2750 psig during a 100% load rejection without reactor trip.

C. The Code Safeties prevent exceeding the safety limit of 2750 psig during a startup accident.

D. The Code Safeties prevent exceeding the safety limit of 2500 psig during a startup accident.

Answer:

C. The Code Safeties prevent exceeding the safety limit of 2750 psig during a startup accident.

Notes:

Answer C is correct, it lists the proper safety limit and the design basis accident.

Answer A is incorrect, it lists the safety setpoint (not the safety limit) and a plausible, but incorrect, accident.

Answer B is incorrect, it lists the proper safety limit and a plausible, but incorrect, accident.

Answer D is incorrect, it lists the safety setpoint (not the safety limit) and the design basis accident.

References:

Technical Specifications bases 82.1.2 amendment #215 History:

New for 2010 RO/SRO exam

RCS Pressure SL B 2.1.2 B 2.0 SAFETY LIMITS (SLs)

B 2.1.2 Reactor Coolant System (RCS) Pressure SL BASES BACKGROUND In SAR, Section 1.4 (Ref. 1), GDC 14, Reactor Coolant Pressure Boundary (RCPB), and GDC 15, Reactor Coolant System Design, address RCPB design and protection, respectively. The ANO-1 discussion regarding how GDC 15 is accomplished states that analysis and evaluation of all normal and abnormal operating conditions and transients are integrally related to all RCS and associated systems design. SAR Chapter 14 (Ref. 2) lists these abnormal operating conditions and transients and terms them abnormalities. In addition, GDC 28, Reactivity Limits (Ref. 1), specifies that reactivity accidents including rod ejection do not result in damage to the RCPB greater than limited local yielding.

The design pressure of the RCS is 2500 psig. During normal operation and abnormalities, the RCS pressure is kept from exceeding the design pressure by more than 10% in order to remain in accordance with the design codes (Ref. 3 and 4).

Hence, the safety limit is 2750 psig. To ensure system integrity, all RCS components were hydrostatically tested at 125% of design pressure prior to initial operation, according to the design code requirements. Inservice leak testing at not less than 2155 psig is also required, prior to MODE 2, following any opening of the reactor coolant system in accordance with ASME code, Section Xl; IWA-5000. When performed at the end of refueling outages, this leak test also satisfies the requirements of IWB-2500, Table IWB-2500-1; Category B-P items B15.10, B15.20, B15.30, B15.40, B15.50, B15.60, and B15.70 for all Class I pressure retaining components (Ref. 5).

APPLICABLE SAFETY ANALYSIS The RCS pressurizer safety valves, operating in conjunction with the Reactor Protection System trip settings, ensure that the RCS pressure SL will not be exceeded.

The RCS pressurizer safety valves are sized to prevent system pressure from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME code for Nuclear Power Plant Components (Ref. 3). The design basis transient that is most influential for establishing the required relief capacity, and hence the valve size requirements and lift settings, is a rod withdrawal event from low power.

The startup event analysis (rod withdrawal at low power) (Ref. 2) is performed using conservative assumptions relative to pressure control devices.

ANO-1 B 2.1.2-1 Amendment No. 215

RO Written Exam Tier 2 Group I

ES-401 PWR Examination Outline Form ES-401-2 PWR Examination Outline Form ES-401 -2 PtanLSems -Tier 2/Group 1 (RO)

K K K K K K A A A A G KIATopic(s) lR # OlD T 1234561234 y p

e K5.05 The dependency of 003 Reactor Coolant Pump X X 2.8* 28 781 RCS flow rates upon the N number of operating RCPs A4.08 RCP cooling water 3.2 29 782 M supplies 2.1.34 changed to 2.2.38 004 Chemical and Volume X Knowledge of conditions and 3.6 30 796 N Control limitations in the facility license K4.03 Protection of ion x exchangers (high letdown 2.8* 31 259 D temperatures will isolate ion 005 Residual Heat Removal X 1(2.01 RHR Pumps 3.0* 32 786 M 006 Emergency Core Cooling 1(6.10 Valves 2.6 33 783 M 007 Pressurizer Relief/Quench X 1(5.02 Method of forming a 3.1 34 561 D Tank steam bubble in the PZR 008 Component Cooling Water X A2.08 changed to A2.01 - 3.3 35 787 N 010 Pressurizer Pressure Control X 1(3.02 RPS 4.0 36 788 N

012 Reactor Protection X K6.10 Permissive circuits 3.3 37 784 N

X 2.1.32Abilityto explain and 3.8 38 785 N apply system limits and precautions 013 Engineered Safety Features X 1(4.10 Safeguards 3.3 39 144 D

Actuation equipment control reset 022 Containment Cooling X A3.01 Initiation of 4.1 40 135 D safeguardsmodeofoperation 025 Ice Condenser Not Selected N/A 026 Containment Spray X KI .01 - ECCS 4.2 41 78 D 039 Main and Reheat Steam X A2.04 Malfunctioning steam 3.4 42 202 D dump 059 Main Feedwater X A3.03 Feed water pump 2.5 43 195 D

suction flow pressure K4.16 Automatic trips for X MFW pumps 3.1 44 789 N 061 Auxiliary/Emergency X A1.04 changed to A1.01 3.9 45 270 D Feedwater S/G level PWR Examination Outline Form ES-401-2 Plant Systems Tier 2/Group 1 (RO)

ES-401 1 Form ES-.401-2

ES-401 PWR Examination Outline Form ES-401-2 K K K K K K A A A A G K/A Topic(s) IR # ID Ty 1234561234 e 062 AC Electrical Distribution X 2.4.35 Knowledge of local 3.8 46 790 N auxiliary operator tasks during an emergency and the resultant operational effects.

x K2.0lMajorsystem loads 47 316 D 063 DC Electrical Distribution X K3.02 Components using 3.5 48 86 D DC control ower 064 Emergency Diesel Generator X K2.0l Air compressor 2.7 49 791 N X KI .05 Starting air system 3.4 50 792 N 073 Process Radiation X K5.0l Radiation theory, 2.5 51 672 R Monitoring including sources, types, units, and effects 076 Service Water X A4.02 SWS valves 2.6 52 793 D X Al .02 Reactor and turbine 2.6 53 794 N building closed cooling water tem eratures 078 Instrument Air X Kl.03 changed to Kl.02 2.7 54 535 D Service air 103 Containment X A4.06 Operation of the 2.7 55 795 D

containment ersonnel airlock K/A Catego Point Totals: 3 3 2 3 3 2 2 2 2 3 3 Group Point Total: 28 ES-401 Form ES-401-2

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0781 Rev: 0 Rev Date: 9/09/2009 Source: New Originator: S. Pullin TUOI: AILP-RO-ICS Objective: 26 Point Value: I Section: 3.4 Type: Heat Removal from Reactor Core System Number: 003 System

Title:

Reactor Coolant Pump

Description:

Knowledge of the operational implications of the following concept as they apply to the RCP:

The dependency of RCS flow rates upon the number of operating RCPs K/A Number: K5.05 CFR

Reference:

41.5/45.7 Tier: 2 RO Imp: 2.8 RO Select: Yes Difficulty: 3 Group: 1 SRO Imp: 3.0 SRO Select: Yes Taxonomy: C Question: 28 28 RO:J SRO:1 Given:

Plant 60% power All RCPs are in service A OTSG BTU LIMIT (K07-E2) alarm is received What is the most likely cause of the alarm?

A. A Thot temperature instrument failing high B. A Feed water temperature instrument failing high C. A OTSG pressure instrument failing low D. A RCS flow instrument failing low Answer:

D. A RCS flow instrument failing low Notes:

D. is correct RCS flow has the largest input to BTU limit A. is incorrect although the Thot feeds this alarm the instrument is failing in the wrong direction B. is incorrect although the Feed water instrument feeds this alarm the instrument is failing in the wrong direction C. is incorrect although the OTSG instrument feeds this alarm the instrument is failing in the wrong direction this one is hard to figure out due to BTU limits is looking at 35 f of superheat as SG pressure goes down with RCS flow and the other instruments staying the same superheat is getting higher

References:

1203.012F change 028 STM 1-64 rev 10 History:

New for 2010 RO/SRO exam

i PROC IWORK PLAN NO.

I PROCEDURE/WORK PLAN TITLE: I PAGE: 12 c 1203.012F ANNUNCIATOR K07 CORRECTIVE ACTION CHANGE: 028 Location: C13 Device and Setpoint: N/A A OTSG BTU LIMIT Alarm: K07-E2 1.0 OPERATOR ACTIONS

1. Verify ICS is not raising load.

2. Check for possible instrument failure.

A. IF there is an ICS input signal failure, THEN GO TO ICS Abnormal Operation (1203.001)

B. IF failure indicated, THEN select an alternate instrument.

3. IF alarm occurs during End of Cycle T-ave reduction, THEN determine Main Steam superheat using 1102.004 Attachment Q.

A. IF Main Steam superheat drops to 35°F, THEN stop Tave reduction AND consult Rx Engineering.

NOTE The parameter causing the BTU limit alarm may not be readily apparent.

Other indications such as cross limits or feedwater-reactor limited may help determine the cause.

4. IF valid BTU limit condition exists, AND NOT due to Tave reduction, THEN either raise reactor power or lower feedwater demand (or both) as necessary to clear alarm.

5. IF necessary, THEN initiate steps to repair ICS or input transmitters.

2.0 PROBABLE CAUSES NOTE The BTU limit is derived from SG heat capacity and superheat considerations.

1. BTU Limit = (Thot + FWtemp + PresssG - 200) RC flow (%)

3.0 REFERENCES

Schematic Diagram Annunciator K07 (E-457)

Integrated Control System STM 1-64 Rev. 10 valves and pump will return to the mode of control previously described.

2.6.3.1 Feedwater Pump With one FW Pump running. the Main FW Block valves are closed Control, and the crossover valve is open. The 70 psid setpoint is being compared to the low auctioneered iW signal. The z\P error signal is used to adjust the respective main feedwater loop demand signal to adjust pump speed to ke the lowest 1 P at setpoint.

When both FW Pumps are running with the crossover valve closed and both main block valves closed, each FW Pump is controlled by its own individual loop P summed with its loop demand signal.

If both FW Pumps are running, with the crossover valve closed, and both main block valves open, each FW Pump is controlled by its respective loop flow error summed with its feedforward loop demand signal.

A characteristic of the ICS is that there are numerous tie-back schemes which enable the ICS to have bump-less transfers. With the main feedwater pumps the tie-back scheme works well for the A pump controls but potentially can cause a feedwater transient when placing the B pump in Auto. When both feed pumps are in HAND with the main block valves closed and the cross-tie valve open, the selected z\P controller looks at the status of the A pump to determine which manual demand to track. If the A pump is latched, the selected zIP controller will have been tracldng the A pump demand signal. If the A pump is tripped, manual demand for the B pump will have been tracked. Thus, if the B pump is placed in AUTO first with the A pump just latched or latched and rolling at minimum speed, the B pump would be driven down toward the minimum demand of the A pump. To address this idiosyncrasy, a caution was placed in the Condensate, Feedwater and Steam System Operation procedure which states: With both MFW Pumps in manual and the Feedwater Pumps Disch Crosstie (CV-2827) open, placing B MFW Pump in AUTO with a significant difference in demand signals between A and B MFW Pumps will cause a feedwater transient.

2.6.4 BTU Limits. The purpose of BTU limits is to monitor for a minimum of 35°F of superheat in the steam leaving the OTSG. To insure that moisture does not carryover from the OTSG to the turbine generator, it is desirable to have a minimum of 35°F of superheat in each pound mass of steam. ICS monitors the superheat of the steam indirectly by monitoring four parameters and calculating the maximum loop feedwater flow allowable. If the loop feedwater demand is greater than the calculated limit, a BTU limit alarm is sounded to alert the operator. The four parameters used in the BTU limit calculation are:

Selected TH Individual OTSG steam pressure Loop feedwater temperature 38

Integrated Control System STM 1-64 Rev. 10 RCS flow in that loop.

Each BTU limit calculator takes these four parameters and determines what the maximum loop feedwater demand is that will not drop superheat to < 35°F. (Refer to figure 64.25)

BTU Limit [(TH + OTSG press + FW Temp) -2001 x RCS Flow %.

RCS flow changes have the largest effect on the calculation, and note that the limit is lowered by either RCS flow, TH, or feedwater temperature being lowered.

FEED W9TER LIMIT

,OFFIJLL POWER POW FLOW 50 1011 RC FLOW T HOT 574 57% 380 505 )11 595 65O 1,05 F PSTEAM 11W) 990 3311 lOll)) 1055 11(10 115(1 200 P010 T FW 151) 7)0) 250 311)) 35)) 4101 451) 5)))) 1)

FIGURE 64.25: BTU LIMIT CURVE Lowering feedwater temperature means that more of the primaiy heat is used to raise the feedwater to saturation temp. Therefore, less energy is available for superheat. The highest value that steam temperature out of an OTSG can be is to approach TH. Therefore, if TH decreases and OTSG saturation temperature is constant, superheat would decrease. If a constant feedwater flow to the OTSG is maintained, less RCS flow means less BTU of heat available to an OTSG and therefore superheat would decrease. If OTSG pressure increases, then saturation temperature will increase, if steam outlet temperature is constant, then superheat will decrease.

2.6.5 High Level Limit The purpose of high level limit is to prevent flooding aspirating steam ports in the OTSG. Operate level for each OTSG is compared to the high level limit setpoint (90%) and an error signal is generated. If that signal is less than the loop feedwater flow error signal, then the low 39

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0782 Rev: 0 Rev Date: 9/09/2009 Source: Modified Originator: S. Pullin TUOI: A1LP-RO-RCS Objective: 23 Point Value: 1 Section: 3.4 Type: RCS Heat Removal System Number: 003 System

Title:

Reactor Coolant Pump

Description:

Ability to manually operate and/or monitor in the control room: RCP cooling water supplies K/A Number: A4.08 CFR

Reference:

41.7/45.5 to 45.8 Tier: 2 RO Imp: 3.2 RO Select: Yes Difficulty: 3 Group: I SRO Imp: 2.9 SRO Select: Yes Taxonomy: C Question: 29 RO: SRO:r Given:

- Plant heat up in progress from refueling outage.

- P-32C and P-32D RCPs are running.

- Seal injection block CV-1 206 is in override for testing

- Seal injection flow has been balanced and is in auto at 16 gpm total flow.

- Non-nuclear ICW to RCP motor cooling flow is 200 gpm.

- Nuclear ICW to RCP seal cooling flow is 35 gpm.

- RCS loop A & B cold leg temps are 275°F.

- RCP lift oil pressure is 1800 psig.

A start of RCP P-32A is attempted but is unsuccessful. Why?

A. Nuclear ICW to RCP seal cooling flow is low.

B. Seal injection flow is low.

C. RCP lift oil pressure is low.

D. RCP motor cooling flow is low.

Answer:

D. RCP motor cooling flow is low.

Notes:

D. is correct to satisfy the starting interlock RCP motor cooling flow needs to be >250 gpm A is incorrect, nuclear ICW to RCPS is greater than 30 gpm.

B is incorrect, seal injection flow is greater than 3 gpm to each RCP.

C is incorrect, RCP lift oil pressure is >1750 psig

References:

1103.006 change 032 History:

Modified from QID 559 Selected for 2010 RO/SRO exam

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0559 Rev: I Rev Date: 1218/06 Source: Direct Originator: Cork/Possage TUOI: AILP-RO-RCS Objective: 23 Point Value: 1 Section: 3.4 Type: RCS Heat Removal System Number: 003 System

Title:

Reactor Coolant Pump System

Description:

Knowledge of the physical connections and/or cause-effect relationships between the RCPS and the following systems: RCP bearing lift oil pump.

KIA Number: Ki .13 CFR

Reference:

41.2 to 41.9 / 45.7 to 45.8 Tier: 2 RO Imp: 2.5 RO Select: No Difficulty: 3 Group: I SRO Imp: 2.5 SRO Select: No Taxonomy: M Question:

RO:j SRO:J Given:

- Plant heatup in progress from refueling outage.

- P-32A and P-32B RCPs are running. Q..d i) 1

- Seal injection flow has been balanced and is in auto at 16 gpm total flow.

- Non-nuclear ICW to RCP motor cooling flow is 275 gpm.

- Nuclear ICW to RCP seal cooling flow is 35 gpm.

- RCS loop A & B cold leg temps are 370°F.

- RCP lift oil pressure is 1600 psig.

A start of RCP P-32C is attempted but is unsuccessful. Why?

A. Nuclear ICW to RCP seal cooling flow is low.

B. Seal injection flow is low.

C. RCP lift oil pressure is low.

D. RCS cold leg temps are low.

Answer:

C. RCP lift oil pressure is low.

Notes:

C is correct, lift oil pressure must be> 1750 psig for pump start interlock to be met.

D is incorrect, RCS cold legs must be greater than 375°F to start the fourth RCP, not the third.

A is incorrect, nuclear ICW to RCPS is greater than 30 gpm.

B is incorrect, seal injection flow is greater than 3 gpm to each RCP.

References:

1103.006, Chg. 026-01-0 History:

New for 2005 RO exam by Pullin, but not used. Modified version of 615.

New for 2007 RO Exam.

PROC.IWORK PLAN NO. I PROCEDUREIWORK PLAN TITLE: I PAGE: 6 of 56 1103.006 REACTOR COOLANT PUMP OPERATION CHANGE: 032 5.28 During cooldown, the following RCP limits apply:

• <271°F no more than two RCPs may be operated

<166°F no RCP5 may be operated 5.29 During heatup, the following RCP limits apply:

• <241°F no more than two RCPs may be operated

• <316°F no more than three RCPs may be operated, however due to hydraulic lift of the core, no more than three RCPs may be operated until RCS temperature is >430°F

. <106°F no RCPs may be operated 5.30 RCP motor and pump vibration limits are as follows:

• P-32B or D motor vibration; more than one channel >20 mils after startup stabilization

• P-32A or C motor vibration; more than one channel >0.8 in/sec after startup stabilization

• RC pump vibration; more than one channel >25 mils after startup stabilization 5.31 Plant startup conditions could result in exceeding the Steam Generator Design Limit of 60°F Tube to Shell T (tubes hotter) 5.32 Simultaneous operation of the normal and Emergency HP Oil Lift Pump (P-63 and P-80) is undesirable. Reduced oil pressure and cavitation can occur.

6.0 SETPOINTS The following conditions must be satisfied to start an RCP from the control room.

6.1 Rx power <22%.

6.2 RCP seal injection flow >3 gpm.

If <3 gpm, alarms RCP SEAL INJ FLOW LO (K08-A7).

RCP P-32A Seal Injection Flow (FS-1280)

RCP P-32B Seal Injection Flow (FS-128l)

RCP P-32C Seal Injection Flow (FS-l282)

RCP P-32D Seal Injection Flow (FS-l283) 6.3 RCP motor cooling flow >250 gpm (non-nuclear ICW).

If <250 gpm alarms RCP MOTOR COOLING FLOW LO (K08-E6).

P-32A MTR Air LO CLR ICW RTN Flow (PDIS-2260)

P-32B MTR Air LO CLR ICW RTN Flow (PDIS-2261)

P-32C MTR Air LO CLR ICW RTN Flow (PDIS-2262)

P-32D MTR Air LO CLR ICW RTN Flow (PDIS-2263)

i PLAN NO.

I PROCEDUREIWORK PLAN TITLE:

PAGE: 7 of 56 1103.006 REACTOR COOLANT PUMP OPERATION CHANGE: 032 6.4 RCP seal cooling flow >30 gpm (nuclear ICW)

If <30 gpm alarms RCP SEAL COOLING FLOW LO (K08-E7).

P-32A Seal CLR ICW RTN Flow (PDIS-2250)

P-32B Seal CLR ICW RTN Flow (PDIS-225l)

P-32C Seal CLR ICW RTN Flow (PDIS-2252)

P-32D Seal CLR ICW RTN Flow (PDIS2253) 6.5 RCP start interlock on low oil reservoir level 6.5.1 Upper Reservoir Oil Level Low

-2.0 for P-32A, C, and D

-1.6 for P-32B RCP A Upper Lube Oil Level Lo (LS-6535)

RCP B Upper Lube Oil Level Lo (LS-6536)

RCP C Upper Lube Oil Level Lo (LS-6537)

RCP IJ Upper Lube Oil Level Lo (LS-6538) 6.5.2 Lower Reservoir Oil Level Low

-1.5 for P-32A, C, and D

-1.2 for P-32B RCP A Lower Lube Oil Level Lo (LS-6560)

RCP B Lower Lube Oil Level Lo (LS-6561)

RCP C Lower Lube Oil Level Lo (LS-6562)

RCP D Lower Lube Oil Level Lo (LS-6563) 6.6 Computer alarms on high and low oil reservoir level 6.6.1 Upper Reservoir Oil Level High

+2.0 for P-32A, C, and D

+1.6 for P32B 6.6.2 Upper Reservoir Oil Level Low

-2.0 for P-32A, C, and D

-1.6 for P-32B 6.6.3 Lower Reservoir Oil Level High

+1.5 for P-32A, C, and D

+1.2 for P-32B 6.6.4 Lower Reservoir Oil Level Low

-1.5 for P-32A, C, and ID

-1.2 for P-32B 6.7 RCP HP oil lift pressure >1750 psig.

If <1750 psig alarms RCP LIFT OIL TROUBLE (K08-CB)

(1000 psig for P-32B)

RCP P-32A HP Lift Oil Press (PS-6530).

RCP P-32B HP Lift Oil Press (PS-6526).

RCP P-32C HP Lift Oil Press (PS-6532)

RCP P-32D HP Lift Oil Press (PS-6533)

I PROC.IWORK PLAN NO.

I PROCEDUREIWORK PLAN TITLE: I PAGE: 8 of 56 1103.006 REACTOR COOLANT PUMP OPERATION CHANGE: 032 6.8 RCP reverse rotation <12.7 gpm return oil flow/pump start permitted If >12.7 gpm alarms plant computer (not applicable for P-32B)

RCP P32-A REVERSE ROTATION Computer Alarm (FSE51O)

RCP P-32A Reverse Rotation Starting Interlock (FS-6515).

RCP P32-C REVERSE ROTATION Computer Alarm (FS6512)

RCP P-32C Reverse Rotation Starting Interlock (FS-65l7)

RCP P32-D REVERSE ROTATION Computer Alarm (FS6513)

RCP P-32D Reverse Rotation Starting Interlock (FS-6518) 6.9 If starting first RCP, RCS to SG Downcomer iT 50°F.

RC Loop A Cold Leg Temp (TS-l017)

RC Loop B Cold Leg Temp (TS-1045)

A Stm Gen Downcomer Temp (TI-2665)

B Stm Gen Downcomer Temp (TI-2615) 6.10 If starting third RCP, RCS temperature >241°F.

RC Loop A Cold Leg Temp (TS-l017)

RC Loop B Cold Leg Temp (TS-l045) 6.11 If starting fourth RCP, RCS temperature >430°F.

RC Loop A Cold Leg Temp (TS-l0l7)

RC Loop B Cold Leg Temp (TS-l045)

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0796 Rev: 0 Rev Date: 9/15/2009 Source: New Originator: S. Pullin TUOI: Al LP-RO-TS Objective: 5 Point Value: 1 Section: 3.2 Type: Reactor Coolant System Inventory Control System Number: 004 System

Title:

Chemical and Volume Control System (CVCS)

==

Description:==

Knowledge of conditions and limitations in the facility license.

KIA Number: 2.2.38 CFR

Reference:

41 .7/41.10/43.1/45.13 Tier: 2 RO Imp: 3.6 RO Select: Yes Difficulty: 3 Group: I SRO Imp: 4.5 SRO Select: Yes Taxonomy: C Question: RO:I 30 SRO:F30 REFERENCE PROVIDED Which of the following Boric Acid Addition Tank level and concentration versus RCS Tave would require entry into TRM 3.5.1 ?

A. 8,700 ppm Boron, BAAT level 36 inches 400 F Tave B. 9,500 ppm Boron, BAAT level 46 inches 450 F Tave C. 10,000 ppm Boron, BAAT level 50 inches, 500 F Tave D. 12,000 ppm Boron, BAAT level 56 inches, 550 F Tave Answer:

C. 10,000 ppm Boron, BMT level 50 inches 500 F Tave Notes:

C. is correct due to the values fall below and to the right of reference curve TRM figure 3.5.1-1 A, B, and D are incorrect due to the values fall above and to the left of reference curve TRM figure 3.5.1-1 REFERENCE PROVIDED FOR THIS QUESTION

References:

1104.003 change 046 TRM 3.5.1 rev 16 History:

New for 2010 RO/SRO exam

Makeup and Chemical Addition Systems 3.5.1 TRM 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

TRM 3.5.1 Makeup and Chemical Addition Systems TRO 3.5.1 The Makeup and Chemical Addition System shall be OPERABLE with the following requirements:

a. Two makeup pumps shall be OPERABLE except as specified in TS 3.5.2, Emergency Core Cooling Systems (ECCS) Operating, and TS 3.5.3, Emergency Core Cooling Systems (ECCS) Shutdown,

b. The boric acid addition tank (BAAT) shall be OPERABLE, containing at least the equivalent of the boric acid volume and concentration requirements of TRM Figure 3.5.1-1, Boric Acid Addition Tank Volume and Concentration Vs RCS Average Temperature as boric acid solution with a temperature of 10°F above the crystallization temperature for the concentration in the tank, and

c. One boric acid pump associated with the BAAT shall be OPERABLE.

d. System piping and valves necessary to establish a flow path from the boric acid addition tank to the makeup system shall be OPERABLE and shall have a temperature of 10°F above the crystallization temperature for the concentration in the tank.

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

I.

Condition entry is not required when the flow path from the boric acid pump(s) to the Makeup Tank is unavailable during procedurally controlled activities.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of TRO not A.1 Restore Makeup and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> met. Chemical Addition System to OPERABLE status.

B. Required Action and B.1 Initiate a condition report to Immediately associated Completion document the condition Time not met. and determine any limitations for continued operation of the plant.

ANO-1 TRM 3.5.1-1 Rev.

I PROCJWORK PLAN NO.

1104.003 I PROCEDUREIWORK PLAN TITLE:

CHEMICAL ADDITION I PAGE: 67 of 127 CHANGE: 046 ATTACHMENT H Page 1 of 1 Volume of BAAT vs. Depth of Liquid 8000

-H t-**+ -H-HH-H- Ht t-7000  : I

-:-- -:----ttt- -:--4---:-H- h-H-:- --:----- --:J--H-H 6000 5000 U,

E C

4000 (U

C, 3000 2000 1000 E

0 0 20 30 40 50 60 70 80 90 100 Depth in Inches (LT-1604) 1.0 To calculate the BAAT (T-6) level drop corresponding to a certain feed volume:

1.1 Read initial BAAT level and determine initial volume from graph.

1.2 Subtract feed volume from initial tank level.

Example: It is desired to feed 530 gallons of boric acid.

A. Initial BAAT level = 82. (From graph, 7100 gal.)

B. Initial volume - feed volume = 7100 - 530 = 6570 gal.

C. Final level, from graph, corresponding to 6570 gal. = - 74.

J PROCJWORK PLAN NO.

J PROCEDUREIWORK PLAN TITLE: I PAGE: 66 of 127 f 1104.003 CHEMICAL ADDITION ATTACHMENT G CHANGE: 046 Page 1 of 1 BAAT Volume and Concentration Vs. RCS T-ave (Ref. TRM Figure 3.5.1-1) 8000 7000 6000 5000 4000 3000 2000 1000 0

200 300 400 500 600 8700 ppm 09500 ppm 10,000 ppm 12,000 ppm

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0259 Rev: 0 Rev Date: 9-2-99 Source: Direct Originator: D. Slusher TUOI: ANO-1-LP-RO-MU Objective: 07 Point Value: 1 Section: 3.2 Type: Reactor Coolant System Inventory Control System Number: 004 System

Title:

Chemical and Volume Control System

==

Description:==

Knowledge of CVCS design feature(s) and/or interlock(s) which provide for the following:

Protection of ion exchangers (high letdown temperature will isolate ion exchangers)

K/A Number: K4.03 CFR

Reference:

CFR: 41.7 Tier: 2 RO Imp: 2.8 RO Select: Yes Difficulty: 2 Group: I SRO Imp: 2.9 SRO Select: Yes Taxonomy: K Question: RO:1 3 31 SRO:J What is the function of the temperature interlock associated with RCS letdown?

A. Prevents letdown fluid from flashing to steam when pressure is reduced by closing CV-1221 (letdown isolation).

B. Prevents exceeding letdown piping thermal limits by shutting CV-1213 & 1215 (letdown cooler inlet MOV).

C. Prevents degrading T36AIB resin by shutting CV-1221 (letdown isolation).

D. Prevents exceeding letdown cooler capacity by shutting CV-1 213 & 1215 (letdown cooler inlet MOV).

Answer:

C. Prevents degrading T36A!B resin by shutting CV-1221 (letdown isolation).

Notes:

A is incorrect, this is the function of the letdown coolers.

B is incorrect, interlock doesnt close the inlets and piping limits will not be exceeded before the resin is damaged.

C is correct D although the letdown cooler capacity is exceeded when temperature is exceeded the interlock doesnt close the inlet valves.

References:

1104.002 Rev 051-02-0 STM1-04 Rev 5 History:

Used in 1999 exam.

Selected for 2010 RO/SRO exam

i PROCJWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: I PAGE: 10 of 359 1104.002 MAKEUP & PURIFICATION SYSTEM OPERATION CHANGE: 065 5.0 LIMITS AND PRECAUTIONS 5.1 Do not start or continue to run a Makeup Pump (P36A, B or C) with the RCS in a solid water condition except as directed by emergency procedures.

5.2 Maintain Makeup Tank (T4) pressure above 10 psig.

(4.3.3) 5.3 Restricting flow through an operating Makeup Pump (P-36A, P-36B or P-36C) to < 55 gpm can cause pump damage. Accounting for instrument accuracy, HPI should be maintained such that flow through at least one HPI line is maintained 90 gpm when recirc valve is closed.

5.4 Maximum flow through a makeup pump is 500 gpm for normal operation.

5.5 Maximum flow through a makeup pump is 525 gpm for emergency operation.

5.6 Maximum flow through a Letdown Cooler (E-29A & B) is 87.5 gpm per cooler.

5.7 Allowing flow through a primary Makeup Filter (F-3A or F3B) in excess of 80 gpm can lead to filter damage.

5.8 Allowing flow through a purification DI in excess of 123 gpm can compact the resin and restrict letdown flow.

5.9 Restricting flow through a Purification Demineralizer (T36A or T36B) to < 25 gpm can cause channeling of resin and reduce efficiency of demineralizer.

5.10 Maximum purification demineralizer inlet temperature is 135°F.

5.11 Placing a purification demineralizer into service that has not been borated will result in a reduction in RCS boron concentration.

5.12 Ensure clean waste system is aligned to receive waste from letdown system prior to positioning Letdown 3-Way Valve (CV-1248) to BLEED.

Otherwise letdown line will overpressurize.

5.13 When makeup pumps are subject to HPI actuation, maintain MU tank pressure/level relationship within limit of Exhibit A. Exceeding the limit reduces the time available for isolating the MU tank after HPI actuation.

5.14 When venting the makeup tank, the waste gas system shall be aligned to compress the gas for storage unless samples indicate negligible activity in the makeup tank.

5.15 MU Tank T4 Relief Valve (PSV1249) is not designed to relieve water.

For uncontrollable high MU tank water level, open MUT Vent Valve (CV1257)

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0786 Rev: 0 Rev Date: 9/14/2009 Source: Modified Originator: S. Pullin TUOI: Al LP-RO-ELECD Objective: 11 Point Value: I Section: 3.4 Type: Heat Removal From Reactor Core System Number: 055 System

Title:

Residual Heat Removal System

==

Description:==

Knowledge of bus power supplies to the following: RHR pumps.

KIA Number: K2.01 CFR

Reference:

41.7 Tier: 2 RO Imp: 3.0 RO Select: Yes Difficulty: 3 Group: 1 SRO Imp: 3.2 SRO Select: Yes Taxonomy: K Question: RO:j 32 32 SRO:j Given:

- Plant is in Mode 6

- P-34B Decay Heat pump is running Which of the following would cause a loss of Decay Heat Removal?

A. A-i voltage of 2475 volts B. A-2 voltage of 2475 volts C. B-5 voltage of 428 volts D. B-6 voltage of 428volts Answer:

D. B-6 voltage of 428volts Notes:

B Decay Heat Removal Pump is powered from A-4 via A-2. An undervoltage on the A buses or B buses will trip A-409 (A4 feeder breaker). The undervoltage setpoint for A-4 is 2450 volts. The undervoltage setpoint for B-6 is 429 volts. Therefore, a,b, and c are incorrect.

References:

OP-1107.002 Change 025 History:

Modified from QID 0293 Selected for 2010 RO/SRO Exam

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0293 Rev: 2 Rev Date: 12/14/06 Source: Direct Originator: D Slusher TUOI: Al LP-RO-ELECD Objective: 11 Point Value: 1 Section: 3.4 Type: Heat Removal From Reactor Core System Number: 005 System

Title:

Residual Heat Removal System

==

Description:==

Knowledge of bus power supplies to the following: RHR pumps.

K/A Number: 1(2.01 CFR

Reference:

41.7 Tier: 2 RO Imp: 3.0 RO Select: No Difficulty: 3 Group: I SRO Imp: 3.2 SRO Select: No Taxonomy: Ap Question: RO:j SRO:j Given:

- Plant is in Mode 5

- P-34A Decay Heat pump is running Which of the following would cause a loss of Decay Heat Removal?

A. A-I voltage of 2425 volts Q%J%)

B. A-2 voltage of 2425 volts C. 8-5 voltage of 435 volts D. B-6 voltage of 435 volts Answer:

A. A-i voltage of 2425 volts Notes:

A Decay Heat Removal Pump is powered from A-3 via A-I. An undervoltage on the A buses or B buses will trip A-309 (A3 feeder breaker). The undervoltage setpoint for A-3 is 2450 volts. The undervoltage setpoint for B-5 is 429 volts. Therefore, b, c and d are incorrect.

References:

1107.002, Chg. 023-00-0 History:

Developed for 1999 exam.

Selected for 2005 Jon Gray RO re-exam.

Modified and USED in 2007 RO Exam.

PROC.IWORK PLAN NO. PROCEDUREIWORK PLAN TITLE:

PAGE: 8 of 80 -

1107.002 ES ELECTRICAL SYSTEM OPERATION CHANGE: 025 5.6 Diesel generator load limits;

• Limit diesel generator load to 2750KW continuous load.

• Additional loads, beyond those automatically sequenced on, may be started if needed provided that continuous diesel generator load is maintained <2750KW.

5.7 When racking out 4160V bus breakers, personnel shall use the protective equipment specified in Electrical System Operations (1107.001), Exhibit I, Electrical Safety Requirements.

5.8 Opening the DC Control Power Breaker in the following breaker cubicles results in loss of Bus-Protective Relays:

• A-309 Al Feed to A3

• A-409 A2 Feed to A4 5.9 Load Center Transformers are NOT capable of supplying full load of two buses when crosstied. Loading must be restricted.

5.10 When racked down and disengaged from the lifting mechanism, 4l60V breakers no longer meet seismic requirements.

5.11 Load Center Breaker Handling Jib Cranes do NOT meet seismic requirements when NOT secured in the stowed position.

5.12 Motor Control Centers and Load Centers require a seismic evaluation by Design Engineering to determine operability if the following conditions are exceeded:

• Two breakers removed

• One breaker removed and two breakers racked out

• Three breakers racked out 5.13 All 4l60V TEST breaker operations, including racking up or down, shall be performed by Electrical/Relay Department personnel.

5.14 Except in an emergency, all 4l60V breaker removal and reinstallation operations shall be performed by Electrical/Relay Department personnel.

6.0 SETPOINTS 6.1 Bus A3 and A4 undervoltage: 2450V nominal 6.2 Bus B5 and B6 undervoltage: 429.6V nominal

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0783 Rev: 0 Rev Date: 9/10/2009 Source: Modified Originator: Possage TUOI: Al LP-RO-ESAS Objective: 20 Point Value: 1 Section: 3.3 Type: Reactor Pressure Control System Number: 006 System

Title:

Emergency Core Cooling System

Description:

Knowledge of the effect of a loss or malfunction on the following will have on the ECCS: Valves KIA Number: K6.1 0 CFR

Reference:

41.7 / 45.7 Tier: 2 RO Imp: 2.6 RO Select: Yes Difficulty: 3 Group: 1 SRO Imp: 3.3 SRO Select: Yes Taxonomy: K Question: 33 33 RO:j SRO:

Given Degraded Power condition is present with a LOCA RCS pressure 1580 psig Reactor Building pressure 2 psig Diesel Generator #1 failed to start No other failures are present Which component would be automatically actuated to it ES position?

A. B Letdown cooler outlet CV-1216 would close B. RCP Motor Air and Lube Oil Cooling Isolation valve CV-2221 would close C. BWST outlet valve CV-1408 would open D. HPI Pump P-36B would start Answer:

C. BWST outlet valve CV-l 408 would open Notes:

C. Is the correct answer. CV-1408 is ES actuated open and will have power available to open..

A. Is incorrect, although it should close on ES, with the given information, it would not have power available to close.

B. Is incorrect, CV-2221 would have power to close but would not close until reactor Building pressure reached the setpoint for Channels 5 & 6 D. Is incorrect, B HPI pump would not start unless there was a failure of P-36C and the student was given that no other failures are present.

References:

STM 1-04 Rev. 9 STM 1-43 Rev. 12

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

History:

Modified from exma bank ANO-OpsUniti -06296 Selected for 2010 RO/SRO exam.

editQuestions Page 1 of Editing ANO:OpsUnitl :ANO-OpsUnitl-06296 You have 1 question in your editor cart.

l2pt Question:

Given Degraded Power condition is present with a LOCA Diesel Generator #1 failed to start 0 AAI 4,; (Li:

1 No other failures are present a Which component would be automatically actuated to it ES position/status if RCS pressure subsequently dropped below 1590 psig?

A. B Letdown cooler outlet CV-1216 would close B. Penetration Room Ventilation fan VEF-38A would start C. HPI Pump P-36B would start D. Letdown coolers outlet CV-1221 would close Answer: A. B. C C.

• D.

Select:

• Open Reference Question - Closed Reference Question Select: Handout Required with Exam

• Handout Not Required with Exam Point Value: 1.0 points Cognitive Level:

1: Fundamental Knowledge or Memory

• 2: Comprehension or Analysis 3: Synthesis or Evaluation Question Comments:

Review Comments: (Edited with Question Reviewer.)

http ://exambank.entergy.comleditQuestions.jsp 9 10 2009

Primary Makeup And Purification STM 1-04 Rev. 9 2.2.3 Leak Detection Tube to shell leaks inside the LD Coolers is detected through radiation elements monitoring the Nuclear ICW System. Because primary pressure is higher than ICW pressure when the coolers are in service, leakage will be from the primary to the ICW side of the coolers. Small leaks will be detected by process monitor RE- 2236.

Larger leaks may cause ICW Surge Tanks to overflow. Corrective actions for LD Cooler leaks are addressed in annunciator corrective actions and the Excess RCS Leakage Abnormal Operating Procedure.

For additional information on the ICW System see STM 1-43.

2.3 Letdown Heat These motor operated valves are located in the Reactor Building Exchanger Outlet Valves, and on the outlet side of the letdown coolers. They are operated from CV-1214 and CV-1216 Control Room Panel C18. During an Engineering Safeguards Actuation Signal (ESAS) Channel 1 Initiation, both valves close automatically. During ES actuation, these valves provide reactor building isolation. The valves are of the split disc gate valve type (system pressure leaking past one disc aids in seating the other disc).

On the bottom of each valve is a connection for local leak rate testing.

The testing connection allows pressure to be placed between the valve discs to assure positive seating.

2.4 N-I 6 Expansion The Expansion Tank (refer to figure 4.06) is an enlargement Tank from a 2 1/2 line to a 12 line and back to a 2 1/2 line. This expansion increases the transport time and allows for the decay of Nitrogen 16 (N 16), a high energy Gamma emitter. N 16 is produced in the reactor as an activation product and has a very short (7 second) half life. Increasing the LD transport time has the positive effect of reducing radiation levels outside containment. The tank is located in the letdown cooler room.

2.5 Letdown isolation CV-1221 is the first valve outside of the containment building valve CV-1 221 and is operated from Control Room Panel C16. CV-1221 is a motor operated gate valve interlocked with Temperature Switch TS-1221.

At an increasing letdown temperature of 13SF CV-1221 will close.

This closure is provided to protect the resins of the purification demineralizers (T-36A/B) from being exhausted prematurely due to high temperature.

CV-1221 is also closed by actuation of ES Channel 2 and provides Letdown isolation in the Upper North Piping Penetration Room (UNPPR). CV-1221 is the only single line isolation in the letdown flow path.

Instructions for reopening CV-1221 after closure due to high LD temperature trip are included in OP 1104.02, Makeup and Purification System Operation. Recovery from closure consists primarily of correcting the cause of the overheating, bypassing the LD demineralizers until flow can be reestablished and temperatures brought down to normal and then returning the proper DI to operation.

Intermediate Cooling Water STM 1-43 Rev. 12 indication on panel C09 (FI-2222) and provides standby pump auto start signal through FS-2222 discussed earlier in this section.

CR]) return temperature indication and high alarm functions are provided by TE-2222 located on the return line. CR]) return temperature can be read on panel C09 using TI-2222 or on the plant computer (T2222).

TS-2222 will cause annunciator alarm K08-B1 CR]) Cooling Return Temp Hi to alarm when return temperature reaches setpoint of> 160°F. This alarm also indicates inadequate SW cooling of the Non-Nuclear ICW cooler.

The CRD cooling water return line combines with the RCP Motor Air and LO Coolers return line to form a common 8-inch return line. CR]) and RCP ICW return line exits the RB at penetration #60 in the USPPR. This common return line like the supply line is provided with isolation valves for system isolation during an ESAS event. Both RB isolation valves are 8-inch, motor-operated gate valves. Inside isolation valve, CV-2221 receives a closed signal from ES channel 6 through HS-2221 located on panel C 16. CV-2221 is powered from vital bus B6 1, breaker B-61 92.

Outside isolation valve, CV-2220, receives a closed signal from ES channel 5 through HS-2220 located on panel C- 18. CV-2220 is powered from vital bus B52, breaker B-522 1.

Non-Nuclear ICW flow from the CR]) and RCPs tie into the common 10-inch return header to ICW cooler E-28A. Return flow from the Isophase Bus Cooling coils also ties into this return line.

2.6.2 MFP! RCP ICW (Refer to Figures 43.03 and 43.04)

Supply Line The first main supply line, which taps off the discharge line of P 33k is a 10-inch line, which provides cooling water flow to the following components:

Main Feed Water Pump L.O. Coolers.

Reactor Coolant Pump (RCP) Motor Air Coolers.

RCP Motor Bearing L.O. Coolers.

RCP High Pressure Lift Oil System Coolers.

RCP Backstop L.O. System Coolers. Note: RCP P-32B does not utilize a backstop lube oil system.

This supply line is provided with a means to isolate ICW flow to the MFP / RCP by closing isolation valve ICW-l 1. ICW-1 1 is a 10-inch butterfly valve located in the Main Chiller room. Downstream of ICW- 11 the line splits into two 10-inch lines which provide cooling water to the MFP/RCP and the other line is used to divert or bypass ICW flow to the MFPs/RCPs during shutdown conditions. ICW flow is diverted back to the Non-Nuclear ICW header through bypass valve ICW-23. This valve is normally throttled during plant operation to balance ICW flow.

The 10-inch supply line to the MFPs and RCPs splits into two separate lines. The first supply line provides cooling water flow to 14

Primary Makeup And Purification STM 1-04 Rev. 9 Design conditions on the shell side are: 150 psig, 300 °F, 110000 lbm/hr, 220 gpm 2.22 HPI Valves CV-1 227, These eight MOVs can be operated from Panels C16 and C18.

CV-1228, CV-1278, CV They are automatically opened on Engineering Safeguard Actuation.

1279, CV-1219, CV-1220, The HPI valves are actuated from the same channel as the associated CV-1 284, CV-1 285 HPI pump. ES channel 1 opens CV-1219, CV-1220,CV-1278, and CV-1279. ES channel 2 opens CV-1227, CV-1228, CV-1284, and CV-1285. High pressure injection enters the RCS on the discharge side of the reactor coolant pumps. Refer to figure 4.23.

Each injection line has flow instrument installed which has a readout on C 16 and Cl 8. The control room indicator has a low flow cutoff at 10 gpm to prevent indication when there is no flow. This indication is due to errors in loop flow.

There are high flow setpoints of 450 gpm total HPI flow per train and >140 gpm on each injection line. This warns the operator that insufficient flow may be going to the core due to high flow through an HPI line that has a break. A low flow alarm at 200 gpm is to warn operator of minimum flow requirements.

CLEAN wsro CV-1235 HIGH PRESSURE INJECTION TIS 1220 TO 069 TO 1068 A&B A&B P321)

MU661) MU341) MU4SI) 101062 101061 CZ-45 TIS 1219 A&13 A&B 132C MUI9A MU2OA CV1219 MU12II MU2 Mii1231 MU66C MU34C MU45C 0S2 2 MU25 FE 1232 CV1285 MUI9B MU2OII P01231 MU24 ES2 CV1284 MUI3OS MU23 ESI !l I -

P36C I P01212 CV1279 MU1307 MUI9C M 20C F I 29 P01211 ES I T1S1227 TEIO6I, T0I0(5 CV 1278MU 1306 A&13 A&13 ES2 P328 C1227 MU1214 ML1233 MU66B MU34B MU41l FE 1230 T01064 TE1063 TIS 228 F01228 A13 AB ES2 L

P32A CVI228 MU1215 Ml 1234 MU66A MU34A MU4 P01228 FIGURE 04.23: [WI LINE INSTRUMENTS & PIPING 33

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0561 Rev: I Rev Date: 8/10/05 Source: Direct Originator: S.Pullin TUOI: AILP-RO-RCS Objective: 21 Point Value: 1 Section: 3.5 Type: Containment Integrity System Number: 007 System

Title:

Pressurizer Relief Tank/Quench Tank System

==

Description:==

Knowledge of the operational implications of the following concepts as they apply to the PRTS:

Method of forming a steam bubble in the PZR.

KIA Number: K5.02 CFR

Reference:

41.5 I 45.7 Tier: 3 RO Imp: 3.1 RO Select: Yes Difficulty: 3 Group: 2 SRO Imp: 3.4 SRO Select: Yes Taxonomy: Ap Question: RO:j 34 SRO:J A plant startup is in progress with a steam bubble being drawn in the Pressurizer.

- Initial Quench Tank pressure is 3 psig.

- RCS pressure 75 psig.

- Pressurizer temperature 320°F.

Which of the following assures that venting and steam bubble formation is complete in the Pressurizer?

A. Quench Tank pressure 7.6 psig after a 3 minute blow of the ERV.

B. Quench Tank pressure 6.2 psig after a 3 minute blow of the ERV.

C. Quench Tank pressure 4.8 psig after a 3 minute blow of the ERV.

D. Quench Tank pressure 3.5 psig after a 3 minute blow of the ERV.

Answer:

D. Quench Tank pressure 3.5 psig after a 3 minute blow of the ERV.

Notes:

D is correct with Quench Tank pressure rise less than or equal to I psig.

All other choices contain greater than I psig pressure rise which indicates nitrogen is still being vented to the Quench Tank.

References:

1103.005, Chg. 036 History:

New for 2005 RO exam, later modified for replacement.

Selected for 2010 RO/SRO exam.

NOTE Venting and bubble formation is considered complete when both of the following conditions are met:

• A threeminute blow through the ERV results in Quench Tank pressure rise of 1 psig.

• A saturation pressure/temperature relationship exists in the PZR.

7.2.9 WHEN RC pressure rises to near 70 psig, THEN repeat steps 7.2.5 through 7.2.7 as necessary until bubble forms.

7.3 System Pressurization CAUTION The pressurizer spray block valve shall remain closed until the AT between the pressurizer and the RCS is 250°F to prevent exceeding design criteria of the spray and surge lines.

7.3.1 WHEN RCS is > 200°F, THEN open Spray Block Valve (CV1009) 7.3.2 Spray valve and heater banks may be cycled as necessary for heat-up and pressurization as outlined in Plant Startup (1102.002), Heatup and Pressurization to 350° & 500 PSIG section.

(4.3.7)

NOTE ERV Isolation (Cv-l000) is subject to binding if heatup continues with CV-l000 closed.

7.3.3 Verify ERV Isolation (CV-1000) remains open during heatup.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0787 Rev: 0 Rev Date: 9/14/2009 Source: Originator: S. Pullin TUOI: A1LP-RO-MSSS Objective: 9 Point Value: 1 Section: 3.8 Type: Plant Service Systems System Number: 008 System

Title:

Component Cooling Water System

Description:

Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of CCW Pump KIA Number: A2.01 CFR

Reference:

CFR: 41.5 / 43.5 / 45.3 / 45.13 Tier: 2 RO Imp: 3.3 RO Select: Yes Difficulty: 3 Group: 1 SRO Imp: 3.6 SRO Select: Yes Taxonomy: C Question: 35 RO:j SRO:1 Given:

- 80% power,

- P33A and P33B ICW pumps in service.

- P33C (ICW Pump) out of service

- P33B (ICW Pump) trips What impact would this have on plant operations, and what actions are required per 1104.028, ICW System Operating Procedure?

A. Loss of Non-Nuc ICW, open all ICW cross connect valves CV-2238, CV-2239, CV-2240 and CV-2241 B. Loss of Non-Nuc ICW, close A to B cross connect valves CV-2238 and CV-2240 C. Loss of Nuc ICW, open all ICW cross connect valves CV-2238, CV-2239, CV-2240 and CV-2241 D. Loss of Nuc ICW, close A to B cross connect valves CV-2238 and CV-2240 Answer:

C. Loss of Nuc 1GW, open all ICW cross connect valves CV-2238, CV-2239, CV-2240 and CV-2241 Notes:

C is correct P33C supplies the Nuc 1GW loads, OP-I 104.028 has the operator open the suction and discharge cross connect valves to supply both loops with one pump prior to reducing loads.

A is incorrect due to Non Nuc 1GW loads were never lost B is incorrect due to Non Nuc ICW loads were never lost D is incorrect due to procedure has you open the valves and not close them

References:

OP-1104.028 Change 026 History:

New question, selected for 201 ORO/SRO exam.

i PROCJWORK PLAN NO.

I PROCEDUREIWORK PLAN TITLE: I PAGE: 2 of 111 I 1.0 1104.028 PURPOSE ICW SYSTEM OPERATING PROCEDURE CHANGE: 026 To provide procedure for operation of the intermediate cooling water system.

2.0 SCOPE This procedure is provided for the startup, normal operation, emergency operation, and shutdown of the ICW and CR]) cooling water systems.

This procedure contains Temp Mod controls in Attachment B, Temporary Installation of a Service Water Outlet at ICW Cooler E-28C.

3.0 DESCRIPTION

The ICW system is composed of two independent closed loop cooling systems which provide an intermediate cooling water barrier between the cooled components and the Service Water system. The purpose for closed loop systems is to prevent direct contact between a radioactive system and the Service Water system.

The system uses three parallel recirculation pumps (P-33A, B, C) and three parallel heat exchangers (E-28A, B, C). The pumps circulate the ICW to various components and back through the heat exchangers which are cooled by Service Water running through the tubes. P-33A and E-28A provide cooling for the Non-Nuclear loop components and P-33C and E-28C provide cooling for the nuclear loop components. P-33B and E-28B are swing components which can be used by either loop. In normal alignment, Non-Nuc ICW is cooled by Loop II Service Water and Nuc ICW is cooled by Loop I Service Water. Both ICW loops are continuously monitored by radiation detectors to warn operators of radioactivity in the ICW system.

The Non-Nuclear loop normally has a higher activity level due to activation of ICW chemicals while over the Rx vessel head in the CR0 cooling loop. There is an ICW Surge Tank (T-37A & B) associated with each loop that provides NPSH for pumps and a surge volume for the loops. This is also where makeup is added to the system from the condensate transfer system.

The CR0 cooling system has two parallel recirculation pumps, CRD Pumps (P-79A and P-79B), which take a suction on the Non-Nuclear ICW loop downstream of E-28A and provide cooling water to CRD motors. It returns to the Non-Nuclear loop on the inlet to E-28A.

The RCP Seal Cooling Pumps (P-ll4A & B) provide added system pressure and flow for RCP seal cooling. The pumps take a suction on the Nuclear Loop inside the Reactor Building and return to the Nuclear Loop inside the Reactor Building.

3.1 Nuclear Loop cools

• Spent Fuel Coolers (E-27A & B)

• RCP Seal Return Coolers (E-26A & B)

• Waste Gas Compressors (C-9A & B)

• Waste Gas Compressor Aftercoolers (E-40A & B)

• Vacuum Degasifier Seal Water Cooler (E-53)

• Pressurizer Sample Cooler (E-30)

• Steam Generator Sample Cooler (E-3lA)

• Letdown Coolers (E-29A & B)

• RCP Seal Water Coolers (E-25A, B, C, D)

i PROCJWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: I PAGE: 58 of 111 I 20.0 1104.028 ICW SYSTEM OPERATING PROCEDURE Contingency Actions for Loss of Two ICW Pumps CHANGE: 026 CAUTION Operation of one ICW pump with the cross-connect valves open will result in pump operation at runout conditions. Pump cavitation can occur and there is elevated risk for motor breaker trip until ICW loads are reduced.

20.1 Place tripped ICW pump(s) in PULL-TO-LOCK.

20.2 Open the following valves:

• ICW Pump Suction Crossconnect CV-2240

• ICW Pump Suction Crossconnect CV-224l ICW Pump Discharge Crossconnect CV-2238

• ICW Pump Discharge Crossconnect CV-2239 20.3 Close the following valves to isolate letdown:

• Letdown Orifice Block Bypass (CV-1223)

• Letdown Orifice Block (CV-l222) 20.4 Isolate both Letdown Coolers (E-29A and E-29B) by closing the following valve pairs from C04:

• E-29A HS-2216 for Letdown Cooler Inlet Valve (CV-2216) and RC to Letdown Cooler E-29A (CV-12l3)

• E-29B HS-2217 for Letdown Cooler Inlet Valve (CV-2217) and

• RC to Letdown Cooler E-29B (CV-12l5) 20.5 Isolate both SFP Coolers (E-27A, E-27B) by closing the following:

• SFP Clr E-27A ICW Outlet (ICW-121A)

• SFP Clr E-27B ICW Outlet (ICW-l2lB) 20.6 Return one Letdown Cooler to service by opening one of the following valve pairs from C04:

• E-29A HS-2216 for Letdown Cooler Inlet Valve (CV-2216) and RC to Letdown Cooler E-29A (CV-1213)

• E-29B HS-2217 for Letdown Cooler Inlet Valve (CV-2217) and RC to Letdown Cooler E-29B (CV-12l5) 20.7 Verify combined ICW flow is 3100 gpm.

20.8 Establish letdown by opening Letdown Orifice Block (CVl222) 20.9 IF letdown isolated on high temperature, THEN perform Recovery of Letdown Following High Letdown Temperature section of Makeup and Purification System (1104.002).

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0788 Rev: 0 Rev Date: 9/14/2009 Source: New Originator: S. Pullin TUOI: AILP-RO-RPS Objective: 5 Point Value: 1 Section: 3.3 Type: Reactor Pressure Control System Number: 010 System

Title:

Pressurizer Pressure Control System (PZR PCS)

Description:

Knowledge of the effect that a loss or malfunction of the PZR PCS will have on the following:

RPS KIA Number: k3.02 CFR

Reference:

41.7 /45.6 Tier: 2 RO Imp: 4.0 RO Select: Yes Difficulty: 2 Group: I SRO Imp: 4.1 SRO Select: Yes Taxonomy: K Question: 36 36 RO:j SRO:I Given:

- 100% power,

- A MFW Pump trips

- PZR Spray valve (CV-1 008) will not open.

What effect would this pressurizer control system malfunction have on the plant?

A Reactor trip due to AMSAC B. Reactor trip due to anticipatory trip from RPS on loss of MFW pumps C. Reactor trip due to High Power/Imbalance/Flow D. Reactor trip due to High RCS Pressure Answer:

D. Reactor trip due to High RCS Pressure Notes:

A is incorrect because total feedwater flow will remain above trip setpoint B is incorrect because only one MFW pump is tripped C is incorrect because the flow in this coice refers to RCS flow D is correct, without the spray valve opening RCS pressure will rise to the trip setpoint

References:

QP-1202.001 Change 31 History:

New selected for 2010 RO/SRO exam.

CHANGE 1202.001 REACTOR TRIP 031 PAGE 1 of 25 ENTRY CONDITIONS

• An automatic Rx trip or DSS trip.

• Failure of RPS to trip the Rx upon reaching a limit listed below:

- High power 104.9%

- High power/pumps one pump per loop .. 55%

OR 0 pumps in one loop.. 0%

- High power/imbalance/flow COLR Figure

- High RCS temp 618 °F (T-hot)

- High RCS press 2355 psig

- Low RCS press 1800 psig

- Variable low RCS press COLR Figure

- High RB press 18.7 psia

- Turbine trip Rx power 43% Turbine is tripped

- Both MFW pumps trip Rx power 9% AND both MFW pumps tripped.

• PZR level dropping < 100, AND no indication of recovery.

• PZR level > 290.

• Any MSIV closure at power.

• Either SG level < 15 or> 95%,

AND no indication of recovery.

• A system degradation that requires manual Rx trip based on operator judgment.

• Abnormal Operating Procedure requirement.

• IF a system degradation occurs while shutdown, above D HR operation, THEN perform applicable steps.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0784 Rev: 0 Rev Date: 9/10/2009 Source: New Originator: Possage TUOI: A1LP-RO-RPS Objective: 11 PointValue: I Section: 3.7 Type: Instrumentation System Number: 012 System

Title:

Reactor Protection System

==

Description:==

Knowledge of the effect of a loss or malfunction of the following will have on the RPS:

Permissive circuits K/A Number: K6.1 0 CFR

Reference:

41.7 / 45.7 Tier: 2 RO Imp: 3.3 RO Select: Yes Difficulty: 3 Group: 1 SRO Imp: 3.5 SRO Select: Yes Taxonomy: K Question: RO:J 37 SRO:j 37 Given:

The plant is at 100% power l&C is troubleshooting RPS 6 RPS is in Manual Bypass The Shutdown Bypass 5% bistable in Channel A has been pulled from the cabinet.

What would be the effect of a failure in the B RPS permissive circuitry that caused a short which de-energizes the B RPS Cabinet?

A. RPS would be in a 2 out of 3 coincidence trip logic B. RPS would be in a 2 out of 2 coincidence trip logic C. Reactor Trip would occur D. High Flux trip bistable tripped in Channel A Answer:

C. Reactor Trip would occur Notes:

C. Is correct. The conditions given would result in the A Channel being tripped, when B is de-energized it would also be tripped and make up the logic to trip the reactor.

A and B are incorrect because the logic to trip the reactor has already been met.

D is incorrect, pulling the Shutdown Bypass 5% bistabie would not cause a high flux trip bistable to trip in RPS.

References:

STM 1-63 Rev. 7 History:

Modified from Exam Bank ANO-OPS1-1670 Selected for 2010 RO/SRO exam

Reactor Protection System STM 1-63 Rev 7 Figures And Diagrams!Tables Etc.

FIGURE 63.01: CHANNEL TRIP CONTACT STRING

.11 SOC HOST PRESSURE UUIOSLUCTCT FROST ISO DRYABLE O1TIERCTLGAOULO IOPRESOWFM000 11010 ONE CHANNFL SF0105017555 OTU1000NOIPASS AT ONE lOAF RIOT! FLUX 150 POTABLE CBINTUOLGCE III OThER EIISSSNFLS L

CHANNEL 017555 EFXSWOC1I CTTANNFL

+/-

EI7AOS RELAY TOOII REMOYAL 10011 FlUX TRIP BITTARLE I I iç HIGH PRESSURE 157 POThOLE  ::

MOOULF-IN-1XST FCIFSISWET 111011 TXT.IPRRATUEE TRIP HOTSOLE -C RERRREOC PRUSSJRE1IOP CONYACF BUFFER PRESSURE SWELl!

TERHOJE 1117.055 RUmBLE SIFIICH PINE CONTROL PRESSURE SWOflIFO m

OWOSEL EFSEF NOTE:

CONTACT FOSFETONS ARE SHOWN FOR OPERATION CONOrrIONS.

42

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0785 Rev: 0 Rev Date: 9/10/2009 Source: New Originator: Possage TUOI: A1LP-RO-RPS Objective: 19 PointValue: I Section: 2.0 Type: Generic K/A System Number: 012 System

Title:

Reactor Protection System

==

Description:==

Ability to explain and apply system limits and precautions.

K/A Number: 2.1.32 CFR

Reference:

41.10/43.2/45.12 Tier: 2 RO Imp: 3.8 RO Select: Yes Difficulty: 3 Group: 1 SRO Imp: 4.0 SRO Select: Yes Taxonomy: K Question: RO: 38 SRO:i Given:

The plant is at 100% power B RPS is INOPERABLE due to a failed High Temperature Trip Bistable All other RPS channels OPERABLE Which of the following is NOTa required action perT.S. 3.3.1?

A. Place channel in bypass within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> B. Place channel in a trip condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> C. Prevent bypass of remaining channels within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> D. Open all CRD trip breakers within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Answer:

D. Open all CRD trip breakers within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Notes:

D is correct, with only one RPS channel inoperable T.S. does not require CRD trip breakers to be opened A, B and C are all incorrect. T.S. 3.3.1 reuires any one of listed conditions be performed for the condition given.

References:

OP-1105.001 Change 024 TS 3.3.1 History:

New Selected for 2010 RO/SRO exam

4.2.11 Reactor Protection System Channel D Test (1304.040).

4.2.12 Reactor Protection System Channel 1) Calibration (1304.044).

4.2.13 CR1) System Operating Procedure (1105.009) 4.2.14 Emergency Operating Procedures (1202.XXX).

4.2.15 Source Range Channels Test (1304.055) 4.3 NRC COMMITMENTS None.

5.0 LIMITS AND PRECAUTIONS 5.1 Do not place an RPS protection channel in manual bypass without first obtaining permission from the Shift Manager/CRS and notifying Control Room personnel.

When testing an RPS protection channel, only the EFIC channel associated with the RPS channel being tested may be in MAINTENANCE BYPASS. TS 3.3.1 provides guidance when an RPS channel is bypassed or contains inoperable functions.

5.3 Placing two RPS protection channels in test simultaneously will result in a reactor trip unless one is in channel bypass.

5.4 Only one RPS channel shall be key locked in the untripped state at any one time.

5.5 Only one RPS channel bypass key shall be accessible for use in the control room.

5.6 The key-operated shutdown bypass switch associated with each RPS channel shall not be used during power operation except for testing.

5.7 In the event that one of the trip devices in either of the sources supplying power to the CRDMs fails in the untripped state, perform required actions for applicable TS 3.3.4 conditions.

5.8 Do not apply power to the CRDM5 without using applicable section(s) of CRD System Operating Procedure (1105.009).

RPS Instrumentation 3.3.1 3.3 INSTRUMENTATION 3.3.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1 Four channels of RPS instrumentation for each Function in Table 3.3.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.1-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One channel inoperable. A.1 Place channel in bypass or 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> trip.

OR A.2 Prevent bypass of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> remaining channels.

B. Two channels B.1 Place one channel in trip. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> inoperable.

AND B.2.1 Place second channel in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> bypass.

OR B.2.2 Prevent bypass of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> remaining channels.

C. Three or more channels C.1 Enter the Condition Immediately inoperable, referenced in Table 3.3.1-1 for the Q Function.

Required Action and associated Completion Time of Condition A or B not met.

ANO-1 3.3.1-1 Amendment No. 215

RPS Instrumentation 3.3.1 CONDITION REQUIRED ACTION COMPLETION TIME D. As required by Required D.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action C.1 and referenced in Table 3.3.1-1.

D.2 Open all control rod drive 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (CRD) trip breakers.

E. As required by Required E.1 Open all CRD trip 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action C.1 and breakers.

referenced in Table 3.3.1-1.

F. As required by Required F.1 Reduce THERMAL POWER 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action C.1 and < 45% RTP.

referenced in Table 3.3.1-1.

G. As required by Required G.1 Reduce THERMAL POWER 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action C.1 and < 10% RTP.

referenced in Table 3.3.1-1.

SURVEILLANCE REQUIREMENTS

---

----------------NOTE-------------------

Refer to Table 3.3.1-1 to determine which SRs apply to each RPS Function.

SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ANO-1 3.3.1-2 Amendment No. 215

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0142 Rev: 0 Rev Date: 10/28/97 Source: Direct Originator: G. Giles TUOI: AA51002-012 Objective: 21 Point Value: 1 Section: 3.2 Type: RCS Inventory Control System Number: 013 System

Title:

Engineered Safety Features Actuation System(ESFAS)

==

Description:==

Knowledge of ESFAS design feature(s) and/or interlock(s) which provide for the following:

Safeguards equipment control reset.

K/A Number: K4.10 CFR

Reference:

41.7 Tier: 2 RO Imp: 3.3 RO Select: Yes Difficulty: 2 Group: 1 SRO Imp: 3.7 SRO Select: Yes Taxonomy: K Question: RO:1 39 SRO:I 39 Under what conditions can the Control Board Operator bypass or defeat a component automatically actuated by ESAS?

A. Bypassing or defeating a component automatically actuated by ESAS is not allowed.

B. The Control Board Operator, after careful consideration, determines that the component is no longer required.

C. ONLY when procedurally directed by the Emergency Operating or the Abnormal Operating procedures.

D. After it is determined that the component is no longer needed and approval is obtained from the SM/CRS.

Answer:

D. After it is determined that the component is no longer needed and approval is obtained from the SM/CRS.

Notes:

[A] is incorrect, provisions are made for this action.

[B] is partially correct, the component must not be needed but the CBO cannot make this decision on his own.

[C] is only one of the directions where a component can be bypassed/reset, CRS/SS permission is the other.

[D] contains all correct elements, lack of need and supervisory (SRO) permission.

References:

OP-1202..012 Change 008 History:

Taken from Exam Bank QID # 4791 Used in A. Morris 98 RO Re-exam Previously used under K/A: 3.2 / Reactor Coolant System Inventory Control / 013 / Engineered Safety Features Actuation System / A4.02 / Ability to manually operate and/or monitor in the control room: Reset of ESFAS channels. / CFR: 41.7 / 45.5 to 45.8 / RO: 4.3 / SRO: 4.4 Used on 2004 RO/SRO Exam (K/A T2 Gi 013 K4.06)

Selected for the 2010 RO/SRO exam

CHANGE 1202.012 REPETITIVE TASKS 008 PAGE 22 of 50 Page 1 of 3 NOTE Obtain Shift Manager/CRS permission prior to overriding ES.

10. Verify proper ESAS actuation:

A. Verify BWST Outlets open (CV-1 407 and 1408).

1) IF CV-1407 or 1408 fails to open, THEN override AND stop associated HPI, LPI, and RB Spray pumps until failed valve is opened.

B. Verify SERV WTR to DGI and DG2 CLRs open (CV-3806 and 3807).

C. if y RCP is running, THEN perform the following:

1) IF ES Channel 5 or 6 has actuated, THEN perform the following:

a) IF SCM is adequate, THEN trip all running RCPs due to loss of ICW.

b) iF SCM is <adequate, THEN check elapsed time since loss of adequate SCM fjQ. perform the following:

(1) IF 2 minutes have elapsed, THEN trip all RCPs.

(2) IF >2 minutes have elapsed, THEN perform the following:

(a) Leave currently running RCPs on.

(b) jf RCS press> 150 psig, THEN notify CRS to GO TO 1202.002, LOSS OF SUBCOOLING MARGIN procedure AND perform contingency for failure to trip RCPs within 2 minutes.

(c) Restore RCP services (RT 8) while continuing.

2) IF neither ES channel 5 or 6 has actuated, THEN dispatch an operator to perform Service Water And Auxiliary Cooling System (1104.029) Exhibit B, Restoring SW to ICW Following ES Actuation, while continuing.

a) WHEN ICW Cooler SW Outlets and Bypasses are aligned per 1104.029, Exhibit B, THEN override AND open one Service Water to ICW Coolers Supply (CV-3811 or 3820).

(10. CONTINUED ON NEXT PAGE) f 1202.012 I RTIO Rev 3-16-06

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0135 Rev: I Rev Date: 4/7/05 Source: Direct Originator: B. Short TUOI: Al LP-RO-ESAS Objective: 20 Point Value: I Section: 3.5 Type: Containment Integrity System Number: 022 System

Title:

Containment Cooling System

==

Description:==

Ability to monitor automatic operation of the CCS, including: Initiation of safeguards mode of operation.

K/A Number: A3.0l CFR

Reference:

41.7 / 45.5 Tier: 2 RO Imp: 4.1 RO Select: Yes Difficulty: 2 Group: I SRO Imp: 4.3 SRO Select: Yes Taxonomy: K Question: RO:1 40 SRO:J A LOCA has occurred.

Reactor Building (RB) pressure is 47 psia.

Which ESAS channels have actuated the RB cooling units and what is the correct RB cooling alignment?

A. ES channels 3 & 4, VSF-1A, IB, 1C, & 1D running with service water aligned to the cooling coils.

B. ES channels 3 & 4, VSF-IA, IB, IC, ID, & IE running with chilled water aligned to the cooling coils.

C. ES channels 5 & 6, VSF-1A, IB, 1C, & ID running with service water aligned to the cooling coils.

D. ES channels 5 & 6, VSF-1A, IB, 1C, ID, & IE running with chilled water aligned to the cooling coils.

Answer:

c. ES channels 5 & 6, VSF-1A, IB, 1C, & ID running with service water aligned to the cooling coils.

Notes:

ESAS channels 5 & 6 actuate RB cooling fans VSF-IA through ID and also cause the bypass dampers to drop which allows air to bypass the retum air duct and chilled water coils and flow directly to the service water coils that were aligned by ES channels 5 & 6. Thus (c) is the correct answer. (a), (b) & (d) combine other ventilation alignments with other ES channels that are incorrect.

References:

STM 1-09, Rev. 9 History:

Developed for use in 98 RO Re-exam Selected for 2005 RO exam.

Selected for 2010 RO/SRO exam

Reactor Building Ventilation STM 1-09 Rev. 9 Cooling Units VSF-1A through 1D each have an associated ES signal from either Channel S or 6. During normal operation, the four units are nmning with chilled water as the cooling medium. On an ES actuation signal, all four units receive a start signal and a bypass damper opens allowing air to bypass the return air duct and chilled water coils allowing flow directly to the service water coils. Service water valves to the coils are opened by ESAS Ch 5 or 6 and chilled water to the RB is automatically secured. The lower pressure drop caused by bypassing the chilled water coils and return plenum, permits the single speed fan to handle the quantity of air necessary for emergency cooling. This precludes the necessity of a two-speed motor with the additional controls, power source and wiring.

Unit Control CS Power Supply ES Actuating Switch Location Signal:

VSF-1A HS-7410 C18 480v ES Bus ES-S B523 VSF-1B HS-7411 C18 480v ES Bus ES-5 B533 VSF-1C HS-7412 C16 48OvESBus ES-6 B623 VSF-1D HS-7413 C16 480v ES Bus ES-6 B633 VSF-1E HS-7419 C19 480vB714 None 2.1.1.2 Supply Fan Back- Each supply fan (VSF-1A thru D) has a single blade, butterfly draft Dampers damper (CV-7470 thru 7473) at the discharge of the fan that opens CV7470 7473-when the fan starts. These dampers are called back-draft dampers because they prevent reverse flow through the fan when it is not running. Each damper has a Limitorque motor operator that is controlled from the same hand switch as the supply fan. They are powered from MCC B5252 for CV-7470, B5332 for CV-7471, B6212 for CV-7472 and B6332 for CV-7473. Damper position indication is provided on Control Room panels C-i 6 or C-i 8.

Refer to figure 9.01, 9.02 & 9.03 2.1.1.3 VCC-IA IE-Chilled Water The Chilled Water Cooling Coils for the RB Cooling Units are single stage coils supplied from Main Chill Water. Isolation Valves Cooling Coils for Main Chill Water (CV-6202 & CV-6203) are air operated outside the RB with a motor operated valve (CV-6205) for the return line inside the RB. Check valve AC-60 is used for double isolation in the supply line inside the RB. The Containment Isolation valves for Chill Water are closed by ES Channel 5 & 6 signals.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0078 Rev: 0 Rev Date: 6/29/98 Source: Direct Originator: JCork TUOI: Al LP-RO-ELECD Objective: 11 .e Point Value: I Section: 3.5 Type: Containment Integrity System Number: 026 System

Title:

Containment Spray System (CSS)

==

Description:==

Knowledge of the physical connections and/or cause-effect relationships between the CSS and the following systems: ECCS.

K/A Number: Ki .01 CFR

Reference:

41.2 to 41.9 / 45.7 to 45.8 Tier: 2 RO Imp: 4.2 RO Select: Yes Difficulty: 2 Group: 1 SRO Imp: 4.2 SRO Select: Yes Taxonomy: K Question: RO:1 41 SRO:I 41 If an ESAS occurs simultaneously with a Loss of Offsite Power, the start of RB Spray pumps is delayed by 35 sec. Why?

A. To allow the EDGs to come up to speed.

B. To allow SW pumps to start for spray pump cooling.

C. To prevent overload of the EDGs.

D. To prevent water hammer of the spray headers.

Answer:

C. To prevent overload of the EDGs.

Notes:

With an ES signal present, ES loads will sequence on to the EDG to prevent overload, therefore C is correct. (a), (b) and (d) are reasons for other aspects of RB spray operation but are not applicable to the basis for the time delay.

References:

1107.002, Chg. 025 History:

Developed for 1998 RO/SRO Exam.

Used in A. Moms 98 RO Re-exam Selected for 2005 Jon Gray RO re-exam.

Selected for the 2010 RO/SRO exam.

PROC.IWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: I PAGE: 3 of 80 J 1.0 1107.002 PURPOSE ES ELECTRICAL SYSTEM OPERATION CHANGE: 025 To provide instructions for operating the engineered safeguard 4160V and 480V AC electrical distribution system.

2.0 SCOPE This procedure is used for normal and infrequent operation of the 4lEOV and 480V ES distribution system including normal, emergency, and alternate AC power sources where those instructions differ significantly than those more generic instructions in Electrical System Operations (1107.001).

This procedure establishes operating guidelines and requirements to meet NRC Generic Letter 91-11, LCOs for Vital Instrument Buses and Tie Breakers.

3.0 DESCRIPTION

Two 4160V AC engineered safeguard buses provide power to the engineered safeguard equipment, including the 480V AC ES distribution system, through 4l60V/480V transformers.

The normal power source to bus A3 and A4 is from non-ES 4l60V buses Al and A2 respectively. The emergency power source is from 4160V AC, 2750KW diesel generators, one for each bus. Emergency power is supplied automatically on loss of normal power. The Alternate AC source is a 4400KW diesel generator manually placed into service.

Normally the buses are separated and independent; however, bus tie breakers are provided for abnormal situations. Some ES loads can be powered from either bus. To maintain bus separation and independence as required by 10CFR5O Appendix R, motor operated disconnects (MODs) are provided for the B HPI pump and B service water pump.

To prevent overload due to simultaneous starting currents, ES loads are automatically sequenced onto the ES buses. This automatic sequencing occurs whether the bus is on the normal source or the emergency source.

The 480V AC engineered safeguard distribution system consists of two 480V AC load centers, B5 and B6, each containing a 1000KVA 4l60V/480V step-down transformer. B5 and BE are powered from buses A3 and A4, respectively, through the step-down transformer.

Bus B5 supplies motor control centers MCC B51, B52, B53 and B57 (MCC B53 is supplied from MCC B52).

Bus B6 supplies motor control centers MCC BEl, B62, B63, B64 and B65 (MCC B63 is supplied from MCC B62. MCC B64 is supplied from MCC B65).

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0202 Rev: 0 Rev Date: 11/23/98 Source: Direct Originator: R. Walters TUOI: AILP-RO-EOP Objective: 9 Point Value: I Section: 3.4 Type: RCS Heat Removal System Number: 039 System

Title:

Main and Reheat Steam System

==

Description:==

Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (b) based on predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Malfunctioning steam dump.

KIA Number: A2.04 CFR

Reference:

41.5 / 43.5 / 45.3 / 45.13 Tier: 2 RO Imp: 3.4 RO Select: Yes Difficulty: 3 Group: 2 SRO Imp: 3.7 SRO Select: Yes Taxonomy: A Question: RO:j 42 SRO:I 42 Given:

- A plant startup is in progress with the reactor critical below the point of adding heat.

- B OTSG Turbine Bypass Valve (CV-6688) fails full OPEN and is unable to be closed with the handjack.

- Tave 524 degrees and dropping

- Pressurizer level 205 inches and dropping

- RCS pressure 2120 psig and dropping What is the proper course of action?

A. Initiate MSLI for the B OTSG and maintain the reactor critical using A OTSG Turbine Bypass Valve to control RCS temperature and pressure.

B. Continue the reactor startup maintaining startup rate <1 DPM while continuing to monitor primary and secondary plant parameters.

C. Go directly to 1203.003, OVERCOOLING for actions to mitigate the oversteaming of the B OTSG.

D. Trip the reactor and follow the guidance of 1202.001 REACTOR TRIP.

Answer:

D. Trip the reactor and follow the guidance of 1202.001 REACTOR TRIP.

Notes:

(A.) is incorrect. You would not want to isolate a OTSG and maintain the reactor critical.

(B.) is incorrect. With the reactor below the point of adding heat with a stuck open TBV, this would not be possible.

(C.) is incorrect. This will be the ultimate tab that you will end up in, however, it is necessary to trip the reactor first and progress through the Reactor Trip EOP.

(D.) is correct. Taking the conservative action of tripping the reactor is appropriate due to being below the minimum temperature for criticality and the inability to maintain SUR below I DPM.

References:

1102.008 (Rev 023), Approach to Criticality, pages 4&5 History:

Developed for use in 98 RO Re-exam.

Used in 2001 RO/SRO Exam.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

Used on 2004 RO/SRO Exam.

Selected for 2010 RO/SRO exam QID: 0203 Rev: 0 Rev Date: 11/23/98 Source: Direct Originator: B. Short TUOI: AA51002-008 Objective: 8.8 Point Value: 1 Section: 3.4 Type: RCS Heat Removal System Number: 039 System

Title:

Main and Reheat Steam System

==

Description:==

Ability to manually operate and/or monitor in the control room: Emergency feedwater pump turbines.

KIA Number: A4.04 CFR

Reference:

41.7 / 45.5 to 45.8 Tier: 2 RO Imp: 3.8 RO Select: No Difficulty: 4 Group: 2 SRO Imp: 3.9 SRO Select: No Taxonomy: An Question: RO:I SRO:j Red powered EFW Pump Turbine (K-3) Steam Admission Valve Bypass Valve (SV-2663) has failed to open during regularly scheduled surveillance testing. What are the required operator actions?

a. Deenergize SV-2663 closed and declare P-7A inoperable.

b. Declare SV-2663 inoperable and manually open the valve.

c. Declare SV-2663 inoperable and deenergize CV-2663 closed.

d. Deenergize SV-2617 open and declare P-7B inoperable.

Answer:

c. Declare SV-2663 inoperable and deenergize CV-2663 closed.

Notes:

(a.) is incorrect. The red powered valve being inoperable does not affect the operability of the green train P 7A. .

(b.) is incorrect. SV-2663 is a solenoid operated valve and can not be manually operated.

(C.) is correct. With CV-2663 deenergized closed, the green powered valve CV-261 7 is still available to operate P-7A. .

(d.) is incorrect. P-7B is the electric driven pump and is not affected by the steam supply valve operability.

References:

1106.006 (Rev 58)

History:

Developed for use in 98 RO Re-exam

4.2 REFERENCES

USED IN CONJUNCTION WITH THIS PROCEDURE 4.2.1 Soluble Poison Concentration Control (1103.004) 4.2.2 Reactivity Balance Calculation (1103.015) 4.2.3 CRD Operating Procedure (1105.009) 4.2.4 Plant Preheatup and Precritical Checklist (1102.001 4.2.5 Power Operation (1102.004) 4.2.6 NI & RPS Operating Procedure (1105.001) 4.2.7 Unit 1 Technical Specifications 4.2.8 Loss of Neutron Flux Indication (1203.021) 4.2.9 Plant Startup (1102.002) 4.2.10 Infrequently Performed Tests or Evolutions EN-OP116.

4.2.11 Reload Criticality and Low Power Physics Test (1302.020) 5.0 LIMITS AND PRECAUTIONS 5.1 Operators performing/supervising the reactor startup should not rely on the critical rod position predicted by the estimated critical position calculation, but anticipate criticality any time during rod withdrawal, boron dilution or RCS temperature changes.

5.2 Maintain at least a 1.5% shutdown margin if any condition, physical or administrative, delays approach to criticality.

5.3 Do not simultaneously change reactivity by more than one means while subcritical or prior to point of adding heat.

5.4 Operators performing/supervising the reactor startup should use all pertinent instrumentation available to monitor indication of approaching criticality. The tendency to become fixed on one indication should be avoided.

5.5 Maximum continuous SUR is 1 DPM. Prompt change associated with attaining this SUR shall be <1.5 DPM.

5.6 Reactor coolant temperature shall be above 525°F when the reactor is critical (TS 3.4.2).

5.7 During approach to criticality, safety rod groups shall be at upper limit and regulating rods shall be positioned as prescribed per Regulating Rod Insertion Limits curves of the COLR and (TS 3.2.1).

5.8 During startup when intermediate range instruments come on scale, flux level shall be maintained in the source range until overlap between intermediate range and source range instruments is greater than or equal to one decade (SR 3.3.10.1 Bases).

5.9 Reactor shall not be made critical until at least 2 of the 3 (TS 3.4.9 and emergencypowered pressurizer heater groups are operable TRM 3.4.9).

rizer Code Safety 5.10 Reactor shall not be made critical until both Pressu Valves (PSV-lOOl and PSV-l002) are operab le (TS 3.4.10) startup 5.11 The licensed Operators performing/supervising the reactor shall perform no other duties during reactor startup .

g the reactor 5.12 The licensed Control Room Operators performing/supervisin not conduc t shift relief until the reactor is critical startup shall shutdo wn by 1.5% Ak/Ic except during physic s testing.

at l% power or ies in 5.13 Prior to commencing the reactor startup, a review of activit d shall be conduc ted to ensure that distrac tions to progress or planne the startup will be minimized.

be 5.14 During the reactor startup, access to the control room shall established, limited to ensure that a professional environment, once is maintained without distraction or interruption.

the control 5.15 The Shift Manager shall oversee the reactor startup from ional enviro nment is mainta ined.

room and ensure that a profess startup, 5.16 If unexpected situations/conditions arise during the reactor ing/sup ervisin g the reactor startup shall then the Operators perform vative action to place the reactor in a safe condit ion.

take conser between 5.17 During startup when withdrawing regulating groups, the overlap tial groups shall be betwee n 15% and 25% except for physics two sequen testing. (TS 3.2.1)

Control Room to 5.18 Reactor Engineering personnel shall be present in the perform l/M plots.

monitor the approach to criticality and to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior 5.19 SR 3.2.1.3 requires verification of SDM 1% AK/K within to achieving critical ity.

t Reactor 5.20 If the reactor has been shutdown <48 hours, then contac verify that the Fuels and Analys is calcula ted RHOBAL Engineering to Critica l Calcul ations.

bias has been incorporated into the Estimated (CRHQN200900107) (CRANO 1200902 37) 0.5% Ak/k but 5.21 If criticality achieved within procedural limits of +

to initiate a NOT within +/- 0.25% Ak/k, then notify Reactor Engineering condition report. (CRANO120090237) 6.0 SETPOINTS ures.

6.1 Observe setpoints in referenced system operating proced

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0195 Rev: 0 Rev Date: 11/24/98 Source: Direct Originator: L. Kilby TUOI: A1LP-RO-FW Objective: 18 Point Value: 1 Section: 3.4 Type: RCS Heat Removal System Number: 059 System

Title:

Main Feedwater System flow

==

Description:==

Ability to monitor automatic operation of the MFW, including: Feedwater pump suction pressure K/A Number: A3.03 CFR

Reference:

41.7 / 45.5 Tier: 2 RO Imp: 2.5 RO Select: Yes Difficulty: 2 Group: I SRO Imp: 2.6 SRO Select: Yes Taxonomy: K Question: RO:1 SRO:j 43 Unit 1 is operating at 100% power with no abnormal conditions or alignments.

B MFP SUCT PRESS LO (K07-C8) annunciator is received.

Where can the Control Room Operators read the B MFW pump suction pressure WITHOUT leaving the control room?

A. The B MFP Lovejoy Operator Control Station (OCS).

B. B MFP Suction Pressure (P1-2830) indicator.

C. B MFP Suction Pressure computer point (P2830)

D. The Operator Information Touchscreen (OIT).

Answer:

C. B MFP Suction Pressure computer point (P2830)

Notes:

is not (a.) & (d.) are incorrect. These panels are located in the control room, however, MFP suction pressure available on these panels.

(b.) is incorrect. This indicator is located outside the control room.

which are (c.) is correct. This computer point is found on the Plant Computer and the SPDS computer both of available in the control room.

References:

STM 1-19, Rev. 11 History:

Developed for use in 98 RO Re-exam Selected for 2005 RO exam Selected for 2010 RO/SRO exam

Feedwater System STMI-19 Rev. 11 PS-2841 provides the second suction pressure trip signal used to satisfy the trip logic. Setpoint for PS-2841 is less than 200 psig.

Refer to table provided on the following page for suction pressure indications associated with the B MFP. Alarm and trip signal setpoints are identical to P-lA for P-lB.

PT-2830 provides suction pressure signal to plant & SPDS computers P1-283 0 provides local suction pressure indication at rack 21.

PS-2830 provides Lo & Lo-Lo alarms (K07-C8 & K07 B8).

Provides Suction Pressure trip signal.

PS-2835 provides suction pressure trip signal to MFP trip logic.

2.3.1.3 Suction Pressure Operation of the MFP s with suction pressure less than 230 psig Trip Logic can cause pump damage. To provide MFP protection and increase plant reliability, the MFP suction pressure logic was modified requiring two separate pressure signals to trip a MFP. To increase plant reliability and inadvertent trips due to suction pressure transients, time delays were installed. During normal operation one of the MFPs will be selected for the preferred pump to trip on low suction pressure. The preferred pump is selected by handswitch HS 6712 located on panel C02. HS-6712 positions are P-lA or P-lB.

Time delays associated with the preferred MFP trip are set at 40 seconds and 50 seconds for the remaining MFP.

The Lo-Lo and < 200 psig pressure switches provide the signals used to trip the preferred MFP and br both MFPs associated with switches discussed in the above section.

If suction pressure drops to <200 psig for greater than 40 seconds the preferred or selected MFP will trip. If Suction pressure remains less than 200 psig for an additional 10 seconds the remaining MFP will trip. Refer to Trip Logic String provided below.

Suction Pressure Trip Logic P ret. Selection Switch P IA DIA (HS-6712)

P IA S uction Press. Trip (PS-28414 200 PSI__

40 Sec K2A PIA Suction Pressfo-Lo 50 Sec (PS-2842)< 230 PSIG PIB P1B Suction Press. Trip PIA Trip K2B (P S -2835)

< 200 PSIG 50 Sec PIB Suction Press. Lo-Lu (PS-2830)<23OPSIG 40 Sec Pref. Selection Switch P 1B (HS-6712) PIB

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0789 Rev: 0 Rev Date: 9/14/2009 Source: New Originator: S. Pullin TUOI: Al LP-RO-FW Objective: 6 Point Value: I Section: 3.4 Type: Heat Removal From Reactor Core System Number: 059 System

Title:

Main Feedwater (MFW) System

==

Description:==

Knowledge of MFW design feature(s) and / or interlock(s) which provide for the following:

automatic trips for MFW pumps.

K/A Number: K4.16 CFR

Reference:

41.7 Tier: 2 RO Imp: 3.1 RO Select: Yes Difficulty: 2 Group: I SRO Imp: 3.2 SRO Select: Yes Taxonomy: K Question: RO:1 SRO: I Given:

- 100% power Which of the following interlocks provide an automatic trip of the Main Feed Water Pump?

A. Main Feed Water Pump suction pressure reading 220 psig for 45 seconds B. Main Feed Water Pump bearing oil pressure reading 15 psig C. Main Feed Water Pump discharge pressure reading 1360 psig D. Main Feed Water Pump vibration reading 14 mils Answer:

C. Main Feed Water Pump discharge pressure reading 1360 psig Notes:

A is incorrect, suction pressure would have to be less than 200 psig for 40 seconds.

B is incorrect, bearing oil pressure of 15 psig would cause an alarm but pressure must be less than 10 psig for a trip.

C is correct, pump discharge pressure of 1350 psig would result in a pump trip D is incorrect, the high vibration trip is bypassed when the pump is in operation

References:

STM 1-24 Rev. 11 History:

New selected for 2010 RO/SRO exam

Main Feedwater Pump Controls STM 1-25 Rev. 11 The pilot valve bleeds oil from the main valve disc, which allows a spring to open the main valve disc. Trip header oil pressure decreases and stop valves close as outlined above.

The overspeed trip valve will also open when auto-stop oil pressure decreases to zero (for instance, when the solenoid trip opens). This will seal in a main feedwater pump trip from the solenoid trip. To pressurize the trip header it is necessary to close the overspeed trip valve.

Closing the overspeed trip valve is accomplished through the use of the reset devices. A local reset button is used to reset (close) the overspeed trip. Depressing the reset button closes the main valve disc and pilot valve. This allows oil pressure to build up above the main valve disc and hold the valve closed. All trips must be reset to maintain the main valve closed; otherwise, the springs will open the main valve.

An overspeed trip reset solenoid valve is installed to allow the overspeed trip to be reset from a remote location. The overspeed trip reset valve will port high-pressure oil to a piston located on the shaft of the reset button. The high-pressure oil moves the piston which closes the pilot and main valve disc the same as depressing the local reset button. When the solenoid is not energized, the solenoid valve aligns the piston to drain and no pressure will be applied to the piston. The solenoid is energized when the trip-reset switch on C02 is positioned to the reset position or the reset switch at the front standard is taken to reset.

The trip lever is used to manually trip the feedwater pump.

Depressing the trip lever opens a drain that bleeds pressure from the top of the main valve disc. The main valve disc opens and the pump trips as above.

3.10 Solenoid Trip Valve The solenoid trip valves interact directly with the trip header to depressurize the trip header in response to various trip signals. The solenoid is energized to open the valve and trip the feedwater pump.

For redundancy, two solenoid trip valves are used in a parallel arrangement. Trips that will trip the solenoid trip valve are:

• Electronic overspeed trip

• Low suction pressure Two pressure switches used (200 and 230 psig) 40 seconds the preferred pump trips 50 seconds the non-preferred pump trips.

• High discharge pressure (two of three pressure switches at 1350 psig)

• Low bearing oil pressure (two of three pressure switches after a 3 second time delay)

Two pressure switches at 10 psig One pressure switch at 15 psig also supplies low pressure alarm

• High vibration (normally bypassed during operation) 12

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0270 Rev: I Rev Date: 11/8/05 Source: Direct Originator: D. Slusher TUOI: A1LP-RO-EFIC Objective: 29 Point Value: 1 Section: 3.4 Type: Heat Removal From Reactor Core System Number: 061 System

Title:

Auxiliary/Emergency Feedwater System

==

Description:==

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the AFW controls including: S/G level.

K/A Number: Al .01 CFR

Reference:

41.5 / 45.5 Tier: 2 RO Imp: 3.9 RO Select: Yes Difficulty: 2.5 Group: 1 SRO Imp: 4.2 SRO Select: Yes Taxonomy: Ap Question: RO:[ 45 SRO:

The EFIC automatic fill rate is designed to prevent overcooling.

With the plant in a degraded power condition and given a SG pressure of 885 psig, determine the proper OTSG fill rate by EFIC for the EFW system:

A. 3/min B. -4/min C. 5/min D. 6/min Answer:

B. 4/min Notes:

OTSG fill rate is adjusted so that OTSG levels raise at 2 inches/minute at OTSG pressure of 800 psig and 8 inches/minute at OTSG pressure of 1050 psig. This limits the overcooling effects of feeding OTSGs with EFW. At 885 psig OTSG fill rate will be 4 inches/minute. b is the correct answer.

References:

1105.005, Chg. 032 History:

Used in 1999 exam.

Direct from ExamBank, QID# 92 used in class exam Modified for 2005 Jon Gray RO re-exam.

Selected for 2010 RO/SRO exam

PROC.IWORK PLAN NO. PROCEDUREIWORK PLAN TITLE:

I PAGE: 7 of 91 1105.005 EMERGENCY FEEDWATER INITIATION AND CONTROL CHANGE: 032 6.0 SETPOINTS 6.1 Initiation Setpoints

• EFW low level initiate 11, delayed 9.9 seconds

• MSLI and EFW initiate on low SC pressure - 600 psig.

• Loss of both MFW pumps with reactor power >7%.

• ESAS Channel 3 or Channel 4 trip.

• MFW Flow in both loops <15% with reactor power >45%. (AMSAC)

• All RCPs OFF (May be bypassed at <10% Power) 6.2 Control Setpoints 6.2.1 SG level

• Low level control 31.

• Natural circulation control - 312.

• Ref lux boiling control - 378.

6.2.2 Rate of SG level rise when RCP5 are off is variable from 2 to 8 inches per minute depending on SC pressure. (2 inches per minute at 800 psig, 8 inches per minute at 1050 psig) 6.2.3 SG z\P - 100 psi determines good (unaffected) SC to allow EFW flow and isolates bad (affected) SC on MSLI actuation.

6.2.4 Atmospheric dump control valves will control SG pressure at 1020 psig at all times if not isolated.

6.3 Low condenser vacuum interlock opens atmospheric dump isolation valves at 21 Hg.

6.4 MSLI actuation opens affected SC atmospheric dump isolation valve.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0790 Rev: 0 Rev Date: 9/14/2009 Source: New Originator: S Pullin TUOI: A1LP-RO-EOP Objective: 9 PointValue: I Section: 3.6 Type: Electrical System Number: 062 System

Title:

A.C. Electrical Distribution

==

Description:==

Knowledge of local auxiliary operator tasks during emergency and the resultant operation effects.

K/A Number: 2.4.35 CFR

Reference:

41.10 / 43.5 / 45.13 Tier: 2 RO Imp: 3.8 RO Select: Yes Difficulty: 3 Group: I SRO Imp: 4.0 SRO Select: Yes Taxonomy: C Question: RO:J 46 SRO:J Given:

Unit I is in a Blackout condition.

Voltage has been recovered on SU#2 and is 155 kV To restore power to A-3 and A-4, what action along with its purpose is required by the Auxiliary Operator?

A. Perform Attachment 1, Blackout Breaker Alignment and UV Relay Defeat, to defeat UV Close Permissive interlocks to allow for starting of equipment necessary to protect the core.

B. Perform Attachment 1, Blackout Breaker Alignment and UV Relay Defeat, to prevent excess current during starting of the motors.

C. Perform Attachment 2, Recovery from Blackout Breaker Alignment and UV Relay Defeat, to allow for starting of equipment necessary to protect the core.

D. Perform Attachment 2, Recovery from Blackout Breaker Alignment and UV Relay Defeat, to allow Unit 2 to tie on non-vital loads on SU#2.

Answer:

A. Perform Attachment 1, Blackout Breaker Alignment and UV Relay Defeat, to defeat UV Close Permissive interlocks to allow for starting of equipment necessary to protect the core.

Notes:

A is correct, with degraded voltage on SU#2, Aft. 1 is required to defeat the UV interlocks.

B is incorrect, Aft. I would have no effect on actual starting current for motors C & D are incorrect, Att. 2 will only be performed when SU#2 voltage is greater than 158 kV.

References:

OP-1202.028 Change 010 History:

New selected for 2010 RO/SRO exam.

CHANGE 1202.008 BLACKOUT 010 PAGE 16 of 29 INSTRUCTIONS CONTINGENCY ACTIONS

44. Dispatch an operator to perform Attachment 1, Blackout Breaker Alignment and UV Relay Defeat.

NOTE Off-site power is considered restored to normal if either of the following conditions exists:

• SUI 22KV

• SU2 1 58KV j[ of the following conditions are met:

- Auto X-FMR energized from 500KV

- Auto X-FMR aligned to SU2 Unit 2 buses powered from SU2

- SU 2 V REG 3% reduction disabled A. IF off-site power is restored to normal, THEN dispatch an operator to perform Attachment 2, Recovery From Blackout Breaker Alignment and UV Relay Defeat AND RETURN TO step 8.

45. WHEN Attachment I is complete, THEN re-energize Al, A2, HI, and H2 by performing the following for each bus:

A. Check associated bus L.O. RELAY TRIP A. Determine AND correct cause of L.O.

alarm clear on K02. RELAY TRIP before energizing bus, while continuing with this procedure (Refer to Electrical System Operation (1107.001),

Re-closing Tripped Bus or MCC Feeder Breakers section).

B. IF buses are to be energized from SU2, THEN notify Unit 2.

C. Turn SYNC switch for associated bus C. Reset breaker anti-pump feature by taking feeder breaker ON handswitch to PULL-TO-LOCK AND releasing.

AND

1) jf neither Al nor A2 is energized, close breaker from handswitch. THEN RETURN TO step 33.

CHANGE 1202.008 BLACKOUT 010 PAGE 17 of 29 INSTRUCTIONS CONTINGENCY ACTIONS

46. Re-energize A3 and A4 by performing the following for each bus.

A. Check associated bus L.O. RELAY TRIP A. Determine AND correct cause of L.O.

alarm clear on K02. RELAY TRIP before energizing bus, while continuing with this procedure (Refer to Electrical System Operation (1107.001),

Re-closing Tripped Bus or MCC Feeder Breakers section).

B. Turn SYNC switch for associated bus B. IF non-vital bus voltage is <3160V, feeder breaker ON THEN dispatch an operator to close A3 and A4 feeder breakers in LOCAL to override Sync-check Relays (A-309 and 409).

close breaker from handswitch.

CAUTION

• During degraded voltage conditions the following problems may occur:

- Motors may trip on overload, overheat due to high running currents, or stall.

- MCC starter may pick up to energize loads.

- AC auxiliary relays may pick up to provide interlock or load energization features.

• Motors should be started one at a time and allowed to reach run speed to minim further ize voltage degradation.

• If both Units are aligned to SU2, coordination between Units is required when startin g loads.

47. Restart only equipment absolutely necessary to protect the core as follows:

A. Verify suction and discharge flow path aligned.

B. Review system operating procedure to ensure essential pump services available.

C. Consider closing centrifugal pump discharge valve before starting to reduce starting current.

(47. CONTINUED ON NEXT PAGE)

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0316 Rev: 0 Rev Date: 9/5/99 Source: Direct Originator: J Haynes TUOI: ANO-1-LP-RO-MU Objective: 3.5 Point Value: I Section: 3.6 Type: Electrical System Number: 062 System

Title:

A.C. Electrical Distribution

Description:

Knowledge of bus power supplies to the following: Major system loads.

KIA Number: K2.01 CFR

Reference:

CFR: 41.7 Tier: 2 RO Imp: 3.3 RO Select: Yes Difficulty: 2 Group: 1 SRO Imp: 3.4 SRO Select: Yes Taxonomy: K Question:

RO:j47 SRO:F 47 Which of the following would explain why a loss of bus Al will cause CV-1206 CRC Pump Seal Injection Block Valve) to close?

(Assume plant is at 100% power)

A. P36A (HPI) pump was the in-service pump.

B. Loss of instrument air to Seal Injection Control Valve, CV-1 207.

C. P36C (HP I) pump was the in-service pump.

D. Loss of instrument air to Pressurizer Level Control valve CV-1235.

Answer:

A. P36A (HPI) pump was the in-service pump.

Notes:

a is correct, if P36A was the in-service pump, then a loss of Al would cause a loss of A3, P-36A would cease to run, and CV-1206 would close when Seal Injection flow dropped to less than 22 gpm.

b is incorrect, CV-l 207 fails open on a loss of Instrument Air.

c is incorrect, a loss of Al would not affect P36Cs power supply, bus A4.

d is incorrect, CV-l 235 fails as-is on a loss of Instrument Air.

References:

1203.026, Change 11 History:

Used in 1999 exam.

Modified from ExamBank, QID# 3716.

Selected for 2010 RO/SRO exam.

CHANGE 1203.026 LOSS OF REACTOR COOLANT MAKEUP 011 PAGE 3 of 17 INSTRUCTIONS SECTION 1 -- LOSS OF HPI PUMP NOTE Indications of loss of HPI suction are:

. Erraticflow,and

. Erratic discharge pressure, and

. Control valves stable

1. jf HPI pump has lost suction, THEN stop the HPI pump.

2. Isolate letdown by performing one of the following:

. Close Letdown Coolers Outlet (CV-1221)

. Close both of the following on C18:

Letdown Coolers Outlet (RCS) (CV-1214)

Letdown Coolers Outlet (RCS) (CV-1216)

NOTE

. With HPI pump off, ICW cooling of RCP seals should provide adequate time to correct HPI pump or control problems, providing no pre-condition exists, such as excessive RCP shaft sleeve leakage.

HPI can provide necessary makeup for normal operations or plant shutdown.

. Reactor Coolant Pump and Motor Emergency (1203.031), Attachment A can be used as an aid to assess seal parameters.

3. Verify RC pump seals are being cooled by 1GW.

A. IF ICW to RCP seals is NOT available, THEN perform Reactor Coolant Pump and Motor Emergency (1203.031), Simultaneous Loss of Seal Injection and Seal Cooling Flow section.

4. Prepare to restart an HPI pump as follows:

A. jf OP HPI pump is unavailable Q STBY HPI pump is unavailable, THEN dispatch an operator to re-align the ES HPI pump per Attachment A of this procedure.

CHANGE 1203.026 LOSS OF REACTOR COOLANT MAKEUP 011 PAGE 4 of 17 SECTION 1 -- LOSS OF HPI PUMP (continued)

B. Place the following valves in HAND AND close:

• RC Pumps Total INJ Flow (CV-1207)

• Pressurizer Level Control (CV-1235)

C. Verify RCP Seal Injection Block (CV-1206) closes.

D. Select Safety System Diagnostic Inst display on SPDS for OP HPI pump AND evaluate suction pressure and flow stability prior to event.

E. IF loss of pump suction was indicated, THEN perform the following:

1) Verify Makeup Tank Outlet (CV-1275) open.

CAUTION Indicated suction pressure could be H 2 gas pressure only and is NOT absolute assurance of adequate volume of water. HPI pump operation with inadequate water volume can damage pump.

NOTE Addition of 600 gallons to the MU tank ensures a volume of water in the tank regardless of level indication.

2) IF CV-1 275 was NOT closed, THEN refill Makeup Tank (T-4) by adding 20 (600 gallons) using current RCS boron concentration.

5. IE STBY HPI pump is available, THEN perform the following:

A. Start Aux lube oil pump for STBY HPI pump.

B. GO TO step 8 to place STBY HPI pump into service.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0086 Rev: 0 Rev Date: 7/11/98 Source: Direct Originator: JCork TUOI: Al LP-RO-ELECD Objective: 37 Point Value: I Section: 3.6 Type: Electrical System Number: 063 System

Title:

D.C. Electrical Distribution

==

Description:==

Knowledge of the effect that a loss or malfunction of the dc electrical system will have on the following: Components using dc control power.

KIA Number: K3.02 CFR

Reference:

41.7 / 45.6 Tier: 2 RO Imp: 3.5 RO Select: Yes Difficulty: 3 Group: 1 SRO Imp: 3.7 SRO Select: Yes Taxonomy: K Question: RO:J 48 SRO:F 48 The plant is at 100% power.

Which of the following DC buses/panels, if cie-energized, would cause a reactor trip?

A. Panel D41 B. Panel RA1 C. MCCD15 D. Panel D21 Answer:

B. Panel RA1 Notes:

Only B is capable of causing a reactor trip due to loss of two RCP contact monitors.

The others would cause a loss of vital equipment capability but as seen in Att. J of 1107.004, they would not cause a trip.

References:

1107.004, Chg. 016 History:

Developed for 1998 RD exam Used in A. Morris 98 RD Re-exam Selected for use in 2005 RO exam, but not used.

Selected for 2010 RO/SRO exam.

PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 39 of 150 1107.004 BATTERY AND 125V DC DISTRIBUTION CHANGE: 016 ATTACHMENT J Page 1 of 15 Consequences and Required Actions For Opening 125V DC Breakers NOTE

• Some breaker operations render equipment inoperable and requires entry into Tech Spec LCO.

• Attachment J is not listed by priority. Locating grounds should begin with circuits of least consequences.

125V DC Bus DOl Breakers BREAKER CONSEQUENCES OF REQUIRED NUMBER DESCRIPTION OPENING ACTION Supply To MCC Dl5 Loss of power to MCC Dl5 DO1-2lA and EFW P7A valves. None DC Power Supply to Loss of Inverter Yll DC If available, place DOl-22A Inverter Yll Supply Inverter Y15 in service.

Supply to Panel Loss of power to PAl. Check PAl breakers DOl-23 PAl Reactor trip if 5o% individually first using power due to loss of PAl section of this (breaker handle not power to RCP Contact attachment. Verify connected- Monitor input to RPS. reactor power <50% and not fused supply) MSIVs open if instrument in 3 RCP operations.

air is not isolated If MSIVs are closed, verify instrument air is isolated.

Emer Supply to Loss of emergency supply Verify D2l is powered from DOl-24 Panel D2l to panel D2l bus D02 (breaker handle not connec ted -

fused supply)

Supply From Battery Disconnects battery Verify battery charger D0l-4l Charger DO3A charger from bus DOl DO3A not in operation.

Supply From Battery Disconnects battery Verify battery charger DOl-42 Charger DO3B charger from bus DOl DO3B not in operation DC Power Supply to Loss of Inverter Yl3 DC If available, place D0l-52B Inverter Y13 Supply Inverter Y15 in service.

DC Power Supply to Loss of Inverter Y15 DC If available, place D0l-53A Inverter Yl5 Supply Inverter Yll (for RS-l) or Inverter Yl3 (for RS-3) in service.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0791 Rev: 0 Rev Date: 9/14/2009 Source: Originator: S. Pullin TUOI: Al LP-RO-EDG Objective: I 9a Point Value: 1 Section: 3.6 Type: Electrical System Number: 064 System

Title:

Emergency Diesel Generators (ED/G)

==

Description:==

Knowledge of bus power supplies to the following: Air Compressor K/A Number: K2.0I CFR

Reference:

41.7 Tier: 2 RO Imp: 2.7 RO Select: Yes Difficulty: 2 Group: 1 SRO Imp: 3.1 SRO Select: Yes Taxonomy: K Question: RO:j 49 SRO:F What is the power supply to Emergency Diesel Generator Starting Air Compressors, C4AI and C4B2?

A. B31 and B41 B. B32 and B42 C. B51 and B61.

D. B52 and B62 Answer:

A. B31 and B41 Notes:

A is correct, the other choices are alternate possibilities.

References:

OP-l 107.001 Change 73 History:

New for 2010 RO/SRO exam.

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INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0792 Rev: 0 Rev Date: 9/14/2009 Source: New Originator: S. Pullin TUOI: A1LP-RO-EDG Objective: 19 Point Value: 1 Section: 3.6 Type: Electrical System Number: 064 System

Title:

Emergency Diesel Generators (ED/G)

Description:

Knowledge of the physical connections and / or cause-effect relationships between the ED/G system and the following systems: Starting air system.

K/A Number: KI .05 CFR

Reference:

41.2 to 41.9 / 45.7 to 45.8 Tier: 2 RO Imp: 3.4 RO Select: Yes Difficulty: 3 Group: I SRO Imp: 3.9 SRO Select: Yes Taxonomy: C Question: RO:150 50 SRO:j Given:

Plant at 100%

Performing #1 EDG monthly surveillance per 1104.036 Supplement 1 The CBOT presses the start pushbutton on dO K01-B2, EDG I OVERCRANK, alarms What is the cause of the alarm and how long did the starting air system attempt to start the engine?

A. #1 EDG did not exceed 300 rpm in 45 seconds and air start motors engaged for 8 seconds.

B. #1 EDG did not exceed 300 rpm in 8 seconds and air start motors engaged for 45 seconds.

C. #1 EDG did not exceed 30 rpm in 45 seconds and air start motors engaged for 2.5 seconds.

D. #1 EDG did not exceed 30 rpm in 8 seconds and air start motors engaged for 8 seconds.

Answer:

A. #1 EDG did not exceed 300 rpm in 45 seconds and air start motors engaged for 8 seconds.

Notes:

A is correct, due to meeting the annunciator logic B, C, and D are variations of the control logic for the starting air to the engine

References:

STM-1-31 rev 10 1203.012A change 038 History:

New 2010 RO/SRO exam

PROCJWORK PLAN NO. PROCEDUREMORK PLAN TITLE: PAGE: 17 of 183 1203.012A ANNUNCIATOR KOl CORRECTIVE ACTION CHANGE: 038 Location: ClO Device and Setpoint: EDG 1 OVERCRANK Alarm: K01-B2 1.0 OPERATOR ACTIONS

1. Place DG1 lockout switch in LOCKOUT position.

2. Reference TS 3.8.1, TS 3.8.2 and TS 3.8.3 for operability requirements.

3. Initiate action to determine cause of over-crank.

4. Operate fuel oil priming pump and verify return fuel sight glass is full.

5. WHEN cause of over-crank is corrected, THEN prove DG1 operable using Emergency Diesel Generator Operation (1104.036), Supplement 1.

6. IF DG1 inoperable, THEN verify proper MOD alignment for Service Water Pump (P-4B) and Makeup Pump (P-36B) per Makeup & Purification System Operation (1104.002) AND Service Water and Auxiliary Cooling System (1104.029).

7. Alarm may be cleared by ANY of the following methods:

• Place DG1 lockout switch in LOCKOUT position

• Depress local RESET button

• Place Local/Maint/Remote switch in MAINT

• Place DG1 Output (A-308) in PULL-TO-LOCK 2.0 PROBABLE CAUSES

• DG1 did not reach minimum speed within 45 seconds

• Loss of fuel oil pump prime

3.0 REFERENCES

• TS 3.8.1, TS 3.8.2 and TS 3.8.3

• Schematic Diagram Annunciator KOl (E-451)

• Schematic Diagram Diesel Generator Engine Control (E-l02)

Emergency Diesel Generators STM-1-31 Rev. 10 STM 1-31-43 EDG STARTING SEQUENCE 82

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0672 Rev: 0 Rev Date: 12/16/06 Source: Repeat Originator: Possage TUOI: A1LP-RO-RMS Objective: 8 Point Value: 1 Section: 3.7 Type: Instrumentation System Number: 073 System

Title:

Process Radiation Monitoring System ts as they apply to the PRM

==

Description:==

Knowledge of the operational implications of the following concep System: Radiation theory, including source s, types, units, and effects.

K/A Number: K5.01 CFR

Reference:

41.5 / 45.7 2 RO Imp: 2.5 RO Select: Yes Difficulty: 2 Tier:

Group: I SRO Imp: 3.0 SRO Select: Yes Taxonomy: K Question: RO:j 51 SRO: I 51 Monitor to monitor for steam What type of detector is used by the Main Condenser Air Discharge Radiation generator tube leaks?

A. Scintillation Detector B. Geiger Mueller Detector C. Ion Chamber Detector D. Beta Radiation Detector Answer:

A. Scintillation Detector Notes:

detector.

A is correct. The Main Condenser Air Dischagre Radiation Monitor is a scintillation B is incorrect. Area Monitors are G-M Detectors C is incorrect. Ion chambers are used for RP surveys D is incorrect. The Penentration Ventilation Monitors are beta sensitive monitors.

References:

STM 1-62 Rev. 11 History:

New for 2007 RO Exam.

Selected for 2010 RO/SRO exam

Radiation Monitoring STM 1-62 Rev. 11 2.2.6 Liquid Radwaste The Liquid Radwaste Monitor is an in-line monitor located in Monitor the liquid Radwaste common discharge line prior to its connection to the flume. The connection is between CZ-58 and CV-4642 and the monitor is physically located on the 335 elevation of the auxiliary building by the discharge flume. Liquid Radwaste passes through the pipe section of the sampler and is monitored by a gamma sensitive scintillation detector (RE-4642). The detector count rate is displayed on the digital rate meter located in the Control Room (C-25, Figure 62.14). There is an input to SPDS and the plant computer as well as a recorder readout on RR-4830.

The Liquid Radwaste Monitor is used to determine radioactive discharge activity during a release and to shut off the discharge should a pre-determined level of radioactivity be reached. On a high radiation level solenoid valve, SV-4642 operates to shut CV-4642, terminating the liquid Radwaste release. An annunciator in the Control Room will alarm on high radiation.

2.2.7 Main Condenser Air The Main Condenser Air Discharge Radiation Monitor is an in-Discharge Radiation line monitor on the combined suction line of the condenser vacuum Monitor pumps. The detector (RE-3632) is a gamma sensitive scintillation detector and is located on a platform above and just south of the condenser vacuum pumps. The detector count rate is displayed on the digital rate meter located in the Control Room (C-25, Figure 62.14). There is an input to both SPDS and plant computer as well as recorder readout on RR-4830.

The purpose of the Main Condenser Air Discharge Radiation Monitor is to detect activity resulting from a steam generator tube leak. On a high radiation, an annunciator in the Control Room alarms.

2.2.8 Waste Gas The Waste Gas Radiation Monitor is an in-line monitor in the Radiation Monitoring gaseous Radwaste system discharge to the vent plenum. This monitor is down stream of gaseous discharge shutoff valve CV-4830 and is located on the 404 elevation of the auxiliary building in the CRD transformer (X-8) room. The detector is a gamma sensitive scintillation detector (RE-4830). The count rate is displayed on a digital rate meter located in the Control Room (C-25, Figure 62.14) and provides an input to both SPDS and plant computer. There is also a recorder readout of radiation level on recorder RR-4830.

On a high radiation level, an annunciator in the Control Room alarms. At this alann setpoint, solenoid valve, SV-4830, operates to shut CV-4830, isolating gaseous Radwaste discharge to the station vent plenum. Also, the following will take place: CV-4820 will be shut by solenoid valve, SV-4820, to isolate the Waste Gas Tanks discharge header; Solenoid valve, SV-4806, operates to open CV 4806, to direct miscellaneous vents from the components in the Auxiliary Building to the Waste Gas Surge Tank.

18

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0793 Rev: 0 Rev Date: 9/15/2009 Source: Direct Originator: S Pullin TUOI: A1LP-RO-MSSS Objective: I Point Value: I Section: 3.4 Type: Heat Removal From Reactor Core System Number: 076 System

Title:

Service Water System (SWS)

==

Description:==

Ability to manually operate and / or monitor in the control room SWS valves KIA Number: A4.02 CFR

Reference:

41.7/45.5 to 45.8 Tier: 2 RO Imp: 2.6 RO Select: Yes Difficulty: 3 Group: 1 SRO Imp: 2.6 SRO Select: Yes Taxonomy: Ap Question: RO: 52 SRO:I When starLing Service Water Pump P-4A after maintenance, you observe the following symptoms.

- Pump start is indicated by normal light indication above pump control HS on.

- Annunciator K10-B3 SW DISCH PRESS HI alarms.

- Valve position indication in the control room indicates proper valve alignment.

- SW Bay levels are 338 feet

- No change in SW flow or discharge pressure indications on the SPDS Diagnostics screen.

- No change in SW Loop pressure indications on control room panel C09.

Which of the following is the most likely cause of these symptoms?

A. The pump discharge valve was not opened when returned to service.

B. The pump did not start when pump breaker closed.

C. P-4A cannot pump into the system because of high system pressure from the other(running) pump.

D. P-4A is running without sufficient NPSH to pump water into the SW System.

Answer:

A. The pump discharge valve was not opened when returned to service.

Notes:

A is the correct answer. With the local discharge valve closed, the SW Pump would not be able to pump water to the loop, but since the discharge pressure switch is between the pump and discharge vlave, therefore a high discharge pressure would be seen.

B is incorrect, if the pump did not start there would not be a high discharge pressure alarm.

C is incorrect, if the maintenance performed had caused low discharge pressure such that the pump was unable to pump water to the loop, there would not be a high discharge pressure alarm.

D is incorrect, with a bay level of 338 feet, suction pressure would be (356.5-338)0.433= 8 psig which is adequate.

References:

OP-1203.012l Change 046 History:

Direct ANO Exam bank QID ANO-OPSI-3284 Selected for 2010 RO/SRO exam

PROCJWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: I PAGE: 29 of 68 1203.01 21 ANNUNCIATOR K10 CORRECTIVE ACTION CHANGE: 046 I I Location: Cl6 Device and Setpoint:

SW Pump P4A running & P4A Disch Press (PS3611) >90 psig SW PUMP SW Pump P4B running & P4B Disch Press (PS3609) >90 psig DISCH PRESS SW Pump P-4C running & P-4C Disch Press (PS36l0) >90 psig HI Alarm: Kl0-B3 OPERATOR ACTIONS

1. Determine which pump is in alarm.

2. IF experiencing a loss of Service Water OR degraded Service Water flow, THEN GO TO Loss of Service Water (1203.030).

3. IF lake temperature is low OR cold weather operations with low ACW/SW demand, THEN consider throttling open ICW Coolers Loop 1 and 2 SW Bypass (SW4026A and SW4026B)

4. IF Talt is installed from ICW Cooler (E28C) outlet, THEN throttle open temporary valve Tl.

5. Place additional SW/ACW loads into service as needed.

2.0 PROBABLE CAUSES NOTE This annunciator has multiple input without reflash.

1. Improper SW Pump discharge alignment

2. Cold lake temperatures causing low ACW/SW demand

3.0 REFERENCES

1. Schematic Diagram Annunciator Kb (E460, sheets 1 3)

2. NRC Commitment P 6186, Provide procedure for cause, action, and how to clear alarms of DHR.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0794 Rev: 0 Rev Date: 9/15/2009 Source: New Originator: S. Pullin TUOI: Al LP-RO-ESAS Objective: 20 Point Value: I Section: 3.4 Type: Heat Removal From Reactor Core System Number: 076 System

Title:

Service Water System

==

Description:==

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the SWS controls including: Reactor and turbine building closed cooling water temperatures.

KIA Number: Al .02 CFR

Reference:

41.5 I 45.5 Tier: 2 RO Imp: 2.6 RO Select: Yes Difficulty: 2 Group: I SRO Imp: 2.6 SRO Select: Yes Taxonomy: K Question: RO:i 53 SRO:1 What would be the effect to service water pressure due to an inadvertent actuation of ES Channel 5?

A. Service Water Pressure would drop due to SW valves to the EDG Coolers opening.

B. Service Water Pressure would drop due to SW valves to the RB Coolers opening.

C. Service Water Pressure would rise due to ACW isolation valve closing.

D. Service Water Pressure would rise due to SW to ICW isolations closing.

Answer:

B. Service Water Pressure would drop due to SW valves to the RB Coolers opening.

Notes:

A is incorrect, SW to EDG Coolers open on diesel start. EDG starts on Channels I or 2 B is correct, ES Channel 5 will align SW to the RB Coolers C is incorrect, ACW isolation valve would close on ES Channel 2 D is incorrect, SW to ICW isoaltion valve will close on ES Channels 1 and 2

References:

STM 1-65 Rev. 5 History:

New, Selected for 2010 ROISRO exam

Engineered Safeguards Actuation System STM 1-65 Rev 5

• CV-4446 closes to prevent the RB Sump from draining to the Auxiliary Building Sump.

• CV-1052 closes to isolate the Quench Tank and CV-1845 and 1054 close the Quench Tank sample isolations.

4.12.2 Low Pressure Low Pressure Injection is also initiated by the 1590 psig low Injection and RCS pressure and the 4 psig high RB pressure, these signals Diverse actuate the following equipment: (Channels 3 & 4)

Containment Isolation

• Both P34A and B start (DH Pumps).

• The LPI Block Valves open, CV- 1400 and 1401.

• CV-1407 and 1408, BWST Outlet Valves open.

• BWST Recirc Isolation Valves CV- 1441 and CV- 1438 will receive a close signal from their associated BWST Isolation.

• CV-1053 closes to isolate the Quench Tank.

• CV-5612 and 5611 close to isolate the RB from the Fire Water System.

• CV-7403, CV-7404 and CV-7401 and CV-7402, RB purge and isolations close.

• CV-7454 and 7453, RB Air Particulate Monitor isolation is closed.

• CV-4400, RB Sump drain to the Auxiliary Sump is closed.

• CV-1 667 isolates nitrogen to the Quench Tank. See Note 1 below.

NOTE 1: Credit is no longer taken for the ES function for CV 1667. N2-47 performs the function of containment isolation.

4.12.3 Reactor Building RB isolation and cooling (Channel 5 and 6 is initiated by Cooling and high Reactor Building pressure of 4 psig, and as its name Isolation implies, its function is to isolate and cool the RB. The following equipment is actuated:

• CV-2234, 2235, 2220 and 2221 close to isolate Non-Nuc ICW to RC Pump Air/LO and CRD Coolers.

• CV-22 14, CV22 15 and CV-2233 close to isolate Nuc ICW to Letdown and RCP Seal Coolers.

35

Engineered Safeguards Actuation System STM 1-65 Rev 5

• CV-6205, CV-6202 and CV-6203 close to isolate the RB Chillers.

• The RB Coolers Inlet and Outlet Valves open to VCC 2A, B, C & D (CV-38l2, CV-3814 and CV-3813, CV-3815).

• RB Cooling Fan A, B, C & D start and SV-7410, SV-741 1, SV-74l2 and SV-7413 (RB Bass Dampers open.

• VEF-38A or B, Penetration Room Fans start.

• CV-2235, CRD Cooling Coil Inlet Isolation Valve closes.

• CV-1065, Quench Tank Cond. Isolation closes.

4.12.4 Reactor Building Reactor Building Spray and Chemical Addition Spray components are actuated when RB pressure reaches 30 psig.

The components actuated are:

• P35A & B RB Spray Pumps start.

• CV-2401 and 2400 RB Spray Blocks open.

• CV-1616 and 1617 open to supply Sodium Hydroxide to the Spray Pumps.

5.0 Technical Specifications The Technical Specification requirements for the Engineered Safeguards Actuation System are found in:

21. Identify the Technical

• 3.5 Instrumentation Systems Specification requirementsfor ESAS. 0 3.5.1 Operational Safety Instrumentation 3.5.1.1 Requirements of Table 3.5.1-1

. 3.5.1.2 Number of channels below that required.

O Table 3.5.1-1 Instrumentation Limiting Conditions for Operation O 3.5.3 Safety Features Actuation Setpoints

• 4.1 Operational Safety Items 0 Table 4.1-1 Instrument Surveillance Requirements.

The Technical Specification requirements for the systems and the components actuated by ESAS are covered in the respective systems System Training Manual.

36

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0535 Rev: 1 Rev Date: 10/13/200 Source: Direct Originator: J.Cork TUOI: A1LP-RO-AOP Objective: 3 Point Value: 1 Section: 3.8 Type: Plant Service Systems System Number: 078 System

Title:

Instrument Air System

Description:

Knowledge of the physical connections and / or cause-effect relationships between the lAS and the following systems: Service Air K/A Number: Ki .02 CFR

Reference:

41.7 / 45.5 Tier: 2 RO Imp: 2.7 RO Select: Yes Difficulty: 3 Group: 1 SRO Imp: 2.8 SRO Select: Yes Taxonomy: K Question:

RO:J SRO:F Instrument Air pressure has dropped to 50 psig.

Which of the following manual or automatic actions should be performed or will occur in response to the low Instrument Air pressure?

Note: All actions for higher pressures have been completed at the required pressure and answer the question considering only the action for the current pressure.

A. Service Air to Instrument Air cross-connect automatically opens.

B. Open Unit I to Unit 2 Instrument Air cross-connect.

C. Trip Reactor, actuate EFW and MSLI on both SGs.

D. Close Letdown Cooler outlet to isolate Letdown.

Answer:

A. Service Air to Instrument Air cross-connect automatically opens.

Notes:

B is incorrect, this was done when pressure dropped to 75 psig.

A is correct, this automatically occurs when pressure drops to 50 psig.

c is incorrect, this would not be done until pressure was less than 35 psig.

D is incorrect, this would not be done until pressure was less than 35 psig.

References:

1104.025, Chg. 014 History:

Developed for 1998 RO exam (similar to QID 102)

Modified question for A. Morris 98 RO Re-exam Modified for J. Gray 2005 re-exam.

Selected for 2010 ROISRO exam.

6.2 Compressor trips on any of the following:

6.2.1 Electrical fault.

6.2.2 After cooler discharge temp high:

• C-3A After Cooler Disch Air Temp High (TS-5405) 125°F

• C-3B After Cooler Disch Air Temp High (TS-5407) 125°F 6.2.3 Lube oil pressure low: 8 psig for >10 seconds

• C-3A Low Lube Oil Press (PS-5434)

• C-3B Low Lube Oil Press (PS-5436) 6.3 Interlocks 6.3.1 Compressor start opens cooling water solenoid valve:

• E-19A After Cooler ICW Inlet (SV-2251)

• E-19B After Cooler ICW Inlet (SV-2250) 6.3.2 50 psig dropping IA pressure opens Inst Air X-over (SV-5400) and closes at -54 psig rising IA pressure.

6.4 SA Compressor Alarms 6.4.1 Compressor cooling water outlet temp high: 125°F.

• C-3A ICW Disch Temp (TS-2261)

• C3B ICW Disch Temp (TS-2260) 6.4.2 Compressor discharge air temp high: 340°F.

• C-3A JJisch Air Temp High (TS-5404)

• C-3B Disch Air Temp High (TS-5406) 6.4.3 After cooler discharge air temp high: 110°F

• C-3A After Cooler Disch Air Temp High (TS-5405)

• C-3B After Cooler Disch Air Temp High (TS-5407) 6.4.4 Lube oil pressure low: 15 psig

• C-3A Low Lube Oil Press (PS-5434)

• C-3B Low Lube Oil Press (PS-5436)

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0795 Rev: 0 Rev Date: 9/15/2009 Source: Direct Originator: S. Pullin TUOI: AILP-RO-RBS Objective: 11 Point Value: 1 Section: 3.5 Type: Containment Integrity System Number: 103 System

Title:

Containment System

==

Description:==

Ability to manually operate and I or monitor in the control room: Operation of the containment personnel airlock door KIA Number: A4.06 CFR

Reference:

41.7 / 45.5 to 45.8 Tier: 2 RO Imp: 2.7 RO Select: Yes Difficulty: 3 Group: I SRO Imp: 2.9 SRO Select: Yes Taxonomy: C Question: RO:j 55 SRO:1 55 Given:

Plant refueling is in progress The Reactor Building Coordinator calls the control room and reports the following:

The inner door of the reactor building personnel hatch will not close The outer door is operable In accordance with Technical Specifications for Refueling Operations, how does this affect fuel movement?

A. Irradiated fuel movement in the reactor building and auxiliary building must be suspended.

B. Irradiated fuel movement in the reactor building must be suspended.

C. Irradiated fuel movement in the auxiliary building must be suspended.

D. Irradiated fuel movement may continue without restriction.

Answer:

D. Irradiated fuel movement may continue without restriction.

Notes:

D is correct, fuel movement may continue in both the Reactor Building and Aux Building provided one of the air lock doors is capable of being closed.

A, B, and C are incorrect due to the outer door being operable.

References:

T.S. 3.9.3 Amendment No. 215 History:

Direct from AND exam bank ANO-OPSI-6622 Selected for 2010 RO!SRO exam.

Reactor Building Penetrations 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Reactor Building Penetrations LCO 3.9.3 The reactor building penetrations shall be in the following status:

a. The equipment hatch is capable of being closed;

b. One door in each air lock is capable of being closed; and

c. Each penetration providing direct access from the reactor building atmosphere to the outside atmosphere either:

1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or

2. capable of being closed by an OPERABLE reactor building isolation valve, except reactor building purge isolation valves, or

3. capable of being closed by an OPERABLE reactor building purge isolation valve with the purge exhaust radiation monitoring channel OPERABLE.

APPLICABILITY: During movement of irradiated fuel assemblies within the reactor building.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more reactor A.1 Suspend movement of Immediately building penetrations not in irradiated fuel assemblies required status. within the reactor building.

ANO-1 3.9.3-1 Amendment No. 215

1 0 CD N) -

hJ) CD 0

m N) 0 3

ES-401 PWR Examination Outline Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems Tier 2/Group 2 (RO)

System # / Name K K K K K K A *A A A G K/A Topic(s) IR # QID TYPe 1 2 3 4 5 6 1 2. 3 4 1(2.05 changed to 1(2.02 001 Control Rod Drive X One-line diagram of power 3.6 56 429 0 supply to trip breakers 002 Reactor Coolant X A2.01- Loss of coolant 4.3 57 604 D inventory A3.03- Charging and 011 Pressurizer Level Control X letdown 3.2 58 797 N 014 Rod Position Indication X K4.05 Rod hold interlocks 3.1 59 308 D 015 Nuclear lnstwmentation x 1(3.04 ICS

- 3.4 60 299 D 016 Non-nuclear Instrumentation X 1(5.01- Separation of control 2.7 61 77 D protection circuits K6.01- Sensors and 017 In-core Temperature Monitor X detectors. 2.7 62 240 D 027 Containment Iodine Removal Not selected N/A 028 Hydrogen Recombiner Not selected N/A and Purge Control 029 Containment Purge Not selected N/A 033 Spent Fuel Pool Cooling . Not selected N/A 034 Fuel Handling Equipment Not selected N/A 035 Steam Generator Not selected N/A 041 Steam Dump/TUrbine Not selected N/A Bypass Control 045 Main Turbine Generator X A4.06- Turbine stop valves 2.8 63 138 D 055 Condenser Air Removal Not selected N/A 056 Condensate Not selected N/A 068 Liquid Radwaste 1(4.01 Safety and

.. environmental precautions for

• handling hot, acidic, and N/A

. radioactive liquids Rejected

. system to 014 Rod Position Indication

K3.05 ARM and PRM 071 Waste Gas Disposal systems Rejected system N/A to 015 Nuclear Instrumentation 072 Area Radiation Monitonng Not selected N/A 075 Circulating Water .X 2.4.11- Knowledge of 4.0 64 798 N abnormal condition procedures 079 Station Air Not selected N/A 086 Fire Protection X - A1.01- Fire header pressure 2.9 65 542 0 K/A Category PointTotals: 0 1 1 1 J((f_it Fj Group PointTotal: 10 ES-401 Form ES-401-2

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0429 Rev: 0 Rev Date: 4/30/2002 Source: Direct Originator: S.PuIlin TUOI: AILP-RO-CRD Objective: 8 Point Value: I Section: 3.1 Type: Reactivity Control System Number: 001 System

Title:

Control Rod Drive System

==

Description:==

Knowledge of bus power supplies to the following: One-line diagram of power supply to trip breakers KIA Number: K2.02 CFR

Reference:

41.7 Tier: 2 RO Imp: 3.6 RO Select: Yes Difficulty: 2 Group: 2 SRO Imp: 3.7 SRO Select: Yes Taxonomy: C Question: RO:J 56 SRO:J If breaker B631 opened while operating at 100% power, the response of the Control Rod Drive system would be:

A. A ratchet trip of all regulating rods since half of the power supply has been removed.

B. No effect on regulating rods, safety rods are held by a single phase (CC) energized.

C. A ratchet trip of the safety rods due to a single phase remaining energized.

D. A trip of all safety rods since the main power has been removed.

Answer:

B. No effect on regulating rods, safety rods are held by a single phase (CC) energized.

Notes:

B is correct. The one-line diagram shows the power supply configuration from A-501 providing power to the CC phase on the DC hold bus which will maintain the safety rods out. Regulating rods are not effected normal movement will be supplied by the Bus 2 power supplied by A-501.

References:

STM 1-02, Control Rod Drive System, page 9, step 2.4 History:

Direct from regular exambank QID 4208.

Selected for use in 2002 RO/SRO exam.

Selected for 2010 RO/SRO exam.

Control Rod Drive System STMI-02 Rev.

UHDH DCRNRNH °1BRDCKE.

A H

-p 0:

IOVIR S TO OPT RODS TO S FN HO DC TO OPT RODS TOOPH ROOD ERR C

A A EC ICC TOOPS COOPS 006PS TOOPS 152 342 152 340 NOTES:

SYMBOLS:

TO BEEHiVE ALL POWER FROST ALL ONE OF TIlE FOLLCMJING TRIP (PSWER MUST CAGE RE A TRIP SI3NAI.S ICRDAXCROIJ) *.CRDAXC050)*(C009XCHDC)*CROA.TCRO WHICH REDUCES HO END A

• CAD DI X(CRDC

• CR0 DIWHIEH IS OAT-OF-i B CLAy AC T CATION UNLESS A MALFUNCTION OCCURS IN TIlE TRIPPING OFFOUTE ALL UHUAHCRDARE OPEN AND ALLGATPD ,RRETUHNEO WIlES A TOUT-OFi TRIP CONDITION OCCURS IA THE FIGURF 02.34: CRD TRIP SIGNAl S 64

Control Rod Drive System STM 1-02 Rev. 8 FiGURE 02.30: CR1.) 120 VOLT POWER SUPPLIES A

MA[N AC FROM VOLTAGE IRANSFC)RMFR TRIP 13631 RLGULATOR 480/120 VAC BKR 3 PHASE/6 PHASE V

CONTROL PWR XFMR TRANSFORMER 10 OR 5.6. 7. 8 120 VAC 13US I

- AND AUX.,

GATE I)RIVES 208 VAC 120 VAC 1O GR 5, 6, 7, 8 TO TO ART AND AUX..

ABT- I ABT-2 3 PROGRAMMERS INSL1 INSL2 A TOGR5.6.7.8 120 VAC BUS2 AN1)AUX..

PROGRAMMERS AND GATE DRIVES CONTROL PWR XFMR TRANSFORMI R A

A SECONDARY AC FROM VOLTAGE TRANSFORMER TRIP A501 REGULATOR 480/120 VAC BKR 3 PHASE/6 PHASE 58

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0604 Rev: 0 Rev Date: 6/30/05 Source: Direct Originator: S.Pullin TUOI: Al LPR-RO-RCS Objective: 5 Point Value: 1 Section: 3.2 Type: Reactor Coolant System lnventroy Control System Number: 002 System

Title:

Reactor Coolant System (RCS) or operations on the RCS and (b)

==

Description:==

Ability to (a) predict the impacts of the following malfunctions based on those predictions, use procedures to correct , control, or mitigate the consequences of t

those malfunctions or operations: Loss of coolan invent ory.

KIA Number: A2.0l CFR

Reference:

41.5 / 43.5 / 45.3 / 45.5 2 RO Imp: 4.3 RO Select: Yes Difficulty: 2 Tier:

Group: 2 SRO Imp: 4.4 SRO Select: Yes Taxonomy: C Question: RO:j 57 SRO:j 57 r Trip.

A reactor trip has occurred and the CRS is directing actions per 1202.001, Reacto Assume all actions have been performed when required by system parameters.

The CBOR reports that Pressurizer level has fallen to 30 and continuing to drop.

Pressurizer Level Control (CV-1 235) is in Auto and fully open.

Which of the following is the proper response?

A Initiate HPI per RT-2.

B. Reduce Letdown by closing Orifice Bypass (CV-l 223).

C. Isolate Letdown by closing Letdown Cooler Outlet (CV-1 221).

D. Operate CV-1235 in HAND to control PZR level 90 to 110.

Answer:

A Initiate HPI per RT-2.

Notes:

Answer A is correct, this is done when level is < 30 per 1202.001.

iate actions.

Answer B is incorrect, this was done early in the procedure, shortly after immed Answer C is incorrect, this was done earlier when level was < 50g.

not help.

Answer D is incorrect, CV-1235 is operating properly in Auto, taking it to hand would

References:

1202.001, Chg. 031 History:

New for 2005 RO exam, modified as a replacement question.

Selected for 2010 RO/SRO exam.

I CHANGE 1202.001 REACTOR TRIP 1 031 PAGE 19 of 25 INSTRUCTIONS CONTINGENCY ACTIONS

26. Check Pressurizer Level Control valve 26. Perform the following:

(CV-1 235) maintains PZR level> 55.

A. IF CV-1 235 fails to respond in AUTO, THEN operate CV-1235 in HAND to control PZR level 90 to 110.

B. IF PZR level is < 55 with no indication of recovery, THEN isolate Letdown by closing either:

Letdown Cooler Outlet (CV-1221),

OR Letdown Cooler Outlets (CV-1214 and 1216).

C. jf PZR level drops below 55, THEN verify Pressurizer Heaters off.

D. E PZR level drops below 30, THEN initiate HPI (RT 2).

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0797 Rev: 0 Rev Date: 9/15/2009 Source: New Originator: S. Pullin TUOI: AILP-RO-MU Objective: 4 Point Value: 1 Section: 3.2 Type: Reactor Coolant System Inventory Control System Number: 011 System

Title:

Pressurizer Level Control System (PZR LCS)

==

Description:==

Ability to monitor automatic operation of the PZR LCS, including: Charging and letdown.

K/A Number: A3.03 CFR

Reference:

41.7 /45.5 Tier: 2 RO Imp: 3.2 RO Select: Yes Difficulty: 3 Group: 2 SRO Imp: 3.3 SRO Select: Yes Taxonomy: Ap Question: RO: 58 SRO: 58 Given:

Plant at 100%

Letdown flow 80 gpm indicated on Fl-I 236 Letdown pressure 50 psig on P1-1237 CV-1 244 and CV-1245 Letdown DI Inlet Isolation valves lose power.

With no operator action what would be the expected automatic response of the pressurizer level control system?

A. Fl-1236 would read 80 gpm, P1-I 237 would read 50 psig and Pressurizer level control valve CV-1235 position would not change.

B. Fl-1236 would read 0 gpm, P1-1237 would read 150 psig and Pressurizer level control valve CV-I 235 position would open.

C. Fl-1236 would read 85 gpm, P1-1237 would read 45 psig and Pressurizer level control valve CV-1 235 position would open.

D. FI-1236 would read 70 gpm, P1-1237 would read 150 psig and Pressurizer level control valve CV-1235 position would close.

Answer:

B. Fl-1236 would read 0 gpm, P1-1237 would read 150 psig and Pressurizer level control valve CV-1235 position would open.

Notes:

B is correct, due to letdown Dl Inlet Isolation Valves fail closed on a loss of power. Which would isolate letdown, letdown pressure would rise to the letdown relief setpoint of 150 psig, causing a LOCA. Pressurizer level would go down causing CV-I235 to open.

A,C, and D are variations of these possible combinations.

References:

STM 1-04 Rev. 9 History:

New for 2010 RO/SRO exam.

Primary Makeup And Purification STM 1-04 Rev. 9 MU-4, is a backup for either the letdown orifice or the control valve.

This valve, when fully open, passes as much flow as the control valve.

The manual valve is opened only if the control valve is shut and secured. Thus, the maximum flow capacity is 170 gpm through the control valve or the manual valve and 45 gpm through the orifice.

This yields a total of 215 gpm. The relief valve downstream of the letdown orifice is set for 150 psig and can pass up to 257 gpm. Even in the unlikely situation that all three paths are open simultaneously, which would require multiple operator error, the flow capacity of the makeup system combined with the relief valve prevents overpressuring the letdown line.

2.7.1 Letdown Orifice This air operated, solenoid actuated pneumatic valve is used to Isolation Valve CV-1 222 isolate the normal letdown stream . Its hand switch is located on Control Room Panel C04 (HS-1222). Loss of Instrument Air to the valve will cause it to fail as is.

2.7.2 Letdown Orifice CV-1223 is an air operated, electrically controlled valve, and is Bypass Valve CV-1 223 throttled from Panel C04 by the operator. Flow indicating controller FIC-l223 is used to electronically control flow around the letdown orifice line and give the operator final control of maximum letdown flow. CV-1223 is equipped with a voltage to pneumatic transducer, EIP-1223. Normal letdown purification flow is more than can be passed through the letdown flow orifice (FO-1222). Loss of Instrument Air to the valve will cause it to fail closed.

2.7.3 Letdown Flow This flow orifice was sized to limit letdown flow to Orifice FO-1 222 approximately 45 gpm at normal RCS pressure. At this flow rate, one complete RCS volume turnover occurs each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The orifice also causes a pressure drop from 2155 psi to about that of current MJU tanic pressure.

2.7.4 Letdown Flow This bypass or parallel orifice can be used to obtain more flow Orifice FO-1 220 during low pressure operations. It is also available for use if the letdown orifice bypass valve, CV-1223 is not available. It is placed in service manually by opening manual letdown valve MU4.

2.7.5 Letdown Temperature element TE-1221 monitors letdown temperature Temperature Element and operates TIS-1221. This temperature switch sends a signal to the and Switch letdown penetration isolation valve, CV-l 221, to close if letdown temperature reaches 135 °F. The interlock is designed to protect the resin of the purification demineralizers from damage due to excessively hot water.

2.8 Pressure Relief PSV-1236 is set for 150 psig and relieves pressure on the LD Valve PSV-1 236 piping should the downstream piping and components be isolated. It discharges to the Auxiliary Building Equipment Drain Tank 12

Primary Makeup And Purification STM 1-04 Rev. 9 (ABEDT). PSV-1236 can pass up to 257 GPM @ 10% above set pressure.

2.9 Letdown Flow This Letdown Flow Element (FE-1236) provides the operator Element, FE-I 236 indication of letdown flow on panel C04, indicator Fl- 1236, SPDS, and feeds the Plant Computer.

2.10 Letdown TE- 1237 is located on the letdown line on the 335 foot elevation Temperature Element of the Reactor Auxiliary Building. This temperature string consists of TE-1 237 a measuring mechanism and a pneumatic transmitting mechanism. It provides pneumatic signals for letdown temperature indicator TI-1237 on Control Room panel C04. The temperature element also feeds a temperature switch and an electro-pneumatic converter, EIP-1237.

The switch, TS-1237 provides electrical signals for the High Letdown Temperature annunciator alarm. The E/P converter supplies an electrical signal to the Plant Computer. Should letdown fluid temperature increase to greater than 1 30F, it causes annunciator Kb A8, LETDOWN TEMP HI to alarm.

2.11 Makeup Prefilter, The Makeup Prefliter is designed to be used in the event that F-25 extra filtering will be needed to filter out crud that would otherwise be entrained in the demineralizers. Crud could cause increasing radiation levels of the Dls and possibly shorten their useful life. RCS transient changes (pressure, temperature, pH and flow) can cause the release of crud within the RCS. F-25 is interconnected with, and normally used with, the decay heat removal system.

The Makeup Prefilter can be placed in service by verifing it isolated from the DI{R system and manually opening MU-S and MU-6 then closing Valve MU-7. Refer to figure 04.07. F-25 is used during plant start up and shut down to prevent excessive buildup of particulates in the demineralizers. It can be used anytime excessive particulate is indicated in the letdown system. The filter may be used during normal steady state operation. The filter is used when the decay heat system is in operation as part of one method of RCS drain down. This drain path is from the decay heat system, through F-25 to the letdown line and ultimately to the Clean Waste Receiver Tanks (CWRT) via the Vacuum Degasifler.

The filter is located upstream of the purification deniineralizers and has a flow capacity of 140 gpm. In parallel with the filter is a differential pressure transmitter, PDIS-1400. At 25 psid across the filter MU Sys. F-25 Filter t\P HI annunciator alarms on control room annunciator Kl0-F7.

13

Primary Makeup And Purification STM 1-04 Rev. 9 remove ionic impurities and have some filtering capability for suspended materials. Each demineralizer contains a bed of mixed cation and anion resins. One unit is normally operating while the second unit is in standby.

The F-3A & B filters are used to keep resin fines and any particulate material that may pass through the Dls from entering the remainder of the purification system and the RCS.

2.13.1 Purification These air operated gate valves are used to place either of the Demineralizer Inlet purification demineralizers in service. CV-1244 is for Demineralizer Valves, CV-1244 and T-36A and CV-1245 is for Demineralizer T-36B. Both valves are CV-1 245 operated from panel C-04 and have solenoid actuated, air operated, single acting cylinder operators. One valve is normally open, the other normally closed. These valves will fail closed on loss of air or power.

2.13.2 Demineralizer This valve is used to bypass the purification demineralizers. Its Bypass Valve, MU-9 use is directed during recovery from high temperature conditions to prevent LD DI resin depletion. During high temperature conditions in the letdown line, this valve should be opened prior to opening the letdown isolation valve (CV-1221). MU-9 is a manual valve.

2.13.3 Purification The HOH mixed bed demineralizers (DPs) are used to remove Demineralizers (Dls), reactor coolant impurities from the letdown stream. Since the reactor T-36A1B coolant may be contaminated with dissolved fission and corrosion products, ion exchange resins are used to clean the reactor coolant.

The resins remove radioactive impurities and reduce the radiation levels that might otherwise be present in the RCS piping.

Normally, the operating demineralizer is saturated with boron at a concentration equal to RCS boron concentration. The standby demineralizer may be unsaturated. This allows use of the standby demineralizer to remove boron late in core life to keep the reactor operating. A positive reactivity addition hazard may occur if the wrong DI is placed in service during power operation. The mixed bed HOH resin will remove the boron from the water passing through it.

This will continue until the HOH resin comes up to an equiliabrium concentration of boron that equals RCS concentration. The RCS water passing into the demineralizer also may have an excess concentration of lithiwn-7, in the form of Lithium Hydroxide (LiOH).

LiOH is used for pH control, thus corrosion control of the reactor coolant system. The HOH resin will also remove the Lithium from the water passing through it. This will continue until the HOH resin comes up to an equilibrium concentration of lithium that equal the RCS concentration.

Maximum and minimum flows through one demineralizer are 123 gpm (to prevent resin compacting) and 25 gpm (to avoid channeling) respectively. Table 4.2 contains design data for the purification demineralizers.

16

Primary_Makeup And Purification STM 1-04 Rev. 9 TO HPI SEAL TO RE7URN LOOP MAJ^{6.5% from its API group average position.}

• Individual fault lamps on the P1 Panel indicate a rod is >5% out of alignment with its group average position.

OUT INHIBIT lamp, when on, indicates that control rods will not respond to out commands. Control rod out inhibits:

• Source range SUR >2 DPM and reactor power <10% and IR <109 amps.

• IR range SUR >3 DPM and reactor power <10%.

• Loss of any safety group (14) out limit and reactor power >40%.

• Any rod group asymmetric fault (any rod >6.5% from group average) and reactor power >40%.

SEQUENCE INHIBIT lamp, when on, indicates that regulating groups cannot be withdrawn in sequence. A sequence monitor provides control input for this indication. The lamp will come on if regulating groups are operated in any of the following conditions.

• Group 5 less than 80% and Group 6 greater than 5%.

• Group 5 less than 95% and Group 6 greater than 20%.

• Group 6 less than 80% and Group 7 greater than 5%.

• Group 6 less than 95% and Group 7 greater than 20%.

• Group 5 less than 95% and Group 7 greater than 5%.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0299 Rev: 0 Rev Date: 9-5-99 Source: Direct Originator: J Haynes TUOI: ANO-1-LP-RO-Nl Objective: 10 Point Value: I Section: 3.7 Type: Instrumentation System Number: 015 System

Title:

Nuclear Instrumentation System

==

Description:==

Knowledge of the effect that loss or malfunction of the NIS will have on the following: ICS KIA Number: K3.04 CFR

Reference:

41.7 / 45.6 Tier: 2 RO Imp: 3.4 RO Select: Yes Difficulty: 3 Group: 2 SRO Imp: 4.0 SRO Select: Yes Taxonomy: An Question: RO: 60 SRO:I Given:

- The plant is at 80% power.

- The NI SASS mismatch alarm is bypassed due to a mismatch.

What would be the predicted plant response if Nl-6 failed to 125%?

A. Control rods move inward, feedwater flows go up.

B. Control rods move inward, feedwater flows go down.

C. Control rods move outward, feedwater flows go up.

D. Control rods move outward, feedwater flow go down.

Answer:

A. Control rods move inward, feedwater flows go up.

Notes:

The mismatch alarm disables the SASS module automatic operation. When NI-6 fails to 125% power, ICS will see NI-6 as the input power. ICS will generate an error to drive rods in. At the same time a cross-limit is generated to keep feedwater balanced with reactor power. Feedwater will go up. Therefore, B, C, and D are incorrect.

References:

STM 1-64, Integrated Control System, rev 10, page 33, step 2.6.1, page 43, step 2.7 History:

Used in 1999 exam.

Direct from ExamBank, QID# 3723 Selected for 2002 RO exam.

Used on 2004 SRO/SRO Exam.

Selected for 2010 RO/SROexam

Integrated Control System STM 1-64 Rev. 10 FIGURE 6422 TOTAL FEEOWATER FLOW vs POW ER F LOW US :55 S 15 Feedwater demand is then modified by feedwater temperature error which is developed by comparing measured feedwater temperature to a feedwater temperature program developed from the demand signal. (Refer to figure 64.23) This characterization of the feedwater demand signal is required to compensate for the change in Btu input from feedwater. If feedwater temperature is low, then feedwater flow should be decreased. The primary purpose of this fhnction is to maintain a constant Btu/lb of steam for large temperature errors such as might be experienced on a bypass of feedwater heaters.

fIbURS 4 2 2 .1)5 1 2.6.1 Cross Limiting One requirement for proper steam production is that the feedwater flow/neutron power ratio must never exceed predetermined limits. Whenever the feedwater control is on automatic, a set of limits is imposed on the feedwater demand to maintain feedwater flow within 5% of the neutron power. The cross limit of feedwater is taken from neutron error in the reactor control subsystem. Greater than a

+/-5% neutron error will modify the feedwater demand signal. If you assume the feedwater demand and the reactor demand signals are 34

Integrated Control System STM 1-64 Rev. 10 together then if power is less than demand by more than 5%, the amount of error greater than 5% will decrease feedwater by that amount. For example, the demand has increased, the reactor is not responding, thus hold back the feedwater demand in order to keep the reactor and feedwater within 5% of each other. Power greater than demand by more than 5%, will increase feedwater demand.

If either limiting action on feedwater does occur, Feedwater is Reactor Limited annunciator will alarm and the ICS will be transferred into the Tracking mode. The occurrence of this limiting action indicates that the neutron power is not able to satisfy its demand. Therefore, by modifying the feedwater demand signal with the neutron error, feedwater is held to within 5% of reactor power.

Since the ICS is in Track, the turbine merely controls header pressure and thus the load can be no greater nor less than 5% of the neutron power.

2.6.2 Load Ratio (ATC) The total feedwater flow demand signal is split by the ICS into Control loop A and B feedwater demand signals by adjustment of the value of a multiplier controller. This controller sets the value of loop A feedwater demand by multiplying the total flow demand by the value of the multiplier. If the multiplier is set at .5, half of the total feedwater flow demand signal becomes loop A feedwater demand.

The loop B feedwater demand is determined by subtracting the loop A demand from the total demand. Changing the multiplier value will change the value of both loop demand signals. The maximum loop feedwater demand signal is 6 x 106 pounds mass per hour.

The value of the multiplier is set by the value of a control signal.

This signal is the algebraic summation of two other signals. One of these signals is the RCS flow mismatch signal and will be zero when all four RCPs are properly operating. This signal will be described under Three Pump Operations. The other signal is the zTc correction signal.

The control of the ratio of feedwater to each OTSG will determine the amount of heat that will be removed from the primary water in the reactor coolant system (RCS) and the relative amount of loading that each OTSG will carry. Therefore, the loading of the OTSGs can be indicated by the relative RCS return temperatures to the reactor (Tcs). If the difference in the Tcs (tT) is controlled near zero, then each OTSG will be loaded properly for the RCS flow through it. A trip of one RCP would give an immediate re-ratioing. An important benefit of keeping ATc low is that quadrant tilts within the reactor may be kept to a minimum.

The actual z.Tc is compared to the ATe setpoint. The difference (ATe Error) is used to generate the ATe correction signal. A zero ATe correction signal will split the signal equally between the ioops.

The operator may choose to manually control the ATc correction signal by placing the Load Ratio Hand/Automatic Station in hand.

The only difference between this station and the other feedwater hand/auto stations is the additional dial and knob located under the 35

Integrated Control System STM 1-64 Rev. 10 both feedwater loop demands. When both OTSGs are above low level limit, the operator may place the ICS in auto. The total flow circuit will then be blocked until low level limit is reached while operating on 3 RCPs during a plant shut down or load reduction.

2.7 Reactor Demand Refer to figure 64.28.

Subsystem The megawatt demand signal that is received by the Reactor Demand Subsystem from the Integrated Master Subsystem has a Low Limit of 15%. It is undesirable to lower reactor power to less than 15% in automatic Therefore, the demand being sent to the reactor demand calculator will not be allowed to go below a value that is translated to 15% by the reactor demand calculator, shouldbe about .15 x 902 or-435 megawatts. However, procedurally the operator will take manual control of the reactor when reduction of reactor power to < 20% is desired.

FIGURL 64.28. RLAC1 OR CONTROL Tsve REACTOR ULD - DEMO

.1......

CALCULATOR Sp CONTROL CROSS LIMITS j FEEDWATER TO REACTOR L)cj HI LIMITS REACTOR DEMAND CROSS LIMITS REACTOR TOFDT j

]

V 4 DIAMOND CRD Since reactor power is not linear with generated megawatts, the reactor demand calculator changes the magawatt demand signal to a reactor demand signal equivalent to 0-125% power. The calculator output span then is 15% to 125% taking into account the low limit on the input.

44

Integrated Control System STM 1-64 Rev. 10 The reactor demand signal is then modified as needed to keep Tave equal to setpoint. Tave is compared to the Tave setpoint which is controlled by the operator at the reactor demand WA station. A 0% to 100% selection is possible. The 0% is equal to 520°F and 100% is equal to 620°F. Therefore, 59% (579°F) is the normal setpoint. If a Tave error exists it is used in both a proportional and integral action to adjust reactor demand.

The adjusted reactor demand signal is limited to between 10%

and 103%. The low limit of 10% is there to allow Tave correction to decrease power a maximum of 5% when trying to establish Tav at setpoint. This could occur if low level limits are set too low. The high limit of 103% allows a Tave correction of 3% when reactor demand is 100%. However, the main purpose of the 103% limit is to prevent an automatic signal from raising power to its RPS trip setpoint.

The adjusted and limited reactor power demand signal is compared to the high auctioneered reactor power signal from the reactor protection system. The difference between the two signals is termed Neutron Error. If actual power is greater than reactor demand, a positive neutron error results. If neutron error is > +1%,

the control rods move into the core to reduce power until neutron error becomes <+.975%. If actual power is less than reactor demand, a negative neutron error results. If neutron error is > -1%, the control rods move out of the core to increase reactor power until neutron error becomes < -.975%.

2.7.1 Cross limits The purpose of crosslimits is to keep the heat production (the reactor) and the heat removal (feedwater) within 5% of each other. In accomplishing this purpose, ICS assumes that reactor demand and feedwater demand are matched. Therefore, if actual reactor power is out from demanded reactor power, it is also out from demanded feedwater flow.

The first of the two crosslimits concerns reactor power which was discussed earlier but will be repeated here. If reactor power is > +/-5% out from reactordemand,thenitisoutfromfeedwaterdemandby>+/-5%.

If actual feedwater flow is equal to its demand, then actual reactor power is > +/- 5% mismatched to feedwater flow. To correct this problem, the amount of mismatch greater than +/- 5% is calculated and sent to adjust total feedwater demand by that amount. An alann Feedwater is Reactor Limited is given. This means that the feedwater demand is being limited by the reactor mismatch (neutron error).

The second crosslimit has to do with feedwater flow. We could have a crosslimit setup identical to the one for the reactor. However, this could put us in the condition of having rods being pulled to raise reactor power when a feedwater flow mismatch occurred, this was determined to be undesirable.

If total feedwater demand is 5% greater than total feedwater flow, then the excess above 5% is used to correct (lower) reactor demand. The basis for this crosslimit is that, if for some reason feedwater flow is not 45

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0077 Rev: 0 Rev Date: 9/29/98 Source: Direct Originator: JCork TUOI: ANO-1-LP-RO-NNI Objective: 5 Point Value: 1 Section: 3.7 Type: Instrumentation System Number: 016 System

Title:

Non-Nuclear Instrumentation System (NNIS)

==

Description:==

Knowledge of the operational implications of the following concepts as they apply to the NNIS:

Separation of control and protection circuits.

K/A Number: K5.01 CFR

Reference:

41.5 / 45.7 Tier: 2 RO Imp: 2.7 RO Select: Yes Difficulty: 3 Group: 2 SRO Imp: 2.8 SRO Select: Yes Taxonomy: C Question: RO:j 1* SRO:J Given:

- Loop A RCS flow 70 E6 Ibm/hr

- Loop B RCS flow 63 E6 Ibm/hr

- Loop A Tave 578°F

- Loop B Tave 580°F

- Unit Tave 579°F Which Tave will be selected by the SASS Auto/manual transfer switch and why?

a. Unit Tave due to Loop B flow

b. Loop A Tave due to Loop B flow

c. Loop B Tave due to Loop B flow

d. Unit Tave, flows are within tolerances Answer:

b. Loop A Tave due to Loop B flow Notes:

SASS will automatically select the Loop Tave for the Loop with the highest RCS flow should either flow drop below 95%. Normal RCS loop flow is -70 E6 lbm/hr, therefore Loop B flow is <95% and SASS will select Loop A flow for Tave control, this control function protects the core from excessive heat transfer based upon flux to flow, therefore, (b) is the only correct response.

References:

STM 1-69 (Rev 5), Non-Nuclear Instrumentation System page 12 step 3.3.5 History:

Modified QID 2517 for 1998 RO/SRO Exam.

Used in A. Morris 98 RO Re-exam Selected for 2002 RQISRQ exam.

Selected for 2010 RO/SRO exam

Non-Nuclear Instrumentation System STMI-69 Rev. 13 3.3.4 Average Th and Tc The SASS selected loop A and B hot leg temperatures are averaged by an average amplifier. A selector switch, located on C03, allows selecting either the loop A hot leg temperature, the loop B hot leg temperature, or the average hot leg temperature.

LOOP A Thot LOOP ATc AVERAG AMP AVERAG AMP LOOP B Tc x x IC LOOP B Thot Normally the average hot leg temperature is selected. The selected temperature inputs into ICS for calculation of OTSG BTU limits. The selected temperature is also displayed on the Th t

0 temperature recorder located on C-i 3.

The SASS selected loop A and B cold leg temperatures are also averaged by an average amplifier. This average cold leg temperature is used for calculation of Unit RCS average temperature and Unit RCS differential temperature.

3.3.5 RCS Average RCS loop A, loop B, and Unit Tave indications are Temperature calculated. The loop average temperatures are displayed on a 520 °F to 620 °F meter which is located on C03 (TI-lo2orri-1o43). Unit Tave is displayed on a recorder located on C-13.

Average amplifiers average the SASS selected Th and Tc signals. RCS loop A average temperature is calculated by the NNIX channel. The NNIY channel calculates RCS loop B average temperature. Unit Tave is calculated from the average Th and Tc (loop A and ioop B Th and Tc are averaged) signals.

SEL. LOOl> A T, 1

TI 1020 Cs AVERAGE 1I 1032 SEL. LOOP B

• 1i SEL. LOOP B T 1043 12

Non-Nuclear Instrumentation System STMI-69 Rev. 13 Loop A average temperature, loop B average temperature, and Unit average temperature provide input to an auto/manual selector switch. The Tave selector switch output supplies the digital Tave indicator on C03 and ICS for temperature control. Unit Tave, loop A Tave, or loop B Tave may be selected by depressing the appropriate button. The selected average temperature will be backlighted. The Tave selector switch selects one of the inputs based on RCS flow.

Normally Unit Tave is selected for display and control. Should either RCS loop flow drop below 95%, the opposite loop Tave is selected for output. For instance, if RCS loop A flow is less than 95% then loop B Tave is selected. In this case, the operator will not be able to select any other average temperature. If both loop flows are less than 95%, then any of the inputs may be selected.

3.3.6 Differential Loop A, loop B, and Unit differential temperatures are calculated Temperatures from the hot leg and cold leg temperature inputs. The loop A SASS selected cold leg and hot leg temperatures are supplied to a difference amplifier and the loop B SASS selected cold leg and hot leg temperatures are supplied to a difference amplifier. The difference amplifiers subtract the cold leg temperature from the hot leg temperature. The resulting output is displayed on the loop A/B differential meter located on C-i 3. The range of indication is 0 °F to 70°F. The average hot leg and average cold leg temperatures (described above) also supply inputs a difference amplifier. The difference amplifier supplies the Unit differential temperature indicator located on C-i 3.

SEL. LOOP A El TDJ M1 SELLOOPA TC1 L1 1019 AVERAGE TDJ 1024 AVERAGE T I TDI 1048 The SASS selected loop A and B cold leg temperatures supply a difference amplifier. The difference amplifier subtracts the loop B cold leg temperature from the loop A cold leg temperature. The resulting differential temperature is displayed on the cold leg differential located on C-i 3. The range of the indicator is -10 °F to

+10 °F. The cold leg differential temperature also inputs into the ics system. The input is used to re-ratio feedwater to the OTSGs in order to maintain the cold leg temperatures equal.

SEL. LOOP A T

__IJ1 SEL. LOOP B T 1033 13

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0240 Rev: 0 Rev Date: 8-17-99 Source: Direct Originator: Don Slusher TUOI: ANO-1-LP-RO-NNI Objective: 25 Point Value: I Section: 3.7 Type: Instrumentation System Number: 017 System

Title:

In-Core Temperature Monitor (ITM) System

==

Description:==

Knowledge of the effect of a loss or malfunction of the following ITM system components:

Sensors and detectors.

K/A Number: K6.01 CFR

Reference:

CFR: 41.7/45.7 Tier: 2 RO Imp: 2.7 RO Select: Yes Difficulty: 2 Group: 2 SRO imp: 3.0 SRO Select: Yes Taxonomy: C Question: RO:I SRO:J Given:

- Plant is at 100% power

- All CET5 indicate 602 °F ICC train B Core Exit ThermocoupJe TE-l 152 fails to 900 °F.

What is the effect of this failure?

A. Core Exit Thermocouple TE-1 152 will be removed from the average.

B. ICC Core Exit Thermocouple indication will go to -627 °F.

C. TRAIN B SUBCLG MARG LO annunciator will alarm.

D. Ba SPDS will switch from ATOG to the ICC display.

Answer:

A. Core Exit Thermocouple TE-1 152 will be removed from the average.

Notes:

CETs are averaged together to generate alarms, indication, or action. Therefore, b, c, and d are incorrect and a is correct since ICCMDS will determine that TE-1152 is unreliable and remove it from the average.

References:

1105.008 Rev 17 History:

Developed for 1999 exam.

Used on 2004 RO/SRO Exam.

Selected for 2010 RO/SRO exam

I PROC JWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 3 of 39 1105.008 INADEQUATE CORE COOLING MONITOR AND DISPLAY CHANGE: 017

• NARR Range Hotleg LVL LT-1l89 LT- 1190 LT-l191 LT-ll92 LT-l193 LT-ll94 LT -1195 LT-1l96

• RCS Pressure PT-1042 PT-1041

• Reactor Coolant Pump Contacts 3.1.1 Reactor Vessel Level Sensors Two level probes, each having nine level sensors and an absolute thermocouple near the top, are installed in the reactor vessel through the head at the center CRDM location (center CRD no longer used) . Level is sensed at approximately 2 intervals from the top of the dome to near the top of the fuel assemblies.

A level sensor consists of two thermocouples connected internally to provide a signal proportional to the temperature difference. One thermocouple is heated by an internal heater element in the probe. The area around the heated thermocouple has a different heat transfer coefficient to the surrounding RCS and, therefore, has a different sensitivity to water or steam. As the water level drops below the level sensor, its zT changes and provides wet or dry indication.

The absolute thermocouple provides head fluid temperature indication from near the top of the head.

TS 3.3.15 includes the reactor vessel level sensors (RVLMS).

3.1.2 Core Exit Thermocouples Twenty-four qualified core exit thermocouples provide temperature indication in a range of 50°F to 2300°F. These instruments are part of the incore detector system and are installed through the bottom of the reactor vessel through the incore instrument guide tubes. All valid CETs are averaged, and each CET is compared to the average. If a significant deviation exists, the CET is flagged SUSPECT.

Failed or suspect CETs are automatically excluded from the average. TS 3.3.15 includes the core exit thermocouples.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0138 Rev: 0 Rev Date: 12/02/98 Source: Direct Originator: B. Short TUOI: AA51002-013 Objective: 9 Point Value: 1 Section: 3.4 Type: RCS Heat Removal System Number: 045 System

Title:

Main Turbine Generator System

==

Description:==

Ability to manually operate and/or monitor in the control room: Turbine stop valves.

KIA Number: A4.06 CFR

Reference:

41.7 / 45.5 to 45.8 Tier: 2 RO Imp: 2.8 RO Select: Yes Difficulty: 3 Group: 2 SRO Imp: 2.7 SRO Select: Yes Taxonomy: K Question: RO: 3 SRO:F63 During the performance of Main Turbine Governor Valve testing, while governor valve #1 was closed in the test position governor valve #3 fails closed. What turbine problems does this impose?

A. Moisture impingement on the turbine blading.

B. Thermal shock to the turbine rotor.

C. Turbine will trip due to low load.

D. Turbine overspeed condition.

Answer:

B. Thermal shock to the turbine rotor.

Notes:

(A) is incorrect. The closure of both valves does not change the quality of the steam.

(B) is correct. Closure of GVI and GV3 with GV2 & GV4 open or closure of GV2 & GV4 with GVI & GV3 open causes thermal shock on the turbine rotor.

(C) is incorrect. The load shifts through the two valves that remain open.

(D) is incorrect. The load will stay essentially the same so that an overspeed condition should not occur.

References:

1106.009 (Change 37)

History:

Developed for use in A. Morris 98 RO Re-exam Selected for 2010 RO/SRO exam

I PROC IWORK PLAN NO. PROCEDUREIWORK PLAN TITLE:

TURBINE STARTUP (WARMUP & ROLL)

PAGE: 56 of 181 1106.009 CHANGE: 037 CAUTION Thermal shock can damage turbine rotor if either of the following occurs:

- Simultaneous closing of GV-l and GV-3 with GV-2 and GV-4 both open.

- Simultaneous closing of GV-2 and GV-4 with GV-1 and GV-3 both open.

• Under no circumstance should more than one Governor Valve be operated at the same time.

12.3.3 Continuously monitor Governor Valve positions.

• Do NOT allow GV-l and GV-3 to be closed simultaneously with GV-2 and GV-4 both open.

• Do NOT allow GV-2 and GV-4 to be closed simultaneously with GV-l and GV-3 both open.

12.3.4 IF plant response becomes erratic when closing or opening Governor Valves, THEN release the pushbutton, allow plant to stabilize and then continue.

CAUTION Governor Valve operation with turbine controls not in ICS auto may cause large load changes.

NOTE

• Both GV CLOSE and GV OPEN pushbuttons will be backlit after GV CLOSE pushbutton is depressed.

• Pushbuttons will be backlit even if governor valve is already closed.

12.4 Slowly close the Governor Valve associated with the servo being repaired/replaced by depressing the GV CLOSE pushbutton.

• GV-l A Governor Valve

• GV-2 B Governor Valve

• GV-3 C Governor Valve

• GV-4 D Governor Valve 12.4.1 WHEN Governor Valve is closed ND associated pushbutton backlights are on, THEN release GV CLOSE pushbutton.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0798 Rev: 0 Rev Date: 9/16/2009 Source: New Originator: S. Pullin TUOI: A1LP-RO-MSSS Objective: 4 Point Value: 1 Section: 3.8 Type: Plant Services System System Number: 075 System

Title:

Circulating Water System

==

Description:==

Knowledge of abnormal condition procedures.

KIA Number: 2.4.11 CFR

Reference:

41.10 / 43.5 / 45.13 Tier: 2 RO Imp: 4.0 RO Select: Yes Difficulty: 2 Group: 2 SRO Imp: 4.2 SRO Select: Yes Taxonomy: C Question: RO: 64 SRO:F Given:

- Plant at 100% power

- Lake Temperature is 65 F

- P-3A, P-3B, and P-3C Circulating Water Pumps are running

- P-3A Circulating Water Pump trips.

- P-3D Circulating Water Pump standby pump was started.

- It is noticed that the condenser waterbox discharge temperature is 10 degrees higher and condenser vacuum is dropping.

- AOP 1203.016, Loss of Condenser Vacuum, has been entered.

Which of the following is the cause for these conditions?

A. The stopping and starting of a circ pump caused fouling to be removed from the tube sheet promoting better heat transfer capabilities.

B. The discharge valve on the tripped pump did not go completely closed and circulating water is short cycling.

C. The debris on the bar grates of the circulating water bays was stirred up during the circ pump swap causing reduced flow.

D. Lake temperature is too high for 3 circulating water pump operation per 1104.008, Circulating Water and Water Box Vacuum System Operation.

Answer:

B. The discharge valve on the tripped pump did not go completely closed and circulating water is short cycling.

Notes:

(A.) is incorrect. Although some fouling can be removed during pump rotations, it should not result in a 10 degree change in waterbox discharge temperature.

(B.) is correct. The discharge valve on an idle pump can allow a significant amount of backflow from the operating pumps if it is not closed completely.

(C.) is incorrect. This condition is normal for a circ pump swap and may contribute to waterbox fouling, however, the service water system would be affected by this condition as well.

(D.) is incorrect. 1104.008 states that 4 CW Pumps are needed when lake temperature is above 67 F

References:

1104.008, Circulating Water System, change 27, pagel 3, Caution

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I-History:

New for 2010 RO/SRO exam

PROCJWORK PLAN NO. I I PROCEDUREIWORK PLAN TITLE:

CIRCULATING WATER AND WATER BOX VACUUM SYSTEM I PAGE: 13 of 72 II OPERATION CHANGE: 027 1104.008 CAUTION

• Stopping CW pump during radwaste release could result in Tech Spec violation for exceeding MPC requirements at site boundary.

• Stopping CW pump during chemical release (NT dump or biocide injection) could result in violation of NPDES requirements.

• Failure of pump discharge CV to close upon stopping of pump will result in short cycling of circ water back to lake which can cause pump reverse rotation and lowering of condenser vacuum.

• Debris in circ water bay will become stirred when Circ Water Pump is stopped. A greater potential of Service Water pump strainer fouling exists when stopping P-3B OR P-3C.

8.4 Circ Water Pump Stop 8.4.1 IF normal pump rotation, THEN verify no radioactive releases in progress on either unit.

A. IF a radioactive release in progress, THEN verify either of the following is performed:

• Do NOT continue until the release is complete, or

• Terminate the release.

8.4.2 IF normal pump rotation AND trench release in progress per Turbine Building Draining System (1104.044), Turbine Building Trench Continuous Release, THEN verify Chemistry notified of change in Circ Water flow configuration.

8.4.3 IF normal pump rotation, THEN verify no radioactive release permits have been submitted to Chemistry from either unit.

A. IF radioactive release permit has been submitted, THEN verify either of the following is performed:

• Release permit is cancelled, or

• Release calculations are re-performed for new estimated dilution flow rate.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0542 Rev: 0 Rev Date: 12/8/2003 Source: Direct Originator: NRC TUOI: Objective: Point Value: I Section: 3.8 Type: Plant Service Systems System Number: 086 System

Title:

Fire Protection System

==

Description:==

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the Fire Protection System controls including: Fire header pressure.

KIA Number: Al .01 CFR

Reference:

41.5/45.5 Tier: 2 RO Imp: 2.9 RO Select: Yes Difficulty: 2 Group: 2 SRO Imp: 3.3 SRO Select: Yes Taxonomy: K Question: RO:] SRO:I 65 You are on watch in the Control Room when the following annunciator alarms:

- K12-A1, FIRE As Fire Water Header pressure drops from 110 psig to 80 psig select the order that fire pumps would start.

A. Jockey FWP P-Il; Diesel Fire Pump P-6B starts second; Electric Fire Pump P-6A starts last.

B. Electric Fire Pump P-6A; Diesel Fire Pump P-6B starts second; Jockey FWP P-il starts last.

C. Electric Fire Pump P-6A; Jockey FWP P-Il starts second; Diesel Fire Pump P-6B starts last.

D. Jockey FWP P-li; Electric Fire Pump P-6A starts second; Diesel Fire Pump P-6B starts last.

Answer:

D. Jockey FWP P-l 1; Electric Fire Pump P-6A starts second; Diesel Fire Pump P-6B starts last.

Notes:

D is correct, Jockey FWP P-Il; Electric Fire Pump P-6A starts second; Diesel Fire Pump P-6B starts last.

The other choices are incorrect based on pressure to start for each one.

References:

STM 1-60, Fire Protection System, rev 8, page 2.

History:

Developed by NRC.

Used on 2004 RO Exam.

Selected for 2010 RO/SRO exam

Fire Protection System STM 1-60 Rev. 8 The Pre-Fire Plans are used by operations and the fire brigade.

They contain detailed information for each fire zone. Copies are located in both control rooms and the fire locker on 386 elevation (for fire brigade use).

The Fire Brigade provides the trained on-site personnel required to attack and control a fire. Three fire brigade members and two support personnel are on-site at all times. The fire brigade has been trained and drilled in fire fighting strategy. Abnormal operating procedures and the Emergency Plan provide for contacting the Russeilville Fire Department and other agencies should the magnitude of the fire exceed on-site fire fighting capabilities.

2.0 COMPONENT DESCRIPTION 2.1 Fire Water System (Refer to Figure 60.1)

Three fire water pumps in the intake structure take their suction from the Service Water bays. The motor-driven jockey pump (P1 1) is a small displacement pump whose function is to makeup for small leaks and thereby maintain fire main pressure. This keeps the high capacity pumps P6A&B from starting inadvertently, reducing wear on the larger pumps and thus increasing their reliability. An actuation of a sprinkler system will cause header pressure to drop significantly, starting the electric motor driven fire pump (P6A) first and, if pressure drops further, the diesel-driven fire pump (P6B) next.

The fire water pumps discharge to the fire water loop, a 12-inch main that surrounds both units and is buried below the frost line for freeze protection. A loop configuration with isolation valves is used so a broken pipe in the loop can be isolated without isolating all of the fire header downstream of the break.

There are other fire water ioops tapping off the yard fire main.

These loops supply water to the Turbine, Auxiliary, Administration, and Reactor Buildings, and to the Switchyard and Transformer areas.

All fire water loops have hose reels spaced at a maximum of 100 ft. apart. Each hose reel has fifty feet of 1 /2 inch hose available for use. Hydrants are also provided at various locations and are spaced approximately 250 ft apart.

2.1.1 Fire Water Pumps The fire water pumps, P6A&B, are vertical mounted, 3 stage, centrifugal pumps rated at 2500 gpm at a discharge pressure of 150 psig. The pumps are identical while the drivers are diverse to ensure a high volume fire water source will be available for all plant conditions. Both pumps use identical relief type recirc valves which are set at 150 psig. Flow through the relief can be checked by a sight glass. The pumps are located in the intake structure and take suction on the Service Water bays.

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INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0482 Rev: 0 Rev Date: 10/7/2003 Source: Direct Originator: J.Cork TUOI: AILP-WCO-CZ Objective: 13 PointValue: 1 Section: 3.9 Type: Radioactivity Release System Number: 068 System

Title:

Liquid Radwaste System (LRS)

==

Description:==

Ability to perform specific system and integrated plant procedures during all modes of plant operation.

K/A Number: 2.1.23 CFR

Reference:

41.10 / 43.5 /45.2 / 45.6 Tier: 3 RO Imp: 4.3 RO Select: Yes Difficulty: 2 Group: SRO Imp: 4.4 SRO Select: Yes Taxonomy: K Question: RO:J SRO:

Which of the following must be performed to release T-16A contents with the Liquid Radwaste Process Monitor (Rl-4642) inoperable?

A. Estimate radiation level every four hours during the release.

B. Have an independent sample obtained and analyzed prior to release.

C. Estimate flow rate at least once every three hours during release.

D. T-16A can NOT be released if Rl-4642 is inoperable.

Answer:

B. Have an independent sample obtained and analyzed prior to release.

Notes:

Answer B contains the requirement from Att. BI of 1104.020. The other answers are incorrect.

2004 Exam Development Note: Randomly selected alternate K/A 2.1.23 to replace 2.1.31 due to lack of CR controls at ANO for the Liquid Radwaste system.

References:

1104.020, Change 49, Att. BI, section 2 History:

Modified regular exambank QJD #2765.

Used on 2004 RO/SRO Exam.

Selected for 2010 RO/SRO

i PROC.IWORK PLAN NO.

1104.020 I PROCEDUREIWORK PLAN TITLE:

I PAGE: 110 of 145 I CLEAN WASTE SYSTEM OPERATION CHANGE: 049 ATTACHMENT Bi Page 2 of 9 1.5 Record the following:

1.5.1 Number of CW Pumps running AND CW pump Disch Press psig 1.6 IF adjustments are made to CW flow, THEN terminate release.

1.7 Submitted to Chemistry for Analysis, Section 2.0.

Date Time Section 1.0 Performed By 2.0 ANALYSIS (Chemistry) 2.1 Sample Tank T16A for release analysis using Sampling Treated Waste Monitor Tank (T-16A/B) (1607.009)

Date/Time /__________

2.2 IF Liquid Radwaste Process Monitor (RI4642) is inoperable OR unavailable as identified in either Request, or Verification of PreRelease Requirements sections of this permit, THEN obtain independent sample of tank contents for analysis.

Date/Time

/___________

2.3 Record selected tank pH 2.4 Review gamma spectroscopy report and Tritium analysis.

2.5 IF release is radioactive AND release desired, THEN generate Preliminary Release Report.

2.6 Check sample results indicate that release of total tank contents will not violate ANO radioactive effluent discharge limit.

2.7 IF Liquid Radwaste Process Monitor (RI4642) is inoperable OR unavailable as identified in either Request, or Verification of Pre-Release Requirements section of this permit, THEN perform independent analysis of computer data input.

Date/Time

/__________

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0800 Rev: 0 Rev Date: 9/16/2009 Source: New Originator: S Pullin TUOI: Al LP-RO-ESAS Objective: 20 Point Value: 1 Section: 2 Type: Generic KM System Number: 2.1 System

Title:

Conduct of Operations

==

Description:==

Ability to locate control room switches, controls, and indications, and to determine they correctly reflect the desired plant lineup.

K/A Number: 2.1.31 CFR

Reference:

41.10 / 45.12 Tier: 3 RO Imp: 4.6 RO Select: Yes Difficulty: 3 Group: SRO Imp: 4.3 SRO Select: Yes Taxonomy: C Question: RO:J 67 SRO:1 67 Given:

LOCA in progress has caused ESAS actuation of Channel 1-4 Which of the following combinations of indications and locations are correct for the given condition?

A. CV-3820, SW TO ICW, green light, on C16; CV-1 270, RCP SEAL BLEEDOFF FROM D RCP, red light, on C18; CV-1 053, QUENCH TANK DRAIN, green light, on C16 B. CV-1233, RCS MAKEUP, red light, on C16; CV-1441, BWST PURIF RECIRC ISOL, green light, on C13; CV-5612 ,FIRE WATER TO RB, green light, on C18.

C. CV-1285, HIGH PRESSURE INJECTION, red light, on C16; CV-1407, BWST OUTLET, red light, on C18; CV-3841, LPJ PUMP BRG CLR E-50 INLET, red light, on C16 D. CV-1408, BWST OUTLET, red light, on C18; CV-7402, RB PURGE INLET, green light, on C18; CV-4804, RB VENT, red light, on C16 Answer:

C. CV-1285, HIGH PRESSURE INJECTION, red light, on C16; CV-1407, BWST OUTLET, red light, on C18; CV-3841, LPI PUMP BRG CLR E-50 INLET, red light, on C16 Notes:

C is correct in that it has the correct indications and panel locations.

A is incorrect in that it has the incorrect indications and correct panel locations.

B is incorrect in that it has the correct indications and incorrect panel locations.

D is incorrect in that it has the incorrect indications and incorrect panel locations.

References:

STM 1-65 Rev 5 ESAS STM 1-05 Rev 16 DHR History:

New selected for 2010 ROISRO exam

Engineered Safeguards Actuation System STM 1-65 Rev 5 4.12

SUMMARY

OF ACTUATION OF ENGINEERED SAFEGUARDS BY ESAS 4.12.1 High Pressure Upon 1590 psig RCS or 4 psig RB pressure, HPI and diverse Injection and containment isolation is actuated. (Channels 1 and 2).

Diverse Containment

• High Pressure Injection Pumps start with a design pressure Isolation and flow of 3000 psig and 300 gpm. The auxiliary oil pumps will run for only approximately 20 seconds after an ES actuation to minimize oil system over-pressurization and leaking out of the oil.

20. IdentiYi all devices actuated

• Two Diesel Generators start and come up to rated speed by ESAS to include post actuation (900 RPM). If needed can supply 2.7 MWe to each 4160 condition or condition .

ES bus. Electncal buses align to ensure separation of vital buses.

• HPI Block Valves open, CV-1228, 1227, 1219, 1284, 1285, 1278, 1279 and 1220, to supply the RCS with water.

• The Letdown Coolers are isolated by the closing of CV-1214, 1216 Letdown Cooler Isolation valves and CV-1221 Letdown Isolation.

• MU Pump Recirc Valves, CV-1301 and 1300 close to allow full flow to the RCS.

• The BWST Outlet Valves, CV-1407 and 1408, open to supply the MU Pumps.

• BWST Recirc Isolation Valves CV-1441 and CV-1438 will receive a close signal from their associated BWST Isolation.

• The MIJ Block valves; CV-1234 and CV-1233 get a close signal to isolate normal MU.

• Service Water Valves, CV-3640 3642, 3644 and 3646 will either open or close to give two independent loops, CV-3643 closes to isolate the ACW System.

• CV-1270, 1271, 1272, 1273, and 1274 close to isolate the Reactor Coolant Pumps Seal Returns.

• CV-3820 and 3811 SW Supply to ICW Coolers close to isolate and give 2 independent Service Water loops.

• CV-4803 and 4804 close the RB vent.

Engineered Safeguards Actuation System -

STM 1-65 Rev 5 4.12

SUMMARY

OF ACTUATION OF ENGINEERED SAFEGUARDS BY ESAS 4.12.1 High Pressure Upon 1590 psig RCS or 4 psig RB pressure, HPI and diverse Injection and containment isolation is actuated. (Channels 1 and 2).

Diverse Containment

• High Pressure Injection Pumps start with a design pressure Isolation and flow of 3000 psig and 300 gpm. The auxiliary oil pumps will run for only approximately 20 seconds after an ES actuation to minimize oil system over-pressurization and leaking out of the oil.

20. Identify all devices actuated

• Two Diesel Generators start and come up to rated speed by ESAS to include post actuation (900 RPM). If needed can supply 2.7 MWe to each 4160 condition or condition .

ES bus. Electncal buses align to ensure separation of vital buses.

• HPI Block Valves open, CV-1228, 1227, 1219, 1284, 1285, 1278, 1279 and 1220, to supply the RCS with water.

• The Letdown Coolers are isolated by the closing of CV-1214, 1216 Letdown Cooler Isolation valves and CV-1221 Letdown Isolation.

• MU Pump Recirc Valves, CV-130l and 1300 close to allow full flow to the RCS.

• The BWST Outlet Valves, CV-1407 and 1408, open to supply the MU Pumps.

• BWST Recirc Isolation Valves CV-l441 and CV-1438 will receive a close signal from their associated BWST Isolation.

• The MU Block valves; CV-l234 and CV-l233 get a close signal to isolate normal MU.

• Service Water Valves, CV-3640 3642, 3644 and 3646 will either open or close to give two independent loops, CV-3643 closes to isolate the ACW System.

• CV-1270, 1271, 1272, 1273, and 1274 close to isolate the Reactor Coolant Pumps Seal Returns.

• CV-3820 and 3811 SW Supply to ICW Coolers close to isolate and give 2 independent Service Water loops.

• CV-4803 and 4804 close the RB vent.

Decay Heat Removal System STM 1-05 Rev.1 6 The recirc flow path ensures a flow of >80 gpm for pump protection during periods of low flow.

If Service Water flow is maintained to the cooler, run time in the recirculation mode is not restricted due to the high volume recirculation flow. No discharge isolation valves are provided. Each decay heat pump has a discharge check valve (DH-2A & 2B) on the discharge line to the DHR cooler. A motor current monitor with a variable alarm setpoint has been provided to detect vortex formation when the system is operating with reduced RCS levels in the DHR mode of operation. There are no interlocks on the DHR Pumps that would prevent a pump start with the suction or discharge valves closed. Engineered Safeguards (ES) signals will start the pumps based on RCS or Reactor Building conditions regardless of valve alignments. Pump operation with the suction lines closed can cause pump damage in a very short time. For this reason, it is very important to check valve alignments when DHRLPI Pumps are prepared for immediate or standby operation.

2.1.3.1 Pump Bearing DHR /LPI pump and motor bearings are lubricated by a slinger Lubncation ring. A loose collar is hung on the pump shaft at each bearing location. The collar or slinger ring hangs down into a bearing oil sump and slings oil onto the rotating parts in the bearing housing. Oil level is checked using Bulls eye sight glasses on the motor and vertical sight glasses at the pump bearings. Trico automatics oilier are provided at the bearings to maintain oil level in the bearing sumps.

The Automatic oilers are clear plastic reservoirs inverted on a supply pipe. Oil level is visible at all times. Experience has shown that the Trico oilers are susceptible to vapor binding and close checks of oil levels are required after extended runs of the associated pump. Level gauges at the bearing sump should be checked to verify proper sump oil levels.

To determine if slinger rings are operating properly observation ports are provided. The observation ports are located on the opposite side of the level indicator. By using a flashlight during pump operation, the slinger ring can be viewed to determine if the slinger rings are operating properly. During pump operation the slinger ring should be rotating on the shaft. If the slinger rings are not rotating on the shaft notify control room personnel at once.

2.1.3.2 Pump Cooling Pump bearing and stuffing box cooling is supplied by the Service Water System. Cooling water flows through the pump oil and stuffing box jacket coolers (E-50A & B) and then flows to the Service Water return line. Service water flow through the cooler is normally isolated when the pump is secured by an air operated control valve.

CV-3840 provides isolation for P-34A and CV-3841 for P-34B. The associated control valve receives an open signal when the pump is started. The control valves can be manually opened when maintenance is performed or when valve fails required stroke time to maintain pump operability. Indication is provided on C-i 6 & C-i 8.

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INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

History:

New for 2010 RO/SRO exam QID: 0799 Rev: 0 Rev Date: 9/16/2009 Source: New Originator: S Pullin TUOI: A1LP-RO-ICS Objective: 11 Point Value: 1 Section: 2.0 Type: Generic K&A System Number: 2.1 System

Title:

Conduct of Operations

==

Description:==

Ability to explain and apply system limits and precautions.

K/A Number: 2.1.32 CFR

Reference:

41.10/43.2/45.12 Tier: 3 RO Imp: 3.8 RO Select: Yes Difficulty: 4 Group: SRO Imp: 4.0 SRO Select: Yes Taxonomy: Ap Question: RO: 6 SRO:

Procedure 1105.004, Integrated Control system limit and precaution states do not operate Reactor Demand H/A station in Auto with both S/Gs on low level limits.

What is the reason for this precaution and does any exception apply?

A. Due T-ave reduction as power lowers rods will pull to maintain T-ave at setpoint, you can operate with Reactor Demand H/A station in Auto with both S/Gs on low level limits if you adjust T-ave setpoint to match reactor power B. Due T-ave reduction as power lowers rods will not move due to T-ave error, you can not operate with Reactor Demand H/A station in Auto with both S/Gs on low level limits C. When S/Gs are on Low Level Limits, the Tave calibrating integral is blocked,, you can operate with Reactor Demand H/A station in Auto with both STGs on low level limits providing you verify calibrating integral is blocked on PDS.

D. When S/Gs are on Low Level Limits, the Tave calibrating integral is released, you can not operate with Reactor Demand H/A station in Auto with both S/Gs on low level limits.

Answer:

A. Due T-ave reduction as power lowers rods will pull to maintain T-ave at setpoint, you can operate with Reactor Demand H/A station in Auto with both S/Gs on low level limits if you adjust T-ave setpoint to match reactor power Notes:

A is correct, due to lowering power with S/G on LLL will cause Tave to ramp down. The Rx Demand station will try to pull rods to maintain 579 F. Limit & Precaution allows this mode of operation only if you reduce Tave setpoint to match Rx power.

B, C and D are incorrect

References:

OP-I 105.004 Change 20 History:

New selected for 2010 RO/SRO exam.

i PROCJWORK PLAN NO.

I PROCEDUREIWORK PLAN TITLE:

I PAGE: 5 of 53 1105.004 INTEGRATED CONTROL SYSTEM CHANGE: 020 5.6 If it is necessary to operate either the startup or low load valve in manual, both startup and low load valves in that Loop should be placed in manual and valves operated in normal sequence.

5.6.1 Do not modulate startup valve unless low load valve is closed.

5.6.2 Do not modulate low load valve unless startup valve is full open.

5.7 Do not operate Reactor Demand H/A station in AUTO with both OTSGs on low level limits unless T-ave setpoint is adjusted to match desired reactor power.

5.8 Operation of either Startup Valve in HAND when SG pressure is 750 psig requires entry into TS 3.7.3 Condition D.

5.9 Operation of either Low Load Valve in HAND when SG pressure is 750 psig requires entry into TS 3.7.3 Condition C.

5.10 Operation of both Startup Valve and Low Load Valve in HAND when SG pressure is 750 psig requires entry into TS 3.7.3 Condition E.

5.11 Due to offset of the prongs on the light bulbs used in the ICS H/A stations, it is necessary to ensure proper alignment prior to replacing bulbs to avoid damage to the H/A station or shorting of ICS circuitry. (CRANO120042382, CRANO120042384)

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0116 Rev: 0 Rev Date: 7/14/98 Source: Direct Originator: JCork TUOI: Al LP-RO-NOP Objective: 7 Point Value: 1 Section: 2.0 Type: Generic K/As System Number: 2.2 System

Title:

Equipment Control

==

Description:==

Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity.

KIA Number: 2.2.1 CFR

Reference:

45.1 Tier: 3 RO Imp: 3.7 RO Select: Yes Difficulty: 2 Group: SRO Imp: 3.6 SRO Select: Yes Taxonomy: K Question: RO: 69 SRO:J During an INITIAL approach to criticality, if criticality is NOT achieved within of the ECC, then insert and A. Plus or minus 1.0% delta k/k control rods to achieve 1.5% SD margin establish hot shutdown conditions B. Plus or minus 1.0% delta k/k regulating groups to achieve 1.0% SD margin notify Reactor Engineering C. Plus or minus 0.5% delta k/k control rods to achieve 1.5% SD margin verify calculation D. plus or minus 0.5% delta k/k regulating groups to achieve 1.0% SD margin verify calculation Answer:

C. plus or minus 0.5% delta k/k control rods to achieve 1.5% SD margin verify calculation Notes:

Answer C is correct per 1102.008.

References:

1102.008, Chg. 023 History:

Used in 1998 RO exam Used in NRC developed RO exam 8/24/92, no. 88 Used in A. Morris 98 RO Re-exam Used in 2001 RO Exam Selected for 2010 RO/SRO exam

i PROC.IWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: I PAGE: 12 of 19 1102.008 APPROACH TO CRITICALITY CHANGE: 023 9.7 Sequentially withdraw regulating groups in 30% increments, per CRD System Operating Procedure (1105.009), Regulating Group Sequential Withdrawal section. Perform the following during rod withdrawal:

• IF unexpected situations/conditions arise, p THEN take conservative actions to place the reactor in a safe condition.

• Continuously monitor available instrumentation for doubling count rate and unplanned criticality.

• IF unexpected count rate/power rise is observed THEN immediately insert control rods to stop rise or if required trip the reactor.

• At 30% rod position increments stop rod withdrawal, p allow count rate to stabilize, and collect data for 1/rn plot.

9.8 IF criticality is achieved within procedural limits of +/- 0.5% zk/k AND NOT within + 0.25% tk/k, THEN notify Reactor Engineering to initiate a condition report AND continue this procedure. (CRANO120090237) 9.9 IF this is a startup immediately following refueling AND a rod index of 300% is within the ECC band AND criticality is NOT achieved by a rod index of 300%,

THEN inform Reactor Engineering and refer to 1302.020 for completion of the approach to criticality.

9.10 IF criticality is NOT achieved within +/-0.5% zk/k of the ECC, THEN insert control rods to obtain 1.5% subcritical conditions, and perform the following:

9.10.1 Inform Reactor Engineering.

9.10.2 Verify boron concentrations.

9.10.3 Verify ECC calculation.

9.10.4 Verify position of all control rods by comparing API to zone or limit position switches.

9.10.5 WHEN cause of ECC error is determined, AND cause corrected, THEN reperform AND re-initial applicable steps of this procedure.

9.11 Record time reactor is made critical

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0801 Rev: 0 Rev Date: 9/17/2009 Source: New Originator: S Pullin TUOI: AILP-RO-TS Objective: 7 Point Value: I Section: 2.0 Type: Generic K&A System Number: 2.2 System

Title:

Equipment Control

==

Description:==

Ability to determine operability and / or availability of safety related equipment.

KIA Number: 2.2.37 CFR

Reference:

41.7 / 43.5 / 45.12 Tier: 3 RO Imp: 3.6 RO Select: Yes Difficulty: 2 Group: SRO Imp: 4.6 SRO Select: Yes Taxonomy: Ap Question: RO:j 70 SRO: 70 REFERENCE PROVIDED Which of the following plant conditions would require entry into LCO 3.2.1 due to exceeding Regulation Rod Insertion Limits per the COLR?

A. 80% Power, 4 RCPs in service, 150 EFPD, Rod Index of 250 %

B. 70% Power, 4 RCPs in service, 300 EFPD, Rod Index of 220 %

C. 60% Power, 3 RCPs in service, 100 EFPD, Rod Index of 265 %

D. 50% Power, 3 RCPs in service, 350 EFPD, Rod Index of 255 %

Answer:

B. 70% Power, 4 RCPs in service, 300 EFPD, Rod Index of 220 %

Notes:

Per the graphs in the COLR answer (B) falls within the Operation Restricted area of the figure and would require entry into LCO 3.2.1.

A, C, and D do not require entry into LCO

References:

ANO-1 Cycle 22 COLR Figures 3-A through 4-B History:

New for 2010 RO/SRO exam

Regulating Rod Insertion Limits 3.2.1 3.2 POWER DISTRIBUTION LIMITS 3.2.1 Regulating Rod Insertion Limits LCO 3.2.1 Regulating rod groups shall be within the physical insertion, sequence, and overlap limits specified in the COLR.

---

NOTE----------------------------------.

Not required for any regulating rod repositioned to perform SR 3.1.4.2.

APPLICABILITY: MODES I and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Regulating rod groups A.1 -____

inserted in restricted Only required when operation region. THERMAL POWER is

> 20% RTP.

Perform SR 3.2.5.1. Once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> AND A.2 Restore regulating rod 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from groups to within acceptable discovery of failure to region. meet the LCO B. Required Action and B.1 Reduce THERMAL 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> associated Completion POWER to less than or Time of Condition A not equal to THERMAL met. POWER allowed by regulating rod group insertion limits.

C. Regulating rod groups C.1 Restore regulating rod 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> sequence or overlap groups to within limits.

requirements not met.

ANO-1 3.2.1-1 Amendment No. 215

ANO-1 CYCLE 22 COLR CALC-ANO1 -N E-08-00006 Figure 3-A Regulating Rod Insertion Limits for Four-Pump Operation From 0 to 200 +/- 10 EFPD (Figure is referred to by Technical Specification 3.2.1)

I 0 50 100 150 200 250 300 Rod Index, % Withdrawn 0 20 40 60 80 100 GROUP 7*

0 20 40 60 80 100 L I GROUP 6*

0 20 40 60 80 100 GROUP 5*

Technical Specification 3.5.2.5(2) requires that operating rod group overlap be 20% +/- 5%

between two sequential groups, except for physics tests.

ANO-1 Rev. 0

ANO-1 CYCLE 22 COLR CALC-ANOI -NE-08-00006 Figure 3-B Regulating Rod Insertion Limits for Four-Pump Operation From 200+/-10 EFPD to EOC (Figure is referred to by Technical Specification 3.2.1) 110.0 100.0 90.0 80.0 0

70.0 60.0 2.

CI 50.0 40.0 0 50 100 150 200 250 300 Rod Index, % Withdrawn 0 20 40 60 80 100 GROUP 7 0 20 40 60 80 100 L II I GROUP 6*

0 20 40 60 80 100 GROUP 5*

Technical Specification 3.5.2.5(2) requires that operating rod group overlap be 20% +/- 5%

between two sequential groups, except for physics tests.

ANO-1 Rev. 0

ANO-1 CYCLE 22 COLR CALC-ANOI -NE-08-00006 Figure 4-A Regulating Rod Insertion Limits for Three-Pump Operation From 0 to 200 +/- 10 EFPD (Figure is referred to by Technical Specification 3.2.1) 100.0 90.0 OPERATION IN THIS 80.0 REGION IS NOT 0 ALLOWED (105.5, 77) 70.0 OPERATION (248.5, 67) 0 RESTRICTED (248.5, 58)

SHUTDOWN MARGIN 50.0 40.0 30.0 PERMISSIBLE OPERATION 20.0 REGION 10.0 0 50 100 150 200 250 300 Rod Index, % Withdrawn 0 20 40 60 80 100 I I I I GROUP 7*

0 20 40 60 80 100 L I I I GROUP 6*

0 20 40 60 80 100 I I I I GROUP 5 Technical Specification 3.5.2.5(2) requires that operating rod group overlap be 20% +/- 5%

between two sequential groups, except for physics tests.

ANO-1 10 Rev. 0

ANO-1 CYCLE 22 COLR CALC-ANOI -NE-08-00006 Figure 4-B Regulating Rod Insertion Limits for Three-Pump Operation From 200+/-10 EFPD to EOC (Figure is referred to by Technical Specification 3.2.1)

-D 0

0 0

a 0) 0)

0 50 100 150 200 250 300 Rod Index, % Withdrawn 0 20 40 60 80 100 I I I I I I GROUP 7*

0 20 40 60 80 100 I I I I I I GROUP 6*

0 20 40 60 80 100 I I I I I I GROUP 5*

Technical Specification 3.5.2.5(2) requires that operating rod group overlap be 20% +/- 5%

between two sequential groups, except for physics tests.

ANO-1 Rev. 0

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0802 Rev: 0 Rev Date: 9/17/2009 Source: New Originator: S. Pullin TUOI: ASLP-RO-RADP Objective: 15 Point Value: 1 Section: 2.0 Type: Generic K&A System Number: 2.3 System

Title:

Radiation Control

==

Description:==

Knowledge of radiation exposure limits under normal or emergency conditions.

KIA Number: 2.3.4 CFR

Reference:

41.12 / 43.4 / 45.10 Tier: 3 RO Imp: 3.2 RO Select: Yes Difficulty: 3 Group: SRO Imp: 3.7 SRO Select: Yes Taxonomy: Ap Question: RO:I 71 SRO:

Given:

- A General Emergency has been declared on Unit 1.

- A Maintenance crew must enter a radiological area with a dose rate of 150 Rem/Hr to protect valuable property.

Which of the following is the MAXIMUM time an individual team member can stay in this area?

A. 4 minutes B. 6 minutes C. 8 minutes D. 10 minutes Answer:

A. 4 minutes Notes:

A is correct, for protecting valuable property 10 Rem is the does limit.

B, C and D exceed 10 Rem limit.

References:

OP-i 903.033 Change 01 9-01-0 History:

New for 2010 RO/SRO exam

PROC IWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 5 of 15 1903.033 PROTECTIVE ACTION GUIDELINES FOR RESCUE!REPAIR

& DAMAGE CONTROL TEAMS CHANGE: 019-01-0 Dose limit*

(rem TEDE) Activity Condition 5 All 10 Protecting valuable property Lower dose not practicable 25 Life saving or protection of Lower dose not practicable large populations

>25 Life saving or protection of Only on a voluntary basis to large populations persons fully aware of the risks involved (refer to Attachment 1 of this procedure for health risks).

• Workers performing services during emergencies should limit dose to the lens of the eye to three times the listed value and doses to any other organ (including skin and body extremities) to ten times the listed value.

6.1.4 Rescue/repair and damage control personnel shall perform their duties in the most safe and efficient manner possible. Once their operations have been completed, they shall follow self-monitoring and personnel decontamination procedures as specified by the Health Physics Supervisor.

6.2 ACTIONS NOTE

[During a Personnel Emergency the Emergency Medical Team may enter Radiologically Controlled Areas without SRDs or Alarming Dosimeters as long as an HP Technician is providing radiological instructions and is monitoring dose rates and time in the area. Prompt medical attention shall take precedence over HP procedures for a seriously injured individual.J 6.2.1 Personnel selected for the rescue/repair and damage control teams should report to the OSC (unless otherwise instructed) for their briefing.

6.2.2 The rescue/repair and damage control team leader shall function under the direction of the Shift Manager/OSC Director.

6.2.3 Immediate Actions A. IF exposure to significant radioiodine concentrations is possible, THEN refer to procedure 1903.035, Administration of Potassium Iodide for guidance.

B. Rescue/repair and damage control teams shall be briefed using Form l903.033B, OSC Team Briefing Form. This form serves as an emergency RWP and Work Order. Instructions for conducting re-entry team briefings are contained in Attachment 3.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0436 Rev: 0 Rev Date: 4/30/2002 Source: Repeat Originator: J.Cork TUOI: AILP-WCO-CZ Objective: 11 Point Value: I Section: 2.2 Type: Generic K&A System Number: 2.3 System

Title:

Radiation Control

==

Description:==

Ability to control radiation releases.

KIA Number: 2.3.11 CFR

Reference:

41.13/43.4/45.10 Tier: 3 RO Imp: 3.8 RO Select: Yes Difficulty: 2 Group: G SRO Imp: 4.3 SRO Select: Yes Taxonomy: K Question: RO:j 72 SRO: 72 The WCO is preparing to commence a liquid release on TWMT T-16A when he notices that there is no tag hanging on T-16A inlet valve CZ-47A (tank was sampled several hours ago).

What action should be taken?

A. Document discrepancy via CR, install tag on CZ-47A, and continue with the release.

B. Terminate the release, install tag on CZ-47A and submit new release permit to nuclear chemistry.

C. Install tag on CZ-47A and continue with the release.

D. Install tag on CZ-47A, inform nuclear chemistry and resample with current release permit.

Answer:

B. Terminate the release, install tag on CZ-47A and submit new release permit to nuclear chemistry.

Notes:

Per 1104.020 Chg 043-05-0 if tag is missing when preparing to perform release, then the operator shall:

Terminate release, Install tag on CZ-47A, and Submit new relesase permit to Chemistry. Therefore:

B is correct, all other answers do not contain the correct information.

References:

1104.020, Chg 043-05-0 History:

Modified regular exambank QID 2761 for 2002 RO exam. Was KA 2.1.32 for Liquid Radwaste System Selected for use on 2007 RO Exam.

Selected for 2010 RO/SRO exam

PROC.IWORK PLAN NO. PROCEDUREIWORK PLAN TITLE:

PAGE: 113 of 145 1104.020 CLEAN WASTE SYSTEM OPERATION CHANGE: 049 ATTACHMENT B1 Page 5 of 9 4.0 Release (Operations)

Unauthorized discharge to Lake Dardanelle via the flume shall be avoided.

CAUTION 4.1 Verify CZ Disch to Flume Flow (CV4642) closed.

4.2 Verify T-l6A Xfer PP (P47A) stopped.

NOTE Tag contains information to remind personnel that tank is isolated for chemistry sample.

4.3 Verify Treated Waste Monitor Tank T16A Inlet (CZ47A) closed AND tagged.

4.3.1 IF tag is missing or has been removed since tank was last sampled, THEN perform the following:

A. Terminate this release.

B. Install tag on CZ-47A.

C. Submit new release permit to Chemistry.

4.4 Verify Treated Waste Monitor Tank T16A Outlet (CZ48A) open.

4.5 Verify F560 inservice by performing the following:

4.5.1 Verify the following valves open:

• CZ74 (LRW Disch Filter F560 Inlet)

• CZ77 (LRW Disch Filter F560 Outlet) 4.5.2 Verify CZ83 (LRW Disch Filter F560 Bypass) closed.

4.6 Verify Treated Waste Discharge Valve to Header from P47B (CZ55B) closed.

4.7 Verify Treated Waste Monitor Tank T16A Recirc Inlet (CZ54A) closed.

4.8 Open Treated Waste Discharge Valve to Header from P-47A (CZ-55A).

4.9 Open Treated Waste Discharge to Circ. Water Flume (CZ-58).

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0803 Rev: 0 Rev Date: 9/17/2009 Source: Originator: S Pullin TUOI: A1LP-RO-EOP Objective: 2 Point Value: 1 Section: 2.0 Type: Generic K&A System Number: 2.4. System

Title:

Emergency procedure / plan

==

Description:==

Knowledge of EOP mitigation strategies.

KIA Number: 2.4.6 CFR

Reference:

41.10/43.5/45.13 Tier: 3 RO Imp: 3.7 RO Select: Yes Difficulty: 2 Group: G SRO Imp: 4.7 SRO Select: Yes Taxonomy: K Question: RO:I 73 SRO:

General rules of the Generic Emergency Operating Guidelines are that symptoms are treated whenever they occur based on priorities.

Which of the following transients has top priority per the GEOG?

A. Overheating

8. Overcooling C. Loss of Subcooling Margin D. Steam Generator Tube Rupture Answer:

C. Loss of Subcooling Margin Notes:

C is correct per the GEOG LOSM has top priority.

References:

Volume I GEOG Part 1, Introduction History:

New for 2010 RO/SRO exam.

A AR EVA NUMBER TECHNICAL DOCUMENT 74-1152414-10 Part I Introduction The Generic Emergency Operating Guideline (GEOG) Bases is a guideline developed from the technical bases contained in Volume 3. The GEOG is intended to demonstrate how the individual sections of the Technical Bases document (TBD) can be assembled into one overall transient mitigation guideline. It represents the vendor-preferred path relative to options included in the TBD.

The GEOG is a procedure nor should it be used as a direct model for a procedure. The development of this document did not rigorously adhere to any set of human factors principles other than to achieve consistency in the use of terms, such as IF-THEN statements (the users of this document have their own plant specific procedure writers guides to control procedure format and content). The GEOG should also not be used as a stand-alone document. All of the TBD volumes must be read and understood before implementing TBD guidance.

GEOG Structure Seven parts comprise the GEOG:

Introduction:

basic information on use.

List of acronyms and abbreviations.

Diagnosis and mitigation: covers entry, diagnosis of abnormal conditions, mitigation of transients and plant stabilization.

• Cooldown: covers cooldown under abnormal conditions of LOCA, HPI cooling, or degraded SGs.

• Repetitive tasks: covers guidance for tasks that may apply in several mitigation or cooldown sections.

• Rules: covers important guidance that always applies after the reactor is shutdown when the stated conditions exist.

• Figures: provides any figures used in the GEOG other than the section flowcharts.

DATE PAGE 12/31/2005 Vol.1, I-i Framatome ANP, In ., an AREVA and Siemens company

A AR EVA NUMBER TECHNICAL DOCUMENT 74-1152414-10 Symptoms are treated whenever they occur, and are treated in order of priority. This precludes the need for repeated steps in the guidelines to require symptom status checks. Symptom checks are specified where their occurrence is more likely or as a transfer check at the completion of a section.

Symptom priorities are, in descending order:

- Loss of SCM

- Upsets in heat transfer (lack of or excessive)

- Steam generator tube rupture ICC is not a symptom, and can only occur following a loss of SCM. The possibility of ICC conditions developing is always monitored when SCM does not exist.

• Rules (Part VI) are used for specific guidance that always applies when the stated conditions exist. This also reduces the need for repeated steps, but more importantly fosters the better response and consistency that is achievable using rule-based behavior.

• The intent of the guidelines is to proceed through the appropriate actions without undue delay and to primarily mitigate transients from the control room when possible. Except for specific hold points or loops, it is not expected that delays will be encountered due to either prolonged attempts to achieve satisfactory results from a lesser impact action or due to attempting significantly time-consuming actions from outside the control room. For example, it is expected that a feedwater pump will be tripped to terminate overfeeding a SG if initial attempts to control flow were unsuccessful, rather than repeated attempts at local valve manipulation.

Transient Mitigation Sections The transient mitigation sections are intended to provide the necessary guidance to bring the plant to a safe and stable condition following the occurrence of a symptom (i.e., abnormal transient).

Once the plant is in a safe stable configuration, the guidance routes to either an appropriate cooldown section, or back to Section ffl.A for completion of VSSV checks, or provides the option to remain in the stable configuration and await station managements decision relative to continued operation or shutdown.

The transient mitigation sections are:

DATE PAGE 12/31/2005 Vol.1, 1-3 Framatome ANP, In ., an AREVA and Siemens company

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0161 Rev: 1 Rev Date: 4/24/2002 Source: Direct Originator: J. Cork TUOI: Al LP-RO-AOP Objective: 4 Point Value: I Section: 2.0 Type: Generic K&A System Number: 2.4 System

Title:

Emergency procedure / plan

==

Description:==

Knowledge of abnormal condition procedures.

K/A Number: 2.4.11 CFR

Reference:

41.10/43.5/45.13 Tier: 3 RO Imp: 4.0 RO Select: Yes Difficulty: 3 Group: G SRO Imp: 4.2 SRO Select: Yes Taxonomy: C Question: RO: 74 SRO: I Given:

- Power escalation is in progress following a shutdown.

- Reactor power is 35%.

- Rod 6 of Group 7 drops.

Which of the following actions should be taken?

A. Insert all regulating rods in sequential mode.

B. Trip the reactor and go to Reactor Trip, 1202.001.

C. Verify plant stabilizes at 320 MWe after ICS runback.

D. Verify SDM within COLR limit within one hour.

Answer:

D. Verify SDM within COLR limit within one hour.

Notes:

[a] would only be performed if power was <2%.

[b] would not be done because only one rod dropped.

[C] power is <360 MWe so there wouldnt be any runback, the value given would require a power increase.

[dJ is the correct answer per TS.

References:

1203.003, Control Rod Drive Malfunction Action, change 023, page 12, step 4 History:

Developed for use in 98 RO Re-exam.

Used in 2001 RO/SRQ Exam.

Selected for 2002 RO/SRO exam. Revised to agree with ITS.

Selected for 2010 RO/SRO exam

CHANGE

[1203.003 CONTROL ROD DRIVE MALFUNCTION ACTION 023 PAGE 12 of 47 SECTION 2 DROPPED ROD REACTOR CRITICAL NOTE

• Technical Specification s defines an inoperable rod as follows:

Safety Rod that is NOT fully withdrawn within one hour, except during performance of rod exercise surveillance (TS 3.1.5). If the Safety Rod is declared inoperable in TS 3.1.5, then TS 3.1.4 must also be entered.

Inability to move control rod (SR 3.1.4.2) or APSR (TS 3.1.6).

Rod can not be located with API, RPI or limit lights (TS 3.1.7).

Not meeting TS 3.1.7 results in not meeting either TS 3.1.4 or 3.1.6.

• The misaligned (>6.5%) rods position is NOT to be used in the calculation of the rod group average position.

4. IF rod is declared inoperable is misaligned >6.5%,

THEN perform the following:

NOTE If the inoperable control rod is fully inserted, then it is not necessary to consider it inoperable for the purposes of shutdown margin calculations because it has inserted its negative reactivity. A control rod is considered to be inoperable if it is not free to insert into the core within the required insertion time, or does not have at least one position indicator channel operable, i.e., cannot be located. (Ref. TS 3.1.4 Bases)

• Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> AND once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter, verify 1.5% available shutdown margin per Reactivity Balance Calculation (1103.015) OR initiate boration to restore SDM to be within COLR limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

A. IF control rod is NOT fully inserted, OR the control rod can NOT be located, THEN use worksheet 4 and use the inoperable rod option (does NOT apply to APSRs).

B. IF rod is fully inserted, THEN use worksheet 4 and do NOT use the inoperable rod option.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0804 Rev: 0 Rev Date: 9/17/2009 Source: Modified Originator: S. Pullin TUOI: A1LP-RO-RBS Objective: 8 Point Value: 1 Section: 2.0 Type: Generic K&A System Number: 2.4 System

Title:

Emergency procedure I plan

==

Description:==

Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

KIA Number: 2.4.50 CFR

Reference:

41.10 / 43.5 / 45.3 Tier: 3 RO Imp: 4.2 RO Select: Yes Difficulty: 3 Group: G SRO Imp: 4.0 SRO Select: Yes Taxonomy: C Question: RO:j 75 SRO:J Given:

- Plant is in cold shutdown.

- All necessary components have been aligned per 1305.006, Integrated ES System Test.

- All ES EVEN Digital Channels actuated per procedure using RB pressure transmitters.

- Annunciator RB SPRAY P35B ES FAILURE KI 1-C7 is in alarm.

What caused the alarm and what is the proper response to this alarm (KI 1-C7)?

A. Flow is < 1050 gpm, no response required, the Spray pump breaker is racked down for this test.

B. Flow is < 1500 gpm, raise RB Spray flow using CV-2400, RB Spray Block valve.

C. Flow is < 1050 gpm, raise RB Spray flow using DH-9, DH-10 Bypass valve.

D. Flow is < 1500 gpm, No response needed, expected alarm due to no flow through the flow transmitter.

Answer:

C. Flow is < 1050 gpm, raise RB Spray flow using DH-9, DH-10 Bypass valve.

Notes:

C is correct for the ES test since the RB Spray pump is recircing on the BWST. Low flow alarm setpoint is 1050 gpm.

A is incorrect, the Spray pumps are operated while the HPI pumps breakers are racked down for this test B is incorrect, although this would be done for an actual ES actuation, this would spray the RB down during this test, hence the valve is closed and tagged.

D is incorrect, the flow transmitter is in service for this test.

References:

OP-I 203.012J Change 37 OP-I 305.006 Change 30 History:

Modified from QID 564 Selected for 2010 RO!SRO exam

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0564 Rev: 0 Rev Date: 4/7/05 Source: Direct Originator: S.Pullin TUOI: AILP-RO-RBS Objective: 8 Point Value: I Section: 3.2 Type: RCS Inventory Control System Number: 013 System

Title:

Engineered Safety Features Actuation

==

Description:==

Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

KIA Number: 2.4.50 CFR

Reference:

45.3 Tier: 2 RO Imp: 3.3 RO Select: No Difficulty: 3 Group: 1 SRO Imp: 3.3 SRO Select: No Taxonomy: C Question: RO:I SRO:I Given:

- Plant is in cold shutdown.

- All necessary components have been aligned per 1305.006, Integrated ES System Test.

- All ES EVEN Digital Channels actuated per procedure using RB pressure transmitters.

- Annunciator RB SPRAY P35B ES FAILURE KI I-C7 is in alarm.

Which of the following is a proper response to this alarm (K11-C7)? PA A. No response required, the Spray pump breaker is racked down for this test.

\J4 si B. Raise RB Spray flow using CV-2400, RB Spray Block valve.

C. Raise RB Spray flow using DH-9, DH-10 Bypass valve.

D. No response needed, expected alarm due to no flow through the flow transmitter.

Answer:

C. Raise RB Spray flow using DH-9, DH-I 0 Bypass valve.

Notes:

C is correct for the ES test since the RB Spray pump is recircing on the BWST.

A is incorrect, the Spray pumps are operated while the HPI pumps breakers are racked down for this test B is incorrect, although this would be done for an actual ES actuation, this would spray the RB down during this test, hence the valve is closed and tagged.

D is incorrect, the flow transmitter is in service for this test.

References:

1203.012J, Chg. 035-00-0 1305.006, Chg. 020-04-0 History:

New for2005 RO exam

PROC.IWORK PLAN NO. I PROCEDUREIWORK PLAN TITLE: I PAGE: 40 of 49 1203.012J ANNUNCIATOR Ku CORRECTIVE ACTION CHANGE: 037 Location: C18 Page 1 of 2 Device and Setpoint (either of the following)

A. P-35B breaker (A-404) is open 55 seconds after ES CH 8 RB SPRAY actuation P35B ES B. RB spray flow <1050 gpm 55 seconds after ES CH 8 FAI LURE actuation Alarm: Kll-C7 1.0 OPERATOR ACTIONS CAUTION Attempting to reclose breaker with protective relay tripped may damage motor and circuit components.

1. IF breaker A-404 open, THEN perform the following:

NOTE Indications such as P1-2408, ESAS actuation alarms on Ku, ES Cabinet pressure indicators, wide range pressure indicators and recorders P1-2412, P1-2413, PR-2413 and SPDS/PMS may be relied upon.

A. IF RB Press >30 psig, THEN perform the following:

1) Verify that ES CH 7 actuated.

2) Check that RB Spray Pump (P-35A) has started.

B. Determine cause of P-35B failure.

2. IF RB Spray P-35B Flow on C16 is low, THEN perform the following:

NOTE Indications such as P1-2408, ESAS actuation alarms on Kll, ES Cabinet pressure indicators, wide range pressure indicators and recorders P1-2412, P1-2413, PR-24l3 and SPDS/PMS may be relied upon.

A. IF RB Press >30 psig, THEN perform the following:

1) Verify that ES CH 7 actuated.

2) Check that RB Spray Pump (P-35A) has started.

B. Determine and correct cause of low flow.

PROCJWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 41 of 49 1203.012J ANNUNCIATOR Ku CORRECTIVE ACTION CHANGE: 037 K1l-C7 Page 2 of 2

3. Refer to TS 3.6.5 for RB Spray Pump operability requirements.

4. IF desired to clear alarm, THEN perform either of the following:

• Clear ES condition Reset ES CH 8 per Engineered Safeguards Actuation System (1105.003)

• Close breaker A-404 Raise RB spray flow to >1050 gpm 2.0 PROBABLE CAUSES Pump P-35B failure to auto start

3.0 REFERENCES

Schematic Diagram Annunciator 1(11 (E-461, sheets 1-3)

PROCJWORK PLAN NO. PROCEDUREIWORK PLAN TITLE:

PAGE: 1 35 of 170 1305.006 INTEGRATED ES SYSTEM TEST CHANGE: 030 SUPPLEMENT 1 Page 38 of 73 3.7.12 Perform the following to vent P-35B discharge piping:

A. Connect hose to Pressure Point (PP-2400) and run end of hose to floor drain.

B. Slowly open PP-2400 Isol Before CV-2400 (BS-2400C) until solid stream of water flows out of hose.

C. Close BS-2400C.

3 .7.13 Start RB Spray Pump (P-35B).

3 7 14

. . Adjust DH-9 to obtain 1500 GPM spray flow.

A. IF necessary to obtain -1500 GPM AND DH-9 is fully open, THEN throttle open DH Test & Recirc Isol (DH-lO) 3.7.15 Stop P-35B AND leave handswitch in NORMAL-AFTER-STOP.

3.7.16 Close CV-l408.

3.8 Align DH Pump (P-34B) for start in DH mode as follows:

3.8.1 Close P-34B Suction From BWST (CV-l437) 3.8.2 Open P-34B Suction From RCS (CV-l435).

3.8.3 Unlock and close B DH Cooler SW Outlet Isol (SW-22B) 3.8.4 Verify LPI Block Valve (CV-l400) closed.

3.8.5 Verify Decay Heat Cooler Outlet (CV-l429) open.

3.8.6 Verify Decay Heat Cooler Bypass (CV-l432) closed.

3.8.7 Start DH Pump (P-34B).

3.8.8 Open LPI Block Valve (CV-1400).

3.8.9 Verify proper alignment by observing LPI P-34B Flow -3500 gpm.

3.8.10 Stop P-34B AND leave handswitch in NORMAL-AFTER-STOP.

3.8.11 Close CV-1400.

0

ES-401 PWR Examination Outline Form ES-401-2 Facility: Arkansas Nuclear One Unit I Date of Exam: 31512010 RO K/A Catpojy Points SRO-Only_Points Tier Group F A2 G* Total KKIKKKKAAAAG

• 1 213 4 5 6 1 2 3 4 Total 3 3 6

1. I 0

Emergency &

Abnormal 3 1 4 2

. 000 00 00 Evolutions 6 4 10 Tier Totals N/A N/A 000 00 00 3 2 5 1

2. 000000000000 Plant 1 OIl 2 3

. 2 ysems 000000000000 4 4 8 Tier Totals 000000000000 0 0 0 0 0 1 2 3 4 7

3. Generic Knowledge and Abilities Categories 2 1 2 2 0 0 0 0 of the RO Note: 1. Ensure that at least two topics from every applicable KA category are sampled within each tier outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals and SRO-only in each K/A category shall not be less than two).

table.

2. The point total for each group and tier in the proposed outline must match that specified in the based on NRC revisions.

The final point total for each group and tier may deviate by +/-1 from that specified in the table The final RO exam must total 75 points and the SRO-only exam must total 25 points.

that do not apply

3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions utions that are not at the facility should be deleted and justified; operationally important, site-specific systems/evol the elimination included on the outline should be added. Refer to Section D.1 .b of ES-401 for guidance regarding of inappropriate K/A statements.

in the group before

4. Select topics from as many systems and evolutions as possible; sample every system or evolution selecting a second topic for any system or evolution.

shall be selected.

5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.

the topics 7* The generic (G) K/As in Tiers I and 2 shall be selected from Section 2 of the K/A Catalog, but evolution or system. Refer to Section D.1 .b of ES-401 for the applicable K/As.

must be relevant to the applicable of each topic, the topics importance ratings (IRs)

8. On the following pages, enter the K/A numbers, a brief description group and tier totals for the applicable license level, and the point totals (#) for each system and category. Enter the A2 or G* on the for each category in the table above; if fuel handling equipment is sampled in other than Category A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate SRO-only exam, enter lion the left side of Column pages for RO and SRO-only exams.

For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, 9.

and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

ES-401 Form ES-401 -2

SRO Written Exam Tier I Group I

ES-401 PWR Examination Outline Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emer enc and Abnormal Plant Evolutions Tier 1/Grou 1 RO APE # / Name / Safety Function K K K A A G K/A Topic(s) IR # QID T 12312 y p

e 000007 (BW/E02&E10; CEIEO2) Reactor X EA2.1- Facility conditions and selection of 4.0 76 588 D Trip Stabilization Recovery / 1

- - appropriate procedures during abnormal and emer enc 0 erations 000008 Pressurizer Vapor Space Not selected N/A Accident! 3 000009 Small Break LOCA / 3 Not selected N/A 000011 Lar eBreakLOCA/3 Not selected N/A 000015/17 RCP Malfunctions /4 Not selected N/A 000022 Loss of Rx Coolant Makeup / 2 X AA2.04- How long PZR level can be 3.8 77 805 N maintained within limits 2.4.31 Knowledge of annunciator alarms, 000025 Loss of RHR System / 4 X indications or res onse rocedures. 4.1 78 806 N 000026 Loss of Component Cooling X AA2.O1- Location of a leak in the CCWS 3.5 79 807 N Water / 8 000027 Pressurizer Pressure Control Not selected N/A System Malfunction / 3 000029 ATWS I 1 Not selected NIA 000038 Steam Gen. Tube Rupture /3 X 2.4.18 Knowledge of the specific bases for 4.0 80 585 N EOPs.

0040 (BW/E05; CE/E05; W/E12) X 2.4.6- Knowledge of symptom based EOP 4.7 81 584 D team Line Rupture Excessive Heat

- mitigation strategies Transfer/4 000054 (CEIEO6) Loss of Main Not selected N/A Feedwater / 4 000055 Station Blackout /6 Not selected N/A 000056 Loss of Off-site Power /6 Not selected N/A 000057 Loss of Vital AC Inst. Bus /6 Not selected N/A 000058 Loss of DC Power / 6 Not selected N/A 000062 Loss of Nuclear Svc Water / 4 Not selected N/A 000065 Loss of Instrument Air /8 2.4.18 Knowledge of the specific bases for N/A EOPs Rejected system to 038 Steam Gen Tube Ru ture W/E04 LOCA Outside Containment /3 Not selected N/A ES-401 Form ES-401-2

ES-401 PWR Examination Outline Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emer enc and Abnormal Plant Evolutions Tier 1/Grou 1 RO APE#/ Name/SafetyFunction K K K A A G KIATopic(s) IR # QID T 12312 y p

e W/E1 I Loss of Emergency Coolant Not selected N/A Recirc. / 4 BW/E04; W/E05 Inadequate Heat Not selected NIA Transfer Loss of Seconda Heat Sink /4 000077 Generator Voltage and Electric Not selected NIA Grid Disturbances / 6 K/A Category Totals: 3 3 Group Point Total: 6 ES-401 Form ES-401-2

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0588 Rev: 0 Rev Date: 6/1/05 Source: Direct Originator: J.Cork TUOI: Al LP-RO-EOPO4 Objective: 11 Point Value: I Section: 4.3 Type: B&W EPEs/APEs System Number: ElO System

Title:

Post-Trip Stabilization

Description:

Ability to determine and interpret the following as they apply to the (Post-Trip Stabilization):

Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

K/A Number: EA2.1 CFR

Reference:

43.5 /45.13 Tier: I RO Imp: 2.5 RO Select: No Difficulty: 3 Group: 1 SRO Imp: 4.0 SRO Select: Yes Taxonomy: C Question: 76 RO:j SRO:1 Given:

- Reactor tripped due to a loss of both MFWPs approximately 15 minutes ago.

- Annunciator K02-B6 A3 L.O. RELAY TRIP is in alarm.

- AFW pump, P-75, is tagged out for maintenance.

- Steam Driven EFW Pump, P-7A, has tripped on overspeed.

- RCS pressure is 2000 psig.

- CETs are 612°F.

- Both OTSG levels are 30.

Which of the following procedures should be in use for the above conditions?

A. 1202.002, Loss of Subcooling Margin B. 1202.004, Overheating C. 1202.011, HPI Cooldown D. 1203.037, Abnormal ES Bus Voltage Answer:

B. 1202.004, Overheating Notes:

Answer B is correct, the Overheating EOP should be entered with CETs> 610°F and all MFW and EFW lost during loss of adequate Subcooling Margin.

Answer A is incorrect, this procedure would have been in use up to the point where CETs became > 610°F.

Answer C is incorrect, this procedure is entered from Loss of Subcooling Margin.

Answer D is incorrect, this procedure is used when ES bus voltage is low but not de-energized.

References:

1202.004, Chg. 006 History:

New for 2005 SRO exam.

Selected for 2010 SRO exam

CHANGE 1202.004 OVERHEATING 006 PAGE 1 of 17 ENTRY CONDITIONS NOTE Throughout this procedure, harsh containment values in brackets [] shall be used, where provided, if either of the following criteria are met:

• Average RB Temp >200°F

• RB Radiation Level 1 o R/hr

• RCS temp rising above either:

580° F T-hot with any RCP on OR 610°F CET temp with all RCPs off, following a Reactor trip.

• CET temp rising above 610°F AND all MEW and EFW is lost during loss of adequate SCM.

• Loss of all feedwater (MFW and EFW) following a Reactor trip.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0805 Rev: 0 Rev Date: 9/21/2009 Source: New Originator: S. Pullin TUOI: AILP-RO-TS Objective: 13 Point Value: I Section: 4.2 Type: Generic APEs System Number: 022 System

Title:

Loss of Reactor Coolant Makeup

==

Description:==

Ability to determine and interpret the following as they apply to Reactor Coolant Makeup: How long PZR level can be maintained within limits.

KIA Number: AA2.04 CFR

Reference:

43.5/45.13 Tier: I RO Imp: 2.9 RO Select: No Difficulty: 4 Group: I SRO Imp: 3.8 SRO Select: Yes Taxonomy: Ap Question: RO:1 SRO:J Given:

- RCS Cooldown in progress

- Tave is 295 F

- RCS Pressure is 440 psig.

- Pressurizer level is 85 inches

- All makeup has been lost

- Pressurizer level is dropping at 5 inches per minute

- Assuming pressurizer level rate of change remains the same When will LCO 3.4.9 Pressurizer, be entered due to low Pressurizer level and what is the bases per Technical Specification for the low level?

A. 2 minutes and to prevent violating NDTT Curve.

B. 4 minutes and to prevent violating LTOP Curve.

C. 6 minutes and to maintain the minimum ES bus powered pressurizer heaters OPERABLE.

D. 8 minutes and to maintain on scale pressurizer level indication.

Answer:

D. 8 minutes and to maintain on scale pressurizer level indication.

Notes:

D is correct, the limit per LCO 3.4.9 is less than or equal to 45 inches and the minimal water level limit has been established to ensure that water level is above the minimum detectable level.

A is incorrect, due to PZR level would be 75 inches which is below the administrative limit per OP-i 102.010 for PZR level, but does not violate the NDTT curve.

B is incorrect, due to PZR level would be 65 inches which is below the administrative limit per OP-I 102.010 for PZR level, but does not violate the LTOP curve.

C is incorrect, due to PZR level would be 55 inches which is at the pressurizer heater cutoff level which would deenergize the ES powered heaters.

References:

T.S. 3.4.9 Amendment 215 History:

New selected for 2010 SRO exam.

Pressurizer 3.4.9 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 Pressurizer LCO 3.4.9 The pressurizer shall be OPERABLE with:

a. Pressurizer water level 45 inches and 320 inches; and

b. A minimum of 126 kW of Engineered Safeguards (ES) bus powered pressurizer heaters OPERABLE.

---

NOTE-------- -------

OPERABILITY requirements on pressurizer heaters do not apply in MODE 4.

APPLICABILITY: MODES 1, 2, and 3, MODE 4 with RCS temperature> 262°F.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressurizer water level not A.1 Restore level to within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> within limits, limits.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 4 with RCS 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> temperature 262°F.

C. Capacity of ES bus C.1 Restore pressurizer heater 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> powered pressurizer capacity.

heaters less than limit.

D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition C not AND met.

D.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ANO-1 3.4.9-1 Amendment No. 215

Pressurizer B 3.4.9 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.9 Pressurizer BASES BACKGROUND The pressurizer provides a point in the RCS where liquid and vapor are maintained in equilibrium under saturated conditions for pressure control purposes to prevent bulk boiling in the remainder of the RCS. Key functions include maintaining required primary system pressure during steady state operation and limiting the pressure changes caused by reactor coolant thermal expansion and contraction during normal load transients.

The pressure control components addressed by this LCO include the pressurizer water level, the required heaters, and their controls. Pressurizer safety valves are addressed by LCO 3.4.10, Pressurizer Safety Valves.

The maximum water level limit has been established to ensure that a liquid to vapor interface exists to permit RCS pressure control during normal operation and proper pressure response for abnormalities. The water level limit thus serves two purposes:

a. Provides pressure control during normal operation; and

b. Prevents the peak RCS pressure from exceeding the safety limit of 2750 psig during an abnormality.

The maximum water level limit thus permits pressure control equipment to function as designed. The limit preserves the steam space during normal operation, so that both sprays and heaters can operate to maintain the design operating pressure. The level limit also prevents filling the pressurizer (water solid) during abnormalities, thus ensuring that pressure relief devices (electromatic relief valve (ERV) or code safety valves) can control pressure by steam relief rather than water relief. If the level limits were exceeded prior to an abnormality that creates a large pressurizer insurge volume leading to water relief, the maximum RCS pressure might exceed the design Safety Limit (SL) of 2750 psig or damage may occur to the ERV or pressurizer code safety valves.

The minimum water level limit has been established to ensure that water level is above the minimum detectable level.

The pressurizer heaters are used to maintain a pressure in the RCS so reactor coolant in the loops is subcooled and thus in the preferred state for heat transport to the steam generators (SG5). This function must be maintained with a loss of offsite power.

Consequently, the emphasis of this LCO is to ensure that the Engineered Safeguards (ES) bus powered heaters are adequate to maintain pressure for RCS loop subcooling with an extended loss of offsite power.

ANO-1 B 3.4.9-1 Amendment No. 215

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0806 Rev: 0 Rev Date: 9/21/2009 Source: New Originator: S. Pullin TUOI: A1LP-RO-AOP Objective: I Point Value: 1 Section: 4.2 Type: Generic APEs System Number: 025 System

Title:

Loss of RHR System

==

Description:==

Knowledge of annunciator alarms, indications, or response procedures.

KIA Number: 2.4.31 CFR

Reference:

41.10 / 45.3 Tier: I RO Imp: 4.2 RO Select: No Difficulty: 3 Group: I SRO Imp: 4.1 SRO Select: Yes Taxonomy: Ap Question: RO: SRO:F78 Given:

- Mode5

- RCS temperature 170 F

- RCS pressure 0 psig

- A RCP seal removed for maintenance

- A Decay Heat in service

- Following alarms are received

- DECAY HEAT FLOW HI/LO (K09-A8)

- DECAY HEAT VORTEX WARNING (K09-D8)

- ISOL VLV OPEN RC PRESS LO (KI 0-E5)

Which section of OP-1203.028, Loss of Decay Heat Removal, will be entered for the given conditions?

A. Section 6, Decay Heat Pump Trip B. Section 7, Suction Valve Closure C. Section 9, Loss of Both DH Systems RCS Pressure Boundary Intact D. Section 10, Loss of Both DH Systems RCS Pressure Boundary Open Answer:

B. Section 7, Suction Valve Closure Notes:

B is correct, with the given alarms K10-E5 would automatically cause the DHR Suction valve to close.

A is incorrect, the DHR Pump would still be running for the given condition. The pump does not automatically stop on valve closure.

C is incorrect, although the RCS is still intact with an RCP seal removed, the transition to loss of both DHR Pumps does not occur until RCS temperature is greater than 280 F D incorrect, the RCS is not open and the transition to loss of both DHR Pumps does not occur until RCS temperature is greater than 280 F

References:

OP-I 203.028 Change 021 op-i 203.0121 Change 046 History:

New selected for 2010 SRO exam.

CHANGE 1203.028 LOSS OF DECAY HEAT REMOVAL 021 PAGE 40 of 82 SECTION 7 SUCTION VALVE CLOSURE ENTRY CONDITIONS One or more of the following:

• DECAY HEAT FLOW HI/LO (K09-A8) alarm

• RCS temp rise

• Train A CET TEMP HI (K09-D6) alarm

• Train B CET TEMP HI (K09-E6) alarm

• CV-1050 AUTO CLOSE (K09-B7) or CV-1410 AUTO CLOSE (K09-B8) alarm

CHANGE LOSS OF DECAY HEAT REMOVAL 021 PAGE 41 of 82 1203.028 SECTION 7 SUCTION VALVE CLOSURE INSTRUCTIONS

1. Stop the running OH pump.

ication (1903.01 0).

2. Notify Shift ManagerlCRS to implement Emergency Action Level Classif

3. Terminate any operation causing pressure rise.

RCS level rise,

4. maintenance activities in the Reactor Building could be affected by THEN perform local evacuation of the affected areas.

5. i.E RCS temp exceeds 280° F, procedure.

THEN GO TO applicable Loss of Both OH Systems section of this NOTE

• Containment closure must be established prior to steam release.

02B provides

• Decay Heat Removal and LTOP System Control (1015.002), Form 1015.0 ry, heatup rate, and estimate of time to 200°F, time to steam release, time to core uncove required makeup rate.

6. jf any of the following conditions occur:

• Time remaining to steam release is, or becomes <1 hour DH removal can Qj be immediately restored

• RCS press >150 psig RCS loops jQ] filled w n (1102.010),

• RCS press >Decay Heat Sys Max Pressure limit of Plant Shutdown and Cooldo Attachment A ure, while continuing with THEN initiate containment closure per Attachment G of this proced this section.

(continued)

I CHANGE LOSS OF DECAY HEAT REMOVAL 1 021 PAGE 42 of 82 1203.028 SECTION 7 SUCTION VALVE CLOSURE NOTE

• CV-1 050 will close autom atically if Core Flood Tank T-2A Outlet (CV-2415) comes off its closed seat or if RCS press exceeds 320 psig.

comes off its

• CV-1410 will close automatically if Core Flood Tank T-2B Outlet (CV-2419) closed seat or if RCS press exceeds 385 psig.

• The auto close interlock is automatically reset when RCS press is <290 psig.

7. Determine and correct cause of valve closure.

below,

8. if RCS press is greater than the applicable limit listed THEN perform the following:

• RCS loops filled --150 psig

• RCS loops filled -- Decay Heat Sys Max Pressure limit of Plant Shutdown and Cool down (1102.010), Attachment A.

A. Initiate containment closure per Attachment G of this procedure.

B. Cycle the ERV as necessary to maintain RCS press within limits.

C. jf RCS press can fjQ[ be reduced below applicable limit, THEN perform the following:

1) Stop the running DH pump.

2) Close at least one of the following Decay Heat Suction valves:

• CV-1050

• CV-1410

• CV-1404

3) GO TO applicable Loss of Both DH Systems section of this procedure.

(continued)

PROC IWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: - I PAGE: 48 of 68 1203.0121 ANNUNCIATOR K10 CORRECTIVE ACTION CHANGE: 046 Location: C16 Device and Setpoint:

ISOL VLV OPEN RC Loop A Unit Press (PS1021) <700 psig along with RC PRESS valve open signal from either of the following:

LO CFT T2A Outlet Limit Switch (ZS2415A)

CFT T2B Outlet Limit Switch (ZS2419A)

Alarm: K10E5 1.0 OPERATOR ACTIONS ss, secure

1. Unless a lossofcoolant accident, or valve testing is in progre depressurization, and verify CFT outlets closed :

A. Core Flood Tank T-2A Outlet (CV-2415)

B. Core Flood Tank T-2B Outlet (CV-2419)

2. IF lossofcoolant accident is indicated, THEN GO TO Emergency Operating Procedure series (].202.XXX).

2.0 PROBABLE CAUSES

1. Valve testing during plant shutdown.

2. CFT outlet open and RC pressure <700 psig

3.0 REFERENCES

1. Schematic Diagram Annunciator 1<10 (E460, sheets 1 3)

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0807 Rev: 0 Rev Date: 9/21/2009 Source: New Originator: S Pullin TUOI: Al LP-RO-AOP Objective: 3 Point Value: I Section: 4.2 Type: Generic APEs System Number: 026 System

Title:

Loss of Component Cooling Water.

==

Description:==

Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water: location of a leak in the CCWS.

KIA Number: AA2.01 CFR

Reference:

43.5/45.13 Tier: I RO Imp: 2.9 RO Select: No Difficulty: 3 Group: I SRO Imp: 3.5 SRO Select: Yes Taxonomy: C Question: RO:j SRO:1 Given:

- Plant at 100%

- The following alarms are received

- ICW COOLER OUTLET TEMP HI (K12-E4)

- RCP BEEDOFF TEMP HI (K08-C7)

- A RCP seal temperature rising

- Skewed RCP Seal Injection flows indicated on C04

- RCS leak rate is 50 gpm Which of the following procedures provide the actions necessary to mitigate the abnormal operating condition?

A. OP-1203.039, Excess RCS Leakage B. OP-1203.026, Loss of Reactor Coolant Makeup C. OP-1203.031, Reactor Coolant Pump and Motor Emergency D. OP-1102.016, Power Reduction and Plant Shutdown Answer:

A. OP-I 203.039, Excess RCS Leakage Notes:

A is correct, because Excess RCS Leakage procedure is the only procedure that combats an intersystem LOCA.

B is incorrect, OP-1203.026 has a section to address makeup & purification system leaks, but with the indications given this is not considered a makeup & purification system leak.

C is incorrect, with the given indications the student could misdiagnose this as a seal failure issue, D is incorrect, with the given leak rate, a rapid plant shutdown would be necessary.

References:

OP-1203.039 Change 11 History:

New selected for 2010 SRO exam

CHANGE EXCESS RCS LEAKAGE 011 PAGE 3 of 14 1203.039

7. Check for primary to secondary leak indicated as follows:

A. Alarm or rising count rate on any of the following:

• A OTSG N-16 Detector (Rl-2691)

• B OTSG N-16 Detector (Rl-2692)

• Main Condenser Radiation Process Monitor (Rl-3632)

• Steam Line A High Range Rad Monitor (Rl-2682)

• Steam Line B High Range Rad Monitor (Rl-2681)

• SPING 2 Unit 1 Radwaste Area (RX-9825)

• SG-A N-16 AVG Leakrate GPM (SGALRGPM)

• SG-A N-16 Leakrate ROC (Rate of Change) GPM/HR (SGAROCI)

• SG-B N-16 AVG Leakrate GPM (SGBLRGPM)

• SG-B N-I 6 Leakrate ROC (Rate of Change) GPM/HR (SGBROCI)

B. Chemistry samples indicate rising secondary activity.

C. A leaking SG may exhibit the following at low feedwater fi ow rates:

• Higher SG level

• Lower FW flow rate

• Lower MEW pump speed D. jf primary to secondary leakage is indicated, THEN GO TO Small Steam Generator Tube Leaks (1203.023).

8. Check RCP seals for proper staging.

A. IF seal degradation OR seal failure is indicated, THEN GO TO Reactor Coolant Pump and Motor Emergencies (1203.031).

9. Check indications of Makeup and Purification System leakage.

• AUX BLDG Sump level rising

• Dirty Waste Drain Tank (T2OA and T2OB) level rising

• Equipment Drain Tank (TI I) level rising

• AREA MONITOR RADIATION HI alarm (KI0-BI)forAUX BLDG area A. IF Makeup and Purification System leakage is indicated, p and Purification THEN GO TO Loss of Reactor Coolant Makeup (1203.026), Large Makeu System Leak section.

CHANGE - -

EXCESS RCS LEAKAGE 011 PAGE 4 of 14 1203.039

10. Monitor RB Sump level.

A. E leakage into RB Sump is indicated,and isolate the leak using RCS Leak Detection THEN continue with efforts to locate (1103.013) fQ continue with this procedure.

11. Monitor Quench Tank (T42) pressure, level, and temperature.

A. jf leakage is indicated into Quench Tank, THEN GO TO Pzr Systems Failure (1203.015).

12. Check indications of RCS leakage into ICW system.

NOTE (CW Surge Tank T-37B Level (PDIS 2229) 0.5 to 2.7 psid (1 psid = 333 gallons)

• Nuclear Loop ICW Surge Tank (T37B) level rising

• Nuclear Loop ICW activity rising

• Indication of Letdown Cooler RCS leak into ICW:

Letdown Cooler ICW Outlet temp rising on PMS:

• 8PlCWtrend

• T2214 for E29A

• T22l5forE29B

• Indication of RCP Seal Cooler RCS leak into ICW:

RCP Seal Temp rising RCP Seal Bleedoff Temp rising Skewed RCP Seal Injection Flows NOTE With small leak rates, suffici ent time should be available to isolate one cooler at a time.

A. IF RCS leak into LETDOWN COOLER is indicated, THEN perform the following:

1) Isolate one or both Letdown Cooler (s) (E29AIB) by closing associated valves

• RC to Letdown Cool ers E29A (C04) (CV-1 213)

AND Letdow n Cooler s Outlet (RCS) E29A (C18) (CV-1214)

• RC to Letdown Coolers E29B (C04) (CV-1215)

AND Letdown Cooler Outlet (RCS) E29B (C18) s (CV-1216)

(12.A CONTINUED ON NEXT PAGE)

CHANGE 7 011 PAGE 10 of 14 1203.039 I EXCESS RCS LEAKAGE NOTE Recommended shutdown rates for RCS leaks inside containment with no additional corn plications are as follows:

• <50 gpm 0.5 to 5% per minute

• 50 gpm 5 to 10% per minute by Tech Spec 3.4.13

14. jf total RCS leakage is in excess of that allowed AND poses an immediate threat to plant operations, THEN perform the following:

A. IF reactor is Critical, THEN commence plant shutdown per Rapid Plant Shutdown (1203.045).

B. IF reactor is shutdown, THEN perform RCS cooldown by one of the foil owing:

1) IF RCS is cooling down due to HP1/break flow, independent of SG cooling, THEN perform Small Break LOCA Cooldowri (1203.041), while continuing with this procedure.

2) jf any RCP is running, THEN perform Forced Flow Cooldown (1203.040), while continuing with this procedure.

3) IF all RCPs are off, THEN perform Natural Circulation Cooldown (1203.01 3), while continuing with this procedure.

15. IF total RCS leakage is in excess of that allowed by Tech Spec 3.4.13 AND poses q immediate threat to plant operations, THEN perform the following:

and A. Bring reactor to cold shutdown per Power Reduction and Plant Shutdown (1102.016)

Plant Shutdown and Cool down (1102.010).

16. Advise Shift Manager to implement Emergency Action Level Classification (1903.010).

17. jf leakage is within Tech Spec 3.4.13 limits, THEN continue with efforts to locate and isolate the leak using RCS Leak Detection (1103.01 3) and proceed as directed by Operations Manager.

18. IF leak is isolated, THEN proceed as directed by Operations Manager.

N

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0585 Rev: 0 Rev Date: 9/21/2009 Source: New Originator: B. Possage TUOI: Al LP-RO-EOP Objective: 9 Point Value: 1 Section: 4.1 Type: Generic EPEs System Number: 038 System

Title:

Steam Generator Tube Rupture

Description:

Knowledge of the specific bases for EOPs.

KlANumber: 2.4.18 CFR

Reference:

41.10/43.1 /45.13 Tier: I RO Imp: 3.3 RO Select: No Difficulty: 3 Group: 1 SRO Imp: 4.0 SRO Select: Yes Taxonomy: Ap Question: RO: 80 SRO:j Given:

- SGTR in progress

- Rx is tripped

- RCS pressure 1350 psig

- RCS Thot 540°F

- Projected dose rate at site boundary at NUE criteria

- B SG level at 395 and rising rapidly

- A SG level stable at 40 Considering the above conditions, which of the following procedural actions will cause higher tube stresses than normal limitations but is acceptable during a SGTR per the EOP technical bases document?

A. Perform a cool down to less than 500°F at 100°F/hr and isolate bad SG.

B. Steam bad SG to maintain bad SG Tube-to-Shell DT <150°F (tubes colder).

C. Steam bad SG to maintain bad SG Tube-to-Shell DT <100°F (tubes hotter).

D. Establish a cool down rate of 250°F/hr to 500°F Thot.

Answer:

B. Steam bad SG to maintain bad SG Tube-to-Shell DT <150°F (tubes colder).

Notes:

B is correct, per Technical Bases during emergency cool downs the tube to shell delta T limits are relaxed.

With the given information an emergency cool down is required at the rate of <1= 240 F/hr.

A is incorrect, this rate is the normal cool down rate.

C is incorrect, this is the normal tube to shell delta T limit.

D is incorrect, this exceeds the allowed emergency cool down limit.

References:

OP-1202.006 Change 11 B&W EOP Technical Bases Document History:

New selected for 2010 SRO exam

CHANGE 1202.006 TUBE RUPTURE 011 PAGE 13 of 42 INSTRUCTIONS CONTINGENCY ACTIONS

17. jf bad SG level is approaching 410 due to 17. GO TO step 18.

leakage OR dose rate Alert criteria is projected at Site boundary, THEN establish emergency cooldown rate of 240°FIhr (4°Flmin) to 500°F T-hot as follows:

A. For good SC, place TURB BYP valves in A. IF TURB BYP valves are not available, HAND THEN operate ATM Dump Control System AND for good SG in HAND to maintain adjust to maintain cooldown rate 240°F/hr. cooldown rate 240°F/hr.

SGA SGB ATM CV-2676 DUMP ISOL CV-2619 ATM CV-2668 DUMP CNTRL CV-2618 I) IF both SGs are bad, THEN steam both SGs.

B. WHEN RCS press is <1700 psig, THEN bypass ESAS.

C. jf only one SG is bad, C. I.E both SGs are bad, THEN steam bad SG only as necessary to THEN steam both SGs.

maintain:

• MSSVs closed

• SC press:

990 psig if using TURB BYP valves 1 040 psig if using ATM Dump Control system

• SG level 410.

• SC Tube-to Shell zT 150°F (tubes colder).

• Desired cooldown rate if good SG TBV or ADV is full open.

A AREVA NUMBER TECHNICAL DOCUMENT 74-1152414-10 3.3.1.2 Tube-to-Shell AT The normal tube-to-shell AT limit for cooldowns is 100°F (tubes colder) and, during an emergency cooldown (3.3.1.1) this limit may be increased to 150°F. Methods to control tube-to-shell AT are discussed in Chapter III.G.

This relaxation is allowed to facilitate an emergency cooldown should it be required.

However, two important points should be considered:

a. Whenever tube-to-shell AT exceeds 100°F a post-transient stress evaluation will be required.

b. Higher tube-to-shell ATs will increase the tensile stresses on the tubes and may lead to higher leak flows. Indications of this occurring have been observed during actual tube leak transients.

Therefore, some judgment is required before a decision is made to increase tube-to-shell AT. Normally, it is recommended that tube-to-shell AT be kept much lower than the normal cooldown AT limit if at all possible. However, there may be cases where an increase in AT is necessary to accommodate an expeditious cooldown which may be accomplished with little or no risk (e.g., decision has already been made to isolate the affected SG and allow it to fill, thus increases in leak flow rate may not significantly impact the transient). As noted in section 3.3.1.1, the use of the emergency cooldown rate to 500°F should not result in excessive tube-to-shell ATs.

3.3.1.3 Cooldown Limits The normal cooldown limit is the Technical Specification limit. With the exception of section 3.3.1.1, this limit should not be exceeded during a plant cooldown when the RCS is subcooled. If the RCS is not subcooled, then this limit does not apply as discussed in Chapter ffl.B.

3.3.1.4 Summary of Limits During Cooldown The following limits should be observed, if at all possible:

a. If section 3.3.1.1 applies, then above 500°F the cooldown rate limit is 240°F/hr DATE PAGE 12/31/2005 Vol.3, ffl.E -17 Framatome ANP, Inc., an AREVA and Siemens company

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0584 Rev: 0 Rev Date: 5/20/05 Source: Direct Originator: J.Cork TUOI: AILP-RO-EOPO3 Objective: 10 Point Value: I Section: 4.1 Type: Generic EPEs System Number: 040 System

Title:

Steam Line Rupture

==

Description:==

Knowledge of symptom based EOP mitigation strategies.

KIA Number: 2.4.6 CFR

Reference:

41.10 / 43.5 / 45.13 Tier: I RO Imp: 3.7 RO Select: No Difficulty: 4 Group: 1 SRO Imp: 4.7 SRO Select: Yes Taxonomy: An Question: RO:j SRO:

A steam line rupture has occurred in the Reactor Building with the following conditions now present:

- ESAS actuated on channels I thru 6.

- All RCPs secured per RT-1 0.

- RB pressure 19 psig and dropping.

- HPI throttled due to existence of adequate SCM.

- RCS pressure is 1050 psig.

- T-hot is 490°F.

- EOP actions have terminated the overcooling.

The SE recommends to the CRS to restore normal operating pressure per RT-14 in order to reset ESAS and re-start RCPs.

As CRS, does this recommendation follow the EOP mitigation strategies?

A. Yes, overcooling event has been terminated.

B. No, this could overstress reactor vessel.

C. Yes, adequate SCM has been restored.

D. No, RB pressure is not within normal limits.

Answer:

B. No, this could overstress reactor vessel.

Notes:

B is correct, trainee must recognize that with RCPs secured and HPI having been initiated that PTS limits apply until an evaluation is performed prior to returning to normal pressure. PTS limits prevent overstressing reactor vessel.

A is incorrect, yes the overcooling has been terminated but normal operating pressure would violate procedure.

C is incorrect, subcooling margin was never lost but normal operating pressure would violate procedure.

D is incorrect, although RB pressure is a concern the overriding concern is with PTS concerns.

THIS QUESTION IS TIED to 43.1

References:

1202.012, chg. 004-03-0, RT-14 History:

New for 2005 SRO exam.

Selected for the 2010 SRO exam

I CHANGE I 1202.012 REPETITIVE TASKS 008 PAGE 34 of 50 Page 1 of 3 NOTE

. PTS limits apply ify of the following has occurred:

HPI on with all RCPs off RCS OlD rate> 1 00°F/hr with Tcold < 355°F RCS C/D rate> 50°F/hr with Tcold < 300° F

. Once invoked, PTS limits apply until an evaluation is performed to allow normal press control.

. When PTS limits are invoked SGTR is in progress, PZR cooldown rate limits apply.

14. Control RCS press within limits of Figure 3.

A. IF PTS limits apply or RCS leak exists, THEN maintain RCS press within limits of Figure 3.

B. IF RCS press is controlled AND will be reduced below 1650 psig, THEN bypass ESAS as RCS press drops below 1700 psig.

C. IF PZR steam space leak exists, THEN limit RCS press as PZR goes solid by one or more of the following:

1) Throttle makeup flow.

2) IF SCM is adequate, THEN throttle HPI flow by performing the following:

a.) Verify both HPI RECIRC valves (CV-1 300 and 1301) open.

b.) Throttle HPI.

3) Raise Letdown flow.

a) IF ESAS has actuated, THEN unless fuel damage or RCS to ICW leak is suspected, restore Letdown flow (RT 13).

4) Verify ERV Isolation open (CV-1 000) cycle ERV (PSV-1 000).

(14. CONTINUED ON NEXT PAGE) 1202.012 RT14 fl Rev 3-17-08

SRO Written Exam Tier I Group 2

ES-401 PWR Examination Outline Form ES-401 -2

-401 PWR Examination Outline Form ES-401-2 Emer enc and Abnormal Plant Evolutions Tier 1/Grou 2 RO E/APE#/Name/SafetyFunction K K K A A G KIATopic(s) IR # QID Typ 12312 e 000001 Continuous Rod Withdrawal/i AA2.05- Uncontrolled rod NIA withdrawal from available indications Rejected system to 005 Inoperable!Stuck Control Rod 000003 Dropped Control Rod / I Not selected N/A 000005 Inoperable/Stuck Control Rod / I X AA2.03 Required actions if more 82 589 D than one rod is stuck or mo erable 000024 Emergency Boration / I X AA2.05 Amount of boron to add g 83 808 M to achieve the re uired SDM 000028 Pressurizer Level Malfunction / 2 Not selected N/A 000032 Loss of Source Range NI /7 Not selected NIA 000033 Loss of Intermediate Range NI /7 Not selected N/A 000036 (BW/A08) Fuel Handling Accident / 8 Not selected NIP.

000037 Steam Generator Tube Leak /3 Not selected N/A 000051 Loss of Condenser Vacuum / 4 Not selected N/A 000059 Accidental liquid RadWaste Rel. / 9 Not selected 000060 Accidental Gaseous Radwaste Rel. / 9 Not selected N/A 0061 ARM System Alarms / 7 Not selected N/A 00067 Plant Fire On-site /8 Not selected N/A 000068 (BW/A06) Control Room Evac. /8 Not selected N/A 000069 (W/E14) Loss of CTMT Integrity /5 Not selected NIP.

000074 (W/E06&E07) lnad. Core Cooling / 4 Not selected N/A 000076 High Reactor Coolant Activity /9 Not selected N/A W/EO1 & E02 Rediagnosis & SI Termination /3 Not selected N/A W/Ei3 Steam Generator Over-pressure / 4 Not selected NIA W/E15 Containment Flooding / 5 Not selected N/A W/E16 High Containment Radiation /9 Not selected N/A BW/A01 Plant Runback /1 Not selected N/A BW/A02&A03 Loss of NNI-XIY /7 X AA2.1 Facility conditions and 4.0 84 591 D selection of appropriate procedures during abnormal and emer enc 0 erations BW/A04 Turbine Trip /4 Not selected N/A BW/A05 Emergency Diesel Actuation /6 Not selected N/A BWIAO7 Flooding /8 Not selected NIP.

BW/E03 Inadequate Subcooling Margin / 4 Not selected ES-401 Form ES-401-2

ES-401 PWR Examination Outline Form ES-401-2 X 2.4.47- Knowledge of EOP 85 592 D BW/E08; W/E03 LOCA Cooldown Depress. /4

- 4.0 implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe mana ement uidelines BW/E09; CE/Al 3; W/E09&El 0 Natural Circ. / 4 Not selected N/A BW/El3&E14 EOP Rules and Endosures Not selected N!A CE/All; W/E08 RCS Overcooling PTS / 4 Not selected N/A CEJA16 Excess RCS Leakage / 2 Not selected N/A CE/E09 Functional Recovery Not selected N/A I KJA Category Point Totals: I I I I I 3 I i I Group Point Total: I4 I ES-401 Form ES-401 -2

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0589 Rev: 0 Rev Date: 6/1/05 Source: Direct Originator: S.Pullin TUOI: A1LP-RO-TS Objective: 4 Point Value: I Section: 4.2 Type: Generic APEs System Number: 005 System

Title:

Inoperable/Stuck Control Rod

==

Description:==

Ability to determine and interpret the following as they apply to the Inoperable/Stuck Control Rod: Required actions if more than one rod is stuck or inoperable.

K/A Number: AA2.03 CFR

Reference:

43.5/45.13 Tier: 1 RO Imp: 3.5 RO Select: No Difficulty: 4 Group: 2 SRO Imp: 4.4 SRO Select: Yes Taxonomy: An Question: RO:J SRO:1 Given:

- Plant is at 40% power.

- Group 4, Rod 4 is stuck and is mis-aligned from the group by 7.5%.

- The rod can not be re-aligned with the group.

Subsequently Group 7 Rod 6 drops to 0% withdrawn.

What are the required action(s) per Technical Specifications for the above conditions?

A. Immediately trip the reactor.

B. Borate to restore SDM within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and perform Linear Heat Rate surveilllance, SR 3.2.5.1, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

C. Borate to restore SDM within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and verify the potential ejected rod worth is within the assumptions of the rod ejection analysis within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

D. Borate to restore SDM within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and place the plant in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Answer:

D. Borate to restore SDM within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and place the plant in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Notes:

Answer D is correct per TS 3.1.4 action C for two inoperable rods.

Answer A is incorrect, this action is performed for two dropped rods.

Answer B is incorrect, this action is performed for one inoperable rod and the time given for the stated condition is incorrect.

Answer C is incorrect, this action is performed for one inoperable rod and the time given for the stated condition is incorrect.

References:

T.S. 3.1.4 amendment 215 Do not include this spec in the student handout!!!

History:

New for 2005 SRO exam.

Selected for 2010 SRO exam

CONTROL ROD Group Alignment Limits 3.1.4 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 CONTROL ROD Group Alignment Limits LCO 3.1.4 Each CONTROL ROD shall be OPERABLE and aligned to within 6.5% of its group average height.

APPLICABILITY: MODES I and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CONTROL ROD A.1.1 Verify SDM to be within the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> inoperable, or not aligned limit provided in the COLR.

to within 6.5% of its group average height, or both.

Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter OR A.1 .2 Initiate boration to restore 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> SDM to within limit.

AND A.2.1 Restore CONTROL ROD 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> alignment.

OR A.2.2.1 Reduce THERMAL 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> POWER to 60% of the ALLOWABLE THERMAL POWER.

AND A.2.2.2 Verify the potential ejected 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> rod worth is within the assumptions of the rod ejection analysis.

AND ANO-1 3.1.4-1 Amendment No. 215

CONTROL ROD Group Alignment Limits 3.1.4 CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.2.3 ----------NOTE------------

Only required when THERMAL POWER is

> 20% RTP.

Perform SR 3.2.5.1. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time for Condition A not met.

C. More than one CONTROL C.1.1 Verify SDM to be within the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ROD inoperable, or not limit provided in the COLR.

aligned within 6.5% of its group average height, or OR both.

C.1 .2 Initiate boration to restore 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> SDM to within limit.

AND C.2 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> SURVEILLANCE_REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify individual CONTROL ROD positions are within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 6.5% of their group average height.

SR 3.1.4.2 Verify CONTROL ROD freedom of movement for 92 days each individual CONTROL ROD that is not fully inserted.

ANO-1 3.1.4-2 Amendment No. 215

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0808 Rev: 0 Rev Date: 9/22/2009 Source: Modified Originator: S.PuIIin TUOI: AILP-RO-POISN Objective: 14 Point Value: I Section: 4.2 Type: Generic Abnormal Plant Evolutions System Number: 0024 System

Title:

Emergency Boration

==

Description:==

Ability to determine and interpret the following as they apply to the Emergency Boration:

Amount of boron to add to achieve the required SDM.

K/A Number: AA2.05 CFR

Reference:

43.5 / 45.13 Tier: I RO Imp: 3.3 RO Select: No Difficulty: 4 Group: 2 SRO Imp: 3.9 SRO Select: Yes Taxonomy: Ap Question: RO:1 SRO: I REFERENCE PROVIDED

- Rx has tripped with three CRDs stuck full out.

- Core lifetime = 150 EFPD

- RCS initial Boron concentration = 810 ppm

- Chemistry reports that the RCS boron concentration is 2200 ppm.

Which of the following contains guidance that must be used, for the above conditions?

A. No action required, SDM is adequate B. 1202.012, RT-12 Emergency Boration C. 1203.01 7, Moderator Dilution D. 1103.01 5, Reactivity Balance Calculation Answer:

B. 1202.012, RT-12 Emergency Boration Notes:

Answer B is correct, using Afl. B-16 from the plant data book, the examinee should determine that adequate SDM has not been established and Emergency Boration must be performed until adequate SDM is established.

Answer A is incorrect, SDM is not adequate.

Answer C is incorrect, although this might seem like a logical choice, this procedure should not be used for these conditions.

Answer D is incorrect, although this might seem like a logical choice, use of the Reactivity Balance Calculation procedure does not have any plant actions in it.

References:

1202.012 RT-12, Chg. 008 CALC-ANOI -NE-08-00007 NOTE: CALC AU. B-16 must be in SRO handout!?!!

History:

Modified from QID 678 Selected for 2010 SRO exam

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0678 Rev: 0 Rev Date: 2121/07 Source: Direct Originator: Cork/Possage TUOI: AILP-RO-POISN Objective: 14 Point Value: I Section: 4.2 Type: Generic Abnormal Plant Evolutions System Number: 024 System

Title:

Emergency Boration

==

Description:==

Ability to determine and interpret the following as they apply to the Emergency Boration:

Amount of boron to add to achieve the required SDM.

K/A Number: AA2.05 CFR

Reference:

43.5 /45.13 Tier: 1 RO Imp: 3.3 RO Select: No Difficulty: 3 Group: 2 SRO Imp: 3.9 SRO Select: No Taxonomy: An Question: RO:J SRO:I Given:

Rx has tripped with two CRD5 stuck full out.

Core lifetime = 250 EFPD r

- All immediate actions have been performed

- RCS initial Boron concentration = 638 ppm Chemistry reports that the RCS boron concentration is 1656 ppm.

(Reference Provided)

Which of the following contains guidance that must be used, if any, for the above conditions?

A. No action required, SDM is adequate.

B. 1203.017, Moderator Dilution C. 1202.012, RT-12 Emergency Boration D. 1202.010, ESAS Answer:

C. 1202.012, RT-12 Emergency Boration Notes:

Answer C is correct, using Att. B-16 from the plant data book, the examinee should determine that adequate SDM has not been established and Emergency Boration must be performed until adequate SDM is established.

Answer A is incorrect, SDM is not adequate.

Answer Ba is incorrect, although this might seem like a logical choice, this procedure should not be used for these conditins.

Answer D is incorrect, use of the ESAS procedure is not required for inadequate SDM.

References:

1202.012 RT-12, Chg. 004-06-0 CALC-A1-NE-2005-003, Rev. 0 NOTE: CALC Aft. B-16 must be in SRO handout!!!!

History:

New for 2007 SRO exam.

CHANGE 1202.012 REPETITIVE TASKS 008 PAGE 29 of 50 Page 2 of 4

12. (Continued).

10) IF Batch Controller output rate <5 gpm THEN perform the following:

a) Stop running Boric Acid pump(s) (P-39A, P-39B).

b) Close CV-1 250.

C) Stop Batch Controller by depressing stop key.

d) GOTOstepB.

11) Adjust Pressurizer Level Control Setpoint to 220.

12) Open BWST Outlet to OP HPI Pump (CV-1407 or 1408).

13) WHEN PZR level is 100, THEN establish maximum Letdown flow.

14) Perform the following as necessary to maintain MU Tank level 55 to 86:

a) Close Batch Controller Outlet (CV-1 250).

b) Stop running Boric Acid Pump(s) (P-39A, P-39B).

c) Place 3-Way valve in BLEED.

d) WHEN MU Tank level is lowered to desired level, THEN perform the following:

(1) Return 3-Way valve to LETDOWN.

(2) Start available Boric Acid Pump(s) (P-39A or B or both).

(3) Open Batch Controller Outlet (CV-1 250).

15) As time permits, determine actual required boration as follows:

a) Obtain required boron concentration from the Plant Data Book ppmB.

b) Calculate batch add required using Plant Computer OR Soluble Poison Concentration Control (1103.004), Attachment A.3, Calculation of Feed Volume For Batch Boration or Dilution, gal.

c) Use 1103.004, Attachment D, Volume of BAAT vs. Depth of Liquid to determine desired final BAAT level, in.

(12. CONTINUED ON NEXT PAGE) 1202.012 U RT12 Rev 9-04-08

CHANGE 1202.012 REPETITIVE TASKS 008 PAGE 30 of 50 Page 3 of 4

12. (Continued).

16) WHEN required amount of boric acid has been added per step 15)

OR as determined by Reactor Engineering, THEN perform the following:

a) Stop Boric Acid pump (P39A and B).

b) Close Batch Controller Outlet (CV-1 250).

c) Verify MU Tank level 55 to 86 Q close BWST Outlet to OP HPI pump (CV-I 407 or 1408).

d) Adjust Letdown flow to desired rate.

(12. CONTINUED ON NEXT PAGE) 1202.012 RT12 Rev 9-04-08 Q

Attachment B-16: Boron Concentration for 1.5% Shutdown Margin During Emergency Boration 2800 I I I I 2700 Burnup 1.5% ak/k SD Boron Conc EFPD 2 flop Rods 3 mop Rods 2600 0 2050 2542 100 1849 2334

% 200 1655 2140 2500 434 1227 1663 N

2400 434+ 1234 1663 N 474+ 1134 1565 2300 +APSRs out E 4.

2200 *S%,

2100 N

2000

\

C.) \

\

1900

\

1800 %4%

N

%4 1700 1600 1500 ._____

1400 ._____

1300 1200 E 2 Inoperable Rods 3 or IVbre Inoperable i]

\

1100 0 50 100 150 200 250 300 350 400 450 500 Cycle Lifetime, EFPD CALC-ANO 1 -NE-08-00007 ANO-1 Cycle 22 Plant Data Book Page 53 of 53 Rev. 000

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0591 Rev: 0 Rev Date: 6/6/05 Source: Direct Originator: S.Pullin TUOI: AILP-RO-ANNI Objective: 1 Point Value: I Section: 4.3 Type: B&W EPEs/APEs System Number: A02 System

Title:

Loss of NNI-X

==

Description:==

Ability to determine and interpret the following as they apply to the (NNI-X): Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

K/A Number: AA2.1 CFR

Reference:

43.5 / 45.13 Tier: I RO Imp: 3.6 RO Select: No Difficulty: 4 Group: 2 SRO Imp: 4.0 SRO Select: Yes Taxonomy: An Question: RO: SRO:á Given:

- Pressurizer Level Control Valve CV-1 235 indicates 50% open.

- RC Pump Seals Total lnj Flow valve CV-1 207 indicates 50% open.

- Letdown flow indication is zero.

- Letdown pressure indication is zero.

- Letdown Orifice Bypass valve CV-1 223 indicates 50% open.

- RCS pressure is 2210 psig and slowly rising.

- Pressurizer Spray valve CV-1008 indicates closed.

What procedure should be in use due to the above conditions?

A. 1203.015, Pressurizer Systems Failure B. 1203.024, Loss of Instrument Air C. 1203.047, Loss of NNI Power D. 1203.012B, ACA for K10-A8 LETDOWN TEMP HI Answer:

C. 1203.047, Loss of NNI Power Notes:

Answer C is correct since the conditions given are representative of a loss of NNI X and Y power.

Answer A is incorrect, this would be in use if Spray valve was failed due to something other than a loss of NM power.

Answer B is incorrect, this would be in use for failed valves due to loss of IA, but the positions given are different than for loss of air alone.

Answer D is incorrect, this is chosen for hi letdown temp but letdown flow would still be indicated while the question states there is none.

References:

1203.047, Chg. 000-01-0 History:

New for 2005 SRO exam.

Selected for 2010 SRO exam

CHANGE 1203.047 LOSS OF NNI POWER 000-01-0 PAGE 2 of 9 INSTRUCTIONS CONTINGENCY ACTIONS NOTE

• MU Tank level recorder is inoperable.

• Pressurizer Level Control valve (CV-1235) and RC Pump seals Total INJ Flow valve (CV-1207) fail as follows:

• both fail to 50% on loss of NNI X AC and NNI X DC

• both fail to 50% on loss of NNI X DC only

• CV-1 207 fails closed on loss of NNI X AC only

• CV-1235 failure position is indeterminate on loss of NNI X AC only

• Automatic Pressurizer Heater, Spray, and ERV controls are inoperable.

• Letdown Flow indication is lost.

• If NNI Y AC power is lost, the following occurs:

• Letdown Orifice Bypass (CV-1 223) fails to 50%

• Letdown_Pressure_indication_is_lost

2. IF any combination of both NNI X and NNI V 2. RETURN TO step 1.

power is lost, THEN perform the following:

A. TnptheRx AND perform 1202.001, REACTOR TRIP in conjunction with this procedure.

B. Manually actuate EFW verify proper actuation and control (1 202.012, RT 5).

C. Trip both MEW pumps.

D. Open BWST Outlet to OP HPI pump (CV-1407 or 1408).

E. Operate TURB BYP valves in HAND to E. IF TURB BYP valves are not available, control SG press 970 to 1020 psig. THEN verify ATM Dump Control System operates to maintain SG press 1000 to 1040 psig.

F. Close RCS Makeup Block (CV-1 233 or 1234)

G. Operate HPI Block (CV-1 220 or 1285) associated with OP HPI pump to maintain PZR level 90 to 110.

(2. CONTINUED ON NEXT PAGE)

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0592 Rev: 0 Rev Date: 6/6/05 Source: Direct Originator: S.Pullin TUOI: Al LP-RO-ASDCD Objective: 2 Point Value: 1 Section: 4.3 Type: B&W EPEs/APEs System Number: E08 System

Title:

LOCA Cool down

==

Description:==

Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe management guidelines KIA Number: 2.4.16 CFR

Reference:

41.10 / 43.5 / 45.13 Tier: 1 RO Imp: 3.0 RO Select: No Difficulty: 4 Group: 2 SRO Imp: 4.0 SRO Select: Yes Taxonomy: Ap Question: RO:j SRO:1 85 Given:

- Rx was shutdown using 1203.045 Rapid Plant Shutdown,

- Due to a RCS leak

- RCS pressure 1720 psig and lowering slowly

- HPI flow 150 gpm

- A & B SG pressure 910 psig

- RCS cool down rate 35°F per hour

- All Turbine bypass valves closed Which procedure should be in use?

A. 1202.001, Overcooling B. 1203.041, Small Break LOCA cool down C. 1203.040, Forced Flow cool down D. 1202.01 0, ESAS Answer:

B. 1203.041, Small Break LOCA cool down Notes:

Answer B is correct with an uncontrolled cool down continuing due to breaklHPl flow, regardless of SG status.

Answer A is incorrect, Overcooling entry conditions have not yet been met Answer C is incorrect, although RCPs are running, there is no control of the cool down.

Answer D is incorrect, although parameters are close to ES actuation setpoints, the ESAS procedure would eventually transition to 1203.041.

References:

1203.039, Chg. 011 History:

New for 2005 SRO exam.

Selected for 2010 SRO exam

CHANGE 1203.039 EXCESS RCS LEAKAGE 011 PAGE 10 of 14 NOTE Recommended shutdown rates for RCS leaks inside containment with no additional complications are as follows:

• <50 gpm 0.5 to 5% per minute

• 50gpm--5tolO%perminute

14. IF total RCS leakage is in excess of that allowed by Tech Spec 3.4.13 AND poses an immediate threat to plant operations, THEN perform the following:

A. IF reactor is Critical, THEN commence plant shutdown per Rapid Plant Shutdown (1203.045).

B. IF reactor is shutdown, THEN perform RCS cooldown by one of the foIl owing:

1) IF RCS is cooling down due to HPI/break flow, independent of SG cooling, THEN perform Small Break LOCA Cooldown (1203.041), while continuing with this procedure.

2) IF any RCP is running, THEN perform Forced Flow Cooldown (1203.040), while continuing with this procedure.

3) I all RCPs are off, THEN perform Natural Circulation Cooldown (1203.01 3), while continuing with this procedure.

15. IF total RCS leakage is in excess of that allowed by Tech Spec 3.4.13 AND poses immediate threat to plant operations, THEN perform the following:

A. Bring reactor to cold shutdown per Power Reduction and Plant Shutdown (1102.016) and Plant Shutdown and Cool down (1102.010).

16. Advise Shift Manager to implement Emergency Action Level Classification (1903.010).

17. IF leakage is within Tech Spec 3.4.13 limits, THEN continue with efforts to locate and isolate the leak using RCS Leak Detection (1103.013) and proceed as directed by Operations Manager.

18. IF leak is isolated, THEN proceed as directed by Operations Manager.

SRO Written Exam Tier 2 Group I

PWR Examination Outline Form ES4OI-2 ES-401 PWR Examination Outline Form ES-401-2 Rant Systems-Tier 2/Gryp I (SRO)

K K K K K K A A A A G KJA Topic(s) IR # OlD T 1 2 3 4 5 6 1 .2 3 4 y p

X A2.02 Conditions which 3.9 86 809 N 003 Reactor Coolant Pump exist for an abnormal shutdown of a RCP in

• . comparison to a normal shutdown of RCP 004 Chemical and Volume Not Selected N!A Control 005 Residual Heat Removal Not Selected NIA 006 Emergency Core Cooling Not Selected NIA 007 Pressurizer Relief/Quench Not Selected NIA Tank V.:

008 ComponentCooling Water . NotSelected NIA A2.02 Spray failures 3.9 87 762 R 010 Pressurizer Pressure Control 012 Reactor Protection Not Selected NIA X A2.06 Inadvertent ESFAS 4.0 88 812 N 013 Engineered Safety Features Actuation actuation 022 Containment Cooling Not Selected NIA 025 Ice Condenser Not Selected N!A 026 Containment Spray Not Selected NIA 039 Main and Reheat Steam Not Selected N!A 059 Main Feedwater Not Selected NIA 2.2.22 Knowledge of 4.7 89 811 N 061 Auxiliary/Emergency Feedwater limiting conditions for operations and safety limi 062 AC Electrical Distribution Not Selected NIA X 2.2.42 Ability to recognize 4.6 90 810 N 063 DC Electrical Distribution

• • . system parameters that are entry-level conditions for Technical Specifications 064 Emergency Diesel Generator Not Selected N!A 073 Process Radiation Not Selected NIA Monitoring . -

076 Service Water Not Selected NIA 078 Instrument Air Not Selected NIA 103 Containment Not Selected NIA K/ACategoryPointTotals: 3 2. Group PointTotal: 5 ES-401 Form ES-401-2

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0809 Rev: 0 Rev Date: 9/23/2009 Source: New Originator: S Pullin TUOI: Al LP-RO-AOP Objective: 6 Point Value: I Section: 3.4 Type: Heat Removal from Reactor Core System Number: 003 System

Title:

Reactor Coolant Pump System (RCPs)

Description:

Ability to (a) predict the impacts of the following malfunctions or operations on the RCP5; and (b) based on those predictions use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Conditions which exist for an abnormal shutdown of an RCP in comparison to a normal shutdown of an RCP KIA Number: A2.02 CFR

Reference:

41 .5/43.5/45.3/45/13 Tier: 2 RO Imp: 3.7 RO Select: No Difficulty: 3 Group: I SRO Imp: 3.9 SRO Select: Yes Taxonomy: C Question:

RO:i SRO:1 86 Given:

- 100% Power,

- C RCP seal bleed off temperature 210 F.

- C RCP motor bearing temperature 185 F and stable,

- C RCP motor inboard vibration alert alarm,

- C RCP seal cavity pressure oscillating from 650 to 1250 psig.

What is the appropriate section and action of 1203.031, Reactor Coolant Pump and Motor Emergency which will mitigate the consequences of these malfunctions?

A. Section 2, Seal Failure, Reduce reactor power to within the capacity of unaffected RCP combination and stop the affected RCP per Reactor Coolant Pump Operation, OPI 103.006.

B. Section 2, Seal Failure, Trip the Reactor and trip the affected RCP.

C. Section 5, Motor / Bearing Trouble, Reduce reactor power to within the capacity of unaffected RCP combination and stop the affected RCP per Reactor Coolant Pump Operation, OPI 103.006.

D. Section 5, Motor / Bearing Trouble, Trip the Reactor and trip the affected RCP.

Answer:

B. Section 2, Seal Failure, Trip the Reactor and trip the affected RCP.

Notes:

B is correct, a seal bleedoff temperature of greater than 200 F with no change in cooling (seal injection or ICW flow) meets the requirements to trip the RCP due to seal failure section.

A is incorrect. The given conditions require an abnormal shutdown of an RCP instead of a normal shutdown of an RCP.

C is incorrect. The given conditions require an abnormal shutdown of an RCP instead of a normal shutdown of an RCP.

D is incorrect. The given conditions do not indicate a bearing problem that warrents stopping the RCP.

References:

OP-1203.031 Change 018 History:

New selected for 2010 SRO exam

CHANGE 1203.031 REACTOR COOLANT PUMP AND MOTOR EMERGENCY 018 PAGE 38 of 38 ATTACHMENT A Page 1 of 1 RCP PARAMETERS Seal DegradationlSeal Failure

1. of the following are criteria to SECURE the affected RCP per Section 1 Seal Degradation

• RCP seal cavity pressure oscillations exceed 800 psi peak-to-peak

• AP across any stage exceeds 2/3 of system pressure on a running RCP OR exceeds 80% of system pressure on an idle RCP.

• 2.5 gpm total seal oufflow, including seal bleedoff (excluding shaft sleeve leakage),

AND a loss of seal injection

• Seal bleed off temp >40°F above 1st stage seal temp

• RCP seal bleed off or seal stage temp reaches 180°F, fQ no interwption of seal injection OR ICW flow.

2. of the following are criteria to TRIP the affected RCP per Section 2 Seal Failure

• 10 gpm rise in RCS leak

[Qa change in seal cavity pressure behavior.

• RCP seal bleed off or seal stage temp reaches 200° F Q no change in seal injection ICW flow.

• L\P across a single stage equal to RCS press, with seal bleed off flow established.

Loss of Cooling Water to RCP Motors or MotorlBearing Trouble

1. IF Motor Bearing Temperature >190°F (167°F for P-32B)

AND continues to rise, THEN SECURE the affected RCP per section 4 andlor section 5 of this procedure.

2. ANY of the following are criteria to SECURE the RCP per section 5 of this procedure:

• P32B, P32C or P32D PUMP SHAFT vibration; more than one channel 25 mils, after startup stabilization

• P32A PUMP SHAFT vibration; more than one channel 28 mils, after startup stabilization

3. of the following are criteria to TRIP the affected RCP per section 4 andlor section 5 of this procedure:

• Motor current exceeds 800 amps

• Winding temperature exceeds 300°F

• Bearing temperature exceeds 225°F (176°F for P32B)

• P-32B or D MOTOR vibration; more than one channel >20 mils after startup stabilization

• P-32A orG MOTOR vibration; more than one channel >0.8 in/sec after startup stabilization

• ANY RC PUMP SHAFT vibration 29 mils after startup stabilization

CHANGE 1203.031 REACTOR COOLANT PUMP AND MOTOR EMERGENCY 018 PAGE 8 of 38 SECTION 2 SEAL FAILURE ENTRY CONDITIONS One or more of the following:

• 1O gpm rise in RCS leak a change in seal cavity pressure behavior.

• RCP seal bleed off or seal stage temp reaches 200°F FQ no change in seal injection ICW flow.

• zP across a single stage equal to RCS press, with seal bleed off flow established.

CHANGE 1203031 REACTOR COOLANT PUMP AND MOTOR EMERGENCY 018 PAGE 9 of 38 I SECTION 2 SEAL FAILURE INSTRUCTIONS if tripping the affected RCP(s) will result in an automatic reactor trip, THEN perform the following:

A. Trip reactor.

B. Trip affected RCP(s).

C. While continuing with follow-up actions, refer to Emergency Operating Procedure (1 202.XXX).

2. jf tripping the affected RCP(s) will JQI cause an automatic reactor trip, THEN perform the following:

A. Trip affected RCP(s).

B. Verify proper CS response.

C. if only I RCP in operation per loop, THEN enter Tech Spec 3.4.4 Condition A (18-hour time clock).

3. jf HPI is required to maintain RCS inventory, THEN trip reactor AND refer to Emergency Operating Procedure (1202.XXX).

(continued)

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0762 Rev: 0 Rev Date: 11/11/200 Source: Repeat Originator: Steve Pullin TUOI: ANO-1-LP-RO-RCS Objective: 6 Point Value: I Section: 3.3 Type: Reactor Pressure Control System Number: 010 System

Title:

Pressurizer Pressure Control System (PZR PCS)

==

Description:==

Ability to (a) predict the impacts of the following malfunctions or operations on the PZR PCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Spray valve failures KIA Number: A2.02 CFR

Reference:

41.5 / 43.5 / 45.3 / 45.13 Tier: 2 RO Imp: 3.9 RO Select: No Difficulty: 4 Group: I SRO Imp: 3.9 SRO Select: Yes Taxonomy: An Question: RO:j SRO:1 87 Given:

-Unit I is operating at 40% power.

-The Unit is in three pump ops due to the failure of P-32B.

-The Pressurizer Spray Control valve (CV-1 008) fails open.

The ATC attempts to close the Pressurizer Spray Isolation valve (CV-1 009) and it will NOT close

-Reactor Coolant Pressure is at 2100 psig and slowly lowering with all Pzr Heaters on.

What is the correct procedure and correct action for this condition?

A. 1202.001 Reactor Trip, and trip the Reactor.

B. 1202.001 Reactor Trip, and stop P-32C.

C. 1203.015 PZR System Failure, and trip the Reactor.

D. 1203.015 PZR System Failure, and stop P-32C.

Answer:

D. 1203.015 PZR System Failure, and stop P-32C.

Notes:

A is incorrect. Since the Power to Pump trip entry conditions are not met.

B is incorret with the correct action but with the incorrect procedure since the Power to Pump trip entry conditions are not met.

C is incorrect with the correct procedure but incorrect action.

D is correct.

References:

1203.015 Pzr System Failure Chg 16 History:

New for the 2009 Retake SRO Exam Selected for 2010 SRO exam REPEAT

CHANGE 1203.015 PRESSURIZER SYSTEMS FAILURE 016 PAGE 17 of 24 SECTION 6 PRESSURIZER SPRAY VALVE (CV-1008) FAILURE INSTRUCTIONS jf failed open, THEN place Pressurizer Spray Control switch in HAND AND attempt to close CV-1 008 (modulating valve).

NOTE CV-1 009 torque switch can be overridden in the OPEN or CLOSE direction by holding the hand switch in the respective position.

A. if CV-1 008 will NOT close, THEN close Pressurizer Spray Isolation Valve (CV-1 009).

B. Verify Pressurizer heaters return RCS pressure to normal.

CAUTION Pressurizer spray shall not be used if the temperature difference between the Pressurizer and the spray fluid is >430°F (TRM 3.4.3). Closing CV-1 009 isolates the CV-1008 bypass spray flow.

C. jf necessary, THEN control spray flow by cycling Pressurizer Spray Isolation Valve (CV-1 009) open and closed.

D. IF both CV-1 008 and CV-1009 do NOT close fjQ. RCS pressure is dropping, THEN perform the following:

1) Verify all PZR heaters ON.

2) Immediately begin reducing load to 40% at 1 0%/mm per Rapid Plant Shutdown (1203.045).

3) jf 4 RCPs are running AND BOTH of the following conditions are met:

• Load is reduced to 675 MWe (75% load)

• Reactor power is 75%,

THEN perform the following:

a. Start C RCP HP Oil Lift Pump (P-63C) and C RCP Backstop Lube Oil Pump (P-81C).

b. Stop C RCP (P-32C).

c. WHEN zero speed is indicated, THEN stop P-63C and P-81C.

(continued)

F CHANGE 1203.015 I PRESSURIZER SYSTEMS FAILURE 016 PAGE 18 of 24 SECTION 6 PRESSURIZER SPRAY VALVE (CV-1 008) FAILURE NOTE In Modes 1 and 2, operation with only one RCP in each loop causes entry into TS 3.4.4 Condition A.

4) !E 3 RCPs running all of the following conditions are met:

• Load is reduced to 360 MWe (40% load)

• Reactor power is 55%,

• C and D RCPs in-service THEN perform the following:

a) Start C RCP HP Oil Lift Pump (P-63C) and C RCP Backstop Lube Oil Pump (P-81C).

b) Stop C RCP (P-32C).

C) WHEN zero speed is indicated, THEN stop P-63C and P-81C.

d) Enter TS 3.4.4 Condition A.

5) IF 3 RCPs running, AND D RCP is secured, THEN perform the following:

a) Trip Reactor.

b) Secure P-32C as follows:

(1) Start C RCP HP Oil Lift Pump (P-63C) and C RCP Backstop Lube Oil Pump (P-81C).

(2) Stop C RCP (P-32C).

(3) WHEN zero speed is indicated, THEN stop P-63C and P-81C.

c) Perform Reactor Trip (1202.001) while continuing with this procedure.

d) Enter TS 3.4.5 Condition A.

6) WHEN conditions permit a reactor building entry, THEN attempt to manually close either CV-1 008 or CV-1 009.

E. Contact Ops Manager.

(continued)

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0812 Rev: 0 Rev Date: 9/24/2009 Source: New Originator: S. Pullin TUOI: Al LP-RO-ESAS Objective: 6 Point Value: I Section: 3.2 Type: Reactor Coolant System Inventory Control System Number: 013 System

Title:

Engineered Safety Features Actuation System ESFAS; and

==

Description:==

Ability to (a) predict the impacts of the following malfunctions or operations on the (b) based ability on those predictions use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: lnadvertant ESFAS actuation.

K/A Number: A2.06 CFR

Reference:

41.5 / 43.5 / 45.3 / 45.13 Tier: 2 RO Imp: 3.7 RO Select: No Difficulty: 3 Group: 1 SRO Imp: 4.0 SRO Select: Yes Taxonomy: Ap Question: RO:j SRO:I 88 Given

- Plant at 100% power

- P-2B Condensate Pump OOS

- Inadvertent actuation of ES Channel #1

- S/u #1 OOS for maintenance LCO 3.8.1 .A 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Time Clock in effect to mitigate What would be the impact to the plant due to this malfunction and what procedure would be used the effects?

Actuation A. #1 Emergency Diesel Generator would start and use OP-I 105.003, Engineered Safeguards System to reset the tripped channel.

B. Red Train High Pressure Injection would occur and use 1202.01 0, ESAS EOP to override HPI C. Loss of power to A-I bus and use 1202.001, Reactor Trip EOP realign D. All Seal Return isolates and use 0P1203.031, Reactor Coolant Pump and Motor Emergencies to seal bleed off.

Answer:

C. Loss of power to A-I bus and use 1202.001, Reactor Trip EOP Notes:

in C is correct, the Unit Aux supply breaker to A-I would open on ES Channel #1 actuation and would result a reactor trip due to a loss of all Main Feedwater.

the A is incorrect, although the EDG would start with a reactortrip the EOP would have priority over securing EDG inadvertant B is incorrect, although HPI would occur the ESAS EOP would not be utilized to secure HPI for an actuation.

D is incorrect, seal return would be realigned to the Quench Tank rather than isolate.

References:

STM 1-32 Rev 33 OP-l 107.001 Change 073 History:

New selected for 2010 SRO exam

Electrical Distribution STMI-32 Re .33 RI S BUS TRIP ON MANUAl LOCK-OIJ1 GENERATOR ACTUATION UNDER MANUAL MANUAL ClOSE. TRIP LOCKOUT TRIP VOLTAGE TRIP TRANSFER TO SU- I TO SU-2 I CLOSES WHEN 1 SPRING REfl RN I TO NORMAL CLOSET)

TEST _l-_ CONTROL +/-CONTROT. SWITCL REMOFE SWITCH -

113(SI. I) lll(SU2)

SWiTCH 186-Al J86Gl-1/l_2

_L CONTROL. (CLOSES ON _1_

- CLOSEON SWI FCK Al BUS LOCK-OUT) ThENERATOT

_l LocK-our) 127-Al SYNC ESX-A3 SELECT CLOSES ON - Xl SYNC SWITCH ESCH.I CLOSES ON ,-

SELECTOR CH.2 FOR BUS Al SWITCH 286-G 1-2 x BKR 212 UNDER VOLTAGE 75%

7 OPEN ON GbNERA1 01 LOCK-OUT) 186-Al 152-113 CLOSE WI-lEN 152-Ill OPEN ON a m BREAKERS Al BUS I CLOSE LOCK-OUT CLOSED IN REMOTE CLOSED IN REMOTE FTGIJRE 32.62: UNIT AUXILIARY TRANSFORMFR I-I N)LR BRK 112(212) 183

ATTACHMENT B Page 1 of 3 Date 4160V SWITCHGEAR (NON-ES) CHECKLIST WARNING could lead to injury or Operating a breaker without knowing the consequences equipment damage. U NOTE

• This attachment assumes Unit 1 in Mode 5 or 6.

breaker position will

• Use of the local breaker status light to determine also indicate control power availab ility.

1.0 Check each listed breaker for the following:

• Breaker in desired position.

• Breaker control power on.

• Breaker racked up (except where specified Racked Down)

• Breaker control selector in REMOTE. (A-116 and A206 may be in either local or remote as desired)

• Breaker labeled properly.

position on 1.1 Log any breaker that is danger tagged or not in desired Lineup Exception Sheet (E-doc 1015.O O1F).

1.2 Notify plant labeling of any label discrepancies.

4160V Bus Al DESIRED ACTUAL TAG INI BREAKER DESCRIPTION POSITION POSITION (v) TIAL NUMBER A-113 Startup Xfmr #1 Feed to Al (E9l)

A-112 Unit Auxiliary Xfmr Feed to Al (E-90) Open A-ill Startup Xfmr #2 Feed to Al (E92) -

A-lb Circ Water Pump P3A (E27l)

A-109 Circ Water Pump P3C (E27l) -

A-l08 Main Chiller VCHlA (E372) Closed A-107 Heater Drain Pump P8A (E-304) Open A-106 Condensate Pump P2C (E-306) --

0 C) r C -

I I I I 0)

I I I rt I- I- I-* I- I- I- Ii 0 0 CD I- I- I- 0 0 O 0 . I) W . 0 0 o

• C) z CD CD C)

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CD I-I CD CD CD 0- 0- 0- (0 C) 0) rt rt rI r1 rt 0 0 0 0 0 CD C, I-a.

m x II x x x C CD I- I-t1 91 C) l- 91 2:1 C C C I-C C),,

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0

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0811 Rev: 0 Rev Date: 9/24/2009 Source: New Originator: S Pullin TUOI: AILP-RO-EFIC Objective: 43 Point Value: I Section: 3.4 Type: Heat Removal from Reactor Core System Number: 061 System

Title:

Auxiliary / Emergency Feewater System

==

Description:==

Knowledge of limiting conditions for operations and safety limits KIA Number: 2.2.22 CFR

Reference:

41.5 / 43.2 / 45.2 Tier: 2 RO Imp: 4.0 RO Select: No Difficulty: 3 Group: I SRO Imp: 4.7 SRO Select: Yes Taxonomy: C Question: RO:j SRO:J 89 Given

- A SG Low level transmitter feeding the D EFIC Channel failed Lo

- B SG Pressure transmitter feeding the C EFIC Channel failed Hi What operator actions are required per Technical Specifications?

A. Place D channel in bypass per 3.3.11 .A B. Place C channel in bypass per 3.3.11 .B C. Trip D channel per 3.3.11.B D. Trip C channel per 3.3.11 .A Answer:

B. Place C channel in bypass per 3.3.11 .B Notes:

B is correct, the Low Level transmitter failing low will result in a trip of the D Channel, 3.3.11 .B requirements for two inoperable channels requires one to be placed in bypass and the other one tripped.

A is incorrect, D Channel is already tripped and placing in bypass would have no effect. TS 3.3.11 .A is only applicable to one inoperable channel. The question asks what to do for two inoperable channels C is incorrect, because it is only half of the action required by 3.3.11 .B D is incorrect because tripping C Channel would result in an EFIC actuation.

References:

TS 3.3.11 Amendment 215 History:

New selected for 2010 SRO exam

EFIC System Instrumentation 3.3.11 3.3 INSTRUMENTATION 3.3.11 Emergency Feedwater Initiation and Control (EFIC) System Instrumentation LCO 3.3.11 The EFIC System instrumentation channels for each Function in Table 3.3.11-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.11-1.

ACTIONS Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Emergency A.1 Place channel(s) in bypass 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Feedwater (EFW) Initiation or trip.

or Main Steam Line Isolation Functions listed in Table 3.3.11-1 with one channel inoperable.

B. One or more EFW Initiation B.1 Place one channel in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or Main Steam Line Isolation bypass.

Functions listed in Table 3.3.11-1 with two channels Q inoperable.

B.2 Place second channel in trip. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> C. One EFW Vector Valve C.1 Restore channel to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Control channel inoperable. OPERABLE status.

D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not metfor Function I .b.

D.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ANO-1 3.3.11-1 Amendment No. 215

EFIC System Instrumentation 3.3.11 CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action and E.1 Reduce THERMAL 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time POWER to 10% RTP.

not met for Function l.a or I .d.

F. Required Action and F.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not metfor Functions I .c, 2, or 3.

F.2 Reduce steam generator 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> pressure to < 750 psig.

SURVEILLANCE REQUIREMENTS

-NOTE Refer to Table 3.3.1 1-1 to determine which SRs shall be performed for each EFIC Function.

SURVEILLANCE FREQUENCY SR 3.3.11.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Perform CHANNEL FUNCTIONAL e t TEST. 0 &2) 31 days SR 3.3.11.2 I &2)

SR 3.3.11.3 Perform CHANNEL 18 months The following notes apply only to the SG Level Low function:

Note 1: If the as-found channel setpoints are conservative with respect to the Allowable Value but outside their predefined as-found acceptance criteria band, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service, If the as-found instrument channel setpoints are not conservative with respect to the Allowable Value, the channel shall be declared inoperable.

Note 2: The instrument channel setpoint(s) shall be reset to a value that is equal to or more conservative than the Limiting Trip Setpoint; otherwise, the channel shall be declared inoperable. The Limiting Trip Setpoint and the methodology used to determine the Limiting Trip Setpoint and the predeflned as-found acceptance criteria band are specified in the Bases.

ANO-1 3.3.11-2 Amendment No. 24,227

EFIC System Instrumentation 3.3.11 Table 3.3.11-1 Emergency Feedwater Initiation and Control System Instrumentation APPLICABLE MODES OR FUNCTION OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE CONDITIONS CHANNELS REQUIREMENTS VALUES

1. EFW Initiation

a. LossofMFWPumps 10%RTP 4 SR3.3.11.1 55.5psig (Control Oil Pressure) SR 3.3.11.2 SR 3.3.11.3

b. SG Level Low- 1,2,3 4 perSG SR 3.3.11.1 9.34 inches SR 3.3.11.2 SR 3.3.11.3

c. SGPressure-Low 1,2,3(a) 4perSG SR3.3.11.1 584.2psig SR 3.3.11.2 SR 3.3.11.3

d. RCP Status 10% RTP 4 SR 3.3.11.1 NA SR 3.3.11.2

2. EFW Vector Valve Control

a. SG Pressure Low 1,2,3 4 perSG SR 3.3.11.1 584.2 psig SR 3.3.11.2 SR 3.3.11.3

b. SG Differential 1,2,3(a) SR 3.3.11.1 150 psid PressureHigh SR 3.3.11.2 SR 3.3.11.3

3. Main Steam Line Isolation

a. SG Pressure Low 4 per SG SR 3.3.11.1 > 584.2 psig SR 3.3.11.2 SR 3.3.11.3 (a) When SG pressure 750 psig.

(b) Except when all associated valves are closed and deactivated.

(c) The SG Level Low Limiting Trip Setpoint in accordance with NRC letter dated September 7, 2005, Technical Specification For Addressing Issues Related To Setpoint Allowable Values, is 10.42 inches.

(d) Includes an actuation time delay of 10.4 seconds.

ANO-1 3.3.11-3 Amendment No. 24,227

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0810 Rev: 0 Rev Date: 9/23/2009 Source: New Originator: S. Pullin TUOI: Al LP-RO-TS Objective: 5 Point Value: 1 Section: 2 Type: Generic Knowledge and Abilities System Number: 063 System

Title:

DC Electrical Distribution for Technical

==

Description:==

Ability to recognize system parameters that are entry-level conditions Specifications KIA Number: 2.2.42 CFR

Reference:

41 .7/41.10/43.2/43.3/45.3 2 RO Imp: 3.9 RO Select: No Difficulty: 3 Tier:

Group: I SRO Imp: 4.6 SRO Select: Yes Taxonomy: C Question: RO:i SRO:j Which of the following conditions requires entry into Technical Specification 3.8.4, DC Sources, Operating and what is the bases for Technical Specification 3.8.4?

A. DO4A, Battery Charger inoperable and D06, Battery operable.

Bases is to insure reactor coolant pressure boundary limits are not exceeded as a result of abnormalities B. 0048, Battery Charger inoperable and DO3B, Battery Charger inoperable.

Bases is to insure reactor coolant pressure boundary limits are not exceeded as a result of abnormalities C. DO4A, Battery Charger inoperable and DO4B, Battery Charger inoperable.

Bases is to insure adequate core cooling is provided, and reactor building operability and other functions are maintained in the event of a postulated DBA D. DO3B, Battery Charger inoperable and D07, Battery operable.

Bases is to insure adequate core cooling is provided, and reactor building operability and other functions are maintained in the event of a postulated DBA Answer:

C. DO4A, Battery Charger inoperable and DO4B, Battery Charger inoperable.

Bases is to insure adequate core cooling is provided, and reactor building operability and other functions are maintained in the event of a postulated DBA Notes:

tem is inoperable C is correct, with both battery chargers on the same train being inoperable, the subsys bases for TS 3.8.4 is to insure adequa te core cooling is provided, and requiring entry into TS 3.8.4. The postulated DBA reactor building operability and other functions are maintained in the event of a A is incorrect, Only one of the two charge s being inoperable does not affect the operability of the subsystem.

The bases used for this option is partially correct.

trains they do not affect the B is incorrect, two battery chargers are inoperable but since they are on different operability of either subsystem. The bases used for this option is partially correct.

operability of the subsystem.

D is incorrect, Only one of the two charges being inoperable does not affect the The bases used for this option is partially correct.

References:

T.S. 3.8.4 Amendment 215 History:

New selected for 2010 SRO exam

DC Sources Operating B 3.8.4 LCO The DC electrical power subsystems, each subsystem consisting of one battery, one of two battery chargers and the corresponding control equipment and interconnecting cabling supplying power to the associated bus within the train are required to be OPERABLE to ensure the availability of the required power to shut down the reactor and maintain it in a safe condition after an abnormality or a postulated DBA. Loss of any train DC electrical power subsystem does not prevent the minimum safety function from being performed (Ref. 4).

An OPERABLE DC electrical power subsystem requires the associated battery to be OPERABLE and connected to the associated DC bus and one of its respective chargers to be operating and connected to the associated DC bus.

APPLICABILITY The DC electrical power sources are required to be OPERABLE in MODES 1, 2, 3, and 4 to ensure safe unit operation and to ensure that:

a. Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of abnormalities; and

b. Adequate core cooling is provided, and reactor building OPERABILITY and other vital functions are maintained in the event of a postulated DBA.

The DC electrical power requirements for MODES 5 and 6 are addressed by the definition of OPERABILITY for each required supported load.

ACTIONS A.1 Condition A represents one train with a loss of ability to completely respond to an event, and a potential loss of ability to remain energized during normal operation. It is therefore imperative that the operators attention focus on stabilizing the unit, minimizing the potential for complete loss of DC power to the affected train. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> limit is consistent with the allowed time for an inoperable DC distribution system train.

If one of the required DC electrical power subsystems is inoperable (e.g.,

inoperable battery, inoperable battery chargers, or inoperable battery chargers and associated inoperable battery), the remaining DC electrical power subsystem has the capacity to support a safe shutdown and to mitigate an accident condition. Since a subsequent worst-case single failure would, however, result in the complete loss of the remaining 125 VDC electrical power subsystems with attendant ANO-1 B 3.8.4-3 Amendment No. 215

DC Sources Operating 3.8.4 3.8 ELECTRICAL POWER SYSTEMS 3.8.4 DC Sources Operating LCO 3.8.4 Both DC electrical power subsystems shall be OPERABLE.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One DC electrical power A.1 Restore DC electrical 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> subsystem inoperable, power subsystem to OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Associated Completion Time not met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE_REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify battery terminal voltage is 124.7 Von float 7 days charge.

SR 3.8.4.2 Verify battery capacity is adequate to supply, and 18 months maintain in OPERABLE status, the required emergency loads for the design duty cycle when subjected to a battery service test or a modified performance discharge test.

ANO-1 3.8.4-1 Amendment No. 215

SRO Written Exam Tier 2 Group 2

PWR Examination Outline Form ES-401-2 ES-401 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems Tier 2/Group 2 (RO)

Type System #1 Name K K K K K K A A A A G K/A Topic(s) IR # QID 1234561234 001 Control Rod Drive Not selected N1A 002 Reactor Coolant Not selected N!A 011 Pressurizer Level Control Not selected NIA 014 Rod Position Indication Not selected N/A 015 Nuclear Instrumentation Not selected NIA X 2.2.40 Ability to apply 4.7 91 599 D 016 Non-nuclear Instrumentation technical specifications for a system 017 In-core Temperature Monitor Not selected N/A 027 Containment Iodine Removal Not selected N/A 028 Hydrogen Recombiner 2.4.23 Knowledge of the N/A and Purge Control bases for prioritizing emergency procedure implementation during emergency operations.

Rejected system replaced with 016 Non-Nuclear Instrumentation 029 Containment Purge Not selected NIA 033 Spent Fuel Pool Cooling Not selected NIA 034 Fuel Handling Equipment X 2.1.40 Knowledge of 3.9 92 600 D refueling administrative requirements.

035 Steam Generator X A2.01 Faulted or ruptured 4.6 93 813 N S/Os.

041 Steam Dump/Turbine Not selected N/A Bypass Control 045 Main Turbine Generator Not selected N/A 055 Condenser Air Removal Not selected NIA 056 Condensate Not selected N/A 068 Liquid Radwaste Not selected NIA 071 Waste Gas Disposal Not selected NIA 072 Area Radiation Monitoring Not selected N/A 075 Circulating Water Not selected N/A 079 Station Air Not selected N/A 086 Fire Protection Not selected NIA K/A Category Point Totals: [ 2 Group Point Total: 3 ES-401 Form ES-401 -2

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0599 Rev: 0 Rev Date: 6/27/05 Source: Direct Originator: J.Cork TUOI: AILP-RO-NNI Objective: 35 Point Value: I Section: 3.7 Type: Instrumentation System Number: 016 System

Title:

Non-Nuclear Instrumentation

==

Description:==

Ability to apply technical specifications for a system.

K!A Number: 2.2.40 CFR

Reference:

41.10 /43.2 I 43.5 /45.3 Tier: 2 RO Imp: 3.4 RO Select: No Difficulty: 4 Group: 2 SRO Imp: 4.7 SRO Select: Yes Taxonomy: Ap Question: RO:j SRO:r 91 REFERENCE PROVIDED The plant is operating at 100% power.

Both PZR level transmitters LT-1 001 and LT-1 002 have failed LOW.

Which of the following actions are required by Technical Specification 3.3.15 and Table 3.3.15-1?

A. Be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

B. Both channels must be restored within 7 days.

C. Restore one channel to operable status within 30 days or be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

D. Restore one channel to operable status within 7 days or be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Answer:

D. Restore one channel to operable status within 7 days or be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Notes:

Answer D is correct in accordance with Table 3.3.15-1 and actions C and E.

Answer A is incorrect, there is still an allowance of 7 days per action C.

Answer B is incorrect, only one channel must be restored.

Answer C is inocrrect, this is a combination of A and E.

References:

T.S. 3.3.15 Amendment 232 Note: T.S. 3.3.15 must be in students handout.

History:

Direct from regular exam bank QID#ANO-OPSI -6623 Selected for 2005 SRO exam.

Selected for 2010 SRO exam.

PAM Instrumentation 3.3.15 3.3 INSTRUMENTATION 3.3.15 Post Accident Monitoring (PAM) Instrumentation LCO 3.3.15 The PAM instrumentation for each Function in Table 3.3.15-1 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS NOTE-Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions A.1 Restore required channel to 30 days with one required channel OPERABLE status.

inoperable.

B. Required Action and B.1 Initiate action to prepare Immediately associated Completion and submit a Special Time of Condition A not Report.

met.

C. One or more Functions C.1 Restore one channel to 7 days with two required channels OPERABLE status.

inoperable.

D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition C not Table 3.3.15-1 for the met. channel.

ANO-1 3.3.15-1 Amendment No. 24,222,232

PAM Instrumentation 3.3.15 CONDITION REQUIRED ACTION COMPLETION TIME E. As required by Required E.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action 0.1 and referenced in Table 3.3.15-1. AND E.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> F. As required by Required F.1 Initiate action to prepare Immediately Action D.1 and referenced and submit a Special in Table 3.3.15-1. Report.

SURVEILLANCE REQUIREMENTS

---

NOTE- -____

These SR5 apply to each PAM instrumentation Function in Table 3.3.15-1.

SURVEILLANCE FREQUENCY SR 3.3.15.1 Perform CHANNEL CHECK for each required 31 days instrumentation channel that is normally energized.

SR 3.3.15.2 --------------------NOTE-----------------------------

Neutron detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION. 18 months ANO-1 3.3.15-2 Amendment No. 245,222

PAM Instrumentation 3.3.15 Table 3.3.15-1 Post Accident Monitoring Instrumentation CONDITIONS REFERENCED FROM FUNCTION REQUIRED CHANNELS REQUIRED ACTION D.1

1. Wide Range Neutron Flux 2 E

2. RCS Hot Leg Temperature 2 E

3. RCS Hot Leg Level 2 F

4. RCS Pressure (Wide Range) 2 E

5. Reactor Vessel Water Level 2 F

6. Reactor Building Water Level (Wide Range) 2 E

7. Reactor Building Pressure (Wide Range) 2 E

8. Penetration Flow Path Automatic Reactor 2 per penetration flow E Building Isolation Valve Position path

9. Reactor Building Area Radiation (High Range) 2 F

10. Deleted

11. Pressurizer Level 2 E

12. a. SG A Water Level Low Range 2 E

b. SG B Water Level Low Range 2 E

c. SG UA Water Level High Range 2 E

d. SG B Water Level High Range 2 E

13. a. SGAPressure 2 E

b. SG B Pressure 2 E

14. Condensate Storage Tank Level 2 E

15. Borated Water Storage Tank Level 2 E

16. Core Exit Temperature (CETs per quadrant) 2 E

17. a. Emergency Feedwater Flow to SG A 2 E

b. Emergency Feedwater Flow to SG B 2 E

18. High Pressure Injection Flow 2 E

19. Low Pressure Injection Flow 2 E

20. Reactor Building Spray Flow 2 E (a) Not required for isolation valves whose associated penetration is isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.

(b) Only one position indication channel is required for penetration flow paths with only one installed control room indication channel.

ANO-1 3.3.15-3 Amendment No. 24&,222

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0600 Rev: 0 Rev Date: 6/27/05 Source: Direct Originator: J.Cork TUOI: AILP-RO-FH Objective: 16 PointValue: I Section: 3.8 Type: Plant Service Systems System Number: 034 System

Title:

Fuel Handling Equipment

==

Description:==

Knowledge of refueling administrative requirements.

KIA Number: 2.1.40 CFR

Reference:

41.10/43.5/45.13 Tier: 2 RO Imp: 2.8 RO Select: No Difficulty: 3 Group: 2 SRO Imp: 3.9 SRO Select: Yes Taxonomy: C Question: RO:1 SRO:E Given:

- Plant is in a Refueling outage.

- Core re-load is in progress.

- Approximately 90% of the core is in the Reactor vessel.

The Main Fuel Handling Bridge has a once-burned fuel assembly and is in the process of indexing over the specified core location when Nl-502 fails to 0.1 cps.

What action should be taken?

A. No action necessary because with NI-501 operating, Tech Spec NI requirements for operablility are met.

B. Contact the Main Fuel Bridge operator and place the assembly in a core location without any adjacent fuel assemblies.

C. Halt operations on the Main Fuel Bridge. Core geometry cannot be changed unless two neutron flux monitors are operable.

D. Verify boron concentration in the Refueling Canal is greater than 2326 ppm and then continue fuel load.

Answer:

C. Halt operations on the Main Fuel Bridge. Core geometry cannot be changed unless two neutron flux monitors are operable.

Notes:

Answer C is correct per 1502.004, 5.3, and T.S. 3.9.2 Answer A is incorrect, although only one is required in Mode 6, two NIs are required during core alterations.

Answer B is incorrect, this is still a core alteration.

Answer D is incorrect, this is simply a requirement for refueling.

References:

1502.004, Chg. 041 T.S. 3.9.2 Amendment 215 History:

Direct from regular exam bank QID#3178 Selected for 2005 SRO exam.

Selected for 2010 SRO exam

4.3 NRC Commitments 4.3.1 P 205, Response to NRC Bulletin 8903, Fuel in temporary core locations shall not reduce the shutdown margin below minimum required limit. Contained in Limits and Precautions and Initial Conditions sections.

4.3.2 P 9071, Emphasize housekeeping requirements. Contained in Limits and Precautions, and Initial Conditions sections.

4.3.3 P 12369, Caution tag source range power supplies.

Contained in Initial Conditions section.

4.3.4 P 12368, Record neutron count rate with each fuel assembly.

Contained in Instructions sections.

4.3.5 P 12366, Deviations from the fuel shuffle sequence require approval of SRO in Charge of Fuel Handling and Reactor Engineer. Contained in Limits and Precautions and Instructions sections.

4.3.6 P 14883, Ensure core offloads are performed after sufficient time for decay of fuel heat load, or when lake temperature is in range to assure existing SFP design temperature limits are not exceeded. Contained in Limits and Precautions, and in Initial Conditions sections.

5.0 LIMITS AND PRECAUTIONS 5.1 During movement of any fuel assemblies within the reactor building, radiation levels shall either be monitored by RE-80l7 or applicable TRN 3.9.1 Condition has been entered and the Required Action to place a portable survey instrument of appropriate range and sensitivity inservice have been performed. (TRM 3.9.1).

5.2 During movement of any fuel assemblies within the auxiliary building, radiation levels shall either be monitored by RE8009 or applicable TRM 3.9.2 Condition has been entered and the Required Action to place a portable survey instrument of appropriate range and sensitivity inservice have been performed. (TRM 3.9.2).

5.3 One source range neutron flux monitor shall be operable in Mode 6.

Two source range neutron flux monitors shall be operable during core alterations (TS 3.9.2).

5.4 One decay heat removal loop shall be operable and in operation in Mode 6 with water level 23 feet above the top of the irradiated fuel seated in the reactor pressure vessel. Refer to TS 3.9.4 for contingencies and exceptions.

Nuclear Instrumentation 3.9.2 3.9 REFUELING OPERATIONS 3.9.2 Nuclear Instrumentation LCO 3.9.2 a. One source range neutron flux monitor shall be OPERABLE, and

b. One additional source range neutron flux monitor shall be OPERABLE during CORE ALTERATIONS.

APPLICABILITY: MODE 6.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required source range A.1 Suspend CORE Immediately neutron flux monitor ALTERATIONS.

inoperable during CORE ALTERATIONS.

A.2 Suspend operations that Immediately would cause introduction into the RCS, coolant with boron concentration less than required to meet the boron concentration of LCO 3.9.1.

B. No OPERABLE source B.1 Initiate action to restore one Immediately range neutron flux monitor, source range neutron flux monitor to OPERABLE status.

AND B.2 Perform SR 3.9.1.1. Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ANO-1 3.9.2-1 Amendment No. 215

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0813 Rev: 0 Rev Date: 9/24/2009 Source: New Originator: S Pullin TUOI: Al LP-RO-EOPO6 Objective: 4 Point Value: I Section: 3.4 Type: Heat Removal from Reactor Core System Number: 035 System

Title:

Steam Generator System (S/GS)

==

Description:==

Ability to (a) predict the impacts of the following malfunctions or operations on the S/G and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Faulted or ruptured S/Gs.

KIA Number: A2.0l CFR

Reference:

41.5 / 43.5 / 45.3 / 45.5 Tier: 2 RO Imp: 4.5 RO Select: No Difficulty: 4 Group: 2 SRO Imp: 4.6 SRO Select: Yes Taxonomy: An Question:

RO:j SRO:I Given:

- Plant at 100% power Simultaneously the following occurs:

- Reactor trips on low RCS Pressure

- N-16 alarm on A Steam Generator

- Steam Line High Range Radiation monitor Rl-2681 in alarm.

- RCS pressure drops to 1300 psig

- CETs indicate 550°F

- Reactor Building and Aux Building sump levels are stable.

Starting with 1202.001, Reactor Trip EOP, which of the following lists the order of EOPs to mitigate this event?

A. 1202.002 Loss of Subcooling Margin and 1202.006 Tube Rupture B. 1202.002 Loss of Subcooling Margin and 1202.010 ESAS C. 1202.006 Tube Rupture and 1202.010 ESAS D. 1202.006 Tube Rupture and 1202.012 RT-10 Answer:

A. 1202.002 Loss of Subcooling Margin and 1202.006 Tube Rupture Notes:

A is correct, The Reactor Trip EOP immediate actions will send the operator to Loss of Subcooling margin, with the only LOCA being a tube rupture the Loss of Subcooling Margin procedure will send the operator to Tube Rupture.

B is incorrect, ESAS would only be entered if RCS pressure dropped below 150 psig.

C and D are incorrect, Reactor Trip would send the operator to Loss of Subcooling Margin EOP first.

References:

OP-1202.001 Change 031 OP-1202.002 Change 006 History:

New selected for 2010 SRO exam

CHANGE 1202.001 REACTOR TRIP 031 PAGE 4 of 25 INSTRUCTIONS CONTINGENCY ACTIONS Check adequate SCM. 3. Check elapsed time since loss of adequate SCM AND perform the following:

A. IF 2 minutes have elapsed, THEN trip all RCPs.

B. IF >2 minutes have elapsed, THEN leave currently running RCPs on.

C. Advise Shift Manager to implement Emergency Action Level Classification (1903.010).

D. GO TO 1202.002, LOSS OF SUBCOOLING MARGIN procedure.

4. Advise Shift Manager to implement Emergency Action Level Classification (1903.010).

5. Reduce Letdown by closing Orifice Bypass (CV-1 223).

6. Open BWST Outlet to OP HPI pump (CV-1407 or 1408).

7. J.[ Emergency Boration is I in progress, THEN adjust Pressurizer Level Control setpoint to 100.

CHANGE I 006 PAGE 6 of 17 L1202.002 LOSS OF SUBCOOLING MARGIN INSTRUCTIONS CONTINGENCY ACTIONS

8. Check SG tube integrity: 8. jf CET SCM is adequate, OR A. None of the following rad monitor no other LOCA indications exist indications rising in alarm: (RB and Aux Bldg sump levels are stable),

THEN GO TO 1202.006, TUBE RUPTURE

• Main Condenser (Rl-3632) procedure.

• OTSG N-16 Gross (RI-2691 and 2692)

• Steam Line High Range (Rl-2681 and 2682).

B. No report from Nuclear Chemistry that SG tube leak exists.

C. rise in unidentified RCS leakage accompanied by:

• Higher than expected SG level

• Lower than expected FW flow rate

9. IE CET SCM is adequate, THEN control RCS press low within limits of Figure 3 (RT 14).

10. Check RCS press remains 150 psig. 10. if RCS press is <150 psig, THEN GO TO 1202.01 0, ESAS procedure.

11. Check SG levels at or approaching one of 11. Re-verify proper EFW actuation and control the following: (RT5).

SCM adequate I SCM <adequate I A. jf all MFW and EFW is lost I 300 to 340 I 370 to 410 I AND either of the following conditions is met, THEN GO TO 1202.004, OVERHEATING procedure.

• CET SCM adequate

• CETtemps610°F

SRO Written Exam Tier 3

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INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0492 Rev: 1 Rev Date: 1214/06 Source: Direct Originator: S.Pullin TUOI: Al LP-RO-EOPO8 Objective: 7 Point Value: I Section: 2.0 Type: Generic Knowledges and Abilities System Number: 2.1 System

Title:

Conduct of Operations operating

==

Description:==

Ability to evaluate plant performance and make operational judgments based on characteristics, reactor behavior, and instrument interpretation.

K/A Number: 2.1.7 CFR

Reference:

41.5/43.5/45.12/45.13 Tier: 3 RO Imp: 4.4 RO Select: No Difficulty: 4 Group: SRO Imp: 4.7 SRO Select: Yes Taxonomy: An Question: RO:1 SRO:l Given the plant conditions following a reactor trip:

- RCS temperature: 605 degrees stable

- RCS pressure: 2300 psig slowly dropping

- ERV: open in AUTO

- OTSG shell temperature: 558 degrees

- OTSG levels 20 inches, steady

- PZR level 180 inches, rising Which of the following actions are required?

A. Trip the running RCP per 1202.002, Loss of Subcooling Margin.

B. Isolate the ERV per 1202.001, Reactor Trip.

C. Select the reflux boiling setpoint per RT-5.

D. Initiate Full HPI per RT 3.

Answer:

B. Isolate the ERV per 1202.001, Reactor Trip.

Notes:

RCS pressure Answer B is correct. A pressurizer steam space leak is indicated by PZR level rising with dropping and no rise in RCS temperature. ERV is open and should have dosed at 2395 psig.

Answer A is incorrect, Tube to Shell delta T of 60 degrees tubes hotter would require this action however the delta T is only 47 degrees in the question.

Answer C is incorrect, although RCS temperature/pressure conditions are close to a loss of subcooling margin which would require selection of Reflux Boiling but SCM is still adequate.

EOP in Answer D is incorrect, Full HPI would be required if the ERV opened in Auto with the Overheating effect but the Overheating entry conditions are not met.

References:

1202.001, Chg. 031 History:

Modified from regular exambank QID#3314.

Used on 2004 SRO Exam.

Modified for use on 2007 SRO Exam.

Selected for 2010 SRO exam.

CHANGE 1202.001 REACTOR TRIP 031 PAGE 21 of 25 INSTRUCTIONS CONTINGENCY ACTIONS

28. Verify ERV, Pressurizer Spray, and 28. Perform the following:

Pressurizer Heaters operate to control RCS press 2050 to 2250 psig. A. IF ERV is open in AUTO AND RCS press <2395 psig, THEN verify ERV Isolation closed (CV-I 000).

B. IF Pressurizer Spray valve is open in AUTO AND RCS press <2155 psig, THEN close Pressurizer Spray Isolation (CV-1 009).

C. IF Pressurizer Heaters fail to operate in AUTO, THEN operate Heaters manually to control RCS press 2050 to 2250 psig.

29. Check at least one RCP running. 29. Perform the following:

A. Verify proper EFW actuation and control (RT 5).

B. IF HI or H2 is energized with normal voltage (6900V)

AND CET SCM is adequate AND RCPs are available, THEN perform the following:

1) Start one RCP in each loop (RT 11).

30. Check RCS T-cold remains > 540°F. 30. IF RCS T-cold is < 540°F AND dropping, THEN GO TO 1202.003, OVERCOOLING procedure.

31. Check adequate SCM. 31. GO TO 1202.002, LOSS OF SUBCOOLING MARGIN procedure.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0814 Rev: 0 Rev Date: 9/24/2009 Source: New Originator: S Pullin TUOI: AILP-RO-FH Objective: 4 Point Value: 1 Section: 2 Type: Generic Knowledge and Abilities System Number: 2.1 System

Title:

Conduct of Operations

==

Description:==

Knowledge of the fuel-handling responsibilities of SROs KIA Number: 2.1.35 CFR

Reference:

41.10/43.7 Tier: 3 RO Imp: 2.2 RO Select: No Difficulty: 3 Group: G SRO Imp: 3.9 SRO Select: Yes Taxonomy: K Question: RO: SRO: 95 Which of the following conditions would require the SRO in charge of fuel handling to order a stop to fuel movement in the Reactor Building?

A. Outage Control Center reports that the reactor has been subcritical for 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br />.

B. National Weather Service declares a Tornado Watch in effect for Conway County.

C. One Control Room Emergency Air Conditioning System (CREACS) inoperable for the past 5 days.

D. Reactor Building Radiation monitor RE-8017 inoperable, and portable survey instrument is being monitored on the fuel handling bridge.

Answer:

A. Outage Control Center reports that the reactor has been subcritical for 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br />.

Notes:

A is correct, the reactor must be subcritical for greater than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to fuel movement.

B is incorrect, Pope, Johnson, Yell and Logan counties in a tornado watch would require stopping fuel movement. Conway county is immediately east of Pope county.

C is incorrect, with one CREACS channel inoperable we have 30 days to repair prior to stopping fuel movement.

o is incorrect, RE-8017 is desired to be operable for monitoring radiation levels on the bridge, however if it becomes inoperable any portable survey instrument is allowed for monitoring rad levels and continue fuel movement.

References:

OP-I 502.004 Change 041 History:

New selected for 2010 SRO exam.

i PROC IWORK PLAN NO.

1502.004 I PROCEDUREIWORK PLAN TITLE:

CONTROL OF UNIT I REFUELING I PAGE: 6 of 52 CHANGE: 041 4.3 NRC Commitments 4.3.1 P 205, Response to NRC Bulletin 8903, Fuel in temporary core locations shall not reduce the shutdown margin below minimum required limit. Contained in Limits and Precautions and Initial Conditions sections.

4.3.2 P 9071, Emphasize housekeeping requirements. Contained in Limits and Precautions, and Initial Conditions sections.

4.3.3 P 12369, Caution tag source range power supplies.

Contained in Initial Conditions section.

4.3.4 P 12368, Record neutron count rate with each fuel assembly.

Contained in Instructions sections.

4.3.5 P 12366, Deviations from the fuel shuffle sequence require approval of SRO in Charge of Fuel Handling and Reactor Engineer. Contained in Limits and Precautions and Instruct ions sections.

4.3.6 P 14883, Ensure core offloads are performed after sufficient time for decay of fuel heat load, or when lake temperature is in range to assure existing SEP design temperature limits are not exceeded. Contained in Limits and Precautions, and in Initial Conditions sections.

5.0 LIMITS AND PRECAUTIONS 5.1 During movement of any fuel assemblies within the reactor building, radiation levels shall either be monitored by RE-8017 or applicable TRM 3.9.1 Condition has been entered and the Required Action to place a portable survey instrument of appropriate range and sensitivity inservice have been performed. (TRN 3.9.1).

5.2 During movement of any fuel assemblies within the auxiliary building, radiation levels shall either be monitored by RE8009 or applicable TRN 3.9.2 Condition has been entered and the Required Action to place a portable survey instrument of appropriate range and sensitivity in-service have been performed. (TRM 3.9.2) 5.3 One source range neutron flux monitor shall be operable in Mode 6.

Two source range neutron flux monitors shall be operable during core alterations (TS 3.9.2).

5.4 One decay heat removal loop shall be operable and in operation in Mode 6 with water level 23 feet above the top of the irradiated fuel seated in the reactor pressure vessel. Refer to TS 3.9.4 for contingencies and exceptions.

5.9 Refueling canal water level shall be maintained 23 feet above the top of the irradiated fuel assemblies seated within the reactor pressure vessel during movement of irradiated fuel assemblies within the reactor building (TS 3.9.6).

5.10 A minimum of 10 feet separation shall be maintained between fuel assemblies when two assemblies are moved simultaneously in the transfer canal (TRM 3.9.3).

5.11 Each required reactor building penetration shall be verified in the required status within 7 days prior to refueling operations and at least every 7 days thereafter (SR 3.9.3.1).

5.12 Irradiated fuel shall not be removed from the reactor until the unit has been subcritical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> (TRM 3.9.3).

5.13 In the event of a complete core offload, the decay heat load to be transferred to the Spent Fuel Pool shall be verified to be within the limits of the spent Fuel Pool Cooling System (TRM 3.7.3).

5.14 No tornado watches shall be in effect for Pope, Yell, Johnson, or Logan counties in Arkansas during movement of any fuel assemblies within the auxiliary building (TP.N 3.9.2) or the Reactor Building (ERANO20021078007 Rev. 0) 5.14.1 Upon issue of a tornado watch for any of these counties, enter TRM 3.9.2 Condition B, cease all fuel handling in the auxiliary building and Reactor Building. Fuel handling in progress will be completed to the extent necessary to place the fuel handling bridge and crane in their normal parked and locked position.

5.15 Loads in excess of 2000 pounds (such as the cask loading pit gate and tilt pit gate, 4000 lbs. each) shall not travel over fuel assemblies in the storage pool (TRN 3.7.2) 5.16 During movement of irradiated fuel assemblies, either two Control Room Emergency Ventilation System (CREVS) trains shall be operable, and one CREVS train shall be capable of automatic operation, or the applicable TS 3.7.9 Condition has been entered and the Required Actions of TS 3.7.9 have been performed. The control room boundary may be opened intermittently under adxnin controls (TS 3.7.9).

5.17 During movement of irradiated fuel assemblies, either two Control Room Emergency Air Conditioning System (CREACS) shall be operable or applicable TS 3.7.10 Condition has been entered and the Required Actions of TS 3.7.10 have been performed.

5.18 During movement of irradiated fuel assemblies, either two channels of Control Room Isolation High Radiation shall be operable or the applicable TS 3.3.16 Condition and the Required Actions of TS 3.3.16 (immediately place one Operable CREVS train in emergency recirculation mode) have been performed.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0646 Rev: 0 Rev Date: 10/23/200 Source: Direct Originator: Cork/Possage TUOI: AILP-RO-EDG Objective: 2 Point Value: 1 Section: 2 Type: Generic Knowledge and Abilities System Number: 2.2 System

Title:

Equipment Control

==

Description:==

Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.

K/A Number: 2.2.25 CFR

Reference:

41.5 /41.7 / 43.2 Tier: 3 RO Imp: 3.2 RO Select: No Difficulty: 3 Group: G SRO Imp: 4.2 SRO Select: Yes Taxonomy: C Question: RO:j SRO:J REFERENCE PROVIDED Given:

- #1 EDG has one Air Start Compressor and its associated Air Receiver Tanks tagged out.

- The remaining Air Start Compressor on #1 EDG trips while EDG is running for a surveillance.

- The Air Receiver Tanks pressure is 145 psig.

In accordance with Technical Specifications, what is the required action for the above conditions?

A. No actions are necessary since the EDG is running and an air start system is not needed.

B. Restore required starting air receiver pressure to within limits in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

C. Declare #1EDG inoperable immediately.

D. Be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Answer:

C. Declare #1EDG inoperable immediately.

Notes:

Answer C is correct, with only one receiver bank and pressure <158 psig the EDG must be declared inoperable per 3.8.3.E.1.

Answer A is incorrrect, although the EDG is running, if it tripped there would not be enough air for a re-start.

Answer 8 is incorrect, this is the action from 3.8.3.D and would be applicable if pressure was between 158 and 175 psig.

Answer D is incorrect, this action is from 3.8.1 .F and would be applicable if the EDG was not made operable within 7 days.

References:

3.8.3 and Bases Amendment 215 History:

Uses QID 447 stem with some modifications, all answers are different, therefore it is a new question.

New question for 2007 SRO exam.

Selected for the 2010 SRO exam

Diesel Fuel Oil and Starting Air 3.8.3 3.8 ELECTRICAL POWER SYSTEMS 3.8.3 Diesel Fuel Oil and Starting Air LCO 3.8.3 The stored diesel fuel oil and starting air subsystem shall be within limits for each required diesel generator (DG).

APPLICABILITY: When associated DG is required to be OPERABLE.

ACTIONS Separate Condition entry is allowed for each DG.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more DG fuel oil A.1 Restore fuel oil volume to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> storage tank(s) with fuel within limits.

volume < 20,000 gallons and> 17,140 gallons.

B. One or more DG5 with B.1 Restore fuel oil total 7 days stored fuel oil total particulates to within limits.

particulates not within limit.

C. One or more DGs with C.1 Restore stored fuel oil 30 days new fuel oil properties not properties to within limits.

within limits.

D. One or more DGs with D.1 Restore required starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> required starting air air receiver pressure to receiver pressure within limits.

< 175 psig and 158 psig.

ANO-1 3.8.3-1 Amendment No. 215

Diesel Fuel Oil and Starting Air 3.8.3 CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action and E.1 Declare associated DC Immediately associated Completion inoperable.

Time not met.

OR One or more DGs with diesel fuel oil or required starting air subsystem not within limits for reasons other than Condition A, B, C,orD.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.3.1 Verify each fuel oil storage tank contains 31 days 20,000 gallons of fuel.

SR 3.8.3.2 Verify fuel oil properties of new and stored fuel oil are In accordance with tested in accordance with, and maintained within the the Diesel Fuel Oil limits of, the Diesel Fuel Oil Testing Program. Testing Program SR 3.8.3.3 Verify each DC required air start receiver pressure is 31 days 175 psig.

SR 3.8.3.4 Check for and remove accumulated water from each 31 days fuel oil storage tank.

ANO-1 3.8.3-2 Amendment No. 215

Diesel Fuel Oil and Starting Air B 3.8.3 ACTIONS (continued)

B. I This Condition is entered as a result of a failure to meet the acceptance criterion of Specification 5.5.13. Normally, trending of particulate levels allows sufficient time to correct high particulate levels prior to reaching the limit of acceptability. Poor sample procedures (bottom sampling), contaminated sampling equipment, and errors in laboratory analysis can produce failures that do not follow a trend. Since the presence of particulates does not mean failure of the fuel oil to burn properly in the diesel engine, particulate concentration is unlikely to change significantly between Surveillance Frequency intervals, and proper engine performance has been recently demonstrated (within 31 days), it is pwdent to allow a brief period prior to declaring the associated DG inoperable. The 7 day Completion Time allows for further evaluation, resampling, and re-analysis of the DC fuel oil.

C.1 With the new fuel oil properties defined in the Bases for SR 3.8.3.2 not within the required limits, a period of 30 days is allowed for restoring the stored fuel oil properties. This period provides sufficient time to test the stored fuel oil to determine that the new fuel oil, when mixed with previously stored fuel oil, remains acceptable, or to restore the stored fuel oil properties. This restoration may involve feed and bleed procedures, filtering, or combinations of these procedures. Even if a DG start and load was required during this time interval and the fuel oil properties were outside limits, there is a high likelihood that the DC would still be capable of performing its intended function.

D.1 With starting air receiver pressure < 175 psig in the required receivers, sufficient capacity for five successive DC start attempts does not exist. However, as long as the receiver pressure is 158 psig, there is adequate capacity for at least one start attempt, and the DC can be considered OPERABLE while the air receiver pressure is restored to the required limit. A period of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is considered sufficient to complete restoration to the required pressure prior to declaring the DC inoperable.

This period is acceptable based on the remaining air start capacity, the fact that the credited DC start is accomplished on the first attempt, and the low probability of an event during this brief period.

E.1 With a Required Action and associated Completion Time not met, or one or more DGs with fuel oil or required starting air subsystem not within limits for reasons other than addressed by Conditions A through D, the associated DG may be incapable of performing its intended function and must be immediately declared inoperable.

ANO-1 8 3.8.3-3 Amendment No. 215

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0815 Rev: 0 Rev Date: 9/24/2009 Source: New Originator: S PulIin TUOI: ASLP-SRO-MNTC Objective: 2 Point Value: 1 Section: 2 Type: Generic K&A System Number: 2.2 System

Title:

Equipment Control

Description:

Knowledge of maintenance work order requirements KIA Number: 2.2.19 CFR

Reference:

41.10 / 43.5 / 45.13 Tier: 3 RO Imp: 2.3 RO Select: No Difficulty: 3 Group: G SRO Imp: 3.4 SRO Select: Yes Taxonomy: K Question: RO: SRO:F97 Given:

- Annunciator K12-B5, P-7A Turbine Trip alarms

- WCO reports that the linkage for the trip throttle valve has broken.

You are the Shift Manager, Per EN-WM-100, Work Request (WR) Generation, Screening and Classification, which work order process should be used to correct this condition.

A. Priority One Work Order B. Priority Two Work Order C. Tool Pouch Maintenance / No work order required D. FIN Team / No work order required Answer:

A. Priority One Work Order Notes:

A is correct, since P-7A inoperability is a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Time Clock, a Priority I work order would be initiated to begin maintenance and work around the clock to completion.

B is incorrect, Priority 2 work orders are entered into the T-3 week schedule and would not be urgent enough to meet the needs of the plant.

C is incorrect, Tool pouch maintenance is not allowed on safety related equipment even though the repairs are skill of the craft.

D is incorrect, FIN Team can not work on safety related equipment with out a work order even though the repairs are skill of the craft.

References:

EN-WM-100 Rev 3 History:

New selected for 2010 SRO exam

significance of the condition identified. For on line work, the priority is determined in accordance with Attachment 9.1. Each priority state is shown below along with guidelines for starting the work.

• Priority 1: Begin immediately following planning of the work order and work around the clock.

• Priority 2: Schedule at earliest opportunity within T-3.

• Priority 3: Schedule at next available system week within the 12 week process or next available system window.

• Priority 4: Schedule as resources allow within the normal process.

• Priority 5: Work only when time allows (fill in activity).

• Priority 8: Outage work where performance is mandatory (required I de-rate)

• Priority 9: Outage work where performance is discretionary (potential)

[10] Power Block Equipment All SSCs required for the safe and reliable operation of the station. It will include all safety-related and balance-of-plant system and components required for the operation of the station, including radioactive waste processing and storage, and switchyard equipment maintained by the station. Systems, structures, or components required to maintain federal or state regulatory compliance should be included in this grouping. This classification does not include buildings or structures that support station staff, such as offices or storage structures, or the HVAC and support systems focused only on habitability of those structures.

[11] Skill Of The Craft A task that workers are familiar with and experienced in performing, which are not complex in the actions required and are common to their craft.

Familiarity may have been gained through training or on the job performance. To perform the task safely and successfully, the worker would not require further instruction or oversight.

[12] Work Instructions A set of work steps included in a work package provided to direct how work is to be accomplished.

[13] Work Request Screening Committee The Work Request Screening Committee, chaired by Scheduling, meets each normal workday and reviews WRs contained on the work screening report (Attachment 9.5). The standard report is WEB based and includes work requests that have not been previously converted to work orders or approved as toolpouch, and includes all work requests generated since the previous meeting.

Scheduling will review the work screening report prior to the meeting and provide

• QUALITYREI.ATED EN-WM-100 REV. 3 NUCLEAR

- Entergy MANAGEMENT INFORMATIONAL USE PAGE 12 OF 28 MANUAL 5.2121 cont.

The individual discovering the deficiency, or another person, can repair it, if qualified to do so, Deficiency utilizing Toolpouch Maintenance, and if:

Identified

• the activity does not affect a safety related function

• there is no risk of a plant transient

• the activity does not require either a procedure, work instructions or material other than consumable

• the activity does not involve work on ASME Code or EQ equipment

• the activity does not alter plant configuration

• the activity is not complex and is within the skill of the personnel

• the activity does not affect a Maintenance Rule function

• the activity requires no additional support beyond that for normal plant access If the requirements for Toolpouch Maintenance are not met, the identifier should generate a WR in lAS. lAS required actions for Work Request Initiation, Operability Screening and Classification are contained in Attachment 9.6.

The identifier of the activity provides the following information:

Yes

• Originator name (Defaults in lAS)

No Further Action

• The date identified (Defaults in lAS)

Necessary

• identification of the component including Equipment Number

• A description of the deficiency (Attach. 9.3)

• Recommended solution, if known

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEARONE-UNITI QID: 0816 Rev: 0 Rev Date: 9/24/2009 Source: New Originator: S Pullin TUOI: Al LP-RO FH Objective: 4 Point Value: I Section: 2 Type: Generic Knowledges and Abilities System Number: 2.3 System

Title:

Radiation Control

==

Description:==

Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

KIA Number: 2.3.14 CFR

Reference:

41.12/43.4 / 45.10 Tier: 3 RO Imp: 3.0 RD Select: No Difficulty: 3 Group: G SRO Imp: 3.8 SRO Select: Yes Taxonomy: C Question: RO: SRO:I During a fuel handling accident Krypton-85 is the major source of gaseous activity released from a damaged Fuel assembly.

Which portion of the body will receive the highest dose after a fuel handling accident?

A. Skin dose from Beta B. Whole body dose from Gamma C. Extremities dose from Beta D. Internal Organ dose from Gamma Answer:

A. Skin dose from Beta Notes:

A is correct, skin dose rates from K-85 are 100 times higher than the whole body gamma dose rates.

B, C, and 0 are all incorrect.

References:

OP-I 203.042 Change 005-03-0 History:

New selected for 2010 SRO exam

CHANGE 1203.042 REFUELING ABNORMAL OPERATION 005-03-0 PAGE 5 of 10 SECTION 1 FUEL HANDLING ACCIDENT

2. E damage to a spent fuel assembly is suspected, THEN perform the following:

WARNING Krypton-85, a beta emitter, is the major source of gaseous activity released from a damaged spent fuel assembly that has decayed >190 days. Skin dose rates from Kr-85 are 100 times higher than the whole body, gamma dose rate. Instruments not sensitive to beta, such as self-reading dosimeters and survey meters with their beta windows closed, will read less than the actual values.

A. Direct RP personnel to proceed to the area and inform them of the beta hazard associated with a damaged spent fuel assembly.

B. Inspect the spent fuel assembly with all available means to determine if damage has occurred.

C. IF even slight spent fuel assembly damage is detected, THEN take actions for confirmed damage per this procedure.

D. IF fuel assembly is not damaged, THEN proceed as directed by Plant Management.

EN DISCUSSION A fuel handling accident has occurred when fuel handling equipment malfunctions or other occurrences result in damage to a spent fuel assembly. Until proven otherwise, it is assumed that one or more fuel pins are ruptured. This assumption is made regardless of how slight the damage to the fuel assembly(ies) appears.

Ruptured spent fuel cladding releases fission product gases in undetermined quantities to the pool water or possibly to atmosphere in case of dry fuel storage handling accident. A mechanical damage type accident is considered the maximum potential source of activity release during refueling operations.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT I QID: 0411 Rev: 0 Rev Date: 12/1/00 Source: Direct Originator: E-PIan TUOI: ASLP-RO EPLAN Objective: 7 Point Value: 1 Section: 2 Type: Generic Knowledges and Abilities System Number: 2.4 System

Title:

Emergency Procedures/Plan

==

Description:==

Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the state, the NRC, or the transmission system operator.

KIA Number: 2.4.30 CFR

Reference:

41.10 / 43.5 / 45.11 Tier: 3 RO Imp: 2.7 RO Select: No Difficulty: 2 Group: G SRO Imp: 4.1 SRO Select: Yes Taxonomy: C Question: RO:1 SRO:I A fire was reported at 0844 in the vicinity of the Old Radwaste Building.

It is now 0920 and the fire is still burning.

What is the Emergency Plan time requirement for notification to the NRC?

A. Notification to the NRC is required within 15 minutes of the declaration of an emergency class.

B. Notification to the NRC is required immediately following notification of the ADH and within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the declaration of an emergency class.

C. Notification to the NRC is required immediately following declaration of an emergency class and notify the ADH within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

D. Notification to the NRC is required within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of the declaration of an emergency class.

Answer:

B. Notification to the NRC is required immediately following notification of the ADH and within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the declaration of an emergency class.

Notes:

Answer [B] is correct since this is the procedural requirement.

Answer [A], [C], [D] are incorrect, these are not in accordance with 1903.011.

References:

1903.OIIY, Emergency Initial Notification Message Change 036 History:

Modified E-Plan exam bank QID#61 for use in 2001 SRO Exam.

Selected for use in 2002 SRO exam.

Selected for 2010 SRO exam

ARKANSAS NUCLEAR ONE Page 1 E-DOC TITLE: E-DOC NO. CHANGE NO.

EMERGENCY CLASS INITIAL NOTIFICATION MESSAGE 1903.O11-Y 036 ACTIONS FOR INITIAL NOTIFICATION he Arkansas Department of Health (ADH) SHALL be notified within 15 minutes of an:

• Emergency Class Declaration

• Emergency Class Change (Upgrade or Downgrade)

• PAR Change The Nuclear Regulatory Commission (NRC) SHALL be notified immediately following notification of the ADH and SHALL NOT exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> following the declaration of an emergency class.

ERDS must be started within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the declaration of an ALERT or higher emergency class.

NOTE

• The material contained within the symbols (*) throughout this form is proprietary or private information.

• The Emergency Telephone Directory contains the emergency telephone numbers that you may need to complete this notification.

• Computer generated Form 1903.011 -Y may be used for notifications. The computer generated form is not an identical copy to the hard copy form, but contains all necessary information.

INSTRUCTIONS (circle/slash) 1.0 Complete Initial Notification Message in accordance with Step 1.1 Computerized Notification Method OR Step 1.2 Manual Notification Method. Computerized Notification Method preferred.

1.1. Computerized Notification Method 1.1.1. IF the Computerized Notification Method fails while performing notifications, THEN go to the Manual Notification Method Step 1.2.

1.1.2. Sign onto the computerized notification system computer using your Entergy logon ID and password. Control Room may use a generic ID and password.

1.1.3. Verify your computer is connected to a local or network printer in your area.

[Start]-)[Settings]-)jPrinters and Faxes]

1.1.4. On the desktop double click the EP Notification icon OR select [Start], [(All) Programs], [EP Notifications], [EP Notifications Version X)(XX] to start notification program.

1.1.5. Enter the appropriate data into the data fields for the Initial Notification Message. Use the [Tab]

key (preferred) or mouse to navigate through the form. Refer to Emergency Class Notification Instructions page 7 of this form as needed.

1.1.6. WHEN the data fields are populated, THEN press the [Create PDF only] button.

1.1.7. IE you receive an error message (i.e. You have not correctly entered all the required data on Tab...),

THEN review the form and make corrections.

Go to Step 1.1.6 above.

1.1.8. WHEN the PDF notification message is displayed on the computer screen, THEN print the message to a local printer.

1.1.9. Give the notification message to the person with ED&C for review and approval.

1.1.10. Once approval has been obtained, then close the PDF notification message on the computer screen by pressing [X] in upper right hand corner of PDF document.

INITIAL ROISRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE UNIT I -

QID: 0750 Rev: 2 Rev Date: 6/23/08 Source: Direct Originator: Spullin TUOI: AILP-RO-AOP Objective: 5 Point Value: I Section: 2.0 Type: Generic KJAs System Number: 2.4 System

Title:

Emergency Procedures / Plan

==

Description:==

Knowledge of local auxiliary operator tasks during an emergency and the operational resultant effects K/A Number: 2.4.35 CFR

Reference:

41.10/43.5/45.13 Tier: 3 RO Imp: 3.8 RO Select: No Difficulty: 3 Group: SRO Imp: 4.0 SRO Select: Yes Taxonomy: C Question: RO:J SRO:J 00 Given:

- Severe Fire on 335 Auxiliary Building on Unit I

- Reactor has been tripped Which of the following actions would the CRS direct the Outside AO to perform and what procedural guidance would be used?

A. Fire fighting tasks per Fire or Explosion procedure 2203.034.

B. Securing Polishers per Reactor Trip/Outage Recovery procedure 1102.006.

C. Placing the Startup Boiler in service per Startup Boiler Operation procedure 1106.022.

D. Throttle CV-2627 EFW Supply to A SG per Fires in Areas Affecting Safe Shutdown procedure 1203.049 Answer:

D. Throttle CV-2627 EFW Supply to A SG per Fires in Areas Affecting Safe Shutdown procedure 1203.049 Notes:

A. is incorrect; due to recent procedures changes have the opposite Unit ACs fighting the fire, but the WCO non licensed operator has fire fighting duties B. is incorrect; under normal Reactor trip conditions this would be an Outside AC action promptly following Rx trip, but it is not in the Reactor procedure C. is incorrect; under normal Reactor trip conditions this would be an Outside AO would perform following Rxtrip D. is correct; this is a new procedure action for the non licensed operators

References:

1203.049 Fires in Areas affecting Safe Shutdown Change 005 History:

Selected for 2010 SRO exam

E CHANGE I 1203.049 FIRES IN AREAS AFFECTING SAFE SHUTDOWN 005 PAGE 139 of 251 Fire Area C (335 Aux Building)

Page 2 of 2 Outside AO Required Actions

5. LE directed by CBOR, THEN close MSIVs by manually opening the following valves:

• IA Vent to MSIV A (IA-2691 B)

• IA Vent to MSIV A (IA-2691C)

• IA Vent to MS IV A (lA-2691 D)

• IA Vent to MSIV A (IA-2691 E)

• IA Vent to MSIV B (lA-2692B)

• IA Vent to MSIV B (IA-2692C)

• IA Vent to MSIV B (IA-2692D)

• IA Vent to MSIV B (IA-2692E)

6. WHEN notified by CBOT that CV-1 405 is de-energized, THEN verify RB Sump Line A Outlet (CV-I 405) closed (A Decay Heat Vault).

A. Notify CBOR that CV-1405 is de-energized and closed.

7. WHEN directed by CBOR AND notified by CBOT that CV-2627 is de-energized, THEN locally close EFW P-7A to SG-A Isol (CV-2627) in UNPPR.

8. WHEN notified by CBOT that CV-I 220 is de-energized, THEN locally verify HPI to P-32D Discharge (CV-1220) is open (UNPPR).

A. Notify CBQR that CV-1 220 is de-energized and open.

9. WHEN directed by CRS, THEN throttle CV-2627 as directed (UNPPR).

10. WHEN notified by CBOT that CV-1 407 is de-energized, THEN verify BWST T-3 Outlet (CV-1407) open (behind Waste Gas Panel on 354 EL).

A. Notify CBOR that CV-1407 is de-energized and open.