ML070810590

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Draft - Outlines (Facility Ltr. Dtd - 12/19/2006) (Folder 2)
ML070810590
Person / Time
Site: Millstone Dominion icon.png
Issue date: 12/19/2006
From: Price J
Dominion Nuclear Connecticut
To: Marvin Sykes
Operations Branch I, NRC/RGN-I/DRS/OSB
Sykes, Marvin D.
Shared Package
ML060800095 List:
References
06-1061
Download: ML070810590 (34)
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Text

,

Dominion Nuclear Connecticut, Inc.

ivlillstone Power Station Rope Fcri-), Road Warerford, CT 06385 L/

Mr. M. D. Sykes, Chief Operational Safety Branch U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406-141 5 D Dominion D E I9 2006 Serial No.

06-1 061 Docket No.

50-423 License No.

N P F-49 MPS Lic/BAK RO DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3 SENIOR REACTOR AND REACTOR OPERATOR INITIAL EXAMINATIONS In a letter dated November 21, 2006, the NRC requested examination outlines associated with the Senior Reactor and Reactor Operator Initial Examinations -

Millstone Unit 3 (scheduled for the weeks of March 12 and 19, 2006) be provided by December 27,2006.

The requested material (examination outlines) is included as Attachment 1 and is being furnished in accordance with Title 10, Section 55.40(b)(3), of the Code of Federal Regulations (1 0 CFR 55.40(b)(3)), by an authorized representative of the facility.

To ensure the integrity of the examinations in accordance with 10 CFR 55.49 and NUREG 1021, Examiners Standard 201, Attachment 1 should be withheld from public disclosure until after the examinations have been completed. No redacted versions are being supplied.

If you have any questions or require additional information, please contact Mr. Jeff T. Spence at (860) 437-2540.

Very truly yours,

- Millstone

~

Letter from Sykes (NRC) to Christian (DNC) dated November 21, 2006,

Subject:

Senior Reactor and 1

4 Reactor Operator Initial Examinations - Millstone Unit 3.

Serial No. 06-1061 Requalification Program Inspection Page 2 of 2 u

Attachments:

7 Commitments made in this letter: None.

cc:

(w/o attachments)

U.S. Nuclear Regulatory Commission Region I Regional Administrator 475 Allendale Road King of Prussia, PA 19406-1415 Mr. V. Nerses NRC Senior Project Manager Millstone Units 2 and 3 U.S. Nuclear Regulatory Commission, Mail Stop 8 C2 One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738 d

Mr. S. M. Schneider NRC Senior Resident Inspector Millstone Power Station

E Serial No. 06-1061 Examination Outlines Millstone Power Station Unit 3 Dominion Nuclear Connecticut, Inc. (DNC)

MP3 2K7 Operating Test Change Summary

1. ES-301-1, Administrative Topics outline for the RO:

Deleted the original Equipment Control Activity (SDM with INOPERABLE Control Rods), and replaced with the following Activity:

Given a maintenance repair recommendation and reference material, recommend a clearance boundary.

2. 2K7 NRC-02, Sim Exam Overview:

Removed the Loss of Main Board Annunciation Event, as it had no identifiable event type for any position. Also, corrected the description of the source of the initial steam leak.

3. ES-D-1, Scenario outline for 2K7-NRC-02:

Removed the Loss of Main Board Annunciation Event. Made minor changes/deletions to the event type associated with all scenario outlines.

4. ES-301-5, Transient and Event Checklist:

Modified the checklist to be position specific with respect to the scenario set.

ES-40 1 PWR Examination Outline Form ES-40 1 -2 G

A 2 G*

Total Total

3. Generic Knowledge and Abilities Categories Note: I.

Ensure that at least two topics from every K/A category are sampled within each tier of the RO and

2.

3.

4.
5.
6.
7. *
8.
9.

SRO-only outlines (i. e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two).

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 If: 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.

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 ion the outline should be added. Refer to ES-401 Attachment2, for guidance regarding the eliinination of inappropriate K/A statements.

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

Absent a plant-specific priority, only those K/As 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.

Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.

The generic (G) K/As in Tiers 1 and 2 shall be selected fiom Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system.

011 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; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note # 1 does not apply). 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 lined to 1 OCFR55.43.

NUREG-1021, Revision 9

j ES-401 2

Form ES-40 I -2 Emergency and Abnorm:

I I

EIAPE # I Name J Safety Function K

I<

1 2

000007 Reactor Trip -

Stabilization - Recovery / 1 000009 Small Break LOCA / 3 00001 1 Large Break LOCA I 3 0000 1511 7 RCP Malfunctions 1 000022 Loss of Reactor Cooling Water I 8 Control System Malfunction I 3 000029 ATWS 1 1 7

000040 (\\hflE 12) Steam Line Rupture - Excessive Heat Transfer 14 000054 Loss of Main Feedwater 14 000055 Station Blackout 16 000056 Loss of Off-site Power 16 000057 Loss of Vital AC Elec.

Inst. Bus I 6 000058 Loss of DC Power 16 1

I I

000062 Loss of Nuclear KIA Topic(s)

Ability to operate / monitor AFW on 8

a reactor trip 13 Determine or interpret Charging Pump flow indication Reason for criteria for shifting to recirculation mode Ability to operate or monitor CCW 15 6

3 Ability to operate or monitor Pressurizer level trend Explain and apply system limits and precautions necessary to restart CCW while 2.1.32 3

Determinelinterpret valve lineup I bypassing a portion of the system 1 Operational implication of expansion I of liquids as temperature increases 1 9 I I Ability to determine occurrence of a turbine I reactor trip.

Operational implications of effects of feedwater introduction on dry SG 4

Ability to operate / monitor operation of HPI under total feedwater loss 4

Determinelinterpret I&C Ops Implications of cooling by natural convection.

LOP Determine i interpret impact on systems of loads Lost.

Reasons for automatic actions on I

18 Determinehterpret fail position on 3

2

ES-40 1 2

Form ES-40 1 -2 ES-40 1 EIAPE # I Name I Safety Function 0000 1 1 Large Break LOCA I 3 000026 Loss of Component Cooling Water I8 000038 Steam Gen. Tube Rupture 1 3 000055 Station Blackout / 6 000065 Loss of Instrument Air / 8 W/E 1 1 Loss of Emergency Coolant Recirc. / 4 KIA Category Totals:

PWR Examination Outline Form ES-40 1-2 normal K

I<

1 2

lant Evc K

A 3

1 utions - Tier l/Group 1 (SRO) d A I G 2

I I IR KIA Topic(s) 3 Ability to determinelinterpret 4.2 consequences of LOCA with loss of ccw for operations and safety limits A bi 1 it y to de t ermi ne/in terpret existence of SGTR, and consequences.

2.2.22 Knowledge of limiting conditions 4.1 4.8 2

2.2.25 Knowledge of bases in Tech Specs 3.7 I

I I

I 2.4.1 1 I Knowledge of Loss of Instrument I 3.6 Air Abnormal Operating Procedure 2

Adherence to appropriate procedures 4.2 1

1 1

1 1

1 -

i i

ES-40 1 3

Form ES-40 1-2 ES-40 I EIAPE # I Name I Safety Function 000005 InoperablelStuck Control Rod I 1 000024 Emergency Boration I 1

000032 Loss of Source Range N1/7 000036 Fuel Handling Accident I 8 000068 Control Room Evac. /

8 W/E 14 Loss of CTMT Integrity I 5 W/E 16 High Containment Radiation 1 9 Loss of Emergency Bus (Site Specific)

RCS Leak (Site Specific)

WA Category Point Totals:

PWR Examination Outline Form ES-40 1-2

ES-40 1 3

Form ES-40 1-2 ES-40 1 PWR Examination Outline Form ES-40 1-2 EIAPE # I Name I Safety Function

i i

i 4

Forin ES-40 1-2 ES-40 1 003 Reactor Predict/monitor temperature

4 (Continued)

Forin ES-40 1-2 ES-40 1 rou,p 1 (RP) Continued PWR Examination Outline Form ES-40 1-2 I

I ES-401 System # I Name Ability to monitor auto operation of Vital Bus amperage Effect of loss of DC on components Operate/monitor battery voltage Ability to adjust exciter volts Knowledge of cause/effect between PRM and Systems Ability to predict the impact and mitigate a detector failure.

62 AC llectrical listriblition 163 DC 3ectrical Iistribution

)63 DC 3ectrical Xstribution

)64 Emergency Iiesel Generator 173 Process iadiation Monitoring 373 Process Radiation Monitoring 076 Service Water 078 Instrument 1

3.0 1

3.5 I

2.8 1

3.3 1

3.6 1

2.7 1

s - Tier 21 A

A 1

2 2

Effect of loss of service water on TPCCW Knowledge of interlocks with 2.5 1

3.2 1

7' ) G 2

2 1

K/A Topic(s)

/ I R / #

ES-40 1 4

Forin ES-40 1-2 ES-40 1 PWR Examination Outline Forin ES-40 1 -2 System # / Name 003 Reactor Coolant Pump 005 Residual Heat Removal

I 012 Reactor Protection 062 AC Electrical Distribution 064 Emergency Diesel Generator K/A Category Point Totals:

t Systei K

K 3

4 G

I IR I

K/A Topic(s) 2.1.23 Perform integrated plant procedures during all modes Predict impact /

mitigate failure modes Predict impact /

mitigate faulty 4.0 2.9 3.6 1 bistable operation 2.1.7 I Evaluate plant I 4.4 performance and make operational 1 judgments 2.4.30 I Knowledge of events I 3.6 that should be reported to outside 3

Group Point Total:

p 5

Form ES-40 1-2 ES-40 1

S-40 1 PWR Examination Outline Form ES-40 1 -2 Plant Systems - Tier 2/Group 2 I

l l

(RO)

I I

I I

I

ystein#/Name I<

I<

I<

K K

K

)O 1 Control Rod Drive 1

~ 1 1 I 1 1 17 In-core Temperature Uonitor AMSAC (Site Specific)

X 1

2 3

4 I<nowledge that CCW 2.S I

must be cut in before I CRDMs are energized. I I 3.5 1 1 I 2.4.46 I Verify alarms are 1 consistent with plant 1

I I

I I

I conditions between SFC and Physical connections 2.7 1

RWST I

2 Predict / monitor SG 3.5 pressure response to operating controls performance and make

2. I.7 Evaluate plant 3.7 1

operational judgments of CNM pump on CNM 4

Predict / mitigate loss 2.6 1

I System I Effect of WGD on 1 3.2 1 1 I PRM/ARM System 4 1 I

I I

I Predict / monitor I 2.7 1 1 I

l l

1 1 changes with damper I

I I operations 1x1 I Ability to inanually 1 N/A 1 1 operate or monitor the SBO Diesel Operational N/A I

i

W' ES-40 1 5

Form ES-40 1-2 ES-40 I PWR Examination Outline Form ES-40 1-2 Form ES-40 1-2 1

judgments AOPs alarms and ARPs 2.4.1 1 Knowledge of 3.6 1

2.4.31 Knowledge of 3.4 I

I 3

Group Point Total:

3

i "I'

?

ES-40 1 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-40 1-3

ES-40 1 Record of Rejected WAS Form ES-40 1-4 Tier I Randomly Group Selected KIA 1/1 APE.040.AKI.04 111 APE.062.AA1.03 111 APE.065.AA1.04 111 APE.062.AA1.01 1/2 I APE.036K3.03 E+---- APE.OOl.A2.01 005.A1.07 I 007.A4.10 212 086.A1.03 313 GEN.2.3.10 1 I2 APE.001.A2.03 21 1 073.A2.02 311 GEN.2.1.13 3 I2 GEN.2.2.5 Reason for Rejection Concern with Nil1 Ductilitv TemDerature on steam break: GFS.

SWP as backup to CCW: KA Typo? (Same as APE.026.AA1.03).

Low operational validity. Simply tests loss of backup function while CCW is still functioning. More appropriate for APE.026 Oversamde: Similar to 078.K4.02. which was also selected.

Ability to operatelmonitor S WP temperature on loss of S WP:

Millstone 3 Main Boards display SWP flow and pressure, not temperature. No Loss of SWP AOP steps direct monitoring of SWP temperature.

Reasons for Fuel Handling AOP actions: Low discriminatory validity; Basis for each action is obvious.

Ability to Interpret Reactor Trip Breaker indication. Low discriminatory validity. Better tested during operating portion of the exam.

Determination of test acceptability by comparison of recorded valve response times with Tech-Spec requirements: Low difficulty.

Indication of lealtv PORV: overlaD with oDeratinn exam Predict impact1Monitor Fire Doors: overlap with Operating Exam Oversample: also selected on SRO portion of exam.

(SRO) Determine actions if auto-safety functions have not taken place: RO level knowledge; Trip the Reactor.

(SRO) Oversample: This KA Statement was already selected on RO portion of the exam.

(SRO) Vital Area Access Control: General Emdovee Knowledrre (SRO) Process for making changes to facility: Recommended for rejection by facility reviewer; not SRO job function at Millstone 3.

(Performed by engineering dept).

ES-401, Page 27 of 33

i Group Form ES-401-4 ES-40 1 Record of Rejected WAS WA Tier I Randomly Selected Paragraph 1. They were selected and rejected.

211 I 061.GEN.2.2.22 1 Knowledge of LCOs and Safety Limits: AFW not part of safety Reason for Rejection 312 313 GEN.2.2.3 1 GEN.2.3.6 3 I3 3 I4 GEN.2.3.2 GEN.2.3.1 GEN.2.3.8 GEN.2.4.43 KA<2.5 n

3 I2 KA<2.5 Emergency Communication System not operated by ROs at Millstone 3.

SRO: Ability to track LCOs (tracked in shift log) Low discriminatory validity.

SRO: Knowledge of ALARA, General Employee Knowledge SRO: Knowledge of 1 OCFR20 Limits, General Employee Knowledge SRO: SRO responsibilities for aux systems (Waste and Handling) outside control room. The Rad Waste PEO operates these systems. There is little SRO interaction, other than Discharge Permits, which is being tested on the operating exam.

GEN.2.2.23 ES-401, Page 27 of 33

Written Exam Development Process (provided per ES-40 1, Section D. 1.b):

Millstone 3 used the systematic sampling methodology described in ES-401, Attachment 1 as the process to develop the written exainination outline.

ES-30 1 Administrative Topics Outline Form ES-301-1 Facility: Millstone 3 Date of Examination: Week of 3/12/06 Examination Level:

RO SRO 0 Operating Test Number:

2K7 Administrative Topic (see Note)

Conduct of Operations RO A.1.1 Conduct of Operations RO A. 1.2 Equipment Control RO A.2 Radiation Control RO A.3 Emergency Plan Type Code*

P, D, R

~~

~

Describe activity to be performed Determine the Maximum Rate of Power Increase and Control Rod Withdrawal Limitations.

WA 2.1.25 Determine the Required Boration Time and Final Control Rod Height For a Rapid Downpower.

KIA 2.1.20 Perform a Shutdown Margin Calculation with Inoperable Control Rods.

KIA 2.2.12 Actions and Expected Response for RMS Equipment failure Alarms.

KIA 2.3.1.

\\IOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are

'Type Codes & Criteria:

retaking only the administrative topics, when all 5 are required.

(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)

(P)revious 2 exams ( 1; randomly selected)

I.

ES-3.01, Page 22 of 27 NUREG-I 021, Revision 8

i ES-301 Administrative Topics Outline Form ES-301-1 acility: Millstone 3 xamination Level:

R O O S R O R Operating Test Number:

2K7 Date of Examination: Week of 3/12/06 Administrative Topic (see Note)

Conduct of Operations SRO A.l.l Conduct of Operations SRO A. 1.2 Equipment Control SRO A.2 Radiation Control SRO A.3 Emergency Plan SRO A.4 N. R Describe activity to be performed Evaluate Technical Specifications and Technical iequirements Manual.

WA 2.1.12 Notifications and Reportability.

KIA 2.1.6

~~

Response to Door Inoperability.

WA 2.2.21 Review and Approve a Radioactive Liquid Waste Discharge Permit.

KIA 2.3.6 Emergency Plan Classification and Protective Action Recommendation for a General Emergency.

K/A 2.4.41 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)

(P)revious 2 exams ( 1; randomly selected)

ES-3.01, Page 22 of 27 NUREG-1021, Revision 8

I

j.

System / JPM Title

a.
b.

S.l / Loss of All Charging Pumps.

S.2 / Energize an AC Emergency Bus Through the RSSA during ECA-0 -0.

( 3 M S S-PT2 0 D )

c.
d.
e.

S.5 / Performance of the Immediate Actions of E-0.

f.
g.
h.

S.3 / Respond to Main Steam Pressure Transmitter Failure 5.4 / Control Rod Out of Alignment.

S.6 / Check if RCPs Should be Stopped.

S.7 / Depressurize the RCS to Refill the Pressurizer.

S.8 / Respond to Containment Sump Blockage.

ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 Type Code*

Safety Function N, A, E, s PI D, A, E, S

P, D, A, E, S

2 -004 6 - 062 4.2 - 039 D, E, S 1-001 7 - 012 4.7 - 002 3 - 010 5 - 026 N, A, E, s D, L, E, S M, A, E, S P, D, E, S Facility:

Millstone Unit 3 Date of Examination: Week of 3/12/06 Exam Level: RO SRO-I SRO-U 0 Operating Test No.:

2K7

i.
1.
k.

P.l / Primary PEO Actions on a Control Room Evacuation.

P.2 / Local Actions on a Loss of Instrument Air.

P.3 / Cross-Connect Service Water to East Switchgear ACUs.

D, R, E D, A, E 6 - 062 8 - 068 N, E 4.2 - 076 (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (L)ow-Power I Shutdown (N)ew or (M)odified from bank including 1(A)

(P)revious 2 exams (R)CA 4-6 14-6 12-3 9/ 8 1 4 1 / 1 1 1 1 / 1 1 1 2 1 2 1 1 1 1 I/

1 3 I 3 / 2 (randomly selected)

(S) im u I ator ES-301, Page 23 of 27

ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 System / JPM Title

a.
b.

S.l / Loss of All Charging Pumps.

5.2 / Energize an AC Emergency Bus Through the RSSA during ECA-0.0.

(3MSS-PT20D)

c.
d.
e.

S.5 / Performance of the Immediate Actions of E-0.

f.
g.

S.3 / Respond to Main Steam Pressure Transmitter Failure S.4 / Control Rod Out of Alignment.

S.6 / Check if RCPs Should be Stopped.

S.7 / Depressurize the RCS to Refill the Pressurizer.

~~

~~

Facility:

Millstone Unit 3 Type Code*

Safety Function N, A, E, S P, D, A, E, S

P, D, A, E, S

D, E, s N, A, E, S D, L, E, s M, A, E, S 2 -004 6 - 062 4.2 - 039 1 - 001 7 - 012 4.1 - 002 3 - 010 Date of Examination: Week of 3/12/06

h.
i.
i.

P.l / Primary PEO Actions on a Control Room Evacuation.

P.2 / Local Actions on a Loss of Instrument Air.

P.3 / Cross-Connect Service Water to East Switchgear ACUs.

Exam Level: RO SRO-I SRO-U 0 Operating Test No.:

2K7 6 - 062 8 - 068 D, R, E D, A, E N, E 4.2 - 076 I

I All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

  • Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 J 4-6 J 2-3 (C)ontrol room (D)irect from bank 9 1 8 1 4 (E)mergency or abnormal in-plant 1 / 1 1 1 (L)ow-Power I Shutdown I l l 1 1 (R)CA 1 / 1 1 1 (N)ew or (M)odified from bank including 1(A) 2 / 2 1 1 (P)revious 2 exams 3 I 3 I 2 (randomly selected)

(S) im u lator ES-301, Page 23 of 27

ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 System / JPM Title Type Code*

Facility:

Millstone Unit 3 Date of Examination: Week of 3/12/06 Safety Function Exam Level: RO SRO-I SRO-U

~

a.
b.

S.5 / Performance of the Immediate Actions of E-0.

c.

S.1 / Loss of All Charging Pumps. (New)

S.6 / Check if RCPs Should be Stopped.

Operating Test No.:

2K7 N, A, E, s Nl A, E, s (ESF)

D, L, El s 2 -004 7 - 012 4.1 - 002

d.

P.l / Primary PEO Actions on a Control Room Evacuation.

D, R, E

e.

P.2 / Local Actions on a Loss of Instrument Air.

D, A, E 6 - 062 8 - 068

~~~~~~

~~

All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

  • Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 14-6 12-3 (C)ontrol room (D)irect from bank 91 81 4 (E)mergency or abnormal in-plant 1 / 1 1 1 (L)ow-Power / Shutdown 1 / 1 1 1 (N)ew or (M)odified from bank including 1 (A) 2 / 2 1 1 (WCA I/ I/ 1 (P)revious 2 exams 3 / 3 I 2 (randomly selected)

(S)imulator ES-301, Page 23 of 27

Appendix D Scenario Outline Forin ES-D-1 Facility: Millstone 3 Scenario No.: 2K7 NRC-01 012-Test No.:

2K7 Examiners:

Operators:

[nitial Conditions:

IC-21, 100% power, Eiid of Life, Equilibrium Xe.

Turnover:

The plant is at 100% power and at end of life. The A Emergency Diesel Generator is out of service for major maintenance. The C TPCCW pump is out of service for oil replacement.

Event No.

1 2

3 4

5 6

7 8

Malf. No MSO5B RXlOA RX14G RD1160 ED09D RC14C RC02C RP07A/B FW20 RC02C RPO6A/B Event Type C (BOP)

N (SRO)

R (SRO)

N (BOP)

R (RO) 1 (RO)

I (BOP)

C (RO)

Event Description Moisture Separator Relieater tube leak and subsequent procedurally required downpower. AOP 3575, Rapid Do wnpo wer.

Controlling channel of PZR level fails as is (3RCS*L459) in coniunctioii with downuower.

D Steam Generator steam flow instrument fails low Control rod nosition indication failure, Data B (rod M4).

(3MSS-FT542).

Loss of Battery Bus 4.

RPCCW leak into umer oil reservoir of C RCP.

A I

Small break LOCA inside CTMT. Safety Injection fails to auto-actuate. AFW pumps fail to auto-start.

Large break LOCA inside CTMT. CDA fails to auto-actuate.

T\\J)ormal, (R)eactivity, (I)nstmment, (C)omponent, (M)ajor Appendix D, Page 38 of 39

EXAM OVERVIEW

Title:

Large Break LOCA ID Number:

2K7 NRC-01 Exam Brief:

Revision: 0

1.

service for planned maintenance. The C TPCCW pump is out of service for an oil replacement. After taking the shift, a tube leak will occur in the B MSR. The crew will use OP 3317, Reheat and Moisture Separator, to identify the MSR leakage and determine required actions, which include a power reduction of 10% followed by taking the MSRs out of service. Plant Management will direct the crew to lower power using AOP 3575, Rapid Downpower, @ 1/2% per minute. During the downpower, controlling PZR Level transmitter, 3RCS-L459, will fail as is. The crew should respond by entering AOP 3571, Response to an Instrument Failure, to address the problem.

3MSS*FT522, will fail low. Again, the crew should enter AOP 3571 Instrument Failure Response to address the failure.

Once actions as specified in AOP 3571 are complete, the Digital Rod Position Indication system will develop a Data B failure on rod M4. AOP 3552, Malfunction of the Rod Drive System, will be used by the crew to mitigate the event.

and will respond using AOP 3563, Loss of DC Bus Power. There are no Main Board actions other than verifying system response. The US should evaluate and enter the appropriate Tech Spec. If attempted, power cannot be restored to the Battery Bus.

Once Tech Specs have been addressed, a leak (tube failure) of Reactor Plant Component Cooling Water (CCP) into the Upper Oil Reservoir of the C RCP will occur. CCP Surge Tank level, though slight, will discernibly decrease and, after a few minutes of leakage, an Oil Reservoir high level alarm will sound. The crew should take action per the ARP and AOP 3554, RCP Trip or Stopping a RCP af Power to attempt to reduce reactor power and take the affected RCP out of service. The crew will not have the time to downpower prior to exceeding RCP motor bearing temperature limits and should elect to trip the plant and the affected RCP. If the crew takes no action the affected RCP will seize and the reactor will trip.

area of the affected RCP). The crew will carry out the immediate actions of E-0, Reactor Trip and Safety Injection, determine that automatic safety injection (SI) actuation has failed, and manually initiate SI [Critical Task]. While in E-0, the crew should recognize that the AFW pumps did not start and take action to start the pumps [Critical Task].

The crew should then transition to E-I, Loss of Reactor or Secondary Coolant, and stop the RHR pumps when procedurally directed. After the crew trips the RHR pumps, the LOCA will rapidly increase in severity forcing the restart of the RHR pumps. As the break size increases an Orange Path will be generated based on CTMT pressure. The crew should respond by transitioning to FR-Z.l, Response to high CTMT Pressure. CDA Train A & 6 will fail to automatically or manually actuate. This will require the crew to manually manipulate individual components [Critical Task]. If The Session will begin with the plant at 100% power and at end of life. The A EDG is out of Once Tech Specs have been addressed, and bistables tripped, the SG steam flow channel, When AOP 3552 actions are complete, the crew will experience a Loss of 125 Volt DC Bus 4 At the time of the reactor trip a small break LOCA on loop 3 occurs (simulating a break in the

the operator uses the pushbuttons to manually actuate CDA, the RHR pumps will not automatically start. The RHR pumps will need to be started using the hand switches on MB2. The RHR pumps will start and provide flow. Once FR-Z.l is complete, the crew transitions back to E-I, the session can be terminated.

2.

The US should classify the event as an ALERT - Charlie One based on failure of the RCS barrier.

3.

Duration of Exam:

75 minutes i

Scenario Outline Form ES-D-1 Appendix D Event No.

1 2

Facility: Millstone 3 Scenario No.: 2K7 NRC-02 Op-Test No.: 2K7 Malf. No Event Event Type Description RX09A I (RO)

Controlling channel of PZR pressure fails high.

RX13E I (BOP)

Controlling chaiinel of C Steam Generator feed flow Examiners:

Operators :

6 Initial Conditions: IC-2 1, 100% power, End of Life, Equilibrium Xe.

N (BOP) to open.

C (RO)

EGO 1 C (RO)

Main Generator trip and automatic reactor trip failure.

Turnover:

The plant is at 100% power and at end of life. The A Emergency Diesel Generator is out of service for routine maintenance. The C TPCCW pump is out of service for oil replacement.

I I

I 7

RP 1 OA/B C (BOP)

M (ALL) Four faulted Steam Generators. Main Steam Isolation fails MS07C/D C (BOP) to automatically actuate. Several RPCCW components fail MS12A/B C (RO) to respond to a Safety Injection signal.

MS02A RP08 RPllH

, XO)

RCS leak. Reactor Vessel Flange 1

)4A I I (RO)

Loop 1 Tavg fails higl-I MV8104 I R (SRO) I 3575, Rapid D O I. I ~ J O M ~.

Eniergeiicy Boration valve tails

EXAM OVERVIEW

Title:

Four Faulted Steam Generators ID Number:

2K7 NRC-02 Exam Brief:

Revision: 0

1.

service for planned maintenance. The C TPCCW pump is out of service for an oil replacement.

Shortly after turnover is complete a Pressurizer Pressure Instrument will fail high. This failure will require the use of AOP 3571, lnstrument Failure Response, to respond to the failure. The procedure should be completed up to and including addressing any required Technical Specifications.

The Session will begin with the plant at 100% power and at end of life. The A EDG is out of After completion of AOP 3571, a feed flow instrument failure will occur on the C Steam Generator. Again, the crew should enter AOP 3571 Instrument Failure Response to address the failure.

Once feed control has been regained and C SG water level stable, Valve stem leakoff from valve 3CHS*MV8438B will significantly increase to about 20 gpm. VCT level will decrease, PDTT level will increase and the annunciator for VCT low pressure will alarm. The crew should enter AOP 3555, Reactor Coolant Leak to mitigate the event. AOP 3555 will direct the crew carry out the actions in attachment E to determine the source. Leakage will be in excess of the Tech Spec limit on Identified RCS leakage.

Once the actions of AOP 3555 are complete and the appropriate Tech Specs addressed, a partial loss of Main Board annunciation will occur due to a problem associated with 3BYS-PNL-5-1. The crew should diagnose the failure and use AOP 3574, Loss of Main Board Annunciation to respond.

Power cannot be restored to the affected annunciators.

After the crew has worked their way through AOP 3574, a narrow range Tc instrument will fail high. No annunciation will occur, but rods will begin stepping in. The crew will utilize AOP 3571, Instrument Failure Response to mitigate the failure.

When AOP 3571 is complete, CONVEX will direct the crew to begin an Emergency Load Reduction decreasing unit electrical output by 300 MWe. The crew will use AOP 3575, Rapid Downpower to accomplish this down power. The emergency boration valve, 3CHS*MV8104, will fail to open and the RO will have to use the RNO steps to achieve boration flow.

Once the crew has completed the downpower, the electrical grid will become unstable resulting in a main generator trip. The reactor will fail to automatically trip [Critical Task], resulting in SG pressures increasing. A steam break will occur upstream of the A S/G inside the MSlV Building.

Once the plant is tripped, the A

& B MSlVs will fail to close and the C and ID SG low set safety valves will stick open. Main Steamline Isolation (MSI) will fail to automatically actuate, necessitating the crew to manually initiate MSI [Critical Task]. Additionally, several RPCCW components will fail to respond as required to the safety injection signal and will have to be manually positioned by the crew.

The crew should proceed through E-0 to E-2 to ECA-2.1. After progressing into the SI Termination steps of ECA-2.1, the C low set safety valve will close. The scenario will end when the crew identifies the safety valve closure and pressure increasing in the C SG, and discusses the transition to E-2, once SI termination is complete.

'b'

2.

The SM/ US should classify this event as an ALERT based on failure of automatic reactor trip (EAI). The event is also classifiable at the ALERT level based on Unisolable Steam Line Break 5

outside CTMT (BAZ).

3.

Duration of Exam:

75 minutes

EXAM OVERVIEW i/

Title:

Four Faulted Steam Generators ID Number:

2K7 NRC-02 Exam Brief:

1.

service for planned maintenance. The C TPCCW pump is out of service for an oil replacement.

Shortly after turnover is complete a Pressurizer Pressure Instrument will fail high. This failure will require the use of AOP 3571 I lnsfrurnenf Failure Response, to respond to the failure. The procedure should be completed up to and including addressing any required Technical Specifications.

The Session will begin with the plant at 100% power and at end of life. The A EDG is out of After completion of AOP 3571 a feed flow instrument failure will occur on the C Steam Generator. Again, the crew should enter AOP 3571 Instrument Failwe Response to address the failure.

Once feed control has been regained and C SG water level stable, Valve stem leakoff from valve 3CHS*MV8438B will significantly increase to about 20 gpm. VCT level will decrease, PDTT level will increase and the annunciator for VCT low pressure will alarm. The crew should enter AOP 3555, Reactor Coolant Leak to mitigate the event. AOP 3555 will direct the crew carry out the actions in attachment E to determine the source. Leakage will be in excess of the Tech Spec limit on Identified RCS leakage.

4 After the crew has worked their way through AOP 3574, a narrow range Tc instrument will fail high. Rods will begin stepping in. The crew will utilize AOP 3571, Instrument Failure Response to mitigate the failure.

When AOP 3571 is complete, CONVEX will direct the crew to begin an Emergency Load Reduction decreasing unit electrical output by 300 MWe. The crew will use AOP 3575, Rapid Downpower to accomplish this down power. The emergency boration valve, 3CHS*MV8104, will fail to open and the RO will have to use the RNO steps to achieve boration flow.

Once the crew has completed the downpower, the electrical grid will become unstable resulting in a main generator trip. The reactor will fail to automatically trip [Critical Task], resulting in SG pressures increasing. A steam break will occur upstream of the A MSlV inside the Main Steam Valve Building. Once the plant is tripped, the A

& B MSlVs will fail to close and the C and D SG low set safety valves will stick open. Main Steamline Isolation (MSI) will fail to automatically actuate, necessitating the crew to manually initiate MSI [Critical Task]. Additionally, several RPCCW components will fail to respond as required to the safety injection signal and will have to be manually positioned by the crew. The crew should proceed through E-0 to E-2 to ECA-2.1. After progressing into the SI Termination steps of ECA-2.1, the C low set safety valve will close. The scenario will end when the crew identifies the safety valve closure and pressure increasing in the C SG, and discusses the transition to E-2, once SI termination is complete.

2.

The SMI US should classify this event as an ALERT based on failure of automatic reactor trip (EAI). The event is also classifiable at the ALERT level based on Unisolable Steam Line Break d

outside CTMT (BA2).

3.

4 Duration of Exam:

75 minutes

Appendix D Scenario Outline Form ES-D-1 Event I

I No.

2 3

4 5

6 (1 Facility: Millstone 3 Scenario No.: 2K7 NRC-03 (spare)

Op-Test No.:

2K7 Event Type*

R (RO)

R (SRO)

N (BOP)

C (RO) c (RO)

Examiners:

Operators:

Event Description Power ascension from 3% to 8% power using OP 3203, Plant Startup. Running RPCCW pump trip. Alignment and start of the standby RPCCW pump.

Small leak develops through pressurizer PORV, Initial Conditions: IC-07 (modified), 3% power, Beginning of Life, No Xe.

I (BOP)

Turnover:

The crew will take the shift with reactor power stable at the point of adding heat (reactor from a refueling outage. OP 3203, Plant Startup is in progress.

ower 3%), following a reactor startup by the previous shift. This is the initial plant startup I

3RCS*PCV455A.

Power Range Nuclear Instrument (NI) Channel 43 Lower C (RO)

N (BOP)

M (ALL)

Malf. No CCOlB Detector fails high requiring FRV Bypass valve controllers to be placed in manual.

A RCP #I seal degradation resulting in high RCP seal leakoff. RCP is tripped using AOP 3554, RCP Trip 01 Reinoving n RcP.fi.0~1 Service At Poiiw..

Loss of offsite power, B Emergency Diesel Generator (EDG) fails to automatically or manually start from the RC07A NI09C C (BOP)

CV13A control room.

A EDG trips resulting in a loss of all AC power.

ED0 1 EGOGB C (RO)

EG08A EG07A FW20C TDAFW pump fails to auto-start. B EDG is started locally. B service water pumps fail to auto-start after B EDG is successfully started locally.

N)ornial,

(

I Appendix D, Page 38 of 39

Title:

Loss of All AC Power ID Number:

2K7 NRC-03 EXAM OVERVIEW Revision: Q Exam Brief:

1.

power 3%), following a reactor startup by the previous shift. This is the initial plant startup from a refueling outage. OP 3203, Plant Startup is in progress and complete up through step 4.2.7. The crew is to raise reactor power from 3% to 6 to 9% in accordance with step 4.2.8. The US should facilitate a brief of the evolution prior to taking the shift. The MP3 simulator briefing room may be used for this purpose.

The crew will take the shift with reactor power stable at the point of adding heat (reactor During the power ascension, the B RPCCW pump will spuriously trip and will not restart.

The crew will use AOP 3561 I Loss of RPCCW, to verify alignment and start the C RPCCW pump on the B Train.

Once the reactor is in MODE 1 (6 to 9%) and the primary plant stable, a small leak will develop through pressurizer PORV, 3RCS*PCV455A. The crew should use ARP MB4A 3-5, PZR RELlEF VALVE DIS TEMP HI, to respond. Correct actions will include closing the A PORV Block valve to isolate the leaky PORV. This event will exercise the US in Tech Spec and TRM use.

Once the reactor is 6 to 9% and the primary plant stable, Power Range Nuclear lnstrument (NI)

Channel 43 Lower Detector will fail high and the appropriate annunciators will alarm. Rod control is in manual so no rod motion will occur, but the auctioneered high PRNl channel inputs to the control circuitry for the FRV Bypass valves. The FRV Bypass valve controllers should be placed in manual and the crew should respond using AOP 3571 I lnstrument failure Response. Tech Specs will need to be addressed.

Once Tech Specs are addressed for the failed NI, A RCP # I seal will slowly degrade. The crew should address the RCP seal degradation using the Main Board 4 Annunciator Response Procedure for RCP HI RANGE LKG FLOWHI. The procedure requirements will have the operator transition to AOP 3554, RCP Trip or Removing a RCP from Service At Power, to stop the affected RCP. The procedure will then direct the crew to commence a plant and reactor shutdown.

After the RCP is tripped and a downpower plan discussed, transmission grid instabilities will result in a loss of offsite power. The B EDG will not start due to its inability to respond to signals from the Sequencer or MB8. Though the A EDG will initially start it will not load and will exhibit degraded frequency due to damaged governor linkage. The crew may conservatively decide to shutdown the EDG; if not the A EDG will trip at step 4 of E-0 and will not be able to be re-started.

Also, at the time of the loss of power, an RCS leak will occur (A RCP #I seal catastrophically fails).

The crew should enter E-0, Reactor Trip and Safety Injection, and once the A EDG trips, transition to ECA 0.0, Loss ofAZZ AC Power. Prior to or at step 4 of ECA-0.0, the crew should diagnose that the TDAFW pump failed to auto-start and manually start the TDAFW [Critical Task].

Plant assistance (a PEO/NLO, Maintenance, Engineering, etc.) should be dispatched to both EDGs to ascertain the reason for the start failures. The B EDG will be able to be started locally using ECA-

J 0.0, Attachment E. The B EDG local start attempts will succeed when the crew is aligning the selected train busses for the Station Blackout Diesel. Service Water will need to be restored to the running EDG [Critical Task]. The crew should then Go To step 26 as per the note prior to step 6.

Low pressurizer level will necessitate a transition to ECA 0.2, Loss of All AC Power - Recovery with SI Required. The session will end when the crew announces the transition to ECA-0.2.

L/

2.

The SM should classify the event as an Site Area Emergency - Charlie Two, Loss of Voltage on Buses 34C and 34D > 15 minutes (EAL PSI).

3.

Duration of Exam:

120 minutes L

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