IR 05000456/1998305
| ML20155E778 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 11/02/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20155E739 | List: |
| References | |
| 50-456-98-305, 50-457-98-305, NUDOCS 9811050177 | |
| Download: ML20155E778 (120) | |
Text
{{#Wiki_filter:__ . . . . __
,- q l .
l
'
l U.S. NUCLEAR REGULATORY COMMISSION REGION 111 ,
Docket Nos: 50-456;50-457 License Nos: DPR-29; DPR-30 Report Nos: 50-456/98305(OL); 50-457/98305(OL) Licensee: Commonwealth Edison Company (Comed) Facility: Braidwood Nuclear Power Station, Units 1 and 2 i Location: RR# 1, Box 79 Braceville, IL 60407 Dates: September 14,1998 Telephoned Examination Results October 20,1998 Inspectors: H. Peterson, Chief Examiner, Rill Approved by: M. Leach, Chief, Operator Licensing Branch Division of Reactor Safety l 9811050177 981102 ~" PDR ADOCK 05000456 V pm
. - . , ~ .- - . - . . . . _ _ _ . ~ - . - - . . - . - _ - ~ . . - . . . - . - - . -. _-. ,
l EXECUTIVE SUMMARY Braidwood Nuclear Generating Station NRC Examination Reports 50-456/98305; 50-457/98305 The Braidwood training department in conjunction with Byron training department developed an i initial operator licensing examination that was administered to one Braidwood Reactor Operator license applicant by NRC examiner Results: ' l e The' applicant passed the retake examination and was issued a Reactor Operator's license.
, Reoort Summarv: e' The training staff's knowledge of the examination development guidelines, attention to . detail during examination development, and the ability to develop technically accurate I . written examination material in accordance with the examination guidelines were ! considered satisfactory. The licensee performed quality assurance reviews and satisfactorily submitted the written examination for NRC approval. . (Section 05.2) ~
~
l
'
e - The inspectors determined the training staff properly administered the written examination. No examination compromise issues were identified and examination security was considered good. (Section 05.3)
e' Taking into account the combined written examination results of Braidwood and Byron, the high failure rate and below average grades on the written examination suggested that the training programs did not well prepare the applicant for the examination. Several apparent knowledge deficiencies were identified through the written examination, including some understanding of system response, knowledge of operator actions, and ' bases for technical specifications.- (Section 05.4) e The examiners accepted four out of the five post written examination comment Overall, the licensee's submital of the post examination documents was considered satisfactory. (Section 05.5) , l l l
.- -- . - . --.--- - - .
-- . -. - . -- .
.
l Reoorts Details l LOperations I 05 Operator Training and Qualification 05.1 General Comments - Initial Ooerator License Examination - An initial licensing examination was administered to one Reactor Operator (RO) - applicant in conjunction with the Byron initiallicense examination at the Byron Statio The written examination was administered by Byron training staff with approval from and observation by the NRC examiners on September 14,199 The licensee developed the initial operator license examination in accordance with guidance prescribed in NUREG 1021, " Operator Licensing Examination Standards for Power Reactors," Interim Revision 8. In general, the examiners reviewed and approved all examination material that the licensee developed prior to its administratio .2 Pre-Examination Activities Examination Scope The licensee developed the examination material in accordance with the prescribed examination development guidelines. The examiners reviewed, revised, and validated the written examination material during the week of August 31,199 Observations and Findinos Written Examination
.
The written examination was primarily developed by Byron training department with site specific information updated to reflect Braidwood Station. The Braidwood training department staff reviewed and verified the written examination prior to NRC submittalin accordance with NUREG 1021 ES 401-6, " Written Examination Quality Assurance Checklist." The examiners reviewed all 127 questions from the originally submitted written examination. The examiners identified that some question deficiencies that required additional corrections and enhancements to better conform with the written examination question development guidance stated in NUREG 1021, " Operator Licensing Examination Standards for Power Reactors," Interim Revision 8. The licensee was informed of the potential changes and improvements needed on the written examination. During the on-site validation week, additional effort was made to ensure the required changes and enhancements were made to properly reflect the examination guidance to allow for exam administration. Following the validation week, the licensee had one week to incorporate the changes and enhancements to the written exarninatio _ _ _ _ _ _ _ _ . ~ _ _ _ _ _ _ _ . ____. ~_ . _ _ _. __
.
i Conclusions 1
,
The training staff's knowledge of the examination development guidelines, attention to : detail during examination development, and the ability to develop technically accurate l l written examination material in accordance with the examination guidelines were i ! considered satisfactory. The licensee performed quality assurance reviews and satisfactorily submitted the written examination for NRC approval. For specific details l on the written examination deficiencies refer to Byron initial examination report ' 50-454/98301(OL); 50-455-98301(OL).
! 05.3 Examination Administration Examination Scope
The written examination was administered on September 14,1998, by Byron training I staff with approval from and observation by the NRC examiners.
' Observations and Findinas l Written Examination The licensee administered the written examination with the approval from and observation by the NRC. The testing facility was appropriate to assure proper examination security. The licensee's examination proctors appropriately implemented i their responsibilities in accordance with the guidance of NUREG 1021, Section ES-40 l
. All appropriate documentation for written examination administration was complete No examination compromise issues were identified.
' Conclusions The inspectors determined the training staff properly administered the written examination. No examination compromise issues were identified and examination security was considered goo O5.4 License Aeolicant Performance Examination Sc_qng An initial licensing written retake examination was administered to one RO applican ' Observations and Findinas Written Examination-The one RO applicant passed the written retake examination. But, initially the RO applicant along with the five Byron initial license applicants all failed the written
- examination. The examiners, taking into account the licensee's post examination l- comments, regraded the applicant's written examination and the applicant subsequently (
, i 4 L _ _ _ _ . ._ _ _- _
- .- - . . . . . .. .
l l l passed the test. Following the examination grading, the licensee submitted a list of
'
questions that were missed by 50% or more of the applicants, and a matrix of potential knowledge weaknesses based on the written examinatio I Conclusions Taking into account the combined written examination results of Braidwood and Byron, the high failure rate and below average grades on the written examination suggested that the training programs did not well prepare the applicant for the examination Several apparent knowledge deficiencies were identified through the written l examination, including some understanding of system response, knowledge of operator actions, and bases for technical specifications. For further details on the written examination concerns refer to Byron initial examination report 50-454/98301(OL); 50-455/98301(OL). I
05.5 Post Examination Activities Scope The examiners reviewed the written examination grading that was performed by the licensee in accordance with Form ES-403-1," Written Examination Grading Quality Assurance Checklist," contained in NUREG-1021, Interim Revision 8. The examiners also reviewed the post written examination comments submitted by the license Observations and Findinos The post examination submital included the necessary documentation as required per the guidance of NUREG-1021, ES-501. The licensee submitted an analysis of the written examination results, which was a list of missed questions by 50% or more of the applicants, Braidwood and Byron, who jointly took the written examinatio . The examiners also reviewed the licensee's submitted post written examination comments. Four out of five comments were accepted by the examiners, and the written examination was graded accordingly. The licensee's comments and the NRC resolution of the comments are detailed in Enclosure 2," Post Written Examination Facility Comments and NRC Resolution." Conclusions The examiners accepted four out of the five post written examination comment Overall, the licensee's submital of the post examination documents was considered satisfactor V. Manaaement Meetinas
- X1 Exit Meeting Summary l
The examiners conducted an exit meeting with members of licensee menagement on September 22,1998, and the licensee was contacted by telephone on October 20,1998, to inform licensee management of the examination results. The licensee acknowledged the ,
l l
t
. .. _ . .. , . . . - . . ~ , . - - . = . . . - . . . . .. .- - -. . . . . , *
l findings presanted and indicated that the materials reviewed were not considered proprietar In addition, members from the Braidwood training department staff attended the Byron senior management meeting conducted with members of Byron and corporate management at the Region 111 office on October 21,1998.- The purpose of this meeting was to hear and discuss Byron's post examination root cause evaluation concerning the high failure rate on the written examination.
l
. . - . . _ _ . . . - . . . . _ - . .. .- - . - , , .
PARTIAL LIST OF PERSONS CONTACTED Licensee . C. Cerovac, Operations Training Superintendent T. Benton, ILT Group Lead P. H.'ppley, NGG Exam Developer M. Brown, Byron Training Instructor - Exam Developer NRC C. Phillips, Senior Resident inspector i INSPECTION PROCEDURES USED NONE ITEMS OPENED, CLOSED, AND DISCUSSED 4 NONE
.
5 t
J s
7
. . _- .. . - _ - - . . .. .- . . .. . . . , . .
LIST OF ACRONYMS USED BwAP Braidwood Administrative Procedure BwOA- Braidwood Abnormal Operating Procedure BwOP- Braidwood Operating Procedure CFR Code of Federal Regulations Comed - Commonwealth Edison Company - DRS Division of Reactor Safety EOP Emergency Operating Procedure-ES Examination Standards ILT Initial Operator Licensing Training IP inspection Procedure JPM Job Performance Measure K/A Knowledge and Abilities LCO Limiting Condition for Operation - LOCA Loss of Coolant Accident NRC Nuclear Regulator Commission NRR NRC Office of Nuclear Reactor Regulation OL Operator Licensing PDR Public Document Room RO Reactor Operator SAT Systematic Approach to Training SG Steam Generator SM- Shift Manager TS Technical Specification
8
.
. ..
_ l Enclosure 2 Facility Post Written Examination Comments and NRC Resolution i 1. EXAMINATION QUESTION RO #3 l LICENSEE COMMENT l l The question asked how was a procedure change procedurally conveyed to members of i the operating crew. Procedure BwAP 350-2 Rev 6," Daily Order Book," requires only the Shift Manager (SM) to read and initial the Daily Orders and was responsible to ensure appropriate operating personnel were notified, as necessary. The question distractor (A) was incorrect because a "memn" was not issued to all crew personnel and the SM may not be the person who places the information in the Daily Order Book.
! ! Distractor (B) was incorrect because the SM was not necessarily informed by memo and it wasn't proceduralized. Distractor (C) was incorrect because individual operators were briefed by the SM (by procedure). Distractor (D) was also incorrect because the Shift Operation Supervisor does not make an announcement at the shift briefings; it was the SM who does. Therefore, there was no definitive correct answe Recommend deleting the questio ) l NRC RESOLUTION The comment was accepted and the question was delete Question His. tory No chales were made to the original question as submitte . EXAMINATION QUESTION RO # 10 LICENSEE COMMENT The question asked the condition for entry into a fuel handling accident given the plant conditions. The entry condition for distractors (B) and (D) directly apply. No procedural differentiation or hierarchy for a backup indication or report to cause the immediate actions to occur. BwOA Refuel-1, Rev. 54, symptoms / entry conditions include: observed dropping of, or damage to, a fuel assembly; observed dropping of an object or a fuel assembly; observed from a potentially damaged fuel assembly. Any of the procedure entry conditions necessitate use of BwOA Refuel-1. Step 3 only requires one of the alarms (AR011 or AR012) in order to execute actions. Expected, conservative operator actions would be to initiate or actuate Unit 1 Containment evacuation alarm when RE-AR012 alarms, followed by a report from personnel in containment or corresponding rise on a duplicate monito Recommend changing the answer key to accept both (B) and (D) as correct answer NRC RESOLUTION The comment was accepted, and both (B) and (D) was noted as correct answer Question History No changes were made to the original question as submitte __ - - . . .= - _
.
, Enclosure 2 l 3. EXAMINATION OUESTION RO # 13 LICENSEE COMMENT The reference cited for this question / answer was BwAP 380-2, Rev. 6E1. This was not valid for the plant conditions stated in the question. This procedure applies for alarms that are in alarm condition >1 shift and for power levels > P-8 set point (30%). The only applicable standard that applies was NOD OPS DEPT STD-BWD which states that "the SER shall be utilized to verify the exact cause of each annunciator..." and
" announcement of repetitive alarms ... are not required with the concurrence of the Unit l
Supervisor... ," BwAP 300-1, Rev.19E1, also applies, giving generic good operating practice guidance to "believe their instruments until the indications are proven to be false." OPS DEPT STDS and BwAP 300-1 are governing guidanc Recommend changing the answer key to eccept both (A) and (C) as correct answer NRC RESOLUTION The comment was accepted, and both (A) und (C) was noted as correct answer Question History: No changes were made to the original question as submitte . EXAMINATION QUESTION SRO # 54 LICENSEE COMMENT Candidate answered question correctly for current Braidwood plant configuration and procedures in effect. Automatic actuations in combination with designated SG levels l would have resulted in both auxiliary feedwater pumps running. The student was advised to use current plant configuration, procedure references and setpoints for the ; purposes of this examination (" freeze" point of training and evaluation process). This was clarified to him by the ILT group lead in preparation for this examination. The proctor at the point of administering the examination directed use of "NEW" SG numbers (setpoints, data) based on the belief that the student had received this specific training (which he had not). The student had received SG RO training which characterized the modification / purpose, not the setpoints/ procedure specifics. Candidate had not attended licensed requal training v1.ich revi3wed abnormal and emergency operating procedure changes relative to new set points. He was in formal ILT remediation at this poin These changes are not in place yet. He will receive this training during September and October 1998 as part of preparation for license activatio Recommend changing the answer key to accept choice (A) as the correct answer.
NRC RESOLUTION The comment was NOT accepted, and the correct answer remains as (B). The original examination question, with the associated answer key and reference based on the New i SG information, was submitted to the NRC. The exam documentation with the NRC , Quality Assurance checklist (ES 401-6), original submittal and subsequent revision, was l accepted and signed by the Braidwood Training Department as part of the NRC exam , submittal. The licensee had several opportunities to review, verify, and correct the written examination prior to final submittal to the NRC for exam administration.
'
. . _ _ _ . . ._ _ _ . _ _ _ _ _ _ . . - _ _ . . _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ ,
.
Enclosure 2 Additionally, the licensee's exam development group specifically noted that the examination referenced the new SG and approved by the Braidwood facilit ' Qgestion Historv Minor wording change on the question stem was made. Changed some information in the question stem that potentially gave information for a previous question. Also, the deleted information was not needed to answer the question. Deleted the word "ONLY" from two distractors, to eliminate " specific determiners" per NUREG 1021, Appendix Remainder of the question stem, distractors, and answer were left the same.
'
, EXAMINATION QUESTION RO # 96 LICENSEE COMMENT The question stated that RCS activity was increasing due to corrosion product activation and to identify the effects of placing the cation bed demineralizer in servic Procedural guidance discriminates the purposes of each ion bed. The cation bed was efficient at removing Lithium which contributes to low pH and further corrosion products
and associated activity. BwOA PRI-4 (High Reactor Coolant Activity) provides guidance / direction for action to be taken in the event of high RCS activity, it directs that the standoy mixed bed demineralizer be placed in service NOT the cation demineralize BwOP CV 8 references use of the Cation Demin for the reduction of pH or fission products, (not for the purpose of reduction of activated corrosion products). Based on procedural guidance and use, the cation demineralizer would not be placed in service, substantiating (A) as a correct answer; (B) would be correct based on the physical properties of the demineralizer, not accounting for procedural guidance and us Recommend -changing the answer key to accept both (A) and (B) as correct answer NRC RESOLUTION The comment was accepted, and both (A) and (B) was noted as correct answer Question Histort No changes were made to the original question as submitte , _ - _ ,. y ,. .- _ # ,,_ . . - . . -
- _ , , . , . .
_ __ , l l ! l l . I '
.
' l ES-401 Site-Specific Written Examination Form ES-401-7 Cover Sheet ,
U.S. Nuclear Regulatory Commission Site-Specific l Written Examination l l
:
Applicant Information Name: MASTER EXAMINATION Region: lll SEPTEMBER 14, 1998 Date: Facility / Unit: Braidwood 1 & 2 )
License Level: RO Reactor Type: W I Start Time: Finish Time: Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top , of the answer sheets. The passing grade requires a final grade of at least 80.00 percen l Examination papers will be collected four hours after the examination start '
~
j Applicant Certification All work done on this examination is my own. I have neither given nor received ai Applicant's Signature Results Examination Value Yf -HXT4 Points Applicant's Score - Points l Applicant's Grade Percent i NI
- . - - - -- .__ ._ . -.
""""
- ()CCANTRZN* FORM NO.888-E ct00 r903 r802 c702 r60i , , , , , '"g, =
.
TO REORDER CALL 1-800 826-7196 ,mm,,
~ "
c50s c40, r30> c202 c10i .- ll :'.':*.'.. ::"." :: '"" pggy 3
~ = -
c92 c8s c7s c6 c5s .".'". - . . . . + . u ., r., . - - - i . i EME PART 2 c43 e33 e2s c1a c02 :::l* II I I T Yl:J g y (T) (F) KEY PART 1 g c%s *Wu c31 *
"""""
1 cAs eB3 me= cDs cEs .
""""
2 c As eBs * cDs cEs r - ;. . . ; .. ; ms,E; AM4 1998 INITIAL LICENSE TRAINING f """" @4 *** j """" cBs cCs cDs cEs 5 c As eB2 *cDs cEa
~
6 asas cB> cC3 cDs cE' f Description: NRC REACTOR OPERATOR FINAL RYAMINATION p~ 7 cAs =O= cCs cDs cEs Q) gog f g,[yg eg ~ 8 adse cBs cC2 cDs cEs p f p 9 cAs *#= cCs cDs cEs ,g { """" h A s @ c hCcEs s
"""" ~
11 c As cBs cCs *ea cEs s d /[/[b W
~
12 c A s sem cCs cDs cEs g ame/Date: MIN M / /
~ """' @ rBs h cDs cEs 14 etr cBs cCs cDs cEs Instructions """"
15 c A s cBs ee. cDs cEs
-
16 c A s eBs me= cDs cE, 1. Use black ink only on all portions of the exam
~
17 c A s me* cCs cDs cE, package EXCEPT for the scantron answer selections,
-
18 .A= cB2 cC2 cD3 cEs i 6
-
19 c A s eBs cCs se* cEs 2, print your name and date in the space provided abov """"" 20 c A s eBs =e= cDs cEn 3. If you have any cuestions or need clarification sm - - 21 c A s =#= cCs cDs cEn during the examination. notify the procto o 22 c As cBs =e c D s cEs
@~ 23 se e B s cCs cDs cEs 4. Conversation during the examination is prohibited si + 24 c As tBa cCs mes cEs except when asking for question clarification from J $' 25 cAs * cCs cDs cE, the procto R 26 **= cBs rCs cDs cE'
5. Cheating on the examination will result in failure 5 27 c As cBs cCs me= cE' of the examination and may result in further
* -
28 c A s sem- cCs cDs cEs
-
29 as,. c B 3 cCs cDa cEs disciplinary actio cB2 =e= cDs cEs 6. Use only #2 pencil to mark"your selection on this N ~ 3031c cA As s cBs cCs =ess cE' exam shee ~ 32 c As eBs me= cDs cEs _
-
33 c A s ses cCs cDa cEs 7. Completely darken the selected answer. If you make
-
34 c A s cBs ae= cDs cEs a mistake, completely erase the darkened selectio ~ 35 c A s eB3 cCs =#a cEn
-
36 c A s eBs cCs e cEs 8. Ensure you do not skip a question or answer which
- W uld place you out of sequenc c A s aes cCs cDs cEs ~
38 c A s cB3 mea cDs r- E ' 9. Do not place any extraneous marks on this exam shee ~ 39 c A s cBs we c D s cEs
-* 40 c A s cBs cCs see cE, 10.You have hour (s) to complete this exa c A s cB2 cCs =e* cE' ,. -
42 c As cBs mew cDs cEs 11. Prior to handing in your exam, verify that you have transferred your answers to thic scantron sheet me* cCs cDs cEs
! " 43 c44 A,# s cB3 cCs cDs cEs Properl ,
l ~ 45 c A s eBs <Cs ee= cE' b I have neither given, received, or Obeerved any aid or
? ~~ 46 c A s eB3 e cDs cE' informa tiOD regarding this exam prior to or during itS ;;; 47 c A s cBs & cDa cEs 48cAs cBs een cDs cE3 administra tion tha t COUld Compromise th: S exam'S g -
49 c A 2 cB2 cCs =#s cEs integriCY. I alSO Understand my Obligat]On to report
. 50 eA , =e- cC3 cD, cE, aRY ex Compromise by otherS prior, during, or Subsequent to the exam administratio CD % aln w JA . M oh signature l l date
. _ .
W NOLLO3tHO.SIH1 0331 :
('l1111111111111!!ll111111 :11111111111111111111111 lill s s @sesee s==aa=aeee s ::::::=n ===e==re e s=s :=@e:e s.tst$stsss sestlisiss esssssssil lsistissss sssesE*fssiass8s e
5555@I555I III55555I5055III5555 5555555I50 I055I5%5I5I2' 9968 . lsissssiss ssssssssss ssssssssss sisissssss sssist??sts 9988
9999999919 999999999f 9tl9999999 9999994969 99f99999991 3988 L mm.... mmmmmmmmm ,mmmmm mmmmmmmmmmmmmmmm mmmm ga gggg [ . e j h ' ~
. . .
B
'
t, th - lilij
. ) t
\ 3-
'
3 I s Ill { : l{
:;i,iI lii . !
L
--ee 1
-.,
.l 1 :
w ?# 'l k n : b i
.
+
'
.
I
.
- .I i .
.
_ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ . . . _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ . _ _ _ _ . . . _ _ _ _ _ _ _ _ _ _ _ _ _ _
- - . - - . - - - . . - . ~_ .. - - - . - . . . - . . _ - .
Question [ Ev1 tuition of requirement for"activt lic:nse An "Activs" licensed NSO (original license obtained in 1996) worked the following schedule at Brcidwood:
- 9/4 - 0700 to 1500 as Unit 1 NSO - 9/7 - 0700 to 1500 as Unit 2 NSO - 9/8 - 0700 to 1500 as Unit 2 NSO -- 9/9 - 0700 to 1200 as Unit 2 NSO and 1200 to 1500 as WEC NSO - 9/10 - 0700 to 1500 as WEC NGO - 9/11 - 0700 to 1500 as Unit 1 NSO - 9/14 - 1500 to 2200 as Unit 2 NSO - 9/12 - 1500 to 2200 as Unit 2 NSO Ths NSO...
s. meets the requirements for maintaining his/her license active for the next quarte b. needs to work an additional FOUR hour shift to maintain his/her license active for the next quarte [ . needs to work an additional EIGHT hour shift to maintain his/her license active for the next quarter, d. needs to work TWO additional EIGHT hour shifts to maintain his/her license active for the next quarte , Answer C Exam Level B Cognitive Level Memory Facility: Braidwood ExamDate: 9/14/98 KA: 2. RO Value: 3.7 SRO Value: 3.8 Section: PWG RO Group: 1 SRO Group: 1 System / Evolution KA Knowledge of conduct of operations requirements, Explanation of Answer Reference Title / Facility Reference Number Revisio L Braidwood Ops Memo #2 97 issued 6/1/97 rev. O Bwd Tsk List Task P1-AM TK-180 Material Required for Examination , Question Source: New Question Modification Method:
. Question Source Comments:
Comment Type Comment > Friday, September 4, it98 Page 1 of 100 -
_ _ . . . _ - . __ . . _ _ _ _ _ _ - - . _ . - - - . _ _ - . ouestion 3 operating D:ily orders . How is a procedura ch ngs, which significantly chang:s normal process 2s, procedurally convsyrd to liCrnsed members of the operating crew? j e. The SM places the applicable information in the Daily Order Book, a issues an additional memo to i i all crew personnel that is initiale l
b. The SM is informed by memo of the addition to the Daily Or r Book, and makes an announcement l of the addition during the shift briefing, j o. The SOS places the applicable information in the y Order Book, and the individual operator is l responsible for reviewing the Daily Order, d. The SOS places the applicable informatio ' th4 Dailyhrder Book, and makes an ann the addition during the shift briefin Answer C Exam Level B Cognitive Levet e ry a cility: Braidwood ExamDate: 9/14/98 KA: 2. RO Value: 3.0 SRO alue: . P#G Ro Group: 1 SROGroup: 1 System / Evolution k KA Knowledge of operator res sibilities Jrin s of plant operatio Explanation of Answer Reference Title / Facility rence Number Page Revisio L. BwAP 350-2 ev. 6 C.71dl 14 Intro to Main Control R Ops Lesson Plan 6 Craidwood Task List Task P1.AM-TK-025 terial Requi for Examiraation wuestion Source: New Question Modification Method: Question Source Comments: Comment Type Comment
.
Fridly, september 4,1998 Page 3 of 10C
..
_._ _ _ _ _ _ 1 Question Y Procedura required usaga An ex mple of a licensed oper: tor evolution that c n be performed WITHOUT having a procedura in I I hind is ... l e. Adjusting rod position following a boration for delta-l contro c. Starting the.1 A Heater Drain Pump.
- c. Placing excess letdown in servic d. Latching and rolling up the main turbine following surveillance trip tes ;
Answer a Exam Level B Cognidve Level Memory Facilky: Braidwood ExamDate: 9/14/98 l
.KA: 2.1.23 RO Value: 3.9 SRO Value: 4.0 Section: PWG RO Group: 1 SRO Group: 1 System / Evolution KA 1 Ability to perform specific system and integrated plant procedures during all modes of plant operatio ;
'
E=? :^15 of Answer Ruerence Title / Facility Reference Number Section Page Revisio L Use Of Procedures For Operating Department BwAP 340-1 C.1.f.3) pg 4,5 rev.12 l
'Braidwood Task List Task P1-AM-TK-022 l Material Required for Examination Question Source: New Question Modification Method Question Source Comments:
Comment Type Comment I i l
,
l I
.
! I I l l I
i' l Friday, September 4.1998 Page 4 of 100 ..
.- . -- ... --.. - . . . . .. . _ . -
Question [ Use of clectrical prints Assuming en cuto-closs signal is continuously pr:snnt in ths circuit for tha 1 A SI pump, which contact will be maintained open in order to prevent the starting relay (SR) from attempting repeated breaker closures onto a faulted bus?
(ti 1-4030-S101 is provided for use.)
c. LC SW e. 52/b e. Y d.LS Answer C Exam Level B Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 2.1.24 RO Value: 2.8 SRO Value: 3.1 Section: PWG RO Group: 1 SRO Group: 1 System / Evolution
'KA Ability to obtain and interpret station electrical and mechanical drawing Explanation of "Y"is an antipump relay that when prevented from energizing interrupts the circuit that energizes the START Answer relay in the AUTO mstart circuit Reference Title / Facility Reference Number Section/Page Revisio L O.
Schematic Diagram Safety injection Pump 1 A 20E-1-4030S101 Print Reading Lesson Plan Chap 3 pg 23 rev. 5 2c,3
'4aterial Required for Examination estion Source: Facility Exam Bank Question Modification Method: Editorially Modified Question Source Comments: Braidwood requal bank Comment Type Comment
!
l l l
.
Frid:y, September 4.1^98 Page 5 of 100
. . __ . . _ . . . . __ ._ _ _ _ _ . . . __ .. _ ._ _
Question h .MOV tagout An operator is prcpiring en OOS thtt d signct:s 1CC685, RCP Thermal Barrier CC Rcturn CNMT isolation valve, as an isolation point.
l l ' hat is the acceptability of using this isolation point? Th3 OOS is... e. acceptable if the MOV is tagged at its control switch, power supply and valve handwheel, b. acceptable if the MOV is tagged at its control switch, power supply and a blocking device is placed on the valv . c. NOT acceptable because the MOV fails to meet isolation requirement . d. NOT acceptable because the valve fails open on a loss of powe Answer a Exam Level B cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 MA: 2.2.13 RO Value: 3.6 SRO Value: 3.8 Section: PWG RO Group: 1 SRO Group: 1 System / Evolution KA Knowledge of tagging and clearar.ce procedure Explanation of Valve is MOV and requirements include tagging control switch, electrical power supply and local handwheel if Answer accessibl Reference Title / Facility Reference Number Section/Page Revisio L BwAP 330-1 Out of Service Process D.4.a pg 12 D.4.c.1) pg 14 F iwood Task List Task P1.AM.TK-010 Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment
.
FridIy, September 4,1998 Page 6 of 100
._, - - -.
. ._ . -- - - - . . - - - .- - .-. .. . ..
Question 7 RCSlev 1discrrpincy during refulling Ths following conditions exist for Unit 1 in pr:ptration for head removal: -
.
l
- Unit shutdown and cooldown initiated 120 hours ago - Lowering of RCS level to the reactor vessel flange is underway l - RCS temperature - 95* - RCS level Control Room indicators: 1Ll-RYO46 - 401'0" 1LI-RYO49 - 402'1" l - RH loop 1 A in operation with " normal" indications What is the appropriate action for these conditions?
a. The lowering of RCS level can continue after verifying appropriate amount of water remove b. The level change must be stopped until the cause for the level discrepancy is determine c. The running RHR pump shall be immediately stopped to prevent cavitatio d. The available S1 Pump is immediately aligned for hot leg injection and shall be started.
l i Answer b Exam Level B Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 2.2.26 RO Value: 2.5 SRO Value: 3.7 section: PWG RO Group: 1 SRO Group: 1 Sy; tem / Evolution KA Knowledge of refueling administrative requirement Explanation of With any level discrepancy, the reason for the discrepancy must be determined before further draining can Answer Continu I ence Title / Facility Reference Number Section/Page Revisio L BwOP RC4 Reactor Coolant System Drain E1
BwGP 1004 Refueling outage lesson plan 12 2 l Material Required for Examination Question Source: Fadlity Exam Bank Question Modification Method: Significantly Modified Question Source Commects: Zion exam bank Comment Type Comment NRC Significant industry Event - L l
.
! (.
.
Friday, Septe-mber 4,1998 Page 7 of 100 l . .- - -
- = . - - - . - - - - . - _ . .- . .
Question [ RO duurs in Control Room during refu; ling Wh t is a responsibility of ths NSO during refu: ling operttions in ths main control room? c. Checking source range counts while a fuel assembly is being placed in the cor b. Verifying direct phone communication with the Fuel Handling Supervisor once per day during fuel movemen c. Maintaining a 1/M plot while reloading fuel during a core shuffl d. Updating the Control Room tag boardper the Nuclear Component Transfer List on an hourly basi Answer a Exam Level B Cognitive Level Memory Facility: Braidwood ExamDate: 9/14/98 KA: 2.2.32 RO Value: 3.5 SRO Value: 3.3 Section: PWG RO Group: 1 SRO Group: 1 System / Evolution KA Knowledge of Ro duties in the control room during 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 instrumentatio Explanation of Answer Reference Title / Facility Reference Number Section/Page Revisio L0 SWAP 2000-38 Reactivity Management F.2.h.5) pg 11 2E2 Braldwood Task List Task P1-QG-TK-051 Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment Friday, September 4,1998 Page 8 of 100
. _ _ . . _ .--. . - - - _ - _ - - - -. .
Qu.suon 7 Radiation cxposura dit:rminition An oper: tor hrs th3 following exposuro history this ysar until today: - Deep Dose Equivalent (DDE) - 210 mrem ommitted Effective Dose Equivalent (CEDE) - 45 mrem dhillow Dose Equivalent (SDE) - 33 mrem i Committal Dose Equivalent (CDE) - 28 mrem Today the operator was required to make two entries into containment:
Entry 1: Gamma dose - 52 mrem; Neutron dose - 24 mrem Entry 2: Gamma dose - 124 mrem How much radiation exposure is available to the operator if he has to make additional entries? His available margin based on the routine Administrative Exposure Control Levels is... c.100 mrem for that day; 2484 mrem for the yea b.100 mrem for that day; 2545 mrem for the yea j c.124 mrem for that day; 2569 mrem for the yea I d.124 mrem for that day; 2614 mrem for the yea Answer b Exam Level B Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 2. RO Value: 2.6 SRo Value: 3.0 Section: PWG Ro Group: 1 SRO Group: 1 System / Evolution Knowledge of 10 CFR: 20 and related facility radiation control requirement Explanation of Limits are 300 mrem routine DDE/ Day and 3000 mrem routine cumulative TEDE/ year. C. Neutron rad not ! Answer counted for daily & yearly; A. All counted for yearly; d. previous DDE+ CEDE only counted for yea Reference Titie/ Facility Referet ce Number
. Section/Page Revisio L Selected BwRPs Lesson Plan Rev.00 2,3,4 Material Required for Examination Question Source: New Question Modification Method:
Question Source Comments: Comment Type Comment . - Friday. September 4.1998 Page 9 of 100
. - - - .~ _ .- - - .- . _ . _ - . _ . - - -- -
, Question /8 Fu*lHandling Accident Response The following conditions exist on Unit 1: - i
- Refueling operations in progress - A HIGH alarm received on radiation monitor 1RE-AR012, Containment Fuel Handling incident
When should the NSO initiate action'and what action should he/she take from the control room?
, l Indication of a fuel handling accident i.s considered when a... l e report is received from personnelin containment.' The operator starts the containment charcoal filter fan b. report is received from personnelin containment. The operator actuates Unit 1 CNMT evacuation alar a. corresponding rise is indicated on monitor 1RE-AR011. The operator starts the containment charcoal filter fan j
., d. Corresponding rise is indicated on monitor 1RE-AR011. The operator actuates Unit 1 CNMT evacuation alar Answer d Exam Level R Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 2.3.10 RO Value: 2.9 SRo value: 3.3 Section: PWG RO Group: 1 SRO Group: 1 System / Evolution KA Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposur Explanation of answer R..erence Title / Facility Reference Number Section/Page Revisio L BwOA REF-1 Lesson Plan Re ,3,4
.
Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment
-
Friday. September 4.1998 Page 10 of 100
-
. .. _ _ _ _ _ . _ . . ._ _ _ . . . . _ _ . _ __ . _ . . _ _ _ _ _ _
, Question // P;rformanca of Status Trees / Function R:storation , The following conditions exist on Unit 1: I
- A reactor trip has occurred and both reactor trip breakers are verified open - The turbine has tripped - BwEP-0 " Reactor Trip OR Safety injection" has been entere BUS 141 ALIVE light is NOT lit with bus voltage at ZERO volts
!
- BUS 142 ALIVE light is lit with bus voltage at 4149 volt Which of the following describes the action (s) the operators is/are required to take?
e. Check SI statu b. Turn on the synchroscope and rnanually close ACB 1412, SAT 142-1 feed breaker.
, e. Manually start 1 A D/G and verify ACB 1413, D/G output breaker, closes.
l ' d. Initiate actions of BwOA ELEC-3 and then check SI statu Answer d Exam Level B Cognitive Level Memory Facility: Braidwood ExamDate: 9/14/98 KA: 2.4.16 RO V61ue: 3.0 SRO Value: 4.0 Sec+1on: PWG RO Group: 1 SRO Group: 1 System / Evolution KA Knowledge of EOP implementation hierarchy and cooNination with other support procedure Explanation of Answer Rsference Title / Facility Reference Number Section/Page Revisio L Reactor Trip or Safety injection BwEP-0 Step 3.b. RNO P '-0 Rx Trip or SI Lesson Plan rev.11 1,3 l Material Required for Examination l Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment l i
~
! Friday, September 4,1998 Page 11 of 100
_ . . - -- . . . . _ . -- . . - . - - . - . . . - - - _ - - . . - . _.
- ousetion /d Applicability of EOP Foldout Paga Following tr nsition to BEP-1 " Loss of Reactor Or Secondtry Coolant", ths US rafars to tha Operator L
' Action Summary, and directs the operator to Cold Leg Recirculation Switchover Criterion. Which of the following describes the complete set of procedures for which the Transfer to Cold Leg Recirculation quirements are applicable?
,
f . l l (NOTE: The following procedures are in the E-1or CA-1 series: l , BwEP-1 " Loss Of Reactor Or Secondary Coolant" ' BwEP ES-1.1 "Si Termination" BwEP ES-1.2 " Post-LOCA Cooldown And Depressurization" l l BwEP ES-1.3 " Transfer To Cold Leg Recirculation" l BwEP ES-1.4 " Transfer To Hot Leg Recirculation" BwCA-1.1 " Loss Of Emergency Coolant Recirculation" BwCA-1.2 "LOCA Outside Containment") a. BwEP-1, BwCA-1.1 and BwCA-1.2 procedure b. BwEP-1, BwEP ES-1.1 and ES-1.2 procedure e. BwEP-1 and BwEP ES-1.2 procedure d. BWEP-1 procedur ' Answer b Exam Level B Cognitive Level Comprehension Fac4!'y' Braidwood ExamDate: 9/14/98 KA: 2.4.20 RO Value: 3.3 SRO Value: 4.0 Section: PWG 1t0 Group: 1 SRoGroup: 1 Sy; tem / Evolution KA Knowledge of operationalimplications of EoP warnings, cautions, and note " plination of swer Reference Title / Facility Reference Number Section/Page Revisio L DwEP-1 Loss of Reactor of Secondary Coolant Lesson Plan rev.11 1,10 Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment
.
l i Friday, September 4,1998 Page 12 of 100 l
Question /) ldintification ofinoper:ble CR cnnunciators The following conditions exist on Unit 1:
- Reactor trip breakers status - OPEN
- RCS Tave - 557'F
- Pzr pressure - 2235 psig
.
Annunciator RCFC VIBRATION HI (1-3-C5) has been in alarm for the past hour due to a vibration condition while maintenance troubleshoots the vibration probe on RCFC 1 Which of the following actions is appropriate for this alarm window? e. The alarm should be acknowledged for each actuation and the SER monitored for valid alarm input b. The alarm should be acknowledged for each actuation and operators stationed locally at each RCFC to monitor vibration, i c. The alarm should have been silenced without acknowledgement after obtaining Unit Operating Engineer's permission and the SER monitored for valid alarm input d. The alarm should have been silenced without acknowledgement with US permission and operators ) stationed locally at each RCFC to monitor vibratio ! Answer C Exam Level B Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 ! KA: 2.4.31 RO Value: 3.3 SRO Value: 3.4 Section: PWG RO Group: 1 SRO Group: 1 System / Evolution KA Knowledge of annunciators alarms and indications, and use of the response instruction .planation of Answer Reference Title / Facility Reference Number Section/Page Revisto L RCFC VIBRATION HI /BwAR 1-3-C5 HANDLING OF MAIN CONTROL BOARD and RADWASTE PANEL ANNUNCIATOR ALARMS / BwAP 380-2 C.4 Brddwood Task List Task P1-AM-TK-033 Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment
.
!
!
FridIy, September 4,1996 Page 13 of 10' I l
. . . -_ - - , - - - . _ .- . - . - - - - - . - -
l Queeuen /f Effect of X:non Tr nsient & com' pensation i A feed pump trip occurrea rcsulting in a rapid power reduction on Unit 1. Powcr was reduced from 100% st:"dy-state conditions using a combination of rods and boratio ) l J l [ fu following conditions exist for Unit 1 following stabilization:
- Reactor Power . - 60%
l - Delta-l target value - +2.0 l - Control Bank D position - 160 steps withdrawn
- Tave - 572*F l - Delta-l - -10.5% -Core Age - MOL l
l ' Wh t actions will be required to maintain the current power level and maintain Delta-1 within its normal l operating band over the next FIVE hours? c. Boration and control rod withdrawal, followed by dilution.
l b. Boration and control rod insertion, followed by dilutio ,
c. Dilution and control rod withdrawal, followed by boration j d. Dilution and control rod inseliion, folloWed by boratio Answer a Exam Level B Cognitive Level Application Facility: Braidwood ExamDate: 9/14/98 KA: 001 A2.06 RO Value: 3.4 SRO Value: 3.7 Section: SYS RO Group: 1 SRO Group: 1 System / Evolution Control Rod Drive System KA Ability to (a) predict the impacts of the following on the Control Rod Drive System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: l Effects of transient xenon on reactivity i Explanation of With delt-l near the negative limit of the band, boration would be initiated to to allow rod withdrawal and hence Answer shifting of power poduction toward positive delta-l (power shift toward top of core). Later as Xenon (neutron ! , poison) builds in, dilution will be initiated to maintain power level I
- _ Reference Title / Facility Reference Numbe Section/Page Revisio L O.
l DELTA l CONSIDERATIONS F.3,5,6 3,4-7 ' BwGP 100-8 BwGP 100-8 Lesson Plan rev 4 1 Material Required for Examination Question Sourc New Question Modification Method l Question Source Comments: Comment Type Comment I l l
.
l l I
.
a l
! l
- !
'
l Friday, September 4,1998 Page 14 of 100
- ._ .- - - - . . - .-
l Quesuon / Application of DC Hold j A problem with ths rod control syst:m requires checking s veral rod bank circuits. The affected power i cabinet repairs are to be made by supplying power from the DC hold supply cabine ! , hich statement describes the proper operation for DC Hold and the associated response in the l t:vant of a reactor trip? l
c. ONE control rod bank group can be placed on DC HOLD, and these rods will drop if the l
'
controls are taken to OFF at the DC Hold cabine b. ONE control rod bank group and ONE shutdown bank group can be placed on DC HOLD, and these rods will drop if the controls are taken to OFF at the DC Hold cabinet.
l e. ONE control rod bank group can be placed on DC HOLD, and these rods will automatically dro d. ONE control rod bank group and ONE shutdown bank group can be placed on DC HOLD, and these rods will automatically dro Answer C Exam Level B cognitive Leve Memory Facility: Braldwood ExamDate: 9/14/98 KA: 001 K1.03 RO Value: 3.4 SRo Value: 3.6 Section: SYS Ro Group: 1 SRO Group: 1 System / Evolution Control Rod Drive System j KA Knowledge of the physical connections and/or cause4ffect relationships between Control Rod Drive System and the following: CRDM Explanation of Only one GROUP of control rods can be placed on HOLD at a time in order to ensure the rods are held without Answer falling. Opening the reactor trip breakers interrupts power to the power cabinet and DC Hold cabinet , so that pow:r to the CRDM is interrupted when the breakers open Reference Title / Facility Reference Number Section/Page Revisio L. F ' Control System Chap 28 A.S.e pg 40 12 1,9 Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment
.
, Fridry. September 4,1998 Page 15 of 100 l
Muestion M RIlationship oflev:Is during refueling operztions i Yhe following conditions exist for Unit 1: - l ! - Mode 5 i l - RCS is draining to Pzr level of 40% I
- IM calibrations have been completed for LT-048, Refuel Cavity level, in preparation for further draining - LI-462 indicates 40%
What is the relationship of Pzr level instrument LI-459 as compared to Ll-048? c. LI-459 and Ll-RYO48 will be offscale hig b. LI-RY048 will be just onscale and LI-459 will be offscale lo l l ! e. LI-459 will read higher than Ll-462 and LI-RYO48 will just be onscal d. LI-RY048 will be offscale high and LI-459 will read lower than LI-46 Answer c Exam Level B cogniuve Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 002 A1.11 RO Value: 2.7 SRO Value: 3.2 Section: SYS RO Group: 2 SRO Group: 2 System /Evoludon Reactor Coolant System KA Ability to predict and/or monitor changes in parameters associated with operating the Reactor Coolant System controls including: Relative levelindications in the RWST, the refueling cavity, the PzR and the reactor vessel during preparation for refueling Explanation of Ll-462 is the cold calibrated Pzr level instrument and will read lower (but more accurately) than the hot Answer calibrated level instruments (LI-459/460/461) at lower RCS temperatures. The refueling cavity level instrument just comes onscale ht 40% Pzr leve F wnce Title / Facility Reference Number Section/Page Revisio L FuCTOR COOLANT SYSTEM DRAIN BwOP RC-4 D.2 pg 4 rev.12E1 BwOP RC-4A5 BwCB % fig 31 BwGP 100-6 Refuel Outage lesson plan rev.12 1,2 Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment i ! , Frid:y, September 4.1998 Page 16 of 100
. -. . . . . - . . - _ . . - . _ . - _ _ - ~ _ - - - - - . _ - . - . - - - . - - . ~ _ . . - . -
Question /7 RCS lerk Detection Syst:ms The following conditions exist for Unit 1: -
- Reactor power - 100% - RCS activity is elevated, but below Technical Specification (CTS) levels - Pzr pressure - 2225 psig - Pzr level - 44% <
l
- Leak rate - 10 gpm in cn attempt to isolate the leakage past the PORV, the Block Valve 1RY80008 was taken to close. The Block Valve failed to close and the operator placed 1RY456 in the CLOSE position. When conditions i stabilize: l - Reactor power - 100% : - Pzr pressure - 2228 psig I
,
- Pzr level - 44%
How would the operator be able to tell if the PORV has closed? j
.
e. Position lights for PCV-456 showing CLOSE indicatio I
n. Verify stable VCT level indicatio c. Level change in RCD d. Lower readings for containment radiation monitors RE-0011 A/0012 Answer b Exam Level R Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98
: 002 A3.01 RO Value: 3.7 SRO Value: 3.9 section: SYS RO Group: 2 SRO Group: 2 system / Evolution . Reactor Coolant System KA Ability to monitor automatic operations of the Reactor Coolant System including:
Reactor coolant leak detection system Explanation of Answer Reference Title / Facility Reference Number Section/Page Revisio L. WAR 12-C-6 rev51E2 ' Braidwood Task List task P1 OA.TK-058 ' Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: I Comment Type Comment _
.
' Friday, september 4.1998 Page 17 of 100
, . - . _ . -
-. .
_ _ _. . . _ - . _ ._ ._ __ _ _ ._ _ _ _ _ . _ _ _ . _ Question Use of Loop Isolition Vdvrs The following conditions exist on Unit 1: -
- RCS Loop C is isolated for maintenance - RCS Loop A had been isolated for maintenance - RCS Loop A Hot Leg Stop Isolation Valve (LSIV) was opened at 1001 - RCS Loop A Bypass Stop Valve was opened at 1005 with relief line flow of 115 gpm verified - RCS Loop A Cold Leg LSIV is closed - RCS temperature - 110*F - RCS Hot Leg Loop temperatures - 108*F (A); 119'F (B); 110*F (C); 125'F (D) - RCS Cold Leg Loop temperatures - 103*F (A); 108'F (B); 90*F (C); 115*F (D) - S/G levels (Narrow Range) - 20% (A); 30% (B); 15% (C); 32% (D)
Wh:t will occur when the operator takes the control switch for MOV-RC8002A (RCS Loop A Cold Leg LSIV) to OPEN at 15097 Th3 valve... c. will travel fully open with NO automatic actuations, b. Will travel fully open, and the AFW pumps get a start signa c. remains closed because the temperature difference interlock remains activ d. remains closed because the timer interlock is still activ Answer a Exam Level R Cogniuve Level Comprehension Facility: Braletwood ExamDate: 9/14/98
: 002 K4.09 RO Value: 3.2 SRO Value: 3.2 section: SYS RO Group: 2 SRO Group: 2 ty: tem / Evolution Reactor Coolant System i KA Knowledge of Reactor Coolant System design feature (s) and or interlock (s) whch provide for the following:
Operation of loop isolation valve Explanation of I Answer Reference Title / Facility Reference Number Section/Page Revisio L Simplified RCS/RC-1 valve interlocks /1 3 R! actor Coolant system lesson plan Chapt:r 12 8 9 Material Required for Examination Question Source: Facihty Exam Bank Question Modification Method: Significantly Moddied Question Source Comments: Question 30/35 on Braidwood 1996 NRC exam is about LSIV interlocks. Premise and answers signifcantly different. Question asked about interlock for opening HL LSI Comment Type Comment
.
I ! l l f , Friday, September 4,1998 Page 18 of 100
, - - , - -
.- - . . __ . - - . .- - -_ -. . . . - _ . - . .
Queouon / f RCP cnd Pzr spray oper'tions The following Unit 1 conditions exist: -
- RCS temperature (Average CETC) - 140*F - RCS pressure -
365 psig , - A bubble has just been drawn in the Pressurizer I - Allloops are filled and vented
- Preparations are in progress to start the first RCP for continuous run - 1C RCP is started l What is the effect on RCS pressure control?
RCS pressure will increase and... both Pzr Sprays will function normally for Pzr pressure contro b. manual cycling of the Pzr heaters will be required for Pzr pressure control, e. PORV RY456 will open on high pressure from high pressure bistable PB456 d. Pzr spray will deliver minimal spray flow for Pzr pressure contro Answer d Exam Lsvol B Cognitive Level Memory Facility: Braidwood ExamDate: 9/14/98 l KA: 003 A1.06 RO Value: 2.9 SRO Value: 3.1 Section: SYS RO Group: 1 SRO Group: 1 ! System / Evolution Reactor Coolant Pump System KA Abilrty to predict and/or monitor changes in parameters associated with operating the Reactor Coolant Pump System controts PZR spray flow l Explanation of Answer l L .ence Title / Facility Reference Number Section/Page Revisto L. BwGP 100-1 Plant Heat up f. 57 pg 20 rev 11 BwGP 100-1 Plant Heat up L sson plan 12 1;2,3 i Material Required for Examination , Question Source: New Question Modification Method:
- Question Source Comments
Comment Type Comment l ! !
~
I ! Friday. September 4,1998 Page 19 of 100 I
.- .. . - . _ .. -- . . .
Question AO RCP Br:cksr & int:rlocks The following conditions exist on Unit 1: .
- Reactor power 26% - Pzr pressure - 2235 psig - Pzr level - 35%
RCP 1 A breaker trips due to sensed undervoltage from bus 157. What is expected as a result of the trip c ths RCP? o. The reactor will trip due to the open RCP breaker, b. The reactor will trip due to RCS loop low flow conditio c. The reactor will be manually tripped by the operato d. A normal plant shutdown will be initiate < Answer c Exam Level R Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 003 K2.01 RO Value: 3.1 SRO Value: 3.1 Section: SYS RO Group: 1 SRO Group: 1 System / Evolution Reactor Coolant Pump System KA Knowledge of electrical power supplies to the following: RcPS Explanation of No AUTO trip is expected due to power < P-8. Administrative direction for a RCP trip in these condiitons is a Answer manual trip Will be initiated.
Reference Title / Facility Reference Number Section/Page Revisio L O.
Chp 13, Reactor Coolant Pump lesson plan C. 4.a 2)/ pg 16 9 8 AC Electrical Distribution lesson plan chp 4 8 10b (' - Memo / special Op Order Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: knment Type Comment FM;y, September 4,1998 Page 20 of 100
- - - . . --
Question M/ Charging & letdown flows (including seal inj ction) Tha following conditions exist on Unit 1: .
- Reactor power - 100% - PZR pressure - 2235 psig - PZR level - 44% stable - CV121 - In MANUAL - CVCS letdown - Isolated due to leak in Letdown Hx - CVCS Excess Letdown - In service with maximum flow of 20 gpm - RCP seal injection - 1 A CV pump aligned to all RCPs - RCP seal leakoff flow - 3 gpm (1 A); 3.5 gpm (1B); 3 gpm (1C); 2.5 gpm (1D)
What flow is indicated on Charging Header Flow indicator, F1-1217 c. 20 gpm b. 32 gpm e. 55 gpm d. 67 gpm Answer b Exam Level R Cognitive Level Application Facility: Braidwood ExamDate: 9/14/98 KA: 004 A3.11 RO Value: 3.6 SRO Value: 3.4 Section: SYS RO Group: 1 SRO Group: 1 System / Evolution Chemical and Volume Control System KA Ability to monitor automatic operations of the Chemical and Volume Control System including: Charging / letdown Explanation of F1-121 Indicates total charging flow (chg header + RCP seal flow, less Chg pump recirc (60 gpm)). Flow Answer balance - Letdown: 20 + 12 = 32 & Chg: 0 + 20 + 12 = 3 R...rence Title / Facility Reference Number Section/Page Revisto L CVCS/ Schematic CV-1 Chp 15a Chemical VolumeControl System lesson plan 10 4,5,9,15 Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment
.
l Frid:y, September 4,1998 Page 21 of 100 l' . l
_ - . - - - - .- .- .. -- Question sU Criculation cf dilution The following conditions exist on Unit 2: - ! - Unit is in MODE 5
- Unit burnup is 5700 EFPH in Cycle 7 - SDM - 1.3% DeltaK/K . - RCS pressure - 400 psig - RCS average temperature - 195'F - RCS boron concentration - 1006 ppm - Differential boron worth - -10.75 pcm/ ppm - PZR level - 32.3% - SR NIS countrma - 10 cps , BOTH channels stable background levels - An inadverterMution at 70 gpm begins at 1300 hours Assuming NO operator action is taken and PZR level remains constant over the time period, when would l the HIGH FLUX AT SHUTDOWN alarm actuate?
l c. No action, because BDPS will actuate prior to actuatio b.1430 hours, c.1505 hour d.1734 hour ' Answer C Exam t.evel B Cognidve Level Application Facility: Braidwood ExamDate: 9/14/98 KA: 004 A4.07 RO Value: 3.9 SRO Value: 3.7 Section: SYS RO Group: 1 SRO Group: 1 i Sy tem /Evoludon Chemical and Volume Control System Ability to manually operate and/or monnor in the control room: Boration/ dilution Explanadon of Dilution rate dc/dt = (500)(C)(Y)/M where M is the RCS mass at the given temperature (200*F). M = 745,537 Answer Ibm; C = 1006 ppm (given); Y=70 gpm (given). The dil rate = 47.2 ppm /hr. HIGH FLUX AT SHUTDOWN 1 alarms at 5 x background = 50 cps. With K1= 0.987 dK/K (p1=-0.01317), calculate K2 = 0.9974 DKr/K (p2=-0.00261). Delta-P = 1056 pcm.1056/-10.75=-98.2 ppm change required. Therefore the time required for the 98.2 ppm dilution is 98.2/47.2 = 2 hours 5 min. Difference in time based on use of Nomograph for RCS at normal pressure & temperature conditions. 'd' would only occur if count rate doubled in any 10 minute period. Assuming count rate increase is linear, for given dilution rate counts would change by 3 every 10 minute Reference Title / Facility Reference Number Section/Page Revisio L. R; actor Makeup Control system lesson plan 8 4,7,11 Source Range Nuclear instrumentation Lesson plan 6 6,10,11 Brcidwood Curve Book Boron dilution rate nomograph
- Material Require 1 for Examination Braidwood CURVE BOOK Figure 1 Question Source: New Question Modification Method:
Question Source Comments: I Comment Type Comment I FridIy, September 4,1998 Page 22 of 100
.. . _ . . -
l Topic ouestion M Boron mixing Th9 following conditions exist on Unit 1: . ! - Reactor power was 95% prior to the event
- A turbine runback resulted in rod insertion with control gods in AUTOMATIC - Annunciator ROD BANK LO-2 INSERTION LIMIT (1-10-A6) is lit
Th3 operators initiated an emergency boration per BwOA PRI-2 " Emergency Boration" and have verified control rods are now withdrawing. Why does the operator energize the Pzr Backup Heaters? This action... c. ensures Pzr boron concentration equalization with RCS by increasing normal spray flo b. counteracts RCS cooldown due the boration by the additional heat from the backup heater c. prevents loss of Pzr level by increasing the volume of fluid maintained in the Pz d. guarantees. adequate subcooling margin is maintained by raising the saturation temperature of the i Pz Answer a Exam Level R Cognitive Level Comprehension Facility: Braldwood ExamDate: 9/14/98 KA: 004 K6.01 Ro Value: SRO Value: 3.3 Section: SYS RO Group: 1 SRO Group: 1 Sy; tem / Evolution Title: Chemical and Volume Control System KA Statement: Knowledge of the of the effect of a loss or malfunction on the following will have on the Chemical and Volume Control System: Spray / heater combination in PZR tu assure uniform boron concentration
%planation of swer Reference Title / Facility Reference Number Section/Page Revisio L BwOA Pri-2 Emergency Boration lesson plan 6 6 R: actor Makeup control system lesson plan 8 12 I
Material Required for Examination Number (s) n Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment I l
.
l , Friday, September 4,1998 Page 23 of 100 l !
Que uon k Recire intirties to S1 Pumps & CV Pumps Tha following conditions exist on Unit 1:
* - A LOCA has occurred - Actions of 1BwEP ES-1.3, ' Transfer To Cold Leg Recirculation, have been complete During alignment,1CV8804A, RH HX to CENT CHG Pumps isolation Valve, failed to open and could NOT be manually opene What is the status of the ECCS system? l c. The RHR discharge headers are cross-tied with only RHR Pump 1B running and supplying suction to l the Si pumps and Centrifugal Charging pumps from the B train connectio l l
b. The RHR discharge headers are cross-tied with both RHR pumps running and supplying suction to thi l SI pumps only from the B train connection. The Centrifugal Charging pumps are stoppe ! e. RHR Pump 1B is discharging through the B Train cold leg injection headers and supplying suction to the SI Pumps. RHR Pump 1A and the Centrifugal Charging pumps are stoppe I d. RHR Pump 1B is discharging through the B Train cold leg injection headers and supplying suction to l the Si pumps and Centrifugal Charging pumps. RHR Pump 1 A is discharging through the A Train cold leg injection header Answer d Exam Level B Cognitive Level Comprehension Facility: Braldwood ExamDate: 9/14/98 KA: 005 K1.12 RO Value: SRO Value: 3.4 Section: SYS RO Group: 3 SRO Group: 3 System / Evolution Residual Heat Removal System KA Knowledge of the physical connections and/or cause-effect relationships between Residual Heat Removal System and the following: Safeguard pumps alanadon of CL recirc lineup has any ONE running RHR pump aligned to provide suction path to all other ECCS pumps (Sl
. ,swer & CENT CHG). The discharge headers between RH trains are required to be separate so that the ONE running RH pump does not operate in runout condition.
Reference Title / Facility Reference Number Section/Page Revisio L O.
Emergency Operating Procedures l Loss of Reactor or secondary coolant / BwEP 1, BwEP ES 1.1- Chp 58 Emergency Core Cooling system Lcsson plan 10 5,7,8,14 Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment
'
Fndry, September 4,1998 Page 24 of 100
. . .- - ~ ~ - - -- . - . . - - - . - - ... .-. ..- Question Feilura of Hx Outlet Vilva Tha following conditions exist on Unit 1: . - Unit is in MODE 4 during cooldown per 1BwGP 100-5 following unit shutdown 38 hours ago - RCS temperature - 340*F_- . - RCS pressure - 345 psig ' - PZR level - 33% - RHR pump 1 A is operating in Shutdown Cooling mode - RH-618 A Hx Bypass Flow Control Valve is in MAN at 3000 gpm - RH-606 A HX Flow Control Valve controller demand is at 20% - CV-128 RHR Ltdn Flow Contr Valve demand is at 100% - PCV-131 is in AUTOMATIC set to maintain 350 psig A sig'nal failure from the controller causes RH-606 to go fully closed. What is the system response to this ,
failure without operator action? , e. PCV-131 will throttle open due to lower RH discharge pressur b. RCS pressure will increase due to RCS heatu e. Pressurizer level will decrease due to increased letdown flo d. RH-610 will throttle open due to lower RH flo Answer b Exam t.evel R Cognitive t.evel Application Facility: Braidwood ExamDate:^ 9/14/98 KA: 005 K4.10 RO value: 3.1 . SRO value: 3.1 section: SYS RO Oroup: 3 SRO Group:. -3 System / Evolution Residual Heat Removal System
*\ . Knowledge of Residual Heat Removal System design feature (s) and or interlock (s) which provide for the following:
Control of RHR heat exchanger outlet flow
- Explanation of RCS pressure will rise as fluid temperature increases due to loss of cooling flow through HX. IF flow Answer decreases system pressure downstream may decrease this will cause PCV-131 to throttle close in an attempt to raise pressure Reference Title / Facility Reference Number Section/Page Revisio L RHR Cooldown/ RH-1 Schematic RH-1 1 Chp 18 Residual Heat Removal system 7 3,4,5,9 Material Required for Examination Question Source New Question Modification Method:
Queetion Source Comments: CommentType Comment
-
Fnday, September 4,1998 Page 25 of 100
.- . _. . . .. . . _ _ _ _ _ _ _ _ . _ .. _ . . _ . _ _ _ . _ _ . - _ _ _ _ . ._
ouestion M Syst:ms response to SI/ Actions Th3 following conditions exist on Unit 1: l - A plant heatup is underway
- MODE 3 has just been entered l - RCS pressure 450 psig i SI Accumulator 1C was drained below required level during the outage for repair work. System configuration has NOT allowed refilling the Accumulator until now. The SI Accumulator line is being flushed in accordance with BOP SI-14 "Si Accumulator Fill Line Flush" (Valve lineup includes: 1S1-8964, Si Test Lines to Radwaste Isolation Valve, and SI-8888, St Pps to Accumulator Fill Valve, are open.1SI 8821A, St Pump to Cold Leg Isolation Valve, and 1SI 8802A, Si to Hot Leg 1 A & 1D isol valve are closed). Si pump 1A running. During the flushing, an inadvertent Si signal is generate What is the status of the ECCS based on the current alignment without operator action?
e.1B Si pump injection flow is directed to the RCS cold legs and 1 A SI pump flow is directed to the Accumulator 1C fillline flus b.1 A Si pump flow is directed to the 1C Accumulator fill line flush and 1B St Pumps is in PULL-TO-LOC c. BOTH SI pump flows are directed to the RCS cold legs and to the Accumulator 1C fill line flus d. BOTH SI pump flows are directed to the RCS cold legs ONL Answer a Exam Level B Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 006 A2.13 RO Value: 3.9 SRO Value: 4.2 Section: SYS RO Group: 2 SRO Group: 2 stem / Evolution Emergency Core Cooling System KA Ability to (a) predict the impacts of the following on the Emergency Core Cooling System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Inadvertent SIS actuation Explanation of SI pumps are operable; Sl8821 A remains closed; Sl8888 and Sl8964 remain ope Answer Reference Title / Facility Reference Number Section/Page Revisio L Plant Heatup BwGP 100-1 F.49 pg 30 11 St Accumulator Fill Line Flush BwOP SI-14 6 Chp 58 Emergency Core Cooling system Lesson plan 10 6,9 Meterial Required for Examination Question Source: New Question Modification Method: Ques 7on Source Comments: Comment Type Comment
.
l Fridry, September 4,1998 Page 26 cf100
,_ ,
. _ _ __ _
l Question N 10CFR50.46 Design Crit:ria l To meet tha 10CFR50.46 crit:ria, tha ECCS Systsm is designed such that under accidsnt conditions it i will maintain... i ' e. total hydrogen production from zirconium-water reaction below maximum value of 5%. b. maximum fuel temperature at the inside surface of the cladding NOT to exceed 2000* c. the core at least 5% dK/K shutdown to prevent an inadvertent return to criticalit d. fuel clad oxidation less than 17% of total clad thickness anywhere within the cor l Answer d Exam Level B Cognitive Level Memory Facility: Braidwood ExamDate: 9/14/98 KA: 006 K3.02 Ro Value: 4.3 SRO Value: 4.4 Section: SYS RO Group: 2 SRO Group: 2 Cystem/ Evolution Emergency Core Cooling System KA Knowledge of the effect that a loss or malfunction of the Emergency Core Cooling System will have on the following: Fuel 1 Explanation of Third selection addresses de. sign criteria for reactivity control per CT Answer Reference Title / Facility Reference Number Section/Page Revisio L CFR50/ 47 Chp 58 Emergency Core Cooling system L:sson plan 10 2 l heatorial Required for Examination ' Question Source: Facility Exam Bank Quastion Modification Method: Editorially Modified Question Source Comments: Comment Type Comment ) l l l i l _ i ! I i . i Friday, September 4,1098 Page 27 of 100
,
Question Evtluation cf flow ECCS pumps The following ' conditions exist on Unit 1: .
- A LOCA has occurred ,
!
_1B SI pump trips and cannot be restarted ) l - Transfer to Cold Leg recirculation is required l ! - RCS pressure is approximately 50 psig -
. What is the approximate total SI pump flow indicated on the main control board and how will this value chinge following transfer of BOTH trains of ECCS to cold leg recirculation? l
<
Total Flow Flow Change i e. 400 gpm Decrease b. 650 gpm increase e, 800 gpm Decrease d.1300 gpm
.
Increase Answer b Exam Level B Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 006 K6.03 RO Value: 3.6 sRO Value: 3.9 section: SYS RO Group: 2 SROGroup: 2 system / Evolution Emergency Core Cooling System KA Knowledge of the of the e#ect of a loss or malfunction on the following will have on the Emergency Core Cooling System: Safety injection Pumps - Explanotion of SI pump design values provid e for 650 gpm flow per pump @ 1300 psig and 1300 gpm @ 600 psig (or less).
answer The flow from the pumps increases since the RH pumps are now providing a suction pressure of approximately 250 psig to the pumps instead of the lower pressure (30 psig or less) provided by the head associated with RWST leve Reference Title / Facility Reference Number Section/Page Revisio L Chp 58, Emergency Core Cooling System Lisson plan . 10 -3,8a i Material Required for Examination j Question Source New Question Modification Method: I Question source Comments-
Comment Type Comment
'
l _
!
l* l l l ! l . Fridry, September 4,1998 Page 28 of 100
.
. _ . . _ . . _ . . ,_._. _ _ _ . ._ _ . _ ._ ..._. _ _ . _ _ Que. mon : M f PRT conditions cau:ing stitrm/r:sponse During shift turnovar for Unit 1, the NSO not:s tha following paramatsrs: -
l RCS Tave - 566.5'F
:r pressure - 2235 psig Pzr level - 38.3%
PRT pressure -- 4 psig PRTlevel '74% PRT temperature '- 98*F One hour later when annunciator 1-12-A7, PRT LEVEL HIGH LOW alarmed, the NSO notes the following , pirameters:
'
l RCS Tave - 566.2*F l l Pzr pressure - 2233 psig l ' Pzrlevel - 38% i PRT pressure - 5.9 psig l PRT level -- 81% I l .PRT temperature - 96*F-l Wh .t condition resulted in the change in parameters? ; l- '
.
l e. PRT PW Supply inside Cnmt isol Valve RY-8030 opene b. PRT to GW Comp Isol Valve RY-469 failed close c. CVCS letdown relief valve CV-8117 lifte .
d. PORV RY-455A opened and reclose j l Answer a Exam Level R Cogniuve Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 2.4.50 RO Value: 3.3 SRO Value: 3.3 Section: SYS RO Group: 3 SRO Group: 3 ) Sy tenVEvolution . Pressurizer Relief Tank / Quench Tank System l KA-Ability to verify system alarm setpoints and operate controls identified in the alarm response manua Explanation of The only input provided that would give a level increase and a temperatue decrease is the makeup from P Answer i Reference Title / Facility Reference Number Section/Page Revisio L O.
, Pressurizer Relief Tank Filling and Venting L BwOP RY-3 3 i
'
i PRT Level High low / BwAR 1-12-A7 51E1
- - Chp 14 Pressurizerlesson plan 9 13,14 Material Required for Examination -
' Question Sourc New Question Modification Method: Editorially Modified Question Source Comments * Ginna 9/90 NRC Exam - l Comment Type Comment
-
i .
i ! l Fnday. September 4,1998 Page 29 of 100
- . ~ - . - - -. .
- . - - - - . - . - - . - - - - - - . - . - . - ~ - . - - - ~ ~ . - - . . . _ .
-
3o l Question Topic Determination of effect of valve positioning
' lit 1 is operating at 100% power in MOL conditions. All systems are functioning normally with rod control in manua What is the effect on plant operations if instrument air supplied to the CVCS letdown Hx component cooling water outlet valve, TGV-CC-130 is lost?
TGV-CC-130 goes fully... c. shut and reactor power decreases due to boration in the CVCS demineralizers.
l b. shut and the CVCS demineralizers are automatically bypassed on temperature signal.
t e. open and reactor power increases due to deboration in the CVCS demineralizers.
( d. open and the CVCS demineralizers are automatically bypassed on temperature signa l i 1 Answer C Exam Level R Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 i KA: 008 A2.05 RO Value: 3.3 sRO Value: 3.5 section: SYS RO Group: 3 SROGroup: 3 ) sy: tan / Evolution Component Cooling Water System i KA Ability to (a) predict the impacts of the following on the Component Cooling Water System and (b) based on those predictions, use l ' procedures to correct, control, or mitigate the consequences of those abnormal operation: Effect of loss of instrument and control air on the position of the CCW valves that are air operated Explanation of The CVCS letdown flow is overcooled and will give up boron to the resins in the CVCS demins (until a new Answer equilibrium value of boron reached in demins).
l r l Ra%ronce Title / Facility Reference Number Section/Page Revisio L. i . of Instrument Air /18wOA Sec-4 Table A l Component cig 2 Ch15n CVCS tesson plan 10 10,14 ' S:rvico Air / Instrument Air Lasson plan review quest 14 8 9 i ! Meterial Required for Examination l Question Source: New Question Modification Method- l Question Source Comments: Comment Type Comment ;
'
~
l i ! i l
I Friday, September 4,1998 Page 30 of 100
.- , , . . . . . -,. - _
. .- . - - - . - . . _ _-- .- . - - - . . _ -
. l Question k Spr:y using Normal end Aux Spray ' Whr.t cra tha param:ters and values us:d by the optrator to ensure the tempcraturo differance between
- ths PZR and the spray fluid are within the specified limit (s) in the PRESSURE AND TEMPERATURE l LIMIT REPORT when initiating PZR spray? l l a. For normal spray, the difference between RCS hot le0l oop temperature and PZR vapor space l temperature limit is 50*F, and for aux spray, the difference between Regenerative Hx charging inlet temperature and PZR vapor space limit is 320* b. For normal spray, the difference between RCS cold leg loop temperature and PZR vapor space 1 temperature limit is 50*F, and for aux spray, the difference between Regenerative Hx charging outlet l
- temperature and PZR vapor space limit is 320*F, c. For normal spray, the difference between RCS hot leg loop temperature and PZR vapor space temperature limit is 320*F, and for aux spray, the difference between Regenerative Hx charging inlet temperature and PZR vapor space limit is 320* !
d. For normal spray, the difference between RCS cold leg loop temperature and PZR vapor space temperature limit is 320*F, and for aux spray, the difference between Regenerative Hx charging outlet temperature and PZR vapor space limit is 320* Answer d Exam Level B cognitive Level Memory Facility: Braidwood ExamDate: 9/14/98 KA: 010 A1.08 RO Value: 3.2 SRO Value: 3.3 Section: SYS RO Group: 2 SRO Group: 2 Sy; tem / Evolution Pressurizer Pressure Control System KA Ability to predict ana/or monitor changes in parameters associated with operating the Pressurizer Pressure Control System controls including: Spray nozzle DT Explanation of Answer i ence Title / Facility Reference Number Section/Page Revisio L. O.
l Pr:ssurizer Temperature Limit Surv/ 18wOS 4.9.2-1 Pr:ssurizer Spray Water Temperature Differ:ntial Limit surv/18wOS 4.9.2-2 18wGP 100-1 Plant heat up lesson plan 12 1,2,3 Chp 14 Pressurizer lesson plan 9 7,8 Material Required for Examination Question Source: New Question Modification Method: Signifcantly Modified Question Source Comments: Kewaunee 2/94 NRc Exam Comment Type Comment !
-
.. l l ! l l
! Fridry, September 4,1998 Page 31 of 100 l
. _ . ._ _ __ _ _ _ ~
Question 3d Ev;lu" tion of Pzr conditions ! The following conditions exist on Unit 1: - l l A load reject from 100% power has occurred l - Reactor power - 80% l - Pzr level - 56% l - Pzr vapor temperature - 655'F l - Pzr liquid temperature - 653*F '
- RCS Tave - 578'F l Whit is the current status of the Pressurizer based on given conditions? !
l l c. Backup and proportional heaters are fully o b. Proportional heaters are modulated o l l c. Pzr spray valves have modulated ope d. Pzr spray valves and Pzr PORVs are ope Anewer C Exam Level B Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 010 K5.01 RO Value: 3.5 SRO Value: 4.0 Section: SYS RO Group: 2 SRO Group: 2 System / Evolution Pressurizer Pressure Control System KA Knowledge of the operationalimplications of the following concepts as they apply to the Pressurizer Pressure Control System: Determination of condition of fluid in PZR, using steam tables Explanation of At 655'F, saturation pressure is 2272 psig. At this pressure, with current PZR level deviation <5% of program Answr level (53%), the sprays are the only component "on".
Reference Title / Facility Reference Number Section/Page Revisio L P oressure Control / RY-2 Pzr Pressure . Setpoints l Chp 14 Pressurizer lesson plan 9 5,6,7 Steam tables Saturation table Material Required for Examination Steam Tables Question Source: Facihty Exam Bank Question Modification Method: Concept Used Question Source Comments: Braidwood 1997 NRc exam Comment Type Comment
. '
t l Friday, September 4,1998 Page 32 of 100
.
. . . - - .-- .- - - - _ . . . - - -- . . - . -
Qumon D Pzr Lzv:1 RIactor Trip The following. conditions exist on Unit 1 with all controls in normal linsup: -
- Reactor power - 30% stable - RCS Tave - 564.5'F - Pzr precsure - 2230 psig . - Pzr level - 36% (Ll-459), 37% (Ll-460), 36% (Ll-461) - Pzr LVL CONT CH SELECT is in 459/460 position The pressurizer level controller 1LK-459 output fails low. What automatic actions will occur as a result of - this failure assuming NO operator action taken?
c. Pzr level will NOT change due to LT-460 being the controlling channel b. The reactor will trip on high Pzr level due to letdown isolatio e. Pzr level will control at 25% due to low output from the controlle d. Pzr level will control at 60% due to low output from the controlle ! Answer b Exam Level B Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98
. KA: 011 K1.04 RO Value: 3.8 SRO Value: 3.9 Section: SYS RO Group: 2 SRO Group: 2 SystemIEvolution Pressurizer Level Control System KA Knowledge of the physical connections and/or cause-effed relationships between Pressurtzer Level Control System and the
.RPS
. Explanation of NOTE that this failure is lik: the failure of the controlling level channel high in that charging flow falls to Answer minimum. At 17% level, letdown isolates charging continues at minimum (52 gpm) and Pzr level rises to high level trip setpoint.).
)
P^rence Title / Facility Reference Number Section/Page Revisio L. i.svel Control schematic RY-3 Pzr level setpts 2 Chp14 Pressurizer lesson plan 9 21 Material Required for Examination Question Source: Facility Exam Bank Question Modification Method: Significantly Modified Question Source Comments: l Comment Type Comment
.
j Friday, September 4,1998 Page 33 of 100
. ___ __ _ . .. . . _ _ _ m _ .. __ . _ _ . _ _ _ . . _ - _ _ _ _ _
Question Oper: tion of BOTH Bypa:s Trip Br:aksrs The following conditions exist on Unit 1: - ' "
- Mode 3 NOT NOP with reactor trip breakers (RTA and RTB) closed '
Testing of reactor trip bypass breakers underway
- Reactor bypass breaker B (BYB)is racked in and closed
- - A7 operator begins to perform test with reactor bypass breaker A (BYA).
What occurs as the operator operates the breaker BYA7 When reactor bypass breaker BYA is... c. locally closed, breaker BYB will trip. RTA and RTB remain close n. racked in to the CONNECT position, breaker BYB will trip. RTA and RTB remain close e. locally closed, all reactor trip and bypass breakers will tri d. is racked in to the CONNECT position, all reactor trip and bypass breakers will tri Answer C Exam Level ' R Cognitive Level Memory Facility: Braidwood ExamDate: 9/14/98 KA: 012 A3.07 RO Value: 4.0 sRO Value: 4.0 section: SYS RO Group: 2 SROGroup: 2 . System / Evolution Reactor Protection System KA Ability to monitor automatic operations of the Reactor Protection System includin , Trip breakers 1 m Explanation of Closure of the second BYB results in SPSS generating a GENERAL WARNING on both trains which would l Answer open all trip and bypass breaker Reference Title / Facility Reference Number Section/Page Revisio L '
setpoints Schematic EF-2 Rx Trip Byp brkr trips 5 Ch 603 SSPS lesson plan 3 6,9 ' Material Required for Examination Question Source: Facility Exam Bank Question Modification Method: Editorially Modified Question Source Comments: Comment Type Comment
~
,
Friday, September 4,1998 Page 34 of 100
- _ . . _ __ _ _ _ _
Question 3 [ Input th t can be bypass & condition The following conditions exist on Unit 2: -
- Unit shutdown is in progress
- Reactor power - 20%
- RCS Tave - 562*F
'
- Pzr pressure - 2235 psig
- Pzr level - 32% ,
- First stage turbine pressure channel PT-506 fails high j What affect does this failure have on operations as unit shutdown is continued, if NO action is taken for the !
channel failure? e. At 10% power, the reactor will trip if the SR MAN BLOCK switches are taken to RESE l b. At 9% power, the reactor will trip if an RCP trip e. At 7% power, the reactor will trip if the TURBINE TRIP pushbuttons are depresse d. At 5% power, the reactor will be manually tripped as required during a normal shutdow Answer d Exam Level B Cognitive Level Comprehension Facility: Braldwood ExamDate: 9/14/98 KA: 012 A4.03 RO Value: 3.6 SRO Value: 3.6 Section: SYS RO Group: 2 SRO Group: 2 Sy; tem / Evolution Reactor Protection System KA Ability to manually operate and/or monitor in the control room: Channel blocks and bypasses Explanation of PT-506 failure results in P13 interlock NOT clearing when turbine power fa!'s below 10%. This also feeds into Answer P7 AT POWER TRIPS *' interlock also remains active. Trips affected: 1) 2 loop loss of flow,2) Pzr low press, 3) Pzr high level,4) RCP brkr open,5) RCP UV,6) RCP UF. At 10% power, the SR NIS should still be auto blocked by P-10 (active). The turbine is normally tripped from ~65 Mwe at 5% power per BwGP.
Reference Title / Facility Reference Number Section/Page Revisio L. O.
Power Descension /1BwGP 100-4 note step F.27 16 ESF Setpoints/ Schematic EF-1/ Permissive Rx Trip 4 Ch60b/ Reactor Protection system 6 4 Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment _ Friday, September 4,1998 Page 35 of 100
Topic ouestion 3 $ OTdT inputs & cffect of ching:s The following conditions exist on Unit 1: -
- Power range NIS reading - 100% - Tcold - 553*F ' - Thot - 608*F - RCS total flow - 372,000 gpm - Pzr pressure - 2245 psig - Pzr level - 69%
How does the setpoint for Over Temperature Delta-T (OTdT) change when a listed parameter is ch nged? (Consider each change individually) ! The setpoint... I c. Increases if Power range NIS output rises to 102%. , I b. decreases if total reactor flow increases to 370,000 gp I c. increases if pressurizer pressure decreases to 2235 psi d. decreases if the Thot rises to 612* Answer d Exam Level R Cogniuve Level Comprehension Facility: Braldwood ExamDate: 9/14/98 KA: 012 K5.01 RO Value: 3.3 SRO Value: 3.8 Section: SYS RO Group: 2 SROGroup: 2 Title: System / Evolution Reactor Protection System Statement: KA Knowledge cf the operationalimplications of the following concepts as they apply to the Reactor Protection System: DNB planation of a - NIS input is only for exceeding +/- delta-l; b - Flow affects when DNB occurs, but is NOT an input to OTdT; Answer c - Pressurize rise increases OTdT. Thot input to dT power for OTdT determination Number (s) n Reference Title / Facility Reference Number Section/Page Revisio L l ESF Setpoints/ EF-2 OTDT S CH 60bl RPS lesson plan 6 3,4 Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment
-
Frid:y, September 4,1998 Page 36 of 100
. .- ._, . - - . - - - - . - . - -.
Qu.etion 37 CNMT SprCy/Ph se B A hutup is in progr:ss on Unit At 0700, the following conditions are noted:
- RCS pressure - 1750 psig - RCS temperature - 480*F - S/G pressures - 565 psig At 0730, the following conditions are noted: - RCS pressure - 1850 psig - RCS temperature - 485'F - S/G pressures - 593 psig If the current trend continues, the FIRST event that the operators should expect to see is the...
e. Pzr PORVs open b. MSIVs close c. Pzr sprays open, d. S/G PORVs open Ans wr b Exam Level B Cognitive Level Comprehension Facility: BraidwoM ExamDate: 9/14/98 Tier: Plant Systems RO Group: 1 sRO Group: 1 013 Engineered Safety Features Actuation System
. Knowledge of Engineered Safety Features Actuation System design feature (s) and or interlock (s) which provide for the following:
K4.03 Main Steam isolation System 3.9 Explanation of RCS (Pzr) pressure rises above the P-11setpoint (1930 psig), which provides permissive for Sl/ Main Steam Anser Line isolation on low S/G pressure, and S/G pressure is less than MSL isolations.
Reference Title / Facility Reference Number Section/ Pace Revisio L O.
ESF Setpoints/ EF-2 CS/ Phase B sig S CS/ MCB indications / CS-1, CS-2 CS Actuation sig 3 Chp 61 ESF lesson plan 5 7,8 Material Required for Examination # Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment
.
Friday, September 4,1998 Page 37 of 100
ousation - FWlsolation - P14 The following conditions exist on Unit 2: .
- RCS temperature 340*F - RCS pressure - 900 psig - '
l
- All MSIVs for the S/Gs are closed-The MSIV Bypass valves are open - The FW-035s, Feedwater Tempering Isolation Valves, are open - The FW-034s, Feedwater Tempering Flow Control Valves, are closed (opened periodically for level control) - Feedwater pump 2C is reset and latched on turning gear - The Start Up Feedwater pump is running .
The level in the S/G 2B rises to 90%. How is the plant affected? c. No actuation occurs because of the position of the MSIV b. The 2C Feedwater pump and Start Up Feedwater pump tri c. The 2C Feedwater pump trips and FW-035 valves close, d. The 2C Feedwater pump and Start Up Feedwater pump trip, the FW-035 valves close, and the MSIV Bypass valves clos Answer Exam Level R cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 MA:. 013 K4.13 ~ RO Value: 3.7 SRO Value: 3.9 - Section: SYS RO Group: 1 SRO Group: 1 System / Evolution Engineered Safety Features Actuation System l
KA ~ Knowledge of Engineered Safety Features Actuation System design feature (s) and or interlock (s) which provide for the following: MFW isolation / reset planation of Having Loop Isolation Stops closed does not defeat P-1 Answer Reference Title / Facility Reference Number Section/Page Revisio L O.
Feedwater simple / FW-1 FWI signals 4 SGWLC/ FW-2 S/U Flowpaths 0 i j Chp61 ESF lesson plan 5 7 Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Coment Type Comment l _
.
i i
!
.. i
!
Frid:y, September 4.1998 Page 38 of 100 i
. _ . . ._ _ _ .._ _ ___ . _ . . ._ _ _ . _ . - _ - _ _ _ . _
oue. tion 3[ ROD BOTTOM Altrm operation During a reactor startup, whnn does the ROD AT BOTTOM alarm becomo activa for each control bank? Ths alarm will actuate for a dropped rod for... c. any Control Bank whenever Control Bank A DRPI output is above 9 step b. each Control Bank whenever that Control Bank demand positi- i is above 3 steps, e.'each Control Bank whenever that Control Bank DRPI output is above 9 step d. Control Banks A, B and C whenever their Control Bank demand position is above 9 steps, and for Control Bank D whenever Control Bank D demand position is above 3 step Answer C Exam Level Cognitive Level Memory Facility: Braidwood ExamDate: 9/14/98 KA: 2.4.31 RO Value: 3.3 SRO Value: 3.4 Section: SYS Ro Group: 2 SRO Group: 1 Sy:temrEvolution Rod Position Indication System KA Knowledge of annunciators alarms and indications. and use of the response instruction Explanadon of Note that the ROD BOTTOM comes direfctly from the DRPI unit with a setpoint of 9 steps; the alarm actuates Answer when rod position is detected at 3 steps (or less).
Reference Title / Facility Reference Number Section/Page Revisio L ROD ct Bottom /1BwAR 1-10-E6 2 Chp 29 rod Position Indication sys L sson plan 9 4,5 Material Required for Examination Nestion Source: New Question Modification Method: Significantly Modded
,estion Source Comments: Millstone 311/90 NRC Exam j Comment Type Comment I d
l l l
~
i
!
, Friday, September 4.1998 Page 39 of 100
--
, _ . . - . . - - - . .. - . . . - --
ouestion N SR NIS discriminitor failura How would tho failura of tha puiss height discriminator to a low valus affect the indication of tha affected i Source Range channel? ' a output would increase due to... l c. electronic filtering which narrows the pulse height windo l b failure in removing the higher amplitude neutron generated pulse l c. increased gamma interaction inside the detecto d. counting of the gamma generated pulses and decay-alpha generated pulse Answer d Exam Level B Cognitive Level Memory Facility: Braidwood ExamDate: 9/14/98 KA: 015 A2.02 RO Value: SRO Value: 3.5 Section: SYS RO Group: 1 SRO Group: 1 System / Evolution Nuclear Instrumentation System KA Ability to (s) predict the impacts of the following on the Nuclear instrumentation System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnomal operation: Faulty or erratic operation of detectors or compensating components Explanation of Pulse height discriminator used to set window to detect those pulses with energy level high enough to be from Answer event associated with neutron detection. Gamma and other interactions such as the alpha decay of fission product daughters is of lower heigth (energy) and disciminator normally electronically removes.
Reference Title / Facility Reference Number Section/Page Revision L O.
Source Range Detector schematic NI-4 4 Chp 31 Source Range NuclearInst 6 3 l Material Required for Examination astion Source: New Question Modification Method: Question Source Comments: Comment Type Comment l l l
,
Fridry. September 4.1998 Page 40 of 100
. - - - _- - - - - _ - - . - - . - .- - ..- ..
p ' Que. con Y[SR NIS -loss of control powtr Ths following conditions exist on Unit 1: .
- RCS at NOT NOP - Reactor trip breakers - closed
, '
- Source Range readings:
N31 - 18 cps N32 - 22 cps l l Wh:t indication would the operator observe if Control Power was lost to the N31 Drawer? The N31 meter would read... c. downscale, the associated drawer bistable lamps NOT lit, and reactor trip breakers closed.
l l b. downscale, the associated drawer bistable lamps lit, and reactor trip breakers ope e.18 cps, the associated drawer bistable lamps NOT lit, and reactor trip breakers close d.18 cps, the associated drawer bistable lamps lit, and reactor trip breakers ope Answer d Exam Level B cognitive t.evel Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 015 K2.01 RO Value: 3.3 SRO Value: 3.7 Section: SYS RO Group: 1 SRO Group: 1 System / Evolution Nuclear Instrumentation System KA Knowledge of electrical power supplies to the following: NIS channels, components, and interconnections Explanation of Control power loss affects bistables which trip but NOT drawer instrument indication which is from instrument l
* w er Power source.
l- Reference Title / Facility Reference Number Section/Page Revision L O.
' Source Range Detector Schematic NM loss of Control power 4 Ch 31 Source Range Nuclear Inst 6 Material Required for Examination l i Question Source: New Question Modification Method: Question Source Comments: i l Comment Type Comment i !
'
.
f l i l Friday, September 4,1998 Page 41 of 100 l
. .. . - -
'ouestion h Ev;l for 1/M - Eightfold incrase The following conditions exi t on Unit 1: - - A reactor startup is about to be performed - All shutdown banks are fully withdrawn ; - All control banks are fully inserted ) - An ECC records the following: 1 Predicted Critical Position (ECP) - 130 steps on CBD Max rod position - 231 steps on CBD Min rod position - 58 steps on CBD The following parameters were recorded during the rod withdrawal:
l ROD HEIGTH - N31 cps N32 cps O on CBA 25 23 178 on CBA 34 31 178 on CBB 80 82 178 on CBC 200 162 80 on CBD 237 184 92 on CBD 260 245 Whrn was the first time the operator was required to determine the Predicted Critical Position? e. At 50 steps on CBA, with N32 as the designated Source Range detecto t At 47 steps on CBC, with N31 as the designated Source Range Detector, c. At 178 steps on CBC, with N31 as the designated Source Range detecto d. At 80 steps on CBD, with N32 as the designated Source Range detecto Answer c Exam Level R cognitive Level Comprehension Facility: Byron ExamDate: 9/14/98 Tier: Plant Systems RO Group: 1 SRO Group: 1 015 Nuclear Instrumentation System K Knowledge of the operational implications of the following concepts as they apply to the Nuclear Instrumentation System: K5.06 Suberitical multiplications and NIS indications 3.4 Explanation of Durir g reactor SU, hold point for ICRR determination is performed for each Control Bank at 50 steps Answer withdrawn. The actual determination of Predicted Critical Position is required at the eight-fold count increase on highest reading SR. Holdpoint occurs on CBC @178. $& steps on CBC is the 0% power Rll value.
Reference Title / Facility Reference Number Section/Page Revision L O.
1BwGP 100-2 Reactor Startup 1BwGP 100-2A1 12 1BwGP 100-2A1 Lesson plan 13 2 Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Fridzy, September 4,1998 Page 42 of 100
. _ , .- - - -- -
Question . NR RTD Fritura effects The following conditions exist on Unit 1: -
- Reactor power - 50%
- RCS Tave - 570*F (A); 569'F (B); 569'F (C); 570*F (D) ,
'
l- - RCS Thot - 585'F (A); 584*F (B); 583*F (C); 585*F (D)
- RCS Tcold - 555*F (A) 554*F (B); 555*F (C); 555'F (D) - Pzr pressure - 2235 psig - Pzr level - 43 %
If loop B Thot output channel fails LOW, what is the response of Pzr level ? L ! Pressurizer level will... a. increases to 60%. ! i ! b. remains the same, e. decreases to 25%. d. decreases to the letdown isolation setpoin ; Answer b Exam Level U Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 016 K3.02 RO Value: 3.4 SRO Value: 3.5 Section: SYS RO Group: 2 SRO Group: 2 System / Evolution Non-Nuclear Instrumentation System KA Knowledge of the effect that a loss or malfunction of the Non-Nuclear instrumentation System will have on the following: PZR LCS Explanation of Tnot fails to 510*F. With loop Tcold of 537'F, loop Tave is now 524*F. Auctioneered HIGH Tave is used for Pzr swer - level progra Reference Title / Facility Reference Number Section/Page Revision L. PZR L' vel Control Schematic RY-3 2 1BwOA Inst-2 lesson plan 15 1 chp 12 RCS lesson plan 8 13 Material Required for Examination Question Source: Facildy Exam Bank Question Modification Method Concept Used Question Source Comments: Zion 2/92 NRC Exam (along with several others). Change includes failure of Thot loop, failure low and conditions instead of dual conditio Comment Type . Comment I !~ Friday, September 4,1998 F' age 43 of 100
.
i oue.uon f CETC frilurnffect on Subcooling Monitor /lconic Display With Unit 1 et 100% pow:r and with normal operating parametsrs, how would tha failuro of the HOTTEST Cora Exit Thermocouple affect the reading of subcooling margin on the SPDS Iconics (CETC/SMM display) for each of the two situations below: Situation 1 - The CETC output fails high slowly Situation 2 - The CETC output fails low slowly e. Situation 1: Subcooling margin will decrease to saturation then indicate superheated, and return to l normal when CETC output reaches 2300* Situation 2: Subcooling margin will increase, then stabilizes when the CETC output is smaller than TEN other TC b. Situation 1: Subcooling margin will decrease to saturation then indicate superheated, and return to normal when CETC output reaches 1200* Situation 2: Subcooling margin will remain constan e. Situation 1: Subcooling margin will increase to saturation then indicate superheated, and return to normal when CETC output reaches 1200* Situation 2: Subcooling margin will decrease, then stabilizes when the CETC output is smaller than l TEN other TC l
d. Situation 1: Subcooling margin will increase to saturation then indicate superheated, and return to normal when TC output reaches 2300* Situation 2: Subcooling margin will remain constan Answer a Exam Level R cognitive Level Comprehension Facility: Braldwood ExamDate: 9/14/98 KA: 017 K4.01 Ro Value: 3.4 SRo Value: 3.7 Section: SYS RC Group: 1 SRo Group: 1 stem / Evolution In-Core Temperature Monitor System KA Knowledge of in-Core Temperature Monitor System design feature (s) and or interlock (s) which provide for the following: Input to subcooling monitors Explanation of Fail high - Since it is initially tha highest, its input will remain active in average until high setpoint reached at Answer 2300*F. Fail low - subcooling margin will slightly increase as temperature falls and input to average remains valid. When it reaches the 11th highest value, the subcooling margin will stabilize and reamin constant (assuming other 10 inputs do not change).
Reference Title / Facility Reference Number Section/Page Revision L. Chrpter 34b inadequate Core Cooling Detection 7 5,6 Material Required for Examination Question Sourc Facility Exam Bank Question Modification Method: Significantiy Modified Question Source Comments: Braidwood 1997 NRC Exam. Differenz in all answer choices - similar premise in theory, but different vardin Comment Type Comment _ l
!
Fridry, September 4,1998 Page 44 of 100
_ . . .- - _ _ . . . - - - . - - . - . . - -. . -. . - . . Queouon RCFC operations requir m:nts The following conditions exist on Unit 2: -
- RCS Temperature - 342*F - Pzr pressure - 375 psig - 2A,2B, and 2D RCFCs are operating in high speed - Unit 2 RCFC Dry Bulb temperatures are recorded as follows:. - 2A RCFC - 119*F - 2B RCFC - 118*F
. - 2C RCFC - 127'F-2D RCFC - 121*F Which of the following identifies the equipment status and actions for the above conditions? c. RCFC 2C must be started because the average of ALL the RCFC temperatures exceeds the limi ] b. RCFC 2C must be started because ONE of the operating RCFCs temperatures is above the limi l
c. NO action is necessary because ALL temperatures are within the required limi j d. NO action is necessary because the average temperature of ALL operating RCFCs is below the limi I
1 Answer d Exam Level R Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 2.1.32 RO va'ce: 3.4 SRO Value: 3.8 Section: SYS RO Group: 1 SRO Group: 1 System / Evolution Containment Cooling System KA Ability to explain and apply a(I system limits and precaution l
~~olanation of Limits on CNMT temperature determined by average of temperatures for OPER TING RCFC outlet temp Reference Title / Facility Reference Number Section/Page Revision L l RCFC Start up 18 WOP VP-5 l U2 Mode 1,2,3 shiftly daily Op surv 2BwOS-0,1 1,2,3 chp 42 Containment Vent system lesson plan 4 6,10a Material Required for Examination Question Source: New Question Modification Methoi Question Source Comments:
Comment Type Comment "
.
Fridly, September 4,1998 Page 45 of 100
._.. _ _
_ _ . . . _ .- . - _ . ! Topic
'
- Question Sequence for securing CNMT Spray Ths following conditions exist on Unit 1:
!
- A LOCA has occurred i
- Transition has been made to BwEP ES-1.3 " Transfer To Cold Leg Recirculation" !
- Containment Spray actuated due to high containment pressure l
- All systems and components operating as expected l Wh:t conditions allow for termination of Containment Spray? ;
.
:
l o. ONE pump is stopped when containment pressure is less than 15 psig. The other pump is stopped when RWST LO-3 levelis reache ! b. ONE pump is stopped when containment pressure is less than 20 psig. The other pump is stopped after it has operated for a period of at least TWO hours
c. BOTH pumps are stopped when containment pressure is less than 15 psig and have operated for a period of atleast1WO hour d. BOTH pumps are stopped when containment pressure is less than 20 psig and RWST LO-3 level is reache Answer C Exam Level B Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 026 A2.08 RO Value: 3.2 SRO Value: 3.7 section: SYS RO Group: 2 SRO Group: 1 Sy; tem / Evolution Containment Spray System KA Ability to (a) predict the impacts of the following on the Containment Spray System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormat operation: safe securing of containment spray when it can be done) Explanation of Answer Title: Reference Title / Facility Reference Number Section/Page Revision L. Contrinment Spray Schematic CS-1/ CS term 3 Loss of Reactor or Sec Coolant /1BwEP-1 1B WOG-1B Ch 59 Containment Spray sys Lesson plan 6 14 Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment
.
! I I Fridiy, September 4,1998 Page 46 of 100
. _ .._ _ ___
Question . h Pump operation int:rlock The following conditions exist on Unit 1: - !
- LOCA is in progress - - Containment pressure - 17 psig - Containment Spray actuated due to high containment pressure - Containment Spray signal has been reset - The actions of BwEP ES-1.3 " Transfer To Cold Leg Recirculation" have been completed - Offsite power is then lost and the D/G output breakers have just closed onto ESF buses How are the Containment Spray Pumps re-started?
I c. The pumps will auto start 15 seconds following closure of the D/G output breaker :. The pumps will auto start 40 seconds following closure of the D/G output breakers, c. If the operator immediately places the CS & PHASE B ISOL switches for both trains to ACTUATE, the pumps will auto start 15 seconds following closure of the D/G output breaker d. If the operator immediately places the PP 1._ TEST switches for both pumps in TEST, the pumps will auto start 40 seconds following closure of the D/G output breaker Answer C Exam Level R Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 026 A4.01 RO Value: 4.6 SRO Value: 4.3 Section: SYS RO Group: 2 SRO Group: 1 l
!
System / Evolution Containment Spray System KA Ability to manually operate and/or mo'nitor in the control room: CSS controls
._xplanation of If the AUTO actuation input signal is absent and actuation input has been reset, manaul actuation is required Answer to get equiptment restarted following a LOSP.
Reference Title / Facility Reference Number Section/Page Revision L. O.
Chp 59 Containment spray sys lesson plan 6 8,9 Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment I
~
l l l Fridiy, September 4,1998 Page 47 of 100
-. -- .-. . . - - . .- . . - - .
Quencon Charcoal Filt:rs r:sponse to d:lugn Annunci: tor 0-33-C3, FILTER 1VP05FA TEMPERATURE HIGH, alarms in the Control Room whila 1VP02CA CNMT Charcoal Filter Fan is operating. The alarm condition is verified locall 'hich of the following describes the actions taken and/or the system response for the Containment ventilation System? e. The deluge valve FP244A will automatically open and the fan will automatically sto b. The control room operator will open the deluge valve FP244A and the local operator will then stop the fa e. The local operator will open the deluge valve FP244A and the fan will automatically sto d. The local operator will open the deluge valve FP244A and the control room operator will then stop the fa Answer c Exam Level R Cognitive Level Memory Facility: Braidwood ExamDate: 9/14/98 MA: 027 A4.03 Ro value: 3.3 SRo value: 3.2 Section: SYS RO Group: 3 SRo Group: 2 i Sy; tem / Evolution Containment lodine Removal System l KA Ability to manually operate and/or monitor in the control room: I CIRS fans Explanadon of Operation of fp components associated with charcoal filter is local. But fan trips when deluge system Answer activate Reference Title / Facility Reference Number Section/Page Revision L. Filt:r 1VP05FA Temperature High
/1Bw:R 1VP01J-1-A1 1 chp 42 Containment vent 7 purge 4 8 Material Required for Examination Question Source: New Question Modification Method:
Question Source Comments: Commerit Type Comment l
.
l Fridzy, September 4,1998 Page 48 of 100 l l
_ _ __ . _ . . _ . . . _ _ _ . _ _ _ . _ _._._ . _ _ . _ . _ _ _ _ l Question k RWST Purification Loops The following conditions exist: - i i - Unit 1 - 20% power with load increase in progress
- Unit 2 - MODE 5 following refueling outage - Unit 2 Spent Fuel Pool Cooling Loop is in servic Spent Fuel Pool Pump 1FC01P is OO i Which of the following is allowed under this situation? - Alignment and operation of...
c. both Unit 1 RWST purification and Unit 2 RWST purification with flow through the Unit 2 Spent Fuel ! Pool Demineralizer and Unit 2 Spent Fuel Pool Filte n. Unit 1 Spent Fuel Pool purification and Unit 1 RWST purification with flow through the Unit 1 Spent l Fuel Pool Demineralizer and Unit 1 Spent Fuel Pool Filter, e. Unit 2 RWST purification with flow through the Unit 1 Spent Fuel Pool Filter and return to Unit 2 I RWS d. Unit 2 RWST purification with flow through the Unit 2 Spent Fuel Pool Demineralizer and Unit 2 Spent ! Fuel Pool Filte Answer d Exam Let R Cognitive Let Memory Facility: Braidwood ExamDate: 9/14/98 l l ~KA: 033 K1.05 RO Value: 2.7 SRO Value: 2.8. Section: SYS RO Group: 2 SRO Group: 2 l System / Evolution Spent Fuel Pool Cooling System 4 l KA Knowledge of the physical connections and/or cause-effect relationships between Spent Fuel Pool Cooling System and the RWST ! _glanation of The lineup allows Unit 2 only to be used for Unit 2 RWST cleanup. Only one unit RWST can be aligned at Answer time due to common input path via Refueling Water Purification Pumps. With the cooling loop inservice only, l the Unit's RWST may be aligned through the same Unit's, demin and filter train. Simultaneous use of Demin/ filter for the same Unit's SFP and RWST is NOT allowed due to concerns of draining RWS Reference Title / Facility Reference Number Section/Page Revision L. S/U purification sys to purity or l Reciculate the RWST/ BwOP FC-7 7 I Fu:1PoolCooling Schematic FC-1 3 j Chp 51 Spent Fuel Pool Cooling , And Cleanup 5 3 ! Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment L
t Fnday, September 4,1998 Page 49 of 100 _ _ .. - __ ,. _ _ _ _ _ _ .
. . . . __ - _ .
i l Question St"cm Dump input malfunction ' The following conditions exist on Unit 1: -
- Reactor power was 65% when the turbine tripped 1
- An ATWS occurred
- The reactor tripped 15 seconds later when B reactor trip breaker was locally opened
- Reactor trip breaker A is failed closed
- RCS Tave - 559'F i
'
- Pzr pressure - 2255 psig l - Steamline header pressure - 1100 psig
- No controls other than control rods and boration controls have been operated Whit is the status of the Steam Dump valves? j Ste:m Dumps are...
c. modulated opon due to steam header pressur b. modulatr i open due to Tave above no-load Tav c. closed because Tave is NOT greater than 3*F above Tre d. Closed because the dumps are NOT arme Answer b Exam Level B Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 041 A3.02 RO Value: 3.3 SRG Value: 3.4 Section: SYS RO Group: 3 SROGroup: 3 Syltem/ Evolution Steam Dump System and Turbine Bypass Control , KA Ability to monitor automatic operations of the Steam Dump System and Turbine 9ypass Control including: RCS pressure, RCS temperature, and reactor power
.planation of The "A" reactor trip breaker provides the arming signal for dumps on normal reactor trip. Since "A" RTB is still Answer closed, the steam dumps respond to event like load rejection, with C-7 load rejection (10% load decrease in 2 minutes sensed on PT-506) arming the dumps. Since the "B" RTP was opened, the steam dump controller does operate on the plant trip controller (No load Tave compared to Auct Hi Tave).
Referenca Title / Facility Reference Number Section/Page Revisin L Ste:m Dumps / Schematic MS-4 4 Chp 24 Steam Dumps Lesson Plan 7 3,4 Material ReqWred for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment
.
l ! l ! Frid:y, September 4.1998 Page 50 of 100
- - . __ _ . . - - _ -- - - - -
Question .N Turbine Control response to Failed impulse Ch:nnil Ths following conditions exist on Unit 1: -
- Reactor power 28% - All systems normal
! - Turbine EHC Panel settings: Turbine REFERENCE DEMAND - 580 MW Turbine REFERENCE - 330 MW
- The GO pushbutton is LIT Whit would be the DEHC System response to a slow failure to ZERO for the turbine impulse pressure channel that feeds into the DEHC?
Turbine load will... I c. decrease until the difference between REFERENCE and impulse pressure exceeds 30%, the operator would then be alerted to select MANUAL contro b. decrease until the difference between REFERENCE DEMAND and impulse pressure exceeds 30%, i then load wili stabilize in MANUAL contro j e. Increase until the difference between REFERENCE and impulse pressure exceeds 30%, then load will stabilize in MANUAL contro d. Increase until the difference between REFERENCE DEMAND and impulse pressure exceeds 30%, the operator would then be alerted to select MANUAL contro A newer C Exam Level R cognitive Level Comprehension Facility: E! dwood ExamDate: 9/14/98
.: 045 K1.20 RO Value: 3.4 SRo Value: 3.6 Section: SYS Ro Group: 3 SRO Group: 3 Sy; tem / Evolution Main Turbine Generator System KA Knowledge of the physical connections and/or cause-effect relationships between Main Turbine Generator System and the Protection system Explanation of When the difference between actual load and turbine impulse pressure (IMP IN) channel exceeds, circuit Answer AUTO transfer impulse feedback to IMP OUT Reference Title / Facility Reference Number Section/Page Revision L TV/GV Control / Schematic EHC-3/ Impulse 1 l Chp 37a Main turbine Control And Protection 5 52 Material Required for Examination Question Source: New Question Modification Method:
Question Source Comments: Comment Type Comment .
i i ! Friday. September 4,1998 Page 51 of 100
. .- _ . . - _
Question S/G Level progr:m -low power The following conditions exist on Unit 1: - ( - Reactor power 35% l - All systems normal i Wh:t failure would cause an INITIAL decrease in feedwater flow to all S/Gs? e. Turbine first stage impulse pressure PT-505 fails lo b. Main steamline pressure PT-507 fails lo e. Turbine first stage impulse pressure PT-506 fails lo I d. Main feedwater header pressure PT-508 fails lo Answer b Exam Level B Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 MA: 2. RO Value: 3.7 SRO Value: 4.4 Section: SYS RO Group: 1 SRO Group: 1 System / Evolution Main Feedwater Syst KA Abilrty to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretatio Explanation of PT-507 felle low causes feed pump speed to decrease which reduces FW pressure. This would initially result Answer in a decrease of flow to all S/G Reference Title / Facility Reference Number Section/Page Revision L Fw EH controls / schematic EHC-6/ DP 1 Chp 27 SGWLC 6 16 Material Required for Examination estion Source: New Questson Modification Method: Question Source Comments: Comment Type Comment l l Frid;y, September 4,1998 Page 52 of 100
_ _ _ _ _ __ _ . - _ . . . _..__ ouestion N Effect of f ilura of S/G stram pressurs chann 1 Th3 following conditions exist on Unit 1:
- Reactor power 100%
l - All systems normal l - FT-512 selected for steam flow input into SGWLC for S/G 1 A l Wh:t is the effect of the pressure transmitter associated with FT-512 failing low? l .1 A S/G level will decrease, .. c. feed pump speed will decrease and S/G level will decrease below the LO-2 setpoin b. feed pump speed is unaffected, and S/G level will return to norma c. feed pump speed will increase and S/G level will return to norma d. feed pump speed is unaffected, and S/G level will decrease below LO-2 setpoin Answer a Exam Level R Cognitive Level Comprehension Facility: Braldwood ExamDate: 9/ 4/98 KA: 059 K1.04 RO Value: 3.4 SRO Value: 3.4 Section: SYS RO Group: 1 3RO Group: 1 Sy tem /Evoludon Main Feedwater System KA Knowledge of the physical connections and/or cause-effect relationships between Main Feedwater System and the foHowing: S/GS water levelcontrol system Explanation of Steam flow is output to summator for FW control system program Delta-P. Delta-P program will decrease Answer causing feed pump speed and FW header pressure to decreas Reference Title / Facility Reference Number Section/Page Revision L FW EH controls / schematic EHC-6/DP 1 F"M.C schematic FW-2/ 512 loop 0 t 27 SGWLC lesson plan 6, 16 Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment
.
Frid;y, September 4,1998 Page 53 of 100
_
* - Question AFW Startup The following ccnditions exi t on Unit 1: . 1 - The reactor tripped from an at-power condition l - An undervoltage condition exists on RCP 1C bus - Power Range NIS channel N42 failed at 100% on the trip !
- ESF bus 141 undervoltage occurred - 1 A D/G automatically started and ACB 1413 is closed - S/G levels lowest readings were - 19% (A); 25% (B); 22% (C); 20% (D)
Wh:t is the status of the Auxiliary Feedwater (AF) Pumps on Unit 1 for these conditions at ONE minute l following the trip? c. Both AF pumps are running, b. The 1 A AF pump is running and the 1B AF pump is NOT runnin ] c. The 18 AF pump is running and the 1 A AF pump is NOT runnin d. NO AF start signal is initiate Answer b Exam Level B Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 06i A3.01 RO Value: SRO Value: 4.2 section: SYS RO Group: 1 SRO Group: 1 System / Evolution Auxiliary / Emergency Feedwater System KA Ability to monitor automatic operations of the Auxilk v / Emergency Feedwater System including: AFW startup and flows Explanation of SG levels are above AF actuation setpoints and the motor driven AF pump starts on the detected undervoltag Answer i once Title / Facility Reference Number Section/Page Revision L. o.
Aux Feedwater System 2 5 Chp 26 AFW sys lesson plan 9 3,5 Chp 9 EDG lesson plan 7 7 Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment l l l _ FridIy, September 4,1998 Page 54 of 100
- - - -__ .- . .- . -.
Questioa AFW flow requirrments for cooldown l in cccordince with ths BEPs, which of ths following describ:s the MINIMUM AFW pump flow and S/G l l configuration necessary to remove all of the reactor decay heat load following a reactor trip from 102% l power to preclude entry into loss of heat sink RED path entry? l 4. The 1 A AF pump supplying 480 gpm to at least ONE S/G with S/G blowdown manually isolate b. The 1B AF pump supplying 245 gpm to each of TWO S/G with S/G blowdown in servic c. The 1 A AF pump supplying 170 gpm flow to each of THREE S/Gs with S/G blowdown manually isolated, d. The 1B AF pump supplying 130 gpm flow to each of FOUR S/Gs with S/G blowdown in servic Answer C Exam Level B Cognitive Level Memory Facility: Braidwood ExamDate: 9/14/98 KA: 061 K5.02 RO Value: 3.2 SRO Value: 3.6 section: SYS RO Group: 1 SRO Group: 1 i System / Evolution Auxiliary / Emergency Feedwater System KA Knowledge of the operational implications of the following concepts as they apply to the Auxiliary / Emergency Feedwater System: Decay heat sources and magnitude l Explanation of Answer Reference Title / Facility Reference Number Section/Page Revision L AFW system lessson plan ch26 9 1,11 Material Required for Examination Question Source: New Question Modification Method: Signifcantly Modifwed l
'
Question Source Comments: Comanche Peak 11/93 NRC Exam Comment Type Comment l
.
i Frid:y, September 4,1998 Page 55 of 100
.. -.- - -- ---- _ -- - - _ - . .-
Question h DC bus bitt ry charg:r Th3 following conditions exist on Unit 1: I
- Reactor power - 100%
, .inv:stigation has located a ground on the 125 VDC Normal supply to the 1 A.D/G. What j tction is required to transfer DC Control Power to the reserve source? ! Tha Reserve power breaker from... o. DC 111 will be closed after opening the Normal power breaker and the Reserve power breaker at the D/G control pane ! b. DC 111 will be closed after swapping the no-blow link at the Normal and Reserve power fuse blocks et the D/G control panel, e. DC 112 will be closed after opening the Normal power breaker and the Reserve power breaker at the D/G control pane d. DC 112 will be closed after swapping the no-blow link at the Normal and Reserve power fuse blocks I ct the D/G control pane Answer b Exam Level B Cognitive Level Memory Facility: Braidwood ExamDate: 9/14/98 KA: 2.1.30 RO Value: 3.9 SRO Value: 3.4 Section: SYS RO Group: 2 SRO Group: 1 System /Evoludon D.C. Electrical Distribution , KA Ability to locate and operate components, including local control Explanation of Answer k. .rence Title / Facility Reference Number Section/Page Revision L VDC system / schematic DC-1 0 DC Control power transfer from Normal to reserve source / BwOP-DC-6A1 51 Chp Ba 125 VDC lesson plan 6 4,6 Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: [ Comment Type Comment
.
i i Friday, September 4,1998 Page 56 of 100
__ _ ._ ___ . _ - . _ ___ _ .- _ _.__ _ _ _ _ _ _ __ __ _._____._ l Question N _ Sequencing of ESF pumps - SI & SI w LOP l Unit 1 wts being synchronized to the grid wh n tha following occurred:
- Trip of 345 KY breakers resulted in deenergizing the SATs
'
- A steamline break occurred that resulted in containment pressure reaching 20 psig 20 seconds after the D/Gs output breakers have closed
_ Wh:n would the 1A SX pump re-start? c. Following start of the 1 A CS Pum b. Between the start of the 1 A CV pump and the 1 A RH pum e. Between the start of 1 A CC pump and the 1 A AF pum d. Coincident with the starting of the 1 A and 1C RCFC Answer c Exam Level B Cognitin Level Memory racility: Braidwood ExamDate: 9/14/98 KA: 064 A3.07 RO Value: 3.6 sRO Value: 3.7 section: SYS RO Group: 2 sROGroup: 2 systemlEvolution Emergency Diesel Generators KA Ability to monitor automatic operations of the Emergency Diesel Generators including: Load sequencing Explanation of The SX pump would be started in this case by the Si signal which is overrides the UV condition. The SX pump Answer starts in following sequence: CV (0 sec); SI ((5 sec); RH (10sec); CS (15-18 secs, if actuation signal present); CC pumps (20 sec); SX pumps (25 sec); AF 1 A pump (35 sec); CS pump (40 sec, if acutaion signal now present but not present at 18 sec) Reference Title / Facility Reference Number section/Page Revision L ~ D'4 RIlaying/ schematic DG-2/ sequencing order 1 L 9 EDGs and Aux sys lesson plan 7 7 Chp 20 Essential Service Water sys L sson plan 7 8 Material Required for Examination Question source: New Question Modification Method: Question source comments: Comment Type ~ ~ Comment
-
Frid:y, september 4.1998 Page 57 of 100
.
- . . . ~ - - - - w n-
. - . . - -- - _ . - - - , -- - . - _ - - . - _ - . . - _ - -
th+ I RCDT oper; tion - cffect of CNMT lsolition The following condition 2 cxist on Unit 1: - f~
- Unit is in MODE 3 - A cooldown had just been initiated - Steam Dump Bypass Interlock control switches have just been taken to BYPASS - No other operator actions have been performed , - The Steam Dump valves fail open and the following parameters are observed: l - RCS temperature - 537'F (A); 539'F (B); 538'F (C); 538'F (D) ! - Pzr pressure - 1820 psig j - Pzr level - 10% ' - S/G pressure - 850 psig (A); 740 psig (B); 800 psig (C); 750 psig (D) l
,
- S/G flow - 1.0 Mlb/hr (A); 1.5 Mlb/hr (B); 1.1 Mlb/hr (C); 1.6 Mlb/hr (D) l
- The level in the RCDT has risen to the alarm setpoint (80%) for REACTOR COOLANT DRAIN TANK UNIT 1 LEVEL Hi-LO Assuming all systems are functioning correctly, what is the status of the RCDT system?
]
l e. BOTH RCDT pumps are running and flow is directed to the Holdup Tank b. BOTH RCDT pumps are running and flow is recirculated back to the RCD n ONE RCDT pump is running and flow is directed to the Holdup Tank NtiTHER RCDT pump is running and NO flow exists for the syste Answm d Exam Level B Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 068 A4.04 RO Value: 3.8 SRO Value: 3.7 section: SYS RO Group: 1 SROGroup: 1 hm/ Evolution Liquid Radwaste System esA Abihty to manually operate and/or monitor in the control room- . Automatic isolation Explanation of Conditions for steem flow & low RCS temp. actuate Si, The coincident CNMT Phase A lsolation signal Answer isolates RCDT valves out. Closure of valve RE9170 cuses pumps to sto Reference Title / Facility Reference Number Section/Page Revision L PRTcnd RCDT/ schematic RY-4 2
- Ch'p 48a Liquid Rad Waste lesson plan 6 11 Ch61 ESF lesson plan 5 7 l Material Required for Examination Question Source New Question Modification Method Question Source Comments:
Comment Type Comment I l
~
l
'
! l I ! l.
! oc Friday, September 4.1998 Page 58 of 100 i
.. - . . . .--
- . - .- ._ - . . - . . - - . . .- . . - . . - .. .-
l Question h CNMT Sump sourc;s ofinput during normal operrtions During et-power operations with syst:ms in th:ir normal clignm:nt, what is a normal-source of wit:r to the C ntainment Floor Sump? i c. SI Accumulator valve leakoffs.
I b. Leakoff from the #3 RCP seals.
f e. Leakoff from the reactor vessel flange, d. Valve packing leakage from the CVCS letdown isolation valve Answer b Exam Level R Cognitive Level Memory Facility: Braidwood ExamDate: 9/14/98 KA: 068 K1.07 RO Value: 2.7 SRO Value: 2.9 Secuon: SYS RO Group: 1 SRO Group: 1 System / Evolution Liquid Radwaste System l KA Knowledge of the physical connections and/or cause-effect relationshios between Uguki Radwaste System and the following- ) Sources ofliquid wastes for LRS ' E v'r > ,of Rx Cavity sump output to CNMT Floor sump, #2 seals directed to RCDT, RV flange to RCDT , valve leakoffs Answer directed to PRT Reference Title / Facility Reference Number Section/Page Revision L Chp 46a Liquid Radwaste System 6 12 i Material Required for Examination l Question Source: New Question Modification Method: Question Source Comments: I Comment Type Comment
.
l~ l Friday, September 4.1998 Page 59 of 100 ._ . . _ . _ _ _ . _ _ ___ . _ .....-. ., _ _ _ _ _ . _ _ . _ _ _ _ - _ _ . . Question 60 Wasts G:s Decay Tank operations L Wh:n cligned for normel operation (BOP GW-1), what is the responso to high pressure sensed at the in-service Gas Decay Tank? I clarm is generated that... e. alerts the operator to manually place a standby Gas Decay Tank in servic b. Indicates auto swap of in-service Gas Decay Tank to selected standby Gas Decay Tank, and alerts the operator to align another standby Gas Decay Tank, e. Indicates auto swap of in-service Gas Decay Tank to selected standby Gas Decay Tank and auto swap of standby Gas Decay Tank to new standby Gas Decay Tan d. shuts down the Waste Gas Compressors and isolates the in-service Gas Decay Tan Answer b Exam t.evel R Cognitive Level Memory Facility: Braidwood ExamDate: 9/14/98 KA: 071 A4.05 RO value: 2.6 SRO value: 2.6 Section: SYS Ro Group: 1 SROGroup: 1 SystemtEvolution Waste Gas Disposal System j KA Ability to manually operate and/or monitor in the control room: 4 oss decay tanks, including vanes, indicators, and sample line ' Explanation of Indicates auto swap to standby WGD Tank at 95 psi Answer Reference Title / Facility Reference Number Section/Page Revision ' L ; Gas waste Sys S/U & Operation / l BwOP GW-6 5 GDT set sw reposition req'd/ OGWO2J-A1 51
.Chgp 46 Gas Radwaste sys lesson plan 6 6 i AAstorial Required for Examination petion Source: New Question Modification Method:
Question Source Comments: Comment Type Comment j I
-
Frkiry, September 4,1998 Page 60 of 100
. . - , , - - - _ - - _ -- - . .--
~_ .. . -. -- - . . . . , . -- - . . - . - . - ..
l l ouestion k Check Source oper1 tion l Ar s Radiation Monitor for Fu:1 Bldg Fu:1 Handling incident (ORE-AR055)is being manually Check Source tested. What is the system response when the monitor's CHECK SOURCE (C/S) pushbutton is d: pressed at the RM-23 panel? l 2. The alarm and automatic action output will be blocked, and the RM-23 amber INTLK LED will be lit.
l l ! b. The alarm and automatic action output will be blocked, and the RM-23 green AVAIL LED will be li c. The alarm will actuate when the alert setpoint value is reached, and the RM-23 amber INTLK LED will be li d. The alarm will actuate when the high setpoint value is reached, and the RM-23 red HIGH LED will be li Answer b Exam Level R Co9nitive Level ' Memory Facility: Braidwood ExamDate: 9/14/98 MA: 072 A4.03 RO Value: SRO Value: 3.1 Section: SYS RO Group: 1 SROGroup: 1 System / Evolution Area Radiation Monitoring System KA Ability to manually operate and/or monitor in the control room: Check source for operability demonstration Explanation of Depressing the C/S blocks the alarm and auto function of the minitor but the AVAlllitght remains li Answer Reference Title / Facility Reference Number Section/Pa9e Revision L. Control Function Channel l Check Source Energized /BwOP AR/PR 11A26 i Rad Monitor Sys lesson plan chp 49 7 3, 8 l
% rial Required for Examination '
Action Source, New Question Modification Method: Question Source Comments: j Comment Type - Comment i l
.
l i Fridiy. September 4,1998 Page 61 of 100
. . . . - . . -. . .
Question ' N ' Loss of FHB OvIrh:ad Crrna rad ' monitor Th3 following conditions exist on Unit 2: . L - Refueling operations are in progress
'
While using the Fuel Handling Building Crane to move new fuel into the Spent Fuel Pool, the radiation monitor ORE-AR039, Fuel Handling Building Crane Monitor, goes into alarm. What action is affected? c. Traverse of the Fuel Handing Building Crane bridge and trolle n. Both lowering and raising the Fuel Handing Building Crane hois e. Traverse of the Fuel Handing Building Crane trolley and raising the hois d. Raising the Fuel Handing Building Crane hois Answer d Exam t.evel B Cognitive Level Comprehension Facility: Breldwood ExamDate: 9/14/98 MA: 072 K3.02 RO Value: 3.1 sRO Value: 3.5 section: SYS Ro Group: 1 SRO Group: 1 system / Evolution Area Radiation Monitoring System KA Knc vledge of the effect that a loss or malfunction of the Area Radiation Monitoring System will have on the following: Fuel handling operations Esplanation of Rad monitor prevents raising hois l Answer Reference Title / Facility Reference Number Section/Page Revision L. Chp 49, Radiation Monitors lesson plan 7 4.a.3) i
"%I Required for Examination .estion Source: New Question Modification Method:
Question Source Comments: Comment Type Comment i e.
l I ,
Frid?y. September 4.1998 Page 62 of 100 i .
- - . . -. .-. . . . -- . . .. . - - - _ ~
, l Question bD Eviluition of eqpt affected for slow loss The following conditions exist on Unit 1: . l - A unit startup is in progress with reactor power raised above 18%. l - Turbine is at 1800 rpm ready to be synchronized to grid.
l - Motor driven feedwater pump is supplying the S/Gs with Feed Reg Bypass valves l in AUT Steam Dump demand in AUTO at 12%.
- Instrument air header pressure begins to slowly drop due to a leak if the leak CANNOT be isolated and instrument air pressure continues to drop, which of the following would occur? (Assume NO operator action taken.)
a. AF recirculation flow to the CST would be lost due to AF recirc,1 AF022A, failing close b.- Pressurizer level would increase due to charging header flow control valve,1CV121, failing ope c. Pressurizer pressure would decrease due to Aux spray isolation,1CV8145, failing ope .d. Feedwater heater 17A extraction steam would isolate due to emergency drain,1HD038A, failing close Answer b Exam Level B Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 078 K3.02 RO Value: 3.4 SRO Value: 3.6 Section: SYS RO Group: 3 SRO Group: 3 Sy: tem / Evolution Instrument Air System KA Knowledge of the effect that a loss or malfunction of the Instrument Air System will have on the following: Systems having pneumatic valves and controls r epianation of Charging flow goes to maximum due to 1CV121 failing cpen, and letdown isol 1CV459 & 1CV460 fail close swer 'a' is incorrect because both 1 A & 1B AF pump recirc valves fail open . 'c' main turbine not directly affected. 'd' not occur because steam dumps fail close Reference Title / Facility Reference Number Section/Page Revision L Loss ofinstrument Air Lesson Plan 18wOA SEC-4 Table A 52 Chp 53 lA/SA lesson plan 8 9 Material Required for Examination Question Sovece: New Question Modification Method:
,Question Source Comments:
Comment Type Comment
-
Friday, September 4,1998 Page 63 of 100 _--
. -. . - . - . - - - - . - - -- _- - --. .. .- .
Question 0 [ Effe::t of loss of DC - CO2 cctuation L With ths firs prot:ction syst:ms in th ir norm:I alignmnnt, what is tho affect of a loss of DC power? ! Loss of DC control power to the... 4. halon control cabinet will cause halon release in the Upper Cable Spreading Room.
l b. battery control panel will cause automatic start of the diesel driven fire pump.
i c. fire detection system will cause start of the motor driven fire pum d. CarDon dioxide system will cause the master EMPC valve to open pressurizing the CO2 heade Anomr d Exam Level B Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 066 K4.06 RO Value: 3.0 SRO Value: 3.3 Section: SYS RO Group: 2 SRO Group: 2 System / Evolution Fire Protection System MA Knowledge of Fire Protodion System design feature (s) and or interlock (s) which provide for the following: Co2 Explanation of EMPCs uses DC control power. On loss of power, the master EMPC valves fail open which in tum cause the Answer master discharge / selector valve to open, charging the affected heade Refenmco Title / Facility Reference Number Section/Page Revision L Chp 57, Fire Protection System lesson plan 5 8 Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment
-
! Friday, September 4,1998 Page 64 of 100
- - -- . . , - - . . . -- . Question bI Evilutta conditions - unwirrtnted rod withdrawal l The f;llowing Conditions exist on Unit 1: -
! ! - Reactor power is 30%. l - Rod controlis in Automatic ! - Tref- 564*F
- Tave values - 564*F (A); 565'F (B); 565'F (C); 564*F (D) - Power Range NI - 31% (N41); 29% (N42),30% (N43); 30% (N44) - Control bank D is at 156 step Which condition would result in continuous rod withdrawal?
c. Turbine first stage pressure PT-505 fails to 100%. b. Power Range channel N41 fails to 20%. e. Loop A Tcold fails 553* d. Tref signal fails 557' Answer a Exam Level B Cognitive Level Comprehension Facility: Braldwood ExamDate: 9/14/98 KA: 001 AA2.05 RO Value: 4,4 SRO Value: 4.6 Section: EPE RO Group: 2 SRO Group: 1 System / Evolution Continuous Rod Withdrawal MA Ability to determine and interpret the following as they apply to Continuous Rod Withdrawal: Uncontrolled rod withdrawal, from available indications Explanation of input to rod control Tref, auctioneered HIGH Tave & Auctioneered high PRNis: PT-505 provides input signal for Answer development of Tref. If it fails high Tref goes to maximum value (581*F) and results in rods being withdrawn to match Tave to Tref. PR failure high compares the rate of change of reactor power to the rate of change of turbine power. Initially high rate of change during failure but rapidly the rate of change falls to zero and so rods may initailly begin to insert but quickly stop motion with no more rate of change. Auctioneered high Tave is used and Tcold failing low will remove this input (if prevolusly auctioneered high). Tref failing low will cause ! rods to move inward to match Tave to Tre l Reference Title / Facility Reference Number Section/Page Revision L. , l Rod control Unit / Schematic RD-2 2 ; Chp 28 Rod control sys Lesson Plan 12 20 I Uncontrolled Rod Motion /18wOA ROD-1 L sson plan 6 3 Material Required for Examination Question Source: *New Question Modification Method: Question Source Comments: Comment Type Comment
.
l
*
, Friday. September 4.1998 Page 65 of 100 l l l
i Question O P/A vs. Group Step Count:rs A Control B nk D rod w s dropped from 156 ct:ps. The P-A conv:rtar wcs I:ft at 156 sinps whsn it was to be reset to ZERO steps as directed by procedure BwOA ROD-3 " Dropped Rod Recovery". ,
ilect the affect of performing the procedure in this manner? l e. While performing the procedure, the C-11 Rod Stop will be received prior to realigning the ro j b. While performing the procedure, the Rod insertion Limit Alarm will be received at a lower rod position ! than required, e. After the procedure is complete, Bank C control rods will begin insertion at a lower value of Control Bank D, d. After the procedure is complete, Bank C control rods will begin insertion at a higher value of Control Bank I l Answer a Exam byel B Cognitive Level Application Facility: Braidwood ExamDate: 9/14/98 KA: 003 AK3.10 RO Value: 3.2 SRO Value: 4.2 Section: EPE RO Group: 2 SRO Group: 1 j System /Evoludon Dropped Control Rod l l KA Knowledge of the reasons for the following responses as they apply to oropped Control Rod-RIL and PDIL l Explanation of The bank overlap units are bypassed when rods are moved with individual bank selector positions. The P to A l Answer converter provides step information to rod position indication including the C-11 circuit. As the individual rod I was withdrawn to approximately 67 steps the C11 circuit would sense that bank D was at 223 steps and block outward motio l Reference Title / Facility Reference Number Section/Page Revision L. O.
RD D:ta logging / rod stops schematic RD-5/RD-1 l P/A & C-11 rod stop 0/0 (. 28 Rod Control sys lesson plan 12 ig,10 Material Required for Examination Question Source: New Question Modification Method: Editorially Modified Question Source Comments: o.C. Cook 6/13/1995 Comment Type Comment ,
.
Friday, September 4,1998 Page 66 of 100
r ouestion. [ 7 Stabilized RCS temper;tura with f;ilura of Sterm Dumps On Unit 1, e loss of cli circultting water pumps his resulted in a reactor trip. All control systcms respond ; cs cxpected. Significant decay heat causes RCS temperature to increase following the tri l
what RCS temperature should temperature stabilize? ;
Temperature should stabilize at... e.550* I
b.557' c. 561* i d. 565* . Answer c Exam Level B Cognitive Level Application Facility: Braidwood ExamDate: 9/14/98 KA: 007 EA1.03 Ro Value: 4.2 sRo Value: 4.1 Section: EPE Ro Group: 2 SRO Group: 2 l systemtEvolution Reactor Trip ;
!
KA Ability to operate and / cr monitor the following as they apply to Reactor Trip: RCS pressure and temperature Explanation of The condenser would NOT be available for steam dumps (either on trip controller or load rejection controlier).
Answer Th S/G pressure would stabilize based on the seocndary PORV opening setpoint normally set at 1115 psi The Main Steam safety valve setting is 1175 psi Reference Title / Facility Reference Number Section/Page Revision L i l Stzm dumps / schematic MS-4/ C-9 4 Chp 24 Steam dumps lesson plan 7 4 Chp 23 Main steam lesson plan 8 3 Material Required for Examination ution Source: New Question Modification Method: Question Source Comments: Comment Type Comment
.
.
Friday. September 4.1998 Page 67 of 100
-_ - . _ _ . . . _- -- . . _ _ - . .--~ - -
Topic
.
( Question h Reactor Trip requirements L
" Unit 2 is operating at full load, which group of conditions will result in an automatic reactor trip either j .,r::ctly or indirectly?
I c. RCP bus frequency (Hz):56.9 (Bus 156) 57.1(Bus 157) 56.9 (Bus 158) 57.2 (Bus 159) n. Power range (%): 107 (N41) 108 (N42) 108 (N43) 109 (N44) e. PZR pressure (psig): 2375 (PT-455) 2380 (PT-456) 2385 (PT-457) 2380 (PT-458) d. S/G C NR level (%): 35 (LT-537) 38 (LT-538) 38 (LT-539) 37 (LT-558) Answer a Exam Level R cognitive Level Memory Facility: Braidwood ExamDate: 9/14/98 KA: 007 EK2.03 RO Value: 3.5 SRO Value: 3.6 Section: EPE RO Group: 2 SRO Group: 2 System / Evolution Reactor Trip KA Knowledge of the interrelations between Reactor Trip and the following: Reactor trip status panel Explanation of Trp condition RCP UF - 2/4 RCP buses < 57.0 Hz. Other trip setpoints: Rx power - 2/4 >109%; Pzr pressure
. Answer Title: 2/4 > 2385 psig iteference Title / Facility Reference Number Section/Page Revision L ESF Setpoints/ schematic EF-1/Rx trip 4 2BwEP-0 Reactor Trip or Si lesson plan 3 6 Chp 60b RPS lesson plan 6 4 Material Required for Examination Question Source: New Question Modification Method: Significantly Modifed Question Source Comments: Comancho Peak 11/94 nment Type Comment
.
l ! Friday, September 4,1998 Page 68 of 100
- .
au tion b f Tril-Pipe conditions With ths RCS ct norm 2l op;rsting pressuro and temperatura, what is the condition of the steam entering
;
ths PRT at normal conditions, if a PORV opens? (Assume an ideal thermodynamic process).
c. Superheated steam at 651* b. Superheated steam at 250* e. Saturated steam-water mixture at 222* d. Saturated steam water mixture at 163* ,
Answer C Exam Level R Cognitive Level Application Facility: Braldwood ExamDate: 9/14/98 ) KA: 008 AK1.01 RO Value: 3.2 SRO Value: 3.7 Section: EPE RO Group: 2 SRO Group: 2 System / Evolution Pressurizer Vapor Space Accident i KA Knowledge of the operationalimplications of the following concepts as they apply to Pressurizer Vapor Space Accident I Thermodynamics and flow characteristics of open or leaking vatves j Explanation of Nominal PRT pressure 3 psig; Hg = 1154 BTU /lb. Saturation temperature 221.9'F. At NOP Pzr pressure 2235 Answer psig with Hg = 1117.7 BTU /lb. Therefore PRT conditions are within saturation parameters.
" l Reference Title / Facility Reference Number Section/Page Revision L I Sterm Tables Chp 14, Pressurizer lesson plan 9 25e l Material Required for Examination Steam Tables Question Source: New Question Modification Method: Signifcantly Modified Question Source Comments: South Texas 9/95 l Comment Type Comment l l
l
.
Friday. September 4,1998 Page 69 of 100
. _ _ . . _ . . ._ . .. ._. - - . _ .
r Question % C:lculation of subcooled margin on leonics ' The following conditions exist on Unit 1: -
- Subcooling Margin output from the SPDS Iconics has failed - 1C RCP and 1D RCP are running Tha Unit Supervisor has asked you to determine the subcooling margin using the same valid inputs as L used by SPDS.
! Wh:t are the parameters used to calculate subcooling margin? , c. RCS wide range pressure from loop C hot leg and core exit thermocouple temperatures.
' b. Pressurizer pressure and core exit thermocouple temperature e. RCS wide range pressure from loop A and loop C hot leg, and RCS loop A and loop C hot leg temperature d. Pressurizer pressure and RCS loop C and loop D hot leg temperature Answer a Exam Level B Cognitive Level Comprehension Facility: Braldwood ExamDate: 9/14/98 KA: 009 EA1.10 RO Value: 3.8 SRO Value: 3.9 section: EPE RO Group: 2 SRO Group: 2 ; sye+ m/ Evolution Small Break LOCA
*
e Ability to operate and / or monitor the following as they apply to Small Break LoCA: I Safety parameter display system i Explanation of I Answer ) Reference Title / Facility Reference Number Section/Page Revision L. F"S Display schematic CX-1/subcooling 1
'
L .4b inadequate Core Cooling Lcsson plan 7 6 Material Required for Examination Question Source: New Question Modification Method: i Question Source Comments: Comment Type Comment l ! ! ! i l Frid y, September 4,1998 Page 70 of 100
- - - -- . .- . - - . - . . - - -
ouestion 7/ ' RCP trip criteria (viluition The following conditions exist during perform:nce of BEP- . ! - Train A ECCS pumps failed to star RCS pressure is 1350 psi . Containment pressure of 7 psi Bus 142 has an overcurrent trip on the normal feeder breake .
- Si actuated due to High Containment Pressur ' -
The highest critical safety function is Yellow on Heat Sin All other equipment and components operated as expecte Bised on above plant conditions, the RCPs should... a. remain running because NO Si pumps or Charging Pumps are runnin h. be stopped because RCS pressure is below the RCP trip criteri e. remain running until Pressurizer level decreases below 34%. ) i d. be stopped because CC flowpath to the RCP motor oil coolers is isolated, j Answer 8 Exam Level B Cognitive Level Application Facility: Braidwood ExamDate: 9/14/98 KA: 011 EA1.03 RO Value: 4.0 SRO Value: 4.0 Section: EPE RO Group: 2 SROGroup: 1 System /Evoludon large Break LOCA KA ' Ability to operate and / or monitor the following as they apply to Large Break LOCA: Securing of RCPs Explanation of The trip criteria is < 1425 psig, with NO cooldown in progress, and HHSI flow > 50 gpm or Si flow > 100 gp Answer F snce Title / Facility Reference Number Section/Page Revision L ., L. .a for 18wEP-0 Trip RCPs 1C 1BwEP-0 lesson plan RCP trip criteria 11 2,5 Material Required for Examination Question Source: New Quostion Modification Method: Signifcantly Modered Qt'estion Source Comments: Watts Bar 3/3/1995
- Comment Type Comment -
,
_ l ! i !- Friday, september 4,1998 Page 71 of 100 .. - = . . . . - .. - - . - -- _ . - - - . _ - . ~- Question 7M Ev:1 loss of cooling flow ! On a loss of s;:1 injection to the RCPs, whtt crit:ria is us:d to det2rmino if ths RCPs should be tripped per BwOA RCP-2 " Loss Of Seal Cooling"? c. High temperatures on the RCP lower bearing outlet temperature n. Time elapsed since loss of sealinjectio c. RCP Thermal Barrier Component Cooling Water low flow alarm d. High vibration condition on the RCP, Answer a Exam level B Cognitive Level Memory Facility: Braidwood ExamDate: 9/14/98 l MA: 015 AA2.10 RO Value: 3.7 SRO Value: 3.7 Section: EPE RO Group: 1 SRO Group: 1 l System / Evolution Reactor Coolant Pump Malfunctions j MA Ability to determine and interpret the following as they apply to Reactor Coolant Pump Malfunctions: When to secure s.0Ps on loss of cooling or sealinjection Explanation of Seal & bearing temperatures are monitored for trip setpoin l An wor ' Roderence Title / Facility Reference Number Section/Page Revision L Less cf seal cooling 1BwOA RCP-2 54 , Losss of Seal Cooling lesson plan 6 4 l Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment l I l l l I Frid:y, September 4,1998 Page 72 of 100 ! l __
___ _ .. _ _ __ .__ __ _ -. a-a- 73 eviior ace seri r:iiura Unit 1 is oper ting ct 100% pow r wh:n the following alarms aro receiv:d/ reported: .
- RCP SEAL LEAKOFF FLOW LCW (1-7-C3)
Tho NSO investigates and reports the following additionalinformation:
- RCP 1 A seal injection flow is 10.7 gpm - #1 Seal Leakoff Flow on 1 A RCP is 0.4 gpm - RCP 1 A Seal Water Outlet Temperature is 140*F and STABLE - RCP 1 A Bearing Outlet Temperature is 145'F and STABLE - Unit 1 RCDT levelindicates 75%
Besed on the above information, which of the following events has occurred? c. RCP 1A #1 Seal has failed closed b. RCP 1 A #1 Seal has failed ope e. RCP 1 A #2 Seal has failed close d. RCP 1 A #2 Seal has failed ope Answer d Exam Level B Cognitive Level Comprehension Facility: Braldwood ExamDate: 9/14/98 KA: 015 AK2.07 RO Value: 2.9 SRO Value: 2.9 Section: EPE RO Group: 1 SRO Group: 1 Sy: tem / Evolution Reactor Coolant Pump Malfunctions KA Knowledge of the interrelations between Reactor Coolant Pump Malfunctions and the following: RCP seals
~ olanation of swer Reference Title / Facility Reference Number Section/Page Revision L O.
RCP seal Failure /1BwOA RCP-1 SSB 1BwOA RCP-1 lesson plan 7 5 Material Required for Examination Question Source: Facility Exam Bank Question Modification Method: Editorially Modified Question Source Comments: Braidwood bank Comment Type Comment
.
FndIy. September 4,1996 Page 73 of 100 l l
. _ _ __ __ _ _ _ _ _ _ -- . Question) VCT level tr:nsmitt::r malfur.ction .Giv:n the following: -
The plant is at 90% power with ALL controls in AUT VCT level transmitter, LT-112, fails HIGH causing a letdown diversio At the time of failure VCT level transmitter, LT-185, reads 50%. What will occur if NO operator action is taken? VCT level decreases... I a. until Auto makeup starts and maintains VCT leve b. with NO auto makeup capability and charging suction shifts to RWS c. faster than auto makeup input and charging suction shifts to RWS d. until charging pumps lose suction and start to cavitat Answer d Exam Level B Cognitive Level Application Facility: Braidwood ExamDate: 9/14/98 KA: 022 AA1.08 RO Value: 3.4 SRO Value: 3.3 Section: EPE RO Group: 2 SRO Group: 2 System / Evolution 1.oss of Reactor Coolant Makeup KA Ability to operate and / or monitor the following as they apply to Loss of Reactor Coolant Makeup: VcT level Explanation of LT 112 provides for AUTO makeup to the VCT. If NO operator action taken, then level will continue to fall until Answer NPSH is lost to the CENT CHG pump (s). Transfer will NOT occur to RWST since both channels are required for swap. An alarm will be generated from LT-185 at 20% leve Reference Title / Facility Reference Number Section/Page Revision L C"OS notes / schematic CV-2/ LT 112 table 3 L 15a CVCS lesson plan 10 11,14 Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment _ - Friday. September 4,1998 Page 74 of 100
->
Question 7 Time / amount E-boration for condition Giv:n the following cft:r e r::ctor trip: . !
-
THREE rods remain withdraw Due to equipment malfunctions boration is only available from the RWST.
1 - Charging flow rate 132 gpm.
L - RCS boron concentration was 1050 prior to the tri gpm letdown in servic Of the listed times, which would be minimum acceptable time that boration from the RWST would have to occur? c.1 Hour b. 2 Hours e. 3 Hours d. 4 Hours Answer b Exam Level B Cognitive Level Application Facility: Braldwood ExamDate: 9/14/98 KA: 024 AA2.05 RO Value: 3.3 SRO Value: 3.9 Section: EPE RO Group: 1 SRO Group: 1 System / Evolution Emergency Boration KA Ability to determine and interpret the following as they apply to Emergency Boration: Amount of boron to add to achieve required SDM Explanation of 18wEP ES-0.1 requires 3600 gallons boration from RWST for each rod not fully inserted, therefore requiring Answer 10,800 gallons. If net change over is 120 gpm, then required time is 10,800/120 = 90 minutes. Other answers based on counting 2 rods and/or borating from CV-8104 @ 57 gpm with total of 1200 gallon P * ence Title / Facility Reference Number Section/Page Revision L JA Prl-2 emergency Boration 55B 18wOA Pri-2 lesson plan 1 4,6 18wEP-0 lesson plan 11 3 Material Required for Examination 18wEP ES-0.1, page 6 (step 5) Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment
.
l l Friday, september 4,1998 Page 75 of 100 l !
I l Question Calc of time to satur; tion /cora boiling l The following conditions cxist on Unit 1: - l I
- - A forced outage is in progress l - The plant was shutdown 8% days ago to repair a steam generator tube lea Draining of the RCS was initiated to allow access to S/G l
- Reactor vessel level is at 397' 1" with Thot at 212* A loss of RHR pumps due to cavitation has occurred i
!
l Which of the following is the smallest amount of flow that meets the minimum makeup flow required to l m:intain current RCS level? c. 80 gp I l b.' 72 gpm.- c. 65 gp d. 59 gp Answer b Exam Level B Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 025 AK1.01 RO Value: 3.9 SRO Value: 4.3 section: EPE RO Group: 2 SROGroup: 2 system / Evolution Loss of Residual Heat Removal System KA Knowledge of the operationalimplications of the following concepts as they apply to Loss of Residual Heat Removal System- . Loss of RHRS during all modes of operaten I Explanation of 81/2 days is 204 afters shutdown. The curve shows minimum flow at approximately 70 gp Answer Reference Title / Facility Reference Number Section/Page Revision L Lo s cf RH cooling /18wOA Prl-10 56
-
3A Pri-10 Lesson plan 4
' Material Required for Examination Figure 18wOA PRI 10-1 . Question Source: New Question Modification Method:
Question Source Comments: Comment Type Comment
-
! l l Friday, Septerrber 4,1998 Page 76 of 100 _ . . _ _ _ _ _ _ _ . _ . _ _ . _ _ _.__ _ ...___ ___. _ . _ . Question N Alternat3 RCS cooling The following conditions exist on Unit 2: .
-
MODE 5 operation during normal cooldown
-
RCS temperature - 195* F
--
RCS pressure - 325 psi <- Train A RH in service, train B RHR tagged out for repairs Whit is the preferred method of core cooling if a loss of RH cooling occurs? Alt:rnate RCS cooling using... c. the Si accumulators.- b. the S/G c. normal charging and RHR letdow d. S1 Pump hot leg injectio Answer b Exam Level B Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 025 AK3.01 RO Value: 3.1 sRO Value: 3.4 section: EPE RO oroup: 2 SRO Group: 2 system / Evolution Loss of Residual Heat Removal System KA Knowledge of the reasons for the following responses as they apply to Loss of Residual Heat Removal System: ! Shift to attemate flowpath ! Explanation of Steaming Intact /non-isolated SGs is the preferred alternate decay heat removal method if the RCS is intac Answer
I ' Reference Title / Facility Reference Number Section/Page Revision L 0, ' 4 8 s of RHR Cooling /18wOA Pri-10 Table A 56 L.rOAPri-10 Lesson Plan 4 Material Required for Examination Question Sourc New Question Modification Method: l Question Source Commente: Comment Type Comment
. - Friday, September 4,1998 Page 77 of 100 ' . - .. . . _ . . . _
__
_ . __ . _ _ _ Queeuen '7[ Ev:luation of CCWleak The fallowing conditions cxi;t on Unit 1: -
- The reactor is shutdow RHR is in shutdown coolin RCS temperature is 300' RCS pressure is 160 psi CCW surge tank levelis decreasing Whtt leak locations will produce these indications?
c. RHR Heat Exchange b. Thermal Bearing Heat Exchanger I c. Letdown Heat Exchanger d. Seal Water Heat Exchanger Answer d Exam Level B Cognitive Level Comprehension Facility: Braldwood ExamDate: 9/14NS MA: 026 AA1.05 RO Value: 3.1 sRO Value: 3.1 section: EPE RO Group: 1 SRO Group: 1 systenVEvolution Loss of Component Cooling Water KA Ability to operate and / or monitor the following as they apply to Loss of Cornponent Cooling Wa'e The CCWS surge tank, including level control and level alarms, and radiation alarm Explanation of The seal water HX would be the only location where the CC pressure would be lower than the process fluid I Answer pressure. RHR HX approx.165 psig; UD Hx pressure should be approximately 160 psig, & Thermal barrier j pressure should be about 160 psi ; Reference Title / Facility Reference Number Section/Page Revision L CCW malfs/18wOA Pri4 Att B 56
*
3A Pri4 lesson plan Att B 6 3 Material Required for Examination Question Source: Facility Exam Bank Question Modification Method: Significantly ModWied Qut stion Source Comments: Zion 7/13/92 Comment Type Comment ! Friday, SepMmber 4,1998 Page 78 of 100
--
. .,- - ... .- - .. - - . - - - _ . . _ -- - Pressura controller st:p chinga ouestion The follow 7[ing conditions exi:t on Unit 2: - - Reactor power is 100% - Pressurizer pressure control is in automati Wh;t is the immediate response of the pressure control system if the Master Pressure Contioller setpoint is inadvertently changed to 2330 psig (step change)?
e. PORV RY455A opens and spray valves open, b. PORV RY455A opens, spray valves open, and all heaters energiz e. Spray valves open and proportional heaters go to minimu d. Spray valves close and proportional heaters go to maximu Answer d . Exam Lowl B Cognitiw Level Application Facility: Braidwood ExamDete: 9/14/98 KA: 027 AA1.01 RO Value: 4.0 SRO Value: 3.9 section: EPE RO Group: 1 SROGroup: 2 System / Evolution Pressurizer Pressure Control Malfunction KA Ability to operate and / or monitor the following as they apply to Pressurtzer Pressure Control Malfunction: PZR heaters, sprays, and PORVs Explanation of Setting the pot setting higher reduces the output from the controller and raises the demanded pressure Answer setpoint. This reduction results in spray valve closure & heaters turning fully o Reference Title / Facility Reference Number Section/Page Revisio L Par Pressure Controllschematic RY.2/PK 456A in Auto 3 Chp 14 Pressurizerlesson plan 9 30 Material Required for Examination Question Source: New Question Modification Method: Significantly Modified
: Question Source Comments: Calvert Cliffs 11/97 Comment Type Comment
.
A f
5 Friday, september 4,1998 Page 79 of 100
_ .__ _ _ _ _ Quescon 70 Non-Controlling ch nnel fritura The following conditions exist on Unit 1: -
- Reactor power is 100% - All systems are in automatic - Pressurizer pressure channels PT-456 and PT-458 reads normal - Channel i Pressurizer Pressure Channel (PT-455) was declared inoperable and taken out of service with the appropriate bistables placed in the tripped condition . - Controlling pressurizer pressure channel (PT-457) fails high Assuming NO operator action, what is the plant response to the channel failure?
e. Both PORVs and both spray valves open resulting in a reactor trip from low pressurizer pressure followed by Si actuatio b. The reactor will trip on high pressure, and safety injection will actuate on low pressure due to spray valve operatio e. Pressurizer proportional heatere will de-energize and spray valves will open resulting in an OTdT runback prior to reactor tripping, and Sl will actuate due to low pressurizer pressur d. Both PORVs and both spray valves remain closed while pressurizer heaters de-energiz Answer b Exam Level B Cognitive Level Application Facility: Braldwood ExamDate: 9/14/98 KA: 027 AA2.15 RO Value: 3.7 SRO Value: 4.0 Section: EPE RO Gr?;p: 1 SRO Group: 2 Sy: tem / Evolution Pressurizer Pressure Control Malfunction KA Ability to determine and interpret the following as they apply to Pressurizer Pressure Control Malfunction: Actons to be taken if PZR pressure instrument fails high slanation of TWO PZR pressure channels will have HIGH PZR PRESSURE bistables actauted resulting in the reactor tri wer The sparys wil have modulated fully open resulting in actual pressure decreasing (PORV 1RY455A would have also opened on the failure of PT-457, but would close when the PZR pressure fell to 2185 psig PT-458 will actaute the low pressure interlock closing the PORV) until Si occurs at 1829 psig.
Reference Title / Facility Reference Number Section/Page Revision L O.
Pzr Pressure Control / schematic RY-2/ PZR press 3 Chp 14 Pressurizer lesson plan 9 30 Material Required for Examination Question Source: New Question Modification Method: Signiricantly Modiruxj Question Source Comments: BV 8/91 Comment Type Comment
-
Frid;y, September 4,1998 Page 80 of 100
-
r oueecon 7/ Failed levelchannellow.- The plant is oper: ting et 100% power with til control syst ms in AUTO. Tha following paramnt:rs are noted: l
' - Letdown Hx outlet flow (F1-132) - 75 gpm - Charging Header flow (Fl-121) - 87 gpm - Total seal injection flow (FI-142 -Fl -45) - 33 gpm .
Whit is the effect on total seal injection flow initially if controlling Pzr level channel LT-459 fails LOW? Total seal injection flow will... l
c. decrease to 0 gp ; b. decrease to approximately 20 gp j o. remain approximately 33 gpm, d. I'! crease to greater than 40 gpm, Answer d Exam Levet B Cogniuve Level Comprehension Facility: Braidwood ExamDate: 9/14/98 ) KA: 028 AK3.05 RO Value: 3.7 SRO Value: 4.1 Saction: EPE RO Group: 3 SRO Group: 3 Sptem/ Evolution Pressurizer Level Control Malfunction l MA Knowledge of the reasons for the following responses as they apply to Pressurizer Level Control Malfunction: Actions contained in EoP for PZR level malfunction Explanation of The failure of the level instrument low increases charging flow and charging dicharge header pressure. Since Answer seal injection flow is normally increased by throttling close on CV182 to increase backpressure, the result is , the same and seal injection flow will increas A .tence Title / Facility Reference Number Section/Page Revision L CVCS notes / schematic CV-2/cycs ratings 2 18wOAInst 2 Att C lesson plan 9 1 Material Required for Examination Question Source: Facility Exam Bank Question Modification Method: Signifcantly Modified Question Source Comments: Braidwood 1996 NRC exam. ModiruHf premise from failed controller to failed level channel Changed location of correct answer LA%rd on different response (increasing flow instead of decreasing flow).
Comment Type Comment
.
Friday, September 4,1998 Page 81 of 100 _ .
..- . .. - - . . ~. -.
Question [A AMS conditions The following conditions exist on Unit 1: -
. At t= 0 sec, Turbine load was decreased below 352 MW(30% power)
At t=240 sec, The running main feedwater pump trippe The reactor did NOT trip due equipment malfunctio ~ At t=250 sec, All feedflow indications decrease to 0% flow
- At t=320 sec, All steam generator levels decrease below 15%.
Br_ sed on this information, AMS would... c. initiate at t=320 se b. Initiate at t=345 se e. initiate at t=360 se d. NOT initiate because C-20 is cleare Answer b Exam Level B Cognitive Level Application Facility: Braidwood ExamDate: 9/14/9o KA: 2.4.48 RO Value: 3.5 sRO Value: 3.8 section: EPE RO Group: 2 sRO Group: 1 system / Evolution Anticipcted Transient Without Scram KA Ability to interpret control room indications to verify the staic and operation of system. and understand how operator actions and directives affect plant and system condition Explanation of AMS remains armed for 6 minutes (360 sec) following decrease below 30%(C-20). The actuation siganiis Answer generated after 3/4 SGs level have fallen 3% below the LO-2 (reactor trip) setpoints of 18% for a period of 25 seconds. C-20 would clear @ t=360sec. AMS actuation occurs at 320 + 25 = 345 se Ra4rence Title / Facility Reference Number section/Page Revision L i / schematic PN-3/ logic 1 schem 2 Chp 60b 6 7 Material Required for Examination Question source: New Question Modification Method: Question source Comments: Comment Type Comment _ Friday. September 4,1998 Page 82 of 100
~ _ _
_ Question b Evtluttion of SR NIS voltags frilura Th3 following conditions exist on Unit 1:
- Reactor startup in progress-Intermediate power range indication: 2.5E-5 amp N35 & 2.8E-5 amp N36 - SOURCE RANGE PERMISSIVE P-6 permissive iighc clear - Source Range Channel N31 high voltage power s'upply fails to HALF its normal value What indication (s) would be available to alert the operator to this failure?
e. None, until power is lowered below the P-6 setpoint, and then the Source Range N31 indication will indicate lower than expecte b. None, until power is lowered below the P-6 setpoint, and then the Source Range N31 indication will indicate higher than expecte c. Annunciator SR HIGH VOLTAGE FAILURE (1-10-81) will remain in alarm when power exceeds P-1 d. Annunciator SR HIGH VOLTAGE FAILURE (1-10-B1) will re-flash when the voltage source fail Answer a Exam Level B Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 032 AK1.01 RO Value: 2.5 SRO Value: 3.1 Section: EP Ro Group: 2 SRO Group: 2 Sy tem /Evoludon Loss of Source Range Nuclear instrumentation KA Knowledge of the operationalimplications of the following concepts as they apply to Loss of Source Range Nuclear Instrumentation: Effects of voltage changes on performance Explanation of Based on Gas filled detector curve (Region lil), the number of events collected would drop (counts drop).
Answer Alarm and voltage input to SR detector is blocked until both IR NIS fall below the P-6 setpoint.
I once Title / Facility Reference Number Section/Page Revision L O.
Set High Volt Failure /18 WAR 1-10-B1 setpts/ notes 1 Source Range detector / schematic NI-4 4 Chp 31 source range nuclear inst Lcsson plan 6 2,3,11,12 Material Requ8 red for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment
.
l
,
Friday, September 4,1998 Page 83 of 100
. . -- . - - - - - . - -
ousetson I[ Evil of failed IR chinn:1 on SU Tha following conditions exists on Unit 2: -
- Plant shutdown is in progres Power range channels indicate: 9% (N41),10% (N42),11% (N43) ,11% (N44) - Intermediate range channel N-36 fails HIG Wh:n this failure occurs, what is the plant response this failure?
c. The reactor will trip on high IR flux, and source range trip will reinstate when N-35 decreases below P- b. The reactor will trip on high IR flux, and source range trip will NOT be automatically reinstate c. The reactor will NOT trip immediately, but will trip when the source range trip is reinstated when N-35 decreases below P-6 , d. The reactor will NOT trip, and source range trip will NOT be automatically reinstate Answer d Exam Level B Cognitive Level Application Facility: Braidwood ExamDate: 9/14/98 KA: 033 AA2,04 RO Value: 3.2 SRO Value: 3.6 section: EPE RO Group: 2 SRO Group: 2 system / Evolution Loss of Intermediate Range Nuclear Instrumentation KA Ability to determine and interpret the following as they apply to Loss of Intermediate Range Nuclear instrumentation: Satisfactory overlap between source-range, intermediate-range and power-range instrumentation Explanation of Since reactor power is < P-10 setpoint (10% power), the IR trip setpoint at 25% EICAwill be exceeded Answer resulting in reactor trip. SR will NOT be reinstated automatically because only one IR channel will fall below P-6 and Two are required to remove P-6.
Reference Title / Facility Reference Number Section/Page Revision L. O.
Ir' mediate Rangelschematic NI-3 4 l L 2 Intermediate range nuclear inst 1 Lesson plan 6 4,8,9,10 l Material Required for Examination ) Question Source: New Question Modificatio 1 Method: Significantly Modified ! Question Source Comments: Watts Bar 8/94 Comment Type Comment l l I
l l l
.
'
!
Frid:y, September 4,1998 Page 84 of 100
. . --
Quesson II Monitors for S/G Tube Irktg2 The following conditions exist on Unit 1: . l
- Reactor power is 75% ; - Troubleshooting has commenced due to reduced condenser vacuum l with the air ejectors out of servic ' - Hogging vacuum pumps are aligned to the main condenser to aid in maintaining vacuum.
- Wh::_t would NOT be an indication of a Steam Generator Tube Leak under these conditions? e. Increasing conductivity levels for the main condenser hotwel b. Increasing radiation level on 1RE-PR027,"SJAE/ Gland Steam Exhaust Monitor". ' e. Decreasing feed flow to ONE S/ d. Increa3ing radiation levels on 1RE-PR08 "S/G Blowdown Monitor".
Answer a Exam Level R Cogniuve Level Comprehension Facility: Braldwood ExamDate: 9/14/98 KA: 037 AA1.02 RO Value: 3.1 SRO Value: 2.9 Secuon: EPE RO Group: 2 GRO Group: 2 Sy: tem /Evoludon Steam Generator Tube Leak KA Ability to operate and / or monitor the following as they apply to Steam Generator Tube Leak: 1 Condensate exhaust system Explanacon of The Hogger discharge is aligned through the Off Gas header which is monitored by 1RE-PR02 Answer Reference Title / Facility Reference Number Section/Page Revisio L ) SGTR lesson plan / Bw0A Sec 8 6 4 ) Ch 49 rad monitors lesson plan 7 14
) **gterial Required for Examination l l
estion Source: New Question Modification Method-Question Source Comments: l Comment Type Comment l l l l l I l
,
i
'
.
I I
!
Frid:y. September 4,1998 Page 85 of 100
. _ _ . . _ _ _ _ _ ___. _ _ _ ._ _ _ _ . _ _ . _ __ ._
%etion ' Loss of subcooling u - Zo-3 "St=m Gen:rstor Tube Rupturs"is being performed in responso to a tube ruptura on 2C S/ . The cooldown has just been completed but the target temperature value selected by the operators was higher than that stipulated in the procedur Whit condition could result because of this error? j o. Loss of RCS subcooling before RCS and ruptured S/G pressures are equalize s. Increase in pressure of the ruptured S/G with resultant lifting of the S/G Safety Valve, c. Increase in pressure of the non-ruptured S/Gs with resultant lifting of their S/G Safety Valve d. Filling the Pressurizer solid during the subsequent depressurizatio Answer a Exam Level B Cognitive Level Application Facility: Braidwood ExamDa'te: 9/14/98 '
l KA: 038 EK3.06 Ro Value: 4.2 SRo Value: 4.5 Section: EPE RO Group: 2 SRoGroup: 2 System / Evolution Steam Generator Tube Rupture KA Knowledge of the reasons for the following responses as they apply to Steam Generator Tube Rupture: Actions contained in EoP for RCS water inventory balance, S/G tube rupture. and plant shutdown procedures Explanation of Anower Reference Title / Facility Reference Number Section/Page Revision L SGTR lesson plan 18wEP-3 12 1 l ERG basis Material Required for Examination Ouestion Source: New Question Modification Method: Editorially Modified estion Source Comments: Salem 6/94 Comment Type Comment
.
Frid::y, september 4,1998 Page 86 of 100
. _ . - . - . _ . .. - - . - . ..-
Quescon U Strmlineisolition The following conditions exi t on Unit 1: -
- The Unit was in MODE 3 at normal operating temperature and pressure prior to the even A faulted steam generator has occurre RCS hot leg temperatures - 547*F (A),544*F (B),545*F (C),547*F (D) - RCS cold leg temperatures - 545'F (A),530*F (B),543*F (C),545*F (D) - S/G pressures - 700 psig (A),635 psig (B),690 psig (C),705 psig (D) - S/G flow - 0.85 MLB/hr (B) - Containment pressure (Channel)- 8 psig (1),7.5 psig (2),7.5 psig (3),8 psig (4)
Based on these conditions, a main steam line isolation should... c. have occurred because of the low pressure in at least ONE S/ b. have occurred because the steamline high negative rate occurred in SIG 1 e. NOT have occurred because Containment pressure is below the setpoint for the CNMT High-2 l pressure signal.
, d. NOT have occurred because THREE S/Gs have pressures above the isolation setpoint and do NOT
- indicate high steam flo Answer a Exam Level B Cognitive Level Application Facility: Braidwood ExamDate: 9/14/98 )
KA: 040 AA1.01 RO Value: 4.6 SRO value: 4.6 Section: EPE RO Group: 1 SRO Group: 1 System / Evolution Steam Line Rupture KA Ability to operate and / or monitor the following as they apply to Steam Line Rupture: l Manual and automatic ESFAS initiation l Sanacon of The steamline isolation signal is generated by the low pressure sensed on 2/3 pressure transmitters in any !
., ewer one SG. CNMT pressure is below the MSLI setpoint of 8.2 psig and steamline negative rate is blocked since l l
In!?ial condition has PZR pressure > P-1 Reference Title / Facility Reference Number Section/Page Revision L ESF S:tpoints/ schematic EF-2/ Stmline iso! 5 Ch 23 Main steam Sys lesson plan 8 5,13,15,16 I Ch 61 ESF lesson plan 5 7 Material Required for Extmination Question Source: New Question Modification Method: Question Source Comments: j Comment Type Comment l l
. .
Friday, September 4.1998 Page 87 of 100
_ _ . _ . _ _ _ . _ . . . ._ - - _ _ _ . ouestion 7[ Evilof Leak 1
The following conditions cxi t on Unit 1 following a trip from 100% pownr: - l
- Pressurizerlevelis 0%
Pressurizer pressure is 1500 psig
- Containment Pressure is 16 psi Tcold is 420*F for all loops, i
Where is the location of the leak?
:
e. On one loop RCS cold le b. On a Main Steam Line inside containment, e. In a Steam Generator Tub d. On a feedwater line between FWRV and Associated FWlV,1FW00 l Answer b Exam Level B Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 j KA: 040 AK1.06 RO Value: 3.7 sRO Value: 3.8 Section: EPE RO Group: 1 SRO Group: 1 system / Evolution Steam Line Rupture I KA Knowledge of the operational implications of the following concepts as they apply to Steam Line Rupture: ) , High4nergy steam line break considerations i Explanation of Secondary LOCA not indicated since Tcold is the same in all loops and RCS tcold is not consistent wth given Answer CNMT pressure for steam / feed break. SGTR not indicated since CNMT pressure is elevated. LOCA condiiton I indcated by consistent Tcold, and CNMT pressure increas Reference Title / Facility Reference Number Section/Page Revisio L BwEP-0 Reactor Trip or Si lesson plan 3 6,7 iPwEP2 Faulted S?g isolation lesson plan 7 2,4 l
'
Material Required for Examination Question Source: New Question Modification Method: Editorially Modired , Question Source Comments: St. Lucie 10/13/97 Comment Type Comment I
<
Frid;y, September 4,1998 Page 88 of 100
i Question [[ Eval of conditions in record:nce with BwOA SEC-3," Loss of Condens:r Vccuum", which of the following s ts of conditions requires the operator to trip the reactor? c. LOW POWER TRIP BLOCKED P-8 annunciator- LIT l Turbine load - 200 MW Condenser pressure - S.2 " HgA
.l b. LOW POWER TRIP BLOCKED P-8 annunciator- LIT Turbine load - 300 MW Condenser pressure - 6.3" HgA - c. LOW POWER TRIP BLOCKED P-8 annunciator- CLEAR Turbine load - 600 MW Condenser pressure - 7.2" HgA d. LOW POWER TRIP BLOCKED P-8 annunciator- CLEAR Turbine load - 900 MW Condenser pressure - 7.8" HgA ' Answer b Exam Level B Cognitive Level Application Facility: Braldwood ExamDate: 9/14/98 l KA: 051 AA2.02 RO Value: 3.9 SRO Value: 4.1 Section: EPE Ro Group: 1 SRO Group: 1 System / Evolution Loss of Condenser Vacuum KA Ability to determine and interpret the following as they apply to Loss of Condenser Vacuum:
Condrtions requiring reactor and/or turbine trip Explanation of P-8 permissive active below 30% power (annunciator lit). At 480 MW and below, the minimum acceptable Answer condenser pressure is 5.5 in HgA. At 600 MW minimum acceptable pressure is 7. 8 in HgA. At 610 MW and greater, minimum acceptable pressure is 8.0 in HG F we Title / Facility Reference Number section/Page Revision L h ..OA Ses -3 loss of condenser vac lesson plan 6 5 Material Required for Examination Figure 18wOA SEC 3-1 Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment
-
1 Friday, September 4,1998 Page 89 of 100
.
,
l Queouon f 0 ' identification of RCP seil LOCA/cooldown Select th3 prim:ry r cson for r:pidly d: pressurizing the strem g:nnrators during a Loss of All A l c. To provide maximum core cooling until power can be restored, l s To minimize RCS inventory loss from RCP seal e. To enhance restoration of S/G level from the diesel driven AF pum d. To increase subcooling of the RC Answer b Exam Level B Cognitive Level Memory Facility: Braidwood ExamDate: 9/14/98 l KA: 055 EK3.02 RO Value: 4.3 SRO Value: 4.6 Section: EPE RO Group: 1 SRO Group: 1 System / Evolution Station Blackout KA Knomkdge of the reasons for the following responses as they apply to Station Blackout: I Actions contained in EOP for loss of offsite and onsite power Explanation of The rapid cooling allows depressuring the RCS reducing the leak rate via the RCP seals l Answer Reference Title / Facility Reference Number Section/Page Revision L O.
Loss of All AC Power /18wCA Caution 2 1B Wog 1B l 1BwCA 0.0 lesson plan 4 Material Required for Examination
. Question Source: New Question Modification Method:
Question Source Comments: Comment Type Comment
)
l i i I l l l l l
-
l I l l
'
Fridry, September 4,1998 Page ',0 of 100 I
. . _ . I oueetion k R:; set of sequencer How would the sequ:ncer opercta if a Saf:ty inj:ction (SI) cctuation occurs while tha s::qu:nc:r is I sequencing loads in response to an ESF bus undervoltage condition? II c. There will be no change in operation; the undervoltage sequence overrides the Si sequenc b. The undervoltage sequencing stops, the sequencer immediately resets and Si loads NOT already running will sequentially star c. The undervoltage sequencing stops, all started loads are shed, and Si loads will sequentially star d. The undervoltage sequencing completes its cycle, then resets to SI mode, and SI loads NOT already running will sequentially star l Answer b Exam t.evel B Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 056 AA1.21 Ro Value: 3.3 SRO Value: 3.3 Section: EPE RO Group: 3 SRO Group: 3 i SystemrEvolution Loss of Off-Site Power l KA Ability to operate and / or monitor the following as they apply to Loss of off. Site Power: J Roset of the ESF load sequencers l Explanation of The UV sequence is stopped and the SARA sequencing is initiated from step I Answer '
Reference Title / Facility Reference Number Section/Page Revision L l Dl3 Relaying schematic DG 2/ SARA & SDRA 1 Ch 9 EDG and Aux sys lesson plan 7 7 Ch 4 AC Electrical distribution lesson plan 8 10,16 j Ch 61 ESF lesson plan 5 7,8 l Material Required for Examination Question Source: New Question f 4odification Method: Significantly Modified Question Source Comments: vogtle 5/91 Comment Type Comment l
!
l l l
- 1
!
!
Frid:y, September 4,1998 Page 91 of 100 ,
;
. . _ -
Question k Evalof electric bus status The following conditions exist on Unit 1: .
- Bus 141 is powered from its normal source - D/G 1 A surveillance is being performed with the D/G paralleled to the bus Wh t would occur if a failure of the undervoltage relay results in a sensed undervoltage condition on Bus 1417 c. SAT feeder breaker ACB 1412 and D/G feeder breaker ACB 1413 remain closed. The Safe Shutdown loads will NOT sequence and CANNOT be manually started from the control roo b. SAT feeder breaker ACB 1412 and D/G feeder breaker ACB 1413 will open. After a 10-second delay, ACB 1413 will close and the Safe Shutdown loads will sequenc e. SAT feeder breaker ACB 1412 will open but D/G feeder breaker ACB 1413 will remain closed. The Safe Shutdown loads will sequence normall d. SAT feeder breaker ACB 1412 will opea but D/G feeder breaker ACB 1413 will remain closed. The Safe Shutdown loads will NOT sequence and CANNOT be manually started from the control roo Answer d Exam Level B Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 056 AA2.46 RO Value: 4.2 SRO Value: 4.4 Section: EPE RO Group: 3 SRO Group: 3 Sy:temnvolution Loss of Off-Site Power Ability to determine and interpret the following as they apply to Loss of off-Site Power That the ED/Gs have started automatically and that the bus tie breakers are closed Explanation of On sensed UV, the SAT feeder breaker opens (and alternate feeder breaker would also have opened if closed)
Answer and the control switches for the safe shutdown loads will be locked ou Reference Title / Facility Reference Number Section/Page Revision L Ch 4 AC Electrical Distribution 8 10,16 l Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment l
~
i l !
i l Frid:y, September 4,1998 Page 92 of 100
_ oue tlon h Eqpt effected on bus to:s On Unit 1 power is lost to 120 VAC instrum:nt Bus 111 . How are the ESF and Safe Shutdown loads affected? a. "A" Train ESF loads will NOT load on an Si signal, but Safe Shutdown loads will load on a UN signa "B" Train loads are NOT affecte b. A" Train ESF loads will load on an Si signal, but Safe Shutdown loads will NOT load on a UN signa "B" Train loads are NOT affecte c. "A" Train ESF loads will NOT load on an Si signal, and Safe Shutdown loads will NOT load on a UN signa "B" Train loads are NOT affecte d. "A" Train AND "B" Train ESF loads will NOT load on an Si signal, but Safe Shutdown loads will load on a UN signa Answer C Exam Level B Cogniuve Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 057 AA2.19 RO Value: 4.0 SRO Value: 4.3 Section: EPE RO Group: 1 SRO Group: 1 System / Evolution Loss of Vital AC Instrument Bus KA Ability to determine and interpret the following as they apply to Loss of Vital AC instrument Bus: The plant automatic actions that will occur on the loss of a vital ac electricalinstrument bus Explanation of Answer Reference Title / Facility Reference Number Section/Page Revision L ' OA Elec 2 Loss of inst bus Table A 7 G ,dO2 SSPS lesson plan 3 11 1BwOA elec 2 lesson plan 6 3,5 I cnd C system notes 1&C1 Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment
.
I l l l Friday, september 4,1998 Page 93 of 100
. - - - -. . .
Que. con Operations required for tr:nsfsr Which of tha following snts of indications cra cvril: bis on the Remots Shutdown Panel? j o. Emergency boration flow, S/G level, and RCS wide range temperatur b. Red and green lights for reactor trip breaker position indication, S/G pressure, and pressurizer level.
l e. Main feedwater flow, letdown flow, and charging line pressur d. Containment pressure, charging fbv. and auxiliary feedwater flow.
! Answer a Exam Level B cognitive Level Memory Facility: Braidwood ExamDate: 9/14/98 Tier: Emergency and Abnormal Plant Evolutions RO Group: 1 SRoGroup: 1 068 Control Room Evacuation AA1. Ability to operate and / or monitor the following as they apply to Control Room Evacuation: AA1.12 Auxiliary shutdown panel controls and indicators 4.4 Explanation of Anewer Reference Title / Facility Reference Number Section/Page Revisio L RSP PLO4/SJ/ schematic PN-1 2 Control Room inaccessbility 18wOA Prl-5 tesson plan Att. A 57B Ch 62 Remote shutdown Panel t.esson plan 3 3,4 Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment l i Frid:y, September 4,1998 Page 94 of 100
ouestion Mrjor action categori:s When inrdequats cora cooling exists, which of th3 following sets of cctions states tha pronai sequsnce of the major action categories to be performed in accordance with BwFR-C.1, " RESPONSE TO INADEQUATE CORE COOLING", for removing decay heat from the core? 4. Rapid secondary depressurization; reinitiation of safety injection; RCP restar b. Reinitiation of safety injection; rapid secondary depressurization; RCP restar c. Rapid secondary depressurization; RCP restart ; reinitiation of safety injectio d. RCP restart; rapid secondary depressurization; reinitiation of safety injectio . Answer b Exam Level B Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: 074 EK1.03 RO Value: 4.5 SRO Value: 4.9 Section: EPE RO Group: 1 SRO Group: 1 system / Evolution inadequate Core Cooling KA Knowledge of the operationalimplications of the fonowing concepts as they apply to inadequate Core Cooling: Processes for removing decay heat from the core Explanation of Answer Reference Title / Facility Reference Number Section/Page Revisio L Function Restoration Procedures BwFR-C.1, C.2, 5 2,3 C.3 lesson plan j i Material Required for Examination l
'
Question Source: New Question Modification Method: Editorialty Modified Question Source Comments: VC Summer 5/94 l Comment Type Comment !
I i
-
Fridry, September 4.1998 Page 95 of 100
. - , - - .- .- ..
I oue. tion 14 Achons for reducing acuvity High coolant cctivity has been datteted cnd ch:mistry has dstsrmined that it is dua to corrosion product i cctivation.
antify the effect of placing the cation demineralizer in servic Th] cation demineralizer... e. will remove lithium so it should NOT be used in this conditio b. will cause the activity level to decrease as soon as it is placed in servic ! e is NOT effective in removing corrosion product activity, d. ls less effective than the mixed bed demineralizer so it is placed in service ONLY if decontamination factor is less than 1 Answer b Exam Level B Cognitive Level Memory Facility: Braidwood ExamDate: 9/14/98 KA: 076 AA2.02 RO value: 2.8 SRO Value: 3.4 Section: EPE RO Group: 1 SRO Group: 1 , System / Evolution High Reactor Coolant Activity ) KA Ability to determine and interpret the following as they apply to High Reactor Coolant Activity: ) Corrective actions required for high fission product activity in RCS ! Explanation of The cation demin is highly effective in removing corrosion products from the coolan i Answer
.
Reference Title / Facility Reference Number Section/Page Revision L BwOP CV-8 1BwOA Pri-4 High coolant Activity lesson plan 1 4,5 ch 152 CVCS lesson plan 10 4 terial Required for Examination , Question Source: New Question Modification Method: I Cuestion Source Comments: Comment Type Comment l l _ Frid y, September 4,1998 Page 96 of 100
_ Queedon 7/ Int:rlocks effecting reestablishment of feed The following conditions cxist on Unit 2: I
- Reactor power was 8% prior to the event belo A failure in the feedwater control system caused ONE S/G level to rise to 83%. - The main turbine trippe S/G levels have returned to their normallevel range - The Startup FW Pump is running Wh:t are all the conditions that would have to be met to feed the S/Gs using the FWO34's Feedwater Tempering Flow Control valves?
c. The FW lsolation Aux Relays would have to be reset and FWO35 Feedwater Tempering isol valves opene b. The reactor trip breakers would have to be cycled, the FW lsolation Aux Relays would have to be reset and FW035 Feedwater Tempering Isol valves opene c. The FW lsolation Main Relays and Aux Relays would have to be reset and FWO35 Feedwater Tempering isol valves opene I d. The reactor trip breakers would have to be cycled and FW lsolation Main Relays and Aux Relays reset l and FWO35 Feedwater Tempering isol valves opene Answer a Exam Level B Cognidve Level Application Facility: Braldwood ExamDato: 9/14/98 KA: EOS EK RO V iue: 3.7 SRO value: 3.9 Section: EPE RO Group: 2 SROGroup: 2 Sy: tem / Evolution Loss of Secondary Heat Sink KA Knowledge of the interrelations between Loss of Secondary Heat Sink and the following: Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual feature Expl:nadon of The P-14 signal, once clear, only mainaitns FW1 signal via the FW isol Aux relays if NO reactro trip signal is Answer present. So reseting the FW lsolation Aux relay allows opeing of FWO35s (normal feed path at low power) and throttling of FWO34s Reference Title / Facility Reference Number Section/Page Revision L. O.
ESF setpoints/ schematic EF-2/ reset FWI 5 Feedwater Simple /SGWLC FW-1,2/ reset FWI 0 Ch 61 ESF lesson plan 6 4,7,8 Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment
-
Frid y, September 4,1998 Page 97 of 100
.
Topic Question f[ Identification of helt r:,rnoval proccss , The following conditions exist on Unit 1:
- A leak developed on the RCS loop C flow instrument pipin Coincident with the RCS leak, on the reactor trip a S/G PORV failed open and was later isolate FR-P.1 was entered to due to an ORANGE PATH conditio Si actuated and has been rese All RCPs are stoppe Conditions required to support an RCP start are me Under the current conditions starting the RCP will...
c. cause excessive thermal stresses in the stagnant loop b. cause a pressure surge that will aggravate the PTS conditio c. provide mixing of the ECCS injection flow thereby decreasing the likelihood of PT d. increase the RCS cooldown rate thereby increase the likelihood of PT Answer C Exam Level B Cognitive Level Comprehension Facility: Braidwood ExamDate: 9/14/98 KA: E06 EK RO Value: 3.6 SRo Value: 4.0 Section: EPE RO Group: 1 SROGroup: 1 system / Evolution Title: Pressurized Thermal Shock KA Statement: Knowledge of the interrelations between Pressurized Thermal shock and the following: Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facilit Explanation of
'swer Reference Title / Facility Reference Number Section/Page Revisin L O.
FRP 18wFR P.1,2, lesson plan 4 3,4 Status Trees ST-l/ Integrity Material Required for Examination Question Source: New Question Modification Method: Question Source Comments: Comment Type Comment
.
Frid;y, September 4,1998 Page 98 of 100
+ ., .--a -~a . - an .. x u x u -..+-a.- .s... +, . >u_>>., ,.,.-a ..s. .
l Question df f Naturel Cire conditions End limits . The following conditions exist on Unit 1: . j s - A natural circulation is in progress per BwEP ES-0.2 " Natural Circulation Cooldown"
- - Pressurizer pressure is being controlled using Aux. Spray and Pzr heaters - As pressure is being lowered through 1300 psig, a rapid increase is noted in Pzr level - Charging and letdown are in manual and are balanced
Whht actions are required to be taken by the operators?
-l c. Repressurize the RC I b. Isolate the SI Accumulator e. Increase the RCS cooldown rat ]
d. Place excess letdown in servic Answer a Exam Level B Cognitive Level Memory Facility: Braidwood ExamDate: 9/14/98 Tier: . Emergency and Abnormal Plant Evolutions RO Group: 1 SROGroup: 1
. E09 Natural Circulation Operations EK3. Knowledge of the reasons for the following responses as they apply to Natural Circulation Operations; l EK Facility operating characteristics during transient conditions, including coolant chemistry and the effects 3.3 of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristic .
. Explanation of Answer Reference Title / Facility Reference Number Section/Page Revision L BwF.P -0 Reactor Trip or SI Lesson plan ' 11 3,4,6
.
Material Required for Examination Question Source: New Question Modification Method: Cuestion Source Comments: Comment Type Comment i
' ,.
.
J
,
FridIy, September 4,1998 Page 99 of 100
. - . . - - . ..
.
Question /N Reason for rapid S/G deprrssurizrtion Why cra the S/Gs d: pres:urized to less than 670 psig cccording to BwCA-1.1," Loss of Emergency Coolant Recirculation"? a. To allow maxirnum AFW flow to the S/G e. To ensure adequate subcooling for restart of the RCP c. To set up conditions for controlled injection to the RCS from the accumulator d. To decrease RCS temperature and pressure which reduces break flow in a LOCA conditio Answer c Exam Level B cogniuve Levet Memory Facility: Braidwood ExamDate: 9/14/98 KA: E11 EA RO Value: 3.9 SRO Value: 4.0 section: EPE RO Group: 2 SROGroup: 2 System / Evolution Loss of Emergency Coolant Recirculation KA Ability to operate and / or monitor the following as they apply to Loss of Emergency Coolant Recirculation: Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual feature Explanation of The concern is maximizing cooling volumes that supply water to RCS. By cooling RCS, depressurization of Answer RCS can be initiated (while maintaining subcooling) to the point where the Si accumulators inject their volumes into the RCS, Reference Title / Facility Reference Number Section/Page Revision L O.
Loss of Emergency Coolant Recirc/1BwCA-1,1 1BWOG1B 18wCA 1.1 and 1.2 lesson plan 7 3 Material Required for Examination Question Source: New Question Modification Method: Editorially Modified Question Source Comments: South Texas 9/92
. nment Type Comment l
l l l l l l l Friday, September 4,1998 Page 100 of100
_ _ _ . _ _ . . _ __ __ _ GENERIC FUNDAMENTALS EIANINATION EOUATIONS AND CONVERSION 8 EANDOUT SHEET
* ~
BOUATIONS .
-
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ______ _________4____________-
, .
= AC,AT . P = P.10"*3 6 = 66h P = P e"I
~
= UAAT ,
A = A.e ^' 6 oc %.scite #" ~ !
~
"
.
CR (1 - K.cri) = CR2 (1 - K.crz) t AToc A,tcire .
.
1/M = CR t/CRx K cr = 1/ (1 p) I p = - (K.cr - 1) /K.cr F = PA SUR = 26.06/r
& e- pA9 *
N' p r sw = idPu A.cr # E = IR ' p.[. Eff. = Net Work Out/ Energy In f 1+ h.cr f v(P2 - P ) + (% t
g ) + g(z z-Z)=0 1
. 4 f = 1 x 10 seconds 2g, ge -
herr = 0.1 seconds *
*
g, = 32.2 lbm-ft/lbf-sec* - CONVERSIONS l _____________________________________ ____ ____________________ 1 Mw = 3.41 x 10' Btu /hr 1 Curie = 3.7 x 10** dps * 1 hp = 2 . 5 4 x 1 0 ' B t u /.h r 1 kg = 2.21 lbm i 1 Btu = 778 ft-lbf 1 gal ., = 8. 35 lba
*C = (5/9) (*F - 32) 1 f t*, t., = 7.48 gal *F .= (9/5) ( *C) + 32 l
l
*
i
!
>
, ,. . . . . .. .
,
l
~ . ..
. . - . - . . - . . - . . - . . ~ . - . _ . ~ .~. . _ - - - . . . .- .,.
.~ . ~ +
nsII clur-t east .
" '
et i
'
6 W' " M* * S a M8 M
.
. rea I
=I ,
te80.eC4> 0 N* 48t80h4> T85. $114J1t-M 50M M
\ DMs_spre4.c-,
- *
,
sit 6 _ d f K* p 7 ; a3"m Ov- W .Li ll GA.C .
, 1
-
,.= .n n H =n*n- 4 . ; o ;
"1 --
' .a.wo aimo swan >====>a =oa
, , i ,, g._t . .
NECle i
- - . - , . - - -
4 , M4I f I A. F. TT8_C
-- -r --
. -
,
m' in '
#L e_ w i M 4 H E ,,
yt _rmL : - -> Q aw
-
I
-- ] m - - -
.
u ' V: w I wou **esseow
" ,' = E_ _=
,,,
* lie o g . ir I e e i 4 __
>
.. -
a 'm
'
:~~ , p-- -- - - J ,
,
,
- .- i ,
,
q '= .m i
.
IIi + 1 = l
.; I ,_ L - - - - - - e. ,9 ---
,
#
.;
f si n . ,
, , -
d. .
.
-,
< ene- ois n t 6LO8'" -
(ttlotP. =
,
ll*N M ,,.
*
a P
*
d __[ g GedhCf B906 g gg, __ v, g,
-
. 'n o40 y n m -
.
rw , . = x n a
,. - gywg ara .see O M M e A4 - to C M E at . gs vp m a As - se o A m cs e, nn c w ai - et C av,ip i*8-')*
ba-see -
,
8****** = = am am re me 4E-es t *' C et - One O W M G K
' U" J- -
c. - et c . . <. .- = C m x
., Oj" c. . c,
.
.
.. = n .
. . s
.
.j ~_,,1
-* - 0-= ** = = . -
.me.a e m
. .Ma.EE- Of
=erti d
". a.: -ei pie . es com as u a u 8 M 56 = 5 P se 4 M.*h 5 . ,1i . = = m a 1 4
'"' ""*
=
' m,osseo
.~ I.' M.' '"'.E..mm
.a cias .
. eu . "Ja".
;
'
6 5 Y x. .- ==i='= m. - eo.*=e come ==+== . I .I w@p e onive osvema ew
**~~" ! :
st - 8E
.
,. -
$ ' % EAT s 5 t
. _ _ _ _ .. ... . . . _ _ . _ _ _ _ . . . . _ __ _ _ _ . _ _ _ _ . . . - _ _ . _
.
*
. .
_
* ,
_"'"'# 'tp.ee Y* M "Y-O. Mt
,
*co
.
-t'
.
,
wem, eso e me
< 4 wism, , ,,
N
* *83 0 " + casas ,. ,
N > _
,* a ,* ses <>s:s vit vil un
** :T- *** s g .
- <
og
.r. * + n- ~ u
:
.E a :str :ce ::xs -:str .:sse v ::s eer . au su
:: %. ::]p w u' vu
-, -,
r_2 ,- ==
,4 W9 w asec o w ensco y g *"
7 7 7 + 7 !! p=E- ;:t L y.g E */ woestyg o il , :ll el , t Yi II ,
-i =o was = *
wnwo rawy NY '-
, ,4 .m , *
j p_. ., w 5i
- ' kla18 +
M[?h N MN
- =-* %.
McHM+
-
_ _ _ _ _ . . Nw**4 . _ _ _ _ _ . ,
-
, ret was , -l o W -
.m e l . Tjpt* .l e- ,. .
E
'im; NUCLEAR SAFETY RELATED ac:.nz: is sHOWN 006 DES DRAW 968 M an, u.a.co, ar..aticero ser.26 ;6i:
'
N 8"*** "'"""""-- Yan*,eavat MTIiN e s' - met cars 'assearms a d 1stoien ossa. se otscairymse " ' 7C
**'"'"' ","JPJ P 21, ***" K .+a QM ausmespusuuowooo sentiose -usertt
.g C00SSONwSAA.W M C e ne= ='m s.gAeDes s+4423 sesues wercur acnannemsuo cas c.srsmans seassaicana a ernias m 3+ 4ess seau writar uso ammuscuttet erur can: (tse es) erops
"**** g mgg v
,
'7- 1een6a so.44ei samu wetze w,o assov ssrs emnem
-
_ e+=se mist s
'
g .' - ,a nswa _
(Ty - g
_, w,,......s,-, . _7 M - A """"'""" 3 , sw essues eeners we m mar. smus caans)
~~
;- -~
K '"* RE n h_; . . s . ,
'
07 L fur- 9'846) fHtd.kn44tw) - s i.,.
~^
. . . . ' . ---.-
_ _
' . BwCB-2 Figuro 12
~Revo0 8/10/87 '
Page 1 of 1 l VlULTIPLE USt t d - l l
-
BORON DILUTION RATE NOM 0 GRAPH l
.
- 30M ,,,,g.P) utiba)
-
, POWER 509,342 2000 -
525 531.756 500 3, ,347
-
U 40 ,763
' Soo 300 636,314 ~ -
200 '345,537- _ _ inoo
-
'" .400
'
-
-
.
200 l
' - .
'
-
.* ion .
- 7 = _
500 1- C
- 400 5 . "go 7
-
a __
= >= m . -
s . = . - l 200 * $~
- 70 w es -
W _ E .
g so
-
8 [ = - g _
= O ,- **- '
" $Q -
= _ , l-
_
- 8 ! "
E - . -
-- 2 30
# I .
30
*
L 500 CY d N -
'
M= TOTAL SYSTEM MASS
/], '
O APPROVED 7984P(080587)18 NOV 271987 enAtowo,9R.,
. --
- . . - - . - . - - . - . - - - - - .- - .. - .. .. - _ - - .. -
, ! ! REV.' 53 LOSS OF CONDENSER VACUUM 1 Bid)A
- ," UNIT 1- .
SEC-3
- -
l . ! ! - i . . t
'
n I
-
.
- - ,
,
! .
,
i C 10
i N
: D , - . . - . .
NOT ! N" ACCEPTABLE S j E8 r { R j
>
j A y / ! B a
'
i l
O g L , l , U
T i
.
E5 P
., R4 ACCEPTABLE E
S S
.U R -'
.__
E ~
I ' y 4-
H
Ao o 100 200 300 400 500 600 700' 800 900 1000 1100 1200
. NEGAWATTS APPROVED FIGURE IBw0A SEC-3-1 APR'211994 TURBINE LOAD -vs- CONDENSER PRESSURE ,
ONeelfs ngyggw Page 9 of 9
._ _ _____ _
- . _ . . . . _ . . ~ . . _ _ _ _ _ _ .
e REV.fi7 LOSS OF RH COOUNG UNIT - 1 1BwOA PRI 10 __
.
.
.
MAKEUP 12 . s
. . . .
11 anys nuva
.; 4 - ss s= 192 10 )
. , .
e u= ,
( 16 = 384 0 , , ,,,,
24 = 576
'8 e - 672 32 = 7A8
. ...
, ,
- ?..
0 = 960
' ..
0 * s ~ ' 5 .
'
-
~ ~ .
'3 't.0 ' .100.0 .300.0 300.0 400.0[510.0 000.0' 700.0 000.0 50 TIE AFTER 901T01504 95I51
..
l _ FIGURE 18wOA PRI 10-3 Minimum Makeup Flow Required to Match Boiloff
'
) Page 11 of 62 APPROVED'
(11/21/96)
,
!
i NOV 261996 BRAIDWOOQ 9es stic acview _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ . _
. . _ . . .__ . . . . _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ . . _ _ . . _ . _ _ . _ _ __ _ _ . . _ .
REACTOR TRIP RESPONSE '1B EP REV/. 10 S - WOG 18 UNIT 1 . .
,
,
,
. . .
.
, . l l ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED l STEP VERIFY ALL' CONTROL RODS FULLY Perform the following:
INSERTED: a. IE two or,more rods are e All rod bottom lights - L71 M fully. inserted, .
~
THEN emergency borate 1-200. GAL (3600 GAL FROM I RWST) for'each rod M
-
-
fully inserted per 1BwCA PRI-2, EMERGENCY
-
.BORATION.,,. . * Within 1 HOUR calculate l
[: Shutdown. Marginp'er
~
!
*!. 1BwOS.1.1 1.1.e-1, !
, ca .w+,
.
-
-
.
SHUTDOWN MARGIN VERIFICATION DURING
- ' SHUTDOWN (1BwCSR 3.1.1.1) .
. . . j
. >
.e i
-
, l ,
.-
? .
,...
> l l .
-
!
-
,
..<'
.
,.,
. .~~ .
.
.
s ,s.
I l' -
,
!
.
E , .
..
'
APPRQVbD I 'JUN 101998 . - SHA10 WOOD oN SITE RCVIEW _ Page 7 of b7 l
- . . .. . - . .
}}