Text

Draft - Outlines
	ML040960319
Person / Time
Site:	Calvert Cliffs
Issue date:	03/05/2004
From:	Sickle J; Constellation Nuclear
To:	Conte R; NRC/RGN-I/DRS/OSB
	Conte R
References
	50-317/04-301, 50-318/04-301
	Download: ML040960319 (52)
	v • d • e

CCNPP 2/04 NRC EXAMDEVELOPMENT Operations Training Unit TO: J. DAntonio, Chief Examiner, NRC Region 1 FROM: D. Lavato, Facility Representative SUBJ: 2/04 Initial License Operator Exam DATE: January 16,2004 I have e n c a e d 19 JPMs, 2 (1 RO, 3 SRO) written examinations with associa.sd documentation, copies of required procedures for pedorming the JPMs and 4 scenarios with the required EOPdAOPs for the 2/04 Initial License Operator Examinations at Calvert Cliffs Nuclear Power Plant per NUREG 1021 DraR Revision 9, ES 20 1 for your review. Copies of modified questions with their originals are also enclosed.

I have included Forms ES-301-3,4, 5 , 6 and ES-401-6 as required.

A copy of the audit exam which you requested is included.

In accordance with Attachment 1 of ES 201, the enclosed materials SHALL BE WITHHELD FROM PUBLIC DISCLOSURE UNTIL EXAMINATIONS ARE COMPLETE.

If you need any additional information, please call me at (410) 495-4586 or Mike Wasem at (410) 495-3638.

Sincerely, D. Lavato cc: 2004 NRC exam file

3/04 N Rc EXAM DEVELOPMENT Operations Training Unit Two copies of each written exam are included. One copy is presented how the examinees will see it, including the handouts (references) they will be allowed to use. The other copy contains information that will help in your review process. This copy also contains the original question that was modified where applicable. I have included a marked-up question to explain the format of the information provided. I have also included marked up copies of the written outlines.

4) E f G tfi J Given the following plant conditions:

-- Unit One has tripped due to a loss P-13000-1

-- 1I4KV bus is energized from 1A Diesel Generator

--,' Pzr level is 100' and'slowly lowering

-- RCS pressure is 1920 PSIA and slowly lowering The RO reports that only 12 Charging Pump is running and Pressure and Inventory is Deing monitoredfor positive trends.

What alternate actions must the CRS direct or verify?

A. Verify SIAS actuation when RCS pressure reaches 1725 PSIA.

B. Manually start 1Iand 13 charging pumps to restore pressurizer level to greater than 1Ol"and locally reset 1Ipressurizer backup heater breaker.

C. Isolate letdown, check that charging pumps automatically start to restore pressurizer level and reset pressurizer proportional heaters by momentarily placing the handswitches to OFF.

D. Verify charging pumps start automatically to restore Pressurizer level to greater than I O I " , verify 12 and 14 pressurizer backup heaters start to restore RCS pressure.

A is incorrect, action should be taken so that SIAS does not actuate.

B is correct, charging pumps must be manually started, and heaters must be reset, and they will not operate below 101" in the pressurizer.

C is incorrect, letdown will automatically isolate, 11 and 13 charging pumps will not automatically start, and the proportional heaters will take a long time to restore RCS pressure.

D is incorrect,l 1 and 13 charging pumps will not automatically start with the normal and alternate 4 KV bus feeder bkrs open and 12 and 14 heaters will not have power available.

References:

EOP basis docs. 41.5, 43.5

CCNPP 2/04 NRC EXAMDEVELOPMENT Operations Training Unit TO: J. DAntonio, Chief Examiner, NRC Region 1 FROM: Mike Wasem, exam author SUBJ: 2/04 Initial License Operator Exam DATE: Feruary 11,2004 I have enclosed copies of the examination materials which include all the corrections and comments gathered during validation the week of February 2. Also, a proposed schedule and a copy of the security agreement with the most up to date signatures, as of today, are included.

In accordance with Attachment 1 of ES 201, the enclosed materials S U L BE WITHHELD FROM PUBLIC DISCLOSURE UNTIL EXAMINATIONS ARE COMPLETE.

If you need any additional information, please call me at (410) 495-3638.

Sincerely, Nb-Mike Wasem cc: 2004 NRC exam file

Here is a quick summary of the changes from the originally submitted materials:

SRO Administrative Topics Outline-Changed A.2 descriptionof activity and WA to better reflect intent of the JPM Control Roodn-Plant Systems Outlines---Changed SystedJPM Title to reflect the change out of one in-plant JPM and the modification of one Control Room JPM SRO written:

Incorporated comments on questions 1,2, 3,4, 6, 7, 9, 1 1, 16, 19 and 21.

Included Electrical System LCOs and basis to handout associated w/l5.

Replaced questions 10 and 25.

RO written:

Incorporated comments on questions 4, 12, 24 29, 32,42,45 and 73.

Replaced questions 18, 50, 55 and 74.

Changes to the written exams resulted in changes to items 6 and 7 of the Written Examination Quality Checklist, Form ES-401-6 to:

Bank Modified New 23/10 811 44/14 Memory CIA 34/11 41/14 Scenarios:

Write ups of critical tasks.

l--correct load reduction target, referenced TS for PORV failure, Actions for possible 2 CPUs being failed and SRO referencing table for OTCC success.

Z-added referencing TS 3.6.6 for loss of CCW

3-reflect SGFP initial lineup change, addded details for expected actions shifting GS and SGFPs to Main Steam

CCNPP 3/04 NRC EXAMDEVELOPMENT u Operations Training Unit TO: J. DAntonio, Chief Examiner, NRC Region 1 FROM: D.F. Lavato, Facility Representative SUBJ: 3/04Initial License Operator Exam Outlines DATE: December 9,2003 I have attached the required outlines for the March 2004 Initial License Operator Examinations at Calved Cliffs Nuclear Power Plant per NUREG 1021 Revision 9 ES 201.

Additional information is attached which includes the Random Selection method used for the Written exam development and a proposed exam week schedule.

In accordance with Attachment 1 ofES 201, the enclosed materials SHALL BE WITHHELD FROM PUBLIC DISCLOSURE UNTIL EXAMINATIONS ARE COMPLETE.

I have included form ES 201-2 with the test outlines. If you need any additional L . information, please call me at (410)495-4586or Mike Wasem at (410) 495-3638.

Sincerely, D.F. Lavato cc: 2004 NRC exam file

Scenario 1 CRS u1 U2 u3 I1 RO I2 I3 I4 RO1 CRO R02 R03 R04

I1 = Instant SRO 1 U1 = Upgrade SRO 1 RO1= Reactor Operator 1 Note: Could all be the same scenarios on DAY 1, since no one participates in more than one (1) scenario on each day and everyone participates in scenarios.

Created on 12/09/03 12:18 PM

Scenario 2 CRS I2 I3 I4 RO R02 R03 R04 CRO RO1 I1

0 = Surrogate (can be same for both scenarios)

Created on 12/09/03 12:I8 PM

u3 A1 .a U.b U.a A.1.b u.c I I1 1.c A. 1.b 1.a A.1.a 1.d I2 1.c A. 1.b 1.d 1.a A. 1.a I3 A.1.b 1.c A. 1.a 1.d 1.a I

I4 1.a A.1.a 1.d 1.c A. 1.b j

RO1 R.d R..a R.h R.c A.1.b 1I

-1 R02 R.h R.d R.a A.1.b R.c 1

R03 R.c A.1.b R.a R.d R.h R04 A. 1.b R.c R.h R.a R.d NRC 1 u1 I1 I4 R03 NRC 2 U2 I2 RO1 R04 U= upgrade I= instant SRO A=Admin

-1 JPM Created on 12/09/0311:43 AM

I 'V P'V I 'V 1 zn I h

In plant R04 NOTE: The Exit will be as determined by the Lead Examiner.

Created on 12/09/0311:43 AM

CCNPP NRC License Examination March 2004 P W R Examination outline Emergency and Abnormal Plant evolutions - Tier 1 Group1 (RO/SRO) Form ES-401-2 E/APE #/Name/Safety Function I 41 KIA Topics I Imp. I SRO#

000007/E02 Reactor Trip - Recovery I1 000008 Pzr Vapor Space Accident 13 000009 Small Break LOCA 13 000011 Large Break LOCA 13 000015/17 RCP Malfunctions 14 I l l 000022 Loss of Rx Coolant Makeup 12 sol

~~ ~ ~ ~

000025 Loss of RHR System 14 x 2.1.12 Ability to apply Technical Specifications for a system L C ~ Seb e/-./R 4.0 1 000026 Loss of Comp. Cooling Water I8 I / - - , A i e-000027 Pzr Press. Ctrl. Sys. Malf. I3 I ,

x 2.4.45 1 Ability to prioritize and interpret the s i g n i f i c f n $ % o t e o r & ad?; u , I 3.6 1 1

000029 ATWS I1 I x I I EA2.01 I Ability t o determine or interpret reactor nuclear instrumentation as .mdeJt?a A W d I 4.7 I 1 000038 SG Tube Rupture13 000040CEO5 Steam Line Rupture 14 000054CE06 Loss of Feedwater 14 000055 Station Blackout 16 I l l 000056 /Loss of Off Site Power I6 I l l 000057 Loss of Vital AC Instrument Bus 16 000058 Loss of DC Power I6 000065 Loss of Instrument Air 18

CCNPP NRC License Examination March 2004 PWR Es;lmination outline Emergency and Abnorinal Plant evolutions -Tier 1 Group 2 (RO/SRo) Form ES-401-2 E/APE #/Name/Safety Function K K K A A G Number KIA Topics Imp. SRO 1 2 3 1 2 #

oooOO1 Continuous Rod Withdrawal /I oooO03 Dropped Control Rod / I I -T 1 I I oooO05 InoperablelStuck Control Rod / I I I I --.- -~ .

u.-107-

-5.

oooO24 Emergency Boration /I x 2.2.22 Kfmwledge of limitingyonditions for operations and safety limits /--3-:-84.1 0 a?(1 oooO28 /Pzr Level Malfunction I2 oooO32 Loss of Source Range NI ff I l l oooO33 Loss of Intermediate Range NI 17 1 oooO36 /Fuel Handling Accident /8 oooO37 SG Tube Leak /3 I l l ations for system op oooO51 Loss of Condenser vacuum /4 I l l oooO59 Accidental Liquid RadWaste Rei. /9 oooO60 Accidental Gaseous Radwaste Re1.19 I I oooO61 ARM System Alarms f f oooO67 Plant Fire on Site /8 oooO68 Control Room Evac. /8

~~

oooO69 Loss of CTMT Integrity /5 I l l oooO74 Inadequate Core Cooling 14 000076 Hiah Reactor Coolant Activity 19 I I I CEIAI 1 RCS OvercoolinglPTS 14 I l l CE/A13 Natural Circulation /4 CE/A16 /Excessive RCS Leakage 12

~ ~~~~~

CUE09 /Functional Recovery I l l KIA Category Totals I Group Point Total

CCNPP NRC License Examination March 2004 PWR Examination outline Plant Systems - Tier 2 Group 1 (RO/SRO) Form ES-401-2 System #/Name l ~ l ~ l ~ l ~ l ~ l ~ W A Topics I Imp. I SRO 003 Reactor Coolant Pump 004 Chemical and Volume Control 005 Residual Heat Removal 007 Pressurizer Relief/Quench Tank 008 Component Cooling Water I I I I I I X

A2'07 Consequences of high or low flow rate and temperature: the flow rate at which the CCW standby pump will start 0 h . 4 010 Pressurizer Pressure Control 012 Reactor Protection 013 ESFAS I 022 Containment Cooling I I I I I I X '2.04 Ability to predict imp ts of and correct or control-Loss of servicewater c 0x4\ v \ - ~ G JrzG fin\ 3.2 I 039 Main and Reheat Steam 059 Main Feedwater 061 Auxiliary Feedwater I I I I I I 076 Service Water T 078 Instrument Air 103 Containment 1 K/A Category Point totals 121 I I 21 I Group Point Total

CCNPP NRC License Examination March 2004 PWR E'uamination outline Plant Systems - Tier 2 Group 2 (RO/SRO) Form ES-401-2 System #/Name K K K K K E A A A A G Number KIA Topics Imp. SRO 1 2 3 4 5 6 1 2 3 4 #

001 Control Rod Drive 002 Reactor Coolant I I I I I I 011 Pressurizer Level Control IIIIII X A2.05 Ability to predict the impads of and correct or control-Loss of pressurizer heaters O E L Ld Lfmr 5-Rx3 3.7 1 014 Rod Position Indication 1 1 1 1 1 1 015 Nuclear Instrumentation I I I I I I 0 1 6 Non-Nuclear Instrumentation I I I I I I 027 Containment Iodine Removal Control 029 Containment Purge 033 Spent Fuel Pool Cooling x

034 Fuel Handling Equipment 2.4.49 immediate operation w- I I 7- 6- c]-

Ability to perform wlo reference to procedures actions that require 001 4.0 1 035 Steam Generator I I I I I I 041 Steam DumplTurbine Bypass Control

~~~

068 Liquid Radwaste I I I I I I 072 Area Radiation Monitoring 079 Station Air TIIIII

~

0 8 6 Fire Protection WA Category Point totals 1 1 Group Point Totat 2

CCNPP NRC License Examination March 2004 P W R Examination outline Generic Knowledge and Abilities Outline (Tier 3)

> mES-401-3

~

1 Category WA # Topic SRO-Only Imp. #

2.1.2 of plant 3.8 1 2.1.15 Ability to manage short-term information such as night and 3.0 1 standing orders -

&rGkr$Nojd 6 m SUI Subtotal 2.2.22 Subtotal 2.3.3 2.3.7 Subtotal 2.4.4 Ability to recognize abnormal indications for system operating parameterswhich are entw level conditions for ememencv and I 4.3 1 1 2.4.18 1 /

Tier 3 Point Total Subtotal

CCNPP NRC License Examination

,L~uAI 2~0-4 PWR Esallhution O L ~;dit-

E c

-MM. line A

h r u p Restoration of power with one EDlG c j ~

OccurrenieTFa t k i n e trip t

LOSS 6 fi u

Effects of f e s w a t e r Introduction on dry S1Gj-f

~

r e u+ l ~ r~L pndl) 1 11, r)( &d!

P I

@, i

~

dS I G T ~

jL 0-3

&.,I 1015-.,

-i;,

~ J G- I ;

40 3.6 3.9 ~I I I~ ~
f 000057 Loss of Vital AC Instrument Bus I6 AA2.17 System and compone@atus, using local or remote controls .
LiXs 66 v m i / S L 031 3.1 I
\LI000058 Loss of DC Power 16 I I component status 3.0 1 000062 Loss of Nuclear Service Water 14 S isolation valves to 3.2 000065 Loss of Instrument Air 18 3.5 WA Category Totals 21 I I 1 1 8 1 Page 2
CCNPP NRC License Examination March 2004 PWR Examination outline Emergency and Abnormal Plant evolutions - Tier 1 Group 2 (ROISRO) Form ES-401-2 E/APE #/Name/Safety Function IK IK IK A A G Number KIA Topics Imp. RO 1 2 #
000001 Continuous Rod Withdrawal 11 6 lyl\
000003 DroDDed Control Rod 11 X M.01 Ability to determine and interpret-Rod p o s M w k t u a l rod dosition 3.7 1 ITp- I 1 I
~~
000005 InoperablelStuckControl Rod ~
000024 Emergency Boration 11 000028 1Pzr Level Malfunction /2 000032 Loss of Source Range NI i7 x 2.2.22 "Knowledge of limitivcondltlons for operation and safety limits kw ~ p & 01 ;3.4 1 000033 Loss of Intermediate Range NI i7 L. I, -1 ,- . I 000036 /Fuel Handling Accident I8 X M2.M Ability to determine and i " t e r ~ e t - u r r ~ ~ ~ ~3.4 ~ i1~ ~ ~ ~ ' ~
000037 SG Tube Leak /3 000051 Loss of Condenser vacuum 14 000059 Accidental Liquid RadWaste Rei. I9 000060 Accidental Gaseous Radwaste Re1.19 000061 ARM Svstem Alarms ff x 2.1.1 Knowledge of condqbt of operations requhements flp& (28 I 3.7 1 000067 Plant Fire on Site 18 X ~ ~ 3 . 0 4 &#&ded in'EOP for plant fireon site .~ L ) A . - 4q c;q~ 3.3 I 000068 Control Room Evac. 18 000069 Loss of CTMT Integrity 15 000074 lnadeauate Core Cooling 14 X EA1.16 Abllityto operate and mmftor RCS incore thermocouple indicato~Q~-)/q- /-.el. &.$ (f&
000076 High Reactor Coolant Activity 19 I Ix I Aw.O1 J Knowledge of interrelations between the Hi h Reactor Coolant Activ3 and the process radiation monitors -
7 lois I - ]-ry- o<q- 2.6 1 r - . I ,.".--
CE/AI IRCS OvercoolinglPTS14 CE/AI 3 Natural Circulation 14 C V A I 6 /Excessive RCS Leakage I2 3.4 1 CE/EO9 /Functional Recovery EK3.4 ' such KOowfedgafoFth~~~asbns'olRO'or SRO function within the control room team in a way that procedures are adhered to and limlts are not violated l n[,- L?, , K..y.3
- 159 I
212 2 Group Point Total 9 Codidential Page 3
CCNPP NRC License Examination Mdrch ~ W I PLLREAAI~\IILI~IW J otriiil, .
Plant Systems - Tier 2 Group 1 (RO/SRO) Form ES-40 1-2 System #/Name K K K 1 2 3 003 Reactor Coolant Pump Q i+:(
3 004 Chemical and Volume Control 53- ,
.*. * \
f,
? 005 Residual Heat Removal : x 008 Component Cooling Water > ~
7 1, 3 .
I 007 Pressurizer RelieflQuench Tank
~ 0 1 2 Reactor Protection 0 2 6 Containment Spray
-.\ \ I 026 Containment Spray I -1 I (4 :)
7 039 Main and Reheat Steam
--$ ?
$\ 1 039 Main and Reheat Steam ?
I 056 Condensate 1- I 059 Main Feedwater 059 Main Feedwater 061 Auxiliary Feedwater 061 Auxiliarv Feedwater I 0 6 2 AC Electrical Distribution 063 DC Electrical Distribution 2 Confidential Page 4
CCNPP NRC License Examination March 2004 PWR Examination outline Plant Systems - Tier 2 Group 2 (ROISRO) Form ES-401-2 System #&Name KIA Topics Imp. RO 001 Control Rod Drive A4.08 Ability to manually operate the mode select for CRDS, operation of 3.7 1 MG sets and control panel CR. o-. (a&.)- oq 002 Reactor Coolant Ability to predictlmonitor reactor power associated with RCS controls -e 01 1 Pressurizer Level Control 014 Rod Position Indication K5.01 Knowledge of operational implications of differences betxeen RPlS 2.7 1 and step counter ('-bD ' &)- 4" 01-1 4 01 5 Nuclear Instrumentation Ability to monitor recognition of audio output 5xpected for a given plant 2.6 condition N I; ILxtm-mM9(5 R >I 01 6 Non-Nuclear Instrumentation 0 1 7 In-Core Temperature Monitor Ability to manually operatelmonitor H2 sampling and analysis of (3.1 1 cntmt atmosphere including alarms and indications f./.,., h/ \/It*-/
I I Knowledge of the design featurefinterlock providing auto isoiaion b4,j rflQ$h.,&fY toAL<i& @I 7
temperature [fi) __
Knowledge that t e effed of I of cooling has on spent fuel
-'L/-3-D!=. oZ3/
3.0 1 034 Fuel Handling Equipment 035 Steam Generator 041 Steam Dumpflurbine Bypass Control Knowledge of operational implications of biolggical hazards of radiation and resulting goal of ALARA I-!, Q L I fi &@a 1 e 00 1, 3.2 1 Knowled e of connecfions and cause/effect betwezn ARM and Control Ffoom ventilation c.0 /J 4- 1. 5 -9g qfy 3.3 Ability to manuall operatelmonitor starting Pumps &&.G 5l/fir&Pl l.7 echanisms of fire m7/
2.9 1 WA Category Point totals 111 01 11 11 21 01 11 01 21 21 01 Confidential Page 6
FormES-401-3 Category WA # Topic RO Imp. I #
2.1.3 Knowledge of shift turnover practices Sflr TU,@Q~'&. 1 i', 3 . a I 1 2.1.10 Knowledge of conditions and limitations in the facility license c &27~&#~1,]j Conduct of OperatiOllS 2.1.14 Knowledge of system status criteria which require the notificationof danimersonnel +i,<Thn <rfirJ<(L?tr/n I 2iiI I 1 I
Subtotal 2.2.28 2.2.33 Equipment
2. r\ 2.2.34 process for determining the internal and core reactivity Qkc L, f&Lj-lb$Ty dc/2.8 1 control I Subtotal
3. 7s 2.3.10 Control 3 2.3.11 Subtotal 2.4.17
4. 2.4.46 Emergency Procedures Subtotal ICI 3PointTotaI Confidential Page 7
2/04 NRC Exam Outline Random Selection Method Overview:
The Written Exam Outlines were developed per NUREG 1021 Draft Revision 9, ES-401 D. 1 and Attachment 1.
Numbered tokens were used as described in the attachment 1 to develop both the RO and SRO Written Test Outlines.
Specific steps:
The RO and SRO Written Exam outlines were made site-specific before any random selection was performed. For Tier 1 categories, the Westinghouse and B&W specific E/APEs were removed from the outlines. For Tier 2, Ice Condenser was removed from the outline.
For the SRO exam, only Categories A2,AA2, EA2 and G were maintained in the pool of available WA statements for selection in Tiers 1 and 2.
Rejected WA statements were replaced with randomly selected WA statements in the basic method specified in Attachment 1.
u
CCNPP NRC License Examination March 2004 PWR Written examination outline FormES-401-2 Tier I Group I RO W A Category Points SRO-Only Points I - I 2 I 18 1 2 2 2 4 2 2 4 5 1 2 2 28 2.
Plant Systems 2 1 0 1 1 2 0 1 0 2 2 0 10 Tier 3 2 3 5 4 2 5 5 3 4 2 38 Totals
3. Generic Knowledge and Abilities Categories Note: 1. Ensure that at least two topics from every WA category are sampled within each tier of the RO outline (i.e. the Tier Totals in each WA category shall not be less than two). Refer to Section D. 1.c for additional guidance regarding SRO sampling.
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 f l from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.
3. Select topics from many systems and evolutions; avoid selecting more thantwo WA topics from a given system or evolution unless they are related to plant specific priorities.
4. Systemdevolutions within each group are idenMied on the associated outline.
5. The shaded areas are not applicable to the categoryltier.
6.
The generic (G) WASin Tiers 1and 2 shall be selected from Section 2 of the WA Catalog, but the topics must be relevant to the applicable evolution or system. The SRO U A s must also be linked to 10 CFR 55.43 or an SRO-level learning objective.
7. On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (R) 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; summarize all the SRO-only knowledge and non-A2 ability categories in the columns labeled K and A. Use duplicate pages for RO and SRO-only exams.
8. For Tier 3, enter the WA numbers, descriptions, importance ratings, and point totals on Form ES-401-3.
9. Refer to ES-401, Attachment 2, for guidance regarding the elimination of inappropriate WA statements.
u Confidential Page 1
CCNPP NRC License Examination March 2004 PWR Examination outline Emergency and Abnormal Plant evolutions - Tier 1 Group 1 e) Form ES-401-2 E/APE #/Name/Safety Function WA Topics I Imp. I RO 1 2 #
000007/E02 Reactor Trip - Recovery 11 X EA1.06 Verification that the control and safety rods are in after the trip 4.4 I 000008 Pzr Vapor Space Accident 13 I I I x AA2.01 RCS pressure and temperature indicators and alarms 3.9 1 000009 Small Break LOCA /3 I l l x I I 1 EA1.07 I Ability to operate and monitor the Containment Cooling Svstem I 3.7 I 1 00001 1 Large Break LOCA 13 X EK2.02 Knowledge of the interrelations between Pumps and the Large Break LOCA 2.6 1 Basic steady state thermodynamic relationship between RCS loops and SlGs resulting from 000015/17 RCP Malfunctions 14 X AK1.04 unbalanced RCS flow 2.9 1 000022 Loss of Rx Coolant Makeup /2 x 2.4.21 Knowledge of parameters and logic used to assess the status of safety functions 3.7 1 000025 Loss of RHR Svstem 14 I l l I NOT SELECTED I I 000026 ~ o s ofComp.
s Cooling Water 18 I I 1 x AK3.02 Automatic actions within the CCWS resulting from ESFAS actuation 3.6 1 000027 Pzr Press. Ctrl. Sys. Malf. I3 I I Ix AK3.03 Actions contained in EOP for PZR PCS malfunction 3.7 I I I I EK1.03 I Effects of boron on reactivity, as it relates to an ATWS I I
~~
000029 ATWS I1 1x1 1 3.6 I 000038 SG Tube Rupture13 X EK3.01 Equalizing pressure on primary and secondary sides of ruptured SIG 4.1 1 000040CE05 Steam Line Rupture 14 X AK1.07 Effects of feedwater introduction on dry S/G 3.4 I 000054CE06 Loss of Feedwater 14 IxI I AK1.01 MFW line break depressurizes the SIG (similar to steam line break) 4.1 I 000055 Station Blackout 16 X EA1.06 Restoration of power with one ED/G 4.1 1 000056 /Loss of Off Site Power 16 x AA2.43 Occurrence of a turbine trip 3.9 1 000057 Loss of Vital AC Instrument Bus AA2.17 System and component status, using local or remote controls 3.1 I
/6 000058 Loss of DC Power 16 I l l 2.1 Ability to use plant computer to obtain and evaluate parametric information on System or component status 3.0 I 000062 Loss of Nuclear Service Water 14 I I Ix AK3'01 The conditions that will initiate the automatic opening and closing of the SWS isolation valves to the nuclear service water coolers 3.2 1 Ability to operate andlor monitor Emergency air compressors as applicable to loss of instrument 000065 Loss of Instrument Air 18 I l l X AA1.04 3.5 1 I I
~~~ ~
WA Category Totals 4 13 1 2 1 l8 Confidential Page 2 1
CCNPP NRC License Examination March 2004 PWR Examination outline Emergency and Abnormal Plant evolutions - Tier 1 Group 2@6$RO) Form ES-40 1-2 I
4 A G Number WA Topics Imp. RO 1 2 #
X AA2.01 Ability to determine and interpret-Rod position indication to actual rod position 3.7 1 D O 0 0 0 5 InoperabTelStuck Control Rod /l ~ I 1 000024 Emergency Boration ll I I WOO28 /Pzr Level Malfunction /2 000032 Loss of Source Range NI 17 x 2.2.22 Knowledge of limiting conditions for operation and safety limits 3.4 1 000033 Loss of Intermediate Range NI /7 1 I 000036 /Fuel Handling Accident 18 X AA2.02 Ability to determine and interpret-occurrence of a fuel handling incident 3.4 1 000037 SG Tube Leak /3 000051 Loss of Condenser vacuum /4 I I 000059 Accidental Liquid RadWaste Rel. /9 I I 000060 Accidental G a s e o G d w a s t e Re1.19 I I 000061 ARM System Alarms /7 x 2.1.1 Knowledge of conduct of operations requirements 3.7 1 000067 Plant Fire on Site /8 AK3.04 Actions contained in EOP for plant fire on site 3.3 1 000068 Control Room Evac. /8 I I 000069 Loss of CTMT Integrity 15 000074 Inadequate Core Cooling /4 X EA1.16 Ability to operate and monitor RCS in-core thermocouple indicators 4.4 1 000076 High Reactor Coolant Activity 19 I Ix AK2.01 Knowledge of interrelations between the High Reactor Coolant Activity and the process radiation monitors 2.6 1 CElA11 RCS OvercoolinglPTS /4 CElAl3 Natural Circulation /4
~~
CE/AI 6 /Excessive RCS Leakage /2 I I X AA1 .I Ability to operatelmonitorcomponents, functions of control and safety systems including instrumentation, signals, interlocks failure modes, autolman features 3.4 1 CUE09 /Functional Recovery EK3.4 Knowledge for the reasons of RO or SRO function within the control room team in 3.3 such a way that procedures are adhered to and limits are not violated 212121 I GrouD Point Total I 1 9 Confidential Page 3
CCNPP NRC License Examination March 2004 PWR Examination outlin Plant Systems - Tier 2 Group 1 @SRQ Form ES-401-2 System #/Name K K K K K K 1 2 3 4 5 6 0 0 3 Reactor Coolant Pump 004 Chemical and Volume Control X 005 Residual Heat Removal X I Ix1
~ _ _
0 0 6 Emergency Core Cooling
+I
~
008 Component Cooling Water System I I IxI I I Ability to predict the impacts of and correct, control or 007 Pressurizer RelieflQuench X A2'02 Tank mitigate abnormal pressure in the PRT 008 Component Cooling Water X K4.09 The standby feature of the CCW pumps 2.7 010 Pressurizer Pressure Control X K6.03 Knowledge of the effect of a loss or malfunction of PZR sprays and heaters will have on RCS PCS 012 Reactor Protection X K3.02 Knowledge that a loss or malfunction of the RPS will have on the TurbineIGenerator 012 Reactor Protection x 2-4-47 Abiliy to diagnose and recognize trends in an accurate and timely matter using reference material 0 1 3 ESFAS X A
l Ability to predcffmonitor changes in RCS pressure and temperature associated with operating ESFAS Knowledge of connections, cause effect relationship K1 between CCS and SWS cooling system K2.01 -
Knowledge of the bus Dower sumlies.. . .
to CS PumDs I
I I I I I 0 2 6 Containment Spray X Ability to predict impacts of and correct, loss of spray pump A2.07 suction when in recirc due to clogged screen, cavitation 039 Mainand Reheat Steam I I 1 I I I x I I I I I Al.09 I Ability to predicffmonitor changes in Main Steam line RMS 2.5 039 Mainand Reheat Steam I I 1 1 I I I I I x I I A4.01 I Ability to manually operatelmonitor MS supply valves 2.9 X A2.04 Ability to predict impacts of loss of condensate pumps... 2.6 K4.08 FRV operation on basis of feedkteam flow mismatch 2.5 I
X I
A3.06 I Abiliy to monitor MFW isolation I K5.01 I Knowledge of the relationship between AFW flow and RCS heattransfer x I 1 I I I Al.03 I Ability to predicffmonitor interactions when units cross tied X I A2'01 I Ability to predict impacts, correct, de-energized loads that degrade or hinder operation I I I x I I A4.01 I Ability to operate or monitor breakers, fuses in the CR 2.8 Confidential Page 4
CCNF'P NRC License Examination March 2004 PWR Examination outli Plant Systems - Tier 2 Group 2 @SRO) Form ES-401-2 I System #/Name I KI a KI h A A A A G Number KIA Topics Imp. RO 1 2 3 4 1 2 3 4 #
001 Control Rod Drive X A4.08 Ability to manually operate the mode select for CRDS, operation of 3.7 1 MG sets and control panel I 0 0 2 Reactor Coolant I I I I K A1 -06 Ability to predictlmonitor reactor power associated with RCS controls 4.0 1 011 Pressurizer Level Control 014 Rod Position Indication K5.01 Knowledge of operational implications of differences between RPlS 2.7 1 and step counter 0 1 5 Nuclear Instrumentation X Ability to monitor recognition of audio output expected for a given plant A3'05 2.6 condition 0 1 6 Non-Nuclear Instrumentation 0 1 7 In-Core Temperature Monitor I 027 Containment Iodine Removal 1 I I I x A4.03 Ability to manually operatelmonitor H2 sampling and analysis of 3.1 1 cntmt atmosphere including alarms and indications K4.03 Knowledge of the design feature/interlock providina auto isolation 3.2 1
'3.03 Knowled e that the effect of loss of cooling has on spent fuel 3.0 1 ternperatgure w 041 Steam Dumpflurbine Bypass Control 0 4 5 Main Turbine Generator 0 5 5 Condenser Air Removal I 068 Liquid Radwaste I I I I K5.04 Knowledge of operational implications of biological hazards of radiation and resulting goal of ALARA 3.2 1 071 Waste Gas Disposal 0 7 2 Area Radiation Monitoring X Kl.04 Knowledge of connections and causeleffect between ARM and 3.3 1 I I I I Control Room ventilation 0 7 5 Circulating Water 1 0 7 9 Station Air H I I 086 Fire Protection H I X A3.01 Ability to manually operatehonitor starting mechanisms of fire pumps 2.9 1 WA Category Point totals 1 1 1 01 11 1 1 0 2 2 0 10 Confidential Page 6
CCNPP NRC License Examination March 2004 PWR Examination outline Generic Knowledge and Abilities Outline (Tier 3)
A0 Form ES-401-3 Category KIA # 1 Topic Imp. #
n 2.1.3 Knowledge of shift turnover practices 3.0 1 2.1.10 Knowledge of conditions and limitations in the facility license 2.7 1
1. 2.1.14 Knowledge of system status criteria which require the 2.5 1 Conduct of notification of plant personnel Operations Subtotal Y&g&; 3 2.2.28 Knowledge of new and spent fuel movement procedures I 2.6 1 2.2.33 Knowledge of control rod programming 2.5 1
2. 2.2.34 Knowledge of the process for determining the internal and 2.8 1 Equipment external effects on core reactivity Control Subtotal 2.3.10 Ability to perform procedures to reduce excessive levels of 2.9 1
3. radiation and guard against personnel exposure Radiation Control 2.3.11 Ability to control radiation releases 2.7 1 I
Subtotal 2.4.17 Knowledge of EOP terms and definitions 1
4. 2.4.46 Ability to verify that alarms are consistent with plant 3.5 1 n
Emergency conditions Procedures I Plan Subtotal Tier 3 Point Total Confidential Page 7
CCNPP NRC License Examination March 2004 PWR Examination outline Emergency and Abnormal Plant evolutions - Tier 1 Group 1 ( Roe Form ES-401-2 E / N E #/Name./Safety Function G Number WA Topics Imp. SRO#
000007/E02 Reactor Trip - Recovery 11 000008 Pzr Vapor Space Accident 13 000009 Small Break LOCA 13 00001 1 Large Break LOCA 13 I l l 000015117 RCP Malfunctions 14 m.10 When to secure RCPs on loss of cooling or seal injection 3.7 1 000022 Loss of Rx Coolant Makeup 12 000025 Loss of RHR System 14 H I x 2.1.12 Ability to apply Technical Specifications for a system 4.0 1 000026 Loss of Comp. Cooling Water 18 I I I I 1 I
~~
000027 Pzr Press. Ctrl. Sys. Malf. 13 x 2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm 3.6 1 000029 A W S I1 I l l EA2.01 Abilitv to determine or interwet reactor nuclear instrumentation as it amlies to a ATWS 4.7 1 000038 SG Tube Rupture13 000040CE05 Steam Line Rupture 14 I I I
~~ ~ ~ ~~
000054CE06 Loss of Feedwater 14 AA2.08 Ability to determine and interpret Steam flow-feed trend recorder 3.3 1 x 2.1.20 Ability to execute procedure steps 4.2 I 000056 /Loss of Off Site Power 16 I I 000058 Loss of DC Power 16 Ability to determine and interpret that a loss of dc power has occurred: verification that AA2.0 4.1 1 substitute power power sources have come on line 000062 Loss of Nuclear Service Water 14 000065 Loss of Instrument Air 18 K/A Category Totals
CCNPP NRC License Examination March 2004 PWR Examination outline Emergency and Abnormal Plant evolutions - Tier 1 Group 2 ( R O B Form ES-401-2 E/APE MNamdSafets Function IKIKI K A I A I G I Number I KIA Topics I Imp. I SRO 000001 Continuous Rod Withdrawal 11 000003 Dropped Control Rod I1 000005 InoperablelStuckControl Rod 11 I I I 000024 Emergency Boration 11 x 2.2.22 Knowledge of limiting conditions for operations and safety limits 4.1 1 000028 1Pzr Level Malfunction 12 000032 Loss of Source Range NI 17 I l l 000033 Loss of Intermediate Range NI /7 000036 /Fuel Handling Accident 18 000037 SG Tube Leak /3 I l l ' 2'1 .33 Abiliy to recognize indications for system operating parameters which are entry-level conditions for technical specifications 4.0 1 000051 Loss of Condenser vacuum /4 I l l 000059 Accidental Liquid RadWaste Rel. /9 000060 Accidental Gaseous Radwaste Re1./9 000061 ARM System Alarms i7 I l l 000067 Plant Fire on Site 18 I l l 000068 Control Room Evac. 18 I l l x 2.2.25 Knowledge of bases in technical specifications for limiting conditions and safety limits 3.7 1 000069 Loss of CTMT Integrity 15 I l l 000074 Inadequate Core Cooling 14 000076 High Reactor Coolant Activity /9 CE1A1 1 RCS OvercoolinglPTS14 I l l CE/AI 3 Natural Circulation /4 CE/A16 /Excessive RCS Leakage 12 X AA2.1 Facility conditions and selection of appropriate procedures during abnormal and emergency conditions 3.5 ,
CEIEO9 /Functional Recovery X EA2.1 Facility conditions and selection of appropriate procedures during abnormal or 4.4 emergency operations 2 3 5
CCNPP NRC License Examination March 2004 PWR Examination outline Plant Systems - Tier 2 Group 1 Form ES-401-2 W
A A A A G Number K/A Topics Imp. SRO 1 2 3 4 #
1 006 Emergency Core Cooling I I I I I I 007 Pressurizer Relief/Quench Consequences of high or low flow rate and temperature: the X A2*07 flow rate at which the CCW standby pump will start 2.8 1 010 Pressurizer Pressure Control I I I I I I I 022 Containment Cooling I I I I I I X Ability to predict impacts of and correct or control-Loss of service water 3.2 1 026 Containment Spray 039 Main and Reheat Steam 056 Condensate E n Feedwater I I I I I I I 061 Auxiliary Feedwater I I I I I I x 2.1.32 Ability to explain and apply all system limits and precautions 3.8 1 076 Service Water 078 Instrument Air 103 Containment Knowledge of system status criteria which require the 2.1 notification of plant personnel 3.3 1 I WA Category Point totals 2 2 Grout, Point Total 4
CCNPP NRC License Examination March 2004 PWR Examination outline Plant Systems - Tier 2 Group 2 ( R O a Form ES-401-2 System m a m e IK WA ToDics 001 Control Rod Drive 0 0 2 Reactor Coolant I 0 1 1 Pressurizer Level Control X A2.05 heatersto predict the impacts of and correct or control-Loss of pressurizer Ability 3.7 1 0 1 5 Nuclear Instrumentation I 016 Non-Nuclear Instrumentation I 017 In-Core Temperature Monitor I 0 2 8 H2 Recombiner and Purge Control 029 Containment Purge I 034 Fuel Handling Equipment x 2.4.49 Ability to perform wlo reference t o procedures actions that require 4.0 1 immediate operation 041 Steam Dumpflurbine Bypass Control 0 4 5 Main Turbine Generator 055 Condenser Air Removal 068 Liquid Radwaste 071 Waste Gas Disposal I 0 7 2 Area Radiation Monitoring I 0 7 5 Circulating Water 0 7 9 Station Air 086 Fire Protection WA Categorv Point totals Ill I Ill I Grouo Point Total I 1 2
CCNPP NRC License Examination March 2004 PWR Examination outline Generic Knowledge and Abilities Outline (Tier 3)
SRO ES-401-3 1~rm Category K/A # Topic I SRO-Only ,-
I ImD. I #
2.1.2 Knowledge of operator responsibilities during all modes of plant 3.8 1 operation 2.1.15 Ability to manage short-term information such as night and 3.0 1 1.
Conduct of standing orders Operations Subtotal 2.2.22 2.
Equipment Control Subtotal 2.3.3 3.
Radiation Control 2.3.7 Subtotal 2.4.4 Ability to recognize abnormal indications for system operating parameters which are entry level conditions for emergency and 4.3 1
4. abnormal operating procedures Emergency Procedures 2.4.18 Knowledge of the specific bases for EOPs 3.6 1 I Plan Subtotal Tier 3 Point Total
Tier I Randomly Reason for Rejection Group Selected WA RO EXAM 111 026AKl. None listed, selected 026AK3.01 111 056AK2.07 Low RO importance, selected 056AA2.43 1I1 057AK1. None listed, selected 057AA2.17 111 065AA1.05 RPS unaffected by loss of IA, selected 065AA1.04 111 027AK3.04 I Routine Rx tri! gressure restor s at he am time or before level, due ti hinh he t r capacity-seeected 02Qk3.83 111 026AK3.01 WS valves to CCWS coolers receive no automatic signals, selected 8.02 211 026K5. None > 2.5, selected 026A2.07 211 039K2. None > 2.5, selected 039A1.09 211 076K6 None > 2.5, selected 076A1.02 211 078K6 None > 2.5, selected 078K4.02 211 004K4.14 Over sampled K4, selected 004K6.04 211 007K3.01 Same WA e e d I st exam, mi i al uestions tg mgtch WA, replaced wTOd8@.O?to p r e s e r v e m nlmber distribution 212 014A3. None listed, selected 014K5.01 IRU high temps not addressed in procedures, spray retired in place, 212 027A2.01 selected 002A1.06 212 029K6. None listed, selected 029K4.03 212 001A4.12 Not applicable, no limit to prevent coil burnout, selected 001A4.08 SRO EXAM 111 055 2.1.6 Not applicable to SRO during SBO, selected 2.1 2 0 1I1 026 2.1.12 Over sampled CC system, selected APE 025 211 059 2.1.14 Over sampled Main Feed, selected 103 (Containment)
ES-301 Control Roodln-Plant Systems Outline Form ES-301-2 Facility: CCNPP Date of Examination: 03/01/2004 Exam Level (circle one): SRO(I) / SRO(U) Operating Test No.:
Y Control Room Systems (8 for RO;7 for SRO-I; z or 3 for SRO-U)
Type Safety System/JPM Title Code* Function
a. CVCSIShift Charging Pumps per OI-2A (Discharge relief lifts on 2 started pump) 004A2.22 3.2/3.1 N S 3
b. ECCSiTest LPSl check valves per STP-0-65-1 006A4.02 4.0/3.8
~
4-primary
c. RHWlnitiate SDC (An RCS leak develops when suction MOVs are opened) 005A1.05 3.3/3.3
d. Emergency Diesel Generatorsnransfer 4 KV bus from an EDG to Offsite Power 064A4.06 3.9/3.9
e. Control Rod Drive System/Realign a CEA (CEA drops when moved) 003AA1.02 3.6/3.4
f. Verify the Containment Environment Safety Function 022A4.01,4.04 3.9/3.8 and 3.W3.2
g. Process RMS/Test Containment Atmosphere Rad. Mon. per STP-0 1 073A4.03 3.V3.2
h. Secure a Circulating Water Pump per OI-14A 075A2.02 2.5/2.7
D 4 I. Align AFW pump speed/flow control to 1(2)C43 secondary 068AA1.12 4.4/4.4 D 2
j. ESFAS/Re-energize ESFAS cabinet per 01-34 013K2.01 3.6/3.8 N,R,A 9
k. Liquid RadWasteIPerform flush of Rad. Monitor per OI-17C.
(Discharge valve shuts during flush) 068A3.02 3.6/3.6 NUREG-1021,Draft Revision 9
ES-301 Control Room/ln-Plant Svstems Outline Form 13-301 -2 Facility: CCNPP Date of Examination: 03/01/2004 Exam Level (circle one): RO / SRO(I) / SRO(U) Operating Test No.:
Control Room Systems (8 for RO; 7 f o e z or 3 for SRO-U)
Type Safety System/JPM Title Code* Function N, A, s
a. CVCS/Shift Charging Pumps per OI-2A (Discharge relief lifts on 2 started pump) 004A2.22 3.2/3.1 3
b. ECCS/Test LPSl check valves per STP-0-65-1 006A4.02 4.0/3.8 4-primary
c. RHWlnitiate SDC (An RCS leak develops when suction MOVs are opened) 005A1.05 3.3/3.3 6
d. Emergency Diesel Generatorsmransfer 4KV loads from an EDG to Offsite Power 064A4.06 3.913.9
e. Control Rod Drive System/Realign a CEA (CEA drops when moved) 003AA1.02 3.6/3.4
f. Verify the Containment Environment Safety Function 022A4.01.4.04 3.9/3.8 and 3.113.2 N, s 7
g. Process RMS/Test Containment Atmosphere Rad. Mon. per STP-0-33-1 073A4.03 3.1/3.2 h.
D 4
1. Align AFW pump speed/flow control to 1(2)C43 secondary 068AA1.12 4.4/4.4 D 2
j. ESFAS/Re-energizeESFAS cabinet per 01-34 013K2.01 3.6/3.8 NR,A 9
k. Liquid RadWaste/Perform flush of Rad. Monitor per 01-17C.
(Discharge valve shuts during flush) 068A3.02 3.6/3.6
NUREG-1021, Draft Revision 9 ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2 Facility: CCNPP Date of Examination: 03/01/2004 Exam Level (circle one): RO / SRO(I) / SRO(U) Operating Test No.:
Control Room Systems (8 for RO; 7 for SRO-I; 2 or 3 for Type Safety SystemlJPM Title Code* Function
a. CVCS/Shift Charging Pumps per OI-2A (Discharge relief lifts on started pump) 004A2.22 3.2/3.1
b. Emergency Diesel Generatorsmransfer 4 W bus from an EDG to Offsite Power 064A4.06 3.9/3.9 C. RHRllnitiate SDC (An RCS leak develops when suction MOVs are opened) 005A1.05 3.3/3.3 N,A,L,S 4-primary d.
e.
1 f.
I In-Plant Systems (3 for RO; 3 for SRO-I; 3 or 2 for R O - d D 4
1. Align AFW pump speed/flow control to 1(2)C43 secondary 068AAl .I2 4.414.4 N RIA I 9
j. Liquid RadWaste/Perform flush of Rad. Monitor per OI-17C.
(Discharge valve shuts during flush) 068A3.02 3.6/3.6
NUREG-1021, Draft Revision 9 ES-301 Administrative Topics Outline Form ES-301-1 Facility: CCNPP Date of Examination: 03/01/2004 1 Examination Level (circle one):@/ SRO Operating Test Number:
Administrative Topic Describe activity to be performed (see Note)
A. 1.a 2.1.19 3.0 Ability to use plant computer to obtain and evaluate parametric Conduct of information on plant system or component status Operations Enter feed flow correction factors into the plant computer per OI-50A, verifying NI calibration.
A.1 .b 2.1.25 2.8 Conduct of Ability to obtain and interpret station reference materials such as graphs, monographs, and tables which contain performance Operations data Verify shutdown margin in mode 3 with one stuck CEA A.2 2.2.13 3.6 Knowledge of tagging and clearance procedures Equipment Control Verify tagout boundaries per NO-1-112 A. 3 Radiation Control A.4 2.4.43 2.8 Knowledge of emergency communications and techniques Emergency Plan Recall station personnel due to a hurricane warning per ERPIP 3.0 Att. 20, D.5.0 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.
NUREG-1021, Draft Revision 9 ES-301 Administrative Topics Outline Form ES-301-1
~~
Facility: CCNPP Date of Examination: 03/01/2004 Examination Level (circle one): RO /@Operating Test Number:
Ad ministrative Topic Describe activity to be performed (see Note)
A.1 .a A.l . I 2 4 0 Conduct of Ability to apply technical specifications to a system Operations Review /Approve an IR for Operability determination A.l .b 2.1.25 3.1 Conduct of Ability to obtain and interpret station reference materials such as graphs, monographs, and tables which contain performance Operations data Verify an ECC A. 2 2.2.13 3.8 Knowledge of tagging and clearance procedures Equipment Control Approve a clearance or restoration per NO-1-112 A. 3 2.3.6 3.1 Knowledge of the requirements for reviewing and approving release permits Radiation Control Review an RCWMT discharge permit A.4 2.4.44 4.0 Knowledge of emergency plan protective action recommendations Emergency Plan Classify an EAL and provide protective action recommendations per the ERPIP 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.
Simulation Facility Calvert Cliffs ScenarioNo.: 1 Op Test No.: 1 Examiners: operators: SRO To evaluate the applicants ability to conduct a unit power decrease and implement the ARUS,Oh,AOPs, as Dbjectives: appropriate, for malfunctioning systems andor controls including failure of a VCT level Transmitter with an intermediate indication on suction switchover, erratic operation of a FW controller, and a dropped CEA. After the unit is stabilized, the System operator directs power be reduced 100 MWe in the next 10 minutes. Once power is stabilized,a high SGFP vibration alarm occurs followed shortly thereafter by a SGFP trip. After it is determined the SGFP cannot be reset, the unit is tripped and EOP-0 entered. The feed system pernubation causes a feed rupture. A blockage in the AFW suction line prevents AFW flow from being established requiring EOP-3 be implemented. The crew will attempt to restore feed and then initiate OTCC. Upon initiation of OTCC, one PORV will not open.
Initial Conditions: The plant is at 100%Power, MOC 500KV Bkr.,552-23 tripped open 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ago, 13 HPSI Pp is 00s.
13 SRW Pp is 0 0 s 12A Travelling Screen -flow switch is jumpered 12 4KV Bus is on the Alternate Feeder hmover: Present plant conditions: 100% power, MOC; Unit 2 is in MODE 5 - no CW Pps and 23 AFW Pp unavailable.
Power history: 100%power for previous 68 days.
Equipment out of service:
1) 500KV Bkr.,552-23 tripped open 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ago. Investigating.
2) 13 HPSI Pp motor bearing failure during STP two (2) days ago. It is nearly reassembled, expected to be returned to service in 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months , T.S.3.5.2 Action Statement entered 49 hours5.671296e-4 days 0.0136 hours 8.101852e-5 weeks 1.86445e-5 months ago.
3) 13 SRW Pp has a broken shaft, expected return tomorrow.
4) 12A Travelling Screen -flow switch is jumpered - will be worked tomorrow.
5) 12 4KV Bus is on Alternate Feeder, 152-1209 - the Normal Feeder, 152-1201, has a loudly buzzing relay, electricians willbe investigating later this shift.
Surveillances due: None Instructions for shift:
1) Maintain 100%power.
2) Perform PMT on 13 HPSI when returned to service.
0401.dococ
Event Event Event me* Description 13 HPSI Pp 00s.
RF-rackout 13 SRW 00s.6 RWT Outlet 1-SI-504-MOV - GREEN light ON, PWR ON light ON.
PORV-402 switch in Auto.
AFW pump suction blockage -
(CD-161) (12 CST suction valve shut and indicating open.)
I RO Several minutes after the crew takes the watch, VCT Level transmitter LT-227 fails low. This causes Chg. Pp suction to shift to the RWT. The RO informs the CRS. After verifying it is an instrument Mure, the CRS directs the RO to shift Chg. Pp suction back to the VCT. RO should note the failure of RWT Outlet (504-MOV) to go fully open. The CRS directs MOV-501 be reopened and MOV-504 to be shut. The AB0 should be dispatched to check VCT level and MOV-504 position locally. The OWC is contacted for assistance.
I CRO FRV Controller 1111begins to operate erratically. The CRO should notice oscillating SGFP WM, MWe, and SG level. The CRS should direct the Controller be shifted to manual and AOP3G implemented. The PDI Control Switch is placed in MAIN FAIL and level controlled on the PDI. The CRS should then direct the controller be placed in bypass override. The CRS contacts the OWC for assistance.
+
3 5
CEDSO12-0 1 C reduction PNL OVR on R
N C
RO RO CRO CRO 1
After the 1111failure, CEA 0 1drops. The RO should acknowledge the alarms, inform the CRS and refer to the ARM. The CRS should implement AOP-1B and address T.S. 3.1.4. The primary will be stabilized and realignment time determined. After notifying maintenance and correcting the cause, realignment will be commenced.
Next, the System Operator calls the Control Room and reports a transformer cooling problem at Waugh Chapel and requests they reduce unit load by 100 W e within the next 10 minutes. The CRS should brief the crew and commence a rapid downpower to ~ 9 0 % .
After power has been stabilized, a SGFP supeMsory alarm is received on 11 SGFP. The alarm is determined to be a valid vibration alarm in the danger C-69 and Status Panel range at about 7 mils. The TBO or PPO and the System Engineer are (Supervisory dispatched to the pump. After a course of action is determined, (reduce load on Inst.) the SGFP), 11SGFP trips. The CRS may direct an attempted reset of 11SGFP pump. The SGFP will not reset. The CRS will direct the RO to trip the unit anc EOP-O be implemented.
6 CDOOS - M A l l After the reactor trip, an unisolable rupture occurs on the feed pump suction 20% no ramp (upstreamof 1-FW-101). The feed and condensate headers will have to be isolated. When the CRO attempts to initiate AFW flow, flow will not be able to be established due to the common AFW Pump suction being blocked. The CRS should evaluate the EOP-O flowchart and determine EOP-3 should be implemented.
M A l l Prior to EOP-3 entry blowdown will be isolated and the RCPs tripped. A rapid cooldown to less than 465OF will be commenced. When both SG levels are less than -350, OTCC will be initiated. When the RO attempts to open PORV-402, it will not open. The crew should evaluate Attachment 17 and determine the potential success of OTCC. After this assessment, the scenario can be 1 terminated.
ity (I)nstnunent, (C)omponent, w a j o r Transient 0401.dococ
SCENARIO 1 OVERVIEW u The candidates will take the shift at 100% power with instructions to maintain power.
Several minutes after the crew takes the watch, VCT Level transmitter LT-227 fails low. This causes Chg. Pp suction to shift to the RWT. The RO informs the CRS. After v e m g it is an instrument failure, the CRS directs the RO to shift Chg. Pp suction back to the VCT. RO should note the failure of RWT Outlet (504-MOV) to go fully open. The CRS directs MOV-501 be reopened and MOV-504 to be shut. The AB0 should be dispatched to check VCT level and MOV-504 position locally. The OWC is contacted for assistance.
FRV Controller 1111begins to operate erratically. The CRO should notice oscillating SGFP RPM, M e , and SG level. The CRS should direct the Controller be shifted to manual and AOP-3G implemented. The PDI Control Switch is placed in MAIN FAIL and level controlled on the PDI. The CRS should then direct the controller be placed in bypass override. The CRS contacts the OWC for assistance.
After the 1111failure, CEA 01 drops. The RO should acknowledge the alarms, inform the CRS and refer to the ARM. The CRS should implement AOP-1B and address T.S. 3.1.4. The primary will be stabilized and realignment time determined. After notifying maintenance and correcting the cause, realignment will be commenced.
Next, the System Operator calls the Control Room and reports a transformer cooling problem at Waugh Chapel and requests they reduce unit load by 100 MWe within the next 10 minutes. The CRS should brief the crew and commence a rapid downpower to ~ 9 0 % .
After power has been stabilized, a SGFP supervisory alarm is received on 11SGFP. The alarm is determined to be a valid vibration alarm in the danger range at about 7 mils. The "30 or PPO and the System Engineer are dispatched to the pump. After a course of action is determined, (reduce load on the SGFP), 11 SGFP trips. The CRS may direct an attempted reset of 11 SGFP pump. The SGFP will not reset. The CRS will direct the RO to trip W the unit and EOP-Obe implemented.
After the reactor trip, an unisolable rupture occurs on the feed pump suction (upstream of 1-FW-101). The feed and condensate headers will have to be isolated. When the CRO attempts to initiate AFW flow, flow will not be able to be established due to the common AFW Pump suction being blocked. The CRS should evaluate the EOP-O flowchart and determine EOP-3 should be implemented.
Prior to EOP-3 entry blowdown will be isolated and the RCPs tripped. A rapid cooldown to less than 465OF will be commenced. When both SG levels are less than -350", OTCC will be initiated. When the RO attempts to open PORV-402, it will not open. The crew should evaluate Attachment 17 and determine the potential success of OTCC. After this assessment, the scenario can be terminated.
0401.dococ
Simulation Facility Calvert Cliffs Scenario No.: 2 Op Test No.: 1 Examhers: operators: -
SRO To evaluate the applicants ability to conduct a unit power decrease, to implement the ARMS, Oh, AOPs, as lbjectives: appropriate, for malfunctioning systems andlor controls including failure of 11 CCW Pp, a PZR spray valve opening, an RCP seal and the ADV controller. After the ADV controller fails the remaining seals fail on the affected RCP resulting in an RCS leak. After EOP-O is entered, the RCS leak causes a SIAS actuation. SIAS A fails to actuate requiring 11 or 12 HPSI Pump to be started manually to establish HPSI flow. As EOP-O progresses, a steam leak begins in CNMNT from 12 SG requiring EOP-8 be implemented. The ADV Controller will not operate from the Control Room,the crew will have to operate the ADVs locally.
nitial Conditions: The plant is at 100% Power, MOC 13 HPSI Pp is 00s.
13 CCW Pp is 00s 12A Travelling Screen -flow switch is jumpered 12 4KV Bus is on the Alternate Feeder hmover: Present plant conditions: 100% power, MOC; Unit 2 is in MODE 5 - no CW Pps and 23 AFW Pp unavailable, Power history: 100% power for previous 68 days.
Equipment out of service:
1) 13 HPSI Pp motor bearing failure during STP two (2) days ago. It is nearly reassembled, expected to be returned to service in 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months . T.S. 3.5.2 Action Statement entered 49 hours5.671296e-4 days 0.0136 hours 8.101852e-5 weeks 1.86445e-5 months ago.
2) 13 CCW Pp has a bad bearing, expected return on night shift.
3) 12A Travelling Screen - flow switch is jumpered - will be worked tomorrow.
4) 12 4KVBus is on Alternate Feeder, 152-1209 -the Normal Feeder, 152-1201, has a loudly buzzing relay, electricians will be investigatinglater this shift.
Surveillances due: STP-O-29 (CEAMovement Test) due by end of shift. SM will discuss with CRS shortly aftei turnover.
Instructions for shift
1) Maintain 100%power.
0402.dococ
Event Ma.
No.
I Event Type*
Event Preload ESFA002-0 1 SIAS A fails to actuate automatically and manually.
ESFAOO~~O~
RF-rackout 13 HPSI Pp 00s.
RF-raCkoUt 13 CCW Pp 00s.
T 1 About 3 minutes after the crew takes the watch, 11 CCW Pp trips. The CRO will acknowledge the alarms, inform the CRS and refer to the ARM. The crew will check for common mode failure and the CRS will direct the CRO to start 12 CCW Pp. The CRS refers to AOP-7C and T.S. 3.7.5. The CRS contacts the OWC for assistance.
2 RCSO16 I RO PRZR Spray Valve IOOE fails open. The RO will note lowering RCS pressure and inform the CRS of the open spray valve. May refer to the PZR pressure ARM. Spray valve control will be shifted from both to IOOF. lOOE will be verified to be going shut and RCS pressure restored. The CRS contacts the OWC for assistance.
3 RCSO11-01 C RO Next, the lower seal (W stage) fails on 11A RCP. The RO notes the alarm, (O-100%over informs the CRS and refers to the ARM. The crew should determine the lower 3 min) seal on 11A RCP has failed. The OWC and system engineer are contacted for assistance.
1 4 MS015 CRO Two minutes after the notifications have been made the ADV controller fails high causing the ADVs to open. The CRO will inform the CRS. The CRS should direct the CRO to take manual control of the ADVs and shut them. The RO should maintain reactor power less than 100%. The CRS refers to AOP-7K.
The CRS contacts the OWC for assistance.
5 RCSO12-01 R RO After the CRS has referenced A 0 P - X the 11A RCP middle seal fails. The (O-100% over N CRO crew should implement ARM guidance and begin an expeditious shutdown.
2 min) The CRS refers to OP-3for guidance on for the shutdown.
RCSO13-01 M All After power has been reduced at least 5%, the 11A upper seal begins to fail (over 2 min) followed shortly thereafter by the failure of the vapor seal. With all the seals RCSOO3 (50 failed, RCS leakage begins via the seals. The CRS will trip the unit based on gpm over 5 high RCP seal temperature and EOP-O is implemented. The CRS will direct reactivity be performed then 11A RCP be tripped. When SIAS actuates, train RCSO14-0 1 A fails to automatically and manually initiate. To establish HPSI flow 11 or (over 4 min) 12 HPSI must be started manually.
MS010-02 M A l l After the first pass through the safety functions is complete, a steam break in (0-25% over CNMNT begins. The CRS should recognize two events are taking place and 3 min) implement EOP-8. Success Paths will be selected and implemented. When the crew attempts to operate the ADVs,they will not operate from the Control Room. To establish heat removal via 11 SG, the ADV will have to be operated Override locally. After 12 SG is isolated, the scenario can be terminated.
ADVs closed
& in manual (R)eactivity (i)nstrument, (C)omponent, m a j o r Transient u
0402.dococ
SCENARIO 2 OVERVIEW W The candidates will take the shift at 100%power with instructions to maintain power.
After the crew has taken the shift, 11CCW Pp trips. The CRO will acknowledge the alarms, inform the CRS and refer to the ARM. The crew will check for common mode failure and the CRS will direct the CRO to start 12 CCW Pp. The CRS should refer to AOP-7C and T.S. 3.7.5. The CRS contacts the OWC for assistance.
PRZR Spray Valve lOOEfails open. The RO will note lowering RCS pressure and inform the CRS of the open spray valve. May refer to the PZR pressure ARM. Spray valve control will be shifted from bothto 1OOF. lOOE will be verified to be going shut and RCS pressure restored. The CRS contacts the OWC for assistance.
Next, the lower seal (first stage) fails on 11A RCP. The RO notes the alarm, informs the CRS and refers to the ARM. The crew should determine the lower seal on 11A RCP has failed. The OWC and system engineer are contacted for assistance.
Two minutes after the notifications have been made the ADV controller fails high causing the ADVs to open. The CRO diagnoses the open ADVs and inform the CRS. The CRS directs the CRO to take manual control of the ADVs and shut them. The RO should maintain reactor power less than 100%. The CRS refers to AOP-7K. The CRS contacts the OWC for assistance.
After the CRS has referenced AOP-7K, the 11A RCP middle seal fails. The crew should implement ARM guidance and begin an expeditious shutdown in accordance with OP-3. The CRS refers to OP-3and directsthe unit shutdown at a rate less than 30o/dhour. The appropriate notifications should be made.
After power has been reduced at least 5%, the 11A RCP upper seal begins to fail followed shortly t h e r d e r by the failure of the vapor seal. With all the seals failed RCS leakage begins via the seals. The CRS will trip the unit based on high RCP Controlled Bleed off temperature and EOP-O implemented. The CRS will direct reactivity be b performed then 11A RCP be tripped. When SIAS actuates, train A fails to initiate automatically and manually.
To establish HPSI flow 11 or 12 HPSI must be started manually.
After the first pass through the safety functions is complete, a steam break in CNMNT begins. The CRS should recognize two events are taking place and implement EOP-8.Success Paths will be selected and implemented.
When the crew attempts to operate the ADVs,they will not operate from the Control Room. To establish heat removal via 11 SG, the ADV will have to be operated locally. After 12 SG is isolated, the scenario can be terminated.
0402.dococ
' SimulationFacility Calvert Cliffs Scenario No.: 3 (Spare) Op Test No.: 1 Examiners: Operators: -
SRO Objectives: To evaluate the applicant's ability to conduct a unit power increase, to implement the ARMS, OIs, AOPs, as appropriate, for maZfunctoning systems and/or controls including a failure of the SRW Controller for the Main Generator H2 Cooler, a failure of a letdown backpressure regulator valve and a PZR pressure control channel and selector switch,then a loss of 14 4KV Bus occurs. The crew will shift charging suction back to the VCT, stabilize the unit and implement AOP-71. Next,a loss of offsite power occurs and the Main Turbine does not Trip. The crew should determine a reactor trip is required, trip the unitand implement EOP-O. The lADG does not start and the OC DG Tie Bus is faulted resulting in a Station Blackout. The crew will implement EOP-7. As actions for EOP-7 are being performed, 11 AFW Pp will trip. The crew will be able to restore AFW using 12 AFW Pump allowing them to remain in the optimal procedure.
Initial Conditions: The plant is at 75% Power, MOC (IC-18) 13 SRW Pp is 00s 12 ChargingPp is 00s 12 4KV Bus is on the Alternate Feeder Turnover: Present plant conditions: 75% power, MOC; Unit 2 is in MODE 5 - no CW Pps and 23 AFW Pp unavailable.
Power history: 100%power for previous 190 days. Then reduced to 6 8 % 3 days ago for 12 SGFP repairs.
Equipment out of service:
1) 13 SRW Pp has a broken shaft, expected repair tomorrow-noon.
2) 12 Charging Pp is 0 0 s due to severe packing leakage. The packing is being replaced. Expected to be ready for PMT in about 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
3) 12 4KV Bus is on Alternate Feeder - the Normal Feeder has a loudly buzzing relay, electricians will be investigating later this shift.
Surveillances due: None.
Instructions for shift:
1) Unit 1 had been at =68% power the previous 3 days to repair a steam leak on 12 SGFP. Both SGFPs are operating in parallel. Continue raising power to 100%per OP-3. Raise power as soon as possible, ESO says needed for grid stability, maximum emergency generation.
2) Perform PMTon 12 ChargingPump when returned to service.
0403.doc 1
Event Malf. I Event Event No. No. I Type* Description Preload 1Goc004 IC DG tie bus faulted.
)Go02-02 LA DG start failure.
Wrackout 13 SRW Pp 0 0 s moo2 M i n e fails to trip.
IF-raCkOUt 12 Chg Pp 00s XF- 12 CH.
?pto 14 bus
?NL OVR Pressurizer Pr&s Controller Selector Switch in channel X.
I WA R RO After the crew takes the watch, the CRS should brief the crew on the power N CRO increase to 100% per OP-3 and that the ESO needs the power as quickly as possible. After the brief the RO and CRO coordinate to raise power while maintaining plant parameters within acceptable limits and Tc on program.
r~o30-01 I CRO After power has been raised at least 5%, the SRW controller for the Main
[closed) Generator HZcooler fails the valve shut. When the high temperature alarm is received, the CRO acknowledges the alarm, informs the CRS and refers to the ARM. The CRO should determine TIC-1608 has failed low causing the CV to go shut, take manual control and restore Hztemperature. The OWC is contacted for assistance.
3 C RO Next, Letdown backpressure valve 1-CVC-201Q fails shut. The RO acknowledges the alarm, informs the CRS and refers to the ARM. The CRS should direct the RO to shift to 1-CVC-201P and restore letdown flow. The OWC is contacted for support.
4 RCSO23-01 I RO Next, PRZR Press Channel lOOX fails high. The RO acknowledges the alarms, (high) informs the CRS and refers to the ARM. The RO should note the lowering RCS pressure and the open spray valves. Pressure control should be shifted to channel Y and the spray valves verilied shut. The RO notes the spray valves did not shut. The CRS should direct the spray controller be taken to manual and thc spray valves closed. The CRS contacts the OWC for assistance.
4KVOO1-04 C CRO After the RCS is stabilized, 14 4KV Bus is lost. The crew should stop any power increase, diagnose the bus loss, shift Charging Pump suction back to the VCT and stabilize the unit. The CRS implements AOP-71. The CRO will perform APO-71 actions as directed by the CRS. The OWC is contacted for support.
swYD002 M ALL About 5 minutes after the bus loss, a loss of offsite power occurs. 1A DG does not start and 14 4KV bus is locked out. The Main Turbine dues not trip automatically. The crew should recognize a reactor trip is required and trip the unit manually. The OC Diesel will not energize the OC Tie Bus due to a fault.
The crew will implement EOP-Oand diagnose a stationblackout exists.
7 AFwOO1~01 C A L L In EOP-7, 11 AFW Pp trips but they can restore AFW using 12 AFW Pump.
1A DG will become available and 11 4KV Bus re-energized. After the bus is re-energized the scenario can be terminated.
N o d , (R)eactivity (I)nstrument, (C)omponent, Wajor Transient 0403.doc 2
SCENARIO 3 OVERVIEW i/ The candidates will take the shift at W5% power with instructions to raise power to 100% per OP-3.
After the crew takes the watch, the CRS should brief the crew on the power increase to 100% per OP-3 and that the ESO needs the power as quickly as possible. After the brief the RO and CRO coordinate to raise power while maintainingplant parameters within acceptable limits and Tc on program.
After power has been raised at least 5%, the SRW controller for the Main Generator H2 cooler fails the SRW valve closed. When the generator H2high temperature alarm is received, the CRO should acknowledge the alarm, inform the CRS and refer to the ARM. The CRO should determine TIC-1608 has failed low causing the CV to go shut, take manual control and restore H2 temperature. The OWC should be contacted for assistance.
Next, Letdown backpressure valve 1-CVC-201Q fails shut. The RO acknowledges the alarm, informs the CRS and refers to the ARM. The CRS should direct the RO to shift to 1-CVC-201P and restore letdown flow. The OWC is contacted for support.
Next, PRZR Press Channel lOOX fails high. The RO acknowledges the alarms, informs the CRS and refers to the ATM. The RO should note the lowering RCS pressure and the open spray valves. Pressure control should be shifted to channel Y and the spray valves verified shut. The RO notes the spray valves did not shut. The CRS should direct the spray controller be taken to manual and the spray valves closed. The CRS contacts the OWC for assistance.
After the RCS is stabilized, 14 4KV Bus is lost. The crew should stop any power increase, diagnose the bus loss, shift Charging Pump suction back to the VCT and stabilize the unit. The CRS implements AOP-71. The CRO will perform APO-71 actions as directed by the CRS. The OWC is contacted for support.
About 5 after the bus loss, a loss of offsite power occurs. 1A DG does not start and 14 4KV Bus is locked L out. The OC Diesel will not energize the OC Tie Bus due to a fault. The Main Turbine will not trip automatically.
The crew should recognize a reactor trip is required and trip the unit manually. The crew will implement EOP-O and diagnose a station blackout exists.
In EOP-7 11 AFW Pp will trip but they can restore AFW using 12 AFW Pump. 1A DG will become available and the 11 4KV Bus energized. After the bus is energized the scenario can be terminated.
0403.doc 3

ML040960319

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