Text

Initial Exam 2008-302 Draft SRO Written Exam
	ML091810120
Person / Time
Site:	Browns Ferry
Issue date:	06/30/2009
From:	; NRC/RGN-II
To:	;
References
	Download: ML091810120 (283)
	v • d • e

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 29S001AA2.0S Tier# 1 Ability to determine and interpret the following as they apply to 1 Group #

Partial or Complete Loss of Forced Core Flow Circulation: Jet pump operability. KJA# 295001AA2.05 Importance Rating 3.1 3.4 I Proposed Question: SRO #1 Given the following conditions:

Unit 1 is starting up following refueling on 12/17/2008.

Power level is being held at 30% to allow maintenance on RFPT lB.

1A Recirc VFD inadvertently tripped at 1400 due to human error locally at the VFD.

1A Recirc Pump was restarted at 1415.

Which ONE of the following describes the requirement for performing 1-SR-3.4.2.1, "Jet Pump Mismatch and Operability?"

Surveillance 1-SR-3.4.2.1, "Jet Pump Mismatch and Operability" is required to be performed NO LATER THAN (1)

a. 1815 on 12/17/2008

b. 1800 on 12/18/2008

c. 1415 on 12/18/2008

d. 1400 on 12/18/2008

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

b. Incorrect. The allowance for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is only appropriate if 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months have passed since the recirc pump was started and power level remained below 25%. In addition, the time limitation begins when the pump is started, not tripped.

c. Incorrect. The allowance for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is only appropriate if 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months have passed since the recirc pump was started AND power level remained below 25%.

d. Incorrect. The allowance for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is only appropriate if 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months have passed since the recirc pump was started and power level remained below 25%. In addition, the time limitation begins when the pump is started, not tripped.

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): U1 TSR 3.4.2, U1 TSB 3.4.2, 1-01-68 (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 X 7/6/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X Comments:

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

3. 4.2 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.2 Jet Pumps LCO 3.4.2 All jet pumps shail be OPERABLE APPLICABiLITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TlME A One or more jet pumps A.1 Be fnMODE 3. 12 hoors inoperable.

(

BFN-UNIT 1 3.4-5 Amendment No. 234

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Jet Pumps 3A2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.2.1 ----------------NOTES------------

1. Not required to be perfooned until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after associated recirculation loop is in operation.

2. Not required to be perfooned until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after > 25°10 RTP.

Verify at least one of the following criteria (a, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months b, ore) is satisfied for each operating redrcuJation loop:

a. Recirculation pump flow to speed ratio differs by ~ 5% fromestablistied patterns, and jet pump loop flow to recirculation pump speed ratio dlffers by 5 5% from established patterns.

b. Each jet pump diffuser to lower plenum differential pressure differs by ~ 20% from established patterns.

c. Each jet pump flow differs by 510% from establ,ished patterns.

BFN-UNIT 1 3.4-6 Amendment No. 234

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet BASES SURVEILLANCE SR 3.4.2.1 (continued)

REQUIREMENTS Note 2 allows this SR not to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER exceeds 25% of RTP. During low flow conditions, jet pump noise approaches the threshold response of the associated flow instrumentation and precludes the collection of repeatable and meaningful data. The 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is an acceptable time to establish conditions appropriate to perform this SR.

REFERENCES 1. FSAR, Section 14.6.3.

2. GE Service Information Letter No. 330, "Jet Pump Beam Cracks," June 9,1980.

3. NUREG/CR-3052, "Closeout of IE Bulletin 80..;07: BWR Jet Pump Assembly Failure," November 1984.

4. NRC No.93-102, "Final Poticy Statement on Technical Specification Improvements," July 23, 1993.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 295006G2.2.42 Tier# 1 Ability to recognize system parameters that are entry-level 1 Group #

conditions for Technical Specifications: Scram KIA # 295006G2.2.42 Importance Rating 3.9 4.6 I Proposed Question: SRO #2 Given the following conditions:

Unit 1 is operating at 100% rated power.

Both Reactor Recirculation Pumps have tripped.

A manual scram was initiated in accordance with 1-AOI-68-1A, "Recirc Pump Trip/Core Flow Decrease OPRMs Operable."

Which ONE of the following describes the relationship between the required actions of 1-AOI-68-1A and the required actions per Technical Specifications for the given conditions and the basis for that action?

The requirement to initiate a scram in accordance with 1-AOI-68-1A is _ _--'(1,..:!1:...J..)___ with respect to the required actions per Technical Specifications. The basis for that action is (2)

(1) (2)

a. consistent the FSAR analysis for a DBA requires both recirc pumps to be operating.

b. consistent to place the plant in a MODE where the LCO does not apply.

c. conservative the FSAR analysis for a DBA requires both recirc pumps to be operating.

d. conservative to place the plant in a MODE where the LCO does not apply.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: 0 Explanation: a. Part (1) is incorrect. A manual scram is conservative with respect to Tech Spec requirements. TSR 3.4.1.8 provides 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to be in Mode 3. Part (2) is also incorrect. The FSAR analysis "assumes" both recirc pumps are in operation but does not "require" this conditions provided adjustments are made to thermal limits with the required timeframe.

b. Part (1) is incorrect. Part (2) is correct. Although the actions required by 1-AOI-68-1A are conservative compared to Tech Specs, the end result is the same. Operation in Mode 3.

c. Part (1) is correct. Part (2) is also incorrect. The FSAR analysis "assumes" both recirc pumps are in operation but does not "require" this conditions provided adjustments are made to thermal limits with the required timeframe.

d. Correct answer

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 1-AOI-68-1A, U1 TSR 3.4.1 (Attach if not previously provided)

U1 TSB 3.4.1 Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 7/6/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43* X Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Recirculation Loops Operating 3A1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LCO 3.4.1 Two recirculatton loops With matched flows shall be in operation.

OR One recirculation r,oop may be in operation provided the following limits are apptfed when the assoclated LCO is applicable:

a. LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR), Single loop operation lim~ts specified in the W

COLR;

b. LCO 3.2.2, "MINIMUM CRITICAl POWER RATIO {MCPR),"

single loop operation limits specified In the COL R;

c. LCO 3.3.1.1,"Reactor Protection System (RPS)

Instrumentation,' Function 2.b (Average Power Range Monnors Flow Biased Simulated Thermal Power - High). Allowable Value of Table 3.3.1.1-1 is reset for single loop operation.

APPLICABILITY: MODES 1 and 2.

BFN-UNIT 1 3.4-1 Amendment No. ~266 December 29, 2006

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Recirculation Loops Operating 3.4.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A Requirements of the lCO A.1 Satilsfy the requi rements 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months oot met. of the LCO.

S. 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 nne of Condition A not met

-OR No recirculation loops in operation.

(

BFN-UNIT 1 3.4-2 Amendment No. ~266 December 29, 2006

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Recire Pump Trip/Core Flow Decrease 1-AOI-68-1A Unit 1 OPRMs Operable Rev. 0002 Page6of12 4.0 OPERATOR ACTIONS 4.1 Immediate Actions None 4.2 Subsequent Actions

[1] IF both Recire Pumps are tripped in modes 1 or 2, THEN (Otherwise N/A)

[1.1] SCRAM the Reactor. D CAUTION INERICJ Failure to restart Reador Recirculation pumps in a timely manner may result in exceeding the differential temperature limit for pump start and subsequently require plant depressurization to avoid exceeding pressure-temperature limits for the reactor vessel. tsER 93-005]

[1.2] RESTART affected Reaetor Recirculation pumps. Refer to 1-01-68 Section 8.0. D

[2] IF the AT between the Rx vessel bottom head temperature and the moderator temperature precludes restart of a Recife pump, OR forced Recirculation flow CANNOT be established for any reason, THEN (Otherwise NA)

[2.1} INITIATE a p~ant cooldown to prevent exceeding the pressure limit for the Rx vessel bottom head temperature indicated on REACTOR VESSEL METAL TEMPERATURE, 1-TR-564 pt 10 (Panel 1-9-47) and based on Tech Specs Figure 3.4.9-1. D

[2.2] INFORM the Unit Supervisor, Tech Spec 3.4.1 requires the Reactor be placed in Mode 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . REFER TO 1-GOI-1 00-12A and Teeh Specs 3.4 .1.B. D

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Recirculation loops Operating B 3.4.1 BASES (continued) lCO Two recirculation loops are required to be in operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure that during a lOCA caused by a break of the piping of one recirculation loop the assumptions of the lOCA analysis are satisfied. With the limits specified in SR 3.4.1.1 not met, the recirculation loop with the lower flow must be considered not in operation. With only one recirculation loop in operation, modifications to the required APlHGR limits (lCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APlHGR),,), MCPR limits (lCO 3.2.2, 'MINIMUM CRITICAL POWER RATIO (MCPR),,), and APRM Flow Biased Simulated Thermal Power-High Setpoint (lCO 3.3.1.1) may be applied to allow continued operation consistent with the assumptions of References 7 and 8.

APPLICABILITY In MODES 1 and 2, requirements for operation of the Reactor Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur.

In MODES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Recirculation Loops Operating B 3.4.1 BASES ACTIONS B.1 (continued)

With no recirculation loops in operation while in MODES 1 or 2 or the Required Action and associated Completion Time of Condition A not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . In this condition, the recirculation loops are not required to be operating because of the reduced severity of DBAs and minimal dependence on the recirculation loop coastdown characteristics. The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 295016G2.2.3 Tier # 1 Knowledge of the design, procedural and operational differences 1 Group #

between units: Control Room Abandonment.

KIA # 295016G2.2.3 Importance Rating 3.8 3.9 I Proposed Question: SRO # 3 Given the following plant conditions:

Unit 1/2 control room has been evacuated due to a toxic gas intrusion.

Control has been established at both Backup Control Panels.

All four Unit 1/2 diesel generators have been started.

Both units have initiated a cooldown at < 90 OF.

The Shift manager has directed both units to initiate and maximize Suppression Pool Cooling.

Which ONE of the following describes the required RHR pump lineup for each unit and the basis for that lineup?

(

The RHR pump lineup is _ _~(,..:!:.1.J-)_ _ _

The basis for the required lineup is to (2)

(1) (2)

A. Unit-1 Div I, Unit-2 Div II equalize loading of all four 4Kv Shutdown Boards B. Unit-1 Div I, Unit-2 Div II ensure PREFERRED pumps for LPCI injection remain available.

c. Unit-1 Div II, Unit-2 Div I equalize loading of all four 4Kv Shutdown Boards D. Unit-1 Div II, Unit-2 Div I ensure PREFERRED pumps for LPCI injection remain available.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: C Explanation: a. Part (1) is incorrect. This would be the correct lineup for LPCI pumps under Common Accident conditions, but is reversed for operation from the Backup Control Panel. Part (2) is correct.

b. Part (1) is incorrect. This would be the correct lineup for LPCI pumps under Common Accident conditions, but is reversed for operation from the Backup Control Panel. Part (2) is incorrect. Although the lineup for Suppression Pool Cooling does not involve PREFERRED RHR pumps for the given unit, actions farther along in the AOI direct opening the breakers for the remaining two RHR pumps so they would not be available for injection.

c. Correct answer.

d. Part (1) is correct. Part (2) is incorrect. Although the lineup for Suppression Pool Cooling does not involve PREFERRED RHR pumps for the given unit, actions farther along in the AOI direct opening the breakers for the remaining two RHR pumps so they would not be available for injection.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 1/2-AOI-100-2, OPL 171.044 (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 7/13/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 X Comments: The basis for the divisional separation is primarily historical. Unit-2 was the first of the three BFN units to be re-started following an extended 3 unit shutdown. As such, Division I was established for use following control room abandonment. When Unit-3 followed, being essentially identical to Unit-2 with a separate control room and DGs, Division I was also used for control room abandonment. The Unit-1 re-start had to take into consideration the current divisional assignment of Unit-2 as well as reliability of available power. As such, Unit-1 was assigned Division" to allow equal loading of the two remaining 4Kv Shutdown Boards not being used by Unit-2. What makes this such a challenging question is that the divisional assignments are opposite of those for a Common Accident scenario.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Control Room Abandonment 1-AOI-100-2 Unit 1; Rev. 0016 Page 150179 4.2 Unit 1 Subsequent Actions (continued)

(15.4) VERtFY OPEN RHR SYSTEM II MtN FLOW VlV, 1-FCV-074-0030, at either Oftt1e folloWing:

480V RMOV 8d 18, Compt R11E. RHR SYSTEM II MIN FlOWVLV.1-8KR-014-oo30 OR. 0

Rx Bldg - SE Quad - E1 541'lOcal contro1 switch RHR SYS II MIN FlOW VAlVE, 1-HS-074-0030B.

(otherwise N/A) 0

[1S.5] PLACE RHR PUMP 18, 1-HS-074-0028C, in CLOSE at 4160V Shutdown 8d C, Compt 17, to start RHR PUMP

18. 0

[15.6] PLACE RHR SYS II SUPPR CH8R1POOL ISOL VlV.

1-HS-074-0071C in OPEN at480V RMOV 8d 18, Compt 11C. 0

[15.7] ESTABLISHRHR system flow between 7,000 and 10,000 gpm as follows:

115.7.1] MONITOR RHR SYS II TOTAL FLOW, 1-Fl-74-79 at Panel 1-25-32. 0 115.7.2] THROITLE. OPEN RHR SYS II SUPPR POOL CLGlTEST VLV using 1-HS-074-0073C at 400V RMOV Bd 18, Compt R11 C. 0

{15.7.3] WHEN RHR SYS II TOTAL FLOW, 1-FI-74-79 indicates between 7,000 and 10,000 gprn. THEN DIRECT tt1e operator to stop throttling 1-HS-074-0073C. o

{15.7.4] VERIFY CLOSED RHR SYSTEM n MIN FLOW VLV. 1-FCV-074-0030, at either of the foIlowiing:

480VRMOVBd 18, Compt. R11E, RHR SYSTEM It MIN FLOW VLV, 1-BKR-074-0030

~ 0

Rx Bldg - SE Quad - E1 541'1oca1 control switch RHR SYSTEM U MIN FLOW VALVE.

1-HS-07<W030B. (otherwise N/A) 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 16 Of 79 4.2 Unit 1 Subsequent Actions (continued)

['15.8] MONITOR SU?PR POOL TEMPERATURE, 1-TI-64-55B, at Pare 1-25-32 and MAINTAIN temperature less than 12()OF. o NOTE Communication betWeen 4160V Shutdown Bd 0 and 480V RMOV Bd 1B is necessary for establishing RHRSW flow and to prevent exceeding 53 amps on RHRSW Pump 02.

[15.9] IF additionalJ SUppression Pool cooling is necessary, THEN (Otherwise NlA)

START a second RHRSWand RHR pump as fOllows:

(15.9.1] PLACE RHRSW PUMP 02 MOTOR O-HS-023-0021C in CLOSE at 4160V Shutdown Bd D, Gomipt 15 to start RHRSW PUmp 02. 0

[15.9.2) THROTTLE OPEN RHR HX 10 RHRSW OUTLET VLV, using l-HS-023-0052C at 480V RMOV Bd 1B, Compt. 15C. 0

[15.9.3] WHEN betWeen 48 and 52 amps on RHR SERVICE WATER PUMP C1, THEN STOP throttling RHR HX 10 RHRSW OUTLET VLV,1-HS-023-0052C. o

{15.9.4} PLACE RHR PUMP 1D, 1-HS-Ol4-0039C, in CLOSE at 4160V Shutdown Bd D. Compt. 16, to start RHR Pump 10. 0

[15.9.5] MONITOR RHR SYS II TOTAL FLOW, 1-FI-14-19 at Panel 1-25-32. 0 115.9.6] THROTTLE OPEN RHR SYS II SUPPR POOl CLGlTEST VLV using 1-HS-074-0013C at 480V RMOV Bd 1B. campt. R11 C. 0 (15.9.71 WHEN RHR SYS II TOTAL FLOW, 1-FI-74-79 is at or below 13,000 gpm, THEN DIRECT operator to stop throttling 1-HS-074-0013C. o

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet SFN Control Room Abandonmem 2-AOI-10C-2 Unit 2 Rev. 0051 Page 16 of9S 4.2 Unit 2 Subsequent Actions (continued)

[15] INITIATE RHR Suppress~on Pool Cooling as follows:

NOTE Communication between 4160V Shutdown Bd Band 480V RMOV Bd 2A is necessary for estabHshing RHRSW flow and to prevent exceeding 53 amps on RHRSW Pump C2.

{'15.*1] At 4160V Shutdown Bd B, compt 15, PLACE RHRSW PUMP C2 MOTOR, O-HS-23-12C. in CLOSE to start RHR SERVICE WATER PUMP C2. 0

['15.2] At 480V RMOV Bd 2A, compt. 18C, THROTTLE OPEN RHR HX 2C OUTLET VLV, 2-HS-023-0040C. 0

[15.3] WHEN between 48 and 52 amps on RHR SERVICE WATER PUMPC2, THEN S*TOP throtmng, RHR HX 2C OUTLET VL V, 2-HS-023-0040C. o

{'15.4] VERIFY OPEN RHR SYSTEM I MINIMUM FLOW VALVE, 2-FCV-74-7, at either of the following:

480V R1VJOV Bd 2D, campI. 5E, RHR SYSTEM I MINIMUM FLOWVLV, OR

Rx Bldg - SW Quad - E! 541' local control switch RHR SYSTEM I MINIMUM FLOWVALVE, 2-HS-74-7B. 0

['15.5] At 4160V ShutdOlMl Sd B., camp!. 17, PLACE RHR PUMP 2C, 2-HS-074-0016C,. in CLOSE to start RHR PUMP2C. o

{'l5.6] At480V RMOV Bd 2A, compt. 11 C, PLACE RHR SYSTEM I SUpp POOL SPRAVrrEST ISOL VLV, 2-HS-74-57C, in OPEN. o

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Control Room Abandonment 2-AOI~1CO-2 Unit 2 Rev. 0051 Page 18 of 95 4.2 Unit 2 Subsequent Actions (continued}

NOTE Communication between 4160V Shutdown Bd A and 480V RMOV Bd 2A is necessary for establishing RHRSW flow and to prevent exceeding 53 amps on RHRSW Pump A2.

['15.9] IF additional Suppression Pool cooling is necessary, THEN (Othel""'Nise NlA)

START a second RHRSWand RHR pump as follows:

[15.9.1J At 4160V Shutdown Bd A, compt 17, PLACE RHRSW PUMP K2 MOTOR, O-HS-023-0005C,in CLOSE to start RHR SERVICE WATER PUMPA2. 0

[1;5.9.2] At 480V RMOV Bel 2A, camp!. HC, THROTTLE OPEN RHR HX 2A OUTLET VLV, 2-HS-023-'0034C. 0

[15.9.3J WHEN between 48 and 52 amps on RHR SERV1CE WATER PUMPA2, THEN STOP throttling RHR HX 2A OUTLET VLV, 2'-HS-023-0034C. o

[15.9.4] At 4'160V Shutdown Bd A, compt. 19, PLACE RHR PUMP 2A, 2-HS-074-0005C, in CLOSE to start RHRPUMP2A 0

['15."10] MAINTAIN RHR system flow at or below 13,000 gpm as follows:

[15.10.1] At Panel 2-25-32, MONITOR RHRSYS I TOTAL FLOW,2-FI!-74-79. 0

£15.10.2] At 480V RMOV Bel 2A, compt 19C5, THROITlE RHR SYSTEM 1TEST VLV, 2-HS-074-oo59C. 0

[15.10.3] WHENRHR SYS I TOTAL FLOW, 2-Fl-74-79 is at or below 13,000 gpm, THEN OIRECT operator to stop throttl.ing 2-HS-74-59C. 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Contra' Room Abandonment 1*AOI*100-2 Unit 1 Rev. 0016 Page 3 of7S 1.0 PURPOSE This instruction provides symptoms and operator actions for safe shut doWn and coo~down to cold cooditions (Mode 4) of Unit 1 Reactor from locations outside the Unit 1 Control Room in the event of a Main COntrol Room evacuation.

1.1 Scope This procedure can NOT be propeTly executed for, and does NOT support, shutting down the Reactor during any type of accident The provisions of this instruction are adequate and. proper for the following EOt entry conditions that may be encountered while executing COntrol Room abandonment

1-EOI-1 Flow Chart, RPV Control Reactor Water Level less than +2.0 inches Reactor Pressure High above 1073 psig..

1-EOI-2 Row Chart, Primary Containment Control Suppression Pool Temperature above 95C'F Suppressloo Pool Level aoove -1 inch 1.2 Responsibilities A The Shift Manager/Unit Supervisor (SRO) has prilTlary responsibility for implementation and coordinatfoo of this instruction.

B. For ALL Situations, the Shift Manager/Unit Supervisor (SRO) makes an assessment of the situation and attempts corrective measures to preclude evacuation. If abandonment becomes necessary, the Shift Manager/Unit Supervisor (SRO) has the authority to assign personnel necessary to implement this instruction.

c. When the Control Room becomes avaUable, the Shift Manager/unit Supeflli.sor (SRO) makes an assessment of the situation, gradually transfers contro1 back to normal, re-establishes the condenser as a heat sink, and returns the Condensate System to service.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 29S02SG2.1.19 Tier # 1 Ability to use plant computer to evaluate system or component 1 Group #

status: High Reactor Pressure.

KIA # 295025G2.1.19 Importance Rating 3.9 3.8 I Proposed Question: SRO # 4 Given the attached printout of U1 PEDS Main Steam/Safety Relief Valves:

Which ONE of the following describes the status of U1 Main Steam Relief Valves (MSRV) and any required actions resulting from this condition?

MSRVs 1-19 and 1-34 are _ _-\,. : (1"=.+)_ _ . The required action for this condition is to (2)

REFERENCE PROVIDED ON THE FOLLOWING PAGE.

(1) (2)

A. INOPERABLE be in Mode 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and Mode 4 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

B. INOPERABLE be in Mode 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to < 150 psig in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

c. OPERABLE enter an "INFORMATION ONLY" LCO in accordance with OPDP-8, "LCO Tracking."

D. OPERABLE maintain Suppression Pool temperature < 95 OF in accordance with 1-01-74, "RHR System."

/"\

.~

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet r ' MAIN SlfAM/SAFFTY RfllfF VALVES G::1@:lrgj Moin Alarms Graphics Trends Points ZoomJL~yers Prlrt Help

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

(

I Proposed Answer: D Explanation: a. Part (1) is incorrect. The MSRVs in question are both OPERABLE. Part (2) is also incorrect. This action is based on Tech Spec 3.4.3.A if more than one MSRV is inoperable.

b. Part (1) is incorrect. The MSRVs in question are both OPERABLE. They are not considered INOPERABLE based solely on leakage. Part (2) is also incorrect. This action is based on Tech Spec 3.S.1.G if two or more ADS MSRVs are inoperable.

c. Part (1) is correct based on (a) and (b) above. Part (2) is incorrect. This action is based on INOPERABLE equipment that does not apply to an LCO.

d. Correct Answer.

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

( Technical Reference(s): U1 TSR 3.4.3,3.5.1, 1-01-74 (Attach if not previously provided)

OPDP-8 Proposed references to be provided to applicants during examination: U-1 ICS computer printout of Main Steam/Safety Relief Valve page Attachment 1.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 7/13/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X Comments: In order to answer the first part of this question, the candidate must first determine whether the MSRVs are ADS or non-ADS valves. He must then determine whether the given indications from the plant computer satisfy the requirements for OPERABILITY according to Tech Spec bases. Therefore, only half of the first part of this question can be answered using RO technical knowledge.

-10

~

I tJ) w E

o LL.

c 0.,.

.- CI) ftiCl)

C.c

.- tn E~

ns ..

>< 0 w3:

c c 0

=..tn CI)

c 3:~

Cl)O Q.

E ns tJ)

Q tJ)

W

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

(

SlRVs 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 SafetylRelief Valves (SIRVs)

LCO 3.4.3 The safety function of 12 SlRVs shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION OOMPLETION llME A. One or more required A1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SlRVs inoperable.

AND A2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

c. BFN-UNIT 1 3.4-7 Amendment No. 234

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

(

ECCS - Operating 3.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION OOMPLETION TIME G. Two or more ADS valves G.l Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months inoperable.

AND OR G.2 Reduce reactor steam 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Required Action and dome pressure to associated Completion ~ 150 psig.

Time of Condition C, D, E, or F not met.

H. Too or more low pressure H.l Enter LCO 3.0.3. ImmedJiately EGGS injectionlspray SUbsystems inoperable for reasons other than Condition A.

OR HPGI System and one or more ADS valves inoperable.

BFN-UNIT 1 3.5-3 Amendment No.~ 240 March 12.2001

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

ECCS - Operating 3.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. HPCI System inoperable. C.l Verify by administratiVe Immediately means RCIC System is OPERABLE.

AND C.2 Restore HPCI System to 14 days OPERABLE status.

D. HPCI System inoperable. D.l Restore HPCI System to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OPERABLE status.

AND OR Condition A entered.

D.2 Restore low pressure 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ECCS injection/spray subsystem to OPERABLE status.

E. One ADS valve E.1 Restore ADS valve to 14 days inoperable. OPERABLE status.

F. One ADS valve F.1 Restore ADS vamve to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable. OPERABLE status.

AND OR Condition A entered. F.2 Restore low pressure 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ECCS injection/spray subsystem to OPERABLE status.

(continued )

BFN-UNIT 1 3.5-2 Amendment NO.-234; 240 March 12, 2001

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

(

SlRVs 63.4.3 BASES (continued)

APPLICABLE The overpressure protection system must accommodate the SAFETY ANALYSES most severe pressuriZation transient. Evaluations have determined that the most severe transient is the closure of all main steam isolation valves {MSIVs). followed by reactor scram on high neutron flux (i.e.* failure oflhe direct scram aSSOCiated with MSIV position) (Ref. 1). For the purpose of the analyses, 12 SlRVs are assumed to operate in the safety mode. The analysis results demonstrate that the desagn S/RV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design pressure (110% x 1250 psig =

1375 psig). This LCO helps to ensure that the acceptance limit of 1375 psig is met during the Design BaSis Event Reference 2 discusses additional events that are expected to actuate the SlRVs. From an overpressure slanqpoint. the design basis events are bounded by the MSIV dosure with flux scram event described above.

S/RVs satisfy Criterion 3 of the NRC Policy Statement (Ref. 4).

LCO The safety function of 12 S/RVs are required to be OPERABLE to satisfy the assumptions of the safety analysis (Refs. 1 and 2).

The requirements of this LeO are applicable only to the capability of the SlRVs to mechanically open to relieve excess pressure when the lift setpoint is exceeded (safety function).

The S1RV setpoints are established to ensure that the ASME Code limit on peak reactor pressure is satisfied. The ASME Code specifications require the lowest safety valve setpoint to be at or below vessel design pressure (1250 psig) and the (continued)

BFN-UNIT 1 B 3.4-18 ReVision 0

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

(

NPG Standard Limiting Conditions for Operation Tracking OPOP-8 Oepartm.ent Rev.OU02 Procedure Page 10 of 42 3.3.7 Documentation A. Immediate Operabi1ity determinations should be documented and sufficienUy addressed so an individual knowledgeable in the teclhnical disdpline associated with the condition woul<J be expected to understand its oasis.

B. The documentation for immedia e determinations need not be extens'ive. Plant record systems, such as operator logs or the corrective action program, are oHen sufficient.

C. Communication of the oasis for an immediate determination to following shins is also important. OperabUity may be oased on conditions that subsequently change. The person making the initial determination mayor may not be p!iesen at the plant when those conditions change. Subsequent personnel need to have sufficient info mation to ensure that if conditions that were important to a previous operability determination have changed that they are abJe to reassess the validity of the determinaoon and/or take appropriate actions.

3.4 TS LCO Evaluations 3.4.1 General Guidelines A. When equipment identifi.edin TS is made or becomes inoperab'le, existing plantfunit conditions may require LCOs be entered.

B. LCOs are entered if, for eXisting plantJunit conditions, TS require actton(s) to be ta'ken.

C. LCOs are exited when the equipment is rreturned to operable status or when the plant/unit is put into a condition where TS no longer require action(s) to be taken.

WON BFN Onll NOTE iTS 3.0.6 and 5.7.2.18 .......:,~.....,~a

=d=ress

== es""'L

C=O e~,~=Ia

tiy"-'e:;..,t~

o su~~rt s~tems and

stI2poDed systems D. TS action requiriements may change as the aggrega e of inoperable equjpment changes. Determination of TS actton(s} and the quantity of LCOs to be entered are based on the aggregate of inoperable systems, equipment, and components.

E. Mumple LCOs shall be entered and [logged if equipment is inoperab.le or removed from service and more than one liS LCO action is required to be taken.

F. If equipment is identifi.ed or made inoperable that does not apply 0 an LCO based on the cunrent plant conditions, an ::Information Only" LCD should be entered into the Unit Log or LeO Tracking Log, as appropriate. The "Information Only" lCO entry should contain [information similar to an "Active" LCO with possibly the exception of the LeO expiration date not being required .

During plant shutdown/outages, it is not required to utilize INFORMATION ONLY LCOs for conditions that are applicable only in other modes which are controlled by other

( plant instructions (Le .. general operating instructions, surveillance instructions etc.).

REFERENCE MATERIAL Provided to CANDIDATE

(

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

( Examination Outline Cross-reference: Level RO 295038EA2.04 Tier# 1 Ability to determine and interpret the following as they apply to Group # 1 High Off-site Release Rate: Source of Off-site Release.

KIA # 295038EA2.04 Importance Rating 4.1 4.5 I Proposed Question: SRO #5 Unit 3 was operating at 100% rated power when a Main Turbine trip resulted in the following indications:

OG PRETREATMENT RADIATION HIGH (9-3A W5) in alarm.

MAIN STEAM LINE RADIATION HIGH-HIGH (9-3A W27) in alarm.

TURBINE BLDG AREA RADIATION HIGH (9-3A W29) in alarm.

TURB BLDG ROOF EXH VENT RADIATION HIGH (9-3A W18) in alarm.

Reactor Water Level (-) 25 inches and recovering.

Reactor Pressure 900 psig and lowering.

MSIVs are open

3-TS-1-60A indication 140 OF and rising.

The Radcon Manager reports dose readings at the site boundary are 1.5 REM TEDE.

Which ONE of the following is the cause of these indications and the action required to mitigate the event?

Plant conditions indicate a (1) has occurred. The Unit Supervisor must

_ _ _ _ _~(2=+)----_ to mitigate this event.

(1) (2)

A. main steam line break in the Turbine Bldg enter 3-EOI-3, "Secondary Containment ControL" B. main steam line break in the Turbine Bldg enter 0-EOI-4, "Radioactivity Release Control."

c. Group I auto isolation failure with fuel damage enter 3-EOI-3, "Secondary Containment ControL" D. Group I auto isolation failure with fuel damage enter 0-EOI-4, "Radioactivity Release Control."

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: B Explanation: a. Part (1) is correct. Part (2) would be correct if the break occurred inboard of the MSIVs rather than outboard. This is only determined by analyzing the annunciators and recognizing that 3-TS-1-60A has not yet reached the Group I setpoint. Steam Tunnel relief panels will blowout to relieve steam to the Turbine Building, thus preventing temperature from isolating the MSIVs as quickly.

b. Correct answer

c. Part (1) is incorrect only because the MSIVs have not received an AUTO closure signal yet. However, the MSL Radiation High-High annunciator indicates the need to immediately close the MSIVs. In addition, 3-TS-1-60A is one indicator used to determine 3-EOI-3 entry conditions, but has not yet reached the required value.

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

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 3-ARP-9-3A, 3-ARP-9-30, O-EOI-4 (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 7/7/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X Comments:

(

ES-401 , Sample Written Examination Form ES-401-5 Question Worksheet BFN Panel 9-3 3-ARP-9-3A Unit 3 3-XA-55-3A Rev. 0036 10of51 SensorlTrip Point:

OG PRETREATMENT HI MOlATION 3-RM-90-151 1595 MRlHR HIGH 3-RA-90-'157A (Page 1 of 2)

Sensor RE-90-157, Turt) Bldg OG pretreatment samp.e chamber.

Location: E156S', T-14 B-UNE Probable A High radiation in the off-gas pretreatment system.

Cause: B. Resin trap faiAure (RWCU or Cood Demin).

C. PossiiJ4e fuel element failure.

D. Sensor malfunction.

Automatic None Action:

Opera.tor A VERIFY high radiation on following:

Action: 1. OFFGAS PRETREATMENT RADIATION recorder, 3-RR-Q0-157 on Panel 3-9-2. o

2. OFFGAS RADIATION ferorder~ 3-RR-90-160 00 Panel 3-9-2. o

3. 06 PRETREATMENT RAD MON RTMR, 3-RM-90-157 on Pan~ 3-9-10. o

4. OFFGAS RAD MON RTMR, 3-RM-90-160 on Panel 3-9-10. o B. CHECK off-gas fiow normal. 0 C. CHECK following radiation reCOfders and associated radiation monitors:

1. MAIN STEAM LINE RADIATION, 3-RR-90-135 on Panel 3-9-2. 0

2. OFFGAS POST-TREATMENT RADIATION, 3-RR-90-265 on Panel 3-9-2. 0

3. STACK GAS RADlATJON, O-RR-90-147 on Panel 1-9-2. 0 D. NOTIFY RADCON. 0 E. REQUEST Chemistry perform radiochemical analysis to determine source. o Continued on Next Page

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panel 9-3 3-ARP-9-3A Unit 3 3-XA-55-3A Rev. 0036

...........j~~@.~1,~~t O'G PRETREATMENT RADIATION HIGH 3-RA-90-151A, Window 5

{Page 2: of .2}

O'perator Action: (Continued)

F. REFER TO' 0-SI-4.8.B.1.a.1 and 1(2)(3}-SR-3.4.6.1 (A) for ODeM compliance and to determine if power level reduction is required. 0 G. IF directed by Unit Supervisor, THEN ,

REDUCE reactor power to maintain off-gas radiation within O'DCM limits. 0 H. IF O'DCM limits are exceeded, THEN REFER TO' EPIP-1. 0

References:

3-45E620-3 3-47E610-90-1 GE 3-729E814-4 3-SIMI-90B

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet EVENT CLASS1RCAllON 'MATRIX OR V~ 00 ~R,crR,EATMeff ~AT1OH HiGH min, RA~oo.151A OPERATING CON04il1OW:

S~tJoM, ~ Mode.? ~orOOring

~ib'l$ln ~ 5.

I I

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panel 9-3 3-ARP-9-3A Unit 3 3-XA-55-3A . Rev. 0036 Pag~~9of51 Sensm/Trip Poiot MAIN STEAM UNE RADIATION 3-RM-90-136 Three times the normal full power backgmuoct HIGH-HIGH 3-RM-90-137 3-RA-90-135G (page 1 of 1)

Sensor Panel 3-9-10 in the control room.

location:

Probable A. Radiation is three times normal full power background.

Ca.use: B. Sensor malfunctions.

C. SI (SR) in progress.

Automatic A. Mechanical vacuum pumps tlip.

Action: B. Vacuum pump suction vaLves 3-FCV-66-36 and 3-FCV-tl6-40 close.

Operator A VERIFY alarm on 3-RM-90-136 tllm 137 on Panel 3-9-10. o Action: B. CONF~RM main steam Hne radiation level on recorder 3-RR-90-135, Panel 3-9-2. o

c. IF alarm is VALID and scram has NOT o*ccurred, THEN PERFORM the following:

IF core flow is above 60.%, THEN 1.. LOWER core flow to between 50-60%. 0

2. MANUALL Y SCRAM the Reactor. 0

3. REFER TO 3-AOI-100-1. 0 D. IF plant conditions DO NOT require the execution of3-C-5, Power/level Control, THEN VERIFY the MSIV's CLOSED. 0 E. NOTIFY RADCON. 0 F. VERIFY actions OF 3-ARP-9-3A window 1 have been completed. 0 G. IF Technical Specifications limits are exceeded, THEN REFER TO EPIP-1. 0

References:

3-47 E61 0-90-1 G E. 730E915-9, 10 3-4-45E620-5

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panet9-3 Unit 3 3-XA-55-3A SensorlTrip Point TURBINE BLDG AREA RADIATION RI..:OO-5A 10 MRfHR HIGH RI-OO-6A 10MRfHR 3-RA-90-1E RI..:OO-7A 10 MRfHR RI..:90-10A 10MRfHR RI-OO-11A 10 MRfHR (Page 1 of 1)

RI-OO-12A 10MRfHR RI-OO-16A 10 MRfHR RI-OO-17A 10MRfHR RI-OO-1BA 10MRfHR RI-OO-19A 10MRfHR RI.:OO-31A 10 MRfHR Sensor RE-90-5, Generator operating floor TBH6H' T-14 D-UNE location: RE-90-6, RFP operating floor TB EI. 617' T-12 F-UNE RE-90-7, TurDine opera.ung floor TB H617' T-15 K-L1NE RE-90-11, Turbine Breeze way TB EI. 58,s' T-12 B-L1NE RE-90-12, FW Heater area TBH58,s' T-12 D-UNE RE-90-15, Decontamination Chmb. TB H 578' T-12 B-UNE R!-90-16, HobNeli pump area TB EI. 557' T-14 C-UNE R!-90-17, Condenser room area TB H 557' T-14 F-UNE R!-90-UI, Condenser corridor TB H 557' T-12 F-lINE R!-90-19, Outside Steam Tunnel TB H 565' T-15 J-UNE Ri-90-31, Raw Cooling Water Pumps TB EI. 557' T-13C-UNE Probable Radaation levels have risen above alarm set point.

Cause:

Automatic None Action:

Operator A. DETERMINE area with high radiation level on Panel 3-9-11. {Ala;rm Action: on Panel 3-9-11 will automatically reset if radiation level lowers beloW setpoint) o B. IF the TSC is NOT manned, THEN USE public address system to evacuate area 'Where high airborne conditions exist o Continued on Next Page

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panet9-3 3-ARP-9-3A Unit 3 3-XA-55-3A Rev. 0036

""~, - ""~,,~~." ~"" """"- "",."~~"~

43 of 51 TURBINE BLDG AREA RADIATION HIGH J-RA_90-1E, Window:l9 (Page 2 of 2)

Operator Action: (Continued)

C. I'F the TSC is manned, THEN REQUEST the TSC to evacuate non-essential personnel from affected areas. 0 Do NOTIFY RADCON. 0 E. MONITOR other parameters provkUng input to this annunciator frequently as these parameters will be masked from alarming while this alarm is sealed in. 0 F. If alarm is due to sensor malfunction, THEN REFER TO 0-01-55. 0

References:

3-45E620-3 3-45E61 0-90-1 GE 730E356-1

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet REV 0025 UNIT 3 Panel. 9-3 3-ARP-9-3D 3-D.-55-3D Paqe 25 SENSOR/TRIP POINT:

M1>~IN STEAM LINE LEAK DETECTION 3-TIS-1-60A TEMP HIGH 3-TS-1-60B throuqh D 160'F 3-TA-1-60

.24 SENSOR LOCATION: Pane.l 3-9'-3, Main Control. Room, El. 617', Panel 3-9-:n.

Main Control Rm, E1 617' PROBABLE CAUSE: 1. Main. Steam, RWCU, Feedwater, RCIe, or HPeI Disch(Only with HPCI in service and elevated Suppression Pool water Temp.] Line Break.

2. Turb or Rx IUdg cooling/ventilation out of service.

3. Sensor malfunct.ion ..

4. Ste*am ';..'a.u1t Exhaust Booster Fan out of service.

AUTOMATIC 1'.CTION: Impendinq MSIT{ Isolation at 169<>F area temp.

OPERATOR ACTION:

1. 3-LI-64-159'A, SUPPR POOL W).TER LEVEL may give erroneous indications due to High Temperatures experienced by the Instrument durinq a Main Steam. hiqh energy line break in Secondary Containment.

2. The f<:)l.lowinq Steps may be performed in any order or concurrently as necessary.

1. CHECK the foliowing temperature indications:

liN STEAM TUNNEL TEMP temperature indica.tor, 3-TIS-1-60A on Panel 3-9-3.

Temperature g'witches 3-TS-1-60B, -6GC, or -60D windowes) on Pa.nel 3-9-:n.

RWCU P:iping in the Main Steam Tunn~i tempexature indicators, 3-TIS-69-E!34A(B} {C~ (D). Auxiliary Instrument Room .Panels 9-83 (84) {8*5) Hl6)

OR res 'HPTURB' mimic.

2. CHECK the foliowing flow ind.ications:

M).IN STEAM LINE FLOW A{B) eCl (Il), 3-FI-46-1 {21 (3) {4} on Panel 3-9-5

RFW FLOW LINE A (S), 3-FI-3-7BlI. (78B} on Pa.nel 3-9-5.

RFP 3A(3B} 3Cl flow indicator's, 3-FI-3-21) (13) (6) on Panel 3-9-6.

3. IE" RCIC is not in service AND 3-FI-7l-1A{B}, RCIC STE1<..M FLOW indi.cates flow, "r.EIER ISOLlo.TE RCIC AND VERIFY Tempera.tures lowering.

CONTINUED OR REX'l' PAGE

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet REV 0025 UNIT 3 Panel 9-3 3-ARP-9-3D 3-XA-5S-3D Page 26 NAIll S'lBAK LID LEAK DETECTION 'l'BMP HIGH 3-TA-1-60 OPERATOR ACTION (Window 24 Continued)

4. CHl!!CK for elevated RAD Levels on the following Instruments:

3-*iRM-90-2.0, CRD-HCU Vlest.

3-iRM-90-29, Suppression Pool.

S. IF HPCI is injecting with elevated Sl:lppress.ion Pool Temperature, 'l'DIIi CONSIDER securing RPCI to deteDnine if i t is the sou.rce of the leak.

6. IF Rx B.ldg main steam tunnel temperatl:lre is ar.o~ 160°F c'n 3-TIS-l-60A on Panel 3-9-3, THEN PERFORM the following:

a. ENTER 3-EOI-3 Flowchart.

r.. VERIFY Rx Zone fans, 3-RS-61j-llA at Panel 3-9-25, in fast speed.

c. VERIFY Steam Vault Exhaust Booster Fan in service.

REFER TO 3-0r-30B.

7. IF turbine bu.ilding main steam tunnel temperatl:lre is above 160*F on 3-TS-1-6m~. -GIOC, o.r -60D on Pa.nel 3-9-21, 'l'BD DISPl'.TCH personnel to 4ell;,' AC 'rarr. Bldg Vent Bd SA (1'B, El 617') to veri.fy 1'B fans and the Mechanical Spaces Exhaust Fan running.

REFERENCES:

3-4SE620-.2; 3-47E61G-l-l i GE 920D351; Tech Specs 3.3.1.1, Reactor Protection System (RPS) Instrumentation; 3.3.6.1. Primary Containment Isolation Instrumentation.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet EVENT ClASSFlCAl10H MATRIX M~!n ~ llrIe ~~ .~~~~

PrilmflY~wth ~

AND

~d~_~<<.~

OO~a" a$~mmu~~tWtty~

AND

~~~<<.~

15~"

OPeRATING COWOi~:

jJn~ MMl$~ UM~al ~

P~<:OO~.

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

_ _ _<ilI<CI!<NIM _ _

. . . . ALLJja~1U!.N)

RV1Mll:$W1'IUII!

' - -_ _ _ _ _. - -_ _ _ _----1 L T;/I\Bl.£ $

'nil!!'

_110IIII II1'G"""""",,,

0IFl=sm:: AAOIOAC'I'l'lllTV ~CLAS:$FICA'DCltiI UMI1'SFOiR GlBERiL BllERae<<::Y srACK"fCGLltGlG*IUUMlIl LM'I'IIt_QI!

IFNmP ........GllltiZtCY CUitSA:AlDli UMT II!ttIRPGIIICIML~"

_.-.or 1'SI.D~"'1Ii_

1tJItIlIi.rn:tl1UMDIfG CUilGFIOAn:N UMT MiGFHf'GaCJfll\LI~

fa:.c..1~t RILD'"IIElI8IENT. . . .

OUlllJ1'Cll..11CHUIi!6T

"'1WI.at __

. . . e<lI!IIIIQl,II""'" rFNlpt ... CiINUW.~

I(J (ft1UffQN) nc. DOH . . . . .8IiT ~1'II:fiUMT

$llIU"""',**",,O'IIX'j hC1IIIl.at_,_ CC1R _ _

NQL DOIJ'J~JidAT QR.'l'CftCnca,.

1ftj1..... GIYIML~1C!'f'

~RCA1DtUMT aCU6\Jrl'-~Cm)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 600000AA2.07 Tier # 1 Ability to determine and interpret the following as they apply to 1 Group #

Plant Fire On Site: Whether malfunction is due to common-mode electrical failures. KIA # 600000AA2.07 Importance Rating 2.6 3.0 I Proposed Question: SRO # 6 Given Unit 2 at 1000/0 rated power:

While performing rounds, the U2 Rx Bldg AUO discovered 2-HS-74-1S8 on 480V RMOV Board 2B in the "SHUTDOWN" position for Suppression Pool Suction valves 2-FCV 74-24 and 2-FCV-74-3S.

Which ONE of the following describes the current status of the suction path for RHR Loop II and the action required?

Operation of the suction valves from the control room is _ _~(,..:!:.1.J-)_ _ _

The Unit Supervisor must _ _ _ _-4(.=2,J....)_ _ _ __

(1) (2)

A. unavailable restore RHR to OPERABLE in 7 days per Tech Spec 3.S.1.A B. unavailable return 2-HS-74-1S8 to NORMAL in 7 days per Appendix R

c. available restore RHR to OPERABLE in 7 days per Tech Spec 3.S.1.A D. available return 2-HS-74-1S8 to NORMAL in 7 days per Appendix R

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: D Explanation: a. Part (1) is incorrect. The NORMAL position of 2-HS-74-1S8 prevents operation of the valves from any location to prevent a common-mode electrical failure due to a fire. The SHUTDOWN position allows operation from the control room. Part (2) is incorrect. RHR Tech Spec OPERABILITY is not affected by the position of 2-HS-74-1S8. This is only an Appendix R issue.

b. Part (1) is incorrect. The NORMAL position of 2-HS-74-1S8 prevents operation of the valves from any location to prevent a common-mode electrical failure due to a fire. The SHUTDOWN position allows operation from the control room. Part (2) is correct.

c. Part (1) is correct. The SHUTDOWN position allows operation from the control room. Part (2) is incorrect. RHR Tech Spec OPERABILITY is not affected by the position of 2-HS-74-1S8. This is only an Appendix R issue.

d. Correct answer

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): FPR Volume 1 Section 4, U2 TSR 3.5.1 (Attach if not previously provided)

( OPL 171.044 Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 7/7/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X Comments: I evaluated this as CIA for two reasons. First, this question is answered without references. Second, what is actually tested involves determining whether the Appendix R switch impacts the OPERABILITY of RHR in accordance with 3.5.1. The bases of TSR 3.5.1 defines a LPCI sub-system as two pumps, piping, and valves to transfer water. The candidate must determine if the suction valves are considered OPERABLE by Tech Specs if they are considered INOPERABLE by Appendix R or vice versa. A determination must be made that RHR will still perform its intended function by Tech Specs by transferring water from the Suppression Pool to the RPV, but will NOT perform its required function by Appendix R which is to prevent a common-mode electrical failure from preventing LPCI injection from RHR Loop II.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet M:Ulual #: Fire Protection R.eport PLANT: BFN UNI'l'{s):1/2/3 PAGE 462 of915 VoL 1 TITLE: Appendix R. Safe Shui'ld.owll Program SECTION: 4 REV: 0 The listed compensatory measure in the Unit 1, 2 & 3 tables due to equipment degredation or the compensat.ory measures due to lack .of spatial separation per 9 .3.1LG. La of the Fire Protection Plan may be removed/revised if:

the affected unit is br.ought to COLD S~, or

an engineering analysis is performed, this program is changed and the Safe Shutdo;o,n Instructions are changed to provide an alternative shutdown path or

a different c.ompensatory measure or combination of measures is established (e.g., additional administrative controls, operator briefings,temporary procedures, interim strategies, operator manual actions, temporary fire barriers, temporary detection or suppression systems). An engineering analysis of the alternative measure should incorporate risk insights regarding the location, quantity and type of combustible material in the fire area; the presence of ignition sources and their likelihood of occurrence; the automatic fire suppression and fire detection capability in the fire area; the manual suppression capability in the fire area, and the human error probability where applicable. (Reference 5)

Compensatory Measure A will be documented and tracked in accordance with Attachment A of this instruction.

COMPENSATOR yo MEASURES A. Restore the equipment function in 7 days or provide equivalent shutdown

( capability by one of the following met.hods.

1) A temporary alteration in accordance with plant procedures that allows the equipment to perform its intended. function, or

2) A fire watch in accordance with the site impairment program in the affected areas/z.ones as specified in Section III.

Note:

Fire watch requirement.s in the Turbine Building (FA #25) and Control Building (FA jt16j may be evaluated on a case by case basis due to the large size of these areas. For example, fire watches in the Turbine Building can be limited to within 20 feet of the south wall (near M-Line wall on EL 565' and 586') or the Intake Pumping Station due to the location of the PJmSW power cables in the areas. No Safe Shutdown circuits are located in any other location within the Turbine building.

Control Building areas, even though n'::lt separated by fire resistive barriers, provide substantial protect. ion against the spread of fire due to installed fire suppression systems and concrete floor slabs and walls.

The potential of fire spread betwe.en control building compartments and the turbine building compartments has been evaluated in Section 3.0 of the lPEEE Fire Induced Vulnerability Evaluations for Unit.s 1-3 (Reference A16 Section 3). These evaluations may be reviewed to det,ermine the extent of fire watches.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet ECCS - Operating 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATlON COOLING (RCte) SYSTEM 3.5.1 ECGS - Operating LCO 3.5.1 Each ECGS injection/spray sUbsystem and the Automatic Depressurization System (ADS) function of six safe1yJreliefvatves shall be OPERABLE.

APPLICABILITY: MODE 1, MODES 2 and 3, except high pressure coolant injection (HPCf} and ADS valves are not required tore OPERABLE With reactor steam dome pressure::; 150 pSig.

ACTIONS

NOTE--------------------------------------------------

LCO 3.QA.bis not applicable to HPCL CONDITION REQUIRED ACTION COMPLETION TIME A One low pressure ECCS A1 Restore low pressure 7 daysPI injectlon/spray subsystem ECCS injection/spray inoperable. subsystem(s) to OPERABLE status.

One low pressure coolant injection (LPCI) pump in both LPCI subsystems inoperable.

(continued)

ll _ Thi'l! Completion Time may be extended to *t4 days on a one-time basis. This temporal)! approval expires June 1, 2005.

BFN-UNIT2 3.5-1 Amendment No. 253., 269, 286, 294 May 9, 2005

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPL171.044 Revision 15 Page 6 of7 INSTRUCTOR NOTES

1. Valve Interlocks Obj. V.B.IO.

a. RHR pump suction from Suppression Pool Obj. V.C.5.

(74-1, 74-12, 74-(1) No automatic closing or opening interlocks 24, 74-35)

(2) Valve cannot be opened unless the corresponding TP-15 and 16 pump Shutdown Cooling suction valve is fully closed (Interlock cannot be bypassed).

(3) App. R mod. Added NORMAL/SHUTDOWN SW Obj. V.D.8.

to control circuit. Prevents Operations from

Switch 2-74-157 (480 RMOV 2A) must be in any location.

SHUTDOWN to operate 2-74-1/12/48.

Switch 2-74-158 (480 RMOV 2B) must be placed in SHUTDOWN to operate 2-74-24/35.

Switch 3-74-24 (480 RMOV 3B) must be in SHUTDOWN to operate 3-74-24.

Switch 1-74-158(480 RMOV IB) must be in SHUTDOWN to operate 1-74-24/35

Switch 1-74-157 (480 RMOV lA) must be in SHUTDOWN to operate 1-74-1/12 (4) "EMERGENCY" position allows operation at Ul: 74-24/35 breaker only, but does not bypass in-line valve U2/3: 74-1112 interlock (5) Interlock with RHR pump - Must be a suction path available to either start pump or pump will trip if running and suction path isolates.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BASES BACKGROUND at 0.2 seconds when offsite power is available and B, C, and D (continued) pumps approximately 7, 14, and 21 seconds afterwards and if offsite power is not available all pumps 7 seconds after diesel generator power is available). When the RPV pressure drops sufficiently, CS System flow to the RPV begins. A full flow test line is provided to route water from and to the suppression pool to allow testing of the CS System without spraying water in the RPV.

LPCI is an independent operating mode of the RHR System.

There are two LPCI subsystems (Ref. 2), each consisting of two motor driven pumps and piping and valves to transfer water from the suppression pool to the RPV via the corresponding recirculation loop.

The two LPCI pumps and associated motor operated valves in each lPCI subsystem are powered from separate 4 kV shutdown boards. Both pumps in a LPCI subsystem inject water into the reactor vessel through a common inboard injection valve and depend on the closure of the recirculation pump discharge valve following a LPCI injection signal.

Therefore, each LPCI subsystem's common inboard injection valve and recirculation pump discharge valve are powered from one of the two 4 kV shutdown boards associated with that subsystem. The ability to provide power to the inboard injection valve and the recirculation pump discharge valve from two independent 4 kV shutdown boards ensures that a single failure of a diesel generator (DG) will not result in the failure of both LPCI pumps in one subsystem.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 700000G2.1.23 Tier# 1 Ability to perform specific system and integrated plant procedures 1 Group #

during all modes of plant operation. Generator Voltage and Electric Grid Disturbances. KIA # 700000G2.1.23 Importance Rating 4.3 4.4 I Proposed Question: SRO # 7 Unit-2 is at 95% rated power with the following conditions:

Trinity 1 500 Kv line has been removed from service to perform maintenance on switchyard breaker 5234.

Main Generator reactive load is 110 Mvars outgoing.

System voltage is 505 Kv.

System frequency is 59.97 Hz.

TVA Transmission System Load Coordinator has just declared an Emergency Load Curtailment Plan (ELCP) for the TVA transmission network.

Which ONE of the following describes the required action per plant procedures and the status of off-site power circuits to the Browns Ferry Nuclear Plant?

Direct the Unit Operator to _ _ _ _~(,.:l::.1J_)_____ . Off-site power circuits to Browns Ferry (2)

(1) (2)

A. raise reactor power to 100% in accordance with are NOT qualified.

O-AOI-57-1F, "Emergency Load Curtailment."

B. raise reactive load to 200 Mvars in accordance with are NOT qualified.

0-AOI-57-1E, "Grid Instability."

c. raise reactor power to 100% in accordance with remain qualified 0-AOI-57-1F, "Emergency Load Curtailment."

D. raise reactive load to 200 Mvars in accordance with remain qualified.

0-AOI-57-1E, "Grid Instability."

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: B Explanation: a. Part (1) is incorrect. Raising power is directed by O-AOI-57-1E only if grid frequency is less than 59.85 Hz. Since the initiating condition for declaring an ELCP is excessive grid loading, raising power may eventually be addressed with the Load Coordinator, but is not required per procedure at this time. Part (2) is correct since entry into 0-AOI-57-1F is based on notification that off-site power sources are no longer QUALIfIED.

b. Correct answer,

c. Part (1) is incorrect as stated in (a) above. Part (2) is incorrect. Although the Trinity 1 line is out of service, two QUALIFIED off-site power sources are available UNTIL the ELCP was declared.

d. Part (1) is correct. System voltage less than 507 Kv requires raising Mvars to 200 in accordance with 0-AOI-57-1E to aid in restoring voltage. Part (2) is incorrect as stated in (c) above.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 0-AOI-57-1 E, 0-AOI-57-1 F (Attach if not previously provided)

U2 TSR 3.B.1 Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 7/12/200B RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X Comments: The actions to raise power and reactive load are both appropriate to assist in mitigating the given conditions, however the stem specifically requires actions per plant procedures.

Only raising reactive load is procedurally directed for the given conditions. The Tech Spec action determination requires the candidate to recall the basis for declaring an ELCP as well as recognition that one 500 Kv transmission line out of service does not necessarily place the unit in 3.B.1.A. In fact, the Trinity 1 line should not have been approved for maintenance if sufficient QUALIFIFED off-site sources were not verified before hand.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet AC SOurces - Operating 3.8.1 3.8 ELECTRICAl POWER SYSTEMS 3.8.1 ACSOufces - Operating The following AC electrical power sources shall be OPERABLE:

a. Two qualified circuits between the offsite transmission network and the on site Class 1E AC Electrical Power Distribution System;

b. Unit 1 and 2. diiesel generators COGs) with two divisaons of 480 V load shed logic and common accident signal logic OPERABLE; and

c. Un~t 3 OG{s) capable of supplying the Unit 3 4.16 kV shutdown board(s) required by LCO 3.EL7, "DistributIon Systems -

Operati ng . "

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS

NOTE---------------------------------

LCO 3.HAbis not appUcable to DG5.

CONDITION REQUIRED ACTION COMPLETION TIME A. One required offsite A.1 Verify power availability 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months circuit inoperable. from the remaining OPERABLE offsite AND transmission netw'ork.

Once per Shours thereafter (continuedl BFN-UNIT 1 3.8-1 Amendment No ..~ 249 December 1, 2003

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet AC SOurces - Operating 3.8.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. OnedMsion of 480 V C.1 Restore requ!red dMslon 7 days load shed logic of 480 V load shed fogic inoperable. to OPERABLE status.

D. One divistoo of common D.1 Restore requfred dMslon 7 days accident signal logJc of common accident inoperable. sdgnal logic to OPERABLE status.

E. Two required offsite E.l Declare required 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months from circuits inoperaole. feature(s) inoperable discovery of when the redundant COndition E required feature(s} are concurrent with inoperable. inoperabiHty" of

( redundant required feature(s)

AND E.2 Restore one required 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months offsite circuit to OPERABLE status.

( continUed)

BFN-UNIT 1 3.8-4 Amendment No. 234

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet AC Sources - Operating B 3.8.1 BASES (continued)

LCO Two qualified circuits between the offsite transmission network and the onsite Class 1E DistribUtion System, four separate and independent Unit 1 and.2 OGs (A, B, C, and D), and the Unit 3 DG(s) needed to support required standby Gas Treatment (SGT) trains and Control Room Emergency Ventilation System (CREVS) trains are required to be OPERABLE. Two divisions of 480 V load shed I'ogic and two divisions of CAS logic are required to be OPE RABLE to support Unit 1 and.2 DG OPERABILITY and post-accident loads. Unit 3 Technical Specfficationswill require theoperabiiity of all Unit 3 DGs and provide appropriate compensatory acttoos for inoperable Unit 3 DGs in support of Unit 3 operations. To support the operation of Unit 1, the Unit '1 LCO for AC Sources - Operating also requires the necessary Unit 3 OG(s) to support SGT and GREVS required by LCO 3.8.7, Distribution Systems-Operating, for supplying the Unit 3 4.16 kVshutdown boards.

These requirements ensure availability of the required power to

( shut down the reactor and mainta,in it in a safe shutdown condition after an abnormal operational transient or a postulated DBA.

Qualified offsFte circuits are those that are described in the FSAR, and are part of the licensing basis for the unit Each offsite circuit must be capable of maintaining rated frequency and voltage, and accepting requlred loads during an accldent, white connected to the 4.16 kV shutdown boards.

(continued)

BFN-UNfT'1 B 3.8-6 Revision Q,52 May 11,2007

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Emergency Load Curtailment O-AOI-57-1F Unit 0 Rev. 0001 Page 4 of 24 1.0 PURPOSE A_ This abnormal operating instruction provides guidance for respondi:ng to the load dispatCher contacting BFN and stating that the oNe dfCUlts are unable to provide QUAliFIED offSite power to Browns Ferry Nudear Plant 2.0 SYMPTOMS A. None 3.0 AUTOMATIC ACTtONS A. None

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet SFN Emergency Load Curtailment O-AOl-57-1F Unit 0 Rev. 0001 Page 5 of 24 4.0 OPERATOR ACTION NOTE If necessary, pface keeping marks may be made directly in the COntrol Room copy of this instructiOn. CONTACT Management services for a repfacement copy When time permits.

4.1 Immediate Action None 4.2 Subsequent Action B. IF the load dispatcher contacts BFN and states the offsite cirwits are unable to provide QUAliFIED offsite power to Browns Feny Nuclear Plant, THEN PERFORM the following: (Otherwise N/A)

[1.11 ENTER LCO 3.8.1 (or 3.8.2 as appropriate) o

[1.2] MAKE appropriate notmcattons in accordance with SPP,-

3.5, REGULATORY REPORTlNG REQUIREMENTS. o

[1.3] tNmATE OPDP-9. EMERGENT ISSUE RESPONSE, (induding WorK Week Manager notificatiOn). o

[1.4] R.EVIEW work schedule and remove worK adM:ties on train components. o

[1*51 INmATE appropriate actions to restore nn components. o

[2] VERIFY that an Emergency load CUrtailment Plan (ElCP) has been dedared and RECORD the date and time beloW.

o Date Time

[3] SECURE all discretionary SWitching activities and maintenance, that could jeopardize the retiabtUty of either generation or the bulk transmission grid for the duration of the power supply alert/emergency bad curtailment 0

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

(

BFN Gnd fnstabil ity O-AOI,..S1-1 E Unit 0 Rev.. 0:006 Page 6 of '18 4.2 Subsequent Action {continued}

NOTES

-1) Changes in reactor power must comply with thermal power limits, rate of change limits and maximum power ilmFts as specified in the unitiZed GOI-1 OO-t2.

2) Fliuduating Frequency andilor Voltages by + 0.15 Hz or + 5kV is indicative of too muc:hgeneraUon for the system load. The Generator will have a tendency to pulsate due to coupling and momentum (with speedJfrequency rising and lowering).

3) Unless oiheM'ise directed by TVA Transmission system Load Coordlnators, IF frequency out of range, THE.N PERFORM step 4.2[5].

4) Unless otherwise directed by TVA Transmission System Load Coord.inators, IF VOltage out of range, THEN PERFORM step 4.2[6].

[4] IF System Frequency and/or vottages are continually fluctuating by + 0.15 Hertz or + 5kV, THEN VERIFYJIN~T;IATE Recire System upper power runbac:k by DEPRESSING 2 (3)-HS-42. UPPER POWER RUNBACK. o

[5] IF grid instability is characterized by system frequency being maintained outside the normal limits of 60.0 + 0.05 Hz. THEN PERFORM the follolNing:

[5.1] IF system frequency is greater than 60.15 Hz, THEN LOWER reactor power by approximately 1%lminute (10 MW(e)/minute) UNTIL system frequency returns to 60.03 Hz. 0

[5.2) IF system frequency is lower than 59.85 Hz AND reactor power is less than rated power, THEN RAISE reactor power by approximately 1%/minute (10 MW(e)/minute) UNTIL system frequency returns to 59.98 Hz. 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Grid Instability (i-AOI*57-1 E Unit 0 Rev. 0006 Page 7 of 18 4.2 Subsequent Action (continued)

[6] IF grid instability is characterized by system voltageberng maintained outside the normal Umits of 525 + 5 KV, THEN PERFORM the followTng steps:

(6_1] IF system voltage is greater than 540KV, THEN r6.1.1] LOWER reactive po\lVer to-15O MVAR, OR UNTI L system voiltage returns to 530KY 0

[6.1.2] CHECK -161KV cap Banks are Out of Service and EVALUATE conditions to determine appropriate actions. REFER TO O-GOf-300-4. 0

[6..2] IF system voltage is lower than 507KV, THEN PERFORM the fOillowing:

f6.3] RAISE reactive power to +200 MVAR, O.R UNTIL system voltage retums to 520KV. o f6A] CHECK 161KV Cap Banks are In Service and EVALUATE conditfoos to determine appropriate actions.

REFER TO O-GOI-300-4. o

[7] MONITOR system flowrates, operating pump amperages, energized boards and 4KV system voltages. REFER TO Attachment 1. o NOTE Unit Board 2C and 3C do NOT have Tap Changer regulation. Initial rising trend jn Pump amps Will be indicated on these boards loads first

{7.1] IF the 500 kV System voltage is degraded and Pump amps are rising through the yellow band., THEN PERFORM. the following:

[7.1.1} S:ECURE Unit 2(3) CCW Pump 2C(3C)- o

[7.1.2} SECURE Unit 2(3) Condensate Pump 2C(3C). . o

[7.1.3j SECURE Unit 2(3) Condensate Booster Pump 2C{3C). o

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 29S009G2.4.18 Tier # 1 Knowledge of specific bases of EOPs. Low Reactor Water Level 2 Group #

KIA # 295009G2.4.18 Importance Rating 3.3 4.0 I Proposed Question: SRO #8 Given the following Unit 2 conditions:

Reactor pressure: 10 psig

Drywell temperature: 250°F

Secondary Containment temperatures:

74-95F 220°F 74-95C & D 245°F 69-835A thru D 260°F 69-29F, G & H 200°F

Reactor water level indications:

U-3-58A & B Erratic

( U-3-52 (-)140 inches U-3-62A (-)160 inches U-3-53, 60 & 206 o inches U-3-55 o inches Which ONE of the following describes the required action and the basis for that action?

Enter _ _ _. . .J("",,1. J-)_ _ _ _ due to _ _ _ _ _ _~(=2)1-------- in the reference legs of U-3-58A & B.

REFERENCE PROVIDED A. 2-EOI-C-1, "Alternate Level Control;" flashing steam B. 2-EOI-C-4, "RPV Flooding;" flashing steam C. 2-EOI-C-1, "Alternate Level Control;" non-condensible gases coming out of solution

( D. 2-EOI-C-4, "RPV Flooding;" non-condensible gases coming out of solution

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: B Explanation: Flashing steam in the reference legs of U-3-S8A/B is indication of a a.

loss of RPV level instrumentation since Curve 8, RPV Saturation Temp is currently unsafe. However, the correct action is to enter 3-EOI-C4, RPV Flooding. This is plausible because U-3-S2 and 62A are approaching TAF, which would indicate a need to transition to 3-EOI-C1, "Alternate Level Control."

b. Correct answer Noncondensible gases are indicative of "Notching" instruments c.

caused by rapid depressurization. In addition, the incorrect EOI flowchart is entered. This is plausible because U-3-S2 and 62A are approaching TAF, which would indicate a need to transition to 3-EOI-C1, "Alternate Level Control."

Non-condensible gases are indicative of "Notching" instruments d.

caused by rapid depressurization. However, the correct EOI flowchart is entered due to a loss of RPV level instrumentation.

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 2-EOI-C-4 Flowchart (Attach if not previously provided)

EOI Program Manual Proposed references to be provided to applicants during examination: EOI Caution 1, Curve 8, Table 6, PIP 95-64 Question Source: Bank # X 259002G2.1.23 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 55.43 X Comments:

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet TABLE 6 SECONDARY CONTMT INSTRUMENT RUNS INSTRUMENT SC TEMP ELEMENTS AND LOCATIONS ELS21 EL593 EL565 RWCUHXRM (74.c95F) (74-95C AND D) (S9-835A THRU D) (69-29F, G. H)

LI-3..58A of OF N/A OF U-3-58B OF Of' NlA N/A U-3-53 OF Of' N/A OF U-3-60 OF Of' N/A N/A U~3*20S OF ,oF N/A OF Uw3*253 OF Of' NlA N/A U-3-52 OF OF OF N/A LI*3-62A OF OF OF NlA U~3-55 OF OF N/A N/A U~3-208A. B of OF N/A OF U-3-208C.D OF Of' N/A N/A

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet CAUTIONS i ,CAUTION #1

.. AN RP\l WATER !)iL INSTRUMENT MAY BE USED TO DETE:RMINE OR TREND LVL WHEN iT R.EADS ABOVE THE MtNIMUM INDICATED L\lL ASSOCIATED 'iJ'VITH THE HIIGHEST 'MAXDtN OR Be EMP.

.. IF OW TEt..'lPS, OR: Be ItREA TEMPS (TABLE S), AS APPUC.I\8LE, ARE OUTSIDE THE SAFE REGION OF eURVE8, THE ASSOCIATED INST.RUIIJ~El\rr M.a::!::8E UNREUA:BLE DUETOBOIUNG tN THE RUN.

rvUNI MUfrlt MAX DW RUN TEMP rvtAX SC INSTRUMENT RANGE INDICATED (FROM XR~64~50 RUN TEMP lVl OR TI~64-52AB) (FROM TABLE 6)

ON SCALE NIA BELaW~50 Nf.A 151 TO 200 EMERGENCy' 1---.-.1-40-,---I------N-lt-~.-----+--2{]- . . T-0-25-*.l\l:-(J---I U~3-:58.A, B

15)5 TO +150 Ni.A, 251 TO 300 NJA 3m TO 350 U*3-53 ON SCALE N.lA BELOW 100 U*34'.1O +5 NJA 15'1 TO 200 NORMAL U-3-206 +15 NlA .201 TO .250 D TO +60 U-3-253 +20 NfA 251 TO 300

( U*3*2DBA B. C, D +3'0 NI'A 301 '10350 U-3-!52 POST ACCIDENT ON SCALE N/A Nil;,.

U**3*;J52A -.288 TO +32

+10 BELO"'; HlO NIA

+15 100 TO 150 N!A SHUTDOIJVN +20 151 T020n Nli\

FLOODUP +30 201 TO 250 N/A OTO+4oo +40 NJ'A 251 TO 3,(10

+50 301 TO 35'0 Ni.A

+65 351 TO 40{l NlA

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet 3-Ll-3-52 & 62 CORRECTION CURVES

-150" TAF -162" - -162" = TAF (RED LINE)

-185" = MSCRWL (GREEN LINE)

-175" -200" = MZIRWL (BLUE LINE)

-215" = TWO-THIRDS CORE HEIGHT (BLACK LINE)

--1 W 1\

W

~

~ ....

--1 -200" 1'-1'-0 0 1\ ""f...1--

W r-- I--r- ACTUAL l- r...

e:( r- I--

LEVEL

()

i\ r-

""I--

0 -225 f'I r;;; r..

~

~ -162" Z I' r..f-r..

I' 1'-"

I""'r- r..

~

i"or-

~ -185"

-250" ~

I"" -200"

~

....... -215"

-268" o 100 200 300 400 500 600 700 800 900 1000 1100 REACTOR PRESSURE (PSIG) PIP-95-64 REV. 12

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet CURVE 8 RPV SATURATION TEMP 400 t- 360 380

~

J 0::

340

....z w

320

~

J 300 0::

00

~ 280 0::

w 260 z

Co.

~ 240 w

220 200 0 50 100 150 200 250 RPV PRESS (PSIG)

CONSTANT ABOVE 250 PSIG

( REFERENCE MATERIAL Provided to CANDIDATE

(

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

'\"

CAUTIONS ....

CAUTION #1 II AN RPV WATER l VllNSTRUMENT MAY BE USED TO DETERMINE OR TREND l Vl ONLY WHEN IT READS ABOVE THE MINIMUM INDICATED lVl ASSOCIATED WITH THE HiGHEST MAX OW OR SC RUN TEMP, IF D\I'J TEMPS, OR SC AREA TEMPS (TABLE 6)., AS APPliCABLE, ARE OUTSIDE THE SAFE REGION OF CURVE 8,

" THE ASSOCIATED INSTRUMENT MAY BE UNRELIABLE DUE TO BOILING IN THE RUN, 1******"'*,,**..** ****'T* ...........

MINIMUM MAX DWRUN TEMP MAXSC i INSTRUMENT RANGE INDICATED (FROM XR-64-50 RUN TEMP LVL OR TI-64-52AB) (FROM TABLE 6)

ON SCALE NlA BELOW 150

-145 N/A 151 TO 200 EMERGENCY U-3-58A,8 -140 NfA 201 TO 250

-155 TO +60

-130 NfA 251 TO 300

-120 N/A 301 TO 350 li-3-53 ON SCALE NlA BELOW 150 U-3-60 +5 NfA 151 TO 200 NORMAL LJ-3-206 +15 NlA 201 TO 250 i OTO +60 U-3-253 +20 N/A 251 TO 300 i 11 U-3,208A, B, C, D +30 NiA 301 TO 350 !

POST U-3-52 U-3-62A ACCIDENT

-268 TO +32 ON SCALE

+10 N!A NlA N/A I

BELOW 100

+15 100 TO 150 NlA SHUTDOVvN +20 151 TO 200 N1A U-3-55 FLOODUP +30 201 TO 250 NfA o TO +400 +40 251 T030n NiA

+50 301 TO 350 NfA

+65 351 TO 400 N/A

, ............... c .... ** ... .. cc . . . . . . . . . . . . . . c **. c...... ***** ** * ** c *** . . . . .c . . . . . . . . . . . . . . . . * . . c .... . ., ..., ........ .......................

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

(

CURVE 8 RPV SATURATION TEMP 400 380 ~ - *

-"" ~

....- ACTION ten 360 .~ REQUIRED z

J 340 ~

~

t-z w 320

/

E

J

~ 300 /

t-en z 280 /

~

w 260 ~ SAFE z

0..

E 240

/

220 V w

t-200 o 50 100 150 200 250 RPV PRESS (PSIG)

CONSTANT ABOVE 250 PSIG

(

3-Ll-3-52 & 62 CORRECTION CURVES

-150" TAF -162" - -162" =TAF (RED LINE)

-185" =MSCRWL (GREEN LINE)

-175" -200" =MZIRWL (BLUE LINE)

-215" =TWO-THIRDS CORE HEIGHT (BLACK LINE)

.....J W r\

W

~

"I'-.

.....J -200" aw f' "'"" "' l"-

.-<< i'

"'t-.

1-01-0 ...

1-0 1"- 1-0 ACTUAL LEVEL

() 1\ "'I"-I"- I--

a -225 ,~

t'-I-o ~ -162" z I' 1'" ,...,... "'" 1-0 I'

~

1'0 r"t'-

1"-1"- " ' ~

-185"

-250" .J'

....... -200" l..".-

....... -215"

-268" o 100 200 300 400 500 600 700 800 900 1000 1100 REACTOR PRESSURE (PSIG) PIP-95-64 REV. 12

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 295014AA2.04 Tier# 1 Ability to determine and interpret the following as they apply to 2 Group #

Inadvertent Reactivity Addition: Violation of fuel thermal limits.

KIA # 295014AA2.04 Importance Rating 4.1 4.4 I Proposed Question: SRO # 9 Given the following Unit 2 conditions:

Reactor power is at 100% rated.

Unit-2 Cycle 15 exposure is calculated at 30,000 MWd/MTU.

During performance of 2-SR-3.7.S.1, "Turbine Bypass Valve Cycling", Bypass valve #4 failed to open as required.

Which ONE of the following describes the most limiting Abnormal Operational Transient analyzed in the Final Safety Analysis Report (FSAR) and the most limiting fuel thermal limit during that transient?

The most limiting transient is _ _ _~(1=..i)'-- _ _ _ and the most limiting fuel thermal limit associated with that transient is _~(2=..),---_ _

(1) (2)

A. Generator Load Reject Linear Heat Generation Rate B. EHC Pressure Regulator Failure Linear Heat Generation Rate

c. Generator Load Reject Minimum Critical Power Ratio D. EHC Pressure Regulator Failure Minimum Critical Power Ratio

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

( I Proposed Answer: C Explanation: a. Part (1) is correct. Part (2) is incorrect. Although the positive reactivity inserted due to the rapid pressure increase will cause a rapid increase in power, the initiation of a reactor scram will act to turn power quickly near the bottom of the core where the LHGR is most limiting. The area of the core most susceptible to transition boiling is at the top of the core which will remain at a higher power level longer as control rods are inserted.

b. Part (1) is incorrect. An EHC Pressure Regulator failure results in a simultaneous closure of the Main Turbine Control Valves much the same as a load reject. However, the primary difference is the speed at which the TCVs close. A load reject results in a fast closure of the TCVs where an EHC failure closes the TCVs slower. The resultant pressure transient and associated reactivity addition is significant, but less severe.

c. Correct Answer

d. Part (1) is incorrect as stated in (b) above. Part (2) is correct. The area of the core most susceptible to transition boiling is at the top of the core which will remain at a higher power level longer as control rods are inserted. Since Transition Boiling is the failure mechanism of concern with regard to Critical Power Ratio, MCPR is the fuel thermal limit of concern during this transient.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): UFSAR Chapter 14.5 Uprated (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 7/16/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 X Comments: Even though the FSAR categorizes the above transients under "Nuclear System Pressure Increase" and the KIA is asking for "inadvertent reactivity addition", the result of the above transients is more of a concern with reactivity than with pressure. The only other option available involves a loss of feedwater heating which was already covered by a different question on the HLT 0707 SRO exam. To avoid double jeopardy or duplication, I took a small liberty in interpretation of the KIA.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN-22

'14.5.2.2 Generator Load Reject (TCV fast Closure) with Turbine Bypass Valve Fa.ilure (LRNBP) 14.5.2.2.1 Transtent Description The most severe transient for a fuU'-powergenerator trip occurs jf the turbine bypass valves fail to operate. Althoug.h the TCV fast closure time is sng.htly ~onger than that of the turbine stop valves, the control varves are cons~dered to be partially closed initially. This results in the generator trip steam supply shutoff being fa.ster than the turbine stop valve steam shutoff.

A generator trip from high power condit~ofls produces a transient sequence similar to the sequence described in section 14.5.2.1 except the turbine bypass valves are assumed to remain closed. The LRNBP event is caused by the fast dosure of all turbine control valves (TCVs) due to significant loss of electrical load on the generator. This wiHcause a sudden reduction in steam 'Row that results in significant vessel pressurization. The turbine bypass system is conservatively assumed lobe inoperable for this event A reactor scram signal is initiated by the TCVs closure.

The LRNBP event is ~dentified as one of the most limiting abnormal operational tran~ents for the BFN licensing analyses {assuming all equipment in service}.

Therefore, this evenlis analyzed to determine the operating limits and to verify the plant safety margins.

This abnormal operating transient is evaluated for each reload core to determine if this event could potentially alter the previous cycle MCPR operating limit. The analyses of this event for the most recent reload cycle is contained in the unit-speCIfiC and cycle-spedfic Reload Licensing Report 14.5.2.2.2 Initial Conditions and Assumptions For GE reload analyses, the analysis described in this section was performed with the ODYN computer code at the limiting powerlilow conditions at normal operation:

100 percent rated power (conSistent with the current licensing methodology) and maximum core fiOW (lGF) conditions. For bounding purposes, norma.! feedW*ater temperature (as opposed to reduced feedwater temperature) is assumed since the reactor steam generation would be IOlNer with a reduced feed'lNater temperature.

The EOC all-rods-out exposure is assumed to conservattve!y bound the control rod insertion effectiveness at any other cycle exposure. For FANP reload analyses, the FANP computer codes and a.naly~s methodology described ;in Section 3.7.7.1 .2 "MCPR Operating Limit Calculation Procedure" are used .

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

(

BFN-22 14.5.2.2.3 Interpretation of Transient Results Rgure14 . 5-5 shows the ~ant-specjfic response to the generator load rejectFon without bypass at '100 percent rated power and 105 percent flow conditions. The neutron flux peaks at 568 percent of initial; the average heat flux peaks at 125 percent of its initial val ue. The peak pressure at the bottom of the vessel is 1283 psia which fS well below the ASME upset code transients limit of 1375 psig while the peak. steam line pressure i:s 1245 psja. The calculated ilCPR at the stated conditioos is 0.19 for GE13 fuel; this result is representative but oot bounding for other GE fuel types.

At rated power, the aCPR for the lRNBP event is one of the most severe resulting from any other pressurizatton event. As power is reduced, the severity of the transient jncreases; but the fuel integrity is protected by the power-flow dependent thermal limits (see Section '14.5.8).

14.5.2.2.4 Generator load Rejed with Turbine Bypass Valve Failure with EOC-RPT-OOS The EOC-RPT-OOS condition etiminates the automatic ReCirculation Pump Trip signal when load Rejection occurs increasing the severity of the transient response.

At power levels below 30 percent of rated power (PIl)'Fa&6), the RPT is always ovpassed in conjunction with the scram on TSVslTCVs dosure. Therefore, these Jow power cases are not affected by the EOC-RPT-DOS condmon.

Figure 14.5-6 shows the transient results for the 100 percent of rated power and 105 percent of rated core flow case. EOC exposure and normal feedwater temperature conditions have been assumed for this transient analysis, the same as in the transient analysis with TBV in service described above.

The neutron flux peaks at 674 percent of initial, the average heat flux peaks at 130 percent of its initial value. The peak pressure at the bottom of the vessel is 1293 psia which is well below the ASME upset code tranSients limit of '1375 psig while the peak steam line pressure IS 1248 psla. The calculated ilCPR of this transient at the stated conditions is 0.23.

The penalty associated with EOC-RPT-OOS is about 0.04 In ACPR. At less than rated core flow., the penalty is smaller because of the relatively reduced beneficial effect of EOC-RPT.

The lmpact of the EOC-RPT-OOS on the transient fuel protecUon at off-rated powerJilow conditions has been addressed with the appropriate revision to the 14.5-"12

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 29S033G2.1.20 Tier# 1 Ability to execute procedure steps. High Secondary Containment Group # 2 Area Radiation Levels.

KIA # 295033G2.1.20 Importance R~ting 4.6 4.6 I Proposed Question: SRO #10 Given the following plant conditions:

Unit 2 is at 100% rated power.

A Reactor Water Cleanup (RWCU) drain line cracks and is spilling into the Reactor Building.

Area Radiation Monitors in the Reactor Building read as follows:

Reactor Building Elevation 593 >1000 mR/hr Reactor Building Elevation 565 West 800 mR/hr Reactor Building Elevation 565 East 850 mR/hr Reactor Building Elevation 565 Northeast >1000 mR/hr All other Reactor Building areas NOT ALARMED

RWCU Leak Detection Temp High (9-30 W 17) is in alarm.

Which ONE of the following describes the required action that MUST be directed by the Unit Supervisor and/or Shift Manager?

REFERENCE PROVIDED A. Enter 2-GOI-100-12A, "Unit Shutdown" and commence a normal shutdown / cooldown.

B. Enter 0-EOI-4, "Radioactivity Release Control" and initiate a Reactor Scram.

C. Rapidly depressurize the reactor, to the Main Condenser with the Main Turbine Bypass Valves per 2-EOI-1, "RPV ControL" D. Emergency Depressurize the reactor per 2-EOI-C2, "Emergency RPV Depressurization."

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

( I Proposed Answer: 0 Explanation: This is plausible because this action requires at least one area

a. greater than Max Safe. However, this is not appropriate since the source of the leak is not yet isolated as indicated by the RWCU Leak Detection Temp High annunciator in alarm.

This is plausible because this action requires an un-isolable leak and

b. entry into an emergency classification. However, this is not appropriate since the source of the leak is not yet isolated as indicated by the RWCU Leak Detection Temp High annunciator in alarm. Further action to reduce the driving head of the leak is also required.

This is plausible because this action requires at least one area

c. greater than Max Safe and another area approaching Max Safe.

However, this is not appropriate since the conditions already exist to perform Emergency Depressurization. Rapid depressurization is only performed to avoid conditions that require Emergency Depressurization. It is not a substitute if ED conditions currently exist.

d. correct answer

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

( -------------------------------

EOI Program Manual Proposed references to be provided to applicants during examination: 2-EOI-3 Flowchart Question Source: Bank # 295033EA2.01 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 55.43 X Comments:

(

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

. - .. . . .u. S\!~ tw.T!\IIt: fIIecK'oA:liH:>

t..rOlHe!\llEA~S'>_SIll!iOY*9:)ro,

!lE OKR ...I9:) £IV EOrIi

~

SI:J_E!i:!lAFflE L

NOL

_m_"'iIl'V _ _ lIUIliDTiClilillall1l!lli_O

~EtJ,.. ~diiI::tf,>> ..;ILC;'~: e~~'), CS."{l L

(

REFERENCE MATERIAL Provided to CANDIDATE

fI) *g

" 8~

I ~

2

~~

). I ~~~

8 ~I~

~

~~

~

,-l,1--rrn--,-,----rTTTTTTTl u

~

'!! .. ,. ,..11 !,'

~ Ii nj i lim!! II ni i

~

~ "ll " , II i '"11111 I-!l-++H-H-I+H-ttt-l

~! III!!! I ! ! ! II II II !

'i IIII!!!!!! Hl!ini i

!I ill I! , .* _I!

J:lhi;l; ;;!Iill

-o fI)

I

,-I 1&1

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 203000G2.4.40 Tier# 2 Ability to apply Technical Specifications for RHRjLPCI Injection 1 Group #

Mode.

KIA # 203000G2.4 .40 Importance Rating 3.4 4.7 I Proposed Question: SRO #11 Unit 2 is being shutdown because the RHR Loop II has been inoperable for 6 days.

RHR Loop II is tagged out for maintenance on the outboard injection valve.

A cooldown is in progress.

Reactor pressure is currently at 150 psig.

In addition to an OPERABLE injection flowpath, which ONE of the following describes the minimum required RHR pumps for an OPERABLE Low Pressure ECCS subsystem, and can RHR Loop I be flushed in preparation for shutdown cooling under this condition?

An OPERABLE Low Pressure ECCS subsystem requires (1) . In order to comply with Technical SpeCifications, flushing RHR Loop I (2) allowable.

(1) (2)

A. one RHR pump is B. one RHR pump is NOT C. two RHR pumps is D. two RHR pumps is NOT

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: D Explanation: a. Tech Spec bases require 2 RHR pumps per subsystem while in mode 3.

Core Spray only requires one pump per subsystem. In addition, flushing RHR requires defeating valve interlocks which would make LPCI Loop I INOPERABLE during the flush. Although Tech Spec SR 3.5.1.2 allows RHR to be lined up for Shutdown Cooling, this does not include the prerequisite flushing evolution.

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

c. Part (1) is correct. Part (2) is incorrect as in (a) above.

d. correct answer

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): U2 TSR section 3.5.1, 2-01-74 (Attach if not previously provided)

( U2 TSB section 3.5.1 Proposed references to be provided to applicants during examination: None Question Source: Bank # X OPL 171-044 #120 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 55.43 X Comments:

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet ECCS - Operating B 3.5.1 BASES BACKGROUND at 0.2 seconds when offsite power is available and B, C, and 0 (continued) pumps approximately 7, 14;. and 21 seconds afterwards and if offsite power is not available all pumps 7 seconds after diesel generator power is available). When the RPV pressure drops sufficiently, CS System ftow to the RPV begins. A full flow test line is provided to route water from and to the suppression pool to allow testing of the CS System without spraying water in the RPV.

LPCI is an independent operating mode of the RHR System.

There are two LPCI subsystems (Ref. 2), each consisting of two motor driven pumps and piping and valves to transfer water from the suppression pool to the RPV via the corresponding recirculation loop_

( The two LPCI pumps and associ.ated motor operated valves in each LPCI subsystem are powered from separate 4 kV shutdown boards. 80th pumps in a LPCI subsystem inject water into the reactor vessel through a common inboard injection valve and depend on the closure of the recirculation pump discharge valve following a LPCI injection signal.

Therefore, each LPGI subsystem's common inboard injection valve and recirculation pump discharge valve are powered from one of the two 4 kV shutdown boards associated with that subsystem. The ability to provide power to the inboard injection valve and the recirculation pump discharge valve from two independent 4 kV shutdown boards ensures that a single failure of a diesel generator (DG) will not result in the failure of both LPCI pumps in one subsystem.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify, for each ECCS injection/spray 31 days subsystem, the piping is filled with water from the pump discharge valve to the injection valve.

SR 3.5.1.2 -----------------NOTE,.-----------------

Low pressure coolant injection (LPG!)

subsystems may be consl,dered OPERABLE duTing alignment and operation for decay heat removal with reactor steam dome pressure less than the Residual Heat Removal (RHR) lOW' pressure permissive pressure in MODE 3, if capable of being manually realigned and not otherwise inoperable.

Verify each EGCS injecttoolspray subsystem 31 days manuat, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwlse secured in position, is in the correct position.

SR 3.5.1.3 Verify ADS air supply header pressure is ~ 81 31 days psig.

SR 3.5.*1.4 Verify the LPCI cross tie valve is closed and 31 days power is removed from the valve operator.

(continued)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Residual Heat Removaf System 2-0i-74 Unit 2 R.ev.0133 Page S8*of 367 8.7 loop I(U) flush for Shutdown CooHng OAtmON In on1er to ensLlre adequate net pos.itive suction head (NPSH) on the CRD pumps, the level in any CST should not be allowed to dropbe~OW 15 feet, as indicated by the Condensate Storage Tank Level indicators on Unit 1 Panel 9-20, NOTES

1) An operations are performed at Panel 2-9-3 unless otherwise noted.

2) A WO is used to place and subsequently remove temporary jumpers between terminal points 84 and 85 on loop I (terminal points 84 and 85, Loop II) in the limit switch compartments of 2-FCV-74-2 and 2-FCV-74-13, loop I (2:-FCV-74-25 and 2-FCV-74-36, Loop II) to bypass interlock with 2-FCV-74-57, loop I (2-FCV-74-71, Loop II).

[1] IF CS&S has been aligned as the keep fill source for N.!() days or more, THEN p.i] REQUEST chemistry to sample. o

[1.2] IF water quality requirements are met, THEN PROCEED TO' Section 8.8. o f1.3] IF water quality requirements are not met, THEN CONTINUE in this section. o (2) VERIFY the following initial conditions are satisfied:

{2.1J RHR Loop I(II} is in Standby Readiness.

REFER TO Section 4.0. REFER TO Tech Specs for RHR opera.bility requirements. o

{2.2] Condensate Transfer System in service to provide RHR ftush water. REFER TO 0-01-2B . o

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

(

21S003G2.1.28 Tier # 2 Knowledge of the purpose and function of major system 1 Group #

components and controls: IRM KIA # 21S003G2.1.28 Importance Rating 4.1 4.1 I Proposed Question: SRO #12 Given the following plant conditions:

Unit 3 is performing a startup and heatup in accordance with 3-GOI-100-1A, "UNIT STARTUP."

IRM 'B' is in BYPASS.

No IRM Range Switches are being manipulated.

All OPERABLE IRMs are on Range 7 when a half-scram occurs on RPS 'A'.

The Reactivity Manager reports the cause of the half-scram was due to a momentary Upscale Trip on IRM 'G;' but, the IRM is currently reading normally.

Whose approval, if any, is required in accordance with 3-GOI-100-1A to bypass IRM 'G', and what are the required actions per Technical Specifications, if any?

A. Plant Manager or designee approval is required.

NO Tech Spec action is required.

B. Plant Manager or designee approval is required.

Place ONE IRM channel in the TRIP condition.

C. NO approvals are required.

NO Tech Spec action is required.

D. NO approvals are required.

Place ONE IRM channel in the TRIP condition.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: C Explanation: Part (1) is incorrect. Plant Manager's permission is NOT required a.

since the bypass is directed by an approved procedure. Part (2) is correct.

Part (1) is incorrect. Plant Manager's permission is NOT required b.

since the bypass is directed by an approved procedure. Part (2) is incorrect. Tech Spec 3.3.1.1 requires 3 OPERABLE channels per trip system. IRM 'G' is not required to be OPERABLE since all other IRMs assigned to trip system "A" are OPERABLE.

c. Correct answer.

d. Part (1) is correct. Part (2) is incorrect. Tech Spec 3.3.1.1 requires 3 OPERABLE channels per trip system. IRM 'G' is not required to be OPERABLE since all other IRMs assigned to trip system "A" are OPERABLE.

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 3-01-92A, 3-ARP-9-5A (33) (Attach if not previously provided)

U3 TSR 3.3.1.1 Proposed references to be provided to applicants during examination: None Question Source: Bank # 215003G2.1.14 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 55.43 X Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet P*aneIS-5 3-ARP-9-SA 3-XA-55-SA Rev. 0037 Pa~43of45 SensorlTrip Point IRM Relay K-16 A. HI-HI ~ 116.4 on 125 scale CHA,C, E, G B. INOP.

HI-HIIINOP 1. Hi voltage low.

2. Mooule unplugged.

3. Function switch NOT in RED BAR OPERATE.

(Page 1 ofl) 4. loss of +/- 24 VDC to monitor Sensor Control Room Panel 3-9-12.

location:

Probable A. Flux level at or above setpoint Cause: B. One or more inoperable com:litions exist C. SI or SR in progress.

D. Malfunction of sensOf.

E. Control rod drop accident

.Automatic A. Half-scram if one sensor actuates (except with Rx Mode Sw. in RUN) .

( Action: 8.. Reactor scram if one sensor per channel actuates, (except with Rx Mode Sw. in RUN).

Operator A. STOP any reactivity changes. o Action: B.. VERIFY alarm by multiple indications. o C. RANGE initiating channel or BYPASS initiating channel.

REFER TO 3-01-92A. o D. With SRO permiSSion, RESET Half Scram. REFER TO 3-01-99 E. IFa:larm is from a control rod drop, THEN REFEiR TO 3-AOH35-1. o F. [NRC/ClIF one or more IRM recorder reading is downscale, THEN CHECK for loss of +/- 24 VDC power. 0 G. NO TlFY Instrument Maintenance that functional tests of any monitors indicating an INOP conditiOn, including a downscale reading, are required: before the instrument can be considered operable. [NRC IE item 8~[!3J 0 H. NOTIFY Reactor Engineer. 0 I. REFER TO Tec,h Spec Table 3.3.1.1-1, TRM Tabies 3.3.4-1 and 3.3.5-1. 0

References:

3-45E62D-6 1 97R114-16 GEK 3-730E915-1O 3-73DE91SRF-12 3-01-92A 3-AOI-85-1 Technical Specifications Tectmical Requirements Manual-TRM

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Unit Startup 3-GOI-1 OO~1A UTlft 3 Rev. 0071 Page 18 of 165 3.2 Coolant and Metaf Temperatures (continued)

3. During Reactor heatup, operators should tlse metal temperatures as a reminder that as metal heats tip, the moderator HEATUP RATE will rise with the same amount of heat input D.Minimizing operatioo with low feedwater flow and temperature Of cold feedwater flow cycling limi".5 thermal duty 00 feedWater nozzles (REFER TO 3-01-3).

3.3 Primary Containment A [l1"Fj Prior to initiating any event which adds, or has the potential to add, heat energy to the suppression chamber, the Unit Supervisor Of 8hlft Manager will e\laluate the necessity of placing suppression pool cooling in service. This is due to the potential ot developing thermal stagnation during sustained heat addittoos.. 111-5'91-1291 B. When containment integrity is required, alrlod: door sealssnould be tested within seven days after each contarnment access per O-TI-360 App A 3.4 Contr04 Rods, Reactivity Control and Relative Instrumentation A [NRC'C] Startups are performed using 3-SR ~3 .1.3.5(A) 10 incorporate Reduced Notch Worth Procedure fRNWP) and Banked Position Withdrawal SeQUence (BPWS) recommended by G.E. [tE 5IlJJlelIn (g*lZ. L£R25C1'301OC"4j B. [lNER.'q Periodic pauses during control rod wtlhdrawa! are necessary to allow for stabilization of neutron le\'el and ooilection of data for estrmating proximity to cmicaity. ~R m-Gile.. SCER 38-0021

c. £INPO'C] Adjustment of Nuclear Instrumentation readings downward. to match other indications without a ful:1investigation and comparison with aU available methods to measure power level may result in non-conservative power readings and protective setpoints. lSOER 9:1-003. SOER-68-JlIl2j D, [lNE!OCJ If SRMs or IRMs exhibit noise spikies duringstartU!p, control roo withdrawal should be suspended and an assessment of 8RM or fRMoperability performed in accordance wtlh 3-01-92 or 3-01-92A, as apPlicable. 1SCER88-t4J2!

E. [lNERIq Activities that can directty affect core readivily are of a critical nature and reQUire strict procedural compliance, aloog wilh conservative actions. [lNPO Sffi 8!HI05. SOER 58-002]

F. [NSRB.'Cl ReactMtycan be added wi1hout moving control rods due to changing plant conditioos (such as lowering moderator temperature, towering xenon concentration, rising Reactor pressure, and rising feedwater flow) especiaUy at law power. Awareness of these conditioos and monitOring core instrumentation for these changes is required. {A2$4J

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet BFN Intermediate Range Monitors 3-0il-92A Unit 3 Rev. 0014 PageS eftS 3.0 PRECAUTIONS AND UMfTATIONS(eontinued}

L. !IUFI An IRM or SRM may be bypassed in the following: conditions:

1. STOP control rod withdrawal and PLACE the channel in bypass when the SRM or IRM first gets nOisy.

2.. STOP contro~ rod withdrawal and PLACE the channel in bypass immediately upon receipt of a si~eevent large noise spike.

These conditions bypass the instrument for an operability assessment based on whether the noise is transitory or sustained. Transitory noise is considered a one time occurrence that does not repeat itself and the channel can be removed from bypass and restored to service.

Sustained noise is when theduraUon exceeds 15 minutes and may result in signal build up until a trip sfgnal is reached. tf a trip or high flux signal was generated, the channel is required to be observed for at least 15 minutes before returning the instrument to service with concurrence from System Engineering.

When the tnitial assessment and recognition of the magnitude of the event has been determined, then control rod Withdrawal may be resumed where it has been left off as long as the minimum number of SRM and IRM channels operable are within the Technical SpeCification limits. [1I-B-Q1-04O}

M. [CIAIC] SPP-1 0.4 requires approval of the Plant Manager or his designee prior to any planned operatbn with IRIvls bypassed unless bypassing is specifically allowed within approved procedures. [ISE-NPS*B2-RfJ1]

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet RPS Instrumentation 3.3.1.1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shaH be OPERABLE.

APPLICABILITY: According to Table 3.3.1.1-1.

ACTIONS

NOTE------------------------------------

Separate Conditioo entry is allowed for each channel.

CONmTION REQUIRED ACTION COMPLETION TIME

(

A. One or more required A.1 Place channef in trip. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months channels inoperable.

OR A.2 -----------NOTE---------

Not applicable for Functions 2.a, 2.b, 2.c, 2.d, or 2.f.

Place associated trip 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months system in trip.

( continued)

BFN-UNIT3 3.3-1: Amendment No. 212, 213, 221 September 27, 1999

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

(

RPSlnstrumentation 3.3.1.1 Table 3.3. 1.1-1 {page 1 of 3)

Reacb:lf Froteclion SY61em Inw~on

.APPLICABLE COOOlIlONS MOClESOR REOOIRED REFERENCED FUNCTiON OTHER CHANNELS FROM SURVEillANCE ALLOWABlE SPECmFiED PERTRI? REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION D.l

a. Neutron Rllx- High 2 SR 3.3.U.1 ;:;12011.25 SR 3.3.U.3 tt;"iEiorn;of full SR 3.3.1.Ui sca!e SR 3.3.1.1.6 SR 3.3.1. 1.9 SR 3.3.1.1.14 H SR 3.3.1.1.1 ;:; t21lf125 SR 3.3.1.104 d;"iEiorn; of full SR 3.3.1.1.9 sca!e SR 3 . 3.U.14

b. loop :2 3 G SR 3,3.U.3 Nil.

SR 3. 3.1.U4 eJa} 3 H SR 3.3.1.1.4 Nil.

SR 3.3.U.14

( 2. Average POWB' Range, Monilon; a, Neutron f'N.lC - High. ""' 3{b) G SR 3.3.U.1 :::; 15%R1P l:saoown) SR 3.3.1.Ul SR 3.3.1.1.7 SR 3.3.U.13 SR 3.3.U.16

b. FIovI Biased SinYJ!ated J{b) F SR 3.3.U.1 :::O.e6W The1ma1 Perl/er - Hiig/1 SR 3.3.1.1 ..2 +f.l6%RTP SR 3.3.1.1.7 and::; 120%

SR 3,3.1.1.13 RTP(c)

SR 3..3.1. L16

c. Neutron Rllx . High SR 3.3.1.U :::120%RTP SR 3.3.U.2 SR 3.J.U.?

SR 3.3.1.1.13 SR 3.3.U.16 Ie} [.66 W +66% - .ee~ VII! RTP when reset br single loopopEf:ata1 per LCO 3.4.1, 'RecircIJlaiion Loops Operating:

BFN-UNIT3 3.3-7 Amendment No. 216 December 23, 1998

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

( Examination Outline Cross-reference: Level RO 21S00SA2.06 Tier# 2 Ability to (a) predict the impacts of the following on the APRMjLPRM 1 Group #

system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal KIA # 215005A2.06 conditions or operations: Recirculation Flow Channels Upscale Importance Rating 3.4 3.5 I Proposed Question: SRO #13 Given the following plant conditions:

Unit-2 is operating at 100% rated power

Control Rod Withdrawal Block (9-SA W7) in alarm.

APRM Flow Bias Off Normal (9-SA W32) in alarm.

2A Recirc Pump Flow indicator 2-FI-68-S indicates upscale in Panel 9-4.

Which ONE of the following describes the required action to clear the above alarm conditions and the final status of 2-FI-68-S once the above action is accomplished?

Direct _ _~(1~)_ __ Flow indicator 2-FI-68-S will be reading _--1,(,."..2)1--_

(1) (2)

A. APRM #3 bypassed per 2-01-92B, normally "Average Power Range Monitoring."

B. APRM #3 bypassed per 2-01-92B, upscale "Average Power Range Monitoring."

c. RBM "A" bypassed per 2-01-92C, normally "Rod Block Monitor."

D. RBM "A" bypassed per 2-01-92C, upscale "Rod Block Monitor."

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: B Explanation: a. Bypassing the APRM is correct, however that will not effect the flow indication on panel 9-4. This indicator will continue to read upscale until repaired.

b. correct answer

c. Bypassing the RBM will clear the rod block but not the flow bias off normal condition. #2 incorrect as stated in (a) above.

d. Bypassing the RBM will clear the rod block but not the flow bias off normal condition. #2 is correct.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 2-01-92B, 2-01-92C (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 X 6/22/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 X Comments:

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Pa.net 9-5 2-ARP-9-5A Unit 2 2-XA-55-SA Rev. 0042 11 of 45 Sensor/Trip Point Relays: 3A~K1 Nuclear Instrumentation CONTROL ROD WITHDRAWAL 3A-K2 Refuel Equipment In Use BLOCK High level In Scram Discharge Volume Scram Discharge Volume High Water YELLOW BAR f7 Level Bypass Rx Mode Switch in Shutdown (Page 1 of2)

PRNM {ANY APRM, OPRM or RBM)

Sensor Panel 2-9-28 Location: Elevation 593' Aux lnstr Room PrcJibable A. One or more sensors at or above set point.

Cause: S. Malfunction of sensor.

C. Control rod drop accident Automatic Rod withdrawal block.

Action:

Operator A. DETERM INE in itiating condition from corresponding rod withdrawal Action: block alarm(s) and REFER TO operator action for alarm(s). []

B. IF aJarm due to inadvertent criticality during incore fueil movements, THEN REFER TO 2-AOI-79-2. []

C. IF alarm is from a control rod drop, THE N REfER TO 2-AOI-85-1. []

D. IF NO corresponding alarm exists, THEN []

1. AT ICS conso~e, DETERMINE if there is a refuel rod block by selecting Siing!e PO'int Menu, Single Value Display, and typing F602, RETURN. 0

2. IF rod block was from Refuel Floor, THEN NOTIFY Refuel Floor Operator to have dummy plug (Refuel floor between cavity and p004, southside) checked. []

3. WHEN IRM switches are below Range 3 with REACTOR MODE SWITCH not in RUN, THEN CHECK SRM detectors NOT FULL IN. []

4.. WHEN REACTOR MODE SWITCH is in START-UP position, THEN CHECK IRM detectors NOT FULL IN. []

E. REFER TO Tech Spec Table 3.3.2.1-1, TRM Tallie 3.1.4-1. []

Continued on Next Pa;ge

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

( BFN Panel 9-5 . 2-ARP-9-SA Unit 2 2-XA-SS-5A Rev. 0042 P~g~,,2 of 45 SensorlTrip Point APRM APRM Channel 1,2,3,4 (1) Either APRM Channel Redrc fLOW BIAS FLOW;:: 107% or failed. Flow;:: 107% or has fa*iled upscale.

OfF NORMAL YELLOW BAR f32 (Page 1 of 1)

Sensor 214 Logic Modules (Voters) located in Panel 2-9-14, Main Control Room.

Location:

Probable A Malfunction of flow circuit instrumentation.

Cause: S. Testing in progress.

C. Malfunchon of Sensor.

Automatic Rod btock.

Action: FLOW s~gnal an APRM bargraph indicates either:> 107% or Failed upsca.le as seen by "iit' on the Display screen.

Operator A VERIFY Rod Block. o Action: B. REQUEST IMs to check for high output or mismatch. o C. VERIFY flow by either the :> 107% FLOW or 'iii indicated on the APRM Display. o D. REFER TO Tech. Spec Tabl,e 3.3.1.1-1 TRM Table 3..3.4-1, Sect.

5.3.'1. o

References:

2-45E620-6. GEK103936 Technica! Specifications Technical Requirements Manuai-TRM

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OA.11U48 Rw~8 3101 1150

k. Recirculation Flow Monitor Obi V,B.9, V.D.5 V,B.l1 (1 ) Each flow monitor channel consists of two flow inputs used to calculate a Total Recirculatlon Flow, one from Recirculation Loop Aand one from Recirculation loop B.

(2) Ea.ch APRM receives the inputs from Displayed 00 t'j,A'l (4 to 20 rna) dffferential pressure Input Status (aP) transmtters used to measure (lsptay the recircuJafioo hJOp (3) Wh~" rna rurrenUs4 J)

,.fa,. rec~rculafloo value 's s~

to zero (OJl%).

(4) The loop flow continues 10 00 calculated when the macun'ent is

( above 20.0 rna (I.e., the value is not clamped at 125.0% flow).

(5) The flow monitor is considered INOP Inpu!s to flow whenever loop current is less than upscale alarm 1.0 rna oj' greater than 26.0. rna. APRM flow alarm is when flow Is (6) Each ABM receives the four Tota~ >107 % Of Recirculation Flow values tram the upsc~lllt, AiPRM channelS to nbf.Q......"""O status of the flow compare blook (1) The Recirculation Flow Monitor Funcnoo provides the following

,*n,",,,,.,,fU~ for each Tmal

FLOW UPSCALE ALARM (generated by the APRM)

FLOW COMPARE ALARM (generated by the RBM).

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet OPl17U48 ReN~8 38 of t50 (8) The table below sllows the Right relationship between the flow unltl1:rainlcomp monitors and each APRM channel, Discuss effects jf Loop.A LoopS flow APRM1 FT~68~5A FT-6a-8~A transmitters fail APRM2 FT~68-5B FT-68-S1B APRM3 FT-68-5C FT*68 81C M Flow indicators (9--

APRM4 FT-68,.5D FT~68*81D 4)

Flow Recorder (9..4)

\1,8.10 \I,C,3 (9) Since !:he APRM (channefs 3 4) ate used to the ;li!!"1.:;IInn output to the flow indiCa1,OtS recorders on panea 9-4, a loss power, or if the APR.M IS unplugged wo;u!d interrupt the flow signals fol" panel 9-4 indic.at.ions, (10) The APRM coo'llensthe4 to 20 ~

input signals, from the transmtters, to digital signals.

(11) Each APRM calculate,s a Totsl V,B,11 Recirculation Flow value: APRM (a) by averaging its OPRM toop A and loop B recircO&3~ion flow valUes.

(0) EachAPRM uses Total Recirculafion FiOlilI ~tie for its flow biasedtfips Ind atarn.

(c) in addition, lotm Flow is also used fOf Mablfrtg OPRM tnp;s: and .,..1<0.._""

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet 0Pl171 "148 RwiRm8 F'a~ 39 at 150 (12)

{13} The alarm of channel is indicated 00 theAPRM and RBM This is lnd#coated 00 the header and 00 the TRIP INOP status screens, (14) All of th:eflow are non-latchlfig .

(15) The flow transmitters are powered RPS from the APRM channel poVtler supplies. If no power is available to.

the APRM instrument. then no recirculation flow indication wiU be avaflable from that instrument (16) Each APRM sends its total f'low RBM signal to both RBMs for the flow compare function and for signal transmission to. the process computer.

(17) Each RBM chassis compares the foor total flows (one from each Inverse video APRM).

(18) Each RBM uses the flow signal it Home means receives lrom its "home" APRM, or primary/normal alterMte APRM if "'home" APRM is APRM input unavailable. AdditfonsJ information is presented in the RBM section of lesson plan.

(

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet 0Pl171.M8

~8 P.-40uf150 (19) Ead'I APRM ~ its own flowU~is>

processed flow signal to a high III 107%or~

~. A rod block and alarm am lsaued on upscale trip.

(2:0) Bypass of the APAM channel bypasses the flow biasing function.

No separate flow signal bypass.

(21) T* Recirculation Row va&ue Is used for its flow biased trips. rod bk:dcs. and alarms. The flow inputs forfihe aPRM functicnlwiI be

~1ISSd.1aIer in 1h6 CPRM section of this lesson. plan.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Average Power Range Monitoling 2-01-92B Unit 2 Rev. 0036 Page 15 of 2.8 6.0 SYSTE.MOP'ERATIONS NOTES

1) Only one APRMJOPRM can be bypassed at a time.

2) All operations are performed on Panel 2-9-5 unless specificalty stated otherwise.

3) In order to prevent inadvertent roo withdrawal block or Reador scram While operating APRM BYPASS selector switch, always ensure the preViously bypassed channei returns to normal status by observing the BLUE bypassed lights on Pane! 2-9-14 Voters are extinguished prior to selecting any other channel to be bypassed. After bypassing a channel, the applicabile BLUE BYPASSED status lights on Pane12-9-14 Voters should be illuminated prior to testing, operating, or working on that channel .

6.1 Bypassing APRMJOPRM Channel CAUTION

[QA.lC] SPP-10.4 requires approval of the Plant Manager or his designee prior to any planned operation with APRMs bypassed unless bypassing is specifi.cally allowed Within approved procedures. !!SE-NPS*92'-R01J

[1] REVIEW all precautions and limitations.

REFER TO Section 3.0. 0

[2] PLACE APRM BYPASS, 2-HS-92'-7B/S3, to desired channel to be bypassed. 0

[3] CHECK BLUE BYPASSED lights illuminated on Panel 2-9-14, Voters. 0

[4] VERIFY white bypass light on Panel 2-9-5 is illuminated. 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Rod Block Monitor 2-o1-92C Unit 2 Rev. 0033 Page 13of16 6.2 Manually Bypassing RBM Channefs CAUTION

[QAIC] SPP-1O.4 requires approval of the Plant Manager or hisdes~gnee prior to any p4anned operation with RBMs bypassed unless bypassing is specificallY allOlNed within approved procedures. ifSE-NPS*92-ROlj

[1] REVIEW all precauUoos and limitations in sectfon 3.0. 0

[2] PLACE RBM BYPASS. 2-HS-92-7B/S2, to desired channel to be bypassed. 0 (3) CHECK BYPASSED light illuminated for channel bypassed 00 Panel 2-9-14 in Inverse Video, and the WHITE BYPASS llght illuminated on Panel 2-9-5. 0 6.3 Returning RBM to Service from Bypassed Condnion

( [1] REVIEW all precautions and limitations .in Section 3.0. o

[2] PLACE RBM BYPASS, 2-HS-92-7B/S2. to neutral (off). 0

[3] CHECK previously bypassed channel BYPASSED lights extinguished. 0

[4] OBSERVE RBM channel display in operation (Inverse Video) and recorder indicating. 0 7.0 SHUTDOWN None S.OI INFREQUENT OPERATIONS None

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet NPG Standard Reactivity Management Program SPP-10.4 Programs and Rev.OOOS Processes Page 12 Oof 48 3.2.5 Operations Shift Managers (cOonftnued,

13. Contacting Reactor Engineering to provide support when their evaluation Oof the planned maneuver or other reactMty evoliUtion warrants Reactor Engmeenng support.

14. Attend Reactivity Management Review Board when requested.

3.2.6 Unit Supervisors A. Are sensitive to the reacti¥ity effects that may result from normal and infrequent evoMions.

B. Ensure that planned 'IliiJfk activities have received appropriate reactivity management reviews and the necessary controls have been implemented inta the 'IliiJfk packages and/or procedures, induding contfogency or compensatory actions as needed.

C. Place emphasiS during turnover and control board walk dOwns on items important to readivity management.

D. Have the authority to terminate any actMty in whk:n the effects on reactivity contra are unknown or non-consellVative.

E. Have the responsibility and authority to trip the unit if there is uncertainty as to the unifs starus with respect to the control of reactMty and control of the plant F. Ensure that the specific deta~s of events or equipment problems related to the control of reactivity are recorded and initiates corredive actions.

G. Maintains a cautious approach tathe adjustment or interpretation of power indication by questioning the reasons behind discrepancies that may exist between power measurements.

H. Evaluate the rerommendatlions providiOO by the Reactivity Control Plan or verbally by 1I1,e on-call RE. However, the election to take actions more conservative than the reoommendations is within the Reac1Mly Management Philosophy.

I. Ensure all control rod movements are made to a deliberate, carefully controlled manner while constantly monitoring nuclear instrumentation and redunda;nt indications of reactor power and neutron flux.

J. Provide direct oversight (line of sight within normal conversation level' distance) for all reactivity manipulations (may be performed by a, dedicated SRO for major reacti¥ity evoMions sud! as reactor startup).

K. Attend Rea,ctjyjty Management Review Board when requested.

3.2.1 Reactor OperatOR A. Place emphasis during turnover and control board surveillance Qn items important 10 reactivity management

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet N PG Standard Reactivity Management Program SPP-10.4 Programs and Rev.OOOS Processes Page 14 of 48 3.9 Reactivity Control During Approach to Criticality (continued}

0, If thearpproach to criticality is suspended for an extended period of time near the point ofcriticaldty, the reactor core shall be made sUfficlently suocritical to avoid an inadvertent criticality.

E. In the event of an unexplained change in reactivity during.m approach to criticality, the approach to criticality sha!1 cease a!l1d the reactor core sha.li be made suffidently sullcriti'cal to prevent an inadvertent cri1icality. Approval of the Plant Manager or his designee is require<l to resume the approach to criti'ca.lity.

F. Procedural controls shall be in place to ensure th,at cycle-specific information is appropriately updated in the process computer and offline software used to track reactivity-related parameters.

3.10 Reactivity Control During Power A.scension

.A. If discrepancffis (outside allowable acceptance criteria) exist belween reactor power level indicators, power ascension shall cease until the situation is investigated.

Approval of the Plant Manager or his deSignee is required to resume power ascension.

B. tndi,cations of core thermaJ power using the neutron monitoring insirumentation shall be compared to alternate indication to velify consistency. COnservative actions in accordance with Technical Specifications shall be taken when indications are inconsistent C. Reactor power level Increases shall be cons'istent with fuel vendor requirements.

D. In the event of an unexplained change in reactiVity, power ascenSion shaH cease.

Limitations on continued operation shall be based upon Technical Specifications and approval of the Plant Manager or his designee is required to resume power ascenSion.

E. When unexpected core power distribuMn indications are encountered, the cause of the power distribution anomaly needs to be thoroughly investigated and understood by all concerned participants (I.e., nuclear fuel, reactor engineers ami fuel vendOir) before the reactor is operated at higher power ievels. In all cases where the reactor is operating outside expected parameter limits, line management should seek the advice of the appropnate technica! staff ami fuel vendor in order to make a conservative decision regarding whether to allow continued reactor operation or a return to power.

3.11 Reactivity Control At Steady-State Conditions A Shiff Manager has the primary authority over the control of reactivity. The Shift Manager or the Unit Supervisor shall giVe direction for all changes in reactivity. Any planned reactivity changes Implemented in the control room, namely the norma!

manipulation of reactivity contrOils, shall he performed under the oversight of a desIgnated SRO.

B. Operation of reacUvity controls and other mechanisms which may affect the reactivity or power level of the reactor shaIl Oinly I)e accomplished with the knowledge and consent of the licensed Operator "at the controJs~ and! with the approval of the On-Duty Unit Supervisor.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 239002G2.4.2 Tier# 2 Knowledge of system set pOints, interlocks and automatic actions 1 Group #

associated with EOP entry conditions: SRVs KJA# 239002G2.4.2 Importance Rating 4.5 4.6 I Proposed Question: SRO #14 A loss of all off-site power and LOCA have occurred on Unit 1 which resulted in the following plant conditions:

RPV pressure 960 psig and stable.

RPV water level (-) 134 inches and lowering.

Drywell pressure 14 psig and rising.

Drywell temperature 285 of and rising.

Suppression Pool level 15 feet

Suppression Pool temperature 175 of and rising.

Execution of 1-EOI-1, "RPV Control" and 1-EOI-2, "Primary Containment Control" is in progress.

Which ONE of the following describes the required actions based on the given conditions?

A. Exit 1-EOI-1 path RCjP and execute 1-EOI-C2, "Emergency Depressurization."

B. Execute 1-EOI-1 Step RCjP-3 and Rapidly Depressurize the RPV.

C. Execute 1-EOI-1 Step RCjP-7 and lower RPV pressure to stay in the safe area of the HCTL Curve.

D. Remain in 1-EOI-1, "RPV Control" and execute 1-EOI-Appendix 11B, "Re-Open MSIVs" concurrently.

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

b. Incorrect. Rapid depressurization is an anticipatory response to prevent conditions which require Emergency Depressurization. DW temp above 280 of precludes anticipation of ED.

c. Incorrect. Although the guidance in this step wi" apply if SP temperature continues to rise, the action to perform Emergency Depressurization wi" eliminate any possibility of reaching the unsafe area of the HCTL curve.

d. Incorrect. Although the actions to re-open MSIVs may eventually be performed as an option for long term decay heat removal, the requirement to perform Emergency Depressurization exists right now and must take priority.

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 1-EOI-1, "RPV Control" (Attach if not previously provided) 1-EOI-2, "Primary Containment Control" Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 7116/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X Comments: I consider this a KIA match by requiring the candidate to recognize entry requirements to EOI-C2 to use SRVs to Emergency Depressurize based on Drywell High Temperature limits (set points).

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

~~(~)

'--_---'0'-vo~=,."..L~_.~_u~",..,J o =--

~~;;r..

-_._- =.~-

1 Z':... _ _ _

(

". ........ '--.~

7I!11'\k;i' ~'-. ~~

.. -~.-' - -

~ .....----

O~-

.~'z::a.=__~~

EOI-1 o::-:==--

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet EOI-2

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 259002A2.01 Tier # 2 Ability to (a) predict the impacts of the following on the Reactor 1 Group #

Water Level Control system and (b) based on these predictions, use procedures to correct, control, or mitigate the consequences of KIA # 259002A2.01 those abnormal operations: Loss of any number of main steam flow inputs, Importance Rating 3.3 3.4 I Proposed Question: SRO #15 Unit-2 is operating at 100% rated power when the following annunciators are received:

REACTOR WATER LEVEL ABNORMAL (9-SA W8)

MAIN STEAM LINE VS. MAIN TURB STEAM FLOW MISMATCH (9-5B W24)

RFWCS INPUT FAILURE (9-6C W14)

Which ONE of the following describes the current status of Reactor Water Level and the required actions to mitigate this failure?

Actual reactor water level is _~(=1)~__ than normal. Coordinate with Instrument Mechanics and direct bypassing _ _ _ _-4.(=.,2)__- - - - -

(1) (2)

A. ,higher Main Steam Line Flow instrument in accordance with 2-01-3, "Reactor Feedwater System ,"

B. higher Feedwater Line Flow instrument in accordance with 2-AOI-3-1, "Loss of Reactor Feedwater or Reactor Water Level High/Low."

c. lower Main Steam Line Flow instrument in accordance with 2-01-3, "Reactor Feedwater System ,"

D. lower Feedwater Line Flow instrument in accordance with 2-AOI-3-1, "Loss of Reactor Feedwater or Reactor Water Level High/Low."

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

( I Proposed Answer: C Explanation: a. Part (1) is incorrect. A failure of the feedwater flow instrument would not result in a steam line vs. turb steam flow mismatch. Both of the other two annunciators would be valid for either failure. If the failure is a steam flow transmitter, RWL would be lower than normal due to steam flow being higher than indicated. Part (2) is correct. The transmitter is bypassed using 2-01-3.

b. Part (1) is incorrect as in (a) above. Part (2) is incorrect for two reasons.

First, the FW flow transmitter is not the cause of these indications. Second, 2-AOI-3-1 directs bypassing the failed instrument, but the actual steps to accomplish that action are directed by 2-01-3, so the procedure reference is incorrect.

c. Correct answer.

d. Part (1) is correct. A failed steam flow transmitter will result in lower RWL due to a reduction in actual feed flow to compensate for a reduction in indicated steam flow. Part (2) is incorrect as stated in (b) above.

(

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 2-01-3, 2-AOI-3-1, ARP 9-6C W 14 (Attach if not previously provided)

ARP 9-5A W8, ARP 9-5B W24 Proposed references to be provided to applicants during examination: None Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 7/17/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Pa~9-5 2-ARP-9-5A Unit 2 2-XA-55-5A Rev. 0042 Page 13 of 45 Sensor/Trip Point REACTOR RFW Control System WATER LEVEL 2-RlY-D{J3-0053ll :S 27 inches ABNORMAL 2-LA-3-53 2-Rl Y-D03-0053LH 2:39 inches Average of valid Narro..... Range Level signals from RFW Control System. Signals originate from 2-LT-3-53, 3~60, 3-206 and 3-253.

(Page 1 of 1)

Sensor 2-RL Y -003-0053U_ in Panel 2-9-97 (Unit 2 Control Room) location: 2-RLY-003-0053LH in Panel 2~9-97 (Unit 2 Control Room) 2-LT-3-53 on Pane! 25-58 {Rx Bldg E! 593', R10-S}

2-LT-3-60 on Panel 25-68(Rx Blag EI 593', R12-P) 2-lT-3-206 on PaneI2-925-426(Rx Bldg EI593', R10-T) 2-lT-3-253 on Panel 25-6D(Rx Bldg EI593', R12-R)

Probable A. Reactor .....ater level high or low.

Cause: B. Malfunction of sensor.

( None (Reactor scram on tow level at +2 in)

Automatic Action: (Main Turbine/RFPT trip and subsequent scram jf ~30o/() Reactor Power on htgh level at +55 in)

Operator A. VALIDATE Reactor water level hiJlow using multip!e indications Action: including Average Narrow Range Level on 2-XR-3-53 recorder, 2-U-3-53, 2-U-3-60, 2-3-206 and 2-U-3-253 on Panel 2-9-5_ o B . IF aJarm is vatid, THEN REFER TO 2-AOI-3-1 or 2-01-3. o C. IF 2-U-3-53, 2-U-3-60, 2-U-3-206 and 2-U-3-253 has failed or is invalid, THEN with SRO permission, BYPASS the affected level instrument REFER TO 2-01-3 Section 8.2. o

References:

0-45N620-6 2-J29E895-1O 2-47E61 0-46-1 2-]29ES9S-1 2-..4.01-3-1 2-01-3

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Panel 9..

Unit 2 2-XA-55-6C Rev. 0016 17 of 41 SensorlTrip Point:

RFWCS 2-RlY-46 A process input to the RFW Control System INPUT FAILURE dedared Bad or Invalid. A signaf is dedared bad when ~t has failed or is off scale.

2-lA-46-5D A signal is declared inlVajid wtIef! it fails the validation process described in 2-01-3 INustration 8.

(Page 1 of 2)

Sensor Panel 2-9-97 (Behind Panel 2-9-5).

LocBtion:

Probable Any of the fomowing inputs will cause tNs annunciator to' alarm:

Cause; A. RFP A, S, or C Discharge Flow bad or invalid.

B. RfW A or B line FlOW bad or invalid.

C. Main Steam Une Flow A. a, C, or D bad or invalid.

D_ Reactor Pressure (Wide Range) A, 8, or C bad or invalid.

E Reactor Pressure (Narrow Range) bad or invalid.

F. Reactor Water level A, e, C or D bad Of .nvalid.

G. Turbine First Stage Pressure bad.

H. RFW A or 8 Une Temperature bad.

Automatic A RFWCS bypasses a bad or invalid signal from the system.

Action: B. Amber light on the following instruments illuminates when the signal has been automatically bypassed:

RFP Discharge Flows, 2-Fl-3-20,13,6 (Panel 2,.9-6).

RFW Une Flows, 2-FI-3-78A, 788 (Panel 2-9-5).

Main Steam Une Flows, 2-FI-46-1, 2, 3, 4(PaneI2-g,-5).

WR Reactor Pressu:re 2-PI-3,-54, 61,201 {Panel 2-9-5).

Reactor Water leveI2-U-3-53, 60,206,253 (Panel 2-9-5).

Operator A. VERIFY RFWCS continues to mainmrll Reactor Wa.ter level.

Action: S. IOENTIFY badtinvalid signal by checking Control Room instrumentation and/or ICS. REFER TO A.TTACHMENT 1 on next page for list of RFWCS instrumentation. REFER TO ICS RX FW l Vl CONTROL SYS display (FWLCS).

C. REQUEST assistance from Si:te Engineering.

D. BYPASS the badfmvalid signal wi~h Unit Supervisor approval. REFER TO 2-01-3.

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

(

BFN Pane'19-5 2-ARP-9-S8 Unit 2 2-XA-55-SB Rev. 0023

~ PCige 28 of 43 SensorlTrip Point MAIN STEAM LINE VS. MAIN TURB 2-Rl Y-046-0007 Total Steam Flow to Turbine First Stage Pressure (flow equivalent) devLate by more STEAM FLOW than :bO.S Mlblhr for 30 seconds. This is MISMATCH determined by the RFW Control System by 2-XA-46-7 signals originating from 2-PT-1-81 and 2-FT-1-13, 25,.36, and 50.

(Page 1 of 1)

Sensor 2-RLY-046-0007 in Pane4 2-9-97(Unit 2 Control Room)

Location: 2-FT-1-13 on 2-lPNl-925-0056A (R.x Bldg EI541 NE QUAD) 2-FT-1-25 on 2-LPNl 925-D056A (Rx Bldg El 541 NE QUAD) 2-FT-1-36 on 2-LPNl-925-0056B (Rx BI£lg EI541 NE QUAD) 2-FT-1-50 on 2-lPNL-925-0056B CRx Bldg Ei 541 NE QUAD) 2-PT-1-81 on Panel 25-11HTurb Bldg EI5B6 T-10-J-Une)

Probable A, Small steam line break.

Cause: B. Unit startup (Mode 2) or shutdown (Mode 3).

( C.

D.

Bypass Valves Open.

Sensor malfunction.

E. Main Steam Une Flow instrument failed (2-fl-46-1 through 46-4 on Panel 2-9-5).

Automatic None Action:

Operator A, IF a Main Steam Une Flow instrument has failed or is illvalid, THEN Action: with SRO permission, BYPASS the affected instrument REFER TO 2-01-3 Section 8.4. o B. CHECK main steam tunnel temperature on LEAK DETECTION SYSTEM TEMPERATURE, 2-TH39-29, Panel 2-'9-21. o C. VERIFY Rx and Tum Bldg fans on fast speed. o D. DISPATCH personnel to determine leak location. o E. IF steam leak: presents personnel safety hazard or radiological proNem, THEN EVACUATE the affected area. o F. IF leak CANNOT be isolated, THEN PLACE Reactor In HOT STANDBY CONDITION (Mode 3} with MSIVs closed, or COLD SHUTDOWN CONDITION (Mode 4). o G. REFER TO Tech Spec Table 3.3.6.1-1. o

References:

2-47E61O-1-1 2-729E895-10 2-45E620-6 2-4 7E61 0-46-1 2-01-3 Tech n ical Specifications

(

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

(

BFN l.oss of Reactor Feedwateror Reactor 2-AOI-3-1 Unlt 2: Water Level HighllowRev. 0020 PageS of 16 5.0 LOW REACTOR WATER LEVEL OR LOSS OF FEEDWATER (continued)

[2] IF FeedWater Flow signal fails (FI-3-78A, FI-3-78B), THEN PERFORM the following:

A With SRO's permission., REFER TO 2-01-3 and BYPASS failed FeedWater Flow Instrument in Unit 1&2 Computer Room or Unit 2 Aux Instrument Room. 0 12_1J IF both FeedWater Flow Instruments faH,THEN VERIFY level control transfers to SINGLE ELEMENT. 0

[3] IF Steam Flow signal fails (FI-46--1,2,3,4}. THEN PERFORM the following:

With SRO's permission, REFER TO 2-01-3 and

( BYPASSmiled Steam Flow Instrument in Unit 1&2 computer Room or Unit 2 Aux Instrument ROOfR o

[3_2] IF three Steam Flow Instruments fair, THEN VERIFY level control transfers to SINGLE ELEMENT. 0

[4] IF Reactor Water Level signal fails (LI-3-53, 60,206,253),

THEN PERFORM the following: o With SRO's permission, REFER TO 2-01-3 and BVPASSfailedleve1instrument on Panel 2-9-5_ o

{42} IF four Reactor Water Level instruments faH, THEN PERFORM the following:

[421] VERIFY ~Icontrol transfers to MANUAL o

[422] MAINTAIN Reactor Water Level in MANUAL mode_ o

[5] VERIFY dosed an Safety/ReHef Valves_ o

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet SFN Reactor ,Feedwater System 2-01-3 Unit 2 Rev. 0118 Page 83 of 209 8.4 Bypassing RFWCS Main Steam Flow Instrumentation CAUTION Bypassing steam flO"N signal for a partiaUy or completely isrnated steam Hne could result in the RFW Control System raising Reactor water Level to Main Turbine/RFPT high water level trip setpelnt when in THREE ELEMENT control. Even though isolated steam line will produce a steam now signal of low value, the signal is sml valid and is used by the RFW Control System when in THREE ELEMENT control, for calculating avefage steam line now.

NOTES

1) RFW Control system will allow up to two Main steam line Flow instruments to be bypassed at a time. Should a third steam Row instrument be bypassed or fail with two other instruments bypassed, RFVVCS control will automatically transfer to SINGLE ELEMENT.

2) Illustration 8 can be referred to for general informati,oo on RFWCS instrumentation.

[1] REQUEST aSs4stance from Tech Support and IMs. 0

[2] OBTAIN Unit Supervisor approval. 0

[3] BYPASS any of the following Main steam line Flow instruments at AW51 Work Station in Unit 1/2 Computer Room or at Panel 2-9-18 in Unit 2 Auxiliary Instrument Room:

LINE A, 2-FI-46-1 0

LINE 8,. 2-FI-46-2 0

LINE C, 2-FI-46-3 0

LINE D, 2-FI-46-4 0

[4] CHECK amber light illumfnated on steam line tlow instrument bypassed (Panel 2-9-5).. 0

[5J IF Steam line flow instrument was bypassed due to signal failure, THEN (Othel1Nise N/A)

CHECK RFWCS INPUT FAILURE annunciation, 2-XA-55-6C Window 14, will reset {N/A if annunCiation was in alarm prior to signa! failure). 0

[6] VERIFY RFVV Control System continues to maintain Reador Water level. 0

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Reactor feedwater System 2 ..01-3 Untt2 Rev. 0118 Pa~84of209 8.5 Bypassing RFWCS RFW Une Flow Instrumentation:

NOTES 1} The RFW COntrol System will allow up to one RFW Line Flow instrument to be bypassed at a time. With one instrument bypassed, should the other Feed Line Flow instrument be bypassed or fall, RFWCS control will automatically transfer to SINGLE ELEMENT.

2} Illustration 8 can be referred to for genera! information on RFWCS instrumentation.

[1] REQUEST assistance from Tech Support and IMs. o

[2] OBTAIN Unit Supervisor approval. o

[3] With Unit Supervisor approval, BYPASS any one of the following Feed Line FJowinstTuments at AW51 Work stat~on in Unit 112 Computer Room or at Panel 2-9-18 in Unit 2 Auxiliary I nstrument Room:

LINE A, 2-FI-3-7BA o

LINE S, 2-FI-3-78S o

[4] CHECK amber light muminated on the Feed Line Flow instrument bypassed (Panel 2-9-5). o

[5] IF Feed Line Flow instrument was bypassed due to signal failure, THEN (OtheIVVise N/A)

CHECK RFWCS INPUT FAILURE annunciation, 2-XA-55-6C Window 14, will reset (NJAif annunciation was in alarm prior to signal failure). 0

[6] VERIFY RFW Control System continues to maintain Reactor Water leveL 0

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

(

BFN Reactor FeedWater System 2-01-3 Unit 2 Rev. 0118 Page 197 of 209 Illustration 8 (Page 3 of 7)

RFWCS InStrumentatiion 2.3 System Operation The RFW Control System will use a Main stearn Une Flow signal provided the system determines signal to be good and valid. A GOOD steam 1IO"N signal is one that has NOT failed and is 00 scale. A VAUD steam flow signal is determined by a validation process described in the next paragraph.

Total steam Row is validated by comparison to Turbine First stage Pressure signal stearn flow equivalent If those two signals dilef by more than .8 x 10' Ibmll1r for more than 3 seconds AND Total Steam Flow is greater than 19%, then each of the steam line flow signals will be ,compared to average steam line flow. If any infNidual signal deviates from the average by more than 1.2 x 106 1bmlhr, then the signal is declared invalid and is automatically bypassed.

2.4 Failure Mechanisms WIllen the RFWCS declares a steam Une FJow signal bad or invalid, the flow instrument isamomatically bypassed. The amber light on the affected instrument will illuminate and the RFWCS INPUT FAILURE annundation (2-XA-55-6C, window 14) will alarm.

When the RFWCS declares the Turrnne First stage Pressure signal to be bad, the RFWCS INPUT FAILURE annunciation (2-XA~5-6C,w1ndOW 14) wil alann.

Thefoltowingevents will automatically transfer RFWCS control to SINGLE ELEMENT:

Total Steam Flow <: 19%.

More than two Steam ~jne flows declared invalid.

If 1 or 2 steam line nows are invalid AND Turbine First Stage Pressure is Invalid.

When Total Steam Ftowto Turbine First stage Pressure devmes by more than

.8 x 1o'Ibmlhrfor more than 30 seconds, MAIN STEAM LINE VS STEAM FLOW MISMATCH annunciation (2-XA-55-5B. window 24) wiftl alarm.

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

(

BFN Reactor FeedWater System 2-01-3 Unit 2 Rev. 0118 Page 199 of 209 Illustration 8 (Page 5 of 7)

RFWCS InstrumentatIon 3.3 System Operation RFW Control System will use a Feedwater line now signal provfded the signal is good and va[id*. A GOOD line now signal is one that has not failed and is on sca~.

A VALID signal is determined by a validation process described in the next paragraph.

Indrvidual Feedwater Line Flows are validated by comparison to each other.

If line flows deviate by more than.8 Mlbmlhr and' Total Steam Flow is :> 19% rated, then each Feedwater line flow is validated agarnst one half of the total of the valid indMdualRFP discharge flows*. If either Feedwater Line FIO\iV signal deviates from the total of the RFP discharge flows by more than .8 x1 06lbm/hr, then Feedwater U ne Flow signal is invalid and bypassed by the system.

RFP discharge flows are used for the auto now balancing feature of the RFIN Control System. Individual flows are subtracted from the operator supplied flow bias

( with resultant error signal sent to individual RFP flow balance blocks in RFWCS.

RFP discharge flow signals are also used for controlling RFP Minimum Flow Valves.

The aSSOciated minimum flow valve will open when RFP discharge flow falls below 2000 gpm and dose when RFP discharge flow exceeds 3000 gpm. RFP Discharge Flows are utilized by the Recire System for 75% Pump Runback (any individual RFP Discharge Flow <: 19% AND Reactor Level <: 27 inches).

3.4 Failure Mechanisms WhenRFWCSdeclares Feedwater Line Flow or RFP Discharge Flow signal bad or invalid, the signal is automatically bypassed (amber light illuminates on the instrument) and RFWCS INPUT FAILURE annunciation {2-XA-55-6C, window 14}

will alarm.

When RFWCS declares aRFW Line Temperature signal bad, RFWCS INPUT FAILURE annunciation (2-XA-9-6C, window 14) wilt alarm.

If both Feedwater line Flows arededared bad or invalid, then RFWCS will automatically transfer to 81 NGLE ELEMENT control.

If both temperature inputs are lost from the associated Feedwater Line or there is a deviation between the two sensors on the sa:rne Feoowater Line of greater than 5°F then the average temperature signal isdeciaroo bad by the system. A default temperature signal of- 380°F is produced for density compensation.

( If RFP Discharge Flow signa. is bad, the minimum flow valve will open.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 239001G2.2.2S Tier # 2 Knowledge of the bases in Technical Specifications for limiting 2 Group #

conditions for operations and safety limits: Main and Reheat Steam.

KIA # 239001G2.2.25 Importance Rating 3.2 4.2 I Proposed Question: SRO #16 Which ONE of the following describes the Technical SpeCification limits for closure times for the Main Steam Isolation Valves (MSIVs) and the bases for that limit?

The MSIVs are designed to close _ _---->..C..=.l)/---_ _ in order to _ _ _ _--->.C.=2).I--_ _ __

(1) (2)

A. faster than 3 seconds minimize the radiological dose following a DBA LOCA.

B. faster than 5 seconds minimize the mass loss following a DBA LOCA.

c. slower than 3 seconds minimize the RPV pressure transient following a Group I Isolation.

D. slower than 5 seconds minimize Primary Containment pressure following a steam line break.

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

( I Proposed Answer: C Explanation: a. Part (1) is incorrect. MSIV closure time is > 3 seconds and < 5 seconds.

Part (2) is incorrect. The accident of concern for minimizing radiological dose is a steam line break outside containment, not a DBA LOCA. In that case the limit is < 5 seconds, not < 3 seconds.

b. Part (1) is correct for a Main Steam Line Break but not for a DBA LOCA.

The mass lost during a DBA LOCA is from the severed Recirculation Loop which is essentially unaffected by MSIV closure times.

c. Correct answer.

d. Part (1) is incorrect. The FSAR analysis for a steam line break assumes a 10.5 second closure time for conservatism but that is not the Tech Spec limit.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): FASR Chapter 14.6 (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 7/18/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 X Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEillANCE FREQUENCY SR 3.6.1.3.5 Verify the isolaUon time of each power In accordance operated, automatic PCIV, except for MSIVs, with the Inservtce is withjn limits. Testing Program SR 3.6.1.3.6 Verify the isolation ti'me of each MSIV is 2:: 3 In accordance seconds and s 5 seconds. with the lnsrvtce Testing Program SR 3.6.1.3.7 Verify each automatic PCN actuates to the 24 months iso~ation position on an actual or simulated isolation signal.

SR 3.6.*1.3.8 Verify a representative sample of reactor 24 montfls instrumentation line EFCVs actuate to the isolation position on a SImulated instrument line break signal.

SR 3.6.1.3.9 Remove and test theexptosive squib from 24 months on a each shear isolation valve of the TIP System. STAGGERED TEST BASIS SR 3.6.1.3.10 Verify leakage rate through each MSN is In accordance

~ 100 seth and that the combjned leakage Vvith the Primary rate for aU tour main steam lines is ::; 150 scfh containment when tested at 2:: 25 psig. Leakage Rate Testing Program SR 3.6.'1.3.11 Verify combined leakage through water In accordance tested lines that penetrate primary with the Primary containment are within the limits spedfied in Containment the Primary containment Leakage Rate Leakage Rate Testing Program. Testing Program BFN-UNIT 1 3.6-16 Amendment Nos. 234, 2§3 294261 September 27, 2006

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet BFN-22 in adldition to those specified for the loss of coolant acadent described in paragraph 14.6.3.1.)

G. The reactor is assumed to beirlitially operating at: the conditions specified in Table 14.6-3. Tables 14.6-4 and 14 . 6-5 provide additi0na4 conditions that apply for the short term containment response and long term containment response, respectively.

b. The reactor is assumed to go suibcritical at the time of acddent initration due to void formatioo in the core region. Scram also occursirlless than one secood from receipt of the* high drywell pressure and low water !level' signals, but the difference in shutdown time bebNeen zero and one second is negligible.

c. The sensible heat released in cooIting the fuel to the normal primary system operating saturation temperature and the core decay heat were indUded in the reactor vessel depressurization calculation. Initial high vessel pressure increases the catculated flow rates out of the break; this is conservatiVe for containment analysis purposes.

d. The main steam isolation valves were assumed to start closmg at 0.5 seconds after the accident, and the valYeS were assumed to be fuRy dosed in the shortest possible time of three seconds fOllIoW'ing dosure initiation.

Actually. the closures of the main steam isolation valYeS are expected to be the result of low water level, so these valves may not receive a signal to dose fOr over four secondS; and the dosing time could be as high as 10 secoods. By assuming rapid dosure of these vatves, the reactor vessei is maintained at a high pressure Which maximizes the discharge of high energy steam and water into the primary containment.

e. For the short term containment response analysis, the feedwater now is

'assumed to coast down to zero at four seconds into the event This conservatism is used because the relatively cold feedwater flow, if considered to continue, tends to depressuliZethe reactor vessel, thereby, reducing the discharge 01 steam and water into the primary containment.

1. For the tong term rontainment response analysis, the reactor feedWater flOW into the reactor continues until all the high energy feedwater (water 1hat wwld contribute to heating the pool) is injected into the vessel.

g. The pressure response of the containment is caf,culated assuming~

1. Thermodynamic equilibrium in the drywell and pressure suppression chamber. Because complete mixing is nearly achieved, the error 14 . 6-13

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN-22 outside the secondary containment. Fig,ure14.6-7 shows the break location. The analysis of the accident is described in three parts as follows:

a. Nuc.ear System Transient Effects This includes analysis of the changes in nuclear system parameters pertinent to fuel performance and the determination of fuel damage.

b. Radioactive Material Release This includes determination of the quantity and type of radioactive material re~eased through the pipe break. and to the environs.

c. Radiological Effects This portion determines the dose effects of the accident to control room and offsite persons.

14.6.5.1 Nuclear System Transient Effects 14.6.5.1.1 Assumptions The following assumptions are used in evaluating response of nuclear system parameters to the steam line break accident outside the secondary containment

a. The reactor is operating at the power associated With maximum mass release.

b. Reactor vessel water level is normal for initial power level assumed at the time the break occurs.

C. Nuclear system pressure is normal for the initial! power ileveL d.. The steam pipeline is assumed to be instantly severed by a circumferential break. The break is physicaUy arranged so that the coolant discharge through the break is unobstructed. These assumptions result in the most severe depressurization rate of the nuclear system.

e. For the purpose of fuel pertormance. the main steam isolation valves are assumed to be closed 10.5 seconds after the break. This assumption is based on the 0.5 second time required for the development of the automatic isolation signal (high differential pressure across the main steam line flow restrictor) and the 10-second closure time for the valves.

For the purpose of radiological dose calculations, the main steam isolation valves are assumed to be dosed at 5.5 seconds after the break. Faster main steam isolation valve closure could reduce the mass loss unti r finally some 14.6-33

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

(

BFN-22 other process tine break would become controlling. However, the resulting rad101ogical dose for this break would be less than the main steam line break with a five second valve cl:osure. Thus, the postulated main steam line break outside the primary containment INith a five second isolation valve closure results in maximum calculated radiological dose and is, therefore, the design basisaccfdent t The mass flow rate through the upstream side of the break is assumed to be not affected by (solation valve closure until the isolation valves are closed far enough to establish Hmiting cf(tlcal flow at the valve location. After limiting Critical flow is established at the isolation valve, the mass flow is assumed to decrease linearly as the valve is closed.

g. The mass flow rate through the downstream side of the breaK is assumed to be not affected by the closure of the isolation valves in the unbroken steam lines until those valves are far enough closed to establish limiting criUcal flow at the valves. After Umiting critical flow is established at the isolation valve positiOns, the mass flow is assumed to decrease linearly as the valves close.

h. Feedwater flow is assumed to decreaselinearty to zero over tile first five seconds to account for the slowing down of the turbine-driven feed pumps in response to the rise in reactor vessel water levet

i. A loss of auxiliary AC power is assumed to occur simultaneous with the break. This resuttsin the immediate loss of power to the recirculation pumps. Recirculation flow is assumed to coast down with a three second time constant 14.6.5.1.2 Sequence of Events The sequence of events follOWing the postulated maiin steam Unebreak is as follows:

The steam now through both ends of the break increases to the value limited by Critical flow conslderations. The fiowfrom the upstream side of the break is limited in'itially by the main steam line flow restrtctor. The flow from the downstream side of the break is limited initially by the downstream break area. The decrease in steam pressure at the turbine inlet initiatesc!osure of the main steam isolation valves VYithin about 200 milliseconds after the break occurs (see Subsection 7.3 "Primary Cootajnmentlsolation System"). Also, main steam isolation vaive closure signals are generated as the differential pressures across the main steam line flow restrictors increase abOve [solation setpoints. The Instruments senSing flow restrictor differential pressures generate isolation signals within about 500 milliseconds after the break occurs.

14.6-34

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO 24S000A2.06 Tier# 2 Ability to (a) predict the impacts of the following on the Main 2 Group #

Turbine Gen. / Aux. system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of KIA # 245000A2.06 those abnormal conditions or operations: Loss of extraction steam Importance Rating 2.9 3.1 I Proposed Question: SRO # 17 Given the following conditions:

Unit 2 had been operating at 100% rated power for 370 days since refueling.

The "A" high pressure heater string, 2A1 and 2A2, have been isolated due to a high level and tube leak in the A2 heater. And all required actions have been completed.

Which ONE of the following describes the highest allowable steady state power limitation and most limiting MCPR limit under the current plant conditions?

Maintain _ _ _ _~(1=+)----. The most limiting MCPR limit is (2)

REFERENCE PROVIDED (1) (2)

A. less than 95% power in accordance with 2-AOI-6-1A, 1.50 "High Pressure Feedwater Heater String/Extraction Steam Isolation. "

B. less than 95% power in accordance with 2-AOI-6-1A, 1.48 "High Pressure Feedwater Heater String/Extraction Steam Isolation. "

c. less than 920 MWe in accordance with 2-01-6, "Feedwater 1.50 Heating and Misc Drains System" Illustration 1.

D. less than 920 MWe in accordance with 2-01-6, "Feedwater 1.48 Heating and Misc Drains System" Illustration 1.

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

( I Proposed Answer: C Explanation : a. Reducing power by 5% is an action required due to the loss of extraction steam, however the limitations in 2-01-6, Illustration 1 are more limiting for the Steady State conditions specified in the stem. This is determined by recognizing that a tube leak into the heater will result in automatic isolation of extraction steam and manual isolation of feedwater flow through the heater. The most limiting MCPR limit is correct for the current plant conditions.

b. Same as (a) above. In addition, the MCPR limit is incorrect. The limit of 1.48 is for Atrium-10 fuel. GE14 fuel is more limiting. The candidate must recognize, from the given conditions, the current cycle exposure as well as the status of Scram Time Testing. Specifically, if all Tech Spec surveillances are current and satisfactory, Nominal Scram Speed (NSS) limits apply and not Tech Spec Scram Speed (TSSS) limits.

c. Correct answer

d. The power/generator load limit is correct for the given conditions. Since all required actions are complete, the feedwater flow must have been isolated and the limitations of 2-01-6, Illustration 1 apply. The MCPR limit is incorrect as in (b) above.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 2-01-6, 2-AOI-6-1A, U2 COLR (Attach if not previously provided)

Proposed references to be provided to applicants during examination: U2 COLR Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X RMS 7/6/2008 Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X Comments:

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet BFN Feedwater Heating and Mise Drains :2-0i-6 Unit :2 System Rev. con Page 10 of 142 3..0 PRECAUTIONS AND LIMITATIONS (continued)

G. When more than one condensate and condensate booster pump is in service, the tube side of two higher low pressure Feedwater heater strings should not be isolated at the same time.

H. When placing isolated Feedwaterheaters in service, Feedwater flow may vary, resulting in reactor water level osdllatiol1S.

L Assuming that throffie steam flow is not changed, the following win occur when a FeedWater heater is removed from service:

1. If a number one Feedwater heater is removed from service, generator output Will be higher. This occurs because the normally extracted steam oow passes through the iow pressure tUrbine.

2. When other than number one Feedwater heater is removed from service, extraction to the next higher FeedWater heater will be higher because the Feedwater temperature rise across the heater is greater than before.

A slightly lower generator output win occur.

3. Turbine thrust bearing 'oading may be higher due to ioadimbalance caused by loss of extraction point

4. Although not likely, a turbine thrust bearingadivelpassive plate reversal could occur due to load imbalance. If this occurs the turbine thrust bearing trip setpoints should be checked. (REFER TO 2-01-47).

5. When operating with FeedWater heater(s) out of service there Will be a loss of effidency.

J. The limitations of mustfatron 1, Maximum Turbine-Generalor Load Allowed When Any Feedwater Heater is Not in Service, shaU be followed when removing Feedwater heaters from service.

K. When extraction steam is shut off to a Feeclwater heater, generator output must be within the limits of Illustration 1, which is only evaluated for a maximum of 3458 MWt. Additionally, PCIOMR constraKnts may require a pOlNer reduction prior to removal of FeedWater heaters. A reduction of core flow alone should prevent a violation of the envelOpe, but some power shap1ng may be necessary.

L. Do not attempt to maintain Feedwater heaters at oormal level when the associated level controller is in manual operation for long periods of time.

1. Operation with low level can result in heater damage.

2. Operation with higtllevel will cause loss of efficiency and may result in turbine damage.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN High Pressu.re Feedwater Heater 2-AOI-6-1A Unit 2 String/Extraction Steam Isolation Rev.001S Page SofS 4.0 OPERATOR ACTIONS 4.1 tmmediate Actions

[1] REDUCE Core Thennal Poltjver to ~ 5% below initial power level to maintain thennai margin. o 4.2 Subsequent Actions

[1] REFER TO 2-01-6 for turbine/heater load restrictions. o

[2] REQUEST Reactor Engineer EVALUATE and ADJUST thenna! Hmits, as required. o CAUTION Failure to reduce core power if fUel is operating at or near the preconditionedenveJope in any region of the core may result in fUel damage.

[3] ADJUST reactor power and flow as directed by Reactor Engineer/Unit Supervisor to stay within required thennal and feedwater temperature limits. REFER TO 2-GOI-1 00-12 or 2-GOI-100-12A for the power reduction. 0

[4] ISOLATE heater drain flow from the feedwater heater string by closing the appropriate FEEDWATER HEATER A2(82)(C2)

DRAIN TO HTR A3(83)(C3), 2-FCV-'B-94(95)(96). 0 (5] VERIFY automatic actions occur. REFER TO Attachment 1. 0

[6] MONITOR TURB THRUST BEARING TEMPERATURE, 2-TR-47-23, for rising metal temperature and posStDle active/passive plate reversal. 0

[7] DETERMINE cause which required heater isolation and PERFORM necessary correcttve action. 0

[8] WHEN the condition which required heater isolation is no longer required, THEN RESTORE affected heater. REFER TO 2-01-6. o

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Feedwater Heating and Mise Drains 2-0J-6 Unit 2 System Rev. 0017 Page 133 of 142 Illustration 1 (Page 1 of 2)

Maximum Turbine-Generator Load Allowed wilen any Feedwater Heater is Not in Service HEATERS OUT (Tube and Shell Side)*"

one HP st~fl9, 920 MWe (79G~)

one lP string 920 MWe{79%)

one HP ami LP string 920 MWe (79%)

HEATERS OUT (EXTRACTION STEAM ONLY)*

.* These MWe limits are only evaluated for a maximun of 3458 Mvvt A new evaluation will be required for any reactor power greater than 3458 Mwt.

2A1, 2B1, 2C1 There are !10 extr~ sim. lsol. 1riws, for Htrs" & 5. 'l'i'hen Htts 4 &. 5 are me,,-Mioned

2A2, 2B2, 2C2 AIry Combtnabons in table this indicates thililheir eX~!l See page two 2A3. 2B3, 2C3 bypass vilhres are opened.

one Vessel of Number 4 Heater from Either ~!!!,:!g~,. f3,.~ or C Two Vessel of Number 4 Heater from Any 2 Strings (A. B. or C) NOT Limiting

All Three Vessels of Number 4 Heater NOT Li!l1i~i~~

one Vessel of Number 5 Heater from Either String A,{3' or C NOT Umit!~~ ,",

Two Vessel of Number 5 H~.ater from Any 2 Strings {A, B! or C) NOT Limitjn~ ,

AU Three Vessefsof Number 5 Heater

~ny 2 Ves~E;ts of No. 5 & No.4 from Any string "

Any 3 Vessels of No. 5 & No.4 from Any string

Any'~vessels of NO.5 & No.4 from.,A,ny string

An'y,~, Vessels of No. 5 & No.4 from Ar/Y. string All of No.5 And All of No. 4 2~, 2A~,.2A5

'i .)\il. ~ Vesset:s of One Stri!.l9 t!:1~5, 2M, "?f\3,?A2~ ,?!-1 952 Tne Ilmitatlons apply to the combinalions indicated and equi ....alent combinations. For instance, the restriction for operation with heaters lA3, 2A4 and 2A.5 out of service is equally appli'l:alJle to operation wilh heaters 283, 284 and 285 or 2C3, 2C4 and 2C5 out of service.

U It is permissible to operate at power levels above 79% as long as the requirements of Sections 8.1 and/or 8.3 are met

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet BFN FeedWater Heating and Mise Drains 2-01-6 Unit 2 System Rev.OOn Page 134 of 142 Illustration 1 (Page 2 of 2)

Maximum Turbine-Generator Load AJlowed when an"y FeedWater Heater is Not in Service HEATERS OUT (EXTRACTION STEAM ONLY)*

MAXIMUM GENERATOR 2A1 2A2 2A3 2B1 2B2 2B3 2C1 2C2 2C3 X

X X

X X 1064 X X X X X X X X X 1064

~

X X X X X X X X X X X X 1064 X X X X x

~i=J X X x x x x x x x X 1064

" ""'~"'"""""""T064'"""""~'""-"'"

X X X X

'"'""-"~1064 M_'_

x x x x x X x x X 1064 X X X X X 1064 X X X X 1008 X X X X X 1064 X X X X X x ___J!!t5.L._

x x x x X 1064

-"""'"""'"""""""foS;r"""'""""""'""-

X X X X X x ,.,-" 1008 . _ _.

x x x x x X X X X X X 1008 X X X X X X 1064 X X X X X X X 1064 X X X X X X 1064 X X X X X X 1008 X X X X X X X 1008 X X X X X X X 1008 X X X X X X X 10013 X X X X X X X X 1008 X X X X X X X X 1008 X X X X X X X X X 1008

The fimiIIaIicns af!IfIiy 10 the oombinations indicated and equivalent ccml:!inaliDns.. For ins:tance. I!1e restriclilln _ operation wilt! healers 21\1 andlA2 ootofsel1lice is equililyapplicilble10 operalicn'lWh IleaWs 2B1 and 2B2 or2C1 and 2C200tof seI1Iice. Tolly1odarify.

Any wmber 1 heater may be 5UbstiiIuted _ any llilher IIIImber 1 heaililf. any rwmber 2 healer may be subs1iMeII _ilny aller rwmber 2 Ileaier. and ally number 3. heille may be ~ for any other number 3. healer". ConIiK:t Systiem Engineer for I,miher GlilrifiiciIliion.

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

( TVA Nucil!a:r Fuel TVA*COLR-BF2CIS Core Operating Limit5 Rl!JKllf Revision. 0, Page 10 ScramS~ MCPRP BOCtON'EOC NSS Table 1 TSSS Table 2 BOCtoEOC NSS Table 3 TSSS Table 4 BOCtoCD NSS Table 5 TSSS Table 6

d. Scram Speed Dependent Limits (TSSS vs . NSS)

MCPRP limits are provided for !viiO different sets of assumed scram speeds. The Tedmical Specification Scram Speed (TSSS) MCPRJ, limits are applicable at aU tunes as long as the scmn time sllfVeillance demon.strates that the times in Technical Specification table 3.1. 4-1 have been met. Nominal Scram Speeds (NSS) may be used as long as the scmn time surveillan.cedemonstrates that the times in the follo\\<ing table are met (Ref 9).

Notch Nominal Scram Spero Position (seconds) 46 0.42 36 0.98 26 1.60 06 2.90 In demonstrating compliance v.rith this table, the same surveillance requirements from Tedmical Specification 3.1.4 apply, except that the definition of SLOW rods should conform to the scram speeds in the table above. If conformance to this table is not demonstrated, TSSS MCPRJ,limits shall be used.

On initial cycle startup, TSSS limits are l1sed until the successfulc01llpletion of scram timing confinns that NSS limits may be used.

b. Fuel Type Dependent Limits Separate MCPRl' limits are provided for the GE14 and AI0 fuel types.

Browns Ferry Nuclear PlantlJnit 2 Cycle 15

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet TVA Nuclur Fuel TVA-COLR.~BF"-C15 Core Opeming Limits Report Revisioo 0, Page 18 Table 1: MCPR,p Limits for DOC to NEOC Exposures - NSS ScraJlll Times (App1icabl'e uo to Core A:verare Exposure of27,.788 MWd I MTU)

MCPRrLimit MCPLLimit EOOS Power EOOS P_EJ'

(~RaIetl)

AlII GEl4 AIO GEl4 OPDoIl OotioJl ("RaI).

100 1.45 1.46 100 1.48 150 69 1.58 1.62 69 1.61 1.64 63 1.61 1.64 63 1.~ l.70 58 1.69 1.7S 1.73 58 1.15 --

la-58 1.78 58 US 1:79 46 1.85 U9 FHOOS 46 1.90 1.97 Semee 30 2.22 235 30 2.:33 2..49 30(>sm~ 2.64 2.79 30 <> sm'.J.') 2.75 2.92 25(>~ 2.89 3.08 25 (> Sll"Q) 3.03 3.24 30~~ 2.51 2.68 30(!f~ 2.60 2.79 25~~~ 2.68 2.S9 25 (!:50%f) 2.81 3.03 100 1.45 1.46 100 1.48 1.50 69 1.:58 1.62 69 1.61 1.64 63 1.61 1.64 63 1.~ 1.70 58 1.69 1.73 58 1.75 -

58 US US 58 1.1S 1.79 ltPTOOS 46 US U9 RPTOOS 1.9(1 46 1.97 30 2.22 235 FHOOS 30 2.33 2.49

(~ 30 (:> SO%p) 2.<64 2.79 30 (:> 5O%fj 2.75 2.92 25 (:>j()"/.1') 2.89 3.08 25 (> 50%f) 3.03 .3.24 30~~ 2.51 2.68 30 (!:sO%f) 2.60 2.79 25~~~ 2.68 2.89 25 (!:SO%f) 2.81 3.1)3 100 1.49 1.50 100 1.51 l.S3 69 1.62 1.~ 69 1.64 1.66 63 1.65 1.67 63 1.68 1.72 58 58 1.71 US 1.75 1.18 58 58 1.77 1.78 un TBVOOS 46 1.85 1.9(1 TBVOOS 46 1.92 1.H 30 2.23 2.36 moos 30 2.35 2.49 30(:>SO.,~ 3.09 3.20 30 (> SO%f) 3.19 HI 25 (:>SO%F) 3.51 3,64 25 (>50%f) 3.62 Hi 30~~ 2.64 2.19 30 (!: 50"411) . 2.12 2.SS 25~~ 2.97 3.18 25 (!f5O%f) 3.G? 3.30 100 1.49 LSO 100 L51 1.53 69 1.62 1.~ 69 1.64 1.66 63 1.65 1.67 63 1.68 1.72 58 51!!

1:71 1.78 1.75 1.78 RPTOOS 58 58 1.77 --

RPTOOS 1.78 Ul TBVOOS 46 1.85 1.90 TBVOOS 46 1.92 1.98 30 2.23 2.36 moos 30 2.35 2.49 30(::>~ 3.09 3.20 30 (>50%f) 3.19 HI 25 (:>j()"Q) 3.51 3.64 25 (:> SO%f) 3-62 3.7&

30~sm-~ 2.64 239 30 (!fsO%f) 2..72 :US 25~j()"Q) 2.97 3.18 25 (S5O%f) 3.G7 3.30 Add G.02 to the above MCPR. limits foc SLO.

BT"tm'1JS Ferry NlIdmr PltmIUlfiJ 2 Cycle 15

(\ REFERENCE MATERIAL Provided to CANDIDATE c

TV A Nuclear Fuel QADocument TVA-cOLR-BF2C15 Core Operatmg Limits Report Revision 0, Page 1 Browns Ferry Nuclear Plant .

Unit 2 Cycle 15 CORE OPERATING LIMITS REPORT

. (COLR)

. TENNESSEE VALLEY AUTHORITY

. Nuclear Fuel Division BWR, Fuel' Engine~ring Department .

Prepared By: Date:Z- A-o'}

Earl E. Riley, '. . Engineering Speci~list BWRFuel Engineering VerifiedB¥:" _. ....E,.;;~--=~::..,.:~.*-*.~.

_"_".A-;-'~'_~ _-_'"-'-.._ _

.' _.----'-'-_ _.-,-._._ _

Date:*?/11 #,1*

Brye.C. Mitchell, Nuclear Engineer.

. BWR Fuel Engineering' Approved By: Date:

. Man~ge.r ngineering Reviewed By: .* ' ~/ *. #v J. M. eck~ S';;;ervisor .

Date:. ;.. - Q-.J- 07

.' ..' Browns Ferry Reactor Engineering

'. Approved ay:

. ~ .. Date: ;by~2.

Revision 0(2/21/2007) Pages Affected: All

TV A Nuclear Fuel TV A-COLR-BF2ClS Core Operating Limits Report Revision 0, Page 2 Revision Log Revision Date Description Affected Pages o 212112007 Initial Release for New Cycle All Browns Ferry Nuclear Plant Unit 2 Cycle 15

TVA Nuclear Fuel TVA-COLR-BF2C15 Core Operating Limits Report Revision 0, Page 3

1. INTRODUCTION This Core Operating Limits Report (COLR) for Browns Ferry Nuclear Plant Unit 2 Cycle 15 is prepared in accordance with the requirements of Browns Ferry Technical Specification 5.6.5. This revision of the COLR supports operation at the current licensed thermal power (CLTP) of3458 MWt which is 105% of original licensed thermal power (OLTP).

The core operating limits presented here were developed using NRC-approved methods (References 2 and 3). One exception to this is an issue with the assumed uncertainty for the GEXL 14 CPR correlation. The NRC has identified that the correlation lacked top-peaked axial power shape data in its formulation and in the calculation of the overall correlation uncertainty. As an interim action, an increased GEXL 14 uncertainty that incorporates a significant penalty has been calculated and applied to the MCPR Safety Limit (SLMCPR) for this cycle.

Results from the reload analyses for Browns Ferry Nuclear Plant Unit 2 Cycle 15 are documented in Reference 1.

The following core operating and Technical Specification limits are included in this report:

a. Average Planar Linear Heat Generation Rate (APLHGR) Limit (Technical Specifications 3.2.1 and 3.7.5)

b. Linear Heat Generation Rate (LHGR) Limit (Technical Specification 3.2.3,3.3.4.1, and 3.7.5)

c. Minimum Critical Power Ratio Operating Limit (OLMCPR)

(Technical Specifications 3.2.2, 3.3.4.l, and 3.7.5)

d. Average Power Range Monitor (APRM) Flow Biased Rod Block Trip Setting (Technical Requirements Manual Section 5.3.1 and Table 3.3.4-1)

e. Rod Block Monitor (RBM) Trip Setpoints and Operability (Technical Specification Table 3.3 .2.1-1)

f. Shutdown Margin (SDM) Limit (Technical Specification 3.1.1)

The Unit 2 Cycle 15 core is composed of AREVA-NP ATRIUMTK I0 and Global Nuclear Fuel GE-14' assemblies. Throughout this document these are referred to as AI0 and GE14 with the trademark implied.

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TV A Nuclear Fuel TV A-COLR-BF2C15 Core Operating Limits Report Revision 0, Page 4

2. APLHGR LIMIT (TECHNICAL SPECIFICATIONS 3.2.1 AND 3.7.5)

The APLHGR limit is determined by adjusting the rated power APLHGR limit for off-rated power, off-rated flow, and SLO conditions. The most limiting of these is then used as follows:

APLHGR limit = MIN ( APLHGRp , APLHGRF , APLHGRs LO )

where: APLHGRp off-rated power APLHGR limit [APLHGRRATED

MAPFAC(P)]

APLHGRF off-rated flow APLHGR limit [APLHGRRATED

MAPFAC(F)]

APLHGRs LO SLO APLHGR limit [ALPHGR RATED

SLO_Multiplier]

The off-rated power and flow corrections to the APLHGR limit only apply to the GE14 fuel in the Browns Ferry Unit 2 Cycle 15 core. For that reason, this multiplier is set to 1.0 as shown below for the A10 fuel.

Rated Power and Flow Limits: APLHGRRATED The APLHGR limits for full power and flow conditions for each type of fuel as a function of exposure are shown in Figures 1-5. The APLHGR limits provided in the COLR figures for the GE14 assemblies are for the most limiting lattice (excluding natural uranium) at each exposure point. The specific values for each GE14 lattice are given in Reference 4. The ATRIUM-10 values are provided in Reference 1.

Bundle Type Rated Power APLHGR Limit GE14-P10DNAB416-16GZ (EDB2600) Figure 1 GE14-P10DNAB416-16GZ (EDB2601) Figure 2 GE14-P10DNAB416-18GZ (EDB2627) Figure 3 GE14-P10DNAB417-18GZ (EDB2628) Figure 4 A10-3920B-14GV70 Figure 5 A10-4227B-15GV80-FBB Figure 5 A10-4239B-15GV80-FBB Figure 5 Al 0-3552B-1 OGV80-FBB Figure 5 Browns Ferry Nuclear Plant Unit 2 Cycle 15

TV A Nuclear Fuel TV A-COLR-BF2C15 Core Operating Limits Report Revision 0, Page 5 Off-Rated Power Corrections: APLHGRp The APLHGR limits for the GE14 fuel lattices are adjusted for off-rated power conditions using the ARTS multiplier, MAPFAC(P). The reduced power multiplier, MAPFAC(P) , for the GE14 fuel is provided in Reference 1. No off-rated power correction is required for the AlO rated APLHGR limits.

Product Line MAPFAC(p)

GE14 Figure 6 AlO 1.0 Off-Rated Flow Corrections: APLHGRF The APLHGR limits for the GE14 fuel lattices are adjusted for off-rated flow conditions using the ARTS multiplier, MAPFAC(F). The reduced flow multiplier, MAPFAC(F) is provided in Reference 1. No off-rated flow correction is required for the AlO rated APLHGR limits.

Product Line MAPFAC(F)

GE14 Figure 7 AlO 1.0 SLO Corrections: APLHGRs LO Single Recirculation Loop Operation (SLO) requires that the rated power APLHGR limit (APLHGRrated) be reduced by applying the following multipliers. The GE14 multiplier is provided in Reference 5. The AlO multiplier is provided in Reference 1.

Product Line SLO Multiplier GE14 0.90

AlO 0.85

The GE14 SLO multiplier of 0.90 is the more limiting ofCLTP and EPU values provided in Reference 5. This value bounds operation at CLTP.

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TV A Nuclear Fuel TV A-COLR-BF2ClS Core Operating Limits Report Revision 0, Page 6 Equipment Out-Or-Service mOOS) Corrections:

The rated APLHGR limits in Figures 1-5 are applicable for operation with all equipment In-Service as well as the following Equipment Out-Of-Service (EOOS) options. This includes combinations of these EOOS options.

In-Service All equipment In-Service (includes 1 SRVOOS)

RPTOOS EOC-Recirculation Pump Trip Out-Of-Service TBVOOS Turbine Bypass Valve(s) Out-Of-Service PLUOOS Power Load Unbalance Out-Of-Service FHOOS (or FFTR) Feedwater Heaters Out-Of-Service (or Final Feedwater Temperature Reduction)

Single Recirculation Loop Operation (SLO) requires the application of the SLO multipliers to the rated APLHGR limits as described previously.

The off-rated power corrections [MAPFAC(P)] in Figure 6 is dependent upon the operating status of the Turbine Bypass Valve (TBV) system. For this reason, separate limits are supplied in these figures to be applied for TBVIS (in service) or TBVOOS (out of service) operation. The MAPF AC(P) limits have no dependency on RPTOOS, SLO, FHOOS/FFTR, orPLUOOS.

The off-rated flow corrections [MAPFAC(F)] in Figure 7 bound both equipment In-Service or EOOS operation.

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TVA Nuclear Fuel TVA-COLR-BF2C15 Core Operating Limits Report Revision 0, Page 7

3. LHGR LIMIT (TECHNICAL SPECIFICATION 3.2.3, 3.3.4.1, and 3.7.5)

The LHGR limit is determined by adjusting the rated power LHGR limit for off-rated power and off-rated flow conditions. The most limiting ofthese is then used, as follows:

LHGR limit = MIN ( LHGRp , LHGRF )

where: LHGRp off-rated power LHGR limit [LHGRRATED

LHGRFAC(P)]

LHGRF off-rated flow LHGR limit [LHGRRATED

LHGRFAC(F)]

The off-rated power and flow corrections to the LHGR limit only apply to the Ala fuel in the Browns Ferry Unit 2 Cycle 15 core. For that reason, these multipliers for the GE14 fuel is set to 1.0, as shown below.

Rated Power and Flow Limits: LHGRRATED The LHGR limit is fuel type dependent. The limits for these types are given below:

Fuel Type LHGRLimit GE14 Figure 8 AlO Figure 9 The Ala LHGR limit is provided in References 1 and 6. The GE14 LHGR limit is provided in References 1 and 7.

Off-Rated Power Corrections: LHGRp The LHGR limits for the Ala fuel are adjusted for off-rated power conditions using the LHGRF AC(P) multiplier which is provided in Reference 1. The LHGRF AC(P) multiplier is dependent on whether the Turbine Bypass system is in-service (TBVIS) or out-of-service (TBVOOS). No off-rated power correction is required for the GE14 rated LHGR limits.

Product Line LHGRFAC(P)

GE14 1.0 Ala Figure 10 Browns Ferry Nuclear Plant Unit 2 Cycle 15

TV A Nuclear Fuel TVA-COLR-BF2C15 Core Operating Limits Report Revision 0, Page 8

(

Off-Rated Flow Corrections: LHGRF The LHGR limits for the A10 fuel are adjusted for off-rated flow conditions using the LHGRF AC(F) multiplier which is provided in Reference 1. No off-rated flow correction is required for the GE14 rated LHGR limits.

Product Line LHGRFAC(F)

GE14 1.0 A10 Figure 11 Equipment Out-Or-Service (EOOS) Corrections:

The rated LHGR limits are applicable for operation with all equipment In-Service as well as the following Equipment Out-Of-Service (EOOS) options. This includes combinations of these EOOS options.

In-Service All equipment In-Service (includes 1 SRVOOS)

RPTOOS EOC-Recirculation Pump Trip Out-Of-Service TBVOOS Turbine Bypass Valve(s) Out-Of-Service PLUOOS Power Load Unbalance Out-Of-Service SLO Single Recirculation Loop Operation FHOOS (or FFTR) Feedwater Heaters Out-Of-Service (or Final Feedwater Temperature Reduction)

The off-rated power corrections [LHGRFAC(P)] in Figure 10 are dependent upon operation of the Turbine Bypass Valve system. For this reason, separate limits are supplied in this figure to be applied for TBVIS or TBVOOS operation. The LHGRF AC(P) limits have no dependency on RPTOOS, PLUOOS, SLO, or FHOOSIFFTR.

The off-rated flow corrections [LHGRFAC(F)] in Figure 11 bound both equipment In-Service or EOOS operation.

(

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TVA Nuclear Fuel TV A-COLR-BF2C15 Core Operating Limits Report Revision 0, Page 9

4. OLMCPR (TECHNICAL SPECIFICATIONS 3.2.2, 3.3.4.1, AND 3.7.5)

The MCPR Operating Limit (OLMCPR) is calculated to be the most limiting of the flow-dependent MCPR (MCPRF) and power-dependent MCPR (MCPRp).

OLMCPR limit = MAX ( MCPRF , MCPRp )

where: MCPRF core flow-dependent MCPR limit MCPRp power-dependent MCPR limit MCPRF limits are provided in Figure 12. MCPRp limits are provided in Tables 1 through 6.

Flow-Dependent MCPR Limits: MCPRF The MCPRF limits are dependent upon:

Core Flow (% of Rated)

Max Core Flow Limit (Rated or Increased Core Flow, ICF)

Fuel Type (GE14 or A10)

The MCPRF limits are provided in Figure 12. For Unit 2 Cycle 15 the same MCPRF limits apply to both the GE14 and A10 fuel types. These limits are valid for all EOOS combinations. No adjustment is required to the MCPRF limits for SLO. The MCPRF limits are found in Reference 1.

Power-Dependent MCPR Limits: MCPRp The MCPRp limits are dependent upon:

Core Power Level (% of Rated)

Technical Specification Scram Speed (TSSS) or Nominal Scram Speed (NSS)

Fuel Type (GE14 or A10)

Cycle Operating Exposure (NEOC, EOC, and CD - as defined in this section)

Equipment Out-Of-Service Options

Two or Single recirculation Loop Operation (TLO vs. SLO)

The MCPRp limits (Ref. 1) are provided in the following tables, where each table contains the limits for all fuel types and EOOS options (for a specified scram speed and exposure range). The MCPRp limits are determined from these tables using linear interpolation between the specified powers.

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TV A Nuclear Fuel TV A-COLR-BF2ClS Core Operating Limits Report Revision 0, Page 10 Exposure Range Scram Speed MCPRp BOCtoNEOC NSS Table 1 TSSS Table 2 BOCto EOC NSS Table 3 TSSS Table 4 BOCto CD NSS Table 5 TSSS Table 6

a. Scram Speed Dependent Limits (TSSS vs. NSS)

MCPRp limits are provided for two different sets of assumed scram speeds. The Technical Specification Scram Speed (TSSS) MCPRp limits are applicable at all times as long as the scram time surveillance demonstrates that the times in Technical Specification table 3.1.4-1 have been met. Nominal Scram Speeds (NSS) may be used as long as the scram time surveillance demonstrates that the times in the following table are met (Ref. 9).

(

Notch Nominal Scram Speed Position (seconds) 46 0.42 36 0.98 26 1.60 06 2.90 In demonstrating compliance with this table, the same surveillance requirements from Technical Specification 3.1.4 apply, except that the definition of SLOW rods should conform to the scram speeds in the table above. If conformance to this table is not demonstrated, TSSS MCPRp limits shall be used.

On initial cycle startup, TSSS limits are used until the successful completion of scram timing confirms that NSS limits may be used.

b. Fuel Type Dependent Limits Separate MCPRp limits are provided for the GE14 and AlO fuel types.

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TVA Nuclear Fuel TVA-COLR-BF2CI5 Core Operating Limits Report Revision 0, Page 11

c. Exposure Dependent Limits Exposures are tracked on a Core Average Exposure basis (not Cycle Exposure). The higher exposure MCPRp limits are always more limiting and may be used for any Core Average Exposure up to the ending exposure.

MCPRp limits are provided for the following exposure ranges (Ref. 1):

BOCtoNEOC NEOC corresponds to 27,788 MWd / MTU BOCto EOC EOC corresponds to 31,075 MWd / MTU BOCtoCD CD corresponds to 32,274 MWd / MTU NEOC refers to a Near EOC exposure point.

The EOC exposure point is not the true End-Of-Cycle exposure. Instead it corresponds to a licensing exposure window that exceeds expected end-of-full-power-life.

The CD (CoastDown) exposure point represents a licensing exposure point that exceeds the expected end-of-cycle exposure including cycle extension options.

d. Equipment Out-Of-Service (EOOS) Options EOOS options included in the MCPRp limits are:

In-Service All equipment In-Service (includes 1 SRVOOS)

RPTOOS EOC-Recirculation Pump Trip Out-Of-Service TBVOOS Turbine Bypass Valve(s) Out-Of-Service RPTOOS+TBVOOS Combined RPTOOS and TBVOOS PLUOOS Power Load Unbalance Out-Of-Service PLUOOS+RPTOOS Combined PLUOOS and RPTOOS PLUOOS+TBVOOS Combined PLUOOS and TBVOOS PLUOOS+TBVOOS+RPTOOS Combined PLUOOS, RPTOOS, and TBVOOS FHOOS (or FFTR) Feedwater Heaters Out-Of-Service (or Final Feedwater Temperature Reduction)

For exposure ranges up to NEOC and EOC, additional combinations of MCPRp limits are also provided that include FHOOS. The CD exposure range assumes application of FFTR, so the CD based MCPRp limits already include FHOOS.

e. Single-Loop-Operation (SLO) Limits The MCPRplimits for SLO are to be increased by 0.02 (Ref. 1).

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TV A Nuclear Fuel TVA-COLR-BF2C15 Core Operating Limits Report Revision 0, Page 12

f. Below Pbypass Limits Below Pbypass (30% rated power), the MCPRp limits are dependent upon core flow.

One set of MCPRp limits applies if the core flow is above 50% of rated with a second set that applies if the core flow is less than or equal to 50% rated.

(

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TVA Nuclear Fuel TVA-COLR-BF2C15 Core Operating Limits Report Revision 0, Page 13

5. APRM FLOW BIASED ROD BLOCK TRIP SETTING (TECHNICAL REQUIREMENTS MANUAL SECTION 5.3.1 AND TABLE 3.3.4-1)

The APRM Rod Block trip setting shall be (Ref. 10):

SRB ::; (0.66(W-~W) + 61%) Allowable Value SRB ::; (0.66(W-~ W) + 59%) Nominal Trip Setpoint (NTSP) where:

SRB = Rod Block setting in percent of rated thermal power (3458 MWt)

W= Loop recirculation flow rate in percent of rated

~W = Difference between two-loop and single-loop effective recirculation flow at the same core flow (~W=O.O for two-loop operation)

The APRM Rod Block trip setting is clamped at a maximum allowable value of 115%

(corresponding to a NTSP of 113%).

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TV A Nuclear Fuel TVA-COLR-BF2C1S Core Operating Limits Report Revision 0, Page 14

(

6. ROD BLOCK MONITOR (RBM) TRIP SETPOINTS AND OPERABILITY (TECHNICAL SPECIFICATION TABLE 3.3.2.1-1)

The RBM trip setpoints and applicable power ranges shall be as follows (refs. 10 & 11):

Allowable Value Nominal Trip Setpoint RBM Trip Setpoint CAY) (NTSP)

LPSP 27% 25%

IPSP 62% 60%

HPSP 82% 80%

L TSP - unfiltered 124.7% 123.0%

- filtered 123.5% 121.8% (1 ),(2)

ITSP - unfiltered 119.7% 118.0%

- filtered 118.7% 117.0% (1),(2)

HTSP - unfiltered 114.7% 113.0%

- filtered 113.7% 112.0% (1),(2)

DTSP 90% 92%

Notes: (1) These setpoints are based upon an Analytical Limit HTSP of 117% (w/o filter) which corresponds to a MCPR operating limit of 1.42(A10/GE14), as reported in section 5.5 of Reference 1. Unit 2 Cycle 15 has had a cycle specific CRWE analysis performed and the table provided in section 5.5 of Reference 1 supercedes the OLMCPR values of references 10 and 12.

(2) The unfiltered setpoints are consistent with a nominal RBM filter setting of 0.0 seconds (reference 1O.b)). The filtered setpoints are consistent with a nominal RBM filter setting:s 0.5 seconds (reference 10.a)).

The RBM setpoints in Technical Specification Table 3.3.2.1-1 are applicable when:

THERMAL POWER Applicable Notes from

(% Rated) MCPR (1) Table 3.3.2.1-1

27% and < 90% <1.72 (a), (b), (f), (h) dual loop operation

< 1.75 (a), (b), (f), (h) single loop operation

90% < 1.47 (g) dual loop operation (2)

Notes: (1) The MCPR values shown correspond to a SLMCPR of 1.08 for dual recirculation loop operation and 1.10 for single loop operation. (Ref. 1).

(2) Greater than 90% rated power is not attainable in single loop operation.

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TV A Nuclear Fuel TV A-COLR-BF2CI5 Core Operating Limits Report Revision 0, Page 15

7. SHUTDOWN MARGIN (SDM) LIMIT (TECHNICAL SPECIFICATION 3.1.1)

The core shall be subcritical with the following margin with the strongest OPERABLE control rod fully withdrawn and all other OPERABLE control rods fully inserted (Ref. 8).

SDM ~ 0.38% dk/k Browns Ferry Nuclear Plant Unit 2 Cycle 15

TV A Nuclear Fuel TV A-COLR-BF2C1S Core Operating Limits Report Revision 0, Page 16

8. REFERENCES

1. ANP-2592 Rev. 0, "Browns Ferry Unit 2 Cycle 15 Reload Analysis for 105% Original Licensed Thermal Power", dated January 2007.

2. Framatome-ANP Analytical Methodology

References:

a) XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984.

b) XN-NF-85-67(P)(A) Revision 1, Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, Exxon Nuclear Company, September 1986.

c) EMF-85-74(P) Revision 0 Supplement I(P)(A) and Supplement 2(P)(A), RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model, Siemens Power Corporation, February 1998.

d) ANF-89-98(P)(A) Revision 1 and Supplement 1, Generic Mechanical Design Criteria for BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.

e) XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, Exxon Nuclear Company, March 1983.

f) XN-NF-80-19(P)(A) Volume 4 Revision 1, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads, Exxon Nuclear Company, June 1986.

g) EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4IMICROBURN-B2, Siemens Power Corporation, October 1999.

h) XN-NF-80-19(P)(A) Volume 3 Revision 2, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company, January 1987.

i) XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, Exxon Nuclear Company, February 1987.

j) ANF-524(P)(A) Revision 2 and Supplements 1 and 2, ANF Critical Power Methodology for Boiling Water Reactors, Advanced Nuclear Fuels Corporation, November 1990.

k) ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3 and 4, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses, Advanced Nuclear Fuels Corporation, August 1990.

1) ANF-1358(P)(A) Revision 1, The Loss of Feedwater Heating Transient in Boiling Water Reactors, Advanced Nuclear Fuels Corporation, September 1992.

m) EMF-2209(P)(A) Revision 2, SPCB Critical Power Correlation, Siemens Power Corporation, September 2003.

n) EMF-2245(P)(A) Revision 0, Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel, Siemens Power Corporation, August 2000.

0) EMF-2361(P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP, May 2001.

p) EMF-2292(P)(A) Revision 0, ATRIUMTK10: Appendix K Spray Heat Transfer Coefficients, Siemens Power Corporation, September 2000.

3. Global Nuclear Fuel Analytical Methodology

References:

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TVA Nuclear Fuel TVA-COLR-BF2CI5 Core Operating Limits Report Revision 0, Page 17 a) NEDE-24011-P-A-I5, "General Electric Standard Application for Reactor Fuel",

September 2005.

b) NEDE-24011-P-A-I5-US, "General Electric Standard Application for Reactor Fuel (Supplement for United States)", September 2005.

4. 0000-0006-1355-MAPL Rev. 0, "Lattice-Dependent MAPLHGR Report for Browns Ferry Unit 2 Reload 12 Cycle 13", February 2003.

5. NEDC-32484P Rev. 6, "Browns Ferry Nuclear Plant Units 1, 2, and 3 - SAFERIGESTR-LOCA Loss-Of-Coolant Accident Analysis", dated February 2005.

6. ANP-2537P Rev. 0, "Mechanical Design Report for Browns Ferry Unit 2 Reload BFE2-I5 ATRlUMTM-IO Fuel Assemblies", dated May 2006.

7. GE-NE-LI2-00889-00-0IP Rev. 0, "GEI4 Fuel Design Cycle-Independent Analyses for Browns Ferry Units 2 and 3", dated January 2002.

8. TVA-COLR-BF2CI4 Rev. 1, "Browns Ferry Nuclear Plant Unit 2, Cycle 14 Core Operating Limits Report (COLR)", dated April 10, 2006.

9. EMF-3238(P) Rev. 0, "Browns Ferry Unit 2 Cycle 15 Plant Parameters Document", dated January 2006.

10. PRNM Setpoint Calculation:

a) Filtered Setpoints - EDE-28-0990 Rev. 3 Supplement E, "PRNM (APRM, RBM, and RPM) Setpoint Calculations [ARTSIMELLL (NUMAC) - Power-Uprate Condition] for Tennessee Valley Authority Browns Ferry Nuclear Plant", dated October 1997.

b) Unfiltered Setpoints - EDE-28-0990 Rev. 2 Supplement E, "PRNM (APRM, RBM, and RPM) Setpoint Calculations [ARTSIMELLL (NUMAC) - Power-Uprate Condition] for Tennessee Valley Authority Browns Ferry Nuclear Plant", dated October 1997.

11. GE Letter LB#: 262-97-133, "Browns Ferry Nuclear Plant Rod Block Monitor Setpoint Clarification - GE Proprietary Information", dated September 12, 1997.

12. NEDC-32433P, "Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Browns Ferry Nuclear Plant Unit 1,2, and 3", dated April 1995.

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TV A Nuclear Fuel TVA-COLR-BF2C15 Core Operating Limits Report Revision 0, Page 18 Table 1: MCPRp Limits for BOC to NEOC Exposures - NSS Scram Times (A pp rleabi e up to C ore A verae Exposure of27,788 MWd / MTU)

MCPRp Limit MCPRpLimit EOOS Power EOOS Power AIO GE14 AIO GE14 Option (% Rated) Option (% Rated) 100 1.45 1.46 100 1.48 1.50 69 1.58 1.62 69 1.61 1.64 63 1.61 1.64 63 1.66 1.70 58 1.69 1.73 58 1.75 ---

58 1.78 1.78 58 1.78 1.79 In-46 1.85 1.89 FHOOS 46 1.90 1.97 Service 30 2.22 2.35 30 2.33 2.49 30 (> 50%F) 2.64 2.79 30 (> 50%F) 2.75 2.92 25 (> 50%F) 2.89 3.08 25 (>50%F) 3.03 3.24 30 (S 50%F) 2.51 2.68 30 (S 50%F) 2.60 2.79 25 (S 50%F) 2.68 2.89 25 (S 50%F) 2.81 3.03 100 1.45 1.46 100 1.48 1.50 69 1.58 1.62 69 1.61 1.64 63 1.61 1.64 63 1.66 1.70 58 1.69 1.73 58 1.75 ---

58 1.78 1.78 58 1.78 1.79 RPTOOS RPTOOS 46 1.85 1.89 46 1.90 1.97 FHOOS 30 2.22 2.35 30 2.33 2.49 30 (> 50%F) 2.64 2.79 30 (> 50%F) 2.75 2.92 (

25 (> 50%F) 2.89 3.08 25 (>50%F) 3.03 3.24 30 (S 50%F) 2.51 2.68 30 (S 50%F) 2.60 2.79 25 (S 50%F) 2.68 2.89 25 (S 50%F) 2.81 3.03 100 1.49 1.50 100 1.51 1.53 69 1.62 1.66 69 1.64 1.66 63 1.65 1.67 63 1.68 1.72 58 1.71 1.75 58 1.77 ---

58 1.78 1.78 58 1.78 1.81 TBVOOS TBVOOS 46 1.85 1.90 46 1.92 1.98 FHOOS 30 2.23 2.36 30 2.35 2.49 30 (> 50%F) 3.09 3.20 30 (> SO%F) 3.19 3.31 25 (> 50%F) 3.51 3.64 25 (>50%F) 3.62 3.78 30 (S 50%F) 2.64 2.79 30 (S SO%F) 2.72 2.88 25 (S 50%F) 2.97 3.18 25 (SSO%F) 3.07 3.30 100 1.49 1.50 100 1.51 1.53 69 1.62 1.66 69 1.64 1.66 63 1.65 1.67 63 1.68 1.72 58 1.71 1.75 58 1.77 ---

58 1.78 1.78 RPTOOS 58 1.78 1.81 RPTOOS 46 1.85 1.90 TBVOOS 46 1.92 1.98 TBVOOS 30 2.23 2.36 FHOOS 30 2.35 2.49 30 (> 50%F) 3.09 3.20 30 (> SO%F) 3.19 3.31 25 (> 50%F) 3.51 3.64 25 (> SO%F) 3.62 3.78 30 (S 50%F) 2.64 2.79 30 (S SO%F) 2.72 2.88 25 (S 50%F) 2.97 3.18 25 (S SO%F) 3.07 3.30 Add 0.02 to the above MCPRp limits for SLO.

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TV A Nuclear Fuel TVA-COLR-BF2C15 Core Operating Limits Report Revision 0, Page 19 Table 1 (Continued): MCPRp Limits for BOC to NEOC Exposures - NSS Scram Times (App rleabieup t0 eore Avera e Exposure of27,788 MWd / MTU)

MCPRpLimit MCPRpLimit EOOS Power EOOS Power AIO GE14 AIO GE14 Option (%Rated) Option (%Rated) 100 1.45 1.46 100 1.48 1.50 69 1.58 1.62 69 1.61 1.64 63 1.73 1.73 63 1.73 1.73 58 --- --- 58 --- ---

58 1.78 1.78 58 1.78 1.79 FROOS PLUOOS 46 1.85 1.89 46 1.90 1.97 PLUOOS 30 2.22 2.35 30 2.33 2.49 30 (> 50%F) 2.64 2.79 30 (>50%F) 2.75 2.92 25 (> 50%F) 2.89 3.08 25 (>50%F) 3.03 3.24 30 (~50%F) 2.51 2.68 30 (~50%F) 2.60 2.79 25 (~50%F) 2.68 2.89 25 (~50%F) 2.81 3.03 100 1.45 1.46 100 1.48 1.50 69 1.58 1.62 69 1.61 1.64 63 1.73 1.73 63 1.73 1.73 58 --- --- 58 --- ---

58 1.78 1.78 RPTOOS 58 1.78 1.79 RPTOOS 46 1.85 1.89 FROOS 46 1.90 1.97 PLUOOS 30 2.22 2.35 PLUOOS 30 2.33 2.49 30 (> 50%F) 2.64 2.79 30 (>50%F) 2.75 2.92 25 (>50%F) 2.89 3.08 25 (> 50%F) 3.03 3.24 30 (~50%F) 2.51 2.68 30 (~50%F) 2.60 2.79 25 (~50%F) 2.68 2.89 25 (~50%F) 2.81 3.03 100 1.49 1.50 100 1.51 1.53 69 1.62 1.66 69 l.64 l.66 63 1.73 1.73 63 1.73 l.73 58 --- --- 58 --- ---

58 1.78 1.78 TBVOOS 58 1.78 l.81 TBVOOS 46 1.85 1.90 FROOS 46 1.92 1.98 PLUOOS 30 2.23 2.36 PLUOOS 30 2.35 2.49 30 (>50%F) 3.09 3.20 30 (> 50%F) 3.19 3.31 25 (>50%F) 3.51 3.64 25 (> 50%F) 3.62 3.78 30 (~50%F) 2.64 2.79 30 (~50%F) 2.72 2.88 25 (~50%F) 2.97 3.l8 25 (~50%F) 3.07 3.30 100 1.49 1.50 100 1.51 1.53 69 1.62 1.66 69 1.64 1.66 63 1.73 1.73 63 l.73 l.73 58 --- --- RPTOOS 58 --- ---

RPTOOS 58 1.78 1.78 58 l.78 l.81 TBVOOS TBVOOS 46 1.85 1.90 46 1.92 l.98 FROOS PLUOOS 30 2.23 2.36 30 2.35 2.49 PLUOOS 30 (>50%F) 3.09 3.20 30 (>50%F) 3.19 3.31 25 (>50%F) 3.51 3.64 25 (>50%F) 3.62 3.78 30 (~50%F) 2.64 2.79 30 (~50%F) 2.72 2.88 25 (~50%F) 2.97 3.l8 25 (~50%F) 3.07 3.30 Add 0.02 to the above MCPRp limits for SLO.

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TV A Nuclear Fuel TVA-COLR-BF2CI5 Core Operating Limits Report Revision 0, Page 20 Table 2: MCPRp Limits for BOC to NEOC Exposures - TSSS Scram Times (App rleabi e up to C ore A verae Exposure of27,788 MWd / MTU)

MCPRpLimit MCPRpLimit EOOS Power EOOS Power AIO GE14 AIO GE14 Option (% Rated) Option (% Rated) 100 1.47 1.49 100 1.50 1.52 69 1.59 1.65 69 1.63 1.66 63 1.63 1.66 63 1.68 1.72 58 1.71 1.75 58 1.77 ---

58 1.79 1.79 58 1.79 1.81 In-46 1.86 1.91 FHOOS 46 1.92 1.99 Service 30 2.23 2.37 30 2.35 2.51 30 (> SO%F) 2.64 2.79 30 (> SO%F) 2.75 2.92 25 (> SO%F) 2.89 3.08 25 (> SO%F) 3.03 3.24 30 (:::: SO%F) 2.51 2.68 30 (::::SO%F) 2.60 2.79 25 (:::: SO%F) 2.68 2.89 25 (::::SO%F) 2.81 3.03 100 1.47 1.49 100 1.50 1.52 69 1.59 1.65 69 1.63 1.66 63 1.63 1.66 63 1.68 1.72 58 1.71 1.75 58 1.77 ---

58 1.79 1.79 58 1.79 1.81 RPTOOS RPTOOS 46 1.86 1.91 46 1.92 1.99 FHOOS 30 2.23 2.37 30 2.35 2.51 30 (> SO%F) 2.64 2.79 30 (> SO%F) 2.75 2.92 25 (> SO%F) 2.89 3.08 25 (> SO%F) 3.03 3.24 30 (:::: SO%F) 2.51 2.68 30 (:::: SO%F) 2.60 2.79 25 (:::: SO%F) 2.68 2.89 25 (:::: SO%F) 2.81 3.03 100 1.51 1.52 100 1.53 1.55 69 1.64 1.65 69 1.66 1.68 63 1.66 1.69 63 1.70 1.74 58 1.73 1.78 58 --- ---

58 1.79 1.79 58 1.79 1.83 TBVOOS TBVOOS 46 1.87 1.93 46 1.94 2.00 FHOOS 30 2.25 2.39 30 2.37 2.52 30 (> SO%F) 3.10 3.20 30 (> SO%F) 3.20 3.31 25 (> SO%F) 3.52 3.64 25 (>SO%F) 3.63 3.78 30 (:::: SO%F) 2.65 2.79 30 (::::SO%F) 2.72 2.88 25 (:::: SO%F) 2.97 3.19 25 (::::SO%F) 3.08 3.30 100 1.51 1.52 100 1.53 1.55 69 1.64 1.65 69 1.66 1.68 63 1.66 1.69 63 1.70 1.74 58 1.73 1.78 58 --- ---

58 1.79 1.79 RPTOOS 58 1.79 1.83 RPTOOS 46 1.87 1.93 TBVOOS 46 1.94 2.00 TBVOOS 30 2.25 2.39 FHOOS 30 2.37 2.52 30 (> SO%F) 3.10 3.20 30 (> SO%F) 3.20 3.31 25 (> SO%F) 3.52 3.64 25 (> SO%F) 3.63 3.78 30 (:::: SO%F) 2.65 2.79 30 (:::: SO%F) 2.72 2.88 25 (:::: SO%F) 2.97 3.19 25 (:::: SO%F) 3.08 3.30 Add 0.02 to the above MCPRp limits for SLO.

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TVA Nuclear Fuel TVA-COLR-BF2CI5 Core Operating Limits Report Revision 0, Page 21 Table 2 (Continued): MCPRp Limits for BOC to NEOC Exposures - TSSS Scram Times Avera e Exposure of27,788 MWd / MTU)

(App l'leableup t0 eore MCPRpLimit MCPRpLimit EOOS Power EOOS Power AIO GE14 AIO GE14 Option (% Rated) Option (% Rated) 100 1.47 1.49 100 1.50 1.52 69 1.59 1.65 69 1.63 1.66 63 1.74 1.74 63 1.74 1.74 58 --- --- 58 --- ---

58 1.79 1.79 58 1.79 1.81 FHOOS PLUOOS 46 1.86 1.91 46 1.92 1.99 PLUOOS 30 2.23 2.37 30 2.35 2.51 30 (> 50%F) 2.64 2.79 30 (>50%F) 2.75 2.92 25 (> 50%F) 2.89 3.08 25 (>50%F) 3.03 3.24 30 (:S:50%F) 2.51 2.68 30 (:S:50%F) 2.60 2.79 25 (:S:50%F) 2.68 2.89 25 (:S:50%F) 2.81 3.03 100 1.47 1.49 100 1.50 1.52 69 1.59 1.65 69 1.63 1.66 63 1.74 1.74 63 1.74 1.74 58 --- --- 58 --- ---

58 1.79 1.79 RPTOOS 58 1.79 1.81 RPTOOS 46 1.86 1.91 FHOOS 46 1.92 1.99 PLUOOS 30 2.23 2.37 PLUOOS 30 2.35 2.51 30 (>50%F) 2.64 2.79 30 (>50%F) 2.75 2.92 25 (>50%F) 2.89 3.08 25 (>50%F) 3.03 3.24 30 (:s:50%F) 2.51 2.68 30 (:S:50%F) 2.60 2.79 25 (:S:50%F) 2.68 2.89 25 (:S:50%F) 2.81 3.03 100 1.51 1.52 100 1.53 1.55 69 1.64 1.65 69 1.66 1.68 63 1.74 1.74 63 1.74 1.74 58 --- --- 58 --- ---

58 1.79 1.79 TBVOOS 58 1.79 1.83 TBVOOS 46 1.87 1.93 FHOOS 46 1.94 2.00 PLUOOS 30 2.25 2.39 PLUOOS 30 2.37 2.52 30 (>50%F) 3.10 3.20 30 (>50%F) 3.20 3.31 25 (> 50%F) 3.52 3.64 25 (>50%F) 3.63 3.78 30 (:S:50%F) 2.65 2.79 30 (:s:50%F) 2.72 2.88 25 (:S:50%F) 2.97 3.19 25 (:S:50%F) 3.08 3.30 100 1.51 1.52 100 1.53 1.55 69 1.64 1.65 69 1.66 1.68 63 1.74 1.74 63 1.74 1.74 58 --- --- RPTOOS 58 --- ---

RPTOOS 58 1.79 1.79 58 1.79 1.83 TBVOOS TBVOOS 46 1.87 1.93 46 1.94 2.00 FHOOS PLUOOS 30 2.25 2.39 30 2.37 2.52 PLUOOS 30 (>50%F) 3.10 3.20 30 (>50%F) 3.20 3.31 25 (>50%F) 3.52 3.64 25 (>50%F) 3.63 3.78 30 (:S:50%F) 2.65 2.79 30 (:S:50%F) 2.72 2.88 25 (:S:50%F) 2.97 3.19 25 (:S:50%F) 3.08 3.30 Add 0.02 to the above MCPRp limits for SLO.

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TVA Nuclear Fuel TVA-COLR-BF2C15 Core Operating Limits Report Revision 0, Page 22

(

Table 3: MCPRp Limits for BOC to EOC Exposures - NSS Scram Times (App rleahI e up to C ore A verae Exposure of31,075 MWd / MTU)

MCPRpLimit MCPRpLimit EOOS Power EOOS Power AIO GE14 AIO GE14 Option (% Rated) Option (% Rated) 100 1.45 1.47 100 1.48 1.50 69 1.58 1.62 69 1.61 1.64 63 1.61 1.64 63 1.66 1.70 58 1.69 1.73 58 1.75 ---

58 1.78 1.78 58 1.78 1.79 In-46 1.85 1.89 FHOOS 46 1.90 1.97 Service 30 2.22 2.35 30 2.33 2.49 30 (>50%F) 2.64 2.79 30 (>50%F) 2.75 2.92 25 (> 50%F) 2.89 3.08 25 (>50%F) 3.03 3.24 30 (:::50%F) 2.51 2.68 30 (:::50%F) 2.60 2.79 25 (:::50%F) 2.68 2.89 25 (:::50%F) 2.81 3.03 100 1.45 1.47 100 1.48 1.50 69 1.58 1.62 69 1.61 1.64 63 1.61 1.64 63 1.66 1.70 58 1.69 1.73 58 1.75 ---

58 1.78 1.78 58 1.78 1.79 RPTOOS RPTOOS 46 1.85 1.89 46 1.90 1.97 FHOOS 30 2.22 2.35 30 2.33 2.49 30 (>50%F) 2.64 2.79 30 (>50%F) 2.75 2.92 25 (>50%F) 2.89 3.08 25 (>50%F) 3.03 3.24 30 (:::50%F) 2.51 2.68 30 (:::50%F) 2.60 2.79 25 (:::50%F) 2.68 2.89 25 (:::50%F) 2.81 3.03 100 1.49 1.51 100 1.51 1.53 69 1.62 1.66 69 1.64 1.66 63 1.65 1.67 63 1.68 1.72 58 1.71 1.75 58 1.77 ---

58 1.78 1.78 58 1.78 1.81 TBVOOS TBVOOS 46 1.85 1.90 46 1.92 1.98 FHOOS 30 2.23 2.36 30 2.35 2.49 30 (> 50%F) 3.09 3.20 30 (>50%F) 3.19 3.31 25 (>50%F) 3.51 3.64 25 (>50%F) 3.62 3.78 30 (:::50%F) 2.64 2.79 30 (:::50%F) 2.72 2.88 25 (:::50%F) 2.97 3.18 25 (:::50%F) 3.07 3.30 100 1.49 1.51 100 1.51 1.53 69 1.62 1.66 69 1.64 1.66 63 1.65 1.67 63 1.68 1.72 58 1.71 1.75 58 1.77 ---

58 1.78 1.78 RPTOOS 58 1.78 1.81 RPTOOS 46 1.85 1.90 TBVOOS 46 1.92 1.98 TBVOOS 30 2.23 2.36 FHOOS 30 2.35 2.49 30 (>50%F) 3.09 3.20 30 (>50%F) 3.19 3.31 25 (>50%F) 3.51 3.64 25 (>50%F) 3.62 3.78 30 (:::50%F) 2.64 2.79 30 (:::50%F) 2.72 2.88 25 (:::50%F) 2.97 3.18 25 (:::50%F) 3.07 3.30 Add 0.02 to the above MCPRp limits for SLO.

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TVA Nuclear Fuel TVA-COLR-BF2CI5 Core Operating Limits Report Revision 0, Page 23 Table 3 (Continued): MCPRp Limits for BOC to EOC Exposures - NSS Scram Times (App rleahIeupto C ore A verae Exposure 0[31,075 MWd / MTU)

MCPRpLimit MCPRpLimit EOOS Power EOOS Power AIO GE14 AIO GE14 Option (% Rated) Option (% Rated) 100 1.45 1.47 100 1.48 1.50 69 1.58 1.62 69 1.61 1.64 63 1.73 1.73 63 1.73 1.73 58 --- --- 58 --- ---

58 1.78 1.78 58 1.78 1.79 FHOOS PLUOOS 46 1.85 1.89 46 1.90 1.97 PLUOOS 30 2.22 2.35 30 2.33 2.49 30 (>50%F) 2.64 2.79 30 (>50%F) 2.75 2.92 25 (>50%F) 2.89 3.08 25 (>50%F) 3.03 3.24 30 (:::50%F) 2.51 2.68 30 (:::50%F) 2.60 2.79 25 (:::50%F) 2.68 2.89 25 (:::50%F) 2.81 3.03 100 1.45 1.47 100 1.48 1.50 69 1.58 1.62 69 1.61 1.64 63 1.73 1.73 63 1.73 1.73 58 --- --- 58 --- ---

58 1.78 1.78 RPTOOS 58 1.78 1.79 RPTOOS 46 1.85 1.89 FHOOS 46 1.90 1.97 PLUOOS 30 2.22 2.35 PLUOOS 30 2.33 2.49 30 (>50%F) 2.64 2.79 30 (>50%F) 2.75 2.92 25 (>50%F) 2.89 3.08 25 (>50%F) 3.03 3.24 30 (:::50%F) 2.51 2.68 30 (:::50%F) 2.60 2.79 25 (:::50%F) 2.68 2.89 25 (:::50%F) 2.81 3.03 100 1.49 1.51 100 1.51 1.53 69 1.62 1.66 69 1.64 1.66 63 1.73 1.73 63 1.73 1.73 58 --- --- 58 --- ---

58 1.78 1.78 TBVOOS 58 1.78 1.81 TBVOOS 46 1.85 1.90 FHOOS 46 1.92 1.98 PLUOOS 30 2.23 2.36 PLUOOS 30 2.35 2.49 30 (> 50%F) 3.09 3.20 30 (>50%F) 3.19 3.31 25 (>50%F) 3.51 3.64 25 (>50%F) 3.62 3.78 30850%F) 2.64 2.79 30 (:::50%F) 2.72 2.88 25 (:::50%F) 2.97 3.18 25 (:::50%F) 3.07 3.30 100 1.49 1.51 100 1.51 1.53 69 1.62 1.66 69 1.64 1.66 63 1.73 1.73 63 1.73 1.73 58 --- --- RPTOOS 58 --- ---

RPTOOS 58 1.78 1.78 58 1.78 1.81 TBVOOS TBVOOS 46 1.85 1.90 46 1.92 1.98 FHOOS PLUOOS 30 2.23 2.36 30 2.35 2.49 PLUOOS 30 (> 50%F) 3.09 3.20 30 (> 50%F) 3.19 3.31 25 (> 50%F) 3.51 3.64 25 (> 50%F) 3.62 3.78 30 (:::50%F) 2.64 2.79 30 (:::50%F) 2.72 2.88 25 (:::50%F) 2.97 3.18 25 (:::50%F) 3.07 3.30 Add 0.02 to the above MCPRp limits for SLO.

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TV A Nuclear Fuel TVA-COLR-BF2C15 Core Operating Limits Report Revision 0, Page 24 Table 4: MCPRp Limits for BOC to EOC Exposures - TSSS Scram Times c

(App rleabie up t0 eoreAvera e Exposure 0[31,075 MWd / MTU)

MCPRpLimit MCPRpLimit EOOS Power EOOS Power AIO GE14 AIO GE14 Option (%Rated) Option (% Rated) 100 1.47 1.49 100 1.50 1.52 69 1.59 1.65 69 1.63 1.66 63 1.63 1.66 63 1.68 1.72 58 1.71 1.75 58 1.77 ---

58 1.79 1.79 58 1.79 1.81 In-46 1.86 1.91 FHOOS 46 1.92 1.99 Service 30 2.23 2.37 30 2.35 2.51 30 (> 50%F) 2.64 2.79 30 (>50%F) 2.75 2.92 25 (>50%F) 2.89 3.08 25 (>50%F) 3.03 3.24 30 (:::50%F) 2.51 2.68 30 (:::50%F) 2.60 2.79 25 (:::50%F) 2.68 2.89 25 (:::50%F) 2.81 3.03 100 1.47 1.50 100 1.50 1.52 69 1.59 1.65 69 1.63 1.66 63 1.63 1.66 63 1.68 1.72 58 1.71 1.75 58 1.77 ---

58 1.79 1.79 58 1.79 1.81 RPTOOS RPTOOS 46 1.86 1.91 46 1.92 1.99 FHOOS 30 2.23 2.37 30 2.35 2.51 30 (>50%F) 25 (>50%F) 2.64 2.89 2.79 3.08 30 (>50%F) 25 (>50%F) 2.75 3.03 2.92 3.24

(

30 (:::50%F) 2.51 2.68 30 (:::50%F) 2.60 2.79 25 (:::50%F) 2.68 2.89 25 (:::50%F) 2.81 3.03 100 1.51 1.52 100 1.53 1.55 69 1.64 1.65 69 1.66 1.68 63 1.66 1.69 63 1.70 1.74 58 1.73 1.78 58 --- ---

58 1.79 1.79 58 1.79 1.83 TBVOOS TBVOOS 46 1.87 1.93 46 1.94 2.00 FHOOS 30 2.25 2.39 30 2.37 2.52 30 (>50%F) 3.10 3.20 30 (>50%F) 3.20 3.31 25 (>50%F) 3.52 3.64 25 (>50%F) 3.63 3.78 30 (:::50%F) 2.65 2.79 30 (:::50%F) 2.72 2.88 25 (:::50%F) 2.97 3.19 25 (:::50%F) 3.08 3.30 100 1.52 1.53 100 1.53 1.55 69 1.64 1.65 69 1.66 1.68 63 1.66 1.69 63 1.70 1.74 58 1.73 1.78 58 --- ---

58 1.79 1.79 RPTOOS 58 1.79 1.83 RPTOOS 46 1.87 1.93 TBVOOS 46 1.94 2.00 TBVOOS 30 2.25 2.39 FHOOS 30 2.37 2.52 30 (>50%F) 3.10 3.20 30 (>50%F) 3.20 3.31 25 (>50%F) 3.52 3.64 25 (>50%F) 3.63 3.78 30 (:::50%F) 2.65 2.79 30 (:::50%F) 2.72 2.88 25 (:::50%F) 2.97 3.19 25 (:::50%F) 3.08 3.30 Add 0.02 to the above MCPRp limits for SLO.

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TV A Nuclear Fuel TVA-COLR-BF2C15 Core Operating Limits Report Revision 0, Page 25 Table 4 (Continued): MCPRp Limits for BOC to EOC Exposures - TSSS Scram Times (App rleabie up to C ore A verae Exposure of31,075 MWd / MTU)

MCPRpLimit MCPRpLimit EOOS Power EOOS Power AIO GE14 AIO GE14 Option (% Rated) Option (% Rated) 100 1.47 1.49 100 1.50 1.52 69 1.59 1.65 69 1.63 1.66 63 1.74 1.74 63 1.74 1.74 58 --- --- 58 --- ---

58 1.79 1.79 58 1.79 1.81 FHOOS PLUOOS 46 1.86 1.91 46 1.92 1.99 PLUOOS 30 2.23 2.37 30 2.35 2.51 30 (> 50%F) 2.64 2.79 30 (>50%F) 2.75 2.92 25 (> 50%F) 2.89 3.08 25 (>50%F) 3.03 3.24 30 (:::50%F) 2.51 2.68 30 (:::50%F) 2.60 2.79 25 (:::50%F) 2.68 2.89 25 (:::50%F) 2.81 3.03 100 1.47 1.50 100 1.50 1.52 69 1.59 1.65 69 1.63 1.66 63 1.74 1.74 63 1.74 1.74 58 --- --- 58 --- ---

58 1.79 1.79 RPTOOS 58 1.79 1.81 RPTOOS 46 1.86 1.91 FHOOS 46 1.92 1.99 PLUOOS 30 2.23 2.37 PLUOOS 30 2.35 2.51 30 (> 50%F) 2.64 2.79 30 (> 50%F) 2.75 2.92 25 (> 50%F) 2.89 3.08 25 (> 50%F) 3.03 3.24 30 (:::50%F) 2.51 2.68 30 (:::50%F) 2.60 2.79 25 (:::50%F) 2.68 2.89 25 (:::50%F) 2.81 3.03 100 1.51 1.52 100 1.53 1.55 69 1.64 1.65 69 1.66 1.68 63 1.74 1.74 63 1.74 1.74 58 --- --- 58 --- ---

58 1.79 1.79 TBVOOS 58 1.79 1.83 TBVOOS 46 1.87 1.93 FHOOS 46 1.94 2.00 PLUOOS 30 2.25 2.39 PLUOOS 30 2.37 2.52 30 (> 50%F) 3.10 3.20 30 (> 50%F) 3.20 3.31 25 (>50%F) 3.52 3.64 25 (>50%F) 3.63 3.78 30 (:::50%F) 2.65 2.79 30 (:::50%F) 2.72 2.88 25 (:::50%F) 2.97 3.19 25 (:::50%F) 3.08 3.30 100 1.52 1.53 100 1.53 1.55 69 1.64 1.65 69 1.66 1.68 63 1.74 1.74 63 1.74 1.74 58 --- --- RPTOOS 58 --- ---

RPTOOS 58 1.79 1.79 58 1.79 1.83

TBVOOS TBVOOS 46 1.87 1.93 46 1.94 2.00 FHOOS PLUOOS 30 2.25 2.39 30 2.37 2.52 PLUOOS 30 (>50%F) 3.10 3.20 30 (> 50%F) 3.20 3.31 25 (>50%F) 3.52 3.64 25 (>50%F) 3.63 3.78 30 (:::50%F) 2.65 2.79 30 (:::50%F) 2.72 2.88 25 (:::50%F) 2.97 3.19 25 (:::50%F) 3.08 3.30 Add 0.02 to the above MCPRp limits for SLO.

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TV A Nuclear Fuel TVA-COLR-BF2C15 Core Operating Limits Report Revision 0, Page 26 Table 5: MCPRp Limits for BOC to CD Exposures - NSS Scram Times (Applicable up to Core Average Exposure of32,274 MWd / MTU)

All Values Include FFTRlFHOOS and Bound Heaters In-Service MCPRpLimit MCPRpLimit EOOS Power EOOS Power AIO GE14 AIO GE14 Option (% Rated) Option (% Rated) 100 1.48 1.50 100 1.48 1.50 69 1.61 1.64 69 1.61 1.64 63 1.66 1.70 63 1.73 1.73 58 1.75 --- 58 --- ---

58 1.78 1.79 58 1.78 1.79 In-46 1.90 1.97 PLUOOS 46 1.90 1.97 Service 30 2.33 2.49 30 2.33 2.49 30 (> SO%F) 2.75 2.92 30 (>SO%F) 2.75 2.92 25 (> SO%F) 3.03 3.24 25 (>SO%F) 3.03 3.24 30 (:sSO%F) 2.60 2.79 30 (:SSO%F) 2.60 2.79 25 (:s SO%F) 2.81 3.03 25 (:SSO%F) 2.81 3.03 100 1.48 1.50 100 1.48 1.50 69 1.61 1.64 69 1.61 1.64 63 1.66 1.70 63 1.73 1.73 58 1.75 --- 58 --- ---

58 1.78 1.79 58 1.78 1.79 RPTOOS RPTOOS 46 1.90 1.97 46 1.90 1.97 PLUOOS 30 2.33 2.49 30 2.33 2.49 30 (> SO%F) 2.75 2.92 30 (>SO%F) 2.75 2.92 25 (>SO%F) 3.03 3.24 25 (>SO%F) 3.03 3.24 30 (:s SO%F) 2.60 2.79 30 (:SSO%F) 2.60 2.79 25 (:s SO%F) 2.81 3.03 25 (:SSO%F) 2.81 3.03 100 1.51 1.53 100 1.51 1.53 69 1.64 1.66 69 1.64 1.66 63 1.68 1.72 63 1.73 1.73 58 1.77 --- 58 --- ---

58 1.78 1.81 58 1.78 1.81 TBVOOS TBVOOS 46 1.92 1.98 46 1.92 1.98 PLUOOS 30 2.35 2.49 30 2.35 2.49 30 (> SO%F) 3.19 3.31 30 (>SO%F) 3.19 3.31 25 (> SO%F) 3.62 3.78 25 (> SO%F) 3.62 3.78 30 (:SSO%F) 2.72 2.88 30 (:SSO%F) 2.72 2.88 25 (:s SO%F) 3.07 3.30 25 (:SSO%F) 3.07 3.30 100 1.51 1.53 100 1.51 1.53 69 1.64 1.66 69 1.64 1.66 63 1.68 1.72 63 1.73 1.73 58 1.77 --- 58 --- ---

58 1.78 1.81 RPTOOS 58 1.78 1.81 RPTOOS 46 1.92 1.98 TBVOOS 46 1.92 1.98 TBVOOS 30 2.35 2.49 PLUOOS 30 2.35 2.49 30 (>SO%F) 3.19 3.31 30 (>SO%F) 3.19 3.31 25 (> SO%F) 3.62 3.78 25 (> SO%F) 3.62 3.78 30 (:s SO%F) 2.72 2.88 30 (:SSO%F) 2.72 2.88 25 (:SSO%F) 3.07 3.30 25 (:SSO%F) 3.07 3.30 Add 0.02 to the above MCPRp limits for SLO.

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TV A Nuclear Fuel TVA-COLR-BF2C15 Core Operating Limits Report Revision 0, Page 27 Table 6: MCPRp Limits for BOC to CD Exposures - TSSS Scram Times (Applicable up to Core Average Exposure 0[32,274 MWd / MTU)

All Values Include FFTRlFHOOS and Bound Heaters In-Service MCPRpLimit MCPRpLimit EOOS Power EOOS Power AIO GE14 AIO GE14 Option (% Rated) Option (% Rated) 100 1.50 1.52 100 1.50 1.52 69 1.63 1.66 69 1.63 1.66 63 1.68 1.72 63 1.74 1.74 58 1.77 --- 58 --- ---

58 1.79 1.81 58 1.79 1.81 In-46 1.92 1.99 PLUOOS 46 1.92 1.99 Service 30 2.35 2.51 30 2.35 2.51 30 (> SO%F) 2.75 2.92 30 (> SO%F) 2.75 2.92 25 (> SO%F) 3.03 3.24 25 (>SO%F) 3.03 3.24 30 (.:sSO%F) 2.60 2.79 30 (.:sSO%F) 2.60 2.79 25 (.:sSO%F) 2.81 3.03 25 (.:sSO%F) 2.81 3.03 100 1.50 1.52 100 1.50 1.53 69 1.63 1.66 69 1.63 1.66 63 1.68 1.72 63 1.74 1.74 58 1.77 --- 58 --- ---

58 1.79 1.81 58 1.79 1.81 RPTOOS RPTOOS 46 1.92 1.99 46 1.92 1.99 PLUOOS 30 2.35 2.51 30 2.35 2.51 30 (> SO%F) 2.75 2.92 30 (>SO%F) 2.75 2.92 25 (>SO%F) 3.03 3.24 25 (>SO%F) 3.03 3.24 30 (.:sSO%F) 2.60 2.79 30 (.:sSO%F) 2.60 2.79 25 (.:sSO%F) 2.81 3.03 25 (.:sSO%F) 2.81 3.03 100 1.53 1.55 100 1.53 1.55 69 1.66 1.68 69 1.66 1.68 63 1.70 1.74 63 1.74 1.74 58 --- --- 58 --- ---

58 1.79 1.83 58 1.79 1.83 TBVOOS TBVOOS 46 1.94 2.00 46 1.94 2.00 PLUOOS 30 2.37 2.52 30 2.37 2.52 30 (>SO%F) 3.20 3.31 30 (> SO%F) 3.20 3.31 25 (>SO%F) 3.63 3.78 25 (> SO%F) 3.63 3.78 30 (.:sSO%F) 2.72 2.88 30 (.:sSO%F) 2.72 2.88 25 (.:sSO%F) 3.08 3.30 25 (.:sSO%F) 3.08 3.30 100 1.54 1.55 100 1.54 1.55 69 1.66 1.68 69 1.66 1.68 63 1.70 1.74 63 1.74 1.74 58 --- --- 58 --- ---

58 1.79 1.83 RPTOOS 58 1.79 1.83 RPTOOS 46 1.94 2.00 46 1.94 2.00 TBVOOS TBVOOS 2.37 2.52 30 2.37 2.52 PLUOOS 30 30 (>SO%F) 3.20 3.31 30 (> SO%F) 3.20 3.31 25 (>SO%F) 3.63 3.78 25 (>SO%F) 3.63 3.78 30 (.:sSO%F) 2.72 2.88 30 (.:sSO%F) 2.72 2.88 25 (.:sSO%F) 3.08 3.30 25 (.:sSO%F) 3.08 3.30 Add 0.02 to the above MCPRp lImits for SLO.

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TV A Nuclear Fuel TV A-COLR-BF2ClS Core Operating Limits Report Revision 0, Page 28 Figure 1 APLHGR Limits for Bundle Type GE14-PI0DNAB416-16GZ (GE14 EDB#2600) 14.00 13.00 I 12.00 I UNACCEPTABLE OPERATION t 11.00

)'W..'

10.00 ~ I~ ~

~ ~

r----.

9.00

~

! 8.00 ~

~

J 7.00 IACCEPTABLE OPERATION I

~

n: '\

C)

J: 6.00 \It

..J Il.

<< 5.00

\

4.00 3.00 2.00 1.00 0.00 0.00 10.00 20.00 30.00 40.00 50.00 60.00 Average Planar Exposure (GWD/MTUj Most Limiting Lattice for Each Exposure Point Average Planar LHGR Average Planar LHGR Average Planar LHGR Exposure Limit Exposure Limit Exposure Limit (GWD/MTU) (kw/ft) (GWD/MTU) (kw/ft) (GWD/MTU) (kw/ft) 0.00 9,41 8.82 10.56 22.05 10.85 0.22 9.51 9.92 10.65 27.56 10.50 1.10 9.61 11.02 10.74 33.07 10.10 ...'

2.20 9.73 12.13 10.85 . 38.58 9.63 I

3.31 9.86 13.23 ........

10.85 44.09 .....

9.10 4,41 10.00 14.33 10.86 49.60 8.57 5.51 10.14 .......

15,43 10.88 55.12 8.02 6.61 10.28 16.53 10.91 60.63 6.24 7.72 10,42 18.74 10.94 63.50 4.93 These values apply to both Turbine Bypass In-Service and Out-Of-Service.

These values apply to both Recirculation Pump Trip In-Service and Out-Of-Service.

These limits are for dual recirculation loop operation. Single Loop Operation (SLO) adjustments are performed as described in Section 2 Browns Ferry Nuclear Plant Unit 2 Cycle 15

TV A Nuclear Fuel TVA-COLR-BF2CIS Core Operating Limits Report Revision 0, Page 29

(

Figure 2 APLHGR Limits for Bundle Type GE14-PI0DNAB416-16GZ (GE14 EDB#2601) 14.00 13.00 12.00 I

I 1

r UNACCEPTABLE OPERATION}

I 11.00 I

~ t--..a

~

10.00 / --....

~

E' 9.00 8.00

~I'--- I

E 7.00

J lACCEPTABLE OPERATION I '\

~

0::

(!)

J: 6.00

..J a.

<I:

5.00 \

4.00 I

3.00 2.00 1.00 0.00 0.00 10.00 20.00 30.00 40.00 50.00 60.00 Average Planar Exposure (GWD/MTU)

Most Limiting Lattice for Each Exposure Point Average Planar LHGR Average Planar LHGR Average Planar LHGR Exposure Limit Exposure Limit Exposure Limit (GWD/MTU) (kw/ft) (GWD/MTU) (kw/ft) (GWD/MTU) (kw/ft) 0.00 9.43 8.82 10.76 22.05 10.88 0.22 9.47 9.92 10.83 27.56 10.50 1.10 9.54 11.02 10.91 33.07 10.10 2.20 9.67 12.13 10.99 38.58 9.66 3.31 9.83 13.23 11.03 44.09 9.13 4.41 10.02 14.33 11.02 49.60 8.59 5.51 10.21 15.43 11.02 55.12 8.03 6.61 10.43 16.53 11.03 60.63 6.38 7.72 10.62 18.74 11.02 63.82 4.92 These values apply to both Turbine Bypass In-Service and Out-Ot-Service.

These values apply to both Recirculation Pump Trip In-Service and Out-Ot-Service.

These limits are tor dual recirculation loop operation. Single Loop Operation (SLO) adjustments are

(

performed as described in Section 2 Browns Ferry Nuclear Plant Unit 2 Cycle 15

TV A Nuclear Fuel TV A-COLR-BF2C15 Core Operating Limits Report Revision 0, Page 30 Figure 3 APLHGR Limits for Bundle Type GE14-PI0DNAB416-18GZ (GE14 EDB# 2627) 14.00 i

I I I 13.00 I -j 12.00 - - - - - - - -

I (UNACCEPTABLE OPERATION l-- --

11.00

.. _ - - - f---

~ ~~ I It-I 10.00

~

E'

-*E

J It:

C)

J:

9.00 8.00 7.00 6.00 i

l ACCEPTABLE OPERATION J

+--. ~'" "\f---

.~

l \

...J

a. I

-+

<I:

5.00 4.00 I

II I 3.00 -_..

---._---t-2.00 -r-----' I 1.00 0.00 i

t-----j-0.00 10.00 20.00 30.00 40.00 50.00 60.00 Average Planar Exposure (GWD/MTU)

Most Limiting Lattice for Each Exposure Point Average Planar LHGR Average Planar LHGR Average Planar LHGR Exposure Limit Exposure Limit Exposure Limit (GWD/MTU) (kw/ft) (GWD/MTU) (kw/ft) (GWD/MTU) (kw/ft) 0.00 9.26 8.82 10.54 22.05 10.85 0.22 9.34 9.92 10.65 27.56 10.49 9.47 .......... !

1.10 ...

11.02 ..........

10.75 33.07 10.09 I I 2.20 9.62 ..... ...............

12.13 10.85 38.58 9.60 u

I 3.31 9.77 13.23 10.85 44.09 9,9 9 .........

I 4.41 9.93 14.33 10.86 49.60 8.56 .'.

5.51 10.09 15.43 10.88 55.12 8.01 6.61 10.25 16.53 10.91 60.63 6.21 7.72 10.41 18.74 10.93 63.42 4.93 These values apply to both Turbine Bypass In-Service and Out-Of-Service.

These values apply to both Recirculation Pump Trip In-Service and Out-Of-Service.

These limits are for dual recirculation loop operation. Single Loop Operation (SLO) adjustments are performed as described in Section 2 Browns Ferry Nuclear Plant Unit 2 Cycle 15

TV A Nuclear Fuel TVA-COLR-BF2CI5 Core Operating Limits Report Revision 0, Page 31 Figure 4 APLHGR Limits for Bundle Type GE14-PI0DNAB417-18GZ (GE14 EDB# 2628) 14.00 I i

13.00 ~~-----

I 12.00 r UNACCEPTABLE OPERATION) 11.00 ".

10.00

. / ~

I~

~

V ~

I 9.00 f - - - -

8.00 I

i

~~

I

'E 7.00

J l ACCEPTABLE OPERATION I 0::

Cl J: 6.00 I

"ta\

...J a.

0:(

5.00 \

4.00 3.00 ".~~

i 2.00 --

1.00 f:---------I-**

I I

0.00 0.00 10.00 20.00 30.00 40.00 50.00 60.00 Average Planar Exposure (GWD/MTU)

Most Limiting Lattice for Each Exposure Point Average Planar LHGR Average Planar LHGR Average Planar LHGR Exposure Limit Exposure Limit Exposure Limit (GWD/MTU) (kwlft) (GWD/MTU) (kw/tt) (GWD/MTU) (kwlft) 0.00 9.39 8.82 10.74 22.05 10.88 0.22 9.43 9.92 10.83 27.56 10.50 1.10 9.50 11.02 10.91 33.07 10.10 2.20 9.63 12.13 11.00 38.58 9.62 3.31 9.80 13.23 11.02 44.09 9.12 4.41 9.99 14.33 11.02 49.60 8.58 5.51 10.19 15.43 11.02 55.12 8.02 6.61 10.41 16.53 11.03 60.63 6.35 7.72 10.64 18.74 11.02 63.74 4.92 These values apply to both Turbine Bypass In-Service and Out-Ot-Service.

These values apply to both Recirculation Pump Trip In-Service and Out-Ot-Service.

These limits are tor dual recirculation loop operation. Single Loop Operation (SLO) adjustments are performed as described in Section 2 Browns Ferry Nuclear Plant Unit 2 Cycle 15

TVA Nuclear Fuel TVA-COLR-BF2C15 Core Operating Limits Report Revision 0, Page 32

(

Figure 5 APLHGR Limits for all ATRIUM-IOTM Fuel (AlO) 14.00 13.00 12.00

............ J UNACCEPTABLE OPERATION}

11.00 ~

10.00 ~

9.00 ~

~

I.... 8.00 ~

'E 7.00

~

i

!r:

l ACCEPTABLE OPERATION

(!)

J: 6.00

...J Il.

<< 5.00 4.00 3.00 2.00 1.00 0.00 0.00 10.00 20.00 30.00 40.00 50.00 60.00 Average Planar Exposure (GWD/MTU)

Average Planar LHGR Exposure Limit (GWD/MTU) (kw/ft)

These values apply to both Turbine Bypass In-Service and Out-Of-Service.

These values apply to both Recirculation Pump Trip In-Service and Out-Of-Service.

These limits are for dual recirculation loop operation. Single Loop Operation (SLO) adjustments are performed as described in Section 2

(

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TVA Nuclear Fuel TVA-COLR-BF2C15 Core Operating Limits Report Revision 0, Page 33 Figure 6 GE14 Power Dependent MAPLHGR Multiplier - MAPFAC(P)

NSSITSSS Insertion Times - All Exposures 1.1 0.9

~

0.8 Turbine Bypass In-Service (TBVIS)

"- ~ ~

~

~I~

ii:' 0.7 g ~

~

- ~

~

~ 0.6

~ '-----

~ Turbine Bypass Out-Of-Service (TBVOOS)

~

0.5

~

~ ~BVIS: ~ 50% Core Flow TBVIS: > 50% Core I 0.4 V~

TBVOOS: < 50% Core TBVOOS: > 50% Core Flow I

0.3 i

0.2 I 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 Power (% Rated)

Turbine Bypass In-Service Turbine Bypass Out-Of-Service Core Power Core Power MAPFAC(P) MAPFAC(P)

(% ratedl t% ratedl 100 0.89 100 0.87 30 0.48 30 0.48 Core Flow> 50% rated Core Flow> 50% rated 30 0.41 30 0.38 25 0.38 25 0.35 Core Flow ~ 50% rated Core Flow ~ 50% rated 30 I 0.46 30 0.43 25 I 0.43 25 0.38 MAPFAC(P) is not dependent upon any Equipment Out-Of-Service except Turbine Bypass.

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TVA Nuclear Fuel TVA-COLR-BF2C15 Core Operating Limits Report Revision 0, Page 34 Figure 7 Flow Dependent MAPLHGR Factor - MAPFAC(F)

(GE14) 1.00 +----+----t---~f---____+/-.._"7-..,.".~~~=+~~~.....,..~"""

0.90 +----+----+-----'lC""""""C it 0.80 +----+----c g

a..

0:(

0.70 +------

0.60 -1"'----+----+----1-------1----+----+-----+--1 30 40 50 60 70 80 90 100 Core Flow (% Rated)

Max Core Flow 102.5% Rated Max Core Flow 107% Rated Core Flow Core Flow MAPFAC(F) MAPFAC(F)

(% rated) (% rated) 30 0.62 30 0.60 71 1.00 75 1.00 102.5 1.00 107 1.00 These values bound both Turbine Bypass In-Service and Out-Of-Service.

These values bound both Recirculation Pump Trip In-Service and Out-Of-Service.

The 102.5% maximum flow line is used for operation up to 100% rated flow.

The 107% maximum flow line is used for operation up to 105% rated flow (ICF).

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TV A Nuclear Fuel TVA-COLR-BF2C15 Core Operating Limits Report Revision 0, Page 35 Figure 8 LHGR Limits for all GE-14 Fuel (GE14) 14.00 I

13.00 12.00 ~ r UNACCEPTABLE OPERATION I 11.00 ~ i 10.00 I

I "~ "-.. II 9.00 ~

~

I'----

~ 8.00

'E 7.00 l ACCEPTABLE OPERATION I I \

\

i 0::

(!) 6.00

x:

...J 5.00 I I 4.00 I I I

(

3.00 2.00

---I- I I

1.00 0.00 0.00 10.00 20.00 30.00 40.00 50.00 60.00 70.00 Pellet Exposure (GWD/MTU)

Pellet LHGR Exposure Limit (GWD/MTU) (kw/ft)

I******************

0.00 13.40 16.00 13.40 I

63.50 8.00 70.00 5.00 These values apply to both Turbine Bypass In-Service and Out-Of-Service.

These values apply to both Recirculation Pump Trip In-Service and Out-Of-Service.

These limits apply to both Two Loop Operation (TLO) and Single Loop Operation (SLO).

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TV A Nuclear Fuel TV A-COLR-BF2C15 Core Operating Limits Report Revision 0, Page 36 Figure 9 LHGR Limits for all ATRIUM-10 Fuel (AlO) 14.00 13.00 i'-.

12.00

~ r UNACCEPTABLE OPERATION' 11.00 ~

10.00 ~

9.00 ~ ........

~

I-

E 8.00 7.00

~

i l ACCEPTABLE OPERATION I II::

(!) 6.00 J:

....I 5.00 4.00 3.00 2.00 1.00 0.00 0.00 10.00 20.00 30.00 40.00 50.00 60.00 70.00 Pellet Exposure (GWD/MTU)

Pellet LHGR Exposure Limit (GWD/MTU) (kw/ft)

These values apply to both Turbine Bypass In-Service and Out-Ot-Service.

These values apply to both Recirculation Pump Trip In-Service and Out-Ot-Service.

These limits apply to both Two Loop Operation (TLO) and Single Loop Operation (SLO).

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TV A Nuclear Fuel TVA-COLR-BF2C15 Core Operating Limits Report Revision 0, Page 37 Figure 10 AIO Power Dependent LHGR Multiplier - LHGRFAC(P)

NSSITSSS Insertion Times - All Exposures 1.10 1.00

~

0.90 i""'"

~

0.80 Turbine Bypass In-Service (TBVIS) ~

~

iL iJ if 0.70 I '-~~ ~ ~

a. ~-- ..........

~

Turbine Bypass Out-Of-Service (TBVOOS) 0.60

~ TBVIS: 550% Core Flow i TBVOOS: !: 50% Core 0.50 TBVIS: ,. 50% Core I V I I TBVOOS: > 50% Core Flow I

/~

DAD 0.30 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 Power (% Rated)

Turbine Bypass In-Service Turbine Bypass Out-Of-Service Core Power Core Power LHGRFAC(P) LHGRFAC(P)

(% rated) (% rated) 100 0.93 100 0.93 30 0.64 30 0.63 Core Flow> 50% rated Core Flow> 50% rated 30 I 0.51 30 I 0.45 25 I 0.46 25 I 0.39 Core Flow::. 50% rated Core Flow::. 50% rated 30 I 0.55 30 I 0.55 25 I 0.50 25 I 0.48 LHGRFAC(P) is not dependent upon any Equipment Out-Of-Service except Turbine Bypass.

Browns Ferry Nuclear Plant Unit 2 Cycle 15

TVA Nuclear Fuel TVA-COLR-BF2C15 Core Operating Limits Report Revision 0, Page 38

(

Figure 11 Flow Dependent LHGR Multiplier - LHGRFAC(F)

(AIO Fuel) 1.00 0.95 LL g

!5 0.90

r:

..J 0.85 +--~~~+-~~~+-~~--If--~~--+~~~-+~~~-+~~~-+-~--l 30 40 50 60 70 80 90 100 Core Flow (% Rated)

Max Core Flow 102.5% Rated Max Core Flow 107% Rated Core Flow Core Flow LHGRFAC(F) LHGRFAC(F)

(% rated) (% rated) 30 0.91 30 0.88 48 1.00 54.4 1.00 102.5 1.00 107 1.00 These values bound both Turbine Bypass In-Service and Out-Of-Service.

These values bound both Recirculation Pump Trip In-Service and Out-Of-Service.

The 102.5% maximum flow line is used for operation up to 100% rated flow.

The 107% maximum flow line is used for operation up to 105% rated flow (ICF).

Browns Ferry Nuclear Plant Unit 2 Cycle 15

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 second 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months shift did not count due to the absence for two hours for FFD testing. Part (2) is incorrect. Although 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months will yield the required 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months of 10CFR 50, OPDP-lO Appendix E 4.0.A.3 requires the same number of complete shifts for activating a license as that required to maintain a license.

c. Part (1) is incorrect. OPDP-10 does not allow credit for absences from shift that exceed a few minutes in duration. Although only two additional hours may meet the total requirement, credit for a shift must extend from turnover to turnover. Therefore, to comply with OPDP-lO Appendix C step 4.0.E Table, three additional shifts are required. Part (2) is correct.

Although only 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months are required to achieve 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months total, Appendix E 4.0.A.3 requires five total shifts to ensure the 40 hour4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months requirement has been met.

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 Examination Outline Cross-reference: Level RO 2S6000G2.2.44 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. Reactor Condensate KIA # 256000G2.2.44 Importance Rating 4.2 4.4

[ Proposed Question: SRO #18 A seismic event has resulted in the following Unit-2 plant conditions:

All control rods are inserted.

RPV level is (-) 12S inches and lowering slowly.

RPV pressure is 4S0 psig with a cool down in progress at <90 OF/hr.

RHR Loop II is lined up for Drywell Spray.

All other ECCS systems are unavailable.

Drywell pressure is 4.8 psig and lowering.

ADS has been inhibited in accordance with 2-EOI-1, "RPV Control" step RC/L-7.

Which ONE of the following describes the required actions to mitigate this event?

A. Enter 2-EOI-C1, "Alternate Level Control" and direct performance of 2-EOI-Appendix GA, "Injection Subsystems Lineup Condensate."

B. Enter 2-EOI-C1, "Alternate Level Control" and direct performance of 2-EOI-Appendix SA, "Injection System Lineup Condensate/Feedwater."

C. Enter 2-EOI-C2, "Emergency Depressurization" and direct performance of 2-EOI-Appendix GA, "Injection Subsystems Lineup Condensate."

D. Enter 2-EOI-C2, "Emergency Depressurization" and direct performance of 2-EOI-Appendix SA, "Injection System Lineup Condensate/Feedwater."

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. Appendix SA is a lineup for injection with RFPs which require MSIVs open. With RPV level below -122 inches! the MSIVs are closed. In addition! given all rods are in! performance of EOI Appendix 8A to bypass the MSIV low water level isolation is not appropriate.

c. Part (1) is incorrect. Direction to perform Emergency Depressurization based on reactor water level is given from EOI-C1 when RPV level drops below -162 inches. Other conditions given in the stem do not require Emergency Depressurization since Drywell Sprays have been initiated and appear to be effective. Part (2) is correct.

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): 2-EOI-1, "RPV Control", 2-EOI-App 5A (Attach if not previously provided) 2-EOI-C1, Alt Lvi Control", 2-EOI-App 6A Proposed references to be provided to applicants during examination: None Question Source: Bank#

Modified Bank # (Note changes or attach parent)

New 7/18/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X Comments: This question is provided without reference to the EOI flowchart. As such, the candidate must utilize available indications and apply knowledge of the EOI mitigation strategy to determine the correct answer.

~

,- .. =£:~ t

...... -"" --"',.-~-

, -.-.---.- =*'1'_ ~

,-~:----:i 1

~.. ~

~

~-.. - ..

":::.4'P,"~~

'Ii ~-----*iI

'I I. ~**--"*--**I!

,"".".. _.0' Jaa4s>lJOM uO!Jsano 9- ~Ot-S3 WJ0::l UOneu!wex3 ua>>!JM aldwes

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet 2-EOI APPENDIX-6A Rev. 4 Pa.ge 20f 2

5. VERIFY OPEN the following heater isolation valves!

2-FCV-3-38, HP HTR 2A2 FW INLET ISOL VLV

2-FC\l'-3-31 , HP HTR 2B2 FW INLET ISOL VLV

2-FCV-3-24, HP HTR 2C2 FW INLET ISOL VLV

2-FCV-3-75, HP HTR 2Al FW OUTLET ISOL VLV

2-FC"1-3-76, HP HTR 2BI FW OUTLET ISOLVLV

2-FC'V-3-77 I HP HTR 2c1 FW OUTLET ISOL 'IlLV.

6. VERIFY OPEN the following RFP suction valves:

2-FC"1-2-83, RFP 2A SUCTION V1UJVE

2-FCV-2-95, RFP 2B SUCTION "VALVE

2-FC"1-2-108, RFP 2C SUCTION V1I.LVE.

7. VERIFY at least one condensate pump running.

8. VERIFY at least one condensate booster plb":J.p rup.ning.

9. ADJUST 2-LIC-3-53, ru-w START-UP LEVEL CONTROL, to control injection (Panel 2-9-5).

10. VERIFY RF""i'l" flc;.;." to RPV.

LAST PAGE

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet 2-E01 APPENDIX-SA Rev. 8

.Pacre 3 of 4

15. RAISE RFFT 211. (2B) (2e) speed UNTIL RFP discharge pressu.re i . s approximately equal to RPV pressure using ANY of the follor,iing methods on Panel 2-9-5:

US1ng individual 2-HS-46-8A (9A) (lOA), RFPT 211. (2B) (2C}

SPEED CONT RAISE/LOWER s."itch in l-Uk""mAL GUVERNOR, OR

Using individual 2-51(;-46-8 (9) (10), RFPT 2A(2B) (2C)

SPEED CONTROL PDS in ~l{UALr OR

Using 2-L1C-46-5, REACTOR WATER LE"V'EL CONTROL PDS, in MANUAL with individual 2-5IC-46-8(9) {lO}, RFPT 211.(28) (2C) SPEED CONTROL PD5 in AUTO.

16. SLOWLY RAISE speed of RFPT UNTIL RFW flow to the RPV i.5 indicated usi.ng ANY of the fol1,;),"ing methods on Panel 2-9-5: --

Using individual 2-HS-46-8A(9A} (1011.), RFPT 2A{2B) (2C}

5PE.ED CONT RAISE/LO'"rlER sitch in l-L~:mAL GO\lERNOR, OR

Using individual 2-51(;-46-8 (9) {10), RFPT 2A(2B) (2C)

SPEED CONTROL PDS in. MfI.NUAL, OR

Using 2-L1C-46-5, REACTOR WATER LE"V'EL CONTROL PDS, in l-lANUll.L ,.;rith individual 2-5IC-46-8 (9) (lO), RFPT 2A(28) (2C} SPEED CONTROL PDS in AUTO.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet 2-EOI APPENDIX-SA Rev. B Paqe <1 of 4

17. ADJUST RFPT speed as necessary to control injection using ANY of the following methods on Panel 2-9-5:

Using individual 2-HS-46-8A(9A) (lOA), RFPT 2A(2B} (2C)

SPEED CONT RAISE/LOWER switch in MANUAL GOVERNOR, OR

Using individual 2-SIC-46-8 (9) (lO) I RFPT ZA,(2B) (2C)

SPEED CONT.RDL PDS in MANUAL, OR

Using 2-LIC-46-5, REACTOR WATER LEVEL CONTROL PDS, in MANUAL with individual 2-SIC-46-8(9) (10), RFPT 2A (28)( 2C) SPEED CONTROL PDS .in AUTO.

18. WHEN *** RPV level is approximately equal to desired level AND automatic level control is deSired, THEN *** PIACE 2-LIC-46-5, REACTOR WATER LEVEL CONTROL PDS, in AUTO with individual 2-SIC-46-8(9J(lO) .. RFPT 2A (28) (2C) SPEED CONTROL PDS in AUTO ..

LAST PAGE

TVA Nuclear Fuel TVA-COLR-BF2C15 Core Operating Limits Report Revision 0, Page 39 (Final)

Figure 12 Flow Dependent MCPR Limit - MCPR(F)

(All Fuel) 1.43~-------r--------r--------r------~~------'--------'--------'---~

1.38 I """

1.33+1-----'"

ii:'

~

()

1.28 I I ~

1.23 +I----+----+----f.-----="

1.18 +-1~~~+--~~"_'+~~~+__~~---+~~~....,.....~~~j__o_.o_~~_+---.-...J 30 40 50 60 70 80 90 100 Core Flow (% Rated)

Max Core Flow 102.5% Rated Max Core Flow 107% Rated Core Flow Core Flow MCPR(F) MCPR(F)

(% rated) (% rated) 30 1.37 30 1.40 72 1.21 78 1.21 102.5 1.21 107 1.21 These values bound both Turbine Bypass In-Service and Out-Of-Service.

These values bound both Recirculation Pump Trip In-Service and Out-Of-Service.

The 102.5% maximum flow line is used for operation up to 100% rated flow.

The 107% maximum flow line is used for operation up to 105% rated flow (ICF).

-)

Browns Ferry Nuclear Plant Unit 2 Cycle 15

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): OPDP-10 (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 7/23/2008 Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X Comments:

)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet NPG Standard license Status Maintenance, OPDP-10 Department Reactivat.ion and Proficiency for Rev.UOOO Procedure Non-licensed Positions Page 14 of 30 Appendix C (Page 1 of .3)

Browns Ferry Nuclear Plant Requirements for Maintaining Active license Status 1.0 PURPOSE The purpose of th!isappendix is to provide administrative instructions in order to comply with 10CFR55_53 (e), ." "actively performing the functions of an operator or senior operator.

2.0 REFE RiENC ESfBACKGROUND A. References

1. 10 CFR 5O.54{rn)(2)(i)

2. 10 CFR 55.4

3. 10 CFR 55.53{e)

4. NUREG-1262 - Preface; pages 11-80

5. Technical Specification B. To ma,intain active status, per 55.53(e), Conditions of License, the licensee shall actively perform the functions of an operator or senior operator on a minimum of seven (7) a-hour or five (5) 12-hour shifts* per calendar quarter.

C. ,A.ctively performing the functions of an operator or senior operator means that an indMd,ual has a position on the shm crew that requires the individual to be iicensed as defined in Technical Speeifical1on, and that the individual cames out and is responsible tor the duties covefed I)y that positilon.

D. Technical Specifications and 10 GFR 50.54 specify the minimum requirement per shift E. Licensed personnel who dO' not meet these requirements are designated as inactive licensees.

3.0 RESPONSIBILITIES A. All licensed personnel who maintain an active license shall comply with these requirements.

B. All licensed personnel who maintain an active ficense and are OFF SHIFT (not part of

a. rotating shift) shall provide on-sMt dO'cumentation quarteny to the Operations Superintendent [Appendix DJ.

C. The Operations Superintendent is responsii)le for administering this program and documentation.

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet NPG Standard Ucen~e Status Maintenance, OPDP-10 Department Reactivation and Proficiency for Rev. 0000 Procedure Non-Licensed Positions Page 150130 Appendix C (Page 2 of 3)

Browns Ferry Nuclear Plant Requirements for Maintaining Active license Status 4.0 INSTRUCTIONS A. Indivrduals assigned to the foMowing positions, AND NO OTHERS, on each Shift, are considered to be acti .... ely performing the functions man operator or senior operator in order to maint<iinactive license status:

Browns Ferry Nuciear

1. Shift Manager

2. UnIt 1 Unit Supervisor !Controi Room SRej

3. Unit 2 Unit Supervisor {Cootrot Room SRej

4. Unit 3 Unit Supervisor {COOtrol Room SRej

5. Unit 1 Board and Desi( Ras

6. Unit 2 Board and Desk Ras

7. Unit 3 Board and Desi( Ras B. To be granted credit for a shift, the individual wlll be present from shift turnover filru shift turnover. Short absences from the Control Room are acceptable (Le., rest room visits). Absences from the Control Room for extended periods (I.e., Fitness-far-Duty testing) will not count towards shift run coons. For these type of cases, the time absence will be made up by working additional time 00 another shift or an additional shift.

C. The shirt period is defined by the schedule worked by the fotating Shift crews. Either 12-;hour or B-hou r shifts is the nomlal. If a 12-hour shift rotation is used, then a minimum of five (5) shi.fts in a licensed position per Quarter, or if an 8-hoor Shift rotation is used, then a minimum of seven (7) shifts in a Ilcensedposition per Quarter tS required in order to remain "active.n D. Technical Specificaoons 110CFR50 for each site contains the requirement for the minimum number of licenses required. Howe....er, only the positions listed for the applicable site as listed in 4.0A above qualify for license maintenance.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet NPGStmdard license Status Maintenance, OPDP-10 Department Reactivation and Proficiency for Rev. 0000 Procedure Non-Ucensed Positions Page 16 ofJO AppendixC (Page JofJ)

Browns Ferry Nuclear Plant Requirements for Maimaining Active License Status 4.0 tNSTRUCTIONS (continued)

E. If the operating crews convert from an 8-hour to a 12-hour, or a 12-hoor to anS-hoor shill: rotation schedule during a calendar quarter, then the nulfl1iber of shifts ~ired to be wooked in a licensed position to be crediteclfor active lcense maintenance on the combination of Shifts (8's and 12'5) will be in accordance with the foMowing:

8-HourShHts TO 12-l-Iour Shifts 12-HourShifls TO 8-HourShiHs

,. Shifts ComPleted # Additional Shifts 11 Shifts Completed ,. Additional Shifts PJior to Change Needed On New Prior to Change Needed On New Schedule Sc:hedule 6 1 4 :2 5 :2 3 3 4 :3 2 5 3 :3 1 6 2 4 0 1 1 5 - -

0 5 - -

F. The individual as&gned to one of the seven (7) positions designated for maintaining an active lioense, shdilliog in and "our on the Narrative log for each shiftWOflked.

G. The Shift Manager on each shift shaIIl vefffy that the data entered into the "Shift staling log" in the Narrative log is correct for their shill:.

H. A Shift Manager shall actively perform the fundions of a Shift Manager a minimum of seven 8-hoor or five 12-hour shifts per calendar quarter to remain current 5.0 OOCUMENTATlON A.Appendix D contains the form "(Active) Ucensed 00-8hift Personnel Quarter1y On-Shift TIme Documentation" that is submitted by active off-shift lcensed individuals each Quarter to the Operations Superintendent.

B. The Control Room logs are the legal record of watc'hstmder assignment

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet NPG Standard License Status Maintenance, OPDP~10 Department Reactivation and Proficiency fOf Rev. 0000 Procedure Non-licensed Positions P*alQe18 of 30 Appendix E (Page 1 of 5)

Browns Ferry Nudear Plant Requirements for Returning an Inactive license to Active Status 1.0 PURPOSE This appendix is intended to provide additional guidance, to return a ticensed individual to an active status.

2.0 REFERENCES

fBACKGROOND A. The Code of Federal Regulation, 10 CFR5553 f(2) specifies relumllg a ticense to active status. The intent of the law is to ensure proficiency in the conduct of licensed actMties prior to assuming licensed duties. The followllg requirements are addressed as part of this law:

1. The Qlulilifications and status of the licensee are current and valid. This requirement ensures the licensee has oompl!etced all required requalification training., .including plant lllOOif.ications and indus!:.ryevents; and second~y, that ~I conditions of his/her license are still being met.

2. This licensee has oomp!eteda minimum of 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months of shift functions under the d,irection of a reactor operator or senior operator, as appropriate, and in the pos.ition to which the individual will be assigned. This ensures '!hat an active license is directing or performi,ng the manipulations of pilant controls, and allows the ina ctive individual to obtain proficiency at hislher watch station. hnclucled within the minimum of 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months is the following:

a. A complete review of turnover procedures by the reador operator or senior reactor operator as appropriate for the position, to ensure that th,e licensee is familiar with current shift turnover practices.

b. A complete tour of the plant, to ensure the individual is aware of changing plant conditions that have occurred since he/she has been inactive. The individual performing the tour w~1 be accompanied by a Licensed Reactor Operator or a Licensed Senior Reactor Operator, as appropriate.

J.G RESPONSIBIUTIES A. All licensed personnel who maintain an active license shall comply with these requirements.. Tile Operations Superintendent is responsible for adminislering the process.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet NPG Standard license Status Maintenance, OPDP-10 Department Reactivation and Proficiency for Rev. 0000

_ __"rocefjDre Noo-Ucensed Posill:ions p.age 19 of 30 AppendixE (Page20f 5)

Browns Ferry Nuclear Plant Requirements for Retumiftg an Inactive license to Active Status 4.0 INSTRUCTIONS A. The f04lowingguidetines are to be llsed when reac1iva~i1ng iii Wcense:

1. POOr to standing the minimum of 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months of shift fundioos, the licensed i1ndivfdual shaill meet with the Operation Training Manager and the Operations Superintendent to discuss htslher current status and any smnaardsandlor expectations. For certain individuals, additional requirements may be imposed (g.reater than those required by law) if directed by the OperationS Superintendent.

2. The following positions are the only ones that qualify for reactivation of a license:

Browns FellY Nuclear

a. Shift Manager

b. Unit 1 Unit SupeMsor ~ Room SROJ

c. Unit 2 Unit SupeMsor ICcnIroiI Room SROJ

d. Unit 3 Unit SUpeMsor ~ Room SROJ

e. Unit 1 Board and Desk ROs

f. Unit 2 Board and Desk RQs Q. Unit 3 Board and Desk ROs

3. The individual shall be under the direct supervision of an active Ik:ensed individual in the position to which the individuail will be assigned. To receive credit for a shift, the individual wil be present from shift turnover thru shift tumover. Short absences from the Control Room are acceptatJae (i.e., rest room visits); hoWever, the totail time in the Control Room under supervision wil total at least 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months (this 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months does not includelle plant tour).

To ensure that the minimum of 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months is obtaiined in the Control Room under supervision,the break-in period wll be seven (7)-8 hour sh~fts or five (5)-12 hour shifts. This applies to al positions used to re-activate Gl. leanse to active stmls.

4. The individual shall make a Narrative Jog entry at the start of the shfft which witt include the f04fo!wingat a minimum:

a. Name and time of assuming shift

b. Shift Position (as identified in 4.0A.2) assumed under direction

c. Name of the operators (Board and Desk), Control Room SRO, *orShift Manager providing supervision.

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

')

NPG Standard Ucmse Status Maintenance, OPDP-10 Department ReiICtivatkm and Profic.iency for Rev* *aO Procedure Non-Uamsed Positions P3Jge 20 of 30 Appendix E (Page J of 5)

Browns Ferry lNudear Plant Requirements for Returning an Inactive Liceftse to Active Status 4.0 INSTRUCTIONS (continued)

5. The individual shall make a Narrative Log enny at the end of the shift indicating they have completed the shift under supervision. A copy of the Narrative log for each shift worked shaill be obtained for processing after the break-in is 'COif'rI\PIete.

This will be the ,entire log for the shift WOOked and not selected entries.

6. The individual shall OOImplete Appendix G for each shift liisliing unit, shift, position

.assumiillg, along with the activities the individual was personally involved in.

Time, Position, Unit, Activity. and Date must be filled out for each activity performed. The position the indiYidllJal is holding must be one of the seven indicated in step 4.0A,2. Appendix H is to be used to acooont for a plant toor and shift turnover briefing. Appendix H is reqUlired to be signed ~ the Operations Superintern1entensuring that all aptJendixes have been reviewed and once reviewed, these appendixes will be submitted with the reactivaiion documentation and will become part of the individuais trainling recorn.

7. If liamse re-activation is lor a. multi~nit site, then the individual shall divide their time between the urnts to ensure adequate brealk-in in all license areas Hley may be assigned. The amount of time in each Conoot Room does rot have to be equalzec! between units, but should be enough t{) ensure that the individual will be ready to assume the shift once their liamse is retumed to a:ctive status.

8. It an individual moves from {)ne unit to another unit dulliingthe same shift for the purpose of breaking-in on the other unit, the individual shall make an log entry indicating that they are moving to Hle other unit to continue their breaik-iiIl.

Another entry, to include the areas in4.0A.4, wml be made when the individual goes under instruction on Hle new unlit. This reqUliremem is not aPrllicable to an iiIldividual being re-activated as a Shift Manager since the break-in wotdd stMI be under the same individual.

9. The individual shalll'eView the turnover procedures with an active reactor operator or senior reactor operator, as appilicaDie. The following are the minimum procedures that will be reviewed:

1iI. Plant Operations Manager, OperationS Superintendent"andfor Operations Support Superintendent will deClide the requirements here.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO G2.1.44 Tier # 3 Conduct of Operations: Knowledge of RO duties in the control room Group #

during fuel handling.

KIA # G2.1.44 Importance Rating 3.9 3.8

[ Proposed Question: SRO #20 Unit-2 is in Mode 5 loading fuel into the reactor when the following indications are received:

FUEL POOL FLOOR AREA RADIATION HIGH (2-XA-55-3A, window 1).

REFUELING ZONE EXHAUST RADIATION HIGH (2-XA-55-3A, window 34).

RX BLDG AREA RADIATION HIGH (2-XA-55-3A, window 22).

REACTOR ZONE EXHAUST RADIATION HIGH (2-XA-55-3A, window 21).

SRM PERIOD (2-XA-55-5A, Window 20).

Which ONE of the following describes the required actions for this condition?

Direct the Unit Operator to enter (1) and take action to (2)

(1) (2)

A. 2-AOI-79-1, "Fuel Damage During Refueling," evacuate ALL personnel from the Refuel Floor.

B. 2-AOI-79-2, "Inadvertent Criticality During Incore evacuate ALL personnel from Fuel Movements," the Refuel Floor.

c. 2-AOI-79-1, "Fuel Damage During Refueling," evacuate ONLY non-essential personnel from the Refuel Floor.

D. 2-AOI-79-2, "Inadvertent Criticality During Incore evacuate ONLY non-essential personnel from the Refuel Floor.

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

) I Proposed Answer: B Explanation: a. Part (1) is incorrect. All four radiation annunciators are common to each AOI, but the SRM Period annunciator is indicative of inadvertent criticality and not fuel damage. Part (2) is correct for inadvertent criticality due to the potential for lethal radiation levels.

b. Correct answer.

c. Part (1) is incorrect as stated in (a) above. Part (2) is incorrect. Evacuation of non-essential personnel ONLY is required for fuel damage during handling.

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

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 2-AOI-79-1, 2-AOI-79-2 (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 7/19/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X Comments: To answer this question without reference requires the candidate to analyze the intent of the two potential AOls. Specifically, what actions are required to correct or mitigate the problem. Since the actions to correct a damaged fuel bundle during handling require personnel to operate refueling equipment located on the Refuel Floor, evacuation of ALL personnel is not possible if the procedure is to be completed. The actions required to control or mitigate an inadvertent criticality are carried out from the control room, therefore no personnel are necessary on the Refuel Floor and evacuating ALL personnel is appropriate.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Inadvertent Criticality During Incore .2-AOI-19-2 Unit 2 Fuel Movements Rev. 0013 Page 3 of8 1.0 PURPOSE This instruction provides the symptoms, automatic actions and operator actions for an inadvertent criticality during incore fuel movements.

2.0 SYMPTOMS A. Possi.ble annunciators in alarm:

1. CONTROL ROD WITHDRAWAL BLOCK (2-XA-S5-5A, W~ndow n

2. SRIM HIGHffNOP (2-XA-5S-5A, Window 13).

3. SRIM PERIOD (2-XA-S5-SA, Window 20).

4. REACTOR CHANNEL A AUTO SCRAM (2-XA-55-SB, Window n S. REACTOR CHANNEL B AUTO SCRAM (2-XA-55-SB, Window 2}-

6. FUEL POOL FLOOR AREA RADIATION HIGH (2-XA-5S-3A, Window 1).

1. REACTOR ZONE EXHAUST RADIATION HIGH (2-XA-55-3A, Window 21).

8. RX BLDG AREA RADIATION HIGH (2-XA-55-3A, Window 22).

9. REFUEUNG ZONE EXHAUST RADIATION HIGH (2-XA-55-3A, Window 34).

10. REACTOR CHANNEL A MAN SCRAM (2-XA-B5-5B., Window 8)..

11. REACTOR CHANNEL B MAN SCRAM (2-XA~55-SB., Window 9).

B. SRM period lights illuminated.

C. Rising count rate on SRIM meters.

D. Rising power level onlRM recorders.

E. Rising radiation level on refuel floor.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Fuel Damage During Refueling 2-AOJ-19-1 Unit 2 Rev. 0011 1"*1Ige 3 of 1 1.0 PURPOSE This instruction provides the symptoms, automatic actions and operator actions for a fuel damage accident 2.0 SYMPTOMS A. Possible annundators in alarm:

1. FUEL POOL FLOOR AREA RADIATION HIGH (2-XA-55-3A, window 1).

2. AIR PARTICULATE MONITOR RADIATION HIGH (2-XA-55-3A, window 2).

3. RX BLDG, TURB BLDG, RF ZONE EXH RADIATION HIGH (2-XA-55-3A, window 4).

4. REACTOR ZONE EXHAUST RADIATION HIGH (2-XA-55-3AT window 21).

5. RX BLDG AREA RADIATION HIGH (2-XA-55-3A, window 22).

6. REFUELING ZONE EXHAUST RADIATION HIGH (2-XA-55-3A, window 34).

B. Gas bubbles visible, in the Spent Fuel Storage Pool and/or Reactor Cavity, attributed to physical fuel damage.

C. Known dropped or physically damaged fuel bundle.

D. Portable CAM in alarm.

E. Radiation level on the Refuel Floor is greater than 25 mrlhr and cause is unknown.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet SFN Inadvertent Criticality During Incore .2-AOI-79-.2 Unit 2 Fuel Movements R.ev.OO13 Page50f8 4.0 OPERATOR ACTIONS 4.1 Immediate Actions

[1] IF unexpected criticality is observed following control rod withdrawal, THEN REINSERT 'the control rod. D

[2] IF all control rods CANNOT be fully inserted, THEN MANUALlY SCRAM the reactor. D

[31 IF unexpected criticality is observed following the insertion of a fuel assembly, THEN PERFORM the following: D

[3.1] VERIFY fue! grapple latched onto 1he fuel assembily handle AND immediately REMOVE the fuel assembly from the reactor core. D

[3.2] IF the reactor can be determined to be subcritical AND no radiological hazard is apparent, THEN PLACE the fuel assembly in a spent fuel storage pool location with the least possible number of surrounding fuel assemblies, leaving the fuel grapple latched to the fuel assembly handle. D

[3.3] IF the reactor CANNOT be determined to be subcritical OR adverse radiologkal conditions exist, THEN TRAVER.SE 'the refueling bridge and fuell assembly away from 1he reactor core, preferably to the area of the catOe chute, AND CONTINUE at Step 4 .. 1[4]. D

[4] IF the reactor CANNOT be determined to be subcritical OR adverse radioJogical conditions exist, TH EN EVACUATE the refuel floor. D

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BFN Fuet Damage Durin 9 Refueling 2-AOI-19-1 Unit 2 Rev. 0017 PageS ofl 4.0 OPERA TOR ACnONS 4.1 Immediate Actions

[1] STOP an fuel handling. o

[2] EVACUATE an non-essentiat personnel from Refuel Floor. o 4.2 Subsequent Actions CAUTION The release of lad ine is of major concern. If gas bubbles are identifi*ed at any time, Iodine release should be assumed until RADCON determines othenvise.

[1] VERIFY secondary containment is intact

(.REFER TO Tech Spec 3.6.4.1) o

[2] IF any EOI entry condition is .met, THEN ENTER the appropriate E01(s). o

[3] VERIFY automatic actions. o

[4] NOnFY RADCON to perform the following:

EVALUATE the radiation levels. 0

MAKE recommendation for personnel access. 0

MONITOR around the Reactor Bui~ding Equipment Hatch, at levels below the Refuel FkJor, for possible spread of the release. 0

[5] REFER TO EPIP-1 for proper notification. o

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO G2.2.39 Tier # 3 Knowledge of less than one hour technical specification action Group #

statements for systems.

KIA # G2.2.39 Importance Rating 3.9 4.5

[ Proposed Question: SRO # 21 Given the following plant conditions:

Unit 2 is at 100% rated power.

RHR Loop II is INOP 2 days into a 7 day action statement per TSR 3.S.1.A.

Which ONE of the following equipment failures would result in the most limiting Technical Specification LCO and the reporting requirements that would result from that failure?

If (1) was declared INOPERABLE, a (2) report to the NRC would be required.

)

(1) (2)

A. Core Spray Loop I 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B. Core Spray Loop I 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

c. Diesel Generator A 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months D. Diesel Generator A 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: A Explanation: a. Correct answer. Two low pressure ECCS subsystems INOP requires entry into Applicability Statement 3.0.3 immediately in accordance with 3.S.1.H.

b. Part (1) is correct. Part (2) is incorrect. A shutdown required by Technical Specification is no longer a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months report. Reporting requirements have been revised such that a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months report is required for this condition.

c. Part (1) is incorrect. An INOP DIG will result in implementation of Applicability Statement 3.0.3, but there is a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months grace period before redundant subsystems, trains and components are declared inoperable in accordance with 3.8.1.B.2. Part (2) is correct.

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): U1 TSR 3.5.1, U1 TSR 3.8.1 (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 7/23/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet ECGS - Operating 3_5_1 35 EMERGENCY CORE COOLING SYSTEMS (ECGS) AND REACTOR CORE ISOLATION COOLING (RGIG) SYSTEM 3.5_1 EGCS - Operating LCO 3_5_1 Each ECCS injecfiontspray subsystem and the Automatic Depressurization System (ADS) function of six safety/rehef valves shall be OPERABLE.

APPlICAB;ILlTY: MODE 1, MODES 2 and 3, except hfgh pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure ::;;150 psilg.

ACTIONS

NOTE-------------------------------------------

LCO 3_0A.b is not applicabte to HPCt CONDITI'ON REQUIRED ACTION COMPLETION TIME A. One low pressure ECCS I A.1 Restore low pressure 1 days injection/spray subsystem ECCS injectionJspray inoperable_ subsystem(s) to OPERABLE status.

OR One fow pressure coolant injection (LPCI) pump tn both LPGI subsystems inoperable_

(continued)

BFN-UNIT1 3.5-1 Amendment No. 234, 24Q, 249 December 1, 2003

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet EGGS - Operating 3.5.1 CONDITION REQUIRED ACTION COMPLETION TIME G. Two or more ADS valves G.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months inoperable.

AND

-OR G.2 Reduce reactor steam 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Required Action and dome pressure to associated Completion :5 150 pslg.

Time ofCondihon G, D, E, or F not met H. Two or more IO'N pressure H.1 Enter lCO 3.0.3. Immediately ECCS injediontspra~'

subsystems inoperable for reasons other than Condition A.

OH HPCI System and one or more ADS var. . .es inoperable.

BfN-UNIT 1 35-3 Amendment No.~ 240 March 12,2001

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet AC Sources - Operating.

3.R1 3.8 ELECTRilCAL POWER SYSTEMS 3.8.1 AC Sources - Operating LCO 3.8.1 The foltowtng AC electrical power sources shall be OPERABLE:

a. Two qualified circuits between the offsile transmission network and the onsite Class 1E AC Electrical Power Distribuhon System;

b. Unit*l and 2 diesel generators (DGs) with two divisions of 480 V load shed logic and common accident signal logic OPERABLE; and

c. Unit 3 DG(s) capable of supplying the Unit 3 4.16 kV shutdown board(s) required by LCO 3.8.7, "Distribution Systems -

Operating ."

APPLICABILITY: MODES 1,2, and 3.

ACTIONS

N 0 TE ----------------------------------------------

LCO 3.0A.b is not applicable to DGs.

CONDITION REQUIRED ACTION COMPLETION TIME A. One required offsite A.1 Verity power availability 1 hoor circuit inoperable. from the remaining OPERABLE offsite AND transmission network.

Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND

{ continued}

BFN-UNIT 1 3.8-1 Amendment No. ~ 249 December 1, 2003

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet AC Sources - Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 Declare req uired 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from feature(s) with no offsite discovery of no power available offsite power to inoperable when the one shutdown redundant required board concurrent feature(s) are inoperable. with inoperabiJity of redundant required feature( s )

AND A.3 Restore required offsite 7 days circuit to OPERABLE status. AND 14 days from discovery of failure to meet LCO B. One required Unit 1 and 2 B.1 Verify power availability 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months DG inoperable. from the offsite transmission network. AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND (continued)

BfN-UNIT1 3.8-2 Amendment No. 234

AC Sources - Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 Declare required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from feature(s), supported by discovery of the inoperable Unit 1 and Condition B 2 DG, inoperable when concurrent with the redundant required inoperability of feature(s) are inoperable. redundant required feature(s)

AND B.3.1 Determine OPERABLE 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Unit 1 and 2 DG(s) are not inoperable due to common cause failure.

) OR B.3.2 Perform SR 3.8.1.1 for I 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE Unit 1 and 2 DG(s).

AND B.4 Restore Unit 1 and 2 DG I 7 days to OPERABLE status.

AND 14 days from discovery of failure to meet LCO (continued)

BFN-UNIT 1 3.8-3 Amendment No. 234

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet NPG Standard Regulatory Reporting Requirements SPP-J.5 Programs and Rev. 0018 Processes Page 18 of 64 Appendix A (Page 2' of 11) 3.0 REQUIREMENTS NOTES 1} Internal management notification requirements for plant events are found in Appendix D. Operations and the Plant Manager (or Duty Plant Manager) are responsible for making these internal management notmcatiom*.

2) NRC NUREG-1022, Supp.ements and subsequent revisions Shoukl be used as guidance for determin!ing reportability of ptant events pursuant to §50.72 and §50.73.

3.1 Immediate 'Notification - NRC TVA is required by §50.72 to notify NRC mmedlatel'y if certain types of events occur. This appendix contains the types of events and ttle allotted time in which NRC must be ootified.

(Refer to Form SPP-3.5-1). Operations is responsible for making ttle reportabi!ity determinations Tor §50.72 and §5D.73 reports. Operations is responsible for making ttle immedlate notification to NRC ,in accordan ce with §50.72.

Notmcation is via the Emergency Notification System. If the Emergency Notification System is not operative, use a telephone, telegraph., mailgram, or facsLmile.

NOTE The NRC Event Notification Worksheet may be used in preparing for notifying the NRC.

,A.. The Immed.iate Notification Cmeria of §50]2 is divided into 1-hour,4-hour, and 8- hour phone calls. Notify ttle NRC Operations Center within the applicable lime limit for any item which is kJentified in the Immediate Notification Cmeria.

B. The fOllowing cmeria req,uire 1-hour notiftcation:

i. (leclmical Specit1caOOns)- Safety limits as defined by the Technical Specifications which. have been violated.

2. §50.72 {a){ 1Xi) - The declaration of any of the Emergencyelasses specified in ttle licensee's approved Emergency PlarL NOTE If it is discovered that a condition existed whichi met the Emergency Plan cmeria but no emergency was dedaredand the basis for the emergency class 00 longer exists at the time of discovery, an ENS notification (and notification of the Operations Duty Specialist), within one hoor of discovery of the undeclared (or misclassifld) event, shaH be made . However, actual declaration of the emergency class is not necessary in ttlese circumstances.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet NPG Standard Regulatory Reporting Requirements SPP-3.5 Programs and Rev. 0018 Processes Page 19 of 64 Appendix A tpage30fl1) 3.1 Immediate Notification - NRC (continued)

3. §51172(b).{1)) - Any deviaUon from the planfs Technical Specifications autllOfized pursuant to §50.54(x).

c. The following criteria require 4-hO'ur notification:

1. §5Q.J2(b)(2)(i) - The initiation cfany nI.ic!ear pJanhtmtdown required by the plant's Technical Specifications.

2. §5o..72(b)(2){iv)(A) - Any event th,at results or should have resulted in Emergency Core Cooling System {ECeS) discharge into the reactor coolant system as a result of a valid signal except when the actuation results from and is part of a pre-planned sequence during testing Dr reactor operation.

3. §50 . 72(b)(2)(t>.r)(B) - Any event or condioon that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

4. §5(l72(b)(2){xt) - Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notmcation to other government agencies has been or will be made. Such an event may include an onsite fatality orinoovertent release of radioactive contaminated materiais.

0. The fOllowing cmena req'uire 8-hour notification:

NOTE The non-emergency events specified below are only reportable if they occurred within three years of the date of discovery.

1. §50.72(b)(3}(H)(A) - Any event or condition that results in the condition olthe nuclear power plant, including its principal safety bamers, tJeing seriously degraded.

2. §5R72(b)(3)(if)(8) - Any event cr condition that results in the nuclear power plant being [nan unanalyzed condition that significantly degrades p1ani safety.

3. §50.J2(b)(3}(iv)(A) - Any event or condition that results* in valid actuation of any of the systems listed in paragraph (b)(3}(iv}(8) [see list belOW], ex.cept When the actuation results from and is part O'f a pre-piannedsequence during testing or reador operation.

a. Reactor protection system (RPS) including: Reactor scram anu reactor trip.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO G2.2.43 Tier # 3 Knowledge of the process used to track inoperable alarms.

Group #

KIA # G2.2.43 Importance Rating 3.0 3.3

[ Proposed Question: SRO #22 Given the following plant conditions:

Unit-2 is at 100% rated power

HPCI OIL FILTER DIFF PRESS HIGH 2-PDA-73-53 (9-3F W34) has alarmed.

Investigation revealed that L1P transmitter 2-PDS-073-0053 has failed and requires replacement.

A Work Order was written on 10/15/2008 to disable annunciator (9-3F W34), replace the transmitter when available, then re-enable the annunciator once post-maintenance testing is completed.

The transmitter has been ordered and is expected to arrive 01/15/2009.

Which ONE of the following describes the required actions per OPDP-4, "Annunciator Disablement?"

A 10CFR50.59 evaluation (1) required prior to disabling the annunciator and this action must be AUDITED every (2)

(1) (2)

A. is NOT 12-hour shift.

B. is NOT 30 days.

c. is 12-hour shift.

D. is 30 days.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: D Explanation: a. Part (1) is incorrect. Per OPDP-4 Appendix A, a 50.59 review is required since the annunciator will be disabled for> 90 days. Part (2) is incorrect.

The Disabled Annunciator list is REVIEWED every shift, but is AUDITED every 30 days to ensure a 50.59 review is still not required.

b. Part (1) is incorrect. A 50.59 review is required. Part (2) is correct. The Disabled Annunciator list is audited to ensure 50.59 compliance every 30 days.

c. Part (1) is correct. Per OPDP-4 Appendix A, a 50.59 review is required since the annunciator will be disabled for> 90 days. Part (2) is incorrect. The Disabled Annunciator list is REVIEWED every shift, but is AUDITED every 30 days to ensure a 50.59 review is still not required.

d. Correct answer.Part (1) is incorrect as stated in (c) above. Part (2) is correct. The Disabled Annunciator list is audited to ensure 50.59 compliance every 30 days.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): OPDP-4, U2 TSR 3.3.5.1 (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 7/27/2008 RMS Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X Comments:

ES-401 Sample Written Examination Form ES-401*5 Question Worksheet BFN Panel 9-3 2-ARP-9-Jf Unit 2 2-XA-55-3F Rev. 0024 38 of 39 SenscrlTrip Point HPCIOll 2-PDS-013-0053 10psid FitTER IDIFF PRESS HIGH 2-PIDA-73-53 f34 (Paget of 1)

Sensor HPCI TUIItline, EI 519', SE oorner RX Bldg.

location:

Probable The inservice filter is dirty.

Cause:

Automatic None Action:

Operator A. OISPATCH personnel to switch to dean filter. o Action: B. INITIATE WO for dirty filter to be charged. o

References:

2~5E62{}-1 2-47E61 0-73-2 GE 130E928-4

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet TVAH STANDARD OPDP-4 DEPARTMENT ANiNONIClATOR DlSABLEMENT Rev. 2 PROCEDURE PagellDof 18 APPEttDIXA Page1of2 TECHNrCAL EVALUATION ANiD 50.59 AJPPUCABtUTY

1. When an annunciator windowflllpll1 is disabled as directed/allowed in an approved plant prooedure (excluding maintenance or surveillance adMHes), a separnte 50.59 review and Technical Evaluation are not required since the procedUre has already been reviewed and approved. The following example wool!:;!, bean alann disabiement per an approved plant iml!lrudioo:

A system operating inslrucOOn direcIB or allows an alarm disablement due to abnormal conditions which are addressed (and restored) by !hat instruction.

NOTE The initiation and processing of a work order OOesNOT constitute in..process mail'ltenaince. Refer to Section 5.0 Definitions.

2. If an annunciator windowiirlput is disabied in support of maintenance or surveillance adivities, a SO.59 review is nat required UNLESS the annunciator willi remain disabled far more than 90 days. If 90 days will ibe exceeded, a SO.59 review shall be completed pOOr to exceeding 90 days. A Teehnical Evaluation is required pfiorto disabiement if alarm functions* will be di_baed for equipment remaining in service (nat removed from service/inoperabie for the maintenance activity).

The following example wooki be oonsidered necessary to support maintenance aotMtiies and requires a Technical Evaluation:

A pump is tagged with a dearance for maintenance. lis suction pressure switch will be depressurized and disabling the associated low pressure alarm will disable !he aiarm function for other equipmentfhat must remain in service.

The fmlDwing examples 1NOUid be considered necessary to support maintenance adivi!lies and do not require II Technical Evaluation provided Ine parameter is the only input to the alann:

A pump is tagged with a dearance far maintenance. lis suction pressure switch will be depresSJUrized and the associated low pressure al8Im cIIisabIed.

An instrument is declared' inoperable, and any required leO adion(sl are entered far calibration in accordance wi!h an approved maintenance instruction. The alann from this instrument is disab&ed.

)

/

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet TVA'NSTANDARD OPDP-4 DEPARTMENT ANINUNClATOR DISABLEMENT Rev. 2 PROCEDURE Page 1I1*of 18 APPENiDIXA Page2.of2

3. When an annullCiator windowlinput must. be disSb!ed due to degqded or inoperable equipment with maintenance NOT in progress, 8 50.59 review is required prior to disabling the 8am EXCEPT when oovered by an approved piant procecllJre (liem 1). A Technical Evaluation is ~ required EXCEPT when oovered by an approved piant prooedure (nem 1) OR when ihe affected alarm function is only monitoring equipment which is inopembiefoot-of-service and !he alarm will be restored pricY to decl8fing the affected equipment operable 011' returning it to service. The following excerpt: from NEI 96-07 is an example of a degraded condition affecting multiple .rm inputs:

A lewl transmitter for one Reactor Coolant Pump (RCP) lower oil reservoir failed while at. power. The trarmmitter 'provides an alarm function, but not an automatic protective action function. The transmitter and associated alarm are described in !he UFSAR as protective features fOr the RCPs, but rID technical specification applies,. loss of Ine transmitter does not resuH in the loss of operability for any technical specification equipment The transmitter fails in a direclion resurtinq in a conmuousalarm inlhe controlTOOm. The alarm circuitry provides a common alarm for both the upper and lower oil reservoir circuits, so nnsmHter failure causes a hanging alarm and a masking of proper operalion of the remaining functional transmiIrer. PrecalJlionary measures are taken to monitor lower reservoir oil level as ouiIined in the alarm manual!

using available alternate means. An interim coolpensatOll'y action is proposed to lift the leads (temporary chenge) from the failed nnsmHterlo restore ihe alarm 1'tinclion for the remaining rullClioning iransmitter. Lilting ihe leads is a compensatory action (tempormy change., that is SlIbjectto 10 CFR 50.59. The 10 CFR SO.S9 screening woukl be appiied to the tempormy change itself (Wilted leads), notlhe degraded condition (failed transmitter) to determine its impact on oHler aspects of the facility described in the UfSAR If screening determines that no other UFSAR-described SSCS would be affectedby' this compensatory action, the temporary change wookl screen out, i.e., not require a 10. CFR SO.59 evaduation.

4. If an annunciator window Of input must be disabled for other reasons (e.g. due to actual plant parameters which are knownlsuspected to be at or exceeding the al!arm setpoint),

then a 50:59 review and Technical Evaduation are required prior to disabtinq the alarm, EXCEPT when covered by an approved plant procedure (nem 1).

)

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

}

TVAN STANDARD OPDP-4 D'EPARTMENT AN.NIJNCIATOR EllSABlEMENT Rev. 2 PROCEDURE PageS of 18 Employees Performing A.ctivities That witl Bring in Alarms A. Provides list of affected alarms to the Unil Operator before starting work.

B. IF the 8Il1nunciator is out of service due to mainten8ll1ce oroiher abnormal condition, THEN Verify WO written which identifiesollt..of-service annunciator.

C. Ensure out-of-servlce indicator is pk:aced on each applicable anmmciator window.

Employees Performing Activities That will Bring in Alarms

[L When maintenance or surveillance activities are complete, then notify OperatioF!s to remove out..of-service indicator fromaffectedannunciaror windows.

Unit Operator/~nee E. Remove out..of-se:rvice indicator 011 alarm windows which have been enabfed, UNLESS there are other inputs to the afte.cted window which remain disabled by OPDP-4 or aoother approved .plant procedure.

3.5 Review and Audit The Ditsabled Anm.H")ciator Book is reviewed during shift llImover (OPDP-1)to ensure disabled alarms are documerned as required.

On a monthly basis, the Disabled Annunciator Book should be audited Io verify that 10CFR50.59 re ....iews have beeF! compteted as required. A mCFRSO.59 review shall be completed for any annunciators disabled for malmenance which wil! exceed 90 days prior to the next review.

4.0 RECORDS 4.1 QA-Records Disabled Alarm Checklist OPDP-4-1 Annunciator Disablement Technical E....aluation OPDP-4-5 4.2 Non-QA Records Disable Alarm Index Sheet OPDP-4-2 Disabted Alarm Com.pensatory Measures OPDP-4-3

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO G2.3.13 Tier# 3 Knowledge of radiological safety procedures pertaining to licensed Group #

operator duties.

KIA # G2.3.13 Importance Rating 3.4 3.8

[ Proposed Question: SRO # 23 Valve line-up checklists are being conducted on Unit 2 and the initial positioning has been completed for a valve located in the drywell.

Independent verification (IV) of the valve should only take 15 minutes, but Radcon reports that the general area dose rate is now 150 mrem/hr.

Which ONE of the following describes the circumstances, if any, that allow the Shift Manager to waive the IV requirement for this valve?

A. Independent verification may be waived by the Shift Manager ONLY if alternate means of verification are available.

B. Independent verification can not be waived; direct Radcon to determine other means of dose reduction.

C. Independent verification may be waived by the Shift Manager ONLY if the system is not an ECCS system.

D. Independent verification may be waived by the Shift Manager solely based on ALARA concerns.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: 0 Explanation: a. Incorrect. Alternate verification methods may be used if available, but are not a prerequisite for waiving IV requirements due to the excessive radiation exposure given in the stem. If the word "only" is removed from the distracter, it would be correct.

b. Incorrect. Being located inside the Drywell implies this valve meets the critical activity requirements of SPP 10.3 Step 3.4.3.A, and is "absolutely necessary for SSCs to function." Even so, excessive radiation exposure to the extent given in the stem warrants waiving IV requirements. If the anticipated dose was closer to the limit, the actions to have Radcon take actions to reduce the dose rate may have been appropriate.

c. Incorrect. According to SPP 10.3, the Shift Manager may authorize deviations from the requirements if needed If the word "only" is removed from the distracter, it would be correct.

d. Correct answer.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): SPP 10.3 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Question Source: Bank # G2.3.2 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 55.43 X Comments:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet WAN STANDARD SPP-10.3 PROGRAMS AND VERIFICATION PROGRAM Rev. 1 PROCESSES Page 50f16 Procedure Preparers

c. The pre parers of site proceduresJinstrucoons and WOfk documents are responsible for the foflowing:

1. Ensuring that rv' or secooo-party vertficaoon requirements are specified as appropriate.

2. Ensuring file type of verification is cieal'tyiden1:ified.

[t stlm Manager The shift manager (SM) shall be responsible for the follofNing:

i. Determining the corrective actions to be taken when discrepancies are discovered.

2. Ensuring that personnel assigned to perform IV and second-party verification are qualified.

3. Authorizing deviations from normal verification practices if needed.

E. Trajning Manager De'tl1elop, conduct, and document training of personnel engaged in ....erification activities.

F. All Personnel Infoml their respedive foreman orsuper'o'lSOf jf tl1ey have beenass~ned a verification which they do not feel qualified to perform. In the event tl1eir respecti'tl1e supervisor is not available, they will contact the SM for resolution before continuing the vermcatioll.

3.2 Qualrfications Indi....iduals assigned I'll or second-party vermcation responsibififies shall meet the following qualification reqUirements.:

A. Technically Qualified to perform file assigned task (experience, position description, familiarity with the taSK, etc., should be considered j as determrned by file responsible manager.

B. Completed training on verlfication program requirements.

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

)

WAN STANDARD 5"-1'0.3 PROGRAMS AND VERIFICATION PROGRAM Rev. 1 PROCESSES Page8of16 3.3.2 Altemat~ V~rification Techniques Alternate verification techniques maybe used by:the verifier where specified by approved procedures, vaWe and !Jreaar line-tlp cl1edtlists, or at the discretion of $hift supervisory penmnneL Exampies include :the following:

A. Use of remote position indicators. (lndicaling lights in :the control room, at Dle switchgear, or allocal eootmls are Dle oormalmelhod of determining motor..q>erated and air-operaled vmve position.)

B. Use of process parameters (e.g., pressure, 1krN, vibration, eurrenl~

voltage, potentia! lamps, etc.).

C. Observation of Dle valve stem to aid in determination of valve position if the valve stem is marked by paint (when fully ,closed) or other positive verification me1hods.

D. Aulttorized scribe marks on valve stems, property labeled willl the throtled position.

E.. FunctiOllld mechanical position indicators.

F. A post maintenanceimodificalion functional test provided Dle testing verifies each component under consideration.

3.3.3 Circuit Breakers Circuit breaker verifICation will include a focal inspection of the breaker, control power sVlilx:hes or fuses, and other equipment as ooIlined below:

A. To verify a bleaker is removed from service, the independent or second-party verifier will ensure control power is isolated (it required) by inspecting appropriate switches, fuses or fuse blocks, and ensure the breaker is I'Kkedoot to the disconnected position, as applicable.

B. To verify a breaker is restored to service, the independent or second-party verifier will ensure eootrol power is energized by inspecting appropriate switches, indicating lights, fuses or fuse btocks, and will ensure tna Neaker is fully racked in wittl closing springs charged as appjieabre. Where practical,the end de\lice should be operated following the reinstallation of iii breaker. The verifier will also ensure fhe cubicle door is in good condition with aU fasteners tight.

3.4 Verificatioo Requirements When determination of these requirements is not dear, fhe responsibre manager witl designa1e :the requirements. If there is disagreement, Ina operations manager will deSignate the requirements.

3.4.1 f\I or second-PBrtY verification is required for those sysiems lis1ed in Appendix A and sl1a11 include :thefo&lrNiing as a minimum:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet TVAN STANDARD SPP40.3 PROGRAMS AND VERIFICAnON PROGRAM Rev. 1 PROCESSES Page 9 of16 A. All valves, breakers, and other components in safety"related systems where an inappropriate positioning could adversay affect systemtp4ant operation or containment integrity.

B. AI! valves, breakers, and other components in frreprotection system major flow paths, including fire fighting water supp$y and storage,carbon dioxide storage systems, fire protection systems, and all components necessary for the system to function and supply extinguishing media to tne fire.

C. Atl valves, breakers, and other components ingaseOilJS and liquid radioactive waste handling and processing systems where an inappropriate positioning couid result in radioactive material release to the environment 3.4,2 Activities Exempt From Independent and Second-Party Verification Regui:rements A. Calcullatioosperformed by quaUfied computersofiware.

B. ,A,ctivities for which veriffcations would be required and one or more of the following cOlliditions exrst

out-of-servfce systemstcnannelsJcomponenm for whfclri configuration contro.l wiil nQtbe maintained and will be verified to be in the proper configuration during the return to operable status.

Activities involving significant radiation exposure. As a guideline, an exposure greater than 10 mrem TEDE to perform the verification WQuld be considered excessive.

Activities 'Occurring during emergencycOIliditions (imminent danger to plant or persoonel) requiring rapid personnel action.

,A,ctivities that coutd jeopardize personnel safety.

Components ilocated within lockedlcoveredfcontrolled alCcess areas provided access to the area has not occurred since Ule last. documented verification.

For these instances, the decision not to perfom1 a verification is to be documented 'On the procedureiinsIructioo or work document 3.4.,3 Independent Verification Requirements

!Vis used to confirm that an activity or condition has been implemented in conformance with specified requirements.. The individual performing the lV must physically check the conditloo without relying 00 observation 'Or verbal confirmation by !he inilial performer. However ,the independent vefifier may be invotved in unl'el!ated portions of the sameactivily. IV is required for the fQnomng:

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet TVAN STANDARD SPP-1'O.3 PROGRAMS AND VERIFICATION P'ROGRAM Rev. 1 PROCESSES Page 100f16 A. Any critical acrnmy that, if done improperly, could remain undetected until that structure, system, or component was called upon to mitigate an accident or llninsient. asdescnbedin the FSAR, Fire Protection Plan, Securtty Plan,or ODCM. Crilicai impllies the actwttyis absolutely necessary for Syslems, Structures, and Compollents ID function.

B. Initial system Yineups, or resIDling components to their required posiliontcondition foUowing an outage where the system status was not maintained.

C. Normal system lille-up periodic check.s conducted during operating conditions. In this case, the individual perform!ngthe periodic check of the original lineup is considered to be the independent verifier and an additional secolld chedis not required. IV of locked components consists of checking fuat required locking devices are present and intact.

D. Installation and removal of temporary alterations covered by the IACF Program.

3.4.4 Second-Party Verification Requirements Second-party verification is used in lieu of I'll for the activities listed below.

When pertorming a second-party verification, an agreement must be reached between the pedormer and tlhe verifierlhat the 3ctivity/manipulation to be performed is correct before performance.

A. Activities where performing an IV 'WOuld by itseM" invalidate the actions or conditions the performer is atiempnng to estabtish.

EXAMPLE Verification of throttled valve position, locked valve position, installation and removal of high voltage !ineor bus PI fuses, installation and removal of fuses in fuse bbocksJciips wlhich are normakly hidden from view, etc.

B. Activities which, if improperly accomplished or incorredy identified, may cause any oHhe following:

ImmedJate plant trip or transient

Safety system actuation

Start of equipment

Equipment failure/damage

Release of raruoaclive material

Personnel injury EXAMPLE Removai or installation of wires, jumpers, or otlher connections; valve, switch, or .breaker manipulations; removal or installation of fuses or circuit cards; etc.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO G2.4.43 Tier # 3 Knowledge of emergency communications systems and techniques.

Group #

KIA # G2.4.43 Importance Rating 3.2 3.8

[ Proposed Question: SRO #24 Given the following plant conditions:

A transient has occurred on Unit 1 resulting in the following annunciators in alarm:

- STACK GAS RADIATION HI (1-RA-90-147B)

- STACK GAS RADIATION HIGH-HIGH (1-RA-90-147A)

- OG PRETREATMENT RADIATION HIGH (1-RA-90-157A)

- RX BLDG,TURB BLDG, RF ZONE EXH RADIATION HIGH (1-RA-90-250A)

Which ONE of the following describes the required operator action?

Declare a/an (1) . Direct a (2) to implement Appendix B, "Unit Operator Notifications. "

REFERENCE PROVIDED A. Notification of Unusual Event Unit 1 Unit Operator B Alert Unit 1 Unit Operator C. Notification of Unusual Event Unit 2/3 Unit Operator D. Alert Unit 2/3 Unit Operator

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet I Proposed Answer: C Explanation: a. Part (1) is correct. The Off-Gas pretreatment radiation high is indicative of a NOUE. However, since the transient has occurred on Unit-1, the normal assignment of the Unit-1 Unit Operator to implement Appendix B is not appropriate. EPIP-2 has a NOTE which allows delegation of that action to a Unit Operator on an unaffected unit.

b. Part (1) is incorrect. Conditions do not yet indicate a severity which justifies an ALERT emergency. Part (2) is incorrect as stated in (a) above.

c. Correct answer.

d. Part (1) is incorrect. Conditions do not yet indicate a severity which justifies an ALERT emergency. Part (2) is correct.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): EPIP-1, EPIP-2 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: EPIP-1 (Section 1 - 4 ONLY without bases)

Question Source: Bank #

Modified Bank # SRO 271000G2.4.36 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 55.43 X Comments:

)

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet SRO 271000G2.4.36 Given the following plant conditions:

A transient has occurred on Unit 1 resulting in the following annunciators in alarm:

- STACK GAS RADIATION HI (1-RA-90-147B)

- STACK GAS RADIATION HIGH-HIGH (1-RA-90-147A)

- OG PRETREATMENT RADIATION HIGH (1-RA-90-1S7A)

- RX BLDG,TURB BLDG, RF ZONE EXH RADIATION HIGH (1-RA-90-2S0A)

Which ONE of the following describes the required operator action?

Declare a/an (1) . Fifteen minutes later you determine you must notify (2) to implement EPIP-13, for dose assessment.

REFERENCE PROVIDED (1) (2)

a. Notification of Unusual Event SED b Alert TSC

c. Notification of Unusual Event Radcon

d. Alert CECC

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICAnON MATRIX EPIP-1 Valid MAIN STEAM UNE RADIAl'jON HIGH-HIGH c*

alarm, RA-90-135C z c

OR ~

Valid OG PRETREATMENT RADIATION HIGH l.!:

alarm, RA-90-151A.

!r!

I....

Reader moderaklrl:emperaWre can maintained lilebN 21:zD F whenever Teclmical SpecificafiOOIS require Mode 4 conditions or during operations in Mode 5.. ::.-

m

~

) OPERATING CONDITION:

5!

-I m

i::a G) m OPERATING CONDITION:

Mode 1 or 2 or3 ~

~

I m I (')

PAGE 23 OF 206 REWSION43

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BROWNS FERRY NOTIFICATION OF UNUSUAL EVENT EPIP-2 3.3 Notificadoll of Site p.ersonnel NOTE Normally Appendix B,"Unit Operator Notifications",is conducted by a Unit 1, Unit Operator, Depending upon the affected unit this action may be delegated to a Unit Operator, on an unaffected unit 3.3.'1 PROVIDE. .. a Unit Operator with a compi;eted copy of Appendix A.

o AND DIRECT ... the Unit Operator to make ;personnel notifications per Appendix S, ~UNIT o

OPERATOR NOTIFICAT10NS".

CAUTION Ongoing or anticipated security events may present a danger to site personnel. Do not conduct the no*tffication of site personnel PA message during an ongoing Of anticipated securityevent. All pertFnent site personne' PA messages will be conducted per AOI-100-B for security e. . .ents.

3.3.2 CONDUCT a PI'ant PA announcement similar to the 0 following: (Dial 687 to obtaf,n the Plant PA)

Let me have your attention please.

This is (name) .

A Notifi cation of Unusual Event, Erne rgency Classification has been declared.

We are currently implementing EPIP-2.

3.3.3 NOTIFY the Plant Manager or designee of the Notification of Unusual Event.

o PAGE 4 Of 14 REvtSlON 28

(

I~yalaNY~

'YIH:I~YIN :I~N:lH:I:I:lH (

TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE EPIP-1 EMERGENCY CLASSIFICATION PROCEDURE REVISION 43 PREPARED BY: RANDY WALDREP PHONE: 2038 RESPONSIBLE ORGANIZATION: EMERGENCY PREPAREDNESS APPROVED BY: TONY ELMS DATE: 06/25/2008 EFFECTIVE DATE: 07/01/2008 LEVEL OF USE: REFERENCE USE QUALITY-RELATED

(

HISTORY OF REVISION/REVIEW REV. REVISED NO. PAGES REASON FOR CURRENT REVISION 42 17-19 IC-53 BFN EPIP-1 revision 42 adjusts the information that supports EAL 1.1-G2 , 1.2-21,27,33,35, G and 1.5-S for changes resulting from engineering calculations that support 94,99,105, Minimum RPV Flooding Pressures (MRFP), and Heat Capacity Temperature 114,115, Limits. Revisions to these calculations were conducted for EOI Program 117,118, Manual Revision 27 (U2C15). The revision to the EOI Program manual adjusts 120,128,132 the EAL supportive information that is in compliance with the REP.

EALs 2.3-A, 2.3-S1, 2.3-S2, 2.3-G1, 2.3-G2, 3.1-G, and 3.2-G were revised to adjust Unit 1 drywell radiation values to support the decision to not start Unit 1 at extended power up-rate (EPU). Calculations ND-N0090-930050 R11 and ND-N0090-930055 R12 support the conditions described above. The two calculations utilized to support the drywell radiation values are not a function of the Emergency Operating Instruction.

This revision does not affect, alter or change the basis supporting the BFN TVA's standard emergency classification and action level scheme. Although this change does modify data and information utilized by existing EALs, the criteria established by NUREG 1.101 Rev. 3 (NUMARC/NESP 007 Revision 2) concerning the development of emergency action levels are not modified or changed. Specific EALs utilize information/data maintained through the implementation/maintenance of the Emergency Operating Instruction (EOI) procedures as well as specific calculations thus establishing thresholds used as entry conditions for emergency classifications. As calculations are revised based upon reactor parameters such as in this case, fuel specifications, the EAL threhold information must also be revised. This revision neither increases nor decreases the effectiveness of the REP. This revision simply adjusts data necessary to maintain the accuracy of applicable Emergency Action Levels.

43 21,97,34, IC-54 Some pages were added which were intentionally left blank (and noted as so) 75,127, to accommodate appropriate double sided printing and filing in procedure 129,188, manuals.

189 EAL 1.2-A - Wording of EAL enhanced to clarify intent of EAL.

EAL 7.3-U - Wording revised to change "greater than" to "exceeds or is predicted to exceed". Additionally, the basis page for this EAL addressed the escalation to the Alert classification. This wording in the basis was also revised to change "greater than" to "exceeding or predicted to exceed".

EAL 7.3-A - Wording in first condition changed from "greater than" to "exceeds or is predicted to exceed". Wording in second bullet of second condition changed from "Affecting equipment required for safe shutdown" to "Equipment required for safe shutdown is affected."

Table 3.1 - Removed from Table 3.1 the maximum safe operating temperature limit value for Core Spray BID Pump Room High Humidity or Temp High specific for Unit 2 and Unit 3.

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPIP-1

(

TABLE OF CONTENTS TABLE OF CONTENTS .............................................................................................................................................1 SECTION I

1.0 INTRODUCTION

..................................................................................................................................................3 1.1 Purpose ................................................................................................................................................................3

2.0 REFERENCES

.....................................................................................................................................................3 2.1 Industry Documents ..............................................................................................................................................3 2.2 Plant Instructions ..................................................................................................................................................3 3.0 INSTRUCTIONS ...................................................................................................................................................4 3.1 Instructions ...........................................................................................................................................................4 4.0 GLOSSARY of ABBREVIATIONS, ACRONYMS, AND DEFINITIONS ............................................................. 7 5.0 EVENT CLASSIFICATION INDEX .................................................................................................................... 15 SECTION II EVENT CLASSIFICATION MATRiX........................................................................................................................17 1.0 Reactor ...............................................................................................................................................................17 2.0 Primary Containment ..........................................................................................................................................25 3.0 Secondary Containment ....................................................................................................................................33 4.0 Radioactivity Release .........................................................................................................................................39 5.0 Loss of Power .....................................................................................................................................................45 6.0 Hazards ..............................................................................................................................................................51 7.0 Natural Events ....................................................................................................................................................69 8.0 Emergency Director Judgment ...........................................................................................................................77 SECTION III BASiS .......................................................................................................................................................................87 1.0 Reactor ...............................................................................................................................................................87 2.0 Primary Containment ........................................................................................................................................107 3.0 Secondary Containment .................................................................................................................................. 125 4.0 Radioactivity Release ....................................................................................................................................... 134 5.0 Loss of Power ...................................................................................................................................................145 6.0 Hazards ............................................................................................................................................................155 7.0 Natural Events .................................................................................................................................................. 180 8.0 Emergency Director Judgment ......................................................................................................................... 187 PAGE 1 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPIP-1 THIS PAGE INTENTIONALLY BLANK PAGE 2 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPIP-1

1.0 INTRODUCTION

1.1 Purpose Provide guidance to the Shift Manager or Site Emergency Director (SED) for proper declaration and classification of emergencies and ensure emergency classifications are consistent with those used by state and local governments and the Nuclear Regulatory Commission (NRC).

The procedure applies to site events that constitute an emergency consistent with those site specific events outlined in NUMARC/NESP-007 August 1992.

The Shift Manager and the SED are the only persons authorized to make the emergency classification determination.

2.0 REFERENCES

2.1 Industry Documents A. NUREG-0654, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants" B. 10 CFR 50.47, Code of Federal Regulations C. Reg Guide 1.101 Rev. 3, "Methodology for Development of Emergency Action Levels 2.2 Plant Instructions A. TVA Radiological Emergency Plan B. EPIP - 2, "Notification of Unusual Event" C. EPIP - 3, "Alert" D. EPIP - 4, "Site Area Emergency" E. EPIP - 5, "General Emergency" F. EPIP-16, "Termination and Recovery Procedure" PAGE 3 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPIP-1 3.0 INSTRUCTION 3.1 Following plant events or transients review EPIP-1 Section II, 1.0 through 8.0 and determine if an event should be classified as an emergency.

NOTE

1. If an emergency action level for a higher classification was exceeded, but the present situation indicates a lower classification, the fact that the higher classification occurred shall be reported to the NRC and the CECC, if staffed, or ODS if the CECC is not staffed. The higher classification should not be declared.

2. If an emergency action level was met but the emergency has been totally resolved, the emergency class that was appropriate shall be reported to the ODS and the NRC but should not be declared.

3.1.1 EPIP-1 Section II, 1.0 through 8.0 captures events in eight major categories as listed on the event classification index.

3.1.2 Each emergency action level (EAL) in a category is given an alpha-numeric designator. The first numeric component of the EAL indicates the section followed by a numeric designator for the specific EAL within the section and an alpha numeric designator for the event class.

Example: 5.2-U These designators provide for cross-reference between the specific EAL and the basis document which provides technical supporting information for the EAL and may aid the Shift Manager/SED in classifying events.

Curves, notes, or tables that support the EAL are located on the face adjacent page within the matrix section of the procedure and are identified within the event classification window on the information bar that precedes the deSignator. The information bar contains the appropriate indication to alert the user that a corresponding curve, note, or table applies to the EAL.

Curves, notes, or tables that contain unit specific information will also be identified within the event classification window by the letter "US" located at the end of the EAL information bar. This information should alert the user that the corresponding curve, note, or table contains unit specific information.

Example I 5.2-U I CURVE I NOTE I TABLE I US PAGE 4 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPIP-1 3.2 If the event is determined to be one of the four emergency classifications, the Shift Manager assumes the responsibility of SED until relieved by the Plant Manager or designee.

3.2.1 Implement one of the following procedures as applicable:

EPIP-2 Notification of Unusual Event EPIP-3 Alert EPIP-4 Site Area Emergency EPIP-5 General Emergency 3.2.2 Continue to review the emergency conditions in the event classification matrix and escalate, terminate, or implement recovery as appropriate.

Refer to EPIP-16 for termination or recovery.

3.3 If the event is determined not to be one of the four event classifications, continue to monitor plant conditions for possible changes that could result in reaching an event classification.

LAST TEXT PAGE 5 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPIP-1

(

THIS PAGE INTENTIONALLY BLANK PAGE 6 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPIP-1 4.0 GLOSSARY of ABBREVIATIONS, ACRONYMS, AND DEFINITIONS The following is a list of terms and phrases found in EPIP-1. Each term or phrase is provided with a meaning, to ensure consistent use and understanding.

TERM/PHRASE MEANING/DEFINITION ADS Automatic Depressurization System AOI Abnormal Operating Instruction Alert Events are in process or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involve probably life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

ARI Alternate Rod Insertion ARM Area Radiation Monitor ARP Alarm Response Procedure ATWS Anticipated Transient Without Scram Auto Automatic Bomb An explosive device BWR Boiling Water Reactor Can/Cannot be The current value or status of an identified parameter relative to that determined specified in the instruction can/cannot be ascertained using all available indications (direct and indirect, singly or in combination).

Can/Cannot be The value of the identified parameter(s) is/is not able to be kept Maintained above/below specified limits. This definition includes making an evaluation Above/Below that considers both current and future system performance in relation to the current value and trend of the parameter(s). "Cannot" does not imply that the actual value of the parameter must first exceed the specified limit.

Can/Cannot be The value of the identified parameter(s) is/is not able to be returned to Restored above/below specified limits within a reasonable time after having exceeded Above/Below the specified limits. This determination includes making an evaluation that considers both current and future system performance in relation to the current value and trend of the parameter(s).

PAGE 7 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPIP-1 (

TERM/PHRASE MEANINGIDEFINITION CAD Containment Atmosphere Dilution CAS Central Alarm Station CDE Committed Dose Equivalent CECC Central Emergency Control Center Ci Curie Civil A group of 20 or more persons violently protesting station operations or Disturbance activities at the site.

Cubic Centimeters Confinement Spent Fuel Storage Cask CONFINEMENT BOUNDARY consisting of the Boundary MPC shell, bottom base plate, MPC lid (including the vent and drain port cover plates), MPC closure ring, and associated welds.

CS Core Spray deg Degrees DG Diesel Generator orywe II The upper portion of the Primary Containment which encloses the Reactor Pressure Vessel.

EAl Emergency Action level ECCS Emergency Core Cooling System ECl Effluent Concentration Limit EOI Emergency Operating Instruction EPA Environmental Protection Agency EPIP Emergency Plan Implementing Procedure EQ Environmental Qualification PAGE 8 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE

( INTRODUCTION EPIP-1 TERM/PHRASE MEANING/DEFINITION Event Assessment of an EVENT commences when recognition is made that one or more of the conditions associated with the event exists. Implicit in this definition is the need for timely assessment, i.e. within 15 minutes.

Explosion A rapid, violent, unconfined combustion or a catastrophic failure of pressurized equipment that imparts energy of sufficient force to potentially damage permanent structures required for safe operation.

F Fahrenheit Fire Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical components do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.

Flammable Gas Combustible gasses maintained at concentrations less than the lower explosive limit. Will not explode due to ignition.

GOI General Operating Instruction General Events are in process or have occurred which involve actual or imminent Emergency substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.

gm Gram HCTL Heat Capacity Temperature Limit Hostage A person(s) held as leverage against the station to ensure that demands will be met by the station.

Hostile Action An act toward a nuclear power plant or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.

Hostile Action should NOT be construed to be acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism based EALs should be used to address such activities, (e.g. violent acts between individuals in the owner controlled area).

PAGE 9 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPIP-1 TERM/PHRASE MEANING/DEFINITION Hostile Force One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

HPCI High Pressure Coolant Injection HR Hour IN Inches ISFSI Independent Spent Fuel Storage Installation KV Kilovolt Large framed A large aircraft with the potential for causing significant damage to the plant; aircraft may be referred to as an airliner.

LCO Limiting Condition for Operation LOCA Loss Of Coolant Accident LPCI Low Pressure Coolant Injection MRFP Minimum RPV Flooding Pressure MCUTL Maximum Core Uncovery Time Limit MIN Minute Modes of Mode Title Reactor Mode Avg. Reactor Coolant Operation Switch Position Temperature (oF) 1 Power Operation Run NA 2 Startup Refuel(a) or NA Startup/Hot Standby 3 Hot Shutdown(a) Shutdown > 212 4 Cold Shutdown(a) Shutdown < 212 5 Refueling(b) Shutdown or Refuel NA (a) All reactor vessel head closure bolts fully tensioned.

(b) One or more reactor vessel head closure bolts less that fully tensioned.

MPC Multi-Purpose Canister (part of ISFSI)

PAGE 10 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPIP-1 TERM/PHRASE MEANING/DEFINITION MPH Miles per Hour mrem Millirem MSIV Main Steam Isolation Valve MSL Main Steam Line MSRV Main Steam Relief Valve NESP National Environmental Studies Project Notification of Events are in process or have occurred which indicate a potential Unusual Event degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.

NUMARC Nuclear Management and Resources Council OCA Owner Controlled Area ODS Operations Duty Specialist 01 Operating Instruction OSC Operations Support Center PA Protected Area PAR Protective Action Recommendation PCIS Primary Containment Isolation System Primary The drywell, the vent system, and the suppression chamber.

Containment Primary System Primary systems comprise the pipes, valves and other equipment connected to the RPV such that a reduction in RPV pressure will affect a decrease in the flow of steam or water being discharged through an unisolable break in the system.

PAGE 11 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPIP-1 TERM/PHRASE MEANING/DEFINITION Projectile An object ejected, thrown, or launched towards a plant structure. The source of a projectile may be offsite or onsite. Damage is sufficient to cause concern regarding the integrity of the affected structure or the operability or reliability of safety equipment contained therein.

Protected Area All areas within the security protected area fence.

PSIG Pounds Per Square Inch Gauge R Rad RCIC Reactor Core Isolation Cooling RCS Reactor Coolant System REP Radiological Emergency Plan RHR Residual Heat Removal RPS Reactor Protection System ('

RPV Reactor Pressure Vessel Sabotage Deliberate damage, misalignment, misoperation of plant equipment with the intent to render equipment inoperable.

SAMG Severe Accident Management Guideline 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 following:

Transient (1) Automatic turbine run back greater than 25% thermal reactor power or (2) Electrical load reduction greater than 25% full electrical load, or (3) Thermal power oscillations greater than 10%, or (4) Reactor scram, or (5) Valid ECCS initiation.

PAGE 12 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPIP-1 TERM/PHRASE MEANING/DEFINITION SI Surveillance Instruction Site Area Events are in process or have occurred which involve actual or likely major Emergency failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts (1) toward site personnel or equipment that could lead to the likely failure thereof or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.

Site Boundary That line beyond which the land or property is not owned, leased, or otherwise controlled by TVA.

Subcritical Reactor power below the heating range and not trending upward.

Suppression The water volume contained in the suppression chamber intended to Pool condense steam from an MSRV actuation or a primary system break inside the drywell, and provide an ECCS system injection water source.

Suppression The structure enclosing the suppression pool water and the atmosphere Chamber above it.

TAF Top of Active Fuel TEDE Total Effective Dose Equivalent Torus The lower portion of the primary containment which encloses the suppression pool. Equivalent to the suppression chamber.

Toxic Gas A gas that is dangerous to life or limb by reason of inhalation or skin contact.

TSC Technical Support Center Valid An indication, report, or condition is considered to be valid when it is conclusively verified by redundant indicators or actual observation by plant personnel.

Visible Damage Damage to equipment that is readily observable without measurements, testing, or analYSis. Damage is sufficient enough to cause concern regarding the continued operability or reliability of affected safety structure, system, or component.

PAGE 13 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPIP-1 TERM/PHRASE MEANINGIDEFINITION Vital Area An area that contains equipment necessary for the safe operations and shutdown of the plant.

WRGERMS Wide Range Gaseous Effluent Radiation Monitoring System yr Year PAGE 14 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPIP-1 5.0 EVENT CLASSIFICATION INDEX SECTION 1.0 REACTOR 1.1 WATER LEVEL 1.2 SCRAM FAILURE 1.3 REACTOR COOLANT ACTIVITY 1.4 MSUOFFGAS RADIATION 1.5 LOSS OF DECAY HEAT REMOVAL SECTION 2.0 PRIMARY 2.1 PRIMARY CONTAINMENT PRESSURE CONTAINMENT 2.2 PRIMARY CONTAINMENT HYDROGEN 2.3 DRYWELL RADIATION 2.4 DRYWELL INTERNAL LEAKAGE 2.5 LOSS OF PRIMARY CONTAINMENT SECTION 3.0 SECONDARY 3.1 SECONDARY CONTAINMENT CONTAINMENT TEMPERATURE 3.2 SECONDARY CONTAINMENT RADIATION SECTION 4.0 RADIOACTIVITY 4.1 GASEOUS EFFLUENT RELEASES 4.2 MAIN STEAM LINE BREAK 4.3 LIQUID EFFLUENT SECTION 5.0 LOSS OF POWER 5.1 LOSS OF AC POWER 5.2 LOSS OF 250V DC POWER SECTION 6.0 HAZARDS 6.1 RADIOLOGICAL 6.2 CONTROL ROOM EVACUATION 6.3 TURBINE FAILURE 6.4 FIRE/EXPLOSION 6.5 TOXIC GASES 6.6 FLAMMABLE GASES 6.7 SECURITY 6.8 VEHICLE CRASH 6.9 SPENT FUEL STORAGE SECTION 7.0 NATURAL EVENTS 7.1 EARTHQUAKE 7.2 TORNADO/HIGH WINDS 7.3 FLOOD SECTION 8.0 EMERGENCY 8.1 TECHNICAL SPECIFICATIONS DIRECTOR 8.2 LOSS OF COMMUNICATION JUDGMENT 8.3 LOSS OF ASSESSMENT CAPABILITY 8.4 OTHER

(

LAST PAGE PAGE 15 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSI FICA TION PROCEDU RE INTRODUCTION EPIP-1 THIS PAGE INTENTIONALLY BLANK PAGE 16 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 REACTOR 1.0 PAGE 17 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1

(

NOTES 1.1-U 1/1.1-A 1 Applicable when the Reactor Head is removed and the Reactor Cavity is flooded.

1.1-S1 Applicable in Mode 5 when the Reactor Head is installed.

1.1-G2 The reactor will remain subcritical under all conditions without boron when:

Unit 1: All control rods are inserted to or beyond position 02.

Unit 2: Any 19 control rods are inserted to position 02, with all other control rods fully inserted.

Unit 3: Any 19 control rods are inserted to position 02, with all other control rods fully inserted.

All control rods except one are inserted to or beyond position 00.

Determined by Reactor Engineering.

CURVES/TABLES:

TABLE 1.1 *G2 MINIMUM ALTERNATE RPV FLOODING PRESS (MARFP)

NUMBER OF OPEN MSRVs MARFP (PSIG) 6 or More 190 5 230 4 290 PAGE 18 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 Uncontrolled water level decrease in Reactor Uncontrolled water level decrease in Spent Fuel c:

Cavity with irradiated fuel assemblies expected to Pool with irradiated fuel assemblies expected to z remain covered by water. remain covered by water. c:

CJ) c:

J::.

r-m OPERATING CONDITION: OPERATING CONDITION m

Mode 5 ALL z

-I Uncontrolled water level decrease in Reactor Uncontrolled water level decrease in Spent Fuel Cavity expected to result in irradiated fuel Storage Pool expected to result in irradiated fuel assemblies being uncovered. assemblies being uncovered.

OPERATING CONDITION: OPERATING CONDITION:

Mode 5 ALL CJ)

=i m

m s::

m

0 (i) m z

OPERATING CONDITION: OPERATING CONDITION: o ALL Mode 1 or 2 or 3 -<

Reactor water level can NOT be determined AND Either of the following exists:

The reactor will remain subcritical without boron G')

under all conditions, and m

~ Less than 4 MSRVs can be opened, or z

~ Reactor pressure can NOT be restored and m

maintained above Suppression Chamber ~

r-pressure by at least

.:. UNIT 1 - 90 psi m

.:. UNIT 2 - 80 psi s:

m

.:. UNIT 3 - 70 psi

It has NOT been determined that the reactor will ~

m remain subcritical without boron under all z conditions and unable to restore and maintain o MARFP in Table 1.1-G2. -<

OPERATING CONDITION: OPERATING CONDITION:

Mode 1 or 2 or 3 Mode 1 or 2 or 3 PAGE 19 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 (

NOTES 1.2 Subcritical is defined as reactor power below the heating range and not trending upward.

CURVESITABLES:

Ea.

2 LU Do Do Do

0 E

0..

~

....J 0..

0::

0..

J II)

SUPPR PL LVL (Fn I11III ACTION REQUIRED IF ABOVE CURVE FOR EXISTING RX PRESS SUPPR PL LVL (FT)

I11III AC"IlCIN IIECIUIRED IF ABOVE CURVE FOR EXISTING RX PRESS PAGE 20 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE

( EVENT CLASSIFICATION MATRIX EPIP-1 Reactor coolant activity exceeds 26 IlCilgm dose c:

equivalent 1-131 (Technical Specification Limits) z as determined by chemistry sample. c:

en c:

):10 r-m OPERATING CONDITION m ALL z

-I Failure of RPS automatic scram functions to bring Reactor coolant activity exceeds 300 IlCilgm dose the reactor subcritical equivalent lodine-131 as determined by chemistry AND sample.

):10 r-Manual scram or ARI (automatic or manual) was m successful.

OPERATING CONDITION:

~

OPERATING CONDITION: Mode 1 or 2 or 3 Mode 1 or 2 Failure of automatic scram, manual scram, and en ARI to bring the reactor subcritical. =i m

m S

m

a (i) m z

OPERATING CONDITION: o Mode 1 -<

Failure of automatic scram, manual scram, and ARI. Reactor power is above 3% G) m 2

AND m Either of the following conditions exists: ~

r-

Suppression Pool temp exceeds HCTL. m Refer to Curve 1.2-G. s:

m

Reactor water level can NOT be restored and maintained at or above -180 inches. ~

m 2

o OPERATING CONDITION: -<

Mode 1 or 2 PAGE 21 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 NOTES CURVES/TABLES:

UNIT 1 CURVE 1.5-S HEAT CAPACITY TEMP LIMIT 11 11

!5 U) 1 SUPPR PL LVL (FT)

ACT10H REQUIRED IF ABOVE CURVE FOR EXISTING RX PRESS UNIT2 CURVE 1.5-S HEA T CAPACITY TE MP LIMIT SUPPR PL LVL (FT)

ACTION REQUIRED IF ABOVE CURVE FOR EXISTING RX PRESS UNIT3 CURVE 1.5-S HEAT CAPACITY TEMP LIMIT 11 11 11 ill 11.5 12 13 14 15 16 17 18 19 SUPPR PL LVL (FT) 1fj AC110N REQUIRED IF ABOVE CURVE FOR EXISTING RX PRESS PAGE 22 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 c EVENT CLASSIFICATION MATRIX c

z c(J)

OR c Valid OG PRETREATMENT RADIATION HIGH

>r-alarm, RA-90-157A. m m

OPERATING CONDITION: z Mode 1 or 2 or 3 -I Reactor moderator temperature can NOT be maintained below 212 0 F whenever Technical Specifications require Mode 4 conditions or during operations in Mode 5. >

r-m

~

OPERATING CONDITION:

Mode 4 or 5 (J)

=i m

m 5

m

u G) m z

OPERATING CONDITION: o Mode 1 or 2 or 3 -<

G) m z

m

~

r m

s:

m

~

m z

o PAGE 23 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1

(

THIS PAGE INTENTIONALLY BLANK

(

PAGE 24 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 PRIMARY CONTAINMENT 2.0 PAGE 25 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 NOTES CURVES/TABLES:

TABLE 2.1-A INDICATIONS OF PRIMARY SYSTEM LEAKAGE INTO PRIMARY CONTAINMENT Primary Containment Pressure Hiqh Alarm Drywell Floor Drain Sump Pump Excessive Operation Drywell CAM Activity Increasing Drywell Temperature High Alarm Chemistry Sample Radionuclide Comparison To Reactor Water CURVE 2.1-S PRESS SUPPR PRESS

~30T7~~~~~~~

(!)

ij)

~ 25+*~rlb~~~~~~--~~--+---~--~~~

I 20 T7.".....,,==+---__--_+--__+----__--+-----_+_---+-_+~

"w 15+*4-~--~---+--~----~--+---~--~~~

! 10+*4-~--~---+--~----~~+---~--~-+~

"a.a. 5~4_~--_I_--_+ ____+----1__--+_--_+_---+-_+~

~

1111.512 13 14 15 16 17 18 19 20 SUPPR PL LVL (FT)

PAGE 26 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 c:

z c:

CJ) c:

):0 r-m m

z

-I Drywell pressure at or above 2.45 psig AND ):0 r-Indication of Primary System leakage into m Primary Containment. Refer to Table 2.1-A. ~

OPERATING CONDITION:

Mode 1 or 2 or 3 Drywell or Suppression Chamber CJ) hydrogen concentration at or above 4% ~

m AND m S

m Drywell or Suppression Chamber ;;U oxygen concentration at or above 5%. G')

m z

OPERATING CONDITION: OPERATING CONDITION: o Mode 1 or 2 or 3 Mode 1 or 2 or 3 -<

Suppression Chamber pressure can NOT be Drywell or Suppression Chamber maintained below 55 psig. hydrogen concentration at or above 6% G) m z

AND m Drywell or Suppression Chamber r

~

oxygen concentration at or above 5%. m s:

m

~

m OPERATING CONDITION: OPERATING CONDITION: z Mode 1 or 2 or 3 o

Mode 1 or 2 or 3 -<

PAGE 27 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 NOTES CURVES/TABLES:

TABLE 2.3-A/2.3-S2 DRYWELL RADIATION LEVELS WITH RCS BARRIER INTACT INSIDE PRIMARY CONTAINMENT UNIT 1 UNIT2 UNIT3 RAD MONITOR RlHR RAD MONITOR RlHR RADMONITOR RlHR 1-RE-90-272A 196 2-RE-90-272A 642 3-RE-90-272A 196 1-RE-90-273A 297 2-RE-90-273A 297 3-RE-90-273A 297 TABLE 2.3-S1/2.3-G2 DRYWELL RADIATION LEVELS WITH RCS BARRIER NOT INTACT INSIDE PRIMARY CONTAINMENT UNIT 1 UNIT2 UNIT3 RAD MONITOR RlHR RAD MONITOR RlHR RAD MONITOR RlHR 1-RE-90-272A 2981 2-RE-90-272A 2263 3-RE-90-272A 2981 1-RE-90-273A 2960 2-RE-90-273A 2960 3-RE-90-273A 2960 TABLE 2.3-G1 DRYWELL RADIATION LEVELS WITH RCS BARRIER NOT INTACT INSIDE PRIMARY CONTAINMENT UNIT 1 UNIT2 UNIT3 RAD MONITOR RlHR RAD MONITOR RlHR RAD MONITOR RlHR 1-RE-90-272A 90091 2-RE-90-272A 68405 3-RE-90-272A 90091 1-RE-90-273A 89450 2-RE-90-273A 89450 3-RE-90-273A 89450 TABLE 2.3/2;5-U INDICATIONS OF LOSS OF PRIMARY CONTAINMENT Unexplained Loss Of Containment Pressure Exceeding SI-4.7.A.2.a Limits Inability To Isolate Any Line Exiting Containment When Isolation Is Required Venting Irrespective Of Offsite Release Rates Per EOls/SAMGs PAGE 28 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 c:

z c:

C/)

c:

):10 r-m m

z

-I Drywell radiation levels at or above the values listed in Table 2.3-N2.3-S2, with the RCS barrier intact inside Primary Containment.

OPERATING CONDITION:

Mode 1 or 2 or 3 Drywell radiation levels at or above the values Drywell radiation levels at or above the values listed in Table 2.3-S1/2.3-G2 with the RCS barrier listed in Table 2.3-N2.3-S2, with the RCS barrier C/)

NOT intact inside Primary Containment. intact inside Primary Containment, ::j AND m Either of the following exists: m

Indications of loss of Primary Containment. s::

m Refer to Table 2.3/2.5-U. ;;a

Primary Containment integrity can NOT be (i) maintained.

m z

o OPERATING CONDITION: OPERATING CONDITION: -<

Mode 1 or 2 or 3 Mode 1 or 2 or 3 Drywell radiation levels at or above the values Drywell radiation levels at or above the values G) listed in Table 2.3-G1 with the RCS barrier NOT listed in Table 2.3-S1/2.3-G2 with the RCS barrier m intact inside Primary Containment. NOT intact inside Primary Containment, z AND m

Either of the following exists: ~

r-

Indications of loss of Primary Containment.

Refer to Table 2.3/2.5-U. m

Primary Containment integrity can NOT be s:

m maintained.

~

m z

OPERATING CONDITION: OPERATING CONDITION: o Mode 1 or 2 or 3 Mode 1 or 2 or 3 -<

PAGE 29 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 NOTES CURVES/TABLES:

TABLE 2.3/2.5-U INDICATIONS OF LOSS OF PRIMARY CONTAINMENT Unexplained Loss Of Containment Pressure Exceeding SI-4.7.A.2.a Limits Inability To Isolate Any Line Exiting Containment When Isolation Is Required Ventinq Irrespective Of Offsite Release Rates Per EOls/SAMGs

(

PAGE 30 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 c:

z OR c:

C/)

c:

Drywell identified leakage exceeds 40 gpm. >

r m

OPERATING CONDITION: OPERATING CONDITION: m Mode 1 or 2 or 3 Mode 1 or 2 or 3 z

~

r m

~

OPERATING CONDITION:

Mode 1 or 2 or 3 C/)

=i m

m S

m AI G) m z

o Ci) m 2

m

~

r m

s:

m

~

m 2

o

(

PAGE 31 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 THIS PAGE INTENTIONALLY BLANK PAGE 32 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 c EVENT CLASSIFICATION MATRIX SECONDARY CONTAINMENT 3.0 PAGE 33 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 NOTES CURVES/TABLES:

TABLE 3.1 MAXIMUM SAFE OPERATING AREA TEMPERATURE LIMITS APPLICABLE PANEL 9-21 MAX SAFE OPERATING AREA TEMPERATURE ELEMENTS VALUE of (UNLESS OTHERWISE NOTED) UNIT 1 UNIT2 UNIT3 RHR AlC Pump Room 74-95A 215 150 155 RHR BID Pump Room 74-95B 150 210 215 HPCI Turbine Area 73-55A 275 270 270 CS AlC Pump and RCIC Turbine Area 71-41A 190 190 190 RCIC Steam Supply Area 71-41B, 41C, 410 195 200 250 HPCI Steam Supply Area 73-55B, 55C, 550 245 240 240 RHR AlC Pump Supply Area 74-95H 245 240 240 RHR BID Pump Supply Area 74-95G 190 240 240 Main Steam Line Leak Detection High (XA-55-3D-24) Panel 9-3 TIS-1-60A 315 315 315 RHR Valve Room 74-95E 175 170 175 RWCU 1501 Logic Channel AlB Temp (XA-55-5B-32/33) Panel 9-5 175 170 175 High 69-835A, B, C, 0 Aux Inst Room RWCU Outbd 1501 Vlv Area 69-29F 220 220 220 RWCU HxArea 69-29G 220 220 220 RWCU Hx Exh Duct 69-29H 220 220 220 RWCU Recirc Pump A Area 69-290 215 215 215 RWCU Recirc Pump B Area 69-29E 215 215 215 RHR A/C Hx Room 74-95C 210 195 200 RHR BID Hx Room 74-950 210 195 200 FPC HxArea 74-95F 160 155 155 TABLE 3.1-G/3.2-G INDICATIONS OF POTENTIAL OR SIGNIFICANT FUEL CLADDING FAILURE WITH RCS BARRIER INTACT INSIDE PRIMARY CONTAINMENT UNIT 1 DRYWELL RADIATION UNIT 2 DRYWELL RADIATION UNIT 3 DRYWELL RADIATION 1-RE-90-272A I > 196 R/HR 2-RE-90-272A I> 642 R/HR 3-RE-90-272A I > 196 R/HR 1-RE-90-273A I > 297 R/HR 2-RE-90-273A I> 297 R/HR 3-RE-90-273A I > 297 R/HR Reactor Coolant Activity Reactor Coolant Activity Reactor Coolant Activity

300 /lCi/gm Dose Equivalent ::: 300 /lCi/gm Dose Equivalent ::: 300 /lCi/gm Dose Equivalent Iodine 131 Iodine 131 Iodine 131 PAGE 34 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 SECONDARY CONTAINMENT TEMPERATURE Description I I I I I c::

z c::

(()

c::

r-m m

z

-I I I I I I r-m

u

-I 3.1-S I I J TABLE I US I

(()

An unisolable Primary System leak is discharging into Secondary Containment =i m

AND m S

m Any area temperature exceeds the Maximum Safe Operating Temperature limit listed in Table 3.1. ;::u (i) m z

OPERATING CONDITION: o Mode 1 or 2 or 3 -<

3.1-G I I I TABLE I US I An unisolable Primary System leak is discharging into Secondary Containment G) m 2

AND m Any area temperature exceeds the Maximum Safe Operating Temperature limit listed in Table 3.1 r-m AND s:

m Any indication of potential or significant fuel cladding failure exists. Refer to Table 3.1-G/3.2-G with RCS Barrier intact inside Primary Containment.

~

m 2

OPERATING CONDITION o

Mode 1 or 2 or 3 PAGE 35 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 NOTES CURVES/TABLES:

TABLE 3.2 MAXIMUM SAFE OPERATING AREA RADIATION LIMITS AREA RAD MONITOR MAX SAFE VALUE MRlHR RHR West Room 90-2SA 1000 RHR East Room 90-28A 1000 HPCI Room 90-24A 1000 CS/RCIC Room 90-26A 1000 Core Spray Room 90-27A 1000 Suppr Pool Area 90-29A 1000 CRD-HCU West Area 90-20A 1000 CRD-HCU East Area 90-21A 1000 TIP Drive Area 90-23A 1000 North RWCU System Area 90-13A 1000 South RWCU System Area 90-14A 1000 RWCU System Area 90-9A 1000 MG Set Area 90-4A 1000 Fuel Pool Area 90-1A 1000 Service Fir Area 90-2A 1000 New Fuel Storage 90-3A 1000 TABLE 3.1-G/3.2-G INDICATIONS OF POTENTIAL OR SIGNIFICANT FUEL CLADDING FAILURE WITH RCS BARRIER INTACT INSIDE PRIMARY CONTAINMENT UNIT 1 DRYWELL RADIATION UNIT 2 DRYWELL RADIATION UNIT 3 DRYWELL RADIATION 1-RE-90-272A I > 196 R/HR 2-RE-90-272A I> 642 R/HR 3-RE-90-272A I > 196 R/HR 1-RE-90-273A I > 297 R/HR 2-RE-90-273A I> 297 R/HR 3-RE-90-273A L> 297 R/HR Reactor Coolant Activity Reactor Coolant Activity Reactor Coolant Activity 2:. 300 IlCi/gm Dose Equivalent 2:. 300 IlCi/gm Dose Equivalent 2:. 300 IlCi/gm Dose Equivalent Iodine 131 Iodine 131 Iodine 131 PAGE 36 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 SECONDARY CONTAINMENT RADIATION Description I I I I I c:

z c:

(J) c:

r m

m z

-I 3.2-A I I I I I Any of the following high radiation alarms on Panel 9-3:

RA-90-1A, Fuel Pool Floor Alarm

RA-90-250A, Reactor, Turbine, Refuel Exhaust

RA-90-142A, Reactor Refuel Exhaust

RA-90-140A, Refueling Zone Exhaust >>

r m

AND

~

Confirmation by Refuel Floor personnel that irradiated fuel damage may have occurred.

OPERATING CONDITION:

ALL 3.2-S I I I TABLE I I An unisolable Primary System leak is discharging into Secondary Containment CJ)

j AND m m

Any area radiation level at or above the Maximum Safe Operating Area radiation limit listed in Table 3.2. S m

0 G')

OPERATING CONDITION: m z

Mode 1 or 2 or 3 o 3.2-G I I I TABLE I US I An unisolable Primary System leak is discharging into Secondary Containment G')

m AND z m

Any area radiation level at or above the Maximum Safe Operating Area radiation limit listed in Table 3.2. ~

r AND m

s:

m

~

Any indication of potential or significant fuel cladding failure exists. Refer to Table 3.1-G/3.2-G with RCS Barrier intact inside Primary Containment.

m z

OPERATING CONDITION o Mode 1 or 2 or 3 -<

PAGE 37 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 THIS PAGE INTENTIONALLY BLANK

(

PAGE 38 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 RADIOACTIVITY RELEASES 4.0

(

PAGE 39 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 NOTES 4.1-U Prior to making this emergency classification based upon the WRGERMS indication, assess the release by eITher of the following:

1. Actual field measurements exceed the limITS in table 4.1-U

2. 0-S14.8.B.1.a.1 release fraction exceeds 2.0 If neITher assessment can be conducted within 60 minutes then the dedaration must be made on the valid WRGERMS reading.

4.1-A Prior to making this emergency classification based upon the WRGERMS indication, assess the release by eITher of the following:

1. Actual field measurements exceed the limITS in table 4.1-A

2. 0-S14.8.B.1.a.1 release fraction exceeds 200 If neITher assessment can be conducted within 15 minutes then the dedaration must be made on the valid WRGERMS reading.

4.1-5 Prior to making this emergency classification based upon the gaseous release rate indication, assess the release by either of the following methods:

1. Actual field measurements exceed the limITS in table 4.1-S.

2. Projected or actual dose assessments exceed 100 mrem TEDE or 500 mrem CDE.

If neither assessment can be conducted within 15 minutes then the declaration must be made based on the valid WRGERMS reading.

4.1-G Prior to making this emergency classification based upon the gaseous release rate indication, assess the release by either of the following methods:

1. Actual field measurements exceed the limITS in table 4.1-8.

2. Projected or actual dose assessments exceed 1000 mrem TEDE or 5000 mrem CDE.

If neither assessment can be conducted within 15 minutes then the declaration must be made based on the valid WRGERMS reading.

CURVES/TABLES' Table.4.1-U RELEASE LIMITS FOR UNUSUAL EVENT TYPE MONITORING METHOD LIMIT DURATION

(

Gaseous Release Rate Stack Noble Gas (WRGERMS) 2.88 X 10 7 ~Ci/sec 1 Hour Gaseous Release Rate O-SI 4.8. B.1.a.1 Release Fraction 2.0 1 Hour Site Boundary Radiation Reading Field Assessment Team 0.10 MREMfHR Gamma 1 Hour Table 4.1-A RELEASE LIMITS FOR ALERT TYPE MONITORING METHOD LIMIT DURATION Gaseous Release Rate Stack Noble Gas (WRGERMS) 2.88 X 10 9 ~Ci/sec 15 Minutes Gaseous Release Rate 0-SI4.8.B.1.a.1 Release Fraction 200 15 Minutes Site Boundary Radiation Reading Field Assessment Team 10 MREMfHR Gamma 15 Minutes Table 4.1-S RELEASE LIMITS FOR SITE AREA EMERGENCY TYPE MONITORING METHOD LIMIT DURATION Gaseous Release Rate Stack Noble Gas (WRGERMS) 5.9 X 10 9 ~Ci/sec 15 Minutes Site Boundary Radiation Reading Field Assessment Team 100 MREMfHR Gamma 1 Hour Site Boundary lodine-131 Field Assessment Team 3.9 X 10 -7 ~CI fcm 3 1 Hour Table 4.1-G RELEASE LIMITS FOR GENERAL EMERGENCY TYPE MONITORING METHOD LIMIT DURATION Gaseous Release Rate Stack Noble Gas (WRGERMS) 5.9 X 10 10 )..lCi/sec 15 Minutes Site Boundary Radiation Reading Field Assessment Team 1000 MREMfHR Gamma 1 Hour Site Boundary lodine-131 Field Assessment Team 3.9 X 10 -6 )..lCI f cm 3 1 Hour PAGE 40 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 GASEOUS EFFLUENT Description 4.1-U I I NOTE I TABLE I I c:

Gaseous release exceeds ANY limit and duration in Table 4.1-U. z c:

CJ) c:

):It r

m OPERATING CONDITION:

m ALL z

-f 4.1-A I I NOTE I TABLE I I Gaseous release exceeds ANY limit and duration in Table 4.1-A.

):It r

m

~

OPERATING CONDITION:

ALL 4.1-S I I NOTE I TABLE I I CJ)

EITHER of the following conditions exists: =i m

Gaseous release exceeds or is expected to exceed ANY limit and duration in Table 4.1-S. m S

m

Dose assessment indicates actual or projected dose consequences :::0 above 100 mrem TEDE or 500 mrem thyroid CDE. Q m

z OPERATING CONDITION: o ALL -<

4.1-G I I NOTE I TABLE I J (j)

EITHER of the following conditions exists: m 2

m

Gaseous release exceeds or is expected to exceed ANY limit and duration in Table 4.1-G.

~

r

Dose assessment indicates actual or projected dose consequences m

above 1000 mrem TEDE or 5000 mrem thyroid CDE. s::

m

~

m OPERATING CONDITION 2 ALL o

(/

PAGE 41 OF 206 REVISION 43

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1

(

NOTES CURVES/TABLES:

PAGE 42 OF 206 REVISION 43

ML091810120

Contents

Text

C=O e~,~=Ia

References:

References:

References:

REFERENCES:

References:

References:

References:

References:

References:

References:

2.0 REFERENCES

References:

1.0 INTRODUCTION

2.0 REFERENCES

1.0 INTRODUCTION

2.0 REFERENCES

Navigation menu

ML091810120

Text

C=O e~,~=Ia

References:

References:

References:

REFERENCES:

References:

References:

References:

References:

References:

References:

2.0 REFERENCES

References:

1.0 INTRODUCTION

2.0 REFERENCES

1.0 INTRODUCTION

2.0 REFERENCES

Navigation menu

Search