Revision as of 02:12, 4 November 2019

Text

Initial Exam 2014-301 Draft SRO Written Exam
	ML14233A497
Person / Time
Site:	North Anna
Issue date:	08/21/2014
From:	; Division of Reactor Safety II
To:	; Virginia Electric & Power Co (VEPCO)
References
	50-338/14-301, 50-339/14-301
	Download: ML14233A497 (74)
	v • d • e

ES-401 Site-Specific SRO Written Examination Form ES-401-8 Cover Sheet U.S. Nuclear Regulatory Commission Site-Specific SRO Written Exam i nation Applicant Information Name:

Date: Facility/Unit: North Anna Power Station Region: I [1 II Ill IV Li Reactor Type: WE CE BW DSE LI Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80.00 percent to pass. You have 8 hours0.333 days 0.0476 weeks 0.011 months to complete the combined examination, and 3 hours0.125 days 0.0179 weeks 0.00411 months if you are only taking the SRO portion.

Applicant Certification All work done on this examination is my own. I have neither given nor received aid.

Applicants Signature Results RO/SRO-OnlyJTotaI Examination Values / / Points Applicants Scores I I Points Applicants Grade I / Percent ES-401, Page 31 of 33

76. 076 015AA2.09 001/BANK/SURRY EXAM 2012/HIGH/3/3.4/3.5//

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Given the following:

0800: Unit 1 is at 100% steady-state power. PCS is out of service.

0810: The following annunciator is received:

1-AR-C-H4 RCP lA-B-C BEARING HI TEMP 0812: The plant operator reports the following lB RCP temperatures:

Motor upper bearing 145°F

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Motor upper thrust bearing - 148°F

Motor lower thrust bearing - 131°F

Motor lower bearing - 142°F

Motor stator 305°F

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Pump radial bearing - 130°F lAW with the Annunciator Response, based on the above conditions, at 0812 (1) What actions are required?

AND (2) What is the Technical Specification basis for prohibiting power operation with less than 3 RCPs in service?

A. (1) The Reactor must be tripped and 1 B RCP stopped (2) The design limit for fuel peak centerline temperature (PCT)

B. (1) Continued operation is allowed, increase monitoring of lB RCP (2) The design limit for fuel peak centerline temperature (PCT)

C. (1) Continued operation is allowed, increase monitoring of lB RCP (2) The design limit for departure from nucleate boiling ratio (DNBR)

D (1) The Reactor must be tripped and lB RCP stopped (2) The design limit for departure from nucleate boiling ratio (DNBR)

Distractor Analysis:

In accordance with 1-OP-5.2, RCP Trip criteria are as follows:

RCP Motor Bearing temperatures greater than 195°F

RCP Lower Seal Water Bearing (Pump Bearing) temperature greater than 225°F

RCP Stator Winding temperature greater than 300° F Based on the given conditions, 1 B RCP meets trip criteria for Stator Winding Temperature.

In accordance with Technical Specification LCO 3.4.4 Bases, The DNB Analyses assume normal three loop operation.. .The Unit is designed to operate with all RCS Loops in operation to maintain DNBR above the limit during all normal operations and anticipated transients.

A) INCORRECT The first part is correct. The design limit for fuel peak centerline temp is plausible as that is the basis for Nuclear Enthalpy Rise Hot Channel Factor (LCO 3.2.2.)

B) INCORRECT This is plausible based on all of the motor and pump bearing temperatures remaining below the RCP Trip criteria. The design limit for fuel peak centerline temp is plausible as that is the basis for Nuclear Enthalpy Rise Hot Channel Factor (LCD 3.2.2.)

C) INCORRECT This is plausible based on all of the motor and pump bearing temperatures remaining below the RCP Trip criteria. The second part is correct.

D) CORRECT See above.

Title:

RCP Malfunctions K/A:

01 5AA2.09 Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): When to secure RCP5 on high stator temperature.

Technical

References:

1-AR-C-H4 RCP lA-B-C BEARING HI TEMP 1-OP-5.2, Reactor Coolant Pump Startup and Shutdown (rev 42)

North Anna Tech Specs (Safety Limits) and 3.4.4 (RCS Loops MODES 1 & 2)

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References provided to applicants: None Learning Objective:

9572 List the motor and bearing temperature limits that require tripping the reactor coolant pump.

SRO-only: 10 CFR 55.43(b)(2)/ 1OCFR55.43(b)(5)

The question requires the applicant to assess plant conditions and to know the content of procedures In order to select a required course of action, meets 1 OCFR55.43(b)(5);

and have knowledge of facility operating limitations in the technical specifications and their basis, meets 1OCFR55.43(b)(2).

77. 077 022AA2.01 OO1INEW/N/A/HIGH/3/3.2/3.8//

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Given the following:

Unit 1 is at 100% power

RCS Tave is stable

Charging flow is rising

VCT level is lowering

PRZR level is lowering

The containment sump pumping rate is rising

Regenerative Heat Exchanger Letdown outlet temperature is rising

Regenerative Heat Exchanger Charging outlet temperature is rising (1) Which of the following is the cause for the indications above?

(2) Following containment entry, if the leakage is determined to be 3.8 gpm from a valve packing, what is the OPERABILITY status of LCO 3.4.13, RCS Operational Leakage? (Assume previous RCS total leakrate was 0.2 gpm)

A (1)Leak in charging line upstream of the Regenerative Heat Exchanger (2) OPERABLE B. (1)Leak in charging line upstream of the Regenerative Heat Exchanger (2) INOPERABLE C. (1)Leak in letdown line upstream of the Regenerative Heat Exchanger (2) OPERABLE D. (1)Leak in letdown line upstream of the Regenerative Heat Exchanger (2) INOPERABLE

Distractor Analysis:

Based on both the letdown and charging temperatures rising and loss of level in both the VCT and PRZR the leak is in the charging line upstream of the Regen Heat Exchanger.

LCO 3.4.13 bases states the following: Leakage past seals and gaskets is not considered pressure boundary leakage.

3.8 gpm is above the limit for unidentified, but in this case a containment entry was made and consistent with the bases identified leakage is described as, from known sources that do not interfere with detection of unidentified leakage and is will within the capability of the RCS makeup system. Identified leakage includes.. .from specifically known and located sources.

Based on the 0.2 gpm previous leakrate and the current 3.8 gpm the leakrates are within the LCO requirements for Identified Leakage.

A) CORRECT See above.

B) INCORRECT The first part is correct. Inoperable is plausible if the values for unidentified leakage are used to evaluate operability. This limit is 1 gpm.

C) INCORRECT Leak in the letdown line is plausible, however with Pressurizer level lowering this is incorrect. The second part is correct D) INCORRECT Leak in the letdown line is plausible, however with Pressurizer level lowering this is incorrect. Inoperable is plausible if the values for unidentified leakage are used to evaluate operability. This limit is 1 gpm.

Title:

Loss of Reactor Coolant Makeup K/A:

022AA2.O1 Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup: Whether charging line leak exists.

Technical

References:

Chemical and Volume Control Lesson Plan/Student Guide References provided to applicants: None Learning Objective:

11969 Given a set of plant conditions, evaluate Chemical and Volume Control System operations in light of the following issues.

Effect of a failure, malfunction, or loss of a related system or component on this system

Effect of a failure, malfunction, or loss of components in this system on related systems

Expected plant or system response based on chemical and volume control component interlocks or design features

Impact on technical specifications

Response if limits or setpoints associated with this system or its components have been exceeded

Proper operator response to the condition as stated SRO-only: 10 CFR 55.43(b)(2)

Analysis of Tech Specs and basis evaluating operability for type of leakage and the impact of that leakage.

78. 078 029EG2.4.8 001 /NEW/N/A/HIGH/3!3 .8/4.5/!

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Given the following:

RCS Pressure was rapidly lowering and Containment Pressure is rising

Based on degrading plant conditions, Operators attempted to trip the Reactor manually

The manual Reactor Trip was unsuccessful and 1-FR-S.1, Response to Nuclear Power Generator/ATWS was entered

Upon entry into 1-FR-S.1, Semi-Vital Bus lB is lost and de-energizes In accordance with OP-AP-104, Emergency and Abnormal Operating Procedures, upon transition out of 1-FR-S.1 what is the requirement for when and how the abnormal operating procedure, 0-AP-lO, Loss of Electrical Power, is used?

A. Parallel procedure usage is not allowed, a transition out of 1-E-O is required prior to implementation of O-AP-1 0.

All steps in 0-AP-1 0, attachment 12 for Loss of Semi-Vital Bus 1 B must be completed.

B. Parallel procedure usage is allowed, 1-E-0 immediate action steps must be completed prior to implementation of 0-AP-1 0.

All steps in 0-AP-1 0, attachment 12 for Loss of Semi-Vital Bus 1 B must be completed.

Cs Parallel procedure usage is allowed, 1-E-O immediate action steps must be completed prior to implementation of 0-AP-1 0.

Only the steps in the AP that ensure success of the EOP are required to be performed.

D. Parallel procedure usage is not allowed, a transition out of 1-E-0 is required prior to implementation of 0-AP-1 0.

Only the steps in the AP that ensure success of the EOP are required to be performed.

Distractor Analysis:

Based on RCS and Containment conditions, a SI is required and E-O will remain in use after step 4. Transition will not occur until after diagnostic steps. The candidate will have to recognize that transition does not occur at step 4 following immediate actions.

OP-AP-104, Emergency and Abnormal Operating Procedures contains a note in section 3.7.

The priority of procedures depends upon the events in progress. Some abnormal operating procedures must be implemented while abnormal operating procedures are in effect. In cases of parallel procedure usage, the EOP receives priority and immediate actions are completed before parallel procedure usage. When using an AOP in parallel with the EOP, only those steps in the AOP that ensure success of the EOP are required to be performed.

A. INCORRECT Parallel procedure usage is allowed. This is plausible because of a statement in OP-AP-104 related to when other procedures can be implemented. Step 3.6.1, Implementation of Status Trees shall begin when directed by the initial emergency response procedure, or when a transition is made to another emergency procedure. It would be plausible that CSF Status trees implementation would be consistent with when AOPs could be implemented. The second part is incorrect. This is plausible because that is consistent with standard AOP implementation where all applicable steps are expected to be completed.

B. INCORRECT The first part is correct. The second part is incorrect. This is plausible because that is consistent with standard AOP implementation where all applicable steps are expected to be completed.

C. CORRECT See above D. INCORRECT Parallel procedure usage is allowed. This is plausible because of a statement in OP-AP-104 related to when other procedures can be implemented. Step 3.6.1, Implementation of Status Trees shall begin when directed by the initial emergency response procedure, or when a transition is made to another emergency procedure. It would be plausible that CSF Status trees implementation would be consistent with when AOPs could be implemented. The second part is correct.

Title:

ATWS K/A:

029EG2.4.8 Knowledge of how abnormal operating procedures are used in conjunction with EOPs Technical

References:

OP-AP-104, Emergency and Abnormal Operating Procedures (rev 2) i-FR-S.i, Response to Nuclear Power Generation/ATWS (rev 17)

O-AP-1 0, Loss of Electrical Power (rev 76)

References provided to applicants: None Learning Objective:

U13593 Explain the guidelines for using the following types of procedures (OP-AP-104; S ER-i 999-2)

Abnormal procedures

Emergency operating procedures SRO-only: 1OCFR55.43(b)(5)

Knowledge of when to implement, hierarchy of implementation, and coordination of abnormal and emergency procedures.

79. 079- 038EA2.17 001/MODIFIED/NAPS BANKIHIGH/2/3.8/4.4//NO Given the following:

Unit2wasinMODEl

RCPs were secured in E-O due to RCP Trip Criteria being met

C SG has been identified as Ruptured and was isolated

RCS cooldown to target temperature is complete

RCS depressurization is complete

Normal charging and letdown are in service

Pressurizer Level is 52%

RVLIS upper range level is 89%

RCS Subcooling is 55°F

RCP cooling and seal flows are normal

Ruptured SIG level is 56% NR and stable In accordance with 2-E-3, Steam Generator Tube Rupture, what is the correct required action and procedure path regarding RCP restart and why?

A. Do NOT start an RCP The required RCP restart conditions for Subcooling are not met. In accordance with 2-E-3, initiate Attachment 2, Natural Circulation Verification.

B. Start 2-RC-P-1A The A RCP is preferred because the rupture is located on the C SG. Use 2-E-3 to start the A RCP, prerequisites and limitations outside of 2-E-3 do not apply.

C Do NOT start an RCP RVLIS level indicates a steam bubble in the upper head region which may condense during the RCP start. Raise pressurizer level and continue on in 2-E-3.

D. Start 2-RC-P-1C Normal pressurizer spray is desired. Use 2-OP-5.2, Reactor Coolant Pump Startup and Shutdown, and observe all precautions, prerequisites and limitations.

Distractor Analysis:

A) INCORRECT RCP restart conditions are met for subcooling (50°F is required). This is plausible because a high value for subcooling is required. This is plausible because the examinee will need to evaluate all the given conditions to determine if restart is acceptable.

B) INCORRECT This is incorrect because RVLIS conditions are not met to start an RCP. This is plausible because cautions in E-3 discuss using non-ruptured SGs. Also, other emergency procedures such as FR-C.1 do not require NOP prerequisites and limitations to be followed.

C) CORRECT 2-E-3 sets the value for additional required actions as less than 95% RVLIS upper range indication. Based on the intial condition of 89% RVLIS level, 2-E-3 directs raising Pressurizer level. These actions ensure prevention of saturated conditions to prevent the potential for SI reinitiation.

D) INCORRECT RCP restart conditions are NOT met, based on the given conditions, without having to take any additional operator actions. This is plausible because In accordance with 2-E-3, step 44 RNO, Try to start RCP to provide normal pressurizer spray. The preferred RCP to start if none are running is C RCP. 2-E-3 directs use of 2-OP-5.2 which requires observance of prerequisites and limitations.

Title:

Steam Generator Tube Rupture K/A 038EA2.1 7 Ability to determine or interpret the following as they apply to a SGTR: RCP restart criteria.

Technical

References:

Background E-3 2-E-3, Steam Generator Tube Rupture (rev 30)

References provided to applicants: None Learning Objective:

13886 Explain the following concepts associated with Reactor Coolant System cooldown to cold shutdown:

Why operation of a reactor coolant pump is desired during the cooldown

Why spreading contamination from the ruptured steam generator should be

___________

minimized SRO-only: 1OCFR55.43(b)(5)

Knowledge of when to implement attachments and appendices, knowledge of diagnostic steps and decision points that involve transitions.

Original Question lD#60225 A steam generator tube rupture has occurred. The crew is in 1-E-3 and has completed an RCS cooldown. The following plant conditions exist.

Reactor Coolant System pressure is 915 psig

CETC temperature is 500°F

Pressurizer level is 25%

1-RC-PT-1456 failed high and actions of l-AP-44 were successfully completed

Ruptured steam generator pressure is stable at 1065 psig

Non-ruptured SG pressures are stable at 650 psig

Ruptured steam generator level is 98%

All RCPs were previously secured due to low subcooling Based on the above information, the crew should A. transition to 1-ECA-3.1, SGTR With Loss of Reactor Coolant Subcooled Recovery

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Desired B. remain in 1-E-3, Steam Generator Tube Rupture C. transition to 1-E-2, Faulted Steam Generator D. transition to 1-ECA-3.3, SGTR Without Pressurizer Pressure Control Answer: A

80. 080 077AG2.2.44 00 !BANKJNAPS/HIGH/4/3 .5/3.6/I

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Given the following:

Both Units are at 100% power

The 1 H Diesel Generator is tagged out for planned maintenance The following alarms and indications occur:

Annunciator 1-T-D3, FREQUENCY 59.8 HERTZ

Annunciator 2-T-D3, FREQUENCY 59.8 HERTZ

Operators have verified the alarms are valid

500 KV switchyard voltage is 498 KV and stable Based on these plant conditions, which of the following identifies the MOST limiting Technical Specification required action? (Reference Provided)

A. Restore one offsite circuit to operable status within 24 hours1 days 0.143 weeks 0.0329 months B. Restore 1 H EDG to operable status within 14 days C. Restore either the offsite circuit or the 1 H EDG to operable within 12 hours0.5 days 0.0714 weeks 0.0164 months D Enter LCO 3.0.3 immediately

Distractor Analysis:

Based on the given conditions, the student will have to recognize that 498KV renders offsite circuits inoperable. LCO 3.8.1 (provided) is used to determine required actions.

The inoperable offsite circuits and the inoperable 1 H Diesel Generator result in LCO 3.8.1 Condition M.

A) INCORRECT Plausible since it would be correct if it were only the two offsite circuits were inoperable.

B) INCORRECT Plausible since it would be correct if the offsite circuits were operable (i.e. candidate is unaware that 498kV renders SWYD, and thus both offsite circuits inoperable).

C) INCORRECT Plausible since it would be correct if it were an EDG and ONE offsite circuit inoperable.

D) CORRECT Conditions (swyd voltage) render two required offsite sources inoperable and with the EDG tagged out three LCO 3.8.1 a. & b AC sources are inoperable. Required Action M states enter LCO 3.0.3.

Title:

Generator Voltage and Electric Grid Disturbances KIA:

077AG2.2 .44 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.

Technical

References:

Tech Spec LCO 3.8.1 0-AP-8, Response to Grid Instability (rev 10)

References provided to applicants:

Technical Specification LCO 3.8.1 (no bases)

Learning Objective:

16298 Explain the following concepts associated with the AC sources operating technical

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specification and bases (TS-3.8.1)

Accident/transient for which protection is afforded Limiting condition for operation

Applicability

Required actions

Surveillance requirements Question History: NAPS 2009 Retake Exam SRO-only: 10 CFR 55.43(b)(2)

Knowledge of facility operating limitations in the TS and their bases. Application of required actions and surveillance requirements.

81. 081 - WE1 1EG2.4.3 001/NEW/N/A/LOW/4/3.7/3.9//

Given the following:

Unit 1 was in MODE 1 when a Small Break LOCA occurred

Several instruments have failed or indications have become erratic

1-ECA-1 .1, Loss of Emergency Coolant Recirculation, is in progress

The SRO is evaluating which instruments per T.S. LCO 3.3.3, Post Accident Monitoring Instrumentation remain available for 1-ECA-1 .1 execution Which of the following instruments evaluated in 1-ECA-1.1 are qualified as Post Accident Monitoring Instruments in accordance with T.S. LCO 3.3.3, Post Accident Monitoring Instrumentation?

A. Pressurizer Pressure and Low Head SI Pump Flow B. Pressurizer Level and Low Head SI Pump Flow C. Pressurizer Pressure and CETC5 D Pressurizer Level and CETC5

Distractor Analysis:

1-ECA-1.1 procedural steps contain references to Post Accident Instrumentation contained in LCO 3.3.3 table 3.3.3-1. This Tech Spec table is considered SRO level knowledge.

Table 3.3.3-1 contains (in part):

High Head Safety Injection Flow

Pressurizer Level

RCS Pressue (Wide Range)

Core Exit Temperature A) INCORRECT RCS Wide Range Pressure is a LCO 3.3.3 post accident indication. Pressurizer Pressure is plausible because it is an indication used in 1-ECA-1.1. This is also the most common primary pressure indication used by operators. Low Head SI Pump flow is plausible because it is used in 1-ECA-1.1 and High Head SI flow is listed in Table 3.3.3-1.

B) INCORRECT Pressurizer Level is correct. Low Head SI Pump flow is plausible because it is used in 1-ECA-1.1 and High Head SI flow is listed in Table 3.3.3-1.

C) INCORRECT RCS Wide Range Pressure is a LCO 3.3.3 post accident indication. Pressurizer Pressure is plausible because it is an indication used in 1-ECA-1.1. This is also the most common primary pressure indication used by operators. CETCs are correct.

D) CORRECT See above

Title:

Loss of Emergency Coolant Recirculation K/A:

WE1 1 EG2.4.3 Ability to identify post-accident instrumentation.

Technical

References:

Tech Spec LCO 3.3.3, Post Accident Monitoring Instrumentation 1-ECA-1.1, Loss of Emergency Coolant Recirculation (rev 19)

References provided to applicants: None Learning Objective:

13840 Explain the following concepts associated with depressurizing the Reactor Coolant System in order to minimize subcooling (1-ECA-1.1)

Purpose of minimizing subcooling

Why pressurizer level is used as termination criterion SRO-only: 10 CFR 55.43(b)(2)

Knowledge of Technical Specifications tables and bases regarding post accident instrumentation.

82. 082 005AG2.2.25 001/NEW/N/A/LO W/2/3 2/4.2//NO

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Given the following:

Unit 1 is at 100% power

Surveillance 1-PT-17.1, Control Rod Operability, is in progress

Control Bank D Control Rod B8 failed to move as required during the surveillance and is currently at 225 steps withdrawn

All other rods in Control Bank D are at 218 steps

l&C troubleshooting has determined Rod Control is functioning properly electrically In accordance with the Technical Specification Required Actions and the COLR what is the minimum Shutdown Margin value that is required to be verified and the Bases for the LCO requirement that is currently not met?

A 1.77% Ak/k Ensure that upon reactor trip, the assumed reactivity will be available and will be inserted.

B. 1.00% Ak/k Ensure that the RCCAs and banks maintain the correct power distribution and rod alignment.

C. 1.77% Ak/k Ensure that the RCCAs and banks maintain the correct power distribution and rod alignment.

D. 1.00% Ak/k Ensure that upon reactor trip, the assumed reactivity will be available and will be inserted

Distractor Analysis:

The value of 1 % Dk/k is plausible and also is the value from Technical Specification 3.1.2 for measured Core Reactivity. LCO 3.1.2 states The measured core reactivity shall be within (+1-) 1 % Dk/k of predicted values Given the values of CBD and stuck Rod B8, all alignment limits are met.

T.S. LCO 3.1.4 contains two parts:

All shutdown and control rods shall be OPERABLE and Individual indicated rod positions shall be within 12 steps of their group step counter demand position From Technical Specification LCO 3.1.4 Bases (LCO Section) is the following paragraph.

The requirements on rod OPERABILITY ensure that upon reactor trip, the assumed reactivity will be available and will be inserted. The rod OPERABILITY requirements (i.e., trippability) are separate from the alignment requirements which ensure that the RCCAs and banks maintain the correct power distribution and rod alignment.

A) CORRECT The COLR states that the value for SDM shall be = 1.77% DkIk. Technical Specification LCO 3.1.4 Bases (LCO Section) states, The requirements on rod OPERABILITY ensure that upon reactor trip, the assumed reactivity will be available and will be inserted.

B) INCORRECT The value of 1% Dk/k is plausible and also is the value from Technical Specification 3.1.2 for measured Core Reactivity. LCO 3.1.2 states The measured core reactivity shall be within (+1-) 1% Dk/k of predicted values. This answer choice contains the bases for alignment requirements. The rod OPERABILITY requirements (i.e.,

trippability) are separate from the alignment requirements which ensure that the RCCAs and banks maintain the correct power distribution and rod alignment.

C) INCORRECT 1.77% Dk/k is the correct minimum Shutdown Margin value. However the answer choice contains the bases for alignment requirements. The rod OPERABILITY requirements (i.e., trippability) are separate from the alignment requirements which ensure that the RCCA5 and banks maintain the correct power distribution and rod alignment.

D) INCORRECT The value of 1% Dk/k is plausible and also is the value from Technical Specification 3.1.2 for measured Core Reactivity. LCO 3.1.2 states The measured core reactivity shall be within (+1-) 1 % Dk/k of predicted values. This answer choice contains the correct Bases for rod OPERABILITY.

Title:

Inoperable/Stuck Control Rod K/A:

005AG2 .2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

Technical

References:

North Anna Core Operating Limits Report COLR-N1C24

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Technical Specification and Bases LCO 3.1.4 (Amendment 231) 1-PT-17.1 Control Rod Operability (T.S. Surveillance 3.1.4.2) (rev 33)

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1-AP-1.3 Control Rod Out Of Alignment (rev 14)

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Refererences provided to applicants: None Learning Objective:

7277 Explain the following concepts associated with the rod group alignment limits technical specification and bases (TS-3.1.4)

Accident/transient for which protection is afforded

Limiting condition for operation

Applicability

Required actions

Surveillance requirements SRO-only: 10 CFR 55.43(b)(2)

Knowledge of Tech Specs and bases that are required to analyze Tech Specs.

83. 083 024AA2.04 OO1INEW/N/A/HIGH/2/3.4/4.2//NO

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Given the following:

Unit 1 was at 100% power when a Reactor Trip occurred

1 -ES-0. 1, Reactor Trip Response, is in progress

The crew is establishing emergency boration due to four control rods indicating greater than 10 steps.

Based on current plant conditions, in accordance with the Technical Requirements Manual (TRM) for Boration Flow Paths, the minimum required BAST borated water volume is (1).

In accordance with the TRM basis for Boration Flow Paths, This expected boration capability requirement occurs at (2)_ from full power equilibrium xenon conditions.

A. (1) 1378 gallons (2) end of life B. (1) 1378 gallons (2) beginning of life C. (1) 6000 gallons (2) beginning of life D (1) 6000 gallons (2) end of life

Distractor Analysis:

Based on the given conditions, the candidate will have to recognize that the unit is in Mode 3 and TRM 3.1 .1, Boration Flow Paths Operating applies. In this MODE, TSR

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3.1.1.3 states:

Verify required BAST borated water volume is > 6000 gallons.

The basis for this TRM states:

.to provide a SHUTDOWN MARGIN as specified in the COLR from expected operating conditions after XENON decay and cooldown to 200°F This expected

.

boration capability requirement occurs at end of life from full power equilibrium xenon conditions and requires 6000 gallons...

A) INCORRECT This is plausible because based on the reactor trip, the candidate may incorrectly identify TR 3.1.2 for Boration Flow Paths Shutdown. The required BAST water

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volume forTR 3.1.2 is 1378 gallons. End of life is correct.

B) INCORRECT This is plausible because based on the reactor trip, the candidate may incorrectly identify TR 3.1.2 for Boration Flow Paths Shutdown. The required BAST water

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volume for TR 3.1 .2 is 1378 gallons. BOL is plausible due to the amount of Kexcess present at the beginning of core life.

C) INCORRECT 6000 gallons is correct. BOL is plausible due to the amount of Kexcess present at the beginning of core life.

D) CORRECT See above

Title:

Emergency Boration K/A:

024AA2.04 Ability to determine and intepret the following as they apply to the Emergency Boration:

Availibility of BWST.

Technical

References:

TRM 3.1.1 and basis TRM 3.1.2 and basis 1-ES-O.1, Reactor Trip Response (rev 31)

References provided to applicant: None Learning Objective:

17476 Explain the following concepts associated with the Boration Flow Paths - Operating technical requirement and bases (TR-3.1 .1)

Accident /transient for which protection is afforded

Technical Requirement

Applicability

Required Actions

Surveillance Requirements SRO-only: 10 CFR 55.43(b)(2)

Knowledge of Technical Requirements Manual and bases that are required to analyze conditions, determine applicability and describe technical basis.

84. 084 033AA2.02 001/BANKJWATTS BAR 2006/HIGH/3/3.3/3.6//

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Given the following:

A Reactor Startup is in progress per 1-OP-i .5, Unit Startup from MODE 3 to MODE 2

No change in boron concentration or control rod movement is in progress and all nuclear instrumentation readings have been stable for the last five minutes.

Operators are preparing to deenergize the source range detectors

Nuclear Instruments read as follows:

SR Channel I (N-31) 1.0x10 4 cps

SR Channel II (N-32) 1.3x10 4 cps

IR Channel I (N-35) 1.1x10 8 amps

IR Channel II (N-36) 1.1x10 10 amps Which one of the following correctly identifies the problem with the Intermediate Range Monitors and the Required Action in accordance with Technical Specification (TS) 3.3.1, Reactor Trip System Instrumentation?

A. IR Channel I (N-35) indicates higher than expected Suspend all actions involving reactivity additions until both IR Channels are OPERABLE.

B IR Channel I (N-35) indicates higher than expected Initiate actions to increase thermal power to greater than P-i 0, Power Range Neutron Flux.

C. IR Channel II (N-36) indicates lower than expected Suspend all actions involving reactivity additions until both IR Channels are OPERABLE.

D. IR Channel II (N-36) indicates lower than expected Initiate actions to increase thermal power to greater than P-iO, Power Range Neutron Flux.

Distractor Analysis:

The Intermediate Range Nis come on scale at 10-11 amps. in accordance with the Excore chart overlap range (Graphic No: LD45) when Source Range is reading 1 .Oxl o cps the Intermediate Range should be between 10-11 amps and 10-10 amps. This was also validated using the most recent startup data. At the Source Range values listed in the question, N-35 is reading to high. A student can also use their knowledge of SR Trip blocks to aid in the understanding of the Excore overlap ranges. P-6 is lxi 0-10 amps IR.

In accordance with Technical Specification LCO 3.3.1 condition F, the correct action for One Intermediate range Neutron Flux channel inoperable is to: Reduce Thermal Power to < P-6 OR Increase Thermal Power to> P-l0.

A) INCORRECT N-35 is reading higher than expected based on current source range NI indications (see above.) Suspend all actions involving reactivity additions is plausible as that is a common required action during lower mode conditions.

B) CORRECT See above C) INCORRECT N-36 is reading consistent with the Source Range NIs. Suspend all actions involving reactivity additions is plausible as that is a common required action during lower mode conditions.

D) INCORRECT N-36 is reading consistent with the Source Range NIs. The action to initiate actions to increase thermal power to greater than P-i 0 is correct.

Title:

Loss of Intermediate Range NI K/A:

033AA2 .02 Ability to determine and interpret the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation: Indications of Unreliable Intermediate-Range channel operation.

Technical

References:

1-OP-i .5, Unit Startup from MODE 3 to MODE 2 (85)

Technical Specification LCO 3.3.1 Ex-core Nuclear Instrumentation System Student Guide References provided to applicants: None Learning Objective:

7789 Explain the following concepts concerning the ex-core monitors.

Ranges of indication

Where each channels output can be monitored 9685 Explain the following concepts associated with the intermediate range neutron flux function of the reactor trip system (RTS) instrumentation technical specification and bases (3.3.1)

Accident/transient for which protection is afforded

Limiting conditions for operation

Applicability

Required Actions

Surveillance requirements SRO-only: 10 CFR 55.43(b)(2)

Knowledge of Technical Specifications application of required actions and bases.

85. 085 WE I 6EG2.4.3 5 001 /MODIFIED/NAPS/LOW/2/3 .8/4.0/7

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Given the following:

Unit 2 is shutdown for refueling

Core on-load is in progress

High and Hi-Hi alarms have just been received on 2-RM-RMS-262, Manipulator Crane

Radiation levels are rising in containment

The refueling SRO reports that a fuel assembly being inserted into the core appears damaged

HP has determined that radiological conditions are acceptable for the containment closure team In accordance with 2-LOG-18, Containment Boundary Breach Log, the maximum time allowed for containment closure is _(1) following the decision to isolate containment.

and The operator assigned to containment closure duties is responsible for closure of the (2).

A. (1) 15 minutes; (2) personnel hatch B. (1) 45 minutes (2) equipment hatch C. (1) 15 minutes (2) equipment hatch D (1) 45 minutes (2) personnel hatch

Distractor Analysis:

2-LOG-18, Precautions and Limitation Operations will ensure that Containment Closure team roster (including Operations personnel hatch operator) will be recorded in the Operations Narrative Logs.

Step 5.9, During movement of irradiated fuel in Containment, to maintain the concept of defense in depth and assuming acceptable radiological protection conditions exist after a fuel handling accident in containment, containment closure will be established within 45 minutes following the decision to isolate containment.

Attachment 2: Containment Closure Flow Chart Fuel Handling Accident > 45 minutes following the decision to isolate containment.

-

A) INCORRECT 15 minutes is incorrect. This is plausible because it correlates to a short duration period that approximates a time which may used during the offload for containment closure. The personnel hatch is correct.

B) INCORRECT 45 minutes is correct. The equipment hatch is referenced in 2-LOG-I 8, however maintenance is responsible to ensuring/verifying closure.

C) INCORRECT 15 minutes is incorrect. This is plausible because it correlates to a short duration period that approximates a time which may used during the offload for containment closure. The equipment hatch is referenced in 2-LOG-i 8, however maintenance is responsible to ensuring/verifying closure.

D) CORRECT See above Title:

High Containment Radiation K/A:

WE1 6EG2.4.35 Knowledge of local auxiliary operator tasks during emergency and the resultant operational effects.

Technical

References:

O-AP-30, Fuel Failure During Handling (rev 14) 2-LOG-18, Containment Boundary Breach Log (rev 10) 2-OP-i 8.1, Operation of the Containment Personnel Air Lock (rev 20)

GMP-GM-103, Establishing and Maintaining Containment Closure Team During Unit Outage. (rev 1)

References provided to applicants: None

Learning Objective:

18032 Perform the following actions of 0-AP-30, Fuel Failure During Handling.

Explain the purpose

Identify the modes of applicability and/or plant conditions

Recognize the symptoms and entry conditions

List the immediate operator actions

Apply applicable Tech Specs/TRMs/EALs/Reportability

Explain the high level actions, major action categories, key mitigating strategies, and their basis.

Recognize plant conditions that result in a transition to or from 0-AP-30.

SRO-only: 10 CFR 55.43(b)(7)

Fuel handling facilities and procedures. Assessment of admin requirements related to containment closure following a fuel handling accident.

Original Question:

ID 7357:

Unit 2 is shutdown for refueling

Core on-load is in progress

High and high-high alarms have just been received on 2-RM-RMS-262, Manipulator Crane.

Radiation levels are increasing in containment.

The Refueling SRO reports a fuel assembly being inserted into the core appears to be damaged.

HP has determined that radiological conditions are acceptable for the containment closure team.

Which one of the following lists required actions per 0-AP-30, Fuel Failure During Handling?

A. Containment closure must be established within 45 minutes B. Containment closure must be established within 60 minutes C. Containment closure must be established within 90 minutes D. Containment closure is NOT required Answer: A

86. 086 004A2. 18 00 1/NEWIN/A/HIGH/2/3. 1/3.1/!

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Given the following:

Unit 1 is at 100% power steady state

Letdown divert valve, I -CH-LCV-1 11 5A is full open

Actual VCT level is 20% and lowering

VCT Auto Make-up has NOT occurred Based on these conditions, which of the following correctly identifies the condition that is causing this event and the Tech Spec/TRM implications of this condition?

A. VCT Level Instrument, 1-CH-LT-1 112 failed high TR 3.3.9, Reg Guide 1.97 Instrumentation, is not met because a key variable that provides information to indicate the operation of safety systems has failed.

B. VCT Level Instrument, 1-CH-LT-1112 failed high LCO 3.3.2, ESFAS Instrumentation, is not met due to the loss of RWST auto swap over actuation instrumentation that protects charging pump suction.

C VCT Level Instrument, 1 -CH-LT-1 115 failed high TR 3.3.9, Reg Guide 1 .97 Instrumentation, is not met because a key variable that provides information to indicate the operation of safety systems has failed.

D. VCT Level Instrument, 1 -CH-LT-1 115 failed high LCO 3.3.2, ESFAS Instrumentation, is not met due to the loss of RWST auto swap over actuation instrumentation that protects charging pump suction.

Distractor Analysis:

If 1-CH-LT-1 115 fails high, the following would occur:

VCT high level alarm

Full divert of LCV-1 11 5A

VCT level will decrease and auto make-up will not occur

Auto swap over to RWST will not occur at 5%

Charging pump suction will be lost as VCT empties A) INCORRECT This is plausible because the student is required to understand which VCT level transmitters affect auto make-up. LT-1 112 failing high would have the same initial effect on LCV-1 11 5A, however auto make-up would occur at 21.5%. TR 3.3.9 is correct.

B) INCORRECT This is plausible because the student is required to understand which VCT level transmitters affect auto make-up. LT-1 112 failing high would have the same initial effect on LCV-1 11 5A, however auto make-up would occur at 21.5%. LCO 3.3.2 includes safety system actuations. This is plausible as LCO 3.3.2 contains RWST level instruments and suction swap over. This VCT level transmitter is not included in the table in 3.3.2, and the SI signal will still reposition charging pump suction to the RWST.

C) CORRECT LT-1115 failing high is correct. TR 3.3.9 is correct.

D) INCORRECT LT-1115 is correct. LCO 3.3.2 includes safety system actuations. This is plausible as LCO 3.3.2 contains RWST level instruments and suction swap over to the containment sump. This VCT level transmitter is not included in the table in 3.3.2, and the SI signal will still reposition charging pump suction to the RWST.

Title:

Chemical and Volume Control System K/A:

004A2.18 Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS system; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations. High VCT Level

References:

1-AR-C-Al, VCT HI-LO Level L-115 (rev 3) 1-AR-C-A4, VCT HI-LO Level L-112 (rev 3)

TR-3.3.9, Reg Guidel .97 Instrumentation (rev 97)

NA-DWG-000-1 1715-CH-012, CVCS VCT 1-CH-TK-2 level indication (rev 7)

References provided to applicants: None Learning Objective:

237 Describe the response of the volume control tank level control subsystem to each of the following level instrumentation failures (SOER-97-1, GL 2008-02)

Leak in the level transmitter reference leg

l-CH-LT-1 112 fails high

1-CH-LT-l 112 fails low

1-CH-LT-l 115 fails high

1 -CH-LT-1 115 fails low SRO-only: 10 CFR 55.43(b)(2)

Knowledge of Technical Requirements and Technical Specification specific instrumentation differences. This is below the line information in tables and relates to knowledge of Technical Specification Bases.

87. 087 013A2.03 001!BANK/NAPS/HIGH/2/4.4/4.7!/

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Given the following:

Initial Conditions:

Unit 1 is shutdown for a refueling outage

RCS cooldown and depressurization have commenced in accordance with 1-OP-3.7, Unit Shutdown From MODE 1 To MODE 5 For Refueling

RCS Temperature - 535°F

RCS Pressure - 1890 psig

All actions for current plant conditions have been completed Multiple failures have just occurred resulting in rapid depressurization of all SGs inside containment Which of the choices below correctly completes the following statements:

An automatic safety injection (1)_ occur.

and In accordance with 1-ECA-2.1, Uncontrolled Depressurization Of All Steam Generators, the Aux Feedwater flow to the SGs will be maintained at a minimum of

_(2).

A (1) will (2)100 GPM to each SG to prevent dry out B. (1) will (2) 340 GPM total to all SGs to maintain heat sink C. (1)willnot (2)100 GPM to each SG to prevent dry out D. (1)will not (2) 340 GPM total to all SGs to maintain heat sink

Distractor Analysis:

In accordance with 1-OP-3.7, for the current plant conditions at Step 5.92, WHEN RCS pressure is <1990 psig, THEN do the following: Place LOW PRZR Pressure SI Block Train A/B switch in BLOCK. When Tave < 543° then place HI Stm Flow SI block switches to block.

Based on these actions, the Lo-Lo Pressurizer Pressure SI and HI Stm Flow SI signals are blocked. However the High Containment Pressure SI signal cannot be blocked.

Based on all SGs depressurizing inside containment an SI signal will occur.

1-ECA-2.1 requires feed flow be reduced to 100 gpm to each SG. The basis for this step is to prevent SG dryout.

1-E-0 and 1-E-1 direct total AFW flow greater than 340 gpm until SG narrow range level is greater than 11% in at least one SG. This is also the limit in 1-F-0 for Heat Sink Red Path.

A) CORRECT see above B) INCORRECT An SI will occur. 340 gpm is plausible because it is the minimum requirement for Feed Flow in 1-E-0, 1-E-1 and 1-F-0. The basis of the 340 gpm is to maintain Heat Sink.

C) INCORRECT An SI will occur (see above) This is plausible because the Lo-Lo Pressurizer Pressure SI signal is blocked. 100 GPM total to prevent SG dry out is correct.

D) INCORRECT An SI will occur (see above) This is plausible because the Lo-Lo Pressurizer Pressure SI signal is blocked. 340 gpm is plausible because it is the minimum requirement for Feed Flow in 1-E-0, 1-E-1 and 1-F-U. The basis of the 340 gpm is to maintain Heat Sink.

Title:

Engineered Safety Features Actuation System (ESFAS)

K/A:

01 3A2.03 Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b) based Ability on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations; Rapid depressurization Technical

References:

i-ECA-2.1 and background, Uncontrolled Depressurization of All Steam Generators (rev 21)

Safety Injection System Student Guide Emergency Contingency Action Procedures Student Guide 1-E-0, Reactor Trip or Safety Injection (rev 46) 1-F-i, Loss of Reactor or Secondary Coolant (rev 27)

References provided to applicants: None Learning Objective:

7694 List the following information associated with automatic safety injection

Automatic safety injection initiation signals including coincidence and setpoint

Safety injection signals which may be manually blocked, including the conditions which must be present to allow blocking them.

Means provided in the control room to determine that safety injection has been actuated.

13842 List the following information associated with i-ECA-2.1, Uncontrolled Depressurization of All Steam Generators.

Purpose of the procedure

Modes of Applicability

Entry Conditions

Major Action Categories

Conditions that result in leaving the procedure SRO-only: 1OCFR55.43(b)(5)

Knowledge of the content of a procedure and the background information which supports specific procedure content.

88. 088 061G2.1 .28 001/BANKJHARRIS2008/LOW/3/4.1/4.1//

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What is the basis for the Emergency Condensate Storage Tank minimum volume required by T.S. LCO 3.7.6 - Emergency Condensate Storage Tank (ECST)?

A. Remove decay heat for 4 hours0.167 days 0.0238 weeks 0.00548 months in MODE 3 followed by a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months cooldown.

B Remove decay heat for 2 hours0.0833 days 0.0119 weeks 0.00274 months in MODE 3 followed by a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months cooldown.

C. Remove decay heat and maintain the unit in MODE 3 for 12 hours0.5 days 0.0714 weeks 0.0164 months .

D. Remove decay heat and maintain the unit in MODE 3 for 10 hours0.417 days 0.0595 weeks 0.0137 months .

Distractor Analysis:

T.S. LCO 3.7.6 Bases state:

The ECST level required is equivalent to a contained volume of >110,000 gallons, which is based on holding the unit in MODE 3 for 8 hours0.333 days 0.0476 weeks 0.011 months , or maintaining the unit in MODE 3 for 2 hours0.0833 days 0.0119 weeks 0.00274 months followed by a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months cooldown to RHR entry conditions within the limits of 100°F/hour.

A) INCORRECT This is plausible because the basis states 8 hours0.333 days 0.0476 weeks 0.011 months (see above), however this is for maintaining in MODE 3, not a combination of maintaining MODE 3 followed by a coo ldown.

B) CORRECT see above C) INCORRECT This is plausible because the basis states that the minimum volume is based on holding the unit in MODE 3 for one continuous time period. 12 hours0.5 days 0.0714 weeks 0.0164 months is a plausible time period consistent with the other times described times (8 hours0.333 days 0.0476 weeks 0.011 months and 4 hours0.167 days 0.0238 weeks 0.00548 months .)

D) INCORRECT This is plausible because the basis states that the minimum volume is based on holding the unit in MODE 3 for one continuous time period. 12 hours0.5 days 0.0714 weeks 0.0164 months is a plausible time period consistent with the other times described times (8 hours0.333 days 0.0476 weeks 0.011 months and 2 hours0.0833 days 0.0119 weeks 0.00274 months .)

Title:

Auxiliary I Emergency Feedwater System K/A:

061G2.1 .28 Knowledge of the purpose and function of major system components and controls.

Technical

References:

Tech Spec LCD 3.7.6 and bases Auxiliary Feedwater Student Guide References provided to applicants: None Learning Objective:

16284 Explain the following concepts associated with the emergency condensate storage tank (ECST) technical specification and bases (TS-3.7.6.)

Accident/transient for which protection is afforded

Limiting condition for operation

Applicability

Required Actions

Surveillance requirements SRO-only: 10 CFR 55.43(b)(2)

Knowledge of Technical Specifications and Bases

89. 089 G2.1 .7 001/NEW/N/A/HIGH/3/4.4!4.7//

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Given the following:

Both Units are at 100% power

1-SW-P-lA trips due to a fault

The crew responds as required to start another SW pump and restore SW flow

Current SW pump discharge pressure is 50 psig

All SW spray valves are open and SW spray bypass valves are closed The SRO is reviewing 0-OP-49.6, SW System Throttling Alignment, to evaluate if T.S.

LCO 3.7.8, Service Water System, is OPERABLE.

In accordance with T.S. LCO 3.7.8, the affected A Service Water System loop

_(1) OPERABLE and the basis for throttling is to (2)

A (1) is not (2) ensure adequate flow to the RS heat exchangers following an accident B. (1)isnot (2) prevent SW pump run out following an accident C. (1)is (2) ensure adequate flow to the RS heat exchangers following an accident D. (1)is (2) prevent SW pump run out following an accident

Distractor Analysis:

In accordance with 0-OP-49.6 (step 4.10.2)

To ensure proper Service Water flows are maintained with only three operable Service Water Pumps and both Service Water Headers A and B in service, the following conditions MUST be satisfied with only one Service Water Pump in service on each header, in accordance with Tech Spec 3.7.8:

All Service Water Spray Valves MUST be open

All Service Water Bypass Valves MUST be closed

The Service Water flow MUST be throttled at the CCHX5 so that each Service Water Pump Discharge pressure is at least 54 psig.

Administrative controls MUST be placed on the position of the Service Water valves used to throttle the four CCHXs

[CO 3.7.8 bases states:

A SW Loop is considered operable when either One SW pump is operable in an operable flowpath provided two SW pumps are operable in the other loop and SW flow to the CC heat exchanger is throttled...

The Bases for LCO 3.7.8 action is:

If one SW System loop is inoperable due to an inoperable SW pump, the flow resistance of the system must be adjusted within 72 hours3 days 0.429 weeks 0.0986 months by throttling component cooling water heat exchanger flows to ensure that design flows to the RS System heat exchangers are achieved following an accident.. ..In this configuration, a single failure disabling a SW pump would not result in loss of the SW System function.

A) CORRECT See above B) INCORRECT Based on the information provided regarding SW system parameters, at 54 psig, SW LCO 3.7.8 is not operable. The basis for throttling is incorrect but plausible. Pump run-out is a concern and would be prevented by throttling. However that is not the basis identified in [CO 3.7.8 bases.

C) INCORRECT This is plausible because the value stated is in the system parameters is 50 psig and the SW spray bypass valves are closed. The second part is correct.

D) INCORRECT This is plausible because the value stated is in the system parameters is 50 psig and the SW spray bypass valves are closed. The basis for throttling is incorrect but plausible. Pump run-out is a concern and would be prevented by throttling. However that is not the basis identified in [CO 3.7.8 bases.

Title:

Service Water K/A:

G2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Technical

References:

T.S. LCO 3.7.8 and bases O-OP-49.6, SW System Throttling Alignment (rev 21)

References provided to applicants: None Learning Objective:

16286 Explain the following concepts associated with the Service Water (SW) system technical specification and bases (TS-3.7.8)

Accident/transient for which protection is afforded

Limiting condition for operation

Applicability

Required actions

Surveillance requirements SRO-only: 10 CFR 55.43(b)(2)

Knowledge of Technical Specification and Bases.

90. 090 1 03G2. 1.20 001 /NEW/N/A/LOW/3/4.6/4.6//

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Given the following:

The unit was operating at 100% power when a Reactor Trip and Safety Injection occurred

The crew transitioned from l-E-1, Loss of Reactor or Secondary Coolant, to 1-ES-i .3, Transfer to Containment Sump Recirculation

Upon entry into i-ES-i .3, the STA reports:

The only running Quench Spray pump tripped

Containment pressure is 29 psia and rising

1-F-0, CSF Status Trees, indicates an orange path for Containment Which of the following is the correct required procedural flowpath?

A. Continue in i-ES-i .3. If a CSF Status Tree red path is reached perform 1-FR-Z.i in parallel.

B Continue in 1-ES-i .3. After completion of Cold Leg Recirc Alignment, transition to 1-FR-Z.1.

C. Transition to i-FR-Z.1. When one Quench Spray pump is restored, return to i-ES-i .3.

D. Transition to 1-FR-Z.i. When containment pressure is less than 28 psia, return to 1-ES-i .3.

Distractor Analysis:

In accordance with OP-AP-104 Section 3.6.1

If an ORANGE path is encountered, the remaining CSF Status Trees shall be monitored. If a RED path is not encountered, the Recovery Procedures in progress shall be suspended and the FR required by the ORANGE path shall be performed.

The appearance of a RED or ORANGE path CSF Status Tree usually implies that some Unit equipment is not available or is significantly degraded.

Certain procedures (e.g., ESs and ECAs) take precedence over the FRs.

Typically, a NOTE will notify the Operator not to implement the FR under specific conditions.

Upon entry into 1-ES-i .3, a note reads:

Step 1 through Step 9 should be performed without delay. FR5 should not be implemented prior to completion of Step 9. Containment Pressure less than 28 A) INCORRECT This is plausible based on the prioritization of Red path conditions. Conditions have continued to degrade and the Red Path signifies immediate attention is required. In the case of Containment FR Red Path, Containment Pressure will have reached 60 psia.

Performing procedures in parallel during an Emergency is required under different circumstances.

B) CORRECT See above C) INCORRECT This is plausible because if transition to 1-ES-i .3 had not occurred and i-E-1 were in effect then transition to 1 -FR-Z. 1, is the correct required path. Restoration of a Quench Spray pump would effectively mitigate the adverse condition in containment.

D) INCORRECT This is plausible because if transition to 1-ES-i .3 had not occurred and 1-E-i were in effect then transition to i-FR-Z.i, is the correct required path. Containment Pressure less than 28 psia is a decision point in i-F-0 that leads to a yellow path. This indicates i-FR-Z.1 is not required.

Title:

Containment K/A:

103G2.1 .20 Ability to interpret and execute procedure steps.

Technical

References:

OP-AP-104, Emergency and Abnormal Operating Procedures (rev 2) 1-F-U, Critical Safety Function Status Trees (rev 7) 1-ES-i .3, Transfer to Cold Leg Recirculation (rev 24)

References provided to applicants: None Learning Objective:

13593 Explain the guidelines for using the following types of procedures (OP-AP-104; SER-1 999-2.)

Abnormal procedures

Emergency operating procedures SRO-only: 1OCFR55.43(b)(5)

Knowledge of when to implement procedures, knowledge of the hierarchy of implementation, and the coordination of emergency procedures.

91. 091 016A2.02 OOIINEW/N!AJLOW/2/2.9/3.2//

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Given the following:

Unit 1 is at 100%

The RO notes that the pressure indication on the A SG PORV manual/auto station reads 0 psig.

l&C investigation finds that the power supply on the pressure instrument card is failed.

In accordance with T.S. LCO 3.7.4, Steam Generator Power Operated Relief Valves (SG PORV5), and basis, the A SG PORV is (1) and the accident that is the limiting event for the PORV is a (2)

A (1) Operable (2) SG tube rupture B. (1) Operable (2) Small Break LOCA C. (1) Inoperable (2) SG tube rupture D. (1) Inoperable (2) Small Break LOCA

Distractor Analysis:

In accordance with Tech Spec LCO 3.7.4 bases:

A SC PORV is considered OPERABLE when it is capable of providing controlled relief of the main steam flow and capable of fully opening and closing, remotely or by local manual operation on demand.

In the SGTR accident analysis presented in Reference 2, the SC PORVs are assumed to be used by the operator to cool down the unit to RHR entry conditions when the SGTR is accompanied by a loss of offsite power, which renders the condenser dump valves unavailable.

A) CORRECT See above B) INCORRECT Operable is correct. Small Break LOCA is plausible because LOCAs are discussed in the applicable safety analysis of other 3.7 components such as AFW and SW related to core decay heat removal.

C) INCORRECT Inoperable is plausible due to the requirement for other main steam system steam/pressure relieving components to operate in auto, for example Main Steam Safety Valves. SC Tube Rupture is correct.

D) INCORRECT Inoperable is plausible due to the requirement for other main steam system steam/pressure relieving components to operate in auto, for example Main Steam Safety Valves. Small Break LOCA is plausible because LOCA5 are discussed in the applicable safety analysis of other Tech Spec components such as AFW and SW related to core decay heat removal.

Title:

Non-Nuclear Instrumentation System (NNIS)

K/A:

01 6A2.02 Ability to (a) predict the impacts of the following malfunctions or operations on the NNIS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of power supply Technical

References:

Main Steam System Student Guide Tech Spec LCO 3.7.4 and bases References provided to applicants: None Learning Objective:

16282 Explain the following concepts associated with the steam generator power operated relief valves (SG PORVs) technical specification and bases (TS-3.7.4.)

Accident/transient for which protection is afforded

Limiting condition for operation

Applicability

Required actions

Surveillance requirements SRO-only: 10 CFR 55.43(b)(2)

Knowledge of Technical Specifications and Bases

92. 092 033G2.4.1 1 OO1INEW/N/A/LOW/2/4.0/4.2//

-

Given the following:

Both Units are at 100% Power

1-AR-E-C6, Spent Fuel Pit Lo Level is LIT

A Non-licensed Operator in the field reports Spent Fuel Pool (SFP) Level is 7 and lowering

0-AP-27, Malfunction of Spent Fuel Pit System, is entered

Spent Fuel Pool level continues to lower

Attempts to makeup to the SFP from either Units blender are unsuccessful In accordance with 0-AP-27, make up to the SFP from the Service Water System requires permission from the (1)

The Tech Spec basis for minimum SFP water level following a fuel handling accident is to (2)

A. (1) Shift Manager (2) reduce gamma exposure B. (1) Shift Manager (2) limit iodine release Ci (1) Manager Nuclear Operations (2) limit iodine release D. (1) Manager Nuclear Operations (2) reduce gamma exposure

Distractor Analysis:

In accordance with O-AP-27 step 12 RNO:

Obtain permission of Manager Nuclear Operations or Operations Manager on Call to use Service Water as a make-up source.

In accordance with LCO 3.7.16 Basis:

The minimum water level in the fuel storage pool meets the assumptions of iodine decontamination factors following a fuel handling accident. The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity.

A) INCORRECT The Shift Manager is incorrect but plausible. In accordance with OP-AA-100, Conduct of Operations, Shift Manager Responsibilities include:

Authorize configuration changes to plant equipment and systems

Assume ultimate responsibility for all reactivity changes

Take actions necessary to optimize safety system availability and plant reliability

Maintain oversight and control during abnormal and emergency conditions Reduce gamma exposure is incorrect but plausible. It would be a common misconception, based on OE regarding loss of refueling cavity level, that the basis of the Tech Spec for Spent Fuel Pool Level is radiological dose concerns.

B) INCORRECT The Shift Manager is incorrect but plausible (see above.) Limit iodine release is correct.

C) CORRECT See above D) INCORRECT Manager Nuclear Operations is correct. Reduce gamma exposure is incorrect but plausible. It would be a common misconception, based on OE regarding loss of refueling cavity level, that the basis of the Tech Spec for Spent Fuel Pool Level is radiological dose concerns.

Title:

Spent Fuel Pool Cooling K/A:

033G2.4.1 1 Knowledge of abnormal condition procedures.

Technical

References:

O-AP-27, Malfunction of Spent Fuel Pit System (rev 23) 1-AR-E-C6, Spent Fuel Pit Lo Level (rev 1)

LCO 3.7.16, Fuel Storage Pool Water Level References provided to applicants: None Learning Objective:

11661 List the following information associated with O-AP-27, Malfunction of Spent Fuel Pit System (SEN-171, OE-8410)

Purpose of the procedure

Modes of applicability

Entry conditions

Condition that would require the fuel building to be evacuated

Action required if fuel movement is in progress during malfunction

Effect that a low level in the spent fuel pit would have on the spent fuel pit cooling pumps

Possible causes of a low spent fuel pit level

Whose permission is required to fill the spent fuel pit using the Fire Protection System

How long before the spent fuel pit temperature would reach 200°F if worst-case conditions existed

Alternate spent fuel pit cooling methods SRO-only: 10 CFR 55.43(b)(2)

Knowledge of Technical Specifications and Bases

93. 093 045A2.13 001/NEW/N/A/HIGH/3/2.1/2.5//

-

Given the following:

Unit 1 Reactor power is 96%, steady state following turbine valve freedom testing

All systems are in automatic control

Turbine control is in IMP IN

The Valve Position Limiter is set at 100%

One Main Steam Dump valve, 1-MS-TCV-1408B, fails 100% open due to a valve positioner failure and is unable to be closed.

In response to this failure, Turbine governor valves will _(1) AND the SRO will direct the crew to (2).

A. (1) open (2) ramp the turbine only if reactor power exceeds 100%

B (1) open (2) ramp the turbine down to less than 96% reactor power C. (1) not change position (2) ramp the turbine only if reactor power exceeds 100%

D. (1) not change position (2) ramp the turbine down to less than 96% reactor power

Distractor Analysis:

In accordance with 1-AP-38, Excessive Load Increase reduce Reactor Power to the power level BEFORE the event started Initial Reactor Power was 96% before the steam dump failure.

When the turbine control is in IMP IN, first stage pressure has an input into governor valve position and will adjust to maintain reactor power. The source of the impulse signal is 1-MS-PT-i 32.

A) INCORRECT The Turbine operating at IMP In will result in the Turbine governor valves opening up.

Ramping the Turbine if reactor power exceeds 100% is plausible based on 1-AP-38 guidance. The procedure A/ER column states, Reactor Power less than or equal to

-

100% power and stable. The RNO states, .reduce power...

. .

B) CORRECT See above C) INCORRECT Not change position is plausible based on the typical 100% power alignment when operating in IMP OUT. Ramping the Turbine if reactor power exceeds 100% is plausible based on i-AP-38 guidance. The procedure NER column states, Reactor Power less than or equal to 100% power and stable. The RNO states, .reduce

- . .

power...

D) INCORRECT Not change position is plausible based on the typical 1 00% power alignment when operating in IMP OUT. Ramp the turbine down to less than 96% reactor power is correct.

Title:

Main Turbine Generator K/A:

045A2.13 Ability to (a) predict the impacts of the following malfunctions or operation on the MT/G system: and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Opening of the Steam Dumps at Low Pressure Technical

References:

1-AP-38, Excessive Load Increase (rev 18)

Student Guide for Main Turbine Control and Protection (75)

References provided to applicants: None Learning Objective:

8902 Explain the Reactivity effects on reactor power when turbine control is in IMP OUT.

8911 Explain the following concepts concerning the Turbine Control Systems impulse pressure feedback signal (IMP IN).

Purpose

Source of the signal

Turbine control mode required for IMP IN operation

How the impulse pressure feedback is placed in service

Three conditions that will automatically shift turbine control to IMP OUT SRO-only: 10 CFR 55.43(b)(1)

Conditions and limitations in the facility license related to potential maximum thermal power conditions.

94. 094 G2.1.35 001/NEW/N/A/LOW/3/2.2/3.9//

-

Given the following:

Unit 1 is in MODE 6

Core Alterations are in progress

Several fuel assemblies have severe bowing which requires use of the manipulator crane to operate with a safety interlock in bypass In accordance with OP-AA-100, Conduct of Operations, the controls the bypass key for the refueling interlocks and authorizes its use.

A. Operations Manager B. Supervisor of Operations Support C Refueling SRO D. Shift Manager

Distractor Analysis:

In accordance with OP-AA-1 00, Conduct of Operations, Attachment 6, Core Alteration Requirements

The Refueling SRO is responsible for overall supervision and coordination of refueling operations, including fuel movement.

The Refueling SRO controls the bypass key for the refueling interlocks and authorizes its use.

OP-AA-100 details Shift Manager Responsibilities

Assume ultimate responsibility for all reactivity changes

Authorize configuration status changes to plant equipment and systems, including maintenance and testing activities that affect Operations A) INCORRECT The Operations Manager is plausible based on the importance and increase in risk associated with refueling safety interlocks in bypass.

B) INCORRECT The Supervisor of Operations Support is the supervisor over the Fuel Handling group including the Fuel Handling Supervisor who has the ability to authorize bypass interlocks in accordance with 1-OP-4.15, Manipulator Crane.

C) CORRECT In accordance with OP-AA-1 00, Conduct of Operations, Attachment 6, Core Alteration Requirements

The Refueling SRO controls the bypass key for the refueling interlocks and authorizes its use.

D) INCORRECT Shift Manager is plausible based on the responsibilities of the Shift Manager described in OP-AA-100. (see above).

Title:

Conduct of Operations K/A:

G2.1 .35 Knowledge of the fuel-handling responsibilities of SROs.

Technical

References:

1-OP-4.15, Manipulator Crane OP-AA-100, Conduct of Operations Student Guide, Fuel Handling System References provided to applicants: None Learning Objective:

9007 Explain the following information associated with the manipulator crane bridge and trolley interlocks.

Relationship between movement of the bridge, trolley, and hoist simultaneously

How the bridge and trolley interlocks maintain the mast within the core area

How the bridge only interlock affects movement from the core area to the transfer canal area.

Individual whose permission is required before any crane interlock can be bypassed SRO-only: 10 CFR 55.43(b)(7)

Administrative requirements associated with Fuel Handling and Refuel Floor SRO responsibilities.

95. 095- G2.1.45 O01/NEW/N/A/HIGH/2/4.3/4.3//

Given the following:

Unit 1 is at 100% Rated Thermal Power (RTP)

Annunciator K-E6, Plant Computer Trouble alarms

The PCS calorimetric program is not functional

The Feedwater UFM remains functional In accordance with TR 3.3.10, Feedwater Ultrasonic Flow Meter Calorimetric, and associated basis, the alternate indications specified to verify thermal power are (1 )_.

Performance of the next required surveillance hand calorimetric heat balance (2)_.

(Reference Provided)

A. (1) Delta T and Power Range Nis (2) allows continued operation > 2893 MWt B (1) Delta T and Power Range NIs (2) requires power to be reduced <2893 MWt C. (1) Power Range NIs and Turbine First Stage Pressure (2) allows continued operation > 2893 MWt D. (1) Power Range NIs and Turbine First Stage Pressure (2) requires power to be reduced <2893 MWt

Distractor Analysis:

TR 3.3.10.b requires the PCS to be functional.

condition C states, PCS calorimetric program not functional for reasons other than condition A. Action C2.2.2: Reduce Thermal Power to <2893 MWt (98.4% RTP) by monitoring alternate power. The required completion time is: Prior to performing the next required power range channel calorimetric heat balance comparison per TS SR 3.3.1.2.

TR 3.3.10 bases for Ci, C.2.1, C.2 states: Thermal Power would be determined by monitoring alternate power indications using the power range nuclear instrumentation (N Is) and RCS loop delta Ts.

Operation at 100% may continue until the next required performance of TS SR 3.3.1.2, Calorimetric Heat Balance Calculation. If the computer calorimetric program is nonfunctional, a manual calorimetric heat balance calculation would be required to meet the requirements of TS SR 3.3.1.2.

If the PCS calorimetric program is not restored to FUNCTIONAL status prior to the performance of the next calorimetric required by TS SR 3.3.1.2, thermal power would be reduced <2893 MWt (98.4% RTP) and a manual calorimetric would be performed.

A) INCORRECT Delta T and Power Range NIs are correct. In accordance with the bases, thermal power would be reduced and a manual (HAND) calorimetric would be performed. This is plausible because the statement earlier in the bases that operation at 100% may continue.

B) CORRECT See above C) INCORRECT Power Range Nis are correct. Turbine first stage pressure is plausible because it is an indication of Reactor Power and would provide an alternate secondary measure of primary conditions. In accordance with the bases, thermal power would be reduced and a manual (HAND) calorimetric would be performed. This is plausible because the statement earlier in the bases that operation at 100% may continue.

D) INCORRECT Power Range Nis are correct. Turbine first stage pressure is plausible because it is an indication of Reactor Power and would provide an alternate secondary measure of primary conditions. Reactor Power cannot remain above 2893 MWt until the next required calorimetric surveillance (HAND) is performed.

Title:

Conduct of Operations K/A:

G2.1.45 Ability to identify and interpret diverse indications to validate the response of another indication.

Technical References K-E6, Plant Computer Trouble TR 33.10, Feedwater Ultrasonic Flow Meter Calorimetric 1-AP-42.1, Loss of Unit 1 Plant Computer System (PCS) 1-PT-24, Calorimetric Heat Balance (Hand Calculation) (rev 46)

References provided to applicants: Technical Requirements Manual TR 3.3.10 (no bases)

Learning Objective:

5127 Explain the following concepts associated with the feedwater ultrasonic flow meter calorimetric and bases (TR 3.3.10)

Accident/transient for which protection is afforded

Technical Requirement

Applicability

Required actions

Surveillance requirements SRO-only: 10 CFR 55.43(b)(1)I 1 OCFR55.43(b)(2)

Conditions and limitations in the facility license related to potential maximum thermal power conditions. Knowledge of Technical Requirements Manual and bases.

96. 096 G2.2.21 001/BANKJNAPS NRC 2009/LOW/2/2.9/4.1//YES

-

Given the following:

Upon completion of maintenance, a post maintenance test was performed but acceptance criteria was not met.

An Operability Determination is in progress.

In accordance with OP-AA-102, Operability Determination, the is responsible for approving prompt Operability Determinations (ODs).

A Shift Manager B. Operations Manager C. Engineering Manager D. OMOC (Operations Manager on Call)

Distractor Analysis:

OP-AA-1 02 states the responsibilities of the Shift Manager as:

5.2.3 The Shift Manager (SM) or designated Operations SRO is responsible for:

Performing immediate and prompt ODs for FAs for degraded or nonconforming SSCs.

A) CORRECT See above B) INCORRECT This is plausible based on the responsibilities of the Manager Nuclear Operations.

OP-AA-1 02 states the Manager Nuclear Operations is responsible for ensuring the CD process is implemented to permit timely disposition of SSC operability or functionality.

C) INCORRECT This is plausible based on the responsibilities of the Engineering Manager. OP-AA-1 02 states the Engineering Manager is responsible for providing technical basis support for ODs and FAs.

D) INCORRECT This is plausible based on OMOC responsibilities for an immediate Operability determination. Attachment 8 for lCD discussed prompt review by the OMOC.

Title:

Equipment Control K/A:

G2.2.21 Knowledge of pre- and post-maintenance operability requirements.

Technical

References:

OP-AA-102, Operability Determination (rev 11)

VPAP-2003, Post Maintenance Testing Program (rev 14)

References provided to applicants: None Learning Objective:

13573 Explain the responsibilities of qualified operators associated with determining operability of systems, structures, and components (OP-AA-102, SOER-98-1)

SRO-only: 10 CFR 55.43(b)(3)

Knowledge of the administrative processes related operating changes in the facility.

97. 097 G2.2.36 001/NEW/N/A/HIGH/3/3.1/4.2//

-

Given the following:

Unit 2 is in MODE 1

Unit 1 is in MODE 5 following completion of a refueling outage

1-QS-P-1A, A QS Pump, is out of service for additional maintenance required for an oil leak that was identified last shift

Unit 1 A SI Accumulator was just found to be inoperable due to water level

A risk evaluation has not been performed

Operations is preparing for a MODE change to MODE 4 In accordance with Technical Specification LCO 3.0.4, a MODE change to MODE 4...

A. is allowable now provided that all applicable LCO actions are entered at the time of entry into MODE 4 for inoperable equipment.

B requires only the Unit 1 A QS Pump be returned to service and declared OPERABLE.

C. requires only the Unit 1 A SI Accumulator be returned to service and declared OPERABLE.

D. requires both the Unit 1 A QS pump and the Unit 1 A SI Accumulator are declared OPERABLE before entry into MODE 4 is allowed.

Distractor Analysis:

LCO 3.0.4 states:

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

When the associated ACTIONS to be entered permit continued operation in the MODE or other specific condition in the Applicability for an unlimited period of time,

After performance of a risk evaluation...

When a specific value or parameter allowance has been approved by the NRC A) INCORRECT This is plausible because this is consistent with LCO 3.0.2. Where if an LCO is not met, the required actions shall be met.

B) CORRECT See above. The QS pump is required in MODE 4 to meet the LCO. However the SI accumulator is not applicable in MODE 4 and therefore a MODE change is acceptable.

C) INCORRECT This is plausible because the SI accumulator is part of the ECCS section of Tech Specs. One train of ECCS is required in MODE 4. It is plausible for a candidate to evaluate Quench Spray as not required for MODE 4 based on temperature and pressure conditions.

D) INCORRECT It is plausible that all inoperable equipment be restored to operable prior to a MODE change.

Title:

Equipment Control K/A:

G2.2.36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions of operations.

References:

Technical Specifications [CO 3.0.4 and Basis References provided to students: None Learning Objective:

7326 Given a copy of Technical Specifications and bases, apply the provisions of LCO 3.0.4 to specific plant conditions to ensure compliance with technical specifications (TS LCO-3.0.4)

SRO-only: 1OCFR55.43(b)(2)

Knowledge of application of generic LCO requirements (LCO 3.0.1 thru LCO 3.0.7)

98. 098 G2.3.6 OO1INEW/N/A/LOW/2/2.0/3.8//

-

Given the following:

0-OP-23.2, Waste Gas Decay Tank and Waste Gas Diaphragm Compressors, has been initiated in preparation for a waste gas release

Operations is reviewing the release permit with HP

In accordance with the Precautions and Limitations of O-OP-23.2, regarding the impending release...

During normal conditions of plant operations, radioactive gases should be provided with a minimum hold up of _(1) days except for low radioactive gaseous wastes resulting from purge and fill operations associated with Refueling and Reactor Startup.

Regarding 1-GW-FCV-101, Char Filt In Fm Decay Tks Cont, IF manual control is required, THEN it is to only be done with _(2)_ permission.

A. (1) 30 days (2) SRO B. (1)3odays (2) Operations Manager C (1) 60 days (2) SRO D. (1) 60 days (2) Operations Manager

Distractor Analysis:

In accordance with 0-OP-23.2:

P&L 4.5 During normal conditions of plant operations, radioactive gases should be provided with a minimum hold up of 60 days except for low radioactive gaseous wastes resulting from purge and fill operations associated with Refueling and Reactor Startup.

P&L 4.13 Regarding 1-GW-FCV-101, Char Filt In Fm Decay Tks Cont, IF manual control is required, THEN it is to only be done with SRO permission.

A) INCORRECT 30 days is plausible because this a common long term time frame. SRO is correct.

B) INCORRECT 30 days is plausible because this a common long term time frame. Operations Manager is plausible because the Ops Manager or Operations Manager on Call permission is required in other cases where changing plant configuration can have a significant. For example aligning SW system to an unthrottled conditon.

C) CORRECT See above D) INCORRECT 60 days is correct. Operations Manager is plausible because the Ops Manager or Operations Manager on Call permission is required in other cases where changing plant configuration can have a significant. For example aligning SW system to an unthrottled conditon.

Title:

Radiation Control K/A:

G2.3.6 Ability to approve release permits Technical

References:

O-OP-23.2, Waste Gas Decay Tank and Waste Gas Diaphragm Compressors (rev 24)

References provided to applicants: None Learning Objective:

11984 Given a set of plant conditions, evaluate Gaseous Waste Disposal System operations in light of the following issues:

Effect of a failure, malfunction, or loss of a related system or component on this system

Effect of a failure, malfunction, or loss of components in this system on related systems

Expected plant or system response based on gaseous waste disposal component interlocks or design features

Impact on the technical specifications

Response if limits or setpoints associated with this system or its components have been exceeded

Proper operator response to the condition as stated SRO-only: 10 CFR 55.43(b)(4)

Radiation hazards that may arise during normal and abnormal situations. Process for gaseous release.

99. 099 G2.4.32 001/NEW/N/A/HIGH/3/3.6/4.0//

-

Given the following:

09:00 Vital bus 1-I deenergizes due to a fault 09:00 Unit 2 Annunciator Panel F H-6, Unit #1 Ann Sys Power Supply Failure is LIT 09:03 Annunciator Loss is verified by the Control Room 09:04 1-AP-6, Loss of Main Control Room Annunciators, is entered 09:05 The Shift Manager is reviewing the EAL matrix In accordance with EPIP-1 .01, Emergency Manager Controlling Procedure, what is the latest time the applicable EAL SU4.1 can be declared which meets the maximum allowable classification time requirement?

A 09:15 B. 09:18 C. 09:20 D. 09:30

Distractor Analysis:

In accordance with the North Anna Power Station Emergency Plan:

Once indications are available to plant operators that an emergency action level has been exceeded, the event is promptly assessed and classified, and the corresponding emergency classification level is declared. This declaration occurs as soon as possible and within 15 minutes of when these indications become available.

SU4.1 Unplanned Loss of most (75%) or all of EITHER:

Annunciators (Panels A thru N)

Indicators associated with safety-related structures, systems and components on Unit 1 (Unit 2)

MCR bench board and Vertical Board for> 15 mins (NOTE 3)

NOTE 3: The SEM should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

A) CORRECT As of 09:00 indications are available that the EAL has been exceeded.

Declaration occurs as soon as possible and within 15 minutes of when these indications become available. 09:15 is correct for the maximum allowable time.

B) INCORRECT 09:18 is plausible based on the verification occurring at 09:03 with declaration maximum allowable time 15 minutes from that time.

C) INCORRECT 09:20 is plausible based on the Shift Manager evaluating EALs at 09:05 with declaration maximum allowable time 15 minutes from that time.

D) INCORRECT 09:30 is plausible based on the EAL threshold. SU4.1 is an unplanned loss for greater than 15 minutes. Declaration maximum allowable time is 15 minutes from the time the threshold is exceeded. However concurrent clock guidance and NOTE 3 do not allow an additional 15 minutes.

Title:

Emergency Procedures/Plans K/A:

G2.4.32 Knowledge of operator response to loss of all annunciators Technical

References:

EPIP-1 .01, Emergency Manager Controlling Procedure (rev 49)

EAL Technical Bases Document (rev 4) 1 -AP-6, Loss of Main Control Room Annunciators (rev 11)

North Anna Power Station Emergency Plan (rev 40)

References provided to applicants: None Learning Objective:

18005 Perform the following actions of 1-AP-6, Loss of Main Control Room Annunciators.

Explain the purpose

Identify the modes of applicability and/or plant conditions

Recognize the symptoms and entry conditions

List the immediate operator actions

Apply applicable Tech Specs/TRM5/EALs/Reportability

Explain the high level actions, major action categories, key mitigating strategies, and their basis

Recognize plant conditions that result in a transition to or from 1-AP-6 SRO-only: 10 CFR 55.43(b)(5)

Assessment of facility conditions and selection of procedure to mitigate, recover or proceed with. SRO only due to requiring candidate to have knowledge of EPIP execution.

100. 100 02.4.44 001/NEWIN/A/LOW/2/2.4/4.4//

-

Given the following:

An event has occurred in the station and the emergency plan has been entered

The Station Emergency Manager (SEM) has relieved the Shift Manager of SEM duties and has activated the TSC

The Recovery Manager (RM) has reported to the LEOF and LEOF activation is in progress

Responsibility for State and Local communications has not been transferred to the LEOF

Conditions in the plant have degraded to the point of requiring entry into a General Emergency In accordance with EPIP-1 .05 Response To General Emergency, the initial

notification to the State of the applicable PAR is required to be made within 15 minutes of (1)

AND At this time, the _(2)_ is responsible for determining the recommendation for offsite protective actions.

A. (1) declaring the General Emergency (2) Recovery Manager B. (1) General Emergency conditions being met (2) Station Emergency Manager C. (1) General Emergency conditions being met (2) Recovery Manager D (1) declaring the General Emergency (2) Station Emergency Manager

Distractor Analysis:

A note in EPIP-1.05 The initial notification of General Emergency classification and the applicable Protective Action Recommendation (PAR) must be made (meaning the state and local Emergency Operations Centers (EOCs) have been provided with the emergency classification level) within 15 minutes of declaring the emergency class.

A note in EPIP-1.05 The Shift Manager may be relieved as Station Emergency Manager lAW the NAPS Emergency Plan.

EPIP-3.02 WHEN the LEOF is activated, THEN do the following:

a. Transfer the following responsibilities to the Recovery Manager (RM), as practical (all responsibilities should be transferred at the same time):

Notifying State and Local governments of emergency status

Recommending off-site protective measures

Performing off-site dose projections

Providing radiological status to the NRC (after the NRC asks that the Health Physics Network (HPN) be established over ENS.

A) INCORRECT The initial notification of General Emergency classification and the applicable Protective Action Recommendation (PAR) must be made (meaning the state and local Emergency Operations Centers (EOC5) have been provided with the emergency classification level) within 15 minutes of declaring the emergency class. Based on the conditions provided above, The Station Emergency Manager still has the responsibility for recommending off-site protective measures because the LEOF has not completed activation.

B) INCORRECT This is plausible because the Emergency Classification is required to be made within 15 minutes of conditions being met. Based on the conditions provided above, The Station Emergency Manager still has the responsibility for recommending off-site protective measures because the LEOF has not completed activation.

C) INCORRECT This is plausible because the Emergency Classification is required to be made within 15 minutes of conditions being met. Based on the conditions provided above, The Station Emergency Manager still has the responsibility for recommending off-site protective measures because the LEOF has not completed activation. This is plausible because the Recovery Manager has reported to the LEOF and activation is in progress.

D) CORRECT See above

Title:

Emergency Procedures / Plan K/A:

G2.4.44 Knowledge of emergency plan protective action recommendations.

Technical

References:

EPIP-1 .05, Response to General Emergency (rev. 23)

EPIP-3.02, Activation of Technical Support Center (rev. 34)

References provided to applicants: None Learning Objectives:

14319 Evaluate a set of plant conditions associated with emergency plan implementing procedures in light of the following issues. (SRO)

Procedure entry conditions

Step bases

Proper procedure usage 12173 Given a set of plant conditions, determine the protective action recommendations.

SRO-only: 10 CFR 55.43(b)(5)

Assessment of facility conditions and selection of procedure to mitigate, recover or proceed with. SRO only due to requiring candidate to have knowledge of EPIP PARs.

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ML14233A497: Difference between revisions

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