Text

Initial Exam 2010-302 Draft SRO Written Exam
	ML110350603
Person / Time
Site:	Sequoyah
Issue date:	02/04/2011
From:	; NRC/RGN-II
To:	; Tennessee Valley Authority
References
	50-327/10-302, 50-328/10-302
	Download: ML110350603 (309)
	v • d • e

I Sequoyah Nuclear Plant SEPT 2010 NRC INITIAL LICENSE SRO Written ExAM Questions76-100

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta

76. Given the following condition:

- Unit 1 is operating at 100% power.

- Pressurizer pressure drops rapidly resulting in a low pressure SI.

- The reactor can NOT be tripped manually from the MCR.

- While performing FR-S.1 Nuclear Power Generation I ATWS, immediate operator actions, the CRC places Turbine Trip HS-47-24 to the TRIP position.

Which ONE of the following identifies...

(1) The direction HS-47-24 is rotated to trip the turbine and the expected light indication that results.

and (2) When ES-0.5, Equipment Verifications, will be first performed?

A. (1) The handswitch is rotated counterclockwise and the red light illuminates.

(2) ES-0.5 would be performed in parallel with the performance of FR-S.1.

B (1) The handswitch is rotated clockwise and the green light illuminates.

(2) ES-0.5 would be performed in parallel with the performance of FR-S.1.

C. (1) The handswitch is rotated counterclockwise and the red light illuminates.

(2) ES-0.5 would be performed after the transition was made to E-0, Reactor Trip or Safety Injection.

D. (1) The handswitch is rotated clockwise and the green light illuminates.

(2) ES-0.5 would be performed after the transition was made to E-0, Reactor Trip or Safety Injection.

Thursday, July 15, 2010 4:47:20 PM 76

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta DIS TRACTOR ANALYSIS:

A. Incorrect. Plausible because most control room switches are rotated counterclockwise to stop the associated equipment, the turbine trip switch is turned to other direction and ES-U. 5 will be directed to be performed by step 6 in FR-S. 1 while continuing performance of the FR.

B. Correct. The handswitch must be rotated clockwise, which is opposite the normal direction of turning a switch to turn-off a piece of equipment, this causes the red indicating light to go out and the green light illuminates.

ES-U. 5 will be directed to be performed by step 6 as a result of the Safety Injection.

C. Incorrect. Plausible because most control room switches are rotated counterclockwise to stop the associated equipment and ES-U. 5 performance would not be performed had an SI not occurred.

D. Incorrect. Plausible because the handswitch must be rotated clockwise which causes the red indicating light to go out and ES-U. 5 performance would not be performed had an SI not occurred.

Thursday, July 15, 2010 4:47:20 PM 77

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta Question Number: 76 Tier: 1 Group I K/A: 029 Anticipated Transient Without Scram (ATWS)

EA2.06 Ability to determine or interpret the following as they apply to a ATWS:

Main turbine trip switch position indication Importance Rating: 3.8 / 3.9 10 CFR Part 55: 43.5 /45.13 10CFR5543.b: 5 KIA Match: The question requires knowledge of operation of the main turbine trip switch during an ATWS and indications available to the Operator during switch manipulation. SRO because it requires knowledge of procedure selection and useage during an emergency condition.

Technical

Reference:

FR-S.1, Nuclear Power Generation /ATWS, Rev 23.

Proposed references None to be provided:

Learning Objective: 0PL271 FR-S.1 Obj. 4 Cognitive Level:

Higher Lower X Question Source:

New X New question generated 03-16-2010 from K/A.

Modified Bank Bank Question History: New question generated 03-16-2010 from K/A for Sept 2010 SQN NRC exam.

Comments:

MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9 Answer: BADDADBDBA ScrambleRange: A-D Thursday, July 15, 2010 4:47:20 PM 78

SQN NUCLEAR POWER GENERATION I ATWS FR-S.1 Rev. 23 STEP IACTIONIEXPECTED RESPONSE I RESPONSE NOTOBTAINED

5. VERIFY Containment Purge isolated:

a. VERIFY containment purge and vent a. PERFORM the following:

dampers (System 30) CLOSED.

[Panel 6K and 6L1 1) ENSURE containment purge supply and exhaust fans STOPPED. [M-9]

2) CLOSE dampers. [M-9J

6. MONITOR for SI signal:

a. CHECK SI signal ACTUATED. a. IF SI signal is required, THEN ACTUATE SI signal.

IF SI signal is NOT required, THEN GO TO Step 7.

b. PERFORM the following WHILE continuing with this procedure:

1) E-0, Reactor Trip or Safety Injection, Steps 1 through 4.

2) ES-0.5, Equipment Verifications.

Page 6 of 16

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0PL271 FR-S.1 Revision 3 Page 1 of 18 NUCLEAR TRAINING TRAINING MATERIALS COVERSHEET OPERATOR TRAINING - LICENSED Program LICENSE TRAINING 0PL271 Course Course No.

FR-SI, NUCLEAR POWER GENERATION /ATWS OPL27IFR-S.1 Lesson Title Lesson Plan No.

INPO Accredited Yes No Multiple Sites Affected fl Yes No

.4 Prepared By A FRoddy 7-Signaire I Date Process Review

/ Signature / Date Lead Instructor/Program Mgr Review

/Signature / Date Line Owner Approval Signature / Date WAN Concurrence (if applicable)

Signature / Date BFN SON VBN CORP Receipt Inspection arid Distribution: / 1, T7#1T!1ig Materials Coordinator Date Standardized Training Material Copies to:

0PL271 FR-S. 1 Revision 3 Page 4 of 18 I. PROGRAM: OPERATOR TRAINING LICENSED -

II. COURSE: LICENSE TRAINING III. LESSON TITLE: FR-S.1, NUCLEAR POWER GENERATION /ATWS IV. LENGTH OF LESSONICOURSE: 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months V. TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of FR-S.1, NUCLEAR POWER GENERATION / ATWS.

B. Enabling Objectives

0. Demonstrate an understanding of NUREG 1122 Knowledges and Abilities associated with FR-Si, NUCLEAR POWER GENERATION /ATWS that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A.

1. Explain the purpose/goal of FR-S. 1.

2. Discuss the FR-S.1 entry conditions.

a. Describe the setpoints, interlocks, and automatic actions associated with FR-S.1 entry conditions.

b. Describe the requirements associated with FR-S.1 entry conditions.

3. Summarize the mitigating strategy for the failure that initiated entry into FR-S.1.

4. Describe the bases for all limits, notes, cautions, and steps of FR-S. 1.

5. Describe the conditions and reason for transitions within this procedure and transitions to other procedures.

6. Given a set of initial plant conditions use FR-S.1 to correctly:

a. Identify required actions

b. Respond to Contingencies

c. Observe and Interpret Cautions and Notes

7. Apply GFE and system response concepts to the performance of FR-S.1 conditions.

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta

77. Given the following:

- A steam generator tube rupture is in progress on Unit 1.

- The Chemistry Lab has confirmed that #1 SG is ruptured.

- The crew is implementing E-3, Steam Generator Tube Rupture.

According to Tech Spec bases, which ONE of the following actions directed by E-3, requires entry into a Tech Spec Action statement?

A. Closing the #1 SG MSIV.

B. Blocking the Low Steamline Pressure SI.

C Closing the TD AFW pump steam supply valve from #1 SG.

D. Cooling down to target incore temperature of 480°F at the maximum rate.

DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible since the action is taken in E-3, but closing the MSIV does not render the MSIV inoperable.

B. Incorrect, Plausible since the action is taken in E-3, however blocking the signal when RCS pressure is below the P-Il setpoint is allowed by a Note in TS Table 3.3-3.

C. Correct, In accordance with T.S.3. 7.1.4 basis, the TD AFWi5 INOPERABLE if either of the steam supplies to the turbine are not operable. Since closing the isolatIon valve from loop #1 will cause the loop #4 valve to open, it will not transfer back to loop #1.

D. Incorrect, Plausible, since the initial cooldown rate may exceed 100°F/hr.

the total amount of temperature drop does not exceed 100°F.

Thursday, July 15, 2010 4:46:41 PM 77

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta Question Number: 77 Tier: 1 Group 1 KIA: 038 Steam Generator Tube Rupture (SGTR)

EG 2.2.40 Ability to apply Technical Specifications for a system.

Importance Rating: 3.4 I 4.7 10 CFR Part 55: 41.10 IOCFR55.43.b: 43.2 / 43.5 KIA Match: Question matches the K/A by requiring the candidate to apply the Tech Spec for specified conditions to determine if Tech Spec LCO entry is required. SRO because condition of lnoperability is only addressed by information in the basis of the specification.

Technical

Reference:

T.S. 3.7.1.2 basis Amendment 206, T.S. Table 3.3-3.

E-3 Steam Generator Tube Rupture, rev 17.

Proposed references None to be provided:

Learning Objective: OPL27IE-3. Obj. B.4 OPT200.AFW Obj B.6.c Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: Bank question T/S 0307.03 001 updated by revising and re-arranging the distractors Comments:

MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9 Answer: CABBBBAABD ScrambleRange: A-D Thursday, July 15, 2010 4:46:41 PM 78

PLANT SYSTEMS BASES Q = Nominal NSSS power rating of the plant (including reactor coolant pump heat), Mwt K = Conversion factor, 947.82 (Btu/sec)

Mwt W = Minimum total steam flow rate capability of the operable MSSVs on any one steam generator at the highest MSSV opening pressure including tolerance and accumulation, as appropriate, in lb/sec. For example, if the maximum number of inoperable MSSVs on any one steam generator is one, then Ws should be a summation of the capacity of the operable MSSVs at the highest operable MSSV r.operating pressure, excluding the highest capacity MSSV. If the maximum number of inoperable MSSVs per steam generator is three then W should be a summation of the capacity of the operable MSSVs at the highest operable MSSV operating pressure, excluding the three highest capacity MSSVs.

hfg = heat of vaporization for steam at the highest MSSV opening pressure including tolerance and accumulation, as appropriate, Btu,lbm N = Number of loops in plant The values calculated from this algorithm must then be adjusted lower to account for instrument and channel uncertainties.

3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The AFW System is configured into three trains. The AFW System is considered OPERABLE when the components and flow paths required to provide redundant AFW flow to the steam generators are OPERABLE. This requires that the two motor-driven AFW pumps be OPERABLE in two diverse paths, each supplying AFW to separate steam generators. The turbine-driven AFW pump is required to be OPERABLE with redundant steam supplies from each of two main steam lines upstream of the MSIVs, and shall be capable of supplying AFW to any steam generator. The piping, valves, instrumentation, and controls in the required flow paths also are required to be OPERABLE.

The AFW System mitigates the consequences of any event with loss of normal feedwater.

The design basis of the AFW System is to supply water to the steam generator to remove decay heat and other residual heat by delivering at least the minimum required flow rate to the steam generators while pumping against the highest credible steam generator pressure.

In addition, the AFW System must supply enough makeup water to replace steam generator secondary inventory lost as the unit cools to MODE 4 conditions. Sufficient AFW flow must also be available to account for flow losses such as pump recirculation and line breaks.

August 14, 2001 SEQUOYAH - UNIT 1 B 3/4 7-2 Amendment No. 115, 155, 196, 206

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TOTAL NO. MINIMUM OF CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT CHANNELS TO TRIP OPERABLE MODES ACTION

f. Steam Line Pressure- 3/steam line 2/steam line 2/steam line 1, 2, 3# 17 Low in any steam line

2. CONTAINMENT SPRAY

a. Manual 2 1** 2 1,2, 3,4 20

b. Automatic Actuation 2 1 2 1,2,3,4 15 Logic

c. Containment Pressure- 4 2 3 1, 2, 3 18

-High-High

3. CONTAINMENT ISOLATION

a. Phase A Isolation

1) Manual 2 1 2 1, 2, 3,4 20

2) From Safety 2 1 2 1,2,3,4 15 Injection Automatic Actuation Logic

- Two switches must be operated simultaneously for actuation.

April 11, 2005 SEQUOYAH - UNIT 1 3/4 3-16 Amendment No.41, 63, 141, 168, 301

TABLE 3.3-3 (Continued)

TABLE NOTATION Trip function may be bypassed in this MODE below P-il (Pressurizer Pressure Block of Safety Injection) setpoint.

Trip function automatically blocked above P-il and may be blocked below P11 when Safety Injection on Steam Line Pressure-Low is not blocked.

ACTION STATEMENTS ACTION 15 - With the number of OPERABLE Channels one less than the Total Number of Channels, be in at least HOT STANDBY within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; however, one channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.2.1.1 provided the other channel is OPERABLE.

ACTION 16 - Deleted.

ACTION 17 - With the number of OPERABLE Channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. The Minimum Channels OPERABLE requirements is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.3.2.1.1.

ACTION 18 - With the number of OPERABLE Channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and the Minimum Channels OPERABLE requirement is met; one additional channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.2.1.1.

ACTION 19 - With less than the Minimum Channels OPERABLE, operation may continue provided the containment purge supply and exhaust valves are maintained closed.

ACTION 20 - With the number of OPERABLE Channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

September 14, 2006 SEQUOYAH - UNIT 1 3/4 3-22 Amendment No. 63, 141, 168, 182, 188, 202, 207, 213, 301, 311

OPT200.AFW Rev. 5 Page 11 of 93 ENABLING OBJECTIVES_(contc

6. Describe the administrative controls and limits for the AFW system.

State Tech Specs/TRM LCOs that govern the Auxiliary Feedwater system a State the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action limit TS LCOs

Given the conditions/status of the AFW system components and the appropriate sections of Tech. Specs., determine if operability requirements are met and what actions are required.

7. Discuss related Industry Events IX. INTRODUCTION:

M. Tech Spec and TRM N. Industry Events:

OPT200.AFW Rev. 5 Page 38 of 93 Auxiliary Feedwater Components

Steam Supply Isolation Valves Operation

FCV-1-15 and the T&T valve are closed. When FCV 1-15 closes, FCV-1-16 opens and the T&T valve re opens after FCV-1-16 is full open.

Steam supplies will not automatically realign from Loop4to Loop 1.

If the TDAFWP steam supply is aligned to Loop 4, it is considered inoperable by Tech Specs because it cant automatically align to an alternate steam supply.

EO 3 X. LESSON BODY:

0PL271 E-3 Revision 0 Page 1 of 24 SEQUOYAH NUCLEAR PLANT NUCLEAR TRAINING TRAINING MATERIALS COVERSHEET OPERATOR TRAINING - LICENSED PROGRAM LICENSE TRAINING 0PL271 COURSE COURSE NO.

E-3, Steam Generator Tube Rupture 0PL271 E-3 LESSON TITLE LESSON PLAN NO.

INPO ACCREDITED YES X NO MULTIPLE SITES AFFECTED YES NO X

/1 PREPARED i--7---.--

- Lacy Pauley Sgnatu(e 1 Date PROCESS REVIEW S-Th

.wod Hy 512L tI inature / Date LEAD INSTRUCTORIPROGRAM MGR. 5JA REVIEW Signature / Date

- JQrU1 tifl5af1 7 PLANTCONCURRENCE ----

Frank Roddy - 3 Si ture / Date TVAN CONCURRENCE (if applicable)

BFN SQN WBN CORP Sigra[urc / T3aie Rec&pt Inspection and Distbution T(in Material5 Coordinator I Date Standardized Training Material Copies to:

WA 40385 [NP 6-2003] Page 1 of 2

0PL271 E-3 Revision 0 Page 3 of 24 I. PROGRAM: OPERATOR TRAINING - LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: E-3, Steam Generator Tube Rupture IV. LENGTH OF LESSON!COURSE: 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months V. TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of E-3, Steam Generator Tube Rupture B. Enabling Objectives

0. Demonstrate an understanding of NUREG 1122 Knowledges and Abilities associated with Steam Generator Tube Rupture that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A.

1. Explain the purpose/goal of E-3.

2. Discuss the E-3 entry conditions.

a. Describe the setpoints, interlocks, and automatic actions associated with E-3 entry conditions.

b. Describe the requirements associated with E-3 entry conditions.

3. Summarize the mitigating strategy for the failure that initiated entry into E-3.

4. Describe the bases for all limits, notes, cautions, and steps of E-3.

5. Describe the conditions and reason for transitions within this procedure and transitions to other procedures.

6. Given a set of initial plant conditions use E-3 to correctly:

a. Identify required actions

b. Respond to Contingencies

c. Observe and Interpret Cautions and Notes

7. Apply GFE and system response concepts to the performance of E-3 conditions.

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta

78. Given the foNowing:

- Unit 2 was operating at 100% power when a loss of main feedwater occurred.

- An automatic trip occurred but steam generator level recovery is delayed due to low initial levels and maintenance that was in progress for the turbine driven auxiliary feedwater pump.

- The crew has transitioned to ES-0.1, Reactor Trip Response, and is implementing FR-H.5, Steam Generator Low Level as a yellow path.

- Total auxiliary feedwater flow currently is 600 gpm.

In accordance with SPP-3.5, Regulatory Reporting Requirements, which ONE of the following identifies both...

(1) the TVA internal notification(s) required to be made directly by the Shift Manager and (2) the maximum time allowed for making the first required notification to the NRC?

A. (1) Site Operations Management, pjy (2) 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B. (1) Site Operations Management, çjjjy (2) 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months C (1) Site Operations Management and Duty Plant Manager (2) 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months D. (1) Site Operations Management and Duty Plant Manager (2) 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Thursday, July 15, 2010 4:45:36 PM 78

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta DISTRACTOR ANALYSIS:

A. Incorrect, Plausible because a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months report is the earliest report required by 10CFR5O. 72(b)(2) and Site Operations management must be notified by the Shift Manager. Incorrect because the Shift Manager is also required to notify the Duty Plant Manager.

B. Incorrect, Plausible because an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months report is required by IOCFR5O. 72(b)(3) and Site Operations management must be notified by the Shift Manager. Incorrect because a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months report is required by 10CFR5O. 72(b)(2) and the Shift Manager is also required to notify the Duty Plant Manager.

C. Correct, A 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months report required by IOCFR5O. 72(b)(2) is the earliest report required to the NRC. Site Operations Management and the Duty Plant Manager must be notified directly by the Shift Manager in accordance with appendix D of SPP-3.5.

D. In correct, Plausible because an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months report is required by IOCFR5O. 72(b)(3) and Site Operations Management and the Duty Plant Manager must be notified directly by the Shift Manager. Incorrect because a report required by IOCFR5O. 72(b)(2) must be made within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Thursday, July 15, 2010 4:45:36 PM 79

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta Question Number: 78 Tier: 1 Group 1 K/A: 054 Loss of Main Feedwater (MEW)

AG2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

Importance Rating: 2.7/ 4.1 IOCFRPart55: 41.10/43.5/45.11 IOCFR55.43.b: 5, Site Specific SRO task in SPP-3.5.

KIA Match: Question matches the K/A because it requires knowledge of the internal notifications and notification to the NRC required for a reactor trip resulting from a loss of main feedwater event. SRO because internal and Regulatory notifications are task specific to a Senior Reactor Operator.

Technical

Reference:

SPP-3.5, Regulatory Reporting Requirements, Rev 0021 Proposed references None to be provided:

Learning Objective: OPL271SPP-3.5 Obj B.3 Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History:

Comments: Written from K/A 3-16-2010. LWP MCS Time: 1 Points: LOO Version: 0 1 2 3 4 5 6 7 8 9 Answer: CDCBCAAABA ScrambleRange:A-D Thursday, July 15, 2010 4:45:36 PM 80

NPG Standard Regulatory Reporting Requirements SPP-3.5 Programs and Rev. 0022 Processes Page 20 of 75 Appendix A (Page 2 of 11)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 3.0 REQUIREMENTS NOTES

1) Internal management notification requirements for plant events are found in Appendix D. The Operations Shift Manager is responsible for notifying Site Operations Management and the Duty Plant Manager. The Duty Plant Manager is responsible for making the remaining internal management notifications.

2) NRC NUREG-1 022, Supplements and subsequent revisions should be used as guidance for determining reportability of plant events pursuant to §50.72 and §50.73.

3.1 Immediate Notification NRC -

TVA is required by §50.72 to notify NRC immediately if certain types of events occur. This appendix contains the types of events and the allotted time in which NRC must be notified.

(Refer to Form SPP-3.5-1 or NRC Form 361). Operations is responsible for making the reportability determinations for §50.72 and §50.73 reports. Operations is responsible for making the immediate notification to NRC in accordance with §50.72.

Notification is via the Emergency Notification System. If the Emergency Notification System is not operative, use a telephone, telegraph, mailgram, or facsimile.

NOTE The NRC Event Notification Worksheet may be used in preparing for notifying the NRC. This Worksheet may be obtained directly from the NRC website (www.nrc.gov) under Form 361, or WA NPG Form SPP-3.5-1 may be used.

A. The Immediate Notification Criteria of §50.72 is divided into 1-hour, 4-hour, and 8- hour phone calls. Notify the NRC Operations Center within the applicable time limit for any item which is identified in the Immediate Notification Criteria.

B. The following criteria require 1-hour notification:

1. (Technical Specifications) Safety Limits as defined by the Technical Specifications which have been violated.

2. §50.72 (a)(1)(i) The declaration of any of the Emergency classes specified in the licensees approved Emergency Plan.

NPG Standard Regulatory Reporting Requirements SPP-3.5 -

Programs and Rev. 0022 Processes Page 21 of 75 Appendix A (Page 3 of 11)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 3.1 Immediate Notification - NRC (continued)

NOTE If it is discovered that a condition existed which met the Emergency Plan criteria but no emergency was declared and the basis for the emergency class no longer exists at the time of discovery, an ENS notification (and notification of the Operations Duty Specialist), within one hour of discovery of the undeclared (or misclassified) event, shall be made. However, actual declaration of the emergency class is not necessary in these circumstances.

3. §50.72(b).(1)) Any deviation from the plants Technical Specifications authorized pursuant to §50.54(x).

C. The following criteria require 4-hour notification:

1. §50.72(b)(2)(i) The initiation of any nuclear plant shutdown required by the plants Technical Specifications.

2. §50.72(b)(2)(iv)(A) Any event that results or should have resulted in Emergency Core Cooling System (ECCS) 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 or reactor operation.

3. §50.72(b)(2)(iv)(B) Any event or condition 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. §50.72(b)(2)(xi) 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 notification to other government agencies has been or will be made. Such an event may include an onsite fatality or inadvertent release of radioactive contaminated materials.

D. The following criteria require 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)(ii)(A) Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.

NPG Standard Regulatory Reporting Requirements SPP-3.5 Programs and Rev. 0022 Processes Page 22 of 75 Appendix A (Page 4 of 11)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 3.1 Immediate Notification - NRC (continued)

2. §50.72(b)(3)(ii)(B) Any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.

3. §50.72(b)(3)(iv)(A) Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) [see list below], except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

a. Reactor protection system (RPS) including: Reactor scram and reactor trip.

NOTE Actuation of the RPS when the reactor is critical is also reportable under §50.72(b)(2)(iv)(B) above.

b. General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIV5).

c. Emergency core cooling systems (ECCS) for pressurized water reactors (PWR5) including: High-head, intermediate-head, and low-head injection systems and the low pressure injection function of residual (decay) heat removal systems.

d. ECCS for boiling water reactors (BWRs) including: core spray systems; high-pressure coolant injection system; low pressure injection function of the residual heat removal system.

e. BWR reactor core isolation cooling system.

f. PWR auxiliary or emergency feedwater system.

g. Containment heat removal and depressurization systems, including containment spray and fan cooler systems.

h. Emergency ac electrical power systems, including: Emergency diesel generators (EDG5).

4. §50.72(b)(3)(v) Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to:

a. Shut down the reactor and maintain it in a safe shutdown condition;

b. Remove residual heat;

NPG Standard Regulatory Reporting Requirements SPP-3.5 Programs and Rev. 0022 Processes Page 43 of 75 Appendix 0 (Page 1 of 2)

Site Event Notification Matrix NOTE The Operations Shift Manager is responsible for notifying Site Operations management and the Duty Plant Manager. The Duty Plant Manager is responsible for the remaining internal management notifications as noted in the matrix.

Notification Requirements EventlCondition Duty Plant Plant Manager Ops. Duty Spec. Site VP Corporate Manager (ODS) Duty Officer*

Reactor/Turbogenerator trip, unscheduled unit power reduction, or Yes# Yes# Reactor trip and Yes Yes nonscheduled unit shutdown; and when unit is restored to full unscheduled service, shutdown only Unplanned entry into a Limiting Condition for Operation with time Yes Yes No Yes Yes duration of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or less.

Classification of the Radiological Emergency Plan (REP) at any Yes Yes Yes Yes Yes entry level.

Personnel injuries that are personnel first aids, potential lost-time Yes Yes Yes Yes Yes injuries, serious recordable injuries, and injuries requiring hospital admittance or transport to an off-site medical facility. (Refer to Appendix J)

Death of any person as a result of injuries received on site or due Yes Yes Yes Yes Yes to medical problems occurring while onsite.

Release of oil or hazardous materials to the discharge canal, Yes Yes No Yes Yes ponds or river and violations of the NPDES permit Any event which may be newsworthy to the public. (1) Yes Yes Yes Yes Yes Any Security Contingency Events or any phone call to NRC Yes Yes Yes Yes Yes Operations Center regarding security issues.

NPG Standard Regulatory Reporting Requirements SPP-3.5 Programs and Rev. 0022 Processes Page 44 of 75 Appendix D (Page 2 of 2)

Site Event Notification Matrix Notification Requirements EventlCondition Duty Plant Plant Manager Ops. Duty Spec. Site VP Corporate Manager (ODS) Duty Officer*

NRC 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months phone calls. Yes Yes Yes for reactor Yes Yes for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months trips, shutdowns, and 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months transport of calls.

contaminated or potentially contaminated victim to hospital and for loss of Prompt Notification System.

Any unusual radiation exposure to personnel, including personnel Yes Yes No Yes Yes contamination events.

Accidental, unplanned or uncontrolled off-site radioactive release. Yes Yes No Yes Yes Any reasonable threat to generation. Yes Yes No Yes Yes Outage critical path extensions exceeding 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Yes Yes No Yes Yes Any reactivity event or unplanned reactivity change. Yes Yes No Yes Yes NOTE:

(1) Consider items specified in Appendix E step 2.2 of this procedure.

Plant Manager should ensure the Plant Managers at the other NPG sites are notified so the Shift Managers and Work Week Managers at the unaffected sites can review their schedules for potential generation-risk activities that may need to be deferred.

If the Corporate Duty Officer cannot be contacted within 15 minutes, the Chief Nuclear Officer should be notified.

Corporate Duty Officer (CDO)

The CDO position is always available and is not part of the REP. This position is not intended to conflict or duplicate any responsibilities of the Central Emergency Control Center (CECC) Director under the REP or the Site Vice President.

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

Ill. Justification for Plant Specific Exemptions The 25 SRO-only questions shall evaluate the additional knowledge and abilities required for the higher license level in accordance with 10 CFR 55.43(b). [NUREG 1021, Section ES-401D.2.dJ The fact that a facility licensee trains its ROs to master certain 10 CFR 55.43 knowledge, skills, and abilities does NOT mean that they can no longer be used as a basis for SRO-only questions. [Operator Licensing Feedback Web page Item 401.36 @ http://www.nrc.gov!reactors/operator-hcensing/op-Iicensing-files/oI-feedback.odfl The SRO-only test item is required to be tied to one of the 10 CFR 55.43(b) items. However, if a licensee desires to evaluate a knowledge/ability that is not tied to one of the 10 CFR 55.43(b) items, then the licensee can classify the knowledge/ability as unique to the SRO position provided that there is documented evidence that ties the knowledge/ability to the licensees SRO job position duties in accordance with the systematic approach to training (SAT).

> Justification: A question that is not tied to one of the 10 CFR 55.43(b) items can still be classified as SRO-only provided the licensee has documented evidence to prove that the knowledge/ability is unique to the SRO position at the site. An example of documented evidence includes:

o The question is linked to a learning objective that is specifically labeled in the lesson plan as being SRO-only (e.g., some licensee lesson plans have columns in the margin that differentiate AO, RO, and SRO learning objectives) [NUREG 1021, ES-401, Section D.2.dj AND/OR o A question is linked to a task that is labeled as an SRO-only task, and the task is NOT listed in the RO task list.

Page 10 of 16

0PL27 1 SPP-3.5 Rev. 2 Page 3 of 19 I. PROGRAM: OPERATOR TRAINING LICENSED -

II. COURSE: LICENSE TRAINING III. LESSON TITLE: REPORTING REQUIREMENTS IV. LENGTH OF LESSON/COURSE: 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (s)

V. TRAINING OBJECTIVES:

A. Tenninal Objective:

Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the actions necessary to comply with regulatory and plant reporting requirements.

B. Enabling Objectives Demonstrate an understanding of NIJREG 1122 Knowledges and Abilities associated with Regulatory and Plant Reporting Requirements that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A.

2. Perform a plant response assessment using the 0-TI-QXX-000-001.0, Event Critique, Post Trip Report, and Equipment Root Cause.

a. State the responsibilities of each control room crewmember. [C.1j

b. Explain the process or Conduct a plant response assessment.

3. For a given condition, determine the regulatory reporting requirements using appropriate reference material.

a. List the tools available to the operator for determining regulatory reporting requirements.

b. Define the key terms used to determine regulatory reporting requirements.

C. State the criteria requiring one-hour notification of the NRC.

d. State the criteria requiring four-hour notification of the NRC.

e. State the criteria requiring eight-hour notification of the NRC.

f. State the criteria requiring 24-hour notification of the NRC.

g. State the criteria requiring 2-day notification of the NRC.

h. State the criteria requiring a written report or LER to the NRC.

I. State the criteria allowing a telephone notification to be made in lieu of a written LER to the NRC.

4. For a given condition, determine plant management reporting requirements using SPP-3 .5.

5. Complete a PER reportability determination per SPP-3. 1.

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta

79. Given the following:

- A loss of offsite power resulted in Unit I only entry into ECA-0.0 Loss Of All AC Power.

Present plant conditions include the following:

- Core Exit TCs 592°F.

- Highest loop T-hot indication is 590°F.

- RCS wide range pressure is 2085 psig.

- Pressurizer level is 12% and stable.

- Containment pressure stabilized at 0.9 psig.

- 1 B DG has been manually started and has energized 6.9 kv shutdown board lB.

- ECA-0.0 Appendix A Locking Out Shutdown Boards Loads has been completed.

Which of the following identifies:

(1) the recovery procedure that Unit I will utilize and (2) when FR procedures are required to be implemented?

A. (1) ECA-0.2 Recovery From Loss Of All AC Power With SI Required.

(2) When directed by ECA-0.2.

B. (1) ECA-0.2 Recovery From Loss Of All AC Power With SI Required.

(2) Upon transition out of ECA-0.0.

C (I) ECA-0.1 Recovery From Loss Of All AC Power Without SI Required.

(2) When directed by ECA-0.1.

D. (1) ECA-0.1 Recovery From Loss Of All AC Power Without SI Required.

(2) Upon transition out of ECA-0.0.

Thursday, July 15, 2010 4:21 :29 PM 79

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta DISTRA CTOR ANAL YSIS:

A. Incorrect. Plausible because the Operator must evaluate subcooling margin, containment pressure and pressurizer level to determine no SI is required and if ECA-O.2 is selected as the recovery procedure FRP will be implemented at step

15.

B. Incorrect. Plausible because the Operator must evaluate subcooling margin, containment pressure and pressurizer level to determine no SI is required. It is plausible that FRP implementation would be resumed upon transition out ot ECA-O.O since power availability has been restored.

C. Correct, The Operators evaluation of subcooling margin, containment pressure and pressurizer level will conclude no SI is required making ECA-O. I the appropriate recovery procedure. FRP implementation will be restored at step #15.

D. Incorrect. Plausible because no SI is required making ECA-O. 1 the appropriate recovery procedure. It is plausible that FRP implementation would be resumed upon transition out ot ECA-O.O since power availability has been restored.

Thursday, July 15, 2010 4:21 :29 PM 80

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta Question Number: 79 Tier: 1 Group 1 KIA: 055 Loss of Offsite and Onsite Power (Station Blackout)

EA2.03 Ability to determine or interpret the following as they apply to a Station Blackout:

Actions necessary to restore power Importance Rating: 3.9 / 4.7 10 CFR Part 55: 43.5/45.13 IOCFR55.43.b: 5 KIA Match: K/A match because the SRO must evaluate plant conditions and select the correct procedure to restore power to critical equipment following loss of offsite power. SRO because procedure selection is evaluated and knowledge of Function Restoration procedure applicability is required.

Technical

Reference:

ECA-0.0 Loss of All AC Power Rev 23, ECA-O.1 Recovery from Loss of All AC Power Without SI Required Rev 8.

Proposed references None to be provided:

Learning Objective: 0PL271 ECA-0.0 Obj B.4 Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History:

Comments: New question written 3-23-2010 from K/A. LWP MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9 Answer: CBCABBADAA ScrambleRange: A-D Thursday, July 15, 2010 4:21 :29 PM 81

SQN LOSS OF ALL AC POWER ECA-O.O Rev. 23 I STEP I I ACTION!EXPECTED RESPONSE I I RESPONSE NOT OBTAINED NOTE Steps 1, 2, and 3 are immediate action steps.

1. SUSPEND FRP implementation and MONITOR status trees for information only.

2. VERIFY reactor TRIPPED: TRIP reactor.

Reactor trip breakers OPEN IF reactor trip breakers will NOT open, THEN

Reactor trip bypass breakers DISPATCH personnel to perform the OPEN or DISCONNECTED following:

Neutron flux DROPPING

OPEN reactor trip breakers and MG set output breakers

[MG Set Room, Aux Bldg, elev 759].

OPEN breakers to Control Rod MG Sets [480V Unit Bds A and B].

3. VERIFY turbine TRIPPED: TRIP turbine.

ALL turbine stop valves CLOSED IF main turbine CANNOT be tripped,

[SSPS status lights on M-6]. THEN CLOSE MSIVs and MSIV bypass valves.

Page 3 of 29

SQN LOSS OF ALL AC POWER ECA-O.O Rev. 23 STEP I ACTIONIEXPECTED RESPONSE I RESPONSE NOT OBTAINED NOTE I ECCS injection flow indicators are unavailable until vital DC loads are restored as directed by step 35.

NOTE 2 IF RCP seal cooling is isolated, further cooling of RCP seals will be established by natural circulation cooldown in subsequent procedures.

36. SELECT recovery procedure:

a. CHECK RCS subcooling based on a. GO TO ECA-0.2, Recovery From core exit T/Cs greater than 40°F. Loss of All AC Power With SI Required.

b. CHECK pressurizer level b. GO TO ECA-0.2, Recovery From greater than 10% [20% ADV]. Loss of All AC Power With SI Required.

c. CHECK ECCS status: c. IF any ECCS pumps are running AND ECCS injection flow exists,

SI pumps in PULL TO LOCK THEN

RHR pumps in PULL TO LOCK GO TO ECA-0.2, Recovery From Loss of All AC Power With SI

COPs in PULL TO LOCK Required.

OR CCPIT ISOLATED.

d. GO TO ECA-0.1, Recovery From Loss of All AC Power Without SI Required.

END Page 28 of 29

RECOVERY FROM LOSS OF ALL AC POWER ECA-O.1 SQN WITHOUT SI REQUIRED Rev. 8 ISTEPI IACTION!EXPECTED RESPONSE I RESPONSE NOTOBTAINED

15. RESUME FRP implementation.

16. MONITOR if letdown can be established:

a. CHECK pressurizer level a. GO TO Step 17.

greater than 20% [35% ADVI.

b. ESTABLISH letdown b. ESTABLISH excess letdown USING EA-62-5, Establishing Normal USING EA-62-3, Establishing Excess Charging and Letdown. Letdown.

17. CONTROL charging and letdown flow to maintain pressurizer level between 20% [35% ADV] and 50%.

Page 13 of 17

RECOVERY FROM LOSS OF ALL AC POWER ECA-O.2 SQN WITH SI REQUIRED Rev. 9 I STEP I j ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED

14. PLACE containment spray pumps in A-AUTO.

15. RESUME FRP implementation.

16. MONITOR if hydrogen igniters and recombiners should be turned on:

a. CHECK hydrogen concentration a. PERFORM the following:

measurement AVAILABLE:

1) PLACE HS-43-200A

Hydrogen analyzers in ANALYZE [M-1O].

have been in ANALYZE for at least 5 minutes. 2) PLACE HS-43-210A in ANALYZE [M-1O].

3) RECORD present time:

4) WHEN hydrogen analyzers have been in ANALYZE for at least 5 minutes, THEN PERFORM substeps 16.b through 16.e.

5) GO TO Step 17.

(step continued on next page)

Page 11 of 12

0PL271 ECA-0.0 Revision 1 Page 3 of 21 I. PROGRAM: OPERATOR TRAINING LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: EMERGENCY OPERATING PROCEDURE ECA-0.0, LOSS OF ALL AC POWER IV. LENGTH OF LESSONICOURSE: 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months V. TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of ECA-0.0, LOSS OF ALL AC POWER.

B. Enabling Objectives

0. Demonstrate an understanding of NUREG 1122 Knowledges and Abilities associated with ECA-0.0, LOSS OF ALL AC POWER that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A.

1. Explain the purpose/goal of ECA-0.0.

2. Discuss the ECA-0.0 entry conditions.

3. Summarize the mitigating strategy for the failure that initiated entry into ECA-0.0.

4. Describe the bases for all limits, notes, cautions, and steps of ECA-0.0.

5. Describe the conditions and reason for transitions within this procedure and transitions to other procedures.

6. Given a set of initial plant conditions use ECA-0.0 to correctly:

a. Identify required actions

b. Respond to Contingencies

c. Observe and Interpret Cautions and Notes

7. Apply GEE and system response concepts to the performance of ECA-0.0 conditions.

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta

80. Given the following:

- A loss of all offsite power has occurred.

- 2A 6.9kv shutdown board is energized by the 2A diesel generator.

- The other three 6.9kV shutdown boards have no power.

- Offsite power restoration is estimated to be complete in 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

- The Unit 1 US is directing unit response per ECA-0.0, Loss of All AC Power.

Which ONE of the following identifies:

(1) required actions to shed DC loads in accordance with ECA-0.0, and (2) the basis for leaving the turbine emergency oil pump in service for 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months when shedding 250V DC loads?

Note:

EA-250-1, Load Shed of Vital Loads After Station Blackout EA-250-2, Load Shed of 250V DC Loads After Station Blackout A. (1) Perform EA-250-2, pjjjy.

(2) To prevent potential for a fire or explosion due to hydrogen loss from the main generators.

B. (1) Perform EA-250-2, pjjjy.

(2) To ensure sufficient turbine heat load has been dissipated from the main turbines.

C. (1) Perform both EA-250-1 and EA-250-2.

(2) To prevent potential for a fire or explosion due to hydrogen loss from the main generators.

D (1) Perform both EA-250-1 and EA-250-2.

(2) To ensure sufficient turbine heat load has been dissipated from the main turbines.

Thursday, July 15, 2010 4:17:53 PM 80

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible to perform only EA-250-2 because one of the Shutdown boards is available to allow charging of one of the 125v Vital batteries and to prevent a fire or an explosion due to hydrogen leakage is the basis for de-pressurizing the main generator prior to load shedding the air side Seal Oil pumps.

B. Incorrect, Plausible to perform only EA-250-2 because one of the Shutdown Boards is available to allow charging of one of the 125v Vital batteries and the basis for the 3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months time limit to load shed the 250v DC Turbine Emergency Oil Pump is to ensure sufficient turbine heat load has been dissipated from the main turbines is correct.

C. Incorrect, Plausible because performing both EA-250-1 and EA-250-2 is correct and to prevent a fire or an explosion due to hydrogen leakage is the basis for de-pressurizing the main generator prior to load shedding the air side Seal Oil pumps.

D. Correct, The conditions (3 shutdown boards de-energized) require both EA-250-1 and EA-250-2 to be performed and the basis for the 3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months time limit to load shed the 250v DC Turbine Emergency Oil Pump is to ensure sufficient turbine heat load has been dissipated from the main turbines.

Thursday, July 15, 2010 4:17:53 PM 81

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta Question Number: 80 Tier: 1 Group 1 K/A: 058 Loss of DC Power AG2.4.18 Knowledge of the specific bases for EOPs.

Importance Rating: 3.3 / 4.0 10 CFR Part 55: 41.10 /43.1 /45.13 IOCFR55.43.b: 5 KIA Match: Specific knowledge of the bases for EOPs as related to the loss of DC power. SRO because it requires seletion of the appropriate procedure.

Technical

Reference:

ECA-0.0, Loss of All AC Power, Rev. 23 EPM-3-ECA-0.0, Basis Document for ECA-0.0 Loss of All AC Power, Rev 11 ;EA-250-1 Load Shed of Vital Loads After Station Blackout Rev 14; EA-250-2, Load Shed of 250V DC Loads After Station Blackout, Rev. 9 Proposed references None to be provided:

Learning Objective: OPT200.DC Obj B; 0PL271 ECA-0.0 Obj B 4.

Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question Comments:

MCS Time: I Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9 Answer: DBBBCCBBBD ScranibleRange:A-D Thursday, July 15, 2010 4:17:53 PM 82

SQN LOSS OF ALL AC POWER ECA-O.O Rev. 23 STEPI IACTION/EXPECTED RESPONSE I IRESPONSENOTOBTAINED I NOTE I To conserve battery capacity, 1 25V and 250V battery board breakers with pink SBO tags must be opened within 45 minutes and DC air side seal oil pump breakers must be opened within 90 minutes of power loss.

NOTE 2 DC load shed will result in loss of all MCR annunciators, permissive lights, SSPS status lights, and numerous control board indications.

NOTE 3 Generator hydrogen pressure must be monitored locally at L-39 panel due to MCR indication being de-energized.

13. MONITOR if DC load shed required:

a. IF off-site power to station service has been lost, THEN DISPATCH an AUO to vent main generator hydrogen and shed 250 V DC loads USING EA-250-2, Load Shed of 250 V DC Loads After Station Blackout.

b. IF at least three 6900 V shutdown boards DE-ENERGIZED (both units combined),

THEN DISPATCH an AUO to shed 125 V DC loads USING EA-250-1, Load Shed of Vital Loads After Station Blackout.

Page 11 of 29

SQN EOl BASIS DOCUMENT FOR ECA-O.0 EPM-3-ECA-0.O PROGRAM LOSS OF ALL AC POWER Rev. 11 MANUAL Page 33 of 94 EOP Step Number: 13 MONITOR if DC load shed required:

ERG Step Number: NIA

Purpose:

To complete a DC load shed if three or more shutdown boards are deenergized.

ERG Basis:

N/A EOP Basis:

Following loss of all AC power, the station batteries are the only source of electrical power.

The station batteries supply the DC busses and the AC vital instrument busses. With no AC power available to charge the station batteries, battery power supply must be conserved to permit monitoring and control of the plant until AC power can be restored.

In order for SQN to satisfy the 4-hour battery coping requirement in accordance with IOCFR5O.63, Station Blackout Rule, it is necessary to strip certain non-essential loads from the 125 V vital batteries and the 250 V station batteries within prescribed time limits.

250V DC battery chargers are normally powered from 480V AB Common Board. This board will be de-energized on a loss of off-site power. Since shutdown boards are de-energized, alternate power sources to the 250V battery charger is not available and 250V load-shedding must be initiated.

If three of the four 6900 V shutdown boards are deenergized, then the majority of the I25VDC vital battery chargers are de-energized. Therefore, vital ac and dc loads must be stripped to extend vital battery life.

Guidance provided in EA-250-2 will depressurize the main generator to between 2 and 3 psig so there is no hydrogen explosion hazard. This is necessary because the air-side seal oil backup pump must be shed within 90 minutes (DCN E20362A), and the effectiveness of the shaft seal oil to prevent hydrogen leakage to atmosphere will be lost as a result. The generator pressure criteria is based on the values on the vendor drawing for the Hydrogen Seal Oil System. DCN M9120A assumes that the emergency oil pump motor is load shed at 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months after the event. Less than 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months is not desirable since the turbine heat load may not have dissipated sufficiently. More than 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months is not desirable since the batterys assumed lifetime may be adversely affected.

SQN LOAD SHED OF VITAL LOADS I EA-250-1 AFTER STATION BLACKOUT Rev. 14 I 1,2 Page 2 of 24 1.0 PURPOSE To provide instructions for shedding vital 125V DC and 120V AC loads to conserve 125V DC batteries during a station blackout.

2.0 SYMPTOMS AND ENTRY CONDITIONS 2.1 Entry Conditions A. ECA-O.O, Loss of All AC Power.

3.0 PRECAUTIONS AND LIMITATIONS 3.1 Precautions A. If the accountability siren sounds the operator should continue performing this procedure. The SM will remain aware of procedure progression and performing personnel location.

B. Disregard caution placards associated with the breakers being opened in this procedure.

32 Limitations A. Load shedding must be completed within 45 minutes following a loss of power to three of the four 6900V AC shutdown boards.

SQN LOAD SHED OF VITAL LOADS I EA-250-1 AFTER STATION BLACKOUT Rev. 14 I 1,2 Page 3 of 24 4.0 OPERATOR ACTIONS 4.1 Section Applicability

1. IF initiating load shed of vital loads, THEN GO TO Section 4.2. El

2. IF restoring from load shed of vital loads, THEN GO TO Section 4.3. El

3. RETURN TO procedure and step in effect.

SQN EA-250-1 LOAD SHED OF VITAL LOADS 1,2 AFTER STATION BLACKOUT I Rev. 14 I Page 4 of 24 4.2 Load Shedding of Vital I2OVAC and 125V DC Loads

1. OBTAIN portable lantern or flashlight from any of the following locations:

Appendix R Tool Basket in MCR

cabinet in SM office

2. OBTAIN Vital Area and Protected Area keys from glass-faced box in SM office.

NOTE Loads in Step 3 may be shed in any order.

3. OPEN breakers marked with glow-in-the-dark tape and pink SBO tags in the following rooms:

a. 125V Vital Battery Board Room I:

120V Vital Instrument Power Board 1-I LI

120V Vital Instrument Power Board 2-I LI

125V Vital Battery Board I. LI

b. 125V Vital Battery Board Room II:

120V Vital Instrument Power Board 1-Il LI

120V Vital Instrument Power Board 2-Il LI

125V Vital Battery Board II. LI

c. 125V Vital Battery Board Room Ill:

120V Vital Instrument Power Board 1-Ill LI

120V Vital Instrument Power Board 2-Ill LI

125V Vital Battery Board Ill. LI (Step Continued on Next Page)

SQN LOAD SHED OF VITAL LOADS I EA-250-1 AFTER STATION BLACKOUT Rev. 14 I 1,2 Page 5 of 24 4.2 Load Shedding of Vital I2OVAC and 125V DC Loads (Continued)

d. 125V Vital Battery Board Room IV:

120V Vital Instrument Power Board 1-IV LI

120V Vital Instrument Power Board 2-IV LI

1 25V Vital Battery Board IV. LI

4. WHEN Step 3 completed, THEN NOTIFY Unit 1 and Unit 2 UOs SBO breakers open.

5. ENSURE SBO breakers OPEN USING Breaker Checklist 1. LI

6. GO TO Section 4.1, step in effect. LI END OF SECTION

TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT EOI PROGRAM MANUAL EMERGENCY ABNORMAL PROCEDURE EA-250-2 LOAD SHED OF 250V DC LOADS AFTER STATION BLACKOUT Revision 9 QUALITY RELATED PREPARED/PROOFREAD BY: D. A. PORTER RESPONSIBLE ORGANIZATION: OPERATIONS APPROVED BY: W. T. LEARY EFFECTIVE DATE: 04/09/10 REVISION DESCRIPTION: Revised Section 4.2 to address PER 223073.

Incorporated other minor changes.

THIS PROCEDURE CONTAINS A TIME CRITICAL ACTION

SQN LOAD SHED OF 250V DC LOADS EA-250-2 1, 2 AFTER STATION BLACKOUT Rev. 9 Page 2 of 11 1.0 PURPOSE To provide instructions for the shedding of 250 V DC loads to conserve 250 V DC batteries during station blackout.

2.0 SYMPTOMS AND ENTRY CONDITIONS 2.1 Entry Conditions A. ECA-0.0, Loss of All AC Power.

3.0 PRECAUTIONS AND LIMITATIONS 3.1 Precautions A. If the accountability siren sounds the operator should continue performing this procedure. The SM will remain aware of procedure progression and performing personnel location.

B. All MCR annunciators may be lost during load shed of TSC inverters.

3.2 Limitations A. All breakers with pink SBO tags must be opened within 45 minutes following loss of power.

B. 250V DC breakers 525 and 526 must remain closed to provide power to the generator air side seal oil backup pump UNTIL main generator hydrogen pressure is less than 3 psig.

C. 250V DC feed breakers 404 and 405 should remain closed to provide power to the main turbine emergency DC oil pump UNTIL the main turbine is at zero speed AND three hours have elapsed.

SQN LOAD SHED OF 250V DC LOADS EA-250-2 1,2 AFTERSTATION BLACKOUT Rev.9 Page 3 of 11 4.0 OPERATOR ACTIONS 4.1 Section Applicability

1. IF initiating 250V DC load shed, THEN GO TO Section 4.2. E

2. IF restoring from 250V DC load shed, THEN GO TO Section 4.3.

3. RETURN TO procedure and step in effect. El

SQN LOAD SHED OF 250V DC LOADS EA-250-2 1, 2 AFTER STATION BLACKOUT Rev. 9 Page 4 of 11 4.2 Shedding Designated 250 V DC Loads

1. OBTAIN portable lantern or flashlight from any of the following locations:

Appendix R Tool Basket in MCR

cabinet in SM office

2. OBTAIN the following keys: [glass-faced-box in SM office]

Vital Area

Protected Area.

3. CHECK main generator hydrogen pressures on [1-L-39] and [2-L-39].

4. IF Unit I OR Unit 2 main generator hydrogen pressure is greater than 3 psig, THEN PERFORM the following on applicable unit(s):

[Turbine Bldg elev 706, near seal oil unit] Unit I Unit 2

a. ENSURE hydrogen vent isolation valve FVLV-35-5611 OPEN. LI LI

b. OPEN hydrogen vent bypass valve FVLV-35-5681.

c. VERIFY generator hydrogen pressure DROPPING [L-39].

SQN LOAD SHED OF 250V DC LOADS EA-250-2 1,2 AFTER STATION BLACKOUT Rev. 9 Page 5 of 11 4.2 Shedding Designated 250 V DC Loads (Continued)

NOTE I Breakers in Step 5 and 6 are marked with pink SBO tags and glow-in-the dark tape. These breakers must be opened within 45 minutes from loss of power.

NOTE 2 The placard (informational posting) on 250V Battery Board 1 and 2 Breaker 401 (TSC Inverter DC supply) does NOT apply when performing this section.

5. OPEN the following breakers on 250V Battery Board 1:

250 V DC BATTERY BOARD ROOM I 250 V DC BATTERY BOARD I BREAKER LOAD OPEN 401 TSClnverterl LI 527 DC EOP 1A (MFPT) Unit I LI 529 Preferred Inverter 1 LI 530 DC EOP lB (MFPT) Unit I LI

6. OPEN the following breakers on 250V Battery Board 2:

250 V DC BATTERY BOARD ROOM 2 25OVDC BATTERY BOARD 2 BREAKER LOAD OPEN 401 TSC Inverter 2 LI 527 DC EOP 2A (MFPT) Unit 2 LI 529 Preferred Inverter 2 LI 530 DC EOP 2B (MFPT) Unit 2 LI

SQN LOAD SHED OF 250V DC LOADS EA-250-2 1,2 AFTER STATION BLACKOUT Rev. 9 Page6ofll 4.2 Shedding Designated 250 V DC Loads (Continued)

7. WHEN Step 5 and 6 are completed, THEN NOTIFY MCR that SBO breakers are open. El

8. OPEN the following breakers and disconnect switches for TSC Inverters and Preferred Inverters: (marked with green EOl tags):

INVERTER BREAKER LOCATION OPEN -I TSC Inverter 2 AC input breaker 250 V DC Battery Board 2-BKRB-250-NX1 Room 2 Preferred Inverter 2 480 V disconnect switch 250 V DC Battery Board 2-XS-250-RQA Room 2, north wall AC input breaker 250 V DC Battery Board TSC Inverter 1 1-.BKRB-250-NW1 Room 1 480 V disconnect switch 250 V DC Battery Board Preferred Inverter 1 1 -XS-250-RQA Room 1, south wall

9. MONITOR main generator hydrogen pressures on F1-L-391 and 12-L-391.

CAUTION Opening DC seal oil pump breakers prior to reducing generator hydrogen pressure to less than 3 psig could cause hydrogen fire or explosion.

NOTE Generator hydrogen pressure must be reduced to less than 3 psig AND breakers for DC air side seal oil pumps opened within 90 minutes of loss of power.

10. NOTIFY MCR that this section is on hold UNTIL generator hydrogen pressure is less than 3psig. El

SQN LOAD SHED OF 250V DC LOADS EA-250-2 1,2 AFTER STATION BLACKOUT Rev. 9 Page 7 of 11 4.2 Shedding Designated 250 V DC Loads (Continued)

Ii. WHEN generator hydrogen pressure is less than 3 psig, Unit I Unit 2 THEN ENSURE hydrogen vent bypass valve lVLV-35-5681 on affected unit is CLOSED. [TB el. 706, near seal oil skid] LI LI

12. WHEN generator pressure is less than 3 psig on both units, THEN OPEN the following breakers:

LOAD LOCATION BREAKER OPEN Gen Air Side Seal Oil Backup 250 V DC Battery Bd Room 1 525 Pump Motor (U-i Norm Feed) 250 V DC Battery Board 1 Gen Air Side Seal Oil Backup 250 V DC Battery Bd Room 2 525 Pump Motor (U-i Alt Feed) 250 V DC Battery Board 2 Gen Air Side Seal Oil Backup 250 V DC Battery Bd Room 2 526 Pump Motor (U-2 Norm Feed) 250 V DC Battery Board 2 Gen Air Side Seal Oil Backup 250 V DC Battery Bd Room 1 526 Pump Motor (U-2 Alt Feed) 250 V DC Battery Board 1

13. CONTACT MCR:

a. DETERMINE time of Station Blackout.

Time:

b. NOTIFY MCR that this section is on hold UNTIL 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months has elapsed since start of event. LI

SQN LOAD SHED OF 250V DC LOADS EA-250-2 1,2 AFTER STATION BLACKOUT Rev. 9 Page 8 of 11 4.2 Shedding Designated 250 V DC Loads (Continued)

14. WHEN main turbine speed at zero on both units AND 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months has elapsed following initiation of SBO event, THEN OPEN the following breakers:

LOAD LOCATION BREAKER OPEN Turbine Emergency Oil Pump 250 V DC Battery Bd Room 1 404 Motor Unit 1 Normal Feed 250 V DC Battery Board 1 Turbine Emergency Oil Pump 250 V DC Battery Bd Room 1 Motor Unit 2 Alternate Feed 250 V DC Battery Board 1 405 n Turbine Emergency Oil Pump 250 V DC Battery Bd Room 2 404 Motor Unit I Alternate Feed 250 V DC Battery Board 2 Turbine Emergency Oil Pump 250 V DC Battery Bd Room 2 405 LI Motor Unit 2 Normal Feed 250 V DC Battery Board 2

15. GO TO Section 4.1, step in effect.

END OF SECTION

SQN LOAD SHED OF 250V DC LOADS EA-250-2 1,2 AFTER STATION BLACKOUT Rev. 9 Page 9 of 11 4.3 Restoring 250 V DC Loads After Return of Station Power

1. NOTIFY US to evaluate and coordinate the following: Unit I Unit 2

Returning 250 V DC battery chargers to service (based on battery condition)

Restoring battery board loads. LI LI

2. NOTIFY US to coordinate returning inverters to service prior to restoring 250 V DC Battery Board I loads in next step.

3. WHEN directed by MCR, THEN CLOSE the following breakers to restore 250 V DC loads:

250 V DC BATTERY BOARD ROOM I 250 V DC BATTERY BOARD I LOAD BREAKER CLOSED q

TSC Inverter 1 401 Li Turbine Emergency Oil Pump Motor Unit 1 404 Li Normal Feed Turbine Emergency Oil Pump Motor Unit 2 405 LI Alternate Feed Gen Air Side Seal Oil Backup Pump Motor Unit 1 525 LI Normal Feed Gen Air Side Seal Oil Backup Pump Motor Unit 2 526 LI Alternate Feed DC EOP 1A (MFPT) Unit 1 527 Li Preferred Inverter 1 529 Li DC EOP lB (MFPT) Unit 1 530 Li

SQN LOAD SHED OF 250V DC LOADS EA-250-2 1,2 AFTERSTATION BLACKOUT Rev.9 Page 10 of 11 4.3 Restoring 250 V DC Loads After Return of Station Power (Continued)

Unit I Unit2

4. NOTIFY US to coordinate returning inverters to service prior to restoring 250 V DC Battery Board 2 loads in next step.

5. WHEN directed by MCR, THEN CLOSE the following breakers to restore 250 V DC loads:

250 V DC BATTERY BOARD ROOM 2 250 V DC BATTERY BOARD 2 LOAD BREAKER CLOSED q

TSC Inverter 2 401 El Turbine Emergency Oil Pump Motor Unit 1 Alternate Feed 404 El Turbine Emergency Oil Pump Motor Unit 2 405 El Normal Feed Gen Air Side Seal Oil Backup Pump Motor Unit 1 525 Alternate Feed Gen Air Side Seal Oil Backup Pump Motor Unit 2 526 El Normal Feed DC EOP 2A (MFPT) Unit 2 527 El Preferred lnverter 2 529 El DC EOP 2B (MFPT) Unit 2 530 El

6. GO TO Section 4.1, step in effect.

END OF TEXT

SQN LOAD SHED OF 250V DC LOADS EA-250-2 1,2 AFTER STATION BLACKOUT Rev. 9 Page 11 of 11 REFERENCES Drawings 45N704-1 & 2, 250 V DC Battery Boards.

0PL271 ECA-0.0 Revision I Page 3 of 21 I. PROGRAM: OPERATOR TRAINING - LICENSED II. COURSE: LiCENSE TRAINING III. LESSON TITLE: EMERGENCY OPERATING PROCEDURE ECA-0.0, LOSS OF ALL AC POWER IV. LENGTH OF LESSONICOURSE: 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months V. TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of ECA-0.0, LOSS OF ALL AC POWER.

B. Enabling Objectives

0. Demonstrate an understanding of NUREG 1122 Knowledges and Abilities associated with ECA-0.0, LOSS OF ALL AC POWER that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A.

1. Explain the purpose/goal of ECA-0.0.

2. Discuss the ECA-0.0 entry conditions.

3. Summarize the mitigating strategy for the failure that initiated entry into ECA-0.0.

4. Describe the bases for all limits, notes, cautions, and steps of ECA-0.0.

5. Describe the conditions and reason for transitions within this procedure and transitions to other procedures.

6. Given a set of initial plant conditions use ECA-0.0 to correctly:

a. Identify required actions

b. Respond to Contingencies

c. Observe and Interpret Cautions and Notes

7. Apply GFE and system response concepts to the performance of ECA-0.0 conditions.

OPT200. DC Rev 3 Page 6 of 163 V. TRAINING OBJECTIVES: (continued)

A R S S U 0 R T 0 0 A X 7. EXPLAIN the operational implication of the following concept as it applies to the DC Systems:

a. System grounds

b. Loss of ventilation during battery charging X 8. Given specific plant conditions, ANALYZE the effect that a loss or malfunction of the DC Systems will have on the following:

a. D/Gs

b. Components using DC control power x X X 9. Given specific plant conditions, ANALYZE the effect that a loss or malfunction of the following will have on the DC Systems:

a. Motors

b. Breakers, relays and disconnects

c. Test instruments X X X 10. EXPLAIN and APPLY DC Systems limits and precautions as covered by the System Operating Instructions..

X X X 11. Given a specific evolution, IDENTIFY the appropriate DC Systems normal operating procedure(s) required to conduct that evolution X X 12. Using the Technical Specifications and Technical Requirements Manual,

a. EXPLAIN applicable DC Systems Tech Spec LCO and Technical Requirements.

b. Given a set of plant conditions/parameters, DETERMINE entry level conditions for DC Systems Tech Spec LCO actions and Technical

. Requirements.

X X X 13. ANALYZE a given set of plant conditions and/or parameters to determine entry conditions into abnormal operating procedures.

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta

81. Given the following:

- Unit-I is operating at 95% power with turbine controls in manual.

- The inter-tie transformer tap changer is in manual for calibration.

- A severe thunderstorm warning is in effect for Hamilton County.

The following indications occur simultaneously:

- An audible alarm sounds and breaker disagreement white lights for 500kv breaker 5028 Franklin and breaker 5048 Widows Creek are lit.

- The following annunciators alarm:

6900v SD BD lA-A FAILURE OR BUS UNDERVOLTAGE I OVERVOLTAGE 6900 SD BD IA-A OVERVOLTAGE STA FREQ EXCESSIVE ERROR

- The operating crew initiates a turbine load reduction to reduce frequency.

- All attempts to contact the Dispatcher are unsuccessful.

- Ten minutes after the event, the CRC announces that frequency recorder 0-REC-241-2 indicates frequency has risen to 60.6 Hz.

Which ONE of the following identifies...

(1) the action required due to the conditions above, and (2) the action required if the station frequency can NOT to restored to normal by the action taken?

A. (1) Continue turbine load reduction to reduce frequency to 60 Hz.

(2) De-energize shutdown boards one-at-a-time and ensure D/G breakers close to energize shutdown boards.

B. (1) Continue turbine load reduction to reduce frequency to 60 Hz.

(2) Declare both offsite power supplies inoperable.

C. (1) Tripthe reactor.

(2) De-energize shutdown boards one-at-a-time and ensure D/G breakers close to energize shutdown boards.

D (1) Trip the reactor.

(2) Declare both offsite power supplies inoperable.

Tuesday, July 13, 2010 8:04:48 AM 81

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta DISTRACTOR ANAL YSIS:

A. Incorrect, Plausible because continue to drop load would be a correct action for frequency errors of a lower magnitude and de-energizing the shutdown boards one-at-a-time to cause DIGs to supply the boards is a correct action directed in the annunciator response if the frequency is low.

B. Incorrect, Plausible because continue to drop load would be a correct action for frequency errors of a lower magnitude and because declaring both offsite lines inoperable is correct.

C. Incorrect, Plausible because tripping the reactor is correct and de-energizing the shutdown boards one-at-a-time to cause DIGs to supply the boards is a correct action directed in the annunciator response if the frequency is low.

D. Correct, The action to trip the reactor and to declare both offsite power sources inoperable are contained in the Annunciator Response for STA FREQ EXCESSIVE ERROR. Declaring the lines inoperable is required because of the Thability to contact the dispatcher.

Tuesday, July 13, 2010 8:04:48 AM 82

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta Question Number: 81 Tier: 1 Group 1 K/A: 077 Generator Voltage and Electric Grid Disturbances AA2.08 Ability to determine and interpret the following as they apply to Generator Voltage and Electric Grid Disturbances:

Criteria to trip the turbine or reactor Importance Rating: 4.3 I 4.4 10 CFR Part 55: 41.5 and 43.5 IOCFR55A3.b: 2 K/A Match: KA is matched because the question requires the ability to interpret the criteria to trip the reactor or turbine as they apply to conditions with a Generator Voltage and Electric Grid Disturbance.

Technical

Reference:

0-AR ECB6-B, Elecrical Control Board 0-XA-55-ECB6-B, Revision 21 Proposed references None to be provided:

Learning Objective: OPTSTG200.SWYD OBJ B.7 Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question Comments:

MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9 Answer: DAD D C C AC B A Scramble Range: A D Tuesday, July 13, 2010 8:04:48 AM 83

19 (C-5)

Source Setpoint SER 2692 High 60.15 STA FREQ Station frequency recorder Low 59.85 EXCESSIVE ERROR Probable 1. System disturbance.

Causes 2. Recorder malfunction.

Corrective [1] MONITOR grid frequency USING one or more of the following Actions indications:

[a] 0-REC-241-2 on 0-RIB-7 (preferred)

IF IO-XS-241-21 selected to non-running unit, THEN SELECT either OFF OR the OPERATING UNIT.

[b] RCP frequency indications on M-5 (backup)

[c] Digital RPM indication on front end standard (backup)

NOTE RCP Underfrequency trip setpoint is 57 Hz (1710 rpm)

[2] IF grid frequency is less than 57.5 Hz (1725 rpm)

OR greater than 60.5 Hz (1815 rpm),

THEN ENSURE Reactor is tripped AND PERFORM E-0, Reactor Trip or Safety Injection.

NOTE EHC speed reference will cause governor valves to move in CLOSED direction if grid frequency rises significantly with turbine controls in automatic.

[3] MONITOR Reactor and Turbine Power.

[a] MAINTAIN Reactor power less than or equal to 100%.

[b] IF uncontrollable load changes occur, THEN ENSURE Reactor is tripped AND PERFORM E-0, Reactor Trip or Safety Injection.

CONTINUED SQN O-AR-ECB6-B Page 25 of 44 0 Rev. 21

19 (C-5)

CORRECTIVE STA FREQ ACTIONS (CONTINUED) EXCESSIVE ERROR NOTE The following step is based upon not exceeding more than 25% of the lifetime limit for off-frequency operation of the [P turbine. Refer to Tl-28 Figure A.26.

[4] MONITOR LP turbine for Off-Frequency Turbine Operation:

[a] IF grid frequency is less than 58.5 Hz (1755 rpm) for more than 2.5 minutes, THEN ENSURE Reactor is tripped AND PERFORM E-0, Reactor Trip or Safety Injection.

[b] IF grid frequency is less than 59.5 Hz (1785 rpm) for more than 15 minutes, THEN ENSURE Reactor is tripped AND PERFORM E-0, Reactor Trip or Safety Injection.

NOTE Steps [5], [6], and [7] are based upon guidance contained in the TPS Transmission Emergency Plan.

[5] IF frequency is greater than 60.1 Hz (1803 rpm), THEN PERFORM the following:

[a] REDUCE load at 1% per minute UNTIL grid frequency reaches 60.1 Hz. (1803 rpm).

[b] ATTEMPT to establish communications with Dispatcher (751-4208).

[c] IF 10 minutes have elapsed AND communications with Dispatcher CANNOT be established, THEN REDUCE load UNTIL grid frequency reaches 60.0 Hz (1800 rpm).

CONTINUED SQN O-AR-ECB6-B Page 26 of 44 0 Rev. 21

19 (C-5)

CORRECTIVE STA FREQ ACTIONS (CONTINUED) EXCESSIVE ERROR

[6] IF frequency is less than 59.9 Hz (1797 rpm), THEN PERFORM the following:

[a] IF reactor power is less than 95%, THEN EVALUATE raising turbine load at 1% per minute UNTIL grid frequency reaches 59.9 Hz. ( 1797 rpm)

OR reactor power is at 100%.

[b] ATTEMPT to establish communications with Dispatcher (751-4208).

[c] IF 5 minutes have elapsed AND communications with Dispatcher CANNOT be established AND reactor power is less than 100%, THEN EVALUATE raising turbine load UNTIL grid frequency reaches 60.0 Hz (1800 rpm) OR reactor power is at 100%.

NOTE The following step attempts to recover frequency by manually dropping loads. Lines with the highest outgoing megawatts should be opened first.

[7] IF frequency drops to 58.0 Hz (1740 rpm) and continues to drop OR remains at 58 Hz for 1.5 minutes, THEN PERFORM the following:

[a] MONITOR grid frequency.

[b] OPEN 161 kv and 500 kv switchyard line PCBs one at a time.

[c] IF grid frequency begins to recover, THEN DO NOT OPEN remaining line PCBs.

[d] IF off-site power is lost to shutdown boards, THEN PERFORM AOP-P.01, Loss of Off-Site Power.

CONTINUED SQN O-AR-ECB6-B Page 27 of 44 0 Rev. 21

19 (C-5)

CORRECTIVE STA FREQ ACTIONS (CONTINUED) EXCESSIVE ERROR NOTE The following step is based upon engineering recommendations for protecting motors during abnormal frequency operation.

[8] IF grid frequency remains less than 57.0 Hz OR greater than 63.0 Hz, THEN EVALUATE placing shutdown boards on diesel generators as follows:

[a] EMERGENCY START diesel generators.

[b] VERIFY normal voltage and frequency on DIG outputs.

[c] DE-ENERGIZE shutdown boards one-at-a time and ENSURE D/G breakers close to energize shutdown boards.

[d] PERFORM AOP-P.01, Loss of Off-Site Power.

CAUTION IF communications cannot be established with the Dispatcher, then the grid should be considered to be in an unanalyzed condition and both off-site power sources declared inoperable.

[9] CONTACT Dispatcher to verify system disturbance and to receive instructions to assist in restoring system to normal frequency. (751-4208)

[10] EVALUATE operability of off-site power sources (LCO 3.8.1.1).

References 45N541, 45B655-ECB6-B SQN 0-AR-ECB6-B Page 28 of 44 0 Rev. 21

SQN TURBINE SHUTDOWN WITHOUT 0-GO-i i REACTOR SHUTDOWN Rev: 29 1&2 Page7of78 3.1 Precautions (Continued)

P. Voltage Control NOTE Failure to comply with the following NERC VAR-002 requirement could lead to a Utility Violation and I or monetary penalties.

i. Operation of the Main Generator without automatic Voltage Control could impact grid voltage requirements. Refer to GOI-6 for MVAR Limits.

2. When the generator is connected to the grid, the Voltage Regulator shall be operated in Automatic unless coordinated with the Transmission Operator.

3. Main Generator operation outside of the Transmission Voltage Schedule requires coordination with the Transmission Operator, an entry into the Operators Narrative Log of the time, reason and that the Transmission Operator notification was made.

4. When direct to modify voltage the Generator Operator shall comply (within plant procedural requirements) or provide an explanation of why the schedule can not be met.

5. While the Main Generator is tied to the grid:

a. The Transmission Operator (SELD) shall be notified of any voltage regulator automatic trips to Manual or urgent transfers between Auto and Manual as soon as practical but within 30 minutes.

b. The Transmission Operator (SELD) shall be notified prior to a planned voltage regulator transfer between Auto and Manual.

c. All position changes (Auto or Manual) of the voltage regulator shall be entered into the Narrative Log and shall include date and time of position change, reasons, anticipated duration and notifications made. [C.5]

SQN TURBINE SHUTDOWN WITHOUT 0-GO-I I REACTOR SHUTDOWN Rev: 29 l&2 Page8of78 3.1 Precautions (Continued)

Q. Reliability Directives and Protective Relay/Equipment Failures NOTE Failure to comply with the NERC VAR-002 requirement could lead to a Utility Violation and / or monetary penalties.

1. Plant Operations shall notify the Transmission Balancing Authority or Transmission Operator of protective relay or equipment failures that create a creditable risk to Plant Generation. A creditable risk to generation represents a potential reduction in transmission system reliability.

2. Reliability Directives tot he Generator Operator are via the Transmission Balancing Authority or Transmission Operator. Required action time may range from immediate to no longer than 30 minutes. Actions shall be taken without delay. The directives my be associated with preventing or clearing Local System issues or neighboring system issues.

3. Plant Operations shall take timely actions as directed by the Transmission Balancing Authority or Transmission Operator to mitigate critical conditions to return the bulk electrical system to a reliable state. Plant Operations shall comply with the Transmission Balancing Authority or Transmission Operator directives unless such actions would violate safety, equipment, regulatory or statutory requirements.

4. Plant Operations shall immediately inform the Transmission Balancing Authority or Transmission Operator of the inability to perform directives so that the TVA Reliability Entities may implement alternate remedial actions.

12 (B-5)

Source Setpoint SER 921(1-7) 6900VSDBD

1. 86S1A Differential Lockout Relay. 1. Tripped position IA-A FAILURE OR

2. DCS1A Loss of Normal 125V DC 2. Complete loss of Control Voltage Relay. voltage. BUS UNDERVOLTAGE

3. DCS1AE Loss of Emergency 125V DC 3. Complete loss of /OVE RVO LTAG E Control Voltage Relay. voltage.

4. DCDAT Loss of Control Voltage to the 4. Complete loss of Degraded and Loss of Voltage Relays voltage.

5. DA-1 Degraded Voltage Alarm Relay. 5. 93.5% for 30 secs.

6. DA-2 Degraded Voltage Alarm Relay. 6. 93.5% for 30 secs.

7. LV Undervoltage Relay. 7. 80% for 1.25 secs.

SER 914(8-10)

8. 59 DAT A 4) solid state overvoltage 8. 105% (7260).

relay.

9. 59 DBT B 4) solid state overvoltage 9. 105% (7260).

relay.

10. 59 DCT C 4) solid state overvoltage 10. 105% (7260).

relay.

Probable 1. lA-A Shutdown Bd Differential.

Causes 2. Loss of normal 125 volt dc control power.

3. Loss of auxiliary 125 volt dc control power.

4. Loss of Control Voltage to the Degraded and Loss of Voltage Relays.

5. Degraded voltage of 93.5% for 30 seconds.

6. Bus voltage less than 80% for 1.25 secs.

7. Voltage sag on 500kV switchyard if backfeeding unit.

8. Bus voltage greater than 105%.

9. DIG voltage too high (if supplying shutdown bd).

10. Intertie Changer Control Switch (tap changer) failurelabnormal Corrective [I] IF reactor trip, THEN Actions GO TO E-0, Reactor Trip or Safety Injection.

[2] IF loss of offsite power, THEN GO TO AOP-P.01, Loss of Offsite Power.

[3] CHECK Voltage on [1-El-57-391 lA-A Shutdown Bd.

[4] IF lA-A Shutdown Board is deenergized, THEN GO TO AOP-P.05, Loss of Unit 1 Electrical Shutdown Boards.

[5] IF lA-A Shutdown Board is energized, but voltage is high, THEN EVALUATE overvoltage condition in accordance with GOl-6 Section F, Onsite and Offsite Power Requirements.

(Continued on Next Page)

SQN 1-AR-MI-B Page 18 of 50 I Rev. 22

12 (B-5) 6900VSD BD IA-A FAILURE OR BUS UNDERVOLTAGE lOVE RVO LTAG E Corrective Action (Continued)

[6] CHECK alarms on O-XA-55-26A AND PERFORM the following:

[a] IF failure/undervoltage from window 0-7 on O-XA-55-26A is valid, THEN PERFORM the corrective actions found in that annunciator response.

[b] IF overvoltage from window B-7 on O-XA-55-26A is valid, THEN PERFORM the corrective actions found in that annunciator response.

[7] IF blown fuse, THEN REPLACE fuse in accordance with O-SO-202-4, Power Availability Checklist.

[8] IF voltage normal, THEN INVESTIGATE DC Control Power failure.

[9] EVALUATE impact on plant Technical Specification requirements of LCOs 33.3. 11, 3.8.2.1, and 3.8.2.2.

References 45N724-1, 45N765-1, 45N765-3, 45N765-15, 45B655-O1B-O SQN 1-AR-MI-B Page 19 of 50 I Rev 22

14 (B-7)

Source Setpoint 6900 SD BD SER914 IA-A

1. 59 DAT A 4) solid state overvoltage relay OVERVOLTAGE

2. 59 DBT B 4) solid state overvoltage relay 105% (7260 volts)

3. 59 DCT C 4) solid state overvoltage relay Probable 1. Grid voltage too high.

Causes 2. D/G voltage too high (if supplying shutdown bd).

3. Intertie Changer Control Switch (tap changer) failure/abnormal Corrective [1] CHECK lO-EI-82-6A1, 6900V SD BD lA-A Voltage to verify Actions overvoltage condition via selector switch IO-XS-82-6A1.

[2] IF overvoltage condition, THEN ADJUST voltage within limits of GOl-6 with the following:

a. lntertie Transformer Bank Tap Changer

b. Units Voltage Regulators

c. Contact SELD for help (Capacitor bank(s) ect.)

[3] EVALUATE impact on plant Technical Specification requirements of 3.8.1.1, 3.8.1.2, 3.8.2.1, and 3.8.2.2.

[4] IF 0-AR-ECB6-A CSS Trans. Auto Tap Changer is in alarm, THEN PERFORM the corrective actions in 0-AR-ECB6-A.

References 45N765-1, 45B655-26A SQN O-AR-M26-A Page 19 of 43 1&2 Rev.26

SQN NORMAL POWER OPERATION 0-GO-5 Unit I & 2 Rev. 0065 Page 10 of 100 31 Precautions (continued)

R. The following limitations are applicable to Unit Two ONLY.

1. In winter months #7 HDTP capacity is not adequate to pump #6 Heater drains when all Condensate Demineralizer pumps are in service. Current practice is to run two Cond DI Pumps and / or throttle the condensate system to reduce backpressure. The preferred method is to throttle condensate pressure instead of running only two Condensate Demineralizer booster pumps at full power due to pump runout concerns.

2. Siemens-Westinghouse analysis has determined that the maximum unit power with one MFP operation is 65% under worst case conditions. The plant could operate higher if plant conditions permit.

3. MFP flow from the lead MFP should not exceed 53.7% of the total flow.

Flow rates above this would result in HP steam flow to the lead MFPT.

Computer points 1(2)U0504 and UO505 can be used to monitor.

S. Voltage Control NOTE Failure to comply with the NERC VAR-002 requirement could result in a Utility Violation and I or monetary penalties.

1. Operation of the Main Generator without Automatic Voltage Control could impact grid voltage requirements. Refer to GOI-6 for MVAR limits.

2. When the Main Generator is connected to the grid, the voltage regulator shall be operated in Automatic, unless coordinated with the Transmission Operator (SELD).

3. Main Generator operation outside of the Transmission Voltage Schedule requires coordination with the Transmission Operator, and notation in the operators Log of time, reason, and that the Transmission Operator notification was made.

4. When directed to modify voltage, the Generator Operator shall comply (within plant procedural requirements) or provide an explanation of why the schedule cannot be met.

5. While the Main Generator is tied to the grid perform the following:

a. The Transmission Operator (SELD) shall be notified of any Voltage Regulator automatic trips to Manual or urgent Manual Transfers between AUTO and Manual as soon as practical but notification shall be within 30 minutes.

SQN NORMAL POWER OPERATION 0-GO-5 Unit I & 2 Rev. 0065 Page 11 of 100 3.1 Precautions (continued)

b. The Transmission Operator (SELD) shall be notified prior to a planned Voltage Regulator transfers between Manual and Auto.

c. All position changes (to and from Auto or Manual) of the Voltage Regulator shall be entered into the Narrative Log along with the date, time of position change, reasons, anticipated duration and notifications made.

T. Reliability Directives and Protective Relay/Equipment Failures NOTE Failure to comply with the NERO VAR-002 requirement could result in a Utility Violation and / or monetary penalties.

1. Plant Operations shall notify the Transmission Balancing Authority (BA) or Transmission Operator of protective relay or equipment failures that creates a creditable risk to Plant Generation. A creditable risk to generation represents a potential reduction in transmission system reliability.

2. Reliability Directives to the Generator Operator are via the Balancing Authority or Transmission Operator. Required action time may range from immediate to no longer than 30 minutes. Actions shall be taken without delay. The directives may be associated with preventing or clearing Local System issues, or neighboring system issues.

3. Plant operations shall take timely actions as directed by the Balancing Authority or Transmission Operator to mitigate critical conditions to return the bulk electrical system to a reliable state. Plant operations shall comply with Balancing Authority or Transmission Operator directives unless such actions would violate safety, equipment, or regulatory or statutory requirements.

4. Plant Operations shall immediately inform the Balancing Authority or Transmission Operator of the inability to perform directives so that the TVA Reliability Entities may implement alternate remedial actions.

SQN NORMAL POWER OPERATION 0-GO-5 Unitl&2 Rev.0065 Page 12 of 100 3.2 Limitations A. When the axial flux difference monitor alarm is inoperable, the AFD must be logged every hour by performing 0-SI-NUC-000-044.0.

(SR4.2.1.1.a.2 &4.2.1.1.b)

B. When both the plant computer and NIS QPTR alarm systems are inoperable, the QPTR must be calculated every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by performing 0-Sl-NUC-000-133.O. (SR 4.2.4.1.b)

C. Do not exceed a load change rate of plus or minus 5% per minute or a step change of 10%.

D. River water temperatures shall be maintained within the limitations of the NPDES permit as specified in 0-PI-OPS-000-666.O.

NOTE Westinghouse should be contacted if the turbine is operated outside of its operating limits as stated below.

E. To prevent high vibratory stresses and fatigue damage to the last stage turbine blading, do not operate the turbine outside of limits listed below:

[W Ltr GP 86-02 (B44 861112002)]

1. At loads less than or equal to 30% (350 MW), the maximum permissible backpressure is 1 .72 psia. (3.5 Hg)

2. At loads greater than 30%, the maximum permissible backpressure is 2.7 psia (5.5 Hg) with a 5 minute limitation before tripping the turbine.

F. Do not allow the generator to become underexcited.

G. In the event of a change in the rated thermal power level exceeding 15% in one hour, notify Chemistry to initiate the conditional portions of 0-SI-CEM-000-050.0, 0-SI-CEM-030-407.2 and 0-SI-CEM-000-415.0 due to the thermal power change.

OPERATIONS OPTSTG200.SWYD

! SWITCHYARD REV 0 STUDENT TRAINING GUIDE PAGE 13 OF 93 ARSS UORT 0 OA X X X 6. EXPLAIN the operational implication of the following concepts as they apply to the Switchyards:

a. Effects of Intertie transformer tap setting on both SWYD voltages

b. Effect of capacitor banks on SWYD voltage X X X 7. Given specific plant conditions, ANALYZE the effect that a loss or malfunction of the Switchyards will have on the offsite power circuits.

X X X 8. Given specific plant conditions, ANALYZE the effect that a loss or malfunction of the following will have on the Switchyards:

a. Air compressors

b. 25OVDC

c. Transmission line X X X X 9. DESCRIBE the design and operational differences between the 500kV and 161 kV switchyards as they apply to PCBs and MOD5.

X X X 10. Given a specific evolution, IDENTIFY the appropriate Switchyard normal operating procedure(s) required to conduct that evolution.

X X X 1 1. Using the Technical Specifications,

a. LIST from memory, Switchyard Tech Spec LCOs having action times one hour,

b. EXPLAIN applicable Switchyard Tech Spec LCO bases.

c. Given a set of plant conditions/parameters, DETERMINE entry level conditions for Switchyard Tech Spec LCO actions.

X X X 12.ANALYZE a given set of plant conditions and/or parameters to determine entry conditions into abnormal or emergency operating procedures.

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta

82. Given the following plant conditions:

- Unit 1 at 95% power when the OATC observes the control rods continuously stepping out at 8 steps/mm.

Which ONE of the following are (1) indications that are consistent for this condition and (2) the required procedural guidance?

Note: FCV-62-138, Emergency Boration Flow Control Valve A. (1) Red light on handswitch for FCV-62-138 LIT and flow indicated on emergency borate flow indicator Fl-62-137A.

(2) Place rod control selector switch to MAN and maintain Tavg within 1°F of Tref in accordance with AOP-C.01, Rod Control System Malfunction.

B (1) Red light on handswitch for FCV-62-1 38 LIT and flow indicated on emergency borate flow indicator Fl-62-137A.

(2) Manually close FCV-62-1 38, and reduce turbine load as necessary to maintain Tavg within 3°F of Tref, in accordance with AOP-C.02, Uncontrolled RCS Boron Concentration Changes.

C. (1) Red and Green lights on handswitch for FCV-62-138 on Panel M-6 LIT and NO flow indicated on emergency borate flow indicator Fl-62-137A.

(2) Place rod control selector switch to MAN and maintain Tavg within 1°F of Tref in accordance with AOP-C.01, Rod Control System Malfunction.

D. (1) Red and Green lights on handswitch for FCV-62-1 38 on Panel M-6 LIT and NO flow indicated on emergency borate flow indicator Fl-62-137A.

(2) Manually close FCV-62-1 38, and reduce turbine load as necessary to maintain Tavg within 3°F of Tref, in accordance with AOP-C.02, Uncontrolled RCS Boron Concentration Changes.

Tuesday, July 13, 2010 8:06:46 AM 82

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta DIS TRACTOR ANAL YSIS:

A. Incorrect, Plausible since first part is correct a slow boration is taking place, however the procedure guidance should be from AOP-C.02 for uncontrolled RCS boration not AOP-C.O1 due to rod control problems.

B. Correct, The indications would indicate that an uncontrolled boration is taking place causing a negative reactivity to be added to the RCS, thus outward rod motion to compensate. The correct procedure for this condition is AOP-C.02.

C. Incorrect, Plausible since Red and Green lights would indicate that the emergency boration valve is not full closed, and with some indicated flow then a boration would be taking place. However the correct procedure would be AOP-C.02.

D. Incorrect, Plausible since Red and Green lights would indicate that the emergency boration valve is not full closed, however with no indicated flow then a boration would not be occurring. The correct procedure for an uncontrolled boration is A OP-C. 02.

Tuesday, July 13, 2010 8:06:46 AM 83

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta Question Number: 82 Tier: 1 Group 2 KIA: 001 Continuous Rod Withdrawal AA2.02 Ability to determine and interpret the following as they apply to the Continuous Rod Withdrawal:

Position of the emergency borate valve Importance Rating: 4.2 / 4.2 10 CFR Part 55: n/a IOCFR55.43.b: 5 K/A Match: Question matches the K/A by having the candidate determine the position of the emergency borate valve while the control rods are withdrawing in an uncontrolled fashion and the selection of the procedure needed to control/mitigate the event.

Technical

Reference:

AOP-C.02 revOO6 1 ,2-47W809-2A Proposed references None to be provided:

Learning Objective: OPT200.CVCS Rev 8 0 bj B.2 & 14 Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: Sequoyah 2000 RO Exam Question Comments:

MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9 Answer: BDBAABDACC ScrambleRange: A-D Tuesday, July 13, 2010 8:06:46 AM 84

SQN UNCONTROLLED RCS BORON CONCENTRATION AOP-C02 CHANGES Rev. 6 STEP ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED 2.0 OPERATOR ACTIONS

1. DIAGNOSE the failure:

GOTO IF._ SECTION PAGE Uncontrolled or unplanned boration in Mode 1 or 2 2.1 4 Uncontrolled or unplanned dilution in Mode 1 or 2 2.2 11 Uncontrolled or unplanned dilution in Mode 3, 4, 5, or 6 2.3 19 Page 3of30

SQN UNCONTROLLED RCS BORON CONCENTRATION AOP-C.02 CHANGES Rev. 6 STEP I ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 Uncontrolled Boration in Mode I or 2 (contd)

NOTE Manually withdrawing control rods in response to an uncontrolled boration from an unknown source is NOT conservative reactivity management.

Automatic rod motion (if available) is acceptable.

MONITOR reactor power and T-avg:

a. VERIFY reactor power STABLE a. IF reactor power is rising or DROPPING. AND T-avg is dropping, THEN GO TO AOP-S.05, Steam or Feedwater Line Break/Leak.

b. CHECK reactor power b. IF reactor power is less than 1 %

greater than 5%. and DROPPING, THEN TRIP reactor and GO TO E-0, Reactor Trip or Safety Injection.

c. CHECK T-avg greater than 541 °F. c. IF T-avg less than 541 °F, (LCO3.1.1.4) THEN TRIP reactor and GO TO E-0, Reactor Trip or Safety Injection.

d. REDUCE turbine load as necessary d. IF turbine is loaded to maintain T-ref within 3°F of T-avg. AND T-avg/T-ref mismatch CANNOT be maintained less than 5°F, THEN TRIP reactor and GO TO E-0, Reactor Trip or Safety Injection.

Page 4 of 30

SQN UNCONTROLLED RCS BORON CONCENTRATION AOP-C.02 CHANGES Rev. 6 r STEP I ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 Uncontrolled Boration in Mode I or 2 (contd)

2. ENSURE CCP suction ALIGNED to VCT:

a. VERIFY VCT level is greater than 13%. a. RESTORE VCT level by automatic or manual makeup USING O-SO-62-7, Boron Concentration Control.

b. ENSURE CCP suction valves b. IF CCP suction CANNOT be realigned aligned as follows: to VCT, THEN

a. LCV-62-132 and LCV-62-133 TRIP reactor and GO TO E-O, OPEN. Reactor Trip or Safety Injection.

b. LCV-62-135 and LCV-62-136 CLOSED.

Page 5of30

SQN UNCONTROLLED RCS BORON CONCENTRATION AOP-C.02 CHANGES Rev. 6 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 Uncontrolled Boration in Mode I or 2 (contd)

3. STOP boration flow:

a. ENSURE Boric Acid Transfer pumps a. STOP boric acid transfer pumps.

in slow speed

b. ENSURE FCV-62-138, Emergency b. STOP boric acid transfer pumps.

Boration Flow Control Valve, CLOSED.

DISPATCH operator to locally close FCV-62-138. [AB el. 690]

c. VERIFY NO emergency boration flow c. STOP boric acid transfer pumps.

indicated on FI-62-137A. [M-6]

DISPATCH operator to locally close FCV-62-138. [AB el. 690]

d. ENSURE FCV-62-140 CLOSED. [M-6] d. STOP boric acid transfer pumps.

e. VERIFY NO boration flow indicated e. STOP boric acid transfer pumps.

on Fl-62-139A. [M-6]

Page 6 of 30

SQN UNCONTROLLED RCS BORON CONCENTRATION AOP-C.02 CHANGES Rev. 6 STEP ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 Uncontrolled Boration in Mode I or 2 (contd)

NOTE CVCS demin beds may cause uncontrolled boration by either of the following:

Rise in letdown temperature (due to failure of letdown temperature controls) resulting in release of boron by in-service resin OR

Placing a demin bed in service which was previously in-service with high RCS boron concentration.

4. CHECK if uncontrolled boration may be due to CVCS demin beds:

a. CHECK for either of the following: a. GO TO Step 5.

rise in letdown HX outlet temperature

[M-6 or lOS point T0145A]

OR

demin bed recently placed in service.

b. PLACE HS-62-79A, Mixed Bed Hi Temp Bypass to V.C. TK position. [M-6]

Page 7 of 30

SQN UNCONTROLLED RCS BORON CONCENTRATION AOP-C.02 CHANGES Rev. 6 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 Uncontrolled Boration in Mode I or 2 (contd)

NOTE A rise in xenon concentration may cause the same symptoms as an uncontrolled boration.

5. CHECK uncontrolled boration source PERFORM the following:

IDENTIFIED and STOPPED.

a. ENSURE boric acid transfer pumps STOPPED.

b. DISPATCH operator to ensure the following valves are CLOSED:

62-929

62-932

[AB el. 690 pen rm]

6. EVALUATE the following Tech Spec and TRM sections for applicability:

TRM 3.1.2.6, Borated Water Sources

LCO 3.5.5, Refueling Water Storage Tank

7. INITIATE RCS dilution USING 0-SO-62-7 as necessary to restore the following:

T-avg

Al (%A Flux)

Control rod position Page 8 of 30

SQN UNCONTROLLED RCS BORON CONCENTRATION AOP-C.02 CHANGES Rev. 6 STEP ACTION!EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 Uncontrolled Boration in Mode I or 2 (contd)

8. NOTIFY Chem Lab to determine the following boron concentrations:

RCS

Pressurizer

9. CONTROL pressurizer heaters and sprays to maintain RCS and pressurizer boron concentrations within 50 ppm.

10. INITIATE actions to determine and correct cause of uncontrolled boration.

11. NOTIFY Plant Management and Reactor Engineering of reactivity event.

Page 9 of 30

SQN UNCONTROLLED RCS BORON CONCENTRATION AOP-C.02 CHANGES Rev. 6 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED I 2.1 Uncontrolled Boration in Mode I or 2 (contd)

12. WHEN dilution is complete AND uncontrolled boration source is isolated, THEN ALIGN blender controls for automatic makeup USING 0-SO-62-7, Boron Concentration Control.

13. IF CVCS demin beds were bypassed in Step 4, THEN EVALUATE actions required to restore demin beds USING 1,2-SO-62-9.

14. GO TO appropriate plant procedure.

END OF SECTION Page 10 of 30

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93IFEO2A3T8AU1811A.UI

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62NJOT A.U2

2. ALL PROBLEM NAMAERS NRC PREFIXED N2 -

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/

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52 93 6 C 6234.0 I A2 62fl378N.Al 52440 62A 3DWN *2 LETDOWN HEAT 3*2 SIS RWST 2936 ORIFICE DESIGN PRESISOPSIG EXCHANGER 4798111 DESIGN TEMP200F NH I COORD DN AND/OR COORD *12 62-NOD 4798 121 0-82-924 COORD CN WA. *1 621130811 02 0-62-926 ii

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2 VENT REV I CHANCE REF 62918 3 Sn 2 REACTOR COOLANT 3 62917 COORD 02 POWERHOUSE FILTER C- OH I CATION UNITS 1 & 2 COORO *12 BED DEMINERALIZER DESIGN PAES200PSIC DESIGN TEMP2S0F 5 CONNECTIVITY DIAGRAM A227A.A2 62A31 2A. 02 B23A.Al A2-A3BDN 01 62-916 CHEMICAL AND VOLUME CONTROL SYSTEM SEQUOYAI-f NUCLEAR PLANT TENNESSEE VALLEY AUTHORITY DRAIN ._J.i ..C. _.1 RESDN DISCHARGE 18161*0 ISSUE K AC SSEEW DUD SH2 5H2 CcORD 03 COORD F2 I ISSUE PER: ODD MADG EBOD ICR 506123* AND 62920 2 0 1 62919 62921 1 8.8. *09888 DONS NOSIBSR7W RD. SONIWDDSW BA.

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OPT200.CVCS Rev 8 Page 7 of 251 V. TRAINING OBJECTIVES:

B. Enabling Objectives, Continued:

ARSS UORT 0 OA X X X X 2. DESCRIBE the location of the following listed CVCS components, including switches, controls, and indications to determine that they correctly reflect the desired plant configuration. Include main control room, auxiliary control room, and local panels as applicable.

a. CCPs

b. Regenerative Hx

c. Letdown orifice isolation valves

d. Letdown Hx

e. VCT divert valve LCV-62-1 18

f. Volume control tank (VCT)

g. Excess letdown Hx

h. Demineralizer divert valve TCV-62-79

i. Mixed bed and cation bed demineralizers

j. Reactor coolant filter

k. Reactor Makeup Control System (blender controls)

I. Seal water injection filters

m. Seal water (return) filter

n. Seal water Hx

o. RCP No. 1 seal bypass valve FCV-62-53

p. Boric acid storage tanks (BATs)

q. Boric acid transfer pumps

r. Emergency borate valve FCV-62-138, alternate borate valve VLV-62-929, and normal boric acid supply to blender valve, FCV 140 S. Emergency borate flow indicators FT-62-137 & FT-62-139

t. Holdup tanks (HUT)

OPT200.CVCS Rev 8 Page 13 of 251 V. TRAINING OBJECTIVES:

B. Enabling Objectives, Continued:

A R S S U 0 R T 0 0 A X X X X 13. Given a specific evolution, IDENTIFY the appropriate CVCS normal operating procedure(s) required to conduct that evolution.

X X X 14. ANALYZE a given set of plant conditions and/or parameters to determine entry conditions into abnormal or emergency operating procedures.

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta

83. Given the following:

- Both units are operating at 100% power

- A LOCA occurs on Unit 1

- Unit 2 remains at 100% power

- During the performance of E-1, Loss of Primary or Secondary Coolant, the STA notes that CNMT rad levels are 110 R/hr and recommends entry into FR-Z.3, High Containment Radiation.

Which ONE of the following identifies both; (1) the expected alignment of the Emergency Gas Treatment System (EGTS) 20 minutes after the initiation of LOCA and (2) the required Tech Spec action for Unit 2 A. (1) Both trains of EGTS are in service (2) LCO 3.6.1.8, Emergency Gas Treatment System, EGTS, Cleanup Subsystem B (1) Both trains of EGTS are in service (2) LCO 3.0.3 C. (1) Only ONE train of EGTS is in service, the other train in Standby (2) LCO 3.6.1.8, Emergency Gas Treatment System, EGTS, Cleanup Subsystem D. (1) Only ONE train of EGTS is in service, the other train in Standby (2) LCO3.0.3 Tuesday, July 13, 2010 8:08:27 AM 83

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta DISTRACTOR ANALYSIS:

A. Incorrect, Plausible since both trains of EGTS automatically start on a phase A of either unit, however after 30 minutes of operation (or per TSC guidance) one train is to be placed in standby lAW EA-65-1. Also T.S. 3.6.1.8 would apply if only one train is OOS however both trains are aligned to the accident unit.

B. Incorrect, Plausible since both trains of EGTS automatically start on a phase A of either unit, however after 30 minutes of operation (or per TSC guidance) one train is to be placed in standby. Also T. S. 3.0.3 applies since both trains of EG TS are unavailable to the non-accident unit.

C. Incorrect, Plausible since the guidance in EA-65-1,(or with TCS guidance) has the operator shut down one train of EG TS and place it in Standby.Also T. S. 3.6. 1.8 would apply if only one train is OOS however both trains are aligned to the accident unit.

D. Correct, Per the guidance in EA-65-1, 30 minutes after initiation of EGTS one train is to be aligned for decay cooling and the other continues in service. Also per SO-65-1, anytime either train is in operation or placed in standby operation it is inoperable for both units.

Tuesday, July 13, 2010 8:08:27 AM 84

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta Question Number: 83 Tier: 1 Group 2 KIA: 060 Accidental Gaseous Radwaste Release AG 2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

Importance Rating: 4.0 I 4.6 10 CFR Part 55: 41.7 IOCFR55.43.b: (2) (5)

K/A Match:

Technical

Reference:

EA-65-1 Rev 1, EA-0-1 rev 10, T.S. 3.6.1.8 and 3.0.3 0-SO-65-1, Rev 19 Proposed references None to be provided:

Learning Objective: OPT200. EGTS Rev 3 Obj 6 Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question for 1009 NRC exam Comments:

MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9 Answer: BBCADCCCDD ScrambleRange:A-D Tuesday, July 13, 2010 8:08:27 AM 85

SQN EMERGENCY GAS TREATMENT 0-SO-65-1 Unit 0, 1, & 2 SYSTEM Rev. 0019 AIR CLEANUP AND ANNULUS Page 6 of 54 VACUUM 3.0 PRECAUTIONS AND LIMITATIONS A. Manual Q Automatic operation of EGTS while in service for either unit makes it inoperable for opposite unit.

B. When EGTS Air Cleanup is placed in service on unit where personnel are in annulus, it becomes an ALARA concern if other unit is in Mode 1-4 and receives Phase A Isolation signal. When EGTS Air Cleanup is placed in service on unit where annulus is open, it becomes an OPERABILITY concern if other unit is in Mode 1-4 and receives Phase A Isolation signal. Assignment and action of dedicated operator within this instruction provides protection for personnel in annulus and will ensure operability of EGTS on other unit. [C.3]

C. The following guidance is provided regarding operability impact when any EGTS decay cooling valve is inoperable and EGTS fans are capable of starting (Refer to Precaution and Limitation M when one EGTS train is Out of Service):

When EGTS decay cooling valve 0-FCV-65-28A or -28B is inoperable, then EGTS Train B is inoperable. When decay cooling valve 0-FCV-65-47A or -47B is inoperable, then EGTS Train A is inoperable. This operability position is based upon the following:

(1) 0-FCV-65-28A and -28B are powered from DC Channel II and IV (Train B) and are supplied from Train B essential air and, therefore, should be tied to operability of EGTS Train B to maintain train separation Similarly, 0-FCV 47A and B are supplied from Train A.

(2) FSAR 6.2.3.3.2 describes the functional requirement for the running EGTS train to provide cooling flow to the inactive filter bank to maintain temperatures within acceptable limits. Therefore, the decay cooling function is treated as a support function which must be provided by the operable train.

D. Either Train A flow divider dampers I (2)-PCO-65-80 &

I (2)-PCO-65-88 or Train B flow divider dampers I (2)-PCO-65-82 &

I(2)-PCO-65-89 will modulate to control vacuum in annulus at 0.5 inch of H

- 0, 2

whenever EGTS Air Cleanup is operating.

E. Normal EGTS damper operating limits are 0.2 to 1.2 inches of H

- - O zP 2

between annulus and outside pressure (atmospheric). Capability to automatically initiate high P swapover to opposite train zP control can only be armed by Phase-A Containment Isolation Signal after AP reduces to less negative than 0.7 inches H

- 0 or 4 minutes, 10 seconds, after Phase-A signal 2

initiation.

F. Do not depress EGTS Air Cleanup damper local RESET switches during an accident. The swapover control circuit would be blocked by reset action and would r be able to reach arming setpoint.

SQN EMERGENCY GAS TREATMENT 0-50-65-1 Unit 0, 1, & 2 SYSTEM Rev. 0019 AIR CLEANUP AND ANNULUS Page 7 of 54 VACUUM 3.0 PRECAUTIONS AND LIMITATIONS (continued)

NOTE When venting containment, it is normal for I (2)-PCO-65-93(-96) and I (2)-PCO-65-94(-95) to swap because annulus pressure drops below swapover setpoint.

G. If normally open damper (2)-PCO-65-93(-96) is closed and 1(2)-PCO-65-94(-95) is open while Annulus Vacuum is in service, failure of normal annulus pressure control may be indicated. Normal pressure control damper should be returned to service to provide normal and backup pressure control.

H. Anytime Containment or Shield Building integrity is required, Annulus Vacuum Fan A or B is required to be in service, maintaining equal to or more negative than 5 inches H

- 0 in annulus.

2 I. Both trains of EGTS Air Cleanup are required to be operable for operation of either unit in modes I through 4.

J. In the event of a Safety Injection on either unit, discharge flowpath of Annulus Vacuum Control Fans will isolate due to ABSCE damper closure. During any changes to Annulus or Aux Bldg ventilation, pressure setpoints may. be reached causing containment vacuum relief valve(s) to OPEN as designed.

K. An ABI or FHAI will isolate discharge flow path of Annulus Vacuum Control Fans for both units, without tripping fans.

L. A containment vent per 0-SO-30-8 may be required during EGTS operation or immediately prior to placing Annulus Vacuum Control Fans in service in order to prevent exceeding 0.3 psi containment differential pressure as specified in Tech Spec 3.6.1.4.

M. With one train (fan) of EGTS out of service (incapable of starting) prior to an accident, the decay cooling valves (0-FCV-65-28A,B and -47A,B) are no longer needed to perform their required function. The decay cooling valves for the filter train that is out of service are not required because the filter train is not running, and it is assumed that the filter train has not been exposed to radioactive material post-accident. Additionally, the operable filter train remains running and does not require decay cooling. Therefore, if one train of EGTS is tagged, both trains decay cooling valves may be tagged closed without affecting the operability of the remaining train.

SQN EGTS OPERATION II EA-65-1 I I

I Rev. I 1,2 Page2ofl7 1.0 PURPOSE To establish proper EGTS operation following reactor trip and safety injection.

2.0 SYMPTOMS AND ENTRY CONDITIONS 2.1 Entry Conditions A. ECA-O.2, Recovery From Loss of All AC Power With SI Required.

B. FR-Z.1, High Containment Pressure.

C. FR-Z.3, High Containment Radiation.

3.0 PRECAUTIONS AND LIMITATIONS 3.1 Precautions A. If the accountabiUty siren sounds, the operator should continue performing this procedure. The SM will remain aware of procedure progress and location of performing personnel.

SQN EGTS OPERATION II EA-65-1 II IRev,1 I 1,2 Page3ofl7 4.0 OPERATOR ACTIONS 4.1 Section Applicability

1. SELECT unit requiring EGTS operation:

Uniti____

Unit2_____

2. IF Unit I selected, THEN GO TO Section 4.2.

3. IF Unit 2 selected, THEN GO TO Section 4.3.

4. WHEN both EGTS trains have been in service greater than 30 minutes, THEN GO TO Section 4.4.

5. RETURN TO procedure and step in effect.

SQN I EGTS OPERATION I

I EA-65-1 I I Rev. I 1,2 Page4ofl7 4.2 EGTS Air Cleanup Fan And Damper Alignment For Unit I

1. ENSURE at least one EGTS fan RUNNING:

EGTS FAN EGTS Fan A-A (O-HS-65-23A)

EGTS Fan B-B (O-HS-65-42A)

2. ENSURE EGTS discharge damper(s) for running fan(s) OPEN:

FAN j EGTS DISCHARGE DAMPER OPqEN A-A O-FCV-65-24 U B-B O-FCV-65-43 U

SQN EA-65-1 EGTS OPERATION I I 1,2 lRev.l Page5ofl7 I

4.2 EGTS Air Cleanup Fan And Damper Alignment For Unit I (Continued)

NOTE In the next two steps, normally the set of dampers (8 1/86 or 83/87) in A-AUTO are OPEN and the other set in A-AUTO/STANDBY are CLOSED. The set in A-AUTO/STANDBY will open if the annulus pressure goes out of the range of 0.2 to 1 .2 inches of water vacuum, and the set in A-AUTO will be closed.

3. ENSURE EGTS Train A ALIGNED for Unit 1:

POSITION TRAIN A EQUIPMENT DESCRIPTION q

1-FCV-65-10 EGTS suction damper OPEN Li HS-65-77 Annulus Vacuum Fan IA STOPPED Li 1-FCO-65-52 Annulus Vacuum Fan IA CLOSED isolation damper Li 1-FCO-65-26 Shield Bldg Exhaust Damper A OPEN Li PCV-65-81 (see note) Shield Bldg Vent Isolation Damper A OPEN Li PCV-65-86 (see note) Annulus Isolation Damper A OPEN Li

4. ENSURE EGTS Train B ALIGNED for Unit 1:

TRAIN B EQUIPMENT DESCRIPTION POSITION q

1-FCV-65-30 EGTS suction damper OPEN Li HS-65-74 Ann ulus Vacuum Fan 1 B STOPPED Li 1-FCO-65-53 Annulus Vacuum Fan lB CLOSED isolation damper Li 1-FCO-65-27 Shield Building Exhaust Damper B OPEN Li PCV-65-83 (see note) Shield Bldg Vent Isolation Damper B OPEN Li PCV-65-87 (see note) Annulus Isolation Damper B OPEN Li

SQN EGTS OPERATION I EA-65-l I

I I

IRev,1 I 1,2 Page6ofl7 4.2 EGTS Air Cleanup Fan And Damper Alignment For Unit I (Continued)

5. ENSURE EGTS Train A ISOLATED to Unit 2:

POSITION TRAIN A EQUIPMENT DESCRIPTION

.1 2-FCV-65-9 EGTS suction damper CLOSED 0

2-FCO-65-46 Shield Building Exhaust Damper A CLOSED 0

2-PCV-65-81 Shield Bldg Vent Isolation Damper A CLOSED 0

2-PCV-65-86 Annulus Isolation Damper A CLOSED 0

6. ENSURE EGTS Train B ISOLATED to Unit 2:

TRAIN B EQUIPMENT DESCRIPTION POSITION q

2-FCV-65-29 EGTS suction damper CLOSED 0

2-FCO-65-45 Shield Building Exhaust Damper B CLOSED 0

2-PCV-65-83 Shield Bldg Vent Isolation Damper B CLOSED 0

2-PCV-65-87 Annulus Isolation Damper B CLOSED 0

SQN EGTS OPERATION I EA-65-1 I I

Rev. I 1, 2 Page 12 of 17 4.4 Establishing Single EGTS Train Operation

1. SELECT unit requiring EGTS operation:

Unitl

Unit2

2. SELECT unit NOT requiring EGTS operation:

Uniti

Unit2

3. IF Unit 1 selected in Step 4.4. 2. (EGTS operation NOT required),

THEN ENSURE the following Unit 1 dampers CLOSED:

UNIT I DAMPERS DESCRIPTION DAMPER CLOSED EGTS Train A suction damper 1-FCV-65-1O LI EGTS Train B suction damper 1-FCV-65-30 El Shield Bldg Exhaust Damper A 1-FCO-65-26 El Shield Bldg Exhaust Damper B 1-FCO-65-27 El Shield Bldg isolation damper 1-PCV-65-81 El Shield Bldg isolation damper 1-PCV-65-83 El Cont. Annulus isolation damper 1-PCV-65-86 El Cont. Annulus isolation damper 1-PCV-65-87 El

SQN EGTS OPERATION I EA-65-1 II I Rev. 1 1, 2 Page 13 of 17 4.4 Establishing Single EGTS Train Operation (Continued)

4. IF Unit 2 selected in Step 4.4. 2. (EGTS operation NOT required),

THEN ENSURE the following Unit 2 dampers CLOSED:

UNIT 2 DAMPERS DESCRIPTION DAMPER CLOSED EGTS Train A suction damper 2-FCV-65-9 LI EGTS Train B suction damper 2-FCV-65-29 LI Shield Bldg Exhaust Damper A 2-FCO-65-45 LI Shield Bldg Exhaust Damper B 2-FCO-65-46 LI Shield Bldg isolation damper 2-PCV-65-81 LI Shield Bldg isolation damper 2-PCV-65-83. LI Cont. Annulus isolation damper 2-PCV-65-86 LI Cont. Annulus isolation damper 2-PCV-65-87 LI

5. IF EGTS Train A flow switch rFS-65-55 BIA1 known to be inoperable, THEN SELECT EGTS Train B to be shut down in Step 4.4. 7. E

6. IF EGTS Train B flow switch rFS-65-31 B/Al known to be inoperable, THEN SELECT EGTS Train A to be shut down in Step 4.4. 7.

7. SELECT EGTS train to be shut down:

EGTS Train A

EGTS Train B

SQN EGTS OPERATION II EA-65-1 I lRev,1 1, 2 Page 14 of 17 4.4 Establishing Single EGTS Train Operation (Continued)

NOTE If phase A containment isolation is RESET, the standby EGTS fan will not auto start from a low flow on in-service fan.

If the EGTS fan HS is in PULL P-AUTO(STANDBY), the fan will not auto start on a phase A from either unit.

8. IF Train A to be shut down, THEN PERFORM the following:

a. PULL rO-HS-65-23A1 to P-AUTO(STANDBY) and momentarily TURN to STOP. LI

b. VERIFY 1O-HS-65-23A1 returns to P-AUTO(STANDBY). LI C. VERIFY Fan A-A STOPPED. LI

d. ENSURE EGTS Fan A discharge damper FO-FCO-65-241 CLOSED. El

e. IF Unit 1 selected for EGTS operation, THEN ENSURE EGTS Fan A-A suction dampers ALIGNED as follows:

UNIT I POSITION q EGTS SUCTION DAMPERS 1-FCV-65-8 OPEN El 1-FCV-65-1O CLOSED El

f. IF Unit 2 selected for EGTS operation, THEN ENSURE EGTS Fan A-A suction dampers ALIGNED as follows:

UNIT 2 POSITION -I EGTS SUCTION DAMPERS 2-FCV-65-7 OPEN El 2-FCV-65-9 CLOSED El

SQN EGTS OPERATION I EA-65-1 I IRev.1 I 1,2 Pagel5ofl7 4.4 Establishing Single EGTS Train Operation (Continued)

8. (Continued)

g. PLACE rO-HS-65-28A1 in P-AUTO and VERIFY EGTS Train A decay cooling Valve A FO-FCV-65-28A1 opens.

h. PLACE IO-HS-65-28B1 in P-AUTO and VERIFY EGTS Train A decay cooling Valve B rO-FCV-65-28B1 opens.

i. CHECK EGTS Train A filter assembly drain loop seal levels greater than or equal to 50%.

[Unit 2, Aux Bldg, Penetration Room, elev 714]

9. IF Train B to be shut down, THEN PERFORM the following:

a. PULL IO-HS-65-42A1 to P-AUTO (STANDBY) and momentarily TURN to STOP. LI

b. VERIFY 1O-HS-65-42A1 returns to P-AUTO(STANDBY). LI

c. VERIFY Fan B-B is STOPPED. LI

d. ENSURE EGTS Fan B discharge damper lO-FCV-65-431 CLOSED. LI

e. IF Unit 1 selected for EGTS operation, THEN ENSURE EGTS Fan B-B suction dampers ALIGNED as follows:

UNIT I POSITION EGTS SUCTION DAMPERS 1-FCV-65-51 OPEN LI 1-FCV-65-30 CLOSED LI

SQN EGTS OPERATION I EA-65-1 I I

IRev,l I 1, 2 Page 16 of 17 4.4 Establishing Single EGTS Train Operation (Continued)

9. (Continued)

f. IF Unit 2 selected for EGTS operation, THEN ENSURE EGTS Fan B-B suction dampers ALIGNED as follows:

UNIT 2 POSITION q EGTS SUCTION DAMPERS 2-FCV-65-50 OPEN LI 2-FCV-65-29 CLOSED LI

g. PLACE IO-HS-65-47A1 in P-AUTO and VERIFY EGTS Train B decay cooling Valve A IO-FCV-65-47A1 OPENS.

h. PLACE IO-HS-65-47B1 in P-AUTO and VERIFY EGTS Train B decay cooling Valve B 1O-FCV-65-47B1 OPENS. LI

i. CHECK EGTS Train B filter assembly drain loop seal levels greater than or equal to 50%.

[Unit 2, Aux Bldg, Penetration Room, elev 714] El

10. GO TO Section 4.1, step in effect.

END OF TEXT

3/4 LIMITING CONDITIONS FOR OPEATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding Specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met and as provided in LCO 3.0.7.

3.0.2 Noncompliance with a Specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Conditions for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.

3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within one hour action shall be initiated to place the unit in a MODE in which the Specification does not apply by placing it, as applicable, in:

1. At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

2. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

3. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual Specifications.

3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made:

a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time;

b. After performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate; exceptions to this Specification are stated in the individual Specifications, or

c. When an allowance is stated in the individual value, parameter, or other Specification.

This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

3.0.5 When a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided: (1) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system(s), subsystem(s), train(s), component(s) and device(s) are OPERABLE, or likewise satisfy the requirements of this Specification. Unless both October 4, 2006 SEQUOYAH - UNIT 1 3/4 0-1 Amendment No. 202, 301, 312

CONTAINMENT SYSTEMS EMERGENCY GAS TREATMENT SYSTEM - EGTS CLEANUP SUBSYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.8 Two independent emergency gas treatment system cleanup subsystems (EGTS) shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one EGTS cleanup subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.1.8 Each EGTS cleanup subsystem shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with the heaters on.

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:

1. Verifying that the cleanup system satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Position C.5.a., C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978 (except for the provisions of ANSI N510 Sections 8 and 9), and the system flow rate is 4000 cfm

+/- 10%.

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%.

3. Verifying a system flow rate of 4000 cfm +/- 10% during system operation when tested in accordance with ANSI N510-1975.

November 2, 2000 SEQUOYAH - UNIT 1 3/4 6-13 Amendment No. 263

SQN EA-0-l EQUIPMENT CHECKS FOLLOWING ESF ACTUATION Rev. 10 1,2 Pagel6of2O 4.6 Placing Ventilation Equipment in Standby CAUTION ABGTS, EGTS, and CREVS fans must be considered inoperable if stopped in PULL P AUTO (Standby). Automatic fan start on low flow will NOT function if start signal (ABI, Phase A, or CR1) has been reset.

NOTE This section should be performed by a licensed operator.

1. WHEN at least 30 minutes have elapsed since start of accident AND TSC is staffed, THEN PERFORM the following [1-M-9]:

a. IF both trains of ABGTS are running THEN NOTIFY TSC to evaluate if one train of ABGTS should be stopped and placed in standby. E

b. IF both trains of EGTS are running THEN NOTIFY TSC to evaluate if one train of EGTS should be stopped and placed in standby. E

c. IF both trains of CREVS are running THEN NOTIFY TSC to evaluate if one train of CREVS should be stopped and placed in stopped standby.

2. IF TSC directs placing one train of ABGTS in standby, THEN PERFORM Appendix A, Placing One Train of ABGTS in Standby.

3. IF TSC directs placing one train of EGTS in standby, THEN PERFORM Appendix B, Placing One Train of EGTS in Standby. El

4. IF TSC directs placing one train of CREVS in standby, THEN PERFORM Appendix C, Placing One Train of CREVS in Standby.

END OF SECTION

SQN EA-0-1 EQUIPMENT CHECKS FOLLOWING ESF ACTUATION Rev. 10 1,2 Pagel9of2O Page lofI APPENDIX B PLACING ONE TRAIN OF EGTS IN STANDBY CAUTION If Phase A signal is reset, EGTS fan in PULL P AUTO (Standby) will NOT automatically start on low flow from running fan.

NOTE This appendix shall be performed by a licensed operator.

1. IF TSC directs placing one train of EGTS in standby, THEN PERFORM the following [O-M-28AJ:

a. PLACE selected EGTS fan handswitch in PULL P AUTO (Standby):

(N/A fan not selected)

Train Handswitch PULL P AUTO A O-HS-65-23A B O-HS-65-42A

b. STOP selected EGTS fan: (N/A fan not selected)

Train Handswitch STOP A O-HS-65-23A El B O-HS-65-42A El

c. MONITOR operation of running EGTS fan from MCR.

d. IF running EGTS fan is lost, THEN ENSURE standby fan is started.

END

SQN HIGH CONTAINMENT RADIATION FR-Z.3 Rev. 7 I STEP I I ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I 1.0 PURPOSE This procedure provides actions to respond to high containment radiation levels.

2.0 SYMPTOMS AND ENTRY CONDITIONS 2.1 ENTRY CONDITIONS FR-O Status Trees:

F-O.5, Containment YELLOW condition:

Upper containment radiation greater than or equal to 100 RIhr.

F-0.5, Containment YELLOW condition:

Lower containment radiation greater than or equal to 1 00 RIhr.

3.0 OPERATOR ACTIONS Page 2 of 4

OPT200.EGTS Rev. 3 Page 3 of 53 PROGRAM: OPERATOR TRAINING II. COURSE: SYSTEMS TRAINING III. TITLE: EMERGENCY GAS TREATMENT SYSTEM (EGTS)

IV. LENGTH OF LESSON: 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months lecture; 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months simulator demonstration; 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months self-study/workshop V. TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of this lesson and others presented, the student should be able to apply the knowledge to support satisfactory performance of the tasks associated with the EGTS system in the plant and on the simulator.

B. Learning Objectives:

0. Demonstrate an understanding of NUREG 1122 knowledges and abilities associated with the Emergency Gas Treatment System that are rated> 2.5 during Initial License Training for the appropriate license position as identified in Appendix A.

1. State the purpose/functions of the EGTS system as described in the FSAR.

2. State the design basis of the EGTS system in accordance with the SQN FSAR.

3. Explain the purpose/function of each major cQmponent in the flow path of the EGTS system as illustrated on a simplified system drawing.

4. Describe the following characteristics of each major component in the EGTS system:

a. Location

b. Power supply (include control power as applicable)

c. Support equipment and systems

d. Normal operating parameters

e. Component operation

f. Controls

g. Interlocks (including setpoints)

h. Instrumentation and Indications

i. Protective features (including setpoints)

j. Failure modes

k. Unit differences

1. Types of accidents for which the EGTS system components are designed

m. Location of controls and indications associated with the EGTS system in the control room and auxiliary control room

OPT200.EGTS Rev. 3 Page 4 of 53 V. TRAINING OBJECT EVES (Contd):

B. Learning Objectives (Contd):

5. Describe the operation of the EGTS system:

a. Precautions and limitations

b. Major steps performed while placing the EGTS system in service

c. Alarms and alarm response

d. How a component failure will affect system operation

e. How a support system failure will affect EGTS system operation

f. How a instrument failure will affect system operation

6. Describe the administrative controls and limits for the EGTS system:

a. State Tech Specs/TRM LCOs that govern the EGTS

b. State the <1 hour action limit TS LCOs

c. Given the conditions/status of the EGTS system components and the appropriate sections of the Tech Spec, determine if operability requirements are met and what actions are required

7. Discuss related Industry Events

a. Both Trains of Emergency Gas Treatment Placed Out of Service for Less than 30 Minutes

b. Misposition of Emergency Gas Treatment System Handswitches VI. TRAINING AIDS:

A. Classroom Computer and Local Area Network (LAN) Access B. Computer projector C. Simulator (if available)

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta

84. Given the following plant conditions:

- Unit 1 is at 100% power.

- Unit 2 is in Mode 6 with preparations being made to reload the core.

- 0-RM-90-103, Spent Fuel Pool Pit Area Monitor, is out of service for maintenance with high rad relays removed. Train B rad monitor block switch 0-HS 136A2 is positioned off and pushed-in..

- As part of performing the Functional Test for 0-RM-90-1 02, Spent Fuel Pool Pit Area Monitor, a source check is to be performed. Train A rad monitor block switch 0-HS 136A1 is positioned to 0-102 and pulled out.

- When the source check is performed the instrument fails to respond.

Which ONE of the following identifies...

(1) the status of the 2 annunciator windows on 0-M-1 2D listed below:

0-HS-90-136A1 HIGH RAD IN CNMT INPUT TO TR A OF SSPS BLOCKED 0-HS-90-136A2 HIGH RAD IN CNMT INPUT TO TR B OF SSPS BLOCKED and (2) how the Tech Specs LCOs listed below will be applied on Unit 2 during the source check of the radiation monitor?

LCO 3.3.3.1, Radiation Monitoring Instrumentation LCO 3.9.12, Auxiliary Building Gas Treatment System.

A. (1) Only, 0-HS-90-1 36A1 HIGH RAD IN CNMT INPUT TO TR A OF SSPS BLOCKED will be LIT (2) Only one of the LCOs will be entered.

B (1) Only, 0-HS-90-1 36A1 HIGH RAD IN CNMT INPUT TO TR A OF SSPS BLOCKED will be LIT (2) Both of the LCOs will be entered.

C. (1) Both of the listed annunciator windows will be LIT.

(2) Only one of the LCOs will be entered.

D. (1) Both of the listed annunciator windows will be LIT.

(2) Both of the LCOs will be entered.

Thursday, July 15, 2010 4:13:54 PM 84

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta DISTRACTOR ANALYSIS:

A. Incorrect, Plausible because only the A Train alarm will be lit since the B Train switch O-HS-90-136A2 is pushed in and the candidate may fail to recall that the SFP area monitors are required attendant equipment for ABG TS Operability for TS 3.9.12.

B. Correct. The A Train alarm will be lit since handswitch O-HS-90-136A1 is pulled out for the FT. The Train B alarm will not be lit since its associated HS is pushed in. O-RM-90-103 is Inoperable (initial condition) and O-RM-90-102 must be declared Inoperable due to its failure to source check; TS 3.3.3.1 must be entered as it requires one of the two monitors Operable. The SFP area monitors are required attendant equipment for ABGTS Operability for TS 3.9.12, therefore both Tech Specs must be entered.

C. Incorrect, Plausible because the B Train alarm would normally be lit with O-RM-90-103s auto function blocked by O-HS-90-136A2 when the monitor is Inoperable and the candidate may fall to recall that the SFP area monitor is required attendant equipment for ABGTS Operability for TS 3.9.12. Incorrect because O-HS-90-136A2 is pushed in so the B Train alarm would not be lit.

D. Incorrect, Plausible because the B Train alarm would normally be lit with O-RM-90-103s auto function blocked by O-HS-90-136A2 when the monitor is Inoperable and entry into both LCOs is correct. Incorrect because O-HS-90-136A2 is pushed in so the B Train alarm would not be lit.

Thursday, July 15, 2010 4:13:54 PM 85

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta Question Number: 84 Tier: 1 Group 2 K/A: 061 Area Radiation Monitoring (ARM) System Alarms AA2.01 Ability to determine and interpret the following as they apply to the Area Radiation Monitoring (ARM) System Alarms:

ARM panel displays Importance Rating: 3.5 I 3.7 10 CFR Part 55: (CFR: 43.5 /45.13)

IOCFR55.43.b: 5 K/A Match: Question matches the K/A by having the applicant interpret indications associated with area rad monitors and SRO because application of TS Operability is dependent on information in the required surveillance of the specification.

Technical

Reference:

O-AR-M12-D E3 and E4 rev 42 T.S. 3.3.3.1, Radiation Monitoring Instrumentation, Amendment 322 T.S. 3.7.8, Auxiliary Building Gas Treatment System, Amendment 263 Proposed references None to be provided:

Learning Objective: OPT200.RM Rev 3; Obj 6.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: New question Comments:

MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9 Answer: BCDBDBBABB ScrambleRange:A-D Thursday, July 15, 2010 4:13:54 PM 86

3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels shown in Table 3.3-6 shall be OPERABLE with their alarm/trip setpoints within the specified limits.

APPLICABILITY: As shown in Table 3.3-6.

ACTION:

a. With a radiation monitoring channel alarm/trip setpoint exceeding the value shown in Table 3.3-6, adjust the setpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable.

b. With one or more radiation monitoring channels inoperable, take the ACTION shown in Table 3.3-6.

c. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.1 Each radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-3.

April 11,2005 SEQUOYAH - UNIT 1 3/4 3-39 Amendment No. 301

TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE ALARM/TRIP MEASUREMENT INSTRUMENT OPERABLE MODES SETPOINT RANGE ACTION

1. AREA MONITOR

a. Fuel Storage Pool 1

151 mR/hr 26 10 - 1O mR/hr Area

2. PROCESS MONITORS 8.5x 10 10- 1O7 cpm

a. Containment Purge 1 1, 2, 3, 4 & 6 28 Ci/cc Air

b. Containment

i. Deleted ii. Particulate Activity RCS Leakage 1 1, 2, 3 & 4 N/A 10 - 7 cpm 1O 27 Detection

c. Control Room 2 ALL MODES 400 cpm** 10 - 7 cpm 1O 29 Isolation and during movement of irradiated fuel assemblies With fuel in the storage pool or building Equivalent to 1.0 x 1 0 iiCi/cc.

December 04, 2008 SEQUOYAH - UNIT 1 3/4 3-40 Amendment Nos. 12, 60, 112, 168, 256, 310, 322

TABLE 3.3-6 (Continued)

ACTION STATEMENTS ACTION 26 With the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirement, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 27 With the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.6.1.

ACTION 28 With the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.9 (MODE 6) and 3.3.2.1 (MODES 1, 2, 3, and 4).

ACTION 29 - a. With one channel inoperable, place the associated control room emergency ventilation system (CREVS) train in recirculation mode of operation within 7 days or be at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With two channels inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months initiate and maintain operation of one CREVS train in the recirculation mode of operation and enter the required Actions for one CREVS train made inoperable by inoperable CREVS actuation instrumentation.

Or place both trains in the recirculation mode of operation within one hour.

If the completion time of Action 29b cannot be met in Modes 1, 2, 3, and 4, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

If the completion time of Action 29b cannot be met during the movement of irradiated fuel assemblies, suspend core alterations and suspend movement of irradiated fuel assemblies.

If the completion time of Action 29b cannot be met in Modes 5 and 6, initiate action to restore one CREVS train.

May 31, 2000 SEQUOYAH - UNIT 1 3/4 3-41 Amendment No. 12, 112, 168, 256

TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECK CALIBRATION TEST REQUIRED

1. AREA MONITOR

a. Fuel Storage Pool S R Q

Area

2. PROCESS MONITORS

a. Containment Purge Air S R Q 1, 2, 3, 4 & 6 Exhaust

b. Containment

i. Deleted ii. Particulate Activity RCS Leakage S R Q 1, 2, 3, & 4 Detection

c. Control Room S R Q ALL MODES Isolation

With fuel in the storage pool or building.

December 04, 2008 SEQUOYAH - UNIT 1 3/4 3-42 Amendment Nos. 12, 112,168, 220, 322

PLANT SYSTEMS 3/4.7.8 AUXILIARY BUILDING GAS TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.7.8 Two independent auxiliary building gas treatment filter trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one auxiliary building gas treatment filter train inoperable, restore the inoperable train to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.8 Each auxiliary building gas treatment filter train shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filter and charcoal adsorber train and verifying that the system operates for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with the heaters on.

b. At least once per 18 months or (I) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:

1. Verifying that the cleanup system satisfies the inpIace testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978 (except for the provisions of ANSI N510 Sections 8 and 9), and the system flow rate is 9000 cfm +/- 10%.

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%.

3. Verifying a system flow rate of 9000 cfm +/- 10% during system operation when tested in accordance with ANSI N510-1975.

November 2, 2000 SEQUOYAH - UNIT 1 3/4 7-19 Amendment No. 12, 263

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%.

d. At least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 3 inches Water Gauge while operating the filter train at a flow rate of 9000 cfm +/- 10%.

2. Verifying that the filter trains start on a Containment Phase A Isolation test signal.

3. Verifying that the system maintains the spent fuel storage area and the ESF pump rooms at a pressure equal to or more negative than minus 1/4 inch water gage relative the outside atmosphere while maintaining a total system flow of 9000 cfm

+/- 10%.

4. Verifying that the heaters dissipate 32 +/- 3.2 kw when tested in accordance with ANSI N51 0-1 975.

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.95% of the DOP when they are tested in-place in accordance with ANSI N510-1 975 while operating the system at a flow rate of 9000 cfm +/- 10%.

f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.95% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1 975 while operating the system at a flow rate of 9000 cfm +/- 10%.

August 18, 2005 SEQUOYAH - UNIT 1 3/4 7-20 Amendment Nos. 12, 88, 103, 122, 263, 303

REFUELING OPERATIONS 3/4.9.12 AUXILIARY BUILDING GAS TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.9.12 One auxiliary building gas treatment filter train shall be OPERABLE.

APPLICABILITY: Whenever irradiated fuel is in the storage pool.

ACTION:

a. With no auxiliary building gas treatment filter train OPERABLE, suspend all operations involving movement of fuel within the spent fuel pit or crane operation with loads over the spent fuel pit until at least one auxiliary building gas treatment filter train is restored to OPERABLE status.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.12 The above required auxiliary buildings gas treatment filter train shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with the heaters on.

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:

1. Verifying that the cleanup system satisfies the inpIace testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978 (except for the provisions of ANSI N510 Sections 8 and 9), and the system flow rate is 9000 cfm +/- 10%.

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5%

when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%.

3. Verifying a system flow rate of 9000 cfm +/- 10% during system operations when tested in accordance with ANSI N510-1975.

April 11,2005 SEQUOYAH - UNIT 1 3/4 9-12 Amendment No. 263, 301

REFUELING OPERATIONS SURVEILLANCE REQU I REMENTS (Continued)

c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1 989 at a temperature of 30°C (86° F) and a relative humidity of 70%.

d. At least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 3 inches Water Gauge while operating the filter train at a flow rate of 9000 cfm +/- 10%.

2. Verifying that the filter train starts on a high radiation signal from the fuel pool radiation monitoring system.

3. Verifying that the heaters dissipate 32 +/- 3.2 kw when tested in accordance with ANSI N51 0-1975.

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.95% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 9000 cfm +/- 10%.

f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.95% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the system ata flow rate of 9000 cfm +/- 10%.

November 2, 2000 SEQUOYAH - UNIT 1 3/4 9-13 Amendment No. 88, 122, 263

31 (E-3)

Source Setpoint O-HS-90-1 36A1 SER 816 (Unit 1 annunciator system) N/A HIGH RAD IN CNTMT 0-H S-go-i 36A1 Retransmitted to U-2 INPUT TO TR A SER 2228 (Unit 2 annunciator system) OF SSPS BLOCKED Probable 1. Operator action.

Causes NOTE Alarm illuminated indicates one of four selectable rad monitors is in test and its automatic functions blocked for Train A.

Corrective [1] VERIFY 0-HS-90-136A1 intentionally placed in the Actions BLOCKED position.

References 45B655-1 2D-0 SQN O-AR-M12-D Page 37 of 42 0 Rev. 41

32 (E-4)

Source Setpoint O-HS-90-1 36A2 SER 817 (Unit 1 annunciator system) N/A HI RAD IN CNTMT O-HS-90-1 36A2 Retransmitted to U-2 INPUT TO TR B SER 2229 (Unit 2 annunciator system) OF SSPS BLOCKED Probable 1. Operator action.

Causes NOTE Alarm illuminated indicates one of four selectable rad monitors is in test and its automatic functions blocked for Train B.

Corrective [1] VERIFY O-HS-90-136A2 intentionally placed in Actions BLOCKED position.

References 45B655-1 2D-O SQN O-AR-M12-D Page 38 of 42 0 Rev41

OPT200.RM Rev. 3 Page 4 of 166 V. TRAINING OBJECTIVES (Contd):

B. Enabling Objectives (Cont d):

5. Describe the operation of the Radiation Monitoring System:

a. Precautions and limitations

b. Major steps performed while placing the system in service

c. Alarms and alarm response

d. How a component failure will affect system operation

e. How a support system failure will affect system operation

f. How a instrument failure will affect system operation

6. Describe the administrative controls and limits for the Radiation Monitoring System:

a. State Tech Specs/TRM LCOs that govern the system.

b. State the l hour action limit TS LCOs

c. Given the conditions/status of the Radiation Monitoring System components and the appropriate sections of the Tech Spec, determine if operability requirements are met and what actions are required

7. Discuss related Industry Events VI. TRAINING AIDS:

A. Classroom Computer and Local Area Network (LAN) Access B. Computer projector C. Simulator (if available)

OPT200.ABGTS Rev. 2 Page 4 of 59 V. TRAINING OBJECTIVES (Contd):

B. Enabling Objectives (Cont d):

5. Describe the operation of the ABGTS as it relates to the following:

a. Precautions and limitations

b. Major steps performed while placing the ABGTS in service

c. Alarms and alarm response

d. How a component failure will affect system operation

e. How a support system failure will affect ABGTS operation

f. How a instrument failure will affect system operation

6. Describe the administrative controls and limits for the ABGTS as explained in this lesson:

a. State Tech Specs/TRM LCOs that govern the ABGTS

b. State the l hour action limit TS LCOs

c. Given the conditions/status of the ABGTS components and the appropriate sections of the Tech Spec, determine if operability requirements are met and what actions are required

7. Discuss related Industry Events:

a. Event

Title:

Auxiliary Building Gas Treatment System Air Heater Inoperable.

1NPO Event # 390-970118-1

b. Event

Title:

Auxiliary Building Gas Treatment System Design Error.

1NOP Event # 327-930805-1 VI. TRAINING AIDS:

A. Computer.

B. Computer Display Projector & Controls.

C. Local Area Network (LAN) Access.

D. Simulator (if available)

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta

85. Given the following:

- Unit I is at 100% power.

- The local leak rate test on the lower containment air lock test performed following a containment entry was NOT satisfactory.

- Plant Engineering has just reported that 0-Sl-SLT-000-160, Primary Containment Total Leak Rate, has just been completed.

- System Engineering reports that Ui overall containment leakage rate Acceptance Criteria has been exceeded.

Which ONE of the following identifies:

(1) the most limiting Tech Spec LCO action time for the condition described and (2) the Tech Spec basis for the maximum allowed total containment leakage (La)?

Reference Provided A. (1) 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (2) To prevent exceeding the design capability of the ABGTS B. (1) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (2) To prevent exceeding the design capability of the ABGTS Cv (1) 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (2) To prevent exceeding 10CFR100 limits at the site boundary D. (1) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (2) To prevent exceeding 10CFR100 limits at the site boundary Tuesday, July 13, 2010 8:13:29 AM 85

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta DIS TRACTOR ANALYSIS:

A. Incorrect, Plausible since the 1 hr LCO is correct, and leakage from ECCS equipment during an accident is processed by the ABGTS train to maintain offsite dose within accident analysis assumptions.

B. Incorrect, Plausible if the candidate determined that the 24 hr LCO action of T.S.

3.6. 1.3.b applied to an inoperable airlock and leakage from ECCS equipment during an accident is processed by the ABGTS train to maintain offsite dose within accident analysis assumptions.

C. Correct, With the overall containment leakage greater than allowable T.S. 3.6.1.3 footnote (2) states that T.S. 3.6.1.1 Containment lntergrity is to be applied which has a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action statement. The containment allowable leakage limit prevents exceeding IOCFR 100 limits..

D. Incorrect, Plausible if candidate determined that the 24 hr LCO action of T. S.

3.6. 1.3.b is the only applicable specification and the containment allowable leakage limit prevents exceeding 10CFR 100 limits..

Tuesday, July 13, 2010 8:13:29AM 86

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta Question Number: 85 Tier: 1 Group 2 KIA: 069 Loss of Containment Integrity AG 2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

Importance Rating: 4.0 / 4.6 10 CFR Part 55: n/aS IOCFR55.43.b: 5 KIA Match: This question matches the KJA by having the candidate analyze data and determine that a loss of containment intergrity exists, this has a 1 hr LCO action statement. The question meets SRO criteria because knowlege of TS basis is tested.

Technical

Reference:

T.S. 3.6.1.3, T.S. 3.6.1.1 and Basis, 0-Sl-SLT-000-160.0 rev 0004 Proposed references None to be provided:

Learning Objective: 0PL271C168 obj 3 0PL273C0530, obj 2 Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question written for 1009 SRO NRC exam Comments:

MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9 Answer: CADBDCAADD ScrambleRange: A- D Tuesday, July 13, 2010 8:13:29 AM 87

CONTAINMENT SYSTEMS c \) I CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE* with both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With one containment air lock door inoperable:

1. Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the OPERABLE air lock door closed.

2. Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days.

3. Otherwise, be in at least HOT STANDBY within the next six hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next six hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

1. An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.

2. Enter the ACTION of LCO 3.6.1.1, Primary Containment when air lock leakage results in exceeding the overall containment leakage rate acceptance criteria.

April 11,2005 SEQUOYAH - UNIT 1 3/4 6-7 Amendment No. 12, 217, 301

CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE* with both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With one containment air lock door inoperable:

1. Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the OPERABLE air lock door closed.

2. Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days.

3. Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With the containment air lock inoperable, except as the result of a inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

1. An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.

2. Enter the ACTION of LCO 3.6.1 .1, Primary Containment when air lock leakage results in exceeding the overall containment leakage rate acceptance criteria.

April 11,2005 SEQUOYAH - UNIT 2 3/4 6-7 Amendment No. 207, 290

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

a. By verifying leakage rates in accordance with the Containment Leakage Rate Test Program.

b. At least once per 6 months by verifying that only one door in each air lock can be opened at a time.

February 5, 1996 SEQUOYAH - UNIT 2 3/4 6-8 Amendment Nos. 40, 167, 207

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a. Deleted.

b. By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.1.3.

c. Perform required visual examinations and leakage rate testing in accordance with the Containment Leakage Rate Testing Program.

March 29, 2000 SEQUOYAH - UNIT 2 3/4 6-1 Amendment No. 117, 167, 183, 193, 207, 245

3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT The safety design basis for primary containment is that the containment must withstand the pressures and temperatures of the limiting design basis accident (DBA) without exceeding the design leakage rates.

The DBAs that result in a challenge to containment OPERABILITY from high pressures and temperatures are a loss of coolant accident (LOCA), a steam line break, and a rod ejection accident (REA). In addition, release of significant fission product radioactivity within containment can occur from a LOCA or REA. In the DBA analyses, it is assumed that the containment is OPERABLE such that, for the DBAs involving release of fission product radioactivity, release to the environment is controlled by the rate of containment leakage. This leakage rate limitation will limit the site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions. The containment was designed with an allowable leakage rate of 0.25 percent of containment air weight per day. This leakage rate, used in the evaluation of offsite doses resulting from accidents, is defined in the Containment Leakage Rate Test Program, as La: the maximum allowable containment leakage rate at the calculated peak containment internal pressure (Pa) resulting from the limiting DBA. The allowable leakage rate represented by La forms the basis for the acceptance criteria imposed on all containment leakage rate testing.

Primary containment INTEGRITY or operability is maintained by limiting leakage to within the acceptance criteria of the Containment Leakage Rate Test Program.

3/4.6.1.2 SECONDARY CONTAINMENT BYPASS LEAKAGE This specification has been relocated.

April 13, 2009 SEQUOYAH - UNIT 1 B 3/4 6-1 Amendment No. 102, 127, 176, 217, 323

PLANT SYSTEMS BASES 3/4.7.8 AUXILIARY BUILDING GAS TREATMENT SYSTEM The OPERABILITY of the auxiliary building gas treatment system ensures that radioactive materials leaking from the ECCS equipment following a LOCA are filtered prior to reaching the environment. The operation of this system and the resultant effect on offsite dosage calculations was assumed in the accident analyses. ANSI N510-1 975 will be used as a procedural guide for surveillance testing. Cumulative operation of the system with the heaters on for 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months over a 31 day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters.

3/4.7.9 SNUBBERS This specification is deleted.

August 18, 2005 SEQUOYAH - UNIT I B 3/4 7-5 Amendment No. 12, 235, 303

SQN PRIMARY CONTAINMENT TOTAL 0-SI-SLT-000-1 60.0 Unit I & 2 LEAK RATE Rev. 0004 Page 6 of 42 Unit Date 4.3 Approvals and Notifications

[1] OBTAIN Senior Reactor Operator (SRO) approval to perform this Instruction on Surveillance Task Sheet.

5.0 ACCEPTANCE CRITERIA The Shift Manager must be notified immediately upon receipt of test results that cause the total leakage calculated in this surveillance instruction to exceed any acceptance criteria stated in A through D. Steps that determine the following criteria are designated by (AC) next to the initials blank.

NOTE La is the Appendix J allowable Containment Leakage and is = 0.25% of Primary Containment Air per Day at Accident Pressure (12 psig). The following determines La in units of SCFH:

0.0025 x Containment Volume x(1 2 psig La + ATM ATM x24 hours/day

= 0.0025 xl .1 9x1 x 26.69595 La = 225.17 SCFH 14.69595x24

OPT200.TS- Intro Revision 2 Page 3 of 30 I. PROGRAM: OPERATOR TRAINING - LICENSED COURSE: LICENSED TRAINING III. LESSON TITLE: TECHNICAL SPECIFICATION - INTRODUCTION IV. LENGTH OF LESSONICOURSE: 2.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months (s)

V. TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of this lesson and others presented, the student shall demonstrate an understanding of the purpose and background of Technical Specifications, and definitions listed in Tech Spec Section 1, Definitions, by successfully completing a written examination with a score of 80%.

B. Enabling Objectives:

Conditions and standard of the Terminal Objective are implied.

1. Describe the relationship between the Technical Specifications and the Facility Operating License.

2. Identify the five (5) sets of Facility Operating License Conditions listed in parts 2A, B, C, D, and E, with which the facility must comply.

3. Explain the purpose of the Tech Spec Limiting Conditions for Operation, and state when their associated Action is applicable.

4. Explain the purpose of the Tech Spec Surveillance Requirements. (C.1)

5. List the six sections of technical specifications per section B of lesson plan.

6. List the subsections for Tech Specs Section 3.0, Limiting Conditions for Operation, per section B of lesson plan.

7. List or identify the cycle-specific core parameters whose operating limits are found in the Core Operating Limits Report.

8. Explain the plant license requirements controlled by the Offsite Dose Calculation Manual.

9. Define the following terms as Defined in Tech Spec Section 1.0, Definitions:

a. Axial Flux Difference

b. Operable Operability B. Enabling Objectives: (continued)

OPT200 .TS-lntro Revision 2 Page 4 of 30 Conditions and standard of the Terminal Objective are implied.

9. (continued):

c. Pressure Boundary Leakage

d. Identified Leakage

e. Unidentified Leakage

f. Containment Integrity

g. Shutdown Margin

h. Operational Modes

i. Core Alteration

10. List or identify the RCS Pressure and Temperature requirements and Low Temperature Over Pressure Protection requirements whose operating limits are established and documented in the RCS Pressure Temperature Limits Report (PTLR).

OPT200.CntmtStructure Rev. 3 Page 3 of 45 PROGRAM: OPERATOR TRAINING II. COURSE: SYSTEMS TRAINING III. TITLE: CONTAINMENT STRUCTURE IV. LENGTH OF LESSON: 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months lecture; 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months simulator demonstration; 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months self-study/workshop V. TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of this lesson and others presented, the student should be able to apply the knowledge to support satisfactory performance of the tasks associated with the Containment Structure system in the plant and on the simulator.

B. Learning Objectives:

0. Demonstrate an understanding of NUREG 1122 knowledges and abilities associated with the Containment Structure system that are rated 2.5 during Initial License Training for the appropriate license position as identified in Appendix A.

1. State the purpose/functions of the Containment Structure system as described in the FSAR.

2. State the design basis of the Containment Structure system in accordance with the SQN FSAR.

3. Explain the purpose/function of each major component in the flow path of the Containment Structure system as illustrated on a simplified system drawing.

4. Describe the following characteristics of each major component in the Containment Structure system:

a. Location

b. Power supply (include control power as applicable)

c. Support equipment and systems

d. Normal operating parameters

e. Component operation

f. Controls

g. Interlocks (including setpoints)

h. Instrumentation and Indications

i. Protective features (including setpoints)

j. Failure modes

k. Unit differences

1. Types of accidents for which the Containment Structure system components are designed

m. Location of controls and indications associated with the Containment Structure system in the control room and auxiliary control room

OPT200.CntmtStructure Rev. 3 Page 4 of 45 V. TRAINING OBJECTIVES (Contd):

B. Learning Objectives (Contd):

5. Describe the operation of the Containment Structure system:

a. Precautions and limitations

b. Major steps performed while placing the Containment Structure system in service

c. Alarms and alarm response

d. How a component failure will affect system operation

e. How a support system failure will affect Containment Structure system operation

f. How a instrument failure will affect system operation

6. Describe the administrative controls and limits for the Containment Structure system:

a. State Tech Specs/TRM LCOs that govern the Containment Structure

b. State the <1 hour action limit TS LCOs

c. Given the conditions/status of the Containment Structure system components and the appropriate sections of the Tech Spec, determine if operability requirements are met and what actions are required

7. Discuss related Industry Events

a. SQN ABSCE rendered inoperable, LER 96-009-00

b. SQN Containment leak rate exceeded TS limit, LER 94-005-00

c. SQN AFW vent line valves not secured closed, LER 93 -028-00

d. SQN personnel airlock outer housing leaking, LER 93-004-0 1

e. Browns Ferry 3 MSIV leakage exceeded the TS limit, LER 94-008-0 1

f. North Anna Power Station 2 breached fire barrier penetrations, LER 88-007-00 VI. TRAINING AIDS:

A. Classroom Computer and Local Area Network (LAN) Access B. Computer projector C. Simulator (if available)

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta

86. Given the following:

- Unit I is operating at 54% power in accordance with GO-S Section 51, Power Ascension From 30% to 100% following a forced outage.

- Power Range Instrument N-41 has been removed from service after failing and the associated bistables have been placed in the tripped condition.

- The operating crew determines the Channel II P-9 interlock has failed to actuate and initiates action to comply with Reactor Trip System Instrumentation LCO 3.3.1.1.

Which ONE of the following identifies...

(1) the affect of the failures on the power ascension and (2) the affect of the failed bistable on the Reactor Trip from Turbine Trip function?

A. (1) The power ascension can continue using GO-S Section 5.1.

(2) Function is inoperable.

B (1) The power ascension can continue using GO-5 Section 5.1.

(2) Function remains operable.

C. (1) Power must be reduced to less than P-9 using GO-5 Section 5.3, Power Reduction From 100% to 30%.

(2) Function is inoperable.

D. (1) Power must be reduced to less than P-9 using GO-5 Section 5.3, Power Reduction From 100% to 30%.

(2) Function remains operable.

Thursday, July 15, 2010 4:10:09 PM 86

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta DISTRACTOR ANAL YSIS:

A. Incorrect, plausible because the Interlock is Inoperable which requires assessing Operability of the turbine trip function and power ascension is not restricted.

B. Correct, Two channels of the P-9 Interlock are Inoperable and Action 8e requires that affected channels (2/3 low autostop oil pressure & 4/4 stop valves closed) for the turbine trip reactor trip function are verified Operable or apply their action. All autostop oil and stop valves are Operable so the turbine/trip reactor trip function is Operable. There is no restriction on increasing power.

C. Incorrect, plausthie because the Interlock is Inoperable which requires assessing Operability of the turbine trip function. If the candidate concludes the turbine trip function is Inoperable power is retricted to less than P-9 by Action 7.

D. Incorrect, plausible because the Interlock is Inoperable which requires assessing Operability of the turbine trip function. If the candidate concludes the turbine trip function is Operable he may conclude power is retricted to less than P-9 due to (noperability of the associated interlock.

Thursday, July 15, 2010 4:10:09 PM 87

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta Question Number: 86 Tier: 2 Group 2 KIA: 012 Reactor Protection System (RPS)

A2.01 Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Faulty bistable operation Importance Rating: 3.1 I 3.6 10 CFR Part 55: 41.5, 43.5, 45.3, 45.5 IOCFR55.43.b: 5 K/A Match: Match because candidate must evaluate the faulty operation of a reactor protection system interlock bistable and SRO due to application of Tech. Spec.

Technical

Reference:

Tech Spec 3.3.1.1 Proposed references None to be provided:

Learning Objective: OPT200.RPS Obj 9 & 13.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History:

Comments:

MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9 Answer: BDCACCACCB ScrambleRange:A-D Thursday, July 15, 2010 4:10:09 PM 88

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1.1 As a minimum, the reactor trip system instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE.

APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLANCE REQUIREMENTS 4.3.1.1.1 Each reactor trip system instrumentation channel and interlock shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-1.

4.3.1.1.2 The logic for the interlocks shall be demonstrated OPERABLE prior to each reactor startup unless performed during the preceeding 92 days. The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be verified to be within its limit at least once per 18 months. Neutron detectors are exempt from response time testing. Each verification shall include at least one train such that both trains are verified at least once per 36 months and one channel per function such that all channels are verified at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the Total No. of Channels column of Table 3.3.1.

February 29, 2000 SEQUOYAH - UNIT 1 3/4 3-1 Amendment Nos. 12, 190, 251

TABLE 3.3-1 (Continued)

TABLE NOTATION

With the reactor trip system breakers in the closed position, the control rod drive system capable of rod withdrawal, and fuel in the reactor vessel.

Above the P-9 (Power Range Neutron Flux) interlock.

Source Range outputs may be disabled above the P-6 (Block of Source Range Reactor Trip) setpoint.

ACTION STATEMENTS ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the reactor trip breakers.

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.3.1.1.1.

c. The QUADRANT POWER TILT RATIO is monitored in accordance with Technical Specification 3.2.4.

September 2, 2005 SEQUOYAH - UNIT 1 3/4 3-5 Amendment Nos. 47, 135, 136, 141, 213, 301, 304

TABLE 3.3-1 (Continued)

ACTION 3 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a. Below the P-6 (Block of Source Range Reactor Trip) setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint.

b. Above the P-6 (Block of Source Range Reactor Trip) setpoint, but below 5% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 5% of RATED THERMAL POWER.

c. Above 5% of RATED THERMAL POWER, POWER OPERATION may continue.

d. Above 10% of RATED THERMAL POWER, the provisions of Specification 3.0.3 are not applicable.

ACTION 4 - With the number of OPERABLE channels one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a. Below the P-6 (Block of Source Range Reactor Trip) setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint.

b. Above the P-6 (Block of Source Range Reactor Trip) setpoint, operation may continue.

ACTION 5 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1 .1 or 3.1.1.2, as applicable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.3.1.1.1.

ACTION 7 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or THERMAL POWER is reduced to less than P-9 within 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months .

September 2, 2005 SEQUOYAH - UNIT 1 3/4 3-6 Amendment No. 47, 141, 304

TABLE 3.3-1 (Continued)

ACTION 8 - With less than the Minimum Number of Channels OPERABLE, declare the interlock inoperable and verify that all affected channels of the functions listed below are OPERABLE or apply the appropriate ACTION statement(s) for those functions.

Functions to be evaluated are:

a. Source Range Reactor Trip

b. Reactor Trip Low Reactor Coolant Loop Flow (2 loops)

Undervoltage Underfrequency Pressurizer Low Pressure Pressurizer High Level

c. Reactor Trip Low Reactor Coolant Loop Flow (1 loop)

d. Reactor Trip Intermediate Range Low Power Range Source Range

e. Reactor Trip Turbine Trip ACTION 9 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. For the affected protection set, the Trip Time Delay for one affected steam generator (Ts) is adjusted to match the Trip Time Delay for multiple affected steam generators (TM) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

c. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.3.1.1 .1.

ACTION 10 With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided that within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , for the affected protection set, the Trip Time Delays (T 5 and TM) threshold power level for zero seconds time delay is adjusted to 0% RTP.

May 16, 1990 SEQUOYAH - UNIT 1 3/4 3-7 Amendment No. 54, 141

TABLE 3. 3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TOTAL NO. MINIMUM OF CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT CHANNELS TO TRIP OPERABLE MODES ACTION

12. Loss of Flow Single Loop

- 3/loop 2/loop in 2/loop in 1 6 (Above P-8) any each operating operating loop loop

13. Loss of Flow Two Loops

- 3/loop 2/loop in two 2/loop in 1 6 (Above P-7 and below P-8) operating each loops operating loop

14. Main Steam Generator Water Level--Low-Low A. Steam Generator Water 3/Stm. Gen. 2/Stm. Gen. 2/Stm. Gen. 1,2 9 Level--Low-Low in any in each (Adverse) operating Operating Stm.Gen Stm.Gen.

B. Steam Generator Water 3/Stm. Gen. 2/Stm. Gen. 2/Stm. Gen. 1,2 9 Level--Low-Low (EAM) in any in each operating operating Stm.Gen. Stm.Gen.

C. RCS Loop T 4 (i/loop) 2 3 1,2 10 D. Containment Pressure 4 2 3 1,2 11 (EAM)

15. Deleted

16. Undervoltage-Reactor Coolant 4-1/bus 2 3 1 6 Pumps

17. Underfrequency-Reactor 4-i/bus 2 3 1 6 Coolant Pumps

18. Turbine Trip A. Low Fluid Oil 3 2 2 i 6 Pressure B. Turbine Stop Valve 4 4 4 1** 7 Closure September 2, 2005 SEQUOYAH - UNIT 1 3/4 3-3 Amendment No. 141, 301, 304

TABLE 3.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TOTAL NO. MINIMUM OF CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT CHANNELS TO TRIP OPERABLE MODES ACTION

19. Safetylnjectionlnputfrom 2 1 2 1,2 12 ESF

20. Reactor Trip Breakers A. Startupand Power 2 1 2 1,2 12,15 Operation B. Shutdown 2 1 2 3*,4* and 5* 16

21. Automatic Trip Logic A. Startup and Power 2 1 2 1, 2 12 Operation B. Shutdown 2 1 2 3*,4* and 5* 16

22. Reactor Trip System Interlocks A. Intermediate Range Neutron Flux, P-6 2 1 2 2, and* 8a B. Power Range Neutron 4 2 3 1 8b Flux, P-7 C. Power Range Neutron 4 2 3 1 8c Flux, P-8 D. Power Range Neutron 4 2 3 1, 2 8d Flux, P-lU E. Turbine Impulse Chamber 2 1 2 1 8b Pressure, P-13 F. Power Range Neutron 4 2 3 1 8e Flux, P-9 G. Reactor Trip P-4 2 1 2 1,2, and

14 July 20, 1987 SEQUOYAH - UNIT 1 3/4 3-4 Amendment No. 54, 56

NUCLEAR INSTRUMENT MALFUNCTION AOP-I.01 SQN Rev. 10 STEP ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED 2.3 Power Range Failure

1. PLACE rod control in MAN.

2. IF power increase is in progress, THEN STABILIZE reactor power at current level.

3. EVALUATE the following Tech Specs for applicability:

3.3.1.1 (3.3.1), Reactor Trip System Instrumentation

3.3.3.5, Remote Shutdown Instrumentation

3.3.3.7, Accident Monitoring Instrumentation

4.2.4.2, QPTR with one PR Channel Inoperable Page 14 of 49

SQN NUCLEAR INSTRUMENT MALFUNCTION AOP-I.O1 Rev. 10 I STEP ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED 2.3 Power Range Failure (contd)

9. IF auto rod control is desired, THEN RESTORE rod control to AUTO.

10. CHECK reactor power greater than 75%. GO TO Step 12.

11. NOTIFY Reactor Engineering to perform 0-Sl-NUC-000-01 1.0, Moveable Detector Determination of Quadrant Power Tilt Ratio.

12. NOTIFY l&C to remove failed power range channel from service USING appropriate Appendix:

POWER RANGE PROT APPENDIX CHANNEL CH N-41 I A N-42 II B N-43 Ill C N-44 IV D

13. GO TO appropriate plant procedure.

END OF SECTION Page 18 of49

SQN NUCLEAR INSTRUMENT MALFUNCTION AOP-I.01 Rev. 10 Page 1 of 7 APPENDIX A REMOVING POWER RANGE CHANNEL N-41 FROM SERVICE NOTES:

Tech Spec LCO 3.3.1.1 (3.3.1) is applicable in Modes 1 and 2 when this channel is inoperable.

. This is a time critical task and shall be performed without interruption.

A. Setup

1. RECORD the date and time the channel failed or was declared inoperable.

Date Time

2. OBTAIN the following approval to perform this appendix.

Unit US (SRO)

NOTE: Steps 3 and 4 may be performed later if required to meet Tech Spec LCO action time.

3. PREPARE 32 orange stickers with AOP number and date.

4. PLACE orange stickers on the following equipment:

ANNUNCIATORS PLACED (-I)

NC-41L NUC OVERPOWER ROD STOP BYPASS [XA-55-4A, A-3]

NIS POWER RANGE UPPER DETECTOR HI FLUX DEVN OR AUTO DEFEAT

[XA-55-4B, B-3]

NIS POWER RANGE LOWER DETECTOR HI FLUX DEVN OR AUTO DEFEAT

[XA-55-4B, C-3]

TS-68-2D REAC COOL LOOPS OVERTEMP z\T TRIP ALERT

[XA-55-6A, A-2]

TS-68-2G REAC COOL LOOPS OVERPOWER LT TRIP ALERT

[)(A-55-6A, B-2]

NC-41U/NC-41K NIS POWER RANGE HIGH NEUTRON FLUX RATE

[XA-55-6A, B-i]

NC-41P NIS PWR RANGE LOW SETPOINT HIGH FLUX LEVEL

[)(A-55-6A, C-i]

TS-68-2E OVERTEMP tT AUTO TURB RNBK BLK C-3 ROD WTD

[XA-55-6A, C-2}

TS-68-2F OVERPOWER zT AUTO TURB RNBK BLK C-4 ROD WTD

[XA-55-6A, D-21 NC-41 R NIS PWR RANGE HIGH SETPOINT HIGH FLUX LEVEL

[XA-55-6A, D-i]

(Step continued on next page)

Page 22 of 49

SQN NUCLEAR INSTRUMENT MALFUNCTION AOP-I.O1 I

Rev. 10 Page 2 of 7 APPENDIX A REMOVING POWER RANGE CHANNEL N-41 FROM SERVICE A. Setup (contd)

4. (Continued)

PLACED INDICATORS

(,

XI-92-5005A, A FLUX CH-1 N-41 [M-4]

XI-92-5005C, RX POWER CH-I N-41 [M-4]

TI-68-2B, OTAT LOOP 1 [M-5]

Tl-68-2A, OPAT LOOP 1 [M-5]

XI-92-5005B, POWER RANGE CH I REACTOR POWER [M-13, N41AJ PLACED STATUS LIGHTS [X 55 5]

NC41N P.R. P8 NC41M P.R. PlO NC41P P.R. HI F LO SET NC41 R P.R. HI PWR HI SET NC41 U/K P.R. HI PWR RATE NC41S P.R. P9 TS-68-2D RC LP1 OTAT REAC TRIP TS-68-2G RC LP1 OPAT REAC TRIP TS-68-2E RC LP1 OTAT TURB RNBK TS-68-2F RC LP1 OPAT TURB RNBK PROT. SET 1 TROUBLE (Step continued on next page)

Page 23 of 49

SQN NUCLEAR INSTRUMENT MALFUNCTION I AOP-I.O1 Rev.1O Page 3 of 7 APPENDIX A REMOVING POWER RANGE CHANNEL N-41 FROM SERVICE A. Setup (contd)

4. (Continued)

PLACED HAND SWITCHES XS-68-2B, LOOP TAVG AT_RECISEL_[M-5]

COMPARATOR CHANNEL DEFEAT [M-13, N37/N46]

POWER MISMATCH BYPASS [M-13, N50]

ROD STOP BYPASS [M-13, N50}

UPPER SECTION [M-13, N50]

LOWER SECTION [M-13, N50j

5. RECORD status of the following:

TRIP STATUS LIGHTS [)(J(..55..5] CHECK (I)

DARK LIT NC42R P.R. HI PWR HI SET NC42UIK P.R. HI PWR RATE NC43R P.R. HI PWR HI SET NC43U/K P.R. HI PWR RATE NC44R P.R. HI PWR HI SET NC44UIK P.R. HI PWR RATE TS-68-25D RC LP2 OTAT REAC TRIP TS-68-25G RC LP2 OPAT REAC TRIP TS-68-25E RC LP2 OTAT TURB RNBK TS-68-25F RC LP2 OPAT TURB RNBK (Step continued on next page)

Page 24 of 49

NUCLEAR INSTRUMENT MALFUNCTION AOP-I.O1 SQN I Rev.1O Page 4 of 7 APPENDIX A REMOVING POWER RANGE CHANNEL N-41 FROM SERVICE A. Setup (contd)

5. RECORD status of the following: (contd)

TRIP STATUS LIGHTS [)(J(..55..5] CHECK (

DARK LIT TS-68-44D RC LP3 OThT REAC TRIP TS-68-44G RC LP3 OPzT REAC TRIP TS-68-44E RC LP3 OThT TURB RNBK TS-68-44F RC LP3 OPAT TURB RNBK TS-68-67D RC LP4 OTL\T REAC TRIP TS-68-67G RC LP4 OPtT REAC TRIP TS-68-67E RC LP4 OThT TURB RNBK TS-68-67F RC LP4 OPAT TURB RNBK CAUTIONS:

If any of the status lights in Step 5 are LIT, completion of this Appendix will initiate a reactor trip signal.

Completion of this Appendix will trip the bistables that supply Ch I inputs to the following:

Overpower rod stop

OTAT turbine runback

OPAT turbine runback

P-7

P-8

P-9

P-b

6. NOTIFY Unit Operator to VERIFY current plant status allows removing this power range channel from service.

UO Initials Page 25 of 49

SQN NUCLEAR INSTRUMENT MALFUNCTION AOP-I.O1 I

Rev.1O Page 5 of 7 APPENDIX A REMOVING POWER RANGE CHANNEL N-41 FROM SERVICE B. Removal of Channel N-41 from Service

1. NOTIFY Unit Operator that Channel I of OThT AND OPAT will be placed in TRIPPED condition. El

2. PLACE the following manual bistable trip switches (at R-2) in TRIP:

TSI41 1 C (OTAT reactor trip) I cv

TS!41 1 D (OThT turbine runback/block rod withdrawal) /

cv

TSI411G (OPT reactor trip) /

cv

TSI411 H (OPT turbine runback/block rod withdrawal) I cv

3. VERIFY the following TRIP LEDs on Eagle 21 Test Panel LIT:

TB/411C

TBI411D

TBI411G

TBI411H El

4. NOTIFY Unit Operator to VERIFY the following trip status lights LIT [XX-55-5]:

TS-68-2D RC LP1 OTAT REAC TRIP

TS-68-2G RC LP1 OPAT REAC TRIP

TS-68-2E RC LP1 OThT TURB RNBK

TS-68-2F RC LP1 OPAT TURB RNBK Page 26 of 49

SQN NUCLEAR INSTRUMENT MALFUNCTION AOP-l.01

. Rev. 10 Page 6 of 7 APPENDIX A REMOVING POWER RANGE CHANNEL N-41 FROM SERVICE STEP COMPLETE eJ)

B. Removal of Channel N-41 from Service (contd)

5. NOTIFY Unit Operator that power range channel N-41 will be placed in TRIPPED condition.

6. REMOVE control power fuses from front of POWER RANGE A CH I drawer [M-13, XX-92-5027J.

cv

7. NOTIFY Unit Operator to VERIFY the following trip status lights LIT [XX-55-5}:

NC41NP.R.P8

NC41M P.R. PlO

NC41P P.R. HI F LO SET

NC41R P.R. HI PWR HI SET

NC41U/K P.R. HI PWR RATE

NC41SP.R.P9

8. VERIFY power range channel N-4l removed from service.

Page 27 of 49

SQN NUCLEAR INSTRUMENT MALFUNCTION AOP-I.O1 Rev.1O Page 7 of 7 APPENDIX A REMOVING POWER RANGE CHANNEL N-41 FROM SERVICE C. Completion

1. RECORD current date and time.

Date Time

2. NOTIFY Operations that Power Range Channel N-41 has been removed from service.

Unit US (SRO)

3. ENSURE Work Order (WO) initiated for Power Range Channel N-41.

WO Number

4. PREPARE copy of this completed appendix and forward to I&C Manager.

Page 28 of 49

OPT200.RPS Rev 5 Page 7 of 173

f. TRAINING OBJECTIVES:

B. Enabling Objectives, Continued:

A R S S U 0 R T 0 0 A X X X 7. EXPLAIN the operational implication of the following concept as it applies to the Reactor Protection System:

a. DNB

b. Power density X X X 8. Given specific plant conditions, ANALYZE the effect that a loss or malfunction of the Reactor Protection System will have on the following:

a. Reactor Trip Bkrs

b. ESFAS

c. Turbine trip bus

d. Steam Dump Controller

e. ReactorTrip Bkrs

f. MFW X X X 9. Given specific plant conditions, ANALYZE the effect that a loss or malfunction of the following will have on the Reactor Protection System:

a. Redundant channels

b. Trip logic circuits

c. Bypass/Bistable trip circuits

d. Test circuits

e. Permissive circuits

f. Trip setpoint calculators X X X X 10. DESCRIBE the design, procedural, and operational differences between units as they apply to control board/control room layout, instrumentation, and controls for the Reactor Protection System.

X X X 11. EXPLAIN and APPLY Reactor Protection System limits and precautions.

X X X X 12. Given a specific evolution, IDENTIFY the appropriate Reactor Protection System normal operating procedure(s) required to conduct that evolution.

OPT200.RPS Rev 5 Page 8 of 173 V. TRAINING OBJECTIVES:

B. Enabling Objectives, Continued:

ARSS UORT 0 OA X X 13. Using the Technical Specifications, Technical Requirements Manual, and the ODCM, X a. LIST from memory, Reactor Protection System Tech Spec LCOs and/or Technical Requirements having action times one hour,

b. EXPLAIN applicable Reactor Protection System Tech Spec LCO, Technical Requirements, and ODCM Control bases.

c. Given a set of plant conditions/parameters, DETERMINE entry level conditions for Reactor Protection System Tech Spec LCO actions, Technical Requirements, and/or ODCM Controls.

X X X 14. ANALYZE a given set of plant conditions and/or parameters to determine entry conditions into abnormal or emergency operating procedures.

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta

87. Given the following conditions:

- Unit I is operating at 100% power when the following annunciator alarms on 1-M-30:

1-RA-421A MN STM LN HI RAD The CRC reports the following Steam Line Radiation monitor readings:

Loop I Loop 2 Loop 3 Loop 4 mr/hr 9.35E-4% 9.95E-3% 1.12E-3% 8.9E-4%

- Pressurizer level is dropping rapidly.

- RCS activity is normal for a core with no fuel leaks.

- The SRO directs a reactor trip and safety injection to be initiated.

- When ready to isolate the affected steam generator, all 4 MSIV5 fail to close.

Which ONE of the following identifies...

(1) the steam generator requiring lower feed water flow to maintain level at setpoint prior to the reactor trip.

and (2) the procedure transition required due to the failure of the MSIVs to close?

Steam Generator Procedure A Loop 2 ECA 3.1, SGTR and LOCA Subcooled Recovery B. Loop 2 ECA 3.2, SGTR and LOCA Saturated Recovery C. Loop I ECA 3.1, SGTR and LOCA - Subcooled Recovery D. Loop 1 ECA 3.2, SGTR and LOCA Saturated Recovery Thursday, July 15, 2010 4:06:33 PM 87

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta DISTRA CTOR ANAL YSIS:

A. Correct, the alarm described is common to all four steam line rad monitors but the indication for loop #2 is a decade higher than the other three loops. E-3 step 4 will direct entry into ECA-3. Ibecause the ruptured S/G cant be isolated from at least one intact S/G due to the failure of all four MSIVs.

B. Incorrect, Plausible because the indication for ioop #2 steam line rad monitor is a decade higher than the other three loops and ECA-3. 2 is one of two procedures that deals with simultaneous occurance of a S/G tube rupture and a LOCA.

C. Incorrect, the alarm engraving may lead the candidtate to identify Loop #1 as the failed loop because it has the same identifier as the loop #1 rad monitor. Incorrect because loop #2 is a decade higher than the other three loops. ECA-3. 1 is the correct contingency instruction for the failed MSIVs.

D. Incorrect, the alarm engraving may lead the candidtate to identify Loop #1 as the failed ioop because it has the same identifier as the loop #1 rad monitor and ECA-3. 2 is one of two procedures that deals with simultaneous occurance of a S/G tube rupture and a LOCA. Incorrect because the rupture is on loop #2 and ECA-3. I is the correct contingency instruction for the failed MSIVs.

Thursday, July 15, 2010 4:06:33 PM 88

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta Question Number: 87 Tier: 2 Group 1 KIA: 039 Main and Reheat Steam System (MRSS)

A2.03 Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Indications and alarms for main steam and area radiation monitors (during SGTR)

Importance Rating: 3.4 / 3.7 10 CFR Part 55: 41.5, 43.5, 45.3, 45.13 IOCFR55.43.b: 5 KIA Match: Match because the candidate must interpret the alarm and steam line radiation monitor indication to identify a ruptured SIG. SRO because knowledge of emergency contingency procedures is required.

Technical

Reference:

AOP-R.01, rev 26; E-3 Steam Generator Tube Rupture Rev 17; ECA-3.1 SGTR and LOCA-Subcooled Recovery Rev 14.

Proposed references None to be provided:

Learning Objective: OPL271AOP-R.01 Obj 0,2, 4; OPL271E-3 Obj 2 & 5; OPL271ECA-3.1 Obj 2.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History:

Comments:

MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9 Answer: ADACCABDDB ScrambleRange:A-D 4

Thursday, July 15, 2010 4:06:33 PM 89

SQN STEAM GENERATOR TUBE LEAK AOP-R.O1 Rev. 26 STEP ACTIONIEXPECTED RESPONSE I RESPONSE NOT OBTAINED 2.1 SIG Tube Leak Requiring Rapid Shutdown (contd)

4. MONITOR indications of leaking S/G:

a. NOTIFY Chem Lab to evaluate Primary to Secondary Leakage USING 1(2)-Sl-CEM-068-1 37.5:

Method 1, Rapid Identification of Leaking Steam Generators

Method 3, Condenser Vacuum Exhaust (CVE) Sampling for Determination of Primary-to-Secondary (PIS) Leakage.

b. NOTIFY RADCON to monitor Turbine Building and site environment:

Steam lines

SIG blowdown

c. IDENTIFY leaking S/G(s)

USING any of the following:

Unexpected rise in any SIG narrow range level OR

S/G sample results OR

RADCON survey of main steamlines and S/G blowdown lines OR

High radiation on any main steamline radiation monitor.

Page 6of68

SQN STEAM GENERATOR TUBE RUPTURE ISTEPI [ACTIONIEXPECTED RESPONSE I RESPONSE NOT OBTAINED I

4. a. CLOSE Ruptured SIG(s) MSIV and e. PERFORM the following:

MSIV bypass valve.

1) CLOSE Intact SIG MSIVs and MSIV bypass valves.

2) DISPATCH operator to perform EA-1-1, Closing MSIVs Locally, for MSIV or MSIV bypass valve which fails to close.

3) ISOLATE steam header:

PLACE condenser steam dumps in OFF. [M-4]

ENSURE steam dump valves CLOSED. [M-4]

CLOSE FCV-47-1 80, HP Steam Seal Supply Isolation. [M-21

ENSURE FCV-47-181 HP Steam Seal Supply Bypass CLOSED. [M-2]

CLOSE MSR HP steam supply isolation valves. [M-21

DISPATCH operator to locally isolate steam header USING EA-1-4, Local Isolation of Steam Header in Turb Bldg.

4) USE Intact SIG(s) atmospheric relief for steam dump.

IF any Ruptured S/G CANNOT be isolated from at least one Intact SIG, THEN GO TO ECA-3.1, SGTR and LOCA -

Subcooled Recovery.

Page7of4l

SQN SGTR AND LOCA SUBCOOLED RECOVERY ECA-3.1 Rev. 14 ISTEPI ACTION/EXPECTED RESPONSE I IRESPONSENOTOBTAINED

16. MONITOR if subcooled recovery appropriate:

a. CHECK RWST level a. REFER TO Curve 10, RWST vs.

greater than 70%. Containment Sump Indicated Level.

IF any of the following conditions exists:

Curve 10 indicates transition to ECA-3.2 is needed (below line)

OR

NO train of containment sump recirculation capability available, THEN GO TO ECA-3.2, SGTR and LOCA

- Saturated Recovery.

b. CHECK Ruptured SIG narrow range b. CONSULT TSC to determine level less than 84% [80% ADV]. if recovery should be completed USING ECA-3.2, SGTR and LOCA -

Satu rated Recovery.

Page 16 of 43

OPL27IAOP-R.O1 Revision I Page 3of47 I. PROGRAM: OPERATOR TRAINING LICENSED-II. COURSE: LICENSE TRAINING III. LESSON TITLE: AOP-R.01, STEAM GENERATOR TUBE LEAK IV. LENGTH OF LESSONICOURSE: 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months V. TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of AOP-R.01, Steam Generator Tube Leak B. Enabling Objectives:

0. Demonstrate an understanding of NUREG 1122 Knowledges and Abilities associated with Reactor Coolant Pump Malfunctions that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A.

1. Explain the purpose/goal of AOP-R.01.

2. Discuss the AOP-R.01 entry conditions.

a. Describe the setpoints, interlocks, and automatic actions associated with AOP-R.01 entry conditions.

b. Describe the ARP requirements associated with AOP-R.01 entry conditions.

c. Interpret, prioritize, and verify associated alarms are consistent with AOP R.01 entry conditions.

d. Describe the Administrative conditions that require Turbine Trip/ Reactor trip due to Reactor Coolant Pump Malfunctions.

3. Describe the initial operator response to stabilize the plant upon entry into AOP Rd.

4. Upon entry into AOP-R.01, diagnose the applicable condition and transition to the appropriate procedural section for response.

5. Summarize the mitigating strategy for the failure that initiated entry into AOP R.01.

6. Describe the bases for all limits, notes, cautions, and steps of AOP-R.01.

0PL271 AOP-R.01 Revision I Page 47 of 47 APPENDIX A APE: 037 Steam Generator (SIG) Tube Leak (Page 2 of 2)

IMPORTANCE K/A NO. ABILITY RO/SRO AA1. Ability to operate and I or monitor the following as they apply to the Steam Generator Tube Leak: (CFR 41.7 / 45.5 / 45.6)

AA 1.01 Maximum controlled depressurization rate for affected S/G. 3.7/3.6 AA1.02 Condensate exhaust system. 3.1*/2.9 AA1.03 Loop isolation valves 3.0*72.9 AA1 .04 Condensate air ejector exhaust radiation monitor and failure 3.6/3.9 indicator AA1 .05 Radiation monitor for auxiliary building exhaust processes 3.3/3.5 AA1 .06 Main steam line rad monitor meters 3.8*73.9*

AA1.07 CVCS letdown flow indicator 3.1/3.2 AA1.08 Charging flow indicator 3.3/3.1 AA1.09 RCS loop pressure indicators 3.3/3.2 AA1.10 CVCS makeup tank level indicator 2.9/3.1 AA1. 11 PZR level indicator 3.4/3.3 AA1.12 Control panel power range channel recorders 2.3*72.5*

AA 1.13 S/G blowdowri radiation monitors 3.9/4.0 AA2. Ability to determine and interpret the following as they apply to the Steam Generator Tube Leak: (CFR: 43.5 / 45.13)

AA2.01 Unusual readings of the monitors; steps needed to verify 3.0/3.4 readings AA2.02 Agreement/disagreement among redundant radiation 3.4/3.9 monitors..

AA2.03 That the expected indication on main steam lines from the 3.4/3.9 S/Gs_should_show_increasing radiation_levels.

AA2.04 Comparison of RCS fluid inputs and outputs, to detect leaks 3.4/3.7 AA2.05 Past history of leakage with current problem 2.8/3.3 AA2.06 S/G tube failure 4.3/4.5 AA2.07 Flowpath for dilution of ejector exhaust air 3.1/3.6 AA2.08 Failure of Condensate air ejector exhaust monitor 2.8/ 3.3 AA2.09 System status, using independent readings from redundant 2.8*73 4*

Condensate_air_ejector_exhaust monitor AA2.10 Tech-Spec limits for RCS leakage 3.2/4.1 AA2.11 When to isolate one or more S/Gs 3.8/3.8*

AA2.12 Flowrateofleak. 3.3/4.1 AA2.13 Which S/G is leaking. 4.1/4.3 AA2. 14 Actions to be taken if S/G goes solid and water enters steam 4.0/4.4 lines AA2. 15 Magnitude of atmospheric radioactive release if cool-down 3.4*74.2 must_be_completed_using_steam_dump_or_atmospheric_reliefs AA2.16 Pressure at which to maintain RCS during S/G cooldown 4.1/4.3

0PL271 E-3 Revision 0 Page 3 of 24 I. PROGRAM: OPERATOR TRAINING - LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: E-3, Steam Generator Tube Rupture IV. LENGTH OF LESSONICOURSE: 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months V TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of E-3, Steam Generator Tube Rupture B. Enabling Objectives

0. Demonstrate an understanding of NUREG 1122 Knowledges and Abilities associated with Steam Generator Tube Rupture that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A.

1. Explain the purpose/goal of E-3.

2. Discuss the E-3 entry conditions.

a. Describe the setpoints, interlocks, and automatic actions associated with E-3 entry conditions.

b. Describe the requirements associated with E-3 entry conditions.

3. Summarize the mitigating strategy for the failure that initiated entry into E-3.

4. Describe the bases for all limits, notes, autions, and steps of E-3.

5. Describe the conditions and reason for transitions within this procedure and transitions to other procedures.

6. Given a set of initial plant conditions use E-3 to correctly:

a. Identify required actions

b. Respond to Contingencies

c. Observe and Interpret Cautions and Notes

7. Apply GFE and system response concepts to the performance of E-3 conditions.

0PL271 ECA-3. I Revision I Page 3 of 24 I. PROGRAM: OPERATOR TRAINING - LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: ECA-3.1, SGTR and LOCASubcooled Recovery IV. LENGTH OF LESSONICOURSE: 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months V. TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of ECA-3.l, SGTR and LOCA Subcooled Recovery.

B. Enabling Objectives

0. Demonstrate an understanding of NUREG 1122 Knowledges and Abilities associated with ECA-3i, SGTR and LOCA Subcooled Recovery that are rated 2.5 during Initial License Training and 3.0 during License Operator Req ualification Training for the appropriate license position as identified in Appendix A.

1. Explain the purpose/goal of ECA-3. 1.

2. Discuss the ECA-3.1 entry conditions.

3. Summarize the mitigating strategy for the failure that initiated entry into ECA-3.1.

4. Describe the bases for all limits, notes, cautions, and steps of ECA-3.1.

5. Describe the conditions and reason for transitions within this procedure and transitions to other procedures.

6. Given a set of initial plant conditions use ECA-3. I to correctly:

a. Identify required actions

b. Respond to Contingencies

c. Observe and Interpret Cautions and Notes

7. Apply GFE and system response concepts to the performance of ECA-3.I conditions.

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta

88. Given the following plant conditions:

- Both Units operating at 100% power.

- ERCW system in normal alignment.

- The following 0-XA-55-27A panel annunciators are LIT:

- UNIT I HEADERA PRESSURE LOW

- UNIT 2 HEADER A PRESSURE LOW

- PUMP J-A DISCH PRESS LOW

- PUMP Q-A DISCH PRESS LOW.

- The following 1-XA-55-15B panel annunciator is LIT:

- ERCW DECK SUMP PUMP A RUNNING.

- ERCW headers 1A and 2A are indicating LOW flow.

- The crew implements AOP-M.01, Loss of Essential Raw Cooling Water.

Which ONE of the following identifies the section of AOP-M.01 to be implemented for the conditions and a mitigating action directed to be taken in response to the conditions?

A. Section 2.8, Supply Header IA/2A Rupture in Yard Area; Stop and Lockout out all A Train ERCW pumps.

B. Section 2.8, Supply Header 1A/2A Rupture in Yard Area; Start additional ERCW pumps to maintain header pressure.

C. Section 2.10, Supply Header A Rupture Upstream of Strainer Inlet Valves OR Loss of Train B ERCW; Stop and Lockout out all A Train ERCW pumps.

D. Section 2.10, Supply Header A Rupture Upstream of Strainer Inlet Valves OR Loss of Train B ERCW; Start additional ERCW pumps to maintain header pressure.

Wednesday, July 07, 2010 11:31:05 AM 88

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta DISTRACTOR ANALYSIS:

A. Incorrect, Plausible because Section 2.8 would be the correct procedure section for a leak downstream of the strainer and the stopping and locking out of the pumps is correct for the leak identified in the stem. Conditions match entry for Section 2.8 except for sump pump running alarm and the Strainer Dp alarm being dark.

B. Incorrect, Plausible because Section 2.8 would be the correct procedure section for a leak downstream of the strainer and starting additional pumps is a mitigating action during performance of Section 2.8. Conditions match entry for Section 2.8 except for sump pump running alarm and the Strainer Dp alarm being dark.

C. Correct, All alarms stated would be lit for a header break upstream of the Train A Strainers. (Section 2.10) Header pressure sensors are located just upstream of the strainers. The sump pump running differentiates the leak upstream from a yard leak (I .e. downstream of the strainer). Mitigating action in AOP section 2.10 is to stop and lock out all A Train pumps.

D. Incorrect, Plausible because all alarms stated would be lit for a header break upstream of the Strainer making Section 2.10 the correct procedure section. While starting additional pumps might restore pressure, the mitigating action is to stop and lock out the A Train pumps to terminate the leakage.

Question Number: 88 Tier: 2 Group I KIA: 076 G 2.4.4 Service Water System (SWS)

Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

Importance Rating: 4.5 / 4.7 IOCFRPart55: 41.10/43.2/45.6 IOCFR55.43.b: 5 K!A Match:

Technical

Reference:

AOP-M.01, Loss of Essential Raw Cooling Water, Rev 22 Wednesday, July 07, 2010 11 31 :05 AM 89

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta Proposed references None to be provided:

Learning Objective: OPL271AOP-M.01, B.4 & 5 Upon entry into AOP-M.1, diagnosis the applicable condition and transition to the appropriate procedural section for response.

Summarize the mitigating strategy for the failure that initiated entry into AOP-M.01 Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: SQN question 076A2.02 090 used on 1/2009 exam with correct answer relocated from D to C Comments: Question 076A2.02 090 orginally SQN question 076 A2.02 053 with modification to correct answer, all distractors, and stem. Used most plausible 2 of original distractors and added requirement to identify mitigating actions. Changed correct answer to D. No significant modification to data in the stem.

Wednesday, July 07, 2010 11:31:05 AM 90

AOP-M.01 SQN LOSS OF ESSENTIAL RAW COOLING WATER Rev. 22 I STEP ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED 2.0 OPERATOR ACTIONS (Continued)

1. (Continued)

IF... GO TO SECTION PAGE Low flow ERCW supply headers IA and 2A, AND STRAINER DIFF PRESS alarms DARK

[M-27A, C-3 and D-2j, AND at least one of the following alarms LIT:

. ERCW DECK SUMP LEVEL HI alarm LIT 2.10 Supply Header A 86

[1-M-15B, A-3] Rupture Upstream of Strainer Inlet Valves OR OR Loss of Train A

. ERCWDECKSUMPPMP RUNNING ERCW

[l-M-15B, D-2 or D-4]

OR

MECH EQUIP SUMP LVL HI alarm LIT

[1-M-15A,_B-6}

Low flow ERCW supply headers 1 B and 2B, AND STRAINER DIFF PRESS alarms DARK

[M-27A, C-6 and D-5],

AND at least one of the following alarms LIT:

. ERCW DECK SUMP LEVEL HI alarm LIT 2.11 Supply Header B 92

[1-M-15B, A-3] Rupture Upstream of Strainer Inlet Valves OR OR Loss of Train B

. ERCWDECKSUMPPMPRUNNING ERCW

[1-M-15B, D-2 or D-4]

OR

MECH EQUIP SUMP LVL HI alarm LIT

[1-M-15A,_B-6]

Loss of flow on ALL ERCW supply headers 2.12 Loss of all ERCW flow 97 in modes 1-4.

END OF SECTION Page 4 of 185

SQN LOSS OF ESSENTIAL RAW COOLING WATER AOP-M.O1 Rev. 22 I STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.0 OPERATOR ACTIONS CAUTION: ERCW header rupture in Aux Bldg could fill passive sump in 15 minutes.

Prompt action may be needed to isolate rupture using this AOP.

1. DIAGNOSE the failure:

IF... GO TO SECTION PAGE ERCW Pump(s) tripped or failed 2.1 ERCW pump failure 5 Controllable ERCW leak: 2.2 Controllable ERCW 8 Rapid isolation of leak is NOT required based on leak safety or equipment damage concerns AND Adequate flow to ERCW components available.

High flow ERCW Supply Header IA 2.3 Supply Hdr 1A Rupture 15 in Aux_Bldg High flow ERCW Supply Header 1 B 2.4 Supply Hdr 1 B Rupture 20 in Aux Bldg High flow ERCW Supply Header 2A 2.5 Supply Hdr 2A Rupture 24 in Aux Bldg High flow ERCW Supply Header 2B 2.6 Supply Hdr 2B Rupture in Aux Bldg Indications of an ERCW Return Header Rupture 2.7 Return Hdr Rupture 40 (must be diagnosed locally), in Aux Bldg NOTE: If alarms listed in remainder of this table are NOT lit, ERCW pressure/flow indications and reports from personnel outside MCR should be used to determine appropriate section.

Low flow ERCW Supply Header 1A and 2A, 2.8 Supply Header 1N2A 56

,ANP Rupture in Yard Area SLAINER DIFF PRESS HIGH alarm LITj

[M-27A, C-3 and/or_D-2]

Low flow ERCW Supply Header 1 B and 2B, 2.9 Supply Header 1 B/2B 72 AND Rupture in Yard Area STRAINER DIFF PRESS HIGH alarm LIT

[M-27A, C-6 and/or D-5}

(step continued on next page)

Page 3 of 185

SQN LOSS OF ESSENTIAL RAW COOLING WATER AOP-M.01 Rev. 22 STEP ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED 2.10 ERCW Supply Header A Rupture. Upstream of Strainer Inlet Valves OR Loss of Train A ERCW CAUTION: If running, the affected CCPs and SI Pumps may experience bearing failure 10 minutes after loss of ERCW.

1. STOP and LOCK OUT all Train A ERCW Pumps.

CAUTION: Pipe rupture inside ERCW Pumping Station could result in safety hazards due to water spray and flooded compartments.

2. IF rupture is inside ERCW pumping station OR location is unknown, THEN DISPATCH operators with radios to perform the following at ERCW Pumping Station:

PERFORM Appendix G.1, ERCW MCC Valves.

ENSURE all watertight doors at ERCW pumping station are CLOSED.

3. OPEN alternate ERCW supply to DIGs to ensure cooling water is available:

1-FCV-67-68 (D/G IA-A)

2-FCV-67-68 (D/G 2A-A)

Page 86 of 185

LOSS OF ESSENTIAL RAW COOLING WATER AOP-M.01 SQN Rev. 22 STEP ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED 2.8 ERCW Supply Header IAI2A Rupture in Yard Area CAUTION I If running, the affected CCPs and SI Pumps may experience bearing failure 10 minutes after loss of ERCW cooling.

CAUTION 2 Loss of 2A ERCW Supply Header affects both Units Train A CCS Heat Exchangers. Both units will be tripped if 2A supply header is isolated.

NOTE: Engineering may be able to identify ruptured yard header using yard piping drawings (17W300 series).

1. DISPATCH personnel to locate rupture.

NOTE: If any of conditions in Step 2 are met, rapid isolation is necessary to prevent major equipment damage.

2. CHECK if rapid isolation is required GO TO Caution prior to Step 4.

based upon any of the following: I rupture Inside ERCW pumping station OR

rupture inside CCW intake pumping station OR

rupture in ERCW pipe tunnel draining to Aux Bldg.

3. GOTOStep7.

I Page 56 of 185

SQN LOSS OF ESSENTIAL RAW COOLING WATER AOP-M.O1 Rev. 22 I STEP ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED 2.8 ERCW Supply Header IAI2A Rupture in Yard Area (contd)

CAUTION If DIG is supplying shutdown board, DIG loading should be monitored to prevent overloading.

NOTE If leak is in yard area, up to four Train A ERCW pumps may be started.

4. START additional Train A ERCW Pumps as required to control header pressure between 78 psig and 124 psig.

5. VERIFY the following: IF rapid isolation of rupture is required, THEN

Adequate header pressure and flow can be maintained. GO TO Step 7.

rapid isolation of rupture is NOT required based upon safety or equipment damage concerns.

6. GO TO Section 2.2, Controllable ERCW Leak.

Page 57 of 185

OPL271AOP-M.01 Revision 0 Page 3 of 45 I. PROGRAM: OPERATOR TRAINING - LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: AOP-M.01 LOSS OF ESSENTIAL RAW COOLING WATER IV. LENGTH OF LESSON!COURSE: 2.0 hour0 days 0 hours 0 weeks 0 months (s)

V. TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of AOP-M.01, LOSS OF ESSENTIAL RAW COOLING WATER B. Enabling Objectives:

0. Demonstrate an understanding of NUREG 1 122 Knowledges and Abilities associated with a Loss of Essential Raw Cooling Water that are rated 2.5 during Initial License Training and 3.0 during License Operator Req ualification Training for the appropriate license position as identified in Appendix A.

1. Explain the purpose/goal of AOP-M.01.

2. Discuss the AOP-M.01 entry conditions.

a. Describe the setpoints, interlocks, and automatic actions associated with AOP-M.01 entry conditions.

b. Describe the ARP requirements associated with AOP-M.01 entry conditions.

c. Interpret, prioritize, and verify associated alarms are consistent with AOP M.01 entry conditions.

d. Describe the Administrative conditions that require Turbine Trip/ Reactor trip due to Loss of Essential Raw Cooling Water.

3. Describe the initial operator response to stabilize the plant upon entry into AOP M.01.

4. Upon entry into AOP-M.01, diagnose the applicable condition and transition to the appropriate procedural section for response.

5. Summarize the mitigating strategy for the failure that initiated entry into AOP M.01.

6. Describe the bases for all limits, notes, cautions, and steps of AOP-M.01.

QUESTIONS REPORT for Friday

89. Given the following conditions:

- Both units are at 100% power when Control Air pressure begins to drop.

- AOP-M.02, Loss of Control Air, Section 2.2, Loss of Nonessential Control Air in MODE 1, 2, or 3, is currently being implemented.

- AUOs are dispatched to determine the cause of the air pressure loss.

- AUO sent to check Auxiliary Air System reports that both Aux Air compressors are running and loaded but the there is no air passing through the Train-A air dryers.

- Air pressures on 1-M-15 indicate:

0-Pl-32-104A, AUX CONT AIR HDR A PRESS - 59 psig and dropping 0-PI-32-105A, AUX CONT AIR HDR B PRESS - 77 psig and rising 0-PI-32-200, CONT AIR HDR PRESS - 68 psig and dropping 0-PI-33-199, SERV AIR HDR PRESS - 40 psig and dropping Which ONE of the following identifies..

(1) the action required by AOP-M.02 and (2) an applicable Tech Spec entry if the action taken is not successful in restoring A Train Aux Air header pressure?

A. (1) Bypass the Train-A Aux Air Dryers.

(2) Enter T.S. 3.7.1 .2.a for one train of AFW INOPERABLE.

B (1) Bypass the Train-A Aux Air Dryers.

(2) Enter T.S. 3.7.1 .2.b for two trains of AFW INOPERABLE.

C. (1) Open 0-FCV-32-82, Aux Cmpsr A-A Aux Bldg Isol.

(2) Enter T.S. 3.7.1 .2.a for one train of AFW INOPERABLE.

D. (1) Open 0-FCV-32-82, Aux Cmpsr A-A Aux Bldg IsoI.

(2) Enter T.S. 3.7.1 .2.b for two trains of AFW INOPERABLE.

Friday, July 16, 2010 10:23:55 AM 89

QUESTIONS REPORT for Friday DISTRA CTOR ANAL YSIS:

A. Correct. Bypassing the malfunctioning air dryer is the correct response for the conditions described and entering LCO Action 3.7. 1.2.a is correct because the A Train MD AFW pump is Inoerable due to the loss of the A Train Aux Air Header pressure.

B. Incorrect, Plausible because bypassing the malfunctioning air dryer is the correct response for the conditions described and the A Train Aux Air Header supplies air pressure to both the A MD AFW pump and the TDAFW pump. Incorrect because the TDAFW pump is not made Inoperable by the loss of air.

C. Incorrect, Plausible because manipulating O-FCV-32-82 is addressed in step 12 and opening the valve would allow control air to be supplied to the aux air system and 3.7. 1.2.a is the correct Tech Spec. Incorrect because the procedure ensures O-FCV-32-82 is closed if header pressure is less than 69 psig.

D. Incorrect, Plausible because manipulating O-FCV-32-82 is addressed in step 12 and the A Train Aux Air Header supplies air pressure to both the A MD AFW pump and the TDAFW pump. Incorrect because the procedure ensures O-FCV-32-82 is closed if header pressure is less than 69 psig and the TDAFW pump is not made Inoperable by the loss of air.

Friday, July 16, 2010 10:23:55 AM 90

QUESTIONS REPORT for Friday Question Number: 89 Tier: 2 Group 1 KIA: 078 Instrument Air System (lAS)

A2.01 Ability to (a) predict the impacts of the following malfunctions or operations on the lAS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Air dryer and filter malfunctions Importance Rating: 2.4 / 2.9 10 CFR Part 55: 41.5/43.5 / 45.3/45.13 IOCFR55.43.b: 5 K/A Match: KA is matched because the question requires knowledge of the impact of an air dryer malfunction on the air system and how procedures are used to correct/mitigate the consequences of the malfunction. SRO because the question requires use of plant conditions to determine Tech Spec applicability.

Note: The instrument air system and the process air system at Sequoyah use the same air compressors, filters and dryers.

Technical

Reference:

AOP-M.02, Loss of Control Air, Revision 16 TS 3.7.1.2 Proposed references None to be provided:

Learning Objective: 0PL271 AOP-M .02

8. Given a set of initial plant conditions use AOP-M.02 to correctly:

b. Identify required actions Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: McGuire 2009 exam question modified 2009 Comments:

Friday, July 16, 2010 10:23:55 AM 91

STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.0 OPERATOR ACTIONS

1. DIAGNOSE the failure:

IF... GOTO SECTION PAGE Auxiliary Air is lost 2.1 4 Unit is in Mode 1, 2, or 3 2.2 11 AND Nonessential Control Air is lost Unit is in Mode 4, 5, or 6 2.3 27 AND Nonessential Control Air is lost Page 3of64

AOP-M.02 SQN LOSS OF CONTROL AIR Rev. 16 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 Loss of Auxiliary Air

1. MONITOR containment pressure MAINTAIN containment pressure between -0.1 psig and + 0.3 psig. USING 0-SO-30-8, Containment Pressure Control.

2. CHECK P1-32-200, control air header PERFORM the following:

pressure, greater than 77 psig.

a. DISPATCH personnel to perform the following:

1) ENSURE Aux Air Compressor RUNNING and LOADED on affected train.

2) VERIFY Aux Air Dryers and Filters operating properly:

Cam Timer Off

One Tower with Purge Air Flow.

3) BYPASS plugged or malfunctioning Aux Air Dryers and Filters:

No air passing through In-Service Dryer

Leakage of some Purge Air flow through In-Service Tower.

b. IF control air header pressure drops to less than 69 psig, THEN ENSURE following valves CLOSED:

0-FCV-32-82, Aux Cmpsr A-A Aux Bldg Isol [Aux Bldg, 734 elev]

0-FCV-32-85, Aux Cmpsr B-B Aux Bldg Isol [Aux Bldg, 734 elev]

Page 4 of 64

SQN LOSS OF CONTROL AIR AOP-M.02 Rev. 16 I STEP ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 Loss of Auxiliary Air (contd)

3. CHECK auxiliary control air header EVALUATE Tech Specs pressure: USING Appendix F, Tech Spec Evaluation.

CHECK P1-32-104, auxiliary control air header A pressure, greater than 70 psig.

CHECK P1-32-105, auxiliary control air header B pressure, greater than 70 psig.

4. DISPATCH personnel to identify and isolate air leak as close to leak as possible.

5. NOTIFY Maintenance to initiate repairs to source of air leakage.

6. EVALUATE valves and equipment properly aligned USING Appendix A, Failure Positions on Loss of Aux Control Air. [C.1J Page 5 of 64

LOSS OF CONTROL AIR AOP-M.02 SQN I Rev. 16 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 7 2.2 Loss of Nonessential Control Air in MODE 1, 2, or 3

1. CHECK loss of non-essential air IF non-essential air was deliberately NOT due to operator action in EA-32-3, depressurized due to performing EA-32-3, Isolating Non-Essential Air to Containment. THEN GO TO Step 12.

2. DISPATCH personnel to check the following:

. Control Air Compressors

Control Air Dryers and Filters

Cooling Water to Control Air Compressors

3. MONITOR S!G levels on program. IF auto reactor trip on low SIG level is imminent, THEN TRIP Reactor and GO TO E-O, Reactor Trip or Safety Injection, WHILE continuing with this procedure.

4. MONITOR P1-33-199, service air header ENSURE O-PCV-33-4, Service Air Isol Valve, pressure greater than 88 psig. CLOSED.

Page 11 of 64

SQN LOSS OF CONTROL AIR AOP-M.02 Rev. 16 STEP ACTION!EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.2 Loss of Nonessential Control Air in MODE 1, 2, or 3 (contd)

CAUTION: LCO 3.7.15 (for Train A MCR Chiller) and TR 3.7.14 (for Train A EBR Chiller) may apply if O-FCV-67-205, Train A ERCW to Station Air Compressors, is OPEN with ERCW temperature greater than 81°F.

5. CHECK RCW supply to Station Air ENSURE ERCW cooling aligned to Station Compressors AVAILABLE: Air Compressors:

RCW Pumps RUNNING

1) OPEN fO-FCV-67-2081 lB ERCW Supply Header Isolation [M-27A].

RCW Pressure greater than 68 psig

[O-Pl-24-22, 1-M-15] 2) IF lB ERCW Supply to Station Air Compressors is Unavailable, THEN OPEN [O-FCV-67-2051 IA Supply Header Isolation

3) REFER to AOP-M.05, Loss of RCW, While Continuing in this procedure

6. DISPATCH Operator to start Control Air Compressors as required to maintain system pressure.

7. VERIFY A and B Control Air Compressors LOAD compressors locally LOADED as required. USING Appendix E, Control Air Compressor Hand Loading.

Page 12 of 64

SQN LOSS OF CONTROL AIR AOP-M.02 Rev. 16 STEP I ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.2 Loss of Nonessential Control Air in MODE 1, 2, or 3 (contd)

8. DETERMINE whether leak is common header or train header, by the indications on the following indicators:

[0-L-66, Turb. Bldg, 685 elev, behind Control Air Dryers]:

0-F 1-32-3, Control Air Header Train A

0-Fl-32-4, Control Air Header Train B NOTE The loss of air pressure to LCV-6-106 may result in a #3 HDT Runback if Turbine Load is above runback setpoint.

9. DISPATCH personnel to identify and isolate air leak as close to leak as possible.

10. VERIFY Control Air Dryers and Filters EVALUATE bypassing plugged or operating properly. malfunctioning Control Air Dryers and Filters.

11. NOTIFY Maintenance to initiate repairs to source of air leakage.

Page 13 of 64

SQN LOSS OF CONTROL AIR AOP-M.02 Rev. 16 STEP ACTION!EXPECTED RESPONSE I RESPONSE NOT OBTAINED 2.2 Loss of Nonessential Control Air in MODE 1, 2, or 3 (contd)

12. MONITOR P1-32-200, control air header PERFORM the following:

pressure, greater than 77 psig.

a. DISPATCH personnel to perform the following:

1) ENSURE Aux Air Compressor RUNNING and LOADED on affected train.

2) VERIFY Aux Air Dryers and Filters operating properly:

Cam TimerOff.

One Tower with Purge Air Flow.

3) BYPASS plugged or malfunctioning Aux Air Dryers and Filters:

No air passing through In-Service Dryer.

Leakage of some Purge Air flow through In-Service Tower.

b. IF Aux Air pressure drops to less than 69 psig, THEN ENSURE following valves CLOSED:

0-FCV-32-82, Aux Cmpsr A-A Aux Bldg lsol [Aux Bldg, el. 734]

0-FCV-32-85, Aux Cmpsr B-B Aux Bldg Isol [Aux Bldg, el. 734]

c. IF Aux Air pressure NOT recovering, THEN GO TO Section 2.1, Loss of Auxiliary Air.

Page 14 of 64

AOP-M.02 SQN LOSS OF CONTROL AIR Rev. 16 STEP I ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED 2.2 Loss of Nonessential Control Air in MODE 1, 2, or 3 (contd)

13. EVALUATE Tech Specs USING Appendix F, Tech Spec Evaluation.

14. CHECK reactor in MODE I or 2. MAINTAIN HOT STANDBY conditions:

USE SIG atmospheric relief valves

USE AFW NOTE MSIV solenoids will remain energized on a loss of air with MSIVs in the open position.

The MSIVs will reopen when air is restored unless MSIV handswitches are placed in the close position.

15. MONITOR MSIVs OPEN. IF MSIVs are drifting closed, THEN

a. ENSURE reactor tripped.

b. PLACE MSIV handswitches in CLOSE.

16. CHECK the following valves OPEN IF control air header pressure is

[M-6, Status Panels 6K and 6L]: greater than 77 psig AND Phase B NOT actuated,

1-FCV-32-11O, Nonessential Air to THEN Rx Bldg U-I Isol OPEN valve(s) as follows:

2-FCV-32-l 11, Nonessential Air to a. MONITOR Containment pressure Rx Bldg U-2 Isol between -0.1 and 0.3 psid.

b. DISPATCH operator with radio to open valves USING EA-32-l, Establishing Instrument Air to Containment.

Page 15 of 64

AOP-M.02 SQN LOSS OF CONTROL AIR Rev. 16 I STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.2 Loss of Nonessential Control Air in MODE 1, 2, or 3 (contd)

17. MAINTAIN RCS pressure STABLE and CONTROLLED:

USE pressurizer heaters.

USE pressurizer sprays.

NOTES:

Charging valves FCV-62-85, -86, -89, and -93 fail OPEN on loss of air.

With normal and excess letdown FCVs failed CLOSED, pressurizer level will continue to rise. Reducing seal injection flow slows rate of level rise.

18. CHECK normal letdown IN SERVICE. PERFORM the following:

a. ENSURE charging isolation valves CLOSED:

FCV-62-90

FCV-62-91

b. IF CCP NOT running, THEN START CCP USING 1 (2)-SO-62-1, Chemical and Volume Control System.

c. DISPATCH operator to control RCP seal injection flow, USING appendix H, Local Control of Seal Injection Flow.

Page 16 of 64

PLANT SYSTEMS AUXILIARY FEEDWATER (AFW) SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 Three auxiliary feedwater trains shall be OPERABLE.

APPLICABILITY: MODES 1 2, and 3, MODE 4 when steam generator is relied upon for heat removal.

ACTION:

a. With one AFW train inoperable in MODE 1, 2, or 3, restore the inoperable AFW train to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With two AFW trains inoperable in MODE 1, 2, or 3, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With three AFW trains inoperable in MODE 1, 2, or 3, immediately initiate corrective action to restore at least one AFW train to OPERABLE status.

d. With the required AFW train inoperable in MODE 4, immediately initiate action to restore the required AFW train to OPERABLE status.

e. LCO 3.0.4.b is not applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.2.1 At least once per 31 days, verify each AFW manual, power operated, and automatic valve in each water flow path, and in both steam supply flow paths to the steam turbine driven pump, that is not locked, sealed, or otherwise secured in position, is in the correct position.

Only one AFW train, which includes a motor driven pump, is required to be OPERABLE in MODE 4.

LCO 3.0.3 and all other LCO ACTIONS requiring MODE changes are suspended until one AFW train is restored to OPERABLE status.

April 11,2005 SEQUOYAH - UNIT 1 3/4 7-5 Amendment No. 12, 115, 206, 301

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.7.1.2.2 At least once per 92 days, verify the developed head of each AFW pump at the flow test point is greater than or equal to the required developed head.

4.7.1.2.3 Once every 18 months, verify each AFW automatic valve that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

4.7.1.2.4 Once yery 18 months, verify each AFW pump starts automatically on an actual or simulated actuation signal.

Not required to be completed for the turbine driven AFW pump until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after steam supply pressure is greater than or equal to 842 psig.

Not applicable in Mode 4 when steam generators are relied upon for heat removal.

Not required to be completed for the turbine driven AFW pump until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after steam supply pressure is greater than or equal to 842 psig. Not applicable in MODE 4 when steam generator(s) are relied upon for heat removal.

August 2, 1995 SEQUOYAH UNIT 1- 3/4 7-6 Amendment No. 12, 77, 114, 206

PLANT SYSTEMS BASES Q = Nominal NSSS power rating of the plant (including reactor coolant pump heat), Mwt K = Conversion factor, 947.82 (Btu/sec)

Mwt W = Minimum total steam flow rate capability of the operable MSSVs on any one steam generator at the highest MSSV opening pressure including tolerance and accumulation, as appropriate, in lb/sec. For example, if the maximum number of inoperable MSSVs on any one steam generator is one, then W should be a summation of the capacity of the operable MSSVs at the highest operable MSSV r.operating pressure, excluding the highest capacity MSSV. If the maximum number of inoperable MSSVs per steam generator is three then W should be a summation of the capacity of the operable MSSVs at the highest operable MSSV operating pressure, excluding the three highest capacity MSSVs.

hfg = heat of vaporization for steam at the highest MSSV opening pressure including tolerance and accumulation, as appropriate, Btu,lbm N = Number of loops in plant The values calculated from this algorithm must then be adjusted lower to account for instrument and channel uncertainties.

3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The AFW System is configured into three trains. The AFW System is considered OPERABLE when the components and flow paths required to provide redundant AFW flow to the steam generators are OPERABLE. This requires that the two motor-driven AFW pumps be OPERABLE in two diverse paths, each supplying AFW to separate steam generators. The turbine-driven AFW pump is required to be OPERABLE with redundant steam supplies from each of two main steam lines upstream of the MSIVs, and shall be capable of supplying AFW to any steam generator. The piping, valves, instrumentation, and controls in the required flow paths also are required to be OPERABLE.

The AFW System mitigates the consequences of any event with loss of normal feedwater.

The design basis of the AFW System is to supply water to the steam generator to remove decay heat and other residual heat by delivering at least the minimum required flow rate to the steam generators while pumping against the highest credible steam generator pressure.

In addition, the AFW System must supply enough makeup water to replace steam generator secondary inventory lost as the unit cools to MODE 4 conditions. Sufficient AFW flow must also be available to account for flow losses such as pump recirculation and line breaks.

August 14, 2001 SEQUOYAH - UNIT 1 B 3/4 7-2 Amendment No. 115, 155, 196, 206

PLANT SYSTEMS BASES The limiting Design Basis Accidents (DBAs) and transients for the AFW System are as follows:

a. Feedwater Line Break (FWLB); and

b. Loss of main feedwater (MFW).

In addition, the minimum available AFW flow and system characteristics are credited for removing decay heat in the analysis of a small break loss of coolant accident (LOCA).

The AFW System design is such that it can perform its function following a FWLB between the MEW isolation valves and containment, combined with a loss of offsite power following turbine trip, and a single active failure of the steam turbine-driven AFW pump (above 50% power) or one motor-driven AFW pump (below 50% power with steam generator low level reactor trip time delay). For 50% power operation and higher, one motor-driven AFW pump is assumed to deliver to the broken MFW header at the pump run-out flow. Sufficient flow would be delivered to the intact steam generator by the redundant motor-driven AFW pump.

For partial power operation (below 50% power with trip time delay active), one motor-driven AFW pump is assumed to fail. All flow from the turbine-driven AFW pump and the redundant motor-driven AFW pump is assumed to deliver to the broken MEW header until the faulted steam generator is isolated. After isolation of the faulted steam generator, sufficient flow is delivered to the intact steam generator by the turbine-driven and redundant motor-driven AFW pump.

The Engineered Safety Feature Actuation System (ESFAS) automatically actuates the AFW turbine-driven pump and associated valves and controls when required to ensure an adequate feedwater supply to the steam generators during loss of power.

The surveillance requirements (SRs) provide a means of ensuring the AFW system components are capable of supplying required flow to the steam generators, the flow path is aligned correctly, and the automatic functions actuate as designed. The automatic functions are verified through either an actual or simulated actuation signal. The actuation signal associated with SR 4.7.1.2.3 (automatic valve actuation) include the AFW actuation test signal and the low AFW pump suction pressure test signal. The actuation signal associated with SR 4.7.1.2.4 (automatic pump start) includes only the AFW actuation test signal.

Each motor-driven auxiliary feedwater pump (one Train A and one Train B) supplies flow paths to two steam generators. Each flow path contains an automatic air-operated level control valve (LCV). The LCVs have the same train designation as the associated pump and are provided trained air. The turbine driven auxiliary feedwater pump supplies flow paths to all four steam generators. Each of these flow paths contains an automatic opening (non-modulating) air-operated LCV, two of November 17, 1995 SEQUOYAH UNIT 1 - B3/4 7-2a Amendment No. 115, 155, 196, 206

PLANT SYSTEMS BASES which are designated as Train A, receive A-train air, and provide flow to the same steam generators that are supplied by the B-train motor-driven auxiliary feedwater pump. The remaining two LCVs are designated as Train B, receive B-train air, and provide flow to the same steam generators that are supplied by the A-train motor-driven pump. This design provides the required redundancy to ensure that at least two steam generators receive the necessary flow assuming any single failure. It can be seen from the description provided above that the loss of a single train of air (A or B) will not prevent the auxiliary feedwater system from performing its intended safety function and is no more severe than the loss of a single auxiliary feedwater pump. Therefore, the loss of a single train of auxiliary air only affects the capability of a single motor-driven auxiliary feedwater pump because the turbine-driven pump is still capable of providing flow to two steam generators that are separate from the other motor-driven pump.

Two redundant steam sources are required to be operable to ensure that at least one source is available for the steam-driven auxiliary feedwater (AFW) pump operation following a feedwater or main steam line break. This requirement ensures that the plant remains within its design basis (i.e., AFW to two intact steam generators) given the event of a loss of the No. 1 steam generator because of a main steam line or feedwater line break and a single failure of the B-train motor driven AFW pump. The two redundant sources must be aligned such that No. I steam generator source is open and operable and the No. 4 steam generator source is closed and operable.

For instances where one train of emergency raw cooling water (ERCW) is declared inoperable in accordance with technical specifications, the AFW turbine-driven pump is considered operable since it is supplied by both trains of ERCW. Similarly, the AFW turbine-driven pump is considered operable when one train of the AFW loss of power start function is declared inoperable in accordance with Technical Specifications because both 6.9 kilovolt shutdown board logic trains supply this function. This position is consistent with American National Standards Institute/ANS 58.9 requirements (i.e., postulation of the failure of the opposite train is not required while relying on the TS limiting condition for operation).

Action e prohibits the application of LCD 3.0.4.b to an inoperable AFW train. There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an AFW train inoperable and the provisions of LCD 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCD not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.

3/4.7.1.3 CONDENSATE STORAGE TANK The CST level required is equivalent to a usable volume of at least 240,000 gallons, which is based on holding the unit in MODE 3 for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , followed by a cooldown to RI-fR entry conditions within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

April 11,2005 SEQUOYAH UNIT 1- B 3/4 7-2b Amendment No. 115, 155, 182, 188, 196, 207, 286, 301

OPL27IAOP-M.02 Revision 1 Page 3of24 I. PROGRAM: OPERATOR TRAINING - LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: AOP-M.02 LOSS OF CONTROL AIR IV. LENGTH OF LESSONICOURSE: 1.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months (s)

V. TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of AOP-M.02, LOSS OF CONTROL AIR.

B. Enabling Objectives Obiectives

0. Demonstrate an understanding of NUREG 1122 knowledges and abilities associated with Loss of Control Air events that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification Training for the appropriate position as identified in Appendix A

1. State the purpose/goal of AOP-M.02.

2. Describe the AOP-M.02 entry conditions.

a. Describe the setpoints, interlocks, and automatic actions associated with AOP-M.02 entry conditions.

b. Describe the ARP requirements associated with AOP-M.02 entry conditions.

c. Interpret, prioritize, and verify associated alarms are consistent with AOP M.02 entry conditions.

d. Describe the Administrative and Tech Spec conditions resulting from a Loss of Control Air.

3. Describe the initial operator response to stabilize the plant upon entry into AOP M.02.

4. Upon entry into AOP-M.02, diagnose the applicable condition and transition to the appropriate procedural section for response.

5. Summarize the mitigating strategy for the condition that initiated entry into AOP M.02.

6. Describe the bases for all limits, notes, cautions, and steps of AOP-M.02.

7. Describe the conditions and reason for transitions within this procedure and transitions to other procedures.

OPL271AOP-M.02 Revision 1 Page 4 of 24 Objectives

8. Given a set of initial plant conditions use AOP-M.02 to correctly:

a. Recognize entry conditions

b. Identify required actions

c. Respond to Contingencies

d. Observe and Interpret Cautions and Notes

9. Describe the Tech Spec and TRM actions applicable during the performance of AOP-M.02.

10. Apply GEE and system response concepts to the abnormal condition prior to, during and after the abnormal condition.

OPL271AOP-M.02 Revision 0 Page 3 of 25 I. PROGRAM: OPERATOR TRAINING LICENSED -

II. COURSE: LICENSE TRAINING III. LESSON TITLE: AOP-M.02 LOSS OF CONTROL AIR IV. LENGTH OF LESSONICOURSE: 1.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months (s)

V. TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of AOP-M.02, LOSS OF CONTROL AIR.

B. Enabling Objectives Obiectives

0. Demonstrate an understanding of NUREG 1122 knowledges and abilities associated with Loss of Control Air events that are rated r 2.5 during Initial License Training and T 3.0 during License Operator Requalification Training for the appropriate position as identified in Appendix A

1. State the purpose/goal of AOP-M.02.

2. Describe the AOP-M.02 entry conditions.

a. Describe the setpoints, interlocks, and automatic actions associated with AOP-M.02 entry conditions.

h. Describe the ARP requirements associated with AOP-M.02 entry conditions.

c. Interpret, prioritize, and verify associated alarms are consistent with AOP M.02 entry conditions.

d. Describe the Administrative and Tech Spec conditions resulting from a Loss of Control Air.

3. Describe the initial operator response to stabilize the plant upon entry into AOP M.02.

4. Upon entry into AOP-M.02, diagnose the applicable condition and transition to the appropriate procedural section for response.

r

5. Summarize the mitigating strategy for the condition that initiated entry into AOP M.02.

6. Describe the bases for amits, notes, cautions, and steps of AOP-M.02.

0PL271 AOP-M.02 Revision 0 Page 4 of 25

7. Describe the conditions and reason for transitions within this procedure and transitions to other procedures.

8. Given a set of initial plant conditions use AOP-M.02 to correctly:

a. Recognize entry conditions

b. Identify required actions

c. Respond to Contingencies

d. Observe and Interpret Cautions and Notes

9. Describe the Tech Spec and TRM actions applicable during the performance of AOPM.O2.

10. ApyGFE and system response concepts to the abnormal condition prior to, during and after the abnormal condition.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross- Level RD SRO reference: x Tier# 2 Group# 1 K/A# 078A2.O1 Importance Rating 2.9 5 tne SSTEM and (b) based on those predcUons, use

? r:quecos of thcse abnorma! operabon:

Proposed Question: SRO 97 1 Pt Given the following:

The VI system on Unit I has become heavily contaminated with oil due to a maintenance problem

The VI Air Dryer packages rapidly clog VVhich ONE (1) of the following describes the impact on VI system operation AND the actions directed by AP-22 (Loss if VI) to mitigate the consequences of the malfunction?

VI header pressure will decrease until...

A. 1 Vl-1 812 (VI Air Dryer Bypass Filter Isol) automatically opens at 85 PSIG. If the valve fails to open, Enclosure 5 (VI Dryer and VI to VS System Isolation) directs the operators to bypass the air dryers locally at 82 PSIG.

B. 1VI-1812 (VI Air Dryer Bypass Filter Isol) automatically opens at 90 PSIG. If the valve fails to open, Enclosure 5 (VI Dryer and VI to VS System Isolation) directs the operators to bypass the air dryers locally at 85 PSIG.

C. lVl-820 (VI to VS Supply) automatically closes at 85 PSIG. As a backup to Thie automatic action, AP-22 directs the operators to manually close 1VI-820 at 82 PSIG.

D. 1VI-820 (VIto VS Supply) automatically closes at 90 PSIG. As a Page 307 of 320 Rev Final

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet backup to the automatic action, AP-22 directs the operators to manually close lVl-820 at 85 PSIG.

Proposed Answer: A Exp)anaton (Optional):

With the air dryers clogged, reducing the air load on the system by Vl-820 closing will reduce the rate of the VI header pressure decrease. However, since Vl-820 is downstream of the air dryers the only automatic action which may be successful in restoring VI header pressure is bypassing the air dryers. With the manual bypass flowpath isolated (1VI-93 closed), the automatic bypass must open to restore header pressure.

lVl-1812 automatically opens when VI header pressure decreases to 85 psig. AP 22 directs the Operators to bypass the air dryers locally using Enclosure 5 if VI header pressure decreases to less than 82 psig. Enclosure 5 has the Operator check that I VI-i 812 is open and has the operator open the manual bypasses (should normally be open) 1\fl-93 and 1VI-94 (in series).

A. (;orrect.

B. orrict: See explanation above. IVI-1812 automatically opens at 85 psig 2 rd the procedure actions are directed when VI header pressure decreases to 82 psig. Plausible because there are automatic actions which occur when V header pressure decreases to 90 psig (1VI-820 closes) and IVI-1 812 aLtomaticaiy opens at 85 psig.

C. Icorrect: See explanation above. Plausible because 1VI-820 will slow the o pressure decrease and there are automatic actions which occur at 85 ps (iVL8i2 opens).

D. hcorrect: See explanation above.lVl-820 closing will not restore VI header p sssu;e. Plausible because lVl-820 will slow the rate of pressure rcase and 1VI-820 does automatically close when VI header pressure cacreases to 90 psig.

Technical Reference(s) AP/1/A/5500/22, LOSS OF (Attach if not previously VLrev. 28 page 3 provided)

Lesson Plan OP-MC-SS-Vl, (Including version or revision Instrument Air, Station Air, #)

Page 308 of 320 Rev Final

ES-401 Same Written Examination Form ES-401-5 Question Worksheet Breathing Air rev. 32 page 193 (Figure 7.27- VI System Composite)

Proposed references to be provided to applicants during None examinetion:

Learnin: Objeotive: OP-MC-AP-22, Obj 2and 4 (As available)

Questlo Sojrce: Bank#

Modfled Bank # (Note changes or attach parent)

New x Questlor History: Last NRC Exam Questio; Ccqniiive Memory or Fundamental Knowledge Level:

Comprehension or Analysis x 10 CFR Part 55 55.41 Content 55.43 43.5 Page 309 of 320 Rev Final

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

i 1h cIo*m o tie SIEM) and (b) based on those predictions, use 0: e o hoae abnorma ooeration:

KA is matched because the candidate must understand the operation of the system to diagnose which automatic action will be successful in restoring VI header gressure and must also be familiar with the procedure mitigating strategy to determine which procedure actions are appropriate as directed by AP-22.

This qu.sHon is comprehension level because the candidate must understand which o i:wo possible automatic actions that could occur would be effective at allow Vi h&ader pressure to be restore.

This is SRO Dny question linked to IOCFR55.43(b)(5), Procedures. This question NOT be answered with system knowledge alone. It can NOT be answerel i knowing procedure immediate actions or entry conditions. To correctly answer the question the candidate must have knowledge of specific procednrl steps.

Page 310 of 320 Rev Final

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta

90. Given the following:

- Unit 1 is at 100% power.

- A LOCA occurred inside containment.

- The crew has just implemented E-1, Loss of Reactor or Secondary Coolant.

- The STA has completed the initial performance of the status trees and reports the highest priority path exists on the CONTAINMENT status tree.

- Containment conditions are as follows:

- Pressure is 2.6 psig and lowering.

- Upper containment Rad Monitors read 85 RIhr.

- Lower containment Rad Monitors read 125 RJhr.

- Containment Sump Level is 58%.

Based on the above conditions, the Unit Supervisor A. is required to IMMEDIATELY implement and complete FR-Z.2, Containment Flooding, then transition back to E-1.

B. will acknowledge entry criteria for FR-Z.2, Containment Flooding, is met but entry into the FR is optional.

C. is required to IMMEDIATELY implement and complete FR-Z.3, High Containment Radiation, then transition back to E-1.

D. will acknowledge entry criteria for FR-Z.3, High Containment Radiation, is met but entry into the FR is optional.

DIS TRACTOR ANALYSIS:

A. Incorrect, Plausible because the containment sump level is elevated and if the required level was present an Orange would be present and require immediate transition to FR-Z. 2. Incorrect because the level of the sump is below the 68%

level required to enter FR-Z.2, Containment Flooding.

B. Incorrect, Plausible because the containment sump level is elevated and if the setpoint was exceeded a yellow path would exist and entry would be optional.

C. Incorrect, Plausible because radiation in lower containment is greater than the threshold level for entering FR-Z.3. Incorrect because the condition is a yellow path and entry is optional.

D. CORRECT, the radiation in lower containment is greater than the threshold level for entering FR-Z.3, High Containment Radiation, and the challenge is a yellow path which allows the performance of the procedure to be optional.

Wednesday, July 07, 2010 1:56:24 PM 90

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta Question Number: 90 Tier: 2 Group 1 KIA: 103 Containment System G 2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

Importance Rating: 4.0 / 4.6 IOCFRPart55: 41.7/43.5/45.12 IOCFR55.43.b: 5 KIA Match: Question matches K/A by having the candidate assess given plant data and then determine the correct procedure actions needed to mitigate the accident.

Technical

Reference:

EPM-4, Users Guide, Rev. 20 1-FR-0, Status trees, Rev. 1 Proposed references None to be provided:

Learning Objective: OPL271EPM-4 B.11 Given a set of conditions, analyze the EOP/FRP implementation.

0PL271 FR-0 B.6 Given a set of initial plant conditions use FR-0 to correctly identify the: a. Identify required actions.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: SQN bank question W/E16 EA2.1 #85 Comments: Modified from Braidwood 2007 NRC SRO exam #85 Wednesday, July 07, 2010 1:56:24 PM 91

CONTAINMENT SQN F-O .5 1-FR-O Rev I CONTAINMENT __h_wLw1KII:E:I PRESSURE LESS THAN 12.0 PSIG YES

, GOTO CONTAINMENT NO - FRZI PRESSURE LESS THAN 2.8 PSIG YES 4

NO r GOTO CONTAINMENT FR-Z.2 SUMP LEVEL LESS THAN 68%

YES Jr GOTO UPPERAND LOWER NO Y FR-Z.3 CONTAINMENT RADIATION MONITORS LESS THAN 100 RIHR YES CSF SAT Page 10 of 16

SQN EOI EPM-4 PROGRAM USERS GUIDE Rev. 20 MANUAL Page 48 of 97 3.10.5 Status Tree Rules of Usage (continued)

6. If any ORANGE challenge is encountered, the person monitoring status trees continues monitoring until all six status trees have been evaluated.

This is necessary because a subsequent RED challenge has priority over any ORANGE challenge. If any RED is encountered, then Rule 3.1O.5.D.4 applies. Otherwise, once it is determined that no RED challenges exist, then the person monitoring status trees informs the procedure reader of the highest priority ORANGE challenge.

7. RED or ORANGE challenges must be addressed immediately by implementing appropriate FRPs in order of priority and per the rules of usage. When the person monitoring status trees informs the procedure reader that a RED or ORANGE challenge exists, the procedure reader immediately suspends the ORP (or lower priority FRP) in progress and implements the appropriate FRP, as indicated at the terminus point of the CSF under challenge.

8. YELLOW challenges may be addressed by implementing appropriate FRPs if desired, but do not require immediate operator action.

Addressing YELLOW challenges is optional since these are usually temporary, off-normal conditions that will be restored to normal status by actions already in progress. In other cases, the YELLOW path might provide an early indication of a developing RED or ORANGE condition.

Following FRP implementation, a YELLOW might indicate a residual off-normal condition. When the person monitoring status trees informs the procedure reader that a YELLOW challenge exists, the procedure reader should evaluate if the YELLOW challenge FRP should be implemented.

This decision will be based on the following:

Whether the procedures in effect will address the challenge as a matter of course.

Whether the procedures in effect are more important at that time based upon available time and current plant conditions.

Whether the challenge is of a nature that it will likely develop into an ORANGE or RED condition if action is not taken early.

0PL271 EPM-.4 Revision I Page 3 of 28 PROGRAM: OPERATOR TRAINING - LICENSED COURSE: LICENSE TRAINING II. LESSON TITLE:

III. LENGTH OF LESSON!COURSE: 4-6 hour(s)

W. TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of HLC Procedures training, the participant shall be able to explain, using classroom evaluations and/or simulator scenarios, the requirements of EMP-4, EOP-E-O, Users Guide.

B. Enabling Objectives:

1. Determine/identify the correct procedural application(s) based on the operating procedures network for normal, abnormal, and emergency evolutions.

2. Analyze an EOP layout and determine (according to EPM-4):

a. correct procedural layout application;

b. if the use of terms is correct (e.g.: Faulted Steam Generator, Shall, Lowering, etc per Appx. B);

c. correct use of symbols and icons.

3. Define EOP warnings, cautions, and notes and, given an EOP condition, determine appropriate usage.

4. Compare and contrast event-based emergency/abnormal operating procedures used in parallel with the symptom-based EOPs.

5. Given an example, apply general guidelines, crew roles and responsibilities for EOP procedural use and determine:

a. format and use of sequenced and non-sequenced sub steps;

b. transition between Action/Expected Response column and the Response Not Obtained column;

c. requirements for task completion prior to proceeding to the next action (and how any exceptions are identified);

d. requirements for task completion still in progress following transition to another procedure or step;

e. actions based on fold-out page use;

f. actions based on hand-out page use;

g. if EOP termination is appropriate based on given conditions.

6. Identify post-accident instrumentation and determine if its use is required.

7. Given plant operating conditions, determine if EOP entry conditions have been met and state the resultant appropriate immediate action steps for those conditions.

0PL271 EPM-4 Revision 1 Page 4 of 28

8. Given plant operating conditions, determine if AOP entry conditions have been met and state the resultant appropriate actions for those conditions.

9. Identify general operating crew responsibilities during emergency operations including appropriate implementation of prudent operator actions.

10. Identify general operating crew responsibilities during emergency operations including requirements for actions outside Technical Specifications/plant licensed conditions (10CFR5O.54x application).

11. Given a set of conditions, analyze the EOPIFRP implementation:

a. identify the basis for the implementation;

b. determine the correct implementation hierarchy;

c. determine if Critical Safety Function Status Trees (CFSTs) implementation is required;

d. identify the status tree colors by priority and summarize each trees purpose;

e. identify conditions which will allow a FRP to be exited once it is entered (a RED or ORANGE condition);

f. state the monitoring frequency of CFSTs and when this can be relaxed;

g. determine correct coordination with other support procedures

h. identify conditions permissible to terminate CFSTs monitoring.

12. Given an operational situation, analyze a crew brief and determine if it meets Management expectations.

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta

91. Given the following:

- Unit I tripped from 100% due to a small break LOCA.

- While implementing E-1, Loss of Reactor or Secondary Coolant, the following data is reported by the crew:

- RCS pressure 1200 psig slowly lowering.

- CETCs are 500°F and slowly lowering.

- CNMT pressure 2.0 psig.

- CNMT temperature 215°F.

- CNMT hydrogen concentration has just been confirmed at 4.2%.

- The crew is implementing EA-268-1, Placing Hydrogen Recombiners In Service.

Which ONE of the following identifies:

(1) the parameter used, in accordance with EA-268-1, to adjust the hydrogen recombiner power setting?

and (2) the REP classification that would be required for the conditions described?

Reference Provided A. (1) Containment humidity (2) Site Area Emergency B. (1) Containment humidity (2) Alert C. (1) Containment pressure (2) Alert D (1) Containment pressure (2) Site Area Emergency Thursday, July 15, 2010 3:22:08 PM 91

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta DISTRACTOR ANALYSIS:

A. Incorrect, Plausible if the candidate confuses CNMT humidity vs pressure for determining the correction factor to adjust the heater controls. Also a Site Area Emergency is correct due to potential loss of 2 barriers.

B. Incorrect, Plausible if the candidate confuses CNMT humidity vs pressure for determining the correction factor to adjust the heater controls. Alert would be correct if candidate only recognizes the RCS barrier potential loss.

C. Incorrect, Plausible since the use of containment pressure is the parameter utilized to modify the reference power setting and Alert would be the correct classification if candidate only recognizes the RCS barrier potential loss. Incorrect because SAE is the correct classification due to potential loss of containment and the RCS.

D. Correct, As per guidance in EA-268-1 the correct power setting is calculated by multiplying the CNMT pressure factor by the Reference Factor to obtain the correct power setting. EPIP-1 requires that a Site Area Emergency be declared for a Loss or Potential loss of any two barriers (RCS potential loss due to entering E- 1 and CNMT potential loss due to hydrogen > 4%.)

Thursday, July 15, 2010 3:22:08 PM 92

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta Question Number: 91 Tier: 2 Group 2 KIA: 028 Hydrogen Recombiner and Purge Control A2.01 Ability to predict and/or monitor malfunctions or operations on the HRPS; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations:

Hydrogen recombiner power setting, determined by using plant data book.

Importance Rating: 3*4* / 3.6*

10 CFR Part 55: 41.5 IOCFR55.43.b: 5 KIA Match: Question matches K/A by having the candidate determine how the power setting for the Hydrogen recombiners are determined and SRO level by having candidate which emergency event classfication applies to data presented.

Technical

Reference:

EA-268-1 rev 4 EPIP-1 rev 44 Proposed references EPIP-1 pages 9, 10 to be provided:

Learning Objective: OPT200.CGCS Obj 5b; OPL271REP Obj 3 Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question written for 1009 NRC exam Comments:

MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9 Answer: DADDCCACBA ScrambleRange:A-D Thursday, July 15, 2010 3:22:08 PM 93

SEQUOYAH EMERGENCY PLAN CLASSIFICATION MATRIX EPIP-1 11 Fuel Clad Barrier 1.2 RCS Barrier I 1 CriticI Sftv Function Status 1. Critical Safety Function Status LOSS Potential LOSS Potential LOSS Core Cooling Red Core Cooling Orange Not Applicable Pressurized Thermal (FR-C.1) (FR-C.2) Shock Red (FR-P.1)

OR OR Heat Sink RED (FR-H.1) and RHR Shutdown Heat Sink RED (FR-Hi)

Cooling not in service and RHR Shutdown

- OR - Cooling not in service 2 Prima p Coolant Activity Level - OR -

LOSS Potential LOSS 2 RCS Lik I LO(A RCS sample activity is Not Applicable LOSS Potential LOSS greater than 300 uCi/gm RCS leak results in Non Isolatable RCS leak dose equivalent 1131 subcooling <40 °F as exceeding the capacity

-OR indicated on Xl-94-101 of one charging pump in I Incore Thermocouple Hi Quad Average or 102 (EXOSENSOR) the normal charging LOSS Potential LOSS alignment Greater than 1200°F on Greater than or equal to Xl-94-101 or 102 700 °F on Xl-94-101 or OR (EXOSENSOR) 102 (EXOSENSOR)

- OR -

RCS leakage results in

4. Reactor Vessel Water Level entry into E-1 LOSS Potential_LOSS

- OR -

Not Applicable VALID RVLIS level I Steam PnPr2tnr Tiib Rupture

<42% on Ll-68-368 or Ll-68-371 with no RCP LOSS Potential_LOSS running SGTR that results in a Not Applicable

- OR -

Safety Injection 5 Continmnt Radiation Monitor actuation LOSS Potential LOSS OR VALID reading of Not Applicable greater than:

Entry into E-3 2.8E+01 Rem/hr on

- OR -

RM-90-271A or -272A 4.

OR LOSS Potential LOSS VALID I-<VL! level Not Applicable 2.9E+01 Rem/hr on <42% on Ll-68-368 or RM-90-273A or -274A Ll-68-371 with no RCP (see instruction note 4) running

-OR - OR -

6. SED Judgment 5. SED Judgment*

Any condition that, in the judgment of the SM or Any condition that, in the judgment of the SM or SED, indicates loss or potential loss of the Fuel SED, indicates loss or potential loss of the RCS Clad Barrier comparable to the conditions listed Barrier comparable to the conditions listed above. above.

Page 9 of 47 Revision 44

SEQUOYAH EMERGENCY PLAN CLASSIFICATION MATRIX EPIP-1 1.3 Containment Barrier INSTRUCTIONS I 1. Critical Safety Function Status Note: A condition is considered to be LOSS Potential LOSS MET if, in the judgment of the SED, the Not Applicable Containment Red condition will be MET IMMINENTLY (i.e.:

(FR-Z.1) with two hours). The classification shall OR be made as soon as this determination is Actions of FR-C.1 (Red Path) made.

are INEFFECTIVE (i.e.: core 1. In the matrix to the left, REVIEW the TCs trending up) initiating conditions in all three barrier

- OR -

columns and circle the conditions that

2. Containment Pressure I are MET.

LOSS Potential_LOSS 2. In each of the three barrier columns, Rapid unexplained pressure Containment Hydrogen IDENTIFY if any Loss or Potential Loss decrease following initial increases to >4% by volume on INITIATING CONDITIONS have been increase on PDI-30-44 or45 H21-43-200 or 210 MET.

Q 3. COMPARE the number of barrier Containment pressure or sump Pressure >2.8 PSIG (Phase B) level not increasing on Ll with < one full train of Losses and Potential losses to the 178 and 179 with a LOCA in containment spray criteria below and make the progress appropriate declaration.

-OR- 4. Containment Radiation Monitors are

3. Containment Isolation Status temperature sensitive and can be V LOSS Potential_LOSS affected by temperature-induced Containment isolation, when Not Applicable currents. These monitors should be required is incomplete and a used for trending only until containment release path to the environment temperature has been stable for exists. approximately 5 minutes after a Steam

- OR - Line Break or LOCA.

4. Containment Bypass

.]* Potential LOSS Note: MONITOR the respective status tree RUPTURED S!G that is also Unexpected VALID increase in criteria if a CSF is listed as an faulted outside containment (E2 area or ventilation RAD INITIATING CONDITION.

and E3) monitors adjacent to OR containment (with LOCA in Emerciencv Class Criteria

>4 hour secondary side release progress).

outside containment from a S/G General Emergency with a S/G tube leak >T/S limits (AOP R.01 App A) LOSS of any two barriers Potential

-OR- LOSS of third barrier

5. Significant Radiation in Containment r.}* Potential LOSS Site Area Emergency Not Applicable VALID reading of greater than:

LOSS or Potential LOSS of any two 3.6E+02 Rem/hr on RM-90-271A and 272 barriers OR 2.8E+02 Rem/hr on Alert RM-90-273A and 274 (see instruction note 4) Any LOSS or Potential LOSS of Fuel Clad

-OR barrier OR

6. SED Judgment Any condition that, in the judgment of the SM or SED, indicates Any LOSS or Potential LOSS of RCS loss or potential loss of the Containment Barrier comparable to barrier the conditions listed above.

Unusual Event LOSS or Potential LOSS of Containment barrier Page 10 of 47 Revision 44

SQN EA-268-1 PLACING HYDROGEN RECOMBINERS IN SERVICE Rev. 4 1,2 Page4ofl2 4.2 Placing Hydrogen Recombiner in Service

1. SELECT applicable unit:

Uniti

Unit2

2. SELECT recombiner to be placed in service:

Train A

TrainB

3. RECORD containment pressure from one of the following instruments: [M-6]

INSTRUMENT PAM PRESSURE PDI-30-45 YES El PDI-30-44 YES El PDI-30-43 NO U PDI-30-42 NO U

4. IF LOSS OF OFFSITE POWER has occurred, THEN PERFORM the following:

a. IF 480V Reactor Vent Boards have NOT been energized, THEN ENSURE all breakers OPEN on 480V Reactor Vent Boards.

b. ENSURE 480V Reactor Vent Boards ENERGIZED USING EA-201-2, Restoring 480V Busses. E

c. ENSURE breakers for hydrogen recombiners CLOSED.

[480V Rx Vent Bd A-A and B-B, Compt 3B]

5. CHECK POWER AVAILABLE light LIT [M-1OJ.

SQN EA-268-1 PLACING HYDROGEN RECOMBINERS IN SERVICE Rev. 4 1,2 Page5ofl2 4.2 Pladng Hydrogen Recombiner in Service (Continued)

6. ENSURE POWER ADJUST potentiometer set at 000:

I TRAIN I POTENTIOMETER I POSITION I A XS-83-5003 000 E1 B XS-83-5004 000

7. PLACE POWER OUT SWITCH in up position (on) and CHECK red light on switch plate LIT.

8. DETERMINE Pressure Factor USING Appendix A, Ice Condenser Containments Recombiner Power Correction Factor vs. Containment Pressure, and RECORD below:

Pressure Factor

9. RECORD reference power from Hydrogen Recombiner Data Plate (Ref. Power): [M-10]

Reference Power KW.

10. CALCULATE required hydrogen recombiner power setting:

a. CALCULATE power setting in KW:

x KW.

Pressure Factor Reference Power Setting (4.2. 8.) (4.2. 9.)

b. RECORD above calculated power setting in the 25 Minute Table, KW Reading column in Step 11. LJ

LI C%1 I.

0

= = = - =

- = = =

>0) ci) z z z LU EC)C) C)C)

C.) (1)

> C: LU LU UI UI UI Co o Z D D D D C I ci) U?Lc) Lf)LC) QLOLO LULDLO W

LU E LUW cc LULU cc WLU c)c)

Z c I I coco coco coco J I I I I I I 8

o C)

H D

O I-LU (/)0 LU LU D LU D LU D LU ILU w D N HC)J- Z F-Co),t Z F-CD Z F-C).t (1) LU LU C) LU C) C) LU z

LU C)C) C)C) C)C) C)C) w 0

E C

occ C) C) occ e) c)

, o C) )

oLcc C) C) a0 o

F-coco F-coco F-c/)co lcoco E Q F-XX _XX 1

0 0 0 0 UJ 8

0 Z Z Z Z ci)

C) ci)

L I.cl) H I I 0

x LUC)C) LUC)C) LUC)C) WC)C)

C C)C) C)C) QQ LOL()

oz I C) C) H - 1 (N (N I c.j c.i 12 D

= = = = = = =

(N z C*4 a

U)

C1

SQN EA-268-1 PLACING HYDROGEN RECOMBINERS IN SERVICE Rev. 4 1,2 Page7ofl2 4.2 Placing Hydrogen Recombiner in Service (Continued)

13. MONITOR and MAINTAIN KW reading as indicated on POWER OUT meter.

NOTE Leaving Temperature Channel Selector Switch in one position for extended period of time can allow film or residue to build up on the contacts, causing inaccurate temperature readings. Rotating Temperature Channel Selector Switch to the right and left three to five times should wipe contacts.

14. SELECT one of the three thermocouples USING TEMPERATURE CHANNEL SELECTOR switch and RECORD present reading:

TEMPERATURE TRAIN SWITCH READING INDICATOR [

A XS-83-5001 TI-83-5001 L B XS-83-5002 TI-83-5002 LI NOTE Hydrogen recombiner temperature should increase to between 1150°F and 1400°F within the first 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months of operation.

15. MONITOR hydrogen recombiner temperature as indicated on TEMPERATURE READOUT meter.

SQN EA-268-1 PLACING HYDROGEN RECOMBINERS IN SERVICE Rev. 4 1,2 Pagel2ofl2 APPENDIX A Page 1 of 1 ICE CONDENSER CONTAINMENTS RECOMBINER POWER CORRECTION FACTOR VS. CONTAINMENT PRESSURE Curves applicable for Pre-LOCA Containment pre-LOCA Containment temperature 110°F in lower pressure of 15.0 psia compartment and 75°F in upper compartment 1.

1.

0 0

0 (5

LL ci) 1.

In I) ci) 0 1

1 14.7 16.7 18.7 20.7 22.7 24.7 (psia) 0 2 4 6 8 10 (psig)

Post-LOCA Containment Pressure

SEQUOYAH EMERGENCY PLAN CLASSIFICATION MATRIX EPIP-1 1.1 Fuel Clad Barrier 1.2 RCS Barrier I 1. Critical Function Status 1. Critical Safety Function Status LOSS Potential LOSS .I* Potential LOSS Core Cooling Red Core Cooling Orange Not Applicable Pressurized Thermal (FR-C.1) (FR-C.2) Shock Red (FR-P.1)

OR OR Heat Sink RED (FR-Ri) and RHR Shutdown Heat Sink RED (FR-H.1)

Coolina not in service and RHR Shutdown

-OR- Cooling not in service

2. Primary Coolant Activity Level - OR -

hi* Potential LOSS 2. RCS Leakaje I LOCA RCS sample activity is Not Applicable

Potential LOSS greater than 300 uCi/gm RCS leak results in Non Isolatable RCS leak dose equivalent 1131 subcooling <40 °F as exceeding the capacity

-OR indicated on Xl-9410i of one charging pump in 1 3. Incore Thermocounle Hi Quad Average or 102 (EXOSENSOR) the normal charging LOSS - Potential_LOSS alignment Greater than 1200 °F on Greater than or equal to XI-94-i01 or 102 700 °F on Xl-94-i01 or OR (EXOSENSOR) 102 (EXOSENSOR)

-OR- RCS leakage results in

4. Reactor Vessel Water Level entry into E-1

.}* Potential LOSS

- OR -

Not Applicable VALID RVLIS level

<42% on LI-68-368 or F 3. Steam Generator Tube Rupture Ll-68-371 with no RCP LOSS Potential_LOSS running SGTR that results in a Not Applicable

- OR -

Safety Injection

5. Containment Radiation Monitor actuation LOSS Potential_LOSS OR VALID reading of Not Applicable greater than:

Entry into E-3 2.8E+01 Rem/hr on

- OR -

RM-90-271A or -272A OR LOSS Potential_LOSS VALID RVLIS level Not Applicable 2.9E+01 Rem/hr on <42% on Ll-68-368 or RM-90-273A or -274A Ll-68-371 with no RCP (see instruction note 4) running

-OR -OR

6. SED Judgment 5. SED Judgment Any condition that, in the judgment of the SM or Any condition that, in the judgment of the SM or SED, indicates loss or potential loss of the Fuel SED, indicates loss or potential loss of the RCS Clad Barrier comparable to the conditions listed Barrier comparable to the conditions listed above. above.

Page 9 of 47 Revision 44

SEQUOYAH EMERGENCY PLAN CLASSIFICATION MATRIX EPIP-1 1.3 Containment Barrier INSTRUCTIONS

1. Critical Safety Function Status Note: A condition is considered to be Potential LOSS MET if, in the judgment of the SED, the Not Applicable Containment Red condition will be MET IMMINENTLY (i.e.:

(FR-Z.1) with two hours). The classification shall OR be made as soon as this determination is Actions of FR-C.1 (Red Path) made.

are INEFFECTIVE (i.e.: core 1. In the matrix to the left, REVIEW the TCs trending up) initiating conditions in all three barrier

-OR- columns and circle the conditions that

2. Containment Pressure I H drogen are MET.

M Potential LOSS 2. In each of the three barrier columns, Rapid unexplained pressure Containment Hydrogen IDENTIFY if any Loss or Potential Loss decrease following initial increases to >4% by volume on INITIATING CONDITIONS have been increase on PDI-30-44 or 45 H2l-43-200 or 210 MET.

Q 3. COMPARE the number of barrier Containment pressure or sump Pressure >2.8 PSIG (Phase B) level not increasing on Ll with < one full train of Losses and Potential losses to the 178 and 179 with a LOCA in containment spray criteria below and make the progress appropriate declaration.

- OR -

4. Containment Radiation Monitors are

3. Containment Isolation Status temperature sensitive and can be

1* Potential LOSS affected by temperature-induced Containment isolation, when Not Applicable currents. These monitors should be required is incomplete and a used for trending only until containment release path to the environment temperature has been stable for exists. approximately 5 minutes after a Steam

-OR- Line Break or LOCA.

4. Containment Bypass

I* Potential LOSS Note: MONITOR the respective status tree RUPTURED S/G that is also Unexpected VALID increase in criteria if a CSF is listed as an faulted outside containment (E2 area or ventilation RAD INITIATING CONDITION.

and E3) monitors adjacent to Q containment (with LOCA in Emergency Class Criteria

>4 hour secondary side release progress).

outside containment from a SIG General Emergency with a SIG tube leak >T/S limits (AOP R.01 App A) LOSS of any two barriers Potential

-OR- LOSS of third barrier

5. Significant Radiation in Containment

.}* Potential LOSS Site Area Emergency Not Applicable VALID reading of greater than:

LOSS or Potential LOSS of any two 3.6E+02 Rem/hr on RM-90-271A and 272 barriers OR 2.8E+02 Rem/hr on Alert RM-90-273A and 274 (see instruction note 4) Any LOSS or Potential LOSS of Fuel Clad

-OR- barrier

6. SED Judgment OR Any condition that, in the judgment of the SM or SED, indicates Any LOSS or Potential LOSS of RCS loss or potential loss of the Containment Barrier comparable to barrier the conditions listed above.

Unusual Event LOSS or Potential LOSS of Containment barrier Page 10 of 47 Revision 44

OPT200.CGCS Rev. 3 Page 3 of 46 PROGRAM: OPERATOR TRAINING IL COURSE: SYSTEMS TRAINING III. TITLE: COMBUSTIBLE GAS CONTROL SYSTEM IV. LENGTH OF LESSON: 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months lecture; 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months simulator demonstration; 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months self-study/workshop V. TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of this lesson and others presented, the student should be able to apply the knowledge to support satisfactory performance of the tasks associated with the Combustible Gas Control system in the plant and on the simulator.

B. Learning Objectives:

0. Demonstrate an understanding of NUREG 1122 knowledges and abilities associated with the Combustible Gas Control system that are rated 2.5 during Initial License Training for the appropriate license position as identified in Appendix A.

I. State the purpose/functions of the Combustible Gas Control system as described in the FSAR.

2. State the design basis of the Combustible Gas Control system in accordance with the SQN FSAR.

3. Explain the purpose/function of each major component in the flow path of the Combustible Gas Control system as illustrated on a simplified system drawing.

4. Describe the following characteristics of each major component in the Combustible Gas Control system:

a. Location

b. Power supply (include control power as applicable)

c. Support equipment and systems

d. Normal operating parameters

e. Component operation

f. Controls

g. Interlocks (including setpoints)

h. Instrumentation and Indications

i. Protective features (including setpoints)

j. Failure modes

k. Unit differences

1. Types of accidents for which the Combustible Gas Control system components are designed

m. Location of controls and indications associated with the Combustible Gas Control system in the control room and auxiliary control room

OPT200.CGCS Rev. 3 Page 4 of 46 V. TRAINING OBJECTIVES (Contd):

B. Learning Objectives (Contd):

5. Describe the operation of the Combustible Gas Control system:

a. Precautions and limitations

b. Major steps performed while placing the Combustible Gas Control system in service

c. Alarms and alarm response

d. How a component failure will affect system operation

e. How a support system failure will affect Combustible Gas Control system operation

f. How a instrument failure will affect system operation

6. Describe the administrative controls and limits for the Combustible Gas Control system:

a. State Tech Specs/TRM LCOs that govern the Combustible Gas Control.

b. State the l hour action limit TS LCOs.

c. Given the conditions/status of the Combustible Gas Control system components and the appropriate sections of the Tech Spec, determine if operability requirements are met and what actions are required.

7. Discuss related Industry Events

a. Containment Hydrogen Analyzer Inoperable

b. Both Hydrogen Recombiners Inoperable, TS 3.0.3

c. Unlocked Manual Containment Isolation Valve Due To Equipment Interface VI. TRAINING AIDS:

A. Classroom Computer and Local Area Network (LAN) Access B. Computer projector C. Simulator (if available)

0PL271 REP Revision 2 Page 3 of 32 I. PROGRAM: OPERATOR TRAINING - LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: NP RADIOLOGICAL EMERGENCY PLAN AND SEQUOYAH EMERGENCY PLAN IMPLEMENTING PROCEDURES IV. LENGTH OF LESSONICOURSE: 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (Hot License Class), 2 -4 hours (LOR)

V. TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of the Radiological Emergency Plan (REP).

B. Enabling Objectives:

Demonstrate an understanding of NUREG 1122 Knowledge and Abilities associated with Radiological Emergency Plan that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A.

1. Discuss the Radiological Emergency Plan

a. Discuss the regulatory bases for the REP

b. State the purpose of the REP.

c. Define and state the purposes of a(n) NOUE, Alert, Site Area Emergency, and General Emergency

d. State the purpose and major job functions of the Technical Support Center (TSC), the Operations Support Center (OSC), the Central Emergency Control Center (CECC) and give the location of each.

e. Describe the role the state and federal agencies play during an event

f. Describe the process of authorizing Emergency Radiological Exposures in accordance with EPIP-15.

g. State the conditions under which onsite personnel would be administered potassium iodide (KI).

h. Describe Chemistry and Radiation Protection tasks during emergency operations.

i. Discuss the termination of a declared Radiological Emergency in accordance with EPIP-16.

2. Determine the required notifications based upon the event, including time requirements.

3. Classify emergency events using appropriate procedures.

0PL271 REP Revision 2 Page 4 of 32

4. Determine protective action recommendations using appropriate procedures.

5 State the duties and responsibilities of the Site Emergency Director (SED).

a. State the duties and responsibilities the SED may not delegate

b. State the conditions under which the SED may order relocation from one assembly point to another.

Discuss medical emergency response per EPIP-1O.

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta

92. Given the following:

- At 1330 a liquid radioactive release was commenced from the Monitor Tank to the cooling tower blowdown.

- At 1450, the following alarm is received:

0-RA-90-122B WDS LIQ EFF MON INSTR MALFUNC

- The radwaste AUO reports that 0-RCV-77-43, Liquid Radwaste Isolation Valve, is OPEN.

- Upon further investigation it is determined that 0-RA-90-122 has had no sample flow since 1345 and should have alarmed at that time.

- A subsequent grab sample indicated release activity exceeds 20 times the 1 OCFR2O limit.

Which ONE of the following identifies:

(1) If the current condition of 0-RCV-77-43 is correct for these conditions.

and (2) Whether NRC notification is required in accordance with SPP-3.5, Regulatory Reporting Requirements.

A. (1)Yes 0-RCV-77-43 is in the correct position.

(2) Yes notification is required.

B (1) No 0-RCV-77-43 should be CLOSED.

(2) Yes notification is required.

C. (1 )Yes 0-RCV-77-43 is in the correct position.

(2) Notification is NOT required.

D. (1) No 0-RCV-77-43 should be CLOSED.

(2) Notification is NOT required.

Saturday, July 17, 2010 1:43:56 PM 92

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta DISTRACTOR ANALYSIS:

A. Incorrect. Plausible because some release path monitors (S/G blowdown) complete isolation for a High Rad but not Instr. Malfunc and notification is required. Incorrect because 0-RCV-77-43 should be closed.

B. Correct. The instrument malfunction should have caused 0-RCV-77-43 to close.

Notification is required by 50.73(a) (2) (vii) (A) and 50. 72(a)(1) (I) as a result of declaration of an NOUE.

C. Incorrect. Plausible because some release path monitors (S/G blowdown) complete isolation for a High Rad but not Instr. Malfunc and notification is required only if the release exceeds 60 minutes. Incorrect because 0-RCV-77-43 should be closed and notification is required.

D. Incorrect, Plausible because the instrument malfunction alarm should have caused the release valve to go closed and notification is required only if the release exceeds 60 minutes. Incorrect because notification is required.

Saturday, July 17, 2010 1:43:56 PM 93

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta Question Number: 92 Tier: 2 Group 2 K/A: 068 Liquid Radwaste System (LRS)

A2.04 Ability to (a) predict the impacts of the following malfunctions or operations on the Liquid Radwaste System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Failure of automatic isolation.

Importance Rating: 3.3 I 3.3 10 CFR Part 55: 41.5 IOCFR55.43.b: 5 K/A Match: The question is a K/A match because the candidate must recognize failure of the liquid discharge monitor to isolate for an instrument malfunction alarm. SRO level because reporting requirements of 10CFR5O.73 must be evaluated.

Technical

Reference:

Dwg 45N677-1, Rev 16; SPP-3.5 Regulatory Reporting Requirements, Rev 22; EPIP-1 Emergency Plan Classification Matrix, Rev 44.

Proposed references None to be provided:

Learning Objective: OPL271SPP-3.5 Obj 3; OPT200.RM Obj 4 & 5 Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question 3-17-2010. LWP Comments:

MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9 Answer: BAADCDABAC ScrambleRange: A-D Saturday, July 17, 2010 1:43:56 PM 94

16 (C-2)

Source Setpoint O-RA-90-1 22B SER 766 (Unit 1 annunciator system) N/A WDS LIQ EFF MON 0-RM-90-122 Monitor INSTR MALFUNC NOTE All Probable Causes are dependent upon local control switch rO-HS-90-122-B1 being placed in Auto.

Probable 1. Instrument power failure.

Causes 2. 0-RE-90-122 monitor pump tripped (low flow)

3. Momentary interruption of 12OVAC to monitor (such as board transfer).

4. Instrument downscale failure.

Corrective [1] IF radwaste release in progress, THEN Actions REQUEST Radwaste AUO to verify 0-RCV-77-43 closed.

[2] CHECK 0-RM-90-122 on 0-M-12 for possible trouble.

[3] IF momentary interruption of 120VAC to monitor, THEN RESET this malfunction alarm by performing 0-SO-90-1 section 8.5.

[4] DISPATCH operator to determine problem at monitor.

[5] IF 0-RM-90-122 is inoperable, THEN

[a] NOTIFY Chemistry Shift Supervisor to comply with DDCM requirements.

[b] PLACE equipment off normal or mop tags, identifying condition.

[c] WHEN ODCM action is satisfied, THEN RESUME the release using appropriate procedures.

[6] COMPLY with ODCM, Section 1.1.1 requirements.

[7] INITIATE WO for maintenance, if required.

References 45B655-1 2B-0, 45N657-1 8, 45N677-1, 47B60 1-90-33 SQN 0-AR-M12-B Page 20 of 40 0 Rev. 29

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NPG Standard Regulatory Reporting Requirements SPP-3.5 Programs and Rev. 0022 Processes Page 29 of 75 Appendix A (Page 11 of 11)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 3.5 Written Report NRC (continued)

b. Remove residual heat;

c. Control the release of radioactive material; or

d. Mitigate the consequences of an accident.

11. §50.73(a)(2)(viii)(A) Any airborne radioactivity release that, when averaged over a time period of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , resulted in airborne radionuclide concentrations in an unrestricted area that exceeded 20 times the applicable concentration limits specified in Appendix B to Part 20, table 2, column 1.

12. §50.73(a)(2)(viii)(B) Any liquid effluent release that, when averaged over a time period of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , exceeds 20 times the applicable concentrations specified in Appendix B to Part 20, table 2, column 2, at the point of entry into the receiving waters (i.e., unrestricted area) for all radionuclides except tritium and dissolved noble gases.

13. §50.73(a)(2)(ix)(A) Any event or condition that as a result of a single cause could have prevented the fulfillment of a safety function for two or more trains or channels in different systems that are needed to:

a. Shut down the reactor and maintain it in a safe shutdown condition;

b. Remove residual heat;

c. Control the release of radioactive material; or

d. Mitigate the consequences of an accident.

NOTE Events covered above may include cases of procedural error, equipment failure, and/or discovery of a design, analysis, fabrication, construction, and/or procedural inadequacy. However, licensees are not required to report an event pursuant to this criterion if the event results from a shared dependency among trains or channels that is a natural or expected consequence of the approved plant design or normal and expected wear or degradation [50.73(a)(2)(ix)(B)}.

14. §50.73(a)(2)(x) Any event that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant including fires, toxic gas releases, or radioactive releases.

NPG Standard Regulatory Reporting Requirements SPP-3.5 Programs and Rev. 0022 Processes Page 20 of 75 Appendix A (Page 2 of 11)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 3.0 REQUIREMENTS NOTES

1) Internal management notification requirements for plant events are found in Appendix D. The Operations Shift Manager is responsible for notifying Site Operations Management and the Duty Plant Manager. The Duty Plant Manager is responsible for making the remaining internal management notifications.

2) NRC NUREG-1022, Supplements and subsequent revisions should be used as guidance for determining reportability of plant events pursuant to §50.72 and §50.73.

3.1 Immediate Notification - NRC TVA is required by §50.72 to notify NRC immediately if certain types of events occur. This appendix contains the types of events and the allotted time in which NRC must be notified.

(Refer to Form SPP-3.5-1 or NRC Form 361). Operations is responsible for making the reportability determinations for §50.72 and §50.73 reports. Operations is responsible for making the immediate notification to NRC in accordance with §50.72.

Notification is via the Emergency Notification System. If the Emergency Notification System is not operative, use a telephone, telegraph, mailgram, or facsimile.

NOTE The NRC Event Notification Worksheet may be used in preparing for notifying the NRC. This Worksheet may be obtained directly from the NRC website (wwwnrc.gov) under Form 361, or WA NPG Form SPP-3.5-1 may be used.

A. The Immediate Notification Criteria of §50.72 is divided into 1-hour, 4-hour, and 8- hour phone calls. Notify the NRC Operations Center within the applicable time limit for any item which is identified in the Immediate Notification Criteria.

B. The following criteria require 1-hour notification:

1. (Technical Specifications) Safety Limits as defined by the Technical Specifications which have been violated.

2. §50.72 (a)(1 )(i) The declaration of any of the Emergency classes specified in the licensees approved Emergency Plan.

SEQUOYAH EMERGENCY PLAN CLASSIFICATION MATRIX EPPP-1 Mode Initiating I Condition Mode Initiating I Condition EAB dose, resulting from an actual or imminent Not Applicable.

release of gaseous radioactivity> I Rem TEDE or

> 5 Rem thyroid CDE for the actual or projected duration of release. (1 or 2 or 3):

1. A VALID rad monitor reading exceeds the values under A General Emergency in Table 7-1 for >15 mm, unless L assessment within that 15 mm confirms that the criterion is L not exceeded.

OR

2. Field surveys indicate >lRem/hr gamma or an 1-131 concentration of 3.9E-06 pCi/cm 3 at the EAB (Fig. 7-A)

OR

3. Dose assessment results indicate EAB dose >1 Rem TEDE or >5 Rem thyroid CDE for the actual or projected duration of the release (Fig. 7-A).

EAB dose resulting from an actual or imminent Not Applicable.

release of gaseous radioactivity >100 mrem TEDE or

>500 mrem thyroid CDE for actual or projected duration of release. (1 or 2 or 3):

1. A VALID rad monitor reading > Table 7-1 values under Site A Area for> 15 mm, unless assessment within that 15 mm L confirms that the criterion is not exceeded.

L

2. Field surveys indicate >100 mrem/hr gamma or an 1-131 conc of 3.gE-07 pCi/cm at the EAB (Fig. 7-A).

OR

3. Dose assessment results indicate EAB dose >100 mrem TEDE or >500 mrem thyroid CDE for actual or projected duration of the release (Fig. 7-A).

Any UNPLANNED release of gaseous radioactivity Any UNPLANNED release of liquid radioactivity that that exceeds 200 times the ODCM Section 1.2.2.1 exceeds 200 times the ODCM Section 1.2.1.1 Limit Limit for >15 minutes. (1 or 2 or 3 or 4) for >15 minutes. (1 or2)

1. A VALID rad monitor reading > Table 7-1 values under Alert 1. A VALID rad monitor reading > Table 7-1 values for >15 minutes, unless assessment within that 15 minutes under Alert for >15 minutes, unless assessment A confirms that the criterion is not exceeded. A within this time period confirms that the criterion is L L

2. Field surveys indicate >10 mrem/hr gamma at the EAB for not exceeded.

L

>15 minutes (Fig 7-A). L OR OR

3. Dose assessment results indicate EAB dose >10 mrem 2. Sample results indicate an ECL >200 times the TEDE for the duration of the release (Fig. 7-A). ODCM limit value for an unmonitored release of OR liquid radioactivity >15 minutes in duration

4. Sample results exceed 200 times the ODCM limit value for an unmonitored release of gaseous radioactivity >15 minutes in duration.

Any UNPLANNED release of gaseous radioactivity Any UNPLANNED release of liquid radioactivity to that exceeds 2 times the ODCM Section 1.2.2.1 Limit the environment that exceeds 2 times the ODCM for >60 minutes. (1 or2or3or4) Section 1.2.1.1 Limit for >60 minutes.

(1 or2)

1. A VALID rad monitor reading > Table 7-1 values under UE for >60 minutes, unless assessment within that 60 minutes A confirms that the criterion is not exceeded. A 1. A VALID red monitor reading > Table 7-1 values L under UE for >60 minutes, unless assessment OR L within this time period confirms that the criterion is

2. Field surveys indicate >0.1 mremlhr gamma at the EAB for L >60 minutes (Fig 7-A)

L not exceeded.

OR

3. Dose assessment results indicate EAB dose >0.1 mrem OR TEDE for the duration of the release (Fig. 7-A).

OR 2. Sample results indicate an ECL >2 times the ODCM

4. Sample results exceed 2 times the ODCM limit value for an limit value for an unmonitored release of liquid unmonitored release of gaseous radioactivity >60 minutes in duration radioactivity >60 minutes in duration.

Page 43 of 47 Revision 44

0PL27 1 SPP-3 .5 Rev. 2 Page 3 of 19 I. PROGRAM: OPERATOR TRAINING LICENSED -

II. COURSE: LICENSE TRAiNING III. LESSON TITLE: REPORTING REQUIREMENTS IV. LENGTH OF LESSON/COURSE: 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (s)

V. TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the actions necessary to comply with regulatory and plant reporting requirements.

B. Enabling Objectives Demonstrate an understanding of NTJREG 1 122 Knowledges and Abilities associated with Regulatory and Plant Reporting Requirements that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A.

2. Perform a plant response assessment using the 0-TI-QXX-000-00l.0, Event Critique, Post Trip Report, and Equipment Root Cause.

a. State the responsibilities of each control room crewmember. [C. 1]

b. Explain the process or Conduct a plant response assessment.

3. For a given condition, determine the regulatory reporting requirements using appropriate reference material.

a. List the tools available to the operator for determining regulatory reporting requirements.

b. Define the key terms used to determine regulatory reporting requirements.

c. State the criteria requiring one-hour notification of the NRC.

d. State the criteria requiring four-hour notification of the NRC.

e. State the criteria requiring eight-hour notification of the NRC.

f. State the criteria requiring 24-hour notification of the NRC.

g. State the criteria requiring 2-day notification of the NRC.

h. State the criteria requiring a written report or LER to the NRC.

I. State the criteria allowing a telephone notification to be made in lieu of a written LER to the NRC.

4. For a given condition, determine plant management reporting requirements using SPP-3 .5.

5. Complete a PER reportability determination per SPP-3.l.

OPT200. RM Rev. 3 Page 3 of 166 PROGRAM: OPERATOR TRAINING COURSE: SYSTEMS TRAINING III. TITLE: RADIATION MONITORiNG SYSTEM IV. LENGTH OF LESSON: 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months lecture; 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months simulator demonstration; 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months self-study/workshop V. TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of this lesson and others presented, the student should be able to apply the knowledge to support satisfactory performance of the tasks associated with the Radiation Monitoring System in the plant and on the simulator.

B. Enabling Objectives:

0. Demonstrate an understanding of NUREG 1122 knowledges and abilities associated with the Radiation Monitoring System as identified in Appendix A.

1. State the purpose/functions of the Radiation Monitoring System as described in the SQN FSAR.

2. State the design basis of the Radiation Monitoring System in accordance with the SQN FSAR.

3. Explain the purpose/function of each major component in the flow path of the Radiation Monitoring System as illustrated on a simplified system drawing.

4. Describe the following characteristics of each major component in the Radiation Monitoring System:

a. Location

b. Power supply (include control power as applicable)

c. Support equipment and systems

d. Normal operating parameters

e. Component operation

f. Controls

g. Interlocks (including setpoints)

h. Instrumentation and Indications

i. Protective features (including setpoints)

j. Failure modes

k. Unit differences

1. Types of accidents for which the components are designed

m. Location of controls and indications in the control room and auxiliary control room

OPT200. RM Rev. 3 Page 4 of 166 V. TRAINING OBJECTIVES (Contd):

B. Enabling Objectives (Contd):

5. Describe the operation of the Radiation Monitoring System:

a. Precautions and limitations

b. Major steps performed while placing the system in service

c. Alarms and alarm response

d. How a component failure will affect system operation

e. How a support system failure will affect system operation

f. How a instrument failure will affect system operation

6. Describe the administrative controls and limits for the Radiation Monitoring System:

a. State Tech Specs/TRM LCOs that govern the system.

b. State the l hour action limit TS LCOs

c. Given the conditions/status of the Radiation Monitoring System components and the appropriate sections of the Tech Spec, determine if operability requirements are met and what actions are required

7. Discuss related Industry Events VI. TRAINING AIDS:

A. Classroom Computer and Local Area Network (LAN) Access B. Computer projector C. Simulator (if available)

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta

93. Given the following plant conditions:

- A release of Waste Gas Tank B is in progress to the Unit 2 stack with 2-RM-90-400 Unit 2 Shield Building Rad Monitor and 0-RM-90-1 18 Waste Gas Effluent Rad Monitor both Operable

- Power is lost to 0-RM-90-1 18 Which of the following identifies; (1) the affect the loss of power will have on the release, and, (2) the requirement to allow any additional release of the tank with radiation monitor 0-RM-90-1 18 remaining out of service?

A (1) The release will automatically terminate.

(2) The release package must be returned to Chemistry before any additional release from the tank due to 0-RM-90-1 18 being Inoperable.

B. (1) The release will automatically terminate.

(2) The existing release package can be used for any additional release from the tank because 2-RM-90-400 remains Operable.

C. (1) A manual termination of the release will be required.

(2) The release package must be returned to Chemistry before any additional release from the tank due to 0-RM-90-118 being Inoperable.

D. (1) A manual termination of the release will be required.

(2) The existing release package can be used for any additional release from the tank because 2-RM-90-400 remains Operable.

Thursday, July 15, 2010 2:13:51 PM 93

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta DISTRACTOR ANALYSIS:

A. CORRECT, the release would be automatically terminated due to an instrument malfunction and the release package must be revised to satisfy ODCM Action 40 requirements for 0-RM-90-1 18 being Inoperable.

B. Incorrect, Plausible because the release would be automatically terminated and 0-RM-90-400 monitors the shield building effluent. Incorrect because the release package must be revised to satisfy ODCM Action 40 requirements for 0-RM-90-1 18 being Inoperable.

C. Incorrect, Plausible because some release point radiation monitor instrument malfunctions only cause an alarm and the release package must be revised to satisfy ODCM Action 40 requirements for 0-RM-90-1 18 being Inoperable.

Incorrect because 0-RM-90-1 18 malfunction causes closure of 0-FCV-77-1 19 release valve.

D. Incorrect, Plausible because some release point radiation monitor instrument malfunctions only cause an alarm and 0-RM-90-400 monitors the shield building effluent. Incorrect because 0-RM-90-1 18 malfunction causes closure of 0-FCV-77-1 19 release valve and the release package must be revised to satisfy ODCM Action 40 requirements for 0-RM-90-1 18 being Inoperable.

Thursday, July 15, 2010 2:13:51 PM 94

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta Question Number: 93 Tier: 2 Group 2 KIA: 071 Waste Gas Disposal System (WGDS)

G2.2.36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

Importance Rating: 3.1 /4.2 10 CFR Part 55: 41.10 IOCFR55.43.b: 2 KIA Match: Question matches the K/A by having the candidate determine what the effect of a loss of power will have on a waste gas release in progress and whether that failure or maintenance issue will prevent the release of gaseous waste or affect the ODCM LCO for gaseous waste.

Technical

Reference:

0-AR-M12-B, (C4) rev 29, 47W611-77-4, Rev 10; 0-SO-77-15 Waste Gas Decay Tank Release, Rev. 16; 0-Sl-CEM-077-410.4,Waste Gas Decay Tank Release, Rev.15; ODCM Offsite Dose Calculation Manual, Rev 55 Proposed references None to be provided:

Learning Objective: OPT200.RM B.5.d Cognitive Level:

Higher X Lower Question Source:

New Modified Bank X Bank Question History: WBN bank # G0712-2 #071 G2.2.36 Comments: Modified for SQN MCS Time: I Points: 100 Version: 0 1 2 3 4 5 6 7 8 9 Answer: A A A A A A A A A A Items Not Scrambled Thursday, July 15, 2010 2:13:51 PM 95

18 (C-4)

Source Setpoint O-RA-90-1 1 8B (Unit 1 annunciator system) N/A WDS GAS EFF MON

- INSTR MALFUNC Probable Causes 1. Instrument power failure.

2. Instrument placed in TRIP ADJ position (Except for monitors equipped with RM-l 000 modules only).

3. Instrument downscale failure or loss of signal.

4. Operate/Calibrate switch set to calibrate (RM-1 000 modules only).

Corrective [1] IF gas release in progress, THEN Actions REQUEST Radwaste AUO to verify 0-RCV-90-1 19 closed.

[2] CHECK 0-RM-90-1 18 on O-M-12 for possible trouble.

[3] IF 0-RM-90-1 18 is inoperable, THEN

[a] NOTIFY the Chemistry Shift Supervisor to comply with ODCM requirements.

[b] PLACE equipment off normal or mop tags, identifying the condition.

[c] WHEN ODCM action is satisfied, THEN RESUME the release using appropriate procedures.

[4] COMPLY with ODCM, Section 1.1.2 requirements.

[5] INITIATE WO for maintenance, if required.

References 45B655-1 2B-0, 45N657-1 8, 45N667-1, SQN 0-AR-M12-B Page 22 of 40 0 Rev. 29

DESCRIPT IDA TVA ND WEST ND NOTED:

RELAY DEGNY YANK A GAS PCV-17-1I5B I. FOR GENERAL NOTES AND REFERENCES GEE SHEET 4lWBll77l.

ANALYZER SARPLING V1.V lOGAN GOX1A GAOl

2. ONE WASTE GAS COMPRESSGR WILL BE RUN CSNYINUOJSLY FS DECAY TANK B GAS PCV77114B 1S37B GDD2A GAGT RITA THE OTHER SERVING AS A BACKUP TO BE SIARTED WHEN I A YZER SAMPLING VLV HEADER PRESSURE EGCEEDS 2 PSIG. THE COMPRESSORS WILL DECAY TANK C GAG PCV77.-1135 WE A_TERRACES FOR UNIFORM WEAR.

I ANPYZER SAMPLING ALA 1D28B 0003A GAUl G. ONE DECAY TANK PAESSJRE ISOLATION VALVE WILL NOAMALLY

[E DECAY TANK B GAS cOX4A GAXI BE OPEN WITH AWOTAER SELECTED FOR STANDBY. ALL OTHERS L.LYZER SAAPLINC VLV PCV77-TTGB 103AM WILL BE CLOSED.

I GAS DECAY TANK E GAS 4. THIS IS A VANS OPERATED NEEDLE VALVE.

I A YZER SAMPLING VLV 77lTAM 1552B GDXSA GAOl IE DECAY TANK F GAS GDGBA GAOl I ANALYZER SAAPLING vLv PCV-77102A 105GB E DECAY TANK 0 GAS CDG7A GAUl ADA/IN I AA YZER SAMPLING ALA PCV771450 IDS4B DECAY TANK H GAS GDXBA (TYP ALL SAC-H I ANA YZER SAMPLING ALA PCA77146B 10019 GAUl

[EE DECAY TANK J GAS CDXNA GAG1 L.YZEA SAMPLING VLV PCV771479 05KB PCV jib TANK S TANK C TANK TANK E FOWERHOUSE TANK F UNITS 1 & 2 MECHAN ICAL TANK LOGIC DIAGRAM WASTE DISPOSAL SYSTEM THAN H SEQUDYAH NUCLEAR PLANT TENNESSEE VALLEY AUTHORITY TANK IAIYIAI. ISSUE I ENGINEERING CHECKER:

J APPROVAL AD ISSUE PER:

J.CCLE L.WALTERS DRAWING MADE 1 N/A -

A CCO FRSH AC-N/A 9 AM-RI REVIEWER: KUR N/A SEI 73RI 4 CCD NO:1 ,247W611774R11

SQN WASTE GAS DECAY TANK RELEASE 0-S0-77-15 Unit 0 Rev. 0016 PagelOof2O Date 6.0 NORMAL OPERATION (continued)

[6] ALIGN the gas discharge header for release as follows:

[6.1] ENSURE Attachment 1, Power Checklist 0-77-1 5.01 has been completed.

NOTE Independent and/or second party verification is only applicable when radiation monitor 0-RM-90-118 is inoperable.

[6.2] ENSURE Attachment 2, Valve Checklist 0-77-15.02 has been completed.

[6.3] VERIFY the following valves are CLOSED.

A. 0-77-748A B. 0-77-748B C. 0-77-748C D. 0-77-748D E. 0-77-748E F. 0-77-748F G. 0-77-748G H. 0-77-748H I. 0-77-748J

SQN WASTE GAS DECAY TANK RELEASE 0-SO-77-15 Unit 0 Rev. 0016 Page 12 of 20 Date 6.0 NORMAL OPERATION (continued)

NOTE Radiation Control Valve 0-FCV-77-1 19 is interlocked with i-FT-30-150 and 2-FT-30-165, and will terminate the release if the fan stops running.

[9] VERIFY the applicable ABGTS fan is running.

UO

[10] IF I0-RE-90-1181 is operable, THEN ENSURE 0-RM-90-118 is in service.

UO

[ii] IF [0-RE-90-1181 is INOPERABLE, THEN ENSURE Compliance with actions of ODCM 1.1.2, AND WO initiated for corrective action.

UO NOTE If the selected unit shield bldg. radiation monitor is inoperable, THEN the following step may be N/A.

[12] VERIFY radiation monitor RM-90-400 is in service and operable for the Shield Building vent indicated in step 6.0[3].

UO

[13] IF the shield bldg vent release path selected in step 6.0[3],

radiation monitor RM-90-400 is inoperable, THEN NOTIFY the U-i SRO to take appropriate actions in accordance with Section 1.1.2 of the ODCM, AND WO initiated for corrective action.

UO

SQN Waste Gas Decay Tank Release 0-SI-CEM-077-410.4 Unit 0 Rev. 0015 Page 14 of 42 6.2 Pre-Release Instructions - Chemistry (continued)

[12] IF the setpoint for 0-RM-90-1 18 in step 6.2[l 1] is greater than step 6.1[3], THEN UPDATE the CDAS CHEM5 setpoint screen in accordance with Appendix H. D NOTE Appendix A is based on the setpoint of PCV-77-117 (2.0 psig) and a temperature of 80°F.

[13] DETERMINE the expected pressure drop across flow orifice F0-FE-77-2301 using Appendix A, the maximum allowable release rate from Step 6.2[1 1] above and hydrogen concentration from Steps 6.2[3] or 6.2[4].

I inches of H 0

2 Performer

[14] RECORD release number and the allowable pressure drop across flow orifice [0-FE-77-2301 as the smaller of Step 6.2[13]

or the administrative limit of 14 inches of water on Appendix B.

[15] IF [0-RE-90-1181 is INOPERABLE, THEN OBTAIN independent verification of Steps 6.2[13] and 6.2[14]

above.

I / I inches of H 0 2 md. Verifier Date Time

SQN Waste Gas Decay Tank Release 0-SI-CEM-077-410.4 Unit 0 Rev. 0015 Page 17 of 42 6.3 Release Instructions Operations (continued)

[8] IF release is to continue with F0-FE-77-2301 INOPERABLE, THEN PERFORM the following substeps.

[8.1] ENSURE that a test gauge (0 20 inches of H

- 0 2

suggested) is installed across f0-FE-77-230I.

[8.2] ENSURE serial number, range and calibration due date of test gauge along with installing Instrument Mechanics initials are recorded in remarks section of Appendix B.

[8.3] ENSURE pressure readings from test gauge are recorded in place of rO-FE-77-2301 readings on Appendix B.

[9] IF F0-RM-90-1181 or IRM-90-4001 alarms, THEN NOTIFY On-shift Chemistry Personnel who will contact the Cognizant Chemist/System Engg. for further guidance in processing tank contents.

[10] WHEN release is complete or stopped, THEN RECORD the following on Appendix B.

[10.1] Release stop time D

[10.2] WGDT psig

[10.3] Initials D

[11] IF radiation monitor setpoint changes were made (Step 6.3[2]), THEN RETURN the radiation monitors to their initial setpoints.

[12] NOTIFY the US/SRO and On-shift Chemistry Personnel that this release is complete.

[13] REVIEW 0-SO-77-15.

[14] ATTACH 0-SO-77-15 to this release package.

[15] TRANSMIT the release package to the Chemistry Laboratory for post release evaluation.

SQN ODCM Revision 55 Page 20 of 171 Table 1.1-2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION (Page 1 of 3)

Minimum Channels Instrument OPERABLE Applicability Action WASTE GAS DISPOSAL SYSTEM

a. Noble Gas Activity Monitor (0-RM-90-1 18A) 1 ***

40

b. Waste Decay Tanks Effluent Flow Rate 1 ***

41 Measuring Device (0-Fl-77-230)

2. CONDENSER VACUUM EXHAUST SYSTEM

a. Noble Gas Activity Monitor (1,2-RM-90-99,119) 1 ***

42

b. Vent Flow Rate Monitor (1 ,2-F-2-256,257) 1 ****

41

3. SHIELD BUILDING EXHAUST SYSTEM

a. Noble Gas Vent Rate Activity Monitor 1** 43 (1 ,2-R-90-400_,Eff)

b. Iodine Sampler (1 ,2-FLT-90-402A or 4028) 1 43,44

c. Particulate Sampler (1 ,2-FLT-90-402A or 402B) 1 ***

43,44

d. Vent Flow Rate Monitor (1 ,2-FM-90-400),(1 ,2- 1 41,43 RI-90-400 Monitor Item 029 or 1 ,2-FI-90-400)

e. Sampler Flow Rate Monitor (1 ,2-RI-90-400 1 45 Monitor Item 028 or 1 ,2-FI-90-400A)

f. Primary Sampling Pumps (1 ,2-PMP-90-452A or 1 46 452B)

4. AUXILIARY BUILDING VENTILATION SYSTEM

a. Noble Gas Activity Monitor (0-RM-90-1O1B) 1

42

b. Iodine Sampler (0-FLT-90-101) 1

44

c. Particulate Sampler (0-FLT-90-101) 1

44

d. Vent Flow Rate Monitor (0-F-30-174) 1

41

e. Sampler Flow Rate Monitor (0-FIS-90-101) 1

45

5. SERVICE BUILDING VENTILATION SYSTEM

a. Noble Gas Activity Monitor (0-RM-90-132B) 1

42

b. Vent Flow Rate Monitor (0-F-90-5132A) 1

41

6. LOWER CONTAINMENT AIRBORNE ACTIVITY FOR VENTING

a. Noble gas Activity Monitor (1 ,2-RM-90-106B,1 12B) 1 Modes 1-4 47 (during venting)

7. CONTAINMENT AIRBORNE ACTIVITY FOR PURGING

a. Noble gas Activity Monitor (1,2-RM-90-1068,112B) 1 (for compartment Modes 14*** 48 being purged)

Ai all times.

Operability of shield building noble gas vent rate activity monitor (Eff) requires primary sample pumps, vent flow rate and Low Rng radiation inputs since the high radiation alarm is only on the effluent channel (iiCils). When the vent flow rate monitor is inoperable only the Low Rng radiation monitor is still capable of providing noble gas indication.

Its associated malfunction alarm on M-30 will alert Operations to any secondary failures such as loss of sample flow or detector failure. Therefore as long as the Low Rng is selected and the malfunction alarm is monitored RE-90-400 can be used to meet ODCM Noble gas and Particulate/Iodine sampler requirements. Mid or High range channels do not have any ODCM requirements. The vent flow rate monitor can be OPERABLE without the noble gas vent rate activity monitor.

During exhaust system operation.

1 ,2-F-2-256 is the low range flow element (0-100 cfm). 1,2 -F-2-257 is a mid (0-1000) cfm flow element which should only be used for abnormal conditions.

SQN ODCM Revision 55 Page 21 of 171 Table 1.1-2 RADIOACTIVE GASEOUS EFFLUENT MOMTORING INSTRUMENTATION (Page 2 of 3)TABLE NOTATION ACTION 40 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank(s) may be released to the environment provided that prior to initiating the release:

a. At least two independent samples of the tanks contents obtained by two technically qualified members of the facility staff are analyzed, and

b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and;

c. At least two technically qualified members of the Facility Staff independently verify the discharge valve lineup.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 41 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 42 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for noble gas activity.

ACTION 43 - Alarm Window M-5C-29, Vent Flow Rate only Inoperable With any inoperable vent flow monitor on a discharge pathway where a fan is operating (Purge A, Purge B, ABGTS, or EGTS), effluent release may continue provided: (a) Low Rng on RI-90-400 is selected instead of Eff; and (b) at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated instrument malfunction, M-30-15, is verified not annunciated; and (c) a reading from Low Rng on RE-90-400 is obtained at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The High Rad alarm on M-30 will not be operable under this alignment. Also Action 41 must be complied with.

Alarm Window M-30-15, Radiation Monitor Inoperable With the low range channel inoperable, effluent releases may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for noble gas activity. Also Action 44 must be complied with.

ACTION 44 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue provided that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the channel has been declared inoperable samples are continuously collected as required in Table 2.2-2. Also Action 45 must be complied with.

ACTION 45 - With the number of channels OPERABLE less than required by minimum channels OPERABLE requirement, effluent releases via this pathway may continue provided the sampler flow rate is verified at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SQN ODCM Revision 55 Page 145 of 171 Figure 7.1 GASEOUS EFFLUENT RELEASE POINTS Shield Building Vent Auxiliary Condenser Building Service Vacuum Vent Building Vent

.J-iot Machine Shops J Office and GenI Area Control Area O-RE-90-132 RCA Access Control 4 Chemistry Lab Titration Room SERVICE TURBINE AUXILIARY SHIELD BUILDING BUILDING BUILDING BUILDING

OPT200. RM Rev. 3 Page 3 of 166 PROGRAM: OPERATOR TRAIMNG II. COURSE: SYSTEMS TRAINING III. TITLE: RADIATION MONITORING SYSTEM IV. LENGTH OF LESSON: 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months lecture; 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months simulator demonstration; 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months self-study/workshop V. TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of this lesson and others presented, the student should be able to apply the knowledge to support satisfactory performance of the tasks associated with the Radiation Monitoring System in the plant and on the simulator.

B. Enabling Objectives:

0. Demonstrate an understanding of NUREG 1122 knowledges and abilities associated with the Radiation Monitoring System as identified in Appendix A.

1. State the purpose/functions of the Radiation Monitoring System as described in the SQN FSAR.

2. State the design basis of the Radiation Monitoring System in accordance with the SQN FSAR.

3. Explain the purpose/function of each major component in the flow path of the Radiation Monitoring System as illustrated on a simplified system drawing.

4. Describe the following characteristics of each major component in the Radiation Monitoring System:

a. Location

b. Power supply (include control power as applicable)

c. Support equipment and systems

d. Normal operating parameters

e. Component operation

f. Controls

g. Interlocks (including setpoints)

h. Instrumentation and Indications

i. Protective features (including setpoints)

j. Failure modes

k. Unit differences

1. Types of accidents for which the components are designed

m. Location of controls and indications in the control room and auxiliary control room

QPT200. RM Rev. 3 Page 4 of 166 V. TRAINING OBJECTIVES (Contd):

B. Enabling Objectives (Contd):

5. Describe the operation of the Radiation Monitoring System:

a. Precautions and limitations

b. Major steps performed while placing the system in service

c. Alarms and alarm response

d. How a component failure will affect system operation

e. How a support system failure will affect system operation

f. How a instrument failure will affect system operation

6. Describe the administrative controls and limits for the Radiation Monitoring System:

a. State Tech Specs/TRM LCOs that govern the system.

b. State the l hour action limit TS LCOs

c. Given the conditions/status of the Radiation Monitoring System components and the appropriate sections of the Tech Spec, determine if operability requirements are met and what actions are required

7. Discuss related Industry Events VI. TRAINING AIDS:

A. Classroom Computer and Local Area Network (LAN) Access B. Computer projector C. Simulator (if available)

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta

1. In accordance with Tech Spec 3.7.7, Control Room Emergency Ventilation System, (CREVS) Bases, which ONE of the following completes the statements below?

The Control Room Emergency Ventilation System (CREVS), when aligned for an emergency, (1) all makeup air for the control room envelope (CRE.)

The system is designed to maintain occupant dose for the (2) period after a design basis accident to less than or equal to 5 REM.

A. filters 7 day B filters 30 day C. isolates 7 day D. isolates 30 day DISTRA CTOR ANAL ISIS:

A. Incorrect, Plausible since the emergency CREVS flow path filters outside air to achieve pressurization of the enclosure and the control room air conditioning system TS 3.7.15 permits 7 days with both trains Inoperable if temperature remains less than 90 degrees. Incorrect because the basis for LCO 3.7.7 describes the design criteria of the CREV maintains the dose for a 30 day period.

B. Correct, The emergency flow path filters outside air to achieve pressurization of the CRE and the basis for LCO 3.7.7 describes the design criteria of the CREVS maintains the dose less than 5 REM for a 30 day period.

C. Incorrect, Plausible because the normal CREVS intake path is isolated when the system is aligned in the accident mode and the control room air conditioning system TS 3.7.15 permits 7 days with both trains Inoperable if temperature remains less than 90 degrees. Incorrect because a filtered makeup path maintains CRE pressure positive and the basis for LCO 3.7.7 describes the design criteria of the CREV maintains the dose for a 30 day period.

D. Incorrect, Plausible because the normal CREVS intake path is isolated when the system is aligned in the accident mode and the basis for LCO 3.7.7 describes the design criteria of the CREVS maintains the dose less than 5 REM for a 30 day period, Incorrect because a filtered makeup path maintains CRE pressure positive.

Saturday, July 17, 2010 1:11:10 PM

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta Question Number: 94 Tier: 3 Group n/a K/A: G 2.1.27 Knowledge of system purpose and/or function.

Importance Rating: 3.9 / 4.0 10 CFR Part 55: 41.7 IOCFR55.43.b: 2 KIA Match: Question matches the K/A by having the candidate identify the function of the CREVS system as outlined in the Tech Spec basis.

Technical

Reference:

Tech Spec 3.7.7 and Bases.

Tech Spec 3.7.15 Proposed references None to be provided:

Learning Objective: OPT200.CBVENT B2, 5, 6 Cognitive Level:

Higher Lower X Question Source:

New X Modified Bank Bank Question History: Question written for 1009 NRC exam Comments:

Saturday, July 17, 2010 1:11:10PM 2

PLANT SYSTEMS 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.7 Two independent control room emergency ventilation systems (CREVS) shall be OPERABLE.*

APPLICABILITY: ALL MODES and during movement of irradiated fuel assemblies ACTION:

MODES 1, 2, 3 and 4

a. With one CREVS inoperable for reasons other than Action b, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With one or more CREVS trains inoperable due to inoperable control room envelope (CRE) boundary, immediately initiate action to implement mitigating actions, and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months verify mitigating actions ensure CRE occupant exposures to radiological and chemical hazards will not exceed limits, CRE occupants are protected from smoke hazards, and restore CRE boundary to OPERABLE status within 90 days. Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With both CREVS inoperable due to actions taken as a result of a tornado warning, restore at least one train to operable status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in a least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d. With both CREVS inoperable for reasons other than Action b. or Action c., be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months MODES 5, 6, and during movement of irradiated fuel assemblies

a. With one CREVS inoperable, restore the inoperable system to OPERABLE status within 7 days or initiate and maintain operation of the operable CREVS in the recirculation mode.

or suspend movement of irradiated fuel assemblies.

b. With both CREVS inoperable or one or more CREVS trains inoperable due to an inoperable CRE bounday, suspend all operations involving movement of irradiated fuel assemblies.

SURVEILLANCE REQUIREMENTS 4.7.7 Each CREVS shall be demonstrated OPERABLE:

a. DELETED

b. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 15 minutes.

The CRE boundary may be opened intermittently under administrative control.

October 28, 2008 SEQUOYAH - UNIT 1 3/4 7-17 Amendment No. 12, 164, 187, 256, 260, 273, 301, 321

B 3/4.7 PLANT SYSTEMS B 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM BASES BACKGROUND The CREVS provides a protected environment from which occupants can control the unit following an uncontrolled release of radioactivity, hazardous chemicals, or smoke.

The CREVS consists of two independent, redundant trains that recirculate and filter the air in the control room envelope (CRE) and a CRE boundary that limits the inleakage of unfiltered air. Each CREVS train consists of a high efficiency particulate air (HEPA) filter, an activated charcoal adsorber section for removal of gaseous activity (principally iodines), and a fan. Ductwork, valves or dampers, doors, barriers, and instrumentation also form part of the system.

The CRE is the area within the confines of the CRE boundary that contains the spaces that control room occupants inhabit to control the unit during normal and accident conditions. This area encompasses the control room, and may encompass other non critical areas to which frequent personnel access or continuous occupancy is not necessary in the event of an accident. The CRE is protected during normal operation, natural events, and accident conditions. The ORE boundary is the combination of walls, floor, roof, ducting, doors, penetrations and equipment that physically form the ORE. The OPERABILITY of the ORE boundary must be maintained to ensure that the inleakage of unfiltered air into the ORE will not exceed the inleakage assumed in the licensing basis analysis of design basis accident (DBA) consequences to ORE occupants.

The ORE and its boundary are defined in the Control Room Envelope Habitability Program.

The CREVS is an emergency system, parts of which may also operate during normal unit operations in the standby mode of operation. Actuation of the CREVS places the system in the emergency radiation state mode of operation. Actuation of the system to the emergency radiation state of the emergency mode of operation, closes the unfiltered outside air intake and unfiltered exhaust dampers, and aligns the system for recirculation of the air October 28, 2008 SEQUOYAH UNIT I

- B 3/4 7-4d Amendment No. 321

CREVS B 3/4.7.7 PLANT SYSTEMS BASES BACKGROUND (continued) within the CRE through the redundant trains of HEPA and the charcoal filters. The emergency radiation state also initiates pressurization and filtered ventilation of the air supply to the CRE.

Outside air is filtered and added to the air being recirculated from the CRE. Pressurization of the CRE minimizes infiltration of unfiltered air through the CRE boundary from all the surrounding areas adjacent to the CRE boundary. The air entering the CRE is continuously monitored by radiation detectors. One detector output above the setpoint will cause actuation of the emergency radiation state.

A single CREVS train operating at a flow rate of 4000 cfm plus or minus 10 percent will pressurize the main control room to 0.125 inch water gauge relative to outside atmosphere. The CRE will be maintained at a slightly positive pressure relative to external areas adjacent to the CRE boundary. The CREVS operation in maintaining the CRE habitable is discussed in the Updated Final Safety Analysis Report (UFSAR), Sections 6.4 and 9.4 (Ref. I and 2).

Redundant supply and recirculation trains provide the required filtration should an excessive pressure drop develop across the other filter train. Normally open isolation dampers are arranged in series pairs so that the failure of one damper to shut will not result in a breach of isolation. The CREVS is designed in accordance with Seismic Category I requirements.

The CREVS is designed to maintain a habitable environment in the CRE for 30 days of continuous occupancy after a DBA without exceeding a 5 rem whole body dose or its equivalent to any part of the body.

APPLICABLE The CREVS components are arranged in redundant, safety related SAFETY ventilation trains. The location of components and ducting within ANALYSES the CRE ensures an adequate supply of filtered air to all areas requiring access. The CREVS provides airborne radiological October 28, 2008 SEQUOYAH UNIT 1

- B 3/4 7-4e Amendment No. 321

CREVS B 3/4.7.7 PLANT SYSTEMS BASES LCO (continued)

In order for the CREVS trains to be considered OPERABLE, the CRE boundary must be maintained such that the CRE occupant dose from a large radioactive release does not exceed the calculated dose in the licensing basis consequence analyses for DBAs.

The LCO is modified by a Note allowing the CRE boundary to be opened intermittently under administrative controls. This Note only applies to openings in the CRE boundary that can be rapidly restored to the design condition, such as doors, hatches, floor plugs, and access panels. For entry and exit through doors, the administrative control of the opening is performed by the person(s) entering or exiting the area. For other openings, these controls should be proceduralized and consist of stationing a dedicated individual at the opening who is in continuous communication with the operators in the ORE. This individual will have a method to rapidly close the opening and to restore the ORE boundary to a condition equivalent to the design condition when a need for CRE isolation is indicated.

APPLICABILITY In MODES 1, 2, 3, 4, 5, and 6, and during movement of irradiated fuel assemblies, the CREVS must be OPERABLE to ensure that the CRE will remain habitable during and following a DBA.

During movement of irradiated fuel assemblies, the CREVS must be OPERABLE to cope with the release from a fuel handling accident.

ACTIONS a. (MODES 1, 2, 3, and 4)

When one CREVS train is inoperable, for reasons other than an inoperable CRE boundary, action must be taken to restore OPERABLE status within 7 days. In this condition, the remaining OPERABLE OREVS train is adequate to perform the ORE occupant protection function. However, the overall reliability is reduced because a failure in the OPERABLE CREVS train could result in loss of CREVS function. The 7 day completion time is based on the low probability of a DBA occurring during this time period, and ability of the remaining train to provide the required capability.

October 28, 2008 SEQUOYAH UNIT 1

- B 3/4 7-4g Amendment No. 321

PLANT SYSTEMS 3/4.7.15 CONTROL ROOM AIR-CONDITIONING SYSTEM (CRACS)

LIMITING CONDITION FOR OPERATION 3.7.15 Two independent control room air-conditioning systems (CRACS) shall be OPERABLE.

APPLICABILITY: ALL MODES and during movement of irradiated fuel assemblies ACTION:

MODES 1,2,3, or4

a. With one CRACS inoperable, restore the inoperable system to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months

b. With both CRACS inoperable, immediately enter LCO 3.0.3.*

MODES 5 or 6, or during movement of irradiated fuel assemblies

a. With one CRACS inoperable, restore the inoperable system to OPERABLE status within 30 days or initiate and maintain operation of the OPERABLE CRACS or suspend movement of irradiated fuel assemblies.

b. With both CRACS inoperable, suspend movement of irradiated fuel assemblies.

SURVEILLANCE REQUIREMENTS 4.7.15 Each CRACS shall be demonstrated OPERABLE:

a. At least once per 18 months by verifying each CRACS train has the capability to remove the assumed heat load.

An allowance to monitor control room temperature every four hours and verify less than or equal to 90 degrees Fahrenheit is permitted for up to seven days in lieu of the immediate entry into LCD 3.0.3. If control room temperature exceeds 90 degrees Fahrenheit or the duration without a train of CRACS being OPERABLE exceeds seven days, the immediate entry into LCO 3.0.3 will be required. This provision is only applicable during maintenance activities planned for the upgrade of the CRACS compressors and controls and expires on March 31, 2005.

May 21, 2004 SEQUOYAH - UNIT 1 3/4 7-44 Amendment No. 273, 292

OPT200.CBVENT Rev. 3 Page 3 of 106 PROGRAM: OPERATOR TRAINING II. COIJRSE: SYSTEMS TRAINLNG III. TITLE: CONTROL BUILDING VENTILATION SYSTEM IV. LENGTH OF LESSON: 3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months lecture; 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months simulator demonstration; 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months self-study/workshop V. TRAINING OBJECTWES:

A. Terminal Objective:

Upon completion of this lesson and others presented, the student should be able to apply the knowledge to support satisfactory performance of the tasks associated with the Control Building Ventilation System in the plant and on the simulator.

B. Enabling Objectives:

0. Demonstrate an understanding of NLTREG 1122 knowledge and abilities associated with the Control Building Ventilation System as identified in Appendix A.

1. State the purpose/functions of the Control Building Ventilation System as described in the SQN FSAR.

2. State the design basis of the Control Building Ventilation System in accordance with the SQN FSAR.

3. Explain the purpose/function of each major component in the flow path of the Control Building Ventilation System as illustrated on a simplified system drawing.

4. Describe the following characteristics of each maj or component in the Control Building Ventilation System:

a. Location

b. Power supply (include control power as applicable)

c. Support equipment and systems

d. Normal operating parameters

e. Component operation

f. Controls

g. Interlocks (including setpoints)

h. Instrumentation and Indications

i. Protective features (including setpoints)

j. Failure modes

k. Unit differences

1. Types of accidents for which the components are designed

m. Location of controls and indications in the control room and auxiliary control room

OPT200.CBVENT Rev. 3 Page 4 of 106 V. TRAINING OBJECTIVES (Contd):

B. Enabling Objectives (Contd):

5. Describe the operation of the Control Building Ventilation System:

a. Precautions and limitations

b. Major steps performed while placing the system in service

c. Alarms and alarm response

d. How a component failure will affect system operation

e. How a support system failure will affect system operation

f. How a instrument failure will affect system operation

6. Describe the administrative controls and limits for the Control Building Ventilation System:

a. State Tech Specs/TRM LCOs that govern the system.

b. State the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action limit TS LCOs

c. Given the conditions/status of the Control Building Ventilation System components and the appropriate sections of the Tech Spec, determine if operability requirements are met and what actions are required

7. Discuss related Industry Events VI. TRAINING AIDS:

A. Classroom Computer and Local Area Network (LAN) Access B. Computer projector C. Simulator (if available)

OPT200.CBVENT Rev. 3 Page 28 of 106 X. LESSON BODY B. System Overview EMERGENCY FLOW PATHS

Fresh or pressurizing air is taken from either of two air intakes, one on the Control Building roof at el. 752 near the south end of the building and the other connected into the fresh air intake at the north end of the building.

Each emergency pressurizing fan is duct connected to both the south and north intakes

Induced air flow to the rooms of the lower floors is provided to replace the air exhausted from the battery rooms to maintain these rooms at a negative pressure relative to the Main Control Room.

Filtered outdoor air during operation of the control room emergency air cleanup system maintains at least 1/8-inch w.g. positive pressure in the main control room relative to outside atmosphere and a slightly positive pressure relative to its control building surroundings. Maintenance of a slightly positive pressure in other rooms in the habitability zone relative to adjoining spaces minimizes the in-leakage of unprocessed air.

During control room isolation, a continuous stream of fresh air is drawn in through the electrical board rooms air handling unit to replace that exhausted from each battery room.

Battery room fans are ESF equipment and are powered by Class 1E buses.

OPT200.CBVENT Rev. 3 Page 30 of 106 Design Bases

Worst Case During Summer

Worst Case During Winter

Performance Requirements EO-2 X. LESSON BODY C. Design Bases

Worst case in Summer

Loss of coolant accident (LOCA) on a calm, hot day

Essential raw cooling water (ERCW) heat sink at its highest temperature

Concurrent, with hot summer conditions lasting for the full duration of the emergency.

Worst Case in Winter

LOCA on a calm, cold day

ERCW heat sink at its lowest temperature.

Concurrent, with the weather conditions holding for the full duration of the emergency.

Performance requirements

Capability to maintain an environment within the main control room habitability system area that is in accordance with the requirements specified in Criterion 19 (10 CFR 50, Appendix A).

Maintained for the duration of the emergency even after suffering any single component or subsystem active failure.

EO-2 X. LESSON BODY C. Design Bases CB Vent. System has several design features to meet its design requirements

Full (100%) capacity redundancy to meet single component or subsystem failure.

Capability to keep the MCR at least a 1/8-inch w.g. positive pressure relative to the outside atmosphere, and

a slightly positive pressure relative to its surroundings, and a slightly positive pressure in other rooms in the habitability zone relative to adjoining spaces during control room isolation emergency operating mode, except during a tornado isolation mode

capability for selecting emergency pressurizing air, during accidents, from intakes on opposite ends of the building. cleanest air selection.

Accessibility of the HVACAC System dampers capability for adjusting flow control or isolation dampers in the event one of these does not fail in the intended fail-safe position during an emergency.

All air-conditioning equipment, essential ventilating equipment, isolation dampers, and ducts are designed to withstand the Safe Shutdown Earthquake (SSE).

The refrigerant condensing units in each equipment room are separated by a concrete partition.

Air cleanup units that purify both make-up and recirculated air flows during emergencies.

Each 100% air cleanup unit in the HVACAC System contains a bank of HEPA filters and a bank of carbon adsorbers.

OPT200.CBVENT Rev. 3 Page 71 of 106 CR1 Auto Equipment Response

Control Building Emergency Air Cleanup Fans & Air Pressurization Fans start.

Fresh air isolation dampers FCV 31A-105A, 106A, 105B, and 106B close

Toilet & Locker Room Fan stops & isolates

Spreading Room Supply Fan stops & isolates

Spreading Room Exhaust Fans stop & isolate

Auxiliary Building Shutdown Board Pressurization Fans stop to prevent in-leakage into the MCR Questionj E05 X. LESSON BODY:

E. System Operations

3. Emergency Operations Automatic equipment operations on Control Room Isolation

Control Building Emergency Air Cleanup Fans start.

Control Building Emergency Air Pressurization Fans start

Redundant isolation dampers close to prevent inflow of unfiltered air into the Control Room and Spreading Room from normal path of induced fresh air FCV 31A-105A, 106A, lO5B, and 106B.

Toilet and Locker Room Fan stops and double isolation dampers FCO 3 1A-103 & 104 close

Spreading Room Supply Fan stops and double isolation dampers in Spreading Room Fan supply duct FCO 31-17 & 102 close

Single isolation damper in Spreading Room Exhaust Fans FCO 3 1-79 & 80 close

Auxiliary Building Shutdown Board Pressurization Fans stop to prevent in leakage into the MCR. covered in more detail in Auxiliary Building Ventilation System Lesson & SD.

OPT200.CBVENT Rev. 3 Page 74 of 106 Technical Specifications

3/4.7.7 Control Room Emergency Ventilation System 7 Day Action Statement HEPA filter and Charcoal Adsorber flow tests 31 days

3/4.3.7.15 Control Room Air-Conditioning Systems 30 Day Action Statement 18 month verification of heat load capability EO-6 X. LESSON BODY:

E. Integrated Operations

2. Technical Specifications Instructor Note. Have students use their copies of Technical Specifications.

Cover these sections of Technical Specifications as related to the Control Room HVAC System

3/4.7.7 Control Room Emergency Ventilation System

7 Day Action Statement

HEPA filter and Charcoal Adsorber flow tests 31 days

3/4.3.7.15 Control Room Air-Conditioning Systems

30 Day Action Statement 18 month verification of heat load capability

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta

95. Given the following:

- Unit-i outage schedule logic changes for a planned 28-day refueling outage are being reviewed by an SRO to determine if they are safety significant in accordance with SPP-7.2) Outage Management, Appendix E, Outage Schedule Logic Change Control.

In accordance with Appendix E, which ONE of the following proposed logic changes would meet the criteria for a safety significant change?

Reference Providedj A*, Reschedule the Unit-I loop #2 MSIV seat inspection from the core empty mid-loop period to the time period during core reload.

B. Increase the cavity level from 711 feet to 712 feet elevation to minimize dose while unlatching control rods.

C. Add a 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months activity to perform preventative maintenance on the turbine driven auxiliary feedwater pump before entry into Mode 4.

D. Change the window for a contract diver to enter the CCW pump intake bay from the end of the outage (before starting CCW pumps) to the beginning of the outage (after securing CCW pumps.)

DIS TRA CTOR ANAL YSIS:

A. Correct. MSIV seat inspection could impact the ability to isolate the contaiment if needed. SPP-7.2 Appendix E B 14.

B. Incorrect. Plausible because cavity level changes do pose a safety significant risk when the level is lowered.

C. Incorrect. Plausible because the TDAFW pump is a significant safety system while operating but is not available until secondary steam pressure is developed.

D. Incorrect. Though this activity is identified as a High Risk when reviewed for (non-outage) SPP-7.3 Work Activity Risk Management Process it does not fall within any criteria the SRO is to evaluate per Appendix E of SPP-7.2.

Thursday, July 15, 2010 11:30:26 AM 95

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta Question Number: 95 Tier: 3 Group n/a KIA: G 2.2.18 Knowledge of the process for managing maintenance activities during shutdown operations, such as risk assessments, work prioritization, etc.

Importance Rating: 2.6 I 3.9 IOCFRPart55: 41.10/43.5/45.13 IOCFR55.43.b: 5 KIA Match: KA is matched because the question requires the candidate to evaluate activities and determine those that require increased measures due to risk. SRO because this is a Site Specific SRO task.

Technical

Reference:

SPP-7.2, Outage Management, Rev. 0020 SPP-7.3, Work Activity Risk Management Process, Rev 6

Proposed references SPP-7.2 Appendix E to be provided:

Learning Objective: OPL27ISPP-7.0

10. Describe the plant activities that require a risk assessment Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: Written 3-18-2010 from K/A. LWP Comments:

MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9 Answer: ACCDBDCADD ScrambleRange:A-D Thursday, July 15, 2010 11:30:26AM 96

NPG Standard Outage Management SPP-7.2 Programs and Rev. 0020 Processes Page 35 of 45 Appendix E (Page 1 of 2)

Outage Schedule Logic Change Control Subsequent to the Independent Outage Safety Review (IOSR) or issuance of Rev. C if an IOSR was not performed, all significant outage schedule changes which involve the modification of logic relation between outage activities will be controlled by the following process.

A. The proposed schedule change is documented on the Outage Schedule Logic Change Form (OSLCF) SPP-7.2-2, Attachment 1 of this procedure.

An OSLCF is initiated by the responsible individual each time a logic change to the outage schedule is required. The initiator must complete items A through D of the OSLCF.

B. A licensed SRO reviews the requested logic change documented on the OSLCF and evaluates whether the change to the outage schedule should be approved and proceed. This review by a SRO is to ensure safety significant changes receive a review consistent with the initial independent outage safety review of the outage schedule. The SRO should consider the criteria below when determining if the change should be approved: If the answer to any of these questions is yes, the reviewer should ensure appropriate actions are taken to minimize the safety risk resulting from incorporating the logic change prior to approval. If the logic change requires a major system window change, or could present a challenge to the defense in depth strategy with safety significance, the Unit Outage Manager will be consulted to convene an Independent Outage Safety Review team to provide additional expert opinions on whether the change should be approved. The plant manager, or designee, will approve a schedule change, deemed safety significant via the following questions.

SRO review questions for consideration:

1. Perturb the stability of RHR parameters (flow, pressure, temperature, etc.) and other operational parameters (operating pump amperes, etc.)?

2. Alter plant configurations that would result in RCS temperature below the minimum value used to analyze reactor shutdown margin?

3. Reduce the reactor cavity or reactor vessel inventory?

4. Reduce the spent fuel pit inventory or challenge SF Cooling redundancy?

5. Reduce the availability of systems or support systems required to provide reactor vessel makeup water consistent with the decay heat generation load?

6. Minimize the availability of low pressure injection?

7. Reduce the availability of alternate sources of reactor vessel makeup water consistent with the decay heat generation rate?

8. Increase the probability of jeopardizing installed temporary equipment that could reduce safety system availability?

9. Isolate of the operable boration flow path?

NPG Standard Outage Management SPP-7.2 Programs and Rev. 0020 Processes Page 36 of 45 Appendix E (Page 2 of 2)

Outage Schedule Logic Change Control

10. Cause leakage of water into the RCS or spentfuel pit which would dilute the boron concentration to a value below the minimum required?

11. Affect bus outages or switchyard outages?

12. Reduce the availability of onsite or offsite electrical power supplies or support systems?

13. Increase the probability of fuel or other core component mishandling or damage?

14. Reduce the ability to isolate containment when required?

15. Affect the non-outage unit?

After reviewing the above questions, the SRO signs the OSLCF for approval if warranted. If the SRO rejects the change, the OSLCF is returned to the originator with an explanation of why it was rejected for enhancement or cancellation as appropriate.

C. Following SRO approval of an OSLCF, the Outage Manager reviews the OSLCF for completeness and determines if additional reviews are required. This determination is based on the following criteria:

1. Logic changes for work activities within a system window which change the sequence of scheduled work and do not pose a potential challenge to the defense in depth strategy can be approved by the Outage Manager.

2. Logic changes which move a work activity scheduled in Modes 5 or 6 (PWR only) to the empty reactor vessel period can be approved by the Outage Manager.

3. Logic changes for work activities on the equipment and systems affecting system operations require review and concurrence by the Operations Management representative in the 0CC, in addition to the original SROs approval. As an example: jE the logic change moves the activity out of its scheduled work window OR has the potential to challenge the stations defense in depth strategy, Operations management concurrence in 0CC is required, as well as Outage Manager and Plant Manager, or outage shift designees.

4. Logic changes for work activities on equipment and systems not identified above can be approved by the Outage Manager.

D. Following approval of the OSLCF, the Outage Manager directs changing the outage schedule in accordance with the OSLCF. Approved OSLCFs are retained by the Outage Manager. OSLCFs that are not approved are returned to the originator for cancellation or further processing.

E. Copies of approved OSLCFs should be distributed to the Outage Manager and to other outage participants that will be affected by the change as deemed necessary.

NPG Standard Outage Management SPP-72 Programs and Rev. 0020 Processes Page 35 of 45 Appendix E (Page 1 of 2)

Outage Schedule Logic Change Control Subsequent to the Independent Outage Safety Review (IOSR) or issuance of Rev. C if an IOSR was not performed, all significant outage schedule changes which involve the modification of logic relation between outage activities will be controlled by the following process.

A. The proposed schedule change is documented on the Outage Schedule Logic Change Form (OSLCF) SPP-7.2-2, Attachment I of this procedure.

An OSLCF is initiated by the responsible individual each time a logic change to the outage schedule is required. The initiator must complete items A through D of the OSLCF.

B. A licensed SRO reviews the requested logic change documented on the OSLCF and evaluates whether the change to the outage schedule should be approved and proceed. This review by a SRO is to ensure safety significant changes receive a review consistent with the initial independent outage safety review of the outage schedule. The SRO should consider the criteria below when determining if the change should be approved: If the answer to any of these questions is yes, the reviewer should ensure appropriate actions are taken to minimize the safety risk resulting from incorporating the logic change prior to approval. If the logic change requires a major system window change, or could present a challenge to the defense in depth strategy with safety significance, the Unit Outage Manager will be consulted to convene an Independent Outage Safety Review team to provide additional expert opinions on whether the change should be approved. The plant manager, or designee, will approve a schedule change, deemed safety significant via the following questions.

SRO review questions for consideration:

1. Perturb the stability of RHR parameters (flow, pressure, temperature, etc.) and other operational parameters (operating pump amperes, etc.)?

2. Alter plant configurations that would result in RCS temperature below the minimum value used to analyze reactor shutdown margin?

3. Reduce the reactor cavity or reactor vessel inventory?

4. Reduce the spent fuel pit inventory or challenge SF Cooling redundancy?

5. Reduce the availability of systems or support systems required to provide reactor vessel makeup water consistent with the decay heat generation load?

6. Minimize the availability of low pressure injection?

7. Reduce the availability of alternate sources of reactor vessel makeup water consistent with the decay heat generation rate?

8. Increase the probability of jeopardizing installed temporary equipment that could reduce safety system availability?

9. Isolate of the operable boration flow path?

NPG Standard Outage Management SPP-7.2 Programs and Rev. 0020 Processes Page 36 of 45 Appendix E (Page 2 of 2)

Outage Schedule Logic Change Control

10. Cause leakage of water into the RCS or spent fuel pit which would dilute the boron concentration to a value below the minimum required?

11. Affect bus outages or switchyard outages?

12. Reduce the availability of onsite or offsite electrical power supplies or support systems?

13. Increase the probability of fuel or other core component mishandling or damage?

14. Reduce the ability to isolate containment when required?

15. Affect the non-outage unit?

After reviewing the above questions, the SRO signs the OSLCF for approval if warranted. If the SRO rejects the change, the OSLCF is returned to the originator with an explanation of why it was rejected for enhancement or cancellation as appropriate.

C. Following SRO approval of an OSLCF, the Outage Manager reviews the OSLCF for completeness and determines if additional reviews are required. This determination is based on the following criteria:

1. Logic changes for work activities within a system window which change the sequence of scheduled work and do not pose a potential challenge to the defense in depth strategy can be approved by the Outage Manager.

2. Logic changes which move a work activity scheduled in Modes 5 or 6 (PWR only) to the empty reactor vessel period can be approved by the Outage Manager.

3. Logic changes for work activities on the equipment and systems affecting system operations require review and concurrence by the Operations Management representative in the 0CC, in addition to the original SROs approval. As an example: E the logic change moYes the activity out of its scheduled work window OR has the potential to challenge the stations defense in depth strategy, Operations management concurrence in CCC is required, as well as Outage Manager and Plant Manager, or outage shift designees.

4. Logic changes for work activities on equipment and systems not identified above can be approved by the Outage Manager.

D. Following approval of the OSLCF, the Outage Manager directs changing the outage schedule in accordance with the OSLCF. Approved OSLCFs are retained by the Outage Manager. OSLCFs that are not approved are returned to the originator for cancellation or further processing.

E. Copies of approved OSLCFs should be distributed to the Outage Manager and to other outage participants that will be affected by the change as deemed necessary.

NPG Standard Work Activity Risk SPP-7.3 Programs and Management Process Rev. 0006 Processes Page 25 of 34 Attachment 2 (Page 1 of 6)

Initial Risk Characterization INITIAL RISK CHARACTERIZATION Initial Plant Specific Safety Risk Characterization (Unless specified applicable to all TVA sites)

If the work involves: Then the risk is characterized as:

All activities associated with freeze seals, due to safety impacts for a

installing freeze seals and industry operating experience All activities associated with or requiring entry into a Confined Space b

Permit required area All activities associated with Diving Operations, due to industry c

operating experience H IG H All Lifting or Rigging activities characterized as a High Hazard Lift,

d. as defined by MMTP-103, Nuclear Power Group Movement of Items Using Overhead Handling Equipment All work that will breech (actively intruding on any pressure retaining e.* parts of the component or system) an un-isolable high energy component or system (>500 psig or >200 F), or a chemically hazardous system.

All work which could challenge condensate, feed water and feed/condensate control systems_operating_margin_at_power.

Component manipulation and placement of clearance tags are not required to be High Risk per this paragraph.

Routine chemical additional/maintenance, such as manipulation of Boron systems during operations may not be considered chemically hazardous.

Routine Condensate Demineralizer backwash / single vessel repairs would not apply.

0PL271 SPP-7.0 Revision No.: 2 Page 3 of 12 PROGRAM: OPERATOR TRAINING - LICENSED COURSE: INITIAL LICENSE TRAINING III. LESSON TITLE: SPP-7.0, WORK MANAGEMENT IV. LENGTH OF LESSONICOURSE: -3 hour V. TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of SPP-7.0, WORK MANAGEMENT.

B. Enabling Objectives

0. Demonstrate an understanding of NUREG 1122 Knowledges and Abilities associated wjth SPP-7.0, WORK MANAGEMENT that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A.

1. State the purpose of this procedure.

2. Describe how the accuracy and adequacy of work documents is assured.

3. Describe the purpose of the Outage Management Organization.

4. Describe the methods used to monitor progress toward meeting pre-outage milestones

5. Describe the coordination responsibilities in planning for a forced outage.

6. Describe the elements included in the process of work management

7. Describe how a safety-conscious culture is promoted for maintenance work

8. Describe the work management philosophy for ownership of maintenance work.

9. Describe the work management philosophy for ownership of emergent work.

10. Describe the plant activities that require a risk assessment

11. State the targets times for pre-outage activities.

12. Describe behaviors of excellence in work management for leaders and workers

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta

96. Given the following:

- Unit I is in Mode 5 with 1 B RHR pump tagged for motor replacement.

- 1A RHR pump is in service and aligned for shutdown cooling.

- RCPs #1 and #2 are in service with RCPs #3 and #4 available if needed.

- All SG levels are between 16% and 20% narrow range with AFW flow in manual.

Which ONE of the following identifies the required action, if any, in accordance with Technical Specifications related to entry into Mode 4?

A. Mode 4 entry is allowed provided that alignment of the 1A RHR Pump in the ECCS cold leg injection Mode is made within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

B Mode 4 entry is allowed provided that the IA RHR pump is capable of being manually aligned to the ECCS mode of operation if needed.

C. Prior to entering Mode 4, one Steam Generator level must be raised to greater than 21% NR in a loop with an RCP running.

D. Prior to entering Mode 4, Risk Assessment must be completed and risk management actions established in accordance with LCO 3.0.4.b.

DIS TRACTOR ANALYSIS:

A. Incorrect, Plausible because realignment within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is a low temperature overpressure protection LCO requirement for ECCS pumps when changing modes.

(LCO 3.4.12)

B. Correct, LCO 3.5.3 requires one RHR pump to be operable for ECCS operation and states that the RHR pump can be aligned for decay heat removal provided it is capable of being manually aligned to the ECCS mode of operation during Mode 4 operation.

C. Incorrect, Plausible because steam generator level are required to be above 21% in the Mode 3 ECCS Tech Spec. (LCO 3.4.1.2)

D. Incorrect, Plausible because use of LCO 3. O.4.b is a mechanism that can be used to change modes when the LCO would not be met and the actions have a specified time for inoperability.

Question Number: 96 Tier: 3 Group n/a KIA: G 2.2.40 Equipment Control Ability to apply Technical Specifications for a system.

Thursday, July 15, 2010 11:06:03AM 96

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta Importance Rating: 3.4 / 4.7 10 CFR Part 55: 41.10/43.2/43.5/45.3 IOCFR55.43.b: 2 K/A Match: KA is matched because the question requires the knowledge of the required actions in Technical Specifications and how the actions would be applied to allow a Mode change. SRO because of the allowances in Tech Spec for considering equipment operable and actions required for changing Modes with equipment is an alignment different than as stated in the LCO statement.

Technical

Reference:

Technical Specifications LCOs:

3.0.4 Amendment 312 3.4.1.2 Amendment 285 3.4.12 Amendment 297 3.5.3 Amendment 326 Proposed references None to be provided:

Learning Objective: 6. Describe the administrative controls and limits for the ECCS:

a. State Tech Specs/TRM LCOs that govern the LCCS

b. State the =1 hour action limit TS LCOs

c. Given the conditions/status of the ECCS components and the appropriate sections of the Tech Spec, determine if operability requirements are met and what actions are required Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question Comments:

MCS Time: I Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9 Answer: BBBCACDBDC ScrambleRange:A-D Thursday, July 15, 2010 11:06:03AM 97

REACTOR COOLANT SYSTEM SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 a. At least two of the reactor coolant and/or residual heat removal (RHR) loops listed below shall be OPERABLE:

1. Reactor Coolant Loop A and its associated steam generator and reactor coolant pump,*

2. Reactor Coolant Loop B and its associated steam generator and reactor coolant pump,*

3. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump,*

4. Reactor Coolant Loop D and its associated steam generator and reactor coolant pump,*

5. Residual Heat Removal Loop A,

6. Residual Heat Removal Loop B.

b. At least one of the above reactor coolant and/or RHR loops shall be in operation. **

APPLICABILITY: MODE 4.

ACTION:

a. With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status, as soon as possible; be in COLD SHUTDOWN within 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b. With no reactor coolant or RHR loop in operation, suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1.1 and immediately initiate corrective action to return the required coolant loop to operation.

- AII reactor coolant pumps and residual heat removal pumps may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided 1) no operations are permitted that would cause introduction of coolant into the RCS with boron concentration less than reqiired to meet SDM of LCO 3.1.1.1, and 2) core outlet temperature is maintained at least 10 F below saturation temperature.

A reactor coolant pump shall not be restarted unless a steam bubble exists in the pressurizer.

V May 22, 2003 SEQUOYAH - UNIT 1 3/4 4-2 Amendment Nos. 12, 157, 285

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS shall be determined to be 4.4.1.3.1 The required reactor coolant pump(s), if not in operation, ents and indicated power OPERABLE once per 7 days by verifying correct breaker alignm availability.

ABLE by verifying 4.4.1.3.2 The required steam generator(s) shall be determined OPER t (wide- range indication) at secondary side water level to be greater than or equal to 10 percen least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

in operation and 4.4.1.3.3 At least one reactor coolant or RHR loop shall be verified to be circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

March 25, 1982 UNIT 1 3/4 4-2a Amendment No. 12 SEQUOYAH -

EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.3 ECCS -SHUTDOWN LIMITING CONDITION FOR OPERATION 3.5.3 One ECCS train shall be OPERABLE.

NOTE An RHR train may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned to the ECCS mode of operation.

APPLICABILITY: MODE 4.

ACTION:

--NOTE

1. LCO 3.O.4b is not applicable to ECCS centrifugal charging subsystem.

2. The required ECCS residual heat removal (RHR) subsystem may be inoperable for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for surveillance testing of valves provided that alternate heat removal methods are available via the steam generators to maintain reactor coolant system Tavg less than 350 F and provided that the 3

required subsystem is capable of being manually realigned to the ECCS mode of operation.

a. With the required ECCS residual heat removal (RHR) subsystem inoperable, immediately initiate action to restore required ECCS RHR subsystem to OPERABLE status.

b. With the required ECCS centrifugal charging subsystem inoperable, within one hour, restore required ECCS centrifugal charging subsystem to OPERABLE status, or be in COLD SHUTDOWN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.5.3 The ECCS train shall be demonstrated OPERABLE per the following applicable Surveillance Requirements of 4.5.2:

SR 4.5.2.b.1 SR 4.5.2.d SR 4.5.2.f SR 4.5.2.g January 28, 2010 SEQUOYAH - UNIT 1 3/4 5-5 Amendment No. 36, 140, 276, 301, 326

REACTOR COOLANT SYSTEM SYSTEM 3/4.4.12 LOW TEMPERATURE OVER PRESSURE PROTECTION (LTOP)

LIMITING CONDITION FOR OPERATION 3.4.12* An LTOP System shall be OPERABLE with a maximum of one centrifugal charging pump isolated capable of injecting into the Reactor Coolant System (RCS) and the accumulators and one of the following pressu re relief capabilities:

specified in

a. Two power operated reliet valves (PORVs) with lift settings within the limits the PTLR, or

b. The RCS depressurized and an RCS vent 3 square inches.

APPLICABILITY: MODE 4 when any RCS cold leg temperature is the LTOP arming temperature specified in the PTLR, MODE 5, MODE 6 when the reactor vessel head is on.

ACTION:

capable of injecting

a. Should any safety injection pump or more than one charging pump be found charging pump is into the RCS, immediately initiate action to verify a maximum of one centrifugal capable of injecting into the RCS.

than or equal to the

b. With an accumulator not isolated when the accumulator pressure is greater isolate the maximum RCS pressure for existing cold leg temper ature allowed in the PTLR, affected accumulator within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , or either; in the

1. Increase RCS cold leg temperature to >the LTOP arming temperature specified PTLR within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , or re for 2 Depressurize the affected accumulator to less than the maximum RCS pressu existing cold leg temperature allowed in the PTLR within 12 hours.

c. With one required PORV inoperable in MODE 4, restore the required PORV to OPERABLE status within 7 days.

OPERABLE

d. With one required PORV inoperable in MODE 5 or 6, restore the required PORV to status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

1) Two charging pumps may be made capable of injecting into the RCS for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for pump swap operations.

RCS

2) Accumulator may be unisolateci when accumulator pressure is less than the maximum in by the PIT limit curves provided pressure for the existing RCS cold leg temperature allowed the PTLR.

inoperable,

3) For the purpose of making the required safety injection pumps and charging pump up to 4 hours after entering MODE 4 from MODE 3, or prior to the following time is permitted:

decreasing temperature on any RCS loop to below 325°F, whichever occurs first.

November 9, 2004 SEOUOYAH UNIT 1 -

3/4 4-29 Amendment No. 157, 213, 294, 297

REACTOR COOLANT SYSTEM ACTION (Continued)

e. With two required PORVs inoperable, or the Actions (a), (b), (c) or (d) not met, or the LTOP System inoperable for any reason other than (a), (b), (c), or (d), depressurize the RCS and establish RCS vent of 3.0 square inches within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

f. LCO 3.0.4b is not applicable when entering MODE 4.

SURVEILLANCE REQUIREMENTS 4.4.12.1 Each PORV shall be demonstrated OPERABLE by:

a. Performance of a CHANNEL FUNCTIONAL TEST*, but excluding valve operation,at least once per 31 days;

b. Performance of a CHANNEL CALIBRATION on each required PORV actuation channel at least once per 18 months; and

c. Verifying the PORV block valve is open for each required PORV at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

4.4.12.2 Verify no safety injection pumps are capable of injecting into the RCS within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entering MODE 4 from MODE 3 prior to the temperature of one or more RCS cold legs decreasing below 325F, and every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

4.4.12.3 Verify a maximum of one charging pump is capable of injecting into the RCS within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entering MODE 4 from MODE 3 prior to the temperature of one or more RCS cold legs decreasing below 325F, and every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

4.4.12.4 Verify each accumulator is isolated at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.12.5 Verify# required RCS vent 3.0 square inches open at least:

a. Once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for unlocked open vent valve(s) and,

b. Once every 31 days for other vent path(s).

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after decreasing RCS cold leg temperatures to c the LTOP arming temperature in the PTLR.

Only required to be met when complying with LCO 3.412.b.

Aprilll,2005 SEQUOYAH - UNIT 1 314 4-30 Amendment No. 157, 213, 294, 297, 301

OPT200.ECCS Rev. 3 Page 3 of 124 PROGRAM: OPERATOR TRAINING II. COURSE: SYSTEMS TRAINING III. TITLE: EMERGENCY CORE COOLING SYSTEMS IV. LENGTH OF LESSON: Initial LicenseTraining: 5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months lecture; 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months simulator demonstration; 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months self-study/workshop V. TRAINiNG OBJECTIVES:

A. Terminal Objective:

Upon completion of this lesson and others presented, the student should be able to apply the knowledge to support satisfactory performance of the tasks associated with the Emergency Core Cooling Systems (ECCS) in the plant and on the simulator.

B. Enabling Objectives:

0. Demonstrate an understanding of NUREG 1122 knowledge and abilities associated with the ECCS that are rated 2.5 during Initial License training for the appropriate license position as identified in Appendix A.

1. State the purpose/functions of the ECCS as described in the SQN FSAR.

2. State the design basis of the ECCS in accordance with the SQN FSAR.

3. Explain the purpose/function of each major component in the flow path of the ECCS as illustrated on the simplified system drawing.

4. Describe the following items for each major component in the ECCS:

a. Location

b. Power supply (include control power as applicable)

c. Support equipment and systems

d. Normal operating parameters

e. Component operation

f. Controls

g. Interlocks (including setpoints)

h. Instrumentation and Indications

i. Protective features (including setpoints)

j. Failure modes

k. Unit differences

1. Types of accidents for which the ECCS components are designed

m. Location of controls and indications associated with the ECCS in the control room and auxiliary control room

OPT200.ECCS Rev. 3 Page 4 of 124 V. TRAINING OBJECTIVES (Contd):

B. Enabling Objectives (Contd):

5. Describe the operation of the ECCS as it relates to the following:

a. Precautions and limitations

b. Major steps performed while placing the ECCS in service

c. Alarms and alarm response

d. How a component failure will affect system operation

e. How a support system failure will affect ECCS operation

f. How a instrument failure will affect system operation

6. Describe the administrative controls and limits for the ECCS:

a. State Tech Specs/TRM LCOs that govern the ECCS

b. State the l hour action limit TS LCOs

c. Given the conditions/status of the ECCS components and the appropriate sections of the Tech Spec, determine if operability requirements are met and what actions are required

7. Discuss related Industry Events:

a. PER 93856, Lessons Learned (operability impact of closing FCV-63-47 or -

48)

b. Closure of Suction Valve for a Safety Injection Pump Places Unit 2 in Limiting Condition for Operation 3.0.3

c. PER 104032, Temperature Control Valve (TCV) stroke testing for ECCS Pump Room Coolers.

VI. TRAINING AIDS:

A. Classroom Computer and Local Area Network (LAN) Access B. Computer projector

REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 At least two of the reactor coolant loops listed below shall be OPERABLE with at least two reactor coolant loops in operation when the Reactor Trip System breakers are closed and at least one reactor coolant loop in operation when the Reactor Trip System breakers are open:*

a. Reactor Coolant Loop A and its associated steam generator and reactor coolant pump,

b. Reactor Coolant Loop B and its associated steam generator and reactor coolant pump,

c. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump,

d. Reactor Coolant Loop D and its associated steam generator and reactor coolant pump, APPLICABILITY: MODE 3 ACTION:

a. With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With only one reactor coolant loop in operation and the Reactor Trip System breakers in the closed position, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months open the Reactor Trip System breakers.

c. With no reactor coolant loop in operation, suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1.1 and immediately initiate corrective action to return the required coolant loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4.1.2.2 The required steam generators shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 21 percent at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.2.3 The required Reactor Coolant loops shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

All reactor coolant pumps may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided (1) no operations are permitted that would cause introduction of coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1.1, and (2) core outlet temperature is maintained at least 10°F below saturation temperature.

May 22, 2003 SEQUOYAH - UNIT 1 3/4 4-la Amendment No. 12, 84, 285

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta

97. In accordance with FHI-3, Movement of Fuel, which ONE of the following identifies the maximum number of NEW and IRRADIATED fuel assemblies within the areas listed below that can be located out of approved storage locations?

New Fuel Assemblies Irradiated Fuel Assemblies within the fuel-handling area within the refueling canal A. 1 2 B 1 3 C. 2 2 D. 2 3 DIS TRACTOR ANAL YSIS:

A. Incorrect, FHI-3 Limitation B allows one unirradiated nuclear fuel assembly within the fuel-handling area and three (not 2) nuclear fuel assemblies within the refueling canal to be out of approved storage locations.

B. Correct, FHI-3 Limitation B allows one unirradiated nuclear fuel assembly within the fuel-handling area and three nuclear fuel assemblies within the refueling canal to be out of approved storage locations.

C. Incorrect, FHI-3 Limitation B allows one (not 2) unirradiated nuclear fuel assembly within the fuel-handling area and three (not 2) nuclear fuel assemblies within the refueling canal to be out of approved storage locations.

D. Incorrect, FHI-3 Limitation B allows one (not 2) unirradiated nuclear fuel assembly within the fuel-handling area and 3 nuclear fuel assemblies within the refueling canal to be out of approved storage locations.

Thursday, July 15, 2010 10:48:44 AM 97

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta Question Number: 97 Tier: 3 Group n/a KIA: G2.3.12 Radiation Control Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Importance Rating: 3.2 / 3.7 IOCFRPart55: 41.12/45.9/45.10 IOCFR55.43.b: 4, 7 KIA Match:

Technical

Reference:

FHI-3, Movement of Fuel, Rev 58 Proposed references None to be provided:

Learning Objective: OPT200. FH 8.5 Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: SQN bank question used on SRO Audit Exam 1/2009 Comments:

MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9 Answer: BBBBBBBBBB ItemsNotScrambled Thursday, July 15, 2010 10:48:44 AM 98

SQN FHI-3 MOVEMENT OF FUEL Rev: 58 1,2 Page 17of 101 III. LIMITATIONS A. The hoist slow speed shall be used prior to entering a confined area (such as a fuel rack, MPC fuel basket or upender) when bottom fuel nozzle of fuel assembly is approximately 10 inches above the confined area and continue in slow speed until bottom fuel nozzle is inside the confined area. The hoist slow speed shall also be used prior to bottom of fuel assembly reaching bottom of the confined area, (when bottom of fuel assembly is approximately 10 inches from bottom) until fuel assembly reaches the bottom.

B. The maximum number of nuclear fuel assemblies allowed out of approved storage locations for Sequoyah Nuclear Plant shall be as listed below: [C5]

1. One unirradiated nuclear fuel assembly shall be allowed within the fuel-handling area, outside of metal shipping containers, or the new fuel storage vault.

2. One nuclear fuel assembly shall be allowed within the spent fuel storage pool boundary when not seated in a storage cell or MPC fuel basket (excluding the inspection, reconstitution or cleaning locations with appropriate evaluation for each configuration that must be performed prior to implementation). The spent fuel storage pool boundary includes the cask pit pool and fuel transfer canal excluding the upender.

3. Three nuclear fuel assemblies shall be allowed within the refueling canal.

The refueling canal includes the fuel transfer tube boundary, the rod cluster control changing fixture and the upender. This allows for two nuclear fuel assemblies to be in the rod cluster control changing fixture while the third nuclear fuel assembly is being transferred through the fuel transfer tube, is in the upender, or is in transit to or from the reactor cavity.

4. One fuel assembly shall be allowed within the reactor cavity.

C. DO NOT allow the RCS and Spent Fuel Pit temperature to decrease below 50 degrees, which is the bound of the criticality analysis. [C.9]

D. Transportation of loads over the Cask Loading Area while spent fuel is being stored in the Cask Loading Pit must be within the guidelines provided in the FSAR.

E. Any assembly identified as leaking or potentially leaking by performance of In-Mast Sipping shall be raised or lowered only at slow speed until the assembly is confirmed as not having a rupture large enough to allow fuel pellets or fragments to be dislodged.

F. Radcon approval is needed when fuel movement is required with either the cask loading pit or the transfer canal empty.

OPT200.FH Rev. 4 Page 3 of 62 PROGRAM: OPERATOR TRAINING II. COURSE: SYSTEMS TRAINING III. TITLE: FUEL HANDLING SYSTEM IV. LENGTH OF LESSON: 10 hour1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months lecture; 10 hour1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months self-study/workshop V. TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of this lesson and others presented, the student should able to apply the knowledge to support satisfactory performance of the tasks associated with the Fuel Handling System in the plant.

B. Enabling Objectives:

0. Demonstrate an understanding of NTiREG 1122 knowledges and abilities associated with the Fuel Handling System that are rated 2.5 during Initial License training for the appropriate license position as identified in Appendix A.

1. State the purpose/functions of the Fuel Handling System as described in the SQN FSAR.

2. State the design basis of the Fuel Handling System in accordance with the SQN FSAR.

3. Explain the purpose/function of each major component Fuel Handling System as illustrated on the simplified system drawing.

4. Describe the following items for each major component in the Fuel Handling System as described in this lesson:

a. Location

b. Power supply (include control power as applicable)

c. Support equipment and systems

d. Normal operating parameters

e. Component operation

f. Controls

g. Interlocks and by-passes

h. Instrumentation and Indications

i. Protective features (including setpoints)

j. Failure modes

k. Unit differences

1. Types of accidents for which the Fuel Handling System components are designed

m. Location of controls and indications associated with the Fuel Handling System.

OPT200.FH Rev. 4 Page 4 of 62 V. TRAINING OBJECTIVES (Contd):

B. Learning Objectives (Contd):

5. Describe the operation of the Fuel Handling system as it relates to the following:

a. Precautions and limitations

b. Major steps performed while refueling.

c. Alarms and alarm response

d. How a component failure will affect system operation

e. How a support system failure will affect Fuel Handling system operation

6. Describe the administrative controls and limits for the Fuel Handling system as explained in this lesson:

a. State Tech Specs/TRM LCOs that govern the Fuel Handling Systems.

b. State the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action limit TS LCOs

c. Given the conditions/status of the Fuel Handling system components and the appropriate sections of the Tech Spec, determine if operability requirements are met and what actions are required

7. Discuss related Industry Events:

a. SQN LER93016 Tilted Fuel Assembly

b. SQN-LER 2-93-3 Equipment Hatch not closed during Fuel Movement

c. SQN-NOV 94-11 Non-conservative Fuel Handling Practices

d. SOER 85-0 1 Reactor Cavity Seal Failure, Connecticut Yankee

e. OE81 12 Movement of irradiated fuel with Ventilation system inop, Dresden 2

f. SOER 94-2 Boron dilution Events in PWRs VI. TRAINING AIDS:

A. Classroom Computer and Local Area Network (LAN) Access B. Computer projector

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta

98. Given the following conditions:

- Both Units operating at 100% power.

- 0600 Due to extremely heavy rainfall, RSOIKEOC issues a Stage I flood warning.

In accordance with AOP-N.03, Flooding, 11 which ONE of the following identifies:

(1) the time the Stage I flood mode actions are required to be completed, and (2) if Stage II actions are required, the strategy used to prepare the Tritiated Drain Collector Tank to prevent a possible release of radioactivity?

Time Tritiated Drain Collector Tank A. 1600 Pressurized to greater than 23 psig 1600 Filled with water C. 2300 Pressurized to greater than 23 psig D. 2300 Filled with water Thursday, July 15, 2010 10:35:03 AM 98

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta DISTRACTOR ANAL YSIS:

After entering AOP-N.03, Flooding Stage I preparations must be implemented and completed within the following 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months and if Stage II actions are required they must be completed within the following 17 hours1.967593e-4 days 0.00472 hours 2.810847e-5 weeks 6.4685e-6 months .

A. Incorrect, The time requirement is correct. However, Stage II actions do require the Tritiated Drain Collector Tank to be filled with water to prevent the tank from floating away and becoming a radiation hazard. Plausible because the time requirement is correct and pressurizing the tank to greater than 23 psig is correct for other tanks during Stage II preparations.

B. Correct, The time requirement is correct and the requirement is to fill Tritiated Drain Collector Tank to be filled with water to prevent the tank from floating away and becoming a radiation hazard.

C. Incorrect, The time requirement is not correct and pressurizing the tank to greater than 23 psig is not correct. Plausible because the time identified is 17 hours1.967593e-4 days 0.00472 hours 2.810847e-5 weeks 6.4685e-6 months which is the time required to complete Phase II actions after Phase II is initiate and pressurizing the tank to greater than 23 psig is correct for other tanks during Stage II preparations.

D. Incorrect, The time requirement is not correct but filling the tank with water is correct. Plausible because the time identified is 17 hours1.967593e-4 days 0.00472 hours 2.810847e-5 weeks 6.4685e-6 months which is the time required to complete Phase II actions after Phase Ills initiate and filling the tank with water is correct.

Thursday, July 15, 2010 10:35:03 AM 99

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta Question Number: 98 Tier: 3 Group n/a KIA: G 2.3.14 Radiation Control Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

Importance Rating: 3.4 / 3.8 10 CFR Part 55:

IOCFR55.43.b: 2, 4, 5 K/A Match: Requires the applicant to know the facility Technical Requirements, implementation of AOP sections and be knowledgeable of provisions to prevent radiation hazards that may arise during normal, abnormal and emergency conditions.

Technical

Reference:

AOP-N.03, Flooding, Rev 28 TRM 4.7.6, Flood Protection Proposed references None to be provided:

Learning Objective: OPL271AOP-N.03 B.4,5, & 8b

4. Upon entry into AOP-N.03, diagnose the applicable condition and transition to the appropriate section for response.

5. Describe the bases for all limits, notes cautions and steps of AOPN.03.

8.b Given a set of initial plant conditions use to correctly (b) Identify required actions.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History:

Comments:

MCS Time: 3 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9 Answer: B B B B B B B B B B Items Not Scrambled Thursday, July 15, 2010 10:35:03 AM 100

FLOODING AOP-N.03 SQN Rev. 30 STEP I ACTIONIEXPECTED RESPONSE I RESPONSE NOT OBTAINED 23 Stagel-Model,2,3,or4 NOTE Stage I preparation completion is required within 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months .

PERFORM Section 2.9, Stage I Preparation, concurrently with this section.

2. CHECK unit in Mode 3 or Mode 4. INITIATE plant shutdown USING one of the following:

AOP-C.03, Rapid Shutdown or Load Reduction OR

0-GO-6, Power Reduction from 30%

Reactor Power to Hot Standby.

3. ENSURE the following plant conditions established USING 0-GO-7, Unit Shutdown From Hot Standby to Cold Shutdown:

RCS temperature less than required value for placing RHR in service.

RCS pressure less than or equal to 350 psig.

Page 15 of 215

SQN FLOODING AOP-N.03 Rev. 30 I STEP ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED 2.9 Stage I Preparation NOTE I Stage I preparation completion is required within 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months .

NOTE 2 This section may be concurrently performed as directed by sections 2.1, 2.2, 2.3, or 2.4. Steps may be performed out of sequence.

1. EVALUATE personnel resources NOTIFY additional personnel to report adequate USING Appendix D, Flood as required to support Stage I preparation.

Mode Preparation Manpower Requirements.

2. DISPATCH personnel to move parts and supplies above probable maximum flood level [720.1 elev] USING Appendix A, Equipment, Supplies, and Parts To Be Moved Above Probable Maximum Flood Level.

Page 100 of 215

SQN FLOODING AOP-N03 Rev. 30 STEP I ACTIONIEXPECTED RESPONSE I RESPONSE NOT OBTAINED 2.10 Stage II Preparation (contd)

2. (Continued)

O-FP-MXX-000-009.O, Flood Preparation Auxiliary Charging System spool pieces

O-FP-MXX-026-003.O, Flood Preparation High Pressure Fire Protection System Installation of Spectacle Orifices in Unit 1 & 2 Aux Bldg Supply Headers

3. NOTIFY Maintenance to install bridge on 714 elev to SFPC and RCP TBB pumps platform USING O-FP-MXX-000-005.O, Flood Preparation -

Access to Spent Fuel Pit Cooling Pumps

4. DISPATCH an operator to FILL CVCS and WDS tanks USING Appendix C, CVCS And WDS Tank Filling Instructions.

NOTE The cold leg accumulators will be vented to the containment atmosphere.

5. NOTIFY RADCON to monitor airborne activity levels in cold leg accumulator rooms during venting operation.

Page 108 of 215

SQN FLOODING I AOP-N.03 Rev.30 Page 1 of4 APPENDIX C CVCS AND WDS TANK FILLING INSTRUCTIONS NOTE I This appendix provides instructions for filling the partially filled and possibly radioactive tanks located below Maximum Probable Flood level of 723.1 elev. Performance of this procedure minimizes the possibility of the tanks collapsing or breaking loose and possible release of radioactivity.

NOTE 2 Steps I through 22 may be performed out of sequence.

1. PLACE Reactor Building Floor and Equipment Drain Sumps IN SERVICE to Tritiated Drain Collector Tank USING 0-SO-77-10.

2. FILL pressurizer relief tank USING 1 ,2-SO-68-5.

3. NOTIFY Maintenance to connect 2 inch passive failure discharge connection on Auxiliary Building Floor and Equipment Drain Sump Pumps discharge USING 0-FP-MXX-000-008.0.

4. OPEN the following valves to fill Floor Drain Collector Tank:

0-77-915

0-77-68 1 NOTE The time required for filling the Tritiated Drain Collector Tank is approximately 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

5. OPEN valve 0-77-914 and FILL Tritiated Drain Collector Tank.

6. ALIGN Laundry and Hot Shower Tank Pump to Waste Condensate Tanks A, B, and C USING 0-SO-77-5.

7. ALIGN Laundry and Hot Shower Tanks A and B and Chemical Drain Tank.

Page 143 of 215

OPL271AOP-N.03 Revision I Page 3 of 41 I. PROGRAM: OPERATOR TRAINING - LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: AOP-N.03, FLOODING IV. LENGTH OF LESSONICOURSE: 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months V. TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of AOP-N.03, FLOODING.

B. Enabling Objectives Objectives

0. Demonstrate an understanding of NUREG 1122 knowledges and abilities associated with a plant Flood that are rated 2.5 during Initial License Training and 3.0 during License Operator Req ualification Training for the appropriate position as identified in Appendix A

1. State the purpose/goal of this AOP-N.03.

2. Describe AOP-N.03 entry conditions.

a. Describe the setpoints, interlocks, and automatic actions associated with AOP-N.03 entry conditions.

b. Describe the ARP requirements associated with AOP-N.03 entry conditions.

c. Interpret, prioritize, and verify associated alarms are consistent with AOP-N.03 entry conditions.

d. Describe the plant parameters that may indicate a plant Flood.

3. Describe the initial operator response to stabilize the plant upon entry into AOP-N.03.

4. Upon entry into AOP-N.03, diagnose the applicable condition and transition to the appropriate procedural section for response.

5. Summarize the mitigating strategy for the failure that initiated entry into AOP-N.03.

6. Describe the bases for all limits, notes, cautions, and steps of AOP-N.03.

0PL271 AOP-N.03 Revision I Page 4 of 41

7. Describe the conditions and reason for transitions within this procedure and transitions to other procedures.

8. Given a set of initial plant conditions use AOP-N.03 to correctly:

a. Recognize entry conditions.

b. Identify required actions.

c. Respond to Contingencies.

d. Observe and Interpret Cautions and Notes.

9. Describe the Tech Spec and TRM actions applicable during the performance of AOP-N.03.

10. Apply GFE and system response concepts to the abnormal condition prior to, during and after the abnormal condition.

OBJECTIVES TO BE COVERED IN THESE SEQUOYAH OPERATOR TRAINING PROGRAMS OBJECTIVE NONLICENSED LICENSE TRAINING NO. OPERATORS RO SRO REQUALISPECIAL

0. X X

1. x x

2. X X

3. X X

4. x x

5. X X

6. X X

7. X X

8. X X

9. X X

10. x x Selected objectives to be covered in:

PowerPoint presentation to be used:

Sequoyah Operator Training Manager I________

Date Sequoyah Operations Manager I________

Date

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta

99. Given the following plant conditions on Unit 1:

Time Condition 0803 PZR level dropping slowly.

0805 Crew enters AOP-R.05, RCS Leak and Leak Source Identification.

0808 Crew trips the reactor and initiates Safety Injection.

0812 ALERT is declared.

0828 SITE AREA EMERGENCY is declared.

Which ONE of the following is the latest time that the f State notification must be made?

A. 0818 B. 0823 C 0827 D. 0843 DIS TRACTOR ANALYSIS:

A. Incorrect, Plausible because the time is 15 minutes from the start of the event.

B. Incorrect, Plausible because the time is 15 minutes from the reactor trip.

C. Correct, The 15 minutes maximum following declaration for notification of state and local authorities.

D. Incorrect, Plausible because the time is 15 minutes from the time of the SAE declaration and the State would have already been required to be notified of the Alert.

Saturday, July 17, 2010 4:07:57 PM 99

QUESTIONS REPORT for 2010 SEPT SRO EXAM Pre-Atanta Question Number: 99 Tier: 3 Group n/a KIA: G 2.4.30 Emergency Procedures / Plan Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

Importance Rating: 2.7 / 4.1 10 CFR Part 55: 41.10/43.5/45.11 IOCFR55.43.b: 5 KIA Match: Applicant is required to have knowledge of the requirements for the time requirement for notifiying the State of implementation of the Radiological Emergency Plan.

Technical

Reference:

EPIP-1, Emergency Plan Classification Matrix, Rev 42 Proposed references None to be provided:

Learning Objective: 0PL271 REP

2. Determine the required notifications based upon the event, including time requirements.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: SQN bank question G2.4.30 001 used on 2007 exam.

Comments:

MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9 Answer: CBCDDDABCA ScrambleRange:A-D Saturday, July 17, 2010 4:07:57 PM 100

SEQUOYAH EMERGENCY PLAN CLASSIFICATION MATRIX EPIP-1 3.0 INSTRUCTIONS (Continued) 3.4.7 The NRC shall be notified within one hour of all classifications. Once made and reported, a declaration cannot be canceled or rescinded even if it is later determined to be invalid. If there is reason to doubt that a given condition has occurred, the SM or SED shall follow indications and proceed with classification, as required by this procedure, until otherwise proven false.

3.4.8 The State shall be notified by the ODS within 15 minutes of any declaration and notified, for information only, within one hour of any classification that was met but not declared as allowed above. If the State is notified of a declaration that is invalidated before the NRC is notified, terminate the classification, if not already done, and report the declaration to the NRC.

3.4.9 The ACCEPTABLE timeframe for initiating notification to the ODS of an emergency declaration is considered to be five (5) minutes. This is the time period between declaration of the emergency and contacting the ODS.

4.0 RECORDS RETENTION 4.1 Records of Classified Emergencies The materials generated in support of key actions during an actual emergency classified as NOUE or higher are considered Lifetime retention Non-QA records.

Materials shall be forwarded to the EP Manager who shall submit any records deemed necessary to demonstrate performance to the Corporate EP Manager for storage.

4.2 Drill and Exercise Records The materials deemed necessary to demonstrate performance of key actions during drills are considered Non-QA records. These records shall be forwarded to the EP Manager who shall retain records deemed necessary to demonstrate six-year plan performance for six years. The EP Manager shall retain other records in this category for three years.

Page 6 of 47 Revision 44

0PL271 REP Revision 1 Page 3 of 32 I. PROGRAM: OPERATOR TRAINING LICENSED II. COURSE: LICENSE TRAINING III. LESSON TITLE: NP RADIOLOGICAL EMERGENCY PLAN AND SEQUOYAH EMERGENCY PLAN IMPLEMENTING PROCEDURES IV. LENGTH OF LESSON/COURSE: 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (Hot License Class), 2 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (LOR)

V. TRAINING OBJECTIVES:

A. Terminal Objective:

Upon completion of License Training, the participant shall be able to demonstrate or explain, using classroom evaluations and/or simulator scenarios, the requirements of the Radiological Emergency Plan (REP).

B. Enabling Objectives:

0. Demonstrate an understanding of NUREG 1122 Knowledge and Abilities associated with Radiological Emergency Plan that are rated 2.5 during Initial License Training and 3.0 during License Operator Requalification Training for the appropriate license position as identified in Appendix A.