ML20236E786
| ML20236E786 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 07/20/1987 |
| From: | Collins S, Dudley N, Keller R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20236E772 | List: |
| References | |
| 50-334-87-08OL, 50-334-87-8OL, NUDOCS 8708030089 | |
| Download: ML20236E786 (94) | |
See also: IR 05000334/1987008
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
REQUALIFICATION EVALUATION REPORT-
EVALUATION REPORT NO. 50-334/87-08 (OL)
FACILITY DOCKET NO. 50-334.
FACILITY LICENSE NO. DPR-66-
LICENSEE: Duquesne Light Company
-P.O. Box 4
Shippingport, Pennsylvania 15077
FACILITY.: Beaver Valley Unit I-
EXAMINATION DATES: -May 20-22,.1987
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. CHIEF EXAMINER: #/IfJ
'N. F. Dudley, Le fReactor Engineer
7-d - 6 7
Date
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REVIEWED BY: //
R.-M. Keller, Ihief, Project Section 1C
7/ Zd/I7
Date
APPROVED BY: $MDbl/IM _
JS.-J. Collins, Deputy Director. Division
7fb87
Date
of Reactor Projects
SUMMARY: Six licensed'. Senior Operators (SRO) and six licensed Reactor
Operators-(RO) were administered written and operating examinations. Two
operators failed the operating examination and nine operators failed the=
written examination. The requalification program was evaluated as unsatisfac-
- tory per the criteria provided in the Operator Licensing Examiner Standards,
NUREG-1021, Chapter ES-601, Paragraph F.1.
Operator weaknesses were extensive and individualized. No apparent root cause
for the high number of failures could be identified.
b
B708030009 870724
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(DR ADOCK 05000334: PDR
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DETAILS-
1. EXAMINATION RESULTS:
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R0 SR0' . TOTAL EVALUATION ~ b
Pass / Fail Pass / Fail Pass / Fail ')
Written Examination 2/4 ~1/5 3/9 Unsatisfactory ,
- Oral Examination 5/1 5/1 10./ 2 Marginal i
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Overall Program Evaluation: Unsatisfactory
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Three (3) out.of the twelve (12) licensed operators' examined during this 1
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' evaluation passed all portions of their NRC-administered examinations. l
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- In accordance with the guidelines provided in NUREG 1021, Operator ,
Licensing Examiner Standards, Chapter ES-601, " Administration of l
NRC Requalification. Program Evaluation", the 25% passing rate for this j
examination indicates an unsatisfactory requalification program. j
1. ' Scope of Evaluation
on May 20, 1987, the NRC administered written examinations to six (6)
SR0s and six (6) R0s licensed at the Beaver Valley Unit I Power
Station.
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On May 21 and May 22, 1987, the.NRC administered operating examina-
tions to the operators who had taken the written examinations. ,
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2. Exit Interview l
The Chief Examiner conducted an exit interview on May 22, 1987. The
following persons were present.
NRC Personnel
N. F. Dudley, Lead Reactor Engineer i
S. M. Pindale, Resident Inspector I
Facility Personnel
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J. D. Sieber, Vice President - Nuclear f
W. S. Lacey, Plant Manager {
A. J.'Morabito, Manager, Nuclear Training )
T. W. Burns, Director Operations Training )
T. P. Noonan, Assistant Plant Manager
L. R. Freeland, Nuclear Operations Supervisor
F. J. Lipchick, Senior Licensing Supervisor
L. G. Schad, Simulator Coordinator
R. J. Brooks, Nuclear Operations Instructor
T. E. Kuhar, Nuclear Operations Instructor
L. R. Freeland, Nuclear Operations Instructor
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Summary of NRC Comments
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The examiner reviewed the number and types of examinations which had
been administered as part of the requalification program
evaluation. Two operators had been evaluated as unsatisfactory
during the operating examination. The operators' names and identi-
.fied major. areas of weakness were provided to the Plant Manager so
that appropriate actions could be taken in accordance with the
requalification program. The examiner stated that the details of the
operating examinations and the results of the written examinations
would be provided to the training department, as soon as the evalu- '
ations were completed, so that appropriate actions could be taken in .
accordance with the facility requalification program. '
The examiner stated that there was no consistency during the
operating examinations in the use of procedures during normal and
abnormal evolutions. Some shifts reviewed procedures prior to
performing evolutions while other shifts occasionally reviewed
procedures after completion of evolutions.
The examiner explained that problems with the simulator had required
running an additional scenario in order to adequately evaluate the
operators' ability to use the Emergency Operating Procedures. The
examiner thanked the simulator staff for their support during the
operating examination.
Summary of Weaknesses Identified on Written Examinations:
The following weaknesses were identified during grading the
written examinations. This information is provided for use in
developing future training programs.
RO EXAMINATION
questionNo. Question Topic Area Class Average
1.01 Parameter which effects shape o 33%
Differential Rod Worth Curve
1.08b Relationship between differential Boron 33%
Worth and Fission Product Concentration
2.6a Design basis for MSIV's shutting on steam 33%
line rupture
3.06a Interlocks on Quench Spray Pump Cut Back 0%
Control Valves
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4.3a Basis for maintaining SG water level 40%
during SG Tube Rupture l
SRO EXAMINATION
Question No. Question Topic Area Class Average
5.03a How Xenon Oscillations are effected 17%
by core age
5.5b Basis for maintaining subcooling margin 33%
during Natural Circulation
7.06 Method of Depressurizing the RCS 48%
7.9b Precautions for feeding hot / dry steam 17%
generators
8.5b How to determine that a Special Operating 0%
Order is effective
8.8a Reporting requirements on loss of 120
( vac vital bus 33%
8.8c Reporting requirements on liquid 14%
waste discharge /use of 10 CFR 20 App B
Table.II
8,10c When deviations from procedures can be 50%
made without SRO approval
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The areas ,rrf indicated knowledge deficiencies on the written examinations
comprised only 9% of each examination. On 91% of the written examinations a
majority of the operators demonstrated adequate knowledge in the area being
examined. There was a wide range of grades on each section of the written
examinations and resulted in a spread of 15% to 25% between the lowest and
highest grade on each section. However the range of final grades was much
narrower with a spread of 11% and 16% on the SRO and R0 exar.,f nations
respectively. The average grades on the written SR0 and RO were 77% and 79%
respectively.
Summary of Weaknesses on Operating Examinations:
The following weaknesses were identified as part of evaluating operator
performance during the operating examinations. Details are provided on
individual operating examination report forms. This information is provided
for use in developing future training programs.
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Understanding the basis for tripping bistables,
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Ability to use reference material to identify bistables required to be
tripped.
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Imprecise orders given to I&C technician for tripping bistables.
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Failure to reference abnormal procedures once alarm response procedures
had been completed.
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Repeat back of information.was not conducted nor required when giving i
orders over the phone.
Changes to the Writcon Eemination:
Not all facility comments resulted in changes to the written examination
answer key, however, all comments were considered during grading of the
examinations.
Multiple choice' questions will continue to be used in future examinations.
Additional efforts will be taken to ensure that questions which have been
successfully' used at other facilities are consistent with the Beaver Valley
facility procedures.
The knowledges required of an operator are defined by NUREG-1122, Knowledges
and Abilities Catalog for Nuclear Power Plant Operators: Pressurized Water
Reactors. Utility enabling objectives are also used to define areas of
required knowledges. An enabling objective must define the conditions of f
performance and be of sufficient detail so that trainees can identify the
required depth of knowledge without reference to a terminal objective or an
administrative procedure. The required understanding and knowledge of normal
and abnormal procedures is defined by NUREG-1122 and the facility enabling i
objectives.
Questtog No. Change Justification
1.04 and 5.02 Change "0" to "A". Due to system flow curve
centrifugal pumps in ;
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"* parallel will result in l
increased head and flow.
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1.07e Change to " Decrease". Response was seen in
simulator.
1.08d Delete "initally decreases Initial response to core j
due to fision product aging not asked for in i
buildup then..." question.
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" Change- Justification
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12.01' ' Change " leakage" to " seal Provides' proper
injection". nomenclature.
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'2.05 ' Change "No release" to -
Corrects; release path
. "past' loop seal and out for waste gas' tank
ventilation stack". ' relief' valves.
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3.03 and ADD "or Annunciator Alarm". Includes additional
6.04, correct answers.
3.04c. and Change "Open" to "Close". Provides concensus
6.08 answer for expected
plant response.
- 3.05b- . Change;to "SI-equipment Specifies the major
becomes inoperable due effect on.the Safety
to loss of control power Injection. systra due-
to breakers". to the loss of,a DC bus.
4.03b! Delete "PFR and". -Pressurizer level will not ~
necessarily stabilize when
equalize, due to-heating and
cooling of RCS.
4.03c. . Delete question. Entry conditions into
Emergency. Functional
Recoveries are not required
to be memorized.
4.04b' LAdd "or vapor' space leak". Includes additional correct
answer.
4.07a and; Change to " Feed and bleed Only understanding of overall
7.09a- by manually initiating HPSI actions taken in accordance
and manually opening'the with FR.H-1 is required.
PORV".
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4.07c'.;and Add " Aux Feedwater > 350 Includes additional
7.09c- gpm". correct answer.
'6.01- Delete question. Question not applicable
to plant specific procedures
for tripping failed NI
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channel.
6.06b Add "or raises sump PH". Includes additional correct
answer.
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Question No. Change Justification
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7.04 Delete question. No correct answer was
provided as a choice.
8.08a Add "if #2 Bus is Annunciator alarm panels
deenergized". are powered from #2 vital
bus. .
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8.10b Add " Protect public Includes basis from 10 CFR
health and safety". 50.54(x).
8.10c Change to "Under No Actions taken in accordance
Ci rcumstance s" . with Station's Administration
Procedures Chapter 4, i
Paragraph 3, can only be l
taken with SRO approval. l
Simulator Performance:
The ability to evaluate the performance of the operators during the operating
examination was hampered by simulator problems. One scenario, which resulted
in two phase flow conditions, was invalidated since the simulator response did
not correspond to the laws of nature. The simulator response was a result of j
the failure to enter constants into the model at the proper time. A second !
problem occurred with the turbine plant control system. It was unclear whether l
an operator's inability to properly control turbine load was a simulator l
problem or an indication of an operator weakness.
NRC Follow up:
During a telephone conference conducted on June 2,1987, the licensee was
informed of the number of written examination failures. The licensee was
requested to take appropriate actions in accordance with its requalification
program and attempt to identify root causes for the large number of failures.
The completed operating examinations, the corrected written examinations, and ,
the final written examinations' question and answer keys were mailed to the !
training department on June 3, 1987.
Conclusion:
The requalification program is evaluated as unsatisfactory. No root cause
could be identified for the large number of failures. Reviews of the
unsatisfactory evaluation were conducted considering the licensee's review
comments, the questions dealing with information presented in the last
requalification cycle, and the questions answered incorrectly by the majority
of individuals. None of the reviews changed the pass fail decision for
individual operators nor identified generic training program weaknesses.
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It appears that the requalification program has been unable to maintain a
minimum level of operator knowledge and ability. The average knowledge level
of the operators appears to be uniform. However, the knowledge of specific j
items varies greatly between operators.
The licensee is requested to identify the root cause of the high failure rate
and develop corrective action as required by NUREG-1021, Chapter 601, Parsgraph
F.2.6(1). The short term and long term' corrective actions should be submitted
to the NRC within 60 days.
ATTACHMENTS:
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1. Written Examination and Answer Key (RO)
2. Written Examination and Answer Key (SRO)
3. Facility Examination Review Coa.ments
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~0; 5. NUCLEAR' REGULATORY COMMISSION
REACTOR OPERATOR'REQUALIFICATION EXAMINATION .y ,
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~ FACILITY: BEAVER VALLEY _1
REACTOR TYPE: PWRLWEC3
DATE ADMINISTERED: 87/05/18
EXAMINER: _ SILK. D.
CANDIDATE:
1HEIBURIIQNS TO CANDIDATE:
. Read the attached instruction page carefully. This-examination replaces
.the : current: cycle facility administered requalification examination.
Retraining requirements for failure of. this examination are the.same as
-for-failure'of a requalification. examination prepared and administered by
your training- staff; Points for each question are . indicated in
. parentheses after'the' question. The passing grade requires at least. 70%.
liin=each category and a final grade of =+. least 80%. Examination- papers
.will'be picked up four.(4) hours after the examination etarts.
% OF-
. CATEGORY % OF CANDIDATE'S CATEGORY
__YALUE_ _IQIAL SQQBE VALUE__. CATEGQBX.
15.00__ 25.00 1. PRINCIPLES OF NUCLEAR POWER
PLANT OPERATION,. THERMODYNAMICS,
HEAT TRANSFER AND FLUID FLOW
_15.-00 _21.99 2. PLANT DESIGN INCLUDING SAFETY
AND. EMERGENCY SYSTEMS
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_15 00 25.00 3. INSTRUMENTS AND CONTROLS
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4. PROCEDURES - NORMAL, ABNORMAL,
_15.00 _25 QQ
EMERGENCY AND RADIOLOGICAL
'
CONTROL
_EO.00_ __
% Totals
Final Grade
'All work done on this examination is my own. I have neither given
nor received aid,
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Candidate's Signature
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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. Restroom trips are to be limited and only one candidate at a time may
leave. You must avoid all contacts with anyone outside the exaraination
room to avoid even the appearance or possibility of cheating.
3. Use black ink or dark pencil onlZ to facilitate legible reproductions. )
4. Print your name in the blank provided on the cover sheet of the
examination.
5. Fill in the date on the cover sheet of the examination (if necessary).
6. Use only the paper provided for answers.
7. Print your name in the upper right-hand corner of the first page of each
section of the answer sheet.
8. Consecutively number each answer sheet, write "End of Category __" as
appropriate, start each category on a ngw page, write onlZ 2n one side
of the paper, and write "Last Page" on the last answer sheet.
9. Number each answer as to category and number, for example, 1.4, 6.3.
10. Skip at least three lines between each answer.
11. Separate answer sheets from pad and place finished answer sheets face
down on your desk or table.
12. Use abbreviations only if they are commonly used in facility litgrature.
13. The point value for each question is indicated in parentheses after the
question and can be used as a guide for the depth of answer required.
14. Show all calculations, methods, or assumptions used to obtain an answer
to mathematical problems whether indicated in the question or not.
15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE
QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
16. If parts c the examination are not clear as to intent, ask questions of
the gxaminer only.
17. You must sign the statement on the cover sheet that indicates that the
work is your own and you have not received or been given assistance in
completing the examination. This must be done after the examination has
been completed.
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18. When you complete your examination, you shall: ;
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Assemble your examination as follows: l
a.
(1) Exam questions on top.
(2) Exam aids - figures, tables, etc.
(3) Answer pages including figures which are part of the answer.
b. Turn in your copy of the examination and all pages used to answer
the examination questions,
c. Turn in all scrap paper and the balance of the paper that you did
not use for answering the~ questions,
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d. Leave the examination area, as defined by the examiner. If after
leaving, you are found in this area while the examination is still
in progress, your license may be denied or revoked.
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L1, ' PRINCIPLES OF NUCLEAR' POWER' PLANT OPERATION. PAGE. 2 1
- ' THERMQDlHAMICS. HEAT TRANSFER AND FLUID FLOW s ,. ,
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QUESTION 1.01 (.50) ;
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Which parameter below will have the MOST effect on the shape of a
. Differential Rod Worth Curve?
A. Core radial flux profile
.B. Core axial flux profile
C. Core axial temperature profile !
D.. Time of core cycle
QUESTION 1.02 ( .50)
.For two equivalent positive reactivity additions to a critical reactor,-
how will the SUR at BOL compare with the SUR at EOL7
A. SUR at BOL will be larger
B. SUR at BOL will be smaller
C. SUR will not change
D. Will depend on the original enrichment of the' fuel
QUESTION 1.03 ( .50)
How does Beta bar effective change over core life?
A. Increase
B. Decrease
C. Remain the same
D. Will depend on the enrichment of the fuel
QUESTION 1.04 ( .50)
Choose the answer that most correctly completes the sentence.
"In a' closed system, two single stage centrifugal pumps operating in
parallel will have , as compared to the same
system with one single stage centrifugal pump operating with one j
pump isolated."
A. a higher head and the same flow rate l
B. the same head and the same flow rate ,
C. the same head and a higher flow rate !
D. a higher head and a higher flow rate i
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TakxMODYNAMICS. MAT TRANS m AND FLUID FLOW
1. PRINCIPLES'OF NUCLEAR POWER PLANT-OPERATION, PAGE- .3 1
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QUESTION -1.05- (2.90)
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-Unit 1 isLin Mode 3, 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />'after a reactor' trip'from 100% power,,with a
boron concentration of 1200-ppm, all shutdown banks withdrawn..and a.
- present' core reactivity of minus 5 percent delta K/K. A dilution of boron
concentration occurs increasing ~ source range counts from'120 cps to 196
cys. During this dilution, xenon reactivity changes add 1000 pcm to the
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core. What-is'the new boron concentration? Assume a constant boron worth'
.of 10 pcm/ ppa. SHOW YOUR WORK.
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QUESTION '1.06 (1.50)
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A variable speed centrifugal pump is operating at 1/4 rated speed in
o closed system.with'the following parameters:
Power = 300 KW
- Pump Delta P.= 50:psid
Flow'= 880 GPM
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' What are the new values for these parameters when the pump speed is
' increased to full. rated speed?
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QUESTION l'.07 (1.50)'
t: .Howido each of.the following parameters change (Increase, Decrease, or
No change) if one main' steam isolation valve closes with the plant at
25% load. -Assume all controls are in automatic and that no trip occurs.
a. Affected loop steam generator level (initial change only)
b. Affected loop cold' leg temperature
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c.. Unaffected loop steam generator level (initial change only)
d. Unaffected loop steam generator pressure
e. Unaffected loop cold leg temperature
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QUESTION 1.08 (2.40) I
. Explain how and why the following changes affect the magnitude of differ- !
cntial: boron worth: l
a. Boron concentration INCREASES
b. Moderator temperature INCREASES
c. Fission product concentration INCREASES
d. Core age INCREASES
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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'1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 4
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW
,:.
QUESTION 1.09 (9.35)
Assuming you are operating at 85% power indicate how the following
changes in plant conditions would affect DNBR (Increase, Decrease, or
Remain constant). Consider each separately.
a. The operator withdraws control rods without changing turbine load
b. Axial flux difference changes from 0% to +5%
c. Steam generator PORV fails open
d. Pressurizer heaters are inadvertently left on
e. Reactor coolant pump speed decreases
QUESTION 1.10 (2.35)
>
Compare the. Calculated Estimated Critical Position (ECP) for a startup
to be performed 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after a trip from 100% power, to the ACTUAL con- ,
trol rod position if the following events / conditions occurred. Consider I
cach independently. Limit your answer to Higher than, Lower than, or l
Same as the ECP.
a. One RCP is stopped two minutes prior to criticality
b. The startup is delayed until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the trip
c. The steam dump pressure setpoint is increased to a value just
below the steam generator PORV setpoint
d. Condenser vacuum is reduced by four inches of mercury
e. All steam generator levels are being raised by 5% as the ECP
is reached
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(***** END OF CATEGORY 01 *****) )
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2. ' PLANT DESIGN INCLUDING SAFETY AND EMEEGENCY SYSTEMS PAGE 5
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QUESTION 2.01 (1.80) l
The plant is at 100% power when a loss of all AC power occurs. Explain
how RCS leakage could develop and worsen as a result of having no AC
power.
QUESTION 2.02 (2.00)
Give two problems that result from operating on excess letdown instead
of the normal letdown.
QUESTION 2.03 (3.00)
I
Figure 1 shows the system line up for the normal standby mode of operation
for the Safety Injection System. Indicate what the system lineup would
.be during the hot / cold recirculation phase by circling the valves which
would be in a different position. Assume charging pumps A and B are
operating.
' QUESTION 2.04 (3.00)
a. Explain the RCS pressure response on Figure 2 at the designated
points for the following condition:
One inch cold leg break
Loss of offsite power occurs when the reactor trips
Minimum safeguards safety injection is assumed
b. In accordance with the Emergercy Operating Procedures (EOP), explain
how the operator will mitigate inadequate core cooling conditions
for a small break LOCA if no high head safety injection pumps are
available?
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(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
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2500
h 2000
1500 - Ot
e
$ 1000 -
E
$ 500 -
m ,
O 25 50 75
Time (Minutes)
RCS PRESSURE FOR 1.0
INCH COLD LEG BREAK
FIGURE 2.
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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE. 6
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. QUESTION 2.05' (2.20)
c..What:three gases are retained in the charcoal delay beds? (0.75)
b. What two gases present in the Waste Gas System must be maintained
within limit? (0.6)
c.' List the two components that can be used to relieve an overpressure
condition in the Surge Tank (1GW-TK-2) and, indicate whether actuat-
ion of these components will result in a direct release to the atmos-
phere. (0.85)
QUESTION 2.06 (3.00)
o. Give two reasons (NOT CONDITIONS) why the MSIV's are required to
close during a steam line rupture. 4
b. Which mode (HSB, HZP, HFP) AND what time in cycle (BOL, M3L, EOL)
will have the mest severe effect on a main steam line break accident.
Explain each separately.
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(***** END OF CATEGORY 02 *****)
.________ _ _ _
3. INSTRUMENTS AND CONTROLS PAGE 7.
s
w.
. QUESTION 3.01' (3.00)
With the plant at 50% power and all systems in automatic, the Turbine
First Stage Pressure Transmitter fails low. With no operator action, ex-
' plain the sequence of events leading to a reactor trip and give the cause
of the trip.
QUESTION 3.02 (2.00)-
State which direction rods would move if the Loop 1 control T hot
RTD opened. Explain your answer
QUESTION 3.03 (2.00)
n. What indication would the operator have that a radiation monitor's
power supply has failed? (0.5)
b. What three automatic' actions are initiated by the Fuel Building Vent
Exhaust Monitor reaching its high-high alarm setpoint? (1.5)
'
QUESTION 3.04 (2.50)
- .
If the Component Cooling Water Pump Discharge Pressure Control Valve
Controller setpoint is decreased by 20 psi, indicate how the following
parameters or valve positions will change (Increase, Decrease, Open,
Close, or Remain the same).
l
a. Thermal barrier flow
< b. Neutron shield tank temperature
c. TCV-1CC-100, CCW Hx Bypass Valve
d. Surge tank level
e. PCV-1CH-145, Low Pressure Letdown Valve
..
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
_ _ _ _ - _ - - _ _ - _ _ _ _ _ _ _ _ - _ _ _ - _ - _ - _ . -
- _ - _ _ -
i
3. ' INSTRUMENTS AND CONTROLS PAGE 8
,: .
QUESTION 3.05 (2.50)
c. What will be.the effect on the following major systems by a loss
of vital bus VB27 (2.0)
1. Condenser Dumps
2. Manual Atmospheric Dumps
3. MFW-
4. Makeup
5 Charging. l
b. How will the loss of 125 VDC Switchboard #2 affect the Safety
Injection System? (0.5)
,
QUESTION 3.06 (3.00)
a. Under what conditions, if any, will the Quench Spray Pump Cut Back
Control Valves open when a Motor Electical Protection Trip is
present? (0.9)
b. Under what conditions, if any, will the Chemical Injection Pump Dis-
charge Valves automatically open? (1.2)
c. What TWO conditons should be met before the Quench Spray Pumps are
secured following a Design Based Accident? (0.9)
l
l
l
!
(***** END OF CATEGORY 03 *****)
L
-
- _ - _ - --- - - _ . - - - - - - - - - _ ----- - - - --------__------- - _ _ _- .
- - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
x-
f
->
S . 'L4 ; PROCEDUprR - NORMAL. ABNORMAL. EMERGENCY AND- PAGE 9
RADIOLOGICAL CONTROL
- .
'
'QUESTIONi 4 . 01'. ' ( 1 '. 50 )
In order to maintain the plant at'100%' power, work must be performed by
Ecn operator-inside the? containment in a-radiation ~ field of 400 mrem /hr -
gamma,.and 0.04 Rad /hr from fast neutrons. How=1ong may the. operator
.beipermitted to work in this" area without exceeding his administrative
-
exposure? limit?...Show all calculations and assume the operator has no
exposure this' quarter.
.
QUESTION. 4.02- (2.50)
'
a. What action-should be taken during a liquid waste discharge when the- -
discharge flow rate decreases below the flow rate listed on the dis-
charge permit?- (1.0)
b. An approved RWDA-L for a release is signed on 5/4/87-at 0830. Due-to
various delays, a discharge can not be initiated until 5/7/87 at 1230.
Explain what-action, if-any, should be taken to begin the release?
(1.0)
c. With the Radiation Monitor Recorder out.of service, what action, if
any, should;be taken in order to proceed with the discharge? (0.5)
QUESTION .4 03 (2,50)
.a. While-cooling down'and.depressurizing the'RCS following a SGTR,.why
is it necessary to ensure that the ruptured steam generator water
, level is not permitted to become too low? (1.5)
b. After SI has been terminated during a SGTR, what indication would
the operator have that RCS pressure and the ruptured steam generator '
- pressure have equilized? (0.5)
.c. Under what condition, while combating a SGTR, would the procedure ECA-
3.1, SGTR with Loss of Reactor Coolant - Subcooled Recovery Desired, 1
be entered? (0.5) ;
i
1
-
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
- _ _
,
!
l
4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 10
R&DIOLOGICAL CONTROL
3:_.
QUESTION 4.04 (1.50)
Answer the following questions regarding AOP-13, Malfunction of Pressur-
izer pressure control. Consider each separately.
I
a. Pressurizer pressure is rapidly increasing to the PORV lift setpoint.
Is it permissible to open two PORY Isolation Valves to ensure that a
PORV is available to reduce pressure? (0.5)
,
b.- What event will give the following simultaneous symptoms? (1.0)
Increasing pressurizer level
Decreasing pressurizer pressure
QUESTION 4.05 (2.50)
'
Answer the following questions regarding AOP-29, Loss of 120 VAC Vital
Bus 1:
a. VITAL BUS 1 TROUBLE ALARM-Al-10 and VITAL BUS 1 BATTERY OPERATION
ALARM A1-18 come in. What check needs to be made to determine
whether to proceed to AOP-29 or to investigate further? (0.5)
b. In the event of a loss of the 120 VAC Vital Bus 1, why is it recom-
mended to restore power to the bus as rapidly as possible? (0.5)
c. How much time is allotted to re-energize the bus before the operator l
is required to initiate a manual reactor trip? (0.5)
d. What conditions must exist to cause a possible SIS, SLI, or CIB
actuation'when transferring power to the auxiliary source after
losing Bus 17 (1.0) l
l
1
QUESTION 4.06 (2.00) i
i
List, in order, the immediate actions of the " Response to Nuclear i
Power Generation ATWS" procedure FR-S.I.
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
I
_ _ _ . _
i
4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 11
RADIOLOGICAL CONTROL :
.
.m
QUESTION 4.07 (2.50)
a. What is the difference between " bleed and feed" and " feed and bleed"
and which is used during FR-H.17 (1.4)
b. During restoration of secondary heat sink, what cautions should be
,
observed if feed flow can be established when all steam generators
are classified as hot / dry steam generators? (0.7)
c. What criteria is used to determine if an adequate heat sink is avail-
able? (0.4) i
e
,
(***** END OF CATEGORY 04 *****)
(************* END OF EXAMINATION ***************) :
_ _ - _ _ _ _ _ _ _ _ - _ _ . _.
_-
_
f = ma- :v--= L s/t Cycle efficiency = (Networt-
! -
out)/(Energy in)
7 = mg s = Vo t + 1/2 at 2
2
/E mc ,
,
- 'KE =.1/2 mv 'a=(Vf - Vg )/t' A = AN A=Aeg
PE:= mgn .
' '
yf.= V, + at w = e/t x = in2/t1/2 = 0.693/t1/2.
~
1/2 eff = [(t 39)(t))
~
- t
b y
, [(t1/2) + (t b)3-
AE = 931. am
~
, ,
I=Ieg
f: Q = mCp at
I = UAa T- I = I g e~"*
, Pwr = W fan I = I, 10-x/TVL
'
TYL = 1.3/u
P:= P 10 sur(t)~
g HVL = -0.693/u
p = p e /T t
~ ,
.SUR = 26.06/T SCR = S/(1 - K,ff)
CR
x =S/(l'-Keffx)
SUR = 250/t* + (s - p)T .
CR)(1 - X,ff)) = CR2 (1 - keff2)
'
T.= (t*/o) + [(s - 9)/lo] M = 1/(1 - K,ff) = CR)/CR g
T . =: t/(s'- 8) M = (1 -'K ,77,)/(1 - K,ff))
T ='(8 - o)/(lo) - K,ff)/K,ff
'
SOM = (
-p = (X,7f-1)/K,fff = aK,f f/K,ff 1* = 10 seconds
I = 0.1-seconds-I
p-= [(t*/(T K,ff)] + [T,ff (1 / + $T)]
. =1d
P = '( reV)/(3.x :10
10
7 I)d)
I)d ' . b=2
y 222 ' '
' I d " ~~
'
2
I = oN R/hr = (0.5 CE)/d (meters)
' Water Parameters Miscellaneous Conversions
'
1 gal. = 8.345 lbm. l curie = 3.7 x 1010 dps .
I gal. = 3.78 liters 1 kg ='2.21 lom .
lift 3 = 7.48 gal. I hp = 2.54 x 10 3 Btu /hr
Density = 62.4 lbm/ft3 1 mw = 3.41 x 106 Btu /hr -
Density = 1 gm/cm3 lin = 2.54 cm
Heat.of vaporization = 970 Btu /lom *F = 9/5'C + 32
.' Heat of fusion = 144 Btu /lbm *C = 5/9 (*F-32)
1.Atm = 14.7 psi = 29.9 in. Hg.
~
-
.
i
D 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 12
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW
'ARSWERS -- BEAVER-VALLEY 1&2 -87/05/18-SILK, D. ,,,,
I
ANSWER. 1 01 ( .50)
b (Core axial flux profile) (0.5)
REFERENCE
BVPS-Rx Th, chapter 8, pgs 14-20
3.1 001 000 K_5.11 3.1
i
ANSWER 1.02 ( .50)
,
b. (0.5)
REFERENCE
BVPS Rx Th, chapter 5, pgs 15-18, 21-25
3.1 001 000 K 5.47 2.9
.
1
ANSWER. 1.03 ( .50)
b.- Decreases (0.5).
REFERENCE
BVPS Rx Th, chapter 5, pgs 15-18, 21-25
3.1 001 000 K 5.47 2.9
J
ANSWER 1.04 ( .50)
A +F (0.5)
REFERENCE
BVPS Thermo Manual, chapter 4, pgs 32-33
Component: 191004 PUMPS K 1.05 2.3
K 1.09 2.4
!
I
i
< ,
l
l
i.
L
L l '. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 13
l IllERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW
ANSWERS -- BEAVER VALLEY 1&2 -87/05/18-SILK, D. ,,
1
l
ANSWER 1.05 (2.90)
rhol = (Keffl - 1)/ Keffi -0.05
Keff1 = 1/((1 - (-0.05)) = 0.9524 [0.5]
120(1 - 0.9524) = 196(1 - Keff2)
Keff2 = 0.9709 [0.5]
rho 2 = (0.9709 -1)/0.9709 = -0.03 [0.25]
delta rho = rho 2 - rhol = -0.03 - (-0.05) = 0.02 = 2000 pcm [0.4]
1000 pcm is due to xenon, so the remaining 1000 pcm is due to boron [0.5]
change in boron concentration for 1000 pcm is:
1000 pcm / 10 pcm/ ppm = 100 ppm [0.25].
new boron concentration = 1200 - 100 = 1100 ppm [0.5)
REFERENCE
BVPS Reactor Theory Manual Chapter 5, page 49; Chapter 9, pages 2-9.
learning objectives 5-1/13;7-1/3
001010K524 004000A404 ...(KA'S)
ANSWER 1.06 (1.50)
Power (2) = Power (1) * (H2/N1)**3 = 300 * 4**3 = 19.2 MW (0.5)
Delta P(2) : Delta P(1) * (N2/N1)**2 = 50 * 4**2 = 800 psid (0.5)
Flow (2) = Flow (1) * (N2/N1) = 880 * 4 = 3520 gpm (0.5)
REFERENCE
BVPS Thermo Manual, chapter 4, pgs 32-33
Component: 191004 PUMPS K 1.05 2.3
K 1.09 2.4
I
1
1
.I
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 14
THERMODYNAMICS. HEt.T TRANSFER AND FLUID FLOW
ANSWERS -- BEAVER VALLEY 1&2 -87/05/18-SILK, D. ., s
ANSWER 1.07 (1.50)
n. Decrease
b. Increase
c. Increase
d. Decrease
e. Nc Change [0.3 each]
DElen% 95
REFERENCE
BVPS Thermo Manual, chapter 7, pgs 26-27
chapter 6, pg 37
3.2 002 000 K 5.01 3.1
5.09 3.7
5.11 4.0
ANSWER 1.08 (2.40)
1. Differential boron worth (DBW) decrease (0.3) because the boron atoms
are competing with each other for neutrons (0.3)
2. DBW decreases (0.3) as moderator density decreasing (moves boron atoms
farther apart) decreasing neutron capture probability in boron atoms
(0.3)
3. DBW decreases (0.3) because poisons are competing with boron atoms (0.3)
4. DBW initi:lly decrences due te fierien prod"et b"4 ! A"r tk-=4dnavaases
(0.3) due to boron depletion (0.3)
REFERENCE
BVPS Rx Th, chapter 8, pgs 34,37,45
3.1 001 000 K 5.20 3.2
5.28 3.8
5.30 3.1
.______--___-______- - _ -
_ __ _ ______ _ ____
l
-
, l'. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 15
. THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW
-ANSWERS ---BEAVER VALLEY 1&2 -87/05/18-SILK, D. ,
ANSWER 1.09 (2.35) ;
a. Decreases
b. Decreases
c. Decreases
d. Increases
e. Decreases (0.47 pts each)
REFERENCE
BVPS Rx Th, chapter 7, pgs 12-18
3.2 002 000 K 5.01 3.1
5.09
'
3.3
3.3 003 000 K 5.01 3.3 i
ANSWER 1.10 (2.35)
.a. Same
b. Higher
c. Higher
d. Same '
e. Lower (0.47 pts each)
REFERENCE i
'
BVPS Rx Th chapter 9, pgs 2,3
3.1 001 000 K 5.18 4.2
i
L
l
l
l
l
l
.
h
l
l
L
I
2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 16 I
ANSWERS -- BEAVER VALLEY 1&2 -87/05/18-SILK, D.
w.
ANSWER 2.01 (1
SES 1N3ECT[0W. 80) ~ '
,
Loss of RCP Le$kmp> could casue degradation of the sealing capacity [0.9]. !
causing leakage along the RCP shaft [0.9]
REFERENCE
BVPS EOP Ex Vol ECA 0.0 pg 3 of 123
'BVPS OH 1.6.1 pas 24-25
Module 1, Failure Mechanisms, LP-LRT-VII-69, pas 6,7; EO-1 ,
(
3.2 002 000 K 1.13 4.1 >
3.7 062 000 K 3.01 3.5
'
ANSWER 2.02 (2.00)
Activity of the coolant and impurities will increase.
Borations and dilutions take longer, limiting rate of power changes.
No hydrogen addition takes place to-prevent corrosion.
[any 2 @ 1.0 each]
REFERENCE
1987 RO Annual License Examination, Session 3, queation'2-5.
CVCS NS-5, p ns-5-1, fig. NS-5-1
004000K106 004020A202 ...(KA'S)
ANSWER 2.03 (3.00)
See Figure 1.1 26 valves 0.11 pts each
REFERENCE
Module 1, SI & CNMT Depress System, LP-LRT-V-48, pgs 8-10; EO-3
-3.2 006 000 K 4.05 4.3
4.06 3.9
006 020 K 4.04 3.8
,
_ -
-
'
y
YV [
'
>
n
4
I
y4"~&"'
wI'~ , @_ ~ ~
I
a - ,.
-
.M]M
I, , .
_ .
,
J ,
L &
-
M'hO,- ,
,
3
e_ - o.2 =,i 1. b "
%)
,
_
. - -
s
-
.d
.
-
.
%,P
-
m u , 7 : ,, l
p
_r
H ,
._== L.
T T
o O -
= a H
O
a
_
%
= w T ,
- -
w
y P5 @_ _._P
H
1
A
= A b.
a ,.
"
c a
Ti . .g, 4 =
,
4 > -- = = L b n%-
emg
=_::o
-
>
o
>
g
4<> X
@ *
. f
i
.
i
e c
g. =.u .
s _
_-g%.A1
=
gpuA.
c
=
gT
4
y,,,.Q mL . 4 s a- -
~ 4= a " r
"
"r ,, = .
4 a
"=m.,_
' '
(
.
" " #
==~
=
c M
3 " -
- =~- E
-
s T
a
S
o
%. w_ m_ o
u
m @ ,ssV: .
.
YE
S
n f ' ,T.
i
0
f
"m
,
,yr-aE4_
_
c
F! s8
-
E O'I
4 N
V1
5
O
. I
O14al,6$_s. s7
W-H
C 5
-
-
- 5
a- 5
-
O
L
u
4~ T
ca
)
W=%cde, J
d c vle4
ju
A
u v
=1
= NA
'
- -
. o
s
,
=- i
e- I
Ve@,,P
s
- _S su g' = YO
o3
v g sg, TN
%g
" wa 4s~ -- -
".
1I
= E
F
m xW A
Aa
, i
xss
-
-
d S
- __ . - - _ _ _ _ - _ _ _
2. -PLANT DESIGN _ INCLUDING SAFETY AND EMEEGENCY SYSTEMS PAGE 17-
ANSWERS -- BEAVER VALLEY 1&2 -87/05/18-SILK, D.
v.
ANSWER 2.04 (3.00)
a. 1. Immediately following reactor trip, the RCS rapidly depressurizes
since only a fraction of the heat previous to the trip is now
being transferred to the primary fluid [0.75]
2. Equilibrium pressure is achieved when decay heat product and ,
removal are matched [0.4] and SI flow matches leak flow [0.35] 1
b. The operator would use the atmospheric steam dumps to cool down and
depressurize the RCS [0.7)
Accumulators'will inject water into the core (0.4]
LHSI will inject water into the core [0.4]
REFERENCE.
Module 1,' Loss of coolant transients, LP-LRT-VII-70, pgs 3-7; EO-4
Recovery from loss of Rx coolant w/o HHSI, LP-LRT-VII-74, pgs 11,12
EO-3,6
3.3 000 009 EK 3.06 3.9
3.11 4.4
3.27 3.6
ANSWER 2.05 (2.20)
a. Xenon, Krypton, Iodine (0.25 pts each)
b. Hydrogen, Oxygen (0.3 pts each)
c. Pressure' control valve and Rupture disc (0.3 pts each)
I' M: ::lence. (0.25 pts)
PAST LOOP 5fN. Mean vidt trM
REFERENCE
Module 4, Gaseous Waste System Review, LP-LRT-V-55, pgs 2,4,15; EO-3b,4,
3.11 071 000 K 1.06 3.1
4.04 2.9
4.06 2.7
System Generic K&A 5 2.4
l
l
!
,
. ___ _ - _ _ . _ _ _ . _ _ _ _ . _ . _._._._____.J
.,
. -_-- _ - -. . . _ _ . ____ _ - _ _ . ___ . _ _ _ .. . - _ . . .
'
- te 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 18
ANSWERS - BEAVER VALLEY-1&2 -87/05/18-SILK, D.
w,
i
ANSWER 2.06' (3.00)
a. 1. Minimize' positive reactivity effects of RCS cooldown associated
with the blowdown- [0. 75] ver
2. . Limit pressure rise within containment during a steam break
in. containment (0.75]' ,
b. Hot Zero Power [0.35].because of the greatest mass in the SG results.
in the largest RCS cooldown [0.4]
,EOL [0.35] because MTC is at its maximum negative value [0.4]
REFERENCE
-Module 2, LOSC Transients, LP-LRT-VII-78, pgs 17,23-26; EO-3
MS.& SGFW system, LP-LRT-V-50, pgs 12-14; EO-4
T/S B 3/4 7-1
T/S B 3/4 7-3
FSAR 14.1-35 to 38
3.5 000 040 EK 3.01- 4.5
EK 2.01 2.5
EK 1.05 4.4
.3.5 039 000 K 4.05 3.7
.
j
l
_
'
! '
4,
zp ,
3 '.' ' INSTRUMENTS AND CONTROLS' PAGE 19
, ANSWERS --~ BEAVER VALLEY 1&2 -87/05/18-SILK, D.
' % "A ,,..
ANSWdN * 3.01 (3.00)
(Governor ~ valves open in an attempt to restore Pimp (increasepower)
oRods insert due to turbine load / reactor power difference and Tave/ Tref
mismatch
oACS cooldown causes inventory to shrink
FZR level; decreases as Tave decreased 3d / 0 -
oPlant trips on low'PZR pressure when~PZR empties (GMG pts each) H
REFERENCE
Module 2, SGWLC system, LP-LRT-III-39, Fig 1, EO-3,4 i
1.26.1.pg 89 Training rystems Descriptions PGS-6,
1.1.5 Fig'l-14 pg 6-53, Objective 1
o 1.1.1 pg 12~
u 1.6.1 pg 60
3.9 016 000 K 1.11 2.3 3.2 011 000 K 4.02 3.2
1,12 3.5 3.5 039 000 K 1.04 3.1
s
AdSWER 3.02 (2.00)'
Rods move in (0.5)
T hot indicates high [0.3] causing Tavg from loop i to increase (0.3]
~Tavg auctioneered high and sent to temp' mismatch channel for rod
control channel. [0.5]
Tref - Tavg signal will go negative causing rods to move in.[0.4]
i
REFERENCE
Reactor Control System NS-10, p NS-10-10 : EO 2
001010A301 ...(KA'S)
.
-
.
t
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . - _ _ _ _ _ _
_
W ) e>:
3. INSTRUMENTS AND CONTRO(,S PAGE 20-
r -ANSWERS -- BEAVER VALLEY'1&2 -87/05.'18-SILK, D.
,: ,.
'!
l ANSWER 3.03 (2.00)
a. FailJgre0nlinhtisextinguishedforthatmonitor(0.5)'
on .n wNu czA rQ ALngn
b. Closes SLCRS: filter bypass. damper-
Open main filter bank. inlet damper
Actuates fuel bids evacuation alarm (0.5 each)
- REFERENCE
LModule 4,..RMS. review, LP-LRT-V-C6 pgs 4,5,25,27; EO-1,2,3,4
6
3.9 072 000 E 2.011 2.3
'
'4.01~ 3.3-
4.02 3.2
4.03 3.2.
.. . Components: 191002 Sensors & Detectors K 1.18 2.6
' ANSWER 3 . '0 4 ' .(2.50)
a. Decrease
'b. Increase.
%:1 CLCSs
'
c.
d. Remain the same
e..- Remain the same. (0.5 pts each)
.,- . REFERENCE.
Module 3,'Rx plant component and Neutron Shield Tank Cooling System,
LP-LRT-V-53, pgs 8-13; EO-4
BVPS-OM 1 15.1 pas 3,15,16
3.10 008 000LK 1.02 3.3-
_ __ _ - __--____ _ _ m
_ _ __ -
3. INSTRUMENTS AND CONTROLS- PAGE 21-
~ ANSWERS --LBEAVER VALLEY 1&2- -87/05/18-SILK, D.
m
LANSWER' 3 '. 0 5 ' (2.50)-
< ;m.1. Not available. [0.4]-
. .
2.-Availaible [0.4]
3. A;+ C available [0.1]: A.in manual.[0.1] ;
B in auto-hold [0.1]
"
. . -
C in auto / manual-[0.1]
4.: Auto:noTavailable-[0.3]
no flow indication [0.1]
57 AH-CH-122 in manual -[0.2] (rans opsd
Master' controller to auto hold [0.2]
b. CCR to RWST. Refrigeration Unit will isolate [0.1]
BIT recirculation isolation Valve SHUT [0.2]
$t Epust ne ~ Wrsupply to SIsocMa
de cones Accumulators fails
Ans e o ac rt shut
loss o' [0.2]edt
co w Madd f# "*'O
REFERENCE
Electrical Dist. Review, LP-LRT-V-59, p 21, 51; EO 4, 7
-000057A219- 000058A203 ...(KA'S)
i
- ANSWER -3.06 (3.00)
{ut.ve Waaur cars)
a. CIB actuated"(0.5).and RWST not at low-low level (0.4)
b. CIB signal present
Chemical addition tank not low-low
Motor' electrical-protection trip not present (0.4 pts each)
c. CNMT pressure < -1.0 psig (0.5).-and CIB reset (0.4)
REFERENCE
Module 1, SI & CNMT Depress.' systems, LP-LRT-V-48, pas 14,15; EO-6
-BVPS EOP E-1 step 10
3.6 026 000 K 4 02 . 3.1
4.04 3.7
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-4. PRQQEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 22
RADIOLOGICAL CONTROL-
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ANSWERS -- BEAVER VALLEY 1&2 -87/05/18-SILK, D. . ;
ANSWER 4.01 (1.50)
With Form 4 on file he is permitted 3 Rem /Qtr (0.5)
0.4 Rem /hr + 0.04 Rad /hr x 10 Rem / Rad = 0.8 Rem /hr (0.5)
3.0 Rem /0.8 Rem /hr =.3.75 hrs = 225 minutes (0.5)
REFERENCE
BVPS-RCH pgs 5-7
..............----------------------------------------------------------- --
Plant Wide Generic 15 3.9
ANSWER 4.02 (2.50)
a. Secure discharge immediately (0.5) and notify Shift Supervisor (0,5)
b. A confirmatory sample should be analyzed to extend the effective per- _
iod of authorization since its 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> limit was exceeded (1.0) l
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c. Every 15 minutes log the readings from the radiation monitor (0.5) !
REFERENCE
Module 4, Liquid Radwaste System Review, LP-LRT-V-54, pg 15, EO-6
.
3.11 068 000 K 4.01 3.4
068 System generic 1 2.7
ANSWER 4.03 (2.50)
c.6
c. Prevents exposing the steam space to cold water [Gv44 0.6
which would deprssurize the SG and increase pri. to sec. leakage [Gv61
Thus preventing reduction in RCS pressure and reinitiation of SI [Or5-]
.
o. 3
b. Water levels in the P R ndcaffected SG will stabilize [0.5]
REFERENCE
Module 3, Operator Response to SGTR, LP-LRT-VII-87, pgs 15,16,10,4, EO-1
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4. PROCEDURES - NOBBAL d BRQBMAL. EMERGENCY AND PAGE 23
RADIOLOGICAL CONTROL
ANSWERS -- BEAVER VALLEY 1&2 -87/05/18-SILK, D. .s .
3.3 000 038 EK 3.01 4.1
EK 3.06 4.2
ANSWER 4.04 (1.50)
a. No [0.5]
e'
b. PORV or Safety Valve is open$po[d 51.0] fact IJAK
REFERENCE
AOP-13, Malfunction of PZR press control, pgs 2,5,4
3.3 010 system generic 1 3.5
3.3 000 008 EA 2.12 3.4
ANSWER 4.05 (2.50)
a. Check #1 DC Bus volts [0.5] (n . A.aro- nao j p r c su. 3rne su naro)
b. Regain control of the FRV (which f ails open) to-avcid SC cverf444 [0.5]
c. 5 minutes [0.5)
d. Another channel CNMT HI or HI-HI pressure bistable trip [0.5]
Bistable on effected channel is NOT bypassed [0.5]
REFERENCE
AOP-29, Loss of 120 VAC Vital Bus 1,.pgs 1,2
3.7 000 057 EK 3.01 4.1
ANSWER 4.06 (2.00)
Verify Rx trip
i Sound the standby alarm and announce the problem
l Verify turbine trip
Verify AFW pumps running
Initiate emergency boration of the RCS
[0.4 each]
REFERENCE
1987 RO Annual License Examination, Session 3, # 4-7
FR-S.1
000029S011 ...(KA'S)
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'4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 24
]1 RADIOLOGICAL CONTRQL
ANSW RS -- BEAVER VALLEY 1&2 -87/05/18-SILK, D. ,
I
ANSWER. 4.07 (2.50)
a. Bleed end feed' Manually initiate HP SI, manu/ ally opens the
PORVs [0.7]
Feed and bleed: Mnnen11;- initiete MP SI, re -it s +ka o n + n = = + 4 r-
veling Of the *0"?: te . ent RCS in . ente:;- [0. ?]
b. Feed only one SG until Thot < 550 F [0.4] then feed all SCs [0.3]
c. >5%NRlevelinatleastoneSG[0d]
A sit FEEC a:ATCg > 350 p [0.Q
REFERENCE
Module 2 , Oper Response to LOSHS, LP-LRT-VII-81, pgs 1,6,7; EO-1,2,4
BVPS IIOP F-0. 3
BVPS ICOP FR-H.1 pgs 15,16
000054K304 ...(KA'S)
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U..S. NUCLEAR REGULATORY COMMISSION ,,,, ,
SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION- l
FACILITY: BEAVER VALLEY 1
'
REACTOR TYPE: PWR-WEC3
DATE ADMINISTERED: 87/05/20
EXAMINER: DUDLEY. N.
CANDIDATE:
INSTRUCTIONS TO CANDIDATE:
Read the attached instruction page carefully. This examination replaces
the current cycle facility administered requalification examination.
Retraining requirements for failure of this examination are the same as
- for failure of a requalification examination prepared and administered by !
your training staff. Points for each question are indicated
'
in
parentheses after the question. The passing grade requires at least 70%
in each category and a final grade of at least 80%. Examination papers
will be' picked up four (4) hours after the examination starts.
% OF
CATEGORY % OF CANDIDATE'S CATEGORY
VALUE TOTAL SCORE VALUE CATEGORY
15.00 25.00 5. THEORY OF NUCLEAR POWER PLANT
OPERATION, FLUIDS, AND
THERMODYNAMICS
15.00 25.00 6. PLANT SYSTEMS DESIGN, CONTROL,
AND INSTRUMENTATION
_15.00 25.00 7. PROCEDURES - NORMAL, ABNORMAL,
EMERGENCY AND RADIOLOGICAL
CONTROL'
15.00 _25100 8. ADMINISTRATIVE PROCEDURES,
CONDITIONS, AND LIMITATIONS
60.00 % Totals
Final Grade
!
All work done on this examination is my own. I have neither given
nor received aid.
Candidate's Signature i
. . _ . . _ - .- - . - . - . . . . . . . . . - - - . . - . . . . . - . . . . . . . . . . - . . _
_____ - _ _ _ .
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS
During the administration of this examination the following rules app'ly:
1. Cheating on the examination means an automatic denial of your application
and could result in more severe penalties.
2. Restroom trips are to be limited and only one candidate 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.
3. Use black ink or dark pencil only to facilitate legible reproductions.
4. Print your name in the blank provided on the cover sheet of the
examination.
5. Fill in the date on the cover sheet of the examination (if necessary).
6. Use only the paper provided for answers.
7. Print your name in the upper right-hand corner of the first page of each
section of the answer sheet.
8. Consecutively number each answer sheet, write "End of Category __" as
appropriate, start each category on a nag page, write 2nlE 2n anc aida
of the paper, and write "Last Page" on the last answer sheet.
9. Number each answer as to category and number, for example, 1.4, 6.3.
10. Skip at least three lines between each answer.
11. Separate answer sheets from pad and place finished answer sheets face
down on your desk or table.
12. Use abbreviations only if they are commonly used in facility literature.
13. The point value for each question is indicated in parentheses after the
question and can be used as a guide for the depth of answer required.
14. Show all calculations, methods, or assumptions used to obtain an answer
to mathematical problems whether indicated in the question or not.
15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE
QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
16. If parts of the examination are not clear as to intent, ask questions of
the 2Xaminar only.
17. You must sign the statement on the cover sheet that indicates that the
work is your own and you have not received or been given assistance in
completing the examination. This must be done after the examination has
been completed.
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18'.-When.you complete your examination, you shall:
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a. Assemble your examination as follows:
.(1) Exam questions on top.
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(2) Exam aids - figures, tables, etc. ;
(3) Answer pages including figures which are part of the answer.
1
b. Turn in your copy of the-examination and all pages used to answer
the examination questions.
c. Turn.in $11 scrap paper and the balance of the paper that you did
not.use for answering the questions,
d. Leave the examination area, as defined by the examiner. If after
. leaving, you are found in this area while'the examination:is still
in progress, your license may be denied or revoked.
.
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5! ' THEORY'OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND 'PAGE 2'
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THERMODYNAMICS
,~.
' QUESTION 5.01. -(2.50)
A reactor.startup is infprogress with rods in manual, Tavs at 547 F,
~
Pressurizer pressure at 2235 psig, steam generator pressure at 1005 psig,
and steam. dump pressure! controller in Pateam-mode at 1005 psig. Reactor
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power is'at 10(E-8) on the intermediate range with a'SUR of 1.0 DPM.
a. At time =.0, steam dump pressure controller set point-is reduced to 995
psig.- . Explain ~how and why the SUR will initially change if no operator
"
action is taken.
.b. If an automatic-' reactor trip does not occur, what'should the SUR be at
' time'= 5 minutes? Briefly explain your answer.
c. If an automatic reactor trip' occurs at time = 10 minutes. A SUR of
e -1/3 DPM: is observed. What is the basis'for this SUR?
- r
QUESTION 5.02 ( .50)
Choose the answer:that most correctly completes the sentence,
e
In the condensate system, when two condensate pumps operate in parallel
.they will have -(choose from below) , as compared t,o when one pump
is operating ~with the other pump isolated.
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A. a higher head and a higher. flow rate.
B. a higher head ~and the same flow rate.
C. the same head and the same flow rate.
.D. the same head and a higher flow rate.
QUESTION 5.03 (2.00)
a.:Do xenon oscillations converge (dampen) more rapidly at BOL or EOL7
Justify your answer in terms of reactivity effects,
b. Would the. magnitude and frequency of xenon oscillations be greater at
50%. power or 100% power? Justify your answer.
QUESTION 5.04 (2.00) l
List the four parameters affecting DNB and state whether the " margin to
DNB" increases or' decreases due to an increase in that parameter.
.
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(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) l
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THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS. AND
~
5. PAGE 3
l
THERMODYNAMICS
9 . l
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QUESTION 5.05. (2.25) {
.a. Calculate.the subcooling margin with the plant at the following f
conditions during a natural circulation cooldown: i
~P steam = 1005.0 psig
P Pzr = 1535.0 psig i
T hot = 550.0 F.
T cold = 530.0 F. !
T core exit thermocouple = 540.0 F. '
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b. What could happen: to natural circulation flow rate if subcooling margin
decreases to less than 20 F7. Briefly explain why.
c. Briefly explain why a natural circulation cooldown rate greater than
25 F/Hr can cause a bubble to form in-the' vessel head while a cooldown
rate of less than 25 F/Hr does not.
QUESTION 5.06 (3.00) !
Unit 1 is in-Mode 3, 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after a reactor trip from 1000 power, with a
boron concentration of 1200 ppm, all shutdown banks withdrawn, and a
present core' reactivity of minus 5 percent delta K/K. A dilution of. boron
concentration occurs-increasing source range counts from 120 cps to 196
cps. During this dilution, xenon reactivity changes add 1000 pcm to the
core. What is the new boron concentration? Assume a constant boron worth
of 10 pcm/ ppm. SHOW YOUR WORK.
-QUESTION- 5.07 (2.75)
Using heat transfer equations, explain how:
a. Boron precipitation during a design base LOCA can cause increased
fuel clad temper 1tures. State all assumptions made and all equations
used. (1.1)
b. :The plugging of steam generator tubes can reduce main generator
electrical ouput. Assume no change in core thermal output, Tavg, or
turbine. steam flow and state all equations used. (1.65)
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(***** END OF CATEGORY 05 *****)
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E. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATIQH PAGE 4
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QUESTION 6.01 ( .60)
DEL ETE
The trip bistables of a failed power range detector are placed in the
trip condition by which of the following?
A. Placing the applicable bistable test switch in the " Test"
position in the Reactor Protection Cabinet.
B. Removing the applicable control and instrument power fuses
on the power range drawers.
C. Placing the applicable Power Mismatch Bypass switch to the
failed position at the Miscellaneous Control and Indication F
Panel. L
D. Placing the applicable Comparator Channel Defeat switch to
the failed channel position at the Detector Current Comparator
Panel.
QUESTION 6.02 ( .60)
Which of the following malfunctions will result in both a low Tavg
indication and a low delta T indication?
A. Hot leg RTD failed high
B. Hot leg RTD failed low
C. Cold leg RTD failed high
D. Cold leg RTD failed low
QUESTION 6.03 (1.00) .
'
The plant is in mode 3 with all shutdown banks withdrawn. If the
intrument power fuses are to be removed from the intermediate range
neutron flux detector N-35 control cabinet, what action must be taken
to prevent a reactor trip?
QUESTION 6.04 (1.80)
'
a. What indication would the operator have that a radiation monitor's
power supply has failed? (0.6)
b. What automatic actions are initiated by the Fuel Building Vent
Exhaust Monitor reaching its high-high alarm setpoint? (1.2)
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6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 5
w.
QUESTION 6.05 (1.50)
a. What will be the effect on the Charging System of a loss of vital
bus VB2?
b. What will be the effect on the Main Feedwater System of a loss of
vital bus VB37
c. What will be the effect on the Safety Injection System of a loss of
125 VDC Switchboard # 27
QUESTION 6.06 (2.50)
a. How and why is quench spray flow automatically reduced to 1100 gpm per
train upon a low-low level in the RWST, assuming that both quench spray
pumps are running? (1.2)
b. inw.is a sodium hydroxide solution added to the quench spray system?
(0.6)
c. Why is 450 rpm of quench spray flow diverted to the suction of the
recirculation pumps? (0.6)
QUESTION 6.07 -(1.50)
a. During solid plant operations, how is RCS pressure contolled?
b. State the basis for maintaining pressure within specific temperature
dependent pressure limits.
c. Would the RH Relief Valve, RV-1RH-721, open if the RHR pressure
interlock point was reached and all systems operated properly?
QUESTION 6.08 (2.50)
If the Component Cooling Water Pump Discharge Pressure Control Valve
Controller setpoint is decreased by 20 psi, indicate how the following
,
parameters or valve positions will change. Answer: Increase, Decrease,
Open, Close, or Remain the same,
l
a. Thermal barrier flow
l b. Neutron shield tank temperature
c. TCV-1CC-100, CCW Hx Bypass Valve
d. Surge tank level
e. PCV-1CH-145, Low Pressure Letdown Valve
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
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6. -PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 6
m,
QUESTION 6.09 '(3.00)
For the following instruments, state what type of failure will lead
to a high steam generator. water level. Briefly describe the mechanisms I
by which the instrument failure will cause high level. i
a.-Steam flow -
b. Steam generator pressure ']
a
c. Feed flow l
d. Steam generator level
e. Turbine impulse pressure
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(***** END OF CATEGORY 06 *****)
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7. PROCEDURES ~ NORMAL.~ ABNORMAL. EMERGENCY AND PAGE 7
RADIOLOGICAL CONTBQL
, .s
-QUESTION 7.^1 ( 60).
,
During a natural. circulation cooldown conducted in accordance with
.ES-0.2, " Natural' Circulation Cooldown" which one of the following
criteria determines the requirements _for the amount of RCS subcooling
which must be maintained?
'A. RCS cooldown rate.
B. Decay heat rate, resulting from reactor power history.
C. Pressurizer level.
, D.. Number of CRDM fans running.
QUESTION 7.02 ( .60)
During a reactor startup and power escalation, when is it allowed to
place the rod control system in automatic?
A. When the Reactor is critical.
B. When Tavg is within 2 F of Tref
C. When the main feedwater regulating valves are in automatic
D. When reactor power is above 15%
QUESTION 7.03- ( .60)
According to AOP-25 " Loss of Reactor' Plant River Water", which one
of the following actions should be taken if neither of the RP River
Water pumps can restore header pressure while the plant is at 100%
Power?
'A. Reduce power to prevent temperature alarms on the RCPs.
B. Reduce power to no load Tavg.
C. Commence a reactor plant shutdown.
D. Trip the reactor.
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(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 8
RADIOLOGICAL CONTROL
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QUESTION 7.04 ( .60) 4
Whic h below statements describe the RCP trip criteria following a
SI with normal containment conditions?
A. Less than 30 degrees subcooling AND par level less than 4%.
B. No charging /HHSI pump flow indicated AND no CCR flow to RCP
indicated.
C. No charging /HHSI pump flow indicated AND RCS pressure less than
1380 psig.
D. RCS/ Highest SG DP less than 145 psid AND par level less than 4%.
QUESTION 7.05 ( .60) !
The Response Not Obtained for the first immediate action of EOP-FR-S.1
" Response to Nuclear Power Generation / ATWS" is to manually trip the
reactor. Select the next action to be taken if the reactor will not
trip.
A. Place rods in Manual and insert them into the core.
B. Trip the turbine and verify steam dumps open.
C. Emergency borate the RCS.
D. Dispatch operator to locally trip reactor.
QUEST. ION 7.06 (1.40)
What are the three methods, in order of preference, of depressurizing
the Reactor Coolant System after a Steam Generator tube rupture?
-QUESTION 7.07 (1.50)
While touring through the auxiliary building, you come to an area
that has been roped off and posted with a Radiation Area sign. Five
feet within the area is a valve that produces a 2500 mrem /hr field
at 18' inches. How should this area be posted? Justify your answer.
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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'7. PROCEDURES NORMAL. ABNORMAL. EMERGENCY AND PAGE~ 9 i
RADIOLOGICAL CONTROL
v b
-QUESTION 7.08 (1.80)
a. An approved RWDA-L for a release is signed on 5/4/87 at 0830. Due to-
various delays, a discharge can not be initiated.until.5/7/87:at 1230.
Explain what action, if any, should be taken to begin the release? i
(1.0)
b. With the . Radiation Monitor Recorder out of service, what action, if
any, should be taken in order to proceed.with the discharge? (0.6)' <
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. QUESTION 7.09 (2.50)
e..What is.the difference between " bleed and feed" and " feed and bleed"
and which is used during FR-H.17 (1.4)
b..During restoration of secondary heat sink, what cautions should be
observed-if feed flow.can be established when all steam generators
are classified as hot / dry steam generators? (0.7)
.c. What criteria is used to determine if an adequate heat sink is avail-
able? (0.4)
QUESTION 7.10 (2.00)
AnswerLthe following questions regarding AOP-13, Malfunction of
Pressurizer. pressure control. Consider each separately.
,
a. Pressurizer pressure is rapidly increasing to the PORV lift
setpoint. Is it permissible to open two PORY Isolation Valves to
ensure that a PORV is available to reduce pressure? (0.4)
b. What event will give the following simultaneous symptoms? (0.6)
Increasing pressurizer level
f Decreasing pressurizer pressure
c. One pressurizer pressure protection channel fails high.
. Immediately after the order has been given to trip the bistables
on.the failed channel a second pressure protection channel fails i
low. What action, if any, should be taken? (1.0)
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7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 10
RADIOLOGICAL CONTROL i
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QUESTION-~ 7.11 (3.00)
i
a. Indicate why each of the requirements provided below must be met
in' order to. terminate SI in accordance with E-0. .
1. Total feed flow to intact SG > 350 gym (0.6)~ f
'2.-RCS pressure stable or increasing (0.6) j
3. PRZ level > 5% (0.6) j
b. What'TWO plant conditions require manual reinitiation of SI? (1.2) j
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(***** END OF CATEGORY 07 *****)
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8 ', ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS. .-PAGE 11 l
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QUESTION 8.01- ( .60)
-Unit J is' operating at 100% thermal power with,AFD in the target band when q
the SS discovers an error in the calculation of the cumulative penalty- !
,
-deviation time. Recalculation shows a cumulative penalty deviation time
of 1.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> during.the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. .According to the provided. >
LTechnical Specifications, which of the following actions is required?
,
.A. Remain at'100% power as long as'AFD is within the target band.
B. Reduce' thermal power to less than 90%, within 15 minutes.
, .C. Reduce thermal power to less than 50%, within 30 minutes.
D. Commence a shutdown within one hour.
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QUESTION 8'.02 (. 60)
Which on'e of the following situations requires action to be taken in one
' hour in accordance with the Technical Specifications?
A. In Mode 1 with one full-length rod immovable.
B. In Mode 5 with one pressurizer code safety valve inoperable.
C. In Mode 2 with two. pressurizer PORV's failed shut.
D. In Mode.1 with two charging pumps inoperable.
QUESTION 8.03 ( .60)
Which of.the following is the correct definition for Heat Flux Hot Channel
Factor?
A. The' ratio of the integral of linear power along a rod with
the highest integrated power to the average rod power.
B. The maximum local heat flux on the surface of a fuel rod at i
core. elevation Z divided by the average fuel rod heat flux.
C. The ratio of peak power density to average power density in
the horizontal plane at core elevation Z.
D. Maximum excore detector calibrated current divided by average
excore detector calibrated current, for the upper or lower
detectors, whichever is greater.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
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8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 12
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QUESTION 8.04 ( .60) ;
Which one~of the following conditions requires immediate action in
-cccordance with the Technical Specifications, if the plant is in' Mode'37
A. The shutdown margin is reported to be 1.8 % delta K/K.
B. One train of heat tracing on the BAT becomes inoperable.
C. Two~ reactor coolant pumps trip.
D. Two of the three charging pumps fail.
QUESTION 8.05 (1.50)
n. What' actions should the NSOF or NSS take if a lead seal.on a
valve is reported to be broken?
b. How can an operator determine whether a Special Operating Order that
was written four days ago is still effective?
QUESTION 8.06 (1.00) t
When must a supervisor be present for work in a subatmospheric
containment?
QUESTION 8.07 (2.00)
'
The plant is operating at 75% power and the latest leak rate data shows:
12.2 GPM - Corrected RCS leakage rate
1.5 GPM - Leakage into the Pressurizer Relief Tank
1.5 GPM - Leakage into the Primary Drains Transfer Tank
.
3.2 GPM - Leakage through SI-23, RCS Loop 1A, cold leg isolation
(Previous leakage rate was 1.6 GPM)
0.4 GPM - Primary to secondary leakage in SG #1
0.2 GPM - Primary to secondary leakage in SG #2
0.2 GPM - Primary to secondary leakage in SG #3
4.0 GPM - Leakage rast RCP seals
What RCS leakage limits, if any, have been exceeded? Refer to attached
Technical Specifications.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
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'8. : ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 13
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LQUESTION '8.08.
(2.10)
- For each of.the following' events indicate what type of notification,
if any,,should.be made to the NRC. Consider each event separately and
JUSTIFY your decision.
a. A 120 vac vital' bus is accidently deenergized while in mode 1.
.lf. CIA is initiated as part of a surveillance test while in mode 4.
c. The chemist reports that due to a calculational error 1.1 X.10E-6
uCi/ml of'soluable I-133 had.been released during the last liquid
waste discharge which had lasted 15. minutes.
QUESTION 8.09 (3.00)
For each of the following situations indicate whether-the equipment
should be' considered operable in accordance with the Technical
Specifications and. JUSTIFY your answer.
a.1 Diesel Generator #2, if the air supply valves from the starting air
tanks are found shut.
,
b. A charging pump, if the control switch is in " pull-to-lock". ,
c. Safety equipment on Emergency Bus 1DF, if the #2 Diesel Generator
is inoperable.
!
QUESTION. 8.10 (3.00)
,
Answer the following concerning Adherence to Operating Procedures.
a. What two conditions do not require a procedure to be present at j
the location, open and readable?
b. What specific instanceLwould allow deviation from procedures, i
license conditions or Technical Specifications WITH the approval j
of an'SRO? j
c.'What specific-instance would allow deviation from procedures, i
license conditions or Technical Specifications WITHOUT the approval
of aus SRO7
)
j
I
(***** END OF CATEGORY 08 *****)
(************* END OF EXAMINATION ***************) !
l
'
, .
'
b
L- .- f'=-m:- v= s/t . Cycle efficiency = (Netsorx
"
out)/(EnergyLin)
- w = mg .
s.= Vg t + 1/2 at 2
<
E ='mc m
'
KE = 1/2 mv .
a=(Vf - Vg )/t A = AN A=Aeg -
PEj=mgn .
I
'
Lyf = V, +'at w'= e/t-
A= in2/t1/2 = 0.693/t 1/2
F - t
1/2'If " E(*1D)IIb )) j
' , [(t1/2) + (tb )) ,
. AE = 931. am
, ,
!=Ie' g j
- Q' = mCpat
d = UA 4 T- 1 = I g e~"*
'
Pwr =.W an I~= I 10-x/TVL
, f n
l TVL = 1.3/u
5
- P = Po l0 "#5*) HVL = -0.693/u
P = Po e*II .
- SUR = 26.06/T SCR = S/(1 - K,ff)
CR
x = S/(1 - Keffx)
- SUR = 25p/ t* + (s-p)T'.
CR)(1 - Keffl) = CR2 (.1 - keff2)
.- T=(t*/o)'+[(s-p)/ho] M = 1/(1 - K,ff) = CR /CR j g
T = 1/(o - 8)' M = (1 - K,ffo)/(1 - K,ff))
T = (8 - o)/(Ap)
'
SDM = ( - K,ff)/Keff
a=(K,ff-1)/K,f[=d,ff/K,ff t" = 10 seconds
I = 0.1 seconds-I
i = [(t=/(T K,ff)] + [T,77/ (1 + b)]
10
.
I 3d) =Id
P = '( r+ V )/(3. x '10 y
.
I)dy ' .2.,2
, 7d~~~'
222 ~
2
r = oN R/hr = (0.5 CE)/d (meters)
Water Parameters Miscellaneous Conversions
~
1 gal. = 8.345Libm. l curie = 3.7 x 1010 dps .
Igaj.=3.78' liters 1 kg = 2.21 lem
1 ft = 7.48 gal. I hp = 2.54 x 103 Btu /br
Density = 62.4-1bm/ft3 1 mw = 3.41 x 106 Etu/hr
,
'
Density = 1 gm/cm3 lin = 2.54 cm ,
Heat of vaporization = 970 Stu/lom 'F = 9/5'C + 32 l'
Heat of fusion = 144 Btu /lbm *C = 5/9 (*F-32)
~'
1 Atm =: 14.7 psi = 29.9 in. Hg.
'
.
w, . ,, .. ..8 -. s.. , , . . , . . . . . - -.. .
- - . . . -- ---w
- _ _ _ _ _ - _ _ _ _ _
Nuclear Regulatory Commission Part 20, App. B
APPENOlX D-CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND--Continued
(See footnoles at end of Appendix 2'
isotope * Table i Tatde la
Element (atomic numtm) Col 1-Aa Col. 2- g,4 Col. 2-
(pO/m0 *'
[C#)*m'n (#0'#*0 g
nodine (53) a 125.. S 5 x 10 * 4 x 10-' 8x10-" 2 x 10-'
I 2x10" 8 x 10' 6 x 10-' 2 x 10"
1 126 .S e x to-' 5 x 10" 9 x 10-" 3 x 10"
g 6 3 x 10-' 3 x 10" 1 x t0-8 9 x t0"
6129 S 2 x 10-* 1 x 10" 2 x to-" 6 x 10-8
1 7 x 10" 6 x 10-s 2 x 10- 8 2 x 10"
I 131 S 0 x 10" 6 x 10-8 1 x 10'" 3 x 10-'
I 3 x 10-' 2x t0-8 1 x t0-' 6 x 10 *
1132 S 2 r to' 2 v 10 8 3 v 10 8 8 x try'
I 0 x 10" 6 x 10- o 3 x 10-' 2x l0"
6133 S 3 x 10 ' 2 x 10" 4 x 10-" 1 x 10 '
I 2x10" 1x10 8 7 x to" 4 x 10 8
1 134 S 6 x 10" 4 x 10" 6 x 10-' 2 x 10-8
I 3 x 10-' 2 x 10-' 1 x 10-' 6 x 10-*
1 135.-- S 1 x 10" 7 x 10-' 1 x 10-' 4 x 10-*
i 4 x 10 ' 2 x 10- s g x go-e 7x 10 '
1rwAum (77) Ir 190 S 1 x 10-* 6 x 10-a 4 x 10" 2 x 10-*
4 4 x 10" 5 x 10-8 1 x 10-* 2 x 10-'
k 192 S 1 x t0-' 1 x 10 8 4 x to-' 4 x 10-8
1 3 x 10-8 1 x 10-e 9 x 10-" 4x10"
er 194 . S 2 x 10" 1 x 10- 8 8x10" 3 x 10"
i 2 x 10-' 9 x t0-* 5 x 10-' 3 x 10-s
tron (26) . Fe 55 S 9 x 10" 2 x 10-' 3 x 10-8 8x10"
1 1 x 10-8 7 x 10-8 3 x 10-* 2 x 10-8
Fe 59. S 1 x 10-' 2 x 10-' 5 x 10-' 6 x 10-8
4 5 x 10-e 2 x 10- 8 2 x t0-' 6 x 10-8
Krypton (36) Kr85m Sub 6x 10-8 1 x 10-'
Kr 85. Sue 1 x 10" 3 x 10-'
Kt 67 Sub 1 x 10-' 2 x *0-*
Kr 86. Sub 1 x 10-* 2 x 10**
Lanthanum (57). LJ 140 - S 2x10" 7 x 10-* 5 x 10" 2 x 10"
I 1 x 10" 7 x 10-* 4 x 10-' 2 x 10* *
Lead (82) PD 203 S 3 x 10- * 1 x 10-s g x io-e 4 xion
i 2 x 10-* l x 10-8 6 x 10-' 4 x t0-*
PD 210 S 1 x 10- * 4 x 10-* 4 x 10-" 1 x 10-'
I 2x 10-" 5x 10-' 8x10-" 2 x 10-*
Pb 212 S 2 x 10* * 6 x 10-' 6 x 10-" 2 x 10-'
1 2 x 10 8 6 x 10- * 7 x 10 " 2 x 10"
Lutetsum (71) ... . ..
Lv 177.. S 6 x 10" 3 x 10-a 2 x 10-8 1 x 10"
1 5 x 10" 3 x 10 8 2 x 10-' 1 x 10"
Manganese (25l._ Mn 52 S 2 x 10" 1 x 10 8 7 x 10-' 3 x 10-*
I 1 x t0-8 0 x 10" 6x t0-' 3 x 10"
g,,3:",. t,th
Mn 54 5 4 x 10" 4 x 10-8 f x to-' 1 x 10"
l 4 x 10-e 3 x to a g x to-e g x gon
un 56. . . . . . S 8 x 10" 4 x 10-8 3 x 10-8 1 x 10"
l 5 x 10" 3 x 10-' 2 x 10-8 1 x 10"
Mercury (80) .. Hg 197m . _. S 7 x 10" 6 x 10-8
3 x 10". , 2 xgo.
gx 10".
l 8 x 10-' 5 x 10.a 3 x go
Hg 19 7.,, . . . S 1 x 10-* 9 x 10- 8 4 x 10 8 3 x to"
1 ( 3 x 10-' 1 x 10-8 9 x.10 * 5 x to-*
'
Hg 203 S 7 x 10 ' 5x t0" 2 x 10-' 2 x 10"
i l x to" 3 x 10 8 4 x 10-' 1 x 10 *
Molyodenum (42) ... .
Mo 99 . .. S 7x10" 5 x to" 3 x 10" 2 x 10"
l 2 x 10-' 1 x 10-8 7 x 10-8 4 x 10"
Neodymium (60) .. .
Nd 144 . S 6 x 10-" 2 x 10* * 3 x 10 7 x 10"
1 3 x 10- * 2xt0s gxto-n a x io-a
Nd 147 _ S 4 x 10 ' 2 x 10- 8 1x to" 6 x 10"
1 2 x 10" 2 x 10-8 8 x 10" 6 x 10-'
Ped 149.., . S 2 x 10-8 6 x 10" 6 x 10-8 3 x 10-*
1 1 x 10- s 8 x 10 8 5 x 10-8 3 x 10-*
Neptunsum (93) .. .
No 237 .., .
S 4 x 10- " 9 x 10" 1 x 10 3 x 10
i 1 x 10 " 9 x 10- * 4 x 10 " 3 x 10
Np 239 _ . S 8 x t0" e s 10 8 3 x 10" 1 x 10"
l ! 7 x 10 - ' 4 x 10- 8 2x10-* 1x10"
Neckel (26) . N.59 .. .S i 5 x 10 ' ' 6 a 10- 8 2 x 10 * 2 x 10"
277
.
i
I
_3 /4. 2 POWER DISTRIBUTION LIMITS
AX1AL FLUX DIFFERENCE (AFD)
LIMITING CONDITION FOR OPERATION +~
3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained
within a i 7% target band (flux difference units) about the target flux
. difference.
APPLICABILITY: MODE 1 ABOVE 50% RATED THERMAL POWER *
ACTION: )
1
a. With the indicated AXIAL FLUX DIFFERENCE outside of the + 7% I
target band about the target flux difference and with THERMAL
POWER:
1. Above 90% of RATED THERMAL POWER, within 15 minutes:
a) Either restore the indicated AFD to within the target
band limits, or
b) Reduce THERMAL POWER to less than 90% of RATED THERMAL
9 2.
POWER.
Between 50% and 90% of RATED THERMAL POWER:
a) POWER OPERATION may continue provided:
1) The indicated AFD has not been outside of the
1 7% target band for more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> penalty
deviation cumulative during the previous 24
hours, and
2) The indicated AFD is within the limits shown on
Figure 3.2-1. Otherwise, reduce THERMAL POWER te
less than 50% of RATED THERMAL POWER within 30
minutes and reduce the Power Range Neutron Flux-
High Trip Setpoints to < 55% of RATED THERMAL
POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />,
b) Surveillance testing of the Power Range Neutron Flu)
Channels may be performed pursuant to Specification
4.3.1.1.1 provided the indicated AFD is maintained
within the limits of Figure 3.2-1. A total of if
hours operation may be accumulated with the AFD
outside of the target band during this testing withcut
penalty deviation.
- See Special Test Exception 3.10.2
Amendment No. $, 17
BEAVER VALLEY - UNIT 1 3/4 2-1
l
,
I
. _ - _ _ A
-
-
.
. . . . . . . _ .
. _ . _ . _ . . 1
. . .
4
,. .
.q
!
POWER DISTRIBUTION LIMITS
.
l
r
4- LIMITING CONDITION FOR OPERATION (Contin *ued)
\ ../
.
'b. THERMAL POWER shall not be increased above 90% of RATED' THERMA -i
POWER unless the indicated AFD is within the + 75 target band-
-
.
and ACTION 2.a).1), above has been satis.fied.~ j
c. THERMAL POWER shall not be increased above 50% of RATED THERMAL
POWER unless the indicated AFD has not been outside of the
'
t,7*. target band for more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />' penalty deviation. o
cumulative.during the previous 24, hours.
- '
~
SURVEILLANCE REQUIREMENTS
!
i
4.2.1.1. The indicated AXIAL FLUX OIFFERENCE shall be determined to be
within its limits during POWER OPERATION above 15% of RATED THERMAL POWER
by: .
,
- a. Monitoring the indicated AFD for each OPERABLE excore channel:
,
1. At least once per 7 days when the AFD Monitor Alarm is
OPERA 8LE, and
2. At least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after
A restoring the AFD Monitor Alam to OPERABLE status.
..-
b. Monitoring and logging the indicated AXIAL FLUX DIFFERENCE for
each OPERA 8LE excere channel at least once per hour for the
first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least once per 30 minutes thereafter,
when the AXIAL FLUX DIFFERENCE Monitor Alarm is inoperable.
The logged values of the indicated AXIAL FLUX OIFFERENCE shall
be assumed to exist during the interval preceding each logging.
4. 2.1. 2 The indicated AFD shall bc considered outside of its + 7% target
.
band when at least 2 of 4 or 2 of 3 OPERABLE excore channels are indi
the AFD to be outside the target band., POWER OPERATION outside of the + ' ~
'
target band shallibe accumulated on a time basis of:
a. One minute penalty deviation for each one minute of POWER
OPERATION outside of the target band at THEFAL POWER levels
equal to or above 50% of RATF.D THERMAL POWER, and
i
.! One-half minute penalty deviation for each ene minute of PCWER
.: b.
OPERATION outside of the target band at THERMAL POWER levels
belcw 50% of RATED THERMAL PCWER.
J.
' <
'
!
..
3/4 2-2 /cendment No. 9
,d SEAVER VALLEY - UNIT 1
I . l
l
J
--__ _ - _ _ _ _
. .
,
.
- POWER DISTRIBUTION LIMITS ~
.
~
. - '- 't
_ _ .
- SURVEILLANCE REQUIREMENTS (Continued)
_
_ _ .
__-
-.
- - - -
~
4.2.1.3 The target flux difference of each OPERABLE excore channet
shall be detemined by measurement at least once per 92 Effective Full
Power Days. The provisions of Specification 4.0.4 are not applicable.
4.2.1.4 The target flux difference shall be updated at least once per
.
31 Effective Full Power Days by either detemining the target flux
difference pursuant to 4.2.1.3 above or by linear interpolation between
the most recently measured value and 0 percent at the end of the cycle
The provisions of Specification 4.0.4 are not applicable.
~
life.
l
-
.
.
e
e
h
o
Y
?
e
.
0
l
Amendment No, f J
, BEAVER VALLEY - UNIT 1
3/4 2-3
l
.
8
- - - - - - - - _ . - _ - - - _ _ _
._ z
7 .
'
.
.
. .
. -
-
1
gpg
- ::::L:::: ::::J:nt' ::::*-
- - :::={::::' n:: :=:c= ::af.:=:.- ::-
- a.:tr
+
g-- - : =- := :d=:::f=- :::: :;::l::::::
- = m - :::d:*=::::.::: ::::::::h::.-
-.' - - -
-'::li i.I*g is ". :-4 - :;i..E
- p::=.; 5.
--
_-_t.ii. , _iE_i ;; ;
- =. p_._4. _ ..
. . ._
. .
.
. . . ., ,. . . . . - . - _ . . ... , - -
. . ;
m:A .-: 4
- J-
_. _
w
-.
-m _
_: - H<: 4.
- .
__
.
3
m::s . _ . - i
-
s
e :::w.
-:
..e
- _
> at::a:2 m :
7 -
.
.
100
. - -..
E gUNACCEPTA8
- OPERATION -
LEj( 11,90t h,(11,90)iEUNACCEPTABLE ~~-
..
l
"
-'"-""
'
!
80 -
-*! . . ,
.. 1
y _
-
.
'
_
j
.
- I
\
1
'I
-
.iiEACCEPTABLEf: OPERATION
~ -
I
-
60 , n -- .
_
.
-
__
. ( 31.501_ - (31,50)
.
.
. -
"EE E I l -
M M
- __
.
_
20 _ _ _
. .m
-- ::::- . : -- t :=.t.:::: qm
. . . . . t .- -
_m .
__
4 --- .
.-.*
- -
w
0
50 40 30 20 10 0 10 20 30 40 50
FLUX DIFFERENCE (tl) %
i- FIGURE 3.21 AX1AL FLUX OIFFERENCE LIMITS AS A FUNCTION OF RATED -
THERMAL POWER
4
I
3/4 2-4 Amendment No.
a
<
BEAVER VALLEY - UNIT.1 '
k
.
t
_ __ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ - _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ . _ _ _ .
_- -_-- . _ - _ _ - - - _ _ _- _ _ _ _ _ _ _ _ _ -_
-
'
?
POWER DISTRIBUTION LIMITS
.
~
REAT FLUI BOT CHANNEL FACTOR-F q (Z)
- -
- _ ' f. . -. _ ,
,._
-- . .
LIMITING CONDITION FOR OPERATION
.
3.2.2 F q(Z) shall be limited by the following relationships:
,
Fq (Z) 4 [2.32] [K(Z)] for P >0.5 .
P !
>
Fq (Z) i ((4.64)] [K(Z)] for P4 0.5
where P = THIRMAL POkT.R
RATED mnMAL POWER
and K(Z) is the function obtained from Figure 3.2-2 for
a given core height location.
APPLICA3ILIIT: h0DE 1
ACTION:
s
.
1 With F (Z) exceeding its limit:
v Q
a. Reduce TIGIMAL POWER at less; 1% for each 1% nF (Z) exceeds the
limit within 15 minutes and similiarly reduce the Power Range
Neutron Flux-Eigh Trd; Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />;
POWER OPERATION may ps,ceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />;
subsequent POWER OPERATION may proceed provided the Overpower
.6 T Trip Setpoints have been reduced at least 1% for each 1%
Fn_(Z) exceeds the limit. The Overpower 4T Trip Satpoint
rtduction shall be performed with the reactor suberitica1.
b. Identify and correct the cause of the out of limit condition
prior to increasing THERMAL POWER; THIR$1 POWER may then be
increased previded Fq (Z) is de=enstrated through incore
mapping to be withf.n its li=1t.
t
i
,
_) ,
stiver viu.Er - mT 1 us 2-3 ^== * == 3o-
. qs
-
.
l ,
.
_ . _ _ _ __-_- __ - _ - _ _ . _ _ _ _ _ .
r
G Eh'D'S~R:3UT!CN Lito!T5
SURVEILLAt:CE RECUIRD;EtiTS
- ~
.
,,
.
4.2.E.1 The provisions of Specification 4.C.4 are not applicable."
-- --
.4.2.2.2 F,y shall be evaluateo to cetemine if F g (Z) is within its
limit by: -
a. . Using the movable incere detectors te obtain a power distribu-
-
tion map at any THERMAL POWER greater tnan 5'. of RATED THERMAL
POWER.
.b. . Increasing the measured F,y component of the pcwer distribution
cap. by 31 to account for manufacturing tolerances anc furtner
increasing;tne value by 5 to account for measurement uncer-
tanties.
'
c. Comparing the F , computed (F C) obtained in b', aoove to:
1. The F'xy limits for RATED THETelAL PCWER (FRTP ) for the
xy
appropriat'e measured core planes given in e anc f belew,
and ..
.
2. The relationship:
l RTP
F
xy
=Fty [lv0.2(1-P)1
1
(- wnere Fxy l'is\the limit for fractional THET0tAL PCWER
operation expressed as a functicn of F f and P is
the fraction of RATED THET@iAL PCWER a t anien 2 was
measured. ,y
xy according to the follcwing senecule:
d. Remeasuring F
C RTP
1. When F
xy
is greater than the F xy limit for the accrepria te
g
measured core plane but less nan :.e F xy rela ticr.:ni;,
additicnal power distribution a;s shall ta taken anc
0
F
xy ccccared to F xy RTP anc Fxy'
' : !
a) Either within 24 neurs af ter ex:eecing by 2C'. of .
RATED THEPf!AL PCWER cr greate , ne THEP"A'. ?CWER l
C
at which F was last cete 'tec, or
Xy
b) At least once per 31 EF?D, wnicrever cccurs first.
Seaver Valley " nit 1 3/4 2-5
v Amendment No. 73
.
. . _ _ _ . . _ . _ _ . _ _ _ - _ _ _ _ _ _ _ . _ -
_ _ _
_ _ _ _ _ - - -
POWET. DISTR:30 TION LIMITS,
.
-
SURVEILLANCE REQUIREMENTS (Continued)- __ _.
'"-
'
.C
2. When the F,C is less than or' ecual to the F,Rlimit for the
~~
~' appropriate measured core plane, additional power distribution
maps shall be taken and F C Compared to F RTP and F' - at least
once per 31 EFPD. *# *# #7
e. The F limit for Rated Themal Power (FNE) shall be provided . for
xy y
.
all ccre planes containing tank "D" control rocs and all unrodcec
core planes in a Radial Feaking Facter Limit Report per specification
6. 9 .1.14. .
i
f. The F
xy limits of e, above, are not applicabIe in the following core
plane regions as measured in percent' of core height from the bottom
of the fuel:
1. Lower core region from 0 to 15t, inclusive.
2. Upper core region from 85
3. Grid plane. regions at 17.6$o2%, 100%
32.1inclusive.
1 2t,
'
46.4 + 2%, 60.'6 ; 2% anc 74.9 1 2%, inclusive
4. Core plane regions. within 2 2% of core heign: {I 2.85 inches)
. bout the bank cemand pcsition of *he tank C" centrol
k- rods.
g. With F,C exceeding F , the effects of F
xy en F; (Z) shall be
evaluated to determine if Fq (Z) is within its limit.
4.2.2.3 When F Z is measured pursuant to.5pecificatien 4.10.2.2, an
overal9 m(ea)sured F be obtained frem a pcr,er distribution
n (Z) shall
map and incraased cy 3t to account for manufacturing tolerances anc
further increased by 5% to acccunt fer measurement uncertainty.
EEAVER VALLEY - UNIT 1 3/4 2-6a
-( s
.
Amendment No. 73
I
_ _ _ - _ _ _ _ - _ _
. _ _ . _ _ _ _ _ _ .
.
K(I) - MORMALIZED FQ (2)
+-
( AS A FUNCTION OF CORE HEIGHT
-
K-LOOP
.
BEAVER YALLEY - UNIT l .
.
.s,:
.=.
-Q :=
.
- :::
-
un
6.0. 1.0) _.~
l. 0 =%, ~~._
-
mm.
a t -- - (10.6, 0. 94 5_
. .
_
g,3 __
_-
_
N
-
[ _-
n:
-
), 6 : .,
r
.=.
,.
b
y -_
.-
~
f
8 (12 s 074 ;
i
~
-C..
-x
_ _ _
-
J
'-
._
O.2 -
_ :::::
=
'
i~:="",
-. --==5
. - ,_2
'~_~'
2 4 6 5 10 12
!
-*
j COPS HEIGHT (FT)
i -
1 (,, ) -
Figure 3.2-2 AMENL ME::T .'!O . . -
. .
l
3.1VER "ALLET - LTNIT 1 3/+ 2-7
_ _ _ _ _ _ _ _ - _ - _ _
_ - _ - _ _ _ _ _ _ _ - _ _ _ __ _ - _ _
t.'
n
3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE
LEAKAGE DETECTION SYSTEMS
m.
I
LIMITING CON 0! TION FOR OPERATION
,3.4.6.1~ The following Reactor Coolant System leakage detection systems
shall be OPERABLE:
a. The containment atmosphere particulate radioactivity monitoring
system,
b. The containment sump discharge flow measurement system or narrow
range -level instrument, and
c. Containment atmosphere gaseous radioactivity monitoring system. !
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
a. With one of the above required radioactivity monitoring leakage i
detection systems inoperable, operations may continue for up to
30 days provided:
1. The other two above required leakage detection systems are
OPERABLE, and
2. Appropriate grab samples are cotainec and analy:ed at least
once per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s:
otherwise, be in at least HOT STANDSY within the next 5 neurs anc
in COLD SHUTDOWN within the follewing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
t. Witn t."e containment sump discharge ficw measure.ent sys em anc
narrcw range level instrument inoperable, restore at least One
inoperable system to OPERABLE status witr.in 7 cays or ce in a-
least HOT STANDBY within the next 6 hcurs anc in' CCLD SHUTCCWN
within the follcwing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
c. The provisiens of s; edification 3.3.4 are n:t a:clicacie in Mcces
1, 2 anc 3.
l
SL'oVE!LLANCE RECUIR?ENTS
4.4.6.1 The leakage detection systems snali :e dem: stra:e: CPEF,A5;.I :y:
a. Containment a *: eschere par-icG a*e an: :asecus ment:: ring
system-performance of CHANNEL CHECX, CHANNE;. CALI5 RAT:0" a :
-
[ CHANNEL FUNCT:CNAL TEST a: :re faecuent e: s:eci#ie: in
Tacle '. 3-3,
EEAVER VALLEY - UNIT 1 3/4 a-1; .
Amencment N:,3C
. s.
,
__ _____.___ _ _
._
_ - _ - _ .
_ _.
[EACTOR COOLANT SYSTDi
SURVEILLANCE REQUIREMENTS (Continued)
-
..-
b. Containment sump discharge flew measurement system-perfomance
of CHANNEL CALIBRATION TEST at least once per 18 months.
c. Logging the narrow range level indication every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. l
l
\
l
.
.
..
.
,
\
BEAVER VALLEY - L,'l*T 1 2 /4 4 ;; Amendment '4c .30
%.
_ _ _ _ - _ _ _ _ _ _ . _ _ _ _ ._. -- - - - - - - - - . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - --
- -__ __- - __ _ _
C C T.~ ~ * D
CO. J d-
[ ,: .
~
. .- . .- . - -
LIMfTING CONDITION FOR OPERATION
_
.
' 3.4.6.2 Reactor Coolant System leakage s' hall be limitad to:
a. No PRESSURE SOUNDARY LEAXAGE,
.
b.- 1 GPM UNIDENTIFIED LEAKAGE,
c. 1 GPM total primary-to-secondary leakage through all steam
generators not isolated from the Reactor Coolant System and
.
~
$00 gallons per day through any one steam generator not isolated
from the Reactor Coolant System,
d. 10 GPM IDENTIFIED LEAKAG, E from the Reactor Coolant' System, and
.?
e. 28 GPM CONTROLLED LEAKAGE at a Reactor Coolant System-
pressure of 2230 +20 psig. ,
- /
APP!.ICABILITY: MODES 1, 2, 3 and 4.
j )
W ACTION: ,
a '.
With any PRESSURE BOUNDARY LEAKAGE, be in at.least ACT STANCB
within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b. With any Reactor Coolant System leakage greater than any ene
of the above limits, excluding PRESSURE SOUNDARY LEAKAGE,
reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be
in at lease HOT STANDBY within the .next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD
SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS -
.
j
-
4.4.6.2 Reactor Coolant System lekkages shall be demonstrated to be
! within each of the above limits by:
- -
a. Monitoring the containment atmosphere particulate and.gasecus
- i
!
,
radioactivity monitor at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. :
1, l
~l L,
,
.I ~' - 3/4 4-13
BEAVER VALLEY - UNIT 1
_
,__ _. ___ ._ __ ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
"CP:T"^
. ' ' 10
.
- ^
'
. PY ::: C.
..
., : .
T.ha _J,J REA30R COOLANT SYSTEM-
-
SURVEILLANCE REQUIREMENTS (Cont'.nued)
b. Monitoring the containment sump discharge at least once per 12
hours.
,
. c. Measurement of the CONTROLLED LEAKAGE to the reactor coolan
pump seals when. the Reactor Coolant System pressure is
2230 + 20 psig at least oncq per 31 days with the modulating
valve ~ full open,
'i,
Perfonnan'ce of a Reactor Coolant System water inventory balance
at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during steady state operatien, and
.
e..
Monitoring the reactor head flange leakoff temperature at ,least '
once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. .
4
e .
.-
'#
,
9
6
0
e
S
4
'
!
.
BEAVER VALLEY - UNIT 1
3/4 4-14
- ._ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_ - _ _ _ _ _ _ _ .-_
..
l
-
b- PRESSURE ISOLATION VALVES
'
LIMITING CONDITION FOR OPERATION'
i
i 3.4.6.3 Reactor coolant system pressure isolation valves shall be
operational. l
_.
. APPLICA9ILITY' Modes 1.,2, 3 and 4.' ,
c )
Action: l
1. All pressure isolation valves listed in Table 4.4-3 shall
be functional as a pressure isolation device, except as ;
specified in 2. Valve leakage shall not exceed the amounts .
indicated.
2. In the event that integrity of any pressure isolation valve
-
specified in Table 4.4-3 cannot be demonstrated, reactor
operation may continue, provided that at least two
valves in each high pressure line having a non-functional
valve are in,and regafn in, the mode corresponding to the
isolated condition.iaJ
\. 3. 'If Specification 1 and 2 cannot be met, an orderly shutdown
shall be initiated and the reactor shall be in the cold
shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4 The provision of specification 4.0.4 is not applicable for
entry into Mode 3 or 4
.
.,._. n
i
I
'! (8) Motor operated valves shall be placed in tne closed ,,atition d ;c.-,c-
j supplies deenergized.
1
i
.i
BEAVER VALLEY - UNIT 1 3/4 4-14a Order dated April 20,1981
-
l
.
R_E. ACTOR COOLANT SYSTEMS s.
SURVEILLANCE REQUIREMENT
. .
. . . .
_.
(8) on each valve listed in Table
4.4.6.3.1 Pe i test
4.4-3diesha:faktiaaccompHih.dpriorto.nteringmod.1afterevery l '
time the plant is placed in the cold shutdown condition for
refueling, after each time the plant is placed in a cold shutdown
condition for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if testing has not been accomplished in
the proceeding 9 months and prior to returning the valve to
service after maintenance, repair or replacement work is
performed.
4.4.6.3.2 Whenever integrity of a pressure isolation valve listed in Table
4.4-3 cannot be demonstrated the integrity of the remaining valve
in each high pressure line having a leaking valve shall be
determined and recorded daily. In addition, the position of the
other closed valve located in the high pressure piping shall be
recorded daily.
l') To satisfy ALARA requirements, leakage may be measured indirectly (as
from the performance of pressure indicators) if accomplished in i
accordance with approved procedures and suoported by computations
showing that the method is capable of demonstrating valve compliance
'
with the leakage criteria.
BEAVER VALLEY - UNIT 1 3/4 4-14b Of/df/d W d/Mfl7/ W / W 7 '
AMENDMENT NO. 101
__
- - _ - _ - _ _ _ _
.
[
TABLE 4.4-3
- REACTOR COOLANT SYSTEM PRES $URE ISOLATION VALVES _
- . . . - Maxhum(a) { b)
System Valve No. Allowable Leakaoc
Loop 1, cold leg SI-23 < 5.0 GPM
SI-12 E 5.0 GPM
-
Loop 2, cold leg SI-24 < 5.0 GPM
SI-11 '"~'{5.0GPM
Loop 3, cold leg SI-25 ~< 5 0 GPM
SI-10 55.0GPM
.
4
(a) 1. Leakage rates less than or equal to 1.0 gpm are considered acceptable.
2. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm
,
are considered acceptable if the latest measured rate has r. t exceede
i
j
the rate determined by the previous test by an amount that reduces
!
the margin between measured leakage rate and the maximum pemissible
rate of 5.0 gpn by 50% or greater.
l
I
- 3. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm
are considered unacceptable if the latest measured rate exceeded the
l rate determined by the previous test by an amount that reduces the
margin between measured leakage rate and the maximum permissible rate
of 5.0 gpm by 50t oc greater,
i 4 Leakage rates greater than 5'.0 gpm are considered unacceptable.
(b) Minimum test differential pressure shall not be less that 150 psid.
i
4
3/4 4-14e Order dated April 20, 1986
. - _
- 14
-L: ' THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE
THERdQDYNAMICS
ANSWERS -- BEAVER VALLEY 1&2 -87/05/20-DUDLEY, N. m.,
ANSWER 5.01 (2.50)
a) SUR increases [0.4] due to the added reactivity from the primary
cooldown when the steam dumps open. [0.6]
'
b) SUR will be zero. [0.4] The positive SUR will cause power to increase
until the negative reactivity effects of the power coefficient make the
net reactivity of the core equal to zerto. [0.6]
c) The longer lived delayed neutrons groups. [0.5]
REFERENCE
BVPS Reactor Theory Manual Chapter 5, pages 9-12,26-30;
Chapter 6, pages 45-53.
learning objectives 5-1/4;6.1/2,4
192003K106 192003K107 ...(KA'S)
ANSWER 5.02 ( .50)
'
4F' [0. 5]
A
REFERENCE
BVPS Thermodynamics Manual Chapter 4, pages 31-33. i
191004K109 ...(KA'S)
,
ANSWER 5.03 (2.00)
l
l a) EOL [0.25]
The negative power coefficient of reactivity tends to dampen the
oscillations. [0.5] This coefficient is more negative at EOL. [0.25]
b) 100% power [0.25]
l The higher neutron flux at 100% power can make more xenon faster. [0.5]
Rapid increases in xenon concentration increase the magnitude and
frequency of the oscillations. [0.25]
(small effects of temperature due to Tavg program)
i
REFERENCE
BVPS Reactor Theory Manual Chapter 6, page 51; Chapter 7, page 17.
learning objectives 6.1/5;7-1/6
001050A206 192006K106 ...(KA'S)
1
l
l
i.
l
l
t
_ - _ _ _ _ - _ .
5. THEORY OF NUCLEAB POWER PLANT OPERATION. FLUIDS. AND PAGE 15 l
THERMODYNAMICS l
l
ANSWERS -- BEAVER VALLEY 1&2 -87/05/20-DUDLEY, N. ,,..
1
l
ANSWER 5.04 (2.00)
1) power [0.25] decrease [0.25]
2) RCS flow [0.25] increase [0.25]
3) RCS pressure [0.25] increase [0.25]
4) RCS temperature [0.25] decrease [0.25]
REFERENCE
BVPS Thermodynamics Manual Chapter 7, page 17.
002000A103 002000A105 002000A106 193008K105 ...(KA'S)
.
i
ANSWER 5.05 (2.25)
a. P sat = 1535.0 + 15.0 = 1550.0 psia [0.25]
T sat = 600.6 F [0.25]
subcooling margin = 600.6 - 540.0 = 60.6 F [0.25]
b. Core boiling at DNB [0.25]t o , .f f
or Steam binding of the coolant loops may occur, [G-Ge] ,, y
which will cause flow to be reduced or cease completely. [W)
c. Almost no flow travels through the vessel head region. [0.3] The
vessel head cools down through ambient heat losses only. [0.45]
REFERENCE
BVPS Thermodynamics Manual Chapter 7, pages 20-26.
000017K101 000074A201 193001K101 193008K115 193008K125
...(KA'S)
_ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ ____ _ _____-
- - _ _ _ _ _ - _
5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 16
THERMODYNAMICS
ANSWERS -- BEAVER VALLEY 1&2 -87/05/20-DUDLEY, N. ,, ; .
ANSWER .5.06 (3.00)
rhoi = (Keffl - 1)/ Keffi = -0.05
Keffl = 1/((1'- (-0.05)) = 0.9524 [0.5] ;
120(1.- 0.9524) =.196(1 Keff2)
Keff2 = 0.9709 [0.5]
rho 2 :-(0.9709 -1)/0.9709 = -0.03 [0.25] j
delta rho = rho 2 - rhol = -0.03 - (-0.05) = 0.02 = 2000 pcm [0.5]-
1000 pcm is due.to xenon, so the remaining 1000 pcm is due to boron [0.5] l
change in1 boron concentration for 1000 pcm is-
- 1000 pcm / 10 pcm/ ppm =-100 ppm [0.25]
new boron concentration = 1200 - 100 = 1100 ppm [0.5]
REFERENCE
BVPS Reactor Theory Manual Chapter 5, page 49; Chapter'9, pages 2-9.
learning objectives 5-1/13;7-1/3
'001010K524 004000A404 ...(KA'S)
ANSWER .5.07 (2.75)
a. using Q = U A (Tclad - Tcoolant) [0.15]
boron precipitation causes U to decrease [0.5]
assume Q constant
then (Tclad - Tcoolant) increases [0.2]
,
assume Tcoolant constant
then Tclad increases [0.25]
.b. using Q = U A (Tavg - Tsat) [0.15]
tube plugging causes A to decrease [0.5]
using given conditions:
Tsat decreases [0.2]
if Tsat decreases then Psat decreases [0.2]
using Q = m (4h) steam [0.15]
if Psat decreases then using the given conditions:
(oh) steam decreases [0.2]
if (4h) steam decreases then Q and MWe decreaser, [0.25]
REFERENCE
BVPS Thermodynamics Manual Chapter 7, pages 1-3;
Chapter 3, pages 3-9. ,
- _ - - - _ _ _ _ _ _ _ ___
5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 17
THERMODYNAMICS
ANSWERS -- BEAVER VALLEY 1&2 -87/05/20-DUDLEY, N. ,,,,
Various Topics LP-LRT-VIII-85; Enabling Obj. 3
002020K501 193001K101 193003K125 .. (KA'S)
'
l
4
l
-
e
o
"-- - - - - - - . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ , _ _ _ _ _
- _ - . - _ __
, ,
6. PLANT SYSIEMS_ DESIGN, CONTROL. AND INSTRUMENTATION PAGE 18
ANSWERS -- BEAVER VALLEY 1&2 -87/05/20-DUDLEY, N.
i
w.
I
'
ANSWER 6.01 0FLErf 60)
B. Pull fuses [0.6] l
i
REFERENCE
Excore NI System, NS-8, p 28-30
)'
012000K406 015000A403 ...(KA'S)
1
-
,
ANSWER 6.02 ( .60)
B. Hot leg RTD fai16d low [0.8)
REFERENCE
'
b0 20 $. KA'S)
ANSWER 6.03 (1.00) ,
i
' place N-35 level trip switch in bypass [1.0)
REFERENCE
BVPS OM - Chapter 2, pages 16,17
012000A403 ...(KA'S)
.
ANSWER 6.04 (1.80)
'
a. Fail green lightA 4-10
og MVJ/CZAT04 is ,p/ADxtinguished
,mcs's M A%'N for that
w &'. monitop
Y FA IwCf * [0.6]
b. Closes SLCRS filter bypass damper
Open main filter bank inlet damper
Actuates fuel bldg evacuation alarm [0.4 pts each)
REFERENCE
' Module 4, RMS review, LP-LRT-V-56 pgs 4,5,25,27; EO-1,2,3,4
Components: 191002 Sensors & Detectors K 1.18 2.6
07200K201 072000K401 ...(KA'S)
. . . . . . . .
_ _ _ _ _ _ - _ _ _ _ _ _ -
6. PLANT SYSIEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE- 19 l
ANSWERS -- BEAVER VALLEY.1&2 -87/05/20-DUDLEY, N.
v,
1
-ANSWER 6.05 (1.50)
.o. AM-CH-122 in manual
'
[0.25] ;
Master Controller in-auto-hold [0.25] i
b. A + B available [0.2] : A auto / manual [0.1]
B manual [0.1]
'
C auto hold [0.1]
c. CCR to RWST refrigeration unit isolates [0.1] !
Bit recirculation isolation valve shuts [0.2]
Nitrogen supply to SI accumulators fail shut (0.2]
REFERENCE
Electrical Distribution Review LP-LRT-V-59, p 21, 31; EO 4, 7
p 000057A219 000058A203 ...(KA'S)
ANSWER 6.06 (2.50)
a. cut-back control valves close (MOV-1QS-103 A,B) [0.6]
prevents containment pressure from becoming excessively negative (0.7]
b. improves the removal of radioactive iodine from the containment
atmosphere 3 [0.6]
OR R A tsts Sune PH
c. provides' recirculation pumps with adequate NPSH [0.6]
REFERENCE
BVPS Training Systems Descriptions AS-16, pages 3-5
BVPS OM - Chapter 13, 1.13.1, page 4
026000A301 026020K401 026020K402 ...(KA'S)
i
ANSWER 6.07 (1.50)
e. regulation of CVCS letdown flow (PCV-CH-145) [0.5] ,
!
b. reactor vessel brittle fracture protection (NDT) [0.5] !
c. yes. [0.5] i
REFERENCE
Chapter 10, page 5
BVPS Training Systems Descriptions NS-14, pages 13-15
_______________________o
_
i
l
D
6. PLANT' SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 20
ANSWERS -- BEAVER VALLEY 1&2 -87/05/20-DUDLEY, N. j
,: .
i
005000K104 005000K401- 005000K501 ...(KA'S)
' ANSWER .6.08 (2.50)
a. -Decrease
-h. Increase
c. Opem CLOSED
d. Remain the same
e. Remain the same (0.5 pts each)
REFERENCE
Module 3, Rx plant component and Neutron Shield Tank Cooling System,
LP-LRT-V-53, pgs 8-13; EO-4
3.10 008 000 K 1.02 3.3
008000K102 ...(KA'S)
-ANSWER 6.09 (3.00)
a. High [0.3] causes a steam flow error which opens FWRV [0.3]
b. High [0.3] causes a steam flow error which opens FWRV [0.3]
c. Low [0.3) causes a feed flow error which opens FWRV [0.3]
d. Low [0.3] causes a level error which opens FWRV [0.3] ,
e. High [0.3] causes a level error due to change in level program [0.3] !
(no change [0.3] if program level is assumed to be at maximum level
prior to instrument ~ failure [0.3])
h
REFERENCE i
Sream Generator Feedwater System PGS-10, p PGS-10-39 to PGS-10-41; EO 2
1987 SRO Annual Licensing Examination, Session 3, #6-3
059000A211 ...(KA'S)
.
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
t
7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 21
RADIOLOGICAL CONTROL l
ANSWERS -- BEAVER VALLEY 1&2 -87/05/20-DUDLEY, N. ,, ; .
i
!
ANSWER 7.01 ( 60)
D.-Number.of CRDM fans [0.6)
' REFERENCE
ES-0.2 step 12.b
Natural Circulation Cooldown, LP-LRT-VII-94, p 12; EO 3
000009K326 000056K302 000074A201 ...(KA'S)
!
ANSWER 7.02 ( .60)
.
D. Rx power > 15% [0.'6]
REFERENCE
OH 1.1.4, p 11
RCS NS-10-3, Obj. 1
001010A401 ...(KA'S)
ANSWER 7.03 ( .60)
,
C. Rx Shutdown.[0.6]
REFERENCE
AOP-25 issue 3/rev 0, p 2
RP River Water System LP-LRT-V-52, Enabling Obj. 11
075000A202 ...(KA'S)
.
ANSWER 7.04 ( .60)
DEL E TE
B. HHSI and CCR pump flow [0.5)
REFERENCE
E-3 p. 2,3
Operator Response to SGTR LP-LRT-VII-87, Enabling Ojj. 8
000074K304 ...(KA'S)
______-
_ _ _ _ - -
7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 22
RADIOLOGICAL CONTROL
'
' ANSWERS -- BEAVER VALLEY 1&2 -87/05/20-DUDLEY, N. ,,,..
ANSWER 7.05 ( .60)
.A. Rods ~in manual and drive into core (0.5]
REFERENCE
kb8:
'
O ...(KA'S) l
ANSWER 7.06 -
(1.40)
Normal Spray [0.4]
PORV [0.4]
Auxiliary Spray [0.4]
Correct order [0 7]
1
REFERENCE
Operator Response to SGTR LP-LRT-VII-87; Enabling Obj. 1; p 8, 9
000038K306 ...(KA'S)
,
' ANSWER 7.O'7 (1.50)
I D(squared) = 1 d(squared) [0.3]
i = (2500 mrem /hr)(2.25 ft sq)/25 ft sq [0.3]
1: 225 mrem /hr [0.15]
100 mrem /hr < 225 mrem /hr < 1000 mrem /hr
The area should be posted as an High Radiation Area [0.75]
REFERENCE
10CFR20,205
BVPS-RCH pgs 5-7
194001K103 ...(KA'S)
_ - - _ - _ _ _ _ _ _ _ _ _ __ - _______ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
-__ -
_ _ _ _
7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 23
RADIOLOGICAL CONTROL
ANSWERS'-- BEAVER VALLEY 1&2 -87/05/20-DUDLEY, N. m
,
ANSWER 7.08 (1.60) '
a.-A' confirmatory sample should be analyzed to extend the effective
period ~of authorization (since its 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> limit was exceeded). [1.0]
c. Every 15 minutes-log the readings from the radiation monitor [0.6]
REFERENCE
Module 4, Liquid Radwaste System Review, LP-LRT-V-54, pg 15, EO-6
068 System generic 1 2.7
068000K401 ...(KA'S) .
i
' ANSWER 7.09 (2.50)
a. Sleed and f::d: Manually initiate HP SI, manuually opens the I
PORVs [0. 7] .
Feed and bleed: Manually initist: "P OI, pcruits the autca.atic
cycling of the PORY: te vent RCS invent ry [0.?]
b. Feed only'one SG until Thot < 550 F [0.4] then feed all SGs [0.3]
e. 2
c. > 5% NR level in at least one SG [Bv4*]
3 60 y A FW FLO W [0. 2] ;
REFERENG '
Module 2 , Oper-Response to LOSHS, LP-LRT-VII-81, pgs 1,6,7; EO-1,2,4
000054K304 ...(KA'S) j
ANSWER 7.10 (2.00)
a. No [0.4]
b. PORV or Safety Valve is open [0.6] l
I
c. Stop bistable tripping from first channel failure [0.6] l
and be in hot standby in one hour [0.4]
REFERENCE
AOP-13, Malfunction of PZR press control, pgs 2,5,4
3.3 010 system generic 1 3.5
1
. . . - . . - . . . . - . . . . . . .. . . - - - - - . - ---------------------------Q
.. - - . _
- _ _ _ _ . _ _ _ - _ . - -
'
7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 24
RADIOLOGICAL CONTROL
ANSWERS -- BEAVER VALLEY 1&2 -87/05/20-DUDLEY, N. ,,,,
000008A212 ...(KA'S)
ANSWER 7.11 (3.00)
a. 1. Ensure continued secondary reat removal capability. [0.6)
2. Ensure RCS subcooling stable of increasing [0.33
and SI flow is effective in increasing RCS inventory. [0.3]
3. Indicates sufficient RCS inventory if there is' verified hot leg
or core exit subsooling present. [0.6)
b. RCS subcooling less than the value obtained from the attachment. [0.6)
Pressurizer level cannot be maintained above 5% [0.6)
REFERENCE i
EOP Executive Volume 1.53b.4, p 39 of 57
.1987 SRO Annual License Examination, Session 3, Question 7-3
L 006050S007 006050S010 ...(KA'S)
l
?
1
C _ _ _ _ ___
_ _ _ _--
1.
8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 25
ANSWERS --~ BEAVER VALLEY 1&2 -87/05/20-DUDLEY, N.
- .
' ANSWER '8.01 ( .60)
C [0.6)
. REFERENCE
.TS p. 3/4 2-1 to 2-4
Contorl Rod Misoperation, IE Information Notice 86-07; Enabling Obj. 3
001050A206 ...(KA'S)-
l
ANSWER 8.02 ( .60)
A. immovable Rod [0.6]
REFERENCE ,
TS p 3/4 1-7, 4-5, and 1-18
.001050G005' ...(KA'S)
ANSWER 8.03- ( .60)
B [0.'6]
REFERENCE
'TS'p 3/4 2-5
001000K553- ...(KA'S)
ANSWER 8.04 ( .60)
C. Two RCP trip [0.6)
REFERENCE
TS p:3/4 1-1, 1-9, and 4-2b
000074G005. ...(KA'S)
8. ADMINISTRATIVE PROCEDURES. CONDITIONS._AND LIMITATIONS PAGE 26
ANSWERS -- BEAVER VALLEY 1&2 -87/05/20-DUDLEY, N.
- ,
ANSWER 8.05 (1.50)
l
a. Check the Lead Seal Log [0.1]
Check valve in proper position [0.4]
Have new lead seal installed [0.25]
b. Effective if Order is numbered (year and sequential number) [0.75]
REFERENCE
' Administrative Review; LP-LRT-VI-39; Terminal Objective
194001A103 ...(KA'S)
ANSWER 8.06 (1.00)
For initial assessment of non-routine work [0.5]
All times unless more critical work requires his presence. [0.5]
REFERENCE
Standing Night Orders, Containment Entries-Supervisor Involvement, from
R. Druga, dated January 8, 1987
1987 SRO Annual License Examination, Session 3; Question 8-7
194001K114 ...(KA'S)
ANSWER 8.07 (2.00)
Primary to Secondary leakage (576 gallon / day on SG #1) [1.0]
Total unidentified leakage (1.2 gpm unidentified) [1.0]
REFERENCE
002020G008 002020S005 ...(KA'S)
ANSWER 8.08 (2.10)
f " 2 Da
a. One hour [0.3]"due to loss of emergency assessment capability [0.4]
b. No notification [0.3] due to planned testing of ESF system [0.4]
c. No notification [0.3] since release was less than 2 times MPC averaged
over one hour. [0.4]
_ __ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _- _____
- - _ - _ _ _ _ - . _ ._.
i
8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 27 I
ANSWERS -- BEAVER VALLEY 1&2 -87/05/20-DUDLEY, N.
s. 1
REFERENCE
Reporting Requirements LP-LRT-VI-41; Enabling Obj. 1 and 2
_
Liquid Radwaste System Review LP-LRT-V-54; Enabling Obj. 9
068000A404 068000G008 194001A116 ...(KA'S)
ANSWER 8.09 (3.00)
a. No [0.5] not all necessary attendant auxiliary components related for
the system to perform its function are available. [0.5]
.b. No [0.5] pump is not capable of performing its specified function.[0.5)
c. Yes [0.5] its normal power supply and redundant systems are
operable. [0.5]
REFERENCE
Technical Specifications p 3/4 0-1
006050G005 062000G005 063050G008 064000K105 ...(KA'S)
ANSWER 8.10 (3.00)
a. Emergency procedure immediate action steps [0.5)
Routine procedures that are frequently repeated [0.5]
b. When no action consistent with the procedures provides equivalent
protection and is immediately apparent. [p* ]
PuoLic HCAL T H AND SAFETY EC.f]
c. Action le necessary te prevent pereennel injurr [0.5]_er equirrent
dart;e. [0.5]
udpeg f/C C T4 L u M !.or 8 & C E S
REFERENCE
BV-1, Station Admin Proc, ch. 4, pg 40, 41
Team Work and Diagnostic Skills Retraining LP-TWD-2; Enabling
Objective 1
194001A103 194001A111 ...(KA'S)
_ _ _ _ - _ _ _ _ - - _
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TEST CROSS REFERENCE PAGE 1
' QUESTION VALUE REFERENCE
________ ______ __________ <
05.01 2.50 DUDOOOl636
05.02 .50 DUDOOOl637
05.03 2.00 DUDOOOl638
05.04 2.00 DUDOOOl639
05.05 2.25 DUDOOOl64
05.06 3.00 DUDOOO164
05.07 2.75 DUDOOO1642
______
15.00
06.01 .60 DUDOOOl601
06.02 .60 DUDOOO1606
06.03 1.00 DUDOOOl645
06.04 1.80 DUDOOOl652
06.05 1.50 DUDOOOl661
06.06 2.50 DUDOOOl644
06.07 1.50 DUDOOO1646
06.08 2.50 DUDOOO1653
06.09 3.00 DUDOOOl654
______
15.00
07.01 .60 DUDOOOl600 1
07.02 .60 DUDOOOl602 l
07.03 .60 DUDOOOl603 l
07.04 .60 DUDOOO1604
07.05 .60 DUDOOOl607
07.06 1.40 DUDOOO1647
07.07 1.50 DUDOOO1648
07.08 1.60 DUDOOOl649
07.09 2.50 DUDOOOl650
07.10 2.00 DUDOOOl651
07.11 3.00 DUDOOOl655
______
14.00
08.01 .60 DUDOOO1595
08.02 .60 DUDOOO1596
08.03 .60 DUDOOO1597
08.04 .60 DUDOOO1598
08.05 1.50 DUDOOO1593
08.06 1.00 DUDOOO1658
08.07 2.00 DUDOOO1592
08.08 2.10 DUDOOO1594
08.09 3.00 DUDOOO1589
08.10 3.00 DUDOOO1591
_ _ _ _ - . .
14.00
______
__h_p*_
60.00
-
_-_
.1dV4 .
'Af .
Telephon 412) 393 6000
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Nuclear Group
P.O. Box 4
shippingport, PA 15077-0004
May 22, 1987 ,
ND3VPN: 5024
Robert M. Keller, Chief
Section 1C (Operator Licensing)
Division of Reactor Projects
U.S. Nuclear Regulatory Commission
Region 1
631 Park Avenue
King of Prussia, PA 19406
Reference: Beaver Valley Power Stations, Unit #1 and #2
Docket Numbers 50-334, 50-412
Operator Written Examination Report
Dear Mr. Keller:
Please find enclosed comments generated by the Training Section
concerning the R.0, and S.R.0. examinations administered May 19 and
20. 1987 at our facility.
As you review the specific exam comments, you will see that certain
generic concerns are evident on our part. We would like to resolve these
issues as they challenge our candidates unfairly. These concerns cause
excessive amounts of time to be used and in some cases adds unneeded
confusion in an already stressful situation. In addition, our candidates
are being tested extensively in areas that are not required knowledge from
memory. The specific areas of concern are:
. Multiple choice questions that do not have any correct answers to
choose from.
. Questions were asked of candidates concerning steps frcm normal and
abnormal procedures that are required to be present and followed.
These steps are not immediate manual actions and are not required to
be reproduced from memory. It should be remembered that operators
have a very large number of approved procedures which they are
required to use when operating a licensed facility. We train our
personnel on the content, interpretation, and proper use of these
procedures and to have these procedures present when they are
performing their operational duties. It is not expected that they
would perform these large numbers of operations without the use of
the latest revisions of their controlled operating copies.
. Questions appeared on the four exams that were the same but had
different answers depending on which exam was being reviewed.
. . ..
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Operator Writterc Examination Report
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May 22, 1987:
m Page 2 , ,,7 _,
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.. Requalification' Exam questions derived from' enabling objectives that- -1
are more specific than written and-do not take into consideration the
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terminal objective that sets the conditions and-standards for
. performance ~of the enabling-objective.
. A requalification exam question:was asked that indicated a calculated-
value was required,-when in fact it was not possible to, calculate a
specific value- and was 'really not . required by the key. This imposed.
a time consideration on our candidates, i.e.. time that could'have
been spent working on the remainder of the exam questions.
.'Our intent is-to work with the NRC in improving the examination'
process. Therefore, we respectfully request that you place sufficient
_
emphasis on these comments and the comments attached to insure that our
candidates have a fair-and equitable outcome of these examinations.
Very truly yours,
J.-D. Sieber
Vice President Nuclear
cc: .J. O. Crockett w/o attachment
"-
T. W. Burns ."
a
W. S. Lacey "
Central File (2) "' "
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l Unit 1 Senior Reactor Operator Written
Recualification Examination Cuwrits
Administered 5/20/87 ,
CUESTION OmMDFF
5.Olb 'Ihe question asked for a calculation of what SUR would be 5
minutes after a particular transient. 'Jhis is not possible to
calculate since SUR would be continuously changing due to
doppler faa4hek. An explanation that SUR would be less than
the initial SUR should be acceptable for full credit.
5.02 'Ibe answer should be "A. a higher head and a higher flow
rate."
ty sks,c wc l
o . . . ._
E ~ ~ ~ '
'
> R,1l : Pm?
,
,- ',
4
f lou ?
Also, see question 1.04 en the RO examination.
'
5.05a should be used to calculate subcooling margin since it
is highest temperature. h answer should be 50.6*F.
5.05b Alternate reasonable answers should be acceptable since a i
siWling margin of less than 20'F does not necessarily mean '
there is boiling in the core. As long as there is 91hling,
natural circulation flow should be unaffected.
6.01 None of the choices for this multiple choice question are
correct. h answer given on the key is not correct since
only the control power fuses are rumoved as per AOP-10, l
" Malfunction of Nuclear Instrumentation." The question should
be deleted.
)
6.04a Annunciator A4-70, " RADIATION HolmOR KMER SUPPLY FAIIURE"
will also alarm upon power supply failure of a radiation
ronitor. See the attachment. '1his should also be an
acceptable answer.
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Unit 1 Senior Reactor Operator Written
Reaualification Examination Cum s tus
Administered 5/20/87 *-
(continued)
i
OUESTION OCNMENT
6.06b '1here is another reason for adding NaGI to the quench spray
system. Na0R will also raise sump pH to hWsate for the
boric acid aMM by Safety Injection. 'Ihis should be added to l
the key. See the attachment.
l
7.04 None of the choices for this multiple choice question are
correct. See the attachment for tte correct RCP trip
critaria- 'Ihis question should be deleted.
7.09 'Ihis question is the same as question 4.07 on the RO
examination. Refer to the orament for question 4.07.
P
8.01 No action statement addresses the situation presented in the
question. 'Iherefore, this is a judgement call. A case can be
made for ej thar answer C or D. If the candidate reasons that
sinco the sitation is not acktressed in the Tech Spec action
statements that Applicability Statement 3.03 applies, then he
would choc6e answer D. If he reasons that since the 100 is
presently met, he does not need to refer to the action state-
ments; but, the intent of the Tech Spec is violated, then he
would cficose the appropriate action basM on the basis for the
60 penalty minute restriction. In this case, he would choose
answer C. If this situation actually comrred, the NSS would
be able to discuss the situation with plant management,
including the Licensirg Department and arrive at an inted
pretaticn of the Tech Spec. Since this was not available to
the candidates, a.nd no justification is asked for by the
questicn, either answer C or D should be acceptable for full
credit.
8.08a Since the question does not specify which Vital Bus is de-
energized, the answer oculd be diffetmt depending on the
justification. 'Ihe cany Vital Bus that could cause the
condition on the armv.r key would be #2 sirm all Control Room
annunciators would be lost. For the other Vital Busses, the
answer would be "no notification".
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Unit 1 Senior Reactor Operator Written
- Pw==1ificatim Pramb ation
i G------;ta -
Administered 5/20/87 $-
(ocntinued).
..
QUESTION CQ9ENT
8.10b&c' 'Ihe words, "to protect the health and safety.of the public",
should be added to the key for part b. Part c should be-
deleted since there'is no specific instance in which.this
would be allowed.. See the attachments. 'Diis question-
appeared on an examination given at Beaver valley in' July,
1986. It was deleted from that examination during the exam
. review.
,,
"
L6.05a,b,c 'Ibese questions require that parts of W - twes that are
7.02 required to be sos =A. and followed when performing an
- .7.03 : operation be last W from memory. Operators.at Beaver
7.08b'- Valley are Det required to memorize these re ares. ' Relying
7.09b on nuannTy leads 'to mistakes. Since this is not required
7.10a- knowledge,: these questions should be deleted.
8.05a
18.08'
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Unit 1 Reactor Operator Written
Poaualification Examination Cwmmius
Administered 5/20/87 '
QUJSTION N
1.07e CharxJe to decrease for unaffected loop cold leg temperature
based on simulator response observed by examiners.
1.08b Change to D N decreases as moderator temperature increases
because thermal neutron energy level will increase. Since
boron is a 1/V absorber, this will cause the absorption cross
section to decrease. B erefore, D N will decrease.
1.08d Change to DN will increase as core ages because the coolant
systnm boron cuw/w& tion will decrease. B erefore, less
ocupetition arri DN will increase.
2.01 Change to "Ioss of RCP seal iniectioD could cause. . ."
2.03 We received no answer for question 2.03. We are supplying the
correct answer.
2.05b me word " limit" in this part of the question is not defined,
therefore, the candidate could interpret this a mber of
different ways. If the limit is taken to mean 'lechnical
Specification limits, any number of answers are possible.
'Ibch Spec 2.11.2.1, 2, 3, 4, and 5 apply to Radiation Waste
Discharge Authorization - Gaseous (NDA-G) . Any gas addressed
on this permit should be an acceptable answer since these
gases are processed in our gaseces waste disposal system. me
operator utilizes this permit to make a gaseous discharge.
2.05c 2e answer key should be changtd from "no release" to "yes a
release will occur". Se attached drawing will show the path
for the release.
3.01 Since the question dcas not specify which impulse pressure
channel fails, three potential failures are possible:
1. Impulso chanNd to Hic fails. Dic will shift to imp
out mode. No system transient.
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Unit J Reactor Operator Written
Recualification Wa=4 nation Otzments
x
Administered 5/20/87 1
(continued) I
!
OUESTION OCMENT l
3.01 2. Inpulse pressure channel not selected fails. No
(continued) system transient. '
3. . Inpulse pressure c:hannel selected fails. Rods insert
due to turbine loa 4/ reactor power difference and
Tave/ Tref mismatch. .RCs cooldown causes inventory to !
shrink. PZR level decramaan. PZR pressure de- l
cramaan. Reactor trip is possible on low pzr l
pressure.
'l
HggE: 'A reactor trip may not occur since this transient
will result'in a 15'F decrease Tave and approximately.
a 150 to 200 psi decrease in'pzr pressure. %e '
reactor trip setpoint of 1945 psig.will probably not
be reached. Since this is a rate sensitive trip, it
may occur but student does not have enough
information to determine if it will.
Request that all three answers be acceptable.
3.03a % e answer key should also include the alarm on annunciator
window A4-70, " Rad Manitor Power Supply Failure". S e alarm ;
response is attached. During the Exam Review Meetity, ANN ;
window A4-72 was reported not A4-70.
3.04c Request this part be deleted. Se response of the TCV will
depend on the dcminant effect. As system flow decramaan, the
Hx outlet terqperature will decrease which will open the valve,
but the rMad flow will also cause return water temperature
to increase. This will cause the TCV to close. Since the
valve may open or close impending on actual system resgs,e, !
we request this part be deleted.
3.05 Request the question be deleted. W e answer to part A is
found in a table in IRT module which is not required to be
menorized. S e answer to part B is found in follow-up actions
to AOP-38, Extended Loss of 125 VDC Switchboard #2.
3.06a Request the fact the cutback valves are normally open be
considered in gradirg.
b; _ - - _ _ _ _ _ _ _ _ - _ ___ -
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Unit 1 Reactor Operator Written
Reaualification Examination C-+rds
Administered 5/20/87 m
(continued)
GJESTLON CGem
4.01 Accept as an alternate answer 1.25 RenVO.8 RenVhr = 1.56 hours6.481481e-4 days <br />0.0156 hours <br />9.259259e-5 weeks <br />2.1308e-5 months <br />
if the student assumes NRC Form 4 is not on file. The i
question does not specify. !
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4.02a Question dices not specify if the flowrate which decreases is
liquid waste dischanye flow or cooling tower blowdown flow.
Request the following answers be acceptable.
1. If blowdown flow a=mai, answer is correct.
2. If liquid waste M=hme flow a===4 no action
required.
4.03a Request the last sentence in answer, "Thus preventing
reduction. . ." be deleted. The first two lines are
sufficient to answer the question.
4.03b- Request the answer be changed to ' Mater level in affected SG
will stabilize." Pzr level respc a will also depend on
charging / letdown relationship.
4.03c Request this part be deleted since there are 12 entry
conditi'ns, aid this is not required operator knowledge.
4.05a Request this part be deleted since this is a note in the
AOF29 which is not required to be memorized by operators.
4.05b Request answer be change to, "Since many control systems are
affected, it is advantageous to restore power as rapidly as
possible to regain complete control of the plant." The answer
given is not correct. The FRV for A SG will go to auto-hold
and B/C SG will stay in auto.
4.07a Feed and bleed is the initiation of HHSI and then opening the
MRVs. Bleed and feed is the cpening of the MRVs and then
initiation of HHSI. This distinction is not relevant since ;
FR.H-1 only uses feed and bleed initiated by the operator. '
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Unit 1 Reactor Operator Written
Recualification Dcaminatio_r Twme.:nts
Administered 5/20/87 m
(continued)
GFSTION 0: MBT
4.07c Since the question does not specify which gMare to use,
re.yest the followirg additional answer be acceptable for full
cra. lit: ">350 gpm". 'Ihe red path for heat sink recpires < 5%
NR level An3 < 350 9pn AEW. 'Iherefore, the absence of either
one implies a heat sink exists.
3.05b 'Ihese questions require that parts of gu-Mnes that are i
3.06c. required to be present and followed when performixg an
4.02c oper 6.icn be reproduced fran memory. Operators at Beaver
4.03c. Vall., are Dot required to memorize these gMares. Relying
4.04a on memory leads to mistakes. Since this is not required
4.05a,c,d knowleckye, these questions should be deleted.
! 4.07b
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