ML12073A209

Initial Exam 2011-302 Draft SRO Written Exam 1 of 3

ML12073A209
Person / Time
Site:	Watts Bar
Issue date:	04/29/2011
From:	Operator Licensing and Human Performance Branch
To:	Tennessee Valley Authority
References
50-390/11-302
Download: ML12073A209 (134)

Text

WBN 10-2011 NRC SRO EXAM as Submitted 08/1512011 REFERENCE PACKAGE

1. Steam Tables

2. AOl-i 1, Appendix A, Condenser Vacuum ICS Graph, (1 page)

3. O-Sl-0-3, Weekly Log, Appendix A, (2 pages)

4. ICS AFD TARGET DISPLAY, (1 page)

5. ECA-1-1, Loss of RHR Sump Recirculation, (2 pages)

6. Tech Spec 3.6.12, Ice Condenser Doors (5 pages)

7. EPIP-1, Emergency Plan Classification Logic, (1 page)

8. AOl-30.2 APP B, Fire Safe Shutdown Elevation Diagrams 2.0 AB EL 772.0, 776.0 &

763.5 ELEVATION DIAGRAM (1 page)

W o A & B and C CONDENSER VACUUM LO-LO AND LO ALARMS cm OH o zr

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• 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 LOAD (MWe)

WBN Weekly Log 0-SI-0-3 Unit 0 -

Rev. 0045 Page 20 of 43 Appendix A (Page 9 of 24)

SR 3.8.4.3 SR 38 5 1 SR 3.8.7.1 SR 5.8.8.1 SR 3.8.9.1 NO (1OA)

SR 3.8.10.1 5-OLOSED N

A- AVAW_NBLE I- E(OPEIL-)BLE 0-FNL-2364 4SOV AC VITA.

DISCONNECT PNL CHANNELS O-SW-2364S(VL FROM (NC) S-XSW-235-7ND22-S, OUTPUT 2 125V DC TRANSFER SWITCH 125V VITAL SELECT CHARGERS 7-SW-S BAIT IV NOTE 2 (NC)

((SO) 15JR23(CB353 2-EETE.237--I-CE3U1 I I

(NC) ((AC) 2BITE.23T.4Ct71 T4C4 1 (NC)

L-BER-OW-1C132 0C) (NO)

¶-XSW-235-4 12S SAC V1A1. (SOP. t;:____;:3;..i ALT 2-XSW-235-4 i; NSTR POWER ND (-IV 120 VAC V1T14.

INSTR POWER IT JR____

0 i

A (SOP.

INC)

TRANSFER SD 2-IV l.SW.2354 120 VAC VITAL TRANSFER 2SW-23TA 120 Vac VITAL NC NOTE POWER SD 1-IS DISC POWER SD 2-IVOISC 1-SD255-4l2SVac VITAL 2-SO 235-4 125 Sac VITAL INSTR POWER BOARD (-IV NOTE POWER BOARD 2-IS NOlt (1) In Modes 5&6 ctlly one train of ac/dc PWR is reouired, if this train is not reouired this caoe may be N/A.

(2) When 7-S or 9-S char9er is cDCnected to Bati Bd th rity assoc. train Transfer Switches are closed and Pt bkrs are open.

INITIALS OF DATA COLLECTOR: DATE REMARKS: /1 I,

WBN Weekly Log 0-Sl-0-3 Unit 0 Rev. 0045 Page 21 of 43 Appendix A (Page 10 of 24)

SR 3.8.9.1 NC 48OVAux Bldg Corn MCC C SR 3.8.10.1 I

- I NCI 125V VITAL H I r 9 V 125VViICS BATTV C82 NC 125V VITAL NC See note2 I (101) BATT SD V (102) 125V DC 01ST PNL 5A 125V DC 01ST PNL 58 I

ANO NO (301) (302) i I

I 1

j)

I I

i I

I I

I r;

I I I I 1 INO 1 NO NO: 11401 (2/0)1 I I (4/0),

I___I....I 0-DPL-236-1 0-DPL-236-3 0-DPL-236-2 0-DPL-236-4 125V VITAL 125V VITAL 125V VITAL 125V VITAL BATT BD I BATT 80111 BATT BD II BATT 00 IV CLOSED NOTE 1: Only one battery board may be lend from the 5th battery board at any one time.

AVA)LABLE NOTE 2: 0-BKR-236-51102 must be OPEN when 5th battery is inservice for battery bd I, II, III, or IV INOPERABLE INITIALS OF DATA COLLECTOR: W DATE REMARKS: F

Main .Jarms Graphics Trends Points Zoom/Layers Print Help 06-.JUL-2011 07:46:16 SELECT FUNC. KEY OR 11JRII -OFI CODE [)O{HOU SE:>

110-P ioo- POWER LEVEL 99+4 o GIRL BANK D (STEPS) 2200 7

AF[) MIS / + I) w 80 CHANNEL 41 E AFD MIS CHANNEL 42 AFD MIS CHANNEL 43 7+7 R i-7

+ L AFD MIS CHANNEL 44 /

60 NIS ACTUAL AFD 7.5 L NI S TARGET AFD 7.5  ;

E V

AFD LOW LIMIT E AFD HIGH LIMIT L 2O:

0 iOZ CONTROL BAND LOW L IH

/0 CONTROL BAND HIGH LIII I I II I II II I I II III II I I I I I

-20 -1 -10 - 0 10 1 20

-3 -30 -2 AXIAL FLUX D I FFERENC E TINE OUT OF BAND AGCIJM (HIM) 0+O J TT(Y) il= AIII)IIYtl IS SEC LVL= I PRIlI IEACK CPI I I I lODE I

WBN Loss of RHR Sump Recirculation ECA-ti Uniti Rev0012 Step Action/Expected Response Response Not Obtained

19. CHECK SI termination criteria:

a. RVLIS greater than 60% a. IF RVLIS is less than or equal to with NO RCP running, setpoint, THEN OR **

GO TO Step 25.

RVLIS greater than 63%

with ANY RCP running.

b. RCS subcooling greater than b. ESTABLISH minimum ECCS flow required from table: for decay heat removal:

1) REFER TO Figure 1, Minimum SI Flow For Decay Heat Versus Time After Trip.

2) **GOTOStep25 RCS PRESS BETWEEN REQUIRED SUBCOOLING 285 AND 585 psig 115°F [135°F ADVJ 585 AND 1085 psig 102°F [1 23°F ADV]

1085 AND 1885 psig 97°F[117°FADV]

Greater than 1885 psig 94°F [1 14°F ADVJ

20. RESET Phase A and Phase B.

Page 12 of 35

WBN Loss of RHR Sump Recircuation ECA-1 .1 Unit I -

Revs Qi2 Figure 1 (Page 1 oN)

Minimum SI Flow for Decay Heat vs. Time After Trip 700.0 .

rr ii rr 1

--.-f. I 600.0 c

500.0

\

1-44-1

• 1 --

400.0 1

4-4-p I w -

- - - - I 300.0 L

. I -___

200.0 100.0 0.0 10 z: 100 Time After Trip (Minutes) 1000 10000 Page 29 of 35

Ice Condenser Doors

3. 6.12 3.6 CONTAINMENT SYSTEMS 3.6.12 Ice Condenser Doors LCO 3.6.12 The ice condenser inlet doors, intermediate deck doors, and top deck doors shall be OPERABLE and closed.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS NOTE Separate Condition entry is allowed for each ice condenser door.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more ice A.l Restore inlet door to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> condenser inlet OPERABLE status.

doors inoperable due to being physically restrained from opening.

B. One or more ice B.1 Verify maximum ice bed Once per condenser doors temperature is 27°F. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> inoperable for reasons other than AND Condition A or not closed. B.2 Restore ice condenser 14 days door to OPERABLE status and closed positions.

(continued)

Watts Bar-Unit 1 3 . 631

Ice Ccndenser DQrs

3. 6.12 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Restore ice condenser 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> associated door to OPERABLE status Completion Time of and closed positions.

Condition B not met.

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 AND Condition A or C not met. D.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 3.6.12.1 Verify all inlet doors indicate closed by 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> the Inlet Door Position Monitoring System.

SR 3.6.12.2 Verify, by visual inspection, each 7 days intermediate deck door is closed and not impaired by ice, frost, or debris.

(continued>

Watts BarUnit 1 3.6-32 Amendment 3

Ic Coidnser Door a

3. 6.12 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.12.3 Verify, by visual inspection, each inlet NOTE door is not impaired by ice, frost, or The 3 month debris, performance due September 9, 1996 (per SR 3.0.2) may be extended until October 21, 1996.

3 months during first year after receipt of license AND 18 months SR 3.6.12.4 Verify torque required to cause each NOTE inlet door to begin to open is The 3 month 675 in-lb. performance due September 9, 1996 (per SR 3.0.2) may be extended until October 21, 1996.

3 months during first year after receipt of license AND 18 months (continued)

Watts Bar-Unit 1 3.633 Amendment 3

Ice. Conden.sr Docrs

3. 6.12 SURVEILLANCE REQUIREMENTS (Continued)

SURVE ILLANCE FREQUENCY SR 3.6.12.5 Perform a torque test on a sampling of NOTE 50% of the inlet doors. The 3 month performance due September 9, 1996 (per SR 3.0.2) may be extended until October 21, 1996.

3 months during first year after receipt of license AND 18 months SR 3.6.12.6 Verify for each intermediate deck door: 3 months during first

a. No visual evidence of structural year after deterioration; receipt of license

b. Free movement of the vent assemblies; and AND

c. Free movement of the door. 18 months (continued)

Watts Bar-Unit 1 3.634 Amendment 3

Lce Condenser Doora

3. 6.12 SURVEILLANCE REQUIREMENTS (Continued)

SURVEILLANCE FREQUENCY SR 3.6.12.7 Verify, by visual inspection, each top 92 days deck door:

a. Is in place;

b. Free movement of top deck vent assembly; and

c. Has no condensation, frost, or ice formed on the door that would restrict its opening.

Watts Bar-Unit 1 3.6-34a mendment 3

WBN Emergency Plan Classification Logic EPIP-1 Unit 0 Rev. 0035

. Page 49 of 51 Attachment 7 (Page 6 of 7)

Mode InitiatinglCondition Mode InitiatinglCondition Refer to Fission Product Barrier Matrix or Refer to Gaseous Effluents (7.1)

Gaseous Effluents (7.1)

Refer to Fission Product Barrier Matrix or Refer to Gaseous Effluents (7.1)

Gaseous Effluents (7.1)

UNPLANNED increases in Radiation levels All Major damage to Irradiated Fuel, Loss of within the Facility that impedes Safe water level that has or will uncover Irradiated Operations establishment maintenance Fuel outside the Reactor Vessel of Cold Shutdown (1 and 2)

(1 or 2)

1. VALID alarm on 0-RE-90-101B

1. VALID, area Radiation Monitor readings 0-RE-90-102 g survey results exceed 15 mrem/hr in the Control Room or CAS O-RE-90-lO3or 1-RE-90-1301131 or 1-RE-90-112 or 1-RE-90-400 or

2. (aandb) 2-RE-90-400

a. VALID area radiation monitor readings 2. (aorb) exceed values listed in Table 7-2

a. Plant personnel report damage of

b. Access restrictions impede operation of Irradiated Fuel sufficient to rupture Fuel systems necessary for Safe Operation Rods or the ability to establish Cold Shutdown b. Plant personnel report water level drop has or will exceed makeup capacity See UNUSUAL EVENT Note Below such that Irradiated Fuel will be uncovered All UNPLANNED increase in Radiation levels All UNPLANNED loss of water level in Spent Fuel within the Facility Pool Reactor Cavity or Transfer Canal with fuel remaining covered (1 and 2 and 3)

1. VALID area Radiation Monitor readings increase by a factor 1000 over normal levels 1. Plant personnel report water level drop in Spent Fuel Pool, or Reactor Cavity, or Note: In Either the UE 0rALERTEAL, the SED Transfer Canal must determine the cause of Increase in Radiation Levels and Review Other 2. VALID alarm on 0-RE-90-1 02 or INITIATING/CONDITIONS forApplicabillty (e.g., 0-RE-90-1 03 or 1 -RE-90-59 or 1 -RE-90-60 a dose rate of 15 mrern/hr in the Control Room could be caused by a release associated with a 3. Fuel remains covered with water.

DBA).

WBN Fire $af ShutdQwn EIevaton AOI-30.2 APP B Unit 0 Diagrams Rev. 0000 Page 5 of 16 2.0 AB EL 772.0, 776.0 & 763.5 ELEVATION DIAGRAM Auxiliary Building El 772.0, 776.0 & 763.5 Diagram A5 ROOM ROOM NAME PROCEDURE ROOM ROOM NAME PROCEDURE 772.0Al 480V Rx MDV Bd Rm AOl-30.2 0.2 772.0AlO Mech Equip Rm ADI-30.2 0.11 1A 772,0A2--E 480V Rx MDV ADI-30.2 0.3 772.0All 480V Xfmer Rm 2B ADI-30.2 0.12 Bd Rm lB (East) 772.0A2W 480V Rx MDV Bd Rm ADI-30.2 C.4 772.0Al2 480V Xfmer Rm 2A ADI-30,2 0.13 lB (West) 772.0A3 125V Vital Batt Rm II ADI-30.2 0.5 772.0A13 125V Vital Batt Rm IV ADI-30.2 0.14 772.0A4 125V Vital BattRm I ADI-30.2 0.6 772.0A14 l25V Vital Batt Rm Ill ADI-30.2 0.15 772.0A5 480V Xfmer Rm lB ADI-30.2 0.7 772.0Al5E 480V Rx MDV Bd Rm ADI-30.2 0.16 2B (East) 772.0A6 480V Xfmer Rm 1A ADI-30.2 0.8 772.0A15W 480V Rx MDV Bd Rm ADI-30.2 0.17 2B (West) 772.0A7 Mech Equip Rm ADI-30.2 0.9 772.0A16 480V Rx MDV Bd Rm ADI-30.2 0.18 2A 72.0A8 5th Vit Batt & Bd Rm ADI-30.2 0.10 776.0Al Elev Mach Rm ADI-30.2 0.60 p772.0A9 HEPA Filter Plenum Rm ADI-30.2 0.44 763.5Al Ice Equip Rm ADI-30.2 0.45 757.0Al3 (Next Page)

WBN 10-2011 NRC SRO Exam As Submitted 811512011

76. 011 EG2.4.41 076 During a LOCA, which ONE of the following identifies...

(1) how many of the EPIP-1, Emergency Plan Classification Logic, Fission Product Barrier Matrix contain a decision point based directly on RVLIS and (2) the RVLIS threshold level that first requires a classification declaration be made?

Barrier Matrix Threshold level A. 1 <44%

B. 2 <44%

C. 1 <33%

D 2 <33%

DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible because many of the criteria only appear in one of the barriers and 44% is the void content of the RCS while the RCPs are in service that would result in a vessel level of less than 33% during a LOCA if the RCPs were to trip.

B. Incorrect, Plausible because RVLIS level appears in both Fuel Clad Barrier (as a Potential LOSS) and in the RCS Barrier (as a LOSS) and 44% is the void content of the RCS while the RCPs are in service that would result in a vessel level of less than 33% during a LOCA if the RCPs were to trip.

C. lncorrect Plausible because many of the criteria only appear in one of the barriers and less than 33% being the RVLIS value during a LOCA that will require a declaration is correct.

D. Correct, Valid RVLIS level less than 33% appears in both 1.1 Fuel Clad Barrier and in 1.2 RCS Barrier. RVLIS appears in the Fuel Clad Barrier as a Potential LOSS and in the RCS Barrier as a LOSS.

Question Number: 76 Tier: 1 Group 1 Page 1

WBN 10-2011 NRC SRO Exam As Submitted 811512011 K/A: 011 EG2.4.41 Large Break LOCA Emergency Procedures / Plan Knowledge of the emergency action level thresholds and classifications.

Importance Rating: 2.9 / 4.6 10 CFR Part 55: 41.10/43.5/45.11 IOCFR55.43.b: 6,7 K/A Match: K/A is matched because the question requires knowledge of the emergency action level threshold value for reactor vessel level and the barriers that are affected by the Reactor Vessel Level Indicating System indicating below the minimum required.

Question is SRO because it requires detailed knowledge of the procedures used to evaluate plant conditions to determine Emergency Classifications and this determination is an SRO function.

Technical

Reference:

EPIP-1, Emergency Plan Classification Logic, Fission Product Barrier Matrix, Revision 0035 FR-0, Status Trees, Revision 0014 Proposed references None to be provided:

Learning Objective: 3-OT-PCDO48C

1. Classify emergency events

16. Recognize conditions which constitute activation of the emergency response facilities regardless of the time of day when an emergency has been declared.

Cognitive Level:

Higher Lower X Question Source:

New X Modified Bank Bank Question History: New question for the WBN 10/2011 NRC exam Comments:

Page 2

WBN Emeeney PIanCIassificatn Logic EP-1 Unit 0 Rev. 0035 Page 12 of 51 Attachment I (Page 3 of 4)

1. Critical Safety Function Status ii. Critical Safety Function Status Potential LOSS LOSS Potential LOSS Core Cooling Red (FR-Cl) Core Cooling Orange Not Applicable Pressurized Thermal Shock (FR-C.2) Red (FR-P.1)

OR OR Heat Sink Red (FR-H.1) Heat Sink Red (FR-Hi)

(RHR Not in Service) (RHR Not in Service)

-OR- -OR

2. Primary Coolant Activity Level 2. RCS Leakae/LOCA

.1* Potential LOSS Potential LOSS RCS sample activity is Not applicable RCS Leak results in Loss of Non Isolatable RCS Leak Greater Than 300 ICi/gm subcooling (<65°F Exceeding The Capacity of dose equivalent iodine-131 Indicated), [85°F ADVJ One Charging Pump (CCP)

In the Normal Charging Alignment.

OR RCS Leakage Results In Entry Into E-1 0R.

3. Incore TCs Hi Quad Average 3. Steam Generator Tube Ruoture toss Potential LOSS LOSS Potential LOSS Greater Than 1200°F Greater Than 727°F SGTR that results in a Not Applicable safety injection actuation OR Entry into E-3

-OR- -OR-

4. Reactor Vessel Water Level 4. Reactor Vessel Water Level

.}* Potential LOSS Potential LOSS Not Applicable VALID RVLIS level <33% VALID RVLIS level <33% Not Applicable (No RCP running) (No RCP Runnino

-OR-

5. Containment Radiation Monitors LOSS Potential LOSS VALID reading increase of Not Applicable Greater Than:

293 R/hr On 1-RM-90-271 -OR and 272 OR 261 R/hr On 1-RM-90-273 and 274 (see instruction note 51

-OR-

6. Site Emergency Director Judgment 5. Site Emeroencv Director Judment Any condition that, in the Judgment of the SM/SED, Any condition that, in the Judgment of the SMJSED, Indicates Loss or Potential Loss of the Fuel Clad Barrier Indicates Loss or Potential Loss of the RCS Barrier ComarabIe to the Conditions Listed Above. Comparable to the Conditions Listed Above.

WBN - Emergency Plan Classification Logic EPIP-1 Unit 0 Rev. 0035 Page 13 of 51 Attachment I (Page 4 of 4)

Modes: 1,2,3,4

1. Critical Safety Function Status INSTRUCTIONS I1* Potential LOSS NOTE:

Not Applicable Containment (FR-Z.1) Red A condition is considered to be MET if in the OR judgment of the Site Emergency Director, the Actions of FR-C.1 (Red condition will be MET imminently (i.e., within Ito Path) are INEFFECTIVE 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in the absence of a viable success path). S (i.e.: core TCs trending up) The classification shall be made a soon as this S

determination is made.

-OR- In the matrix to the left, review the INITIATING 0 2 Cnntinmnt PrrcstirIF n CONDITIONS in all columns and identify which, N if any, INITIATING CONDITIONS are MET.

LOSS Potential_LOSS P Circle these CONDITIONS.

Rapid unexplained Containment Hydrogen 2. For each of the three barriers, identify if any R decrease following initial Increases to >4% by LOSS or Potential LOSS INITIATING 0 increase volume CONDITIONS have been MET. D 2E 3. If a CSF is listed as an INITIATING U Containment pressure or Pressure >2.8 PSIG CONDITION; the respective status tree criteria C Sump level increasing (Phase B) with < One full will be monitored and used to determine the T (with LOCA in progress) train of Containment spray EVENT classification for the Modes listed on the B

-OR- classification flowchart.

A

3. Containment Isolation Status 4. Compare the barrier losses and potential losses R LOSS Potential_LOSS to the EVENTS below and make the appropriate R declaration.

Containment Isolation is Not Applicable 5. Containment High Range Radiation Monitors Incomplete (when required) (HRRMs) are temperature sensitive and can be AND a Release Path to the R affected by both temperature induced currents Environment Exists and insulation resistance temperature effects. M

-OR Following the initial increase in containment A 4 flnntinmnt ç temperature the HRRM monitors can give T LOSS Potential_LOSS erratic indication for up to 1 minute. Steady R Unexplained VALID state temperature effects on cable insulation RUPTURED SIG is also FAULTED outside CNTMT increase in area or resistance for the HRRM signal cable is x ventilation RAD monitors in dependent on containment temperature and OR U areas adjacent to CNTMT could result in a shift in monitor output Prolonged (>4 Hours) I (with LOCA in progress) indication. With a containment excursion Secondary Side release outside CNTMT from a S/G temperature to 327 °F (HELB), the output of with a SGTL> T/S Limits the HRRMs could potentially have up to a 25 RIhr indicated offset for duration of 10 minutes

-OR- until the containment air return fans are started 5: Signifibant Radioactivity in Containment and temperature starts to reduce. (Caution:

1.)* Potential LOSS Should the containment air return fans not Not Applicable VALID Reading increase of start, containment temperatures could Greater Than: remain elevated resulting in potential 5290 RIhr on 1-RM-90-271 false HRRM indicated readings).

and 1-RM-90-272 EVENTS OR UNUSUAL EVENT ALERT 4710 RJhr on 1-RM-90-273 Loss Potential LOSS of Any LOSS or Potential Containment Barrier LOSS of Fuel Clad barrier and 1-RM-90-274 OR (see instruction note 5) Any LOSS Potential

-OR- LOSS of RCS barrier

6. Site Emergency Director Judgment Any condition that, in the Judgment of the SM/SED, SITE AREA EMERGENCY GENERAL EMERGENCY Indicates Loss or Potential Loss of the CNTMT Barrier LOSS or Potential LOSS of LOSS of any two barriers any two barriers and Potential LOSS of third Comparable to the Conditions Listed Above.

barrier

WBN Status Trees FR-0 Unit I Rev. 0014 Attachment I (Page 2 of 8)

Monitoring Critical Safety Functions CORE COOLING FR-C RED GO TO FR-Cl COLOR PROC GO TO FR-C.l GOTO ORANGE FR-C.2 GO TO ORANGE,C)

FR-C,2 YELLOQ GO TO FR-C.3 RAGE GO TO FR-C.2 ELL0W GO TO FR-C.3 0

GREE OK Page 5 of 11

Watts Bar RADIOLOGICAL 3-OT-PCD-048C Nuclear Plant EMERGENCY PLAN Rev. 11 TRAINING Page 4 of 48 V. TRAINING OBJECTIVES

1. Classify emergency events.

2. Recognize the reasons for having the Radiological Emergency Plan (REP).

3. Identity the functions of the onsite emergency response facilities.

4. Formulate Protective Action Recommendations (PARs).

5. Use the WBN Emergency Plan Implementing Procedures (EPIPs).

6. State three Site Emergency Director responsibilities that cannot be delegated.

7. Identify Operations responsibilities for the following emergency response positions:

• Site Emergency Director (who is initially the SM)

• Operations Manager in the TSC

• Control Room Communicator in the Control Room

• Operations Communicator in the TSC

• OSC Operations Advisor

• Operations emergency response team assignments

• NOMS Logkeeper in the Control Room (when available)

• Technical Advisor

• Designated Phone Talker

8. Recognize how AUOs are dispatched and controlled during radiological emergencies.

9. Recognize REP communications guidelines (OPDP-1).

10. Demonstrate effective communication techniques used in emergency response.

11. Identify lessons learned from TVA/industry events, drills and exercises.

1 2. Recall where radios can and cannot be used at WBN (BP-364).

13. Use the Integrated Computer System (ICS).

14. Identify all locations where the Emergency Paging System (EPS) may be activated from and demonstrate the use of the EPS to include the printed report from the TSC.

1 5. Using WBN EPIPs 2, 3, 4, and 5, recognize who is responsible to activate the Emergency Paging System.

1 6. Recognize conditions which constitute activation of the emergency response facilities regardless of the time of day when an emergency has been declared.

17. Identify and use the back-up Emergency Response Organization call lists used when the Emergency Paging System has failed.

Watts Bar RADIOLOGICAL 3-OT-PCD-048C Nuclear Plant EMERGENCY PLAN Rev. 11 TRAINING Page 5 of 48 V. TRAINING OBJECTIVES (continued)

18. Recognize entry conditions for Severe Accident Management Guidelines (SAMGs).

19. Use the Radiological Emergency Notification Directory (REND).

20. Use the Satellite Phone to make calls during emergencies.

21. Identify the WBN REP procedure addressing MERT responsibilities, offsite agreement support, and emergency phone numbers.

22. Review Operations drill critique items.

23. Perform dose assessments using ICS for WBN EPIP-13.

24. Interpret MET data obtained in the TSC from the CECC computer.

25. Identify specific actions of OSC Emergency Responders in the OSC teams staging area (EPT 309.000).

26. Understand the critical times associated with

• Event Declaration

• Offsite Notification

• Facility Staffing

• Printed EPS Report

Clarification Gtiidaneefor SRO-oniy Questions RevI (0311112010)

F. Procedures and limitations involved in initial core loading, alterations in core configuration, control rod grogramming, and determination of various internal and external effects on core reactivity. [10 CFR 55.43(b)(6)]

Some examples of SRO exam items for this topic include:

• Evaluating core conditions and emergency classifications based on core conditions.

• Administrative requirements associated with low power physics testing processes.

• Administrative requirements associated with refueling activities, such as approvals required to amend core loading sheets or administrative controls of potential dilution paths and/or activities.

• Administrative controls associated with the installation of neutron sources.

• Knowledge of TS bases for reactivity controls.

G. Fuel handling facilities and procedures. [10 CFR 55.43(b)(7)]

Some examples of SRO exam items for this topic include:

• Refuel floor SRO responsibilities.

• Assessment of fuel handling equipment surveillance requirement acceptance criteria.

• Prerequisites for vessel disassembly and reassembly.

• Decay heat assessment.

• Assessment of surveillance requirements for the refueling mode.

• Reporting requirements.

• Emergency classifications.

This does not include items that the RO may be responsible for at some sites such as fuel handling equipment and refueling related control room instrumentation operability requirements, abnormal operating procedure immediate actions, etc. For example, an RO is required to stop the refueling process when communication is lost between the control room and the refueling floor, therefore, this is a task that is both an RO and SRO responsibility and is not SRO-only.

Page 9 of 16

WBN 10-2011 NRC SRO Exam As Submitted 811512011

77. 025 AG2.4.21 077 Given the following:

- Unit I was shutdown 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ago for a refueling outage.

- RHR pump I B-B is tagged for maintenance.

- Current RCS temperature is 135°F and 80 psig.

- RHR pump IA-A trips and cannot be restarted.

- RCS temperature begins to rise.

- The crew is performing AOl-I 4, Loss of RHR Shutdown Cooling, and Section 3.9, RCS Alternate Cooling Method with RX Vessel Head Installed, has been initiated.

Which ONE of the following identifies...

(1) the cooling method that is directed to be attempted first after AOl-14 Section 3.9 is implemented and (2) the condition that will result in an REP declaration being required?

A. (1) Establish RCS feed and bleed using a CCP and a pressurizer PORV.

(2) RCS incore temperature > 200°F B (1) Establish natural circulation in the RCS using the steam generators.

(2) RCS incore temperature > 200°F C. (1) Establish RCS feed and bleed using a CCP and a pressurizer PORV.

(2) RHR not established for> 15 minutes D. (1) Establish natural circulation in the RCS using the steam generators.

(2) RHR not established for> 15 minutes Page 3

WBN 10-2011 NRC SRO Exam As Submitted 8115/2011 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because RCS feed and bleed with a CCP and a pressurizer PORVIs a cooling process implemented by Section 3.9 of the AOl and because EPIP- I requiring a declaration following the loss of RHR when the incore temperatures exceed 200°F is correct.

B. Correct, Section 3.9 will check for conditions to establish cooling by natural circulation in the RCS and EPIP-1 will require an ALERT to be declared following the loss of RHR when the incore temperatures exceed 200°F.

C. Incorrect Plausible because RCS feed and bleed with a CCP and a pressurizer PORV is a cooling process implemented by Section 3.9 of the AOl and because exceeding 15 minutes is a time frame used in several conditions for requiring a declaration of the REP (e.g. electrical board not available for> 15 minutes, fire lasting >15 minutes, rad assessments not completed within 15 minutes, etc.)

D. Incorrect, Plausible because establishing natural circulation is correct and because exceeding 15 minutes is a time frame used in several conditions for requiring a declaration of the REP (e.g. electrical board not available for> 15 minutes, fire lasting >15 minutes, rad assessments not completed within 15 minutes, etc.)

Question Number: 77 Tier: 1 Group 1 K/A: 025 AG2.4.21 Loss of Residual Heat Removal System Emergency Procedures I Plan Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

Importance Rating: 4.0 I 4.6 IOCFRPart55: 41.7/43.5/45.12 IOCFR55.43.b: 5, 6 K/A Match: K/A is matched because the question requires knowledge of core cooling and heat removal processes directed by the AOl following a loss of the RHR system and is SRO because of requiring detailed knowledge of the procedure content (including flowpath through the procedure) to prevent radioactive releases and the requirements for implementation of the Radiological Emergency Plan following a loss Page 4

WBN 10-2011 NRC SRO Exam As Submitted 811512011 KIA Match: K/A is matched because the question requires knowledge of core cooling and heat removal processes directed by the AOl following a loss of the RHR system and is SRO because of requiring detailed knowledge of the procedure content (including flowpath through the procedure) to prevent radioactive releases and the requirements for implementation of the Radiological Emergency Plan following a loss of the RHR system.

Technical

Reference:

AOl-14, Loss of RHR Shutdown Cooling, Revision 0037 EPIP-1, Emergency Plan Classification Logic, Revision 0035 Proposed references None to be provided:

Learning Objective: 3-OT-A0l400

5. Explain Alternate RHR Cooling methods.

3-OT-PCD048C

1. Classify emergency events.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank question A0l1400.05 001 modified Comments:

Page 5

WBN Loss of RH Shutdown Cooling AOI-14 Unit I Rev. 0037 Step Action/Expected Response Response Not Obtained 39 RCS Alternate Cooling Method With Rx Vessel Head Installed [C.4, C.7)

CONTINUE attempts to restore RHR Shutdown Cooling flow.

2. COMPLETE the following: [c.21

• EVACUATE non-essential personnel from cntmt.

• NOTIFY STA to implement Tl-68.002 for cntmt closure.

• IF any RCP shaft uncoupled, THEN NOTIFY STA to install RCP shaft restraint device(s) to limit pump shaft leakage.

• NOTIFY RP to provide monitoring and Rad protection guidance for workers involved in cntmt closure.

3. IF at least one RCP is running, THEN USEsteamdumpsorSGPORV operation to restore cooling, AND

• • GOTOStepl8

4. CHECK at least TWO SIGs narrow IF S/G narrow range NOT available, range levels greater than 29%. -

-? THEN GO TO Note prior to Step 12.

V Page 49 of 83

WBN Loss of RHR Shutdown Cooling AOi-14 Unit I Rev. 0037 Step _Action/Expected Response Response Not Obtained 3.9 RCS Alternate Cooling Method With Rx Vessel Head Installed [C.4, C.7] (continued)

5. CHECK RCS intact and capable of IF RCS NOT capable of being being pressurized & transferring heat: pressurized,

• Pzr COLD CAL level THEN indicator 1-LI-68-321 on scale. GO TO Note prior to Step 12.

• RCP shafts coupled.

• Pzr PORVs or associated block valves capable of being CLOSED.

• Pzr safeties installed.

• S/G primary manways installed.

• RCS intact with SG tubes filled.

6. CHECK RCS press less than COPS: ENSURE PORV and associated block

• valve OPEN.

REFER TO Appendix D.

WHEN press less than COPS, THEN 7.

L ENSURE pzr PORVs CLOSED and ENSURE PORV or associated block valve CLOSED.

GO TO Caution prior to Step 8.

IF any PORV can NOT be closed, HSs in AUTO: THEN

• 1-HS-68-334A. CLOSE associated block valve.

• 1 -HS-68-340A.

IDri 1fl f R V -

WN Loss of RHR Shutdown Cooling AOI-14 Unit 1 Rev. 0037 Step Action/Expected Response Response Not Obtained 3.9 RCS Alternate Cooling Method With Rx Vessel Head Installed [C.4, C.7] (continued) 8.

CAUTION

/ RCS pressure control must be established prior to the onset of core boiling to avoid the potential loss of natural circulation due to steam binding of the S1G U-tubes.

ESTABLISH RCS press greater than 100 psig and less than COPS:

• REFER TO Appendix D.

• IF pzr bubble exists, THEN USE heaters and sprays.

• lFwatersolid, THEN USE charging and letdown.

• IF less than water solid AND pzr subcooled, THEN USE charging and letdown to raise level and maintain press control.

Page 51 of 83

WN Loss of Shutdown ooUng AOI-14 Unit I Rev. 0037 Step Action/Expected Response Response Not Obtained 3.9 RCS Alternate Cooling Method With Rx Vessel Head Installed [CA, C.7] (continued)

NOTE When initially establishing natural circulation, RCS temperature response will be delayed until the necessary delta-T is established.

9. ESTABLISH natural circulation: IF natural circulation can NOT be

a. MAINTAIN RCS press greater established,

- -) THEN than 100 psig and less than **

GO TO Note prior to Step 12. 13cLif COPS.

b. MAINTAIN at least two SGs NR level greater than 29%.

c. OPEN S/G PORV5 for the selected S/Gs.

d. USE AFW and S/G blowdown for feed and bleed of SIG.

e. CHECK RCS temp stable or dropping.

10. IF RCP support conditions exist, OR can be established, THEN START one RCP:

• REFER TO SOI-68.02, Reactor Coolant Pumps.

11. **GOTOStep18 Page 52 of 83

• WN Loss f i-I Shutdown tooling AOl-14 Unit I Rev. 0037 Step Action/Expected Response Response Not Obtained 3.9 RCS Alternate Cooling Method With Rx Vessel Head Installed [C.4, C.7] (continued)

NOTE The following steps only apply if natural circulation can NOT be established and Reactor Vessel head is installed.

12. ENSURE both pocket sump pumps STOPPED [M-15]:

1-HS-77-410.

1-HS-77-411.

13. DETERMINE appropriate step to initiate a feed and bleed cooling method:

IF the following feed and bleed path is to be used: THEN Normal charging to RCS; bleed through manual pzr GO TO Step 17.

PORV control with RCS intact, Normal charging to RCS; bleed through SIG HL GO TO Step 15.

manway with nozzle dam removed, Normal charging to RCS; bleed through pzr GO TO Step 16.

manway or removed PORV/Safety valve flanges, Gravity feed to RCS; bleed through S/G CL GO TO Step 14.

manway with nozzle dam removed, Gravity feed to RCS; bleed through RCP shaft GO TO Step 14.

leakage (shaft uncoupled and NOT blocked),

Page 53 of 83

WBN Loss of RHR ShUtdown Collng AOl-14 Unit I Rev. 0037 Step LActb0fhExpDt& Response Response Not Obtained 3.9 RCS Alternate Cooling Method With Rx Vessel Head Installed [c.4, C.7] (continued)

NOTE Core boiling is necessary for long-term core heat removal during the loss of RHR cooling. RCS inventory must be maintained.

14. PERFORM the following to cool RCS via the S!G CL manway or RCP shaft:

a. DISPATCH Operator to operate 1-FCV-63-1 locally OR electrically [480V Rx MOV Bd IAI-A].

b. OPEN the following valves: b. ENSURE 1-FCV-74-1

• and 1-FCV-74-2 CLOSED.

1-FCV-74-1, Loop 4 Hot Leg To RHR Suction.

RESTORE power, and OPEN

• 1-FCV-74-2, Loop 4 Hot Leg bypass valves:

To RHR Suction.

1) 1-FCV-74-9, RHR System Isol Bypass.

2) 1 -FCV-74-8, RHR System Isol Bypass.

Step continued on next page.

Page 54 of 83

WBN Loss of RHR Shutdown Cooling AOI-14 Unit I Rev. 0037 Step Action/Expected Response Response Not Obtained 39 RCS Alternate Cooling Method With Rx Vessel Head Installed [c.4, C.7] (continued)

NOTE 1-FCV-63-1 should be slowly throttled open and flow monitored with communications maintained between MCR and local Operator throughout this evolution.

c. CONTROL 1-FCV-63-1 c. Locally CONTROL 1-FCV-63-1.

electrically [480V Rx MOV Bd 1A1-A, c/bA] using breaker operation.

d. MONITOR RCS level stable or d. THROTTLE OPEN 1-FCV-63-1.

rising.

RVLIS

• 1-LT-68-399A, NR Level

• 1-LT-68-399B, WR Level

e. MONITOR RCS temp stable or e. IF RCS temp control can NOT be dropping. established, THEN GO TO Step 13.

f. **GOTOStepI8 Page 55 of 83

• WBN Emergency Plan Classification Logic EPlP4 Unit 0 Rev. 0035

. Page 42 of 51 Attachment 6 (Page 3 of 4)

Mode InitiatinglCondition Mode InitiatinglCondition 5,6 Note: Additional information will be provided later Not Applicable pending NRC Guidance on Shutdown EALs Refer to Gaseous Effluents (7.1)

Loss of water level in the Rx vessel that has Not Applicable or will uncover fuel in the Rx vessel (1 and 2 and 3 and 4)

1. Loss of RHR capability

2. Rx vessel water level < el. 718

3. Incore TCs (if available) indicate RCS temp.

>200 F

4. RCS is vented/open to CNTMT Note: If CNTMT open, refer to Gaseous Effluents (7.1)

Inability to maintain Unit in Cold Shutdown 5,6 or UNPLANNED loss of Offsite and Onsite AC De- Power for >15 minutes (1 and 2) Fuel lAand lB 6.9KV Shutdown Bds

1. RHR capability is not available for RCS de-energized for >15 minutes Cooling

2. Incore TCs (if available) indicate RCS temp.

>2000 F Note: If CNTMT open, refer to Gaseous Effluents (7.1) 5,6 Note: Additional information will be provided later 5,6 or UNPLANNED loss of All Offsite Power for pending NRC Guidance on Shutdown EALs De- >15 minutes Fuel (1 and 2)

1. C and D CSSTS not available For

>15 minutes.

if -fI 1

1l7%ft4 4ti

2. Either Diesel Generator is supplying power to its respective Shutdown Board C1&LvL

A011400.05 001 Ii oN Given the following plant conditions:

- The Unit is being cooled down and depressurized for an outage.

- 1 B-B RHR pump is tagged for required repairs.

- RCS temperature is 100°F.

- RCS pressure is 225 psig.

- All RCPs are off.

- SIG levels are all 38%.

- The running IA-A RHR pump trips.

Which of the following is required per AOl-i 4, Loss of RHR Shutdown Cooling?

a. Maintain RCS pressure < 100 psig, and establish natural circulation using AFW and Steam Dumps.

b Maintain RCS pressure> 100 psig, and less than COPS, and open SIG PORVs to establish natural circulation using AFW and S/G blowdown.

c. Establish RCS feed and bleed using normal Charging and a Pzr PORV.

d. Immediately start an RCP and establish a secondary heat sink by steaming the S/Gs.

3-CT-AOl 1400 Revil Page 4 of 181 PROGRAM WATTS BAR OPERATOR TRAINING II. COURSE A. LICENSE TRAINING B. NON-LICENSE TRAINING III. TITLE AOl-i 4, LOSS OF RHR SHUTDOWN COOLING IV. LENGTH OF LESSON A. License Training 2 Hours B. NOTP 1 Hour V. TRAINING OBJECTIVES AR SS U OR T 0 0 A x x X X 1. Demonstrate knowledge of the Purpose/Goal of AOI-14, Loss of RHR Shutdown Cooling.

X X X X 2. Identify 3 factors that effect Severity of the Loss of RHR Shutdown Cooling.

X X X 3. Explain possible Alarms for Loss of RHR Shutdown Cooling.

X X X X 4. Describe 5 ways that RHR Cooling can be lost.

X X X 5. Explain Alternate RHR Cooling methods.

X X X X 6. Identify 5 Causes for Loss/Degradation of RHR capability in PWRs in the industry, per SOER 85-4.

X X X 7. Demonstrate ability/knowledge of AOl, to correctly:

a. Recognize Entry conditions.

b. Respond to Action steps.

c. Respond to Contingencies (RNO column).

d. Respond to Notes & Cautions.

X X X X 8. Describe AUO actions for venting a RHR pump when it becomes air bound.

Watts Bar RADIOLOGICAL 3-OT-PCD-048C Nuclear Plant EMERGENCY PLAN Rev. 11 TRAINING Page 4 of 48 V. TRAINING OBJECTIVES

1. Classify emergency events.

2. Recognize the reasons for having the Radiological Emergency Plan (REP).

3. Identify the functions of the onsite emergency response facilities.

4. Formulate Protective Action Recommendations (PARs).

5. Use the WBN Emergency Plan Implementing Procedures (EPIPs).

6. State three Site Emergency Director responsibilities that cannot be delegated.

7. Identify Operations responsibilities for the following emergency response positions:

• Site Emergency Director (who is initially the SM)

• Operations Manager in the TSC

• Control Room Communicator in the Control Room

• Operations Communicator in the TSC

• OSC Operations Advisor

• Operations emergency response team assignments

• NOMS Logkeeper in the Control Room (when available)

• Technical Advisor

• Designated Phone Talker

8. Recognize how AUOs are dispatched and controlled during radiological emergencies.

9. Recognize REP communications guidelines (OPDP-1).

10. Demonstrate effective communication techniques used in emergency response.

11. Identify lessons learned from TVA/industry events, drills and exercises.

1 2. Recall where radios can and cannot be used at WBN (BP-364).

13. Use the Integrated Computer System (ICS).

14. Identify all locations where the Emergency Paging System (EPS) may be activated from and demonstrate the use of the EPS to include the printed report from the TSC.

1 5. Using WBN EPIPs 2, 3, 4, and 5, recognize who is responsible to activate the Emergency Paging System.

1 6. Recognize conditions which constitute activation of the emergency response facilities regardless of the time of day when an emergency has been declared.

17. Identify and use the back-up Emergency Response Organization call lists used when the Emergency Paging System has failed.

Watts Bar RADIOLOGICAL 3-OT-PCD-048C Nuclear Plant EMERGENCY PLAN Rev. 11 TRAINING Page 5 of 48 V. TRAINING OBJECTIVES (continued) 1 8. Recognize entry conditions for Severe Accident Management Guidelines (SAMGs).

19. Use the Radiological Emergency Notification Directory (REND).

20. Use the Satellite Phone to make calls during emergencies.

21. Identify the WBN REP procedure addressing MERT responsibilities, offsite agreement support, and emergency phone numbers.

22. Review Operations drill critique items.

23. Perform dose assessments using ICS for WBN EPIP-1 3.

24. Interpret MET data obtained in the TSC from the CECC computer.

25. Identify specific actions of OSC Emergency Responders in the OSC teams staging area (EPT 309.000).

26. Understand the critical times associated with

• Event Declaration

• Offsite Notification S Facility Staffing

• Printed EPS Report

Clarification GuidancefGr SRO-only Questions Rev 1(03/1112010)

F. Procedures and limitations involved in initial core loading, alterations in core configuration, control rod rogramminq, and determination of various internal and external effects on core reactivity. [10 CFR 55.43(b)(6)]

Some examples of SRO exam items for this topic include:

• Evaluating core conditions and emergency classifications based on core conditions.

• Administrative requirements associated with low power physics testing processes.

• Administrative requirements associated with refueling activities, such as approvals required to amend core loading sheets or administrative controls of potential dilution paths and/or activities.

• Administrative controls associated with the installation of neutron sources.

• Knowledge of TS bases for reactivity controls.

G. Fuel handling facilities and irocedures. [10 CFR 55.43(b)(7)]

Some examples of SRO exam items for this topic include:

• Refuel floor SRO responsibilities.

• Assessment of fuel handling equipment surveillance requirement acceptance criteria.

• Prerequisites for vessel disassembly and reassembly.

• Decay heat assessment.

• Assessment of surveillance requirements for the refueling mode.

• Reporting requirements.

• Emergency classifications.

This does not include items that the RO may be responsible for at some sites such as fuel handling equipment and refueling related control room instrumentation operability requirements, abnormal operating procedure immediate actions, etc. For example, an RO is required to stop the refueling process when communication is lost between the control room and the refueling floor, therefore, this is a task that is both an RO and SRO responsibility and is not SRO-only.

Page 9 of 16

Clarificatian Guidance for SRO-only Qiestions Rev 1(0311112010)

C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]

Some examples of SRO exam items for this topic include:

• 10 CFR 50.59 screening and evaluation processes.

• Administrative processes for temporary modifications.

• Administrative processes for disabling annunciators.

• Administrative processes for the installation of temporary instrumentation.

• Processes for changing the plant or plant procedures.

Section IV provides an example of a satisfactory SRO-only question related tothistopic.

D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

[10 CFR 55.43(b)(4)]

Some examples of SRO exam items for this topic include:

• Process for gaseous/liquid release approvals, i.e., release permits.

• Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.

• Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.

SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.

E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)J This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.

The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:

Page 6 of 16

WBN 10-2011 NRC SRO Exam As Submitted 8115/2011

78. 026 AA2.06 078 Given the following:

- Unit 1 is operating at 100% power with the two components listed below out of service and tagged for maintenance:

- CCP lB-B

- CCSpump lB-B

- Component Cooling Water pump IA-A trips due to motor failure.

- AOl-I 5, Loss of Component Cooling Water (CCS), is implemented.

In accordance with AOl-i 5, which ONE of the following identifies...

(1) the maximum time the Reactor Coolant Pumps can be allowed to remain in service and (2) if the AOl-i 5 Attachments listed below require implementation?

Attachment 1 - Alignment of ERCW to CCP IA-A Lube Oil Coolers Attachment 2 - Alignment of CCS Train B to SEP HX B Max time Attachments A 10 minutes Only Attachment I performance is required.

B. 10 minutes Performance of both Attachments is required.

C. 12 minutes Only Attachment 1 performance is required.

D. 12 minutes Performance of both Attachments is required.

Page 6

WBN 10-2011 NRC SRO Exam As Submitted 811512011 DISTRACTOR ANAL YSIS:

A. Correct, AOl-15 has a Caution stating RCPs can be operated for up to 10 minutes after loss of CCS flow and during performance of the AOl, a step will direct the performance of Attachment 1, but the step directing the performance of Attachment 2 will not be performed because the if..then condition is not met due to the 2A header being available.

B. lncorrect Plausible because the time being 10 minutes is correct and both Attachments would be performed if the 2A header was not available.

C. Incorrect, Plausible because 12 minutes is a time in the procedure section being performed but it is the time that a CCP may survive (not the time required to remove the RCPs) and only Attachment I being performed is correct.

D. Incorrect, Plausible because 12 minutes is a time in the procedure section being performed but it is the time that a CCP may survive (not the time required to remove the RCP5) and both Attachments would be performed if the 2A header was not available.

Question Number: 78 Tier: 1 Group 1 KIA: 026 AA2.06 Loss of Component Cooling Water Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water:

The length of time after the loss of CCW flow to a component before that component may be damaged Importance Rating: 2.8*! 3.1*

IOCFRPart55: 43.5/45.13 IOCFR55.43.b: 5 KIA Match: The K/A is matched because the question requires knowledge of the length of time after the loss of CCW flow to the RCPs before they are required to be removed to prevent being damaged and is SRO because it requires the knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps to mitigate the damage due to the loss of Component Cooling Water.

Technical

Reference:

AOl-15, Loss of Component Cooling Water (CCS),

Page 7

WBN 10-2011 NRC SRO Exam As Submitted 8115/2011 Technical

Reference:

AOl-i 5, Loss of Component Cooling Water (CCS),

Revision 0032 Proposed references None to be provided:

Learning Objective: 3-OT-A011500

11. Demonstrate ability/knowledge of AOl, by:

a. Recognizing Entry conditions.

b. Responding to Actions.

c. Responding to Contingencies (RNO).

d. Responding to Notes/Cautions.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank question AOIi 500 002 modified for the WBN 10/2011 NRC exam.

Comments:

Page 8

WBN Loss of Component Cooling Water AOl-15 Unit I (CCS) Rev. 0032 Step Action/Expected Response Response Not Obtained 3.0 OPERATOR ACTIONS 3.1 Diagnostics NOTE The loss of CCS heat sink (e.g., loss of ERCW to CCS heat exchanger) should be evaluated as a loss of CCS flow (Subsection 3.2) with equipment and CCS temperatures monitored closely.

IF GOTO Page Loss of CCS flow, Subsection 3.2 6 OR Surge Tank level less than 60% or dropping uncontrolled.

Surge Tank level greater than 72% or rising Subsection 3.3 20 uncontrolled, OR CCS Rad Monitor alarm.

Loss of CCS due to loss of AC power train. AOl-35 Page 5of33

WBN Loss of Component Cooling Water AOI-15 Unit I (CCS) Rev. 0032 Step Action/Expected Response Response Not Obtained 3.2 Loss of CCS FlowIOut-leakage CHECK CCS pumps status:

a. CHECK any CCS pump a. **

GO TO CAUTION prior to TRIPPED or running pump NOT Step 2.

pumping forward:

• ERCW/CCS Motor tripout alarm,

• Low header pressure (Train A or B),

• Multiple low flow alarms.

t4

b. CHECK at least one U-i Train A b. START available U-i Train A puyve header supply pump RUNNING CCS pump. tv-J !

AND pumping forward:

• iA-A

• lB-B

c. CHECK any Train B header c. START available Train B CCS supply pump RUNNING AND pump pumping forward:

OR

• c-s

• 2B-B REFER TO SOI-70.D1, Component Cooling Water (CCS),

to align CCS pump 1 B to Train B header as necessary.

d. PLACE any non-operable or tripped CCS pump in STOP/PU LL-TO-LOCK.

Step continued on next page Page6of33

WBN Loss of Component Cooling Water AOl-15 Unit I (CCS) Rev. 0032 Step Action/Expected Response Response Not Obtained 3.2 Loss of CCS FlowIOut-leakage (continued)

1. (continued from previous page)

e. CHECK TWO U-i Train A header e. ENSURE at least one of the supply pumps RUNNING: following CLOSED to avoid

• IA-A excessive flow:

• 18-B

• RHR HXA, i-FCV-70-156, OR

• SEP HX A, O-ECV-70-197.

f. CHECK flows returned to f. **

GO TO CAUTION prior to NORMAL. Step 2.

g. CHECK A and B side Surge Tank g. IF Surge Tank level less than 57%,

levels between 57% and 85%. THEN GO TO CAUTION prior to Step 2.

IF Surge Tank level greater than 85%, THEN GO TO Subsection 3.3.

h. **

GOTO Step 15.

Pae7of33

WBN Loss of Component Cooling Water AOl-15 Unit I (CCS) Rev. 0032 Step Action/Expected Response Response Not Obtained 3.2 Loss of CCS FlowIOut-leakage (continued)

CAUTION A closed Surge Tank vent valve may cause a positive or negative tank pressure, giving an erroneous level indication.

2. CHECK i-FCV-70-66, Ui Surge Tank OPEN 1-FCV-70-55, UI Surge Tank Vent, OPEN. Vent.

3. IF Surge Tank level less than 57%,

THEN ENSURE 1-LCV-70-63, Ui Surge Tank Makeup LCV, OPEN (Refer to SOI-70.01 as required if makeup NOT available).

4. MONITOR A and B side Surge Tank STOP affected CCS pumps.

levels greater than 10%.

5. IF RHR Shutdown Cooling is in service, THEN GO TO AOI-14, Loss of RHR Shutdown Cooling.

Page 8of33

WBN Loss of Component Cooling Water AOl-15 Unit I (CCS) Rev. 0032 Step Action/Expected Response Response Not Obtained 3.2 Loss of CCS FlowIOut-leakage (continued)

CAUTION COP may survive for only 10 to 12 minutes after loss of COS to lube oil cooler.

6. MONITOR the following for Unit 1 OCS Train A:

• U-I OCS Train A level

• ERCW flow to OCS HX A IF loss of either is imminent, THEN **

GO TO Step 7.

PERFORM the following:

a. ENSURE COP lB-B is INITIATE alignment of ERCW to RUNNING. COP lA-A lube oil heat exchanger USING Attachment 1 (may use placard posted locally in COP room lA-A).

GO TO Substep c.

b. ENSURE COP lA-A is STOPPED.

c. ISOLATE charging and letdown.

Step continued on next page Page9of33

WBN Loss of Component Cooling Water AOI-15 Unit I (CCS) Rev. 0032 Step _Action/Expected Response Response Not Obtained 32 Loss of CCS FlowIOut-leakage (continued)

6. (continued from previous page)

d. STOP and LOCKOUT the following pumps:

TBBPsI-A&1-B,

• OCS pumps lA-A & lB-B,

• CSpump lA-A,

• RHRpump lA-A,

• SI pump lA-A,

• CCP lA-A.

yjvS.

CAUTION RCPs can be operated for up to 10 minutes after loss of CCS flow.

e. TRIP Reactor.

f. STOP RCPs.

g. GO TO E-0, Reactor Trip or Safety Injection, WHILE continuing this Instruction.

h. INITIATE alignment of ERCW to CCP lA-A lube oil heat exchanger USING Attachment 1 (may use placard placed locally in COP room lA-A).

Step continued on next page Page 10 of 33

WBN Loss of Component Cooling Water AOl-15 Unit I (CCS) Rev. 0032 Step Action/Expected Response Response Not Obtained 3.2 Loss of CCS FlowIOut-leakage (continued)

6. (continued from previous page)

CAUTION CCS should NOT be reestablished to RCP seals on a total loss of cooling due to probable damage to the seals. ECA-O.O, Loss of Shutdown Power, has guidance to isolate RCP seals.

i. IF CCS Train B is available AND WHEN ERCW cooling is aligned to COP lB-B is in service, THEN COP lA-A, THEN GO TO Substep k. EVALUATE performing the following based on time thermal barrier and ROP seal injection flow lost:

a. STARTCCP IA-A.

b. STOPCCP lB-B.

j. IF thermal barrier flow lost and RCP seal injection flow NOT reestablished, THEN REFER TO ECA-O.O, Loss of Shutdown Power, to isolate RCP seals.

k. IF CCS Train A, Unit 1 & 2, is NOT available, THEN ALIGN CCS Train B to SFP HX B USING Attachment 2.

Page 11 of 33

A011500002 Lkfl3N ArJ< ThsI1Or1)

Given the following plant conditions:

- AOI-15, Loss of Component Cooling Water (CCS), is in progress.

- BOTH trains of CCS flow indicate 0 gpm.

Per AOl-I 5, All RCP5 MUST be stopped...?

a. immediately upon loss of CCS flow to motor oil coolers.

b. immediately upon loss of CCS flow to RCP thermal barriers.

c. within 10 minutes of a total loss of CCS flow to motor oil coolers.

d. within 10 minutes of a total loss of CCS flow to RCP thermal barriers.

3-OT-AOII 500 Rev 11 Page 4 of 69 I. PROGRAM Watts Bar Operator Training II. COURSE A. License Training B. Non-License Training III. TITLE AOl-i 5, Loss of Component Cooling Water (CCS)

IV. LENGTH OF LESSON A. License Training 1.5 Hours B. Non-License Training 1.5 Hours V.TRAINING OBJECTIVES ARSS UORT 0 OA X X X X 1. Describe the Purpose/Goal of this AOl.

X X X 2. Identify Alarms associated with Loss of CCS.

X X X 3. Describe Auto Actions designed to compensate for loss of CCS.

X X X 4. Describe Action if Surge Tank Level is not maintained.

X X X X 5. Describe affect on plant operation if Surge Tank Level is not maintained.

X X X X 6. Determine the Purpose of AOl Appendix A.

X X X 7. Determine Action for Loss of an ESF Equipment header.

X X X X 8. Given a Rx Bldg hdr leak, determine components affected and actions to take upon header isolation.

3-OT-AQIl 500 Rev 11 Page 5 of 69 V. TRAINING OBJECTIVES (continued)

A RS S U ORT 0 OA X X X X 9. Given indications for a CCS Hx, determine if the Hx has a tube leak.

X X X 10. Give 3 sources of potential In-leakage to the CCS.

X X X 11. Demonstrate ability/knowledge of AOl, by:

a. Recognizing Entry conditions.

b. Responding to Actions.

c. Responding to Contingencies (RNO).

d. Responding to Notes/Cautions.

X X X X 12. Given a loss of Component Cooling Water is in progress demonstrate the process for performing NAUO actions for AOl-i 5 Attachment associated withALIGNMENT OF ERCW TO CCP lA-A LUBE OIL COOLERS.

VI. TRAINING AIDS A. Marker board & markers B. Multimedia/Overhead Projector(s)

VII. MATERIALS Attachments:

VIII. REFERENCES ENGINEERING SYSTEM DESCRIPTION(S)

Number Title Rev.

N3-70-4002 Component Cooling System (CCS) 15 A1 :1I Section Title Amend.

9.2.2 Component Cooling System 7

Clarification Guidance for SRO-only Questions Rev 1(0311112010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location?

No Can the question be answered solely by knowing immediate operator actions? Yes RD question No I

j Can the question be answered solely by knowing entry conditions for ADPs or plant parameters that require direct entry to major EOPs?

No Can the question be answered solely by knowing the purpose, overall sequence of events, or question overall mitigative strategy of a procedure?

No Does the question require one or more of the following?

o Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps SRO-only Knowledge of diagnostic steps and decision points in the Elestion EDPs that involve transitions to event specific sub-procedures or emergency contingency procedures o Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures

[No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

WBN 10-2011 NRC SRO Exam As Submitted 8/1512011

79. 038 EG2.4.18 079 Given the following:

- Unit 1 was operating at 100% power when a Safety Injection occurred due to a tube rupture in SG #2.

- The crew determined a Target lncore Temperature of 466°F and has initiated the rapid RCS cooldown in accordance with E-3, Steam Generator Tube Rupture.

- Before reaching the Target Incore Temperature, SG #2 pressure begins dropping in a uncontrolled manner.

Which ONE of the following identifies...

(1) the basis of cooling the RCS to a Target Incore Temperature of 466°F and (2) the action the SRO is required to take in accordance with the emergency instructions?

A. (1) Ensures adequate RCS subcooling is maintained after the subsequent RCS depressurization to the ruptured SG pressure determined from the E-3 table.

(2) Immediately transition to ECA-3.1, SGTR and LOCA Subcooled Recovery.

B (1) Ensures adequate RCS subcooling is maintained after the subsequent RCS depressurization to the ruptured SG pressure determined from the E-3 table.

(2) Continue in E-3 until the Target Incore Temperature is reached, then transition to ECA-3.1, SGTR and LOCA Subcooled Recovery.

C. (1) Provides the maximum amount of RCS temperature reduction without exceeding Pressurized Thermal Shock limits of the RCS prior to the RCS depressurization.

(2) Immediately transition to ECA-3.1, SGTR and LOCA Subcooled Recovery.

D. (1) Provides the maximum amount of RCS temperature reduction without exceeding Pressurized Thermal Shock limits of the RCS prior to the RCS depressurization.

(2) Continue in E-3 until the Target Incore Temperature is reached, then transition to ECA-3.1, SGTR and LOCA Subcooled Recovery.

Page 9

WBN 10-2011 NRC SRO Exam As Sub mitted 811512011

82. 003 AA2.02 082 Given the following:

- The Unit 1 reactor was at 85% when one Control Bank D, Group 2 rod dropped into the core.

- The crew has taken the actions in accord ance with AOl-2, Malfunction of Reactor Control System, and reduced reactor power to 73%.

- The control rod has been repaired.

- During recovery of the dropped rod, reac tor power increases to the maximum allowed before the rod is fully recovered.

- The operators reconnect the lift coils for the appropriate rods in order to reduce reactor power.

- Annunciator 86-A, CONTROL ROD UR GENT FAILURE, has NOT been reset.

- OAC postions 1-RBSS, ROD BANK SEL ECT, from the CBD to the MAN position.

Which ONE of the following identifies...

(1) which control rods, if any, will move when the IN-HOLD-OUT switch lever is placed to the IN position and (2) the basis for the reactor power red uction required by Tech Spec 3.1.5, Group Rod alignment Limits?

A. (1) No rod motion will occur.

(2) To ensure AFD remains within limits to prevent exceeding core design limits for hot channel factors.

B. (1) Bank D Group 2 rods ONLY.

(2) To ensure core design limits for loca l LHR are not exceeded due to the misaligned rod.

C (1) No rod motion will occur.

(2) To ensure core design limits for loca l LHR are not exceeded due to the misaligned rod.

D. (1) Bank D Group 2 rods ONLY.

(2) To ensure AFD remains within limits to prevent exceeding core design limits for hot channel factors.

Page 17

WBN 10-2011 NRC SRO Exam As Submitted 811512011 DISTRA CTOR ANAL YSIS:

A. lncorrect Plausible since the first part is correct with a Rod Urgent Failure alarm in all rod motion is blocked while in Auto or Manual. Also the second part is not correct for a reason in Tech Spec basis for a misaligned rod but plausible because AFD would be changing as the rod is being withdrawn during recovery of the rod B. Incorrect, Plausible since during the recovery of the dropped rod an Urgent Failure alarm is generated in power cabinet 2 BD due to having all lift coils disconnected but demanding a signal for movement while withdrawing the dropped rod. This will not prevent rod movement in the other group due to the position of the selector switch being in the CBD position. However with the selector switch in the MAN position all rod motion is stopped. Also the reason for 75% power is correct according to Tech Spec 3.1.5 bases.

C. Correc1, With the rod control selector switch in the MAN position and Rod Control Urgent Failure alarms will stop/prevent all rod motion. The candidate will have to determine the difference between system response based on the input signals from the bank selector switch. If the switch is positioned to CBD as is the case when the rod in recovered the power cabinet for the other group would receive an Urgent Failure Alarm and block rod movement for that group, however the other group would still function and allow the rod to be recovered. The candidate must recognize the position of the selector switch, and its affect on rod motion. Also per Tech Spec 3.1.5 bases, The reduction of power to 75% RTP ensures that local LHR increases due to a misaligned RCCA will not cause the core design criteria to be exceeded.

D. lncorrecl, Plausible since during the recovery of the dropped rod an Urgent Failure alarm is generated in power cabinet 2 BD due to having all lift coils disconnected but demanding a signal for movement while withdrawing the dropped rod. This will not prevent rod movement in the other group due to the position of the selector switch being in the CBD position. Also the second part is not correct for a reason in Tech Spec basis for a misallgned rod but plausible because AFD would be changing as the rod is being withdrawn during recovery of the rod.

Question Number: 82 Tier: 1 Group 2 K/A: 003 AA2.02 Dropped Control Rod Ability to determine and interpret the following as they apply to the Dropped Control Rod:

Signal inputs to rod control system Importance Rating: 2.7 I 2.8 Page 18

WBN 10-2011 NRC SRO Exam As Submitted 811512011 10 CFR Part 55: 43.5/45.13 IOCFR55.43.b: 2 KIA Match: This question matches the K/A by having the candidate determine the Urgent Failure Signals affect on rod control depending on the position of the rod bank selector switch during a dropped rod recovery. SRO by having the candidate recall from the Tech Spec bases the reason for reducing reactor power to 75% during a dropped rod recovery.

Technical

Reference:

Tech Spec 3.1.5 and bases AOl-2, Malfunction of Reactor Control System, Revision 0038 Proposed references None to be provided:

Learning Objective: 3-OT-SYSO85A

20. Differentiate between the Rod Urgent Failure and Non-Urgent Failure alarms. Explain the cause and effect of the alarms and how resetting of alarms is accomplished.

26. Discuss applicable Technical Specifications, Technical Requirements, and Bases.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank question and a McGuire question combined and modified for use at WBN Comments:

Page 19

WBN Malfunction of Reactor Control System AOl-2 linItI Rev. 0038 Step Action/Expected Response Response Not Obtained 1

3.3 Dropped RCCA (continued)

NOTE Boric acid requirements can be determined using Reactivity Briefing Sheet.

24. ALIGN RCCA to affected bank IF RCCA can NOT be aligned, position: THEN:

(p) USING rod control in Bank a. RECONNECT hft coils (toggle Select, position affected RCCA to down) of bank.

bank affected position recorded b. RESET CONTROL ROD URGENT in Step 18 FAILURE alarm [86-A] with

• (p) BORATE RCS to maintain 1-RCAR.

T-ave and T-ref within 3°F. c. SET affected group step counter to

• MAINTAIN less than or equal to original value.

75% Reactor power. d. RESET the computer to its original value USING the UPDATE function.

e. REFER TO Tech Specs 3.1.5, Rod Group Alignment Limits.

• 3.1.6, Shutdown Bank Insertion Limits.

3.1.7, Control Bank Insertion Limits.

• 3.1.8, Rod Position Indication.

f. NOTIFY Plant Management and Reactor Engineering.

g. RETURN TO Instruction in effect.

25. RECONNECT lift coils (toggle down) disconnected in Step 19.

Page 18of48

WBN Malfunction of Reactor Control System [AOI-2 Unit I IRev. 0038 Step Action/Expected Response Response Not Obtained 3.3 Dropped RCCA (continued)

26. ENSURE the following agree with values recorded in Step 18:

• Bank overlap counter

• Group step counters

• Computer points

27. RESET CONTROL ROD URGENT FAILURE alarm [86-A] using ROD CONTROL ALARM RESET pushbutton 1-RCAR.

28. PLACE control rods in MAN.

29. (p) RESTORE T-ave and T-ref to within 3°F.

NOTE Computer constant K0015 contains the current monthly full out rod position for all rod banks.

30. WHEN plant stabilized, THEN REFER TO 1-Sl-85-2, Reactivity Control Systems Movable Control Assemblies, for affected bank.

(Modes 1 and 2)

Page 19 of 48

WBN Malfunction of Reactor Control System AOl-2 Unit I Rev. 0038 Step Action/Expected Response Response Not Obtained 3.3 Dropped RCCA (continued)

CAUTION Allowing at least 5 minutes between any rod control input (i.e., T-ave, T-ref, or NIS) changes and placing rods in AUTO, will help prevent undesired control rod movement.

31. WHEN auto rod control desired, THEN:

a. ENSURE T-ave and T-ref within 1°F.

b. ENSURE zero demand on control rod position indication [1-M-4].

c. PLACE rods in AUTO.

32. RETURN TO Instruction in effect.

End of Subsection Page 20 of 48

Rd.Grcu Ai-ign4nntLimits 3.1.5 3.1 REACTIVITY CONTROL SYSTEMS 3.1.5 Rod Group Alignment Limits LCO 3.1.5 All shutdown and control rods shall be OPERABLE, with all individual indicated rod positions within 12 steps of their group step counter demand position.

APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more rod(s) A.1.1 Verify SDM is 1.6% 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> untrippable. k/k.

OR A.1.2 Initiate boration to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore SDM to within limit.

AND A.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 /> B. One rod not within B.l Restore rod to within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> alignment limits, alignment limits.

OR B.2.1.lVerify SDM is 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1.6% k/k.

OR (continued)

Watts BarUnit 1 3.18

Rod Group Alignment Limits 3.1.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2.l.2 Initiate boration to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore SDM to within limit.

AND 3.2.2 Reduce THERMAL POWER 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to < 75% RTP.

AND B.2.3 Verify 8DM is Once per 1.6% k/k 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AND B.2.4 Perform SR 3.2.1.1. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> AND B.2.5 Perform SR 3.2.2.1. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> AND B.2.6 Reevaluate safety 5 days analyses and confirm results remain valid for duration of operation under these conditions.

C. Required Action and C.l 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 B not met.

(continued)

Watts Bar-Unit 1 3.19

Rod Group Alignment Limits 3.1.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

0. More than one rod not 0.1.1 Verify SDM is 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> within alignment 1.6% k/k.

limit.

OR 0.1.2 Initiate boration to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore required SDM to within limit.

AND D.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 SURVE ILLANCE FREQUENCY SR 3.1.5.1 Verify individual rod positions within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> alignment limit.

AND Once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter when the rod position deviation monitor is inoperable SR 3.1.5.2 Verify rod freedom of movement 92 days (tripability) by moving each rod not fully inserted in the core 10 steps in either* direction.

(continued)

Watts Bar-Unit 1 3 .110

P4 Grop*Alignmeiit Limits 3.1.5 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.5.3 Verify rod drop time of each rod, from the Prior to fully withdrawn position, is < 2.7 seconds reactor from the beginning of decay of stationary criticality gripper coil voltage to dashpot entry, with: after initial fuel loading

a. Tavg 551°F; and and each removal of the

b. All reactor coolant pumps operating. reactor head Watts BarUnit 1 3.111

Rod Group Alignment Lnnfts B 3.1.5 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.5 Rod Group Alignment Limits BASES BACKGROUND The OPERABILITY (e.g., trippability) of the shutdown and control rods is an initial assumption in all safety analyses that assume rod insertion upon reactor trip. Maximum rod misalignment is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available SDM.

The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, Reactor Design, and GDC 26, Reactivity Control System Redundancy and Capability, (Ref. 1), and 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors (Ref. 2).

Mechanical or electrical failures may cause a control rod to become inoperable or to become misaligned from its group. Control rod inoperability or misalignment may cause increased power peaking, due to the asymmetric reactivity distribution and a reduction in the total available rodworth for reactor shutdown. Therefore, control rod alignment and OPERABILITY are related to core operation in design power peaking limits and the core design requirement of a minimum SDM.

Limits on control rod alignment have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.

Rod cluster control assemblies (RCCAs), or rods, are moved by their control rod drive mechanisms (CRDM5). Each CRDM moves its RCCA one step (approximately 5/8 inch) at a time, but at varying rates (steps per minute) depending on the signal output from the Rod Control System.

The RCCAs are divided among control banks and shutdown banks. Each bank be further subdivided into two groups to provide for precise reactivity control may (Shutdown Banks C and D have only one group each). A group consists (continued)

Watts Bar-Unit 1 B 3.1-24 Revision 51

Rod Group Alignment Limits B3.1.5 BASES BACKGROUND of two or more RCCAs that are electrically paralleled to (continued) step simultaneously. Except for Shutdown Banks C and D, a bank of RCCAs consists of two groups that are moved in a staggered fashion, but always within one step of each other. There are four control banks and four shutdown banks.

The shutdown banks are maintained either in the fully inserted or fully withdrawn position. The control banks are moved in an overlap pattern, using the following withdrawal sequence: When control bank A reaches a predetermined height in the core, control bank B begins to move out with control bank A. Control bank A stops at the position of maximum withdrawal, and control bank B continues to move out. When control bank B reaches a predetermined height, control bank C begins to move out with control bank B. This sequence continues until control banks A, B, and C are at the fully withdrawn position, and control bank D is approximately halfway withdrawn. The insertion sequene is the opposite of the withdrawal sequence. The control rods are arranged in a radially symmetric pattern, so that control bank motion does not introduce radial asymmetries in the core power distributions.

The axial position of shutdown rods and control rods is indicated by two separate and independent systems, Which are the Bank Demand Position Indication System (commonly called group step counters) and the Analog Rod Position Indication (ARPI) System.

The Bank Demand Position Indication System counts the pulses from the rod control system that moves the rods. There is one step counter for each group of rods. Individual rods in a group all receive the same signal to move and should, therefore, all be at the same position indicated by the group step counter for that group. The Bank Demand Position Indication System is considered highly precise (+/- I step or +/- 5/8 inch). If a rod does not move one step for each demand pulse, the step counter will still count the pulse and incorrectly reflect the position of the rod.

The ARPI System provides an accurate indication of actual control rod position, but at a lower precision than the step counters. This system is based on inductive analog signals from a series of coils spaced along a hollow tube with a center to center distance of 3.75 inches, which is six (continued)

Watts Bar-Unit 1 B 3.1-25 Revision 51

Rod Group Alignment Limits B3.1.5 BASES BACKGROUND steps. The normal indication accuracy of the ARPI System is (continued) +/- 6 steps (+/- 3.75 inches), and the maximum uncertainty is +/- 12 steps

(÷ 7.5 inches). With an indicated deviation of 12 steps between the group step counter and ARPI, the maximum deviation between actual rod position and the demand position could be 24 steps, or 15 inches.

APPLICABLE Control rod misalignment accidents are analyzed in the SAFETY ANALYSES safety analysis (Ref. 3). The acceptance criteria for addressing control rod inoperability or misalignment are that:

a. There be no violations of:

1. specified acceptable fuel design limits, or

2. Reactor Coolant System (RCS) pressure boundary integrity; and

b. The core remains subcritical after accident transients other than a main steam line break (MSLB).

Two types of misalignment are distinguished. During movement of a control rod group, one rod may stop moving, while the other rods in the group continue.

This condition may cause excessive power peaking. The second type of misalignment occurs if one rod fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition requires an evaluation to determine that sufficient reactivity worth is held in the control rods to meet the SDM requirement, with the maximum worth rod stuck fully withdrawn.

Three types of analysis are performed in regard to static rod misalignment (Ref. 4). The first type of analysis considers the case where any one rod is completely inserted into the core with all other rods completely withdrawn. With control banks at their insertion limits, the second type of analysis considers the case when any one rod is completely inserted into the core. The third type of analysis considers the case of a completely withdrawn single rod from a bank inserted to its insertion limit. Satisfying limits on departure from nucleate boiling ratio in both of these cases bounds the situation when a rod is misaligned from its group by 12 steps.

(continued)

Watts Bar-Unit 1 B 3.1-26

Rod Group Alignment Limits B3.1.5 BASES APPLICABLE Another type of misalignment occurs if one RCCA fails to insert upon a SAFETY ANALYSES reactor trip in response to a main steam pipe rupture and remains stuck (continued) fully withdrawn. This condition is assumed in the evaluation to determine that the required SDM is met with the maximum worth RCCA also fully withdrawn (Ref. 5). The reactor is shutdown by the boric acid injection delivered by the ECCS.

The Required Actions in this LCO ensure that either deviations from the alignment limits will be corrected or that THERMAL POWER will be adjusted so that excessive local linear heat rates (LHRs) will not occur, and that the requirements on SDM and ejected rod worth are preserved.

Continued operation of the reactor with a misaligned control rod is allowed if the heat flux hot channel factor (FQ(Z)) and the nuclear enthalpy hot channel factor(FH) are verified to be within their limits in the COLR and the safety analysis is verified to remain valid. When a control rod is misaligned, the assumptions that are used to determine the rod insertion limits, AFD limits, and quadrant power tilt limits are not preserved. Therefore, the limits may not preserve the design peaking factors, and FQ(Z) and FH must be verified directly using incore power distribution measurements. Bases Section 3.2 (Power Distribution Limits) contains more complete discussions of the relation of FQ(Z) and FH to the operating limits.

Shutdown and control rod OPERABILITY and alignment are directly related to power distributions and SDM, which are initial conditions assumed in safety analyses. Therefore they satisfy Criterion 2 of the NRC Policy Statement.

LCO The limits on shutdown or control rod alignments ensure that the assumptions in the safety analysis will remain valid. The requirements on OPERABILITY ensure that upon reactor trip, the assumed reactivity will be available and will be inserted. The OPERABILITY requirements also ensure that the RCCAs and banks maintain the correct power distribution and rod alignment.

The requirement to maintain the rod alignment to within plus or minus 12 steps is conservative. The minimum misalignment assumed in safety analysis is 24 steps (15 inches), and in some cases a total misalignment from fully withdrawn to fully inserted is assumed.

(continued)

Watts Bar-Unit 1 B 3.1-27 Revision 104 Amendment 82

-Rod Group-AlignmentUmits B3.1.5 BASES LCD some cases a total misalignment from fully withdrawn to (continued) fully inserted is assumed.

Failure to meet the requirements of this LCD may produce unacceptable power peaking factors and LHRs, or unacceptable SDMs, all of which may constitute initial conditions inconsistent with the safety analysis.

APPLICABILITY The requirements on RCCA OPERABILITY and alignment are applicable in MODES 1 and 2 because these are the only MODES in which neutron (or fission) power is generated, and the OPERABILITY (i.e., trippability) and alignment of rods have the potential to affect the safety of the plant. In MODES 3, 4, 5, and 6, the alignment limits do not apply because the control rods are bottomed and the reactor is shut down and not producing fission power. In the shutdown MODES, the OPERABILITY of the shutdown and control rods has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the RCS. See LCD 3.1.1, SHUTDOWN MARGIN (SDM) Tavg > 2OOF, for SDM in MODES 3 and 4, LCO 3.1.2, Shutdown Margin (SDM)-Tavg2OO°FforSDM in MODE 5, and LCD 3.9.1, Boron Concentration, for boron concentration requirements during refueling.

ACTIONS A.1.1 and A.1.2 When one or more rods are untrippable, there is a possibility that the required SDM may be adversely affected. Under these conditions, it is important to determine the SDM, and if it is less than the required value, initiate boration until the required SDM is recovered. The Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is adequate for determining SDM and, if necessary, for initiating boration to restore SDM.

In this situation, SDM verification must include the worth of the untrippable rod, as well as a rod of maximum worth.

(continued)

Watts Bar-Unit I B 3.1-28

Rod rouAignment Lmts B 3.1.5 BASES ACTIONS A2 (continued)

If the untrippable rod(s) cannot be restored to OPERABLE status, the plant must be brought to a MODE or condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

B.1 When a rod becomes misaligned, it can usually be moved and is still trippable. If the rod can be realigned within the Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, local xenon redistribution during this short interval will not be significant, and operation may proceed without further restriction.

An alternative to realigning a single misaligned RCCA to the group average position is to align the remainder of the group to the position of the misaligned RCCA. However, this must be done without violating the bank sequence, overlap, and insertion limits specified in LCO 3.1.6, Shutdown Bank Insertion Limits, and LCO 3.1.7, Control Bank Insertion Limits. The Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> gives the operator sufficient time to adjust the rod positions in an orderly manner.

B.2.1.1 and B.2.1.2 With a misaligned rod, SDM must be verified to be within limit or bóration must be initiated to restore SDM to within limit.

In many cases, realigning the remainder of the group to the misaligned rod may not be desirable. For example, realigning control bank B to a rod that is misaligned 15 steps from the top of the core would require a significant power reduction, since control bank D must be moved fully in and control bank C must be moved in to approximately 100 to 115 steps.

(continued)

Watts Bar-Unit 1 B 3.1-29

Rod Group Alignment Lrmits B 3.1.5 BASES ACTIONS B.2.1.1 and B.2.1.2 (continued)

Power operation may continue with one RCCA trippable but misaligned, provided that SDM is verified within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

The Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> represents the time necessary for determining the actual unit SDM and, if necessary, aligning and starting the necessary systems and components to initiate boration.

B.2.2, B.2.3, B.2.4, B.2.5, and B.2.6 For continued operation with a misaligned rod, RTP must be reduced, SDM must periodically be verified within limits, hot channel factors (FoR) and FH) must be verified within limits, and the safety analyses must be re-evaluated to confirm continued operation is permissible.

Reduction of power to 75% RTP ensures that local LHR increases due to a misaligned RCCA will not cause the core design criteria to be exceeded (Ref. 6).

The Completion Time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> gives the operator sufficient time to accomplish an orderly power reduction without challenging the Reactor Protection System.

When a rod is known to be misaligned, there is a potential to impact the SDM.

Since the core conditions can changewith time, periodic verification of SDM is required. A Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient to ensure this requirement continues to be met.

Verifying that F (Z) and FH are within the required limits ensures that current 0

operation at 75% RTP with a rod misaligned is not resulting in power distributions that may invalidate safety analysis assumptions at full power. The Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allows sufficient time to obtain an incore power distribution measurement and to calculate FQ(Z) and FtH.

Once current conditions have been verified acceptable, time is available to perform evaluations of accident analysis to determine that core limits will not be exceeded during a Design Basis Event for the duration of operation under these conditions. A Completion Time of 5 days is sufficient time to obtain the required input data and to perform the analysis.

(continued)

Watts Bar-Unit I B 3.1-30 Revision 104 Amendment 82

Rod Group Alignment Lknits -

B3.1.5 BASES ACTIONS B.2.2. B.2.3, B.2.4, B.2.5, and B.2.6 (continued) to obtain the required input data and to perform the analysis.

0.1 When Required Actions cannot be completed within their Completion Time, the unit must be brought to a MODE or Condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, which obviates concerns about the development of undesirable xenon or power distributions. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging the plant systems.

D.1.1 and D.1.2 More than one control rod becoming misaligned from its group average position is not expected, and has the potential to reduce SDM. Therefore, SDM must be evaluated. One hour allows the operator adequate time to determine SDM.

Restoration of the required SDM, if necessary, requires increasing the RCS boron concentration to provide negative reactivity, as described in the Bases of LCO 3.11. The required Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for initiating boration is reasonable, based on the time required for potential xenon redistribution, the low probability of an accident occurring, and the steps required to complete the action. This allows the operator sufficient time to align the required valves and start the boric acid pumps. Boration will continue until the required SDM is restored.

D.2 If more than one rod is found to be misaligned or becomes misaligned because of bank movement, the unit conditions fall outside of the accident analysis assumptions. Since automatic bank sequencing would continue to cause misalignment, the unit must be brought to a MODE or Condition in which the LCO requirements are not applicable.

(continued)

Watts Bar-Unit I B 3.1-31

Rod Group fgnment Limits-B 3.1.5 BASES ACTIONS D.2 (continued)

To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.5.1 REQUIREMENTS Verification that individual rod positions are within alignment limits at a Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> provides a history that allows the operator to detect a rod that is beginning to deviate from its expected position. If the rod position deviation monitor is inoperable, a Frequency of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> accomplishes the same goal. The specified Frequency takes into account other rod position information that is continuously available to the operator in the control room, so that during actual rod motion, deviations can immediately be detected.

SR 3.1.5.2 Verifying each control rod is OPERABLE would require that each rod be tripped.

However, in MODES 1 and 2, tripping each control rod would result in radial or axial power tilts, or oscillations. Exercising each individual control rod every 92 days provides increased confidence that all rods continue to be OPERABLE without exceeding the alignment limit, even if they are not regularly tripped.

Moving each control rod by 10 steps will not cause radial or axial power tilts, or oscillations, to occur. The 92 day Frequency takes into consideration other information available to the operator in the control room and SR 3.1.5.1, which is performed more frequently and adds to the determination of OPERABILITY of the rods. Between required performances of SR 3.1.5.2 (determination of control rod OPERABILITY by movement), if a control rod(s) is discovered to be immovable, but remains trippable and aligned, the control rod(s) is considered to be OPERABLE. At any time, if a (continued)

Watts Bar-Unit 1 B 3.1-32

RodGroup Alignment Limits B 3.1.5 BASES SURVEILLANCE SR 3.1.5.2 (continued)

REQUIREMENTS control rod(s) is immovable, a determination of the trippability (OPERABILITY) of the control rod(s) must be made, and appropriate action taken.

SR 3.1.53 Verification of rod drop times allows the operator to determine that the maximum rod drop time permitted is consistent with the assumed rod drop time used in the safety analysis. Measuring rod drop times prior to reactor criticality, after initial fuel loading and reactor vessel head removal, ensures that the reactor internals and rod drive mechanism will not interfere with rod motion or rod drop time, and that no degradation in these systems has occurred that would adversely affect control rod motion or drop time. This testing is performed with all RCPs operating and the average moderator temperature> 551°F to simulate a reactor trip under actual conditions.

This Surveillance is performed prior to initial criticality and during a plant outage, due to the plant conditions needed to perform the SR and the potential for an unplanned plant transient if the Surveillance were performed with the reactor at power.

REFERENCES 1. Title 10, Code of Federal Regulations, Part 50, Appendix A, General Design Criterion 10, Reactor Design, and General Design Criterion 26, Reactivity Control System Redundancy and Capability.

2. Title 10, Code of Federal Regulations, Part 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors.

3. Watts Bar FSAR, Section 15.0, Accident Analyses.

4. Watts Bar FSAR, Section 15.2.3, Rod Cluster Control Assembly Misalignment.

5. Watts Bar FSAR, Section 15.4.2, Major Secondary System Pipe Rupture.

(continued)

Watts Bar-Unit 1 B 3.1-33

QUESTIONS REPORT /Lçi) for ILT EXAM BANK MARCH 2007 IN J

2. A010200.05013 Given the following plant conditions:

,il ,

&tE:S 7 /

43,4)

- The Unit is at 98% power.

- One Control Bank D, Group 1 rod dropped fully into the core.

- The dropped rod recovery is in progress per AOI-2, Malfunction of Reactor Control System.

- The dropped rod is being withdrawn, resulting in reactor power increasing to 74%.

- To address the above conditions, operators have reconnected the lift coil(s) for the appropriate rod(s), per AOl-2.

- Annunciator 86-A, CONTROL ROD URGENT FAILURE, has not been RESET.

With 1-RBSS, Rod Bank Select, in the CBD position, which control rods, if any, will move if the In-Hold-Out switch lever is placed to the IN position prior to resetting the Urgent Failure? . 17 o

s. /0 a Group 1 rods, only.

b. Group 2 rods, only.

c. No rod motion will occur.

d. All Bank D rods.

a. CORRECT. When the rod was being recovered an Urgent Failure was generated in Bank D Group 2 because motion was demanded, and with all Group 2 lift coils disconnected no motion was sensed. Thus, after reconnecting the lift coils, the Urgent Failure is still present and prevents rod motion in Group 2.

b. Incorrect. In Individual Bank Select for Bank D, Group 2 rods are NOT capable of motion, since the Control Rod Urgent Failure alarm affected Power Cabinet 1 BD (the power cabinet for Group 2 rods). The Control Rod Urgent Failure alarm originated from the 1BD power cabinet. With all of the Group 2 lift coils disconnected, the 1 BD sensed an Urgent Failure when the Group 1 rod was withdrawn. Reconnecting the lift coils will not reset the Control Rod Urgent Failure alarm on Power Cabinet 1 BD.

c. Incorrect. Plausible, since the Control Rod Urgent Failure alarm does block rod movement for one of the groups in Bank D (Group 2), but Group 1 Control Bank D rods will move on demand.

d. Incorrect. Plausible, since applicant may fail to recall that the effect of the standing Control Rod Urgent Failure alarm is group specific until this alarm is reset Group 2 rods are blocked from movement. Group 1 rods WILL move, since they are on a separate power cabinet. An URGENT FAILURE exists on the 1 BD power cabinet due to the previous rod withdrawal.

Tuesday, May 17, 2011 4:01:54 PM 2

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross- Level RO SRO reference:

Tier# 1 Group# 2 K!A# 003 AA2.02 Importance Rating 2.8 Ability to determine and interpret the following as they apply to the Dropped Control Rod: Signal inputs to rod control system Proposed Question: SRO 82 Given the following Unit 1 initial conditions:

• Reactor power is at 40%

• Power range NIS indicate:

o 40% (N41), 41% (N42), 41% (N43), 41% (N44)

• Tave for each loop indicates:

o 567°F (N), 567°F (B), 568°F (C), 568°F (D)

• Turbine power is at 481 MWe

• Rod control is in automatic

• Group demand counters and DRPI indicate Control Bank D at 140 steps.

Control Bank D Rod L-12 drops fully into the core and the following conditions now exist:

• Power range NIS indicate:

o 40% (N41), 40% (N42), 42% (N43), 38% (N44)

• Tave for each loop indicates:

o 564°F (A), 564°F (B), 563°F (C), 564°F (D)

• Turbine power is 478 MWe Assuming NO operator action, which ONE of the following describes the effect on the rod control system, and the technical specification action required?

A. Rods withdraw due to the Tave-Tref mismatch. Verify Shutdown Margin requirements are met or initiate boration to ensure Shutdown Margin is met, to ensure accident analysis assumptions remain valid.

B. Rods withdraw due to the Power Range NIS Mismatch Rate signal. Verify Shutdown Margin requirements are met or initiate boration to ensure Shutdown Margin is met, to ensure accident analysis assumptions remain valid.

Page 208 of 260 Draft 7

ES-401 Sample Wntten Examination Form ES-401-5 Question Worksheet C. Rods withdraw due to Power Range NIS Mismatch Rate signal. Verify AFD requirements are met to ensure that fuel design limits and hot channel factors are maintained within limits.

D. Rods withdraw due to the Tave Tref mismatch. Verify AFD requirements are met to ensure that fuel design limits and hot channel factors are maintained within limits.

Proposed Answer: A Explanation (Optional):

A. Correct. Tave deviation is higher than 1 .5 degrees F and rods will withdraw. TS action is correct.

B. Incorrect. Power mismatch is not high enough to overcome the Tave mismatch, and power mismatch is based on rate of change with turbine power, which is minimal C. Incorrect. Incorrect bases and also incorrect reason for rod withdrawal.

Plausible because power mismatch is an input and AFD would be a concern above 50% power D. Incorrect. Incorrect basis but AFD would be a concern at higher power, as well as action required (>50%)

Technical OP-MC-lC-IRX, Rev 23 (Attach if not previously Reference(s) provided)

AP/14 Rev 10 AP-14 Basis Document Rev 6 TS 3:1.4 Proposed references to be provided to applicants during None examination:

Learning Objective: OP-MC-IRX-Obj 5 (As available)

Question Source: Bank # X Modified Bank (Note changes or attach

1. parent)

New Question History: Last NRC Exam 2002 McGuire

(

Page 209 of 260 Draft 7

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet

( Question Cognitive Memory or Fundamental Knowledge Level:

Comprehension or Analysis X 10 CFR Part 55 55.41 Content:

55.43 2 Comments:

Stem not modified but distractors all different from original KA met because inputs to rod control are the evaluated parameters.

SRO level because the effect of the failure has implications in TS basis that the applicant must determine Page 210 of 260 Draft 7

rviiori Page 4of57 I. PROGRAM Watts Bar Operator Training II. COURSES A. License Training B. Non-License Training Ill. TITLE Rod Control and Motor Generator Sets IV. LENGTH OF LESSON A. Licensed Training 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> B. Non-Licensed Training 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> V. TRAINING OBJECTIVES NOTOctives will be identified in the text as Objective RC1 etc.

A R S S U 0 R T 0 0 A X X X X 1. State the design basis (purpose) of the control rod drive system.

X X X X 2. State the number of RCCAs and their compositions.

X X X X 3. Identify the number of banks, groups per bank, and rods per group for the shutdown control rods.

X X X X 4. Describe the sequence of shutdown bank withdrawal or insertion including mode of control and speed.

X X X X 5. Identify the number of banks, groups per bank, and rods per group for the control banks.

X X X X 6. Describe how the rod drive mechanism moves rods on withdrawal, rest, or insertion.

X X X X 7. Describe the effects of normal control rod motion on RCS Tavg.

X X X X 8. Describe the controls for the control rods, including mode selector switch, speeds, and bank overlap.

X X X 9. Sketch the control rod drive control logic from the input signals to the cyclers.

X X X X 10. Identify and explain the input channels to the automatic rod control system.

X X X II. Explain how the rod control inputs serve to position the control rods on a given change in any one.

X X X X 12. Describe the operation of the rate comparator circuit.

X X X X 13. Discuss the purpose of the non-linear gain circuit.

X X X X 14. Discuss the purpose of the variable gain circuit.

rvisiuit ,

Page 5of57 A R S S U 9:-

0 0 A X X X X 15. Draw and explain the gull wing program.

X X X X 16. Briefly describe the purpose of each type of control rod system cabi net.

X X X X 17. Briefly explain how to start up the motor generator sets.

X X X X 18. Explain the purpose of the maintenance hold system for the control rod system.

X X X X 19. Describe the power supplies for the control rod drive system.

X X X 20. Differentiate between the Rod Urgent Failure and Non-Urgent Failure alarms. Explain the cause and effect of the alarms and how resetting of alarms is accomplished.

X X X X 21. List each of the rod control stops/interlocks and give its purpose.

X X X 22. For the rod position indicators, state the sources of signals, type of indication, and all alarms generated by each circuit.

X X X 23. Given a failure of the controlling input instrumentation for rod control and no operator action, describe the effects of rod motion on the plant, if any.

X X X X 24. Explain how a normal reactor trip occurs and how to perform an emergency reactor trip from outside the main control room.

X X X 25. Explain the bases, input, alarms, and operator actions relative to the rod insertion limits.

X X X 26. Discuss applicable Technical Specifications, Technical Requirements, and Bases.

rvisiurj Page 6 of 57 TRAINING OBJECTIVES MG SETS NOTE: Objectives will be identified in the text as Objective MG1 etc.

A R S S U 0 R T 0 0 A X X X X 1. Describe the power supply to the Control Rod Drive Mechanisms.

X X X X 2. Identify the power supply to the MG Sets.

X X X X 3. Explain what the bypass breakers are used for.

X X X X 4. Explain the function of the protective relaying equipment provided to each MG Set.

X X X X 5. Describe the position indication or annunciations the Operator has in the Main Control Room for the reactor trip and bypass breakers.

X X X X 6. Describe the daily routine checks an AUO makes on the MG Sets and CRD Equipment Room as specified in Electronic Logs.

X X X X 7. Explain how to place a MG Set in service.

X X X X 8. Explain how to take a MG Set out of service normally.

X X X X 9. Explain how to locally trip the reactor in the event of an ATWS.

- Clarification Guidance for -SRO-only Questions -

Rev 1(0311112010)

Figure 1: Screening for SRO-only linked to 10 CFR 5&43(b)(2)

(Tech Specs)

Can question be answered solely by knowing hour TS/TRM Action?

1 I Yes I RD question No Can question be answered solely by knowing the L0ITRM information listed above-the-line?

No Can question be answered solely by knowing the TS Safety Limits? RD queslion No Does the question involve one or more of the following for TS, TRM, 0rDDCM?

• Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)

• Application of generic LCD requirements (LCD 30.1 thru o

3.0.7 and SR 4.Oi thru 40.4)

Knowledge of TS bases that is required to analyze TS Yes I 1I SRO-only question required actions and terminology No I Question might not be linked to I 10 CFR 55.43(b)(2) for S RD-only PageS of 16

WBN 10-2011 NRC SRO Exam As Submitted 811512011

83. 028 AA2.07 083 Given the following:

- Unit 1 was operating at 100% power with 1-RBSS, ROD BANK SELECT, in MAN.

- 1-XS-68-339E, PZR LEVEL CONTROL CHANNEL SELECT, is selected to Ll-68-339 & 335.

- Performance of 1-Sl-68-33, Measurement Of Reactor Coolant Pump Seal Injection Flow, Section 6.2, Determination of Seal Leakage, is in progress.

- The CR0 has adjusted 1-HIC-62-89A, CHRG HDR-RCP SEAL FLOW CONTROL, as required and is now ready to record the seal injection flow rates for each of the RCPs.

- The Auctioneered High Tavg signal fails LOW.

Which ONE of the following identifies...

(1) how the RCP seal injection flow indication will respond due to the Auctioneered High Tavg signal failure and (2) the Bases for Tech Spec LCO 3.5.5 requiring performance of the test?

A (1) Remain the same (2) To ensure sufficient CCP flow to the RCS through ECCS injection lines during an accident.

B. (1) Remain the same (2) To ensure CCP flow to the RCP seals remains within 8-13 gpm after an actuation of the ECCS during an accident.

C. (1) Decrease (2) To ensure sufficient CCP flow to the RCS through ECCS injection lines during an accident.

D. (1) Decrease (2) To ensure CCP flow to the RCP seals remains within 8-13 gpm after an actuation of the ECCS during an accident.

Page 20

WBN 10-2011 NRC SRO Exam As Submitted 8/1512011 DIS TRACTOR ANALYSIS:

A. Correct, As identified in LCD 3.5.5 the surveillance requirement requires the pressurizer level control valve to be fully open to perform the test. This condition is established in the Surveillance Instruction by taking manual control of the level control valve and positioning it fully open for the test. While the Tavg signal (used to determine pressurizer program level setpoint) failure would normally cause the valve to close, the valve remain full open due to being in manual leaving RCP seal injection flow unaffected. Also, the bases background for T/S 3.5.5 Seal Injection Flow states The restriction on reactor coolant pump (RCP) seal injection flow limits the amount of ECCS flow that would be diverted from the injection path following an accident. (Also, see below)

B. Incorrect, Plausible because the seal injection flow remaining the same is correct and while the RCPs seal flow is designed to be maintained during an accident the bases is to limit the flow to the seals not to ensure the seals have flow.

C. Incorrect, Plausible because the seal injection flow dropping would be correct if the level control valve had been in automatic and the bases is correct.

D. Incorrect, Plausible because the seal injection flow dropping would be correct if the level control valve had been in automatic and while the RCPs seal flow is designed to be maintained during an accident the bases is to limit the flow to the seals not to ensure the seals have flow.

The intent of the LCD limit on seal injection flow is to make sure that flow through the RCP seal water injection line is low enough to ensure that sufficient centrifugal charging pump injection flow is directed to the RCS via the injection points (Ref. 2).

The LCD is not strictly a flow limit, but rather a flow limit based on a flow line resistance. In order to establish the proper flow line resistance, a pressure and flow must be known. The flow line resistance is determined by assuming that the RCS pressure is at normal operating pressure and that the charging pump discharge pressure is greater than or equal to the value specified in this LCO. The charging pump discharge header pressure remains essentially constant through all the applicable MODES of this LCO. A reduction in RCS pressure would result in more flow being diverted to the RCP seal injection line than at normal operating pressure. The valve settings established at the prescribed charging pump discharge header pressure result in a conservative valve position should RCS pressure decrease. The additional modifier of this LCD, the pressurizer level control valve being full open, is required since the valve is designed to fail open for the accident condition. With the discharge pressure and control valve position as specified by the LCD, a flow limit is established. It is this flow limit that is used in the accident analyses.

The limit on seal injection flow, combined with the charging pump discharge header pressure limit and an open wide condition of the pressurizer level control valve, must be met to render the ECCS OPERABLE. If these conditions are not met, the ECCS flow will not be as assumed in the accident analyses.

Question Number: 83 Tier: 1 Group 2 Page 21

WBN 10-2011 NRC SRO Exam As Submitted 8115/2011 K/A: 028 AA2.07 Pressurizer Level Control Malfunction Ability to determine and interpret the following as they apply to the Pressurizer Level Control Malfunctions:

Seal water flow indicator for RCP Importance Rating: 2.6 / 2.9 IOCFRPart55: 43.5/45.13 IOCFR55.43.b: 2 K/A Match: The question matches the K/A because it requires the ability to determine how a pressurizer level control system malfunction will affect the RCP seal water flow indications for the RCPs while the plant is in alignment to perform a Surveillance Requirement. SRO because it requires knowledge of the plant alignment requirements for performance of the Surveillance Requirement and also the bases of the applicable Tech Spec.

Technical

Reference:

1-Sl-68-33, Measurement Of Reactor Coolant Pump Seal Injection Flow, Revision 0012 Tech Spec 3.5.5 Bases Proposed references None to be provided:

Learning Objective: 3-OT-T!S0305

2. Determine the bases for each specification, as applicable, to the ECCS.

Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question for the WBN 10/2011 NRC exam.

Comments:

Page 22

WBN 10-2011 NRC SRO Exam As Submitted 8/1512011 DISTRA CTOR ANAL YSIS:

A. Incorrect, Plausible because establishing sufficient subcooling of the RCS so that the primary system will remain subcooled after the RCS pressure is decreased to stop the primary to secondary leakage is correct. Procedure transition plausible because a transition to ECA-3. I will be made but not until after the cooldown is complete and if any steam generator other than the ruptured steam generator had faulted, then an immediate transition would be required but the transition would be to E-2.

B. Correct, The step is to establish sufficient subcooling of the RCS so that the primary system will remain subcooled after the RCS pressure is decreased to stop the primary to secondary leakage. Procedurally if the ruptured steam generator pressure starts to drop uncontrolled during the cooldown, E-3 will be continued complete the cooldown to Target incore temperature and after the target temperature is reached step will address the need to make the transition to ECA-3. I.

C. Incorrect, Plausible because the RCS is being rapidly cooled and the bases discusses the concern for a PTS condition and how the target temperature table is built to preclude a PTS condition. Procedure transition plausible because a transition to ECA-3. I will be made but not until after the cooldown is complete and if any steam generator other than the ruptured steam generator had faulted, then an immediate transition would be required but the transition would be to E-2.

D. Incorrect, Plausible because the RCS is being rapidly cooled and the bases discusses the concern for a PTS condition and how the target temperature table is built to preclude a PTS condition. The second part is plausible because the correct procedure path is to continue the cooldown in E-3 and make the transition to ECA-3. I after the cooldown is complete.

Page 10

WBN 10-2011 NRC SRO Exam As Submitted 8115/2011 Question Number: 79 Tier: 1 Group 1 K/A: 038 EG2.4.18 Steam Generator Tube Rupture Knowledge of the specific bases for EOPs Importance Rating: 3.3 / 4.0 10 CFR Part 55: 41.10 / 43.1 I 45.13 IOCFR55.43.b: 5 K/A Match: This question matches the K/A by requiring the candidate to apply the basis for the step in E-3 to perform a rapid cooldown of the RCS.

SRO by requiring the knowledge of specific EOP step basis, and applying the information to make the correct procedure selection.

Technical

Reference:

WOG E-3 Background HP-Rev 2, Step 5 E-3, Steam Generator Tube Rupture, Revision 0023 Proposed references None to be provided:

Learning Objective: 3-OT-EOPO300

5. Given a set of plant conditions, use E-3, ES-3.1, ES-3.2, and ES-3.3 to correctly diagnose and implement: Action Steps, RNOs, Foldout Pages, Notes and Cautions.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank question EOPO300 010 modified for the for the WBN 10/2011 NRC exam Comments:

Page 11

WBN Steam Generator Tube Rupture E-3 Unit I Rev. 0023 Step Action/Expected Response Response Not Obtained

15. ENSURE major steam flowpaths ISOLATE secondary pathways to limit from the ruptured S/G isolated: depressurization and contamination by

a. TD AFW pump steam supply INITIATING Attachment 3 (E-3),

from Ruptured S/G CLOSED Steamline Isolation (MCR), and (if applicable). Attachment 4 (E-3), Steamline Isolation (Local).

b. Ruptured SIG MSIV and bypass valve CLOSED, OR Intact S/G MSIVs and bypass valves CLOSED.

16. CHECK Ruptured S/G pressure **

GO TO ECA-3.1, SGTR and greater than 690 psig. LOCA Subcooled Recovery.

2Qo.a

17. DETERMINE target incore temp for RCS cooldown:

IF Ruptured S/G pressure is between listed values, THEN USE lower value:

RUPTURED SIG TARGET INCORE TEMP (°F)

PRESSURE (PSIG) 1100 491°F[471°FADV]

1000 479°F [459°F ADV]

900 466°F [446°F ADV]

800 451°F [431°F ADVI 700 434°F [414°F ADVj 690 433°F [413°F ADVI Page 10 of 46

WBN Steam Generator Tube Rupture Unit I Rev. 0023 Step Action/Expected Response Response Not Obtained CAUTION

• The 1500 psig RCP trip criteria is NOT applicable during or after a controlled RCS cooldown and depressurization.

• If total feed flow CAPABILITY of 410 gpm is AVAILABLE, FR-H.1, Loss of Secondary Heat Sink, should NOT be implemented.

• Excessive steam dump cooldown rate will cause MSIV isolation due to the rate sensitive signal.

• If RCPs are NOT running, a false red or orange path may be indicated for FR-P.1 during the following steps. T-cold in the ruptured loop should be disregarded until Step 43.

18. INITIATE RCS cooldown to target incore temp, determined from Step 17.

a. DUMP steam to condenser from a. IF condenser steam dumps NOT Intact S/G(s) at maximum available, THEN achievable rate:

USE Intact S/G PORVs at IF dumps are in Tavg mode, maximum achievable cooldown THEN: rate.

1) PLACE steam dump controls OFF.

IF an Intact S/G is NOT available,

2) PLACE steam dump mode THEN switch in STEAM PRESSURE. PERFORM one BUT NOT BOTH of

3) ENSURE steam dump the following:

demand indicator 1-XI-1-33

• USE Faulted S/G, reading zero.

4) PLACE steam dump OR controls ON. * **

GO TO ECA-3.1, SGTR

5) PLACE steam dump LOCA Subcoole Recovery.

controller in MAN, AND FULLY OPEN three cooldown valves

( 25% demand). d. i)

Step continued on the next page ol Pagellof46 4 1 (L

WBN Steam Generator Tube Rupture E-3 Unit I Rev. 0023 Step Action/Expected Response Response Not Obtained

18. (continued)

b. WHEN RCS pressure is less than 1962 psig (P-Il), THEN:

• BLOCK low pzr pressure SI.

• BLOCK low steam pressure SI.

c. WHEN Tavg is less than 550°F (P-12), THEN BYPASS Lo-Lo Tavg interlock.

d. WHEN incore temp is less than target temp, THEN STOP RCS cooldown, AND MAINTAIN incore temperature less than or equal to target.

e. CONTINUE with Step 19 of this Instruction.

19. MONITOR Intact S/G levels:

a. At least one S/G NR level a. ENSURE feed flow greater than 29% [39% ADV]. greater than 410 gpm.

b. S/G NR levels less b. IF NR level in any unisolated S/G than 50% and controlled. continues to rise with no feed flow, THEN STOP RCS cooldown, AND GO TO Step 2.

Page 12 of 46

WBN Steam Generator Tube Rupture E-3 Unit I Rev. 0023 Step Action/Expected Response Response Not Obtained

20. CONTROL Intact S/G NR levels between 29% and 50%

[39% and 50% ADV].

21. MONITOR pzr PORVs and block valves:

a. Pzr PORVs CLOSED. a. WHEN RCS pressure less than 2335 psig, THEN ENSURE pzr PORV or associated block valve CLOSED.

IF PORV fails open AND associated block valve can NOT be closed, THEN GO TO ECA-3.1, SGTR and LOCA Subcooled Recovery.

b. At least one block valve OPEN. b. OPEN one block valve UNLESS it was closed to isolate an open PORV.

22. CHECK pzr safety valves CLOSED: IF RCS pressure is less than 2485 psig,

• EVALUATE tailpipe and pzr safety valve failed open, THEN temperatures and acoustic **

monitors. GO TO ECA-3.1, SGTR and LOCA Subcooled Recovery.

Page 13 of 46

WBN Steam Generator Tube Rupture E-3 Unit I Rev. 0023 Step Action/Expected Response Response Not Obtained CAUTION If offsite power is lost after SI reset, manual action will be required to restart the SI pumps and RHR pumps due to loss of SI start signal.

23. RESET SI, AND NOTIFY IMs to block auto SI USING IMI-99.040, Auto SI Block.

CHECK the following:

• SI ACTUATED permissive DARK.

• AUTO SI BLOCKED permissive LIT.

24. RESET Phase A and Phase B.

25. ENSURE cntmt air in service:

a. Aux air pressure greater a. DISPATCH Operator to than 75 psig [M-15]. aux air compressors:

1) ENSURE affected compressor(s) running.

2) ENSURE affected train isolation valve CLOSED:

• Train A, O-FCV-32-82.

• Train B, O-FCV-32-85.

b. Cntmt air supply valves OPEN

[M-1 5]:

• I -FCV-32-80.

• 1-FCV-32-102.

• I-FCV-32-IIO.

Page 14 of 46

WBN Steam Generator Tube Rupture E-3 Unit I Rev. 0023 Step Action/Expected Response Response Not Obtained

26. DETERMINE if RHR pumps should be stopped:

a. CHECK RHR suction a. **

GO TO Step 27.

aligned from RWST.

b. CHECK RCS pressure b. ENSURE RHR pumps RUNNING.

greater than 150 psig.

GO TO Step 27.

c. CHECK RCS pressure c. ENSURE CCS aligned to RHR heat stable or rising, exchanger:

• 1-FCV-70-153 OPEN

• 1-FCV-70-156 OPEN.

CLOSE SFP heat exchanger A CCS supply 0-FCV-70-1 97.

GO TO Step 27.

d. STOPRHRpumpsAND PLACE in A-AUTO.

e. MONITOR RCS pressure e. Manually RESTART RHR pumps.

greater than 150 psig.

27. CHECK target incore temperature:

a. VERIFY incore temperature a. DO NOT CONTINUE this instruction less than target temperature. UNTIL incore temperature less than target temperature.

b. STOP RCS cooldown.

c. MAINTAIN incore temperature less than target temperature.

Page 15 of 46

WBN Steam Generator Tube Rupture E-3 Unit I Rev. 0023 Step Action/Expected Response Response Not Obtained

28. MONITOR Ruptured SIG pressure MAINTAIN Ruptured S/G at least stable or rising. 250 psig greater than the pressure of the SIG(s) used for cooldown:

• Slowly DUMP steam from S/G(s) used for cooldown.

• MAINTAIN RCS cooldown rate less than 1000 F in one hour.

IF the Ruptured S/G depressurizes to less than 250 psig above the pressure of the S/G(s) used for cooldown, THEN GOTOECA-3.1,SGTRand LOCA Subcooled Recovery.

29. CHECK RCS subcooling IF subcooling is less greater than 85°F [1 05°F ADV]. than 65°F [85°F ADVJ, THEN GO TO ECA-3.1, SGTR and LOCA Subcooled Recovery.

IF subcooling is STABLE OR DROPPING, THEN GO TO ECA-3.1, SGTR and LOCA Subcooled Recovery.

DO NOT CONTINUE this instruction UNTIL subcooling is greater than 85°F [1 05°F ADV].

Page 16 of 46

STEP DESCRIPTION TABLE FOR E-3 Step 6 STEP: Initiate RCS Cooldown PURPOSE: To establish sufficient subcooling in the RCS so that the primary system will remain subcooled after pressure is decreased to stop primary-to-secondary leakage.

BASIS:

The principal goal of the E-3 guideline is to stop primary-to-secondary leakage and to establish and maintain sufficient indications of adequate coolant inventory. These indications include a pressurizer level indication to trend coolant inventory and RCS subcooling to ensure that the indicated pressurizer level is reliable. This step is designed to establish sufficient subcooling in the RCS so that the primary system will remain subcooled after RCS pressure is decreased in subsequent steps to stop primary-to-secondary leakage.

Since, in order to stop this leakage, the RCS pressure must be decreased to a value equal to the ruptured steam generator pressure, the temperature at which this cooldown is terminated is dependent upon the ruptured steam generator pressure. A table should be constructed for various ruptured steam generator pressures showing the fluid temperature corresponding to 20°F subcooling at each of these pressures, including allowances for subcooling uncertainties with normal or adverse containment conditions. The cooldown should be based on the core exit TCs since these also provide the input for SI termination and reinitiation. The 20°F subcooling is provided as operating margin to accommodate fluctuations in RCS temperature, perturbations in ruptured steam generator pressure, interpolation between listed ruptured steam generator pressures, and overshoot during RCS depressurization.

As previously demonstrated (see Step 3), the pressure of the intact steam generators must be maintained less than the pressure of the ruptured steam generators in order to maintain RCS subcooling. Since flow from the ruptured steam generator should be isolated, this pressure differential is established by dumping steam only from the intact steam generators. Steam dump to the condenser is preferred to minimize radiological releases and conserve feedwater supply. However, the PORVs on the intact steam generators provide an alternative steam release path. If no intact steam generator is available, RCS temperature should be controlled by adjusting feed flow to a faulted steam generator or by releasing steam from a ruptured steam generator. This latter method will result in continued primary-to-secondary leakage and is best handled in ECA-3.l, SGTR WITH LOSS OF REACTOR COOLANT-SUBCOOLED RECOVERY DESIRED.

E-3 Background 76 HP-Rev. 2, 4/30/2005 HE3BG.doc

STEP DESCRIPTION TABLE FOR E-3 Step 6 ACTIONS:

o Determine required core exit temperature o Dump steam to condenser at maximum rate o Dump steam from intact SG PORV5 at maximum rate o Control feed flow to faulted SG to cooldown RCS o Control steam release and feed flow to stabilize RCS temperature when required temperatures are reached o Transfer to ECA-3.1 INSTRUMENTATION:

o SG pressure indication o Core exit ICs o Main steamline isolation and bypass valve position indications o Condenser status indications o Steam dump valve position indication o SG PORVs position indication CONTROL/EQUIPMENT:

o Steam dump valves o SG PORV5 o Feed flow control valve o Plant specific controls to dump steam from intact SGs by other means E-3 Background 77 HP-Rev. 2, 4/30/2005 HE3BG .doc

STEP DESCRIPTION TABLE FOR E-3 Step 6 KNOWLEDGE:

o It is not intended for the operator to reevaluate the required core exit temperature or precisely interpolate between values listed in the table.

o When the required core exit temperature is reached, the intact steam generator pressure (or feed flow to a faulted steam generator) should be controlled to maintain that temperature.

o Cooldown of the RCS should be completed before continuing in the guideline.

o Natural circulation flow in the ruptured loops may stagnate during this cooldown. The hot leg temperature in that loop may remain significantly greater than the intact loops. In addition, safety injection flow into the cold leg may cause the cold leg fluid temperature to decrease rapidly in that same loop. Steps to depressurize the RCS and terminate SI should be performed as quickly as possible after the cooldown has been completed to minimize possible pressurized-thermal shock of the reactor vessel.

o RCS cooldown should proceed as quickly as possible and should not be limited by the 100°F/hr Technical Specification limit. Integrity limits should not be exceeded since the final temperature will remain above 350°F.

o The RCP trip criteria (Step 1) does not apply after a controlled cooldown is initiated.

o If more than one steam generator is ruptured, the lowest ruptured steam generator pressure should be used to determine the required core exit temperature. If cooldown to a target core exit temperature is already in progress when a subsequent SGTR is diagnosed the operator should stop the cooldown until the subsequent ruptured steam generator is isolated since continuing the cooldown would lower the pressure in the newest ruptured steam generator and result in unnecessary releases prior to its isolation from the intact steam generators. The target core exit temperature should be reexamined to determine if the temperature should be reduced based on the subsequent ruptured steam generator pressure. If a RCS depressurization is in progress, although it does not impact the pressure in the newest ruptured steam generator, for the sake of simplicity it should be stopped and the plant stabilized by the operator until the newest ruptured steam generator is isolated.

E-3 Background 78 FtP-Rev. 2, 4/30/2005 HE3BG .doc

STEP DESCRIPTION TABLE FOR E-3 Step 6 PLANT-SPECIFIC INFORMATION:

o If no intact SG is available, the operator must decide between feeding a faulted SG or steaming a ruptured SG for RCS cooldown to RHR conditions. One must weigh the concerns of reactor vessel thermal stresses, increased discharge to containment, and stresses on the SG tubes against increased radiological releases from the ruptured SG and the potential for SG overfill on an event specific basis. Refer to Section 3.2, Key Utility Decision Points.

o For some plant designs, the probability of having no intact steam generators available may be sufficiently small to warrant removal of the associated contingency actions. The benefit of these actions should be weighed against the increased burden on operator training and complexity of the E-3 guideline.

o Alternative means of dumping steam from the intact steam generators, such as steam flow to the turbine-driven AFW pumps, should be evaluated on plant specific basis.

o (0.05) SG saturation pressure to preclude a PTS condition, including allowances for normal channel accuracy and post accident transmitter errors for pressure instrument. Refer to Background Document for E-3.

o (G.Ol) Temperature corresponding to 20°F subcooling at the ruptured steam generator pressure, including allowances for normal channel accuracy.

Allowances for normal channel accuracy should be based on RCS subcooling uncertainty.

o (G.02) Temperature corresponding to 20°F subcooling at the ruptured steam generator pressure, including allowances for normal channel accuracy and post-accident transmitter errors, not to exceed 100°F.

Allowances for normal channel accuracy and post accident transmitter errors should be based on RCS subcooling uncertainty.

E-3 Background 79 HP-Rev. 2, 4/30/2005 HE3BG.doc

STEP DESCRIPTION TABLE FOR E-3 Step 5 Check Ruptured SG(s) Pressure - GREATER THAN (0.05) PSIG PURPOSE: o To identify a secondary side break in the ruptured steam generator and transfer the operator to the appropriate contingency guideline o To minimize possible pressurized thermal shock of the reactor vessel due to rapid cooldown below 350°F in subsequent steps BASIS:

Subsequent steps direct the operator to dump steam from the intact steam generators to cool the RCS as rapidly as possibly in order to establish adequate subcooling margin. The temperature at which this cooldown is terminated depends on the pressure in the ruptured steam generators. If this pressure is less than (0.05) psig, this cooldown could result in an ORANGE priority on the Integrity Status Tree. To avoid this condition the operator is transferred to ECA-3.1, SGTR WITH LOSS OF COOLANI-SUBCOOLED RECOVERY DESIRED, which limits the cooldown rate to less than 100°F/hr.

A ruptured steam generator pressure less than the saturation pressure corresponding to a temperature for precluding pressurized thermal shock (PTS) conditions is also a possible indication of a steam break associated with the affected steam generator. For such an event, the ECA-3.1 guideline is more appropriate since primaryto-secondary leakage cannot be terminated until cold shutdown.

The basis for determining footnote (0.05) starts with determining a temperature that will preclude PTS since the cooldown rate of 100°F can be exceeded in E-3. Violating the PTS limitation would transition the operator to FR-P.2 and stop the cooldown, which contradicts the instructions provided in Step 6b of E-3. To prevent this occurrence, footnote (1.02) should be selected as the starting temperature. To this value, an assumed 40°F temperature rise across the core and the uncertainty of the core exit temperature indication should be added to the value. If this value exceeds 350°F, it should be used as the initial temperature input. If this value does not exceed 350°F, then 350°F should be used as the minimum initial temperature input. A 20°F margin should then be added to this initial temperature value, along with the uncertainty in the RCS subcooling indication. This temperature should then be converted to a saturation pressure, and the uncertainty in SG pressure indication considered.

E-3 Background 69 HP-Rev. 2. 4/30/2005 HE3BG.doc

STEP DESCRIPTION TABLE FOR E-3 Step 5 A pressure based on this temperature input was chosen to prevent unnecessary transitions from E-3 at higher pressures when it is still desirable to continue with E-3 and to minimize possible pressurized thermal shock of the reactor vessel. Since there is no check on the reactivity condition, there is no guarantee that return to criticality will not occur during RCS cooldown for plants with BIT removed or boron concentration reduced. A plant specific evaluation may be required to determine the optimized RCS temperature used as the basis for Footnote (0.05) for plants with BIT removed or reduced BIT boron concentration.

Under the unlikely case that recriticality occurs, the RCS cooldown would result in a challenge to the Critical Safety Functions, i.e., a criticality condition on the Subcriticality Status Tree. The operator will be directed to the Guideline FR-S.1, RESPONSE TO NUCLEAR POWER GENERATION/ATWS, or the Guideline FR-S.2, RESPONSE TO LOSS OF CORE SHUTDOWN, to initiate emergency boration of the RCS and obtain adequate shutdown margin. After the adequate shutdown margin is assured, the operator will be directed to go back to E-3 Guideline to continue the recovery actions.

ACTIONS:

o Check ruptured SG pressure o Transfer to ECA-3.l INSTRUMENTATION:

SG pressure indication CONTROL/EQUIPMENT:

N/A KNOWLEDGE:

N/A PLANT-SPECIFIC INFORMATION:

(0.05) SG saturation pressure to preclude a PTS condition, including allowances for normal channel accuracy and post accident transmitter errors for pressure instrument. Refer to Background Document for E-3.

E-3 Background 70 HP-Rev. 2, 4/30/2005 HE3BG .doc

1\ jvW) 6cAtc77V/J After a ruptured SIG is isolated, E-3, Steam Generator Tube Rupture, directs an RCS cooldown to a specific target Incore TC temperature, derived from a fable In E-3.

Which of the following is the BASIS for the target temperature from the E-3 table?

a. Allows a maximum amount of RCS temperature reduction without exceeding the Pressurized Thermal Shock limits.

b. Minimizes back leakage from the ruptured S/G until the subsequent RCS depressurization can be initiated.

c. Ensures adequate subcooling (including instrument inaccuracies) is maintained during the subsequent RCS depressurization.

d. Prevents void formation in the S/G tubes when depressurizing the RCS with Aux Spray or Prz PORVs.

3-OT-EO P0300 Revl3 Page4of 121 pages I. PROGRAM:

Watts Bar Operator Training II. COURSE:

A. License Training B. License Operator Requalification III. TITLE:

E-3, Steam Generator Tube Rupture IV. LENGTH OF LESSON:

A. License training 3 Hours License operator REQUAL time will be determined after objectives are identified.

V. TRAINING OBJECTIVES:

000< I D .

(/) Cl)

X X X 1. Explain why timely operator response is important in mitigating the effects of a SGTR accident.

X X X 2. Given a set of plant conditions, the operator will be able to identify which SGs, if any, are ruptured by evaluating the symptoms of a ruptured SG.

X X X 3. Describe the major actions of E-3.

X X X 4. Explain the basis for controlling the ruptured SG NR level greater than 29%.

X X X 5. Given a set of plant conditions, use E-3, ES-3.1, ES-3.2, and ES-3.3 to correctly diagnose and implement: Action Steps, RNOs, Foldout Pages, Notes and Cautions.

X X X 6. Explain the basis for cooling the RCS to a target incore temp prior to depressurization of the RCS.

3-OT-EO P0 300 Rev 13 Page5of 121 pages V. TRAINING OBJECTIVES: (continued) 0 0 0 D U)

U)

X X X 7. Given a set of plant conditions, including ruptured SG press, determine the target incore temp for RCS coo ldown.

X X X 8. Given a set of plant conditions, evaluate the conditions to determine if natural circulation exists and take appropriate action to initiate, restore, or maintain natural circulation.

X X X 9. Describe the most effective method of collapsing a steam bubble in the reactor vessel head, SOER 83-02, recommendations 13c and 13b.

X X X 10. Describe the action(s) taken if a RCP cannot be restarted to help cooldown and depressurize, SOER 83-02, recommendation 13d.

X X X 1 1. Explain why it is especially important to monitor Shutdown Margin while cooling down using procedure ES-3.1.

X X X 12. Explain why it is undesirable for the safety valves on a ruptured steam generator to open during a tube rupture event and explain how the possibility of their opening is reduced, SOER 83-2, recommendation 15.

X X X 13. Explain why the cold leg accumulators are isolated when RCS press drops to less than 1000 psig (assuming RCS subcooling and inventory requirements are met).

X X X 14. Describe the advantages and disadvantages of ES-3.1, Post SGTR Cooldown Using Backfill, (SOER 83-02, recommendation 14).

3-OT-EOPO300 Rev 13 Page6of 121 pages V. TRAINING OBJECTIVES: (continued) 0 0 0 D 0::

Cl) C,)

X X X 15. Describe the consequences of letting the RCS go solid (i.e., excessive use of Safety Injection) during a steam generator tube rupture (SOER 83-02, recommendation 13a).

X X X 16. Explain why it is important to cooldown to. Cold Shutdown as quickly as possible (<1 00°F/hr) when performing procedure ES-3.2 or ES-3.3.

X X X 17. Deleted.

CtarificaUo Guidance fr SRO-only Questions Rev 1(0311112010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location? -,.question No Can the question be answered solely by knowing 1 Yes I I immediate operator actions?

No

] RO question Can the question be answered solely by knowing entry conditions for AOPs or plant parameters question that require direct entry to major EQ Ps?

No Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

No Does the question require one or more of the following?

• Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps SRQ-only

• Knowledge of diagnostic steps and decision points in the estion EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No I Question might not be linked to I 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

WBN 10-2011 NRC SRO Exam As Submitted 811512011

80. 056 AA2.74 080 Given the following:

- Unit 1 is operating at 100% power with 1-FCV-68-332, BLOCK VALVE FOR PORV 334, closed as required by Tech Specs due to PZR PORV 334 being inoperable but capable of being cycled.

- The following sequence of events occur:

1300 - Both 161 kV Offsite power supplies are lost.

1400 - Annunciator 91-A, PZR PORV/SAFETY OPEN, alarms due to PORV 340A opening and the PORV sticks open in mid-position during a pressure transient.

1401 - TheOACreports:

1-TI-68-331, PORV 340A & 334 TAILPIPE TEMP, rising, PORV 340A GREEN and RED indicating lights DARK, and 1 -FCV-68-333A, BLOCK VALVE FOR PORV 340A has been closed.

1500 - Both offsite power supplies are restored.

Which ONE of the following identifies...

(1) why the indicating lights on 1 -HS-68-340A, PZR PORV 340AA, are DARK at 1401 and (2) if Tech Specs allow continued operation in Mode 1 for an unlimited period of time with the current status of the pressurizer PORVs?

A. (1) Due to the loss of offsite power.

(2) Continued operation allowed.

B. (1) Due to the loss of offsite power.

(2) Continued operation NOT allowed.

C. (1) Due to the valve being at mid-position.

(2) Continued operation allowed.

D (1) Due to the valve being at mid-position.

(2) Continued operation NOT allowed.

Page 12

WBN 10-2011 NRC SRO Exam As Submitted 8/1512011 DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because there are circuits that would not have power while offsite power was lost. Also, plausible because there are conditions with both PORVs failed and isolated that will allow the unit to continue to operate in Mode I for an unlimited time period.

B. lncorrect Plausible because there are circuits that would not have power while offsite power was lost. Also plausible because the unit being required to be placed in MODE 3 within 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> of the PORV 340A failure is correct.

C. Incorrect, Plausible because neither the RED nor the GREEN indicating light being lit is due to the PORV being stuck in the mid position and there are conditions (both PRO Vs inoperable but capable of being cycled) with both PORVs isolated that will allow the unit to continue to operate in Mode I for an unlimited time period.

D. Correct, with PORV 340A stuck in the mid position neither the RED nor the GREEN indicating light will be lit and the status of PORV 340A requires the plant be placed in MODE 3 within 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> of the failure.

Question Number: 80 Tier: 1 Group 1 K/A: 056 AA2.74 Loss of Off-Site Power Ability to determine and interpret the following as they apply to the Loss of Offsite Power:

PORV position Importance Rating: 3.6 / 3.7 10 CFR Part 55: 43.5 /45.13 IOCFR55.43.b: 2 K/A Match: K/A is matched because the question requires the ability to determine the status of PORV indications during a loss of offsite power and is SRO because the questions requires knowledge of Tech Spec information below the line.

Technical

Reference:

Tech Spec LCO 3.4.11, Pressurizer PORVs, Amendment 55 1-45W600-68-1 R12 Proposed references None Page 13

WBN 10-2011 NRC SRO Exam As Submitted 8/15/2011 to be provided:

Learning Objective: 3-OT-T/S0304

4. Given plant conditions and parameters correctly determine the applicable Limiting Conditions for Operations or Technical Requirements for the various components of the RCS.

3-OT-SYSO68C

11. Describe the indication an operator has that a PORV is open or leaking through.

Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question for the WBN 10/2011 NRC exam.

Comments:

Page 14

Pressurizer PORVs 3.4.11 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.11 Pressurizer Power Operated Relief Valves (PORVs)

LCO 3.4.11 Each PORV and associated block valve shall be OPERABLE.

APPLICABILITY: MODES 1,2, and 3.

ACTIONS Separate Condition entry is allowed for each PORV.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more PORVs A.1 Close and maintain power to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> inoperable and capable of associated block valve.

being manually cycled.

B. One PORV inoperable B.1 Close associated block valve. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and not capable of being manually cycled.

B.2 Remove power from associated 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> block valve.

AND B.3 Restore PORV to OPERABLE 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> status.

(continued) 12vtt1J/i 6k14 C4cQ4,I fJl Watts Bar-Unit 1 3.4-22

pressurizer POVs 3.4.11 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. One block valve Ci Place associated PORV in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> inoperable, manual control.

AND C.2 Restore block valve to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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 of Condition A, B, AND or 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 /> E. Two PORVs inoperable E.1 Close associated block valves. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and not capable of being manually cycled.

E.2 Remove power from associated 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> block valves.

AND E.3 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND E.4 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. Two block valves F.1 Place associated PORVs in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> inoperable, manual control.

AND (continued)

Watts Bar-Unit 1 3.4-23

Pressurizer PORVs 3.4.11 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME F. (continued) F.2 Restore one block valve to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> OPERABLE status.

G. Required Action and G.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 F not AND met.

0.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SURVEILLANCE REQUIREMENTS

. SURVEILLANCE FREQUENCY SR 3.4.11.1 NOTE Not required to be met with block valve closed in accordance with the Required Action of Condition B or E.

Perform a complete cycle of each block valve. 92 days SR 3.4.11.2 Perform a complete cycle of each PORV. 18 months Watts Bar-Unit 1 3.4-24

3 -OT-T/S03 04 Revision 4 Page 3 of 14 I. PROGRAM WATTS BAR OPERATOR TRMNING II. COURSE A. License Training B. Licensed Requalification III. TITLE T/S 3.4, Reactor Coolant System, Bases, and Technical Requirements Manual IV. LENGTH OF LESSON A. License Training 1 Hour Licensed Requalification time will be determined after objectives are identified.

V. TRAINING OBJECTIVES AR S S U OR T 0 0 A

)0. Demonstrate an understanding of NUREG 1122 knowledges and abilities associated with the Reactor Vessel that are rated >2.5 during Initial License Training and >3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A.

x x 1. Demonstrate the ability to extract specific information from the Technical Specifications and Technical Requirements, as they pertain to RCS.

. Determine the bases for each specification, as applicable, to the RCS.

x 3. Given plant conditions/parameters correctly determine the OPERABILITY of components associated with RCS.

. Given plant conditions and parameters correctly determine the applicable Limiting Conditions for Operations or Technical Requirements for the various components of the RCS.

3-OT-SYSO68C Revision 13 Page5of45 I. PROGRAM Watts Bar Operator Training II. COURSES A. License Training B.. Non-License Training III. TITLE PZR, PZR Pressure Control System! PZR Level Control System, and PRT IV. LENGTH OF LESSON A. License Training 4 Hours B. Non-License 6 Hours V. TRAINING OBJECTIVES 0 0 D 0 C,) (I)

X X X X 1. Identify the three (3) main purposes of the Pressurizer.

X X X X 2. Describe the major components of the Pressurizer.

X X X X 3. Describe the purposes of the Manual Bypass Pressurizer Spray Throttle Valves.

X X X X 4. Identify the normal setpoint required to auto open the PZR Relief Valves (PORVs).

X X X X 5. Identify each setpoint and resulting automatic action for the Pressurizer Pressure Program.

X X X 6. State the basis for the low pressure reactor trip, as stated in Tech Specs Section 2.1.1.

X X X 7. State the basis for the high pressure reactor trip, as stated in Tech Specs Section 2.1.1.

X X X 8. Describe the operation of the master pressure controller.

X X X 9. Describe what control room indication would alert the operator that the pressurizer spray valves were open.

X X X 10. Describe the method of control for the power operated relief valves.

X X X 1 1. Describe the indication an operator has that a PORV is open or leaking through.

3-OT-SYSO68C Revision 13 age6of45 0 0<

DO C I

< Q: Cl) Cl)

X X X X 12. Identify the program setpoints, and describe any automatic actions relative to the pressurizer level program.

X X X X 13. Describe the basis for the program setpoints of the pressurizer level program circuit.

X X X X 14. Explain the basis for programming the level vs. maintaining the level constant in the pressurizer.

X X X X 15. Describe the response to a deviation from pressurizer level program.

X X X X 16. Explain the purpose of the PRT.

X X X X 17. Identify the components which drain into the Pressurizer Relief Tank.

X X 18. Deleted.

X X 19. Deleted X X X X 20. Describe the in-plant location of major system components, instrumentation, controls, and piping/header arrangements.

X X X X 21. Describe the flow path of sources of supply, discharges, vents, drains, leakoff, and connections/penetrations that intertie this system to other systems.

X X X X 22. Explain the operation of major system components.

X X X X 23. Deleted X X X 24. Deleted

Clarification Guithncefor SRG-onl-y Questions Rev 1(0311112010)

Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)

(Tech Specs)

Can question be answered solely by knowing 1 Yes hour TSITRM Action? RD question No Can question be answered solely by knowing the Yes LCD/TRM information listed above-the-line? RD question No Can question be answered solely by knowing the Yes TS Safety Limits? RD question No Does the question involve one or more of the following for TS, TRM, or DDCM?

• Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)

• Application of generic LCD requirements (LCD 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4) Yes SRD-only Knowledge of TS bases that is required to analyze TS question required actions and terminology No Question might not be linked to 10 CFR 55.43(b)(2) for S RD-only Page 5 of 16

WBN 10-2011 NRC SRO Exam As Submitted 811512011

81. W/E04 EA2.1 081 Given the following:

- During performance of ECA-1 .2, LOCA Outside Containment, the crew determines RCS pressure is rising after RHR Train B cold leg injection valve 1-FCV-63-94 is closed.

- The crew then stops and locks out RHR pump 1 B-B and coses its suction valve.

Which ONE of the following identifies the required procedure transition?

A. ES-I .1, SI Termination B E-1, Loss of Reactor or Secondary Coolant C. ECA-1 .1, Loss of Emergency Coolant Recirculation D. ES-i .2, Post LOCA Cooldown and Depressurization DISTRA CTOR ANAL YSIS:

A. lncorrect, Plausible because ES-I. I is a sub-procedure in the LOCA series of emergency procedures and would be a transition that could be required subsequent to the E-I transition depending on the RCS pressure trend.

B. Correct, the RCS pressure rising indicates that the leak has been terminated and with the RCS pressure rising the transition to E-I is directed by the step in ECA-I.I.

C. lncorrect Plausible because if the RCS pressure had been dropping after the valve closure, then the transition would be to ECA-I. I D. Incorrect, Plausible because ES-I.2 is a sub-procedure in the LOCA series of emergency procedures and would be a transition that could be required subsequent to the E-I transition depending on the RCS pressure trend.

Question Number: 81 Tier: 1 Group 1 K/A: WIEO4 EA2.1 LOCA Outside Containment Ability to determine and interpret the following as they apply to Page 15

WBN 10-2011 NRC SRO Exam As Submitted 811512011 K/A: W/E04 EA2.1 LOCA Outside Containment Ability to determine and interpret the following as they apply to the (LOCA Outside Containment)

Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

Importance Rating: 3.4 / 4.3 10 CFR Part 55: 43.5 /45.13 IOCFR55.43.b: 5 K/A Match: K/A is matched because the question requires the ability to assess plant conditions to determine the proper procedure transition during a LOCA outside containment event. The question is SRO because it requires Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

Technical

Reference:

ECA-1 .2, LOCA Outside Containment, Revision 0005 WOG ECA-1 .2 Background, Revision 2 E-1, Loss of Reactor or Secondary coolant, Revision 0016 Proposed references None to be provided:

Learning Objective: 3-Oi-ECAO1O1

08. Given a set of plant conditions, use procedures ECA-1 .1 and 1.2 to identify any required procedure transition.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: VOGTLE 2010 bank question WEO4EA2.1 01 used on the VOGTLE 2010 exam with wording changes in stem and to allow use at WBN. Stem conditions modified but no choices changed.

Comments:

Page 16

WBN LOCA Outside Containment ECA-1.2 Unit I Rev. 0005 Step Action/Expected Response Response Not Obtained 3.0 OPERATOR ACTIONS

1. ENSURE RHR suction from RCS CLOSED:

• 1-FCV-74-1 and 1-FCV-74-9.

AND

• 1-FCV-74-2 and 1-FCV-74-8.

2. ENSURE SI pumps hot leg injection 1 -FCV-63-1 56 and 1 -FCV-63-1 57 CLOSED.

3. ENSURE RCS letdown isolated:

• Letdown isolation 1-FCV-62-69 and 1-FCV-62-70 CLOSED.

• Excess letdown isolation 1-FCV-62-54 and 1-FCV-62-55 CLOSED.

4. ENSURE RHR hot leg injection 1-FCV-63-172 CLOSED.

5. CHECK RCS press **

GOTOStepl4.

DROPPING or stable.

6. CLOSE RHR crosstie valve 1-FCV-74-33 or 1-FCV-74-35.

7. CLOSE RHR Train A cold leg IF 1-FCV-63-93 failed OPEN, injection valve 1-FCV-63-93. THEN GOTOSteplO.

Page3of5

WBN LOCA Outside Containment - ECA-12 Unit I Rev. 0005 Step Action/Expected Response Response Not Obtained

8. CHECK LOCA isolated: OPEN 1-FCV-63-93.

• RCS press rising. GO TO Step 10.

9. ISOLATE RHR Train A:

a. STOP RHR pump A-A, AND PLACE in PULL TO LOCK.

b. CLOSE RHR suction valve 1 -FCV-74-3.

c. GOTO Step 15.

10. CLOSE RHR Train B cold leg injection IF 1-FCV-63-94 failed OPEN, valve 1 -FCV-63-94. THEN GO TO Step 13.

11. CHECK LOCA isolated: OPEN 1-FCV-63-94.

• RCS press rising. **

GOTO Step 13.

12. ISOLATE RHR Train B:

a. STOP RHR pump B-B, AND PLACE in PULL TO LOCK.

b. CLOSE RHR suction valve 1 -FCV-74-21.

c. GO TO Step 15.

Page4of5

- WBN LOCA Outside Containment ECA1 .2 Unit I Rev. 0005 Step Action/Expected Response Response Not Obtained

13. ENSURE RHR crosstie valves 1-FCV-74-33 and 1-FCV-74-35 OPEN.

14. IDENTIFY break location: NOTIFY TSC of failure to identify break location.

• Radiation Protection surveys.

• RHR pipe break lights [M-6].

• ECCSpumpflows.

o Aux bldg flood alarms [M-15; light panel, Aux Bldg 757].

• Radiation area monitor recorders 1-RR-90-1 and O-RR-90-12A.

15. DETERMINE appropriate Instruction: NOTIFY TSC of failure to isolate break.

• IF LOCA outside cntmt isolated, GO TO ECA-1.1 Loss of RHR THEN Sump Recirculation.

GO TO E-1, Loss of Reactor or Secondary Coolant.

End of Section Page 5 of 5

WBEN Loss of Rator orSecondry COO Int E-1 Uniti Rev.0016 Step Action/Expected Response Response Not Obtained

11. CHECK SI termination criteria:

a. CHECK RCS subcooling a. **

GO TO Caution greater than 65°F [85°F ADV}. prior to Step 12.

b. CHECK secondary heat sink b. ENSURE no higher available with either: priority exists,

• Total feed flow to Intact S/Gs THEN greater than 410 gpm, **

GO TO FR-H.1, Loss of OR Secondary Heat Sink.

• At least one Intact S/G NR level greater than 29% [39% ADV].

c. CHECK RCS pressure c. **

GO TO Caution stable or rising. prior to Step 12.

d. CHECK pzr level greater d. RESTORE pzr level:

than 15% [33% ADVJ.

1) ATTEMPT to stabilize RCS pressure with normal pzr sprays.

2) **

GO TO Caution prior to Step 12.

e. GO TO ES-1.1, (4

SI Termination.

Page 8 of 24

• WBN LOS Of Ráctor or Secondary Coolant -1 Uniti Rev. 0016 Step Action/Expected Response Response Not Obtained

23. DETERMINE if RCS cooldown and depressurization is required:

a. CHECK RCS pressure a. IF RHR pump injecting to RCS, greater than 150 psig. THEN GO TO Step 24.

GO TO ES-I .2, Post LOCA Cooldown and Depressurization.

24. PREPARE for switchover to RHR cntmt sump:

a. ENSURE power restored to 1-FCV-63-1 USING Appendix B (E-1), 1-FCV-63-1 Breaker Operation.

b. CHECK RWST level b. **

GO TO Step 19.

less than 34%.

c. GO TO ES-I .3, Transfer to Containment Sump.

Page 13of24

3. RECOVERY/RESTORATION TECHNIQUE The objective of the recovery/restoration technique incorporated into guideline ECA1.2 is to provide actions to identify and isolate a LOCA outside containment.

The following subsection provides a summary of the major categories of operator actions for guideline ECA-1.2, LOCA OUTSIDE CONTAINMENT.

3.1 High Level Action Summary A high level summary of the actions performed in ECA-1.2 is given on the following page in the form of major action categories. These are discussed below in more detail.

o Verify Proper Valve Alignment The first instruction given to the operator is to verify that all normally closed valves in lines that penetrate containment are closed. If a normally closed valve is open, this action may isolate the break.

o Identify and Isolate Break The operator then attempts to identify and isolate the break by sequentially closing all normally open valves in paths that penetrate containment.

o Check If Break Is Isolated RCS pressure is monitored to determine if the break has been isolated. A significant increase in RCS pressure indicates the break is isolated and the operator is sent to guideline E-1, LOSS OF REACTOR OR SECONDARY COOLANT. If the break is not isolated, the operator transfers to guideline ECA-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION, for further recovery actions.

ECA-1.2 Background 7 HP-Rev. 2, 4/30/2005 HECA12BG.doc

STEP DESCRIPTION TABLE FOR ECA-1.2 Step j..

STEP: Check If Break Is Isolated PURPOSE: To determine if the LOCA outside containment has been isolated from previous actions BASIS:

This step instructs the operator to check RCS pressure to determine if the break has been isolated by previous actions. If the break is isolated in Step 2, a significant RCS pressure increase will occur due to the SI flow filling up the RCS with break flow stopped.

The operator transfers to E-1, LOSS OF REACTOR OR SECONDARY COOLANT, if the break has been isolated, for further recovery actions. If the break has not been isolated, the operator is sent to ECA-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION, for further recovery actions since there will be no inventory in the sump.

ACTIONS:

o Determine if RCS pressure is increasing o Transfer to ECA1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION, Step 1 o Transfer to E-1, LOSS OF REACTOR OR SECONDARY COOLANT, Step 1 INSTRUMENTATION:

RCS pressure indication CONTROL/EQUIPMENT:

N/A KNOWLEDGE:

It should be noted that for some breaks SI flow may cause an RCS pressure increase independent of break isolation. It should also be noted that for larger breaks, RCS repressurization may be delayed following break isolation.

Additionally, if the RCS is saturated or a cooldown is in progress, RCS repressurization will proceed more slowly. Other means of verifying break isolation should be checked. For example, increasing RVLIS trend due to injection flow, decreasing trends in local abnormal conditions and local observation (if practical) may be useful.

ECA-1.2 Background 14 HP-Rev. 2, 4/30/2005 HECA12BG.doc

HL45RSRONRCEXAM 99.

The crew is implementing EOP 19112-C, ECA-l .2 LOCA Outside Containment.

- The COLD LEG INJECTION FROM SIS HV-8835 has been closed.

- The SS determines the leak is now isolated.

The SS will transition to...

A. 19011-C, ES-l.l SI Termination.

B. 19010-C, E-l Loss of Reactor or Secondary Coolant.

C. 19111-C, ECA-l.l Loss of Emergency Coolant Recirculation.

D. 19012-C, ES-I .2 Post-LOCA Cooldown and Depressurization.

99

hil OPERATIONS 3-OT-ECAO1O1

! EMERGENCY CONTINGENCY ACTIONS, ECA-1 1 & 1.2 Rev. 9 INSTRUCTOR GUIDE Page 3 of 42 I. PROGRAM:

Watts Bar Operator Training II. COURSE:

A. License Training B. License Requalification III. TITLE:

Emergency Contingency Actions, ECA- 1.1 and 1.2 1V. LENGTH OF LESSON:

License Certification 2 Hours License operator REQUAL time will be determined after objectives are identified.

Non-License operator REQUAL time will be determined after objectives are identified.

V. TRAINU%G OBJECTIVES:

AR S S U OR T 0 0 A X X X )O. Deleted.

X X X )l. Identify and explain the major actions of procedures ECA-l.l and 1.2.

X X X )2. Given the time of reactor trip, be able to use ECA-l .1 figure 1 to identify the minimum required SI flow.

X X X )3. Explain the purpose of establishing minimum SI flow as determined by figure 1.

X X X )4. Identify problems which might result from dropping RCS press to less than 250 psig prior to CLA isolation.

OPERATIONS 3-OT-ECAO1O1 I EMERGENCY CONTINGENCY ACTIONS ECA-t1 & 12 Rev. 9 INSTRUCTOR GUIDE Page 4 of 42 V. TRAINING OBJECTWES: (continued)

AR S S U OR T 0 0 A X X X )5. Explain the action taken and the basis for that action if RWST level decreases to 8%.

X X X )6. Identify and explain the limitation on charging flow after the RWST is empty and the CCP suction is aligned to the VCT.

X X X )7. Identify the limitations for continued RCP operation at low RCS pressures.

X X X )8. Given a set of plant conditions, use procedures ECA-l.l and 1.2 to identify any required procedure transition.

X X X )9. Discuss the reasons that ECA- 1.1, Loss of RHR Sump Recirculation, is given priority over FR-Z. 1, High Containment Pressure for directing Containment Spray Operation.

X X X

10. Determine appropriate operator actions/system response for the Containment Spray System with SI actuated under each of the following conditions:

• RWST LEVEL LO RECIRC INTLK Alarm e RWST LEVEL LO-LO Alarm

• Containment Pressure> 13.5 psig

• Containment Pressure between 2.0 & 13.5 psig

. Containment Pressure <2.0 psig

hiI OPERATIONS 3-OT-ECAO1O1 TA EMERGENCY CONTINGENCY ACTIONS ECA-1.1 & 1.2 Rev 9 INSTRUCTOR GUIDE Page 5 of 42 V.TRAINING OBJECTWES: (continued)

AR S S U OR T 0 0 A X X X

11. Given a set of plant conditions, use ECA-1.1 and ECA-1.2 to correctly diagnose and implement: Action Steps, RNOs, Notes and Cautions.

VI. TRAINING AIDS:

A. White Marker Board.

B. Classroom PC C. Projector.

VII. MATERIALS:

A. Appendix A - Instructor Guide for Fundamental Overview of ECA 1-1.

B. Attachments, Handouts: One copy of each of the following for each participant:

1. Attachment 1- WBN Emergency Contingency Actions, ECA- 1.1

2. Attachment 2- WBN Emergency Contingency Actions, ECA-l .2

3. Attachment 3- Background Information for ECA-1 .1

4. Attachment 4- Background Information for ECA- 1.2

5. Attachment 5- ECA 1.1 and ECA 1.2 Power Point presentation

6. Attachment 6- Operating Experience 0E23 154 Watts Bar Unplanned Loss of Reactor Coolant

2 OPERATIONS 3-OT-ECAO1O1 EMERGENCY CONTINGENCY ACTIONS, ECA.rl J & 1 2 Raw INSTRUCTOR GUIDE Page 28 of 42 X. LESSON BODY INSTRUCTOR NOTES

Purpose:

To check if an excessive containment hydrogen concentration is present.

Basis: This step instructs the operator to obtain a current hydrogen concentration measurement. Depending upon the magnitude of the hydrogen concentration, the operator will either continue with ECA 1.1, turn on the hydrogen recombiners or notify the TSC to determine additional recovery actions before continuing with the instruction.

47) CONSULT TSC for long term plant operation.

Purpose:

To consult with the plant engineering staff for further actions.

Basis: This ECA provides generic steps for cooldown and depressurization of the plant to atmospheric conditions following a loss of emergency containment recirculation capabilities.

Subsequent actions are plant specific and plant operators, TSC personnel and plant management need to make decisions about long term plant operation and any repairs necessary for plant restart.

The presence of acidic water from the LOCA may lead to chloride induced stress corrosion of the recirculation loop piping.

B. Discussion of ECA-l .2,.LOCA Outside Containment Use latest revision of Emergency Instructions.

1. Purpose This Instruction provides actions to identify and Procedure Use and isolate a LOCA outside containment. Adherence:

Reinforce procedure usage and

2. Symptoms and Entry Conditions placekeeping standards during

a. Symptoms presentation of LP.

OPERATIONS 3-OT-ECAO1O1 I I Tf EMERGENCY CONTINGENCY ACTIONS, ECA-li, & 1.2 Rev 9 INSTRUCTOR GUIDE Page 29 of 42 X. LESSON BODY INSTRUCTOR NOTES Abnormal Auxiliary Building radiation due to LOCA outside containment.

b. Transitions

1) E-O, Reactor Trip or Safety Injection.

2) E-l, Loss Of Reactor Or Secondary Coolant.

3. Major Action Categories Objective 1

a. Verify proper valve alignment.

b. Identify and isolate break.

c. Check if break is isolated.

4. Steps, Purposes, and Bases NOTE: Outline is changed to correspond to procedure step

1) ENSURE RHR suction from RCS numbers.

CLOSED.

Purpose:

To ensure that normally closed valves are closed.

Basis: This step instructs the operator to Operator Fundamentals:

verify that all normally closed valves in low Understanding plant design.

pressure lines and other plant specific lines that penetrate containment are closed. The valving connecting the RHR system to the RCS is of particular interest since the RHR system is a low pressure system connected to the high pressure RCS. Therefore a rupture or break outside containment is most probable to occur in the low pressure RHR piping.

2) ENSURE SI pumps hot leg injection 1-FCV-63-156 and 1-FCV-63-157 CLOSED.

Purpose:

To ensure that normally closed valves are closed.

Basis: Same as Step 1.

3) ENSURE RCS letdown isolated.

OPERATIONS 3-OT-ECAO1O1

. EMERGENCY CONTINL ENCY ACTIONS, ECA-1 1, & 1.2 Rev.

INSTRUCTOR GUIDE Page 30 of 42 X. LESSON BODY TNSTRUCTOR NOTES

Purpose:

To ensure that normally closed valves are closed.

Basis: Same as Step 1.

4) ENSURE RHR hot leg injection 1-FCV-63-172 CLOSED.

Purpose:

To ensure that normally closed valves are closed.

Basis: Same as Step 1.

5) CHECK RCS press DROPPING or stable.

Purpose:

To determine if actions performed to this point isolated the break.

Basis: Plant specific step added to address the possibility that prior actions may have isolated the break.

6) CLOSE RHR crosstie valve 1-FCV-74-33 or 1 -FCV-74-35.

Purpose:

To attempt to identify and isolate a LOCA outside containment.

Basis: This step instructs the operator to Procedure use and sequentially close and open all normally adherence:

opened valves in paths that penetrate Reinforce procedure usage containment to identify and isolate the standard.

break outside containment.

Again as in the previous steps, the valving connecting the RHR system to the RCS is of primary interest since the probability of a break occurring outside containment is most probable to occur in the low pressure RHR system piping.

7) CLOSE RHR Train A cold leg injection valve 1-FCV-63-93.

Purpose:

To attempt to identify and isolate a LOCA outside containment.

Basis: Same as Step 6.

8) CHECK LOCA isolated:

bi1 OPERATIONS 3-OT-ECAO1O1 EMERGENCY CQNTIfLGENCY ACTIONS, ECA1ir& L2 Rev. 9 INSTRUCTOR GUIDE Page 31 of 42 X. LESSON BODY INSTRUCTOR NOTES

Purpose:

To determine if the LOCA outside containment has been isolated by previous actions.

Basis: This step instructs the operator to check RCS pressure to determine if the break has been isolated by a previous action. If the break is isolated, a significant RCS pressure rise will be observed due to the SI flow filling up the RCS with break flow stopped.

9) ISOLATE RHR Train A:

Purpose:

To attempt to identify and isolate a LOCA outside containment.

Basis: Same as Step 6.

10) CLOSE RHR Train B cold leg injection valve 1 -FCV-63-94.

Purpose:

To attempt to identify and isolate a LOCA outside containment.

Basis: Same as Step 6.

11) CHECK LOCA isolated:

Purpose:

To determine if the LOCA outside containment has been isolated by previous actions.

Basis: Same as Step 8

12) ISOLATE RHR Train B:

Purpose:

To attempt to identify and isolate a LOCA outside containment.

Basis: Same as Step 6.

13) ENSURE RHR crosstie valves 1-FCV 33 and 1-FCV-74-35 OPEN.

Purpose:

To properly realign RHR if not the source of leakage.

I OPERATIONS 3-CT- ECAO1 01 EMERGENCY CONTINGENCY.ACTIQNS, ECA-ti, & 12 Rev.9 INSTRUCTOR GUIDE Page 32 of 42 X. LESSON BODY INSTRUCTOR NOTES Basis: Plant specific step added to reopen the cr0sstie valves if leak path is not found in RHR system.

14) IDENTIFY break location:

Purpose:

To attempt to identify and isolate a LOCA outside containment.

Basis: Same as Step 6. Plant specific guidance on indications/actions for leak identification have been added to this step.

15) DETERMINE appropriate instruction:

Purpose:

To direct the operator to the proper instruction after leakage has either been isolated or cannot be isolated.

Basis: If the break can be identified and Objective 8 isolated, then the operator is directed to E- 1, for further recovery actions. If the break cannot be identified and isolated, then the operator is directed to ECA- 1.1, where actions are taken to minimize break flow and initiate makeup to the RWST.

C. Operating Experience On 7/7/06 at Watts Bar, an unplanned loss of Volume Control Tank inventory occurred during preparations for the transfer of resins from the mixed bed demineralizer to the spent resin storage tank.

One of two possible valves was leaking through to the in-service mixed bed vessel, and upon venting the cation bed to the Tritiated Drain Collector Tank, the loss of inventory occurred.

C[arification Guithnce for SRO..anly Questions Rev 1(03/1112010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, Yes flowpath, logic, component location? RD question I Canthequestionbeans Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?

No Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

No Does the question require one or more of the following?

• Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps SRO-only

• Knowledge of diagnostic steps and decision points in the ll*estion EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures

• Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No I Question might not be linked to 1

10 CFR 55.43(b)(5) for SRO-only Page 8 of 16