TMI-14-002, Draft - Outlines (Folder 2)
| ML14111A294 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 01/07/2014 |
| From: | Libra R Exelon Generation Co |
| To: | Peter Presby Operations Branch I |
| Jackson D | |
| Shared Package | |
| ML13333A179 | List: |
| References | |
| TAC U01895, TMI-14-002 | |
| Download: ML14111A294 (60) | |
Text
=~Exelon Generation@
Three Mile Island Unit 1 Route 441 South P.O. Box 480 Middletown, PA 17057 717-948-8000 Office www.exeloncorp.com January 7, 2014 TMI-14-002 USNRC, Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 THREE MILE ISLAND NUCLEAR STATION, UNIT 1 (TMI-1)
RENEWED OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289
SUBJECT:
SUBMITTAL OF INITIAL OPERATOR LICENSING EXAMINATION OUTLINES Enclosed are the examination outlines, supporting the Initial License Examination scheduled for the week of April 7, 2014, at Three Mile Island Unit 1.
This submittal includes all appropriate Examination Standard forms and outlines in accordance with NUREG 1021, "Operator Licensing Examination Standards," Revision 9 Supplement 1.
In accordance with NUREG 1021, Revision 9 Supplement 1, Section ES-201, "Initial Operator Licensing Examination Process," please ensure that these materials are withheld from public disclosure until after the examinations are complete.
Should you have any questions concerning this letter, please contact Mike Fitzwater of Regulatory Assurance at (717) 948-8228. For questions concerning examination materials, please contact Rich Megill, Exam Author, at (717) 948-2093.
Respectfully,
~
Rick W. Libra Site Vice President, Three Mile Island Unit 1 Exelon Generation Co., LLC RWUmdf
Enclosures:
(Mailed to Peter Presby, Chief Examiner, NRC Region I)
Examination Security Agreements (Form ES-201-3)
Administrative Topics Outlines (Form ES-301-1)
Control Room/In-Plant Systems Outline (Form ES-301-2)
PWR Examination Outline (Form ES-401-2)
Generic Knowledge and Abilities Outline (Tier 3) (Form ES-401-3)
Statement detailing method of Written Outline generation Scenario Outlines (Form ES-D-1)
Record of Rejected K/As (Form ES-401-4)
Completed Checklists:
Examination Outline Quality Checklist (Form ES-201-2)
Transient and Event Checklist (Form ES-301-5)
Three Mile Island Unit 1
=::r-Exelon GenerationJ) Route 441 South P.O. Box480 Middletown, PA 17057 717-948-8000 Office www.exeloncorp.com cc: (without attachments)
Chief, NRC Operator Licensing Branch NRC Senior Resident Inspector-- TMI Unit 1
ES-401 PWR Examination Outline FORM ES-401-2
!Facility Name: Date of Exam:
RO K/A Category Points SRO-Only Points I
Tier Group K K K K K K A A A A G Total A2 G* Total 1 2 3 4 5 6 1 2 3 4 *
- 1. Emergency 1 3 3 3 3 3 3 18 3 3 6 Abnormal 2 2 2 1 N/A 2 1 N/A 1 9 2 2 4 Plant Evolutions Tier Totals 5 5 4 5 4 4 27 5 5 10 1 3 2 3 3 3 2 2 3 2 2 3 28 3 2 5 2.
2 1 1 1 1 1 1 1 1 1 1 0 10 0 2 1 3 Plant Systems Tier Totals 4 3 4 4 4 3 3 4 3 3 3 38 5 3 8 1 2 3 4 1 2 3 4
- 3. Generic Knowledge and Abilities 10 7 Categories 3 3 2 2 2 2 2 1 Note: 1. Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO and SAO-only outlines (i.e., except for one category in Tier 3 of the SAO-only outline, 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 RO exam must total 75 points and the SAO-only exam must total25 points.
- 3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate KIA statements.
- 4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5. Absent a plant-specific priority, only those KIAs havin9 an importance rating (lA) of 2.5 or higher shall be selected.
Use the RO and SAO ratings for the RO and SAO-only portions, respectively.
- 6. Select SAO topics for Tiers 1 and 2 from the shaded systems and KIA categories.
7.* The generic (G) KIAs in Tiers 1 and 2 shall be selected from Section 2 of the KIA Catalog, but the topics must be relevant to the applicable evolution or system Refer to Section D.1.b of ES-401 for the applicable KIAs.
- 8. On the following pages, enter the KIA numbers, a brief description of each topic, the topics' importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SAO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SAO-only exams.
- 9. For Tier 3, select topics from Section 2 of the KIA catalog, and enter the KIA numbers, descriptions, IRs, and point totals(#) on Form ES-401-3. Limit SAO selections to KIAs that are linked to 10 CFR 55.43.
ES-401, 21 of 33
ES-401 2 Form ES-401-2 PWR Examination Outline Form ES-401 Emergency and Abnormal Plant Evolutions- Tier 11Group 1 (RO) 1~--------------------------~~r-~~--~ ------------~~~--------------------~---.---11 EIAPE #I Name I Safety Function K K KIA Topic(s) IR #
2 0 2.6 Reactor Trip - Stabilization - Recovery I 1 2
Vital System Status Verification 1 1 Post Trip Stabilization I 1 0 3.7 Pressurizer Vapor Space Accident I 3 3.4 0
li i concepts as they apply to Pump Malfunctions (Loss of RC Flow): Basic steady state _
2 9 the,rmod1mamic relationship between RCS loops and S/Gs resulting from unbalanced Loss of Rx Coolant Makeup I 2 3.2 Whether charging line leak exists 0 interrelations between the Loss of Residual Heat Removal System Loss of RHR System I 4 3.2 2 : LPI or Decay Heat Removai/RHR pumps and I or monitor the following as they apply to the Loss of 3.6 r Water: SWS as a backup to the CCWS 2.8 4
Steam Gen. Tube Rupture I 3 4.2 Steam Line Rupture - Excessive Heat Transfe the interrelations between the Steam Line Rupture and the following:
2.6 Excessive Heat Transfer I 4 Loss of Main Feedwater I 4 0 and monitor the following as they apply to a Station Blackout 4.3 of power from offsite Loss of Off-site Power I 6 3.2 Loss of Vital AC lnst. Bus I 6 0 4.1 Loss of DC Power I 6 2.8 Loss of Nuclear Svc Water I 4 0 Loss of Instrument Air I 8 4.1 4.2 Ge,nerat<~r Voltage and Electric 3.3 U";[lJrO>li"Jr.F!S I6 18 ES-401, 22 of 33
ES-401 3 Form ES-401-2 EIAPE #I Name I Safety Function IR #
2 Continuous Rod Withdrawal I 1 0 Dropped Control Rod I 1 0 Inoperable/Stuck Control Rod I 1 01 2.5 Emergency Boration I 1 0 Pressurizer Level Malfunction I 2 3.7 concepts as they apply to Loss of Source Range Nl I 7 01 Effects of voltage changes on 2.5 Loss of Intermediate Range Nl I 7 0 Fuel Handling Accident I 8 0
Refueling Canal Level Decrease I 8 Steam Generator Tube Leak I 3 0 Loss of Condenser Vacuum I 4 0 Accidental Liquid RadWaste Rei. I 9 0 Accidental Gaseous Radwaste Rei. I 9 0 ARM System Alarms I 7 0 and I or monitor the following as they apply to the Plant Fire on Plant Fire On-site I 9 8 3.6 for the following responses as they apply to the Control Control Room Evac. I 8 01 System response to reactor trip 3.9 Shutdown Outside Control Room I 8 Loss of CTMT Integrity I 5 0 lnad. Core Cooling I 4 0 High Reactor Coolant Activity I 9 0 Plant Runback I 1 0 Loss of NNI-X I 7 of the operational implications of the following concepts as they apply to Loss of NNI-Y I 7 02 NNI-Y): Nonnal, abnormal and emergency operating procedures 3 with (Loss of NNI-Y).
Turbine Trip I 4 0 of the interrelations between the (Emergency Diesel Actuation) and the and functions of control and safety systems, including Emergency Diesel Actuation I 6 01 i , signals, interlocks, failure modes, and automatic and manual 4
Flooding I 8 0 Inadequate Subcooling Margin I 4 0 LOCA Cooldown I 4 following as they apply to the Natural Circulation Cooldown I 4 to appropriate procedures and operation within 3.5 and and I or monitor the following as they apply to the (EOP Rules):
3 EOP Rules 2.8 behavior characteristics of the facility.
4 EOP Enclosures 2 2 9 ES-401, 23 of 33
ES-401 4 Form ES-401-2 Form ES-401 KIA Topic(s) IR #
RCPS design feature(s) and/or interlock(s) which provide for i : Adequate cooling of RCP motor and seals; Knowledge of the 2.8; Reactor Coolant Pump loss or malfunction on the following will have on the RCPS: RCP 2
2.7 seal water supply eves design feature(s) and/or interlock(s) which provide for Interrelationships and design basis, including fluid flow splits in Chemical and Volume Control (e.g., charging and seal injection flow); Knowledge of 3.3; 4 2 Residual Heat Removal 3.7 Emergency Core Cooling 3.3 Pressurizer Relief/Quench Tank 2.6 Component Cooling Water 2.5 3.8 3.3 Safety Features 3.6; 2
3.4 2.5; 2
4.1 4.1; 2
3.5 3.6; Main and Reheat Steam 2 3.1 3.4 of the effect that a loss or malfunction of the AFW will have on RCS; Knowledge of events related to system operation/status 4.4; Auxiliary/Emergency Feedwater must be reported to internal organizations or external agencies, such as 2
2.7 the NRC, or the transmission system operator.
to (a) predict the impacts of the following malfunctions or operations on distribution system; and (b) based on those predictions, use to correct, control, or mitigate the consequences of those 2.9 melfllrlct;,,ns or operations: Consequences of improper sequencing when or from an inverter DC Electrical Distribution 2.5 of the effect of a loss or malfunction of the following will have on Emergency Diesel Generator : Fuel oil storage tanks 3.2 operate and/or monitor in the control room: Radiation Process Radiation Monitoring control panel 3.7 predict and/or monitor changes in parameters (to prevent exceeding limits) associated with operating the SWS controls including: Reactor 2.6 building closed cooling water temperatures of the physical connections and/or cause-effect relationships lAS and the following systems: Cooling water to compressor 2.6 predict and/or monitor changes in parameters (to prevent exceeding associated with operating the containment system controls 3.7 F1;i1Jlllinc:luclina: Containment pressure, temperature, and humidity 28 ES-401, 24 of 33
ES-401 5 Form ES-401-2 IES-401 PWR Examination Outline Form ES-401-~
Plant Systems -Tier 2/Group 2 (RO)
!System # I Name KKK 1 2 3 K
4 K!K 5 6
~~
1 :..
- 3 4 KIA Topic(s) lA #
~Knowledge of the following operational implications as they apply to the CADS:
pot Control Rod Drive
~ I CADS circuitry, including effects of primary/secondary power mismatch on rod motion 3.2 1
- b L *~"" ~*" " "' " '"
po2 Reactor Coolant ~ 0 r,'J Charging pump ptt Pressurizer Level Control liM IOWCOnirOIS oo*rn' - 3.5 1 pt4 Rod Position Indication l,c. 0 I
0 of bus power supplies to the following: NIS channels, components, pt5 Nuclear Instrumentation 3.3 1 i
pt6 Non-nuclear Instrumentation
~~ 0 pt7 In-core Temperature Monitor c~ 0 P27 Containment Iodine Removal Ill i
- _ (a) predict the impacts of the following malfunctions or operations on and (b) based on those predictions, use Procedures to correct, mitigate the consequences of those malfunctions or operations: High 3 1 in the filter system f-
~~~t:drogen Recombiner and Purge 0 p29 Containment Purge 0 P33 Spent Fuel Pool Cooling P34 Fuel Handling Equipment p35 Steam Generator
.. fl i
li 1 to predict and/or monitor changes in parameters (to prevent exceeding i
- i associated with operating the Fuel Handling System controls I i :Water level in the refueling canal 2.9 0
1 0
fA~~~wl~~~~"
p4* Steam DumpfTurbine Bypass Control
~ 8i effect of a loss or malfunction on the following will have on the
- Co~troller and positioners, including ICS, S/G, CADS 2.7 1 1:>45 Main Turbine Generator I! ~ 0 m~
p55 Condenser Air Removal 0 lil li!W;Jilll' uy~ effect that a loss or malfunction of the CARS will have on the i condenser 2.5 1 li I , *~""'-- ... "'"'" ~- '""" ,~. . .
P56 Condensate 0 P68 Liquid Radwaste 1!2 "u'o" rario isolation 3.6 1 p71 Waste Gas Disposal
~ I feature(s) and/or interlock(s) which provide for the Isolation of waste gas release tanks 2.9 1 p72 Area Radiation Monitoring P75 Circulating Water 0
3 iL ~
of the physical connections and/or cause-effect relationships ARM system and the following systems: Fuel building isolation 3.6 1 0
iF p79 Station Air m 0 p86 Fire Protection 0
~
10
,KIA Category Totals:
r Point Total:
ES-401 , 25 of 33
ES-401 2 Form ES-401-2 PWR Examination Outline Form ES-401 Emergency and Abnormal Plant Evolutions- Tier 11Group 1 (SRO) lr--------------------------r~--r-~~ ~----------------------------------r---.---11 EIAPE #I Name I Safety Function KIA Topic(s) IR #
Reactor Trip - Stabilization - Recovery I 1 determine and interpret the following as they apply to the (Vital System Vital System Status Verification I 1 vo,cilio.tir*n\* Facility conditions and selection of appropriate procedures during 4 and emergency operations.
0 Post Trip Stabilization I 1 Pressurizer Vapor Space Accident I 3 0 0
Knowledge of the operational implications of EOP warnings, cautions, and notes. 4.3 RCP Malfunctions I 4 0
RCP Malfunctions (Loss of RC Flow) I 4 Loss of Rx Coolant Makeup I 2 0 Loss of RHR System I 4 0 0
Pressure Control System 0
0 Steam Gen. Tube Rupture I 3 0 Steam Line Rupture - Excessive Heat interpret control room indications to verify the status and operation of a Excessive Heat Transfer I 4 understand how operator actions and directives affect plant and system 4.4 and interpret the following as they apply to the Loss of Main Loss of Main Feedwater I 4 3.7
- Status of MFW pumps, regulating and stop valves 0
Loss of Off-site Power I 6 of the specific bases for EOPs. 4 Loss of Vital AC lnst. Bus I 6 0 Loss of DC Power I 6 0 Loss of Nuclear Svc Water I 4 2.5 Loss of Instrument Air I 8 0 0
GAn.,lmlrlr Voltage and Electric 0
""'"rn"n'""" I 6 0 0 0 p Point Total: 6 ES-401, 22 of 33
ES-401 3 Form ES-401-2 Form EIAPE #I Name I Safety Function IR #
2 Continuous Rod Withdrawal I 1 0 Dropped Control Rod I 1 0 Inoperable/Stuck Control Rod I 1 0 Emergency Boration I 1 0 Pressurizer Level Malfunction 1 2 0 Loss of Source Range N I I 7 0 Loss of Intermediate Range Nl I 7 0 Fuel Handling Accident I 8 Refueling Canal Level Decrease 1 8 4 Steam Generator Tube Leak I 3 0 Loss of Condenser Vacuum I 4 0 Accidental Liquid RadWaste Rei. I 9 0 and recognize trends in an accurate and timely manner utilizing Accidental Gaseous Radwaste Rei. 1 9 4.2 control room reference material.
ARM System Alarms I 7 0 Plant Fire On-site I 9 8 0 to perfonn specific system and integrated plant procedures during all modes Control Room Evac. I 8 4.4 Shutdown Outside Control Room I 8 Loss of CTMT Integrity I 5 0 lnad. Core Cooling I 4 0 High Reactor Coolant Activity 1 9 0 Plant Runback I 1 0 Loss of NNI-X I 7 0
Loss of NNI-Y I 7 Turbine Trip I 4 0 Emergency Diesel Actuation I 6 0 Flooding I 8 0 Inadequate Subcooling Margin I 4 0 LOCA Cooldown I 4 0
Natural Circulation Cooldown I 4 determine and interpret the following as they apply to the (EOP Rules):
EOP Rules and selection of appropriate procedures during abnormal and 4 EOP Enclosures 0 0 0 4 ES-401, 23 of 33
ES-401 4 Form ES-401-2 PWR Examination Outline Form ES-401 Plant Systems ** Tier 2/Group 1 (SAO)
KIA Topic(s) lA #
0 Chemical and Volume Control 0 Residual Heat Removal 0 Emergency Core Cooling 0 to analyze the effect of maintenance activities, such as degraded power Pressurizer Relief/Quench Tank 4.2 on the status of limiting conditions for operations.
Component Cooling Water 0 0
predict the impacts of the on
- and (b) based on those predictions, use procedures to correct, 3.7 mitigate the consequences of those malfunctions or operations
eered Safety Features 0
0 0
Containment Spray 0 predict the impacts of the following malfunctions or operations on
- and (b) based on predictions, use procedures to correct, control, or Main and Reheat Steam consequences of those malfunctions or operations
- Increasing 3.6
, its relationship to increases in reactor power 0
1 Auxiliary/Emergency Feedwater 0 AC Electrical Distribution 0 of the bases in Technical Specifications for limiting conditions for DC Electrical Distribution 4.2 safety limits.
Emergency Diesel Generator 0 Process Radiation Monitoring 0 i to (a) predict the impacts of the following malfunctions or operations on and (b) based on those predictions, use procedures to correct, Service Water 3.1 0
0 Category Totals: 5 ES-401, 24 of 33
ES-401 5 Form ES-401-2 ES-401 1 PWR Examination Outline Form ES-401-2 Plant Systems- Tier 2/Group 2 (SAO) 1~~
~ ~ ~ ~ ~
- I K
System #I Name KIA Topic(s) lA #
5 001 Control Rod Drive t! i;~
0 002 Reactor Coolant
~ ~l 0
~11 Pressurizer Level Control
~ ~ 0
~
~14 Rod Position Indication 0 i to (a) predict the impacts of the following malfunctions or operations on NIS; and (b based on those predictions, use procedures to correct, control, p15 Nuclear Instrumentation 3.5 1 ii the consequences of those malfunctions or operations: Faulty or operation of detectors or compensating components
>_(a) predict the impacts of the following malfunctions or operations on I and (b) based on those predictions, use procedures to correct, p16 Non-nuclear Instrumentation 3.1 1 mitigate the consequences of those malfunctions or operations:
failure p17 In-core Temperature Monitor 0 P27 Containment Iodine Removal
~\1 0
~~~trol ~
Recombiner and Purge 0
' ~t p29 Containment Purge 0 p33 Spent Fuel Pool Cooling 0 p34 Fuel Handling Equipment p35 Steam Generator
- 0 0
1-fl p41 Steam Dump/Turbine Bypass Control 0 p45 Main Turbine Generator 0 p55 Condenser Air Removal 0 p56 Condensate 0 p68 Liquid Radwaste I 0 p71 Waste Gas Disposal I I 0 p72 Area Radiation Monitoring i 0 p75 Circulating Water I 0 p79 Station Air ~ 0 rt~t~:~~!
~
p86 Fire Protection diagnose and recognize trends in an accurate and timely manner 4.2 1 utilizing the appropriate control room reference material.
IKIA Category Totals: 0 0 0 0 0 0 0 2 0 0 Group Point Total; 3 ES-401, 25 of 33
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3
!Facility Name: Date of Exam RO _§RO-Only Category KIA# Topic IR # IR #_
2.1. 20 Ability to interpret and execute procedure steps. 4.6 1 4.6 2.1. 36 ""v .. *cY"c of procedures and limitations involved in core alterations. 3 1 4.1
[1. 2.1. 42 ""v*"cY"c of new and spent fuel movement procedures. 2.5 1 3.4
[Conduct of 2.1. 34 .,, of primary and secondary plant chemistry limits. 2.7 3.5 1
!Operations 2.1. 41 '" IcY,.~ of the refueling process. 2.8 3.7 1 2.1.
ISubtotal 2_
Ability to perform pre-startup pi .:Jcedures for the facility, including operating those controls 2.2. 01 'with plant equipment that could affect reactivity.
4.5 1 4.4 2.2. 13 .,, of tagging and clearance procedures. 4.1 1 4.3 2.2. 22 ""v.vl.,uy" of limiting conditions for operations and safety limits. 4 1 4.7
[2.
Ability to analyze the effect of maintenance activities, such as degraded power sources, on Equipment 2.2. 36 3.1 4.2 1 the~tatl!s of limiting conditions for operations.
Control Ability to interpret control room indications to verify the status and operation of a system, 2.2. 44 and understand how operator actions and directives affect plant and system conditions.
4.2 4.4 1 2.2.
[Subtotal 2 2.3. 11 Ability to control radiation releases. 3.8 1 4.3
""u"l"uY" of radiation monitoring systems, such ; fixed radiation monitors and alarms, 2.3. 15 portable survey instruments, personnel monitoring dqu.pm.,, "* etc.
2.9 1 3.1
- 3. 2.3. 04 Knowledge of radiation exposure limits under normal or emergency conditions. 3.2 3.7 1
[), .J:.
I ""uvvl.,uy" of radiation or contamination hazards that may arise during normal, abnormal, 2.3. 14 3.4 3.8 1 Control 1or ~* ""' ,.c,v: conditions or activities.
[2.3.
[2.3.
[Subtotal 2
[2.4. 03 lAbility to identify post-accident instrumentation. 3.7 1 3.9
[2.4. 17 Knowledge of EOP terms and definitions. 3.9 1 4.3
- 4. [2.4. 40 Knowledge of SRO responsibilities in emergency plan implementation. 2.7 4.5 1 Emergency
~rocedures [2.4.
1 Plan 2.4.
2.4.
Subtotal 1 1
Tier 3 Point Total 7 ES-401, Page 26 of 33
ES-401 Record of Rejected KlA's Form ES-401-4 Randomly Selected Tier I Group Reason for Rejection KA 0011AK1.19 1I 2 replaced by A03 I The subject KIA isn't relevant at the subject facility.
AK1.2 006 I K6.19 replaced Topic oversampled. 006IK6.19 overlaps with KIA 005 I K3.05 2I 1 by 064 I K6.08 on the Written Exam.
073 I A4.01 replaced Topic oversampled. 0731A4.01 overlaps with KIA 057 I AK3.03 2I 1 by 004 I K4.05 and 068 I A3.02 on the Written Exam.
078 I A3.01 replaced Topic oversampled. 078 I A3.01 overlaps with KIA 065 I 2.4.45 2I 1 by 039 I A3.02 and 078 I K1.04 on the Written Exam.
003 I 2.2 .37 replaced 2I 1 The subject KIA isn't relevant at the subject facility.
by 061 I 2.4.30 017 I K3.01 replaced Topic oversampled. 017 I K3.01 overlaps with 038 I EA2.09 on 212 by 055 I K3.01 the Written Exam.
079 I 2.2.39 replaced 212 The subject KIA isn't relevant at the subject facility.
by 015 I K2.01 008 I 2.4.9 replaced Topic oversampled. 008 I 2.4.9 overlaps with 026 I AA 1.03, 008 2 I 1 SRO by 063 I 2.2.25 I A2.08, 013 I K2.01 and 026 I K1.02 on the Written Exam.
010 I A2.01 replaced Topic oversampled. 010 I A2.01 overlaps with 056 I AA 1.03, 2 I 1 SAO by 012 I A2.03 016 I A2.01, and 028 I 2.4.6 on the Written Exam.
034 I K5.03 replaced Topic oversampled. 034 I K5.03 overlaps with 025 I AK2.02, 2 I 2 SRO by 016 I A2.01 and G 2.1.36 on the Written Exam.
Three Mile Island 2014 NRC Initial License Written Examination Written Examination Outline Methodology The written examination outline was developed using NKEG Version 1 .1 software by Westinghouse. Generation of RO and SRO outlines was done randomly from the database containing NUREG 1122 Rev 2 Supp 1 catalog, as pre-suppressed for TMI system design.
Pre-suppression was done by reviewing previously approved, suppressed knowledges and abilities and suppressing them against the appropriate Tier 1 and Tier 2 categories. In accordance with NUREG 1021 Revision 9 supp 1 ES-401 D.1.b Generic 2.2.3 and 2.2.4 were suppressed, also the Generic not specifically listed for sampling of Tiers 1 and 2 were suppressed, except for the Tier 3 sampling.
KIA rejections are listed on ES-401-4 as required by NUREG 1021.
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Three Mile Island Date of Examination: April 2014 Examination Level: RO [8J SAO D Operating Test Number: 289-2014-301 Administrative Topic Type Describe activity to be performed (See Note) Code*
Calculate an Estimated Critical Boron Concentration in accordance with 11 03-15B, ESTIMATED CRITICAL CONDITIONS.
Conduct of Operations M/R 2.1.25 (3.9): Ability to interpret station reference materials such as graphs, curves, tables, etc.
Perform OP-TM-300-202, QUADRANT POWER TILT AND CORE IMBALANCE USING THE OUT-OF-CORE DETECTOR SYSTEM, and compare to COLA limits.
Conduct of Operations M/R 2.1.37 (4.3): Knowledge of procedures, guidelines, or limitations associated with reactivity management.
Reactor Coolant Pump electrical print reading to determine pump operation.
Equipment Control N/R 2.2.41 (3.9): Ability to obtain and interpret station electrical and mechanical drawings.
Radiation Control Category not selected for RO applicants.
ERO Notification - Alt path Emergency Procedures/Plan M/S 2.4.39 (3.9): Knowledge of RO responsibilities in emergency plan implementation.
NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.
- Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank(~ 3 for ROs; ~ 4 for SROs & RO retakes)
(N)ew or (M)odified from bank (?. 1)
(P)revious 2 exams(~ 1; randomly selected)
ES 301, Page 22 of 27
I '
2014 TMI RO NRC EXAMINATION CONDUCT OF OPERATIONS: Given a set of plant conditions, calculate the Estimated Critical Boron Concentration for reactor startup. Modified Bank JPM.
CONDUCT OF OPERATIONS: Given a failed computer, perform OP-TM-300-202, QUADRANT POWER TILT AND CORE IMBALANCE USING THE OUT-OF-CORE DETECTOR SYSTEM, and compare to COLA limits. Modified Bank JPM. Calculation is RO/SRO common.
EQUIPMENT CONTROL: Given a set of conditions and Reactor Coolant Pump electrical prints, determine if and where the Reactor Coolant Pump can be operated. New JPM.
RADIATION CONTROL: Given a General Emergency declaration and a failure of the ERO Notification using the World Wide Web, initiate activation of the ERO using the Live Everbridge Agent. Modified Bank JPM:
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Three Mile Island Date of Examination: April 2014 Examination Level: RO 0 SRO [gj Operating Test Number: 289-2014-301 Administrative Topic Type Describe activity to be performed (See Note) Code*
Review and Approve an Estimated Critical Boron Concentration.
Conduct of Operations M/R 2.1.25 (4.2): Ability to interpret station reference materials such as graphs, curves, tables, etc.
Review and Approve OP-TM-300-202, QUADRANT POWER TILT AND CORE IMBALANCE USING THE OUT-OF-CORE DETECTOR SYSTEM, and compare to Conduct of Operations M/R COLR limits.
2.1.37 (4.6): Knowledge of procedures, guidelines, or limitations associated with reactivity management.
Reactor Coolant Pump electrical print reading to determine pump operation and Tech Spec implications.
Equipment Control N/R 2.2.41 (3 . 9): Ability to obtain and interpret station electrical and mechanical drawings.
Implement the Requirements of ODCM for RMS Operability.
Radiation Control M/R 2.3.15 (3.1) Knowedge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
Emergency Action Level Identification, Event Declaration, and Protective Action Recommendation.
Emergency Procedures/Plan D/R 2.4.44 (4.4): Knowledge of emergency plan protective action recommendations.
NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.
- Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (.s_ 3 for ROs; .s_ 4 for SROs & RO retakes)
(N)ew or (M)odifiecl from bank(~ 1)
(P)revious 2 exams (.s. 1; randomly selected)
ES 301 , Page 22 of 27
2014 TMI SRO NRC EXAMINATION CONDUCT OF OPERATIONS: Review and approve an Estimated Critical Boron Concentration calculation with multiple errors. Modified Bank JPM.
CONDUCT OF OPERATIONS: Given a failed computer, perform OP-TM-300-202, QUADRANT POWER TILT AND CORE IMBALANCE USING THE OUT-OF-CORE DETECTOR SYSTEM, and compare to COLA limits and make final approval recommendation. Modified Bank JPM. Calculation is RO/SRO common.
EQUIPMENT CONTROL: Given a set of conditions and Reactor Coolant Pump electrical prints, determine if and where the Reactor Coolant Pump can be operated and any Tech Spec implications.
NewJPM.
EQUIPMENT CONTROL: Given a set of conditions with a gas release in progress and various RMS components out of service, determine the actions to be taken lAW facility procedures. Modified Bank JPM RADIATION CONTROL: Given a set of conditions, determine and declare the appropriate Emergency Action Level and make a Protective Action Recommendation lAW with the TMI Emergency Plan. Bank JPM.
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Three Mile Island Date of Examination: APR 2014 Exam Level: RO 1Zl SR0-1 D SRO-U D Operating Test Number: 289-2014-301 Control Room Systems@ (8 for RO); (7 for SR0-1); (2 or 3 for SRO-U, including 1 ESF)
Safety System I JPM Title Type Code*
Function
- a. Manually Initiate ESAS (006 A2.12) UDIAIS 2
- b. Restore Seal Injection with a Loss of ICCW (003 K6.02) NIA/S 4P
- d. Perform Emergency Operations of Reactor Building Emergency Cooling MIA/S 5 Water (Sys 022) A4.04
- e. Energize 1E Bus from SBO (Sys 064 A4.01) UDIAIS 6
- f. Respond lAW OP-TM-MAP-C0101 with Failure (Sys 072) A3.01 A/PIS 7
- g. Initiate and Isolate a Reactor Building Purge (Sys 029 K1.01) DIUAIS 8
- h. Feed from the "C" RCBT and the BAMT (001 A4.02) NIS 1 In-Plant Systems@ (3 for RO); (3 for SR0-1); (3 or 2 for SRO-U)
- k. lnit EB IP (Sys 004) G2.1.30 DIEIR 1
@ All RO and SR0-1 control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO I SR0-1 I SRO-U (A)Iternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank -:::_91 -:::_8 I -:: _ 4 (E)mergency or abnormal in-plant ::: 1 I > 1 I ::: 1 (EN)gineered safety feature - I - I ::: 1 (control room system (L)ow-Power I Shutdown >1I > 1 I ;:: 1 (N)ew or (M)odified from bank including 1(A) ;::21 >2 I ;:: 1 (P)revious 2 exams -:: _ 3 I -:::_3 I -:: _ 2 (randomly selected)
(R)CA ::: 1 I > 1 I ::: 1 (S)imulator ES-301 , Page 23 of 27
THREE MILE ISLAND 2014 NRC RO EXAMINATION JPM A- Manually Initiate ESAS. Alternate Path due to no ESAS component response when manual pushbuttons are operated requires the candidate to navigate out of the normal section and into an alternate section of OP-TM-211-901.
Safety significance: Failure to manually initiate ESAS will result in a loss of Reactor Coolant System Inventory.
JPM B- Restore Seal Injection with a Loss of ICCW. New JPM. Candidate will be directed to restore Seal Injection after a trip of the "B" Makeup Pump. Alternate Path 1 due to "A" Makeup Pump not starting, candidate will be required to navigate to a different section of OP-TM-AOP-041. Alternate Path 2 will occur with a loss of ICCW. With no Seal Injection and no ICCW, there is a navigation point (IAAT) to trip the reactor lAW OP-TM-EOP-001 and trip RCP's.
Safety significance: Failure to complete this JPM will result in an extended loss of RCP Seal cooling.
JPM C- Transfer Feedwater Pump From ICS to the Motor Speed Changer. Bank JPM.
Safety significance: Failure to complete this JPM successfully will lead to a Feedwater Pump Trip and a Plant Runback/possible Reactor Tnp.
JPM D- Perform Emergency Operations of Reactor Building Emergency Cooling Water. Modified Bank JPM. Alternate path. While Initiating RB Emergency Cooling, one valve will not open and cooler oulet pressures are too high, so the candidate will be required to navigate out of the normal section and into an alternate section of OP-TM-534-901.
Safety significance: Failure to properly control RB Emergency Cooling will result in elevated RB Pressures and Temperatures.
JPM E - Energize the 1E Bus from the SBO. Bank JPM. Involves restoring the "B" train ES Bus from Station Blackout Diesel power to normal Off-Site power.
Safety significance: Improper operation could lead to loss of 1 train of ES components through loss of ES bus, causing significant degradation in coping capability.
JPM F- Respond lAW OP-TM-MAP-C0101 Alarm Response with Failure. Previous NRC. Alternate Path JPM. Automatic actions do not occur lAW the alarm response, requiring navigation to another procedure.
Safety significance Failure to place control tower on "recirculation" following high airborne contamination in the Control Room may result in unnecessary dose for the personnel that must remain to operate the plant.
JPM G - Initiate and Isolate a Reactor Building Purge. Bank JPM. Alternate Path. RB Purge is initiated and is followed by an alarm, requiring navigation to an alarm response and aRB purge isolation.
Safety significance: Failure to complete this JPM will result in an uncontrolled release to the public to occur.
JPM H- Feed from the "C" RCBT and the BAMT. New JPM. A Feed is calculated, lined up, and performed from multiple sources simultaneously.
Safety significance: Failure to complete this JPM correctly will result in the wrong amount of Boron being inserted 1nto the RCS, which will cause a reactivity event.
JPM I - Return a Battery Charger to Service. Bank JPM.
Safety significance: Failure to complete the JPM properly will result in a degraded condition of the Station Batteries.
JPM J- Perform a Cooldown Outside of the Control Room. Bank JPM. Candidate will perform IMA's of OP-TM-EOP-020.
Safety significance: Failure to complete the JPM properly will result in an unstable critical condition for the Reactor and/or the Secondary Plant.
JPM K- Initiate Emergency Boration In Plant. Bank JPM. RCA entry required. Candidates will initiate an Emergency Boration from outside of the Control Room.
Safety significance: Failure to complete the task will result in insufficient shutdown boron concentration.
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Three Mile Island Date of Examination: APR 2014 Exam Level: RO D SR0-1 [gl SRO-U D Operating Test Number: 289-2014-301 Control Room Systems@ (8 for RO); (7 for SR0-1); (2 or 3 for SRO-U, including 1 ESF)
Safety System I JPM Title Type Code*
Function
- a. Manually Initiate ESAS (006 A2.12) UDIAIS 2
- b. Restore Seal Injection with a Loss of ICCW (003 K6.02) NIA/S 4P
- d. Perform Emergency Operations of Reactor Building Emergency Cooling MIA/S 5 Water (Sys 022) A4.04
- e. Energize 1E Bus from SBO (Sys 064 A4.01) UDIAIS 6
- f. Respond lAW OP-TM-MAP-C0101 with Failure (Sys 072) A3.01 A/PIS 7
- g. Initiate and Isolate a Reactor Building Purge (Sys 029 K1.01) DIUAIS 8 h.
In-Plant Systems@ (3 for RO); (3 for SR0-1); (3 or 2 for SRO-U)
- k. lnit EB IP (Sys 004) G2.1.30 DIEIR 1
@ All RO and SR0-1 control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO I SR0-1 I SRO-U (A)Iternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank ::_91 ::_8 I ::_ 4 (E)mergency or abnormal in-plant ~1 I >1 I ~1 (EN)gineered safety feature - I - I ~ 1 (control room system (L)ow-Power I Shutdown ~1 I >1 I ~1 (N)ew or (M)odified from bank including 1(A) ~21 ~2 I ~ 1 (P)revious 2 exams ::_ 3 I ::_3 I :::;. 2 (randomly selected)
(R)CA ~ 1I ~1 I ~1 (S)imulator ES-301, Page 23 of 27
THREE MILE ISLAND 2014 NRC RO EXAMINATION JPM A- Manually Initiate ESAS. Alternate Path due to no ESAS component response when manual pushbuttons are operated requires the candidate to navigate out of the normal section and into an alternate section of OP-TM-211-901.
Safety significance: Failure to manually initiate ESAS will result in a loss of Reactor Coolant System Inventory.
JPM B- Restore Seal Injection with a Loss of ICCW. New JPM. Candidate will be directed to restore Seal Injection after a trip of the "B" Makeup Pump. Alternate Path 1 due to "A" Makeup Pump not starting, candidate will be required to navigate to a different section of OP-TM-AOP-041. Alternate Path 2 will occur with a loss of ICCW. With no Seal Injection and no ICCW, there is a navigation point (IAAT) to trip the reactor lAW OP-TM-EOP-001 and trip RCP's.
Safety significance: Failure to complete this JPM will result in an extended loss of RCP Seal cooling.
JPM C- Transfer Feedwater Pump From ICS to the Motor Speed Changer. Bank JPM.
Safet~ significance: Failure to complete this JPM successfully will lead to a Feedwater Pump Trip and a Plant Runback/possible Reactor Tnp.
JPM D- Perform Emergency Operations of Reactor Building Emergency Cooling Water. Modified Bank JPM. Alternate path. While Initiating RB Emergency Cooling, one valve will not open and cooler oulet pressures are too high, so the candidate will be required to navigate out of the normal section and into an alternate section of OP-TM-534-901.
Safety significance: Failure to properly control RB Emergency Cooling will result in elevated RB Pressures and Temperatures.
JPM E- Energize the 1E Bus from the SBO. Bank JPM. Involves restoring the "B" train ES Bus from Station Blackout Diesel power to normal Off-Site power.
Safety significance: Improper operation could lead to loss of 1 train of ES components through loss of ES bus, causing significant degradation in coping capability.
JPM F- Respond lAW OP-TM-MAP-C0101 Alarm Response with Failure. Previous NRC. Alternate Path JPM. Automatic actions do not occur lAW the alarm response, requiring navigation to another procedure.
Safety significance Failure to place control tower on "recirculation" following high airborne contamination in the Control Room may result in unnecessary dose for the personnel that must remain to operate the plant.
JPM G - Initiate and Isolate a Reactor Building Purge. Bank JPM. Alternate Path. RB Purge is initiated and is followed by an alarm, requiring navigation to an alarm response and aRB purge isolation.
Safety significance: Failure to complete this JPM will result in an uncontrolled release to the public to occur.
JPM H - Not selected for SROs.
JPM I - Return a Battery Charger to Service. Bank JPM.
Safety significance: Failure to complete the JPM properly will result in a degraded condition of the Station Battenes.
JPM J- Perform a Cooldown Outside of the Control Room. Bank JPM. Candidate will perform IMA's of OP-TM-EOP-020.
Safety significance: Failure to complete the JPM properly will result in an unstable critical condition for the Reactor and/or the Secondary Plant.
JPM K- Initiate Emergency Boration In Plant. Bank JPM. RCA entry required. Candidates will initiate an Emergency Boration from outside of the Control Room.
Safety significance: Failure to complete the task will result in insufficient shutdown boron concentration.
Appendix D Scenario Outline Form ES-D-1 Facility: Three Mile Island Scenario No.: 1 Op Test No.: 289-2014-301 Examiners: Operators:
Initial Conditions: * (Temporary IC-xxx)
- 100% Power, MOL.
- SBO OOS For Maintenance, expected to return to service in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> .
- Crane work is occurring on the West side of the Plant to stage new piping .
Turnover: Maintain 100% Reactor Power Critical Tasks:
- Isolate Possible RCS Leak Paths (CT-3)
- Establish Natural Circulation RC Flow (CT-12)
- Turbine Trip (CT-18)
Event No. Malt. No. Event Type* Event Description 1 DHR32 TSCRS BWST Level Lowers, entry into OP-TM-MAP-E0204.
2 MS03A CCRS lsolable Steam Leak in Turbine Bldg, entry into OP-TM-AOP-051.
CARO (ARO: Isolate Steam Leak) 3 IC23 ICRS SG/RX Demand Fails to Zero Volts, entry into OP-TM-AOP-070.
IURO IARO (URO/ARO: Coordinate to stabilize plant in ICS HAND control) 4 TU01D NCRS High Vibrations on Main Turbine, entry into OP-TM-MAP-K0201 and 1102-4.
R URO (URO: Lower Reactor Power, ARO: Manually Adjust Feedwater)
NARO 5 ESOBA ICRS Inadvertent 1600# ESAS Signal, entry into OP-TM-AOP-046.
IURO IARO (URO: Immediate Manual Actions, ARO: Restores Letdown) 6 EDR14 TSCRS Loss of 1D 4k Bus, entry into OP-TM-AOP-013.
EGO?A CARO (ARO: Places LO-P-6 in PTL) 7 ED12 ICRS Loss of Stator Coolant Pumps, Main Turbine fails to automatically trip.
IURO (URO: Trip Reactor, ARO: Adjust Main Feedwater) 8 ED01 MCRS Loss of Offsite Power, entry into OP-TM-AOP-020.
MURO MARO 9 EG07B CCRS EG-Y-1 B Trips, Station Blackout.
CURO (URO: Isolates Cooling paths to RCP's)
- (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor Scenario Set-up NRC Scenario 1 Three Mile Island NRC Scenario #1 Event #1: When the crew has accepted the watch, the Lead Examiner will cue the lowering level of the BWST due to a crane piercing the tank at the 53.5 ft level.
The crew will diagnose the low level in the BWST by lowering level on the BWST level indicators (CC and CR) and annunciator E-3-4 in alarm. Although it initially drops rapidly, the crew will identify it as steady at approximately 54.5 ft. The CRS will review and declare the following Tech Spec: 3.3.1.1.a.
Once the Tech Spec has been declared, the scenario can continue.
Event #2: The Lead Examiner will cue the lsolable Steam Leak in the Turbine Building.
The operators will diagnose a Secondary Side Steam Leak based on a lowered efficiency of the Secondary Plant (Megawatts, Header Pressure, OTSG pressures, etc.)
Upon a Steam Leak in the Turbine Building, the following Critical Safety Functions are affected:
CSF 1, Reactivity and Reactor Power Control: Maintain control of the fission process, maintain the capability to shutdown the reactor and the capability to maintain the reactor in a shutdown condition. Control energy production and reactor power distribution based on design limits and current core heat removal capability. A secondary side steam leak will bypass steam away from the Turbine. If the steam leak is large enough, electrical generation will lower. ICS will raise reactor power to raise MWe back up to ICS demanded MW. If the steam leak causes Tave to lower, ICS will also try to raise reactor power to recover Tave.
CSF 7, Electrical Power: Provide electrical power as required to accomplish the other Critical Safety Functions. Provide AC and DC power for emergency equipment operation and instrumentation systems. A Steam Leak may affect power to non-safety related equipment in the Turbine Building.
CSF 9, Fire Protection and Remote Shutdown Capability: Maintain means to prevent detect and suppress fires, as well as the capability to perform a plant shutdown without access to the Control Room. A Steam Leak may cause fire alarms to actuate. This could be the first indication of a Steam Leak to the Control Room. May cause sprinkler systems to actuate or fire dampers to close.
The crew will diagnose the Steam Leak and the CRS will enter OP-TM-AOP-051, Secondary Side High Energy Leak. OP-TM-AOP-051 is entered for Steam Leaks that affect large portions of the plant and therefore it is not obvious to the operator what needs to be done initially to isolate the leak.
Scenario Set-up NRC Scenario 1 The affected Building is determined and from there, OP-TM-AOP-051 systematically attempts to isolate the leak remotely from the Control Room while taking steps to minimize the adverse effects of a steam environment on safety related equipment. The OP-TM-AOP-051 mitigation strategy for a Steam Leak in the Turbine Building is as follows:
- Attempt to isolate the leak from the Control Room.
- Shutdown and Cooldown the plant in a controlled manner to minimize pressure surges that could make the leak worse. Shutdown may have to be done quickly or the plant may have to be tripped depending on the circumstances.
The ARO will isolate Steam Leak by closing the appropriate valve, MS-V-5B. This can be performed because the steam supplies to the Main Feedwater Pumps are as follows:
- Below 25% power, Main Steam, only, supplies the Main Feedwater Pumps.
- Between 25% and 40% power, Main Steam supplements Extraction Steam as supplies to the Main Feedwater Pumps.
- Above 40% power, Extraction Steam, only, supplies the Main Feedwater Pumps.
Once the Steam Leak has been isolated, the scenario can continue.
Event #3: The Lead Examiner will cue the failure of the SG/RX Demand Station to Zero Volts. This will cause an ICS transient, which if not responded to swiftly, will cause a Reactor Trip.
The crew will diagnose the ICS failure by a rapid change in RCS pressure, Reactor Power rising, multiple annunciator alarms, and/or changes in indications at multiple ICS stations. Entry into OP-TM-AOP-070, PRIMARY TO SECONDARY HEAT TRANSFER UPSET will be required based on RCS pressure not being controlled in ICS AUTO.
"RCS pressure is not being controlled" requires the operator to make a subjective determination, based on their skills, training, and experience. A determination that RCS pressure is being controlled should include the following elements: 1) The reason for the transient is understood 2) RCS pressure response is consistent with the expected response for the event 3) Automatic or manual control in accordance with normal operating procedures is effectively controlling RCS pressure. A conservative assessment (i.e. concluding that RCS pressure is not being controlled) is appropriate when the three conditions above cannot be satisfied. ICS failures are one class of events that can lead to an upset in primary to secondary heat transfer. Most ICS failures can be mitigated by use of the appropriate manual control normal operating procedures.
This entry into AOP-070 is unique from the other scenarios because Reactor Power will lower while RCS pressure rises. The URO will have to place the Diamond Panel in manual to control Reactor Power or RPS will trip the reactor on high RCS pressure.
Once the plant is stabilized in ICS HAND control, the scenario can continue.
Scenario Set-up NRC Scenario 1 Event #4: The Lead Examiner will cue the High Vibrations on Main Turbine.
The crew will diagnose the High Vibrations on Main Turbine by Annunciator K-2-1 in alarm, and multiple PPC points in alarm. The crew will commence a power reduction to
< 45% with ICS in manual to trip the Main Turbine. This is the reactivity manipulation for the scenario.
Once sufficient reactivity manipulation has occurred, the scenario can continue.
Event #5: The Lead Examiner will cue the Inadvertent 1600# ESAS Signal. The crew must quickly recognize the condition and perform the required Immediate Manual Actions to minimize the RCS pressure transient and pressurizer in-surge due to HPI.
Additionally, while at power, immediately reducing HPI also minimizes the possibility of a reactor trip on high RCS pressure. The following Critical Safety Functions are affected by an inadvertent ESAS signal:
CSF 1, Reactivity & Reactor Power Control: Maintain control of the fission process, maintain the capability to shutdown the reactor and the capability to maintain the reactor in a shutdown condition. Control energy production and reactor power distribution based on design limits and current core heat removal capability. HPI Actuation will insert negative reactivity as borated water from the BWST will be injected into the RCS. ICS will pull control rods to maintain ULD demand and RCS Tave. Potential all rods out condition. As primary side power decreases, ICS cross limits will lower feedwater flow in an attempt to match primary to secondary heat removal.
CSF 2, Reactor Vessel Inventory Control: Provide the means to maintain the core covered with sub cooled water. HPI Actuation will isolate RCS letdown and normal makeup. HPI will cause pressurizer level to rise. MU tank level will lower (MUT level/pressure requirements). BWST level will lower (BWST TS level).
CSF 3, RCS Integrity: Maintain the capability to control heatup and cooldown rates and control RCS pressure to prevent reactor vessel brittle fracture or LTOP events. Maintain RCP seal cooling to prevent excessive loss of RCS inventory through RCP seals. HPI Actuation will cause RCS pressure to rise. Seal Injection is not affected by an ES actuation. AOP-046 has the operator secure the remaining MU pump if the MU-V-16s cannot be closed from the control room.
This action immediately terminates HPI and seal injection. The thermal barrier heat exchangers provide adequate seal cooling when the Sl is secured. AOP-046 ensures adequate thermal barrier cooling prior to terminating seal injection.
Once MU-V-36 and 37 are opened locally and the appropriate MU-V-16s are closed, a makeup pump is restarted and Sl is re-established.
CSF 4, Core Heat Removal: Provide the capability to remove core heat production at all times. HPI Actuation will inject cold BWST into the RCS. The RCS will also cool due to the negative reactivity insertion from the injected BWST water lowering core power.
Scenario Set-up NRC Scenario 1 CSF 5, Containment Integrity: Provide means to prevent or minimize fission product release to the environment. (1) Maintain containment pressure below design and (2) Provide capability to isolate the containment when required. An inadvertent HPI Actuation will start the Reactor River system, and the RB cooling fans will be running in slow speed. Normal cooling to the RB cooling fans will be isolated. RB Temperature and pressure will lower due to the actuation of the RR system. The degree of temperature reduction would depend of river temperature. Building Spray System valves will open to align the BWST to the RB, but the BS pumps would not start unless a 30# signal is present. Closure of the containment isolation valves under the HPI signal would not adversely affect their associated systems.
CSF 7, Electrical Power: Provide electrical power as required to accomplish the other Critical Safety Functions. Provide AC and DC power for emergency equipment operation and instrumentation systems. An inadvertent HPI Actuation will start the emergency diesel generators unloaded.
CSF 8, Auxiliary Emergency Systems: Provide equipment cooling (closed cooling
& ventilation), and other support requirements to accomplish the other Critical Safety Functions. Provide Instrument Air for operation of EFW, ADVs, RCP Support Systems and some containment isolation valves. An inadvertent HPI Actuation will start support systems to support ECCS and RB cooling systems.
The DC and DR pumps will start to support MUP and DHP cooling which would be running during an inadvertent actuation. ES selected NS and NR pumps will start. Two NR and NS pumps are normally running. There is a potential for three NS pumps running which would start an overcooling of the NS system.
CSF 9, Fire Protection & Remote Shutdown Capability: Maintain means to prevent, detect, and suppress fires, as well as the capability to perform a plant shutdown without access to the Control Room. An inadvertent HPI Actuation would trip FS-P-2.
The crew will diagnose the Inadvertent "A" 500# ESAS Signal by multiple annunciators in alarm, "A" Train components in their ES actuated state, and/or "A" EDG running, while all primary indications appear steady or rising (RCS pressure not at 500#}.
The URO will perform the Immediate Manual Actions of OP-TM-AOP-046, INADVERTANT ESAS.
The ARO will restore letdown lAW OP-TM-211-950 (performing the appropriate portion of the procedure when restoring from isolation following an ESAS signal).
Once the plant is stabilized and Letdown is restored, the scenario can continue.
Event #6: The Lead Examiner will cue the Loss of the 1D 4Kv Bus.
Upon a Loss of the 1D 4Kv Bus, the following Critical Safety Functions are affected:
Scenario Set-up NRC Scenario 1 CSF 1, Reactivity & Reactor Power Control: Maintain control of the fission process, maintain the capability to shutdown the reactor and the capability to maintain the reactor in a shutdown condition. Control energy production and reactor power distribution based on design limits and current core heat removal capability. A loss of the 1D 4Kv bus will not affect reactor shutdown capability.
Emergency boration must be performed using "B" Train as MU-V-14A, MU Pump Suction Valve from BWST, will be unavailable.
CSF 2, Reactor Vessel Inventory Control: Provide the means to maintain the core covered with sub cooled water. A loss of the 1D 4Kv bus will cause Train A Emergency Core Cooling Systems (HPI and LPI) to be inoperable. Train B is unaffected.
CSF 3, RCS Integrity: Maintain the capability to control heatup and cooldown rates and control RCS pressure to prevent reactor vessel brittle fracture or LTOP events. Maintain RCP seal cooling to prevent excessive loss of RCS inventory through RCP seals. A loss of the 1D 4Kv bus will cause RC-V-1, Pressurizer Spray Valve, and the Emergency power to Pressurizer Group 8 Heaters to be unavailable.
CSF 4, Core Heat Removal: Provide the capability to remove core heat production at all times. A loss of the 1D 4Kv bus will cause EF-P-2A and DHR Train A to be inoperable. EF-P-2B, EF-P-1, and DHR Train Bare available.
CSF 5, Containment Integrity: Provide means to prevent or minimize fission product release to the environment. {1) Maintain containment pressure below design and (2) Provide capability to isolate the containment when required. A loss of the 1D 4Kv bus causes Reactor Building Emergency Cooling Train A and Building Spray Train A to be unavailable. Train B is available as well as Normal Reactor Building Cooling.
CSF 7, Electrical Power: Provide electrical power as required to accomplish the other Critical Safety Functions. Provide AC and DC power for emergency equipment operation and instrumentation systems. A loss of the 1D 4Kv bus due to a bus fault will prevent use of the Emergency Diesel Generators.
Prolonged loss of AC increases the risk of a Loss of DC and Vital busses.
CSF 8, Auxiliary Emergency Systems: Provide equipment cooling (closed cooling
& ventilation), and other support requirements to accomplish the other Critical Safety Functions. Provide Instrument Air for operation of EFW, ADVs, RCP Support Systems and some containment isolation valves. A loss of the 1D 4Kv bus will cause Train A of all emergency cooling support systems (Control Building cooling, Nuclear Services Closed Cooling Water cooling, Intermediate Closed Cooling water cooling, River Water Systems, Equipment ventilation, etc.)
to be unavailable. Train B is unaffected.
CSF 9, Fire Protection & Remote Shutdown Capability: Maintain means to prevent, detect, and suppress fires, as well as the capability to perform a plant shutdown without access to the Control Room. A loss of the 1D 4Kv bus will cause the Relay Room Cardox and Screen House detection systems to be inoperable. Fire watches will be initiated.
The operators will diagnose a Loss of the 1D 4Kv Bus based on a loss of running equipment powered from the 1D 4Kv bus (Secondary Closed Cooling Water Pumps, Secondary River Water Pumps, Intermediate Closed Cooling Water Pumps, etc.), LO-P-6 starting, and half of the Control Room lighting out.
Scenario Set-up NRC Scenario 1 The CRS will enter OP-TM-AOP-013, LOSS OF 1D 4160V BUS. The ARO will place the Control Switch for LO-P-6 to Pull-To-Lock to minimize loading on the Battery. The CRS will evaluate and declare Tech Spec 3.7.2.f:
- T.S. 3.7.2- The reactor shall not remain critical unless all of the following requirements are satisfied:
o 3.7.2.f- The Engineered Safeguards bus, switchgear, load shedding, and automatic diesel start systems shall be operable except as provided in Specification 3.7.2.c above and as required for testing.
Once LO-P-6 is in Pull-To-Lock and the Tech Spec has been declared, then the scenario can continue.
Event #7: The lead examiner will cue the Loss of Stator Coolant Pumps, causing a lack of Main Generator Stator Cooling.
The crew will diagnose a Loss of Stator Cooling Pumps by Main Annunciators L-2-7, GEN STATOR STBY CLG PUMP RUN, and Main Annunciator L-1-7, GEN STATOR CLG LOSS RUNBACK, in alarm, and no operating Stator Coolant Pump indications on PCL.
The crew will enter OP-TM-MAP-L01 07 and OP-TM-MAP-L0207. The ARO will attempt to start the standby pump from PCL and will identify that it will not start. The crew will identify that the Turbine Control Valves are not closing as expected and, if the condition continues for more than 3.5 minutes, that the Main Turbine did not automatically trip.
- lAW OP-TM-MAP-L01 07:
o Automatic Actions:
- Turbine Control Valves close at 23.4% vmo/min.
- Turbine trips after 2 minute time delay should turbine not runback to approximately 95% power.
- Turbine trips after 3.5 minute time delay should turbine not runback to approximately 29.6% power.
The CRS will enter OP-TM-EOP-001, Reactor Trip. The URO will trip the reactor (CT-18). The Main Turbine will fail to automatically trip and the URO will trip it manually.
The ARO will adjust Main Feedwater flow to avoid an overcooling event.
Once the Reactor and Main Turbine have been tripped and a symptom check has been performed, then the scenario can continue.
Event #8: The lead examiner will cue the Loss of Offsite Power.
Scenario Set-up NRC Scenario 1 The crew will diagnose a Loss of Offsite Power by the Control Room lighting going dark for 10 seconds, followed by half of the lights returning once the B Emergency Diesel Generator powers the 1E 4Kv bus, as well as several alarms indicating a loss of the 4 bus and 8 bus, and a loss of seal injection and letdown. The CRS will direct entry into OP-TM-AOP-020, Loss of Station Power.
OP-TM-AOP-020 addresses two types of Loss of Station Power events. (1) Loss of Offsite Power with one or both diesels supplying the ES 4160V busses and (2) a loss of offsite power with no diesels supplying the ES 4160V busses.
- If one or both diesels are supplying the ES busses, the procedure walks the operators through ensuring RCS cooling and RCP seal cooling is established. As long as one diesel is available, the reactor can be stabilized and the operators can methodically walk through powering additional busses and starting additional equipment.
The ARO will establish Natural Circulation by feeding the OTSG's lAW Rule 4 (CT-12) and verify that Natural Circulation exists lAW Guide 10, Natural Circulation.
- lAW Guide 10:
o Natural Circulation exists when the following conditions exist:
- RCS differential temperature develops and stabilizes
<50 F.
- Thot < 600F.
- lncore temperatures stabilize and track Thot indications.
- Primary to Secondary heat transfer is demonstrated by steaming or feeding OTSGs.
- Adequate SCM exists.
Once Natural Circulation has been established, then the scenario can continue.
Event #9: The lead examiner will cue the Loss of EG-Y-1 B.
The crew will diagnose a Loss of Offsite Power by the Control Room lighting going dark as well as several alarms indicating a loss of EG-Y-1 B. The CRS will direct entry into OP-TM-AOP-020, Loss of Station Power, Section 4.0, Station Blackout.
OP-TM-AOP-020 addresses two types of Loss of Station Power events. (1) Loss of Offsite Power with one or both diesels supplying the ES 4160V busses and (2) a loss of offsite power with no diesels supplying the ES 4160V busses.
- In a Station Blackout (no power to either 4160V ES bus), immediate attempts are made to energize the ES busses. Actions are taken to prevent inventory loss, including protecting the RCP seals. The operators take action to minimize DC loads to maximize station battery life.
Scenario Set-up NRC Scenario 1 The URO will place Intermediate Closed Cooling Water Pumps in Pull-To-Lock and ensure that RCP Seal Return and Seal Injection Valves are closed (CT-3}.
The scenario can be terminated when the Main Turbine has been tripped, Natural Circulation has been established, Intermediate Closed Cooling Water Pumps have been placed in Pull-To-Lock, and RCP Seal Return and Seal Injection lines have been isolated.
Scenario Set-up NRC Scenario 1 B&W Unit EOP Critical Task Description Document, 47-1229003-04:
CT Isolate Possible RCS Leak Paths- All isolable leaks should be isolated, if possible.
There are several possible leaks in the RCS which may be isolated by closing certain valves.
These may include:
- Pressurizer Spray Valve and Spray Block Valve
- Pressurizer Vent Valves.
- Pressurizer Sample Valves.
- Hot Leg High Point Vent Valves.
- Reactor Vessel Head Vent Valves.
Station Blackout procedures include provisions for minimizing primary inventory losses, including:
- Isolating all Letdown flows.
- Closing Seal Return Valves.
- Isolating other known leak paths.
Protecting against RCP Seal LOCA outside of the following limits should be considered grounds for failure of the critical task:
- Isolating Seal Return Lines while any Reactor Coolant Pump #1 seal inlet temperature > 235°F.
- Isolating Intermediate Closed Cooling Water System while any Reactor Coolant Pump #1 seal inlet temperature > 235°F.
Safety Significance: Elevated RCP seal temperatures will result in increased seal leakage of approximately 21 gpm I pump. To avoid seal damage and excessive seal leakage, do not restore RCP seal injection.
To avoid water hammer, thermal barrier cooler damage, and RCS leakage to the ICCW system, do not restore RCP thermal barrier cooling.
Either Seal Injection restoration or Intermediate Closed Cooling Water restoration would lead to a reduction of RCS inventory.
Cues:
- 1. Computer indication for RCP #1 Seal Inlet Temperature Performance Indicators:
- 1. Operation of associated RCP Seal Valve controls.
- 2. Operation of associated RCP Seal Pump controls.
Feedback:
- 1. Indications of Valve status indications associated with RCP Seals.
- 2. Indications of Pump status indications associated with RCP Seals.
Scenario Set-up NRC Scenario 1 B&W Unit EOP Critical Task Description Document, 47-1229003-04:
CT Establish Natural Circulation RC Flow- Whenever forced RC flow is not available, NC flow should be established. Maintaining primary to secondary heat transfer via NC eliminates the need to add RC to the RB as would occur with the back up feed and bleed HPI core cooling mode.
o If primary to secondary heat transfer has been lost, then establish and maintain appropriate SG levels in accordance with Rule 4.0.
o Reduce SG pressure using the TBVs/ADVs to establish a positive primary to secondary side ~T of - 50°F.
o RCS pressure should be maintained constant or slightly increasing using MU or HPI. RCS pressure should not be increased if PTS guidance is invoked.
o Trying to establish Natural Circulation RC flow outside of the following limits should be considered grounds for failure of the critical task:
- Establish Emergency Feedwater flow lAW Rule 4 to each OTSG (less than 515 gpm total), with a target band of 50-85% in the Operating Range.
- Establish Natural Circulation prior to transitioning into OP-TM-EOP-009, HPI Cooling.
Safety Significance: Enhances the transient mitigation capability of the plant by maintaining SGs operable and eliminates the need to add RC to the RB as with HPI Cooling.
Cues:
- Low RC flow alarm
- Verbal alert by plant staff that all RCPs have tripped
- SCM monitor and associated alarms
- P-T display and associated alarms Performance Indicators:
- Operation of EFW/FW pump and valve controls
- Operation of TBV/ADV controls
- Operation of MU/HPI pump and valve controls Feedback:
- Verbal verification that natural circulation has been established
- SG pressure
- RC temperature Scenario Set-up NRC Scenario 1 B&W Unit EOP Critical Task Description Document, 47-1229003-04:
CT Turbine Trip- Whenever conditions exist such that a reactor trip is required, then the normally redundant actions of tripping the reactor and main turbine should be accomplished immediately. Tripping the main turbine provides assurance of a redundant trip signal to the main turbine electro-hydraulic control unit.
- Due to Excessive Heat Transfer concerns, tripping the Main Turbine outside of the following limit should be considered grounds for failure of the critical task:
- Trip the Main Turbine to ensure that it does not cause a continuous excessive cooldown rate or Tcold to be less than 329°F.
- Once the reactor is shutdown, prompt isolation of the turbine steam flow path is significant. Until the major steam flow path through the turbine to the condenser is isolated, RCS heat removal will be much larger than heat generation, and a rapid RCS cooldown will continue. (Source: OP-TM-EOP-0011)
Safety Significance: When the reactor is tripped (shutdown), steam flow to the main turbine must be stopped in order to maintain the appropriate primary to secondary heat balance. When the appropriate primary to secondary heat balance is established, the normal heat removal systems are available for plant control thus enhancing the transient mitigation capability of the plant. If the turbine steam flow path is not isolated after a rapid reduction in reactor heat generation (reactor trip), extremely rapid RCS cooldown is possible. Prompt operator action can minimize the extent of this overcooling and potential consequences to RCS pressure boundary. A prolonged rapid cooldown will complicate plant control and could challenge OTSG tube integrity or Reactor Vessel integrity.
Cues:
- Visual indications (closed generator output and exciter breakers, main turbine stop and control valves are not closed)
- P-T display and associated alarms
- Verbal alert by plant staff that all main turbine stop and control valves are not closed immediately following actuation of a reactor trip signal
- Verbal alert by plant staff that main alternator output/exciter breakers are not open immediately following actuation of a reactor trip signal Performance Indicators:
- Operation of control room manual main turbine trip pushbutton
- Main turbine trip alarm
- Main turbine-generator exciter alarms
- Main turbine-generator breaker status alarms Scenario Set-up NRC Scenario 1 B&W Unit EOP Critical Task Description Document, 47-1229003-04:
CT Turbine Trip (cont)
Feedback:
- RC temperature and pressure
- SG level and pressure
- Mega-Watt electric indication
- Main turbine-generator breaker status indications
- Verbal notification by plant staff of main turbine trip status Scenario Set-up NRC Scenario 1 Industry Experience:
- 1. OE28735 - Main Turbine High Vibration Trips (Palo Verde Unit 1 and 3) (5/5/09)
- 2. Fort St. Vrain Loss of all AC Power (Blackout) (1 0/27/83)
- 3. SOER 99-1 Loss of Grid (12/99)
- 5. OE25328 - Engineered Safeguards Actuation System (ESAS) Relay Failure Contributes to lnadvertant Safety System Actuation (Three Mile Island Unit 1)
(6/27/07)
- Diesel Generator 1A loss (Risk Increase Factor)
- Loss of Offsite Power (Initiating Event)
Appendix D Scenario Outline Form ES-D-1 Facility: Three Mile Island Scenario No.: 2 Op Test No.: 289-2014-301 Examiners: Operators:
Initial Conditions: * {Temporary IC-165)
- 100% Power, MOL
- SBO OOS For Maintenance, expected to return to service in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />
- Crane work is occurring on the West side of the Plant to stage new piping .
Turnover: Maintain 100% Reactor Power Critical Tasks:
- Minimize SCM (CT-7)
- Limit Uncontrolled Radiation Release (CT-21)
- Reduce Steaming/Isolate Affected SGs (CT-22)
Event No. Malt. No. Event Type* Event Description 1 MSR01 CCRS MSIV Inadvertent Closure, entry into OP-TM-PPC-L2204.
CURO CARO (URO: Lowers power in ICS Auto, ARO: Opens MS-V-1A}
2 RW02C TSCRS NR-P-1 C Trips, NR-P-1 B Fails to Auto-Start, entry into OP-TM-MAP-B0105, and OP-TM-MAP-B0205 CARO (ARO: Starts NR-P-1 B from CR) 3 MU19D NCRS Reactor Coolant Pump Seal Leakage, entry into OP-TM-AOP-040.
~~ 1--CURO CARO
~URO: Lower Reactor Power, ARO: Manually Control Feedwater Pumps) 4 ED09D TSCRS Loss of Vital Bus D, entry into OP-TM-AOP-018 CARO (ARO: Restore Control Building Ventilation) 5 IC20 ICRS Total FW Demand Fails to Zero Volts, ICS Transient, entry into OP-TM-AOP-070.
IURO (URO/ARO: Coordinate to stabilize plant in ICS HAND control)
IARO 6 TH17B TSCRS -30 gpm "B" OTSG Tube Leak (TS), entry into OP-TM-EOP-005 R URO (URO: Guide 9)
NARO 7 TH16B MCRS -800 gpm "B" OTSG Tube Rupture with an elevated offsite dose, entry into OP*-TM-EOP-005 MURO MARO 8 Override CCRS MU-P-1A Overcurrent.
CURO (URO: Secure MU-P-1A}
. (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor NRC Scenario 2 Scenario Set-up Three Mile Island NRC Scenario #2 Event #1: When the crew has accepted the watch, the Lead Examiner will cue the Closure of MS-V-1 A. The crew will identify this by the green closed light lit, white test light lit (during travel) and the red open light not lit (after travel is complete) (CC). The crew will enter OP-TM-PPC-L2204, which will direct lowering power less than 90% and reopening MS-V1A.
When MS-V-1A has been opened, the scenario can continue.
Event #2: When the crew has accepted the watch, the Lead Examiner will cue trip of the "C" Nuclear River Pump. "B" Nuclear River Pump fails to auto-start in standby, leaving only one (1) Nuclear River Pump running. One Nuclear River Pump may not be sufficient to cool both the Nuclear Service Closed Cooling System (NSCCW) and the Intermediate Closed Cooling System (ICCW).
It is considered a loss of NSCCW if NSCCW temperatures reach 100°F, and the following Critical Safety Functions are affected:
CSF 4, Core Heat Removal: Provide the capability to remove core heat production at all times: Loss of Nuclear Services cooling function: RC pumps must shutdown. Natural Circulation will be used RCS heat removal.
CSF 8, Auxiliary Emergency Systems: Provide equipment cooling (closed cooling and ventilation), and other support requirements to accomplish the other Critical Safety Functions. Provide Instrument Air for operation of EFW, ADVs, RCP Support Systems, and some containment isolation valves: Loss of Nuclear Services cooling function: Other CSFs are affected as follows: (1) the reliability of safety related power sources and instrumentation system is degraded by the loss of the control building chillers and (2) the reliability of the decay closed pump motors and emergency feed pump motors is degraded by the loss of cooling to the area ventilation coolers.
CSF 10, Chemistry Control: Provide the means to monitor and control primary and secondary water chemistry in order to ensure the long term reliability of plant systems and limit the potential release of radioactive materials: Loss of Nuclear Services cooling function would result in the loss of the capability to obtain an RCS or OTSG sample.
It is considered a loss of ICCW if ICCW temperatures reach 120°F, and the following Critical Safety Functions are affected:
CSF 1, Reactivity and Reactor Power Control: Maintain control of the fission process, maintain the capability to shutdown the reactor and the capability to maintain the reactor in a shutdown condition. Control energy production and reactor power distribution based on design limits and current core heat removal capability. Loss of Intermediate Component Cooling: The reactor is tripped in the event of loss of cooling to the CAD stators in order to prevent stator damage.
Loss of CAD stator cooling would not prevent CAD insertion on RPS actuation.
Maintaining reactor shutdown is not affected by loss of IC component cooling.
NRC Scenario 2 Scenario Set-up CSF 3, RCS Integrity: Maintain the capability to control heatup and cooldown rates and control RCS pressure prevent reactor vessel brittle fracture or LTOP events. Maintain RCP seal cooling to prevent excessive loss of RCS inventory through RCP seals. Loss of Intermediate Component Cooling: One of two RCP seal cooling methods is lost. Loss of seal injection would require RCP shutdown.
If Sf is lost, overheating of RCP seals is likely. If seal injection maintained, solid operation may be required due to the loss of letdown.
The crew will diagnose the trip of NR-P-1C by an amber disagreement light on the NR-P-1 C control switch and Annunciator alarms A-1-5 and A-2-5. The ARO will manually start "B" Nuclear River Pump to provide sufficient cooling for NSCCW and ICCW. The CRS will identify and declare the following Tech Spec: 3.3.2.
When NR-P-1 B is running and the Tech Spec has been declared, the scenario can continue.
Event #3: The Lead Examiner will cue the Reactor Coolant Pump Seal Leak on RC-P-1D. lAW OP-TM-AOP-040, RCP #1 SEAL FAILURE, Basis Document:
The objective is to shutdown the RCP before any further significant RCP damage occurs. The goal is to attempt to reduce power to allow stopping pump without having to trip the reactor. If tripping the RCP would challenge RPS, then the reactor will be tripped, the IMAs and symptom check performed and then the affected RCP tripped.
These actions should be performed promptly. If seal leakoff flow exceeds 8 gpm or temperature exceeds limits at seal water to radial bearing or #1 seal inlet, the RCP must be shutdown immediately. Once an RCP is tripped, #1 seal return is isolated (close MU-V-33) and the #2 seal becomes the primary seal.
The crew will diagnose the #1 Seal Leak on RC-P-1 D by the PPC alarm, the digital recorder (PC), and Lab Seal D/P indicated abnormal (CC). The CRS will announce entry into OP-TM-AOP-040 and will determine that, based on Seal Leakage being greater than 6 gpm but less than 8 gpm, a power reduction will be made in order to secure the RCP.
The URO will have to reduce power with ICS in AUTO to secure the RCP. The ARO will place Main Feedwater Pumps to HAND control per 1102-4, Power Operation.
Once the "D" Reactor Coolant Pump is tripped, then the scenario can continue.
Event #4: The Lead Examiner will cue the Loss of Vital Bus "D".
The effects of a loss of VBD which are significant to plant safety or operation are numerous. For each effect the required compensatory action is described in OP-TM-AOP-018.
NRC Scenario 2 Scenario Set-up This procedure stabilizes the plant and performs compensatory actions for equipment failures. It is considered a loss of Vital Bus "D" if the OTSGs are being used for RCS heat removal and an unplanned deenergization of VBD has occurred, and the following Critical Safety Functions are affected:
CSF 1, Reactivity and Reactor Power Control: Maintain control of the fission process, maintain the capability to shutdown the reactor and the capability to maintain the reactor in a shutdown condition. Control energy production and reactor power distribution based on design limits and current core heat removal capability. Loss of VBD: Nl-4 and Nl-8 are lost, but the remaining channels of nuclear instrumentation and incore detectors provide sufficient information to control power level and reactor power distribution.
CSF 6, Radiation Control and Control Room Habitability: Monitor and control the release of radiation to the environment. Maintain access to critical plant equipment and use of the Control Room. Loss of VBD: RM-L-6, RM-A-7, RM-A-4, and RM-A-6 are deenergized. Compensatory actions will be taken lAW ODCM requirements. Access to plant equipment and Control Room is not affected.
CSF 7, Electrical Power: Provide electrical power as required to accomplish the other Critical Safety Functions. Provide AC and DC power for emergency equipment operation and instrumentation systems. Loss of VBD: VBD is deenergized. Ability to accomplish other Critical Safety Functions is not compromised.
CSF 8, Auxiliary Emergency Systems: Provide equipment cooling (closed cooling and ventilation), and other support requirements to accomplish the other Critical Safety Functions. Provide Instrument Air for operation of EFW, ADVs, RCP Support Systems and some containment isolation valves. Loss of VBD:
Ventilation will be lost to CB 322' Battery Rooms, Inverter Rooms, ES 480V Switchgear Rooms, and Remote Shutdown Area. Compensatory actions will be performed lAW OP-TM-AOP-034, "Loss of Control Building Cooling." Instrument air is not affected.
CSF 9, Fire Protection and Remote Shutdown Capability: Maintain means to prevent, detect and suppress fires, as well as the capability to perform a plant shutdown without access to the Control Room. Loss of VBD: Loss of PRF annunciators disables alarms PRF-5-1, Relay Room Fire, PRF-7-1, IWFS!TS Bldg Fire, PRF-7-6, UPS Fire, and PRF-7-7, Process Center Fire. HVB-4-10 remains available to annunciate a fire in the Relay Room, and Relay Room C02 fire suppression remains operable. Sprinkler systems remain operable in IWFS!TS Bldg, UPS room, and Processing Center.
The crew will diagnose the loss of Vital Bus "D" by the "D" Reactor Protection System Cabinet being deenergized, Nl-8 indication deenergized (CC), Multiple annunciator alarms, including one for a failed inverter, "D" powered HSPS lights lit, and a loss of the right monitor of the Position Monitor Panel. The ARO will place Radiation Monitor Interlock switches to Defeat, and restore Control Building, Auxiliary Building, and Fuel Handling Building Ventilation. The CRS will identify and declare the following Tech Spec: 3.5.5.2.
When the radiation monitor interlock switches are in defeat, ventilation is running and the Tech Spec has been declared, the scenario can continue.
NRC Scenario 2 Scenario Set-up Event #5: The Lead Examiner will cue Total FW Demand Fails to Zero Volts. This will cause an ICS transient, which if not responded to swiftly, will cause a Reactor Trip. The crew will diagnose the ICS failure by a rapid change in RCS pressure, multiple annunciator alarms, and/or changes in indications at multiple ICS stations. Entry into OP-TM-AOP-070, PRIMARY TO SECONDARY HEAT TRANSFER UPSET will be required based on RCS pressure not being controlled in ICS AUTO. "RCS pressure is not being controlled" requires the operator to make a subjective determination, based on their skills, training, and experience. A determination that RCS pressure is being controlled should include the following elements: 1) The reason for the transient is understood 2) RCS pressure response is consistent with the expected response for the event 3) Automatic or manual control in accordance with normal operating procedures is effectively controlling RCS pressure. A conservative assessment (i.e. concluding that RCS pressure is not being controlled) is appropriate when the three conditions above cannot be satisfied. ICS failures are one class of events that can lead to an upset in primary to secondary heat transfer. Most ICS failures can be mitigated by use of the appropriate manual control normal operating procedures.
This entry into OP-TM-AOP-070 is different from the others because FeedWater will rise significantly and the crew will need to establish ICS controls in manual at the Feed Pump and Feedwater Valve ICS controllers.
Once the plant is stabilized in ICS HAND control, the scenario can continue.
Event #6: The Lead Examiner will cue the "B" OTSG Tube Leak. Any OTSG tube leak causes an abnormal increase in the release of radioactive materials to the environment.
The most fundamental objective is to minimize this release. The prioritized objectives of this procedure are:
- Maintain core cooling.
- Minimize the activity release to the atmosphere (minimize release duration, rate and concentration of radioisotopes, particularly iodine)
- Minimize the integrated tube leakage 30 seconds of MSSV actuation can release 75% of the iodine for the entire event. The main condenser provides for iodine removal due to the ability of water to absorb the iodine (at least temporarily} through the condensers 104 "partitioning factor". This phenomenon reduces off site does consequences to the point that the radiation limits listed in the EOP can not be reached regardless of the tube leak size and the number of fuel pin leaks. MSSV actuation provides an opportunity for failure resulting in an unisolable path for reactor coolant directly to the environment.
The crew will diagnose an OTSG tube leak based on RM-G-27, RM-A-5, and RM-A-15 indications (PR), Annunciator C-1-1 in alarm, and/or pressurizer level lowering (CC).
The CRS will announce entry into OP-TM-EOP-005, OTSG TUBE LEAKAGE. This is a reactivity manipulation event. The URO will perform reactor shutdown with ICS in HAND. The ARO will place control Main Feedwater in HAND and may lineup to feed to the RCS from the "B" RBCT for inventory control. The CRS will evaluate and declare Tech Spec 3.1.6.3 When sufficient reactivity manipulation has been observed, the scenario can continue.
NRC Scenario 2 Scenario Set-up Event #7/8: The Lead Examiner will cue the "B" OTSG Tube Rupture. The CRS will continue in OP-TM-EOP-005, OTSG Tube Leak.
The URO will minimize SCM to lower pressure and therefore lower the OTSG tube leak rate. Guide 8 and OS-24 give direction to maintain SCM 30-?0F, but as close to 30F as possible. The ARO will preferentially steam the "B" OTSG:
Upon receiving the elevated offsite dose projections, the CRS should invoke 10CFR 50.54x and raise the cooldown rate lAW OP-TM-EOP-005, Step 3.31.
The URO will secure MU-P-1A due to an overcurrent condition. This will leave the "B" Train of HPI operational, which should be sufficient to maintain Pressurizer level once RCS pressure has lowered.
The scenario can be terminated the "B" OTSG has been isolated, HPI flow is terminated, minimum Makeup flow is established, and SCM has been minimized.
NRC Scenario 2 Scenario Set-up B&W Unit EOP Critical Task Description Document, 47-1229003-04:
CT Minimize SCM - HPI must be throttled to minimize SCM while maintaining margin> 30°F this minimizes primary to secondary leakage and reduces dose on the secondary side of the plant as well as minimizing release to the public. If HPI is allowed to raise OTSG pressure above 1000 psig after OTSG is full, a liquid RCS release to atmosphere would occur. Task failure would be to not throttle and challenge this.
- Minimizing SCM outside of the following limits should be considered grounds for failure of the critical task:
- Do not allow SCM to fall below 25F.
- Do not allow SCM to rise above ?OF sustained.
Safety Significance: Except when RCP NPSH limits are applicable and are more restrictive, RCS pressure should be maintained close to, but above, the minimum SCM to minimize RCS-SG L1P. The reason for minimizing RCS-SG L1P is to reduce the leak flowrate from primary to secondary to as low as possible.
Therefore, this procedure (minimizing SCM) is desirable whenever possible during SGTR mitigation.
Reducing the leak flowrate from the RCS to the secondary side of a SG reduces RCS losses and when accomplished with an impaired steam system (e.g.,
weeping MSSV and MSL leak) should reduce integrated radiation releases from the impaired system. If the level of the leaking SG can be maintained within normal operating limits, then the SG will remain available for continued use during the cooldown, thus enhancing the transient mitigation capability of the plant.
Cues:
- 1. SCM monitor
- 2. SPDS displays and associated alarms
- 3. P-T display and associated alarms Performance Indicators:
- 1. Operation of MU/HPI pump and valve controls
- 2. Operation of normal or auxiliary spray valve controls Feedback:
- 2. RCS pressure and temperature
- 3. MU/HPI pump and valve status indications
- 4. Normal and auxiliary spray valve status indications NRC Scenario 2 Scenario Set-up B&W Unit EOP Critical Task Description Document, 47-1229003-04:
CT Limit Uncontrolled Radiation Release- During SGTR mitigation:
- Attempt to control PZR level such that it does not continually decrease by increasing MU or HPI flow and reducing letdown flow
- If the reactor has not tripped, then perform controlled shutdown to prevent lifting MSSVs.
- Isolate non-essential steam loads from affected SG(s).
- If required to prevent exceeding [SG overfill setpoint] or [radiation limit], use emergency cooldown rate limit to 500'F.
The typical plant design allows for 40 cycles of an emergency cooldown to 500'F Thot at 240°F/hr. This rate is allowed for any SGTR event. However, it is recommended that the use of this emergency cooldown rate be limited to situations where:
a) the affected SG level(s) will reach the SG level limit before the SG can be isolated using the normal cooldown rate, including the use of SG drains if available, or b) activity release rates are projected to reach the integrated limit before 500°F Thot at the normal cooldown rate.
The emergency cooldown rate is recommended for the two cases noted because several large SGTRs and/or a relatively high percentage of failed fuel already exist.
In these cases, it is most important to prevent liquid discharge through the MSSVs and limit the duration of high activity release rates.
- Limiting Uncontrolled Radiation Release outside of the following limits should be considered grounds for failure of the critical task:
- Do not allow affected OTSG MSSV's to lift.
- Do not allow affected OTSG to go dry.
Safety Significance: If pressurizer level can be controlled, then the operator's ability to perform a controlled shutdown is greatly enhanced. Whenever possible power should be reduced as quickly as possible, but in a controlled manner, to well within the turbine bypass system capacity before tripping the reactor to prevent lifting of the MSSVs. This includes cases where maximum MU or HPI flow and letdown isolation are required to keep up with the tube leak and maintain pressurizer level. Power reduction is intended to minimize atmospheric radiation releases due to SG safety valve operation. Also, if a reactor trip can be averted through controlled operations, then ability to mitigate the transient is expected to be enhanced as normal transition from power operations to a controlled cooldown occurs.
Cues:
- 1. Main steam line radiation alarm
- 2. SG high level alarm
- 3. Pressurizer low level alarm
- 4. Verbal alert by plant staff that a SGTR is occurring and the reactor has not tripped NRC Scenario 2 Scenario Set-up B&W Unit EOP Critical Task Description Document, 47-1229003-04:
CT-21 -Limit Uncontrolled Radiation Release (cont)
Performance Indicators:
- 1. Operation of MU/HPI pump controls
- 2. Operation of MU/HPI valve controls
- 3. Operation of letdown valve controls
- 4. Operation of TBV/ADV controls Feedback:
- 1. Pressurizer level
- 2. Letdown flow
- 3. MUIHPI flow
- 4. RCS temperature
- 5. Verbal alert by plant staff of pressurizer level status
- 6. Verbal alert by plant staff of RCS cooldown rate NRC Scenario 2 Scenario Set-up B&W Unit EOP Critical Task Description Document, 47-1229003-04:
CT Reduce Steaming/Isolate Affected SGs (includes use of SG drains)-
Steam affected SGs to maintain level< [overfill setpoint]. If steaming alone cannot prevent SG fill, then use SG drains (if available) to maintain SG level below [overfill setpoint]. Isolate SG(s) if steaming and draining cannot prevent overfill and maintain RCS and isolated SG pressures < 1000 PSIG by use of [primary and secondary relief paths].
- Isolating Affected SGs outside of the following limits should be considered grounds for failure of the critical task:
- Do not allow isolation to occur with RCS pressure > 1000 psig.
Safety Significance: The more probable tube rupture scenario is a tube leak in one SG with both SGs available. The preferred mitigation strategy is therefore isolation of the affected SG following the initial cooldown and depressurization to
<1 000 PSI G. This limits the radiological consequences of the event, but does require cooldown to DHRS operation using one SG.
Both SGs are always used in the initial cooldown and depressurization to < 1000 PSIG. Prevention of MSSV lift on the affected SG(s) is integral to the goal of minimizing off-site release, and assurance requires RCS temperatures at or below 500°F in order to maintain SCM when RCS pressure is < 1000 PSI G.
Once this initial cooldown and RCS depressurization to <1 000 PSIG is completed, then SG isolation can be considered.
There are limitations on continued steaming of a SG with a SGTR. These limitations consider the overriding concerns of SGTR transients that dictate the isolation of the SG(s) and initiation of HPI cooling, if necessary. These limits are based on integrated radiation dose reaching predetermined values and SG filling due to tube leakage despite steaming to achieve maximum allowable cooldown rate.
SGs isolated due to SG fill criteria pose concerns related to liquid passing through MSSVs. MSSVs should be prevented from passing liquid, since their failure to reseat becomes more probable. For this reason, RCS and SG pressures are maintained <1 000 PSIG by use of [primary and secondary relief paths]. These relief paths may include such things as letdown, PZR vents, HPVs, the PORV, TBVs and ADVs.
Cues:
- 1. Rising OTSG level
- 2. Rad Monitor Alarms
- 3. Lowering Pressurizer level
- 4. Lowering RCS Pressure
- 5. Automatic initiation of HPI NRC Scenario 2 Scenario Set-up B&W Unit EOP Critical Task Description Document, 47-1229003-04:
CT Reduce Steaming/Isolate Affected SGs {includes use of SG drains) {cont)
Performance Indicators:
- 1. Operation of TBV/ADV controls Feedback:
- 1. SG(s) level and pressure
- 2. RCS pressure
- 3. MFW/EFW flow
- 4. MFW/EFW pump and valve status indication
- 5. TBVI ADV status indication NRC Scenario 2 Scenario Set-up Industry Experience:
- TMI Reactor Trip (11/2/06) Main Steam Safety Valves remained open longer than expected. (I R 552591)
- Indian Point 2 (2/15/00) - Steam Generator Tube Failure (380 litres per minute)
- Palo Verde 2 (3/14/93) - Steam Generator Tube Leak ranged between 11 and 39 litres per day, suddenly turned to 900 litres per minute tube rupture.
- Steam Generator Tube Rupture (Initiating Event)
Appendix D Scenario Outline Form ES-D-1 Facility: Three Mile Island Scenario No.: 4 Op Test No.: 289-2014-301 Examiners: Operators:
Initial Conditions: * (Temporary IC-238)
- 85% Power, MOL
- I&C Maintenance is occurring on HSPS, Train B
- SBO OOS for Maintenance, expected to return to service in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> .
Turnover: Maintain 85% Reactor Power
~
" rv v Critical Tasks:
- PORV Control for Heat Transfer (CT-13)
\Y
- Shutdown Reactor- ATWS (CT-24)
- Restore Feed to a Dry OTSG (CT-26)
Event No. Malt. No. Event Type* Event Description 1 lAOS TSCRS Instrument Air Leak Requiring Isolation of "A" Side 2-Hour Air and
~ CARO "A" EFW Valves, entry into OP-TM-AOP-028 (ARO:Start IA-P-1 A/B) 2 IC38B CCRS Invalid "B" OTSG Low Level, "B" EFW inadvertent actuation.
CARO (ARO: defeats invalid signal, secures EF-P-2B) 3 RD10B ICRS Uncontrolled Inward Rod Motion.
IURO IARO (URO: Assumes Manual Control of Control Rods) 4 TSCRS MU-V-18 Fails Closed, entry into Guide 9.
CURO (URO: Controls Pressurizer Level with HPI valve) 5 FW15A CCRS "A" Main Feed Pump Trips, Manual runback required.
R URO NARO (URO: Runback in Manual, ARO: Runback in Manual) 6 TH18B CCRS Sheared Shaft on RC-P-1 B CURO CARO (URO: Secures RC-P-1 B, ARO: re-ratios Main Feedwater) 7 RD28 MCRS "B" Main Feed Pump Runs to 0 rpm, ATWS, Lack of Primary to RD32 M URO Secondary Heat Transfer. / C c...~ C ~ *[
MARO tu~ S-r~ ~A-~5- 0 8 v CCRS EFW Control Valves fail to operate.
CARO (ARO: Establish PSHT via Condensate Booster Pump flow) 9 CCRS MU-P-1 A/C will not start, MU-P-1 B trips.
CURO (URO: Establish PORV control for Heat Transfer)
. (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor Scenario Set-up NRC Scenario 4 Three Mile Island NRC Scenario #4 Event #1: When the crew has accepted the watch, the Lead Examiner will cue the Loss of Instrument Air.
The operators will diagnose the Loss of Instrument Air based on a report from the field and slightly lower lA pressure.
Upon a Loss of Instrument Air, the following Critical Safety Functions are affected:
CSF-1, Reactivity & Reactor Power Control: Maintain control of the fission process, maintain the capability to shutdown the reactor and the capability to maintain the reactor in a shutdown condition. Control energy production and reactor power distribution based on design limits and current core heat removal capability. Loss of lA: If lA pressure < 60 psig, then the reactor will be tripped. At above 60 psig, lA will be sufficient for plant control and proper reactivity management. Reactor trip capability is not impacted by loss of lA.
Emergency boration can be performed via BWST or by local operation of MU-V-51 when emergency berating from BAMT. Emergency boration via the RBAT is inoperable (several air operated valves w/o handwheel). DC-V-20A
& B are closed to prevent flooding new fuel storage area with non-borated water.
CSF-3, RCS Integrity: Maintain the capability to control heatup and cooldown rates and control RCS pressure prevent reactor vessel brittle fracture or LTOP events. Maintain RCP seal cooling to prevent excessive loss of RCS inventory through RCP seals. Loss of lA: Normal RCS pressure control (spray &
heaters) is not affected by loss of lA. Loss of lA will result in a loss of letdown.
This will eventually cause problems with RCS pressure control (i.e., with seal injection and no letdown, pressurizer level will increase until solid ops is required). This is mitigated by providing a method to restore letdown without lA lAW OP-TM-211-950. Emergency pressure control functions (PORV & ES powered heaters) are not affected by loss of lA. Both methods (ICCW to thermal barrier & seal injection) of RCP seal cooling are adversely affected by loss of lA.
IC-V-3, IC-V-4, and MU-V-20 fail closed on loss of air (after depletion of local air reservoir). Local operator action to block OPEN IC-V-3, 4, and MU-V-20 is needed to maintain redundant means of seal cooling. Loss of ICCW more than 10 minutes after loss of seal injection, or Loss of Seal Injection for more than 10 minutes without ICCW will result in unacceptable RCP seal & thermal barrier temperatures which will preclude restoration of RCP seal cooling. (Reference 6.1) Therefore, the Loss of lA procedure is designed to ensure that both Sl and ICCW are not lost.
Scenario Set-up NRC Scenario 4 start if air to IBis lost. MS-V-13A & B fail open. The pump will remain operating until lA pressure is restored. EF-P-1 remains operable during a loss of lA event.
CO-V-13 is closed at less than 60 psig to prevent loss of EFW inventory caused by dumping the CSTs to the hotwell. CO-V-8 and hotwell makeup level control is supplied by BUIA.
CSF-5, Containment Integrity: Provide means to prevent or minimize fission product release to the environment. (1) Maintain containment pressure below design and (2) Provide capability to isolate the containment when required.
Loss of lA: Containment cooling. Industrial cooler operation will be degraded.
RBEC will be initiated lAW MAP N-1-6 if RB air temperature is high.
The CRS will enter OP-TM-AOP-028, LOSS OF INSTRUMENT AIR. The ARO will start IA-P-1A and/or IA-P-1 B from the Control Room prior to Instrument Air reaching 60psig (setpoint below which a manual Reactor Trip must occur). The CRS will determine isolation points and order the appropriate valves to be closed. The CRS will evaluate and declare T.S. 3.4.1.1.a.(2) based on the following Limit/Precaution in 1104-25, Instrument and Control Air System:
Both "A" and "B" trains of two-hour BUIA must be pressurized and inservice prior to exceeding 250°F in RCS. (Tech. Spec. Definition of OPERABLE (1.3) applied to EFW.)
When IA-P-1 A and/or IA-P-1 B are/is running, the air leak has been isolated and the Tech Spec has been declared, the scenario can continue.
Event #2: The Lead Examiner can cue the Inadvertent Low Level Signal on the "B" OTSG with EFW actuation.
The crew will diagnose the Inadvertent Low Level Signal by EF-P-2B and EF-P-1 running with no valid reason (OTSG level is greater than 10" in the Startup Range and OTSG pressure is greater than 600 psig).
The ARO will respond per OP-TM-424-901, and defeat the HSPS signal, secure the running EFW pump (EF-P-2B), and re-enable HSPS.
When HSPS is re-enabled, the scenario can continue.
Event #3: The Lead Examiner will cue the Uncontrolled Inward Rod Motion. This will cause an ICS transient.
The crew will diagnose the uncontrolled inward rod motion by an inward signal shown on the Diamond panel and the Position Indication Panel, Reactor Power lowering, RCS pressure and temperature lowering.
Note: OP- TM-AOP-064, UNCONTROLLED ROD MOTION, was deleted when the Digital Control Rod System was installed, and so the only way to address the situation is with OP- TM-AOP-070, PRIMARY TO SECONDARY HEAT TRANSFER UPSET.
Scenario Set-up NRC Scenario 4 Entry into OP-TM-AOP-070, PRIMARY TO SECONDARY HEAT TRANSFER UPSET will be required based on RCS pressure lowering.
"RCS pressure is not being controlled" requires the operator to make a subjective determination, based on their skills, training, and experience. A determination that RCS pressure is being controlled should include the following elements:
- 1) The reason for the transient is understood
- 2) RCS pressure response is consistent with the expected response for the event
- 3) Automatic or manual control in accordance with normal operating procedures is effectively controlling RCS pressure.
A conservative assessment (i.e. concluding that RCS pressure is not being controlled) is appropriate when the three conditions above cannot be satisfied.
ICS failures are one class of events that can lead to an upset in primary to secondary heat transfer. Most ICS failures can be mitigated by use of the appropriate manual control normal operating procedures.
This entry into AOP-070 is unique from the other scenarios because Reactor Power will lower due to a rod insertion signal, and not due to a failed ICS input.
Once the plant is stabilized in ICS HAND control, the scenario can continue.
Event #4: The Lead Examiner will cue the Closure of MU-V-18.
The crew will diagnose MU-V-18 closing by the green closed light lit and the red open light not lit (CC), lowering Pressurizer level indications (CC), and if left unattended long enough, MAP G-2-5, Pressurizer Level Hi/Lo in alarm.
The URO will establish Pressurizer Makeup manually via HPI Control Valve, MU-V-168 lAW OP-TM-EOP-010, Guide 9, RCS Inventory Control:
If normal makeup flow has not been established via MU-V-17 or MU-V-217 and through MU-V-18 (i.e. MU24-FI < 20 GPM), then MU-V-168 (or MU-V-16D) is used for normal MU flow. MU-V-16D is only used when the MU discharge cross connects are not in the normal lineup.
Once Pressurizer level is being restored and Makeup is being controlled manually, the scenario can continue.
Event #5: The Lead Examiner will cue the Loss of the "A" Main Feedwater Pump.
Scenario Set-up NRC Scenario 4 The crew will diagnose the Loss of FW-P-1A by an immediate drop in Feedwater flow, OTSG level decreasing rapidly, steam header pressure increasing, a neutron cross-limit alarm coming in, and the remaining feedwater pump speed increases causing feedwater flow to recover somewhat.
OP-TM-MAP-H01 01, Plant Run back, will be entered and a manual run back will be performed to lower Reactor Power to approximately 68%. this will require coordination between the URO and ARO, ensuring that the Control Rods and Feedwater are run back in symphony. This is the reactivity manipulation for the scenario.
Once sufficient reactivity manipulation has been observed, the scenario can continue.
Event #6: The Lead Examiner will cue the Sheared Shaft on RC-P-1 B.
The crew will diagnose the sheared shaft on RC-P-1 B by MAP alarm F-3-1 in alarm and zero amps indicated on RC-P-1 B.
OP-TM-MAP-F0301, RC LOOP FLOW LO, and OP-TM-226-152, Shutdown RC-P-1 B, will be entered to secure RC-P-1 Band tore-ratio Main Feedwater.
The URO will secure RC-P-1 Band the ARO will re-ratio Main Feedwater due to the "A" and "B" Feed flow master ICS control stations being in HAND.
Once sufficient reactivity manipulation has been observed, the scenario can continue Event #7/8/9: The Lead Examiner will cue the B" Main Fee mp Lowering to 0 rpm.
The crew will diagnose the loss of Main Feedwater by aJl\f'r mediate drop in Feedwater flow, OTSG level decreasing rapidly, and a rapid rise irir:+-RCS pressure and temperature. The URO will identify that an ATWS has occurred and will perform the Immediate Manual Actions of OP-TM-EOP-001, REACTOR TRIP.
Memorized operator action is appropriate because operator response time can significantly alter the consequences of an ATWS or a failure of the turbine to trip when required.
The reactor protection system is designed to prevent fuel clad or RCS pressure boundary failure. If RCS conditions are outside of the RPS envelope and RPS fails to de-energize the control rod drive mechanisms, prompt operator response can minimize the potential for fuel damage or an RCS pressure boundary failure EOP-001 is the primary entry point to the EOP network. This procedure is designed to address or direct procedures to address events from a reactor trip with no adverse plant conditions or equipment failures as well as events which challenge the fission product barriers.
Scenario Set-up NRC Scenario 4 The first priority (with the exception of entry into OP-TM-EOP-005 at power to mitigate an OTSG tube leak) to mitigate the consequences of a significant plant upset is to ensure the reactor is shutdown. Success of all subsequent EOP action is based on reducing core heat generation to reactor decay heat generation rates.
- Once the reactor is shutdown, prompt isolation of the turbine steam flow path is significant. Until the major steam flow path through the turbine to the condenser is isolated, RCS heat removal will be much larger than heat generation, and a rapid RCS cooldown will continue.
-When the turbine steam flow path is isolated, plant conditions are evaluated to determine if any symptoms of a core cooling upset are present.
Based on a symptom of low subcooling margin, excessive heat transfer, lack of primary to secondary heat transfer or primary to secondary leakage, rule based actions and entry into other sections of the EOP network is performed.
The crew will identify a Lack of Heat Transfer based on the following definition from OS-24, Conduct of Operations During Abnormal and Emergency Events:
LOHT is the inability of either OTSG to remove sensible heat from the RCS.
LOHT can be confirmed if one of the following sets of conditions exists:
- lncore temperatures or Thot rising above 580°F and at least one RC Pump operating
- lncore temperatures rising and NO FEEDWATER available
- lncore temperatures rising and RCS circulation can not be confirmed The CRS will direct entry into OP-TM-EOP-004, LACK OF PRIMARY TO SECONDARY HEAT TRANSFER.
RCP philosophy lAW the GEOG and the OP-TM-EOP-004 Basis Document:
One or two RCPs (one in each loop) should be left running to reduce heat input to the RCS yet provide for heat transfer as soon as FW is restored to either OTSG.
The step intent is to reduce heat input to the RCS while maintaining forced flow in both RC loops.
Impact to pressurizer spray flow should be considered when selecting RCPs for shutdown.
Although one RCP is allowed, our procedure directs one RCP in each loop to allow for even flow between the loops (avoiding the reverse directional flow in the opposite loop),
and to prevent a loss of forced RCS flow if the only running RCP were to trip.
Scenario Set-up NRC Scenario 4 The ARO will recognize that EFW control valves are inoperable and will establish primary to secondary heat transfer using Condensate Booster Pumps lAW OP-TM-EOP-004, LACK OF PRIMARY TO SECONDARY HEAT TRANSFER:
If feedwater is NOT available, then efforts to establish EFW should continue. If this event occurs when the condensate booster pumps could provide a continuous feedwater supply, then the booster pumps may be used alone to feed the OTSGs.
HPI COOLING will be initiated (EOP-009) when RCS pressure approaches the PORV setpoint. After initiating HPI COOLING, actions to restore feedwater (main or emergency) should continue.
Additionally, when the conditions are met during the scenario the CRS will direct entry into OP-TM-EOP-009, HPI COOLING.
HPI will not be adequate, which will force the URO to manually control the PORV to maintain RCS pressure while minimizing inventory losses until Primary to Secondary Heat Transfer exists.
Termination: The scenario can be terminated when the Reactor is shutdown, and Primary to Secondary Heat Transfer has been established via Condensate Booster Pump flow.
Scenario Set-up NRC Scenario 4 B&W Unit EOP Critical Task Description Document, 47-1229003-04:
CT PORV Control for Heat Transfer- During action to restore primary to secondary heat transfer, RCS pressure must be continually controlled. RCS pressure is controlled by manually opening and closing the PORV. This prevents excessive PORV cycling which could occur if the PORV were allowed to operate automatically.
During mitigation of LHT, manually cycle the PORV as necessary to maintain RCS pressure between the PORV setpoint or RV P-T limit and minimum SCM (if subcooled) or 1600 PSIG (if saturated).
- Performing PORV Control for Heat Transfer outside of the following limit should be considered grounds for failure of the critical task:
- PORV automatically lifts three (3) times.
Safety Significance: During mitigation of LHT the PORV should be manually cycled to control RCS pressure between the PORV setpoint or RV P-T limit and 1600 PSIG if the RC is saturated or minimum allowable SCM if the RC is subcooled. The PORV opening values prevent challenges to the pressurizer safety valves and the RV P-T limit while the PORV closing values maintain a positive primary to secondary side AT (SGs remain heat sinks).
Cues:
- 1. High RCS pressure alarm
- 2. SCM monitor and associated alarms
- 3. SPDS displays and associated alarms
- 4. P-T display and associated alarms
- 5. Verbal alert by plant staff that RCS pressure has reached the PORV setpoint Performance Indicators:
- 2. RCS pressure and temperature
- 3. Verbal indication by plant staff that PORV flow has been initiated Scenario Set-up NRC Scenario 4 B&W Unit EOP Critical Task Description Document, 47-1229003-04:
CT Shutdown Reactor- ATWS- Actuation of the manual reactor trip pushbutton, to backup the automatic trip and/or provide the necessary reactor trip, anytime the reactor trips or should have tripped. In the event the reactor fails to trip, in response to automatic and manual demands, then perform the following: Deenergize CRDMs
- Shutting down the reactor due to an ATWS outside of the following limit should be considered grounds for failure of the critical task:
- Not deenergizing the CRDM power supplies:
- 1L-02 Safety Significance: Without taking the proper actions, there exists a potential challenge to the Reactor Coolant System pressure boundary due to high RCS pressure.
An ATWS could occur due to a failure of the RPS to initiate a reactor trip signal upon one of the reactor trip parameters reaching its trip limit or the control and safety rods failing to insert once the RPS trip signal is given automatically or manually. A Diverse Scram System (DSS) is provided, independent of the RPS, to minimize the potential for an ATWS event. However, the operator must recognize and react to any of the reactor trip parameters that exceeds its limit but does not cause a reactor trip.
In this situation, the manual reactor trip button has been actuated but reactor power is not less than the plant specific reactor power level for verification of a reactor trip. Therefore, the reactor has not been shut down and there has been a failure of all or most of the control and safety rods to insert into the reactor core.
Given that RPS, DSS and the manual reactor trip have failed to trip the reactor, then immediate actions to shut down the reactor by the alternate methods should be initiated. These methods include trip of CRDM breakers and maximum rate of boron addition to the RCS. Once the control and safety rods are successfully tripped into the core, or sufficient boric acid has been added to provide an adequate shutdown margin, the reactor will be shut down.
This should be achieved prior to taking additional mitigating actions because post-trip transient mitigation, from this point forward, is based on the assumption that the reactor is shutdown (subcritical).
Scenario Set-up NRC Scenario 4 B&W Unit EOP Critical Task Description Document, 47-1229003-04:
CT Shutdown Reactor- ATWS- Continued Cues:
- 1. RPS channel alarms
- 2. RCS Power, Pressure and Temperature indications
- 3. P-T display and associated alarms
- 4. Verbal alert by plant staff that reactor shutdown requirements have not been met Performance Indicators:
- 1. Operation of control rod drive feeder breakers Feedback:
- 1. Nuclear Instruments
- 2. Control rod status indication
- 3. Control rod drive breaker status indication
- 4. Verbal indication from plant staff of reactor shutdown status Scenario Set-up NRC Scenario 4 B&W Unit EOP Critical Task Description Document, 47-1229003-04:
CT Restore Feed to a Dry OTSG- If a RCP is running, establish FW to the SG(s) and control FW flow to maintain RCS cooldown rate within limits. EFW flow is established at less than 450 GPM total flow and MFW flow is established at less than 200,000 LBM/HR total flow.
- Restoring Feed to a Dry OTSG (sustained) outside of the following limits should be considered grounds for failure of the critical task:
- To minimize OTSG stress, do not exceed MFW flow greater than than 200,000 LBM/HR total flow (sustained).
- To ensure the main feedwater nozzles remain full and to prevent cavitation type damage to the Main Feedwater nozzles if they are not full of subcooled fluid, do not fall below less than 160,000 LBM/HR total flow (sustained).
Safety Significance:
If it is decided to perform the cooldown by using trickle feeding, it will be necessary to control the rate of FW addition to the SGs to maintain RCS cooldown limits. The FW flow rate should be adjusted to get the desired cooldown rate. If possible EFW should be used to limit SG thermal stresses. If MFW is used with the MFW nozzles, it will only be effective with forced flow.
Once heat transfer is restored in the SG, feed rates can be adjusted as necessary to control the cooldown and SG tube-to-shell ~T.
Cues:
- 1. Low SG level alarms
- 2. Low SG pressure alarms
- 3. Verbal alert by plant staff that no SG is available for heat transfer Performance Indicators:
- 1. Operation of EFW /MFW pump controls
- 2. Operation of EFW /MFW valve controls Feedback:
- 1. EFW/MFW flow
- 2. SG level and pressure
- 3. RCS pressure and temperature
- 4. Verbal alert by plant staff of EFW /MFW flow status Scenario Set-up NRC Scenario 4 Industry Experience:
- Harris Nuclear Plant Manual Scram Due to Loss of Feedwater (12/14/99)
- Oconee 1 Loss of Feedwater (5/26/00)