ML023240268
| ML023240268 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 09/27/2002 |
| From: | Pisano L Constellation Nuclear |
| To: | Conte R NRC/RGN-I/DRS/OSB |
| Conte R | |
| References | |
| 50-220/02-303 50-220/02-303 | |
| Download: ML023240268 (61) | |
Text
8113/02 NMP#1 EXAM OUTLINE COMMENTS 0
RO Admin, Test A, A.2.2 questions are leading - review ST and determine acceptability.
0 SRO Admin, Test A,. A. 1.1 J PM will be revised to 2 questions because of confusion with the contractors.
0 SRO Admin, Test A, A.3 will be revised to review a rad waste discharge permit. The proposed discharge was an environmental concern not a rad concern.
0 SRO Admin, Test A, A.4, will include an EP classification as well as a PAR recommendation.
0 RO Admin, Test B, A.1.2, too close to Test A, testing the same concept will be replaced concern over potential exam compromise.
0 RO Admin, Test B, A.1.3, the question basically asks the scram actions that will be tested more appropriately and extensively on C part of the exam (dynamic scenarios).
This will be replaced.
0 RO Admin, Test B, A.2.2, rewriter questions to be less leading - review ST and take actions as appropriate.
RO Admin, Test B, A.3.1, GET level of knowledge - replace.
0 RO Admin, Test B, A.3.2, too close to SRO Test A, replace.
0 RO Admin, Test B, A.4, Typo task involves a medical emergency.
0 SRO Admin, Test B, A.1.1, A.1.2 and A.2 - all three of these JPMs are the same as those listed on SRO admin Test A.
0 SRO Admin, Test B, A.3.1, same concept as tested on RO admin test A.
0 SRO Admin, Test B, A.3.2, will be revised to review a rad waste discharge permit. The proposed discharge was an environmental concern not a rad concern.
0 SRO Admin, Test B, A.4, will include an EP classification as well as a PAR recommendation.
0 Submitted 2 JPM sets will try to get away using only one version.
0 Scenario # 1, impact of fault on Power Bus 102?
0 Scenario Critical Tasks need to be developed that meet NUREG 1021,App D, sect D criteria-not listed for all scenarios submitted except for #3.
0 Scenario #3 success path for ATWS?
0 Scenario #4, what is the plant impact or impact to operators by failing core spray lVs to open until reactor pressure has decreased to 250 psig?
0 RO written, tier3, Conduct of Ops 2.2.31 & Equip Control, 2.2.2 maybe better tested on part C of the exam.
SRO written, tier3, Rad Controls, 2.3.1 and 2.3.9 may not be SRO only subject material depends on what is developed.
0 SRO written, tier 1, group 2, 295034A2, and tier 2, group 1, 226001, EOP basis questions may not be valid SRO only topics.
SRO written, tier 2, group 3, 201003 2.1.14 may not be valid SRO only topic.
ES-410-10, SRO written, tier 1, group 1, 295009 AK3.02, doesn't appear to be a valid reason for rejection.
P.O. Box 63 Lycoming, New York 13093 0
Constellation Nuclear Nine Mile Point Nuclear Station A Member of the Constellation Energy Group July 15, 2002 NMP-97932 Mr. Hubert J. Miller Regional Administrator USNRC Region I 475 Allendale Road King of Prussia, PA 19406 ATTENTION: Mr. John Caruso
SUBJECT:
NINE MILE POINT UNIT 1 INITIAL OPERATOR EXAMINATION OUTLINE SUBMITTAL Mr. Miller:
In response to the NRC Corporate Notification Letter dated June 4, 2002, arrangements were made for the administration of licensing examinations at Nine Mile Point, Unit 1 during the week of September 30, 2002. The examinations are being prepared based on the guidelines in Revision 8, Supplement 1, of NUREG 1021, "Operator Licensing Examination Standards for Power Reactors." To meet the examination schedule, Nine Mile Point Nuclear Station is required to furnish the examination outlines by July 17, 2002. Enclosed are the examination outline and quality checklists.
Please withhold these examination materials from public disclosure until after the examinations have been completed.
Nine Mile Point Nuclear Station has used an industry standard and widely available commercial product to develop the exam outlines. The written exam outlines for the Nine Mile Point Unit 1 RO and SRO exams were randomly generated, as required by ES-401, by WD Associates, utilizing the BWR Owners Group sponsored "BWR K/A CATALOG Version 1.07" Test Outline Generator Program. This software program was developed by WD Associates, for randomly generating written exam outlines.
The following K/A statements were not included in the random generation process, as allowed by ES 401, D.l.b:
"* K/A Catalog was pre-screened prior to random outline generation. Non applicable system K/A statements were manually suppressed. Enclosed in the outline submittal package is a marked up K/A Catalog, identifying the K/A statements that were excluded.
"* All K/A statements with an importance rating of below 2.5 were pre-screened by the software program during the random outline generation process.
"* Generic K/A statements for which it is not possible to develop Tier 1 and Tier 2 system/evolution questions were pre-screened.
Page 2 NMP-97932 July 17, 2002 The random outline was generated for both the RO and SRO written exams. The program randomly selected SRO only K/A topics per NUREG 1021, Rev. 8, Supplement 1, from the facility specific K/A Catalog database. The program next randomly selected the remaining K/A topics to satisfy the SRO exam requirements. These test items are common to both the RO and SRO outlines. These common K/A topics were then inserted into the RO outline, followed by the remaining RO K/A topics. The random generation process did not result in the necessity to replace any K/A statements that can result when the software program generates the RO and SRO outlines together. These outlines were then saved with password protection for exam material development.
If you have any questions regarding the examination outline submittal, please contact Mr. Jerry Bobka (Facility Contact) at 315-349-2569 or Mr. David J. Russell (Initial Training Supervisor) at 315-349-4568.
Sincerely, Louis E. Pisano Manager Nuclear Training LEP/crr
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ES-401 Record of Rejected KIAs Form ES-401-10 (R8, SI)
Tier /
Randomly Selected Reason for Rejection Group K/A SRO Tier 1/Group 295009 AK3.02 Replaced with 295009 AA1.01 - Ability to operate and/or monitor the following as they apply to Low Reactor Water Level: Reactor 1
feedwater better match for question written SRO Tier 1/Group 295038 EK1.03 Original K/A not appropriate for >LOD1 question replaced with 295024 2.1.23 - Ability to perform specific system and integrated plant 1
procedures during different modes of plant operation added as SRO only was RO only better SRO only as written SRO Tier 1/Group 295030 EK3.02 Original K/A not applicable to NMPCU1 design replaced with 295003 2.4.41 - Knowledge of the emergency action level threshold and 1
classification SRO Tier 1/Group 295038 2.1.32 Original K/A not appropriate for >LOD1 question replaced with 295038 2.3.9 - Knowledge of the process for performing a containment 1
purge SRO Tier 2/Group 223001 K6.10 Original K/A not applicable to NMPCU1 design replaced with 216000 2.4.6 - Knowledge of symptom based EOP strategies to SRO 1
exam only removed from RO exam better SRO only as written SRO Tier 2/Group 226001 K5.02 Original K/A not appropriate for >LOD1 question replaced with 226001 K5.06 - Knowledge of the operational implications of the 1
following concepts as they apply to the RHR/LPCI: CTMT Spray Mode: Vacuum breaker operation SRO Tier 2/Group 239002 K2.01 Original K/A not appropriate for >LOD1 question replaced with 239002 A1.01 - Ability to predict and/or monitor changes in parameters 1
associated with operating the SRVs controls including: Tail pipe temperature SRO Tier 2/Group 215001 A1.03 Original K/A double jeopardy with RO only question in same tier and system replaced with 215001 A1.01 - Ability to predict and/or 3
monitor changes in parameters associated with operating the Traversing In-core Probe controls including: Radiation levels SRO Tier 1/Group 295018 2.4.49 Original K/A not applicable to SRO only replaced with 2.4.9 - Knowledge of low power/shutdown implications in accident (eg. LOCA or 2
loss of RHR) mitigation strategies SRO Tier 1/Group 295034 2.4.49 Original K/A not applicable to SRO only replaced with 2.4.18 - Knowledge of the specific bases for EOPs 2
SRO Tier 2/Group 212000 2.4.49 Original K/A not applicable to SRO only replaced with 2.4.21 - Knowledge of the parameters and logic used to assess the status of 1
safety functions including:
SRO Tier 2/Group 226001 2.4.49 Original K/A not applicable to SRO only replaced with 2.4.22 - Knowledge of the bases for prioritizing safety functions during I
abnormal/emergency operations RO Tier 1/Group 295009 AK3.02 Replaced with 295009 AA1.01 - Ability to operate and/or monitor the following as they apply to Low Reactor Water Level: Reactor 1
feedwater better match for question written RO Tier 1/Group 2
295038 EK1.03 Original K/A not appropriate for >LOD1 question replaced with 295029 2.1.30 - Ability to locate and operate components/including local controls
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RO Tier 1/Group 295030 EK3.02 Original K/A not applicable to NMPCU1 design replaced with 295030 AA2.02 -Ability to determine and interpret the following as they 2
apply to Control Room Abandonment: Reactor water level added to RO as a both question was SRO only K/A RO Tier 1/Group 295038 2.1.32 Original K/A not appropriate for >LOD1 question replaced with 295038 2.3.9 - Knowledge of the process for performing a containment 2
purge RO Tier 2/Group 223001 K6.10 Original K/A not applicable to NMPCU1 design replaced with 223001 K4.05 - Knowledge of Primary CTMT and Auxiliaries design 1
feature(s) and or interlock(s) which provide for the following: Maintains proper suppression pool to drywell differential pressure RO Tier 2/Group 226001 K5.02 Original K/A not appropriate for >LOD1 question replaced with 226001 K5.06 - Knowledge of the operational implications of the 2
following concepts as they apply to the RHR/LPCI: CTMT Spray Mode: Vacuum breaker operation RO Tier 2/Group 239002 K2.01 Original K/A not appropriate for >LOD1 question replaced with 239002 AI.01 - Ability to predict and/or monitor changes in parameters 1
associated with operating the SRVs controls including: Tail pipe temperature RO Tier 2/Group 215001 A1.03 Original K/A double jeopardy with K6.04 question replaced with 215001 Al.01 - Ability to predict and/or monitor changes in parameters 3
associated with operating the Traversing In-core Probe controls including: Radiation levels RO Tier 1/Group 295024 2.1.23 Removed was RO only better SRO only as written replaced with 295005 2.1.33 - Ability to recognize indications for system operating 1
parameters which are entry-level conditions for technical specifications to RO exam RO Tier 2/Group 216000 2.4.6 Removed from RO exam better SRO only as written replaced with 216000 2.4.10 - Knowledge of annunciator response procedures 1
RO Tier 2/Group 202001 2.1.27 Original K/A not appropriate for >LOD1 question replaced with 204000 2.1.32 - Ability to explain and apply system limits and 2
precautions RO Tier 2/Group 290001 A4.06 NMPCU1 has RB not a Fuel Building replaced with 290001 A4.02 - Ability to manually operate and/or monitor in the control room:
2 Reactor building area temperatures RO Tier 3 2.2.3 Original K/A removed not applicable to NMPCU1 replaced with 2.4.47 - Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material RO Tier 3 2.3.11 Original 2.3.11 question written made better SRO only replaced with 2.3.4 - Knowledge of radiation exposure limits and contamination control/including permissible levels in excess of those authorized original added to SRO only I
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ES-401 BWR SRO Examination Outline Form ES-401-1 (R8, SI)
Facility:
NMPC UI Date of Exam: 09130/02 Exam Level: SRO 1
K/A Category Points Tier Group K
K K
K K
K A
A A
A G
Point 1
2 3
4 5
6 1
2 3
4 Total
- 1.
1 35 2
6 4
6 26 Emergency &
2 3
3 2
3 3
3 17 Abnormal Tier 6
8 4
9 7
9 43 Evolutions Totals 1
2 1
2 2
2 113 2
3 1
4 23
- 2.
2 1
1 1
1 1
2 1
1 1
1 2
13 Plant 3
1 0
0 1
00 1
0 0
0 1
4 Systems Tier 4
2 3
4 3
3 5
3 4 2
7 40 Totals I
Cat1 Cat2 Cat3 Cat4
- 3. Generic Knowledge and Abilities 4
4 4
5 17 Note: 1.
Ensure that at least two topics from every K/A category are sampled within each tier (i.e., the "Tier Totals" in each K/A category shall not be less than two).
- 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 NRC revisions. The final exam must total 100 points.
- 3.
Select topics from many systems; avoid selecting more than two or three K/A topics from a given system unless they relate to plant-specific priorities.
- 4.
Systems/evolutions within each group are identified on the associated outline.
- 5.
The shaded areas are not applicable to the category/tier.
6.*
The generic K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system.
- 7.
On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importance ratings for the SRO license level, and the point totals for each system and category. K/As below 2.5 should be justified on the basis of plant-specific priorities. Enter the tier totals for each category in the table above.
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ES-401 BWR SRO Examination OutlineForm ES-401-1 (R8, S)
Emeraencv and Abnormal Plant Evolutions - Tier 1/Group I E/APE # / Name / Safety Function K
K K
A A
G K/A Topic(s)
Imp.
Points 1
2 3
1 2
295003 Partial or Complete Loss of AC Pwr / 6 X
2.4.41 - Knowledge of the emergency action level threshold and 1
classification 295007 High Reactor Pressure / 3 X
X AK2.05 - Knowledge of the interrelations High Reactor Pressure and 3.1 1
the following: Shutdown cooling RO001,SRO002 2.1.33 - Ability to recognize indications for system operating parameters 4.0 1
which are entry-level conditions for technical specifications RO009,SRO00I 295009 Low Reactor Water Level / 2 X
X AKI.05 - Knowledge of the operational applications of the following 3.4 1
concepts as they apply to the Low Reactor Water Level: Natural circulation RO002,SRO003 AA1.01 - Ability to operate and/or monitor the following as they apply to 3.9 Low Reactor Water Level: Reactor feedwater R0003,SRO004 295010 High Drywell Pressure / 5 X
AA1.03 - Ability to operate and/or monitor the following as they apply to 2.6 1
High Drywell Pressure: Nitrogen makeup R0004,SRO005 295013 High Suppression Pool Temp./ 5 X
X AKI.04 - Knowledge of the operational applications of the following 3.2 1
concepts as they apply to the High Suppression Pool Temp: Complete condensation R0020,SRO006 AK2.01 - Knowledge of the interrelations between High Suppression 3.7 Pool Temp and the following: Suppression pool cooling R0021,SRO007 295014 Inadvertent Reactivity Addition /
X AAI.07 - Ability to operate and/or monitor the following as they apply to 4.1 1
Inadvertent Reactivity Addition: Cold water injection R0005,SROO08 295015 Incomplete SCRAM / I X
X AK2.04 - Knowledge of the interrelations between Incomplete SCRAM 4.1 1
and the following: RPS R0006,SRO009 AK3.01 - Knowledge of the reasons for the following responses as they apply to Incomplete SCRAM: Bypassing rod insertion blocks 3.7 R00007,SRO010 295016 Control Room Abandonment / 7 X
X AA1.03 - Ability to operate and/or monitor the following as they apply to 3.1 1
Control Room Abandonment: RPIS R0022,SRO012 AA2.02 - Ability to determine and interpret the following as they apply to Control Room Abandonment: Reactor water level 4.3 1
295017 High Off-site Release Rate / 9 X
X AA2.01 - Ability to determine and interpret the following as they apply to
- 4.
1 High Off-site Release Rate: Off-site release rate 2.2.25 - Knowledge of bases in technical specifications for limiting conditions for operations and safety limits 295023 Refueling Accidents / 8 X
X AA1.06 - Ability to operate and/or monitor the following as they apply to Refueling Accidents: Neutron monitoring R0034,SRO016 AA2.05 - Ability to determine and interpret the following as they apply to 3.4 I
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Refueling Accidents: Entry conditions of emergency plan 4.6 295024 High Drywell Pressure / 5 X
X EA2.01 - Ability to determine and interpret the following as they apply to 4.4 1
High Drywell Pressure: Drywell pressure R0008,SR0017 2.1.23-Ability to perform specific system and integrated plant procedures during different modes of plant operation SR0024
- 4.
1 295026 Suppression Pool High Water Temp. / 5 X
X EK2.03 - Knowledge of the interrelations between Suppression Pool 3.6 1
High Water Temp and the following: Suppression chamber pressure EK3.03 - Knowledge of the reasons for the following responses as they apply to Suppression Pool High Water Temp: Suppression pool spray 3.8 1
295030 Low Suppression Pool Water Level / 5 X
2.1.32 - Ability to explain and apply system limits and precautions a
1 295037 SCRAM Condition Present and Power X
X EK2.13 - Knowledge of the interrelations between SCRAM Condition 4.1 1
Above APRM Downscale or Unknown / 1 Present and Power Above APRM Downscale or Unknown and the following: Alternate boron injection methods RO011,SRO022 EA1.01 - Ability to operate and/or monitor the following as they apply to 4.6 1
SCRAM Condition Present and Power Above APRM Downscale or Unknown: RPS RO012,SRO023 295038 High Off-site Release Rate / 9 X
2.3.9 - Knowledge of the process for performing a containment purge 3.4 1
500000 High Containment Hydrogen Conc. /5 X
EKI.01 - Knowledge of the operational applications of the following 3.9 1
concepts as they apply to the High Containment Hydrogen Conc:
Containment integrity R0013,SRO026 K/A Category Totals:
i3 I5 I2 I6 I4 I 6I Group Point Total:
26
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ES-401 BWR SRO Examination OutlineForm ES-401-1 (R8, SI)
Emergency and Abnormal Plant Evolutions - Tier l/Group 2 E/APE # / Name I Safety Function K
K K
A A
G K/A Topic(s)
Imp.
Points 1
2 3
1 2
295001 Partial or Complete Loss of Forced Core X
X AA1.05 - Ability to operate and/or monitor the following as they apply to 3.3 1
Flow Circulation / 1 & 4 Partial or Complete Loss of Forced Core Flow Circulation:
Recirculation flow control R0014,SRO028 AA2.06 - Ability to determine and interpret the following as they apply to Partial or Complete Loss of Forced Core Flow Circulation: Nuclear boiler instrumentation SR0027 295002 Loss of Main Condenser Vacuum / 3 X
AK3.02 - Knowledge of the reasons for the following responses as they 3.4 1
apply to Loss of Main Condenser Vacuum: Turbine trip ROO16,SRO029 295004 Partial or Total Loss of DC Pwr / 6 X
X AK1.04 - Knowledge of the operational applications of the following 2.9 1
concepts as they apply to the Partial or Total Loss of DC Pwr: Effect on battery discharge rate AK2.01 - Knowledge of the interrelations between Partial or Total Loss 3.1 1
of DC Pwr and the following: Battery charger 295012 High Drywell Temperature / 5 X
AK1.02 - Knowledge of the operational applications of the following 3.2 1
concepts as they apply to the High Drywell Temperature: Reactor power level control R0019,SRO032 295018 Partial or Total Loss of CCW / 8 X
X AA1.03 - Ability to operate and/or monitor the following as they apply to 3.4 1
Partial or Total Loss of CCW: Affected systems so as to isolate damaged portions R0023,SRO034 2.4.9 - Knowledge of low power/shutdown implications in accident (eg.
3 LOCA or loss of RHR) mitigation strategies 295022 Loss of CRD Pumps / 1 X
AK2.07 - Knowledge of the interrelations between Loss of CRD Pumps 3.6 1
and the following: Reactor pressure (scram assist) R0024,SRO035 295028 High Drywell Temperature I 5 X
X EK3.04 - Knowledge of the reasons for the following responses as they 3.8 1
apply to High Drywell Temperature: Increased drywell cooling EA2.04 - Ability to determine and interpret the following as they apply to High Drywell Temperature: Drywell pressure p
1 295033 High Secondary Containment Area X
EK1.02 - Knowledge of the operational applications of the following 4.2 1
Radiation Levels / 9 concepts as they apply to the High Secondary Containment Area Radiation Levels: Personnel protection 295034 Secondary Containment Ventilation High X
X EA2.02 - Ability to determine and interpret the following as they apply to
'42 I
Radiation / 9 Secondary Containment Ventilation High Radiation: Cause of high radiation levels 2.4.18 - Knowledge of the specific bases for EOPs I
295036 Secondary Containment High Sump/Area Water Level / 5 EK2.02 - Knowledge of the interrelations between Secondary Containment High Sump/Area Water Level and the following: Post-2.9 I
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X
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accident sampling system 600000 Plant Fire On Site / 8 X
X AAI.05 - Ability to operate and/or monitor the following as they apply to 3.1 Plant Fire On Site: Plant and control room ventilation systems 2.4.30 - Knowledge of which events related to system operations/status should be reported to outside agencies K/A Category Point Totals:
3 3
2 1 3 3
3 Group Point Total:
7I
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ES-401BWR SRO Examination OutlineForm ES-401-1 (R8, S1)
Plant Systems - Tier 2/Group 1
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System # Name K
K K
K K
K A
A A
A G
K/A Topic(s)
Imp.
Points 1
2 3
4 5
6 1
2 3
4 206000 HPCI X
X A2.05 - Ability to (a) predict the impacts of 3.8 1
the following on the HPCI and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: DC failures R0039,SRO044 A3.07 - Ability to monitor automatic 3.8 1
operations of the HPCI including: Lights and alarms R0040,SRO045 209001 LPCS X
X A1.03 - Ability to predict and/or monitor 3.9 I
changes in parameters associated with operating the LPCS controls including:
Reactor water level R0041,SRO046 A4.09 - Ability to manually operate and/or 3.5 1
monitor in the control room: Suppression pool level R0042,SRO047 212000 RPS X
2.4.21 - Knowledge of the parameters and 4.3 logic used to assess the status of safety functions including:
215004 Source Range Monitor X
X K4.06 - Knowledge of Source Range 3.2 1
Monitor design feature(s) and or interlock(s) which provide for the following: IRM/SRM interlock R0044,SRO049 K5.03 - Knowledge of the operational 2.8 1
implications of the following concepts as they apply to the Source Range Monitor:
Changing detector position R0045,SRO050 215005 APRMI LPRM X
X K2.02 - Knowledge of electrical power 2.8 1
supplies to the following: APRM channels R0046,SRO051 A1.03 - Ability to predict and/or monitor changes in parameters associated with 3.6 I
operating the APRM/LPRM controls including: Control rod block status R0047,SRO052 216000 Nuclear Boiler Instrumentation X
2.4.6 - Knowledge of symptom based 4
I EOP strategies SR0055 X
A3.03 - Ability to monitor automatic operations of the ADS including: ADS valve acoustical monitor noise R0050,SRO053 3.8 218000 ADS 1
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223001 Primary CTMT and Auxiliaries X
K1.10 - Knowledge of the physical 3.1 connections and/or cause-effect relationships between Primary CTMT and Auxiliaries and the following: Plant air systems R0052,SR0054 223002 PCIS/Nuclear Steam Supply Shutoff X
2.4.6 - Knowledge of symptom based EOP 4-0 strategies 226001 RHR/LPCI: CTMT Spray Mode X
X K5.06 - Knowledge of the operational 2.8 implications of the following concepts as they apply to the RHR/LPCI: CTMT Spray Mode:
Vacuum breaker operation 2.4.22 - Knowledge of the bases for 4
prioritizing safety functions during abnormal/emergency operations 239002 SRVs X
X K3.01 - Knowledge of the effect that a loss or 4.0 malfunction of the SRVs will have on the following: Reactor pressure control A1.01 - Ability to predict and/or monitor 3.4 changes in parameters associated with operating the SRVs controls including: Tail pipe temperature 241000 Reactor/Turbine Pressure Regulator X
K1.01 - Knowledge of the physical 3.9 connections and/or cause-effect relationships between Reactor/Turbine Pressure Regulator and the following: Reactor power R0057,SR0061 259002 Reactor Water Level Control X
X K4.13 - Knowledge of Reactor Water Level 3.6 Control design feature(s) and or interlock(s) which provide for the following: FWRV lockup R0059,SR0062 A2.03 - Ability to (a) predict the impacts of 3.7 the following on the Reactor Water Level Control and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of reactor water level input R0060,SR0063 261000 SGTS X
K3.03 - Knowledge of the effect that a loss or 3.4 malfunction of the SGTS will have on the following: Primary containment pressure:
Mark ]&ll R0061,SR0064 262001 AC Electrical Distribution X
K6.03 - Knowledge of the effect that a loss or 3.7 malfunction of the following will have on the AC Electrical Distribution: Generator trip R0074,SR0065 264000 EDGs A3.05 - Ability to monitor automatic operations of the EDGs including: Load f
/
X 3.5 I
I I
I I
I I
I I
I shedding and sequencing R0063,SR0066 KlA Category Point Totals:
2 11 1 2
2I 2 1
3 2
=
Group Point Total:
2 i,
K ES-401BWR SRO Examination OutlineForm ES-401-1 (R8, SI)
Plant Systems - Tier 2/Group 2 System # I Name K
K K
K K
K A
A A
A G
K/A Topic(s)
Imp.
Points 1
2 3
4 5
6 1
2 3
4 201001 CRD Hydraulic X
K6.03 - Knowledge of the effect that a loss or 2.9 1
malfunction of the following will have on the CRD Hydraulic: Plant air systems R0037,SRO067 201006 RWM X
A4.03 - Ability to manually operate and/or 3.0 1
monitor in the control room: Latched group indication R0066,SRO068 202001 Recirculation X
X K3.08 - Knowledge of the effect that a loss 2.9 1
or malfunction of the Recirculation will have on the following: Shutdown cooling system R0067,SRO070 2.2.25 - Knowledge of the bases in technical 37 specifications for limiting conditions for operations and safety limits 204000 RWCU X
2.1.32 - Ability to explain and apply system 3.8 1
limits and precautions R0068,SRO071 219000 RHR/LPCI: Torus/Pool Cooling Mode X
X K2.02 - Knowledge of electrical power 3.3 1
supplies to the following: Pumps R0069,SRO072 A1.03 - Ability to predict and/or monitor 2.9 1
changes in parameters associated with operating the RHR/LPCI: Torus/Pool Cooling Mode controls including: System pressure R0070,SRO073 245000 Main Turbine Gen. and Auxiliaries X
K5.02 - Knowledge of the operational 3.1 1
implications of the following concepts as they apply to the Main Turbine Gen. and Auxiliaries: Turbine operations and limitations 271000 Offgas X
K4.08 - Knowledge of Offgas design 3.3 1
feature(s) and or interlock(s) which provide for the following: Automatic system isolation 272000 Radiation Monitoring X
A3.06 - Ability to monitor automatic 3.4 1
operations of the Radiation Monitoring including: Ventilation system isolation indications X
X K1.01 - Knowledge of the physical connections and/or cause-effect relationships between Control Room FIVAC and the following: Radiation monitors 3.5 1
290003 Control Room HVAC
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400000 Component Cooling Water KIA Category Point Totals:
R0080,SR0077 K6.02 - Knowledge of the effect that a loss or 2.9 malfunction of the following will have on the Control Room HVAC: Component cooling water systems R0081,SR0078 A2.02 - Ability to (a) predict the impacts of 3.0 the following on the Component Cooling Water and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High/low surge tank level R0083,SR0079 Group Point Total:
1 13
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ES-401BWR SRO Examination OutlineForm ES-401-1 (R8, SI)
Plant Systems - Tier 2/Group 3 System # / Name K
K K
K K
K6 Al A2 A3 A4 G
K/A Topic(s)
Imp.
Points 1
2 3
4 5
201003 Control Rod and Drive Mechanism X
X K4.04 - Knowledge of Control Rod and 3.7 1
Drive Mechanism design feature(s) and or interlock(s) which provide for the following: the use of either accumulator or reactor water to SCRAM the control rod R0056,SR080 2.1.14 - Knowledge of system status 13 criteria which require the notification of plant personnel 215001 Traversing In-core Probe X
A1.01 - Ability to predict and/or monitor 2.9 1
changes in parameters associated with operating the Traversing In-core Probe controls including: Radiation levels 233000 Fuel Pool Cooling and Cleanup X
K1.14 - Knowledge of the physical 2.5 1
connections and/or cause-effect relationships between Fuel Pool Cooling and Cleanup and the following: Reactor building ventilation R0086,SRO083 K/A Category Point Totals:
1 0
0 1
0 0
1 0
0 0
1 Group Point Total:
4 Plant-Specific Priorities System / Topic Recommended Replacement for...
Reason Points 295009 AA1.01 295009 AK3.02 Better K/A match to question written 1
295024 2.1.23 295038 EK1.03 Written as RO only but is better SRO only 1
question as written
-I-iI I
i I4I
/
Plant-Specific Priority Total (limit 10):
t
(
Generic Knowledge and Abilities Outline (Tier 3)
Facility:
NMPCU1 Date of Exam:
09/30102 Exam Level: SRO Category K/A #
Topic Imp.
Points 2.1.11 Knowledge of less than one hour technical specification action 8
1 statements for systems SR0084 2.1.6 Ability to supervise and assume a management role during plant 43 1
transients and upset conditions SR0085 Conduct of 2.1.18 Ability to make accurate, clear and concise logs, records, status 3.0 1
Operations boards, and reports R0088,SR0086 2.1.9 Ability to direct personnel activities inside the control room 4.0 1
R0089,SR0087 Total 4
2.2.28 Knowledge of new and spent fuel movement procedures 5
1 2.2.5 Knowledge of the process for making changes in the facility as I 1 described in the safety analysis report SR0089 Equipment 2.2.2 Ability to manipulate the console controls as required to operate 3.5 1
the facility between shutdown and designed power levels Control R0091,SRO090 2.2.24 Ability to analyze the affect of maintenance activities on LCO 3.8 1
status R0092,SRO091 Total 4
2.3.1 Knowledge of10CFR20 and related facility radiation control 3.0 1
requirements SR0092 2.3.9 Knowledge of the process for performing a containment purge 3,4 1
Radiation SR0093 Control 2.3.11 Ability to control radiation releases SR0094 3.2 1
2.3.2 Knowledge of facility ALARA program R0095,SR0095 2.9 1
Total 4
2.4.47 Ability to diagnose and recognize trends in an accurate and 3.7 1
timely manner utilizing the appropriate control room reference material R0094,SR0096 2.4.24 Knowledge of loss of cooling water procedures R0097,SR0097 3.7 1
Emergency 2.4.5 Knowledge of the organization of the operating procedures 3.6 1
Procedures/
network for normal, abnormal and emergency evolutions Plan 2.4.27 Knowledge of fire in the plant procedures R0099,SR0099 3.5 1
2.4.22 Knowledge of the bases for prioritizing safety functions during 4.0 1
abnormal/emergency operations Total 5
Tier 3 Point Total (SRO) 17 ES-401
Facility:
NMPCU1 Date of Exam:
09130102 Exam Level:
RO K/A Category Points Point Tier Group K
K K
K A
A A A G
Total 1
2 3
4 5
6 1
213 4
1 2
3 1
4 2
1 13 Emergency &
2 4
4 3
4 2
2 19 Abnormal 3
0 1
1 1
0 1
4 Plant Tier 6
8 5
9 4
4 36 Evolutions Totals I
I I........
1 3
1 3
3 3
114 3
3 3
1 28
- 2.
2 2
2 2
2 2
2 2
1 1
2 1
19 Plant 3
1 0
0 0
0 1
1 1
0 0
0 4
Systems Tier 6
3 5
5 5
4 7
5 4
5 2
51 Totals Cati1 Cat 2 Cat 3 Cat 4 13
- 3. Generic Knowledge and Abilities Cat11 Cat2 _______C_______
13
__3_
3 3
12 5
Note: 1.
Ensure that at least two topics from every K/A category are sampled within each tier (i.e., the "Tier Totals" in each KIA category shall not be less than two).
- 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 NRC revisions. The final exam must total 100 points.
- 3.
Select topics from many systems; avoid selecting more than two or three K/A topics from a given system unless they relate to plant-specific priorities.
- 4.
Systems/evolutions within each group are identified on the associated outline.
- 5.
The shaded areas are not applicable to the category/tier.
6.*
The generic K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system.
- 7.
On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importance ratings for the SRO license level, and the point totals for each system and category. K/As below 2.5 should be justified on the basis of plant-specific priorities. Enter the tier totals for each category in the table above.
ES-401 BWR RO Examination Outline Form ES-401-2 (R8, $1)
ES-401 BWR RO Examination OutlineForm ES-401-2 (R8, SI)
Emergency and Abnormal Plant Evolutions - Tier I/Group 1 E/APE # Name / Safety Function 1
K K
A A
G K/A Topic(s)
Imp.
Points 1 2 3
1 2
295007 High Reactor Pressure / 3 X
X AK2.05 - Knowledge of the interrelations High Reactor Pressure and 2.9 1
the following: Shutdown cooling ROOO1,SRO002 2.1.33 - Ability to recognize indications for system operating parameters 3.4 1
which are entry-level conditions for technical specifications R0009,SRO001 295009 Low Reactor Water Level 1 2 X
X AK1.05 - Knowledge of the operational applications of the following 3.3 1
concepts as they apply to the Low Reactor Water Level: Natural circulation R0002,SR0003 AA1.01 - Ability to operate and/or monitor the following as they apply to 3.9 1
Low Reactor Water Level: Reactor feedwater R0003,SR0004 295010 High Drywell Pressure / 5 X
AA1.03 - Ability to operate and/or monitor the following as they apply to 2.6 1
High Drywell Pressure: Nitrogen makeup R0004,SR0005 295014 Inadvertent Reactivity Addition/ 1 X
AA1.07 - Ability to operate and/or monitor the following as they apply to 4.0 1
Inadvertent Reactivity Addition: Cold water injection R0005,SR0008 295015 Incomplete SCRAM / I X
AK2.04 - Knowledge of the interrelations between Incomplete SCRAM 4.0 1
and the following: RPS R0006,SR0009 AK3.01 - Knowledge of the reasons for the following responses as they 3.4 1
apply to Incomplete SCRAM: Bypassing rod insertion blocks R00007,SRO010 295024 High Drywell Pressure / 5 X
EA2.01 - Ability to determine and interpret the following as they apply to 4.2 1
High Drywell Pressure: Drywell pressure R0008,SRO017 295025 High Reactor Pressure / 3 X
EA2.01 - Ability to determine and interpret the following as they apply to 4.,
1 High Reactor Pressure: Reactor pressure RO010 295037 SCRAM Condition Present and Power X
X EK2.13 - Knowledge of the interrelations between SCRAM Condition 3.4 1
Above APRM Downscale or Unknown / I Present and Power Above APRM Downscale or Unknown and the following: Alternate boron injection methods RO011,SR0022 EA1.01 - Ability to operate and/or monitor the following as they apply to 4.6 1
SCRAM Condition Present and Power Above APRM Downscale or Unknown: RPS R0012,SR0023 500000 High Containment Hydrogen Conc. / 5 X
EK1.01 - Knowledge of the operational applications of the following 3.3 1
concepts as they apply to the High Containment Hydrogen Conc:
Containment integrity R0013,SR0026 K/A Category Totals:
2 f 3 i1 4
I2 1 I Group Point Total:
7 13
(
o7
(
(
ES-401 BWR RO Examination OutlineForm ES-401-2 (R8, SI)
Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 E/APE # / Name / Safety Function K
K K
A A
G K/A Topic(s)
Imp.
Points 1
2 3
1 2
295001 Partial or Complete Loss of Forced Core X
X AA1.05 - Ability to operate and/or monitor the following as they apply to 3.3 1
Flow Circulation / 1 & 4 Partial or Complete Loss of Forced Core Flow Circulation:
Recirculation flow control RO014,SR0028 AA2.02 - Ability to determine and interpret the following as they apply to Partial or Complete Loss of Forced Core Flow Circulation: Neutron 31 1
monitoring R0015 295002 Loss of Main Condenser Vacuum / 3 X
AK3.02 - Knowledge of the reasons for the following responses as they 3.4 1
apply to Loss of Main Condenser Vacuum: Turbine trip R0016,SR0029 295004 Partial or Complete Loss of DC Pwr / 6 X
X AK1.04 - Knowledge of the operational applications of the following 2.8 1
concepts as they apply to the Partial or Total Loss of DC Pwr: Effect on battery discharge rate AK2.01 - Knowledge of the interrelations between Partial or Total Loss 3.1 1
of DC Pwr and the following: Battery charger 295012 High Drywell Temperature / 5 X
AK1.02 - Knowledge of the operational applications of the following 3.1 1
concepts as they apply to the High Drywell Temperature: Reactor power level control R0019,SR0032 295013 High Suppression Pool Temp. / 5 X
X AK1.04 - Knowledge of the operational applications of the following 2.9 1
concepts as they apply to the High Suppression Pool Temp: Complete condensation R0020,SR0006 AK2.01 - Knowledge of the interrelations between High Suppression 3.6 1
Pool Temp and the following: Suppression pool cooling R0021,SRO007 295016 Control Room Abandonment / 7 X
X AA1.03 - Ability to operate and/or monitor the following as they apply to 3.0 1
Control Room Abandonment: RPIS R0022,SR0012 AA2.02 - Ability to determine and interpret the following as they apply to 4.2 1
Control Room Abandonment: Reactor water level 295018 Partial or Complete Loss of CCW / 8 X
AA1.03 - Ability to operate and/or monitor the following as they apply to 3.3 1
Partial or Total Loss of CCW: Affected systems so as to isolate damaged portions R0023,SR0034 295022 Loss of CRD Pumps / I X
AK2.07 - Knowledge of the interrelations between Loss of CRD Pumps 3.4 1
and the following: Reactor pressure (scram assist) R0024,SR0035 295026 High Suppression Pool Water Temp. / 5 X
X EK2.03 - Knowledge of the interrelations between Suppression Pool 3.2 1
High Water Temp and the following: Suppression chamber pressure 3.5 1
EK3.03 - Knowledge of the reasons for the following responses as they apply to Suppression Pool High Water Temp: Suppression pool spray 295028 High Drywell Temperature / 5 X
EK3.04 - Knowledge of the reasons for the following responses as they apply to High Drywell Temperature: Increased drywell cooling 3.6 I
(
295029 High Suppression Pool Water Level / 5 X
2.1.30 Ability to locate and operate components/including local controls 3M9 1
295033 High Sec. Cont. Area Rad. Levels / 9 X
EK1.02 - Knowledge of the operational applications of the following 3.9 1
concepts as they apply to the High Secondary Containment Area Radiation Levels: Personnel protection 295038 High Off-site Release Rate 19 X
2.3.9 - Knowledge of the process for performing a containment purge 2.5 1
600000 Plant Fire On Site / 8 X
AAI.05 - Ability to operate and/or monitor the following as they apply to 3.0 1
Plant Fire On Site: Plant and control room ventilation systems K/A Category Point Totals:
I4 I4 I3 I4 I2 I2 J Group Point Total:
1 19
(
(
ES-401 BWR RO Examination OutlineForm ES-401-2 (R8, SI)
Emergency and Abnormal Plant Evolutions - Tier 1 /Group 3 E/APE # / Name / Safety Function K
K K
A A
G K/A Topic(s)
Imp.
Points 1
2 3
1 2
295021 Loss of Shutdown Cooling / 4 X
AK3.01 - Knowledqe of the reasons for the following responses as they 3
1 apply to Loss of Shutdown Cooling: raising reactor water level 295023 Refueling Accidents / 8 X
AAI.06 - Ability to operate and/or monitor the following as they apply to 3.3 1
Refueling Accidents: Neutron monitoring R0034,SRO016 295036 Secondary Containment High X
X EK2.02 - Knowledge of the interrelations between Secondary 2.6 1
Sump/Area Water Level / 5 Containment High Sump/Area Water Level and the following: Post accident sampling system 2.4.50 - Ability to verify system alarm setpoints and operate controls 1
identified in the alarm response manual K/A Category Point Totals:
°
]0 1 1 [I1
°0 [i 1 Group Point Total:
4
(
(
(
ES-401 BWR RO Examination OutlineForm ES-401-2 (R8, SI)
Plant Systems - Tier 2/Group 1 System # / Name K
K K
K K
K A
A A
A G
K/A Topic(s)
Imp.
Points 1
2 3
4 5
6 1
2 3
4 201001 CRD Hydraulic X
K6.03 - Knowledge of the effect that a loss or 3.0 1
malfunction of the following will have on the CRD Hydraulic: Plant air systems R0037,SR0067 202002 Recirculation Flow Control X
K5.02 - Knowledge of the operational 2A implications of the following concepts as they apply to the Recirculation Flow Control:
Feedback signals R0038 206000 HPCI X
X A2.05 - Ability to (a) predict the impacts of 3.5 the following on the HPCI and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: DC failures R0039,SR0044 A3.07 - Ability to monitor automatic 3.9 operations of the HPCI including: Lights and alarms R0040,SR0045 209001 LPCS X
X A1.03 - Ability to predict and/or monitor 3.8 changes in parameters associated with operating the LPCS controls including:
Reactor water level R0041,SR0046 A4.09 - Ability to manually operate and/or 3.6 monitor in the control room: Suppression pool level R0042,SR0047 212000 RPS X
A1.09 - Ability to predict and/or monitor 2.7 changes in parameters associated with operating the RPS controls including:
Individual relay status R0043 215004 SRM X
X K4.06 - Knowledge of Source Range 3.2 Monitor design feature(s) and or interlock(s) which provide for the following: IRM/SRM interlock R0044,SR0049 K5.03 - Knowledge of the operational 2.8 implications of the following concepts as they apply to the Source Range Monitor:
Changing detector position R0045,SRO050 X
X K2.02 - Knowledge of electrical power supplies to the following: APRM channels R0046,SR0051 A1.03 - Ability to predict and/or monitor changes in parameters associated with operating the APRMILPRM controls 2.6 3.6 1
(
(
including: Control rod block status R0047,SR0052 216000 Nuclear Boiler Instrumentation X
X K3.30 - Knowledge of the effect that a loss 3,2 or malfunction of the Nuclear Boiler Instrumentation will have on the following:
Recirculation system R0048 2.4.10 - Knowledge of annunciator response 3.0 procedures 218000 ADS X
X K5.01 - Knowledge of the operational 1
implications of the following concepts as they apply to the ADS: ADS logic operation R0051 A3.03 - Ability to monitor automatic operations of the ADS including: ADS valve acoustical monitor noise R0050,SR0053 223001 Primary CTMT and Auxiliaries X
X K1.10 - Knowledge of the physical 3.0 connections and/or cause-effect relationships between Primary CTMT and Auxiliaries and the following: Plant air systems R0052,SR0054 K4.05 - Knowledge of Primary CTMT and Auxiliaries design feature(s) and or 2*
interlock(s) which provide for the following:
Maintains proper suppression pool to drywell differential pressure 223002 PClS/Nuclear Steam Supply Shutoff X
A4.01 - Ability to manually operate and/or monitor in the control room: Valve closures R0054 239002 SRVs X
K3.01 - Knowledge of the effect that a loss or 3.9 malfunction of the SRVs will have on the following: Reactor pressure control A1.01 - Ability to predict and/or monitor 3.3 changes in parameters associated with operating the SRVs controls including: Tail pipe temperature 241000 Reactor/Turbine Pressure Regulator X
K1.01 - Knowledge of the physical 3.8 X
connections and/or cause-effect relationships between Reactor/Turbine Pressure Regulator and the following: Reactor power R0057,SR0061 K1.09 - Knowledge of the physical
- 3.
connections and/or cause-effect relationships between Reactor/Turbine Pressure Regulator and the following: Combined intermediate valves R0058 259002 Reactor Water Level Control X
X K4.13 - Knowledge of Reactor Water Level Control design feature(s) and or interlock(s) 3.5 I
(
\\
(
which provide for the following: FWRV lockup R0059,SR0062 3.6 A2.03 - Ability to (a) predict the impacts of the following on the Reactor Water Level Control and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of reactor water level input R0060,SR0063 261000 SGTS X
K3.03 - Knowledge of the effect that a loss or 3.2 malfunction of the SGTS will have on the following: Primary containment pressure:
Mark I&Il R0061,SR0064 A4.09 - Ability to manually operate and/or monitor in the control room: Ventilation valves/dampers R0062 264000 EDGs X
A2.06 - Ability to (a) predict the impacts of P'A 1
the following on the EDGs and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
opening normal and/or alternate power to emergency bus R0064 A3.05 - Ability to monitor automatic 3.4 1
operations of the EDGs including: Load shedding and sequencing R0063,SR0066 K/A Category Point Totals:
]3 i1 13 3 J3 11 41 1 1
]3 I3 H I Group PointTotal:
28
(
(
(
ES-401 BWR RO Examination OutlineForm ES-401-2 (R8, SI)
Plant Systems - Tier 2/Group 2 System # / Name K
K K
K K
K A
A A
A G
K/A Topic(s)
Imp.
Points 1
2 3
4 5
6 1
2 3
4 201003 Control Rod and Drive Mechanism X
K4.04 - Knowledge of Control Rod and 3.6 1
Drive Mechanism design feature(s) and or interlock(s) which provide for the following:
the use of either accumulator or reactor water to SCRAM the control rod R0065,SR080 201006 RWM X
A4.03 - Ability to manually operate and/or 3.0 monitor in the control room: Latched group indication R0066,SR0068 202001 Recirculation X
K3.08 - Knowledge of the effect that a loss 2.8 or malfunction of the Recirculation will have on the following: Shutdown cooling system R0067,SR0070 204000 RWCU X
2.1.32 - Ability to explain and apply system 3.4 1
limits and precautions R0068,SR0071 219000 RHR/LPCI: Torus/Pool Cooling Mode X
X K2.02 - Knowledge of electrical power 3.1 1
supplies to the following: Pumps R0069,SR0072 A1.03 - Ability to predict and/or monitor 2.9 changes in parameters associated with operating the RHR/LPCI: Torus/Pool Cooling Mode controls including: System pressure R0070,SR0073 226001 RHR/LPCI: CTMT Spray Mode X
K5.06 - Knowledge of the operational 2.6 implications of the following concepts as they apply to the RHR/LPCI: CTMT Spray Mode:
Vacuum breaker operation 245000 Main Turbine Gen. and Auxiliaries X
K5.02 - Knowledge of the operational 2.8 implications of the following concepts as they apply to the Main Turbine Gen. and Auxiliaries: Turbine operations and limitations 256000 Reactor Condensate X
AI.03 - Ability to predict and/or monitor Z8 changes in parameters associated with operating the Reactor Condensate controls including: System pressure R0073 262001 AC Electrical Distribution X
K6.03 - Knowledge of the effect that a loss or 3.5 malfunction of the following will have on the AC Electrical Distribution: Generator trip R0074,SR0065
(
263000 DC Electrical Distribution I
X K11.01 - Knowledge of the physical
(
(
connections and/or cause-effect relationships between DC Electrical Distribution and the following: AC electrical distribution R0075 271000 Offgas X
K4.08 - Knowledge of Offgas design 3.1 feature(s) and or interlock(s) which provide for the following: Automatic system isolation 272000 Radiation Monitoring X
A3.06 - Ability to monitor automatic 3.4 operations of the Radiation Monitoring including: Ventilation system isolation indications 286000 Fire Protection X
K2.02 - Knowledge of electrical power 2ZS 1
supplies to the following: Pumps 290001 Secondary CTMT X
A4.02 - Ability to manually operate and/or 33 1
monitor in the control room: Reactor building area temperatures 290003 Control Room HVAC X
X K1.01 - Knowledge of the physical 3.4 connections and/or cause-effect relationships between Control Room HVAC and the following: Radiation monitors R0080,SR0077 2.7 K6.02 - Knowledge of the effect that a loss or malfunction of the following will have on the Control Room HVAC: Component cooling wa%5sstems R0081,SR0078 300000 Instrument Air X
K3.PY1 - Knowledge of the effect that a loss or malfunction of the Instrument Air will have on the following: Containment air system 400000 Component Cooling Water X
A2.02 - Ability to (a) predict the impacts of 2.8 the following on the Component Cooling Water and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High/low surge tank level R0083,SR0079 K/A Category Point Totals:
I2
[2 2
2 [ 2 2
2 2 1111 2 I Group Point Total:
19 io A/Af61
~a~/(5 F
ES-401 BWR RO Examination OutlineForm ES-401-2 (R8, SI)
Plant Systems - Tier 2/Group 3 System # / Name K
K K
K K
K A
A A
A G
K/A Topic(s)
Imp.
Points 2
3 4
5 6
1 2
3 4
215001 Traversing In-core Probe X
X K6.04 - Knowledge of the effect that a loss or 3.1 malfunction of the following will have on the Traversing In-core Probe: primary containment isolation system R0085 A1.01 - Ability to predict and/or monitor changes in parameters associated with 2.8 operating the Traversing In-core Probe controls including: Radiation levels 233000 Fuel Pool Cooling and Cleanup X
X K1.14 - Knowledge of the physical 2.5 connections and/or cause-effect relationships between Fuel Pool Cooling and Cleanup and the following: Reactor building ventilation R0086,SRO083 A2.1 5 - Ability to (a) predict the impacts of 2a I
the following on the Fuel Pool Cooling and Cleanup and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High system temperature R0087 K/A Category Point Totals:
[1 0
0 0
0 1
W1 [1 0
0 J0 Group Point Total:
4 System / Topic Recommended Replacement for...
Reason 295009 AA1.01 295009 AK3.02 Better K/A match to question written 295030 AA2.02 295030 EK3.02 Was SRO only K/A added to as a common 290001 A4.02 290001 A4.06 NMPCU1 has a Reactor building not a Fuel building swapped to applicable K/A t
+
Li gL~,/'CIA
(
(
Plant-Specific Priorities
(
Plant-Specific Priority Total: (limit 10)
(
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Facility:
NMPCU1 Date of Exam:
09/30102 Exam Level: RO Category K/A #
Topic Imp.
Points 2.1.18 Ability to make accurate, clear and concise logs, records, status 2.9 1
boards, and reports R0088,SR0086 2.1.9 Ability to direct personnel activities inside the control room 2.5 1
Conduct of R0089,SR0087 Operations 2.1.31 Ability to locate control room switches, controls and indications 442 1
and to determine that they are correctly reflecting the desired plant lineup Total 3
2.2.2 Ability to manipulate the console controls as required to operate 4.0 1
the facility between shutdown and designed power levels R0091,SRO090 2.2.24 Ability to analyze the affect of maintenance activities on LCO 2.6 1
status R0092,SRO091 Equipment 2.2.33 Knowledge of control rod programming 2.5 1
Control Total 3
2.3.2 Knowledge of facility ALARA program R0095,SR0095 2.5 1
2.3.4 Knowledge of radiation exposure limits and contamination 2.5 1
control/including permissible levels in excess of those authorized Total 2
2.4.47 Ability to diagnose and recognize trends in an accurate and 3.4 1
timely manner utilizing the appropriate control room reference material R0094,SR0096 2.4.24 Knowledge of loss of cooling water procedures R0097,SR0097 3.3 1
Emergency 2.4.5 Knowledge of the organization of the operating procedures 2.9 1
Procedures/
network for normal, abnormal and emergency evolutions Plan 2.4.27 Knowledge of fire in the plant procedures R0099,SR0099 3.0 1
2.4.22 Knowledge of the bases for prioritizing safety functions during 3.0 1
abnormal/emergency operations Total 5
Tier 3 Point Total (RO) 13
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Nine Mile Point # I Date of Examination: 09/30/2002 Examination Level (circle one): SRO Operating Test Number: Cat A Test A Administrative Topic/Subject Describe method of evaluation:
Description
- 2. TWO Administrative Questions A. ]
Reactor Plant 2.2.9, 2.2.36 Startup JPM - During a Plant startup, given a completed "Reactivity Maneuver Form" Requirements and some requested and required changes, perform the administrative duties required for the SSS.
References:
GAP-OPS-05 Fuel Handling 2.1.8, 2.2.31, 2.4.40 JPM - Perform SRO administrative actions with a report of a "dropped" fuel bundle during a Refueling.
References:
N I-SOP-12 A.2 Surveillance 2.1.14, 2.2.12 Testing JPM - Given a completed Surveillance Test, NI-ST-MIA, Liquid Poison Pump
- l 1 Operability Test, complete the "Acceptance Criteria" and "SSS Review" sections as applicable.
References:
N I-ST-M I A, Technical Specifications A.3 Control of Releases 2.3.6 Q. I. Given the following:
A portable sump pump has been approved for the removal of accumulated water The sump pump discharge is aligned to the Radwaste system After receiving permission from the SSS, an operator starts the sump pump and begins the discharge Two minutes later, it is discovered that the sump pump is discharging to the storm sewer system due to a valve lineup error The operator turns off the sump pump
- a. Based on the above conditions, what actions/notifications are required?
- b. What actions must be taken prior to resuming the discharge?
References:
GAP-ENV-01 Section 3.1
9 9
Radiation Work 2.3.10 Permit Q.2. The plant is at rated power, with the following:
"* TIP traces are in progress for LPRM calibration
"* As TIP detector #3 is being withdrawn from the core, the drive mechanism fails with the detector in the TIP Room but not yet in its chamber shield
"* The SSS has directed that the detector be withdrawn to the chamber shield using the manual handcrank, if possible
"* The SSS has also directed a brief entry into the TIP Room to ensure no mechanical interference exists in the associated TIP drive tube
"* TIP Room Area Radiation Monitors are in an alarmed condition
- a. What individual(s) is(are) responsible for approving entry into the TIP Room?
- b. Identify all areas where the operator can get a key that will grant access to the TIP Room?
References:
GAP-RPP-08 Sections 3.2, 3.4, and 3.5; GAP-RPP-02 Section 3.7 A.4 Emergency Action 2.4.29, 2.4.41 Levels and JPM - Emergency Plan Classification.
Classifications
References:
r~~i~~ni
~Adm~inistrative Tonics OutlineFomE-0l Facility: Nine Mile Point # I Date of Examination: 09/30/2002 Examination Level (circle one): SRO Operating Test Number: Cat A Test B Administrative Topic/Subject Describe method of evaluation:
Description I. ONE Administrative JPM, OR
- 2. TWO Administrative Questions A. 1 Reactor Plant 2.2.9, 2.2.36 Startup JPM - During a Plant startup, given a completed "Reactivity Maneuver Form" Requirements and some requested and required changes, perform the administrative duties required for the SSS.
References:
GAP-OPS-05 Fuel Handling 2.1.8, 2.2.31, 2.4.40 JPM - Perform SRO administrative actions with a report of a "dropped" fuel bundle during a Refueling.
References:
N I-SOP-12 A.2 Surveillance 2.1.14, 2.2.12 Testing JPM - Given a completed Surveillance Test, NI-ST-M1A, Liquid Poison Pump
- 11 Operability Test, complete the "Acceptance Criteria" and "SSS Review" sections as applicable.
References:
N I -ST-M I A, Technical Specifications A.3 Radiation Work 2.3.10 Permit Q.I. The plant has experience an emergency, with the following:
"* Access to a posted Locked High Radiation Area is required to respond to the emergency
"* A Radiation Work Permit is currently not available for the area
"* The operator assigned to enter the area by the SSS has an authorized delta exposure of 250 mRem
- a. What support is the SSS required to direct as part of this task?
- b. What conditions would be necessary to use the master keys in the "break-to-enter" key box in the Control Room to access the Locked High Radiation Area?
References:
GAP-RPP-08 Sections 3.2, 3.3, and 3.5; GAP-RPP-02 Section 3.2 Form ES-3 0 1-1
A.3 Control of Releases 2.3.6 Q.2. Given the following:
"* A large amount of non-radioactive water has accumulated in the Circulating Water Pumphouse
"* Operations Management has decided to discharge this water to Lake Ontario
"* Operations personnel have begun preparing the Work Order for this discharge
"* During your review of the Work Order, you notice that the Chemistry Analysis will expire during the planned discharge
- a. Should the discharge be permitted?
- b. If a Work Order for a planned discharge has been reviewed and found acceptable, what individual(s) is(are) responsible for approving the Work Order?
References:
GAP-ENV-01, Sections 2.5 and 3.1; GAP-PSH-01 A.4 Emergency Action 2.4.29, 2.4.41 Levels and Classifications JPM -Emergency Plan Classification.(
References:
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Nine Mile Point # I Date of Examination: 09/30/2002 Examination Level (circle one): RO Operating Test Number: Cat A Test A Administrative Topic/Subject Describe method of evaluation:
Description
- 2. TWO Administrative Questions A. I Plant Parameter 2.1.2, 2.1.9 Verification Q.1. A plant startup is in progress, with the following:
- No control rods have been withdrawn yet
- An RO has been assigned as the Additional Qualified Individual The SSS has directed control rod withdrawal to achieve criticality
- a. Prior to the movement of the first in-sequence control rod, what communications are required between the ATC RO and the Additional Qualified Individual?
- b. How does the Additional Qualified Individual denote verification of control rod movements?
References:
GAP-OPS-05 Section 3.4 and ATTACHMENT 1 2.1.2, 2.1.9 Q.2. The plant is at rated power, with the following:
Weekly control rod testing is in progress in accordance with N 1-ST-WI, Control Rod Exercising Operability Test An additional RO has been assigned as the Additional Qualified Individual
- a. What is the minimum number of "short breaks" that must be taken by the ATC RO and Additional Qualified Individual during the performance of N 1-ST-WI ?
- b. Technical Specifications requires weekly performance of N 1-ST-W1.
What other plant conditions require the performance of Nl-ST-WI and what is the frequency of performance when these conditions exist?
References:
GAP-OPS-05 Section 3.4.5; N I-ST-WI
A.l OATC Duties 2.1.1, 2.1.2 Q.l. The plant has experienced a hydraulic Anticipated Transient Without Scram (ATWS), with the following:
"* The CSO has left the Control Room to supervise alternate control rod insertion activities
"* An unexpected annunciator alarms on a Main Control Room back panel
- b. What are the responsibilities of the person responding to this alarm?
References:
Operations Manual Sections 3.3.5 and 4.1 2.1.1, 2.1.9 Q.2. Given the following:
Panel A l Annunciator 3-4, "CONDENSER VACUUM BELOW 24" HG" alarms The ATC RO verifies Main Condenser vacuum is NOT degrading The CRS contacts I&C to troubleshoot the alarm
"* I&C reports that transmitter XS-LVA-1 has failed The SSS directs I&C to defeat the XS-LVA-1 input signal to Annunciator 3-4
- a. What actions must be performed by Operations Branch personnel when l&C reports that the input signal from XS-LVA-1 has been defeated?
- b. What actions, if any, must be performed by Operations Branch personnel if the input signal from transmitter XS-LVA-2 is also defeated?
References:
GAP-OPS-01 Section 3.10.7; N I-ARP-A I Window 3-4 A.2 Surveillance 2.2.12 Testing Q.1. Given the following:
0 45 minutes into the performance of NI-ST-M4A, Emergency Diesel Generator 102 And PB 102 Operability Test, the CRS directed the shutdown of EDG 102 due to rising Service Water and Lube Oil temperatures All steps in N 1-ST-M4A were completed satisfactorily prior to the shutdown of EDG 102 a
The SSS has directed the restoration of EDG 102 to its normal standby lineup
- a. Would EDG 102 be considered OPERABLE or INOPERABLE?
- b. What is the status of N1-ST-M4A?
References:
GAP-SAT-0I Section 3.6.4; N I -ST-M4A A.1I OATC Duties 2.1.1,2.1.2
A.2 Surveillance 2.2.12 Testing Q.2. Given the following:
"* NI-ST-C13, Reactor Shutdown Cooling System Valve Leakage Test, is being performed Corrected leakage rate values for applicable valves are as follows:
38-02 (step 8.1.27):
4.2 gpm 38-01 (step 8.1.28):
4.0 gpm 38-12 (step 8.3.26):
3.9 gpm 38-13 (step 8.3.27):
3.9 gpm 38-13 (step 8.4.19):
3.7 gpm 38-216 (step 8.1.16f):
4.2 gpm 38-216 (step 8.2.22):
0.35 gpm 38-01 (step 8.2.14):
3.5 gpm
- a. Based on the above conditions, do any of the above valves need to be declared INOPERABLE?
- b. Based on the above conditions, is total leakage through Reactor Shutdown Cooling system valves acceptable?
References:
N I -ST-C 13 A.3 Radiation Work 2.3.10 Permit Q. 1. The plant has experience an emergency, with the following:
"* Access to a posted Locked High Radiation Area is required to respond to the emergency
"* A Radiation Work Permit is currently not available for the area
"* The operator assigned to enter the area by the SSS has an authorized delta exposure of 250 mRem
- a. What support from the Radiation Protection Department does the operator require to enter the area?
- b. Identify all areas where the operator can get a key to gain access to the area?
References:
GAP-RPP-08 Sections 3.2, 3.3, and 3.5; GAP-RPP-02 Section 3.2 2.3.10 Q.2 Given the following:
A Reactor Operator has just returned from an assignment to provide outage support at Calvert Cliffs The RO received an exposure of 150 mRem TEDE during this assignment The exposure of the RO at NMPI prior to the assignment was 860 mRem TEDE
- a. Based on the above information, when would the RO be required to receive a dose extension?
- b. What individual(s) would be responsible for approving the dose extension?
References:
GAP-RPP-07 Sections 3.2
Emergency Plan 2.4.39 JPM - Search and Rescue, during Plant fire with a reported missing person
References:
EPIP-EPP-03 ATTACHMENT 1 A.4
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Nine Mile Point # I Date of Examination: 09/30/2002 Examination Level (circle one): RO Operating Test Number: Cat A Test B Administrative Topic/Subject Describe method of evaluation:
Description
- 2. TWO Administrative Questions A. 1 Plant Startup 2.1.1 Activities Q. 1. Preparations are in progress to perform a plant startup, with the following:
"* NI-OP-43A, Plant Startup, has been entered
"* No actions have yet been performed in NI-OP-43A
- a. What evolution(s) in NI-OP-43A require a briefing by the Senior Manager (or a higher level manager) prior to their performance?
- b. Prior to reactor power exceeding 25%, a "Heightened Level Of Awareness" (HLA) briefing is required. What plant risk necessitates this briefing?
References:
N I-OP-43A Attachment 15; GAP-SAT-03 Section 3.4; GAP-OPS-05 Section 3.2 Plant Parameter 2.1.2 Verification Q.2. A plant startup is in progress in accordance with NI-OP-43A, Plant Startup, with the following:
"* An additional RO has been assigned as the Additional Qualified Individual
"* The reactor is critical
"* The Main Turbine is at rated speed and the Main Generator has just been synchronized to the grid
"* Reactor Engineering has directed the continuation of power ascension by raising Reactor Recirculation flow
- a. Prior to raising reactor power, what communications are required between the ATC RO and the Additional Qualified Individual?
References:
GAP-OPS-05 Section 3.5; NI-OP-43A
OATC Duties A.I I _______________________________________________
2.1.2,2.1.17, 2.1.20 Q.I. The plant is at rated power, with the following:
"* A small steam leak develops in the Drywell, resulting in a rise in Drywell pressure
"* Operator actions are unable to halt the rising Drywell pressure condition
"* The CRS directs the ATC RO to place the Reactor Mode Switch in the "SHUTDOWN" position
"* All control rod fully insert
- a. What information should be reported by the ATC RO to the CRS following the placement of the Reactor Mode Switch in the "SHUTDOWN" position?
- b. Based on the above conditions, what Immediate Operator Actions, if any, are required?
References:
Operations Manual Section 3.4; NI-SOP-I 2.1.1, 2.1.9 Q.2. The plant is at rated power, with the following:
"a Panel F2 Annunciator 3-2, "MAIN STM LINE BREAK AREA TEMP HIGH" alarms
"* Operations personnel investigating the alarm could detect no steam leakage
"* The CRS contacts I&C to troubleshoot the alarm
"* I&C reports that temperature element TE-01-50 has failed upscale
"* The SSS directs I&C to defeat the TE-01-50 input signal to Annunciator 3-2
- a. What actions must be performed by Operations Branch personnel when l&C reports that the input signal from TE-01-50 has been defeated?
- b. What actions, if any, must be performed by Operations Branch personnel if the input signal from temperature element TE-01-51 is also defeated?
References:
GAP-OPS-01 Section 3.10.7, N I-ARP-F2 Window 3-2
A.2 Surveillance 2.2.12 Testing Q.]. Given the following:
"* N1-ST-MIA, Liquid Poison Pump 11 Operability Test, is being performed
"* Liquid Poison Storage Tank level is 1,550 gallons
"* Following the start of Liquid Poison Pump 11, its discharge relief valve opens
"* Liquid Poison Pump 11 was shutdown and its relief valve re-closed
- a. Based on the above conditions, what actions are required?
- b. When the test is resumed, Liquid Poison Pump 11 is operated for 6 minutes and then shutdown. Liquid Poison Tank level following pump shutdown is 1,361 gallons. Based on these conditions, what actions, if any, are required?
References:
N I -ST-M 1 A 2.2.12 Q.2. Given the following:
"* Repairs have just been completed on Emergency Service Water Pump 11
"* Applicable sections of NI-ST-Q13, Emergency Service Water Pump Operability Test, are being performed to verify pump operability
"* The running current value recorded in Step 8.1.7.c for Emergency Service Water Pump II is 135 amps
- a. Based on the above conditions, what actions are required?
- b. Would these results require SSS notification of an INOPERABLE component?
References:
N I -ST-Q 13 A.3 Radiation Work 2.3.10 Permit Q.]. Given the following:
"* A Reactor Operator has been dispatched to the Reactor Building to evaluate a small leak in the discharge header of one of the Core Spray pumps
"* While investigating the leak, the Electronic Dosimeter worn by the RO falls off and can not be recovered
"* Following the fall, the Electronic Dosimeter can be heard to be alarming
- a. Based on the above conditions, what must the RO do?
- b. How often, and when, should the RO expect to receive information related to his/her exposure?
References:
GAP-RPP-07, Sections 3.5 and 3.9
A.3 Radiation Work 2.3.10 Permit Q.2 The plant is at rated power, with the following:
"* TIP traces are in progress for LPRM calibration
"* As TIP detector #3 is being withdrawn from the core, the drive mechanism fails with the detector in the TIP Room but not yet in its chamber shield
"* The SSS has directed that the detector be withdrawn to the chamber shield using the manual handcrank, if possible
"* The SSS has also directed a brief entry into the TIP Room to ensure no mechanical interference exists in the associated TIP drive tube
"* TIP Room Area Radiation Monitors are in an alarmed condition
- a. What support from the Radiation Protection Department is required to enter the TIP Room?
- b. Identify where the operator can get a key to gain access to the TIP Room?
References:
GAP-RPP-08 Sections 3.2, 3.3, and 3.5; GAP-RPP-02 Section 3.7 A.4 Emergency Plan 2.4.3 97iP'f JPM -Se ire t,
14L,-s2ý4 h a r
pvi ied missing person
References:
EPIP-EPP-9, ATTACHMENT I (Alternate Path)
Facility:
Nine Mile Point # I Date of Examination:
9/30/02 Exam Level (circle one): RO & SRO Operating Test No.:
B.1 Control Room Systems System / JPM Title Type Code*
Safety Function JPM 11 Torus Cooling Mode/ Torus Cooling Using Containment Spray Pumps Ill and 112. (Alternate Path)
N/A/S 5
K/A 219000 A4.02 3.7/3.5;Task 2000190501; N I -EOP-I Attachment 16 JPM 12 A. C. Electrical Distribution/ Respond to PB 102/103 Bus Under Voltage Low Annunciator (PRA) (01-OPS-SJE-264-1-03)
D/S 6
K/A 262001 A2.05 2.9/3.3 ; Task 3449020401; NI-ARP-A4 (1-6 & 4-2)
JPM 13 Control Rod Drive Hydraulic System/ Control Rod Coupling Check (Alternate Path).
M/A/S/L K/A 201003 A2.02 3.7/3.8; Task 2010060201; N I-OP-5 Section H.9.0 JPM 14 Plant Ventilation Systems/ Manual Start of RB EMER Ventilation. (Alternate Path)
N/A/S 9
K/A 288000 A4.01 3.1/2.9; Task 2610030101; NI-OP-10 Section H. 1.0 JPM 15 APRM/LPRM/ Bypass LPRM (28-25B) Input to APRM 18.
(Alternate Path)
N/A/S 7
K/A 215005 A4.04 3.2/3.2; Task 2159060401; NI-OP-38C Section H.3.0 JPM 16 Reactor Water Level Control System/ Return Vessel Level Control to Automatic Following Maintenance on a Level Column.
N/S 2
K/A 259002 A4.03 3.8/3.6; Task 2590040101; N1-OP-16 Section F.10 JPM 17 Main Turbine Generator and Auxiliary Systems/ Place Main Generator Amplidyne Back in Service and Adjust Vars.
N/S 4
K/A 245000 A4.02 3.1/2.9; Task 2450090101 ; N I -OP-32 Section H.6.0 B.2 Facility Walk-Through JPM 18 Isolation (Emergency) Condenser/ Supply Lake Water to EC Makeup Tanks using the Electric Fire Pump (01-OPS-PJE-200-1-73)
K/A 207000 A1.09 3.7/3.7; Task 2009140504; N I-OP-21A Section 3.0 M/R 4
JPM 19 Reactor Feedwater System/ Lineup to Flood Reactor Vessel using the Fire System. (01-OPS-PJE-200-1-66)
D/R 2
K/A 295031 EAI.12 3.9/4.1; Task 2009130504; NI-EOP-1 Attachment 19 JPM 20 Fire Protection System/ Manually Initiate Cardox Flooding of Aux. Control Room. (01-OPS-PJE-286-1-07).
K/A 286000 A1.05 3.2/3.2; Task 2869110101; NI-OP-21C Section H.2 D
8
- Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)lternate path, (C)ontrol room, (S)imulator, (L)ow-Power, (R)CA
Facility:
Nine Mile Point # I Date of Examination:
9/30/02 Exam Level (circle one): RO & SRO Operating Test No.:
B.1 Control Room Systems System / JPM Title Type Code*
Safety Function JPM I Torus Cooling Mode/ Transfer Torus Water to the Waste Collector Tank. (01-OPS-SJE-200-1-04)
K/A 219000 A4.12 3.9/3.8;Task 2269090401/2000230501; D/S 5
N 1-EOP-1 Attachment 15 JPM 2 A. C. Electrical Distribution/ Shift source of power for PB 101 from R1014 to R101 1.
N/S 6
K/A 262001 A4.01 3.4/3.7; Task 2620020101; NI-OP-30 Section H.10.0 JPM 3 Control Rod Drive Hydraulic System/ Switching CRD Pumps (Alternate Path).
N/A/S K/A 201001 A4.01 3.1/3.1; Task 2010020101; N1-OP-5 Section F.3.0 JPM 4 Plant Ventilation Systems/ Shift Reactor Building Operating Exhaust and Supply fans from # I I's to # 12's. (Alternate Path)
N/A/S 9
K/A 288000 A4.01 3.1/2.9; Task 2880040101; NI-OP-10 Section F.1.0 and F.2.0 JPM 5 Traversing Incore Probe/ Secure TIP on receipt of Containment Isolation. (Alternate Path)
N/A/S 7
K/A 215001 A4.03 3.0/3.1; Task 2159090401; NI-OP-39 Section H.1.0 JPM 6 Reactor Feedwater System/ Change operating Motor Driven Feedwater Pumps at power.
N/S 2
K/A 259001 A4.02 3.9/3.7; Task 2590040101; NI-OP-16 Section F.2.0 JPM 7 Main Turbine Generator and Auxiliary Systems/ Manual Turbine trip. (Alternate Path)
N/A/S/L 4
K/A 245000 A2.01 3.7/3.9; Task 2450070101; N I -SOP-4 B.2 Facility Walk-Through JPM 8 Isolation (Emergency) Condenser/ Perform initiation of EC's from the Remote Shutdown Panel #11. (01-OPS-PJE-200-1-64)
K/A 207000 A 1.09 3.7/3.7; Task 2000140401/2079010201; N I -SOP-9.1 D/R 4
JPM 9 Low Pressure Core Spray System/ Lineup Raw Water to Core D/R Spray Pump.
(Used on 2
PRA: Supply containment spray raw water to Core Spray 2000 NRC K/A 209001 A1.08 3.3/3.2; Task 2009170504; N I-EOP-I Attachment 5 Exam)
JPM 10 Emergency Plant Evolutions/ Diesel Fire Pump Start with No Control Power.
K/A 295031 EAI.08 3.8/3.9; Task 2009050501; NI-OP-21A Sect. H.4.4 N/R 8
- Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)ltemate path, (C)ontrol room, (S)imulator, (L)ow-Power, (R)CA
Scenario Outline Nine Mile Point 1 Scenario No. I Operating Test No. 1 Examiners:
Candidates:
Objectives:
Evaluate candidates ability to perform routine operating tasks, raise and/or lower reactor power and to respond to the following failures:
- 1.
APRM #13 fails upscale,
- 2.
Electrical Pressure Regulator oscillation
- 3.
Mechanical Pressure Regulator failure
- 4.
Recirculation Pump Seal leak
- 5.
Fault on Power Bus 102
- 6.
Failure of Diesel Generator 102 Output Breaker to close
- 7.
- 8.
Core Spray Pump 122 trip
- 9.
Core Spray Pump 112 Injection Vales fails closed.
Evaluate the candidates' ability to execute normal, abnormal and emergency procedures while ensuring compliance with Technical Specifications.
This scenario will be classified as an General Emergency 2.1.3 Initial Conditions:
- 1.
IC 24, at 100% power Turnover:
- 1.
Plant is operating at 100% Reactor Power.
- 2.
APRM # 14 out of service. Repair in progress and expected to be returned to service within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
- 3.
All required log entries have been completed including Technical Specification references.
- 4.
Swap stabilizing valves from A and B to E and F.
Event Malf.
Type Event Description No.
No.
N (BOP) Switching Stabilizing Valves - A and B to E and F.
(N1-OP-5, Section F.4.1) 2 NM 19C I
(RO/SRO) APRM # 13 fails upscale.
Form ES-D3-1 Appendix D
Form ES-D-1 Scenario Outline 3
RR33 &
C (BOP) Recirculation Pump Seal Leak (50% lower, 25% upper
_34 over 10 min)
R C TC06 C
--TCO08 (RO/SRO) Reduce power to remove and isolate recirc pump (RO) Electrical Pressure Regulator (EPR) Oscillating.
C (RO) Mechanical Pressure Regulator (MPR) fails.
7 ED07 C
(BOP) Fault on Power Board (PB 102) 8 DG02A 9
RR29 10 CSOID &
CS03B C
(BOP) Diesel Generator 102 Starts, but won't close into the Bus.
M (RO/BOP/SRO) LOCA (30% over 5 minutes)
C 11 (BOP) Core Spray Pump 122 Trips and Core Spray Pump 112 Injection Valve won't open.
Reactor Vessel Water Level can't be maintained greater than 109", Enter SAP's for Containment Flooding.
I
TARGET QUANTITIVE ATTRIBUTES (PER SCENARIO; SEE SECTION D.4.D)
- 1.
Total Malfunctions (5-8)
- 2.
Malfunctions after EOP entry (1-2)
- 3.
Abnormal events (2-4)
- 4.
Major transients (1-2)
- 5.
EOP's entered/required substantive actions (1-2)
- 6.
Eenc s requiring substantive actions (0-2)
- 7.,7Critical tasks 7
(2-3)
( b. i Actual Attributes Actual Attributes 8
3 1
2 1
4 5
6 Initials Appendix D Initials
Scenario Outline Scenario 1 - Description Summary The scenario begins with the Crew assuming the shift with the Plant at rated power and #14 APRM out of service due to having failed on the previous shift. The Crew is directed to perform NI-OP-5, Section F.4.1 and swap the Stabilizing Valves from A and B to E and F. When this evolution is completed, #13 APRM fails upscale which produces a half scram that cannot be reset. The Crew will review Tech. Specs.
Once Tech. Specs. has been consulted, the Crew will note that pressure is oscillating and that the MPR Servo position is also oscillating. The Crew determines that the MPR is failing and will place the EPR in control per NI OP-3 1. When Plant conditions have been stabilized following the MPR problem, the EPR will fail low which causes the Turbine Control Valves to close and the Reactor to Scram. During the scram one of the Reactor Recirculation Pump seal packages is damaged and starts leaking, this problem gets progressively worse for the rest of the scenario.
After the scram actions have been completed PB 102 develops a fault, DG 102 starts but will not close in on the Bus. The Crew will take action to cross tie PB 16B and PBI6A and restore loads. When the loads have been restored a large break LOCA occurs. Core Spray Pump 122 trips and Core Spray Pump 112 injection valve won't open.
Reactor Vessel level will not be able to restore and maintain greater than -109" and the Crew will have to enter SAP's for Containment Flooding.
Major Procedures:
NI-EOP-2, 4 & 8, NI-SAP-1 & 2 EAL Classification:
General Emergency, EAL 2.1.3 Crew has entered the SAP's and is flooding Containment.
Form ES-D-1 Termination Criteria:
Appendix D
Scenario Outline Nine Mile Point 1 Scenario No. 2 Operating Test No. 1 Examiners:
Candidates:
Objectives:
Evaluate candidates ability to perform routine operating tasks, raise and/or lower reactor power and to respond to the following failures:
- 1.
Feedwater Booster Pump #11 trip,
- 2.
Reactor trip bus MG 141 trip,
- 3.
Steam line rupture in Turbine Building,
- 4.
- 5.
RWCU fails to isolate when Liquid Poison is initiated, and
- 6.
One MSIV fails open when isolation attempted.
Evaluate the candidates' ability to execute normal, abnormal and emergency procedures while ensuring compliance with Technical Specifications.
This scenario will be classified as Site Area Emergency 2.2.2 Initial Conditions:
- 1.
IC 24, 100% Reactor Power Turnover:
- 1.
Plant is operating at 100% Reactor Power.
- 2.
Electrical Maintenance have completed #11 FWBP motor winding work, post maintenance testing will be performed at the beginning of shift.
- 3.
Crew to place #11 FWBP in service and remove #12 FWBP.
- 4.
All required log entries have been completed including Technical Specification references.
Event Malf.
Type Event Description N. No.
1 N
(BOP) Place #11 Feedwater Booster Pump in service and remove
- 12.
(Nl-OP-16, Section F.2.0) 2 FW2A C
(BOP/SRO) # 11 Feedwater Booster Pump trip 3
RP1B C
(BOP/RO/SRO) Reactor trip bus MG 141 trip Appendix D Form ES-D3-1
Scenario Outline 4
R (RO/SRO) Power reduction due to FW heating loss.
5 MS12 M
(BOP/RO/SRO) Steam line rupture in Turbine Building (25% over 8 min).
6 RP5A I
(BOP/RO/SRO) RPS failure to scram 7
CU13 C
(BOP) RWCU failure to isolate on Liquid Poison initiation requiring manual isolation 8
MS13A C
(BOP/SRO) MSIV fails open MS01-01 (111)
TARGET QUANTITIVE ATTRIBUTES (PER SCENARIO; SEE SECTION D.4.D)
Actual Attributes Initials
- 1.
Total Malfunctions (5-8) 6
- 2.
Malfunctions after EOP entry (1-2) 2
- 3.
Abnormal events (2-4) 4
- 4.
Major transients (1-2) 1
- 5.
EOP's entered/required substantive actions (1-2) 2
- 6.
EOP contingencies requiring substantive actions (0-2) 0
- 7.
Critical tasks (2-3)
- a.
b.
Form ES-D3-1 Appendix D
Scenario Outline Scenario 2 - Description Summary The scenario begins with the Crew assuming the shift with the Plant at rated power. Maintenance on #11 Feedwater Booster Pump is completed and is ready for post maintenance testing.
The crew is directed to perform NI-OP-16 section F.9.0 to place #11 Feedwater Booster Pump in service and remove #12 Feedwater Booster Pump from service. When this evolution is completed #11 Feedwater Booster Pump will trip on motor overload. The crew will review technical specification and place plant in a 15 day LCO for loss of a HPCI component.
Once the T.S. determination has been made a loss of RPS MG set 141 occurs. A half scram and Feedwater Heating will occur and require crew to reduce power per N1-OP-16. The crew will investigate the cause and transfer RPS MG set 141 to maintenance bus per NI-OP-48.
Following power reduction and resetting the half scram and Feedwater Heating a Main Steam Line leak develops in the Turbine Building. As area radiation monitors alarm the crew will implement EPIP-EPP-21 and directs TB evacuation and emergency power reduction. As rad levels continue to rise the crew will attempt a manual scram.
The ASSS will enter EOP-3 for failure to scram and direct execution of EOP-3.1. MSIV 111 will fail open when MSIVs are directed to be shut and crew must successfully shut one MSIV in each line.
When torus temperature approaches 1 OF Liquid Poison will be initiated and require RWCU to be manually isolated due to failure of the system to auto isolate.
Control rods can be inserted by venting the scram air header and/or removing the RPS fuses.
Major Procedures:
N I -OP-16,40,48, N 1-SOP-1 &4, N 1-EOP-2,3 & 3.1 EAL Classification:
Site Area Emergency, EAL 2.2.2 RPV level and pressure controlled in prescribed band as directed by EOP-2..
Appendix D Form ES-D3-t Termination Criteria:
Scenario Outline Nine Mile Point 1 Scenario No. 3 Operating Test No. 1 Examiners:
Candidates:
Objectives:
Evaluate candidates ability to perform routine operating tasks, raise and/or lower reactor power and to respond to the following failures:
- 1.
Channel 12 backup scram and SDV Vent and Drain Valves fail to reset,
- 2.
RRP 13 GEMAC fails,
- 3.
Loss of Line 1 & 4,
- 4.
DG-103 trips,
- 5.
1 ERV fails open,
- 6.
- ATWS,
- 7.
Liquid Poison Pump #12 trips after start.
Evaluate the candidates' ability to execute normal, abnormal and emergency procedures while ensuring compliance with Technical Specifications.
This scenario will be classified as Alert EAL 6.1.3 Initial Conditions:
- 1.
IC 24, 100% Reactor Power Turnover:
- 1.
Plant is operating at 100% Reactor Power.
- 2.
Manual Scram Instrument Channel test, N1-ST-W15, Sections 8.1 & 8.2 to be completed during this shift
- 3.
All required log entries have been completed including Technical Specification references.
Event Malf Type Event Description No.
~No.
I N
(BOP) Manual Scram Instrument Channel test, N1-ST-W15, Sections 8.1 & 8.2.
2 Overrides I
(SRO) During performance of this test Channel 12 backup scram F-5, and SDV Vent and Drain Valves fail to reset.
- Lamps, page 65, R-031-09 3
(SRO) Severe Thunderstorm Warning is issued by "Power Control".
Appendix D Form ES-D3-1
Scenario Outline 4
RR69C C
(RO/SRO) RRP 13 GEMAC fails, lowering 5
ED01A &
C (BOP/SRO) Loss of Line 1 & 4 ED01B 6
DG01A C
(BOP/SRO) DG-103 trips.
7 AD05 C
(RO/BOP/SRO) 1 ERV fails open.
&AD06 8
R (RO/SRO) Emergency Power Reduction and manual scram to prevent exceeding 11 0°F in the Torus.
9
- RD33A, M
(RO/BOP/SRO) Hydraulic ATWS B3, C & E (Stick all selected at 20) 10 LPOIB C
(BOP/SRO) Liquid Poison Pump #12 trips after start.
TARGET QUANTITIVE ATTRIBUTES (PER SCENARIO; SEE SECTION D.4.D)
Actual Attributes Initials
- 1.
Total Malfunctions (5-8) 7
- 2.
Malfunctions after EOP entry (1-2) 1
- 3.
Abnormal events (2-4) 4
- 4.
Major transients (1-2) 1
- 5.
EOP's entered/required substantive actions (1-2) 3
- 6.
EOP contingencies requiring substantive actions (0-2) 1
- 7.
Critical tasks (2-3)
- a.
Spray Containment with Service Water to prevent exceeding PSP.
- b.
Scram and closure of MSIV's prior to reaching 400 psig.
Prevent exceeding maximum cooldown rate.
Appendix D Form ES-D-1
Scenario Outline Scenario 3 - Description Summary The Crew will assume the shift with Reactor Power at 100%. The weekly surveillance, N1-ST-W15, Section 8.1 and 8.2 are due to be done on this shift. During the surveillance test the Channel 12 "Red" BU SCRAM/SDV Vent and Drain Valve light does not illuminate as required. The Crew will enter Tech. Specs. and determine any LCO actions required.
Following the Tech. Spec. determination, Power Control will contact the Control Room to inform them of a Severe Thunderstorm Warning for the area for the next hour. At this time the Crew will notice that reactor power is lowering and enter Ni -SOP-02 to attempt to stabilize the power change. The Crew will diagnose the failure of RRP 13 GEMAC, take manual control and return power to the previous value.
When Reactor Power has been stabilized, a loss of off-site power will occur resulting in the automatic START of both Diesel Generators. An electrical fault on Power boards 103 and 17B will cause # 103 Diesel Generator to trip on overcurrent. Following completion of SOP-5, #115 kV power will be restored. However, power boards 103 and 17B will remain unavailable.
Then ERV 111 inadvertently opens. All efforts to close the ERV by the Crew are not successful. The Crew will place Torus Cooling in service and perform an Emergency Power reduction. Prior to exceeding Torus temperature of 11 0°F the Crew will Scram the Reactor and attempt to inject Liquid Poison. The first attempt will have the selected Liquid Poison pump tripped shortly after it starts.
When the scram is initiated, the control rods fail to insert. Shortly after the SCRAM, the Turbine will trip and the bypass valves fail closed. The crew will be forced to Terminate and Prevent injection to lower power and maintain pressure below the HCTL curve.
Major Procedures:
N 1-OP-01, N I-SOP-5,10, N 1-EOP-2,4 & 8 EAL Classification:
Alert EAL 6.1.3 Core Spray injecting to restore RPV level, RPV depressurized and reactor shutdown.
Form ES-D-1 Termination Criteria:
Appendix D
Scenario Outline Nine Mile Point 1 Scenario No. 4 Operating Test No. 1 Examiners:
Candidates:
Objectives:
Evaluate candidates ability to perform routine operating tasks, raise and/or lower reactor power and to respond to the following failures:
- 1.
Failure of #12 EC LCV to open.
- 2.
FZWLMS fails downscale.
- 3.
Steam Leak in Primary Containment.
- 4.
Power Board # 16B failure to transfer.
- 5.
Containment Spray #121 trip Evaluate the candidates' ability to execute normal, abnormal and emergency procedures while ensuring compliance with Technical Specifications.
This scenario will be classified as an Alert, EAL 3.1.1.
Initial Conditions:
- 1.
IC 24, 100% power / 103% Rod line
- 2.
Containment Spray Pump #111 and RAW Water Pump out of service for PM's.
- 3.
- 12 FZWLI out of service due to an accurex problem.
Turnover:
- 1.
The Plant is operating at 100% power.
- 2.
Equipment out of service:
- a. #111 Containment Spray Pump for PM.
- c. #12 FZWLI for accurex problem.
- 3.
Complete N1-ST-Q4 Section 8.4, Quarterly Surveillance of the Emergency Condenser LCV's.
- 4.
All appropriate Equipment Log entries have been made.
Event Malf.
Type Event Description No.
No.
1 N
(BOP/SRO) Complete NI-ST-Q4 Section 8.4, Quarterly Surveillance of the Emergency Condenser LCV's.
Appendix D Form ES-D-1
Scenario Outline 2
EC05B C
(BOP) Emergency Condenser #12 LCV fails to operate.
3 RR52 I
(RO/SRO) RPS Channel #11 Hi/Lo Rosemount fails downscale.
4 RP04A I
(RO/SRO) Failure of RPS Channel #11 to trip on the failure of the Rosemount Instrument.
5 ECQI C
(BOP/RO) Steam Leak in Containment (10% over 8 minutes) 6 R
(RO) Power Reduction 7
EC01 M
(BOP/RO/SRO) Steam Leak in Containment (100% over 3 minutes) 8 Reactor Scram 9
ED19 C
(BOP/SRO) Power Board 16B Section B fails on the transfer.
10 CS01C C
(BOP/SRO) Containment Spray Pump 121 trips after start.
11 AD07C C
(RO/SRO) ERV 113 fails to open TARGET QUANTITIVE ATTRIBUTES (PER SCENARIO; SEE SECTION D.4.D)
Actual Attributes Initials
- 1.
Total Malfunctions (5-8) 6
- 2.
Malfunctions after EOP entry (1-2) 3
- 3.
Abnormal events (2-4) 4
- 4.
Major transients (1-2) 1
- 5.
EOP's entered/required substantive actions (1-2) 3
- 6.
EOP contingencies requiring substantive actions (0-2) 1
- 7.
Critical tasks (2-3)
- a.
b.
Appendix D Form ES-D-1
Scenario Outline Scenario 4 - Description Summary The crew will assume the shift with the Plant operating at 100% power, the Containment Spray Pump # 111 and the Raw Water Pump out of service for PM's. The FZWLI #12 is also out of service due to accurex problems. All are anticipated to be back in service prior to the end of the shift. The crew will complete N1-ST-Q4 Section 8.4, Quarterly Surveillance of the Emergency Condenser LCV's. While performing this evolution the crew will find that Emergency Condenser #12 LCV will not open and they must enter and interpret Technical Specifications regarding this problem.
Once the LCV problem has been addressed the RPS channel 11 Rosemount instrument channel fails downscale and RPS Channel #11 does not trip. The Crew will consult Technical Specifications and direct the appropriate actions be taken, which will include insertion of a manual scram on the "A" RPS.
After T.S. have been reviewed a small steam leak in the primary containment causes the containment parameters, such as temperature, pressure and humidity to rise slowly. As they attempt to determine the cause of the problem, the SRO will direct a power reduction. As power is reduced the steam leak gets larger and the crew is forced to scram the reactor. Power Board 16 Section B remains deenergized following the power transfer, the crew will attempt to recover the loads of PB 16 Section B.
The crew then attempts to lower containment pressure by starting Containment Spray Pump #121, however, it fails to start and with Containment Spray Pump #111 out of service, a blowdown will have to be performed. Eventually containment pressure will exceed Pressure Suppression Pressure and they will has to perform a blowdown. During the blowdown one of the ERV's does not open and the crew must use another Relief Valve to facilitate the blowdown.
Major Procedures:
N1-OP-30, N1-EOP-02, N1-EOP-04, N1-EOP-08 EAL Classification:
ALERT based on EAL 3.1.1 RPV depressurized with level in the assigned band.
Form ES-D-1 Termination Criteria:
Appendix D
Scenario Outline Nine Mile Point 1 Scenario No. 5 Operating Test No. 1 Examiners:
Candidates:
Objectives:
Evaluate candidates ability to perform routine operating tasks, raise and/or lower reactor power and to respond to the following failures:
- 1.
Loss of #11 Reactor Building EVS fan,
- 2.
Feedwater heater level control valve fails closed,
- 3.
Fuel Failure,
- 4.
Loss of Instrument Air,
- 5.
A RWCU leak,
- 6.
Loss of CRD pumps, and
- 7.
Core Spray inside IV's failure to open properly.
Evaluate the candidates' ability to execute normal, abnormal and emergency procedures while ensuring compliance with Technical Specifications.
This scenario will be classified as an Alert 3.1.1 Initial Conditions:
- 1.
IC 8, 48% power during power ascension.
- 2.
EDG 102 OOS for oil change, clearance hung.
Turnover:
- 1.
EDG 102 is out of service under clearance for an oil change.
- 2.
Continue startup to raise power to 60% using recirc flow.
- 3.
Once at 60% power perform N 1-ST-M8 Section 8.1 RB EVS Operability for Loop
- 11.
- 4.
All appropriate Equipment Log entries have been made.
Event Malf No.
Type Event Description 1
R
[(RO/SRO) Raise recirc flow to achieve 60% power 2
1 1(BOP) Reactor Building Emergency Ventilation System N
Operability, Nl-ST-M8, Section 8.1 for Loop #11.
3 HV02A C { (RO/SRO) #11 RB EVS fan trip.
Form ES-D-1 Appendix D
Scenario Outline FW35 (BOP/SRO) Feedwater Heater 124 drain valve fails closed RXO1 leading to a power increase and Fuel Failure. (10% over 10 minutes.) and entry into N1-SOP-02, Unexplained Power Change and verification of string isolation per N1-OP-16 5
IA01 C
(RO/BOP) Loss of Instrument Air (50% over 8 minutes) 6 cu01 M
(RO/BOP/SRO) As the Reactor SCRAM is inserted a Reactor Water Cleanup leak (30% over 10 minutes) will develop.
RD35A & B C
(BOP) Following the leak the running CRD Pump will trip and the standby CRD Pump will be unable to be started.
CS03 A,B,C C
(BOP/SRO) Core Spray inside IV's fail to open until Reactor
& D pressure is < 250 psig.
9 TARGET QUANTITIVE ATTRIBUTES (PER SCENARIO; SEE SECTION D.4.D)
Actual Attributes Initials
- 1.
Total Malfunctions (5-8) 7
- 2.
Malfunctions after EOP entry (1-2) 2
- 3.
Abnormal events (2-4) 3
- 4.
Major transients (1-2) 1
- 5.
EOP's entered/required substantive actions (1-2) 2
- 6.
EOP contingencies requiring substantive actions (0-2) 0
- 7.
Critical tasks (2-3)
- a.
b.
Appendix D Form ES-D-1
Scenario Outline Scenario 5 - Description Summary The Crew will assume the shift with the plant at 50% power and EDG 102 OOS for an oil change with a clearance hung. They will be directed to raise power using recirculation flow to 60% power and then perform N1-ST-M8 section 8.1 for Loop #11.
Approximately 2 minutes after the RB fia is started an overcurrent trip will occur. The crew will access the trip and review T.S. 3.4.4. After the T.S. have been reviewed, failure of the 124 Feedwater heater LCV closed will occur which cause a power excursion and some amount of fuel failure. The Crew will enter NI-SOP-02 and OP-16 to verify the string isolation. After actions for the string isolation are complete an air leak occurs on the discharge of the #11 Instrument Air compressor. As the Crew is responding to the loss of Instrument Air header pressure by entering NI-SOP-06. The Crew will respond to this by initiating a manual SCRAM of the Reactor.
Following the reactor scram, a RWCU leak in the primary containment will occur along with a trip of the operating CRD pump. The crew will enter N1-EOP-4 for Primary Containment Control.
To maintain level in the desired band per N1-EOP-2, Core Spray will be initiated but the inboard IVs will fail to open until pressure is less than 250 psig.
Major Procedures:
N1-ST-M8, NI-OP-16, NI-SOP-2,4, Nl-EOP-2 & 4 EAL Classification:
Alert EAL 3.1.1 Reactor shutdown with RPV level maintained in prescribed band with CS injecting.
Form ES-D3-1 Termination Criteria:
Appendix D