Text

Initial Exam 2013-301 Draft SRO Written Exam
	ML14091A150
Person / Time
Site:	Farley
Issue date:	04/01/2014
From:	; Division of Reactor Safety II
To:	; Southern Nuclear Operating Co
References
	Download: ML14091A150 (323)
	v • d • e

The following SRO questions were part of the initial 10 question presubmittal and have had minor changes made.

80. 024AA2.06 Restated the first part of the stem and put in a timeline based on validator comments that the sequence of events was confusing. No change of content or intent.

84. 054AA2.01 Modified the second part of distracters B and D from a LOCA event to a Steam Generator Tube Rupture event - based on comments during the 10 question presubmittal review.

Also added the title of TS 3.7.3 in the stem of the question. It was not previously included.

One KA needs changed, per our discussion.

On Feb 7, 2013 we discussed changing KA 028G2.1.1 Hydrogen Recombiner and Purge Control System (HRPS) Knowledge of conduct of operations requirements to 029 Containment Purge System. The 028 was too hard to write to and we agreed to write a question to the ctmt purge KA. This was approved and needs to be changed on the sample plan or we need authorization to put this on FORM 4014 Record of Rejected KAs.

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

76. 001A2.17 076 Unit 1 is at 98% power.

The OATC pulls Control Rods 2 steps and the following indications occur:

Control Bank D Group Step Counters indicate 227 steps.

On DRPI, Control Bank D, rod P8 drops to 144 steps.

On DRPI, all other Control Bank D rods indicate 228 steps.

Other Control Room parameters indicate Control Rod P8 has dropped to 144 steps.

Which one of the following completes the statements below?

(1) , will come into alarm due to the rod malfunction.

Per the BASES of Tech Spec 3.1.4, Rod Group Alignment Limits, a power reduction to < 75% power is required to ensure (2) .

A. 1) FF5, COMP ALARM ROD SEQ/DEV OR PR FLUX TILT

2) total available rod worth is within safety analyses limits B. 1) FE2, CONT ROD BANK POSITION LO-LO

2) total available rod worth is within safety analyses limits C. 1) FF5, COMP ALARM ROD SEQ/DEV OR PR FLUX TILT

2) the local Linear Heat Rate increases will not exceed core design criteria D. 1) FE2, CONT ROD BANK POSITION LO-LO

2) the local Linear Heat Rate increases will not exceed core design criteria Tuesday, March 19, 2013 11:38:42 AM 1

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 A. Incorrect 1) Correct, FF5 will come into alarm due to Control Rod P8 being misaligned by >12 steps.

2) Incorrect, since total available rod worth has not changed.

Plausible because a candidate may make a judgment that the control rod that is misaligned is not OPERABLE as described in TS bases 3.1.4, and has to be assumed unavailable for shutdown margin. In actuality, as long as the control rod is trippable it is available for shutdown margin (no information is given in the stem that indicates the control rod is untrippable). A ramp down to a lower power level will require less negative reactivity for a reactor shutdown (and less negative reactivity from total available rod worth) and a candidate may assume this is required by the TS Bases and safety analyses.

B. Incorrect 1) Incorrect, plausible because a single rod is below the Lo-Lo Control Rod insertion limit. If a candidate thought the Rod Insertion limit was calculated from DRPI, this would be a plausible choice. The Rod Insertion limit is actually calculated from the P/A Converter input (in the Rod Control Logic cabinet) providing rod position and median delta T providing Reactor power. DRPI has no input.

2) Incorrect, see A.2.

C. Correct 1) Correct, see A.1.

2) Correct, the TS Bases reason for reducing Reactor power to <75% is to ensure Linear Heat Rate increases on the fuel assemblies outside of the dropped rod area are within limits to ensure fuel integrity.

D. Incorrect 1) Incorrect, see B.1.

2) Correct, see C.2.

K/A: 001A2.17 Control Rod Drive System A2 Ability to (a) predict the impacts of the following malfunction or operations on the CRDS- and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

(CFR: 41.5/43.5/45.3/45.13)

A2.17 Rod-misalignment alarm . . . . . . . . . . . . . . . . . 3.3 3.8 Importance Rating: 3.3 3.8 Technical

Reference:

FNP-1-ARP-1.6, v70 FNP-1-AOP-19.0, v29 Tuesday, March 19, 2013 11:38:42 AM 2

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 Tech Specs v190/186 Tech Spec Bases v58 References provided: None Learning Objective: DETERMINE AND APPLY the information from the LCO BASES sections: BACKGROUND, APPLICABLE SAFETY ANALYSIS, ACTIONS, SURVEILLENCE REQUIREMENTS, for any Technical Specifications or TRM requirements associated with AOP-19, Malfunction of Rod Control System, components and attendant equipment.

(OPS-62520S01) 10CFR55.43 (b) 2 Question History: New question Basis for meeting K/A: A Control Rod misalignment has occurred during a rod movement. The candidate has to predict the alarm that will be received due to the rod misalignment, based on the plant conditions. Using Tech Spec Bases information, the candidate has to determine why a ramp to < 75% power is required as a result of the misaligned rod.

SRO justification: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

Knowledge of TS bases that is required to analyze TS required actions and terminology.

The candidate has to apply knowledge of the reason for the ramp to < 75% power due to the Tech Spec action to be

<75% within two hours.

2013 NRC exam From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 flowchart for 10 CFR 55.43(b)(2) :

1) can NOT be answered by knowing less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Tech Specs.

2) can NOT be answered by knowing information listed "above-the-line".

3) can NOT be answered by knowing the TS Safety Limits or their bases.

4) Does involve one or more of the following for TS, TRM or ODCM:

Knowledge of TS bases that is required to analyze TS required actions and terminology Tuesday, March 19, 2013 11:38:42 AM 3

03/15/13 00:31:18 UNIT 1 FNP-1-ARP-1.6 LOCATION FE2 SETPOINT: Variable with Reactor Power as measured by E2

'T and TAVG. CONT ROD BANK ORIGIN: Rod Insertion Limit Computer POSITION LO-LO Input is not from DRPI PROBABLE CAUSE NOTE: y Zinc Addition System injection will result in a continuous RCS dilution of as much as 1.7 gph, which may result in a reduction in shutdown margin if compensated for by inward rod motion instead of boration.

y This annunciator has REFLASH capability.

1. Reactor Coolant System Boric Acid Concentration too low to ensure Reactor Protection under Accident conditions due to; A. Plant Transient B. Xenon Transient C. Dilution of RCS AUTOMATIC ACTION NONE OPERATOR ACTION

1. Check indications and determine that actual control bank rod position is at the low-low insertion limit.

1.1 Click on Rod Supervision button on Applications Menu.

1.2 Click on Rod Insertion Limits button.

1.3 Determine if low-low insertion limit exceeded.

2. Emergency borate the reactor coolant system in accordance with FNP-1-AOP-27.0, EMERGENCY BORATION. {CMTs 0008555, 0008900}

3. IF a plant transient is in progress, THEN place turbine load on "HOLD".

4. Refer to FNP-1-UOP-3.1, POWER OPERATIONS.

5. Refer to the Technical Specifications section on Reactivity Control.

References:

A-177100, Sh. 292; U-2606l0; U266647 PLS Document; Technical Specifications; DCP 93-1-8587; {CMT 0008887}

Page 1 of 1 Version 70.0

Rod Group Alignment Limits 3.1.4 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Rod Group Alignment Limits LCO 3.1.4 All shutdown and control rods shall be OPERABLE, with all individual indicated rod positions within 12 steps of their group step counter demand position.

APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more rod(s) A.1.1 Verify SDM to be within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months untrippable. the limits provided in the COLR.

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

AND A.2 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months B. One rod not within B.1 Restore rod to within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months alignment limits. alignment limits.

OR B.2.1.1 Verify SDM to be within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the limits provided in the COLR.

OR (continued)

Farley Units 1 and 2 3.1.4-1 Amendment No. 146 (Unit 1)

Amendment No. 137 (Unit 2)

Rod Group Alignment Limits 3.1.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2.1.2 Initiate boration to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore SDM to within limit.

AND B.2.2 Reduce THERMAL 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months POWER to 75% RTP.

AND B.2.3 Verify SDM to be within Once per the limits provided in the 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months COLR.

AND B.2.4 Perform SR 3.2.1.1. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months AND B.2.5 Perform SR 3.2.2.1. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months AND B.2.6 Re-evaluate safety 5 days analyses and confirm results remain valid for duration of operation under these conditions.

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

Farley Units 1 and 2 3.1.4-2 Amendment No. 146 (Unit 1)

Amendment No. 137 (Unit 2)

Rod Group Alignment Limits 3.1.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. More than one rod not D.1.1 Verify SDM to be within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months within alignment limit. the limits provided in the COLR.

OR D.1.2 Initiate boration to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore required SDM to within limit.

AND D.2 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify individual rod positions within alignment limit. In accordance with the Surveillance Frequency Control Program SR 3.1.4.2 Verify rod freedom of movement (trippability) by In accordance with moving each rod not fully inserted in the core the Surveillance 10 steps in either direction. Frequency Control Program SR 3.1.4.3 Verify rod drop time of each rod, from the fully Prior to reactor withdrawn position, is 2.7 seconds from the criticality after beginning of decay of stationary gripper coil voltage each removal of to dashpot entry, with: the reactor head

a. Tavg 541°F; and

b. All reactor coolant pumps operating.

Farley Units 1 and 2 3.1.4-3 Amendment No. 185 (Unit 1)

Amendment No. 180 (Unit 2)

Rod Group Alignment Limits B 3.1.4 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.4 Rod Group Alignment Limits BASES BACKGROUND The OPERABILITY (e.g., trippability) of the shutdown and control rods is an initial assumption in all safety analyses that assume rod insertion upon reactor trip. Maximum rod misalignment is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available SDM.

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

may be a reduction in total available rod worth - but this is Mechanical or electrical failures may cause a control rod to become for a situation with an inoperable or to become misaligned from its group. Control rod INOPERABLE control rod, not inoperability or misalignment may cause increased power peaking, a misaligned control rod. This due to the asymmetric reactivity distribution and a reduction in the lends plausibility for the total available rod worth for reactor shutdown. Therefore, control rod distracter "a power reduction alignment and OPERABILITY are related to core operation in design power peaking limits and the core design requirement of a minimum to < 75% power is required to SDM.

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

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

The RCCAs are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control. A group consists of two or more RCCAs that are electrically paralleled to step simultaneously. A bank (continued)

Farley Units 1 and 2 B 3.1.4-1 Revision 0

Rod Group Alignment Limits B 3.1.4 BASES BACKGROUND of RCCAs consists of two groups that are moved in a staggered (continued) fashion, but always within one step of each other. There are four control banks and two shutdown banks. All control banks and shutdown banks contain two rod groups.

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

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

The Bank Demand Position Indication System counts the pulses from the rod control system that moves the rods. There is one step counter for each group of rods. Individual rods in a group all receive the same signal to move and should, therefore, all be at the same position indicated by the group step counter for that group. The Bank Demand Position Indication System is considered highly precise (+/- 1 step or

+/- 5/8 inch). If a rod does not move one step for each demand pulse, the step counter will still count the pulse and incorrectly reflect the position of the rod.

The DRPI System provides a highly accurate indication of actual control rod position, but at a lower precision than the step counters.

This system is based on inductive analog signals from a series of coils spaced along a hollow tube with a center to center distance of 3.75 inches, which is six steps. To increase the reliability of the system, the inductive coils are connected alternately to data system A or B. Thus, if one system fails, the DRPI will go on half accuracy with (continued)

Farley Units 1 and 2 B 3.1.4-2 Revision 0

Rod Group Alignment Limits B 3.1.4 BASES BACKGROUND an effective coil spacing of 7.5 inches, which is 12 steps. Therefore, (continued) the normal indication accuracy of the DRPI System is +/- 4 steps (all coils operable and 1 step added for manufacturing and temperature tolerances), and the maximum uncertainty is +/- 10 steps (only one data system A or B coils operable). With an indicated deviation of 12 steps between the group step counter and DRPI, the maximum deviation between actual rod position and the demand position could be 22 steps.

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

a. There be no violations of:

1. specified acceptable fuel design limits, or

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

b. The core remains subcritical after accident transients that result in a reactor trip, except for the MSLB.

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

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

(continued)

Farley Units 1 and 2 B 3.1.4-3 Revision 0

Rod Group Alignment Limits B 3.1.4 BASES APPLICABLE Another type of misalignment occurs if one RCCA fails to insert upon SAFETY ANALYSES a reactor trip and remains stuck fully withdrawn. This condition is (continued) assumed in the evaluation to determine that the required SDM is met with the maximum worth RCCA also fully withdrawn.

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

Continued operation of the reactor with a misaligned control rod is allowed if the heat flux hot channel factor (FQ(Z)) and the nuclear N

enthalpy hot channel factor ( FH ) are verified to be within their limits in the COLR and the safety analysis is verified to remain valid. When a control rod is misaligned, the assumptions that are used to determine the rod insertion limits, AFD limits, and quadrant power tilt limits are not preserved. Therefore, the limits may not preserve the N must be verified directly by design peaking factors, and FQ(Z) and FH incore mapping. Bases Section 3.2 (Power Distribution Limits)

N contains more complete discussions of the relation of FQ(Z) and FH to the operating limits.

Shutdown and control rod OPERABILITY and alignment are directly related to power distributions and SDM, which are initial conditions assumed in safety analyses. Therefore they satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

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

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

Maintaining rods within the limits of the TS ensures that the safety analyses remains (continued) valid, Farley outside the limits, the Units 1 and 2 safety analyses B 3.1.4-4 Revision 0 are no longer valid.

Rod Group Alignment Limits B 3.1.4 BASES LCO Failure to meet the requirements of this LCO may produce (continued) unacceptable power peaking factors and LHRs, or unacceptable SDMs, all of which may constitute initial conditions inconsistent with the safety analysis.

APPLICABILITY The requirements on RCCA OPERABILITY and alignment are applicable in MODES 1 and 2 because these are the only MODES in which a self-sustaining chain reaction occurs, and the OPERABILITY (i.e., trippability) and alignment of rods have the potential to affect the safety of the plant. In MODES 3, 4, 5, and 6, the alignment limits do not apply because the control rods are fully inserted and the reactor is shut down, with no self-sustaining chain reaction. In the shutdown MODES, the OPERABILITY of the shutdown and control rods has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the RCS. See LCO 3.1.1, "SHUTDOWN MARGIN (SDM), " for SDM in MODES 3, 4, and 5 and LCO 3.9.1, "Boron Concentration," for boron concentration requirements during refueling.

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

In this situation, SDM verification must account for the absence of the negative reactivity of the untrippable rod(s), as well as a rod of maximum worth.

A.2 If the untrippable rod(s) cannot be restored to OPERABLE status, the plant must be brought to a MODE or condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

(continued)

Farley Units 1 and 2 B 3.1.4-5 Revision 0

Rod Group Alignment Limits B 3.1.4 BASES ACTIONS A.2 (continued)

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

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

An alternative to realigning a single misaligned RCCA to the group average position is to align the remainder of the group to the position of the misaligned RCCA. However, this must be done without violating the bank sequence, overlap, and insertion limits specified in LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, "Control Bank Insertion Limits." The Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months gives the operator sufficient time to adjust the rod positions in an orderly manner.

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

In many cases, realigning the remainder of the group to the misaligned rod may not be desirable. For example, realigning control bank B to a rod that is misaligned 15 steps from the top of the core would require a significant power reduction, since control bank D must be moved fully in and control bank C must be moved in to below 90 steps.

Power operation may continue with one RCCA trippable but misaligned, provided that SDM is verified within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months represents the time necessary for determining the actual unit SDM and, if necessary, aligning and starting the necessary systems and components to initiate boration.

(continued)

Farley Units 1 and 2 B 3.1.4-6 Revision 0

Rod Group Alignment Limits B 3.1.4 BASES ACTIONS B.2.2, B.2.3, B.2.4, B.2.5, and B.2.6 (continued)

For continued operation with a misaligned rod, RTP must be reduced, SDM must periodically be verified within limits, hot channel factors N ) must be verified within limits, and the safety analyses (FQ(Z) and FH must be re-evaluated to confirm continued operation is permissible.

Reduction of power to 75% RTP ensures that local LHR increases due to a misaligned RCCA will not cause the core design criteria to be exceeded. The Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months gives the operator sufficient time to accomplish an orderly power reduction without challenging the Reactor Protection System.

When a rod is known to be misaligned, there is a potential to impact the SDM. Since the core conditions can change with time, periodic verification of SDM is required. A Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient to ensure this requirement continues to be met.

N are within the required limits ensures that Verifying that FQ(Z) and FH current operation at 75% RTP with a rod misaligned is not resulting in power distributions that may invalidate safety analysis assumptions at full power. The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months allows sufficient time to obtain flux maps of the core power distribution using the incore flux N .

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

The following accident analyses are required to be reevaluated:

1. Rod Cluster Control Assembly Insertion Characteristics;

2. Rod Cluster Control Assembly Misalignment;

3. Loss Of Reactor Coolant From Small Ruptured Pipes or From Cracks In Large Pipes Which Actuates The Emergency Core Cooling System; (continued)

Farley Units 1 and 2 B 3.1.4-7 Revision 0

Rod Group Alignment Limits B 3.1.4 BASES ACTIONS B.2.2, B.2.3, B.2.4, B.2.5, and B.2.6 (continued)

4. Single Rod Cluster Control Assembly Withdrawal At Full Power;

5. Major Reactor Coolant System Pipe Ruptures (Loss Of Coolant Accident);

6. Major Secondary System Pipe Rupture; and

7. Rupture Of A Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection).

C.1 When Required Actions cannot be completed within their Completion Time, the unit must be brought to a MODE or Condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , which obviates concerns about the development of undesirable xenon or power distributions. The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging the plant systems.

D.1.1 and D.1.2 More than one control rod becoming misaligned from its group average position is not expected, and has the potential to reduce SDM.

Therefore, SDM must be evaluated. One hour allows the operator adequate time to determine SDM. Restoration of the required SDM, if necessary, requires increasing the RCS boron concentration to provide negative reactivity, as described in the Bases or LCO 3.1.1. The required Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for initiating boration is reasonable, based on the time required for potential xenon redistribution, the low probability of an accident occurring, and the steps required to complete the action. This allows the operator sufficient time to align the required valves and start the boric acid pumps. Boration will continue until the required SDM is restored.

D.2 If more than one rod is found to be misaligned or becomes misaligned because of bank movement, the unit conditions fall outside of the (continued)

Farley Units 1 and 2 B 3.1.4-8 Revision 0

Rod Group Alignment Limits B 3.1.4 BASES ACTIONS D.2 (continued) accident analysis assumptions. Since automatic bank sequencing would continue to cause misalignment, the unit must be brought to a MODE or Condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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

SURVEILLANCE SR 3.1.4.1 REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.1.4.2 Verifying each control rod is OPERABLE would require that each rod be tripped. However, in MODES 1 and 2, tripping each control rod would result in radial or axial power tilts, or oscillations. Exercising each individual control rod provides increased confidence that all rods continue to be OPERABLE without exceeding the alignment limit, even if they are not regularly tripped. Moving each control rod by 10 steps will not cause radial or axial power tilts, or oscillations, to occur. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. Between required performances of SR 3.1.4.2 (determination of control rod OPERABILITY by movement), if a control rod(s) is discovered to be immovable, but remains trippable and aligned, the control rod(s) is considered to be OPERABLE. At any time, if a control rod(s) is immovable, a determination of the trippability (OPERABILITY) of the control rod(s) must be made, and appropriate action taken.

(continued)

Farley Units 1 and 2 B 3.1.4-9 Revision 52

Rod Group Alignment Limits B 3.1.4 BASES SURVEILLANCE SR 3.1.4.3 REQUIREMENTS (continued) Verification of rod drop times allows the operator to determine that the maximum rod drop time permitted is consistent with the assumed rod drop time used in the safety analysis. Measuring rod drop times prior to reactor criticality, after reactor vessel head removal, ensures that the reactor internals and rod drive mechanism will not interfere with rod motion or rod drop time, and that no degradation in these systems has occurred that would adversely affect control rod motion or drop time. This testing is performed with all RCPs operating and the average moderator temperature 541°F to simulate a reactor trip under actual conditions.

Testing is performed with the rods fully withdrawn (225 to 231 steps inclusive). The fully withdrawn position used for determining rod drop times shall be greater than or equal to the fully withdrawn position used during subsequent plant operation.

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

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10 and GDC 26.

2. 10 CFR 50.46.

3. FSAR, Section 15.2.3.

4. FSAR, Section 15.2.3.2.2.C.

Farley Units 1 and 2 B 3.1.4-10 Revision 0

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

77. 004A2.02 077 Unit 1 is stable at 100% power with the following conditions:

AOP-16.0, CVCS Malfunction, is in progress due to a loss of Pressurizer level control.

At the step to "Check Pressurizer level", Pressurizer level is 76% and rising.

Which one of the following completes the statements below?

The Pressurizer is (1) per Tech Spec 3.4.9, Pressurizer.

Per AOP-16.0, the Control Room crew is required to (2) .

(1) (2)

A. NOT OPERABLE trip the Reactor and go to EEP-0.0, Reactor Trip or Safety Injection B. NOT OPERABLE reduce Reactor power/TAVG as necessary using UOP-3.1, Power Operation C. OPERABLE trip the Reactor and go to EEP-0.0, Reactor Trip or Safety Injection D. OPERABLE reduce Reactor power/TAVG as necessary using UOP-3.1, Power Operation Tuesday, March 19, 2013 11:38:42 AM 4

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 A. Correct 1) Correct, per TS 3.4.9, the Pressurizer is not OPERABLE when level is

>63.5%.

2) Correct, per AOP-16.0 a Reactor trip and transition to EEP-0 is required. This does not meet the normal Reactor trip setpoint for entry into EEP-0 and is a procedure transition directed in AOP-16 strictly to mitigate the rising Pressurizer level.

B. Incorrect 1) Correct, see A.1.

2) Incorrect, Plausible since this would be a correct answer if Pressurizer level was between 60-75%.

C. Incorrect 1) Incorrect, per TS 3.4.9 the Pressurizer is not OPERABLE when level is

>63.5%. Plausible since the Pressurizer high level trip setpoint is 92%.

This setpoint is commonly applied for Pressurizer operability and is incorrect.

2) Correct, see A.2.

D. Incorrect 1) Incorrect, see C.1.

2) Incorrect, see B.2.

K/A: 004A2.02 Chemical and Volume Control System:

Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

(CFR: 41.5/ 43/5 / 45/3 / 45/5)

A2.02 Loss of PZR level (failure mode) . . . . . . . . . . . 3.9 4.2 Importance Rating: 3.9 4.2 Technical

Reference:

FNP-1-AOP-16.0, v18 Tech Specs v190/186 Tech Spec Bases v58 References provided: None Learning Objective: EVALUATE plant conditions and DETERMINE if transition to another section of AOP-16, CVCS Malfunction, or to another procedure is required. (OPS-62520K02)

Question History: New question Tuesday, March 19, 2013 11:38:42 AM 5

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 K/A match: The question tests the ability to predict the impacts of a loss of Pressurizer level control after a CVCS malfunction causes Pressurizer level to rise uncontrollably. The impacts are a Tech Spec RAS and a Reactor Trip. Based on the failure, the candidate has to select the proper procedure to mitigate the malfunction SRO justification: Normally knowledge of entry conditions into EEP-0 would be RO knowledge. The normal Reactor trip setpoint for high Pressurizer level is 92% and would be RO knowledge, but in this question selection of EEP-0 is NOT from a direct entry setpoint, but is a procedure transition directed by AOP-16 at 75% Pressurizer level in an effort to control Pressurizer level. This makes this question an SRO procedure selection question rather than a direct EEP-0 entry RO question.

10 CFR 55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations [10 CFR 55.43(b)(5)], involving BOTH:

1) assessing plant conditions and then

2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

One area of SRO level knowledge is knowledge of content of the procedure vs. the procedure's overall mitigative strategy or purpose. The applicants knowledge can be evaluated at the level of 10 CFR55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:

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

2013 NRC exam From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 flowchart for 10 CFR 55.43(b)(5) :

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations [10 CFR 55.43(b)(5)], involving BOTH:

1) assessing plant conditions and then

2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

Using the flowchart, this question can:

NOT be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location.

Tuesday, March 19, 2013 11:38:42 AM 6

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

NOT be answered solely by knowing immediate operator actions.

NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs.(this is not a direct entry condition into EEP-0)

NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

CAN be answered with knowledge of ONE or MORE of the following:

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

Tuesday, March 19, 2013 11:38:42 AM 7

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

a. Pressurizer water level 63.5% indicated; and

b. Two groups of pressurizer heaters OPERABLE with the capacity of each group 125 kW and capable of being powered from an emergency power supply.

APPLICABILITY: MODES 1, 2, and 3.

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

Pressurizer water level limit does not apply during:

a. THERMAL POWER ramp > 5% RTP per minute; or

b. THERMAL POWER step > 10% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressurizer water level A.1 Be in MODE 3 with 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months not within limit. reactor trip breakers open.

AND A.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months B. One required group of B.1 Restore required group of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months pressurizer heaters pressurizer heaters to inoperable. OPERABLE status.

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

C.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Farley Units 1 and 2 3.4.9-1 Amendment No. 146 (Unit 1)

Amendment No. 137 (Unit 2)

Pressurizer 3.4.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 Verify pressurizer water level is 63.5% indicated. In accordance with the Surveillance Frequency Control Program SR 3.4.9.2 Verify capacity of each required group of pressurizer In accordance with heaters is 125 kW. the Surveillance Frequency Control Program SR 3.4.9.3 Verify required pressurizer heaters are capable of In accordance with being powered from an emergency power supply. the Surveillance Frequency Control Program Farley Units 1 and 2 3.4.9-2 Amendment No. 185 (Unit 1)

Amendment No. 180 (Unit 2)

1/16/2013 18:21 FNP-1-AOP-16.0 UNIT 1 CVCS MALFUNCTION Revision 18.0 Step Action/Expected Response Response NOT Obtained 10.2 Go to procedure and step in effect

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ NOTE: The intent of the following steps is to borate through the RCP seals and reduce reactor power. The power reduction and cooldown will lower TAVG and will either reduce the rate of pressurizer level increase or stop the pressurizer level increase through RCS shrinkage, and afford time to restore a letdown path.

Maximum ramp rates specified in FNP-1-UOP-3.1 are not applicable.

Pressurizer level, response to boration through the RCP seals, and time required to restore a letdown path will dictate the maximum ramp rate. Depending on pressurizer level a reactor trip may be required.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ 11 Verify CHG FLOW FK-122 -

MANUALLY CLOSED 12 Check Reactor - CRITICAL 12 Proceed to step 16.0.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ NOTE: Tech Spec. 3.4.9 actions apply when pressurizer level exceeds 63.5%.

The requirement for a reactor trip at 75% prior to reaching the automatic trip setpoint at 92% allows more time to establish a cooldown before the pressurizer fills solid.

If a reactor trip is required, the remainder of this procedure should be performed in conjunction with FNP-1-ESP-0.1, REACTOR TRIP RESPONSE.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ 13 [CA] Check PRZR level - LESS 13 Trip the reactor and go to THAN 75% FNP-1-EEP-0, REACTOR TRIP OR SAFETY INJECTION.

14 [CA] Maintain PRZR level -

20-60%.

Procedure transition based on 14.1 Reduce reactor power/TAVG as necessary using FNP-1-UOP-3.1, POWER OPERATION Pressurizer level.

Page 11 of 23

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

78. 006G2.2.22 078 Unit 1 is in Mode 3 with the following conditions:

RCS pressure is 1800 psig.

An RCS heatup is in progress.

1A Accumulator has been declared INOPERABLE.

1A Accumulator pressure is 655 psig.

1A Accumulator level is 41%.

Per Tech Specs 3.5.1, Accumulators, which one of the following completes the statements below?

Restore 1A Accumulator to OPERABLE by (1) to within the Tech Spec required range.

Tech Spec 3.5.1 BASES states that (2) .

A. 1) lowering 1A Accumulator pressure

2) the maximum pressure limit prevents injecting nitrogen into the RCS during a LOCA B. 1) lowering 1A Accumulator pressure

2) the maximum pressure limit prevents the accumulator relief valve from actuating C. 1) raising 1A Accumulator level

2) the minimum level limit prevents injecting nitrogen into the RCS during a LOCA D. 1) raising 1A Accumulator level

2) the minimum level limit ensures peak clad temperature remains below 2200°F during a LOCA Tuesday, March 19, 2013 11:38:42 AM 8

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 This is not a true 2+2 question since the 2nd part of each question has to be plausible and related to the first part. There are four distinct answer choices.

The answer choices "lowering pressure" and "raising level" were chosen to ensure that there were not two possible correct answer choices. The question asks which one of those two actions is required to restore the parameter to within limits. Under certain conditions, lowering level could also lower pressure. By the same token, raising level could raise pressure. The two answer choices were chosen to preclude the action of one correcting the condition of the other.

A. Incorrect 1) Correct, the TS SR 3.5.1.3 limits for Accumulator pressure are > 601 psig and < 649 psig.

2) Incorrect, this is a common misconception that too much nitrogen pressure will cause gas injection into the RCS during a LOCA. During a large break LOCA, nitrogen gas will be injected into the RCS since the RCS pressure is so much lower than the Accumulator pressure.

B. Correct 1) Correct, see A.1.

2) Correct, TS 3.5.1 Bases states "The maximum nitrogen cover pressure limit prevents accumulator relief valve actuation, and ultimately preserves Accumulator integrity."

C. Incorrect 1) Incorrect, the TS SR 3.5.1.2 limits for Accumulator level are > 31.4%

and < 58.4%, which is already met in the stem conditions with level at 41%. Plausible since this setpoint could be seen as lower than we normally maintain our safety related tank levels (for example the Condensate Storage Tank or Refueling Water Storage Tank).

2) Incorrect, Plausible since, along with the first part, it would seem that a higher level would prevent nitrogen injection during a LOCA.

D. Incorrect 1) Incorrect, see C.1.

2) Correct, this statement could be correct, but in conjunction with the first part, it makes this whole distracter incorrect. Accumulator injection, along with sufficient level, does help prevent exceeding the peak cladding temperature limit of 2200°F.

K/A: 006G2.2.22 Emergency Core Cooling System (ECCS):

Knowledge of limiting conditions for operations and safety limits.

(CFR: 41.5 / 43.2 / 45.2)

IMPORTANCE RO 4.0 SRO 4.7 Importance Rating: 4.0 4.7 Tuesday, March 19, 2013 11:38:42 AM 9

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 Technical

Reference:

Tech Specs v190/186 Tech Spec Bases v58 References provided: None Learning Objective: RECALL AND APPLY the information from the LCO BASES sections: BACKGROUND, APPLICABLE SAFETY ANALYSIS, ACTIONS, SURVEILLENCE REQUIREMENTS, for any Technical Specifications or TRM requirements associated with the Emergency Core Cooling System components and attendant equipment alignment, to include the following (OPS-62102B01): 10CFR55.43 (b) 2 3.5.1 Accumulators 3.5.2 ECCSOperating 3.5.3 ECCSShutdown 3.5.4 Refueling Water Storage Tank (RWST) 2.1.1 Reactor Core Safety Limits Question History: FNP Bank question ECCS-62102B01 05 K/A match: This question tests the candidate's knowledge of the Tech Spec ECCS Accumulator pressure and level limits. In addition, it tests the TS Bases reason for the limit.

SRO justification: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

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

Knowledge of TS bases that is required to analyze TS required actions and terminology.

All of the required knowledge is TS below the line limitations. The first part of the question requires the candidate to know the Surveillance requirements for Accumulator level and pressure and determine the action required to restore the Accumulator to OPERABLE status.

The second part of the question tests the candidate's knowledge of the TS Bases for the pressure and level limits.

2013 NRC exam From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 flowchart for 10 CFR 55.43(b)(2) :

1) can NOT be answered by knowing less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Tech Specs.

2) can NOT be answered by knowing information listed "above-the-line".

3) can NOT be answered by knowing the TS Safety Limits or their bases.

Tuesday, March 19, 2013 11:38:42 AM 10

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

4) Does involve one or more of the following for TS, TRM or ODCM:

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

Knowledge of TS bases that is required to analyze TS required actions and terminology.

Tuesday, March 19, 2013 11:38:42 AM 11

Accumulators 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1 Accumulators LCO 3.5.1 Three ECCS accumulators shall be OPERABLE.

APPLICABILITY: MODES 1 and 2, MODE 3 with RCS pressure > 1000 psig.

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

In MODE 3, with RCS pressure > 1000 psig, the accumulators may be inoperable for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to perform pressure isolation valve testing per SR 3.4.14.1.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One accumulator A.1 Restore boron 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable due to boron concentration to within concentration not within limits.

limits.

B. One accumulator B.1 Restore accumulator to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months inoperable for reasons OPERABLE status.

other than Condition A.

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

C.2 Reduce RCS pressure to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 1000 psig.

D. Two or more D.1 Enter LCO 3.0.3. Immediately accumulators inoperable.

Farley Units 1 and 2 3.5.1-1 Amendment No. 162 (Unit 1)

Amendment No. 155 (Unit 2)

Accumulators 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify each accumulator isolation valve is fully open. In accordance with the Surveillance Frequency Control Program SR 3.5.1.2 Verify borated water volume in each accumulator is In accordance with 7555 gallons (31.4%) and 7780 gallons (58.4%). the Surveillance Frequency Control Program SR 3.5.1.3 Verify nitrogen cover pressure in each accumulator is In accordance with 601 psig and 649 psig. the Surveillance Frequency Control Program SR 3.5.1.4 Verify boron concentration in each accumulator is In accordance with 2200 ppm and 2500 ppm. the Surveillance Frequency Control Program AND

NOTE-------

Only required to be performed for affected accumulators Once within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution volume increase of 12%

level, indicated, that is not the result of addition from the refueling water storage tank SR 3.5.1.5 Verify power is removed from each accumulator In accordance with isolation valve operator when RCS pressure is the Surveillance 2000 psig. Frequency Control Program Farley Units 1 and 2 3.5.1-2 Amendment No. 185 (Unit 1)

Amendment No. 180 (Unit 2)

Accumulators B 3.5.1 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.1 Accumulators BASES BACKGROUND The functions of the ECCS accumulators are to supply water to the reactor vessel during the blowdown phase of a loss of coolant accident (LOCA), to provide inventory to help accomplish the refill phase that follows thereafter, and to provide Reactor Coolant System (RCS) makeup for a small break LOCA.

The blowdown phase of a large break LOCA is the initial period of the transient during which the RCS departs from equilibrium conditions, and heat from fission product decay, hot internals, and the vessel continues to be transferred to the reactor coolant. The blowdown phase of the transient ends when the RCS pressure falls to a value approaching that of the containment atmosphere.

In the refill phase of a LOCA, which immediately follows the blowdown phase, reactor coolant inventory has vacated the core through steam flashing and ejection out through the break. The core is essentially in adiabatic heatup. The balance of accumulator inventory is then available to help fill voids in the lower plenum and reactor vessel downcomer so as to establish a recovery level at the bottom of the core and ongoing reflood of the core with the addition of safety injection (SI) water.

The accumulators are pressure vessels partially filled with borated water and pressurized with nitrogen gas. The accumulators are passive components, since no operator or control actions are required in order for them to perform their function. Internal accumulator tank pressure is sufficient to discharge the accumulator contents to the RCS, if RCS pressure decreases below the accumulator pressure.

Each accumulator is piped into an RCS cold leg via an accumulator line and is isolated from the RCS by a motor operated isolation valve and two check valves in series.

The accumulator motor operated isolation valves are maintained in the open position with power to the valve removed when pressurizer pressure is 2000 psig. Should the valves be inadvertently closed below 2000 psig, the requirements of this LCO would ensure that the valves would be returned to their correct position in a timely manner or the plant would be taken out of the Mode of Applicability. The valves will (continued)

Farley Units 1 and 2 B 3.5.1-1 Revision 0

Accumulators B 3.5.1 BASES BACKGROUND automatically open, however, as a result of an SI signal. These features (continued) and requirements ensure that the accumulators will be available for injection.

The accumulator size, water volume, and nitrogen cover pressure are selected so that two of the three accumulators are sufficient to partially cover the core before significant clad melting or zirconium water reaction can occur following a LOCA. The need to ensure that two accumulators are adequate for this function is consistent with the LOCA assumption that the entire contents of one accumulator will be lost via the RCS pipe break during the blowdown phase of the LOCA.

APPLICABLE The accumulators are assumed OPERABLE in both the large and SAFETY ANALYSES small break LOCA analyses at full power (Ref. 1). These are the Design Basis Accidents (DBAs) that establish the acceptance limits for the accumulators. Reference to the analyses for these DBAs is used to assess changes in the accumulators as they relate to the acceptance limits.

In performing the LOCA calculations, conservative assumptions are made concerning the availability of ECCS flow. In the early stages of a LOCA, with or without a loss of offsite power, the accumulators provide the sole source of makeup water to the RCS. The assumption of loss of offsite power is also considered to determine if it is most limiting, and if so, imposes a delay wherein the ECCS pumps cannot deliver flow until the emergency diesel generators start, come to rated speed, and go through their timed loading sequence. In cold leg break scenarios, the entire contents of one accumulator are assumed to be lost through the break.

The limiting large break LOCA is a double ended guillotine break in the cold leg. During this event, the accumulators discharge to the RCS as soon as RCS pressure decreases to below accumulator pressure.

As a conservative estimate, no credit is taken for ECCS pump flow until an effective delay has elapsed. This delay accounts for the diesels starting and the pumps being loaded and delivering full flow. The delay time is conservatively set with an additional 2 seconds to account for SI signal generation. During this time, the accumulators are analyzed as providing the sole source of emergency core cooling. No operator action is assumed during the blowdown stage of a large break LOCA.

(continued)

Farley Units 1 and 2 B 3.5.1-2 Revision 0

Accumulators B 3.5.1 BASES APPLICABLE The worst case small break LOCA analyses also assume a time delay SAFETY ANALYSES before pumped flow reaches the core. For the larger range of small (continued) breaks, the rate of blowdown is such that the increase in fuel clad temperature is terminated solely by the accumulators, with pumped flow then providing continued cooling. As break size decreases, the accumulators and centrifugal charging pumps both play a part in terminating the rise in clad temperature. As break size continues to decrease, the role of the accumulators continues to decrease until they are not required and the centrifugal charging pumps become solely responsible for terminating the temperature increase.

This LCO helps to ensure that the following acceptance criteria established for the ECCS by 10 CFR 50.46 (Ref. 2) will be met following a LOCA:

a. Maximum fuel element cladding temperature is 2200°F;

b. Maximum cladding oxidation is 0.17 times the total cladding thickness before oxidation;

c. Maximum hydrogen generation from a zirconium water reaction is

0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react; and

d. Core is maintained in a coolable geometry.

Since the accumulators discharge during the blowdown phase of a LOCA, they do not contribute to the long term cooling requirements of 10 CFR 50.46.

For both the large and small break LOCA analyses, a nominal contained accumulator water volume is used. The contained water volume is the same as the deliverable volume for the accumulators, since the accumulators are emptied, once discharged. For large breaks, an increase in water volume can be either a peak clad temperature penalty or benefit, depending on downcomer filling and subsequent spill through the break during the core reflooding portion of the transient. The safety analysis assumes values of 7331 gallons for the accumulator, and 337 gallons for the accumulator discharge line. To allow for instrument inaccuracy, values of 7,555 gallons and 7,780 gallons are specified.

These values include the volume of water in the accumulator discharge line.

(continued)

Farley Units 1 and 2 B 3.5.1-3 Revision 0

Accumulators B 3.5.1 BASES APPLICABLE The minimum boron concentration setpoint is used in the post LOCA SAFETY ANALYSES boron concentration calculation. The calculation is performed to assure (continued) reactor subcriticality in a post LOCA environment. Of particular interest is the large break LOCA, since no credit is taken for control rod assembly insertion. A reduction in the accumulator minimum boron concentration would produce a subsequent reduction in the available containment sump concentration for post LOCA shutdown and an increase in the maximum sump pH. The maximum boron concentration is used in determining the cold leg to hot leg recirculation injection switchover time and minimum sump pH.

The large and small break LOCA analyses are performed at the minimum nitrogen cover pressure for small break LOCA and nominal nitrogen cover pressure for large break LOCA, since sensitivity analyses have demonstrated that higher nitrogen cover pressure results in a computed peak clad temperature benefit. A sensitivity study is performed for the BE LOCA (large break LOCA) to determine the sensitivity of PCT to accumulator pressure. This study, in addition to several others, is incorporated into a PCT response surface in order to generate a 95/95 PCT. The maximum nitrogen cover pressure limit prevents accumulator relief valve actuation, and ultimately preserves accumulator integrity.

The effects on containment mass and energy releases from the accumulators are accounted for in the appropriate analyses (Ref. 2).

The accumulators satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO The LCO establishes the minimum conditions required to ensure that the accumulators are available to accomplish their core cooling safety function following a LOCA. Three accumulators are required to ensure that 100% of the contents of two of the accumulators will reach the core during a LOCA. This is consistent with the assumption that the contents of one accumulator spill through the break. If less than two accumulators are injected during the blowdown phase of a LOCA, the ECCS acceptance criteria of 10 CFR 50.46 (Ref. 2) could be violated.

For an accumulator to be considered OPERABLE, the isolation valve must be fully open, power removed above 2000 psig, and the limits established in the SRs for contained volume, boron concentration, and nitrogen cover pressure must be met.

Farley Units 1 and 2 B 3.5.1-4 Revision 0

Accumulators B 3.5.1 BASES APPLICABILITY In MODES 1 and 2, and in MODE 3 with RCS pressure > 1000 psig, the accumulator OPERABILITY requirements are based on full power operation. Although cooling requirements decrease as power decreases, the accumulators are still required to provide core cooling as long as elevated RCS pressures and temperatures exist.

This LCO is only applicable at pressures > 1000 psig. At pressures 1000 psig, the rate of RCS blowdown is such that the ECCS pumps can provide adequate injection to ensure that peak clad temperature remains below the 10 CFR 50.46 (Ref. 2) limit of 2200°F.

The Accumulator Applicability is modified by a Note which takes exception to the LCO requirements for the Accumulators to be OPERABLE in MODE 3 with RCS pressure above 1,000 psig for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during the performance of isolation valve testing required by SR 3.4.14.1. The applicability of the Note is restricted solely to the isolation valve testing required by SR 3.4.14.1. In order to perform the required isolation valve testing, the Accumulators must be isolated and various parameters (e.g., pressure, level) must be adjusted. The exception provided by this Note allows operation in MODE 3 with RCS pressure above 1,000 psig for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with Accumulators not configured per the requirements of the LCO such that the Actions for an inoperable Accumulator are not applicable.

In MODE 3, with RCS pressure 1000 psig, and in MODES 4, 5, and 6, the accumulator motor operated isolation valves are closed to isolate the accumulators from the RCS. This allows RCS cooldown and depressurization without discharging the accumulators into the RCS or requiring depressurization of the accumulators.

ACTIONS A.1 If the boron concentration of one accumulator is not within limits, it must be returned to within the limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . In this Condition, ability to maintain subcriticality or minimum boron precipitation time may be reduced. An average boron concentration for the injected water is assumed in the Best Estimate LOCA (large break LOCA) analysis. One accumulator up to 100 ppm below the minimum boron concentration limit, however, will have no effect on available ECCS water and an insignificant effect on post-LOCA core subcriticality. The large main steam line break analysis predicts that the accumulators would discharge following the event. However, their impact is minor and not a design limiting event. Thus, 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allowed to return the boron concentration to within limits.

(continued)

Farley Units 1 and 2 B 3.5.1-5 Revision 0

Accumulators B 3.5.1 BASES ACTIONS B.1 (continued)

If one accumulator is inoperable for a reason other than boron concentration, the accumulator must be returned to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . In this Condition, the required contents of two accumulators cannot be assumed to reach the core during a LOCA.

Due to the severity of the consequences should a LOCA occur in these conditions, the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time to open the valve, remove power to the valve, or restore the proper water volume or nitrogen cover pressure ensures that prompt action will be taken to return the inoperable accumulator to OPERABLE status. The Completion Time minimizes the potential for exposure of the plant to a LOCA under these conditions. The 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowed to restore an inoperable accumulator to OPERABLE status is justified in WCAP-15049-A, Rev. 1 (Ref. 3).

C.1 and C.2 If the accumulator cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and RCS pressure reduced to 1000 psig within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

D.1 If more than one accumulator is inoperable, the plant is in a condition outside the accident analyses; therefore, LCO 3.0.3 must be entered immediately.

SURVEILLANCE SR 3.5.1.1 REQUIREMENTS Each accumulator valve should be verified to be fully open. This verification ensures that the accumulators are available for injection and ensures timely discovery if a valve should be less than fully open. If an isolation valve is not fully open, the rate of injection to the RCS would be reduced. Although a motor operated valve position should not change with power removed, a closed valve could result in not meeting accident analyses assumptions. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

(continued)

Farley Units 1 and 2 B 3.5.1-6 Revision 52

Accumulators B 3.5.1 BASES SURVEILLANCE SR 3.5.1.2 and SR 3.5.1.3 REQUIREMENTS (continued) The borated water volume and nitrogen cover pressure are verified for each accumulator. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.5.1.4 The boron concentration should be verified to be within required limits for each accumulator since the static design of the accumulators limits the ways in which the concentration can be changed. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. Sampling the affected accumulator within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a 12%

level, indicated, increase (approximately 1% of tank volume) will identify whether inleakage has caused a reduction in boron concentration to below the required limit. It is not necessary to verify boron concentration if the added water inventory is from the refueling water storage tank (RWST), when the water contained in the RWST is within the accumulator boron concentration requirements. This is consistent with the recommendation of NUREG-1366 (Ref. 4).

SR 3.5.1.5 Verification that power is removed from each accumulator isolation valve operator when the pressurizer pressure is 2000 psig ensures that an active failure could not result in the undetected closure of an accumulator motor operated isolation valve. If this were to occur, only one accumulator would be available for injection given a single failure coincident with a LOCA. Therefore, each isolation valve operator is disconnected by a locked open disconnect device. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR allows power to be supplied to the motor operated isolation valves when RCS pressure is < 2000 psig, thus allowing operational flexibility by avoiding unnecessary delays to manipulate the breakers during plant startups or shutdowns.

(continued)

Farley Units 1 and 2 B 3.5.1-7 Revision 52

Accumulators B 3.5.1 BASES SURVEILLANCE SR 3.5.1.5 (continued)

REQUIREMENTS Should closure of a valve occur below 2000 psig, the SI signal provided to the valves would open a closed valve in the event of a LOCA.

REFERENCES 1. FSAR, Chapter 15.

2. 10 CFR 50.46

3. WCAP-15049-A, Rev. 1, April 1999.

4. NUREG-1366, February 1990.

Farley Units 1 and 2 B 3.5.1-8 Revision 22

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

79. 008AG2.2.22 079 Unit 1 is in Mode 2 at 3% power when the following occurs:

PCV-444B, PRZR PORV, fails open.

The OATC performs actions of AOP-100, Instrumentation Malfunction.

PCV-444B cannot be manually closed.

The OATC closes MOV-8000B, PRZR PORV ISO.

RCS pressure is rising.

Which one of the following completes the statements below per Tech Spec 3.4.11, Pressurizer Power Operated Relief Valves (PORVs)?

PRZR PORV ISO, MOV-8000B is required to be closed with power (1) .

Entry into Mode 1 (2) allowed by Tech Specs.

(1) (2)

A. removed is NOT B. maintained is NOT C. removed IS D. maintained IS Tuesday, March 19, 2013 11:38:42 AM 12

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 LCO 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 (continued) 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.

A. Correct 1) Correct, per TS 3.4.11, Condition B, the PORV block valve must be closed and power removed from the MOV.

2) Correct, with the PORV INOPERABLE, and TS 3.4.11 not having specific guidance that a Mode change is allowed with this failure, a mode change is not allowed per TS 3.0.4. Operation for an unlimited amount of time is not allowed for this condition.

B. Incorrect 1) Incorrect, plausible since this is the action required for a leaking PORV.

2) Correct, see A.2.

C. Incorrect 1) Correct, see A.1.

2) Incorrect, plausible since this would be acceptable for a leaking PORV.

Operation for an unlimited amount of time with a leaking PORV isolated and power maintained to the Block Valve is allowed, so TS 3.0.4 allows a Mode change for that condition.

D. Incorrect 1) Incorrect, see B.1.

2) Incorrect, see C.2.

K/A: 008AG2.2.22 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open)

Knowledge of limiting conditions for operations and safety limits.

l (CFR: 41.5 / 43.2 / 45.2)

Tuesday, March 19, 2013 11:38:42 AM 13

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 IMPORTANCE RO 4.0 SRO 4.7 Importance Rating: 4.0 4.7 Technical

Reference:

Tech Specs v190/186 Tech Spec Bases v58 References provided: None Learning Objective: RECALL AND APPLY the information from the LCO BASES sections: BACKGROUND, APPLICABLE SAFETY ANALYSIS, ACTIONS, SURVEILLENCE REQUIREMENTS, for any Technical Specifications or TRM requirements associated with the Pressurizer System components and attendant equipment alignment, to include the following (OPS-62101E01): 10CFR55.43 (b) 2 3.4.9, Pressurizer 3.4.10, Pressurizer Safety Valves 3.4.11, Pressurizer Power Operated Relief Valves 13.4.2, Pressurizer 13.4.4, Safety Valves - Shutdown Question History: Modified from Vogtle 2012 NRC Exam question K/A match: A Pressurizer PORV has stuck open (Relief Valve Stuck Open). The candidate displays "knowledge of limiting conditions for operations" by determining the TS 3.4.11 requirements for the stuck open PORV, and in addition the candidate is questioned on TS 3.0.4 requirements - is a Mode change allowed?

SRO justification: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).

The first part of the question is < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months TS information for actions required for a stuck open PORV. This is RO knowledge. The second part of the question meets the K/A at the SRO level. SRO knowledge is required to determine if a Mode change is allowed by applying TS 3.0.4. The plant cannot be operated in this condition for an unlimited amount of time, so a Mode change is not allowed per TS 3.0.4.

2013 NRC exam From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 Tuesday, March 19, 2013 11:38:42 AM 14

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 flowchart for 10 CFR 55.43(b)(2) :

1) can NOT be answered by knowing less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Tech Specs.

The first part of the question is < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months TS information for actions required for a stuck open PORV. The second part of the question meets the K/A at the SRO level.

2) can NOT be answered by knowing information listed "above-the-line".

3) can NOT be answered by knowing the TS Safety Limits or their bases.

4) Does involve one or more of the following for TS, TRM or ODCM:

Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).

Tuesday, March 19, 2013 11:38:42 AM 15

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

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

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS

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

Separate Condition entry is allowed for each PORV.

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

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

manually cycled.

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

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

Operation for an unlimited amount of time is not allowed in this conditon, but would be allowed if the candidate thought CONDITION A was correct.

Farley Units 1 and 2 3.4.11-1 Amendment No. 170 (Unit 1)

Amendment No. 163 (Unit 2)

Pressurizer PORVs 3.4.11 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. One block valve C.1 Place associated PORV 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable. in manual control.

AND C.2 Restore block valve to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OPERABLE status.

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

D.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months E. Two PORVs inoperable E.1 Close associated block 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and not capable of being valves.

manually cycled.

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

AND E.3 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND E.4 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Farley Units 1 and 2 3.4.11-2 Amendment No. 146 (Unit 1)

Amendment No. 137 (Unit 2)

Pressurizer PORVs 3.4.11 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME F. More than one block valve F.1 Place associated 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable. PORVs in manual control.

AND F.2 Restore one block valve 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to OPERABLE status.

AND F.3 Restore remaining 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months block valve to OPERABLE status.

G. Required Action and G.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition F not AND met.

G.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1 -------------------------------NOTES-----------------------------

1. Not required to be met with block valve closed in accordance with the Required Action of Condition B or E.

2. Not required to be performed prior to entry into MODE 3.

3. Not required to be performed for Unit 2 for the remainder of operating cycle 16 for block valve Q2B31MOV8000B.

Perform a complete cycle of each block valve. In accordance with the Surveillance Frequency Control Program Farley Units 1 and 2 3.4.11-3 Amendment No. 185 (Unit 1)

Amendment No. 180 (Unit 2)

Pressurizer PORVs 3.4.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.2 ----------------------------NOTE---------------------------------

Not required to be performed prior to entry into MODE 3.

Perform a complete cycle of each PORV during In accordance with MODE 3 or 4. the Surveillance Frequency Control Program SR 3.4.11.3 Perform a complete cycle of each PORV using the In accordance with backup PORV control system. the Surveillance Frequency Control Program SR 3.4.11.4 ----------------------------NOTE---------------------------------

Required to be performed only for Unit 2 for the remainder of operating cycle 16.

Check power available to the Unit Two PORV block 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months valve Q2B31MOV8000B.

Farley Units 1 and 2 3.4.11-4 Amendment No. 185 (Unit 1)

Amendment No. 180 (Unit 2)

LCO Applicability 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2, 3.0.7, and 3.0.8.

LCO 3.0.2 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6.

If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated.

LCO 3.0.3 When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to place the unit, as applicable, in:

a. MODE 3 within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months ;

b. MODE 4 within 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months ; and

c. MODE 5 within 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months .

Exceptions to this Specification are stated in the individual Specifications.

Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LCO 3.0.3 is not required.

LCO 3.0.3 is only applicable in MODES 1, 2, 3, and 4.

LCO 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; Not allowed to operate for an unlimited amount of time with Inoperable PORV and closed Block valve. (continued)

Farley Units 1 and 2 3.0-1 Amendment No. 173 (Unit 1)

Amendment No. 166 (Unit 2)

LCO Applicability 3.0 3.0 LCO APPLICABILITY LCO 3.0.4 b. After performance of a risk assessment addressing inoperable (continued) systems and components, consideration of the results, determination of the acceptability of entering the MODE or other No Risk Assessment specified condition in the Applicability, and establishment of risk has been performed management actions, if appropriate; exceptions to this Specification are stated in the individual Specifications, or Not allowed per the c. When an allowance is stated in the individual value, parameter, or TS 3.4.11 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.

LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY.

LCO 3.0.6 When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered.

Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, an evaluation shall be performed in accordance with specification 5.5.15, "Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.

LCO 3.0.7 Test Exception LCO 3.1.8 allows specified Technical Specification (TS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements remain unchanged. Compliance with Test Exception LCOs is optional. When a Test Exception LCO is desired to be met but is not met, the ACTIONS of the Test Exception LCO shall be met. When a Test Exception LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall be made in accordance with the other applicable Specifications.

Farley Units 1 and 2 3.0-2 Amendment No. 173 (Unit 1)

Amendment No. 166 (Unit 2)

QUESTIONS Original Vogtle 2012 NRC Exam question REPORT that was used to develop 008AG2.2.22 for HL-17 NRC Exam

1. 008AG2.1.19 001 Initial conditions:

- Unit 1 is at 100% power, all systems normal.

- ALB12-D03, PRZR PRESS LO PORV BLOCK illuminates.

- ALB12-E01, PRZR RELIEF DISCH HI TEMP illuminates.

- All PRZR Relief temperatures are rising.

- PRZR PORV 455A has been placed in the closed position.

- PRZR PORV 455A indicates half red / hollow and half green / solid on the IPC computer point.

- PRZR pressure cycling between 2185 psig to 2200 psig.

Current conditions:

- PRZR PORV 455A Block Valve has been closed.

- PRZR pressure is 2208 psig and slowly rising.

- All PRZR Relief temperatures have stabilized and begun to lower.

Which one of the following correctly completes the following statement?

Per Tech Spec 3.4.11, "Pressurizer Power Operated Relief Valves (PORVs)", PRZR PORV 455A Block Valve ___(1)___ required to be de-energized and based on the currrent conditions, if an RCS pressure transient were to occur, "B" PORV 456A will lift at an RCS pressure of ____(2)___ psig.

A. (1) is NOT (2) 2345 B. (1) is NOT (2) 2335 C. (1) is (2) 2345 D. (1) is (2) 2335 Tuesday, March 19, 2013 1:27:11 PM 1

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

80. 024AA2.06 080 Unit 1 has been in operation at 100% power for 3 months after a Refueling Outage.

The following conditions exist:

At 1000:

A Reactor Trip occurs.

At 1400:

RCS Tavg is 547°F and stable.

Source Range counts have been slowly increasing for the last two hours and is currently 320 cps.

STP-29.1, Shutdown Margin Calculation (TAVG 547°F), has determined that Shutdown Margin is (-)1550 pcm.

Which one of the following completes the statements below?

Source Range counts are rising due to (1) .

Per the BASES of Tech Spec 3.1.1, Shutdown Margin (SDM), one of the accidents that the SDM requirement must protect against is (2) .

A. 1) TCV-3083, LTDN HX CCW TEMP CONT, valve has failed open

2) an uncontrolled Control Rod withdrawal B. 1) Xenon decay

2) an uncontrolled Control Rod withdrawal C. 1) TCV-3083, LTDN HX CCW TEMP CONT, valve has failed open

2) an inadvertent boron dilution D. 1) Xenon decay

2) an inadvertent boron dilution Tuesday, March 19, 2013 11:38:42 AM 16

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 TCV-3083 is the temperature control valve for CCW cooling to the Letdown Heat Exchanger. This valve is controlled automatically per a setpoint demanded by a controller on the MCB. This valve fails open on a loss of air. If the valve fails open, it will cause more CCW cooling flow through the Letdown HX and cause Letdown temperature to go down, thus causing more boron absorption in the Letdown demins and causing Source Range counts to rise.

A. Incorrect 1) Correct, TCV-3083 failing open will cause excessive cooling of the Letdown flow, and a resultant increased absorption of boron in the Letdown demineralizers. This will equate to a dilution and cause Source Range counts to rise.

2) Incorrect, plausible because one of the accident scenarios evaluated for adequate SDM is a Control Rod ejection accident. This distracter is very similar. Both of the items mentioned will add positive reactivity. In addition, a candidate may think that Control Rods stepping out would reduce the SDM, but actually SDM should be basically unaffected. If Control rods step out during an uncontrolled Control Rod withdrawal, they are still available to insert negative reactivity upon a Reactor trip, and SDM should be unaffected.

B. Incorrect 1) Incorrect, Xenon will build in for the same hours as the square root of the original power level, i.e. if original power level is 100% (and stable for an extended period of time), xenon will build in for 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months . After that point Xenon will turn and start to decay away. In this case the Reactor trip occurred 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ago, so Xenon is still steadily building in.

Plausible that a candidate would determine Xenon would be decaying away since that is what it would be doing after a 10 hour1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months time frame. This could add positive reactivity and cause Source Range counts to increase.

2) Incorrect, see A.2.

C. Correct 1) Correct, see A.1.

2) Correct, one of the TS Bases accident conditions evaluated to determine required SDM to prevent core damage is an inadvertent boron dilution.

D. Incorrect 1) Incorrect, see B.1.

2) Correct, see C.2.

K/A: 024AA2.06 Emergency Boration:

Ability to determine and interpret the following as they apply Tuesday, March 19, 2013 11:38:42 AM 17

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 to the Emergency Boration:

(CFR: 43.5 / 45.13)

AA2.06 When boron dilution is taking place . . . . . . 3.6 3.7 Importance Rating: 3.6 3.7 Technical

Reference:

FNP-1-STP-29.1, v38 Tech Specs v190/186 Tech Spec Bases v58 COLR v24-1 / 22-1 References provided: None Learning Objective: DETERMINE AND APPLY the information from the LCO BASES sections: BACKGROUND, APPLICABLE SAFETY ANALYSIS, ACTIONS, SURVEILLENCE REQUIREMENTS, for any Technical Specifications or TRM requirements associated with AOP-27, Emergency Boration, components and attendant equipment. (OPS-62521A01) 10CFR55.43 (b) 2 Question History: New question Basis for meeting K/A: This question tests the candidate's ability to evaluate conditions relating to a boron dilution event causing Shutdown Margin to be inadequate. An Emergency Boration is required due to Shutdown Margin being less than the COLR required minimum. The candidate has to determine that a boron dilution event is occurring. In addition, the candidate is tested on knowledge of TS Bases accident analyses for SDM. One of the Bases requirements of adequate SDM is to allow the operator sufficient time to stop an inadvertent dilution prior to potential core damage.

SRO justification: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

Knowledge of TS bases that is required to analyze TS required actions and terminology.

TS Bases knowledge is tested for analysis of accidents for which adequate SDM is a limiting condition.

2013 NRC exam From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 flowchart for 10 CFR 55.43(b)(2) :

1) can NOT be answered by knowing less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Tech Specs.

2) can NOT be answered by knowing information listed "above-the-line".

3) can NOT be answered by knowing the TS Safety Limits or their bases.

4) Does involve one or more of the following for TS, TRM or ODCM:

Tuesday, March 19, 2013 11:38:42 AM 18

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

Knowledge of TS bases that is required to analyze TS required actions and terminology Tuesday, March 19, 2013 11:38:42 AM 19

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after a trip from 100%

power, Xenon is still building in, thus adding negative reactivity.

SDM 3.1.1 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM)

LCO 3.1.1 SDM shall be within the limits provided in the COLR.

APPLICABILITY: MODE 2 with keff < 1.0, MODES 3, 4, and 5.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SDM not within limit. A.1 Initiate boration to restore Immediately SDM to within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.1.1 Verify SDM to be within limits. In accordance with the Surveillance Frequency Control Program Farley Units 1 & 2 3.1.1-1 Amendment No. 185 (Unit 1)

Amendment No. 180 (Unit 2)

SDM B 3.1.1 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM)

BASES BACKGROUND According to GDC 26 (Ref. 1), the reactivity control systems must be redundant and capable of holding the reactor core subcritical when shut down under cold conditions. Maintenance of the SDM ensures that postulated reactivity events will not damage the fuel.

SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences (AOOs). As such, the SDM defines the degree of subcriticality that would be obtained immediately following the insertion or trip of all shutdown and control rods, assuming that the single rod cluster assembly of highest reactivity worth is fully withdrawn.

The system design requires that two independent reactivity control systems be provided, and that one of these systems be capable of maintaining the core subcritical under cold conditions. These requirements are provided by the use of movable control assemblies and soluble boric acid in the Reactor Coolant System (RCS). The Rod Control System can compensate for the reactivity effects of the fuel and water temperature changes accompanying power level changes over the range from full load to no load. In addition, the Rod Control System, together with the boration system, provides the SDM during power operation and is capable of making the core subcritical rapidly enough to prevent exceeding acceptable fuel damage limits, assuming that the rod of highest reactivity worth remains fully withdrawn. The Chemical and Volume Control System can control the soluble boron concentration to compensate for fuel depletion during operation and all xenon burnout reactivity changes and can maintain the reactor subcritical under cold conditions.

During power operation, SDM control is ensured by operating with the shutdown banks fully withdrawn and the control banks within the limits of LCO 3.1.6, "Control Bank Insertion Limits." When the unit is in the shutdown and refueling modes, the SDM requirements are met by means of adjustments to the RCS boron concentration.

Farley Units 1 and 2 B 3.1.1-1 Revision 0

SDM B 3.1.1 BASES APPLICABLE The minimum required SDM is assumed as an initial condition in SAFETY ANALYSES safety analyses. The safety analysis (Ref. 2) establishes an SDM that ensures specified acceptable fuel design limits are not exceeded for normal operation and AOOs, with the assumption of the highest worth rod stuck out on a trip.

For MODE 5, the primary Safety Analysis that relies on the SDM limits is the boron dilution analysis.

The acceptance criteria for the SDM requirements are that specified acceptable fuel design limits are maintained. This is done by ensuring that:

a. The reactor can be made subcritical from all operating conditions, transients, and Design Basis Events;

b. The reactivity transients associated with postulated accident conditions are controllable within acceptable limits (departure from nucleate boiling ratio (DNBR), fuel centerline temperature limits for AOOs, and less than 200 cal/gm, thus meeting the NRC acceptance criteria of 280 cal/gm average fuel pellet enthalpy at the hot spot for the rod ejection accident); and

c. The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

An Operating Procedure (Ref. 5) assures sufficient operator action time for the mitigation of an uncontrolled boron dilution event (Ref. 3) in MODE 5. This procedure is independent of SDM and uses the RHR system flowrate, and the calculated critical boron concentration to specify a minimum allowable boron concentration.

The most limiting accident for the SDM requirements is based on a guillotine break of a main steam line (MSLB) inside containment initiated at the end of core cycle life with RCS average temperature at no-load operating temperature, as described in the accident analysis (Ref. 2). The increased steam flow resulting from a pipe break in the main steam system causes an increased energy removal from the affected steam generator (SG), and consequently the RCS. This results in a reduction of the reactor coolant temperature. The resultant (continued)

Farley Units 1 and 2 B 3.1.1-2 Revision 0

SDM B 3.1.1 BASES APPLICABLE coolant shrinkage causes a reduction in pressure. In the presence SAFETY ANALYSES of a negative moderator temperature coefficient, this cooldown (continued) causes an increase in core reactivity. As RCS temperature decreases, the severity of an MSLB decreases until the MODE 5 value is reached. The most limiting MSLB, with respect to potential fuel damage before a reactor trip occurs, is a guillotine break of a main steam line inside containment initiated at the end of core life.

The positive reactivity addition from the moderator temperature decrease will terminate when the affected SG boils dry, thus terminating RCS heat removal and cooldown. Following the MSLB, a post trip return to power may occur; however, no fuel damage occurs as a result of the post trip return to power, and that the Safety Limit (SL) requirement of SL 2.1.1 is met.

In addition to the limiting MSLB transient, the SDM requirement must also protect against:

a. Inadvertent boron dilution; and

b. Rod ejection.

Each of these events is discussed below.

In the boron dilution analysis (Ref. 3), the required SDM defines the reactivity difference between an initial subcritical boron concentration and the corresponding critical boron concentration. These values, in conjunction with the configuration of the RCS and the assumed dilution flow rate, directly affect the results of the analysis. This event is most limiting at the beginning of core life, when critical boron concentrations are highest. For each cycle of operation at Farley Nuclear Plant, the minimum boron concentrations that are required in MODES 4 and 5 to allow 15 minutes operator action time are given in the Nuclear Design Report for that cycle.

The ejection of a control rod rapidly adds reactivity to the reactor core, causing both the core power level and heat flux to increase with corresponding increases in reactor coolant temperatures and pressure. The ejection of a rod also produces a time dependent redistribution of core power.

(continued)

Farley Units 1 and 2 B 3.1.1-3 Revision 0

SDM B 3.1.1 BASES APPLICABLE SDM satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). Even though SAFETY ANALYSES it is not directly observed from the control room, SDM is considered (continued) an initial condition process variable because it is periodically monitored to ensure that the unit is operating within the bounds of accident analysis assumptions.

With Tavg less than 200°F, the reactivity transients resulting from a postulated steam line break cooldown are minimal, and a 1% delta k/k SHUTDOWN MARGIN provides adequate protection.

LCO SDM is a core design condition that can be ensured during operation through control rod positioning (control and shutdown banks) and through the soluble boron concentration.

The MSLB (Ref. 2) and the boron dilution (Ref. 3) accidents are the most limiting analyses that establish the SDM value of the LCO. For MSLB accidents, if the LCO is violated, there is a potential to exceed the DNBR limit and to exceed 10 CFR 100, "Reactor Site Criteria,"

limits (Ref. 4). For the boron dilution accident, if the LCO is violated, the minimum required time assumed for operator action to terminate dilution may no longer be applicable.

APPLICABILITY In MODE 2 with keff < 1.0 and in MODES 3, 4, and 5, the SDM requirements are applicable to provide sufficient negative reactivity to meet the assumptions of the safety analyses discussed above. In MODE 6, the shutdown reactivity requirements are given in LCO 3.9.1, "Boron Concentration." In MODES 1 and 2, SDM is ensured by complying with LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, Control Bank Insertion Limits.

ACTIONS A.1 If the SDM requirements are not met, boration must be initiated promptly. A Completion Time of Immediately is adequate to ensure prompt operator action to correctly align and start the required (continued)

Farley Units 1 and 2 B 3.1.1-4 Revision 0

SDM B 3.1.1 BASES ACTIONS A.1 (continued) systems and components. It is assumed that boration will be continued until the SDM requirements are met.

In the determination of the required combination of boration flow rate and boron concentration, there is no unique requirement that must be satisfied. Since it is imperative to raise the boron concentration of the RCS as soon as possible, the flowpath of choice would utilize a highly concentrated solution, such as that normally found in the boric acid storage tank, or the refueling water storage tank. The operator should borate with the best source available for the plant conditions.

In determining the boration flow rate, the time in core life must be considered. For instance, the most difficult time in core life to increase the RCS boron concentration is at the beginning of cycle when the boron concentration may approach or exceed 2000 ppm.

For example, if the emergency boration path is used, the CVCS is capable of inserting negative reactivity at a rate of approximately 65 pcm/min when the RCS boron concentration is 1000 ppm and approximately 75 pcm/min when the RCS boron concentration is 100 ppm.

SURVEILLANCE SR 3.1.1.1 REQUIREMENTS In MODES 1 and 2, SDM is verified by observing that the requirements of LCO 3.1.5 and LCO 3.1.6 are met. In the event that a rod is known to be untrippable, however, SDM verification must account for the worth of the untrippable rod as well as another rod of maximum worth.

In MODES 3, 4, and 5, the SDM is verified by performing a reactivity balance calculation, considering the listed reactivity effects:

a. RCS boron concentration;

b. Control bank position;

c. RCS average temperature;

d. Fuel burnup based on gross thermal energy generation; (continued)

Farley Units 1 and 2 B 3.1.1-5 Revision 0

SDM B 3.1.1 BASES SURVEILLANCE SR 3.1.1.1 (continued)

REQUIREMENTS

e. Xenon concentration;

f. Samarium concentration; and

g. Isothermal temperature coefficient (ITC).

Using the ITC accounts for Doppler reactivity in this calculation because the reactor is subcritical, and the fuel temperature will be changing at the same rate as the RCS.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 26.

2. FSAR, Section 15.4.2.

3. FSAR, Section 15.2.4.

4. 10 CFR 100.

5. Letter from D.E. McKinnon to L.K. Mathews, Operating Procedure for Mode 4/5 Boron Dilution, 90 AP*-G-0041, July 6, 1990.

Farley Units 1 and 2 B 3.1.1-6 Revision 52

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

81. 028G2.1.1 081 Unit 2 is in Mode 5 with the following conditions:

CTMT Purge was secured for a Local Leak Rate Test (LLRT) and the LLRT Tagout has been cleared.

R-24A, Containment Radiation Monitor, has been declared INOPERABLE.

Prior to placing the CTMT Purge system in operation per SOP-12.2, Containment Purge and Pre-Access Filtration System, which one of the following requirements must be met?

A. R-24A is required to be restored to OPERABLE status.

B. Gaseous Release Permits are required to be issued.

C. Alternate sampling is required to be performed per the ODCM.

D. CTMT Pre-Access Filtration must always be placed in service prior to placing CTMT Purge in service.

Tuesday, March 19, 2013 11:38:42 AM 20

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 FNP-2-SOP-12.2 v38.0 A. Incorrect SOP-12.2 requires that either R-24A or B is in service prior to starting Containment Purge. A candidate could think that both rad monitors are required to be in service prior to placing Containment Purge in service. In MODE 5 with no CORE ALTERATIONS in progress, there are no TS requirements for R-24A or B to be in service. Plausible since this answer would be a correct statement if CORE ALTERATIONS or fuel movement were in progress B. Correct Current Gaseous Release Permits are required to be issued. This is plural because there is also a release permit that is in effect at all times when the Plant Vent Stack exhaust fans are running. Step 4.4.9 has the operator verify this. This is to comply with the ODCM requirement for sampling. The Gaseous Release Permit establishes that the Containment atmosphere has been sampled and its activity is within limits bounded by the ODCM, Table 3-3. (FNP-0-CCP-213.0)

C. Incorrect Plausible since SOP-12.2 has the operator verify R-14 and 22 are aligned for normal operation OR applicable ODCM or TS actions are being performed. These actions may include alternate sampling requirements.

Neither R-14 or R-22 is out of service. In addition, SOP-12.2 has alternate sampling requirements for R-14, R-22, or R-29B (step 2.4.3 of Appendix 3). Plausible that the candidate may think these requirements are for R-24A and R-24B.

In MODE 5 with no CORE ALTERATIONS in progress, there are no TS requirements for R-24A or B to be in service.

D. Incorrect Plausible, SOP-12.2 states:

"IF the activity level within containment is determined to be excessive by sample analysis, THEN operate the containment pre-access filtration system per Section 4.1 as necessary to reduce the activity level as specified by Health Physics."

Placing Pre-Access Filtration in service is not always required, but is required when CTMT activity is excessive.

K/A: 028G2.1.1 Hydrogen Recombiner and Purge Control System (HRPS)

Knowledge of conduct of operations requirements.

(CFR: 41.10 / 45.13)

IMPORTANCE RO 3.8 SRO 4.2 Importance Rating: 3.8 4.2 Technical

Reference:

FNP-2-SOP-12.2, v38 References provided: None Learning Objective: RECALL AND APPLY the information from the LCO BASES Tuesday, March 19, 2013 11:38:42 AM 21

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 sections: BACKGROUND, APPLICABLE SAFETY ANALYSIS, ACTIONS, SURVEILLENCE REQUIREMENTS, for any Technical Specifications or TRM requirements associated with Miscellaneous Ventilation System components and attendant equipment alignment, to include the following (OPS-62107D01): 10CFR55.43 (b) 2 13.3.4, Radiation Monitoring Instrumentation 13.7.5, Area Temperature Monitoring (Unit 2 Only)

Question History: FNP Bank question from HLT-28, 2004 NRC Exam CMNT VENT-62107A02 01 K/A match: Candidate is questioned on the requirements of a Containment atmosphere release. Per the Farley ODCM, a Containment atmosphere release is not considered an effluent release, but there are ODCM requirements for sampling to ensure that the effects will not cause an effluent release to exceed limits in the ODCM. Candidate has to have knowledge of the ODCM and actions required for conducting a release of the Containment atmosphere.

SRO justification: 10 CFR 55.43(b)(2):

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

Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).

Requires knowledge of Surveillance Requirements of the ODCM - the ODCM requires sampling of the Containment atmosphere prior to release. This is performed by CHEMISTRY, who then issues a release permit to allow the Containment atmosphere release.

2013 NRC exam From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 flowchart for 10 CFR 55.43(b)(2)

1) can NOT be answered by knowing less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Tech Specs.

2) can NOT be answered by knowing information listed "above-the-line".

3) can NOT be answered by knowing the TS Safety Limits or their bases.

4) Does involve one or more of the following for TS, TRM or ODCM:

Application of Required Actions (Section 3) and Surveillance Requirements (SR)

(Section 4) in accordance with rules of application requirements (Section 1)..

Tuesday, March 19, 2013 11:38:42 AM 22

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 June 13, 2011 Version 38.0 FARLEY NUCLEAR PLANT SYSTEM OPERATING PROCEDURE FNP-2-SOP-12.2 S

A See pages 2, 5, and page 2 of 5 of F Appendix 3. E T

Y CONTAINMENT PURGE AND PRE-ACCESS FILTRATION SYSTEM R

E L

A T

E D

PROCEDURE LEVEL OF USE CLASSIFICATION PER NMP-AP-003 CATEGORY SECTIONS Continuous: ALL

Reference:

NONE Information: NONE Approved:

David L Reed (for)

Operations Manager Date Issued: February 23, 2012

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 TABLE OF CONTENTS Procedure Contains Number of Pages Body ........................................................13 Appendix 1 ................................................5 Appendix 2 ................................................2 Appendix 3 ................................................5 Page 1 of 1 Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 FARLEY NUCLEAR PLANT UNIT 2 SYSTEM OPERATING PROCEDURE SOP-12.2 CONTAINMENT PURGE AND PRE-ACCESS FILTRATION SYSTEM 1.0 PURPOSE This procedure provides the Initial Conditions, Precautions and Limitations, and Instructions for operation of the Containment Purge and Pre-Access Filtration System.

Instructions are included in the following sections:

4.1 Containment Pre-Access Filtration Operation 4.2 Containment Radiation Monitor R-11 and R-12 Operation 4.3 Containment/Atmosphere Differential Pressure Instrument PDI-3317 Operation 4.4 Placing Containment Purge System in Operation 4.5 Removing Containment Purge System from Operation 4.6 Restoring Containment Purge Following a Safety Injection or Hi Hi Radiation System 4.7 Placing Containment Mini-Purge System in Operation 4.8 Removing Containment Mini-Purge System from Operation 4.9 Restoring Containment Mini-Purge Following a Safety Injection or Hi Hi Radiation Signal Appendix 1 Contingency Operation of Containment Mini-Purge with Supply Fan Unavailable Appendix 2 Contingency Operation of Containment Main Purge with Supply Fan Unavailable with Containment Integrity Not Required Appendix 3 Batch Releases of Containment Atmosphere Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 2.0 INITIAL CONDITIONS 2.1 The electrical distribution system is energized and aligned for normal operation per FNP-2-SOP-36.0, PLANT ELECTRICAL DISTRIBUTION LINE-UP, with exceptions noted.

2.2 The compressed air system is in service and aligned for normal operation per FNP-2-SOP-31.0, COMPRESSED AIR SYSTEM.

2.3 The containment purge and pre-access filtration system valves and electrical distribution systems are aligned per system checklist FNP-2-SOP-12.2A with exceptions noted.

2.4 The auxiliary building main exhaust system is in service per FNP-2-SOP-58.0, AUXILIARY BUILDING HVAC SYSTEM.

2.5 Radiation monitors (R-14, R-22, R-24A and R-24B) are aligned for normal Plausibility for operation per FNP-2-SOP-45.0, RADIATION MONITORING SYSTEM, or ODCM Alternate applicable ODCM or technical specifications action statement is being met.

actions 2.6 The plant hot water heating system is aligned for normal operation as required by seasonal conditions per FNP-2-SOP-59.0, PLANT HOT WATER HEATING SYSTEM (applicable only to section 4.4).

3.0 PRECAUTIONS AND LIMITATIONS 3.1 The following rules apply to operation of the Auxiliary Building Main Exhaust fans, in relation to the CTMT Purge system:

3.1.1. Both Auxiliary Building Main Exhaust Fans must be and should only be placed in service when CTMT main purge is placed in service to prevent an overpressurization or excessive vacuum from being developed inside the exhaust plenum. (NEL 99-0078; REA 98 1808).

3.1.2. There is one exception to this rule: WHEN the Spent Fuel Pool Ventilation system is NOT operating, and the CTMT Main Purge system is in operation in Lo-Speed, THEN only one main exhaust fan is required (NEL 99 0078) 3.1.3. For the normal situation when the Spent Fuel Ventilation and Radwaste Ventilation fans are running, THEN both Auxiliary Building Main Exhaust fans shall be in operation to operate CTMT Main Purge.

3.2 A gaseous waste release permit must be obtained prior to purge or mini-purge of CTMT.

Correct answer Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 NOTE: FNP-2-UOP-1.1 and FNP-2-UOP-2.2 control the associated links as necessary to inhibit CTMT Main Purge damper operation when RCS temperature is > 200°F.

3.3 The CTMT main purge system can only be used when RCS temperature is 200°F.

3.4 Evaluation on a case by case basis may be necessary for the release of smoke or other chemicals not considered VOCs that could contaminate the charcoal adsorbers or HEPA filter.

3.5 The main and mini-purge supply and exhaust dampers must not be opened with CTMT TO ATMOS 'P greater than 0.4 psid.

3.6 The mini-purge supply and exhaust fans must not be started with CTMT TO ATMOS 'P greater than 0.25 psid.

3.7 IF unable to commence a batch release of the CTMT per Appendix 3 prior to exceeding 0.4 psid, THEN go to FNP-2-SOP-10.0 Appendix 1 for instructions for venting the CTMT.

3.8 Following a CTMT ILRT, CTMT purge must be verified vented prior to being placed in service. (AI 2009207430)

Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 4.0 INSTRUCTIONS 4.1 CONTAINMENT PRE-ACCESS FILTRATION OPERATION:

NOTE: The local control panels for the filtration units are at the following locations:

2A CTMT Pre-Access Filtration Unit - 155 ft by 2C MCC

2B CTMT Pre-Access Filtration Unit - 121 ft by 2D MCC 4.1.1 IF desired to place CTMT pre-access filtration in service, THEN start the desired filtration unit(s):

2A CTMT Pre-Access Filtration Unit N2T50M001A

2B CTMT Pre-Access Filtration Unit N2T50M001B 4.1.2 IF desired to remove CTMT pre-access filtration from service, THEN stop the filtration unit(s):

2A CTMT Pre-Access Filtration Unit N2T50M001A

2B CTMT Pre-Access Filtration Unit N2T50M001B 4.2 CONTAINMENT RADIATION MONITOR R-11 AND R-12 OPERATION:

4.2.1 Align R-11/12 flow path for operation per FNP-2-SOP-45.0.

4.2.2 Remove R-11/12 flow path from operation per FNP-2-SOP-45.0.

4.3 CONTAINMENT/ATMOSPHERE DIFFERENTIAL PRESSURE INSTRUMENT PDI-3317 OPERATION:

4.3.1 IF desired to place CTMT TO ATMOS 'P N2C14PDI3317 in service, THEN open the following:

CTMT 'P ISO Q2E14MOV3318A

CTMT 'P ISO Q2E14MOV3318B 4.3.2 IF desired to remove CTMT TO ATMOS 'P N2C14PDI3317 from service, THEN close the following:

CTMT 'P ISO Q2E14MOV3318A

CTMT 'P ISO Q2E14MOV3318B Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 4.4 PLACING CONTAINMENT PURGE SYSTEM IN OPERATION 4.4.1 IF CTMT Purge system was removed from operation for a short duration for maintenance or testing, THEN place the CTMT Purge system back in service by performing steps 4.4.13 through 4.4.21.

4.4.2 IF main purge supply and exhaust fans were tripped by temperature switches TSH-2280A, TSH-2280B, TSH-2281A, or TSH-2281B, THEN refer to FNP-2-FVP-9.0, CONTAINMENT AND PURGE EQUIPMENT ROOM AREAS.

4.4.3 Verify auxiliary building main exhaust system in service.

4.4.4 Verify plant vent stack radiation monitor R-14 is in service per FNP-2-SOP-45.0, RADIATION MONITORING SYSTEM.

4.4.5 Verify one of the following plant vent stack radiation monitors are in service:

x R22 per FNP-2-SOP-45.0, RADIATION MONITORING SYSTEM OR x R29B per FNP-0-CCP-641, OPERATIONS OF THE PLANT VENT STACK MONITORING SYSTEM 4.4.6 Verify radiation monitor R-24A or R-24B in service per FNP-2-SOP-45.0, RADIATION MONITORING SYSTEM.

Only one rad monitor, Either R24A or R24B 4.4.7 Verify CTMT PURGE SUPP/EXH FANS & DAMPERS handswitch is required selected to STOP.

4.4.8 Verify the following are in operation OR CR written for repair:

x Recorders RR0100 (points RR14 and RR22), RR24A, and RR24B OR x Recorders RR0100 (point RR14), RR24A, RR24B, and printer R29B Correct Answer 4.4.9 Verify gaseous waste release permit has been issued for CTMT purge operation.

4.4.10 IF the activity level within containment is determined to be excessive by sample analysis, THEN operate the containment pre-access filtration system per section 4.1 as necessary to reduce the activity level as specified by Health Physics.

Plausibility for placing Pre-Access Filtration in service Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 4.4.11 Verify that the alarm setpoints for R-24A and R-24B are less than or equal to the maximum allowable setpoints listed in Part II of the gaseous waste release permit. (See FNP-2-CCP-213.1, GASEOUS EFFLUENT RADIATION MONITORING SYSTEM SETPOINTS, for current alarm setpoints.)

4.4.12 Verify closed the following CTMT purge filter cooling valves:

x N2P13V293 x N2P13V294 4.4.13 Verify stopped MINI PURGE SUPP/EXH FAN.

4.4.14 Place CTMT PURGE DMPRS handswitch HS-3196 to FULL.

4.4.15 Place CTMT PURGE DMPRS handswitch HS-3198 to FULL.

CAUTION: Operation of CTMT main purge in High Speed with only one auxiliary building main exhaust fan available AND SFP ventilation in service will over pressurize the exhaust plenum. (OR 99750.2) 4.4.16 IF only one Auxiliary Building Main Exhaust Fan is available THEN secure SFP ventilation per FNP-2-SOP-58.0, AUXILIARY BUILDING HVAC SYSTEM, and then continue on in this procedure.

(Refer to P&L 3.13.1.)

NOTE: Step 4.4.18 below needs to be performed without delay after performing step 4.4.17, to avoid overpressurizing the exhaust plenum. Consideration should be given to the use of two operators in communication, if available, to perform these two steps.

CAUTION: Placing CTMT main purge supply and exhaust fans in service, with only one auxiliary building main exhaust fan in service will over pressurize the exhaust plenum, unless SFP Ventilation system has been secured. (NEL 99-0078) 4.4.17 Start CTMT PURGE SUPP/EXH FANS & DAMPERS in LOW.

4.4.18 IF available, THEN Immediately start the second Auxiliary Building Main Exhaust Fan, Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 4.4.19 Perform the following to checkout the operating system:

x Walkdown fans and ductwork.

x Visually inspect for damage.

x Report any problems to the control room.

4.4.20 IF increased flow is required AND both Auxiliary Building Main Exhaust Fans are operating, THEN shift CTMT PURGE SUPP/EXH FANS &

DAMPERS to HI.

4.4.21 Notify the Shift Radiochemist that CTMT main-purge is in service.

4.5 REMOVING CONTAINMENT PURGE SYSTEM FROM OPERATION CAUTION: Both Auxiliary Building Main Exhaust Fans must be and should only be placed in service when CTMT main purge is placed in service to prevent over pressurizing the exhaust plenum. (NEL 99-0078) 4.5.1 Notify the Shift Radiochemist prior to securing CTMT purge.

4.5.2 IF the unit is in Modes 5, 6 or Defueled THEN notify HP prior to securing CTMT purge.

4.5.3 Reduce the number of Aux Bldg Mn Exh Fans in service to one.

4.5.4 Place CTMT PURGE SUPP/EXH FANS & DAMPERS handswitch to STOP.

4.5.5 Place CTMT PURGE DMPRS handswitch HS-3196 to CLOSE.

4.5.6 Place CTMT PURGE DMPRS handswitch HS-3198 to CLOSE.

4.5.7 Check the following containment purge dampers closed:

x HV-3198A & D x HV-3198B & C Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 NOTE: If CTMT Purge system is to be removed from operation for a short duration for maintenance or testing, steps 4.5.8 and 4.5.9 are NOT required.

4.5.8 IF desired AND approved by the Shift Supervisor, THEN open the following CTMT purge filter cooling valves:

x N2P13V293 x N2P13V294 NOTE: R-24A or R-24B is required to be operable in Modes 1-4, AND both channels are required to be operable during core alterations and moving of irradiated fuel in containment.

4.5.9 IF desired, THEN turn OFF radiation monitors R-24A and R-24B and their respective recorders.

4.6 RESTORING CONTAINMENT PURGE FOLLOWING A SAFETY INJECTION OR HI HI RADIATION SIGNAL:

4.6.1 Verify Safety Injection or Hi Hi Radiation signal has been reset.

4.6.2 Place the following CTMT PURGE DMPRS handswitches to CLOSE for at least 2 seconds:

x HS-3196 x HS-3198 4.6.3 Place the following CTMT PURGE DMPRS handswitches to FULL:

x HS-3196 x HS-3198 4.6.4 Restart CTMT purge supply and exhaust fans per steps 4.4.17 through 4.4.21.

Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 4.7 PLACING CONTAINMENT MINI-PURGE SYSTEM IN OPERATION 4.7.1 IF CTMT mini-purge system was removed from operation for a short duration for maintenance or testing, THEN go to step 4.7.12 to place the CTMT mini-purge system back in service.

4.7.2 Verify auxiliary building main exhaust system in service per FNP-2-SOP-58.0, AUXILIARY BUILDING HVAC SYSTEM.

4.7.3 Verify that plant vent stack radiation monitors R-14 and R-22 are aligned for normal operation per FNP-2-SOP-45.0, RADIATION MONITORING SYSTEM, OR alternate sampling being performed as required per ODCM.

4.7.4 Verify radiation monitor R-24A OR R-24B in service per FNP-2-SOP-45.0, RADIATION MONITORING SYSTEM.

4.7.5 Verify CTMT PURGE SUPP/EXH FANS & DAMPERS handswitch selected to STOP.

4.7.6 Verify recorders RR0100 (points RR14 and RR22), RR24A, and RR24B in operation, OR CR written for repair.

4.7.7 Verify gaseous waste release permit has been issued for CTMT mini-purge operation.

Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 4.7.8 Verify that the alarm setpoints for R-24A and R-24B are less than or equal to the maximum allowable setpoints listed in Part II of the gaseous waste release permit. (See FNP-2-CCP-213.1, GASEOUS EFFLUENT RADIATION MONITORING SYSTEM SETPOINTS, current alarm setpoints.)

4.7.9 IF the activity level within containment is determined to be excessive by sample analysis, THEN operate the containment pre-access filtration system per section 4.1 to reduce the activity level as specified by the Health Physics department.

4.7.10 Verify the following are current prior to initiating mini-purge flow:

x FNP-2-STP-18.5, CONTAINMENT MINI PURGE AND EXHAUST VALVE INSERVICE TEST x FNP-2-STP-18.6, CONTAINMENT MINI PURGE AND EXHAUST ISOLATION TEST MODES 1, 2, 3, 4 4.7.11 IF opened while removing CTMT mini-purge system from operation, THEN close the following containment purge filter cooling valves:

x N2P13V293 x N2P13V294 4.7.12 Perform the following for containment purge:

4.7.12.1 Place the following CTMT PURGE DMPRS handswitches to MINI:

x HS-3196 x HS-3198 4.7.12.2 Verify the following dampers are OPEN:

x Q2P13HV2866C x Q2P13HV2866D x Q2P13HV2867C x Q2P13HV2867D x N2P13HV3198B x N2P13HV3198C Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 4.7.13 Perform the following for containment mini-purge:

4.7.13.1 Start MINI PURGE SUPP/EXH FAN.

4.7.13.2 Verify the following dampers are OPEN:

x N2P13HV3203A x N2P13HV3203B 4.7.14 Verify the following dampers are CLOSED:

x Q2P13HV3196 x Q2P13HV3197 x Q2P13HV3198A x Q2P13HV3198D 4.7.15 Perform the following to checkout the operating system:

x Walkdown fans and ductwork.

x Visually inspect for damage.

x Report any problems to the control room.

4.7.16 Notify the Shift Radiochemist that CTMT mini-purge is in service.

4.8 REMOVING CONTAINMENT MINI-PURGE SYSTEM FROM OPERATION 4.8.1 Stop MINI PURGE SUPP/EXH FAN.

4.8.2 Place the following CTMT PURGE DMPRS handswitches to CLOSE:

x HS-3196 x HS-3198 4.8.3 Notify the Shift Radiochemist that CTMT mini-purge has been secured.

NOTE: If CTMT Mini Purge is to be removed from operation for a short duration for maintenance or testing, step 4.8.4 is NOT required.

4.8.4 IF desired AND approved by the Shift Supervisor, THEN open the following containment purge filter cooling valves:

x N2P13V293 x N2P13V294 Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 NOTE: R-24A or R-24B is required to be operable in Modes 1-4, AND both channels are required to be operable during core alterations and moving of irradiated fuel in containment.

4.8.5 IF desired, THEN turn off radiation monitors R-24A, R-24B, and their respective recorders.

4.9 RESTORING CONTAINMENT MINI-PURGE FOLLOWING A SAFETY INJECTION OR HI HI RADIATION SIGNAL:

4.9.1 Verify Safety Injection or Hi Hi Radiation Signal has been reset.

4.9.2 Place the following CTMT PURGE DMPRS handswitches to CLOSE for at least 2 seconds:

x HS-3196 x HS-3198 4.9.3 Place the following CTMT PURGE DMPRS handswitches to MINI:

x HS-3196 x HS-3198 4.9.4 Start MINI PURGE SUPP/EXH FAN.

4.9.5 Perform the following to checkout the operating system:

Walkdown fans and ductwork.

Visually inspect for damage.

Report any problems to the control room.

Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 4.9.6 Notify the Shift Radiochemist that CTMT mini-purge is in service.

4.9.7 Verify the following damper positions:

x Q2P13HV3203A, open x Q2P13HV3203B, open x Q2P13HV3196, closed x Q2P13HV3197, closed x Q2P13HV3198A, closed x N2P13HV3198B, open x N2P13HV3198C, open x Q2P13HV3198D, closed x Q2P13HV2866C, open x Q2P13HV2866D, open x Q2P13HV2867C, open x Q2P13HV2867D, open

5.0 REFERENCES

5.1 P & IDs:

5.1.1 D-205010, sheets 1 and 2, Containment Cooling and Purge System P & ID 5.1.2 D-207783, Elem. Diag., Containment Mini-Purge Fans 5.1.3 D-207204, Elem. Diag., Containment Purge Iso. Dampers Train A 5.1.4 D-207199, Elem. Diag., Containment Purge Iso. Dampers Train B 5.1.5 D-207236, Elem. Diag., Containment Purge Air Handling Unit Fan 5.1.6 D-207237, Elem. Diag., Containment Purge Exhaust Fan 5.1.7 D-204654, Conn. Diag., Containment Purge Starter Panels 5.2 FSAR Section 6.2.3 Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 APPENDIX 1 CONTINGENCY OPERATION OF CONTAINMENT MINI-PURGE WITH SUPPLY FAN UNAVAILABLE Performed by: Date Reviewed by: Date This appendix consists of 5 pages Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 APPENDIX 1 CONTINGENCY OPERATION OF CONTAINMENT MINI-PURGE WITH SUPPLY FAN UNAVAILABLE 1.0 Verify the following: (OR 1-98-498) 1.1 The version of this procedure has been verified to be the current version.

(OR 1-98-498) 1.2 This procedure has been verified to be the correct unit for the task.

(OR 1-98-498) 2.0 Prepare MINI PURGE EXH FAN for contingency operation.

2.1 Open BKR FD-G5.

NOTE: FNP-0-EMP-1906.1, INSTALLATION AND REMOVAL OF TEMPORARY ELECTRICAL ALTERATIONS, must be used to provide detailed guidance for installing the following jumper.

2.2 Record the following information for test instrument(s) used in accordance with FNP-0-EMP-1906.1:

Instrument Serial Number:

Calibration Due Date:

/ 2.3 Jumper the 42 interlock contact (C2 to X1) for BKR FD-G5.

EM/EM CV 2.4 Place a caution tag on MINI PURGE SUPP/EXH FAN MCB handswitch to indicate that the MINI PURGE EXH FAN is being controlled locally at BKR FD-G5.

Page 1 of 5 Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 3.0 Operate MINI PURGE EXH FAN to control CTMT to atmosphere 'P.

3.1 Monitor Narrow Range CTMT Pressure Avg. as indicated by computer point PC1501 and log on Data Sheet 1 on an hourly basis.

CAUTION: CTMT to atmosphere 'P must be maintained -0.5 - +0.2 psid.

3.2 IF at any time during operation of the MINI PURGE EXH FAN, CTMT to atmosphere 'P exceeds +0.2 psid, THEN notify the Shift Supervisor.

3.3 WHEN CTMT to atmosphere 'P approaches +0.2 psid, THEN start the MINI PURGE EXH FAN.

3.3.1 Verify auxiliary building main exhaust system is in service per FNP-2-SOP-58.0, AUXILIARY BUILDING HVAC SYSTEM.

3.3.2 Verify that plant vent stack radiation monitors R-14 and R-22 are aligned for normal operation per FNP-2-SOP-45.0, RADIATION MONITORING SYSTEM.

3.3.3 Verify radiation monitor R-24A OR R-24B in service per FNP-2-SOP-45.0, RADIATION MONITORING SYSTEM.

3.3.4 Verify CTMT PURGE SUPP/EXH FANS & DAMPERS handswitch selected to STOP.

3.3.5 Verify Recorders RR0100 (points RR14 and RR22), RR24A, and RR24B in operation or CR written for repair.

3.3.6 Verify gaseous waste release permit has been issued for CTMT mini-purge operation.

3.3.7 Verify that the alarm setpoints for R-24A and R-24B are less than or equal to the maximum allowable setpoints listed in Part II of the gaseous waste release permit. (See FNP-2-CCP-213.1, GASEOUS EFFLUENT RADIATION MONITORING SETPOINTS, for current alarm setpoints.)

Page 2 of 5 Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 3.3.8 IF the activity level within containment is determined to be excessive by sample analysis, THEN operate the containment pre-access filtration system per procedure section 4.1 as necessary to reduce the activity level as specified by Health Physics.

3.3.9 Verify the following are current prior to initiating mini-purge flow:

x FNP-2-STP-18.5, CONTAINMENT MINI PURGE AND EXHAUST VALVE INSERVICE TEST x FNP-2-STP-18.6, CONTAINMENT MINI PURGE AND EXHAUST ISOLATION TEST MODES 1, 2, 3, 4 3.3.10 IF opened while removing CTMT mini-purge from operation, THEN close the following CTMT purge filter cooling valves:

x N2P13V293 x N2P13V294 3.3.11 Position the following CTMT PURGE DMPRS handswitches to MINI:

x HS-3196 x HS-3198 3.3.12 Start MINI PURGE EXH FAN by locally closing BKR FD-G5.

3.3.13 Notify the Shift Radiochemist that CTMT mini-purge is in service.

Page 3 of 5 Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 3.4 WHEN CTMT to atmosphere 'P approaches -0.5 psid, THEN stop the MINI PURGE EXH FAN.

3.4.1 Open BKR FD-G5.

3.4.2 Place the following CTMT PURGE DMPRS handswitches to CLOSE:

x HS-3196 x HS-3198 3.4.3 Notify the Shift Radiochemist that CTMT mini-purge has been secured.

NOTE: If CTMT Mini Purge is to be removed from operation for a short duration for maintenance or testing, step 3.4.4 is NOT required.

3.4.4 IF desired AND approved by the Shift Supervisor, THEN open the following CTMT purge filter cooling valves:

x N2P13V293 x N2P13V294 NOTE: R-24A or R-24B is required to be operable in Modes 1-4, AND both channels are required to be operable during core alterations and moving of irradiated fuel in containment.

3.4.5 IF desired, THEN turn off radiation monitors R-24A and R-24B, AND their respective recorders.

4.0 WHEN contingency operation NOT required, THEN restore MINI PURGE EXH FAN to normal operation.

4.1 Verify open BKR FD-G5.

NOTE: FNP-0-EMP-1906.1, INSTALLATION AND REMOVAL OF TEMPORARY ELECTRICAL ALTERATIONS, must be used to provide detailed guidance for removing the jumper in the following step.

/ 4.2 Remove jumper installed in step 2.3 from the 42 interlock contact (C2 to X1) for CV BKR FD-G5.

EM 4.3 Remove caution tag placed in step 2.4.

Page 4 of 5 Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 DATA SHEET 1 DATE TIME 'P 'P 'P 'P 'P 'P 'P 'P 'P 0000 0100 0200 0300 0400 0500 0600 0700 0800 0900 1000 1100 1200 1300 1400 1500 1600 1700 1800 1900 2000 2100 2200 2300 Page 5 of 5 Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 APPENDIX 2 CONTINGENCY OPERATION OF CONTAINMENT MAIN PURGE WITH SUPPLY FAN UNAVAILABLE WITH CONTAINMENT INTEGRITY NOT REQUIRED Performed by: Date Reviewed by: Date This appendix consists of 2 pages Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 APPENDIX 2 CONTINGENCY OPERATION OF CONTAINMENT MAIN PURGE WITH SUPPLY FAN UNAVAILABLE WITH CONTAINMENT INTEGRITY NOT REQUIRED 1.0 Procedure Verifications 1.1 The version of this procedure has been verified to be the current version.

(OR 1-98-498) 1.2 This procedure has been verified to be the correct unit for the task.

(OR 1-98-498) 2.0 Defeat Purge Supply Fan Starting Capability as follows:

2.1 Verify containment integrity NOT required.

2.2 Verify equipment hatch is open.

NOTES: The Purge Supply Fan control circuit must be energized for the Purge Exhaust Fan damper to open and provide a flow path for the exhaust fan. The intent of lifting the motor leads at the starter panel is to provide personnel safety for any one working on the Purge Supply Fan while allowing the Purge Supply Fan control circuit to remain energized. The Purge Supply damper will also open with the Purge Exhaust damper when Purge is placed in service.

The cable scheme number for the Hi speed motor leads is 2VXEA06Q and for the Low speed motor leads is 2VXEA06R as shown on D-204654.

2.3 Submit a CR to have Electrical Maintenance lift, tape and tag the Purge Supply Fan Hi and Low speed motor leads at the CTMT PURGE SUPPLY FAN STARTER PANEL N2P13L002-N.

CR number Page 1 of 2 Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 2.4 WHEN ready to implement the work order, THEN perform the following per the appropriate section(s) in the body of this procedure:

2.4.1 Secure Main Purge.

2.4.2 IF desired, THEN start Mini Purge system.

2.5 Direct EM to lift, tape and tag the Purge Supply Fan motor leads per the work order.

2.6 Place a caution tag on CTMT PURGE SUPP/EXH FANS & DAMPERS MCB handswitch to indicate that annunciator BF3 may alarm during purge operation due to the supply fan motor leads being lifted.

2.7 WHEN the Purge Supply Fan motor leads have been lifted AND the associated tagout completed for this stage, THEN perform the following per the appropriate sections in the body of this procedure:

2.7.1 Verify Mini Purge is secured.

2.7.2 Place Main Purge in service.

3.0 Restoration of Purge Supply Fan Following Maintenance 3.1 WHEN ready to restore purge supply fan starting capability, THEN perform the following per the appropriate section(s) in the body of this procedure:

3.1.1 Secure Main Purge.

3.1.2 IF desired, THEN start Mini Purge system.

3.2 Direct EM to untag, untape, and reland the Purge Supply Fan motor leads per the work order.

3.3 Clear the caution tag on CTMT PURGE SUPP/EXH FANS & DAMPERS handswitch that indicates that annunciator BF3 may alarm during purge operation due to the supply fan leads being lifted.

3.4 WHEN the Purge Supply Fan motor leads have been relanded AND the associated tagout cleared, THEN perform the following per the appropriate section in the body of this procedure:

3.4.1 Verify Mini Purge is secured.

3.4.2 Place Main Purge in service.

Page 2 of 2 Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 APPENDIX 3 BATCH RELEASES OF CONTAINMENT ATMOSPHERE Verified By: Date Performed by: Date Reviewed by: Date This appendix consists of 5 pages Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 APPENDIX 3 BATCH RELEASES OF CONTAINMENT ATMOSPHERE 1.0 Procedure Verifications 1.1 The version of this procedure has been verified to be the current version.

(OR 1-98-498) 1.2 This procedure has been verified to be the correct unit for the task.

(OR 1-98-498) 2.0 Initial Conditions NOTES: Provide Shift Radio Chemist with request for Batch Release permit at least 12-24 hours in advance of determined time batch release will take place. (Batch Release should take place before CTMT to ATMOS 'P reaches 0.4 psid.)

Use computer point PC1501 NARROW RANGE CTMT AVG PRESS (preferred) OR the average of the following computer points for monitoring CTMT pressure.

3/4 PT0951 3/4 PT0952 3/4 PT0953

IF purge is to started for other than routine batch releases or is requested to run longer than expected to meet termination criteria, THEN document the reason (CTMT entry, maintenance, chemistry or other reason) in CONTROL ROOM LOG and on the release permit.

(AI 2009202457) 2.1 Once CTMT to ATMOS 'P is approximately 0.25 psid, then notify the Shift Radio Chemist to prepare a Batch Gaseous Waste Release Permit for the Unit 2 CTMT Atmosphere.

2.2 Verify auxiliary building main exhaust system in service per FNP-2-SOP-58.0, AUXILIARY BUILDING HVAC SYSTEM.

/ 2.3 Verify gaseous waste release permit has been issued for CTMT mini-purge IV operation.

Page 1 of 5 Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 2.4 Verify either 2.4.1, 2.4.2 OR 2.4.3 is satisfied:

2.4.1 R-14 is in service and aligned for normal operation per FNP-2-SOP-45.0, RADIATION MONITORING SYSTEM.

AND R-29B is in service and aligned per FNP-0-CCP-641, OPERATIONS OF THE PLANT VENT STACK MONITORING SYSTEM.

OR-2.4.2 R-22 is in service and aligned for normal operation per FNP-2-SOP-45.0, RADIATION MONITORING SYSTEM.

AND R-29B is in service and aligned per FNP-0-CCP-641, OPERATIONS OF THE PLANT VENT STACK MONITORING SYSTEM.

OR-2.4.3 IF R-14 and R-22 OR R-29B is inoperable, THEN Alternate sampling is being performed as required per ODCM.

Plausibility for Alternate Sampling Page 2 of 5 Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 NOTE: Any containment purge isolation valve made inoperable due to a failed rad monitor shall be close and deactivated in accordance with the time frame indicated in TS 3.6.3.

2.5 Verify Containment Purge Rad Monitor operation:

2.5.1 Verify one of the following radiation monitors in service per FNP-2-SOP-45.0, RADIATION MONITORING SYSTEM.

R-24A OR

R-24B 2.5.2 Verify alarm setpoint for the above selected radiation monitor is less than or equal to the maximum allowable setpoint listed in Part II of the gaseous waste release permit. (See FNP-2-CCP-213.1, GASEOUS EFFLUENT RADIATION MONITORING SYSTEM SETPOINTS, current alarm setpoints.)

/ R-24A IV

/ R-24B IV 2.6 Verify items listed below are in operation for operable radiation monitors:

Recorder RR0100 (point RR14 and/or RR22)

Recorder RR24A OR RR24B

Printer R29B Page 3 of 5 Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 3.0 Performing Batch Release of CTMT Atmosphere NOTE: Batch releases should be terminated as soon as the criterion for termination is satisfied. (Historical data indicates on average, the criterion will be satisfied in approximately ninety minutes.) AI 2009202457 3.1 Record initial pressure as indicated by Computer Point PC1501 NARROW RANGE CTMT AVG PRESS, on Part III of the Batch Gaseous Waste Release Permit.

3.2 Open N2P13V294, PURGE FILTER COOLING OUTLET VALVE.

3.3 Record start date/time data in Part III of the Batch Gaseous Waste Release Permit.

3.4 WHEN performing the following valve manipulations, THEN note the start time for recording purposes:

3.4.1 Place the following CTMT Purge DMPRS hand switches to MINI to initiate CTMT Batch Release:

HS-3196

HS-3198 3.5 WHEN CTMT DIFF PRESSURE decreases to THEN perform the following, noting the fan start time and containment-to-atmosphere delta pressure for recording purposes:

/ 3.5.1 Close N2P13V294, PURGE FILTER COOLING OUTLET IV VALVE.

3.5.2 Start MINI PURGE SUPP/EXH FAN.

3.5.3 Record fan start date/time data in Part III of the Batch Gaseous Waste Release Permit.

3.5.4 Record CTMT TO ATMOS 'P (Computer Point PC1501 NARROW RANGE CTMT AVG PRESS) at which the mini purge fans were started in Part III of the Batch Gaseous Waste Release Permit.

NOTE: During highly active or busy periods consider using a timer to ensure the release is terminated when anticipated. AI 2009202457 3.6 After CTMT DIFF PRESSURE has leveled off, OR when at desired value, THEN secure the CTMT Mini Purge System per section 4.0 of this Appendix.

Page 4 of 5 Version 38.0

08/15/12 08:03:04 UNIT 2 FNP-2-SOP-12.2 4.0 Securing CTMT Mini Purge System after CTMT Batch Release Complete 4.1 Stop the MINI PURGE SUPP/EXH FAN.

4.2 Place the following CTMT Purge DMPRS hand switches to CLOSE, noting the time for recording purposes:

/ HS-3196 IV

/ HS-3198 IV 4.3 Record the following data in Part III of the Batch Gaseous Waste Release Permit:

Final pressure as indicated by Computer Point PC1501 NARROW RANGE CTMT AVG PRESS

Release stop date/time data 4.4 Notify the Shift Radio Chemist that CTMT Mini-purge has been secured.

4.5 Ensure the Batch Gaseous Waste Release Permit is reviewed by a member of OPS supervision.

Page 5 of 5 Version 38.0

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

82. 029EA2.04 082 The following conditions exist on Unit 1:

At 1000:

A Reactor Trip was initiated, but was unsuccessful.

The crew entered FRP-S.1, Response to Nuclear Power Generation/ATWT.

The Rover was dispatched to open Reactor Trip breakers.

At 1005:

Per FRP-S.1, an Emergency Boration is in progress from the 1A BAT (Boric Acid Tank).

FK-122, CHG FLOW, is in AUTO.

FI-122A, CHG FLOW, indicates 35 gpm.

All Reactor Trip and Reactor Trip bypass breakers are open.

The Shift Manager is evaluating emergency classifications.

Which one of the following completes the statements below?

At 1005, Charging flow (1) adequate for the Emergency Boration.

The Shift Manager is required to declare a(n) (2) emergency classification.

Reference Provided (1) (2)

A. IS Alert B. is NOT Alert C. IS Site Area D. is NOT Site Area Tuesday, March 19, 2013 11:38:42 AM 24

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 A. Incorrect 1) Incorrect, per FRP-S.1, minimum Emergency Boration flow from the BATs is 30 gpm with minimum Charging flow of 40 gpm.

Plausible: A candidate determining that Charging flow is adequate is plausible because minimum BAT flow is 30 gpm.

2) Incorrect, per NMP-EP-110 Classification Matrix, a Site Area Emergency is required.

Plausible if the candidate determines that the actions to dispatch an operator to manually open the RX trip breakers is considered manually tripping the Reactor.

There is also plausibility that a candidate may choose the Alert classification because the Reactor is shutdown prior to the Shift Manager declaring the initial classification. The Shift Manager has to declare the emergency classification within 15 minutes. The candidate may determine that since Reactor trip breakers are open prior to the end of the 15 minute window, the Manual Trip was successful by opening the trip breakers.

This is not a direct look-up since the key to the correct classification is how "Manual Trip" is interpreted. The interpretation of Manual Trip is explained in places other than the reference that is provided to the candidate.

B. Incorrect 1) Correct, minimum Charging flow required during Emergency Boration from the BATs is 40 gpm.

2) Incorrect, see A.2.

C. Incorrect 1) Incorrect, see A.1.

2) Correct, having to drive control rods manually and locally tripping the reactor during FRP-S.1 will be classified as a Site Area Emergency.

D. Correct 1) Correct, see B.1.

2) Correct, see C.2.

K/A: 029EA2.04 Anticipated Transient Without Scram (ATWS) - Ability to determine or interpret the following as they apply to a ATWS:

EA2.01 CVCS centrifugal charging pump operating indication . . . . . . . . . . . . . . . . . . 3.2* 3.3*

Importance Rating: 3.2* 3.3*

Technical

Reference:

FNP-1-FRP-S.1, Response to Nuclear Power Generation/ATWT, V 27 NMP-EP-110-GL01 FNP EALs-IC's, Threshold Values and Tuesday, March 19, 2013 11:38:42 AM 25

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 Basis, Ver 2 References provided: Applicable portion of NMP-EP-110-GL01 that only allows evaluation of Site Area and Alert classifications Learning Objective: Using plant procedures/references, ANALYZE a set of plant conditions and DETERMINE the proper classification of the emergency condition as being a NOUE, Alert, Site Area, or General Emergency. (OPS-63002C01).

Question History: New question K/A match: During an ATWT event, a candidate has to evaluate indications for Charging pump flow and determine if is acceptable during an Emergency Boration per FRP-S.1.

SRO justification: III. Justification for Plant Specific Exemptions UNIQUE to the SRO position:

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:

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.d]

AND/OR

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.

The SRO is solely responsible for determining Classifications at FNP, the objective listed above is an SRO only objectives.

2013 NRC exam III. Justification for Plant Specific Exemptions UNIQUE to the SRO position:

DOES NOT MATCH one of the 10 CFR 55.43(b) items but for the Emergency Plan implementation, FNP has classified the knowledge/ability as unique to the SRO position as documented within SAT process as ties the knowledge/ability to the licensees SRO job position duties.

Tuesday, March 19, 2013 11:38:42 AM 26

Reference provided to candidate for this question - from NMP-EP-110 classification Matrix

3/15/2013 13:29 FNP-1-FRP-S.1 UNIT 1 RESPONSE TO NUCLEAR POWER GENERATION/ATWT Revision 27.0 Step Action/Expected Response Response NOT Obtained 3 Verify AFW pumps - RUNNING.

3.1 MDAFWPs - RUNNING

[] 1A amps > 0

[] 1B amps > 0 3.2 TDAFWP - RUNNING IF NECESSARY TDAFWP STM SUPP FROM 1B(1C) SG

[] MLB-4 1-3 lit

[] MLB-4 2-3 lit

[] MLB-4 3-3 lit TDAFWP SPEED

[] SI 3411A > 3900 rpm TDAFWP SPEED CONT

[] SIC 3405 at 100%

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ NOTE: 2500 gallons of emergency boration is required for each control rod not fully inserted, up to a maximum of 17,309 gallons.

[CA] Emergency boration should continue until an adequate shutdown margin is established.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ 4 Initiate Emergency Boration of the RCS.

4.1 IF safety injection in progress, THEN go to step 4.7.

4.2 Verify at least one CHG PUMP -

RUNNING.

Step 4 continued on next page.

Page 4 of 18

3/15/2013 13:29 FNP-1-FRP-S.1 UNIT 1 RESPONSE TO NUCLEAR POWER GENERATION/ATWT Revision 27.0 Step Action/Expected Response Response NOT Obtained 4.3 Start a boric acid transfer 4.3 Perform the following.

pump.

4.3.1 Align charging pump suction BATP to RWST.

[] 1A

[] 1B RWST TO CHG PUMP

[] Q1E21LCV115B open

[] Q1E21LCV115D open VCT OUTLET ISO

[] Q1E21LCV115C closed

[] Q1E21LCV115E closed 4.3.2 Proceed to step 4.5.

4.4 Align normal emergency 4.4 Perform the following.

boration.

Align charging pump suction EMERG BORATE to RWST.

TO CHG PUMP SUCT

[] Q1E21MOV8104 open RWST TO CHG PUMP

[] Q1E21LCV115B open

[] Q1E21LCV115D open VCT OUTLET ISO

[] Q1E21LCV115C closed

[] Q1E21LCV115E closed OR Align manual emergency boration flow path.

BORIC ACID TO BLENDER

[] Q1E21FCV113A open MAN EMERG BORATION

[] Q1E21V185 open (100 ft, AUX BLDG rad-side chemical mixing tank area)

Step 4 continued on next page.

Page 5 of 18

3/15/2013 13:29 FNP-1-FRP-S.1 UNIT 1 RESPONSE TO NUCLEAR POWER GENERATION/ATWT Revision 27.0 Step Action/Expected Response Response NOT Obtained 4.5 Establish adequate letdown.

4.5.1 Verify 45 GPM letdown orifice - IN SERVICE.

LTDN ORIF ISO 45 GPM

[] Q1E21HV8149A open 4.5.2 Verify one 60 GPM letdown orifice - IN SERVICE.

LTDN ORIF ISO 60 GPM

[] Q1E21HV8149B open

[] Q1E21HV8149C open 4.6 Establish adequate charging flow.

IF boration is from boric acid storage tank, THEN verify charging flow - Actual Charging flow is 35 GREATER THAN 40 GPM. gpm, <40 gpm required OR IF boration is from the RWST, THEN verify charging flow -

GREATER THAN 92 GPM.

Step 4 continued on next page.

Page 6 of 18

3/15/2013 13:29 FNP-1-FRP-S.1 UNIT 1 RESPONSE TO NUCLEAR POWER GENERATION/ATWT Revision 27.0 Step Action/Expected Response Response NOT Obtained 4.7 Verify emergency boration flow adequate.

4.7.1 IF normal emergency boration flow path aligned, This is plausibility THEN check emergency for Charging flow boration flow greater than being adequate.

30 GPM.

This is BORIC ACID BORATION flow, EMERG BORATE not Charging flow

[] FI 110 4.7.2 IF manual emergency boration flow path aligned, THEN check boric acid flow greater than 30 GPM.

MAKEUP FLOW TO CHG/VCT

[] BA FI 113 4.7.3 IF boration is from the RWST, THEN verify charging OR HHSI flow - GREATER THAN 92 GPM.

4.8 Check pressurizer pressure 4.8 Verify PRZR PORVs and PRZR LESS THAN 2335 psig. PORV ISOs - OPEN. IF NOT NOT, THEN open PRZR PORVs and PORV ISOs as necessary until pressurizer pressure less than 2135 psig.

Page 7 of 18

NMP-EP-110-GL01 - FNP EALs - ICs, Threshold Values and Basis Version 3.0 This is the Site Area classification background SS2 Initiating Condition (Back to Hot IC p. 105)

Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Trip Once a Reactor Protection System Setpoint Has Been Exceeded AND Manual Trip Was NOT Successful.

Operating Mode Applicability: Power Operation (Mode 1)

Startup (Mode 2)

Threshold Value: (1)

NOTE: A successful manual trip for purposes of declaration is any action taken from the MCB that rapidly inserts the control rods. This can be accomplished by tripping the reactor using the Reactor Trip switches on the MCB OR by de-energizing both Rod Drive Motor Generator sets from the MCB.

NOTE Failure of both MCB Rx Trip switches to trip the reactor meets the TV criteria of a setpoint being exceeded with no automatic trip occurring.

1. Indications exist that a reactor protection system setpoint was exceeded and automatic trip did not occur, and a manual trip did not result in the reactor being made subcritical (NOTE).

Basis:

Automatic and manual trip are not considered successful if action away from the reactor control console was required to trip the reactor.

The Reactor should be considered subcritical when reactor power level has been reduced to less than 5% power and SUR is negative.

Under these continued power generation conditions, the reactor may be producing more heat than the maximum decay heat load for which the safety systems are designed. A Site Area Emergency is indicated because conditions exist that may lead to imminent loss or potential loss of both fuel clad and RCS. Although this IC may be viewed as redundant to the Fission Product Barrier Degradation IC, its inclusion is necessary to better assure timely recognition and emergency response. Escalation of this event to a General Emergency would be via Fission Product Barrier Degradation or Emergency Director Judgment ICs. A manual reactor trip is considered to be a trip input to the automatic Reactor Protection System.

45

NMP-EP-110-GL01 - FNP EALs - ICs, Threshold Values and Basis Version 3.0 This is the Alert Classification background SA2 Initiating Condition (Back to Hot IC p. 105)

Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Trip Once a Reactor Protection System Setpoint Has Been Exceeded AND Manual Trip Was Successful.

Operating Mode Applicability: Power Operation (Mode 1)

Startup (Mode 2)

Hot Standby (Mode 3)

Threshold Value: (1)

NOTE: A successful manual trip for purposes of declaration is any action taken from the MCB that rapidly inserts the control rods. This can be accomplished by tripping the reactor using the Reactor Trip switches on the MCB OR by de-energizing both Rod Drive Motor Generator sets from the MCB.

NOTE Failure of both MCB Rx Trip switches to trip the reactor meets the TV criteria of a setpoint being exceeded with no automatic trip occurring.

1. Indication(s) exist that a reactor protection setpoint was exceeded and an automatic trip did not occur, and a manual trip resulted in the reactor being subcritical.

Basis:

This condition indicates failure of the automatic protection system to trip the reactor. This condition is more than a potential degradation of a safety system in that a front line automatic protection system did not function in response to a plant transient and thus the plant safety has been compromised, and design limits of the fuel may have been exceeded. An Alert is indicated because conditions exist that lead to potential loss of fuel clad or RCS. Reactor protection system setpoint being exceeded, rather than limiting safety system setpoint being exceeded, is specified here because failure of the automatic protection system is the issue. A manual reactor trip is considered to be a trip input to the automatic Reactor Protection System or de-energizing the MG sets should initiate a manual trip.

The Reactor should be considered subcritical when reactor power level has been reduced to less than 5% power and SUR is negative.

50

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

83. 051AG2.1.7 083 Unit 1 is at 28% power with the following conditions:

At 1000:

Condenser pressure is 1.7 psia and slowly rising (vacuum degrading).

KK1, TURB COND VAC LO, is in alarm.

HP Gland steam supply pressure is 3.5 psig.

Main Turbine LP Gland pressures are as follows:

- #3 LP Gland pressure is 2.0 psig.

- #4 LP Gland pressure is 4.8 psig.

- #5 LP Gland pressure is 2.5 psig.

- #6 LP Gland pressure is 0.5 psig.

At 1015:

Condenser pressure is 2.1 psia and still slowly rising (vacuum degrading).

KK2, TURB COND VAC LO-LO, is in alarm.

Which one of the following completes the statements below per AOP-8.0?

The action required at 1000 to stabilize Condenser pressure is to (1) .

The action required at 1015 is to (2) .

Procedure titles are as follows:

AOP-3.0, Turbine Trip Below P-9 Setpoint AOP-8.0, Partial Loss of Condenser Vacuum A. 1) bypass the #6 Gland Seal Regulator

2) perform AOP-3.0 in parallel with AOP-8.0 B. 1) bypass the #6 Gland Seal Regulator

2) perform AOP-3.0 ONLY C. 1) throttle closed V528, SPILLOVER VALVE BYPASS

2) perform AOP-3.0 in parallel with AOP-8.0 D. 1) throttle closed V528, SPILLOVER VALVE BYPASS

2) perform AOP-3.0 ONLY Tuesday, March 19, 2013 11:38:42 AM 28

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 A. Correct 1) Correct, per AOP-8.0, step 4.2 - if individual gland seal pressures are not in the band of 1-5 psig (#6 is .5 psig), the RNO action is to bypass the affected gland seal regulator.

2) Correct, per AOP-8.0, step 2. This is a Continuing Action throughout the procedure. Step 2.1.2.1 of the RNO actions requires performing AOP-3.0 in parallel with AOP-8.0.

B. Incorrect 1) Correct, see A.1.

2) Incorrect, performing AOP-3.0 is required when the Turbine is tripped

<35% power. Plausible since the candidate may think AOP-3.0 is the highest priority, and since the Turbine is tripped, actions of AOP-8.0 are no longer required.

C. Incorrect 1) Incorrect, plausible if the candidate does not fully understand the function of the Spillover Valve and its Bypass valve. There is a common misconception that this valve controls the Gland Sealing Steam pressures on both the LP Turbine and the HP Turbine. In actuality, it only controls the HP Turbine Gland pressure.

2) Correct, see A.2.

D. Incorrect 1) Incorrect, see C.1.

2) Incorrect, see B.2.

K/A: 051AG2.1.7 Loss of Condenser Vacuum Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

(CFR: 41.5 / 43.5 / 45.12 / 45.13)

IMPORTANCE RO 4.4 SRO 4.7 Importance Rating: 4.4 4.7 Technical

Reference:

FNP-1-AOP-8.0, v22.1 References provided: None Learning Objective: EVALUATE plant conditions and DETERMINE if transition to another section of AOP-8.0, Partial Loss of Condenser Vacuum or to another procedure is required.

(OPS-62520H02)

Question History: New question Tuesday, March 19, 2013 11:38:42 AM 29

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 Basis for meeting K/A: Plant instrumentation conditions are given for a malfunction causing a Loss of Condenser Vacuum event. The candidate has to demonstrate the ability to determine the appropriate actions to perform per AOP-8.0, and ability to determine proper procedure implementation while in AOP-8.0.

SRO justification: 10 CFR 55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations [10 CFR 55.43(b)(5)], involving BOTH:

1) assessing plant conditions and then

2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

The candidate has to have knowledge of the transition criteria in AOP-8.0 (to go to AOP-3.0 upon a Turbine Trip condition), but in addition knowledge of how AOP-3.0 will be implemented (in parallel with AOP-8.0, or alone) is required.

This also falls under the 10 CFR 55.43(b)(5) condition of:

Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.

2013 NRC exam From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 flowchart for 10 CFR 55.43(b)(5) :

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations [10 CFR 55.43(b)(5)], involving BOTH:

1) assessing plant conditions and then

2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

Using the flowchart, this question can:

NOT be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location. (PT 1 IS sys/fund knowledge, but PT 2 requires procedural knowledge)

NOT be answered solely by knowing immediate operator actions.

NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs.(Knowledge of how AOP-8.0 is implemented in parallel with AOP-3.0 is required.)

NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

CAN be answered with knowledge of ONE or MORE of the following:

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

Tuesday, March 19, 2013 11:38:42 AM 30

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps. (Knowledge of how AOP-8.0 is implemented in parallel with AOP-3.0 is required.).

Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures.

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.

Tuesday, March 19, 2013 11:38:42 AM 31

05/02/12 14:30:32 UNIT 1 FNP-1-AOP-8.0 March 21, 2012 Version 22.1 FARLEY NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE See pages 2, 3, 7 FNP-1-AOP-8.0 S

A F

E T

Y PARTIAL LOSS OF CONDENSER VACUUM R

E L

A T

E D

PROCEDURE LEVEL OF USE CLASSIFICATION PER NMP-AP-003 CATEGORY SECTIONS Continuous: ALL

Reference:

NONE Information: NONE Approved:

David L. Reed (for)

Operations Manager Effective Date: March 26, 2012

05/02/12 14:30:32 FNP-1-AOP-8.0 UNIT 1 PARTIAL LOSS OF CONDENSER VACUUM Version 22.1 Table of Contents PROCEDURE CONTAINS NUMBER OF PAGES Body 9 Figure 1 1

[] A A A 1 Page 1 of 1

05/02/12 14:30:32 FNP-1-AOP-8.0 UNIT 1 PARTIAL LOSS OF CONDENSER VACUUM Version 22.1 A. Purpose This procedure provides actions to prevent a turbine trip following a partial loss of condenser vacuum.

This procedure is applicable in Modes 1 and 2.

B. Symptoms or Entry Conditions I. This procedure is entered when a partial loss of condenser vacuum is indicated by any of the following:

a. Rapidly falling condenser vacuum.

b. Loss of one or both circulating water pumps.

c. Actuation of TURB COND VAC LO annunciator KK1

1.485 psia when < 25% turbine power

2.901 psia when > 47.9% turbine power

Varies Linearly Between 25% (1.485 psia) and 47.9% (2.901 psia)

d. Loss of the 12Kv linede-energizes 16 cooling tower fans.

1 Page 1 of 9

05/02/12 14:30:32 FNP-1-AOP-8.0 UNIT 1 PARTIAL LOSS OF CONDENSER VACUUM Version 22.1 Step Action/Expected Response Response Not Obtained

° NOTE:

DEHC point CNDP1 displays condenser pressure in Hga absolute on the point detail page.

On all other pages CNDP1 displays condenser pressure in psia.

IPC points PT0214 and PT0215 display condenser pressure in psia.

MCB recorder PR 4029 displays condenser pressure in psia.

Main turbine trip will occur at 4.351 psia

SGFP trip will occur at 5.9 psia

DEH Condenser Vacuum Display AND Figure 2 provides condenser vacuum setpoints for various turbine power levels.

1

__ 1 [CA] Monitor Condenser pressure 1 1.1 1.1 IF annunciator KK1, TURB COND VAC 1.1 LO in alarm OR setpoint exceeded, THEN increased monitoring of condenser pressure is required:

1.485 psia when < 25% turbine power

2.901 psia when > 47.9% turbine power At 10:15 - Annunciator

Varies Linearly Between 25% KK2 is in alarm, so RNO (1.485 psia) and 47.9% (2.901 psia) actions are required -

2nd part of the question 2

__ 2 [CA] Monitor turbine trip criteria. 2 2.1 2.1 Check condenser pressure less than 2.1 Perform the following.

annunciator KK2, TURB COND VAC LO-LO setpoint for existing turbine power 2.1.1 using MWe or PT-446/447: 2.1.1 IF reactor power greater than 35%,

THEN trip the reactor.

2.1.1.1

1.885 psia when < 25% turbine power 2.1.1.1 Go to FNP-1-EEP-0, REACTOR TRIP OR SAFETY INJECTION.

3.8 psia when > 55.9% turbine power 2.1.1.2

Varies Linearly Between 25% 2.1.1.2 Perform the remainder of this (1.885 psia) and 55.9% (3.8 psia) procedure in conjunction with FNP-1-EEP-0, REACTOR TRIP OR SAFETY INJECTION.

° Step 2 continued on next page

__Page Completed 8 ProcedureStepsMain Page 2 of 9

05/02/12 14:30:32 FNP-1-AOP-8.0 UNIT 1 PARTIAL LOSS OF CONDENSER VACUUM Version 22.1 Step Action/Expected Response Response Not Obtained 2.1.2° 2.1.2 IF reactor power less than 35%,

THEN place MAIN TURB EMERG Reactor power is 28%, so TRIP to TRIP for 5 seconds.

these are the actions 2.1.2.1 required. 2.1.2.1 Perform FNP-1-AOP-3.0, TURBINE TRIP BELOW P-9 SETPOINT in parallel with this procedure.

2.1.3 2.1.3 Proceed to step 3.

NOTE:

Loss of the 12Kv line will result in the loss of 16 cooling tower fans. See FIGURE 1.

On a loss of cooling tower fans, condenser Circ Water inlet temperature should be monitored as a leading indicator for condenser vacuum impact. These points may be found on the Circ Water system mimic on the IPC.

Any ramp in progress should not be stopped until condenser outlet temperature is less than 120F 3

__ 3 Stabilize condenser vacuum using any or 3 all of the following actions based on plant conditions, and the rate at which vacuum is worsening.

3.1 3.1 IF the rate of condenser pressure increase 3.1 is significant and approaching annunciator KK1, TURB COND VAC LO setpoint, THEN reduce load prior to reaching annunciator KK1 setpoint,.

3.2 3.2 IF condenser pressure increases above 3.2 annunciator KK1, TURB COND VAC LO setpoint, THEN reduce turbine load as necessary to maintain pressure at or below annunciator KK1 setpoint.

3.3 3.3 IF available, 3.3 THEN start additional cooling tower fans.

° Step 3 continued on next page

__Page Completed 8 ProcedureStepsMain Page 3 of 9

05/02/12 14:30:32 FNP-1-AOP-8.0 UNIT 1 PARTIAL LOSS OF CONDENSER VACUUM Version 22.1 Step Action/Expected Response Response Not Obtained 3.4 3.4 IF the loss of condenser vacuum is due to 3.4 the loss of the electrical ring bus, THEN notify ACC to restore the ring bus.

3.5° 3.5 IF the loss of condenser vacuum is due to 3.5 loss of the electrical ring bus, AND condenser vacuum has been restored, THEN return to procedure and step in effect.

NOTE: Normal SJAE alignment is one section per SJAE. Starting a second section on a SJAE may worsen vacuum if SJAEs are malfunctioning.

3.6 3.6 Verify proper operation of on service 3.6 Swap SJAEs or place additional SJAE SJAEs. sections in service as required to obtain proper SJAE operation using FNP-1-SOP-28.5, CONDENSER AIR REMOVAL SYSTEM.

3.7 3.7 IF available, 3.7 THEN start an additional CW PUMP.

4

__ 4 Dispatch personnel to check main turbine 4 gland sealing steam pressures.

4.1 4.1 Check HP Gland seal header pressure 4.1 Perform the following:

maintained at ~125 psig.

4.1.1 GS STM PRESS 4.1.1 IF HP gland seal header pressure

[ ] PI 4069B abnormal due to HP regulator malfunction, THEN transfer control to the HP regulator control valve bypass.

° Step 4 continued on next page

__Page Completed 8 ProcedureStepsMain Page 4 of 9

05/02/12 14:30:32 FNP-1-AOP-8.0 UNIT 1 PARTIAL LOSS OF CONDENSER VACUUM Version 22.1 Step Action/Expected Response Response Not Obtained a) Open N1N32V527, HP Regulator Control Vlv Bypass breaker (MCC located South of SJAE's).

Gland Seal HP Bypass MOV

[ ] N1R17BKRHBBD2

° b) Manually control HP gland sealing steam pressure. (east side of HP turbine)

HP Regulator Control Vlv Bypass

[ ] N1N32V527 throttled open (east side of HP turbine)

GS HP Reg Control Vlv Outlet Iso

[ ] N1N32V538 - throttled closed (east side of HP turbine)

Control HP gland sealing steam pressure 120-130 psig.

[ ] N1N32PI505 (east side of HP turbine)

° Step 4 continued on next page

__Page Completed 8 ProcedureStepsMain Page 5 of 9

05/02/12 14:30:32 FNP-1-AOP-8.0 UNIT 1 PARTIAL LOSS OF CONDENSER VACUUM Version 22.1 Step Action/Expected Response Response Not Obtained 4.1.2° 4.1.2 IF supply pressure low due to lack of steam supply, THEN verify gland sealing steam supply alignment based on intended supply source.

For main steam supply as the intended supply source, verify main steam aligned.

[ ] N1N11V606, MN STEAM ISO TO GLAND SEAL HP REG - OPEN (Turbine Bldg overhead above 155, above purity monitor)

For Unit 1 auxiliary steam supply as the intended source, verify auxiliary steam aligned.

[ ] N1N11V612, MS TO AS HDR MANUAL ISO - OPEN (Turb bldg 155 south side)

[ ] N1P20V907, MAIN STEAM TO AUX STM HDR PCV - ADJUSTED to maintain ~185 psig (Turb bldg 155 south side)

[ ] N1P20V525, MS TO AS ISO (MOV) -

OPEN

[ ] N1P20V519, ASB STEAM SUPP TO AUX BLDG (MOV) - OPEN

[ ] N1P20V549, AS SUPPLY TO GS STEAM SYSTEM - OPEN (Turb bldg 155 above spillover valve)

[ ] N1P20V905, AS SUPPLY TO GS STEAM SYSTEM - OPEN

For Unit 2 Auxiliary steam supply as the intended source verify proper alignment using FNP-1-SOP-55.1, AUXILIARY STEAM AND CONDENSATE SYSTEM.

° Step 4 continued on next page

__Page Completed 8 ProcedureStepsMain Page 6 of 9

05/02/12 14:30:32 FNP-1-AOP-8.0 UNIT 1 PARTIAL LOSS OF CONDENSER VACUUM Version 22.1 Step Action/Expected Response Response Not Obtained 4.2° 4.2 Check individual low pressure gland seal 4.2 IF any gland sealing steam regulator failed, regulators maintain 1-5 psig gland seal THEN perform the following:

pressures.

These are the actions being performed 4.2.1 Bypass the affected gland seal regulator 4.2.1 at 10:00 - 1st part of the question (reference FNP-1-SOP-28.4, GLAND SEALING STEAM SYSTEM or posted operator guidance).

NOTE: The high-pressure turbine becomes self-sealing at higher loads (about 25% load). At higher loads, the GS regulator closes off and the spill over valve opens to dump the excess steam leaking out of the HP turbine.

4.3 4.3 Check HP gland steam supply pressure 4.3 Perform the following:

maintained GREATER THAN 2 psig 4.3.1 (normal pressure 2-10 psig). 4.3.1 Verify gland sealing steam spillover operation. (155 Turb bldg behind MCC 1AA)

[ ] N1N32V528, SPILLOVER BYPASS -

HP CYL HDR

[ ] PI 4069A Plausible THROTTLED CLOSED as necessary to maintain 2-10 psig on HP gland seals Distracter [ ] N1N32V544 SPILLOVER VLV -

THROTTLED CLOSED as necessary to maintain 2-10 psig on HP gland seals 4.3.2 4.3.2 IF gland steam spillover isolated and HP gland seal regulator not maintaining pressure, THEN transfer control to HP gland seal regulator bypass (Turb bldg 189 east side of HP turbine).

[ ] N1N32V543, HP REGULATOR BYPASS - THROTTLED OPEN to maintain 2-10 psig on HP gland seals

[ ] N1N32V541, HP REGULATOR INLET ISO - THROTTLED CLOSED while bypass throttled open to maintain 2-10 psig on HP gland seals

° Step 4 continued on next page

__Page Completed 8 ProcedureStepsMain Page 7 of 9

05/02/12 14:30:32 FNP-1-AOP-8.0 UNIT 1 PARTIAL LOSS OF CONDENSER VACUUM Version 22.1 Step Action/Expected Response Response Not Obtained 4.4° 4.4 Check HP gland steam supply pressure 4.4 Verify proper operation of gland sealing maintained LESS THAN 10 psig. steam HP regulator and spillover valve using FNP-1-SOP-28.4, GLAND SEALING STEAM SYSTEM 5

__ 5 Dispatch personnel to verify SJAE steam 5 Manually adjust regulator to maintain greater supply regulator - OPERATING than 120 psig.

PROPERLY. (155 ft, TURB BLDG)

[ ] N1N51PI569A ~120 psig for steam jet air [ ] N1N51V595A for steam jet air ejector A on ejector A on service (155 Turb Bldg) service (155 Turb Bldg)

[ ] N1N51PI569B ~120 psig for steam jet air [ ] N1N51V595B for steam jet air ejector B on ejector B on service (155 Turb Bldg) service (155 Turb Bldg) 6

__ 6 Dispatch personnel to verify on service 6 Start standby GSSC EXH FAN using GSSC EXH FAN - OPERATING FNP-1-SOP-28.4, GLAND SEALING PROPERLY. (155 ft, TURB BLDG) STEAM SYSTEM.

[ ] N1N32PI503 ~16-22.5 inches water vacuum (155 Turb Bldg) 7

__ 7 Verify condenser vacuum breakers closed. 7 7.1 7.1 Verify COND VAC BKR VLVS 7.1 N1N51V519A/519B - CLOSED.

7.2 7.2 Verify COND VAC BKR MAN ISO 7.2 N1N51V518A and N1N51V518B -

CLOSED. (155 ft, TURB BLDG) 7.3 7.3 Verify condenser vacuum breaker water 7.3 seal - ESTABLISHED.

8

__ 8 Dispatch personnel to inspect main 8 condenser for leaks.

9

__ 9 Dispatch personnel to vent condenser 9 water boxes.

10

__ 10 Check hotwell level - LESS THAN 4 ft. 10 Manually control hotwell level.

° Step 10 continued on next page

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05/02/12 14:30:32 FNP-1-AOP-8.0 UNIT 1 PARTIAL LOSS OF CONDENSER VACUUM Version 22.1 Step Action/Expected Response Response Not Obtained 10.1° 10.1 Verify hotwell fill valve controlling properly.

HOTWELL FILL

[ ] CP 4055F adjusted 10.2 10.2 IF required, THEN establish hotwell flush using FNP-1-SOP-21.0, CONDENSATE AND FEEDWATER SYSTEM.

11

__ 11 IF condenser vacuum returned to normal, 11 Return to step 1.

THEN go to procedure and step in effect.

°

-END-

__Page Completed 8 ProcedureStepsMain Page 9 of 9

05/02/12 14:30:32 FNP-1-AOP-8.0 UNIT 1 PARTIAL LOSS OF CONDENSER VACUUM Version 22.1 FIGURE 1

° NOTE: Cooling tower fans within the bold borders will be deenergized if the electrical ring bus is deenergized.

West East West East West East Tower Tower Tower Tower Tower Tower Fan # Fan # Fan # Fan # Fan # Fan #

Breaker # Breaker # Breaker # Breaker # Breaker # Breaker #

1A9 1A10 1B9 1B10 1C8 1C9 EV07 EU07 EX07 EW07 EY06 EZ07 1A8 1A11 1B8 1B11 1C7 1C10 EU06 EV08 EW06 EX08 EZ06 EY07 1A7 1A12 1B7 1B12 1C6 1C11 EV06 EU08 EX06 EW08 EY05 EZ08 1A6 1A13 1B6 1B13 1C5 1C12 EU05 52-E103 EW05 52-E203 EZ05 EY08 1A5 1A14 1B5 1B14 1C4 1C13 EV05 52-E104 EX05 52-E204 EY04 52-E303 1A4 1A15 1B4 1B15 1C3 1C14 EU04 52-E105 EW04 52-E205 EZ04 52-E304 1A3 1A16 1B3 1B16 1C2 1C15 EV04 52-E106 EX04 52-E206 EY03 52-E305 1A2 1A17 1B2 1B17 1C1 1C16 EU03 52-E107 EW03 52-E207 EZ03 52-E306 1A1 1A18 1B1 1B18 Cooling Tower C EV03 52-E108 EX03 52-E208 Cooling Tower A Cooling Tower B NORTH

° S

1 ProcedureStepsMain Page 1 of 1

05/02/12 14:30:32 FNP-1-AOP-8.0 UNIT 1 PARTIAL LOSS OF CONDENSER VACUUM Version 22.1 FIGURE 2 1 ProcedureStepsMain Page 1 of 1

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

84. 054AA2.01 084 Unit 1 is at 37% power with the following conditions:

FCV-488, 1B SG FW FLOW, fails open and sticks open.

Given the following Tech Spec title:

3.7.3, Main Feedwater Stop Valves and Main Feedwater Regulation Valves (MFRVs) and Associated Bypass Valves With NO OPERATOR ACTIONS, which one of the following completes the statements below?

A Reactor Trip (1) occur.

Per the BASES of Tech Spec 3.7.3, the primary reason for automatic closure of the Main Feedwater Regulating valves is to prevent (2) .

(1) (2)

A. will NOT an excessive mass addition to the Steam Generator during a Steamline Break event B. will NOT a radiological release through the feedline penetration during a Steam Generator Tube Rupture event C. WILL an excessive mass addition to the Steam Generator during a Steamline Break event D. WILL a radiological release through the feedline penetration during a Steam Generator Tube Rupture event Tuesday, March 19, 2013 11:38:42 AM 32

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 For the conditions in the stem, the SG high level setpoint of 82% will be exceeded and this initiates a Turbine trip, SGFP trip, and FW isolation. Since Reactor power is >35%,

an automatic Reactor trip occurs due to the Turbine trip. If Reactor power was <35% an automatic Reactor trip would not occur due to the Turbine trip.

It is plausible that a candidate would recall the AMSAC Turbine Impulse enabling setpoint of 40% as being the setpoint associated with a Turbine trip initiating a Reactor trip.

A. Incorrect - 1) Incorrect, an automatic Reactor trip would occur due to a Turbine trip when above 35% power.

Plausible since power is 37% and AMSAC system enabling setpoint is 40% power. A candidate may incorrectly conclude that an automatic Reactor trip would not occur since Reactor power is <40%.

2) Correct, the Bases for TS 3.7.3 states "The design basis of the MFRVs and Main FW Stop Valves is primarily established by the analyses for the large SLB. Failure of a Main FW Stop Valve and MFRV, or Main FW Stop Valve and MFRV bypass valve to close following an SLB or an excess feedwater event can result in additional mass and energy being delivered to the steam generators, contributing to cooldown. This failure also results in additional mass and energy releases following an SLB or FWLB event."

B. Incorrect - 1) Incorrect, see A.1.

2) Incorrect, Plausible since most valves that isolate Containment are required to close due to radiological release concerns during a LOCA. Previous question statistics indicate that this response is plausible.

C. Correct - 1) Correct, SG high-high level of 82% initiates a Turbine trip, which initiates a Reactor trip when >35% power.

2) Correct, see A.2.

D. Incorrect - 1) Correct, see C.1.

2) Incorrect, see B.2.

K/A: 054AA2.01 Loss of Main Feedwater (MFW)

Ability to determine and interpret the following as they apply to the Loss of Main Feedwater (MFW):

(CFR: 43.5 / 45.13)

AA2.01 Occurrence of reactor and/or turbine trip . . . . . . . . .

. . . 4.3 4.4 Tuesday, March 19, 2013 11:38:43 AM 33

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 Importance Rating: 4.3 4.4 Technical

Reference:

Tech Specs v190/186 Tech Spec Bases v58 References provided: None Learning Objective: Given a set of Plant Conditions ACCESS those conditions and DETERMINE the ability of plant equipment and structures to meet their intended, designated function (OPS-52302A06)

ANALYZE plant conditions and DETERMINE if actuation or reset of any Engineered Safety Features Actuation Signal (ESFAS) is necessary. (OPS-52530A05)

Question History: Modified FNP Bank AOP-13.0-62520M01 001 K/A match: A FW supply valve has failed open. This causes a high-high level on the associated SG. The high-high level trips the SGFPs and causes a FW isolation signal (Loss of Main Feedwater), as well as a Turbine trip. The candidate has to determine if an automatic Reactor trip occurs due to the high-high SG level.

SRO justification: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

Knowledge of TS bases that is required to analyze TS required actions and terminology.

Tech Spec Bases provides the background information that the primary reason for the FW isolation signal is to minimize the mass in Containment that can turn to steam and potentially challenge the Containment barrier. The FW Isolation signal minimizes the mass of water added to a SG during accident conditions.

2013 NRC exam From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 flowchart for 10 CFR 55.43(b)(2)

1) can NOT be answered by knowing less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Tech Specs.

2) can NOT be answered by knowing information listed "above-the-line".

3) can NOT be answered by knowing the TS Safety Limits or their bases.

4) Does involve one or more of the following for TS, TRM or ODCM:

Knowledge of TS bases that are required to analyze TS required actions and terminology.

Tuesday, March 19, 2013 11:38:43 AM 34

Main FW Stop Valves and MFRVs and Associated Bypass Valves B 3.7.3 BASES BACKGROUND MFRVs and associated bypass valves, or the Main FW Stop Valves (continued) isolate the feedwater line penetrating containment, and ensure that the consequences of events do not exceed the capacity of the containment heat removal systems.

The MFRVs and the Main FW Stop Valves are part of the Condensate and Feedwater System as described in the FSAR, Section 10.4.7 (Ref. 1).

APPLICABLE The design basis of the MFRVs and Main FW Stop Valves is primarily SAFETY ANALYSES established by the analyses for the large SLB. Although the Main FW Stop Valves are not specifically credited in the accident analyses, these islation valves provide a diverse backup isolation function to the MFRVs. Closure of the MFRVs and associated bypass valves, or Main Feedwater Stop Valves, may also be relied on to terminate an SLB for core response analysis and excess feedwater event upon the receipt of a steam generator water level high high signal.

Failure of a Main FW Stop Valve and MFRV, or Main FW Stop Valve and MFRV bypass valve to close following an SLB or an excess feedwater event can result in additional mass and energy being delivered to the steam generators, contributing to cooldown. This failure also results in additional mass and energy releases following an SLB or FWLB event.

The MFRVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO This LCO ensures that the MFRVs and their associated bypass valves or Main FW Stop Valves will isolate MFW flow to the steam generators, following an excess feedwater event or main steam line break. These valves will also isolate the nonsafety related portions from the safety related portions of the system.

This LCO requires that three MFRVs and associated bypass valves and three Main FW Stop Valves be OPERABLE. The MFRVs and the associated bypass valves and the Main FW Stop Valves are considered OPERABLE when isolation times are within limits and they close on the appropriate signal(s).

(continued)

Farley Units 1 and 2 B 3.7.3-2 Revision 0

Original FNP Bank question that was modified to develop QUESTIONS REPORT054AA2.01.

for ILT 36 COMBINED

1. 054AG2.1.27 086 Unit 1 is in Mode 3 with a Startup in progress, when the following event occurs:

FCV-489, 1B SG FW BYP FLOW, air line is severed between the controller and the positioner.

Which one of the following completes the statements below?

Per SOP-0.13, Recording Limiting Conditions For Operations, a(n) (1) be written.

The safety related function of the Main Feedwater Regulating Bypass valve is to prevent (2) during accident conditions.

(1) (2)

A. Administrative LCO may radiological release through the feedline penetration B. Administrative LCO may excessive mass added to the steam generator C. Mandatory LCO must radiological release through the feedline penetration D. Mandatory LCO must excessive mass added to the steam generator Tuesday, March 19, 2013 1:18:39 PM 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

85. 055EG2.2.12 085 Unit 1 is in Mode 3 with the following conditions:

A Grid disturbance caused a loss of all offsite power.

During the event, the 1-2A DG started and then tripped.

Operating DG status is as follows:

- 1C DG is supplying Unit 1.

- 1B DG is supplying Unit 1.

- 2B DG is supplying Unit 2.

ACC reports that the following lines have been restored:

- Webb 230 KV line

- Snowdoun 500KV line

#1 Auto Bank Transformer is out of service.

Which one of the following completes the statements below for an evaluation of Unit 1 Tech Specs for the current conditions?

Perform REQUIRED ACTION of Tech Spec 3.8.1, AC Sources -

Operating, (1) .

A 25% extension of the COMPLETION TIME (2) allowed for the initial performance of SR 3.8.1.1 (STP-27.1, A.C. Source Verification).

Reference Provided (1) (2)

A. CONDITION A IS B. CONDITION B IS C. CONDITION A is NOT D. CONDITION B is NOT A Station Blackout has occurred due to a loss of grid event. Once the plant has been stabilized, Tech Specs have to be evaluated for the 1-2A DG trip and for the loss of offsite power supplies to the plant. For the evaluation of offsite power supplies, see Figure 1 of STP-27.1. The 4 Startup Transformers (2 for Unit 1 and 2 for Unit 2) are all powered from the 230KV side of the High Voltage Switchyard (HVSYD) .

There are 4 offsite lines on the 230KV side of the HVSYD, and 2 offsite lines on the 500KV side of the HVSYD. For the two 500KV offsite lines to be considered available, there has to be two Autotransformers to step down voltage from 500KV to 230KV. If only one of the Autotransformers is out of service, then only one of the 500KV offsite lines can be counted as available for offsite sources.

For the given conditions, there is one 230KV line in service and one 500KV line in service. There are normally two Auto bank transformers that will allow flow between the Tuesday, March 19, 2013 11:38:43 AM 35

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 230KV and 500KV sides of the High Voltage Switchyard, but for this situation there is only one Auto bank transformer available. When only one Auto bank transformer is available, only one 500KV line can be used as an off-site source. Therefore, two off-site sources are available, one from the 230KV side and one from the 500KV side.

A. Incorrect - 1) Incorrect, plausible since CONDITION A is entered if there is only one offsite line available in the HVSYD (at least two are required). For the conditions given, the candidate may think that there is only one Autotransformer available, and it is out of service. In addition, since the Startup Transformers are powered from the 230KV side of the Switchyard, a candidate may think that only the 230KV source is available and that the 500KV sources cannot be counted as offsite sources. This question requires detailed knowledge of the SR evaluation methodology, STP-27.1.

Selection of CONDITION A may be more plausible than CONDITION B because there are two DG's supplying Unit 1 power. That may lead a candidate to conclude that both DG sets are OPERABLE for Unit 1, so CONDITION B can't be correct and CONDITION A has to be correct.

2) Incorrect, plausible since a 25% grace is allowed for each subsequent performance of the SR. A candidate may not know that it is not allowed for the first performance. There is OE from our plant where there was a misconception that the 25% grace was always applicable.

B. Incorrect - 1) Correct, the 1-2A DG trip makes the Unit 1 'A' Train DG set INOPERABLE.

2) Incorrect, see A.2.

C. Incorrect - 1) Incorrect, see A.1.

2) Correct, per Tech Spec Bases for 3.0.2, a 25% grace is available for SR's that are required to be performed on a "once per X amount of time" basis. This is not applicable for the first performance of the SR.

D. Correct - 1) Correct, see B.1.

2) Correct, see C.2.

K/A: 055EG2.2.12 Loss of Offsite and Onsite Power (Station Blackout):

2.2.12 Knowledge of surveillance procedures.

(CFR: 41.10 / 45.13)

IMPORTANCE RO 3.7 SRO 4.1 Importance Rating: 3.7 4.1 Tuesday, March 19, 2013 11:38:43 AM 36

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 Technical

Reference:

FNP-1-STP-27.1, A.C. Source Verification, v37.1 Tech Specs v190/186 Tech Spec Bases v58 References provided: Tech Specs 3.8.1, pages 3.8.1-1 to 3.8.1-3 (with Conditon A

& B only)

Learning Objective: RECALL AND APPLY the information of the generic LCO requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4) including the BASES of the generic section , for any Technical Specifications or TRM requirements (OPS-62302A02): 10CFR55.43 (b) 2 Question History: New question K/A match: A Station Blackout event has occurred with component failures. Once the transient is controlled, the candidate is required to evaluate Tech Specs for the current conditions and determine if the required surveillance procedure has to be completed within the stated time, or if a 25% grace is allowed. The Tech Spec evaluation requires knowledge of the methodology of the STP-27.1 evaluation of offsite sources available.

SRO justification: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

Knowledge of TS bases that is required to analyze TS required actions and terminology.

Tech Spec Bases provides the background and explicit instructions for implementation of the 25% grace period allowed for certain performances of SR's.

2013 NRC exam From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 flowchart for 10 CFR 55.43(b)(2)

1) can NOT be answered by knowing less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Tech Specs.

2) can NOT be answered by knowing information listed "above-the-line".

3) can NOT be answered by knowing the TS Safety Limits or their bases.

4) Does involve one or more of the following for TS, TRM or ODCM:

Knowledge of TS bases that are required to analyze TS required actions and terminology.

Tuesday, March 19, 2013 11:38:43 AM 37

Reference provided to candidate AC Sources Operating for this question - TS 3.8.1 3.8.1 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources Operating LCO 3.8.1 The following AC electrical sources shall be OPERABLE:

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

b. Two diesel generator (DG) sets capable of supplying the onsite Class 1E power distribution subsystem(s); and

c. Automatic load sequencers for Train A and Train B.

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

ACTIONS

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

LCO 3.0.4b is not applicable to DGs.

CONDITION REQUIRED ACTION COMPLETION TIME A. One required offsite circuit A.1 Perform SR 3.8.1.1 for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months inoperable. required OPERABLE offsite circuit. AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND A.2 Declare required 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from feature(s) with no discovery of no offsite offsite power available power to one train inoperable when its concurrent with redundant required inoperability of feature(s) is inoperable. redundant required feature(s)

AND (continued)

Farley Units 1 and 2 3.8.1-1 Amendment No. 170 (Unit 1)

Amendment No. 163 (Unit 2)

AC Sources Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3 Restore required offsite 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months circuit to OPERABLE status. AND 13 days from discovery of failure to meet LCO B. One DG set inoperable. ------------------NOTE-------------------

LCO 3.0.4c is applicable when only one of the three DGs is inoperable.

B.1 Perform SR 3.8.1.1 for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months the required offsite circuit(s). AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND B.2 Declare required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from feature(s) supported by discovery of the inoperable DG set Condition B inoperable when its concurrent with required redundant inoperability of feature(s) is inoperable. redundant required feature(s)

AND B.3.1 Determine OPERABLE 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months DG set is not inoperable due to common cause failure.

OR (continued)

Farley Units 1 and 2 3.8.1-2 Amendment No. 170 (Unit 1)

Amendment No. 163 (Unit 2)

AC Sources Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.3.2 Perform SR 3.8.1.6 for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE DG set.

AND B.4 Restore DG set to 10 days OPERABLE status.

AND 13 days from discovery of failure to meet LCO C. Two required offsite C.1 Declare required 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months from circuits inoperable. feature(s) inoperable discovery of when its redundant Condition C required feature(s) is concurrent with inoperable. inoperability of redundant required features AND C.2 Restore one required 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months offsite circuit to OPERABLE status.

Farley Units 1 and 2 3.8.1-3 Amendment No. 146 (Unit 1)

Amendment No. 137 (Unit 2)

UNIT 1 FNP-1-STP-27.1 Ver 37.1 Farley Nuclear Plant Page Number 2/15/2012 06:14:57 A.C. SOURCE VERIFICATION 12 of 12 FIGURE 1 Onsite Startup Page 1 of 1 Transformers 1A and 1B

AC Sources Operating 3.8.1 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources Operating LCO 3.8.1 The following AC electrical sources shall be OPERABLE:

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

b. Two diesel generator (DG) sets capable of supplying the onsite Class 1E power distribution subsystem(s); and

c. Automatic load sequencers for Train A and Train B.

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

ACTIONS

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

LCO 3.0.4b is not applicable to DGs.

CONDITION REQUIRED ACTION COMPLETION TIME A. One required offsite circuit A.1 Perform SR 3.8.1.1 for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months inoperable. required OPERABLE offsite circuit. AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months This surveillance thereafter is also performed when Condition A AND is required A.2 Declare required 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from feature(s) with no discovery of no offsite offsite power available power to one train inoperable when its concurrent with redundant required inoperability of feature(s) is inoperable. redundant required feature(s)

AND (continued)

Farley Units 1 and 2 3.8.1-1 Amendment No. 170 (Unit 1)

Amendment No. 163 (Unit 2)

AC Sources Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3 Restore required offsite 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months circuit to OPERABLE status. AND 13 days from discovery of failure to meet LCO B. One DG set inoperable. ------------------NOTE-------------------

LCO 3.0.4c is applicable when only one of the three DGs is inoperable.

Condition B is required because the 1-2A DG is B.1 Perform SR 3.8.1.1 for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months tripped. For the "A" Train the required offsite DG's to be OPERABLE, circuit(s). AND both 1C and 1-2A DGs must be OPERABLE. Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND B.2 Declare required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from feature(s) supported by discovery of the inoperable DG set Condition B inoperable when its concurrent with required redundant inoperability of feature(s) is inoperable. redundant required feature(s)

AND B.3.1 Determine OPERABLE 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months DG set is not inoperable due to common cause failure.

OR (continued)

Farley Units 1 and 2 3.8.1-2 Amendment No. 170 (Unit 1)

Amendment No. 163 (Unit 2)

AC Sources Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.3.2 Perform SR 3.8.1.6 for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE DG set.

AND B.4 Restore DG set to 10 days OPERABLE status.

AND 13 days from discovery of failure to meet LCO C. Two required offsite C.1 Declare required 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months from circuits inoperable. feature(s) inoperable discovery of when its redundant Condition C required feature(s) is concurrent with inoperable. inoperability of redundant required features AND C.2 Restore one required 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months offsite circuit to OPERABLE status.

Farley Units 1 and 2 3.8.1-3 Amendment No. 146 (Unit 1)

Amendment No. 137 (Unit 2)

SR Applicability B 3.0 BASES SR 3.0.1 Surveillances, including Surveillances invoked by Required Actions, do (continued) not have to be performed on inoperable equipment because the ACTIONS define the remedial measures that apply. Surveillances have to be met and performed in accordance with SR 3.0.2, prior to returning equipment to OPERABLE status.

Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2. Post maintenance testing may not be possible in the current MODE or other specified conditions in the Applicability due to the necessary unit parameters not having been established. In these situations, the equipment may be considered OPERABLE provided testing has been satisfactorily completed to the extent possible and the equipment is not otherwise believed to be incapable of performing its function. This will allow operation to proceed to a MODE or other specified condition where other necessary post maintenance tests can be completed.

SR 3.0.2 SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances and any Required Action with a Completion Time that requires the periodic performance of the Required Action on a "once per . . ." interval.

SR 3.0.2 permits a 25% extension of the interval specified in the Frequency. This extension facilitates Surveillance scheduling and considers plant operating conditions that may not be suitable for conducting the Surveillance (e.g., transient conditions or other ongoing Surveillance or maintenance activities).

The 25% extension does not significantly degrade the reliability that results from performing the Surveillance at its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the SRs. The exceptions to SR 3.0.2 are those Surveillances for which the 25% extension of the interval specified in the Frequency does not apply.

These exceptions are stated in the individual Specifications. An example of where SR 3.0.2 does not apply is the Containment Leakage Rate Testing Program.

(continued)

Farley Units 1 and 2 B 3.0-14 Revision 35

SR Applicability B 3.0 BASES SR 3.0.2 (continued) As stated in SR 3.0.2, the 25% extension also does not apply to the initial portion of a periodic Completion Time that requires performance on a "once per ..." basis. The 25% extension applies to each performance after the initial performance. The initial performance of the Required Action, whether it is a particular Surveillance or some other remedial action, is considered a single action with a single Completion Time. One reason for not allowing the 25% extension to this Completion Time is that such an action usually verifies that no loss of function has occurred by checking the status of redundant or diverse components or accomplishes the function of the inoperable equipment in an alternative manner.

The provisions of SR 3.0.2 are not intended to be used repeatedly merely as an operational convenience to extend Surveillance intervals (other than those consistent with refueling intervals) or periodic Completion Time intervals beyond those specified.

SR 3.0.3 SR 3.0.3 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a Surveillance has not been completed within the specified Frequency. A delay period of up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified Frequency, whichever is greater, applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2, and not at the time that the specified Frequency was not met.

This delay period provides adequate time to complete Surveillances that have been missed. This delay period permits the completion of a Surveillance before complying with Required Actions or other remedial measures that might preclude completion of the Surveillance.

The basis for this delay period includes consideration of unit conditions, adequate planning, availability of personnel, the time required to perform the Surveillance, the safety significance of the delay in completing the required Surveillance, and the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the requirements.

(continued)

Farley Units 1 and 2 B 3.0-15 Revision 35

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

86. 058AA2.02 086 Unit 1 is at 100% power when the following conditions occur:

VC4, 1B BATT CHG FAULT OR DISC, alarms.

EEO5, 1B BATTERY CHARGER SUPPLY BREAKER, has tripped open.

1B DC Bus Voltage is 130 Volts.

Given the following Tech Spec titles:

3.8.4, DC Sources-Operating

3.8.9, Distribution Systems-Operating Which one of the following completes the statement below?

is(are) required to be implemented.

Reference Provided A. NEITHER Tech Spec 3.8.4 nor Tech Spec 3.8.9 B. ONLY Tech Spec 3.8.4 C. ONLY Tech Spec 3.8.9 D. BOTH Tech Spec 3.8.4 and Tech Spec 3.8.9 Tuesday, March 19, 2013 11:38:43 AM 38

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 The Bases of TS 3.8.4 in the LCO section states:

An OPERABLE DC electrical power subsystem requires all required batteries and respective chargers to be operating and connected to the associated DC bus(es).

The Bases of TS 3.8.9 in the LCO section states:

OPERABLE DC electrical power distribution subsystems require the associated buses to be energized to their proper voltage from either the associated battery or charger.

A. Incorrect - REQUIRED ACTIONS of TS 3.8.4 are required to be implemented.

Plausible that a candidate could determine that no LCO is required since the bus has power available and the Batteries are rated to carry the bus for a minimum of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months with no Charger aligned. There are questions on this exam and in our exam bank where the correct answer is neither or no action, however in this case this is not correct.

B. Correct - Both a Battery and Charger in operation and supplying proper voltage and connected to the DC bus are required to ensure OPERABILITY per TS 3.8.4.

C. Incorrect - Both a Battery and Charger in operation and supplying proper voltage and connected to the DC bus are required to ensure OPERABILITY per TS 3.8.4. Per TS 3.8.9. Either a Battery or Charger supplying proper voltage is required to ensure OPERABILITY per TS 3.8.9. It is plausible that a candidate may determine the wrong requirements for this Tech Specs.

D. Incorrect - This is a plausible answer choice since both of these TS's have requirements for the DC system OPERABILITY.

K/A: 058AA2.02 Loss of DC Power Ability to determine and interpret the following as they apply to the Loss of DC Power:

(CFR: 43.5 / 45.13)

AA2.02 125V dc bus voltage, low/critical low, alarm . . . . . .

3.3* 3.6 Importance Rating: 3.3 3.6 Technical

Reference:

Tech Specs v190/186 Tech Spec Bases v58 References provided: Tech Specs 3.8.4 and 3.8.9 (the TS only and not the SR's),

pages 3.8.4-1, 3.8.9-1, and 3.8.9-2 Learning Objective: RECALL AND APPLY the information from the LCO BASES sections: BACKGROUND, APPLICABLE SAFETY ANALYSIS, ACTIONS, SURVEILLENCE REQUIREMENTS, for any Technical Specifications or TRM requirements Tuesday, March 19, 2013 11:38:43 AM 39

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 associated with the DC Distribution System components and attendant equipment alignment, to include the following:

(OPS-62103C01) 10CFR55.43 (b) 2 3.8.4, DC Sources - Operating 3.8.5, DC Sources - Shutdown 3.8.6, Battery Cell Parameters 3.8.9, Distribution Systems - Operating 3.8.10, Distribution Systems - Shutdown Question History: 058AA2.02 89 from Farley Test Bank K/A match: A Battery Charger has tripped off causing an alarm and a loss of Operability per the TS for DC Sources - Operating.

The candidate has to interpret the effects of the loss of a Battery Charger with bus voltage still being maintained by the Battery, and determine the proper application of Tech Spec requirements per TS Bases.

SRO justification: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

Knowledge of TS bases that is required to analyze TS required actions and terminology.

Both of these Tech Specs have a portion related to the DC electrical system. The requirements for OPERABILITY are not defined in the TS. TS Bases knowledge is required to be able to determine the TS applicability.

2013 NRC exam From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 flowchart for 10 CFR 55.43(b)(2)

1) can NOT be answered by knowing less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Tech Specs.

2) can NOT be answered by knowing information listed "above-the-line".

3) can NOT be answered by knowing the TS Safety Limits or their bases.

4) Does involve one or more of the following for TS, TRM or ODCM:

Knowledge of TS bases that are required to analyze TS required actions and terminology.

Tuesday, March 19, 2013 11:38:43 AM 40

Reference provided to candidate for this question - TS 3.8.4 & 3.8.9 DC Sources Operating 3.8.4 3.8 ELECTRICAL POWER SYSTEMS 3.8.4 DC Sources Operating LCO 3.8.4 The Train A and Train B Auxiliary Building and Service Water Intake Structure (SWIS) DC electrical power subsystems shall be OPERABLE.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One Auxiliary Building DC A.1 Restore the Auxiliary 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months electrical power subsystem Building DC electrical inoperable. power subsystem to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for 1B OPERABLE status. Auxiliary Building DC electrical power subsystem inoperable due to inoperable battery for cycle 19 only B. One Auxiliary Building DC B.1 Restore the battery 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months electrical power subsystem connection resistance to with battery connection within limit.

resistance not within limit.

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

C.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months D. One required SWIS DC D.1 Restore the battery 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months electrical power connection resistance to subsystem battery within the limit.

connection resistance not within limit.

E. One required SWIS DC E.1 Declare the associated Immediately electrical power Service Water System subsystem inoperable. train inoperable.

OR Required Action and associated Completion Time of Condition D not met.

Farley Units 1 and 2 3.8.4-1 Amendment No. 164 (Unit 1)

Amendment No. 137 (Unit 2)

Distribution Systems Operating 3.8.9 3.8 ELECTRICAL POWER SYSTEMS 3.8.9 Distribution Systems Operating LCO 3.8.9 Train A and Train B AC, DC, and AC vital bus electrical power distribution subsystems shall be OPERABLE.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more AC electrical A.1 Restore AC electrical 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months power distribution power distribution subsystems inoperable. subsystem(s) to AND OPERABLE status.

16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months from discovery of failure to meet LCO B. One or more AC vital B.1 Restore AC vital bus 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months buses inoperable. subsystem(s) to OPERABLE status. AND 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months from discovery of failure to meet LCO C. One Auxiliary Building DC C.1 Restore Auxiliary 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months electrical power distribution Building DC electrical subsystem inoperable. power distribution AND subsystem to OPERABLE status. 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months from discovery of failure to meet LCO Farley Units 1 and 2 3.8.9-1 Amendment No. 146 (Unit 1)

Amendment No. 137 (Unit 2)

Distribution Systems Operating 3.8.9 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A, B, or AND C not met.

D.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months E. One Service Water Intake E.1 Declare the associated Immediately Structure (SWIS) DC Service Water train electrical power distribution inoperable.

subsystem inoperable.

F. Two trains with inoperable F.1 Enter LCO 3.0.3. Immediately distribution subsystems that result in a loss of safety function.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.9.1 Verify correct breaker alignments and voltage to In accordance with required AC, DC, and AC vital bus electrical power the Surveillance distribution subsystems. Frequency Control Program Farley Units 1 and 2 3.8.9-2 Amendment No. 185 (Unit 1)

Amendment No. 180 (Unit 2)

DC Sources Operating 3.8.4 3.8 ELECTRICAL POWER SYSTEMS 3.8.4 DC Sources Operating LCO 3.8.4 The Train A and Train B Auxiliary Building and Service Water Intake Structure (SWIS) DC electrical power subsystems shall be OPERABLE.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One Auxiliary Building DC A.1 Restore the Auxiliary 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months electrical power subsystem Building DC electrical inoperable. power subsystem to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for 1B OPERABLE status. Auxiliary Building DC CONDITION A of TS 3.8.4 is electrical power reqired due to the loss of the subsystem inoperable due to 1B Battery Charger (see inoperable battery for Bases) cycle 19 only B. One Auxiliary Building DC B.1 Restore the battery 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months electrical power subsystem connection resistance to with battery connection within limit.

resistance not within limit.

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

C.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months D. One required SWIS DC D.1 Restore the battery 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months electrical power connection resistance to subsystem battery within the limit.

connection resistance not within limit.

E. One required SWIS DC E.1 Declare the associated Immediately electrical power Service Water System subsystem inoperable. train inoperable.

OR Required Action and associated Completion Time of Condition D not met.

Farley Units 1 and 2 3.8.4-1 Amendment No. 164 (Unit 1)

Amendment No. 137 (Unit 2)

DC Sources Operating 3.8.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify battery terminal voltage is 127.8 V on float In accordance charge. with the Surveillance Frequency Control Program SR 3.8.4.2 Verify no visible corrosion at battery terminals and In accordance connectors. with the Surveillance OR Frequency Control Program Verify post-to-post battery connection resistance of each cell-to-cell and terminal connection is 150 microhms for the Auxiliary Building batteries and 1500 microhms for the SWIS batteries.

SR 3.8.4.3 Verify battery cells, cell plates, and racks show no In accordance visual indication of physical damage or abnormal with the deterioration. Surveillance Frequency Control Program SR 3.8.4.4 Remove visible terminal corrosion, verify battery In accordance cell- to-cell and terminal connections are coated with with the anti-corrosion material. Surveillance Frequency Control Program SR 3.8.4.5 Verify post-to-post battery connection resistance of In accordance each cell-to-cell and terminal connection is 150 with the microhms for the Auxiliary Building batteries and Surveillance 1500 microhms for the SWIS batteries Frequency Control Program Farley Units 1 and 2 3.8.4-2 Amendment No. 185 (Unit 1)

Amendment No. 180 (Unit 2)

DC Sources Operating 3.8.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.6 -------------------------------NOTE-------------------------------

This Surveillance may be performed in MODE 1, 2, 3, 4, 5, or 6 provided spare or redundant charger(s) placed in service are within surveillance frequency to maintain DC subsystem(s) OPERABLE.

Verify each required Auxiliary Building battery In accordance with charger supplies 536 amps at 125 V for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the Surveillance and each required SWIS battery charger supplies Frequency Control 3 amps at 125 V for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Program SR 3.8.4.7 -----------------------------NOTES-------------------------------

1. The performance discharge test in SR 3.8.4.8 may be performed in lieu of the service test in SR 3.8.4.7 once per 60 months.

2. The modified performance discharge test in SR 3.8.4.8 may be performed in lieu of the service test at any time.

3. This Surveillance shall not be performed for the Auxiliary Building batteries in MODE 1, 2, 3, or 4.

Verify battery capacity is adequate to supply, and In accordance with maintain in OPERABLE status, the required the Surveillance emergency loads for the design load profile described Frequency Control in the Final Safety Analysis Report, Section 8.3.2, by Program subjecting the battery to a service test.

Farley Units 1 and 2 3.8.4-3 Amendment No. 185 (Unit 1)

Amendment No. 180 (Unit 2)

DC Sources Operating 3.8.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.8 -------------------------------NOTE-------------------------------

This Surveillance shall not be performed for the Auxiliary Building batteries in MODE 1, 2, 3, or 4.

Verify battery capacity is 80% of the manufacturer's In accordance with rating when subjected to a performance discharge the Surveillance test or a modified performance discharge test. Frequency Control Program AND 18 months when battery shows degradation or has reached 85% of expected life or 17 years, whichever comes first Farley Units 1 and 2 3.8.4-4 Amendment No. 185 (Unit 1)

Amendment No. 180 (Unit 2)

DC Sources Operating B 3.8.4 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.4 DC Sources Operating BASES BACKGROUND The station DC electrical power system provides the AC emergency power system with control power. It also provides both motive and control power to selected safety related equipment and preferred AC vital bus power (via inverters). As required by 10 CFR 50, Appendix A, GDC 17 (Ref. 1), the DC electrical power system is designed to have sufficient independence, redundancy, and testability to perform its safety functions, assuming a single failure. The DC electrical power system also conforms to the recommendations of Regulatory Guide 1.6 (Ref. 2) and IEEE-308 (Ref. 3).

The 125 VDC electrical power system consists of two main systems.

The Auxiliary Building System and the Service Water Intake Structure (SWIS) System. The Auxiliary Building 125 VDC system consists of two independent and redundant subsystems (Train A and Train B) which supply DC power to various ESF systems throughout the plant.

Each Auxiliary Building subsystem (train) consists of a 125 VDC battery, an associated full capacity battery charger and all associated control equipment and interconnecting cabling. Each Auxiliary Building 125 VDC train is normally supplied by the associated battery charger (A or B). In the event of an A or B battery charger failure, battery charger C, the full capacity swing battery charger, may supply power to either train. Either train may be considered OPERABLE when supplied from battery charger C. Battery charger C input and output breakers are interlocked to prevent supplying power to a DC bus from the opposite train. Both the Auxiliary Building 125 VDC source subsystems (Train A and B) are required OPERABLE by this LCO.

The SWIS 125 VDC system provides a reliable source of power for controls, power loads, annunciation and alarms primarily for the safety-related Service Water System. The SWIS 125 VDC system consists of four battery/battery charger subsystems. Each subsystem consists of a 125 VDC battery and full capacity battery charger. The subsystems are divided into Train A and Train B which are shared between the two units. Each of the 4 subsystems can supply 100% of the required capacity for the associated train. Subsystems 1 and 2 are associated with Train A, with subsystem 1 being the normal (continued)

Farley Units 1 and 2 B 3.8.4-1 Revision 0

DC Sources Operating B 3.8.4 BASES BACKGROUND supply, and subsystem 2 the standby supply. Subsystems 3 and 4 (continued) are associated with Train B, with subsystem 3 being the normal supply and subsystem 4 the standby supply. Each train has a manual transfer switch which is used to select which of the two available SWIS subsystems supplies that train. One SWIS subsystem is required OPERABLE for each train.

During normal operation, the 125 VDC load is powered from the battery chargers with the batteries floating on the system. In case of loss of normal power to the battery charger, the DC load is automatically powered from the station batteries.

The Train A and Train B DC electrical power subsystems provide the control power for its associated Class 1E AC power load group, 4.16 kV switchgear, and 600 V load centers. The DC electrical power subsystems also provide DC electrical power to the inverters, which in turn power the AC vital buses.

The DC power distribution system is described in more detail in Bases for LCO 3.8.9, "Distribution System Operating," and LCO 3.8.10, "Distribution Systems Shutdown."

Each train of 125 VDC batteries is separately housed in a ventilated room apart from its charger and distribution centers. Each subsystem is located in an area separated physically and electrically from the other subsystem to ensure that a single failure in one subsystem does not cause a failure in a redundant subsystem. There is no sharing between redundant Class 1E subsystems, such as batteries, battery chargers, or distribution panels.

The Auxiliary Building batteries are stationary type consisting of 60 individual lead-calcium cells electrically connected in series to establish a nominal 125VDC power supply. Under both normal and accident conditions the batteries are capable of providing the required voltage for component operation considering an aging factor of 25%

and minimum electrolyte temperature of 60°F. The battery float voltage is 2.20V per cell average and 132V total terminal voltage.

During an LOSP or LOSP with SI, the Auxiliary Building batteries supply safety-related loads for a period of less than one minute duration without charger support. The design is such that subsequent to LOSP, the battery chargers are re-energized by the Diesel Generators within one minute.

(continued)

Farley Units 1 and 2 B 3.8.4-2 Revision 0

DC Sources Operating B 3.8.4 BASES BACKGROUND Although not a requirement for the mitigation of design basis events, (continued) each battery is capable of providing LOSP or LOSP plus SI loads for a period of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months assuming the single failure loss of the battery charger aligned at the onset of the event. During such an occurrence, the redundant train battery with its connected charger remains fully capable of providing DC power to redundant train safety-related loads.

The batteries also have the capacity to supply normal operating loads for a period of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months without charger support as discussed in the FSAR Chapter 8.3 (Ref. 4). The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months period of time is adequate to allow alignment of the spare battery charger to the affected battery without disrupting continued operation.

The SWIS batteries are stationary type consisting of individual lead-calcium cells electrically connected in series to establish a nominal 125 VDC power supply. They are sized to furnish the anticipated vital loads without dropping below a total battery voltage of 105 V. Under both normal and accident conditions the batteries are capable of providing the required voltage for component operation considering an aging factor of 25% and a minimum electrolyte temperature of 35°F.

The battery float voltage is 2.20 V per cell average and 132 V total.

Each SWIS battery subsystem has adequate capacity to carry its loads without charger support for a period of at least 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months as discussed in the FSAR, Chapter 8.3 (Ref. 4).

Each Train A and Train B DC electrical power subsystem has ample power output capacity for the steady state operation of connected loads required during normal operation, while at the same time maintaining its battery bank fully charged. Each battery charger has adequate capacity to restore its battery to full charge after the battery has been discharged while carrying steady-state normal or emergency loads. The time required to recharge the battery to full charge is compatible with the recommendation of the battery manufacturer (Ref. 4).

APPLICABLE The initial conditions of Design Basis Accident (DBA) and SAFETY ANALYSES transient analyses in the FSAR, Chapter 6 (Ref. 6), and in the FSAR, Chapter 15 (Ref. 7), assume that Engineered Safety Feature (ESF) systems are OPERABLE. The DC electrical power system provides normal and emergency DC electrical power for the DGs, emergency auxiliaries, and control and switching during all MODES of operation.

(continued)

Farley Units 1 and 2 B 3.8.4-3 Revision 0

DC Sources Operating B 3.8.4 BASES APPLICABLE The OPERABILITY of the DC sources is consistent with the initial SAFETY ANALYSES assumptions of the accident analyses and is based upon meeting the (continued) design basis of the unit. This includes maintaining the DC sources OPERABLE during accident conditions in the event of:

a. An assumed loss of all offsite AC power or all onsite AC power; and

b. A worst case single failure.

The DC sources satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Both the Auxiliary Building 125 VDC source subsystems (Train A and B) and two SWIS 125 VDC source subsystems (one in each train) including a battery charger for each Auxiliary Building and SWIS battery and the corresponding control equipment and interconnecting cabling supplying power to the associated bus within the train are required to be OPERABLE to ensure the availability of the required power to shut down the reactor and maintain it in a safe condition after an anticipated operational occurrence (AOO) or a postulated DBA. Loss of any train DC electrical power subsystem does not prevent the minimum safety function from being performed (Ref. 4).

An OPERABLE DC electrical power subsystem requires all required batteries and respective chargers to be operating and connected to the associated DC bus(es).

APPLICABILITY The DC electrical power sources are required to be OPERABLE in MODES 1, 2, 3, and 4 to ensure safe unit operation and to ensure that:

a. Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of AOOs or abnormal transients; and

b. Adequate core cooling is provided, and containment integrity and other vital functions are maintained in the event of a postulated DBA.

The DC electrical power requirements for MODES 5 and 6 are addressed in the Bases for LCO 3.8.5, "DC Sources Shutdown."

Farley Units 1 and 2 B 3.8.4-4 Revision 0

DC Sources Operating B 3.8.4 BASES ACTIONS A.1 Condition A represents one train of Auxiliary Building DC electrical power with a loss of ability to completely respond to an event, and a potential loss of ability to remain energized during normal operation. It is, therefore, imperative that the operator's attention focus on stabilizing the unit, minimizing the potential for complete loss of DC power to the affected train. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months limit is consistent with the allowed time for an inoperable DC distribution system train.

[For Unit 1 only for cycle 19] The second Completion time for Condition A represents the 1B train of Auxiliary Building DC electrical power subsystem due to an inoperable battery. With the 1B Auxiliary Building battery inoperable, the DC bus is being supplied by the OPERABLE battery charger. Any event that results in a loss of the AC bus supporting the battery charger will also result in the loss of DC to that train. Recovery of the AC bus, especially if it is due to a loss of offsite power, will be hampered by the fact that many of the components necessary for the recovery (e.g., diesel generator control and field flash, AC load shed and diesel generator output breakers, etc.) rely upon the battery. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months limit allows sufficient time to effect restoration of the inoperable battery given that the majority of the conditions that lead to battery inoperability (e.g., loss of battery charger, battery cell voltage less than 2.02 volts, etc.) are identified in Specifications 3.8.4, 3.8.5, and 3.8.6 together with additional specific completion times.

If one of the required DC electrical power subsystems is inoperable (e.g., inoperable battery, inoperable battery charger(s), or inoperable battery charger and associated inoperable battery), the remaining DC electrical power subsystem has the capacity to support a safe shutdown and to mitigate an accident condition. Since a subsequent worst case single failure would, however, result in the complete loss of the remaining 125 VDC electrical power subsystems with attendant loss of ESF functions, in the case of the Auxiliary Building DC power subsystem, continued power operation should not exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is based on Regulatory Guide 1.93 (Ref. 8) and reflects a reasonable time to assess unit status as a function of the inoperable DC electrical power subsystem and, if the Auxiliary Building DC electrical power subsystem is not restored to OPERABLE status, to prepare to effect an orderly and safe unit shutdown.

(continued)

Farley Units 1 and 2 B 3.8.4-5 Revision 25

DC Sources Operating B 3.8.4 BASES ACTIONS B.1 and D.1 (continued)

Conditions B and D represent one Auxiliary Building or SWIS DC electrical power subsystem with connection resistance not within the specified limit. Consistent with the guidance in IEEE-450, connection resistance not within the limit is an indication that the affected battery requires attention to restore the resistance to within the limit but is not a basis on which to declare the battery inoperable. Therefore, the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time allowed to restore the battery connection resistance to within the required limit is a reasonable time considering that variations in connection resistance do not mean the battery is incapable of performing its required safety function, but is an indication that the battery requires maintenance.

C.1 and C.2 If the inoperable Auxiliary Building DC electrical power subsystem cannot be restored to OPERABLE status or the connection resistance restored to within the limit within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply.

To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging plant systems. The Completion Time to bring the unit to MODE 5 is consistent with the time required in Regulatory Guide 1.93 (Ref. 8).

E.1 If a required SWIS DC electrical power subsystem is inoperable or the connection resistance is not restored to within the limit and the associated Completion Time has expired, the Service Water System train supported by the affected SWIS DC electrical power subsystem must be declared inoperable. The capability of the affected SWIS DC electrical power subsystem to fully support the associated train of Service Water is not assured. Therefore, consistent with the definition of OPERABILITY, the associated train of Service Water must be declared inoperable immediately, thereby limiting operation in this condition to the Completion Time associated with the affected Service Water System train.

(continued)

Farley Units 1 and 2 B 3.8.4-6 Revision 25

DC Sources Operating B 3.8.4 BASES (continued)

SURVEILLANCE SR 3.8.4.1 REQUIREMENTS Verifying battery terminal voltage while on float charge for the batteries helps to ensure the effectiveness of the charging system and the ability of the batteries to perform their intended function. Float charge is the condition in which the charger is applying a voltage to the battery to maintain it in a fully charged condition during normal operation. The float voltage of 2.2 V per cell or 132 V overall is higher than the nominal design voltage of 125 V and is consistent with the manufacturer's recommendations for maintaining a full charge.

Verifying that terminal voltage is 127.8 V provides assurance that the average of all cell voltages is maintained greater than 2.13 V.

Maintaining float voltage at the higher value of 2.2 V per cell prolongs cell life expectancy. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.8.4.2 Visual inspection to detect excessive corrosion on the battery terminals or connectors, or measurement of the post to post resistance of these items provides an indication of the need for cleaning and/or retorqueing.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.8.4.3 Visual inspection of the battery cells, cell plates, and battery racks provides an indication of physical damage or abnormal deterioration that could potentially degrade battery performance. The presence of physical damage or deterioration does not necessarily represent a failure of this SR, provided an evaluation determines that the physical damage or deterioration does not affect the OPERABILITY of the battery (its ability to perform its design function).

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

(continued)

Farley Units 1 and 2 B 3.8.4-7 Revision 52

DC Sources Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.4 and SR 3.8.4.5 REQUIREMENTS (continued) Visual inspection and post to post resistance measurements of battery terminals or connectors provide an indication of the need for cleaning and/or retorqueing. The anticorrosion material is used to help ensure good electrical connections and to reduce terminal deterioration. The visual inspection for corrosion is not intended to require removal of and inspection under each terminal connection.

The removal of visible corrosion is a preventive maintenance SR. The presence of visible corrosion does not necessarily represent a failure of this SR provided visible corrosion is removed during performance of SR 3.8.4.4.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR requires that each required battery charger be capable of supplying 536 amps (Auxiliary Building chargers) and 3 amps (SWIS chargers) at 125 V for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . These requirements are based on the design capacity of the chargers (Ref. 4). According to Regulatory Guide 1.32 (Ref. 10), the battery charger supply is required to be based on the largest combined demands of the various steady state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrespective of the status of the unit during these demand occurrences. The minimum required amperes and duration ensures that these requirements can be satisfied.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This surveillance is modified by a Note which clarifies that it may be performed in any mode of operation provided certain conditions are met. The design is such that any battery charger may be tested while a spare or redundant battery and/or charger is in service in its place.

The spare or redundant battery and/or charger must be within the 18 month surveillance frequency to maintain the DC subsystem(s) to which they are aligned OPERABLE. This operational flexibility maintains TS OPERABILITY of the applicable battery and DC train while testing the normally aligned charger.

(continued)

Farley Units 1 and 2 B 3.8.4-8 Revision 52

DC Sources Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.7 REQUIREMENTS (continued) A battery service test is a special test of battery capability, as found, to satisfy the design requirements (design load profile) of the DC electrical power system. The discharge rate and test length should correspond to the design load profile requirements as specified in Reference 4.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR is modified by three Notes. Note 1 allows the performance of a performance discharge test in lieu of a service test once per 60 months. Note 2 allows the performance of a modified performance discharge test in lieu of a service test at any time.

The modified performance discharge test is a simulated duty cycle consisting of just two rates: the one minute rate published for the battery or the largest current load of the duty cycle, followed by the test rate employed for the performance test, both of which envelop the duty cycle of the service test. Since the ampere-hours removed by a rated one minute discharge represents a very small portion of the battery capacity, the test rate can be changed to that for the performance test without compromising the results of the performance discharge test. The battery terminal voltage for the modified performance discharge test should remain above the minimum battery terminal voltage specified in the battery service test for the duration of time equal to that of the service test.

A modified discharge test is a test of the battery capacity and its ability to provide a high rate, short duration load (usually the highest rate of the duty cycle). This will often confirm the batterys ability to meet the critical period of the load duty cycle, in addition to determining its percentage of rated capacity. Initial conditions for the modified performance discharge test should be identical to those specified for a service test.

The reason for Note 3 is that performing the Surveillance for the Auxiliary Building batteries would perturb the electrical distribution system and challenge safety systems.

(continued)

Farley Units 1 and 2 B 3.8.4-9 Revision 52

DC Sources Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.8 REQUIREMENTS (continued) A battery performance discharge test is a test of constant current capacity of a battery, after having been in service, to detect any change in the capacity determined by the acceptance test. The test is intended to determine overall battery degradation due to age and usage.

A battery modified performance discharge test is described in the Bases for SR 3.8.4.7. Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.4.8. The modified performance discharge test may be used to satisfy SR 3.8.4.8 while simultaneously satisfying the requirements of SR 3.8.4.7 at any time. The performance discharge test may be used to satisfy 3.8.4.8 while simultaneously satisfying the requirements of SR 3.8.4.7 once per 60 months.

The acceptance criteria for this Surveillance are consistent with IEEE-450 (Ref. 9). This reference recommends that the battery be replaced if its capacity is below 80% of the manufacturer's rating. A capacity of 80% shows that the battery rate of deterioration is increasing, even if there is ample capacity to meet the load requirements.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. If the battery shows degradation, or if the battery has reached 85% of its expected life or 17 years, whichever comes first, the Surveillance Frequency is reduced to 18 months. Degradation is indicated, according to IEEE-450 (Ref. 9),

when the battery capacity drops by more than 10% relative to its capacity on the previous performance test or when it is > 10% below the manufacturer's rating.

This SR is modified by a Note. The reason for the Note is that performing the Surveillance for the Auxiliary Building batteries would perturb the electrical distribution system and challenge safety systems.

(continued)

Farley Units 1 and 2 B 3.8.4-10 Revision 52

DC Sources Operating B 3.8.4 BASES (continued)

REFERENCES 1. 10 CFR 50, Appendix A, GDC 17.

2. Regulatory Guide 1.6, March 10, 1971.

3. IEEE-308-1971.

4. FSAR, Section 8.3.

5. None.

6. FSAR, Chapter 6.

7. FSAR, Chapter 15.

8. Regulatory Guide 1.93, December 1974.

9. IEEE-450-1980.

10. Regulatory Guide 1.32, February 1972.

Farley Units 1 and 2 B 3.8.4-11 Revision 25

Distribution Systems Operating 3.8.9 3.8 ELECTRICAL POWER SYSTEMS 3.8.9 Distribution Systems Operating LCO 3.8.9 Train A and Train B AC, DC, and AC vital bus electrical power distribution subsystems shall be OPERABLE.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more AC electrical A.1 Restore AC electrical 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months power distribution power distribution subsystems inoperable. subsystem(s) to AND OPERABLE status.

16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months from discovery of failure to meet LCO B. One or more AC vital B.1 Restore AC vital bus 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months buses inoperable. subsystem(s) to OPERABLE status. AND 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months from discovery of failure to meet LCO C. One Auxiliary Building DC C.1 Restore Auxiliary 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months electrical power distribution Building DC electrical subsystem inoperable. power distribution AND subsystem to OPERABLE status. 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months from discovery of failure to This provides plausibility that this meet LCO TS requires an LCO as well as TS 3.8.4 Farley Units 1 and 2 3.8.9-1 Amendment No. 146 (Unit 1)

Amendment No. 137 (Unit 2)

Distribution Systems Operating 3.8.9 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A, B, or AND C not met.

D.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months E. One Service Water Intake E.1 Declare the associated Immediately Structure (SWIS) DC Service Water train electrical power distribution inoperable.

subsystem inoperable.

F. Two trains with inoperable F.1 Enter LCO 3.0.3. Immediately distribution subsystems that result in a loss of safety function.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.9.1 Verify correct breaker alignments and voltage to In accordance with required AC, DC, and AC vital bus electrical power the Surveillance distribution subsystems. Frequency Control Program Farley Units 1 and 2 3.8.9-2 Amendment No. 185 (Unit 1)

Amendment No. 180 (Unit 2)

Distribution Systems Operating B 3.8.9 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.9 Distribution Systems Operating BASES BACKGROUND The onsite Class 1E AC, DC, and AC vital bus electrical power distribution systems are divided into two redundant and independent AC, DC, and AC vital bus electrical power distribution trains.

The AC electrical power subsystem for each train consists of a primary Engineered Safety Feature (ESF) 4.16 kV bus and secondary 600 and 208/120 V buses, distribution panels, motor control centers and load centers. Each train of 4.16 kV ESF buses has at least one separate and independent offsite source of power as well as an onsite diesel generator (DG) source. Each 4.16 kV ESF bus is normally connected to a preferred offsite source. If all offsite sources are unavailable, the onsite emergency DG supplies power to the 4.16 kV ESF bus(es).

Control power for the 4.16 kV breakers is supplied from the Class 1E batteries. Additional description of this system may be found in the Bases for LCO 3.8.1, "AC Sources Operating," and the Bases for LCO 3.8.4, "DC Sources Operating."

The secondary AC electrical power distribution system for each train includes the safety related load centers, motor control centers, and distribution panels shown in Table B 3.8.9-1.

The 120 VAC vital buses are arranged in two load groups per train and are normally powered from the inverters. The alternate power supply for the vital buses are Class 1E constant voltage source transformers powered from the same train as the associated inverter, and its use is governed by LCO 3.8.7, "Inverters Operating." Each constant voltage source transformer is powered from a Class 1E AC bus.

There are two independent 125 VDC electrical power distribution subsystems (one for each train).

The list of all required distribution buses is presented in Table B 3.8.9-1.

APPLICABLE The initial conditions of Design Basis Accident (DBA) and transient SAFETY ANALYSES analyses in the FSAR, Chapter 6 (Ref. 1), and in the FSAR, Chapter 15 (Ref. 2), assume ESF systems are OPERABLE. The AC, (continued)

Farley Units 1 and 2 B 3.8.9-1 Revision 0

Distribution Systems Operating B 3.8.9 BASES APPLICABLE DC, and AC vital bus electrical power distribution systems are SAFETY ANALYSES designed to provide sufficient capacity, capability, redundancy, and (continued) reliability to ensure the availability of necessary power to ESF systems so that the fuel, Reactor Coolant System, and containment design limits are not exceeded. These limits are discussed in more detail in the Bases for Section 3.2, Power Distribution Limits; Section 3.4, Reactor Coolant System (RCS); and Section 3.6, Containment Systems.

The OPERABILITY of the AC, DC, and AC vital bus electrical power distribution systems is consistent with the initial assumptions of the accident analyses and is based upon meeting the design basis of the unit. This includes maintaining power distribution systems OPERABLE during accident conditions in the event of:

a. An assumed loss of all offsite power or all onsite AC electrical power; and

b. A worst case single failure.

The distribution systems satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO The required power distribution subsystems listed in Table B 3.8.9-1 ensure the availability of AC, DC, and AC vital bus electrical power for the systems required to shut down the reactor and maintain it in a safe condition after an anticipated operational occurrence (AOO) or a postulated DBA. The AC, DC, and AC vital bus electrical power distribution subsystems are required to be OPERABLE.

Maintaining the Train A and Train B AC, DC, and AC vital bus electrical power distribution subsystems OPERABLE ensures that the redundancy incorporated into the design of ESF is not defeated. Therefore, a single failure within any system or within the electrical power distribution subsystems will not prevent safe shutdown of the reactor.

OPERABLE AC electrical power distribution subsystems require the This TS is not required associated buses, load centers, motor control centers, and distribution because of this statement. panels to be energized to their proper voltages. OPERABLE DC As long as the DC bus is electrical power distribution subsystems require the associated buses energized, it doesn't to be energized to their proper voltage from either the associated matter if the Charger is available.

(continued)

Farley Units 1 and 2 B 3.8.9-2 Revision 0

Distribution Systems Operating B 3.8.9 BASES LCO battery or charger. OPERABLE vital bus electrical power distribution (continued) subsystems require the associated buses to be energized to their proper voltage from the associated inverter via inverted DC voltage or Class 1E constant voltage transformer.

In addition, tie breakers between redundant safety related AC, DC, and AC vital bus power distribution subsystems, if they exist, must be open.

This prevents any electrical malfunction in any power distribution subsystem from propagating to the redundant subsystem, that could cause the failure of a redundant subsystem and a loss of essential safety function(s). If any tie breakers are closed, the affected redundant electrical power distribution subsystems are considered inoperable. This applies to the onsite, safety related redundant electrical power distribution subsystems. It does not, however, preclude redundant Class 1E 4.16 kV buses from being powered from the same offsite circuit.

APPLICABILITY The electrical power distribution subsystems are required to be OPERABLE in MODES 1, 2, 3, and 4 to ensure that:

a. Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of AOOs or abnormal transients; and

b. Adequate core cooling is provided, and containment OPERABILITY and other vital functions are maintained in the event of a postulated DBA.

Electrical power distribution subsystem requirements for MODES 5 and 6 are covered in the Bases for LCO 3.8.10, "Distribution Systems Shutdown."

ACTIONS A.1 With one or more required AC buses, load centers, motor control centers, or distribution panels, except AC vital buses, inoperable, and a loss of safety function has not yet occurred, the remaining AC electrical power distribution subsystems are capable of supporting the (continued)

Farley Units 1 and 2 B 3.8.9-3 Revision 0

Distribution Systems Operating B 3.8.9 BASES ACTIONS A.1 (continued) minimum safety functions necessary to shut down the reactor and maintain it in a safe shutdown condition, assuming no single failure.

The overall reliability is reduced, however, because a single failure in the remaining power distribution subsystems could result in the minimum required ESF functions not being supported. Therefore, the required AC buses, load centers, motor control centers, and distribution panels must be restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Condition A worst scenario is one train without AC power (i.e., no offsite power to the train and the associated DG inoperable). In this Condition, the unit is more vulnerable to a complete loss of AC power.

It is, therefore, imperative that the unit operator's attention be focused on minimizing the potential for loss of power to the remaining train by stabilizing the unit, and on restoring power to the affected train. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months time limit before requiring a unit shutdown in this Condition is acceptable because of:

a. The potential for decreased safety if the unit operator's attention is diverted from the evaluations and actions necessary to restore power to the affected train, to the actions associated with taking the unit to shutdown within this time limit; and

b. The potential for an event in conjunction with a single failure of a redundant component in the train with AC power.

The second Completion Time for Required Action A.1 establishes a limit on the maximum time allowed for any combination of required distribution subsystems to be inoperable during any single contiguous occurrence of failing to meet the LCO. If Condition A is entered while, for instance, a DC bus is inoperable and subsequently restored OPERABLE, the LCO may already have been not met for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . This could lead to a total of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months , since initial failure of the LCO, to restore the AC distribution system. At this time, a DC circuit could again become inoperable, and AC distribution restored OPERABLE. This could continue indefinitely.

The Completion Time allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." This will result in (continued)

Farley Units 1 and 2 B 3.8.9-4 Revision 0

Distribution Systems Operating B 3.8.9 BASES ACTIONS A.1 (continued) establishing the "time zero" at the time the LCO was initially not met, instead of the time Condition A was entered. The 16 hour1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months Completion Time is an acceptable limitation on this potential to fail to meet the LCO indefinitely.

B.1 With one or more AC vital buses inoperable, and a loss of safety function has not yet occurred, the remaining OPERABLE AC vital buses are capable of supporting the minimum safety functions necessary to shut down the unit and maintain it in the safe shutdown condition. Overall reliability is reduced, however, since an additional single failure could result in the minimum required ESF functions not being supported. Therefore, the required AC vital bus must be restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months by powering the bus from the associated inverter via inverted DC or Class 1E constant voltage transformer.

Condition B represents one or more AC vital buses without power; potentially both the DC source and the associated AC source are nonfunctioning. In this situation, the unit is significantly more vulnerable to a complete loss of all noninterruptible power. It is, therefore, imperative that the operator's attention focus on stabilizing the unit, minimizing the potential for loss of power to the remaining vital buses and restoring power to the affected vital bus.

This 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months limit is more conservative than Completion Times allowed for the vast majority of components that are without adequate vital AC power. Taking exception to LCO 3.0.2 for components without adequate vital AC power, that would have the Required Action Completion Times shorter than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months if declared inoperable, is acceptable because of:

a. The potential for decreased safety by requiring a change in unit conditions (i.e., requiring a shutdown) and not allowing stable operations to continue; (continued)

Farley Units 1 and 2 B 3.8.9-5 Revision 0

Distribution Systems Operating B 3.8.9 BASES ACTIONS B.1 (continued)

b. The potential for decreased safety by requiring entry into numerous Applicable Conditions and Required Actions for components without adequate vital AC power and not providing sufficient time for the operators to perform the necessary evaluations and actions for restoring power to the affected train; and

c. The potential for an event in conjunction with a single failure of a redundant component.

The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time takes into account the importance to safety of restoring the AC vital bus to OPERABLE status, the redundant capability afforded by the other OPERABLE vital buses, and the low probability of a DBA occurring during this period.

The second Completion Time for Required Action B.1 establishes a limit on the maximum time allowed for any combination of required distribution subsystems to be inoperable during any single contiguous occurrence of failing to meet the LCO. If Condition B is entered while, for instance, an AC bus is inoperable and subsequently returned OPERABLE, the LCO may already have been not met for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . This could lead to a total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months , since initial failure of the LCO, to restore the vital bus distribution system. At this time, an AC train could again become inoperable, and vital bus distribution restored OPERABLE. This could continue indefinitely.

This Completion Time allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." This will result in establishing the "time zero" at the time the LCO was initially not met, instead of the time Condition B was entered. The 16 hour1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months Completion Time is an acceptable limitation on this potential to fail to meet the LCO indefinitely.

C.1 With Auxiliary Building DC bus(es) in one train inoperable, the remaining Auxiliary Building DC electrical power distribution subsystems are capable of supporting the minimum safety functions (continued)

Farley Units 1 and 2 B 3.8.9-6 Revision 0

Distribution Systems Operating B 3.8.9 BASES ACTIONS C.1 (continued) necessary to shut down the reactor and maintain it in a safe shutdown condition, assuming no single failure. The overall reliability is reduced, however, because a single failure in the remaining DC electrical power distribution subsystem could result in the minimum required ESF functions not being supported. Therefore, the required DC buses must be restored to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months by powering the bus from the associated battery or charger.

Condition C represents one train without adequate DC power; potentially both with the battery significantly degraded and the associated charger nonfunctioning. In this situation, the unit is significantly more vulnerable to a complete loss of all DC power. It is, therefore, imperative that the operator's attention focus on stabilizing the unit, minimizing the potential for loss of power to the remaining trains and restoring power to the affected train.

This 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months limit is more conservative than Completion Times allowed for the vast majority of components that would be without power.

Taking exception to LCO 3.0.2 for components without adequate DC power, which would have Required Action Completion Times shorter than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , is acceptable because of:

a. The potential for decreased safety by requiring a change in unit conditions (i.e., requiring a shutdown) while allowing stable operations to continue;

b. The potential for decreased safety by requiring entry into numerous applicable Conditions and Required Actions for components without DC power and not providing sufficient time for the operators to perform the necessary evaluations and actions for restoring power to the affected train; and

c. The potential for an event in conjunction with a single failure of a redundant component.

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time for DC buses is consistent with Regulatory Guide 1.93 (Ref. 3).

(continued)

Farley Units 1 and 2 B 3.8.9-7 Revision 0

Distribution Systems Operating B 3.8.9 BASES ACTIONS C.1 (continued)

The second Completion Time for Required Action C.1 establishes a limit on the maximum time allowed for any combination of required distribution subsystems to be inoperable during any single contiguous occurrence of failing to meet the LCO. If Condition C is entered while, for instance, an AC bus is inoperable and subsequently returned OPERABLE, the LCO may already have been not met for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . This could lead to a total of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months , since initial failure of the LCO, to restore the DC distribution system. At this time, an AC train could again become inoperable, and DC distribution restored OPERABLE. This could continue indefinitely.

This Completion Time allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." This will result in establishing the "time zero" at the time the LCO was initially not met, instead of the time Condition C was entered. The 16 hour1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months Completion Time is an acceptable limitation on this potential to fail to meet the LCO indefinitely.

D.1 and D.2 If the inoperable distribution subsystem(s) addressed by Conditions A, B, or C cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging plant systems.

E.1 With one SWIS DC electrical power distribution subsystem inoperable, the Service Water System train supported by the affected SWIS DC electrical power distribution subsystem must be declared inoperable.

The capability of the affected SWIS DC electrical power distribution subsystem to fully support the associated train of Service Water is not assured. Therefore, consistent with the definition of OPERABILITY, the associated train of Service Water must be declared inoperable immediately, thereby limiting operation in this condition to the Completion Time associated with the affected Service Water System train.

(continued)

Farley Units 1 and 2 B 3.8.9-8 Revision 0

Distribution Systems Operating B 3.8.9 BASES ACTIONS F.1 (continued)

With two trains with inoperable distribution subsystems that result in a loss of safety function, adequate core cooling, containment OPERABILITY and other vital functions for DBA mitigation would be compromised, and immediate plant shutdown in accordance with LCO 3.0.3 is required.

SURVEILLANCE SR 3.8.9.1 REQUIREMENTS This Surveillance verifies that the required AC, DC, and AC vital bus electrical power distribution systems are functioning properly, with the correct circuit breaker alignment. The correct breaker alignment ensures the appropriate separation and independence of the electrical divisions is maintained, and the appropriate voltage is available to each required bus. The verification of proper voltage availability on the buses ensures that the required voltage is readily available for motive as well as control functions for critical system loads connected to these buses. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. FSAR, Chapter 6.

2. FSAR, Chapter 15.

3. Regulatory Guide 1.93, December 1974.

Farley Units 1 and 2 B 3.8.9-9 Revision 52

Distribution Systems Operating B 3.8.9 Table B 3.8.9-1 (page 1 of 1)

AC and DC Electrical Power Distribution Systems TYPE VOLTAGE TRAIN A* TRAIN B*

AC Safety 4160 V SWGR 1/2 F and 1/2 K 1/2 G and 1/2 L Buses 600 V LC 1/2 D, K**, and R** 1/2 E, L**, and S**

DC Buses 125 V SWGR 1/2 A 1/2 B 125 V Dist. Panels 1/2 M 1/2 N Vital AC Buses 120 1/2 A and 1/2 B 1/2 C and 1/2 D

Each train of the AC and DC electrical power distribution systems is a subsystem.

- Indicates buses shared between Units 1 and 2.

Farley Units 1 and 2 B 3.8.9-10 Revision 0

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

87. 059A2.12 087 Unit 1 tripped from 100% power with the following conditions:

A Safety Injection occurred due to a Steam Dump malfunction.

Status of AFW pumps is as follows:

- 1A MDAFW pump is Tagged Out.

- 1B MDAFW pump is tripped.

- The TDAFW pump trip/throttle linkage is broken.

The Safety Injection signal has been reset.

FRP-H.1, Response to Loss of Secondary Heat Sink, is in progress with the crew attempting to restore SG level using a Condensate pump.

Which one of the following completes the statements below?

Implementation of FRP-H.1, Attachment 1, Main Feedwater Bypass Valves Automatic Closure Defeat, (1) required to open the Main Feedwater Bypass valves.

Criteria to exit FRP-H.1 is met (2) .

(1) (2)

A. IS ONLY after at least one SG NR level is > 31%

B. is NOT ONLY after at least one SG NR level is > 31%

C. IS when stable Condensate flow is established and SG wide range level is rising D. is NOT when stable Condensate flow is established and SG wide range level is rising Tuesday, March 19, 2013 11:38:43 AM 41

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 FRP-H.1 is in progress with no AFW pumps available. MSIV closure has occurred due to a Steam Line low pressure MSIV closure signal. The normal course of FRP-H.1 is to attempt to establish condensate flow using a Condensate pump. These actions will continue until either a) Condensate flow is established and SG level is rising or B)

Bleed and Feed criteria is met.

A. Incorrect - 1) Correct, Attachment 1 is required to be implemented due to the SI signal in progress. Feed Reg Bypass valves will not open unless this attachment is implemented..

2) Incorrect, plausible since step 10.2 RNO of FRP-H.1 states to continue feeding SGs to restore at least one SG narrow range level to greater than 31%, it gives the impression that 31% level is required to exit the procedure.

B. Incorrect - 1) Incorrect, plausible since SI has been reset. The candidate may think that resetting SI will allow the operator to take control of Feed Reg Bypass valves, thus implementation of Attachment 1 is not required. In addition, if there was no SI, this could be a correct answer. At one time Attachment 1 was only implemented if an SI had occurred.

2) Incorrect, see A.2.

C. Correct - 1) Correct, see A.1.

2) Correct, per step 10.1 -10.3 of FRP-H.1 and the associated note, as long as stable feed flow has been established and SG wide range level is rising, the heat sink critical safety function is satisfied and FRP-H.1 can be exited.

D. Incorrect - 1) Incorrect, see B.2.

2) Correct, see C.2.

K/A: 059A2.12 Main Feedwater (MFW) System Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

(CFR: 41.5 / 43.5 / 45.3 / 45.13)

A2.12 Failure of feedwater regulating valves . . . . . 3.1* 3.4*

Importance Rating: 3.1 3.4 Tuesday, March 19, 2013 11:38:43 AM 42

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 Technical

Reference:

OPS Lesson Plan - Reactor Protection, Fig. 17 OPS-62201I/52201I/40302F/ESP-52201, v2.0 FNP-1-FRP-H.1, v27.0 References provided: None Learning Objective: ASSESS the facility conditions associated with the (1)

FRP-H.1, Response to Loss of Secondary Heat Sink; (2)

FRP-H.2, Response to SG Overpressure; (3) FRP-H.3, Response to SG High Level; (4) FRP-H.4, Response to Loss of Normal Steam Release Capabilities; (5) FRP-H.5, Response to SG Low Level, and based on that assessment:

(OPS-62533F01)

SELECT the appropriate procedures during normal, abnormal and emergency situations. 10CFR55.43 (b) 5

DETERMINE if transition to another section of the procedure or to another procedure is required

DETERMINE if the critical safety functions are satisfied Question History: Modified from September 2009 Sequoyah NRC Exam Retake #89 K/A match: For the conditions given in FRP-H.1, Feed Reg valves and bypass valves are failed closed due to the SI signal. They are required to be opened to be able to mitigate the accident and feed the SG's from Condensate flow. Candidate has to determine if any additional actions are required to open the valves (implementation of Attachment 1) after the SI signal has occurred and been reset. In addition, candidate has to display knowledge of how to use FRP-H.1 and when the procedure can be exited.

SRO justification: 10 CFR 55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations [10 CFR 55.43(b)(5)]

One area of SRO level knowledge is knowledge of content of the procedure vs. the procedure's overall mitigative strategy or purpose. The applicants knowledge can be evaluated at the level of 10 CFR55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:

Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific subprocedures or emergency contingency procedures.

Tuesday, March 19, 2013 11:38:43 AM 43

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 2013 NRC exam From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 flowchart for 10 CFR 55.43(b)(5) :

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations [10 CFR 55.43(b)(5)], involving BOTH:

1) assessing plant conditions and then

2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

Using the flowchart, this question can:

NOT be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location.

NOT be answered solely by knowing immediate operator actions.

NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs.

NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

CAN be answered with knowledge of ONE or MORE of the following:

Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps. (the first part of the question relates to this, but is not clearly SRO)

Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific subprocedures or emergency contingency procedures (the second part of the question relates to this and is clearly SRO)

Tuesday, March 19, 2013 11:38:43 AM 44

REACTOR PROTECTION OpsSsp009 T

AVG TB I TB II TB III 412 422 432 D D D 1 2 1 2 1 2 LOW T HIGH T AVG AVG 2 /

3 P-4 REACTOR TRIP A MCB ALARM With a Reactor trip and SI signal generated, FRV's cannot be opened until the seal in signal is broken by removing the P-4 signal HI-HI SG (cycling Reactor Trip SI WTR LVL breakers). This is why jumpers have to be installed per Attachment 1 of FRP-H.1. MANUAL RESET R L FEEDWATER ISOLATION FIGURE 17 - Tavg Protection (Low Tavg Feedwater Isolation, High Tavg Alarm)

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Modified this question from 2009 Sequoyah NRC Exam to develop 059A2.12 question QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

88. 064G2.2.44 088 Unit 1 is at 100% power when the following occurs:

WE1, 1F, 1H, OR 1K 4KV BUS BKR AUTO TRIP, is in alarm.

WE3, 1H 4KV BUS UV OR LOSS OF DC, is in alarm.

The following indications are noted on the handswitch for 4160V Breaker DF-13-1, SUPPLY TO 1H 4160V BUS:

- The amber light is lit.

- The green light is lit.

The AC PWR AVAIL lights for 1F 4160V bus are illuminated.

The AC PWR AVAIL lights for 1H 4160V bus are NOT illuminated.

Which one of the following completes the statements below?

The 1C DG (1) autostart due to the DF-13-1 malfunction.

For Unit 1, a REQUIRED ACTION statement of Tech Spec 3.8.1, AC Sources - Operating, (2) required to be implemented.

(1) (2)

A. WILL is NOT B. WILL IS C. will NOT is NOT D. will NOT IS This is a difficult question due to the complexity of the DG autostart design and operation.

For the first part of the question: There is one dedicated 4075kw DG available for B Train on each Unit (1B DG and 2B DG). Each DG will tie to the G 4160V bus of its respective Unit. Their operation is simple and clearly defined.

For the A Train, there is one 2850kw DG (1C DG) and one 4075kw DG (1-2A DG), both of which are shared between the Units. They can each go to Unit 1 or Unit 2 based on the conditions on each Unit, i.e. LOSP, SI with LOSP, which Unit had the LOSP first etc. The 1C DG will go to the opposite Unit from the 1-2A DG. When powering the busses, they can even divorce from one Unit and tie to the other Unit based on conditions. The 1C DG will tie to U-1 or U-2 H 4160V bus, which is in turn tied to the F 4160V bus by breaker DF-13. The 1-2A DG will tie to U-1 or U-2 F 4160V bus. The intent is to ensure the F 4160V bus is energized to supply safety related loads.

The autostart for the DG's can be even more complicated since the DG operation is different now than from the original design of the plant . For the A Train DGs' autostart Tuesday, March 19, 2013 11:38:43 AM 45

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 for a Unit 1 LOSP, the 1-2A DG gets an autostart signal from the UV relays located on the B1F Sequencer. The 1C DG will autostart from the UV relays located on the B1H Sequencer. It is a common misconception that both A Train DG's start from an UV signal on the B1F Sequencer. This is because the B1H Sequencer has very few functions any more (original design was to power River Water pumps that used to be safety related, but are no longer safety related and are not powered up during an LOSP). Once the B1H Sequencer starts the 1C DG, the B1F Sequencer tells 1C DG where to go and sequences the starting of loads. Essentially B1F Sequencer takes over all other functions. As a result, the misconception is that the B1F Sequencer performs all actions associated with starting both DG's, deciding which Unit each DG ties to, and sequencing all loads.

For this question, the 1C DG will start due to the LOSP sensed by the B1H Sequencer, 1-2A DG will not start because there is no LOSP sensed by the B1F Sequencer. The 1C DG will start and run unloaded since the B1F Sequencer has not told it to do anything, and the 1H 4160V bus will remain de-energized.

For the second part of the question: this is also difficult to evaluate. The 1H bus is no longer considered safety related, but it is required as a support system for operation of the A Train DG set. The 1C DG output breaker will tie to the 1H bus, which has to be able to be tied to the 1F 4160V bus through breaker DF-13-1, to power up safety related loads. The plausibility for no Tech Spec entry required is due to the fact that 1H bus is no longer safety related, but is a support system for operability of the 1C DG.

Tech Spec 3.8.1 Bases knowledge or Surveillance test (STP-27.2) knowledge is required to be able to evaluate and answer this question correctly.

A. Incorrect - 1) Correct, 1C DG will autostart. See detailed information above.

2) Incorrect, plausible since the 1H 4160V bus is no longer safety related and not required, except for the fact that it is a support system for the 1C DG. In addition, a candidate may incorrectly determine that the more appropriate LCO would be Electrical Power Systems TS 3.8.9, Distribution Systems - Operating. Since there is no cascading of TS, there would be no LCO for TS 3.8.1.

B. Correct - 1) Correct, see A.1.

2) Correct, a REQUIRED ACTION statement would be implemented since the A Train DG set (composed of 1-2A DG and the 1C DG) is INOPERABLE due to a loss of a support system to the 1C DG.

C. Incorrect - 1) Incorrect, plausible because of the common misconception that the DGs start based on a LOSP on the F 4160V bus and not the H 4160V bus. See detailed information above.

2) Incorrect, see A.2.

D. Incorrect - 1) Incorrect, see C.1.

Tuesday, March 19, 2013 11:38:43 AM 46

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

2) Correct, see B.2.

K/A: 064G2.2.44 Emergency Diesel Generator (ED/G) System Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

l (CFR: 41.5 / 43.5 / 45.12)

IMPORTANCE RO 4.2 SRO 4.4 Importance Rating: 4.2 4.4 Technical

Reference:

OPS Lesson Plans - Diesel Generators and Auxiliaries, v1 Figure 17 - OPS-62102I/52102I/40102C/ESP-52102I OPS Lesson Plans - Diesel Generator Sequencers, v3 Figure 10 - 62103F/52103F/40102D/ESP-52103F FNP-1-STP-27.2 v26 Tech Specs v190 Tech Spec Bases v58 References provided: None Learning Objective: RECALL AND APPLY the information from the LCO BASES sections: BACKGROUND, APPLICABLE SAFETY ANALYSIS, ACTIONS, SURVEILLENCE REQUIREMENTS, for any Technical Specifications or TRM requirements associated with the Diesel Generator and Auxiliaries System components and attendant equipment alignment, to include the following (OPS-62102I01): 10CFR55.43 (b) 2 3.8.1, AC Sources - Operating 3.8.2, AC Sources - Shutdown 3.8.3, Diesel Fuel Oil, Lube Oil, Starting Air Question History: New question K/A match: This question requires the candidate to interpret Control Room indications for a tripped breaker supplying the 1H 4160V bus. An evaluation has to be performed to understand how that affects the operation of 1C DG, and then when TS are evaluated, understand how that affects the operability of the DG system.

SRO justification: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

Knowledge of TS bases that is required to analyze TS required actions and terminology.

Tuesday, March 19, 2013 11:38:43 AM 47

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).

Tech Spec Bases knowledge is required to understand that although the 1H 4160V bus is not safety related and not required by TS, it is a support system for the 1C DG and is required for operability of the 1C DG.

In addition, STP-27.2 provides guidance that the 1H 4160V bus is a support system for the 1C DG.

2013 NRC exam From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 flowchart for 10 CFR 55.43(b)(2)

1) can NOT be answered by knowing less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Tech Specs.

2) can NOT be answered by knowing information listed "above-the-line".

3) can NOT be answered by knowing the TS Safety Limits or their bases.

4) Does involve one or more of the following for TS, TRM or ODCM:

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

Knowledge of TS bases that is required to analyze TS required actions and terminology Tuesday, March 19, 2013 11:38:43 AM 48

DIESEL GENERATORS AND AUXILIARIES OpsDgs027 1C DG autostarts DIESEL BUS UV on a UV signal DIESEL BUS UV 1-2A 1F from the 1H bus 1-2A 2F 1C 1H 1C 2H 2-3 1-2 3-1 2-3 1-2 3-1 SWING DIESELS SI UNIT SELECTOR SWITCH SI UNIT 1 UNIT 2 1 1/2 2 2-SEC. TDPU 2-SEC. TDPU AUTOSTART DEDICATED DIESELS SIGNAL (SEE FIG. 18)

DIESEL BUS UV 1B 1G 2B 2G 2-3 1-2 3-1 UNIT 1 UNIT 2 SI 2-SEC. TDPU AUTOSTART SIGNAL (SEE FIG. 18)

FIGURE 17 - Emergency Diesel-Generator Auto-Start Logic OPS-62102I/52102I/40102C/ESP-52102I - Version 1

DIESEL GENERATOR SEQUENCERS RUN RUN RUN OpsDgs039 1-2A If breaker 2B 1B RUN DF-13 is 1C 2C tripped, 1C DG cannot supply the 1F bus if required.

DG08 DG08 DHO7 DF13 DG13 DJ06 DHO7 DF13 DG13 DJ06 DF08 DF08 2H 2K 2F 2G 2L 2J 1H 1K 1F 1G 1L 1J DF02 DG02 DF02 DG02 TRAIN TRAIN TRAIN TRAIN A B A B UNIT UNIT 2 1 NOTES DGs 1-2A, 1B, 2B, AND 1C START. DG 1-2A ALIGNS TO BUS 2F, AND THE B2F ESS SEQUENCER

RUNS. BUS 2H REMAINS DEENERGIZED UNTIL MANUALLY ENERGIZED BY THE OPERATOR. DG 1C ALIGNS TO BUS 1F THROUGH DH07-1 AND DF13-1. THE B1F LOSP SEQUENCER RUNS. DG 1B ALIGNS TO BUS 1G, AND THE B1G LOSP SEQUENCER RUNS. DG 2B ALIGNS TO BUS 2G, AND THE B2G ESS SEQUENCER RUNS.

FIGURE 10 - Dual-Unit LOSP, SI Unit 2, No Failures OPS-62103F/52103F/40102D/ESP-52103F- Version 3

10/10/11 13:06:42 FNP-1-STP-27.2 UNIT 1 2.3 Administrative Control The following AC electrical buses and load centers are not included in Technical Specifications. There are no acceptance criteria associated with these buses and load centers. Verification of availability of these buses and load centers is included to maintain administrative control only.

2.3.1 4160 volt Buses H and J (3850-4220V) 2.3.2 600 volt Load Centers A, C, 1-2H, and 1-2J (550-605V) 3.0 Initial Conditions 3.1 The revision of this procedure has been verified to be the current revision.

(OR 1-98-498) 3.2 This procedure has been verified to be the correct unit for the task. (OR 1-98-498) 3.3 The electrical distribution system is energized for normal operation per FNP-1-SOP-36.0, PLANT ELECTRICAL DISTRIBUTION LINE-UP, with exceptions noted.

4.0 Precautions and Limitations 4.1 To prevent an inadvertent de-energization of the electrical buses , permission must be obtained from the Shift Supervisor prior to operating any circuit breaker.

If the 1H bus is unavailable, the 4.2 1H and 1J 4160V buses are required support equipment for the 1C and 2C diesel 1C DG cannot generators. Any operability issues regarding the 1H or 1J 4160V buses should be perform its evaluated against the requirements for the 1C or 2C diesel generator, as safety function appropriate.

to energize the 1F and 1K 4.3 1F 600 V LC will not receive a load shed signal when aligned to supply 1D (1E) 4160V busses. 600 V LC. This signal to DF12 (DG12) is blocked when breaker ED12 (EE12) is racked in and closed. If 1F 600 V LC is aligned to supply a 600 V LC other than 1D or 1E, the load shed signal to 1F 600 V LC will not be blocked.

4.4 When locally verifying a breaker position, the following items should also be checked:

Proper light bulb illuminated Spring charge indication No obvious physical damage to breaker Page 2 of 7 Version 26.0

AC Sources Operating 3.8.1 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources Operating LCO 3.8.1 The following AC electrical sources shall be OPERABLE:

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

b. Two diesel generator (DG) sets capable of supplying the onsite Class 1E power distribution subsystem(s); and

c. Automatic load sequencers for Train A and Train B.

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

ACTIONS

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

LCO 3.0.4b is not applicable to DGs.

CONDITION REQUIRED ACTION COMPLETION TIME A. One required offsite circuit A.1 Perform SR 3.8.1.1 for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months inoperable. required OPERABLE offsite circuit. AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND A.2 Declare required 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from feature(s) with no discovery of no offsite offsite power available power to one train inoperable when its concurrent with redundant required inoperability of feature(s) is inoperable. redundant required feature(s)

AND (continued)

Farley Units 1 and 2 3.8.1-1 Amendment No. 170 (Unit 1)

Amendment No. 163 (Unit 2)

AC Sources Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3 Restore required offsite 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months circuit to OPERABLE status. AND 13 days from discovery of failure to meet LCO B. One DG set inoperable. ------------------NOTE-------------------

LCO 3.0.4c is applicable when only one of the three DGs is inoperable.

The "A" Train DG set consists B.1 Perform SR 3.8.1.1 for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the 1C DG and the 1-2A DG. the required offsite The "B" Train DG set consists circuit(s). AND of the 1B DG.

Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND B.2 Declare required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from feature(s) supported by discovery of the inoperable DG set Condition B inoperable when its concurrent with required redundant inoperability of feature(s) is inoperable. redundant required feature(s)

AND B.3.1 Determine OPERABLE 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months DG set is not inoperable due to common cause failure.

OR (continued)

Farley Units 1 and 2 3.8.1-2 Amendment No. 170 (Unit 1)

Amendment No. 163 (Unit 2)

AC Sources Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.3.2 Perform SR 3.8.1.6 for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE DG set.

AND B.4 Restore DG set to 10 days OPERABLE status.

AND 13 days from discovery of failure to meet LCO C. Two required offsite C.1 Declare required 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months from circuits inoperable. feature(s) inoperable discovery of when its redundant Condition C required feature(s) is concurrent with inoperable. inoperability of redundant required features AND C.2 Restore one required 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months offsite circuit to OPERABLE status.

Farley Units 1 and 2 3.8.1-3 Amendment No. 146 (Unit 1)

Amendment No. 137 (Unit 2)

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

89. 071G2.2.25 089 Which one of the following describes TRM 13.12.4, Gas Storage Tanks, limitations on the quantity of radioactivity permitted in the Waste Gas Decay Tank(s) and the basis for the limit?

The quantity contained in each Waste Gas Decay Tank shall be < (1) to ensure an uncontrolled release will not exceed a whole body dose of 0.5 REM to (2) .

A. 1) 10 curies, excluding tritium and dissolved or entrained noble gases

2) an operator at the Waste Gas Control Panel B. 1) 70,500 curies of noble gases (considered as Xe-133)

2) an operator at the Waste Gas Control Panel C. 1) 10 curies, excluding tritium and dissolved or entrained noble gases

2) an individual at the exclusion area boundary D. 1) 70,500 curies of noble gases (considered as Xe-133)

2) an individual at the exclusion area boundary Tuesday, March 19, 2013 11:38:43 AM 49

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 TR 13.12.2 The quantity of radioactive material contained in any outside temporary tank, excluding liners being used to solidify radioactive wastes, shall be limited to < 10 curies, excluding tritium and dissolved or entrained noble gases.

TR 13.12.4 Gas Storage Tanks TR 13.12.4 The quantity of radioactivity contained in each gas storage tank shall be limited to < 70,500 curies of noble gases (considered as Xe-133).

A. Incorrect - 1) Incorrect, plausible since the limit for TR 13.12.2 is < 10 curies. The candidate could confuse this curie limit with the correct one.

2) Incorrect, plausible since minimizing dose to operators in the field is a concern. If a WGDT were to rupture, an operator in the field at the Waste Gas Control Panel could receive significant dose.

B. Incorrect - 1) Correct, this is the TRM limit for curie content of a WGDT.

2) Incorrect, see A.2.

C. Incorrect - 1) Incorrect, see A.1.

2) Correct, this is correct as stated in the Bases of TRM 13.12.4. This is the dose limit to protect an individual at the exclusion area boundary.

D. Correct - 1) Correct, see B.1.

2) Correct, see C.2.

K/A: 071G2.2.25 Waste Gas Disposal System (WGDS)

Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

(CFR: 41.5 / 41.7 / 43.2)

IMPORTANCE RO 3.2 SRO 4.2 Importance Rating: 3.2 4.2 Technical

Reference:

Technical Requirements Manual v24 Technical Requirements Manual Bases v9 References provided: None Learning Objective: RECALL AND APPLY the information from the LCO BASES sections: BACKGROUND, APPLICABLE SAFETY ANALYSIS, ACTIONS, SURVEILLENCE REQUIREMENTS, Tuesday, March 19, 2013 11:38:43 AM 50

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 for any Technical Specifications or TRM requirements associated with Waste Gas System components and attendant equipment alignment, to include the following (OPS-62106B01): 10CFR55.43 (b) 2 13.12.1, Waste Gas Monitoring Instrumentation 13.12.3, Waste Gas Monitoring 13.12.4, Gas Storage Tanks Question History: New question K/A match: This question tests the knowledge of an SRO on the information contained in the bases of TRM 13.12.4.

SRO justification: Facility operating limitations in the TS and their bases.

[10 CFR 55.43(b)(2)]

Knowledge of TS bases that is required to analyze TS required actions and terminology.

Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).

Knowledge of the TRM Bases is required to determine the maximum dose to a member of the public during a WGDT rupture.

2013 NRC exam From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 flowchart for 10 CFR 55.43(b)(2) :

1) can NOT be answered by knowing less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Tech Specs.

2) can NOT be answered by knowing information listed "above-the-line".

3) can NOT be answered by knowing the TS Safety Limits or their bases.

4) Does involve one or more of the following for TS, TRM or ODCM:

Knowledge of TS bases that is required to analyze TS required actions and terminology.

Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).

Tuesday, March 19, 2013 11:38:43 AM 51

Gas Storage Tanks TR 13.12.4 13.12 Explosive Gas and Storage Tank Radioactivity Monitoring (EGSTRAM) Program TR 13.12.4 Gas Storage Tanks TR 13.12.4 The quantity of radioactivity contained in each gas storage tank shall be limited to 70,500 curies of noble gases (considered as Xe-133).

APPLICABILITY: At all times.

ACTIONS

NOTES----------------------------------------------------------

1. TR 13.0.3 is not applicable.

2. TR 13.0.4c is applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A.1 Quantity of radioactive A.1 Suspend all additions of Immediately material in any gas radioactive material to the storage tank exceeding tank.

limit.

AND A.2 Reduce the tank contents to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months within the limit.

TECHNICAL REQUIREMENT SURVEILLANCES SURVEILLANCE FREQUENCY TRS 13.12.4.1 Verify quantity of radioactive material contained in Once per 7 days each gas storage tank to be less than the limit. when radioactive materials have been added to the tank during the previous 7 days.

Farley Units 1 and 2 13.12.4 - 1 Version 13.0 Technical Requirements

Gas Storage Tanks TR 13.12.4 TECHNICAL REQUIREMENT SURVEILLANCES SURVEILLANCE FREQUENCY TRS 13.12.4.2 In the event of confirmed major fuel failure (> 1%), Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months verify quantity of radioactive material contained in when radioactive each waste gas storage tank to be less than the limit. materials have been added to the tank during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Farley Units 1 and 2 13.12.4 - 2 Version 8.0 Technical Requirements

EGSTRAM Program B 13.12 B 13.12 EXPLOSIVE GAS AND STORAGE TANK RADIOACTIVITY MONITORING (EGSTRAM) PROGRAM BASES TR 13.12.1 Waste Gas Monitoring Instrumentation This instrumentation monitors (and controls) the concentrations of potentially explosive gas mixtures in the waste gas holdup system. The FUNCTIONALITY and use of this instrumentation are consistent with the requirements of General Design Criteria 60 and 63 of Appendix A to 10 CFR Part 50.

TR 13.12.2 Liquid Holdup Tanks Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, (to paragraphs 20.1001 -20.2401), Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area.

TR 13.12.3 Waste Gas Monitoring This Technical Requirement is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas holdup system is maintained below the flammability limits of hydrogen and oxygen. During recombiner operation, an automatic control feature is included in the system to prevent the oxygen concentration from reaching these flammability limits. The automatic control feature includes isolation of the source of oxygen (the recombiner oxygen supply), to reduce the concentration below the flammability limit. When the recombiner is not operating and thus the recombiner oxygen supply is isolated, a grab sample can be taken to measure oxygen levels in the waste gas system. Maintaining the concentration of oxygen below the flammability limit when hydrogen is above 4% by volume provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

TR 13.12.4 Gas Storage Tanks Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tanks contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem. This is consistent with Standard Review Plan 15.7.1, "Waste Gas System Failure".

Farley Units 1 and 2 B 13.12-1 Version 9.0 Technical Requirements Bases

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

90. 073A2.01 090 Unit 1 is at 100% power with the following conditions:

#4 Waste Gas Decay Tank (WGDT) release is in progress.

CTMT Mini-purge supply and exhaust fans are running.

The Radwaste Ventilation system is running.

Subsequently, the following occurs:

R-14, PLANT VENT, radiation monitor is oscillating erratically.

The HIGH ALARM and LOW ALARM lights are LIT.

The Shift Supervisor has declared R-14 INOPERABLE.

Which one of the following completes the statements below?

The WGDT release (1) terminated.

Per the ODCM, (2) is(are) Applicable.

REFERENCE PROVIDED (1) (2)

A. IS ACTION 35 ONLY B. IS ACTION 35 and ACTION 37 C. is NOT ACTION 35 ONLY D. is NOT ACTION 35 and ACTION 37 Per the ODCM, Table 3-1, Condition 2a, either R-14 or R-22 has to be operable for a Continuous release from the Plant Vent Stack. That is met by R-22 being operable, so Action 37 is not required -

ACTION 37 - 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 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Per the ODCM, Table 3-1, Condition 3 is not met due to R-14 being out of service, so Action 35 is required at all times -

ACTION 35 - 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 tank's contents are analyzed, and

b. At least two technically qualified members of the Facility Staff independently verify the discharge line valving, and (1) Verify the manual portion of the computer input for the release rate calculations performed on the computer, or (2) Verify the entire release rate calculations if such calculations are performed Tuesday, March 19, 2013 11:38:43 AM 52

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 manually.

Otherwise, suspend release of radioactive effluents via this pathway.

This question targets the actions required per the ODCM for a normal release of Radwaste Vent, Ctmt Minipurge, and Waste Gas Decay Tanks with a failed radiation monitor, along with systems knowledge of the isolation of a WGDT.

A. Correct - 1) Correct, R-14 going into alarm whether the monitor is failed or erratic will cause RCV-14 to close and the release to be terminated.

2) Correct, ACTION 35 is applicable at all times, but is only required when releasing a WGDT. The candidate has to determine that either R-14 or R-22 will fulfill the requirements of the Plant Vent Stack Noble Gas Activity Monitor for evaluation of the requirement to implement ACTION

37. It is also applicable at all times, but either monitor operable will meet the requirement. R-22 is operable.

B. Incorrect- 1) Correct, see A.1.

2) Incorrect, but plausible if the applicant were to misread Table 3-1 and either miss or not properly apply the "Minimum Channels OPERABLE" column properly, this would be selected. Either R-14 or R-22 will fulfill the requirements of the Plant Vent Stack Noble Gas Activity Monitor for evaluation of the requirement to implement ACTION 37.

C. Incorrect- 1) Incorrect, the release will terminate. Plausible since the R-14 is reading erratically and both the high and low lights are LIT. The LOW alarm light will cause a MCB alarm and not isolate RCV-14, but the HIGH alarm light being lit should trigger a MCB alarm and an auto-isolation of RCV-14. The applicant may not understand how this condition affects the automatic functions of the rad monitor, which is to close RCV-14.

2) Correct, see A.2.

D. Incorrect- 1) Incorrect, see C.1.

2) Incorrect, see B.2.

K/A: 073A2.01 Process Radiation Monitoring (PRM) System Ability to (a) predict the impacts of the following malfunctions or operations on the PRM system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

(CFR: 41.5 / 43.5 / 45.3 / 45.13)

A2.01 Erratic or failed power supply........................ 2.5 2.9*

Tuesday, March 19, 2013 11:38:43 AM 53

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 Importance Rating: 2.5 2.9*

Technical

Reference:

FNP-ODCM, v24 References provided: FNP-ODCM , Table 3-1 (page 3-3)

Learning Objective: RECALL AND APPLY the information from the LCO BASES sections: BACKGROUND, APPLICABLE SAFETY ANALYSIS, ACTIONS, SURVEILLENCE REQUIREMENTS, for any Technical Specifications or TRM requirements associated with Waste Gas System components and attendant equipment alignment, to include the following (OPS-62106B01): 10CFR55.43 (b) 2 13.12.1, Waste Gas Monitoring Instrumentation 13.12.3, Waste Gas Monitoring 13.12.4, Gas Storage Tanks Question History: New question Basis for meeting K/A: R-14, a Process Radiation Monitor, has failed and is INOPERABLE. The RO portion of this question asks if the release is terminated which is the "predict the impacts of the erratic rad monitor" KA match. The SRO must display knowledge of how to read and apply the ODCM to ensure the correct actions are applied. This question provides a scenario in which a normal release will be in progress with a failed rad monitor and the actions, if any, required.

SRO justification: 10 CFR 55.43(b)(2):

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

Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).

Requires application of REQUIRED ACTIONS of the ODCM.

2013 NRC exam From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 flowchart for 10 CFR 55.43(b)(2)

1) can NOT be answered by knowing less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Tech Specs.

2) can NOT be answered by knowing information listed "above-the-line".

3) can NOT be answered by knowing the TS Safety Limits or their bases.

4) Does involve one or more of the following for TS, TRM or ODCM:

Application of Required Actions (Section 3) and Surveillance Requirements (SR)

(Section 4) in accordance with rules of application requirements (Section 1)..

Tuesday, March 19, 2013 11:38:43 AM 54

Reference provided to candidate for this question - ODCM Table 3-1 FNP-ODCM Table 3-1 Radioactive Gaseous Effluent Monitoring Instrumentation OPERABILITY Requirementsb Minimum Channels Instrument OPERABLE Applicability ACTION

1. Steam Jet Air Ejector Noble Gas Activity Monitor (RE-15) 1 MODES 1,2,3,4 37

2. Plant Vent Stack

a. Noble Gas Activity Monitor (RE-14 or RE-22) 1 At all times 37a

b. Iodine Sampler 1 At all times 39 c Particulate Sampler 1 At all times 39

d. Flowrate Monitor 1 At all times 36

3. GASEOUS RADWASTE TREATMENT SYSTEM Noble Gas Activity Monitor (RE-14), with Alarm and Automatic Termination of Release 1 At all times 35

a. For continuous releases.

b. All requirements in this table apply to each unit.

3-3 Version 24 01/10

WASTE GAS FIGURE 2 - Waste Gas System OPS-62106B/52106B/40303B Ver 2

FNP-ODCM Table 3-1 Radioactive Gaseous Effluent Monitoring Instrumentation OPERABILITY Requirementsb Minimum Channels Instrument OPERABLE Applicability ACTION

1. Steam Jet Air Ejector Noble Gas Activity Monitor (RE-15) 1 MODES 1,2,3,4 37

2. Plant Vent Stack

a. Noble Gas Activity Monitor (RE-14 or RE-22) 1 At all times 37a

b. Iodine Sampler 1 At all times 39 c Particulate Sampler 1 At all times 39

d. Flowrate Monitor 1 At all times 36

3. GASEOUS RADWASTE TREATMENT SYSTEM Noble Gas Activity Monitor (RE-14), with Alarm and Automatic Termination of Release 1 At all times 35

a. For continuous releases.

b. All requirements in this table apply to each unit.

3-3 Version 24 01/10

FNP-ODCM Table 3-1 (contd) Notation for Table 3 ACTION Statements ACTION 35 - 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 tank's contents are analyzed, and

b. At least two technically qualified members of the Facility Staff independently verify the discharge line valving, and (1) Verify the manual portion of the computer input for the release rate calculations performed on the computer, or (2) Verify the entire release rate calculations if such calculations are performed manually.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 36 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flowrate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 37 - 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 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 39 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue provided samples are continuously collected with auxiliary sampling equipment as required in Table 3-3.

3-4 Version 24 01/10

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

91. 077AG2.4.31 091 Unit 1 is at 100% power with the following conditions:

Several substations are separated from the grid resulting in the following plant conditions:

Unit 1 Generator Voltage is 20.45 kV.

The following alarms have actuated:

WE2, 1F 4KV BUS OV-OR-UV OR LOSS OF DC.

VE2, 1G 4KV BUS OV-OR-UV OR LOSS OF DC.

Grid frequency has fallen to 59.6 hertz and is stable.

4160V Bus voltages are 3840 Volts.

This condition has existed for the past hour.

Which one of the following completes the statements below?

The Generator temperatures will (1) .

AOP-5.2, Degraded Grid, will require the crew to (2) .

(1) (2)

A. RISE immediately enter AOP-17.1, Rapid Turbine Power Reduction B. LOWER immediately enter AOP-17.1, Rapid Turbine Power Reduction C. RISE place the unit in mode 3 in the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months using UOP-3.1, POWER OPERATION D. LOWER place the unit in mode 3 in the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months using UOP-3.1, POWER OPERATION AOP-5.2 v14.0 The FNP HV switchyard is connected to the Grid at a location where there is limited generation, and a loss of "various substations" could result in either a loss of load OR a loss of supporting generation (supply).

Generator (and Bus) voltage and frequency display that an OVERLOAD condition has occurred (more load than generation capacity), this would result in HIGHER currents on the Main Generator windings, and VR thus resulting in Greater I2R losses (MORE HEAT generation). If the condition were to persist, the insulation within the Main Generator begins to break down resulting in a reduction in its RESISTANCE and eventual failure.

Annunciator response procedures for NA1 & NB1, v18, provides the following direction:

Tuesday, March 19, 2013 11:38:43 AM 56

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 UNIT 1 procedures state: unlimited operation is allowed between 57 Hz and 61.8 Hz.

Both WE2 and VE2 will direct AOP-5.2 entry, and both are ENTRY CONDITIONS for AOP-5.2. (RO knowledge).

AOP-5.2, v14.0, step 3 will evaluate bus voltage.

IF <3850V then logging voltage every 30 mins is required per step 3.2 RNO actions.

and aligning equipment for the most reliable conditions. IF the conditions persist for > 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , then step 11 will require a planned shutdown to be conducted within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

UNIT 2 contains the following guidance:

NOTE: Per FNP-2-SOP-28.1, Turbine Generator Operation, the accumulated time operating between 58.5 Hz and 59.5 Hz should not exceed 60 minutes.

IF the frequency remains above 57 Hz and approaches the time band limits (see above note), THEN operator action is required to prevent turbine damage by removing the turbine from the grid.

A. Incorrect 1) Correct, temperatures will rise due to the high current conditions. See above.

2) Incorrect, reducing Reactor power per AOP-17, would exasperate the Grid voltage condition; Reducing Turbine load would result in reducing Grid voltage further and potentially causing a Degraded Grid LOSP condition. The potential damage to the Generator is NOT instantaneous and delaying or slowing the progression of the degraded grid would permit more time for ACC/PCC to correct the condition, before increasing the risk to the plant and/or the public by dropping a significant power supply thus potentially causing an LOSP condition for FNP and a Blackout condition for the grid. (See TS B3.3.5 discussion).

Plausible: The low frequency condition may lead one to believe that the conditions are dire and that immediate action is required to protect the main generator. An immediate load reduction might be warranted for a HIGH voltage condition (although only by the direction of PCC/ACC; not directed by AOP-5.2 for a high voltage condition). Also, if examinee properly assesses the temp impact, then one might believe a rapid load reduction is required to protect the Main Turbine from damage.

NOTE: immediately is part of answer choice A & B since AOP-17.1 could be required AFTER UOP-3.1 is initiated if a delay is encountered to comply with TS 3.3.5)

B. Incorrect 1) Incorrect, temperatures would go up.

Plausible: This would be the correct temperature response and ACTION if 4160V bus voltages were >4220 Volts; caused by a loss of LOAD vs Generation. Lower current flows from the generator and from the voltage Tuesday, March 19, 2013 11:38:43 AM 57

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 regulator would reduce I2R losses.

2) Incorrect, see A.2 Plausible: The low voltage conditions on the 4160V bus would result in an increased current draw on each of the loads within the plant therefore, sustained operation with degraded voltage, one might believe that the downpower is necessary to protect the plant components if they had incorrectly assessed the temperature impact on the Main Generator.

C. Correct 1) Correct, see A.1.

2) Correct, see above procedure quotes/summary. Although it would appear that an immediate response is required for protection of the Main Generator, TS and the procedure direct a more orderly approach to try to restore proper voltage and frequency, and then if necessary take the Unit offline and prevent a challenge to safety systems by inducing an LOSP.

D. Incorrect 1) Incorrect, see B.1.

2) Correct, see C.2.

Plausible: This would be selected if one were to improperly asses the temperature impact, and properly recall the transition requirement within the procedure.

K/A: 077AG2.4.31 Generator Voltage and Electric Grid Disturbances Knowledge of annunciator alarms, indications, or response procedures.

(CFR: 41.10 / 45.3)

IMPORTANCE RO 4.2 SRO 4.1 Importance Rating: 4.2 4.1 Technical

Reference:

FNP-0-ARP-2.1, v35.0 FNP-0-ARP-2.2, v32.2 FNP-1-ARP-1.13, v18.1 FNP-1-AOP-5.2, v15.0 References provided: NONE Learning Objective: EVALUATE plant conditions and DETERMINE if transition to another section of AOP AOP-5.1, Contingency Electrical Alignments and AOP-5.2, Degraded Grid or to another procedure is required. (OPS-62521N02)

Question History: 2011 HLT-34 NRC Exam question # 90 Tuesday, March 19, 2013 11:38:43 AM 58

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 K/A match: Generator Voltage is degraded below operational limits of SOP-36.8, the candidate must interpret these indications and determine that the Main Generator is overloaded vs underloaded and a generator overheat condition will occur.

The candidate must evaluate the indications and determine the actions required by the required procedures.

SRO justification: 10 CFR 55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations [10 CFR 55.43(b)(5)], involving BOTH:

1) assessing plant conditions and then

2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

One area of SRO level knowledge is knowledge of content of the procedure vs. the procedure's overall mitigative strategy or purpose.

AOP-5.1 has two different strategies; based on the conditions of the degraded grid (HI or LOW). To answer this correctly the candidate must have knowledge of:

the fundamental temperature response to an overload on the Generator.

the transitions required 2013 NRC exam From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 flowchart for 10 CFR 55.43(b)(5) :

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations [10 CFR 55.43(b)(5)], involving BOTH:

1) assessing plant conditions and then

2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

Using the flowchart, this question can:

NOT be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location. (PT 1 IS sys/fund knowledge, but PT 2 requires procedural knowledge)

NOT be answered solely by knowing immediate operator actions.

NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs.

NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure. (PT 2 is procedure selections based on evaluation of conditions and not the generic overall mitigative Tuesday, March 19, 2013 11:38:43 AM 59

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 strategy of the procedure.

CAN be answered with knowledge of ONE or MORE of the following:

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

Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps. (Requires knowledge of the TIME requirements for the implementation of the transitions (using transition as equivalent to attachment and appendices of same procedure since AOP-5.2 will be conducted in parallel to UOP-3.1).

Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures. (Where "event specific" in this case is the Shutdown guidance)

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures. (The plant shutdown time requirements vs immediately are incorporated within the decision point.)

Tuesday, March 19, 2013 11:38:43 AM 60

10/18/12 9:26:20 FNP-1-AOP-5.2 UNIT 1 DEGRADED GRID Version 15.0 Step Action/Expected Response Response Not Obtained 9

__ 9 Contact Alabama Control Center For Voltage Control Strategies NOTE: Step 10 is a continuing action step.

10

__ 10 [CA] Check Status of 4160 V Buses 1F And 1G:

10.1 10.1 4160 V Buses 1F and 1G voltages - LESS 10.1 Return to step 2.

THAN 4200 V 10.2 10.2 4160 V Buses 1F and 1G voltages - LESS 10.2 Perform the following:

THAN 3850 V 10.2.1 10.2.1 IF no PCC notification exists, THEN go to procedure and step in effect.

10.2.2 10.2.2 Return to step 2.

11°

__ 11 Evaluate need for plant shutdown (refer to TS 3.3.5, Action F.1):

11.1 11.1 Check reactor - MODE 1 OR 2 11.1 Proceed to Step 12.

11.2 11.2 Check duration of degraded voltage - 11.2 Return to Step 2.

GREATER THAN ONE HOUR 11.3 11.3 Initiate actions to place Reactor in MODE 3 within the next 6 Hours using:

FNP-1-UOP-3.1, POWER OPERATION

FNP-1-UOP-2.1, SHUTDOWN OF UNIT FROM MINIMUM LOAD TO HOT STANDBY

__Page Completed 6 ProcedureStepsMain Page 7 of 8

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

92. G2.1.20 092 Unit 1 had an extended Loss of all AC power with the following conditions:

At 1000:

Power has been restored to the 1F and 1K 4160V busses.

The Shift Supervisor is preparing to exit ECP-0.0, Loss of ALL AC Power, and is at the step to "Evaluate plant conditions".

SCMM is 39°F.

Pressurizer level is 16% and slowly falling.

CTMT pressure is 5.8 psig.

SI equipment did NOT automatically actuate when power was restored.

At 1002:

ECP-0.0 has been exited and the applicable recovery procedure has been entered.

The Shift Supervisor is informed that a RED path exists for Heat Sink on the IPC CSF status trees.

Which one of the following completes the statements below?

At 1000, entry into (1) is required.

At 1002, per the applicable procedure in effect, the Control Room crew (2) required to enter FRP-H.1, Loss of Secondary Heat Sink.

Procedure titles are as follows:

ECP-0.1, Loss of All AC Power Recovery Without SI Required.

ECP-0.2, Loss of All AC Power Recovery With SI Required.

(1) (2)

A. ECP-0.1 IS B. ECP-0.1 is NOT C. ECP-0.2 IS D. ECP-0.2 is NOT Tuesday, March 19, 2013 11:38:43 AM 61

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 ECP-0.0 Step 31 31 Evaluate plant conditions.

31.1 Check SI not required. 31.1 Go to FNP-1-ECP-0.2, LOSS OF ALL AC POWER RECOVERY WITH SI

Check SUB COOLED MARGIN REQUIRED.

MONITOR indication - GREATER than 16F{45F} SUBCOOLED IN CETC MODE.

Check Pressurizer level -

GREATER THAN 13%{43%}.

Check SI equipment - HAS NOT ACTUATED UPON AC POWER RESTORATION such that SI flow occurred.

31.2 Go to FNP-1-ECP-0.1, LOSS OF ALL AC POWER RECOVERY WITHOUT SI REQUIRED.

A. Incorrect - 1) Incorrect, plausible since SI equipment has not actuated and other parameters at 10:00 would direct a transition to ECP-0.1 if NON-Adverse numbers were applied (See step 31 of ECP-0.0). This would be a correct answer if Containment pressure were < 4 psig.

2) Incorrect, plausible since this is the normal strategy when a FRP red path condition is detected. There is a note above step 1 of ECP-0.2 that states FRP's are not implemented until after completion of step 13.

B. Incorrect - 1) Incorrect, see A.1.

2) Correct, there is a note above step 1 of ECP-0.2 that states FRP's are not implemented until after completion of step 13..

C. Incorrect - 1) Correct, entry into ECP-0.2 is required (See step 31 of ECP-0.0).

2) Incorrect, see A.2.

D. Correct - 1) Correct, see C.1.

2) Correct, see B.2.

Tuesday, March 19, 2013 11:38:43 AM 62

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 K/A: G2.1.20 Ability to interpret and execute procedure steps.

(CFR: 41.10 / 43.5 / 45.12)

IMPORTANCE RO 4.6 SRO 4.6 Importance Rating: 4.6 4.6 Technical

Reference:

FNP-1-ECP-0.0 v26 FNP-1-ECP-0.2 v19 References provided: None Learning Objective: ASSESS the facility conditions associated with the (1)

ECP-0.0, Loss of All AC Power; (2) ECP-0.1, Loss of All AC Power Recovery, Without SI Required; (3) ECP-0.2, Loss of All AC Power Recovery, With SI Required, and based on that assessment: (OPS-62532A01)

SELECT the appropriate procedures during normal, abnormal and emergency situations. 10CFR55.43 (b) 5

DETERMINE if transition to another section of the procedure or to another procedure is required

DETERMINE if the critical safety functions are satisfied Question History: New question K/A match: This question requires the candidate to interpret the conditions given and choose the correct recovery procedure.

In addition, once the procedure transition is made, the candidate has to properly execute implementation of the recovery procedure, ECP-0.2.

SRO justification: 10 CFR 55.43(b)(5)

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

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

- Procedure selection is required on the first and second parts of this question.

Tuesday, March 19, 2013 11:38:43 AM 63

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures.

- Knowledge of diagnostic steps is required on the first part of this question.

2013 NRC exam From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 flowchart for 10 CFR 55.43(b)(5) :

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations [10 CFR 55.43(b)(5)], involving BOTH:

1) assessing plant conditions and then

2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

Using the flowchart, this question can:

NOT be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location.

NOT be answered solely by knowing immediate operator actions.

NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs.

NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

Be answered with knowledge of ONE or MORE of the following:

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

Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures.

Tuesday, March 19, 2013 11:38:43 AM 64

1/22/2013 14:14 FNP-1-ECP-0.0 UNIT 1 LOSS OF ALL AC POWER Revision 26 Step Action/Expected Response Response NOT Obtained

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ NOTE: If RCP seal cooling was previously isolated, further cooling of the RCP seals will be established by natural circulation cooldown as directed in subsequent procedures.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ 31 Evaluate plant conditions.

31.1 Check SI not required. 31.1 Go to FNP-1-ECP-0.2, LOSS OF ALL AC POWER RECOVERY WITH SI Check SUB COOLED MARGIN REQUIRED.

MONITOR indication - GREATER than 16 16F{45 F{45F} SUBCOOLED IN CETC MODE.

Check pressurizer level - Transition is required, based on GREATER THAN 13%{43%}. conditions, to ECP-0.2 Check SI equipment - HAS NOT ACTUATED UPON AC POWER RESTORATION such that SI flow occurred.

31.2 Go to FNP-1-ECP-0.1, LOSS OF ALL AC POWER RECOVERY WITHOUT SI REQUIRED.

-END-Page 43 of 44

1/22/2013 14:14 FNP-1-ECP-0.2 UNIT 1 LOSS OF ALL AC POWER RECOVERY WITH SI REQUIRED Revision 19 Step Action/Expected Response Response NOT Obtained CAUTION CAUTION:: Critical safety function status trees should be monitored for information only. No Function Restoration Procedure should be implemented until completion of step 13.

Once transition is made to ECP-0.2, FRP's are not implemented until some recovery 1 Reset SI signals. actions are performed.

1.1 Verify SI - RESET. 1.1 Perform the following:

[] MLB-1 1-1 not lit (A TRN) 1.1.1 IF any train will NOT reset

[] MLB-1 11-1 not lit (B TRN) using the MCB SI RESET pushbuttons, THEN place the affected train S821 RESET switch to RESET. (SSPS TEST CAB.)

1.1.2 IF a failure exists in SSPS such that SI cannot be reset, THEN reset SI using FNP-1-SOP-40.0, RESPONSE TO INADVERTENT SI AND INABILITY TO RESET OR BLOCK SI, Appendix 2.

1.2 IF DF08-1 is closed AND LOSP has occurred, THEN place B1F Sequencer Test Trip Override Switch (TTOS)in OFF:

1.2.1 Obtain B1F Sequencer TTOS key from SSS office.

1.2.2 Insert key into B1F Sequencer Test Trip Override Switch.

1.2.3 Place B1F Sequencer Test Trip Override Switch in OFF.

1.3 Reset B1F sequencer. (139 ft, AUX BLDG A train SWGR room)

Step 1 continued on next page.

Page 2 of 16

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

93. G2.1.37 093 Unit 1 is at 75% power and ramping up with the following conditions:

A reactivity plan has been developed and approved for the reactivity additions.

A Reactivity Management SRO has been stationed due to frequent reactivity additions.

Which one of the following completes the statements below per NMP-OS-001, Reactivity Management Program?

During the ramp up, the limit for Control Rod withdrawal is a maximum of (1) per rod pull.

The Reactivity Management SRO (2) authorize changes to the reactivity plan.

(1) (2)

A. three steps CANNOT B. three steps CAN C. four steps CANNOT D. four steps CAN Tuesday, March 19, 2013 11:38:43 AM 65

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 From NMP-OS-001 Step 6.3.9.3 -

Plant Farley and Plant Vogtle 1&2:

When withdrawing control rods in MODE 1, the OATC shall stop rod withdrawal at least every three steps and check for expected response on NIs, DRPI, and reactor coolant temperature (i.e. pull and wait).

Step 6.4.2-While the Reactivity Management SRO provides direct oversight and approval of the planned reactivity additions, the SS must approve the reactivity plan as well as any changes to the plan.

A. Correct - 1) Correct, the limit for Control Rod withdrawal is 3 steps per rod pull.

2) Correct, the Reactivity Management SRO cannot modify the reactivity plan. The SS has to approve the reactivity plan in addition to any changes to the reactivity plan.

B. Incorrect- 1) Correct, see A.1.

2) Incorrect, plausible since the candidate may think the Reactivity Management SRO has taken over the total responsibility for reactivity management.

C. Incorrect- 1) Incorrect, plausible because NMP-OS-001 has a 4 step per rod pull limit for Vogtle 3 & 4 (see step 6.3.9.4 of NMP-OS-001).

2) Correct, see A.2.

D. Incorrect- 1) Incorrect, see C.1.

2) Incorrect, see B.2.

K/A: G2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management.

(CFR: 41.1 / 43.6 / 45.6)

IMPORTANCE RO 4.3 SRO 4.6 Importance Rating: 4.3 4.6 Technical

Reference:

NMP-OS-001, v17 References provided: None Learning Objective: STATE and EXPLAIN the responsibilities of the Reactivity Management SRO as described in NMP-OS-001, Reactivity Management Program. (OPS-62303P03)

Tuesday, March 19, 2013 11:38:43 AM 66

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 Question History: Modified from Farley Bank question RX MGMT-53203P02 04 K/A match: The candidate is questioned on the limitations for Control Rod withdrawal during a routine ramp, and also the responsibilities of the Reactivity Management SRO during reactivity manipulations during a ramp.

SRO justification: This question meets the criteria for SRO by requiring the candidate to have knowledge of administrative procedures (NMP-OS-001, Reactivity Management Program) that specifies the responsibilities of the SRO (Reactivity Management SRO) during a ramp when reactivity manipulations are being performed.

10 CFR 55.43(b)(5)

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

The applicants knowledge can be evaluated at the level of 10 CFR55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.

10 CFR 55.43(b)(6)

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

Could also fall under the SRO criteria of this CFR, although not specifically listed as one of the examples. The fact that it is not listed does not exclude it from inclusion if it meets the criteria of 10 CFR 55.43(b)(6). The question is about "Procedures and limitations involved in control rod Tuesday, March 19, 2013 11:38:43 AM 67

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 reactivity".

2013 NRC exam Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.

NMP-OS-001 is an administrative procedure that provides guidance on reactivity management during a routine ramp up per UOP-3.1.

This question could also meet the SRO criteria under :

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

Although not listed as an example for this category in the "Clarification Guidance for SRO-only Questions", this question does include procedural administrative requirements and controls associated with external effects on core reactivity.

Tuesday, March 19, 2013 11:38:43 AM 68

Southern Nuclear Operating Company Nuclear NMP-OS-001 Management Reactivity Management Program Version 17.0 Procedure Page 14 of 39 6.3.9.3 Plant Farley and Plant Vogtle 1&2:

When withdrawing control rods in MODE 1, the OATC shall stop rod withdrawal at least every three steps and check for expected response on NIs, DRPI, and reactor coolant temperature (i.e. pull and wait). This requirement may be suspended during evolutions that have adequate procedural direction and oversight during the withdrawal of the control rods (i.e. recovery of a dropped rod, rod control surveillances). It should be noted that many of the fuel-damaging events in the industry have resulted from misoperation of control rods. Careful monitoring of the rod control system is essential.

Peer check for rod manipulations should confirm placement of hand on rod motion switch in a manner that allows physical confirmation of intended direction of rod motion.

6.3.9.4 Plant Vogtle 3&4:

During Mode 1 operation, control rods are normally in automatic control. If continuous manual control rod withdrawal is required above the point of adding heat, the OATC shall stop rod withdrawal and check for the expected response on NIs, DRPI, and reactor coolant temperature (i.e. pull and wait). Withdrawal is Plausibility for 4 steps limited in M1, M2, and AO banks to 4 steps, and MA, MB, MC, MD banks are for distracter. limited to 20 steps.

The OATC shall monitor/report Gray Bank and Axial Offset automatic rod movement to control temperature (M-banks), power (when rod control is in low power mode) and axial flux position (AO Rods). The OATC shall report Tave-Tref deviation and proper M-Bank rod insertion or Delta Flux Deviation and proper operation of the AO banks as they withdraw or insert.

Peer check for manual rod manipulations should confirm the mouse pointer is placed over the desired direction (i.e. in or out). Control rods will move in or out based on the current sequence. If moving a bank or group of rods, the peer checker will also confirm that the proper bank or group is selected.

6.4 Reactivity Management SRO 6.4.1 An additional active licensed SRO shall be dedicated to provide direct and intrusive oversight of reactivity manipulations allowing the SS to maintain the overall perspective of unit operation. A Reactivity Management SRO may also be stationed at the request of the Shift Supervisor during power maneuvers involving frequent reactivity manipulations.

6.4.2 While the Reactivity Management SRO provides direct oversight and approval of the planned reactivity additions, the SS must approve the reactivity plan as well as any changes to the plan.

Modified this question from FNP Bank to develop G2.1.37 QUESTIONS REPORTquestion for 6 FNP Bank Questions

1. RX MGMT-53203P02 004 Unit 2 is performing a Rx Startup using UOP-1.2, Startup of Unit from Hot Standby to Minimum Load.

Which one of the following describes the the requirements of stationing and the authority given to a Reactivity Management SRO as stated in NMP-OS-001, Reactivity Management Program?

The Reactivity Managment SRO is REQUIRED to be stationed while power is A. being raised from the POAH to stable at power operations, and may authorize reactivity additions and changes to the reactivity plan; the SS is not REQUIRED to approve the changes.

B. being raised from the POAH to stable at power operations, and may ONLY approve adding positive reactivity that complies with that already briefed with the SS; Changes can not be approved by the Reactivity Management SRO.

C. approaching criticality, may authorize reactivity additions and changes to the reactivity plan; the SS is not REQUIRED to approve the changes.

D. approaching criticality, may ONLY approve adding positive reactivity that complies with that already briefed with the SS; Changes can not be approved by the Reactivity Management SRO.

Tuesday, March 19, 2013 1:38:36 PM 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

94. G2.2.21 094 Unit 1 is at 100% power with the following conditions:

The packing was replaced on MOV-8812B, CTMT SUMP TO 1B RHR PUMP.

The MOV was left in the closed position.

The Return to Service Tagout is in progress and power has been restored to the MOV.

Which one of the following states the minimum action(s) required to restore MOV-8812B to OPERABLE after the valve packing replacement?

MOV-8812B is OPERABLE when .

A. the MOV's auto open function is tested per its Surveillance Test Procedure B. the MOV has been satisfactorily time stroked per its Surveillance Test Procedure C. the MOV is stroked open and closed from its remote handswitch per the guidance of the Tagout D. the MOV is manually stroked open and closed per the guidance of the Tagout, with no leakage verified A. Incorrect - plausible since this is a safety function of the MOV, and there is a test (STP-11.13) to ensure the valve will auto open when required. Incorrect since no actions have been performed on the MOV that would affect its auto open signal.

B. Correct - Since maintenance was performed on the MOV that could affect its stroke time, and thus the ability to perform its safety function, a Surveillance test that time strokes the valve must be performed.

C. Incorrect - plausible because this is guidance that may be included on the Tagout for a MOV that was manually operated with no maintenance performed that would affect is stroke time. If the MOV had only been manually operated (and no work performed on the MOV), this would be the correct answer.

D. Incorrect - plausible since one of the post maintenance testing actions performed for manual valves that have been repacked is to verify that the valve operates smoothly with no binding, and to also verify no leakage.

K/A: G2.2.21 Knowledge of pre- and post-maintenance operability requirements.

(CFR: 41.10 / 43.2)

Tuesday, March 19, 2013 11:38:43 AM 69

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 IMPORTANCE RO 2.9 SRO 4.1 Importance Rating: 2.9 4.1 Technical

Reference:

FNP-1-STP-11.6, v40.1 NMP-AD-012, v12.0 FNP-0-SOP-0.13, v26.0 References provided: None Learning Objective: RECALL AND APPLY the information from the LCO BASES sections: BACKGROUND, APPLICABLE SAFETY ANALYSIS, ACTIONS, SURVEILLENCE REQUIREMENTS, for any Technical Specifications or TRM requirements associated with the Residual Heat Removal System components and attendant equipment alignment, to include the following (OPS-62101K01): 10CFR55.43 (b) 2 3.4.3, RCS Pressure and Temperature (P/T) Limits 3.4.6, RCS Loops - MODE 4 3.4.7, RCS Loops - MODE 5, Loops Filled 3.4.8, RCS Loops - MODE 5, Loops Not Filled 3.4.12, Low Temperature Overpressure Protection (LTOP)

System 3.4.14, RCS Pressure Isolation Valve (PIV) Leakage 3.5.2, ECCS - Operating 3.5.3, ECCS - Shutdown 3.9.4, Residual Heat Removal (RHR) and Coolant Circulation - High Water Level 3.9.5, Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level 13.5.1, Emergency Core Cooling System (ECCS)

Question History: New question K/A match: This question tests the candidate's knowledge of requirements for return to service of a Safety Related piece of equipment after maintenance has been completed. This tests knowledge of what restores operability to a MOV that has had work performed on it that can affect its stroke time.

SRO justification: This question meets the criteria for SRO by requiring the candidate to have knowledge of administrative procedures (FNP-0-SOP-0.13, Recording Limiting Conditions for Operations, and NMP-AD-012, Operability Determinations and Functionality Assessments) that specifies how to return a piece of equipment to service after maintenance and how to use that in conjunction with Tech Specs to declare a system OPERABLE.

Tuesday, March 19, 2013 11:38:43 AM 70

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 10 CFR 55.43(b)(5)

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

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.

Could also be covered by 10 CFR 55.43(b)(2) as knowledge of Surveillance requirements to return a piece of Safety Related equipment to service.

10 CFR 55.43(b)(2):

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

2013 NRC exam Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.

Application of Required Actions (Section 3) and Surveillance Requirements (SR)

(Section 4) in accordance with rules of application requirements (Section 1).

Tuesday, March 19, 2013 11:38:43 AM 71

03/15/13 2:07:03 UNIT 1 FNP-1-STP-11.6

_____ 5.2.13 Verify CTMT SUMP TO 1B RHR PUMP, Q1E11MOV8811B closed.

_____ 5.2.14 Independently verify CTMT SUMP TO 1B RHR PUMP, IV Q1E11MOV8811B closed.

NOTE: Instructions for time stroke of CTMT SUMP TO 1B RHR PUMP, Q1E11MOV8812B during outages can be found in Appendix 1.

5.2.15 IF the following conditions exist,

[] 1A RCS LOOP TO 1B RHR PUMP Q1E11MOV8702A closed

[] 1A RCS LOOP TO 1B RHR PUMP Q1E11MOV8702B closed THEN, time stroke CTMT SUMP TO 1B RHR PUMP, Q1E11MOV8812B as follows:

x Open CTMT SUMP TO 1B RHR PUMP, Q1E11MOV8812B and record the opening time in the ACTUAL STROKE TIME column of Table 1.

x Close CTMT SUMP TO 1B RHR PUMP, Q1E11MOV8812B and record the closing time in the ACTUAL STROKE TIME column of Table 1.

ACCEPTANCE CRITERIA: Stroke times are less than or equal to the Maximum Allowable Times listed in Table 1.

5.2.16 Verify CTMT SUMP TO 1B RHR PUMP, Q1E11MOV8812B closed.

5.2.17 Independently verify CTMT SUMP TO 1B RHR PUMP, IV Q1E11MOV8812B closed.

Version 40.1

Southern Nuclear Operating Company Nuclear NMP-AD-012 Operability Determinations and Management Version 12.0 Functionality Assessments Procedure Page 7 of 27 3.9 Immediate Determination of Operability (IDO) - An initial determination of operability is made after confirming the existence of a degraded, nonconforming or unanalyzed condition.

The determination should be made without delay and in a controlled manner using the best available information. The information should be sufficient to conclude that there is a reasonable expectation that the SSC is operable. The IDO provides justification of operability while a PDO is in progress. Instructions for processing an IDO in Maximo are found in NMP-AD-012-GL03. Instructions for processing a PDO are found in NMP-AD-012-GL01.

3.10 Necessary Support Function - A function of an SSC that is required to support operability of an SSC described in TSs. SSCs that provide necessary support functions (i.e., Support SSCs) may or may not be described in TSs. If they are described in TS, then the term operable applies to them. If they are not described in TSs, then the term operable does not apply to them directly; however, these SSCs are still within the scope of the Operability Determination process because their lack of functionality can affect the operability of SSCs described in TS. Refer to Attachment 2, Section 6 of NMP-AD-012-GL01 for instructions.

3.11 Nonconforming Condition - A nonconforming condition is a condition of an SSC that involves a failure to meet the CLB or a situation in which quality has been reduced because of factors such as improper design, testing, construction, or modification. The following are examples of nonconforming conditions:

x An SSC fails to conform to one or more applicable codes or standards (e.g., the CFR, operating license, and TSs, UFSAR, and/or licensee commitments).

x An as-built or as-modified SSC does not meet the CLB.

x Operating experience or engineering reviews identify a design inadequacy.

x Documentation required by NRC requirements such as 10 CFR 50.49 is unavailable or deficient.

3.12 Operability Declaration - An operability declaration is a decision by a senior licensed operator on the operating shift crew that there is a reasonable expectation that an SSC can perform its specified safety function.

3.13 Operable/Operability - In accordance with the TS, a system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its Specified Safety Functions, and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function(s) are also capable of performing their related support function(s).

In order to be considered operable, an SSC must be capable of performing the safety functions specified by its design, within the required range of design physical conditions, initiation times, and mission times. In addition, TS operability considerations require that an SSC meet all surveillance requirements (as specified in Surveillance Requirement (SR)

Applicability SR 3.0.1). An SSC that does not meet an SR must be declared inoperable. For Operability Determination purposes, the mission time is the duration of SSC operation that is credited in the design basis for the SSC to perform its Specified Safety Function.

Procedure Number Ver SHARED Farley Nuclear Plant FNP-0-SOP-0.13 26.0 1/17/2013 RECORDING LIMITING CONDITIONS FOR Page Number 20:10:24 13 of 55 OPERATIONS 3.4 Initiation of LCO/TR Status Sheet For An Admin LCO/TR Complete Section I of the LCO/TR Status Sheet per Section 3.1 except as modified These are the by the following subsections.

actions required to 3.4.1 LCO/TR Initiated restore OPERABILITY after Enter date and time of equipment malfunction.

work is performed.

3.4.2 Required Actions The STP that time strokes the MOV3.4.2.1 is State any mode restraints or restraints on equipment operation. The the functional test. completion section will be marked N/A.

3.5 Restoration of LCO/TR s 3.5.1 As a minimum, the following items should be addressed in determining when an LCO/TR may be closed out:

3.5.1.1 Field work on the initiating condition resolved/corrected.

3.5.1.2 A functional test or other method has demonstrated operability.

3.5.2 For certain conditions, an LCO/TR may be closed out based on completion of functional testing without having the applicable WO or surveillance test reviewed by an ATL. The intent is that the extent of field work is understood and that functional testing adequately tests all field work performed.

Examples include, but are not limited to:

3.5.2.1 7300 and NI A and B pack testing, provided acceptance criteria is satisfactorily met as determined by the journeyman prior to coming out of test.

3.5.2.2 Diesel Generator maintenance activities such as 24 month outages and minor leak repairs.

3.5.2.3 Radiation monitor activities such as pump lubes or paper drive inspections provided Control Room drawer alignment is performed satisfactory.

3.5.2.4 Radiation monitor functional test, calibration or isotopic provided acceptance criteria is satisfactorily met as determined by the journeyman.

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

95. G2.2.23 095 Unit 2 is at 100% power with the following conditions:

The 2B RHR pump is being Tagged Out for a scheduled oil change and maintenance inspection.

Maintenance is expected to take 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months to complete.

Which one of the following completes the statements below?

The LCO required is a (1) LCO.

In addition to a Control Room LCO Log entry, an LCO/TR Status Sheet (2) required to document the LCO.

(1) (2)

A. Voluntary IS B. Voluntary is NOT C. Mandatory IS D. Mandatory is NOT SOP-0.13 3.0.1 Initiation of Hard Copy LCO/TR ACTION statement When a unit fails to meet an applicable LCO/TR and requires initiation of an ACTION statement, or equipment malfunction requires an Admin LCO/TR, the SS or a licensed individual shall:

a. Complete Figure 4, "LCO/TR Status Sheet"Section I.

3.0.2 Short Term LCO's Initiation of short term (log or computer entry) LCO/TR ACTION statement Completion of an LCO/TR Status Sheet is NOT required for the performance of surveillances or short term entry into an LCO/TR which will be cleared prior to turning over the Control Room Log for the shift.

3.1.3 Type of LCO/TR IF the LCO/TR becomes effective due to circumstances beyond the Shift Supervisors control (i.e., a covered piece of equipment breaks), then this will be indicated by entering Mandatory (M).

IF the LCO becomes effective due to the Shift Supervisors decision (i.e., a covered piece of equipment is voluntarily removed from service for maintenance etc.), then this will be indicated by entering Voluntary (V).

If a piece of equipment is removed from service that is not required in the present plant mode, but is required in an higher plant mode or if it reduces the redundancy, then an Administrative LCO may be written.

Tuesday, March 19, 2013 11:38:43 AM 72

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 This will be indicated by entering Administrative (A).

A. Correct 1) Correct, per SOP-0.13, if a piece of equipment is OPERABLE and voluntarily removed from service, it is a Voluntary LCO.

2) Correct, an LCO/TR Status sheet is required if the LCO will be in effect for longer than one shift. Since the LCO will be in effect for 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , a LCO/TR Status Sheet is required to be filled out.

B. Incorrect 1) Correct, see A.1.

2) Incorrect, plausible because per SOP-0.13 it is acceptable to only make an entry into the Control Room Log for LCO's that are of short duration (will be restored prior to the end of the shift).

C. Incorrect 1) Incorrect, Mandatory LCO's are required when an unplanned event occurs that makes a piece of equipment INOPERABLE. If the equipment was failed when removed from service, it is a Mandatory LCO. Plausible that a candidate may think it is a Mandatory LCO since he may associate Voluntary LCO's with Admin LCO's. Since this is an actual Tech Spec entry, where REQUIRED ACTIONS are implemented, he may determine it is a Mandatory LCO

2) Correct, see A.2.

D. Incorrect 1) Incorrect, see C.1.

2) Incorrect, see B.2.

K/A: G2.2.23 Equipment Control - Ability to track Technical Specification limiting conditions for operations.

(CFR: 41.10 / 43.2 / 45.13)

IMPORTANCE RO 3.1 SRO 4.6 Importance Rating: 3.1 4.6 Technical

Reference:

FNP-0-SOP-0.13, Recording Limiting Conditions for Operations, v26.0 References provided: None Learning Objective: STATE AND DESCRIBE the process for recording and tracking the failure to meet the Limiting Condition for Operation and Technical Requirements (OPS-52302A07)

Question History: New question K/A match: Candidate is given conditions where an LCO is required, Tuesday, March 19, 2013 11:38:43 AM 73

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 addition the candidate has to determine what tools are acceptable for tracking the status of the LCO.

SRO justification: This question meets the criteria for SRO by requiring the candidate to have knowledge of administrative procedures (FNP-0-SOP-0.13, Recording Limiting Conditions for Operations) that specifies how to document and track Tech Spec LCO's (implementation of Tech Specs).

10 CFR 55.43(b)(5)

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

The applicants knowledge can be evaluated at the level of 10 CFR55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.

2013 NRC exam Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.

Tuesday, March 19, 2013 11:38:43 AM 74

Procedure Number Ver SHARED Farley Nuclear Plant FNP-0-SOP-0.13 26.0 1/17/2013 RECORDING LIMITING CONDITIONS FOR Page Number 20:10:24 7 of 55 OPERATIONS 3.0 LCO/TR STATUS SHEET PREPARATION NOTE A Shift Support Supervisor (SSS) or a licensed individual can enter the required information on Figures 4, 5, 6, and 7; however, a SS must be informed.

For T.S. LCO Required Actions with completion times less than or equal to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , an administrative reminder shall be initiated, and the Shift Manager shall be notified of the required action and its associated completion time.

3.1 Initiation of Hard Copy LCO/TR ACTION statement.

When a unit fails to meet an applicable LCO/TR and requires initiation of an ACTION statement, or equipment malfunction requires an Admin LCO/TR, the SS or a licensed individual shall:

3.1.1 Complete Figure 4, "LCO/TR Status Sheet"Section I.

3.1.2 For all unplanned LCO entries and emergent work, an INDEPENDENT VERIFICATION by a licensed SRO will be performed, and will validate the correct LCO has been entered.

3.1.3 Notify the Shift Supervisor (SS) and Shift Manager (SM) 3.1.4 Log the status sheet on Figure 7, "LCO/TR Status Log". The LCO/TR Status Sheet shall be located in the LCO Book.

3.1.5 In the Control Room Log, enter bracketed (< >) information which is applicable to the Tech Spec/TRM/ODCM entry.

3.1.6 For mandatory LCO entries, inform, without delay, the Operations Manager or one of the Operations staff superintendents.

3.1.7 For T.S. LCO Required Actions with completion times less than or equal to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when the LCO is entered, initiate an appropriate administrative reminder to ensure that required actions are completed within the time requirements.

(AI 2009208864) 3.1.8 For T.S. LCO Required Actions with completion times less than or equal to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when the LCO is entered, notify the Shift Manager of the required actions and associated completion times. This does not remove the primary responsibility for impl menting T.S. from the Shift Supervisor. (AI 2009208864) e 3.1.9 IF required, notify Security.

3.1.10 IF required, update EOOS.

Procedure Number Ver SHARED Farley Nuclear Plant FNP-0-SOP-0.13 26.0 1/17/2013 RECORDING LIMITING CONDITIONS FOR Page Number 20:10:24 8 of 55 OPERATIONS 3.2 Initiation of short term (log or computer entry) LCO/TR ACTION statement.

If the LCO Completion of an LCO/TR Status Sheet is NOT required for the performance of will go past surveillances or short term entry into an LCO/TR which will be cleared prior to turning over the end of the Control Room Log for the shift. In this case, the following steps should be performed.

the shift, a hard copy 3.2.1 In the Control Room Log, enter bracketed (< >) information which is applicable LCO/TR to the Tech Spec/TRM/ODCM entry.

status 3.2.2 For planned work, the releasing SRO will be responsible and required to write sheet must the LCO entered for the specified work when the work order is released. The be used in applicable units SS will peer check the written LCO. The SS is expected to addition to annotate that the LCO has been peer checked in Control Room Log.

the log [AI 2008202372]

entry. The 3.2.3 For mandatory LCO entries, inform the Operations Manager or one of the CCW Operations staff superintendents.

pump will be out for 3.2.4 To provide a tool to assist in ensuring accurate and timely application of Tech 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . Spec actions, an INDEPENDENT VERIFICATION by a licensed SRO will be performed for all unplanned LCO entries and emergent work. This should be completely independent of the preparer, and will validate the correct LCO has been entered. [AI 2009204242]

x Example: The SRO tasked with performing the independent verification should be provided the conditions necessitating the unplanned LCO entry. Using plant knowledge and Technical Specifications, the SRO should arrive at the same conclusion as the person requesting the independent verification.}

3.2.5 For T.S. LCO Required Actions with completion times less than or equal to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when the LCO is entered, initiate an appropriate administrative reminder to ensure that required actions are completed within the time requirements.

(AI 2009208864) 3.2.6 For T.S. LCO Required Actions with completion times less than or equal to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when the LCO is entered, notify the Shift Manager of the required actions and associated completion times. This does not remove the primary responsibility for implementing T.S. from the Shift Supervisor. (AI 2009208864)

Procedure Number Ver SHARED Farley Nuclear Plant FNP-0-SOP-0.13 26.0 1/17/2013 RECORDING LIMITING CONDITIONS FOR Page Number 20:10:24 9 of 55 OPERATIONS 3.3 Initiation of LCO/TR Status Sheet For An LCO/TR.

Complete Section I of the LCO/TR Status Sheet in accordance with the following sections:

3.3.1 LCO/TR Number The LCO/TR Number is assigned by unit-year-sequence. Unit 1 and Shared LCOs are recorded in the Unit 1 Status Log.

e.g.:

1-2000-002 The 2nd LCO/TR for the year 2000 Unit 1 3.3.2 LCO / TR Initiated The date and time the LCO was not met. If a time cannot be determined for the initiating event, the time of discovery should be used.

Record the date as mm/dd/yyyy.

x Where:

mm = month (01-12) dd = day (01-31) yy = year (2000-2099)

Record the time using a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months format (hh-mm).

x Where:

hh = hours (00-23) mm = minute (00-59) 3.3.3 Type of LCO/TR If the LCO/TR becomes effective due to circumstances beyond the Shift Supervisors control (i.e., a covered piece of equipment breaks), then this will be indicated by entering Mandatory (M).

Type of LCO - IF the LCO becomes effective due to the Shift Supervisors decision (i.e., a Mandatory or covered piece of equipment is voluntarily removed from service for Voluntary. maintenance etc.), then this will be indicated by entering Voluntary (V).

If a piece of equipment is removed from service that is not required in the present plant mode, but is required in an higher plant mode or if it reduces the redundancy, then an Administrative LCO may be written. This will be indicated by entering Administrative (A).

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

96. G2.3.14 096 Unit 1 is shutdown with the following conditions:

STP-18.4, Containment Mid-Loop and/or Refueling Integrity Verification and Containment Closure, has been completed to allow refueling operations.

The Containment Equipment Hatch is open.

During fuel movement, a spent fuel assembly has dropped from the Manipulator Crane to the bottom of the cavity.

Both R-24A & B, CTMT PURGE, are in alarm.

AOP-30, Refueling Accident, actions are in progress.

Which one of the following completes the statements below?

The ACCEPTANCE CRITERIA of STP-18.4 for Refueling Integrity states that the Containment Equipment Hatch is capable of being closed within (1) of notification.

Per AOP-30, actions are required to place (2) in service without delay.

(1) (2)

A. four hours PRF (Penetration Room Filtration)

B. four hours CREFS (Control Room Emergency Filtration System)

C. two hours PRF (Penetration Room Filtration)

D. two hours CREFS (Control Room Emergency Filtration System)

Tuesday, March 19, 2013 11:38:43 AM 75

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 A. Incorrect - 1) Incorrect, plausible since some actions of Tech Specs and the ODCM allow a four hour time frame for completion.

2) Incorrect, plausible since AOP-30.0 directs to start PRF if it is a high radiation condition in the Spent Fuel Pool area. Since PRF also performs a function for atmospheric cleanup for a high radiation condition and leakage into the Penetration rooms, it is plausible that a candidate would determine that it should be started here.

B. Incorrect - 1) Incorrect, see A.1.

2) Correct, CREFS is required to be started without delay. This is directed at step 2 of AOP-30.0. In addition, there is a Caution statement at the beginning of Attachment 1 of AOP-30 (Control Room Isolation with CREFS in Service) to place CREFS in service without delay to ensure continued Control Room habitability.

C. Incorrect - 1) Correct, per step 2.3.1 of STP-18.4. The Acceptance Criteria states :A Maintenance Closure Response Team (MCRT) is available and briefed to effect closure within two hours of notification.

2) Incorrect, see A.2.

D. Correct - 1) Correct, see C.1.

2) Correct, see B.2.

K/A: G2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.(CFR: 41.12 / 43.4 / 45.10)

IMPORTANCE RO 3.4 SRO 3.8 Importance Rating: 3.4 3.8 Technical

Reference:

FNP-1-AOP-30, v19.0 FNP-1-STP-18.4, v39.0 References provided: None Learning Objective: RECALL AND APPLY the information from the required actions (section 3) and surveillance requirements (section 4) in accordance with rules of application requirements (Section 1) for any Technical Specifications or TRM requirements (OPS-62302A01): 10CFR55.43 (b) 2 Question History: New question Tuesday, March 19, 2013 11:38:43 AM 76

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 K/A match: Candidate is given conditions where a fuel handling accident has occurred. Knowledge of the radiation hazards is displayed by choosing the actions that should be taken to mitigate the radiation hazard.

SRO justification: 10 CFR 55.43(b)(2):

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

This question requires knowledge of the Surveillance Requirement of STP-18.4 Acceptance Criteria that states : A Maintenance Closure Response Team (MCRT) is available and briefed to effect closure within two hours of notification.

2013 NRC exam From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 flowchart for 10 CFR 55.43(b)(2)

1) can NOT be answered by knowing less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Tech Specs.

2) can NOT be answered by knowing information listed "above-the-line".

3) can NOT be answered by knowing the TS Safety Limits or their bases.

4) Does involve one or more of the following for TS, TRM or ODCM:

Application of Required Actions (Section 3) and Surveillance Requirements (SR)

(Section 4) in accordance with rules of application requirements (Section 1)..

Tuesday, March 19, 2013 11:38:43 AM 77

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08/18/12 13:17:30 UNIT 1 FNP-1-AOP-30.0 May 23, 2012 Version 19.0 FARLEY NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE FNP-1-AOP-30.0 S

A F

E T

Y REFUELING ACCIDENT R

E L

A T

E D

PROCEDURE LEVEL OF USE CLASSIFICATION PER NMP-AP-003 CATEGORY SECTIONS Continuous: ALL

Reference:

NONE Information: NONE Approved:

David L Reed (for)

Operations Manager Effective Date: June 4, 2012

08/18/12 13:17:30 FNP-1-AOP-30.0 UNIT 1 REFUELING ACCIDENT Version 19.0 1B Table of Contents PROCEDURE CONTAINS NUMBER OF PAGES Body 8 Attachment 1 1 A

A A 1 Page 1 of 1

08/18/12 13:17:30 FNP-1-AOP-30.0 UNIT 1 REFUELING ACCIDENT Version 19.0 1B A. Purpose This procedure provides actions for response to fuel handling accident or a loss of refueling cavity water level.

This procedure is applicable at all times.

B. Symptoms or Entry Conditions

1. This procedure is entered when a fuel handling accident causes damage to a fuel assembly in conjunction with a high radiation indication on any of the following:

[] R-2 CTMT 155 ft

[] R-5 SFP ROOM

[] R-24A(B) CTMT PURGE

[] R-25A(B) SPENT FUEL BLDG EXH

2. This procedure is entered when a dry storage activity causes damage to a fuel assembly in conjunction with a high radiation indication on radiation monitor R-5(SFP ROOM).

3. This procedure is entered when rapidly falling refueling cavity level is observed.

4. This procedure may be entered at the discretion of the Shift Supervisor when any abnormal fuel handling incident occurs.

1 Page 1 of 8

08/18/12 13:17:30 FNP-1-AOP-30.0 UNIT 1 REFUELING ACCIDENT Version 19.0 1B Step Action/Expected Response Response Not Obtained q

NOTE: x A high alarm from R-25A or B will automatically isolate fuel building HVAC and start penetration room filtration.

x The Refueling Supervisor, or Cask Supervisor for dry storage activities, shall have sole responsibility for evaluating placement of fuel in step 1.

x The Refueling Supervisor, or Cask Supervisor for dry storage activities, shall have sole responsibility for evacuation of the refueling or dry storage crew.

1

__ 1 Secure all Fuel Handling operations: 1 1.1 1.1 IF fuel movement in progress and 1.1 conditions permit, THEN place fuel assembly in a safe location by one of the following:

[ ] Return fuel assembly to assigned location in the core.

[ ] Place fuel assembly in upender with manipulator crane gripper disengaged and raised to the GRIPPER UP DISENGAGED position.

[ ] Place fuel assembly in the spent fuel rack.

[ ] Place fuel assembly in the multi-purpose canister (MPC) fuel basket.

[ ] Lower the MPC into the cask storage area or cask wash area, as appropriate.

1.2 1.2 Verify CTMT and spent fuel room 1.2 upenders - LOWERED.

1.3 1.3 Verify CONVEYER CONTROL switch 1.3 in ON at CONTAINMENT BUILDING panel.

1.4 1.4 Verify CARRIAGE AT PIT lamp is ON. 1.4 (155 ft, AUX BLDG spent fuel room control panel) 1.5 1.5 Close fuel transfer tube gate valve. 1.5 (155 ft, AUX BLDG spent fuel room) q Step 1 continued on next page

__Page Completed 7 ProcedureStepsMain Page 2 of 8

08/18/12 13:17:30 FNP-1-AOP-30.0 UNIT 1 REFUELING ACCIDENT Version 19.0 1B Step Action/Expected Response Response Not Obtained 1.6 q 1.6 Dispatch personnel to close all spent fuel 1.6 area fuel handling hatches.

1.7 1.7 IF required, 1.7 THEN have Maintenance personnel close the containment equipment hatch.

2

__ 2 [CA] While continuing with the remainder 2 of this procedure, perform ATTACHMENT 1, CONTROL ROOM ISOLATION WITH CREFS IN SERVICE to ensure control room isolation with CREFS placed in service.

CAUTION: Dangerously high dose rates will result from uncovering reactor internals.

3

__ 3 IF reactor internals are being moved, 3 THEN place them in a safe location.

q S

__Page Completed 7 ProcedureStepsMain Page 3 of 8

08/18/12 13:17:30 FNP-1-AOP-30.0 UNIT 1 REFUELING ACCIDENT Version 19.0 1B Step Action/Expected Response Response Not Obtained q

CAUTION: Rapidly falling refueling cavity level will affect the containment and the spent fuel room.

4

__ 4 Evacuate all non-essential personnel from 4 accident area.

5

__ 5 Verify all access doors to accident area - 5 CLOSED.

6

__ 6 Evaluate actuating the plant emergency 6 alarm.

7

__ 7 [CA] IF fuel damage has resulted in 7 elevated containment radiation levels, THEN perform the following.

7.1 7.1 Verify containment ventilation isolation. 7.1 7.1.1 7.1.1 Verify containment purge dampers - 7.1.1 CLOSED

[] 3197

[] 3198D

[] 3198C

[] 3196

[] 3198A

[] 3198B q

Step 7 continued on next page

__Page Completed 7 ProcedureStepsMain Page 4 of 8

08/18/12 13:17:30 FNP-1-AOP-30.0 UNIT 1 REFUELING ACCIDENT Version 19.0 1B Step Action/Expected Response Response Not Obtained 7.1.2 q 7.1.2 Verify containment mini purge 7.1.2 dampers - CLOSED.

CTMT PURGE DMPRS MINI-2866C & 2867C FULL-3198A & 3198D

[ ] 2866C

[ ] 2867C CTMT PURGE DMPRS MINI-2866D & 2867D ULL-3196 & 3197 BOTH-3198B & 3198C

[ ] 2866D

[ ] 2867D 7.1.3 7.1.3 Stop MINI PURGE SUPP/EXH FAN. 7.1.3 7.1.4 7.1.4 Stop or verify stopped CTMT PURGE 7.1.4 SUPP/EXH FANs 8

__ 8 [CA] IF SFP Ventilation tripped 8 This step is for starting PRF if the accident OR had been in the Spent Fuel Pool Room. The accident is in Containment, so PRF start is SFP EXH FAN SUCT DMPR not required. Adds plausibility to the HV3990A(B) closed, distracter.

THEN perform the following.

8.1 8.1 Verify PRF exhaust fan A(B) running. 8.1 8.2 8.2 Verify PRF recirculation fan A(B) 8.2 running.

8.3 8.3 Verify SFP to PRF FLTR UNIT 8.3 3538A(B) open.

9

__ 9 Check Refueling Cavity Level - STABLE. 9 Proceed to step 19.

q S

__Page Completed 7 ProcedureStepsMain Page 5 of 8

08/18/12 13:17:30 FNP-1-AOP-30.0 UNIT 1 REFUELING ACCIDENT Version 19.0 1B Step Action/Expected Response Response Not Obtained 10 q

__ 10 Monitor affected area radiation. 10

Monitor containment radiation.

[ ] R-2 CTMT 155 FT

[ ] R-24A CTMT PURGE

[ ] R-24B CTMT PURGE

Monitor spent fuel room radiation.

[ ] R-5 SPENT FUEL ROOM

[ ] R-25A SPENT FUEL BLDG EXH

[ ] R-25B SPENT FUEL BLDG EXH 11

__ 11 IF dry storage activities are in progress in 11 the SFP area and R-5, SFP ROOM, is in alarm, THEN notify Maintenance to initiate appropriate section of FNP-0-MP-112.1, DFS EQUIPMENT MALFUNCTION GUIDANCE.

12

__ 12 Evaluate event classification and 12 notification requirements using NMP-EP-110-GL01, FNP EALS - ICS, THRESHOLD VALUES AND BASIS, NMP-EP-110, EMERGENCY NOTIFICATIONS and FNP-0-EIP-8, NON-EMERGENCY NOTIFICATIONS.

NOTE: Entry into a previously evacuated area should be made using FNP-0-EIP-14, PERSONNEL MOVEMENT, RELOCATION, RE-ENTRY AND SITE EVACUATION.

13

__ 13 Evaluate the need for Chemistry to sample 13 refueling cavity and spent fuel pool for activity.

14

__ 14 Evaluate the need for Chemistry to sample 14 the spent fuel pool and reactor cavity for boron concentration.

q S

__Page Completed 7 ProcedureStepsMain Page 6 of 8

08/18/12 13:17:30 FNP-1-AOP-30.0 UNIT 1 REFUELING ACCIDENT Version 19.0 1B Step Action/Expected Response Response Not Obtained 15 q

__ 15 Evaluate placing spent fuel pool 15 purification system in service using FNP-1-SOP-54.0, SPENT FUEL PIT COOLING AND PURIFICATION SYSTEM.

16

__ 16 Evaluate placing containment pre-access 16 filtration system in service per FNP-1-SOP-12.2, CONTAINMENT PURGE AND PRE-ACESS FILTRATION SYSTEM.

17

__ 17 Consult Operations Manager to evaluate 17 further plant response.

18

__ 18 Go to procedure and step in effect. 18 NOTE: Efforts to determine source of inventory loss should be performed in parallel with makeup efforts below. Possible sources of inventory loss include damaged nozzle dams, reactor cavity seal or spent fuel pool liner, or RHR system malfunctions. Refer to FNP-1-AOP-12.0, RESIDUAL HEAT REMOVAL SYSTEM MALFUNCTION for RHR malfunctions.

19

__ 19 Place CTMT SUMP PUMPs Q1G21P019A 19 and Q1G21P019B in PULL-TO-LOCK.

(BOP) 20

__ 20 Align RHR system. 20 20.1 20.1 Verify only one RHR PUMP - STARTED 20.1 IN COOLDOWN ALIGNMENT.

20.2 20.2 [CA] Monitor running RHR pump for 20.2 evidence of cavitation.

20.3 20.3 Align remaining RHR pump to fill 20.3 Align one charging pump or the refueling refueling cavity from RWST using water purification pump to fill the refueling FNP-1-SOP-7.0, RESIDUAL HEAT cavity.

REMOVAL SYSTEM.

q S

__Page Completed 7 ProcedureStepsMain Page 7 of 8

08/18/12 13:17:30 FNP-1-AOP-30.0 UNIT 1 REFUELING ACCIDENT Version 19.0 1B Step Action/Expected Response Response Not Obtained 21 q

__ 21 Check RWST level - GREATER THAN 21 Consult Operations Manager to evaluate 3 ft. method for RWST makeup.

22

__ 22 Evaluate the need for Chemistry to sample 22 the spent fuel pool and reactor cavity for boron concentration.

23

__ 23 Evaluate event classification and 23 notification requirements using NMP-EP-110-GL01, FNP EALS - ICS, THRESHOLD VALUES AND BASIS, NMP-EP-110, EMERGENCY NOTIFICATIONS and FNP-0-EIP-8, NON-EMERGENCY NOTIFICATIONS.

24

__ 24 Consult Operations Manager to evaluate 24 further plant response.

25

__ 25 Go to procedure and step in effect. 25 q

-END-

__Page Completed 7 ProcedureStepsMain Page 8 of 8

08/18/12 13:17:30 FNP-1-AOP-30.0 UNIT 1 REFUELING ACCIDENT 1BVersion 19.0 ATTACHMENT 1 CONTROL ROOM ISOLATION WITH CREFS IN SERVICE q

CAUTION: This Attachment should be completed without delay to ensure continued control room habitability.

NOTE: A control room isolation should occur and the TSC HVAC System should shift to its emergency (recirculation) mode upon receipt of a high radiation alarm from R-35A or R-35B.

1

__ 1 Ensure a control room is isolated by closing or verifying closed the following dampers:

1.1 1.1 CONTROL RM UTILITY EXH This Attachment places CREFS in

[ ] HV-3628 service

[ ] HV-3629 1.2 1.2 COMPUTER RM HVAC SUPP

[ ] HV-3626

[ ] HV-3627 1.3 1.3 COMPUTER RM HVAC RTN

[ ] HV-3622

[ ] HV-3623 1.4 1.4 CONTROL RM HVAC SUPP

[ ] HV-3624

[ ] HV-3625 1.5 1.5 CONT RM EXH FAN INLET DMPRS

[ ] HV-3649A/B/C 2

__ 2 Place one of the redundant Control Room Emergency Filtration/Pressurization system (CREFS) in service per FNP-0-SOP-56.0, CONTROL ROOM HVAC SYSTEM.

3

__ 3 Stop or verify stopped the Control Room Utility Exhaust Fan.

4

__ 4 Verify the TSC HVAC System in the emergency (recirculation) mode of operation per FNP-0-SOP-56.1, TECHNICAL SUPPORT SYSTEM HVAC SYSTEM.

q

-END-

__Page Completed 1 Page 1 of 1

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

97. G2.3.6 097 Unit 1 is operating at 100% power with the following conditions:

A release of the #2 Waste Monitor Tank (WMT) is planned.

R-18, LIQ WASTE DISCH, is INOPERABLE.

Chemistry has taken two (2) independent samples of the #2 WMT and reports the activity is <1.4 x10-5 µCi/mL and is within the normal limits for a release.

Two Shift Radio-Chemists have verified the manual input for the computer generated release rate calculation.

Which one of the following completes the statement below?

A WMT release (1) permitted (2) per the ODCM.

(1) (2)

A. is NOT until R-18 is returned to service B. is NOT until the activity is lowered less than <1x10-7 µCi/mL C. IS but as a minimum two qualified plant personnel are required to verify the discharge lineup only D. IS but as a minimum two qualified plant personnel are required to verify the discharge lineup and an SRO is required to verify the entire release rate calculation A. Incorrect A release could be held until RE-18 is operable, and this may be chosen as the conservative action, but it is not required per the ODCM. The ODCM directs the ACTION shown in Table 2-1, which permit a release if all conditions can be completed.

Plausible: if a release was on-going when R-18 became inoperable then the release must be immediately stopped. (SOP-50.0, v66.0, P&L 3.3)

B. Incorrect The release limit of 1x10-7 µCi/mL is that for SGBD per action 29 when R-23B is INOPERABLE, not applicable to the R-18 release path, therefore a release is permitted at this activity. The activity is NOT too high for release assuming all the requirements of the ODCM are met.

Plausible: Two consecutive samples must be taken, and must be within release limits (the release limit provided exceeds the limit which is applicable only to SGBD Release path).

C. Correct This answer choice describes the actions required by ACTION 28 of Table 2-1 of the ODCM which states the effluent releases may continue Tuesday, March 19, 2013 11:38:44 AM 78

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 provided that prior to initiating the release:

a. "Two separate samples must be analyzed,

b. two independent qualified members of facility staff verify the discharge line valving and verify:

1) the manual portion of the computer input for the release rate calculations OR

2) the entire release rate calculations if performed manually.

A liquid release permit is required to be reviewed by the SSS per SOP-50.1 appendices 1&2 to verify the above requirements are satisfied.

D. Incorrect The second part is incorrect, the SRO is not required to verify the calculation. The ODCM requires that the release rate calculations are verified by TWO qualified facility personnel, and the discharge flowpath also verified by TWO qualified facility personnel; NEITHER requires an SRO or licensed operator and both are normally not performed by an SRO or licensed operator. The System operators normally complete the valve lineup and verification while the Chemists normally complete the calculations and verifications. Further, the "entire release rate calculation" is not "REQUIRED" unless the calculation is done manually, but again is not REQUIRED to be performed by an SRO.

All actions of ACTION 28 of the ODCM have been completed except the release lineup has to be verified.

Plausible: It could be assumed that the SRO has to perform the verification of the release rate calculations. This is not the case, the ODCM requires two technically qualified personnel. There is no requirement for one of those to be an SRO.

K/A: G2.3.6 Ability to approve release permits.

(CFR: 41.13 / 43.4 / 45.10)

IMPORTANCE RO 2.0 SRO 3.8 Importance Rating: 2.0 3.8 Technical

Reference:

FNP-ODCM, v24 References provided: None Learning Objective: OPS-62106A01RECALL AND APPLY the information from the LCO BASES sections: BACKGROUND, APPLICABLE SAFETY ANALYSIS, ACTIONS, SURVEILLENCE REQUIREMENTS, for any Technical Specifications or TRM requirements associated with Liquid Tuesday, March 19, 2013 11:38:44 AM 79

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 and Solid Waste System components and attendant equipment alignment, to include the following

5.5.1, Offsite Dose Calculation Manual (ODCM)

Question History: 2011 HLT-34 NRC Exam question # 97 This is one of the previous 2 Farley NRC Exams.

K/A match: The SRO must approve a release if ODCM actions are required prior to the release. For a normal release the Chemistry Department provides the final approval of the permit, but the affected unit's SS (SRO) must also provide approval for the release. This question provides a scenario in which the release cannot be allowed until ODCM actions are accomplished as determined by the SRO, and the applicant must recall what completed actions allow approving the release.

SRO justification: 10CFR55.43 (b) (2)

Some examples of SRO exam items for this topic include:

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

Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).

2013 NRC exam 10 CFR 55.43(b)(2)

Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 flowchart:

1) can NOT be answered by knowing less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Tech Specs.

2) can NOT be answered by knowing information listed "above-the-line".

3) can NOT be answered by knowing the TS Safety Limits or their bases.

4) Does involve one or more of the following for TS, TRM or ODCM:

Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1). (Required actions for INOPERABLE R-18 rad monitor have to be implemented per the ODCM.)

Tuesday, March 19, 2013 11:38:44 AM 80

FNP-ODCM CHAPTER 2 LIQUID EFFLUENTS 2.1 LIMITS OF OPERATION The following Liquid Effluent Controls implement requirements established by Technical Specifications Section 5.0. Terms printed in all capital letters are defined in Chapter 10.

2.1.1 Liquid Effluent Monitoring Instrumentation Control In accordance with Technical Specification 5.5.4.a, the radioactive liquid effluent monitoring instrumentation channels shown in Table 2-1 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits specified in Section 2.1.2 are not exceeded. The alarm/trip setpoints of these channels shall be determined in accordance with Section 2.3.

2.1.1.1 Applicability This limit applies at all times.

2.1.1.2 Actions With a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above control, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, declare the channel inoperable, or change the setpoint to a conservative value.

With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 2-1. Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful, explain in the next Radioactive Effluent Release Report pursuant to Section 7.2 why this inoperability was not corrected in a timely manner.

This control does not affect shutdown requirements or MODE changes.

2.1.1.3 Surveillance Requirements Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL OPERATIONAL TEST (COT) operations at the frequencies shown in Table 2-2.

2-1 Version 24 01/10

FNP-ODCM 2.1.1.4 Basis The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The Alarm/Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in Section 2.3 to ensure that the alarm/trip will occur prior to exceeding the limits of Section 2.1.2. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

2-2 Version 24 01/10

FNP-ODCM Table 2-1 Radioactive Liquid Effluent Monitoring Instrumentation OPERABILITY Requirementsa Instrument Minimum Channels OPERABLE ACTION

1. Gross Radioactivity Monitors Providing Automatic Termination of Release

a. Liquid Radwaste Effluent Line (RE-18) 1 28

b. Steam Generator Blowdown Effluent Line (RE-23B) 1 29

2. Flowrate Measurement Devices

a. Liquid Radwaste Effluent Line

1) Waste Monitor Tank No. 1 1 30

2) Waste Monitor Tank No. 2 1 30

b. Discharge Canal Dilution Line 1 30 (Service Water)

c. Steam Generator Blowdown Effluent Line 1 30

a. All requirements in this table apply to each unit.

2-3 Version 24 01/10

FNP-ODCM Table 2-1 (contd) Notation for Table 2 ACTION Statements ACTION 28 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue provided that prior to initiating a release:

a. At least two independent samples are analyzed in accordance with Section 2.1.2.3, and

b. At least two technically qualified members of the Facility Staff independently verify the discharge line valving and (1) Verify the manual portion of the computer input for the release rate calculations performed on the computer, or (2) Verify the entire release rate calculations if such calculations are performed manually.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 29 - 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 analyzed for gross radioactivity (beta or gamma) at a MINIMUM DETECTABLE CONCENTRATION no higher than 1 u 10-7 PCi/mL.

a. At least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when the specific activity of the secondary coolant is greater than 0.01 PCi/gram DOSE EQUIVALENT I-131.

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the specific activity of the secondary coolant is less than or equal to 0.01 PCi/gram DOSE EQUIVALENT I-131.

ACTION 30 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue, provided that the flowrate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases.

Pump curves may be used to estimate flow.

2-4 Version 24 01/10

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

98. G2.4.6 098 Unit 1 tripped from 100% power with the following conditions:

A Safety Injection was actuated due to a SG Tube Rupture on the 1B SG.

The crew is performing EEP-3.0, Steam Generator Tube Rupture.

The following results are reported:

HV3369B, 1B SG MSIV, will not close.

HV3370B, 1B SG MSIV, will not close.

1B SG pressure is 950 psig.

Which one of the following completes the statements below?

The Control Room crew is required to (1) .

The recovery strategy for the selected procedure is to (2) .

Procedure titles are as follows:

EEP-3.0, Steam Generator Tube Rupture ECP-3.1, SGTR with Loss of Reactor Coolant Subcooled Recovery Desired (1) (2)

A. continue in EEP-3.0 complete an RCS cooldown first; then perform an RCS depressurization B. transition to ECP-3.1 complete an RCS cooldown first; then perform an RCS depressurization C. continue in EEP-3.0 start an RCS cooldown and then perform an RCS depressurization during the cooldown D. transition to ECP-3.1 start an RCS cooldown and then perform an RCS depressurization during the cooldown Tuesday, March 19, 2013 11:38:44 AM 81

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 A. Incorrect - 1) Incorrect, plausible since SG pressure is at normal pressure. There is a transition point in EEP-3.0 that sends you to ECP-3.1 if SG pressure is

<250 psig. A candidate may think that since SG pressure is higher than 250 psig, actions will continue in EEP-3.0 (step 5 of EEP-3.0).

2) Incorrect, plausible since this is the normal strategy for actions in EEP-3.0, but EEP-3.0 is not the correct procedure for performing the actions.

B. Incorrect - 1) Correct, per EEP-3.0, if at least one MSIV is not closed in the ruptured SG, go to ECP-3.1 (step 3.7 of EEP-3.0).

2) Incorrect, see A.2.

C. Incorrect - 1) Incorrect, see A.1.

2) Correct, this is the strategy for ECP-3.1..

D. Correct - 1) Correct, see B.1.

2) Correct, see C.2.

K/A: G2.4.6 Knowledge of EOP mitigation strategies.

(CFR: 41.10 / 43.5 / 45.13)

IMPORTANCE RO 3.7 SRO 4.7 Importance Rating: 3.7 4.7 Technical

Reference:

FNP-1-EEP-3.0 v27 References provided: None Learning Objective: ASSESS the facility conditions associated with the EEP-3, Steam Generator Tube Rupture and based on that assessment: (OPS-62530D01)

SELECT the appropriate procedures during normal, abnormal and emergency situations. 10CFR55.43 (b) 5

DETERMINE if transition to another section of the procedure or to another procedure is required

DETERMINE if the critical safety functions are satisfied Question History: New question K/A match: The second part of the question deals with knowledge of EOP mitigation strategies - specifically how the cooldown is performed and how the depressurization is performed. It is different between EEP-3.0 and ECP-3.1.

Tuesday, March 19, 2013 11:38:44 AM 82

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 SRO justification: 10 CFR 55.43(b)(5)

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

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

Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures.

2013 NRC exam From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 flowchart for 10 CFR 55.43(b)(5) :

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations [10 CFR 55.43(b)(5)], involving BOTH:

1) assessing plant conditions and then

2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

Using the flowchart, this question can:

NOT be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location.

NOT be answered solely by knowing immediate operator actions.

NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs.

NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure. (the second part of the question is major mitigative strategy and meets the K/A. The first part of the question meets the SRO requirement for procedure selection)

Be answered with knowledge of ONE or MORE of the following:

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

Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures.

Tuesday, March 19, 2013 11:38:44 AM 83

1/22/2013 14:18 FNP-1-EEP-3 UNIT 1 STEAM GENERATOR TUBE RUPTURE Revision 27 Step Action/Expected Response Response NOT Obtained 3.7 Verify at least one SG main 3.7 Perform the following.

steam isolation and bypass valve for ruptured SG(s) - 3.7.1 Place associated test CLOSED. switch to TEST position.

>>¥¥¥¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥£¥¥¥¥¥¥¥£¥¥¥¥¥¥¥ >>¥¥¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥ Ruptured SG 1A 1B 1C Ruptured SG 1A 1B 1C 1A(1B,1C) SG 1A(1B,1C) SGSG MSIV - TRIP MSIV - TEST Q1N11HV []3369A []3369B

[]3369A []3369C

[]3369B

[]3369C Q1N11HV []3369A/

[]3369A/[]3369B/ []3369C/

[]3369B/

[]3369C/

[]3370A []3370B

[]3370A []3370C

[]3370B

[]3370C 70A 70B 70C

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ ¥¥¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥º 1A(1B,1C)SG MSIV - BYPASS BYPASS 3.7.2 IF at least one main steam Q1N11HV []3368A []3368B

[]3368A []3368C

[]3368B

[]3368C isolation and one bypass

[]3976A []3976B

[]3976A []3976C

[]3976B

[]3976C valve for ruptured SG

¥¥¥¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥º closed, THEN proceed to step 4 IF NOT go to FNP-1-ECP-3.1, SGTR WITH LOSS OF REACTOR COOLANT SUBCOOLED RECOVERY DESIRED.

CAUTION CAUTION:: [CA] To prevent excessive RCS cooldown, AFW flow to any ruptured SG that is also faulted, should remain isolated during subsequent recovery actions unless the SG is needed for RCS cooldown.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ NOTE: [CA] Maintaining ruptured SG(s) narrow range level greater than 31%{48%} prevents SG depressurization during RCS cooldown.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ 4 [CA] WHEN ruptured SG(s) narrow range level greater than 31%{48%}, The first two pages are related to part one THEN perform the following. of the question.

Step 4 continued on next page.

Page 10 of 54

1/22/2013 14:18 FNP-1-EEP-3 UNIT 1 STEAM GENERATOR TUBE RUPTURE Revision 27 Step Action/Expected Response Response NOT Obtained 4.1 [CA] Isolate AFW flow to 4.1 Perform the following.

ruptured SG(s) using FCVs.

4.1.1 If AFW flow from MDAFWP(s)

>>¥¥¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥£¥¥¥¥¥¥£¥¥¥¥¥¥¥ NOT isolated, Ruptured SG 1A 1B 1C THEN isolate flow to MDAFWP TO ruptured SG(s) using MDAFWP 1A(1B,1C) SG []

SG [] [] isolation valve(s).

Q1N23HV 3227A 3227B 3227C in MOD in MOD MOD in MOD MOD >>¥¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥£¥¥¥¥¥¥¥£¥¥¥¥¥¥¥ Ruptured SG SG 1A 1B 1C MDAFWP TO MDAFWP TO 1A(1B,1C) SG SG 1A(1B,1C)

FLOW CONT [] [] [] SG ISO HIC 3227AA 3227BA 3227AA 3227CA 3227BA Q1N23MOV []3764A

[]3764A[]3764D

[]3764D[]3764F

[]3764F closed closed closed closed closed OR OR OR

[]3764E

[]3764E[]3764B

[]3764B[]3764C

[]3764C TDAFWP TO ¥¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥º 1A(1B,1C) SG []

SG [] []

Q1N23HV 3228A 3228B 3228C 4.1.2 If AFW flow from TDAFWP(s) in MOD in MOD MOD in MOD MOD NOT isolated, THEN isolate flow to TDAFWP TO ruptured SG(s) using AFW 1A(1B,1C) SG SG stop valve(s).

FLOW CONT [] [] []

HIC 3228AA 3228BA 3228AA 3228CA 3228BA >>¥¥¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥£¥¥¥¥¥¥£¥¥¥¥¥¥ closed closed closed closed closed Ruptured SG 1A 1B 1C

¥¥¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¢¥¥¥¥¥¥¢¥¥¥¥¥¥¥º AFW TO 1A(1B,1C) SG [] BKR SG BKR[] BKR BKR[] BKR BKR STOP VLV FU-U4 FU-U5 FU-I2 supply bkr (key (key (key Q1R17BKR Z-81) Z-82) Z-83)

AFW TO 1A(1B,1C) [] [] []

SG STOP VLV 3350A 3350B 3350C Q1N23MOV Closed ClosedClosed ClosedClosed Closed

¥¥¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¢¥¥¥¥¥¥¢¥¥¥¥¥¥º CAUTION CAUTION: : Major steam flow paths from the ruptured SG(s) (i.e. MSIVs and bypasses, TDAFW steam supply and SG atmospheric relief valve) should be isolated before initiating cooldown 5 Check ruptured SG(s) pressure - 5 Go to FNP-1-ECP-3.1, SGTR WITH GREATER THAN 250 psig. LOSS OF REACTOR COOLANT SUBCOOLED RECOVERY DESIRED.

Plausibility for remaining in EEP-3, SG pressure in the stem is 950 psig.

Page 11 of 54

1/22/2013 14:18 FNP-1-EEP-3 UNIT 1 STEAM GENERATOR TUBE RUPTURE Revision 27 Step Action/Expected Response Response NOT Obtained CAUTION CAUTION: : [CA] To ensure proper SI flow to the reactor, the RHR pumps must be manually restarted if they are secured and RCS pressure falls below 275 psig{435 psig}.

The rest of the pages are related to part two of the question.

CAUTION CAUTION: : Pump damage may occur if RHR pumps are operated on miniflow for longer than three hours with no CCW supplied to the RHR heat exchangers.

13 [CA] Check if LHSI Pumps should be stopped.

13.1 RHR pumps - ANY RUNNING WITH 13.1 Proceed to step 14.

SUCTION ALIGNED TO RWST.

13.2 Check RCS pressure - GREATER 13.2 Perform the following.

THAN 275 psig{435 psig}

13.2.1 Establish CCW flow to RHR 1C(1A) LOOP heat exchangers.

RCS NR PRESS CCW TO

[] PI 402B 2A(2B) RHR HX

[] PI 403B [] Q1P17MOV3185A open

[] Q1P17MOV3185B open 13.2.2 Proceed to Step 14.

13.3 Stop any RHR Pumps running with suction from the RWST.

14 Check if Cooldown should be stopped.

14.1 Check hottest core exit T/Cs - 14.1 Do NOT proceed until hottest LESS THAN REQUIRED core exit T/Cs - LESS THAN TEMPERATURE. REQUIRED TEMPERATURE.

14.2 Stop RCS cooldown 14.3 Maintain core exit T/Cs - LESS Once the pre-determined Core THAN REQUIRED TEMPERATURE. Exit temperature is reached during the cooldown, the cooldown is stopped.

Page 22 of 54

1/22/2013 14:18 FNP-1-EEP-3 UNIT 1 STEAM GENERATOR TUBE RUPTURE Revision 27 Step Action/Expected Response Response NOT Obtained 15 Check ruptured SG(s) pressure - 15 Perform the following.

STABLE OR RISING.

15.1 Maintain RCS cold legs cooldown rate less than 100 F 100 in any 60 minute period.

15.2 Dump steam from intact SGs to maintain ruptured SG pressure at least 250 psig above pressure of intact SGs used for cooldown.

15.3 IF ruptured SG(s) pressure NOT maintained greater than 250 psig above pressure of intact SGs used for cooldown, THEN go to FNP-1-ECP-3.1, SGTR WITH LOSS OF REACTOR COOLANT SUBCOOLED RECOVERY DESIRED.

16 Check SUB COOLED MARGIN MONITOR 16 Go to FNP-1-ECP-3.1, SGTR WITH indication - GREATER THAN LOSS OF REACTOR COOLANT 36 36F{65 F{65F} SUBCOOLED IN CETC SUBCOOLED RECOVERY DESIRED.

MODE.

17 Reduce RCS pressure to minimize break flow and refill pressurizer.

17.1 IF normal pressurizer spray 17.1 Proceed to Step 18. OBSERVE available, CAUTIONS AND NOTES PRIOR TO THEN open all available normal STEP 18.

pressurizer spray valves.

1A(1B) LOOP SPRAY VLV Now the RCS

[] PK 444C depressurization is

[] PK 444D performed.

Step 17 continued on next page.

Page 23 of 54

1/22/2013 14:18 FNP-1-EEP-3 UNIT 1 STEAM GENERATOR TUBE RUPTURE Revision 27 Step Action/Expected Response Response NOT Obtained 17.2 Reduce RCS pressure with maximum available spray until ANY of the following conditions occur, then stop RCS pressure reduction.

RCS pressure less than ruptured SG(s) pressure AND pressurizer level greater than 13%{43%}

OR RCS pressure within 300 PSI of ruptured SG(s) pressure AND pressurizer level greater than 43%{50%}

OR Pressurizer level greater than 73%{66%}

OR SUBCOOLED MARGIN MONITOR indication less than 16 16F{45 F{45F} subcooled in CETC mode.

17.2.1 Verify both normal 17.2.1 Stop 1A AND 1B RCPs to stop pressurizer spray valves - spray flow.

CLOSED.

RCP 1A(1B) LOOP [] 1A SPRAY VLV [] 1B

[] PK 444C

[] PK 444D 17.2.2 Verify auxiliary spray 17.2.2 Isolate auxiliary spray valve - CLOSED. line.

RCS PRZR CHG PMPS TO AUX SPRAY REGENERATIVE HX

[] Q1E21HV8145 [] Q1E21MOV8107 closed

[] Q1E21MOV8108 closed 17.2.3 Proceed to Step 20.

OBSERVE CAUTION PRIOR TO STEP 20.

End of EEP-3.0 info, the next several pages are from ECP-3.1 Page 24 of 54

1/22/2013 14:15 FNP-1-ECP-3.1 UNIT 1 SGTR WITH LOSS OF REACTOR COOLANT SUBCOOLED RECOVERY Revision 23 DESIRED Step Action/Expected Response Response NOT Obtained 12.2.2 Control TDAFWP flow.

TDAFWP FCV 3228

[] RESET reset TDAFWP SPEED CONT

[] SIC 3405 adjusted

>>¥¥¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥ Intact SG 1A 1B 1C TDAFWP TO 1A(1B,1C) SG SG Q1N23HV []3228A []3228B []3228C in MOD in MOD in MOD TDAFWP TO 1A(1B,1C) SG SG FLOW CONT HIC []3228AA []3228BA

[]3228AA []3228CA

[]3228BA []3228CA adjusted adjusted adjusted adjusted adjusted adjusted

¥¥¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥º

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ NOTE: [CA] Comparison of Curve 61 and/or 61A with existing RCS boron concentration should be performed to verify adequate shutdown margin during cooldown to cold shutdown.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ 13 Begin RCS cooldown to cold shutdown. An RCS cooldown to cold 13.1 WHEN P-12 light lit (543 F),

(543 shutdown is initiated and THEN perform the following. maintained.

13.1.1 Block low steam line pressure SI.

STM LINE PRESS SI BLOCK - RESET

[] A TRN to BLOCK

[] B TRN to BLOCK Step 13 continued on next page.

Page 11 of 61

1/22/2013 14:15 FNP-1-ECP-3.1 UNIT 1 SGTR WITH LOSS OF REACTOR COOLANT SUBCOOLED RECOVERY Revision 23 DESIRED Step Action/Expected Response Response NOT Obtained 13.1.2 Verify blocked indication.

BYP & PERMISSIVE STM LINE ISOL.

SAFETY INJ.

[] TRAIN A BLOCKED light lit

[] TRAIN B BLOCKED light lit 13.1.3 Bypass the steam dump interlock.

STM DUMP INTERLOCK

[] A TRN to BYP INTLK

[] B TRN to BYP INTLK 13.1.4 Adjust steam header pressure controller to control cooldown rate.

STM HDR PRESS

[] PK 464 adjusted 13.2 [CA] Maintain RCS cold legs cooldown rate - LESS THAN 100 100F IN ANY 60 MINUTE PERIOD.

RCS COLD LEG TEMP

[] TR 410 13.3 IF RHR system is in service in shutdown cooling alignment per FNP-1-SOP-7.0, RESIDUAL HEAT REMOVAL SYSTEM, THEN use RHR to cool the RCS.

Step 13 continued on next page.

Page 12 of 61

1/22/2013 14:15 FNP-1-ECP-3.1 UNIT 1 SGTR WITH LOSS OF REACTOR COOLANT SUBCOOLED RECOVERY Revision 23 DESIRED Step Action/Expected Response Response NOT Obtained

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ NOTE: The steam dumps will be interlocked closed when RCS TAVG reaches P-12 (543 (543F). This interlock may be bypassed for A and E steam dumps with the STM DUMP INTERLOCK switches.

Excessive opening of steam dumps can cause a high steam flow LO-LO TAVG main steam line isolation signal.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ 13.4 IF condenser available, 13.4 Dump steam to atmosphere.

THEN dump steam to condenser from intact SGs. 13.4.1 Direct counting room to perform FNP-0-CCP-645, MAIN BYP & PERMISSIVE STEAM ABNORMAL COND ENVIRONMENTAL RELEASE.

AVAIL

[] C-9 status light lit 13.4.2 IF normal air available, THEN control atmospheric STM DUMP relief valves to dump steam

[] MODE SEL A-B TRN in STM PRESS from intact SGs, IF NOT NOT,, dump steam using STM DUMP FNP-1-SOP-62.0, EMERGENCY INTERLOCK AIR SYSTEM.

[] A TRN in ON

[] B TRN in ON 1A(1B,1C) MS ATMOS REL VLV STM HDR [] PC 3371A adjusted PRESS [] PC 3371B adjusted

[] PK 464 adjusted [] PC 3371C adjusted 13.4.3 IF no intact SG available, AND RHR system NOT in service, THEN dump steam from a faulted SG.

13.4.4 IF no intact or faulted SG available, AND RHR system NOT in service, THEN dump steam from a ruptured SG.

13.5 IF RHR system is NOT in service, THEN begin preparation of RHR system for cooldown using FNP-1-SOP-7.0, RESIDUAL HEAT REMOVAL SYSTEM.

Page 13 of 61

1/22/2013 14:15 FNP-1-ECP-3.1 UNIT 1 SGTR WITH LOSS OF REACTOR COOLANT SUBCOOLED RECOVERY Revision 23 DESIRED Step Action/Expected Response Response NOT Obtained

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ NOTE: Step 14 is a continuing action. FNP-1-ECP-3.2, SGTR WITH LOSS OF REACTOR COOLANT SATURATED RECOVERY DESIRED, transition criteria specified in the RNO apply anytime during the cooldown to cold shutdown.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ 14 [CA] Check subcooled recovery allowed.

14.1 [CA] Check RWST level - 14.1 IF CTMT SUMP LVL LI 3594A less GREATER THAN 25 ft. than TABLE 1 requirements, THEN go to FNP-1-ECP-3.2, SGTR RWST WITH LOSS OF REACTOR COOLANT LVL SATURATED RECOVERY DESIRED.

[] LI 4075A

[] LI 4075B 14.2 [CA] Check ruptured SG narrow 14.2 Consult TSC staff to evaluate range level - LESS THAN need for transition to 91%{76%}. FNP-1-ECP-3.2, SGTR WITH LOSS OF REACTOR COOLANT SATURATED RECOVERY DESIRED.

15 Check SUB COOLED MARGIN MONITOR 15 Proceed to step 29.

indication - GREATER THAN 16 16F{45 F{45F} SUBCOOLED IN CETC MODE.

Page 14 of 61

1/22/2013 14:15 FNP-1-ECP-3.1 UNIT 1 SGTR WITH LOSS OF REACTOR COOLANT SUBCOOLED RECOVERY Revision 23 DESIRED Step Action/Expected Response Response NOT Obtained 16 Check SI in service. 16 Proceed to Step 24.

Check HHSI flow - GREATER THAN 0 gpm.

A TRN HHSI FLOW

[] FI 943 OR Check any RHR PUMP - STARTED IN SI MODE.

>>¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥¥ Started RHR PUMP 1A 1B RWST TO 1A(1B) RHR PUMP Q1E11MOV [] 8809A [] 8809B open open 1C(1A) RCS LOOP TO 1A(1B) RHR PUMP PUMP Q1E11MOV [] 8701A [] 8702A closed closed closed closed

[] 8701B [] 8702B closed closed closed closed

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¥º Page 15 of 61

1/22/2013 14:15 FNP-1-ECP-3.1 UNIT 1 SGTR WITH LOSS OF REACTOR COOLANT SUBCOOLED RECOVERY Revision 23 DESIRED Step Action/Expected Response Response NOT Obtained CAUTION CAUTION:: To prevent pressurizer PORV failure, cycling of pressurizer PORVs should be minimized.

CAUTION CAUTION:: The PRT may rupture causing abnormal containment conditions while using pressurizer PORVs.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ NOTE: Reactor vessel steam voiding may occur during pressure reduction while on natural circulation. This will cause a rapid rise in pressurizer level.

As the RCS cooldown

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ continues, an RCS 17 Reduce RCS pressure to refill depressurization is initiated.

pressurizer.

17.1 IF normal pressurizer spray 17.1 IF any PRZR PORV available, available, THEN open only one PRZR PORV, THEN open all available normal IF NOT NOT, ,

pressurizer spray valves. THEN proceed to Step 18.

1A(1B) LOOP SPRAY VLV

[] PK 444C

[] PK 444D 17.2 Check pressurizer level - 17.2 Perform the following.

GREATER THAN 25%{50%}.

17.2.1 WHEN pressurizer level greater than 25%{50%},

THEN perform Step 17.3.

17.2.2 Proceed to Step 18.

OBSERVE NOTE PRIOR TO STEP 18.

Step 17 continued on next page.

Page 16 of 61

1/22/2013 14:15 FNP-1-ECP-3.1 UNIT 1 SGTR WITH LOSS OF REACTOR COOLANT SUBCOOLED RECOVERY Revision 23 DESIRED Step Action/Expected Response Response NOT Obtained

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ NOTE: Pressurizer level may fall when SI flow is reduced in subsequent steps. Establishing a pressurizer level greater than 25%{50%} will minimize returning to this step to refill the pressurizer.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ 17.3 Stop RCS pressure reduction.

17.3.1 Verify both normal 17.3.1 Stop RCP in loop with pressurizer spray valves - normal pressurizer spray CLOSED. valve open.

1A(1B) LOOP >>¥¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥¥ SPRAY VLV 1A LOOP 1B LOOP

[] PK 444C SPRAY VLV VLVSPRAY VLV VLV

[] PK 444D Spray Valve Valve PK 444C PK 444D

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ RCP [] 1A [] 1B

¥¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¥º 17.3.2 Verify both PRZR PORVs - 17.3.2 Close PRZR PORV ISO for any CLOSED. open PRZR PORV.

Page 17 of 61

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6

99. WE02EA2.1 099 Unit 1 has tripped from 100% power with the following conditions:

A Safety Injection occurred after the Reactor Trip due to a Steam Dump malfunction.

All MSIV's are closed.

The crew is evaluating SI termination criteria in EEP-0.0, Reactor Trip or Safety Injection.

The SCMMs indicate 105°F in CETC mode.

All SG Narrow Range levels are 43% and rising.

RCS pressure is 1820 psig and slowly rising.

Pressurizer level is 5% and rising.

Which one of the following describes the required actions regarding SI Termination?

A. Immediately transition to ESP-1.1, SI Termination, to terminate SI flow.

B. Maintain SI flow until PZR level recovers, then transition to ESP-1.2, Post LOCA Cooldown and Depressurization.

C. Maintain SI flow until PZR level recovers, then terminate SI flow in EEP-0.0, Reactor Trip or Safety Injection.

D. Immediately terminate SI flow in EEP-0, Reactor Trip or Safety Injection, then transition to ESP-1.1, SI Termination.

EEP-0 Termination step 16 Check SI termination criteria.

16.1 Check SUB COOLED MARGIN 16.1 Proceed to step 23.

MONITOR indication - GREATER THAN 16 F SUBCOOLED IN CETC MODE.

16.2 Check secondary heat sink 16.2 IF neither condition is available. satisfied, Total feed flow to SGs - THEN proceed to step 23.

GREATER THAN 395 gpm.

AFW FLOW TO 1A(1B,1C) SG

[] FI 3229A

[] FI 3229B

[] FI 3229C AFW TOTAL FLOW

[] FI 3229 OR Narrow range level in at least one SG - GREATER THAN Tuesday, March 19, 2013 11:38:44 AM 85

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 31%.

16.3 Check RCS pressure - STABLE OR 16.3 Proceed to step 23.

RISING.

1C(1A) LOOP RCS WR PRESS

[] PI 402A

[] PI 403A 16.4 Check Pressurizer level - 16.4 Perform the following.

GREATER THAN 13%.

16.4.1 Stabilize RCS pressure to allow Pressurizer level to rise.

1A(1B) LOOP SPRAY VLV

[] PK 444C adjusted

[] PK 444D adjusted 16.4.2 Return to step 16.1.

A. Incorrect - In ESP-0.1, as long as you are greater than 4% Pressurizer level, a SI is not required. Plausible since normally there is a transition to ESP-1.1 when SI termination criteria is met in other procedures, and normally SI flow is terminated in ESP-1.1. For the given conditions, ESP-1.1 will be implemented, but after SI flow is actually terminated in EEP-0.

B. Incorrect - Plausible since normally when SI termination criteria is not met, there is a transition to ESP-1.2. In EEP-0, step 21 normal charging flow is being established. At step 21.2.3 RNO a transition to ESP-1.2 exists for the case where Pzr level cannot be maintained 25-50%, then HHSI flow is realigned and ESP-1.2 entered.

For the given conditions, SI termination is not met due to Pressurizer level being < 13%. EEP-0 provides an exception to allow Pressurizer level to rise until all SI termination criteria is met, and then terminate SI.

C. Correct - PZR level is less than the required 13%. The RNO action for step 16.4 is to stabilize pressure and allow level to recover (maintain SI flow until level recovers to >13%). SI Termination Criteria will be met in EEP-0 so SI flow will be terminated in this procedure.

D. Incorrect - PZR level is < the required 13%; SI termination criteria is not yet satisfied, so it is incorrect to immediately terminate SI. EEP-0 provides a loop back to allow time for SI flow to refill the Pressurizer, and then if all SI termination criteria is met, terminate SI flow in EEP-0, and transition to ESP-1.1.

Tuesday, March 19, 2013 11:38:44 AM 86

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 Ability to determine and interpret the following as they apply to the (SI Termination)

(CFR: 43.5 / 45.13)

EA2.1 Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

IMPORTANCE RO 3.3 SRO 4.2 Importance Rating: 3.3 4.2 Technical

Reference:

FNP-1-EEP-0.0, v44.0 FNP-1-ESP-1.1, v25 References provided: None Learning Objective: ASSESS the facility conditions associated with the (1)

EEP-0, Reactor Trip or Safety Injection and (2) ESP-0.0, Rediagnosis and based on that assessment:

SELECT the appropriate procedures during normal, abnormal and emergency situations. 10CFR55.43 (b) 5 DETERMINE if transition to another section of the procedure or to another procedure is required DETERMINE if the critical safety functions are satisfied (OPS-62530A01)

Question History: FNP Bank - 006A2.13 01 Same as Harris 2009 NRC Exam K/A match: The question provides information from plant instrumentation on SI termination criteria while in EEP-0.

The candidate has to interpret the information and determine the actions required and the procedure required to perform SI termination actions.

SRO justification: 10 CFR 55.43(b)(5)

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

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

Tuesday, March 19, 2013 11:38:44 AM 87

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures.

2013 NRC exam From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 flowchart for 10 CFR 55.43(b)(5) :

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations [10 CFR 55.43(b)(5)], involving BOTH:

1) assessing plant conditions and then

2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

Using the flowchart, this question can:

NOT be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location.

NOT be answered solely by knowing immediate operator actions.

NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs.

NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

Be answered with knowledge of ONE or MORE of the following:

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

Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.

Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures.

Tuesday, March 19, 2013 11:38:44 AM 88

12/17/2012 13:16 FNP-1-EEP-0 UNIT 1 REACTOR TRIP OR SAFETY INJECTION Revision 44.0 Step Action/Expected Response Response NOT Obtained 16 Check SI termination criteria.

16.1 Check SUB COOLED MARGIN 16.1 Proceed to step 23.

MONITOR indication - GREATER THAN 16 F SUBCOOLED IN CETC 16 MODE.

16.2 Check secondary heat sink 16.2 IF neither condition is available. satisfied, THEN proceed to step 23.

Total feed flow to SGs -

GREATER THAN 395 gpm.

AFW FLOW TO 1A(1B,1C) SG

[] FI 3229A

[] FI 3229B

[] FI 3229C AFW TOTAL FLOW

[] FI 3229 OR Narrow range level in at least one SG - GREATER THAN 31%.

16.3 Check RCS pressure - STABLE OR 16.3 Proceed to step 23.

RISING.

1C(1A) LOOP RCS WR PRESS

[] PI 402A

[] PI 403A 16.4 Check pressurizer level - 16.4 Perform the following.

GREATER THAN 13%.

16.4.1 Stabilize RCS pressure to allow pressurizer level to rise.

A continuous loop will be 1A(1B) LOOP performed here until Pressurizer SPRAY VLV

[] PK 444C adjusted level is > 13%. [] PK 444D adjusted 16.4.2 Return to step 16.1.

Page 22 of 35

12/17/2012 13:16 FNP-1-EEP-0 UNIT 1 REACTOR TRIP OR SAFETY INJECTION Revision 44.0 Step Action/Expected Response Response NOT Obtained CAUTION CAUTION:: Automatic reinitiation of SI will not occur since reactor trip breakers are not reset.

17 Verify SI reset. 17 Perform the following:

[] MLB-1 1-1 not lit (A TRN) 17.1 IF any train will NOT reset

[] MLB-1 11-1 not lit (B TRN) using the MCB SI RESET pushbuttons, THEN place the affected train S821 RESET switch to RESET.

(SSPS TEST CAB.)

17.2 IF a failure exists in SSPS such that SI cannot be reset, THEN reset SI using FNP-1-SOP-40.0, RESPONSE TO INADVERTENT SI AND INABILITY TO RESET OR BLOCK SI, Appendix 2.

18 Stop all but one CHG PUMP.

19 Verify RCS pressure - STABLE OR 19 Go to FNP-1-ESP-1.2, POST LOCA RISING. COOLDOWN AND DEPRESSURIZATION.

1C(1A) LOOP RCS WR PRESS

[] PI 402A

[] PI 403A Plausibility for transition to ESP-1.2.

Page 23 of 35

12/17/2012 13:16 FNP-1-EEP-0 UNIT 1 REACTOR TRIP OR SAFETY INJECTION Revision 44.0 Step Action/Expected Response Response NOT Obtained 20 Isolate HHSI flow.

20.1 Verify charging pump miniflow valves - OPEN.

1A(1B,1C) CHG PUMP MINIFLOW ISO

[] Q1E21MOV8109A

[] Q1E21MOV8109B

[] Q1E21MOV8109C CHG PUMP MINIFLOW ISO

[] Q1E21MOV8106 20.2 Close HHSI isolation valves.

SI termination is actually HHSI TO RCS CL ISO performed here, prior to

[] Q1E21MOV8803A going to ESP-1.1.

[] Q1E21MOV8803B 21 Establish normal charging.

21.1 Manually close charging flow control valve.

CHG FLOW

[] FK 122 Step 21 continued on next page.

Page 24 of 35

12/17/2012 13:16 FNP-1-EEP-0 UNIT 1 REACTOR TRIP OR SAFETY INJECTION Revision 44.0 Step Action/Expected Response Response NOT Obtained 21.2 Verify charging flow path aligned.

21.2.1 Verify charging pump discharge flow path -

ALIGNED.

CHG PUMP DISCH HDR ISO

[] Q1E21MOV8132A open

[] Q1E21MOV8132B open

[] Q1E21MOV8133A open

[] Q1E21MOV8133B open CHG PUMPS TO REGENERATIVE HX

[] Q1E21MOV8107 open

[] Q1E21MOV8108 open 21.2.2 Verify only one charging line valve - OPEN.

RCS NORMAL CHG LINE

[] Q1E21HV8146 RCS ALT CHG LINE

[] Q1E21HV8147 Step 21 continued on next page.

Page 25 of 35

12/17/2012 13:16 FNP-1-EEP-0 UNIT 1 REACTOR TRIP OR SAFETY INJECTION Revision 44.0 Step Action/Expected Response Response NOT Obtained

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ NOTE: The RCS may be approaching solid plant conditions. In the event the RCS must be operated water solid, charging flow should be adjusted to maintain subcooling instead of pressurizer level.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ 21.2.3 [CA] Control charging flow 21.2.3 Perform the following.

to maintain pressurizer level 25%-50%{50%-60%}. IF pressurizer level cannot be maintained 21.2.3.1 Control charging flow. greater than 25%{50%},

THEN perform the CHG FLOW following.

[] FK 122 adjusted

1) Establish HHSI flow using ATTACHMENT 7, RE-ESTABLISHING HHSI FLOW.

2) Go to FNP-1-ESP-1.2, POST LOCA COOLDOWN AND DEPRESSURIZATION.

IF solid plant operation required, THEN control charging flow to maintain subcooling greater than 16 16F{45 F{45F}.

CHG FLOW

[] FK 122 adjusted 22 Go to FNP-1-ESP-1.1, SI TERMINATION, step 6. Plausibility for SI termination to be performed in ESP-1.1, 23 Begin monitoring critical you are directed there by this safety function status trees.

procedure.

Page 26 of 35

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 100. WE06EG2.1.27 100 A LOCA has occurred on Unit 1 with the following conditions:

There has been no HHSI or LHSI flow for an extended period of time.

ESP-1.2, Post LOCA Cooldown and Depressurization, was in progress when conditions required a transition to FRP-C.2, Response to Degraded Core Cooling.

A SG depressurization is in progress per FRP-C.2.

Subsequently, core cooling is re-established:

1A RHR pump is started.

LHSI flow is 1800 gpm.

Which one of the following completes the statements below?

The SG depressurization is performed at (1) per FRP-C.2.

Once the SG depressurization is complete and core cooling is re-established, a transition will be made to (2) .

A. 1) a maximum attainable rate

2) EEP-1.0, Loss of Reactor or Secondary Coolant B. 1) <100°F/hour cooldown rate

2) EEP-1.0, Loss of Reactor or Secondary Coolant C. 1) a maximum attainable rate

2) ESP-1.2, Post LOCA Cooldown and Depressurization D. 1) <100°F/hour cooldown rate

2) ESP-1.2, Post LOCA Cooldown and Depressurization Tuesday, March 19, 2013 11:38:44 AM 89

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 A. Incorrect 1) Incorrect, plausible because this is the action that would be performed in FRP-C.1.

2) Correct, for the conditions given, a transition is made to EEP-1 after core cooling is re-established and the SG depressurization is complete.

B. Correct 1) Correct, this is the required cooldown rate for FRP-C.2.

2) Correct, see A.2.

C. Incorrect 1) Incorrect, see A.1.

2) Incorrect, plausible since before a cooldown is performed in FRP-C.2, if conditions are met at step 8.3, an exit to "the procedure and step in effect" would be performed, and the procedure previously in effect was ESP-1.2.

D. Incorrect 1) Correct, see B.1.

2) Incorrect, see C.2.

K/A: WE06EG2.1.27 Degraded Core Cooling Knowledge of system purpose and/or function.

(CFR: 41.7)

IMPORTANCE RO 3.9 SRO 4.0 Importance Rating: 3.9 4.0 Technical

Reference:

FNP-1-FRP-C.2, v18 References provided: None Learning Objective: ASSESS the facility conditions associated with the (1)

FRP-C.1, Response to Inadequate Core Cooling; (2)

FRP-C.2, Response to Degraded Core Cooling; (3)

FRP-C.3, Response to Saturated Core Cooling, and based on that assessment: (OPS-62533C01)

SELECT the appropriate procedures during normal, abnormal and emergency situations. 10CFR55.43 (b) 5

DETERMINE if transition to another section of the procedure or to another procedure is required

DETERMINE if the critical safety functions are satisfied Question History: New question K/A match: The K/A requires a knowledge of the function of actions taken during a Degraded Core cooling scenario (WE06).

Tuesday, March 19, 2013 11:38:44 AM 90

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 The candidate is questioned on the procedural actions taken to perform a SG depressurization as it relates to how the depressurization is performed (to cooldown and depressurize the RCS at <100°F/hr). In addition, knowledge of the procedurally directed transition is required.

SRO justification: 10 CFR 55.43(b)(5)

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

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

Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures.

2013 NRC exam From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 flowchart for 10 CFR 55.43(b)(5) :

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations [10 CFR 55.43(b)(5)], involving BOTH:

1) assessing plant conditions and then

2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

Using the flowchart, this question can:

NOT be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location.

NOT be answered solely by knowing immediate operator actions.

NOT be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs.

NOT be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

Be answered with knowledge of ONE or MORE of the following:

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

Knowledge of when to implement attachments and appendices, including Tuesday, March 19, 2013 11:38:44 AM 91

QUESTIONS REPORT for ILT 36 SRO NRC Exam Ver6 Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures.

Tuesday, March 19, 2013 11:38:44 AM 92

1/7/2013 14:07 UNIT 1 FNP-1-FRP-C.2 3-03-2011 Revision 18 FARLEY NUCLEAR PLANT FUNCTION RESTORATION PROCEDURE FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING S

A

² F PROCEDURE USAGE REQUIREMENTS-per FNP-0-AP-6 SECTIONS E

¨¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ T Continuous Use ALL Y

¨¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ Reference Use R

¨¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ E Information Use L

©° A T

E D

See pages 11,15,16,21 Approved:

David L Reed (for)

Operations Manager Date Issued: 03/24/12

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Table of Contents Procedure Contains Number of pages Body..............................................22 Figure 1..........................................1 Attachment 1......................................1 Attachment 2......................................3 Attachment 3......................................1 Page 1 of 1

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 A. Purpose This procedure provides actions to restore core cooling.

B. Symptoms or Entry Conditions I. This procedure is entered from the Core Cooling Critical Safety Function Status Tree, on an Orange condition.

Page 1 of 22

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained CAUTION CAUTION:

Further degradation of core cooling can occur if any running RCP is stopped before being directed by this procedure even if normal support conditions are lost.

1 Monitor RWST level.

RWST LVL

[] LI 4075A

[] LI 4075B 1.1 [CA] WHEN RWST level less than 12.5 ft, THEN go to FNP-1-ESP-1.3, TRANSFER TO COLD LEG RECIRCULATION.

2 Verify proper SI valve alignment using ATTACHMENT 2, SI VALVE ALIGNMENT FOR COLD LEG INJECTION.

Page 2 of 22

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained 3 Check any HHSI flow - GREATER 3 Perform the following.

THAN 0 gpm.

3.1 Verify all available charging A TRN pumps started.

HHSI FLOW

[] FI 943 CHG PUMP

[] 1A amps > 0 HHSI [] 1B amps > 0 B TRN RECIRC [] 1C amps > 0 FLOW

[] FI 940 3.2 Verify charging pump MOV disconnects closed using ATTACHMENT 3, CHARGING PUMP MOV DISCONNECTS.

3.3 Verify proper SI alignment.

CHG PUMPS TO REGENERATIVE HX

[] Q1E21MOV8107 closed

[] Q1E21MOV8108 closed RWST TO CHG PUMP

[] Q1E21LCV115B open

[] Q1E21LCV115D open VCT OUTLET ISO

[] Q1E21LCV115C closed

[] Q1E21LCV115E closed HHSI TO RCS CL ISO

[] Q1E21MOV8803A open

[] Q1E21MOV8803B open CHG PUMP SUCTION HDR ISO

[] Q1E21MOV8130A open

[] Q1E21MOV8130B open

[] Q1E21MOV8131A open

[] Q1E21MOV8131B open CHG PUMP DISCH HDR ISO

[] Q1E21MOV8132A open

[] Q1E21MOV8132B open

[] Q1E21MOV8133A open

[] Q1E21MOV8133B open Step 3 continued on next page.

Page 3 of 22

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained 3.4 IF HHSI flow now established, THEN proceed to step 4, IF NOT NOT,

, establish HHSI bypass SI flow.

CHG PMP RECIRC TO RCS COLD LEGS

[] Q1E21MOV8885 open HHSI TO RCS CL ISO

[] Q1E21MOV8803A closed

[] Q1E21MOV8803B closed 3.5 IF HHSI flow now established, THEN proceed to step 4, IF NOT NOT,

, perform the following.

3.5.1 Open HHSI isolation valves.

HHSI TO RCS CL ISO

[] Q1E21MOV8803A

[] Q1E21MOV8803B 3.5.2 Align charging pump suction header isolation valves based on 1B charging pump status.

>>¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥¥¥¥¥ 1B Charging Pump Pump Aligned As Aligned As Status A Train pump B Train pump pump pump CHG PUMP SUCTION HDR ISO Q1E21MOV [] 8130A [] 8130A open closed

[] 8130B [] 8130B open closed

[] 8131A [] 8131A closed open

[] 8131B [] 8131B closed open

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¥¥¥¥º Step 3 continued on next page.

Page 4 of 22

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained 3.5.3 Align charging pump discharge header isolation valves based on 1B charging pump status.

>>¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥¥¥¥¥ 1B Charging Pump Pump Aligned As Aligned As Status A Train pump B Train pump pump pump CHG PUMP DISCH HDR ISO Q1E21MOV [] 8132A [] 8132A open closed

[] 8132B [] 8132B open closed

[] 8133A [] 8133A closed open

[] 8133B [] 8133B closed open

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¥¥¥¥º

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ NOTE: Continuing efforts to establish SI flow should not interfere with performance of the remainder of this procedure.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ 3.6 If HHSI flow NOT established, THEN Continue efforts to establish SI flow.

HHSI flow LHSI flow Any form of RCS injection.

Page 5 of 22

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained CAUTION CAUTION:

Pump damage may occur if RHR pumps are operated on miniflow for longer than three hours with no CCW supplied to the RHR heat exchangers.

4 Check LHSI status.

4.1 Verify CCW flow to RHR heat exchangers - ESTABLISHED.

CCW TO 1A(1B) RHR HX

[] Q1P17MOV3185A open

[] Q1P17MOV3185B open 4.2 Check RCS pressure - LESS THAN 4.2 Proceed to step 5.

275 psig{435 psig}.

1C(1A) LP RCS NR PRESS

[] PI 402B

[] PI 403B Step 4 continued on next page.

Page 6 of 22

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained 4.3 Check both RHR flows - GREATER 4.3 Verify LHSI properly aligned.

THAN 1.5x10 3 gpm.

RHR PMP 1A(1B) [] 1A amps > 0 RHR HDR [] 1B amps > 0 FLOW

[] FI 605A 1A(1B) RHR HX TO RCS

[] FI 605B COLD LEGS ISO

[] Q1E11MOV8888A open

[] Q1E11MOV8888B open RWST TO 1A(1B) RHR PUMP

[] Q1E11MOV8809A open

[] Q1E11MOV8809B open 1A(1B) RHR HX DISCH VLV

[] HIK 603A open

[] HIK 603B open 1A(1B) RHR HX BYP FLOW VLV

[] FK 605A closed

[] FK 605B closed 1A(1B) RHR TO RCS HOT LEGS XCON

[] Q1E11MOV8887A open

[] Q1E11MOV8887B open Page 7 of 22

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained 5 Check RCS vent paths.

5.1 Check any PRZR PORV ISO - 5.1 Restore power to PRZR PORV ISO POWER AVAILABLE. valves unless de-energized for inoperable PORVs not capable of being manually cycled.

5.2 Verify both PRZR PORVs - 5.2 Perform the following.

CLOSED.

5.2.1 Close PRZR PORVs.

5.2.2 IF any valve can NOT be closed, THEN close its PORV ISO valve.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ NOTE: The purpose of the following step is to establish an available PORV flowpath for mitigation of overpressure conditions, without relying on the PRZR code safety valves. A failed open PORV must not be unisolated. A leaking PORV which is isolated with power available to the isolation valve should remain isolated until needed to reduce RCS pressure or mitigate an RCS overpressure condition. Any leaking PORV should be re-isolated when not in use.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ 5.3 Check at least one PRZR PORV 5.3 Open any PRZR PORV ISO not ISO - OPEN. required to isolate an open or leaking PORV.

5.4 Verify reactor vessel head vent valves - CLOSED.

RX VESSEL HEAD VENT OUTER ISO

[] Q1B13SV2213A

[] Q1B13SV2213B RX VESSEL HEAD VENT INNER ISO

[] Q1B13SV2214A

[] Q1B13SV2214B Page 8 of 22

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained 6 Check RCP status.

6.1 Check at least one RCP - 6.1 Proceed to Step 8.

STARTED.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ NOTE: Normal support conditions for running RCPs are desired, however, RCP operation must continue even if support conditions cannot be maintained.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ 6.2 Verify No. 1 seal support conditions established.

6.2.1 [CA] Maintain seal injection flow - GREATER THAN 6 gpm.

6.2.2 Verify No. 1 seal leakoff flow - WITHIN FIGURE 1 LIMITS.

6.2.3 Verify No. 1 seal differential pressure -

GREATER THAN 200 psid.

6.3 Verify CCW - ALIGNED.

CCW FROM RCP THRM BARR

[] Q1P17HV3045 open

[] Q1P17HV3184 open 6.4 Check RCP thermal barrier - 6.4 Verify CCW flow isolated.

INTACT.

CCW FROM RCP RCP THRM BARR THRM BARR [] Q1P17HV3045 closed CCW FLOW [] Q1P17HV3184 closed HI

[] Annunciator DD2 clear Step 6 continued on next page.

Page 9 of 22

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained 6.5 Check CCW to RCP oil coolers - 6.5 Verify CCW - ALIGNED.

SUFFICIENT.

CCW TO RCP CLRS CCW FLOW [] Q1P17MOV3052 open FROM RCP OIL CLRS CCW FROM RCP LO OIL CLRS

[] Annunciator DD3 clear [] Q1P17MOV3046 open

[] Q1P17MOV3182 open 6.6 Check RCP oil level -

SUFFICIENT.

RCP 1A(1B,1C) BRG UPPER/LOWER OIL RES LO LVL

[] Annunciator HH1 clear

[] Annunciator HH2 clear

[] Annunciator HH3 clear

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ NOTE: Since RCP damage may occur when operating RCPs without normal support conditions established or under highly voided RCS conditions, the intent of the following step is to save one RCP (which provides the best pressurizer spray capability) for future use, if all three RCPs are running.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ 7 Check if one RCP should be stopped.

7.1 Check ALL RCPs - STARTED 7.1 Proceed to Step 9.

7.2 Stop RCP 1B.

7.3 Proceed to Step 9.

Page 10 of 22

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained 8 Check core cooling.

8.1 Check REACTOR VESSEL LEVEL 8.1 IF SI established, indication - GREATER THAN 0% THEN return to step 2, UPPER PLENUM. IF NOT NOT,

, proceed to step 9.

8.2 Check core exit T/Cs - LESS 8.2 IF core exit T/Cs falling, THAN 700 700F. THEN return to step 2, IF NOT NOT,

, proceed to step 9.

8.3 Go to procedure and step in This provides plausibility for the distracter to effect.

go to ESP-1.2. Once the cooldown is 9 Check SI accumulator discharge initiated at step 12, you will always exit to valve status. EEP-1.

9.1 Check power to discharge 9.1 Close accumulator discharge valves - AVAILABLE. valve disconnects using ATTACHMENT 1.

1A(1B,1C) ACCUM DISCH ISO

[] Q1E21MOV8808A

[] Q1E21MOV8808B

[] Q1E21MOV8808C 9.2 Check discharge valves - OPEN. 9.2 IF accumulators have NOT discharged, 1A(1B,1C) ACCUM THEN open discharge valves.

DISCH ISO

[] Q1E21MOV8808A 1A(1B,1C) ACCUM

[] Q1E21MOV8808B DISCH ISO

[] Q1E21MOV8808C [] Q1E21MOV8808A

[] Q1E21MOV8808B

[] Q1E21MOV8808C 10 Monitor CST level.

10.1 [CA] Check CST level greater 10.1 Align AFW pumps suction to SW than 5.3 ft. using FNP-1-SOP-22.0, AUXILIARY FEEDWATER SYSTEM.

CST LVL

[] LI 4132A

[] LI 4132B 10.2 Align makeup to the CST from water treatment plant OR demin water system using FNP-1-SOP-5.0, DEMINERALIZED MAKEUP WATER SYSTEM, as necessary.

Page 11 of 22

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained CAUTION CAUTION:

To prevent potential release of radioactive material to the atmosphere, a faulted or ruptured SG should only be used if no intact SG is available.

11 Check intact SG levels.

11.1 Check narrow range levels - 11.1 Verify total AFW flow to GREATER THAN 31%{48%}. intact SGs greater than 395 gpm.

AFW FLOW TO 1A(1B,1C) SG

[] FI 3229A

[] FI 3229B

[] FI 3229C AFW TOTAL FLOW

[] FI 3229 Step 11 continued on next page.

Page 12 of 22

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained 11.2 [CA] WHEN intact SG narrow range level 31%-65%{48%-65%},

THEN maintain intact SG narrow range level 31%-65%{48%-65%}.

11.2.1 Control MDAFWP flow.

MDAFWP FCV 3227 RESET

[] A TRN reset

[] B TRN reset MDAFWP TO 1A/1B/1C SG B TRN

[] FCV 3227 in MOD

>>¥¥¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥ Intact SG 1A 1B 1C MDAFWP TO 1A(1B,1C) SG SG Q1N23HV [] 3227A [] 3227B 3227A [] 3227C 3227B 3227C in MOD in MOD in MOD MDAFWP TO 1A(1B,1C) SG SG FLOW CONT HIC []3227AA []3227BA

[]3227AA []3227CA

[]3227BA

[]3227CA adjusted adjusted adjusted adjusted adjusted adjusted

¥¥¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥º Step 11 continued on next page.

Page 13 of 22

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained 11.2.2 Control TDAFWP flow.

TDAFWP FCV 3228

[] RESET reset TDAFWP SPEED CONT

[] SIC 3405 adjusted

>>¥¥¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥ Intact SG 1A 1B 1C TDAFWP TO 1A(1B,1C) SG SG Q1N23HV [] 3228A [] 3228B 3228A [] 3228C 3228B 3228C in MOD in MOD in MOD TDAFWP TO 1A(1B,1C) SG SG FLOW CONT HIC []3228AA []3228BA

[]3228AA []3228CA

[]3228BA

[]3228CA adjusted adjusted adjusted adjusted adjusted adjusted

¥¥¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥º Page 14 of 22

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained CAUTION CAUTION:

Performance of step 12 will cause accumulator injection which may result in a red path on the INTEGRITY status tree. This procedure should be completed before transition to FNP-1-FRP-P.1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITIONS.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ NOTE: After the low steam line pressure SI is blocked, excessive opening of steam dumps can cause a high steam flow LO-LO TAVG main steam isolation signal.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ 12 Reduce pressure in all intact SGs to 160 psig.

12.1 WHEN P-12 light lit (543 F),

(543 The SG THEN perform the following.

depressurization starts here. Once you 12.1.1 Block low steam line pressure SI.

STM LINE PRESS SI BLOCK - RESET get to this step, you never go back in this

[] A TRN to BLOCK

[] B TRN to BLOCK 12.1.2 Verify blocked indication.

procedure, you just go BYP & PERMISSIVE STM LINE ISOL. forward. Step 19 to EEP-1 is the final SAFETY INJ.

[] TRAIN A BLOCKED light lit

[] TRAIN B BLOCKED light lit 12.1.3 Bypass the steam dump transition out.

interlock.

STM DUMP INTERLOCK

[] A TRN to BYP INTLK

[] B TRN to BYP INTLK Step 12 continued on next page.

Page 15 of 22

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained 12.1.4 Adjust steam header The cooldown rate is maintained <100°F/

pressure controller to control cooldown rate.

STM HDR PRESS hr throughout this

[] PK 464 adjusted procedure to prevent 12.2 [CA] Maintain RCS cold leg cooldown rate - LESS THAN 100 100F IN ANY 60 MINUTE PERIOD.

going to FRP-P.1.

12.3 IF condenser available, 12.3 Dump steam to atmosphere.

THEN dump steam to condenser from intact SGs. 12.3.1 Direct counting room to perform FNP-0-CCP-645, MAIN BYP & PERMISSIVE STEAM ABNORMAL COND ENVIRONMENTAL RELEASE.

AVAIL

[] C-9 status light lit 12.3.2 IF normal air available, THEN control atmospheric STM DUMP relief valves to dump steam

[] MODE SEL A-B TRN in STM PRESS from intact SGs, IF NOT NOT,, dump steam using STM DUMP FNP-1-SOP-62.0, EMERGENCY INTERLOCK AIR SYSTEM.

[] A TRN in ON

[] B TRN in ON 1A(1B,1C) MS ATMOS REL VLV STM HDR [] PC 3371A adjusted PRESS [] PC 3371B adjusted

[] PK 464 adjusted [] PC 3371C adjusted 12.3.3 IF no source of air available, THEN locally control SG atmospheric relief valves with handwheel to dump steam from intact SGs.

(127 ft, AUX BLDG main steam valve room)

>>¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥ Intact SG SG 1A 1B 1C Q1N11PCV [] 3371A 3371A[] 3371B 3371B[] 3371C 3371C

¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥º 12.4 Check all intact SG pressures 12.4 Return to Step 11. OBSERVE

- LESS THAN 160 psig. CAUTION PRIOR TO STEP 11.

Step 12 continued on next page.

Page 16 of 22

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained 12.5 Check at least two RCS hot leg 12.5 Return to Step 11. OBSERVE temperatures - LESS THAN CAUTION PRIOR TO STEP 11.

385 385F.

RCS HOT LEG TEMP

[] TR 413 12.6 Stop SG pressure reduction.

STM HDR PRESS

[] PK 464 adjusted 1A(1B,1C) MS ATMOS REL VLV

[] PC 3371A adjusted

[] PC 3371B adjusted

[] PC 3371C adjusted OR 1A(1B,1C) MS ATMOS REL VLV

[] Q1N11PCV3371A closed

[] Q1N11PCV3371B closed

[] Q1N11PCV3371C closed CAUTION CAUTION:

Pump damage may occur if RHR pumps are operated on miniflow for longer than 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months with no CCW supplied to the RHR heat exchangers.

13 Verify RHR pumps - STARTED.

RHR PUMP

[] 1A amps > 0

[] 1B amps > 0 Page 17 of 22

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained 14 [CA] Check if SI accumulators should be isolated.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ NOTE: Step 14.1 is a continuing action.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ 14.1 [CA] Check at least two RCS 14.1 Perform the following.

hot leg temperatures - LESS THAN 385 385F. 14.1.1 WHEN at least two RCS hot leg temperatures are less RCS HOT LEG TEMP than 385 385F,

[] TR 413 THEN perform steps 14.2 and 14.3 to isolate accumulators.

14.1.2 Proceed to step 15.

OBSERVE CAUTION PRIOR TO STEP 15.

14.2 Verify SI reset. 14.2 Perform the following:

[] MLB-1 1-1 not lit (A TRN) 14.2.1 IF any train will NOT reset

[] MLB-1 11-1 not lit (B TRN) using the MCB SI RESET pushbuttons, THEN place the affected train S821 RESET switch to RESET. (SSPS TEST CAB.)

14.2.2 IF a failure exists in SSPS such that SI cannot be reset, THEN reset SI using FNP-1-SOP-40.0, RESPONSE TO INADVERTENT SI AND INABILITY TO RESET OR BLOCK SI, Appendix 2.

Step 14 continued on next page.

Page 18 of 22

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained 14.3 Close all SI accumulator 14.3 Perform the following.

discharge valves.

14.3.1 Vent any SI accumulator 1A(1B,1C) ACCUM that cannot be isolated.

DISCH ISO

[] Q1E21MOV8808A ACCUM

[] Q1E21MOV8808B N2 VENT

[] Q1E21MOV8808C [] HIK 936 open

>>¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥ SI ACCUM 1A 1B 1C 1A(1B,1C) ACCUM ACCUM N2 SUPP/VT ISO Q1E21HV [] 8875A 8875A[] 8875B 8875B[] 8875C 8875C open open open

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥º 14.3.2 IF an accumulator can NOT be isolated or vented, THEN consult the TSC staff to determine contingency actions.

CAUTION CAUTION:

Core cooling may degrade during subsequent steps. FNP-1-CSF-0.2, CORE COOLING status tree should be closely monitored.

15 Stop all RCPs.

RCP

[] 1A

[] 1B

[] 1C Page 19 of 22

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained 16 Reduce pressure in all intact SGs to atmospheric pressure.

16.1 Maintain RCS cold leg cooldown rate - LESS THAN 100 F IN ANY 100 60 MINUTE PERIOD.

16.2 IF condenser available, 16.2 Dump steam to atmosphere.

THEN dump steam to condenser from intact SGs. 16.2.1 Direct counting room to perform FNP-0-CCP-645, MAIN BYP & PERMISSIVE STEAM ABNORMAL COND ENVIRONMENTAL RELEASE.

AVAIL

[] C-9 status light lit 16.2.2 IF normal air available, THEN control atmospheric STM DUMP relief valves to dump steam

[] MODE SEL A-B TRN in STM PRESS from intact SGs, IF NOT NOT,, dump steam using STM DUMP FNP-1-SOP-62.0, EMERGENCY INTERLOCK AIR SYSTEM.

[] A TRN in ON

[] B TRN in ON 1A(1B,1C) MS ATMOS REL VLV STM HDR [] PC 3371A adjusted PRESS [] PC 3371B adjusted

[] PK 464 [] PC 3371C adjusted 16.2.3 IF no source of air available, THEN locally control SG atmospheric relief valves with handwheel to dump steam from intact SGs.

(127 ft, AUX BLDG main steam valve room)

>>¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥ Intact SG 1A 1B 1C Q1N11PCV [] 3371A 3371A[] 3371B 3371B[] 3371C 3371C

¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥º Page 20 of 22

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained 17 Verify any SI flow established. 17 Perform the following.

Verify any HHSI flow - 17.1 Continue efforts to establish GREATER THAN 0 gpm. SI flow.

A TRN HHSI flow HHSI FLOW

[] FI 943 LHSI flow HHSI Any form of RCS injection.

B TRN RECIRC FLOW 17.2 Return to Step 16.

[] FI 940 OR Verify any LHSI flow -

GREATER THAN 1.5x10 3 gpm.

1A(1B)

RHR HDR FLOW

[] FI 605A

[] FI 605B 18 Check core cooling. 18 Return to step 16.

REACTOR VESSEL LEVEL indication - GREATER THAN 0%

UPPER PLENUM.

At least two RCS hot leg temperatures - LESS THAN 350 350F.

RCS HOT LEG TEMP

[] TR 413 19 Go to FNP-1-EEP-1, LOSS OF REACTOR OR SECONDARY COOLANT, step 13.

-END-Page 21 of 22

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 START STEP CONTINUOUS ACTION 1 1.1 [CA] WHEN RWST level less than 12.5 ft, THEN go to FNP-1-ESP-1.3, TRANSFER TO COLD LEG RECIRCULATION.

6 6.2.1 [CA] Maintain seal injection flow - GREATER THAN 6 gpm.

10 10.1 [CA] Check CST level greater than 5.3 ft.

11 11.2 [CA] WHEN intact SG narrow range level 31%-65%{48%-65%},

THEN maintain intact SG narrow range level 31%-65%{48%-65%}.

12 12.2 [CA] Maintain RCS cold leg cooldown rate - LESS THAN 100F IN ANY 60 MINUTE PERIOD.

14 [CA] Check if SI accumulators should be isolated.

14.1 [CA] Check at least two RCS hot leg temperatures -

LESS THAN 385F.

Page 22 of 22

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 FIGURE 1 NO. 1 SEAL NORMAL OPERATING RANGE 2400 2250 2200 2000 1800 Normal Operating Range 1600 No.1 Seal 1400 Differential Pressure (PSID, See 1200 Note 1) 1000 800 600 400 200 0

0 .2 1 2 3 4 5 6 No. 1 Seal Leak Rate (GPM)

Note 1: For No. 1 Seal Differential Pressures greater than 400 psid, use RCS pressure in psig.

Page 1 of 1

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained ATTACHMENT 1 1 Close the following disconnects.

>>¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥

'A' Train Disconnects

¥¥¥¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥£¥¥¥¥¥£¥¥¥¥¥¥¥¥¥¥¥¥¥ Disconnect Description Position Key Location Position TPNS No.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ Q1R18B031-A Disconnect FU-Z3 MOV 8808C-A 8808C-A ON Z-85 Z-85139' hallway-hallway-across from

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥MCC MCC 1A Q1R18B032-A Disconnect FU-Z2 MOV 8808A-A 8808A-A ON Z-84 Z-84

¥¥¥¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¢¥¥¥¥¥¢¥¥¥¥¥¥¥¥¥¥¥¥¥º

>>¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥

'B' Train Disconnects

¥¥¥¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥£¥¥¥¥¥£¥¥¥¥¥¥¥¥¥¥¥¥¥ Disconnect Description Position Key Location Position TPNS No.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ 139' hallway-hallway-Q1R18B035-B Disconnect FV-S2 MOV 8808B-B 8808B-B ON Z-86 Z-86across from chemistry sample room

¥¥¥¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¢¥¥¥¥¥¢¥¥¥¥¥¥¥¥¥¥¥¥¥º 2 Verify accumulator discharge valves MCB indication - POWER AVAILABLE.

1A(1B,1C) ACCUM DISCH ISO

[] Q1E21MOV8808A

[] Q1E21MOV8808B

[] Q1E21MOV8808C 3 Notify control room of accumulator discharge valve disconnect status.

-END-Page 1 of 1

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained ATTACHMENT 2 SI VALVE ALIGNMENT FOR COLD LEG INJECTION 1 Verify proper HHSI alignment.

CHG PUMPS TO REGENERATIVE HX

[] Q1E21MOV8107 closed

[] Q1E21MOV8108 closed RWST TO CHG PUMP

[] Q1E21LCV115B open

[] Q1E21LCV115D open VCT OUTLET ISO

[] Q1E21LCV115C closed

[] Q1E21LCV115E closed HHSI TO RCS CL ISO

[] Q1E21MOV8803A open

[] Q1E21MOV8803B open CHG PUMP SUCTION HDR ISO

[] Q1E21MOV8130A open

[] Q1E21MOV8130B open

[] Q1E21MOV8131A open

[] Q1E21MOV8131B open CHG PUMP DISCH HDR ISO

[] Q1E21MOV8132A open

[] Q1E21MOV8132B open

[] Q1E21MOV8133A open

[] Q1E21MOV8133B open Page 1 of 3

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained ATTACHMENT 2 2 Verify proper RHR alignment.

RWST TO 1A(1B) RHR PUMP

[] Q1E11MOV8809A open

[] Q1E11MOV8809B open CTMT SUMP TO 1A(1B) RHR PUMP

[] Q1E11MOV8811A closed

[] Q1E11MOV8812A closed

[] Q1E11MOV8811B closed

[] Q1E11MOV8812B closed 1C(1A) RCS LOOP TO 1A(1B) RHR PUMP

[] Q1E11MOV8701A closed

[] Q1E11MOV8702A closed

[] Q1E11MOV8701B closed

[] Q1E11MOV8702B closed RHR TO RCS HOT LEG XCON

[] Q1E11MOV8887A open

[] Q1E11MOV8887B open 1A(1B) RHR HX TO CHG PUMP SUCT

[] Q1E11MOV8706A closed

[] Q1E11MOV8706B closed 1A(1B) RHR HX DISCH VLV

[] HIK 603A open

[] HIK 603B open 1A(1B) RHR HX BYP FLOW VLV

[] FK 605A closed

[] FK 605B closed Page 2 of 3

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained ATTACHMENT 2 3 Align charging pump miniflow valves based on RCS pressure.

1C(1A) LOOP RCS WR PRESS

[] PI 402A

[] PI 403A

>>¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥¥¥ RCS PRESSURE <1300 psig 1300 psig-psig psig->1900 psig psig 1900 psig 1A(1B,1C)CHG PUMP PUMP MINIFLOW ISO Q1E21MOV []8109A []8109A []8109A closed as is open

[]8109B []8109B []8109B closed as is open

[]8109C []8109C []8109C closed as is open CHG PUMP MINIFLOW ISO Q1E21MOV []8106 []8106 []8106 closed as is open

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¥¥º

-END-Page 3 of 3

1/7/2013 14:07 FNP-1-FRP-C.2 UNIT 1 RESPONSE TO DEGRADED CORE COOLING Revision 18 Step Action/Expected Response Response NOT Obtained ATTACHMENT 3 CHARGING PUMP MOV DISCONNECTS 1 Close the following disconnects.

>>¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥

'B' Train Disconnects

¥¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥£¥¥¥¥¥¥£¥¥¥¥¥¥¥¥¥¥¥¥ Disconnect Description Position Key Location Position TPNS No.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ Q1R18B034-B Disconnect FV-J2 MOV 8132B-B Q1R18B034-B 8132B-B ON Z-89

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥139' 139' hallway hallway Q1R18B041-B Disc for MOV 8130B-B Q1R18B041-B ON Z-408 Z-408across from

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥Chemistry Chemistry Q1R18B042-B Disc for MOV 8131B-B Q1R18B042-B ON Z-412 Z-412Sample Room

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ Q1R18B043-B Disc for MOV 8133B-B Q1R18B043-B ON Z-416 Z-416

¥¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¥¥¥¥º 2 Close the following disconnects.

>>¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥

'A' Train Disconnects

¥¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥£¥¥¥¥¥¥¥¥£¥¥¥¥¥¥£¥¥¥¥¥¥¥¥¥¥¥¥ Disconnect Description Position Key Location Position TPNS No.

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ Q1R18B030-A Disconnect FU-J2 MOV 8132A-A Q1R18B030-A 8132A-A ON Z-203 Z-203

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ Q1R18B038-A Disc for MOV 8130A-A Q1R18B038-A ON Z-405 Z-405139' hallway hallway

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥across across from Q1R18B039-A Disc for MOV 8131A-A Q1R18B039-A ON Z-410 Z-410MCC 1A

¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥ Q1R18B040-A Disc for MOV 8133A-A Q1R18B040-A ON Z-415 Z-415

¥¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¢¥¥¥¥¥¥¢¥¥¥¥¥¥¥¥¥¥¥¥º 3 Notify control room of charging pump discharge valve disconnect status.

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ML14091A150

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