Text

Initial Exam 2008-302 Draft RO Written Exam
	ML091810062
Person / Time
Site:	Browns Ferry
Issue date:	06/30/2009
From:	; NRC/RGN-II
To:	; Tennessee Valley Authority
References
	50-259/08-302, 50-260/08-302, 50-296/08-302
	Download: ML091810062 (445)
	v • d • e

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet Examination Outline Cross-reference: Level SRO 203000G2.4.20 Tier # 2 Knowledge of operational implications of EOP warnings, cautions 1 Group #

and notes. RHR/LPCI: Injection Mode KIA # 203000G2.4.20 Importance Rating 3.8 4.3 I Proposed Question: # 1 A loss of all off-site power and main steam line break have occurred on Unit 3 resulting in the following plant conditions:

RPV pressure 50 psig and steady.

RPV level (-) 100 inches and steady, being maintained by RHR Loop I at 22,000 gpm.

Drywell pressure 19 psig and rising.

Drywell temperature 225 of and rising.

Suppression Pool Temp 205 of and rising.

Both loops of Core Spray are unavailable.

Which ONE of the following describes the appropriate action to take and the basis for that action?

Initiating Drywell Sprays with RHR Loop II is (1) because _ _4-(2=..)t....-__

(1) (2)

A. ina ppropriate RHR Loop II must be aligned for LPCI injection to ensure adequate core cooling.

B. inappropriate RHR Loop II must be aligned for LPCI injection UNTIL RPV level is restored to between (+) 2 and (+) 51 inches.

c. appropriate RHR Loop II is not required for LPCI injection to ensure adequate core cooling.

D. appropriate this will ensure adequate NPSH for RHR Loop I under the current plant conditions.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: C Explanation : a. Part (1) is incorrect. Initiating OW Spray IS appropriate. It can be interpolated that a OW pressure of 19 psig corresponds to a Suppression Chamber pressure above 12 psig, ("'5 psid) which requires OW Sprays. In addition, due to the relatively small L1T between SP temp and OW temp, sprays will not reduce containment pressure low enough to exceed Curve

2 NPSH limits. This choice becomes plausible due to the wording of Caution #4, which warns of the potential for exceeding NPSH limits with a reduction in containment pressure. Only a detailed analysis of the containment conditions, based on fundamental heat transfer dynamics, can eliminate that potential. Part (2) is incorrect. As long as RPV level can be maintained above (-) 162 inches, RHR Loop II can be aligned for containment control.

b. Part (1) is incorrect. Part (2) is incorrect. As long as RPV level can be maintained above (-) 162 inches, RHR Loop II can be aligned for containment control. Although the required level band is +2 to +51 inches, as long as adequate core cooling is assured based on current conditions, using RHR for containment control is allowed.

c. Correct answer.

d. Part (1) is correct as stated in (a) above. Part (2) is incorrect. Initiating Orywell Sprays under this condition will actually reduce NPSH to the running RHR pumps, however as stated in (a) above, the small L1T between SP temp and OW temperature would prevent containment pressure from dropping low enough to challenge NPSH limits.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 3-EOI-1 and 3-EOI-2 flowcharts (Attach if not previously provided)

Proposed references to be provided to applicants during examination: 3-EOI-2 Flowchart Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New RMS 6/16/2008 Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: This has been evaluated as CIA due to the multiple plant conditions which must be evaluated against procedural requirements. The KIA is met by structuring the question to ensure RHR LPCI injection Mode is considered and maintained while performing other actions in accordance with EOls.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from 3-EOI-2 flowchart path DW/T:

2 PUMP NPSH AND VORTEX LIMITS L

INITIATE OW SPRAYS USING ONLY PUMPS NOT REQUERED TO ASSURE ADEQUATE CORE COOLING BY CONTINUOUS INJ (APPX 178)

L DWfT-10

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet CURVE 2 RHR NPSH LIMITS 245 235 I

15-r P.slG SAFE"II.

I

- ~

225 215 10 PSIG SAFE.

1 1 ILL 0

Q..

! 205

~

5PSIG SAFE. .............

~

t:= 195

~

..J Q.. o PSIGSAFE*

a:: 185

"'-1"-

Q.. ~

Q..

l 175 en ~

165 155 "-

145 A

500 2500 4500 6500 8500 10500 12000 RHR PUMP FLOW (GPM) *SUP?R CHMBR PRESS OPERATION OF RHR OR CS WITH SUCTION FROM THE SUPPR PL MAY RESULT IN EaU IPMENT DA~GE IF:

PUMP FLOW IS ABOVE THE NPSH LIMIT (CURVE 1 OR 2)

OR

SUPPR PL LVL IS BELOW THE VORTEX LIMIT (10 FT).

REDUCING PC PRESS WILL REDUCE THE AVAJLABLE NPSH FOR PUMPS TAKING SUCTION FROM THE SUPPR PL

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

\.

203000G2.4.34 Tier # 2 Knowledge of RO tasks performed outside the main control room 1 Group #

during an emergency and the resultant operational effects:

RHR/LPCI Injection Mode. KIA # 203000G2.4.34 Importance Rating 4.2 4.1 I Proposed Question: RO # 2 Given the following Unit 1 plant conditions:

Unit 1 and 2 control rooms have been abandoned due to a toxic gas release.

Control has been established at Backup Control Panel 1-25-32.

RHR Loop II is in Suppression Pool Cooling with both RHR pumps running.

Reactor Pressure is 110 psig and lowering due to the cooldown.

Reactor water level is (-) 48 inches and lowering slowly.

The Unit Supervisor has directed that RHR be lined up for LPCI injection in accordance with 1-AOI-l00-2, "Control Room Abandonment."

Which ONE of the following describes the location where this lineup is performed and the method of monitoring injection flow?

Operating the LPCI Injection valves can be accomplished from _ _ _~(=.1),---___ and injection flow is monitored by (2)

(1) (2)

A. 480V RMOV Board lB RHR Total Flow indication from Panel 1-25-32 B. 480V RMOV Board lB RHR pump amps from 4KV Shutdown Board C

c. Backup Control Panel 1-25-32 RHR Total Flow indication from Panel 1-25-32 D. Backup Control Panel 1-25-32 RHR pump amps from 4KV Shutdown Board C

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: A Explanation: a. Correct answer.

b. Part (1) is correct. Part (2) is incorrect. Pump amperage is used to verify RHRSW pump flow during operation from 1-25-32, not RHR pump flow.

c. Part (1) is incorrect. RHR injection valves cannot be controlled from 1 32. Only RCIC injection flow can be controlled from 1-25-32. Part (2) is correct.

d. Part (1) is incorrect as stated in (c) above. Part (2) is incorrect. Pump amperage is used to verify RHRSW pump flow during operation from 1 32, not RHR pump flow.

ES-401 Sample Written Examination Forni ES-401-5 Question Worksheet Technical Reference(s): 1-AOI-100-2 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/07/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Control Room Ahandonment 1-AOI-1 00-2 Unit 1 Rev. 0016 Page 14 of 19 4.2 Unit 1 Subsequent Actions (continued)

U-2 Unit Operato,r complete Attachment 7 o

[13] Upon completion of attachments, RE-ESTABUSH communication using the best avallable means and AWAIT further instructions. 0

[14] tF CRO Pump 18 is to be aligned to, Unit 1, THEN PERFORM the following:

[14.1] OPEN 1-FCV-085-0008 using CRO PUMP 1B UNIT 1 DlSCHARGE,1-I-lS-085~OO08C at 480V RMOV Bd 1C Compt3R 0

[14.2] PLACE GRO PUMP 18, 1-I-lS*OB5-0002C in CLOSE at 4KV SD Bd A, Compt 13 to start CRID Pump 1R 0

[15] fNIHATE RHR Suppression Pool Cooling as toflows.:

NOTE Communication between 4160V Shutdown Boam C and 480V RMOV Bd 1B is necessary for establishing RHRSW flow and to prevent exceeding 53 amps on RHRSW Pump B2.

[15.1] PLACE RHRSW PUMP 82 MOTOR, O-HS-023-0019C in CLOSE at 4160V Shutdown Bd C, Compt 16 to start RHRSW Pump 82. 0

[15.2] THROTTLE OPEN RHR HX 1B RHRSW OUTLET VLV, using 1-HS-023-0045C at 480V RMOV Bd 1B, Compt 14C2. 0

[15.3] WHEN between 48 and 52 amps on RHR SERVICE WATER PUMP B2, THEN STOP throtfling, RHR HX 1B RHRSW OUTLET \ltV, 1-HS-023-0046C. 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Control Room Abandonment 1-AOI-100-2 Unit 1 Rev. 0016 Page 17 o.f79 4,.2 Unit 1 Subsequent Actions (continued}

[16] MAIHTAIN Drywell temperature less than 160°F as follows:

[16.1] MONfTOR DRYWELL A!R TEMPERATURE, 1-TIS-64-52AA at Panel 1-25-32, 0

[16.2] OPER/HE Drywel! BIow-efS as required. 0 Orywel! Compt. SWitch Blower SWitch No. No.. PosiOOn 480V Shutdown Sd 1.1\

1.4.-1 OW CLG UNIT 1A1 BLOWER, 1-HS-070-0037C 2C CLOSE 0 l.A.-2 OW CLG UNIT 1A2 BLOWER, l-HS-070-0038C 2D CLOSE 0 480V Shutdown Sd 18 1B-1 OW CLG UNIT 1B1 BLOWER, 1-HS-070-0042C 2C CLOSE 0 1B-2 OW CLG UNiT 1B2 BLOWER, 1-HS-070-0043C 2D CLOSE 0

,/

480V RMOV Bd 1A

\

1.0\-3 OW CLG UNIT 1A3 BLOWER, 1-HS-070-0039C 17A START 0 1A-4 OW CLG UNIT 1M BLOWER, 1-HS-070-0040C 18A START 0 18-3 COOLERS started by AUO in Attachment 4 Part B.

18-4 480V RMOV Bd 1C lAS OW CLG UNIT 1A5 BLOWER, 1-HS-Q70-0041C 1A START 0 1B-5 OW ClG UNIT 1B5 BLOWER. 1-HS-070-0046C 11A START 0

[17] IF Reactor makeup from RHR LPC! is desired AND RHR Pumps are operating in Suppression Pool Cooling, THEN ESTABlISHRHR system How to the Reactor as follows:(otherwise N/A) o

[17.1] MONJTOR RHR SYS II TOTAL FLOW, 1-FI-74-79 at Pane! 1-25-32. o

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Control Room Abandonment 1-AOI-1 00-2 Unit 1 Rev. 0'016 Page 18 of 79 4.2 Unit 1 Subsequent Actions (continued}

[112] CLOSERHR SYS II SLJPPR POOL CLGJTEST VLV using 1-HS-074-0073C at 430V RMOV Bd 18, Compt R11C. 0 117.3] OPEN RHR SYS II LPCIINBD INJECT VLV, using 1-HS-074-0067C at 480V RMOV Bd 1B, CampI. R10A. 0

[17 A] THROTTLE OPEN RHR SYS II LPCI OUTBD INJECT VlV, using 1-HS-074-0066C at 480VRMOV Bd 1B, Compt.3A, as necessary to maintain Reactor Water Level between +.2 and +50 inches. 0 NOTE Step 4.2{18] will prepare the condensate system for injection in preparation for tripping RCIC when pressure is less than 50 psig.

[18] IF Reactor makeup from Condensate System is desired, THEN PERfORM the follOWing: (Otherwise NlA} o

[18.1] DISPATCH an operator to Unit 1 Turb Bldg Ei 617' at RFW START-UP LCV, 1-lCV-003-G053, and ESTABLISH communication'S. 0

[18.2] MONnOR RX WATER LEVEl A, 1-U-3-46A and RX WATER LEVEL B, 1-U-346B, at Panel 1-25-32, 0

[18.3] DIRECT operator to THROTTLE RFW START~LJP lCV-3-53 BYPASS, 1-BYV-003-0053, as necessary to mainta'in Reador Water Leve! between +2 and

+50 inches. 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 20S000A3.02 Tier # 2 Ability to monitor automatic operation of Shutdown Cooling 1 Group #

including: Pump trips.

KIA # 205000A3.02 Importance Rating 3.2 3.2 I Proposed Question: RO # 3 Given the following Unit 1 plant conditions:

RHR Loop I is in Shutdown Cooling Mode with 1A and 1C RHR Pumps running.

Unit 1 is in Mode 4 at 185 OF and lowering slowly.

RHR SYSTEM I MIN FLOW INHIBIT switch, 1-HS-74-148 is in INHIBIT.

RHR SYS I LPCI INBD INJECT VALVE, 1-FCV-74-53 is fully open.

RHR SYS I LPCI OUTBD INJECT VALVE, 1-FCV-74-52 is throttled open.

480V RMOV Board B de-energizes due to an electrical fault.

Which ONE of the following describes the status of RHR Loop I following the loss of 480V RMOV

( Board B and the final position of the RHR Inboard Injection valve 1-FCV-74-53?

RHRPumps1Aand1C ____~(~1~)------- The RHR SYS I LPCI INBD INJECT VALVE, 1-FCV-74-53 _ _ _--4-'(2=.+)----

(1) (2)

A. trip on loss of a suction path closes due to Group II logic signal B. trip on loss of a suction path fails open on loss of power

c. remain in operation closes due to Group II logic signal D. remain in operation fails open on loss of power

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: A Explanation: a. Correct answer.

b. Part (1) is correct. The candidate must determine that 480V RMOV Bd B supplies RPS B which, when lost, causes a Group II isolation of the RHR Shutdown Cooling suction valves. Part (2) is incorrect. RHR Loop II inboard injection valve is powered from 480V RMOV Bd B, so the Loop I valve (480V RMOV Bd A) still has power to automatically close.

c. Part (1) is incorrect. When RPS B is lost due to the 480V RMOV Bd Bloss, the Outboard Shutdown Cooling suction valve closes which causes a loss of suction path to BOTH RHR pumps. Part (2) is correct. The RHR Loop I inboard valve is powered from 480V RMOV Bd A so will have power to automatically close when the Group II logic signal is received.

d. Part (1) is incorrect. When RPS B is lost due to the 480V RMOV Bd Bloss, the Outboard Shutdown Cooling suction valve closes which causes a loss of suction path to BOTH RHR pumps. Part (2) is incorrect. RHR Loop II inboard injection valve is powered from 480V RMOV Bd B, so the Loop I valve (480V RMOV Bd A) still has power to automatically close.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 1-01-74, 1-01-99, Illustration 1 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/08/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Residual Heat Removal System 1-01-74 Unit 1 Rev.OOS5 Page 15 of 260

3. 0 PRECAUTIONS AND UMITATIONS (continued)

4. If U nit 1 reactor pre.ssufeex.ceeds 100 psig Of a Group i I oo<ition occurs 00 Unit 1 while Shutdown Cooling is iii operation, the following wnl occur for the given condition:

a. (100 psig) RHR SHUTDOWN COOLING SUCT OUTBD and INBD ISOl VLVs,1-FCV-74-47 and 1-FCV-74-48, close, thus tripping operating Unit 1 RH R PUmpl!L .

b. (Group II) RHR SYS I and IIlPCllNBD INJECT VALVEs, 1-FCV-74-53 and 1-FCV-74-67. close and Unit 1 RHR SHUTDOWN COOLING SUCT OUTBD and INBD ISOL VLVs, 1-FCV-74-47 and 1-FCV-74-48, close, thus tripp~ng operating Unit 1 RHR Pumps.

5. To reopen RHR SYS I{II) LPCIINBD INJECT VALVE, 1-FCV-14-53(67),

after a loss of Shutdown Cooling from one of the al)ove conditions, RHR SYS 1(11) SO ClG INBD lNJECT(INJ) ISOL RESET, 1-XS-74-126(132) pus41-t)utton must be depressed after either of the following occur:

a. Isolation signal has been reset

b. l-FCV-74-410r 1-FCV-74-48 are fully closed.

6. If after a GROUP II Isolation, RHR SYS I{II) LPCI1NBD INJECT VALVE, 1-FCV-74-53(61J is given an OPEN Signal prior to depressing the RHR SYS 1(11) SD ClG INBD INJECT(INJ) ISOL RESET, l-XS-74-126(132) then the valve will travelfufl open and full close unless given a close signal prior to traveling ful! open.

1. The RHR spray/cooling valves, 1-FCV-74-57(71), will receive an auto closure ~gnai in the presence of a lPClinitiation ~gnal and they are interlocked to prevent opening if the in-line torus spray valve, 1-FCV-74-.58(72), is NOT ful~ closed. The in-line vawe interlock can be by-passed if the following conditions exist

a. Reactor level is >2/3 core height t). LPClinitiation signal is present

c. Select reset switch is in the SELECT position.

The requirements for'"2/3 core he~ght and a LPCI ,initiation ~gnal may be by-passed using the keylock bypass swikh, 1-XS-74-12.2I130.

8. If primary containment cooling is desired with reactor level at <213 core height, the keylock bypass switch must be placed in BYPASS before the select reset switch is placed in SELECT to ensure relay logic is made up.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet SFN loss of Power to One RPS Bus 1-AOI-99-1 Unit 1 Rev.V017 Page 5 of9 3.0 AUTOMA TIC ACTIONS NOTES An overview ottne automatic actions is provided here. A detailed list ofthe actlons is provided in 1-01-99, Illustration 1 which lists the actions mat occur when RPS buses are de-energized on a transfer of power supply.

A. An RPS trip logic A(B) half scram occurs.

B. One-half of the PC IS Group 1 trip logic is de-energized.

C. Inboard (outboard) isolation of PCIS Group 2, shutdown cooling mode of RHR.

D. tnboard and olUtboardisolation of PCIS Group 3, RWCU on loss of RPS A and oufboardisolation on loss of RPS B.

E. Inboard and outboard isolation of PCIS Group 6, primary containment vent and purge,. reactor building venlilatiofl.

F. Group B, TIP.

G. Control Bay Emergency Pressurization System A & B start H. Standby Gas Treatment System starts.

4.0 OPERATOR ACTIONS 4.1 Immediate Action

[1] STOP all testing that could cause RPS half scrams or PCIS Logic isolation signals.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Reactor Protection System 1-01-99 Unit 1 Rev. DOll Page 50 of 68 Illustration 1 (Page 4 of4}

RPS Bus A Of" B Power Transfer C. loss of power to RPS Bus B only will result in iheiolilowi~ events in addition 10 1t1ose listed for RPS Bus A or S power loss VALVE FUNCTION/SYSTEM ACTION 1-fCV-074-0047 RHR SD CLG SUCT OUTSO ISOL 'ilL'll CLOSES 1-FCV-074-OO67 RHR SYS II lPCIINBO INJECT 'ilL'll CLOSES 1-FCV-075-oo58 PSC PUMP SUCTION OOTBO ISOL CLOSES 1-FCV-On-OO15B DRYWElL EQ DR SUMP OUTBD FCV CLOSES 1-FCV-On-OOO2S ORWEll FO SUMP OUTSD ISOLATION VtV CLOSES 1-FCV-iJ69-oo02 RWCU OUTBD SUCT ISOl 'ilL'll CLOSES 1-FCV-iJ69-OO12 RWCU SYS RETURN ISOL VtV CLOSES 1-FCV-001-OO15 MAIN STEAM LINE A OUTBD ISOl VlV AC control! power DE-ENERGIZES 1-FCV-OO 1-0027 MAIN STEAM LINE B OUTSD ISOI.. VlV AC control! power DE-ENERGIZES l-FCV-OO 1-0038 MAtN STEAM UNE C OUTBO ISOl VLV AC comrol power DE-ENERGIZES l-FCV-001-oo52 MAIN STEAM UNE 0 OUTBO ISOL VLV AC comro! power DE-ENERGlZES l-FCV-OO 1-0056 MAIN STEAM LINE DRAIN OUTeO iSOL VLV CLOSES l-FCV-043-oo14 REACTOR RECIRC OUTeD ISOlATION VLV CLOSES

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 20S000A3.03 Tier # 2 Ability to monitor automatic operation of Shutdown Cooling 1 Group #

including: Lights and alarms.

KIA # 20S000A3.03 Importance Rating 3.S 3.3 I Proposed Question: RO # 4 Given the following plant conditions:

Unit 2 is aligned with RHR Loop I in Shutdown Cooling and RHR Loop II in standby readiness.

A leak occurs in the RPV, which results in the following conditions:

- RPV level at 0 inches and slowly lowering

- Drywell Pressure at 3.0 psig and slowly rising

- RHR Pumps 'A' and 'C' TRIPPED Which ONE of the following describes the minimum actions required to align RHR Loop II for injection to the RPV?

A. After FCV-74-47 OR FCV-74-48 is closed, push the RHR SYS II SD CLG INBD INJECT ISOL RESET 2-XS-74-132.

B. After FCV-74-47 AND FCV-74-48 are closed, start RHR Loop II pumps, reset PCIS, and open the inboard injection valve.

C. After FCV-74-47 OR FCV-74-48 is closed; reset PCIS, push the RHR SYS II SD CLG INBD INJECT ISOL RESET 2-XS-74-132, and open the inboard injection valve.

D. After FCV-74-47 AND FCV-74-48 are closed, reset PCIS, push the RHR SYS II SD CLG INBD INJECT ISOL RESET 2-XS-74-132, and open BOTH injection valves.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: A Explanation: a. Correct answer.

The valve alignment is correct and resetting 2-XS-74-132 is correct.

b. However, resetting PCIS, re-opening FCV 74-47 and re-starting RHR pumps are NOT required.

The valve alignment is incorrect. Both 74-47 and 74-48 must be

c. closed. Resetting 2-XS-74-132 is correct. However, resetting PCIS is not required to re-start RHR pumps.

d. The valve alignment is correct and resetting 2-XS-74-132 is correct.

However, resetting PCIS is not required to re-start RHR pumps and the injection valves will open automatically.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 2-01-74 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank # RO 295021G2.4.50 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 3/25/2008 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Residual Heat Removal System 2-01-74 Unit 2 Rev. (13)

Page 22 of 367 3.5 INTERLOCKS (cootinued)

5. The RI-IR outboard LPGI injection valves, 2-FCV-74-52(66), have throttling capability. They receive an auto open signal:in the presence of a lPGI initiation signal when Reactor pressure is ~450 psig and are mtenocked open under these conditions for 5 minutes, or until the appropriate LPCl SYS r (SYS II) OUTBD INJ VLV BYPASS SEL keylock Swltc'h, 2-HS-74-155A(155B), is placed in the BYP.A.SS position. AdditionaUythese valves are .interlocked to prevent opening when reactor pressure is

>450 psig if its in-line companion valve 2-FCV-74-53(67) is not fully closed.

6. If Unit 2 reactor pressure exceeds 100 psig or a Group II isolation occurs on Unit 2 while Shutdown Coolingisin operation, the following will occur for the given condition:

(100 psig) RHR SHUTDOWN COOLING SUCT OUTSD and INBD ISOL VLVs, 2-FCV-74-47 and 2-FCV-74-48, close, thus tripping opernting Unitt 2 RHR Pumps.

(Group tl) RHR SYS ~ and II LPCIINBO INJECT VALVEs, 2-FC'.,.'-74-53 and 2-FCV-74-61. close and Unit 2

( RHR SHUTDOWN GaOLING SUCT OUTSO and INBD ISOLVLVS, 2-FCV-74-47 and 2-FCV-14-48, close, thus trtpping operating Unit 2 RHR Pumps.

If RHR SYS !(II) lPCllNBD INJEGT VLV, 2-FCV-74-53{67) is giving an OPEN signal prior to resetllng the RHR SYS 1(11) SD CLG INBO INJECT lSOl, after a GROUP lilsoiation. The valve travels fun open and full close unless given a dose signal prior to traveling full. open.

To reopen RHR SYS 1(11) LPCIINSD INJECT VLV, 2-FCV-74-53(67),

after a loss of Shutdown Cooling from one of the abov conditions, RHR SYS 1(11) SO ClG INBD INJECT ISOL RESET pushbutton is required to be depressed after either of following occur:

(1) Isolation ~gnal has been reset OR (2) 2-FCV-14-47 or 2-FGV-74-48 is fully closed.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 206000A2.04 Tier # 2 Ability to (a) predict the impacts of the following on the HPCI 1 Group #

system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal KIA # 203000A2.04 conditions or operations: A.C. Failures Importance Rating 2.7 3.0 I Proposed Question: # 5 Unit 2 is at 100% rated power when the following annunciators are received:

"HPCI 120 VAC POWER FAILURE" (9-3F W7),

"HPCI LOGIC POWER FAILURE" (9-3F W3).

Which ONE of the following describes the current HPCI status and the action required to return HPCI to a normal standby lineup?

The HPCI system (1) initiate and inject if required. ________~(2~)--------to restore HPCI to a normal standby lineup.

(1) (2)

A. will Restart Div II ECCS inverter per 2-AOI-S7-11, "Loss of Power to an ECCS ATU PaneljECCS Inverter."

B. will Transfer 2S0V RMOV Board A to ALTERNATE per 0-01-570, "DC Electrical System."

c. will NOT Restart Div II ECCS inverter per 2-AOI-S7-11, "Loss of Power to an ECCS ATU PaneljECCS Inverter."

D. will NOT Transfer 2S0V RMOV Board A to ALTERNATE per 0-01-570, "DC Electrical System."

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: D Explanation : a. HPCI will not initiate or inject with a loss of 120V power. In addition, the logic power failure annunciator would not occur is only the ECCS inverter was de-energized.

b. HPCI will not initiate or inject with a loss of 120V power. However, the transfer of the 250V RMOV Board to alternate is the correct action to take.

c. Past (1) is correct, however the logic power failure annunciator would not occur is only the ECCS inverter was de-energized.

d. correct answer

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 2-AOI-57-11 (Attach if not previously provided) 0-01-57D Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New RMS 6/20/2008 Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL171.042 Page 44

a. **The Division II ECCS ATU inverter (powered **SOER 83-3 from 250V RMOV Board A) supplies power to (Recommendation the flow controller and the following instruments 11 )

on Panel 9-3: Obj. V.B.6 (1 ) PI-73-31A, Pump Discharge Note that these loads are supplied (2) PI-73-28A, Booster Pump Suction from the HPCI Inverter on U-1 (3) PI-73-4A, Steam Supply (4) PI-73-21A, Turbine Exhaust (5) FIC-73-33, Flow Ind Controller Obj. V.B.6 Obj. V.C.6 (6) If Division II ECCS inverter output is lost, HPCI 120VAC FAILURE (XA Note: on U-1 may 3F-7) would alarm and flow controller indicate a loss of fails downscale, Control valve closes if logic bus B open. If accompanied by HPCI LOGIC POWER FAILURE (XA-55-3F-3),

would indicate a loss of power from 250V RMOV Board A.

(7) If Division I ECCS inverter and converter output is lost, HPCI will not initiate from DIV I logic. LlS-3-58A and B will be lost.

b. Relay Logic Bus A (Div I) is powered from 250V XA-55-3F(3)

RMOV Board 2B. It supplies power to half of the low level circuit. It also supplies isolation logic Channel A (Div I). If lost, HPCI can still initiate and isolate on all signals.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN DC Electrical System 0-01-51D Unit 0 Rell.0114 PaQe 193 of 241 8.7 Tnmsfer of Power Supplies to 250V Reactor MOV Boards PI REVIEW all Precautions and limitations En Section 3.0. o CAUTIOH When any unit 250VOC RMOV Board A Of B is transferred to alternate power suppty, it is possible that a transfer of EHC from ReactOf Pressure Control to Header Pressure Control to occur due to a los of power to 2 of the 4 reactor press instruments. IPER 1139297]

NOTES

1) Tripping of the normal or alternate feeder breakers to 250VReactOf MOV Boards on overcurrent results in lockout of both breakers.

2) Both nOfITlal and alternate feeder breakers to a 250'1 Reactor MOV Board are electrically interlocked to prevent simultaneous closure (paralleling) of DC sources.

3) The normal and aitemate feeder breakers are located on the 250V Reactor MOV Board whi;ch they supply.

4) Trip Test push-buttons are used only fOf testing racked out nOfmai and alternate feeder breakers.

5) Transfer requires tvva operators due to the distance betw'een the normal and altemate

( feeder breakers.

6) Prior to transferring any 25OVDCRMOV Boord to the alternate supply, Precaution and Limitation 3.0A must be complied with.

7) Transfer of 250V RMOV BD 3A will cause annunciation of the fol!OINing alarms:

3-XA-55-3C, Window t, RCIC RELAY LOGIC POWER FAILURE

3-XA-55-3C, Window 32, ADS BlOWNDOVVN POWER fAILURE

3-XA-B5-3E, Windaw 23, 480V RX MOV SD 0 BACKUP SWIN EMER POSN

3-XA-55-3F, Window 28, HPCI GLAND SEAL CONDENSER HOTWELL lEVEL lOW

3-XAcS5-5B, Window 34, PNl 9-47 FUSE FAILURE

3-XA-55-4B, Window 22, 4160V RPT SO 3-H CONTROL ABNORMAL

[21 CHECK power availability of the EMERGENCY (NORMAL) supply breaker as follows:

A Voltmeter indicates greater than or equal to 250 volts. o B. V04tageRelay is reset o

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BfN DC Electrical System 0-01-570 Unit 0 Rev. 0114 Page 194 of 241 8.1 Transfer of Power Suppl:ies to 250V Reactor MOV BoardS (continued)

[3] PLACE NORMIEMBRG TRANSFER SWITCH in the AlT(NOR) position. D

[4] HOlD the EMERGENCY(Normal) supply BREAKER CONTROL SWITCH in CLOSE. D

[5] TRIP the Norma&(Emergency) supplY BREAKER CONTROL SWITCH. D

[6] CHECK Emergency(Normal) SUPPtY BREAKER is CLOSED. D

[7] CHECK NORMAl(EMERGENCy) SUPPLY BREAKER is OPENITRIP. D

[8] RELEASE the Breaker Control Switches. D

[9] CHECK control room panels for iIflIY abnormal indications.

REFER TO Caution and Notes beginning this section. D

ES-401 Sample Written Examination Form ES-401-S Question Worksheet BFN Loss Of Power to an ECCS ATU 2-AOI-S7-11 Unit 2 PanellECCS Inverter Rev. 0008 Page 4 of 31 1.0 PURPOSE This aboormal operating instruction provides symptoms, automatic actions, operator actions, technica. specification requ:irements, and reportabmty requirements resulting from a loss of power to EGCS ATU Panel 9-81 or 9-82 or loss of an ECGS inverter.

NOTES

1) Each Inverter provides eectrical power to divisional logics plus one of the two redundant power supplies to its divisional ATU cabinet (REFER TO Illustration 5). The powersuppties to ECCS ATU Panel 9-81 (Div. 0, EGGS ATU Pane 9-82 (Dtv.lI) and the ECCSinverters are as foltoYtls*:

Panel 9-81

Dtvision I ECGS inverter 250V RMOV Board 26, compartment BA.

Dtvision I 250124vdc converter 250V RMOV Board 28, compartment 1B1.

Panel 9-82

Dtvision II ECCS inverter 250VRMOV Board 2A, compartment 11A1

Division II 250124vdc converter 250V RMOV Board 2A, compartment 9A1.

Power will be lost to an ECGS ATU Panel due to the loss of the respective 250V RMOV board listed aboVe,openingidoss of both of the breakers listed above, loss of ECGS ATU Panel internal fuses, or simultaneous loss of both redundant 24vdc power supplies in each ECCS ATU panel.

2) The total loss of power to the EGGS A TU panels results in power loss to all instrumentation on:

Division I Panel 9-81 (Aux Instrument Room)

Division II Panel 9-82 (Aul( Instrument Room)

RHR system I(II} containment spray valVe operation will require the use oHhe manual override Switch 2-XS-74-122(130) upon a loss of Panel 9-81(82) due to a toss of the two-thirds core height level channel.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 209001Kl.13 Tier # 2 Knowledge of the physical connections and/or cause-effect 1 Group #

relationships between Core Spray and the following: Leak detection.

KIA # 209001 K1.13 Importance Rating 2.8 3.0 I Proposed Question: RO # 6 Unit-1 is operating at 100% rated power when CORE SPRAY SYS I SPARGER BREAK (9-3C W14) alarms on Panel 9-3.

Which ONE of the following describes the principle of operation of the Core Spray Leak Detection instrument due to a Core Spray pipe break between the RPV wall and the core shroud?

The pressure sensed in the Core Spray pipe will be _ _~(1'=-J-)_ _ causing a _----10(.=.2)1--_ ~P to be sensed by the Leak Detection ~P transmitter.

(1) (2)

A. higher lower B. higher higher

c. lower lower D. lower higher

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: C Explanation: a. Part (1) is incorrect. The reference leg of the ~P transmitted normally senses pressure inside the core shroud below the steam separators and compares it to the pressure just above the core plate. If the Core Spray pipe breaks between the RPV wall and the core shroud, the ~P transmitter reference leg will now sense pressure in the steam dome above the steam dryers, which is rv 7 psig lower due to the pressure drop across the dryers and separators. Part (2) is correct. The reference leg pressure would lower, which would reduce the ~P sensed by the ~P transmitter.

b. Part (1) is incorrect as stated in (a) above. Part (2) is incorrect. However, this would be consistent with Part (1) if the break occurred in the variable leg rather than the reference leg of the ~P transmitter.

c. Correct answer.

d. Part (1) is correct. Reference leg pressure lowers due to the break. Part (2) is incorrect. This would be correct if the break occurred in the variable leg rather than the reference leg of the ~P transmitter.

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ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): OPL 171.045, Core Spray System (Attach if not previously provided)

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

Proposed references to be provided to applicants during examination: None Question Source: Bank#

Modified Bank # (Note changes or attach parent)

New 09/09/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

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

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Note: Delta P Instrument is connected to give a positive reading at rated conditions in order to provide fail-safe instrument.

On a Core Spray piping break the actual delta P increases while the indicated delta P decreases Above Core Plate Pressure

">: ...e----- Standby Liquid Control

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.045 pages 29 & 30 of 49:

1. Leak Detection Core Spray piping penetrates the drywell, reactor vessel and shroud. If a pipe break occurred between vessel wall and the shroud, Core Spray function would be lost. Pipe break detection system monitors the integrity of the Core Spray piping and alarms in Control Room.

a. Pressure 1 (P1) is greater than PS due to the jet pump driving force.

b. P1 is greater than P2 due to the pressure drop across the core plate.

c. P2 is greater than P3 due to the pressure drop across the core. (This 8.P is small.)

d. P3 is greater than P4 by 7 psi due to the pressure drop across the steam separators.

e. P4 is greater than PS by 7" of water due to the pressure drop across the steam dryer.

f. The low side of the detector senses above-core plate pressure (P2) plus the pressure due to the height of water in the vessel. Under normal conditions the high side of the detector senses core exit pressure (P3) plus pressure due to the height of water in the sensing leg. With the plant operating at rated conditions the detector reads +3.S psid. P3 isslightly less than P2 due to the 8.P across the core. Therefore, the pressure differential detected is mainly due to the height of cold water (13S0F) in the high leg of piping. If the Core Spray piping breaks between the reactor vessel and the shroud, piping is now sensing PS instead of P3, and the high-side pressure at the detector would decrease by 7 psig. Sensed low-side pressure will remain the same. This would cause the 8.P to decrease, causing an alarm to sound at 2 psid decreasing (following a 1S-sec time delay). During cold shutdown conditions this alarm will normally be in.

This is due to low-side pressure being greater than high side pressure (negative 8.P).

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panel 1-9-3 1-ARP-9-3C Unit 1 1-XA-55-3C Rev.lI019 Page 17 of 38.

SellsoriTrip Point:

CORE SPRAY SYSI 1-PDIS-075-OO28 2 psig lowering .tiP (15 second time delay)

SPARGER SiREN<

1-PDA~75-28 (page 1 of 1)

Sensor 1-LP NL-925-OO57 Location: Rx Bldg, El 565" R-3 S-UNE Probable A.. Indication of Core Spray piping break imide primary containment Cause: B. Low core flow.

C. Matrllllclion of sensor.

Automatic None AcitOll:

Operator A.. DISPATCH personnEl! to 1-LPNL-925-D057 to check CSS SYS 1 Hi ActiOll: .dP, 1-PDIS-075-0028. COMPARE with CSS SYS II Hi.dP, 1-PDIS-075-0056, on same pane!. (The norma! reading shoulid be

( ap.proxtmate"/3.S psid.) o B. If necessary, THEN DISPATCH IMs 10 VERIFY instrument operation. o C. If there are indications of a broken Core Spray header, THiEN CONSIDER the associated Core Spray system INOPERABLE and TAKE appropriate action as required by Tech Spec 3.5.A {TS SectiOll 3.5.1}. o D. If there are no indkaoons of a Core Spray header break, THEN REFER TO Tech Spec table 3.2.8 (TS Section 3.3.5.1). o

References:

1-45E620-2-1 G E 730E930-2 & -8 47W600-59 1-47E61 0-75-1 (Technical Specifi'caoons 3.5.1 and 3.3.5.1)

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

( Examination Outline Cross-reference: Level SRO 211000K6.03 Tier# 2 Knowledge of the effect that a loss or malfunction of the following 1 will have on the SLC System: AC Power.

Group #

KIA # 211000K6.03 Importance Rating 3.2 3.3 I Proposed Question: RO # 7 Given the following Unit 3 plant conditions:

480 V Shutdown Board 3A tripped due to an electrical fault.

An A TWS has occurred requiring the initiation of Standby Liquid Control (SLC).

Which ONE of the following describes the SLC pump which should be started and the status of the squib valves once the appropriate pump is started?

The OATC should start the (1 ) SLC Pump. Once started, (2) should fire.

(1 ) (2)

A. 3A both squib valves.

( B. 3A only the "A" squib valve.

C. 3B both squib valves.

D. 3B only the "B" squib valve.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: C Explanation: a. Part (1) is incorrect. SLC Pump 3A is powered from 480V Shutdown Board 3A, which is de-energized. Part (2) is correct. Both squib valves receive power from two sources. Each squib has two primers on separate power supplies.

b. Part (1) is incorrect as stated in (a) above. Part (2) is incorrect as described in (a) above.

c. Correct answer.

d. Part (1) is correct. 3B SLC pump has an operable power supply. Part (2) is incorrect as stated in (a) above.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 3-01-63, SLC System (Attach if not previously provided)

OPL 171.039, SLC System Proposed references to be provided to applicants during examination: None

~

Question Source: Bank # 211000K5.04 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Standby liquid Control System 3-01-63 Unit 3 Rev. 0020 Page 60f30 3.0 PRECAUTIONS AND UMITAliONS A. The Unit SRO/RO or Shift Manager are the only persons authorized to inject SlC solution.

B. SlC Pump Operation

t. 3Aand 3B SLC PUMP HAND SWiTCHES, 3-HS-063-0006AA and 3-HS-063-00H6B,are for pump starting onty. The squib valves will not fire when using these control switches.

2. Starting either SLC pump from the control room fires both squib valves.

3. The SLC pumps are interlocked so that only one pump can be run at a time. Operation of both SLC pumps simultaneously may result in overpressurization of the system.

4. [!!IF] SLC pump abnormal noise (similar to uncoupled or no load condition),

lack of nomlal test tank perturbations, or smell of bumt packing may indicate that the pump is air bound. These positive displacement pumps do not deliver flow if air bound. [Incident Investigation II-B-90-134)

(

C. SlC System Heating

1. The use of heat tracing is optional. Fuses 3-FU2-063-0005AB for the nonnal heat trace circuit {480V RMOV Bel 3A, Compt 12A) and 3-FU2-063-0005BB for the alternate heat trace circuit (480V RMOV Bd 3B, Compt SA) will have to be installed if heat trace is to be used.

2. The SLC tank heaters are set to cut off at approximately 830 gallons in the tank.

D. Adequate mixing time (20 minutes) is required to be strictly enforced to ensure representative sampling. Excessrve mixing times should be avoided (i.e.

approximately 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ).

E. When SLC is being air mixed, the SlC system is to be considered INOPERABLE due to the possibility of air entrapment in the SlC pumps rendering them air bound. The SlC system witi' be OPERABLE when the air mix IS no longer rn operation.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.039 Pages 15 & 16:

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1. SLC Pumps a) Two 100% capacity, triplex, positive displacement piston pumps are installed in parallel.

b) 'A' pump is powered from 4S0V Shutdown Board A.

c) 'B' pump is powered from 4S0V Shutdown Board B.

d) Electrically interlocked so that only one pump will run at a time. This prevents system overpressu rization.

e) The pumps are manually started from the main control room using the key-lock switch on panel 9-5, or locally, using the Test Permissive Transfer Switch at Panel 25-19.

f) A control room start signal will fire the explosive valves. A local start will not fire the explosive valves.

2. Explosive Valves a) Two 100% capacity explosive (Squib) valves, FCV 63-SA and B, are installed in parallel.

b) Provide a zero leakage seal between the boron solution and the reactor.

( c) Each valve contains two firing primers, powered by the 250V DC control power from the 4S0V Shutdown Boards A and B, (unit specific).

d) Either primer is capable of actuating the valve.

e) The primer is fired by taking the main control room handswitch, HS-63-6A, to the START PUMP A or START PUMP B position. This forces the ram outward, which shears the end cap off the valve fitting, allowing flow to pass through the valve.

f) After firing, the ram remains extended. This prevents the sheared cap from obstructing flow through the valve.

g) The primer requires a minimum current of 2 amps to fire, and fires within 2 milliseconds after this circuit is applied. All the explosion by-products are retained in the trigger explosive chamber.

h) Each valves firing circuit continuity is monitored by a blue indicating light on Panel 9-5 and a current meter located in the back of Panel 9-5.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 212000A1.06 Tier # 2 Ability to predict and/or monitor changes in parameters associated 1 Group #

with operating the RPS controls including: Reactor Power.

KIA # 212000A1.06 Importance Rating 4.2 4.2 I Proposed Question: RO # 8 Unit 1 is performing 1-SR-3.3.1.1.8(1l), "Reactor Protection System Manual Scram Functional Test" with the following conditions:

A manual scram was inserted on RPS Channel "A".

All four (4) SCRAM SOLENOID GROUP A LOGIC RESET, red indicating lights extinguished.

The OATC failed to properly reset the V2 scram on RPS "A" and SCRAM SOLENOID GROUP A LOGIC RESET red indicating lights 2 and 3 remained extinguished.

A manual scram was then inserted on RPS Channel "B".

Which ONE of the following describes the final STEADY STATE condition of the plant to this event and the reason for that condition?

Inserting a manual V2 scram on RPS Channel "B" will cause _ _....J.(,.=.1J-)_ _ of the control rods to insert. This is caused by (2)

(1) (2)

A. 50% Scram Discharge Volume high level.

B. 50 0/0 Backup Scram Valve actuation.

c. 100% Scram Discharge Volume high level.

B. 100% Backup Scram Valve actuation.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: C Explanation: a. Part (1) is correct. Initially, 50% of the control rods insert, but within a few seconds the Scram Discharge volume fills enough to initiate a full scram.

Since the stem asks for STEADY STATE conditions, 100% of the control rods will insert. Part (2) is correct.

b. Part (1) is incorrect. Initially, 50% of the control rods begin to insert, but the remainder will begin to insert from a full scram signal before the first 50% reach full-in. Part (2) is incorrect. Backup Scram Valves will only actuate as a result of the SDV high level, which has already initiated a scram.

c. Correct answer.

d. Part (1) is correct as stated in (a) above. Part (2) is incorrect. Backup Scram Valves will only actuate as a result of the SDV high level, which has already initiated a scram.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 1-SR-3.3.1.1.8(11) (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/10/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: NOTE: Each HCU accumulator hold approximately 5 gallons of water which enters the SDV on a scram. With Approximately 90 control rods inserting, 450 gallons of water are directed to the SDV, which will initiate a full scram at 46-50 gallons. It doesn't take long. Simulator times were on the order of 2 to 3 seconds.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.028 page 16 of 50:

a. SCRAM discharge volume high level, (1) Initiates a SCRAM while adequate volume is available to receive SCRAM discharge water to assure that all operable drives will fully insert (2) Level is sensed by two mechanical float switches and two electronic level switches (RTD's) in each instrument volume.

(3) East Instrument Volume

LS-8S-4SE (A 1-float)

LS-8S-4SF (81-float) so gal - float

LS-8S-4SG (A2-thermal)

LS-8S-4SH (82-thermal) 46 gal - thermal (4) West Instrument Volume

LS-8S-4SA (A 1-thermal)

LS-8S-4S8 (81-thermal) 46 gal - thermal

LS-8S-4SC (A2-float)

LS-8S-4SD (82-float) so gal - float (S) When the instrument volume fills up to the setpoint, the sensors open contacts in both RPS Trip Systems. Therefore, one SDV that is full would initiate a full SCRAM

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 212000KS.Ol Tier # 2 Knowledge of the operational implications of the following concepts 1 Group #

as they apply to RPS: Fuel Thermal Time Constant.

KIA # 212000K5.01 Importance Rating 2.7 2.9 I Proposed Question: RO # 9 Which ONE of the following Reactor Protection System scram signals is delayed for six (6) seconds once the setpoint has been exceeded and the basis for that delay?

The scram signal for _ _~(.:::..1)1--_ _ _ is delayed for six seconds to take into consideration the (2)

(1) (2)

A. APRM Flow-biased STP (.66W +66 % ) fuel thermal time constant.

B APRM Flow-biased STP (.66W +66%) nominal MSIV closure time.

( C. APRM High Flux < 120% RTP fuel thermal time constant D. APRM High Flux < 120% RTP nominal MSIV closure time.

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

( I Proposed Answer: A Explanation: a. Correct answer.

b. Part (1) is correct. Part (2) is incorrect. The high flux trip of 120% is designed to provide protection against an MSIV closure and the resultant pressure transient.

c. Part (1) is incorrect. The 120% high flux trip is not delayed by RPS. The flow-biased high flux scram is designed for slow power increases such as feedwater heating problems. Part (2) is correct, but not for the setpoint given in Part (1).

d. Part (1) is incorrect. The 120% high flux trip is not delayed by RPS. The flow-biased high flux scram is designed for slow power increases such as feedwater heating problems. Part (2) is incorrect, but the MSIV closure time is relevant with regard to the 120% scram setpoint since the basis for that setpoint is to protect against a MSIV closure event.

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): TSR 3.3.1.1 (Attach if not previously provided)

OPL 171.148, PRNM Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/10/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

(

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

(

OPUl1.148 Revl5ioo6 5801150 (3) The re~ctor mode s,,\liteh provides the

'"RUN" \IS, "OUT of RUN" S'gflllli to deterrrunelf the SET DOWN function is app4ied, This setdown flJl1!ction refers to the reactor mode switch 1ul'Vction, (4) Each APRM provides a digital input that monitors the state of the Reactor Mode SwitCh, The input is. actIve (signal present) when ttleswitch is in the "RUN" position and inaC'livQ for ail other $witeh positions, (5)

(6) The slgnruis used to that V,S,12 only 0l!!6 APRM channel IS bypassed.

r~ll'&'~\>\i~1?}.l.~.* ft~mgnaJ IS interpreted as

( tor theAPftMs Once an APRM channel ts all trip 1unetion *nputs to the voters are also (7) The APRM m"'*.."'ww:..... ";::"",,;:,,U\iilo1l:!

from the Two*Out*Of~Foor indjclt#ng APRM we1:1 as lPRM and OPRM) is bypassed.

(8) The APRM ca4culates theavet'ags neutron V.B.13 flu*)(, The flux is the V.CA

~\lerage of non bypassed LPRM with gain correction app4iedso that the signal corresponds t'O rsac!or power. The averaged LPHM \falue is f~(*.lsted to read In units 'Of "percent of rated core therma1 power",

(9) The SfP signal is a result of applying a 6 second filter to the average l1ux Signal which approximates the tirT'lerespoose of reactor thermal power.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet RPS Instrumentation 3.3.1.1 Table 33.1.1-1 lpagel of 3)

Reactor PmIedioo Sy$tem InsIrumel1lalial APPUCABLE C8HDlTICNS MODES OR. REQUiRED REF-RENCED FUNCTICN OT:'i-lER CHANNaS FROM SURVEilLANCE AllOWA9U SPECIRED PER TRIP REQUIRED REO'.JIREMENTS VALUE CONDITIONS SYSTEM ACTICN D.l

1. w.E!'1T,ediaie Range Mo!¥~ .

a. !ieulmn Rut( - High G SR 3.3.1.1.1 :; ~20n25 SR 3.3.1.1.3 ~i::m. of rut SR3.3.1. 1.;5 liGiiIle SR 3.3.1.1.6 SR3.3.1.1.9 SR 3.3.1.1.14 H SR 3.3.1.1.1 :; 12Oi125 SR 3.3.1.1.4 di...i&i::m. of 1141 SR 3.3.1.Ul< liGiiIle SR 3.3.1.1.14

b. ,1oDp 2 3 6: S,R 3.3.1.1.3 SR 3.3.1.1.14 5'fal 3 H SR 3.3.1.1.4 SR 3.3.1.t.14

2. Average P _ Range r.mtcrs

a. !ieutron Au:.: - High, 2 lIb} :3 SR 3.3.1.1.1 Se1J:Iov.n SR3.3.1.1.6 SR 3.3.1.1..7 SR 3.3.1.1.n SR 3.3.1.1.113

b. FtiJwBia6ed S'muIated SR 3.3.1.1. :;0.00 IN rnermal Pewer - High SR 3.3.1. 1.2 +OO'!.RTP SR 3.3.1.1..7 and:; lXl%

SR3.3.1.U3 R,-p(cj SR 3.3.1.Ue

c. !ieulmn Aux. - High SR 3.3.1.1.1 :; 120% RTP SR 3.3.1.1.2 SR 3.3.1.1.1 SR 3.3.1.1.13 SR3.3.1.U6 BFN-UNIT 1 Amendment No. 236, ~2, 269 March 06, 2007

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet RPS Instrumenmtion B 3.3.1.1 BASES APPLICABLE 2.b. Average Power Range Monitor Flow Biased Simulated SAFETY ANALYSES, Thermal Power - High lCO,and APPLIGABIUTY The Average Power Range Monitor Flow Biased Simulated (continued) Thermal Power - High Function monitors neutron flux to approximate the THERMAL POWER be,ing transferred to the reactor coolant The APRM neutron flux is electronically filtered with a time constant representative of the fuel heat transfer dynamics to generate a signal proportional to, the THERMAL POWER in the reactor. The trip level is varied as a function of redrcolation drive flow (i.e., at lower 'core flows, the setpoint is reduced proportional to the reduction in pDwer experienced as core flow is reduced with a fIXed control rod pattern) but is clamped at an upper limit that is always lower than Dr equal to the Average Power Range Monitor Fixed Neutron Flux - High Function Allowable Value. The Average Power Range Mooi:tor Flow Biased Simulated Thermal Power - High Function provides protection aga.insUransients where THERMAL POWER increasessJowly (such as the loss offeedwater heating event) and protects the fuel cladding integrity by ensuring that the MCPR Sl is not exceeded. During these events, the THERMAL POWER increase does nDt significantly lag the neutron flux response and, because ofa IDWe!r trip setpoint, wild initiate a scram before the high neutron flux scram.

For rapid neutron flux increase events, the THERMAL POWER lags the neutron flux and the Average Power Range Mooi:tor Fixed Neutron Flux - High Function will provide a scram sigmai before the Average Power Range Monitor Flow Biased Simulated Thermai Power - High Function setpoint is exceeded.

( continued)

BFN-UNIT 1 B 3.3-12 Revision 0, 40, 45 February 27, 2001

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet RPS Instrumentation B3.3.1.1 BASES APPLICABLE 2.b. Average Power Range Monitor Flow Biased Simulated SAFETY ANALYSES, Thermal Power - High (contrnued)

LCO,and APPLICABILITY Each APRMchannel uses one total drive flow signal representative of tota:1 core flow.. The total drive flaw signal is generated by the flow processing logic, part of the APRM channel. by summing up the flow calculated from two flow transmitter signal inputs, one from eadl of the two recirculation loop flows. The flow processing Iogft OPERABILITY is part of the APRM channel OPERABILITY requirements for this function.

The damped Allowable Value is based on analyses that take credit for the Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function for the mitigation of the foss of feedwater heating event The THERMAL POWER time constant of < 7 seconds is based on the fuel! heat transfer

( dynamics and provides a signal proportional to the THERMAL POWE R. The term "W" Ln the equation for determining the Allowable Value is defined as total recirculation flow in percent ofratoo.

The Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function is required to be OPERABLE in MODE 1 when there is the possibility of generating excessive THERMAL POWER and potentially exceeding the SL applicable to high pressure and core flow conditions (MCPR SL). During. MODES 2 and 5, other IRM and APRM Functions provide protection for fuel cladding integrity.

(continued)

BFN-UNIT 1 B 3.3-13 Revision G;-4O October 26, 2006

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet RPS tnstrumentation 83.3.1.1 BASES APPLICABLE 2.(;:. Average Power Range Monitor Fixed Neutroo Flux - High SAFETY ANALYSES, LCO, and The Average Power Range Monitor Fixed Neutron Flux - High APPLICABILITY Function is capable of generating a trip signal to prevent fuel damage or excessiveRCS pressure. For the overpressurization protection anafysis of Reference 4, the Average Power Range Monitor Fixed Neutron Flux - High F unction is assumed to tenninate the main steam isolation valve (MS1V) dosureevent and, along with the safety/relief valves (SfRVsl. limits the peak reactor pressure vessel (RPV}

pressure to less than the ASME Code t;imits. The control rod drop accident (CRDA) analysis (Ref. 5) takes credit for the Average Power Range Monitor Fixed Neutron Flux - High Function to terminate the CRDA.

The Allowable Value is based on the Analytical Limit assumed in the CRDA analyses.

( The Average Power Range Monitor Fixed Neutron Flux - High Function is required to be OPERABLE in MODE 1 where the potential consequences of the anatyzed transients could result in the SLs (e.g .. MCPR and ReS pressure) being exceeded.

Although the Average Power Range Monitor Fixed Neutron Flux

- High Function is assumed in the CRDA anaJysis, which is applicable in MODE 2, the Average Power Range Monitor Neutron Flux - High. (Setdown) Function conservatively bounds the assumed trip and, together with the assumed IRM trips .. provides adequate protection. Therefore, the Average Power Range Monitor Fixed Neutron Flux - High Function is not required in MODE 2.

(continued)

BFN-UNIT 1 B 33-14 Revision G;-40 October 26, 20G6

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

( Examination Outline Cross-reference: Level SRO 21S003G2.4.S Tier# 2 Knowledge of how abnormal operating procedures are used in 1 Group #

conjunction with EOPs: Intermediate Range Monitors KIA # 215003G2.4.8 Importance Rating 3.8 4.5 I Proposed Question: RO # 10 Given the following Unit 1 plant conditions:

A reactor scram has occurred.

All control rods did NOT fully insert following the scram.

1-EOI-1, "RPV Control" has been entered based on low RPV level.

RPV level is (-) 15 inches and rising with feed water injection.

1-EOI-2, "Primary Containment Control" entry is NOT required.

All Intermediate Range Monitors are fully inserted and reading on Range 5 and lowering.

Which ONE of the following describes the appropriate action to monitor and control reactor power?

(1) path RC/Q of 1-EOI-1, "RPV Control" and control reactor power using (2)

(1) (2)

A. Remain in 1-AOI-100-1, "Reactor Scram" and 1-01-85, "CRO System."

B. Remain in 1-EOI Appendix 10, "Insert Control Rods using Reactor Manual Control System."

c. Exit 1-AOI-100-1, "Reactor Scram" and 1-01-85, "CRO System."

O. Exit 1-EOI Appendix 10, "Insert Control Rods using Reactor Manual Control System."

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: C Explanation: a. Part (1) is incorrect. With IRMs on Range 7 or below and no entry conditions for EOI-2, the reactor is subcritical with no boron injection. This requires exiting RC/Q and entry into AOI-100-1. Part (2) is correct. Actions in AOI-100-1 and 01-85 are authorized even if RC/Q is NOT exited as required so long as those actions do not interfere with EOI actions. In this case, they would not.

b. Part (1) is incorrect as stated in (a) above. Part (2) is. incorrect. Actions in EOI Appendix 1D involve bypassing RWM and other more drastic actions that are not necessary if the reactor is subcritical without boron injection.

Therefore, those actions should NOT be performed.

c. Correct answer.

d. Part (1) is correct. The given conditions indicate sub-criticality and no requirement for boron injection. Retainment Override step RC/Q-3 directs RC/Q exited and AOI -100-1 entered for power control. Part (2) is incorrect as stated in (b) above.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 1-EOI-1, 1-AOI-1 00-1, 1-01-85 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/11/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.202 page 49:

1. Step RC/Q-3

a. The second retainment override statement directs the operator to transfer reactor power control actions if present plant conditions are such that the reactor is subcritical and no boron has been injected. As used in EOls, the term "subcritical" means that reactor power is below the heating range and not increasing. If it is determined that the reactor is subcritical without having injected any boron into the RPV, an exit to AOI-l00-l, Excerpt from OPL171.201 page 25:

a. Determination of Shutdown Margin (1) EOI-l, C-l,C-2, C-4, C-5, requires that a determination of the ability of the reactor to remain subcritical under all conditions without boron, be made.

(2) During ATWS conditions, when the reactor is subcritical, the conditions of EOI Note 1 should be evaluated. If necessary, Reactor Engineering should be requested to determine if the reactor will remain subcritical under all conditions

( without boron injection.

(3) This request should be made as soon as possible after it is known that rod insertion is no longer possible in order to facilitate later actions in the EOls, if needed.

Excerpt from OPL171.201 page 30:

b. Subcritical (1) When used in the EOls, subcritical means reactor power below the heating range and not trending upward.

(2) (Reactor power on range 7 and lowering of the IRMs with the IRMs inserted.)

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet BFN Reactor Scram 1-Aot-100..:1 Unit 1 Rev. 0003 Page 10 of 61 4.2 Subsequent Actions (continued)

[8] [lN1"OICJ CHECK ali control rods are fully inserted as indicated 00 the full core display or on the ICS NSSS FULL CORE DISPLAY and REQUEST PRINT ROD POSITION LOG on the lCS NSSS menu. [lNPO SOER fII).{lOO] 0 NOTE Step 4.2[8.1] may require support from off-site organizations and an extended period may elapse before results are obtained.

[8.1] IF a.1 rods are NOT inserted to Position 02 or beyond, THEN DIRECT Reactor Engineer to commence determination that the reactor will remain subcriticalunder all cond~tions wi1hout boron. 0

[91 [lNPOK:] IF any control rod fails to fully insert and it is required to Re-scram, THEN PERfORM the following, as requ1red. [lNPO SOER8I!).6IIIJ6j

[9.1] RESET the scram per Steps 4.2[23] thru 4.2[23.10). o

[9.2] VERIFY WEST and EAST eRD DISCH VOL WTR LVL HIGH HALF SCRAM annunciators (1-XA-55-4A, window 1 and 1-XA-554A, window 29) are reset. o

[9.3] INITIATE a manual scram.REPEAT Step 4.2(9), as necessary, as long as rod motion is observed. o

[10] [lNI"OICJ IF any contro~ rod fails to fully insert and it is required to Drive Controt Rods, THEN REFER TO 1-01-85. [lNPO SOER~ o

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet BFN Control Rod Drive System 1-01-85 Unit 1 Rev. 0005 Page 4 of 119 Table of Contents (continued) 6.8 Operations with Rod Worth Minimizer InsertM!ithdraw Errors .................................... 56 6.9 Reactor Vessel Level Instrumentation System (RVUS) Backfi*11 System Operations ......................................................................................................................... 58 6.10 CRD pump operation at elevated flow ..............................................................*......... 60 7.0 SHUlDOWN .............................. u .................................................................................. 63 7.1 Control Rod Drive Hydraulic System Shutdown .......................................................... 63 8.0 aN:FREQUENT OPERAT*'O'NS ...................................................................................... 64 8.1 filling and Venting the CRD Hydraulic System ........................................................... 64 8.2 Precharging Hydraulic Control Unit Accumulators ...................................................... 69 8.3 Reducing HCU Accumulator Nitrogen Pressure ......................................................... 77 8.4 Recharging Hydraulic Control Unit Accumulators ., ......................................,.............. 80 8.5 Dra.ining Hydraulic Control Unit Accumulators ............................................................ 86 8.6 Removing a Hydraulic Control Unit from Service ....................,.................................... 88 8.7 Returning a Hydraulic Control Unit to Service ............................................................. 96 8.8 Venting a Hydraulic Control UniL .................................................................................. 100 8.9 Timing Adjustment of Control Rods ........................................................................... 106 8.10 CRD Internal Ban Check Valve FlUsh ............................................................................ 110 8.11 Condensate Purge Alignment of CRD Hydraulic System.......................................... 114 8.12 Securing Condensate Purge Alignment of CRD Hydraulic System .......................... 118 8.13 Rod Drift Alarm Test ., ............................................................................................... 120 8.14 Reactor Manual Control System Timer Test and Operational TesL .......................... 122 8.15 Control Rod Diffkult to Withdraw .............................................................................. 124 8.16 Control Rod DifficuJtto Insert ..................................................................................... 132 8.17 Manual Bypass*of the Rod Worth Minimizer ............................................................. 135 8.18 Reinitialization of the Rod Worth Minimizer ................................................................ 136 8.19 Control Rods Which fail to fully Insert After Scram .................................................. 137 8.20 CRD FlUsh and Exercise During auta*ges ................................................................... 146 8.21 Alternate Procedure for Removing a Hydraulic Control Unit from Service or Restoring a Hydraulic Control Unit: to Service When No fuel is in the Reactor Vessel ............................................................................................ 148 8.22 Manual Actuation of the ATWS (ARIlRPT) ................................................................ 151 8.23 Manual Operation of 1-fCV-85-11A(B) Using 1-PCV-85-11 ...................................... 152

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet 1-EOI APPENDJX-1D BFN INSERT CONTROL ROnS USING Rev. 0 UNIT 1 REACTOR MANUAL CONTROL SYSTEM Page 2ef3 LOCATION: Unit 1 Control Room, Panel 1-9-5 ATTACHMENTS: 1. Core Positfon Map This EOI' Appendix may be executed concLirrently with EOl Appendix 1A or 18at SRO's discretion when time and manpower perm it

1. VER!FY at least one CRD pump in service.

Crosing 'I-SHV-085-0586, CHARGING WTR ISOL, valve may reduce the effectiveness of EOI Appendix 1A or 1B.

2. IF .................. Reactor Scram Of ARt CANNOT be reset, THEN ........... DISPATCH personnel to dose 1-SHV-085-0S8S, CHARGING WTR ISOL (RB NE, E! 565 ft}.

3. VER1FY REACTOR MODE SWITCH in SHUTDOWN.

4. BYPASS Rod Worth Minimizer.

S. REFER TO Attachment 2 and INSERT control rods in the area of highest power as follows:

3. SELECT control rod.

b. PLACE eRD NOTCH OVERRIDE switch in EMERG ROO IN position UNTIL control rod is NOT moving inward.

c. REPEAT Steps 5.a and 5.b for each control rod to be inserted.

6. WHEN .......... NO further control rod movement is possible or desired, THEN ........... DISPATCH personnel to VERiFY open 1-SHV-085-0586, CHARGING WTR ISOL (RB NE, EI565 ft).

END OF TEXT

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 21S004A2.02 Tier # 2 Ability to (a) predict the impacts of the following on the Source 1 Group #

Range Monitor system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of KIA # 215004A2.02 those abnormal operations: SRM INOP condition.

Importance Rating 3.4 3.7 I Proposed Question: RO # 11 Given the following Unit 1 plant conditions:

A reactor startup is in progress following refueling, with ALL 8 Reactor Protection System (RPS) Shorting Links installed.

The reactor is in Mode 2.

IRM "G" is on Range 7, all other IRMs are on Range 8.

An electronic failure in the 'B' Source Range Monitor (SRM) drawer results in a SRM HIGH/INOP (9-SA W13) alarm.

Which ONE of the following describes the plant response and required action(s), if any, to continue the startup?

The SRM failure will initiate a (1) . The startup may continue (2) bypassing SRM "B" in accordance with 1-01-92, "Source Range Monitor System."

(1) (2)

A. Control Rod Withdraw Block after B. SRM HIGH/INOP alarm ONLY without

c. Control Rod Withdraw Block without D. SRM HIGH/INOP alarm ONLY after

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: A Explanation: a. Correct answer.

b. Part (1) is incorrect. If the reactor was in Mode 1, this would be correct with IRM "G" still on range 7. Part (2) is incorrect. Rod withdrawal is not possible with the SRM INOP until it is bypassed.

c. Part (1) is correct. If the reactor was in Mode 1, this would be incorrect.

Part (2) is incorrect. Rod withdrawal is not possible with the SRM INOP until it is bypassed.

d. Part (1) is incorrect. If the reactor was in Mode 1, this would be correct with IRM "G" still on range 7. Part (2) is incorrect. Rod withdrawal is not possible with the SRM INOP until it is bypassed.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 1-01-92, Source Range Monitor System (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # 215004A3.03 attached New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Original question RO 215004A3.03:

Given the following plant conditions:

A reactor startup is in progress following refueling, with ALL 8 Reactor Protection System (RPS) Shorting Links removed.

The reactor is approaching criticality.

An electronic failure in the 'B' Source Range Monitor (SRM) drawer results in an SRM HIGH/HIGH output signal.

Which ONE of the following describes the plant response?

A. A Rod Out Block ONLY.

B. A Rod Out Block and 1/2 Scram ONLY.

C. A "SRM HIGH/HIGH" alarm ONLY.

D. A Full Reactor Scram.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet SFN Source Range MoOOors 1-01-92 Unit 1 Rev. DOOG Page 14 of 14 IItustramm 1

{Page 1 of 1}

SRM Trip Outputs TRIP SIGNAl SETPOINT ACTION SRMHigh 6.8 X 104 counts per second Rod block un4ess IRMs on Range 8 (or higher) or REACTOR MODE SWITCH in RUN SRM loop A. Module unplugged Rod block unless IRMs on Range 8 (or higher) or REACTOR B. Mode sWitch not in operate MODE SWITCH in RUN C. HV power suppiy low voltage D. Loss of +1-24 vae SRM Downscale 5 counts per second Rod block unless IRMs on range 3 (or higher) or REACTOR MODE SWITCH in RUN SRM Detector 145 counts per second Rod block unless detector full-in, Wrong Posmon IRMs on range 3 (or higher), or

( REACTOR MODE SWITCH in RUN SRM High-High 2 x 105 counts per second Scram if shorting links removed

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 21S00SK2.02 Tier# 2 Knowledge of electrical power supplies to the following: APRM/LPRM Group # 1 (APRM Channels)

KJA# 215005K2.02 Importance Rating 2.6 2.8 I Proposed Question: RO # 12 Given the following Unit 1 plant conditions:

Operating at 100% rated power.

The normal feeder breaker to 4KV Shutdown Board C inadvertently trips.

The alternate breaker from Shutdown Bus 1 fails to close.

4KV Shutdown Board C is now being powered from C Diesel Generator.

Which ONE of the following describes the effect on Average Power Range Monitors (APRM) and Rod Block Monitors (RBM) due to this electrical transient?

A. Power Range Neutron Monitoring (PRNM) is not affected by this transient.

(

B. All APRM channels generate a Critical Fault and RBM B generates a Non-critical fault.

C. All APRM channels generate a Non-critical Fault and RBM B generates a Critical fault.

D. All APRM channels and RBM channels generate a Non-critical fault.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: C Explanation: a. Incorrect. APRM channels generate non-critical faults due to a loss of RPS B. RBM B generates a Critical Fault. The loss of RPS B can be determined because of the time required for C DIG to start and tie to the 4KV SID board. The RPS MG set flywheel cannot hold speed and voltage long enough to prevent a UV trip.

b. Incorrect. APRM channels generate non-critical faults and RBM B generates a Critical Fault since its interface panel has lost power.

c. Correct answer.

d. Incorrect. RBM B generates a Critical Fault since its interface panel has lost power.

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 1-01-92B, PRNM System (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/11/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Average Power Range Monitoring 1-01-92B Unit 1 Rev. 0008 Page 8of27 3.0 PRECAUTIONS AND LIMITATIONS (continued)

N. The OPRM channellcell provides an INOPERATIVE ALARM (Inverse Video) when the quantity of operating OPRM cells is less than 23. When the number of LPRMs in a cell is reduced to less than 2, the OIP RM Cell is considered Inoperable. The OPRM function is disabled when the reactor mode switch is in a position other than RUN or the Reactor is operating outside of the OPRM Auto Enable region.

O. The.Qperators Display Assembly, which normally monitors LPRM or APRM function, wil! automatically switch over to OPRM monitoring when the reactor is placed in the region of potential instabflity. The region of potential instability is bounded by at least 25% power and less than or equal to 60% total recire drive flow (OPRM Auto Enable region) from any one of the channels.

P. The message "OPRM TRIP ENABLED" will be displayed for each APRM when entering. the power/flow region where instabiHty can occur. The message will be replaced with "ANTICIPATED1NSTABIUTY" whenever a Pre Trip (alarm) setpoint has been reached by any of the OPRM algorithms. If an osciflation trip exists, as defined by the OPRM trip setpoints, the message will be repl.aced with "INSTAB,IUTY DETECTED" and when two of these types of trips occur, an RPS automatic scram is received.

Q. The operator has the ability to transfer the display back from OPRM to APRM by depressing the "ETC" softkey.

R There are a total of four Qperators Display ~ssemblies (ODAs), two for the APRMsJOPRMs and two for the RBM. Each APRM ODA provides indication for two APRMs/OPRMs. All four ODAs are powered by 1& C BUS "A".

S. The following are power supplies for the APRMIOPRM:

Panel 1-9-14 is made lip of 5 Chassis, 4 APRMs and 1 RBM.

There are five Quadruple .how yoltage .E.ower ~upplies, one per bay on Pane! 1-9-'14 .

Each QLVPS receives power from both RPS busses. l VPS1 and 2 are fed rromRPS A, LVPS 3 and 4 are fed fromRPS B.

For each QLVPS, LVPS 1 and 4 feed the APRM and RBM A Chassis and LVPS 2 and 3 feed LPRM and RBM B Chassis.

Each Voter is powered from the RPS bus it serves, such that those assigned to RPS sub-channels B1 and B2 are powered from RPS Band those assigned to RPS sub-channels Ai and A2 are powered fromRPS A. These power supplies are seen at the bottom of the panels on the QLVPS and are indicated energized by the illuminated green lights.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Average Power Range Monitoring 1-01-92B Unit 1 Rev.OOOS Page 9 of 21 3.0 PRECAUTIONS AND UMITAnONS (continued)

T. A loss of an RPS A or B will resua in a GmicaJ Fault on RBM A or B (respectively), and a Non-Critica~ Fault on APRM channels (All APRM and lPRM Chassis continue 10 operate).

U. lPRM Alamls high at 100% and downscale at 3%. The 5O~d box above the ba,rgraph indicates that the setP04nt marker is presently exceeded while the holow box indicates a past condition. A past condition can be reset by entering the TRIP STATUS dis~ay and pressing the RESET MEMORY softkey.

V. The total number o.f LPRMs that may be bypassed (failed) is 23. If the number of bypassed (fa.iled) LPRM inputs exceeds the minimum number required in the APRM average, <<20 total or <: 3 per level) an APRM INOP CONDITION is applied, resulting in a Rod Wi1hdrawal Block and a trouble alaml on the APRM channel djs~ay in Inverse Video. This APRM INOP CONDITION is not an automatic trip but does render the associated APRM inoperable.

W. Bypassed lPRMs are not used by PRNM system for calibration, flux, or tfip points. LPRMsare normally bypassed from Panel 1.:9-14 using "BYPMV ON" or "BYP/HV OFF".

BYPfHVON LPRM is manually bypassed with the voltage on the detector_Indication of the detector output is available, but the signal is not included for any input to the APRMlOPRMIRBM functions.

BYPfHVOFF LPRM is manually bypassed with the voltage off. No detector output In addition, LPRMs may be bypassed as lndkated below:

BYP/IV LPRM is automatically bypassed as a remotely initiated IN (current to Voltage check)proress is in progress.

BYP/CAL LPRM is automatically bypassed while the lPRM is being calibrated (CALIBRATE) o.r that the ca:l,ibration is being checked {CAL CHECK}.

BYP/SUS'O LPRM is bypassed while undergoing TRIP CHECK BYPIFAULT LPRM is automatically bypassed as a LPRM self-test fault is detected.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.148 page 69:

OPL't11.148 RIiIVkllil{);1'\ 8 t'l9of 150 (g) QLVPS, four provide there is power 10 l"I"IOdules, Am tlluminated indicating lamp t$ only anlndicatkoo *01 power to the l VPS module and IS not an indication that the l VPSmoduie is functioning properly.

(11) In case of one powersupplyiaiture, the APRM self testing process will identify it a$ a non-cribcaf fault. and the n01'l"l'l81 01 APHM wm continuec (2) D4scu~s the ooPA.NM~

(a) Voters far channelsi and 31000 (b) Non critical fault on all APRM, l?RM. and RBM instruments.

(c) Critical fatllt 0(1 RaM channel A since it's interlace panel has lost A Rod block is initiated from RaM cnaa'1nel "N'.

(3) 105801 RPS*S" is similar.

(4) If one votef ~ powered dow-'n wftb trip stgRl1 Fe,.". and bypassed, tmm tt'pstgna' wltll remain in and bypass

{cunel.on sUIt works.

If powered down with no trips and with no bypallts ,the X and Y relays are deone'rglzed which will Input to RPS (112 scram).

(5) The loss 01 a lVPS sup~y's output causes a alarm

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 217000K3.04 Tier # 2 Knowledge of the effect that a loss or malfunction of the RCIC 1 Group #

system will have on the following: Adequate Core Cooling.

KIA # 217000K3.04 Importance Rating 3.6 3.6 I Proposed Question: RO # 13 Given the following Unit 2 plant conditions:

A Group I isolation occurred from Steam Tunnel high temperature.

Feedwater flow on "A" feedwater line indicated upscale.

HPCI and RCIC automatically initiated and RPV level was slowly restored until both HPCI and RCIC tripped on high RPV level.

RPV level again began to lower.

With RPV level (-) 25 inches and lowering, the Board Unit Operator noticed that 2-71-8, RCIC Steam Supply Valve suddenly lost light indications for valve position.

Which ONE of the following describes the status of the RCIC system and, based on that status, the effect on RPV level recovery efforts?

The RCIC Steam Supply Valve, 2-71-8 will (1) without power when RCIC receives another initiation signal. Based on this response, RPV level will (2)

(1) (2)

A. remain open lower due to RCIC injection into the "A" feedwater line.

B. remain open rise due to RCIC injection into the "B" feedwater line.

c. remain closed lower due to HPCI injection into the "A" feedwater line.

D. remain closed rise due to HPCI injection into the "8" feedwater line.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: C Explanation: a. Part (1) is incorrect. 2-71-8 will close on a high level trip. Part (2) is incorrect. RPV level will lower, but not based on RCIC injecting into the "A" FW line. HPCI injects into the "A" FW line.

b. Part (1) is incorrect. 2-71-8 will close on a high level trip. Part (2) is incorrect. RPV level will not rise even though RCIC does inject into the "B" FW line. RCIC will not run with 2-71-8 failed closed.

c. Correct answer.

d. Part (1) is correct. 2-71-8 will remain closed and RCIC will not initiate. Part (2) is incorrect. HPCI injects into the "A" FW line, which is isolated and broken. HPCI will run at rated flow but the water will be injected into the steam tunnel and flow into the Reactor Building, not the RPV.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 2-01-71, OPL 171.040, OPL 171.042 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/12/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Reactor Core Isoiation Coo'ing 2-01-71 Unit 2 Rev. 0055 Page 90f70 3.0 PRECAUTIONS AND LIMITATIONS A. Turbine controls provide for automatic shutdown of the RCICturbine upon receiving any of the folrowing signafs {REFER TO Section 8.4 for auto actions):

1. High RPVwater level (+51 in.); 579 in. above vessel zero. The RCIC TURBINE STEAM SUPPLY VLV, 2-FCV-71-8, and RCIC PUMP MIN FLOW VALVE, 2-FCV-71-34, will close at +51 in. and will RE-OPEN when RCIC re-initiatesat -45 in.RPV water level.

2. Turbine overspeed (Mechanical, '122.3% of rated speed).

3. Pump low suction pressure (10 inches Hg vacuum}.

4. Turbine high exhaust pressure (50 psrg).

5. Any isolation signal.

6. Remote manual trip (RC1C TURBINE TRIP push-button., 2-HS-71-9A, depressed).

B. RCIC turbine steam supply wifl isolate from the followrng signals (REFER TO 2-AOI-64-2C for auto actions):

1. RCIC steamline space temperature at s;;180°F Torus Area or s;;180"'F RCIC Pump Room.

2. RCIC turbine hig.h steam flow (150% flow, 3-second time delay.)

3. RCIC turbine steam line low pressure (73 pS4g).

4. RCIC turbine exhaust diaphragms ruptured (10 psig).

5. Remote manual isolation (RCIC AUTO-IN IT MANUAL ISOLATION push-buttO'n., 2-HS-71-54, depressed,only if RC~C initiation signal is present).

C. The RCIC turbine will auto initiate on RPV Low-LO'w Water Level, -45 in.

(REFER TO Section 5.1 for autO' actions.)

O.ln the presence of a RCIC initiatiO'n signal, the RCIC PUMP MIN FLOW VALVE, 2-FCV-71-34, opens when system flow is below 60 gpm and closes when flO'w is above 120 gpm. The valve win NOT auto open on low flow if an initiation signal is NOT present E. RGIC PUMP MfN FLOW VALVE, 2-FCV-71-34, will open on receipt of an initiation signal even with RCIC turbine manually tripped resulting in slowly draining CST to Suppression Chamber.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.040, RCIC page 20:

m. Injection Valve (FCV-71-39) (Normally closed)

Powered from 250VDC RMOV Board C. Pump discharges through a thermal sleeve into B FW line downstream of the outboard isolation check valve.

Excerpt from OPL171.042, HPCI page 11:

a. Water path (1) Normal condensate path from CST to the HPCI pump, to the A Feedwater line and into the reactor vessel

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 218000K6.03 Tier# 2 Knowledge of the effect that a loss or malfunction of the following 1 Group #

will have on the ADS System: Nuclear Boiler Instrument System (level indication). KIA # 218000K6.03 Importance Rating 3.8 3.9 I Proposed Question: RO # 14 Given the following Unit 1 plant conditions:

A steam line break in the drywell has occurred.

All control rods inserted on the scram.

RPV level is (-) 70 inches and steady with automatic HPCI injection.

All attempts to initiate Drywell Sprays have been unsuccessful.

Drywell temperature is 390 OF and rising.

RPV Saturation Temp (Curve 8) has been exceeded.

RPV level instrumentation has become erratic.

Which ONE of the following describes the required Automatic Depressurization System (ADS) operation and the basis for that requirement?

ADS must be manually _ _ _(1.-=1:..J-)--_ in order to _ _ _ _ _-\,.:(2~)----_

(1) (2)

A. inhibited prevent adding more energy to the Primary Containment.

B. initiated establish the conditions necessary to flood the RPV.

c. inhibited prevent a further loss of RPV water inventory.

D. initiated return to the SAFE area of the RPV Saturation Temp (Curve 8).

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

( I Proposed Answer: B Explanation: a. Part (1) is incorrect. No conditions currently exist that would require ADS to be inhibited. All rods are in and RPV level was steady before indication was lost. Part (2) is incorrect. Adding more heat to the containment is certainly not desirable at this point, but ADS must be initiated to flood the RPV in accordance with 1-EOI-C4, "RPV Flooding."

b. Correct answer.

c. Part (1) is incorrect as stated in (a) above. Part (2) is incorrect. Losing additional inventory is a secondary concern without RPV level indication.

Establishing the conditions to flood the RPV to a condition where adequate core cooling is assured is the priority.

d. Part (1) is correct. Part (2) is incorrect. Once Curve 8 has been exceeded and RPV level indication is lost, reducing pressure and returning to the SAFE area of the curve will not help restore RPV level instruments. The drywell must be cooled down and RPV level instrument reference legs must be refilled.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 1-EOI-C4, "RPV Flooding." (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/13/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from l-EOI-C4:

0---

WHEN THE RE./\CTOR. Will RE~J\IN suaCRrnCAl V>lITHOUr aORON UNDERALtCONDITIONS USEE NOTE)

THEN CONTINUE L

YES L OPEN All ADS VllJs L

NOL YES L

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from l-EOI-l Path RC/L:

I.BOVE

+2 IN.

MAiNTAiN RPV WATER LVL ABOVE -IB.2 !N.

L Rc/L-6 WHILE EXECUTING THE FOLLOWING STEPS:

If RPV WATER LVL DROPS BELOW -'120 iN.

THEN QB. INHIBlTADS THE ADS TIMER HAS !N!T!ATED L

RG/L-7 CAUTiON

3 ELEVATED SUPPR CHMBR PRESS MAY TRIP RCIC

6 HPCI OR RCIC SUCTION TEMP ABOVE 140 OF II I L AUGMENT RPV WATER LVL CONTROL AS NECESSARY WiTH ONE OR MORE OF THE FOLLOWING lNJ SOURCES; INJ SOURCE I APPX I INJ PRESS

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from l-EOI-CS, "Level/Power Control,"

PC WATER LVL CANNOT BE MAINTAINED BELOW 105 FT STOP INJ INTO THE RPV FROM SOURCES QB. EXTERNAl TO THE PC NOT REQUIRED FOR ADEQUATECORECOOL~G SUPPR CHMBR PRESS CANNOT BE MAINTAINED BELOW 55 PSiG L

C5-1 INHIBIT ADS L

C5-2 NO L BYPA.SS THE FOLLOWING fSOLAT!ON !NTERlOCKS:

MSIV LOW LOW LOW RPV WATER LVL (APPX SA)

RBVENTILATION LOWRPVWATER LVL (APPX8E)

L

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 223002K3.20 Tier# 2 Knowledge of the effect that a loss or malfunction of the PCISjNSSS 1 Group #

system will have on the following: Standby Gas Treatment system.

KIA # 223002K3.20 Importance Rating 3.3 3.4 I Proposed Question: RO # 15 Given the following Unit 1 plant conditions:

Reactor power is at 100% with all systems in a normal lineup.

Instrument Mechanics (1M) are performing calibrations on the drywell pressure sensors when an inadvertent Group 2 and Group 6 PCIS isolation signal is received on Unit 1.

Which ONE of the following describes the response of the Standby Gas Treatment (SGT) system and the effect these PCIS isolations will have on plant operation?

Standby Gas Treatment (SGT) trains _~(1'-1-) __ will be operating. As a result of this PCIS malfunction, the reactor will (2)

(1) (2)

A. A, Band C continue to operate until PCIS is restored to a normal condition.

B. A, Band C scram due to steam tunnel high temperature and Group I isolation.

C. A and B only continue to operate until PCIS is restored to a normal condition.

D. A and B only scram due to steam tunnel high temperature and Group I isolation.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: A Explanation: a. Correct answer.

b. Part (1) is correct. Even though only two trains are required for 100%

capacity and "C" SGT is powered from Unit 3, all three trains will initiate on a PCIS Group 2 or Group 6 isolation. Part (2) is incorrect. This was an issue due to the loss of Reactor building ventilation until a booster fan was recently installed in the steam tunnel to maintain temperature following a loss of normal ventilation.

c. Part (1) is incorrect. Even though only two trains are required for 100%

capacity and "C" SGT is powered from Unit 3, all three trains will initiate on a PCIS Group 2 or Group 6 isolation. Part (2) is correct. Other than steam tunnel temperature, no other systems effected by the spurious isolations pose an immediate threat to continued operation.

d. Part (1) is incorrect. Even though only two trains are required for 100%

capacity and "C" SGT is powered from Unit 3, all three trains will initiate on a PClS Group 2 or Group 6 isolation. Part (2) is incorrect. This was an issue due to the loss of Reactor building ventilation until a booster fan was recently installed in the steam tunnel to maintain temperature following a loss of normal ventilation.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): OPL 171.017, PCIS (Attach if not previously provided)

OPL171.018, SGT OPL171.067, HVAC Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/12/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.017 pages 16 & 18:

{

" 2. Group 2 This group includes Residual Heat Removal (RHR), drywell floor/equipment drain sump, and PSC Head Tank Pump suction valves.

The signals which will initiate a Group 2 isolation are:

RPV low level (+2"or Level 3)

Drywell High Pressure (+2.45 psig)

6. Group 6 This group provides for isolations of systems associated with Primary containment atmosphere control and sampling. Systems/lines isolated are as follows:

Nitrogen/Air Purge

Drywell/Suppression Chamber Exhaust

Hydrogen/Oxygen Sample Lines

Post Accident Sample System (PASS) Lines

Drywell Air Compressors

Drywell Leak Detection The signals which will initiate a Group 6 isolation are:

RPV Low Level (+2" or Level 3)

Drywell High Pressure (2.45 psig)

Reactor Bldg Vent Hi Radiation (72 mr/hr)

Refuel Zone Hi Radiation (72 mr/hr)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.018 pages 21:

1. Initiation Signals

a. System automatically starts with one or more of These signals on any the following signals. unit, will start all three SGT trains when the (1) High drywell pressure (2.45 psi) control switch is in AUTO (2) Low reactor water level: +2.0" (3) High radiation, Reactor Zone Ventilation System (72 mRem/hr) 2 monitors in the same channel or 2 monitors downscale (RE-142, -143) one in each channel.

(4) High radiation, Refueling Zone (72 mRem/hr) 2 monitors in the same channel or 2 monitors downscale (RE-140, -141) one in each channel.

b. All 3 SGT trains auto-start on initiation and run until manually stopped.

Train A gets its signals

c. 2 of the 3 trains can provide design flow from DIV. I.

conditions.

Train 8 gets its signals from DIV. II.

d. SGT will auto start with initiation signal as soon as power is available (slight delay until DIG powers Train C gets its signals SO 8d.) from both divisions.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.067 page 17:

1. Main Steam Vault Booster Fan

a. Centrifugal booster fan installed in the Main Steam Vault exhaust duct to provide cooling during hot weather months or when normal ventilation is lost.

b. This prevents unnecessary isolation of Scram Frequency MSIVs due to ambient overheating. Reduction Effort

c. Power Supplies:

Unit-2 480V RMOV Bd 2C Compartment SA Unit-3 480V RMOV Bd 3C Compartment SA

d. Fan operation is controlled from the breaker cubicle with a maintenance control switch located at the fan motor.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 239002A4.02 Tier # 2 Ability to manually operate and/or monitor in the control room: SRV 1 Group #

Tail pipe Temperatures.

KIA # 239002A4.02 Importance Rating 3.6 3.7 I Proposed Question: RO # 16 Unit-1 is operating at 100% rated power when the following annunciator alarmed:

MAIN STEAM RELIEF VALVES OPEN 1-FA-1-1 (9-3C W24)

Which ONE of the following describes the primary sensor that initiated the annunciator and one of the secondary indications which could be used to verify its accuracy?

The annunciator is initiated by the MSRV _ _ _~(1"..".)_ _ _ _ and can be verified using the MSRV (2)

(1) (2)

A. valve position indication tail pipe temperature B valve position indication tail pipe flow monitor

c. tail pipe flow monitor tail pipe temperature D. tail pipe temperature tail pipe flow monitor

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: C Explanation: a. Part (1) is incorrect. Position indication is only indicating that the MSRV has received a demand to open, not its actual status. Part (2) is correct.

Verifying the tail pipe temperature is responding to the demand is positive proof that the MSRV is passing steam from the RPV to the tail pipe.

b. Part (1) is incorrect as stated in (a) above. Part (2) is incorrect, but is also proof that the MSRV is passing steam from the RPV to the tail pipe. This would certainly be an indication to verify MSRV status, but it is also the input to the annunciator given in the stem of the question. Therefore, it is incorrect for the conditions given.

c. Correct answer.

d. Part (1) and (2) are incorrect. These are the two methods to positively verify MSRV operation, but the annunciator is based on tail pipe flow read by acoustic sensors in each tailpipe. Since a leaking (simmering) MSRV can also indicate tail pipe temperatures equal to an open MSRV, it is not used to annunciate an open MSRV.

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 1-ARP-9-3C Window 25 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/12/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

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

(

BFN Panel 1-9-3 Unit 1 1-XA-55-3C SensorITrip Point:

MAIN STEAM RELIEF VALVES 1-FMT-1-4 OPEN MSRV Tailpipe Flow 1-FA-1-1 Monitor (Page 1 of 1}

Sensor 1-FMT-1-4, Panel 1-9-3 location: Main Control Rm EI 617' Probable Main Steam Re1iefVaive is open or leaking.

Cause:

Automatic None Acti.oo:

Operator A.. CHECK MSRV Temp Recorder. 1-TR-1-1 on Pane! 1-9-47forraised Action: temp and MSRV Tailpipe Flow Monitor, 1-FMT-1-4 on Pane11-9-3 for ftow indications. o R REfER TO 1-AOI-1-1. o C. If alarm is due to sensor malfunction, THEN REfER TO 0-01-55 and OPDP-4. o

References:

1-47E610-'t-1 1-45E620-2'-1 BE 730E929-2

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 239002A4.04 Tier# 2 Ability to manually operate and/or monitor in the control room: SRV 1 Group #

Suppression Pool temperature.

KIA # 239002A4.04 Importance Rating 4.3 4.3 I Proposed Question: RO # 17 Given the following Unit-3 plant conditions:

A reactor scram has occurred due to a spurious Group I isolation.

Over 50% of the control rods failed to insert on the scram.

Reactor pressure is being maintained 800 to 1000 psig using MSRVs in accordance with 3-AOI-Appendix 11A, "Alternate RPV Pressure Control Systems, MSRVs."

Which ONE of the following describes the method of operating the MSRVs and the basis for the prescribed method?

MSRVs are opened in a _ _--\(.=-l),/--_ _ to ensure _ _ _ _(>.=2:+.)_ _ __

(1) (2)

A. numerical order by switch UNID that both CAD tanks receive an equal loading.

B. numerical order by switch UNID even heat distribution in the Suppression Pool.

c. specific order per the Appendix that both CAD tanks receive an equal loading.

D. specific order per the Appendix even heat distribution in the Suppression Pool.

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

( I Proposed Answer: 0 Explanation: a. Part (1) is incorrect. Numerical order by switch position is a common procedure violation while performing Appendix llA. Part (2) is incorrect for the given conditions. If Drywell Control Air is lost, MSRV operation is to ensure sufficient CAD tank volume is maintained. This is addressed in Appendix 11A.

b. Part (1) is incorrect as stated in (a) above. Part (2) is correct for the given conditions, but not for the answer given in Part (1) of this response.

c. Part (1) is correct. A specific order of opening is prescribed in the procedure. Part (2) is incorrect based on the given conditions. Drywell Control Air is available under these conditions unless a specific problem has occurred to cause it to be lost. No such problem was given, therefore Part (2) is incorrect.

d. Correct answer.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 3-AOI-Appendix 11A (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/12/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: I evaluated this question as CIA because the candidate must determine, based on conditions in the stem,. if Drywell Control Air is available.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet 3-EOI APPENDIX-11A Rev. :2

. . . .~. . . . . . . . . . . . . . . . . . . . . . . . . . . . .page . 1 of 3 3 - EOl APPENDIX -l.lA ALTERNATE RPV PRESSURE CONTROL SYSTEMS MSRVs UJCATION, unit .3 ccmtrol RCIOJri ATTJ..CHI>IENTS: None r---'-'--"'--~'-" ..,.,., ..... ,......

1. IF ** , .* DrY1h-.all control Air is NOT available, THEt;r " ,. EXECUTE EOI 1Ippendix SQ, CROSSTIE CAD TO DRY"RELL CONTROL AIR, COl'lCURRENTLY with this procedur .....

2., IF '" ~ ~ ~ ~ suppression pool Ifrrilel is at or below 5 ~ 5 ft .

THEN *** CLOSE MSRVs and CONTRO.I.. RPV pressure using other optiollS .

.3

OPEN MaRVs usin'g the following seque,nce to control RPV pre£,sure as directed by SR;:):

a. 1 3-PCV-1-179 rom STI>l LINE A RELIEF 'inU.,vE.

h. :2 3-PCV-1-1S0 !-1l-l' STl-! LINE D RELIEF '1lll.AL"'lE"

c. 3. 3-PCV-1-4 /-1m STM LINE A RELIEF VALVE.

d. 4 3-PCV-1-1l Mli STM LINE C RELIEF VALVE.

e. 5 3-PCV-1-:23 I-IN STM LINE B RELIEF YALVE ..

f. 6 3-PCV-1-42 /-lUi STM LINE D RELIEF 14J.LVE ..

"'"w,r...

.g. 7 3-PCV-1-30 Mli sm LINE C RELIEF VALVE.

h. 8 3 - PCV-1-19 Ml{ STIll LINE B RELIEF VALVE.

i. 9 3-PCV-1-5 IDl STM LINE A RE.LIEF 'IlALVE.

j. 1D 3-PCV-1-41 r-nr STI>l LINE D RELIEF 'lAINE.

k. I 11 3-PCV-1-22 Ml-l STI-I LINE B RELIEF ;!ALVE.

I. 1"'" I 3-PCV-1-1S 1Ml-l S~ LINE B RELIEF V*ALVE.

m.. 13 1.3 - PCV 34 1*!'IIn STM LINE C RELIEF W"LVE.

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet 3-EOI APPENDIX-l1A Rev. :2

4. IF .*... Drywell control Air header supplied from CAD system A shows indications of being depressurized as determined by Appendix 8G, THEN *** OPS. MSRVs supplied by CAD system B using the following sequence to control RW pressure as directed by SRa:

a. 6 3-PCV-1-42 MN STM LINE D RELIBF VALVE.

b. 7 3-PCV-1-30 MN STM LINE C RELIBF VALVE.

c. 4 3-PCV-1-31 MN STM LINE C RELIBF VALVE.

d. 1 13 3-PCV-1-34 MN STM LINE C RELIBF VALVE.

e. 110 13-PCV-1-41 MN STM LINE D RELIBF VALVE.

f. I2 13-PCV-1-180 MN STM LINE D RELIBF VALVE.

g. 112 13-PCV-1-18 MN STM LINE B RELIBF VALVE.

5. IF ..... Drywell control Air header supplied from CAD system B shows indications of being depressurized, as determined by Appendix 8G, THEN *** OPSN MSRVs supplied by CAD System A using the following sequence to control RW pressure as directed by SRa:

a. I9 13-PCV-1-S IMN STM LINE A RELIBF VALVE.

b. 111 13-PCV-1-22 IMN STM LINE B RELIBF VALVE.

C. 5 13-PCV-1-23 IMN STM LINE B RELIBF VALVE.

d. :3 13-PCV-1-4 IMN STM LINE A RELIBF VALVE.

e. 8 13-PCV-1-19 IMN STM LINE B RELIBF VALVE.

f. 1 13-PCV-1-179 IMN STM LINE A RELIBF VALVE.

LAST PAGB

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

(

3-EOI APPENDIX-IIA Rev. 2

~= 'l ()f "

6. IF ..... BOTH Drywell Control Air headers are depressurized, THEN ... PERFORX the following as directed by EOI-I, RPV Control, RC/P Section:

PLACE ea.ch MSRV control switch in CLOSE/AUTO, and PLACE 3-XS-1-202, MSRV AUTO ACTUATION LOGIC INHIBIT, to INHIBIT.

AND

MINIMIZE MSRV cycling by using sustained openings for RPV depressurization.

(

LAST PAGE

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 259002A4.01 Tier# 2 Ability to manually operate and/or monitor in the control room: 1 Group #

Reactor Water Level Control, All individual component controllers in the manual mode. KIA # 259002A4.01 Importance Rating 3.8 3.6 I Proposed Question: RO # 18 Given the following Unit 1 plant conditions:

A reactor startup is in progress from cold conditions.

The reactor is critical with a heat up in progress at 180 OF.

The heat up rate is currently 80 °F/hr.

1A and 1B Condensate pumps and 1A Condensate Booster pump are operating.

RPV level is (+) 30 inches and steady.

Which ONE of the following describes the appropriate method used to return RPV level to the normal control band under these conditions, and the reason for using that method?

A. CRD SYSTEM FLOW CONTROL in MANUAL can be used to raise injection flow to as high as 80 gpm.

B. RWCU BLDN FLOW CaNT in MANUAL can be reduced to reject less water to the main condenser due to thermal expansion from the heat up.

C. RFW SU LVL CaNT in AUTOMATIC can be used to control level and prevent distracting the OATC during the startup.

D. CNDS FLOW CONTROL SHORT CYCLE in MANUAL will raise Condensate Booster pump discharge pressure to raise injection flow.

ES-401 , Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: B Explanation: a. Incorrect. Although the CRD flow controller can be adjusted to 80 gpm, this is NOT done in manual. In addition, doing so under this condition will eventually result in too much inventory and has a negative impact on CRD cooling water flow while its adjusted that high.

b. Correct answer. Also the preferred method.

c. Incorrect. Placing the 5jU Level Control valve in automatic allows uncontrolled addition of water to the RPV which amounts to an uncontrolled reactivity addition under these conditions. Distracting the OATC takes a back seat to reactivity control.

d. This method will also raise RPV level but is not easily controlled. Therefore, this method is limited to controlling injection during a cooldown following a reactor shutdown.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 1-01-3, 1-01-69, 1-01-85, 1-GOI-100-1A (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/13/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Unit Startup 1-GOt-100-1A Unit 1 Rev. 0011 Page 93* .of 173 5.0 INSTRUCnON STEPS (c.ontinued)

(45) IF Reactor pressure is less than 750 psig AND a RFP is NOT being used t.o maintain Reactor water level, THEN MAINTAIN Reactor water level between 28 inches ami 50 inches as indicated on RX VESSEL lEVELlTOTAL FW FLOW recorder, 1-XR-3-53, and less than 48 inches on 1-U-3-208A(B)(C)(DJ using the following vessel makeup aoo level control systems: (NiA if RFP is being used to mainta,in Reactor water level)

CRD System (40 to 65 gpm). (Control Rod Drive Hydfau~c System Startup section of 1-01-85).

CRD System {up to 80 gpm). (CRD Pump Operation at Elevat.ed Flow section of 1-01-85).

RWCU System. (1-01-69).

Condensate System. (1-01-2).

(R)

-"'71n-:i:::'"tia"'l7s-*.- Date Time

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet SfN Control Rod Drive System 1~1-85 Unit 1 Rev. 0005 Page 60 of 179 6,10 CRD pump operation at elevated flow

[1] VERIFY CRD System in service in accordance with Section 5.1. o

[2] REViEW all Precautions andi limitations in Section 3.1. o CAUTIONS 1} Elevateclflow rates are I~kely to reduce drive water and CRD cooling water DIP.

2) Elevated flows for extended periods (> 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) are NOT recommended due to CRD Graphilar Seal erosion.

[3] PERfORM the following steps concurrently, as required, to establish a maximum of SO gpm as indicated on CRO SYSTEM FLOW, 1-FlC-S5-11:

THROTTLE CRO PUMP DISCH THROTTlING VlV, 1-THV-085-0527, to maintain pressure less than Of equal to 1500 psig as indicated on CRD ACCUM CHG INTR

( HDR PRESS, 1-PI-85-13A o

ESTABUSH the following by a;lternatefy adjusting the tape setpoint of the CRD SYSTEM FlOilv' CONTROL, 1-FIC-S5-11', and the throttled posUion of the CRO DRIVE WATER PRESS CONTROL VlV, 1-HS-85-23A:

CRD CLG INTR HDR DP, 1-PDI-85-18A, of approximately 20 psid. 0

CRD DRIVE 'NTR HDR DP, 1-PDI-85-17A, between 250 psjd and 270 psid. 0

[4] CHECK CRO STABILIZING FLOW, 1-Fl-85-22 is approximately 6 gpm at 1-LPNl-925-0018B. o

[4.11 IF CRD stabiliz~ng Frow adjustment is necessa,ry, THEN CONTACT Technical Support and REQUEST performance of O-TI-20 in ordierto adjust stabilizer needle valve setUngs. o

[5] VER.FY CRD DRrVE WTR HDR FLOW, 1-FI-85-15A is approximately 0 gpm. o

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN .Reactor Feedwater System 1-01-3 Unit 1 Rev. {)009 Page 10 of 210 7.1 RFPtRFPT Shutdown (continued I

[10.4] With RFW SU lVl CONT, 1-LlC-3-53, in MANUAL ESTABUSH approximately 70% demand signal on RFW SU l Vl CONT, HJC-3-53, using R,A,!SEllOWER push-buUons. 0

[10.5] VERIFY the RFYV START-UP LCV, 1-LCV-003-G053 is oJ)n via communications with the OJ)rator at the valve. 0

[111.6] IF required, THEN HAVE the OJ)ratof open the RFW START-UP lCV 3-53 BYPASS, 1-8Y'I/-003-0533. (N/A Otherwise) 0

[10.7] ADJUST the CNDS FLOW CONTROL SHORT CYCLE, 1-FIC-2-29, as needed to raise or lower Condensate header pressure enabling vesse.llevel control through the operating feeclpump. 0

[10.8] WHEN reactor pressure is approximately 270 psJg, THEN CLOSE 1-FCV-3-19(12)(5), RFP 1A{1 8)(1 C)

DISCHARGE VALVE using handswitch 1-H S-3-1 9,A,(12A)(5A). o

[10.9] ADJUST the CNDS FLOW CONTROL SHORT CYCLE, 1-FIC-2-29,as needed to raise or lower Condensate header pressure enabling vessel level control. 0

[10.10] IF the RFW START-UP LCV, 1-LCV-003-0053,failsto perform as expected AND makeup to the vessel is desired, THEN RE-OPEN 1-FCV-3-19(12){5)., RFP 1A(18)(1C)

DISCHARGE VAL V E. using handswitch 1-HS-3-19,A,(t2A)(5A) to re-estab.ish feed to the vessel. (Otherwise N/A) 0

[10.11] WHEN steady level has been obtained from the RFW START-UP lCV, f-LCV-003-0053, THEN DEPRESS RFPT tA.(1B)("fC) TRIP, 1-HS-3-125A(151A)(176A), to trip RFPT being removed from sefVice. 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet SFN Reactor Water Cleanup System 1-01~

Unit 1 Rev.OOJ7 Page 64 of 126 6.5 Slowdown Operation (continuedl NOTE Blowdown to the Main Condenser is preferred to reduce Radwaste processing reQulrements.

CAUTIONS 11} During blowdown operation, failure to maintain the Non-Regenera!llve Heat Exchanger O'utlet temperature less than 1::)ooF as indicated on 1-TR-69-6, will cause s. reduction O'f resin efficiency and possible resin damage.

2) Opening RWCU MAIN CONDR BDV, l-FCV-069-0016, and RWGU BLOWDOWN TO RADWASTE, 1-FCV-69-17, simultaneously during I)IDwdown operationS will result rn a loss of condenser vacuum.

3) Failure to monitor Non-Regenerative Heat Exchanger inlet and outlet temperature to maintain less than 436°F across the heat exchanger during blowdawn operations wiU cause damage to the Non-Regenerative Heat Exchanger. [5FPER OJ..OO1800"DOOl

4) Failure to manilor Non-Regenerative Heat Exchanger O'uttet temperature closely during blowdownoperations could result in an automatic RWCU isolation as sensed by RWCU NON REGEN HTX DISCH TEMP, 1-TI'8-069-0011.

5) RWCU BLDN PRESS CNTL VLV, 1-PCV-069-0015, will automaticallyiso!.ateon 5 psig low upstream pressure or 140 psig high downstream pressure.

6) Failure to closely monitor Reactor level dunng blowdown operations could result in uncontrollable level fluctuations.

[4] IF iJlowdown is to the Main Condenser, THEN OPEN RWCU SLOWDOWN TO MAIN CNDR,. using 1-HS-69-16A o

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 261000K4.03 Tier # 2 Knowledge of SGT design feature(s) and/or interlock(s) which 1 Group #

provide for the following: Moisture Removal.

KIA # 261000K4.03 Importance Rating 2.5 2.7 I Proposed Question: RO # 19 Regarding the Standby Gas Treatment (SGT) train, which ONE of the following describes the temperature set point that will trip the relative humidity heater and the basis for maintaining relative humidity within the SGT train?

The "SGT FILTER BK RH HTR CONT TEMPERATURE" annunciator will alarm and trip the Relative Humidity heater at a set point of (1) of. Moisture is controlled within the SGT train to __________~(2~)---------

(1) (2)

A. 180 prevent lowering the adsorption properties of the charcoal.

B. 180 prevent damaging the charcoal and clogging the HEPA filter.

c. 80 prevent lowering the adsorption properties of the charcoal.

D. 80 prevent damaging the charcoal and clogging the HEPA filter.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: A Explanation: a. Correct answer.

b. Part (1) is correct. The heater is tripped at 180 of. Part (2) is incorrect.

The basis of why the heater is tripped to prevent high temperatures is to prevent physical damage to the charcoal which would degrade it's mechanical filtration capability and also clog the HEPA filter, but it is NOT the basis for maintaining relative humidity.

c. Part (1) is incorrect. This value is the CHARCOAL BED VESSEL TEMP HIGH (9-52 W9) alarm for the charcoal bed in the Off-gas system, NOT the Standby Gas Treatment system. Part (2) is correct. This is the basis for removing moisture before the SGT flow enters the charcoal filter.

d. Part (1) is incorrect as stated in (c) above. Part (2) is incorrect as stated in (b) above.

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): OPL 171.030, OPL 171.018 (Attach if not previously provided)

( 1-ARP-9-53 Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/13/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet SGTSystem B3.6.4.3 BASES BACKGROUND The sizing of the SOT System equipment ;;md COiI'nPOnents is (continued) based on the results of an infiltration analysis. The internal pressure of the SOT System boundary region is maintained at a negative pressure of 0.25 inches water gauge when the system is in OfJeration. The Secondanr Containment membrane limits infiltration to oot more than the design now requirements for the SOT System under neulfal (< 5 mph) wind condHions. This adlows the SGT System 10 evacuate the entire secoooanr oontinment volume to at least a negative 0.25 inches water gauge relative to outside the membrane.

The moisWre separator is provided to remove entrained water in the air, Whilelhe elecmc heater reduces the reRative humidity of the airstream to less than 70% {Ref. 2). The prefilter removes large particulate matter. whMe the HEPA tilter remo¥es fine particulate matter and protects the charcoal from fooling.

The charcoad adsorber removes gaseous elemental iodine and organic iodides (however, no credit is taken in the radiological dose analyses for the charcoal), and the final HEPA tilter COllects any carbon fines exhausted from the charcoal adsorber.

The SOT System automatically starts and operates in response to actuation signals indicative of conditions or an accident that could require operation of the system. Following initiation, the three charcoal fiRer train fans start and run until manually stopped. Two of the three subsystems can provide designOow conditions.

(continued)

BFN-UNlT 1 B 3.6-115 Revision .:1-, 29 January 25, 2005

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.030 Appendix D:

Appendix D - CHARCOAL ADSORPTION PROCESS:

Activated, granulated charcoal provides a tremendous amount of surface area contact with many exposed chemically active sights.

Charcoal bed adsorption in our offgas system (including SBGT, containment purge units) is very effective in delay of noble gasses and even better delay on the halogen family (iodines).

Halogens such as iodine are delayed by chemical interaction with active sights in the activated charcoal. Halogens are very reactive with the hydrocarbon sights and actually combine chemically remaining as long as the halogen remains a halogen. Note that this is entirely dependent on the chemical properties and not on the nuclear properties. The charcoal beds will stop the halogens whether or not they are radioactive. The radioactive decay process of iodine normally beta decays to barium. Barium has totally different chemical properties and can actually combine with other iodine atoms. However barium will no longer remain chemically combined with the charcoal and thus would be released.

The compounds which barium would form will be particulates which can be filtered out in the post filter. Not all halogens atoms pausing through the charcoal with actually combine chemically, but would still be delayed by the second interactive process described next.

Noble gasses and halogens also are delayed by electron cloud interaction with hydrocarbons. The hydrocarbon molecules normally present a slightly positive charge to the passing offgas molecules. The noble gasses and halogens are very electron rich thus having a slightly negative charge. This causes the entrained atoms and compounds to "stick" to the charcoal like Velcro. However, the gasses will still tumble along but be significantly delayed. The delay due to the chemical properties allows radioactive decay prior to release.

Note that halogens are very reactive prior to reaching the charcoal bed, so that a lot of the atoms will be already combined chemically with other atoms. This can prevent the chemical combination but will still be delayed by the electrostatic "Velcro" sticking.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.018 page 19:

i. SGT FILTER BK A (B,C) RH HTR >180°F Annunciation (U-1 and U-CONT TEMPERATURE TA 65-12A 2 only) Turns off R-H

(-34A, -60) heater control Excerpt from OPL171.030 page 50:

y. Charcoal Bed vessel temperature high (80°F) Panel 9-53

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panel 9-53 1-ARP-9-53 Unit 1 1-XA-55-53 Rev. 0014 l:l aOe-J30{39 SensorITrip Point:

CHARCOAL BED VESSEL 1-TRS-66-115 OO°F TEMP HIGH t-TA-66-115 (Page 1ot2)

Sensor 1-TE-066-011S A Ihru G location: Off-Gas Bldg Probable A. OFFGAS REHEATER OUTLET TEMP CONTROL, 1-TIC-66-109, aut of Cause: adjustment.

B. Msoroer vault air conditioning failure.

C. Carbon bed wetting due to high off-gas moisture.

D. large quantities. of adsorbed radioactive gas.

E. Hz ignition in Off-Gas System.

F. Fire in charcoal vessel.

G. sensor 1T'..alfunction.

(

Automatic None ActiOfl:

Operator A. VERIFY the fOflowing 00 Panel 1-9-53:

Action: 1 . 1-TRS-66-115/( 1-7), adsorber vessel and vaun temperature, high. o

2. 1-TIC-66-109, OfFGAS REHEATER OUTlET TEMP CONTROL, 1-TIC-66-1IJ9, in AUTO and set at nOF.

3. 1-TRS-66-115f8. ABSORBER \tAUl T temperature, at aboul 70°F. o

4. REHEATER OUTlET (DEW POINT) temperature, 1-TRS-66-10613, <: 48°F. o

5. Hz ANALYZER A (B)".taH2,1-XR~66-103l1(f2).. IF Hz ignition is suspected, THEN REFER TO 1-.11.01-66-1. o B. VERIFY Carbon Bed Rad Mooitor, 1-RI-90-280, Panel 1-9-10 for normal reading. 0 Continued on Next Page

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 262001KS.Ol Tier# 2 Knowledge of the operational implications of the following concepts 1 Group #

as they apply to the AC Electrical Distribution: Principle involved with paralleling two A.C. sources. KIA # 262001 K5.01 Importance Rating 3.1 3.4 I Proposed Question: RO # 20 Given the following plant conditions:

Diesel Generator (DG) 3EA is running in parallel with the grid during the monthly load surveillance test.

The DG Mode Selector Switch for 3EA DG is in the PARALLEL WITH SYSTEM position.

3EA DG load is currently 2400 KW and steady.

Which ONE of the following describes the expected response of 3EA DG if the DG Mode Selector Switch was moved to the UNITS IN PARALLEL position, and the basis for that response?

The 3EA DG would (1) . This is a result of (2) speed droop control.

(1) (2)

A. trip on Overload (SlX) zero B. trip on Overload (SlX) automatic C. continue to operate normally automatic D. continue to operate normally zero

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: A Explanation: a. Correct answer.

b. Part (1) is correct. Operating in "Parallel with System" allows automatic speed droop control which monitors DG speed, frequency AND load. When placed in "Units in Parallel" position, the speed droop control becomes "Zero" and the DG only monitors speed and frequency. Since the DG speed setpoint is maintained slightly above grid frequency during the surveillance, the governor will attempt to increase grid frequency by sending more and more fuel to the DG. The result is an overload condition. Part (2) is incorrect. The DG was in automatic speed droop control BEFORE the mode switch was moved out of "Parallel with System".

c. Part (1) is incorrect. The "Parallel with System" position is the only position that uses automatic speed droop control. Part (2) is incorrect. If the mode switch is placed in "Units in Parallel" or "Single Unit", speed droop control is set to zero and the DG will respond to grid frequency.

d. Part (1) is incorrect as stated in (c) above. Part (2) is correct for the switch position but would not result in continued operation of the DG.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 0-01-82, OPL 171.038 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/15/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.038 page 20 and 21:

a. Speed Droop Control (Zero Droop Operation) b.

(1) When the generator is the only power supply to a bus, it is desirable to have the speed governor maintain constant speed and frequency regardless of load on the bus. This is "Zero Droop" operation. The regulator system simply compares generator output frequency and setpoint frequency and actuates the fuel supply to maintain setpoint frequency. In effect for Single Unit and Units in parallel.

(2) Zero Droop Operation results in the speed/frequency remaining constant as KW load is increased.

(3) If the generator were to be tied to the grid, when in Single Unit or Units in parallel, as soon as the output breaker is shut the speed regulator senses output frequency, but now the generator output frequency is fixed by the other machines on the grid. If the diesel speed setpoint is higher than grid frequency, the zero droop governor will keep advancing the fuel supply to the diesel in order to try and raise grid/DG output frequency to the governor's setpoint. This will cause the diesel to overload. (495 amps.)

(4) Droop operation is in effect for Parallel with System. In droop mode the load carried by the diesel is sensed in addition to the output frequency. If the speed setpoint is higher than grid frequency, when the output breaker is shut the governor will see generator output frequency as being too low and start advancing fuel. This will cause the generator load to pick up. As load picks up, it sends a negative speed signal back to the regulator which cancels out the difference between grid frequency and setpoint frequency. When this happens the governor will stop advancing fuel and the engine will steady out at a certain amount of load. To pick up additional load the speed setpoint is adjusted upwards and the load builds up until it has canceled out the additional speed setpoint adjustment. If a droop mode generator was the sole supply to a board, its frequency versus kilowatt load would droop.

(5) Droop mode operation of the governor is controlled by the electronic governor and is in use only when the generator mode is "PARALLEL WITH SYSTEM."

(6) The governor control, when in parallel with the grid, serves to control the KILOWATT loading on the machine.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Standby Diesel Generator System 0-01-82 Unit 0 Rev, 0094 Page 72 of H8 8.1 ParalIe:1 wfth System Operation at Panel 9-23 (continued)

CAUTION On'ly ooe Unit 1 and 2 Diesel Generator at a time is allowed to be operated in parallel wi1h system.

{51 PULL an(j PLACE the associated Diesel Generator mode selectorswitcn in PARALLELED WITH SYSTEM. o Diesel Handswitch Name Handswitch No, Pane.

A DG A MODE SELECT O-HS-82-A/SA 0~'9-23-7 B DG B MODE SELECT O-HS-82-BlSA 0~9-23-7 C DG C MODE SELECT O-HS-82-<:/5A 0-9-23-3 D DG 0 MODE SELECT O-H5-32-D/5A 0-9-23-8

( CAUTION Fa1!ure of the PARALLELED WITH SYSTEM light to illuminate in the folvowing step could iocliicate that the OG is still M1 SINGLE UNIT operation and result in overload when the DG output breaker is closed.

[11 RELEASE the Diesel Generator mode selector switch and OBSERVE PARALLELED WITH SYSTEM light illuminated. o fS] ADJUST Diesel Generator frequency using the associated Diesel Generator governor control switch to obtain a synchroscope needle rotation of one revolution every 15 to 20 seconds in the FAST direction. o Diesel HandswitchName Handswitch No. Panel A DG A GOVERNOR CONTROL O-HS-82-Af3.A. 0-:9-23-7 B DG B GOVERNOR CONTROL O-HS-82-B,f3A 0-9-23-7 C DG C GOVERNOR CONTROL O-HS-82-C!3A 0-9-23-8 D DG D GOVERNOR CONTROL O-H5-82-Dl3.A 0-9-23-8

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 262002Al.02 Tier # 2 Ability to predict and/or monitor changes in parameters associated Group # 1 with operating the UPC CAC/DC) controls including: Motor Generator outputs. KIA # 262002A 1.02 Importance Rating 2.5 2.9 I Proposed Question: RO # 21 Given the following plant conditions:

Unit 3 is in a normal lineup.

The following alarm is received:

- UNIT PFD SUPPLY ABNORMAL (9-8B W35)

It is determined that the alarm is due to a Unit 3 Unit Preferred AC Generator Over-voltage condition Which ONE of the following describes the result of this condition?

Unit 3 Breaker 1001 (1) Unit 2 Breaker 1003 _ _-->.=(2=+-)__; and the Motor-Motor-Generator (MMG) set _ _ _ _ _ _~(!:<.,3)~-----_

(1) (2) (3)

A. trips OPEN; is interlocked OPEN; automatically shuts down.

B. is interlocked OPEN; trips OPEN; automatically shuts down.

C. trips OPEN; is interlocked OPEN; continues to run without excitation.

D. is interlocked OPEN; trips OPEN; continues to run without excitation.

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet I Proposed Answer: C Explanation: Part (1) and Part (2) are correct. Part (3) is incorrect. The MMG set does

a. not automatically shut down.

Part (1) and Part (2) are incorrect. The breaker lineup is backward. Part (3)

b. is incorrect. The MMG set does not automatically shut down.

c. Correct answer.

d. Part (1) and Part (2) are incorrect. The breaker lineup is backward. Part (3) is correct. The MMG set does not automatically shut down.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): OPL171.102 (Attach if not previously provided) 3-ARP-9-8B, (W 35)

Proposed references to be provided to applicants during examination: None Question Source: Bank # 262002A 1.02 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 3/25/2008 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: In order to answer this question correctly the candidate must determine the following:

1. The 1001 and 1003 breakers from an MMG set will trip on over voltage or under frequency at the output of the MMG.

2. Unit 2 MMG Breakers are interlocked to prevent alternate power to unit 1 and 3 at the same time.

3. When an over voltage condition exists at the Generator Output, the 1001 breaker from the MMG Set trips.

4. Excitation is lost and the MMG Set continues to run.

5. The "Hold to build up voltage" switch must be depressed to restore voltage.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.102 page 20 & 21:

b. MMG Sets (Unit 2&3)

(1) The MMG is normally driven By the AC motor, powered from 480V Shutdown Board A. Should this supply fail, the AC motor is automatically disconnected and the DC motor starts, powered from 250V Battery Board. The DC motor has an alternate power supply from another 250V Battery Board. Transfer to the alternate DC source is manual. Under frequency on the generator output will trip the DC motor. Transfer of the MMG set back to the AC motor is manual.

(2) The 1001 and 1003 breakers from an MMG set will trip on over voltage or under frequency at the output of the MMG. Also Unit 2 MMG Breakers are interlocked to prevent alternate power to unit 1 and 3 at the same time.

(3) When an under frequency or over voltage condition exists at the Generator Output the following occurs (a) BB panel 10 breakers from the MMG Set trip.

U2 1001 (U2) 1003 (U1 &3)

U3 1001 (U3) 1003 (U2)

(b) Excitation is lost and the MMG Set continues to run. (The Hold to build up voltage switch must be depressed to restore voltage.)

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet BFN Panel 9-8 ; 3-ARP-9-8B Unitl l-XA-55-8B 'Rev. 0014 SensorfTrip Point:

UNITPFD SUPPLY ReIaySE - loss of nonnm DC power sooroe.

ABNORMAL ReIayTS- DC bansferswitch in emergency positioo.

Relay2MS - DC motor slarts.

Relay 3k- AC generator ovelVtlHlilge, under frequency, or loss of output voltage.

(Page 1 of 1)

Relay 13K - difference between b!lHe1y vcftage and open circuit DC motor vattage greater than -7V DC or greaterhn +12\1 DC.

Sensor 250V DC Battery Bd 3 Location: 8593' Probable A. Loss of nonnm DC power source, UNIT PFD DC NORMfEMERG SELECTOR, Cause: 3-HS-252-02 inEMERG.

B. DC power transfer.

C. DC motor starts.

D. AC generalm' overvoHage.

E_ AC generator under frequency.

F. AC generator loss of output voHage..

G. Relay fairure.

H. Sensor malfunction..

t Voltage differelltial between battery and motor is greater than -7V DC or greater than +1211 DC.

J. Burned Qut light butb 00 volage indicating switch which activates Relay 13K.

Automatk: None Action:

Operator A IF 120V AC Unit Preferred is lost, THEN Acti<m: REfER TO 3-A~-57-4. o B. After delemlining situation, REFER TO appropriate pomoo of O-OI-57C. o C. IF voltage differential behlteen battery and motor is greater fflan

-7V DC or greater than +12\1 DC, THEN PLACE boUllocal and remote OFf..,AUTO-MANUAl select switch in MANUAl to prevent possible damage to DC drive motor. o D. IF burned out light bulb is cause of alarm, THEN INITIATE WO to change out the light btJlb due to it being hard wired. o

References:

0-45E641-2 3-45E.62D-11 3-33D0015A4585

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 263000Al.0l Tier# 2 Ability to predict and/or monitor changes in parameters associated 1 Group #

with operating the DC Electrical Distribution controls including:

Battery charging/discharging rates. KIA # 263000A 1.01 Importance Rating 2.5 2.8 I Proposed Question: RO # 22 Which ONE of the following describes the analyzed time in which the 250V Battery Board can carry a full electrical load without a battery charger connected, and the analyzed transient or accident that is the basis for that time?

The 250V DC battery board can carry a full load for _--l(-.=1:.J...)__. This capability is required to provide vital equipment power during a(an) (2)

(1) (2)

A. 30 minutes. Appendix R fire.

B. 30 minutes. Design Basis LOCA.

c. 60 minutes. Appendix R fire.

D. 60 minutes. Design Basis LOCA.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: B Explanation: a. Part (1) is correct. Part (2) is incorrect. There are time sensitive actions during an Appendix R fire that must be taken to ensure the battery chargers are not lost, but the battery capacity is not the primary concern as it is during a DBA LOCA.

b. Correct answer.

c. Part (1) is incorrect. This is the time the battery is expected to carry electrical loads during normal operation. Part (2) is correct.

d. Part (1) is incorrect. This is the time the battery is expected to carry electrical loads during normal operation. Part (2) is incorrect. There are time sensitive actions during an Appendix R fire that must be taken to ensure the battery chargers are not lost, but the battery capacity is not the primary concern as it is during a DBA LOCA.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 0-01-570, OPL 171.037 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 06/15/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet SFN DC Electrical System O-OI-57D UnitS Rev. 8117 Page 16 of 241 3.0 PRECAUTIONS AND LIMITATIONS (continued)

E. All safety requirements concerning smOKtog, fires or sparks should be obsefved when in the Battery-Battery Board Rooms because of potential accumulation of hydrogen in Hammabile amounts.

F. 250V Unit Battery Charger 1,2A,2B and :3 Emergency ON select switch bypasses battery charger emergency load shed contacts. Placing the select switch in Emergency ON reestablishes charger operafioos with an acddent signal present and Diesel Generator voltage alVaftlalJle. Battery Charger 4 supply breaker, 480V Shutdown Board 3B, Compt 6D, receives a trip signal from the load shed logic and the breaker must be manually re-closed after a 40 second time delay to restore the charger to service. The annunciaiionc:ircuit forthe 250V Unit Battery Charger 3 does NOT work when the EMERIOFFION Select Swit.ch is in the EMER Position.

G. 11I1e! Neutron monitoring battery chargers are NOT stand alone power supp~es and shall only be operated while connected to the neutron mooooring batteries.

tBFFER~

H. Within 30 minutes after the loss of the normal charger to a 250V Unit Battery another charger Shall be placed in service to that battery and load reduced so that the battery is NOT discharging.

( I. jNRCtCl Upon return to service of 24V DC Neutron Monitoring Battery A or B.

Instrument Maintenance must perform functional tests on SRMs and IRMs that are powered from the affected battery board On that. the IRMs and SRMs are normaldy inoperable after entering RUN mode due to lack of testing, these tests are NJA for the IRMs and the SRMs if the Unit is in RUN Mode and the IRMs and SRMs are inoperable). Prior toc.alling the IRMs and SRMs operable, the tests have to be performed. {NRC IE 5!\6peeI FaIIIIW-up fl\em a6..o1OJ

.J. To return equipment to service following a fa~ure or trip, the shutdown section of this instruction should be performed on the equipment failed. The initial conditions may NOT be applicable in this case.

K. BNlRClCj The transfer of 250VDC control power to a 4kV Shutdown Board with a diesel generator operafing may cause an inadvertent start of a RHRSW pump.

!lEft 8&l21J'25i

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.037 page 13:

(2) Batteries The 250 volt batteries are 120-cell lead-calcium type. The Unit Batteries (Mfg type LCUN-33) have a manufacturer 1 minute discharge rating of 2080 amps and an 8-hour discharge rating of 2320 amp-hours to a 21 OV DC minimum (required ECCS components must operate with as low as 200V). Two batteries can carry maximum expected load under DBA (Design Basis Accident) conditions without recharging for 30 minutes.

The Plant/Station Batteries (Mfg type LCR-33) have a manufacturer 1 minute discharge rating of 2240 amps and a 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months discharge rating of 2320 amp-hours.

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet DC Sources - Operating B3.8.4 BASES BACKGROUND Each battery charger for a DC electrical power subsystem has (continued) ample power output capadty for the steady strte opera1ion of connected loads required during 00ffI'I;id operation, while at the same time maintaining its battery bank fUlly charged. Each batlery charger has sufficient capacity to restore the battery from the design mininlUm cnarge to its fully charged state Witmn 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months while supplying normal steady state loads (Ref. 4).

APPliCABLE The initial conditions of Design Basis Accident (DBA) and SAFETY ANALYSES transient analyses in the FSAR, Chapter 6 (Ref. 5) and Chapter 14 (Ref. 5), assume that Engineered Safety Feature (ESF) systems are OPERABLE. The DC electrical power system provides oorrmd and emergency DC electrical power for the DGs, emergency aux~iaries, and con,trol and switching during aM MODES of operatioo. The OPERABilITY of the DC subsystems is consistent with the initial assumptioos of the accident analyses and is based upon meeting the design basis of the unit. This includes maintaining DC sources OPERABLE during accident conditions in the event of:

3. An assumed loss of all offsite AC power or aM onsite AC power sources; and

b. A posrulated worst case single failure.

The DC sources satlsfy Crnelioo 3 of the NRC POlicy Statement (Ref. 11).

(continued)

BFN-UNIT 1 B3.8-60 Revision 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Safe Shutdown Instructions O-SSI-001 Unit 0 Rev. 0000 Page 88 of 97 Attachment 1 (Page 6 of 13)

TBO-31 Alignments of the 480V Control Bay Vent Board(s) and/or 480V HVAC Board B are necessary to ensure power is available to the guaranteed ventilation equipment.

TBO-32 Control Bay Chillers and Air Handling Units are necessary to maintain coo&ed areas at or below their design temperature limits.

TBO-33 An Electric Board Room AHU and Control Bay Chiller are placed in service to provide necessary cooling to Electric Board Rooms which have equipment required to support safe shutdown.

TBO-34 The CAD system is aligned to supply nitrogen to the main steam relief valve actuators as a backup supply to the accumulators for long term operation.

TBO-35 R HRSW pumps a re started to provide RH RSW flow to .RH R Heat Exchangers to be placed in service for decay heat removal.

TBO-36 FCV-23-34, 40, 46, or 52 wi!! be opened to provide a cooling water flow path

( through the RHR Heat Exchanger for decay heat removal.

TBD-37 Battery Chargers are placed in service to provide long term DC power availability for designated MSRV solenoids and RPV instrumentation.

TBO-38 Lights are tumed off in these rooms to reduce the heat load in the rooms.

TBO-39 I & C Bus transformers that are not designated safe shutdown equipment are removed from service to reduce the heat load in the associated rooms.

TB0-40 These doors are opened to provide natural circulation with adjacent areas.

This will prevent or prolong the necessity of providing portable ventilation to these rooms.

TB0-41 Battery and Board Room Exhaust Fan 'f A or 1B is started to provide ventilation to vital rooms on EJ. 593'. Board Room Supply Fan 1A or 1B is started to provide ventilation to Unit 1 EL 606' Mechanical Equipment Room.

3EA andJor 3EB Oiesel Auxiliary BoardRoom Exhaust Fans are started to support operation of Diesel Auxiliary Boards on Unit 3.

TBD-42 This(These) Board(s} is(are} aligned to its(their) a Item ate power supply due to potentiaf fire induced damage to the norma! power supply.

TBD-43 480V RMOV Board B is aligned to remove power to FCV-23-46 and/or 52.

This will allow unimpeded manual valve operation.

(

"\.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 263000KS.Ol Tier# 2 Knowledge of the operational implications of the following concepts 1 Group #

as they apply to the DC Electrical Distribution: Hydrogen Generation during battery charging. KIA # 263000K5.01 Importance Rating 2.6 2.9 I Proposed Question: RO # 23 Which ONE of the following is a concern to plant operation if the Plant/Station Battery Rooms HVAC units are not operating properly?

A. The design limit for hydrogen concentration in the rooms may be reached when the batteries are being charged.

B. Electrical Maintenance will not be able to obtain accurate Cell specific gravity readings if temperature is above 90 of.

C. The lead-calcium batteries tend to release toxic gas into the atmosphere above 90 of, and access to the room would be limited.

D. The Quarter& Battery SR frequency is lowered to weekly when temperatures are above the 70 of to 90 F temperature range.

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

( I Proposed Answer: A Explanation: a. Correct answer.

b. Incorrect. This would be correct for temperatures below 60 of.

c. Incorrect. Lead-calcium batteries suffer degraded performance at high temperatures but do not release toxic gas as a result.

d. Incorrect. This would be correct if temperatures were below the temperature range, not above it.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 0-01-570 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank # 263000K5.01 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet SFN DC Electrical System 0-01-'510 Un if 0 Rev. 0117 Page 16 of 247 3.0 PRECAUnONS AND LIMITATIONS (continued)

E. All safety requirements concerning smoking, fires or sparks should be observed when in the Battery-Battery Board Rooms bectlt.JSeof potential accumulation of hydrogen in flammatlle amounts.

F. 250V Unit Battery Charger 1,2A,2B and 3 Emergency ON select switch bypasses batterycnarger emergency load shed contacts. P1acrng the select switch in Emergency ON reestablJshes charger operations with an accident signal preselltand Dieset Genemtor voHage availahle. Battery Charger 4 suppty breaker, 480V Shutdown Board 3B, Compt 60, receives atr.ip signal from the load shed logic ane! the breaker must be manually fe-closed after a 40 second time delay to restore the charger to service. The annunciation circuit for the 250V Unit Battery Charger 3 does NOT work when the EMERJOFF/ON Seled Switch is in the EMER Position.

G. !1I1q Neutron monitoring baUery chargers are NOT stand alone power suppi'ies and shad! only be opemted while* connected to the neutron roonitoring batteries.

!5fPER 9411%2]

H. Within 30 minutes after the loss of the norma! charger to a 250V Unit BaUery another charger shall be placed in service to th.at baUeryand load reduced so that the battery is NOT discharg:ing.

I. !NlI',CfCj Upon return to service of 24 V DC Neutron Monitoring Battery l\ or B, Instrument Maintenance must perform functional tests on SRMs and tRMs that are powered tronl the affected battery board (tn lIat the IRMs and SRMs are normally inoperable after entering RUN mode due to lack of testing, these tests are N/A for the IRMs and lie SRMs jUhe Un.it is in RUN Mode and ttle IRMs and SRMs are inopemtlle). Prior to calling the IRMs and SRMs operable., the tests have to be performed. INRC IE !n6peCl fOllOW-up iterm 6-4OJ J. To return equipment to service following a failure or trip,the shutdown section of lIis instruction should be performed on the equipment fa*iled. The initial conditions may NOT be applicable in this case.

K. iNROCj The transfer of 250VDC control. power to a 4kV Shutdown Board with a diesel generator opemting may cause an inadvertent start of a RHRSW pump.

ilBHI80:!H'2Ei

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Battery CeU Parameters B 3.8.6 BASES ACTIONS AJ, .A.2, and A3 (continued)

Continued operation is only permuted for 31 days before hattery cell parameters must be restored to within Category A ami B Umits. Taking into consideration that, while battery capacity is degraded, sufficient capacity exists to perform the intended function and to altow time to fuEly restore the battery cell parameters to normal limits, this time is acceptable for operation prior to declaring the associated DC IJattery inoperable.

When any baUery parameter is outside the Category C limit for anY' connected cell, sufficient capacity to supply the maximum expected load requirement is not ensured and the corresponding DC electrical power subsystem must be declared inoperable. Additionally, other potentially extreme conditions, such as not completing the Required Actions of Condition A within the required Compiletion Time or average electrotyte temperature of representative cens falling below 60~F for each Unit and Shutdown Board battery (except Shutdown Board battery 3EB) and 40¢F for Shutdown Board hattery 3EB and each DG battery, also are cause for immediately declaring the associated DC electrical power suhsystem inoperable.

(continued)

BFN-UN!T 1 B 3.8-77 Revision 0

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet Examination Outline Cross-reference: Level SRO 264000A3.0S Tier # 2 Ability to monitor automatic operation of the EDGs including: Load 1 Group #

shedding and sequencing.

KIA # 264000A3.0S Importance Rating 3.4 3.S I Proposed Question: RO # 24 Given the following Unit 3 plant conditions:

Operating at 100% rated power with all equipment in a normal lineup.

A total loss of all off-site power occurs in conjunction with a large break LOCA.

Drywell pressure peaks at 22 psig and is subsequently lowered to 2.3 psig using Drywell Sprays.

RPV pressure lowered to 400 psig and is stable.

RPV level drops below (-) 122 inches.

Assume no operator actions.

Which ONE of the following describes the expected response of the RHR pumps and SGT system as a result of these conditions?

When the DG output breakers close, RHR pumps will start _ _(~1'-1-)_ _

The B SGT fan (2)

(1) (2)

A. in 7 seconds auto starts ONLY if A SGT fan fails to start B. in 7 seconds auto starts in 40 seconds

c. immediately auto starts ONLY if A SGT fan fails to start D. immediately auto starts in 40 seconds

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: 0 Explanation: a. Part (1) is incorrect. This time is associated with starting Core Spray pumps. Part (2) is correct. Part (2) is incorrect. The B SGT fan will auto start in 40 seconds regardless of whether A SGT fan starts.

b. Part (1) is incorrect. This time is associated with starting Core Spray pumps. Part (2) is correct.

c. Part (1) is correct. Part (2) is incorrect. The B SGT fan will auto start in 40 seconds regardless of whether A SGT fan starts.

d. Correct answer.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): O-AOI-57-1A (Attach if not previously provided)

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Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/15/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: I rated this question as CIA due to conditions in the stem which indicate that conditions for a Common Accident Signal (CAS) may not be met since DW pressure dropped below the LOCA setpoint of 2.45 psig before RPV level dropped below -122 inches. The candidate must evaluate plant conditions to determine whether a Load Shed logic trip has been initiated.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet SFN loss of Offsite Power (161 and 500 O-AOI-51-1A UnifO KV)JStation Blackout Rev. 0011 Page 1 Of 11 3,0 AUTOMATIC ACTIONS (continued)

V. Unit 112 480\1' Load Shed occurs on 11 loss ofoffsite power incoojundion with a LOCA signal:

1. One RBCCW pump auto restarts (after 40 seconds on U1 and U2).

2. Drywell Blowers auto restart on oon-accident unit (after 40 seconds).

Drywell Blowers with their respective auto restart inhihit switches in the INHIBIT position will not auto restart.

3. DrjWell coolers are manualty restarted on the accident unit A Drywall Blower with its auto restart lnhibtt switch in the IN HIBIT position can be manually restarted after a ten minute time deiay.

4. SGT TRAINS A &. B trip, but wtll AUTO RESTART in 40 seconds when an inmation signal is present.

5. Loss of Control Ba.y Ch,illed Water Pumps A 8. B. (may be restarted after 10 minutes with use of bypass SWitch).

W. Unit 3 480\1' load shedding occurs as follows:

1. Division I 480\1' load sheckUng will occur when an accident Signal is present

( and diesel generator voltage is available on the 4160V shutdown board supplying the 480V shutdown board 3A as follows:

Iil. RBCCW pump 3A trips

b. Drywel! t)lowers 3A 1 &. 3A2 trip

c. After a 40 second time delay, with the control switch in Nomlal After start, R8CCW pump 3A restarts

d. After a 40 second time delay, Drywall blowers 3A1 and 3A2 can be manualiy restarted

e. Drywel! lJiowers 3A3, 1M and 3A5 cannot be restarted untU the load shed signal is corrected

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.038 page 38:

a. If normal voltage is available, load will sequence on as follows: (NVA)

Time After Accident SID Board SID Board SID Board SID Board A C B D 0 RHR/CS A 7 RHRlCS B 14 RHR/CS C 21 RHR/CS D 28 RHRSW* RHRSW* RHRSW* RHRSW*

RHRSW pumps assigned for EECW automatic start

b. If normal voltage is NOT available: (DGVA)

(1) After 5-second time delay, all4kV Shutdown Board loads except 4160/480V transformer breakers are automatically tripped.

(2) Diesel generator output breaker closes when diesel is at speed.

(3) Loads sequence as indicated below Time After Accident SID Board SID Board SID Board SID Board A B C D 0 RHR A RHR C RHR B RHR D 7 CSA CS C CS B CS D 14 RHRSW* RHRSW* RHRSW* RHRSW*

RHRSW pumps assigned for EECW automatic start

c. Certain 480V loads are shed whenever an accident signal is received in conjunction with the diesel generator tied to the board. (see OPL 171.072)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 300000K4.02 Tier # 2 Knowledge of Instrument Air system design feature(s) and/or Group # 1 interlock(s) which provide for the following: Cross-over to other air systems. KIA # 300000K4.02

\NOTE: Instrument Air at BFN is referred to as Control Air.\

Importance Rating 3.0 3.0 I Proposed Question: RO # 25 Given the following plant conditions:

The Unit 3 Control Air system is aligned with the "G" Air compressor running and loaded.

Subsequently, the Unit 3 Control Air system pressure falls to approximately 60 psig due to a rupture.

Which ONE of the following describes the final plant configuration pertaining to the Control Air system.

A. All Unit's Control Air system pressure will drop and a manual scram will be required for all units.

B. The effects on Unit 1 and 2 will not be as severe as on Unit 3 because of the automatic unit separation capability.

C. The effect on Unit 3 will not be experienced on Unit 1 and 2 because of the alignment of the closed manual header isolation valve between Unit 2 and 3.

D. All Unit's Control Air system pressures will drop, all Control Air Compressors will full load, and the Service Air Compressor will unload due to the Surge Condition experienced on Unit 3.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: B Explanation: Incorrect. 2-PCV-032-3901 is installed on the main control air header

a. between Unit 2 and Unit 3 and l-PCV-032-3901 is installed on the main control air header between Unit 2 and Unit 1. Each valve automatically close when Control Air header pressure on either side of the valve drops below 65 psig. This action prevents a control air failure on anyone Unit from resulting in a multi-unit scram. Illustration 3 provides more information on this feature.

b. Correct answer.

c. Incorrect. The valve between Unit 2 and 3 is not a manual isolation valve.

d. Incorrect. Partially true. All three unit's control air pressure will begin to drop, but will stabilize when the PCV between Unit 2 and Unit 3 closes.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank # OPL171.054.10 minor format changes Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

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ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Original question OPL171.054.10:

The Unit 3 Control Air system is aligned with the "G" Air compressor running and loaded.

Subsequently, the Unit 3 Control Air system pressure falls to approximately 60 psig due to a rupture.

Determine which one of the following describes the final plant configuration pertaining to the Control Air system.

A. Both Units Control Air system pressure will drop and a manual scram will be required for both units.

B. The effects on Unit 2 will not be as severe as on Unit 3 because of the automatic unit separation capability.

C. The effect on Unit 3 will not be experienced on Unit 2 because of the alignment of the closed manual header isolation valve between Unit 2 and 3.

D. Both Units Control Air system pressures will drop,all Control Air Compressors will full load, and the Service Air Compressor will unload due to the Surge Condition experienced on Unit 3.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Control Air System 0.:01-32 Unit 0 Rev. 0114 Page 12 of 105 3.0 PRECAUTIONS AND UMITATIONS(continued)

O. Control Air Compressor A, S, C, 0 Air ShutoffVal. . . es To Unload,ers, O-SHV-032-4004A (S,C,D) and O-SHV-032-4005A (B,C,D) can be used as an alternate method for manually loading Control Air Compressors A, S, C, D.

Closing anyone of these valves wi,1I cause the applcicable compressor to Half Load. Closing both vaives on the compressor wH/, result in that compressor reaching Full Load. Section 8.4, Alternate Method for Manually Loading Control Air Compressors A, S, C, 0, provides instruction for utilizing these valves to load A,. S, C, D cornpressors.

P. Air flow should never be established through an air dryer unless power i:s supplied to the dryer.

Q. Section 6.5, Moisture Trap blowdown, is performed once per shift.

R. [QAlCj Header isolation vaives 1-32-586 and 1-32-2318 should be dosed during multi-unit operation so that a Control Air failure in Unit 1 will NOT result in the possibility of a sc.ram of Unit 2. !CA.QR BFPl?tGOO31.

S. 2-PCV-032-3901 is installed on the main control air header bet\Neen Unit 2 and

( Unit 3 and 1-PCV-032-3901 is installed on the main control air header between Unit 2 and Uni:t1. Each valve automaticaHy dose when Control Air header pressure on either side of the valve drops below 65 psig. This action prevents lit control air failure on anyone Unit from resulting in at multi-unit scram.

Hlustration 3 provides more information on this feature.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Control Air System 0-01-32 Unit 0 Rev. 0'114 Page 72 of 105 Illustration 3 (Page 1 of 2)

Control Air System U nit Separation Hgure 1 (Normal Alignment)

Telt CorJl *

.1IIiir.....oIII.~

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2-P CV-032-39 01 Figure 2 (Low Header Pressure Alignment)

Tesl t...3°m .

[........ ]

I RlV-03>-"""'~ UNIT Ie( 3

2. PCV..oS2.3Q01

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 400000Kl.03 Tier# 2 Knowledge of the physical connections and/or cause-effect 1 Group #

relationships between Component Cooling Water system and the following: Radiation monitoring system. KIA # 400000K1.03 Importance Rating 2.7 3.0 I Proposed Question: RO # 26 Unit 2 is at rated power with the following indications:

RECIRC PUMP MTR 'A' TEMP HIGH (9-4A W13).

RECIRC PUMP MTR 'B' TEMP HIGH (9-4B W13).

RBCCW EFFLUENT RADIATION HIGH (9-3A W17).

RBCCW SURGE TANK LEVEL HIGH (9-4C W6).

RX BLDG AREA RADIATION HIGH (9-3A W 22).

Recirc Pump Motors "2A" and '2B' Winding and Bearing Temperature Recorder (2-TR-68-84) are reading 170 of and 188 of respectively and rising.

RBCCW Pump Suction Header Temperature Indicator (2-TIS-70-3) is reading 94 of and rising.

RWCU NON-REGENERATIVE HX DISCH TEMP HIGH (9-4C W17).

Area Radiation Monitor RE-90-13A and RE-90-14A are in alarm reading 55 mr/hr and

( rising.

Which ONE of the following describes the actions that should be taken in accordance with plant procedures?

REFERENCE PROVIDED Enter 2-EOI-3, "Secondary Containment Control" and _ _ _ _ _ _ _ _ __

A. Trip and isolate '2B' Recirc Pump.

Enter 2-AOI-68-1A "Recirc Pump Trip/Core Flow Decrease OPRMs Operable."

B. Trip and isolate '2B' Recirc Pump.

Commence a normal shutdown in accordance with 2-GOI-100-12A, "Unit Shutdown."

C. Trip RWCU pumps and isolate the RWCU system.

Commence a normal shutdown in accordance with 2-GOI-100-12A, "Unit Shutdown."

D. Trip RWCU pumps and isolate the RWCU system.

Close RBCCW Sectionalizing Valve 2-FCV-70-48 in accordance with 2-AOI-70-1, "Loss of RBCCW."

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

( I Proposed Answer: D Explanation: a. Incorrect. Although 2B Recirc pump motor temperature is high, insufficient RBCCW cooling to the motor could be the cause. This can be verified by the temperature of the 2A Recirc pump. Although not as high as the 2B pump, it is still well above normal for this condition, which leads to the conclusion that the cause of the high temperature is common to BOTH Recirc pumps and NOT an individual pump issue. The procedure, 2-AOI-68-1A would be correct if the Recirc pump was the cause 0 the problem.

b. Incorrect. The reason above explains why the Recirc pump is not the problem. Commencing a plant shutdown is directed by 2-EOI-3, but only if the cause of the leak into secondary containment is NOT isolated. In this case, there is no information given that the corrective actions taken will not be successful.

c. Incorrect. The action to trip and isolate RWCU is correct for the given indications. The action to commence a shutdown is incorrect for the reasohs given in (b) above.

d. Correct answer.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 2-EOI-3 flowchart, 2-ARP-9-4A & B (W13) (Attach if not previously provided) 2-ARP-9-3A (W17)

Proposed references to be provided to applicants during examination: 2-EOI-3 flowchart Question Source: Bank #

Modified Bank # RO 400000G2.4.31 Attached New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: This question was used on the last NRC exam but has been significantly modified. It has been changed from Unit 3 to Unit 2 and the conditions in the stem have changed to provide a different source of the leak. Therefore, the answer is now different from the previous version of this question. I did not take credit for a previous NRC exam question in calculating the total number of questions from the last exam. If this is inappropriate, even given the substantial changes to the question, please let me know and I'll develop a different question.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Original Question RO 400000G2.4.31:

Unit 3 is at rated power with the following indications:

RECIRC PUMP MTR 'B' TEMP HIGH (3-XA-9-4B, Window 13).

RBCCW EFFLUENT RADIATION HIGH (3-XA-9-3A, Window 17).

RBCCW SURGE TANK LEVEL HIGH (3-XA-9-4C, Window 6).

RX BLDG AREA RADIATION HIGH (3-XA-9-3A, Window 22).

Recirc Pump Motor '3B' Winding and Bearing Temperature Recorder (3-TR-68-84) is reading 1700 F and rising.

RBCCW Pump Suction Header Temperature Indicator (3-TIS-70-3) is reading 1040 F and rising.

RWCU NON-REGENERATIVE HX DISCH TEMP HIGH (3-XA-9-4C, Window 17).

Area Radiation Monitor RE-90-13A and RE-90-14A are in alarm reading 55 mr/hr and rising.

Which ONE of the following describes the actions that should be taken in accordance with plant procedures?

REFERENCE PROVIDED Enter 3-EOI-3, "Secondary Containment Control" and _ _ _ _ _ _ _ _ __

A. Trip and isolate '3B' Recirc Pump.

Enter 3-AOI-68-1A "Recirc Pump Trip/Core Flow Decrease OPRMs Operable."

B. Trip and isolate '3B' Recirc Pump.

Commence a normal shutdown in accordance with 3-GOI-100-12A, "Unit Shutdown."

. C. Trip RWCU pumps and isolate the RWCU system.

Commence a normal shutdown in accordance with 3-GOI-100-12A, "Unit Shutdown."

D. Trip RWCU pumps and isolate the RWCU system.

Close RBCCW Sectionalizing Valve 3-FCV-70-48 to isolate non-essential loads and maximize cooling to '3B' Recirc Pump.

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

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BFN Panel 9-4 .2-ARPc9-4A Unit 2 2-XA-55-4A Rev.U034

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SensoclTrip Point: Alarm is from 2-TR-63-58, Panel 2-9-21 RECIRC 2-TE-esc61A RECIRC PMP MTR :2A-TH'R BRG UPPER FACE {1QOOF)

PUMP MTR 2A 2-TE-6S-61C RECIRC PMP MTR 2A-THR BRG LOWER FACE (190"F)

TEMP HIGH 2-TE:..e8-t\1ERECIRC PM? MTR :2A-UPPER GUIDE B'<~G (100"f) 2-TA-68-58 2-TE-6S-61N REGliRC PMP MTR 2A-lOWER GUIDE BRG (100'F) 2-TE-6S-tl1G RECIRC PM? MTR :2A-MOTOR WiNDING A (25ljoF) 2-TE-6S-61J REG!RC PMP MTR 2A-~:)TOR WINDING B {255'".!f}

(Page 1 of 2) 2-TE:..e8-t\1L REGIRC PM? MTR :2A-MOTOR WiNDING C 1255"1")

2-TE-6S-tlH RECIRG PMP MTR 2A-SEAL NO.. :2 CAVJTY (NOTE 1) 2-TE:..e8-61 U REGIRC PMP MTR 2A-SEAL NO.1 C.t..V1TY (NOTE 1) 2-TE~54 RECIRC PMP MTR :2A-CLG WTR FROM SEAL Cill (14G"F) 2-TE-6S-57 RECIRC PPO'P MTR 2A-CLG WTR FROM BRG (14G"F)

NOTE 160"F with the original water seal (Model SU) installed or 1&l"F with the new water seal {Model N7500} lnstatled.

Sensor Temperarureelements are located on Recirculation pump motor. Elevation 563.'12, location: Unit 2 drywelL Probable A. Possible bearing failure.

Cause: B. Possible motor overload.

C Insufficient cooling wateL D. Possible seal faiilure.

E. High drywel! temperature.

Automatic None Actioo:

Operator A CHECK following on Panel 2-9-4:

Actioo.:

RBCC'N PUMP SUCTION HDR TEMP temperature indicating switch, 2-TIS-7Q*3 normal (summer 70-:95"F, winter SO-BO*F). o

RBCCW PRI CTMT OUTLET handslll'itch, 2-HS-70-47A (2-FCV-70-47) OPEN. o B. CHECK the temperature of the cooling water leavmglhe sea.land bearing rooters <140'F on RECIRC PMP MTR 2A'NINDING AND BRG TEMP ternperature recorder, 2-TR-S3-5B on Panel 2-9-21. 0 C. REDOCE Recire pump speed until the temperature drops below alarm setpoint 0 D. CONTACT System engineering to PERFORM a complete a.ssessment and mooitomg of aU seal oonditiOfls parncularly seal leakage, temperature, and pressure of ali stages for Recife Pump seal temperatures wn excess of 200'E 0 Continued on Neld Page

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

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BfN Panet 9-4XA-55-4B 2-ARP-9-4B Unit .2 Rev. 0034

JJaSJ~Jlor~

SensorlTnp Point:

RECiRC PUMP MTR 2B AI_arm is fmm 2- TR~68-84. Pane~ 2-9-21 TEMP HIGH 2-TE-68-nA RECIRC PMP MTR 2B-THiR aRG UPPER FACE {UlO"F}

2-TA-68-84 2-TE-68-73C REC!RC PMP MIR 2.S-THR BRG LOWER FACE (1 QO"'F) 2-TE-68-73E RECIRC PMP MTR 2B-UPPER GUIDE BRG (100"1'")

2-TE-68-73N RECIRC PMP MTR ::lB-LOWER GUIDE BRG (100"F)

(Page 1 of2}

2-TE-68-73G RECiRC PMP MTR 2B-MOTOR WINDING A (255"F) 2-TE-68-73J RECIRC PMP MTR 2B-MOTORW'lNDING 8 (255"1'")

HE-6B-73L RECIRC PMP MTR 2B-MOTOR WtNOiiNG C (255"1'")

2-TE-68-73T RECIRC PMP MTR 2B~SEAL NO.2 CAVITY {Note 1}

2-TE-68-73U RECIRC PMP MTR 2.S-SEAL NO. t CAVITY (Note 1}

2-TE-68~e7 RECIRC PMP MTR 2B-CLGWTR FROM SEAL CLB {140"F) 2-TE-68-70 RECIRC PMP MTR 2B-CLG WTR FROM BRG (140"F)

NOTE 160"F with the Ofigina! water sea! (Mode! SUl illStaUed or 100"F with the new water seal (Model N7S00) installed.

Sensor Temperature elefllenl.s are located on Redrculation pump motor, Elevation 563.12, location: Una 2 drywell.

Probable A.. Possib4e bearing failure.

Cause: B. Possib4e motor oveooad.

C. Insufficient cooling water.

D. Possible seal faiiure.

E. High dl)'Vlell temperature.

Automabc None Action:

operator A CHECK following 011 Panel 2c9-4:

Action:

RBCCW PUMP SUCTION KDR TEMP temperature indlcating switch, 2-TlS-70-3 oomlal (sumrner 70-95"F. 'iMnter 60-80°F). o

RBCCW PRI CTMT OUTLET handswitch, 2-HS-70-47A..

{2-FCV-70-47) OPEN. o B. CHEC K the temperature of the cooling water leavioglhe sea.1 and bearing coolers <: 14QoF on RECIRC PMP MTR 2B WINDING AND BRG TEMP temperature recorder, 2-TR-68-84 on Pane! 2 2 1 . 0 Continued on Next Page

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ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panel 9-3 Unit 2 2-XA-55-3A SensorfTrip Point:

RBCCW EFFLUENT RADIATION HI HI-HI HIGH RM-90-1311J (NOTE 2) (NOTE 2) 2-RA-90-131A NOTE: HI alarm from recorder. HI HI alarm from drawer SOUD MAGENTA 11'1 (Page 1 of2}

Sensor RE-9Il..U1A Reccw HX RxBldg, EI 593, R-2 R-LiNE Location:

Probable HX tube teak into RBCCW system_

Cause:

Automatic None Action:

Operator A. DETERMINE cause of alarm by observing. following:

Action: 1. ReCCWand RCW EFFLUENT RAmAllON recorder, 2-RR-90-131/132 Red pen 00 Panel 2-9'-2. o

2. ReCCW EFFLUENT OFFLINE RAO MONradiation mooii:or.

2-RM-90.. 131D on Panel 2-9.. 10_

R NOTIFY Chemistl)l to sample RBCCW for total gamma activity. 0 C_ DETERMINE if source of teak is RWCU Noo-regeneratlve, Fuel Pool Cooling, Reactor Water Sample or RWCU Recife Pump A or B Seal Water heat exchanger(s). 0 0 .. [NE~JCl CHECK the following for indica.tion of Reactor Recirculation Pump Sea! Heat. Exchanger leak:

1. LOWERING Pressure in react;:)f Recircuiation pump 2A(2S).

No.1 or:2 SEAl, 2-PI-68-64A or 2-PI-68-63A (2-PI-68-76A or 2-PI-68-75A) on Pane! 2"9-4.. 0

2. Risil1g Temperature on ClG WTRFROM SEAL CLG TE"68-54, 00 RECIRC PMP flATR 2A. WINDING AND BRG TEMP temperature recorder, 2-TR-68-58, on Panel 2-9-21. 0

3. Hisil'lg Temperature on CLG WTR FROM SEAL CLG TE-68-67, 00 RECIRC PMP flATR 26 WINDING ANDBRG TEMP temperature recorder, 2-TR-68-M, on Pane~ 2-9-21. 0 Corrtinued on Next Page

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ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BfN Panel 9-3 2-ARP-9-3A Unit2 2-XA-55-3A Rev.Oi036

. . ,Page?6of@

RBCCW EFflUENTRADIATKlN HIGH 2-RA-90-131A, Window 17 (Page2of2)

Operator Action: (Continued)

E. If it is determlned the source of leakage is from Reactor Recire Pump A(B), THEN, at the discretion anne Shffi Manager,

1. ISOLATE Reactor Recirculatioo Loop A(B). REfER TO

~~. 0 NOTE Coolido'Ml is required to prevent hangers or shock suppressers from exceeding their maximum travei range

2. WHEN primary system pressure is fess IIlan 125 PS19, THEN, at the discretion of the ShIft Manager, lit. ISOLATE RBCCW Systenl to preclude damage to RBCCW piping. flEN 6'9-re4. GE SIL~59i o HOTE2 Chemlab should be contacted for current setpolnts per O--Tf.-4S.

References:

2-45E620-3 2-47E610-90-3 GE 2~729E814-3

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

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, REFERENCE MATERIAL Provided to CANDIDATE

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,-- - -- ,.,.*----1 UNIT2 EOI-3 IQJ~E-=5:i.-';';:; I BROWNS FERRY NUCLEAR PLANT

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 201001A2.04 Tier# 2 Ability to (a) predict the impacts of the following on the CRD system 2 Group #

and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: KIA # 201001A2.04 Scram conditions.

Importance Rating 3.8 3.9 I Proposed Question: RO # 27 A scram ha~ just occurred on Unit 2 with the following conditions:

All blue Scram lights on the Full Core Display are energized.

Three control rods indicate 48 in red and all remaining control rods indicate (- -) in green on the Full Core Display.

All Accumulator Trouble lights and Rod Drift lights on the Full Core Display are energized.

All IRMs are inserted and on Range 3 and lowering.

Which ONE of the following describes the current status of the CRD hydraulic system and the action(s) necessary to insert the remaining control rods?

CRD system flow is being directed to the charging water header and (1)

In order to insert the remaining control rods, the OATC must first (2)

REFERENCE PROVIDED (1) (2)

A. Scram Discharge Volume close 2-85-586 in accordance with 2-EOI-Appendix lD.

B. Scram Discharge Volume reset the RPS scram signal in accordance with 2-AOI-l00-1.

c. cooling water header close 2-85-586 in accordance with 2-EOI-Appendix lD.

D. cooling water header reset the RPS scram signal in accordance with 2-AOI-l00-l.

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

( I Proposed Answer: B Explanation: a. Part (1) is correct. Part (2) is incorrect. Since the IRMs are below range 7, all actions to insert control rods are done using abnormal procedures, not emergency procedures.

b. Correct answer.

c. Part (1) is incorrect. Following a scram, the flow is prevented from going to the cooling water header because the CRD flow control valve closes. Part (2) is incorrect. Since the IRMs are below range 7, all actions to insert control rods are done using abnormal procedures, not emergency procedures.

d. Part (1) is incorrect as stated in (c) above. Part (2) is correct. Resetting the scram will allow the flow control valve to re-open and establish the required drive water pressure to allow control rod insertion. This is accomplished using 2-AOI-100-1.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 2-AOI-1 00-1, (Attach if not previously provided)

~---------

2-EOI-1 Flowchart path RC/Q Proposed references to be provided to applicants during examination: 2-EOI-1 flowchart Question Source: Bank#

Modified Bank # (Note changes or attach parent)

New 09/15/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

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

RCIQ*1 WHILE EXECUTING THE FOLLOWING STEPS:

IF THEN 0---- THE ReAC'l"OR WILt REMAINSUBGRITICAl WiTHOUT aORON UNDER ALL CONDITIONS (SEIENOTE)

'1. STOP BORON INJ UNLESS REQUIRED BY OTHER .PR:Of'::Et'tURES

2. Il!lUT Re/Q AND ENTER AOI"1~1. REACl'ORSCAAM THE R:>:: IS SUBCRITIGAL EXIT RCIO .AND AND IENTEftAOI~H:>O*i

REACTOR SCRAM

!:!QBORON~IAS BEEN VNJECTED

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BAt Reactor Scram 2-A0I-16o-1 Unit 2 Rev.6OS7 Page 10 of 60 4.2 Subsequent Actions {conooued}

NOTE Step 4.2[8.1] may require support from off-site organizations aoo an extended period may e~3pse before resutts are obtajned.

[8.11 IF aN rods are NOT inserted to Posmon 02 or beyond, THEN DIRECT Reactor Engineer to commeoce determination that reactor will remain subclitical under all conditions wittlout boron. 0

[9] [H'OIq IF any control rod fails to fulty insert and it is required to R8-'scram, THEN PERFORM the following, as required: {lNPO SOEA 8IIl-OO!ij

[9.1} RESET scram per Steps 4.2[22]lhru 4.2[22.10], o

[9.2] VERIFY WEST and EAST CRD DISCH VOl WTR LVl HIGH HALF SCRAM annunciators (2-XA-55-4A-1 and 4A-29) reset. o

[9,3] INITIATE manual scram. o

[9.4] REPEAT Step 4.2(9), as necessary, as long as rod mooon is observed. o

[10J [H'OIq IF any control rod fails to fulty insert and it is required to Drive Control Rods, THEN REFER TO 2-01-85. {lNPO SOER e&II06i o

[11] IF ANY PCIS isolatioo signal is received, THEN VERIFY PCIS isoI1ations using ANY of the following::

Containment Isolation Status System 00 Panel 2-9-4 o

PCIS Mimic and individilJal control switch indications o

ICS o

2-01-64 o

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.00S page 48:

1. Scram

a. Following a scram, but before the SDV is full, the control rod will be in an over travel-in position since there will still be a large differential pressure across the piston.

b. Therefore, the green (full in) light on Panel 9-5 will be on but there will be no rod position readout displayed.

c. After the SDV is full, there will be no differential pressure across the piston, and rod will settle into the 00 position.

Excerpt from OPL171.00S page 18:

(a) Runout protection

i. During a scram, the HCU accumulators will be fully discharged.

ii. The CRD pump will try to recharge all the accumulators at once. Flow through the charging header will cause the flow control valves to close.

REFERENCE MATERIAL Provided to CANDIDATE c

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

( Examination Outline Cross-reference: Level SRO 201003A4.02 Tier# 2 Ability to manually operate and/or monitor in the control room: 2 Group #

Control Rod and Drive Mechanism, CRD mechanism position.

KIA # 201003A4.02 Importance Rating 3.5 3.5 I Proposed Question: RO # 28 Which ONE of the following combinations of the alarms and indications numbered below characterizes the possibility of an uncoupled control rod?

1. "Red" (- -) on the Full Core Display.

2. CONTROL ROD OVERTRAVEL (9-5A W14).

3. "Red" (48) on Four Rod Display.

4. CONTROL ROD DRIFT (9-5A W28)

5. Blank rod position indication on Four Rod Display.

A. 1, 2,4 B. 2, 4, 5 C. 2,3,4 D. 1, 2, 5

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: B Explanation: a. Incorrect. The full core display will indicate blank just like the four rod display.

b. Correct answer.

c. Incorrect. The four rod display indicates blank.

d. Incorrect. The full core display will indicate blank just like the four rod display.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 1-ARP-9-5A W14 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank # OPL 171.006.12 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panel 9-5 1-ARP-9-5A Unit 1 1-XA-55-5A Rev. 0012 SeflSodTrip Point:

CONTROL ROO Relay 3A~K3 Switch 50 Receive alarm when a control rod has been OVERTRAVEl in RPIS probe withdrawn past full out position.

(Page 1 of 1)

Sensor Panel 1*9-28 Location: Bev.593' COlltroi bldg.

Aux. Inst Room Probable A. Roo is UI'lCOOpied.

Cause: B. M:atfunction ofsensoF.

C. Fuse 1-FU1-85-SXA failure.

Automatic The digital read out arid background light wil go rut.

Action:

Operator A. VAUDATE afaml as follows:

Actj,on: 1. Full core d~y will haw 00 digital readout o

2. Background Iiglht extinguiShed. o

3. Rod DRIFT light on. o B. IF alarm is valid, THEN REfER TO '1-A0i-8S-2. 0 C. NOTIFY Reador Engineer. 0 D. REFER TO Tech Spec 3.1.3, 3.10.3.5, 3.3.2.1, Table 3J4.2.F and TRM Tab4e 3.. 3.5-1. 0

References:

1-45E620~6-1 1-730E321.10

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Uncoupled Control Rod 1-AOI-85-2 Unit 1 Rev. MOO Paoe4of8 1.0 PURPOSE This aboormal operating instruction provides symptoms, automatic actions and operator actions for an oocoupded control rod .

.2.0 SYMPTOMS NOTE If a control rod is un!OOUpled and being 'Withdrawn to any PQsition other than position 48. the Rod Position Information System widl display normal oonb"ol rod movement Power must be monftored to determine if the controil roo is faltowing its associated drive.

Nuclear instrumelltidioo does NOT resPQnd tQ control rod movement

CONTROL ROO OVERTRAVEl annunciator (1-XA-55-5A, Window 14) in alarm.

Digital display and recl badklighting for the unOOlJpled control rocl on the full core display is extinguished.

3.0' AUTOMATIC ACTIONS None

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

(

SFN Panel 9-5 1-ARP-9-SA Unit 1 1-XA-55-5A Rev. 0012 Sef1sorfTrip Point:

CONTROL ROO RetayslA -1<37A,B,C,& D Picked up by same relays DRIFT for each indiyfdUa~ rod

-K37 A has 50 rods drift.

-K37B has 43 rods

-K37C has 44 rods

-K37D has 48 rods (Page 1 of 1)

Sensor 1-PNLA-009-0028 Location: Bev.593' Aux. (nst Room Control Bay 1C Probable A. Rod NOT selected, NOT being driVef1, and leaves eVef1 number reed Cause: switch (drift).

1. Intemalleuage of scram valves.

2. Excessive cooling waler pressure.

B. Rod settie timer timed 001, and rod passes odd number reed SIIWdl.

1. Drift-in foIfowing insert :signal;

a. Failure of insert valve to close.

2. Drift-ouf following wHhdraw signal; 8.. Failure of wiil:hdraw valve to close.

b. Failure of rod to latch aftef" withdraw signal.

C. Defective RiPIS alarm cirelli.

D. Readm" SCRAM or Rod SCRAM.

E. loss of 'I or more Rod Posifioo Indications.

Automatic None Action:

Operat<< A. Determine which oontrol rod is driftiing from Full Core otspIay. o Action: B. IF control rod is drifting in, THEN REFER TO 1-AOI-85-5.. o C. IF control rod is drifting out, THEN REFER TO 1-A0I-85-6. o D. REFER TO Tech Spec Seclloo 3.1.3, 3.10.8. o Technical Specifications 1-45E620-6-1 1-730E321-16

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 201006KS.l0 Tier # 2 Knowledge of the operational implications of the following concepts 2 Group #

as they apply to the RWM: Withdraw error.

KIA # 201006K5.10 Importance Rating 3.2 3.3 I Proposed Question: RO # 29 Unit 2 is starting up with the following plant conditions:

Total steam flow is at 20 0/0 of rated flow.

Rod Worth Minimizer (RWM) Group 22 is latched with limits from 00-04. (a double asterisk on the pull sheet applies for this group)

The OATC selects the first rod in Group 22 and takes the ROD MOVEMENT CONTROL switch to the ROD OUT NOTCH position.

The selected rod triple notches to position 06.

Which ONE of the following describes the RWM response to this condition and the reason for that response?

(

The RWM ROD BLOCK (9-5B W35) _----4(.=,1)_-. The reason for that response is (2)

(1) (2)

A. will alarm a Withdraw Block is applied due to power below the Low Power Set Point (LPSP).

B. will alarm a Withdraw Block is applied since the single notch restraint limit for Group 22 control rods has been exceeded.

c. will NOT alarm Notch 06 is the Alternate Withdraw Limit for that control rod and will NOT result in a Withdraw Error.

D. will NOT alarm a Withdraw Block is NOT applied with power above the Low Power Set Point (LPSP).

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

( I Proposed Answer: A Explanation: a. Correct answer.

b. Part (1) is correct. On Unit-2, the LPSP is 24%. On Units 1 & 3 the LPSP is 16%. This is due to EPU differences. Part (2) is incorrect. Although the Group 22 rods are "single notch restricted" by the double asterisk on the pull sheet, this restriction is administrative and does NOT initiate any control rod blocks if exceeded. It is merely an added control to help enforce the Reduced Notch Worth Procedure (RNWP) restraints.

Part (1) is incorrect. This would be true for either Unit 1 or Unit 3, but not

c. for Unit 2. Part (2) is incorrect. For group limits, RWM recognizes the Nominal Limits only. The Nominal Limit is the insert or withdraw limit for the group assigned by RWM. The Alternate Limit is no longer recognized by the RWM as an Acceptable Group Limit..

d. Part (1) and Part (2) are incorrect. This would be the correct answer for either Unit-lor Unit 3.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 2-01-85, 1-01-85,3-01-85 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/15/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet BRt ContrOl Rod Drive System 2~m*85 Unit 2 Rev. 0104 Paoe 17 of 181 3.2 ReletOf'Manual Control System {confinuedt J. Whenever !here is fuel in !he vessel, a peer check verification is required on alii cootrol rod selections, identtflCaOOn offinal position, and verification of rlnal position following movement, except as governed bylhe AOls andJor EOls.

Peer dleck verification is required to be performed by an SRO, RO, STA, or Reactor Engineer.

K. While dri'iling a Cootrof Roo, if at any time a control rod moves unexpectedly more than two notdles from its intended position, !he control rod should be cormnl1OOsly inserted using Ule "EMERGENCY IN" switd'l. Notify the Control Room Unit SUpervisor, Reactor Engineer, and obtain the Shift Manager's permission pnorw resuming rod movement. If rod insertion to Position 00 is required and coreUlermad power is s; 10%, en1Jyillto LCO 3.1.6 may be required.

3.3 Rod Worth Minimizer (RWM)

A. The RWM System Rod Test/TolJiCh screen flJlfldioo allows any one rod to be selec1ed and moved to .any position only if all rnher control rods are fully inserted . To get out of the rod test, the pushbutton needs to be depressed again (orherwise any single roo in any group can be selected and withdrawn).

B. ~ When !he RWM is bypassed, a second lcensed operator, or other qualified member of the technical staff, is required to venfy!he Control Rod Sequence is followed. {lNPO SOER.a4-002]

C. 2-8R-3.3.2.1.7 is used to document inc:lependent verification of the RWM whenever the reactor is in startup or run, beliow 10% power.

D. ~ Activities that can directly affect core reactilVity are of a critical nature.

Strict procedural compliance and conservative actions are required to be followed. IINPO SOER.a4-OO2]

E For RWM to enforce, Totail Feedwater Flow or Total Steam Flow is required to be -< 24%. To take RWM out of seMce automatically, Low Power Set Point (LPSP), Total Steam Row AND Total Feedwater Flow is reQiUired to be > 24%.

The loW Power Alarm Point (lPAP) for the RWM is 27%, as sensed by Toml Steam Flow. When !he RWM is operating in the transition zone, between the LPS? (24%) and Ule lPAP (27%), no rod blod<s are aPlPliedas a resiJJtof insert or wittldraw errors, but ttle RWM wig continue to provide alarm indications and error displays.

The mooitolirlg functions ofUle RWM are automatically bypassed at power levels above the LP AP .

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

(

BFN Control Rod Drive System 2-01-85 Unit.2 Rev.U104 Page 18 of 181 3.3 Rod Worth Minimizer (RWM) (eontinued)

F. All ttle RWM blocks are applied in the event ora system hardwSire or sofiware fSiiltlif. when power is Mlow the lPAP. At .any Rx power, when a ross of les 2A occurs., a select block occurs due to the loss of power and cannot be bypassed using the RWM Bypass key.

G. An insert error occurs if:

1. A rod in the currently Latched group is inserted past the insert limjt for this group.

2. A rod in a group lower than the one that is presently latched is inserted P.!lst the withdraw limit for the lower group.

H. A withdraw error occurs if:

1. A rod in the currently latched group is wittldrawn past ttle withdraw 'limit for the group.

2. A rod in a group lower than the one currently latched is withdrawn past ttle withdraw limit for its group.

3. A rod in a group higher than the one currently latched is withdrawn past the insert limit for its group.

1. A select error occurs it

1. With ttle reactor operaling below 100 lPAP, a roo other than one contained in the currently latched group is selected, unless conditions for latching up or diownare met.

2. With a rod block applied, any rod other than an error roo is selected.

3. When operating in the Sequence Control Mode, a rod is s~jpped.

J. An insert !Jiock occurs if:

1. With two inserterrorsexlsting, a rod is moved to cause a third insert error.

2. A withdraw error has been made, a withdraw block applied, and a rod other than the withdraw error rod is seiected.

1. A withdraw error is made.

2. With thre1l insert. errors existing and an insert hlock present, a rod other than ooe of the insert errors is seleded.

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet SFN Control Rod Drive System 2-01-85 Unit 2 Rev. 0104 Page 19 of 181 3.3 Rod Worth Minimf:zef'{RWM) (continued)

L A select bleck ocrurs if:

1. The RWM Bypass SWitch is in normal aoolhe RWM program is NOT running; i.e., following return to normal from bypass and the program has NOT been initialized.

2. The RWM Bypass Switch is in normal aoo the program stops due to software elrof.

M. For group limits only, RWM recognizes the NiOIminal Limits only. The Nominal limn is the insert or withdraw limit for the group assigned, by RWM. The Alternate Umit is no longer recognized by the RWM as an Acceptable GroopUmit N. Dumg RWM latching, the latched group will be the highest numbered group with 2 or less insert errors and having at least 1 rod withdrawn past its insert limns. WHh Sequence Control ON, latching occurs as follows. (Namally, startups are perfOrmed with Sequence Control ON).

1. RWM wiN latch dOwn when all rods in the presently 1a1dled group haVe been inserted to the group insert limi and a rod inlhe next lower group is selected.

2. RWM wil latch up when a rod within the next higher group is selected, provided that 00 more than two insert errors result.

Wlih Sequence Control OFF, latctung occurs as follows:

3. For non-repeating groups, latching occurs as described above.

4. For repeating groups, latching occurs to the next setup or set down based on rod movement as opposed to rod selection.

O. Latching occurs at

1. System initialization.

2. FollOWing a "System DiagnostJic" request.

3. When operator demands entry or termination of "Rod Test."

4. When power drops below LPAP.

5. When power drops below lPSP.

Ii Every five secooos in the transition zone.

7. Following any full control rod scan When power is below lPAP.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet 8Ft<< Control. Rod Drive System 1-01-85 Unit 1 Rev.IlQ05 Page 17 of 179 3.3 Rod Worth Minimizer (RWM)

A. The RWM system Rod TestITouchs.creenfunction allows anyone roo to be selected and moved to any posnion only if all other control rods are ful'y inserted. To get out of the rod test, the pushbutton needs to be depressed again (otherwise any single rod in any group can be selected and withdrawn).

R iNiERlC] When the RWM is bypassed, a second licensed operator or other qualified member of the technical staff is reQUired to verify the Control Rod Sequence is followed. [iNPO SOER-!l&-OC<2.J (Not required with no fuel in RPV)

C. 1-SR-3.3..2.t.7is used to document lndepelldentverificationofthe RWM whenever the reactor is in STARTUP or RUN, below 10% power.

D. !NiERlC] Activities that can directly affect core reactivity are of acritica! natUlfeand strict procedural compliance, along with conservative actions, must be followed*.

!1NPO SOER-8<l-002!!

E. For RWM to enforce, Total FeedwaterFlow or Total steam Flow must be <16%.

To take R'WM out of service automatically, Low Power Set Point (LPSP), Total Steam Flow and Total Feedwater Flow must be :>16%. The Low Power Alarm Point (LPAP) for the RWM Is (21%) as sensed by Total Steam Flow. When the RWM is operatlng !nthe transruon zooe, between the lPSP (16%) and the LPAP (21%), no rod blocks will be applied as a result of insert or withdraw errors, but the RWM will continue to provide alarm indications and error displays.

F. The rnoftitoring functions of the RWM are automatically bypassed at power

'levels above the LPAP.

G. AU the RWM blocks wil! be applied in the event of a system hardware Of software failure when power is below the LPAP. At any Reactor power, when a loss of ICS 1A occurs, a select bJock will occur due to the ioss of power and cannot be bypassed using the RWM Bypass key.

H. An insert error occurs if:

1. A rod in the currently latched group is inserted past the insert limit for this group, OR

2. A rod in a group lower than the one that is presently latched is inserted past the withdraw limit for the lower group.

ES-401 Sample Written Examination Form ES-401-S Question Worksheet Examination Outline Cross-reference: Level SRO 202002A4.0S Tier# 2 Ability to manually operate and/or monitor in the control room: 2 Group #

Recirculation Flow Control, Reactor level.

KIA # 202002A4.05 Importance Rating 3.4 3.4 I Proposed Question: RO # 30 Unit 2 is at 100% rated power with the following plant conditions:

2C Reactor Feedwater Pump (RFP) tripped due to thrust bearing wear.

RPV level lowered to (+) 24 inches before recovering to normal.

Which ONE of the following describes the Recirculation System response and the reason for that response?

The Reactor Recirculation Pumps (RRP) will _ _ _ _(,..=l:.+-)_ _ _ _

The reason for that response is that (2)

(1) (2)

A. runback to 28% flow total RFP flow is < 19% AND one RRP discharge valve is <90% open.

B. runback to 28% flow total RFP flow is < 19% OR one RRP discharge valve is <90% open.

c. runback to 75% flow individual RFP flow is <19% AND RPV level is

<27 inches.

D. run back to 75% flow individual RFP flow is <19% OR RPV level is

<27 inches.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: C Explanation: a. Part (1) is incorrect. A single RFP trip from rated conditions will not result in a scram, therefore total RFP flow will not lower to <19%. Part (2) is incorrect. Even if total RFP flow dropped to <19%, either condition is all that is necessary to initiate the runback, therefore the "AND" makes this response incorrect.

b. Part (1) is incorrect as stated in (a) above. Part (2) would be the correct reason if Part (1) were correct for these conditions, but it is not.

c. Correct answer.

d. Part (1) is correct. Part (2) is incorrect. Both low RFP flow AND low RPV level are required to initiate the 75% runback.

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 2-01-68 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/15/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

(

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

(

BFN ReactOf" Recirculation System 2-01-68 Unit 2 Rev. 0124 Paue 14 of 166 3.0 PRECAUTIONS AND LIMITATIONS (continued)

M. Recire pump controller limits are asfoClows:

1. When any individual RFP flow is <19% and reador water level is below 21 inches, Of if Reactor Scram occurs, speed limit is set to 15%

(-1130 RPM speed); and if speed is greater than 75% (-1130 RPM speed), recire speed will run back to 15% (,...,1130 RPM speed).

2. When total reed water flow is < 19% (15 sec CAl or recife pump discharge valve is ~ss than 90% open, speed limit is set to 28% (--480 RPM speed);

and if speed is greater than 28% (-480 RPM speed), recire Sjpeed will run back to 28%(-480 RPM speed).

N_ Observe the following recire pump seal limifalions:

1. The RBCCW System and eRD System provide cooing water to tile recife pump seals. As long as one of Ihese systems is suppomng recire pump seals, the seals should remain functional and undamaged. If both systems fail to support recirc pump seals at oormal operating conditions, recire pump seals willi begin to over heat in approximately seven minutes.

(Refer to 2-01-70 for RBCCW TCV adjustments).

2. When reeire pump seal temperatures are in excess of 200°F a com~te assessment and monitoring of all seal conditions should be made; particularly seal leakage, temperature, and pressure of all stages.

3. {NERICI For Model SU seals only, recire pump seat life may be extended by minimizing operation of the reeire pumps at suction pressures below 300 psi, and minimiizing recire pump starts below 300 psi and ensuring recire pumps are NOT operated wHh air inlhe seal cavities bylhoroughly venting pump seals. BFN Unit :2 currently has 1141500 Seals installed in both recirculation pumps. [GE SIL-203I

4. 1ll'CJ Following recirculation pump seal maintenance, reciradation pump

'tents and drains are to remajnopen until seal purge is established and the seals are ftushed and vented. This prevents small, parttdes from entering the pump cavity and seals due to flOOding without seal purge established.

(BII'PER951~

5. Seal purge ailignment is to be secured (isaated) anytime drains backup and water is introduced onto the floor, and Radiation Protection ootirred.IJPER 1171121 O. Aduation of 2-FSV-43-10 could cause erratic readings on Jet Pump No. 1 Flow Indication, 2-FI-68-15, if Reactor Recirculation System is in operation.

P. Keying a Radio whjle 'the recire drive control cabinet door is open has caused recire drive trips.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 226001K2.02 Tier # 2 Knowledge of electrical power supplies to the following: RHR/LPCI: 2 Group #

CTMT Spray Mode, Pumps.

KIA # 226001 K2.02 Importance Rating 2.9 2.9 I Proposed Question: RO # 31 A LOCA has occurred on Unit 2 with the following plant conditions:

RPV level is (-) 50 inches being restored with HPCI.

RPV pressure is 700 psig and steady.

Orywell pressure is 14 psig and lowering.

RHR Loops I and II are operating in Orywell Spray mode in accordance with 2-EOI Appendix 178, "RHR SYSTEM OPERATION ORYWELL SPRAYS."

Unit 1 RPV level has just dropped below (-) 122 inches.

Which ONE of the following describes the Unit 2 RHR system response and the actions required to restore Orywell Sprays on Unit 2?

Unit 2 RHR pumps (1) will trip. RHR pumps _ _ _~(2=..,),--_ _ to re-establish Orywell Sprays on Unit 2.

(1) (2)

A. 2A and 2C 28 and 20 will continue to run and 2A and 2C can be restarted in 60 seconds

8. 2A and 2C 28 and 20 will continue to run and 2A and 2C can NOT be restarted until the Unit 1 CAS clears

c. 2A, 28, 2C & 20 28 and 20 can be restarted in 60 seconds. 2A and 2C can NOT be restarted until the Unit 1 CAS clears O. 2A, 28, 2C & 20 28 and 20 can be restarted in 60 seconds. 2A and 2C can ALSO be restarted in 60 seconds

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: D Explanation: a. Part (1) is incorrect. Although these are the NON-PREFERRED pumps on Unit 2, all RHR pumps will trip since Unit 2 does not currently have an accident signal. Part (2) is incorrect. .Unit 2 does not have an accident signal so 2B and 2C RHR pumps will trip.

b. Part (1) is incorrect as stated in (a) above. Part (2) is incorrect. This would be correct if Unit 2 also had an accident signal at the time of the Unit 1 accident signal.

c. Part (1) is correct. Part (2) is incorrect. All RHR pumps on the non-accident unit can be manually restarted after 60 seconds.

d. Correct answer.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): OPL 171.044 pages 50 and 51 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/15/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.044 pages 50 and 51:

a. Accident Signal (1) LOCA signals are divided into two separate signals, one referred to as a Pre Accident Signal (PAS) and the other referred to as a Common Accident Signal (CAS) .

PAS

-122" Rx water level (Levell)

OR 2.45 psig DW pressure

CAS

-122" Rx water level (Levell)

OR 2.45 psig DW pressure AND <450 psig Rx pressure (2) If a unit receives an accident signal, then all its respective RHR and Core Spray pumps will sequence on based upon power source to the SD Boards.

(3) All RHR and Core Spray pumps on the non-affected unit will trip (if running) and will be blocked from manual starting for 60 seconds.

(4) After 60 seconds all RHR pumps on the non-affected unit may be manually started.

(5) The non-preferred pumps on the non-affected unit are also prevented from automatically starting until the affected unit's accident signal is clear.

(6) The preferred pumps on the non-affected unit are locked out from automatically starting until the affected unit accident signal is clear OR the non-affected unit receives an accident signal.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 233000A3.02 Tier # 2 Ability to monitor automatic operation of the Fuel Pool 2 Group #

Cooling/Cleanup system including: Pump tripes).

KIA # 233000A3.02 Importance Rating 2.6 2.6 I Proposed Question: RO # 32 Unit 2 is at 100% power with the following plant conditions:

SIX MONTHS following a refueling outage.

An electrical problem caused a trip of BOTH Fuel Pool Cooling Pumps approximately one hour ago.

Pump restart is delayed.

The Fuel Pool Temperature was 90 of when the pumps tripped.

Which ONE of the following is the calculated time to reach the fuel pool temperature Operating Limit and Technical Requirements Manual Limit?

The Operating Limit will be reached in (1) . The TRM limit will be reached in _~(2::.<.)__

REFERENCE PROVIDED (1 ) (2)

A. 16.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months . 32.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months .

B. 26.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months . 58.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

C. 26.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months . 64.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months .

D. 43.75. 52.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months .

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: B Explanation: Incorrect since this data is derived from information at 30 days after

a. the start of the outage.

(125 - 90)/2.1 =16.7 and 16.7 + 25/1.6 =32.3 Correct answer. Uses data for 180 days after the start of the outage.

b.

(125 - 90)/1.3 =26.9 and 26.9 + 25/0.8 = 58.2 Incorrect since this uses data from 180 days after the start of the

c. outage but the formula is performed such that the time to reach 1250 F is added to 25 0 F and then divided by 0.8.

(125 - 90)/1.3 = 26.9 and (26.9 + 25)/0.8 = 64.9 Incorrect since this used data for 180 days but the X and Y values are

d. switched.

(125-90)/0.8 =43.75 and 43.75 + 25/1.3 =52.9

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 2-AOI-78-1, Table 1 (Attach if not previously provided)

TRM section 3.9.2 Proposed references to be provided to applicants during examination: 2-AOI-78-1 Table 1 w/o example TRM section 3.9.2 Question Source: Bank # 233000A 1.07 minor changes to format Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Student Handout Question #32 NOTE This table is based on 2 year refuel cycle and a core "off-load" of.=:::: 300 fuel bundles.

Tab~e 1 spent Fuel Poot Heat-up Rate at normal Fuel Pool level.

COLUMNA COLUMNS COLUMNC COLUMN D Decay Time Days Rate to 125 Rate 125 to 150 Max Temp degrees lhr degrees I hr X y 0 2.7 22 180 @90 hrs 30 2.1 1.6 168@ 100 hrs 180 1.3 0.8 152@ 144 hrs 365 1.0 0.8 152@ 144hrs 730(2 yr cycle) 1.0 0.8 152@ 144hrs The information provided above is intended to cover all possib~ event scenarios.

The heat up rates given are for a starting SFSP temperature of 90 degrees and are for the first hour without any cooling.. They will provide a conservative estimate of the time to reach the given temperature.

4.2 Subsequent Actions (continued)

Use the following f01rmula to detemnine time to reach 125°F.

Use Column A (# O1f days since the beginning of the last refueling outage) and B to determine current heatup rate.

= TIME (rn hours) FOR FUEL POOl TO REACH 125"'F.

X (hearup rate determined from columns A and B eF I hr>>

Use the following formula to determine time to reach 150°F.

Use Column A (# of days since the beginning of the last refueling outage) and C to determine current heatup rate Time to reach 125l>F+ 25<>F =TIME (in hours) FOR FUEL POOL TOREACH 150"'F (cak:ulated above) Y (heatup rate determined from columns A and C (OF I hrH

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Spent Fuel Pool 'Water Temperature TR 3.9.2 TR 3.9 REFUELING OPERATfONS TR 3.92 Spent. Fuel Pool Water Temperature lCO 3.92 Fuel pool water temperature shall be s 150" F, APPlICABiUTY: Whenever irrad'iated fuel is in the fuel pool

NOTE-----------------------------------------

TRM lCO 3.0.3 is not applicable_

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A_ Fuel pool water A,1 Initiate actions to lower Immediately temperature> 15{l° F. the pool temperature, TECHNICAL SURVEiLLANCE REQUIREMENTS SURVEILLANCE FREQUENCY TSR 3.9.2.1 Whene ....er irradiated fuel is stored in the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months spent fuel pool, the temperature shan be measured and recorded daily.

BFN-UNIT2 3.9-3 TRM Revision 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 24S000K4.0S Tier # 2 Knowledge of Main Turbine Gen./Aux. system design feature(s) 2 Group #

and/or interlock(s) which provide for the following: Turbine protection. KIA # 245000K4.05 Importance Rating 2.9 3.0 I Proposed Question: RO # 33 I.

Which ONE of the following describes the reason for Extraction Non-Return Valve closure when a turbine trip signal is received?

A. Protect the heater tubes from excessive vibration when the steam flows back to the turbine.

B. Protect the turbine casing from over-pressurization when the steam flows back to the turbine.

C. Protect the moisture separators from over-pressurization on CIV closure.

D. Protect the turbine from overspeed when the steam flows back to the turbine.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: 0 Explanation: Incorrect. Heater tubes normally operate in a steam environment.

a.

Turbine casing pressure would lower to condenser vacuum when the

b. turbine tripped.

Steam flow to the moisture separators stops when the turbine trips.

c.

d. Correct answer.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 1-01-47 (Attach if not previously provided)

( --------------------------------

Proposed references to be provided to applicants during examination: None Question Source: Bank # 295005AK3.05 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.010 page 28:

E. Extraction Non-return Valves

1. Purpose To protect the turbine from an over-peed condition, which might occur when the turbine is tripped. A subsequent lowering of pressure in the turbine and heaters, due to vacuum in the condenser will cause hot water from the heater to flash to steam. The reverse steam flow back through the extraction steam piping to the Main Turbine could cause blade damage.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 268000K3.04 Tier # 2 Knowledge of the effect that a loss or malfunction of the Radwaste 2 Group #

system will have on the following: Drain Sumps.

KIA # 268000K3.04 Importance Rating 2.7 2.8 I Proposed Question: RO # 34 Given the following plant conditions:

The Radwaste system Waste Collector Pump has failed due to bearing damage.

Waste Collector Tank level is upscale at 38,000 gallons.

Which ONE of the following describes the action required to correct this problem and the effect on plant operation until it is completed?

To correct this problem perform the following: (1) . Until this action is complete, the (2) sump levels will continue to rise.

(1) (2)

A. Lineup to pump the Waste Collector Drywell, Reactor Building and Tank to the Waste Surge Tank. Turbine Building Floor Drain B. Cross-tie Waste Collector Pump Drywell, Reactor Building and suction to the Waste Surge Pump. Turbine Building Floor Drain C. Lineup to pump the Waste Collector Drywell, Reactor Building and Tank to the Waste Surge Tank. Turbine Building Equipment Drain D. Cross-tie Waste Collector Pump Drywell, Reactor Building and suction to the Waste Surge Pump. Turbine Building Equipment Drain

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: D Explanation: a. Part (1) is incorrect. The Waste Collector Pump is required to complete that lineup. Part (2) is incorrect. Equipment Drains are directed to the Waste Collector Tank, not Floor Drains.

b. Part (1) is correct. A cross-tie line is provided with a normally closed manual valve. Part (2) is incorrect as stated in (a) above.

c. Part (1) is incorrect as stated in (a) above. Part (2) is correct.

d. Correct answer.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): OPL171.084 page 18 and TP-3 (Attach if not previously provided)

(

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/15/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.084 page 18:

(2) Waste Collector pump and waste surge pump (a) 440 GPM centrifugal pumps (b) Draw suction on waste collector and surge tanks (c) Suction can be cross connected (d) Both discharge to waste filter

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ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 271000K6.02 Tier # 2 Knowledge of the effect that a loss or malfunction of the following 2 Group #

will have on the Off-Gas system: Process radiation monitoring system. KIA # 271000K6.02 Importance Rating 3.0 3.2 I Proposed Question: RO # 35 Unit 2 is in Mode 2 during a reactor startup at rv 2% power.

The Off-gas system is aligned as follows:

Adsorber control switch (HS-66-113) AUTO

"A" SJAE is in service.

Adsorber train "A" inlet (FCV 66-113A) CLOSED

Adsorber bypass valve (FCV 66-113B) OPEN

Off-gas'system isolation valve (FCV 66-28) OPEN Which ONE of the following describes the effect on the Off-gas system alignment should one of the OG Post-Treatment radiation monitors fail upscale?

The Adsorber INLET valve (66-113A) will (1) . The Adsorber BYPASS valve (66-113B) will (2) The Off-gas isolation valve (66-28) will (3)

(1) (2) (3)

A. open close remain open B. open close close C. remain closed remain open remain open D. remain closed remains open close

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: A Explanation: a. Correct answer.

b. Part (1) and (2) are correct. Part (3) is incorrect. The Off-gas isolation valve (66-28) requires BOTH channels to initiate an isolation.

c. Part (1) and (2) are incorrect with the Adsorber control switch (HS-66-113) in AUTO. Part (3) is correct.

d. Part (1) and (2) are incorrect with the Adsorber control switch (HS-66-113) in AUTO. Part (3) is incorrect. The Off-gas isolation valve (66-28) requires BOTH channels to initiate an isolation.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 2-AOI-66-2 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank # 271000K4.07 minor format changes Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Offgas Post-Treatment Radiation HI-HI- 2 -AOI--66-2 Unit.l HI Rev. 0020.

Page 5 of9 3.0 AUTOMATIC ACTIONS A. If the OFFGAS TREATMENT SELECT hanclswitch, 2-XS-66-113, Panel 9-53, is in AUTO when High radiation condmon exists it wi!l automaticaO,.* align, or ensure alignment of, the charcoal adsorbers to thelreatffient mode, Le., the charcoal inlet varve wil! receive an open signal and the charcoal bypass valve wit! receive a close signaL R OFFGAS SYSTEM ISOLATION VALVE, 2-FGV-66-28, aufomaUcallycloses on any combination of Off Gas Post Treatment Hi Hi Hi, downscale, or inop simultaneously in both channels of the O.G. post treatment radiation monitoring system after 5 seconds. 2-FGV-066-0028 wiM not perform ifs design function to automatically close, when it is mechanically restrained open due to plant conditions.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Original question 271000K4.07:

Unit 2 is in Mode 2 during a reactor startup at 2% power tv The Off-gas system is aligned as follows:

- Adsorber control switch (HS-66-113) in AUTO

- "A" SJAE in service

- Adsorber train "A" inlet (FCV 66-113A) closed

- Adsorber bypass valve (FCV 66-1136) open

- Off gas system isolation valve (FCV 66-28) open How will the Off-gas system alignment be affected should one of the OG Post-Treat radiation monitors fail high?

A. Adsorber inlet valve (66-113A) remains closed, Adsorber bypass valve (66-1136) remains open, Off-gas isolation valve (66-28) closes

6. Adsorber inlet valve (66-113A) opens, Adsorber bypass valve (66-1136) closes, Off-gas isolation valve (66-28) closes C. Adsorber inlet valve (66-113A) remains closed, Adsorber bypass valve (66-1136) remains open, Off-gas isolation valve (66-28) remains open D. Adsorber inlet valve (66-113A) opens, Adsorber bypass valve (66-1136) closes, Off-gas isolation valve (66-28) remains open

ES*401 Sample Written Examination Form ES*401*5 Question Worksheet Examination Outline Cross-reference: Level SRO 288000G2.2.44 Tier# 2 Ability to interpret control room indications to verify the status and 2 Group #

operation of a system, and understand how operator actions and directives affect plant and system conditions: Plant Ventilation. KIA # 288000G2.2.44 Importance Rating 4.2 4.4 I Proposed Question: RO # 36 Unit 2 is operating at 100% rated power during the summer with the following conditions:

Control Bay Chiller "A" breaker indicates closed and reading 32 amps on Panel 9-20.

Control Bay Chiller "B" breaker indicates closed and reading 0 amps on Panel 9-20.

Control Bay chilled water inlet temperature is indicating 42 OF on the ICS computer.

Which ONE of the following describes the effect of transferring the CONTROL PANEL MODE SELECT switch from "DIGITAL" to "ANALOG" on the CONTROL BAY CHILLER A LOCAL DISPLAY PANEL?

Control Bay Chiller "A" will _ _ _ _ _ _ _ __

A. continue to run, but chilled water inlet temperature will lower to 40 OF.

B. continue to run, but will no longer supply performance data to the ICS computer.

D. immediately trip, then will automatically restart in the Analog Mode.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: C Explanation: a. Incorrect. The chiller will trip, however the chilled water temperature setpoint while operating in Analog mode is 2 of lower at 40 of.

b. Incorrect. The chiller will trip, however the Chiller will not supply digital data to ICS while running in Analog Mode

c. Correct answer.

d. Incorrect. The chiller will trip as stated in (c) above. The chiller must be manually shutdown prior to transferring to Analog control, then must be manually restarted.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 0-01-31 (Attach if not previously provided)

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

Proposed references to be provided to applicants during examination: None Question Source: Bank#

Modified Bank # (Note changes or attach parent)

New 09/15/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet SFN Control Bay and Oft:..Gu* Treatment 0-01-31 Unit 'I} Bu~ldling Air Condiitioning System Rev,0126 Page 1,(): of 183 3.2 Uniiit 112 Control Bay Chillers A Compress<< sump heaters :are required to 00 energi;red for <lJ minimum of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to starting a chmIeL This prevents compressor damage caused by Uquid refiiger.am in too compressor :at startup. When the heaters are on, the bottom end {end dooest to too chilier control paned} of the compressors w~1 be nrm . This requirement may be modified by the System Engineer taking into account the length of time fhe compressor ns stmtdOWlfl and the owruoo air temperatures.

B. Prior to :any draining being performed on the Unit 112 Control Bay Chilled Water S').fStem, Chemistry and Environmental shook! be notified of the planned system draining.

C. When transferring between analog and digft<ill1 contro~ modes on <lJ Chillef, the mimer must NOT be runnang. Therefore, \Whenevef CONTROL PANEL MODE SELECT, Il..HS-031 ..2110AA{2210BA), swrtl:::hes are to be manipulated, ttle associated Chimer is required to be shutdown.

D. The ChineI' Control Switch. located on the local display pal1el for each Control Bay Chiller Control Pi3lnei has three posiUoos; STOPIRESET, AUTO LOCAL, and AUTO REMOTE.. The STOP/RESET position shws down the chiller when in local control.. The AUTO LOCAL position is used to stlrt the chUler when in 10c311 control. The AUTO REMOTE posilion iiS NOT used.. The control switm should NOT be seleded ito the AUTO REMOTE oogliol1.

E.. 100 Control Bay Chilkld Water PUmps A or B are tripped, by kJad shed sjgnal.

PUmps may 00 restarted after 10 minutes with the use of "CHW PUMP LOAD SHED BYPASS SW foreach pump. (A PUMP "O-HS-031-2101 E" location O-LPNL-925-0:165 PANEL D}(B PUMP "D-HS-031..2201 E" Location O-LPNiL-925-0165 PANEL E) If 1 & 2 CaNT. BAY CHW PUMP A{B)

TRANSFER SWITCH, O-XS-031-2101(2201) is in LOCAL POSITION load shoo logic for these pumps is bypassed.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BfN Control Bay iJlnd Off:'GiliS Treatment ,0-01-31 Unitt Building Air Condiitiooing System Rev. 0126 PaJge 55 of 283 5.16 Startup of Unit 1/2 Control Bay Chiller A Inloca:i Digital Contrm Mode (continued)

(6] VERFY the foUO\i\Iing switch positions: (Unit 1/2 Gootrul Bay Chiller A COntrd Paoon:

ANALOG CONTROL MODE SElECT, O-HS-031-2110AB{HS2), in .ANAlOG STOP. 0

CONTROL PANEL MODE SELECT, O-HS-031-2110AA(HS1), in DIGITAL CON.TROL 0

UNIT 1&2 CB CHIlLER A REMOTE HS APPENDIX R DISC SW, O-HS-031-2110 in LOCAL 0

CONTROL BAY CHillER A, O-BKR-031-211O, in ON. 0

CONTROL BAY CH~lLER A LOCAL DISPLAY PANEL O-PMC-031-2100A Display Window is illumfinated. 0 NOTE Step 5.16171 starts the chUiler. There is approximately a 2 minute Ume delay between when the switch is placOO in AUT01LOCAl and when the chiller ad:ualf,y starts.

171 At CONTROL BAY CHILLER A LOCAL DISPLAY PANEL, O-PMC-031-2100A, PLACE switch in AUTO LOCAL o (8J CHECK thie foUowfing parameters for Chiller A:

1&2 CB CHW PUMP A INLET PRESSURE, D-P&-031-0194, is appro::dmately 15-30 pSig. 0

1&2 CB CHW PUMP A(8) DISCHARGE PRESSURE, O-P&-031-0195, is appro::dmatelY 85-105 psig. 0

On 0-PMC-031-2100A (Menu-PO point F), Evap Leaving Water Temp efo.'entuaMy !Jowers to 42 +/- 2QF. 0

ES-401 Sample Written Examination Form ES-401-S Question Worksheet BfH Control Bayandl Off:'Gn Trellunent 0-01-31' UllitO Building Air CooditiooinQ System Rev. 0126 Paoe 03 of .Z8J 5.t8 Startup of Unit. 112 Control Bay Chiner A tn Analog Contro~ Mooe (contimmd)

NOTE Step 5.18[Bl srnrt5 the chmler at {Unit 112 Control Bay Chi!llerA Control Pane4}. There is approximately a 5 minute time delay between when the switch is ~ced in AUTO and whoo the chiller actually srnrt5.

(8] PLACE ANALOG CONTROL MODE SELECT, O-HS-031-2110AB{HS2), in ANAlOG START. 0

[9] RECORD switch manipulati<Jns in Steps 5.1B[7} and 5,13[8] in Narrative l<Jg., 0

[10] CHECK the fdl<J'.I'I'rng parameters on Chiller A:

1&2 CB CHW PUMP A INLET PRESSURE,

()"PI-031-0194, is approximately 15-30 pstgi. 0

'" 1&2 Cs. CHW PUMP A(B) DISCHARGE PRESSURE,

()"PHl3Hl195, is approximately 85-105 psig. 0

'" VERifY Evaporamr Leaving WaterTernp evenwallly 1(J'NBfS to 40 +/- 5 Q F on 1&2 CS CHILLER A CHW OUTLET TEMP IND,0-TI-031-0023. 0

APPENDIX D: Test Data Changed (TRANSACT Table)

CR NRC00028451 Report Name:

Quarterly Load - Office = "V" Records for Changed to:

Docket No = 05000259 PA NO = 122C91A Week_Ending = 10/4/08 & 10/11/08 Initials = KHD Docket No . = 05000260 PA NO = 122C91A Week_Ending = 10/4/08 & 10/11/08 Initials = KHD Docket No = 07007001 PA NO = 333240A Week_Ending = 10/4/08 & 10/18/08 Initials = MY9 Docket No = 05000259 PIC CD = 19 Week_Ending = 10/4/08 Initials = IFH LFARB #1 - Office = "V" Records for Changed to:

Docket No = 07000143 PA NO = 333C91A Week_Ending = 10/4/08 Initials = GAQ Docket No = 07000027 PA NO = 333C91A Week_Ending = 10/4/08 Initials = SGQ Docket No = 05000250 PA NO = 122C91A Week_Ending = 10/4/08 Initials = JS8 Docket No = 05000251 PA- NO = 122C91A Week_Ending = 10/18/08 Initials =MKZ Docket No = 05000259 PA- NO = 122C91A Week_Ending = 10/4/08 & 10/11/08 Initials = CRY &TLR

LFARB #2 Report - Office = "V" Records for Changed to:

Docket No = 05000391 PA NO = 122C92B Week_ending = 10/4/08 & 10/11/08 Initials = THN Docket No = 05000325 PIC CD = 19 Week_Ending . = 10/4/08 & 10/18/08 Initials = RNA LFARB #3 Report - Office = "M" Records for Changed to:

Docket No = 00000700 PIC CD = 19 TACNo = MD2953 Week_Ending = 10/4/08 & 10/18/08 Initials = G9B &RML LFARB #4 Report - Office = "D" Records for Changed to:

Docket No = 05000219 PA NO = 122122A Week_Ending = 10/11/08 & 10/18/08 Initials = DYA Docket No = 05000244 PA NO = 122122A Week_Ending = 10/4/08, 10/11/08 & 10/18/08 Initials = DLP Docket No = 05000313 PA NO = 171105AC TACNo = MD7067 & MD7178 Week_Ending = 10/4/08 & 10/11/08 Initials = ADW Docket No = 05000368 PIC CD = 19 TACNo = MD5250 & MD7068 Week_Ending = 10/4/08 & 10/18/08 Initials = ADW Casework Fee Memo Report - Office = "D" Records for Changed to:

Docket No = 07007011 PIC CD = 19 W eek_Ending = 10/11/08 Initials = KWJ

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

( Examination Outline Cross-reference: Level SRO 290001Al.0l Tier # 2 Ability to predict and/or monitor changes in parameters associated 2 Group #

with operating the Secondary Containment controls including:

System Lineups. KIA # 290001A1.01 Importance Rating 3.1 3.1 I Proposed Question: RO # 37 Unit 3 Reactor Building Ventilation is running with the Standby Gas Treatment (SGT) Systems in a normal standby lineup.

An event occurs which results in the following Unit 3 plant conditions:

Reactor Water Level increasing from a low of +8 inches.

Drywell Pressure 1.5 psig and steady.

Reactor Building Exhaust duct radiation 62 mR/hr and increasing slowly.

Refuel Floor Exhaust duct radiation 65 mR/hr and steady.

Which ONE of the following describes the Secondary Containment ventilation lineup for these conditions?

The SGT systems are (1) . Reactor Building ventilation is (2) and Refuel Floor Ventilation is (3)

(1) (2) (3)

A. running with suction from secured with running with the Reactor Building. dampers isolated dampers open.

B. running with suction from running with secured with the Refuel Floor. dampers open dampers isolated.

c. in a normal standby running with running with lineup. dampers open dampers open.

D. running with suction from running with running with the HPCI system. dampers open dampers open.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: C Explanation: a. Part (1) and (2) are incorrect. No isolation setpoint has been exceeded. If any of the given conditions were above the set pOint, Part (1) and (2) would be correct and Part (3) would be incorrect. With the given conditions, only Part (3) is correct. Refuel Floor ventilation is operating properly.

b. Part (1) and (3) are incorrect. No isolation setpoint has been exceeded. If any of the given conditions were above the set point, Part (1) and (3) would be correct and Part (2) would be incorrect. With the given conditions, only Part (2) is correct. Reactor Building ventilation is operating properly.

c. Correct answer.

d. Part (1) is incorrect. If Drywell pressure was above 2.45 pSig, this would be a correct answer, but Part (1) and (2) would become incorrect. With the given conditions, Part (1) and (2) are incorrect.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s):OPL 171.067 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank # 290001 K1.08 minor format changes Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 ,X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Original question 290001K1.08:

Unit 3 Reactor Building Ventilation is running with the Standby Gas Treatment (SGT) Systems in a normal standby lineup. An event occurs with the following conditions present on Unit 3:

- Reactor Water Level increasing from a low of +8 inches

- Drywell Pressure 1.5 psig and steady

- Exhaust duct radiation 62 mR/hr and increasing slowly

- Refuel Floor radiation 55 mR/hr and steady Which ONE of the following describes the Secondary Containment ventilation lineup for these conditions?

The SGT systems are ....

A. running with their exhaust to the Main Stack; Reactor Building supply and exhaust dampers closed.

B. running with their exhaust to the Reactor Building Vent Stack; Reactor Building supply and exhaust dampers closed.

C. in their standby lineup; Reactor Building ventilation remains in the normal lineup exhausting to the Reactor Building Vent Stack.

D. in their standby lineup; Reactor Building supply and exhaust dampers are isolated from the Reactor Building Vent Stack.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.067 page 16:

1. System Isolation (Group 6)

a. The isolation signals are low reactor water level +2", high drywell pressure 2.45 psig, and high radiation in exhaust duct (72 MR/hr Refuel zone or Reactor zone).

b. On auto isolation signal (except Refuel Zone high radiation) the unit reactor zone supply and exhaust fans trip, all refuel zone supply and exhaust fans trip, unit reactor zone and all refuel zone supply and exhaust isolation dampers close, and dampers to Standby Gas Treatment System open and SGT train blowers auto start.

c. Note: Damper logic is as follows: PCIS Group 6 with A SGT running opens 64-41 and 45; PCIS Group 6 with B SGT running opens 64-40 and 44.

d. On isolation due to high radiation in Refuel Zone all refuel zones isolate, refuel zone supply and exhaust fans trip, SGT starts and aligns dampers for refuel zone only.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 290003Kl.04 Tier # 2 Knowledge of the physical connections and/or cause-effect 2 Group #

relationships between Control Room HVAC system and the following: Nuclear Steam Supply Shut off System (NSSSS/PCIS). KIA # 290003K1.04 Importance Rating 3.2 3.3 I Proposed Question: RO # 38 Given the following Control Room Emergency Ventilation (CREV) system conditions:

CREV Train "A" was started to prove operability following maintenance on the charcoal trays using the STOP-AUTO-START switch on Panel 9-22.

The SYSTEM PRIORITY SELECTOR SWITCH is selected for "TRAIN-8".

Which ONE of the following describes the CREV system response should a valid CREV initiation signal be received?

On a valid initiation, CREV Train "8" would _ _ _~(1:..J-)_ _ _ and CREV Train "A" would (2)

(1) (2)

( A. initiate shutdown.

8. initiate NOT shutdown.

C. NOT initiate shutdown.

D. NOT initiate NOT sh utdown.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: B Explanation: a. Part (1) is correct. CREV Train B will initiate without a time delay since the CREV UNIT PRIMARY SELECTOR SWITCH is selected for "TRAIN-B". Part (2) is incorrect. CREV will not automatically shutdown with a valid initiation signal present.

b. Correct answer.

c. Part (1) is incorrect. Normally, when an auto initiation signal is received, the TRAIN selected for "secondary" begins its start sequence but will not finish if the Primary CREV train is running. This is sensed by looking at the tlP across the HEPA filter. Since Train B was selected as the Primary CREV unit, the start sequence does not look at the tlP. Part (2) is incorrect. CREV will not automatically shutdown with a valid initiation signal present.

d. Part (1) is incorrect as stated in (c) above. Part (2) is correct. This would be incorrect if CREV Train A was started using the AUTO-INITIATE TEST switch, as would be the case during the periodic surveillance test.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 0-01-31 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank # OPL 171.067.49 minor format modifications Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Original question OPL171.067.49:

CREV Train A is running for testing to prove operability following maintenance on the charcoal trays.

The CREV UNIT PRIMARY SELECTOR SWITCH is selected for "8".

DETERMINE which of the following is the proper sequence of events and/or operator actions should a valid CREV initiation signal be received.

A. CREV Train 8 will initiate on a valid start signal, Train A will shutdown, no operator action required.

B. CREV Train 8 will initiate on a valid start signal and Train A must be manually shutdown.

C. Neither Train would initiate on a valid start signal since Train A is being tested and would shutdown automatically. Train 8 must be manually started.

D. CREV Train 8 will NOT initiate on a valid start signal. Train 8 must be manually started and Train A must be manually shutdown.

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Control Bay IiIld Off-Gas Treatment 0-01-31 UoitB Building Air Conditioning System Rev. 0126 Paoe 21 of 283 3.6 CREV and CREV instrumentation opet"ability issues (continued)

B. The main oontroll'OOifl'1l boundary may be opened: intermiHentty Ufl(!er administration oon1rols. For openings other than normal entry and exit., these controls consist of stiIIioo4ng a dedicated individual at the opening who is in continuous oommunication wiIh the lmIin con1rol room and whose task is to dose the opening When main control room isolation is indicated. With 00ttl CREVs inoperable in Modes 1, 2 or 3, for other than a control room boundary issue, enter LCO 3.0.3 Immediately. With. two CREY subsystems inoperable during OPDRVs, ini~teadtoo to suspend OPDRVs. Reference TS 3.7.3.

C. When there is an automatic actuation of CREVS, the fonowing automatic isolation dampers and hatch is required to be closed for CREVS to be considered operable.

1. O-FCO-31-150S, O-FCO-31-150D. O--FCO-3;1-150E, O-FC0-31-150F.

O-FCO-31-150G.

2. Removable equipment hatch in U-3 Medlamcal EqUipment Room, floor Elevation 617'.

0; The CREV system utilizes 15.45 kW Duct heaters to control mtJisture buildup in the charcoat adsorber. A malfunction ofUle 15.45 kW wct heater makes the apPlicable CREY unit inoperable. [Rererence Fuoc1ional Evaluation in PER 74959 and 75680]

E One of Ule UNIT 1 & 2 Control Bay SUpply Fans and ooe of the UNIT 3 Controi Bay Supply Fans andl their associated power and control circuits is required to be operable for CREVS instrumentation {control bay high radiation)to be considered operable. Reference Tech Spec 3.3.7.1.

F. CREY UNIT PRIMARY SELECTOR, O-XSW-031-7214 may be Placed, in either the "AD or "8'" position, depending 00 the operability status ofttie CREY trains.

When a CREV train is inoperable, it wiN NOT be selected as lead. When both CREV trains are operable, the preferred position for eREV UNIT pmMARY SElECTOR, O-XSW-031-7214, is in TRAIN "A" Which makes Ule N' CREY the lead train. In the event that "A" CREY is INOP, CREV UNIT PRIMARY SELECTOR, O-XSW-031-7214 is required to be P1acedl in the TRAIN "S" position so the "S" CREV ,,111 inmate, without a time deiay, as the lead train.

G. When one of the eREV trains is inoperable for testing, the CREY UN IT PRIMARY SELECTOR SWITCH, 0-XSW-031-7214 is required to be aliglnedl to the train Which is NOT under testing conditions to ensure the non-test train will initiate under an actual initiation signal.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 29S001AK1.03 Tier # 1 Knowledge of the operational implications of the following concepts 1 Group #

as they apply to the Partial or Complete Loss of Forced Core Flow Circulation: Thermal Limits. KIA # 295001AK1.03 Importance Rating 3.6 4.1 I Proposed Question: RO # 39 Which ONE of the following describes the response of thermal limits due to the trip of a single recirculation pump from 100% rated power?

A single recirculation pump trip from 100% rated power will cause the value of Critical Power to (1) and the Critical Power Ratio will (2)

(1) (2)

A. lower lower B. lower rise

c. rise lower D. rise rise

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: B Explanation: a. Part (1) is correct. As the actual power goes down, the power required to cause the onset of transition boiling also goes down. Part (2) is incorrect.

Although actual power goes down and critical power goes down, the power required to cause the onset of transition boiling does NOT go down as far as actual power due to the higher void fraction. Therefore, the Critical Power Ratio rises.

b. Correct answer

c. Part (1) is incorrect. As actual power goes down, void fraction increases.

This causes critical power to lower, although not as significantly as actual power. Part (2) is incorrect as stated in (a) above.

d. Part (1) is incorrect as stated in (c) above. Part (2) is correct. Since actual power drops farther than critical power, the Critical Power Ratio gets larger.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): GFES Thermal Limits (Attach if not previously provided)

OPL 171.007, Recirculation System LP Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 9/2/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

(

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

( Excerpt from OPL171.007 Page 28 of 86:

\..

b. The protective action of a scram is the When voids collapse, normal means of terminating this transient. both actual power At end of cycle conditions, control rods may and critical power go all be fully withdrawn, while thermal neutron up, however, actual flux has shifted upwards in the core. This power goes up more will delay the effect of negative reactivity than critical power.

from a control rod scram. In order to This results in being provide a means for adding additional closer to CPR limits.

negative reactivity, an EOC-RPT system has been designed to trip recirc pumps to allow additional void formation.

Table 9-1 Factors Affecting Critical Power FACTOR CRITICAL BUNDLE CPR POWER POWER INLET SUBCOOLING:

INCREASES t t ,1.

DECREASES ,1. ,1. t MASS FLOW RATE:

INCREASES t t ,1.

DECREASES ,1. ,1. t PRESSURE:

INCREASES ,1. t ,1.

DECREASES t ,1. t LOCAL PEAKING FACTOR INCREASES ,1. ++ ,1.

DECREASES t ++ t AXIAL POWER DISTRIBUTION INCREASES ,1. ++ ,1.

DECREASES t ++ t STEADY STATE AND TRANSIENT The primary design objective is to maintain nucleate boiling and avoid OTB. The CPR thermal limit is set to maintain adequate margin between nucleate bOiling and OTB. The steady state and transient MCPR thermal limits are derived from this single design basis requirement.

Transients caused by single operator error or equipment malfunction shall be limited so that, considering uncertainties in monitoring the core operating state, more than 99.9% of the fuel rods are expected to avoid OTB.

The transients most likely to limit operation because of MCPR considerations are:

Turbine trips or generator load rejections without bypass valve capability

Loss of feedwater heating or inadvertent high pressure coolant injection

Feedwater controller failure (maximum demand)

MAXIMUM FRACTION OF LIMITING CRITICAL POWER RATIO (MFLCPR)

The process computer calculates CPR data evaluating core conditions to ensure limits are not exceeded. One of the most useful forms of this data output is a ratio called the "fraction of limiting critical power ratio" (FLCPR). This ratio compares the flow-adjusted operating (steady-state) maximum CPR for the fuel bundle to the actual bundle CPR. From this, the maximum fraction of limiting critical power ratio (MFLCPR - pronounced "miffle-sipper"), which is the maximum fraction of limiting critical power ratio (MFLCPR) and is the ratio of the flow-adjusted CPR operating limit for that fuel type to the bundle CPR, is developed. For most nuclear plants the MFLCPR ratio takes the following form:

MFLCPR = CPR Limit X K f CPR Equation 9-16

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet Examination Outline Cross-reference: Level SRO 295003AK1.02 Tier # 1 Knowledge of the operational implications of the following concepts 1 Group #

as they apply to the Partial or Complete Loss of AC: Loa.d Shedding.

KIA # 295003AK1.02 Importance Rating 3.1 3.4 I Proposed Question: RO # 40 Given the following Unit-2 conditions:

A Loss of Off-Site power and LOCA has occurred.

480V Load Shedding Logic has actuated.

The Unit Operator immediately clears the "A" RBCCW pump white disagreement light while surveying panel 9-4.

No other actions were performed.

Which ONE of the following describes the effect on the RBCCW system?

A. "A" pump auto starts after 40 sec; "B" pump can be manually started immediately.

B. "B" pump auto starts after 43 sec; "A" pump can be manually started after 40 sec.

C. "A" pump auto starts after 40 sec; "B" pump auto starts after 43 secs.

D. "B" pump auto starts after 43 sec; "A" pump can be manually started immediately.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: B Explanation: a. Part (1) is incorrect. This is true only if the breaker control switch was left in the "normal-after-start" position. Part (2) is incorrect. All RBCCW pumps are prevented from starting for 40 seconds following a load shed regardless of switch positions.

b. Correct answer.

c. Part (1) is incorrect as stated in (a) above. Part (2) is correct.

d. Part (1) is correct. Part (2) is incorrect. All RBCCW pumps are prevented from starting for 40 seconds following a load shed regardless of switch positions.

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 2-01-70, Rev 59 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # OPL 171.072.03 attached New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Original question: OPL171.072.03 480V Load Shedding Logic has actuated on Unit 2 when the operator clears the "A" RBCCW pump white disagreement light while surveying pnl. 9-4 with no other actions.

Which one of the following statements describes the effect on the RBCCW system?

A. "A" pump auto starts after 40 sec; "B" pump auto starts 3 sec. later.

B. "B" pump auto starts after 43 sec; "A" pump can be manually started after 40 sec.

C. "B" pump auto starts after 40 sec; "A" pump can be manually started immediately.

D. "A" pump auto starts after 40 secs; "B" may then be manually started.

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Reactor Building Closed Cooling Water 2-01-10 Unit 2 System Re.... OO59 Page 10 of 62 3.0 PRECAUTIONS AND UMITAnONS (cootinued)

P. With an accident signm present (low Reactor water level or high Drywall pressure) on Unit 1 or Unit 2 and any Diesel Generator output breaker closjng 00 U-1 and 2 Shutdown boards, the foItowing occurs:

1. Unit '1 and Unit 2 RBCCW pumps trip.

2. Unit 1 and Unit 2 Drywell blowers tnp.

l. 1-FCV-70-48 and 2-FCV-70-48 close when power is restered (close signal present for 40 seconds). This auto closure is bypassed if 2(1 )-XS-70-48 at 480 RMOV board 2(1)B is in the EMERGENCY position.

4. After a 40 second time delay, the following occurs:

a. With the control switch in Norma! After Start, RBCCW Pump A restarts for Unit 1 and Unit 2.

b. If RBCCW pump A fails to start, RBCCW Pump B will automatically start after a 3 second time delay for Unit 1 and Unit 2 (with the comol switch in normal after start).

c. The Drywell Blowers on the unit w~thout the accident will auromatically restart (Unit 2 blowers will have staggered auto start times). Unit 2 Drywell blowers with their respective Auto Start Inhibit switch in the INHIBIT position w~1 not auto start, but can, however, be manuaBy started after a ten minute time delay_

d. The Drywel! Blowers A1, B 1, A2. and B2 on the unit with the accident may be manum!y restarted after 40 seconds.

e. The Drywell Blowers AJ, B3, A4. 84, AS and 8500 the unit with the accident will remain tripped.

Q. Rotork valve operator indicators have three indications. "Fu!f~ Open," "Full Closed,." and "Mid Positions. The I*Mid Position" mere.y indicates that the valve is neither "Full Open" nor "Full Closed." It does not represent a percentage Open Of Closed.

R. Temperature Control Valves are required to be isolated very SLOWlY (controlled manner)to 'ensure no erratic system perturbations result in ESF initiations. !PER 01-005343-000]

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 29S004AA2.03 Tier# 1 Ability to determine and interpret the following as they apply to a 1 Group #

Partial or Total Loss of DC Power: Battery Voltage.

KIA # 295004AA2.03 Importance Rating 2.8 2.9 I Proposed Question: RO # 41 The following plant conditions exist:

Complete loss of offsite power on Unit 2.

All 4KV Shutdown boards are being supplied by their Diesel Generators.

DW pressure: 2 psig slowly rising.

RPV level: (-) 140 inches and stable.

RPV pressure: 800 psig and stable.

HPCI and RCIC are injecting to the vessel.

250V Reactor MOV BD 2A UV (9-8C W4) is in alarm.

What ONE of the following describes the actions required to restore 2A 250v Reactor MOV Board voltage to normal?

On 250V Battery Charger _~(=..J1)"---_" perform the following: _ _ _~_~(2=.<),--_ __

(

(1) (2)

A. 2A Place the Emergency ON select switch in "Emergency ON."

B. 2A Manually re-close the normal feeder breaker following a 40 second time delay.

c. 1 Place the Emergency ON select switch in "Emergency ON."

D. 1 Manually re-close the normal feeder breaker following a 40 second time delay.

(

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

( I Proposed Answer: A Explanation: a. correct answer.

b. Part (1) is correct. Part (2) is incorrect. This action would be correct for Battery Charger 4, but is incorrect for Battery Charger 2A.

c. Part (1) is incorrect. This would be the correct battery charger for 250V Reactor MOV Board lA. (wrong unit/train issue) Part (2) is correct and would allow Battery Charger 1 to restart as well.

d. Part (1) is incorrect as stated in (c) above. Part (2) is incorrect. This action would be correct for Battery Charger 4, but is incorrect for Battery Charger 2A.

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 2-ARP-9-8C, 0-01-570 (Attach if not previously provided)

OPL 171.037, DC Distribution Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # RO 295004AK3.01 attached New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Original question RO 295004AK3.01:

The following plant conditions exist:

- Complete loss of offsite power on Unit 2

- All 4kv SID boards are being supplied by their Diesel Generators

- DW pressure: 2 psig slowly rising

- RPV level: -140 inches and stable

- RPV pressure: 800 psig and stable

- HPCI and RCIC are injecting to the vessel What actions are required to restore 2A 250v Battery Charger?

A. The battery charger can only be energized when the accident signal clears.

B The battery charger can be re-energized by placing the emergency bypass switch to bypass.

C. The battery charger is energized, it automatically restarts forty seconds following an accident signal and requires no manual actions to restore it.

D. The battery charger is energized, it requires no manual actions to restore it.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN DC Electrica' System 0-01-51D Unit 0 Rev. 0117 Page 16 of 247 3.0 PRECAUTIONS AND LIMITATIONS {continued}

E. All safety requirernents concerning smoking, fires or sparks should be observed when in the Battery-Battery Board Rooms because of potential accumulation of hydrogen in flammable amounts.

F. 250V Untt Battery Charger 1,2A,2B and 3 Emergency ON select switch bypasses battery charger emergency load shed contacts. Placing the select switch in Emergency ON reestablishes charger operations with an accident signal present and Diesel Generator voltage availa.ble. Battery Charger 4 supply breaker, 480V Shutdown Board 38, Compt 6D, receives a trip signal from the load shed logic and the breaker must be manually re-closed after a 40 second time delay to restore the charger to service.. The annunciation circuit for the 250V Unit Battery Charger 3 does NOT work when the EMER/OFF/ON Select Switch is in the EMER Position.

G. [1I1e] Neutron monitoring battery chargers are NOT stand alone power supplies and shallonJy be operated while connected to the neutron monitoring batteries.

[aFPER 9408621 H. Within 30 minutes after the loss of the normal charger to a 250V Unit Battery another charger shat! be placed in servi:ce to that battery and load reduced so that the battery is NOT discharg.ing.

L [NRClGj Upon retum to service of24V DC Neutron Monitoring Battery A or B, Instrument Maintenance must perform fun ctionaf tests on SRMs and IRMs that are powered from the affected battery board (In that thelRMsand SRMs are normally inoperable afierentering RUN mode due to lack of testing, these tests are N/A for the IRMs and the SRMs if the Unit is in RUN Mode and the lRMs and SRMs are inoperable). Prior to calling the IRMsand SRMs operable, the tests have to be performed. [NRC IE Inspect Follow-up llan s,o-4Q!

J. To return equipment to sefVice following a failure or trip, the shutdown section of this instruction should be performed on the equipment failed. Theinitiaf conditions may NOT be applicab~e in this case.

K. [NRCiGJ The transfer of 250VDC contml power to a 4kV Shutdown 80ard with a diesel generator operating may cause an inadvertent start of a RHRSW pump.

[tER 88:D21125j

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.037, DC Distribution, page 14 of 70:

They also supply alternate control power for Units 1 and 2 4kV Shutdown Boards; however, on Unit 3, the A, C, and D 4kV Shutdown Boards receive both normal and alternate control power from the 250V DC Unit Systems. (3EB receives alternate control power only.) The 250V DC RMOV Boards are supplied from the Unit Battery Board as follows: BB-l supplies 250V RMOV Boards lA, 2C, 3B. BB-2 supplies 250V RMOV Bds 2A, lC, 3C. BB-3 supplies 250V RMOV Boards 3A, lB, 2B.

Excerpt from OPL171.037, DC Distribution, page 31 of 70:

250V Batte!y Charger Normal Source Alternate Source (Charger Service bus) 1 480V SD Bd lA, Comp 6D 480V Common Bd 1, Comp 3A 2A 480V SD Bd 2A Comp 6D 480V Common Bd 1, Comp 3A 2B 480V SD Bd 2B, Comp 6D 480V Common Bd 1, Comp 3A 3 480V SD Bd 3A, Comp 6D 480V Common Bd 1, Comp 3A 4 480V SD Bd 3B, Comp 6D 480V Common Bd 1, Comp 3A 5 480V Com Bd 1 Comp 5C (no alternate) 6 480V Com Bd 3 Comp 3D (no alternate)

The 2B spare charger DC output can be directed to any of four feeders. Three DC outputs can be connected to*

battery board 1, 2, or 3. The fourth DC output is connected to output transfer switch (BBR 4) to batteries 4, 5, or 6.

Mechanical interlock permits closing only one output feeder at a time. (A slide bar is utilized in battery board room 2 and a Kirk key interlock is used in battery board room 4.)

250V DC battery chargers 1, 2A and 2B will load shed upon receipt of a Unit 1 or Unit 2 accident signal and any Unit 1/2 shutdown board being supplied by its respective diesel generator or cross tied to a Unit 3 shutdown board and a unit three Diesel Generator. 250 VDC Battery Charger 3 will load shed on a unit 3 load shed signal. The load shedding feature can be bypassed by placing the "Emergency" switch on the charger to the "EMERG" position.

Station Battery charger 4 does not have load shed logic; however, battery charger 4 will deenergize when 3B 480 SID Board deenergizes and will return when the 480V SID Board voltage returns.

They also supply alternate control power for Units 1 and 2 4kV Shutdown Boards; however, on Unit 3, the A, C, and D 4kV Shutdown Boards receive both normal and alternate control power from the 250V DC Unit Systems. (3EB receives alternate control power only.) The 250V DC RMOV Boards are supplied from the Unit Battery Board as follows:

BB-l supplies 250V RMOV Boards lA, 2C, 3B.

BB-2 supplies 250V RMOV Bds 2A, lC, 3C.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 29S00SAA2.04 Tier # 1 Ability to determine and interpret the following as they apply to a 1 Group #

Main Turbine Generator Trip: Reactor Pressure KIA # 295005AA2.04 Importance Rating 3.7 3.8 I Proposed Question: RO # 42 Given the following plant conditions:

Reactor power is 38% power.

Main turbine load is 23%.

Turbine bypass valves are partially open.

Which ONE of the following describes the response of the reactor if the Main Turbine Generator inadvertently trips?

The reactor will ____________

A. scram on High Reactor Pressure.

B. immediately scram on Turbine Stop Valve closure.

C. continue to operate at 38% power with Bypass Valves open.

D. continue to operate above 38% power due to a loss of Feedwater heating.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: A Explanation: a. Correct answer.

b. Incorrect due to low turbine first stage pressure. If turbine load was slightly higher, the reactor would scram on TSV closure.

c. Incorrect due to power slightly greater than Bypass Valve capacity. If power were lower, than the reactor would continue to operate.

d. Incorrect due to power slightly above Bypass Valve capacity. If power were slightly lower and the reactor did not scram, the reduction in Feedwater temperature would cause a small power increase.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 1-01-99, Reactor Protection System (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank # 295005AA2.05 Attached Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

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

( Bank Question 29S00SAA2.0S Given the following plant conditions:

Reactor power is 38% power.

Main turbine load is 23%.

Turbine bypass valves are partially open.

Which one of the following describes the response of the reactor if the Generator Breaker inadvertently OPENS?

A. Reactor immediately scrams on turbine stop valve 10% closure.

B. Reactor scrams on high reactor pressure.

C. Reactor continues to operate at 38% power.

D. Reactor continues to operate and power decreases to 30%.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BfN Reactor Protection System 1-01-99 Unit '1 Rev. 0033 Page 52 of 68 Illustration 2 (Page 2 of 2)

Unit 1 Reactor Scram Initiation Signal Scram Setpoint Bypass J. OPRMTRIP Anyone of the three Reactor is NOT operating in algorithflls, period, growth, or the AUTO ENABLE Region of amplitude for an operabJe the Povrerfffow Map.

OPRM cell has exceeded its trip value conditions:

K. LowRPV Watef +2.0" NlA Level (Level 3)

L Hi RPV Pressure 1088 psig NlA M. Hi DW Pressure 2.45 psig N/A N. MSIV dosure 900/0 open (3 Main Steanl NOT in RUN lines)

O. Scram Discharge .. Thefmallevel s'llllitches Mode Swit:ch in SHUTOOlNN Instrument 49 gallons or REfUEL with keylock VoiumeHiHi (lS..85-45A,B,G,H) switch in BYPASS

.. Float level s'Witches 45 gallons

{LS-85-45C, D,E,f)

P. TSV Closure 90% open (3 TSVs) < 3O%Rx Power

(~'148.5 psiS 1st stage pressure)

Q. TCV Fast Closure 40% mismatch (amps to < 30% Rx Power (load reject) cross-t.tl1der pressure); (~ "148.5 psiS 1st stage 850 psig EHC RETSat pressur,e)

TCV (1 or 3) & (2 or 4)

R LossofRPS N/A NlA POYtIer S. Scran Channel Key-locked in AUTO Panels N/A Test Switches 1-9-15 & 1-9-17

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 295006AK1.03 Tier # 1 Knowledge of the operational implications of the following concepts 1 Group #

as they apply to a Scram: Reactivity Control KIA # 295006AK1.03 Importance Rating 3.7 4.0 I Proposed Question: RO # 43 Unit 2 has received a scram signal but some of the control rods failed to fully insert. The Unit Supervisor has directed you to insert control rods as directed by 2-01-85, "Control Rod Drive System."

Which ONE of the following control rod insertion processes can ONLY be accomplished in the main control room as directed by 2-01-85, "Control Rod Drive System?"

A. Removal and replacement of RPS scram solenoid fuses.

B. Venting and re-pressurizing the Scram Pilot Air Header.

C. Insertion of control rods by venting the over piston area.

D. Control rod insertion using raised cooling water differential pressure.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: 0 Explanation: a. Incorrect answer. This action must be performed from the Aux Instrument Room.

b. Incorrect answer. This action must be performed in the Reactor Building.

c. Incorrect answer. This action must also be performed in the Reactor Building.

d. Correct answer.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 2-01-85 Section 8.19 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank# 295006AK 1.03 Attached Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Control Rod Drive System 2"()1-35 Unit 2 iRev. 0104 Page 134 of 181 8.19 Control Rods Which Fail to FUllY INSERT After Scram NOTE The operator should determine the most effective method to insert rods from the fOllowing sections:

Removal and Replacement of RPS Scram Solenoid Fuses (Section 8.19{1]).

Venting and Repressurizing the Scram Pilot Ak Header (Section B.19[2J).

Individually Scram Control Rods (Section 8.. 19[3]).

Insert Control Rods using Reactor Manual Control System (Section 8.19{4]).

Manual Insertion of Control Rods by Venting the OYer Piston Area (Section 6.19{5]).

Control Rod Insertion using Raised Cooling Water Differentied Pressure (Section 8.19[6]).

[1] IF Removal and Replac.erJlent of RPS Scram Solenoid Fuses is desired, THEN PERFORM the following:

[1.1] OBTAIN fuse pullers and PROCEED TO Unit 2 Auxiliary Instrument Room. o

{1.2) LOCATE terminal strip CC inside Panel 9-15, Bay 2, RPS CHANNEL A Panel (Rear). o

{1.3] REMOVE the fcllowing RPS Bus "1\ fuses (located at 00tt0m of terminal strip ce, Panel 9-15) AND DOCUMENT removal 00 131ustratioo 6. o FUSE LOCATION FUSE FUSEID CC4FU 5A-F18A 2-FU1-08&OO37AA CC-5FU 5A-f18E 2-'FU1-085-0039A12 CC-6FU 5A-F18C 2-FU1-00&0039A/3 CC-1FU 5A,.F18G 2-FU1-065-0039A14

[1.4] LOCATE terminal strip CC inside Panel 9-17, Bay 2, RPS CHANNEl B Panel (Rear). o

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BfN Control Rod Drive System 2-01-85 Unit 2 Rev. 0104 Page 135 of 181, 8.19 Control Rods Which Fail to fUllY INSERT After Scram (cordinued)

[1.5] REMOVE 1he following RPS Bus "B'" fuses (located at bottom of terminal strip CC, Panel 9-17) AND DOCUMENT rE!fOO\Ial 00 Illustration 6. o FUSE LOCATION FUSE FUSE 10 CC-4FU 5A-f18B 2-FU1-065-0031BA CC-5FU 5A-F18F 2..FU1-00&0039B12 CC-6fU 5A-F18D 2-FU1-005-0039B13 CC-1FU 5A-f18N 2-FU1-085-0039B14 (1.6] WHEN All fuses are removed, THEN NOTIFY Unit Opemtor. o

[1.7] WH'EN Shift ManagerlUnit Supervisor directs replacement of fuses,THEN REPLACE fuses listed in Steps 8.19[1.3) and 8.19{1i.51 above AND DOCUMENT replacement on Illustration 6. 0

{1.8] WHEN ALL fuses are replaced, THEN NOTIFY unit Operator. o 121. IF Venting and Repressurizing 1he Scram Pilot Air Header is desired, THEN PERFORM the following.:

p.1] CLOSE 2-85-331, CONT AIR SPLY HDR SOY (located on RX 81dg North wall near Scram Air Header Pressure Regu~). 0 p..2] OPEN instrument drain valves for the following pressure switch and gauge (located on 2-PNl-925-0018B, East end):

2-PS-85-38, CRD SCRAM VALVE PILOT AIR HEADER PRESS. o

2-P1-85-38. CRn SCRAM VAlVE PILOT AIR HEADER PRESS. o

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BfN Control Rod Drive System 2-01-85 Unit 2 Rev. 0104 Page 138 of 181 8.19 Coo1rol Rods Whim Fail to FULLY INSERT After Scram (continued)

[4.6] REFER TO Illustration 4 and DEPRESS the appropriate CRD Rod select pushbutton on 2-XS-85-40. 0

[4.7] CHECK backlit CRO ROO SELECTpushbutton ,is brightly illuminated and VJhite indicating light on Full Core Display illuminated. 0 14.8) CONTINUOUSlY INSERT control rod to Position 00, by holding CRD CONTROl SWITCH, 2-HS-85-48, in ROD IN OR CRO NOTCH OVERRIDE SWITCH, 2-HS-85-47, in EMERG ROD IN. 0

[4 . 9] IF controloodis difficult to insert, THEN REFER TO Section 8.16. o

[4. 10] REPEAT S1eps 8.19[4.6] through 8.19[4.8] for each Control Rod to be inserted. 0

[4.11] PLACE Rod Worth Minimizer Normal Bypass Switch in NORMAL in accordance INith Section 8.18. 0

{4.12] IlNPOIC} PLACE the Reactor Mode Switch in SHUTDOWN. [IN PO SOER 80-006 reoommendation 9] 0 14.13) VERIFY OPEN CHARGING WATER SHUTOFF, 2-SHV-085-05a6 (RB. EL 565 NE Comer). 0

[5] IF Manuallnsertioo of Control Rods by Venting fue Over Piston Area is desired, THEN PERFORM the foMowing:

[5.1] OBTAiN fue following equipment

Catwalk key from Unit 2 Control Room Key Cabinet. 0

Square valve stem operators (one L-shaped, T-shaped, 10" crescent and speed wrench). 0

50 feet high temperature hose with quick diSCOl1illle'd fitting on one end. 0

[52] ESTABlISH communications with Control Room. o

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

(

BfN Control Rod Drive System 2-01-85 Unit 2 Rev. 0104 Paqe 139 of 181 8.19 Control Rods Which Fail to FULLYINSERT After Scram (continued)

[53] REFER TO Illustration 4 and OBTAIN recotTlmended control rod insert sequence from Unit Operator. o

[5.4] REFER TO Hlustration 5 and PERFORM the following to vent each eRD in insert sequence recoml1"tef1ded by Unit Operator.

NOTE ONLY those CRD modules on the extreme North and Sooth ends of each row contaln EOI Identification Tags.

[5.4.1] UNLOCK AND CLOSE WITHDRAW RISER (SOL 2-ISV-085-615. 0

[5.4.2J DIRECT end of vent hose without quick disconnect coupling to Radwaste floor drain AND SECURE hose to floor drain cover. 0

{5.4.3J PROCEED TO catwalk with tools and equipment 0

[5.4.4J CONNECT quick disconnect end of vent hose to coupiing WITHDRAW RISER VENT TEST CONN, 2-85-623. 0

{5.4.5} IF valve stem cap is insta!~ for WITHDRAWAL RISER VENT. 2-VIV-085-614, THEN REMOVE '. . alve stem cap from WITHDRAWAL RISER VENT. 2-VIV-085-614. o CAUTION Opening of WITHDRAW RISER VENT valve more than two turns may result in burn and contamination hazard to personnel at the valve or floor drain area.

(5.<t6] SLOWlY OPEN WITHDRAW RISER VENT, 2-VTV-085-614, using T- or L-shaped wrench. o

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 295016G2.4.41 Tier # 1 Knowledge of emergency action level thresholds and classifications; 1 Group #

Control Room Abandonment KIA # 295016G2.4.41 Importance Rating 2.9 4.6 I Proposed Question: RO # 44 Which ONE of the following describes the emergency action level required by EPIP-l, "Emergency Classification Procedure" when the control room must be abandoned and AOI-l00-2, "Control Room Abandonment" is entered?

If the control room is evacuated and backup control from Panel 25-32 is NOT established within (1) , EPIP-l classifies the event as a/an (2)

(1) (2)

A. 20 minutes Alert B. 20 minutes Site Area Emergency

c. 15 minutes Alert D. 15 minutes Site Area Emergency

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

( I Proposed Answer: B Explanation: a. Part (1) is correct. Part (2) is incorrect. An Alert is declared immediately upon entering AOI-100-2.

b. Correct Answer.

c. Part (1) is incorrect. 15 minutes is plausible because it is the time limit associated with upgrading an emergency from a NOUE to an Alert in the event of a fire. (EAL 6.4-U1) Part (2) is incorrect as stated in (a) above.

d. Part (1) is incorrect as stated in (c) above. Part (2) is correct.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 1-AOI-100-2, EPIP-1 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None

~------

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 8/27/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet BFN Unit 1 ContTOiJ Room Abandcmmeot i1-A01- too,2 Rev. 0016, Page 4 of 19 1.2 Responsibilities (continued)

NOTE Evacuation or anticipated evacuation is classified by EPIP-1 as an Alert. If the control room is evacuated and backup control from Panel 1-25-32 is NOT established within 20 minutes.

EPIP-1 classifies the event as a Site Area Emergency. Oetaiils are COI"Itained in EPlP-1 and Technical Bases.

D. 'If Unit 3 Cooitrol Room is NOT affected, Unit 3 Unit Supervisor (SRO) assumes responsibiJity for EPIP implementation.

E. If ALL Control Rooms are affected, Shift ManagerlUnit Super:viSOf (SRO) assumes responsib~ity for EPIP implementation.

F. Responsib~ity for completing panel cheddsts is assigned: to individuals stationed in the area of equipment to be checked. Attachment 1 provides backup control station assignments.

2.0 SYMPTOMS A. Dense smoke in Unit '112 Control Room.

B. Toxic gas released through ventilation system.

C. A fire in the Unit 112 Control Room NOT meeting Appendix R entry conditions.

3.0 AUTOMATIC ACTIONS None

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

(

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 CCf1lim1ed fire in A~N ~ area fis~ed jn Unanticipated e)(plm!~ .t.'i~in the protected area Table6.4-'J1 resulting ~f! vi.siili!e oomagefo ANY penna:nent AND structure Of equi~lent NOT extinguished '1iithin 15 immures.

OPERA.TiNG COODITfON: OPERATIN*G CONDITION:

ALL ALL Fife or explos'on &n .A.NY plant area Ifsred ~f!

Tal'lle6.4-A affecting saiei'IS'f$rem peoor;r18nre OR Fl~eor e>:pfosion CBIJS'illg *....isible damage to perni'J!inent.structure of safety 'sysrel11S in ANY

~am area ~isted in Tab!.,,: SA-A.

OPERA.TiNG CONDITION:

.AlL PAGE 56 OF 201 REVISION 42

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 295018AK2.01 Tier# 1 KNOWLEDGE OF THE INTERRELATIONS BETWEEN Partial or Total 1 Group #

Loss of CCW and the following: System Loads.

KIA # 295018AK2.01 Importance Rating 3.3 3.4 I Proposed Question: RO # 45 Which ONE of the following describes the response to a partial loss of RBCCW?

Should the system discharge header pressure drop to less than (1) , isolation valve FCV-70-48 would close, causing RBCCW System loads (2) to lose RBCCW cooling.

1. RBED sump HX 6. DWED sump HX

2. FPC HX 7. RWCU non-regenerative HX

3. Recirc pump seal coolers 8. Recirc pump motor coolers

4. RR system sample coolers 9. DW coolers

5. RWCU pump seal coolers (1 ) (2)

A. 47 psig; 3,6,8,9.

B. 57 psig; 3,6,8,9.

C. 47 psig; 1,2,4,5,7.

D. 57 psig; 1,2,4,5,7.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: D Explanation: a. Part (1) is incorrect. The pressure is 10 psig below the setpoint. Part (2) is incorrect. This is a list of loads that do NOT lose cooling.

b. Part (1) is correct. Part (2) is incorrect as stated in (a) above.

c. Part (1) is incorrect as stated in (c) above. Part (2) is correct. This is the list of loads that will lose cooling.

d. Correct answer.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 2-01-70, RBCCW System (Attach if not previously provided)

( OPL 171.047, RBCCW LP Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # OPL 171.047.17 Attached New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content 55.41 X 55.43 Comments:

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Original question OPL171.047.17:

Concerning the RBCCW system, should the system discharge header pressure drop to less, than

_ _ _ , the non-essential loop isolation valve FCV would close, causing which ones of the following to lose RBCCW.

1. RBED sump HX 6. DWED sump HX

2. FPC HX 7. RWCU non-reg HX

3. DW control AlC 8. RR pump mtr & seal clrs

4. RR system sample clrs 9. DW atmos clrs

5. RWCU pump seal clrs A. 48 psig; 48; 3,6,8,9.

B. 57 psig; 47; 1,2,3,5,7.

C. 47 psig; 47; 1,2,3,4,5,7.

D. 57 psig; 48; 1,2,4,5,7.

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

(

BfN Reactor Building Closed Cooling Water 2-01-70 Unit 2 System Rev. 0059 Page 7 of 62 3.0 PRECAUTIONS AND LIMITATIONS A The Spare RBCCW Pump and RBCCW Heat Exc.hanger are common to all three (3) units. When it becomes necessary for the Spare RBCCW Pump and RBCCW Heat Exchanger to be placed in service, operators on all three units are required to communicate and be fulty aware of RBCCW conditions.

B. Whenrel110ving an RBCCW heat exchanger from service, RBCCW flow' is required to be stopped prior to stopprllg RCW or EECW cooling water flow.

C. Am} water removed from RBCCW is coos~dered potentially contaminated.

D. On low RBCCW Pump discharge header pressure (57 psig), nonessential equipment isolation valve 2'-FCV-7048 automatfcaJly doses and is required to be reopened manually using 2'-HS-70-48A This interlock is bypassed when 2-XS-I0-48 (48OV RMOV board 28, compartment SA} is placed in EMERGENCY.

E. V\t'hen pladng an R8CCW heat exchangerinsefVire, prior to opening an RBCGW heat exchanger inlet valve. placing the R8CCW SECTIONAUZING VLV TRANSFER switch at the 480V reactor MOV board 28, compartment 5A to EMERG wi~j temporarily bypass the header tow pressure auto closure for 2-FCV-70-48.

F. [NRClC] When the RCW, EECW, or RBCCW supp~ied to anyRBCCW heat exchanger is plJt into service or taken out of ser....ice. the Chemistry Laboratory Shift Supervisor is required to be notified so any required sampling can be initiated or stopped as determined by the status of RCW, EECW, and RBCCW.

[NRC lER.25!MlIS01!1]

G. [CAQIR}Cl When the RBCCW system is drained for more than 30 days, the Technical Support Supervisor is required to be notified Fn order that a specific lay-up configuration is established by the System Engineer to prevent system degradation. AU planned draining of RBCGW is required to be coordinated with the Chemistry Unit Supervisor and the Radwaste Coordinator. ~n addition, the Chemical lab should be notified of any leaks discovered which atlowsRBCCW to discharge to floor drains. !CAQR 'SFPOO0249]

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.047 page 10 of 41.

1. RBCCW Heat Loads

a. Essential loop loads

Orywell Blowers(10)

Reactor recirculation pump motor coolers (2)

Reactor recirculation pump seal coolers (2)

Orywell equipment drain sump heat exchanger (1)

b. Non-essential loop loads

Reactor Building equipment drain sump heat exchanger (1 )

Reactor water cleanup pump seal water coolers and bearing oil coolers (2)

RWCU Non-regenerative heat exchangers (2)

Fuel pool cooling heat exchangers (2)

Reactor recirculation pump discharge sample cooler (1)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 295019AA1.04 Tier# 1 Ability to operate and/or monitor the following as they apply to a 1 Group #

Partial or Total Loss of Instrument Air: Service Air isolation valves.

INOTE: Instrument Air at BFN is referred to as Control Air.1 KIA # 295019AA 1.04 Importance Rating 3.3 3.2 I Proposed Question: RO # 46 Units 2 and 3 are operating at 100% power when a leak develops in the Control Air header, causing pressure to lower slowly. All available compressors are in service.

Which ONE of the following statements describes the response of the Service Air System?

Service Air to Control Air Crosstie Valve (0-FCV-33-1) will OPEN at (1) and will (2) when Control Air pressure drops below 30 psig.

(1 ) (2)

A. 70 psig fail open B. 70 psig fail closed C. 85 psig fail open D. 85 psig fail closed

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: D Explanation: a. Part (1) is incorrect. 70 psig is the Control Air header low pressure alarm set point. Part (2) is incorrect. O-FCV-33-1 fails closed on low pneumatic pressure.

b. Part (1) is incorrect as stated in (a) above. Part (2) is correct.

c. Part (1) is correct. Part (2) is incorrect as stated in (a) above.

d. Correct answer.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 0-01-33, Service Air System (Attach if not previously provided)

OPL 171.054, Control Air LP Proposed references to be provided to applicants during examination: None Question Source: Bank# OPL 171.054.17 attached Modified Bank #

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Original question OPL 171.054.17:

Units 2 and 3 are operating at 100% power when a leak develops in the control air header, causing pressure to depressurize slowly. All available compressors are in service.

Which ONE of the following statements describes the operation of the Service Air to Control Air Crosstie Valve (33-1)?

A. The valve will open at 90 psig and closes when air pressure drops to 15 psig.

B. The valve will open at 85 psig and closes when air pressure drops to 30 psig.

C. The valve will open at 80 psig and remains open until air pressure is restored to above 90 psig.

D. The valve will open at 65 psig and closes when air pressure rises above 85 psig.

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet BFN Unit 0 Service Air System I~~

Rev. 0064 Page 10 of 94 3.0 PRECAUTIONS AND UMiTAJIONS (coomued)

D. Service Air Isolation Valve, O-FCV-33-1 will open on low Control Air SYstem pressure. This provides a service air backup to the Contrcl Air System.

E. During a loss of Service Air the Amertap system may release 1he c.ondenser tube cleaning bams to the river.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.054 page 27 of 72:

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(a) Service air supply valve from control air header (O-FCV-33-1). Can be operated from panel 1-9-20 and/or 3-9-20. The switch positions are CLOSE-AUrO-OPEN, with position indication lamps just above each control switch. The valve automatically opens if control air pressure falls to 85 psig and closes at ~ 30 psig (due to insufficient air pressure to keep the valve open).

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 295021G2.2.36 Tier # 1 Ability to analyze the effect of maintenance activities, such as 1 Group #

degraded power sources, on the status of limiting conditions of operations: Loss of Shutdown Cooling KIA # 295021 G2.2.36 Importance Rating 3.1 4.2 I Proposed Question: RO # 47 Unit 3 is in Cold Shutdown with the following plant conditions:

Both reactor recirc pumps are removed from service for maintenance.

RHR Loop II is in shutdown cooling with 3B RHR pump running.

At 08:00, RHR Loop II was taken out of shutdown cooling to adjust the packing on the RHR pumps.

At 09:30, RHR Loop II was returned to the shutdown cooling mode of operation.

During this time RHR Loop I remained in standby.

At 12:00, the Unit Supervisor is informed that an RHR Loop I surveillance needs to be performed that will require declaring RHR Loop I inoperable.

Which ONE of the following describes the EARLIEST time and the MAXIMUM duration that RHR Loop I may be made inoperable for surveillance testing?

RHR Loop I may be made inoperable _ _----'(....,.1..... ) ___ and can remain inoperable (2)

REFERENCE PROVIDED (1 ) (2)

A. immediately as long as RHR Loop II is operable.

B. immediately for no longer than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

C. after 16:00 as long as RHR Loop II is operable.

D. after 16:00 for no longer than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: A Explanation: a. Correct answer

b. Part (1) is correct. Part (2) is incorrect. Since a separate entry condition is allowed for each RHR Shutdown Cooling subsystem, as long as one RHR Shutdown Cooling subsystem remains in operation, the LCO is met.

Specifically, there are TWO RHR Shutdown Cooling subsystems per RHR loop.

c. Part (1) is incorrect. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months time frame is for BOTH RHR Shutdown Cooling subsystems inoperable. Part (2) is correct since the LCO is met with RHR Loop II in shutdown cooling.

d. Part (1) is incorrect as stated in (c) above. Part (2) is incorrect as stated in (b) above.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): U3 TSR 3.4.8 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: U3 TSR 3.4.8 Question Source: Bank # 295021 G2.2.22 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown LCO 3.4.8 Two RHR shutdown cooling subsystems shall be OPERABLE, and, with no recirculation pump in operation, at least one RHR shutdown cooling subsystem shall be in operation.

NOTES---------------------------------------

1. Both required RHR shutdown cooling subsystems and recirculation pumps may not be in operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period.

2. One required RHR shutdown cooling subsystem may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for performance of Surveillances.

APPLICABILITY: MODE 4.

ACTIONS

NOTE---------------------------------------------------

Separate Condition entry is allowed for each RHR shutdown cooling subsystem.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or two required RHR A.1 Verify an alternate 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months shutdown cooling method of decay heat subsystems inoperable. removal is available for each inoperable required RHR shutdown cooling Once per subsystem. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter (continued)

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

( CONDITION REQUIRED ACTION COMPLETION TIME

(

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet RHR Shutdown Cooling System - Cold ShlSdown B 3.4.8 B 3.4 REACTOR COOLANT SYSTEM {RCS)

B 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Co4d Shutdown BASES BACKGROUNO Irradiated fuel in the shutdown reador core generates heat during #Ie decay of fission products and increases the temperature of the reac1Df coolant. This decay heat must be removed to maintain the temperature of the reador coolant S 212"F. This decay heat removal is in preparation for performing refueling or maintenance operations, or for keeping the reactor in tlle Cold Shutdown condition.

The RHR System has two loops with each loop consisting! of two motor driven pumps, two heat exchangers. and associated piping and valves. There are two shutdown cooIfllg subsystems per RHR System loop. Both loops have a cammon suction from the same recirculation loop. The four redundant. manually cootroNed shutdown cooling subsystems of the ,RHR System provide decay heat removaL Each pump discharges the reactOif coolant, after circulation through the respective heat exchanger, to the reactor via the associated recirculation loop. The RHR heat exchangers transfer heat to the RHR Service Water System. Anyone of the four RH R shutdown cooling subsystems can provide the required decay heat removal function.

(continued)

BFN-UN1T 3 B3.4-49 Revision 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet RH R Shutdown Cooling System - Cofd Shutdown B 3.4.8 BASES (continued)

APPLICABLE Decay heat removal byoperatron of the HHR System in the SAFETY ANALYSES shutdown cooling mode is not required for mitigation of any event or accident evaluated in the safety analyses. Decay heat removal is, however, an important safety function that must be accomplished or core damage could result The HHR Shutdown Cooling System meets Criterion 4. of the NRC Policy Statement (Ref. 'I).

LCO Two RHR shutdown cooling subsystems are required to be OPERABLE, and when no recirculation pump is in operation, one RHR shutdown cooling SUbsystem must be in operation.

An OPERABLE RHR shutdown cooling subsystem consists of one OPERABLE RHR pump, one heat exchanger, one RHRSW pump capable of providing cooling to the heat exchanger, and the associated piping and valves. The subsystems have a common suction source and are anowed to have common discharge piping. Since pip.ing is a passive component that is assumed not to fail, it is allowed to be common to the subsystems. In MODE 4, the RHR cross tie valve (FCV-14-46}

may be opened to allow pumps in one loop to discharge through the opposite recirculation loop to make a complete subsystem. Additionally, each shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or focal) in the shutdown cooling mode for removal of decay heat In MODE 4, one RHR shutdown cooling subsystem can provide the required cooling, but two subsystems are required to be OPERABLE to provide redundancy. Operation of one subsystem can maintain or reduce the reactor coolant temperature as required. However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is requlred.

(continued)

BFN-UNIT 3 B 3.4-50 Revlsion 0

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ES-401 Sample Written Examination Form ES-401-5 Question Worksheet RHR Shutdown Cooting System- Cold Shutdown B3.4.8 BASES LCO Note 1 permits both required RHR shutdown cooling (continued) subsystems and recirculation pumps to not beio operation for a period of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months tn an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. Note 2 allows one RHR shutdown cooling subsystem to be inoperable for performance of Surveillance tests. These tests may be on the affected RHR System or on some other plant system or component that necessitates placing the RHR System in an inoperable status during the performance. This is permrfied because the core heat generation can be low enough and the heatup rate slow enough to allow some changes to the RHR swsystems or 01her operations requiring RHR flow interruption and loss of redundancy.

APPUCABIUTY In MODE 4, the RHR Shutdown Gooling System must be OPERABLE and shall be operated in the shutdown cooling mode to remove decay heat to maintain coolant temperature below 212"F. Otherwise, a recirculation pump is required to be in operation.

In MODES 1 and 2, and in MODE 3 with reactor steam dome pressure greater than or equal to the RHR low pressure permissive pressure, this LCO is not applicable. Operation of the RHR System in the shutdown roofing mode is not allOwed above this pressure because the RGS pressure may exceed the design pressure of the shutdown cooling piping. Decay heat removal at reactor pressures greater than or equal to the RHR low pressure permissive pressure is typicalty accomplished by condensing the steam in the main condenseL Additionally, in MODE 2 below this pressure, the OPERABILITY requirements for the Emergency Core Cooling Systems (ECCS) (LCO 3.5.1, "ECeS-Operating") do not al'low placing the RHR shutdown cooling subsystem into operation.

(contrnued)

BFN-UNIT 3 B 3.4-51 Revision 0

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/'

REFERENCE MATERIAL Provided to CANDIDATE

3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown LCO 3.4.8 Two RHR shutdown cooling subsystems shall be OPERABLE, and, with no recirculation pump in operation, at least one RHR shutdown cooling subsystem shall be in operation.

NOTES---------------------------------------

1. Both required RHR shutdown cooling subsystems and recirculation pumps may not be in operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period.

2. One required RHR shutdown cooling subsystem may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for performance of Surveillances.

APPLICABILITY: MODE 4.

ACTIONS

NOTE---------------------------------------------------

Separate Condition entry is allowed for each RHR shutdown cooling subsystem.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or two required RHR A.1 Verify an alternate 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months shutdown cooling method of decay heat subsystems inoperable. removal is available for each inoperable required RHR shutdown cooling Once per subsystem. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter (continued)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.8.1 Verify one required RHR shutdown cooling 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months subsystem or recirculation pump is operating.

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

( Examination Outline Cross-reference: Level SRO 288000G2.2.44 Tier # 2 Ability to interpret control room indications to verify the status and 2 Group #

operation of a system, and understand how operator actions and directives affect plant and system conditions: Plant Ventilation. KIA # 288000G2.2.44 Importance Rating 4.2 4.4 I Proposed Question: RO # 36 Unit 2 is operating at 100% rated power during the summer with the following conditions:

Control Bay Chiller "A" breaker indicates closed and reading 32 amps on Panel 9-20.

Control Bay Chiller "B" breaker indicates closed and reading 0 amps on Panel 9-20.

Control Bay chilled water inlet temperature is indicating 42 OF on the ICS computer.

Which ONE of the following describes the effect of transferring the CONTROL PANEL MODE SELECT switch from "DIGITAL" to "ANALOG" on the CONTROL BAY CHILLER A LOCAL DISPLAY PANEL?

Control Bay Chiller "A" will _ _ _ _ _ _ _ __

A. continue to run, but chilled water inlet temperature will lower to 40 OF.

B. continue to run, but will no longer supply performance data to the ICS computer.

C. immediately trip, causing a loss of chilled water to the Control Bay Air Handling Units.

D. immediately trip, then will automatically restart in the Analog Mode.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: C Explanation: a. Incorrect. The chiller will trip, however the chilled water temperature setpoint while operating in Analog mode is 2 of lower at 40 of.

b. Incorrect. The chiller will trip, however the Chiller will not supply digital data to ICS while running in Analog Mode

c. Correct answer.

d. Incorrect. The chiller will trip as stated in (c) above. The chiller must be manually shutdown prior to transferring to Analog control, then must be manually restarted.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 0-01-31 (Attach if not previously provided)

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Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/15/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content 55.41 X 55.43 Comments:

(

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

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BFN Control Bay and Off-'Gas Treatment Il-OI-J1 Unit 0 Building Air Condjtioning System Rev. 0126 Page 16 of 283 3.2 Unit 112 Control Bay Chillers A Compressor sump heaters are required to be energized for a minimum of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to starting a chilleL This prevents compressor damage caused by eliQuid refrigerant nn the compressor at startup. When the heaters are on, the bottom end (end closest to the chiiler control pane4) of the compressors will be waml. This requirement may be modified by the System En9ineer taking into account the length of time the compressor was shutdown and the outside air temperatures.

R Prior to .any draining being performed on the Unit 112 Control Bay Chilled Water System, Chemistry and Environmental should be ootified of the planned system draini09*

C. When transferring between analog and digital control modes on a Chiller, the chiller must NOT be running. Therefore, whenever CONTROL PANEL MODE SELECT, (l-HS-031-2110M(2210BA), switches are to be manipulated, the associated ChiUeris required to be shutdown.

D. The Chiller Control Switch located on the !Iocal display panel for each Control Bay Chililer Control Panel has three positions; STOP/RESET, AUTO LOCAl, and AUTO REMOTE. The STOP/RESET position shuts down the chiller when in loca~control. The AUTO LOCAL posmoo is used to start the ch~ler when in local control. The AUTO REMOTE position is NOT used. The control switch should NOT be selected to the AUTO REMOTEoosmon.

E. The Control Bay Chilied Water Pumps A or B are tripped by load shed signal.

Pumps may be restarted after 10 minutes with the use of "CHW PUMP LOAD SHED BYPASS SW for each pump. (A PUMP "O-HS-031-2101 E" location O-lPNL-925-n165 PANEL D)(B PUMP "O-HS--031-2201E" location O-LPNL-925-0165 PANEL E) If 1 8. 2 CONT. BAY OHW PUMP A(B)

TRANSFER SWITCH, O-XS-'031-2101(2201) is in LOCAL POSITION lOad shed logic for these pumps is bypassed.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Control Bay and Off-Gas Treatment 0-01:.31 Unit 0 Build'ing Air Conditioning System Rev. 0126 Paue 55 of 283 5~16 Startup of Unit 1/2 Control Bay Chiller A 1ft l.oCBI Digital Co~

Mode (cominued)

[6] VERIFY the following switch positions: (Unit 112 Control Bay Chiller A Cootr04 Panet}:

ANALOG CONTROL MODE SELECT, o-HS-031-2110AB(HS2), in ANAlOG STOP. Oi

CONTROL PANEL MODE SELECT, o-HS-031-211OM(HS1), in DIGITAL CONTROL Oi

UNIT 1&2 CB CHillER A REMOTE HS APPENDIX R DISC SW, O-HS-031-2110 in LOCAL Oi

CONTROL BAY CHILLER A, O-BKR-031-2110, In ON.. 0;

CONTROL BAY CHillER A lOCAl DISPLAY PANEl O-PMC-031-2100A Display Window is illuminated. 0 NOTE Step 5.16[1] starts the chiller. There is approximately a 2 minute time delay between when the switch is placed in AUTOILOCAL and when the chil&er .actuajly starts.

[7} At CONTROL BAY CHILLER A LOCAL DISPLAY PANEL, O-PMC-031-2100A, PLACE swttch in AUTO LOCAL Oi

[8] CHECK. the fotklwing parameters for ChHler A:

1&2 CB CHW PUMP A INLET PRESSURE, o-PI-031-0194, is approximately 15-30 pSig. 0

1&2 CB CHW PUMP A(B) DISCHARGE PRESSURE, o-PI-031-0195, is approximately 85-105 psig. 0

On O-PMC-031-2100A (Menu-PO point Fl, Evap Leaving Water Temp evernualy lOwers to 42 +/- 2l;)F. 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Control Bay and Off-Gas Treatment O-Ot-J1 Unit 0 Building Air CODditiOfling System Rev. 0126 Page 63 01 283 5.18 Startup of Unit 1/2 Control Bay Chiller A In Analog Control Mode (continued)

NOTE Step 5.18(8] starts the chiller at (Unit 1.12 Control Bay Chiler A Control Panel). There is approxim~ a 5 minute time delay between when the switch is placed in AUTO and when the chiller actuallY starts.

[8] PLACE ANALOG CONTROLlMODE SELECT, O-H8-031-2110AB(HS2}. in ANALOG START. 0

[9] RECORD switch manipulations in Steps 5.18[7] and 5.18[8] in Nammv*e Log. 0

[10] CHECK. tile following parameters on Chiller A:

1&2 CB CHW PUMP A :INLET PRESSURE, o-Pi-031-0194, is approXBmately 15-30 psig. o

1&2 CB CHW PUMP A(B) DISCHARGE PRESSURE, o-PI-031-0195, is approximately 85-105 psig.

VERIfY Evaporator Leaving water Temp eventually lowers to 40 +/- 5°F on 1&2C8 CHILLER A CHW OUTLET TEMP IND, O-T1-031-0023. 0

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

( Examination Outline Cross-reference: Level SRO

\

295023AK3.03 Tier# 1 Knowledge of the reasons for the following responses as they apply 1 Group #

to Refueling Accidents: Ventilation isolation.

KIA # 295023AK3.03 Importance Rating 3.3 3.6 I Proposed Question: RO # 48 Unit 1 is performing a fuel pool cleanout when a failure of the Main Grapple hooks result in an irradiated fuel bundle being dropped into the reactor vessel.

Unit 1 Ventilation Radiation Monitors read as follows:

- Channel A Reactor Zone Detector A 1-RM-90-142A 100 MR/HR Reactor Zone Detector B 1-RM-90-142B 100 MR/HR Refuel Zone Detector A 1-RM-90-140A 40 MR/HR Refuel Zone Detector B 1-RM-90-140B 60 MR/HR

- Channel B Reactor Zone Detector A 1-RM-90-143A 45 MR/HR Reactor Zone Detector B 1-RM-90-143B 62 MR/HR Refuel Zone Detector A 1-RM-90-141A 100 MR/HR Refuel Zone Detector B 1-RM-90-141B 100 MR/HR Based on the above conditions, which ONE of the following describes the response of the plant ventilation system?

Refuel Zone Supply and Exhaust fans on _ _ _ _ _~C1=..J-)----_ and Reactor Zone Supply and Exhaust fans on _ _ _ _-----'C.=2.J-)_ _ _ _ __

(1) (2)

A. Unit 1 only will trip all three units will trip.

B. Unit 1 only will trip Unit 1 only will trip.

e. all three units will trip. all three units will trip.

D. all three units will trip. Unit 1 only will trip.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: 0 Explanation: a. Part (1) is incorrect. All Refuel Zones are tied together and will trip in high radiation. Part (2) is incorrect. Reactor Zone ventilation is separate for each unit. Only the effected unit's RB ventilation will trip.

b. Part (1) is incorrect as stated in (a) above. Part (1) is correct.

c. Part (1) is correct. Part (2) is incorrect as stated in (a) above.

d. Correct answer.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): OPL 171.067, HVAC (Attach if not previously provided)

U1 TSB Section 3.3.6.2 Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # 295023AK2.05 Attached New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: This question was considered MODIFIED due to the original question being incorrect. It implied that ONLY Unit1 Refuel Zone trips when all three units would trip. The correct answer was derived by elimir)ating the three distracters due to incorrect statements. That made the required answer the "most correct" choice rather than "the" correct choice.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Original question: 295023AK2.05 Unit 1 is performing a fuel pool cleanout when a failure of the Reactor Building overhead crane results in an irradiated LPRM string raised above fuel pool water level.

Unit 1 Ventilation Rad Monitors read as follows:

- Channel A Reactor Zone Detector A 1-RM-90-142A 100 MR/HR Reactor Zone Detector B 1-RM-90-142B 100 MR/HR Refuel Zone Detector A 1-RM-90-140A 40 MR/HR Refuel Zone Detector B 1-RM-90-140B 60 MR/HR

- Channel B Reactor Zone Detector A 1-RM-90-143A 45 MR/HR Reactor Zone Detector B 1-RM-90-143B 62 MR/HR Refuel Zone Detector A 1-RM-90-141A 100 MR/HR Refuel Zone Detector B 1-RM-90-141B 100 MR/HR Based on the above conditions, which ONE of the following describes the response of plant ventilation systems?

A. Unit 1's Refueling AND Unit 1's Reactor Zone Supply and Exhaust Fans trip.

B. All Refueling AND Unit 1's Reactor Zone Supply and Exhaust fans trip and both CREV units start.

C. All Reactor Zone Supply and Exhaust fans trip and the CREV System starts and all SGTS fans start.

D. All SGTS fans start, the preferred CREV unit and the Unit 1/3 Board Room Emergency Supply Fans start.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.067 page 18 of 71:

1. System Isolation (Group 6)

a. The isolation signals are low reactor water level +2", high drywell pressure 2.45 psig, and high radiation in exhaust duct (72 MR/hr Refuel zone or Reactor zone).

b. On auto isolation signal (except Refuel Zone high radiation) the unit reactor zone supply and exhaust fans trip, all refuel zone supply and exhaust fans trip, unit reactor zone and all refuel zone supply and exhaust isolation dampers close, and dampers to Standby Gas Treatment System open and SGT train blowers auto start.

c. Note: Damper logic is as follows: PCIS Group 6 with A SGT running opens 64-4-1 and 45; PCIS Group 6 with B SGT running opens 64-40 and 44.

d. On isolation due to high radiation in Refuel Zone all refuel zones isolate, refuel zone supply and exhaust fans trip, SGT starts and aligns dampers for refuel zone only.

Excerpt from OPL171.017 page 22 of 56:

n. Reactor Building Ventilation Exhaust Radiation High

Reactor Building ventilation exhaust radiation is monitored by two sets of two (four detectors total) gamma-sensitive radiation monitors located on the reactor building ventilation exhaust duct.

One channel at the high trip setpoint or two channels in a combination of downscale and/or inop will cause isolation.

o. Refuel Zone Ventilation Exhaust Radiation High

Refuel zone ventilation exhaust radiation is monitored by two sets of two (four detectors total) gamma-sensitive radiation monitors located on the refuel zone ventilation exhaust duct.

One channel at the high trip setpoint or two channels in a combination of downscale and/or inop will cause isolation.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Secondary Containment Isolation Instrumentation 83.3.6.2 BASES APPUCABLE 2. Drywell Pressure - H~gh (PIS~64-56A-D) (continued)

SAFETY ANALYSES, LCO, and The Allowable Value was chosen to be the same as the ECCS APPLICABILITY Drywell Pressure - High Function Allowable Va.lue (LeO 3.3.5.1) since this is indicative of a loss of coolant accident (LOCA).

The Drywell Pressure - High FunctiiOn is required to be OPERABLE in MODES 1. 2. and 3 where considerable energy exists in the RCS; thus, there is a probability of p~pe breaks resulting in significant re~eases of radioactive steam and gas.

This Function is not required in MODES 4 and 5 because the probability and consequences of these events are low due to the Res pressure and temperature limitations of these MODES.

3, 4. Reactor Zone Exhaust and RefuelLng Floor Radiation -

High (RM-90-140, 141,142,.143)

High secondary containment exhaust radiation is an indication of possible gross failure of the fuel cladding. The release may have originated from the primary containment due to a break in the Reps. When Exhaust Radiation - High is detected, secondary containment isolatj;on and actuation of the SGT System are initiated to limit the release of fission products as assumed in the FSAR safety analyses (Ref. 4 ).

The Exhaust Radiation - High signals are initiated from radiation detectors located on the reactor zone ventilation exhaust and the common refueling zone. There are two radiation monitors and two divisional trip systems for each unit (Units 1, 2, and 3). Each monitor has one channel of Reactor Zone Exhaust Radiation - High and one channel of Refueling Floor Radiation - High. Each monitor's channels provide signals to its associated divisional trip Systertl_ Each channel has two radiation elements which monitor the

( continued)

BFN-UNIT 1 83.3-228 Revision ~35 February 14, 2006

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Secondary Containment Isolation Instrumentation B3.3.6.2 BASES APPLICABLE 3. 4. Reactor Zone Exhaust and Refueling Floor Radiation -

SAFETY ANAl..YSES, High (RM-90-140, 141. 142, 143) (continued)

LCO,and APPlIICABILlTY ventilation exhaust hoUl of which must be OPERABLE or tripped fa hi! channel to be OPERABLE. Bo4h radiation elements must provide a High signal to trip the assoaated channel (two-out-of-two). However, the oo1put refays from the divisional trip systems are arranged in logic systems such that if either channel for a zone trips, a secondary containment isolation signal is initiated (one-out-of-two). Six channels of Reactor Zone Exhaust Rad~tion - High Function and six channels of Refueling Floor Radiation - Higil Function are available (two channels of each Function from each unit) and are required to be OPERABLE to ensure that no single instrument failure can preclude the isclation function.

The Allowable Values are chosen to provide timely detection of nuclear system process banier leaks inside containment but are far enough above background levels to avoid spurious isolation.

The Reactor Zone Exhaust and Refueling Floor Radiation -

High Functions are required to be OPERABLE in MOOES 1, 2, and 3 where considerable energy exists; thus, there is a probability of pipe breaks resulting. *in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES; thus, these Functions are not required. In addition, the Functions are also required to be OPERABLE during OPORVs because the capability of detecting radiation releases due to 1'l.e failures (due to fuel uncovery) must be provided to ensure that offsite dose limits are oot exceeded.

(continued)

BFN-UNIT 1 B 3 ..3-229 Revision g, 21, 29, 35 February 14, 2006

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 29S024EA1.07 Tier # 1 Ability to operate and/or monitor the following as they apply to High 1 Group #

Drywell Pressure: PCIS/NSSSS.

KIA # 295024EA 1.07 Importance Rating 3.8 3.9 I Proposed Question: RO # 49 Unit 2 is at 100% rated power with the following conditions:

DRYWELL NORM OPERATING PRESS HIGH (9-3B W 19) in alarm.

DRYWELL PRESS APPROACHING SCRAM (9-3B W 30) in alarm.

DRYWELL PRESSURE ABNORMAL (9-5B W 31) in alarm.

DRYWELL FD SUMP LEVEL ABN (9-4C W 2) in alarm.

Drywell venting is in progress using 2-AOI-64-1, "Drywell Pressure and/or Temperature High, or Excessive Leakage into Drywell. II

Drywell pressure is 2.2 psig and steady.

Assuming no further operator action, which ONE of the following describes the plant response if 480V Shutdown Board 2A de-energized due to an electrical fault?

The Drywell vent lineup would be _--->o.C:=..J1)'---_. Drywell pressure would _ _ _---'C-.=2<+-)_ _

(1) (2)

A. unaffected lower due to non-essential RBCCW loads isolating.

B. unaffected rise due to RPS A de-energizing.

c. isolated lower due to non-essential RBCCW loads isolating.

D. isolated rise due to RPS A de-energizing.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: D Explanation: a. Part (1) is incorrect. RPS A would de-energize, causing the vent lineup to isolate, since RPS supplies PCIS isolation valve logic which would fail closed on loss of power. Part (2) is incorrect. Under normal circumstances, DW pressure would lower due to non-essential loads isolating on low RBCCW discharge pressure. This acts to increase cooling flow to the drywell blowers. However, with the vent lineup isolated and a leak in the drywell, pressure would begin to rise again.

b. Part (1) is incorrect as stated in (a) above. Part (2) is correct. With the vent lineup isolated and a leak in the drywell, pressure would begin to rise again.

c. Part (1) is correct. RPS A would de-energize, causing the vent lineup to isolate. Part (2) is incorrect as stated in (a) above.

d. Correct answer.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 2-AOI-64-1, ARPs 9-58, 9-38 (Attach if not previously provided) 2-AOI-99-1 Proposed references to be provided to applicants during examination: None Question Source: 8ank #

Modified 8ank # (Note changes or attach parent)

New 09/03/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Drywell Pressure and/or Temperature 2-A01-64-1 Unit 2 High. or Excessive leakage into Rev.D02l DrvweI . Paoe5of12 2~ Symptoms for High 0ryweI1 Temperature

DRYWEll NORM OPERATING PRESS HIGH (2-XA-55-3B. Window i9)

., Drywefl temperature rising, as indicated on DRYWELl TEMPERATUREIPRESSURE. 2-XR-064-050 (Panel 2-9-3)

., OrytNelpressure rising, as indicated on DRYWELl TEMPERATUREJPRESSURE, 2-XR-064-050 (Panel 2-9-3) 2.4 Symptoms for Drywell Excessive leakage

., DRYWEll NORM OPERATING PRESS HIGH (2-XA-55-3B, Windtwtl19)

ORYWEll FD SUMP LEVEl ABN (2-XA-55-4C, Wmdcw 2)

DRYWEll EQPT DR SUMP LEVEL ABN (2-XA-56-4C, Window 9)

RBCCW SURGE TANK LEVEL LOW (2-XA-55-4C,Wmdow 13)

., ORYWELl EQPT DR SUMP TEMP HIGH (2-XA-55-4C, Window 16}

REACTOR WATER LEVEL ABNORMAl. (2-XA-55-5A, Window 8)

RECIRC PUMP A NO. :2 SEAL LEAKAGE HIGH 2-FA-68-55 (2-XA-55-4A, Window 18)

RECIRC PlJMPA NO.1 SEAl LEAKAGEABN 2-FA-68-62 (2-XA-5&-4A, Window2:5)

., RECIRC PUMP B NO. :2 SEAL LEAKAGE HIGH 2-FA-68-6B (2-XA-55-4B, Window 18)

RECIRC PUMP B NO.1 SEAl LEAKAGEABN 2-FA-68-14. (2-RA-55-4.B.

Window 25)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BfN 2-XA-55-JB Unit 2 SensorfTfip Point DRYWEllNORM 2-PS-64-135 OPERATING or 1.6 psig.

PRESS HIGH 2-PA-64-135 2-PS-64-136 (Page 1of1)

Sensor Panel 9-19 location: Elevation 593' Auxiliafy Instrument Room Probable A. Excessive N., makeup.

cause: 8. Low pressure front moving throtIgh area.

C. DJyweI OP ~ ma!IfuncOOn.

D. DrywetI blowers failure.

E Steam or water leak inside drywel.

F. PossiIbIe DIywe/1 Cootrol.A.ir System In leakage.

G. sensor malfmdion.

H. Reactor start-up.

A. CHECK dryweilpressure and temperature for rise AND CHECK weather report for atmospheric pressure. []

B. CHECK to see if [)rywell OP Compressor is running. IF ~ DP

~ is runnilg, THEN STOP t'OIll're5SOf. []

C. CHECK N.2 makeup valves to S~ Chamber and [)rywell closed. []

O. CHECK iDrywei CooIroI Air System Row Elements 2-F~32~ (I<< Bldg 565' R10-S) and 2-FIQ-032-0015 (I<< Bldg 565' R20-lD) <: 1.7 SCFM. []

E IF pressure rise is due to norrnaII startup, THEN REFER TO 2-01-64 for IlOITTIll venting instructions. []

F. IF Orywell pressure is high, THEN REFER TO 2-AOf-64-1. []

2-45E711-21

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet 2-XA-55-3B L-ARP-9-3B Rev. 0020

". . ."." ........ """...""""."" . . "... ...................... ..... ...... .. ;.p!9!;!~.Qf.~ ...................... .

SensortTrip Point:

DRYWELL PRESS P1S-64-58E APPROACHING P1S-64-58F t.96psig SCRAM Pl~64-58G 2-PA-64-58 P1S-64-58H RED BAR f"3O (Page 1 of 1)

Sensor PIS-64-58E, 580, PIs-.64-58F,58H location: Aux Inst. Rm. AuxlnstRm.

Panei 9-81 Panel 9-82 Probable A. DfYW11 pressure rising.

CauSe: S. Dl)"Nejj cooler(s) failure.

C. steam or water leak inside Dry'Welt O. Loss of RBGCW to Drywell coolers.

Automatic Nrn1e Action:

Operator A. CHECK rontaimnent pressure ami temperature using muHipJe Action: indications. o

8. REFER TO 2-AOI-64-1. o Refereoces: 45N620-3 2-47E61 0-64-1 47W600-57

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

(

Panel 9-5 i 2-ARP 58 2-XA-55-58 ,Rev. 0025

" ......"L~il~~,~f'9" SensorfTrip Point ORYWEll 2-PS-64-56E 1.98 psig rising PRESSURE 2-PS-64-56F O. f psig*lowefing ABNORMAL 2-PA-64-56 (Page 1011)

Sensor Panel 25-5B location: Elevation 593 Cot R-3 S-UNE Probable A Drywell OP air cornpressor failure.

Cause: S. Loss of RBCCW.

C. Breach of PrirnalY Containment.

1. Dsywell vent valves open or leaKing.

2. Dsywe/I varnum breai\er open or leaking.

O. LOCA E Sensor malfunction.

Automatic NOne Action:

Operator A. VERIFY alarm using multiple iooications. o Action: B. IF RBCCW has been lost, THEN REFER TO 2-AOl-70-1. o C. REFER TO 2-AOI-64-1. o 2-45iE620-6 2-47E610-64-1 2-730E915-17 2-AOI-70-1 2-AOI-64-1

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Loss of Power to One RPS Bus 2-.401-99-1 Unit 2 , Rev. 0025 Page40f3 J.a AUTOMA HC ACTIONS NOTE An overview of automatic actions for RPS Bus A(B) is provided here_ A de1a~ed list of actions is provided in 2-01-99, laustration 1, which Nsts actions Ihal occur when RPS buses are de-energized on a transfer of power supply.

A RPS trip logic A(8) half-scram occurs.

B. PCIS Group 1 half-trip logic de-energizes.

C. PCIS Group 2 isolation, RHR Shutdown Coolin;g Mode:

1. Bus A inboard.

2. Bus B outboard.

0_ PCIS Group 3 isolation, RWCU:

1. Bus A inboard and outboard.

2. Bus B outboard.

E. PCIS Group 6 isolation, Primary Containment Vent and Purge and Reactor Buiiding Ventilation:

1. Bus A or B inboard and outboard.

F. Group 8 isolation, TIP.

G. Control Room Emergency 'Ilentilation System start.

H. Standby Gas Treatment System starts.

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

( Examination Outline Cross-reference: Level SRO 29S02SG2.1.31 Tier# 1 Ability to locate control room switches/ controls and indications and 1 Group #

to determine that they are correctly reflecting the desired plant lineup: High Reactor Pressure. KIA # 295025G2.1.31 Importance Rating 4.6 4.3 I Proposed Question: RO # 50 Given Unit 1 at 100% rated power:

Which ONE of the following describes the unit response if the Max Combined Flow Limit setting was inadvertently reduced from 125% to 75% over two minutes?

Main Generator load would _ _~(1!:..J.)_ _ _ _' Main Turbine Bypass Valves would

_ _ _.....,(2'::.J,)'---_ _ and reactor pressure would _ _ _----l.(>L,3)t-_ __

(3) remain the same rise

c. remain the same remain closed remain the same D. lower* open rise

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

( I Proposed Answer: B Explanation: a. Part (1) is incorrect. Generator load will remain the same until the MCFL setting drops below 100%. At that point the TCVs will begin to close. Part (2) is incorrect. If the Generator Load Set setting was being lowered, this would be correct. Part (3) is incorrect, but is consistent with the expected response if the Generator Load Set setting was being lowered.

b. Correct answer.

c. Part (1) is incorrect as stated in (a) above. Part (2) is correct. Part (3) is incorrect but is consistent with the expected response as the MCFL is lowered from 125% to 100%. Once the MCFL is lowered past 100%, the conditions in (b) would occur.

d. Part (1) is correct as the MCFL lowers below 100%. Part (2) is incorrect as stated in (a) above. Part (3) is correct for the given conditions but is not consistent with Bypass Valves opening.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 1-01-47, OPL 171.228 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/04/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

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

(

BFN Turbine-Generator System 1-01-41 Unit 1 Rev. 0013 Page 12 of 220 3.0 PRECAUTIONS AND LIMITATIONS A A Turb~ne-Genef3tor trip is NOT to be reset before the cause of the trip is dearly established and correcfive actions considered.

B. If the hydrogen seal oil is Iiost. the Turbine is to be tripped ;md the hydrogen dumped and the Generator purged immediatety in accordance with 1-01-35.

C. The Turbtne is to be tripped if stator coolant conducfivity exceeds 9'.9 JUfllho.

O. Do NOT select a set speed which is lower than current Turbine speed. If deceleration is desired, trip the Turbina ;md ~ace ootummg gear.

E. The EHC Control System can be used in either Reactor Pressure control or Header Pressure control. While in Header Pressure control, a single header pressure input failing high could cause the bypass valves to open. While in Reactor Pressure control, a singfe Reactor Pressure mput failing high will NOT affect the bypass valves. For this reason, Reactor Pressure rontrol is the preferred mode of operation for the EHC Control System.

F. The CO LR Thermal Limit analysis allows for a Turbine Bypass Valve andlor the Recifc Pump Trip to be out of service. Themfore, the EOG-RiPT logic can remain disabled should a Turbine Bypass Valve become inoperative. The Unit 1 TRM COLR should be referred to for the appropriate Thermal limits and off-rated corrections when either Turbine Bypass out-of-service conditions exist or when the Recirculation Pump Trip is out-of-service.

G. The following pertain to the Max Combined Flow limit:

1. Max combined f!low limit setting of 150% (upper limit) precludes exceeding thermallmils during a single Turbine control valve dosure.

2.. The max combined flow upper and lower setting limits are 50% and 150%.

Normaly it is set at 125%.

3. The max combined flow limit setting is adjustable only on the EHC WORK STATION computer (Paneils 1-9-7 and 1-9-31).

4. Max combined flow limit setpoint can be found on the folliowing computer screens:

liI. On ICS. EHC TURBINE CONTROL (EHCTC) screen.

b. On EHC WORK STATION, TURBINE CONTROL screen.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.228 page 13 of 81:

a. 'The Maximum Combined Flow Limiter Obj. V.B.3 limits the valve opening for the control 50-150% setting valves and bypass valves and is only , 01-47 sets this at adjustable through the EHC 125% flow demand

, Workstation.

b. In the event the maximum combined flow limit becomes active, the MAX CF indicator will illuminate at the MCR panel and the EHC Workstation.

c. All of the above parameters are low signal The low limit and selected on a low signal select block. The high limit values are output of the low signal select block will set for 0% and be the lowest input value provided the 100%.

value is not higher than the high limit or lower than the low limit.

d. The lowest input to the low signal select bus and the parameter that is in control is illuminated on the MCR panel and the EHC workstation.

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

( Examination Outline Cross-reference: Level SRO 295026EK3.01 Tier # 1 Knowledge of the reasons for the following responses as they apply 1 Group #

to Suppression Pool High Water Temp: Emergency/Normal depressurization. KIA # 295026EK3.01 Importance Rating 3.8 4.1 I Proposed Question: RO # 51 Which ONE of the following describes the Suppression Pool temperature limit and bases for injecting Standby Liquid Control into the reactor?

Standby Liquid Control must be injected prior to exceeding (1) 0 F. This is based on (2) prior to Emergency Depressurization being required due to high Suppression Pool temperature.

(1) (2)

A. 95 ensuring the reactor is subcritical B. 95 maintaining Suppression Pool pH

c. 110 ensuring the reactor is subcritical D. 110 maintaining Suppression Pool pH

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: C Explanation: a. Part (1) is incorrect. This is the entry condition for Primary Containment Control based on SP temperature. Part (2) is correct.

b. Part (1) is incorrect as stated in (a) above. Part (2) is incorrect for the given conditions. SLC injection is required for pH control of the Suppression Pool but is based on Drywell high radiation, not Suppression Pool high temperature.

c. Correct answer.

d. Part (1) is correct. Although "prior to 95 0 F" is also "prior to 110 0 F",

there is no procedure guidance to inject SLC before 95 0 F based on SP temperature. Part (2) is incorrect as stated in (b) above.

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet Technical Reference(s): EOIPM SECTION O-V-C page 116 (Attach if not previously provided) 1-EOI-1 flowchart, OPL 171.203 Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/04/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part,55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.203 page 42 of 66.

1. Step SP/T-4

a. This before decision step has the operator evaluate current and future efforts to reduce suppression pool temperature, in relation to the current value and trend of suppression pool temperature, to determine if a reactor shutdown is necessary.

b. The before decision step requires that this determination and subsequent actions be performed before suppression pool temperature reaches 11O o P, the temperature at which boron injection would be required if the reactor was not sub critical.

c. Calculations have determined that if suppression pool temperature reaches 11 oop before boron injection is initiated, there is no assurance that the reactor will be subcritical when emergency RPV depressurization is required due to exceeding the Heat Capacity Temperature Limit.

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

(

EXECUTE RCi0-9, RC/0-1O, AND RC/Q-20 CONCURRENTLY RC/Q-8 WHEN PERIODiC APRM OSCillATIONS GREATER THAN 25% PEAK-TO-PEAK PERS,ST L

RClQ-9 BORON IN.! UI REQUIRED (EO 1-1 , ReiN))

L RCIQ-11 INITIATE SLC (APPX 31'.)

L RC/Q-12 INHIBIT ADS

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet BFN Panel 9-7 1-ARP-9-7C Unit 1 1-XA-5~7C Rev. 0017

!____.__ ~.. _.~~_._.._~w_.~j____ ***_ .* _______ ._ _ _.__ ** _ **** _** _ . _________ ._____________._*..*._. __ *___ ************ _....::p._!lJI!_~_~!!!. ________ . ._. __~._.

SensorJTnp Point:

ORYWELL RADIATION HIGH 1-RE-90-212A 11] RlHR 1-RE-90-213A 10 RlHR 1-RA-'90-272 1-RE-90-2128 No set point (later)

SOLID MAGENTA f"'1'5 1-RE-90-273B No set point (later)

(Page 1 of 1)

Sensor 1-RM-90-272A, Panel 1-9-54 lociltioo: 1-RM-90-273A, Panel 1-9-55 1-RM~90-2728, Panel 1-9-54 1-RM-:90-273B, Panel 1..;9-55 Probable A. Noise spikes.

Cause: B. High radiation (post accident monitor).

Automatic None Action:

Operator A. VERIfY alarm on 1-RR-90-272 on Panel 1-9-54 and 1-RR-90-273 Action: on Panel 1-9-55. o B.. CHECK 1~R-90-256 for rising indication. o C. ATTEMPT to isolate equipment*to stop source. o D. IF the alann is determinee to be valid, THEN, PERFORM the following within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months at alarm:

OPEN UPSTREAM MSL DRAJN TO CONDENSER 1-FCV-001-OO58. o

OPEN DOWNSTREAM MSt. DRAIN TO CONDENSER 1-FCV-001-OO59. o

ENSURE 1-PCV-001-0147 is Closed bytakirlg STEAM SEAl REGULATOR, 1-HS-1-147 to CLOSE. (Panel1-9-7} o

CLOSE MAIN STEAM TO OG PREHEATER 1ft,. (SOL, 1-SHV-001-0741.(SJAE RM B) o

CLOSE MAIN: STEAM TO OG PRE HEATER 1 B ISOl, 1-SHV-001-0743.(SJAE RM B) o Continued on Niext Page

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet BFN Panel 9-1 1-ARP-9-7C Unit 1 1-XA-55-7C Rev. 0011 21 E. IF All the fOllowing conditions exist

Alarm is determined to be valid. []

The reactor will remain sudClitical without boron injection under all conditions []

Leakage of primary coolant into primary containment is indicated []

THEN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of alarm, INJECT SlC for aHemate soorce term control by pladng SlC PUMP 1N1B, 1-HS-63-6A in the START A OR START B position. []

F. REFER TO EPIPs. []

G. IF started at Operator Action step 5, THEN WHEN SlC tank reaches "0", STOP the running SLC Pump. 0

References:

1-45E620-9-1, 2 0-47E61 0-90-2 Technical Specifications 3.3.3.1

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 29S028EK2.03 Tier # 1 Knowledge of the interrelations between High Orywell Temperature 1 Group #

and Reactor Water Level indication.

KIA # 295028EK2.03 Importance Rating 3.6 3.8 I Proposed Question: RO # 52 A LOCA has occurred on Unit 2 resulting in the following conditions:

2-EOI-1, "RPV Control" and 2-EOI-2, "Primary Containment Control" are being executed

Drywell Sprays could not be initiated due to logic failures.

Drywell pressure 15 psig and slowly rising.

Drywell temperature 305 of and steady.

Suppression Pool level 15.5 feet.

Suppression pool temperature 140 of and steady.

ADS was manually initiated due to high Drywell temperature.

The six ADS valves have now closed due to low reactor pressure

Normal range level indicates (+) 34 inches.

Emergency range level indicates (+) 58 inches.

Shutdown Floodup level indicates (+) 30 inches.

Which ONE of the following describes the current status of RPV level instrumentation?

Reactor water level (1) be determined. The Shutdown Floodup instrument (2) be used for trend indication.

REFERENCE PROVIDED (1) (2)

A. CAN CAN B. CAN CANNOT

c. CANNOT CAN D. CANNOT CANNOT

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: B Explanation: a. Part (1) is correct. Although the "Action Required" region of Curve 8 has been entered, no indication of erratic level instruments have been given in the stem. Caution 1 part B states instruments MAY be unreliable. Part (2) is incorrect. The Shutdown Floodup level indicates below the Minimum Indicated Level on Caution 1 which implies actual level may be below the variable leg instrument tap.

b. Correct answer.

c. Part (1) is incorrect. Both Normal and Emergency range level indicators are reliable under the given conditions. Part (2) is incorrect as stated in (a) above.

d. Part (1) is incorrect as stated in (c) above. Part (2) is correct. The Shutdown Floodup level indicates below the Minimum Indicated Level on Caution 1 which implies actual level may be below the variable leg instrument tap.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 2-EOI-2 Curve 8 and Caution 1 (Attach if not previously provided)

OPL 171.201 Proposed references to be provided to applicants during examination: EOI Curve 8 and Caution 1 Question Source: Bank #

Modified Bank # OPL 171.203.82 Attached New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Original question OPL171.203.82:

(

A LOCA coupled with an inability to spray the drywell has resulted in the operators Emergency Depressurizing the reactor.

The following conditions are present:

DW pressure 15 psig and slowly rising DW temperature 305 degrees and steady Suppression Pool level 15.5 feet Suppression pool temperature 140 degrees and steady The six ADS valves have just closed due to low reactor pressure The normal and emergency systems range level indicators are reading off-scale high The 3-55 level indicator is reading +50" EOI-1 & 2 are being executed What additional actions (if any) should be taken?

A. Reactor level CAN be determined, continue to execute EOI-1 and EOI-2.

B. Reactor level CAN be determined, enter C-1.

C. Reactor level CANNOT be determined, enter C-4.

D. Reactor level CANNOT be determined, enter C-1.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet CURVE 8 RPV SATURATION TEMP 400 380

t 360

~

~ 340

~ 320 w

~

l 300

~~ 280 ct::

w 260 z

a.

IE 240 w

I-220 200 0 50 100 150 200 250 RPV PRESS (PSIG)

CONSTANT ABOVE 250 PSIG

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

'i ., .

..**... ' i .. . ,

I cA:UTlO.NS CAUTION #1

., AN RPV WATER LVl iNSTRUMENT MAY BE USED TO DETERMINE OR TREND l\lL ONLY WHEN iT READS ABOVE THE rvm>lIMUM INDICATED LVL ASSOCIATED WITH THE HIGHEST MAX DW OR se RUN TEMP, e IF DIlV TEMPS, OR se AREA TEMPS (TABLE 6). AS APPUCABlE, ARE OUTSIDE THE SAFE REGION OF CURVE 8, THE ASSOCIATED INSTRUMENT Fl.,1AY BE UNRELIABLE DUE TO BOIUNG IN THE RUN.

, .*... n . .... " .... " .....

MINIMUM MAX OW RUN TEMP"""'" MAXSC INSTRUMENT RANGE INDICATED (FROM XR-64-50 RUN TEMP LVL OR TI-64-52AB) (FROM TABLES) -

ON SCALE N/A BELOW 150

-145 N/A 151 TO 200 EMERGENCY u-s-seA. B -140 NfA 201 TO 250

-155 TO +60

-130 N/A 251 TO 300

-120 N/A 301 TO 350 U-S-53 ON SCALE N/A BELOW 150 U.3--60 +5 N1A 151 TO 200 NORMAL U-3,.,20B +15 NlA 201 TO 250

( I U-3-253 Ll-3-208A, S, C, 0 OTO +60

+20

+30 I N/A N/A I 251 TO 300 301 TO 350 POST U-3-52 ACC1DENT ON SCALE N1A NiA U-S-62A

-268 TO +32 NlA Ii +10 BELOW 100 N/A

+'15 100 TO 150 SHUTDOWN +20 151 TO 200 N/A U-S-55 FLOODUP +30 201 TO 250 N/A OTO +400 +40 N/A 251 TOSOa

+50 301 TO 350 NfA

+65 351 TO 400 NJA

.*"".. ".1"... "".. ".,,, " *.. ""."" ... .. "."..... " ...,"" .

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.201 Page 32 of 117:

1. Caution #1

a. RPV water level instrument systems sense liquid level in the vessel downcomer region by measuring differential pressure (dP) between a variable leg water column and a reference leg water column. The reference leg remains full of water from steam condensing in the chamber located at the top of the reference leg water column. Excess condensate drains back into the RPV. To ensure reference leg water remains gas free a trickle flow of CRDH water is continuously injected into the 4 primary reference legs.

b. When water level in the reactor vessel lowers, variable leg height of water decreases, sensed dP increases, and indicated RPV water level lowers. The converse occurs when water level in the reactor vessel increases; variable leg height of water increases, sensed dP decreases, and indicated RPV water level increases.

c. Changes in height or density of water in the instrument reference leg can cause changes in indicated RPV water level. For example: if actual RPV water level is constant at some on-scale value and the instrument reference leg head of water (height and/or density) decreases, sensed dP decreases and indicated RPV water level increases. Under extreme conditions, a high and increasing drywell or containment temperature can decrease the density of water in the reference leg such that the instrument falsely indicates an on-scale and steadily increasing water level even though the actual RPV water level is decreasing and well below the elevation of the instrument variable leg tap.

d. It is important to note that the information presented in Caution #1 is not just a simple accommodation for inaccuracies in RPV water level indication which occur when plant conditions are different from those for which the instruments are calibrated. Rather, the caution defines conditions under which the displayed value and the indicated trend of RPV water level cannot be relied upon.

e. Part B of Caution #1 identifies the limiting conditions beyond which water in instrument legs may boil. Water in the RPV water level instrument legs is maintained in a liquid state by cooling action of the surrounding atmosphere and pressure in the reactor vessel. Water in the instrument legs will boil, however, if its temperature exceeds saturation temperature for the existing RPV pressure.

f. Boiling is a concern in both horizontal and vertical reference and variable instrument leg runs. Boil-off from reference leg water inventory reduces the reference head of water, decreases dP sensed by the instrument, and results in an erroneously high indicated RPV water level. BOiling in the instrument's variable leg exerts increased pressure on the variable leg side of the dP cell. This effect results in a lower sensed dP and an erroneously high indicated RPV water level.

g. Part B of Caution #1 references the RPV Saturation Temperature Curve (Curve 8)

The RPV Saturation Temperature Curve is generic, based simply on the properties of water. The axis for RPV pressure is plotted from atmospheric pressure to the pressure setpoint of the lowest lifting MSRV. Note that the temperature axis of the RPV Saturation Temperature Curve is not simply drywell temperature. Depending upon the relative location of instrument reference legs and variable legs, indications from monitors near instrument runs must be considered. '

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

h. Because BFN does not have the capability of directly reading temperature indications near instrument runs located in secondary containment, the RPV Saturation Temperature Curve (Curve 8) is supplemented with Table 6, Secondary Containment Instrument Runs. Table 6 identifies the temperature elements and general locations for the instrument runs to each RPV water level instrument.

i. Caution 1 part B says instruments "may be unreliable" if Curve 8 is exceeded.

This means instruments may continue to be used until and unless erratic indication is observed since momentary excursions (expected in some post LOCA situations) into curve 8 unsafe region will not result in boiling. If, however, indications of boiling are observed then that instrument is unusable until the instrument lines can be cooled and refilled.

j. Part A of Caution #1 allows the operator to determine if each indicated RPV water level range is reliable by being above the Minimum Indicated Level for each of a series of instrument run temperature ranges. Engineering calculations have determined that when indicated RPV water level is above the Minimum Indicated Level, the operator is assured that actual RPV water level is above the instrument variable leg tap, and trends are valid.

k. The Minimum Indicated Level is defined to be the highest RPV water level instrument indication which results from off-calibration instrument run temperature conditions when RPV water level is actually at the elevation of the instrument variable leg tap. Separate levels are provided for each RPV water level instrument.

I. The table in Part A is structured to give a Minimum Indicated Level corresponding to several temperature ranges for each of the RPV water level instrument ranges.

This yields more usable instrument range than would be available if single values were used.

m. There are two items to note concerning this table: 1) Except for the Shutdown Floodup instrument, all drywell temperatures are not applicable, because there is very little vertical pipe run in thedrywell. This means very little error can be caused by elevated drywell temperatures (until boiling occurs), and 2) The "MAX SC RUN TEMP" is the highest temperature reading which can be obtained from Table 6, Secondary Containment Instrument Runs.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 295030EA2.03 Tier# 1 Ability to determine and interpret the following as they apply to Low 1 Group #

Suppression Pool Water Level: Reactor Pressure.

KIA # 295030EA2.03 Importance Rating 3.7 3.9 I Proposed Question: RO # 53 Which ONE of the following describes the condition where Emergency RPV Depressurization is required based on Suppression Pool level and the reason for that requirement?

Suppression Pool water level below (1) requires Emergency RPV Depressurization based on level below the _ _ _ _--.4.(2!!:,,)1--_ _ _ __

(1) (2)

A. 12.75 feet downcomer pipe exits.

B. 12.75 feet HPCI Exhaust Line exit.

c. 11.50 feet downcomer pipe exits.

D. 11.50 feet HPCI Exhaust Line exit.

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

( I Proposed Answer: C Explanation: a. Part (1) is incorrect. This level (12.75 feet) corresponds to uncovering the HPCI exhaust line. Below this level, HPCI operation must be prevented because the exhaust line high pressure isolation setpoint is above the Primary Containment design pressure. Therefore, Part (2) is incorrect for this condition, but is correct for 11.50 feet.

b. Part (1) is incorrect as stated in (a) above. Part (2) is incorrect. However, this is the correct basis for 12.75 feet, but the stem asks for a condition requiring Emergency RPV depressurization. This condition does not warrant that action.

c. Correct answer.

d. Part (1) is correct. Below this level, the downcomer pipe exits are uncovered, which bypasses the suppression function of the Suppression Chamber in the event of a LOCA. Therefore, Emergency RPV Depressurization is required to place the RPV in the lowest energy state while sufficient Suppression Pool level is available to allow MSRV operation.

( Part (2) is incorrect as stated in (a) above.

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): EOIPM Section O-V-D page2 102 & 104 (Attach if not previously provided)

OPL 171.203 pages 59-60 Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/05/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

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

(

NO I WHILE EXECUTING STEPS SP/L-5 THROUGH SP/L-8:

IF THEN SlJPPR PL LVLCAiNNor BE SECUREI-ll"C, IRRESPECilVE OF MAINTAINED ABOVE ~2.75 Fi ADEOUATE CORE COOUNG L

SPIL4 MAINTAIN SUPPRPL LIlL ABOVE 11>.5 FT (APPX i8)

L SPIL*5 YES L SPIL-8

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.203 page 49 of 66:

1. Step SP/L-4

a. This retainment override step applies while performing Steps SP/L-5 through SP/L-8.

b. The operator is directed to secure HPCI operation if suppression pool level cannot be maintained above the HPCI exhaust discharge device.

c. Suppression pool level of 12.75 ft. corresponds to HPCI exhaust discharge device.

d. Operation of the HPCI System with its exhaust discharge device not submerged will directly pressurize the suppression chamber. If this condition exists, HPCI operation is secured, irrespective of maintaining adequate core cooling, to prevent the failure of primary containment from overpressurization.

Excerpt from OPL171.203 page 50 of 66:

2. Step SP/L-5

a. This action step directs the operator to maintain suppression pool level, using methods in EOI Appendix 18, above 11.5 ft.

b. Calculations have determined that failure of containment, failure of equipment necessary for safe shutdown of the plant, and loss of pressure suppression function of containment, are prevented when suppression pool level is maintained above 11.5 ft.

c. 11.5 ft. corresponds to the bottom of the downcomer openings

d. If suppression pool level cannot be maintained above 11.5 ft., the operator continues at Step SP/L-7. The RPV is not permitted to remain at pressure if suppression of steam discharged from the RPV to the suppression pool cannot be assured.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 29S031EK3.01 Tier # 1 Knowledge of the reasons for the following responses as they apply 1 Group #

to Reactor Low Water Level: Automatic Depressurization System actuation. KIA # 295031 EK3.01 Importance Rating 3.9 4.2 I Proposed Question: RO # 54 Given the following Unit 1 plant conditions:

HPCI 120VAC POWER FAILURE (9-3F W7) is in alarm.

A LOCA has occurred initiating a scram on Low Reactor Water Level.

Reactor water level (-) 122 inches and lowering

Drywell pressure 1.8 psig and steady

A Pre-Accident Signal (PAS) has just been received and all ECCS equipment respond as designed.

Assume NO operator actions.

Which ONE of the following describes the time that must elapse before ADS automatically initiates and the basis for this response?

ADS will initiate in _~(......

1),-- This actuation is in response to a _ _ _~(2!::.,)f..- _ __

(1) (2)

A. 265 seconds LOCA inside the Drywell B. 360 seconds LOCA inside the Drywell

c. 265 seconds LOCA outside the Drywell D. 360 seconds LOCA outside the Drywell

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: D Explanation: a. Part (1) is incorrect. This time delay is associated with -122 inches received without a high DW pressure (>2.45 psi g), which is given in the stem.

However, once this timer times out, if ECCS pumps are running, a 95 second timer initiates and must time out before ADS initiates. This makes the total time 360 seconds. Part (2) is incorrect. This is the basis for ADS initiation with BOTH high DW pressure AND low RPV level.

b. Part (1) is correct as stated in (a) above. Part (2) is incorrect as stated in (a) above.

c. Part (1) is correct as stated in (a) above. Part (2) is correct. ADS initiation in the absence of high DW pressure is due to decay heat boil-off following a LOCA outside the Drywell with MSIV isolation.

d. Correct answer.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): OPL 171.043 pages 13 & 14 of 30 (Attach if not previously provided) 1-ARP-9-3F Window 7 Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/06/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: The addition of "Assume NO operator action" was added due to procedural guidance which would inhibit ADS initiation under this condition. In this condition, 1-EOI-1 flowchart path RC/L would allow ADS to be inhibited below -100 inches. In addition, 1-EOI-C1 would be entered below approximately -120 inches and direct that ADS be inhibited. In fact, there are no foreseeable circumstances where ADS would be allowed to auto initiate by procedure.

The HPCI 120VAC Power Failure annunciator is to provide realistic conditions where ADS would auto initiate. If HPCI were operable, ADS would not be required under these conditions.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.043 pages 13 & 14 of 30:

(

1. Automatic Depressurization Initiation Logic

a. The following conditions must be met before automatic depressurization will occur:

1) Two coincident signals of high drywell pressure (+2.45 psig) and low low low reactor vessel water level (-122")

OR

-122" for 265 sec.

2) A confirmatory low reactor vessel water level signal (+2") (Tech Spec Value 0")

3) Anyone of the four RHR pumps or either A or B and either Cor D Core Spray pumps running NOTE:

This signal comes from pressure switches on the discharge of the pumps which give permissives in the logic above a set pressure of 100 psig for RHR pumps and 185 psig for the Core Spray pumps.

RHR CS PS-74-8A and 8B (Pump PS-75-7 A) (Pump A)

PS-74-31A and 31 B PS-75-35 (Pump B) (Pump B)

PS-74-19A and 19B PS-75-16 (Pump C) (Pump C)

PS-74-42A and 42B PS-75-44 (Pump D) (Pump D)

4) A 95-second timer must be timed out

b. The high drywell pressure signal seals in immediately upon receipt of the signal

1) Must be manually reset after the signal has cleared

2) Indicative of a breach in the process system barrier inside the drywell

c. The reactor vessel low water level signals (-122" and +2") indicate that fuel is in danger of becoming overheated

1) The -122" water level signal would not normally occur unless the HPCI System had failed

2) These signals do not seal

3) The -122" water level initiation setpoint is selected to open the SRVs and depressurize the reactor vessel in time to allow fuel cooling by the Core Spray and LPCI Systems following a LOCA, in the event that the other makeup systems (Feedwater, CRD Hydraulic, RCIC, and HPCI) fail to maintain vessel level

4) The -122" setpoint will also initiate 265 second timers that seal in and will run even if water level is restored to >-122". The timers can be reset (if Rx. Level >-122") using pushbuttons in the auxiliary instrument room.

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

5) Once these timers have timed out, the drywell pressure contacts are bypassed, but other relays (that are not sealed in) must still sense reactor level <-122"

6) If so, and the other conditions are met <<+2" and low pressure pumps running), the 95 second timers will start.

7) This feature is based on a LOCA outside of the drywell which has been isolated. Level is below -122" and inventory is boiling off due to decay heat.

8) General Electric calculations have determined that the core will remain covered for 15 minutes after the -122" level is reached. Our system will initiated within the 15 minutes calculated by GE

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panel 9-3 . 1-ARP-9-3F Unit 1 XA-55-3F Rev. 0016

. Page 10 of 39 SensorJTrip Point:

HPCI120VAC Relay 23A-K50 Loss of 120VAC from mil II ECCS ATU inverter POWER FAilURE and loss of power to HPCI Flow Ind. Controller (1 ~FIC-73-33).

(Page 1 of 1)

Sensor 1-PNLA-009-00 19 Loca1jon: Aux~iary instrument Room, EI 593' Probable A Fatted fuse l-FU2-073-OO33C, Panel 1-9-82 Cau.se: B. OJV II ECCS ATU inverter fai~ure C. loss of 250V OC power supply to OIV II ECCS ATU inverter. (25011 Reador MOV Sd 1A Compt 11A1)

Automatic A_ HPCI controller loses pcywer. {HPCI is inoperable)

Actloo: B. If HPCI is in service, HPCI TURBINE STOP VALVE, 1-FCV-73-18, closes and HPCI controller loses power.. (HPCI becomes inoperable}

C. If fuse 1-FU2-073-0033C Is cleared, 1-P1-064-OO67B wlillose power and become inQPerable Operator A DISPATCH personnel to check the following:

Action:

Inverter ruse 1-FU2-073-0033C, Panel 1-9-82 o

OIV II ECCS ATU inverter o

RMOV Bd 1A. Compt 11At o B. REFER TO Tech Spec 3.3.5.1 and Table 3.3.5 . 1-1. 0 C. REFER TO Tech Spec 3.3.1.1, Table 3.3.3.1-1, and TRM 3.3.5 for iooperabSe indicator 1-PI-0IJ.iI-0067S. 0

References:

1-45E620-i-2 1":730E:928-2 and-4 Technical Specifications 3.5.1, 3.5.2,3.3.3.1

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

( Examination Outline Cross-reference: Level SRO 29S037EA1.04 Tier# 1 Ability to operate and/or monitor the following as they apply to 1 Group #

SCRAM condition present and Power Above APRM Downscale or Unknown: Standby Liquid Control. KIA # 295037EA1.04 Importance Rating 4.5 4.5 I Proposed Question: RO # 55 An ATWS has occurred on Unit 1. The following conditions exist after the OATC has initiated Standby Liquid Control (SLC) injection using the 3A SLC pump:

Pump Running Red Light On

Squib Valve Continuity Lights Off

SLC SQUIB VALVE CONTINUITY LOST (9-5 W20) in alarm

SLC Pressure 1200 psig

Reactor Pressure 1000 psig

RWCU is in service.

Given these control board indications, Which ONE of the following is the appropriate action(s)?

Standby Liquid Control_----l...(1!..J.)__ injecting to the RPV. Perform the foliowing: _ _--l(=2)L--_

(1 ) (2)

A. is NOT Initiate SLC pump 3B.

B. is NOT Fire the squid valves locally.

C. IS Manually isolate RWCU.

D. IS Fire the squid valves locally.

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: C Explanation: a. Part (1) is incorrect. Loss of continuity would indicate a SLC system failure BEFORE SLC is manually initiated. This is expected indication AFTER SLC is initiated. Part (2) is incorrect. The given conditions of SLC pressure above RPV pressure indicate that SLC Pump 3A is injecting.

b. Part (1) is incorrect as stated in (a) above. Part (2) is incorrect since the given indications are correct for SLC manual initiation. This would be a correct answer if discharge pressure was NOT above RPV pressure.

c. Correct answer. RWCU should have automatically isolated on SLC initiation.

1-EOI-Apendix 3A requires RWCU manually isolated if auto isolation does not occur.

d. Part (1) is correct. Part (2) is incorrect since the given indications are correct for SLC manual initiation. This would be a correct answer if discharge pressure was NOT above RPV pressure.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 1-EOI-Appendix 3B, 1-ARP-95B W20 (Attach if not previously provided)

(

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # OPL171.201.1 (Attached)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Original question OPL171.201.1:

An ATVVS has occurred and the following conditions exist after the Unit Operator has initiated SLC injection using the "A" pump:

Red Light On

Squib Continuity Lights Off

Flow Light On

Alarm "SLC Injection Flow to Reactor"

Alarm "SLC Squib Valve Continuity Lost"

SLC Pressure 1200 psig

Reactor Pressure 1000 psig

Tank Level 50%, lowering Given these control board indications, DETERMINE which of the following is the appropriate action.

A. Start SLC Pump B and continue running SLC Pump A.

B. Initiate Alternate SLC Injection.

C. Stop SLC Pump A and start SLC Pump B.

D. Continue running SLC Pump A.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panel 9-5 1-ARP-9-5B Unit 1 1-XA-55-5B Rev. 0016 SensadTrip Point:

SlC SQUIB VALVE 1-XM-63-8A and sa fall 3.0 1M CONTINUITY lOST 1-EA-63-8 rise 7.0ma (Page 1 of 1)

Sensor Behind Panel 1-9-5 location: Mam Control Room Probable A Cleared fuse.

cause: B, Loss of power supp4y.

C. mown PhotDheJic Bulb.

D. Sensor malfunction.

E. SLC pump.start from the control room.

Allitormrtk: None ActiiQA!

A. IF SLC has been initiated, THEN REfER TO 1-E0l-1 or 1-1\.01-79-2. o

5. IF SLC has NOT been initiated, THEN PERFORM the foilowing:

1. CHECK amber indicating Itgtds on Panel 1-9-5 to delennme which vahle i!gniOOn circuit failed. o

2. CHECK sensor and amp meter in bact of Panel 1-9-5. o J. DISPATCH penonnel to the SLC tank, RB EL 639', to investigate squib valve Mring connection. o C. REFER TO Tech Spec 3.1.7. o

References:

147E610-63-1 1-45E620-6-2

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

(

1-E01 APPENDIX-JA SfN UNIT 1 SLC INJECTION Rev. 0 Page 1012 ILOCATION:

ATTACHMENTS:

Unit 1 Colltrot Room None uLJl

1. UNLOCK and PLACE 1-HS-63-6A, SlC PUMP 1AI1 B. control switCh in START-A or START-B posRiorl.

2. CHECK SLC System for injection by observing the following:

Selected pump starts, as indi(3ted by red light illuminated above pump control switch.

Squib valves fire, as indicated by SQUIS VALVE A and S CONTINUITY blue Ughts extinguished,

SLC SQUIS VAlVE CONTINUiTY LOST 1-EA-63-S Annunciator in alarm on Panel 1-9-5 (1-XA-55-5B, Window 20).

1-PI-63-7A, SlC PUMP DISCH PRESS, iil1di(3tes above RPV pressure.

System now, as indicated by 1-1l-6l-11, SLC flOW, red light illuminated on Pane11-9-5,

SLC INJECTION FLOW TO REACTOR 1-FA-63-11 , Annunciator in alarm on Panel 1-9-5 {1-XA-55-5B, Window 14)..

3. IF .............. _.. Proper system operation CANNOT be vemied, THEN ........... RETURN to S1ep1 and START other SLC pump.

4. VERIfY RWCU isolation by observing the following:

RWCU Pumps 1A and 1B tripped

1-fCV-69-1. RWCU INBD SUCT ISOLATION VALVE dosed

.1-FCV-69-2, RWCU OUTBD SUCT ISOLATION VALVE closed

1-fCV-69-12, RWCU RETURN ISOLATION VALVE dosed.

5. VERIfY ADS inhibited.

6. MONITOR reactor power for downward trend.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 29S038EK2.03 Tier # 1 Knowledge of the interrelations between High Off-site Release Rate 1 Group #

and the following: Plant Ventilation Systems.

KIA # 295038EK2.03 Importance Rating 3.6 3.8 I Proposed Question: RO # 56 Given the following plant conditions:

a Site Area Emergency has been declared due to gaseous effluent releases above 100 mrem TEDE.

High radiation has been detected in the air inlet to the Unit 3 Control Room.

Radiation Monitor RE-90-259B is reading 275 cpm above background.

Which ONE of the following describes the Control Room Emergency Ventilation (CREV) System response?

A. NEITHER CREV unit will automatically start at the current radiation level.

B. BOTH CREV units will automatically start with suction from the normal outside air path to Elevation 3C.

C. The Selected CREV unit will automatically start and will continue to run until Control Bay Ventilation is restarted; then, it will automatically stop.

D. The Selected CREV unit will automatically start. The Standby CREV unit will begin to auto-start; but, will ONLY run if the selected CREV unit fails to develop sufficient flow.

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

( I Proposed Answer: 0 Explanation: This is incorrect. This is plausible because the Tech Spec initiation setpoint

a. is 270 cpm, which is less than the given radiation level. However, the actual CREV initiation setpoint is 221 cpm.

This is incorrect. This is plausible since both CREV units receive a start

b. signal on a valid initiation. However, the CREV unit NOT selected will experience a 30 second time delay on initiation and will only complete its start sequence if the selected CREV unit fails to start.

This is incorrect. This is plausible because the start sequence is correct.

c. However, once initiated, CREV must be manually secured. There is no automatic shutdown capability, only trips.

d. Correct answer.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank # RO 290003A3.01 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 3/25/2008 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Control Bay and Off-Gas Treatment 0-01-31 Unit I} BUi.ld1ng Air Conditioning System Rev.1}126 Page 21 of 283 3.6 CREV and CREV instrumentation operability issues (continued}

B. The main control room boundary may be opened intermittently under administration controls. For openings other than oormal entry and eXit,lhese controls consist of s1ationing a dedicated indiVidual at the opening who is in continuouscornmuni*cation with the main control room and whose task is to dose the opening when main contrOij room isolation is indicated. With both CREYs iooperable in Modes 1, 2 or 3, for other thana control room boundary issue, enter LCO 3.0.3 Immediately. With two CREY subsystems !inoperable during OPDRVs, initiate adton to suspend OPDRVs. Reference TS 3.7.3.

C. When there is an automaiic actuation of GREYS,the foltowing automatic isolation dampers and hatch is required to be closed for GREVS to be considered operable.

1. O--FCO-31-150S, O-FCO-31-150D, O-FCO-31-150E, O-FCO-31-150F, O-FCO-31-150G.

2. Removable equipment ihatch in U-3 Mechanical Equipment Room, floor Elevation 617' .

The CREV system utlllzes 15.45 kW Duct heaters to control moisture buildup in

( D.

the charcoal adsorber. A maifunc:1fon of the 15.45 kW duct heater makes the applicable CREV unit inoperable. [Reference Func1iooal Evaluation in PER 74959 and 75680]

E. One of the UNIT 1 8. 2 Contro! Bay Supply Fans and one of the UNIT 3 Controi Bay Supply Fans and their associated power and contra! circuits is required to be operable for CREVS instrumentatioo {control bay high radiatlon)to be considered operable. Reference Tech Spec 3.3.7.1 ..

F. CREV UNIT PRIMARY SELECTOR, O-XSW-031-1214 may be placed rneither the "A" or "8" position, depending on the operability status of the GREY trains.

When a CREV train is inoperable, il wUl NOT be selected as lead. When both GREY trains are operable, the preferred position for CREV UNIT PRIMARY SELECTOR, O-XSW-031-7214, is in TRAIN ~Au wIllch makes the ~A'" CREY the lead train. In the event that "AU CREV is INOP, CREV UNIT PRIMARY SELECTOR, O-XSW-031-7214 iis required to be piaced in the TRAIN "Su position so the US" GREV will initiale, without a time delay, as the lead train.

G. When one of the CREV trains is inoperable for testing, the CREV UNIT PRIMARY SELECTOR SWITCH, O-XSW*03t-1214 is required to be aligned to the train which is NOT under testing conditions to ensure the non-test train will initiate under an actual initiation signal. .

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet Excerpt from OPL171.013, HVAC page 34:

a. Flow switches are provided, one for each division/unit, to start the standby unit if the selected unit does not start or trips off. The selected unit not starting is sensed by low differential pressure across the common HEPA filter in the Unit 2 vent tower. Low differential pressure exists when a fan is not operating; this signal will normally be present. The circuit for each unit is such that its initiation sequence is begun upon either of the following:

(1) Unit is selected as primary unit and CR1 relay for that division is energized.

(2) Other unit is selected as primary unit, low differential pressure exists across the common HEPA filter, and CR1 relay for that division has been energized for approx. 30 seconds.

b. With this circuit design, when an accident signal is initially received, the selected unit will enter its initiation sequence immediately and the other unit will enter its initiation sequence approx. 30 seconds later. Once the selected unit fan has been started (taking approx. 75 seconds -- 70 for the damper and 5 for the fan), the low differential pressure signal will no longer be present in the standby unit circuitry and its damper will return to the fail-close position.

c. If the selected unit fails to start properly, it will itself be turned off by the trips noted above, and the standby unit will continue in its initiation sequence. The time delay for startup of the standby unit will be selected to ensure that regardless of the primary unit failure, both fans will not be running at the same time.

d. If the selected unit starts properly, but then trips at a later time, the standby unit will only be missing the low differential pressure signal to receive its start signal. The standby unit will start when the selected system has completed its shutdown process and the fan has been de-energized.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 600000AA1.0l Tier# 1 Ability to operate and/or monitor the following as they apply to 1 Group #

Plant Fire On Site: Respirator air pack.

KIA # 600000AA1.01 Importance Rating 3.0 2.9 I Proposed Question: RO # 57 Concerning a fully pressurized (60 minute) Self-Contained Breathing Apparatus (SCBA), which ONE of the following describes the value of the protection factor and the significance of receiving an alarm while wearing a SCBA?

The protection factor provided by a SCBA is _ _----'-"'(1=+)_ _ _ . An alarm sounding while wearing the above SCBA indicates that (2) remain(s) before all air is expired.

(1) (2)

A. 10,000 approximately 5 minutes B. 10,000 approximately 10 minutes C. 1,000 approximately 5 minutes D. 1,000 approximately 10 minutes

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: B Explanation: a. Part (1) is correct. Part (2) is incorrect. Approximately 5 minutes of time remaining is applicable to a 30 minute tank. This could be correct if the tank was NOT fully pressurized before use.

b. Correct answer.

c. Part (1) is incorrect. This is lower by a factor of 10. Approximately 5 minutes of time remaining is applicable to a 30 minute tank. This could be correct if the tank was NOT fully pressurized before use.

d. Part (1) is incorrect as stated in (c) above. Part (2) is correct. A 60 minute tank will alarm at approximately 10 to 15 minutes.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): SCBA 063.002 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/08/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from SCBA 063.002 page 7 of 21:

c. Capabilities and Limitations of the SCBA: Objective 5

1. Capabilities:

a. Offers highest assigned protection factor against gases, vapors and particulates (PF = 10,000).

b. An alarm will sound when air supply drops to ~ 114 of Work Expectation -

remaining pressure (10-15 minutes on 60 minute tanks Conservative Decision and ~5 minutes on 30 minute tanks). Making -Your response is to exit the area immediately even though you may think you have time to finish the job ..

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 700000G2.4.34 Tier # 1 Knowledge of RO tasks performed outside the main control room 1 Group #

during an emergency and the resultant operational effects:

Generator Voltage and Electric Grid Disturbances. KIA # 700000G2.4.34 Importance Rating 4.2 4.1 I Proposed Question: RO # 58 Given the following plant conditions:

A significant voltage transient on the grid initiated a fault on Unit Station Service Transformer (USST) 2A.

A fire on USST 2A resulted in actuation of fire suppression systems.

Subsequently, all off-site power was lost due to continued voltage transients on the grid.

Which ONE of the following describes the required operator actions to restore Electric Fire Pump B to service and the location where these actions are performed?

Proceed to _ _ _~(1~)_ _ _ and perform the following: _ _ _ _ _-I.'(2f:::.,j)!.....-.._ _ __

(1) (2)

A. 4KV Shutdown Board B Place the NORMAL/EMERGENCY switch for Fire Pump B to EMERGENCY and then back to NORMAL.

B. 4KV Shutdown Board B Re-close the breaker for Fire Pump B.

c. 4KV Shutdown Board C Place the NORMAL/EMERGENCY switch for Fire Pump B to EMERGENCY and then back to NORMAL.

D. 4KV Shutdown Board C Re-close the breaker for Fire Pump B.

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

( I Proposed Answer: A Explanation: a. Correct answer.

b. Part (1) is correct. This appears easier than actual to determine, however the pumps assigned to 4KV SjD boards B & C are typically backwards. For instance, RHR pump B and Core Spray pump B are powered from 4KV SjD Board C and vice versa. Application of the same logic to the Fire Pumps is a common error. Part (2) is incorrect. Re-closing the breaker will not work.

The breaker will re-open unless the closing coils are reset. This is accomplished by placing the NORMjEMER switch to EMER and back to NORM.

c. Part (1) is incorrect. Refer to the explanation in (b) above. Part (2) is correct. The closing coils must be reset by interrupting DC control power to the breaker using the NORMjEMER switch. Once accomplished, the Fire Pump will automatically re-start as long as the initiation signal is still present.

d. Part (1) is incorrect. Refer to the explanation in (b) above. Part (2) is incorrect. Re-closing the breaker will not work. The breaker will re-open unless the closing coils are reset. This is accomplished by placing the NORMjEMER switch to EMER and back to NORM.

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

Technical Reference(s): O-AOI-57A (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/08/2008 Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet BFN Loss of Offsite Power (161 and 500 O-A0I-51-1A Unit 0 KV)lStation Blackout Rev. 0011 Page 180111 4.2 Subsequent Actions (continued)

NOTE INERiGJ H' eledrical-drivenfire pumps were running due to an automatic inittation. the 52Y relay will klck-out the pump breaker; thus preventing the pump(s) from being restarted. In order to restart the fire pumps, Step 4.2[21] should be perfonned to momentarily inte:rrupt DC conu-ol power to the dosing coifs of the fire pump breakers. lNRC IE 88-075]

[21 J [NERIq IF electrical-driven fire pumps were ruMing due to :an automat;c initiation signal prior to loss of off-site power AND the automatic initiation signal is still present, THEN PERFORM the fcllowing to restart fire pumps necessary for fire suppression: (Otherwise N/A)

[21.1] PLACE the NORMALIEMERGENCY switch for Fire Pump A (B) ee} 00 4kV Shutdown Bd. A (B) (C) rompt 11(11 )(1 0) to EMERGENCY and back to NORMAL 0

[21.2] VERIFY that the fire pump(s) start [HER IE 88-075J 0

[22) MONnOR Batt Bd amps and VOL TS on Panels 9-8. PLACE the battery charger back in service within 30 minutes after loss of the charger to the battery. 0 NOTE Step 4.2(22.1) will ensure adequate voltage is available to operate switchya,rd breaikers.

[22.1] TRANSFER breakercootrol power from NORMAL (Battery Board 4) to ALTERNATE (Battery Board 2) as folkJws (at Panel 9..24):

OPEN BKR 642 NOR FOR FROM BATT BD 4 BKR219. o

CLOSE BKR 641 ALT FOR FROM BATT 802 8KR501. o

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN loss of Offsite Power {161 and 500 O-A0I.51-1A Unit 0 KV)/Station Blaekout Rev. 0011 Page n of 11 IUustration 2 (Page 1 of 1) loss of Diesel Generator Cooling 1.0 lOSS OF DIESEL GENERATOR ENGINE COOLING sao Unit OGs lost on SBO No available #3 EECW Pumps with the following! DGs lost Unit 1 AandC D, JAand 36 Unit 2 BandO C, 3Aand 38 Unit 3 3Aand3C C, Dand3B 2 I) lOSS OF DIESEL GENERATOR ROOM COOLING SBO Unit OGs lost on SBO No availabfe Diesel Generator Room Cooling with the f04lowing OGs lost Unit 1 AandC *0 Unit 2 Band 0 "A Unit 3 3Aand 3C *No Room Cooling

Diesel Generator Room Cooling requires power to 1 OSl Aux Board on U11U2 and U3 3.I) EECW PUMP DIESEL GENERATOR POWER SUPPLIES OG A B C 0 3A 3B 3C 3D North A1 C1 A3 C3 EECW South B3 03 B1 01 EECW 4, 0 EECW VALVE POWER SUPPLIES O-FCV-061-0048 D1 X-TIE OSl Aux Board B 4KV SO SO 0 (Nor") 4KV SO BO B (All)

O-FCV-061-0049 C1 X-TIE DSl Awe Board A 4KV SO BO A (Nor) 4KV SO BD B (All)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 29S008AK3.04 Tier # 1 Knowledge of the reasons for the following responses as they apply 2 Group #

to High Reactor Water Level: Reactor Feed Pump Trip.

KlA# 295008AK3.04 Importance Rating 3.3 3.5 I Proposed Question: # RO 59 Which ONE of the following describes the logic arrangement for the Reactor Feed Pump High Water Level Trip and the Technical Specification basis for that trip?

The Reactor Feed Pump (RFP) High Water Level Trip logic is _ _-\,.:(1""J,)_ _ and is designed to prevent damage to the (2)

(1) (2)

A. one-out-of-two-twice RFP turbines B. one-out-of-two-twice Main Turbine C. two-out-of-two-once RFP turbines D. two-out-of-two-once Main Turbine

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

( I Proposed Answer: D Explanation: a. One-out-of-two-twice is the RPS logic arrangement which will initiate a scram after the Main Turbine trips due to high reactor Water Level.

In addition, although tripping the RFP turbines will protect them from damage, the basis for the trip is to protect the Main Turbine from damage.

b. Logic is incorrect, Basis is correct. Protecting the Main Turbine is the basis for the High Reactor Water Level Trip.

c. Logic is correct. Basis is incorrect.

d. Correct answer

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): U1 TSB 3.3.2.2 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New RMS 6/16/2008 Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments: The logic arrangement is properly covered in the lesson plan but the basis for the trip is not. However, an objective does exist in the lesson plan which requires knowledge of the bases for RFP limitations contained in the Operating Procedure 01-3. High RPV Level trip is addressed within 01-3. I rated this question as MEM because no significant analysis should be required to answer the question correctly.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Feedwater and Main Turbine High Water Level Trip Instrumentation B 3_3.2.2 B 3.3 INSTRUMENTATION B 3.3...22 Feedwater and Main Turbine High Water Level T tip Instrumentation BASES BACKGROUND The feedwater and main tumine high water level trip instrumentation is designed to detect. a potential fatlure of 1I1e Feedvlater Level Control Systenl that causes excessive feedwater flow .

With excessive feedwaterflow, the water !evel in the reactor vessel tises toward the high water level reference point.

causrng the trip of the three feedwater pump turbines and the main tumine.

Reactor Vessel Water level - High signals are provKled by level sensors that sense the difference between the pressure

( due to a constant column of water (reference leg) and the pressure due to the actual water level in the reactor vessel (variable leg). Two channels of Reactor Vessel Water level -

High instrumentation per ttip system are provided asill1plJl to a two-out-of-'two initiation fogre that ttips the three feedwater pump turbines and the main tumine. There are two trip systems, either of which wnll inttiate a. trip. The channels indude eliectronicequiprnent, LS-3-20BA, LS-3-200B.,

LS-3-200C, and LS-3-208D (e.g., trip units) that compares measured input s.gmds with pr~establf:shed setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a main feedwaterand rurbre tr~ signal to the trip logic.

A trrp of the feedwater pump turbines limits further increase in reactor vessel water level by limiting further addition of feedwater to the reactor vessel.. A trip of the main turbine and dosure of the stop valves protects the turbine from damage due to water enteting the turbine.

(continued)

BFN-UN~T 1 B 3.3-72 Revision 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Feedwater and Main Turbine High Water level Trip Imrumentatioo B3.3.22 BASES (continued)

APPLICABLE The feedwater and main turbine high water level trip SAFElY ANALYSES insilrumentation is assumed to be capable of providing a turbine trip in the design basis transient analysis for a feedwater controller failure, maximum demand event (Ref. 1i). The reactor vessel high water level trip indirectly initiates a reader scram from the main turbine trip (above 30% RTP) and trips the feedwater pumps, thereby terminating the event. The reader scram mitigates the reduction in MCPR Feedwater and main turbine high water level trip instrumentation satisfies Criterion 3 of the NRC Poticy Statement (Ref. 3).

LCO The LCO requires two channels of the Reactor Vessel Water Level - High instrumentation per trip system to be OPERABLE to ensure that 00 single instrument failure w~1 prevent the feedwater pump turbines and main turbine trip on a valid Reactor Vessel Water Level - High signal. Both channels in either trip system are needed to provide trip signals in order for the feedwater and main turbine trips to occur. Each channel must have its setpoint set willin the specified Allowable Value of SR 3.3.2.2.3. The Allowable Vatoo is set ID ensure that 1he thermal Imts are not exceeded during the event. The actual setpoint is calibrated to be consistent with the applicable setpoint rnethodbIogy assumptions. Nominal trip setpoints are specified in the setpoint calculations. The nominal seipoints are selected to ensure that Ile setpoines do not exceed the Allowable Value between S1UCCeSSive CHANNEL CALIBRATIONS. Operation with a trip seCpoint less OOIlSEHVaUve than the nominal trip setpoint. but within its AlIOV\i'abIe Value, is acceptable.

{rominued}

BFN-UNfT 1 B 3.3-73 Revision 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 295009AK2.02 Tier # 1 Knowledge of the interrelations between Low Reactor Water Level 2 Group #

and the following: Reactor water level control.

KIA # 295009AK2.02 Importance Rating 3.9 3.9 I Proposed Question: RO # 60 Given the following Unit 1 plant conditions:

A scram has occurred and all control rods did not fully insert.

Reactor power 28%

Reactor water level (-) 85 inches and stable

Reactor Water Level Band (-) 55 to (-) 100 inches using HPCI

Reactor Pressure 985 psig and rising slowly

Reactor Pressure band 800 to 1000 psig using MSRVs Which ONE of the following describes the Reactor Water Level response to opening a MSRV and the reason for that response?

Opening a MSRV will cause indicated level to be (1) due to (2)

(1) (2)

A. lower lower ilP across the dryers and separators.

B. lower higher ilP across the dryers and separators.

C. higher lower ilP across the dryers and separators.

D. higher higher ilP across the dryers and separators.

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

( I Proposed Answer: 0 Explanation: . a. Part (1) is incorrect. Opening an MSRV causes a reduction in pressure outside the shroud where level is measured. This causes reference leg pressure to lower, measured ~p to lower, and indicated level to rise. Part (2) is incorrect. ~p across the dryers and separators gets higher.

b. Part (1) is incorrect as stated in (a) above. Part (2) is correct. Pressure outside the core shroud will lower when the MSRV is opened, but that will cause indicated level to rise as stated in (a) above. ~p across the dryers and separators gets higher.

c. Part (1) is correct. Part (2) is incorrect. ~p across the dryers and separators gets higher.

d. Correct answer.

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): OPL 171.003 page 30 of 54 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/06/2005 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fu.ndamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.003 page 29 and 30:

\. h. Steam flow effect on reactor water level (1) Steam flowing through the dryers is forced to change direction several times, resulting in a pressure drop across the dryers.

(2) At 100 percent steam flow, the pressure drop is 7 inches of water. On a reactor scram, this AP "goes away", due to the void collapse, causing lower back pressure on the recirc pumps and jet pumps. Water from the annulus is relocated to the core area, causing sensed (and indicated) water level (in the annulus) to drop.

(3) Therefore, at 100 percent steam flow, P1 is 7 inches of water less than P2.

(4) The level outside the dryer skirt (down comer region) is 7 inches higher than inside the skirt.

(5) Since the vessel level instruments compare the reference column height to the down comer (variable column) height, setpoints are adjusted to compensate for this error.

NOTE: The principles discussed here apply to this question, but from a different perspective. The effect discussed in the lesson plan applies to an increase in RPV pressure where the question relates to a decrease in pressure as the MSRV is opened. The theory of operation is identical.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 295012AK1.02 Tier # 1 Knowledge of the operational implications of the following concepts 2 Group #

as they apply to High Drywell Temperature: Reactor Power Level Control. KIA # 295012AK1.02 Importance Rating 3.1 3.2 I Proposed Question: # RO 61 Given a condition where a rising Drywell Average Air Temperature CANNOT be restored or maintained, which ONE of the following temperatures will require initiating a reactor scram and the bases for that required action?

Before Drywell Average Air Temperature exceeds -l.1L OF, a manual reactor scram is required in accordance with EOI-2, "Primary Containment Control?"

A. 150 B. 160

( c. 200 D. 280

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: B Explanation: a. Incorrect. 150 of is the Tech Spec limitation requiring a normal shutdown, but not a manual scram.

b. Incorrect. 160 of is the limit requiring entry into EOI-2.

c. Correct answer. 200 of requires entry into EOI-1 which initiates a manual scram.

d. Incorrect. 280 of is the Drywell Average Air Temperature limitation requiring Emergency Depressurization

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): ARP-9-3B Window 3 & 16 (Attach if not previously provided)

EOI-2 Flowchart Tech Spec bases 3.6.1.4 Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New RMS 6/16/2008 Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: I rated this question as CIA because the candidate must analyze the temperature and apply that to differing procedure guidance to ascertain the correct action.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN P*anef 1-9-3 Unit 1 1-XA-55-38 SensorlTrip Point Alarms off 1-TR-80-1 DRYWELL ATMOSPHERIC 1-TE-OOO-OOO1 through -0012 148"F TEMP H1GH 1-TE-OaO-0013, -0014 200"'F 1-TA-80-1 1-TE-OOO-OO 15 185"F l-TE-OOO-0016 through -0025 '14B"F 1-TE-OOO-0026 370"F (page 1 of 1) 1-TE-D80-G021 370°F 1-TE-DBO-G028 370°F 1-TE-080-G029 26soF 1-TE-080-0030 through -OG32 148"F Sensor Multiple locations in Drywell Location:

Probable A. Drywell Cooler(s} failure Cause: B. Loss of cooling water (RBCCW} to Drywell Coolers C. Sensor malfunction Automatic None Action:

Operator k CHECK Drywall temperature and pressure. o Action: S. 'VERIFY OPEN RBCCW PRI CTMT OUTLET VALVE, 1-HS-70-47A, Panel 1-9-4. o C. START all availab!e Drywell Coolers. o D.REFER TO 1-AOI-64-1. o

References:

1-4SE620'-3-2 1-47ES10-00-1 0-47W600-90 GE Drawing 1-730E933-'t 1-47E605-173A

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panel 1-9-3 1-ARP-9-38 Unit 1 1-XA-55-38 Rev. 0020 19 of 38 SensorfTrip Point DRYWELL TEMP HfGH 1~TE-064-0052C ~ 154"F (Alarm comes off recorder)

TA-64-52 (Page ... of 1)

Sensor 1-TE-064-0052A Location: Rx Bldg (DI)IW'e1il)

El584' AZ225° Probable A. DryweIICooler(s) failure Cause: B. Loss of cooling water (RBCCW) to Orywell Coolers C. Sensor malfunction Automatic None Action:

Operator A. CHECK DlYWeli temperature and pressure using multipile Action: indications. a B. VERIFY Drywell Coclers running and START spare Drywell Cocler(s). a C. VERIFY OPEN RBCCW PRI CTMT OUTLET VALVE, 1-HS-I0-41A, 1-PLNA-009-OO04. a O. START additional RCW pumps. REFER TO 1-01,-24. o E. IF high Orywell temperature continues, THEN REFER TO 1-AOI-64-1. o F. If Drywetl temperature is due to a loss of RBCCW, THEN:

REFER TO 1-AOI-I0-1. a G. If temperature is above 16(]<>f, THEN ENTER 1-EOI-2 Flowchart. a

References:

1-45E620-3-2 1-47E61 0-64-1 0-41W6OO-90

ES-401 Sample Written Examination Form ES-401-S Question Worksheet r-I I

j

~

\'----1.......------111>.------\

I L

L

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

(

Drywetl Air Temperature 3.6.1.4 3.6 CONTAINMENT SYSTEMS 3 . 6.1.4 DryweU Air Temperature LCO 3.6.1 A Drywell average air temperature shall be :s: 150"F APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDrTION REQUIRED ACTION COMPLETION TIME A Drywell average air A.1i Restore drywell average 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months temperature not within air temperature to within limit. limit B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated COmpletion Time not met. AND B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months BFN-UNIT 1 3.6-17 Amendment No. 234

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

( DryweII Air Temperature B 3.6.1-4 B 3.6 CONTAINMENT SYSTEMS B 3.6.1-4 Dlyweli Ak Temperature BASES BACKGROUND The drywall contains the reactor vessel and piping, which add heat to theairspacR DryweU coolers remove heat and maintain a suitable environment. The average akspace temperature affects the calculated response to postulated Design Basis Accidents (DBAs). The lillllitatioo 00 the drywell average air temperature was deve~oped as reasonable, based on operating experi,ence. The limitation on drywell air temperature is used in the Referen*ce 1 safety analyses.

APPLICABLE Primary contaimnent perionnance is evaluated for a SAFETY ANALYSES spectrum of break sizes for postulated loss of cooLant accidents (LOCAs) (Ref. 1). Among the inputs to the design basis analysis is the initial drywall average air temperature (Ref 1).

Analyses assume an initial average dryweU air temperature of 150"F. Thislimttatioo ensures that the safety analysis remains va.lid by maintaining the expected initial conditions and ensures that the peak drywell temperature does not exceed the maximum allowable temperature of 336"F (Ref. 2). Exceeding this temperature may result in the degraciation. of the primary containment structure under accident IDoos. Equipment inside primary containment required to mitigate the effects of a DBA is designed to operate and be capable af operating under environmentalconditioos expected for the accident Dryy.rell air temperature satisfies Criterion 2 of the NRC Policy Statement (Ref. 3).

(continued)

BFN-UNfT 1 Revision 0,50 May 03, 2007

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 295020AA1.0l Tier # 1 Ability to operate and/or monitor the following as they apply to 2 Group #

Inadvertent Containment Isolation: PCIS/NSSSS.

KIA # 295020AA 1.01 Importance Rating 3.6 3.6 I Proposed Question: RO # 62 Given the following Unit 2 plant conditions:

An inadvertent Group 6 isolation occurs due to a bag of contaminated trash being brought too close to the Unit 2 Reactor Zone Ventilation Radiation Monitors.

When the bag is removed, the NUMAC radiation monitor readings return to normal.

Which ONE of the following describes the MINIMUM operator actions required to reset the Group 6 isolation?

The NUMAC radiation monitors* (1) reset. The PCIS isolation indication on Control Room Panel 9-4 (2) reset.

(1) (2)

A. must be manually must be manually B. must be manually will automatically C. will automatically must be manually D. will automatically will automatically

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: C Explanation: a. Part (1) is incorrect. NUMAC monitors are widely used in each unit. Some applications require manually resetting trips after the condition clears, while others reset automatically. In the case of Reactor Building High Radiation NUMAC monitors, the trip automatically resets. Part (2) is correct. The PCIS logic contacts re-c1ose automatically, but the trip relay must be reset by manual action from Panel 9-4.

b. Part (1) is incorrect as stated in (a) above. Part (2) is incorrect. Only the relay contacts associated with High Radiation re-c1ose automatically. The trip relay must be manually reset.

c. Correct answer.

d. Part (1) is correct as stated in (a) above. Part (2) is incorrect as stated in (a) and (b) above.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): OPL 171.148 pages 29 and 52 (Attach if not previously provided)

OPL 171.033 page 26 Proposed references to be provided to applicants during examination: None Question Source: Bank # OPL 171.033.10 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Original question OPL171.033.10:

An inadvertent Group 6 isolation occurs due to a contaminated bag of trash being brought too close to the Unit 2 Reactor Zone Ventilation Radiation Monitors.

When the bag is moved, the monitors return to normal readings.

What is the MINIMUM operator actions to reset the Group 6 isolation?

A. Reset rad monitors at both NUMAC drawers.

B. Reset PCIS at Panel 9-4.

C. Reset rad monitors on the NUMAC drawer in alarm and reset PCIS at Panel 9-4.

D. Reset rad monitors at both NUMAC drawers and reset PCIS at Panel 9-4.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPl111.148 Rfwiaion 8 Fle.ge 29 01150 12.. lPRM alarms (Panel S-5) V.8.2

a. LPRM Upscale and LPRM ~ Total . . . . tHo 125%

(1) The upscale and ~_ '1,8.4 point ~ ere daptayed inskfe ,z. Va-+

the bargmphs and a.1tI.lUI Abr:INa ead'I indieation

disptQ8d .at.IcMt the lPRM bararaph bargraphS. The SOlid box above ...

bargraph ifdcata that ... _ point DownIcaIe . . . .

marker. preaenty EMleaded YA'lh than or equal to a holfow box indicates a pal 3% of scale AND condtion. upscale

gl'Hler than or equal to (2) A past oond!tion maybe raat by 100 % of acate.

entering the TRIP STATUS d_.y and pressing tha AESETMEMOAY softkey.

(3) Trip memory indicates trip conditions Obj V.D.S that have occurred In the: past but ara no longer in the "trlppecf' ccndition. The trip memory Is cleared oneHher of two occasions:

(a) ~reset OR (0) The inllrument ~

chanoN from INaP 10 OPER.

(4) The*tm . . . of each LPRM Deteckw .~totheRBM indnJl'NII1I8, (5) The alann ..... 01 . . . LPRM Deteclor. indlcat8d on the APRM and LPRM ins1rument front panel display and on i. aIIOCiated operator's display (ODA).

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

(

Ol"'l171,t48 Ra~mon8 52Df 150 o.

(1) ctjm;i~~.. to lceat~d 00 Pm, 9-14 that TP~8 (2) status of either be tn n('jnmlvideo, blank, or inverse video.

(3) abnormal

....."""....n 'I~e~ ,RPS tn,p.s Uli'{)rass condiUon, faults, (4) Area of display eii11erbe blank. or normal video wnenconditlon is deared.

(5) When normal video is ai.splayed, the operator must reset the memory for that

( NUMAC instrul'T'I&I1t to clear the normal Video display.

So, for instance, it a rod block is displayed for APRM channel 1 in inverse video, then cleared, both the normal video on the OCA AND on the APRM instrument must be reset to clear the normal video display.

(6) Three bargraphs are presented on the APRMchassis as a default display for the OPEA and INOP-CAL. mode: APRM Flux, STP(Simula,1ed Thermal Power) and Flow with the numerical vwues (to the right of each graph) in *percent of rated".

Each bargraph contains tha al arm and trip set point markers (up or down arrows) for the various lirnltsthatare being moo itored by the instrument.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.033 page 26:

a. Trips (1) Reactor zone and refueling zone monitors work independently of each other for trip actuation (2) High radiation trip setpoint is 72 mr/hr Rad-monitor auto for the refueling and reactor zones resets when alarm is clear

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

( 2-EOI APPBNDIX-SF R9V. 5

................................... page 1 of 3 2-I:OI APPENDIX-SF RESTORING REFUEL ZONE AND RlUt.c:rOR ZONE VENTJ:LATJ:ON FDS FOLLOWJ:llG GROUP 6 J:SOLATJ:ON LOCATION: unit .2 control Room ATTACHMBNTS: None 1

VBR1I'Y PCI S Res9 t .

2. PLACK R9fuel Zone V9ntilation in service as follows (panel 2-9-25):

a. V'BR1n 2-HS-64-3A, REFUEL ZONE FANS AND DAMPERS, control switch is in OFF.

NOTE: Wh9n Refuel Zone supply and exhaust fans start, Refuel zone supply and exhaust damp9rs open automatica.lly.

b. PLAa 2-HS-64-3A, REFUEL ZOMEl FANS AND DAMPERS, control switch to SLOW A (SLOW B) *

c. C'lIBCK two SPLY/EXH A(B) gre9n lights above 2-HS-64-JA. REFUEL ZONE FANS AND DAMPBRS, control switch extinguish and two SPLY/BXH A(B) red lights i 11umina te *

d. V'B.R:rnoPD the following dampers:

2-FCO-64-5, REFUEL ZOMEl SPLY OUTBD ISOL DMPR

2 - FCO 6

REFUEL ZOMEl SPLY INBD ISOL DMPR

2-FCO-64-9, REFUEL ZOMEl BXH OUTBD ISOL DMPR

2-FCO-64-10, REFUEL ZONE EXH INBD ISOL DMPR.

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

( Examination Outline Cross-reference: Level SRO 295034EA1.03 Tier # 1 Ability to operate and/or monitor the following as they apply to Group # 2 Secondary Containment Ventilation High Radiation: Secondary containment ventilation. KIA # 295034EA1.03 Importance Rating 4.0 3.9 I Proposed Question: RO # 63 Given the following Unit 2 conditions:

RX ZONE EXH RADIATION MONITOR DNSC (9-3A W35) is in alarm

Reactor Zone Radiation detector 2-RE-90-142A has failed to a DOWNSCALE condition.

Which ONE of the following subsequent instrumentation failures will cause a Reactor Zone Isolation?

Radiation monitor (1) will initiate a Reactor Zone isolation if it fails (2)

(1) (2)

A. 2-RE-90-142B upscale B. 2-RE-90-143B downscale

(

C. 2-RE-90-143C upscale D. 2-RE-90-142C downscale

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

( I Proposed Answer: B Explanation: a. Part (1) is correct if the monitor fails downscale, but Part (2) makes this choice incorrect.

b. Correct answer. The logic is 2 of 2 taken once. This requires either two upscale trips in either channel or one downscalejINOP trip in both channels.

The downscalejINOP condition initiates a trip due to preventing an upscale trip from occurring.

c. Part (1) is incorrect. This monitor is in the same division as 142A which requires both monitors upscale to initiate a trip.

d. Part (1) is incorrect. The downscalejINOP trip logic is 1 of 2 taken twice.

Both channels downscalejINOP in the same division will NOT initiate a trip.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): OPL 171.033, Process Radiation Monitoring (Attach if not previously provided)

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Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # OPL 171.033.21 Attached New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Original question OPL171.033.21:

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One Reactor Zone Radiation detector 2-RE-90-142A (normally powered from RPS "A") has failed to a DOWNSCALE condition.

Which subsequent instrumentation failure will cause a Reactor Zone Isolation and a Primary Containment Group 6 Isolation?

A. 2-RE-90-142B (Reactor Zone powered from RPS A) upscale B. 2-RE-90-143B (Reactor Zone powered from RPS B) downscale C. 2-RE-90-143B (Reactor Zone powered from RPS B) upscale D. 2-RE-90-142B (Reactor Zone powered from RPS A) downscale

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPU7l.033 pages 26 & 27 of 78:

a. Trips (1) Reactor zone and refueling zone monitors work independently of each other for trip actuation (2) High radiation trip setpoint is 72 mr/hr for the refueling and reactor zones (3) Trip logic for the refueling and the reactor zones is identical, and the following combinations will generate a trip (a) Two high level trips in the same channel, (division)

-OR-(b) One downscale trip in each channel (division)

-OR-(c) One monitor INOP in each channel (division)

-OR-(4) Loss of RPS power to either channel (5) Automatic actions (a) Refuel Zone Trip (i) Isolate Refuel Zone on all 3 units (ii) Start SGT and opens SGT suction to refuel zone (iii) Group 6 PCIS (iv) Start CREVs (v) Isolate fresh air paths to Control Bay Elev 3C (b) Reactor Zone - Same as Refuel Zone, plus:

(i) Isolate affected unit reactor zone (ii) Open SGT suction to affected unit reactor zone

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Panel 9-3 3-ARP-9-3A 3-XA-55-JA Rev. DOll Pi.lO~. ~~...~. .~j ..... .

SensOOTrip Point:

RXZONE EXH RADIATION 3-RE-90-142A .20MRJHR MONITOR ONSC 3-RA.-90-142B 3-RE-90-142B .20MRfHR 3-RE-90-1438 .20MRfHR 3-RE-90-1438 .20MRlHR (Page 1012)

Sensor Rx Bldg, E1654' (Refuel Floor), R-18 P-UNE Locati1)O: Pa!1Jel9-42 Prooable A. Sensor malfunction.

Cause: B. Loss of power to aetectoL Au:tomatic A. One dO'l'lflScale trip: None Action: B. [NRc/C} Two out of two taken once upscale tlips or one alit of 1'11'0 taken twice downscale trips or taking both OPERl!NOP switches oui of OPERATE wiil cause Itle fullowing to Gccur:

1. ReactGr zone and refuel zone isotate.

2. SGTS iniitiares..

3. Control Room Emergency Pressurization units start.

4. H20 2 anatyzers isolate and pumps trip.

5. Drywell CAM, 3-RM-OO-256, isolates and pump trips.

5. Drywel! and Supp Chbr purge and exhaust valves close. INReNeo 8711l331!lOZ.

LER 8J.!ii28]

NOTE Trips un Itle Reactor Zone Radialion monitors witl amomatlcalty reset when the alarming condftion resets.

Operator A. VERIFY alarm condition on Itle following:

Action: 1. REACTOR ZONE EXHA,UST RADIATION recorder, 3-RR-90-140 on Panel 3-9-2. o

2. RX & REFUEL ZONE EXH CH A RAD MON RTMR, 3-RM-90-14lli142 on Panel 3-9-10. o

3. RX & REFUEL ZONE EXH CH B RAD MON RTMR, 3-'RM-90~141H43 on Pane! 3-9~10. o Corrtinuedon Next Page

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

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Panel 9-3 . l-ARP-9-3A 3-XA-55..3A *R.ev. 0037

.Ptlg~!it 9t~1.

RX ZONE EXH RADIATION MONITOR DNSC 3~RA-90-142B, Window 35 (Pagel of 2)

Operator Action: (Continued)

B. CHECK power supply to the radiation moniklrs. 0 C. CHECK other radialian monitors for radiation levels below release limits. 0 D. REFER TO Technical Specffication Section 3.3.62. and 3.3.7.1. 0 E. IF Groop 6 isoiation occurs, THEN REFER TO 3-AOI-64-2d. 0

References:

341E620-3 341E61 0-90-1 GE 3-130E921-21 Technical Specifications 3.3~6.2 and 3.3.7.1

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 295035EA2.02 Tier # 1 Ability to determine and interpret the following as they apply to 2 Group #

Secondary Containment High Differential Pressure: Off-site release rate. KIA # 295035EA2.02 Importance Rating 2.8 4.1 I Proposed Question: # RO 64 Given the following plant conditions:

Unit 2 is at 100% power.

During the backwash of a Reactor Water Cleanup (RWCU) Demineralizer, the Backwash Receiving Tank ruptured.

The RWCU system has been isolated.

Secondary Containment conditions are as follows:

- ALL Reactor and Refuel Zone radiation monitors trip on high radiation.

- NO Standby Gas Treatment (SGT) train can be started.

Refuel zone pressure: (-) 0.12 inches of water Reactor zone pressure: (+) 0.02 inches of water

- AREA RADIATION LEVELS RB EL 565 W, 565 E, 565 NE: 250 mr/hr RB EL 593 upscale RB EL 621 upscale Which ONE of the following describes the required action and the type of radioactive release in progress?

REFERENCE PROVIDED A. Initiate a shutdown per 2-GOI-100-12A, "Unit Shutdown." An Elevated radiation release is in progress.

B. Initiate a shutdown per 2-GOI-100-12A, "Unit Shutdown." A Ground-level radiation release is in progress.

C. Scram the reactor and Emergency Depressurize the RPV. An Elevated radiation release is in progress.

D. Scram the reactor and Emergency Depressurize the RPV. A Ground-level radiation release is in progress.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: B Explanation: The release from the Reactor Building is NOT elevated. This is a.

plausible because the required actions are correct except the differential pressure results in a ground-level release.

b. correct answer.

Conditions DO NOT warrant a scram at this point. In addition, the c.

release from the Reactor Building is NOT elevated. This is plausible if the candidate fails to recognize that a primary system is NOT discharging to the Reactor Building.

Conditions DO NOT warrant a scram at this point. This is plausible if d.

the candidate fails to recognize that a primary system is NOT discharging to the Reactor Building.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 2-EOI-3 Flowchart (Attach if not previously provided)

Proposed references to be provided to applicants during examination: 2-EOI-3 Flowchart Question Source: Bank # 295035EA2.02 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam x Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: In order to answer this question correctly the candidate must determine the following:

1. Which area(s) are above or approaching Max Safe.

2. Based on Item #1 above, determine the appropriate action.

3. Whether plant conditions indicate an elevated or ground-level release.

c NOTE: EOI-3 steps SC/R-8 and SC/R-9 apply, requiring shutdown per 2-GOI-100-12A because two (2) or more areas are above max safe rad levels; but, a primary system is NOT discharging to the Reactor Building. Insufficient Reactor Building-to-atmosphere dp (greater than -0.25 inches of water) indicates loss of secondary containment integrity. The positive Reactor Zone pressure is causing an unmonitored and uncontrolled ground-level release of radioactive contaminants.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BROWNS fERRY EMERGENCY CLASSIFfCATION PROCEDURE EPI*P-1 TERMIPHRASE MiEANINGfDEFINITION Projectile An object ejected, thrown, or launched towards a ptant structure. The source of a projectile may be offsite Of onsite. Damage is sufficient to cause concern regarding the integrity of the affected structure orihe operability or reliability of safety equipment contained therein.

Protected Area AJI areas within the security protected area renee.

PSIG Pounds Per Square Inch Gauge R Rad RGIG Reactor Core Isolation Cooling RCS Reactor Coolant System REP Radiological Emergency Plan RHR Residual Heat Removal RPS Reactor Protection System RPV Reactor Pressure Vesset Sabotage Deliberate damage, misalig.nment, misoperalion of plant eqUipment with the intent to render equipment inoperable.

SAMG Severe Accident Management GuideJine SEC Second Secondary The spaces immediately adjacent to or surrounding, the primary containment containment from which the Reactor Building Ventilation System and the Standby Gas Treatment System provides a filtered elevated release.

SED Site Emergency Director SGTS Standby Gas Treatment System Significant An unplanned event involVing one or more of the foltowing:

Transient (1) Automatic turbine runback greater than 25% thermal reactor power or (2) Electricallcad reduchongreater than 25% full electrical load, or (3) Thermal power osdllations greater than 10%, or (4) Reactor scram, or (5) Valid ECCS initiation.

PAGE 11 Of 2{)1 REVISION 42

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet OPL 171.067, HVAC Lesson Plan V. Training Objectives:

A. Terminal Objective Upon completion of this lesson, the operator will demonstrate satisfactory knowledge of Plant Ventilation, Heating and Air Conditioning (HVAC) Systems by scoring at least 80% (70% NLO) on a written exam.

B. Enabling Objectives (HLT/LOR)

1. Identify the purposes of the HVAC systems and how these purposes are accomplished.

2. Identify the relationships between HVAC systems and the following, evaluate how loss or malfunction of the following effects HVAC, and evaluate how loss or malfunction of HVAC effects the following:

a) AC Electrical Distribution b) Secondary containment c) Standby Gas Treatment (SGT) d) EECW e) Process Radiation Monitoring f) Control Air g) Process instrumentation (drywell pressure, reactor level) h) Static pressure control i) Fire Protection j) Control Room Habitability k) Area Temperatures (Reactor Bldg and Control Bay)

OPL 171.067 Page 15

a. Reactor zone and refueling zone static Obj. V.B.2 pressure is maintained at .25 inch water Obj. V.C.2 negative by use of larger exhaust fans and static pressure regulation. Dampers mounted in each fan inlet are designed to gradually close or open in response to static pressure regulators to maintain building pressure and ensure an elevated, monitored release.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL 171.067 Page 20

2. Normal ventilation flow path, ~P control, and Obj. V.B.2! V.B.S isolation capability ensure secondary Obj. V.C.8N.B.4 containment integrity during normal operation. ObjV.C.19 Loss of the normal HVAC system under isolation T.S. 3.6.4.1, 3.6.4.2 (PCIS) conditions does not affect secondary containment integrity if SGT auto starts and has a suction path to maintain pressure at -0.25 EFFECTIVE inches H20 inside the reactor building. This COMMUNICATION ensures leakage is into the building, and releases are elevated (stack) and monitored.

However if secondary containment integrity is lost, leakage may be at ground levels, and site boundary doses may be higher than calculated in the accident analysis.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO 295036G2.4.34 Tier # 1 Knowledge of RO tasks performed outside the main control room 2 Group #

during an emergency and the resultant operational effects:

Secondary Containment High Sump/Area Water Level. KIA # 295036G2.4.34 Importance Rating 4.2 4.1 I Proposed Question: RO # 65 Which ONE of the following EOI-3, "Secondary Containment Control" entry conditions CANNOT be determined using control room indications and the appropriate operator action if it has been exceeded?

The entry condition is (1) Operate ALL available floor drain and eqUipment drain sump pumps from the _ _ _ _~(2=+)----

(1) (2)

A. SECONDARY CNTMT FLOOR DRAIN main control room panel 9-4.

SUMP WATER LVL ABOVE 66 IN.

B. SECONDARY CNTMT FLOOR DRAIN radwaste control room.

SUMP WATER LVL ABOVE 66 IN.

c. ANY SECONDARY CNTMT AREA main control room panel 9-4.

WATER LVL ABOVE 2 IN.

D. ANY SECONDARY CNTMT AREA radwaste control room.

WATER LVL ABOVE 2 IN.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: B Explanation: a. Part (1) is correct. This indication is only available from the radwaste control room or loca"y at the sump. Part (2) is incorrect. The floor drain and equipment drain sump pumps operated from panel 9-4 are for the Drywe", not the Reactor Building.

b. Correct answer.

c. Part (1) is incorrect. This indication is available from high level annunciators for each identified area in the reactor building. If the annunciator is in alarm, the level in that area is above 2 inches. Part (2) is incorrect. Rx BLDG floor and equipment drain sump pumps are operated from the radwaste control room.

d. Part (1) is incorrect as stated in (c) above. part (2) is correct. Rx BLDG floor and equipment drain sump pumps are operated from the radwaste control room.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 2-EOI-3 flowchart (Attach if not previously provided) 1-ARP-9-4C (attached annunciator)**

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/07/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments: ** Other annunciators indicating >2 inches are also available in the control room. This is only one example.

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

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A.'JYS:ec~'A?<Y f:ml!.1T

!>c"U:J\ ',!k\l'Eil'l,LVL mc;,,'!;l

omt RESTORE ANLlMAINTAIN FLOOR OPJ-iJNSUMP ~VA!'ER L'Vb SELOIWei IN. USING ALL AV..DJLA.BLESiJMP Put.tlPS L

SC/l-4 NOL L

ISOLA,.

INTOTH

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet c BFN Panel 9-4 1-ARP..:9-4C Unit 1 1-XA-55-4C Rev. 0016 SensodTrip Paint:

SUPPR CHMBR RM FLOOD lEVEL HIGH 1-LS-n-25F 2::2 Inches ofWarer 00 the fioor 1-LA-77-25F rT (Page 1 of 1)

Sensor Sensoc is IoCBted near the Hoor of the Suppression Chamber room, Coluirnn R-4 loeation: N-Line.

Probable Greater than two inches {)f water on the floor.

Cause:

Automatic None Action:

Operator A. DISPATCH personnd Ie* vilWally check 1tle suppression chamber Action: room. []

B. IF alarm is valid, THEN PERfORM the following:

CHECK the floor drain sump pumpsruming. []

CHECK the floor drains for properdlllifl8lge. []

lif possible, THEN DETERMINE the source of the leak and the leak rate. []

EtffER 1-EOI-3 flowchart. []

NOTE The fioor drain and equipment drain sump pumps may need to be locked out to prewmt Radwaste Hooding.

NOTIFY Rad~ operatcrto monior drain caiBect<<tanlt and waste collector tank levels. []

NOTIFY Radiation Protection. []

ReferenCfi: 0-47E610-77-1 0-47ESOO-6

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO G2.1.2 Conduct of Operations Tier # 3 Knowledge of operator responsibilities during all modes of plant Group #

operation.

KIA # G2.1.2 Importance Rating 3.0 4.0 I Proposed Question: RO # 66 Given the following conditions involving Foreign Material Exclusion (FME):

An outage worker was placing a plastic FME cover on a vacuum breaker inside the Torus.

He inadvertently dropped the cover into the Suppression Pool.

The cover immediately sank to the bottom of the Torus.

The cover was still visible from the catwalk.

Which ONE of the following describes the required actions for this situation?

Work in the Torus (1) . The FME cover MUST be retrieved (2)

(1) (2)

A. MUST be stopped. before job closeout.

B. MUST be stopped. immediately.

C. may continue. before job closeout.

D. may continue. immediately.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: A Explanation: a. Correct answer.

b. Part (1) is correct. The job supervisor may eventually allow work to continue since the cover is visible, but all work is stopped until a retrieval plan has been finalized. Part (2) is incorrect. Torus work requires an underwater inspection by divers for closeout requirements. The cover can be retrieved then if it doesn't interfere with other work being done.

c. Part (1) is incorrect. The job supervisor may eventually allow work to continue since the cover is visible, but all work is stopped until a retrieval plan has been finalized. Part (2) is correct. Torus work requires an underwater inspection by divers for closeout reqUirements. The cover can be retrieved then if it doesn't interfere with other work being done.

d. Part (1) is incorrect. The job supervisor may eventually allow work to continue since the cover is visible, but all work is stopped until a retrieval plan has been finalized. Part (2) is incorrect. Torus work requires an underwater inspection by divers for closeout requirements. The cover can be retrieved then if it doesn't interfere with other work being done.

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ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): SPP-6.5, Foreign Material Control (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank# OPL 171.217.10 minor format changes Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: The requirement to stop work if FME control is lost has only recently been changed.

Previously, the Job Supervisor made the determination of whether to stop work or not based on the specific situation. New restrictions, based on continuing problems with FME control, have led to more restrictive requirements. Notice that the original question had distracters related to initiating a Problem Evaluation Report (PER). Only under certain conditions was a PER required. Now a PER is written for any FME issue. For that reason, I removed that section of the question because this fact has been widely taught to the entire plant population and would no longer be discriminatory.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Original Question OPL717.217.10:

An outage worker was placing a plastic FME cover on a vacuum breaker inside the Torus when he inadvertently dropped the cover into the Suppression Pool. The cover immediately sank to the bottom of the Torus, but was still visible from the catwalk. Select the actions required for this situation.

A. Work in the Torus must be stopped to immediately retrieve the FME cover. Initiating a PER is required.

B. Work in the Torus must be stopped to immediately retrieve the FME cover. Initiating a PER is NOT required.

C. Work in the Torus may continue. Initiating a PER is required. Retrieve the FME cover before job closeout.

D. Work in the Torus may continue. Initiating a PER is NOT required. Retrieve the FME cover before job closeout.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet WAN Standard foreigo Material CO'ntrO'I SPP-O.5 Programs aod Rev,0012 Processes Page 140'f 51 3.3 Work PerlomJance (continued' CAUTION faiture to' remO've mainternmce residue from radioactive components will result in production of radiation source term hazards to personnel.

13. While perfuFmiflg grinding O'rcutting activities, precautiO'ns shall !be made to reduce the mnount of detms produced. II. may be necessary to' ereei shielding to prevent deMs from maintenance activities from !becoming foreign material mtile system iI'Iqueslion or anO'ther crose by. Setting upa '{acuum during activities can be a. viable melhod for reducing debris.

14. Immediately following system breaches, the system or component shall be cleaned of roceign debris. Vacuuming is the Pfeierred method of deaning .

Cleaning shall be done beroce removal of pipe dams, pJ;ugs, O'r barriers, and should be done again after the devicces have been removed.

15. Stop all work jf FME control is IDst and investigate in accordance wilh Section 35.

16. For small items (screws, nuts, ll!lashers, etc.) that cannot be made fail-safe, consider the use O'f re-sealabie fail~sare oontainers, dO'uble bagging or other methods to lessen the prO'babllity of a loss of FME contro!event.

11. Temporary Tie Wraps shou4d be bright in rolO'r, non Metallic (not containing metallic locking tabs) and if at all poss&ble, float in order to fliIcifiitate recovery.

18. Use ofwife brushes on systems that oome in rontactwith the Reactor Coolant System fs prorubilted.

NOTE For Work involving. electrical and I 8. C oomponents or systems further guidance is pro-yi!ded in Appendix A.

J. Ensure any partsJparticles (particulartyvai'le stellite l1ard~fadng) are creaned from the system !before closure, 3.4 Suspension of the Job A. Often plant conditions change and work. within a FMEA is stopped for a period oftime.

Anytime continuous or immediate access through operung is not required or if work is stopped during iii shift, between shifts, or for a period O'f time the following actions are req.uired.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet TV.AN Standard Foreign Material Control SPP-6.5 PfOgrams and Rev. non Processes Page 16 of 51 3.4 Suspension of the Job (continued)

Tile following mateFiaisare unacceptable for use as tempomry plugs, covers or seals:

paper products or rags inserted into pipes or openings.

3.5 Recovering from Loss of FME Controls A.. For ail loss of FME control events, a PER shalt be initialed, and the following notifications shall be made:

1. Nolify the FME monttor.

2. Notify the Work Supervisor.

3. For evems associated with systems fuat connect with the Reactor Cool am, or Spent Fuel Cooling Systems, notify ShJft Operations or the Refuel F!oor Sentor Reactor Operator (RFF SRO).

4. Notify the Department Mamlger over the group performing the tasK, S. DuringoutBges, notify the Outage Contra! Center.

B. If the foreign material can be easily retrieved (Le., without further disassembMy of the system or component)., Stop, Review your actions, then the retrievai may be performed with slow deliberate moves to prevent lodging the foreign material deeper into the system or creating additional debris.

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C. If the fo.reign material can not be easily remeved, Stop, Notify the Work Supervisor and FME monitor tocoordinaie deveioprnen! of a retFieval pian.

D. If the missing item tan not be accounted for, the job supervisor shall determine jf work should be stopped and the item retrieved or it the search and retrieval wUl be performed just before closure.

E. Initiate a. request for Site Engineering to eVliduate the effects of the foreign material on the system(s.) or componenl(s) if not retrieved.

F. A Technical Evaluation in accordance with the requlrements of SPP-9.3 shatl be completed for items thai filII into the reactor vessel, reactor internals, spent fuel poolltransfer canal or into systems with a direct path to reactor vessel and which cannot be retrieved.

3.6 Completing the Job Closeout A Before Ii system opening is c!osed, the respansib4e supelVisorldesignee shal! ensure all PERfS ideniifying Foreign Material have been evaluated and the system determined to be acceptab0e for closure,

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO G2.1.41 Conduct of Operations Tier# 3 Knowledge of the refueling processes.

Group #

KIA # G2.1.41 Importance Rating 2.8 3.7 I Proposed Question: RO # 67 Which ONE of the following describes the proper orientation of a fuel bundle within a control rod cell while performing a Core Load Verification in accordance with O-GOI-100-3C, "Fuel Movement Operations During Refueling", Attachment 67 A. Channel spacer buttons are adjacent to the control blade and adjacent to each other.

B. The bundle serial number is readable as viewed from the center of the reactor core.

C. Channel fasteners for each bundle in the cell are aligned to the outSide corners.

D. Orientation boss on the lifting bails point towards the center of the reactor core.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: A Explanation: a. Correct answer.

b. Incorrect. The bundle serial number is readable as viewed from the center of the cell, not the core.

c. Incorrect. Channel fasteners for each bundle in the cell are grouped in the center of the cell.

d. Incorrect. Orientation boss on the lifting bails point towards the center of the cell, not the core.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 0-GOI-100-3C (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 06/17/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet SFN Fuel Movement Operations During 6~GOI-1 0O-3C Unit 0 Refueling Rev. 0062 Page 102 of 127 Attachment 6 fPage2of3}

Core Verification Verification of Bundle Orientation Verification of .bundle olientation requires two ,eviewers. The two reviewers must be famj*liar with the fuel design features listed below and with Figure 1.

A separate video scan is performed to vemy that all bund4es are oliented properly. TIle camera is positioned above the core to aNow one complete control cell to be viewed at a time.

The reviewers will be provided a core map which has the video path marKed with the beginning and endAng paints.

Experience has demonstrated that certain design features are clearly visible, so that any misoriented fue-I assembly will be readlly distinguished during core vermcauoo. Five separate visual indications of proper fuel assembly orientation forinleoor cells exist:

A. Channej fasteners for each bundle in the cen are grouped in Ihe center of the celt B. Orientation OOSS 00 the lifting bails point towards the center of the cam.

C. Channel spacer buttoos are adjacent to the control blade and adjacent to each other.

D. The bundle seliaJ number is readable as viewed from tne center of thecel!.

E_ There is c.eU-to-celi symmetry.

Additional care must beexerdsed when viewing the partial celts around the core peliphelY-Tne fuel assemblies in these oelis should havelhe same orientation as if the core contained a complete control cell including this fueL These IYundles may also be checked against approved c'Ore maps.

For peripheral bundle locations, pay particular attention to the channel spactng "between the peripheral bundle and the face adjacent bumUes. The spacing should be present and symmetric, and if not, further investigation and review is needed.

It was olJservedlhat this spadng did not exist when a bundle was misseated. The spacing loss was a result of the bundle leaning and touching its adjacent bundle due to oot being inserted into ttle peripheral support pioce.

\.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO G2.2.17 Equipment Control Tier# 3 Knowledge of the process for managing maintenance activities Group #

during power operations.

KIA # G2.2.17 Importance Rating 2.6 3.8 I Proposed Question: RO # 68 Which ONE of the following describes the management level required to provide FINAL approval of maintenance with a RED RISK classification in accordance with BP-336, "RISK DETERMINATION AND RISK MANAGEMENT?"

A. Risk SRO B. Shift Manager C. Operations Manager D. Plant Manager

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: 0 Explanation: a. Incorrect. Risk SRO approves Yellow Risk.

b. Incorrect. Provides approval for all normal GREEN activities.

c. Incorrect. Operations Manager is a member of the Critical Evolutions Review Committee, but does not provide final approval.

d. Correct answer.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): BP-336 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 06/17/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet YY.4.X STANDA.RD BltoWN!> FEKKYNVCLLUPL>\....--r BP...l,jij B~JSlNI!SS IWlKDETERl!IN.4.TION A.">> .lUSK J.L>\.....AGElu:::-.--r P2~6 0(44 PRACTICE Rni,i** 9M7 4.1.2 Classification of Repelitive ActMtres Of Tests 4.12.1 Repetitive activities or tests which have been previoos!y risk assessed and have not been revised since the last pelfarmance will not require a lleW assessment to be peffom1ed.

4.1.2.2 Poor to being inclooedin the schedule,. the work cordrol Operations Representative will review the previous assessment to ensure that it is stiU corred.

4.1.3 Schedule Risk Assess(f"';ent 4.1.3.1 The Operations Shift Mana:geriUnit Supervisor shal1 have ultimate responsibility for ensuring that the impact on O'Veralilllant risk is evaluated before systems or components are removed from service for maintenance Of surveillance activities.

4.2 Green Risk 42.1 Green Rlsk adiv:ities wm be performed in accordance with the NPG Hurnan Performance Pocket Guide.

4.3 Yellow Risk 4.3.1 Yef;JowR.isk acmritiP...s wil! be approved by the Risk SRO, or designee, and will be documented by signing the Work Implementation Schedule.

4.3.2 if the risk ctassffication is due to:.

PSAJPRA. - activfiies should oot be performed on equipment remaining in service whkch would cause the plant to enter the next level of fisk. Consideration shall be given to protecting this equipment as dee(f"';ed necessary by the Risk. SRO, refer to Atta.chmem 5.

SENTI NEt - activ:ities \I1iU be managed in Ii.ccordance with SPP-7.1.

DRAM - activities wil! be managed in accordance wfth SPP-7.2.

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

(

n'.!N Sr~"'D!l.KD BROWNS r.EKK~- NIJC[L,K Fl.A1',"! BF...lJ<6

&'SlNES;S mKDET~1IN.'nox Al',l) .!t!S:K M,\..'I'AGEMENr P..,S nfH PIIL!CTICE Rwm".. OOQ;T 4.4 Orange Risk 4.4.1 The Critical Evolutions Re...~w Cornmittee will approve ali Orange Risk. acti ....ities.

4A.2 If the risk. classification is due 10:

PSAIPRA - actMl:ie:s shall not be: performed on equipment remaining in service vffltch \IlOuld cause the plant to emerthe next Ie....el of risk.. Equipment shal be protected as deemed by the Risk SRO, refer to Attachment 5.

SENTINEL -activities shan oot be performed on equipment remaining in service which \IlOuld cause the plant to enter the next level of risk. Equipment shall be protected as deemed by the Risk SRO, refer to Attachment 5.

ORAM - adNitie:s MI! be managed in a,coordance with SPP-7.2.

ACTIVITY RISK. ASSESSMENT (AITACHMENT 2)-

acti ....ities whtch screen to orange risk Mil be rnanaged in accordance wilh the barriers and actions prescribed per the appiicable Attachment 2. A Responsible Task Lead will be assigned in accordance with SPP-1.1.

Management oversight should be considered by the Crmcm Evo.uUens Review Committee.

GRID RELIABILITY - na activities will be scheduled that \IlOuld increase the probabi~ity of loss af offsite power 0If statton blaiCkout. No activities willi be scheduled on equipment that is used to m~tigate the consequences of a loss of offsite power or station brackout. Any equipnlent thai already has been removed frofn servi,ce that meets the above criteria will be returned to functionat status as S00I18S possible.

Refer to Atta.chment 4 for systems/equipment that meets these cmeria,

PLANNED POWER REDUCnONS- activities wi~! be coordinated with load dispaich and readNity management ovemght winl be provided by Reactor Engineering and Operations. Outage Management win have overall responsibility for schedure development and ;preparal.ion. Management oversight w1t~ be provided as designated by the Critical Evolutiom Review Committee.

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

(

n*,~=- STANDl.R,D BROWNS FEKKY NUCUAKPLAZ'I"r BP-33i5 BUSINESS RISK Dn~lINAnON A.'I1)!tlSK M,Oi.4cG£MlZ'l"r P,.,9.U4 PRACTICE K ....:;i""QIlG7

REDUCED MARGi~ - activities will be a.ssigred to a Responsitle Task Lead per SPP-7.1 .. other equiprllent shal) be protected as deemed by the Risk SRO. refer to Atta.chment 5. Activities will be worked around the c£ock. Management oversight shoold be considered by the Critical Evolutions Review Committee.

TIME REMAINING ON SID lCO (:::. 7 DAYS BUT >72 HOURS) - actvfities will be worked aroondlhe clock and a Responsible Task lead win be assigned in accordance with SPP-7.1. Management o.... ersightshall be assigned as deemed by the Critical Evolutions Review Committee or Plant Manager/designee..

COMPLEXITY (~ TRIP, ACTUATION OR ISOLATION

& MULTIPLE LOCATIONS &:> ANNUAL FREQUENCY) - activities will be managed in accordance with the barners and actions prescribed pel" the app4icabre Attachment 2. A Responsible Tasl<

Lew wml be assigned in acwrdal1ce with SPP-1.1 as deemed by the Critical Evolutions Review Committee.

COMPLEX INTEGRATED .ACTIViTIES {'h'HICH CREATE RISK TO SHUTDOWN SAFETY OR GENERATION} - actMtiies will be assigned to a Responsible Task Lead per SPP-7.1 .. Management oversight gnatl be assigned as deemed by the Critical Evolutions Re..rew Committee or Plant Manager/designee.

OUTAGE ACTIViTIES - activities wil! be assigned to a Responsible Task Lead perSPP-7.1 . Management oversight snail be assigned as deemed by the Critical Ev-olutions R.eview Committee or Plant ManagerJdesignee. These activities may be managed per SPP-7.2 and CRAM as apptlcabte.

4.4.3 Level of Oversight and Super.,;jsion wiU be determined by the Critical Evokrlions Re ....iew Committee.

4.5 Red Risk 4.5.1 The Plant Manager, or designee, will appro.... e allRed Risk activities. The Crnical EvolutionsRevrew Committee will approve all Red Risk activities.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO G2.2.35 Equipment Control Tier# 3 Ability to determine Technical Specification Mode of Operation.

Group #

KIA # G2.2.35 Importance Rating 3.6 4.5 I Proposed Question: RO # 69 Given the following plant conditions:

Unit 2 is coming out of an outage making preparations for re-start.

Reactor Coolant temperature is 150 OF.

Mode Switch is in the REFUEL position.

As the Unit Operator, you receive a phone call from the Refuel Floor SRO to inform the Control Room that the Reactor Pressure Vessel (RPV) Head is fully tensioned.

Which ONE of the following describes the correct operating MODE in accordance with Technical Specifications?

A. Mode 2 B. Mode 3 C. Mode 4 D. Mode 5

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

( I Proposed Answer: C Explanation: a. Correct answer.

b. Incorrect. Reactor coolant temperature is less than 212 of.

c. Incorrect. This would be correct for the given conditions IF the Mode switch was in the SHUTDOWN position.

d. Incorrect. With the RPV head fully tensioned, the plant is no longer in Mode 5.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): U3 TSR Table 1.1-1 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/17/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Definitions 1.1 Table 1.1-1 (page 1 of 1)

MODES REACTOR MODE AVERAGE REACTOR MOOE TITlE SWITCH POSITION COOLANT TEMPERATURE (OF) 1 Power Opemtioo RtJn NtA 2 Startup Refuel(a) or StartupJHot NA standby

3 Hot Shotdown(a) Shutdown >212 4 Cold Shlltdao!m{a) ShLJtdown :'::212 5 Refueliflg{b) Shutdown or Refue! NA (a) Ali reactor vessel head closure OOtts fully tensioned.

(b) One or more reacwr -ves.sel head closure !:Jolts less than fuSly tensioned.

BFN-UN£T2 Ameoolnent No. 253

ES-401 Sample Written Examination Form ES-401*S Question Worksheet Examination Outline Cross-reference: Level SRO G2.2.7 Equipment Control Tier # 3 Knowledge of the process for conducting special or infrequent tests.

Group #

KIA # G2.2.7 Importance Rating 2.9 3.6 I Proposed Question: RO # 70 Which ONE of the following satisfies the criteria that identifies a "Complex Infrequently Performed Test or Evolution?" (CIPTE)

A. Primary System/Reactor Coolant System Barrier Hydrostatic Pressure Test.

B. Rod Worth Minimizer functional test prior to startup in accordance with I-GOI-IOO-IA.

C. HPCI Main and Booster Pump Set Developed Head and Flow Rate Test at Rated Reactor Pressure per 3-SR-3.S.1.7.

D. Functional test conducted per the Work Order to verify operability of the RHR Inboard Injection Valve following maintenance.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: A Explanation: a. Correct answer.

b. Incorrect. This is covered by an approved procedure, is not complex and does not pose an operational risk.

c. Incorrect. This procedure is complex and poses a potential operational risk, but it is covered by an approved procedure and performed routinely.

d. Incorrect. This procedure is not covered by an approved abnormal or normal procedure and does pose an operational risk, but is a simple evolution exempted from the criteria of a CIPTE.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): SPP-2.2 (Attach if not previously provided)

( SPP-8.1 Proposed references to be provided to applicants during examination: None Question Source: Bank#

Modified Bank # OPL 171.078.26 attached New Question History: Last NRC Exam Question Cognitive Lev~l: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

(

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet Original question OPL171.078.26:

.Infrequently performed tests or evolutions that have the potential to significantly degrade the plant's margin of safety that warrant additional management oversight and control are called "Complex Infrequently Performed Tests or Evolutions." (i.e., CIPTE)

Which ONE of the following is a criteria that identifies a CIPTE?

A. Tests/evolutions not specifically covered by existing normal or abnormal operating procedures.

B. Data taking, for example gauge reading, annunciator observations, data compilation, and inspection or inventory type tests.

C. Critical procedures such as Emergency Operating Instructions (EOIs), Abnormal Operating Instructions (AOIs) and Annunciator Response Procedures CARPs).

D. Functional tests conducted by work control documents such as stroking valves, bumping motors, calibrations, visual inspections, leak/pressure tests, and electrical continuity checks.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet NPG Standard Administration of Site Technical SPP-2.2 Programs and Procedmres Rev.OOiS Processes Page 28 of 42 5.0 DEFINmONS Affected Organization Review - A review by organizations for Impacts and implementation requirements in their area of responsibility asa result of a procedure or process change.

Approval Authority - The approval authority for a procedure :is the manager/supervisor of the organizationalunlt responsiible for the procedure. This authority shall not be routinely delegated to a lower level within the organization. Higher leve! managers may approve procedures within the area of responsibility.

Audit Trail - A QA record generated by electronic workflow roUfing which provides the procedure e~edronic review and approval information.

CompJex Infrequently Performed Tests or Evolutions -Infrequeotly performed tests or evolutions that have the potential to signif.icantty degrade the plant's margin of safety that warrant additional management oversight and contro!. The be~w criteria shall be used to identify these ty.pes of tests/e\loMions:

A. Testslevolutions not specifically covered by existing nomlal or abnormal operating procedures.

B. TestsJevolutions that are seldom performed even though covered by existing normal or abnormal procedures (for example., plant startup after a prolonged outage or after any outage that involves significant changes to systems, eqUipment, or procedtJres related to the core, reactivity control, or reactor protection).

C. Special,infrequently performed surveillance testing that involves complicated sequencing or placing the plant in an unusual configuration.

D. Testslevolutions that require the use of special test procedures in conjunction with existing procedures.

Critical Step - A critical step is a worK-related step or action that, if perfomledl incorrectly, wm Significantly harm plant equipment Of signfficantly impact plant operation. A step, action, or phase of a task is considered critical.. if it satisfies am of the fol:lowing conditions:

A. The consequences of incorrect performance are intolerable to reactor safety, generation, or to planteql..lipment (see DEFINITION of Intolerable Consequences to the Plant).

B. The consequences are realized immediately.

NOTE "Immediately~ should be construed as "during implementation of the procedure."

For exarnpJe, if the incorrect perfom1ance of Step A coul{j cause an immediate intolerable consequence when step Bis performed, then step A should be considered a crincal step.

ES-401 Sample Written ExaminatiQn Form ES-401-5 Question Worksheet WAN STANDARD SPP-8;.1 PROGRAMS AND CONDUCT OF TESTING Rev.*

PROCESSES .Page 5of15 1..0 PURPOSE This procedure describes Ule requiremenls., responsibilities, and administrative oonirols necessary fm" oonduding test activities at TVA Nudear's (TVAN's) facilities.

This procedure provides guidelines to help ensure that:

A. Test activities are periormed in a manner to reduce the possibUily of an undesired maniptJlatioo or actuation of plant equipment B. Test activities are performed by qualified personnet C. Personnel who pertiorm test actNities understand their duties and responsibilities.

A. This procedure applies to peI"SOOI1el involved in the administration, supervision, and performance of operations phase fest activities on installed equipment at TVAN's facilities.

Test activies on equipment oot under jurisdictiOOllI control of the Plant Manager on unlicensed units are excluded from this procedure.

B. This procedure contains the minimum requirements for conducting !he following tests except as noted in fiem 2.0.C below.

1. Surveillance tests

2. Functionaa tests

3. Postmaintenan~ificalion tests

4. Comp4ex Infrequienity Performed Tests or Evolutions (CIPTEs) induding Specra!

Tests

5. Technical refated Instructions that peffonn a test (te.. , Technicallnstrudions, etc.)

6. ContractorNendor test C. The following actNities must also be performed in accordance with this procedure, bt.rt do not require Form SPP-I:t 1-1, "Test DirectDr Assignment Sheet: Form SPP-B.1-2, "Chronological Test Log (CTl),~ or pretmef unless requred by Operations.

1. Simple functional tests conducted by work comroI documenl8, such as stding va4ves, bumping motDrs, simple calibrations, and visual ilspectionsOf" site implementing procedures, such 8S leak/pressure tests, electrical c:onmuily.

2. Tests 'that do net requre changes in plant SIta1us and/or configuration SUlCI"! as data taking (e .. g., gauge reading, annunciator obse4"Vations, Gata compilation),

inspection Of" inventory type tests, or inspections.

3. Simple tests performed 00 a routine basis such 8S RadioChem lab sampling instructions.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO G2.3.15 Radiation Control Tier# 3 Knowledge of radiation monitoring systems.

Group #

KIA # G2.3.15 Importance Rating 2.9 3.1 I Proposed Question: RO # 71 Which ONE of the following Process Radiation Monitor systems will NOT be adversely affected by a loss of Reactor Protection System (RPS) bus A?

A. Main Steam Line radiation monitors.

B. Refuel Zone Ventilation radiation monitors.

C. Reactor Zone Ventilation radiation monitors.

D. Control Room Emergency Ventilation radiation monitors.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: D Explanation: a. Incorrect. Two channels are powered from each RPS bus.

b. Incorrect. Two channels are powered from each RPS bus.

c. Incorrect. Two channels are powered from each RPS bus.

d. Correct answer.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): OPL 171.033 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank # OPL 171.033.18 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.033 pages 25 and 35:

1. Reactor Building/Refuel Zone Ventilation Radiation Monitoring System (RM-90-140, 141, 142, and 143)

a. Purpose (1) Indicates whenever abnormal amounts of radioactive material exists in the exhaust plenum of the reactor building/refuel zone and isolate the inlet and exhaust air flows

b. Four gamma sensitive GM instrumentation channels monitor the radiation from the reactor zone exhaust and four identical channels monitor the radiation from the refueling zone (1) These are physically located on the side of the ventilation ducts on the refuel floor (2) Monitors RM-90-140(A & B) and 90-142(A & B) are fed from RPS'A' (3) Monitors RM-90-141(A & B) and 90-143(A & B) are fed from RPS'B'

2. Control Room Ventilation Radiation Monitoring System (90-259 A and B)

(1) The external power supplies for the control room radiation monitor assemblies are:

(a) O-RM-90-259A - Panel 1-9-9 Cabinet 2 (Unit 1 I&C A)

Breaker 222 (b) O-RM-90-259B - Panel 3-9-9 Cabinet 3 (Unit 3 I&C B)

Breaker 325

3. Main Steam Line Radiation Monitoring System (1) RPS "A" supplies "A" (RM 90-136-A1) and "C" (RM-90-137-A2) radiation monitors. RPS "B" supplies "B" (RM-90-138-B1) and "0" (RM-90-139-B2) radiation monitors

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO G2.3.4 Radiation Control Tier# 3 Knowledge of radiation exposure limits under normal and Group #

emergency conditions.

KIA # G2.3.4 Importance Rating 3.2 3.7 I Proposed Question: RO # 72 Given the following plant conditions:

A Fuel Pool cleanout is in progress of Unit 2.

A failure of the Refuel Bridge monorail hoist allowed a bucket of irradiated stellite ball bearings to be raised above the fuel pool water level.

The Refuel Floor radiation levels initiated a Group 6 isolation on all three units.

The AUO operating the Refuel Bridge received a dose of 12 rem while manually lowering the bucket below the water level.

Which ONE of the following describes whether or not this constitutes an emergency exposure in accordance with EPIP-15, "Emergency Exposures" and the basis for this conclUSion?

In accordance with EPIP-15, "Emergency Exposures," this (1) constitute an Emergency Exposure. The basis for this conclusion is (2)

(1) (2)

A. DOES only 25 rem is allowed for equipment problems.

B. does NOT only planned exposures are covered under EPIP-15 limits.

c. DOES spontaneous actions taken to mitigate a problem are covered by the EPIP-15 exposure limits.

D. does NOT this exposure is classified as a Planned Special Exposure.

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

( I Proposed Answer: B Explanation: a. Incorrect. The 25 rem limit is not applicable for this situation.

b. Correct answer.

Incorrect. Spontaneous actions taken to mitigate a problem are NOT

c. covered by the EPIP-15 exposure limits.

d. Incorrect. This exposure was not planned, it was spontaneous.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): EPIP-15, "Emergency Exposure". (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank # OPL 171.000.04 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BROWNS FERRY EMERGENCY EXPOSURES EPIP-15

1.0 INTRODUCTION

1.1 Purpose This pl"ooedlJre pro'lliides9IJiid",n;;:e for aufuorizations af pEI"sonnel dose Ilimits undEl" emergency conditions as described in EPA-400. 'Mal'lUal of Proteclive Aclio;n Guides and Protective Actions fOr NtJclear incidents".

These limits ap~ only to emergencyexpOSUl"e authorizations and not to spontaneous reactions by iodividua!s attempting to mitigate an emergency snuation. This procedure pl"lWidesguidance for Ihe awroYa# and administration of radiation exposures received during emergencies in excess of 10 CFR 20. 12!)1 entitled "Oooupaoonal Dose Limits fOr AidUilts". This procedure does not pl"ovide diJreeoon fOE the a-pproyaland administrafio;nfor eX.posures received <ilJrlng 04her aclNmes inllclving radialion dose to indiYiduals.

NOTE: FlX the pUi!pOse af thfS implementing procedure, radiatio;n exposure as expressed! in unit of Rli1r and slJb-unis,thefeof,are ~UJiy.rn to dose (rad) and doseequi'"alent (rem} based an ANSI N 13 .. ;1 development and terminology. ArMY acute diose greater than 10 rem is generally denoted tn urns af lad since that level is considered as the accident range of personnel exposure. Any dose less than that level is considered as the protective range. af personnel exposure.. For purposes oHms procedlJre the

=

assumptions *ct t rad 1 rem is assumed! for ali levels of exposure.

2.0 REFERENCES

A. EPA-400-R92-DO 1, "Manual of Protective Action Geidies and P*rotecliye AOOons for Nuclear If'lCidients" B. m CFR 20. 12Cl1, Code ot Federal Re9IJ!a!ions 2.2 P1ant Instructions A. TVA Radiological Emergency Plan B. EPlP-iO, "Medical Emergency Procedure" C. EPIP-14, "Radiologioa! Control Procedures" D. SPf' 5.10, "Radiological Respiraiory Protection Program" E. spp 5.1, "Radiological Controls" PAGE 1 OF 1 REVISION 0000

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BROWNS FERRY EMERGENCY EXPOSURES EPIP-15 3.3.7 Persoonel shall not enler an)' Mea where doSE' rates are unknown or not measurable wfihinstrumenls and dosimetry immedia'.ely available.

NOTE: The \larue below corresponds to the ratio of external (measuredl dose raie to estimate TEDE dose, in accordance wfih default .... alues in TVA's !)ase Assessment model. When accident specific nUClide assessment are avaiJiable. more defirlilt4ve dose assessmenlsshouid be perlormed to adjust the correclion factors.

3.3.8 Unfil isotopic assessments of airba.rne radioactivity are availalile. an administrative correction factor of.2 soould be used 10 eslimate TEDE exposures in airborne activity areas:

Estimated TEDE = !)asmeter Reading; X 2 3.'11 DOSE! limits for Workers During: Emergencies 3A.1 Doses toalll workers diuring emergencies shookt to the extent prdcable be limited to 1G eFR 20.1201 limits. There are .. howe ...er. some emergency situations for which higher emerg:enCYE!xposures may be jlJstifJed. Whenever these situations are justified and ALARA considerations have been evaluated fuefoliOlfling limits can be adminisaed.

3A.2 Radiiaoon Protection (RP) considers ~ to-date annual 3CCTued dose to indi\lidl.llals when establishing !he maximum dose limits for workers during emergencies. The b-date annual accrued dose wootd be soolracted from the applicab:.!e emergency dose limit to determine the authorized allowable dose fodhe emergency.

3.4.3 Dose limits for the Protection of Valuable Properly Dose Limit (Rem) R~ptor

'!1J Whole Body (TEDE}

3{} lens of the E:;e tOO AI! Olher Organs 3.4.4 Dose limits for lii'esaviillS Aclivit.ies and the Protection of large Popul.a1Jons Doose Limit 4Rem} R~ptor 25 Wi"~ Body (TEDE) 75 lens of the E:;e 250 Ali Other Organs PAGE 30F7 REVlStOiNOOOO

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO G2.4.20 Emergency Procedures/Plans Tier # 3 Knowledge of operational implications of EOP warnings, cautions Group #

and notes.

KIA # G2.4.20 Importance Rating 3.8 4.3 I Proposed Question: RO # 73 Given the following Unit 1 plant conditions:

A reactor scram has occurred.

HPCI is needed to maintain reactor water level.

Suppression pool temperature is 145 of.

Which ONE of the following describes the reason HPCI is operated with a suction from the CST if possible?

A. The HPCI turbine exhaust pressure is likely to exceed the Primary Containment Pressure limit.

B. The HPCI pump shaft seals are not designed to operate at temperatures in excess of 140 of and may fail.

C. The suppression pool provides insufficient NPSH to the HPCI pump and cavitation may occur at rated flow.

D. The HPCI lube oil will exceed allowable temperatures and the HPCI function could be lost due to damaged bearings.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: D Explanation: a. Incorrect. This condition is possible! but is an issue with SP level below 12.75 feet! not high temperature.

b. Incorrect. HPCI pump shaft seals are capable of operating at higher temperatures even though they are cooled by water from the suppression pool.

c. Incorrect. NPSH will be lower at high SP temperatures! but will remain within the allowable levels for HPCI operation.

d. Correct answer. High lube oil temperatures degrade the viscosity of the lube oil and could result in bearing damage.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 1-EOI-1 Flowchart (Attach if not previously provided)

EOIPM Proposed references to be provided to applicants during examination: None Question Source: Bank # OPL171.201.11 minor format changes Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Original Question OPL171.201.11:

A reactor scram has occurred and HPCI is needed to maintain reactor water level. Suppression pool temperature is 14SoF.

SELECT the statement that correctly describes the reason HPCI is operated with a suction from the CST if possible.

A. The suppression pool provides insufficient NPSH to the HPCI pump and cavitation may occur at rated flow.

B. The HPCI pump shaft seals are not designed to operate at temperatures in excess of 1400 F and may fail.

C. The HPCI lube oil will exceed allowable temperatures and the HPCI function could be lost due to damaged bearings.

D. The HPCI turbine exhaust pressure is likely to exceed the turbine trip setpoint.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.201 page 38:

1. Caution #6 "Operating HPCI or RelC Turbines with suction temperatures above 140 of may result in equipment damage"

a. The HPCI and RCIC Lube Oil Coolers are cooled by routing part of the pump discharge fluid to the cooler. At elevated temperatures in the suppression pool, the turbine lube oil may get too hot to provide adequate lubrication.

b. Only during EOI operations will the system be needed at such an extreme suppression pool temperature. Therefore, the EOls are an appropriate location for this caution.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO G2.4.23 Emergency Procedures/Plans Tier# 3 Knowledge of the basis for prioritizing emergency procedure Group #

implementation during emergency operations.

KIA # G2.4.23 Importance Rating 3.4 4.4 I Proposed Question: RO # 74 Given the following plant conditions:

An accident has occurred on Unit 1 which has resulted in entry into Severe Accident Management Guidelines (SAMG).

SAMGs are being implemented from the Technical Support Center (TSC).

Which ONE of the following procedure classifications is inappropriate to be used in conjunction with SAMG implementation?

A. Abnormal Operating Instructions (AOI).

B. Emergency Plan Implementing Procedures (EPIP).

C. Emergency Operating Instruction (EOI) Flowcharts.

D. Emergency Operating Instruction (EOI) Appendices.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: C Explanation: a. Incorrect. Although some AOI guidance may conflict with SAMG guidance, this is NOT prohibited. However, SAMG implementation takes precedence over AOI guidance IF a conflict exists.

b. Incorrect. EPIP implementation is authorized and will certainly be required under the given conditions.

c. Correct answer.

d. Incorrect. SAMG procedures have several Appendices specific to SAMG implementation, but some EOI Appendices are still appropriate and are used.

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): OPL 171.212 page 6, 0-SSI-001 (Attach if not previously provided)

OPL 171.201 page 22 Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 09/18/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Excerpt from OPL171.212 page 6:

A. EOI Transition into SAMG - Loss of Coolable Geometry

1. The SAMGs are entered, then the core geometry is assumed to be changed and NOT coolable. The EOI strategies are employed for accidents inside BFN design basis. When accidents progress to a point where BFN design basis is exceeded, SAMG entry will be required.

3. At each of these specific points, we cannot assume a coolable geometry exists and SAMG entry is required.

4. Once the SAMGs are entered, the EOI flowcharts no longer apply because the configuration of the core may no longer be amenable to adequate cooling. All EOI flowcharts will be exited and will not be referred to again. Any subsequent EOI entry condition which is received will NOT result in EOI entry.

6. Other procedures (AOls, ARP's, EPIPs, etc.) have event specific entry conditions and may be used to supplement SAMGs. The control room staff may continue to use several such procedures in response to lower-level plant alarms, lineups, etc.

Actions that contradict any direction provided by the SAM Team shall NOT be performed.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level SRO G2.4.31 Emergency Procedures/Plans Tier# 3 Knowledge of annunciators alarms, indications or response Group #

procedures.

KIA # G2.4.31 Importance Rating 4.2 4.1 I Proposed Question: RO # 75 Unit 3 is performing a Reactor Startup with the following conditions:

RPV pressure is at 750 psig.

Control Rod withdrawal is in progress.

Reactor power is at Range 6 on the IRMs.

The Woodward Governor for 3A RFP fails upscale and the Reactor scrams on APRM High-High.

The Operating Crew stabilizes the Unit.

After the scram is reset the OATC notes the following annunciators:

- DRYWELL PRESSURE HIGH HALF SCRAM (9-4A W8).

- DRYWELL TEMP HIGH (9-3B W16).

- DRYWELLjSUPPR CHAMBER RADIATION HIGH (9-7C W15)

- OG PRETREATMENT RADIATION HIGH (9-3A W5)

Which ONE of the following actions should the Unit Supervisor direct to be completed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ?

A. Inject Standby Liquid Control.

B. Place SJAEs on Auxiliary Boiler Steam supply.

C. Open 3-FCV-1-56 Main Steam Line Drain.

D. Place Steam Seals on Auxiliary Boiler Steam supply.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: A Explanation: a. Correct answer.

b. Incorrect. Transferring SJAEs is not required.

c. Incorrect. 3-FCV-1-58 and 59 should be opened, not 3-FCV-1-56.

d. Incorrect. There is no requirement to place steam seals on Aux Boiler steam. The requirement is to adjust the regulator to zero.

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank # OPL 171.03362 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet SFN Panel 9-7 3-ARP-9-1C UnitJ 3-XA-55-1C Rev. 0026 Page 20 of 42 SensorJTrip Point:

DRYWELUSUPPR CHAMBER 3-RM-90-272A 100 RlHR RADIATION HIGH 3-RM-OO-213A 160 RlHR 3-RA-90-272 3-RM-OO-272B 9.99 E6 RfHR (Disabled Alarm input) 3-RM-OO-273B 9.99 ES RfHR (Disabled Alarm Input)

(Page 1 012)

Sensor 3-RR-90-272, Panel 3-9-54, Main Control Room.

Location: 3-RR-90-273, Panel 3-9-55, Main Control Room.

Probable A. H~h radiation levels.

Cause: B. Sensor malfunction.

C. Noise spikes.

Automatic None Action:

Operator A. VERIfY a.larm on 3-RR-OO-272 (Panel 3-9-54), and 3-RR-90-213 Action:. (Panel3-9-5S). o B. CHECK 3-RR-90-2:56 klr rise. o C. ATTEMPT to isolate equipment to stop source. o D. IF the alarm is determined to be valid, THEN PERFORM the following within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the alarm:

OPEN UPSTREAM MSL DRAIN TO CONDENSER 3-FCV-OG1-OO58. o

OPEN DOWNSTREAM MSt DRAIN TO CONDENSER 3-fCV-001-OO59. 0

ENSURE 3-PCV-001-1J147 is Closed by-taking STEAM SEAL REGU\fi,TOR, 3-HS-1-147 to CLOSE. (PaneI9-7) 0 CQntinued on Next Page

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ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panel 9-7 3-ARP-9-7C Unit 3 3-XA-55-7C Rev. 0026 Page 21 of 42 ORYWELLlSUPPR CHAMBER RADIATlON HtGH 3-RA.-90-272, Window '15 (Page 2 of 21 Operator Action: (Continued}

E. IF ALL the fol!owi!l1g coo(iitions exist (E.l ,E.2, and E.3):

1. Alarm is dererm~ned to be valid, .AND

2. The reactor will remain subcritical wfthout boron injection under all conditions, AND

3. leakage of primary coo4'ant mlo primary containment is indicated, THEN Wi!t1in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of alarm, fNJECT SLC for alternate source term control by placing SLC PUMP 3AI3B, 3-HS-63-6A in the START A or STA.RT B position. 0 F.. REFER TO EP1Ps. 0 G. IF started at Operator Action Step E, THEN WHEN SLC tank reaches 0', STOP the running SLC Pump 0

References:

345E620-9 047E61 0-90-2 NESSD 3R-09Q-213A-OQ Technical Specifications Section 3.3.3.1 NESSD 3R-090-272A-OO

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panel 9-1 3-ARP-9-1A Unit 3 3-XA-55-lA Rev. DOll PaSJe 1I) ot51 SensoclTrip Point:

OG PRETREATMENT HI RADIATION 3-RM-90-157 15.95 RlHR HIGH 3-RA-90-151A (Page 1 of 2)

Sensor RE-9G-157, Tum Bldg OG pretreatment sample chamber, Location: E1565', T-14 B-UNE Probable A. High radiatioo in the off-ga.s pretreatment system.

Cause: B. Resin trap failure {RWCU or COM Demin).

C. Possible fuel element failure.

D. Sensor malfunction.

Automatic None Action:

OperatOf" A. VERIFY high radiation 00 following:

Action: 1. OFFGAS PRETREATMENT RAD1ATION recorder, 3-RR-90-157 00 Panel 3-9-2. o

2. OFFGAS RADIATION recorder, 3-RR-90-160 on Panel 3-9-2. o

3. OG PRETREATMENT RA.D MON RTMR, 3-RM-90-157on Panel 3-9-10. o

4. OFFGAS RAD MON RTMR, 3-RM-90-160 on Panel 3-9-10. o B. CHECK off-gas flow normal. 0 C. CHECK following radiatioo recorders and assoc:iated radiation monitors:

1. MAIN STEAM LINE RA.D£AT!ON, 3-RR-90-135 00 Pi!lneI3-9-2. 0

2. OFFGAS POST-TREATMENT RA.DI.A.nON, 3-RR-90-265 on Panel 3-9-2. 0

3. STACK GAS/CONT RM RADIATION FROM STACK GAS, O-RR-90-147 on Panel 1-9-2. 0 D. NOTIFY RADCON. 0 E. REQUEST Chemistry peIfoml radiochemical analysis to determine source. 0 Continued on Nex.t Page

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BfN PanelS-3 3-ARP-9-3A Unit 1 l-XA-55-3A Rev. DOl7

.Pil9~1J ot?1 OG PRETREATMENT RADIATfON H~GH 3-RA-90-157A, Window 5 (Pagel of 21 Operator Action: (Continued}

F. If Offgas System Isolation Valve, 3-fCV-6-23 is manually restrained in the OPEN position and it has been determined that this is a valid alarm, THEN UNRESTRAIN Offgas System isolation Valve, 3-FCV-66-28.

G. REFER TO 0-5!-4..8.8.1.a.1 and 1(2)(3)-SR-3.4.6.1{A)forODCM compliance and to determine if power level reduction is required. 0

!-t If directed by Unit SupeNisor, THEN REDUCE reactor power to maintain off-gas radiation wrthin ODCM limits. 0 I. IF OOCM limits are exceeded,. THEN REFERfO EPIP-1. 0

References:

3-45E620-3 3-47E61 0-00-1 GE 3-7.29E814-4 3-SlMI-OOB

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet SFN Panel 9-3 J..ARP..:9-lB Unit 3 J-XA-55-JS Rev. 0018 Sl!flSOfiTrip Point:

DRYWEU_

TEMP HIGH l1E-64-S2C ~1*54"'f 3-TA-64-52 (Alarm oomes off recorder 3-XR-64-50)

{Page 1 of 1}

Sensor l1E-64-52A Location: Rx Bldg (Drywall)

E!5EW 225° (AZ)

Probable A Ofyweil cooler(s) failure.

cause: S.. loss of RBCCW to DTyweIl CooIer(s).

C. Possible leak in Dfywelt Automatic Action:

A CHECK Orywetl temperatures and pressures using multiple indications. []

S.. VERIFY Drywel! coolers mooing and START spare Orptell CooIer(s). []

C. CHECK OPEN RBCCW PRI CNTMl" OUTlEf VALVE, 3-HS-70-47A (Panel 3-9-4). []

D. START additionm RCW pumps. []

E. IF high Drywell temperature continues, THEN REFER TO 3-A0I-64-1. []

F. IF high Drywel! temperature is due to a loss of RBCCW, THEN REFER TO 3-A0I-70-1. []

G. IF temperature is above 100Of, THEN ENTER 3-EOI-2 Aowchalt. []

References.: 3-45E620-3 3-47E610-64-1 47W600-90

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panel 9-4 Unit J 3-X-55-4.A SensoriTrip Point:

DRY\lVEll 3-PIS-064-0056A 2.45 psi PRESSURE HIGH 3-PIS-064-0056B positive HAlFSGRAM 3-PIS-064-OO58C pressure in 3-PIS-064-00580 thedryweiL (Page 1 of 1)

Sensor 3-PNLA-009-0083,OO84, 0085, 0086 U3 Au:>:: Inst Room Location:

Probable A. :<=2.45 psig in fuedrywelL Cause: B. Sensor malfunction.

C. 51 or SR in progress.

Automatic A. Half Sicram if one sensor actuates.

Action: B. Reactor scram :and group 2, 8 and 8 PCIS if one sensor per chann~ actuates.

Operator A. VERIFY alarm by mutliple indications. . []

Action: B. IF dryweil pressure is :<=2.45 psig AND reactor has NOT scrammed, THEN MANUALLY SCRAM !he reactor. ENTER 3-E01-1 FlOWCHART &.

3-EOI-2 FLOWCHART. []

C. DISPATCH personnel to the pressure switches to check for abnormalcondioon. []

O. IF alarm is NOT valid or initiating condition is corrected, THEN wih SRO permission, RESET Ha/:f Scram. REFER TO 3-01-99. []

References*: 3-45N620-5 GE 730E915 FSAR. Section 7.:U.S and 13.6.2

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1.0 INTRODUCTION

2.0 REFERENCES

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