ML20055E015
| ML20055E015 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 10/28/2019 |
| From: | Todd Fish Operations Branch I |
| To: | Exelon Generation Co |
| Shared Package | |
| ML19151A455 | List: |
| References | |
| CAC 000500 | |
| Download: ML20055E015 (28) | |
Text
Appendix D Scenario Outline Form ES-D-1 0
Facility: Nine Mile Point Unit 2 Scenario No.: NRC-1 Op-Test No.: LC2 18-1 Examiners: Operators:
Initial Conditions: The plant is operating at 95% power with service water pump SWP*P1 E out of.
service for pump bearing replacement.
Turnover: The crew will perform N2-0SP-CSL-Q@002, LPCS Pump and Valve Operability and System Integrity Test (section 6.2 only), then lower reactor power to 90% using reactor recirculation flow.
Critical Tasks: See page 2 Event Malf. Event Event No. No. Type* Description N/A N-BOP, Perform N2-0SP-CSL-Q@002, LPCS Pump and Valve Operability SRO, and System Integrity Test (section 6.2 only) with failure of CSL 1 TS-SRO suction valve to re-open.
N2-0SP-CSL-Q@002, T.S. 3.5.1 N/A R-ATC Lower reactor power to 90% using reactor recirculation flow.
0 2 SRO N2-0P-101D CU08 I-BOP, RWCU fails to automatically isolate on RWCU flow mismatch SRO, caused by cleanup RWCU non-regen heat exchanger tube leak.
3 TS-SRO ARPs, T.S. 3.3.6.1 PC28A C-BOP, Loss of Drywall Cooling.
4 SRO TS-SRO N2-S0P-60, T.S. 3.3.6.1 FW13 I-ATC, Feedwater Master Controller Failure - High.
5 SRO N2-S0P-06 TU02 C-ATC Rising Main Turbine vibrations require scram.
6 SRO ARP's, N2-S0P-21, N2-S0P-101C MS04 M-AII Main Steam Line Break in Primary Containment with loss of condensate and feedwater system.
7 N2-EOP-RPV RR27 I-All All RPV level Instruments fail upscale.
8 N2-EOP-C4 I RH22A C-AII RHR 'A' injection valve (2RHS*MOV24A) loss of power.
9 N2.-EOP-C4 RH02B C-AII 2RHS*MOV24B will fail to open, requiring manual line up and inject 0 10 with 2RHS*MOV40B.
N2-EOP-C4
- (N}ormal (R}eactivity, (l}nstrument, (C}omponent (M}ajor
0 Facility: Nine Mile Point Unit 2
- 1. Malfunctions after EOP entry (1-2)
Event 8, 9, 1O Scenario No.: NRC-1 3
Op-Test No.: LC2 18-1
- 2. Abnormal events (2-4) 4 Events 3, 4, 5, 6
- 3. Major transients (1-2)
Event7
- 4. EOPs entered/requiring substantive actions (1-2) 2 N2-EOP-RPV, N2-EOP-PC
- 5. Entry into a contingency EOP with substantive actions ~1 per scenario set)
N2-EOP-C4
- 6. Pre-identified Critical Tasks (> 2) 2
- ,, ; . :cRITIC::ALTASk~USTIFICATION:'
CT-1.0, Given the plant with RPV water level unknown, open 7 SRVs to Critical Task 1.0 is identified as critical blowdown the reactor in accordance with N2-EOP-C4. because with Reactor water level unknown, the status of core cooling is unknown. An RPV depressurization is required to allow low pressure injection systems to establish conditions to cool the core. This protects the fuel cladding integrity.
CT-2.0, Given the plant with RPV water level unknown, establish injection Critical Task 2.0 is identified as critical and flood the 'RPV in accordance with N2-EOP-C4. because with Reactor water level unknown, the status of core cooling is unknown. RPV flooding is required to establish conditions to cool the core.
This protects the fuel cladding integrity.
0 0
Appe.ndix D Scenario Outline . Form ES-D-1 0
Facility: Nine Mile Point Unit 2 Scenario No.: NRC-2 Op-Test No.: LC2 18-1 Examiners: Operators:
Initial Conditions: The plant is operating at rated power with 21AS-C3C out of service for unloader valve replacement.
Turnover: The crew will perform a Live Bus Transfer of 2NNS-SWG013 to 2NNS-SWG012 followed by performance of N2-0SP-RMC-W @001, Control Rod Movement and Position Verification Test.
Critical Tasks: See page 2 Event Malf. Event Event No. No. Type* Description N/A N-BOP, Live Bus Transfer of 2NNS-SWG013 to 2NNS-SWG012.
- 1 SRO N2-0P-71B RD11 I-ATC, Perform N2-0SP-RMC-W@001, Control Rod Movement and SRO, Position Verification Test with Rod Position Indication Failure.
2 TS-SRO 0 RD1B C-ATC N2-0SP-RMC-W@001, N2-0P-96, T.S. 3.1.3 CRD Pump Trip on Low Suction Pressure.
3 SRO ARP's, N2-S0P-30 ED05B C-BOP, Loss of 2ENS*SWG102 (Electrical Fault).
4 SRO TS-SRO ARPs~ T.S. 3.5.1, 3.8.4, 3.8.8, 3.6.4.3 FW22B1 R~ATC, First Point Feed Water Heater (1 B) Tube Leak.
5 C-BOP SRO N2-S0P-08, N2-S0P-101 D MT01 M-AII Seismic Event Causes a Break in the Suppression Pool Wall and subsequent RPV Slowdown.
6 N2-EOP-RPV, N2-EOP-PC, N2-EOP-C2 RP03 I-All When manual scram attempted, Control rods fail to insert using RPS, RRCS initiation required to insert the control rods.
7 N2-S0P-101C MS12 C-AII Trip of running EHC Pump with a Failure of Standby to Start.
8 '
N2-EOP-RPV
- (N)ormal, (R)eactivitv, (l)nstrument, (C)omponent (M)aior O'
Facility: Nine Mile Point Unit 2 Scenario No.: NRC-2 Op-Test No.: LC218*1 0 1. Malfunctions after EOP entry (1-2)
Event 7, 8
- 2. Abnormal events (2-4) 2 4
Events 2, 3, 4, 5
- 3. Major transients (1-2) 1 Event 6
- 4. EOPs entered/requiring substantive actions (1-2) 2 N2-EOP-RPV, N2-EOP-PC
- 5. Entry into a contingency EOP with substantive actions (?.1 per scenario set) 1 N2-EOP-C2
- 6. Pre-identified Critical Tasks (> 2) 3
- '.*C~!Yl~A,L]'A~i< ~~~~RIP,J!8N*{.?\(::**.,**.'.:* ?*..*'.(,}\ :. . \. {\s'}: . .":'::\ *.:* ~ij1t1C:~.~:t~~~J.usrif1fATIO~J",/.r*:
- CT-1.0, Given the plant operating in the "Exit Region" of the power to flow Critical Task 1.0 is identified as critical .
map due to a RCS-FCV runback, the crew will insert the first four CRAM rods because without operator action the in accordance with N2-SOP-29. reactor would be operating in a high power (rodline) low core flow condition which is a condition that could cause core power oscillations which is a precursor to .
fuel damage.
CT-2.0, Given the failure of RPS to initiate a successful reactor scram, the Critical Task 2.0 is identified as critical crew will manually initiate ARCS in accordance with N2-SOP-101 C. because the reactor a reactor scram is required "before" the blowdown is initiated to shut down the reactor and reduce the steam generation rate. With the failure of RPS to function, manual action is required to shutdown the reactor.
0 CT-3.0, Given the plant with suppression pool water level that cannot be maintained above the 192' elevation, the crew will commence a RPV blowdown before suppression pool level reaches the 192' elevation in accordance with N2-EOP-PC and N2-EOP-C2.
Critical Task 3.0 is identified as critical because 192' is the minimum indicated suppression pool level. An on-scale indication is required to ensure that the actual suppression pool level is above the top of the SRV discharge devices. If the SRVs were opened with the discharge devices exposed, steam would pass directly into the suppression chamber airspace, bypassing the suppression pool. The resulting pressure increase could exceed the maximum pressure capability of the primary containment.
0
Appendix D Scei'lario Outline Form ES-D-1 0
Facility: Nine Mile Point Unit 2 Scenario No.: NRC-4 Op-Test No.: LC2 18-1 Examiners: Operators:
Initial Conditions: The plant is operating at rated power with 'C' Narrow Range Level Transmitter failed high.
Turnover: The crew will perform N2-0SP-RHS-Q@006, RHR System Loop C Pump and Valve Operability Test and System Integrity Test.
Critical Tasks: See page 2 Event Malf. Event Event No. No. Type* Description N/A N-BOP, Perform N2-0SP-RHS-Q@006 Surveillance Test.
1 SRO N2-0SP-RHS-Q@006 NM19A I-ATC, RBM "A" lnop requires bypassing.
2 SRO, TS-SRO ARP's, N2-0P-92, T.S. 3.3.2.1 CS01A I-BOP, Inadvertent HPCS Initiation.
0 3 SRO, TS-SRO ARP's, N2-0P-33, T.S. 3.5.1 RD05 R-ATC, Control Rod Drift Out.
C-BOP 4
SRO, ARP's, N2-SOP-8, T.S. 3.1.3 TS-SRO Remote I-BOP, CCS-TIK104 Auto Setpoint Failure.
5 CW27 SRO N2-0P-14, N2-SOP-14 NM12B I-ATC, APRM #2 Failure Downscale.
6 SRO, ARP's, N2-0P-92 RD17Z, M-AII Loss of 2NNS-SWG011 & Remaining Condensate Pumps, Scram, ED04A, ATWS, RCIC Trip, RPV Slowdown, Re-inject with Preferred ATWS Injection Systems.
7 FW01B, RC06 N2-EOP-RPV, N2-EOP-PC, N2-EOP-C2, N2;.EOP-C5
8 N2-0P-36, Attachment 1 RH01 C-AII RHR Pump trip during Reflood.
9 N2-EOP-RPV 0
. (N)ormal, (R)eactivitv, (l)nstrument, (C)omoonent, (M)aior
0 Facility: Nine Mile Point Unit 2
- 1. Malfunctions after EOP entry (1-2)
Event 8, 9 Scenario No.: NRC-4 2
Op-Test No.: LC2 18-1
- 2. Abnormal events (2-4) 5 Events 2, 3, 4, 5, 6
- 3. Major transients (1-2) 1 Event 7
- 4. EOPs entered/requiring substantive actions (1-2) 2 N2-EOP-RPV, N2-EOP-PC
- 5. Entry into a contingency EOP with substantive actions (?.1 per scenario set) 2 N2-EOP-C2, N2-EOP-C5
- 6. Pre-identified Critical Tasks (> 2) 5 CRITICAL TASK DESCRIPJIONS: '
CRITICAL TASK JUSTIFICATION:
- , *,; , , e'r ~ * , :*, > ,* ,;': ,' ', * >
CT-1.0, Given the plant at rated power with a control rod drifting out, the Critical Task 1.0 is identified as critical crew will reduce reactor power to approximately 85% in accordance with N2- because without a power reduction, SOP-8 and N2-SOP-101D. APRM power will rise above the licensed limit and present a challenge to thermal limits and be a precursor to fuel damage.
N2-SOP-OB requires a power reduction to approximately 85%. This power reduction is performed to avoid any power peaking concerns due to the unplanned control rod pattern change.
CT-2.0, Given the plant with a high power ATWS and degraded high pressure Critical Task 2.0 is identified as critical preferred injection sources, the crew will inhibit ADS in accordance with N2- because with lowering RPV level; the EOP-CS. ADS System, if not disabled, would automatically open all 7 ADS valves and 0 allow the low pressure EGGS pumps to inject if not terminated and prevented.
With a high power A TWS in progress, the pressure transient and resultant uncontrolled injection of relatively cold water would result in fuel damage.
CT-3.0, Given a failure of the reactor to SCRAM and RPV Slowdown Critical Task 3.0 is identified as critical required, the crew will terminate and prevent all injection sources except because without operator action, the boron, CRD, and RCIC in accordance with N2-EOP-C2. manual RPV blowdown combined with a high power A TWS in progress would cause the uncontrolled injection of relatively cold water which would result in fuel damage.
CT-4.0, Given a failure of the reactor to SCRAM with an RPV blowdown Critical Task 4.0 is identified as critical required, the crew will open all 7 ADS valves in accordance with N2-EOP-C2. because without operator action, reactor pressure would remain to high to facilitate the only remaining preferred injection source to inject into the vessel. This would prevent RPV water level from being restored and therefore prevent adequate core cooling from being assured. The intent is to get at least 7 SRV's (ADS or non-ADS) open.
CT-5.0, Given a failure of the reactor to SCRAM and the RPV has been blown Critical Task 5.0 is identified as critical down per N2-EOP-C2, the crew will resume injection when RPV pressure because without operator action, no
, lowers below the MSCP to restore and maintain RPV water level above the injection will occur. The sources that are MSCWL in accordance with N2-EOP-C5. available for injection must be manually lined up and therefore failure to perform this step would cause RPV water level to continue to lower below the level at which adequate core cooling is assured.
0
ES-301 Administrative Topics Outline Form ES-301-1 0
Facility: Nine Mile Point Unit 2 Date of Examination: December 2019 Examination Level: RO Operating Test Number: LC2 18-1 NRG Administrative Topic (see Note) Type Describe activity to be performed Code*
Determine Containment Water Level Conduct of Operations D,R N2-EOP-6.23, KA 2.1.25 (3.9)
Determine Heatup Rate During Startup Conduct of Operations D,R N2-0SP-RCS-@001, KA 2.1.43 (4.1)
Perform Off-Site AC Breaker Alignment Verification Equipment Control N,S N2-0SP-LOG-W001, KA 2.2.15 (3.9) 0 Radiological and Heat Stress Requirements Related to Operator Work In High Radiation P,R Areas - Valve leak in RWCU Pump Room Radiation Control (2015 NRG)
RP-AA-460, RP-AA-203, KA 2.3.7 (3.5)
Emergency Plan , ,
0 < 'S -, a,, ;/"
NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).
- Type Codes and Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (S 3 for ROs; s 4 for SROs and RO retakes)
(N)ew or (M)odified from bank (.:! 1)
(P)revious 2 exams (S 1, randomly selected) 0
ES-301 Administrative Topics Outline Form ES-301-1 0
Facility: Nine Mile Point Unit 2 Date of Examination: December 2019 Examination Level: SRO Operating Test Number: LC218-1 NRC Administrative Topic (see Note) Type Describe activity to be performed Code*
I Determine the Significance of a Reactivity Conduct of Operations Event and Actions Required D,R OP-AA-300, N2-0P-96, K/A 2.1.37 (4.6)
Reactivate SRO Licenses Conduct of Operations D,R OP-AA-105-102, KA 2.1.4 (3.8)
Perform Off-Site AC Breaker Alignment Verification Equipment Control N,S N2-0SP-LOG-W001, Technical Specifications, KA 2.2.15 (4.3) 0 Radiological and Heat Stress Requirements Related to Operator Work In High Radiation P,R Radiation Control Areas - Valve leak in RWCU Pump Room (2015 NRG)
RP-AA-460, RP-AA-203, KA 2.3.7 (3.6)
Security Event Re-Classification Notification Emergency Plan D,R EP-CE-111, EP-AA-1013, KA 2.4.41 (4.6)
NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).
- Type Codes and Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (S 3 for ROs; s 4 for SROs and RO retakes)
(N)ew or (M)odified from bank (;.: 1)
(P)revious 2 exams (S 1, randomly selected) 0
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 0 Facility: Nine Mile Point Unit 2 Date of Examination: December 2019 Exam Level: RO/SR0-1/SRO-U Operating Test No.: LC2 18-1 NRC
'r /'z . ,
, ', :,';,,,,;< '/,. ,
- a. Transfer Feedwater Level Control to FWS-LV55A at Approximately 2% Power IAW N2-0P-3 D,L,S 2 KIA 259001 A4.05 4.0/3.9) N2-0P-3
- b. Manual Initiation of Control Building Special Filter Train (Alternate Path) A,D,EN,S 9 KIA 290003 A4.01 3.2/3.2) N2-0P-53A
- c. Main Steam Line Warmup Operation (Alternate Path)
A,D,L,S 3 KIA 239001 A4.02 3.2/3.2) N2-0P-1 & N2-SOP-83
- d. Restore SDC with Failure to Inject Requiring Tripping Pump A,D,L,S (Alternate Path) 4 KIA 205000 A4.01 3.7/3.7 N2-0P-31 0 e. Suppression Pool Fill Utilizing CSH Pump KIA 295030 EA 1.03 3.4/3.4 N2-0P-33 M,S 5
- f. Unload and Secure 2EGS*EG1 D,S 6 KIA 264000 A4.02 (3.4/3.4) and A4.04 (3.7/3.7) N2-0SP-EGS-M@001
- g. Enter a Substitute Rod Position in the RWM D,P,S 7
KIA 201006 A4.06 (3.2/3.2) N2-0P-95A (NRG 2015)
- h. Temper SW Using Circ Water N2-0P-11 (Alternate Path) (RO Only) A,D,S 8 KIA 400000 A4.01 3.1 / 3.0 N2-0P-11 rn~P,lan(Syst~his'; (~ fo(RO); (;ff~[ SflO*J); .(:3 'or
, e' *~ ,", .,,.J* ' ,; ( _,, ,f ' ,;' *" ' ~ , ,' o* ~ '; ,
2,fc;;r $.RO.;,ur .
- i. Vent Control Rod Overpiston D,E,R 1 KIA 295015 AA.1.01 (3.8/3.9) N2-EOP-6.14
- j. Place Battery Charger 2BVS-CHGR1 A 1 in Service. (Alternate Path) A,N 6 KIA 263000 A1 .01 (2.5/2.8) N2-0P-73A
- k. Reset a Reactor Protection System Electrical Protection Assembly (EPA) (Alternate Path) A,D 7 0 KIA 212000 A4.14 3.8/3.8 N2-SOP-97
~ *
- ' ' "',,' '~ -'*,""f~ 4 c ~
All RO and SRO-I control room (anc:! in-plant) systems must be diff~rent 'and serve different s~fety'
<*( c'< * ,.,~ ;: 1 c ~. C ,
fuhctlons; 'all 5'SRO-U systems must serve different safety functions; in-plant.systems l3.iid func:Hons may
.overlap>jhose test~d iri the bont~ol room .. ' ' ' ' . ' '
SRd~u 0 ' Criteria for RO 1sifo.:1, ./
- Type.Codes. ',
(A)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank , :59/:58/:54 (E)mergency or abnormal in-plant ~1/~1/~1 (EN)gineered safety feature ~1 / ~1 / ~1 (control room system)
(L)ow-Power / Shutdown ~1/~1/~1 (N)ew or (M)odified from bank including 1(A) ~2/~2/~1 (P)revious 2 exams s 3 / s 3 / s 2 (randomly selected)
(R)CA ~1/~1/~1 (S)imulator Pairings:
'a' alone
'b' then 'h'
'c' alone
'd' alone
'e' then 'f'
'g' alone 0
0
ES-401 1 Form ES-401-1 0 IIFacility: Nine Mile Point Unit 2 Date of Exam:
Tier Group RO KIA Category Points SRO-Only Poirits K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* Total A2 G* Total
- 1. 1 3 3 4 4 3 3 20 4 3 7 Emergency and N/A N/A 2 1 2 1 1 1 1 7 2 1 3 Abnormal Plant Evolutions Tier Totals 4 5 5 5 4 4 27 6 4 10
.1 3 1 3 2 3 3 2 3 2 2 2 26 3 2 5 2.
Plant 2 1 1 1 1 1 1 1 1 .1 1 2 12 0 2 1 3 Systems Tier Totals 4 2 4 3 4 4 3 4 3 3 4 38 5 3 8
- 3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 3 2 2 3 1 2 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 0 3.
point total for each group and tier may deviate by +/-1 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 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 KlAs 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) KlAs 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 KlAs.
- 8. On the following pages, enter the KIA numbers, a brief description of each topic, the topics' IRs 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 ca~egory 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, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to KlAs 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 0 **
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 0 ES-401 BWR Examination Outline Form ES-401-1 Emeraency 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 #
295001 (APE 1) Partial or Complete Loss of 3 Knowledge of the operational 3.6 1 Forced Core Flow Circulation / 1 & 4 implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Thermal Limits 295003 (APE 3) Partial or Complete Loss of 2.1.28 Knowledge and the purpose of major 4.1 2 AC Power/6 system components and controls.
295004 (APE 4) Partial or Total Loss of DC 1 , Knowledge of the reasons for the 2.6 3 Power/ 6 following responses as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Load shedding: Plant-Specific 295005 (APE 5) Main Turbine Generator Trip / 4 Knowledge of the interrelations between 3.3 4 3 . MAIN TURBINE GENERATOR TRIP and
,.* the followina: Main aenerator protection 295006 (APE 6) Scram / 1 6 Ability to determine and/or interpret the 3.5 5 following as they apply to SCRAM:
Cause of reactor SCRAM 295016 (APE 16) Control Room Abandonment 1 Knowledge of the interrelations between 4.4 6 17 CONTROL ROOM ABANDONMENT and the following: Remote shutdown panel:
0 295018 (APE 18) Partial or Complete Loss of CCW/8 7
Plant-Specific Knowledge of the reasons for the following responses as they apply to 3.1 7 PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER:
Cross-connecting with backup systems 295019 (APE 19) Partial or Complete Loss of 2 Ability to operate and/or monitor the 3.3 8 Instrument Air/ 8 following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Instrument air system valves: Plant-
.. Specific 295021 (APE 21) Loss of. Shutdown Cooling / 1 Knowledge of the reasons for the 3.3 9 4 . following responses as they apply to LOSS OF SHUTDOWN COOLING:
Raisina reactor water level 295023 (APE 23) Refueling Accidents / 8 2 Knowledge of'the reasons for the 3.4 10
. following responses as they apply to REFUELING ACCIDENTS: Interlocks associated with fuel handling equipment 295024 High Drywell Pressure/ 5 4 Ability to determine and/or interpret the 3.9 11 following as they apply to HIGH DRYWELL PRESSURE: Suppression chamber pressure: Plant-Specific J
295025 (EPE 2) High Reactor Pressure / 3 1 . Knowledge of the interrelations between 4.11 12 HIGH REACTOR PRESSURE and the
,. followina: RPS 295026 (EPE 3) Suppression Pool High Water 1 Ability to operate and/or monitor the 4.1 13 0 Temperature/ 5 following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Suppression pool cooling
295028 (EPE 5) High Drywell Temperature 2 Ability to operate and/or monitor the 3.9 14 (Mark I and Mark 11 only) I 5 following as they apply to HIGH DRYWELL TEMPERATURE: Drywell ventilation system 295030 (EPE 7) Low Suppression Pool Water 2 Knowledge of the operational 3.5 15 Level/ 5 . implications of the following concepts as they apply to LOW SUPPRESSION POOL WATER LEVEL: Pump NPSH
-. Knowledge of the specific bases for 2.4.. ts EOPs.
295031 (EPE 8) Reactor Low Water Level / 2
3.3 16
. ' Knowledge of local auxiliary operator 295037 (EPE 14) Scram Condition Present 2~4.35 tasks during emergency and the resultant 3.8 17 and Reactor Power Above APRM Downscale . operational effects or Unknown / 1 295038 (EPE 15) High Offsite Radioactivity 3 ,,
Ability to determine and/or interpret the 3.5 18 Release Rate I 9
- following as they apply to HIGH OFF-SITE RELEASE RATE: Radiation levels 600000 (APE 24) Plant Fire On Site/ 8 2 Knowledge of the operational 2.9 19 implications of the following concepts as they apply to Plant Fire On Site: Fire
' Fighting 700000 (APE 25) Generator Voltage and 5 Ability to operate and/or monitor the 3.9 20 Electric Grid Disturbances / 6 following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: Engineered safety
. ' ' features 0 295003 (APE 3) Partial or Complete Loss of 4 Ability to determine and/or interpret the 3.7 76 AC Power/6
- following as they apply to PARTIAL OR COMPLETE LOSS OF A.G. POWER:
System lineups 295016 (APE 16) Control Room Abandonment 2.4.34 Knowledge of RO tasks performed 4.1 77 17 . outside the main control room during an emergency arid the resultant operati6nai effects.
295018 (APE 18) Partial or Complete Loss of 2.1.23 Ability to perform specific system and 4.4 78 CCW/8 integrated plant procedures during all modes of plant operation 295019 (APE 19) Partial or Complete Loss of 2.4.4' Ability to recognize abnormal indications 4.7 79 Instrument Air/ 8 for system operating parameters that are
- entry-level conditions for emerge*ncy and abnormal operatino procedures.
295025 (EPE 2) High Reactor Pressure/ 3 5 Ability to determine and/or interpret the 3.6 80
- following .as they apply to HIGH .
REACTOR PRESSURE: Decay heat qeneration 295031 (EPE 8) Reactor Low Water Level/ 2 3 Ability to determine and/or interpret the 4.2 81
. following as they apply to REACTOR LOW WATER LEVEL: Reactor pressure 295037 (EPE 14) Scram Condition Present 7 Ability to determine and/or interpret the 4.2 82 and Reactor Power Above APRM Downscale following as they apply to.SCRAM cir Unknown / 1 *
- CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE 0 OR UNKNOWN: Containment conditions/isolations KIA Cateoory Totals: 3 3 4 4 3/4 3/3 Group Point Total: 20/7
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) IA #
295002 (APE 2) Loss of Main Condenser 2.1.31 Ability to locate control room switches, 4.6 21 Vacuum /3 controls and indications and. to determine that they are correctly reflectina the desired plant lineup 295009 (APE 9) Low Reactor Water Level / 2 1 Knowledge of the reasons for the 3.2 22 following responses as they apply to LOW REACTOR WATER LEVEL:
Recirculation pump run back: Plant-Specific 295012 (APE 12) High Drywell Temperature/ 2 " Knowledge of the interrelations between 3.6 23 5 HIGH DRYWELL TEMPERATURE and the following: Drvwell cooling 295029 (EPE 6) High Suppression Pool Water 1 Knowledge of the operational 3.4 24 Level/ 5 implications of the following concepts as they apply to HIGH SUPPRESSION POOL WATER LEVEL: Containment integrity 500000 (EPE 16) High Containment Hydrogen 3 Knowledge of the interrelations between 3.3 25 Concentrations I 5 HIGH CONTAINMENT HYDROGEN
. CONCENTRATIONS the following:
Containment Atmosphere Control System 0 295034 (EPE 11) Secondary Containment Ventilation High Radiation / 9 2 Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT VENTILATION HIGH 3.7 26 RADIATION: Cause of high radiation levels 295036 (EPE 13) Secondary Containment 4 Ability to operate and/or monitor the 3.1 27 High Sump/Area Water Level/ 5 following as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL: Radiation monitoring:
Plant-Specific 295015 (APE 15) Incomplete Scram / 1 2 Ability to determine and/or interpret the 4.2 83 following as they apply to.
INCOMPLETE SCRAM: Control rod position 295032 (EPE 9) High Secondary Containment 2 Ability to determine and/or interpret the 3.5 84 Area Temperature/ 5 follqwing as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE: Equipment operabilifv 295034 (EPE 11) Secondary Containment :2;4;5 Knowledge of EOP mitigation 4.7 85 Ventilation High Radiation/ 9 strategies.
KIA Category Point Totals: 1 2 1 1 1/2 1/1 Grouo Point Total: 7/3 0
ES-401 BWR Examination Outline Form ES-401-1 Plant Systems-Tier 2/Grouo 1 (RO/SRO)
System # I Name K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G* KIA Topic(s) IR #
203000 (SF2, SF4 RHR/LPCI) 7 Knowledge of RHR/LPCI: INJECTION 3.7 28 RHR/LPCI: Injection Mode MODE (PLANT SPECIFIC) design
, feature(s) and/or interlocks which provide for the following: Emergency generator load sequencing 205000 (SF4 SCS) Shutdown Cooling 6 , Ability to (a) predict the impacts of the 3.4 29 following on the SHUTDOWN
, COOLING SYSTEM (RHR SHUTDOWN COOLING MODE); and (b) based on
- those predictions, use procedures to correct, control, or mitigate the consequences of those
- abnormal conditions or operations:
SDC/RHR pump trips 209001 (SF2, SF4 LPCS) 8 Knowledge of the physical connections 3.2 30 Low-Pressure Core Spray and/or cause-effect relationships between LOW PRESSURE CORE SPRAY SYSTEM and the following:
'. A.C. electrical power 209002 (SF2, SF4 HPCS) 1 Knowledge of the effect that a loss or 3.9 31 High-Pressure Core Spray malfunction of the HIGH PRESSURE 0 *. CORE SPRAY SYSTEM (HPCS) will have on following: Reactor water level:
- BWR-5,6 211000 (SF1 SLCS) Standby Liquid 5 Ability to manually operate and/or 4.1 32 Control monitor in the control room: Flow indication: Plant-Specific 212000 (SF7 RPS) Reactor Protection 2 Knowledge of the operational 3.3 33 implications of the following concepts as they apply to REACTOR PROTECTION SYSTEM: Specific logic arrangements 215003 (SF7 IRM) 2 Knowledge of the effect that a loss or 3.6 34 Intermediate-Range Monitor malfunction of the following will have on the INTERMEDIATE RANGE MONITOR
. (IRM) SYSTEM: 24/48 volt D.C. power:
Plant-Specific 215004 (SF7 SRMS) Source-Range 1 Knowledge of electrical power supplies 2.6 35 Monitor to the following: SRM channels/detectors 215005 (SF7 PRMS) Average Power 4
- Knowledge of the operational 2.9 36 Range Monitor/Local Power Range implications of the following concepts as Monitor . they apply to AVERAGE POWER RANGE MONITOR/LOCAL POWER
' ' RANGE MONITOR SYSTEM: LPRM detector location and core symmetry 215005 (SF7 PRMS) Average Power 5 Knowledge of the operational 3.6 37 Range Monitor/Local Power Range ** implications of the following concepts as Monitor they apply to AVERAGE POWER RANGE MONITOR/LOCAL POWER 0 RANGE MONITOR SYSTEM: Core flow effects on APRM trip setpoints
ES-401 6 Form ES-401-1 0 217000 (SF2, SF4 RCIC) Reactor 2 Ability to monitor automatic operations 3.6 38 Core Isolation Cooling of the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) including:
Turbine startup 218000 (SF3 ADS) Automatic 3 Knowledge of the physical connections 3.7 39 Depressurization and/or cause-effect relationships between AUTOMATIC DEPRESSURIZATION SYSTEM and the following: Nuclear boiler instrument system 209001 (SF2, SF4 LPCS) 5 Knowledge of the effect that a loss or 2.8 40 Low-Pressure Core Spray malfunction of the following will have on the LOW PRESSURE.CORE SPRAY SYSTEM: ECCS room cooler(s) 223002 (SF5 PCIS) Primary 1 Ability to monitor automatic operations 3.4 41 Containment Isolation/Nuclear Steam of the PRIMARY CONTAINMENT Supply Shutoff ; ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF including:
System indicating lights and alarms 239002 (SF3 SRV) Safety Relief 3 Ability to predict and/or monitor changes 2.8 42 Valves in parameters associated with operating the RELIEF/SAFETY VALVES controls including: Air supply: Plant-Specific 2.1:~8 Knowledge of the purpose and function 259002 (SF2 RWLCS) ReactorWater 4.1 43 Level Control of major system components and
,' controls 0 261000 (SF9 SGTS) Standby Gas Treatment 3* Ability to (a) predict the impacts of the foliowing on the STAND BY GAS
- TREATMENT SYSTEM; and (b) based 2.9 44 on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High train
.** temperature 262001 (SF6 AC) AC Electrical 1
- Knowledge of the effect that a loss or 3.5 45 Distribution malfunction of the A.C. ELECTRICAL DISTRIBUTION will have on following:
Major system loads 262002 (SF6 UPS) Uninterruptable 1
- Ability to (a) predict the impacts of the 2.6 46 Power Supply (AC/DC) following on the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
Under voltage 262002 (SF6 UPS) Uninterruptable 2.l.30 Ability to locate and operate 4.4 47 Power Supply (AC/DC) components, including local controls.
263000 (SF6 DC) DC Electrical 2 Ability to manually operate and/or 3.2 48 Distribution monitor in the control room: Battery voltage indicator: Plant-Specific 264000 (SF6 EGE) Emergency 5 Knowledge of EMERGENCY 3.2 49 Generators (Diesel/Jet) EDG GENERATORS (DIESEUJET) design 0
feature(s) and/or interlocks which provide for the following: Load shedding and sequencing
ES-401 7 Form ES-401-1 0 300000 (SF8 IA) Instrument Air 2 Knowledge of the effect that a loss or 3.3 50
- malfunction of the INSTRUMENT AIR SYSTEM will have on the following:
Systems having pneumatic valves and controls 300000 (SF8 IA) Instrument Air 13 Knowledge of the effect that a loss or 2.8 51
. malfunction of the following will have on the INSTRUMENT AIR SYSTEM: Filters 400000 (SF8 CCS) Component 2 Ability to predict and / or monitor 2.8 52 Cooling Water changes in parameters associated with operating the CCWS controls including:
CCW temperature 400000 (SF8 CCS) Component 4 Knowledge of the physical. connections 2.9 53 Cooling Water . and / or cause-effect relationships between CCWS and the following:
Reactor coolant system in order to determine source(s) of RCS leakage
- into CCWS 209002 (SF2, SF4 HPCS) 8 Ability to (a) predict the. impacts of the 3.2 86 High-Pressure Core Spray following on the HIGH PRESSURE CORE SPR.AY SYSTEM (HPCS);: and (b) based on those predictions,. use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
Inadequate system flow: BWR-5,6 0 215004 (SF7 SRMS) Source-Range Monitor 5 Ability to (a) predict the impacts of the following on the SOURCE RANGE 3.5 87 MONITOR (SRM) SYSTEM; and (b) based on th.ose predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
Faulty or erratic operation of
, detectors/system 218000 (SF3 ADS) Automatic 2.4.9 Knowledge of low power/ shutdown 4.2 88 Depressurization *. ' implications in accident (e'.g. LOCA or loss of RHR) mitigation strategies.
262001 (SF6 AC) AC Electrical 2.1.32 Ability to explain ahd apply all system 4.0 89 Distribution limits and precautions.
263000 (SF6 DC) DC Electrical 1 Ability to (a) predict the impacts of the. 3.2 90 Distribution following on the D.C. ELECTRICAL
. DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Grounds KIA Category Point Totals: 3 1 3 2 3 3 2 3/3 2 2 2/2 Group Point Total: 26/5 0
ES-401 8 Form ES-401-1 0 ES-401 BWR Examination Outline Form ES-401-1 Plant Svstems Tier 2/Group 2 <RO/SRO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* KIA Topic(s) IR #
201003 (SF1 CROM) Control Rod and Drive Knowledge of annunciators 2.4.31 alarms, indications or response 4.2 54 Mechanism procedures 202002 (SF1 RSCTL) Recirculation Flow 6 Knowledge of the effect that a 2.9 55 Control loss or malfunction of the following will have on the RECIRCULATION FLOW CONTROL SYSTEM:
Reactor/turbine pressure reoulatino system: Plant-Specific 215002 (SF7 RBMS) Rod Block Monitor 4 Ability to monitor automatic 3.6 56 operations of the ROD BLOCK MONITOR SYSTEM including:
Verification or proper functioning/ operability: BWR-3,4,5 223001 (SF5 PCS) Primary Containment and Knowledge of the effect that a 3 loss or malfunction of the 3.4 57 Auxiliaries PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES will have on following:
Containment/drywell pressure:
Plant-Specific 234000 (SF8 FH) Fuel-Handling Equipment 1 Knowledge of the physical 3.2 58 connections and/or cause-effect 0 relationships between FUEL HANDLING EQUIPMENT and the followino: Fuel 241000 (SF3 RTPRS) Reactor/Turbine 13 Ability to manually operate 2.9 59 Pressure Regulating and/or monitor in the control room: Turbine inlet pressure 245000 (SF4 MTGEN) Main Turbine 10 Knowledge of MAIN TURBINE 2.6 60 Generator/Auxiliary GENERATOR AND AUXILIARY SYSTEMS design feature(s) and/or interlocks which provide for the following: Extraction steam 259001 (SF2 FWS) Feedwater 6 Ability to predict and/or monitor 2.7 61 changes in parameters associated with operating the REACTOR FEEDWATER SYSTEM controls including:
Feedwater heater level 268000 (SF9 RW) Radwaste Ability to (a) predict the impacts 2.9 1 62 of the following on the RADWASTE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: System rupture 272000 (SF7, SF9 RMS) Radiation Monitoring Knowledge of the operational 1 implications of the following 3.2 63 concepts as they apply to RADIATION MONITORING 0 SYSTEM: Hydrogen injection operations effect on process radiation indications: Plant-Specific
ES-401 9 Form ES-401-1 0 286000 (SF8 FPS) Fire Protection 2
Knowledge of electrical power 2.9 64 suoolies to the followinq: Pumps 288000 (SF9 PVS) Plant Ventilation Ability to evaluate plant 2.1.7 performance and make 4.4 65 operational judgments based on
- operating characteristics, reactor behavior, and instrument interpretation.
201006 (SF7 RWMS) Rod Worth Minimizer Ability to (a) predict the impacts 91 4 3.3 c of the following on the ROD WORTH MINIMIZER SYSTEM
- (RWM) (PLANT SPECIFIC); .and (t>) based on those predictions, use procedures to correct, control, of mitigate the consequences of those
- abnormal conditions or op~rations: Stuck rod: Plant Specific (Not78WR6) 230000 (SF5 AHR SPS) RHR/LPCI:
10 : Ability to (a) predict the impacts 3.0 92 Torus/Suppression Pool Spray Mode of the following on the RHR/LPCI:
TORUS/SUPPRESSION POOL SPRAY MODE; and (b) based on those.predictions, use procedures to correct, control, or mitigate the consequences of 0 those abnormal conditions or operations: Nuclear boiler instrument failures Ability to explain and apply all 245000 (SF4 MTGEN) Main Turbine 2.1.32 system limits and precautions. 4.0 93 Generator/Auxiliary KIA Cateqorv Point Totals: 1 1 1 1 1 1 1 1/2 1 1 2/1 Group Point Total: 12/3 0
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 0 Facilitv: Nine Mile Point Unit 2 Date of Exam:
Category KIA# Topic RO SRO-onlv IR # IR #
Knowledge of industrial safety procedures (such as rotating 2.1.26 equipment, electrical, high temperature, high pressure, caustic, 3.4 66 chlorine, oxygen and hydrogen).
Knowledge of procedures and limitations involved in core 2.1.36 alterations 3.0 67 Ability to use procedures to determine the effects on reactivity
- 1. Conduct of 2.1.43 of plant chanaes 4.1 68 Operations Knowledge of industrial safety procedures (such as rotating 2.1.26 equipment, electrical, high temperature, high pressure, caustic, 3.6 94 chlorine, oxvaen and hydroaenl.
1.,* ;**, ,;* *,
Subtotal .. 3 *' * .. '" < 1 Knowledge of the process for controlling equipment 2.2.14 confiauration or status. 3.9 69 Ability to interpret control room indications to verify the status 2.2.44 and operation of a system, and understand how operator 4.2 70 actions and directives affect plant and svstem conditions
- 2. Equipment Control Knowledge of the process for managing maintenance activities 2.2.18 durinq shutdown operations.
- 3.9 95 Knowledge of the bases in Technical Specifications for limiting 2.2.25 conditions for operations and safety limits 4.2 96 0
Subtotal 2 ,. 2 Knowledge of radiological safety principles pertaining to 2.3.12 licensed operator duties 3.2 71 2.3.15 Knowledge of radiation monitoring systems 2.9 72
- 3. Radiation Ability to use radiation monitoring systems, such as fixed 2.3.5 radiation monitors and alarms, portable survey instruments, 2.9 97 Control personnel monitoring equipment, etc.
2.3.14 Knowledge of radiation or contamination hazards that may 3.8 98 arise during normal, abnormal, or emergency conditions or activities.
Subtotal *., 2 2 Knowledge of operational implications of EOP warnings, 2.4.20 cautions and notes. 3.8 73 2.4.27 Knowledge of "fire in the plant procedures. 3.4 74 Knowledge of the RO's responsibilities in emergency plan 2.4.39 implementation. 3.9 75
- 4. Emergency Procedures/Plan Knowledge of abnormal condition procedures 2.4.11 4.2 99 Knowledge of emergency plan protective action 2.4.44 recommendations. 4.4 100 Subtotal . ~/ ,',. ' 3 .. *' 2 Tier 3 Point Total **) ... 10 *.* .*;' ; 7 0
ES-401 Record of Rejected K/As Form ES-401-4 0
Tier/ Randomly Selected Reason for Rejection Group KIA
.**.*o*.. i' * . *** ,* , ';'..*, ."'.,"';..,t.. ,...:* , .. / / .. /<.*. \.**. i . ... :'.:c'; *. . . ** ;::,,'
- The followina.fopics/ .KlAs. werer~xcluded: from. the :sv$.tematic and. rc;tndorn 'Samplino proces$:
Rejected Emergency and Abnormal Plant Evolution 295027 (EPE 4) High Containment Temperature (Mark Ill 1/1 Containment Only). Nine Mile Point 2 containment design is Mark II.
Rejected Emergency and Abnormal Plant Evolution 295011 (APE 11) High Containment Temperature (Mark Ill 1/2 Containment Only). Nine Mile Point 2 containment design is Mark II.
Rejected Plant System 207000 (SF4 IC) Isolation 2/1 (Emergency) Condenser. : Nine Mile Point 2 has no isolation condenser.
Rejected Plant System 206000 (SF2, SF4 HPCIS) High Pressure Coolant Injection. Nine Mile Point 2 is a BWR 5 2/1 and has no HPCI. HPCS instead (209002) was included in the random drawing.
0 2/2 Rejected Plant System 201004 (SF7 RSCS) Rod Sequence Control. Nine Mile Point 2 has no RSCS.
Rejected Plant System 201005 (SF1, SF7 RCIS) Rod 2/2 Control and Information. Nine Mile Point 2 has no RCIS system.
,";_' -~* ..-;,,.~-.,_-* ': __,"'**:'*,µ"*.',._ :',.-'-,".1'* -~**,, ' .. ,- *,,:- .' -/\ ',:*:_:-* , '~-;-~-:- -~**; "'\ ': ,.::l>:""f(,,,'., _ _.,'._*::..,,,,_ '.-.;/*:*:.: ,. ,*,
.The following KIAswere*resfiropledby th~NRC*ourir:ig the imtial_draftina of.the outline: *.
295025 EK2.06 Rejected 295025 EK2.06 - Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following:
RO Question #12 HPCI Plant Specific. NMP2 does not have HPCI.
1/1 Randomly reselected 295025 EK2.01 - Knowledge of the interrelations between HIGH REACTOR PRESSURE and the followina: RPS.
295034 G2.2.4 Rejected this KIA as this KIA is for a multi-unit site. NMP2 1/2 is considered a single unit site. Randomly reselected SRO Question #85 G2.2.36 262002 A2.04 Rejected this KIA as it is specific to a BWR 1. NMP2 is a BWR5.
RO Question #46 2/1 0 Randomly reselected 262002 A2.01 - Under voltage
ES-401 Record of Rejected KlAs Form ES-401-4 0
234000 K1 .07 Rejected this KlA as it is specific to a Mark Ill containment.
NMP2 is a Mark II containment design.
RO Question #58 2/2 Randomly reselected 234000 K1 .01 - Fuel G2.3.1.2 Rejected KIA due to oversampling. This is the same KIA 3 as RO question #71 ,
SRO Question #97 Randomly reselected G2.3.5 G2.3.15 Rejected KIA due to oversampling. This is the same KIA 3 as RO question #72 SRO Question #98 Randomly reselected G2.3.14 G2.4.20 Rejected KIA due to oversampling. This is the same KIA 3 as RO question #73 SRO Question #99 Randomly reselected G2.4.11
- ,,**: ' ' . c.,.,',J':,. ,*::~':., ,/',,,, .;,,, '. ;,;,.,;:,,,,;;*,:**,,.'.,,, T**/£;:</c ,*, ,,,,*
The followJrig K/As were resampled *cturinQ construction ofthe written exam: ,". , , . : , .**
Question #6 An acceptable question could not be developed without Control testing minutia due to limited interrelation between Control 295016 0 Room Abandonment AK2.03 - Knowledge Room HVAC and Control Room Abandonment.
Randomly resampled KIA 295016 Control Room Abandonment AK2.01 - Knowledge of the interrelations of the interrelations 1/ 1 between CONTROL ROOM ABANDONMENT and the between CONTROL following: Remote shutdown panel: Plant-Specific.
ROOM ABANDONMENT and the following:
Control room HVAC Question #7 An acceptable question could not be developed at a discriminating RO level due to the simplicity of the KIA and 295018 Partial or limited plant-specific guidance.
Complete Loss of ccw Randomly resampled KIA 295018 Partial or Complete Loss of CCW AK3.07 - Knowledge of the reasons for the AK3.04 - Knowledge following responses as they apply to PARTIAL OR of the reasons for COMPLETE LOSS OF COMPONENT COOLING WATER:
the following Cross-connecting with backup systems.
1/ 1 responses as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT 0 COOLING WATER:
Starting standby pump
ES-401 Record of Rejected K/As Form ES-401-4 0
Question #11 An acceptable question could not be developed due to lack of operationally relevant use of Drywell radiation levels 295024 High Drywell related to high Drywell pressure at the RO job level. Also Pressure would have potentially overlapped with Question #97.
EA2.08 - Ability to Randomly resampled KIA 295024 High Drywell Pressure determine and/or EA2.04 - Ability to determine and/or interpret the following interpret the 1/ 1 as they apply to HIGH DRYWELL PRESSURE:
following as they Suppression chamber pressure: Plant-Specific.
apply to HIGH DRYWELL PRESSURE:
Drywell radiation levels Question #16 An acceptable question could not be developed due to mismatch between the generic KIA and the given 295031 Reactor Low evolution. Also, this generic KIA is already used on Water Level Question #85.
2.2.36 - Ability to Randomly resampled KIA 295031 Reactor Low Water analyze the effect of Level 2.4.18 - Knowledge of the specific bases for EOPs.
1/ 1 maintenance 0 activities, such as degraded power sources, on the status of limiting conditions of operations Question #25 An acceptable and discriminating question could not be developed without testing minutia due to limited 295032 High operationally relevant tie between Fire Protection and the Secondary given evolution. Additionally, N2-EOP-SC concepts are Containment Area oversampled on the exam (Questions #26, .#27, #84, #85),
Temperature preventing development of a question without overlap.
EK2.03 - Knowledge Randomly resampled KIA 500000 High Containment of the interrelations Hydrogen Concentration EK2.03 - Knowledge of the 1/2 between HIGH interrelations between HIGH CONTAINMENT HYDROGEN SECONDARY CONCENTRATIONS the following: Containment CONTAINMENT Atmosphere Control System.
AREA TEMPERATURE and the following:
Fire protection system 0
ES-401 Record of Rejected K/As Form ES-401-4 0
Question #40 ADS I SRV concepts are oversampled (Questions #39, 40, 42, 88), with 218000 triple sampled. The KlAs for #40 and 218000 ADS
- 42 in particular were causing direct overlap issues.
K6.05 - Knowledge Randomly resampled KIA 209001 Low Pressure Core of the effect that a Spray K6.05 - Knowledge of the effect that a loss or loss or malfunction malfunction of the following will have on the LOW 2/1 of the following will PRESSURE CORE SPRAY SYSTEM: ECCS room have on the cooler(s).
AUTOMATIC DEPRESSURIZATI ON SYSTEM: A.G.
power: Plant-Specific Question #44 An acceptable question could not be developed because "high fuel pool ventilation radiation" does not have an 261000 Standby impact at NMP2.
Gas Treatment Randomly resampled KIA 261000 Standby Gas Treatment A2.12 - Ability to (a)
A2.03 - Ability to (a) predict the impacts of the following on predict the impacts the STANDBY GAS TREATMENT SYSTEM; and (b) of the following on based on those predictions, use procedures to correct, the STANDBY GAS control, or mitigate the consequences of those abnormal 0 TREATMENT SYSTEM; and (b) based on those conditions or operations: High train temperature.
2/1 predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High fuel pool ventilation radiation: Plant-Specific 0
ES-401 Record of Rejected K/As Form ES-401-4 0
Question #47 An acceptable question could not be developed because there are no Immediate Actions related to this system at 262002 NMP2.
Uninterruptable Power Supply Randomly resampled KIA 262002 Uninterruptable Power (AC/DC) Supply (AC/DC) 2.1.30 - Ability to locate and operate components, including local controls.
2.4.49 - Ability to perform without 2I 1 reference to procedures those actions that require immediate operation of system components and controls Question #57 An acceptable question could not be developed without overlapping/oversampling concepts tested in Questions 223001 Primary
- 42 and #50.
Containment and Auxiliaries Randomly resampled KIA 223001 Primary Containment 0 K3.08 - Knowledge of the effect that a and Auxiliaries K3.03 - Knowledge of the effect that a loss or malfunction of the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES will have on following:
loss or malfunction Containment/drywell pressure: Plant-Specific.
of the PRIMARY 2/2 CONTAINMENT SYSTEM AND AUXILIARIES will have on following:
Pneumatically operated valves internal to containment/drywell:
Plant-Specific Question #65 An acceptable question could not be developed due to lack of applicable reference materials (graphs, curves, tables, 288000 Plant etc.) related to Plant Ventilation systems.
Ventilation Randomly resampled KIA 288000 Plant Ventilation 2.1.7 -
2.1.25 - Ability to 2/2 Ability to evaluate plant performance and make operational interpret reference judgments based on operating characteristics, reactor materials, such as behavior, and instrument interpretation.
graphs, curves, tables, etc.
0
ES-401 Record of Rejected K/As Form ES-401-4 0
Question #66 The given generic KIA is already used on the SRO exam and is not "generic" in nature, as preferred for a Tier 3 2.1.32 - Ability to question (would lead to a Tier 2 question).
explain and apply all system limits and Randomly resampled KIA 2.1.26 - Knowledge of industrial 3
precautions. safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen).
Question #69 An acceptable question could not be developed that was both discriminating and at the RO level.
2.2.23 - Ability to track Technical Randomly resampled KIA 2.2.14 - Knowledge of the 3 Specification limiting process for controlling equipment configuration or status.
conditions for operations.
Question #77 An acceptable question could not be developed for this evolution with the given generic KIA without testing minutia 295016 Control due to limited relevant surveillance procedures.
Room Abandonment 0 1/ 1 2.2.12 - Knowledge of surveillance Randomly resampled KIA 295016 Control Room Abandonment 2.4.34 - Knowledge of RO tasks performed outside the main control room during an emergency and procedures the resultant operational effects.
Question #79 An acceptable question could not be developed for this evolution with the given generic KIA due to lack of related 295019 Partial or Immediate Actions.
Complete Loss of Instrument Air Randomly resampled KIA 295019 Partial or Complete Loss of Instrument Air 2.4.4 - Ability to recognize abnormal 2.4.49 - Ability to indications for system operating parameters that are entry-perform without level conditions for emergency and abnormal operating 1/ 1 reference to procedures.
procedures those actions that require immediate operation of system components and controls 0
- o ES-401 Record of Rejected KlAs Form ES-401-4 Question #85 An acceptable question could not be developed for this evolution with the given generic KIA without being overly 295034 Secondary similar to Question #84 (both related to N2-EOP-SC and Containment Technical Specifications).
Ventilation High Radiation Randomly resampled KIA 295034 Secondary Containment Ventilation High Radiation 2.4.6 - Knowledge of EOP 2.2.36 - Ability to mitigation strategies.
analyze the effect of 1/2 maintenance activities, such as degraded power sources, on the status of limiting conditions for operations Question #87 A discriminating question could not be developed for the original KIA due to lack of SRO level material to test for just 215004 Source a failed recorder (other instruments still satisfy all Range Monitor requirements, just need to get maintenance to fix recorder).
A2.06 - Ability to (a) 0 predict the impacts of the f9llowing on Randomly resampled KIA 215004 Source Range Monitor A2.05 - Ability to (a) predict the impacts of the following on the SOURCE RANGE MONITOR (SRM) SYSTEM; and (b) the SOURCE based on those predictions, use procedures to correct, RANGE MONITOR control, or mitigate the consequences of those abnormal (SRM) SYSTEM; conditions or operations: Faulty or erratic operation of 2/ 1 and (b) based on detectors/system.
those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Failed recorder 0
ES-401 Record of Rejected KlAs Form ES-401-4 0
Question #59 An acceptable question could not be developed that was both discriminating and not minutia.
241000 Reactor/Turbine Randomly resampled KIA 241000 Reactor/Turbine Pressure Regulating Pressure Regulating A4.13 - Ability to manually operate and/or monitor in the control room: Turbine inlet pressure.
2/2 A4.12 - Ability to manually operate and/or monitor in the control room:
Turbine acceleration 0
0