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{{Adams | |||
| number = ML20149K131 | |||
| issue date = 07/18/1997 | |||
| title = NRC Operator Licensing Exam Rept 50-293/97-06 (Including Completed & Graded Tests) for Tests Administered on 970505-09 | |||
| author name = | |||
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) | |||
| addressee name = | |||
| addressee affiliation = | |||
| docket = 05000293 | |||
| license number = | |||
| contact person = | |||
| document report number = 50-293-97-06, 50-293-97-6, NUDOCS 9707290264 | |||
| package number = ML20149K125 | |||
| document type = EXAMINATION REPORT, TEXT-INSPECTION & AUDIT & I&E CIRCULARS | |||
| page count = 103 | |||
}} | |||
See also: [[see also::IR 05000293/1997006]] | |||
=Text= | |||
{{#Wiki_filter:=. . - - | |||
+ | |||
Enclosure | |||
U.S. NUCLEAR REGULATORY COMMISSION | |||
REGION I | |||
. License No. DPR-35 - | |||
; Report No. 97-006 | |||
Jocket No.- 50-293 . | |||
Licensee: Boston Edison Company | |||
800 Boylston Street | |||
Boston, Massachusetts 02199 | |||
Facility:~ Pilgrim Nuclear Power Station | |||
. | |||
Exemination Period: May. 5 - 9,1997 | |||
Examiners: D. Florek, Seninr Operations Engineer . | |||
C. Sisco, 0,serations Engineer , | |||
S. Dernh., Examiner in Training l | |||
' | |||
S. Willoughby, Contract Examiner | |||
, Approved by: G. Meyer, Chief | |||
Operations and Human Performance Branch ; | |||
Division of Reactor Safety j | |||
1 | |||
i | |||
9707290264 97071g w | |||
" ' | |||
PDR ADOCK 05000293 | |||
V pg , | |||
ma | |||
--7 ; | |||
EXAMINATION SUMMARY | |||
Examination Report 50-293/97-006 (OL) | |||
Initial examinations were administered to six senior reactor operator (SRO) instant | |||
applicants during the period of May 5 -9,1997, at the Pilgrim Nuclear Power Station. | |||
OPERATIONS | |||
Five of six applicants passed the examination. One SRO instant applicant failed the written | |||
and operating portion of the examination. The five applicants that passed were well | |||
prepared for the examinations. The applicants consistently understood and implemented | |||
the emergency operating procedures well. Some weak areas of understanding were | |||
identified during the written exam and operating test. | |||
Two of the applications were found to be deficient in that the applicants had not performed | |||
the five significant control manipulations on the plant as required by 10 CFR 55.31(a)(5). | |||
The applicants' qualification records did not support performance of five significant control- | |||
manipulations. The root ceuse for this problem appeared to be that the BECO program | |||
guidance inappropriately permitted multiple significant control manipulation credit for a | |||
single, extended power change. | |||
l | |||
l | |||
. | |||
; | |||
ii | |||
_ | |||
_ | |||
Details | |||
05.1 Operator initial Examinations | |||
a. Scope | |||
] | |||
The examiners administered initial examinations to six instant SRO applicants in | |||
accordance with NUREG-1021, " Examiner Standards," Revision 7. | |||
I | |||
b. Observations and Findinas | |||
The results of the initial examinations are summarized below: | |||
i | |||
SRO | |||
PASS / Fall | |||
' | |||
j | |||
Written 5/1 | |||
Operating 5/1 | |||
Overall 5/1 | |||
The Boston E-jison Company (BECO) staff reviewed the written examination and | |||
assisted in the validation of the operating examination during the week of | |||
April 21,1996. The BECO staff provided comments on the examination that ) | |||
significantly improved the examination. The BECO staff, who were involved with | |||
the examination review, signed security agreements to ensure that the initial | |||
examinations were not compromised. | |||
In a letter, dated May 16,1997 (see Attachment 2), BECO provided six comments | |||
on the written examination. The NRC accepted two of the six comments. As a | |||
result, one question was deleted from the examination and two correct answers | |||
were accepted in one question. The NRC resolution of facility comments is | |||
summarized in Attachment 3.' | |||
The following summarizes the written examination questions that were missed by at | |||
least three applicants, indicating a weakness in the understanding of the subject. | |||
Ques 3 Knowledge of the normal indication for the core spray line | |||
break detection monitor. | |||
Ques 33 Knowledge of the method to move an MSIV by use of the | |||
MSIV test push-button. | |||
T Ques 36 Knowledge of the air ejector off gas radiation rnonitor signals | |||
that willinitiate the 13-minute timer, | |||
i | |||
. | |||
-- . . . - | |||
, | |||
, | |||
9 | |||
2 | |||
Ques 38 ' Ability to use technical specifications related to inoperable | |||
IRMs.- | |||
Ques 43 Ability to determine procedure' entry to a given set of. , | |||
; | |||
conditions. | |||
I | |||
Ques 61 Knowledge of the number of drifting rods in a nine-rod array | |||
that require placing the mode switch in shutdown. . | |||
- Ques 76 Ability to determine the method and reason for depressurizing | |||
the reactor to a given set of conditions. ! | |||
Ques 85 Knowledge of the method to track the duration of | |||
surveillances. | |||
4 | |||
During the operating test, at least two applicants performed poorly in each of the , | |||
'following areas: ! | |||
Refueling operations | |||
Recognizing a loss of control room annunciators | |||
- The above test items represent areas of weak understanding or performance and are | |||
provided to enable improvement of the training program. | |||
During.the dynamic simulator test, the following item was significant and a | |||
consistent positive observation. j | |||
l | |||
Knowledge and understanding of the emergency operating procedures | |||
(EOPs). | |||
During the development and administration of the examination, the examiners noted | |||
the following item for further BECO consideration of possible procedure | |||
improvements. | |||
Emergency Operating Procedure 5.3.21 page 34 of 58 indicated that the | |||
installation of the jumper in panel C915 from jumper location DD-24 to DD- | |||
25 defeated the high drywell pressure and low RPV levelisolation signals for | |||
MO-47, Shutdown Cooling Outboard Isolation Valve. This jumper also | |||
affected isolation signals for MO-29B LPCIinjection valve. The procedure | |||
did not provide a note that this valve was also affected by installation of the | |||
jumper. | |||
: c. Conclusions j | |||
Five_of six of the applicants were well prepared for the examination, and as a result, j | |||
five applicants passed the examination. One SRO instant applicant failed the l | |||
' | |||
examination. ; | |||
I | |||
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. _. -. . _ __ _ . | |||
, | |||
3 | |||
05.2 Reactivity Manipulations | |||
a. Scope | |||
The inspector reviewed the BECO records to determine how the applicants complied | |||
with 10 CFR 55.31(a)(5). This section of 10 CFR requires that applicants must | |||
perform five significant control manipulations on the plant that affect reactivity or | |||
power level. | |||
b. Findinos | |||
BECO Document 0-RO-04 "NRC Licensed Nuclear Plant RO/SRO initial | |||
Qualification," dated August 1996, required a minimum of five significant reactivity | |||
manipulations with amplification that effort should be made to diversify the | |||
reactivity manipulations. Ten examples were identified for meeting the requirement. | |||
Four of the examples related to 10% power changes with control reds or | |||
recirculation flow. The inspector considered each of the ten examples as an | |||
appropriate significant control manipulation. | |||
Based on review of the individual applicant qualification records, four of the six | |||
applicants had performed five significant cont'ol | |||
r manipulations on the plant, | |||
although one of these four applicants did not have diverse manipulations. | |||
Based on review of the individual applicant qualification records, the NRC examiner | |||
identified on May 5,1997, that two of the six applicants had not performed five | |||
significant control manipulations on the plant. Although the BECO guidance | |||
specified the minimum conditions for a manipulation, the minimum conditions had | |||
inappropriately been used to credit more than one manipulation when a single, | |||
extended power change occurred. For example, one applicant reduced power with | |||
recirculation flow from 100% to 68% over 56 minutes. BECO considered this to be | |||
three of the five significant control manipulations. The NRC staff disagreed with | |||
BECO and considered this to be one significant control manipulation. Another | |||
applicant reduced power from 100% to 50% initially with recirculation flow and | |||
then later with control rods over 77 minutes. BECO considered this to be all five of | |||
the required five significant control manipulations. The NRC staff disagreed with | |||
BECO and considered this to be two significant control manipulations. BECO was | |||
informed of the examiner's conclusion and informed that this would not impact the | |||
administration of the remainder of the examination. The resolution of this issue was | |||
pursued after the examination was administered. | |||
The final applications submitted on April 18,1997, indicated that these two | |||
applicants had performed their five required significant control manipulations. After | |||
the NRC staff review of the supporting data for the application, the .NRC staff | |||
concluded that one applicant had performed two of the five significant control | |||
manipulations and the other applicant had performed three of the five significant | |||
control manipulations. These two applicants did not meet the requirements of 10 | |||
CFR 55.31(a)(5). | |||
m | |||
7 | |||
4 | |||
In discussions with BECO and the NRC on June 4,1997, the NRC reiterated the | |||
NRC position and informed BECO that these two applicants passed the examination | |||
but would not be issued licenses until the applicants and BECO submitted revised | |||
Form NRC-398s after five significant control manipulations were performed on the | |||
plant. BECO acknowledged the NRC staff finding and indicated that they had | |||
initiated actions to have the applicants perform additional significant control | |||
manipulations on the plant after the examination was administered and would | |||
submit revised Form NRC-358s. | |||
BECO submitted revised Form NRC-398s in a letter dated June 1,1997. BECO also | |||
provided the details of how the applicants satisfied the 10 CFR requirement for | |||
significant control manipulations. Based on the revised applications and supporting | |||
data, the NRC subsequently issued licenses for these individuals. | |||
c. Conclusion | |||
The BECO guidance and examples of how to meet the requirements of 10 CFR | |||
55.31(a)(5) were acceptable. However, the BECO practice of giving multiple | |||
significant control manipulation credit for a single, eendad power change was not | |||
acceptable. The examiner concluded that BECO had violated 10 CFR 55.31(a)(5), | |||
which requires that applicants for operator licenses must have performed five | |||
significant control manipulations on the plant that affects affect reactivity or power l | |||
level. With the multiple manipulations removed, one SRO applicant had performed I | |||
two significant control manipulations, and another SRO applicant had performed | |||
three significant control manipulations. (VIO 97-06-01) | |||
E8 Review of UFSAR Commitments | |||
A recent discovery of a licensee operating their facility in a manner contrary to the | |||
updated final safety analysis report (UFSAR) description highlighted the need for a | |||
special focused review that compares plant practices, procedures, and/or | |||
parameters to the UFSAR descriptions. While performing the examination activities | |||
discussed in this report, the examiners reviewed portions of the UFSAR that related | |||
to the selected examination activities, questions or topic areas. The particular | |||
section reviewed was Table 5.2.4. The specific question reviewed was consistent | |||
with the UFSAR. | |||
. _ _ _ - _ . | |||
- | |||
. _ _. _ __ ._. .. | |||
f | |||
5 | |||
V. Manaaement Meetinas | |||
X1 Exit Meeting Summary | |||
At the conclusion of the examination, the examiners discussed their observations of the | |||
examination process with members of BECO management. BECO acknowledged the ! | |||
examiners' observations. The BECO personnel present at the exit included the following: | |||
! | |||
J. Alexander, Training Manager . | |||
' | |||
M. Briggs, Principal Instructor | |||
K. DiCroce, Sr. Regulatory Affairs Engineer | |||
L. Olivier, Vice President Nuclear l | |||
M. Santiago, Operations Training Manager 1 | |||
(T. Sullivan, Plant Manager . | |||
.T, Trepanier, Operations Department Manager ! | |||
T. Venkataraman, QA Group Manager | |||
, | |||
NRC Personnel- | |||
S. Dennis, Operations Engineer l | |||
D. Florek, Sr. Operations Engineer ] | |||
' | |||
R. Laura, Senior Resident inspector | |||
C. Sisco, Operations Engineer | |||
! | |||
Attachments: ! | |||
! | |||
1. SRO Examination and Answer Key | |||
2. Facility Comments on Written Examinations ; | |||
3. NRC Resolution of Facility Comments i | |||
l | |||
4. Simulation Facility Report I | |||
1 | |||
4 | |||
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ATTACHMENT 1 | |||
SRO Examination and Answer Key | |||
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U. S. NUCLEAR REGULATORY COMMISSION | |||
SITE SPECIFIC EXAMINATION | |||
SENIOR OPERATOR LICENSE | |||
REGION 1 | |||
APPLICANT'S NAME: | |||
FACILITY: Pilarim 1 | |||
REACTOR TYPE: BWR-GE3 | |||
DATE ADMINISTERED: May 5,1997 | |||
INSTRUCTIONS TO APPLICANT: | |||
Use the answer sheets provided to document your answers. Staple this cover sheet | |||
on top of the answer sheets. Points for each question are indicated in parentheses | |||
after the question. The passing grade requires a final grade of at least 80%. | |||
Examination papers will be picked up four (4) hours after the examination starts. | |||
TEST VALUE APPLICANT'S SCORE FINAL GRADE % | |||
100.00 | |||
All work done on this examination is my own. I have neither given nor received aid. l | |||
l | |||
l | |||
Applicant's Signature | |||
7:. . - . . . . -. . . . - . . | |||
. .. - - -- . _ . . . | |||
i | |||
SENIOR REACTOR OPERATOR Prga 2 | |||
ANSWER SHEET | |||
Multiple Choice (Circle or X your choice) | |||
. lf you change your answer, write your selection in the blank. | |||
MULTIPLE CHOICE O23 a .b cd __ | |||
001 a b c d 024 a b c d | |||
002 a b'c d _ | |||
O25 a b c d | |||
003 ' a b c d 026 a b c d | |||
004 a bc d 027 a b c d | |||
005 a b c d 028 a b cd | |||
006 a. b c d 029 a b c d ; | |||
007 a b c d 030 a b c d | |||
008 a b c d 031 a b c d | |||
009 a b c d 032 a b c d | |||
010 a b c d 033 a b c d | |||
'011 a b c d 034 a b c d | |||
, . 012 a b c d 035 a b c d | |||
013 a b c d 036 a b c d | |||
1 014 a b c d 037 a b c d | |||
015 a b c d 038 a b c d | |||
016 a b c d 039 a b c d | |||
017 a b c d 040 a b c d | |||
018 a b c d- 041 a b c d | |||
019 - a b c d 042 a b c d | |||
020 a b cd 043 a b c d | |||
021 a b c d 044 a b c d | |||
022 a. b c d 045 a b c d | |||
f | |||
r, | |||
- . . . . . -- . - .- . . .. -. -. . . -- . . . . | |||
SENIOR REACTOR OPERATOR - Pzgn'3 | |||
ANSWER SHEET- | |||
Multiple Choice (Circle or X your choice) | |||
If you change your answer, write your selection in the blank. , | |||
l | |||
1 | |||
046'.a b c d 069 a. b c d l | |||
i | |||
i | |||
047 a'b c d_ 070 a b c-d- ) | |||
048 ab c d 071 a b c d | |||
049 a bc d 072 a b c d ) | |||
050 a b c -d 073 ab c d | |||
i | |||
074 a b c d l | |||
051 a b-c'd | |||
,. | |||
052' a b c d 075 a b c d | |||
I | |||
053 a b c d 076 a b c d i | |||
1 | |||
054 a b c d 077 a b c d | |||
055 a b c d 078 a b c d | |||
056. a b c d 079 a b c d 1 | |||
057 a b c d 080 a b c d | |||
058 a b c d 081 a b c d | |||
- 059' a b c d 082 a b c d | |||
, 060 a b c d 083 a b c d | |||
061 a b c d 084 a b c d | |||
062 a b c d 085 a b c d : | |||
063 a b c d 086 a b c d | |||
064 a b c d 087 a.b c d | |||
065 a b c d 088 a b c d | |||
066 a b~ c d 089 a b c d | |||
067 a-b c d 090 a b cd | |||
068 -a b c d 091 a b c d | |||
I | |||
a | |||
__ | |||
( | |||
SENIOR REACTOR OPERATOR Prg3 4 | |||
ANSWER SHEET | |||
Multiple Choice (Circle or X your choice) | |||
If you change your answer, write your selection in the blank. | |||
. | |||
092 a - b c. d | |||
093 a b c-d _ | |||
094 a b c d _ | |||
095 a b c d _ | |||
096 a b c d _ | |||
097 a b c d _ | |||
098 a b c d _ | |||
099 a b c d _ | |||
100 a b c d _ | |||
l | |||
l | |||
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l | |||
! | |||
l | |||
1 | |||
1 | |||
1 | |||
! | |||
! | |||
( * * * * * * * * * * END OF EX AMIN ATION * * * * * * * * * * ) | |||
l | |||
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- . - . . ~ . . - - -. . - -. . - .-... .- - -.- . . | |||
7.. . | |||
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: | |||
Pagn.5 | |||
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- | |||
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS | |||
During the administration of this examination the following rules apply: | |||
_ | |||
; | |||
1. Cheating on the examination means an automatic denial of your application | |||
and could result in more severe penalties. | |||
. 2. . After the examination has been completed, you must sign the statement on ' | |||
the cover sheet indicating that the work is your own and 'you have not | |||
received of given assistance in completing the examination. This must be | |||
done after you complete the examination. | |||
, | |||
3. Restroom trips are to be limited and only one applicant at a time may leave. | |||
You must avoid all contacts with anyone outside the examination room to | |||
avoid even the appearance or possibility of. cheating. , | |||
! 4. Use black ink or dark pencil ONLY to facilitate legible reproductions. | |||
5. Print your name in the blank provided in the upper right-hand corner of the | |||
examination cover sheet and each answer sheet. | |||
, | |||
< | |||
~ 6. Mark your answers on the answer sheet provided. USE ONLY THE PAPER | |||
> | |||
PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAGE. | |||
' | |||
7. The point value for each' question is indicated in parentheses after the ; | |||
question. | |||
' | |||
,- | |||
t 8. If the intent of a question is unclear, ask questions of the examiner only. | |||
: | |||
9. When turning in your examination, assemble the completed examination with | |||
examination questions, examination aids and answer sheets in addition, | |||
, turn in all scrap paper. | |||
' | |||
10. Ensure allinformation you wish to have evaluated as part of your answer is | |||
on your answer sheet. Scrap paper will be disposed of immediately | |||
; following the examination. | |||
; | |||
i 11. To pass the examination, you must achieve a grade of 80% or greater. | |||
12. There is a time limit of four (4) hours for completion of the examination. | |||
13. When you are done and have turned in your examination, leave the | |||
; examination area (EXAMINER WILL DEFINE THE AREA). If you are found in | |||
this area while the examination is stillin progress, your license may be | |||
i denied 'or revoked. | |||
, | |||
, | |||
A | |||
4 | |||
t' | |||
e | |||
$ | |||
,, - .4 , ...-.a -, ,,m . .._ . ,, - , , m | |||
.. - . ._ . -. | |||
SENIOR REACTOR OPERATOR Pcgs 7 | |||
. QUESTION: 001 (1.00) | |||
The HPCI system has automatically initiated due to a low reactor water | |||
level. Drywell pressure remains within normal limits. The HPCI turbine | |||
slowly lowers reactor pressure. Reactor pressure continues to decrease | |||
and reaches 80 psig. | |||
Which ONE of the following is the expected automatic response? | |||
a. Group IV isolation, but no HPCI turbine trip and no Group Vil | |||
isolation | |||
b. Group IV isolation and HPCI turbine trip, but no Group Vil | |||
isolation | |||
c. Group Vilisolation and HPCI turbine trip, but no Group IV | |||
isolation | |||
; | |||
d. Group IV solation, Group Vil isolation, and HPCI turbine trip | |||
l | |||
I | |||
OUESTION: 002 (1.00) | |||
, | |||
Which ONE of the following signals will NOT require resetting the Trip | |||
and Throttle Valve to restart the RCIC turbine? | |||
! a. RCIC Turbine Mechanical Overspeed | |||
b. Reactor high water level | |||
c. Manual trip pushbutton on 904 panel | |||
; d. High Steam Supply line differential pressure | |||
l | |||
-. - - . _ _ _ _ _ _ _ _ _ _ _ - - _ _ | |||
. -_. ___ _ _ . _ . ._ | |||
- SENIOR REACTOR OPERATOR _ P ga 8 | |||
9 | |||
QUESTION: 003 (1.00) | |||
With the plant operating at 100% power, the 'A' Core Spray Line Break | |||
Detection Monitor is reading approximately -3.0 psid. | |||
This reading is: | |||
, | |||
a. Indicative of an 'A' Core Spray Line break inside the shroud. | |||
b. indicative of an 'A' Core Spray Line break outside the shroud. | |||
: c. normal due to the differential pressure across the dryers and | |||
separators being approximately -3.0 psid at 100% power. | |||
d. normal due to changes in water density after the instrument was | |||
calibrated to read zero under cold conditions. | |||
QUESTION: 004 (1.00) | |||
The following conditions exist: | |||
- SBLC Tank Temperature 45 Degrees F | |||
- | |||
SBLC Tank Volume 4000 gallons | |||
- SBLC Tank Concentration 9.1% weight % | |||
- B-10 Isotope Enrichment 53 % | |||
What is(are) the MINIMUM required action (s) that you as the NWE should | |||
immediately initiate? | |||
a. Perform a SBLC flow test. | |||
b. Determine whether the sodium pentaborate solution meets the | |||
original design criteria. | |||
c. Perform a SBLC flow test and determine whether the sodium | |||
pentaborate solution meets the original design criteria. | |||
d. Immediately commence a plant shutdown such that the plant can | |||
reach cold shutdown within 24 hours. | |||
., _ -. - .- | |||
SENIOR REACTOR OPERATOR Paga 9 | |||
QUESTION: 005 (1.00)- | |||
Given the following conditions: | |||
- The plant is in cold shutdown. | |||
* | |||
- An RHR system test is in progress. | |||
- The LPCI Override Control Switch (S178) has been taken to | |||
MANUAL OVERRIDE. | |||
- Drywell pressure is at O psig and steady. | |||
- As part of the test, reactor water level is simulated at -60 | |||
inches. | |||
- The next step in the procedure is to take the control switch | |||
for the Torus Spray Valve MO-1001-37B to open. | |||
Which ONE of the following explains why MO 1001-37B will NOT | |||
open when the control switch is taken to open? | |||
a. The 15 minute time delay is not timed out. | |||
b. Drywell pressure is at atmospheric. | |||
c. The RPV Level Override Keylock Switch (S188) is not in MANUAL j | |||
OVERRIDE. | |||
d. The 5 minute time delay has not timed out and the MO-1001-28B i | |||
! | |||
is not closed. | |||
l | |||
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, _ _ _ . . __ _ -_ _ _ ._ .. . _ _ . . _ .__ . _ _ _ . _ | |||
1: | |||
SENIOR REACTOR OPERATOR Pegs 10 | |||
l | |||
QUESTION: 006 (1.00) | |||
, | |||
The following conditions exist: | |||
- The plant is operating at 100% power. | |||
- | |||
The "A" SBGT Fan is in AUTO. - | |||
- | |||
The "B" SBGT Fan is in STBY. 3 | |||
- A valid SBGT Initiation signal occurs. | |||
- The "A" SBGT Fan initially starts, runs for 10 seconds, then | |||
trips for an unknown cause. | |||
Which ONE of the following describes the expected automatic response of | |||
the "B" SBGT Fan? | |||
'l | |||
The "B" SBGT Fan will: | |||
a. start when the initiation signal is received, run for 65 | |||
. seconds and then stop, then restart. | |||
b. start immediately after the "A" SBGT Fan trips and continue | |||
! | |||
running uninterrupted | |||
c. start after 65 seconds and continue to run uninterrupted. | |||
d. start when the initiation signal is received and continue | |||
running uninterrupted. j | |||
> | |||
l | |||
l | |||
i | |||
I | |||
w - . . . | |||
_ _ _ _ _ _ . - _ . ._ _ . _ _ _ _ _ ._ _ _ _ _ . _ . _ _- | |||
4 | |||
SENIOR REACTOR OPERATOR Prgs 11 | |||
. | |||
QUESTION: 007 (1.00). , | |||
The following conditions exist: | |||
; | |||
- - | |||
The plant is at 100% power. | |||
- it is determined that the 3A SRV will NOT open under ANY | |||
' | |||
condition. | |||
- | |||
Which ONE 'of the following states the MINIMUM action REQUIRED by | |||
' | |||
Technical Specifications? | |||
; | |||
a. Place the plant in Cold Shutdown within 24 hours. | |||
b. Reduce reactor coolant pressure below 104 psig within 24 hours. | |||
c. Provided HPCI is operable, enter 14 day LCO. When this LCO is | |||
expired, place the plant in Cold Shutdown within 24 hours. | |||
d. Provided HPCI is operable, enter 14 day LCO. When this LCO is | |||
expired, reduce reactor coolant pressure below 104 psig within | |||
' | |||
24 hours. | |||
. | |||
QUESTION: 008 (1.00) | |||
During a high drywell pressure condition, a valid ADS signal exists and | |||
. | |||
. the ADS system has initiated. With the initiation signal still present | |||
- | |||
both initiation Signal Timer Reset Pushbuttons are depressed. | |||
Which ONE of the following describes the expected automatic response of | |||
the ADS system? | |||
j All ADS valves will: | |||
a. remain open. | |||
b. close and remain closed indefinitely. | |||
c. close and remain closed for 105 seconds then reopen. | |||
d. close and remain closed for 11 minutes then reopen. | |||
. | |||
1 | |||
= | |||
_. _ . . - _. .. . . . _ . _ _ __ . _ _ . _ _ _ _ _ _ __ _ | |||
. | |||
SENIOR REACTOR ' OPERATOR P:ga 12 | |||
_ | |||
4 | |||
^ | |||
QUESTION: OO9 (1.00) | |||
The following conditions exist: | |||
- The "A" and "B" Reactor Feed Pumps are in service. | |||
- - Both Reactor Recirculation Speed demands are at 60%. - | |||
. | |||
- An instrument failure causes the Feedwater Regulating Valves to | |||
reduce feedwater flow to 3 Mlbm/hr with reactor water level | |||
reaching a minimum of + 17 inches. | |||
4 | |||
- The operator takes manual control of the Feedwater Regulating | |||
Valves and is returning water level to normal with a current | |||
' | |||
level of + 18 inches. | |||
' | |||
- No operator action is taken on the Reactor Recirculation | |||
System. | |||
Which ONE of the following describes the expected response of the l | |||
. Recirculation Flow Controllers? ' | |||
'The Recirculation Flow controllers will demand lowering speed to | |||
a. 44% without a rate limitation signal. | |||
! | |||
.b. 44% at a rate of 1.5% per second. | |||
c. 26% without a rate limitation signal. | |||
d. 26% at a rate of 1.5% per second. | |||
'l | |||
' QUESTION: 010 (1.00) | |||
An Emergency Diesel Generator (EDG) has started due to a LOCA signal. | |||
Which ONE of the following signals will cause an EDG trip? | |||
a. Engine Overspeed | |||
i | |||
b. Engine Low Lube Oil Pressure | |||
c. Engine High Lube Oil Temperature | |||
, | |||
d. Engine Crankcase High Vacuum | |||
- . _ | |||
. ,- | |||
- - - .-- . .. . -. -..-. | |||
SENIOR REACTOR OPERATOR: ' Piga 13 ) | |||
L | |||
l | |||
QUESTION: 011 (1.00) l | |||
l | |||
The following conditions exist: | |||
- HPCI is injecting water from the CST to the RPV. l | |||
- The HPCI Suction Valves From Suppression Chamber MO-2301-35 and- | |||
MO-2301-36 Control Switches are in Auto. | |||
- The CST Low Level Alarm comes in. | |||
1 | |||
Which ONE of the following describes the expected response of the HPCI | |||
system? | |||
The HPCI Suction From CST MO-23016 will receive a close signal- | |||
) | |||
a. as soon as both the MO-2301-35 and the MO-2301-36 valves reach | |||
. full open. | |||
- b. as soon as both the MO-2301-35 and MO-2301-36 valves come off | |||
their closed seats, | |||
c. as soon as either the MO-2301-35 or the MO-2301-36 valve | |||
reaches full open. l | |||
l | |||
d. at the same time the MO-2301-35 and MO 2301-36 valves receive ; | |||
an open signal. ) | |||
I | |||
QUESTION: 012 (1.00) | |||
Which ONE of the following states where the RCIC turbine receives stearn | |||
and where the RCIC pump discharges? | |||
a. Steam from "C" Main Steam Line and Discharge to "A" Feedwater | |||
: Line | |||
b. Steam from "D" Main Steam Line and Discharge to "B" Feedwater | |||
: Line | |||
c. Steam from "D" Main Steam Line and Discharge to "A" Feedwater | |||
Line | |||
' | |||
d. Steam from *C" Main Steam Line and Discharge to "B" Feedwater | |||
Line | |||
i | |||
1 | |||
l | |||
__ _ _ _ | |||
- .. ._ . . . . . . _ _ . . _ __ . _ . _ _ _ _ . | |||
Pegs 14 | |||
l SENIOR REACTOR OPERATOR | |||
I | |||
OUESTION: 013 (1.00) | |||
I | |||
' | |||
Which of the,following SRM rod block (s) is(are) bypassed by moving IRM | |||
range switches from Range 2 tc Range 3? l | |||
a. SRM Downscale Rod Block only - | |||
b. SRM Inoperable Rod Block only ; | |||
I | |||
c. SRM Downscale Rod Block and Detector Retract Not Permitted Rod | |||
Block | |||
d. SRM inoperable Rod Block and SRM Downscale Rod Block | |||
u | |||
QUESTION: 014 (1.00) , | |||
With Reactor Power at'100%, an SRV spuriously lifts. Action to close ! | |||
the valve are successful. Immediately after valve closure, the j | |||
downstream temperature is checked. | |||
Which ONE of the following is an expected approximate downstream | |||
temperature? | |||
a. 212 degrees F | |||
b. 295 degrees F.. | |||
c. 375 degrees F .l | |||
d. 525 degrees F | |||
i | |||
_ . | |||
... -. . . . _ . | |||
SENIOR REACTOR OPERATOR P gs 15 | |||
-QUESTION: 015 (1.00) | |||
The following conditions exist: | |||
- A half scram exists on RPS "A" due to APRM testing. | |||
- A fire caused a loss of RPS Bus "B" and a full scram. | |||
' | |||
- The half scram testing was stopped and APRMs were returned to | |||
normal.~ | |||
- The SCRAM DISCHARGE INSTRUMENT VOLUME HI LEVEL SCRAM BYPASS | |||
switch is then taken to bypass. | |||
Which ONE of the following describes when the RPS "A" half scram may be | |||
reset? | |||
a. immediately. | |||
b. after the air dump test switch is placed in isolate. | |||
c. after the SDIV vent and drain valves come fully open. | |||
d. after RPS "B" is energized. | |||
! | |||
l | |||
QU' - ION: 016 (1.00) | |||
The mode switch is in RUN. Which ONE of the following scram signals is | |||
automatically bypassed 2 seconds after taking the mode switch to | |||
SHUTDOWN? | |||
, i | |||
a. Mode switch in shutdown | |||
b. Main steam isolation valve closura | |||
c. Turbine stop valve closure | |||
d. Scram discharge volume high level | |||
. . - -. . . . , . .. . . . -. - . . , . . . - - | |||
' | |||
SENIOR REACTOR OPERATOR. Pags 16 | |||
: | |||
QUESTION: 017 (1.00). | |||
. The ATWS logic system has automatically initiated due to low reactor | |||
water level. | |||
~ | |||
Which ONE of the following actuations will be delayed by 9 seconds? | |||
a. Rod insertion | |||
b. Reactor Recirc Pump Field Breaker Trip | |||
' | |||
c. Reactor Recirc Pump Drive Motor Breaker Trip | |||
d. Reactor Feed Pump Trip | |||
{ | |||
QUESTION: 018.(1.00) | |||
: | |||
' | |||
. During a reactor shutdown, the control rod selected on the Rod Select | |||
Matrix is NOT in the rod group of the latched step. As reactor power | |||
' decreases, at what point will this condition cause an insert and | |||
withdraw block? | |||
a. Steam Flow drops below 35% | |||
b. All APRM readings drop below 20% | |||
c. Steam Flow or Feed Flow drops below 20% | |||
d. Steam Flow and Feed Flow drop below 35% | |||
, | |||
9 | |||
J | |||
' | |||
i | |||
SENIOR REACTOR OPERATOR P:gs 17 | |||
QUESTION: 019 (1.00) | |||
The following conditions exist: | |||
- The plant is operating at 100% power. | |||
- APRM "C".is bypassed for maintenance. | |||
- APRM "E" then fails giving a constant reading of 95% regardless | |||
of input. | |||
- A half scram already exists on RPS "B" | |||
Which ONE of the following meets the action REQUIRED? | |||
a. Initiate insertion of operable rods and complete insertion of | |||
all operable rods within sixteen hours. | |||
b. Reduce power level to IRM range and place mode switch in the | |||
startup/ hot stand'y position within eight hours. | |||
c. Reduce turbine load and close main steam isolation valves | |||
within eight hours, | |||
d. Reduce power to less than 45% of design. | |||
QUESTION: 020 (1.00) | |||
A TIP trace is being performed when a high drywell pressuro signal | |||
occurs. Select the expected automatic action. | |||
a. The shear valve fires with the detector stillin the core. | |||
b. The ball valve closes with the detector stillin the core. | |||
c. The detector withdraws into its shield and the ball valve | |||
closes, | |||
d. The detector withdraws into its shield and the shear valve | |||
fires. | |||
, | |||
e- A ,m. asa _ -ire.d e 4 # ww .e4.1 - ,a 4 4 +4 - | |||
4- beM iJ---'a- | |||
, | |||
SENIOR REACTOR OPERATOR Pi:gs 18 | |||
. | |||
QUESTION: 021 (1.00) | |||
. | |||
' | |||
While operating at 80% power, an instrument failure causes the throttle | |||
pressure sensed by the EPR to fail high. No operator action is taken. | |||
Which ONE of the following is the expected result? i | |||
a. The reactor would scram on a high pressure scram signal. t | |||
b. The MPR would take control and pressure would increase by | |||
approximately 10 psi. | |||
c. The reactor would scram on a low pressure scram signal. | |||
' | |||
d. The reactor would scram on a MSIV closure scram signal. | |||
QUESTION: 022 (1.00) | |||
At 500 psig during a reactor startup and heatup, the #1 Bypass Valve | |||
(BPV) comes partially open. 1 | |||
i | |||
' | |||
Which ONE of the following errors is the cause? | |||
Failure to maintain the: | |||
, | |||
l | |||
a. EPR 40-80 psig below reactor pressure l | |||
l | |||
b. EPR 40-80 psig above reactor pressure ; | |||
I | |||
c. MPR 40-80 psig below reactor pressure | |||
d. MPR 40-80 psig above reactor pressure | |||
! | |||
.i | |||
l | |||
1 | |||
I | |||
. . . . - | |||
SENIOR REACTOR OPERATOR P:ga 19 | |||
- | |||
QUESTION: 023 (1.00)' | |||
With the plant operating at 20% power both Reactor Recirculation Pumps | |||
irip. The operator manually scram the reactor. Post scram, fuel zone | |||
level indicators read: | |||
a. falsely high since less flow exists through the jet pumps than | |||
existed during calibration conditions. | |||
, b. falsely low since less flow exists through the jet pumps than ; | |||
existed during calibration conditions. | |||
c. falsely low due to decreased density of the water in the vessel | |||
against calibrated conditions, | |||
d. falsely high due to decreased density of the water in the | |||
vessel against calibrated conditions. | |||
QUESTION: 024 (1.00) | |||
1 | |||
With the "A" Loop of RHR in Lo oling, RPV level decreased to 12 l | |||
inches. The Shutdown Cooling Outtsumo I:,ulation Valve MO-1001-47 l | |||
stopped in mid-stroke. All other valves have responded as expected. i | |||
! | |||
Which ONE of the following is REQUIRED in order to open the "A" Loop ! | |||
LPCI Injection Valve #2 MO-1001-29A? | |||
a. The MO-1001-47 valve must be closed, | |||
b. The MO-1001-29A must be manually reset, | |||
c. Reactor coolant pressure must be greater than 76 psig, | |||
d. The Group ll isolation signal must clear and the Group 11 logic l | |||
must be reset. | |||
I | |||
l | |||
l | |||
l | |||
- | |||
' | |||
. | |||
l | |||
l | |||
1 | |||
-. . . - . - - | |||
SENIOR REACTOR OPERATOR PJgn 20 | |||
QUESTION: 025 (1.00) | |||
A reactor scram has occurred. Electrical busses A-5 and A-6 have | |||
transferred to the Start-up Transformer. Which ONE of the following | |||
-describes the drywell cooler response? | |||
a. The running drywell coolers will trip and start after a 45 | |||
second time delay. The drywell coolers in standby remain in | |||
standby, | |||
b. The running drywell coolers will trip. The drywell coolers in | |||
standby will start after a 45 second time delay. | |||
c. The running coolers will stay in service. The drywell coolers | |||
in standby willimmediately start when A-5 and A-6 are | |||
reenergized. | |||
d. The running coolers will stay in service. The drywell coolers | |||
in standby will start after a 45 second time delay. | |||
! | |||
QUESTION: 026 (1.00) l | |||
Primary Coolant Temperature is 245 degrees F when Shutdown Cooling is | |||
placed in service, immediately thereafter, a fire disables the Shutdown | |||
Cooling Outboard Isolation Valve MO-1001-47 motor operator. The valve 1 | |||
is in the open position. I | |||
Which ONE of the following meets the MINIMUM REQUIRED action? | |||
a. Verify the ability to manually close the MO-1001-47 valve, then | |||
reestablish shutdown cooling. ) | |||
) | |||
b. Verify the ability to close the MO-1001-50 valve, then | |||
reestablish shutdown cooling, | |||
c. Close either the MO-1001-47 or MO-1001-50 valve and open the | |||
respective breaker. | |||
d. Station an operator to manually close the MO-1001-50 valve if | |||
required and continue in shutdown cooling. | |||
,._ | |||
, | |||
1 | |||
l | |||
l | |||
SENIOR REACTOR OPERATOR Paga 21 | |||
QUESTION: 027 (1.00) | |||
l | |||
When valving in a CRD hydraulic control accumulator, the 305-102 | |||
'(5Nithdraw Riser Isolation Valve) and the 305-112 (Scram Discharge Riser | |||
isolation Valve) are required to be open prior to opening the 305-101 1 | |||
l | |||
(Insert Riser isolation Valve). This prevents- | |||
a. .a single rod scram when opening the 305-101 valvo. I | |||
b. excessive scram time of that rod in the event of a reactor | |||
scram. l | |||
! | |||
c. damage to the accumulator in the event of a reactor scram. I | |||
d. damage to the drive mechanism in the event of a reactor scram. 1 | |||
1 | |||
) | |||
Il | |||
l | |||
QUESTION: 028 (1.00) l | |||
l | |||
While operating at 100% power, a control rod is determined to be i | |||
uncoupled. Attempts to couple the rod have been unsuccessful. l | |||
l | |||
Which ONE of the following states the MINIMUM REQUIRED actions? | |||
a. Verify that the control rod can be moved with drive pressure | |||
and maintain the control rod at the target position. | |||
b. Fully insert the control rod and hydraulically disarm the CRD. | |||
l | |||
' | |||
c. Fully insert the control rod and electrically disarm the l | |||
directional control valves. l | |||
l | |||
d. Fully insert the control rod, electrically disarm the ) | |||
directional control valves and then declare the rod inoperable. i | |||
! | |||
l | |||
1 | |||
I | |||
, SENIOR REACTOR OPERATOR Pzga 22 i | |||
QUESTION: 029-(1.00) | |||
With the plant at pnwer, it is determined that the MO-1001-37B (B Loop | |||
Torus Spray) and MO-1400-25A (A Loop Core Spray Inboard injection) | |||
valves have failed their operability test. Both volves are currently | |||
closed. | |||
- | |||
The maximum time allowed before the plant must be in COLD SHUTDOWN is: | |||
a 24 hours (1 day). | |||
96 hours (4 days), | |||
c. ~ 168 hours (7 days).- | |||
d.192 hours (8 days). | |||
QUESTION: 030 (1.00) | |||
A tagout, which has been in effect on the "A" Reactor Recirculation Pump | |||
for 7 days, has just been cleared. The "A" Reactor Recirculation Pump | |||
is started and immediately manually tripped. On the second start | |||
p attempt, the pump starts and runs for 10 minutes and then is manually | |||
, | |||
tripped. | |||
When is the SOONEST that another start of the "A" Reactor Recirculation | |||
' | |||
Pump may be attempted? | |||
a. Immediately | |||
b.15 minutes after the second trip | |||
c. 45 minutes after the second trip | |||
d. 4 hours after the second trip | |||
i | |||
! | |||
SENIOR REACTOR OPERATOR Pags 23 | |||
I | |||
QUESTION: 031 (1.00) | |||
l | |||
The plant is operating at 100% power when the "B" Reactor Recirculation i | |||
Pump trips. No operator action is taken. | |||
Which ONE of the following describes the initial steady state to final | |||
steady state change in the "A" Reactor Recirculation Loop Jet pump flow | |||
and the reason for the change? | |||
The "A" Reactor Recirculation Loop Jet pump flow will: l | |||
l | |||
a. increase due to lower core pressure drop. I | |||
b. increase due to decreased core voiding. l | |||
l | |||
c. decrease due to higher core pressure drop. | |||
d. decrease due to increased core voiding. | |||
1 | |||
QUESTION: 03 (1.00) | |||
While operating at 0% power, it is determined that th Main Steam Line | |||
High Flow switches o the "B" Main Steam Line will N trip under a high | |||
flow condition. | |||
Which ONE of the following the MINIMUM RE IRED action? r | |||
a. Direct l&C personnel to m ually trip t inop blejpi Eiles. f | |||
b. Direct l&C personnel to manu ly i ert a ha f | |||
isolation on the "B" Group 1 Cha 1. (j p s | |||
c. Initiate an orderly shutdown db in Cold Shutdown Condition | |||
within a MAXIMUM of 30 h urs afte he instrument failure. | |||
d. Initiate an orderly shutd n and have th Main Steam Lines | |||
isolated within a MAXI UM of 10 hours a r the instrument | |||
failure. | |||
; | |||
. . . - .-. - . . . - . - - - . ~ . . . . | |||
SENIOR REACTOR OPERATOR. P:ge 24 | |||
QUESTION: 033-(1.00) | |||
Depressing a.n outboard MSIV test pushbutton will: | |||
a. energize the AC test valve and vent air from the underside of | |||
the piston. | |||
b. energize the AC test valve and admit air to the underside of | |||
the piston. | |||
, | |||
c. deenergize the AC test valve and vent air from the underside of | |||
the piston. - | |||
' d. deenergize the AC test valve and admit air to the underside of r | |||
the piston. | |||
QUESTION: 034 (1.00) | |||
Which ONE of the following conditions requires Rod Block Monitor | |||
Operability? | |||
a. MCPR is 1.35 and Reactor Power is 25%. | |||
b. MCPR is 1.45 and Reactor Power is 75%. | |||
c. MCPR is 1.55 and Reactor Power is 95%. | |||
d. MCPR is 1.60 and Reactor Power is 100%. | |||
t | |||
w- | |||
. SENIOR REACTOR OPERATOR P g) 25 | |||
QUESTION: 035 (1.00) | |||
A loss of 120V Bus A (Y-3) occurs. | |||
Which ONE of the following describes the effect on the RWCU system? | |||
; | |||
a. Half of the logic for closing the MO-2 and MO-5 valves is made | |||
up. | |||
b. MO-2 goes closed. As soon as MO-2 comes off the open seat, the | |||
operating RWCU pump (s) will trip. MO-5 remains open. | |||
c. MO 2 goes closed. As soon as MO 2 comes off the open seat, the , | |||
operating RWCU pump (s) will trip and MO-5 will go closed. j | |||
d. MO-5 goes closed. As soon as MO-5 comes off the open seat, the | |||
operating RWCU pump (s) will trip. MO-2 remains open. | |||
QUESTION: 036 (1.00) | |||
The OFF GAS ISOL CH PRM SEL switch is in position 2. Which ONE of the | |||
following conditions of the Air Ejector Off Gas Radiation Monitors will 4 | |||
cause tiie 13 minute timer to initiate? | |||
a. Hi radiation signal on both channels | |||
b. Hi Hi radiation signal on one channel | |||
c. Hi radiation signal on one channel and Downscale Trip on the ! | |||
other channel | |||
] | |||
d. Downscale trip on one channel and inop trip on the other | |||
channel | |||
1 | |||
_. . - | |||
_. _ . . _ | |||
, | |||
l | |||
l- SENIOR REACTOR OPERATOR P ga 26 | |||
! | |||
QUESTION: 037 (1.00) | |||
l | |||
A reactor startup is in progress with reactor power in the intermediate | |||
range. IRM "A" then starts to intermittently swing upscale and then | |||
downscale. | |||
Which ONE of the following conditions on IRM "A" will cause a Rod Block | |||
but NOT cause a Half Scram? | |||
The IRM reads: | |||
a.1 (on the 0-40 scale) while on range 1. * | |||
, | |||
b. 3 (on the 0-40 scale) while on range 3. i | |||
1 | |||
I | |||
c. 36 (on the 0-40 scale) while on range 5. | |||
d. 39 (on the 0-40 scale) while on range 7. ! | |||
QUESTION: 038 (1.00) | |||
With the Mode Switch in Startup, at 1200 on 5/5/97, the Downscale Trips | |||
for IRM Channels "A", "B", and "E" are made inoperable. ] | |||
Which ONE of the following is the LATEST that one of these channels must | |||
be placed in a tripped condition? | |||
a. 1300 on 5/5/97 i | |||
b.1200 on 5/6/97 | |||
c.1200 on 5/12/97 | |||
d. 1300 on 5/12/97 | |||
, | |||
- - | |||
l | |||
SENIOR REACTOR OPERATOR PIga 27 ' | |||
OUESTION: 039 (1.00) | |||
The plant was. operating at 100% power with the "B" CRD pump in service. | |||
Subsequently, a valid LOCA signal generated a scram. The plant | |||
responded as expected except, the startup transformer feeder breaker | |||
to bus A 5 failed to close. A-5 has been automatically energized from the | |||
shutdown transformer. | |||
Which ONE of the following describes the status / availability of the CRD pumps? | |||
a. "B" CRD pump is running. | |||
"A" CRD pump can be started since no load shed signal was | |||
generated. | |||
b. "B" CRD pump is not running. | |||
"A" and "B" CRD pumps cannot be started due to load shed | |||
signal. | |||
c. "B" CRD pump is not running. | |||
"A" and "B" CRD pumps can be started since no load shed signal | |||
was generated. | |||
, | |||
d. "B" CRD pump is running. | |||
"A" CRD pump cannot be started due to load shed signal. | |||
l | |||
1 | |||
l | |||
l | |||
. . . -. - - - . . . . | |||
' | |||
- SENIOR REACTOR OPERATOR Pign 28 | |||
, | |||
QUESTION: 040-(1.00)- | |||
Which ONE of the following administrative precautions related to valves | |||
are required when lining up RHR "A" loop for shutdown cooling? | |||
a. MO-1001-7A "RHR PUMP A TORUS SUCTION" red tag closed. | |||
MO-1001-7C "RHR PUMP C TORUS SUCTION" red tag closed. , | |||
MO-1001-43A "RHR PUMP A SHUTDOWN COOLING SUCTION yellow tag | |||
closed. | |||
MO-1001-43C "RHR PUMP C SHUTDOWN COOLING SUCTION yellow | |||
tag closed. | |||
b. MO-1001-7A "RHR PUMP A TORUS SUCTION" red tag closed. | |||
MO-1001-7C."RHR PUMP C TORUS SUCTION" red tag closed. | |||
MO-1001-438 "RHR PUMP B SHUTDOWN COOLING SUCTION red tag | |||
closed. | |||
MO-1001-43D "RHR PUMP D SHUTDOWN COOLING SUCTION red tag | |||
closed. | |||
c.1001-6A "RHR PUMP C SUCTION VALVE FROM THE TORUS" red tag | |||
closed. | |||
1001-366A "RHR PUMP A SUCTION VALVE FROM THE TORUS" red tag | |||
closed. | |||
MO-1001-43B "RHR PUMP B SHUTDOWN COOLING SUCTION red tag | |||
closed. | |||
MO-100143D "RHR PUMP D SHUTDOWN COOLING SUCTION red tag | |||
closed. | |||
d.1001-6A "RHR PUMP C SUCTION VALVE FROM THE TORUS" yellow tag | |||
closed. | |||
1001-366A "RHR PUMP A SUCTION VALVE FROM THE TORUS" yellow | |||
tag closed. | |||
MO-1001-43A "RHR PUMP A SHUTDOWN COOLING SUCTION yellow tag | |||
closed. ; | |||
MO 1001-43C "RHR PUMP C SHUTDOWN COOLING SUCTION yellow tag | |||
closed. | |||
_ _ | |||
.. _ _ , | |||
_- _ | |||
SENIOR REACTOR OPERATOR Pags 29 | |||
1 QUESTION: 041 (1.00) | |||
_ | |||
Given the following conditions: | |||
. | |||
The plant is in cold shutdown | |||
No recirculation pumps are in service | |||
RHR pump "A"is in shutdown cooling | |||
RWCU is in service | |||
Reactor shutdown level instrument indicates 40 inches | |||
Which ONE of the fo:iowing describes reactor coolant temperature | |||
indication if the "A" RHR pump trips. Assume no operator action, | |||
a. Recirc loop "A" temperature indicator is representative of | |||
reactor coolant temperature. | |||
b. Recirc loop "B" temperature indicator is representative of | |||
reactor coolant tempesture. | |||
c. RWCU bottom head drain temperature indicator is representative | |||
of reactor coolant temperature. | |||
d. No temperature indicator is representative of reactor coolant | |||
temperature, | |||
, | |||
i | |||
d | |||
-. . .-- _ . . . . . - . _. -- - | |||
, | |||
SENIOR REACTOR OPERATOR Prgs 30 | |||
. | |||
- QUESTION: 042 (1.00) | |||
.The following conditions exist: | |||
!' - EOP-02 is being executed _. . | |||
- | |||
- | |||
The Mode Switch is in Shutdown and ARI has been initia'ted. | |||
' | |||
- The MSIVs are closed. | |||
- Reactor power is 2.5% and no boron has been injected. | |||
- ~ Alternate _Depressurization is required by EOP-04. | |||
- Four SRVs can be opened. | |||
Which ONE of the following actions should be taken to control reactor ' | |||
I water level? | |||
4 | |||
a. Secure all sources of injection. When pressure decreases below | |||
200 psig, slowly inject with LPCI. | |||
b. Secure all sources of injection. When pressure decreases below | |||
:. 400 psig, slowly inject with the Condensate Pumps. | |||
c. Secure all sources of injection except CRD and RCIC. When | |||
, pressure decreases below 200 psig, continue injection flow rate | |||
with RCIC and CRD. | |||
d. Secure all sources of injection except CRD and RCIC. When | |||
pressure decreases below 270 psig, slowly inject with the | |||
. | |||
Condensate Pumps. | |||
: | |||
: | |||
4 | |||
; | |||
l | |||
.. | |||
0 | |||
. . - .. . . . --._. . . . . - . . ~ . _ . - - - - - . . . . - . | |||
1 | |||
i | |||
SENIOR REACTOR OPERATOR Pcgs.31 i | |||
, | |||
l | |||
.) | |||
QUESTION: 043 (1.00) | |||
The following conditions exist: l | |||
- A manual' scram was inserted from 20% power. l | |||
- No other scram signals exist. ! | |||
l | |||
- Reactor power is on intermediate range 6 and decreasing. | |||
- Three control rods are at position 06. All other rods are , | |||
fully inserted. | |||
Which ONE of the following is the required action? | |||
a.' ' enter PNPS 2.1.6. No EOP entry is required. | |||
b. ' enter EOP-01, then exit EOP-01 and enter EOP-02 at R-1. , | |||
; | |||
c. enter PNPS 2.1.6, " Reactor Scram", then exit PNPS 2.1.6 and | |||
. | |||
enter EOP-02 at R 1. | |||
d. enter PNPS 2.1.6, " Reactor Scram", then enter EOP-02 and | |||
execute concurrently with PNPS 2.1.6. | |||
i | |||
l | |||
: | |||
. . | |||
l | |||
i | |||
4 | |||
i | |||
1 | |||
l | |||
I | |||
1 | |||
: | |||
: | |||
: | |||
l | |||
; | |||
I | |||
i | |||
! | |||
- , . .- ., - - . . ., -- | |||
< SENIOR REACTOR OPERATOR pig 3 32- | |||
s | |||
QUESTION: 044 (1.00) | |||
The following conditions exist: | |||
- EOP-02 is being executed. | |||
-- Boron is being injected with the SBLC system. ' | |||
- Initial SBLC tank level was 4100 gallons. ; | |||
Reactor Water Levelis being lowered to reduce reactor power. ; | |||
- Current SBLC tank level is 3000 gallons. | |||
- Torus water temperature is 112 degrees F. | |||
- Reactor water level is'-100 inches. | |||
Which ONE of the following actions is REQUIRED 7 | |||
a. Reise reactor water level to the +12 to +45 inch band and | |||
perform Alternate Depressurization. | |||
' | |||
b. Raise reactor water level to the + 12 to +45 inch band. Do not | |||
perform Alternate Depressurization. ! | |||
! | |||
c. Maintain reactor water level at its current value and perform | |||
. Alternate Depressurization. | |||
d. Maintain reactor water level at its current value. Do not | |||
perform Alternate Depressurization. | |||
QUESTION: 045 (1.00) { | |||
l | |||
While operating at 100% reactor power, reactor pressure starts to | |||
oscillate approximately 10 psi peak to peak and pressure control is | |||
shifting alternately from the EPR to the MPR and back to the EPR. | |||
Which ONE of the following actions are REQUIRED? | |||
a. DJace the EPR control switch to off, | |||
b.' Reduce reactor power to approximately 75%. | |||
. c. Raise the MPR setpoint to prevent pressure control from | |||
. | |||
swapping between regulators. | |||
, | |||
d. Lower.the MPR setpoint to allow the MPR to take control of | |||
pressure. | |||
. . . -. . .- . | |||
. . - - . _ -. - - - -_ - -. - - . .. | |||
SENIOR REACTOR OPERATOR Pcga 33 | |||
, | |||
t | |||
QUESTION: 046 (1.00) | |||
With the plant in Cold Shutdown, some solvent that is improperly stored | |||
in a Control Room locker ignites. The Nuclear Watch Engineer makes the | |||
decision to evacuate the Control Room and to call for off-site | |||
assistance to put out the fire. Control is established at remote - | |||
shutdown stations 20 minutes after the Control Room evacuation. | |||
What is the MINIMUM event level classification? | |||
a. Unusual Event | |||
b. Alert | |||
' | |||
c. Site Area Emergency | |||
d. General Emergency | |||
, | |||
QUESTION: 047 (1.00) | |||
A LOCA has occurred. Which ONE of the following REQUIRES exiting the | |||
RPV level control leg of EOP-017 | |||
a. Reactor water level is -165 inches and increasing with Reactor | |||
Pressure at 200 psig. | |||
b. Reactor water level is -125 inches and decreasing with Reactor | |||
Pressure at 175 psig. | |||
c. Reactor water level is -125 inches and increasing with Reactor | |||
Pressure at 100 psig. | |||
d. Reactor water level is -125 inches and decreasing vdth Reactor | |||
Pressure at 75 psig. | |||
,. | |||
, | |||
. -. . - .. -.. - - _.. .. . - .. - - . - . . ~ . - - - . | |||
- SENIOR REACTOR OPERATOR P gs 34 | |||
; | |||
QUESTION: 048 (1.00) | |||
~ Which ONE of the following conditions REQUIRES Alternate Reactor- , | |||
Pressure Vessel Depressurization assuming a primary system is ' , | |||
dischaiging into secondary containment? | |||
. | |||
a. RCIC torus piping area temperature is 300 degrees F and RCIC | |||
' | |||
, | |||
turbine area temperature is 195 degrees F. | |||
b. HPCI compartment water levelis 8 inches and HPCI turbine area , | |||
temperature is 195 degrees F. .: | |||
c. RHR "B" and "D" pump area temperature is 300 degrees F and RHR | |||
."A" and "C" pump area temperature is 195 degrees F. | |||
, d. Main Steam Tunnel area temperature is 300 degrees F and RHR "A" | |||
and "C" pump area temperature is 220 degrees F. | |||
, | |||
QUESTION: 049 (1.00) | |||
Following a Nitrogen Line leak in the drywell, AO-4356 (Nitrogen / Air | |||
Isolation Valve to the Drywell) was closed. By calculation, how many | |||
times over the next eight hours can each SRV be actuated? | |||
NOTE: COUNT EACH OPEN AND CLOSE CYCLE AS ONE ACTUATION | |||
a. 5 , | |||
b. 10 | |||
c. 20 | |||
4 | |||
d. 40 | |||
. | |||
J | |||
: | |||
. | |||
,a ,. - y a- | |||
- - - . - -- .. .. .-- | |||
SENIOR REACTOR OPERATOR P ga 35 | |||
QUESTION: 050 (1.00) | |||
in the event that torus water level cannot be maintained above 95 | |||
inches, HPCl is secured in order to prevent: | |||
a. exceeding the Primary Containment Pressure Limit. ' | |||
b. exceeding the Pressure Suppression Pressure. | |||
c. exceeding the Heat Capacity Temperature Limit, | |||
d. isolating HPCI on high exhaust pressure. | |||
' QUESTION: 051 (1.00) | |||
The following conditions exist: | |||
- Reactor pressure is 10 psig. | |||
- Drywell pressure is 4 psig. | |||
- ' Torus bottom pressure is 15.2 psig. | |||
- | |||
Torus water level is 303 inches. | |||
Select the correct action and its reason. | |||
. | |||
Under these conditions: | |||
a. Alternate RPV Depressurization is required to prevent SRV Tail | |||
Pipe failure, | |||
b. Suppression Chamber Spray initiation is required using enly | |||
those RHR pumps not required to provide adequate co cooling. | |||
c. Suppression Chamber Spray initiation is not allowed sinc. the | |||
Torus Spray Sparger is covered. | |||
d. Suppression Chamber Spray initiation is not ai. owed since the | |||
Torus-Drywell Vacuum breakers are covered. | |||
; | |||
. | |||
, | |||
' SENIOR REAC f0R OPERATOR P;gs 36 | |||
QUESTION: 052 (1.00) | |||
A trip of the "A" Reactor Recirculation Pump has occurred. The plant is | |||
operating in Region ll of the Power / Flow Map after the immediate actions | |||
2 | |||
of 2.4.17 have been completed. | |||
Which ONE of the following is REQUIRED 7 | |||
Exit Region 11 by: | |||
a. manually scramming the reactor. | |||
b. restarting the "A" Reactor Recirculation Pump. | |||
c. increasing the speed of the "B" Reactor Recirculation Pump. | |||
d. inserting control rods in reverse order of the pull sheet. | |||
l | |||
QUESTION: 053 (1.00) ! | |||
l | |||
The following conditions exist: l | |||
l | |||
- | |||
Torus water level is 105 inches. | |||
- Torus water temperature is 180 degrees F. | |||
- Reactor pressure is 700 psig. | |||
Which ONE of the following states whether Alternate RPV Depressurization | |||
is required, not required, or prohibited and the reason. | |||
Under these conditions, Alternate RPV Depressurization is | |||
a. not required since primary containment limits are not exceeded, | |||
b. required to ensure the energy released during an RPV blowdown I | |||
can be accepted. | |||
. | |||
. c. required since the downcomers are now exhausting to the torus | |||
free air space. | |||
d prohibited since the SRV Tail Pipes are now exhausting to the | |||
torus free air space. | |||
! | |||
. . . - . | |||
SENIOR REACTOR OPERATOR Paga 37 | |||
OUESTION: 054 (1.00) | |||
To initiate a reactor scram when the control room has been evacuated, it | |||
is undesirable to deenergize the RPS busses as the means of scramming | |||
because: | |||
, | |||
a. ' nuclear instrumentation needed to monitor reactor power will | |||
become denergized. | |||
b. pressure control using turbine bypass valves will be lost after | |||
the scram. | |||
c. RPV level control will unnecessarily transfer from feedwater to | |||
HPCI. | |||
I | |||
d. groups I, ll, Ill, and VI isolations will be defeated. j | |||
QUESTION: 055 (1.00) | |||
The following conditions exist: | |||
- The plant is operating at 75% power. | |||
- At 0800 one Safety Relief Valve opened. | |||
- At 0802 EOP-03 has been entered due to torus water temperature | |||
reaching 80 degrees F. | |||
At what point should a Reactor Recirculation pump speed reduction and | |||
manual reactor scram be performed? | |||
a. Immediately when it is determined that the SRV cannot be | |||
reclosed. | |||
b. At 0810. | |||
c. When torus temperature reaches 120 degrees F. | |||
d. When the " unsafe"_ region of the Heat Capacity Temperature Limit | |||
curve is entered. | |||
- | |||
- ...-. . . .. .. .- - . - . - - . - - .. - . . . . | |||
I | |||
l | |||
' | |||
SENIOR REACTOR OPERATOR P:go 38 ' | |||
i | |||
' | |||
, | |||
QUESTION: 056 (1.00) | |||
Drywell spray was initiated in accordance with EOP-03. As drywell 1 | |||
temperature and pressure are decreasing, the unacceptable region on the | |||
Drywell Spray initiation Limit curve is entered at a Drywell temperature | |||
of 250 degrees ~ F. | |||
Which ONE of the following is the REQUIRED action? | |||
a. Secure drywell spray when drywell pressure drops below 2.2 | |||
psig, | |||
b. Secure drywell spray when torus bottom pressure drops below 2.2 | |||
psig. | |||
c. Adjust drywell spray as necessary to maintain operation within | |||
the Drywell Spray Initiation limit curve. ; | |||
l | |||
d. Immediately secure drywell spray. l | |||
l | |||
QUESTION: 057 (1.00) | |||
A loss of feedwater heating has occurred. Which ONE of the following is | |||
the REQUIRED immediate operator action? | |||
Run back Reactor Recirculation flow until: | |||
a. reactor power has been reduced 25% below its pretransient level | |||
without regard to current total core flow, | |||
b. total core flow has been reduced to 36 Mlb/hr without regard to | |||
the current reactor power level, | |||
c. reactor power has been reduced to at least 25% below its | |||
pretransient level AND total core flow has been reduced to at | |||
least 36 Mlb/hr. | |||
d. reactor power has been reduced 25% below its pretransient level | |||
OR total core flow has been reduced to 36 Mlb/hr. | |||
-- - | |||
_ , . , , . .. . . | |||
! | |||
I | |||
- SENIOR REACTOR OPERATOR Paga 39 - | |||
, | |||
QUESTION: 058 (1.00)' | |||
A startup is in progress with reactor pressure at 900 psig when the "A" | |||
CRD pumps trips and the "B" CRD pump cannot be started. .Two accumulator | |||
alarms, both in the same nine rod array, illuminate. | |||
D | |||
Which ONE of the following is the required. action? | |||
, | |||
s. Manually scram the reactor. | |||
b. Determine the cause of the alarms. If both alarms are due to | |||
low gas pressure then manually scram the reactor, | |||
c. Fully insert one of the rods with an accumulator alarm and 1 | |||
- | |||
disarm its directional control valves. | |||
, d. Enter LCO to be in Cold Shutdown within 24 hours.' | |||
QUESTION: 059 (1.00)- | |||
While operating at 100% power, a recirculation pump seal failure causes | |||
EOP-03 entry on high drywell pressure and high drywell temperature. | |||
Following initiation of suppression chamber spray, drywell pressure | |||
- stabilizes at 4 psig, torus bottom pressure stabilizes at 8 psig, and | |||
drywell temperature stabilizes at 175 degrees F. | |||
Which ONE of the following actions is REQUIRED? | |||
a. Declare an Unusual Event and initiate drywell spray in | |||
' | |||
i | |||
accordance with the Primary Containment Pressure leg of EOP-03. | |||
b. Declare an Alert and in:tlate drywell spray in accordance with | |||
the Drywell Temperature leg of EOP-03. | |||
. c. - Declare an Unusual Event. Do not initiate drywell spray. | |||
d. Declare an Alert. Do not initiate drywell spray. | |||
. .- | |||
J | |||
J | |||
- - - - . , - - , . . _ , . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ | |||
- - ._ . . . _ . _ _ _ . . . . - _. . _ _ | |||
: SENIOR REACTOR OPERATOR Prgs 40 | |||
1 | |||
1 | |||
) | |||
. | |||
- QUESTION: 060 (1.00) I | |||
1 | |||
With the plant operating on the 65% load line, condenser vacuum starts | |||
to decrease. Reactor Recirculation Flow is reduced in accordance with -i | |||
plant procedures. ! | |||
After the Reactor Recirculation Flow reductions the plant will be | |||
operating, | |||
a. in the scram region. | |||
b. In the exit region.- | |||
I | |||
c. In the caution zone. | |||
d. above the MELLA line. | |||
< | |||
QUESTION: 061 (1.00) | |||
l | |||
The plant is operating at 100% power when control rods start to drift. | |||
i The MAXIMUM number of control rods in a nine rod array that are allowed | |||
to drift WITHOUT REQUIRING tho mode switch to be placed in shutdown is: | |||
a. one rod without regard to whether the rods are drifting in or | |||
out. | |||
b. two rods if rods are drat;ng in and one rod if rods are | |||
- drif ting out. , | |||
l | |||
c. two rods without regard to whether the rods arr liiting in or I | |||
out. | |||
[ d.~ three rods if rods are drifting in and two rods if rods are | |||
drifting out, | |||
t | |||
0 | |||
: | |||
; | |||
; | |||
l | |||
<- | |||
ta | |||
, , - - | |||
.,. . . , , - . - | |||
.. . - - _-_ _ . - _ _ _ | |||
SENIOR REACTOR OPERATOR Pcgs 41 | |||
OUESTION: 062 (1.00)- | |||
The plant is operating at power when a total loss of TBCCW occurs, | |||
immediate actions are complete in accordance with plant procedures. | |||
Which ONE of the following describes RPV pressure and level control? | |||
a. ' RCIC is being used in the' level control mode and HPCI is being | |||
.used in the pressure control mode. | |||
b. HPCIis being used in the level control mode and RCIC is being | |||
used in the pressure control mode, | |||
c. HPCI is being used in the level control mode and SRVs are being | |||
used to control pressure. RCIC remains shutdown. | |||
d. RCIC is being used in the level control mode and SRVs are being | |||
used to control pressure. HPCI remains shutdown. | |||
QUESTION: 063 (1.00) | |||
The plant is operating at 100% power when a loss of Bus A5 occurs. | |||
Which ONE of the following action (s) is(are) required? , | |||
a. Reduce reactor power to maintain steam tunnel temperature below | |||
' | |||
160 degrees F. | |||
b. If steam tunnel temperature exceeds 160 degrees F scram the | |||
reactor and close the MSIVs. | |||
c. If steam tunnel temperature exceeds 160 degrees F scram the | |||
reactor. Maintain the MSIVs open. | |||
d. If steam tunnel temperature exceeds 160 degrees F commence a | |||
normal plant shutdown. | |||
I | |||
SENIOR REACTOR OPERATOR Pcga 42 | |||
I | |||
, | |||
QUESTION: 064 (1.00). | |||
: | |||
With the plant at 100% power on the 100% load line, reactor water level j | |||
; | |||
starts to decrease due to unknown causes. . Level is currently + 25 inches | |||
and is trending down at 1/2 inch per minute. I | |||
l | |||
Which ONE of the following is the required action assuming water level | |||
continues to fall? : | |||
: | |||
a. Insert rods using the RPR rods until below 70% load line, then l | |||
reduce core flow to 36-40 Mlb/hr. .) | |||
I | |||
b. Insert rods using the RPR rods until below 70% load line, then | |||
' | |||
reduce recirculation pump speed to minimum. | |||
c. Reduce recirculation pump speed to minimum, then insert rods as | |||
necessary to exit the caution zone, | |||
d. Reduce core flow to 36-40 Mlb/nr, then insert rods using the | |||
RPR rods until below 70% load line, then reduce recirculation | |||
pump speed to minimum. | |||
. | |||
! | |||
QUESTION: 065 (1.00) | |||
The plant is operating at 100% power with the "B" Reactor Recirculation | |||
Pump scoop tube locked when a reactor scram occurs. | |||
Which ONE of the following actions are REQUIRED? | |||
a. Direct a licensed operator to manually position the "B" Reactor | |||
Recirculation MG set scoop tube to minimum speed. | |||
b. Direct any member of the operating crew to manually position | |||
the "B" Reactor Recirculation MG set scoop tube to minimum | |||
speed. | |||
c. Unlock the scoop tube, if possible, then run the "B" Reactor | |||
Recirculation pump to minimum speed. | |||
d. Trip the "B" Reactor Recirculation Motor Generator Set. | |||
, | |||
, - , , , | |||
SENIOR REACTOR OPERATOR Pigs 43 | |||
QUESTION: 066 (1.00) | |||
Given the following conditions: | |||
- A fuel leak occurs and as a result the reactor is manually | |||
scrammed. | |||
- Due to the fuelleak, the CRD HCU east and west areas radiation | |||
levels reach 1200 mR/hr and 1250 mR/hr respectively. | |||
- The west Scram Discharge Volume vent and drain valves have | |||
failed open. | |||
Under these conditions, Alternate RPV depressurization is: | |||
a. not required since the CRD HCU east and west areas are | |||
considered the same area, | |||
b. required in order to protect secondary containment from | |||
failing. | |||
c. required to allow the scram to be reset and the primary system | |||
leak isolated, | |||
d. not required since there is no primary system discharging into | |||
secondary containment, | |||
i | |||
1 | |||
; | |||
i | |||
. | |||
i | |||
! | |||
! | |||
l | |||
l | |||
. - . . _ - . .. _.... | |||
t | |||
' SENIOR REACTOR OPERATOR Prga 44 | |||
4 | |||
OUESTION: 067 (1.00) | |||
. | |||
The following conditions exist: | |||
- | |||
A seismic event has caused the torus suction lines to both Core | |||
Spray loops to crack downstream of the Core Spray Suction (MO- | |||
1401-3) valves. | |||
- | |||
These cracks result in the water level in the SE and NW | |||
; Quadrants to reach 8 inches and 10 inches above the floor | |||
respectively. | |||
- Efforts to lower the water level are only able to maintain | |||
level. | |||
- There is no primary system discharge into secondary | |||
containment. | |||
! | |||
Which ONE of the following is required by EOP-047 ; | |||
1 | |||
, | |||
a. Isolate the Core Spray suction from the torus. | |||
1 | |||
b. Maintain the Core Spray suction aligned to the torus, | |||
c. Perform Alternate RPV Depressurization. | |||
d. Transfer the Core Spray suction for both loops to the CST. | |||
; | |||
OUESTION: 068 (1.00) i | |||
Which ONE of the following conditions violates secondary containment | |||
integrity? | |||
a. Both drywell personnel access doors are open. | |||
b. Reactor water cleanup MO-1201-2 (RWCU Suction) valvo is failed ' | |||
open. | |||
~ | |||
c. Reactor building ventilation is secured due to dampers failing | |||
closed. | |||
d. One refuel floor exhaust isolation damper is failed open with | |||
the other refuel floor exhaust isolation damper open and fully | |||
operable. | |||
- | |||
-. _ | |||
: SENIOR REACTOR OPERATOR Pega 45 | |||
OUESTIONi 069 (1.00) | |||
The following conditions exist: | |||
- A reactor startup is in progress | |||
- The Reactor Mode Switch is in "Startup/ Hot Standby" | |||
- Reactor pressure is 850 psig | |||
The main turbine is tripped | |||
-- A valid Group Iisolation has occurred | |||
- All systems operated as designed | |||
Which ONE of the following conditions caused the Group Iisolation? | |||
a. Low main steam line pressure | |||
b. Two main steam lines isolating | |||
c. High main steam tunnel temperature | |||
d. High reactor water level | |||
: | |||
1 | |||
l | |||
QUESTION: 070 (1.00) i | |||
l | |||
A steam leak in the drywell has occurred and the control room crew has | |||
entered EOP-01 and EOP-03. TI-9019 and TRU-9044 on panel C903 are l | |||
' | |||
broken. In accordance with the data contained in the attached 2.1.27, | |||
which ONE of the following is the instrument run temperature for the "A" | |||
channel instruments? | |||
a. 208 degrees F | |||
b. 210 degrees F | |||
c. 216 degrees F | |||
d, 220 degrees F | |||
! | |||
t - , .- - , -, | |||
. . . . . . .. . _ . . | |||
_ .. .. . . . - - - . . - . . - - . . . . . - | |||
: | |||
SENIOR REACTOR OPERATOR Pag 3 46 , | |||
! | |||
i | |||
OUESTION: 071 (1.00) , | |||
! | |||
, | |||
initiating suppression chamber spray prior to torus bottom pressure l | |||
' | |||
reaching 16 psig prevents fatigue failure of | |||
l | |||
a. SRV Tail Pipes. - | |||
b. Torus Drywell vacuum breakers. | |||
. | |||
c. downcomers. l | |||
- | |||
: | |||
d. the Reactor Building-Torus vacuum breakers. | |||
! | |||
l | |||
l | |||
QUESTION: 072 (1.00) | |||
a | |||
l The following conditions exist: | |||
- A core off-load is in progress. | |||
- The Refuel Bellows Seal Rupture alarm is received followed 2 | |||
minutes later by the Fuel Pool Low Level alarm. | |||
- Currently an irradiated bundle has been removed from the core | |||
but is still above the reactor vessel. | |||
Which ONE of the following is the REQUIRED action? | |||
a. Immediately evacuate the refuel floor and leave the bundle | |||
hoisted above the reactor vessel, | |||
b. Return the bundle to the in-core position that it came from, j | |||
c. Place the bundle in the nearest open in-core position. | |||
d. Place the bundle in the nearest open fuel pool position. | |||
i | |||
- . -_ .-. _ | |||
SENIOR REACTOR OPERATOR pig 3 47 | |||
OUESTION: 073 (1.00) | |||
The following conditions exist: | |||
- | |||
The reactor is shutdown. | |||
- | |||
At 1600 all RPV water level indication was lost due to | |||
electrical problems and EOP-16 was entered. | |||
- At 1630 conditions to flood the RPV were established with A, B, | |||
and D SRVs open and RPV pressure 52 psig above torus pressure. | |||
- At 1640 electrical power was restored and water level can be | |||
determined. | |||
Which ONE of the following actions are REQUIRED? | |||
a. Immediately exit EOP-16 and enter EOP-01 at L-1 and P-7. | |||
b. Continue vessel flooding until 1819 then immediately exit EOP- | |||
16 and enter EOP-01 at R-1, | |||
c. Concurrently execute EOP-16 and EOP-01 at L-1 and P-7. | |||
d. Continue vessel flooding until 1819 then stop all injection. | |||
Verify that RPV level decreases before the MCUTL is reached, i | |||
then exit EOP-16 and enter EOP-01 at R-1. | |||
I | |||
I | |||
, | |||
, | |||
l | |||
l | |||
.. - . .. .- . - . .... | |||
SENIOR REACTOR OPERATOR- P gs 48 | |||
QUESTION: 074 (1.00) | |||
" | |||
The following conditions exist: | |||
A failure to scram has occurred. | |||
' | |||
.- | |||
- No boron has been injected. ' | |||
- | |||
Reactor power is 30%. | |||
; - The Main Turbine is tripped. | |||
- | |||
The Main Condenser is available. , | |||
l - | |||
.orus water level is normal. | |||
Due to difficulty in establishing suppression pool cooling, the | |||
' | |||
- | |||
. Heat Capacity Temperature Limit (HCTL) was exceeded. | |||
- | |||
Which ONE of the following states the proper method of controlling | |||
reactor pressure? , | |||
j | |||
a. Reactor pressure should be reduced using the main turbine 1 | |||
bypass valves to stay below the HCTL curve. | |||
b. Reactor pressure should be reduced using the SRVs to stay below l | |||
' | |||
the HCTL curve. | |||
, | |||
c. Alternately depressurize using the main turbine bypass valves. i | |||
! | |||
d. Altemately depressurize using the SRVs. | |||
, | |||
, | |||
OUESTION: 076 (1.00) | |||
; | |||
, | |||
Which ONE of the following actions allow the operator to disregard NPSH | |||
limits? | |||
a. After a successful reactor scram, Core Spray is being used to l | |||
maintain level between -125 to +45 inches. | |||
b. After a successful reactor scram, LPCI is being used to | |||
maintain level between + 12 to +45 inches. | |||
c. Durin0 an ATWS, LPCI is being used to maintain level between | |||
-155 to -140 inches after level was lowered until reactor power | |||
dropped below 3%. | |||
d. During an ATWS with Alternate RPV Depressurization required and | |||
all SRVs INOPERABLE, LPCI is being used for injection. | |||
, | |||
) | |||
, , | |||
. . . _ _ _ _ _ | |||
SENIOR REACTOR OPERATOR pig 3 49 | |||
QUESTION: 076 (1.00) | |||
The following conditions exist: | |||
- A steam leak occurs just upstream of the Main Turbine Stop | |||
Valves with ooth MSIVs in the "A" main steam line failing to | |||
close. | |||
- A reactor scram is successful in inserting all rods fully. | |||
- Both Main Stack Process Radiation Monitors have been reading | |||
2.5E+4 for the last 25 minutes. | |||
- Off-site release rate projections are 2 R/ hour Whole Body at | |||
the site boundary. | |||
Select the correct action and its reason. | |||
Under these conditions the preferred method of depressurizing the RPV is | |||
using: | |||
a. SRVs because of the scrubbing potential of the torus water, | |||
b. SRVs because the heat removal capability is greater than the | |||
Main Turbine Bypass Valves, | |||
c. Main Turbine Bypass Valves because the hotwell is the preferred | |||
heat sink. | |||
d. Main Turbine Bypass Valves because the heat removal capability | |||
is greater than the SRVs. | |||
s | |||
.- | |||
t | |||
y_ ,, - - | |||
e - ,- -, ,. - | |||
- . .- _- _ - . _ _ - . . . _ _ | |||
i | |||
SENIOR REACTOR OPERATOR Pcgs 50 | |||
, | |||
4 | |||
QUESTION: 077;(1.00) | |||
- With a Reactor Building Vent Radiation Hi-Hi Alarm present, EOP-04 , | |||
directs the operator to verify secondary containment isolation of 1 | |||
Reactor Building Heating and Ventilation and the initiation of Standby | |||
' | |||
. Gas Treatment System. j | |||
.' | |||
\ | |||
- | |||
'This verification will ensure that: | |||
! | |||
a, the Reactor Building atmosphere is contained at a positive | |||
pressure until it can be treated and released. | |||
I | |||
b. a trected and controlled ground release of the activity is j | |||
provided. j | |||
i c. a treated and controlled elevated release of the activity is | |||
provided. l | |||
d. both the primary and secondary containments are maintained at a | |||
; | |||
slightly negative pressure, | |||
i | |||
' | |||
l | |||
. 1 | |||
i OUESTION: 078 (1.00) | |||
The following conditions exist: i | |||
1 | |||
1 | |||
- A reactor startup is in progress with RPV pressure at 500 psig. | |||
" - It is determined the "A" Channel of Group i PCIS has one | |||
reactor high water level switch (16A-K105A) that will NOT trip. | |||
The MINIMUM time allowed to place 16A-K105A in the tripped condition is: | |||
a. one hour, | |||
b. two hours. | |||
c. six hours, | |||
d. twelve hours. | |||
0- | |||
- SENIOR REACTOR OPERATOR Pcge 51 | |||
- QUESTION: 079 (1.00) , | |||
~ The following conditions exist: | |||
- ' A successful automatic reactor scram occurred on high reactor ; | |||
pressure. | |||
- The main condenser is ava'ilable but not currently in service. | |||
- The operator is attempting to stabilize pressure between 900- | |||
1060 psig using SRVs. | |||
l | |||
Re-establishing the main condenser as a heat sint:: | |||
a.- is not allowed. 1 | |||
i | |||
b. is preferred but is allowed only if no valid MSIV isolations | |||
exist. | |||
c. is required immediately after valid MSIV isolation signals are | |||
overridden. | |||
d. is only allowed if the SRVs become unavailable. ; | |||
l | |||
l | |||
QUESTION: 080 (1.00) l | |||
With the plant operating at 100% power, the control room becomes | |||
uninhabitable because of toxic gas. Evacuation is ordered and only the | |||
immediate Actions of Pf4PS Procedure 2.4.143 were carried out. | |||
At this point reactor water level is being maintained by: | |||
a. Reactor Feed Pumps and CRD. | |||
b. RCIC and CRD. | |||
: | |||
c. HPCI and CRD. | |||
d. CRD only. | |||
> | |||
M | |||
SENIOR REACTOR OPERATOR Prgs 52 | |||
OUESTION: 081 (1.00)- * | |||
With the plant at 100% power, an MPR and EPR f ailure caused the turbine | |||
stop valves to close and the turbine bypass valves to remain closed. | |||
Reactor pressure peaked at 1330 psig at which time the reactor scrammed | |||
on high flux. - | |||
Select the statement below that correctly describes the transient. | |||
a. No safety limit violation occurred. The Stop Valve closure | |||
scram was the only RPS trip failure. J | |||
b. No safety limit violation occurred. The Stop Valve closure | |||
scram was not the only RPS trip failure, | |||
c. A safety limit violation occurred. The Stop Valve closure | |||
scram was the only RPS trip failure. | |||
d. A safety limit violation occurred. The Stop Valve closure | |||
scram was not the only RPS trip failure. | |||
l | |||
1 | |||
QUESTION: 082 (1.00) | |||
Following a reactor scram, the Mode Switch should be taken to Shutdown | |||
as soon as possible in order to: | |||
a. disable the low steam pressure isolation. | |||
b. enable the high reactor water level isolation. | |||
c. insert another scram signal for 2 seconds, | |||
d. allow MSIV closure without generating a scram signal. | |||
_ | |||
SENIOR REACTOR OPERATOR P:ga 53 ' | |||
OUESTION: 083 (1.00) | |||
The following conditions exist: | |||
- The reactor was shutdown at 0230. | |||
- Due to loss of level indications, EOP-16 was entered at 1030. | |||
- At 1100 flooding conditions were established with 3 SRVs open. | |||
- Flooding was stopped as soon as Flooding Cornpletion Time was | |||
reached, | |||
t | |||
1 | |||
Assuming RPV levelinstruments do not respond, which ONE of the | |||
16116 wing is the LATEST time at which injection must be reinitiated? | |||
- a. 1214 | |||
b. 1217 | |||
c. 1254 | |||
- | |||
d. 1257 | |||
OUESTION: 084 (1.00) | |||
A worker in the Emergency Response Organization had 100 mrem TEDE for | |||
the current year and 2.5 Rem TEDE lifetime prior to the declaration of | |||
an emergency. Which ONE of the following is the MAXIMUM TEDE this | |||
worker can receive over the course of the emergency without special | |||
authorization? | |||
a. 2.4 Rem | |||
b. 2.5 Rem I | |||
c. 4.9 Rem | |||
d. 5.0 Rem i | |||
! | |||
; | |||
l | |||
~_ | |||
_ - . . . _. . __ _ _ _ _ _. . __ _ _ _ _ . _ _ _ _ _ _ _ _ _ | |||
e | |||
-SENIOR REACTOR OPERATOR Pago 54 ; | |||
i | |||
l | |||
1 | |||
l | |||
. | |||
QUESTION: 085 (1.00) j | |||
1 | |||
~ | |||
A surveillance on the Reactor Water Cleanup High Flow Isolation is due. l | |||
. | |||
Which ONE of the following describes how the duration of the | |||
surveillance is tracked and when the inoperability clock begins and | |||
ends? | |||
a. The surveillance is tracked in the NOS Logbook. | |||
The clock starts when the system is removed from service and | |||
ends when the system is returned to normal lineup. | |||
b. The surveillance is tracked in the NOS Logbook. | |||
The clock starts when the system is removed from service and | |||
ends when the NWE signs off the surveillance. l | |||
c. The surveillance is tracked using an LCO Maintenance Planning | |||
i | |||
Checklist. The clock starts when the system is removed from | |||
I | |||
i service and ends when the system is returned to normal lineup. | |||
d. The surveillance is tracked using an LCO-Maintenance Planning | |||
Checklist. The clock starts when the syt. tem is removed from : | |||
-, service and ends when the surveillance is signed off by the I | |||
l | |||
' work group. | |||
T | |||
5 | |||
s | |||
- . | |||
- - .- - - -. ._. - -. . | |||
SENIOR REACTOR OPERATOR P:gs 55 | |||
, | |||
QUESTION: 086 (1.00) | |||
With the plant at 5% power, a closed motor operated valve located in the | |||
drywell must be tagged in the closed position. | |||
Which ONE of the following is the proper method for determining that the | |||
valve is in the closed position? | |||
a. The position should be first verified by the indirect method | |||
before power is isolated. The isolation of the power supply | |||
may then be performed. Independent verification of the power | |||
supply is not required. | |||
b. The position should be first verified and independently | |||
verified by the indirect method before power is isolated. The | |||
isolation of the power supply may then be performed and | |||
independently verified. | |||
c. The first verifier should enter the drywell for verification of | |||
valve position. The independent verifier may perform an | |||
indirect verification of remote valve position. | |||
d. The first verifier and the independent verifier should make | |||
separate drywell entries for verification of valve position. | |||
QUESTION: 087 (1.00) | |||
Which ONE of the following may enter the Controls Area without receiving | |||
permission from the NWE/NOS or Control Room Operator? | |||
a. Operations Department Manager | |||
' | |||
b. NRC Resident inspector | |||
ej. Station Director | |||
d. The Outside Nuclear Plant Reactor Operator (NPRO) | |||
1 | |||
.-. . . . .- | |||
SENIOR REACTOR OPERATOR P:gs 55 | |||
QUESTION: 086 (1.00) . | |||
With the plant at 5% power, a closed motor operated valve located in the | |||
drywell must be tagged in the closed position. | |||
Which ONE of the following is the proper method for determining that the | |||
valve is in the closed position? | |||
a. The position should be first verified by the indirect method | |||
before power is isolated. The isolation of the power supply. | |||
may then be performed independent verification of the power | |||
supply is not required, | |||
b. The position should be first verified and independently | |||
verified by the indirect method before power is isolated. The | |||
isolation of the power supply may then be performed and | |||
independently verified. | |||
c. The first verifier should enter the drywell for verification of | |||
valve position. The independent verifier may perform an | |||
indirect verification of remote valve position. | |||
d. The first verifier and the independent verifier should make | |||
separate drywell entries for verification of valve position. l | |||
l | |||
QUESTION: 087 (1.00) | |||
Which ONE of the following may enter the Controls Area without receiving | |||
permission from the NWE/NOS or Control Room Operator? | |||
a. Operations Department Manager | |||
b. NRC Resident inspector | |||
c. Station Director | |||
d. The Outside Nuclear Plant Reactor Operator (NPRO) | |||
i | |||
l | |||
l | |||
l | |||
l | |||
l | |||
I | |||
_ _ | |||
. . . .- . . . - . - . . - _ | |||
SENIOR REACTOR OPERATOR Pigs 56 | |||
- QUESTION: 088 (1.00) | |||
Absent a basis to assign a longer duration, Which ONE of the following | |||
is the normal duration of a temporary modification? | |||
a. Installation until the end of the shift | |||
b.- 6 weeks following installation | |||
c. 6 months following installation | |||
d. installation until the end of the refueling outage | |||
QUESTION: 089 (1.00) | |||
The MINIMUM amount of parallel watchstanding REQUIRED in order to | |||
reactivate an NRC reactor operator license is: | |||
a. 8 hours. | |||
i | |||
b. 40 hours. ) | |||
i | |||
c. seven 8 hour shifts. | |||
l | |||
d. five 12 hour shifts. | |||
l | |||
l | |||
QUESTION: 090 (1.00) | |||
1 | |||
Which ONE of the following is the MINIMUM REQUIRED protective equipment l | |||
for handling Sodium Hypochlorite outdoors? 1 | |||
Safety Goggles and: | |||
a. Rubber Gloves | |||
b. Rubber Gloves and Apron, Rubber Safety Boots, Forced Air | |||
Respirator i | |||
c. Rubber Gloves and Apron, Rubber Safety Boots | |||
d. Rubber Gloves and Apron, Rubber Safety Boots, Respirator | |||
i | |||
SENIOR REACTOR OPERATOR Prgs 57 | |||
QUESTION: 091 (1.00) | |||
Which ONE of the following conditions would allow a fail open air | |||
operated valve to be DANGER tagged in the closed position? | |||
' | |||
a. The valve is gagged in the closed position with a device to | |||
ensure it does not change state. | |||
' | |||
b. The DANGER tag is only for equipment protection and no | |||
maintenance will be performed under this tagout.. | |||
c. The air supply to the valve is also DANGER tagged in the open | |||
position. | |||
d. A " Human Red Tag" is assigned to monitor the status of air to | |||
the valve. | |||
- | |||
QUESTION: 092 (1.00) | |||
Which ONE of the following conditions PROHIBIT the use of a " Human Red | |||
Tag"? | |||
a. The only qualified tagger available to be a " Human Red Tag" is | |||
- a member of the work group. | |||
b. Two isolation points are required to provide isolation. | |||
' | |||
c. The work is expected to take 2 hours to complete. | |||
d. The work is expected to take 1 hour to complete with only 1/2 | |||
hour left in the current shift. | |||
l | |||
l | |||
I | |||
1 | |||
+ | |||
- _ . _ . _ | |||
1 | |||
. | |||
. | |||
1 | |||
SENIOR REACTOR OPERATOR Prgs 58. ' | |||
l; | |||
. | |||
l | |||
QUESTION: 093 (1.00) 1 | |||
1 | |||
An offsite fire department is responding to the site during a fire in a. / | |||
- vital area. ) | |||
' Which ONE of the following describes the security reouirements in order ! | |||
- to allow access to the protected area / vital area? l | |||
i | |||
a.' The fire truck and firemen must be searched prior to entering ; | |||
~ the protected area. No additional search is required prior to | |||
entering the vital area provided security escorts the' team. | |||
b. No search is required of the fire truck or firemen prior to | |||
entering the protected area provided security escorts the team, , | |||
however both the truck and firemen must be searched prior to > | |||
entering the vital area. : | |||
l | |||
c. No search is required of the f;c - truck or firemen prior to l | |||
entering the protected area or vital area provided security | |||
escorts the team. l | |||
l | |||
d. No search is required of the fire truck or firemen prior to | |||
entering the protected area or vital area provided security and | |||
operations department escort the team. ) | |||
QUESTION: 094 (1.00) | |||
You are working in a Hot Particle Control Zone (HPCZ) in a double set of | |||
protective clothing. Which ONE of the following is the proper method of | |||
removing the protective clothing when exiting the area? | |||
a. Remove both sets of protective clothing at the step off pad at | |||
the exit of the HPCZ. | |||
b. Remove both sets of protective clothing at the step off pad at | |||
. the exit of the buffer zone. | |||
c. Remove the outer set of protective clothing at the step off pad | |||
at the exit of the HPCZ and the' inner set of protective | |||
clothing at the step off pad at the exit of the buffer zone, | |||
d. Remove the outer set of protective clothing at the step off pad | |||
at the exit of the buffer zone and the inner set of protective | |||
clothing at the step off pad at the exit of the HPCZ zone. | |||
4 | |||
;... .- , | |||
_ | |||
l | |||
U | |||
SENIOR REACTOR OPERATOR Pzg3 59 l | |||
l | |||
l | |||
l | |||
QUESTION: 095 (1.00) l | |||
An ALERT has been declared. Which ONE of the following describes the l | |||
REQUIRED emergency notification? I | |||
l | |||
l | |||
a. The NRC must be notified within 15 minutes after the | |||
declaration of the ALERT. State and local agencies must be 1 | |||
' | |||
notified immediately thereafter, not to exceed 1 hour. | |||
b. State and local agencies must be notified within 15 minutes | |||
after the declaration of the ALERT. The NRC must be notifind | |||
immediately thereafter, not to exceed 1 hour. | |||
c. The NRC must be notified within 1 hour after the declaration of | |||
the ALERT. State and local agencies must be notified | |||
immediately thereafter, not to exceed 1 hour and 15 minutes, | |||
d. State and local agencies must be notified within 1 hour after | |||
the declaration of the ALERT. The NRC must be notified | |||
immediately thereafter, not to exceed 1 hour and 15 minutes | |||
QUESTION: 096 (1.00) | |||
Which ONE of the following describes the required manning of the Fire | |||
Brigade? | |||
The Fire Brigade shall consist of five members: | |||
a. including the Brigade Leader. Two of these persons may also be | |||
part of the crew required for safe shutdown of the plant. | |||
b. including the Brigade Leader. These persons may not be part of | |||
, | |||
the crew required for safe shutdown of the plant. | |||
c. excluding the Brigade Leader. Two of these persons may also be | |||
part of the crew required for safe shutdown of the plant. | |||
d. excluding the Brigade Leader. These persons may not be part of | |||
the crew required for safe shutdown of the plant. | |||
- - . . . . - .. | |||
1 | |||
SENIOR REACTOR OPERATOR P:ga 60 | |||
l | |||
OUESTION: 097 (1.00) | |||
l | |||
During an emergency, a reasonable action that departs from Technical l | |||
Specifications must be taken immediately. | |||
1 | |||
in accordance with PNPS procedures, which ONE of the following MUST ; | |||
approve taking this action? 1 | |||
a. An on shift licensed Reactor Operator and on shift licensed | |||
Senior Reactor Operator i | |||
i | |||
b. A licensed Senior Reactor Operator only | |||
c. A licansed Senior Reactor Operator and the Operations | |||
Department Manager | |||
d. A licensed Senior Reactor Operator and the Operations | |||
Department Manager and the Plant Manager j | |||
i | |||
l | |||
I | |||
l | |||
OUESTION: 098 (1.00) | |||
You have worked the foDowing schedule: | |||
- Thursday 1st scheduled day off l | |||
- | |||
Friday 2nd 7 am to 7 pm I | |||
- Saturday 3rd 7 am to 7 pm ) | |||
- | |||
Sunday 4th 7 am to 3 pm | |||
- | |||
Monday 5th 7 am tn 3 pm | |||
- Tuesday 6th 7 am to 9 pm | |||
- | |||
Wednesday 7th 7 am to 3 pm | |||
- | |||
Thursday 8th 7 am to ? | |||
Which ONE of the following represents the LATEST you can be required to | |||
work on Thursday the 8th, without special approval being granted? | |||
(Assume turnover time is NOT included) | |||
a.3pm | |||
b. 5pm | |||
c. 7 pm | |||
d. 9 pm | |||
i | |||
1 | |||
- . . . _ . - . ~ . . | |||
l | |||
l | |||
SENIOR REACTOR OPERATOR Pzga 61 | |||
QUESTION: 099 (1.00) | |||
The only individual available for a call-in for TSC staffing informed | |||
the Nuclear Watch Engineer (NWE) on the phone that he has consumed | |||
alcohol within the previous 5 hours. | |||
Which ONE of the following describes the individuals ability to work in | |||
the TSC? : | |||
a. not permitted to work in the TSC. | |||
b. permitted to work in the TSC provided the individual informs | |||
the NWE that alcohol has not impaired his ability to work in | |||
the TSC. A blood alcohol concentration test is at the NWE | |||
discretion, based on the NWE phone discussions with the ; | |||
individual. | |||
c. permitted to work in the TSC only if a blood alcohol | |||
concentration test is performed upon arrival on site and | |||
the concentration is less than 0.04. | |||
1 | |||
d. permitted to work in the TSC only if a breathalyzer l | |||
test is performed upon arrival on site and the blood to | |||
alcohol ratio is greater than or equal to 4.0. | |||
! | |||
l | |||
QUESTION: 100 (1.00) | |||
A procedure is currently being performed which requires the installation l | |||
of a jumper. It is discovered that the procedure directs the jumper I | |||
placement in a position that would cause an unexpected ESF actuation. A | |||
change to the jumoer position is required. | |||
Which ONE of the following is the required method to revise this | |||
procedure to change the jumper position? l | |||
a. Editorial correction | |||
b. SRO Change | |||
c. Minor Revision | |||
d. Major Revision ] | |||
( * * * * * * * * * * END OF EXAMIN ATION * * * * * * * * * *) | |||
-- - | |||
.. . - - - | |||
, | |||
, | |||
' BOSTON EDIS0N RTYPE H6.02 | |||
. | |||
PILGRIM NUCLEAR POWER ST. TION | |||
Procedure No. 2.1.27 | |||
DRYWELL TEMPERATURE INDICATION | |||
i | |||
REQUIRED REVIEWS REVIEWERS AND APPROVERS | |||
hE. A$ oves /NNW 6b7M | |||
"" Writ *" '**''' | |||
Thi"k '*d""*' | |||
STAR | |||
Act | |||
8) | |||
'wcynical Reviewer | |||
'gg)q,V | |||
Date | |||
Review *# 9 # ^ * '' u' 9/#6/ | |||
Validator ' | |||
Date' W | |||
SAFETY REVIEW E0"!9ED/ | |||
/i f | |||
Procedure A' ner | |||
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e/d | |||
/Dgte | |||
NOT REQUIRED | |||
N/+ | |||
QAD Man'ager | |||
ORC REVIEW REQUIRED / Date | |||
MT REQ'J:"C0 | |||
AA L lo /k 194 | |||
- | |||
ORCC{plirman ' | |||
Date | |||
0m Ibb smlk fobt9/94 | |||
sporpible anager / 1Dhte | |||
Effective Date: /0/d8 9% | |||
020095 2.I.27 Rev. 3 | |||
- -. .. . | |||
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REVISION LOG | |||
REVISION 3 Date originated 5/94 | |||
Paaes Affected Descriotion | |||
4 Add PDC 92-58 to References. | |||
~5,7,10,12,14,15 Revise Kaye nomenclature and delete channel points from old | |||
Kaye recorder in accordance with PDC 92-58. | |||
Editorial 2C Date Originated 3/93 | |||
Paaes Affected Descriotion- | |||
7,9,11,13 Delete references to Station Honeywell Computer System as it | |||
is obsolete. | |||
Editorial 2B Date Originated | |||
1 | |||
Paoes Affected pescription i | |||
1 | |||
4,5,7 Editorial corrections to reflect new E0P numbers and entry l | |||
conditions and to add new Editorial Correction rev bar l | |||
identifications. I | |||
Editorial 2A Date Originated | |||
Paaes Affected Qgscriotion | |||
4,5,7 Incorporated editorial corrections to Main Control Room | |||
Panel Labels per PDC 87-78C. | |||
REVISION 2 Date Originated | |||
Paaes Affected Descriotion | |||
All Reformat to comply with PNPS 1.3.4-1.2. | |||
. | |||
2.1.27 Rev. 3 | |||
Page 2 of 15 | |||
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, .IABLE OF CONTENTS | |||
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EA9A | |||
1.0 PURPOSE AND SC0PE................................................. 4 | |||
2.0 REFERENCES........................................................ 4 | |||
3.01 . DEFINITIONS....................................................... 4 | |||
4.0 DISCUSSION........................................................ 4 | |||
5.0 PRECAUTIONS AND LIMITATIONS....................................... 6 | |||
6.0 PREREQUISITES..................................................... -6' | |||
7.0 PROCEDURE......................................................... 7 | |||
8.0 ATTAC HME N T S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 | |||
' ATTACHMENT 1 - TE-5050 TEMPERATURE ELEMENTS - BULK DRYWELL | |||
TEMPERATURE ESTIMATE..................................... 9 | |||
ATTACHMENT 2 - TE-8125 TEMPERATURE ELEMENTS - BULK DRYWELL | |||
TEMPERATURE DETERMINATION............................... 10 | |||
ATTACHMENT 3 - TE-5050 TEMPERATURE ELEMENTS - INSTRUMENT RUN | |||
T EM PE RATUR E E ST I MAT E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 | |||
ATTACHMENT 4 - TE-8125 TEMPERATURE ELEMENTS . INSTRUMENT RUN | |||
TEMPERATURE DETERMINATION............................... 12 | |||
ATTACHMENT 5 - TE-5050 TEMPERATURE ELEMENTS - LOCATION INFORMATION..... 13 | |||
ATTACHMENT 6 .TE-8125 TEMPERATURE ELEMENTS - LOCATION INFORMATION..... 14 | |||
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2.1.27 Rev. 3 | |||
Page 3 of 15 | |||
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1.0 - PURPOSE AND SCOPE - | |||
' | |||
This Procedure provides instructions for determining Drywell bulk temperature when | |||
, , | |||
the Emergency Operating Procedures _(EOPs) require measurement of this parameter. | |||
- | |||
2.0 REFERENCES | |||
, | |||
2.1 DEVELOPMENTAL | |||
[1] PNPS Technical Specifications Table 3.2.H | |||
4 | |||
[2] PNPS Technical Specifications Tables 3.2.H and 4.2.H | |||
- | |||
[3] PDC 87-78C, Improvements to Labels, Nameplates on Main Control Room Panels | |||
i | |||
[4] PDC 92-58, Kaye Recorder Replacement ! | |||
2.2 IMPLEMENTING | |||
; | |||
[l] PNPS 2.2.49, " Primary Containment Cooling System" | |||
[2] PNPS 8.7.1.4.2, ' Primary Containment Integrated Leak Rate Test" | |||
3.0 DEFINITIONS | |||
None | |||
4.0 DISCUSSION | |||
[1] The following sections of the Emergency Operating Procedures require | |||
measurement of Drywell temperature: | |||
, (a) E0P-1, RPV Control: RPV Water Level Instrument Run temperatures associated | |||
with the RPV Saturation Temperature Figure of Caution 1. | |||
(b) E0P-2, Failure to Scram: RPV Water Level Instrument Run temperatures | |||
: associated with the RPV Saturation Temperature Figure of Caution 1. | |||
(c). E0P-3, Primary Containment Control: | |||
. | |||
(1)' Entry condition _(150"F) | |||
, | |||
(2) Drywell temperature path | |||
(3) RPV Water Level Instrument Run temperatures associated with the RPV | |||
, | |||
Saturation Temperature Figure of Caution 1. | |||
(4)- Figure 5: (SPDS 031) DSIL (Orywell Spray Initiation Limit) | |||
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2.1.27 Rev. 3 | |||
Page 4 of 15 | |||
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4.0 913GUSSION (Continued) | |||
(d) E0P-4, Secondary Containment Control: RPV Water Level Instrument Run , | |||
temperatures associated with the RPV Saturation Temperature Figure of | |||
_ | |||
. | |||
Caution 1. | |||
(e) E0P-16, RPV Flooding | |||
E0P-26, RPV Flooding, Failure To Scram: | |||
(1) Temperatures near the RPV Water Level Instrument Reference Leg | |||
vertical runs. | |||
[2] Drywell temperature is normally monitored in the Control Room by using | |||
TRU-9044, DRYWELL TEMP / PRESS Recorder, and TI-9019, DW TEMP Indicator, on | |||
Panel C903. TRU-9044 receives its input from a single temperature element | |||
located at a relatively low elevation in the Drywell. TI-9019 receives its | |||
input from a single temperature element located just below the neck of the | |||
Drywell. Both of these temperature elements measure ambient Drywell air space | |||
temperature. | |||
[3] The TE-5050A through P temperature elements are used to evaluate Drywell | |||
tem >erature with respect to Technical Specifications limits (refer to | |||
Technical Specifications Table 3.2.H). The Drywell locations of these | |||
elements are listed in Attachment 5. These elements are used to monitor , | |||
Drywell temperature for Technical Specifications requirements because of their I | |||
reliability, location, and their redundancy (dual-element RTDs). In addition, I | |||
these temperature elements are the primary elements used for the Primary | |||
Containment Integrated Leak Rate Test. | |||
[4] Local Drywell air temperature indication is supplied by the TE-8125 series I | |||
temperature elements. The TE-8125 series temperature indication consists of | |||
20 RTDs located throughout the Drywell which provide input to the Kaye Temp. 1 | |||
Computer (refer to Attachment 6). | |||
[5] When TRU-9044 and TI-9019 are not available, selected Drywell temperature | |||
elements are used to estimate an average temperature near the RPV water level | |||
instrument runs and an average bulk Drywell temperature. Temperatures near | |||
the RPV water level instrument runs are monitored by those thermal elements | |||
which are located in the upper elevations of the Drywell since mest of the | |||
instrument runs are found in this region of the Drywell. Bulk average Drywell ' | |||
temperature is a weighted average temperature based on the volume of the | |||
Drywell. By averaging more readings from the lower region of the Drywell | |||
(which contains most of the Drywell air space) than from the upper region of | |||
the Drywell, a representative average Drywell temperature is obtained. More | |||
sophisticated methods to calculate a_ weighted average Drywell temperature are | |||
available, as part of the ILRT Procedure, PNPS 8.7.1.4.2. The method outlined | |||
, | |||
in this Procedure, however, attempts to balance the complexity and time | |||
. | |||
consuming aspects of the sophisticated approach against the requirement to | |||
rapidly obtain a value for Drywell temperature suitable for use in the E0Ps. | |||
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2.1.27 Rev. 3 | |||
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Page 5_ of 15 | |||
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5.0 PRECAUTIONS AfC LIMITATIONS | |||
[1] The Drywell temperature shall be maintained within the following limits when | |||
the reactor coolant temperature is above 212 F. | |||
(a) Above elevation 40': $ 194*F ; | |||
(b) Equal to or below elevation 40': s 150*F | |||
Upon determination that the Drywell temperature at any elevation has exceeded | |||
the above limits, the Drywell temperature at each elevation shall be logged | |||
every 30 minutes. The Drywell temperature shall be reduced to within the | |||
above limits within 24 hours; otherwise corrective action shall be as | |||
specified in Technical Specifications Sections 3.2.H.2 and 3.2.H.3. | |||
i (Tech Spec 3.2 H.1) | |||
[2] If the Drywell temperature has exceeded either limit of Technical | |||
Specifications Section 3.2.H.1 for greater than 24 hours, an engineering | |||
, | |||
evaluation shall immediately be initiated to assess potential damage and | |||
render a determination of ability of safety related equipment to perform its | |||
intended function. | |||
If either limit of Technical Specifications Section 3.2.H.1 has been exceeded | |||
for greater than 24 hours, ferther action to justify continued operation shall | |||
be determined by an engineering evaluation which must be completed within one | |||
week. (Tech Spec 3.2.H.2) | |||
[3] If the requirements of Technical Specifications Section 3.2.H.2 have not been I | |||
met, an orderly shutdown shall be initiated and the reactor shall be in a cold | |||
shutdown condition within 24 hours. (Tech Spec 3.2.H.4) | |||
[4] If the Drywell temperature at any elevation exceeds 215*F and the temperature | |||
cannot be reduced to below 215 F within 30 minutes, a reactor shutdown shall | |||
be initiated and the reactor shall be in cold shutdown condition within | |||
24 hours. (Tech Spec 3.2.H.4) | |||
i | |||
[5] When reactor coolant temperature is above 212*F, the Drywell air temperature ' | |||
limits will be determined by reading the instruments listed in Techaical | |||
Specifications Table 3.2.H. These instruments shall be logged once per shift, | |||
and each reading compared to the limits of Technical Specifications | |||
Section 3.2.H.1. (Tech Spec 4.2.H.1) | |||
J | |||
6.0 PREREQUISIT_E.ji | |||
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None | |||
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2.1.27 Rev. 3 | |||
Page 6 of 15 | |||
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-7.0 50CEDURE | |||
. , . | |||
[1] DETERMINE bulk Drywell temperature using one of the following methods (listed | |||
in order of preference): | |||
(a) SELECT the higher of the valves indicated on TI-9019, DW TEMP Indicator, | |||
and TRU-9044, DRYWELL TEMP /F 'SS Recorder (Panel C903). | |||
CAUTION , | |||
i | |||
j The instruments listed below are not environmentally qualified for use in a harsh ' | |||
environment. Under accident conditions, they should only be used if either j | |||
TI-9019 or TRU-9044 is not available for use. | |||
1 | |||
(b) Highest probable Drywell temperature from EPIC points DRY 002 or DRY 004. | |||
(c) For a more representative bulk teaperature, AVERAGE the TE-5050 series RTDs | |||
using the computer points in accordance with Attachment 1. | |||
(d) For a more representative bulk temperature, AVERAGE the TE-8125 series RTDs | |||
using the Kaye Temp. Computer in accordance with Attachment .2. | |||
1 | |||
(e) All of the TE-5050A through P series RTDs can be read locally at Panel C85, ' | |||
Reactor Building El. 23' East, for Attachment I data. | |||
[2] DETERMINE RPV water level instrument run temperature using one of the ! | |||
' | |||
following methods (listed in order of preference): | |||
(a) SELECT the higher of the values indicated on TI-9019 and TRV-9044 ! | |||
(Panel C903). l | |||
(b) AVERAGE the TE-5050 series RTDs in ecordance with Attachment 3. | |||
(c) AVERAL .he TE-8125 series RTDs using the Kaye Temp. Computer in accordance | |||
with " '.achment 4. | |||
1 | |||
(d) The TE-5050A through P RTDs can be read locally at Panel C85, Reactor l | |||
Building El. 23' East, for Attachment 3 data. | |||
[3] Additional information on Drywell temperature elements and location is | |||
contained in PNPS 2.2.49, " Primary Containment Cooling System". 1 | |||
1 | |||
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2.1.27 Rev. 3 | |||
Page 7 of 15 | |||
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8.0 ATTACMENTS | |||
ATTACHMENT 1 - TE-5050 TEMPERATURE ELEMENTS - BULK DRYWELL TEMPERATURE ESTIMATE | |||
ATTACHMENT 2 - TE-8125 TEMPERATURE ELEMENTS - BULK DRYWELL TEMPERATURE. DETERMINATION | |||
ATTACHMENT 3 - TE-5050 TEMPERATURE ELEMENTS - INSTRUMENT RUN TEMPERATURE ESTIMATE | |||
ATTACHMENT 4 - TE-8125 TEMPERATURE ELEMENTS - INSTRUMENT RUN TEMPERATURE | |||
DETERMINATION | |||
ATTACHMENT 5 - TE-5050 TEMPERATURE ELEMENTS - LOCATION INFORMATION | |||
ATTACHMENT 6 - TE-8125 TEMPERATURE ELEMENTS - LOCATION INFORMATION | |||
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2.1.27 Rev. 3 | |||
Page 8 of 15 j | |||
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ATTACHMENT 1 | |||
Sheet 1 of 1 | |||
TE-5050 TEMPERATURE ELEMENTS | |||
BULK DRYWELL TEMPERATURE ESTIMATE | |||
[1] SELECT one temperature element in each group of temperature elements Alg! | |||
RECORD its temperature. | |||
[2] COMPUTE the average temperature as follows: | |||
(a) Average - (A + B + C + D + E + F)/6 | |||
T'.ME | |||
TE-5050 EPIC | |||
GROUP COMPUTER POINT | |||
ELEMENT # TEMPERATURE ( F) | |||
A | |||
DRY 002 | |||
A -----OR---- -------------- ------- ------- ------- ------- ------- ------- | |||
B DRY 004 | |||
i | |||
. | |||
E | |||
DRY 010 | |||
B -----OR---- ---- - --- ------- ------- ------- ------- -- ---- ------- | |||
G DRY 014 | |||
C -----OR---- -------------- ------- ------- ------- ------- ------- ------- | |||
H DRY 116 | |||
L l | |||
DRY 122 ' | |||
D -----0R---- -------------- ------- ------- ------- ------- ------- ------- | |||
M DRY 124 | |||
K | |||
DRY 120 | |||
E -----0R---- -------------- ------- ------- ------- ------- ------- ------- | |||
J DRY 118 | |||
N | |||
DRY 126 | |||
F -----OR---- -------------- ------- ------- ------- ------- ------- ------- | |||
p DRY 130 | |||
- | |||
AVERAGE | |||
Performed By Date Reviewed By Date | |||
2.1.27 Rev. 3 | |||
Page 9 of 15 | |||
ATTACHMENT 2 | |||
Sheet 1 of 1 | |||
TE-8125 TEMPERATURE ELEMENTS | |||
BULK DRYWELL TEMPERATURE DETERMINATION | |||
[1] SELECT one temperature element in each group of temperature elements 88Q | |||
, | |||
RECORD the temperature indicated on the Kaye Temp. Computer. | |||
[2] COMPUTE the average temperature a: follows: | |||
(a) Average - (A + B + C + 0 + E + F)/6 | |||
TlME . | |||
TE-8125 | |||
GROUP | |||
ELEMENT # TEMPERATURE (*F) | |||
3 | |||
A ----0R----- -------- --------- --------- --------- --------- -------- l | |||
l | |||
4 l | |||
l | |||
l | |||
9 i | |||
: | |||
8 .... 0R.... ........ ......... ......... ......... ......... ........ 1 | |||
10 | |||
.. | |||
11 | |||
C -----OR---- -------- --------- --------- --------- --------- -------- | |||
12 | |||
13 | |||
D -----OR---- -------- --------- --------- --------- --------- -------- | |||
14 | |||
15 | |||
E -----OR---- ----- -- --------- --------- --------- --------- -------- | |||
16 | |||
17 | |||
^ | |||
F -----OR---- -------- ------- *- --------- --------- --------- -------- | |||
18 | |||
AVE: GE | |||
Performed By' Date Reviewed By Date | |||
2.1.27 Rev. 3 | |||
Page 10 of 15 | |||
ATTACHMENT 3 | |||
Sheet 1 of 1 | |||
TE-5050 TEMPERATURE ELEMENTS | |||
INSTRUMENT RUN TEMPERATURE ESTIMA1E | |||
[1]. DETERMINE.the rack (s) of concern 8tlQ RECORD the indicated temperature for each | |||
element in that group. | |||
[2] COMPUTE the average temperature. | |||
Instrument Runs for Rack 2205 | |||
A Channel Instruments | |||
TLME | |||
!TE-5050 EPIC | |||
ELEMENT COMPUTER POINT | |||
# TEMPERATURE ( F) | |||
. | |||
A | |||
........... ...... | |||
DRY 002 | |||
. ..... | |||
Zw . . . ....... ....... ....... ....... ....... | |||
j | |||
........... ........ . ..... . .. .. ....... ....... ....... ....... ....... | |||
AVERAGE | |||
Instrument Runs for Rack 2206 | |||
B Channel Instruments | |||
TLME | |||
TE-5050 EPIC | |||
ELEMENT COMPUTER POINT | |||
# TEMPERATURE (*F) | |||
..... ..... ...... . ..... . $b. ....... ....... ....... ....... ....... | |||
D DRY 008 2.Ir | |||
........... .................. ....... ....... ....... ....... ....... | |||
4..... | |||
E DRY 010 LIO . | |||
AVERAGE | |||
Performed By Date Reviewed By Date | |||
2.1.27 Rev. 3 | |||
Page 11 of 15 | |||
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ATTACHMENT 4 | |||
Sheet 1 of 1 | |||
1 | |||
TE-8125 TEMPERATURE ELEMENTS | |||
INSTRUMENT RUN TEMPERATURE DETERMINATION | |||
[1] DETERMINE the rack (s) of concern A151 RECORD the indicated temperature for each l | |||
element in that group. ! | |||
[2] COMPUTE the average temperature. | |||
Instrument Run for Rack ?205 | |||
A Channel Instrument | |||
T;:ME | |||
TE-8125 | |||
ELEMENT # TEMPERATURE (*F) | |||
4 | |||
........... | |||
[b | |||
........ ......... ......... ......... ......... ........ | |||
- | |||
10 -,g | |||
AVERAGE | |||
- | |||
Instrument Runs for Rack 2206 | |||
B Channel Instrument | |||
T::ME | |||
TE-8125 | |||
ELEMENT # TEMPERATURE (*F) | |||
4 | |||
........... | |||
2.[h | |||
........ ......... ......... ......... ......... ........ | |||
9 | |||
Q{[] . | |||
AVERAGE | |||
Performed By Date Reviewed By Date | |||
2.1.27 Rev. 3 | |||
Page 12 of 15 | |||
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ATTACHMENT 5 | |||
Sheet 1 of 1 | |||
. | |||
LTE-5050 TEMPERATURE ELEMENTS LOCATION INFORMATION | |||
, | |||
. | |||
Temperature EPIC Elevation Azimuth Area | |||
Element Point ID (feet) (degrees) Monitored | |||
; | |||
TE-5050A DRY 002 86 0 2' out from vessel below | |||
an exh. register | |||
TE-50508- DRY 004 89 180 l' out from vessel. | |||
- TE-5050C DRY 006 86 50 4' out from vessel above | |||
supply register | |||
- TE-50E00 DRY 008 90 330 2' below head exh. hole. | |||
TE-5050E DRY 010 60 270 2' out from bio-shield. | |||
' | |||
4 out from bio-shield. | |||
~ | |||
-TE~-5050F DRY 012 60 90 | |||
TE-5050G DRY 014 40 270 10' out from bio-shield under | |||
Main Steam Line. | |||
TE-5050H DRY 116 40 90 10' out from bio-shield under | |||
Main Steam Line. | |||
TE-5050J DRY 118 35 0 l' from CRD area inside wall. | |||
TE-5050K DRY 120 35 180 l' from CRD area inside wall. l | |||
l | |||
TE-5050L DRY 122 22 205 13' out from CRD area outside ! | |||
wall. | |||
TE-5050M DRY 124 22 45 13' out from CRD area outside , | |||
wall. l | |||
TE-5050N DRY 126 15 270 8' out trom CRD rea nutside | |||
wall. | |||
TE-50500 DRY 128 15 0 On CRD area outside wall. | |||
N | |||
TE-5050P DRY 130 12 125 10' out from CRD area outside | |||
wall. | |||
~ | |||
. | |||
1 | |||
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2.1.27 Rev. 3 i | |||
Page 13 of 15 ) | |||
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ATTACHMENT 6 | |||
Sheet 1 of 2 | |||
4 | |||
TE-8125 TEMPERATURE ELEMENTS LOCATION INFORMATION | |||
' | |||
Temperature Elevation Azimuth Area | |||
Element (feet) (dearees) Monitored | |||
TE-8125-1 90 285 Head Exhaust | |||
TE-8125-2 90 210 Head Exhaust | |||
, | |||
TE-8125-3 85 180 3' out from vessel | |||
, | |||
TE-8125-4 85 0 4' out from vessel | |||
above ductwork | |||
TE-8125-5 82 300 In exhaust duct | |||
TE-8125-6 82 45 In exhaust duct | |||
TE-8125-7 80 270 In annulus | |||
: TE-8125-8 80 90 In annulus | |||
TE-8125-9 54 270 6' out from ! | |||
bio-shield l | |||
' | |||
l | |||
; | |||
TE-8125-10 54 90 6' out from ' | |||
. | |||
bio-shield | |||
TE-8125-ll 40 270 10' out from | |||
, bio-shield under Main | |||
, | |||
Steam Line | |||
TE-8125-12 40 90 10' out from | |||
bio-shield under Main | |||
Steam Line , | |||
1 | |||
l | |||
TE-8125-13 25 315 On CRD area outside | |||
wall | |||
! | |||
TE-8125-14 '25 135 On CR0 area outside | |||
wall | |||
TE-8125-15 19 225- 14' out from CRD | |||
area outside wall | |||
TE-8125-16 19 45 14' out from CRD | |||
m es outside wall. | |||
2.1.27 Rev. 3 | |||
Page 14 of 15 | |||
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ATTACHMENT 6 | |||
Sheet 2 of 2 | |||
TE-8125 TEMPERATURE ELEMENTS | |||
LOCATION INFORMATION | |||
Temperature- Elevation Azimuth Area | |||
Element (feet) idearees) Monitored | |||
TE 8125-17~ 14 265 On CR0 area outside wall | |||
TE-8125-18. 14 110 On CRD area outside wall | |||
TE-8125-19 29 180 l' from CRD area | |||
4 | |||
inside wall | |||
TE-8125-105 29 0 l' from CR0 area | |||
inside wall | |||
. | |||
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2.1.27 Rev. 3 | |||
Page 15 of 15 | |||
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SENIOR REACTOR OPERATOR Pegs 1 | |||
ANSWER KEY | |||
MULTIPLE CHOICE O23 c | |||
001 b 024 b | |||
002 b 025 d | |||
003 d 026 c | |||
004 c 027 d | |||
005 b 028 c | |||
l | |||
006 a 029 6 at A | |||
007 b 030 c ! | |||
1 | |||
008 c. 031 a } | |||
s 4 | |||
000 a M - | |||
010 a 033 a | |||
011 a 034 b | |||
012 a 035 b | |||
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SENIOR REACTOR OPERATOR Pags 2 | |||
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' SENIOR REACTOR OPERATOR | |||
: P:gs 3 > | |||
ANSWER KEY | |||
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002 : b | |||
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-_ (* " * * * * * * * END OF FXAMINATION * " "' ' * ) | |||
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ATTACHMENT 2 | |||
Facility Comments on Written Examination | |||
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1 | |||
1 | |||
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. . _ _ _ .. . . . _ _ . . , ._ .. . . .. . _ . . . .._. . | |||
N | |||
, | |||
e 10CFR50.55 | |||
h | |||
Boeton Edinon | |||
b | |||
s. | |||
Pilgrim Nudear Power Station | |||
Rocky Hdi Road | |||
- Plymouth, Massachusetts 02360 | |||
L J. Olivier | |||
, Vice President Nuclear Operations | |||
;. and Station Director | |||
May 16,1997 | |||
BECo Ltr. 2.97-054 | |||
4 | |||
U.S. Nuclear Regulatory Commission | |||
Region I ' | |||
: . 475 Allendale Ruad | |||
King of Prussia, PA 19406 | |||
Docket No. 50-293 , | |||
License No. DPR-35 | |||
Pilarim's 1997 NRC Written Examination Comments | |||
a | |||
1 The written examination administered on May 5,1997, was considered to be an in-depth examination, | |||
which fairly tested the six (6) SRO candidate's knowledge in the appropriate areas. After thorough | |||
, . analysis of the content of the examination, it is clear that the use of misleading information, use of the | |||
double negative context, and the asking of subjects not important to public health and safety were | |||
avoided. | |||
However, specific requests on several written exam questions are submitted for your consideration in | |||
Enclosure 1. Enclosure 2 contains the reference documentation associated with each of the requests. | |||
' | |||
- Your consideration of these requests is greatly appreciated. | |||
. | |||
e | |||
L. J. Olivier | |||
- PMKINRCEXCO. | |||
Enclosure ' | |||
- cc:. See ne'xt page | |||
< | |||
.b | |||
, | |||
-, , | |||
- - . -. | |||
. | |||
! | |||
' | |||
cc- Mr. Don Florek | |||
Region 1. | |||
475 Allendale Road . | |||
King of Prussia, PA .19406 | |||
Mr. Alan Wang, Project Manager | |||
Project Directorate 1-3 | |||
: Division of Reactor Projects - 1/11 | |||
Mail Stop: 14B2 | |||
U. S. Nuclear Regulatory Commission | |||
1 White Flint North | |||
11555 Rockville Pike | |||
Rockville, MD 20852. | |||
' | |||
' U.S. Nuclear Regulatory Commission | |||
Attention: Document Control Desk | |||
Washington, DC 20555 | |||
' Senior Resident inspector ., | |||
Pilgrim Nuclear Power Station | |||
! | |||
I | |||
= | |||
1 | |||
Y | |||
, | |||
9 | |||
ENCLOSURE 1 | |||
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... | |||
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l | |||
ENCLOSURE 1 | |||
' | |||
1 | |||
1. Question # 32 | |||
; While operating at 100% power, it is determined that the Main Steam Lin'e High Flow | |||
switches on the "B" Main Steam Line will t10T trip under a high flow condition. | |||
Which ONE of the following is the MINIMUM REQUIRED action? | |||
a. Direct I&C personnel to manually trip the inoperable switches. I | |||
, | |||
b. Direct l&C personnel to manually insert a half Group i isolation on the "B" Group I - | |||
Channel. | |||
c. Initiate an orderly shutdown and be in Cold Shutdown Condition within a | |||
i MAXIMUM of 30 hours after the instrument failure. | |||
' | |||
d. Initiate an orderly shutdown and have Main Steam Lines isolated within a | |||
MAXIMUM of 10 hours after the instrument failure. | |||
. | |||
ANSWER: d. | |||
; | |||
p 4 | |||
DISCUSSION: l | |||
The stem of this question states,"... it is determined that the Main Steam Line High Flow | |||
' | |||
switches on the "B" Main Steam Line will NOT trip under a high flow condition..." | |||
There are two trip systems associated with Group I PCIS, designated "A" and "B". Trip System | |||
"A" has inputs from MSL High Flow switches comprising two instrument channels, and Trip | |||
System "B" has eight inputs from MSL High Flow switches comprising two instrument channels, | |||
each steam line is equipped with four switches each, one for each instrument channel | |||
(Enclosure 2, Attachment 1, page 1). | |||
, | |||
; . The stem of the question states that all flow switches on the "B" Main Steam Une are | |||
- | |||
inoperable. Since this is the case, there are less than two operable instrument channels for | |||
9 -- | |||
both PCIS logic trip systems. (See Enclosure 2, Attachment 1, Page 2) | |||
; | |||
< . | |||
; Since there are less than the minimum operable instrument channels for both trip systems,. | |||
1 Attachment 1, page 3 states, "If the minimum number of operable instrument channels cannot | |||
' | |||
be met for both trip systems, place at least one trip system (with the most inoperable channels) | |||
' | |||
in the tripped condition within one hour or initiate thc appropriate action required by Table | |||
3.2.A listed below for the affected trip function." | |||
, | |||
Table 3.2.A requires action "B", which states, " Initiate an orderly load reduction and have Main | |||
' | |||
. Steam Lines isolated within 8 hours". | |||
, | |||
Since there is no grace period (of one hour) for the two trip system inoperability (vice the one | |||
trip system inoperability), there is no obvious correct response. | |||
Page 1 of 10 | |||
- - _. . -_ . . ..- .- - . | |||
. . - . . _ _ . . -. - | |||
REQUEST (Question # 321: | |||
Since the correct response is not offered as a choice in the responses, we request that this , | |||
question be deleted from the examination. | |||
REFERENCE: | |||
PNPS Technical Specifications, Table 3.2.A and associated notes (Enclosure 2, | |||
Attachment 1). | |||
. | |||
W | |||
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Page 2 of 10 | |||
- .. - . .. .- . .- - - . . . . . . . . - - - - - - - . - . -. - - - | |||
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! 2. Question # 76 l | |||
The following conditions exist: | |||
I | |||
l | |||
- A steam leak occurs just upstream of the Main Turbine Stop Valves with both )' | |||
j MSIV's in the "A" main steam line failing to close. | |||
- A reactor scram is successful in inserting all rods fully. | |||
' | |||
- . Both Main Stack Process Radiation Monitors have been reading 2.5+E4 for the last | |||
' | |||
. | |||
25 minutes. | |||
- Off-site release rate projections are 2 R/ hour Whole Body at the site boundary. | |||
; | |||
, Select the correct action and its raason. | |||
Under these conditions, the preferred method of depressurizing the RPV is using: | |||
a. SRVs because of the scrubbing potential of the torus water. 'l | |||
l | |||
; b. SRVs because the heat removal capability is greater than the Main Turbine Bypass i | |||
; Valves. l | |||
c. Main Turbine Bypass Valves because the hotwell is the preferred heat sink. . | |||
1 | |||
d. Main Turbine Bypass Valves because the heat removal capability is greater than | |||
the SRVs. | |||
: | |||
j ANSWER: b. | |||
;; l | |||
! I | |||
DISCUSSION: | |||
- , | |||
Because both answer "a" and "b" select the SRVs as the correct mechanism of depressurizing, ! | |||
l | |||
the question then becomes discriminatory as to the basis for doing so. Appendix B of the | |||
Emergency Procedure Guidelines states that Contingency #2, Emergency RPV | |||
Depressurization may be required to: | |||
Minimize radioactivity release from the RPV to the primary containment and secondary | |||
containment, or to areas extemal to the primary containment and secondary ' | |||
: containment. | |||
Additionally, Appendix G states that the purpose of the Radioactivity Release guideline is to | |||
limit radioactivity release into areas cutside the primary and secondary containments. | |||
Since distracter "a" implies that SRV's are used because they discharge to the primary | |||
containment, "a" can be construed as the correct answer. That is, given the situation provided, | |||
the fact that the SRVs tiischarge to the containment via the torus is more significant than the | |||
fact that the SRV's heat removal capability is greatedhan the bypass valves. | |||
Since Appendix B also provides generic guidance that SRV's are used because of their heat | |||
. removal capability, "b" is also correct. . | |||
Page 3 of 10 | |||
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_ . . . . .. .. . -. _ . _ ._-_ _ . _ _ . . _ _ _ _ _ _ | |||
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l | |||
REQUEST: (Question # 76) | |||
Because both answers "a" and "b" are correct per the EPGs, we request that answers "a" and | |||
"a" both be accepted as correct, and the question be retained in the examination. | |||
REFERENCEi | |||
1. Emergency Procedure Guidelines Appendix B, Section 11, Contingency #2 (OEl l | |||
Document 8390-4B, [ Enclosure 2, Attachment 2]). l | |||
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Page 4 of 10 | |||
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4, 1 | |||
l 3. - Qitestion #'28 ) | |||
s , | |||
, | |||
, While operating at 100% power, a control rod is determined to be uncoupled. Attempts | |||
; to couple the rod have been unsuccessful. | |||
. Which ONE of the following states the MINIMUM REQUIRED actions. | |||
o | |||
a. Verify the control rod can be moved with drive pressure and maintain the control rod ; | |||
' | |||
' | |||
at the target position. | |||
i | |||
~ | |||
b. 'Fu!!y insert the control rod and hydraulically disarm the CRD. | |||
; | |||
c. Fully insert the control rod and electrically disarm the directional control valves.- j | |||
d. Fully hsert the control rod, electrically disarm the directional control valves and then | |||
declare the rod inoperable. | |||
# | |||
ANSWER: c. J | |||
I | |||
i, | |||
DISCUSSION: | |||
' | |||
The only differentiation between distracter "d" and the correct answer "c" is whether the | |||
rod is declared inoperable. ] | |||
If a control rod was uncoupled, it would be declared inoperable by Technical | |||
Specifications when the inoperability was discovered. The control rod would then be | |||
inserted and electrically disarmed to ensure control rod movement was precluded. | |||
Taking this action does not eliminate the fact that the control rod was inoperable but does | |||
' | |||
allow relief from the requirements of the associated Technical Specification actions for an l | |||
uncoupled control rod. The control rod that was uncoupled would still be administratively I | |||
controlled as an inoperable control rod, even though the action statement of Technical | |||
Specification 3.3.F does not have to be applied. At PNPS, if an action has to be taken | |||
* | |||
on the part of Technical Specifications, the equipment inoperability is traced through the | |||
application of an " Active LCO"in the LCO log. | |||
i | |||
from a Tech Spec consideration only, the rod is not inoperable. However from an | |||
administrative and practical standpoint, the rod is indeed inoperable, and the Active LCO | |||
: is maintained to control the status of the rod. Therefore, if the cand:date approached the | |||
' question from this perspective, distracter "d" can also be considered as an acceptable | |||
, | |||
answer, | |||
;, While Procedure 2.2.87, 5.2.1[3] does state that a rod fully inserted and electrically | |||
L disarmed is not inoperable, it references Tech Spec 3.3.A.2 that concems rods that | |||
; cannot be moved with drive pressure. This statement does not apply to the conditions | |||
' identified in the question. | |||
. | |||
u | |||
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Page 5 of 10 | |||
,. | |||
i ! | |||
.m__ _ m.______m . _ _ . . _ _ _ _ _ . _ . . . _ _ . . ._ , . , , , . , , - , , | |||
._ | |||
.,.,7 ,,, ...~. | |||
. - . . .-. .. . .. . . - - .. .- . . . . -. .. | |||
. | |||
REQUEST: (Question # 28) | |||
We request that distracter "d" also be accepted as correct. | |||
REFERENCE: | |||
, | |||
1. PNPS 1.3.34.2 (See Enclosure 2, Attachment # 3) | |||
. - 3.0[1]" Active LCO" Definition | |||
. ' 4.0 " Discussion" | |||
2. Operations Department Manager (Tom Trepanier, (508) 830-8364) | |||
!. | |||
; | |||
e | |||
id | |||
d | |||
. | |||
, | |||
A ' | |||
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Page 8 of 10 | |||
, | |||
-. . - . - - - . . | |||
+ | |||
4. Question # 50 | |||
in the event that torus water level cannot be maintained above 95 inches, HPCI is | |||
secured in order to prevent: | |||
a. exceeding the Primary Containment Pressure Limit. | |||
b. exceeding the Pressure Suppression Pressure, | |||
c. exceeding the Heat Capacity Temperature Limit. | |||
d. isolating HPCI on high exhaust pressure. | |||
ANSWER: a. | |||
DISCUSSION: | |||
The operators at PNPS are provided a " supplemental" approved handout for the study of | |||
the EOP procedures (Enclosure 2, Attachment 4). In this handout, the basis for the 95 | |||
inches torus level securing of HPCIis not stressed as the PCPL. The fact the exhaust | |||
will become uncovered is stressed, and HPCI will then directly pressurize the | |||
containment. The wording for the PCPL statement is "may exceed the PCPL", and not | |||
"the basis for the uncovery is the PCPL". Wnen this question is considered, the fact that | |||
the primary containment would pressurize is a valid line of thought. From this direction, | |||
scrutinization of the choices through the use of the supplied EOPs would lead a | |||
candidaa to choose the most limiting curve between the PCPL and the PSP. This would | |||
of course be the PSP curve. Based on this line of reasoning, response "b"is considered | |||
also to be a valid response. | |||
REQUEST: (Question # 50) | |||
We request that distracter"b" also be considered as correct. | |||
REFERENCE: | |||
EOP-03 Supplemental Training Materials / Flow Charts (Enclosure 2, Attachment 4) | |||
. | |||
. | |||
, | |||
Page 7 of 10 | |||
_ . . . ._ _ _ ___ _~ . . _. _ _ _ ._ | |||
. | |||
5. Question # 29 | |||
: | |||
With the plant at power, it is determined that the MO-1001-37 (B loop Torus Spray) and | |||
MO-1400-25A (A Loop Core Spray Inboard Injection) valves have failed their operability | |||
test. Both valves are currently closed. | |||
The maximum time allowed before the plant must be in COLD SHUTDOWN is: | |||
a. 24 hours ( 1 day) | |||
b. 96 hours (4 days) | |||
c. 168 hours (7 days) | |||
t | |||
d. 192 hours (8 days) | |||
ANSWER: b. | |||
DISCUSSION: | |||
PNPS 2.2.125, " Containment isolation System" lists the valves that are considered to be | |||
primary containment isolation valves. An identical listing is contained within the FSAR. | |||
Included in this listing are both the MO-1400-25A and the MO-1001-378 (see | |||
Enclosure 2, Attachment 5). As containment isolation valves, the administrative | |||
requirements require at least one valve in the line to be deactivated in the isolated | |||
position, unless the valve receives any signals other than the isolation signal. Whether | |||
the valve receives any other signals (other than the isolation signal) determines whether | |||
the valve has to be deactivated electrically or otherwise administratively controlled. If the | |||
requirements of this procedure are not met (and the questinn does not provide this | |||
information), an orderly shutdown shall be initiated and the reactor shall be in Cold | |||
Shutdown within 24 hours. , | |||
This question was designed to test the applicants' ability to determine: | |||
1) the impact of an Inoperable 378 valve on "LPCl* operability; | |||
2) the impact of an inoperable 378 valve on the Containment Cooling Loop's Operability | |||
and, | |||
3) the overall effect of 1 and 2 when coupled with an inoperable Core Spray system. | |||
At least one applicant, (during a followup interview), interpreted this question as a test of ! | |||
his ability to recognize that: | |||
1) Both valves are PCIS valves | |||
2) That at a minimum, the 25A would need to be deactivated since it receives an Auto | |||
' Open signal and, | |||
3) Determine the corrective actions for failed PCIS valves. | |||
Since the questions asks for the maximum time allowed before the phnt must be in | |||
COLD SHUTDOWN, if a candidate were to assume that the question is testing his | |||
Page 6 f 10 | |||
g | |||
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$ | |||
- | |||
_ | |||
. . _ - . . . _ . . . _ _ _ _ _ . . _ _ . . . - . ._ . . _ . . . . . _ _ | |||
knowledgs of PCIS, thin it is rs: sort:bla that ths candidits would choso "c" es tha | |||
correct response, given that no other actions are taken. | |||
REQUEST: (Question # 29) | |||
, | |||
Due to the two different ways that this question can be Interpreted, we request that both "a" | |||
and "b" be accepted as correct. | |||
REFERENCE: | |||
' | |||
PNPS Procedure 2.2.125 (Enclosure 2, Attachment 5) | |||
i | |||
i | |||
1 | |||
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: | |||
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Page 9 of 10 | |||
. _ | |||
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l | |||
6. Question # 27 | |||
l | |||
'When' valving in a CRD hydraulic control accumulator, the 305-102 (Withdraw Riser . 1 | |||
Isolatio'n Valve) and the 305-112 (Scram Discharge Riser Isolation Valve) are required to | |||
be open prior to opening the 305-101 (insert Riser Isolation Valve). This prevents: | |||
af a single rod scram when opening the 305-101 valve. | |||
b. excessive scram time of that rod in the event of a reactor scram, | |||
c. damage to the accumu5 tor in the event of a reactor scram. | |||
d. damage to the drive mechanism in the event of a reactor scram. | |||
ANSWER: d. . | |||
DISCUSSION: | |||
While it is stated in PNPS 2.2.87 that valve misoperation dunng the isolation or | |||
restoration of a HCU can cause " severe damage to the mechanism", the isolation of the | |||
102 (by itself) can also delay control rod insertion following a scram signal. As seen in | |||
Enclosure 2, Attachment 6, with the 102 valve shut, the exhaust path from the | |||
- mechanism is isolated. Since the question does not state the position of the associated | |||
rod for the HCU being restored, the candidate could reasonably assume that the rod is in | |||
a position other than fully inserted. If the exhaust path is isolated, any scram signal will | |||
not permit the mechanism to scram at " normal" rates, if the control rod inserts at all. | |||
REQUEST: (Question # 27) | |||
' | |||
Due to the fact that response "b" contain the phrase " excessive scram time of the rod in | |||
the event of a reactor scram", we request that response "b" be also accepted as a correct | |||
answer. | |||
, | |||
REFERENCE: | |||
Figure 4 from PNPS Training Material (Enclosure 2, Attachment 6) | |||
Page 10 of 10 | |||
. | |||
.--,,-~---e | |||
ATTACHMENT 3 | |||
NRC Resolution of Facility Comments | |||
Ques 28 Disagree with BECO comment. The question stem requested " MINIMUM l | |||
REQUIRED actions" and the applicants had Technical specifications. The | |||
question clearly related to interpretation of technical specification-required | |||
actions. As specified in Technical Specification 3.3.A.d, control rod drives | |||
that are fully inserted and electrically disarmeo shall not be considered | |||
inoperable. Therefore answer d is incorrect. There was no change to the | |||
answer key. | |||
Ques 29 Agree with BECO comment. There was insufficient information provided in | |||
the question to rule out consideration of containment isolation system | |||
technical specifications. The answer key was revised to accept a or b as | |||
correct answers. | |||
Ques 32 Agree with BECO comment that there is no correct answer to the question | |||
as written. There was no comment provided to this question during the | |||
preexam review. The question was deleted from the examination. l | |||
Ques 50 Disagree with BECO. The question asks for the reason HPClis secured at a | |||
decreasing torus level of 95 inches. Enabling objective 10 required the | |||
applicant to " state the significance of torus levelless than 95 inches as | |||
regards the HPCI system." The significance, stated in O-RO-03-04-05, Rev | |||
, | |||
1, IG 3 is to prevent exceeding the primary containment pressure limit ? | |||
d | |||
(PCPL). The BECO response to the question is attempting to reword the | |||
-question to determine the first EOP-03 curve limit reached if HPCI exhaust is | |||
not secured at a decreasing torus level of 95 inches. This was not the | |||
, | |||
question asked. There is only one correct answer to the question asked. | |||
While the pressure suppression pressure (PSP) will be exceeded, it has a l | |||
- relatively small consequence. The PCPL will be exceeded, which has a large j | |||
consequence, primary containment failure, and is the stated reason in the l | |||
reference material for securing HPCI at a torus level of 95 '"ches. There was | |||
no change to the answer key. | |||
Ques 7 Disagree with BECO comment. As described in 0-RO-03 04-07, Rev 1, IG 9, | |||
the purpose of performing alternate depressurization under the conditions of | |||
' the question is to reduce the driving head and flow of any primary leak by | |||
rapidly reducing the pressure. In 0-RO-03-04-09, Rev 1, IG 18 the SRVs are | |||
used because the heat removal capability (40% power) is greater than the | |||
main turbine bypass valves and the RPV will be depressurized sooner. Tha | |||
basis for venting containment, when required, using the torus vents | |||
considers the scrubbing potential of the torus water to support the torus | |||
method as the preferred method. Venting of the primary containment was | |||
not required based on the conditions given in the question. Therefore, the | |||
only correct answer to this question was answer b. There was no change in | |||
the answer key. | |||
+ | |||
I | |||
ATTACHMENT 4 | |||
SIMULATION FACILITY REPORT | |||
Facility License: DPR 35 | |||
Facility Docket No: 50 293 | |||
Operating Test Administration: May 6-9,1997 | |||
1 | |||
This form is to be used only to report observations. These observations do not constitute i | |||
audit or inspection findings and are not, without further verification and review, indicative f | |||
of a noncompliance with 10 CFR 55.45(b). These observations do not affect NRC ; | |||
certification or approval of the simulation facility other than to provide information that | |||
may be used in future evaluations. No licensee action is required in response to these | |||
observations. | |||
IIfM DESCRIPTION | |||
1 | |||
None | |||
, | |||
l | |||
l | |||
l | |||
1 | |||
}} |
Latest revision as of 17:12, 7 August 2022
ML20149K131 | |
Person / Time | |
---|---|
Site: | Pilgrim |
Issue date: | 07/18/1997 |
From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
To: | |
Shared Package | |
ML20149K125 | List: |
References | |
50-293-97-06, 50-293-97-6, NUDOCS 9707290264 | |
Download: ML20149K131 (103) | |
See also: IR 05000293/1997006
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Enclosure
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
. License No. DPR-35 -
- Report No.97-006
Jocket No.- 50-293 .
Licensee: Boston Edison Company
800 Boylston Street
Boston, Massachusetts 02199
Facility:~ Pilgrim Nuclear Power Station
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Exemination Period: May. 5 - 9,1997
Examiners: D. Florek, Seninr Operations Engineer .
C. Sisco, 0,serations Engineer ,
S. Dernh., Examiner in Training l
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S. Willoughby, Contract Examiner
, Approved by: G. Meyer, Chief
Operations and Human Performance Branch ;
Division of Reactor Safety j
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EXAMINATION SUMMARY
Examination Report 50-293/97-006 (OL)
Initial examinations were administered to six senior reactor operator (SRO) instant
applicants during the period of May 5 -9,1997, at the Pilgrim Nuclear Power Station.
OPERATIONS
Five of six applicants passed the examination. One SRO instant applicant failed the written
and operating portion of the examination. The five applicants that passed were well
prepared for the examinations. The applicants consistently understood and implemented
the emergency operating procedures well. Some weak areas of understanding were
identified during the written exam and operating test.
Two of the applications were found to be deficient in that the applicants had not performed
the five significant control manipulations on the plant as required by 10 CFR 55.31(a)(5).
The applicants' qualification records did not support performance of five significant control-
manipulations. The root ceuse for this problem appeared to be that the BECO program
guidance inappropriately permitted multiple significant control manipulation credit for a
single, extended power change.
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Details
05.1 Operator initial Examinations
a. Scope
]
The examiners administered initial examinations to six instant SRO applicants in
accordance with NUREG-1021, " Examiner Standards," Revision 7.
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b. Observations and Findinas
The results of the initial examinations are summarized below:
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PASS / Fall
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Written 5/1
Operating 5/1
Overall 5/1
The Boston E-jison Company (BECO) staff reviewed the written examination and
assisted in the validation of the operating examination during the week of
April 21,1996. The BECO staff provided comments on the examination that )
significantly improved the examination. The BECO staff, who were involved with
the examination review, signed security agreements to ensure that the initial
examinations were not compromised.
In a letter, dated May 16,1997 (see Attachment 2), BECO provided six comments
on the written examination. The NRC accepted two of the six comments. As a
result, one question was deleted from the examination and two correct answers
were accepted in one question. The NRC resolution of facility comments is
summarized in Attachment 3.'
The following summarizes the written examination questions that were missed by at
least three applicants, indicating a weakness in the understanding of the subject.
Ques 3 Knowledge of the normal indication for the core spray line
break detection monitor.
Ques 33 Knowledge of the method to move an MSIV by use of the
MSIV test push-button.
T Ques 36 Knowledge of the air ejector off gas radiation rnonitor signals
that willinitiate the 13-minute timer,
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Ques 38 ' Ability to use technical specifications related to inoperable
IRMs.-
Ques 43 Ability to determine procedure' entry to a given set of. ,
conditions.
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Ques 61 Knowledge of the number of drifting rods in a nine-rod array
that require placing the mode switch in shutdown. .
- Ques 76 Ability to determine the method and reason for depressurizing
the reactor to a given set of conditions. !
Ques 85 Knowledge of the method to track the duration of
surveillances.
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During the operating test, at least two applicants performed poorly in each of the ,
'following areas: !
Refueling operations
Recognizing a loss of control room annunciators
- The above test items represent areas of weak understanding or performance and are
provided to enable improvement of the training program.
During.the dynamic simulator test, the following item was significant and a
consistent positive observation. j
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Knowledge and understanding of the emergency operating procedures
(EOPs).
During the development and administration of the examination, the examiners noted
the following item for further BECO consideration of possible procedure
improvements.
Emergency Operating Procedure 5.3.21 page 34 of 58 indicated that the
installation of the jumper in panel C915 from jumper location DD-24 to DD-
25 defeated the high drywell pressure and low RPV levelisolation signals for
MO-47, Shutdown Cooling Outboard Isolation Valve. This jumper also
affected isolation signals for MO-29B LPCIinjection valve. The procedure
did not provide a note that this valve was also affected by installation of the
jumper.
- c. Conclusions j
Five_of six of the applicants were well prepared for the examination, and as a result, j
five applicants passed the examination. One SRO instant applicant failed the l
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examination. ;
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05.2 Reactivity Manipulations
a. Scope
The inspector reviewed the BECO records to determine how the applicants complied
with 10 CFR 55.31(a)(5). This section of 10 CFR requires that applicants must
perform five significant control manipulations on the plant that affect reactivity or
power level.
b. Findinos
BECO Document 0-RO-04 "NRC Licensed Nuclear Plant RO/SRO initial
Qualification," dated August 1996, required a minimum of five significant reactivity
manipulations with amplification that effort should be made to diversify the
reactivity manipulations. Ten examples were identified for meeting the requirement.
Four of the examples related to 10% power changes with control reds or
recirculation flow. The inspector considered each of the ten examples as an
appropriate significant control manipulation.
Based on review of the individual applicant qualification records, four of the six
applicants had performed five significant cont'ol
r manipulations on the plant,
although one of these four applicants did not have diverse manipulations.
Based on review of the individual applicant qualification records, the NRC examiner
identified on May 5,1997, that two of the six applicants had not performed five
significant control manipulations on the plant. Although the BECO guidance
specified the minimum conditions for a manipulation, the minimum conditions had
inappropriately been used to credit more than one manipulation when a single,
extended power change occurred. For example, one applicant reduced power with
recirculation flow from 100% to 68% over 56 minutes. BECO considered this to be
three of the five significant control manipulations. The NRC staff disagreed with
BECO and considered this to be one significant control manipulation. Another
applicant reduced power from 100% to 50% initially with recirculation flow and
then later with control rods over 77 minutes. BECO considered this to be all five of
the required five significant control manipulations. The NRC staff disagreed with
BECO and considered this to be two significant control manipulations. BECO was
informed of the examiner's conclusion and informed that this would not impact the
administration of the remainder of the examination. The resolution of this issue was
pursued after the examination was administered.
The final applications submitted on April 18,1997, indicated that these two
applicants had performed their five required significant control manipulations. After
the NRC staff review of the supporting data for the application, the .NRC staff
concluded that one applicant had performed two of the five significant control
manipulations and the other applicant had performed three of the five significant
control manipulations. These two applicants did not meet the requirements of 10
CFR 55.31(a)(5).
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In discussions with BECO and the NRC on June 4,1997, the NRC reiterated the
NRC position and informed BECO that these two applicants passed the examination
but would not be issued licenses until the applicants and BECO submitted revised
Form NRC-398s after five significant control manipulations were performed on the
plant. BECO acknowledged the NRC staff finding and indicated that they had
initiated actions to have the applicants perform additional significant control
manipulations on the plant after the examination was administered and would
submit revised Form NRC-358s.
BECO submitted revised Form NRC-398s in a letter dated June 1,1997. BECO also
provided the details of how the applicants satisfied the 10 CFR requirement for
significant control manipulations. Based on the revised applications and supporting
data, the NRC subsequently issued licenses for these individuals.
c. Conclusion
The BECO guidance and examples of how to meet the requirements of 10 CFR
55.31(a)(5) were acceptable. However, the BECO practice of giving multiple
significant control manipulation credit for a single, eendad power change was not
acceptable. The examiner concluded that BECO had violated 10 CFR 55.31(a)(5),
which requires that applicants for operator licenses must have performed five
significant control manipulations on the plant that affects affect reactivity or power l
level. With the multiple manipulations removed, one SRO applicant had performed I
two significant control manipulations, and another SRO applicant had performed
three significant control manipulations. (VIO 97-06-01)
E8 Review of UFSAR Commitments
A recent discovery of a licensee operating their facility in a manner contrary to the
updated final safety analysis report (UFSAR) description highlighted the need for a
special focused review that compares plant practices, procedures, and/or
parameters to the UFSAR descriptions. While performing the examination activities
discussed in this report, the examiners reviewed portions of the UFSAR that related
to the selected examination activities, questions or topic areas. The particular
section reviewed was Table 5.2.4. The specific question reviewed was consistent
with the UFSAR.
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V. Manaaement Meetinas
X1 Exit Meeting Summary
At the conclusion of the examination, the examiners discussed their observations of the
examination process with members of BECO management. BECO acknowledged the !
examiners' observations. The BECO personnel present at the exit included the following:
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J. Alexander, Training Manager .
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M. Briggs, Principal Instructor
K. DiCroce, Sr. Regulatory Affairs Engineer
L. Olivier, Vice President Nuclear l
M. Santiago, Operations Training Manager 1
(T. Sullivan, Plant Manager .
.T, Trepanier, Operations Department Manager !
T. Venkataraman, QA Group Manager
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NRC Personnel-
S. Dennis, Operations Engineer l
D. Florek, Sr. Operations Engineer ]
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R. Laura, Senior Resident inspector
C. Sisco, Operations Engineer
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Attachments: !
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1. SRO Examination and Answer Key
2. Facility Comments on Written Examinations ;
3. NRC Resolution of Facility Comments i
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4. Simulation Facility Report I
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ATTACHMENT 1
SRO Examination and Answer Key
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U. S. NUCLEAR REGULATORY COMMISSION
SITE SPECIFIC EXAMINATION
SENIOR OPERATOR LICENSE
REGION 1
APPLICANT'S NAME:
FACILITY: Pilarim 1
REACTOR TYPE: BWR-GE3
DATE ADMINISTERED: May 5,1997
INSTRUCTIONS TO APPLICANT:
Use the answer sheets provided to document your answers. Staple this cover sheet
on top of the answer sheets. Points for each question are indicated in parentheses
after the question. The passing grade requires a final grade of at least 80%.
Examination papers will be picked up four (4) hours after the examination starts.
TEST VALUE APPLICANT'S SCORE FINAL GRADE %
100.00
All work done on this examination is my own. I have neither given nor received aid. l
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Applicant's Signature
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SENIOR REACTOR OPERATOR Prga 2
ANSWER SHEET
Multiple Choice (Circle or X your choice)
. lf you change your answer, write your selection in the blank.
MULTIPLE CHOICE O23 a .b cd __
001 a b c d 024 a b c d
002 a b'c d _
O25 a b c d
003 ' a b c d 026 a b c d
004 a bc d 027 a b c d
005 a b c d 028 a b cd
006 a. b c d 029 a b c d ;
007 a b c d 030 a b c d
008 a b c d 031 a b c d
009 a b c d 032 a b c d
010 a b c d 033 a b c d
'011 a b c d 034 a b c d
, . 012 a b c d 035 a b c d
013 a b c d 036 a b c d
1 014 a b c d 037 a b c d
015 a b c d 038 a b c d
016 a b c d 039 a b c d
017 a b c d 040 a b c d
018 a b c d- 041 a b c d
019 - a b c d 042 a b c d
020 a b cd 043 a b c d
021 a b c d 044 a b c d
022 a. b c d 045 a b c d
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SENIOR REACTOR OPERATOR - Pzgn'3
ANSWER SHEET-
Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blank. ,
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046'.a b c d 069 a. b c d l
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047 a'b c d_ 070 a b c-d- )
048 ab c d 071 a b c d
049 a bc d 072 a b c d )
050 a b c -d 073 ab c d
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074 a b c d l
051 a b-c'd
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052' a b c d 075 a b c d
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053 a b c d 076 a b c d i
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054 a b c d 077 a b c d
055 a b c d 078 a b c d
056. a b c d 079 a b c d 1
057 a b c d 080 a b c d
058 a b c d 081 a b c d
- 059' a b c d 082 a b c d
, 060 a b c d 083 a b c d
061 a b c d 084 a b c d
062 a b c d 085 a b c d :
063 a b c d 086 a b c d
064 a b c d 087 a.b c d
065 a b c d 088 a b c d
066 a b~ c d 089 a b c d
067 a-b c d 090 a b cd
068 -a b c d 091 a b c d
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SENIOR REACTOR OPERATOR Prg3 4
ANSWER SHEET
Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blank.
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092 a - b c. d
093 a b c-d _
094 a b c d _
095 a b c d _
096 a b c d _
097 a b c d _
098 a b c d _
099 a b c d _
100 a b c d _
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( * * * * * * * * * * END OF EX AMIN ATION * * * * * * * * * * )
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS
During the administration of this examination the following rules apply:
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1. Cheating on the examination means an automatic denial of your application
and could result in more severe penalties.
. 2. . After the examination has been completed, you must sign the statement on '
the cover sheet indicating that the work is your own and 'you have not
received of given assistance in completing the examination. This must be
done after you complete the examination.
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3. Restroom trips are to be limited and only one applicant at a time may leave.
You must avoid all contacts with anyone outside the examination room to
avoid even the appearance or possibility of. cheating. ,
! 4. Use black ink or dark pencil ONLY to facilitate legible reproductions.
5. Print your name in the blank provided in the upper right-hand corner of the
examination cover sheet and each answer sheet.
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~ 6. Mark your answers on the answer sheet provided. USE ONLY THE PAPER
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PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAGE.
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7. The point value for each' question is indicated in parentheses after the ;
question.
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t 8. If the intent of a question is unclear, ask questions of the examiner only.
9. When turning in your examination, assemble the completed examination with
examination questions, examination aids and answer sheets in addition,
, turn in all scrap paper.
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10. Ensure allinformation you wish to have evaluated as part of your answer is
on your answer sheet. Scrap paper will be disposed of immediately
- following the examination.
i 11. To pass the examination, you must achieve a grade of 80% or greater.
12. There is a time limit of four (4) hours for completion of the examination.
13. When you are done and have turned in your examination, leave the
- examination area (EXAMINER WILL DEFINE THE AREA). If you are found in
this area while the examination is stillin progress, your license may be
i denied 'or revoked.
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SENIOR REACTOR OPERATOR Pcgs 7
. QUESTION: 001 (1.00)
The HPCI system has automatically initiated due to a low reactor water
level. Drywell pressure remains within normal limits. The HPCI turbine
slowly lowers reactor pressure. Reactor pressure continues to decrease
and reaches 80 psig.
Which ONE of the following is the expected automatic response?
a. Group IV isolation, but no HPCI turbine trip and no Group Vil
isolation
b. Group IV isolation and HPCI turbine trip, but no Group Vil
isolation
c. Group Vilisolation and HPCI turbine trip, but no Group IV
isolation
d. Group IV solation, Group Vil isolation, and HPCI turbine trip
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OUESTION: 002 (1.00)
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Which ONE of the following signals will NOT require resetting the Trip
and Throttle Valve to restart the RCIC turbine?
! a. RCIC Turbine Mechanical Overspeed
b. Reactor high water level
c. Manual trip pushbutton on 904 panel
- d. High Steam Supply line differential pressure
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QUESTION: 003 (1.00)
With the plant operating at 100% power, the 'A' Core Spray Line Break
Detection Monitor is reading approximately -3.0 psid.
This reading is:
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a. Indicative of an 'A' Core Spray Line break inside the shroud.
b. indicative of an 'A' Core Spray Line break outside the shroud.
- c. normal due to the differential pressure across the dryers and
separators being approximately -3.0 psid at 100% power.
d. normal due to changes in water density after the instrument was
calibrated to read zero under cold conditions.
QUESTION: 004 (1.00)
The following conditions exist:
- SBLC Tank Temperature 45 Degrees F
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SBLC Tank Volume 4000 gallons
- SBLC Tank Concentration 9.1% weight %
- B-10 Isotope Enrichment 53 %
What is(are) the MINIMUM required action (s) that you as the NWE should
immediately initiate?
a. Perform a SBLC flow test.
b. Determine whether the sodium pentaborate solution meets the
original design criteria.
c. Perform a SBLC flow test and determine whether the sodium
pentaborate solution meets the original design criteria.
d. Immediately commence a plant shutdown such that the plant can
reach cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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QUESTION: 005 (1.00)-
Given the following conditions:
- The plant is in cold shutdown.
- An RHR system test is in progress.
- The LPCI Override Control Switch (S178) has been taken to
MANUAL OVERRIDE.
- Drywell pressure is at O psig and steady.
- As part of the test, reactor water level is simulated at -60
inches.
- The next step in the procedure is to take the control switch
for the Torus Spray Valve MO-1001-37B to open.
Which ONE of the following explains why MO 1001-37B will NOT
open when the control switch is taken to open?
a. The 15 minute time delay is not timed out.
b. Drywell pressure is at atmospheric.
c. The RPV Level Override Keylock Switch (S188) is not in MANUAL j
OVERRIDE.
d. The 5 minute time delay has not timed out and the MO-1001-28B i
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is not closed.
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SENIOR REACTOR OPERATOR Pegs 10
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QUESTION: 006 (1.00)
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The following conditions exist:
- The plant is operating at 100% power.
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The "A" SBGT Fan is in AUTO. -
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The "B" SBGT Fan is in STBY. 3
- A valid SBGT Initiation signal occurs.
- The "A" SBGT Fan initially starts, runs for 10 seconds, then
trips for an unknown cause.
Which ONE of the following describes the expected automatic response of
the "B" SBGT Fan?
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The "B" SBGT Fan will:
a. start when the initiation signal is received, run for 65
. seconds and then stop, then restart.
b. start immediately after the "A" SBGT Fan trips and continue
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running uninterrupted
c. start after 65 seconds and continue to run uninterrupted.
d. start when the initiation signal is received and continue
running uninterrupted. j
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SENIOR REACTOR OPERATOR Prgs 11
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QUESTION: 007 (1.00). ,
The following conditions exist:
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The plant is at 100% power.
- it is determined that the 3A SRV will NOT open under ANY
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condition.
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Which ONE 'of the following states the MINIMUM action REQUIRED by
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Technical Specifications?
a. Place the plant in Cold Shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b. Reduce reactor coolant pressure below 104 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
c. Provided HPCI is operable, enter 14 day LCO. When this LCO is
expired, place the plant in Cold Shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
d. Provided HPCI is operable, enter 14 day LCO. When this LCO is
expired, reduce reactor coolant pressure below 104 psig within
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24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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QUESTION: 008 (1.00)
During a high drywell pressure condition, a valid ADS signal exists and
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. the ADS system has initiated. With the initiation signal still present
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both initiation Signal Timer Reset Pushbuttons are depressed.
Which ONE of the following describes the expected automatic response of
the ADS system?
j All ADS valves will:
a. remain open.
b. close and remain closed indefinitely.
c. close and remain closed for 105 seconds then reopen.
d. close and remain closed for 11 minutes then reopen.
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SENIOR REACTOR ' OPERATOR P:ga 12
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QUESTION: OO9 (1.00)
The following conditions exist:
- The "A" and "B" Reactor Feed Pumps are in service.
- - Both Reactor Recirculation Speed demands are at 60%. -
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- An instrument failure causes the Feedwater Regulating Valves to
reduce feedwater flow to 3 Mlbm/hr with reactor water level
reaching a minimum of + 17 inches.
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- The operator takes manual control of the Feedwater Regulating
Valves and is returning water level to normal with a current
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level of + 18 inches.
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- No operator action is taken on the Reactor Recirculation
System.
Which ONE of the following describes the expected response of the l
. Recirculation Flow Controllers? '
'The Recirculation Flow controllers will demand lowering speed to
a. 44% without a rate limitation signal.
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.b. 44% at a rate of 1.5% per second.
c. 26% without a rate limitation signal.
d. 26% at a rate of 1.5% per second.
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' QUESTION: 010 (1.00)
An Emergency Diesel Generator (EDG) has started due to a LOCA signal.
Which ONE of the following signals will cause an EDG trip?
a. Engine Overspeed
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b. Engine Low Lube Oil Pressure
c. Engine High Lube Oil Temperature
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d. Engine Crankcase High Vacuum
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QUESTION: 011 (1.00) l
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The following conditions exist:
- HPCI is injecting water from the CST to the RPV. l
- The HPCI Suction Valves From Suppression Chamber MO-2301-35 and-
MO-2301-36 Control Switches are in Auto.
- The CST Low Level Alarm comes in.
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Which ONE of the following describes the expected response of the HPCI
system?
The HPCI Suction From CST MO-23016 will receive a close signal-
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a. as soon as both the MO-2301-35 and the MO-2301-36 valves reach
. full open.
- b. as soon as both the MO-2301-35 and MO-2301-36 valves come off
their closed seats,
c. as soon as either the MO-2301-35 or the MO-2301-36 valve
reaches full open. l
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d. at the same time the MO-2301-35 and MO 2301-36 valves receive ;
an open signal. )
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QUESTION: 012 (1.00)
Which ONE of the following states where the RCIC turbine receives stearn
and where the RCIC pump discharges?
a. Steam from "C" Main Steam Line and Discharge to "A" Feedwater
- Line
b. Steam from "D" Main Steam Line and Discharge to "B" Feedwater
- Line
c. Steam from "D" Main Steam Line and Discharge to "A" Feedwater
Line
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d. Steam from *C" Main Steam Line and Discharge to "B" Feedwater
Line
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Pegs 14
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OUESTION: 013 (1.00)
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Which of the,following SRM rod block (s) is(are) bypassed by moving IRM
range switches from Range 2 tc Range 3? l
a. SRM Downscale Rod Block only -
b. SRM Inoperable Rod Block only ;
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c. SRM Downscale Rod Block and Detector Retract Not Permitted Rod
Block
d. SRM inoperable Rod Block and SRM Downscale Rod Block
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QUESTION: 014 (1.00) ,
With Reactor Power at'100%, an SRV spuriously lifts. Action to close !
the valve are successful. Immediately after valve closure, the j
downstream temperature is checked.
Which ONE of the following is an expected approximate downstream
temperature?
a. 212 degrees F
b. 295 degrees F..
c. 375 degrees F .l
d. 525 degrees F
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SENIOR REACTOR OPERATOR P gs 15
-QUESTION: 015 (1.00)
The following conditions exist:
- A half scram exists on RPS "A" due to APRM testing.
- A fire caused a loss of RPS Bus "B" and a full scram.
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- The half scram testing was stopped and APRMs were returned to
normal.~
- The SCRAM DISCHARGE INSTRUMENT VOLUME HI LEVEL SCRAM BYPASS
switch is then taken to bypass.
Which ONE of the following describes when the RPS "A" half scram may be
reset?
a. immediately.
b. after the air dump test switch is placed in isolate.
c. after the SDIV vent and drain valves come fully open.
d. after RPS "B" is energized.
!
l
QU' - ION: 016 (1.00)
The mode switch is in RUN. Which ONE of the following scram signals is
automatically bypassed 2 seconds after taking the mode switch to
SHUTDOWN?
, i
a. Mode switch in shutdown
b. Main steam isolation valve closura
c. Turbine stop valve closure
d. Scram discharge volume high level
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'
SENIOR REACTOR OPERATOR. Pags 16
QUESTION: 017 (1.00).
. The ATWS logic system has automatically initiated due to low reactor
water level.
~
Which ONE of the following actuations will be delayed by 9 seconds?
a. Rod insertion
b. Reactor Recirc Pump Field Breaker Trip
'
c. Reactor Recirc Pump Drive Motor Breaker Trip
d. Reactor Feed Pump Trip
{
QUESTION: 018.(1.00)
'
. During a reactor shutdown, the control rod selected on the Rod Select
Matrix is NOT in the rod group of the latched step. As reactor power
' decreases, at what point will this condition cause an insert and
withdraw block?
a. Steam Flow drops below 35%
b. All APRM readings drop below 20%
c. Steam Flow or Feed Flow drops below 20%
d. Steam Flow and Feed Flow drop below 35%
,
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SENIOR REACTOR OPERATOR P:gs 17
QUESTION: 019 (1.00)
The following conditions exist:
- The plant is operating at 100% power.
- APRM "C".is bypassed for maintenance.
- APRM "E" then fails giving a constant reading of 95% regardless
of input.
- A half scram already exists on RPS "B"
Which ONE of the following meets the action REQUIRED?
a. Initiate insertion of operable rods and complete insertion of
all operable rods within sixteen hours.
b. Reduce power level to IRM range and place mode switch in the
startup/ hot stand'y position within eight hours.
c. Reduce turbine load and close main steam isolation valves
within eight hours,
d. Reduce power to less than 45% of design.
QUESTION: 020 (1.00)
A TIP trace is being performed when a high drywell pressuro signal
occurs. Select the expected automatic action.
a. The shear valve fires with the detector stillin the core.
b. The ball valve closes with the detector stillin the core.
c. The detector withdraws into its shield and the ball valve
closes,
d. The detector withdraws into its shield and the shear valve
fires.
,
e- A ,m. asa _ -ire.d e 4 # ww .e4.1 - ,a 4 4 +4 -
4- beM iJ---'a-
,
SENIOR REACTOR OPERATOR Pi:gs 18
.
QUESTION: 021 (1.00)
.
'
While operating at 80% power, an instrument failure causes the throttle
pressure sensed by the EPR to fail high. No operator action is taken.
Which ONE of the following is the expected result? i
a. The reactor would scram on a high pressure scram signal. t
b. The MPR would take control and pressure would increase by
approximately 10 psi.
c. The reactor would scram on a low pressure scram signal.
'
d. The reactor would scram on a MSIV closure scram signal.
QUESTION: 022 (1.00)
At 500 psig during a reactor startup and heatup, the #1 Bypass Valve
(BPV) comes partially open. 1
i
'
Which ONE of the following errors is the cause?
Failure to maintain the:
,
l
a. EPR 40-80 psig below reactor pressure l
l
b. EPR 40-80 psig above reactor pressure ;
I
c. MPR 40-80 psig below reactor pressure
d. MPR 40-80 psig above reactor pressure
!
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SENIOR REACTOR OPERATOR P:ga 19
-
QUESTION: 023 (1.00)'
With the plant operating at 20% power both Reactor Recirculation Pumps
irip. The operator manually scram the reactor. Post scram, fuel zone
level indicators read:
a. falsely high since less flow exists through the jet pumps than
existed during calibration conditions.
, b. falsely low since less flow exists through the jet pumps than ;
existed during calibration conditions.
c. falsely low due to decreased density of the water in the vessel
against calibrated conditions,
d. falsely high due to decreased density of the water in the
vessel against calibrated conditions.
QUESTION: 024 (1.00)
1
With the "A" Loop of RHR in Lo oling, RPV level decreased to 12 l
inches. The Shutdown Cooling Outtsumo I:,ulation Valve MO-1001-47 l
stopped in mid-stroke. All other valves have responded as expected. i
!
Which ONE of the following is REQUIRED in order to open the "A" Loop !
LPCI Injection Valve #2 MO-1001-29A?
a. The MO-1001-47 valve must be closed,
b. The MO-1001-29A must be manually reset,
c. Reactor coolant pressure must be greater than 76 psig,
d. The Group ll isolation signal must clear and the Group 11 logic l
must be reset.
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SENIOR REACTOR OPERATOR PJgn 20
QUESTION: 025 (1.00)
A reactor scram has occurred. Electrical busses A-5 and A-6 have
transferred to the Start-up Transformer. Which ONE of the following
-describes the drywell cooler response?
a. The running drywell coolers will trip and start after a 45
second time delay. The drywell coolers in standby remain in
standby,
b. The running drywell coolers will trip. The drywell coolers in
standby will start after a 45 second time delay.
c. The running coolers will stay in service. The drywell coolers
in standby willimmediately start when A-5 and A-6 are
reenergized.
d. The running coolers will stay in service. The drywell coolers
in standby will start after a 45 second time delay.
!
QUESTION: 026 (1.00) l
Primary Coolant Temperature is 245 degrees F when Shutdown Cooling is
placed in service, immediately thereafter, a fire disables the Shutdown
Cooling Outboard Isolation Valve MO-1001-47 motor operator. The valve 1
is in the open position. I
Which ONE of the following meets the MINIMUM REQUIRED action?
a. Verify the ability to manually close the MO-1001-47 valve, then
reestablish shutdown cooling. )
)
b. Verify the ability to close the MO-1001-50 valve, then
reestablish shutdown cooling,
c. Close either the MO-1001-47 or MO-1001-50 valve and open the
respective breaker.
d. Station an operator to manually close the MO-1001-50 valve if
required and continue in shutdown cooling.
,._
,
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SENIOR REACTOR OPERATOR Paga 21
QUESTION: 027 (1.00)
l
When valving in a CRD hydraulic control accumulator, the 305-102
'(5Nithdraw Riser Isolation Valve) and the 305-112 (Scram Discharge Riser
isolation Valve) are required to be open prior to opening the 305-101 1
l
(Insert Riser isolation Valve). This prevents-
a. .a single rod scram when opening the 305-101 valvo. I
b. excessive scram time of that rod in the event of a reactor
scram. l
!
c. damage to the accumulator in the event of a reactor scram. I
d. damage to the drive mechanism in the event of a reactor scram. 1
1
)
Il
l
QUESTION: 028 (1.00) l
l
While operating at 100% power, a control rod is determined to be i
uncoupled. Attempts to couple the rod have been unsuccessful. l
l
Which ONE of the following states the MINIMUM REQUIRED actions?
a. Verify that the control rod can be moved with drive pressure
and maintain the control rod at the target position.
b. Fully insert the control rod and hydraulically disarm the CRD.
l
'
c. Fully insert the control rod and electrically disarm the l
directional control valves. l
l
d. Fully insert the control rod, electrically disarm the )
directional control valves and then declare the rod inoperable. i
!
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, SENIOR REACTOR OPERATOR Pzga 22 i
QUESTION: 029-(1.00)
With the plant at pnwer, it is determined that the MO-1001-37B (B Loop
Torus Spray) and MO-1400-25A (A Loop Core Spray Inboard injection)
valves have failed their operability test. Both volves are currently
closed.
-
The maximum time allowed before the plant must be in COLD SHUTDOWN is:
a 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (1 day).
96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> (4 days),
c. ~ 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> (7 days).-
d.192 hours0.00222 days <br />0.0533 hours <br />3.174603e-4 weeks <br />7.3056e-5 months <br /> (8 days).
QUESTION: 030 (1.00)
A tagout, which has been in effect on the "A" Reactor Recirculation Pump
for 7 days, has just been cleared. The "A" Reactor Recirculation Pump
is started and immediately manually tripped. On the second start
p attempt, the pump starts and runs for 10 minutes and then is manually
,
tripped.
When is the SOONEST that another start of the "A" Reactor Recirculation
'
Pump may be attempted?
a. Immediately
b.15 minutes after the second trip
c. 45 minutes after the second trip
d. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the second trip
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SENIOR REACTOR OPERATOR Pags 23
I
QUESTION: 031 (1.00)
l
The plant is operating at 100% power when the "B" Reactor Recirculation i
Pump trips. No operator action is taken.
Which ONE of the following describes the initial steady state to final
steady state change in the "A" Reactor Recirculation Loop Jet pump flow
and the reason for the change?
The "A" Reactor Recirculation Loop Jet pump flow will: l
l
a. increase due to lower core pressure drop. I
b. increase due to decreased core voiding. l
l
c. decrease due to higher core pressure drop.
d. decrease due to increased core voiding.
1
QUESTION: 03 (1.00)
While operating at 0% power, it is determined that th Main Steam Line
High Flow switches o the "B" Main Steam Line will N trip under a high
flow condition.
Which ONE of the following the MINIMUM RE IRED action? r
a. Direct l&C personnel to m ually trip t inop blejpi Eiles. f
b. Direct l&C personnel to manu ly i ert a ha f
isolation on the "B" Group 1 Cha 1. (j p s
c. Initiate an orderly shutdown db in Cold Shutdown Condition
within a MAXIMUM of 30 h urs afte he instrument failure.
d. Initiate an orderly shutd n and have th Main Steam Lines
isolated within a MAXI UM of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> a r the instrument
failure.
. . . - .-. - . . . - . - - - . ~ . . . .
SENIOR REACTOR OPERATOR. P:ge 24
QUESTION: 033-(1.00)
Depressing a.n outboard MSIV test pushbutton will:
a. energize the AC test valve and vent air from the underside of
the piston.
b. energize the AC test valve and admit air to the underside of
the piston.
,
c. deenergize the AC test valve and vent air from the underside of
the piston. -
' d. deenergize the AC test valve and admit air to the underside of r
the piston.
QUESTION: 034 (1.00)
Which ONE of the following conditions requires Rod Block Monitor
Operability?
a. MCPR is 1.35 and Reactor Power is 25%.
b. MCPR is 1.45 and Reactor Power is 75%.
c. MCPR is 1.55 and Reactor Power is 95%.
d. MCPR is 1.60 and Reactor Power is 100%.
t
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. SENIOR REACTOR OPERATOR P g) 25
QUESTION: 035 (1.00)
A loss of 120V Bus A (Y-3) occurs.
Which ONE of the following describes the effect on the RWCU system?
a. Half of the logic for closing the MO-2 and MO-5 valves is made
up.
b. MO-2 goes closed. As soon as MO-2 comes off the open seat, the
operating RWCU pump (s) will trip. MO-5 remains open.
c. MO 2 goes closed. As soon as MO 2 comes off the open seat, the ,
operating RWCU pump (s) will trip and MO-5 will go closed. j
d. MO-5 goes closed. As soon as MO-5 comes off the open seat, the
operating RWCU pump (s) will trip. MO-2 remains open.
QUESTION: 036 (1.00)
The OFF GAS ISOL CH PRM SEL switch is in position 2. Which ONE of the
following conditions of the Air Ejector Off Gas Radiation Monitors will 4
cause tiie 13 minute timer to initiate?
a. Hi radiation signal on both channels
b. Hi Hi radiation signal on one channel
c. Hi radiation signal on one channel and Downscale Trip on the !
other channel
]
d. Downscale trip on one channel and inop trip on the other
channel
1
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l- SENIOR REACTOR OPERATOR P ga 26
!
QUESTION: 037 (1.00)
l
A reactor startup is in progress with reactor power in the intermediate
range. IRM "A" then starts to intermittently swing upscale and then
downscale.
Which ONE of the following conditions on IRM "A" will cause a Rod Block
but NOT cause a Half Scram?
The IRM reads:
a.1 (on the 0-40 scale) while on range 1. *
,
b. 3 (on the 0-40 scale) while on range 3. i
1
I
c. 36 (on the 0-40 scale) while on range 5.
d. 39 (on the 0-40 scale) while on range 7. !
QUESTION: 038 (1.00)
With the Mode Switch in Startup, at 1200 on 5/5/97, the Downscale Trips
for IRM Channels "A", "B", and "E" are made inoperable. ]
Which ONE of the following is the LATEST that one of these channels must
be placed in a tripped condition?
a. 1300 on 5/5/97 i
b.1200 on 5/6/97
c.1200 on 5/12/97
d. 1300 on 5/12/97
,
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SENIOR REACTOR OPERATOR PIga 27 '
OUESTION: 039 (1.00)
The plant was. operating at 100% power with the "B" CRD pump in service.
Subsequently, a valid LOCA signal generated a scram. The plant
responded as expected except, the startup transformer feeder breaker
to bus A 5 failed to close. A-5 has been automatically energized from the
shutdown transformer.
Which ONE of the following describes the status / availability of the CRD pumps?
a. "B" CRD pump is running.
"A" CRD pump can be started since no load shed signal was
generated.
b. "B" CRD pump is not running.
"A" and "B" CRD pumps cannot be started due to load shed
signal.
c. "B" CRD pump is not running.
"A" and "B" CRD pumps can be started since no load shed signal
was generated.
,
d. "B" CRD pump is running.
"A" CRD pump cannot be started due to load shed signal.
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- SENIOR REACTOR OPERATOR Pign 28
,
QUESTION: 040-(1.00)-
Which ONE of the following administrative precautions related to valves
are required when lining up RHR "A" loop for shutdown cooling?
a. MO-1001-7A "RHR PUMP A TORUS SUCTION" red tag closed.
MO-1001-7C "RHR PUMP C TORUS SUCTION" red tag closed. ,
MO-1001-43A "RHR PUMP A SHUTDOWN COOLING SUCTION yellow tag
closed.
MO-1001-43C "RHR PUMP C SHUTDOWN COOLING SUCTION yellow
tag closed.
b. MO-1001-7A "RHR PUMP A TORUS SUCTION" red tag closed.
MO-1001-7C."RHR PUMP C TORUS SUCTION" red tag closed.
MO-1001-438 "RHR PUMP B SHUTDOWN COOLING SUCTION red tag
closed.
MO-1001-43D "RHR PUMP D SHUTDOWN COOLING SUCTION red tag
closed.
c.1001-6A "RHR PUMP C SUCTION VALVE FROM THE TORUS" red tag
closed.
1001-366A "RHR PUMP A SUCTION VALVE FROM THE TORUS" red tag
closed.
MO-1001-43B "RHR PUMP B SHUTDOWN COOLING SUCTION red tag
closed.
MO-100143D "RHR PUMP D SHUTDOWN COOLING SUCTION red tag
closed.
d.1001-6A "RHR PUMP C SUCTION VALVE FROM THE TORUS" yellow tag
closed.
1001-366A "RHR PUMP A SUCTION VALVE FROM THE TORUS" yellow
tag closed.
MO-1001-43A "RHR PUMP A SHUTDOWN COOLING SUCTION yellow tag
closed. ;
MO 1001-43C "RHR PUMP C SHUTDOWN COOLING SUCTION yellow tag
closed.
_ _
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_- _
SENIOR REACTOR OPERATOR Pags 29
1 QUESTION: 041 (1.00)
_
Given the following conditions:
.
The plant is in cold shutdown
No recirculation pumps are in service
RHR pump "A"is in shutdown cooling
RWCU is in service
Reactor shutdown level instrument indicates 40 inches
Which ONE of the fo:iowing describes reactor coolant temperature
indication if the "A" RHR pump trips. Assume no operator action,
a. Recirc loop "A" temperature indicator is representative of
reactor coolant temperature.
b. Recirc loop "B" temperature indicator is representative of
reactor coolant tempesture.
c. RWCU bottom head drain temperature indicator is representative
of reactor coolant temperature.
d. No temperature indicator is representative of reactor coolant
temperature,
,
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d
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,
SENIOR REACTOR OPERATOR Prgs 30
.
- QUESTION: 042 (1.00)
.The following conditions exist:
!' - EOP-02 is being executed _. .
-
-
The Mode Switch is in Shutdown and ARI has been initia'ted.
'
- The MSIVs are closed.
- Reactor power is 2.5% and no boron has been injected.
- ~ Alternate _Depressurization is required by EOP-04.
- Four SRVs can be opened.
Which ONE of the following actions should be taken to control reactor '
I water level?
4
a. Secure all sources of injection. When pressure decreases below
200 psig, slowly inject with LPCI.
b. Secure all sources of injection. When pressure decreases below
- . 400 psig, slowly inject with the Condensate Pumps.
c. Secure all sources of injection except CRD and RCIC. When
, pressure decreases below 200 psig, continue injection flow rate
d. Secure all sources of injection except CRD and RCIC. When
pressure decreases below 270 psig, slowly inject with the
.
Condensate Pumps.
4
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0
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1
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SENIOR REACTOR OPERATOR Pcgs.31 i
,
l
.)
QUESTION: 043 (1.00)
The following conditions exist: l
- A manual' scram was inserted from 20% power. l
- No other scram signals exist. !
l
- Reactor power is on intermediate range 6 and decreasing.
- Three control rods are at position 06. All other rods are ,
fully inserted.
Which ONE of the following is the required action?
a.' ' enter PNPS 2.1.6. No EOP entry is required.
b. ' enter EOP-01, then exit EOP-01 and enter EOP-02 at R-1. ,
c. enter PNPS 2.1.6, " Reactor Scram", then exit PNPS 2.1.6 and
.
enter EOP-02 at R 1.
d. enter PNPS 2.1.6, " Reactor Scram", then enter EOP-02 and
execute concurrently with PNPS 2.1.6.
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< SENIOR REACTOR OPERATOR pig 3 32-
s
QUESTION: 044 (1.00)
The following conditions exist:
- EOP-02 is being executed.
-- Boron is being injected with the SBLC system. '
- Initial SBLC tank level was 4100 gallons. ;
Reactor Water Levelis being lowered to reduce reactor power. ;
- Current SBLC tank level is 3000 gallons.
- Torus water temperature is 112 degrees F.
- Reactor water level is'-100 inches.
Which ONE of the following actions is REQUIRED 7
a. Reise reactor water level to the +12 to +45 inch band and
perform Alternate Depressurization.
'
b. Raise reactor water level to the + 12 to +45 inch band. Do not
perform Alternate Depressurization. !
!
c. Maintain reactor water level at its current value and perform
. Alternate Depressurization.
d. Maintain reactor water level at its current value. Do not
perform Alternate Depressurization.
QUESTION: 045 (1.00) {
l
While operating at 100% reactor power, reactor pressure starts to
oscillate approximately 10 psi peak to peak and pressure control is
shifting alternately from the EPR to the MPR and back to the EPR.
Which ONE of the following actions are REQUIRED?
a. DJace the EPR control switch to off,
b.' Reduce reactor power to approximately 75%.
. c. Raise the MPR setpoint to prevent pressure control from
.
swapping between regulators.
,
d. Lower.the MPR setpoint to allow the MPR to take control of
pressure.
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. . - - . _ -. - - - -_ - -. - - . ..
SENIOR REACTOR OPERATOR Pcga 33
,
t
QUESTION: 046 (1.00)
With the plant in Cold Shutdown, some solvent that is improperly stored
in a Control Room locker ignites. The Nuclear Watch Engineer makes the
decision to evacuate the Control Room and to call for off-site
assistance to put out the fire. Control is established at remote -
shutdown stations 20 minutes after the Control Room evacuation.
What is the MINIMUM event level classification?
a. Unusual Event
b. Alert
'
c. Site Area Emergency
d. General Emergency
,
QUESTION: 047 (1.00)
A LOCA has occurred. Which ONE of the following REQUIRES exiting the
RPV level control leg of EOP-017
a. Reactor water level is -165 inches and increasing with Reactor
Pressure at 200 psig.
b. Reactor water level is -125 inches and decreasing with Reactor
Pressure at 175 psig.
c. Reactor water level is -125 inches and increasing with Reactor
Pressure at 100 psig.
d. Reactor water level is -125 inches and decreasing vdth Reactor
Pressure at 75 psig.
,.
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- SENIOR REACTOR OPERATOR P gs 34
QUESTION: 048 (1.00)
~ Which ONE of the following conditions REQUIRES Alternate Reactor- ,
Pressure Vessel Depressurization assuming a primary system is ' ,
dischaiging into secondary containment?
.
a. RCIC torus piping area temperature is 300 degrees F and RCIC
'
,
turbine area temperature is 195 degrees F.
b. HPCI compartment water levelis 8 inches and HPCI turbine area ,
temperature is 195 degrees F. .:
c. RHR "B" and "D" pump area temperature is 300 degrees F and RHR
."A" and "C" pump area temperature is 195 degrees F.
, d. Main Steam Tunnel area temperature is 300 degrees F and RHR "A"
and "C" pump area temperature is 220 degrees F.
,
QUESTION: 049 (1.00)
Following a Nitrogen Line leak in the drywell, AO-4356 (Nitrogen / Air
Isolation Valve to the Drywell) was closed. By calculation, how many
times over the next eight hours can each SRV be actuated?
NOTE: COUNT EACH OPEN AND CLOSE CYCLE AS ONE ACTUATION
a. 5 ,
b. 10
c. 20
4
d. 40
.
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SENIOR REACTOR OPERATOR P ga 35
QUESTION: 050 (1.00)
in the event that torus water level cannot be maintained above 95
inches, HPCl is secured in order to prevent:
a. exceeding the Primary Containment Pressure Limit. '
b. exceeding the Pressure Suppression Pressure.
c. exceeding the Heat Capacity Temperature Limit,
d. isolating HPCI on high exhaust pressure.
' QUESTION: 051 (1.00)
The following conditions exist:
- Reactor pressure is 10 psig.
- Drywell pressure is 4 psig.
- ' Torus bottom pressure is 15.2 psig.
-
Torus water level is 303 inches.
Select the correct action and its reason.
.
Under these conditions:
a. Alternate RPV Depressurization is required to prevent SRV Tail
Pipe failure,
b. Suppression Chamber Spray initiation is required using enly
those RHR pumps not required to provide adequate co cooling.
c. Suppression Chamber Spray initiation is not allowed sinc. the
Torus Spray Sparger is covered.
d. Suppression Chamber Spray initiation is not ai. owed since the
Torus-Drywell Vacuum breakers are covered.
.
,
' SENIOR REAC f0R OPERATOR P;gs 36
QUESTION: 052 (1.00)
A trip of the "A" Reactor Recirculation Pump has occurred. The plant is
operating in Region ll of the Power / Flow Map after the immediate actions
2
of 2.4.17 have been completed.
Which ONE of the following is REQUIRED 7
Exit Region 11 by:
a. manually scramming the reactor.
b. restarting the "A" Reactor Recirculation Pump.
c. increasing the speed of the "B" Reactor Recirculation Pump.
d. inserting control rods in reverse order of the pull sheet.
l
QUESTION: 053 (1.00) !
l
The following conditions exist: l
l
-
Torus water level is 105 inches.
- Torus water temperature is 180 degrees F.
- Reactor pressure is 700 psig.
Which ONE of the following states whether Alternate RPV Depressurization
is required, not required, or prohibited and the reason.
Under these conditions, Alternate RPV Depressurization is
a. not required since primary containment limits are not exceeded,
b. required to ensure the energy released during an RPV blowdown I
can be accepted.
.
. c. required since the downcomers are now exhausting to the torus
free air space.
d prohibited since the SRV Tail Pipes are now exhausting to the
torus free air space.
!
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SENIOR REACTOR OPERATOR Paga 37
OUESTION: 054 (1.00)
To initiate a reactor scram when the control room has been evacuated, it
is undesirable to deenergize the RPS busses as the means of scramming
because:
,
a. ' nuclear instrumentation needed to monitor reactor power will
become denergized.
b. pressure control using turbine bypass valves will be lost after
the scram.
c. RPV level control will unnecessarily transfer from feedwater to
HPCI.
I
d. groups I, ll, Ill, and VI isolations will be defeated. j
QUESTION: 055 (1.00)
The following conditions exist:
- The plant is operating at 75% power.
- At 0800 one Safety Relief Valve opened.
- At 0802 EOP-03 has been entered due to torus water temperature
reaching 80 degrees F.
At what point should a Reactor Recirculation pump speed reduction and
manual reactor scram be performed?
a. Immediately when it is determined that the SRV cannot be
reclosed.
b. At 0810.
c. When torus temperature reaches 120 degrees F.
d. When the " unsafe"_ region of the Heat Capacity Temperature Limit
curve is entered.
-
- ...-. . . .. .. .- - . - . - - . - - .. - . . . .
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SENIOR REACTOR OPERATOR P:go 38 '
i
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,
QUESTION: 056 (1.00)
Drywell spray was initiated in accordance with EOP-03. As drywell 1
temperature and pressure are decreasing, the unacceptable region on the
Drywell Spray initiation Limit curve is entered at a Drywell temperature
of 250 degrees ~ F.
Which ONE of the following is the REQUIRED action?
a. Secure drywell spray when drywell pressure drops below 2.2
psig,
b. Secure drywell spray when torus bottom pressure drops below 2.2
psig.
c. Adjust drywell spray as necessary to maintain operation within
the Drywell Spray Initiation limit curve. ;
l
d. Immediately secure drywell spray. l
l
QUESTION: 057 (1.00)
A loss of feedwater heating has occurred. Which ONE of the following is
the REQUIRED immediate operator action?
Run back Reactor Recirculation flow until:
a. reactor power has been reduced 25% below its pretransient level
without regard to current total core flow,
b. total core flow has been reduced to 36 Mlb/hr without regard to
the current reactor power level,
c. reactor power has been reduced to at least 25% below its
pretransient level AND total core flow has been reduced to at
least 36 Mlb/hr.
d. reactor power has been reduced 25% below its pretransient level
OR total core flow has been reduced to 36 Mlb/hr.
-- -
_ , . , , . .. . .
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- SENIOR REACTOR OPERATOR Paga 39 -
,
QUESTION: 058 (1.00)'
A startup is in progress with reactor pressure at 900 psig when the "A"
CRD pumps trips and the "B" CRD pump cannot be started. .Two accumulator
alarms, both in the same nine rod array, illuminate.
D
Which ONE of the following is the required. action?
,
s. Manually scram the reactor.
b. Determine the cause of the alarms. If both alarms are due to
low gas pressure then manually scram the reactor,
c. Fully insert one of the rods with an accumulator alarm and 1
-
disarm its directional control valves.
, d. Enter LCO to be in Cold Shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.'
QUESTION: 059 (1.00)-
While operating at 100% power, a recirculation pump seal failure causes
EOP-03 entry on high drywell pressure and high drywell temperature.
Following initiation of suppression chamber spray, drywell pressure
- stabilizes at 4 psig, torus bottom pressure stabilizes at 8 psig, and
drywell temperature stabilizes at 175 degrees F.
Which ONE of the following actions is REQUIRED?
a. Declare an Unusual Event and initiate drywell spray in
'
i
accordance with the Primary Containment Pressure leg of EOP-03.
b. Declare an Alert and in:tlate drywell spray in accordance with
the Drywell Temperature leg of EOP-03.
. c. - Declare an Unusual Event. Do not initiate drywell spray.
d. Declare an Alert. Do not initiate drywell spray.
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- - ._ . . . _ . _ _ _ . . . . - _. . _ _
- SENIOR REACTOR OPERATOR Prgs 40
1
1
)
.
- QUESTION: 060 (1.00) I
1
With the plant operating on the 65% load line, condenser vacuum starts
to decrease. Reactor Recirculation Flow is reduced in accordance with -i
plant procedures. !
After the Reactor Recirculation Flow reductions the plant will be
operating,
a. in the scram region.
b. In the exit region.-
I
c. In the caution zone.
d. above the MELLA line.
<
QUESTION: 061 (1.00)
l
The plant is operating at 100% power when control rods start to drift.
i The MAXIMUM number of control rods in a nine rod array that are allowed
to drift WITHOUT REQUIRING tho mode switch to be placed in shutdown is:
a. one rod without regard to whether the rods are drifting in or
out.
b. two rods if rods are drat;ng in and one rod if rods are
- drif ting out. ,
l
c. two rods without regard to whether the rods arr liiting in or I
out.
[ d.~ three rods if rods are drifting in and two rods if rods are
drifting out,
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SENIOR REACTOR OPERATOR Pcgs 41
OUESTION: 062 (1.00)-
The plant is operating at power when a total loss of TBCCW occurs,
immediate actions are complete in accordance with plant procedures.
Which ONE of the following describes RPV pressure and level control?
a. ' RCIC is being used in the' level control mode and HPCI is being
.used in the pressure control mode.
b. HPCIis being used in the level control mode and RCIC is being
used in the pressure control mode,
c. HPCI is being used in the level control mode and SRVs are being
used to control pressure. RCIC remains shutdown.
d. RCIC is being used in the level control mode and SRVs are being
used to control pressure. HPCI remains shutdown.
QUESTION: 063 (1.00)
The plant is operating at 100% power when a loss of Bus A5 occurs.
Which ONE of the following action (s) is(are) required? ,
a. Reduce reactor power to maintain steam tunnel temperature below
'
160 degrees F.
b. If steam tunnel temperature exceeds 160 degrees F scram the
reactor and close the MSIVs.
c. If steam tunnel temperature exceeds 160 degrees F scram the
reactor. Maintain the MSIVs open.
d. If steam tunnel temperature exceeds 160 degrees F commence a
normal plant shutdown.
I
SENIOR REACTOR OPERATOR Pcga 42
I
,
QUESTION: 064 (1.00).
With the plant at 100% power on the 100% load line, reactor water level j
starts to decrease due to unknown causes. . Level is currently + 25 inches
and is trending down at 1/2 inch per minute. I
l
Which ONE of the following is the required action assuming water level
continues to fall? :
a. Insert rods using the RPR rods until below 70% load line, then l
reduce core flow to 36-40 Mlb/hr. .)
I
b. Insert rods using the RPR rods until below 70% load line, then
'
reduce recirculation pump speed to minimum.
c. Reduce recirculation pump speed to minimum, then insert rods as
necessary to exit the caution zone,
d. Reduce core flow to 36-40 Mlb/nr, then insert rods using the
RPR rods until below 70% load line, then reduce recirculation
pump speed to minimum.
.
!
QUESTION: 065 (1.00)
The plant is operating at 100% power with the "B" Reactor Recirculation
Pump scoop tube locked when a reactor scram occurs.
Which ONE of the following actions are REQUIRED?
a. Direct a licensed operator to manually position the "B" Reactor
Recirculation MG set scoop tube to minimum speed.
b. Direct any member of the operating crew to manually position
the "B" Reactor Recirculation MG set scoop tube to minimum
speed.
c. Unlock the scoop tube, if possible, then run the "B" Reactor
Recirculation pump to minimum speed.
d. Trip the "B" Reactor Recirculation Motor Generator Set.
,
, - , , ,
SENIOR REACTOR OPERATOR Pigs 43
QUESTION: 066 (1.00)
Given the following conditions:
- A fuel leak occurs and as a result the reactor is manually
scrammed.
- Due to the fuelleak, the CRD HCU east and west areas radiation
levels reach 1200 mR/hr and 1250 mR/hr respectively.
- The west Scram Discharge Volume vent and drain valves have
failed open.
Under these conditions, Alternate RPV depressurization is:
a. not required since the CRD HCU east and west areas are
considered the same area,
b. required in order to protect secondary containment from
failing.
c. required to allow the scram to be reset and the primary system
leak isolated,
d. not required since there is no primary system discharging into
i
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' SENIOR REACTOR OPERATOR Prga 44
4
OUESTION: 067 (1.00)
.
The following conditions exist:
-
A seismic event has caused the torus suction lines to both Core
Spray loops to crack downstream of the Core Spray Suction (MO-
1401-3) valves.
-
These cracks result in the water level in the SE and NW
- Quadrants to reach 8 inches and 10 inches above the floor
respectively.
- Efforts to lower the water level are only able to maintain
level.
- There is no primary system discharge into secondary
containment.
!
Which ONE of the following is required by EOP-047 ;
1
,
a. Isolate the Core Spray suction from the torus.
1
b. Maintain the Core Spray suction aligned to the torus,
c. Perform Alternate RPV Depressurization.
d. Transfer the Core Spray suction for both loops to the CST.
OUESTION: 068 (1.00) i
Which ONE of the following conditions violates secondary containment
integrity?
a. Both drywell personnel access doors are open.
b. Reactor water cleanup MO-1201-2 (RWCU Suction) valvo is failed '
open.
~
c. Reactor building ventilation is secured due to dampers failing
closed.
d. One refuel floor exhaust isolation damper is failed open with
the other refuel floor exhaust isolation damper open and fully
-
-. _
- SENIOR REACTOR OPERATOR Pega 45
OUESTIONi 069 (1.00)
The following conditions exist:
- A reactor startup is in progress
- The Reactor Mode Switch is in "Startup/ Hot Standby"
- Reactor pressure is 850 psig
The main turbine is tripped
-- A valid Group Iisolation has occurred
- All systems operated as designed
Which ONE of the following conditions caused the Group Iisolation?
a. Low main steam line pressure
b. Two main steam lines isolating
c. High main steam tunnel temperature
d. High reactor water level
1
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QUESTION: 070 (1.00) i
l
A steam leak in the drywell has occurred and the control room crew has
entered EOP-01 and EOP-03. TI-9019 and TRU-9044 on panel C903 are l
'
broken. In accordance with the data contained in the attached 2.1.27,
which ONE of the following is the instrument run temperature for the "A"
channel instruments?
a. 208 degrees F
b. 210 degrees F
c. 216 degrees F
d, 220 degrees F
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SENIOR REACTOR OPERATOR Pag 3 46 ,
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OUESTION: 071 (1.00) ,
!
,
initiating suppression chamber spray prior to torus bottom pressure l
'
reaching 16 psig prevents fatigue failure of
l
a. SRV Tail Pipes. -
b. Torus Drywell vacuum breakers.
.
c. downcomers. l
-
d. the Reactor Building-Torus vacuum breakers.
!
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QUESTION: 072 (1.00)
a
l The following conditions exist:
- A core off-load is in progress.
- The Refuel Bellows Seal Rupture alarm is received followed 2
minutes later by the Fuel Pool Low Level alarm.
- Currently an irradiated bundle has been removed from the core
but is still above the reactor vessel.
Which ONE of the following is the REQUIRED action?
a. Immediately evacuate the refuel floor and leave the bundle
hoisted above the reactor vessel,
b. Return the bundle to the in-core position that it came from, j
c. Place the bundle in the nearest open in-core position.
d. Place the bundle in the nearest open fuel pool position.
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SENIOR REACTOR OPERATOR pig 3 47
OUESTION: 073 (1.00)
The following conditions exist:
-
The reactor is shutdown.
-
At 1600 all RPV water level indication was lost due to
electrical problems and EOP-16 was entered.
- At 1630 conditions to flood the RPV were established with A, B,
and D SRVs open and RPV pressure 52 psig above torus pressure.
- At 1640 electrical power was restored and water level can be
determined.
Which ONE of the following actions are REQUIRED?
a. Immediately exit EOP-16 and enter EOP-01 at L-1 and P-7.
b. Continue vessel flooding until 1819 then immediately exit EOP-
16 and enter EOP-01 at R-1,
c. Concurrently execute EOP-16 and EOP-01 at L-1 and P-7.
d. Continue vessel flooding until 1819 then stop all injection.
Verify that RPV level decreases before the MCUTL is reached, i
then exit EOP-16 and enter EOP-01 at R-1.
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SENIOR REACTOR OPERATOR- P gs 48
QUESTION: 074 (1.00)
"
The following conditions exist:
A failure to scram has occurred.
'
.-
- No boron has been injected. '
-
Reactor power is 30%.
- - The Main Turbine is tripped.
-
The Main Condenser is available. ,
l -
.orus water level is normal.
Due to difficulty in establishing suppression pool cooling, the
'
-
. Heat Capacity Temperature Limit (HCTL) was exceeded.
-
Which ONE of the following states the proper method of controlling
reactor pressure? ,
j
a. Reactor pressure should be reduced using the main turbine 1
bypass valves to stay below the HCTL curve.
b. Reactor pressure should be reduced using the SRVs to stay below l
'
the HCTL curve.
,
c. Alternately depressurize using the main turbine bypass valves. i
!
d. Altemately depressurize using the SRVs.
,
,
OUESTION: 076 (1.00)
,
Which ONE of the following actions allow the operator to disregard NPSH
limits?
a. After a successful reactor scram, Core Spray is being used to l
maintain level between -125 to +45 inches.
b. After a successful reactor scram, LPCI is being used to
maintain level between + 12 to +45 inches.
c. Durin0 an ATWS, LPCI is being used to maintain level between
-155 to -140 inches after level was lowered until reactor power
dropped below 3%.
d. During an ATWS with Alternate RPV Depressurization required and
all SRVs INOPERABLE, LPCI is being used for injection.
,
)
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SENIOR REACTOR OPERATOR pig 3 49
QUESTION: 076 (1.00)
The following conditions exist:
- A steam leak occurs just upstream of the Main Turbine Stop
Valves with ooth MSIVs in the "A" main steam line failing to
close.
- A reactor scram is successful in inserting all rods fully.
- Both Main Stack Process Radiation Monitors have been reading
2.5E+4 for the last 25 minutes.
- Off-site release rate projections are 2 R/ hour Whole Body at
the site boundary.
Select the correct action and its reason.
Under these conditions the preferred method of depressurizing the RPV is
using:
a. SRVs because of the scrubbing potential of the torus water,
b. SRVs because the heat removal capability is greater than the
Main Turbine Bypass Valves,
c. Main Turbine Bypass Valves because the hotwell is the preferred
heat sink.
d. Main Turbine Bypass Valves because the heat removal capability
is greater than the SRVs.
s
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SENIOR REACTOR OPERATOR Pcgs 50
,
4
QUESTION: 077;(1.00)
- With a Reactor Building Vent Radiation Hi-Hi Alarm present, EOP-04 ,
directs the operator to verify secondary containment isolation of 1
Reactor Building Heating and Ventilation and the initiation of Standby
'
. Gas Treatment System. j
.'
\
-
'This verification will ensure that:
!
a, the Reactor Building atmosphere is contained at a positive
pressure until it can be treated and released.
I
b. a trected and controlled ground release of the activity is j
provided. j
i c. a treated and controlled elevated release of the activity is
provided. l
d. both the primary and secondary containments are maintained at a
slightly negative pressure,
i
'
l
. 1
i OUESTION: 078 (1.00)
The following conditions exist: i
1
1
- A reactor startup is in progress with RPV pressure at 500 psig.
" - It is determined the "A" Channel of Group i PCIS has one
reactor high water level switch (16A-K105A) that will NOT trip.
The MINIMUM time allowed to place 16A-K105A in the tripped condition is:
a. one hour,
b. two hours.
c. six hours,
d. twelve hours.
0-
- SENIOR REACTOR OPERATOR Pcge 51
- QUESTION: 079 (1.00) ,
~ The following conditions exist:
- ' A successful automatic reactor scram occurred on high reactor ;
pressure.
- The main condenser is ava'ilable but not currently in service.
- The operator is attempting to stabilize pressure between 900-
1060 psig using SRVs.
l
Re-establishing the main condenser as a heat sint::
a.- is not allowed. 1
i
b. is preferred but is allowed only if no valid MSIV isolations
exist.
c. is required immediately after valid MSIV isolation signals are
overridden.
d. is only allowed if the SRVs become unavailable. ;
l
l
QUESTION: 080 (1.00) l
With the plant operating at 100% power, the control room becomes
uninhabitable because of toxic gas. Evacuation is ordered and only the
immediate Actions of Pf4PS Procedure 2.4.143 were carried out.
At this point reactor water level is being maintained by:
a. Reactor Feed Pumps and CRD.
d. CRD only.
>
M
SENIOR REACTOR OPERATOR Prgs 52
OUESTION: 081 (1.00)- *
With the plant at 100% power, an MPR and EPR f ailure caused the turbine
stop valves to close and the turbine bypass valves to remain closed.
Reactor pressure peaked at 1330 psig at which time the reactor scrammed
on high flux. -
Select the statement below that correctly describes the transient.
a. No safety limit violation occurred. The Stop Valve closure
scram was the only RPS trip failure. J
b. No safety limit violation occurred. The Stop Valve closure
scram was not the only RPS trip failure,
c. A safety limit violation occurred. The Stop Valve closure
scram was the only RPS trip failure.
d. A safety limit violation occurred. The Stop Valve closure
scram was not the only RPS trip failure.
l
1
QUESTION: 082 (1.00)
Following a reactor scram, the Mode Switch should be taken to Shutdown
as soon as possible in order to:
a. disable the low steam pressure isolation.
b. enable the high reactor water level isolation.
c. insert another scram signal for 2 seconds,
d. allow MSIV closure without generating a scram signal.
_
SENIOR REACTOR OPERATOR P:ga 53 '
OUESTION: 083 (1.00)
The following conditions exist:
- The reactor was shutdown at 0230.
- Due to loss of level indications, EOP-16 was entered at 1030.
- At 1100 flooding conditions were established with 3 SRVs open.
- Flooding was stopped as soon as Flooding Cornpletion Time was
reached,
t
1
Assuming RPV levelinstruments do not respond, which ONE of the
16116 wing is the LATEST time at which injection must be reinitiated?
- a. 1214
b. 1217
c. 1254
-
d. 1257
OUESTION: 084 (1.00)
A worker in the Emergency Response Organization had 100 mrem TEDE for
the current year and 2.5 Rem TEDE lifetime prior to the declaration of
an emergency. Which ONE of the following is the MAXIMUM TEDE this
worker can receive over the course of the emergency without special
authorization?
a. 2.4 Rem
b. 2.5 Rem I
c. 4.9 Rem
d. 5.0 Rem i
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e
-SENIOR REACTOR OPERATOR Pago 54 ;
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QUESTION: 085 (1.00) j
1
~
A surveillance on the Reactor Water Cleanup High Flow Isolation is due. l
.
Which ONE of the following describes how the duration of the
surveillance is tracked and when the inoperability clock begins and
ends?
a. The surveillance is tracked in the NOS Logbook.
The clock starts when the system is removed from service and
ends when the system is returned to normal lineup.
b. The surveillance is tracked in the NOS Logbook.
The clock starts when the system is removed from service and
ends when the NWE signs off the surveillance. l
c. The surveillance is tracked using an LCO Maintenance Planning
i
Checklist. The clock starts when the system is removed from
I
i service and ends when the system is returned to normal lineup.
d. The surveillance is tracked using an LCO-Maintenance Planning
Checklist. The clock starts when the syt. tem is removed from :
-, service and ends when the surveillance is signed off by the I
l
' work group.
T
5
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SENIOR REACTOR OPERATOR P:gs 55
,
QUESTION: 086 (1.00)
With the plant at 5% power, a closed motor operated valve located in the
drywell must be tagged in the closed position.
Which ONE of the following is the proper method for determining that the
valve is in the closed position?
a. The position should be first verified by the indirect method
before power is isolated. The isolation of the power supply
may then be performed. Independent verification of the power
supply is not required.
b. The position should be first verified and independently
verified by the indirect method before power is isolated. The
isolation of the power supply may then be performed and
independently verified.
c. The first verifier should enter the drywell for verification of
valve position. The independent verifier may perform an
indirect verification of remote valve position.
d. The first verifier and the independent verifier should make
separate drywell entries for verification of valve position.
QUESTION: 087 (1.00)
Which ONE of the following may enter the Controls Area without receiving
permission from the NWE/NOS or Control Room Operator?
a. Operations Department Manager
'
b. NRC Resident inspector
ej. Station Director
d. The Outside Nuclear Plant Reactor Operator (NPRO)
1
.-. . . . .-
SENIOR REACTOR OPERATOR P:gs 55
QUESTION: 086 (1.00) .
With the plant at 5% power, a closed motor operated valve located in the
drywell must be tagged in the closed position.
Which ONE of the following is the proper method for determining that the
valve is in the closed position?
a. The position should be first verified by the indirect method
before power is isolated. The isolation of the power supply.
may then be performed independent verification of the power
supply is not required,
b. The position should be first verified and independently
verified by the indirect method before power is isolated. The
isolation of the power supply may then be performed and
independently verified.
c. The first verifier should enter the drywell for verification of
valve position. The independent verifier may perform an
indirect verification of remote valve position.
d. The first verifier and the independent verifier should make
separate drywell entries for verification of valve position. l
l
QUESTION: 087 (1.00)
Which ONE of the following may enter the Controls Area without receiving
permission from the NWE/NOS or Control Room Operator?
a. Operations Department Manager
b. NRC Resident inspector
c. Station Director
d. The Outside Nuclear Plant Reactor Operator (NPRO)
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SENIOR REACTOR OPERATOR Pigs 56
- QUESTION: 088 (1.00)
Absent a basis to assign a longer duration, Which ONE of the following
is the normal duration of a temporary modification?
a. Installation until the end of the shift
b.- 6 weeks following installation
c. 6 months following installation
d. installation until the end of the refueling outage
QUESTION: 089 (1.00)
The MINIMUM amount of parallel watchstanding REQUIRED in order to
reactivate an NRC reactor operator license is:
a. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
i
b. 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />. )
i
c. seven 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shifts.
l
d. five 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts.
l
l
QUESTION: 090 (1.00)
1
Which ONE of the following is the MINIMUM REQUIRED protective equipment l
for handling Sodium Hypochlorite outdoors? 1
Safety Goggles and:
a. Rubber Gloves
b. Rubber Gloves and Apron, Rubber Safety Boots, Forced Air
Respirator i
c. Rubber Gloves and Apron, Rubber Safety Boots
d. Rubber Gloves and Apron, Rubber Safety Boots, Respirator
i
SENIOR REACTOR OPERATOR Prgs 57
QUESTION: 091 (1.00)
Which ONE of the following conditions would allow a fail open air
operated valve to be DANGER tagged in the closed position?
'
a. The valve is gagged in the closed position with a device to
ensure it does not change state.
'
b. The DANGER tag is only for equipment protection and no
maintenance will be performed under this tagout..
c. The air supply to the valve is also DANGER tagged in the open
position.
d. A " Human Red Tag" is assigned to monitor the status of air to
the valve.
-
QUESTION: 092 (1.00)
Which ONE of the following conditions PROHIBIT the use of a " Human Red
Tag"?
a. The only qualified tagger available to be a " Human Red Tag" is
- a member of the work group.
b. Two isolation points are required to provide isolation.
'
c. The work is expected to take 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to complete.
d. The work is expected to take 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to complete with only 1/2
hour left in the current shift.
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SENIOR REACTOR OPERATOR Prgs 58. '
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QUESTION: 093 (1.00) 1
1
An offsite fire department is responding to the site during a fire in a. /
- vital area. )
' Which ONE of the following describes the security reouirements in order !
- to allow access to the protected area / vital area? l
i
a.' The fire truck and firemen must be searched prior to entering ;
~ the protected area. No additional search is required prior to
entering the vital area provided security escorts the' team.
b. No search is required of the fire truck or firemen prior to
entering the protected area provided security escorts the team, ,
however both the truck and firemen must be searched prior to >
entering the vital area. :
l
c. No search is required of the f;c - truck or firemen prior to l
entering the protected area or vital area provided security
escorts the team. l
l
d. No search is required of the fire truck or firemen prior to
entering the protected area or vital area provided security and
operations department escort the team. )
QUESTION: 094 (1.00)
You are working in a Hot Particle Control Zone (HPCZ) in a double set of
protective clothing. Which ONE of the following is the proper method of
removing the protective clothing when exiting the area?
a. Remove both sets of protective clothing at the step off pad at
the exit of the HPCZ.
b. Remove both sets of protective clothing at the step off pad at
. the exit of the buffer zone.
c. Remove the outer set of protective clothing at the step off pad
at the exit of the HPCZ and the' inner set of protective
clothing at the step off pad at the exit of the buffer zone,
d. Remove the outer set of protective clothing at the step off pad
at the exit of the buffer zone and the inner set of protective
clothing at the step off pad at the exit of the HPCZ zone.
4
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SENIOR REACTOR OPERATOR Pzg3 59 l
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QUESTION: 095 (1.00) l
An ALERT has been declared. Which ONE of the following describes the l
REQUIRED emergency notification? I
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a. The NRC must be notified within 15 minutes after the
declaration of the ALERT. State and local agencies must be 1
'
notified immediately thereafter, not to exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
b. State and local agencies must be notified within 15 minutes
after the declaration of the ALERT. The NRC must be notifind
immediately thereafter, not to exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
c. The NRC must be notified within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the declaration of
the ALERT. State and local agencies must be notified
immediately thereafter, not to exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 15 minutes,
d. State and local agencies must be notified within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after
the declaration of the ALERT. The NRC must be notified
immediately thereafter, not to exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 15 minutes
QUESTION: 096 (1.00)
Which ONE of the following describes the required manning of the Fire
Brigade?
The Fire Brigade shall consist of five members:
a. including the Brigade Leader. Two of these persons may also be
part of the crew required for safe shutdown of the plant.
b. including the Brigade Leader. These persons may not be part of
,
the crew required for safe shutdown of the plant.
c. excluding the Brigade Leader. Two of these persons may also be
part of the crew required for safe shutdown of the plant.
d. excluding the Brigade Leader. These persons may not be part of
the crew required for safe shutdown of the plant.
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SENIOR REACTOR OPERATOR P:ga 60
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OUESTION: 097 (1.00)
l
During an emergency, a reasonable action that departs from Technical l
Specifications must be taken immediately.
1
in accordance with PNPS procedures, which ONE of the following MUST ;
approve taking this action? 1
a. An on shift licensed Reactor Operator and on shift licensed
Senior Reactor Operator i
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b. A licensed Senior Reactor Operator only
c. A licansed Senior Reactor Operator and the Operations
Department Manager
d. A licensed Senior Reactor Operator and the Operations
Department Manager and the Plant Manager j
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OUESTION: 098 (1.00)
You have worked the foDowing schedule:
- Thursday 1st scheduled day off l
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Friday 2nd 7 am to 7 pm I
- Saturday 3rd 7 am to 7 pm )
-
Sunday 4th 7 am to 3 pm
-
Monday 5th 7 am tn 3 pm
- Tuesday 6th 7 am to 9 pm
-
Wednesday 7th 7 am to 3 pm
-
Thursday 8th 7 am to ?
Which ONE of the following represents the LATEST you can be required to
work on Thursday the 8th, without special approval being granted?
(Assume turnover time is NOT included)
a.3pm
b. 5pm
c. 7 pm
d. 9 pm
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SENIOR REACTOR OPERATOR Pzga 61
QUESTION: 099 (1.00)
The only individual available for a call-in for TSC staffing informed
the Nuclear Watch Engineer (NWE) on the phone that he has consumed
alcohol within the previous 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
Which ONE of the following describes the individuals ability to work in
the TSC? :
a. not permitted to work in the TSC.
b. permitted to work in the TSC provided the individual informs
the NWE that alcohol has not impaired his ability to work in
the TSC. A blood alcohol concentration test is at the NWE
discretion, based on the NWE phone discussions with the ;
individual.
c. permitted to work in the TSC only if a blood alcohol
concentration test is performed upon arrival on site and
the concentration is less than 0.04.
1
d. permitted to work in the TSC only if a breathalyzer l
test is performed upon arrival on site and the blood to
alcohol ratio is greater than or equal to 4.0.
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QUESTION: 100 (1.00)
A procedure is currently being performed which requires the installation l
of a jumper. It is discovered that the procedure directs the jumper I
placement in a position that would cause an unexpected ESF actuation. A
change to the jumoer position is required.
Which ONE of the following is the required method to revise this
procedure to change the jumper position? l
a. Editorial correction
b. SRO Change
c. Minor Revision
d. Major Revision ]
( * * * * * * * * * * END OF EXAMIN ATION * * * * * * * * * *)
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' BOSTON EDIS0N RTYPE H6.02
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PILGRIM NUCLEAR POWER ST. TION
Procedure No. 2.1.27
DRYWELL TEMPERATURE INDICATION
i
REQUIRED REVIEWS REVIEWERS AND APPROVERS
hE. A$ oves /NNW 6b7M
"" Writ *" '**
Thi"k '*d""*'
Act
8)
'wcynical Reviewer
'gg)q,V
Date
Review *# 9 # ^ * u' 9/#6/
Validator '
Date' W
SAFETY REVIEW E0"!9ED/
/i f
Procedure A' ner
'
e/d
/Dgte
NOT REQUIRED
N/+
QAD Man'ager
ORC REVIEW REQUIRED / Date
MT REQ'J:"C0
AA L lo /k 194
-
ORCC{plirman '
Date
0m Ibb smlk fobt9/94
sporpible anager / 1Dhte
Effective Date: /0/d8 9%
020095 2.I.27 Rev. 3
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REVISION LOG
REVISION 3 Date originated 5/94
Paaes Affected Descriotion
4 Add PDC 92-58 to References.
~5,7,10,12,14,15 Revise Kaye nomenclature and delete channel points from old
Kaye recorder in accordance with PDC 92-58.
Editorial 2C Date Originated 3/93
Paaes Affected Descriotion-
7,9,11,13 Delete references to Station Honeywell Computer System as it
is obsolete.
Editorial 2B Date Originated
1
Paoes Affected pescription i
1
4,5,7 Editorial corrections to reflect new E0P numbers and entry l
conditions and to add new Editorial Correction rev bar l
identifications. I
Editorial 2A Date Originated
Paaes Affected Qgscriotion
4,5,7 Incorporated editorial corrections to Main Control Room
Panel Labels per PDC 87-78C.
REVISION 2 Date Originated
Paaes Affected Descriotion
All Reformat to comply with PNPS 1.3.4-1.2.
.
2.1.27 Rev. 3
Page 2 of 15
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, .IABLE OF CONTENTS
.
EA9A
1.0 PURPOSE AND SC0PE................................................. 4
2.0 REFERENCES........................................................ 4
3.01 . DEFINITIONS....................................................... 4
4.0 DISCUSSION........................................................ 4
5.0 PRECAUTIONS AND LIMITATIONS....................................... 6
6.0 PREREQUISITES..................................................... -6'
7.0 PROCEDURE......................................................... 7
8.0 ATTAC HME N T S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
' ATTACHMENT 1 - TE-5050 TEMPERATURE ELEMENTS - BULK DRYWELL
TEMPERATURE ESTIMATE..................................... 9
ATTACHMENT 2 - TE-8125 TEMPERATURE ELEMENTS - BULK DRYWELL
TEMPERATURE DETERMINATION............................... 10
ATTACHMENT 3 - TE-5050 TEMPERATURE ELEMENTS - INSTRUMENT RUN
T EM PE RATUR E E ST I MAT E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
ATTACHMENT 4 - TE-8125 TEMPERATURE ELEMENTS . INSTRUMENT RUN
TEMPERATURE DETERMINATION............................... 12
ATTACHMENT 5 - TE-5050 TEMPERATURE ELEMENTS - LOCATION INFORMATION..... 13
ATTACHMENT 6 .TE-8125 TEMPERATURE ELEMENTS - LOCATION INFORMATION..... 14
.
2.1.27 Rev. 3
Page 3 of 15
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1.0 - PURPOSE AND SCOPE -
'
This Procedure provides instructions for determining Drywell bulk temperature when
, ,
the Emergency Operating Procedures _(EOPs) require measurement of this parameter.
-
2.0 REFERENCES
,
2.1 DEVELOPMENTAL
[1] PNPS Technical Specifications Table 3.2.H
4
[2] PNPS Technical Specifications Tables 3.2.H and 4.2.H
-
[3] PDC 87-78C, Improvements to Labels, Nameplates on Main Control Room Panels
i
[4] PDC 92-58, Kaye Recorder Replacement !
2.2 IMPLEMENTING
[l] PNPS 2.2.49, " Primary Containment Cooling System"
[2] PNPS 8.7.1.4.2, ' Primary Containment Integrated Leak Rate Test"
3.0 DEFINITIONS
None
4.0 DISCUSSION
[1] The following sections of the Emergency Operating Procedures require
measurement of Drywell temperature:
, (a) E0P-1, RPV Control: RPV Water Level Instrument Run temperatures associated
with the RPV Saturation Temperature Figure of Caution 1.
(b) E0P-2, Failure to Scram: RPV Water Level Instrument Run temperatures
- associated with the RPV Saturation Temperature Figure of Caution 1.
(c). E0P-3, Primary Containment Control:
.
(1)' Entry condition _(150"F)
,
(2) Drywell temperature path
(3) RPV Water Level Instrument Run temperatures associated with the RPV
,
Saturation Temperature Figure of Caution 1.
(4)- Figure 5: (SPDS 031) DSIL (Orywell Spray Initiation Limit)
.
2.1.27 Rev. 3
Page 4 of 15
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4.0 913GUSSION (Continued)
(d) E0P-4, Secondary Containment Control: RPV Water Level Instrument Run ,
temperatures associated with the RPV Saturation Temperature Figure of
_
.
Caution 1.
E0P-26, RPV Flooding, Failure To Scram:
(1) Temperatures near the RPV Water Level Instrument Reference Leg
vertical runs.
[2] Drywell temperature is normally monitored in the Control Room by using
TRU-9044, DRYWELL TEMP / PRESS Recorder, and TI-9019, DW TEMP Indicator, on
Panel C903. TRU-9044 receives its input from a single temperature element
located at a relatively low elevation in the Drywell. TI-9019 receives its
input from a single temperature element located just below the neck of the
Drywell. Both of these temperature elements measure ambient Drywell air space
temperature.
[3] The TE-5050A through P temperature elements are used to evaluate Drywell
tem >erature with respect to Technical Specifications limits (refer to
Technical Specifications Table 3.2.H). The Drywell locations of these
elements are listed in Attachment 5. These elements are used to monitor ,
Drywell temperature for Technical Specifications requirements because of their I
reliability, location, and their redundancy (dual-element RTDs). In addition, I
these temperature elements are the primary elements used for the Primary
Containment Integrated Leak Rate Test.
[4] Local Drywell air temperature indication is supplied by the TE-8125 series I
temperature elements. The TE-8125 series temperature indication consists of
20 RTDs located throughout the Drywell which provide input to the Kaye Temp. 1
Computer (refer to Attachment 6).
[5] When TRU-9044 and TI-9019 are not available, selected Drywell temperature
elements are used to estimate an average temperature near the RPV water level
instrument runs and an average bulk Drywell temperature. Temperatures near
the RPV water level instrument runs are monitored by those thermal elements
which are located in the upper elevations of the Drywell since mest of the
instrument runs are found in this region of the Drywell. Bulk average Drywell '
temperature is a weighted average temperature based on the volume of the
Drywell. By averaging more readings from the lower region of the Drywell
(which contains most of the Drywell air space) than from the upper region of
the Drywell, a representative average Drywell temperature is obtained. More
sophisticated methods to calculate a_ weighted average Drywell temperature are
available, as part of the ILRT Procedure, PNPS 8.7.1.4.2. The method outlined
,
in this Procedure, however, attempts to balance the complexity and time
.
consuming aspects of the sophisticated approach against the requirement to
rapidly obtain a value for Drywell temperature suitable for use in the E0Ps.
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2.1.27 Rev. 3
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Page 5_ of 15
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5.0 PRECAUTIONS AfC LIMITATIONS
[1] The Drywell temperature shall be maintained within the following limits when
the reactor coolant temperature is above 212 F.
(a) Above elevation 40': $ 194*F ;
(b) Equal to or below elevation 40': s 150*F
Upon determination that the Drywell temperature at any elevation has exceeded
the above limits, the Drywell temperature at each elevation shall be logged
every 30 minutes. The Drywell temperature shall be reduced to within the
above limits within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; otherwise corrective action shall be as
specified in Technical Specifications Sections 3.2.H.2 and 3.2.H.3.
i (Tech Spec 3.2 H.1)
[2] If the Drywell temperature has exceeded either limit of Technical
Specifications Section 3.2.H.1 for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, an engineering
,
evaluation shall immediately be initiated to assess potential damage and
render a determination of ability of safety related equipment to perform its
intended function.
If either limit of Technical Specifications Section 3.2.H.1 has been exceeded
for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, ferther action to justify continued operation shall
be determined by an engineering evaluation which must be completed within one
week. (Tech Spec 3.2.H.2)
[3] If the requirements of Technical Specifications Section 3.2.H.2 have not been I
met, an orderly shutdown shall be initiated and the reactor shall be in a cold
shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. (Tech Spec 3.2.H.4)
[4] If the Drywell temperature at any elevation exceeds 215*F and the temperature
cannot be reduced to below 215 F within 30 minutes, a reactor shutdown shall
be initiated and the reactor shall be in cold shutdown condition within
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. (Tech Spec 3.2.H.4)
i
[5] When reactor coolant temperature is above 212*F, the Drywell air temperature '
limits will be determined by reading the instruments listed in Techaical
Specifications Table 3.2.H. These instruments shall be logged once per shift,
and each reading compared to the limits of Technical Specifications Section 3.2.H.1. (Tech Spec 4.2.H.1)
J
6.0 PREREQUISIT_E.ji
.
None
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2.1.27 Rev. 3
Page 6 of 15
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-7.0 50CEDURE
. , .
[1] DETERMINE bulk Drywell temperature using one of the following methods (listed
in order of preference):
(a) SELECT the higher of the valves indicated on TI-9019, DW TEMP Indicator,
and TRU-9044, DRYWELL TEMP /F 'SS Recorder (Panel C903).
CAUTION ,
i
j The instruments listed below are not environmentally qualified for use in a harsh '
environment. Under accident conditions, they should only be used if either j
TI-9019 or TRU-9044 is not available for use.
1
(b) Highest probable Drywell temperature from EPIC points DRY 002 or DRY 004.
(c) For a more representative bulk teaperature, AVERAGE the TE-5050 series RTDs
using the computer points in accordance with Attachment 1.
(d) For a more representative bulk temperature, AVERAGE the TE-8125 series RTDs
using the Kaye Temp. Computer in accordance with Attachment .2.
1
(e) All of the TE-5050A through P series RTDs can be read locally at Panel C85, '
Reactor Building El. 23' East, for Attachment I data.
[2] DETERMINE RPV water level instrument run temperature using one of the !
'
following methods (listed in order of preference):
(a) SELECT the higher of the values indicated on TI-9019 and TRV-9044 !
(Panel C903). l
(b) AVERAGE the TE-5050 series RTDs in ecordance with Attachment 3.
(c) AVERAL .he TE-8125 series RTDs using the Kaye Temp. Computer in accordance
with " '.achment 4.
1
(d) The TE-5050A through P RTDs can be read locally at Panel C85, Reactor l
Building El. 23' East, for Attachment 3 data.
[3] Additional information on Drywell temperature elements and location is
contained in PNPS 2.2.49, " Primary Containment Cooling System". 1
1
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2.1.27 Rev. 3
Page 7 of 15
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8.0 ATTACMENTS
ATTACHMENT 1 - TE-5050 TEMPERATURE ELEMENTS - BULK DRYWELL TEMPERATURE ESTIMATE
ATTACHMENT 2 - TE-8125 TEMPERATURE ELEMENTS - BULK DRYWELL TEMPERATURE. DETERMINATION
ATTACHMENT 3 - TE-5050 TEMPERATURE ELEMENTS - INSTRUMENT RUN TEMPERATURE ESTIMATE
ATTACHMENT 4 - TE-8125 TEMPERATURE ELEMENTS - INSTRUMENT RUN TEMPERATURE
DETERMINATION
ATTACHMENT 5 - TE-5050 TEMPERATURE ELEMENTS - LOCATION INFORMATION
ATTACHMENT 6 - TE-8125 TEMPERATURE ELEMENTS - LOCATION INFORMATION
4
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2.1.27 Rev. 3
Page 8 of 15 j
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ATTACHMENT 1
Sheet 1 of 1
TE-5050 TEMPERATURE ELEMENTS
BULK DRYWELL TEMPERATURE ESTIMATE
[1] SELECT one temperature element in each group of temperature elements Alg!
RECORD its temperature.
[2] COMPUTE the average temperature as follows:
(a) Average - (A + B + C + D + E + F)/6
T'.ME
TE-5050 EPIC
GROUP COMPUTER POINT
ELEMENT # TEMPERATURE ( F)
A
DRY 002
A -----OR---- -------------- ------- ------- ------- ------- ------- -------
B DRY 004
i
.
E
DRY 010
B -----OR---- ---- - --- ------- ------- ------- ------- -- ---- -------
G DRY 014
C -----OR---- -------------- ------- ------- ------- ------- ------- -------
H DRY 116
L l
DRY 122 '
D -----0R---- -------------- ------- ------- ------- ------- ------- -------
M DRY 124
K
DRY 120
E -----0R---- -------------- ------- ------- ------- ------- ------- -------
J DRY 118
N
DRY 126
F -----OR---- -------------- ------- ------- ------- ------- ------- -------
p DRY 130
-
AVERAGE
Performed By Date Reviewed By Date
2.1.27 Rev. 3
Page 9 of 15
ATTACHMENT 2
Sheet 1 of 1
TE-8125 TEMPERATURE ELEMENTS
BULK DRYWELL TEMPERATURE DETERMINATION
[1] SELECT one temperature element in each group of temperature elements 88Q
,
RECORD the temperature indicated on the Kaye Temp. Computer.
[2] COMPUTE the average temperature a: follows:
(a) Average - (A + B + C + 0 + E + F)/6
TlME .
TE-8125
GROUP
ELEMENT # TEMPERATURE (*F)
3
A ----0R----- -------- --------- --------- --------- --------- -------- l
l
4 l
l
l
9 i
8 .... 0R.... ........ ......... ......... ......... ......... ........ 1
10
..
11
C -----OR---- -------- --------- --------- --------- --------- --------
12
13
D -----OR---- -------- --------- --------- --------- --------- --------
14
15
E -----OR---- ----- -- --------- --------- --------- --------- --------
16
17
^
F -----OR---- -------- ------- *- --------- --------- --------- --------
18
AVE: GE
Performed By' Date Reviewed By Date
2.1.27 Rev. 3
Page 10 of 15
ATTACHMENT 3
Sheet 1 of 1
TE-5050 TEMPERATURE ELEMENTS
INSTRUMENT RUN TEMPERATURE ESTIMA1E
[1]. DETERMINE.the rack (s) of concern 8tlQ RECORD the indicated temperature for each
element in that group.
[2] COMPUTE the average temperature.
Instrument Runs for Rack 2205
A Channel Instruments
TLME
!TE-5050 EPIC
ELEMENT COMPUTER POINT
- TEMPERATURE ( F)
.
A
........... ......
DRY 002
. .....
Zw . . . ....... ....... ....... ....... .......
j
........... ........ . ..... . .. .. ....... ....... ....... ....... .......
AVERAGE
Instrument Runs for Rack 2206
B Channel Instruments
TLME
TE-5050 EPIC
ELEMENT COMPUTER POINT
- TEMPERATURE (*F)
..... ..... ...... . ..... . $b. ....... ....... ....... ....... .......
D DRY 008 2.Ir
........... .................. ....... ....... ....... ....... .......
4.....
E DRY 010 LIO .
AVERAGE
Performed By Date Reviewed By Date
2.1.27 Rev. 3
Page 11 of 15
, -
ATTACHMENT 4
Sheet 1 of 1
1
TE-8125 TEMPERATURE ELEMENTS
INSTRUMENT RUN TEMPERATURE DETERMINATION
[1] DETERMINE the rack (s) of concern A151 RECORD the indicated temperature for each l
element in that group. !
[2] COMPUTE the average temperature.
Instrument Run for Rack ?205
A Channel Instrument
T;:ME
TE-8125
ELEMENT # TEMPERATURE (*F)
4
...........
[b
........ ......... ......... ......... ......... ........
-
10 -,g
AVERAGE
-
Instrument Runs for Rack 2206
B Channel Instrument
T::ME
TE-8125
ELEMENT # TEMPERATURE (*F)
4
...........
2.[h
........ ......... ......... ......... ......... ........
9
Q{[] .
AVERAGE
Performed By Date Reviewed By Date
2.1.27 Rev. 3
Page 12 of 15
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ATTACHMENT 5
Sheet 1 of 1
.
LTE-5050 TEMPERATURE ELEMENTS LOCATION INFORMATION
,
.
Temperature EPIC Elevation Azimuth Area
Element Point ID (feet) (degrees) Monitored
TE-5050A DRY 002 86 0 2' out from vessel below
an exh. register
TE-50508- DRY 004 89 180 l' out from vessel.
- TE-5050C DRY 006 86 50 4' out from vessel above
supply register
- TE-50E00 DRY 008 90 330 2' below head exh. hole.
TE-5050E DRY 010 60 270 2' out from bio-shield.
'
4 out from bio-shield.
~
-TE~-5050F DRY 012 60 90
TE-5050G DRY 014 40 270 10' out from bio-shield under
TE-5050H DRY 116 40 90 10' out from bio-shield under
TE-5050J DRY 118 35 0 l' from CRD area inside wall.
TE-5050K DRY 120 35 180 l' from CRD area inside wall. l
l
TE-5050L DRY 122 22 205 13' out from CRD area outside !
wall.
TE-5050M DRY 124 22 45 13' out from CRD area outside ,
wall. l
TE-5050N DRY 126 15 270 8' out trom CRD rea nutside
wall.
TE-50500 DRY 128 15 0 On CRD area outside wall.
N
TE-5050P DRY 130 12 125 10' out from CRD area outside
wall.
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Page 13 of 15 )
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ATTACHMENT 6
Sheet 1 of 2
4
TE-8125 TEMPERATURE ELEMENTS LOCATION INFORMATION
'
Temperature Elevation Azimuth Area
Element (feet) (dearees) Monitored
TE-8125-1 90 285 Head Exhaust
TE-8125-2 90 210 Head Exhaust
,
TE-8125-3 85 180 3' out from vessel
,
TE-8125-4 85 0 4' out from vessel
above ductwork
TE-8125-5 82 300 In exhaust duct
TE-8125-6 82 45 In exhaust duct
TE-8125-7 80 270 In annulus
- TE-8125-8 80 90 In annulus
TE-8125-9 54 270 6' out from !
bio-shield l
'
l
TE-8125-10 54 90 6' out from '
.
bio-shield
TE-8125-ll 40 270 10' out from
, bio-shield under Main
,
Steam Line
TE-8125-12 40 90 10' out from
bio-shield under Main
Steam Line ,
1
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TE-8125-13 25 315 On CRD area outside
wall
!
TE-8125-14 '25 135 On CR0 area outside
wall
TE-8125-15 19 225- 14' out from CRD
area outside wall
TE-8125-16 19 45 14' out from CRD
m es outside wall.
2.1.27 Rev. 3
Page 14 of 15
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ATTACHMENT 6
Sheet 2 of 2
TE-8125 TEMPERATURE ELEMENTS
LOCATION INFORMATION
Temperature- Elevation Azimuth Area
Element (feet) idearees) Monitored
TE 8125-17~ 14 265 On CR0 area outside wall
TE-8125-18. 14 110 On CRD area outside wall
TE-8125-19 29 180 l' from CRD area
4
inside wall
TE-8125-105 29 0 l' from CR0 area
inside wall
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2.1.27 Rev. 3
Page 15 of 15
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SENIOR REACTOR OPERATOR Pegs 1
ANSWER KEY
MULTIPLE CHOICE O23 c
001 b 024 b
002 b 025 d
003 d 026 c
004 c 027 d
005 b 028 c
l
006 a 029 6 at A
007 b 030 c !
1
008 c. 031 a }
s 4
000 a M -
010 a 033 a
011 a 034 b
012 a 035 b
013 c 036 d I
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014 b 337 c
015 d 038 d
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016 a 039 d
017 b 040 b
018 c 041 d
019 b 042 c
020 c 043 d
021 d 044 b- ,
- 022 d 045 d
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SENIOR REACTOR OPERATOR Pags 2
ANSWER KEY
.046 c 069 c
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047 d 070)('C'
048 ' d 071 c-
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049 c 072 c
050 a 073' a
051 c 074 d
052 d 075 d
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053 b 076 b
054 b 077 c
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055 a 078 a l
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056 a 079 b l
057 d 080 a
g 058 a 081 d
059 d 082 a I
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060 c 083 d
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061 a 084 d
062 a 085 a
063 c 086 b
064 d- 087 a
065 d 088 c
066 b 089 b
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067 a 090 e
'068 d 091 a
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- P:gs 3 >
ANSWER KEY
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002 : b
- 093 'c'
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094 c
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. 095 b
096 b l
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097 b-
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098 - b
- 099 c
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-_ (* " * * * * * * * END OF FXAMINATION * " "' ' * )
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ATTACHMENT 2
Facility Comments on Written Examination
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Boeton Edinon
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Pilgrim Nudear Power Station
Rocky Hdi Road
- Plymouth, Massachusetts 02360
L J. Olivier
, Vice President Nuclear Operations
- . and Station Director
May 16,1997
BECo Ltr. 2.97-054
4
U.S. Nuclear Regulatory Commission
Region I '
- . 475 Allendale Ruad
King of Prussia, PA 19406
Docket No. 50-293 ,
License No. DPR-35
Pilarim's 1997 NRC Written Examination Comments
a
1 The written examination administered on May 5,1997, was considered to be an in-depth examination,
which fairly tested the six (6) SRO candidate's knowledge in the appropriate areas. After thorough
, . analysis of the content of the examination, it is clear that the use of misleading information, use of the
double negative context, and the asking of subjects not important to public health and safety were
avoided.
However, specific requests on several written exam questions are submitted for your consideration in
Enclosure 1. Enclosure 2 contains the reference documentation associated with each of the requests.
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- Your consideration of these requests is greatly appreciated.
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L. J. Olivier
- PMKINRCEXCO.
Enclosure '
- cc:. See ne'xt page
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cc- Mr. Don Florek
Region 1.
475 Allendale Road .
King of Prussia, PA .19406
Mr. Alan Wang, Project Manager
Project Directorate 1-3
- Division of Reactor Projects - 1/11
Mail Stop: 14B2
U. S. Nuclear Regulatory Commission
1 White Flint North
11555 Rockville Pike
Rockville, MD 20852.
'
' U.S. Nuclear Regulatory Commission
Attention: Document Control Desk
Washington, DC 20555
' Senior Resident inspector .,
Pilgrim Nuclear Power Station
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ENCLOSURE 1
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ENCLOSURE 1
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1. Question # 32
- While operating at 100% power, it is determined that the Main Steam Lin'e High Flow
switches on the "B" Main Steam Line will t10T trip under a high flow condition.
Which ONE of the following is the MINIMUM REQUIRED action?
a. Direct I&C personnel to manually trip the inoperable switches. I
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b. Direct l&C personnel to manually insert a half Group i isolation on the "B" Group I -
Channel.
c. Initiate an orderly shutdown and be in Cold Shutdown Condition within a
i MAXIMUM of 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> after the instrument failure.
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d. Initiate an orderly shutdown and have Main Steam Lines isolated within a
MAXIMUM of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after the instrument failure.
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ANSWER: d.
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DISCUSSION: l
The stem of this question states,"... it is determined that the Main Steam Line High Flow
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switches on the "B" Main Steam Line will NOT trip under a high flow condition..."
There are two trip systems associated with Group I PCIS, designated "A" and "B". Trip System
"A" has inputs from MSL High Flow switches comprising two instrument channels, and Trip
System "B" has eight inputs from MSL High Flow switches comprising two instrument channels,
each steam line is equipped with four switches each, one for each instrument channel
(Enclosure 2, Attachment 1, page 1).
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- . The stem of the question states that all flow switches on the "B" Main Steam Une are
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inoperable. Since this is the case, there are less than two operable instrument channels for
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both PCIS logic trip systems. (See Enclosure 2, Attachment 1, Page 2)
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- Since there are less than the minimum operable instrument channels for both trip systems,.
1 Attachment 1, page 3 states, "If the minimum number of operable instrument channels cannot
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be met for both trip systems, place at least one trip system (with the most inoperable channels)
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in the tripped condition within one hour or initiate thc appropriate action required by Table
3.2.A listed below for the affected trip function."
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Table 3.2.A requires action "B", which states, " Initiate an orderly load reduction and have Main
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. Steam Lines isolated within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />".
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Since there is no grace period (of one hour) for the two trip system inoperability (vice the one
trip system inoperability), there is no obvious correct response.
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REQUEST (Question # 321:
Since the correct response is not offered as a choice in the responses, we request that this ,
question be deleted from the examination.
REFERENCE:
PNPS Technical Specifications, Table 3.2.A and associated notes (Enclosure 2,
Attachment 1).
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! 2. Question # 76 l
The following conditions exist:
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- A steam leak occurs just upstream of the Main Turbine Stop Valves with both )'
j MSIV's in the "A" main steam line failing to close.
- A reactor scram is successful in inserting all rods fully.
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- . Both Main Stack Process Radiation Monitors have been reading 2.5+E4 for the last
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25 minutes.
- Off-site release rate projections are 2 R/ hour Whole Body at the site boundary.
, Select the correct action and its raason.
Under these conditions, the preferred method of depressurizing the RPV is using:
a. SRVs because of the scrubbing potential of the torus water. 'l
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- b. SRVs because the heat removal capability is greater than the Main Turbine Bypass i
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c. Main Turbine Bypass Valves because the hotwell is the preferred heat sink. .
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d. Main Turbine Bypass Valves because the heat removal capability is greater than
the SRVs.
j ANSWER: b.
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DISCUSSION:
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Because both answer "a" and "b" select the SRVs as the correct mechanism of depressurizing, !
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the question then becomes discriminatory as to the basis for doing so. Appendix B of the
Emergency Procedure Guidelines states that Contingency #2, Emergency RPV
Depressurization may be required to:
Minimize radioactivity release from the RPV to the primary containment and secondary
containment, or to areas extemal to the primary containment and secondary '
- containment.
Additionally, Appendix G states that the purpose of the Radioactivity Release guideline is to
limit radioactivity release into areas cutside the primary and secondary containments.
Since distracter "a" implies that SRV's are used because they discharge to the primary
containment, "a" can be construed as the correct answer. That is, given the situation provided,
the fact that the SRVs tiischarge to the containment via the torus is more significant than the
fact that the SRV's heat removal capability is greatedhan the bypass valves.
Since Appendix B also provides generic guidance that SRV's are used because of their heat
. removal capability, "b" is also correct. .
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REQUEST: (Question # 76)
Because both answers "a" and "b" are correct per the EPGs, we request that answers "a" and
"a" both be accepted as correct, and the question be retained in the examination.
REFERENCEi
1. Emergency Procedure Guidelines Appendix B, Section 11, Contingency #2 (OEl l
Document 8390-4B, [ Enclosure 2, Attachment 2]). l
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l 3. - Qitestion #'28 )
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, While operating at 100% power, a control rod is determined to be uncoupled. Attempts
- to couple the rod have been unsuccessful.
. Which ONE of the following states the MINIMUM REQUIRED actions.
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a. Verify the control rod can be moved with drive pressure and maintain the control rod ;
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at the target position.
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b. 'Fu!!y insert the control rod and hydraulically disarm the CRD.
c. Fully insert the control rod and electrically disarm the directional control valves.- j
d. Fully hsert the control rod, electrically disarm the directional control valves and then
declare the rod inoperable.
ANSWER: c. J
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DISCUSSION:
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The only differentiation between distracter "d" and the correct answer "c" is whether the
rod is declared inoperable. ]
If a control rod was uncoupled, it would be declared inoperable by Technical
Specifications when the inoperability was discovered. The control rod would then be
inserted and electrically disarmed to ensure control rod movement was precluded.
Taking this action does not eliminate the fact that the control rod was inoperable but does
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allow relief from the requirements of the associated Technical Specification actions for an l
uncoupled control rod. The control rod that was uncoupled would still be administratively I
controlled as an inoperable control rod, even though the action statement of Technical
Specification 3.3.F does not have to be applied. At PNPS, if an action has to be taken
on the part of Technical Specifications, the equipment inoperability is traced through the
application of an " Active LCO"in the LCO log.
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from a Tech Spec consideration only, the rod is not inoperable. However from an
administrative and practical standpoint, the rod is indeed inoperable, and the Active LCO
- is maintained to control the status of the rod. Therefore, if the cand:date approached the
' question from this perspective, distracter "d" can also be considered as an acceptable
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answer,
- , While Procedure 2.2.87, 5.2.1[3] does state that a rod fully inserted and electrically
L disarmed is not inoperable, it references Tech Spec 3.3.A.2 that concems rods that
- cannot be moved with drive pressure. This statement does not apply to the conditions
' identified in the question.
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REQUEST: (Question # 28)
We request that distracter "d" also be accepted as correct.
REFERENCE:
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1. PNPS 1.3.34.2 (See Enclosure 2, Attachment # 3)
. - 3.0[1]" Active LCO" Definition
. ' 4.0 " Discussion"
2. Operations Department Manager (Tom Trepanier, (508) 830-8364)
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4. Question # 50
in the event that torus water level cannot be maintained above 95 inches, HPCI is
secured in order to prevent:
a. exceeding the Primary Containment Pressure Limit.
b. exceeding the Pressure Suppression Pressure,
c. exceeding the Heat Capacity Temperature Limit.
d. isolating HPCI on high exhaust pressure.
ANSWER: a.
DISCUSSION:
The operators at PNPS are provided a " supplemental" approved handout for the study of
the EOP procedures (Enclosure 2, Attachment 4). In this handout, the basis for the 95
inches torus level securing of HPCIis not stressed as the PCPL. The fact the exhaust
will become uncovered is stressed, and HPCI will then directly pressurize the
containment. The wording for the PCPL statement is "may exceed the PCPL", and not
"the basis for the uncovery is the PCPL". Wnen this question is considered, the fact that
the primary containment would pressurize is a valid line of thought. From this direction,
scrutinization of the choices through the use of the supplied EOPs would lead a
candidaa to choose the most limiting curve between the PCPL and the PSP. This would
of course be the PSP curve. Based on this line of reasoning, response "b"is considered
also to be a valid response.
REQUEST: (Question # 50)
We request that distracter"b" also be considered as correct.
REFERENCE:
EOP-03 Supplemental Training Materials / Flow Charts (Enclosure 2, Attachment 4)
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5. Question # 29
With the plant at power, it is determined that the MO-1001-37 (B loop Torus Spray) and
MO-1400-25A (A Loop Core Spray Inboard Injection) valves have failed their operability
test. Both valves are currently closed.
The maximum time allowed before the plant must be in COLD SHUTDOWN is:
a. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ( 1 day)
b. 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> (4 days)
c. 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> (7 days)
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d. 192 hours0.00222 days <br />0.0533 hours <br />3.174603e-4 weeks <br />7.3056e-5 months <br /> (8 days)
ANSWER: b.
DISCUSSION:
PNPS 2.2.125, " Containment isolation System" lists the valves that are considered to be
primary containment isolation valves. An identical listing is contained within the FSAR.
Included in this listing are both the MO-1400-25A and the MO-1001-378 (see
Enclosure 2, Attachment 5). As containment isolation valves, the administrative
requirements require at least one valve in the line to be deactivated in the isolated
position, unless the valve receives any signals other than the isolation signal. Whether
the valve receives any other signals (other than the isolation signal) determines whether
the valve has to be deactivated electrically or otherwise administratively controlled. If the
requirements of this procedure are not met (and the questinn does not provide this
information), an orderly shutdown shall be initiated and the reactor shall be in Cold
Shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. ,
This question was designed to test the applicants' ability to determine:
1) the impact of an Inoperable 378 valve on "LPCl* operability;
2) the impact of an inoperable 378 valve on the Containment Cooling Loop's Operability
and,
3) the overall effect of 1 and 2 when coupled with an inoperable Core Spray system.
At least one applicant, (during a followup interview), interpreted this question as a test of !
his ability to recognize that:
1) Both valves are PCIS valves
2) That at a minimum, the 25A would need to be deactivated since it receives an Auto
' Open signal and,
3) Determine the corrective actions for failed PCIS valves.
Since the questions asks for the maximum time allowed before the phnt must be in
COLD SHUTDOWN, if a candidate were to assume that the question is testing his
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knowledgs of PCIS, thin it is rs: sort:bla that ths candidits would choso "c" es tha
correct response, given that no other actions are taken.
REQUEST: (Question # 29)
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Due to the two different ways that this question can be Interpreted, we request that both "a"
and "b" be accepted as correct.
REFERENCE:
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PNPS Procedure 2.2.125 (Enclosure 2, Attachment 5)
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6. Question # 27
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'When' valving in a CRD hydraulic control accumulator, the 305-102 (Withdraw Riser . 1
Isolatio'n Valve) and the 305-112 (Scram Discharge Riser Isolation Valve) are required to
be open prior to opening the 305-101 (insert Riser Isolation Valve). This prevents:
af a single rod scram when opening the 305-101 valve.
b. excessive scram time of that rod in the event of a reactor scram,
c. damage to the accumu5 tor in the event of a reactor scram.
d. damage to the drive mechanism in the event of a reactor scram.
ANSWER: d. .
DISCUSSION:
While it is stated in PNPS 2.2.87 that valve misoperation dunng the isolation or
restoration of a HCU can cause " severe damage to the mechanism", the isolation of the
102 (by itself) can also delay control rod insertion following a scram signal. As seen in
Enclosure 2, Attachment 6, with the 102 valve shut, the exhaust path from the
- mechanism is isolated. Since the question does not state the position of the associated
rod for the HCU being restored, the candidate could reasonably assume that the rod is in
a position other than fully inserted. If the exhaust path is isolated, any scram signal will
not permit the mechanism to scram at " normal" rates, if the control rod inserts at all.
REQUEST: (Question # 27)
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Due to the fact that response "b" contain the phrase " excessive scram time of the rod in
the event of a reactor scram", we request that response "b" be also accepted as a correct
answer.
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REFERENCE:
Figure 4 from PNPS Training Material (Enclosure 2, Attachment 6)
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ATTACHMENT 3
NRC Resolution of Facility Comments
Ques 28 Disagree with BECO comment. The question stem requested " MINIMUM l
REQUIRED actions" and the applicants had Technical specifications. The
question clearly related to interpretation of technical specification-required
actions. As specified in Technical Specification 3.3.A.d, control rod drives
that are fully inserted and electrically disarmeo shall not be considered
inoperable. Therefore answer d is incorrect. There was no change to the
answer key.
Ques 29 Agree with BECO comment. There was insufficient information provided in
the question to rule out consideration of containment isolation system
technical specifications. The answer key was revised to accept a or b as
correct answers.
Ques 32 Agree with BECO comment that there is no correct answer to the question
as written. There was no comment provided to this question during the
preexam review. The question was deleted from the examination. l
Ques 50 Disagree with BECO. The question asks for the reason HPClis secured at a
decreasing torus level of 95 inches. Enabling objective 10 required the
applicant to " state the significance of torus levelless than 95 inches as
regards the HPCI system." The significance, stated in O-RO-03-04-05, Rev
,
1, IG 3 is to prevent exceeding the primary containment pressure limit ?
d
(PCPL). The BECO response to the question is attempting to reword the
-question to determine the first EOP-03 curve limit reached if HPCI exhaust is
not secured at a decreasing torus level of 95 inches. This was not the
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question asked. There is only one correct answer to the question asked.
While the pressure suppression pressure (PSP) will be exceeded, it has a l
- relatively small consequence. The PCPL will be exceeded, which has a large j
consequence, primary containment failure, and is the stated reason in the l
reference material for securing HPCI at a torus level of 95 '"ches. There was
no change to the answer key.
Ques 7 Disagree with BECO comment. As described in 0-RO-03 04-07, Rev 1, IG 9,
the purpose of performing alternate depressurization under the conditions of
' the question is to reduce the driving head and flow of any primary leak by
rapidly reducing the pressure. In 0-RO-03-04-09, Rev 1, IG 18 the SRVs are
used because the heat removal capability (40% power) is greater than the
main turbine bypass valves and the RPV will be depressurized sooner. Tha
basis for venting containment, when required, using the torus vents
considers the scrubbing potential of the torus water to support the torus
method as the preferred method. Venting of the primary containment was
not required based on the conditions given in the question. Therefore, the
only correct answer to this question was answer b. There was no change in
the answer key.
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ATTACHMENT 4
SIMULATION FACILITY REPORT
Facility License: DPR 35
Facility Docket No: 50 293
Operating Test Administration: May 6-9,1997
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This form is to be used only to report observations. These observations do not constitute i
audit or inspection findings and are not, without further verification and review, indicative f
of a noncompliance with 10 CFR 55.45(b). These observations do not affect NRC ;
certification or approval of the simulation facility other than to provide information that
may be used in future evaluations. No licensee action is required in response to these
observations.
IIfM DESCRIPTION
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None
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