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
NRC/RGN-I/DRS/OSB
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 13thappeared 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 A number of proposed WAS appeared to be testing set points which tend to be overly simplistic non-discriminating (295019, 295037, 207000).

0 Some topics appeared to be generally simplistic and may not discriminate (295012) 0 Several topics involved alarm response which is generally better examined during the dynamic scenarios (295008,218000,295002,201 001).

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

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

0 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).

0 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 No low power scenarios were proposed.

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

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

P.O. Box 63 Lycoming, NY 13093 Constellation Energy@

Nine Mile Point Nuclear Station 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-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

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 NMP-97990 August 12,2004 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 15-349-1508.

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 Type Code* Safety Function 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 1 RODS WHICH DOUBLE NOTCH. REACTIVITY Rod does not withdraw and drive water pressure will be CONTROL 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 2 WATER INTO CORE SPRAY LOOP 77. RX WATER INVENTORY PRA: Supply Cont Spray raw water to core spray CONTROL Task: 2269020507 Nl-EOP-1, Attachment 5
c. RESPOND TO STUCK OPEN ERVAT P O M R . N. S 3 When fuses are pulled in F panel the ERV closes. RX PRESSURE Task: 23990 7 040 I CONTROL Nl-OP-1; H.8.0 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).
d. VENT THE PRIMARY CONTAINMENT VIA DRYWELL 5 THROUGH RBEVS AT POWER (VENT VIA TORUS CONTAINMENT WHEN DRYWELL VENTING IS INEFFECTIVE). INTEGRITY 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: 2829020101, 2009050501 N7-OP-9; H.1.3, H,1.4 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 SRO-U)

System / JPM Title Type Code* Safety Function

e. Nl-ST-M4 FOR EDGIUZ (DG OPERABILITY) 6 Modify to required DG shutdown once loaded based on ELECTRICAL annunciator response to degraded/failed component.

PRA: StarVLoad a diesel generator.

Task: 2640030201, 2640020101, 2640030101 N7-ST-M4

f. ACTIONS IN CONTROL ROOMPRIOR TO CONTROL 7 ROOM EVACUATION INSTRUMENTS 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 M, S 8 Service water pump can be started however service water PLANT pressure can be improved but cannot be restored SERVICE requiring override actions per N1-SOP-7, Path A. SYSTEMS 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 D, s 9 Task: 28800#Ol01 RO ONLY RADIOACTIVITY N1-OP-49; E. 1.U. RELEASE SRO DO NOT PERFORM 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 Type Code* Safety Function

i. MANUALLY VENTSCRAMAIR HEADER PER EOP-3.1. 1 PRA: EOP-3.1 REACTIVITY Task: 2009230504 C0NTR0L EOP-3.1; Section 2
j. PERFORM RPVINJECTION FOR SAFE SHUTDOW 2 OUTSIDE CONTROL ROOM RX WATER Task: 2009070403 INVENTORY CONTROL SOP-9.1, Atfachment 4.
k. TRANSFER BATTERY BOARD I7 LOADS TO BATTERY D 6 BOARD 12 ELECTRICAL Task: 2000450501 Nl-UP-47A; H.9.0
  • Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)lternate path, (C)ontrol room, (S)imulator, (L)ow-Power, ( R K A NUREG-1021, Draft Revision 9

NRC EXAM l

-/' Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point I Scenario No.: NRC-01 Op-Test No.: NRC-01 Examiners: 0perat o rs:

Initial Conditions: 100% power.

urnove e r back to normal Lineup with R.1014 ( h e 4) closed and to t r ~ ~ s fPB101 reaker ~i~~~$1 open.

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

Event Malf. No. Event Event No. Type* Description I 1 The crew will transfer PBlOl back to line 4.

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

SRO 2 TC06, EPR oscillations will cause fluctuations in reactor power, Electrical Pressure reactor pressure, and reactor water level. The crew will Regulator RO place the MPR in service and manually control reactor Fails - SRO pressure and raise the EPR out of the way.

Oscillates ARP A2-2-4, Nl-OP-31; H.I.0, NISOP-2, Tech Specs I 3 RRSE, Recirc C The crew will resDond to a RRP15 MG set hiah temDemture that continues to'degrade. The crew will rem"oe RRPI 5 Pump 15 BOP MG Slot from service and take appropriate actions including those Temperatur SRO actions to support 4-loop operation. When reactor power is e Increase between 45% and 90% the thermal limit penalty must be (35% ramp 3 min.) applied because no backup pressure regulator is available.

ARP F2-2-5, N l - O f - I ; F.4;H.2; H.3, R The crew will be required to lower reactor power to support removal of RRP15 from service. When manual control of RO

=I RRP15 is established, recirc flow control reduction by the SRO BOP must be coordinated with the RO.

Nl-SOP-I. 7 ED18. AC C Loss of PB16A and PB 161. The crew will recognize and Electrical Power Board respond to a reduction of drywell cooling and a loss of an BOP IAC and RBCLC Pump.

Fault (PB 16 SRO Section A)

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

NRC EXA&

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

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

(15%, 30 During and after the blowdown the crew must maintain RPV sec ramp, RELATIVE water level using high pressure and low pressure systems.

1 5 min. until active 25% N1-EOP-I; Att. 4 and 16, Nl-EOP-2, Nl-EOP-4, 1 min. ramp) 7 After containment sprays are placed into service, t Spray containment spray pump 111 trips requiring that the other Pump Trip BOP available containment spray pump be placed into service 111 (TRUE) SRO (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 EOP-8.

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

  • (N)ormal, (R)eac ivity, (I)nstrument, (C)omponent, (M)ajor TARGET QUANTITATIVE ATTRIBUTES (PER SCENARIO; SEE SECTION D.5.d) I ACTUAL ATTRIBUTES I EXAM DEVELOPER I %K
6. EOP contingencies requiring substantive actions (0-2)
7. Critical tasks (2-3) 2ofll NUREG-I 021, Draft Revision 9

NRC EXAM Atmendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point I Scenario No.: NRC-02 Op-Test No.: NRC-01 Examiners : 0perat0rs:

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).

Event Malf. No. Event Event No. Type* Description N The crew will perform NI-ST-Q4, Reactor Coolant System Isolation Valves Operability Test, quarterly surveillance on BOP the EC Loop 11 IVs per Section 8.1. After several valves are SRO 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 C The crew will respond to ECI Ivent radiation monitor alarms Emergency Condenser BOP and diagnose that a tube leak exjsts. The crew will isolate L

Tube Leak ECI 1 to stop the release.

111 (100%;

ARP Kl-1-2, EAL MATRIX, Tech Specs, OP-13 H.IO.0 ramp 5 minutes)

The crew will respond to a failure of the steam supply to the Second second stage reheater. The unbalanced condition requires Stage BOP Reheaters isolating second stage reheaters.

SRO 112 Steam Supply ARP A2-3-5, SOP-1.3, OP-41 H.7.0 Closes TU02, Main R The unbalanced condition on the main turbine results in Turbine Hi turbine vibration. If power is not lowered in response to the Vi bration RO turbine vibration, it will be lowered to 80% to allow isolation of Bearing #5 SRO and #6 the second stage reheaters.

(53%, no S 0P-I.3

\-

3ofll NUREG-1021, Draft Revision 9

NRC EXAh TU02, Main Turbine Hi The main turbine vibration degrades and a vacuum leak Vi bration develops from the vibration. Because of the lowering main Bearing #5 condenser vacuum and the rising turbine vibration, the crew and #6 (goo/,; will insert a manual reactor scram and trip the main tutbine.

no ramp)

MCOI, Main SOP-I, SOP31.3 Condenser Air In-Leakage (100%; ramp 2 minutes)

I RPOSA,B I M ATWS. When the crew scrams the reactor control rods fail RPS AIB to insert requiring actions for an ATWS with power about failure t o ALL 25%. Crew will be able to manually insert control rods using scram RMCS. Manual scrams will be successful in inserting control RP09 ARVATWS I rods but repetitive scrams will be required.

N 1-SOP-I, N I -EOP-2, N 1-EOP-3, N I -EOP-4, N I -EOP-I air header exhaust PRA: Execute EOP-3. I.

Port blocked RD33A-E Control I

Rod Bank Blocked Bank 1 , 2 ,

394,s Position 48, 24, 48, 24,48 See event 2 II C BOPSRO ERVI 11 fuse blown: failed closed and wont open because of a burned out solenoid.

ISee event 6 C Loss of main condenser vacuum. Loss of main condenser MCOI ALL as a heat sink. Challenge HCTL.

Main Cond Nl-EOP-3 Air In leakage 4 of 11 NUREG-1021, Draft Revision 9

NRC EXAM 9 LPOINB C The crew will be required to respond to a failure of the liquid LP11/12 BOP poison pump to continue to tun once started. Shortly after pump trip one LP pump is started (1-2 seconds) it will trip requiring the crew to start the other pump.

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

  • (N)ormal, 1 (R)eactivity, PRA: Inject poison solution into the reactor vessel.

(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) 5ofll NUREG-I 021, Draft Revision 9

NRC EXAM u Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point 1 Scenario No.: NRC-03 Op-Test No.: NRC-01 Examiners: 0perato rs:

Initial Conditions: 700% power.

Turnover: N7-ST-M4B, mergemy Diesel Generator 703 AND PB 703 Operability Test, completed sa tisfactory last shift. Substitute Reactor Building Ventilation Supply. . - and Exhaus Fans fron system 1 to system 7 7 .

Event Malf. Event Event No. No. Type* Description 1 N I

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

BOP OP-10 F.3.2, F.7.0, and F.2.0.

2 HVOIA. RB Exhaust C

I Reactor Building Exhaust Fan 11 trips and exhaust fan 12 Fan Trip 11 will not start. Start RBEVS in response to a degraded BOP (TRUE) Reactor Building negative pressure (0 psig).

RB ARP Ll-3-4, L7-1-5, EOP-5, OP-70 H.I.0, Tech Specs Exhaust Fan 12 &

L Outlet Damper 3 FW37. 13 C I

The crew will respond to FCV 13 oscillations. Later in the FCV Oscillation( scenario the crew will be required to manually adjust 13 FCV 50%; ramp BOP to maintain RPV level below the high level trip when reactor

= 1 minute) RO power is lowered.

C 345KV Power BOP Grid Transient (FINAL ARP A6-2-6, A6-3-3, SOP-33.A.3, Multiple Tech Specs, VALUE. OP-45; E.3.0 338, no 5 EG16, R The crew will be required to lower power to maintain Generator Cooling isophase bus duct temperatures within limits. When reactor RO Fan Leads power is lowered the temperatures stabilize then lower.

Trlp SRO (FINAL ARP A7-3-5, SOP-1.3, OP-32; H.4.0 VALUE:

50, 1 MINUTE) v 6 of 11 NUREG-I 021, Draft Revision 9

NRC EXAA EG 11, When the power reduction has been made, the grid 345KV Power Grid conditions degrade requiring removal of the main generator Transient from service because of the low frequency.

(FINAL VALUE SOP-33.A.3 (continued),SOP-33.7, SOP-7 328; 1 minute ED01B C Loss of offsite power with EDG102 fail to start and cannot be Loss of started. EDG102 and PB102 loss impact SOP5 execution Offsite ALL and RPV level and containment control actions.

115KV Power SOER 99-1; Loss of Grid South Oswego - SOER 03-1; Emergency Power Reliability Line 1 EDOIA NMP LER: Loss of grid (Summer 2003)

Loss of N1-SOP-5, EOP-2, EAL Matrix, Offsite 115KV Power JAF-Line 4 DGOIA.

Diesel Generator 102 Failure to Start (TRUE)

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 7 o f 11 NUREG-1021, Draft Revision 9

NRC EXAM 8ofll NUREG-1021, Draft Revision 9

NRC EXAM v Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point 1 Scenario No.: NRC-04 (ALT) Op-Test No.: NRC-01 Examiners: Operators:

Initial Conditions: 100% power.

ta perform ~~~~~1~ ~ ~ ~ o of ~~ ~

4 -a ~ n ~EVS c- ~

Operability

~ ~ , Test 122 Containment Spray Pump removed from service. EPR in controi.

Event Malf. Event Event No. No. Type* Description Nl-ST-M8, RBEVS Operability Test is scheduled for its RBEVS Train flow monthly performance. During the test, when the RBEVS fan meter BOP is started, the fan flow indicator will fail downscale and the downscaIe SRO train must be declared inoperable. The SRO must make a 11-M040-AO-053. T.S. determination.

Set at 0 0 N1-ST-M8, Tech Specs Loss of Reactor Trip Bus 141 and a coincident single control Reactor Trip Bus rod scram. When control rod 22-19 scrams its CRD MG Set RO Mechanism becomes stuck at position 02. The crew will Trips (141) BOP enter Nl-SOP-40.1 Loss of RPS and restore power to the (TRUE) SRO bus, then the scram may be reset. The crew must lower RD06, Rod power, asses recovering the control rod and the SRO must 22-19 Failure - make a T.S. determination.

Scrammed (IO Sec.

TD)

(TRUE) Nl-SOP-40. I , Tech Specs Overrride for control rod 22-19 position 02.

A small fuel leak will develop from the abnormal rod pattern.

RXOI, Fuel Rising reactor coolant activity levels will require entry into the Cladding RO Emergency Plan and Emergency Power Reduction.

Failure - SRO 10% Ramp

- 5 minutes, TUA- 1 EAL Matrix, EPIP-EPP-I 8, N I-SO P-I .I , N I -SOP-25.2 Min.

9ofll NUREG-I 021, Draft Revision 9

NRC EXAM ED04, AC C During the emergency power reduction the crew will transfer Power normal house service to the reserve transformers. When the Board RO Electrical transfer is made both PB 11 supply breakers will trip. The Fault BOP crew can recover PB 11. The trip of PB 11 will cause a trip (PBIl), SRO of RRPs 11 and 12, this will result in entry into the restricted clears in 3 secs 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, N I -

SOP-? .

When the reactor is scrammed (If the crew does NOT scram Emergency the reactor this malfunction will iequire a reactor scram) a Condenser BOP Tube Leak piping rupture will occur in EC Loop 11. The steam isolation 111 SRO valves fail to fully close.

ECOBA, EC STM IV Nl-SOP-1.3, Nl-SOP-I, 111 Fail to Close =

80%

EC08B. EC STM IV 112 Fail to Close =

80%

RD33E, C A bank of control rods will fail to fully insert requiring the crew Control to perform actions to manually inseft control rods. (if PBI 1 Rod Bank RO 5, Insert was not transferred previously it will fail to automatically Fail SRO transfer and must be manually transferred.)

position (48)

ED26, Nl-EOP-2, Nl-EOP-3, Nl-EOP-4; Att. 2 and 4, NI-OP-12, Failure of H.l.O and G.0 PB 11 to Auto Transfer (TRUE)

RXOl , Fuel M Rising reactor coolant activity and radiation levels will require Cladding Failure - ALL a blowdown. When EOP-8 is entered the crew must enter 100% the path for all control rods not inserted. This will require the Ramp - 15 crew to terminate and prevent injection prior to emergency minutes 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 ACTUAL EXAM FAClLlTY (PER SCENARIO; SEE SECTION D.5.d) ATTRIBUTES DEVELOPER R,EVIEW
1. Total malfunctions (5-8) 8

.rc

2. Malfunctions after EOP entry (1-2) 1 1O A J f I I
3. Abnormal events (2-4)
4. Major transients (1-2) 4 1
5. EOPs enteredkequiring substantive actions (1-2) 2 II 6. EOP contingencies requiring substantive actions (0-2) 1 I

1 I 1

7. Critical tasks (2-3) 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)

K Group 1 Tier

1. 1 3 Emerge ncy 2 0

& Abnormal Plant Tier 3 Evolutions Totals

2. 1 3 Plant 2 0 Systems Tier 3 Totals Note: 1. 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.

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. 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.
4. Systems/evolutions within each group are identified on the associated outline.
5. The shaded areas are not applicable to the category/tier.
6. 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.
7. 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.
8. For Tier 3, enter the WA numbers, descriptions, importance ratings, and point totals on Form ES-401-3.
9. 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)

FS-An1 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier l/Group 1 (RO) --

ElAPE # / Name / Safety Function K K K A A 1 2 3 1 2 GI WA Topic(s) IR 295001 Partial or Complete Loss of Forced 0 AK3.05 Knowledge of the reasons for the following 3.2 1 Core Flow Circulation / 1 & 4 5 responses as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Reduced loop operating requirements.

295003 Partial or Complete Loss of AC / 6 AA2.04 Ability to determine and/or interpret the 3.5 1 following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: System lineups.

295003 Partial or Complete Loss of AC / 6 X 2.1.32 Ability to explain and apply system limits and 3.4 1 precautions.

295004 Partial or Total Loss of DC Pwr / 6 AK3.01 Knowledge of the reasons for the following 2.6 1 responses as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER:

Load shedding.

295005 Main Turbine Generator Trip / 3 I I 0I I I AK2.09 Knowledge of the interrelations between MAIN TURBINE GENERATOR TRIP and the 4.0 1 following: Feedwater - HPCI: BWR-2.

AA1.02 Ability to operate and/or monitor the 3.9 1 following as they apply to SCRAM:

Reactor water level control system.

AA1.07 Ability to operate and/or monitor the 4.2 1 following as they apply to CONTROL ROOM ABANDONMENT:

Control room/local control transfer mechanisms.

AKI .01 Knowledge of the operational implications of 3.5 1 the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Effects on componentkystem operations.

295019 Partial or Total Loss of Inst. Air I 8 AA2.01 Ability to determine and/or interpret the 3.5 1 following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR:

295021 Loss of Shutdown Cooling / 4 3.0 1 all modes of plant operation.

295023 Refueling Acc Cooling Mode / 8 171 III AKI .01 Knowledge of the operational implications of the following concepts as they apply to REFUELING ACCIDENTS: Radiation exposure hazards.

3.6 1 295024 High Drywell Pressure / 5 EK3.04 Knowledge of the reasons for the following 3.7 1 responses as they apply to HIGH DRYWELL PRESSURE: Emergency depressurization.

295025 High Reactor Pressure / 3 EK2.08 Knowledge of the interrelations between 3.7 1 HIGH REACTOR PRESSURE and the following:

295026 Suppression Pool High Water Temp. 0 EK3.05 Knowledge of the reasons for the following 3.9 1 15 5 responses as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Reactor SCRAM.

295028 High Drywell Temperature / 5 IIIII 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) 295030 Low Suppression Pool Wtr Lvl I 5 Ill X 2.1.23 Ability to perform specific system and integrated plant procedures during different modes of plant operation. I I 3'9 295031 Reactor Low Water Level / 2 295037 SCRAM Condition Present and I4 I 0 EKI .01 Knowledge of the operational implications of the following concepts as they apply to REACTOR LOW WATER LEVEL: Adequate core cooling.

EA2.03 Ability to determine and/or interpret the I I 4.6 4.3 1 Power Above APRM Downscale or Unknown 3 following as they apply to SCRAM CONDITION

/ I PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: SBLC tank level.

295038 High Off-site Release Rate / 9 600000 Plant Fire On Site / 8 Ill 0 4

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

2.4.31 Knowledge of annunciators alarms and 4.1 1 indications / and use of the response instructions.

WA Category Totals: 3 2 4 2 4 5 Group Point Total: I 20 Page 3 of 9 NUREG-1021, Draft Revision 9

UNIT 1 NRC (RO)

L 295015 Incomplete SCRAM / 1 0 AA1.03 Ability to operate andlor monitor the following 3.6 1 3 as they apply to INCOMPLETE SCRAM: RMCS.

295022 Loss of CRD Pumps / 1 0 AA1.02 Ability to operate and/or monitor the following 3.6 1 2 as they apply to LOSS OF CRD PUMPS: RPS.

295032 High Secondary Containment Area 0 EK3.02 Knowledge of the reasons for the following 3.6 1 Temperature / 5 2 responses as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE:

Reactor SCRAM.

295036 Secondary Containment High 0 EA2.02 Ability to determine and/or interpret the 3.1 1 Sump/Area Water Level / 5 2 following as they apply to SECONDARY CONTAINMENT HIGH SUMPIAREA WATER LEVEL:

Water level in the affected area.

WA Category Point Totals: 0 0 1 2 2 2 Group PointTotal: 7 Page 4 of 9 NUREG-1021. Draft Revision 9

UNIT 1 NRC (RO)

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

I System ## / Name 205000 Shutdown Cooling WA Topic(s)

K5.02 Knowledge of the operational IR 2.8 1

implications of the following concepts as they apply to SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE):

Valve operation.

206000 HPCl K2.01 Knowledge of electrical power supplies 3.2 1 to the following: system valves.

207000 Isolation (Emergency) K5.09 Knowledge of the operational 3.7 1 Condenser implications of the following concepts as they apply to ISOLATION (EMERGENCY)

CONDENSER: Cooldown rate: BWR-2,3.

207000 Isolation (Emergency) A I .03 Ability to predict and/or monitor 3.3 1 Condenser changes in parameters associated with operating the ISOLATION (EMERGENCY)

CONDENSER controls including:

Steam flow: BWR-2,3.

209001 LPCS K4.10 Knowledge of LOW PRESSURE CORE 2.8 1 SPRAY SYSTEM design feature(s) and/or interlocks which provide for the following:

Testability of all operable components.

~~

209001 LPCS A I .08 Ability to predict and/or monitor 3.3 1 changes in parameters associated with operating the LOW PRESSURE CORE SPRAY SYSTEM controls including:

I System lineup.

.L 21 1000 SLC A2.03 Ability to (a) predict the impacts of the 3.2 1 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 A4.08 Ability to manually operate and/or 4.2 1 monitor in the control room: System initiation.

212000 RPS K1.01 Knowledge of the physical connections 3.7 1 and/or cause-effect relationships between REACTOR PROTECTION SYSTEM and the following: Nuclear instrumentation.

21 5003 IRM K4.05 Knowledge of INTERMEDIATE RANGE 2.9 1 MONITOR (IRM) SYSTEM design feature(s) and/or interlocks which provide for the following: Changing detector position.

21 5004 Source Range Monitor A I .05 Ability to predict and/or monitor 3.6 1 changes in parameters associated with operating the SOURCE RANGE MONITOR (SRM) SYSTEM controls including:

SCRAM, rod block, period alarm trip setpoints.

Page 5 of 9 NUREG-1021, Draft Revision 9

UNIT 1 NRC (RO) 215005 APRM / LPRM A3.05 Ability to monitor automatic operations 3.3 1 of the AVERAGE POWER RANGE MONlTORlLOCAL POWER RANGE MONITOR SYSTEM including:

Flow converter/comparator alarms.

215005 APRM / LPRM I 2.1.33 Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications.

3.4 1 218000 ADS 2.4.31 Knowledge of annunciators alarms and 3.3 1 indications I and use of the response instructions.

~~

223002 PCIS/Nuclear Steam A I .02 Ability to predict and/or monitor 3.7 1 Supply Shutoff changes in parameters associated with operating the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF controls including:

Valve closures.

239002 SRVs 0 K1.08 Knowledge of the physical connections 4.0 1 8 and/or cause-effect relationships between RELIEFISAFETY VALVES and the following:

Automatic depressurization system.

239002 SRVs K3.02 Knowledge of the effect that a loss or 4.2 1 malfunction of the RELIEFEAFETY VALVES will have on following:

Reactor Over-pressurization 259002 Reactor Water Level K6.03 Knowledge of the effect that a loss or 3.1 1 Control malfunction of the following will have on the REACTOR WATER LEVEL CONTROL SYSTEM: Main steam flow input.

259002 Reactor Water Level 2.4.50 Ability to verify system alarm setpoints 3.3 1 Control and operate controls identified in the alarm response manual.

261000 SGTS I A2.05 Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT 3.0 1 SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Fan trips.

262001 AC Electrical 0 K1.O1 Knowledge of the physical connections Distribution 1 andlor cause-effect relationships between A.C. ELECTRICAL DISTRIBUTION and the following: Emergency generators (dieseldet)

I 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)

-I 264000 EDGs K3.03 Knowledge of the effect that a loss or 4.1 1 malfunction of the EMERGENCY GENERATORS (DIESEL/JET) will have on following: Major loads powered from electrical buses fed by the emergency generator(s).

300000 Instrument Air A2.01 Ability to (a) predict the impacts of the 2.9 1 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.

400000 Component Cooling A4.01 Ability to manually operate and I or 3.1 1 Water monitor in the control room:

CCW indications and control.

WA Category Point Totals: Group Point Total: 26 Page 7 of 9 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)

System # I Name WA Topic(s) IR #

201001 CRD Hydraulic K3.03 Knowledge of the effect that a loss 3.1 1 or malfunction of the CONTROL ROD DRIVE HYDRAULIC SYSTEM will have on following: control rod drive mechanisms.

- ~

201006 RWM A3.03 Ability to monitor automatic 3.1 1 operations of the ROD WORTH MINIMIZER SYSTEM (RWM) including:

Annunciator and alarm signals.

202002 Recirculation Flow Control 2.4.6 Knowledge symptom based EOP 3.1 1 mitigation strategies.

214000 RPlS A2.01 Ability to (a) predict the impacts of 3.1 1 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.

216000 Nuclear Boiler Inst. K4.05 Knowledge of NUCLEAR BOILER 3.9 1 INSTRUMENTATION design feature@)

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

226001 CTMT Spray Mode 2.1 2 3 Ability to perform specificsystem 3.9 1 and integrated plant procedures during different modes of plant operation.

239001 Main and Reheat Steam A2.05 Ability to (a) predict the impacts of 3.9 1 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.

241000 Reactornurbine Pressure A I . I 5 Ability to predict and/or monitor 3.1 1 Regulator changes in parameters associated with operating the REACTORnURBlNE PRESSURE REGULATING SYSTEM controls including:

Maximum combined flow limit.

259001 Reactor Feedwater A4.04 Ability to manually operate and/or 3.1 1 monitor in the control room:

System valves.

272000 Radiation Monitoring K6.01 Knowledge of the effect that a loss 3.0 1 or malfunction of the following will have on the RADIATION MONITORING SYSTEM: Reactor Protection System.

288000 Plant Ventilation K5.02 Knowledge of the operational 3.2 1 implications of the following concepts as they apply to PLANT VENTILATION SYSTEMS: Differential pressure control.

Page 8 of 9 NUREG-1021. Draft Revision 9

UNIT 1 NRC (RO) 0 A3.01 Ability to monitor automatic 3.3 1 1 operations of the CONTROL ROOM HVAC including:

Initiation/reconfiguration.

I WA Category Point Totals: 10 Page 9 of 9 NUREG-1021, Draft Revision 9

UNIT 1 NRC (RO)

ES-401 Geneic Knowledge and Abilities Outline (Tier 3) (RO) FOITTIES-401-3 Facility: Nine Mile Point Unit I Date of Exam: November 18,2004 (tentative)

I Category

1. 2.1 .I 0

I Topic Knowledge of conditions and limitations in the RO IR 2.7 Conduct of facility license.

0perat i o ns 2.8 3.4 lineups.

Subtotal 2.2.12 Knowledge of surveillance procedures. 3.0 Equipment 2.2.13 Knowledge of tagging and clearance 3.6 Control procedures.

2.2.26 Knowledge of refueling administrative 2.5 requirement s .

I Subtotal

3. Knowledge of the process for performing a contain ment purge, Radiation Cont roI 2.3.10 Ability to perform procedures to reduce 2.9 excessive levels of radiation and guard against personnel exposure.

Subtotal Knowledge of event based EOP mitigation 3.1 strategies.

Emergency Procedures /

Plan 2.4.24 I

Subtotal Knowledge of loss of cooling water procedures. PRA: LOSS O f ESW 3.3 1 Tier 3 Point Total 1 of1 NUREG-1021, Draft Revision 9

UNIT 1 NRC (RO)

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

ES-40 1 Record of Rejected K/As (RO) Form ES-401-4 I I

--- Tier / GrouD Randomly Selected WA Reason for Reiection 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:

T IG I 295024 EK3.03 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.

T IGI 295026 EK3.03 Knowledge of the reasons for the following responses as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Suooression .. DOOIsDrav.

Have containment spray, not suppression pool spray.

T IGI 295028 2.4.30 Knowledge of which events related to system operations/status should be reported to outside aaencies. IMPORTANCE RO 2.2 SRO 3.6.

Offsite notifizations are SRO Only. Importance rating is ~ 2 . 5 .

T I GI 295030 2.2.25 Knowledge of bases in technical specifications for limiting conditions for operations

-u- and safety limits.

TS bases are SRO Only T IGI 600000 2.4.30 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 .

T I G2 NA 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.

T I G2 295009 2.4.30 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 .

T2G1 206000 A2.15 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.

T2G1 206000 A2.16 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.

T2G1 207000 K5.01 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.

T2G1 239002 K3.03 Knowledge of the effect that a loss or malfunction of the SAFETY/RELIEF VALVES will have on the followina. Abilitv to raDidlv ~ . the reactor.

deDressurize Double jeopardy with 2f9002 Ai.08.

T2G1 261000 K5.01, K5.02 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.

Io f 2 NUREG-1021, Draft Revision 9

UNIT 1 NRC (RO)

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

T2G1 I 300000 A3.01, A3.02 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.

2.1.17 Ability to make accurate /clear and concise verbal reports.

Better evaluated during the operating test (simulator scenarios).

2.1 2 1 Abilitv to obtain and verifv controlled orocedure CODY.

Better evaluated during the walkthrough examinat& (JPMs).

T3 2.2.6 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.

T3 2.2.23 Ability to track limiting conditions for operations. (CFR: 43.2 I45.13)

SRO-On1y.

T3 2.4.34 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.

T2G1 206000 A2.02 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.

T2G2 272000 A I .02 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.

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.

2of2 NUREG-1021, Draft Revision 9

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