ML20199E095
| ML20199E095 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 12/23/1998 |
| From: | Peebles T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Fulmer S SOUTHERN NUCLEAR OPERATING CO. |
| References | |
| NUDOCS 9901200309 | |
| Download: ML20199E095 (102) | |
Text
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December 23, 1998 Southern Nuclear Operating Company, Inc.
ATTN: Mr. Scott Fulmer, Plant Training and Emergency Preparedness Manager Joseph M. Farley Nuclear Plant P. O. Drawer 470 Ashford, AL 36312
SUBJECT:
MEETING
SUMMARY
- TRAINING MANAGERS' CONFERENCE CONDUCTED ON NOVEMBER 5,1998 - JOSEPH M. FARLEY NUCLEAR PLANT
Dear Mr. Fulmer:
This letter refers to the Training Managers' Conference conducted at the Richard B. Russell Building on November 5,1998. Representatives from all utilities in Region 11 participated in the meeting.
Enclosure 1 was the agenda used for the Training Managers' Conference, and Enclosure 2 is the list of attendees. We appreciate the participation of you and your staff and believe that the goal of providing an open forum for discussion of operator licensing issues was met. Mr. Gallo, Chief of the Operator Licensing and Human Factors Branch, Office of Nuclear Reactor i Regulation (NRR), made a presentation as noted in the slides in Enclosure 3. I Additionally, I am enclosing our preliminary schedule for FY 1999 and FY2000 as Enclosure 4.
Please review the schedule and supply comments to my staff or myself.
If you have any questions regarding the content of this letter, please contact me at (404) i 562-4638. !
l Sincerely, 1 Original signed by Thomas A. Peebles Thomas A. Peebles, Chief Operator Licensing and Human Performance Branch Division of Reactor Safety Docket Nos.: 50-348 and 50-364 i License Nos.: NPF-2 and NPF-8 l
Enclosures:
As noted I cc w/encls: 200M3 \
M. Stinson, General Manager, Joseph M. Farley Nuclear Plant i
9901200309 981223 PDR ADOCK 05000348 V PDR LY3
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Distribution w/encis:
PUBLIC i B. Michael, DRS l l l l
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Ril:DRS Ril:DRS M 9 BMICHAEL: TP BLES 12/ tr /98 12/g/98 l Doc Name:
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, Revised November 3,1998 l
L TRAINING AND OPERATIONS MANAGERS' CONFERENCE U.S. Nuclear Regulatory Commission, Region II Atlanta, Georgia Meeting Agenda November 5,1998 Richard B. Russell Building Auditorium Thursday.11/5/98 l
8:00 a.m. Conference Registration l
l 8:20 a.m. Introduction Thomas A. Peebles, Chief, Operator Licensing & Human Performance Branch 8:30 a.m. Opening Remarks William Travers, ,
Executive Director of Operations 8:50 a.m. Welcome / Issues Raised the Last Meeting Bruce S. Mallett, Director Division of Reactor Safety 9:15 a.m. break 9:45 a.m. Other Issues Robert M. Gallo, Chief Operator Licensing Branch NRR 10:15 a.m. Lessons learned from Recent Exams Charlie Payne l Sampling exam criteria 11:00 a.m. Examination Communications Ron Aiello l Exam Development & Coordination 11:30 a.m. Lunch .
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1:00 p.m. Written Examination Questions and Answers Rick Baldwin / George Hopper 2:30 p.m. JPM Examples of questions Rick Baldwin / George Hopper 3:30 p.m. Open Session - Other Issues Training Managers j 4:00 p.m. Meet with Principle examiners All j 4:30 p.m. Adjorn i
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License Applicant Adm-inistrative Walkthrough Examination--NRC-1 .
Examiner Sheet A.1: Shift Staffing' i Question 1: A licensed RO has been off-shin for 6 months to assist in outage scheduling. He is informed that he is needed to join a shift crew in 2 weeks to fill in for a vacationing shift RO. He had a satisfactory physical examination 18 months ago and has maintained satisfactory performance in the licensed operator requalification training I program. In anticipation of rejoining a shift crew, the RO has spent two 8-hour l shins in the control room as the OATC during the past month under the direction of l a shift RO.
What additional requirements must be met by the RO before he may fill in for the vacationing RO?
References Allowed? YES1NO Answer: The RO must complete an additional 24 (40 - 16) hours of"under dir~ection" on-shift time that must include a plant tour and shift turnover.
Reference:
SSP-12.1, Conduct of Operations, p. 61 and 62 K/A 2.1.4 (2.3/3.4)
Applicant Response: SAT UNSAT i
.. -. . . -.. .- . . _ . - . . - . - - - . . - - . - ~ - ~ - ..-
ATTENDEES AT THE NRC REGION ll TRAINING MANAGERS CONFERENCE NOVEMBER 5,1998 i l
Sid Crouch ATTSi i David Lane Sonalysts, Inc. l Bill Fitzpatrick INPO CP&L Rick Gamer HR Supv Ops Tmg William Noll BK Ops Tmg Supv Max Herrell BK Tmg Mgr Scott Poteet RB Exam Team Leader Ralph Mullis BK Ops Mgr j Tony Pearson BK Ops Tmg Anthony Williams RB Tmg Mgr Crystal River - FPC ,
Wes Young CR Supv OpsTng i Tom Taylor CR Dir Nuc Ops Tmg Ivan Wilson CR Ops Mgr l Ktn McCall CR Mgr Ops Tmg '
Duke Power l Gabriel Washbum OC Req Team Leader Ronnie B. White, Jr MG Tmg Mgr W. H. " Soap" Miller CT Site Tmg Mgr !
P ul Stovall OC Mgr Oper Tmg i Bentley Jones OC Tmg Mgr l James Teofilak CT Ops Trng Mgr Alan Orton MG Ops Trng Mgr Richard Bugert Corp Ops Tmg Spec EElk Tmg Mgr Maria Lacal TP Dennis L. Fadden SL Services Mgr -
Jo Magennis Corp Trng Assessment Spec Tom Bolander SL Exam Development Steve McGarry TP Maint Tmg Supv Southem Nuclear (SNC) i John C. Lewis -HT Trng & EP Mgr Bill Oldfield FA Nuc Ops Tm Supv Steve Grantham HT Ops Tmg Supv Scott Fulmer FA Mgr Tmg & EP j Joel Deavers FA Sr Pit Inst Bob Brown VG Trng Mgr Dan Scukanec VG Ops Tmg Supv
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4 l Viroinia Power ;
I Steve Crawford NA Sr Inst Nuc ;
Harold McCallum SR Supv Ops Tmg l l
i TVA Dick Driscoll SQ Tmg Mgr
! Walt Hunt SQ Ops Trng Mgr i Denny Campbell BF SRO Ops Inst l Jack Cox WB Tmg Mgr John Roden WB Ops Tmg Mgr Tom Wallace WB Ops Supt V. C. Semmer- SCE&G l
AlKoon SM Ops Tmg Supv NRC Particioants Tom Peebles R 11 Operator Lic. Br. Ch.
Rick Baldwin R ll Sr. Examiner George Hopper R ll Sr. Examiner Ron Aiello R 11 Sr. Examiner Charlie Payne R 11 Sr. Examiner William Travers NRC Executive Dir. Ops.
R. M. Gallo NRR Br. Ch. OL j Bruce Mallett R 11 Div. Dir. Reactor Safety l
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l FY 99 INITIAL EXAM SCHEDULE AND RESULTS 1
Drcember 14,1998 I
D:te Plant Chief Pass # Pass # Pass # Pass # !
l 9/28/98 Sequoyah GTH 4 4 4 4 f 10/5/98 Harris RFA 2 2 5 5 7 7 11/30/98 Oconee GTH 2 2 5 5 7 7 l 11/30/98 St Lucie & RSB 6 3 9 12/14/98 1/25/99 McGuire & DCP 6 3 2 11 2/8/99 2/8/99 C. River & GTH 6 6 12 2/22/99 2/8/99 B. Ferry MEE 4 1 5 l 3/29/99 Surry & RSB 5 2 4 11 4/12/99 l
! 1 4/12/99 Watts Bar & MEE 6 3 5 14 4/26/99 5/10/99 Failey GTH 7 1 8 i
5/24/99 Catawba & PMS 8 5 3 16 6/7/99.
6/28/99 St. Lucie RSB 1 4 5 07/26/99 Robinson MEE 3 2 2 7 l 08/30/99 Turkey Pt & RFA 20 20
_9/13/99 136 RESULTS TO DATE 4 4 5 5 9 9 18 18 100 100 100 100 No initial exams scheduled for: Brunswick, North Anna and Vogtle FY 00 region 11 write part of Summer & Hatch i
FY 00 INITIAL EXAM SCHEDULE AND RESULTS ecember 14,1998 RO SRO-l SRO-U TOTAL Dr.te Plant Chief Pass # Pass # Pass # Pass 0 I 9/27/99 Summer GTH 6 6 region ll write
! 10/18/99 Hatch DCP 10 2 12
! region ll write 12/13/99 Vogtle RSB 3 5 2 10 l 2/14/00 Brunswick & DCP 12 3 15 l 2/28/00 1
03/**/00 Oconee 7 10 704/10/00 Harris (mayo to/co) 10 705/03/00 St. Lucie GTH 6 5 11 705/**/00 B. Ferry 6 3 3 12 705/03/00 McGuire 4 8 12 706/07/00Farley RSB 10 2 12 707/26/00 Ciystal River RFA 3 3 3 9 r:gion ll write?
708/**/00 Sequoyah 4 2 2 8 709/04/00 Surry? 10 709/11/00 Nodh Anna 12 0 0 42 0 46 0 27 0 149
- d signates tentative No initial exams scheduled for: Catawba Robinson Turkey Point Watts Bar
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Operator Licensing lSSueS .
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l l Region ll -
j Training Managers' Conference November 5,1998 '
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i Robert M. Gallo, Chief i
Operator-Licensing and j
Human Performance Branch l
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l OPERATOR LICENSING ISSUES o
Part 55 Rulemakings _
Status Schedule o
Final Revision 8 of NUREG-1021 o
Examination Quality and Results
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o Generic Fundamentals Exam o
Requal Inspections ilP-710.01) o Recent Information Notices Exam Integrity i:lN 98-15)
Sampling Plans ilN 98-28)
Eligibility ilN 98-37)
RECENT LESSONS LEARNED by
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Charlie Payne Southeastern Training Manager's Conference November 5,1998 t
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- t PC1LLCY CLARIFICATIONS I
O In general, the NRC prefers that the written .
exam be administered after the operating tests are complete.
Allows more time to finalize test.
More flexibility if delays occur.
Less stressful on candidates. 3 t
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PC1LICY C1ARIFICATIONS 9 In general, license class sizes of greater than H candidates will be scheduled for 2 weeks as follows:
t exam week 1
off-week for documentation of week 1 performance 2"d exam week
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POLICY C_LARIFICATIC>NS 9 Examination submittals - 2 copies of draft and final exams l' written, JPMs, and simulator scenarios). Electronic copy is also desired.
9 Written exams submittals will be reviewed by following a sampling process. When criteria are met, review will be stopped and licensee 3 called.
l Criteria - 10 unacceptable questions out of 30 questions sampled ,
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RECENT LE_SSONS LEARNED SRO-only Questions intended to sample those K/As specific to SRO duties (above and beyond those needed by an RO).
Jurpose is to meet the requirements of 10 CFR 55.43(b) (items (1) - (7)).
K/A catalog cross-references K/As to associated portions of 10 CFR 55.
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2.0 GENERIC KNOWLEDGES AND ABILITIES 2.1 Conduct of Operations 2.1.1 Knowledge of conduct of operations requirements.
(CFR: 41.10 / 45.13)
-- IMPORTANCE RO 3.7 SRO 3.8 2.1.2 KnowNie of operator responsibilities during all modes of plant operation.
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(CFh: 41.10 / 45.13) -
IMPORTANCE RO 3.0 SRO 4.0 2.1.3 Knowledge of shift turnover practices.
(CFR: 41.10 / 45.13)
IMPORTANCE RO 3.0 SRO 3.4 2.1.4 Knowledge of shift staffing requirements.
M (CFR: 41.10 /43.2)
IMPORTANCE RO 2.3 SRO 3.4 2.1.5. Ability to locate and use procedures and directives related to shift staffing and activities.
(CFR: 41.10 / 43.5 / 45.12)
IMPORTANCE RO 2.3 SRO 3.4 2.1.6 Ability to supervise and assume a management role during plant transients and upset conditions.
4 (CFR: 43.5 / 45.12 / 45.13)
IMPORTANCE RO 2.1 SRO 4.3 2.1.7 Ability to evaluate plant performance and make operationaljudgments based on operating characteristics / reactor behavior / and instrument interpretation.
(CFR: 43.5 /45.12 /45.13)
IMPORTANCE RO 3.7 SRO 4.4 2.1.8 Ability to coordinate personnel activities outside the control room.
(CFR: 45.5 /45.12 /45.13)
IMPORTANCE R O 3.8 SRO 3.6 2-1 E* REG-1123, hev. 2 l
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2.1 Conduct of Operations (continued) l-
- 2.1.19 Ability to use plant computer to obtain and evaluate parametric information on l system or component status.
! (CFR: 45.12)
IMPORTANCE RO 3.0 SRO 3.0 2.1.20 Ability to execute proc' edure steps.
(CFR: '41.10 / 43.5 / 45.12)
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l 2.1.21 Ability to obtain and verify controlled procedure copy.
l (CFR: 45.10 /45.13)
IMPORTANCE RO 3.1 SRO 3.2 2.1.22 Ability to determine Mode of Operation.
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(CFR: 43.5 / 45.13) '
IMPORTANCE R O 2.8 SRO 3.3 l
2.1.23 Ability to perform specific system and integrated plant procedures during -
( different modes of plant operation.
! (CFR: 45.2 /45.6)
2.1.24 Ability to obtain and interpret station electrical and mechanical drawings.
(CFR: 45.12 / 45.13)
' 2.1.25 Ability to obtain and interpret station reference materials such as graphs /
monographs / and tables which contain performance data. '
(CFR: 41.10 / 43.5 / 45.12)
2.1.26 Knowledge of non-nuclear safety procedures (e.g. rotating equipment / electrical /
l high temperature / high pressure / caustic / chlorine / oxygen and hydrogen).
l (CFR: 41.10 / 45.12) l IMPORTANCE RO 2.2 SRO 2.6 l
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l 2.4 Emergency Procedures / Plan (Continued) ;
2.4.32 Knowledge of operator response to loss of all annunciators.
(CFR: 41.10 / 43.5 / 45.13)
IMPORTANCE RO 3.3 SRO 3.5 2.4.33 Knowledge of the process used track inoperable alarms.
. (CFR: 41.10 / 43.5 / 45.13)
IMPORTANCE R O 2.4 SRO 2.8 2.4.34 Knowledge of RO tasks performed outside the main control room during emergency operations including sysIem geography and system implications.
l (CFR: 43.5 / 45.13)
IMPORTANCE R O 3.8 SRO 3.6 2.4.35 Knowledge of local auxiliary operator tasks during emergency operations including system geography and system implications.
--- -> (CFR: 43.5 / 45.13)
IMPORTANCE R O 3.3 SRO 3.5 l
2.4.36 Knowledge of chemistry / health physics tasks during emergency operations.
(CFR: 43.5)
IMPORTANCE R O 2.0 SRO 2.8 2.4.37 Knowledge of the lines of authority during an emergency.
- (CFR
- 45.13)
IMPORTANCE " R O 2.0 SRO 3.5 l 2.4.38 Ability to take actions called for in the facility emergency plan / including (if required) supporting or acting as emergency coordinator.
(CFR: 43.5 /45.11) l IMPORTANCE R O 2.2 SRO 4.0
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2.4.39 Knowledge of the RO's responsibilities in emergency plan implementation. !
(CFR: 45.11)
IMPORTANCE RO 3.3 SRO 3.1 2.4.40 Knowledge of the SRO's responsibilities in emergency plan implementation.
r (CFR: 45.11) l IMPORTANCE RO 2.3 SRO 4.0 2.4.41 Knowledge of the emergency action level thresholds and classifications.
. r (CFR: 43.5 / 45.11)
l NURE:;- n23, Rev 2 2-14 l
RECENT LE_S_ SONS LEARNED SRO-only Questions l Cont'd)
- c. SRO-only dluestions will be based on following categories: A.2, G2.1, G2.2, G2.3, and G2.4.
differences between SRO and RO outlines shifts only 11 K/As from Tier 2 to Tiers 1 & 3.
- Otler 14 flexible.
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l ES-401 BWR SRO Examination Outline Form ES-401-1 fFacility: Date of Exam: Exam Level:
K/A Category Points Tier Group Point K K K K K K A A A A G Total 1 2 3 4 5 6 1 2 3 4 3*
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gr Er Emergency &
EEE Bi gg f@ @ 26 C,,L. ,. ,
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Abnormal 2 NW Q, p@^ Q3.. ' ._ 17 3;pe v
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w Evolutions Tier Ep j,i[ . , .E g g4 43 +7 Totals M -%
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1 23 2.
P1 ant 13 Systems 3 4 Tier 40 ,,,7f Tctals
- 3. Generic Knowledge and Cat 1 Cat 2 Cat 3 Cat 4 Abilities 17 +$
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 .
i plant-specific priorities.
l Systems / evolutions within each group are identified on the associated outline.
The shaded areas are not applicable to the category / tier.
I NUREG-1021 10 of 39 Interim Rev. 8, January 1997 t .
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lt ES-401 BWR R0 Examination Outline Form ES-401-2 l Facility: Date of Exam: Exam Level- 4 I
K/A Category Points I
! Tier Group Point !
l K K K K K K A A A A G
' - Total 1 2 3 4 5 6 1 2 3 4 1-Einergency &
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- 2. i 2 19 i Plant l Systems 3 4 l Tier 51 Totals
- 3. Generic Knowledge and Cat 1 Cat 2 Cat 3 Cat 4 Abilities 13 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 j or three K/A topics from a given system unless they relate to l: plant-specific priorities.
Systems / evolutions within each group are identified on the associated outline.
The shaded areas are not applicable to the category / tier.
s NUREG-1021 16 of 39 Interim Rev. 8, January 1997
RE_C_ENT LE_S_S1DNS LEARNED Sam aling Criteria '
intent of prodess is to avoid exam predictability.
a so to avoid excessive use of repeat test items.
first use systematic process to develoa samale alan using topics from K/A catalog, then use i facility c uestion resources to accompiish the plan. . i 7 i
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RE_C_ENT LE_S_ SONS LEARNED
. j Sampling Criteria (Cont'd}
each topic in each tier & group should be sampled at east once unless insufficient questions exist to do this. If all topics have been sampled once and other questions need to be selected, the process should be ,
systematic and unbiased.
i final sample alan should have a fairly even 3alance across all Ks & As.
8
_BECENT LESS_ SONS LEARNED
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Samalinc Criteria (Cont'd) up to 25 questions from last two NRC exams, facility licensee exams, tests & quizzes (except final audit test) may be used.
Chief Examiner (CE) has the option to unilaterally shift or change the selected K/As.
i u3 to 5 site-s3eci"ic ariorities may be identified wit 7 CE concurrence (K/A value may be < 2.5 ;
with suf"icient 'ustification).
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_C'TH ER Record Keeping 3er 10 CFR 55 are required to provide evidence that the applicant has successfully comp eted the facility licensee's requirements to be licensed as an operator.
this includes successful manipulation of the controls of their facility. As a minimum,5 sic nificant control manipulations which affect reactivity or power level.
10
OTHER
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Record Keeping (Cont'd)
I this informatipn shou'd be retained anc available for inspection from time of license application to license expiration.
t 11
OTHER -
Requal Control Mani3ulations if have arogram based on SAT process, list in 10 CFR 55.59(cX3) does not need to be strictly followed.
should have sometling similar based on alant JTA and specific plant priorities.
some manipulations are individual operator oriented, most would be team oriented.
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OTHER Requal Control Manipulations (Cont'd) credit for accomplishment should only be given for _ active _ participation in the manipulation.
NOTE: control manipulations are not synonymous with reactivity manipulations.
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Examination Communications l Examination Development I &
3 i Coordination By I
i Ronald F. Aiello
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Facility Suggested improvements
- 1. The exam development team and leac examiner sloulc meet at the aeginning of t7e cevelopment arocess to estaa is, common grounds for the ceveloament and execution of the examination:
- Changes and interaretations to the NUREG.
- Scoae of the exam development and administration process.
- Lessons learned from t1e last exam ac ministered.
' 2. Vlove c ue cates "or :le out ine anc t7e exam 3ack to 90 and 60 cays arior to area week. This will provice more time for
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examination review ay t1e examiner (s).
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- 3. SSNTA continue with efforts to stanc ardize document formats for examination tools (JPMs and scenarios).
- 4. Examiners maintain a list of who (plant) does the exam process the best. This s1ould probably ae bro <en cown to eac, 3ortion of the examination. Provide your ratings to the utilities in Region 2, so we l can meet your expectations anc imarove.
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- 5. T1e 3rinciaal anc t1e utility rearesentative shou c meet early to estaalis, a wor <ing re ations1'ia anc exaectations. If possible this s1ould incluce samales of cuestions,
- JPMs, etc.
- 6. The exam should have no outstancing issues /c uestions that arise and need repair at the last minute. These issues should a l have been identified by the prea wee <, to allow time to make changes that ;
meet all the criteria. l
- 7. The chief examiner should exalain up front all the forms in 1021 tlat need to be completed.
- 8. Always check on badcing arior to coming on site.
- 9. As soon as a Chief Examiner is assigned !
to an exam, the Facility Rea and the C1ief s1ould verify t1e ability to communicate via all channels (inc uding e-mail). When we converted to Lotus Notes, the facility was suddenly unable to send e-mai to his Chief Examiner. This
became somewlat of a lindrance and s1ould be avoided if 30ssible.
- 10. It would be helpful if the Chie" Examiner could arovide his schec u e to the Faciity Rep. This includes aroviding updates for any changes to the Chief Examiner's schedule along the way. The facility rep needs to be aware of when the Chief Examiner is available to assist in exam preaaration activities.
- 11. A face-to-face meeting should be promatly scleduled in order for the Chief to communicate his expectations to the <
Faciity Rep. The face-to-face requirement could ae waivec if the Clief and the Faciity Rea have areviously wor <ed together and t1e Facility Rea is conficent that he/she understancs the '
Clief's ex3ectations. In any case, a
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! conference call would be the minimum to l satisfy t1is im aortant first ste 3.
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j 12. The Chief and the Facility Rep should
- wor < togetler to establish a firm schedule j for the exam week (s). This will ensure the l most efficient schedule is developed (with j resaect to crew composition and
. personnel movement) to minimize the l amount of exam material requirec. l
) 13. The Chief Examiner and Utility Rep MUST l remain fixed during t1e entire 180 day l 3eriod. Handing off the responsibility is
! aoth cisru ative and destructive to communication. The expectations of the clief examiner must be j definec/ communicated early.
i 14. T1e " time ine" must ae enhanced to l identify saecific times and dates for i communication / working meetings between
l l the examiner and the utility re3. These
{ meetings should be " face to face" to j assure expectations are understood, and being met, early on.
l 15. If an examiner and a utility rep have not l worked together before, the timeline for j "deliverab es" must be expanded. I
! Working meetings (face to face) must be
- established for the examiner to review l 5-10 questions,1 scenario,1 jpm,5 jam knowledge questions,5 admin questions,
- etc. to assure that the standards and l exaectations are clear early in the process
! anc tlat the utiity can aroduce a product
- that meets the expectation.
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- 16. 398 and 396 forms neec to be available
- electronica y. We took the time to i develoa an electronic version ourselves j but I wou d arefer that the e ectronic
! master copies came directly from the NRC
so that we have more conficence that everytling is exactly t1e same. We would all benefit from this im arovement.
- 17. =A face to face working meeting of eight (8) to twelve (12) hours, approximately two (2) weeks before the thirty (30) day submittal must be established to resolve any issues BEFORE the submittal. The exam materials should be reviewed, line by line, at this meeting to communicate all changes necessary.
- 18. Ex3ectations must be established early so that the utility clearly understands the rules and the examiners exaectations. Small samples of develoament must be reviewed early to assure expectations are 3eing met.
A face to face meeting, arior to the 30 day submitta , to reso ve any/all issues must be schedu ed such tlat adequate time (sucgest 2 weeks) is available to reso ve
i comments be" ore the 30 cay limit. No one wants to see 30 questions reviewec anc the exam rejectec. S3ending ~24 hours in 3-4 face to face meetings is a small/ smart l price to aay to avoid hundreds of lours of re-development, the emotional stress on cancidates when the exam must be rescheduled and the impact on the plant when candidates are not licensed to meet plant needs. l
- 19. The "new" SSNTA format for JPM level of !
cetail is NOT what you have liked in the 3ast arid neecs to be either acceated by the NRC as a standard or optimum format, or modified, or rejected. The JPMs we su amitted to you were in t7e "ormat and leve of cetai you had found acceataale in t7e past, and we were surarisec to finc that i t7ey needed significant last-minute rework (additiona level o" cetai).
- 20. Maybe Chief Examiners could senc some copies of good written questions, JPMs, and JPM questions up front that could help ;
a new developer survive the exam writing l process and see where you as an examiner are coming from.
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l Facility General Comments l
- 1. T1e limited number of NRC license examiners puts the Region and t1e sites at a cisacvantage with resaect to getting l
timely interchange. If the examiner is out of the o"fice on a tria for several weeks, the time you have to provide t7e licensee with i "eecbac< is very limitec anc results in a l real struggle to ensure a quality exam.
T1e limited resources and interaction time increases t1e risk o" lower c uality.
l 2. Region 11 examiners have been very
! arom at in getting bac< to us w1en we have l
a cuestion even t1ough you may 3e at a remote ocation.
- 3. The cua ity of the communications has 3een gooc. A of t1e examiners tlat I lave ta <ed to lave been thorough, arecise and lave aerformec listening clec<s to
verify tlat the correct messages were sent anc received. I would however, like to see more communications by e-mail where approariate. That would 1elp ensure the clarity of the communications even more.
- 4. With respect to the exam saecifically; there were a number of changes that were made and we had to transmit those by expensive overnig1t or next day delivery. If we coulc figure out a secure e-mail method it would save all of us numerous headaches as well as dollars.
- 5. During my "irst face-to-face meeting with the Chief Examiner (to review c raft exam material), I gainec much-neeced insight into his exaectations. T1is alleviated much stress on my part and, from tien on, the arocess went much more smoothly. T1e C1ief was very le aful curing subsecuent te ephone conversations anc our seconc
- j. l l l l
l meeting in Atlanta. He was very easy to
) work with and very uncerstancing
! concerning my inexperience in this i arocess. His patient guidance was the key !
l -to our success in t1is endeavor. Nex: time, I wit, a I we've learned, we'll do even better.
i l 6. The biggest problem that I encountered
! c uring that exam came from the written l portion tlat was aeing developec by the l contractor. Since he had written exams j 3efore, I assumed that the quality o" questions he was submitting to us were the c ua ity of questions that were acceptable to
-the NRC. We reviewed his work, made technical corrections and assumec that the questions woulc be acceated by the NRC.
I lac very litt e communication with the NRC on the subject of the written exam anc a c reat deal of communication on t7e subject of the 03erating exam. W1en t1e submittal was fina ly made, the focus went
to the written exam and most of t1e communication was mace over saea<er 31ones (about 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />). It became a very aainful process and coulc have been avoided hac I not out so much fait, in the contractor's exam writing exaerience and communicated more on the subject of the written exam with the clief. We had a 100% aass rate on t7e exam, but t1e exam reaort was arutal in the area of t1e written exam.
- 7. Know the chief examiners expectations from the beginning (prior to any cevelo ament).
- 8. Never assume you know what you're coing
- the chief is just a ahone cal away.
- 9. It's better to deliver material anc review it in oerson rather than over the alone. I alan for four or five trias between t7e start of
develoament anc arep week. It may sound ike over<i , but it wor <s (It's a so l safer in the area of security). l 10.'Su omit materia ear y (especial y the written). I like to have the written exam a done deal arior to the actual submittal date.
- 11. You can never talk to the Chief Examiner ;
too much. During the process, I talk to him I more than I ta k to my mother.
- 12. The bottom line is that frequent communication, personal contact, and
'early submittal of materials is the key to a successful NRC exam.
- 13. ALL 3 rob ems / changes must be resolved at the level of tie examiner anc the utility re 3. In no case s1ould aroolems/ chances be reported / escalated to senior management of the utility or NRC unless
i j both the examiner and t1e utility rep are at j an abso ute, and mutually acreed, j impasse.
4 i
! 14. We often felt that we were working in t7e dar<, writing questions on toaics you may l not want (as we were waiting for comment l on our skyscrapers), aossibly wasting
! resources, but seeing no other option to j meet our required cast-in-stone deadlines.
l 15. It's difficult to keep JPMs short and i alausible at the same time.
l 16. What is a good " admin JPM", especially for j ROs? l i
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The unit is operating at 20% power with all systems in automatic. Bank 'D' control rods
- are at 120 steps. Control Bank C' rod H6 drops to the bottom of the core. No rod control urgent failure alarms occur.
Where will thermal power and RCS Tavg stabilize in response to the dropped rod without any operator action?
A. Reactor thermal power will be lower than prior to the dropped rod; RCS Tavg will be more than 5'F lower than the temperature prior to the dropped rod.
B. Reactor thermal power will be lower than prior to the dropped rod; RCS Tavg will be within l'F of the temperature prior to the dropped rod.
C. Reactor thermal power will be the same as prior to the dropped rod; RCS Tavg will be within l'F of the temperature prior to the dropped rod.
! D. Reactor thermal power will be the same as prior to the dropped rod; RCS Tavg l will be more than 5'F lower than the temperature prior to the dropped rod.
1 .
l Answer:
C Reactor thermal power will be the same as prior to the dropped rod; RCS Tavg will be within l'F of the temperature prior to the dropped rod.
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, Reference Page SRO Question 10 RO Question 10 SRO Tier / Group -
111 RO Tier / Group 112 SRO Importance 3.7 RO Importance 3.2 10CFR55.43(b) 10CFR55.41 8 fremAddressed Item Addressed KA Number 000003AK1.01 KA Statement Knowledge of the operational implications of the following concepts as they apply to Dropped Control Rod: Reason for turbine following reactor on dropped rod event
~
SHNPP Objective AOP-LP-3.1-2 RECOGNIZE automatic actions that are associated with AOP-001, Malfunction of Rod Control and Indication Systems References AOP-LP-3.1 AOP-001 AOP-001 Malfunction of Rod Control and Indication Systems SD-104 Rod Control System Question Source New Justification (A) Select if he does not recognize that rods will step out to restore temperature and power was restored due to the previous decrease in temperature.
(B) Select if he recognizes that rods will step out, but the decreased temperature adds positive reactivity to restore power.
(C) CORRECT - Power will initially decrease du'e to the dropped rod. As power decreases, temperature will decrease. As temperature decreases, positive reactivity is added to restore power. Bank D rods in auto will cause rods to step out. Rods stepping out will restore power and temperature to the original value.
(D) Select if he recognizes that power was restored due to the previous decrease in temperature, but does not recognize that rods will step out to restore temperature and W
e
INITIAL EXAMINATIONS QUESTIONS AND ANSWERS TRAINING MANAGERS CONFERENCE NOVEMBER 5,1998 RICK BALDWIN GEORGE HOPPER I
Initial Written Examinations
Reference:
All written examinations are written IAW ES-401, " Preparation of Site-Specific Written Examinations for Power
! Reactors." Using ES-401-1, ES-401-2, ES-401-3, ES-401-4, BWR/PWR, RO/SRO EXAMINATION OUTLINES, and ES-401-6 Written Examination Quality Assurance Checkoff Sheet."
OBJECTIVES .
O BETTER EXAMINATION PRODUCT e LESS NRC/ FACILITY REWORK 9 SHARED EXPECTATIONS t
i 3
t
l SESSION OBJECTIVE:
To review validity concepts affecting the NRC -
t i
written examination for the purpose of:
instructing licensee personnel toward I
construction of more VALID and CONSISTENT NRC license examinations. l t
4
COVERAGE O 3 Levels of Validity i
G 3 Levels of Knowledge l i
l 9 Discrimination, Sampling l G Psychometrics .
i 5
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[
VALIDITY -
A valid test is one which tests what it intends to test.
In training examinations, testing specific skills and knowledge outlined and taught in
! the objectives.
In licensing examinations, testing specific skills and knowledge that SHOULD have l been outlined in the objectives.
6
.-. i 3 LEVELS OF VALIDITY p
G Content
~
I 9 Operational ;
+
3 i
G Discriminant i
4 i
7
CONTENTVALIDITY i
Addresses K/A coverage and sampling plan coverage.
_________-______-_--____-_-__-__________-____________-____________-----_a
-r OPERATIONAL VALIDITY '
Addresses two aspects:
- 1. Is the test item important to be known as a part of the operator's job?
- 2. Does the test item require the candidate to perform a job RELATED mental or l
physical operation?
9
_ __-_______ _________ _________________ ____ - _ _ _ _ _ _ __ a
DISCRIMINANT VALIDITY Addresses:
9 The cut score is the performance level that we use for making a pass / fail decision 80 percent.
O The exam must be written at a level of difficulty that intends to discriminate at the 80 percent level.
9 The question, its stem and distractor, interplay, by DESIGN, at least 80 percent of the candidates taking the exam should answer the item correctly. '
m
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VALIDITY
SUMMARY
- 1. The exam must be content valid, encompassing job safety significance and sampling. .
- 2. The test item should be operationally oriented: a expected mental or psychomotor requirement of the job. The items should be written at the comprehension or analysis level vice simple memory. Items that measure problem solving, prediction, analysis which are essential to job performance.
11
VALIDITY
SUMMARY
- 3. The exam must discriminate at a moderate level of difficulty, set by the cut score.
Meaning the test items as written should provide opportunity for at least 80 percent of the candidates taking the test should answer the item correctly.
12
3 LEVELS OF KNOWLEDGE Bloom's Taxonomy e Analysis, Applibation, Synthesis O Comprehension 9 Fundamental (simple memory) k 13
LEVEL OF KNOWLEDGE G Bloom's Taxonomy, NRC Reference Benchmark to classify levels of knowledge. .
O Bloom's Taxonomy, a classification scheme that classifies items by depth of mental performance required to answer the items.
O Bloom's Taxonomy, can be applied to written, scenarios or JPM questions.
m
LEVELS ;
LEVEL 1 l
Fundamental, using simple mental processes, recall or recognition of discrete bits of information. .
i.e. setpo.in ts, definitions, or specific facts.
15
____-___. _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ - _ _ - _ _ _ -a
l LEVEL 2 Comprehension, involves understanding
- material through relating it to its own parts or other material
i.e. including rephrasing information in different words, recognizing relationships
, including consequences or implications.
16 t
-l LEVEL 3 -
Analysis, synthesis, and application testing is more active and product-oriented testing
! which involves the multi-part mental process ,
of assembling, sorting, or integrating the parts so that the whole, and the sum can be used to: predict and event or outcome, solve a problem or create something new.
i.e. using knowledge to solve problems.
t 17 l
L t
- DETERMINANTS OF DISCRIMINATION
G Level of examination knowledge O Level of examination difficulty i
9 Passing Score i
9 Item bank use
. i 18 i
I
l NATURE OF EXAMINATIONS AND TESTS O TESTS are samples of PERFORMANCE 9 infer overall performance based on a sample 9 Sample must be broad-based to make confident inference O Sample must NOT be fully predictable or inferences cannot be made on untested areas.
O ltems MUST discriminate otherwise it has little or NO value. ,
19
PSYCHOMETRICS .
Items may have one or more of the following psychometric errors:
- 1. Low level of kngwiedge (fundamental)
- 2. Low operational validity (not job related)
- 3. Low discriminatory validity ( hard or easy)
- 4. Implausible distractors
- 5. Confusing language or ambiguous questions
- 6. Confusing or inappropriate negatives
- 7. Collection of true/ false statements
- 8. Backwards logic 20
]
006 Emergency Core. Cooling System-/ JPM 236 Recovery From Safetylnjection
.and Solid Water Conditions'. '- ;- '- '
Question 2:
Given the following plant conditions:
Unit 2 was operating at 100% power.
The plant experienced a large break LOCA with a failure of the ECCS system.
FR-C.1, " Response to Inadequate Core Cooling," is being implemented.
Core exit TCs are 720*F and increasing.
At this point FR-C.1 directs the crew to depressurize intact steam generators.
a.) What is the basis for the direction in FR-C.] to depressurize intact steam generators?
b.) Why is this action taken?
References Allowed? YES X NO_
Answer:
a'.') To reduce RCS pressure below 125 psig b.) To allow the ECCS accumulators and RHR pumps to inject water to the RCS.
Reference:
KA: 006G4.1S [ 2.7 / 3.6 ) Knowledge of specific bases for EOPs.
OPL271C398 pg 12-15 Applicant Response: SAT UNSAT l
- 1 j
INADEQUATE CORE COOLING Rev. 8 STEP ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED l
NOTE . Blocking low steamline pressure Si as soon as pressurizer pressure is less than 1920 psig will prevent an inadvertent MSIV closure and keep the condenser available for steam dump.
1
. After the low steamline pressure SI signalis blocked, main steamline isolation will occur if the high steam pressure rate is exceeded. r
. S/G depressurization at the maximum rate may cause S/G narrow range levels to drop to less than 10% {25% ADV). This is acceptable and expected for this inadequate core cooling condition.
- 14. DEPRESSURIZE Intact S/Gs to reduce RCS pressure to less than 125 psig:
- a. WHEN RCS pressure less than 1920 psig, THEN PERFORM the following:
- 1) BLOCK low steamline pressure St. 7
- 2) CHECK STEAMLINE PRESS ISOUSI BLOCK RATE ISOL ENASLE permissive LIT.
IM-4A, A4]
- b. DUMP steam to condenser b. DUMP steam at maximum rate at maximum rate. USING intact S/G atmospheric relie1(s).
IF local control of atmospheric relie1(s) is necessary, THEN j
DISPATCH personnelto dump steam i USING E A-1-2, Local Control of
~
S/G PORVs.
(Step cont:nued en next oage.)
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.026,Conraintnent Spray System /.JPM.# 57AP Respond to High Contamment
. Pressure 4PJace'RHR Spmy.in Service'! - - -
Question 2:
Given the following plant conditions:
Unit I has tripped from 100% power due to a LOCA.
Containment pressure is 3.0 psid Transfer of Containment Spray pump suction to the containment sump is being perfonned in accordance with ES-1.3, Transfer to RHR Containment Sump. ,
a.) Why must both CS pumps be placed in PULL-TO Lock while transferring suction to the
! containment sump?
l b.) What does placing both CS pumps in PULL-TO-Lock prevent?
References Allowed? YES X NO l I Answer:
a.) While shifting to the containment sump, both the RWST and the containment sump suction valves to the CS pumps will be closed at the same time. !
b.) Placing the CS pumps in PULL-TO-Lock will prevent running a CS pump without a source of l l water.
Reference:
K/A: 026G4.18 [ 2.7 / 3.6 ] Knowledge of specific bases for EOPs OPL271CO24 pg 14-1S CCD NO:1-47W611-72-1 ES-1.3,pages Il-13 OPL271C388 pg 9 Applicant Response: SAT UNSAT l
t I
e
License Applicant Administrative Walkthrough Examination--NRC-1
- Examiner Sheet A'.1: Shift Staffing '
Question 1: A licensed RO has been off-shift for 6 months teassist in scheduling an upcoming
, outage. He had his last physical examination 18 months ago and has had -
satisfactory performance in the licensed operator requalification training program.
He is informed that he is needed to join a shift crew in 3 days to fill in for a vacationing Unit OATC.
Can the RO fillin for the vacationing RO? Why or why not? w References Allowed? YES _.2L NO -
Answer: No. The RO must first reactivate his license by completing at least 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of "under direction" on-shift time. '
Reference:
SSP-12.1, Conduct of Operations, p. 61 and 62 K/A 2.1.4 (2.3/3.4)
Applicant Response: SAT UNSAT e
--, , ,m .-
License Applicant Administrative Walkthrough Examination--NRC-1 Examiner Sheet A.1:' Shift Staffing Question 1: A licensed RO has been off-shift for 6 months to assist in outage scheduling. He is informed that he is needed to join a shift crew in 2 weeks to fill in for a vacationing shift RO. He had a satisfactory physical examination 18 months ago and has maintained satisfactory performance in the licensed operator requalification training program. In anticipation of rejoining a shift crew, the RO has spent two 8-hour shifts in the control room as the OATC during the past month under the direction of a shift RO.
What additional requirements must be met by the RO before he may fill in for the vacationing RO?
References Allowed? YES J'_ NO Answer: The RO must complete an additional 24 (40 - 16) hours of"under direction" on-shift time that must include a plant tour and shift turnover.
Reference:
SSP-12.1, Conduct of Operations, p. 61 and 62 K/A 2.1.4 (2.3/3.4)
Applicant Response: SAT UNSAT l
l Given the following conditions:
- 1. The reactor has experienced a Steam Generator Tube Rupture.
- 2. All systems responded as expected.
. 2. The performance of E0P-04 is in progress.
- 3. One Steam Generator has been isolated.
- 4. All RCPs have been stopped.
I 5. RCS cooldown using natural circulation is in progress.
Which One of the following describes the concern associated with the isolated
.SG pressure prior to placing the RCS on SDC?
- a. The isolated SG pressure would be too low due to excessive cooldown causing RCS water to enter the SG and reducing RCS inventory. ,
- b. The strategy during the performance of E0P-04 is to maintain the affected SG pressure slightly less that RCS pressure to prevent secondary water entering the RCS.
- c. Since the RCS cooldown rate was maintained greater than 30deg/hr. the affected loop has not been cooled sufficiently to allow SG depressurization.
- d. The affected SG pressure is high due to thermal stratification of the secondary water.
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/
Reactor Operator Examination
(
I
- 65. Given the following conditions:
The reactor has expenenced a Steam Generator Tube Rupture.
All systems responded as expected.
The performance ofEOP-04 is in progress. l One steam generator has been isolated. l RCS cooldown using natural circulation is in progress. l Which ONE of the following describes the concem associated with the affected SG pressure prior to placing the RCS on SDC? j l
- a. The SG pressure would be too low due to u cessive cooldown causing RCS water to enter the SG and reducing RCS inventory.
- b. The SG pressure would be slightly less than RCS pressure causing water to enter the RCS resulting in a dilution.
1
- d. The SG pressure would be too high due to thermal straification of the secondary water.
4 i
i e
Which one of the following describes the response of the Unit I charging pumps following receipt of an automatic SIAS signal, coincident with a Loss of Offsite Power?
- a. One charging pump is automatically started on each emergency bus 5 minutes after it is energized by the diesel.
- b. All charging pumps are automatically started immediately after their respective bus is energized.
- c. The operator must manually start one charging pump on each emergency bus 5 minutes after it is energized by the diesel.
- d. One charging pump is automatically started onto each emergency bus, immediately after it is energized by the diesel. ~
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- 19. Which ONE of the following describes the response of the Unit-1 charging pumps following receipt of an automatic SIAS signal, coincident with a Loss of Offsite Power? Assume normal electrical lineup and all equipment is operable.
- a. Only one charging pump is automatically started on each emergency bus 5 minutes after it is energized bythe diese!.
- b. All charging pumps are automatically started immediately after their respective bus is energized by the diesel.
- c. All charging pumps are automatically started 5 minutes after their respective buses are energized by the diesel.
- d. Only one charging pump is automatically started onto each emergency bus immediately after it is energized bythe diesel.
/
Charging pumps are running on Unit 1 an SIAS is present. (Assume no operator action)
Which one of the following lists the charging pump response when the BAM t6nks
. are emptied?
The charging pumps will;
- a. trip on low oil pressure.
- b. trip on low suction pressure.
- c. automatically align to the RWT.
- d. continue to run and become gas bound.
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( Reactor Operator Exa.~ir.a: ion l
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- 27. Charging pumps are runnmg on Unit I and an SIAS is present. (Assume no operator action)
Which ONE of the following lists the charging pump response when the BAM tanks are emptied?
l The charging pumps will:
- a. trip on thermal overload.
b., trip on low su:: tion pressure.
l
- c. automatically align to the RWT.
- d. continue to run and become gas bound. '
i I
Given the following conditions:
Unit 1 CEDM fan HVE-21A is in AUTO after START l Unit 1 CEDM fan HVE-21B is in AUTO after STOP.
l Unit 1 CEDM fan HVE-21A trips on overcurrent.
Which ONE of the following completely lists the logic that will start HVE-21B?
- a. The trip signal from HVE-21A.
- b. The trip signal from HVE-21A concurrent with a low flow signal.
- c. A low flow signal.
- d. The trip signal from HVE-21A concurrent with a low flow signal an6 air inlet temperature to the cooling coils is greater than 100 deg F.
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4 Reactor Operator Examina: ion
(
l l'- 59. Given the following conditions:
Unit 1 CEDM fan HVE-21 A is in AUTO after START.
l Unit 1 CEDM fan HVE-21B is in AUTO after STOP.
- Unit 1 CEDM fan HVE-21 A trips on overcurrent. .
Which ONE of the following lists the signals required by the logie needed to start HVE-21B?
a.- The trip signal from HVE-21 A.
I
- b. The trip signal from HVE-21 A concurrent with a low flow signal.
- c. A low flow signal.
l d. The trip signal from HVE-21 A concurrent with a low flow signal and air inlet temperature signal to the cooling coils is greater than 100 F.
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- - - , . w ,-.- . . - , , , , . . , - - , + - - - , - . - .
- 13. Given the following plant conditions:
. Unit 1 was at 73% power
. A reactor trip / safety injection on low steam line pressure occurred 21 minutes ago
. Average Core Exit TC temperature is 375'F
. RCS pressure is 225 psig
. All S/G pressures are DECREASING slowly
. #2 and #3 S/G ievels are 5% NR and DECREASING slowly
. #1 S/G levelis 6% NR, and INCREASING slowly
. #4 S/G levelis STEADY at 2% NR
. Total feedwater flow is 340 gpm
. PZR levelis 37% and INCREASING
. RCS T-cold temperature is 325'F and DECREASING slowly
. Containment pressure is 5 psid and INCREASING slowly ,
At this po nt, which DNE of the following Critical Safety Functions is the MOST degraded?
- a. Heat Sink
- b. Core Cooling
- c. Containment
- d. Pressurized Thermal Shock Answer. A K/A: 000040K101 [4.1/4.4]
Reference:
E-0, Foldout Page Objective: OPL271C395, B.1 Level: Analysis Source: 000040K101 001 History: Stem and distracters a and d modified (7/7/93)
Note: Provide PTS curve with this question.
Justification:
- a. Correct answer because all S/G levels are Jess than 10% NR and total feeowater flow is less than 440 gpm.
- b. Incorrect because RCS temperature is 325'F (core exit T/Cs less than 1200*F). l
- c. Incorrect because containment pressure is less than 12.0 psid.
d, incorrect because RCS temoerature is 325'F (T-cold is greater tnan 250*F) 1 l
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- 13. Given tne following plant tenditions:
. Unit 1 was at 73% power A reactor trip / safety injection on low steam line pressure occurred 21 minutes aco
~
. Averace Core Exit TC temperature is.37pF
~
. RCS pressure is 225'psig ISSP 7# i
. All S/G pressures are DECREASING slowly
- 2 and #3 S/G levels are 5% NR and DECREASING slowly
. *1 S/G levelis 5% NR, and INCREASING slowly i
. #4 S/G levelis STEADY at 2% NR l
. Totalfeedwatepflow is 340 gpm PZR levelis)7% and INCREASING
. RCS T-cold temperature is 325'F and DECREASING slowly
. Containment pressure is 5 psid and INCREASING slowly , I At inis poin . which ONE of the following Cntical Safety Functions is the MOST degraded?
- a. Heat Sink
- b. Core Cooling
. Containment
- d. Pressurized Thermal Shock Answer: A K/A: 000040K101 (4.1/4.4J
Reference:
E-0, Foldout Page Objective: OPL271C39,5, B.1 7
Level: Analysis ,
a f
Source: E 000040K101 001 History: Stem and distracters a and d modified (7/7/9S)
Note: Provide PTS curve with this question.
Just:ii:ation:
- a. Cone:t answer because at SIG 4evels are Jess tnan 10% NR and totalieeowater fiow is less than e-40 gpm.
- b. Incorrect because RCS temperature is 325'F (: ore eot T/Cs less inan 1200*:)
- . Incorre: De:ause containment pressure is fess inan 2.0 psid.
l 1
i d incorrect oe:ause RCS temperature is 325? (~-50:d is greater inan 250*:)
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- 20. Given the following plant conditions:
. The control room has been evacuated due to a fire
. All controis have been transferred per AOP-C 04
. MDAFW pumps 1 A-A and 1B-B are injecting into the steam generators
. The TDAFW pump has been shut down i
. Steam generator pressures and levels are decreasing Which ONE of the following describes the response of the auxiliary feedwater system?
- a. The TDAFW pump will automatically restart when 2/4 steam generators reach low low level.
- b. The MDAFW pump level control valves will automatically control steam generatof
~
levels at 33%.
- c. The MDAFW pump level control valves will have to be manually adjusted using the Manual Output Adjust in the L-381 cabir'et.
- d. The discharge pressure for the MDAFW pumps will have to be manually adjusted by throttling the manual valves at the l CVs.
Answer: B K/A: 000068A102 [4.3 / 4.5]
Reference:
AOP-C.04, page 11 Objective: OPL271C423, B.4 Level: Comprehension Source: 000068A102 001 1
History: Used on 9/97 RO NRC exam l Text modrfied to correct grammar errors. Distracters a, b, c, and d reordered (7/22/98). Distracter be restructured (7/29/98)
Note: Selected fromM exam bank with minor modification of text
p h
- 20. Given the following p: ant condr:: ens:
)
. The control room has been evacuated due to a fire
. All controls have been transferred per AOP-C.04
. MDAFW pumps 1 A-A and 1B-B are injecting into the steam generators
. The TDAFW pump has been shut down
. Steam generator pressures and levels are decreasing
@ r d, wh M M k oh@ >M /r4 -C-@ 3 Which ONE of tne fo!!owing describes te retocash: Wr.: o ,aciy lc;d ater. system?
~
- a. The TDAFW pump will automatically resta-t when 2/4 steam generators reach low low level. e ~ d k aa" OviIN 03'y e ( -M/ .
(C o w
- b. The MDAFW pump level control valves will automatcally control steam generator ,
, levels at 33%.
The MDAFW pump level =entrol valves will have to be manually aojusted using tne Manual Output Adjust in tne L-381 cautnet.
fW d. The discharge pressure for the MDAFW pumps will have to be manually adjusted by
,,,I y g [ tnrottling the manual valves at the LCVs.
., Al M
, f[j, Answer: B K/A: 00005BA102 (4.3 / 4.5]
Reference. AOP-C.04, page 11 Objective: OPL271C423, B.4 Level: Comprehension Source: 00005BA102 001 H: story: Used on 9/97 RO NRC exam Tex't modtfied to correct grammar errors. Distra~.ers a b, c. and d reordered (7/22/98). Distracter be restructured (7/29/95)
Note: Selected fromMexam bank with minor modification of text
~
l l
+
- 23. Given the following plant conditions:
. FR C.1,' inadequate Core Cooling", has been entered due to a RED path on Core Cooling
. Core exit temperatures (TCs) are 1250*F and increasing
. NO Feedwater / Aux Feedwater is available l
. At step 12, the CRO checks the S/G NR levels and reports all are <10%.
As the SRO you should; (Select ONE of the following)
- a. Go to FR-H.1," Loss of Secondary Heat Sink".
- b. Depressurize allintact S/Gs to atmospheric pressure to dump accumulators.
Answer: C K/A: 000074K307 .[4.0/4.4]
Reference:
FR.C.1, pages 10 & 17 Objective: OPL271C398 Level: Comprehension Source: MExam Bank 101. 000074K307 001 History: Used on HLC 9807 practice exam Distracters b and c reordered Note: Selected fro 6 exam bank without modification of text l
I 1
1 1
I I
l
4 g INADEQUATE CORE COOLING FR-C.1 7q Rev. 8 '
I
! STEP ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED i-CAUTION Use of a Faufted or Ruptured S/G during performance of the following steps l may compound the emergency situation. When NO Intact S/Gs are available, l
a Fautted or Ruptured S/G may be used. '
- 12. MAINTAIN intact S/G narrow range r l levels:
- a. Greater than 10% {25% ADV) a. MAINTAIN total feed flow greater tnan 440 gpm UNTIL level greater tnan 10% {25% ADV) in at least one S/G.
IF total feed flow greater than 440 gpm can NOT be established, THEN PERFORM theiollowing:
- 1) CONTINUE attempts to establish heat sink in at least one S/G.
, 2) GO TO Note prior to Step 21.
- b. Between 10% {25% ADV) and 50%.
l l
l
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\
Page 10 of 19
___ -- - .~. .
1 FR-C.1 w
g INADEQUATE CORE COOLING Rev. 8 STEP ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED l.
NOTE RCP damage due to absence or loss of normal support conditions is an acceptable consequence in this procedure.
l 21. CHECK if RCPs should be started:
- a. , CHECK core exh T/Cs a. GO TO Step 22. l greater than 1200*F. l 1
- b. CHECK if idle RCS toop available: b. PERFORM the following:
l l 1) S/G narrow range level a) OPEN pressunzer PORVs and block greater than 10% [25% ADV] valves.
- 2) RCP in associated loop b) IF core exTt T/Cs remain i AVAILABLE AND STOPPED. greater than 1200*F, THEN OPEN reactor vessel head vents:
, . FSV-66-394
. FSV-68-395 )
. FSV-68-396 ;
. FSV-68-397.
c) GO TO Step 22.
l
- c. START RCP in one idle loop.
- d. GO TO Substep 21.a.
l -
Page 17 of 19
e
- \
- 23. Given the following plant conddions:
- FR-C.1, inade:;uate Core Cooling", has been entered due to a RED path on Core l Cooling
. Core exit temperatures (TCs) are 1250*F and increasing
. NO Feedwater / Aux Feedwater is available
. At step 12, the CRO checks the S/G NR levels and reports all are <10%.
As the SRO you should: (Select ONE of the following)
L
- a. Go to FR-H.1,
- Loss of Secondary Heat Sink".
- b. Depressunze allintact S/Gs to atmospheric pressure to dump accumulators.
~
g PoR\l.s u ~34 o d \/M.wS
- d. .JheparWaNCO recJ oud Geod4WR-4eveHrr eny-2-S/Gs4s4ess-thart-608/, -
Answer: C, K/A: 000074K307 [4.0/4.4]
Reference:
FR-C.1, pages 10 & 17 Objective: OPL271C398 Level: Comprehension Source: 000074K307 001 ,
History: Used on HLC 9807 practice exam Distracters b and c reordered Note: Selected from Mexam bank without modification of text
.p.h yW.6!</E/$ f
- W
- 49. Given tne following plant conditions:
. Reactor power is at 20% dunng a unit shutdown
. Intermediate Range N-36 failed high
. Operators placed the level trip bypass switch for N-25 to the bypass position Which ONE of the following describes the effect of this failure and action during the remainder of the shutdown?
- a. The reactor will automatically trip when the Power Range channels decrease below the P-10 setpoint,
- b. Entry from Mode 1 to Mode 2 is prohibited with an inoperable intermediate Range channel, so the unit must be manually tripped prior to Mode 2 entry.
- c. .Both Source Range channels, N-31 and N-32, must be manually energized when the operable intermediate Range channel (N 35) decreases below the P-6 setpoint.
- d. Source Range channel N-32 must be manually energized when the operable Intermediate Range channel (N-35) decreases below the P-6 setpoint; Source Range channel N-31 will automatically energize.
Answer C K/A: 015000K407 [3.7/3.B]
Reference:
AOP-1.01, page 10 ES-0.1, page 13 Objective: OPL271 C352, B.4 Level: Comprehension Source: 015000K407 001 History: Not used on 9/97 or 5/98 NRC exams. Not used on practice exam.
Distracters c and d reordered Note: Selected fromg exam bank without modification of text
a 0 - - ,
^
ph g/#E2- / WWM W5 Q~/ M ., I i ! ;
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- 49. Given the following plant conditions: ,
(
. Reactor power is at 20% during a unit shutdown \
. Intermediate Range N-35 failed high _\ i Gpera crs pied m: ::sPeypassswiterriorth351:rthirDypasspesitiE J l Which ONE of the following describes the effect of this failure and action during the remainder of l the shutdown? I
- a. The reactor will automatically trip when the Power Range channels decrease below the F-10 setpoint.
1
- b. Entry from Mode 1 to Mode 2 is prohibited with an inoperable Intermediate Range channel, so the unrt must be manually tripped prior to Mode 2 entry.
- c. .Both Source Range channels, N-31 and N-32, must be manually energized when the operabie Intermediate Range channel (N-35) decreases below the P-5 setpoint.
- d. Source Range channel N-32 must be manually energized when tne operable intermed: ate Range channel (N-35) decreases below the P-5 setpoint; Source Range channel N-31 will automatically energize.
Answer. C K/A: 015000K407 [3.7/3.8)
Reference:
AOP-Lol, page 10 ES-0.1, page 13 Objective: OPL271 C352. B.4 Level: Comprehension Source: 010000K407 001 History: Not used on 9/97 or 5/98 NRC exams. Not used on practice exam.
Distra:ters c and d reordered Note: Selected fromMexam bank without modification of text g J .5 M '
~/
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5 i
- 35. Given the following plant conditions:
. . Unit 2 is operating at 29% power in accordance with 0-GO-6, Power Reduction From !
30% Reactor Power to Hot Standby l
-.- . Unit 2 will be going to Cold Shutdown for maintenance l
. Intermediate Range N-36 has just failed high ;
l l
Which ONE of the following actions must be performed before reducing reactor power below
'l 10%?
. a. Manually energize N-31 and N-32 I
L b. Place N-36 LevelTrip switch in BYPASS.
- c. Remove N-36 instrument power fuses. t
- d. Manually trip the reactor to prevent an automatic reactor trip.
Answer; B i
K/A: 000033K302 [3.6/3.9) ;
Reference:
AOP-l.01, page 10 & 13 i
Objective: OPL271C352, B.4 Level: Analysis i Source: New question (Developed 7/15/98) ,,
7 Justification: i i
I i
- a. Incorrect because manually restoring N-31 and N-32 to operation in the power range would L destroy the source rarige detectors. j
- b. Correct because placing the level trip switch in BYPASS prevents high reactor trip when i the low power reactor trip signal is reinstated at the F-10 setpoint (10% power).
- c. Incorrect because action does not bypass the trip signal.- f
- d. Incorrect because a manual reactor trip for the given conditions is not required. Placing N-36 level trip switch in BYPASS allows an orderly reactor shutdown. l l
4 y .~, _ - . .
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- 50. Given the following piant conditions:
. Large Break LOCA is in progress
. RCS pressure is 550 psig
. Exosensor indicates 25'F superheat
. No RCPs are operating Which ONE of the following indications would the operator use along with RCS pressure to accurately substantiate core cooling?
- a. Reactor Coolant Tavg value.
- b. Average value of all core exit thermocouples.
- c. Hottest Reactor Coolant wide range That value.
- d. Average value of five hottest core exit thermocouples.
Answer: D l K/A 017000A402 [3.8 / 4.1]
Reference:
FR-0, page 3 OPL271C044, page 7, A.1.c Objective: OPL271 C044, B.1.b Level: Memory Source: 017000A402 001 History: Used on HLC 9809 practice exam Distracters a, b, c, and d reordered (7/22/98)
Note: Selected from xam bank without modification of text 1
i
l l
1 I
- 50. Given tne following plant condmons:
1 1
- Large Break LOCA is in progress
. RCS pressure is 550 psig !
. Exosensor indicates 25'F superheat l
. No RCPs are operating Which DNE of the following indications would the operator use along wrth RCS pressure to accuratefy substantiate core cooling?. , ,,
u_ . - -- _ - ,_-. , - .
~,a.9 Y 9ead5 C %en^f 5;; A!NTN ^FA ~4,k.SQ, , , ,,,) WM' #
, t Tc. Abtr rote c.7 ~s -
- b. Average value of all core exit thermocouples,
- c. Hottest Reactor Coolant wide range Thot value. -
i a
- d. Average value of five hottest core exrt thermocouples.
Answer. D s j
N5 " -I 0;^ M#
K/A 017000A402 [3.8 / 4.1)
Reference:
FR-0, page 3
- 6 2-OPL271C044, page 7 A.1.c Objective: OPL271 C044, B.1.b Level: Memory Source: 017000A402 001 History: Used on HLC 9809 practice exam Distracters a, b, c, and d reordered (7/22/98)
Note: Selected from exam bank without modification of tert l
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- 88. Given the following plant conditions:
. Unit 2 operating in accordance with 0-GO-5 Normal Power Operation at 73% with a powerincrease to 100% in progress ;
.. . Chemistry reports Unit 2 RCS loop 1 accumulater boron concentration is 2390 ppm
. Current time is 0100 I Which ONE of the following actions must be taken? [
t !
I~ a. Immediately stop the power increase. ,
i
- b. Continue the power increase while restoring loop 1 accumulator boron concentration to 2400 to 2700 ppm boron within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
- c. If loop 1 accumulator boron concentration is NOT restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, be in HOT:
l' < STANDBY by 0700. ;
- d. If loop 1 accumulator boron concentration is NOT restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, reduce pressurizer pressure to 1000 psig or less by 1300.
Answer: A i K/A: 2.1.1 [3.7/3.8) i
Reference:
SSP 12.1, Page 31 ;
Objective: OPL271C209. B.2 ,
Level: Comprehension
- Source: New question (Developed 7/20/98) .
f Note: Provide copy of Technical Specification 3.5.1.1 with the question (exam) 4 Justification:
- a. Correct becausegConduct of Operation (SSP-12.1) restricts power increase when in an LCO action of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or less. RCS loop 1 accumulator boron concentration cf 2390 ppm boron places Unit 1 in a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> LCO.
- b. Incorrect because power increase is not allowed when in a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> LCO action statement. '
c.- incorrect because if loop 1 boron concentration is not restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, the Unit 1 must be in HOT standby by 0800.
- d. Incorrect because if loop 1 boron concentration not restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, pressurizer pressure must be reduced to 1000 psig or less by 1400.
l
_ _ _ = _ _ _ _ _ _ - _ _ _ _ _ _ - _ - - _ - - . . - - - - - _ . . __
h 3 /4. 5 EMERGENCY CORE COOLING SYSTEMS j 3 /4. 5.1 ACCUMULATORS COLD LEG TNJECTION ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1.1 Each cold leg injection accumulator shall be OPERABLE with:
- a. The isolation valve open, A contained borated water volume of between 7615 and 8094 gallons of : a132 b.
barated water, l r
- c. Between 2400 and 2700 ppm of boron,
- d. A nitrogen cover-pressure of batween 600 and 683 psig, and RIS4
- e. Power removed from isolation valve when PC5 pressure is above 2000 psig.
APPLICABILITY: MODES 1, 2 and 3.*
ACTION:
- a. With one cold leg injection accumulator inoperable, except as a result of boron concentration not within limits, restore the inoperable accumulator to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure.to 1000 psig or less within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- b. With one cold leg injection accumulator inoperable due to the boron concentration not within limits, restore boron concentration to within fl.imits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to 1000 psig or less within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- Pressurizer pressure above 1000 psig.
l December 27, 1994 Amendment No. }!3, 131, 133
- UNIT 2 3/4 5-1
s o
- y .v
.j .) Of.. .S &ol Ifh)'
- 88. Given tne following pl ant conditions:
. Unit 2 operating in a:Cordance w:th 0-GO 5, Normal Power Operation at 73*4 with a powerincrease to 100% in progress
. Chemistry reports Unit 2 RCS toop 1 accumulator boron concentration is 23E0 opm
,g Current time is 0100 4 /
- Which ONS of the following actions must be taken?
- a. immediately stop the power increase. ,
- b. Continue the power increase while restoring loy o ,1 accumulator boron concentration to 2400 to 2700 ppm boron.C.1 un 4 Nw 72 4 ou /~$.,
- c. 11 loop 1 accumulator boron concentration is NOT restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, be in HOT STANDBY by 0700.
- d. If loop 1 a:Cumulator boron concentration is NOT restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, reduce pressurizer pressure to 1000 psig or less by 1300.
Answer: A K/A: 2.1.1 {3.7/3.8]
Reference:
SSP-12.1, Page 31 Objective: OPL271C209, B.2 Level: Comprehension Source: New question (Developed 7/20/98)
Note: Provide copy of Technical Spe:Hi ation 3.5.1.1 with tne question (exam)
.?
Just#ication:
- a. Correct becaus onduct of Operation (SSP-12.1) restricts power increase when in an LCO acaon or o nours or less. RCS foop 1 a::umulator coron con =emration of 2390 ppm boron places Unit 1 in a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> LCO.
- b. Incorrect because power increase is not allowed wnen in a 1 nour LCO action statement,
- c. in:ctre:t because if loop 1 boron concentration is not restored wrinin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, the Unit 1 l must be in HOT standby by 0800.
- d. in:orrect because if loop 1 boron concentration not restored wnnin 1 nour, oressurizer l
pressure must be reduced to 1000 osig or less by 1400.
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l 006 Emergency Core Ccoling System-/ JPM 136 Recovery From Safetylnjection
.and Solid Water Co'nilitions '. J .* '> ' '
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Question 2:
Given the following plant conditions:
Unit 2 was operating at 100% power.
The plant experienced a large break LOCA with a failure of the ECCS system.
FR-C.1, " Response to Inadequate Core Cooling," is being implemented.
Core exit TCs are 720*F and increasing.
l At this point FR-C.1 directs the crew to depressurize intact steam generators.
l a.) What is the basis for the direction in FR-C.] to depressurize intact steam generators? )
b.) Why is this action taken?
l References Allowed? YES X NO l
Answer:
a.) To reduce RCS pressure below 125 psig b.) To allow the ECCS accumulators and RHR pumps to inject water to the RCS.
Reference:
KA: 006G4.18 ( 2.7 / 3.6 ) Knowledge of specific bases for EOPs.
l OPL271C398 pg 12-15 i
Applicant Response: SAT UNSAT I
.7 i I l
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~
~
INADEQUATE CORE COOLING Rev. 8 STEP ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE . Blocking low steamline pressure Si as soon as pressurizer pressure is less than 1920 psig will prevent an inadvertent MSIV closure and keep the condenser available for steam dump.
. After the low steamline pressure SI signal is blocked, main steamline isolation will occur if the high steam pressure rate is exceeded. :
. S/G depressurization at the maximum rate may cause S/G narrow range levels to
~
drop to less than 10% [25% ADV). This is acceptable and expected for this inadequate core cooling condhion.
- 14. DEPRESSURIZE Intact S/Gs to reduce RCS pressure to less than 125 psig:
~
- a. WHEN RCS pressure less than 1920 psig, THEN PERFORM the following:
1
- 1) BLOCK low steamline pressure St. g i
- 2) CHECK STEAMLINE PRESS '
ISOUSl BLOCK RATE ISOL '
ENABLE permissive L1T.
[M-4A, A4]
- b. DUMP steam to condenser b. DUMP steam at maximum rate at maximum rate. USING Intact S/G atmospheric relief (s).
IF local control of atmospheric relief (s) is necessary, THEN DISPATCH personnel to dumo steam USING E A-1-2, Local Control of S/G PORVs.
(Steo continued on next page )
Page 13 of 19
i 026, Containment Spray System /,JPM # 57AP Respond to High Contamment Presstir'e4PJace'RHR Sphy.in Seivice'#: ~
Question 2:
Given the following plant conditions:
Unit I has tripped from 100% power due to a LOCA.
Containment pressure is 3.0 psid Transfer of Containment Spray pump suction to the containment sump is being performed in accordance with ES-1.3, Transfer to RHR Containment Sump.
a.) Why must both CS pumps be placed in PULL-TO-1 ock while transferring suction to the containment sump?
b.) What does placing both CS pumps in PULL-TO-Lock prevent?
References Allowed? YES X NO Answer:
a.) While shifting to the containment sump, both the RWST and the containment sump suction valves to the CS pumps will be closed at the same time, b.) Placing the CS pumps in PULL-TO-Lock will prevent running a CS pump without a source of water.
Reference:
K/A: 026G4.18 [ 2.7 / 3.6 ) Knowledge of specific bases for EOPs OPL271CO24 pg 14-18, CCD NO:1-47W611-72-1, ES-1.3, pages Il-13, OPL271C388 pg 9 Applicant Response: SAT UNSAT
License Applicant Administrative Walkthrough Examination--NRC-1
- Examiner Sheet l
K.1:? Shift StafHng 'e -
~
~'
l
(
Question 1: A licensed RO has been off-shift for 6 months to assist in scheduhng an upconun, outage. He had his last physical exarr.ination 18 months ago and has had i satisfactory performance in the licensed operator requalification training program.
l He is informed that he is needed to join a shift crew in 3 days to fill in for a i
vacationing Unit OATC. :
_ Can the RO fill in for the vacationing RO? Why or why not? % i References Allowed? YES.2L.NO -
Answer: No. The RO must first reactivate his license by completing at least 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of t "under direction" on-shift time.
t l
Reference:
SSP-12.1, Conduct of Operations, p. 61 and 62 K/A 2.1.4 (2.3/3.4) i Applicant Response: SAT UNSAT I
?
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f 4
i
, License Applicant Administrative Walkthrough Examination--NRC-1 j Examiner Sheet l
A.1: Shift Staffing'- a t
Question 1: A licensed RO has been off-shiR for 6 months to assist in outage scheduling. He is '
infoimed that he is needed to join a shift crew in 2 weeks to fill in for a vacationing
, shift RO. He had a satisfactory physical examination 18 months ago and has i
maintained satisfactory performance in the licensed operator requalification training I program. In anticipation of rejoining a shin crew, the RO has spent two 8-hour j
shiRs in the control room as the OATC during the past month under the direction of a shin RO..
s What additional requirements must be met by the RO before he may fill in for the !
i r vacationing RO? l J
4
- References Allowed? YES _2L NO Answer
- The RO must complete an additional 24 (40 - 16) hours of"under direction" on-shift time that must include a plant tour and shift turnover.
Reference:
SSP-12.1, Conduct of Operations, p. 61 and 62 K/A 2.1.4 (2.3/3.4)
Applicant Response: SAT UNSAT
? \
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