ML012690520
| ML012690520 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 08/13/2001 |
| From: | Peal R Susquehanna |
| To: | Conte R NRC/RGN-I/DRS/OSB |
| Conte R | |
| References | |
| 50-387/01-301, 50-388/01-301 50-387/01-301, 50-388/01-301 | |
| Download: ML012690520 (112) | |
Text
2001 SSES NRC Written Examination -August 10, 2001 52.'
Prior to Written Examination Administration Prior to administration of the written examination to the candidates, a change to the answer key was made. The question is included in Enclosure A. Typographical errors were also identified and corrected but are not included in this report.
" Developers identified one answer key change and corrected the question and key; the NRC lead examiner was notified.
o Developers identified typographical errors and corrected them.
Administered Examination to the Candidates During the administration of the examination the proctors made minor clarifications to nine questions in response to questions asked by the candidates. Refer to Enclosure B for the "Questions/Comments during Administration."
o During administration candidate questions were captured and nine clarifications were made to the examination. These clarifications were minor. Refer to questions R02/SRO2, R03/SRO3, R030/SRO27, R038, R041/SR033, R053/SRO42, R060, R079/SRO58, and R086/SRO64.
Post Examination Review with Candidates / Concurrent and Independent Review As a result of the post-examination review with the students and an independent and concurrent review by several instructors, changes were made to eight questions on the examination. Of these changes, the changes to five questions were considered significant and the changes to three questions were considered not significant. Refer to Enclosure C for "Post Examination Comments" and the resolution of these comments. The concurrent and independent review by instructors identified question comments similar to those identified by the candidates and also identified some questions comments not identified by the candidates.
o Three questions had changes that were identified as not significant because although the questions had the wrong answer this was identified and reflected on the answer key before grading the examination. This required that the correct answer be changed but did not affect the question and answers for each of these questions. Refer to questions R023/SRO21, R054/SRO43, and R050. The comments that resulted in these changes were identified by the candidates in two instances and were identified by the instructors for all three questions.
" One question had a change considered significant change because there was no correct answer and the question was deleted. Refer to question R092/SRO68. The comments that resulted in this question change were not identified during the review with the candidates but was identified by the instructors. It is not known if there is a knowledge deficiency in this area but this will be further evaluated to determine if there is one.
o Four questions had changes considered significant because there were two correct answers for each of these four questions. Refer to questions R095/SRO70, R033/SR029, R067, and R055/SR044. The comments that resulted in these changes were identified by both the candidates during the post-examination review and by the instructors during the concurrent and independent examination review.
2001 SSES NRC Written Examination - August 10, 2001 Examination Analysis Results:
The examination analysis confirmed the changes previously identified and showed no additional examination changes that warranted review. The examination analysis also revealed three topics where a knowledge deficiency may be present. RO 15/SRO 13 revealed a possible deficiency in limiting safety system settings and the bases for the settings for all candidates. R088/SR066 revealed a possible deficiency in required on-shift time to maintain an active license for the reactor operator candidates. RO66/SRO51 revealed a possible deficiency in RWM operation for the reactor operator candidates.
We/R T re*
6rIQRA4 ZNRA16CS bliRDe PAIRIAlf
?Rao R /#cp
I/F/"
-~-i.
' C l..g s k /..............
SSE$
NR n Wd.on..m.at.
0M n-S C.....Oflg dmnstato Applicant...
Level......
SRO Applicant.
Name Questio Type:
omo RO only 0
WSOnl 1 1'11'..1:..'.'.
e,,.,,.......
Question~~~~~~....
RO(ne.ubr f R nyetrNA SRO~~ ~~
(enter.
nu beifROolyene.NA Qusto (enter......
verbtim)
'tte (2~r
~
4..................4L
.k Response:.............
0....
.hhl adiioa guidance. Intutdteaplcn.od.isbs.ihifrmto rvdd id...
NO.ihol.diina.udnc.Poiedteflown.epos.etr
.ebtm...
.flN
~
C 4~
Change I clrificaions ade.toexamiation lPK ffile on... the.whie.boar aile toth.atetinofth.aplcat Proctor:...../..........
I:..
]~
Post-Examinatio Comet for initiate
- u.
eq 1
SSES 2001 REACTOR OPERA TOR NRC LICENSE EXAMINATION QUESTION 2.
Unit I is operating at 35% power when a loss of turbine lubricating 61ip curs. Which one of the following RPV pressure responses would occur over the first five (5) minutes of this event? Assume all system operate as designed.
RPV pressure will...
- a.
be controlled at approximately 955 psig.
- b.
be initially controlled at 1005 psig then lower to 915 psig.
- c.
rise to 1116 psig then lower and be maintained at 1106 psig.
- d.
initially rise to 1106 psig then cycle between 1070 psig and 1106 psig.
Page 2 of 100
I
..............0....
Applicant Name:
Question Type:
vtommon 0 RO only 0 SRO only I..
Question M RO (enter number, if SRO only enter N/A)
SRO (enter number, if RO only enter N/A)
Question (enter verbatim):
r 7 Respoyse:
0 Vu06held additional guidance. Instructed the applicant to do his best with information provided.
FW'tid NOT withhold additional guidance. Provided the following response (enter verbatim).
AAxccAL 9Cc~a'~
(,At~J 6\\so V
z-h j A
£rcs~rc~~
Changes I clarifications made to examination:
~4fled on the white board fraled to the attention of the applicants Proctor:
El Post-Examination Comments form initiated E4~j, 4 1 i13czaeie
" a I")t*.
SSES 2001 REACTOR OPERATOR NRC LICENSE EXAMINATION QUESTION 3.
The Unit I is operating at 100% power when a failure of the in-service HC Pressure Regulator causes the controller output to lower to zero. Which one of the following will occur?
- a.
The MSIVs will isolate when reactor pressure lowers to 860 psig.
- b.
'The reactor will scram on either high APRM power or high RPV pressure.
- c.
ThllI1epFessTre W1 rise about 4 psig and be controlled by the standby regulator.
- d.
Tjhe w*l ower about 4 psig and be controlled by the standby regulator.
Page 3 of 100
...... x",
Applicant Name.
- 1.
L estion Type:
.I--,........................
_M K I Question #:
RO so (enter number, if SRO only enter N/A)
'1r0144 SRO (enter number, if RO onlv enter N/A)(~/i
)
.~
$ ~
.2 Z...
V h !6P.......e..(......?
Response
E*
th held additional guidance. Instructed the applicant to do his best with information provided.
P6id NOT withhold additional guidance. Provided the following response (enter verbatim).
- 6,*.*s
- M*.-*",*
C*,k.r Changes I clarifications made to examination:
fd tified on the white board FCalled to the attention of the applicants Proctor:
Proctor:
El Post-Examination Comments form initiated S am' P-n"S Sianature Date
SSES 2001 REACTOR OPERATOR NRC LICENSE EXAMINATION QUESTION 30.
An alarm condition exits on the Uiiit I Reactor Building Stack Monitor oni 0C630. You are directed to go to panel IC600 and determine the source of the high radiation condition. Which onc,fthe -fof;ow--
Pr-ýý e-Z***4,- ai, --,=tt, 47(*
=*,",
- n4V
- t l
- a.
Zone,ailroad Access Shaft.
- b.
Main Steam Line Radiation Monitors.
- c.
Radwaste Building Ventilation System.
- d.
Standby Gas Treatment Ventilation Exhaust.
Page 30 of 100
I I Applicant Level:
L] RU U bSW 77 Applicant Name:
Question Type:
0(
es..
n......
.......R Question #:
RC SF Question (enter verbatim):
24
"?A",
U
-X -X,
Response
0 Wjthheld additional guidance. Instructed the applicant to do his best with information provided.
POW NOT withhold additional guidance. Provided the following response (enter verbatim).
a' o
X..
a Changes / clarifications made to examination:
ified on the white board EiCalled to the attention of the applicants P ro c to r:................
..-1
.1.-1
- ý................
Proctor:
U Post-Examination Comments form initiated
&SYIUAJPYN*.,,..c
SSES 2001 REACTOR OPERATOR NRC LICENSE EXAMINATION QUESTION 38.
12c_ O An ATWS has occurred on Unit 2 and reactor power is approximately 19%. The operator attempts to fully insert control rod 32-27 with the CONTINUOUS-IN switch and the rod fails to move. The operator then notes the following conditions:
Drive Water D/P 240 "ig.'.
Reactor Mode Switch Shutdown CRD Flow 120 gpm CRD Flow Control Valves Closed CRD pump 2A and 2B Running RWM Normal Control Rod 32-27 will not move because the...
- a.
RWM is enforcing an insert block.
- b.
Reactor Mode Switch is enforcing a rod block.
- c.
Drive Water DIP is much lower when rod movement is attempted.
d, CRD flow control valves are closed shutting off CRD flow to the HCUs.
Page 38 of 100
FApplicant Level:
1 U KU K+/-!KU
- ,~.
Applicant Name:
X" I...
Question Type:
oR"Common 0 RO only 0 SRO only x:....
Question M RO (enter number, if SRO only enter N/A)
'-S (enter number, if RO only enter N/A)
X...................
Question (enter verbatim):
4 3 4 {
- j 4
/J.&
y r rt &
t 624---
~
/~4zrŽ~~e 6 ~c~nrc~s-Vt
.2 M
Response
o VYhheld additional guidance. Instructed the applicant to do his best with information provided.
00id NOT withhold additional guidance. Provided the following response (enter verbatim).
X,
E Changes I c l 1ifications made to examination:
9d0rtified on the white board
- i~alled to the attention of the applicants I
I
' ' ' ' W,,
I I
, 4
.-.X.:.:.:.XWX.:.:.:
X X XX Proctor:
U Post-Examination Comments form initiated
_jntný'
SSES 2001 REACTOR OPERATOR NRC LICENSE EXAMINATION QUESTION 41.
Unit 1 is operating at 100% power with the following conditions:
SO-1 49-002, Quarterly RHR System Flow Verification, in progress.
"A" RHR Pump discharging 12,200 gpm through HV-151-F024A, Test Return Valve.
During the test a valid LPCI initiation signal and loss of 250 VDC Power occurs. %ich one of the following is the response of the RHR system when RPV pressure lowers to,..
psig.
[iJ F015A, LPCI F024A, Test Return I F007A, Min Flow 1 "A" RqRIgmp I
Page 41 of 100
-e,-
v- ý e"
- 1":'
U KU Applicant Name:
X........
-X.'.
I,'.
Question Type:
1;?dommon 0 RO only 0 SRO only Question M RO (enter number, if SRO only enter N/A)
SRO Iq (enter number, if RO only enter N/A)
Question (enter verbatim):
I........
Re*ionse:
l held additional guidance. Instructed the applicant to do his best with information provided.
EP-id NOT withhold additional guidance. Provided the following response (enter verbatim).
_Vs-c6 r-1"vWL -
Qcz' sCUJrs I...........................................................
Changes / clarifications made to examination:
ntified on the white board aled to the attention of the applicants Proctor:
Li Post-Examination Comments form initiated NL IIr
SSES 2001 REACTOR OPERATOR NRC LICENSE EXAMINA TION QUESTION 53.
Unit 1 was at 100% power when a LOCA occurred. The following events take place at the indicated times after the LOCA:
Time = 2 seconds, High Drywell Pressure setpoint reached, ECCS pumps started and operate on minimum flow.
Time = 20 seconds, RPV water level lowers to -129 inches.
Time = 48 seconds, RPV water level recovers to -90 inches.
Time = 60 seconds, RPV water level lowers to -129 inches.
Which one of the following is the time remaining before ADS initiates?
- a.
42 seconds
- b.
44 seconds
- c.
82 seconds
- d.
102 seconds Page 53 of 100
ApDlicant Level:
I rK U bKU Applicant Name:
Question Type:
U Common IAO only 0 SRO only I~~~~~~~~~
~~~
Qusto N.
etrnmei R
nyetrNA QustonMRo e-O (enter number, if SRO only enter N/A) i::
~~
~
S R (e nteri n um b e r ifi R O~::~
o nly::::::
e nter: N /A):i:i:~:i:
- i::i::
- i:i::i:::::::::::::*:iii::iilii iiiiiiiiii iiilliliii li li iiiii iii,'
'iii! *i:*
f:ii,'i:i,'
'i'i:!i*
- ' 'i*':i:i:ililii
,i'ii,:i,'i,:i,:i:i,
,'i'ii,
- ii'i'ii,:i,'
Response
D thh eld additional guidance. Instructed the applicant to do his best with information provided.
QuidNOT withhold additional guidance. Provided the following response (enter verbatim).
pm 4 9~1
~YeS QAp) c4e
& c s c Changes / clarifications made to examination:
'4ified on the white board Ifrhald to the attention of the applicants Proctor:
0a Post-Examination Comments form initiated Signature Date
SSES 2001 REACTOR OPERATOR NRC LICENSE EXAMINATION QUESTION 60.
-V
(*) i* Lj The following conditions occurred on Unit 1 following a reactor scram from 95% power:
RPV water level lowered to zero (0) inches before rising.
Reactor Water Level Control System remains in AUTOMATIC.
NO operator actions have been taken for the Reactor Water Level Control System.
With RPV water level at +5 inche§ the PCO depressespndsdhe SETPOINT SETDOWN reset pushbutton (HS-C32-1S08)JDbs-,
S{-<ený and then releases it Which one of the following describes where RPV water level will stabilize and why?
- a.
+5 inches because it can't be reset.
- b.
+13 inches because level hasn't recovered.
- c.
+18 inches because it can't be reset until level is)+.*einches.
- d.
+35 inches because level will reset and return to the normal setpoint.
Page 60 of 100
AppliýiW Lýevel:
[] SRO SApplicant Nam.
Question Type:
Common
[] RO only
[] SRO ol Question #:
RO (enter number, if SRO only enter N/A SRO (enter number, if RO only enter N/A)
Question (enter verbatim
Response
Ufthheld additional guidance. Instructed the applicant to do his best with information provided.
idNOT withhold additional guidance. Provided the following response (enter verbatim).
Changes I clarifications made to examination:
IY1tifedon the white board Iv~lle tothe attention of the applicant Proctor:
O] Post-Examination Comments form initiated EZJ JIL'C
SSES 2001 REACTOR OPERATOR NRC LICENSE EXAMINATION QUESTION 79.
For Unit 1 which one of the following are the OFFGAS PRE-TREATMENT and OFFGAS POST-TREATMENT PROCESS RADIATION MONITOR sample point locations?
(1) The Pre-Treatment.*ontrtakes a sample from...
(2) The Post-Treatment.nor takes a sample from...
A 50,,, ý'"._s
- a.
(1) The inlet to the Steam Jet Air Ejectors.
(2) The inlet to the Offgas HEPA Filter.
- b.
(1) The inlet to the Offgas Recombiners.
(2) The outlet from the Charcoal Adsorbers.
- c.
(1) The outlet from the Offgas Recombiners.
(2) The outlet from the Offgas HEPA Filter.
- d.
(1) The outlet from the Steam Jet Air Ejectors.
(2) The inlet to the Charcoal Adsorbers.
Page 79 of 100
I Applicant Name:
Question Type:
RO 82ommon
(
u RO only t
SRO only Q~ stion i M.......
... Ro.....
.. (enter numb~er, if SROo nly enter N/A).................
SRO I
(enter number, if RO only enter N/A)
Question (enter verbatim:
f
Response
II WVVh eld additional guidance. Instructed the applicant to do his best with information provided.
4lJid NOT withhold additional guidance. Provided the following response (enter verbatim).
Changes I clarifications made to examination:
d l~$ffied on the white board p-Called to the attention of the applicants Proctor:
UI Post-Examination Comments form initiated
-qinn~qt,,rp nntp
SSES 2001 REACTOR OPERATOR NRC LICENSE EXAMINATION QUESTION 86.
Both units are operating at 100% power when a Unit 1 Railroad Access Shaft Radiation Monitor reaches 6 mr/hr. Which one of the following is the Reactor Building HVAC response?
1?es.-% (tC t-A N*
X.
70ne
-am A4 rErF I Zone 3 HVAC Fans SBGT CREOASS
- a.
STOPANDISOLATE NO CHANGE STARTS NO CHANGE
- b.
RECIRC INITIATES STOP AND ISOLATE NO CHANGE STARTS
-Mum STOP AND ISOLATE NO CHANGE NO CHANGE NO CHANGE mum
- d.
RECIRC INITIATES STOP AND ISOLATE STARTS STARTS Page 86 of 100
I X.: : X.. X-4 IX Applicant Name:
Question Type:
[] Common U] RO only
- RO only Question #.,
R0O (enter number, if SRO only enter N/A)(
- R*
/
J*
cy j
SRO (enter number, if RO only enter N/A)
Question (enter verbatim):***)
9'
- J 5
Wiheld additional guidance. Instructed the applicant to do his best with information provided.
U] Did NOT withhold additional guidance. Provided the following response (enter verbatim).
Changes / clarifications made to examination:
U] Identified on the white board U] Called to the attention of the applicants,*----
Proctor:
I U Post-Examination Comments form initiated Sinn Jt ime1 L ll c"ite
~/1 (~J~~4 Rinnnfijrp nitp.
WR / -ri-,Ek ECY#Q A4?
):ýAe 14iry ?oar 4ckR*q 40oMMONJr$'
- 0 Pos g
ZAM-/1 IR I TT2EA ii i i i ii ii.i i i..!ii i ii.iii iiii!ii i !i ii..i
!iiiii.
.ii i i i i i ii i i i i i i i ii i i i i i i i ii i i i i i i i !
iiiiii
- i***ii!iii~*i~ii**
ii~iii~ii*i~iiiiiiiiiiiii~iii~iiiii*
- ~
~ii~i~iiiiiiiiii~iiiiii~~ii~i~!iii!!ii!*i~!i~i!!i!i*!i!
ii**i!~iiiiiiii~i~iiiiiii~
iii~ ~i i
- i I~ ~
~~...
AppicntLevl:
...O..SO i~i~~iii~ii~iiiiiiiiiiiii~i*i---------
iiii*i*~iii*i*iiiii*~iiii~iiii*~iii~ii~i~**iiii**i~ii*ii~~iiiiii**~iiii*ii~~ii**ii*iii~iiii~ii~~ii*iii
! Quesion Tye:
- om on
[]...only....ROonly W:*i:i:iAi::
io
.u r
.R.
.ny e t r N A O
f R..o nl...e... r.....
- i i*:*:*:i*:i**:::
i::i:*:*:
!:*:*i::::i*!::
ii~*:*::::::;:::*!:*ii*:!**:::::::*:!*:!
- i*:*:*:*:*ii*:*ii~*:
- !i*:!:;~i::
i! !* i~ ~i* i~;* ~i*;~i;;ii;
!* i;*~i;*ii;.....................::::::::::::::::::.:::.:::.::.:::.::.:::.:::.::..::::.
t..............
......................................................................................................................................A.
- t.
.nw r e
- e.
.s.A
.................B............i:
Reference:
e eyrfee c b lw
- !*;ii*i;**i;;;;*i**i;;*i~;;!;**i*
- ~
i*;**i*;****i**iiiii;;;iii~iii~~iiiiii;iii;*;;;iiiii~i;~~ii~ii~
~iiiii~*;*i*i;;i;*!i*;;;;!i;..............
i;i*i;*iii;i~ii**!**;i;;;;;;i;;;!*;ii;i~!**!;;;;;;!;***
~i;iiii~ii*;~i;;iii*;;i;*;;!i~ii*iii;;
Com ent
.o m n w.
Recommendation:~~.
o icu sth m terwihth N C Che b
e o r ce d n t th e
%g r.......a-d%.......
- . -c p tw
-orc an w r.
] Deet he
....... u n..
ak c.aifcaton to the xuestion.
l ii~i~iiii iiiiiiiiiiiiiiiilii!l !ii~i!iiiiiii! il iiil iiiiiii~iiiiiiiiiiiii~iii iiiii~iiiiii! iiiiiiiiiiiiiiiiii iiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiii~i ~iiiiiii~ ~iii iiiiiiiiii iiiiiiiiiiiii iiiiii i~
iii iiii i iii ii to mn a i n ov d
)
- U %
- J**?J J *
- i i ~ * : *:*::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::
- iii~~..%-.-.%
X.,X~
- ?i~
- ~~i*ii~~iii~~i?**
~iiii~i ii~~*
~i!*~i!**~ii~iiii~iii*i iiiiii iiii=!l e.................s...
to
- up r
/ clrii ato
- ad to ex miato: -
.iii~
~iiiiiiiiii iiiiiiiiii~iiii~iii~i~~iiii
.iiii
.iiii~iiiiiiii~ii~
iiii
.ii~iiii i**!i*
- iiiiiiii~ ~i~
i~ ~ii~
iiiiil*i iii~
iiiii ii!!*ii~ii~iiiiiiiiii i~iiiil iii~iiiiiiiliiiiiiii
?iiii~iiii~ii!ii~il ii~iiii X
- t.
s c o m m e n t:
Ch ng mad in IN the maste examination...p 2
OPERATIONS QUESTION AND ANSWER INPUT FORM 13 RO 15 SRO,.
(A)
SY017 L-5 Course (B) 20 Objective (C) Question Type (check one)
W Multiple Choice Matching Li Free Format (Essay)
(D) Bank Operations X
OP002 (E) 1 2
3 4
5 6
7 8
Keywords Ca:tegory To*c 1 Topc 2 JTA Setting Other Objs Quiz Only I
Retired L<9 characters)
EIAPE Loss of Vacuum (F) Point Value:
I (G)Answer Time: ----
]
(Minutes)
(H) Cognitive Level:
X I
(Check one) 2 3
4 5
Memory Comprehension Application Analysis Problem Solving (I) Review Date (YYMM):
(J) QUESTION:
Which one of the following is the bases for a reactor scram on a main turbine trip above 30% reactor power?
- a.
Provides a backup to the RPV pressure and APRM high scrams.
- b.
Ensures RPV water level remains above the dryer separator skirt.
- c.
Protects the reactor from the pressure effects of a loss of heat sink.
- d.
Anticipates a positive reactivity addition from a loss of feedwater heating.
(K) ANSWER:
c.
6"r_
Fonn STCP-QA-3251)
Rev 3. (8/95)
Page I of I File No. R I -2
OPERATIONS QUESTION AND ANSWER INPUT FORM (L) REQUIRED MATERIALS:
None (M) K&A NUMBER/RATING: 295002, AKI.03/ 3.6 (N) NOTES:
.:'~5......
~
JUSTIFICATION:
Scramming the reactor above 30% power prevents pressure transients caused by loss of the main condenser as the heat sink because turbine bypass capacity is limited. (Turbine trips on low vacuum, 21.7" Hg Vac, before MSIVs close or any other auto actions).
DISTRACTER A:
These scrams backup the turbine trip scram.
DISTRACTER B:
This is part of the bases for the low level scram.
nTQTT? A CT1'P.R r.
This is a concern at all powers and is not the bases for this scram EXAM OUTLINE CROSS-REF:
IVA TEXT:
ILEVEL:
LEVEL:
TIER:
1 1
GROUP:
1 2
AK1.U3 - Knowledqe ot me opera ional implicati5uio 01 1uIIUWInc they apply to LOSS OF CONDENSER VACUUM: loss of heat sink' (0)
REFERENCES:
Technical Specifications Bases B 3.3.1.1.8 (P) POSITIONS:
R-RO S-SRO A-ASO N -NPO T -STA (check one or more boxes)
X IXI I
I (Q) Prepared by ED BOWLES (R) Reviewed by:
W
- cc. ( ]-ýý Form STCP-QA-325 1)
Rev. 3, (8/95)
Page Iof I File No. RI 1-2
IApplicant Level:
RI (K U 6KU Compplicant Nam e:
[ Question Type:
'\\,
common
[] RO only
[] SRO only SQuestion M:
RO 2.,3 (enter number, if SRO only enter N/A)
SRO (enter number, if RO only enter N/A)
[ Answer: (crcle the answer key response)
A B
D ze
. A..
l...
i
.....I...
i......
1 Recommendation: (The grader is encouraged to discuss the matter with the NRC Chief Exppfner before proceeding with the grading)
VChange the correct answer.
L Do NOT change the correct answer.
W A ept two correct answers.
El Delete the question ake clarifications to the question.
Changes / clarifications made to examination: (provide a description)
{t.................
C c ~
.V.A z............
..C
~
Pharoectr:
ctions made ionI examination:ation vide escriptionp Sigh nture Da t Aý (
&Lp 6
ameý J
S ig natu re...te................
Ccx)/
y~v) LP--y
OPERATIONS QUESTION AND ANSWER INPUT FORM R0 23 SRO~Aff (A)
SY017 M-1 Course (B)
Objective (C) Question Type (check one)
~M ultiple Choice Matching Free Format (Essay)
(D) Bank Operations OP002 K5y (E) 1 2
3 words:
Cateeo r Topic 1 Topic 2 characters)
OPS ONXXXXXX (F) Point Value:
j (G)Answer Time:
(Minutes)
(I) Review Date (YYMM):
4 5
6 7
8 JTA Setting OtherObjs.
Quiz Only Retired (H) Cognitive Level:
I (Check one) 2 X3 4
5 Memory Comprehension Application Analysis Problem Solving (J) QUESTION:
Unit 2 is shutdown with the following conditions:
"2A" RBCCW and "2A" TBCCW are aligned to ESW.-
Loop "A" of ESW is isolated from the Diesel Generators (DGs).,/
- 4. loss of off-site power occurs DG output breaker 1A20404 fails to close.
"B" ESW Pump fails to start Assuming NO operator actions, which one of the following is required?
- a.
Trip ALL the DGs in four and one half (4.5) minutes.
- b.
Trip DG' "B" and DG "D" in four and one half (4.5) minutes
- c.
Trip DGs "A", "B" and "C" in four and one half (4.5) minutes and DG "D" in eight (8) minutes.
- d.
Trip DGs "A" and "C" in four and one half (4.5) minutes and DGs "B" and "D" in eight (8) minutes.
1P Is (K) ANSWER:
C-.
Foenn s'cP-QA-325D Rev. 3, (8/95)
Page I of l I r
ýp jO AJ r.ý -Jý-
File No. R 11-2 JIY
OPERATIONS QUESTION AND ANSWER INPUT FORM (L) REQUIRED MATERIALS:
None (M) K&A NUMBER/RATING: 295018, AK3.01/3.5 If (N) NOTES:
JUSTIFICATION:
On a loss of cooling water the diesels jl'tt be tripped in 4.5 minutes if loaded and 8 minutes if unloaded. DG D does load (its output breaker does not close. So the all the DGs are without cooling water, but D is unloaded. So A,B,C are tripped in 4.5 min. and D must be tripped in 8 min.
DISTRACTERA:5 DG Diaingunloaded.
0-s \\p Q-D Lv DISTRACTER
- r anrj etripped and D may run 8 minutes.,
DISTRACTER 1V.-V B must be.g~5jý1minutes.
K/A TEXT:
they apply to P.ARTIL OR. CEOMPLETE LOSS OF CCW: Effects on component/system operation.
SOUCE MODIFINLE ED:R R
NRS-E:
TEW:X 1CFR55:
COMMENTS:
(0)
REFERENCES:
EO-1 00-030, Caution on pages 2 (P) POSITIONS:
R-RQ S-SRO A-ASO N-NPO T -STA (check one or more boxes)
X I
X I
I I
I Ar @i..s (Q) Prepared by ED BOWLES (R) Reviewed by:
Form STCP-QA-325 I)
Rev. 3, (8/95)
Page t of I File No. RI 1-2
PROCEDURE COVER SHEET NUCLEAR DEPARTMENT PROCEDURE UNIT 1 RESPONSE TO STATION BLACKOUT EO-100-030 Revision 17 Page 1 of 12 QUALITY CLASSIFICATION:
APPROVAL CLASSIFICATION:
(X)
QA Program
(
)
Non-QA Program (X)
Plant
)
Non-Plant
( )
Instruction EFFECTIVE DATE:
11/02/00 PERIODIC REVIEW FREQUENCY:
2 Year PERIODIC REVIEW DUE DATE:
12/30/02 RECOMMENDED REVIEWS:
Training Procedure Owner David M. Kapuschinsky Responsible Supervisor Thomas R. Markowski Responsible FUM:
Manager-Nuclear Operations Responsible Approver Manager-Nuclear Operations FORM NDAP-QA-0002-1, Rev. 2, Page 1 of 1
EO-100-030 Revision 17 Page 2 of 12 I
SYMPTOMS AND OBSERVATIONS This procedure is entered concurrently with other Emergency Operating Procedures whenever following conditions are met:
1.1 All offsite power supplying Unit 1 Auxiliary Busses and Unit 1 ESS Busses is lost, AND 1.2 All four Unit 1 ESS Buses remain de-energized.
- 2.
OPERATOR ACTIONS CONFIRM 2.1 PERFORM following:
2.1.1 CLASSIFY SBO event along with other existing plant conditions in accordance with EP-PS-1 00, Emergency Director (ED) - Control Room.
2.1.2 REFER to Attachment B to establish ES priorities.
2.1.3 OPERATE HPCI in accordance with EO-100-032.
2.1.4 OPERATE RCIC in accordance with EO-100-033.
2.2 MONITOR plant parameters on available instrumentation as identified on Attachment A.
CAUTION DIESEL GENERATOR FAILURE IS IMMINENT IF OPERATED WITHOUT ESW COOLING LONGER THAN 4.5 MINUTES LOADED OR 8 MINUTES UNLOADED.
2.3 MANUALLY ATTEMPT to start all Diesel Generators OG501A(B)(C)(D), as follows:
2.3.1 At 0C653, PLACE D/G A(B)(C)(D) GOV MODE SEL 2.3.2 HS-00055A(B)(C)(D) in ISOCH.
To start diesel at 0C653, DEPRESS DG A(B)(C)(D) START HS-00051A(B)(C)(D) push button.
Applicant Level:R Applicant Name:
Question Type:
IPommon
[] RO only LI SROol Question #:
RO (enter number, if SR onyenter N/A)
SRO I*~ (enter number, if RO only enter N/A)
S
Reference:
(enter the answer key reference below)4 Comment: (enter the comment below)
Recommendation: (The grader is encouraged to discuss the matter with the NRC Chief Examiner before proceeding with the grading)
LIC~
ge the correct answer.
IZ Do NOT change the correct answer.
V~cpt two correct answers.
LI Delete the question
~JAaeclarifications to the question.
Changes I clarifications made to exam.rination: (provide a description)
R eference.s) to suport c.hange
,lrifcation eo examination:
Justification for rejection of an applicant's comment:
iiili i iiiiiiiiiiiiii
!ii.*.*
! iili iiii ii~ii i!i ii**
~ i iii*:iiiiiiiiiii i~iiiiiiiiiiiiiiiiiiiiiii!~iiii*ii~...
iii~
~ *iiii~
i~iiiiiiii~ i!iiiiiiiiiiiiii~iiii~ ~iiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiii*iiiiiiiiiiiiiiiiiiiiii:iiiiiii Proctor:
Vthange made in INK on the master examination copy Signature Date 4 ~
u~i~
~Ac~c A~
0~
o
~
'y-S~fl
OPERATIONS QUESTION AND ANSWER INPUT FORM Ro 33SRO (A)
AD045 Course (B)
Objective (C) Question Type (check one)
W Multiple Choice Matching Free Format (Essay)
(D) Bank OperationsoP002 0P002 LJq I Keý LL,:9 (E) 1 2
3 words:
Category Topic 1 Topic 2 characters)
OPS I ONXXXXXX (F) Point Value:
1 (G)Answer Time:
(Minutes)
(I) Review Date (YYMM):
4 5
6 7
8 JTA Setting Other Obis.
Quiz Ony Retired (H) Cognitive Level:
X 1
(Check one) 2 3
4 5
Memory Comprehension Application Analysis Problem Solving (J) QUESTION:
Plant conditions are as follows:
Reactor has been in Cold Shutdown for 2 days following power operation.
Reactor water level is +87 inches.
Both reactor recirc pumps are tagged out of service.
Shutdown cooling has isolated and the shutdown cooling suction valves cannot be opened.
Which one of the following operator actions will reverse or prevent reactor vessel stratification AND provide alternate decay heat removal?
- a.
Place Reactor Water Cleanup in service in recirculation.
- b.
Insert a manual scram to maximize Control Rod Drive flow to the RPV.
- c.
Start a second Control Rod Drive pump and maximize cooling water D/P.
- d.
Begin rejecting water with Reactor Water Cleanup while injecting with CRD.
(K) ANSWER:
a.
,4cce{~~9 lcu ý,- A/-ý's W E6 ý c -
t 4Q.
-p~ c Pr
~J-00%o//
Form STCP-QA-325D-Rev. 3, (8/95)
Fi1e No. RI 1-2 Page I of I
OPERATIONS QUESTION AND ANSWER INPUT FORM (L) REQUIRED MATERIALS:
None (M) K&A NUMBER/RATING: 295021, AK2.02/3.2 (N) NOTES:
JUSTIFICATION:
By recirculating water from the bottom vessel drain and discharging into the feed system the RWCU system will help circulate water in the RPV and cool by discharging heat to the Non-Regen HX.
DISTRACTER B:
Although this will add cool water it will add to stratification.
DISTRACTER C:
Although this will add cool water it will add to stratification.
DISTRACTER D:
This will remove water and inject cooler water but will not provide circulation and is NOT a method of cooling specified in Attachment B of ON-149-001.
EXAM OUTLINE LEVEL:
TIER:
1 1
GROUP:
3 2
K/A TEXT:
AK.2 - Knowledge O inerreans eeen COOLING and the-followinq: Reactor water cleanup.
QUESTION BANK-......
SOURCE:
MODIFIED:
10CFR55:
- 4.
43(b).5 COMMENTS:
i~~*
- i i ii i i* ii i
- ii!i~ :i~:: : i'*i" :*: : :
- :--= ->: ::. --:>::--: "*::::* : :: :**::i:i!:* :*** :* :*::: ::2 :::: *:*:: :::: :::::::::: ::::: :5:: :. :::5 **::; :: ::::: : ::::.:...::..:...:::: :.: :::::::::::.....:.
(0)
REFERENCES:
ON-149-001, Sect. 3.4 and App. B (P) POSITIONS:
R-RO S-SRO A-ASO N NPO T -STA (check one or more boxes)
X IX I
I 1
(Q) Prepared by ED BOWLES (R) Reviewed by:
,er. cit Form STCP-QA-325D Rev. 3, (8/95)
Page 1 of I File No. R I 1-2
ON-149-001 Revision 17 Page 24 of 31 The most preferred methods for all reactor modes would be one or both of the following to maintain reactor coolant temperature at desired value:
Simple feed and bleed operation using Reactor Water Clean-up, RHR Shutdown Cooling in level control letdown operati6n and Fuel Pool Cooling and Cleanup, as available and as required to letdown; and using CRD, Condensate System, Condensate Transfer, Core Spray and RHR to makeup. It should be noted when selecting a makeup source that although condensate is preferred, suppression pool water may be used without causing severe excursions to reactor coolant.
Suppression Pool water is sampled monthly by Chemistry and high water quality is maintained.
Use of available heat exchangers in Reactor Water Cleanup System and Fuel Pool Cooling and Cleanup System, as available and as required.
If these methods fail to maintain desired temperature, other methods must be used.
Determination of method or methods employed is based on reactor decay heat load, plant status, availability of systems and nature of loss.
In Mode 3 with Primary and Secondary containment established, reactor pressure may be maintained greater than 20 psig (to clear column of water from SRV discharge downcomers if steam flow is routed or diverted there) and less than 98 psig (below reactor high pressure isolation). Steam may be routed to the main condenser or suppression pool and methods of makeup previously discussed may be used to maintain level. This method may also be used in Mode 4 with Primary and Secondary containments established. Reactor pressure and temperature may be allowed to rise until within pressure limits cooling by boiling as above. One factor to be considered in Mode 4 is that using this method will result in entering the Emergency Plan. The major factor that must be determined before using this method is that Primary and Secondary containments are established in order to prevent release of radioactive materials to the environment.
'N
1.
2.
3.
4.
5.
6.
(
7.
SYSTEM/EQUIPMENT AVAILABILITY DETERMINATION SYSTEMS STAI Primary Containment (Mode 3 or 4 only) avail Secondary Containment avail Flowpath from reactor to Condenser avail w/vacuum maintained by SJAE RPS Channel ALIA2 avail RPS Channel B1/B2 avail Methods to MIU to RX
- a.
CRID avail
- b.
Condensate avail
- c.
Condensate Transfer (1)
Keepfill avail (2)
SDC Flush avail (3)
- Skimmer Surge Tank avail
- d.
RHR avail
- e.
Core Spray avail Methods of Letdown from RX
- a.
- RWCOJ.... *-C (1)
Main Condenser avail (2)
Radwaste avail
- b.
RHR avail
- c.
SRVs to Supp Pool avail Attachment B ON-149-001 Revision 17 Page 29 of 31 US (Circle One) unavail unavail unavail CHECKED unavail unavail unavail unavail unavail unavail unavail unavail unavail unavail unavail unavail unavail Page 1 of 2
I SYSTEM/EQUIPMENT AVAILABILITY DETERMINATION 8.
9.
10.
SYSTEMS
- Fuel Pool Gates
- Cask Storage Pit Gates
- Method of Cooling
- a.
FPC and Cleanup
- b.
RWCU Recirculation
- c.
- d.
U-2 FPC and Cleanup
- e.
U-2 RHR in FPC Assist STATUS installed installed Attachment B ON-149-001 Revision 17 Page 30 of 31 (Circle One) not installed not installed unavail unavail unavail unavail unavail
.CHECKED avail avail avail avail avail Applicable in Mode 5 and level >22 feet above flange.
Page 2 of 2
Applicant Level:
Applicant Name:
SQuestion Type:
LI Common 0* only LI SRO only SQuestion #:
RO (enter number, if SRO only enter N/A)
SRO (enter number, if RO only enter N/A)
Answer: (circle the answer key response)
A S
Reference:
(enter the answer key reference below)
Comment: (enter the comment below)
Recommendation: (The grader is encouraged to discuss the matter with the NRC Chief Ex~iner before proceeding with the grading) ange the correct answer.
LI Do NOT change the correct answer.
Accept two correct answers.
%I Delete the question
-~-ake clarifications to the question.
Changes T clarifications made to examination: (provide a description)
[ Reference(s) to SUng~ort chanale I clarification made to examination:
C*-*Z 7*il
~II1 0 flit-I 3-:1 C-c-ooi
~~~~ei.
C-L(C-I Justification for rejection of an applicant's comment:
Proctor:
l*fhange made in INK on the master examination copy.
Signature..
OPERATIONS QUESTION AND ANSWER INPUT FORM RO 50 (A)
SY017 E-9 (B) 8 (337)
Course Objective (C
Question Type (check one)
~M ultiple Choice Matching Free Format (Essay)
(D) Bank Operations LIXZ OP002 [
(E) 1 2
3 4
5 6
7 8
Kewords "
Category Topic 1 Topic 2 JTA Setting Other Objs.
Quiz Ony Retired (L9 characters)
PCS PCSINST (F) Point Value:
I (G)Answer Time: [-
-]
(Minutes)
(I) Review Date (YYMM): [-
-]
(H) Cognitive Level:
1 (Check one)
X 23 4
5 Memory Comprehension Application Analysis Problem Solving (J) QUESTION:
Unit 1 is operating at 65% power while a surveillance test is being performed on the recirculation drive flow instruments.
During the surveillance the Mode Switch for the A Flow Unit is placed in zero (0) without first bypassing the flow unit.
Which one of the following will occur and what action is required?
- a.
Several control room annunciators alarm and a rod block occurs, NO half scrams occur, bypass the A Flow Unit.
- b.
Several control room annunciators alarm and a full scram occurs, enter ON-1 00-101, SCRAM and take the immediate actions.
- c.
Several control room annunciators alarm and a rod block and half scram occur, bypass the A Flow Unit and reset the half scram.
- d.
Control room annunciator APRM/RBM FLOW REFERENCE OFF NORMAL activates, NO rod block or trips occur, bypass the A Flow Unit.
(K) ANSWER:
- d.
CL.
C.<xvjuŽ~~
Q1'LA~tLb\\.
-ý-L 1" ý 000
ý 0c CI /1 614tca 00-A4-4 /Qý,
&4 lqýtr-OK 1')O"VdLý Form STCP-QA-325D Rev. 3, (8/95)
Page 1 of I Rý F, N'L Fite No. RI 1-2
OPERATIONS QUESTION AND ANSWER INPUT FORM (L) REQUIRED MATERIALS:
(M) K&A NUMBER/RATING: 216000 / A2.05 / 2.8 (N) NOTES:
0_'Aj
(:ýy O002Z" o &tecA'Tb~
z
~ 4,~f te4Ai
/7W A66/
JUSTIFICATION:
During the surveillance test the mo ch for the flow unit is placed in zero (0) to simulate a high flow con which causes the auctionering circuit to shift to the other flow instru hichh in this condition is un-affected. No blocks or trips occur same mode switch manipulation is an operator action on a flow unit u-f11re.)
DISTRACTER A:
_1r/
AUP A,,7l/a-)IG DISTRACTER B:
p o<o[)
t DITRCERC LA EXAM OUTLINE LEVEL:
RO......
C R O SS-R E F:
T IE R :
2 GROUP:DIFIED K/A TEXT:
A2.05 - Ability to (a) predict the impacts of the following on the NUCLEAR BOILER INSTRUMENTATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Surveillance testing.
.a Q U E T I O....................
ii:::i:
i:::
! :i:
- -:"*~o M S
.` #`
(0)
REFERENCES:
ON-164-001, Sect. 3.0 and 5.0 (P) POSITIONS:
R-RO S-SRO A-ASO N-NPO T -STA (check one or more boxes)
X IX I
I IXI (R) Reviewed by:
/
-. (A'Z (Q) Prepared by Phil Ballard Form STCP-QA-325D Rev. 3, (8/95)
Page I of 1 File No. RI 1-2
AR-1 03-001 Revision 21 Page 15 of 39 C05 APRM/RBM SETPOINT:
See Probable Cause FLOW/REFERENCE OFF NORMAL ORIGIN:
Relays 1K1 or 1K3 for (C05)
Flow Units A, B, C or D
- 1.
PROBABLE CAUSE:
1.1 Flow signal from Reactor Recirc Loops _> 1141125% sensed by Flow Units A, B, C, or D.
1.2 Differential flow > 10% between Flow Units A and C or C and D or A and B or B and D.
1.3 Inop condition in Flow Units A, B, C or D.
1.3.1 Mode switch not in operate.
1.3.2 Any internal module unplugged.
1.3.3 Upscale trip.
1.4 Faulty or failed instrument.
- 2.
OPERATOR ACTION:
2.1 If Flow Unit inoperative or out of calibration, ATTEMPT to bypass to remove Rod Block in accordance with ON-164-001 Recirc Drive Flow Instrument Failure.
2.2 If alarm caused by actual high Recirc loop flow, REDUCE flow.
2.3 If Flow Unit INOP, COMPLY with TS 3.3.1.1 and REVIEWTS Bases 3.3.1.1 section 2.b.
- 3.
AUTOMATIC ACTION:
Rod Block to Reactor Manual Control System.
- 4.
REFERENCE:
4.1 M 1 -C51-19(47) 4.2 E-323 SH 34 43 IOM 305 4.4 TS 3.3.1.1
ROD OUT BLOCK
( H03 )
AR-104-001 Revision 16 Page 38 of 41 H03 See Probable Cause Rod Drive Control Cabinet SETPOINT:
ORIGIN:
I 1.0 PROBABLE CAUSE:
1.1 Rod Block caused by any of following:
1.1.1 APRM Upscale - 0.58W + 50% 1.1.6 1.1.2 APRM Inop 1.1.7 1.1.3 Flow Unit Upscale-1.1.8 114/125 flow.
1.1.4 Flow Unit Inop 1.1.9 1.1.5 Flow Unit Comparitor Trip -10% between Flow 1.1.10 Units RBM Upscale <0.58W + 52%(1)
RBM Inop Scram Disch High Level
< 35.9 gallons Scram Disch Volume High Level Scram - Bypass RWM Rod Block 1.3.2 1.2 Rod Block cause by any of following in Run position only:
1.2.1 APRM Downscale -
5%
1.2.2 RBM Downscale -
5%
1.3 Rod Block caused by any of following in Startup or Refuel position only:
1.3.1 SRM Upscale -
2x105 cps 1.3.7 IRM Downscale-5/125% of SRM Downscale - 3 cps scale 1.3.3 SRM Inop 1.3.8 IRM Detector Position 1.3.4 SRM Detector Position Wrong Wrong 1.3.9 Service Platform Hoist 1.3.5 IRM Upscale-108/125% of loaded.
scale.
1.3.10 Refueling Platform over 1.3.6 IRM Inop core-Startup position.
1.4 Rod Block caused by any of following in Refuel position only:
1.4.1 Refuel Platform over core and Fuel Grapple not Full up.
1.4.2 Refuel Platform over core and Fuel Grapple loaded.
1.4.3 Refuel Platform over core and Frame Hoist loaded.
1.4.4 Refuel Platform over core and Trolley Hoist loaded.
1.4.5 No Rod selected.
1.4.6 One Control Rod withdrawn.
1.5 Rod Block caused by RSCS when in Run or Startup position.
2.0 OPERATOR ACTION:
2.1 DETERMINE cause of Rod Block observing appropriate annunciator.
2.2 REFER to appropriate Alarm Response procedure.
3.0 AUTOMATIC ACTION:
Stops control rod movement on receipt of annunciator.
4.0
REFERENCE:
4.1 Ml-C12-90(28) 4.2 MI-H12-778(1) 4.3 IOM 305 4.4 E-323 SH 35 (1) 4.5 TS Amendment 176
)
-4ANNEL AI/A2 AUTO SCRAM API DIV ¶ TRI P RPS MAN SCRAM CHANNEL Al/A2 SWTCIW-ARMED NEUTRON pON ClIAN A PRIMARY CONTAINMENT WIl PRESS TRIP Ry VESSEL HI PRESS TRIP i
Py VESSEL LO LEVEL TRIP MN STM LINE WI PAOIATION TRIP TUPe CV FAST CLOSURE TPIP PPS CI-IANNEL AVA2 MAN SCPR U
BACKUP/GROUP PILOT SCRAM SYSTEM A POWER FAILU*E MSrV CHAN A/C NOT FULLY OPEN TRIP TUP9 STOP,Iv CLOSURE TRIP PPS Cl-IAN AL'A2 SCRAM OSCI-VOL WI-I WTP LEVEL TRIP SCRAM DISClIARpE VOLUME NOT DRAINED PIOSY TPOUBLE I
I 'iI SYSTEM TRIP I IIImImiiii--
I II SPOS PDC TROB8LE TURB CV FAST CLS & STOP VLV TRIP BYPASS APRm CHAN AIC E TRIP t peM JPSCALE OR INCO ROD BLOCK PBM DXv0NSCALE PECIPC LOOP A PP"T eKP A2051 DC TPIP RECIPC LOOP A RPT BKP 1A2&__02 DC TRIP PPT SYS LOOP A TRIP L
I r APPM,/RBM FLC.W PEFERENCE OFF NORMAL
~F*
I I
APPM CW-IAN,DF UPSCALE DP* INOP TRIP UPSCALE O O
APPM UOWNPSCAL E
U F
LPPM UPSCALE L PPM DOWNS CALE I
CONTROL ROD DRIVE I
RPT SYS LOOP A OUT OF SEpVRCa POD POSITION INDICATIONt SYSTEM NOR i Iri PDCS INOR POD BLOCK I!
I CRD PANEL 1C 007 HI TEMP !I 4 ROD DISPLAY(
INOR CRD ACCUMULATOR TROUBLE
A Cý DJ Ej Fl HI 2
2 5 m
TCD651C)
I =u 7-4z
I SApplicant Name:
SQ~u~est~ion Type:
L?-tommon El RO' only
[] SRO only i Xff-'iiiiiii iii~
-iiiiiiiiiiiiii
- iii iiiiiiii*iiiiii'i'i ii
- i*i~ ~ii'*ii~l~i* ii~ i**~iii~iiiiii
- iiiiiiiii',~ i*iiiiii!'iiii',i',i',iiiiiiiii
- n.
e.......
Reference:
.ete.h anwr.e.rfrnc.e.w Qumesint#:
Rete th5eZ comnent nubelS Oonywne NA ExaROe (efore proceedin ift the onlyenteng/A
[] Chage th corret anser.
I~o..........the.crrectanswe.
e p r
Answer:ficaionfrcl rejectinse key resaplicnse)
Aomment:D Proctor:
El Change made in INK on the master examination copy P ' tNp Signature Date
OPERATIONS QUESTION AND ANSWER INPUT FORM 14I RO 52 SRO (A)
SY017 C-5 Course (B)
(B)
Objective (C) Question Type (check one)
~
Multiple Choice lMatching Free Format (Essay)
(D) Bank Operations W OP002 I
(E) 1 2
3 4
5 6
7 8
FKe owfrds TopCaeor To ic I mo_ Cc2 JTA Settin2 Other Objs" Quiz Onty I
Retired I(L9 chatacters)
Systems ROIC (F) Point Value:
I (G)Answer Time:
(Minutes)
(H) Cognitive Level:
1 (Check one)
X 2
3 4
5 Memory Comprehension Application Analysis Problem Solving (I) Review Date (YYMM):
(J) QUESTION:
Unit 1 was at 30% power when a reactor scram occurred on a loss of vacuum after circulating water was lost. After the initial scram actions were taken the following occurred:
Reactor Core Isolation Cooling (RCIC) was placed in pressure control mode per OP-1 50-001.
Workers in the Reactor Building bump Instrument Rack 1C004 causing a Division 1 Low RPV Level Trip (-30 inches).
%Nhich one of the following is the effect on RCIC and the reasons for that effect?
- a.
No effects because RCIC will NOT realign after being manually placed in this line-up.
- b.
No effects, RCIC remains in pressure control mode, because only one division is effected.
- c.
RCIC automatically aligns for RPV injection because only one division is required for system initiation.
- d.
RCIC automatically aligns for RPV injection after RPV level lowers to actuate the Division 2 RPV Level Trip.
(K) ANSWER:
c.
Form STCP-QA-325 D Rev. 3. (8/95)
Page 1 of r File No. R I 1-2
OPERATIONS QUESTION AND ANSWER INPUT FORM (L) REQUIRED MATERIALS:
None (M) K&A NUMBER/RATING: 217000, K1.02/3.5 (N) NOTES:
JUSTIFICATION:
Tripping one division of RCIC initiation will cause RCIC to automatically shift from pressure control mode to injection mode.
DISTRACTER A:
RCIC will inject.
DISTRACTER B:
RCIC will inject.
DISTRACTER D:
RCIC does NOT have to wait for the second initiation signal it will inject with only one division actuated.
one.................
d iv..
isi act ate...................
EXAM OUTLINE LEVEL:
TIER:
2 2
GROUP:
1 1
K.TEXT:
now ed qe of the pysical connections and/or cause-ete relationships KIA TEXT:
between REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM and the following: Nuclearboilersystem.
QUESTIOI N BANK:
SOURCE:
MODIFIED:
X NEW:
- i~~i~~i* !
- i!ii****i i.!i*.* ;.
~1!i i~**i*......
i ii i* i.....
- ...: :i*ii!iii~i~
- 10 CFR55 5:ý0.
COMMENTS:
a i (O)
REFERENCES:
ON-150-001, Sect. 3.2 and SY017, C-5 (P) POSITIONS:
R-RO S-SRO A-ASO N-NPO T -STA (check one or more boxes)
X I
X (R) Reviewed by:
/s c
(Q) Prepared by ED BOWLES Form STCP-QA-325D Rev. 3. (8/95)
Page I of I File No. RI 1-2
1*(~
1 Applcant Level:
ý6 CiKu I
Applicant.Name:..
I
Reference:
(enter the answer key reference below)
I ii
.i iii
.i
.iiiiii.i.iiiiiiiiiii
... iiiii iiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiii~iiiiiiii~~i~ ~~~i iiii~iiiiiii~iiiiiiiiiiiiiiiiiiiiiiiiiiiii iii~*i*iiiiiiiiiiiiiiiiii~i~iiii ii~ ~
~iiiiiiiiiiiiiii~i~ ~~~;
Comment: (enter the comment below)X 1
Q.:
C L....
.. :::;:.: :.: : :: ::.=
= :.: : : : : : : :: ::a e:
Recommendation: (The grader is encouraged to discuss the matter with the NRC Chief EXyrhiner before proceeding with the grading)
PChange the correct answer.
0 Do NOT change the correct answer.
P Apt two correct answers.
0 Delete the question ake clarifications to the question.
Changes / clarifications made to examination: (provide a description)
SReference(s) to support change I clarification made to examination:
F 6Ž I
I Justification for rejection of an applicant's comment:
Proctor:
Vi-hange made in INK on the master examination copy Signature Date Y.\\\\
\\
1.
OPERATIONS QUESTION AND ANSWER INPUT FORM 43 RO 54 SRO 9 (A)
SYO17 E-6 Course (B)
Objective (C) Question Type (check one)
X-ý Multiple Choice Matching Li Free Format (Essay)
(D) Bank Operations MX OP002 (E) 1 2
3 4
5 6
7 8
K ~ords:
Categor Topic 1 Topic 2 JTA Setting Other CObs.
Quiz Only Retired IL<9 characters)
PCS ATM OSCTLI (F) Point Value:
] (G)Answer Time:
(Minutes)
(H) Cognitive Level: [-X 1
Memory (Check one)L 2
Comprehensic 3
Application 4
Analysis 5
Problem Solvi on ing (I) Review Date (YYMM):
(J) QUESTION:
With Unit 1 at 85% power when a load reject and loss of off-site power occur. The diesel generators start and power their associated buses.
Which one of the following describes the effect on Drywell Cooling?
The operating drywell unit coolers trip, then...
- a.
remain shutdown until manually started.
- b.
restart when the diesels start and are cooled by RBCW.
- c.
restart when the diesels start and are cooled by RBCCW.
- d.
restart when the diesels start and run without cooling water.
(K) ANSWER:y.
C-
-ý) CL'-
c~~~~
K D~r E,
kx (/~
Form STCP-QA-325D Rev. 3, (8/95)
Page I of I File No. RI 1-2
OPERATIONS QUESTION AND ANSWER INPUT FORM (L) REQUIRED MATERIALS:
None (M) K&A NUMBER/RATING: 223001 / K2.09 / 2.7 / 2.9 (N) NOTES:
JUSTIFICATION:
The power supplies to the drywell unit coolers trip but the fans restart and after cooling water is verified available cooling can be restored to the drywell.
DISTRACTER A:
The fans trip and restart.
(0)
REFERENCES:
ON-104-001 (P) POSITIONS:
R-RO S-SRO A-ASO N-NPO T -STA (check one or more boxes)
X I X (Q) Preparedby Phil Ballard (R) Reviewed by:
P, t.&-,
Form STCP-QA-3251)
Rev. 3, (8/95)
Page 1 of I File No. Ri1I-2
+
UNIT 1 RESPONSE TO LOSS OF ALL OFFSITE POWER ON-1 04-001 Revision 13 Page 1 of 17 QUALITY CLASSIFICATION:
APPROVAL CLASSIFICATION:
(X)
QA Program
( ) Non-QA Program (X)
Plant
( ) Non-Plant
( )
Instruction',:
EFFECTIVE DATE.:'.
06/15/01 PERIODIC REVIEW FREQL(FNCY
- 2 Year PERIODIC REVIEWEU OAT.
06130/03 RECOMMENDED REVIEWS.
Procedure Shift Technical Advisor-C Shift Responsible Supervisor Shift Supervisor-C Shift Responsible FUM:
Manager-Nuclear Operations Responsible Approver Manager-Nuclear Operations FORM NDAP-QA-0002-1, Rev. 3, Page 1 of 1
ON-1 04-001 Revision 13 Page 4 of 17
- 3.
OPERATOR ACTIONS 3.1 RECORD date and time of event.
CHECKED i
Date lime (1)
I Shift Supervision 3.2 ENSURE proper Diesel Generator operation in accordance with OP-024-001.
3.3 If Diesel Generator A, B or C fail to start, PERFORM Attachment A approximately 30 minutes into LOOP EVENT to ensure 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> capacity of 250V DC batteries 1ID650 and 1D660:
3.4 REFER to EP-PS-100, Emergency Director-Control Room, to classify situation.
3.5 ENSURE ESW Pumps 0P504A, B, C, D operating in accordance with OP-054-001 after Diesel Generators start and re-energize Emergency Busses.
3.6 On direction of Shift Supervision, TRANSFER following systems cooling water supply from Service Water to Emergency Service Water in accordance with OP-i 11-001:
3.6.1 Reactor Building Closed Cooling Water 3.6.2 Turbine Building Closed Cooling Water 3.7 RESTORE Reactor Building Closed Cooling Water in accordance with ON-1 14-001 Loss of RBCCW.
3.8 RESTORE Turbine Building Closed Cooling Water in accordance with ON-1 15-001 Loss of TBCCW.
3.9 RESTORE Instrument Air System in accordance with ON-1 18-001 Loss of Instrument Air.
3.10 RESTORE RPS in accordance with ON-158-001 Loss of RPS.
3.11 CHECK shift of Drywell Cooling from Reactor Building Chilled Water to Reactor Building Closed Cooling Water 3.11.1 RBCCW Supply Viv FV-18771D OPEN.
3.11.2 RBCCW Retum VIv FV-18771C OPEN.
ON-104-001 Revision 13 Page 5 of 17 CHECKED 3.11.3 Chilled Water Supply Vlv to Drywell Coolers FV-1 8771 B CLOSE.
3.11.4 Chilled Water Return Vlv from Drywell Coolers FV-18771A CLOSE.
3.11.5 If desired, ALIGN control switch for Drywell Cooling Water Control Valves to the RBCCW position at panel 1C279.
3.12 ENSURE following Drywell Cooler Isolation Valves OPEN:
3.12.1 A CIrs CIg WtrOB Iso Valves HV-18781A1&A2.
3.12.2 A Cirs CIg Wtr IB Iso Valves HV-18782B1&B2.
3.12.3 B CIrs CIg Wtr IB Iso Valves HV-18782A1&A2.
3.12.4 B CIrs CIg Wtr OB Iso Valves HV-18781 B1&B2.
3.13 ENSURE Containment Instrument Gas Isolation valves OPEN:
3.13.1 Instr Gas Cmp OB Suct Iso SV-12605.
3.13.2 Instr Gas to Contn Iso SV-12651.
3.14 RESTART Control Rod Drive System by starting CRD Pump in accordance with ON-1 55-007 Loss of CRD System Flow.
3.15 RESTORE affected systems to desired status in accordance with ON-158-001 Loss of RPS.
3.16 ENSURE following running as necessary:
3.16.1 Turbine Generator Turning Gear Oil Pump (1 P111).
3.16.2 Turbine Bearing Lift Pumps (1 P1 09A-J).
3.16.3 Turbine Generator Emergency Bearing Oil Pump (1 P112).
3.16.4 Turbine Generator Turning Gear (when Main Turbine completed coastdown) (1S 103).
3.16.5 RFP Emergency Oil Pumps (1P125A,B,C).
ON-1 04-001 Revision 13 Page 16 of 17 After the main steam isolation valves close, decay heat raises reactor pressure to the lowest relief valves setting. Reactor water level decreases to the initiation setpoint for the RCIC and HPCI Systems, which actuate to restore reactor water level. If a Primary System boundary break should occur coincidentally with the loss of Offsite Power sources, the low pressure Core Spray, RHR and Automatic Depressurization Systems will initiate automatically to depressurize the RPV and restore reactor vessel level.
In the case where no break has occurred, immediate action must be taken to restore Drywell Cooling to prevent a high drywell pressure. With Unit Aux Buses 11A and 11 B de-energized, Drywell Cooling shifts from RBCW to RBCCW. It is important to confirm that the Drywell Cooler isolation valves and containment instrument gas isolation valves are open once the ESS Buses have been energized by the diesel generators. ESW is aligned to supply the RBCCW and/or TBCCW heat exchangers if adequate cooling to the diesel generators and ECCS equipment is available. The loads of the RBCCW and TBCCW heat exchangers are large enough to permit operation of 2 ESW pumps per loop.
With a complete loss of offsite power, the fuel pool cooling pumps will trip and Reactor Building HVAC Zones 1, 2, and 3 will isolate and go into the recirculation mode.
Moisture generated through evaporation and/or boiling of the fuel pool will spread to the secondary containment. Excessive moisture could cause adverse effects to safety related systems and components. ON-135-001 provides instructions for mitigating a Loss of Fuel Pool Cooling event.
Procedures ON-104-201, ON-1 04-202, ON-1 04-203 and ON-1 04-204 contain a list of the 4.16 KV Bus loads and the 480V Load Center loads that are energized on a loss of Offsite Power by the Standby Diesel Generators. ON-1 03-003 contains a list of all loads off the Unit Auxiliary Buses, and ON-003-001 contains a list of all loads off of Startup Bus 10.
FSAR 8.3.2.1.1.4 specifies each 250V battery has capacity without its charger to independently supply required loads for four (4) hours per FSAR Table 8.3-7. Table 8.3-7 shows various non-I E loads terminating at specified times to ensure four (4) hour capacity, however, plant design does not automatically shed these loads. This procedure sheds non-1 E loads at 30 minutes to ensure a four (4) hour battery capacity, per design.
[
1 I Applicant Level:
1R"RO 0: b-*
Applicant Name:
Question Type:
aCommon U RO only 0 SRO only Question #:
- 'z5 (enter number, if SRO only enter N/A)
SRO T
(enter number, if RO only enter N/A)
Answer: (cirde the answer key response)
A B
C
Reference:
(enter the answer key reference below)
Comment: (enter the comment below)
J........
Recommendation: (The grader is encouraged to discuss the matter with the NRC Chief Examiner before proceeding with the grading)
U C nge the correct answer.
U Do NOT change the correct answer.
ccept two correct answers.
EU Delete the question WMake clarifications to the question.
Changes / clarifications made to examination: (provide a description)
Mj.§j~
-Se~
r I
SReference(s) to support change / clarification made to examination:
.1.1 Justification for rejection of an applicant's comment:
Proctor:
Y"Change made in INK on the master examination copy Signature Date 1
V
OPERATIONS QUESTION AND ANSWER INPUT FORM RO 55 SRO 4f (A)
SYO17 H-2 Course (B)
Objective (C) Question Type (check one)
[
Multiple Choice Matching Free Format (Essay)
(D) Bank Operations OP002 (E) 1 2
3 4
5 6
7 8
rds ICatey Toic 1 To ic 2 JTA Seth Other Objs QuizO Retired I L9 characters)
ISystems MSIVs I
(F) Point Value:
1 (G)Answer Time: ---
]
(Minutes)
(I) Review Date (YYMM):
]
(H) Cognitive Level:
X I
Memory
.(Check one) 2 Comprehension 3
Application 4
Analysis 5
Problem Solving (J) QUESTION:
Unit 1 has scrammed and the MSIVs isolated. The cause of the isolation has been corrected and the MSIV isolation logic reset. With RPV pressure greater than 600 psig which one of the following is required to re-open the MSIVs?
- a.
Drain the steam lines, bypass the MSIVs with the steam drains, lower the DIP to less than 200 psid then open the inboard then the outboard MSIVs.
- b.
Open the inboard MSIVs, drain the steam lines, bypass the outboard valves with the steam drains, lower the DIP to less than 50 psid then open the outboard MSIVs.
- c.
Drain the steam lines, open the outboard MSIVs, bypass the inboard valves with the steam drains, lower the D/P to less than 200 psid then open the inboard MSIVs.
- d.
Open the outboard MSIVs, drain the steam lines, bypass the inboard valves with the steam drains, lower the D/P to less than 50 psid then open the inboard MSIVs.
(K) ANSWER: d.
4We-cPLI C4 1
CU, (ký") I tý 6 'ý- 11--A4-cc-A Z4W'cz 411Z.
File No. R 11-2 Form STCP-QA-325D Rev. 3, (M/95)
Page I of I
OPERATIONS QUESTION AND ANSWER INPUT FORM (L) REQUIRED MATERIALS:
None (M) K&A NUMBERfRATING: 223002, A4.03/3.6 (N) NOTES:
R.....
JUSTIFICATION:
Per procedure and system knowledge the outboard MSIVs are opened first, then the lines must be drained, then the D/P lowered to <50 psid then the inboards opened.
DISTRACTER A:
The lines are drained after the outboard valves are opened to allow steam drains downstream to drain the lines, once drained the steam line drains are used to bypass the inboard MSIVs. The DIP must be lowered to <50 psid.
DISTRACTER B:
The outboard MSIVs are opened first to allow steam line drains to drain the lines downstream of the inboard MSIVs.
DISTRACTER C:
The lines are drained after the outboard valves are opened to allow steam drains downstream to drain the lines. The D/P must be lowered to <50 Dsid.
EXAM OUTLINE CROSS-REF:
7EL:
- R:
Y to manually operate andlor monitor in the (0)
REFERENCES:
OP-1 84-001, Sect. 3.2 (P) POSITIONS:
R-RO S-SRO A-ASO N-NPO T -STA (check one or more boxes)
X I X (Q) Prepared by ED BOWLES AZ (L~
(R) Reviewed by:
Forim STCP-QA-325D Rev. 3, (8/95)
Page 1 of I File No. R 11-2
4 PROCEDURE COVER SHEET NUCLEAR DEPARTMENT PROCEDURE MAIN STEAM LINE ISOLATION AND QUICK ON-184-001 RECOVERY Revision 6 Page 1 of 9 QUALITY CLASSIFICATION:
APPROVAL CLASSIFICATION:
(X)
QA Program
( )
Non-QA Program (X)
Plant
(
)
Non-Plant
(
)
Instruction EFFECTIVE DATE:
09/28/00 PERIODIC REVIEW FREQUENCY:
2 Year PERIODIC REVIEW DUE DATE:
9-30-02 RECOMMENDED REVIEWS:
Procedure Owner:
Shift Technical Advisor-C Shift Responsible Supervisor:
Shift Supervisor-C Shift Responsible FUM:
Manager-Nuclear Operations Responsible Approver:
Manager-Nuclear Operations FORM NDAP-QA-0002-1, Rev. 2, Page 1 of 1
ON-184-001 Revision 6 Page 5 of 9 CONFIRM 3.16 3.17 CLOSE Mn Stm SJAE Iso HV-10107.
When directed by Shift Supervision AND initiating event is determined and cleared, RESET NSSSS Main Steam Line Isolation by depressing:
3.17.1 3.17.2 NOTE:
3.18 To OPEN 3.18.1 3.18.2 3.18.3 3.18.4 3.19 ALIGN for 3.19.1 3.19.2 3.19.3 3.19.4 3.19.5 3.19.6 Mn Stm Line Div 1 Iso Reset HS-B21-1S32 Reset push button.
Mn Stm Line Div 2 Iso Reset HS-B21-1 S33 Reset push button.
If primary containment integrity in jeopardy, it is acceptable to open MSIV's (IB first) with AP > 200 PSID. This action will not damage MSIV's. If conditions permit, equalizing around MSIV's is preferred.
IB MSIV's PLACE following control switches to AUTO:
Mn Stm Line A IB Iso HV-141-F022A.
Mn Stm Line B IB Iso HV-141-F022B.
Mn Stm Line C IB Iso HV-141-F022C.
Mn Stm Line D IB Iso HV-141-F022D.
steam line pressurization as follows:
PLACE AC MOV OL Byps HS-B21-1S37A to TEST.
PLACE DC MOV OL Byps HS-B21-1S37B to TEST.
OPEN Mn Stm Line IB Drain HV-141-F016.
OPEN Mn Stm Line OB Drain HV-141-F019.
ENSURE Mn Steam Line Warm Up HV-141-F020 OPEN.
After 2 minutes, PLACE AC MOV OL Byps HS-B21-1S37A to NORM.
ON-1 84-001 Revision 6 Page 6 of 9 CONFIRM 3.19.7 After 2 minutes, PLACE DC MOV OL Byps HS-B21-1S37B to NORM.
3.20 OBSERVE main steam line pressure INCREASING on Main Stm Press PR-10101C.
CAUTION OPENING MSIV'S WITH LARGE DIFFERENTIAL PRESSURE WILL CAUSE RPV PRESSURE TO DROP RAPIDLY AND RPV LEVEL TO SWELL.
3.21 When differential pressure across MSIVs is between 50 psid and 200 psid, OPEN OB MSIV's by PLACING following control switches to AUTO:
3.21.1 Mn Stm Line A OB Iso HV-141-F028A 3.21.2 Mn Stm Line BOB Iso HV-141-F028B 3.21.3 Mn Stm Line C OB Iso HV-141-F028C 3.21.4 Mn Stm Line D OB Iso HV-141-F028D 3.22 If necessary, OPEN any or all of the following drain valves:
3.22.1 OPEN following by depressing Drip Leg Drn HS-10112 OPEN push button:
- a.
Drip Leg Drn HV-10112A1.
- b.
Drip Leg Dm HV-10112B1.
- c.
Drip Leg Drn HV-10112C1.
- d.
Drip Leg Drn HV-10112D1 3.22.2 OPEN BPV Hdr Drip Leg Dm Byps HV-10108A by depressing HS-10108A OPEN push button.
3.22.3 OPEN MSV Bst Dm HV-1 0101 A,B,C,D by depressing common OPEN push button.
OP-1 84-001 Revision 18 Page 8 of 17 3.2 OPENING OF MAIN STEAM ISOLATION VALVES WITH DIFFERENTIAL PRESSURE ACROSS MAIN STEAM ISOLATION VALVES.
3.2.1 Prerequisites
- a.
Primary Containment Instrument Gas System in operation in accordance with OP-125-001.
- b.
Instrument Air System in operation in accordance with OP-1 18-001.
- c.
DC power available in accordance with OP-102-001 (125V).
- d.
DC power available in accordance with OP-1 88-001 (250V).
- e.
AC power available in accordance with OP-105-001 (480V).
- f.
No Main Steam Line Isolation signals present.
- g.
All Main Steam Isolation Valves closed.
- h.
Reactor Protection System in operation in accordance with OP-158-001.
3.2.2 Precautions
- a.
Do not open any Main Steam Isolation Valve with differential pressure of more than 200 psid across the valve. Under normal operating conditions, valve opening should be accomplished only after Main Steam pressure has equalized to less than 50 psid across valve.
- b.
If MSIV Isolation occurred, ensure both Inboard and Outboard MSIV's closed and all MSIV control switches in CLOSE position before MSIV isolation logic reset.
OP-184-001 Revision 18 Page 9 of 17 NOTE:
Any following condition will cause Main Steam Line Isolation:
- a.
Main Steam Line pressure < 861 psig with Reactor Mode Switch in Run.
- b.
Main Condenser Vacuum < 9 inches HgV (20.9 HgA).
(Can be bypassed with Reactor Mode Switch not in Run.)
- c.
- d.
Main Steam Line High radiation (15 times normal).
- e.
Main Steam Tunnel temperature > 1770F.
- f.
Main Steam Tunnel differential temperature > 990F.
- g.
Main Steam Line High flow (134%).
- h.
Logic Power Failure.
3.2.3 RESET Main Steam Line Isolation by depressing MN STM LINE DIV 1(2) ISO RESET HS-B21-1S32(S33) push button.
3.2.4 With differential pressure across Main Steam Isolation Valves, OPEN each Valve as follows:
- a.
PLACE each MN STM LINE A, B, C, D OB ISO HV-141-F028A, B, C, D control switch to AUTO.
- b.
CHECK each Outboard MSIV OPENS by observing Red indicating light ILLUMINATED.
- c.
CHECK OPEN MN STM LINE IB DRAIN HV-141-F016.
- d.
CHECK OPEN MAIN STM LINE OB DRAIN HV-141-F019.
- e.
JOG OPEN MN STM LINE DRAIN TO CDSR HV-141-F021.
- f.
ALLOW drain lines to heat up and drain any condensation to Main Condenser.
- g.
CLOSE MN STM LINE DRAIN TO CDSR HV-141-F021.
4 OP-184-001 Revision 18 Page 10 of 17
- h.
JOG OPEN MN STEAM LINE WARMUP HV-141-F020 to allow Main Steam Line pressure to increase.
CLOSE following by depressing DRP LEG DRN HS-101 12:
(1)
Drip Leg DRN HV-10112A1.
(2)
Drip Leg DRN HV-10112B1.
(3)
Drip Leg DRN HV-10112C1.
(4)
Drip Leg DRN HV-10112D1.
CLOSE BPV HDR DRIP LEG DRN BYPS HV-10108A by depressing HS-1 01 08A AUTO push button.
- k.
DEPRESS common CLOSE push button for MSV BST DRN HV-10101 A, B, C, D to close following:
(1)
MSV-1 Before Seat Drain HV-1 01 01A.
(2)
MSV-2 Before Seat Drain HV-10101B.
(3)
MSV-3 Before Seat Drain HV-10101C.
(4)
MSV-4 Before Seat Drain HV-10101D.
If pressure not equalizing across Inboard MSIV's, CLOSE following:
(1)
(2)
(3)
- m.
If Reactor Pressure higher than EHC pressure, RAISE EHC pressure by adjusting Main Turbine PRESSURE SETPOINT SELECTOR to prevent any rapid change in Reactor pressure.
- n.
When differential pressure across Inboard MSIV's equalized or < 50 psid, OPEN Inboard MSIV's by placing each MN STM LINE A, B, C, D lB ISO HV-141-F022A. B, C, D control switch to AUTO.
OP-184-001 Revision 18 Page 11 of 17
- o.
CHECK each Inboard MSIV OPENS by observing Red indicating light ILLUMINATED.
- p.
If closed, OPEN:
(1)
(2)
(3)
- q.
OPEN any or all drain valves, as necessary, listed in step 3.2.4.i of this procedure.
OPERATIONS QUESTION AND ANSWER INPUT FORM
'-~
SYO17 K-6 Course (B)
Objective (C) Question Type (check one)
[
Multiple Choice Matching Free Format (Essay)
(D) Bank Operations LXI1 0P002 []
(E) 1 2
3 4
5 6
7 8
Keywords :
1Cateo T
Tic 2 JTA Setting Other Objs Quiz Only Retir L<9 characters)
ISystems RWM Tic2 (F) Point Value:
I (G)Answer Time:
(Minutes)
(I) Review Date (YYMM):
(H) Cognitive Level: [ Ii (Check one) 2 3
X 4
5 Memory Comprehension Application Analysis Problem Solving (J) QUESTION:
Control rods are being withdrawn during a plant startup. The following conditions exist:
The Rod Worth Minimizer (RWM) is in operation Control rods are being withdrawn in group 4 Only one control rod remains to be withdrawn in group 4 The operator attempts to select and withdraw a control rod in group 5 Which one of the following describes the response of the control rod in group 5, the response of the RWM, and the required action?
The control rod...
- a.
will NOT withdraw, a control rod withdraw block will be applied to this rod.
Select the correct control rod in group 4.
- b.
will NOT withdraw, a select block will be applied to the control rod in group 5.
Bypass the RWM then select the correct control rod in group 4.
- c.
will withdraw to its withdraw limit, the last rod in group 4 will be identified as an insert error.
Promptly insert the control rod in group 5 to position 00.
- d.
will withdraw only one notch, then control rod withdrawal blocks will be applied to all other control rods.
Position the control rod in group 5 to its intended position.
(K) ANSWER:
)
Form STCP-QA-325D Rev. 3, (8/95)
Page I of I File No. R 11-2
OPERATIONS QUESTION AND ANSWER INPUT FORM (L) REQUIRED MATERIALS:
None (M) K&A NUMBER/RATING: 201006, A2.05/3.1 4(N) NOTES:
JUSTIFICATION:
A rod bkwkwill not be applied until the rod moves out rrent position, then blocks will b-e
- e to all rods. This is considered unintended rod motion becase ainc od is selected and moved one notch. The correct act move the control rod a ded position.
DISTRACTER A:
moves out of its current position.
DISTRACTER B:
A select block will not be applied DISTRACTER W "D
This rod ca be withdrawn to its withdraw limitq blocks will be applied after it leaves its initial po If a controlros positioned, the correct action is to promptly insert the control ro sition 00. This control rod is considered unintended rod moti cause an inco ntrol rod is selected and moved one nntnr.h LEVEL:
CROSS-REF:
K/A TEXT:
LEVEL:
TIER:
2 GROUP:
2 V0 - AiIiTY tto preict*d the impacts oT Toiiowing on ruu. vvu. I n. MII) based on those predictions, use procedures to correct, control or mi isequences of those abnormal conditions or operations: Out of seqL (0)
REFERENCES:
Op-131-001, SY017, K-6 it-Pe Dw.-
o"$6 (P) POSITIONS:
R-RO S-SRO A-ASO N-NPO T -STA (check one or more boxes)
X I X (Q) Prepared by ED BOWLES (R) Reviewed by:
Form STCP-QA-3251)
Rev. 3, (8/95)
Page I of I File No. RI 1-2
The RWM automatically latches into a rod group according to a specific rule. It will select the group, which is the highest group having less than three insert errors and having at least one rod withdrawn past its insert limit. When this occurs, rods contained in the currently latched group or any lower group that are withdrawn past their withdraw limit will be classified as withdraw errors, as will those rods in groups above the currently latched group that are withdrawn past their insert limit.
Upward (or downward) rod group latching also occurs when, upon completion of rod movements in the currently latched group, the operator selects the next in-sequence rod in the next higher (or lower) rod group. For upward latching, this occurs even with existing insert errors, as long as the maximum insert error rule, described above, is met. For downward latching, when this occurs, any rods in higher groups, which are not at the insert limit, are identified as withdraw errors.
Operation With Two Insert Errors If, while withdrawing rods at power levels below the LPSP, it becomes necessary or desirable to leave one or two rods at positions lower than their withdraw limits, the next higher group can be latched, and the startup allowed to proceed. This may be necessary due to operational problems with the specific rods. It is accomplished as follows:
When the group is reached that contains this one or two rods, which are to be left at some intermediate position, the rod or rods, are withdrawn to their respective positions. (It is assumed here that both rods are in the same rod group, although it is possible for them to be in different rod groups. It is also assumed that the rod group(s) in question is/are below the 50 percent rod density point.)
All other rods in the group are withdrawn to the withdraw limit position for that group.
Any rod in the next higher group is then selected. The RWM program latches (or shifts groups) up to this group, the ROD GROUP display shows the higher group number, and the one (or two) rod(s) remaining inserted in the lower group are identified as insert error(s).
23 031 D; Rod Worth Minimizer Revision 00 S:\\Training\\Operations Training Directory\\Training Material\\Systems\\TM-OP-031 D, Rod Worth Minimizer\\TM-OP-031 D-ST, Rod Worth Minimizier.doc
The withdrawal sequence can then proceed normally, so long as no more than two insert errors are allowed to occur.
There are times when the systems that interface with the RWM do not function properly. There is usually a warning that system status is not normal which enables the operator to perform actions to correct the situation prior to reaching off-normal conditions. The first line of defense is the alarm response procedures. These procedures provide guidance to various conditions in the form of probable causes, operator actions, automatic actions, and references, which are applicable to these conditions.
ROD POSITION INDICATION SYSTEM INOP (AR-103-001-H03)
ROD OUT BLOCK (AR-104-001-H03)
The details of these procedures can be found in the individual procedures.
There may be times when the RWM is required to be bypassed due to equipment failure or unforeseen circumstances. Procedure NDAP-QA-0338, Reactivity Management and Controls Program, Controls for Reactivity Control Systems (RWM, RSCS, and RBM), provides direction for bypassing the RWM.
The procedure contains a flowcharted form to authorize bypassing of the RWM.
ON-155-004, RPIS Failure, provides direction for placing substitute data, bypassing, and re-initializing the RWM when RPIS failures occur. These failures range from a failed reed switch to complete failure of the RPIS system.
TEST MODE OPERATION The RWM is tested in accordance with the following procedures:
SO-131-001, Startup Operability Demonstration Startup/Following System Failure This procedure demonstrates the ability of the RWM to indicate a control rod selection error and block rod withdrawal of an out-of-sequence control rod in Mode 2 at <10 percent of rated thermal power.
24 031 D; Rod Worth Minimizer Revision 00 S:\\Training\\Operations Training Directory\\Training Material\\Systems\\TM-OP-031 D, Rod Worth Minimizer\\TM-OP-031 D-ST, Rod Worth Minimizier.doc
The program will change up (latch the next higher group) during an up-power evolution, when all the rods in the currently latched group and in all lower groups have been withdrawn to their respective group withdraw limits, and a rod in the next higher group is selected. There is an exception to the requirement that all rods in previous groups be at their respective group's withdraw limit.
The RWM makes provision for a maximum of two rods to be at other than their group withdraw limit without it impacting the system's upward latch (these would appear as "insert errors," as described below).
The number of the currently latched group is displayed in the LATCHED GROUP box on the Main Screen of the RWM Touchscreen Display (Figure 5). An evaluation of the current rod position distribution, to determine the group which should be latched, is performed by the RWM program during system initialization and at various other times during normal program operation. This ensures that the proper group is latched at all times.
Loaded Sequence Rod movement sequence loaded within RWM System memory.
Low Power Set Point (LPSP)
The Low Power Set Point (LPSP) is the core average power level below which the RWM program enforces adherence to the operating sequence of rod withdrawals or insertions. When Reactor power is above the LPSP, the RWM program does not impose any constraints on operator requested rod movements. That is, the operating sequence ceases to be enforced by the RWM, above the LPSP. Rod blocks due to hardware failure, however, can occur at any power level.
Main Steam Line (MSL) flow is measured by the Feed Water Level Control (FWLC) System to determine when the plant is operating at 22 percent of Rated Thermal Power (RTP). This monitored parameter is inputted to the RDCS and PICSY to activate the LPSP. The setpoint can be adjusted by varying the trip value in the MSL flow sensor.
When the Reactor is operating below the LPSP, the BELOW LPSP box on the RWM Touchscreen Display changes color, from yellow to red, indicating that the RWM is enforcing the loaded sequence.
35 031 D; Rod Worth Minimizer Revision 00 S:\\Training\\Operations Training Directory\\Training Material\\Systems\\TM-OP-031 D, Rod Worth Minimizer\\TM-OP-031 D-ST, Rod Worth Minimizier.doc
I Apýiiýilievel:
0 RO
[] SRO SApplicant Name:
I ~
oly[] SRO ony Q uestio nl en erN/ )
Qusto Type (ete Comn R
nlyer if SRO only etrNA
- J~!*::!:*:::::*:*::!:*i:*::i~
~~i iii~:i~~*::i:i:i:::.............................................
- !::i:::::::::::::::::::::::::::!:i:i:i~:!*:i:*:iii:iii:ii~i::*:*:*
........ *: ::::::::*::::i*!::ii:*~~iii:iiii:i:i:i:i**::*:!::::::*:::::::*i::
- i:::::::*::::::::::*:::**::*::*:::::::*::*.
iiiiiiiiiiii~iiiiiiiiiii~i~iiiiii**i =**~~ii`ii~iii!*i~iiiiiiiiiiiiiiiiiiiiiii~iii~ ~~i ii ii~i~iiiiiiiiiiiiii~iii i ii*iiiiiiiiiiiiii~iiiiiiiii~i~**~ `*~~i~~iii~~~iiiiiii~
~i~~
QuestioneMce: (on(enterenumberrifeSROeonlynenterlN/A Comment: (enter the comment below)
OýJ-17F-000, Recommendation: (The grader is encouraged to discuss the matter with the NRC Chief Examiner before proceeding with the grading)
El Cge the correct answer.
El Do NOT change the correct answer.
c t two correct answers.
El Delete the question 1Make clarifications to the question.
Changes / clarifications made to examination: (provide a description)
{c~L~IL~-\\\\~
'(~ 1ýýo
-~-1~-)
6-
ý W-~~
~
iA-t-Fr( t I
.i~iii~
~ i ii i*
i i~iiiiii* !i!
' :**:** liiiii~ ~i* !il~i~
]~ ~
~
~
~
~
~ ~~~~~....
eecest chng I.lrfcto ae oeaiain Roctorenes tosupotChangeae clarificantion mastert examination:cp Signa
-8 tur Da8t6
/,4, n!0 -
6wwffiaý
OPERATIONS QUESTION AND ANSWER INPUT FORM RO 67 (A)
SY017 L-9 Course (D) Bank Operations [
oP002 LU O
c)
Objective (C) Question Type (check one) 1 ' n Multiple Choice Matching Free Format (Essay)
(E) 1 2
3 4
5 6
7 8
F Keor& 7s Category Topic 1 Topic 2 JTA Setting Other Obis.
..Quz Ony Retired (I9 characters)
I RC RECCONT J
(F) Point Value: [
(G)Answer Time:
(Minutes)
(I) Review Date (YYMM):
(H) Cognitive Level:
1 Memory (Check one) 2 Comprehension X3 Application 4
Analysis 5
Problem Solving (J) QUESTION:
Unit 1 is operating at 80% power with 76 Mlbm/hr core flow when a spurious feedwater flow signal causes a recirculation flow control runback. After the runback the following conditions exist:
APRMs oscillating between 44% and 48% power Core flow is 42 Mlbm/hr Green lights are illuminated above RX RECIRC LIMITER 1 RUNBK RESET pushbutton.
One (1) center region C-level LPRM upscale alarm is sealed in.
Two (2) peripheral A-level LPRM downscale alarms are sealed in.
In accordance with ON-164-002, Recirc Drive Flow Instrument Failure, and the Power/Flow Map, which one of the following actions is required?
- a.
Raise core flow to at least 44 Mlbm/hr.
- b.
Place the reactor mode switch in SHUTDOWN.
- c.
Monitor for power instabilities and wait for RE instructions.
- d.
Insert control rods in accordance with the cram array to less than 40% power.
(K) ANSWER: a.
Form STCP-QA-325D Rev. 3, (8/95)
Page I of I
~2~-cZ'P~i s aCZCi SVT3.4 o
I Coy,- 'ý- &
File No. RI I-2
OPERATIONS QUESTION AND ANSWER INPUT FORM (L) REQUIRED MATERIALS: Power to Flow Map (NDAP-0338)
,M) K&A NUMBER/RATING: 202001, A3.04/ 3.2 4(N) NOTES:
JUSTIFICATION:
The Green lights are illuminated above RX RECIRC LIMITER I RUNBK RESET pushbutton, indicating that the runback (caused by the spurious feedwater flow signal) can be reset, allowing flow to be raised.
DISTRACTER B:
The plant is in the immediate exit region, an immediate shutdown is NOT required.
DISTRACTER C:
You canit wait for RE.
T T*rII A fl1mr+i-n n
"rL..
I.
t*
LJL LJ Ll ~
ltkJllJ I tie proceuure s Tirst step is
£rAiD#soAfew should be done first and EXAM OUTLI]
CROSS-REF:
K/A TEXT:
I Lj+/-EVikL I TIER:
2 GROUP:
A3.UZ -- AD i int-l,*finn-I h*
2 natic operations oT tfe (0)
REFERENCES:
ON-164-002, Sect. 3.4, NDAP-QA-0338-10 (P) POSITIONS:
R-RO S-SRO A-ASO N-NPO T -STA (check one or more boxes)
X x I I
I (Q) Prepared by ED BOWLES (R) Reviewed by:
4'( C,1-t-Form STCP-QA-325D Rev. 3, (8/95)
Page I of I File No. RI 1-2
- i PROCEDURE COVER SHEET CORE FLUX OSCILLATIONS ON-178-002 Revision 10 Page 1 of 5 QUALITY CLASSIFICATION:
APPROVAL CLASSIFICATION:
(X)
QA Program
( )
Non-QA Program (X)
( )
Plant
( )
Non-Plant Instruction EFFECTIVE DATE:
07/03/01 PERIODIC REVIEW FREQUENCY:
2 Year PERIODIC REVIEW DUE DATE:
9-30-03 RECOMMENDED REVIEWS:
Procedure Owner.
Shift Technical Advisor-F Shift Responsible Supervisor:
Shift Supervisor-F Shift Responsible FUM:
Manager-Nuclear Operations Responsible Approver Manager-Nuclear Operations FORM NDAP-QA-0002-1, Rev. 3, Page 1 of I
a' ON-1 78-002 Revision 10 Page 3 of 5
- b.
Peak to peak oscillations trending towards 1Cw/cm 2 on LPRM's.
C.
Two (2) or more LPRM upscale lights flashing and clearing on a one to five second period.
- d.
Two (2) or more LPRM downscale lights flashing and clearing on a one to five second pedod.
3.4 If either Region !1 of Power/Flow Map entered with _> 50% of required LPRM upscale alarms operable, OR
- Abnormal flux oscillations determined to be associated with plant systems (FW, EHC, RECIRC, etc.) less than scram limits specified in step 3.3 observed PERFORM the following:
3.4.1 INITIATE TRA.
3.4.2 MONITOR APRM's and LPRM's for oscillations.
3.4.3 PERFORM following, as required, to rapidly suppress oscillations:
- a.
PROMPTLY INSERT control rods JAW RE Instructions in CRC Book to exit Region II.
OR CAUTION (1)
WITH ONE REACTOR RECIRCULATION PUMP IN OPERATION RATED PUMP SPEED IS LIMITED TO *80% PER TRO 3.4.4.
B.
ON-178-002 Revision 10 Page 4 of 5 CHU L-,,._.D
_CAUTION (2)
EXCEEDING THE CORE FLOW VALUE SPECIFIED IN THE CRC BOOK MAY CAUSE FUEL PRECONDITIONING LIMIT VIOLATIONS.
- b.
INCREASE the speed of the Operating Recirc PP(s) to exit the instability region, without exceeding the Core Flow Value in the RE Instructions in the CRC book.
3.4.4 If flux oscillations continue after performance of preceding steps, SCRAM Reactor lAW ON-1 00-101, Scram.
3.5 When conditions permit, NOTIFY Duty Manager and Reactor Engineering.
3.6 FORWARD completed copy of this procedure to following for review and retention:
3.6.1 Shift Supervisor D ate Signature Date 3.6.2 Nuclear Operations Supervisor-Shift O perations Date Signature Date 3.6.3 Manager-Nuclear Operations
/
Signature Date 3.6.4 DCS Supervisor
- 4.
REFERENCES 4.1 GE SIL 380, Rev. 1 4.2 NRC Bulletin 88-07 4.3 INPO SER 14-88 4.4 TS 3.4.1 4.5 BWROG March 18, 1992 Letter Re: Implementation Guidance for Stability Interim Corrective Actions 4.6 TRO 3.4.4
PAGI-.-~.
"Attache F-REACTOR ENGINEERING INSTRUCTIONS
-'Re'sIon
.5 Page 55 of 69 Unrit Sequence SURVEILLANCE INSTRUCTIONS Accep!
forUt ORE S7 CONTROL VALVE TESTING SHALL BE PERFORMED AT POWER LEVELS LESS THAN OR EQUAL TO
% RATED POWER.
- 1. Load reduction to be done with recirc flow.
NOTE. All other Turbine Valve Testing can be
- 2.
Perform Control Valve Testing.
performed at current power levels.
- 3.
Restore power at.
MW-hr.
(N/A if a ramp is in progress.)
WE CONTROL ROO EXERCISING PERMITTED AT s
% RATED POWER EACH CONTROL ROO SHOULD BE INSERTED ONE NOTCH FIRST THEN WITHDRAWN TO PREVIOUS POSmON.
POWERPLEX is programmed assuming the following plant conditions:
Recirc Loops in Operation 0 DUAL 0 SINGLE Bypass Valves 0 OP a' INOP EOC RPT Instrumentation 0 OP 0 INOP Scram Time Dep MCPROL 0 REALISTIC
. 0 INT 0 MAX Exposure Dep MCPROL 0 BOC to MWD/M`
0 BOC to EOC SHUTDOWNvCRAM ARRA Y INSTRUCTIONS ORE STAj SHUTDOWN INSTRUCTIONS
.1.
Use Emergency Power ReductionrShutdown Instructions.(Form NDAP-QA-03384) with shutdown control rod sequence-RWM shutdown control rod sequence is controling.
- 2.
Use startup sequence sheets in reverse order. RWM Startup Control rod sequence is controling.
(Document in 'NOTES'. column of-startup sheets.)
Initials Sequence Exposure RWM pointer is at CRAM RODS ARE INCLUDED IN THE EMERGENCY POWER REDUCTIONISHUTDOWN INSTRUCONS LOCATED BEHIND SHUTDOWN TAB UTILIZING:
CRAM Array I Shutdown Sequence Sheets UNIT OPERA TION INSTRUCTIONS E STA 1.Contact Reactor Englrueef.
- 2. If R.E..cannot be reached. insert the following rod(s), as necessary to reduce-power to below the APRM Rod Block line.
ROD FROM TO PCO PCO DATE I TIME Reselect and confirm previous moves:
I I
PC0 PCO DATE / TIME FORM NDAP-QA-0338-Re"
- Pae I. 6f (File A-23)
M-
.7.
~
opCAF # 9ýo -ý%
PAGE R_ý UNIT ONE, POWER vs. FLOW MAP Attachment K NDAP-QA-0338 Revision 5 Page 63 of 69 120 110 100 90 80 I
120 110 100 o*
I-
-0 70 60-50 40-4.:..
A on
.eO o p
30 Spo
- d.
i.-
I
- ~*
(h mdia. Scra
,R-" f'-'-
Ap r
Two.
umL Na0% a Circuati.
.Imdat ca
.eu l
4 30 20 10.
0-
- .I..
4
.4...
.1.~~~
I.
90 80 70 60 50 40 30 20 10 I
I-l 0
10 20 30 40 so 60 TO s80 Total Core Flow (Mlb i
lr)
FORM NDAP-QA-0338-10, Rev. /, Page 1 of 1 (File A-23).
K4, 0 90 i0o 110 NFE4B-NA-06 Rev:. 11l
PROCEDURE COVER SHEET PPL SUSQUEHANNA, LLC NUCLEAR DEPARTMENT PROCEDURE CORE FLUX OSCILLATIONS ON-1 78-002 Revision 10 Page 1 of 5 QUALITY CLASSIFICATION:
APPROVAL CLASSIFICATION:
(X)
QA Program
( ) Non-QA Program (X)
Plant
()
Non-Plant
( )
Instruction EFFECTIVE DATE:
07/03/01 PERIODIC REVIEW FREQUENCY:
2 Year PERIODIC REVIEWDE 'DATE, 9-30-03 RECOMMENDED REVIEWS:
ProdureOwer.
Shift Technical Advisor-F Shift Responsible Supervisor:
Shift Supervisor-F Shift Responsible FUM:
Manager-Nuclear Operations Responsible Approver.
Manager-Nuclear Operations FORM NDAP-QA-0002-1, Rev. 3, Page 1 of 1
ON-178-002 Revision 10 Page 2 of 5
- 1.
SYMPTOMS AND OBSERVATIONS 1.1 DOWNSCALE LPRM alarms cycling on and off as indicated by annunciator or Full Core Display Status lights.
1.2 UPSCALE LPRM alarms cycling on and off as indicated by annunciator or Full Core Display Status lights.
1.3 APRM readings oscillating as indicated on SIP Panel lC652.
1.4 LPRM readings in vicinity of a selected control rod oscillating as indicated on SIP Panel 1C652.
1.5 Inadvertent entry to Region II of Power/Flow Map.
- 2.
AUTOMATIC ACTIONS 2.1 Possible intermittent APRM UPSCALE Rod Block Alarms.
2.2 Possible APRM UPSCALE Scram or Half-Scram.
- 3.
OPERATOR ACTIONS CHECKED 3.1 As time permits, RECORD date and time of event.
I Shift Supervision 3.2 ENSURE non-peripheral rod selected to monitor LPRM's for oscillations.
3.3 IMMEDIATELY SCRAM Reactor lAW ON-1 00-101, Scram, per TS 3.4.1 if any following conditions observed:
3.3.1 Region I of Power/Flow Map entered.
.3.3.2 Region II of Power/Flow Map entered, with less than 50% of required LPRM upscale alarms operable.
3.3.3 Any of following:
- a.
Peak to peak oscillations trending towards 10% on APRM's (oscillations measured from minimum peak to maximum peak).
i DaWt e Ti me
ON-178-002 Revision 10 Page 3 of 5 CHECKED
- b.
Peak to peak oscillations trending towards 1Ow/cm 2 on LPRM's.
- c.
Two (2) or more LPRM upscale lights flashing and clearing on a one to five second period.
- d.
Two (2) or more LPRM downscale lights flashing and clearing on a one to five second period.
3.4 If either:
Region II of Power/Flow Map entered with Ž: 50% of required LPRM upscale alarms operable, OR
- Abnormal flux oscillations determined to be associated with plant systems (FW, EHC, RECIRC, etc.) less than scram limits specified in step 3.3 observed PERFORM the following:
3.4.1 INITIATE TRA.
3.4.2 MONITOR APRM's and LPRM's for oscillations.
3.4.3 PERFORM following, as required, to rapidly suppress oscillations:
- a.
PROMPTLY INSERT control rods lAW RE Instructions in CRC Book to exit Region II.
OR CAUTION (1)
WITH ONE REACTOR RECIRCULATION PUMP IN OPERATION RATED PUMP SPEED IS LIMITED TO
- 80% PER TRO 3.4.4.
ON-1 78-002 Revision 10 Page 4 of 5 CHECKED CAUTION (2)
EXCEEDING THE CORE FLOW VALUE SPECIFIED IN THE CRC BOOK MAY CAUSE FUEL PRECONDITIONING LIMIT VIOLATIONS.
- b.
INCREASE the speed of the Operating Recirc PP(s) to exit the instability region, without exceeding the Core Flow Value in the RE Instructions in the CRC book.
3.4.4 If flux oscillations continue after performance of preceding steps, SCRAM Reactor lAW ON-1 00-101, Scram.
3.5 When conditions permit, NOTIFY Duty Manager and Reactor Engineering.
3.6 FORWARD completed copy of this procedure to following for review and retention:
3.6.1 Shift Supervisor
/
Signature Date 3.6.2 Nuclear Operations Supervisor-Shift Operations
/
Signature Date 3.6.3 Manager-Nuclear Operations
/
Signature Date 3.6.4 DCS Supervisor
- 4.
REFERENCES 4.1 GE SIL 380, Rev. 1 4.2 NRC Bulletin 88-07 4.3 INPO SER 14-88 4.4 TS 3.4.1 4.5 BWROG March 18, 1992 Letter Re: Implementation Guidance for Stability Interim Corrective Actions 4.6 TRO 3.4.4
ON-178-002 Revision 10 Page 5 of 5
- 5.
DISCUSSION This procedure specifies actions required to reduce potential for fuel damage resulting from uncontrolled power oscillations, and is in compliance with Limiting Conditions of Operation as specified in TS 3.4.1.
If power oscillations occur and are not suppressed immediately, the MCPR Safety Limit may be violated.
The Reactor is most susceptible to power (flux) oscillations when operating in Regions I and II as identified on the Power/Flow Map. It is important to note that Regions I and II are not absolute with regards to preventing instabilities. As operating conditions approach Region II of the Power/Flow Map heightened awareness must be employed to ensure unstable operation does not occur or is mitigated. The instability regions may be approached during startup; shutdown; sequence exchanges; recirculation pump(s) trip(s) or runbacks; loss of feedwater heating events; inadvertent HPCI/RCIC injection etc.
Thermal hydraulic instabilities may be occurring if any of the following is observed:
- 1.
PEAK to PEAK oscillations are TRENDING TOWARDS, 10% on APRM's.
- 2.
PEAK to PEAK oscillations are TRENDING TOWARDS, 10 w/cm 2 on LPRM's.
- 3.
Two (2) or more LPRM UPSCALE lights cycling with one to five second period.
- 4.
Two (2) or more LPRM DOWNSCALE lights cycling with one to five second period.
A Reactor scram must be initiated if any thermal hydraulic instability is confirmed, regardless if a limit (i.e., 10 wcrm2 or 10% APRM) is actually exceeded.
Immediate Operator actions are required to suppress power oscillations which may lead to high local neutron flux levels without an automatic scram.
Region I, if entered, has a high probability of thermal hydraulic instabilities occurring.
This region requires an immediate manual scram to prevent safety limits from being violated.
Region II, has a lower probability of thermal hydraulic instabilities occurring than Region I, although if entered still requires immediate action. This region must be immediately exited if entered.
Operation near the Region II boundary should be minimized to provide the largest margin to potential core instabilities.
F A #c2ooo--5676 PAGE 4 OF 6 PROCEDURE CHANGE PROCESS FORM
- 1. PCAF NO. 2000- 5980
- 12. PAGE 1 OF 4
- 13. PROC. NO.
ON-164-002 REV. 18
- 4. FORMS REVISED R
RR, R_
R R
R_
- 5. PROCEDURE TITLE Loss ofReactor Recirculation Flow
- 6. REQUESTED CHANGE PERIODIC REVIEW NO El YES INCORPORATE PCAFS E-NO YES
- 2000-3365 REVISION E-PCAF
- DELETION E-ONLY)
- 7.
SUMMARY
OF /REASON FOR CHANGE Deleted the restriction on single loop operation limposed by PCAF 200 See edpages 3 and 4.
Siemens Power Company has corrected analysis erors that hav red h
- ction on single loop operation.
Revsed Core Operating Limits Reports (COLRs) for both units
.Cnin uodU<
- 8.
DETERMINE COMMITTEE REVIEW RE PORCREVIEW?.(REQD FOR '";*
YES 9
PO:!pRC9MTG#ý"
PLANT NDAP S, SICTE S, ANI'
"!i
- b '-"
FPSWAFTY EVALUAT'11'
'S
- y3/4-..
ERC REVIEW? (REQ'D F NO.
E$
Y 10.,ERCMTG#
N/A QUALITY NON-PLANT BLOCKS 11 THRU 142FFORM...,,...
- 15.
4-
" /.r
. 16 TRAINING REQUIRED?
PREPARE1 ETN DATE El NO 9 YES (TYPE) HotBox'00-169.:
(Print or Type) 17.
RESPONSIBLE SUPERVISOR SIGNATURE ATTESTS THAT RESPONSIBLE SUPERVISOR HAS "
CONDUCTED QADR AND TECHNICAL REVIEWUNLESS OTHERWISE 121*51/ýco DOCUMENTED IN'BLOCK 14 OR ATTACHED REVIEW FORMS.
...--DAT-k~CROSS DISCIPUNEtREVIEW (IF REQUIRED) HAS,'BEEN COMPLETED BY SIGNATURE IN BLOCK 14 OR ATTACHED REVIEW FORMS.
18` ý F M APPROVAL.
DATE ENTER N/A IF FUM l, FORM NDAP-QA-0002-8, Rev. 6, Page 1 of.2 (Electronic Form)
PCAF #2000-5676 PAGE 5 OF 6
(
PROCEDURE CHANGE PROCESS FORM
- 1. PCAF NO. 2000-5980
- 12. PAGE 2 OF 4
- 13. PROC. NO.
ON1-164-002 REV. 18
- 11. A 50.59 and 72.48 Evaluation per NDAP-QA-0726 Is required to be attached or referenced for all procedure changes except Expedited Reviews and Administrative Corrections. Either 11 a, b, or c must be checked "YES" and the appropriate form attached or referenced.
- a. 50.59 and 72.48 Screening Determination (Form NDAP-QA-0726-5)
[
YES F
N/A
- b. 50.59 or 72.48 Safety Evaluation (Note: 50.59 Safety Evaluations prepared on r
YES I
N/A Form NDAP-QA-0726-1 Rev. 5 or earlier also require a 50.59 & 72.48 Screening Determination)
Safety Evaluation No.
- c. Expedited Review/Administrative Correction-50.59 and 72.48 Evaluation not.
YES N/A Required
- 12. Is a Surveillance Procedure Review Checklist required per NDAP-QA-0722?
- YES
.1NO
- 13. Is a Special, Infrequent or Complex Test/Evolution Analysis Form required per v,
-' YES i
NO NDAP-QA-0320? (SICT/E form does not need to be attached.)
- 14. Reviews may be documented below or by attaching Document Review Forms NDAP-QA-0101-1.
REV.IEW-
,.-+-
REVIEWED'BY WITH," '
REVIEW.
NO COMMENTS +
DATE.
TECHNICAL REVIEW i
REACTOR ENGINEERING/NUCLEAR FUELS`..
AL IST*
OPERATIONS NUCLEAR SYSTEMS ENGINEERING, -
NUCLEAR MODIFICATIONS MAINTENANCE HEALTH PHYSICS NUCLEAR TECHNOLOGY CHEMISTRY
- ", +
OTHER Required for changes that affect, or have potential for affecting core reactivity, nuclear fýue, core po'wer level Indication or Impact the thermal power heat balance. (r)
Required for changes to Section Xl Inservice Test Acceptance Criteria.
+..
FORM NDAP-QA-0002-8, Rev. 6, Page 2 of 2 (Electronic Form)
I
r I
PROCEDURE CHANGE PROCESS FORM
- 1. PCAF NO. 2000- 5503
- 12.
PAGE 1 OF 3
- 13. PROC. NO.
ON-164-002 REV. 18
- 4. FORMSREVISED
-_R R
R R
R R
- 5.
PROCEDURE TITLE LOSS OF REACTOR RECIRCULATION FLOW
- 6.
REQUESTED CHANGE PERIODIC REVIEW Z
NO r-YES INCORPORATE PCAFS [
NO L] YES REVIs~o.ION LPCAF DELETION [i 0
- 7.
SUMMARY
OF / REASON FOR CHANGE Admin change. This change is made in fulfillment of CRA 264438 and p on on pump operating characteristics only. This information is already incorpo as a ment to the recirc pump operating procedure and is being restated at the appropriate times of this procedure. The.
PCAF incorporates an NSE recommendation to alternatively e th ng a reactor recirc pump near the 30% limiter equates to an approximate pump speed of 5 M.
mp "high oscillation" range (460 to 485 RPM) should be avoided when operating near th Continued
- 8.
DETERMINE COMMITTEE REVIEW REQUI PORC REVIEW? (REQ'D FOR.
- YES
- 9.
PORC MTG#
NA PLANT NDAP'S, SICTIE'S. AND FUP'S WISAFETY EVALU.AIO1NS ERC REVIEW? (REQ'D FO NO Li YES
- 10. ERC MTG#
NA NON-PLANT NDAP'S)
BLOCKS 11 THRU 14 2 OF FORM
- 15.
Eric Mill
- 3321 I 7/21100::1; 16 TRAINING REQUIRED?
PREPAR Eh.
ETN DATE "
NO Li YES (TYPE) NA'*,
(Print or 17.
18.
19.
RESPONSIB E SUPE VISOR SIGNATURE ATTESTS THAT RESPONSIBLE SUPERVISOR HAS CONDJOTED.QADR AND TECHNICAL REVIEW UNLESS OTHERWISE 24/0 DOCUMENTED IN BLOCK 14 OR ATTACHED REVIEW FORMS..
NA*TE
-CROSS DISCIPUNE REVIEW(OF REQUIRED) HAS BEENCOMPLETED D BY SIGNATURE IN BLOCK 14 OR ATTACHED REVIEW FORMS.'
FUM APPROVAL DATE RESPONSIBLE APPROVER
,A INITIALS DATE ENTER N/A IF FUM FORM NDAP-QA-0002-8, Rev. 5, Page I of 2 (Electronic Form)
.r PROCEDURE CHANGE PROCESS FORM
- 1. PCAF NO. 2000-5503
- 12. PAGE 2 OF 3
- 13.
PROC. NO. ON-164-002 REV. 18
- 11.
A 50.59 and 72.48 Evaluation per NDAP-QA-0726 is required to be attached or referenced for all procedure changes except Expedited Reviews and Administrative Corrections. Either I la, b, or c must be checked "YES' and the appropriate form attached or referenced.
- a. 50.59 and 72.48 Screening Determination (Form NDAP-QA-0726-5)
YES N/A
- b. 50.59 or 72.48 Safety Evaluation (Note: 50.59 Safety Evaluations prepared on YES N/A Form NDAP-QA-0726-1 Rev. 5 or earlier also require a 50.59 & 72.48 Screening Determination)
Safety Evaluation No.
- c.
Expedited Review/Administrative Correction-50.59 and 72.48 Evaluation not YES N/A Required
- 12. Is a Surveillance Procedure Review Chec*list required per NDAP-QA-0722?
[]YES NO
- 13. Is a Special, Infrequent or Complex7TsEvolution Analysis Form required per
-lYES
.NO NDAP-QA-0320? (SICT/E fonidoesh 668eed to be attached.)
- 14. Reviews may be documented ilow or b ttaching Document Review Forms NDAP-QA-0101-1 REVIEWiED BYWIT REVIEW NO COMMENTS QADR AI.
TECHNICAL RE~nVIEW
'7; i
7 3 REACTOR ENGINEERING/NUCLEFA_
FUELS-r*
- OPERATIONS NUCLEAR SYSTEMS ENGINEERIG___NG_-_..-.
NUCLEAR MODIFICATIONS MAINTENANCE HEALTH PHYSICS NUCLEAR TECHNOLOGY
_______________i-__
CHEMISTRY OTHER Required for changes that affect or have potential for affecting core reactivity, nuclear fel core power level indication or impact the thermal Odr heat balance. (5)
Required for changes to Section Xl Inservice Test Acceptance Criteria.
FORM NDAP-QA-0002-8, Rev. 5, Pa 2 of2 (Electronic Form)
PROCEDURE COVER SHEET NUCLEAR DEPARTMENT PROCEDURE LOSS OF REACTOR RECIRCULATION FLOW ON-164-002 Revision 18 PagelIof 9 QUALITY CLASSIFICATION:
APPROVAL CLASSIFICATION:
(X)
QA Program
( ) Non-QA Program (X)
Plant
( ) Non-Plant
( )
Instruction EFFECTIVE DATE:
10/27/99 PERIODIC REVIEW FREQUENCY:
2 Years PERIODIC REVIEW DUE DATE:
RECOMMENDED REVIEWS:
Procedure Owner Dayne R. Brophy Responsible Supervisor Grant Femsler Responsible FUM:
Manager-Nuclear Operations Responsible Approver General Manager-SSES FORM NDAP-QA-0002-1, Rev. 2, Page 1 of 1
ON-1 64-002 Revision 18 Page 2 of 9 SYMPTOMS AND OBSERVATIONS 1.1 Reactor power begins to decrease immediately.
1.2 Reactor vessel water level increases.
1.3 Any following indication on Standby Information Panel 1C652:
1.3.1 Recirculation loop flow(s) decreases in affected loop(s).
1.3.2 Jet pump flow(s) for affected loop(s) and total jet pump flow decreases.
1.4 Any following annunciator on Operating Unit Benchboard 1C651:
1.4.1 RECIRC M-G GEN A(B) LOCKOUT TRIP.
1.4.2 RECIRC M-G A(B) DRIVE MTR TRIP.
1.4.3 RECIRC PUMP A(B) HI PRESS/LO LEVEL TRIP.
1.4.5 RECIRC A(B) FLOW LIMIT RUNBACK
- 2.
AUTOMATIC ACTIONS 2.1 Possible main turbine trip and reactor scram because of RPV level swell caused by the trip of both Reactor Recirculation Pumps, depending on plant operating conditions.
2.2 Reduction in Reactor power, core flow, steam flow, feedwater flow, and generator output corresponding to recirc runback flow decrease.
- 3.
OPERATOR ACTIONS CHECKED 3.1 RECORD date and time of event.
. Shift Supervision 3.2 If both Recirculation Pumps trip, IMMEDIATELY SCRAM reactor.
i Date Time
.2oo 5980 ON-164-002 Revision 18 Page 3 of 9 CHECKED 3.3 If one Reactor Recirculation Pump trips:
NOTE:..*.,
If jet pump flow in operating loop is < 38 x 106 Ibm/hr computer generated core flow is not correct.
3.3.1 If jet pump flow in operating loop is < 38 x 106 Ibm/hr, ADD idle and operating loop flows together to determine actual core flow.
3 3 2 i
PLOT position on Power/Flow Map, Form
~ ~>NDAP-QA-33.8-1 0.
",3 33 PERFORM appropriate action as specified on Power/Flow
- 3. 3 Map..O we...o,
?.
3 ENSURE thermal power REDUCED to < 70% rod line.
REDUCE operating pump speed to 80% rated pumhp'-'
s:peed (80%
134 rpm) in accordance vwth OP7164-001 3 OMPLY With,OLR Section 8.0Limitsi 33- 'CMPLY-WMT~dh.pec LCOs*3" R~
T
.. -:..,*-:... *.:..,.:CA UTIO N ONR PRESSURE LOCKNG OF RECIRCULTbONISA 0VVES CMORE THAN APPROXIMATELY 5,MINUTES.
-3.38For
.stopped' pumpPLACE RECIRC A(B)'MOV OL BYPS
.HV-143-FO31A(B)/F032A(B) key switch to TEST pos.t. ion,-..
339 ENSURE RECIRC PUMP A(B) DSCH BYPS HV-143-F032A(B) OPEN.
4-.
=¥,.;3
-A 0;.
.CLOSE.RECIRC PUMP A(B) DSCH HV-143-FO3IA(B
~*-..-.*,*
..i Within 5 minutes,"OPEN RECIRC PUMP A(B)DSCH HV-143-FO31A(B).
3.312 After 2 minutes, PLACE RECIRC A(B) MOV OL BYPS HV-143-FO31A(B)/F032A(B) key switch to NORM position....
-).
4 ';;'
i:
L :'
.~
I.
,e00o - 5980
(
ON-164-002 Revision 18 Page 4 of 9 CHECKED 3.3.13 Prior to restart of pump, NOTIFY Duty Reactor Engineer to perform an evaluation of core thermal limits and preconditioning.
CAUTION DO NOT ATTEMPT TO. RESTART RECIRC PUMP IF OPERATING ABOVE 70% ROD LINE OR IF ANY FLUX OSCILLATIONS ARE OBSERVED.
0%
,i A When cause of trip corrected, RESTART Reactor.
Recirculation Pump in accordanic with'OP-164-001 Reactor Recirculation System.
"": ;3 i
"i If....pump will be out of service > hour, COMPLY with "
GO-100-009 Single Recirculation Loop Operation.
ýIi 3.
n to 1event-of Reactor Recirculation Pump runback
" PLOT position'on.n Power/Flow Map, Form
ý;.NDAP-ýQA-0338-I0.,,.ý9R Z PERFORM approprato action specified on Power/FIpw.
- -..43-,
':DETERMINE which limiter initiated.unback:-
Umter #1(300%) limiting.
m.(teen t ilgrd
ýý"!above RX RECIRC LIMtER IRUNBIK RESET~kt'ý,'--
~~ ~~
~
~
-1S-B31-1S15A(B) pusb~itn
[
- b.
ULmiter #2 (45%) limiting by Green light illuminated
,,above LOSS OF FW PP RUNBK RESET
,HS-B31-1$12A(B) pushbutton.
344** 71rvý-. ENSURE both pumps run back to value associatedwith
-'4 controlling limiter.
,..r 7'
7
~.,-.
."y'..*
1
(
ON-164-002 Revision 18 Page 5 of 9 CHECKED 3.4.5 OBSERVE following plant parameters WITHIN LIMITS corresponding to new power level:
- a.
Power to flow limits
- b.
Condenser vacuum
- c.
Feedwater flow/steam flow
- d.
RPV water level 3.4.6 DETERMINE signal that initiated runback from following:
- a.
Limiter #1 (30%) runback initiated by any following condition:
(1)
Total feedwater flow
- 20% for
> 15 seconds.
(2)
RECIRC PUMP A(B) DSCH HV-143-F031A(B) not fully open.
(3)
RPV water level < level 3.
- b.
Limiter #2 (45%) runback initiated by:
(1)
Any Circulating Water Pump protective trip.
(2)
RPV low water level (+ 30") and any of following:
(a)
Feedwater flow A, B, or C decrease to
- 20%.
(b)
Any Condensate Pump discharge pressure < 100 psig.
i'
- PCAF# #000
-5503 PAGLJ.I..
ON-164-002 Revision 18 Page 6 of 9 CHECKED (c)
Auto isolation of Feedwater Heaters String A, B or C due to high level in Feedwater Heaters I or 2.
ENSURE REACTOR RECIRC PUMP A(B) SPEEDS SY-B31-1R621A(B) IN MANUAL 3.4.7
.3.4.8 *!
For Irmiter #1 runback PERFORM following for one or both pumps as required:
CAUTIONO(
)
I 1ONTROL WITH THE RECIRC PUMP SPEED CONTROLLERS, MINIMIZE*
VTO AVOID INADVERTENT ENTRY INTO REGIONS I OR 11 OF THE CAUTION (2)
TROLWITH THE RECIRC PUMP SPEED CONTROLLER.S IANTNED AT APPX. 500 RPM. SPEED OSCILLATIONS.,
PUMPIS OPERATED BETWEEN 460 TO'485 RPM.
- ~~'~;~To prevent pump speed from changing When "
i'Umiter#1 reset, ENSURE GEN IA(IB) DEMAND
`-.ýadjusted suchthat GEN 1A(IB) SPEED de when controller DEMAND is decreased.
- "l DEPRESS RX RECIRC UIMITER I RUNBK RESET HS-B31-1S15A(B) pushbutton.
MONITOR GEN 1A(1B) SPEED Sl-14032A(B).
If speed increases rapidly. TRIP scoop tube on affected generator by depressing SCOOP TUBE.,
A(B) LOCK OR RESET HS-B31-1S03A(B) TRIP '`
pushbutton.
If previously illuminated, OBSERVE Green light above RX RECIRC LIMITER I RUNBK RESET HS-B31-1$1SA(B) pushbutton EXTINGUISHED.
ON-164-002 Revision 18 Page 7 of 9 CHECKED 3.4.9 For Limiter #2 runback PERFORM following for one or both pumps as required:
CAUTION WHEN ESTABLISHING CONTROL WITH THE RECIRC PUMP SPEED CONTROLLERS, MINIMIZE LOWERING CORE FLOW TO AVOID INADVERTENT ENTRY INTO REGIONS I OR II OF THE POWER/FLOW MAP.
- a.
To prevent pump speed from changing when Limiter #2 reset, ENSURE GEN 1A(1B) DEMAND adjusted such that GEN IA(1 B) SPEED decreases when controller DEMAND is decreased.
- b.
DEPRESS RECIRC A(B) LOSS OF FW PP RUNBK RESET HS-B31-1 S 12A(B) pushbutton.
- c.
MONITOR GEN 1A(1B) SPEED SI-14032A(B).
- d.
If speed increases rapidly, TRIP scoop tube on affected generator by depressing SCOOP TUBE A(B) LOCK OR RESET HS-B31-1S03A(B) TRIP pushbutton.
- e.
OBSERVE Green light above RECIRC A(B) LOSS OF FW PP RUNBK RESET HS-B31-1 S12A(B) pushbutton EXTINGUISHED.
3.4.10 CHECK RECIRC A(B) FLOW LIMIT RUNBACK annunciator CLEARED.
3.4.11
.NOTIFY Reactor Engineering.
ON-1 64-002 Revision 18 Page 8 of 9 3.5 FORWARD completed copy of this procedure to following for review and retention:
3.5.1 Shift Supervisor
/
3.5.2 3.5.3 Operations Supervisor-Nuclear Manager-Nuclear Operations Signature Date Signature
/
Date Signature Date Signature Date 3.5.4 DOS Supervisor
- 4.
REFERENCES 4.1 FSAR Section 5.4.1 Reactor Recirculation Pumps 4.2 FSAR Section 15.3 Decrease in Reactor Coolant Flowrate 4.3 GE SIL No. 380 4.4 M-143 Reactor Recirculation 4.5 Memo PLI-42281, S.A. Somma to A.M. Price, "Indicated Core Flow Anomaly,"
October 1, 1985 4.6 NRC Bulletin 88-07 Supplement 1 4.7 OP-164-001, Reactor Recirculation System
ON-164-002 Revision 18 Page 9 of 9
- 5.
DISCUSSION Loss of Reactor recirculation flow can be caused by the unexpected tripping of one or both Reactor Recirculation Pumps. Tripping of both pumps has the greatest impact on plant operation. At high power levels, the Main Turbine may trip automatically because of RPV water level swell and result in a Reactor scram. If the Reactor does not scram automatically following trip of both Reactor Recirculation Pumps, the reactor is immediately scrammed manually to avoid potential for core flux oscillations.
The Reactor may be operated at reduced power with one Reactor Recirculation Pump out of service and the other driving half of the jet pumps. The idle Reactor Recirculation Pump is kept hot by reverse flow through the loop and is not started unless the idle loop temperature is within 50°F of the operating loop temperature, and the Reactor is operating below the 70% rod line.
Loss of one Reactor Recirculation Pump during Reactor operation does not initiate any Reactor Protection System or safeguards systems actuation because fuel thermal margins are maintained. Flow in the idle jet pumps reverses in approximately 6 seconds, and flow in the operating jet pumps increases to about 143% of its normal flow if the operating pump is at 84% pump speed. If the jet pump flow in the operating loop is less than 38 x 106 Ibm/hr, then flow through the idle jet pumps is not reverse flow.
In this case the total core flow logic of automatically subtracting loop flows is not correct.
If total core flow decreases into Region I of Power/Flow Map, the Reactor is manually scrammed to avoid potential for core flux oscillations. The operator reduces flow in the operating loop to single pump flow criteria. Single pump flow criteria is based on reducing the operating pump speed to less than 80% rated pump speed (80% = 1.344 rpm), in accordance With OP-164-001.
Automatic trips of Reactor Recirculation Pumps can come from EOC-RPT trip (CV fast closure, > 506 psig; SV closure, < 5.5% closed) or ATWS (RPV low level 2, Reactor high pressure). The ATWS and EOC RPT trips open the RPT breakers which in turn cause the MG set drive motor to trip. A LPCI Initiation signal with low Reactor pressure
(< 236 psig) trips the Reactor Recirculation Pumps by closing the discharge and discharge bypass valves.
Two speed limiters limit recirc flow by limiting generator speed on the Reactor Recirc MG sets. Limiter #1 limits speed to 30% and provides NPSH protection for the recirc pumps. Umiter #2 limits speed to 45% and prevents spurious scrams due, to transients on the Condensate/Feedwater System and Circulating Water System.
K3~9
[
I1 Applicant Name:
- i.
... 'a".
iiiiiiiiiiiiiiiiiiiiiiiiiii~iiiiiiiiiiiiii~i i~!
!**i!iiiiiii~iiiiiiiiiiiiiiiiiiiiiii* !iiii ;!*i~ ~*!;~i ii~iiiiiiii~iiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiiii!iii~ii~iiiii Q
u s t o y e Q-uestio..n-Tye Com o 0 (etr onlyer i0 SRO only etrNA QS t nMRO (ente r nu mber, if SRO only enter N/A)
JAnswer: (circle the answer key response)
A B
C D
!i~iii!ii~i!iii~ii!iii~ii~~iiii*;i~i!iii~i...................
X -
1 -.~;;*ii;;!i~*i!**i;~
~i~ ~**;i ii~iiiiiiiiiii~~ii~i~iiiiiiiii~~ii!*i~i;!i*!i~**i!i~~ii~iiiiiiiiiiiiiiiiiii~
~i*ii~i~~~iiiii**iii~~iiii~i~~ii i
ee n
e a n s w e r Comment: (enter the comment below)
Q'QV&4)
/74&
-C 26
~(c
~c Recommendation: (The grader is encouraged to discuss the matter with the NRC Chief Examiner before proceeding with the grading) 0 Change the correct answer.
"Do NOT change the correct answer.
LI64ept two correct answers.
0 Delete the question ake clarifications to the question.
Changes / clarifications made to examination: (provide a description)
Reference(s) to support change I clarification made to examination:
Justification for rejection of an applicant's comment:
r~
4i T
% ý I............
I...............................
Proctor:
LI Change made in INK on the master examination copy P,- J7 N
-- ?
J" Signature Date
OPERATIONS QUESTION AND ANSWER INPUT FORM R0 69 SRO-f (A)
SY017 C-1 Course (B)
Objective (C) Question Type (check one)
X 1 Multiple Choice Matching Free Format (Essay)
(D) Bank Operations F5K 0P002 (E) 1 2
3 4
5 6
7 8
Keywords Category Topic 1 Topic 2 JTA Setting Other Obis.
Quiz Only Retired
(<9 characters)
RHR SY-017 C-1 C
SDC (F) Point Value:
(G)Answer Time:
(Minutes)
(I) Review Date (YYMM):
(H) Cognitive Level:
X I
(Check one)L 2
3 4
5 Memory Comprehension Application Analysis Problem Solving (J) QUESTION:
A shutdown and cooldown is in progress on Unit 1. Per OP-149-002, RHR Shutdown Cooling, the required level band is established then Shutdown Cooling (SDC) is placed into service with one Reactor Recirculation (RR) loop shutdown.
One (1) hour after establishing SDC, the operating RR pump trips. Which one of the following describes the action to be taken for reactor water level?
Adjust reactor water level...
- a.
to a new band of +35 to +50 inches.
- b.
to a new band of +90 to +100 inches.
- c.
maintaining level within the band established before the event.
- d.
maintaining level above the band established before the event.
(K) ANSWER:
c.
- L' ro ý
'.S.Dtr'-
4-Pý VN 7 I
Form STCP-QA-325D Rev. 3, (8/95)
Page I of I File No. RI 1-2
OPERATIONS QUESTION AND ANSWER INPUT FORM (L) REQUIRED MATERIALS:
None
'M) K&A NUMBER/RATING: 205000 / K6.03 1 3.1 / 3.2 4(N) NOTES:
JUSTIFICATION:
A level of +90 to +100 inches is established before starting SDC and is maintained throughout SDC operations - with RR flow and without RR flow.
DISTRACTER A:
+90 to +100 inches is established.
DISTRACTER B:
Once +90 to +100 inches is established before starting SDC, this level is maintained throughout SDC operations - with RR flow and without RR flow.
DISTRACTER D:
+90 to +100 inches is established before establishing SDC and is maintained throughout SDC operations.
EXAM OUTLINE LEVEL:
~'
CROSS-REF:
TIER:
2 2
GROUP:
2 2
i.
K6.03 - Knowledge of the effect that a loss or malfunction of the following will have on the SHUTDOWN COOLING SYSTEM: Recirculation System.
QUESTION BANK:
~
~
~
SOURCE:
MODIFIED:
X NEW:
N:
- ` ``..:.`
`.
10CFR55:
41(b)(10)
X.2
.N...................
COMMENTS:
The original question asked for the required reactor water level to secure all RR flow and why. The question was changed to the action to be taken for reactor water level if the operating RR pump trips once the conditions of OP-149-002 are established. The answer and all distracters changed.
(O)
REFERENCES:
OP-1 49-002, 3.1.1, 3.1.2 (P) POSITIONS:
R-RO S-SRO A-ASO N-NPO T -STA (check one or more boxes) x I x I
I I x I
(Q) Prepared by Phil Ballard (R) Reviewed by:
F CE Form s'rcP-QA-3251)
Rev. 3, (8/95)
Page I of I Filc No R 11-2
Applicant Level:
1 0- 0 Applicant 11ý
ý".'
[-Question Type:
. X..'..:1..
I....
Question RO (enter number, if SRO only enter N/A)
SRO (enter number, if RO only enter N/A)
Answer: -(circle the answer key response)
(A B
C D
Reference:
(enter the answer key reference below) 6 0-
'ýsý
(ý 1 -7 S c--ý- - %
I 0.:.;o.'.-
. I I..........
Comment: (enter the comment below) k, L
. I...
Recommendation: (The grader is encouirged to discuss the matter with the NRC Chief Examiner before proceeding with the grading) 0 Change the correct answer.
E0 DoNOT change the correct answer.
0 Accept two correct answers.
S':elete the question El Make clarifications to the question.
Changes / clarifications made to examination: (provide a description)
IReference(s) to support change/Iclarificationmdte mnain--'
- -O t-(-
Justification for rejection of an applicant's comment:
AI Proctor:.
EV
?.ang m ade i the master examination copy*.k**
.*-.*b\\-*,'*
-=,.
i'\\\\"1 Signature D ate
OPERATIONS QUESTION AND ANSWER INPUT FORM (R-A RO 92 SRO-,"
(A)
SY017 J-1 Course (D) Bank Operations X
OP002 M (3)
I O.e Objective (C) Question Type (check one)
M ultiple Choice Matching li Free Format (Essay)
(E) 1 2
3 4
5 6
7 8
Keyords-Cat Toic 1 Toic2 JTA Sellin 0lherObs Quiz On Retired (L9 characters)
Systems RPV (F) Point Value: L--I (G)Answer Time:
(Minutes)
(1) Review Date (YYMM):
(H) Cognitive Level:
1 Memory (Check one) 2 Comprehension 3
Application X
4 Analysis 5
Problem Solving (J) QUESTION:
During a reactor heatup the following temperature readings are recorded on Attachment A of SO-1 00-011, Reactor Vessel Temperature and Pressure Recording:
- 0800-2420 F 0815 - 2630 F 0817-Startup was temporarily halted
- 0830 - 2396F
- 0845-268°F
- 0900 - 311°F Per GO-1 00-002, which one of the following is the maximum allowable temperature at 0915?
a 329°F b
339°F
- c.
353°F d
363 0F (K) ANSWER:
a.
Q e~A (A)\\
'I I As
ýý V :>
£fV-%,
C ~
&'J~
'J~XC L
~L0 F
,2
-IF7.
File No. RI 1-2 Form STCP-QA-325 D Rev. 3, (8/95)
Page 1 of I
OPERATIONS QUESTION AND ANSWER INPUT FORM (L) REQUIRED MATERIALS:
None (M) K&A NUMBERIRATING: PWG, 2.1.32/3.4 (N) NOTES:
JUSTIFICATION:
Temperature cannot raise more than 90°F per hour. Because heatup rate was allowed to lower (0830) and temperature to lower to 239°F the temperature cannot exceed 329°F during the next hour.
DISTRACTER B:
This would exceed 90OF/hr from 0830.
DISTRACTER C:
This would exceed 90°F/hr from 0830.
Fl)qT'RACrTP*R f.l I This would exceed 90 0F/hr from 17A10 JkiAM UUTLINE CROSS-REF:
VIA TPY*T.
LAVEL:
KU I
1-KU TIER:
3 3
GROUP:
a 2.1.32 - Ability to explain and applY sy-stem limits andr (O)
REFERENCES:
GO-100-002, Secj6* and SO-100-0 11, pg. 6 (P) POSITIONS:
R-RO S-SRO A-ASO N-NPO T -STA (check one or more boxes)
X I x I
I I
I (Q) Prepared by ED BOWLES (R) Reviewed by:
R,8e (7ý_
Form STCP-QA-325D Rev. 3, (8/95)
Page I of I File No. RI 1-2
PROCEDURE COVER SHEET REACTOR VESSEL TEMPERATURE AND PRESSURE RECORDING SO-100-011 Revision 12 Page 1 of 17 QUALITY CLASSIFICATION:
(X)
QAProgram
(
)
tI4on-QA Program APPROVAL CLASSIFICATION:
(X)
Plant
(
)
Noi o
n-Plant
( )
Instruction EFFECTIVE DATE:
PERIODIC REVIEW FREQUENCY:
N/A PERIODIC REVIEW DUE DATE:
N/A RECOMMENDED REVIEWS:
Procedure Owner:
Jay Barnes Responsible Supervisor:
David T. Walsh Responsible FUM:
Manager-Nuclear Operations Responsible Approver:
General Manager-SSES FORM NDAP-QA-0002-1, Rev. 2, Page 1 of 1
SO-100-011 Revision 12 Page 6 of 17 6.1.2 On Reactor Coolant System Temperature and Pressure Log (Attachment A), RECORD following information every 15 minutes:
- a.
Recirc loop A temperature
- b.
Recirc loop B temperature
- c.
Reactor Vessel Bottom Head Drain Temperature
- d.
Reactor Vessel Pressure
- e.
Rx Steam Dome Temperature (Applicable only when Rx Coolant Temperature > 212 0F)
CAUTION IF 15 MINUTE TEMPERATURE CHANGE > 250F, ACTION SHOULD BE TAKEN TO REDUCE HEATUPICOOLDOWN RATE. CONTINUING HEATUP OR COOLDOWN AT THIS RATE WILL LEAD TO A TS VIOLATION.
NOTE:
Calculated temperature change is change in temperature that occurred in previous 15 minutes.
6.1.3 CALCULATE temperature changes for following and RECORD on Reactor Coolant System Temperature and Pressure Log (Attachment A).
- a.
Recirc loop A
- b.
Recirc loop B C.
Reactor Vessel Bottom Head Drain
- d.
Rx Steam Dome Temperature (Applicable only when Rx Coolant Temperature > 212 0F)
(I) 6.1.4 On Reactor Coolant System Temperature and Pressure Log (Attachment A), CONFIRM compliance with SR 3.4.10.1 once every 30 minutes (refer to TS Figure 3.4.10-1 Minimum Reactor Vessel Metal Temperature vs. Reactor Vessel Pressure, Attachment B) by verifying following applicable statements for Heatup and Cooldown Events:
- a.
Reactor Vessel Pressure and Temperature are to right of curve C.
Attachment A SO-100-011 Revision 12 Page 16 of 17 REACTOR COOLANT SYSTEM TEMPERATURE AND PRESSURE LOG All Rx Coolant Temp and Press Data shall be recorded until Heatup, Cooldown or Inservice Leak and Hydrostatic testing is complete. TS Required Actions should only be entered if Rx Steam Dome Temperature AT's are > 100OF in any one hour.
However, ALL AT's should be maintained <250 in any 15 minute period.
SOURCE 1I*LC32-I KU0 NFP02 RTHB31.1R650 INA TR-831-IR650 JNA TR-B21-IR008 IN NRT01 P NT02 IP T. 8) 1NFA05-NA ERE NFP03 NRT51 NRT53 NLT01 PPLICABLE NRT52 NRT54 I
II RECIRC RECIRC BTM IOD X STM REACTOR LOOP A OOP B BOTTOM HEAD RAIN OME CONFIRM rIME/DATE PRESSURE ECIRC LOOP ELTA ECIRC LOOP DELTA DRAIN TEMP F PELTA X STM DOME ELTA COMPLIANCE AS NECESSARY PSIG TEMP F EMP "F TEMP IF rEMP F
EMP "F EMP "F EMP IF W/SR 3 4 10 1
__________ I______________
REACTOR VESSEL FLANGE AND TOP HEAD FLANGE TEMP LOG Rx Vessel Flange and Rx Vessel Top Head Flange Temp need only be recorded when in Mode 4 with Rx Vessel head bolting studs under tension and coolant temp s100"F or at least 30 minutes during tensioning c Rx Vessel head bolting studs.
TR.B21-rR.B21.
IR006 IR006 (POINT #4)
POINT #2)
IX VESSEL RX VESSEL XONFIRM
- LANGE rOP HEAD
^OMPLIANC IME/DATE EMP LOCAL FLANGE TEMP
'/SR 34 10 S NECESSARY IC007 LOCAL C007o 34 108,0.
34 109
.1:
- 1 Sniry Review and Confirmatlon above recorded data is accurate, compliant, and complete.
SHIFT SUPERVISION Page 3 of 3 DATE Time
This I
a
-g- "'
77, Applicant Name:
N.":
V. -X I.. X N.:
X....'.
Question Type:
2ýbommon El RO only 0 SRO only..........
- F.......
X 1*
x, X,
Question M RO (enter number, if SRO only enter N/A)
I S RO (enter number, if RO only enter N/A) ax, X
FA-n s w er: (circle the answer ke.y response)
A B
D
- 6.
- 6.
Reference:
(enter the answer key reference below)
Comment: (enter the comment below) 16r 41,- Jd Q/O b*e
,6S ý0 Recommendation: (The grader is encouraged to discuss the matter with the NRC Chief Examiner before proceeding with the grading)
I *acenge the correct answer.
[] Do NOT change the correct answer.
ccpt two correct answers.
LI Delete the question Mke clarifications to the question.
Changes / clarifications made to examination: (provide a description)
C,
~
AM E(-
EL-~o.
SReferenc e(s) to support change /clarification made tý examination:
AýýIk Justificatio for re.ecion...anapplcants.coment X X a
Proctor:
Pthange made in INK on the master examination copy Signature Date
OPERATIONS QUESTION AND ANSWER INPUT FORM 10 RO 95 SRO -9 (A)
AD045 Course (B)
Objective (C) Question Tvye (check one) 1 Multiple Choice Matching ni Free Format (Essay)
(D) Bank Operations
]
OP002 L_
(E) 1 2
3 4
5 6
7 8
K
ý ords =
Cateo Toi 1 Toi 2 JTA Sethry 01her Obs.
Quiz Onl Retie L<9 characters)
OPS ONXXXXXX (F) Point Value:
]
(G)Answer Time:
(Minutes)
(H) Cognitive Level:
1 (Check one)!
2 3
X4 5
Memory Comprehension Application Analysis Problem Solving (I) Review Date (YYMM): [--
(J) QUESTION:
Unit 1 is operating at 95% power when the High Pressure Coolant Injection (HPCI) system initiates on a spurious high drywell pressure signal. Which one of the following sets of parameters would result from this transient?
- a.
RISE NO CHANG LOE
'LOWIER b
. 'O CHANG1E LW IE OCAG
- c.
RIE NO:CHANGE RISE LOWER
- d.
NO CHANGELOE LOWER NO CHANGE (K) ANSWER:
c.
ý C.C-Q-eý-
I&-%-'
(
o
~
ce d
A'z W) c~
r~
c
~~
G~L~CZ~
A-
~
K W
C~CJ\\
~~ Q~~U4~ -
, VF "fJ-K zs' Y r W Form STCP-QA-325D Rev. 3, (8/95)
Page I of I
~,6.
File No. RII-2
OPERATIONS QUESTION AND ANSWER INPUT FORM (L) REQUIRED MATERIALS:
None (M) K&A NUMBER/RATING: 4(N) NOTES: 2.2.34/2.8 M
z
.JIT T_4"'TVTtrnrA'T I
- "i'T
?"i'.
DISTRACTER A:
DISTRACTER B:
DISTRACTER D:
An inadvertent HPCI injection will cause a rise in power from the cooler feedwater, a reduction in feedwater flow to maintain RPV water level with the additional HPCI flow, increased main generator output from the rise in power, and no change in core flow.
Generator output will rise because of the rise in power.
Reactor power will rise, core flow will not lower and feedwate flow will lower Reactor power will rise, core flow will not lower generator output will rise and feedwate flow will lower (0)
REFERENCES:
Chapter 13, FSAR (P) POSITIONS:
R-RO S-SRO A-ASO N-NPO T -STA (check one or more boxes)
I X 1 x I
I
[
I (Q) Prepared by ED BOWLES (R) Reviewed by:
cL Form STCP-QA-325D Rev. 3, (8/95)
Page 1 of I File No. R 11-2
REAL TIME DATE I*Environmnent i~nitial Conditions
]
RUN TIME 17:24:00 Aug 10 06o 0:0s 0
ýA
ýE00
.I c u-*
F-HE-L -1 FD-SP-L-AY-ICOKTR DKD lF-TElIRST IcrcI I"NIIRU-N I MONITORED PARAMETER TRENDS MPT line 1:
ZAONRC51IR603AA APRM FLUX CHAN A (RED) line 2:
ZAOHRC5llR603CA APRH FLUX CHAN E (RED) 10 Minutes at 1 sample per 5 seconds 10 Hinutes at 1 sample per 5 seconds-I................
3 1,43 3:23 5:33
- 43 3:3 3 3
1:4, 3
3:43
$ 4b
- a line 3:
ZAOWRC511R603BA APR4H FLUX CHAN B (RED) line 4:
ZAOXRI00O1[2]
GENERATOR GROSS OUTPUT RED=N BLU 10 Minutes at 1 sample per 5 seconds 10 Minutes at 1 sample per 5 seconds k
I I.
.I..
....*.... I ' ' '.........
.. I I..
.I
. I..
In
.... *..... I.... t.....I 3
1:41 5: 2 3:93
$:43 10:3 IS I4*11 31:313 5:)
3:43 3:13 "r-J InGr MALFUNCTIONS HP1 52004
<17~s Fie >
TOP PAGE UP
(,OsFa ormat Z
jýj QART 7
WOTI OM PAGE DOWN LINE UP LINE DOWN
- p J,.
F'
SCO07 D Information Rev. 1, 12/27/00 Page 9 of 11 The opening of a relief is a relatively minor transient for the reactor and core. Variations in plant parameters are shown on Figure 6. The sudden increase in steam flow leaving the reactor vessel causes a mild depressurization transient. The pressure regulator senses the system pressure decrease and within a few seconds, closes the turbine control valves far enough to stabilize reactor pressure at a slightly lower pressure. Reactor power settles out at nearly the initial power level.
Turbine megawatts will decrease from diversion of steam through the SRV to the suppression pool (a 720,000 lbm/hr). The decrease in MW output will be proportional to the percent change in turbine steam flow. Since the steam flow through the SRV is not measured at the steam line flow restrictors, the feedwater control system will see a steam/feed flow mismatch. Final steady-state reactor vessel water level will be slightly lower than its initial value, such that the level error signal present will offset the steam/feed flow mismatch signal.
This transient is classified as a decrease in moderator temperature due to the small decrease in pressure and the corresponding PS'I,4t relationship. However this, is not the critical parameter during this transient. The major concern with a stuck open relief is the resultant heating of the suppression pool to its design temperature and resultant decrease of the pool heat capacity. If the temperature of the pool gets too high, condensation of steam during a LOCA or ah ADS blowdown will not be complete, possibly causing containment pressure to exceed its design limits. If a stuck open relief cannot be closed, the reactor will have to be shut down and cooled down. Any of the following conditions will require a scram; It becomes evident that the SRV will not close.
" The SRV has not closed after two minutes have elapsed.
" Suppression pool temperature has reached 105 *F.
Shutting down the reactor will ensure the plant is not operating without sufficient heat capacity in the pool. After the scram, the MSIVs should remain open and the cooldown performed by steaming through the bypass valves to the main condenser. It is possible that with the relief open, the cooldown may exceed the limit of 100 °F in an hour period.
VI.
Inadvertent Initiation of HPCI As stated earlier, this transient is actually classified as an increase in reactor coolant inventory in the FSAR. Because it is the single event analyzed in that category, and because it does result in an increase in core inlet subcooling, it is included here. The inadvertent initiation of HPCI is assumed to occur as a result of operator error.
Figure 7 illustrates the net effect of this transient in terms ofrsteam flow, feedwater flow, and water level. In Figure 7a, plant is operating at steady-state. In Figure 7b, HPCI has inadvertently been initiated and HPCI has reached rated flow. Though there is no measured steam flow feed flow mismatch, there is actually a greater rate of feed into the vessel, since the HPCI injection is greater than its steam usage. Reactor vessel water level will rise as a result This creates a water level error signal causing the feedwater flow to be reduced. The indicated steam/feed flow mismatch signal will offset the higher water level signal to maintain a constant water level in the vessel. See Figure 7.
G:VNTGOPS\\PendingOraft Fi~e\\scDO7sd\\dscOO7d\\SCO07 0.doc
SC007 D Information Rev. 1, 12/27/00
(
Page 10 of 11 A plot of major plant parameters for this transient is shown on Figure 8.
n HPQI W*I--
.steam flow out of the vessel increase.HPCI steam flow is only about 180,000 Ibhi 1.3 percent of rated steam flow. -Thusc, no significant pressure drop will be seen in th" reactor.
In about 25 seconds, HPCI will begin injecting into the core. The combination of HPCI and feedwater will be close to 120 percent of rated feedwater flow. Reactor water level will begin to increase.
Core inlet subcooling increases due to extra injection and cooler injection of HPCI flow into the downcomer. Reactor power increases.A-_Stie6ai floW to0 the turbine does not increase1 isignificantly with the power, since the inriease in power is needed to heat the. cooew ter"M0' treate the extra steam required for the RHiCI turbine.'.
The rising water level will cause the feedwater control system to reduce feedwater flow.
Feedwater flow will decrease until feed and HPCI flows combined equals the steam flow. The final steady-state conditions will be:
i1.
Higher reactor power (Turbine power remains about the same) 7
- 2.
Higher vessel water level
- 3.
Lower feedwater flow rate
- 4.
Indicated steam/feed flow mismatch
- 5.
Lower MCPR (due to higher heat flux)
VII.
Inadvertent RHR Shutdown Cooling Operation Since the RHR system is a low pressure system, it could not be operated in the shutdown cooling mode while the reactor is at operating conditions. If the reactor were critical or near critical on a startup or shutdown, misoperation of the shutdown cooling mode of RHR could result in a moderator temperature decrease. This would cause a slow insertion of positive reactivity into the core, causing a slow increase in the fission rate. This flux increase would be controlled by the operator in the same manner normally used to control the fission rate in the source or intermediate range. If for some reason no operator action is taken, the fission rate increase will be terminated by a scram before any fuel damage could occur.
VIII.
Summary A loss of feedwater heating due to a loss of extraction steam to one or more feedwater heaters will result in colder feedwater entering the reactor. The analysis is shown on Figure 9. The increase in core inlet subcooling will cause reactor power to increase. An APRM reactor scram is possible under the most severe conditions. In addition to the higher MWth, generator MW. wili increase due to the increased steam flow through the turbine. Key parameters of interest in the transient include:
- 1.
Reactor power - Increases due to a rise in core inlet subcooling
- 2.
Reactor pressure - Increases because of greater steam generation
- 3.
MCPR - Decrease GAN'TGOP5\\Pending-Draft Fude\\s Q7s~~d\\scOO7d\\SCOO7 D.doc