ML20155G471
| ML20155G471 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 06/23/1998 |
| From: | Graesser K COMMONWEALTH EDISON CO. |
| To: | Hironori Peterson NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| Shared Package | |
| ML20155F633 | List: |
| References | |
| BYRON-98-0195, BYRON-98-195, NUDOCS 9811090063 | |
| Download: ML20155G471 (59) | |
Text
Commonwealth lide.on Osmpany liyron ba.cratanr, Nation 4150 North German Church Road liyron, 11.61010 9794 Tel HI523 45411 June 23,1998 LTR: Byron 98-0195 FILE: 1.06.5110 Mr. Hironori Peterson U.S. Nuclear Regulatory Commission Region 11I 801 Warrenville Road Lisle,'ll 60532-4351
Dear Mr. Peterson:
Enclosed is the integrated examination outline (written and operating test) which Byron Generating Station is submitting for review, comment, and approval for the initial License Examination scheduled for the week of September 14,1998, at Byron Generating Station.
This outline has been developed in accordance with Interim Revision 8 of NUREG-1021 (" Operator Licensing Examiner Standards").
Please ensure that these materials are withheld from public disclosure, until after the examinations are complete.
If you have any questions or concerns regarding this outline, please contact Mick Brown at (815) 234-5441 extension 3133.
Sincerely,
- O j
Ke.n rae er l
Site Vice-Pre ident i
Byron Nuclear Generating Station ec:
No Enclosures Regulatory Assurance T. Gierich T. Schmidt E. Bendis S. Pettinger P. Hippely R. Franklin R. Brown Class File gg g0g 4
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pA98byltrs\\98 0195. doc A t!nicom Company U N0$00k )
ES 201 Examination Outline Form ES-201-2 Quality Assurance Checklist
<=r FecBity: Of M OM/ r / 4elo 2 Date of Examination:
fdPr /4, /19f initials Itern Task Description a
b c
1.
- a. Verify that the oudine(s) fit (s) the appropriate model per ES-401.
[
- b. Assess whether att six knowledge and four abiity categories are appropriately sampled.
- c. Assess whether the oudine over-emphasizes any systems, evolutions, or generic topics.
- d. Assess whether the repetition from previous examination oudines is excessive.
$6 2.
- a. Using Form ES-301-5, verify tut the proposed scenario sets cover the required number of f //ok h normal evolutions. Instrument and component faRurea, and major transients.
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I
- b. Assess whether there are enough scenario sets (and spares) to test the projected number M
and mix of applicants in accordance with the expected crew composition and rotation schedule without compromising exam integrity; ensure each applicant can be tested usig at y
least one new scenario and scenarios wil not be repeated over successive days.
- c. To the extent possible, assess whether the oudine(s) conform (s) with the qualitative and
. h b y
quantitative criteria specified on Form ES-301-4 and described in Appendix D.
/ v 3.
- a. Verify that the outline (s) contain{s) the required number of control room and in-plant tasks and verify that no more than 30% of the test materialis repeated from the last NRC Ud.
d examination.
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- b. Verify that the tasks are distributed among the safety function groupings as specified in ES-301; one task shall require a low-power or shutdown condition, one or two shah require the applicant to implement an alternate path procedure, and one should require entry to the RCA.
j
- c. Verify that the required administrative topics are covered, with emphasis on performance-ja [
based activities.
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- d. Determine if there are enough different outlines to test the projected number and mix of
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applicants and ensure that no more than 30% of the items are duplicated on successive days.
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appropriate exam section.
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- b. Assess whether the 10 CFR 55.41/43 and 55.45 sampling is appropriate.
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- c. Ensure the: K/A importance ratings (except for plant specific priorities) are at least 2.5.
- d. Check for duplication and overlap among exam sections.
- e. Check the entire exam for balance of coverage.
Printed Na e / Signature Date
- a. Author b ** D (= - N**a / V74r 6 b*/4f M* # A
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- b. FacBity Reviewer (*)
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- c. Chief Examiner
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- d. NRC Supervisor h8Z VW 469C(( [
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(*) Not applicable for NRC-developed examinations MMl b fAlth f*f4 tid 4 (vW(,/k h /2 N 0 '"'1 # N " N ' ^
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g ES-301 -
Administrative Topics Outline Form-ES-301-1 Facility: Byron I & 2 Date of Examination:
September 14,1998 Examination Level: -RO Operating Test Number:
1 Administrative Describe method of evaluation:
Topic / Subject 1.
ONE Administrative JPM, CR Description 2.
TWO Administrative Questions Al Plant Parameter Verification /
- 1. JPM K, A 2.1.7 3.7/4.4 Complete An Estunated Critical Condition Checklist i
Security Familiarity / Respond
- 1. JPM K/A 2.4.28 2.3/3.3 To Telephoned Bomb T1ucat A2 Clearance And Tagging /
- 1. JPM K/A 2.2.13 3.6/3.8 Identify And Replace Blown Fuse i
A.3 Protection From Radiation
- 1. JPM K/A 2.3.10 2.9/3.3 -
Exposure / Prepare For Entry Into High Radiation Area >
1000 mr/hr A.4 Emergency Plan / Emergency '
- 2. a. K/A 2.3.I 2.6/3.0 Emergency Exposures-Plan Directions
- 2. b. K/A 2.4.29 2.6/4.0 Emergency facilities NUREG-1021 Interim Rev. 8, January 1997
.. ~
.. =. -
a ES-301 Administrative Topics Outline Form-ES-301 1 Facility: Byron I & 2 Date of Examination:
Septemberl4,1998 Examination Level: SRO-l Operating Test Number:
1 Administrative Describe method of evaluation:
' Topic / Subject 3.
ONE Administrative JPht OR Description 4.
TWO Administrative Questions Al Plant Parameter Verification /
- 1. JPM K/A 2.1.7 3.7/4.4 Review A Completed Estimated Critical Condition Conduct of 0perations / Sluft
- 1. JPM K/A 2.1.3 3.0/3.4; 2.1.4 2.3/3.4 Turnover with Statling Complications A2 Clearance AndTagging/
- 1. JPM K/A 2.2.13 3.6/3.8 Identify And Replace Blown Fuse A.3 Protection From Radiation
- 1. JPM K/A 2.3.10 2.9/3.3 Exposure / Prepare For Entry Into High Radiation Area >
1000 mr/hr A.4 Emergency Plan / GSEP
- 1. JPM K/A 2.4.38 2.2/4.0 Classification And Protective Action Recommenadtions NUREG 1021 Interim Rev. 8, January 1997
e ES-301 Individual Walk-Through Test Outline Form-ES-301-2 o
Facility: Byron I & 2 Date of Examinaticn:
September 14,1998 Examination Level: SRO-1 Operating Test Number:
1 System / JPM Title / Type Codes
- Safety Planned Followup Questions:
Function K/A/G -Importance - Description 1.
NuclearInstrumentation /
Vil a.
015000K505 4.1/4.5 Determination of criticality Response to Source Range b.
015000A102 3.58.6 Evaluation for flux doubling NI failure - MODE 5 N
S L
2.
Rod Control System / Rod I
a.
001000K402 3.8/3.8 Startup Reset pushbutton response Cluster Control Exercise (N-b.
001000A302 3.78.6 Rod Insertion Limits 41)
D S
3.
RCS Pressure Control /
Ill a.
002000K105 3.2/3.4 PRT Pressure greater than RCDT pressure Depressurize the RCS per b.
000008A108 3.88.8 Effect ofleaking safety ofincrease in PRT pressure ES-1.2 Post-LOCA Cooldown and Depressurization N
A S
4.
V a.
026000A105 3.18.4 Spray Additive Tank eductor flow ManualCS Actuation b.
026000A4014.5/4.3 AUTO start conditions for CS Pump i
N A
S 5.
CVCS / Establish Excess 11 a.
004000A212 4.1/4.3 Effect of SI on excess letdown / seal return Letdown to the RCDT(N-b.
004000K104 3.48.8 Impact of excess letdown (high pressure) on RCPs 11)
M S
6.
AC Electrical / Establish VI a.
062000K403 2.88.1 ACB interlock configuration (Auto transfer setup)
Shutdown Electrical Lineup b.
062000A206 3.4*/3.9 Aligning bus with SAT to be deenergized (cross-ties)
N S
7.
IV a.
003000A303 3.28.1 #2 sealindications on RCP startup Start a Reactor Coolant b.
012000A306 3.70.7 Reactor trip on loss of RCP Pump (0030390101.1 M)
N S
L 8.
Secondary Heat Removal-IV a.
061000A104 3.9/3.9 AF Suction pressure protstion AFW / Local Emergency b.
061000K302 4.2/4.4 AF required flow Start of 1B AFW Pump (N-56 mod)
M A
P R
9.
Diesel Generator System /
VI a.
064000K105 3.4/3.9 Starting Air Compressor operation Local Abnonnal Start of a b.
064000A406 3.9/3.9 Emergency stopping of EDG D/G (N-35b)
D P
- 10. Component Cooling Water Vill a.
008000A203 3.08.2 Maximum CCW Heat Exchanger outlet Temperature System / Locally align U-l b.
008000A303 3.0/3.1 CCW system minimum flow requirements CC System for Post-LOCA Condition -Train Separation N
P R
Type Codes: (D) Direct from bank, (M)odified from bank, (N)ew, (A)lternate Path, (C)ontrol Room, (S)imulator, (P)lant, (L)ow Power (R)CA NUREG-1021 Interim Rev. 8, January 1997
~
m ES-301 Individual Walk-Through Test Outline Form-ES 3012 Facility: Byron I & 2 Date of Examination:
September 14,1998 Examination Level: RO Operating Test Number:
1 System / JPM Title / Type Codes
- Safety Planned Followup Questions:
Function K/A/G -Importance - Description 1.
- Nuclear Instrumentation / -
Vil a.
015000K505 4.1/4.5 Determination of criticality Response to Source Range b.
015000A102 3.5S.6 Evaluation for flux doubling l
NI failure - MODE 5 N
S L
2.
Rod Control System / Rod I
a.
001000K402 3.88.8 Stantup Reset pushbutton response Cluster Control Exercise (N-b.
001000A302 3.78.6 Rod Insertion Limits 41)
D S
l 3.
RCS Pressure Control /
111 a.
002000K105 3.2/3.4 PRT Pressure greater than RCDT pressure Depressurize the RCS per b.
000008A108 3.88.8 Effect of leaking safety of increase in PRT pressure ES-1.2 Post-LOCA Cooldown ami Depressurization N
A S
4.
V a.
026000A105 3.1/3.4 Spray Additive Tank eductor flow Manual CS Actuation b.
026000A4014.5/4.3 AUTO start conditions for CS Pump N
A S
5.
CVCS / Establish Excess 11 a.
004000A212 4.1/4.3 Effect of Si on excess letdown / seal return Letdown to the RCDT(N-b.
004000K104 3.4/3.8 Impact of excess letdown (high pressure) on RCPs 11)
M S
6.
AC Electrical / Establish VI a.
062000K403 2.88.1 ACB interlock configuration (Auto transfer setup)
Shutdown Electrical Lineup b.
062000A206 3.4*/3.9 Aligning bus with SAT to be deenergized (cross-ties)
N S
7.
IV a.
003000A303 3.2/3.1 #2 sealindications on RCP startup Start a Reactor Coolant b.
012000A306 3.7/3.7 Reactor trip on loss of RCP Pump (0030390101.1 M)
N S
L 8.
Secondary Heat Removal-IV a.
061000A104 3.9/3.9 AF Suction pressure protection AFW / Local Emergency.
b.
061000K302 4.2/4.4 AF required flow Start of IB AFW Pump (N-i 56 mod)
M A
P R
9.
Diesel Generator System /
VI a.
064000K105 3.4/3.9 Staiting Air Compressor operation Local Abnormal Start of a b.
064000A406 3.98.9 Emergency stopping of EDG D/G (N-35b)
D P
- 10. Component Cooling Water Vill a.
008000A203 3.0/3.2 Maximum CCW Heat Exchanger outlet Temperature System / Locally align U-l b.
008000A303 3.05.1 CCW system minimum flow requirements CC System for Post-LOCA Condition -Train Separation N
P R
Type Codes: (D) Direct from bank, (M)odified from bank. (N)ew, (A)lternate Path, (C)ontrol Room, (S)imulator, (P)lant, (L)ow Power. (R)CA NUREG-1021 Interim Rev. 8, January 1997
The JPMs are planned to be performed under the following conditions:
JPMs 1 & 7 are to be run in RCS low temperature and pressure conditions (MODE 5).
He desired conditions are approximately 188'F and 360 psig. This can be done with either RCS solid or Pzr bubble drawn with RCS loops filled. Aho, the Shutdown Banks are fully withdrawn and Source Range charmel N-31 selected as audio count rate channel.
JPM 1, SRN! N-31 failed low, requires no particular plant conditions and is therefore suited for coupling with JPM 7. The actions are performed, as directed by BOA INST-1. The evaluated steps are 1, Attachment C steps 1-3,5.
JPM 7 is normal startup of the first RCP as directed by BGP 100-1 step F.28 and directed by BOP RC-1. The steps of RC-1 should be checked through step 18 and steps marked as complete through (& to include) step 3. The evaluated actions are steps 4 through 26. NOTE that RC-1 indicates Temporary Procedure 98-0-68 exists.
JPMs 2,5 & 6 are to be run in at-power condition.
The desired conditions are nonnal for the current power level.
JPM 2. The actions of iBOS 1.3.1.2-1 steps F.1 - F.2 and F.5-F.6 are evaluated for Shutdown Bank E and Control Bank A.
JPM 5. The actions of IBOP CV-15 Step F.1 are performed with excess letdown aligned to the RCDT.
JPM 6. The actions of BGP 100-4 Step F.15 are performed to align normal loads from the UATs to the SATs.
JPMs 3 & 4 are to be run in Post-LOCA condition (about 700 gpm leak).
The desired conditions are:
1.
Break flow equal to injection flow with TWO CV pumps running in HHSI alignment.
2.
MSIVs open (either block Steamline SI, or reset if actuated, and open MSIVs) 3.
RCS pressure between 1000 psig and 1450 psig 4.
CNMT pressure is expected to be ADVERSE and should indicate that at some point 20 psig was exceeded 5.
Pzr level about 40% but less than 50% (level that is termination point for actions of JPM 3) 6.
All RCPs stopped 7.
SI reset, Phase A reset and air restored to CNMT, Phase B reset (if actuated)
JPM 3, the Pressurizer spray valves (and controllers, if required) are to be failed closed. The major steps of BEP-1 are complete through STEP 12. The steps of BEP ES-1.2 are complete through STEP 7. A cooldown should be set up as appropriate (20 F/hr to 50 F/hr). IBEP ES-1.2 steps 9 & 10 are the evaluated actions for the JPM.
JPM 4, CNMT Spray actuation failed, CS007 valves closed (NOT failed), CS019 valves closed (NOT failed), One CS pump running, if possible. The evaluated actions are those of ATTACHMENT B of BEP-0. Would like to have one CS pump l
running with associated CS007 valve closed. The other pump can be NOT nmning. This forces the RNO actions of step 1.b.
CS019 valves should be close (if this does not affect CS pump in step 1.b). One CS pump is restarted per step 1.b, if possible, and the other (or both) is started in step 2.a.
l l
l JPMs 8,9,10 In-plant JPMs JPM 8 is local emergency start of IB AFW per N-56 with modification. The initial conditions are that no feedwater flow exists to the SGs and a fire exist in such areas that control room controls and RSP controls are or may be affected. IBOA i
ELEC-5 was entered for guidance on local actions of which one is to locally start the Diesel Driven Feedwater Pump. The evaluated actions are those of BOA ELEC-5 ATTACHMENT D. At step 2.d, the engine will fail to start, requiring RNO action, select alternate battery bank.
JPM 9 is abnormal start of 2A DG per N-35b. The evaluated actions of BOA ELEC-3 ATTACHMENT D steps 1-7.
JPM 10 is local alignment of CCW for Post-LOCA alignment. The initial conditions are a LOC A on Unit I with possible failures that could affect CCW alignment. Actions provide for train separation (U-1) with B CC Pp available. The evaluated actions are those local actions of BOP CC-14, section 2. (The actions of BOP CC-8 are completed prior to this JPM.)
~ _. -............ -. - -... _.. _ -
RAW d Modes a
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9 9
S' C
BATT 112 E
36.5 BATT 111 -
0.4 INVERTER 114_
15.4 Dynan i-1 B AF PUMP W
15.1
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p g
Dynandes 2-1 INVERTER 111 12.9 A SX PUMP 10.7 Dynamics 2-2 N
1SX005 & 2SX005 E 8.8 g
e Dynandes 1-1,2-3 A AF PUMP Q 6.2 MN Dynandes 2-3 B RH PUMP E5.8 g
M B CV PUMP
] 3.4 Dynamics 1-1,1-3, 2-3 4
Dynamics 2-3 A RH PUMP E 3.2 b.
e Dyna es 2 p i;2
_ E 3.0 BDG A DG g 2.2 E
OB WW PUMP l1.9 0A WW PUMP l1.7 l1.3 08 SX MU PUMP OA SX MU PUMP l1.2 Dynamics 2-2 18 SX PUMP -l1.1
- 1.0 18 St PUMP 1A S1 PUMP - 1.0
1 I
l i
l l
CDF CONTRIBUTION BY INITIATING EVENTS l
Dynamics 2-2 y
LOSP; 18.2%
LOSS OF 125 VDC BUS; 15.6 /N i
l GENERAL TRANSIENT; 7.7%
Dynamics 1-1
\\
l i
i SMALL-BREAK LOCA; 7.4% x Dynamics Spare Jpm #3 1
SECONDARY BREAKS; 4.7%N Dynamic l-1, 2-1 i
3 LARGE-BREAK LOCA; 4.0%
l Dynamic l-3,2-3 Jpm #4, #10 h
LOSS OF SX; 9.6%
/
DUAL-UNIT LOSP; 30.7%
yn mic 2-2 OTHERS; 2.2%
2 Dynamic l-2 i
9 ES-301 Transient and Event Checklist Form ES-301-5 OPERATING TEST NO.:1 Applicant Evolution Minimum Scenario Number Type Type Number 1
2 3
4 Reactivity 1
it i/
1/
5/
Normal 1
/1
/1 it 15 RO instrument 2
ais 2i6 at 4 2i3
- f45 Component 2
47i2.7 3
3.5.6 i 2.8 2.4.7 f i.4.7 Major 1
6,8 / 6.8 5/5 57/5.7 6/6 Reactivity 1
1 1
1 5
Normal 0
As RO Instrument 1
3 2
3 2
Component 1
4.7 23.4.7 3,56 2.4.7 Major 1
6.8 5
5.7 6
SRO-l Reactivity 0
Normal 1
1 1
1 5
As SRO Instrument 1
3.5 2.6 34 2.3 component 1
2.47 23.457 2,3.5.68 1.2.47 Major 1
68 5
5.7 6
Reactivity 0
MA MA MA MA Normal 1
MA MA MA MA SRO-U instrument 1
MA MA MA MA Component 1
MA N/A MA N/A Maior 1
MA MA MA MA Instructions:
(1)
Enter the operating test number and Form ES-D-1 event numbers for each evolution type.
(2)
Reactivity manipulations must be significant as defined in Appendix D.
NOTE: Scenario Number 4 is a ' spare' scenario and is represented on ES-301-5 for ALL Operating Tests for comparison purposes only in Examination Outline submittal.
The '/" in the cells for the "R0" applicant type represents the position the applicant is expected to fill during the scenario. The events are listed for the identified position: R0 / BOP.
Author 1
Chief Examineb/Mv'M M68
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ES-301 Competencies Checklist Form ES-301-6 Operating Test: 1 Applicant #1 Applicant #2 Applicant #3 RO/SR04/S O U RO/SRO4/SRO-U R0(BOP)/SRO4/SRO-U Competencies SCENARIO SCENARIO SCENARIO 1
2 3
4 1
2 3
4 1
2 3
4 Understand and Interpret 28 27 24 14, 34, 2-s,7 3,54 12, 2.58 37 2,44 1,3, 8'7 d'57
- '8'7 Annunciators and Alarms Diagnose Events 2-8 27 28 14, 34, 2-s,7 3,s.8 12, 2.s 8 37 2,44 1,3, 8-7 84 48-7 487 and Conditions Understand Plant 18 17 18 17 14, 25,7 13, 17 12, 1,37 12, 17 u
58 s-8 u
and System Response Comply With and 14 1-7 18 17 1,3,4 2.s,7 13, 17 12, 1,37 12, 17 4-7 58 u
Use Procedures (1)
Operate Control 14 17 18 17 1,3,4 2.s 7 13, 17 12, 1,37 12, 17 l
$8
~8' Boards (2) s.8 78 Communicate and 18 17 18 17 18 25,7 13, 17 14 1,37 12, 17 58 InteractWith the Crew Demonstrate Supervisory 14 17 18 17 MA MA MA N/A MA MA MA MA Ability (3)
Comply With and 24 1,2,4 3
2 MA MA MA MA MA MA N/A N/A Use Tech. Specs. (3)
Notes:
(1) includes Technical Specification compliance for an RO.
(2) Optionalfor an SRO-U.
(3) Only applicable to SR0s.
Instructions:
Circle the applicant's license type and enter the event numbers that test the competency for each scenario in the set.
NOTE: OPERATING TEST NO.: 1. Scenario Number 4 is a ' spare" scenario and is represented on ES-301-5 for ALL Operating Tests for comparison purposes only in Examination Outline submittal. The order oflisting for candidates is SRO, RO and BOP by position.
Author:
M/
c Chief Examiner: \\VdY E. h.
w
a.
i ES-301 Transient and Event Checklist Form ES-301-5 OPERATING TEST NO.: 2 Applicant Evolution Minimum Scenario Number Type Type Number 1
2 3
4 Reactivity 1
ii il 3i Si Normal 1
i1 i12 i3 is l
RO instrument 2
3/4 34i4 its 2ia Component 2
2.8 / 2.5 sis 68 247i2.6 247i1.4.7 j
Major 1
67i6.7 7i7 7/7 6is Reac6vity 1
1 1
3 5
Nonnal 0
As RO instrument 1
3 3.4 1
2 Component 1
2.8 6
24.7 2.47 Major 1
6.7 7
7 6
SRO-l Reactivity 0
Normal 1
1 1,2 3
5 As SRO instrument 1
3.4 34 1.5 2.3 Component 1
25.8 5.6,8 2.46.7 124.7 Major 1
6 7
7 6
Reactivity 0
MA MA MA MA Normal 1
MA MA MA MA SRO-U Instrument 1
MA MA NA MA Component 1
MA MA MA MA Major 1
MA MA MA MA l
Instructions:
(1)
Enter the operating test number and Form ES-D-1 event numbers for l
each evolution type.
(2)
Reactivity manipulations must be significant as defined in Appendix D.
NOTE: Scenario Number 4 is a ' spare" scenario and is represented on ES-301-5 for ALL l
Operating Tests for comparison purposes only in Examination Outline submittal.
l The "/" in the cells for the "R0"' applicant type represents the position the applicant is expected to i
fill during the scenario. The events are listed for the identified position: RO / BOP.
mb Author-Chief Examiner:W4AA~
vw-4 NUREG-1021 2 of 26 Interim Rev. 8, January 1997
l
.~
l ES-301 Competencies Checklist Form ES-301-6 l
Operatir.g Test: 2 Applicant #1 Applicant #2 Applicant #3 RO/SRO-!!SRO U R0/SRO4/SROM R0(BOP)/SRG4/SRO-U Competencies SCENARIO SCENARIO SCENARIO i
1 2
3 4
1 2
3 4
1 2
3 4
l Understand and Interpret 24 34 1.2 14 2.3 3-7 1.2 12 2.
44 2.s-7 1.3 4~7 8'7 84 47 d 8-7 4'7 487 Annunciators and Alarms l
l Diagnose Events 24 34 1.2 14, 2.3 37 1.2.
12, 2,
48 2.57 1.3, 4-7
&7 S8 4.7 4.57 47 4.e,7 I
and Conditions Understand Plant 1-8 14 1-7 17 1-3 1-7 14 17 1.2 1.2 13 1-7 84 7
4-7 d4 57 l
and System Response i
ComplyWith and 14 14 17 1-7 13 1.3-7 14 17 1.2.
1.2 1-3 1-7 84 7
4'7 44 57 Use Procedures (1)
Operate Control 14 14 1-7 17 1-3 1,3-7 14 17 1.2.
1.2.
13, 17 S8 7
4-7 48 57 l
Boards (2)
Communicate and 18 1a 17 1-7 13 17 14, 17 1.2.
1,2, 1 -3, 17
- 8 7
4'7 44 57 InteractWith the Crew Demonstrate Supervisory 14 14 1-7 17 f#A NA t#A MA t#A t#A t#A t#A Ability (3)
Comply With and 2.4 34 1,2 2
t#A t#A N/A t#A t#A t#A N/A N/A Use Tech. Specs. (3)
Notes:
l (1) Includes Technical Specification compliance for an RO.
l (2) Optional for an SRO-U.
(3) Only applicable to SR0s.
Instructions:
Circle the applicant's license type and enter the event numbers that test the competency for each scenario in the set.
NOTE: OPERATING TEST NO.: 2. Scenario Number 4 is a ' spare' scenario and is represented on ES-301-5 for ALL l
Operating Tests for comparison purposes only in Examination Outline submittal. The orPr of listing for candidates is SRO, R0 and BOP by position.
- Author, m
u/%
M,y u, M m T ChiefExaminer:
m i
l
Simulation Facility. Byron I & 2 Scenario No.:
1 Op Test No.: 1 e
Examiners:
Operators:
SEQ B0 DDP Objectives:
In accordance with plant procedures:
1.
raise reactor power.
2.
respond to trip of Component Cooling water Pump with failure of auto-start of standby pump due to header pressure instrument failure.
3.
respond to failure of controlling pressu:izer level channel 4.
respond to degraded operation of Charging Pump.
5.
respond to failure of the steam header pressure transmitter supplying feedwater pump speed control.
5 6.
respond to a faulted steam generator with failure of ESF signal for Steamline Isolation J
7.
perform actions for a loss of heat sinit E
Initial Conditions:
1.
IC-18, Reactor Power 75%, MOL, BGP 100-3, step F.61.
- 2. ' PZR level control is selected to 459/460.
3.
IB CV Pump runmng; 1 A CV Pump in standby 4.
SG IB level LT-557 is out of service.
5.
Turnover:
- 1. ' Reactor power is 75%.
2.
Raise reactor power at 5 MW/hr after shift turnover.
3.
IB CV Pump is currently in senice following pump seal replacement testing. l A CV Pump is in standby.
4.
Narrow range level transmitter LT-557 on SG B is out of senice. All required actions lave been performed per BOA INST-2 and the AR has been initiated.
5.
The Diesel Driven Aux Feedwater Pump is OOS for replacement of a fuel injector. The pump has been OOS for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and is expected to be returned to senice by the end of the shift.
6.
I A Heater Drain Pump is out of senice for motor bearing replacement.
7.
A thunderstorm warning is in effect for Stephenson. Winnebago and Ogle counties for the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
I I
W
=
c p&
M w
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wy-
Event Malf.
Event Event No.
No.
Type
- Description FW43 Preload FW44 1 A and IB AFW Pumps fail to start RX06H, O MRF -
RX125 trip RX072 trip RXil3 trip Preload RX146 trip Narrow Range Steam Generator level LT-557 out of service RP-34, OLTT RP-35, OLTT RP40, OtlT Preload RP41, otrr Failure of MSIV automatic closure signal Preload MS01 A.100%
RO 1
N BOP Raise reactor power (5 MW/ min)
CC01B BOP 1 A CC Pump trips coincident with IB CC Pump discharge pressure switch 2
CCO2B,144 C
SRO failing as-is.
RO 3
RX13A,0,30 I
SRO Controlling pressurizer level (1LT-459) fails downscale on a 30 second ramp.
RO Running charging pump IB experiences impeller degradation of 50% over a 5 4
CV29B,50%
C SRO minute period causing reduction in flow and pressure BOP 5
RX05,0 1
SRO Steam Header Pressure Detector PT-507 fails low RO MSO9, 3.5 SRO 6
MLBH M
BOP Main Steam IIcader Break with failure of MSIV auto closure RO SRO-7 (MS01A)
C BOP 1 A Steam Generator MSIV failure to close both autonntic and manual RO SRO 8
(FW43)
M BOP 1 A MD AFW pump trip results in loss of feed.
- (N)ornal, (R)eactivity (I)nstrument, (C)omponent, (M)ajorTransient fagt{ <f f-
$: y y: 8 J t b r:e 2 wl jt) g s,w 4&m
~d"
SCENARIO #1-1 SCENARIO SUMM ARY:
Reactor power is 75% The IB AF pump is OOS due to injection pump replacement, and has been OOS for the past 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
The AF pump is expected to be returned to senice in the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. SG IB Narrow Range Steam Generator Level Transmitter LT-557 is out of service. All required actions have been performed and an AR has been initiated. 1 A Heater Drain Pump is out of service for motor bearing replacement. Power is to be increased after shift turnover at $MW/ min.
After power has been increased approximately 5%, the 1 A CC Pump will trip and the standby 1B CC Pump will NOT auto start due to failure of the associated discharge pressure instrmnent. BOA PRI-6 will be entered and the operator will be required to recognize failure of auto start of the standby pump, diagnose failed pressure instrument and start the IB CC Pump. The SRO will address Tecimical Specification, Condition B.
Tu o minutes after the actions are determined for the inoperable CC pump, the controlling channel of pressurizer level (LT-459) will fail downscale on a 30 second ramp. BOA INST-2, ATTACHMENT C will be entered to restore pressurizer level and select an operable channel. Letdown will be placed back in senice and heaters restored. Channel LT-459 will be taken OOS and Technical Specification 3.3.1 Table 3.3.1-1 Function 9 actions (Condition J) addressed.
Following Technical Specification actions, the 1B Charging Pump experiences impeller degradation (to 50%) resultirq; in reduction of charging flow and discharge pressure. Low charging line flow alann actuates and pump amps will decrease. Operator response should include placing the 1 A Charging Pump insenice and stopping the 1B Pump. The SRO addresses Technical Specifications 3.5.2 (Condition A) and 3.5.5 (Condition A) and TRM 3.1.d.
After addressing the failed Charging Pump actions, steam header pressure transmitter I'T-507 will fait downscale causing feedwater pumps to go to minimum flow. Manual control will have to be taken of the feedpump speed controller to restore nonna!
f;ed flow (FW header pressure).
A Main Steam header break occurs at the steaa, header crosstie with a failure of the Steamline isolation automatic closure signal and a failure of the l A MSIV to close. The I A AFW pump will trip on overcurrent. Feed flow is lost to the SGs. FR-H.1 will be entered and an alternate method of feedwater will be required. With i A steam generator being faulted, the actions of BEP-2 will be required following transition from FR-It 1. The scenario ends following transition to EP-1 after isolating the fauted SG in EP-2.
ERG-Based Critical Tasks k
1.
EP A:
Isolate the faulted SG before transition out of EP-2.
2.
FR-H.1 - A: Establish feedwater flow into at least one SG before RCS bleed and feed is required.
l 1
1 Simulation Facility - Byron I & 2 Scenario No.-:
2 Op Test No.: 1 Exanuners:
Operators:
SRO BO BOP Objectives:. In accordance with plant procedures:
1.
reduce reactor power (for shutdown).
2.
respond to Pressurizer Pressure Master controller failure with a failure of the PORV to close.
3.
respond to a loss of instrument air to the containment.
4.
respond to a steam generator tube leak.
5.
respond to a loss of Control Room dP.
)
6.
perform emergency actions for a SGTR with out Pressurizer Pressure control.
i Initial Conditions:
1, IC-21, Reactor Power 100%, Steady state BOL, BGP 100-3, step F.61.
2.
PZR pressure contro.' selected to 455/456.
3.
Pressurizer pressure et annel irr-458 is OOS with bistables tripped.
l 4.
PORV PCV-456 cowol is in manual control (CLOSE) and Block Valve RY-8000B is closed and deenergized.
1 5.
I A HDP OOS; 250' Meteorological Tower OOS.
Turnover:
1.
Reactor Power is 100%, Steady state power at BOL. BGP 100-3 is in effect.
, 2.
Unit 2 is at 100% power.
3.
The block valve (RY-8000B) for PCV-456 is closed and deenergized. A leak had developed on PCV-456.
When the block valve was shut it tripped after the closed indication was observed. Electrical Maintenance is investigating. Valve has been OOS for 62 hours7.175926e-4 days <br />0.0172 hours <br />1.025132e-4 weeks <br />2.3591e-5 months <br />.
4.
Pressurizer pressure instrument IPT-458 is failed. All actions of BOA INST-2, ATTACRMENT B have been completed. PT-14 for channel has been initiated. Currently no report of expected time for release to senice.
5.
Heater Drain Pump 1 A is out of senice to meggar motor leads.
6.
The 250 foot Meteorological Tower is out of senice.
7.
High wind warnings have been issued for Stephenson, Winnebago and Ogle counties for the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Event Malf.
Event Event No.
No.
Type
BOP FWI valves fail to close (FWI signal failure)
OR ZA00UREM002P2,0 OR ZA00trREM012PI, O Preload OR ZAOOUREM012P2,0 250' Met Tower OOS RX22B,2500 RF RXO44 trip RF RXO45 trip RF RXO46 trip RF RX025 trip Preload RF RX141 trip Pressurizer pressure instrument 458 is OOS bistables tripped RF Block Valve MOV-RC8000B (PCV 456) was closed to cycle the valve. When Preload En06so Open the operators attempted to open valve, the motor operator tripped on overload.
SRO R
RO 1
N BOP Reactor shutdown due to inoperable PORV j
I RO Pressurizer Pressure Master Controller failure to maximum output / PORV 455 A RX t 5,2500 2
Till1 A. 25%
C SRO fails to close requiring the block valve RY8000A to be closed.
IA03,5000 C
SRO Loss of instrument air to the containment.
TH03D, 25 C
BOP Steam Generator ID Tube Leak - 25 gpm.
RO TH03D,500 BOP Steam Generator ID Tube Rupture - Increases requiring a reactor trip and St.
5 M
SRO FWI signal fails requiring manual operation of FW components.
BOP ZAOOPDIVC038 6
0 I
SRO Main Control Room differential pressure failure with decreasing pressure.
RO RF ggp Power is lost to the power supply for the Block Valve MOV-RC8000A (PORV 7
ED058C Open C
SRO 455 A) while valve is closed (control switch taken to OPEN).
- (N)ormal, (R)eactivity (I)nstrument, (C)omponent, (M)ajor Transient
- - _ - - - - - - -.. _.-.- _.-.. _ _. _ ~ - ---. _ - -.
SCENARIO 01-2 SCENARIO
SUMMARY
The scenario will begin at 100% power. Pressure transmitter lirr-458 is out of service. An AR has been initiated and all required actions have been taken per BOA INST-2. The block valve for PORV 456 is closed and deenergized due to thermal overloads tripping on the valve operator. The 250' elevation Meteorological Tower Instrumentation has failed and action of TRM 3.3.c is being tracked.
Operations Management will call and report that the PORV and block valve cannot be repaired within required time (16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> from now). This will require a Unit shutdown be commenced immediately.
Following the initial down power (>5%), a failure of the Pressurizer Pressure Controller will cause PORV PCV-455 A and the spray valves to open. PORV 455A fails open since the PORV enabling pressure channel Irr-458 is failed HIGH. Failure of PORV
- 455A to close will require closure ofits block valve MOV-RY8000A. The spray valves will close after manual control is taken of the pressurizer pressure controller. SRO will address Technical Specification 3.4.11 (Condition E) and TRM 3.3.k and 3.4.d.
)
After pressurizer pressure is returned to the normal band using manual control, instrument air will be lost to the containment.
Actions will be performed in accordance with the BOA SEC-4 for loss ofinstmment air to the containment. The loss ofinstrument air l
will prevent operation of the pressunzer normal spray valves or the auxiliary spray valve.
A tube leak will occur in SG ID. The crew will have to diagnose the tube leak and perform actions per the BOA PRI l and BOA SEC-8. This will include determination ofleak size (~ 25 gpm) and SRO will address Technical Specification 3.4.13 (Condition A). A trip of the unit should be initiated due to complications of loss of pressure control with a leak.
The reactor trip will cause Steam Generator Tube Rupture to increase requiring a SI. EOP actions of BEP-0 and BEP 3 will
)
be performed. In BEP-0, the feedwater system will not isolate due to a failure of the feedwater isolation signal requiring manual j"
action to isolate feedwater, and at step 21, the operator will recognize failure to maintain positive Control Room pressure and will perform actions of BOP VC-14. PORV Block valve (RY8000A) electrical power feed will trip if the valve is taken to open. Without 1
- pressunzer PORVs, pressurizer normal spray valves and Aux. Spray valves, pressurizer pressure control is lost. This will require that BCA-3.3, "SGTR Without Pressurizer Pressure Control", be performed during actions for the SGTR. The scenario ends with the establistunent of RCS cooldown either in BCA-3 "I or BCA-3.1, as appropriate.
- ERG Based Critical Tasks:
(
l.
EP A:
Isolate feedwater faw into and steam flow from the ruptured SG before a transition to ECA-3.1 occurs.
2.
EP B:
Establish /maint 6n and RCS temperature so that transition from EP-3 does not occur because RCS temperature is in either of the following cond'.tions.
Too high to maintain minimum required subcooling OR Below the RCS temperature that causes an extreme (red-path) or a severe (orange path) challenge to the
+
suberiticality and/or integrity CSF.
3.
ECA-3.3 A Tenninate Si before a water release occurs through the SG PORV or SG Safeties.
'+.
Simulation Facility Byron I & 2.
Scenario No.:
1
. Op Test No.:- 1 Examiners:
Operators:
In accordance with plant procedures:
- 1. - increase reactor power.
2.
respond to a trip of one feed pump. -
3.
respond to a failed primary RTD 4.
respond to failure of digital rod position indication channel.
5.
respond to failed steam flow channel.
6.
respond to TWO dropped control rods with failure of the reactor to trip.
7.
respond to failure of running Charging Pump.
- 8.
peiform actions for a large break LOCA where containment phase B actuation fails to operate.
Initial Conditions:
1.
IC-190, Reactor Power 67%, MOL, BGP 100-3, step F.61.
2.
IB CV Pump running; 1 A CV Pump in standby.
3.
1 A Containment Spray pump OOS.
4.
I A Motor driven Feedwater Pump OOS 5.
250' Meteorological Tower OOS.
Turnover:
1.
Reactor power is 67% with power increase to continue following turnover at SMW/hr.
2.
Unit 2 is operating at 100% power.
3.
1B CV Pump is currently in senice following pump seal replacement testing. I A CV Pump is in standby.
4.
I A Containment Spray pump is 00S due to higinibrations during testing. An AR has been initiated and the pump is expected to be release to senice in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- 5.
1 A Feedwater Pump is OOS with electrical supply cleared due to electrical wiring problem. Pump will NOT be ready to return to service this week.
6.
The 250 foot Meteorological Tower is out of senice, j
7.
Thunderstorm warning is in effect for the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
s
~.
_ ~.
.. ~
6
~
- Event
. M alf.
Event Event Fo.
No..
Type
- Description RO Preload RP01 C
SRO Reactor protection system failure of automatic trip Preload CS01 A -
I A Containment Spray pump OOS.
RPISA MRF RO Preload RP83 open C
SRO Failure of Auto Start for the l A CV ptunp.
OR EDIICS01PB, Failure of CS Pump to start and Phase B to actuate on High-High Containment Preload TRIP Pressure SRO R
RO I
N BOP Raise reactor power (5 MW/ min)
BOP 1
3 FWO2A C
SRO 1B (ttubine driven) Main FW Pump trips.
RX18A,630 I
RO Tcold RTD fails high resulting in higher Ta, input. Coincident with rod 3
RD13AK10 C
SRO motion, a DATA A failure occurs for rod K10.
RX03B,4.8 BOP 4
MLBII I
SRO Steam Flow Transmitter FT-513 (input to controlling SGWLC) fails upscale RD02K10 RO.
RD02F06 BOP 5
(RP01)
M SRO TWO dropped control rod - Failure of the reactor to trip on negative rate trip.
RO 6
CV01B C
SRO Running Cimrging Pump trip (to occur on reactor trip).
- TH06, RO 50,000 to SRO Large break LOCA inside contaisunent. (Put on a ramp for contairunent pressure 7
400,000 M
BOP to rise above HI H1 setpoint.)
(OR)
C SRO Failure of Phase B and containment spray system to actuate automatically.
0(N)ormal, (R)cactivity (I)nstrument, (C)omponent, (M)ajor Transient i
I I
SCENARIO 01-3 SCENARIO
SUMMARY
The scenario begins with power at 67% with the l A Containment Spray pump out of senice due to high vibration during the
- hst run. The 250' elevation Meteorological Tower instrumentation has failed and action of TRM 3.3.c is being tracked. Power is to be raised at 5 MW/hr.
After reactor power is increased at least 5%, the IB Main Feedwater will trip. The crew should respond per BOA SEC-1 and reduce power to within the capacity of one feedpump (540 to 600 MWe or ~ 60% power).
Prior to power stabilization following the FW Pump trip, a loop Tcold RTD will fail high resulting in increased auctioneered high Tave. 'This will result in demand inward inotion of the control rods. The crew will perform actions of BOA ROD-1 due to the rod motion, and will be directed to BOAINST-2, ATTACHMENT A. Coincident with the RTD failure, a DATA A failure (DRPI l
coil fails open) will occur on CBD rod K10. Alarm Response procedures and/or BOA ROD 3 will be entered. SRO will revkw Technical Specification 3.3.1, Table 3.3.1-1, Function 6 & 7 (Condition D) and 3.1.7 (Condition A).
i Following completion of actions for taking the failed RTD instrument out of senice, the 1 A SG selected flow channel l-instrument fails high. This results in indication ofincreased steam flow and initial opening of the l A SG Feed Reg Valve to attempt l
to match feed flow to steam flow. An equilibrium level should be reached if manual control is not taken expediently. The operator l
l will perfonn the actions of BOA INST-2, Attachment H.
Following stabilization of SG levels after control is returned to auto, TWO rods trip into the core. The reactor fails to l
automatically trip on the PR Negative Rate trip and the operator will have to manually trip the reactor (also required on drop of more -
l than one rod). The running Charging Pump trips coincident with the reactor trip, and the operator must start the 1 A CV Pump to j
provide charging flow. After perfonning the first FOUR steps of BEP-0, transition will be made to BEP ES-0.1.
l
. After Step 5 is perfonned in BEP ES-0.I, a large break LOCA will occur requiring BEP-0 to be re-entered and actions of BEP-0 and i
BEP-1 to be perfonned. Phase B will fail to occur and containment spray will fail to initiate (i.e., the CS pumps will NOT start) j requiring the crew to manually actuate CS and Phase B Isolation. The scenario ends after transition to cold leg recirculation.
l l
ERG-Based Cdtical Tasks:
i 1.
EP A: Manually trip the reactor from the Control Room before transition to ES-0.1.
i l
2.
EP E: Manually actuate at least the nummum required complement of containment cooling equipment before an extreme I
(red path) challenge develops to the containment CSF.
l l
3.
EP 1: htmually start I A CV pump before transition out of EP-0.
l l
r
}
4
V Simulation Facility Byron I & 2 Scenario No.:
1 Op Test No.: 2 Examiners:
Operators:
Sgg BQ BOP Objectives:
In accordance with plant procedur::s:
1.
lower reactor power.
2.
respond to 120 VAC instrument bus loss of power.
3.
respond to Tref programmer failure low.
4.
respond to a steam generator atmospheric dump valve failing to mid position.
5.
respond to main turbine bearing high sibration.
6.
perform actions for depressurization of all steam generators compounded with failure of automatic feedwater isolation to one SG.
7.
respond to trip of Charging /HHSI pump with failure of the other pump to start on the Si sequence.
8.
perform actions for isolation of faulted SGs Initial Conditions:
1.
IC-21, Reactor power is 100% BOL in BGP 100-3, step F.61.
2.
IB CV Pump rumung; 1 A CV Pump in standby.
Tumovet.
Reactor power is 100% Steady state power at BOL. BGP 100-3 is in effect. Power is to be reduced to allow performance of TV/GV surveillance.
- 2. - Unit 2 is operating at 100% power.
)
3.
IB CV Pump is currently in service checking for post-test leakage. I A CV Pump is in standby.
4.
Weather conditions are high wind warnings for Stephenson, Winnebago and Ogle counties for the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
5.
Night orders indicate that security has received threats of sabotage against Comed nuclear plants, i
i
r Event Malf.
- Event Event No.
No.
Type
- Description RF FW-Iso.
Failure of feedwater Reg valve to close on FWI for SG IC.
RF FW 151 Preload REMOVED C
BOP Failure of Feedwater Isolation Valve to close for SG IC.
RP15A RF RP83 Preload OPEN C
RO I A CV Pump fails to sequence on SI signal. Manual start available.
MS01 A.100%
RO l
MSolB,100%
MS01C,100%
)
Preload Msolo,100%
RO l
R BOP 1
N SRO Lower reactor power.
ED11A C
SRO 120 VAC Instmment Bus 111 inverter failure.
RO 3
RX12,557 i
SRO Tref programmer output fails low.
BOP Steam Pressure Instrument IPT-MSO41 fails high failing Atmospheric Dump 4
Ms04 A, 50%
1 SRO valve to mid position on SG 1 A.
TUOll,15, BOP 5-600 see C
SRO Turbine bearing #9 increasing vibration over 5 minutes with turbine trip required.
M SRO Turbine Trip due to High Vibrations Main steam rupture downstream of the MSIV on 1 A SG with a failure of the MSIVs to close. The crew will be able to locally close valves on two loops.
NOTE: MS01 for two SGs (IB & IC) should be removed sequentially fHw g dispatch of opemon to My close valves. Operator shouM close MS08A. 4 RO f rst two valves in order if sequence provided by crew.
]
MLmi BOP 7
(RF FW.150)
M SRO SG FW fails to isolate cv01B On the Si signal, the running Charging Pump trips. NO CV pump starts on Si j
8 (RP15A)
C RO sequence.
- (N)ormal, (R)eactivity (1)nstrument, (C)omponent, (M)ajor Transient
.._.~._ ___ __._ _ __._ _ _ _
SCENARIO 02-1 SCENARIO
SUMMARY
The scenario will begin at approximately 100% power. After shift turnover the crew will lower load in order to perform main turbine Throttle Valve / Governor Valve surveillance testing.
Following at least a 5% reactor power change, the inverter to Vital Instrument Bus 11I fails causing a loss of power to the bus. The actions of BOA ELEC-2, ATTACHMENT A will be performed and bus re-energized from CVT. The SRO will address Technical Specification 3.8.7 (Condition A) and 3.8.9 (Condition B). (See next page for major affected components.)
When power is restored to Bus 111, the output of the Tref programmer will fail to the low value of 557'F. This will affect rod control causing rods to step in on Tave-Tref error. Actions of BOA ROD 1 may be performed due to the rod motion. The crew will restore Tave to actual Tref (BCB 1 Figure 33 or equivalent may be used).
After power is stabilized following the Tref failure, steam pressure transnntter IYT-041 (or PORV controller) on SG 1 A will
. fail high causing MS PORV MS018A (atmospheric relief valve) to open. When manual control is taken of the atmospheric dump valve it will only partially close (50% open). The crew will have to locally isolate the atmospheric dump valve (MS019C). SRO will address Technical Specification 3.7.4 (Condition A).
Following isolation of the SG IC MS PORV, a vibration for the turbine #9 bearing will increase over a period of time. The crew will perform actions of BOA TG 1. The reactor will be manually tripped and the turbine will be tripped as vibration continues to rise toward the trip setpoint. On the trip the crew will enter BEP-0 e
When the turbine is tripped a steam mpture will occur on the steam header. All MSIVs will fail to close in response to the l
stesm leak and efforts to close the MSIVs from the control room will be unsuccessful. Additionally the feedwater isolation signal for the steam generator will fail. On the Si signal, the running Charging /HHSI Pump will trip and the 1 A CV ptunp will not auto start when the SI sequencing occurs. The crew will manually start the CV pump. Transition is made ECA-2.1 due to all SGs depressurizing. After step 8 of ECA 2,1 the local efforts to close two of the MSIVs (preferably IB & IC SGs) will be successful. The reports of success will be staggered. The scenario ends following innsition to BEP-2, identification and attempted isolation of remaining faulted SGs and decision to transition to BEP-1.
ERG Based Critical Tasks 1.
EP-0 -I: Establish flow from at least one high-head ECCS pump before transition out of E-0.
2.
EP P: Manually actuate main steandine isolation before a severe (orange-path) challenge develops to either the subcriticality or the integrity CSF or before transition to ECA-2.1, whichever happens first.
- 3., ECA-2.1 - A: Control the AFW flow rate to not less than 25 gpm per SG in order to minimize the RCS cooldown rate before a severe (orange-path) challenge develops to integrity CSF.
+
4.
EP A: Isolate the faulted SG before transition out of EP-2.
J e
e e.
_,,.m.
7.-.. - -,.-,- -...--
l Simulation Facility Byron I & 2 -
Scenario No.:
2 Op Test No.: a Examiners:-
Operators:
't BD BOP Objectives:
In accordance with plant procedures:
1.
reduce reactor power.
2.
transfer to Feed Reg Bypass valves.
3.
respond to a pressurizer pressure instmment failing high.
4.
recognize and perform required action for P 7 failing to perform as required.
5.
respond to a senice water pump trip..
6.
respond to 4160V ESF bus lockout.
7.
perfonn required actions for a loss of off-site power including restoring power to one ESP bus.
8.
Respond to a failure of MSIV's to close from a Main Steam Line Isolation Signal Initial Conditions:
1.
IC-191, Reactor power 21%
2.
BGP 100-4 is complete through step F.16.
3.
Pzr pressure control is selected to 455/456 4.
IB SX Pump is running for Unit due to 2B SX pump OOS.
Turnover:
1.
Reactor power is 17% during a reactor startup.
l 2.
BGP 100-4 is complete through step F.16, 3.
Continue unit shutdown and transfer feedwater control to the Feed Reg Bypass Wives.
4.
The IB Diesci Generator is inoperable due to failure to meet time requirements for rated voltage and speed.
Maintenance is investigating governor for problem. All required actions have been completed on prior shift.
Next surveillance (SR 3.8.1.1) due in four hours.
5.
The IB SX pump is running. The 2B SX Pump is out of service due to seal leak, and so the 2A SX Pump is in senice.
6.
Unit 2 is shutdown.
4 e
A
9 Event Malf.
Event Event No.
No.
Type
- Description Preload RPl7A & B 1
P-7 fails as is.
IOR ZDIMSil Normal ZDIMS2 BOP Preload Normal C
SRO Failure of Main Steam line Actuation Signal.
RO R
BOP 1
N SRO Load decrease from 21%
BOP Transfer the Feedwater Control from the Main Feed Reg Valves to the Feed Reg 2
N SRO Bypass Valves.
RO 3
RX21 A,2500 1
SRO Pressurizer pressure instrument IFI'-455 fails high.
I RO RX10A,800 I
BOP PT-505 turbine first stage pressure fails high in conjunction with P-7 at ~10%
4 (11Pl7A & B)
SRO power. At-power Trips fail to be clear when power is lowered below 10%
SRO 5
SWolB C
BOP Senice water pump IB trips.
ED07A C
RO SRO Loss of off-site power. Fault on Bus 6. Reactor trip required due to loss of SX.
7 EDISD M
BOP Power will be available by electrical crosstic to Unit 2, Bus 242 IOR ZDIMSit Ncirmal ZDIMS2 BOP 1
8 Nortnal C
SRO Failure of Main Steam line Actuation Signal.
O(N)ormal, (R)eactivity (1)nstniment, (C)omponent, (M)ajor Transient j
SCENARIO #2-2 SCENARIO
SUMMARY
The scenario will begin at approximately 21% power with load decrease to shutdown conditions to continue. The IB DG is inoperable with maintenance investigating.
' During load decrease the crew will tmnsfer the Feedwater Control from the Main Feed Reg Valves to the Feed Reg Bypass Valves.
Following transfer to Bypass FRVs, pressurizer pressure instmment IIrr-455 will fail upscale causing both spray valves to open. Actions are taken per BOA INST-2, ATTACHMENT B. The crew will select an operable Pzr pressure channel for control and then attempt to close the spray valves. SRO will address Technical Specifications 3.3.1 Tabic 3.3.1-1 Functions 6,8 (Conditions D &
J),3.3.2 Table 3.3.2-1 Function 1.d (Condition K).
Just prior to reaching 10% reactor power or turbine power, one channel ofinput to P-13 First Stage Pressure FT-505 fails high. This provides an input to P-7. Action s are taken per BOA INST-2, ATTACHMENT D. When power is reduced to < 10%, the IM will be required to defeat the P-13 input to P-7, in order to block (bypass) the "at-power" reactor trips. SRO will address TRM
. 3.3.z Table 3.3.z-l Functions b & e (Condition B).
After the crew determines the required course of action for PT-505 failure, Senice Water Pmnp IB trips. The crew will start the i A SX pump as directed by BOA PRI-7 and direct investigation of the tripped pump. SRO will address Technical Specification 3.7.8 (Condition A).
After the Technical Specifications are addressed for the senice water pump, a lockout on ESF 141 will result in the loss of I division of electrical power. The crew will have to start equipment on other trains w here required. Loss of this bus will contribute to.
the loss of all ac when off-site power is lost. 1 A DG will have to be stopped since its output breaker will not close to the bus and no SX pumps will be available on the unit. SX will be crosstied to Unit 2 as directed by BOA PRI-7, ATTACHMENT A. A problem with Unit 2 cross connect valve (s) will delay completion until after initiation of the following event.
A loss of offsite power occurs when Bus section 6 fault recurs. The fault trips infeeds to SATs and results in all 4KV AC buses deenergized. DG 1 A output breaker will NOT close to bus 141 due to lockout and bus 142 is deenergized. With the loss of all AC power, BCA-0.0 is irnplemented. A failure of Main Steam line Actuation Signal will require closing the individual MSIV from their control switches. Power to bus 142 will be restored by crosstie to Unit 2 bus 242. Scenario ends with actions with transition to recovery procedure at step 54 of BCA-0.0.
ERG Based Critical Tasks 1.
EP 0 - C:
Energize at least one ac emergency bus before transition out of EP-0 unless transition is to ECA-0.0, then energize bus before placing safeguards equipment in FTL.
2.
ECA-0.0 - H: isolate RCP seal injection before a charging pump stans or is staned.
r.-
Simulation Facility Bymn I & 2 Scenaris No.:
2 Op Test No.: 2 Exammers:
Operators:
SRO EO D9P Objectives: ' In accordance with plant procedures:
1.
respond to failure of a Pressurizer Level instrument 2.
respond to an RCS leak.
3.
reduce reactor power 4.
respond to failure of a CV valve diverting letdown from VCT.
5.
respond to a failure of the concenser hotwell level controller
- 6. ' respond to trip of a Condensate / Condensate Booster Pump set.
7.
perform actions forlarge break LOCA 8.
respond to trip and failure to start for LHSI pumps.
Initial Conditions:
1.
IC-16,53% power MOL, BGP 100-3, Step F.61.
2.
PZR level centrolis set to 459/460.
3.
Remove 1 A AF pump from senice.
4.
Remove ! A Charging Pump from senice.
5.
Remove 1 A Motor Driven feedpump from senice.
Turnover: -
1.
Current power is $3%, MOL. Power will be raised following final verification of calorimetric data.
- 2. ' Unit 2 is shutdown for refueling. -
1 3.
A high wind warning has been issued for Stephenson, Winnebago at.d Ogle counties for the next six hours.
4.
I A AF Pump is OOS due to grounds. An AR has been initiated. The expected time OOS is at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Tagging clearance has been hung but has not been verified.
5.
I A Charging Pump is OOS for preventative maintenance. The pump is expected to be released to senice in the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.-
6.
Motor Driven feedpump OOS for bearing replacement. The pump will be OOS for the entire shift.
2 i
__..__.___m._m Event Malf.
Ev:nt Event No.
No.
Type
Preload FWOI Motor Driven Feedwater pump OOS.
Preload RH01A C
RO Trip of I A RH Pump RP15F Preload RF RPS$ open C
RO Failure of 1B RH Pump to auM start on SI sequence i
RF Fwo49,100 i
FWO50,100 I
Preload Fwo32, o C
BOP RO Pressurizer level channel iLT-459 fails downscale causing letdown isolation.
I RX13A, O I
SRO Control will be transferred to another channel and letdown restored.
2 TH06D, 20 C
SRO RCS leak ID RCS Cold Leg loop leak - 15 gpm.
SRO R
RO 3
N BOP Reduce power due to RCS leakage.
ICVI12A VCT Divert valve fails to DIVERT position. NOTE: Valve may bc RO returned to VCT position by failing air locally. If crew directs so, Sim Operator 4
CV4,100 -
C SRO will change fail position to 0.
FW37,48" FW38, 48" BOP Hotwell level controllers (LT-CD037, LT-CD038) fail high causing actual 0101cDo42 5
,4s 1
SRO hotwell level to decrease.
3 BOP 6
FW22C C
- TH06D, RO I
540000 BOP 7
(RPISA)
M SRO Large break LOCA with failure of RHR pumps to auto start
- i f
- (N)ormal, (R)eactivity (I)nstrument, (C)omponent, (M)ajor Transient
SCENARIO 02-3 SCENARIO
SUMMARY
The scenario will begin at 53% power with I A AFW Pump,1 A Charging Pump, and the Motor Driven Feedwater Pump out of service. Preparations may be made for load increase when informed of completion of calorimetric data verification Pressurizer level channel ILT-459 failure low will cause letdown to isolate, heaters to shutoff and associated alanns.
Charging flow rate will increase causing an increase in actual PZR level. The operators will have to select an operable level instrument then restore letdown and heateis. SRO will address Technical Specification 3.3.1 Table 3.3.1-1 Function 9 (Condition J).
A small RCS leak (15 gpm) will occur on RCS D loop requiring the crew to detennine the leak size, take action to maintain pressurizer level. BOA PRl-1 will be entered and SRO will address Tecimical Specification 3.4.13 (Condition A). The crew will initiate a unit shutdown due to the leak. (Operations Management will provide direction for shutdown if necessary.)
During the power reduction and following at least 5% reactor power change, the VCT letdown divert valve ICVI 12A will fail to the DIVERT position directing letdown flow to the Hold Up Tank. This will cause a reduction in actual VCT level and automatic makeup on low level. An operator may be dispatched to locally check the valve and will report failure of solenoid to deenergize. The crew may direct isolation of air to the valve that will cause it to fail to the VCr position.
Following actions to maintain VCT level, the normal hotwell level controllers (ILC-CD037) ILC-CD038 fails high. This results in actual hotwell level decrease and actuation of several alarms associated with opening of ICD 141, Emergency Overflow Valve, and lowering hotwell level. Isolation of ICD 141 by closing manual isolation valve will be required or swap to alternate channel.
Aner hotwell ecnditions are restored, a Condensate / Condensate Booster Pump trips. BOA SEC-1, ATTACHMENT B is entered and actions performed. The operator should start the standby CD/CB pump and check feedwater flow conditions stable.
Inunediately following completion of actions for tripped CD/CB Pumps, a large break LOCA occurs resulting in Si actuation. BEP-0 actions will be performed. The l A RH Pump will trip when it attempts to start, and the 1B RH Pump will fail to automatically stad on the SI sequencing. The operator will manually stan the IB RH Pump.
The crew will transition to BEP-1 following diagnosis in BEP-0. When the RWST level falls below 46%, the crew will transition to BEP ES-1.3, and align the "B" train for recirculation mode. The scenario will end following alignment of Containment Spray system for recirculation.
ERG-Based Critical Tasks 1.
EP H:
Manually start at least one low-head ECCS pump before transition out of EP-0.
2.
ES-1.3 - A: Transfer to cold leg recirculation and establish ECCS recirculation flow that at least meets the assumptions of the LOCA analysis (one train).
. _ _._ _=--.. _ - _.
6 Simulation Facility Byron I & 2 Scenario No.:
Spare Op Test No.: 1/2
~
Exanuners:
Operators:
SR,Q BQ BOP Objectives:
In accordance with plant procedures:
1.
respond to FW Pump trip.
2.
respond to the failure of a Narrow Range RTD.
3.
respond to a dropped / misaligned rod.
4.
respond to failure of a containment pressure channel.
5.
respond to RCP seal failures.
6.
reduce reactor power.
7.
respond to RCS LOCA with failure of selected ECCS valves.
Initial Conditions:
1.
IC-18, Reactor Power 75%, MOL, BGP 100-3, Step F.61.
Turnover:
1.
Reactor power is 75% Power is to be increased later in the shin at the direction of the BPO.
I f
-._ - ~ -.
~ - - _ _ -...
Event
- Malf, Event Event No.
No.
Type
- Description l
RF RWST suction valves, VCT outlet valves, liHSI injection header discharge RP-49, Oltr -
valves, and normal charging MOVs fail to reposition on the Si actuation.
Preload RP.75, otfr (CVI12B, C, D & E; S18801 A & B; CV8105 & CV8106)
Preload FWOIA 1 A Motor driven Main FW Pump trips on start.
BOP l
1 FWO2B C
SRO IC Turbine driven Main FW Pump trips.
RD031108 RxtsB,630, C
RO Primary Cold Leg Narrow Range RTD fails high (30 sec ramp), when rods begin l
2 30 ue I
SRO moving in a rod in bank D will ratchet in resulting in rod position deviation.
l BOP l-3 CH08C,60 I
SRO Containment pressure channel FT-CS936 fails high.
RO l-BOP l
4 CV28A C
SRO l
l R
RO 5
N BOP Power reduction to remove RCP from senice.
A faHum msults in LOCA conMon da D 1 A seals.
RO 6
CV27A,300 M
BOP NOTE: Simulator operator will enter LOCA on I minute ramp RO SRO 7
(RF)
C BOP Charging system valves fail to automatically transfer upon SI initiation.
- (N)ormal, (R)eactivity (1)nstrument, -
(C)omponent, (M)ajor Transient I
r l
L I
j SCENARIO CSpare SCENARIO SUMM ARY:
Reactor power is 75% After the crew assumes the shift a main feed pump will trip. If the motor-driven feed pump start is attempted, the pump will fail to statt. The crew should respond per BOA SEC-1 and reduce power to within the capacity of one feedpump (540 to 600 MWe or ~ 60% power).
After power is reduced to within the capability of the feedwater system and prior to stabilization, a Narrow Range Cold Leg RTD will fail high causing Tave to increase. This will cause control rods to move inward. When movement is initiated a control rod H8 will ratchet in due to movable gripper failure. Initial actions are taken per BOA INST-2 ATTACIBIENT A for the RTD failure and the Technical Specification 3.3.1 Table 3.3.1-1 Functions 6 & 7 (Conditions D) will be addressed. Actions of BOA ROD-3 are performed for the misaligned rod and the appropriate Technical Specification 3.1.4 (Condition A & B) and 3.1.6, if applicable, are addressed.
After conditions are stabilized from the control rod mispositioning and actions are taken for the failed temperature instrument, a containment pressure channel will fail high. The crew will perform actions of BOA INST-2, ATTACHMENT J. The SRO will address Tecimical Specification 3.3.2, Table 3.3.2-1, Functionsl.c,2.c,3.b.(3),4.c (Conditions F & G).
Following completion of actions for the failed containment pressure channel, #2 seal failure for RCP 1 A will occur. Actions are addressed in BOA RCP-1. Power reduction will be initiated (if required, at the direction of Operations Management). During the power reduction and after power is lowered at least 5%, a failure of Control Rod Drive circuit will prevent rod movement in automatic or manual. The crew will evaluate conditions and should determine a reactor trip is required due to faihire of reactivity control system with a reactor shutdown required.
After power is lowered at least 5%, the #1 seal on RCP 1 A fails resulting in a LOCA through the RCP seals. The actions of BEP-0 and BEP-1 will be performed. When a SI is required, the charging pumps will fail to automatically transfer to the RWST and the discharge valves in the IRISI injection line will fail to open, requiring the operators to manually open the valves and close valves for nonnal charging path. Actions will be taken for the small break LOCA that occurs on the RCP. Scenario ends with initiating of RCS cooldown, if RCS cooldown and depressmization is required.
ERG Hased Critical Tasks 1.
EP 1:
Establish flow from at least one high-head ECCS pump before transition out of EP-0.
2.
EP C:
Trip all RCPs so that CET temperatures do not become superheated when forced circulation in the RCS stops.
l l
l
ES-401 PWR SRO Examination Outline Form ES-401-3 l
Facility: Byron 1 & 2 Date of Exam:
September 14,1998 Exam Level: SRO K/A Category Points Tier Group Point K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G
Total 1.
1 4
2 4
5 7
2 24
~
~
't esgny
~"
Emergency &
2 2
1 3
5 3
gj ;gh 2
16 g
- g;,g; Abnormal Plant l7 g
gg hg. gj 1
1 h!
Evolutions 3
1 3
h Tier Totals 6
3 8
11 11 4
43 1
2 1
1 1
2 1
3 2
2 4
19 2.
2 1
1 2
3 1
1 2
2 1
1 2
17 p
Systems 3
1 1
1 1
4 Tier Totals 4
2 4
4 3
1 )3 6
4 3
6 40
- 3. Generic Knowledge and Abilities Cat 1 Cat 2 Cat 3 Cat 4 17 6
6 1
4 v
Note:
Attempt to distribute topics among all K/A categories; select at least one topic from every K/A categosy within each tier.
Actual point totals must match those specified in the table.
Select topics from many systems; avoid selecting more than two or three K/A topics from a given system unless they relate to plant-specific priorities.
Systems / evolutions within each group are identified on the associated outline.
The shaded areas are not applicable to the category / tier.
- b ftCM MX jWf 0 I
/SSI
- #'L "
X /yn p
4 ck/
h Nf W NUREG-1021 23 of 39 Interim Rev. 8, January 1997
r
-- - ' - - ~ - -
Facility: Byron Exam Date:
9/14/98 Examination Level: SRO Section Title Generic Knowledge and Abilities SRO Group 1
System / Evolution K/A SRO KA Statement Level Question Topic Conduct of Operations 2.1.1 3.8 Knowledge of conduct of operations requirements. B Evaluation of requirement for" active" license 2.1.1 3.8 Knowledge of conduct of operations requirements. B Direction of NLO personnel 2.1.2 4.0 Knowledge of operator responsioilities during all B Operating Daily Orders modes of plant operation.
2.1.13 2.9 Knowledge of facility requirements for controlling S US responsibility on CNMT entry vital / controlled access.
2.1.14 3.3 Knowledge of system status criteria which require B MOV Stroke Time actions the notification of plant personnel.
2.1.23 4.0 Ability to perform specific system and integrated B Proceudre required usage plant procedures during all modes of plant operation.
Equipment Control 2.2.12 3.4 Knowledge of surveillance procedures.
S Equipment operability during surveillance 2.2.13 3.8 Knowledge of tagging and clearance procedures.
B MSIV tagout to prevent opening 2.2.22 4.1 Knowledge oflimiting conditions for operations and S Technical Specification 3.0.3 application safety limits.
2.2.23 3.8 Ability to track limiting conditions for operations.
S Timing for Tech Spec required Shutdown 2.2.26 3.7 Knowledge of refueling administrative requirements.B RCS te';al discrepancy during refueling 2.2.32 3.3 Knowledge of RO duties in the control room during B RO duties in Control Room during refueling fuel handling such as alanns from fuel handling area, communication with fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation.
Radiition Control 2.3.1 3.0 Knowledge of 10 CFR: 20 and related facility B Radiation exposure determination radiation control requirements.
Emergency Procedures / Plan 2.4.16 4.0 Knowledge of EOP implementation hierarchy and B Performance of Status Trees / Function Restoration coordination with other support procedures.
Tuesday, June 16,1998 4:1c:27 PM Page 1 Prepared by WD Associates. Inc.
j s
i Facility-Byron Exam Datz
-9/14/98 Examination Lsv l: SRO 4
Section Title Generic Knowledge and Abilities SRO Group 1
Syst;m/ Evolution K/A SRO KA Statement Level Question Topic Emergency Procedures / Plan 2.4.20
- 4.0 Knowledge of operationalimplications of EOP B Applicability of Operator Action Summary Page wamings, cautions, and notes.
2.4.29 4.0 Knowledge of the emergency plan.
S Hazmat Spill Response 2.4.31 3.4 Knowledge of annunciators alarms and indications, B Identification of inoperable CR annunciators and use of the response instructions.
Tuesday, June 16,1998 4:18:27 PM Page 2
- Prepared byWD Associates,Inc.
. x m
m
...m i
1 l
Facility-Byron Exam Date:
9/14/98 Examination Level: SRO Section Title Plant Systems SRO Group 1
l System / Evolution K/A SRO KA Statement Level Question Topic Control Rod Drive System 001000A206 3.7 Effects of transient xenon on reactivity B Effect of Xenon Transient & compensation j
001000K103 3.6 CRDM B Application of DC Hold
~
Reactor Coolant Pump 003000A106 3.1 PZR spray flow B RCP and Pzr spray operations Systsm j
003000K110 3.2 RCS S SG temperature effect upon start of RCP Chsmical and Volume Control 004000A407 3.7 Boration/ dilution B Calculation of RCS Boron dilution I
System j
Engintered Safety Features 013000A301 3.9 Input channels and logic B CNMT Spray / Phase B i
Actuation System l
013000G110 3.9 Knowledge of conditions and limitations in the S Conditions for MODE change - ESFAS function 2
facility license.
inop Rod Position Indication 014000K502 3.3 RPIS independent of demand position S DRPI vs. Demand Position Syst:m Nucl:ar Instrumentation 015000A202 3.5* Faulty or erratic operation of detectors or B SR NIS discriminatorfailure i
Syst:m compensating components j
015000K201 3.7 NIS channels, components, and interconnections B SR NIS-loss of control power 015000K405 4.5 Reactor trip S Work on PR NIS affect on SR NIS Containment Spray System 026000A208 3.7 Safe securing of containment spray when it can be B Sequence for securing CNMT Spray i
done) 026000G110 3.9 Knowledge of conditions and limitations in the S CNMT Spray / Spray Additive operability
{
facility license.
requirements-Basis Main Feedwater System 059000G107 4.4 Ability to evaluate plant performance and make B SG Level program -Iow power operationaljudgments based on operating characteristics, reactor behavior, and instrument interpretation.
Auxiliary / Emergency 061000A301 4.2 AFW startup and flows B AFWStartup FeedwaterSystem 061000K502 3.6 Decay heat sources and magnitude B AFW flow requirements for cooldown I
i ' Tuesday, June 16,1998 4.18:29 PM Page 3
. Prepared by WD Associates, Inc.
F
i-f Facility-Byron Exam Data:
9/14/98 Examination Leval: SRO l
Section Title Plant Systems j
SRO Group 1
j Syst:m/ Evolution K/A SRO KA Statement Level Question Topic j
D.C. Electrical Distribution 063000G130 3.4 Ability to locate and operate components, including B DC bus battery charger local controls.
Liquid Radwaste System 068000A404 3.7 Automaticisolation B RCDT operation - effect of CNMT isolation I
Area Radiation Monitoring 072000K302 3.5 Fuel handling operations B Loss of FHS Overhead Crane rad monitor System J
i l
't
'i Tuesday, June 16,1996 4:18:30 PM Page 4 Prepared by WD Assocsotes,Inc.
Facility-Byron Exam Datz 9/14/98 Examination Leval: SRO Section Title Plant Systems SRO Group 2
Syst:m/ Evolution K/A SRO KA Statement Level Question Topic R: actor Coolant System 002000A111 3.2 Relative level indications in the RWST, the refueling B Relationship oflevels during refueling operations cavity, the PZR and the reactor vessel during preparation for refueling 002000G110 3.9 Knowledge of conditions and limitations in the S Conditions for loops operable /in operation facility license.
Em:rgency Core Cooling 006000A213 4.2 Inadvertent SIS actuation B Systems response to SI/ Actions Syst:m 006000K302 4.4 Fuel B 10CFR50.46 Design Criteria 006000K603 3.9 Safety injection Pumps B Evaluation of flow ECCS pumps Pressurizer Pressure Control 010000A108 3.3 Spray nozzle DT B Spray using Normal and Aux Spray Syst m 010000G112 4.0 Ability to apply technical specifications for a S DNB Limits system.
010000K501 4.0 Determination of condition of fluid in PZR, using B Evaluation of Pzr conditions steam tables Prasurizer Level Control 011000K104 3.9 RPS B Pzr Level Reactor Trip System R actor Protection System 012000A403 3.6 Channel blocks and bypasses B input that can be bypassed & plant conditions 012000K402 4.3 Automatic reactor trip when RPS setpoints are S Basis for OTdT with input exceeded for each RPS function; basis for each Non-Nuclear instrumentation 016000K302 3.5* PZR LCS B NR RTD Failure effects System Spent Fuel Pool Cooling 033000A201 3.5 Inadequate SDM S Safety Analysis on dilution during refueling Syst:m A.C. Electrical Distribution 062000K201 3.4 Major system loads S Evaluation of Electrical Supplies Emergency Diesel Generators 064000A307 3.7 Load sequencing B Sequencing of CNMT Spray pumps - S1 & SI w LOP Process Radiation Monitoring 073000K401 4.3 Release termination when radiation exceeds S Containmen Ventilation (Purge)
System setpoint Tuesday, June 16,1998 4:18:31 PM Page5 Prepared by WD Associates, Inc.
e e
Facility-Byron Exam Date:
9/14/98 Examination Level: SRO Section Title Plant Systems SRO Group 2
System / Evolution K/A SRO KA Statement Level Question Topic fig Protection System 086000K406 3.3 CO2 B Effect ofloss of DC - CO2 actuation
. Tuesday, June 16,1998 4:18:32 PM Page 6 Prepared byWD A%,Inc.
6
i 1
Facility-Byron Exam Data:
9/14/98 Examination Levet SRO Section Title Plant Systems l
SRO Group 3
}
System / Evolution K/A SRO KA Statement Level Question Topic 1
Residual Heat Removal 005000A202 3.7 Pressure transient protection during cold shutdown S Requirements / Operation of PORVs at low RCS Syst;m pressures 005000K112 3.4 Safeguard pumps B Recirc interties to St Pumps & CV Pumps St::am Dump System and
-041000A302 3.4. RCS pressure, RCS temperature, and reactor B Steam Dump input malfunction Turbine Bypass Control power Instrument Air System 078000K302 3.6 Systems having pneumatic valves and controls B Evaluation of eqpt affected for slow loss of IA -
~x Tuesday, June 16,1998 4:18:32 PM '
Page 7 Prepared by WD A-i=ses, Inc.
.e-
i l
Facility-Byron Exam Date:
9/14/98 Examination Level: SRO l
Section Title Emergency and Abnormal Plant Evolutions i
l SRO Group 1
l System / Evolution K/A SRO KA Statement Level Question Topic l
Continuous Rod Withdrawal 000001A205 4.6 Uncontrolled rod withdrawal, from available B Evaluate conditions - unwarranted rod withdrawal j
indications
{
000001K103 4.0 Relationship of reactivity and reactor power to rod S Reactivity effect with positive MTC movement i
j Dropped Control Rod 000003K310 4.2 RIL and PDiL B P/A vs. Group Step Counters j'
Inoperable / Stuck Control Rod 000005K106 3.8 Bases for power limit, for rod misalignment S Reason for power reduction i
]
Largs Break LOCA 000011A103 4.0 Securing of RCPs B RCP trip criteria evaluation 000011A204 3.9 Significance of PZR readings S Pzrlevel requirements 000011)(312 4.6 Actions contained in EOP for emergency LOCA S Use of Adverse containment (large break)
R: actor Coolant Pump 000015A210 3.7 When to secure RCPs on loss of cooling or seal B Evalloss of cooling flow (CCW) j Milfunctions injection 000015K207 2.9 RCP seals B Eval of RCP seal failure Emergency Boration 000024A205 3.9 Amount of boron to add to achieve required SDM B Time / amount E-boration for condition l
Loss of Component Cooling '000026A105 3.1 The CCWS surge tank, including level control and B Evaluation of CCWleak Wat:r level alarms, and radiation alarm
)
i 000026G224 3.8 Ability to analyze the affect of maintenance S Determine Tech Spec limitation on eqpt outage j
activities on LCO status.
j Anticipated Transient Without 000029G448 3.8 Ability to interpret control room indications to verify B. AMS conditions Scram the status and operation of system, and understand I
how operator actions and directives affect plant j
and system conditions.
j Stsam Line Rupture 000040A101 4.6 Manual and automatic ESFAS initiation B Steamline isolation 000040K106 3.8 High-energy steam line break considerations B Eval of Leak 4
Loss of Condenser Vacuum 000051A202 4.1 Conditions requiring reactor and/or turbine trip B Eval of conditions h
I Stition Blackout 000055A104 3.9 Reduction of loads on the battery S Evolutions required to be perfonned to reduce DC loads j
i Tuesday, June 16,1998 4:18:34 PM Page 8 Prepared by WD Associates. Inc.
l
Facility-Byron Exam Data:
9/14/98 Examination Level: SRO Section Title Emergency and Abnormal Plant Evolutions SRO Group 1
Systim/ Evolution K/A SRO KA Statement Level Question Topic Station Blackout 000055K302 4.6 Actions contained in EOP forloss of offsite and B Identification of RCP seal LOCA/cooldown onsite power Loss of Vital AC Instrument 000057A219 4.3 The plant automatic actions that will occur on the B Eqpt affected on bus loss Bus loss of a vital ac electrical instrument bus Control Room Evacuation 000068A121 4.1 Transfer of controls from control room to shutdown B Operations required for transfer panel orlocal control Inadequate Core Cooling 000074K103 4.9 Processes for removing decay heat from the core B Major action categories High Reactor Coolant Activity 000076A202 3.4 Corrective actions required for high fission product B Actions for reducing activity activity in RCS Pressurized Thermal Shock 00WE08K202 4.0 Facility's heat removal systems, including primary B Identification of heat removal process coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.
Natural Circulation Operations 00WE09K301 3.6 Facility operating characteristics during transient B Natural Circ conditions and limits conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics.
Tuesday, June 16,1998 4.18:35 PM Page 9 Prepared by WD Asscoates, Inc.
O
i, Facility-Byron Exam Date:
9/14/98 Examination Levet SRO Section Title Emergency and Abnormal Plant Evolutions 1
SRO Group 2
l j
System / Evolution K/A SRO KA Statement Level Question Topic
)
I l
Perctor Trip 000007A103 4.1 RCS pressure and temperature B Stabilized RCS temperature with failure of Steam l
1 Dumps j
000007A201 4.3 Decreasing power level, from available indications S EP-0 indications i
Pressurizer Vapor Space 000008G222 4.1 Knowledge of limiting conditions for operations and S Evaluation of PORV leak - Tech Spec Limits j
Accid:nt safety limits.
j Small Break LOCA 000009A110 3.9* Safety parameterdisplay system B Calculation of subcooled margin on iconic Display l
000009K316 4.1 Containment temperature, pressure, humidity and S Effects of Adverse containment l
levellimits Loss of Reactor Coolant 000022A108 3.3 VCTlevel B VCT level transmitter malfunction Makrup Loss of Residual Heat 000025K101 4.3 Loss of RHRS during all modes of operation B Calc of time to saturation / core boiling R:moval System 000025K301 3.4 Shift to altemate flowpath B Altemate RCS cooling i
Pressurizer Pressure Control 000027A101 3.9 PZR heaters, sprays, and PORVs B Pressure controller step change Malfundion j
000027A215 4.0 Actions to be taken if PZR pressure instrument fails B Non-Controlling channel failure high Loss of Source Range 000032K101 3.1 Effects of voltage changes on performance B Evaluation of SR NIS voltage failure Nucle:rInstrumentation i ' Loss ofintermediate Range 000033A204 3.6 Satisfactory overlap between source-range, B Eval of failed IR channel on SU l
Nucle rinstrumentation intermediate-range and power-range instrumentation St:am Generator Tube Leak 000037G132 3.8 Ability to explain and apply all system limits and S SG tube leak increase actions
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prect.utions.
Stram Generator Tube 000038K306 4.5 Actions contained in EOP for RCS water inventory B Loss of subcooling l
Rupture ba!ance, S/G tube rupture, and plant shutdown procedures 1
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5 Facility-Byron Exam Data:
9/14/98 Examination Level: SRO 1
Section Title Emergency and Abnormal Plant Evolutions i
SRO Group 2
Syst:m/ Evolution K/A SRO KA Statement Level Question Topic i
Loss of Secondary Heat Sink 00WE05K201 3.9 Components, and functions of control and safety B Interlocks affecting reestablishment of feed i
systems, induding instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Loss of Emergency Coolant 00WE11 A101 4.0 Components, and functions of control and safety B Reason for rapid SG depressurization 1
Recirculation systems, including l istrumentation, signals, interlocks, failum modes, and automatic and manual featurer,.
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Exam Data:
9/14/98 Examination Lcv l: SRO Section Title Emergency and Abnormal Plant Evolutions SRO Group 3
System / Evolution K/A SRO KA Statement Level Question Topic Pressurizer Level Control 000028K305 4.1 Actions contained in EOP for PZR level malfunction B Failed level channellow.
Malfunction Loss of Off-Site Power 000056A121 3.3* Reset of the ESF load sequencers B Reset of sequencer 000056A246 4.4 That the ED/Gs have started automatically and that B Eval of electric bus status the bus tie breakers are closed i
.Tuonday, June 16,1998 4:18:38 PM
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ES-401 PWR RO Examination Outline Form ES-401-4 Facility: Byron I & 2 Date of Exam:
September 14,1998 Exam Level: RO K/A Category Points Kl K2 K3 K4 K5 K6 Al A2 A3 A4 G
Total m 6 a
s 1.
I 2
2 2
lq ;,^
NE g 4
6 16 Emergency &
': s '
";'s
's 5 ':t 6
2 i,'
',5 1
17 2
3 2
3
'q', v^
- s Abnormal Plant
'2 Evolutions 3
1
- jf!; :'g:' '~'E '
1 I
EE 3
4
<4
,s, s e
Tier Totals 5
4 6
'f ;
11 9
1 36 1
3 2
1 2
2 1
1 2
3 4
2 23 2.
2 2
2 2
2 1
2 2
3 2
2 20 p
Systems 3
2 1
1 1
1 1
1 8
Tier Totals 7
2 4
5 4
2 3
5 7
7 5
51
- 3. Generic Knowledge and Abilitics Cat i Cat 2 Cat 3 Cat 4 13 5
3 2
3 Note:
Attempt to distribute topics among all K/A categories; select at least one topic from every K/A category within each tier.
Actual point totals must match those specified in the table.
Select topics from many systems; avoid selecting more than two or three K/A topics from a given system unless they relate to plant-specific priorities.
Systems / evolutions within each group are identified on the associated outline.
The shaded areas are not applicable to the category / tier.
NUREG-1021 30 of39 Interim Rev. 8, January 1997
PWR RO Exrminstien Outlins Facility Byron Exam Data:
9/14/98 Examination Levet RO Section Title Generic Knowledge and Abilities RO Group 1
System / Evolution K/A RO KA Statement Level Question Topic Conduct of Operations 2.1.1 3.7 Knowledge of conduct of operations requirements. B Evaluation of requirement for" active" license t
2.1.1 3.7 Knowledge of conduct of operations requirements. B Direction of NLO personnel 2.1.2 3.0 Knowledge of operator responsibilities during all B
Operating Daily Orders modes of plant operation.
2.1.14 2.5 Knowledge of system status criteria which require B MOV Stroke Time actions the notification of plant personnel.
2.1.23 3.9 Ability to perform specific system and integrated B~
Procedure required usage plant procedures during all modes of plant operation.
Equipment Control 2.2.13 3.6 Knowledge of tagging and clearance procedures. B MSIV OOS hang to prevent opening 2.2.26 2.5 Knowledge of rctueling administrative B
RCS level discrepancy during refueling requirements.
2.2.32 3.5 Knowledge of RO duties in the control room during B RO duties in Control Room during refueling fuel handling such as alarms from fuel handling area, communication with fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation.
Radiation Control 2.3.1 2.6 Knowledge of 10 CFR: 20 and related facility B
Radiation exposure determination radiation control requirements.
2.3.10 2.9 Ability to perform procedures to reduce excessive R Fuel Handling Accident Response levels of radiation and guard against personnel exposure.
Em:rgency Procedures 2.4.16 3.0 Knowledge of EOP implementation hierarchy and B Performance of Status Trees / Function Restoration
/ Plan coordination with other support procedures.
2.4.20 3.3 Knowledge of operationalimplications of EOP B
Applicability of EOP Operator Action Summary Page wamings, cautions, and notes.
2.4.31 3.3 Knowledge of annunciators alarms and B
Identification of inoperable CR annunciators indications, and use of the response instructions.
Tuesday, June 16,1998 4:02:58 PM '
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Facility-Byron Exam Date:
9/14/98-Examination Level: RO j
Section Title Plant Systems t
i RO Group 1
j System / Evolution K/A RO KA Statement Level Question Topic Control Rod Drive 001 A2.06 3.4 Effects of transient xenon on reactivity B
Effect of Xenon Transient & compensation j
System 001 K1.03 3.4 CRDM B
Application of DC Hold f
Reactor Coolant Pump 003 A1.06 2.9 PZR spray flow B-RCP and Pzr spray operations
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System 003 K2.01 3.1 RCPS R
RCP Breaker & interlocks Chemical and Volume 004 A3.11 3.6 Charging / letdown R
Charging & letdown flows (including seal injection)
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j Control System l
004 A4.07 3.9 Boration/ dilution B
Calculation of RCS Boron dilution l
004 K6.01 3.1 Spray / heater combination in PZR to assure R
Boron mixing uniform boron concentration
[
Engineered Safety 013 A3.01 3.7* Input channels and logic B
CNMT Spray / Phase B Features Actuation System l
013 K4.13 3.7 MFWisolation/ reset R
FW isolation - P14 NuclearInstrumentation 015 A2.02 3.1 Faulty or erratic operation of detectors or B
SR NIS discnminatorfailure
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System compensating components 1
l 015 K2.01 3.3 NIS channels, components, and interconnections B SR NIS-loss of control power
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015 K5.06 3.4 Subcritical multiplications and NIS indications R
Eval for 1/M - Eightfold increase L
In-Core Temperature 017 K4.01 3.4 Input to subcooling monitors R
CETC failure effect on Subcooling Monitor / Iconic Monitor System Display l
Containment Cooling 2.1.32 3.4 Ability to explain and apply all system limits and
-R RCFC operations requirements System precautions.
j Main Feedwater 2.1.7 3.7 Ability to evaluate plant performance and make B
SG Level program -low power t
System operationaljudgments based on operating characteristics, reactor behavior, and instrument j
interpretation.
j 059 K1.04 3.4 S/GS water level control system R
Effect of failure of SG steam pressure channel l
Auxiliary / Emergency 061 A3.01 4.1 - AFW startup and flows B
AFW Startup Feedwater System i
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PWR RO Examination Outline Facility-Byron Exam Data:
9/14/98 Examination Level: RO Section Title Plant Systems RO Group 1
System / Evolution K/A RO KA Statement Level Question Topic Auxiliary / Emergency 061 K5.02 3.2 Decay heat sources and magnitude B
AFW flow requirements for cooldown Feedwater System Liquid Radwaste 068 A4.04 3.8 Automaticisolation B
RCDT operation - effect of CNMT isolation System 068 K1.07 2.7 Sources ofliquid wastes for LRS R
CNMT Sump sources of input during normal operations W1.ste Gas Disposal 071 A4.05 2.6* Gas decay tanks, including valves, indicators, and R Waste Gas Decay Tank Operations System sample line l
Ara 3 Radiation 072 A4.03 3.1 Check source for operability demonstration R
Check Source operation Monitoring System 072 K3.02 3.1 Fuel handling operations B
Loss of FHB Overhead Crane rad monitor n
Tuesday, June 16,1998 4:03:01 PM Page 3 Prepared byWD Associates,Inc. ~
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Facility Byron Exam Date:
9/14/98-Examination Leval: RO Section Title Plant Systems RO Group 2
System / Evolution K/A RO KA Statement Level Question Topic R: actor Coolant 002 A1.11 2.7 Pdative level indications in the RWST, the B
Relationship of levels during refueling operations 4
System refueling cavity, the PZR and the reactor vessel during preparation for refueling 002 A3.01 3.7 Reactor coolant leak detection system R
RCS leak Detection Systems 002 K4.09 3.2 Operation of loop isolation valves.
R Use of Loop isolation Valves Em:rgency Core 006 A2.13 3.9 Inadvestent SIS actuation 8
Systems response to SI/ Actions j
Cooling System j
006 K3.02 4.3 Fuel B
10CFR50.46 Design Criteria 1
006 K6.03 3.6 Safety injection Pumps B
Evaluation of flow ECCS pumps Pressurizer Pressure 010 A1.08 3.2 Spray nozzle DT B
Spray using Normal and Aux Spray Control System i
010 K5.01 3.5 Determination of condition of fluid in PZR, using B
Evaluation of Pzr conditions steam tables Pressurizer Level 011 K1.04 3.8 RPS B
Pzr Level Reactor Trip 1
Control System R actor Protection 012 A3.07 4.0 Trip breakers R
Operation of BOTH Bypass Trip Breakers System 012 A4.03 3.6 Channel blocks ::vi bypasses B
Input that can be bypassed & plant conditions 012 K5.01 3.3* DNB R
OTdT inputs & effect of changes Rod Position Indication 2.4.31 3.3 Knowledge of annunciators alarms and R
ROD BOTTOM Alarm operation System indications, and use of the response instructions.
Non-Nuclear 016 K3.02 3.4* PZR LCS B
NR RTD Failure effects 4
Instrumentation System Containment Spray 026 A2.08 3.2 Safe securing of containment spray when it can B
Sequence for securing CNMT Spray System be done) 026 A4.01 4.5 CSS controls R
Pump operation interlocks Spent Fuel Pool Cooling 033 K1.05 2.7* RWST R
RWST Purification Loops System 2
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Facility-Byron Exam Data:
9/14/98 Examination Level: RO j
Section Title Plant Systems
]
RO Group
.2 j
SystenVEvolution K/A-RO KA Statement Level Question Topic 4
D.C. Electncal 2.1.30 3.9 Ability to locate and operate components, B
DC bus battery charger i
Distribution including local controls.
I Emergency Diesel.
064 A3.07 3.6* Load sequencing B
Sequencing of CNMT Spray pumps - SI & SI w LOP Generators Fira Protection System 086 K4.06 3.0 CO2 B
Effect of loss of DC - CO2 actuation f
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PWR RO Examinati::n Outline Facility-Byron Exam Datx 9/14/98--
Examination Level: RO Section Title Plant Systems RO Group 3
System / Evolution K/A RO KA Statement Level Question Topic
' R::sidual Heat Removal 005 K1.12 3.1 Safeguard pumps B
Recirc interties to Si Pumps & CV Pumps System 005 K4.10 3.1 Control of RHR heat exchanger outlet flow R
Failure of Hx Outlet Valve Pressurizer Relief 2.4.50 3.3 Ability to verify system alarm setpoints and R
PRT conditions causing alam1/ response Tenk/ Quench Tank operate controls identified in the alarm response System manual.
Component Cooling 008 A2.05 3.3* Effect of loss of instrument and control air on the R Determination of effect of valve positioning Water System position of the CCW valves that are air operated Containment lodine 027 A4.03 3.3* CIRS fans R
Charcoal Filters response to deluge R:moval System Steam Dump System 041 A3.02 3.3 RCS pressure, RCS temperature, and reactor B
Steam Dumpinput malfunction and Turbine Bypass power Control Main Turbine Generator 045 K1.20 3.4 Protection system R
Turbine Control response to Failed impulse Channel System Instrument Air System 078 K3.02 3.4 Systems having pneumatic valves and controls B
Evaluation of eqpt affected for slow loss of IA Tuesday, June 16,1998 4:03:04 PM Page 6 Prepared byWD Associates,Inc.
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PWR RO Examinatien Outlina Facility-Byron Exam Date:
9/14/98 Examination Level: RO Section Title Emergency and Abnormal Plant Evolutions j
RO Group 1
System / Evolution K/A RO KA Statement Level Question Topic 1
R: actor Coolant Pump 015 AA2.10 3.7 When to secure RCPs on loss of cooling or seal B
Evalloss of cooling flow (CCW)
Malfunctions injection l
015 AK2.07 2.9 RCP seals B
Eval of RCP seal failure Emergency Boration 024 AA2.05 3.3 Amount of boron to add to achieve required SDM B Time / amount E-boration for condition 1
Loss of Component 026 AA1.05 3.1 The CCWS surge tank, including level control and B Evaluation of CCWleak Cooling Water level alarms, and radiation alarm 4
i Pressurizer Pressure 027 AA1.01 4.0 PZR heaters, sprays, and PORVs B
Pressure controller step change l
Control Malfunction 027 AA2.15 3.7 Actions to be taken if PZR pressure instrument B
Non-Controlling channel failure
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fails high Stram Line Rupture 040 AA1.01 4.6 Manual and automatic ESFAS initiation 8
Steamline isolation 040 AK1.06 3.7 High-energy steam line break considerations B
Eval of Leak l
Loss of Condenser 051 AA2.02 3.9 Conditions requiring reactor and/or turbine trip B
Eval of conditions l
Vacuum Station Blackout 055 EK3.02 4.3 Actions contained in EOP forloss of offsite and B
Identification of RCP seal LOCA/cooldown onsite power Loss of Vital AC 057 AA2.19 4.0 The plant automatic actions that will occur on the B Eqpt affected on bus loss instrument Bus loss of a vital ac electricalinstrument bus Control Room 068 AA1.21 3.9 Transfer of controls from control room to shutdown B Operations required for transfer Evacuation panel orlocal control Inadequate Core 074 EK1.03 4.5 Processes for removing decay heat from the core B Major action categories Cooling High Reactor Coolant 076 AA2.02 2.8 Corrective actions required for high fission product B Actions for reducing activity Activity activity in RCS Pressurized Thermal E08 EK2.2 3.6 Facility's heat removal systems, including primary B Identification of heat removal process Shock coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.
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PWR RO Examination Outline Facility-Byron Exam DatI 9/14/98-Examination Level: RO Section Title Emergency and Abnormal Plant Evolutions RO Group
-1 System / Evolution K/A RO KA Statement Level Question Topic Nitural Circulation E09 EK3.1 3.3 Facility operating characteristics during transient B Natural Circ conditions and limits Operations conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics.
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PWR RO Examin:.tian Outlina Facility: Byron Exam Date:
9/14/98 Examination Level: RO Section Title Emergency and Abnormal Plant Evolutions RO Group 2
System / Evolution K/A RO KA Statement Level Question Topic Continuous Rod 001 AA2.05 4.4 Uncontrolled rod withdrawal, from available B
Evaluate conditions - unwarranted rod withdrawal Withdrawal indications Dropped Control Rod 003 AK3.10 3.27 Rlland PDIL B
P/A vs. Group Step Counters Reactor Trip 007 EA1.03 4.2 RCS pressure and temperature B
Stabilized RCS temperature with failure of Steam Dumps 007 EK2.03 3.5 Reactor trip status panet R
Reactor Trip requirements Pressurizer Vapor 008 AK1.01 3.2 Thermodynamics and flow characteristics of open R Tail-Pipe conditions Space Accident orleaking valves Small Break LOCA 009 EA1.10 3.8* Safety parameter display system B
Calculation of subcooled margin on iconic Display Large Break LOCA 011 EA1.03 4.0 Securing of RCPs B
RCP trip criteria evaluation Loss of Reactor Coolant 022 AA1.08 3.4 VCTlevel B
VCT level transmitter malfunction Makeup Loss of Residual Heat 025 AK1.01 3.9 Loss of RHRS during all modes of operation B
Calc of time to saturation / core boiling Removal System 025 AK3.01 3.1 Shift to altemate flowpath B
Altemate RCS cooling Anticipated Transient 2.4.48 3.5 Ability to interpret control room indications to B
AMS conditi<as Without Scram verify the status and operation of system, and understand how operator actions and directives affect plant and system conditions.
Loss of Source Range 032 AK1.01 2.5 Effects of voltage changes on performance B
Evaluation of SR NIS voltage failure Nuclear instrumentation Loss of Intermediate 033 AA2.04 3.2 Satisfactory overiap between source-range, B
Eval of failed IR channel on SU Range Nuclear intermediate-range and power-range Instrumentation instrumentation Stram Generator Tube 037 AA1.02 3.1* Condensate exhaust system R
Monitors for SG Tube leakage Luk Stram Generator Tube 038 EK3.06 4.2 Actions contained in EOP for RCS water inventory B Loss of subcooling Rupture balance, S/G tube rupture, and plant shutdown procedures Tuesday, June 16,1998 4:03:08 PM
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PWR RO Examination Outline Facility-Byron Exam Data:
9/14/98-Examination Lev;l: RO Section Title Emergency and Abnormal Plant Evolutions RO Group 2
System / Evolution K/A RO KA Statement Level Question Topic Loss of Secondary Heat E05 EK2.1 3.7 Components, and functions of control and safety B Interlocks affecting reestablishment of feed Sink systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Loss of Emergency E11 EA1.1 3.9. Components, and functions of control and safety B Reason for rapid SG depressurization Coolant Recirculation systems, including instrumentation, signals, interiocks, failure modes, and automatic and manual features.
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PWR RO Examination Outlins Facility-Byron Exam Data:
9/14/98 Examination Lavnl: RO Section Title Emergency and Abnormal Plant Evolutions RO Group 3
System / Evolution K/A RO KA Statement Level Question Topic Pressurizer Level 028 AK3.05 3.7 Actions contained in EOP for PZR level B
Failed level channel low.
Ccntrol Malfunction malfunction Loss of Off-Site Power 056 AA1.21 3.3* Reset of the ESF load sequencers B
Reset of sequencer 056 AA2.46 4.2 That the ED/Gs have started automatically and B
Eval of electric bus status that the bus tie breakers are closed Tuesday, June 16,1998 4:03:09 PM Page 11
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