ML20126B695
| ML20126B695 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 06/06/1985 |
| From: | Dimmock L, Mcmillen J, Sly G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20126B683 | List: |
| References | |
| 50-263-OL-85-01, 50-263-OL-85-1, NUDOCS 8506140185 | |
| Download: ML20126B695 (60) | |
See also: IR 05000429/2005003
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
ReportNo(s). 50-263/0L85-01
DocketNo(s).50-263 License No. DPR-22
Licensee: Northern States Power Company
414 Nicollet
Minneapolis, MN 55401
Facility Nane: Monticello Nuclear Generating Plant
Examination Administered At: Monticello
Examination Conducted: April 29 through May 3, 1985
Examiner (s):
/Did
L. Dinunock f/4/rf
Date
)<W
G. 51y f/C/W
Dit6
Approved By: . I' . M hief 6/4/ff
Date'
Operator Licensing Section
Examination Suninary
Examination administered on April 29 through May 3, 1985 (Report No. 50-263/0L85-01)
Seven candidates took the examinations for an SRO license. One candidate took
the examinations for an instructor certification with an R0 license.
Results: Six candidates successfully completed the examinations.
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$N63 PDR
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REPORT DETAILS
1. Examiners
L. Dimmock, Region:III
G. Sly, PNL
2. Examination Review Meeting
An examination was held insnediately following the written exam on April 30,
1985. The following utility representatives were in attendance:
Eugene Earney, Training Superintendent
Douglas Antony, Superintendent Operations
Mike Perry, Operator Instructor
Bob McGillic, Operator Training
Additionally, the following NRC representatives were present:
L. Dimmock NRC - Region III
G. Sly PNL
The comments raised by the utility during the exam review and their
resolution are attached.
3. Exit Meeting ,
An exit meeting was held at the facility the evening of May 2, 1985. The
utility representatives were as follows:
Bob McGillic
Eugene Earney
The NRC representative was as follows:
L. Dimmock NRC-Region III
G. Sly PNL
The utility was informed that six of the eight candidates were clear
passes on the oral /simulctor exams.
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FACILITY COMMENTS
QUESTION 5.2.c
Coment: Response may be based on the Tech. Spec. requirement of must be
within 2%.
Response: Not accepted, No indication to adjustments of APRM vs. heat balance
- as to AGAFs only frequency (once every 3 days) was found in T.S.
(T.S. Table 4.1.2, pg. 34) note #4.
QUESTION 5.3.b
Coment: Actual level may change due to pressure increase. Indicated level
will not change because Rx level control will maintain level at
setpoint (constant).
Response: Coment accepted, answer key has been changed to read "not change,
because Rx level control will maintain level at setpoint
(constant).
QUESTION 5.16.c
Coment: May receive response stated Keff * 1*
Response: Partial credit equal to 0.25 will be given for: Keff * I '
The reason for only partial credit is that the equations do not
completely define or explain the terms.
QUESTION 6.1.a. b.
Coment: a. If new initiation signal is from 2# drywell. Will not restart
until low low water level, if new signal is from low low water
will restart.
b. Will not restart until low low.
Response: a. Coment not accepted. With reference material provided there
is no indication that the initiating events have any effect on
a mechanical overspeed trip.
b. Coment accepted. Answer key has been changed to read:
"will," if initiating event is low low water or "will not," if
initiating event is high drywell." If no reason is given
"will not" is the correct answer because the use of the words
" level recovery" implies that level is normal.
Reference: Design Change 82M089, Rev. 4, Appendix lib /c, Figure 1.
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QUESTION 6.3.a
Comment: Could be answered - Loss of 125VDC or loss of level transmitter or
associated level instrument failure.
Response: Since the question asks for a system, the only other possible
answer for full credit would be the 125VDC system. The answer key
has been changed to give credit for both instrument air and 125VDC
systems.
QUESTION 6.6.b
Connent: If channel 'B' selected answer would be nothing.
Response: If candidate were to indicate that channel 'B' was selected and
that the response would be nothing the candidate was given 25% of
full credit for the knowledge of knowing that only one level
channel was used in the feedwater level control system. The intent
of the question and normal operations assume that channel ' A' is
selected. A clarification note was added to the answer key to
preclude this problem in the future.
QUESTION 6.9.b
Comment: May have each valve being one effect or each pump trip being one
effect.
Response: Credit is to be given for each individual component if they are
system consistent (i.e., M02029A and RHR pump A). A clarification
note was idded to the answer key.
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QUESTION 6.10
Comment: Also alarm at 24" Hg on panel C07.
Response: The answer key was modified to accept either 25" Hg or 24" Hg for
the low vacuum alarm setpoint. Credit was not given for including
both low vacuum alarm - 25" Hg and low vacuum alarm - 24" Hg
because they are both low vacuum alarm indications.
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QUESTION 6.11
Comment: The loop with the highest riser pressure is the selected loop.
Response: Comment accepted. Answer key not changed because comment is
implied in the response given.
QUESTION 7.6.a
Comment: a. Scram could be cause of short duration shutdown. C.1.0006
states answer a.1 but not answer a.2.
Response: Comment accepted. Answer key has been changed to read:
a. 1. When the plant is expected to be restarted after a short
duration (1.0) when no major maintenance has been performed
(1.0).
or
2. When the plant is expected to be restarted af ter a short
duration (0,5) following a scram (0.5) and the cause of
the scram is known (0.5) and the cause remedied (0.5).
(either response is acceptable for full credit.)
Reference: C.1 Starting Procedures, pg. C.1-0001.0.0 and C.1.0006.0.0
QUES 110N 7.7
Comment: The first sentence of question is misleading - not ever on range 8
and within critical band.
Response: Comment accepted. A note has been added to reference to change
" range 8" to " range 2, 3, or 4".
Reference: None provided.
QUESTION 7.9
Comment: The above answer only applies to C.1 startup and C.3 shutdown
procedures to change other procedures need 2 SR0s, one duty SS and
must go through temp change procedure as outlined in Admins.
Controls and cannot change intent of procedures.
Response: Comment no accepted. The question implies that you are performing
steps out of written order, not changing any action step. Answer
stands as is.
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QUESTION 7.12.b
Comment: b. If relief valve is stuck open.
Response: Not accepted. Comment is implied in current response (i.e., relief
valve failure).
QUESTION 8.4
Comment: To much emphasis on memorization.
Response: Comment noted. Excess verbage was removed from answer key content
and point values remained the same.
Reference: None required.
QUESTION 8.6.c
Cpmment: c. By volume F memo. C.1.002
Response: Comment accepted. Comment added to existing answer. Either is
acceptable for full credit.
QUESTION 8.8
Comment: See 4 AW1 13.1.1, pg. 6.
Response: a. According to 4 awl-13.1.1 the answer should read: "No (0.25),
you are allowed to exceed an individual criteria by plus or
minus 2 days."
b. March 29.
Reference: 4 awl-13.1.1, Rev. 4, pg. 6 of 17.
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- I' ilesign Lnange CCA,jsy
Appendix II b
Figure 1
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Rev. 4 Design change Ib. 82M089
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.O Addendum tb.
Appendix II c -
Safety Evaluation
All Safety Evaluations for Design Changes shall be submitted
'Ibe content
Department 3ANIof a Safety Evaluation is determined on
4.1.1. bythis
the fom.
c onPower Produ
This
systan mcriification will be perfomcri during the current outage while the IKI
is inoperabic.
.
The logic
prevent change
a restart willa prevent,Elthe
until loJ- IKI systcm fran cycling, since it will
manual restart is desired. actor water level signal is initiatcd or
Electric, since the present logic allcus a IIPCI restart as soon as th -
still prescnt).Thn. Lexet of retMuy foe bre,&ahigh level trip signal cl
to N.\ tvd of redundawy for the, HP4 Autrt- bp., y h trip Mm is 4 IM mual
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or function, as shown on Figure 1.This modification will not affect any
A pre-opwircd.
properly test will be perforal to verify that the mcxilfication lus been
(
Rmmul of the test circuitry does not affect the operability of the HPCI
systen or the ability to perfom routine surveillance.
.
This design change does not create a possibility for an accident or on malfunct
of a different type than evaluatal previously in the USAR or subsequent su
This design change does not increase the probability of occurrence nor ease
incr.
analy:cd in the USAR or other subnittals.the consequences of any
the margin of safety defined in the bases for any Technical SpecificationT .
It is concitded that this modification does not represent an unreviewed
safety question as definal in 10CFR50.59. .
Prepared By
C Q ugv- Date_ f,7-ff
Reviewed By [lj /k k Date
'M 9-/05f4
Design Review Report Attached yes O
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U. S. NUCLEAR REGULATORY C0Ht11SS10N
SENIOR REACTOR OPERATOR LICENSE EXAttlNATION
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Facility: MONTICELLO
Reactor Type: BWR
Date Administered: April 30,1985
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Examiner: G. A. Sly
Candidate: N M rftl $N ta W * NV
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INSTRUCTIONS TO CANDIDATE:
Use separate paper for the answers. Write answers on one side only.
Staple question sheet on top of the answer sheet. Points for each question
are indicated in parenthesis after the ouestion. The passing orade reautres
at least 70% in each category and a final grade of at least 80%. Examination
papers will be picked up six (6) hours after the examination starts.
Category 1 of Candidate's % of
Value Total Score Cat. Value Cateoory
25 25 5. Theory of Nuclear Power Plant
Operation, Fluids and
Themndynamics
25 25 6. Plant System Design, Control
and Instrumentation
"
25 _25 7. Procedures - Noma 1, Abnomal .
Emergency, and Radiological
Control
25 25 8. Administrative Procedures,
Conditions, and Limitations
100 TOTALS
Final Grade 1
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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5.0 THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS AND THERMODYNAMICS 25 PTS
5.1 Explain under what conditions and how it is possible to receive
an increase in power when rods are inserted into the core. (1.5)
Answer:
During power decreases from 100% to 88% (0.25) by insertion of
shallow rods (rods below core midplane) (0.25). Movement of
rods decreases void formation (0.25) in the lower portion of the
control cell (0.25) which results in a power increase in the
middle and upper portion of the core (0.25) offsetting the
reactivity decrease from rod insertion (0.25). (1.5)
Reference: C.2 Power Operation, pg. 21
- Section 5 Continued on Next Page -
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5.2 The process computer is temporarily out of order and it is
necessary for you to manually calculate three (3) gain
adjustment factors (GAF) for the APRMs. Your reactor operator
has provided you with the following information.
Feedwater flow is 6.25 Mlb/hr
Rx Pressure is 1000 psig
APRM 1 reads 94.5% power
APRM 2 reads 97.5% power
APRM 3 reads 93.5% power
a. By using Figure 4 and the information provided calculate the
GAFs for the three (3) APRMs. (Show all work for credit.) (1.0)
( b. Is APRM 1 power reading conservative or non-conservative? (0,5)
c. For each of the GAFs calculated in part a: state whether
you would or would not order an adjustment of the GAF and
3,1ve the reason for your action. (1.5)
Answer:
a. From Figure 4 core power = 1600 MW thermal (0.25)
% rated thermal power = 1600/1670 = 95.81% (0.25)
GAF = % rated /APRM reading (0.25)
GAF 1 = 1.014
GAF 2 = 0.983
GAF 3 = 1.025
(0.25 for GAF calculations) (1.0)
b. non-conservative (0.5)
c. GAF 1 = would adjust (0.25) due to >1.01 (0.25)
GAF 2 = would not adjust (0.2b) due to >0.98 (0.25)
GAF 3 = would adjust (0.25) due to >1.02 (0.25). (1.5)
Reference: C.2 Power Operation, pg. 32, 33
- NOTE: Feeedwater temp. = 350*F was given during exam. ***
- Section 5 Continued on Next Page -
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S.3 Would the following conditions (increase, decrease or not change)
indicated level? h?
a. increase in drywell temperature (0.75)
b. increase in reactor pressure (0.75)
Answer:
a. increase (0.25), due to reference leg being less dense
(0.5). (0.75)
b. not change (0.25), because Rx level control will maintain
level at setpoint (constant).(0.5). (0.75)
Reference: B.S.7 Reactor Level Control, pg. 8
- Section 5 Continued on Next Page -
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5.4 With a reactor that is critical, low in the intermediate range,
- will an equal amount of positive or negative reactivity
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insertion (e.g., 10E-4 del ta k/k) produce stable periods of
equal mgnitude? Explain you answer. (1.5)
Answer:
No (0.5), the positive insertion period will be larger than the
negative insertion (0.5) due to delayed neutrons (0,5). (1.5)
Reference: MTC Reactor Kinetics pg. 26
Equation Sheet
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- Section 5 Continued on Next Page -
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S.5 State how fuel pin centerline temperature will change (increase,
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decrease, or remain the same) with each of the following
conditions.
a. A 0.001 inch thick layer of corrosion product deposits on
the clad surface. (0.5)
b. A fuel bundle reaches DNB. (0.5)
Answer:
a. increase (0,5)
b. increase (0.5)
Reference: G. E. Thermo Book for Monticello, Chapter 8, pg. 5
- Section 5 Continued on Next Page -
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5.6 Which of the following describes the changes to the steam that
occurs between the inlet and the outlet of a REAL turbine? (1.0)
a. Enthalpy DECREASES, Entropy DECREASES, Quality DECREASES
b. Enthalpy INCREASES, Entropy INCREASES, Quality INCREASES
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c. Enthalpy CONSTANT, Entropy DECREASES, Quality DECREASES
d.
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Enthalpy DECREASES, Entropy INCREASES, Quality DECREASES
Answer: d. (1.0)
- Reference
- G. E. Thermo Book for Monticello, Chapter 6
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- Section 5 Continued on Next Page - '
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5.7 Which of the following is not a characteristic of Subcritical
HJTtTp11 cation? (1,0)
a. The subcritical neutron level is directly proportional to
the neutron source strength.
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b. Doubling the indicated count rate by reactivity additions
will reduce the margin to criticality by approximately
one-hal f.
c. For equal reactivity additions, it takes longer for the next
equilibrium count rate to be reached, as K-eff approaches
unity.
d. If ten (10) notches of rod withdrawal increases the SRM
count rate by 10 cps, then twenty (20) notches of rod
withdrawal will increase the SRM count rate by 20 cps.
( Assume constant rod worth.)
Answer: d. (1.0)
Reference: Subcritical Reactor Theory, pg. 8-12
- Section 5 Continued on Next Page -
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'5.8 What is the change in reactivity associated with a change in
R eTf from 0.920 to 1.004? (1.0) l
Answer: - (0.7 equation, 0.3 numbers, O'.1 math) (1.0) l
P = K2 - K1' = 1.004 - 0.92 =
0.091
Kl(K2) 1.004(0.92) ,
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Reference: MTC Neutron Kinetics, pg. 16
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- Section 5 Continued on Next Page -
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5.9 Which of the following statements about Sm-149 is TRUE? (1.0)
a. It is removed from an operating reactor by burnout and
radioactive decay.
b. When a reactor is restarted after a temporary shutdown
Sm-149 concentration increases for several days.
c. It has less effect on reactor operation than Xe-135 due to
its smaller fission yield and smaller microscopic neutron
cross section.
d. The equilibrium concentration of Sm-149 at 50% FP is about
two thirds of the equilibrium concentration of 100% FP.
Answer: c. (1.0)
Reference: MTC Fission Products Poisoning Effect, pg. 22, 23
- Section 5 Continued on Next Page -
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5.10 Regarding MCPR (Minimum Critical Power Ratio):
a. What phenorenon could exist if a fuel bundle were operated
at a MCPR less than one and what would very likely be the
consequences of the phenomena? (1.0)
b. How does the margin to MCPR change (increase, decrease, or
remain constant) when inlet subcooling decreases? (0,5)
Answer:
a. Transition boiling may occur (0,5) which could resul t in
clad failure (0.5). (1.0)
b. Decreases (0.5)
Reference: G. E. Thermo Book for Monticello, Chapter 9. pg. 77,80
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- Section 5 Continued on Next Page -
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5.11 Arrange the following list in order from greatest ability to
thermalize a neutron to least. (1.0)
a. Deuterium (deleted)
b. Hydrogen
c. Carbon
d. Uranium
e. Ordinary Water
Answer: (0.25 for each)
b-e-a-c-d (1,0)
Reference: Neutron Physics M8102L-005, pg. 38
- NOTE: item a. was deleted from question during exam. each
worth 0.25 pts instead of 0.2 pts. ***
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- Section 5 Continued on Next Page -
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5.12 For the following conditions, state if the fuel temperature
coefficient becomes more negative or less negative. I
a. Increase in fuel temperature (0,5)
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b. Increase in moderator temperature (0.5)
c. Increase in void fraction (0.5)
- Answer:
a. less (0.5)
b. more (0.5)
,
c. more (0,5)
Reference: MTC BWR Inherent Reactivity Coefficients, pg. 23, 12
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- Section 5 Continued on Next Page -
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5.13 Assume that the reactor is being started up from COLD SHU100WN
and a R0D DROP ACCIDENT occurs. Of the void, doppler, and
moderator temperature coefficients, which will act FIRST, SECOND
and THIRD to limit the rapid power rise? Explain your answer. (1.5)
Answer: (0.25 for order, 0.25 for reason)
FIRST - D0PPLER - Acts to turn power due to a larger number of
neutrons being absorbed in the fuel. (0,5)
SECOND - MODERATOR TEMPERATURE - Must get to saturation to see
voids. (0.5)
THIRD - VOID - Same as moderator temperature coefficient. (0,5)
Reference: MTC BWR Inherent Reactivitiy Coefficients
General BWR Reactor Theory
- Section 5 Continued on Next Page -
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5.14 In a reactor fueled with U-235 and U-238?
a. Which nuclide is most likely to fission upon absorbing a
TasFneutron? (0.5)
b. Which nuclide produces most of the power from fast fissions
in a BWR core? (0.5)
c. Which is most likely to ' fission upon absorbing a thermal
neutron? (0.5)
Answer:
a. U-235 (0.5)
b. U-238 (0.5)
c. U-235 (0.5)
Reference: MTC Neutron Kinetics, pg. 31
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5.15 The moderator / fuel ratio of BWR is selected to have an "undermoderated
Core".
a. What is meant by and "undermoderated core" and how does this affect
operations? (1.0)
,
b. What is the basis for detennining the amount of
undermoderation at BOL? (1.0)
Answer:
a. Undermoderated means that the ratio of moderator / fuel is
less than for an optimum condition for Keff (0.5). This
results in assuring negative moderator temperature and void
reactivity coefficients (0.5). (1.0)
b. Should be large enough so that the condition at EOL could
still not resul t in an excessive positive moderator
temperature coefficient (0.5). The amount of
.undermoderateration decreases during core life due to burnup
of the fuel and lower positions of control rods (0.5). (1.0)
Reference: MTC Reactor Physics M8102-L-006, pg. 29
Student handout, pg. 26
- Section 5 Continued on Next Page -
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5.16 Explain or define the following terms:
a. Prompt critical (0.5)
b. Reactor period (0.5)
c. Critical (0.5)
Answer:
a. Reactor critical on prompt neutrons alone (reactivity > beta). (0.5)
b. Time in seconds required for power to change by a factor
'e'. (0.5)
c. Number of fission neutrons in one generation is equal to the
number of fission neutrons in next generation. (half credit
for Keff = 1) (0.5)
Reference: MTC Neutron Kinetics M8102L-007, pg. 30,17, 9
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5.17 A motor driven centrifugal pump is operating at rated flow. You
then start closing down the discharge valve. How will the
following be affected (increase, decrease, or no change)?
a. Flow (0.5)
b. Discharge pressure (0.5)
c. Motor amps (0.5)
d. Net positive suction head. (0.5)
Answer: (0.5 for each response)
a. Decreases (0.5)
b. Increases (0.5)
c. Decreases (0.5)
d. Increases (0.5)
Reference: G. E. Thermo Book for Monticello, Chapter 7
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6.0 PLANT SYSTEM DESIGN, CONTROL AND INSTRUMENTATION 25 PTS
6.1 State whether HPCI will or will not auto initiate upon receipt
of a valid auto initiation after the following HPCI pump trip or
isolations.
a. Mechanical hydraulic overspeed trip (0.5)
b. High reactor level and then level recovery (normal) (0.5)
c. Reactor low pressure isolation (126 psig) and then
increasing to 200 psig. (0.5)
d. Aux oil pump failure and control switch in pull to lock. (0.5)
Answer:
a. will (0.5)
b. will, if new initiction signal is low-low water level, or
will not, if no reason or initiation signal is 2# drywell. (0,5)
c. will (0.5)
d. will not (0.5)
Reference: B.3.2 HPCI, pg. 3, 4, 8
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6.2 Give the proper APRS valve response (open, close, or remain the
same) to the initiating events. Assume each event to be unique.
Explain your answer.
a. Sensing bellow failure and receipt of automatic APRS
initiation signals present. (0.75)
b. APRS blowdown in progress (valves open) and the reactor
operator depresses the Div. A. timer reset button. (0.75)
c. APRS blowdown in progress (valves open) and the reactor
operator depresses both drywell pressure reset buttons. (0.75)
d. Failure of the Div. B logic (faulty K6B relay) and receipt
of automatic initiation signals present. (0.75)
Answer:
Open (following 120 sec. timer) (0.25), since bellows
'
a.
failure only effects sel f-actuation mode (safety valve)
(0,5). (0.75)
b. Remain the same (0.25) need to depress both Div. A and Div.
B to reset timer once blowdown has began (0,5). (0.75)
c. Remain the same (0.25) once logic seals in, you must depress
both timer resets to reset the logic (0.5). (0.75)
d. Open (following 120 sec. timer) (0.25), because only one
active division is necessary to activate APRS (0.5). (0.75)
Reference: B.3.3 Automatic Pressure Relief, pg. 4
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6.3 Concerning the design of the Standby Liquid Control System.
Fill in the blank with the proper word or phrase.
a. Loss of the system will al so make level
indication (L1-II-66) on C05 in the control room indicate
zero. (0.25)
b. The Standby Liquid Control System is designed to inject
enough boron to produce a shutdown margin in less
than minutes. (0.5)
c. Local initiation, under emergency conditions, of the SBLC
pumps fire the explosive valves. (0.25)
Answer:
a. instrument air or 125 VDC (0.25)
b. 3% delta K/K (0.25),125 minutes (0.25) (0.5)
c. will not (0.25)
Reference: B.3.5 Standby Liquid Control, pg. 5, 2
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- Section 6 Continued on Next Page -
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6.4 State whether the following statements concerning the SRV
Low-Low Set System is True or False.
a. A 10 second time delay was installed to inhibit the
reopening of a SRV while an elevated water leg is in the
discharge line. (0,5)
b. The Low-Low Set System consists of valves RV2-71F, G, H. (0,5)
c. Following a relief valve opening and subsequent closure the
reactor operator can override, if necessary, the 10 second
time delay by positioning the manual control switch in the
open position. (0.5)
d. The SRV Low-Low Set Logic is initiated upon receipt of one
out of two twice logic consisting of a Rx SCRAM signal, high
reactor pressure signal, and the control switches being in
auto. (0.5)
Answer:
a. True (0.5)
b. False (E, G, H) (0,5)
c. False (0.5)
d. False (two of two once) (0,5)
Reference: B.3.3 Automatic Pressure Relief, pg.12
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6.5 Concerning the Reactor Protection System:
a. List four (4) Reactor Protection System (RPS) trips that are
necessary in the Refuel mode with the reactor water
temperature less than 212*F. (2.0)
b. What is the purpose behind the Delayed Scram Reset
Interlock? (1.0)
Answer:
a. 1. Mode switch in shutdown (0.5)
2. Manual (0.5)
3. High Flux IRM (0.5)
4. Scram Discharge Volume High Level (0.5)
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b. Allow for the control rods to complete their scram travel. (1,0)
Reference: Technical Specification 3.1/4.1 pg 29
CRD Lesson Plans
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6.6 Explain what effect the following failures would have on reactor
l evel . Why? (Assume 3-element control)
a. 'B' feedwater line flow signal fails high. (1.0)
b. Channel ' A' reactor level detector signal fails low. (1.0)
c. Loss of control signal to 'B' feedwater control valve. (1.0)
Answer:
a. Causes reactor level to DECREASE (0.25) due to the Level
Control System having a STEAM FLOW / FEED FLOW ERROR, STEAM
FLOW < FEED FLOW (0.5) resulting in a CLOSURE SIGNAL TO THE
FEEDWATER CONTROL VALVES (0.25). (1.0)
b. Causes reactor level to INCREASE (0.25) due to the Level
Control System having a LEVEL ERROR, LEVEL SET > INDICATED
. LEVEL (0.5) resulting in an OPEN SIGNAL TO THE FEEDWATER
CONTROL VALVES (0.25). (1.0)
c. Reactor level should REMAIN CONSTANT (0.25) because the 'B'
FEEDWATER CONTROL VALVE WILL LOCK-UP (0.75). (1.0)
Reference: B.5.7 pg. 24, 25
Inel Exam Bank Q. 5122
- NOTE: in part b of question indicate that channnel A is
selected. ***
- Section 6 Continued on Next Page -
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6.7 If normal auxiliary electrical power is available at the time of ECCS
initiations:
a. What are the time delays associated with pump starts for each RHR and
core spray pump? (1.5)
b. Also give the power source for each. (1.5)
Answer:
Time Essential Bus 15 Essential Bus 16
5 seconds RHR Pump A RHR Pump B
10 seconds RHR Pump C RHR Pump D
15 seconds Core Spray Pump A Core Spray Pump B
(0.25 for each response) (3.0)
Reference: B.3.4 RHR, pg. 15
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6.8 The uninterruptible AC Power 120/240 V Distribution Panel Y10 is
normally supplied from 480 V LC-104 through a Battery Charger
and a Static Inverter. If the Static Inverter fails... (CHOOSE
ONE) (1.0)
a. ... the 125 vdc battery will maintain power to the Panel Y10
for up to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />,
b. ...the power supply can be manually transferred to the
alternate 480 v LC-103 / AC Transformer 11 by actuating a
transfer switch,
c. ...the power supply will automatically transfer to the
alternate 480 v LC-103 / AC Transformer 11.
4
d. ...the power supply can be manually transferred to the
al ternate 480 v LC-103 / al ternate Static Inverter by
actuating a transfer switch.
Answer: b. (1.0)
Reference: B.9.13 Instrument AC and Uninterruptible Dist. Sys.,
pg.1, Figure 1.
- Section 6 Continued on Next Page -
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26
6.9 Regarding the Residual Heat Removal (RHR) System while operating
in the Shutdown Cooling (SDC) Mode:
a. State why it is necessary to prevent the RHR pumps'
discharge from decreasing below 300 gpm. (1.0)
b. List three (3) ways in which the system will be affected if
reactor pressure increases to above 75 psig. (1.5)
Answer:
a. To prevent a loss of reactor water inventory to the Torus
(0.8) through the Minimum Flow Valve (0.2). (1.0)
b. SDC PCIS Valves (M02029 & M02030) - Auto Close (0,5)
All running RHR Pumps - Trip (0.5)
Head Spray Valve (M02026 & M02027) - Auto Closes (0.5)
(full credit was given if each valve or pump was listed as
an individual affect, up to three components)
Reference: B.3.4 RHR pg. 19, 23, 29, 31
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6.10 As condenser vacuum decreases from a normal operating vacuum to
atmospheric pressure, what interlocks, trips, or alarms are
expected and what are the setpoints for each? (2.0)
Answer: (0.4 indication, 0.1 setpoint)
Low vacuum alarm 25" Hg or 24" Hg (0.5)
Scram 23.5" Hg (0.5)
Turbine trip 20" Hg (0.5)
Bypass valve trip 7" Hg (0.5)
Reference: B.6.3 Main Condenser, pg. 11, 12a, 13
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6.11 Explain how the LPCI Loop Select Logic functions once an
initiation signal is received. (Include in your explanation of
the effects if the recirculation pumps are running and/or
shutdown. Limit your answer to selection of one loop only). (2.5)
Answer:
. If only one pump is running then the logic trips the
operating pump. (0.5)
. The logic then requires that reactor pressure be below 900
psig before continuing to the 2 second time delay prior to
loop selection. (0.5)
. If one or both recirc. pumps are running or tripped the logic
is delayed for 2 seconds prior to selecting the loop for
injection. (0.25)
. Af ter the 2 second time delay, 0.5 seconds are necessary to
check the break detection logic. (0.25)
. Loop selection is made by comparing the two riser pressures.
If loop 11 riser delta pressure is greater than loop 12 delta
riser pressure by a preset differential then loop 11 is
selected. If loop 11 pressure IS NOT greater than loop 12
pressure by the preset differential, then loop 12 is selected
for injection. (1.0)
Reference: B.3.4 RHR, pg. 16, 17
- End of Section 6 -
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1
7.0 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL CONTROL 25 PTS ,
!
7.1 During a normal reactor shutdown you are cautioned (when the I
outside temperature is below 38*F) not to allow the temperature i
of the discharge water to go below 55'F to preclude.... (choose '
one) (1.0)
a. icing in the cooling tower.
b. excessive subcooling of the circulating water.
c. exceeding the lower limit on the discharge water temperature {
meter.
d. overloading of the cooling tower fan motors.
Answer: a. (1.0)
Reference: C.3 Shutdown Procedure, pg. 009.0.0
- Section 7 Continued on Next Page -
30
7.2 Which of the following is not always a Group I isolation (choose
'one'77 (1.0)
a. Low-Low water level
b. Main steamline low pressure
c. High temperature in the main steamline tunnel
d. High flow in the main steamline.
Answer:
b. (must be in run) (1.0)
Reference: C.4.II Primary Containment Isolation pg 21
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7.3 You enter C.4.III Abnormal Conditions for Service Water System
Failure and are unable to restore the service water system
operation. Operator actions state "D0 NOT initiate drywell
sp ray. . . " This is not done because... (choose the most correct
one) (1.0)
a. The cooling water necessary for essential plant systems will
be diverted from the drywell spray (RHR) system.
b. Implosion of the drywell is possible due to the rapid
depressurization of the drywell.
c. Initiation of drywell spray will produce erroneous level
signals and the operator won't be able to monitor level
'
accurately.
d. Initiation of drywell spray will make the vacuum breakers
inoperable if a vessel blowdown were necessary leading to
overpressurization of the torus.
Answer: b. (1.0)
Reference: C.4 III Abnormal Conditions, pg. 063
- Section 7 Continued on Next Page -
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7.4 Assume a pipe break INSIDE the CONTAINMENT:
a. What are the five immediate actions to be performed to
initiate a RHR loop into the LPCI mode from the shutdown
cooling mode? Valve numbers NOT required. (2.0)
b. Which Reactor Vessel Level indicators should an operator use
during a rapid depressurization? (List two and be
specific.) (1.0)
c. If APRS does not automatically ' initiate on low-low reactor
level and cannot be manually initiated, how many relief
valves should be opened to depressurize the reactor? (0.5)
Answer:
a. 1. Open RHR cross-tie valve (M0-2033) (0.4)
2. Close shutdown cooling suction valve (M0-1988/89) (0.4)
3. Open torus suction valve (M0-1986/87) (0.4)
4. Reset S/D Cooling Group 2 Isolation (0.4)
5. Open LPCI injection valves (M0-2012 to 2015) (0.4)
b. 1. Core Flooding, 400 inch, Yarway (0.5)
2. Operating Range, 60 inch, GE/MAC (0.5)
^
c. As many as possible, up to the number used for APRS (less
than or equal to 3). (0.5)
Reference: MNGP Vol . C.4-0109, 0111, 0113
Inel Exam Bank Q. 9717
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7.5 Prior to any reactor startup the reactor operators must complete
a PRE-STARTUP CHECKLIST. At Monticello, contingencies have been
made for a " modified short fonn" of the PRE-STARTUP CHECKLIST.
a. Under what conditions could the modified short form
PRE-START 0FCHECKLIST be used? (2.0)
b. Who must (by title) authorize the use of the PRE-STARTUP
UiECKLIST? (0.5)
Answer:
a. When the plant is expected to be restarted after a short
duration shutdown (1.0) when not major maintenance has been
performed (1.0). (2.0)
or (for full credit)
When the plant is expected to be restarted after a short
duration (0.5) following a scram (0.5) and, if the nature of
the scram is known (0.5) and the cause remedied (0.5). (2.0)
b. Operations Superintendent. (0,5)
Reference: C.1 Startup Procedures, pg. C.1-0001.0.0
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7.6 a. List the conditions that will initiate the ATWS trip and the
action (s) it produces. (1.0)
b. The ATWS event procedure instructs the LPE&R0 to initiate
the SBLC system if certain conditions exist. What are these
conditions? (1.5)
c. Once SBLC is initiated when can you terminate the injection?
Why? (1.0)
Answer:
a. 1135 psig (0.25) or Low-Low Rx water (-47" or -48") level
after a 9 sec. time delay (0.25). The ATWS trip opens the
recirc. MG field breakers (0.25) and opens the ARI valves
(0.25). (1,0)
b. Unable to maintain the reactor subcritical (0.5) AND
RPV water level cannot be maintained (0.5) OR (1,0)
Suppression pool water temperature cannot be maintained
less than 110*F (0.5). (0.5)
c. Once SBLC is initiated the complete charge is to be injected
(0.5). To ensure S/D margin maintained as C/D, dilution,
poison decay and reactivity coefficient feedback take place
(0.5). (1.0)
Reference: MSP Op. Manual C.4.1-11 & B.3.5
Inel Exam Bank Q.9718
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7.7 You have just reached range 8 on the IRM's and are still pulling
rods, and ' within the predicted critical band. According to
Startup Procedure (C.1); what are the Ifmitations imposed on
control ' rod movement for:
a. non - HI-LITED rod (1.0)
b. HI-LITED rod (0.5)
c. HI-LITED and circled rod (1.0)
Answer:
a. None on rods outside rod sequence step on which criticality
is expected (0.5). Single notch for rods in rod sequence
step on which criticality is expected (0.5). (1.0)
b. Single notch withdrawal only (0.5). (0.5)
c. No rod movement (0.5) until steam flow through bypass valves
(0.5). (1.0).
Reference: Startup Procedure: C.1-0020.0.0
- NOTE: Range 8 isn't valid should change to range 2, 3, or 4. ***
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7.8 According to the LOSS OF COOLANT ACCIDENT procedures, what are the three
(3) conditions when the automatic action of an Emergency Core Cooling
System may be manually overriden? (3.0)
Answer:
. Continued operation of the system will result in an unsafe
plant condition. (1.0)
. It is known or positively determined that the automatic
action as initiated by a spurious or erroneous signal and it
is verified that operation of the system is not required. (1.0)
. Approved procedures specifically allow manual override under
specific conditions and those conditions are verified to be
satisfied. '
(1.0)
Reference: MNGP Ops. Manual, C.4-107
Inel Exam Bank Q. 5141
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1
7.9 under normal use, steps within a given procedure are generally
to be performed in sequential order. When would it be
permissible to violate the sequential performance of action
-
steps? Thrte (3) responses required. (1.5)
J
Answer:
1. Steps are not omitted (0.5)
2. Steps are performed in the manner described (0.5)
3. Steps which are called for at or prior to reaching specific
operating conditions are performed before passing beyond
i
these conditions (0.5). (1.5)
Reference: C.1 - Startup Procedures C.1-0002.0.0
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38
7.10 You are the Emergency Director during a radiological accident
and have decided to use the Emerge'cy Exposure Guidelines during
the corrective actions. What are four (4) of the criteria you
would use to select the personnel for the job, all other things
being equal, i.e., skill and job familiarity? (3.0)
1
' Answer:
l
1. Personnel receiving increased exposure should be volunteers
or professional rescue personnel ..
2. Personnel should be broadly familiar with the consequences
of exposures received under emergency conditions.
3. Women in their reproductive years should not take part.
4. Exposures under these conditions should be limited to one in
a lifetime.
5. Internal exposure should be minimized by the use of
appropriate respiratory equipment, and contamination should
be controlled by the use of appropriate protective
clothing.
6. Volunteers above the age of 45 are recommended.
7. Personnel shall be emergency workers. (3.0)
(4 required 0 .75 each)
Reference: MNGP Vol . A.2.401
Inel Exam Bank Q. 9722
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7.11 What are the emergency start signals for the standby diesel ? (1.0)
Answer:
1. Low-Low water level
2. High drywell pressure
3. Loss of voltage on safeguards buses 15 or 16.
4. Degraded voltage on safeguards buses 15 or 16.
(4 required at 0.25 pts each) (1.0)
Reference: C.4.II Primary Containment Isolation pg 22
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7.12 Considering the Relief Valve Failure Manual (C.4.III.0):
a. Under what conditions should and should not the reactor I
operator manually operate the valve in an attempt to reseat
it ? (1.0) '
b. Under what conditions would it become necessary to manually
scram the reactor ? (0.5)
Answer:
a.- Relief Valve Leaking - should not (0.5) (as dictated by <
100,000 lbs/hr) (0.5)
Relief Valve Failure -
should (0.5) (as dictated by >
100,000 lbs/ hrs) (0.5)
b. If the leakage rate remains greater than 100,000 lbs/hr (0.5)
Reference: C.4.Ill.0 Relief Valve Failure Manual pg 113-115
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41
8.0 ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 25 PTS
8.1 There are two (2) different safety limits specified for core
thermal power in technical specifications. State each safety
limit and state when each would apply to reactor operations? (2.0)
Answer:
1. MCPR >1.07 (0.5) when Rx pressure >800 psia and core flow is
>10% of rate (0.5). (1.0)
2. Thermal power shall not exceed 25% of rated (0.5) when Rx
pressure 1800 psia or core flow 110% of rated (0.5). (1.0)
Reference: Tech. Specs. Section 2
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8.2 Briefly explain why each of the following recirculation system
limitations are necessary. l
a. With both pumps running, the speed of the faster pump may
not exceed 130% of the speed of the slower pump for a core
power less than 80%. (1.0)
b. The operating pump must be reduced to 50% speed or less
prior to restarting the tripped pump. (1.0)
c. Recirc flow shall not be increased unless the coolant
temperature difference between the bottom head region and
upper region of the vessel is less than 145*F. (1.0)
Answer:
a. To enhance the capability of the LPCI Loop Selection Logic
to detect some limited low probability breaks in the recirc.
loop. (1.0)
b. To prevent excessive jet pump vibration. (1.0)
c. To preclude excessive thermal stresses on the reactor bottom
head-to-support skirt transition and/or CRD stub tubes.
(Either component for full credit.) (0.25 partial credit
for cold water reactivity accident.) (1.0)
Reference: MNGP Ops. Manual, B.1.4
- Section 8 Continued on Next Page -
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8.3 According to Administrative Control Directive 4 ACD-3.6, Work
Request Authorization:
a. What are CRITICAL SYSTEMS?- (1.0)
b. For work on Critical Systems, what two (2) individuals must i
give approval prior to work on the Critical System? (1.0)
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Answer: ,
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a. Systems or equipment that are required to be operable by
Technical Specifications or are critical to continued 1
operation of the plant. (1.0)
Also accept: 1. Safety systems, structures, and components
identified on Q-list extension
2. Fire Protection
3. Systems, equipment, instruments or
structures identified by WRA coordinator.
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b. WRA coordinator and the superintendent, operations. (1.0)
Reference: Inel Exam Bank, Q. 9731
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8.4 Under what conditions can a system be considered OPERABLE for
the purpose of satisfying the requirements of applicable LCO's
if the system has been determined to be IN0PERABLE solely
because its normal power source is INOPERABLE. (1.5)
Answer:
It may be considered operable for the purpose of satisfying the
. requirements of its applicable Limiting Condition for Operation
provided: (1) its corresponding emergency power source is
operable (0.5); and (2) all of its redundant system (s) (0.2),
subsystem (s) (0.2), train (s) (0.2), component (s) (0.2), and
device (s) (0.2) are Operable, or likewise satisfy the
requirements of this paragraph. (1.5)
Reference: Tech. Specs., pg. 1.0-3
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8.5 What is the Technical Specification reason for requiring plant
shutdown (cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) when a jet pump has
been declared inoperable? (Two reasons required.) (1.0)
Answer: (only 2 required for credit)
1. Provide leak path past the core. (0,5)
2. Provide for an increase in blowdown flow area following a
3. Could not insure 2/3 core coverage. (0.5)
Reference: Tech. Specs. Bases 3.6/4.6, pg. 153, 154
- Section 8 Continued on Next Page -
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46
8.6 Concerning the plant STARTUP PROCEDURES:
a. Other than an item that can be brought into conformance with
the checklist, what two (2) criteria would you as Shift
Supervisor use to determine if rod withdrawal may be
initiated to bring the reactor critical when a checklist
required for the startup contains a CIRCLED ITEM, a
condition not in conformance with the checklist
requirements? (1.5)
b. Administratively, what must be done to proceed with rod
withdrawal if the CIRCEED ITEM in part (a) cannot be brought
into conformance with the checklist, but meets the criteria
above to allow the startup? (1.5)
c. What must be issued to implement a change to the control rod
withdrawal sequence? (0.5)
Answer:
a. It does not conflict with the Tech. Specs.
It does not interfere with safe operation. (1.5)
(2 at 0.75 each)
b. The item must be initialed by the Shift Supervisor (0.5) and
another SR0 (0.5) and submitted as a temporary change
(0.5). (1.5)
c. Management memo or by volume F memo. (0.5)
Reference: MNGP Ops. Manual C.1-285
Inel Exam Bank Q. 5129
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8.7 Fill in the blank and provide the reason for the following
technical specification primary system boundary limits: (A
phrase may be needed to complete each condition).
a. The average rate of reactor coolant temperature change
during normal heatup or cooldown shall not exceed
. Why? (1.0)
b. During heatups and cooldowns the following temperature shall
be recorded at least every until . (0.5)
c. At what locations must the temperatures be recorded in part
'b' above. (1.0)
Answer:
a. 100%"F/hr (0.25), when averaged over a one-hour period
(0.25) to prevent excessive stresses on the vessel wall s
(0.5). (1.0)
b. 15 minutes (0.25) until three consecutive readings at each
location are within 5*F (0.25). (0.5)
c. 1. Rx vessel adjacent to shell flange
2. Rx vessel bottom drain
3. Recirculation loops A and B
4. Reactor vessel bottom head.
(0.25 for each) (1.0)
Reference: Tech. Specs. 3.6/4.6, pg. 121
Tech. Specs. Bases 3.6/4.6, pg. 145, 146
- Section 8 Continued on Next Page -
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ , . _ _ . -
_ _ - .
.
48
8.8 A weekly surveillance has been performed on the following days:
February 27
March 4
March 12
March 20
a. Has the surveillance interval been exceeded for this set of
dates? (Yes/No) Explain your answer. (1.0)
b. When is the maximum allowable date that the next
surveillance can legally be performed? (0.5)
Answer:
'
a. No (0.25), you are allowed to exceed an individual criteria
by plus or minus 2 days (0.75). (1.0)
b. March 27. (0.5)
Reference: 4AW1 13.1.1, pg 6.
- NOTE: Question should state the first date as being performed
on the required surveillence date to limit candidate
confusion. ***
4
1
- Section 8 Continued on Next Page -
,
, - ., -
m , - , ,y..-_._.-.-,wr - .-y. ,. _ . ,
__,,,m-,, _ . .
49
8.9 State whether the following items would constitute a core alteration.
Yes/No.
a. Removal of an NDT sample located inside the core shroud (0.5)
b. Removal of the steam separator units (0.5)
c. Removal of a SRM while in run (0.5)
d. Removal of a control rod drive during refuel (0.5)
Answer:
a. yes (0.5)
b. no (0.5)
c. no - (0.5)
d. yes (0.5)
Reference: Tech. Specs, pg. 1
- Section 8 Continued on Next Page -
_., . _ - - . _ - - . . - - _ , , . - - - . . - , - . . - . _ , _ _ .-,.____ ._-,-. . . - . - . - - .
. - . -
_ _ _ _ _ - - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
50
8.10 Define or Explain each of the following technical specification
terms:
a. Immediate (0.5)
b. Limiting Condition for operation (LCO) (0.5)
c. Limiting Control Rod Pattern (LCRP) (1.0)
d. Identified Leakage (1.0)
e. Site Boundary (0.5)
Answer:
a. Immediate - Immediate means that the required action will be
initiated as soon as practicable considering the safe
operation of the unit and the importance of the required
action. (0.5)
b. Limiting Conditions for Operation (LCO) -
The limiting
conditions for operation speci fy the minimum acceptable
level s of system performance necessary to assure safe
startup and operation of the facility. (0.5)
c. Limiting Control Rod Pattern (LCRP) - A limiting control rod
pattern for rod withdrawal error (RWE) exists when: a)
thermal power is below 90% of rated and the MCPR is less
than 1.70 (0.5), or b) thermal power is 90% of rated or
above and the MCPR is less than 1.40 (0.5) (1.0)
d. Identified Leakage - Identified leakage shall be:
1) Reactor coolant leakage into drywell collection systems,
such as pump seal or valve packing leaks, that is
captured and conducted to a sump or collecting tank
, (0.5), or
!
l 2) Reactor coolant leakage into the drywell atmosphere from
! sources which are specifically located and known not to
l be Pressure Boundary Leakage or which do not
significantly impair the methods used to detect reactor
'
coolant leakage (0.5). (1.0)
e. Site Boundary - Means a line within which the land is owned,
leased, or otherwise controlled by the licensee. (0.5)
i
Reference: Tech. Spec. Definitions, pg.1-Sa
!
,
- Section 8 Continued on Next Page -
- _ _ _ _ _ , _ - - - -- _- - _ _ ____ - _ __ _ _ _ _ _ -. .. _ _ , -
51
8.11 Maximum suppression pool level during normal operation is
limited to 72,910 cubic feet by Technical Specifications. What
is the bases for this maximum level? (1.5)
Answer:
The design volume of the suppression chamber (water and air) was
obtained by considering that the total volume of reactor coolant
to be condensed is discharged to the suppression chamber (0.5)
and that the drywell volume is purged to the suppression chamber
(0.b). Using the maximum water level val ue, containment
pressure during the DBA is approximately 41 psig which is less
than the design value of 61 psig (0.5). (1.5)
Reference: Tech. Specs., Pri. Cont. Bases, pg. 3.7/4.7-175
- Section 8 Continued on Next Page -
._. _. _ ._. __ _ _. _ __
l
.
52
'
8.12 Temporary changes to a safety related Operation Manual, which do
not change the intent of the permanent procedure can be made
with whose concurrence? (1.0)
i
Answer:
Two SR0's are required. (1.0)
Reference: 4 AWI-4.1.2 pg. 4
-
End of Section 8 -
END OF EXAM ,
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.___._______....______.._______...__.......___...________.._-___.______.__
EQUATION SHEET
7 ____....____. __________ ..... ________. ________________________,.._______ ,
(
'
. .
Where mi = m2
(density)i(velocity)i(area)1 = (density)2(velocity)2(area)2
_____..___..._________.._________ . ____________________________ ...______
KE = mv2 PE = mgh PEi +KEi +P1Vi = PE +KE where Y = specific
7 2 +P
2 Y22
volume
P = Pressure
.__..._____... __________... _______________ ._.... .....___________......
Q = icp(Tout-Tin) Q = UA (T ave -Tstm) Q = 5(hi-h2 I
P = Pg10(SUR)(t) p p et/T SUR = 26.06 T = (B-p)t
i P
t
delta K = (Keff-1)/Keff CRg(1-Keff1) = CR2 (1-Keff2) CR = S/(1_KeffI
M = (1-Keffi) SDM = (1'Keff) x 100%
(1-Keff2) K
eff
,
..______.______.. __...____________________._____ _________________.._____
decay constant = In (2) = 0.693 A = Ag e-(decay constant)x(t)
t t
1/2 1/2
.__....__..___ ___________________............... _________________....... ,
Water Parameters Miscellaneous Conversions
. I gallon = 8.345 lbs 1 Curie = 3.7 x 1010 dps
1 gallon = 3.78 liters 1 kg = 2.21 lbs
1 ft3 = 7.48 gallons 1 hp = 2.54 x 103Btu /hr
3
Density = 62.4 lbg/f t 1 MW = 3.41 x 106 Btu /hr
Density = 1 gm/cm
,
1 Btu = 778 ft-lbf
Heat of Vaporization = 970 Btu /lbm Degrees F = (1.8 x Degrees C) + 32
Heat of Fusion = 144 Btu /lbm 1 inch = 2.54 centimeters
1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 f t-1bm/1bf-sec 2
_____.______...__...___......_____..........______... _____________...____
%
6
1.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -