ML050250112

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Draft - Outlines (Facility Ltr. Dated 08/12/2004) (Folder 2)
ML050250112
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Site: Nine Mile Point Constellation icon.png
Issue date: 03/14/2005
From: Evans T
Constellation Energy Group
To: Conte R
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Download: ML050250112 (29)
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August 16,2004 NMP#1 OUTLINE COMMENTS Overall, the outlines received from the licensee on August 1 3th appeared to meet expectations with one exception there were no low power scenarios proposed ES-301, D.4.b (04%

proposed). Comments provided on May 16* per telecom to the licensee.

Written Exam Some general feedback or cautions were provided. Some of these topics may still result in acceptable exam questions depending on exam developers ingenuity.

0 0

0 0

0 0

0 A number of proposed WAS appeared to be testing set points which tend to be overly simplistic non-discriminating (29501 9, 295037, 207000).

Some topics appeared to be generally simplistic and may not discriminate (29501 2)

Several topics involved alarm response which is generally better examined during the dynamic scenarios (295008,218000,295002,201 001).

Several topics appeared to test simple power supplies which may be acceptable in limited numbers (206000).

Ability to manually operate and/or monitor valves in the CR may be better examined during the dynamic scenarios or JPMs (259001).

For TS SRO questions make sure not direct look-up - make integrated TS calls which are better for an SRO for ROs okay to ask more direct simple TS (295013).

Several SRO maybe just testing system knowledge (295004,295028).

Operatina Exam (Draft In-house Comments)

JPMs - Admin, Simulator, and Plant JPM topics appeared to be acceptable. One RO topic on applying yellow and red stickers on equipment in the CR may be acceptable but will it be overly simplistic and tie up too much time in the Simulator?

0 Also asked how many of the Admin. JPM topics are new. Dave W. stated all proposed scenarios are new?

Scenarios 0

0 0

No low power scenarios were proposed.

Good idea to have extra malfunctions beyond minimum just in case another applicant is aggressive and hogs the show.

Make sure CTs are well scripted and comply with APP. D guidance.

Constellation Energy@

Nine Mile Point Nuclear Station P.O. Box 63 Lycoming, NY 13093 NMP-97990 August 12,2004 Mr. Hubert J. Miller Regional Administrator USNRC Region I 475 Allendale Road King of Prussia, PA 19406 ATTENTION: Mr. John Caruso, Senior Examinerhspector

SUBJECT:

NINE MILE POINT UNIT 1 INITIAL OPERATOR EXAMINATION OUTLINE SUBMITTAL Mr. Miller:

As requested by NRC letter dated June 17,2004, the attached package contains the examination outlines for Senior Reactor Operator (SRO) and Reactor Operator (RO) Initial Examinations scheduled for November, 2004. The examinations are being prepared based on the guidelines in Draft Revision 9 of NUREG 1021, "Operator Licensing Examination Standards for Power Reactors." Enclosed are the following examination outline documents:

ES-201-2, Examination Outline Quality Checklist

ES-401-4, Record of Rejected WAS - SRO ES-301-1, Administrative Topics Outline (RO)

ES-301-1, Administrative Topics Outline (SRO)

ES-301-2, Control RoodIn-Plant Systems - RO ES-301-2, Control RoodIn-Plant Systems - SRO

  • ES-301-3, Operating Test Quality Checklist

ES-301-4, Simulator Scenario Quality Checklist ES-301-5, Transient and Event Checklist Preliminary Exam Week Schedule (proposed)

  • Forms ES-401-6, ES-301-3, ES-301-6 are blank. These forms cannot be completed until the examination is finalized. The completed forms will accompany the examination submittal due September 10,2004.

Page 2 August 12,2004 NMP-97990 Please withhold this examination material from public disclosure until after the examinations have been completed.

Nine Mile Point Nuclear Station has used the methodology outlined in ES-401 Attachment 1 "Example Systematic Sampling Methodology." The written examination outlines for the Nine Mile Point Unit 1 RO and SRO examinations and the topics were randomly generated using the method described in ES-401, Attachment 1. These outlines were then saved with password protection on a non-networked computer.

If you have any questions regarding this examination outline submittal, please contact Gregg Pitts (General Supervisor Operations Training) at 3 15-349-1864 or Michael Jaquin (Initial Training Supervisor) at 3 15-349-1 508.

Manager Nuclear Training TAE/crr

~

-~

ES-301 Control Room/ln-Plant Svstems Outline Form ES-301-2 Facility: NINE MILE POINT 7 Date of Examination: 17/1/2004 Examination Level (circle one): RO/SRO Operating Test Number: NRC-07 Control Room Systems (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U)

System / JPM Title A by itself at low power.

6 by itself from LOCA conditions.

C then F is directed after pulling fuses to close ERV.

0, E, H concurrent with each other from power operation.

G performed by itself at power.

a. ACTIONS FOR AND WTHDRAWAL OF CONTROL RODS WHICH DOUBLE NOTCH.

Rod does not withdraw and drive water pressure will be raised as required to withdraw it. When rod withdraws, it continues to withdraw with response per F3-2-6, CONTROL ROD DRIFT, for a rod drift in outward direction.

Task:2000360401, 2010050401 Nf-OP-5; H.27.0

b. LINEUP AND INJECT CONTAINMENT SPRAY RA W WATER INTO CORE SPRAY LOOP 77.

PRA: Supply Cont Spray raw water to core spray Task: 2269020507 Nl-EOP-1, Attachment 5

c. RESPOND TO STUCK OPEN ERVAT POMR.

When fuses are pulled in F panel the ERV closes.

Task: 23990 7 040 I LER 2000-004, Manual Reactor Scram Due To Stuck Upen ERV and Failed Vacuum Breaker DER-NM-2004-2268, Manual Scram Due To ERV723 failure T o 0 Close During PMT (5/4/2004).

Nl-OP-1; H.8.0

d. VENT THE PRIMARY CONTAINMENT VIA DRYWELL THROUGH RBEVS AT POWER (VENT VIA TORUS WHEN DRYWELL VENTING IS INEFFECTIVE).

Unable to establish an effective vent path from the drywell the torus will be vented via the RBEVS. Drywell vent path must be closed to ensure containment function is not bypassed should a LOCA occur; directly pressurize torus air space from drywell if both venting lineups are established.

PRA: Vent primary containment through RBEVS Task: 28290201 01, 2009050501 N7-OP-9; H.1.3, H,1.4 Type Code*

N. S Safety Function 1

REACTIVITY CONTROL 2

RX WATER INVENTORY CONTROL 3

RX PRESSURE CONTROL 5

CONTAINMENT INTEGRITY 1 Of3 NUREG-1021, Draft Revision 9

Facility: NINE MILE POINT 7 Date of Examination: 7 1/7/2004 Examination Level (circle one): RO/SRO Operating Test Number: NRC-07 Control Room Systems (8 for RO; 7 for SRO-I; 2 or 3 for System / JPM Title SRO-U)

Type Code*

e. Nl-ST-M4 FOR EDGIUZ (DG OPERABILITY)

Modify to required DG shutdown once loaded based on annunciator response to degraded/failed component.

PRA: StarVLoad a diesel generator.

Task: 2640030201, 26400201 01, 2640030101 N7-ST-M4

f. ACTIONS IN CONTROL ROOMPRIOR TO CONTROL ROOM EVACUATION When reactor mode switch placed to shutdown the reactor does not scram - presses manual scram pushbuttons to scram the reactor. When vessel isolation switches placed to isolate MSlVs do not close - manually closes MSIVs.

Task: 2009070403 N?-SOP-9. I

g. RESPOND TO A LOSS OF SERVICE WATER Service water pump can be started however service water pressure can be improved but cannot be restored requiring override actions per N1 -SOP-7, Path A.

PRA: Respond to a service water pump trip PRA: Respond to a loss of service water Task: 2769020401. 2000350401 N l -S Of-7

h. START CONTROL ROOM VENTILATION SYSTEM Task: 28800#Ol01 N1-OP-49; E. 1.U.

M, S D, s RO ONLY SRO DO NOT PERFORM Safety Function 6

ELECTRICAL 7

INSTRUMENTS 8

PLANT SERVICE SYSTEMS 9

RADIOACTIVITY RELEASE 2 of 3 NUREG-1021, Draft Revision 9

Facility: NINE MILE POINT 7 Date of Examination: 7 7/7/2004 Examination Level (circle one): RO/SRO Operating Test Number: NRC-07 In-Plant Systems (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

System / JPM Title

i. MANUALLY VENTSCRAMAIR HEADER PER EOP-3.1.

Task: 2009230504 EOP-3.1; Section 2 PRA: EOP-3.1

j. PERFORM RPVINJECTION FOR SAFE SHUTDOW OUTSIDE CONTROL ROOM Task: 2009070403 SOP-9.1, Atfachment 4.
k. TRANSFER BATTERY BOARD I7 LOADS TO BATTERY BOARD 12 Task: 2000450501 Nl-UP-47A; H.9.0 Type Code*

D Safety Function 1

REACTIVITY C 0 N T R 0 L 2

RX WATER INVENTORY CONTROL 6

ELECTRICAL

  • Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)lternate path, (C)ontrol room, (S)imulator, (L)ow-Power, (RKA NUREG-1021, Draft Revision 9

NRC EXAM l-/'

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point I Examiners:

Initial Conditions: 100% power.

Scenario No.: NRC-01 Op-Test No.: NRC-01 0 pe ra t o rs :

urnove to t r ~ ~ s f e r PB101 back to normal Lineup with R.1014 ( h e 4) closed and reaker

~i~~~

$1 open.

122 Containment Spray Pump removed from service. EPR in control.

Event Malf. No.

No. I 1

2

TC06, Electrical Pressure Regulator Fails -

Oscillates I

3

RRSE, Recirc Pump 15 MG Slot Temperatur e Increase (35% ramp 3 min.)

ED18. AC

I

Electrical Power Board Fault (PB 16 Section A)

(TRUE)

Event Type*

BOP SRO RO SRO C

BOP SRO R

RO SRO C

BOP SRO Event Description The crew will transfer PBlOl back to line 4.

N1-OP-30; H. 11.0 (all steps)

EPR oscillations will cause fluctuations in reactor power, reactor pressure, and reactor water level. The crew will place the MPR in service and manually control reactor pressure and raise the EPR out of the way.

ARP A2-2-4, Nl-OP-31; H.I.0, NISOP-2, Tech Specs The crew will resDond to a RRP15 MG set hiah temDemture that continues to'degrade. The crew will rem"oe RRPI 5 from service and take appropriate actions including those actions to support 4-loop operation. When reactor power is between 45% and 90% the thermal limit penalty must be applied because no backup pressure regulator is available.

ARP F2-2-5, Nl-Of-I; F.4;H.2; H.3, The crew will be required to lower reactor power to support removal of RRP15 from service. When manual control of RRP15 is established, recirc flow control reduction by the BOP must be coordinated with the RO.

Nl-SOP-I. 7 Loss of PB16A and PB 161. The crew will recognize and respond to a reduction of drywell cooling and a loss of an IAC and RBCLC Pump.

A4 I, A4-4-2 1 of 11 NUREG-1021, Draft Revision 9

7 n Loop Rupture (15%, 30 sec ramp, RELATIVE 1 5 min. until active 25%

1 min. ramp) t Spray Pump Trip 11 1 (TRUE)

  • (N)ormal, (R)eac ALL BOP SRO NRC EXA&

When the actions for the loss of PB16A and reduced drywell cooling are taken, a gradually increasing LOCA will occur.

The crew must enter EOP-2, EOP-4, and eventually EOP-8.

During and after the blowdown the crew must maintain RPV water level using high pressure and low pressure systems.

N1-EOP-I; Att. 4 and 16, Nl-EOP-2, Nl-EOP-4, After containment sprays are placed into service, containment spray pump 11 1 trips requiring that the other available containment spray pump be placed into service (Containment Spray Pump 112 is removed from service and not available). With insufficient containment spray available for the size of the break both the drywell design temperature and Pressure Suppression Pressure can be exceeded.

When it is determined that the Pressure Suppression Pressure will be exceeded, the crew will perform a blowdown per EO P-8.

Nl-EOP-4, Nl-EOP-8, Nl-EOP-I; Att. 17, EAL Matrix

ivity, (I)nstrument, (C)omponent, (M)ajor TARGET QUANTITATIVE ATTRIBUTES ACTUAL EXAM (PER SCENARIO; SEE SECTION D.5.d)

I ATTRIBUTES I DEVELOPER I %K

6. EOP contingencies requiring substantive actions (0-2)
7. Critical tasks (2-3) 2 o f l l NUREG-I 021, Draft Revision 9

NRC EXAM Atmendix D Scenario Outline Form ES-D-1 L

Facility: Nine Mile Point I Examiners :

Initial Conditions: 700% power.

Turnover: Perform NI-ST-Q4, Reactor Coolanf System lsolation Valves Operabilify Test, Quarterly surveillance on the EC Loop I I lVs per Secfion 8.7.

  1. I12 Containment Spray Pump 00s for repair. TS 3.3.7.b (day1 of 15 day LCO).

Scenario No.: NRC-02 Op-Test No.: NRC-01 0 pe rat0 rs :

Event No.

Malf. No.

Emergency Condenser Tube Leak 111 (1 00%;

ramp 5 minutes)

Second Stage Reheaters 112 Steam Closes Supply TU02, Main Turbine Hi Vi bration Bearing #5 and #6 (53%, no Event Type*

N BOP SRO C

BOP BOP SRO R

RO SRO Event Description The crew will perform NI-ST-Q4, Reactor Coolant System Isolation Valves Operability Test, quarterly surveillance on the EC Loop 11 IVs per Section 8.1. After several valves are tested one valve will fail to indicate full open (dual indication).

ST-Q4, Tech Specs DER-NM-2004-2578, Valve Failed To Indica& Open When Stroke Timed The crew will respond to ECI I vent radiation monitor alarms and diagnose that a tube leak exjsts. The crew will isolate ECI 1 to stop the release.

ARP Kl-1-2, EAL MATRIX, Tech Specs, OP-13 H.IO.0 The crew will respond to a failure of the steam supply to the second stage reheater. The unbalanced condition requires isolating second stage reheaters.

ARP A2-3-5, SOP-1.3, OP-41 H.7.0 The unbalanced condition on the main turbine results in turbine vibration. If power is not lowered in response to the turbine vibration, it will be lowered to 80% to allow isolation of the second stage reheaters.

S 0 P-I. 3

\\-

3 o f l l NUREG-1021, Draft Revision 9

TU02, Main Turbine Hi Vi bration Bearing #5 and #6 (goo/,;

no ramp)

MCOI, Main Condenser Air In-Leakage (100%; ramp 2 minutes) I RPOSA,B I

M RPS AIB failure to scram RP09 ARVATWS air header exhaust Port blocked RD33A-E Control Rod Bank Blocked Bank 1,2, 394,s Position 48, 24, 48, 24,48 I

I ALL I

C See event 2 I BOPSRO I

See event 6 MCOI Main Cond Air In leakage C

ALL NRC EXAh The main turbine vibration degrades and a vacuum leak develops from the vibration. Because of the lowering main condenser vacuum and the rising turbine vibration, the crew will insert a manual reactor scram and trip the main tutbine.

SOP-I, SOP31.3 ATWS. When the crew scrams the reactor control rods fail to insert requiring actions for an ATWS with power about 25%. Crew will be able to manually insert control rods using RMCS. Manual scrams will be successful in inserting control rods but repetitive scrams will be required.

PRA: Execute EOP-3. I.

N 1-SO P-I, N I -EO P-2, N 1-EO P-3, N I -EO P-4, N I -EO P-I ERVI 11 fuse blown: failed closed and wont open because of a burned out solenoid.

Loss of main condenser vacuum. Loss of main condenser as a heat sink. Challenge HCTL.

Nl-EOP-3 4 of 11 NUREG-1021, Draft Revision 9

9 LPOINB LP11/12 pump trip C

BOP NRC EXAM The crew will be required to respond to a failure of the liquid poison pump to continue to tun once started. Shortly after one LP pump is started (1-2 seconds) it will trip requiring the crew to start the other pump.

PRA: Inject poison solution into the reactor vessel.

N7-OP-72; H.7.O k

1

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
5. EOPs enteredkequiring substantive actions (1-2)
6. EOP contingencies requiring substantive actions (0-2)
7. Critical tasks (2-3) 5 o f l l NUREG-I 021, Draft Revision 9

NRC EXAM u

Appendix D Scenario Outline Form ES-D-1 L

v Facility: Nine Mile Point 1 Scenario No.: NRC-03 Op-Test No.: NRC-01 Examiners:

Initial Conditions: 700% power.

Turnover: N7-ST-M4B, mergemy Diesel Generator 703 AND PB 703 Operability Test, completed sa tis factory last shift. Substitute Reactor Building Ventilation Supply and 0 pera to rs :

Exhaus Event No.

1 2

3 5

Fans fron Malf.

No.

HVOIA. RB Exhaust Fan Trip 11 (TRUE)

R B Exhaust Fan 12 &

Outlet Damper FW37. 13 FCV Oscillation(

50%; ramp

= 1 minute) 345KV Power Grid Transient (FINAL VALUE.

338, no

EG16, Generator Cooling Fan Leads Trlp (FINAL VALUE:

50, 1 MINUTE) system 1 Event Type*

N BOP C

BOP C

BOP RO C

BOP R

RO SRO to system 77.

Event Description I Substitute Reactor Building Ventilation Supply and Exhaust Fans from system 12 to system 11.

OP-10 F.3.2, F.7.0, and F.2.0. I Reactor Building Exhaust Fan 11 trips and exhaust fan 12 will not start. Start RBEVS in response to a degraded Reactor Building negative pressure (0 psig).

ARP Ll-3-4, L7-1-5, EOP-5, OP-70 H.I.0, Tech Specs I

The crew will respond to FCV 13 oscillations. Later in the scenario the crew will be required to manually adjust 13 FCV to maintain RPV level below the high level trip when reactor power is lowered.

ARP A6-2-6, A6-3-3, SOP-33.A.3, Multiple Tech Specs, OP-45; E. 3.0 The crew will be required to lower power to maintain isophase bus duct temperatures within limits. When reactor power is lowered the temperatures stabilize then lower.

ARP A7-3-5, SOP-1.3, OP-32; H.4.0 6 of 11 NUREG-I 021, Draft Revision 9

NRC EXAA EG 11, 345KV Power Grid Transient (FINAL VALUE 328; 1 minute ED01 B Loss of Offsite 11 5KV Power South Oswego -

Line 1 EDOIA Loss of Offsite 11 5KV Power JAF-Line 4 DGOIA.

Diesel Generator 102 Failure to Start (TRUE)

C ALL When the power reduction has been made, the grid conditions degrade requiring removal of the main generator from service because of the low frequency.

SOP-33.A.3 (continued),SOP-33.7, SOP-7 Loss of offsite power with EDG102 fail to start and cannot be started. EDG102 and PB102 loss impact SOP5 execution and RPV level and containment control actions.

SOER 99-1; Loss of Grid SOER 03-1; Emergency Power Reliability NMP LER: Loss of grid (Summer 2003)

N1-SOP-5, EOP-2, EAL Matrix, Reactor coolant leak.

EOP-2, EOP-4, EAL Matrix, EOP-7, EOP-8 Core Spray injection valves fail to automatically open and must be manually opened to restore and maintain RPV level above TAF following the RPV blowdown.

EOP -1 (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor 7of 11 NUREG-1021, Draft Revision 9

NRC EXAM 8 o f l l NUREG-1021, Draft Revision 9

NRC EXAM v

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point 1 Examiners:

Operators:

Initial Conditions: 100% power.

Scenario No.: NRC-04 (ALT)

Op-Test No.: NRC-01 ta perform ~~~~~1~

~

~

~

o

~

~

a n

c

~

of ~

4

~

~

~

~

EVS Operability Test 122 Containment Spray Pump removed from service. EPR in controi.

Event No.

Malf.

No.

RBEVS Train flow meter downs ca I e Set at 0 0 1 1 -M040-AO-053.

Reactor Trip Bus MG Set Trips (141)

(TRUE)

RD06, Rod 22-1 9 Failure -

Scrammed (IO Sec.

TD)

(TRUE)

Overrride for control rod 22-19 position

02.

RXOI, Fuel Cladding Failure -

10% Ramp

- 5 minutes, Min.

TUA-1 Event Type*

BOP SRO RO BOP SRO RO SRO Event Description Nl-ST-M8, RBEVS Operability Test is scheduled for its monthly performance. During the test, when the RBEVS fan is started, the fan flow indicator will fail downscale and the train must be declared inoperable. The SRO must make a T.S. determination.

N1-ST-M8, Tech Specs Loss of Reactor Trip Bus 141 and a coincident single control rod scram. When control rod 22-1 9 scrams its CRD Mechanism becomes stuck at position 02. The crew will enter Nl-SOP-40.1 Loss of RPS and restore power to the bus, then the scram may be reset. The crew must lower power, asses recovering the control rod and the SRO must make a T.S. determination.

Nl-SOP-40. I, Tech Specs A small fuel leak will develop from the abnormal rod pattern.

Rising reactor coolant activity levels will require entry into the Emergency Plan and Emergency Power Reduction.

EAL Matrix, EPIP-EPP-I 8, N I-SO P-I. I, N I -SOP-25.2 9 o f l l NUREG-I 021, Draft Revision 9

ED04, AC Power Board Electrical Fault (PBIl),

clears in 3 secs Emergency Condenser Tube Leak 111 ECOBA, EC STM IV 111 Fail to Close =

80%

EC08B. EC STM IV 1 12 Fail to Close =

80%

RD33E, Control Rod Bank 5, Insert Fail position (48)
ED26, Failure of PB 11 to Auto Transfer (TRUE)

C RO BOP SRO BOP SRO C

RO SRO RXOl, Fuel Cladding Failure -

100%

Ramp - 15 minutes M

ALL NRC EXAM During the emergency power reduction the crew will transfer normal house service to the reserve transformers. When the transfer is made both PB 11 supply breakers will trip. The crew can recover PB 11. The trip of PB 11 will cause a trip of RRPs 11 and 12, this will result in entry into the restricted area of the Power to Flow map and the reactor should be manually scrammed.

Nl-SOP-30. I, Nl-SOP-30, Nl-SOP-1.3, Tech Specs, NI-SOP-?

When the reactor is scrammed (If the crew does NOT scram the reactor this malfunction will iequire a reactor scram) a piping rupture will occur in EC Loop 11. The steam isolation valves fail to fully close.

Nl-SOP-1.3, Nl-SOP-I, A bank of control rods will fail to fully insert requiring the crew to perform actions to manually inseft control rods. (if PBI 1 was not transferred previously it will fail to automatically transfer and must be manually transferred.)

Nl-EOP-2, Nl-EOP-3, Nl-EOP-4; Att. 2 and 4, NI-OP-12, H.l.O and G.0 Rising reactor coolant activity and radiation levels will require a blowdown. When EOP-8 is entered the crew must enter the path for all control rods not inserted. This will require the crew to terminate and prevent injection prior to emergency depressurization.

NI-EOP-8, EAL Matrix 10 of 11 NUREG-1021, Draft Revision 9

NRC EXAM

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor TARGET QUANTITATIVE ATTRIBUTES (PER SCENARIO; SEE SECTION D.5.d)
1. Total malfunctions (5-8)
2. Malfunctions after EOP entry (1-2)

ACTUAL EXAM FACl LlTY ATTRIBUTES DEVELOPER R,EVIEW 8

.rc 1

1 O A J f

I 3. Abnormal events (2-4)

I 4

4. Major transients (1-2)
5. EOPs enteredkequiring substantive actions (1-2) 1 2

I 6. EOP contingencies requiring substantive actions (0-2) 1 1

I 1

7. Critical tasks (2-3)

I I

3 I

11 of 11 NUREG-I 021, Draft Revision 9

UNIT 1 NRC (RO)

ES-40 1 BWR Examination Outline Form ES-401-1 Facility: Nine Mile Point Unit 1 Date of Exam: November 18,2004 (tentative)

Tier

1.

Eme rg e n cy

& Abnormal Plant Evolutions

2.

Plant Systems K

1 1

3 2

0 Tier 3

1 3

2 0

Tier 3

Group Totals Totals Note: 1.

2.
3.
4.
5.
6.
7.
8.
9.

Ensure that at least t w topics from every WA category are sampled within each tier of the RO outline (i.e., the Tier Totals in each KIA category shall not be less than tvlit)).

Refer to Section D.l.c for additional guidance regarding SRO sampling.

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.

Select topics from many systems and evolutions; avoid selecting more than two WA topics from a given system or evolution unless they relate to plant-specific priorities.

Systems/evolutions within each group are identified on the associated outline.

The shaded areas are not applicable to the category/tier.

The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the WA Catalog, but the topics must be lelevant to the applicable evolution or system. The SRO K/As must also be linked to 10 CFR 55.43 oran SRO-level learning objective.

On the following pages, enter the K/A 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; summarize all the S R O a l y knowledge and non-A2 ability categories in the columns labeled K and A. Use duplicate pages for RO and SRO-only exams.

For Tier 3, enter the WA numbers, descriptions, importance ratings, and point totals on Form ES-401-3.

Refer to ES-401, Attachment 2, for guidance regarding the elimination of inappropriate K/A statements.

Page 1 of 9 NUREG-1021. Draft Revision 9

UNIT 1 NRC (RO)

ElAPE # / Name / Safety Function K

K K

A A

1 2

3 1

2 Form ES-401-1 FS-An1 BWR Examination Outline 295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 Emergency and Abnormal Plant Evolutions - Tier l/Group 1 (RO) 0 5

295026 Suppression Pool High Water Temp.

15 0

5 IR -

3.2 3.5 3.4 WA Topic(s)

GI 1

AK3.05 Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Reduced loop operating requirements.

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

1 295003 Partial or Complete Loss of AC / 6 295003 Partial or Complete Loss of AC / 6 295004 Partial or Total Loss of DC Pwr / 6 X

2.1.32 Ability to explain and apply system limits and precautions.

1 2.6 1

AK3.01 Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER:

Load shedding.

4.0 3.9 1

1 295005 Main Turbine Generator Trip / 3 I I 0 I I I

AK2.09 Knowledge of the interrelations between MAIN TURBINE GENERATOR TRIP and the following: Feedwater - HPCI: BWR-2.

AA1.02 Ability to operate and/or monitor the following as they apply to SCRAM:

Reactor water level control system.

AA1.07 Ability to operate and/or monitor the following as they apply to CONTROL ROOM ABANDONMENT:

Control room/local control transfer mechanisms.

4.2 3.5 1

1 AKI.01 Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Effects on componentkystem operations.

3.5 1

AA2.01 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR:

all modes of plant operation.

295019 Partial or Total Loss of Inst. Air I 8 295021 Loss of Shutdown Cooling / 4 3.0 1

171 I I I 295023 Refueling Acc Cooling Mode / 8 3.6 1

AKI.01 Knowledge of the operational implications of the following concepts as they apply to REFUELING ACCIDENTS: Radiation exposure hazards.

EK3.04 Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL PRESSURE: Emergency depressurization.

3.7 1 -

1 1

295024 High Drywell Pressure / 5 295025 High Reactor Pressure / 3 3.7 3.9 EK2.08 Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following:

EK3.05 Knowledge of the reasons for the following responses as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Reactor SCRAM.

I I I I I 295028 High Drywell Temperature / 5 X 2.4.6 Knowledge symptom based EOP mitigation strategies.

3.1 1

Page 2 of 9 NUREG-1021, Draft Revision 9

UNIT 1 NRC (RO)

X 0

3 0

4 X

2 4

5 I l l 295030 Low Suppression Pool Wtr Lvl I 5 EA2.03 Ability to determine and/or interpret the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: SBLC tank level.

EA2.04 Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE: Source of off-site release.

I 4 I 295031 Reactor Low Water Level / 2 4.3 4.1 295037 SCRAM Condition Present and Power Above APRM Downscale or Unknown

/ I 600000 Plant Fire On Site / 8 WA Category Totals:

295038 High Off-site Release Rate / 9 3

2 4

I l l I 3'9 I 2.1.23 Ability to perform specific system and integrated plant procedures during different modes of plant operation.

I 4.6 I EKI.01 Knowledge of the operational implications of the following concepts as they apply to REACTOR LOW WATER LEVEL: Adequate core cooling.

1 1

2.4.31 Knowledge of annunciators alarms and indications / and use of the response instructions.

Group Point Total:

I 20 Page 3 of 9 NUREG-1021, Draft Revision 9

L 295015 Incomplete SCRAM / 1 295022 Loss of CRD Pumps / 1 295032 High Secondary Containment Area Temperature / 5 295036 Secondary Containment High Sump/Area Water Level / 5 WA Category Point Totals:

UNIT 1 NRC (RO) 0 AA1.03 Ability to operate andlor monitor the following 3.6 1

3 0

AA1.02 Ability to operate and/or monitor the following 3.6 1

2 0

EK3.02 Knowledge of the reasons for the following 3.6 1

2 as they apply to INCOMPLETE SCRAM: RMCS.

as they apply to LOSS OF CRD PUMPS: RPS.

responses as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE:

Reactor SCRAM.

following as they apply to SECONDARY CONTAINMENT HIGH SUMPIAREA WATER LEVEL:

Water level in the affected area.

0 EA2.02 Ability to determine and/or interpret the 3.1 1

2 0

0 1 2 2

2 Group PointTotal:

7 Page 4 of 9 NUREG-1021. Draft Revision 9

UNIT 1 NRC (RO)

IR 2.8 3.2 3.7 3.3 2.8 I ES-401 1

1 1

1 1

BWR Examination Outline Form ES-401-1 3.3 3.2 4.2 3.7 2.9 3.6 Emergency and Abnormal Plant Evolutions - Tier 2/Group 1 (RO) 1 1

1 1

1 1

I System ## / Name WA Topic(s) 205000 Shutdown Cooling K5.02 Knowledge of the operational implications of the following concepts as they apply to SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE):

Valve operation.

K2.01 Knowledge of electrical power supplies to the following: system valves.

206000 HPCl 207000 Isolation (Emergency)

Condenser K5.09 Knowledge of the operational implications of the following concepts as they apply to ISOLATION (EMERGENCY)

CONDENSER: Cooldown rate: BWR-2,3.

A I.03 Ability to predict and/or monitor changes in parameters associated with operating the ISOLATION (EMERGENCY)

CONDENSER controls including:

Steam flow: BWR-2,3.

207000 Isolation (Emergency)

Condenser K4.10 Knowledge of LOW PRESSURE CORE SPRAY SYSTEM design feature(s) and/or interlocks which provide for the following:

Testability of all operable components.

209001 LPCS 209001 LPCS

~~

A I.08 Ability to predict and/or monitor changes in parameters associated with operating the LOW PRESSURE CORE SPRAY SYSTEM controls including:

System lineup.

I A2.03 Ability to (a) predict the impacts of the following on the STANDBY LIQUIDCONTROL SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: A.C. power failures.

21 1000 SLC

.L A4.08 Ability to manually operate and/or monitor in the control room: System initiation.

21 1000 SLC 212000 RPS K1.01 Knowledge of the physical connections and/or cause-effect relationships between REACTOR PROTECTION SYSTEM and the following: Nuclear instrumentation.

K4.05 Knowledge of INTERMEDIATE RANGE MONITOR (IRM) SYSTEM design feature(s) and/or interlocks which provide for the following: Changing detector position.

21 5003 IRM A I.05 Ability to predict and/or monitor changes in parameters associated with operating the SOURCE RANGE MONITOR (SRM) SYSTEM controls including:

SCRAM, rod block, period alarm trip setpoints.

21 5004 Source Range Monitor Page 5 of 9 NUREG-1021, Draft Revision 9

UNIT 1 NRC (RO) 3.3 3.4 3.3 3.7 4.0 4.2 3.1 3.3 3.0 1

1 1

1 1

1 1

1 1

I 21 5005 APRM / LPRM A3.05 Ability to monitor automatic operations of the AVERAGE POWER RANGE MONlTORlLOCAL POWER RANGE MONITOR SYSTEM including:

Flow converter/comparator alarms.

21 5005 APRM / LPRM I

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

21 8000 ADS 2.4.31 Knowledge of annunciators alarms and indications I and use of the response instructions.

223002 PCIS/Nuclear Steam Supply Shutoff

~~

A I.02 Ability to predict and/or monitor changes in parameters associated with operating the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF controls including:

Valve closures.

239002 SRVs 0

8 K1.08 Knowledge of the physical connections and/or cause-effect relationships between RELIEFISAFETY VALVES and the following:

Automatic depressurization system.

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

Reactor Over-pressurization 239002 SRVs K6.03 Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR WATER LEVEL CONTROL SYSTEM: Main steam flow input.

259002 Reactor Water Level Control 259002 Reactor Water Level Control 2.4.50 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

A2.05 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: Fan trips.

261 000 SGTS I

262001 AC Electrical Distribution 0

1 K1.O1 Knowledge of the physical connections andlor cause-effect relationships between A.C. ELECTRICAL DISTRIBUTION and the following: Emergency generators (dieseldet)

LER: TIE TO LINE#

and MOD OPEN 262002 UPS (AC/DC)

A2.01 Ability to (a) predict the impacts of the following on the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Under voltage.

263000 DC Electrical Distribution I

K2.01 Knowledge of electrical power supplies to the following: Major D.C. loads Page 6 of 9 NUREG-1021, Draft Revision 9

UNIT 1 NRC (RO) 264000 EDGs I

300000 Instrument Air 400000 Component Cooling Water WA Category Point Totals:

K3.03 Knowledge of the effect that a loss or malfunction of the EMERGENCY GENERATORS (DIESEL/JET) will have on following: Major loads powered from electrical buses fed by the emergency generator(s).

A2.01 Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:

Air dryer and filter malfunctions.

A4.01 Ability to manually operate and I or monitor in the control room:

CCW indications and control.

Group Point Total:

4.1 2.9 3.1 Page 7 of 9 1

1 1

26 NUREG-1021, Draft Revision 9

UNIT 1 NRC (RO)

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

IR System # I Name WA Topic(s) 201001 CRD Hydraulic 3.1 3.1 1

~

1 K3.03 Knowledge of the effect that a loss or malfunction of the CONTROL ROD DRIVE HYDRAULIC SYSTEM will have on following: control rod drive mechanisms.

A3.03 Ability to monitor automatic operations of the ROD WORTH MINIMIZER SYSTEM (RWM) including:

Annunciator and alarm signals.

201006 RWM 202002 Recirculation Flow Control 3.1 -

3.1 1 -

1 1

2.4.6 Knowledge symptom based EOP mitigation strategies.

A2.01 Ability to (a) predict the impacts of the following on the ROD POSITION INFORMATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Failed reed switches.

214000 RPlS 216000 Nuclear Boiler Inst.

K4.05 Knowledge of NUCLEAR BOILER INSTRUMENTATION design feature@)

and/or interlocks which provide for the following: Initiation of the emergency core cooling sys tems.

3.9 226001 CTMT Spray Mode 3.9 3.9 3.1 1 -

1 1

2.1 2 3 Ability to perform specificsystem and integrated plant procedures during different modes of plant operation.

A2.05 Ability to (a) predict the impacts of the following on the MAIN AND REHEAT STEAM SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Main steam line high radiation.

A I.I5 Ability to predict and/or monitor changes in parameters associated with operating the REACTORnURBlNE PRESSURE REGULATING SYSTEM controls including:

Maximum combined flow limit.

239001 Main and Reheat Steam 241 000 Reactornurbine Pressure Regulator 259001 Reactor Feedwater 3.1 3.0 3.2 1 -

1 1

A4.04 Ability to manually operate and/or monitor in the control room:

System valves.

K6.01 Knowledge of the effect that a loss or malfunction of the following will have on the RADIATION MONITORING SYSTEM: Reactor Protection System.

K5.02 Knowledge of the operational implications of the following concepts as they apply to PLANT VENTILATION SYSTEMS: Differential pressure control.

272000 Radiation Monitoring 288000 Plant Ventilation Page 8 of 9 NUREG-1021. Draft Revision 9

UNIT 1 NRC (RO) 0 1

I WA Category Point Totals:

1 0 A3.01 Ability to monitor automatic 3.3 1

operations of the CONTROL ROOM HVAC including:

Initiation/reconfiguration.

Page 9 of 9 NUREG-1021, Draft Revision 9

UNIT 1 NRC (RO)

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

FOITTI ES-401-3 Facility: Nine Mile Point Unit I Date of Exam: November 18,2004 (tentative)

I Category

1.

Conduct of 0 pe ra t i o ns Equipment Control I

3.

Radiation Con t ro I Emergency Procedures /

Plan I

Topic 2.1.I 0

Knowledge of conditions and limitations in the facility license.

lineups.

Subtotal 2.2.12 Knowledge of surveillance procedures.

2.2.13 Knowledge of tagging and clearance procedures.

2.2.26 Knowledge of refueling administrative re q u ire men t s.

Subtotal Knowledge of the process for performing a contain men t purge,

2.3.10 Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure.

Subtotal Knowledge of event based EOP mitigation strategies.

2.4.24 Knowledge of loss of cooling water I procedures. PRA: LOSS O f ESW Subtotal 1 Tier 3 Point Total 1 of1 RO IR 2.7 2.8 3.4 3.0 3.6 2.5 2.9 3.1 3.3 NUREG-1021, Draft Revision 9

-u-295024 EK3.03 295026 EK3.03 UNIT 1 NRC (RO)

Facility:

Nine Mile Point Unit 1 Date of Exam: November 18,2004 (tentative)

Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL PRESSURE: Containment venting: Mark-Ill.

Mark I Containment, not Mark 111.

Knowledge of the reasons for the following responses as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Suooression DOOI sDrav.

ES-40 1 Record of Rejected K/As (RO)

Form ES-401-4 295028 2.4.30 Tier / GrouD I Randomly Selected WA I Reason for Reiection Have containment spray, not suppression pool spray.

Knowledge of which events related to system operations/status should be reported to outside aaencies. IMPORTANCE RO 2.2 SRO 3.6.

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:

Per ES-401, Attachment 2 #5: Except as noted in Es-401, Attachment 2, Item 1, all KA statements that are eliminated after they are 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:

295030 2.2.25 600000 2.4.30 NA 295009 2.4.30 206000 A2.15 206000 A2.16 207000 K5.01 239002 K3.03 T I G I T I G I T I G I T I GI T I GI T I G2 T I G2 T2G1 T2G1 T2G1 T2G1 T2G1 Offsite notifizations are SRO Only. Importance rating is ~ 2. 5.

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

TS bases are SRO Only Knowledge of which events related to system operations/status should be reported to outside agencies. IMPORTANCE RO 2.2 SRO 3.6 Offsite notifications are SRO Only. Importance rating is ~ 2. 5.

295034 Secondary Containment Ventilation High Radiation After randomly and systematically selecting 295032, High Secondary Containment Area Temperature, and 295036, Secondary Containment High Sump/Area Water Level, then selected 295034, Secondary Containment Ventilation High Radiation.

Rejected 295034 to avoid over sampling of secondary containment control.

Knowledge of which events related to system operations/status should be reported to outside agencies. IMPORTANCE RO 2.2 SRO 3.6 Offsite notifications are SRO Only. Importance rating is ~ 2. 5.

Ability to (a) predict the impacts of the following on the HIGH PRESSURE COOLANT INJECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of control oil pressure.

No control oil system associated with HPCl per facility design.

Ability to (a) predict the impacts of the following on the HIGH PRESSURE COOLANT INJECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High drywell pressure.

No specific relation between HPCl and high drywell pressure per facility design.

Knowledge of the operational implications of the following concepts as they apply to ISOLATION (EMERGENCY) CONDENSER: Flow measurement across an elbow using differential pressure: BWR-2,3.

Generic Fundamentals Concept.

Knowledge of the effect that a loss or malfunction of the SAFETY/RELIEF VALVES will have on the followina. Abilitv to raDidlv deDressurize the reactor.

261000 K5.01, K5.02

~

Double jeopardy with 2f9002 Ai.08.

Knowledge of the operational implications of the following concepts as they apply to STANDBY GAS TREATMENT SYSTEM:

K5.01 Heat removal mechanisms.....................

2.3* 2.6*

K5.02 Air operated valves: Plant-Specific.....___.

2.3 2.5*

Selected 261000 K5. Rejected K5.01 and K5.02 because importance rating is c 2.5.

I of2 NUREG-1021, Draft Revision 9

UNIT 1 NRC (RO)

Facility:

Nine Mile Point Unit 1 Date of Exam: November 18, 2004 (tentative)

T3 T3 T2G1 I 300000 A3.01, A3.02 Better evaluated during the walkthrough examinat& (JPMs).

Knowledge of the process for making changes in procedures as described in the safety analysis report. (CFR: 43.3 I45.13) IMPORTANCE RO 2.3 SRO 3.3 SRO Only. Importance rating c2.5.

Ability to track limiting conditions for operations. (CFR: 43.2 I45.13) 2.2.6 2.2.23 2.1.17 2.1 2 1 T3 Ability to monitor automatic operations of the INSTRUMENT AIR SYSTEM including:

A3.01 Air pressure.........................

2.3 2.1 A3.02 Air temperature.....................

2.9 2.7 Rejected A3.02 because of inability to evaluate instrument air system air temperature.

Rejected A3.01 because importance rating is < 2.5.

Randomly and systematically selected another KA: 300000 A.1. Rejected because KA listed under A.l is NONE.

Randomly and systematically selected another KA: 300000 A.2. A2.01 selected because only KA available under A.2.

Ability to make accurate /clear and concise verbal reports.

Better evaluated during the operating test (simulator scenarios).

Abilitv to obtain and verifv controlled orocedure CODY.

SRO-On1 y.

Knowledge of RO tasks performed outside the main control room during emergency operations including system geography and system implications.

Better evaluated during the walkthrough examination (In Plant JPMs). One in-plant JPM is required to be local actions in response to an abnormal/emergency condition.

2.4.34 T2G2 201001 K3.01 K3.01 Knowledge of the effect that a loss or malfunction of the CONTROL ROD DRIVE HYDRAULIC SYSTEM will have on following: Recirculation pumps.

No interrelation between CRDH and recirc pumps. Randomly selected new K3 from those available and selected K3.03.

T2G1 T2G2 206000 A2.02 272000 A I.02 A2.02 Ability to (a) predict the impacts of the following on the HIGH PRESSURE COOLANT INJECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Valve closures.

K2 under-sampled (only 1 WA). Randomly selected 206000 HPCI from the 26 T2G1 and 12 T2G2 items, deselected 206000 A2.02, and randomly selected K2.01 from the available K2 WAS for 206000 HPCI.

A1.02 Ability to predict and/or monitor changes in parameters associated with operating the RADIATION MONITORING SYSTEM controls including: Lights, alarms, and indications associated with surveillance testing.

K6 under-sampled (only 1 WA). Randomly selected 272000 Radiation Monitoring from the 26 T2G1 and 12 T2G2 items, deselected 272000 A I.02 and randomly selected K6.01 from the available K6 WAS for 272000 Radiation Monitoring.

2 o f 2 NUREG-1021, Draft Revision 9

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