ML19122A209

From kanterella
Jump to navigation Jump to search

Draft Written Examination and Operating Test Outlines (Folder 2)
ML19122A209
Person / Time
Site: Nine Mile Point 
Issue date: 11/30/2018
From: Peter Presby
Operations Branch I
To: Isham P
Exelon Generation Co
Shared Package
ML18338A499 List:
References
CAC00500, EPID L -2018-OLL-0007, NMP1L3230
Download: ML19122A209 (34)
v • d • e


Text

Exelon Generation.

NMP1L3230 July 2, 2018 U.S. Nuclear Regulatory Commission Attn: Regional Administrator, Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713

Subject:

Nine Mile Point Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-63 NRG Docket No. 50-220 Nine Mile Point Unit 1 Initial License Examination Outlines 10 CFR 55.40

Reference:

(1)

Letter from D. E. Jackson (NRG) to B. Hanson (Exelon Nuclear), dated April 26,

2018, Senior Reactor and Reactor Operator Initial License Examinations - (Nine Mile Point, Unit 1)

As discussed in Reference (1 ), arrangements have been made for the administration of license examinations at Nine Mile Point Unit 1 during the week of December 3, 2018.

The examinations are being prepared based on guidelines in NUREG-1021, "Operator Licensing Examination Standards for Power Reactors," Revision 11.

Enclosed are the approved examination outlines for the Unit 1 Senior Reactor and Reactor Operator initial license examinations. The written examination outline was originally developed by the Chief Examiner, Mr. Peter Presby.

In accordance with NUREG-1021, ES-201, "Initial Operator Licensing Examination Process,", Nine Mile Point Nuclear Station, LLC (NMPNS) requests that the examination materials be withheld from public disclosure until two years after the examinations have been completed. The enclosed materials are appropriately marked in accordance with NUREG-1021.

Should you have any questions regarding the information in this submittal, please contact Greg Elkins, Manager Operations Training, at (315) 349-1261.

Sincerely, James N. Tsardakas Director Site Training, Nine Mile Point Nuclear Station Exelon Generation Company, LLC JNT/RSP

NMP1 Initial License Examination Outlines July 2, 2018 Page 2

Enclosure:

Examination Outlines for the Unit 1 Senior Reactor and Reactor Operator Initial License Examinations cc:

P. Presby, NRC Chief Examiner (with enclosure)

D. Jackson, NRC (without enclosure)

NRC Resident Inspector (without enclosure)

Enclosure Examination Outlines for the Unit 1 Senior Reactor and Reactor Operator Initial License Examinations

ES-201 Examination Preparation Checklist Form ES-201-1 Facility:

Nine Mile Point Unit 1 Date of Examination:

12/03/2018 II Developed by: Written: Facility [Zl NRC D II Operating:

Facility [Zl NRC D Target Task Description (Reference)

Chief Examiner's Date*

Initials

-240

1.

Examination administration date confirmed (C.l.a; C.2.a-b). For NRC-prepared exams,

.IJZ, 04/06/2018 arrangements are made for the facility to submit reference materials (C.1.e; C.3.c; Attachment 3).

-210

2.

NRC examiners and facility contact assigned (C.l.d; C.2.f).

.e_ 04/26/2018

3.

Facility contact briefed on security and other requirements (C.2.c). As applicable, the facility I~ 04/26/2018

-210 contact submits to the NRC any prescreened K/As for elimination from the written examination outline, with a description of the facility's prescreening process (ES-401, D.l.b).

-210

4.

Reference material due for NRC-prepared exams (C.l.e; C.3.c; Attachment 3). O

~ N/A

-210

5.

Corporate notification letter sent (C.2.e).

~

04/30/2018

6.

NRC-developed written examination outline (ES-401-1/2 or ES-401N-1/2 and ES-401-3 or

~01/02/2018

-195 ES-401N-3) sent to facility contact (must be on the exam security agreement) (C.l.e-f; C.2.h; C.3.d-e).

-150

7.

Operating test outline(s) and other checklists due, including Forms ES-201-2, ES-201-3, ES-301-1,

~

07/03/2018 ES-301-2, ES-301-5, and ES-D-1, as applicable (C.1.e-f; C.3.d-e).

-136

8.

Operating test outline(s) reviewed by the NRC and feedback provided to facility license~(C.2.h;

~

07/19/2018 C.3.d-e).

9.

Proposed examinations (written, JPMs, and scenarios, as applicable) and outlines (Forms ES-301-1,

~09/18/2018

-75 ES-301-2, ES-D-1, ES-401-1/2 or ES-401N-1/2, and ES-401-3 or ES-401N-3); supporting documentation (including Forms ES-301-3, ES-301-4, ES-301-5, ES-301-6, ES-401-6, ES-401N-6, and '

any Form ES-201-2 and ES-201-3 updates); and reference materials due (C.l.e-h; C.3.d).

-75

10. Examinations prepared by the NRC are approved by the NRC supervisor and forwarded for facility l#ON/A licensee review (C.1.i; C.2.h; C.3.f-g).

-60

11. Preliminary waiver/excusal requests due (C.l.m; C.2.c; ES-202).

~

10/03/2018

-so

12. Written exam and operating test reviews completed (C.3.f).

~

10/13/2018

-35

13. Examination review results discussed between the NRC and facility licensee (C.1.i; C.1.k-1; C.2.h;

~

10/28/2018 C.3.g). The NRC and the facility licensee conduct exam preparatory week.

xx/xx/2018

-30

14. Preliminary license applications and waiver/excusal requests, as applicable (NRC Form 398) due

,ie 11/02/2018 (C.l.m; C.2.i; ES-202).

-14

15. Final license applications and waiver/excusal requests, as applicable (NRC Form 398), due and

~11/18/2018 Form ES-201-4 prepared (C.l.m; C.2.k; ES-202).

-7

16. Written examinations and operating tests approved by the NRC supervisor (C.2.j-k; C.3.h).

~

11/25/2018

-7

17. Request facility licensee management feedback on the examination (C.2.1).

~

11/25/2018

18. Final applications reviewed; one or two (if more than 10) applications audited to confirm

-7 qualifications/eligibility; and examination approval and waiver/excusal letters sent (C.2.k;

~

11/25/2018 ; ES-202, C.3.j; ES-204).

-7

19. Proctoring/written exam administration guidelines reviewed with facility licensee (C.3.k).

a 11,2512018

-7

20. Approved scenarios and job performance measures distributed to NRC examiners (C.3.i).

~11/25/2018

  • Target dates are based on facility-prepared examinations and the examination date identified in the corporate notification letter.

These dates are for planning purposes and may be adjusted on a case-by-case basis in coordination with the facility licensee.

ES-201 Examination Outline Quality Checklist Form ES-201-2 Facility: Nine Mile Point Unit 1 Date of Examination: December 2018 Initials Item Task Description a

b*

c**

1.
a. Verify that the outline(s) fit(s) the appropriate model in accordance with ES-401 or ES-401 N.

pp1 ~,J ~

w

b.

Assess whether the outline was systematically and randomly prepared in accordance with Pff ~ ~

R Section D.1 of ES-401 or ES-401 N and whether all KIA categories are appropriately sampled.

I Pf1

~

T C.

Assess whether the outline overemphasizes any systems, evolutions, or generic topics.

,~

T E

d. Assess whether the justifications for deselected or rejected KIA statements are appropriate.

rP1 f""' ~

N

2.
a.

Using Form ES-301-5, verify that the proposed scenario sets cover the required number of Pfl normal evolutions, instrument and component failures, technical specifications, and major

,,.tJ ~

s transients.

I M

b.

Assess whether there are enough scenario sets (and spares) to test the projected number and u

mix of applicants in accordance with the expected crew composition and rotation schedule f Pl,,,,..,J ~

L without compromising exam integrity, and ensure that each applicant can be tested using at A

least one new or significantly modified scenario, that no scenarios are duplicated from the T

applicants' audit test(s), and that scenarios will not be repeated on subsequent days.

0 To the extent possible, assess whether the outline(s) conforms with the qualitative and R

C.

~

quantitative criteria specified on Form ES-301-4 and described in Appendix D and in ff/

Section D.5, "Specific Instructions for the 'Simulator Operating Test,"' of ES-301 (including overlap).

3.
a. Verify that the systems walkthrough outline meets the criteria specified on Form ES-301-2:

(1)

The outline(s) contains the required number of control room and in-plant tasks distributed w

among the safety functions as specified on the form.

Pf1 ~ A, A

(2)

Task repetition from the last two NRC examinations is within the limits specified on the form.

L (3)

No tasks are duplicated from the applicant's audit test(s).

K (4)

The number of new or modified tasks meets or exceeds the minimums specified on the form.

T (5)

The number of alternate-path, low-power, emergency, and radiologically controlled area H

tasks meets the criteria on the form.

R 0

b. Verify that the administrative outline meets the criteria specified on Form ES-301-1:

Pf1 u

(1)

The tasks are distributed among the topics as specified on the form.

fit."' ~

G (2)

At least one task is new or significantly modified.

H (3)

No more than one task is repeated from the last two NRC licensing examinations.

c.

Determine whether there are enough different outlines to test the projected number and mix of applicants and ensure that no items are duplicated on subsequent days.

Pff ft,..,1 ~

4.
a.

Assess whether plant-specific priorities (including probabilistic risk assessment and individual 1¥1 (i¥J ~

plant examination insights) are covered in the appropriate exam sections.

G Pf! /1111 ~

E

b. Assess whether the 1 O CFR 55.41, 55.43, and 55.45 sampling is appropriate.

N Ensure that KIA importance ratings (except for plant-specific priorities) are at least 2.5.

pr, kz)

E C.

R

d. Check for duplication and overlap among exam sections and the last two NRC exams.

Pf1,,,.,., ~

A rn fJlll!I ~

L

e. Check the entire exam for balance of coverage.
f.

Assess whether the exam fits the appropriate job level (RO or SRO).

Pft fWJ :4

~~~ture Date

a.

Author Paul Isham/

'/!_7/!f

b.

Facility Reviewer (*)

Phil Nichols/

/&::,.

'i.1. tt C.

NRC Chief Examiner (#)

P. Pre~J:,.., /, ~__g -:J /' 11. I\\ /)

ll/'l/i~

d.

NRC Supervisor n,..... -rJ 1.1.;_1.r-__ /A'i.JJJ '.J..t\\V in.1_ 1.,,-

\\. J

,2.73 if ""

  • Not applicable for NRG-prepared examination outlines.
  1. The independent NRC reviewer initials items in column "c"; the chief examiner's concurrence is required.

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Nine Mile Point Unit 1 Date of Examination: December 2018 Examination Level: RO Operating Test Number:

LC1 17-1 NRG Administrative Topic Type Describe activity to be performed (see Note)

Code*

Verification Of Active License Status Conduct of Operations M,R OP-AA-105-101, OP-AA-105-102, KIA 2.1.4 (3.3)

DWFDT / DWEDT Leak Rate Determination and Evaluation Conduct of Operations D,R N1-0P-8, KIA 2.1.18 (3.6)

Develop a clearance boundary for the Liquid Poison Equipment Control D,R Test Tank OP-CE-109-101 KA 2.2.13 (4.1)

P,D,R Application of Radiation Exposure Limits IAW RP-AA-Radiation Control 203 - SOC Room (2017 NRG)

RP-AA-203, KIA 2.3.4 (3.2)

Emergency Procedures/Plan 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 (:5 3 for ROs; :;; 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (2: 1)

(P)revious 2 exams (:5 1; randomly selected)

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Nine Mile Point Unit 1 Date of Examination: December 2018 Examination Level: SRO Operating Test Number: LC1 17-1 NRC Administrative Topic Type Describe activity to be performed (see Note)

Code*

Reactivate SRO Licenses Conduct of Operations D,R OP-AA-105-102, KA 2.1.4 (3.8)

Perform Time to Boil Calculation for Reactor Coolant Conduct of Operations D,R System OP-NM-108-117-1002, KIA 2.1.40 (3.9)

Review and Approval of Completed Surveillance Test, N1-ST-Q13, Emergency Service Water Pump and Equipment Control N,R Check Valve Operability Test N1-ST-Q13, KIA 2.2.12 (4.1)

P,D,R Determine Actions for Inoperable Service Water Radiation Control Radiation Monitor (2017 NRC)

N1-ARP-H1, ODCM, KIA 2.3.15 (3.1)

Emergency Event Reclassification and Notification Emergency Procedures/Plan D,R EP-CE-111, EPIP-EPP-01 EAL Flowchart, KIA 2.4.41 (4.6)

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 (:5 3 for ROs; ::;; 4 for SROs & RO retakes)

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

(P)revious 2 exams (:5 1; randomly selected)

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Nine Mile Point Unit 1 Date of Examination: December 2018 Exam Level: RO/SR0-1/SRO-U Operating Test No.: LC1 17-1 NRC Control Room Systems* (8 for RO); (7 for SRO-I) ; (2 or 3 for SRO-U)

System I JPM Title Type Code*

Safety Function

a. Swap CRD Pumps M,A,S 1

KIA 201001 A4.01 (3.1/3.1), N1-0P-5

b. Perform N1-ST-M8, Reactor Building Emergency Ventilation System Operability Test N,S,EN 9

KIA 288000 A4.01 (3.1/2.9), N1-ST-M8

c. Vent the Drywell Prior to Personnel Entry >212 M,S,L,A 5

KIA 223001 A4.03 (3.4/3.4), N1-0P-9

d. Rapid RWCU System Restoration for Level Control (RO Only)

D,S,L 2

KIA 204000 A4.06 (3.0/2.9), N1-EOP-HC

e. Restore Emergency Condenser To Service D,A,EN,S 4

KIA 207000 A4.05 (3.5/3.7), N1-0P-13

f. Swap PB 101 from 1014 to R1011 KIA 262001 A4.01 (3.4/3.7), N1-0P-30 D,S 6
g. Control Rod Exercising Operability Test P,D,A,S 7

KIA 214000 A4.02 (3.8/3.8), N1-ST-W1, N1-0P-5 (2015 NRC)

h. MSIV Stroke test and Limit Switch Test P,S, D 3

KIA 239001 A4.01 (4.2/4.0), N1-ST-Q26 (2015 NRC)

In-Plant Systems* (3 for RO); (3 for SRO-I) ; (3 or 2 for SRO-U)

i. Swap CRD Stabilizing Valves D,R 1

KIA 201001 A2.08 (2.8/2.8), N1-0P-5

j. Lineup Lake Water to Supply the Emergency Condenser D,E,A,R Makeup Tanks using the Electric Fire Pump 4

KIA 207000 A1.01 (3.7/3.8), N1-SOP-21.2

k. Supply Emergency Cooling Water to EDG from the Diesel Fire Pump D,E,R 8

KIA 400000 K1.02 (3.2/3.4), N1-0P-45

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 (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power / Shutdown (N)ew or (M)odified from bank including 1 (A)

(P)revious 2 exams (R)CA (S)imulator Pairings:

A then B Ethen F Criteria for RO/ SRO-I/ SRO-U 4-6 I 4-6 I 2-3

59/:58/:54

.::1/.::1/.::1

.:: 1 /.::1 /.::1 (control room system)

.::1/.::1/.::1

.::2/.::2/.::1

5 3 I :5 3 I :5 2 (randomly selected)

.::1/.::1/.::1

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point Unit 1 Scenario No.: NRC-2 Op-Test No.: LC1 17-1 NRC Examiners:

Operators:

Initial Conditions: The plant is operating at approximately 90% power. Containment Spray Pump 112 is out of service for maintenance. Steam Packing Exhauster 12 is out of service due to high vibrations.

PB 11 is aligned to reserve power in preparation for cross-tying PB 16.

Turnover: Cross-tie PB 16A to PB 16B with PB 168 supplying. Power board 11 will remain aligned to reserve power. Then, raise reactor power to 95% using recirc flow.

Event Malf.

Event Event No.

No.

Type*

Description N-BOP, Cross-tie PB 16A to PB 168 1

N/A SRO N1-0P-30 2

N/A R-ATC, Raise reactor power with recirc.

SRO N1-0P-1 3

C-ATC, Control Rod 26-35 Drifts Out RD02 SRO N1-S0P-5.2 C-Feedwater Booster Pump 11 Trips with Failure of standby FW02A

BOP, 4

SRO Feedwater Booster Pump to Auto-start Override N 1-SOP-16.1, Technical Specifications TS-SRO C-AII Respond to trip of Reactor Protection System (RPS) UPS 172 5

RP25 TS-SRO Technical Specification N 1-SOP-40.1 RWCU break in the Secondary Containment requiring scram; 6

CU11 M-AII RWCU Isolation Valves to isolate N1-EOP-2, N1-EOP-5, N1-EOP-8 C-Mode Switch Fails to Scram 7

Overrides

ATC, SRO N1-SOP-1 (N)ormal, (R)eactivitv, (l)nstrument, (C)omponent, (M)ajor

Facility: Nine Mile Point Unit 1 Scenario No.: NRC-2 Op-Test No.: LC117-1 NRC

1. Malfunctions after EOP entry (1-2) 1 Events 7
2. Abnormal events (2-4) 3 Events 3, 4, 5
3. Major transients (1-2) 1 Event 6
4. EOPs entered/requiring substantive actions (1-2) 2 N1-EOP-2, N1-EOP-5
5. EOP contingencies requiring substantive actions (0-2) 1 N1-EOP-8
6. Preidentified Critical tasks (2-3) 2 CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

CT-1.0: Given an un-isolable RWCU leak outside primary containment and With an un-isolable primary system one general area temperature above the maximum safe limit, the crew will discharging outside of Primary insert a manual reactor scram, in accordance with N1-EOP-5.

Containment resulting in general area temperature above the maximum safe limit, the Reactor must be scrammed.

This reduces the rate of energy production and thus the heat input, radioactivity release, and break flow into the Secondary Containment. This also ensures the Reactor is shutdown prior to need for a blowdown.

CT-2.0: Given an un-isolable RWCU leak outside primary containment and An un-isolable primary system two general area temperatures above the maximum safe limit, the crew will discharging outside of Primary execute N1-EOP-8, RPV Slowdown, in accordance with N1-EOP-5.

Containment resulting in two general area temperatures above the maximum safe limit indicates a wide-spread problem posing a direct and immediate threat to Secondary Containment. A blowdown minimizes flow through the break, rejects heat to the suppression pool in preference to outside the containment, and places the primary system in the lowest possible energy state.

SCENARIO

SUMMARY

The scenario begins at approximately 90% power. Containment Spray Pump 112 is out of service for maintenance. Steam Packing Exhauster 12 is out of service due to high vibrations.

The crew will start by cross-tying PB 16A to PB 168. Then, the crew will raise Reactor power to approximately 95% with Recirculation flow.

During the power ascension, a control rod will begin to drift out. The crew will select the drifting control rod and drive it full in. The crew will dispatch an operator to valve out the affected Hydraulic Control Unit to prevent the control rod from continuing to drift.

Then, Feedwater Booster pump 11 will trip. The standby Feedwater Booster pump will fail to auto-start. The crew will manually start the standby Feedwater Booster pump to restore normal system pressures. The SRO will determine the Tech Spec impact for loss of a redundant HPCI component.

RPS UPS 172 will develop an internal fault and drop out the #12 RPS system and RPS Bus 12.

The crew will respond to. the trip of UPS per N1-SOP-40.1. The SRO will direct the bus be repowered from l&C Bus 130A and will determine the most limiting Tech Spec condition. The BOP and the RO will reset Y2 scram and Y2 isolations and perform recovery actions after the bus is repowered. The SRO will determine Tech Spec 3.1.2, 3.6.11 and 3.4.4 are the limiting 7 day LCO's applicable with the RPS 12 Bus tripped.

A Reactor Water Cleanup system line break will occur in the Secondary Containment downstream of the Supply Isolation Valves. Reactor Water Cleanup will fail to isolate on high area temperature. The crew will attempt to isolate the system, but the valves will fail to fully close. This break will require a scram (Critical Task) and RPV blowdown (Critical Task) due to exceeding the Maximum Safe Value for general area temperatures. The Mode Switch will fail to scram the Reactor, however either RPS pushbuttons or manual ARI actuation will result in successful control rod insertion.

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point Unit 1 Scenario No.: NRC-3 Op-Test No.: LC1 17-1 NRC Examiners:

Operators:

Initial Conditions: The plant is operating at approximately 100% power. Containment Spray Pump 112 is out of service for maintenance. Steam Packing Exhauster 12 is out of service due to high vibrations.

Turnover: Reduce reactor power to 98% with recirc flow. Then, start TBCLC Pump 12 and secure TBCLC Pump 11.

Event Malf.

Event Event No.

No.

Type*

Description 1

N/A R-ATC, Lower reactor power to 98% with recirc flow SRO N1-0P-1 N-Swap Running TBCLC Pumps 2

N/A

BOP, SRO (2017 Scenario 4), N1-0P-24 C-BOP, Powerboard 101 fault 3

ED06 SRO N 1-SOP-1.3, Technical Specifications TS-SRO I-SRO Reactor Pressure Instrument 36-0?C Fails Low 4

RP17B TS-SRO Technical Specifications I-ATC, EPR Oscillation 5

TC06 SRO TS-SRO N1-SOP-31.1, Technical Specifications CW04A 6

CW04B C-AII All RBCLC Pumps Trip CW04C (2015 Scenario 5), N1-SOP-11.1, N1-SOP-1, N1-EOP-2 FW03A C-Motor Driven Feedwater Pumps Fail to Operate and 13 FW Pump 7

FW03B

BOP, clutch disengages FW06 SRO (2015 Scenario 5), N 1-EOP-2 CU01 Coolant Leak Inside Primary Containment 8

M-AII EC01 (2015 Scenario 5), N1-EOP-2, N1-EOP-4 9

VICP201 M-AII Fuel Zone Level Instrument Sporadic Indication 68/69 (2015 Scenario 5), N1-EOP-2, N1-EOP-7 (N)ormal, (R)eactivitv, (l)nstrument, (C)omponent, (M)aior

Facility: Nine Mile Point Unit 1 Scenario No.: NRC-3 Op-Test No.: LC117-1 NRC

2. Malfunctions after EOP entry (1-2) 3 Events 7, 8, 9
3. Abnormal events (2-4) 3 Events 3, 5, 6
4. Major transients (1-2) 2 Events 8 & 9
5. EOPs entered/requiring substantive actions (1-2) 2 N1-EOP-2, N1-EOP-4
6. EOP contingencies requiring substantive actions (0-2) 1 N1-EOP-7
7. Preidentified Critical tasks (2-3) 2 CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

CT-1.0: Given a LOCA in the Drywell with the inability to maintain Initiating Containment Sprays reduces containment parameters within the Pressure Suppression Pressure limit, Primary Containment pressure. This initiate Containment Sprays, in accordance with N1-EOP-4.

reduces stresses on the Drywell and Torus, assists in avoiding "chugging" that may cause fatigue failure of the LOCA downcomers, and avoids the need for a blowdown. These benefits reduce challenges to the fuel cladding, the RPV, and the Primary Containment.

CT-2.0: Given the plant with RPV water level unknown, execute N1-EOP-7, With Reactor water level unknown, the RPV Flooding, in accordance with N1-EOP-2.

status of core cooling is unknown. RPV flooding is required to establish conditions to cool the core. This protects the fuel cladding integrity.

SCENARIO

SUMMARY

The scenario begins at approximately 100% power. Containment Spray Pump 112 is out of service for maintenance. Steam Packing Exhauster 12 is out of service due to high vibrations.

The crew is directed to lower power to 98% with recirc flow, and then swap TBCLC pump 11 per N 1-0P-24 section F.1.

After shifting the TBCLC pumps, a fault will occur causing a loss of Powerboard 101. The crew will respond to a loss of ARP 13 per N 1-SOP-1.3. Additional lost loads include Condensate Pump 12, Feedwater Booster Pump 12 and the Electric Fire Pump.

After the loss of PB 101, Reactor pressure instrument 36-0?C fails low. The SRO will review Technical Specifications for the loss of automatic scram instrumentation.

Next, EPA oscillations begin. The crew will implement N1-SOP-31.2 place the MPR in service and secure the EPA. SRO will address Technical Specifications.

Next, the running RBCLC pumps will trip. The standby RBCLC pump will trip upon being started. The crew will enter N 1-SOP-11.1, RBCLC Failure. The crew will scram the Reactor, trip Recirculation pumps, initiate Emergency Condensers, and shut the MSIVs. The high pressure Feedwater pumps will fail to operate on the scram, complicating Reactor water level control.

Once the crew stabilizes the plant after the scram, a coolant leak will develop inside the Primary Containment. The crew will re-enter N1-EOP-2, RPV Control, and N1-EOP-4, Primary Containment Control. Containment parameters will degrade and the crew will initiate Containment Sprays (Critical Task). The elevated Containment temperature will cause the Fuel Zone level indications to become erratic. With all other Reactor water level indicators downscale, the crew will execute N1-EOP-7, RPV Flooding, to lower Reactor pressure and flood the Reactor to the Main Steam lines (Critical Task).

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point Unit 1 Scenario No.: NRC-4 Op-Test No.: LC1 17-1 NRC Examiners:

Operators:

Initial Conditions: A plant startup is in progress with reactor power approximately 2-3%. Containment Spray Pump 112 is out of service for maintenance. Steam Packing Exhauster 12 is out of service due to high vibrations.

Turnover: Continue power ascension by withdrawing control rods.

Event Malt.

Event Event No.

No.

Type*

Description N/A R-ATC, Raise power with control rods 1

SRO N1-0P-43A, N1-0P-5 RD42 C-ATC, Control Rod Double Notches 2

SRO N1-0P-5 RR06A I-ATC, IRM Downscale Failure 3

SRO N 1-SOP-1.2, RR07A C-BOP, Powerboard 16A Electrical Fault 4

ED12A SRO (2015 Scenario 5), ARP L4-3-6, N1-EOP-4 RR06A C-BOP, Recirc Pump 11 seal failure requiring isolation of the pump 5

SRO N1-SOP-1.2, Technical Specification 3.2.5, 3.1.7.e RR07A TS-SRO PC05 C-BOP, Seismic Event; lsolable Leak on Containment Spray Suction Line 6

SRO CT04A TS-SRO N1-SOP-28, N1-EOP-5, Technical Specifications Second Seismic Event; Torus Break; Multiple Control Rods Fail to PC05 Insert 7

PC04 M-AII N1-EOP-5, N1-EOP-4, N1-SOP-1, N1-EOP-2, N1-EOP-8, N1-EOP-3 CT02B Containment Spray Raw Water Pumps 112 and 121 Trips 8

C-AII CT02C N1-EOP-4 (N)ormal, (R)eactivitv, (l)nstrument, (C)omponent, (M)aior

Facility: Nine Mile Point Unit 1 Scenario No.: NRC-4 Op-Test No.: LC117-1 NRC

1. Malfunctions after EOP entry (1-2) 1 Events 8
2. Abnormal events (2-4) 4 Events 3, 4, 5, 6
3. Major transients (1-2) 1 Event7
4. EOPs entered/requiring substantive actions (1-2) 3 N1-EOP-2, N1-EOP-4, N1-EOP-5
5. EOP contingencies requiring substantive actions (0-2) 2 N1-EOP-3, N1-EOP-8
6. Preidentified Critical tasks (2-3) 2 CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

CT-1.0: Given an un-isolable Torus leak exceeding makeup capacity, scram Lowering Torus water level challenges the Reactor, in accordance with N1-EOP-4.

the pressure suppression function of the Primary Containment. Continued Reactor operation is not allowed with an inoperable Primary Containment. A Reactor scram also allows subsequent mitigating actions, such as Reactor cooldown and/or blowdown.

CT-2.0: Given an un-isolable Torus leak exceeding makeup capacity, If torus water level lowers below the perform an RPV Slowdown, in accordance with N1-EOP-4.

elevation of the ERV discharge holes, opening ERVs would discharge steam directly into the torus airspace. The resulting pressure increase could exceed the maximum pressure capability of the Primary Containment. Since the RPV may not be kept at pressure under these conditions, a blowdown is required.

SCENARIO

SUMMARY

The scenario begins with the plant in a startup at approximately 2-3% power and raising power by control rod withdrawal. While withdrawing control rods, one control rod will double notch to one position past its intended position. The crew will respond per N1-0P-5 section H.9.0 and re-insert the control rod to the intended position.

Next, IRM 11 will fail downscale. The SRO will determine the impact of the failure on Tech Specs and the crew will bypass the IRM.

Next, Powerboard 16A will develop an electrical fault. This will cause a loss of power to three Drywell cooling fans. The crew will start an additional Drywell cooling fan to stabilize Drywell temperature and pressure. The electrical loss will also affect EOG 103 auxiliary equipment.

The US will evaluate for Tech Spec impacts.

Then, the inner seal will fail on Reactor Recirculation Pump 11. A few minutes later, the outer seal will fail, affecting drywell leakage. The crew will remove the pump from service and isolate it. The SRO will review Technical Specifications for drywell leakage and partial loop operation.

Then, a seismic event occurs and results in an isolable leak on Containment Spray pump 121 suction line. The crew will execute N 1-SOP-28, Seismic Event, and N 1-EOP-5, Secondary Containment Control. The crew will close Containment Spray suction valve 121 to isolate the leak. The SRO will address Tech Specs.

Next, a second seismic event occurs and results in an un-isolable Torus break. The crew will re-enter N 1-EOP-5 and enter N 1-EOP-4, Primary Containment Control. The crew will attempt to add water to the Torus, however makeup efforts will be complicated by trip of Containment Spray Raw Water pumps. The crew will be unable to raise Torus water level. The crew will insert a manual Reactor scram (Critical Task). Multiple control rods will fail to fully insert. The crew will enter N1-EOP-2, RPV Control, and transition to N1-EOP-3, Failure to Scram. The crew will be unable to drive control rods in with RMCS. As Torus water level lower further, the crew will perform an RPV Slowdown per N1-EOP-8 (Critical Task). During the RPV Slowdown, the crew will terminate and prevent all injection except boron and CRD and later re-inject to restore/maintain Reactor water level above the top of active fuel.

ES-401 1

Form ES-401-1 Facility:

Nine Mile Point Unit 1 Date of Exam:

December 2018 Tier Group RO KIA Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

Total A2 G*

Total

1.

1 3

4 2

4 3

4 20 3

4 7

Emergency and 2

1 1

1 N/A 2

1 N/A 1

7 2

1 3

Abnormal Plant Evolutions Tier Totals 4

5 3

6 4

5 27 5

5 10

2.

1 3

2 3

3 2

2 2

3 2

2 2

26 3

2 5

Plant 2

1 1

1 1

1 2

0 1

2 1

1 12 0

1 2

3 Systems Tier Totals 4

3 4

4 3

4 2

4 4

3 3

38 4

4 8

3. Generic Knowledge and Abilities 1

2 3

4 10 1

2 3

4 7

Categories 3

2 2

3 2

1 2

2 Note: 1.

Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the "Tier Totals" in each KIA category shall not be less than two). (One Tier 3 radiation control KIA is allowed if it is replaced by a KIA from another Tier 3 category.)

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 +/-1 from that specified in the table based on NRG revisions.

The final RO exam must total 75 points, and the SRO-only exam must total 25 points.

3. Systems/evolutions within each group are identified on the outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate KIA statements.
4. Select topics from as many systems and evolutions as possible. Sample every system or evolution in the group before selecting a second topic for any system or evolution.
5.

Absent a plant-specific priority, only those Kl As having an importance rating {IR) of 2.5 or higher. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and KIA categories.
7. The generic (G) Kl As in Tiers 1 and 2 shall be selected from Section 2 of the KIA catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable Kl As.
8. On the following pages, enter the KIA numbers, a brief description of each topic, the topics' I Rs 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 a category 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.
9.

For Tier 3, select topics from Section 2 of the KIA catalog, and enter the KIA numbers, descriptions, I Rs, and point totals (#) on Form ES-401-3. Limit SRO selections to Kl As that are linked to 10 CFR 55.43.

G* Generic KIAs These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the KIA catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the KIA catalog.

These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the KIA catalog is used to develop the sample plan.

ES-401 2

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

E/APE # / Name/ Safety Function K1 K2 K3 A1 A2 G*

KIA Topic(s)

IR Q#

295001 (APE 1) Partial or Complete Loss of X

AA2.02, Ability to determine and/or interpret 3.1 27 Forced Core Flow Circulation / 1 & 4 the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Neutron monitorino 295003 (APE 3) Partial or Complete Loss of X

AA 1.01, Ability to operate and/or monitor the 3.7 29 AC Power/ 6 following as they apply to PARTIAL OR COMPLETE LOSS OF A.G. POWER: A.G.

electrical distribution system 295004 (APE 4) Partial or Complete Loss of X

AK1.04, Knowledge of the operational 2.8 30 DC Power/ 6 implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Effect of battery discharoe rate on capacity 295005 (APE 5) Main Turbine Generator Trip/

X G.2.4.31, Knowledge of annunciator alarms, 4.2 31 3

indications, or response procedures.

295006 (APE 6) Scram I 1 X

AK2.06, Knowledge of the interrelations 4.2 32 between SCRAM and the following: Reactor power X G2.1.19, Ability to use plant computers to 3.8 76 evaluate system or component status.

295016 (APE 16) Control Room Abandonment X G2.4.35, Knowledge of local auxiliary 3.8 33 17 operator tasks during an emergency and the resultant operational effects.

295018 (APE 18) Partial or Complete Loss of X

AK1.01, Knowledge of the operational 3.5 34 ccw /8 implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Effects on component/system operations 295019 (APE 19) Partial or Complete Loss of X

AK3.03, Knowledge of the reasons for the 3.2 35 Instrument Air/ 8 following responses as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Service air isolations:

Plant-Specific 295021 (APE 21) Loss of Shutdown Cooling/

X G2.2.37, Ability to determine operability 3.6 36 4

and/or availability of safety related equipment.

X AA2.06, Ability to determine and/or 3.3 77 interpret the following as they apply to LOSS OF SHUTDOWN COOLING: Reactor pressure 295023 (APE 23) Refueling Accidents/ 8 X

AK1.01, Knowledge of the operational 3.6 37 implications of the following concepts as they apply to REFUELING ACCIDENTS:

Radiation exposure hazards X G2.2.37, Ability to determine operability 4.6 78 and/or availability of safety related equipment.

295024 High Drywell Pressure / 5 X

EK2.15, Knowledge of the interrelations 3.8 38 between HIGH DRYWELL PRESSURE and the following: Containment spray logic:

Plant-Specific

ES-401 3

Form ES-401-1 295025 (EPE 2) High Reactor Pressure/ 3 X G2.4.8, Knowledge of how abnormal 3.8 39 operating procedures are used in conjunction with EOPs.

295026 (EPE 3) Suppression Pool High Water X

EA 1.03, Ability to operate and/or monitor the 3.9 28 Temperature I 5 following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:

Temperature monitoring X

EA2.01, Ability to determine and/or 4.2 79 interpret the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Suppression pool water temperature 295028 (EPE 5) High Drywell Temperature X

EK3.06, Knowledge of the reasons for the 3.4 40 (Mark I and Mark II only)/ 5 following responses as they apply to HIGH DRYWELL TEMPERATURE: ADS EA2.01, Ability to determine and/or.

X interpret the following as they apply to 4.1 80 HIGH DRYWELL TEMPERATURE:

Drywell temperature 295030 (EPE 7) Low Suppression Pool Water X

EA2.04, Ability to determine and/or interpret 3.5 41 Level/ 5 the following as they apply to LOW SUPPRESSION POOL WATER LEVEL:

Drywell/ suppression chamber differential pressure: Mark I & II 295031 (EPE 8) Reactor Low Water Level/ 2 X

EA 1.13, Ability to operate and/or monitor the 4.3 42 following as they apply to REACTOR LOW WATER LEVEL: Reactor water level control 295037 (EPE 14) Scram Condition Present X

EA 1.01, Ability to operate and/or monitor the 4.6 43 and Reactor Power Above APRM Downscale following as they apply to SCRAM or Unknown / 1 CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Reactor Protection System 295038 (EPE 15) High Offsite Radioactivity X

EK2.08, Knowledge of the interrelations 2.6 44 Release Rate I 9 between HIGH OFF-SITE RELEASE RATE and the following: SPDS/ERIS/CRIDS/GDS:

Plant-Specific.

X G2.4.41, Knowledge of the emergency 4.6 81 action level thresholds and classifications.

600000 (APE 24) Plant Fire On Site/ 8 X

AA2.05, Ability to determine and interpret 2.9 45 the following as they apply to PLANT FIRE ON SITE: Ventilation alignment necessary to secure affected area X

2.1.7, Ability to evaluate plant 4.7 82 performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

700000 (APE 25) Generator Voltage and X

AK2.07, Knowledge of the interrelations 3.6 46 Electric Grid Disturbances / 6 between GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES and the following: Turbine/generator control KIA Cateqorv Totals:

3 4

2 4

3/3 4/4 RO/SRO Group Point Total:

20/7

ES-401 4

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

E/APE #/Name/Safety Function K1 K2 K3 A1 A2 G*

KIA Topic(s)

IR Q#

295002 (APE 2) Loss of Main Condenser X

AK1.03, Knowledge of the operational 3.6 47 Vacuum/ 3 implications of the following concepts as they apply to LOSS OF MAIN CONDENSER VACUUM: Loss of heat sink 295007 (APE 7) High Reactor Pressure/ 2 X

G2.2.42, Ability to recognize system 4.6 83 parameters that are entry-level conditions for Technical Specifications.

295009 (APE 9) Low Reactor Water Level / 2 X

AA 1.04, Ability to operate and/or monitor 2.7 48 the following as they apply to LOW REACTOR WATER LEVEL: Reactor water cleanup 295012 (APE 12) High Drywall Temperature I X

AK3.01, Knowledge of the reasons for the 3.5 49 5

following responses as they apply to HIGH DRYWELL TEMPERATURE: Increased drywall cooling 295013 (APE 13) High Suppression Pool X G2.4.20, Knowledge of the operational 3.8 50 Temperature. I 5 implications of EOP warnings, cautions, and notes.

295015 (APE 15) Incomplete Scram I 1 X

AA 1.05, Ability to operate and/or monitor 2.5 51 the following as they apply to INCOMPLETE SCRAM: Rod worth minimizer: Plant-Specific 295022 (APE 22) Loss of Control Rod Drive X

AK2.07, Knowledge of the interrelations 3.4 52 Pumps/ 1 between LOSS OF CRD PUMPS and the following: Reactor pressure (SCRAM assist): Plant-Specific 295029 (EPE 6) High Suppression Pool Water X

EA2.03, Ability to determine and/or 3.4 53 Level/ 5 interpret the following as they apply to HIGH SUPPRESSION POOL WATER LEVEL: Drvwell/containment water level 295033 (EPE 1 O) High Secondary X

EA2.03, Ability to determine and/or 4.2 84 Containment Area Radiation Levels / 9 interpret the following as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS: Cause of high area radiation 295036 (EPE 13) Secondary Containment X

EA2.03, Ability to determine and/or 3.8 85 High Sump/Area Water Level/ 5 interpret the following as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL: Cause of the high water level

~

Point Totals:

1 1

1 2

1/2 1/1 RO/SRO Group Point Total:

7/3

ES-401 5

Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Svstems-Tier 2/Grouo 1 (RO/SRO)

System # I Name K

K K

K4 K

K A A A A G KIA Topic(s)

IR Q#

1 2

3 5

6 1

2 3

4 205000 (SF4 SGS) Shutdown Cooling X

K3.02, Knowledge of the effect that a loss or 3.2 6

malfunction of the SHUTDOWN COOLING SYSTEM (AHR SHUTDOWN COOLING MODE) will have on following: Reactor water level: Plant-Specific X

A2.02, Ability to (a) predict the impacts of 2.7 86 the following on the SHUTDOWN COOLING SYSTEM (AHR SHUTDOWN COOLING MODE); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Low shutdown cooling suction pressure:

Plant-Specific 206000 (SF2, SF4 HPCIS) -

X K2.01, Knowledge of electrical power 3.2 3

High-Pressure Coolant Injection supplies to the following: System valves:

BWR-2,3,4 X

K4.07, Knowledge of HIGH PRESSURE 4.3 24 COOLANT INJECTION SYSTEM design feature(s) and/or interlocks which provide for the following: Automatic system initiation:

BWR-2,3,4 207000 (SF4 IC) Isolation X

K6.04, Knowledge of the effect that a loss or 3.2 11 (Emergency) Condenser malfunction of the following will have on the ISOLATION (EMERGENCY) CONDENSER:

Plant air systems: BWR-2,3 209001 (SF2, SF4 LPCS)

X K1.09, Knowledge of the physical 3.2 1

Low-Pressure Core Spray connections and/or cause-effect relationships between LOW PRESSURE CORE SPRAY SYSTEM and the following: Nuclear boiler instrumentation X

A2.01, Ability to (a) predict the impacts of 3.4 87 the following on the LOW PRESSURE CORE SPRAY SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Pump trips 211000 (SF1 SLCS) Standby Liquid X

A 1.03, Ability to predict and/or monitor 3.6 13 Control changes in parameters associated with operating the STANDBY LIQUID CONTROL SYSTEM controls including: Pump discharge pressure

ES-401 6

Form ES-401-1 212000 (SF? RPS) Reactor X

K5.02, Knowledge of the operational 3.3 9

Protection implications of the following concepts as they apply to REACTOR PROTECTION SYSTEM:

Specific logic arrangements X 2.4.46, Ability to verify that the alarms are 4.2 88 consistent with the plant conditions.

215003 (SF7 IRM)

X K5.01, Knowledge of the operational 2.6 10 Intermediate-Range Monitor implications of the following concepts as they apply to INTERMEDIATE RANGE MONITOR (IRM) SYSTEM: Detector operation 215004 (SF? SAMS) Source-Range X

K4.02, Knowledge of SOURCE RANGE 3.4 7

Monitor MONITOR (SAM) SYSTEM design feature(s) and/or interlocks which provide for the following: Reactor SCRAM signals 215005 (SF? PRMS) Average Power X

A 1.07, Ability to predict and/or monitor 3.0 14 Range Monitor/Local Power Range changes in parameters associated with Monitor operating the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM controls including:

APRM (gain adjustment factor)

X G2.1.32, Conduct of Operations: Ability to 3.8 22 explain and apply all system limits and precautions.

218000 (SF3 ADS) Automatic X

A2.01, Ability to (a) predict the impacts of the 4.1 16 Depressurization following on the AUTOMATIC DEPRESSURIZATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Small steam line break LOCA 223002 (SF5 PCIS) Primary X

A3.01, Ability to monitor automatic operations 3.4 17 Containment Isolation/Nuclear Steam of the PRIMARY CONTAINMENT Supply Shutoff ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF including: System indicating lights and alarms X

A2.04, Ability to (a) predict the impacts of 3.2 89 the following on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Process radiation monitoring system failures

ES-401 7

Form ES-401-1 239002 (SF3 SRV) Safety Relief X

A2.01, Ability to (a) predict the impacts of the 3.0 15 Valves following on the RELIEF/SAFETY VALVES; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Stuck open vacuum breakers X

K1.07, Knowledge of the physical 3.6 23 connections and/or cause-effect relationships between RELIEF/SAFETY VALVES and the following: Suppression Pool 259002 (SF2 RWLCS) Reactor Water X

A4.01, Ability to manually operate and/or 3.8 20 Level Control monitor in the control room: All individual component controllers in the manual mode X G2.4.9, Knowledge of low power/

4.2 90 shutdown implications in accident (e.g.,

loss of coolant accident or loss of residual heat removal) mitigation strategies.

261000 (SF9 SGTS) Standby Gas X

K6.03, Knowledge of the effect that a loss or 3.0 12 Treatment malfunction of the following will have on the STANDBY GAS TREATMENT SYSTEM:

Emergency diesel generator system X

A2.07, Ability to (a) predict the impacts of the 2.7 25 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:

A.G. electrical failure 262001 (SF6 AC) AC Electrical X

K4.03, Knowledge of A.G. ELECTRICAL 3.1 8

Distribution DISTRIBUTION design feature(s) and/or interlocks which provide for the following:

Interlocks between automatic bus transfer and breakers 262002 (SF6 UPS) Uninterruptable X

K3.10 - Knowledge of the effect that a loss or 2.7 5

Power Supply (AC/DC) malfunction of the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) will have on following: Containment isolation: Plant-Specific 263000 (SF6 DC) DC Electrical X

A3.01, Ability to monitor automatic operations 3.2 18 Distribution of the D.C. ELECTRICAL DISTRIBUTION including: Meters, dials, recorders, alarms, and indicating lights X

K3.03, Knowledge of the effect that a loss or 3.4 26 malfunction of the D.C. ELECTRICAL DISTRIBUTION will have on following:

Systems with D.C. components (i.e. valves, motors, solenoids, etc.)

ES-401 8

Form ES-401-1 264000 (SF6 EGE) Emergency X

A4.04, Ability to manually operate and/or 3.7 19 Generators (Diesel/Jet) EOG monitor in the control room: Manual start, loading, and stopping of emergency generator: Plant-Specific 300000 (SFB IA) Instrument Air X

K2.02, Knowledge of electrical power 3.0 4

supplies to the following: Emergency air compressor X 2.1.30, Conduct of Operations: Ability to 4.4 21 locate and operate components, including local controls.

400000 (SFB CCS) Component X

K1.02, Knowledge of the physical 3.2 2

Cooling Water connections and / or cause-effect relationships between CCWS and the following: Loads cooled by CCWS KIA Category Point Totals:

3 2

3 3

2 2

2 3/ 2 2 2/ RO/SRO Group Point Total:

RI 3

2

ES-401 9

Form ES-401-1 1ES-401 BWR Examination Outline Form ES-401-1 Plant Svstems-Tier 2/Grouo 2 (RO/SRO)

System # I Name K

K K K

K K A A A A G*

KIA Topic(s)

IR Q#

1 2

3 4

5 6

1 2

3 4

201002 (SF1 RMCS) Reactor Manual Control X

A2.01, Ability to (a) predict the 2.8 91 impacts of the following on the REACTOR MANUAL CONTROL SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Rod movement sequence timer malfunctions 201003 (SF1 CRDM) Control Rod and Drive X G2.2.38, Knowledge of conditions 4.5 92 Mechanism and limitations in the facility license.

201006 (SF7 RWMS) Rod Worth Minimizer X

A2.01, Ability to (a) predict the 2.5 54 impacts of the following on the ROD WORTH MINIMIZER SYSTEM (RWM) (PLANT SPECIFIC); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Power supply loss: P-Spec(Not-BWR6) 202001 (SF1, SF4 RS) Recirculation X G2.2.36, Ability to analyze the 4.2 93 effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

202002 (SF1 RSCTL) Recirculation Flow X

A3.03, Ability to monitor automatic 3.1 55 Control operations of the RECIRCULATION FLOW CONTROL SYSTEM including: Scoop tube operation:

BWR-2,3,4 215001 (SF7 TIP) Traversing In-Core Probe X

K4.01, Knowledge of TRAVERSING 3.4 56 IN-CORE PROBE design feature(s) and/or interlocks which provide for the following: Primary containment isolation: Mark I & II (Not-BWR1) 223001 (SF5 PCS) Primary Containment and X G2.4.3, Ability to identify post-3.7 57 Auxiliaries accident instrumentation.

226001 (SF5 AHR CSS) RHR/LPCI:

X K6.11, Knowledge of the effect that 2.8 58 Containment Spray Mode a loss or malfunction of the following will have on the RHR/LPCI:

CONTAINMENT SPRAY SYSTEM MODE: Component cooling water svstems 245000 (SF4 MTGEN) Main Turbine X

A4.09, Ability to manually operate 2.6 59 Generator/ Auxiliary and/or monitor in the control room:

Hvdroaen seal oil pressure

ES-401 10 Form ES-401-1 259001 (SF2 FWS) Feedwater X

K6.06, Knowledge of the effect that 2.7 60 a loss or malfunction of the following will have on the REACTOR FEEDWATER SYSTEM: Plant service water 268000 (SF9 RW) Radwaste X

K1.05, Knowledge of the physical 2.9 61 connections and/or cause-effect relationships between RADWASTE and the following: Drywell equipment drains 272000 (SF?, SF9 RMS) Radiation Monitoring X

K2.05, Knowledge of electrical 2.6 62 power supplies to the following:

Reactor building ventilation monitors: Plant-Specific 288000 (SF9 PVS) Plant Ventilation X

K5.02, Knowledge of the operational 3.2 63 implications of the following concepts as they apply to PLANT VENTILATION SYSTEMS:

Differential pressure control 290001 (SF5 SC) Secondary Containment X

A3.01, Ability to monitor automatic 3.9 64 operations of the SECONDARY CONTAINMENT including:

Secondary containment isolation 290002 (SF4 RVI) Reactor Vessel Internals X

K3.01, Knowledge of the effect that 3.2 65 a loss or malfunction of the REACTOR VESSEL INTERNALS will have on following: Reactor water level KIA Category Point Totals:

1 1

1 1

1 2

0 1/ 2 1

1/ RO/SRO Group Point Total:

12/3 1

2

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

Form ES-401-3 Facility: Nine Mile Point Unit 1 Date of Exam: January 2019 Category KIA#

Topic RO SRO-only IR Q#

IR Q#

2.1.4 Knowledge of individual licensed operator 3.3 66 responsibilities related.to shift staffing, such as medical requirements, "no-solo" operation, maintenance of active license status, 1 OCFR55, etc.

2.1.37 Knowledge of procedures, guidelines, or limitations 4.3 67 associated with reactivity management.

2.1.31 Ability to locate control room switches, controls, 4.3 94

1. Conduct of and indications, and to determine that they Operations correctly reflect the desired plant lineup.

2.1.40 Knowledge of refueling administrative 3.9 95 requirements.

2.1.43 Ability to use procedures to determine the effects on 4.1 68 reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc.

Subtotal 3

2 2.2.2 Ability to manipulate the console controls as required 4.6 69 to operate the facility between shutdown and desianated power levels.

2. Equipment 2.2.21 Knowledge of pre-and post-maintenance 4.1 96 Control operability requirements.

2.2.22 Knowledge of limiting conditions for operations and 4.0 70 safety limits.

Subtotal 2

1 2.3.5 Ability to use radiation monitoring systems, such as 2.9 71 fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

2.3.7 Ability to comply with radiation work permit 3.6 97 requirements during normal or abnormal conditions.

3. Radiation 2.3.13 Knowledge of radiological safety procedures pertaining 3.4 72 Control to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

2.3.14 Knowledge of radiation or contamination hazards 3.8 98 that may arise during normal, abnormal, or emergency conditions or activities.

Subtotal 2

2

4. Emergency 2.4.14 Knowledge of general guidelines for EOP usaae.

3.8 73 Procedures/Plan 2.4.17 Knowledge of EOP terms and definitions.

4.3 99

ES-401 Generic Knowledge and Abilities Outline {Tier 3)

Form ES-401-3 2.4.21 Knowledge of the parameters and logic used to 4.6 100 assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

2.4.32 Knowledge of operator response to loss of all 3.6 74 annunciators.

2.4.22 Knowledge of the bases for prioritizing safety functions 3.6 75 during abnormal/emergency operations.

Subtotal 3

2 Tier 3 Point Total 10 7

ES-401 Record of Rejected K/A's Form ES-401-4 Tier/ Group Randomly Selected KIA Reason for Rejection The systematic and random sampling process utilized the pre-approved Nine Mile Point Unit 1 Kl A suppression list.

The following Kl As were rejected following the systematic and random sampling process:

Question 9 A discriminating question could not be developed without testing generic fundamentals 212000 RPS knowledge.

2/1 K5.01 - Knowledge of the Randomly reselected KIA 212000 RPS K5.02 -

operational implications of Knowledge of the operational implications of the the following concepts as following concepts as they apply to REACTOR they apply to REACTOR PROTECTION SYSTEM: Specific logic PROTECTION SYSTEM:

arrangements.

Fuel thermal time constant Question 14 This facility does not have recirculation flow control valves.

215005 APRM / LPRM Randomly reselected KIA 215005 A1.07 - Ability A 1.06 - Ability to predict to predict and/or monitor changes in parameters and/or monitor changes in associated with operating the AVERAGE parameters associated with POWER RANGE MONITOR/LOCAL POWER 2/1 operating the AVERAGE RANGE MONITOR SYSTEM controls including:

POWER RANGE APRM (gain adjustment factor).

MONITOR/LOCAL POWER RANGE MONITOR SYSTEM controls including:

Recirculation flow control valve position: Plant-Specific Question 23 An acceptable question could not be written for the randomly selected KIA due to limited 239002 SRVs interrelations between SRVs and plant air systems.

K1.05 - Knowledge of the physical connections Randomly reselected KIA 239002 K1.07 -

and/or cause-effect Knowledge of the physical connections and/or relationships between cause-effect relationships between 2/1 RELIEF/SAFETY VALVES RELIEF/SAFETY VALVES and the following:

and the following: Plant air Suppression Pool.

systems: Plant-Specific

ES-401 Record of Rejected K/A's Form ES-401-4 Question 25 There are no interlocks or initiations for RBEVS (SGTS) related to high system pressure at this 261000 SGTS facility. An acceptable question could not be developed without testing minutia.

A2.14 - Ability to (a) predict the impacts of the following Randomly reselected KIA 261000 A2.07 - Ability on the STANDBY GAS to (a) predict the impacts of the following on the 2 I 1 TREATMENT SYSTEM; STANDBY GAS TREATMENT SYSTEM; and (b) and (b) based on those based on those predictions, use procedures to predictions, use procedures correct, control, or mitigate the consequences of to correct, control, or those abnormal conditions or operations: A.C.

mitigate the consequences electrical failure.

of those abnormal conditions or operations:

High system pressure:

Plant-Specific Question 28 295001 was inadvertently sampled twice on the RO exam prior to sampling 295026.

295001 Partial or Complete Loss of Forced Core Flow Reselected 295026 Suppression Pool High Circulation Water Temperature and randomly reselected EA 1.03 - Ability to operate and/or monitor the 1 / 1 AA 1.03 - Ability to operate following as they apply to SUPPRESSION POOL and/or monitor the following HIGH WATER TEMPERATURE: Temperature as they apply to PARTIAL monitoring.

OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: RMCS:

Plant-Specific Question 36 An acceptable question could not be developed for the randomly sampled KIA due to lack of 295021 Loss of Shutdown Technical Specification bases related to loss of Cooling Shutdown Cooling that are RO level.

1 / 1 2.2.25 - Knowledge of the Randomly reselected Kl A 295021 Loss of bases in Technical Shutdown Cooling 2.2.37 - Ability to determine Specifications for limiting operability and/or availability of safety related conditions for operations equipment.

and safety limits.

Question 40 An acceptable question could not be developed for the randomly sampled KIA without 295028 High Drywell overlapping Question 49.

Temperature Randomly reselected KIA 295028 High Drywell 1 / 1 EK3.04 - Knowledge of the Temperature EK3.06 - Knowledge of the reasons reasons for the following for the following responses as they apply to responses as they apply to HIGH DRYWELL TEMPERATURE: ADS.

HIGH DRYWELL TEMPERATURE:

Increased drywell cooling

ES-401 Record of Rejected K/A's Form ES-401-4 Question 63 An acceptable question could not be developed for the randomly sampled KIA without testing 288000 Plant Ventilation minutia due to a lack of operationally relevant references related to Plant Ventilation K5.03 - Knowledge of the temperature control.

2/2 operational implications of the following concepts as Randomly reselected KIA 288000 Plant they apply to PLANT Ventilation K5.02 - Knowledge of the operational VENTILATION SYSTEMS:

implications of the following concepts as they Temperature control apply to PLANT VENTILATION SYSTEMS:

Differential pressure control.

Question 64 An acceptable question could not be developed for the randomly sampled KIA due to lack of 290001 Secondary Secondary Containment controls and associated Containment system lineups.

2/2 A 1.01 - Ability to predict Randomly reselected KIA 290001 Secondary and/or monitor changes in Containment A3.01 - Ability to monitor automatic parameters associated with operations of the SECONDARY CONTAINMENT operating the including: Secondary containment isolation.

SECONDARY CONTAINMENT controls including: System lineups Question 67 An acceptable question could not be developed for the randomly sampled KIA without 2.1.19 - Ability to use plant oversampling plant computer topics (see computers to evaluate Questions 44 & 76). Use of plant computers is 3

system or component also tested extensively on the operating exam.

status.

Randomly reselected KIA 2.1.37 - Knowledge of procedures, guidelines, or limitations associated with reactivity management.

Question 75 The randomly sampled generic KIA is also tested on Question 33. Reselecting for better balance 2.4.35 - Knowledge of local of coverage.

3 auxiliary operator tasks during an emergency and Randomly reselected KIA 2.4.22 - Knowledge of the resultant operational the bases for prioritizing safety functions during effects.

abnormal/emergency operations.

Question 88 An acceptable question could not be developed for the randomly sampled KIA due to lack of RO 212000 Reactor Protection tasks performed outside the control room related to the Reactor Protection System.

2 I 1 2.4.34 - Knowledge of RO tasks performed outside Randomly reselected KIA 212000 Reactor the main control room Protection 2.4.46 - Ability to verify that the alarms during an emergency and are consistent with the plant conditions.

the resultant operational effects.

ES-401 Record of Rejected K/A's Form ES-401-4 Question 91 An acceptable question could not be developed for the randomly sampled KIA without 201002 Reactor Manual overlapping the operating exam. Additionally, the Control KIA did not readily support testing at the SRO level.

A2.02 - Ability to (a) predict the impacts of the following Randomly reselected KIA 201002 Reactor on the REACTOR Manual Control A2.01 - Ability to (a) predict the 2/2 MANUAL CONTROL impacts of the following on the REACTOR SYSTEM; and (b) based on MANUAL CONTROL SYSTEM; and (b) based on those predictions, use those predictions, use procedures to correct, procedures to correct, control, or mitigate the consequences of those control, or mitigate the abnormal conditions or operations: Rod consequences of those movement sequence timer malfunctions.

abnormal conditions or operations: Rod drift alarm Question 93 An acceptable question could not be developed for the randomly sampled KIA due to lack of 202001 Recirculation surveillance procedures for the Recirculation system.

2/2 2.2.12 - Knowledge of surveillance procedures.

Randomly reselected KIA 202001 Recirculation 2.2.36 - Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

Question 83 Editorial error - APE 295008 does not coincide with High Reactor Pressure.

1 / 2 295008 High Reactor Pressure Conferred with Chief Examiner to change 295008 to 295007 to coincide with High Reactor Pressure

Retrieved from "https://kanterella.com/w/index.php?title=ML19122A209&oldid=4929981"