oct 2,2006 nmp outline review comments · oct 2,2006 nmp outline review comments e e e m e e e e e...

Oct 2,2006

NMP Outline Review Comments

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RO Admin 1 - this task is overly leading and is effectively cuing the applicants that there is something wrong with this particular instrumentation. A more appropriate task would be to have the applicants take multiple log readings on one section of the panels for example, and then identify the problems. This is more operationally valid and does not cue an obvious problem. The licensee did not disagree and will look into revising the proposed task.

RO Admin 2 - Appears that the task may be overly simplistic. In discussions with the exam developer he indicated that the applicants would have to determine that the system electrical alignment was wrong for one or more components. Will withhold final judgment until the exam item is presented.

RO Admin 3 - Appears to be overly simplistic but will withhold final judgment until the exam item is presented.

JPM 7 - Appears to be overly simplistic but will withhold final judgment until the exam item is presented.

JPM 8 - Licensee plans to replace task based on his own personal concerns prior to submittal.

JPMs 9, 10 & 11 - Noted all proposed in plant JPMs were being taken from the licensee’s exam bank. Concerned with a relatively small in plant JPM bank may not be discriminating. Licensee proposed that he make JPM 10 alt path.

Scenario #1 - Typo in summary description should be 90% vice 100%.

Scenario #2, Event 8 - very lengthy - licensee indicated may not be required action. It would not be counted then.

Scenario #2 - Some verbage missing from event summary top of pg. 6 of 11.

General Comment - TS calls that do not involve operator actions will not count as a malfunction - i.e., other than for the TS calls for the SRO applicants acting in the CRS position only.

Written Exam, Tier 2/Grp 1, RO 13 - maybe better tested in dynamic exam will withhold final judgment until the exam item is presented.

Written Exam, Tier UGrp 1, RO 117 & 26 - cautioned electrical power supplies make sure you examine at the appropriate level not just asking power supplies.

0 Written Exam, Tier 2/Grp 1, RO 34 - EDG alarm set points - cautioned okay if exam item tests remote alarms but not CR alarms that are extensively examined during the dynamic exam portion.

0 Written Exam, Tier 2/Grp 1, RO 35 - make sure this a licensed operator task and has a valid LO not just NLO knowledge.

Written Exam, Tier 3, RO 2 - maybe be more appropriately tested on dynamic or JPM.

Written Exam, Tier 3, RO 9 - prioritizing EOPs implementation more appropriate for SRO topic.

Record of rejected WAS - reasons for rejections for several WAS in-appropriate e.g., NOT RO level topic 2.1.17 . Please go back and re-consider all justifications for rejection and ensure if rejecting a WA it is for a valid reason.

Written Exam, Tier 3, SRO 6 - recognizing EOPs entry conditions more appropriate for RO topic not SRO level knowlege.

Constellation Energy Nine Mile Point Nuclear Station

U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406- 14 1 5

ATTENTION: Regional Administrator Region I

P O Box63 Lycoming, NY 13093

September 22,2006

SUBJECT: Nine Mile Point Nuclear Station Unit No. 1; Docket No. 50-220

Nine Mile Point Unit 1 Initial Operator Examination Outline Submittal

REFERENCE: (a) Letter from Marvin D. Sykes (NRC), Senior Reactor and Reactor Operator Initial Examinations - (Nine Mile Point, Unit I ) , dated August 23,2006.

In response to Reference (a), arrangements were made for the administration of licensing examinations at Nine Mile Point Unit 1 during the week of December 11,2006. The examinations are being prepared based on the guidelines in Revision 9, of NUREG 102 1 , “Operator Licensing Examination Standards for Power Reactors.” Enclosed are the examination outlines and associated quality checklists for Reactor Operator and Senior Reactor Operator examinations.

The enclosures contain examination information and are exempt from public disclosure in accordance with 10 CFR 2.390. The enclosures are appropriately marked in accordance with 10 CFR 2.390. Please withhold these examination materials from public disclosure until after the examinations have been completed.

Nine Mile Point Nuclear Station has used the methodology outlined in ES-40 1 Attachment 1 ”Example Systematic Sampling Methodology.” The written examination outlines for the Nine Mile Point Unit 1 Reactor Operator and Senior Reactor Operator examinations and the topics were randomly generated using the method described in ES-40 1, Attachment 1. These outlines were then saved with password protection on a non-networked computer.

Should you have questions regarding the information in this submittal, please contact R. I. Brown, General Supervisor Operations Training, at (3 15) 349-453 1.

Very truly yours,

Thomas J(Shortel1 Training Manager

Regional Administrator Region I September 22,2006 Page 2

TJSIRF Enclosures

cc: M. D. Sykes, NRC Resident Inspector, NRC

J. G. Caruso, NRC

ES-301, Rev. 9 Ad minist rat ive Topics Outline Form ES-301-1

Facility: Nine Mile Point 1 Examination Level: RO

Ad m in istra tive Topic (see Note)

Conduct of Operations Plant para meter Verification

Conduct of Operations Shift Turnover RO ADMIN 2

Equipment Control Use of Electrical Prints RO ADMIN 3

Radiation Control Radiation Hazards RO ADMIN 4

D

N

N

Date of Examination: December 11, 2006 Operating Test Number: NA

Describe activity to be performed

Perform RPV Level Instrument Checks per N1-ST-DO, Daily Checks Given control room reactor water level instrument readings for various daily checks required by Technical Specifications (readings may be taken in the simulator), enter the instrument readings into the applicable sections of the Daily Checks and take appropriate actions based on those checks.

2.1 . I8 (2.9) Ability to make accurate I clear and concise logs / records I status boards I and reports.

Perform Control Room System Lineup Verification Evaluate conditions provided in S-PM-DO01 Control Room System Lineup Verification for Control Room Ventilation System and determine the required actions for out of position components. (Panel walkdown is performed in simulator).

GAP-OPS-01; 3.10, N1-ST-DO; 7.0

2.1.31 (4.2) Ability to locate control room switches /controls and indications and to determine that they are correctly reflecting the desired plant lineup.

S-PM-DO01 Attachment 2 2.24 RPS Logic Operation with EOP-3.1 RPS Jumpers Installed Identify and locate electrical drawings related to RPS circuits. Use the drawings to describe the effect of installing RPS jumpers on how scram signal is reset prior to initiating another manual scram, with the jumpers installed under failure to scram conditions. (Perform in simulator and supply with copies for candidate to mark-up).

2.1.24 (2.8) Ability to obtain and interpret station electrical and mechanical drawings. N1-EOP-3.1; Drawing C-19859-C SH 4 and 7

Radiological Requirements Related To Operator Inspection Of Rad And High Rad Areas.

Given radiological conditions and a survey map related to an area to be inspected, ensure the appropriate radiological aspects of the job are met prior to performance of the inspection.

2.3.10 (2.9) Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure. GAP-RPP-01; 3.5, 3.6, 3.7, GAP-RPP-02; 3.1, 3,3, GAP-RPP-08; 3.2, 3.3, NI-PM-M5; 6.0, 8.0

for SROs RO applicants require only 4 items unless they are retaking only

(D)irect from bank ( 3 for ROs, 4 for SROs & RO retakes)

9/25/2006 7: 12:30 AM I o f l 2006 Unit 1 NRC RO Admin Outline Submittal

ES-301, Rev. 9 Administrative Topics Outline Form ES-301-1

Facility: Nine Mile Point 1 II Examination Level: SR

Ad mi n is t ra tive Topic II (see Note)

Conduct of Operations SRO ADMIN 2

Equipment Control SRO ADMlN 3

Emergency Plan SRO ADMIN 5

TY Pe Code*

N

D

N

D

Date of Examination: Operating Test Number: NA

December 11, 2006

Describe activity to be performed

Assess Reportability Requirements For Maintenance Activity Given inaccurate log entries stating 10CFR50.72 and 73 requirements for a given maintenance activity, assess the log entries for accuracy and correctness and identify the correct reporting requirements. (CR NM-2006-4134) 2.1 .I8 (3.0) Ability to make accurate/ clear and concise logs/ records/ status boards/ and reports.

10CFR50.72 and 73. NUREG 1022 REV2

Determine Reactivity Event Severity Level And Supporting Actions In Response To The Event Given plant conditions determine the reactivity event severity level and the appropriate compensatory actions and notifications. 2.1.7 (4.4) Ability to evaluate plant performance and make operational judgments based on operating characteristics / reactor behavior / and instrument interpretation. GAP-OPS-05; 3.1 3

Review Reactor Engineering Thermal Limit Surveillance Given daily surveillance procedure with thermal limits exceeded, identify the Tech Spec required compensatory actions.

2.2.22 (4.1) Knowledge of limiting conditions for operations and safety limits.

N1-RESP-1A and Tech Specs ~~~ ~ ___ ~~ ~

Determine Actions Required For an Inoperable Effluent Radiation Monitor

Given plant conditions, determine operability of an effluent radiation monitor and apply action statements contained in the station ODCM.

2.3.1 1 (3.2) Ability to control radiation releases.

ODCM

Classify Emergency Event and Perform PARs Given plant conditions at an existing Site Area Emergency, determine the plant conditions now meet General Emergency conditions. Reclassify the event and determine PARs, based on conditions provided. (Time Critical) NRC 2002 SRO JPM A4

2.4.44 (4.0) Knowledge of emergency plan protective action recommendations.

EAL Matrix

(CR NM-2004-976)

NOTE All items (5 total) are required for SROs RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required

* Type Codes & Criteria (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs 4 for SROs & RO retakes)

(N)ew or (M)odified from bank ( 1)

9/21/2006 8:12:09 AM 2006 Unit 1 NRC SRO Admin Outline Submittal

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Facility: Nine Mile Point Unit 1 Date of Examination: December 11, 2006 Exam Level: RO SRO-I Operating Test No.: NRC EXAMINATION

11 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

11 System / JPM Title Safety 1 1 Function

JPM I Recircl Start a Recirc Pump At Power Proc N1-OP-1 H 4 0

REACTIVITY CONTROL

and original fan must be placed back in service). ,Proc N1-OP-10 F 1 2, F 2 2, F 3.land ARP L1-1-5 alarm KA 288000 A4 01 3 112.9

I Previous NRC 2002 JPM 4

RADIOACTIVITY RE LEAS E

JPM3 T G b Gen Auxl Place Generator Amplidyne In Service (Amplidyne output fails low and must be removed from service). Proc N1-OP-32 H.6.0, H 5.0

S N A 4

HEAT REMOVAL

ELECTRICAL

REACTOR WATER

CONTROL manual level control) INVENTORY

Proc Nl-PM-Q8, N1-OP-16 KA 259007 A4 03 3 813 6 - _ - - _ _ _ - _ - _ _

JPM 6 Primary ContainmentlVent the Drywell During Power Operation S N Proc N1-OP-9 H 1 0 CONTAINMENT

Scram ly. This JPM is not performed by SRO-I INSTRUMENTATION

8 PLANT SERVICE

SYSTEMS

JPM 8 RBCLC TBCLCl Rotate In Service RBCLC and TBCLC Pumps I S N A (After placing standby RBCLC pump in service and while placing standby TBCLC pump in service, RBCLC temperature control valve fails). Proc N1-OP-11 and N1-OP-24 KA 40000O A4 01 3 1/3 I]

LC1 05-01 NRC Exam 2006 09/21/06 8:lO AM

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Proc N1-EOP-1 Attachment 12 KA 295031 EA1 08 3 813 9 Bank/Previous Ol-OPS-PJE-1-2I/NRC 2002 JPM 9

REACTIVITY

COhTRO-

JPM 10 UPS/ Transfer UPS Loads from 162A to 1626 (Note: These are RPS system power supplies). Proc N1-OP-40 F . l .O KA 212000 A2.02 3.713.9

~ - 1 - 7 R D

INSTRUMENTATION

Valve (11 or 12) Proc N1-OP-16 H.l .O KA 259001 A2.05 3.0/3.0 Bank 01-OPS-PJE-259-1-08 (12 FCV)

I \ u REACTOR WATER

INVENTORY

CONTROL

I I @All control room (and in-plant) systems must be different and serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

Planned Simulator JPM Combinations JPM 1,2 and 8 JPM 3,4 (8 also can run with this set) JPM 5,6 JPM 7 standalone (RO Only)

LC1 05-01 NRC Exam 2006 09/22/06 1:53 PM

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ES-301, Rev. 9 Transient and Event Checklist Form ES-301-5

=acility: Nine Mile Point 1 Date of Exam: December 11,2006 Operating Test No.: All

A P P L I C A N T

RO

SRO-I

E Scenarios v

1 I 2 I 3 I 4 I T 1 M - t N T POSITION

CREW CREW POSITION CREW POSITION CREW POSITION

4,7, 3,5,7,a

2 1 1 1 1

3,6 4,7,9, 3,4,5 3,5,8 4,7,a 5 4 4 2 10 ,7,8

8 8 6 6 6 2 2 2 1

3,7 2 0 2 2

Instructions:

1) Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the “at-the-controls (ATC)” and “balance-of-plant (BOP)” positions; Instant SROs must do one scenario, including at least two instrument or component ( K ) malfunctions and one major transient, in the ATC position.

2 ) Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. r) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a 1 -for-I basis.

3. Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicant‘s competence count toward the minimum requirements specified for the applicant‘s license level in the right-hand columns.

Author:

NRC Reviewer:

LC1 05-01 NRC Exam 2006 09/22/06 2:38 PM

. . .

Event Malf. No. Event No. Type*

1 FWO2C C(TS SRO)

2 N

Appendix D, Rev. 9 Scenario Out I i ne Form ES-D-1

Event Description

Feedwater Booster Pump 13 auto trips. The pump is a HPCl component and must be declared inoperable. SRO enters TS 3.1.8 and the pump must be restored within 15 days.

Transfer pressure control from MPR to EPR per N1-OP-31, F.3.0.

11 i c i l i y : NMPI Scenario No.: 1

RP20B

Op-Test No.: NRC

I (TS SRO)

RRO9C

Drywell High Pressure Transmitter 201.2-476A fails downscale. Transmitter supplies input to RPS, Core Spray and Containment Spray Systems. SRO Tech Spec Entry into LCO 3.6.2 is required.

C (ALL)

R (RO)

5

6

7

8

9

4

MS05 C(B0P)

RR27 C (ALL)

R (SRO RO)

Override M(ALL)

MS04 C (ALL)

RP26B C(ALL)

Recirc Pump 13 Motor Generator Slot temperatures rise. Removal of the pump from service is required, which also requires a power reduction. Actions are taken for the Recirc Pump Trip per SOP-I .3

Steam Seal Regulator Failure. Power reduction reveals a pre- existing failure in the Steam Seal Regulator and results in degraded steam seal header pressure and increased condenser air in- leakage. Regulator Bypass must be manually opened to restore seal pressure.

Recirc Master Controller fails low resulting in Restricted Zone entry. Entry into SOP-I .5. Flow drops to 21 Mlbm/hr and power is 45-50% on APRMs. Cram Rods must be inserted to exit the restricted zone. Power must be reduced to about 30% power to exit region.

Loss of Condenser Vacuum due to Steam Seal Regulator Bypass Valve Failure. Enter SOP-25.1. A turbine trip is required when condenser backpressure exceeds 5 inches with generator load < I 90 MWe. Reactor scrams either manually or automatically.

Steam Leak in Drywell 20% ramp time 1O:OO minutes. After the scram and initial actions are complete, the steam leak develops. Drywell pressure exceeds 3.5 psig and EOP entry is required. Drywell parameters will reach values that require use of Containment Spray.

Drywell High Pressure Transmitter 201.2-476C fails downscale. With the “A” transmitter previously failed the high drywell pressure RPS scram signal, Core Spray and Containment Spray automatic initiation signals are prevented. Crew must take manual actions to initiate these functions.

RR99A I (ALL) 1 RR87 1 RPV level instrument reference legs flash. Crew is required to perform RPV Flooding. Event is classified as SAE 2.1.2

I I I

*(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

9/22/2006 1 :54:43 PM NRC Exam Outline Submittal

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Facility: Nine Mile Point I Scenario No.: NRC-01 Op-Test No.: NRC TARGET QUANTITATIVE ATTRIBUTES I ACTUAL (PER SCENARIO; SEE SECTION D.5.d)

1 . Total malfunctions (5-8) Events 4,5,6,8,9,10 2. Malfunctions after EOP entry (1-2) Events 9 and 10

Event 6 SOP-1.5 and Event 7 SOP-25.1

Event 7 Loss of Vacuum

3. Abnormal events (2-4)

4. Major transients (1 -2)

5. EOPs enteredhequiring substantive actions (1 -2)

Containment

actions (0-2)

Event 6 EOP-2 RPV; EOP-4 Pri

6. EOP contingencies requiring substantive

A ~ ~ R l B U T E S Total Malfunction Count: Major not included in this count. Didn’t count Event 1 and 3, because these only require SRO tech spec use.

Abnormal Events Count: Does not include the SRO TS related events. These are considered separately.

SRO TS Events Event 1 and 3 are SRO Tech Spec evaluation events.

6

2

2

1

2

~

1

7. Critical tasks (2-3) CRITICAL TASK DESCRIPTIONS: CT-1 .O Flood to Main Steam Lines CT-2.0 Containment Spray

2

Scenario 1 Summary

The scenario begins a t J h 5 reactor power, with the Mechanical Pressure Regulator (MPR) in service. The crew will shift pressure control to the Electronic Pressure regulator (EPR) per normal operating procedures. Shortly after assuming the shift, Feedwater Booster Pump 13 automatically trips and the standby booster pump automatically starts. The pump is a HPCl component and must be declared inoperable. SRO enters TS 3.1.8 and the pump must be restored within 15 days. While shifting regulators, one of the four drywell pressure transmitters fails downscale, preventing that channel from actuating protective functions. The transmitter inputs to RPS, Core Spray, Containment Spray and Automatic Depressurization Systems (ADS). Tech Spec entry is required.

Recirc Pump 13 Motor Generator experiences an overheating condition and generator slot temperature rises. The crew will reduce power and remove Recirc Pump 13 from service. As turbine load is reduced, a pre-existing failure in the Turbine Steam Seal regulating system is revealed. The turbine seals are normally self-sealing at high power levels. The component failure is only evident as load is reduced from the Recirc Pump trip. Seal header pressure drops below normal values. The crew restores seal header pressure by manually opening the steam seal bypass valve.

A failure of the Recirc Master Flow Controller results in an unplanned power changes, as Recirc Flow is reduced to minimum. Plant parameters are such that the Restricted Zone of the Power/Flow Map is entered. The crew implements N1-SOP-1.5 and must exit the Restricted Zone by inserting cram rods. The transient is complicated by a failure of the Turbine Steam Seal Regulator Bypass valve which causes a degraded condenser vacuum, due to loss of steal seals. Vacuum lowers and now results in a required turbine trip due to low load (<I90 MWe) and high backpressure (>5 inches). The crew trips the turbine, but the reactor remains at power, since power is now within turbine bypass valve capability. The crew is expected to manually initiate a scram, due to the degrading conditions. If the crew does not initiate a manual scram, a spurious automatic scram will occur.

Several minutes after the reactor is scrammed, a steam leak inside the drywell develops along with a failure of an additional drywell pressure transmitter. The transmitter failure results in loss of function for actions occurring on high drywell pressure. These functions include loss of automatic scram, automatic start of Core Spray and Containment Spray systems and ADS. Following the scram, all RPV water level indicators will become erratic as reference legs flash, due to the elevated drywell temperature. The crew will be required to flood the RPV to the Main Steam Lines

9/22/2006 7:22:40 AM 2 of 11 NRC Exam Outline Submittal

per N1 -EOP-7, RPV Flooding. The crew will also control Primary Containment parameters by implementing NI-EOP- 4, Primary Containment.

Major Procedures Exercised: N1-SOP-1.3, N1-SOP-1.5, N1-SOP-25.1, N1-EOP-2, N1-EOP-4, N1-EOP-7

EAL Classification: SAE 2.1.2 RPV Flooding is required.

Termination Criteria: RPV Flooding conditions are met. Containment Spray initiated and secured when DWP drops below 3.5 psig.

9/22/2006 7:22:40 AM NRC Exam Outline Submittal

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1 ~~ ~~

Appendix D, Rev. 9 Scenario 0 ut I i ne Form ES-D-1

ECO6A

Facility: NMPI Scenario No.: 2 Op-Test No.: NA

C (BOP)

Examiners: Operators Initial Conditions: IC20 100% Reactor Power Turnover: Complete Nl-ST-Q4 Section 8.2. 39-13R, 39-14R, 05-04R, and 05-12 testing is done. Test Steam IVs 39-1 OR and 39-08R. LP Pump 12 is out of service since 0600 today. TS LCO 3.1.2 specification b was entered and those requirements are currently met. Cont. Spray 122 is inoperable.

7

8

9

10

No.

RD33

RP09

MSOI rejected to the torus.

LPOIA

C (ALL) ATWS. Following the scram control rods will not fully insert and power will remain within turbine bypass valve capability, at about 30%. The MSIV’s will close on high temperature and heat will be

C (BOP) Liquid Poison Pump 11 trips after several minutes of injection. With LP Pump 12 initially out of service alternate boron injection from the Reactor Building using hydro pump would be necessary to inject boron.

Control Rod Drive Pump 12 trips during the scram transient. Starting CRD Pump 11 is necessary for driving control rods.

Event is classified as SAE 2.2.2

RD35B C (RO)

Event Description

~ ~ ~ ~ ~~ ~~

Crew performs N1 -ST-Q4, Reactor Coolant System Isolation Valve Operability Test for EC Loop 12 IV’s per Section 8.2.

Steam IV 39-08R fails to fully close during testing. Valve must be declared inoperable and isolated per Tech Spec 3.2.7. EC Loop 12 now remains inop and unavailable when the steam line is isolated (TS 3.1.3.b, 7 day)

ERVI 11 inadvertently opens. The crew enters SOP-I .4. An emergency power reduction to 85% is performed. The ERV will close after the fuses are pulled. Tech Spec 3.1.5 must be entered because the valve is now inoperable. TS 3.2.9 may also require I entry.

4 Recirc Flow Master Controller fails as-is, preventing the power reduction by normal methods. The crew will operate individual Recirc Flow controls at F panel or insert cram rods to complete the emergency power reduction.

5 EC 11 tube leak (50% with 5 minute ramp time). EC 11 isolation is required. Both EC loops are now inoperable. Tech Spec 3.1.3 specification e now applies and an orderly shutdown is required.

6 ~ ~

A steam leak develops in the turbine building condenser area with severity at 15%. Turbine Vibration rises following the load reduction. The crew will initiate a manual scram due to degraded plant conditions or when turbine bearing vibration exceeds 12 mils.

*(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor


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Facilitv: Nine Mile Point 1 Scenario No.: NRC-02 OD-Test No.: NRC TARGET QUANTITATIVE ATTRIBUTES (PER SCENARIO; SEE SECTION D.5.d)

1. Total malfunctions (5-8) Events 3,4,5,7,8,9

Event 8 Liquid Poison Pump Trip Event 9 CRD Pump Trip

2. Malfunctions after EOP entry (1 -2)

ACTUAL ATTRIBUTES

6

2

3. Abnormal events (2-4) 2 Event 3 SOP-1.4 Event 5 EC Tube Leak 4. Major transients (1 -2) Event 6 Turbine High Vibration and Steam Leak into the Turbine building 5. EOPs enteredhequiring substantive

actions (1 -2) EOP-4 Primary Containment

I

1


EOP-3 Failure To Scram actions (0-2)

7. Critical tasks (2-3) CRITICAL TASK DESCRIPTIONS: CT-1 .O Maintain below HCTL CT-2.0 Terminate and prevent RPV injection during ATWS. CT-3.0 Restore and maintain RPV water level above -109 inches.

1

3

Total Malfunction Count: Major is not included in this count.


SRO TS Events Event 2 , 3 and 5 are SRO Tech Spec evaluation events.

Scenario 2 Summary

The crew assumes the shift with the plant operating at rated conditions and Containment Spray Pump 122 removed from service for maintenance. The crew will perform NI-ST-Q4, Reactor Coolant System Isolation Valves Operability Test, on the Emergency Condenser (EC) Loop 12 Isolation Valves (IVs) per Section 8.2. This test consists of stoke time tests for EC Steam Isolation Valves. A valve failure will result in entry into Tech Specs for the failed coolant and containment isolation valve wj// LlbL & b d f

When the surveillance on the EC Loop 12 IVs is addressed, the crew will respond to an inadvertent opening of an ERV. The crew will perform b n emergency power reduction to about 85%. A failure of the Master Recirc Flow Controller will require the crew to either take manual control of the pump MA stations or insert the cram rods to complete the power reduction. The ERV will be closed when the crew pulls the control power fuses. The SRO must also assess the Tech Spec impact of the inoperable ERV.

ECI 1 vent radiation monitor alarms and the crew determines that a tube leak exists, based on confirmed alarms and rising shell water level. The crew will isolate ECI 1 to stop the release. The'SRO reviews Tech Specs and determines with ECI 1 inoperable Tech Spec 3.1.3.b applies. However, with a confirmed EC Tube Leak a plant shutdown is required. Additionally both EC are now inoperable and unavailable for pressure control.

1

Following the power reduction, turbine vibration will rise and steam leak in the turbine building develops. These reach a severity level that requires a turbine trip and a reactor scram.

9/22/2006 1 :59:46 PM NRC Exam Outline Submittal

5 o f 11

the turbine building, the main condenser will only be available as a heat sink for a short period of time after the scram before the MSlVs are closed, on high steam tunnel temperature. The crew will terminate and prevent injection to lower reactor water level and suppress reactor power. When the main condenser is lost as a heat sink, the crew will maintain reactor pressure using the ERVs and will place torus cooling in service. Because of the rising torus water temperature the crew will inject Liquid Poison (LP). When the Liquid Poison Pump1 1 is started, it will trip after several minutes of injection, requiring recognition of its failure and with pump 12 already out of service, alternate boron injection must be directed. The SRO will direct the actions of EOP-3 and EOP-4 including alternate control rod insertion per EOP-3.1. The crew will be able to insert control rods, after starting a CRD pump, using the Reactor Manual Control System (RMCS) and manual reactor scrams will be successful in inserting the control rods. The SRO will be required to reduce the pressure control band to remain within the heat capacity temperature limit, since the loss of both Emergency Condensers adds additional heat to the torus due to additional heat rejected through ERVs.

Major Procedures:

EAL Classification: SITE AREA EMERGENCY 2.2.2

Termination Criteria: restore reactor water level restored to 53-95 inches.

All control rods inserted, EOP-3 exited, EOP-2 entered and crew directed to


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c Scenario 0 ut I i ne Appendix D, Rev. 9

Event Type*

Form ES-D-1

Event Description

Malf. No.

N (BOP)

I TS (RO SRO)

C (BOP)

I TS (SRO)

NM13A

~

At about 900 psig RPV Pressure, place 2"d RWCU Pump in service

IRM 11 INOP trip. Requires bypassing, consulting Tech Specs and resetting RPS Channel 11 trip.

Service Water Pump 11 trips resulting in rising RBCLC and TBCLC temperatures. Service Water Pump 12 is manually started.

REACTOR BUILDING VENT RAD MONITOR fails inoperable. Requires TS 3.6.2.j entry. Input into RB Emergency Ventilation Svstem automatic initiation. Trip channel within 24 hours.

CW02A

C (BOP)

R M I U

ERV Fails to open during RPV Blowdown.

RD02R3 03 1

RR54

EC02

EC08A EC08B

RP05B

RP28

FW03A

FW03B

HV03A

HV03B

AD07A

R I Raise RPV Pressure from 845 psig to 918 psig with EPR in service.

Facility: NMPl Scenario No.: 3 Op-Test No.: NA

Examiners: Operators Initial Conditions IC 6 modified to establish desired RPV pressure. IC226 845 psig and 2% power, with one turbine Bypass Valve partially open. Turnover: Continue plant Startup per N1-OP-43A and complete section E.3.0. After assuming the shift, raise pressure and continue startup. Place 2"d RWCU Pump in service. Raise pressure to 91 8 psig

Event No.

1

2

3

4

5

6

7

8

9

10

11

Control Rod Drifts out. The rod is fully inserted and the HCU is valved out per Annunciator Response Procedure F3-2-6 CONTROL ROD DRIFT.

I (BOP) RPV Narrow range water level transmitter failed upscale. RPV level slowly lowers requiring FW LCV placed in manual and level restored manually. N1 -SOP-I 6.1 is entered.

M (ALL)

C (BOP)

Emergency Condenser Steam Leak into Reactor Building with Failure to Isolate. Requires a manual scram. RPS Fails to trip but manual initiation of ARI pushbutton results in complete rod insertion. More than one Reactor Building General Area temperature exceeds 135"F, requiring an RPV Blowdown. Event is classified as SAE 3.4.1 and SAE 4.4.1

Loss of Feedwater Pumps complicates post blowdown level control. Injecting with Feedwater Booster Pumps or Core Spray is necessarv to maintain RPV Water Level.

RB Emergency Ventilation logic failure. Manual action is required to trip and isolate normal ventilation and start RBEVS.


7 o f 11

Facilitv: Nine Mile Point 1 Scenario No.: NRC-03 Op-Test No.: NRC TAR G ET Q U AN TI TAT IVE ATTR I B UTES (PER SCENARIO; SEE SECTION D.5.d)

1. Total malfunctions (5-8) Events 3,4,6,7,9,10,11 2. Malfunctions after EOP entry (1-2)

ACTUAL 2 and 3 AND 4 and 5 events can ATTRIBUTES be run simultaneously, to

7 improve efficiency.

3 ~ Total Malfunction Count:

3. Abnormal events (2-4) 2 I Event 6 Rod Drift Event 7 FWLC

Event 9 FW Pump trips and Event 10 RBEVS fail to start and Event 11 ERV fa i I ure

Major not included in this count.


EOP-3 Failure To Scram; EOP-8 RPV actions (0-2)

Blowdown

4. Major transients (1 -2) Event 8 EC Steam Leak with Isolation

Failure

actions (1 -2) 5. EOPs enteredhequiring substantive

EOP-2 RPV and EOP-5 Sec Cont 2

1

2

7. Critical tasks (2-3) I 4 CR~TICAL TASK DESCRIPTIONS: CT-1 .O Initiate scram before blowdown CT-2.0 Insert rods using ARI CT-3.0 Blowdown when 2 areas above 135


SRO TS Events Event 3 and 5 are SRO Tech Spec evaluation events.

Scenario 3 Summary

The scenario begins at about 2% reactor power, during plant startup. RPV pressure is 845 psig. The crew will continue the startup raising pressure to 918 psig and establishing pressure control on the Electronic Pressure Regulator (EPR). As pressure is raised to about 900 psig, the second Reactor Water Cleanup Pump is started per normal operating procedures. IRM 11 failure due to an inop trip occurs resulting in a trip if RPS Channel 11. The crew will bypass the failed instrument and reset the resulting RPS channel trip.

Service Water Pump 11 trips occurs requiring a standby pump to be placed in service. Control Rod 30-31 will drift out. The rod is fully inserted and isolated. With power level below RWM setpoint the RWM must be bypassed to insert the drifting rod. The startup now continues the startup by pulling control rods. Reactor Building radiation monitor 11 inop condition occurs. Tech Spec entry into 3.6.2.j is required because the monitor is part of RB Emergency Ventilation initiation instrumentation.

RPV Narrow Range level input to Feedwater Level Control System (FWLC) fails upscale. With the level transmitter failed high, an RPV water level transient results requiring crew to take manual control of level control valve to prevent an automatic protective trip function per N2-SOP-16.1, Feedwater Failures. When level is stabilized the crew will either maintain manual level control or transfer to a redundant Narrow Range transmitter per normal operating procedures and subsequently return FWLC to automatic control.

The major transient begins when an Emergency Condenser steam leak into Reactor Building occurs. Automatic and manual attempts to isolate the leak will be unsuccessful. Entry in EOP-5, Secondary Containment Control is required and the reactor will be manually scrammed. When the manual scram is initiated both RPS channels fail to trip. The crew enters EOP-3 Failure To Scram and takes initial actions to mitigate the event. When the crew manually initiates ARI, all control rods will fully insert and EOP-3 is exited. The crew transitions back to EOP-2 RPV Control. During the scram the operating Feedwater Pumps trip, complicating post scram level control. When Reactor Building general area temperatures exceed 135°F in more than one area, an RPV Blowdown using EOP-8 is required. Reactor Building Ventilation system malfunctions require manually action to mitigate steam leak into the building which could lead to a release. 9/22/2006 1 :56:54 PM 8 o f 11 NRC Exam Outline Submittal

ses other blowdown systems per EOP-8.

Major Procedures Exercised: N1-SOP-16.1, N1-EOP-2, N1-EOP-5, N1-EOP-8

EAL Classification: SAE 3.4.1 Main Steam Line, EC steam line or Reactor Water Cleanup isolation failure AND release pathway, outside normal process system flowpaths from unisolable system exists outside primary containment

SAE 4.1 . I Primary system is discharging into RB resulting in RB general area temperatures > 1 3 5 O F in two or more areas, N1-EOP-5

Termination Criteria: RPV Blowdown is complete and RPV level is maintained above TAF with all rods fully inserted.


9 o f 11

Appendix D, Rev. 9 Scenario Out1 i ne Form ES-D-1

1

2

3

4

5

6

7

8

Event Description

R

N (BOP)

NM19C I(ALL)

RRO6A C RR07A (BOP)

Overrides

TC06 c (RO)

TC08 M (ALL)

ED07 C (BOP)

RR29 C (ALL)

FW03A

CSOl B

CS05D

Raise power with Recirc Flow.

Switch CRD Stabilizing Valves from A and B to E and F per NI-OP- 5, Section F.4.1

APRM 13 fails upscale resulting in half scram and Tech Spec entry. Bypassing channel and resetting half scram is required.

Recirc Pump 11 seal leakage requires pump removal from service. Pump suction valves fails to fully close resulting in partial loop isolation.

Electrical Pressure Regulator Failure Oscillations. The EPR is removed from service and the Mechanical Pressure Regulator (MPR) is placed in service

~~ ~

Mechanical Pressure Regulator Failure Low. The MPR fail low resulting in rapid pressure and power rise. An automatic reactor scram occurs.

Electric Fault on Emergency AC Powerboard PB102. Diesel Generator 102 automatically starts, but does not close in on the bus due to the fault. DG102 must be manually shutdown due to loss of power to the diesel auxiliaries. Downstream 600 VAC Powerboard PBI 6B must be re-energized from an alternate source.

Motor Driven Feedwater Pump 11 trips resulting in a loss of high pressure feed. Additional high pressure injection sources (CRD and Liquid Poison) must be started as directed from EOP. A small LOCA (14%) occurs which reduces vessel inventory and level lowers to top of active fuel. Core Spray Pump 1 12 trips and Core Spray Pump 122 suction strainer becomes plugged. Injection can be restored using Feedwater Booster Pumps, after depressurizing the RPV. Containment Spray system operation is required due to elevated Drywell pressure.


10 of 11

TA RG ET QUANTI TAT I VE ATT R I B UTES (PER SCENARIO; SEE SECTION D.5.d)

1. Total malfunctions (5-8) Events 3.4,5,7.8

ACTUAL

5 ATTRIBUTES Total Malfunction Count:

Major not included in this count.

2. Malfunctions after EOP entry (1-2) Event 7,8 3. Abnormal events (2-4) Event 3,4,5 4. Major transients (1 -2) Event 6 5. EOPs enteredlrequiring substantive

actions (1 -2) EOP-2 RPV and EOP-4 Pri Cont

CRITICAL TASK DESCRIPTIONS: CT-I .O Initiate Containment Spray CT-2.0 Restore and maintain RPV level

CT-3.0 Terminate Containment Sprays above -109 inches

2

3

1 considered separately.

2

Abnormal Events Count: Does not include the SRO TS related events. These are

~ SRO TS Events Event 3

Scenario Alternate Summary

6. EOP contingencies requiring substantive actions (0-2)

Alt level Control of EOP-2 7. Critical tasks (2-3)

Crew assumes the shift with the plant at 95% power with Feedwater Pump 12 out of service for repairs. The crew is directed to restore reactor power to <I 00% following performance of N1 -ST-W 1, Control Rod Exercising and Operability Test. Following the power reduction the crew is directed to swap CRD Stabilizing Valves from A and B to E and F. After this, APRM 13 will fail upscale, producing a half scram that may be reset. Crew will bypass the failed APRM, reset the half-scram and review Technical Specifications for the failed instrument.

1

After the Technical Specifications review, a seal leak develops on Reactor Recirculation Pump 11. Crew will remove that pump from service, attempt to isolate it, and review Technical Specifications. Reactor pressure will then begin to oscillate. Crew will recognize a failing EPR, place the MPR in control, move the EPR to its high stop, and review Technical Specifications for operation without a backup regulator. When the plant is stable, the MPR will fail, causing a reactor scram. Crew enters EOP-2 on low RPV water level.

After scram actions are complete, Powerboard 102 will develop a fault; EDG 102 will start, but its output breaker will not close. The Crew crossties PB 166 and PB 16A and restore loads. Shortly afterwards, FWP 11 will trip, leaving only CRD pumps and Liquid Poison pumps for high-pressure injection.

A medium break LOCA begins to develop inside Primary Containment. Crew enters EOP-4 on high drywell pressure. When Core Spray pumps start, one of the operable pumps will trip. Recognizing its inability to maintain level above TAF with high-pressure injection, the crew aligns alternate injection sources, and enters EOP-8 for blowdown.

Core Spray slowly begins to recover water level when suction strainer clogging occurs causing the running Core Spray Pump to trip. By this time, RPV pressure is low enough for firewater to inject. Crew will restore and maintain RPV water level above -109 inches.

Major Procedures: EAL Classification: Alert EAL 3.1.1 Termination Criteria:

N1-EOP-2, 4, and 8

RPV depressurized with level recovered above -1 09 inches


11 of 11

UNIT $NRC (RO)

ES-401 BWR Examination Outline Form ES-401-1

on NRC revisions The final RO exam must total 75 points and the SRO-only exam must total 25

3 Systems/evolutions w[thin each group are identified on the associated outline, systems or evolutions that do not apply at the facility should be deleted and justified, operationally important, site-specific systems that are not included on the outline should be added. Refer to ES-401, Attachment 2, for guidance regarding the elimination of inappropriate WA statements.

4. Select topics from as many systems and evolutions as possible, sample every system or evolution in the group before selecting a second topic for any system or evolution

5 Absent a plant-specific priority, only those WAS having an importance rating (IR) of 2.5 or higher shall be selected Use the RO and SRO ratings for the RO and SRO-only portions, respectively

6 Select SRO topics for Tiers 1 and 2 from the shaded systems and WA categories. 7.*The generic (G) WAS in Tiers 1 and 2 shall be selected from Section 2 of the WA Catalog, but the

topics must be relevant to the applicable evolution or system 8 On the following pages, enter the WA numbers, a brief description of each topic, the topics'

importance ratings (IR) for the applicable license level, and the point totals for each system and category Enter the group and tier totals for each category in the table above. Use duplicate pages for RO and SRO-only exams. For Tier 3, select topics from Section 2 of the WA catalog, and enter the WA numbers, descriptions, IRs, and point totals on Form ES-401-3. Limit SRO selections to WAS that are linked to 10 CFR

9

ES401, Page 1 of 10 2006 U1 NRC ES 401-1 RO Written Exam Outline All Tiers Dr

Outline Submittal Thursday, September 21, 2006

@ UNIT 1 NRC (RO)

IR

3.2

3.1

3.8

2.6

3.6

4.0

3.4

ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier l/Group 1 (RO)

##

1

1

1

1

1

1

1

295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4

AA2.06 Ability to determine andlor interpret the following as they apply to PARTIAL OR COMPLETE

the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF A.C POWER RO 50

RO 51

RO 52

RO 55

RO 57

RO 60

3.4

3.4

3.4

2.5

3.9

4.1

2.9

1

1

1

1

1

1

1

ES 401, Page 2 of 10 2006 U1 NRC ES 401-1 RO Written Exam Outline All Tiers Dr Outline Submittal Thursday, September 21,2006

UNIT 1 NRC (RO)

3.8

3.7

4.1

4.2

3.5

2.6

295030 Low Suppression Pool Wtr Lvl / 5

RO 63

1

1

1

1

1

1

20

295031 Reactor Low Water Level I 2

295037 SCRAM Condition Present and Power Above APRM Downscale or Unknown

Power Above APRM Downscale or Unknown I 1 R 0 6 6

295038 High Off-site Release Rate I 9

600000 Plant Fire On Site / 8 I R 0 6 8

1 KIA Category Totals: 1 3

~ ~

EK3.01 Knowledge of the reasons for the following responses as they apply to LOW SUPPRESSION POOL WATER LEVEL: Emerqencv depressurization

EA1.07 Ability to operate and/or monitor the following as they apply to REACTOR LOW WATER LEVEL: Safetyhelief valves

EKI .01 Knowledge of the operational implications of the following concepts as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Reactor pressure effects on reactor power

~

EK3.07 Knowledge of the reasons for the following responses as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Various alternate methods of control rod insertion ~~

EA1.06 Ability to operate and/or monitor the following as they apply to HIGH OFF-SITE RELEASE RATE: Plant ventilation

AK2.01 Knowledge of the interrelations between PLANT FIRE ON SITE and the following: Sensors / detectors and valves

Group Point Total:

ES 401, Page 3 of 10 2006 U1 NRC ES 401-1 RO Written Exam Outline All Tiers Dr Outline Submittal Thursday, September 21, 2006

- UNIT 1 NRC (RO)

ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions -Tier I/Group 2 (RO)

E/APE # / Name / Safety Function K/A Topic(s)

RO 69

RO 70

RO 72

LevelA

procedures those actions that require immediate operation of system components and controls

RO 75

K A Category Point Totals. 1 1 2 1 0 2 Group PointTotal: I

IR

3.3

3.6

3.9

~

3.9

3.8

- 3.5

#

1

1

1

- 1

1

- 1

ES 401, Page 4 of 10 2006 U1 NRC ES 401-1 RO Written Exam Outline All Tiers Dr Outline Submittal Thursday, September 21, 2006

UNIT 1 NRC (RO)

~

1 - 4.2

2.8

- 3.6

- 2.8

- 4.2

- 3.5

2.9

- 3.9

- 2.8

- 3.2

- 2.5

3.3

A3.08 Ability to monitor automatic operations of the AUTOMATIC DEPRESSURIZATION SYSTEM including: Reactor pressure

!I 8000 ADS

10 23

1 !23002 PClSlNuclear Steam ;upply Shutoff

7 0 24

K3.11 Knowledge of the effect that a loss or malfunction of the PRIMARY CONTAINMENT ISOLATION SYSTEMlNUCLEAR STEAM SUPPLY SHUT-OFF will have on following: Plant ventilation

A2.09 Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Svstem initiation

1

- 1

- 1

!23002 PCIS/Nuclear Steam ;upply Shutoff

70 25

!39002 SRVs

3 0 26

K2.01 Knowledge of electrical power supplies to the following: SRV solenoids

!39002 SRVs

‘10 27

K3.02 Knowledge of effect that a loss or malfunction of the RELIEFEAFETY VALVES will have on the following: Reactor over pressurization

K1.04 Knowledge of the physical connections and/or cause-effect relationships between REACTOR WATER LEVEL CONTROL SYSTEM and the following: Reactor feedwater flow

1 259002 Reactor Water Level Zontrol

3 0 28

A I .01 Ability to predict and/or monitor changes in parameters associated with operating the STANDBY GAS TREATMENT SYSTEM controls including: System flow

1 261 000 SGTS

‘10 29

A2.03 Ability to (a) predict the impacts of the following A.C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of offsite power

1

- 1

262001 AC Electrical 3istribution

4 0 30

262002 UPS (ACIDC)

RO 31

K3.14 Knowledge of the effect that a loss or malfunction of the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) will have on following: Rx power

A3.01 Ability to monitor automatic operations of the D.C. ELECTRICAL DISTRIBUTION including: Meters, dials, recorders, alarms and indicating lights

1 263000 DC Electrical Distribution

RO 32

K4.03 Knowledge of EMERGENCY GENERATORS (DIESEL/JET) design feature@) and/or interlocks which provide for the following: Speed droop control

1 264000 EDGs

RO 33

2.4.50 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual

1 264000 EDGs

RO 34

E S 401, Page 6 of 10 2006 U l NRC ES 401-1 RO Written Exam Outline All Tiers Dr Outline Submittal Thursday, September 21,2006

UNIT 1 NRC (RO)

A2.09 Ability to (d) predict the impacts of the following on the RADIATION MONITORING SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Low fuel pool level

A2.03 Ability to (a) predict the impacts of the following on the PLANT VENTILATION SYSTEMS ; and (b) based on those predictions, use . procedures to correct, control, or mitigare the consequences of those abnormal conditions or operations: Loss of coolant accident

272000 Radiation Monitoring

RO 47

3.1

3.5 ~~

288000 Plant Ventilation

RO 48

KIA Category Point Totals: I 1

- 1

12

ES 401, Page 9 of 10 2006 U1 NRC ES 401-1 RO Written Exam Outline All Tiers Dr Outline Submittal Thursday, September 21,2006

~-

UNIT I NRC (RO)

~

ES-401 Generic Knowledge and Abilities Outline (Tier 3) (RO) Form ES-401-3

Facility: Nine Mile Point Unit I Date of Exam: December 18,2006

Category

1.

Conduct of 0 pe ra t ion s

I

2.

Equipment Control

Radiation Control

4.

E merge ncy Procedures /

Plan

Topic

2.1 .2

RO 1 Knowledge of operator responsibilities during all modes of plant operation.

.29 Knowledge of how to conduct and verify valve RO 2 lineups.

Ability to recognize and indications for system operating parameters which are entry level conditions for technical specifications.

RO 3

2.2.1 Ability to perform pre-startup procedures for the 3.7 facility/ including operating those controls associated with plant equipment that could affect reactivity.

Ro

changes.

RO 6

2.3.1

RO 7

Knowledge of 10CFR20 and related facility radiation control requirements. ! 2.6

2.4.23

RO 9

2.4.48

RO 10

Knowledge of the bases for prioritizing emergency procedure implementation during emergency conditions. Ability to interpret control room indications to verify status and operation of system/ and understand how operator actions and directives affect plant

ES 401, Page 10 of 10 2006 U1 NRC ES 401-1 RO Written Exam Outline All Tiers Dr

Outline Submittal Thursday, September 21, 2006

UNIT 1 NRC (RO) Facility: Nine Mile Point Unit I Date of Exam:December 18,2006

T2G2

T2/G 1

T2/G1

ES-401 Record of Reiected KlAs (RO) Form ES-401-4

Randomli selected K1 andstatement K1.05 (7/6/06) Randomly selected then rejected because statements are below 2.5. Randomly selected, then rejected G 2.1.2 because statement is not applicable to system. Randomly selected G 2.4.49 (7/6/06) Rejected during overall outline review due to inability to develop an operationally oriented discriminating question. Randomly selected K5.05 (8110106) Rejected during outline review because it is not applicable to the unit. There is no relationship between ADS valves and remote shutdown system. Randomly selected A3.08 (8/10/06)

215001 K3

209001 K5.04

218000 K1.05

I Tier / Group 1 Randomly Selected WA I Reason for Rejection

I Per ES-401 D . l .d: After completing the outline, check the selected WAS for balance of coverage within and across the three tiers. Ensure that every applicable WA category is sampled at least twice within each of the three tiers.

I Facility Outline Review 1 ES 401, Page 1 of 2

2006 U1 NRC - RO Written Record of Rejected KAs Thursday, September 21, 2006

T l l G l

T l l G l

Changes made

295001 AA2.01

295006 AK2.06

Rejected during overall outline review due to overlap with Operating Test events in Scenario 1. Randomly selected AA2.06 (8/28/06) Rejected during overall outline review due to overlap with other topic area (R022) and inability to develop an operationally oriented discriminating question. Randomly selected AK2.04 (8/28/06)

during draft development after outline submittal and NRC review. (9/27/06- 10/25/06)

ES 401, Page 2 of 2 2006 U1 NRC - RO Written Record of Rejected KAs Thursday, September 21, 2006

~

I I I 1

UNIT 1 NRC (SRO)

ES-401 BWR Examination Outline Form ES-401-1

Facility: Nine Mile Point Unit I Date of Exam: December 18,2006

Tier

1. Emergency & Abnormal

Plant Evolutions P

2. Plant

Systems

3. Generic Knowledge and

Note. 1 Ensure that at least two topics from every WA category are sampled within each tier of the RO and SRO-only outlines (I e , except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each WA category shall not be less than two)

2 The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by *I from that specified in the table based on NRC revisions The final RO exam must total 75 points and the SRO-only exam must total 25 points

3 Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified, operationally important, site-specific systems that are not included on the outline should be added Refer to ES-401, Attachment 2, for guidance regarding the elimination of inappropriate WA statements

4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution

5 Absent a plant-specific priority, only those WAS having an importance rating (IR) of 2.5 or higher shall be selected Use the RO and SRO ratings for the RO and SRO-only portions, respectively

6 Select SRO topics for Tiers 1 and 2 from the shaded systems and KIA categories. 7 *The generic (G) WAS in Tiers 1 and 2 shall be selected from Section 2 of the WA Catalog, but the

topics must be relevant to the applicable evolution or system 8 On the following pages, enter the WA numbers, a brief description of each topic, the topics'

importance ratings (IR) for the applicable license level, and the point totals for each system and category Enter the group and tier totals for each category in the table above Use duplicate pages for RO and SRO-only exams. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the WA numbers, descriptions, IRs, and point totals on Form ES-401-3 Limit SRO selections to WAS that are linked to 10 CFR

9

ES 401, Page 1 of 6 2006 U1 NRC ES 401-1 SRO Written Exam Outline All Tiers Dr Outline Submittal Thursday, September 21, 2006

UNIT 1 NRC (SRO)

ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier IiGroup 1 (RO)

WA Topic(s)

AA2.03 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Actual core flow

~~

AA2.04 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: System lineups

AA2.04 Ability to determine and/or interpret the following as they apply to CONTROL ROOM ABANDONMENT: Suppression pool temperature

2.2.28 Knowledge of new and spent fuel movement procedures

EA2.02 Ability to determine and/or interpret the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Suppression pool level

EA2.03 Ability to determine and/or interpret the following as they apply to HIGH DRYWELL TEMPERATURE: Reactor water level

2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls

Group Point Total:

IR

3.3

3.7

- 4.1

3.5

3.9

3.9

4.0

- -

#

1

1

- 1

1

1

ES 401, Page 2 of 6 2006 U1 NRC ES 401-1 SRO Written Exam Outline All Tiers Dr Outline Submittal Thursday, September 21, 2006

ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier I/Group 2 (RO)

EA2.02 Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE: Offsite release rate

WA Topic(s)

4.1

2.2.17 Knowledge of the process for managing maintenance activities during power operations

I

Group Point Total:

I 3.5 I

3

I 3.1 I 2.1.25 Ability to obtain and interpret station reference materials such as graphs / monographs / and tables which contain performance data

ES 401, Page 3 of 6 2006 U1 NRC ES 401-1 SRO Written Exam Outline All Tiers Dr Outline Submittal Thursday, September 21,2006

UNIT 1 NRC (SRO)

ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (SRO)

WA Topic@)

A2.10 Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High suppression pool temperature

2.2.25 Knowledge of bases in technical specifications for limiting conditions for operations and safety limits

2.4.1 1 Knowledge of abnormal conditions procedures

2.2.24 Ability to analyze the affect of maintenance activities on LCO status

A2.12 Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High fuel pool ventilation radiation

~~

Group Point Total:

IR

3.4

3.7

3.6

3.8

- 3.4

#


-

UNIT 1 NRC (SRO)

I ES-401 Generic Knowledge and Abilities Outline (Tier 3) (SRO) Form ES-401-3

1 Facility: Nine Mile Point Unit 1 Date of Exam: December 18, 2006

Category

1.

Conduct of 0 pe ra t i o n s

2.

Equipment Control

3

Radiation Control

Emergency Procedures /

Plan

I Tier 3 Point Total


UNIT 1 NRC (SRO) Facility: Nine Mile Point Unit I Date of Exam:December 18,2006

ES-40 1 Record of Reiected KlAs (SRO) Form ES-401-4

I T i e T G r o u p 7 Randomly Selected WA I Reason for Rejection

Per ES-401, Attachment 1, # I : Review each group and delete those items [Emergency/Abnormal Plant Evolutions (E/APEs) for Tier 1 and systems for Tier 21 that clearly do not apply to the facility for which the examination is being written. They are: T I G I NA

T I G2 NA

295027 High Containment Temperature (Mark Ill Containment Only). Not applicable to facility. 29501 1 High Containment Temperature (Mark 1 1 1 Containment Only). Not applicable

Per ES-401, Attachment 2 #5: Except as noted in ES-401, Attachment 2, Item 1, all KA statements that are eliminated after they have been randomly selected to fill an examination outline shall be documented on Form ES-401-4, "Record of Rejected KAs," or equivalent. They are: T I G I 29501 9 A I

T I G I 295024 A I

T I G2 29501 3 A I

T2G1 218000 G 2.1.21

T I G I

T I G I 295031 G 2.3.5

T I G2

T I G2

T2G1 218000 G 2.1.21

T2G2 223001 G2.1.24

Insufficient number of generic topics selected and excessive number of A I topics selected. Randomly deselected 29501 9 A I to become G. (7/5/06) Insufficient number of generic topics selected and excessive number of A I topics selected. Randomly deselected 295024 A I to become G. (7/5/06) Insufficient number of generic topics selected and excessive number of A I topics selected. Randomly deselected 295013 A I to become G. (7/5/06) Randomly selected, then rejected because it is not linked to 1 OCFR55.43(b). Randomly selected G 2.4.1 1. (7/6/06) Randomly selected, then rejected because they are not linked to IOCFR55.43(b). Randomly selected 2.2.28. (7/6/06) Randomly selected, then rejected because the statement is not applicable to this topic. Randomly selected 2.4.49 (7/6/06) Randomly selected, then rejected because the statements are not applicable to this topic. Randomly selected 2.2.17 (7/6/06) Randomly selected, then rejected. 2.2.27 statement is not applicable to this topic. 2.4.3 is not linked to 10CFR55.43(b). Randomly selected 2.1 2 5 (7/6/06) Randomly selected, then rejected because the statement is not applicable to this topic. Randomly selected 2.4.1 1. (7/6/06) Randomly selected, then rejected because it is not linked to 1 OCFR55.43(b). Randomly selected G 2.2.26. (7/6/06)

295023 G 2.1.24,2.4.42

295022 G 2.4.36, 2.3.9

295036 G 2.2.27, 2.4.3

I I Per ES-401 D . l .d: After completing the outline, check the selected WAS for balance of coverage within and across the three tiers. Ensure that every applicable WA category is sampled at least twice within each of the three tiers. NONE I NONE I NONE Facility Outline Review

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Changes made during draft development after outline submittal and NRC review. I I

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ES 401, Page 1 of 1 2006 U1 NRC - SRO Written Record of Rejected KAs Thursday, September 21,2006

oct 2,2006 nmp outline review comments · oct 2,2006 nmp outline review comments e e e m e e e e e...

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