IR 05000354/1986016
| ML20197B087 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 04/30/1986 |
| From: | Crescenzo F, Keller R, Kister H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20197B075 | List: |
| References | |
| 50-354-86-16, NUDOCS 8605120459 | |
| Download: ML20197B087 (200) | |
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I EXAMINATION REPORT Examination Report No.
86-16(OL)
Facility Docket No:
50-354 Licensee: Public Service Electric and Gas Co.
P.O. Box 236 Hancocks Bridge, New Jersey 08038 Facility:
Hope Creek Examination Dates:
February 24-28, 1986 Chief Examiner:
w M. M7(.
- 3 SIN Franl J. Crfscenz,4teac<or Engineer
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Date (Examiner Reviewed by:
N25W4 Robert M. Keller, Chief Date Pro ects Section 1C l
Approved by:
O Harry BMister, Chief
' Dath '
j Projects Branch No. 1 Summary: Operator Licensing examinations were administered at Hope Creek during the week of February 24, 1986. Eleven Senior Reactor Operator candi-dates, one Instructor Certification candidate and seven Reactor Operator candidates were examined. One Senior Reactor Operator candidate failed the written examination; one failed the written and oral examination; one Reactor Operator failed the written examination and one failed the oral and simulator examination.
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8605120459 860502-
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PDR ADOCK 05000354 V
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.f b.
Strengths noted were as follows:
1.
Safety system knowledge was good with exception of item 3.a.2 above.
2.
Knowledge of control roon back panel operation had improved significantly since the last examination period.
3.
The candidates were quick to use the procedures to correct off-normal conditions. This also is an improvement since the last examination period.
4.
Summary of generic strengths or deficiencies noted from grading of written examination:
The results of the SR0 examination were generally positive with the exception of Section 8.
Both failures were attributable to low scores in this section and two other candidates achieved marginal scores in this section. These low scores can be linked to a combination of three factors:
1.
A weakness in abilities to interpret and apply Technical Specifications.
2.
Three points were deleted from this section (see paragraph 8) thereby increasing the relative importance of the remaining points.
3.
Several candidates delayed completion of this section and were rushed to complete it.
The results of the RO were marginal although the percentage pass rate was higher than the SRO results. This is demonstrated by a group high of 84.05% overall with no candidate achieving greater than 87.7 in Section 3.
In particular, as noted in paragraph 3.a.3 above, almost all of the candidates performed poorly on a question concerning F.W.C.S. failures.
It appears that the candidates were trained to a standard BWR-4 F.W.C.S.
which differs from what Hope Creek has actually installed. This is reflected in the training material and must be corrected prior to use for subsequent classes.
5.
Personnel Present at Exit Interview:
NRC Personnel Frank Crescenzo
,
Facility Personnel R. Salvesen, General Manager, Hope Creek H. Hanson, Manager Nuclear Training G. Connor, Operations Manager, Hope Creek G. Mecchi, Operations Training S. Ketcham, Operations Training 6.
Summary of NRC Comments made at exit interview:
a.
The items noted in Paragraph 3 above were discussed.
b.
The results of a training program inspection documented in Inspection Report 50-354/86-15 were discussed.
7.
Summary of facility comments and commitments made at exit interview:
a.
Plant management is currently working on providing technical specification interpretation guidelines.
b.
The written examinations were too long.
c.
Examiners were professional and competent.
8.
Changes made to written examination during examination review:
All comments to the writtan examinations were resolved during the examination review. The t illowing represent significant changes made as a result of these resoluti,7s.
QUESTION CHANGE REASON 1.02 Delete precit 1ed Plant would refuel prior startup.
to this.
2.03 Item 3 not required unless Pressure detector is RPV assumed depressurized.
upstream of IV's.
2.06 Delete: " Gland Seal steam Clean steam used for isolation".
gland seal steam.
3.02 No change for Part a.
Feed flow is detected at individual feed pump.
3.03 Add rod 39-27 to Part c.
Alternate acceptable answer.
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3.05 Delete: "PAM" from answer.
Not really a range but rather a system.
3.06
"a" is lower Relay Room Correct answer.
"d" is deleted.
New logic is true.
3.10 Delete "f".
"3 running" was not defined as before or after. Question invalid.
5.01 Changed answer setpoints.
Reflects current H.C.
T. S. setpoints.
5.05 Deleted 1.5 points.
Question was beyond scope of examination.
6.03 Same as 2.03 Same as 2.03
6.07 Changed answer.
Reflects updated material.
6.12 Same as 2.06.
Same as 2.06.
8.03 Changed answer to reflect Reference provided.
correct T.S. action.
8.04 Question deleted.
No reference exists.
8.06 Answer changed to reflect Reflects correct T.S.
C.S. subsystem inop.
interpretation.
8.07 Deleted answer for most This term is not clearly limiting.
defined.
8.08 Question deleted.
Reference changed, no longer valid.
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8.11 Answer changed to allow Correct T.S. interpreta-I hour to trip channel.
tion.
Attachments:
1.
Written Examination and Answer Key (SRO)
2.
Written Examination and Answer Key (RO)
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NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
HOPE CREEK
_________________________
REACTOR TYPE:
BWR GE4
_________________________
DATE ADMINISTERED: 86/02/24
_________________________
EXAMINER:
CRESCENZO, F.
APPLICANT
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INSTRUCTIONS TO APPLICANT:
__________________________
Use separate paper for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each cateSory and a final grade of at least 80%.
Examination papers will be picked up six (6)
hours after the examination starts.
% OF CATEGORY
% OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY
________ ___________________________________
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25 0 25.00 1.
PRINCIPLES OF NUCLEAR POWER
___1_0
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PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
___1_0______1_0_
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25 0 25 0
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PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 25.00 25.00
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INSTRUMENTS AND CONTROLS
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1____ _25 00__1__
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PROCEDURES - NORMAL, ABNORMAL, 25 00
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EMERGENCY AND RADIOLOGICAL
___
CONTROL 100.00 100.00 TOTALS
________y______
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FINAL GRADE _________________%
All work done on this examination is my own. I have neither given nor received aid.
EPPL5CEATI5~555 ETUR5~~~~~~~~~~~~~~
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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE
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GUESTION 1.01 (2.50)
a.
At the beginning of a fuel cycle, control rod density is approximately 10 to 12% at equilibrium full power. Approximately one third into the cycle, the control rod density is about 15 to 16% at equilibrium full power. Why is there a difference?
(1.00)
b.
What effects does this increase in control rod density have on the void coefficient of reactivity? Explain your answer.
(1.50)
GUESTION 1.02 (2.00)
The reactivity worth associated with equilibrium samarium concentration a.
is about 1.0% dk/k, however, unlike Xenon, this concentration does not change during normal reactor power level changes. Why is this?(1.00)
b.
When would changes in samarium concentration be considered during reactor cperations? (consider BOL and EOL)
(1.00)
GUESTION 1.03 (2.00)
Indicated reactor water level at 100% power differs from the actual water level above the core (that which is present in the steam separators or within the dryer skirt).
a.
Which level (actual or indicated) is higher and by how many inches?
(1.00)
b. Explain why the above difference occurs.
(1.00)
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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE
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QUESTION 1.04 (2.50)
A. Determine the condenser hotwell subcooling (condensate depression) if the condenser vacuum is 27.9 inches of Mercury and the condensate temperature is 90 F.
(1.00)
B. What is the major disadvantase of condensate depression?
(0.50)
C. How does increased condensate depression affect condensate pump net positive suction head?
(0.50)
D.
Give one example of how you, as an operator, can increase condensate depression.
(0.50)
OUESTION 1.05 (3.00)
Explain the difference between the delayed neutron fraction (BETA-CORE)
a.
and the effective delayed neutron fraction (BETA-EFF).
(1.00)
b. How and why does the delayed neutron fraction change through core life (BOL-EOL)?
(1.00)
c.
How does this change in the delayed neutron fraction affect reactor (1.00)
power control?
QUESTION 1.06 (3.00)
Refering to the attached graph of ' rod worth cold to hot full power,'
explain the shape of the curve using the four terms of the rod worth proportionality equation. Be sure to discuss the reason for changes in each section of the curve, 0-A, A-B, B-C.
(3.00)
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QUESTION 1.07 (1.00)
Which of the following requires a steam condenser to remove the most
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heat energy in order to fully condense the steam!
(1.00)
a. One pound of steam at 300 psia b. Two pounds of steam at 600 psia c.
Two pounds of steam at 1200 psia d. One pound of steam at 15 psia
GUESTION 1.08 (2.50)
Given a large vented tank 30 ft. in diameter and 60 ft. high with a centrifugal pump taking a suction from its base. The pump is located at a vertical elavation corresponding to the bottom of the tank and it requires 5 ft. of net positive suction head (NPSH) to prevent cavitation. The tank is entirely full of' water and is maintained at
60 des F by heaters. The tank is designed such that it could
. withstand 15 psi differential pressure in either direction. Assume
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the vent becomes totally clogged while the pump is in operation.
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Answer the following questions.
a. What is the lowest pressure that the tank will drop to as the pump continues to remove water from the tank?
(0.50)
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-b. Will the pump loose NPSH and begin to cavitate prior to reaching a level of 5 ft. in the tank? EXPLAIN. (State any assumptions)
(1.00)
Could the pump continue to pump water at a level below 5 ft c.
without cavitation if the vent were open? EXPLAIN.
(1.00)
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GUESTION 1.09 (2.50)
Three (3) minutes following a reactor scram from high powere indicated reactor p,ower is 75 on IRM range 4 and decreasing.
a. WHAT will INDICATED power be one (1) minute later?
(1~.5)
(Show calculations)
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(1.0)
b. Explain WHY power decreased at this rate.
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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE
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GUESTION 1.10 (1.00)
Current plant conditions suggest that a severly degraded core condition exists. Several SRV's have been, or are currently, open. The STA reports that the temperature detectors in the SRV tailpipes are reading ERRONEOUS since the readings are significantly higher than those calculated from the Mollier diagram. Would you agree with this conclusion? Justify (1.00)
your answer.
QUESTION 1.11 (1.00)
The RBH has a gain adjustment circuit which raises the gain of the LPRM averagin3 circuit to be equal with the APRM reference signal. Why is this (1.00)
gain adjusment necessary?
GUESTION 1.12 (2.00)
Which of the followin3 situations is correct in most cases?
(2.00)
Explain your choice.
A.
A control rod's worth is greatest when it is fully withdrawn and all other rods remain inserted.
B.
A control rod's worth is greatest when it is fully inserted with all others withdrawn.
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PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
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GUESTION 2.01 (1.00)
Each Hydraulic Control Unit (HCU) has a pair of normally CLOSED Scram valves that provide a path for CRD water during a reactor Scram.
The Scram inlet valves...(CH00SE ONE)
(1.00)
...oPen faster than the Scram outlet valves in order to a.provide adequate drivin3 force to the CR0 mechanism to ensure positive Scram insertion.
b.
...are normally held closed by air pressure and will open by spring pressure when either one of the two associated Scram pilot air valves is deenergized.
c.
...will not prevent their associated CRD's from Scramming if they fail to open, providing that reactor pressure is Sreater than 200 psis.
d.
... connect the Scram accumulator to the under-piston port in the Control Rod Drive.
QUESTION 2.02 (3.00)
For each of the RCIC System component failures listed below, state 1.
Whether or not RCIC will auto inject into the reactor vessel, 2. If it will not inject why, and if it will inject, provide one potential adverse effect or consequence of system operation with the failed component.
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Assume NO OPERATOR ACTION, and the component is in the failed condition at the time RCIC receives the auto initiatin3 si3nal.
a. The Barometric Condenser VACUUM PUMP fails to operate.
(1.00)
b. The MINIMUM FLOW VALVE fails to auto open (STAYS SHUT) when system conditions require it to be open.
(1.00)
The RCIC PUMP DISCHARGE FLOW ELEMENT OUTPUT SIGNAL (to the c.
RCIC flow controller) is failed at its MAXIMUM output.
(1.00)
QUESTION 2.03 (3.00)
Following a valid initiation of HPCI, due to high drywell pressure, a steam leak just upstream of the steam admission valve develops. Describe the response of components, within the HPCI system, assuming the high drywell signal remains. Limit your answer to include all generated logic signals, system valve responses, and final HPCI turbine status.
(3.00)
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PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
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OVESTION 2 04 (1.50)
Answer the following TRUE or FALSE concerning the HPCI system o. All valves, components, and instumentation necessary for auto initiation are independent of AC power sources. (assuming HPCI in (0.50)
standby mode)
b. Following a mechanical overspeed condition on the HPCI turbine, the turbine trip must be reset locally prior to further HPCI operation.
(0.50)
Following a high suppression pool level, the HPCI suction from c. the suppression pool, (HV-F042), does not reach the full open position due to a malfunction. The HPCI suction from the CST, (HV-F004),
will remain opsn and potentially drain CST water to the suppression (0.50)
pool.
GUESTION 2.05 (3.00)
With regard to the Standby '. 49 aid Control System (SLC)
During the filling of the Standby Liquid Control Storage a.
Tank, some of the liquid everflows through the overflow line.
WHERE would the fluid drain to and why?
(1.00)
b. If SLC were manually injected, LIST five (5) indications that the operator could observe in the control room which
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would indicate that the SLC system was injecting.
(Do not include system initiation control switch or manual shutoff valve positions.)
(2.00)
l QUESTION 2.06 (2.00)
List the FOUR automatic actions that thuuld occur, OTHER THAN a Group 1 Isolation, if the Main Steam tine Rad Monitors
'A'
and
'B'
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reach their High-High trip setpoint.-timit your answer to those actions
,
which occur as a DIRECT RESULT of the signal.
(2.00)
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PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
QUESTION 2 07 (3.00)
Briefly describe the auto response of the SW and STACS systems if a slow leak in the TACS system occurs, given the following initial conditions.
- SW pumps A,B operating, C,D in standby.
- SACS loop A operating, SACS B in standby,TACS supplied by SACS loop A.
- All systems operationali no operator actions.
- The leak exceeds normal makeup flow (from demin).
(3.00)
QUESTION 2.08 (1.50)
Given all RHR pumps are available for operation and the RHR system is in the normal standby alignment, describe the pump starting sequences in the event of drywell pressure attaining 1.68 psis concurrent with; (0.75)
a.
Off site power available.
b. Off site power unavailable.
(0.75)
QUESTION 2.09 (2.00)
State whether the following actions WILL or WILL NOT occur if vessel pressure were to increase to 90 psig while in the SDC Mode:
(2 00)
a. Shutdown Cooling Suction Valve (F008) auto closes.
b. All running RHR pumps trip.
c. RHR pump suction valve (F006 A/B) auto closes.
d.
Outboard head spray valve (F023) auto closes.
QUESTION 2.10 (1.00)
What is the function of the static switch in the un-interruptible (1.00)
power supply?
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YSTEMS PAGE
_________________ INCLUDING SAFETY AND EMERGENCY S 2.
PLANT DESIGN ______________________________________
od EML P' c QUESTION 2.11 (3.00)
esm3 s ei'x,Mb With the plant operatinmg at 100% powere recire in Mas er Manual, an cpertor inadvertently decreases the pressure by 5 psi. What will be the initial response and final status of the following parameters due to this action? Utilize the attached EHC logic diagram if necessary. Briefly (3.00)
oxplain for initial response only.
o. TCV position b. BPV position c. Power d.
Pressure.
QUESTION 2.12 (1.00)
If one of the Backup Scram Valves thould fail to operate on receipt of the appropriate signal, what assures that the other valve can (1.00)
perform the necessary function?
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INSTRUMENTS AND CONTROLS
QUESTION 3.01 (2 50)
Answer the followins questions based upon the situation described calow.
The RRCS is fully operational.
The RRCS receives a reactor High Pressure (1071 psis) signal in both complementary logics of a RRCS channel and remains in for 64 seconds.
It takes 57 seconds from the initial reactor High Pressure signal before (2.50)
the APRM levels are downscale.
o.
WHICH of the four logics integrated into RRCS are actuated T=0 seconds?
b.
WHICH logics are actuated at T=10 seconds?
WHICH logics are actuated at T=25 seconds?
c.
d.
WHICH logics are actuated at T=53 seconds?
HOW LONG from T=0 seconds is it before the RRCS can be reset?
e.
QUESTION 3.02 (2.00)
Asssume the FEEDWATER LEVEL CONTROL SYSTEM is being operated in
'A'.
3-ELEMENT control usins reactor LEVEL DETECTOR CHANNEL Reactor power is at 85%, STEADY STATE.
For each of the instrument or control signal failures listed below, STATE HOW REACTOR LEVEL WILL INITIALLY RESPOND (increase, decrease, or remains constant) and BRIEFLY EXPLAIN WHY in terms of WHAT is happening in the Feedwater Control System IMMEDIATELY AFTER THE FAILURE.
(FOR EXAMPLE, your answer should include the following detail,
.
'Causes reactor level to decrease due to a steam flow / feed flow error signal, steam flow < feed flow, resulting in a signal to increase the speed of the reactor feed pump (s)," IF APPLICABLE.)
(1.00)
FEEDWATER line FLOW signal FAILS HIGH.
a.
(1.00)
b.
Channel
'A' REACTOR LEVEL detector signal FAILS LOW.
CATEGORY 03 CONTINUED ON NEXT PAGE xxums)
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INSTRUMENTS AND CONTROLS PAGE
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QUESTION 3.03 (3.00)
Assume the following initial rod position distribution:
All rods in RWM Groups 1-3 are withdrawn to their maximum limits 22-51 in Group 1, 46-55 in except one rod in each Group
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Group 2, and 18-03 in Group 3 (1hese 3 rods are FULLY INSERTED).
All rods in Groups 4 - 10 are FULLY INSERTED to position 00 except for rod 34-27, (Group 4), which is withdrawn one even notch past its pull sheet limit. Assume RSCS bypassed.
Fill in the following table with the Rod / Rod Group number you would expect to see displayed in each RWM window for EACH of the situations (a - c). If nothing will appear, write ' Blank."
(3.00)
SITUATIONS RWM (a)(1.00)
(b)(1.00)
(c)(1.00)
WINDOW Initial Conditions IC with Rod 34-27 IC with Rod 22-51 (IC)
INSERTED to 00 withdrawn to maximum limit.
ROD GROUP
________
________
________
INSERT ERROR
________
________
________
INSERT ERROR
________
________
________
WITHDRAW ERROR
________
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QUESTION 3.04 (1.00)
SELECT which one of the following best describes the operation /per-formance of an IRM during a reactor startup.
(1.00)
a. When the IRM is reading full scale on Range 10, the APRM's should be reading approximately 10% power.
,
b. Shifting from Range 4, indicating 75, to Range 5, will result in an indication of 24, on Range 5.
c. Reactivity feedback, due to the moderator temperature coefficient, should begin at approximately Range 5.
d.
When an IRM channel increases from 25 on range 2 to
.~5 on range 3, j
the indication has increased by one decade.
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QUESTION 3.05 (3.00)
LIST the six (6) ranges of reactor water level indication giving their span, calibration criteria, and whether they are used for RPS trip functions or not. (diagram acceptable if desired)
(3.00)
QUESTION 3.06 (3.00)
Use the attached ADS logic and control schematics, if necessary, to answer the following questions:
a. Where are the ' Logic Status Lights * physically located?
(0.50)
b.
Assume you are told that for all the ADS valves, only one logic status light per valve-per logic channel was on, and.all others were off.
Explain what is occuring. Include in your answer the status of the ADS logics and whether the valves are open or closed.
(1.00)
c. Other than ADS system malfunctions, under what conditions will the ADS valves close FOLLOWING a valid automatic initiation?
(1.00)
d. TRUE or FAl.SE: With RHR pump
'D'
operating, depressing both
' ADS D MAN INIT' and ' ADS H MAN INIT' buttons will initiate ADS.
(0.50)
QUESTION 3.07 (3.00)
Explain what effects the following malfunctions would have on the reactor vessel water level indication.(increase or decrease and why)
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a.
The excess flow check valve in the high pressure line of a level detector is shut and reactor pressure is INCREASING.
(1.00)
b. The excess flow check valve in the low pressure line of a level detector is shut and actual reactor water level is DECREASING.
(1.00)
c. An accident condition exists whereby actual reactor water level is below the lower instrument tap (variable les tap) and the reference (1.00)
les is flashing.
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QUESTION 3.08 (2.00)
Indicate whether the following statements are TRUE or FALSEt c. Recirculation loop flow in loop A ir 35%, in loop B it is 50%. This condition will NOT cause a flow comparatcr trip.
(0.50)
b. While in RUN, transfer of an APRM channel mode switch out of ' OPERATE'
will ALWAYS cause an inop alarm, rod block, and half scram (assume all circuitry functioning properly).
(0.50)
c. With an APRM channel meter expand switch in the ' REVERSE * position, the meter scale is expanded by a factor of 10.
(0.50)
d. While in RUN, with IRM channel A bypassed, APRM channel A fails downscale. This condition should NOT cause a half scram.
(0.50)
QUESTION 3.09 (2.00)
Explain three methods by which the service water pumps may be cperated from OUTSIDE the control room. Be sure to include any interlocks which must be met, which pumps can be operated, and at what seneral locations.
(2.00)
GUESTION 3.10 (1.50)
The recirculation system will respond to selected plant conditions by automatically reducing the recirculation M/G set generator output. From the list of possible conditions below, identify each runback signal by designating a
'F'
or
"I" beside each conditioni an
"F" signifies a full runback signal, an
"I" signifies an intermediate runback signal.
(1.50)
a._____ Secondary condensate pump trip with FW> 85%
b._____ Primary condensate pump trip with FW> 85%
c._____ RFPT trip (3 running) with RPV level less than 30"
.
d._____ Feedwater flow at 18% for 20 seconds e._____ Loss of stator coolin3 water f..____ Loss of cire water pump (3 running) and condenser pressure 4.5'hs (mmmmm CATEGORY 03 CONTINUED ON NEXT PAGE munum)
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INSTRUMENTS AND CONTROLS
QUESTION 3 11 (2.00)
Answer the following questions concerning the 1E 4.16KV system.
c. The diesel generator is operating and its associated breaker closed in parallel with the offsite source. Briefly explain how load changes on the diesel are made.
(0.50)
b. Following a loss of offsite power and its subsequent returne it becomes necessary to restore the 1E 4.16KV bus from the diesel generator breaker to the normal supply breaker. Procedure OP-SO.PB-001 instructs the operator to place the normal feed breaker syne-switch in the ON position and, ' synchronize across and close the normal feed breaker *. Briefly describe the conditions for synchronization for TilIS SPECIFIC CASE. Your answer should include:
1. The required indications of the synchroscope prior to closing the breaker.
2. The relative voltages / frequencies of the 1E bus and the normal feeder bus (i.e. which is higher / faster).
3. The expected response of the synchroscope as speed of the diesel(2.00)
is increased prior to closing the breaker.
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I
!
!
.
.
4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
~~~~ 5656L55fCAt E5NTRUL'~~~~~~~~~~~~~~~~~~~~~~~
-
R
QUESTION 4.01 (2.00)
Procedure OP-EO.ZZ-102 (Containment Control) directs the operator to
" runback recirc and manually scram * if suppression pool temperature cannot be maintained below 110 F or if an SRV has been stuck open for Sreater than 2 minutes.
(0.50)
c. Why is recire runback prior to scram?
b. 110 F was chosen because, among other reasons, it is the ' Boron Injection Temperature.' Explain what is meant by this term.
(1 50)
GUESTION 4.02 (2.25)
During operation at power, a feedwater malfunction occurs causing water level to increase rapidly. According to OP-AB.ZZ-117 ' Reactor High Level" procedure:
a. At what level must the operator close the MSIV's, terminate all injection into the RPV, and verify the reactor has scrammed?
(0.25)
b. Assuming level exceeds 118 inches, as indicated on the upset range, what additional problem could exist? What alarms might annunciate as a result of this problem? What additional precautions must the operator take as a result of this problem?
(2.00)
QUESTION 4.03 (2.25)
a.
What is rated thermal power for Hope Creek?
(0.25)
b. According to OP-IO.ZZ-006, " Power Changes During Operation *, rated thermal power may be excegded under certain conditions. What are these (1.00)
conditions?
c. According to the same procedure, what checks must be performed followins a thermal power change exceeding 15% of rated?
.
(1.00)
(xxxxx CATEGORY 04 CONTINUED ON NEXT PAGE xxxxx)
,
.
.
4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
~~~~ I65dEdG5dAE~CdA5 RUE ~~~~~~~~~~~~~~~~~~~~~~~~
R
____________________
GUESTION 4.04 (2.00)
What are all the entry conditions for the Reactor / Pressure Vessel Control Procedure, OP-EO.ZZ-1017 (2.00)
GUESTION 4.05 (1.25)
Authorization for any personnel to receive a radiation dose greater than regulatory limits is considered an Emergency Exposure Authorization.
a. What are the recommended upper limits for emergency exposure to save station equipment and to save a life?
(0.50)
b. What requirements must be met prior to authorizing an Emergency (0.75)
Exposure?
QUESTION 4.06 (3.00)
Answer the following questions regarding procedure OP-IO-ZZ-003,
"Starup From Cold Shutdown to Rated Power":
During a reactor startup, the procedure requires that the a. SRM detectors be withdrawn to maintain count rate between 10ee2 and 10ee5, however, only 2 detectors may be withdrawn at any one time.
(1.00)
What is the purpose of this limitation?
b. During a reactor heatup/ pressurization, reactor pressure increases above the pressure setpoint AND no turbine bypass valves are open.
According to the procedure, what certain precaution must be taken prior to initiatin3 corrective action and why?
(1.00)
c. 'During low flow conditions, feedwater flow to the reactor should be maintained relatively conctant..." Explain why this caution exists and what action the procedure recommends to assist the operttor $n (1.00)
acheiving a steady feedwater flow.
.
QUESTION 4.07 (1.25)
STATE the immediate operator actions for a failed open or stuck open (1.25)
relief valve.
i
!
(xxxxx CATEGORY 04 CONTINUED ON NEXT PAGE xmax*)
.
4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
R565 LUU5EAL"5UUTRbL'~~~~~~~~~~~~~~~~~~~~~~~
~~~~
____________________
QUESTION 4.08 (3.00)
The purpose of OP-EO.ZZ-202, " Emergency Depressurization," is to rapidly depressurize the reactor pressure vessel. In general, why might it be necessary to rapidly depressurize the reactor vessel? (3 reasons (3.00)
required)
QUESTION 4.09 (1.00)
c. According to OP-SO.SB-001, 'RPS system", why is it necessary to reset a full scram a soon as possible?
(0.50)
b. What precaution must be taken prior to transfer of an RPS bus power supply and why must this precaution be taken?
(0.50)
QUESTION 4.10 (3.00)
According to the E0P's, three viable methods of accomplishing adequate core cooling exist. State these three methods in order of preference, briefly describe how each is accomplished, and how adequate core cooling can be verified durin3 use of each method (if applicable).
(3.00)
QUESTION 4.11 (1.00)
When operating one loop of the RHR system in the Shutdown Cooling Mode, and a loss of flow occurs, EXPLAIN WHY reactor level should be-raised to 100" on the Shutdown Range.
(1.00)
t.
(***** CATEGORY 04 CONTINUED OH NEXT PAGE xxxxx)
,
!
l I
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-
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.
_..._
_
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i
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4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
RA5i5L55iEEt E5 sir 5L
-
~~~~
____________________
QUESTION 4.12 (3.00)
During refueling the following alarms occur!
FUEL POOL LEVEL HI/LO FUEL POOL COOLING SYS LEAKAGE HI FUEL POOL COOLING SYS TROUBLE a. List four automatic actions which would occur (assume pool level continues to decrease)
(2.00)
b. In addition to ensuring that all appropriate automatic actions are completed, what immediate operator actions are necessary?
(1.00)
!
(xxxxx END OF CATEGORY 04 xxxxx)
(xxxxxxxxxxxxx END OF EXAMINATION xxxxxxxxxxxxxxx)
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Saturated Steam: Temperature Table Abs Press.
5pecific Volume Enthalpy Entropy
~
~
Temp lb per Sat.
Sat.
Sat.
Sat.
Sat.
Sat.
Temp Iabr SqIn.
Liquid Evap Vapor Lic vid Evap Vapor Liquid Evap Vapor Fahr s
t vs if h ig hg s,
sgg g
t p
v g _vit 32 8 0 08859 0 016022 33047 33047 0 0179 1075.5 1075.5 0.0000 2.1873 2.1973 32.9 34 0 0 09600 0 016021 3061 9 3061.9 1.996 1074.4 1076 4 0.0041 2.1762 2.1802 34.9 36 0 0 10395 0 016020 2839 0 2839 0 4.008 1073.2 1017.2 0 0081 2.1651 2.1132 36.8 38 I O11749 0 016019 761.41 7634 2 6.018 1072.1 1078.1 0 0122 2.1541 2.1663 30.0 49 8 1.12163 0 016019 2445 8 2445 8 8.027 1071 0 1079.0 0.0162 2.I432 2.1594 40.9 42 0 0 13143 0 016019 2272 4 / 2272.4 10 035 1069 8 1079.9 0 0202 2.1325 2.1527 42.0
0 14192 0 016019 2112 8 2112 8 12.041 10681 10801 0 0242 2 1217 2.1459 44.0 48 0 0 15314 0 016020 19657 19651 14.047 10676 1081.6 0 0282 2.1111 2.1393 46.9 44 I
-
i 48 9 0 16514 0 016071 IR10 0 1810 0 16 051 1066 4 1082.5 0 0321 2.1006 2.1327 48.9 I
59 9 0 17796 0 016023 1704 8 1704 8 18 054 1065.3 1083.4 0.0361 2.0901 2.1262 50.0 52 8 019165 0 016024 1589 2 1589 2 20 057 1064.2 1084.2 0.0400 2.0798 2.1197 52.0
54 I O20625 0 016026 1482 4 I482 4 22.058 1063.1 1085.1 0.0439 2.0695 2.1134 54.0 50 0 0 22183 0 016028 1383 6 1383 6 24 059 1061.9 1086 0 0 0478 2.0593 2.1070 56.0 58 8 023843 0 016031 1797 7 1292 2 26 060 1060 8 1086.9 0 0516 2.0491 2.1006 50 0 80 5 025611 0016033 1207.6 1207 8 28.060 10591 10871 0.0555 2.0391 2.0946 88.9 52 0 0 27494 0 016036 11292 1129 2 30 059 1058.5 1088 6 0.0593 2.0291 2.0885 62.0 MI 029497 0 016039 1056 5 1056 5 32.058 1057.4 1089.5 0.0632 2.0192 2 0824 H.S 84 0 0 31626 0016043 989 0 989.1 34 056 1056.3 1090.4 0.0670 2.0094 2.0764 68 0 88 9 031889 0 016046 926 5 926 5 36.054 1055.2 1091.2 0.0708 1.99 %
2.0704 68 8 j
fil 0 36292 0 016050 868 3 868.4 38.052 1054.0 1092.1 0.0745 1.9900 2.0645 10.0 12 0 0 38844 0 016054 814 3 814 3 40 049 1052.9 1093.0 0.0783 1.9804 2.0587 72.5
'
l 74 8 0 41550 0 016058 764 I 764 I 42 046 1051.8 1093.8 0.0821 1.9708 2.0529 14.9
'
TII 044420 0 016063 717 4 717 4 44 043 10501 10941 0.0858 1.%I4 f.0472 76.9 is t 0 41461 n ul6067 671 R 613 9 46.040 1049.5 1095.6 0 0895 1.9520 2.0415 78.9 00 0 0 50683 0 016072 633 3 633 3 48.037 1048.4 1096.4 0 0932 1.9426 2.0959 88.8 82 5 0 54093 0 016077 595 5 595 5 50 033 10473 1097.3 0 0969 1.9334 2.0303 82.0 M9 057102 0 016082 560 3 560 3 52.029 1046.1 1098 2 0.1006 1.9242 2.0248 M.O 88 0 0 61518 0 016087 227 5 527 5 54 026 1045 0 1099 0 0.1043 1.9151 2.0193 86.0 88 0 0 65551 0 016093 4968 4%8 56 072 1043 9 1099.9 0.1079 1.9060 2.0139 88.0 90 0 0 69813 0 016099 4681 468 1 58 018 1042.7 1100 8 0 1115 1.8970 2.0086 90 0 92 8 074313 0 016105 441 3 441 3 60 014 1041 6 1101 6 0.1152 1.8881 2.0033 97.9 MI O19062 0 016111 416 3 4163 62 010 1040 5 1102 5 01188 1.8792 1.9980 MO
>
98 I O84072 0 016117 392 8 392 9 64 006 1039 3 1103 3 0 1724 1.8704 1.9928 96.9 L
28 0 0 89156 0 016173 310 9 370 9 66 003 10382 1104 2 0 1260 1.8617 1.9876 90 0
_ _ _ - _ _ _ _ _ _ _ _ _ _ _ _
.
.
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Specific volume Enthalpy
[ntropy Temp lb per Sal Sal.
Sal.
Sat.
Set.
Sal.
Temp
Iah Sq in liquiil Ivan Vapor li uld fysp Vapor liquid Evap Vapor Fahr
-
I p
v, veg vg i
h ig h
si sig sg t
100 8 0 94924 0016130 350 4 350 4 67.999 1037.1 11051 0.1295 1.8530 1.9825 188.9 lef t 100789 0 016131 331 1 331 1 69 995 1035 9 1105.9 0.1331 1.8444 1.9775 182.0 184 8 1 06 % 5 0 016144 3131 313 1 71.992 1034 8 1106 8 0.1366 1.8358 1.9725 1RO 106 8 11347 0 016151 296 16 29618 73 99 1033 6 1107.6 0.1402 18273 1.9675 108.0 196 8 12030 0 016158 2R0 28 780 30 7598 1032.5 1108.5 0.1437 1 8188 1.9626 100.0 110 0 1.2750 0 016165 26537 265 39 77.98 1031.4 1109.3 01472 1 8105 1.9577 120 112 8 13505 0 016173 251 37 251 38 79 98 1030.2 1110 2 01501 18021 1.9528 112.0
114 I l4299 0 016180 238 21 23822 81.97 10291 1111.0 0.1542 13938 1.9480 114.0 IlsI 15133 0 016188 225 84 225 85 83 97 1027 9 1111.9 01577 11856 1.9433 118.0 j
113 e 16009 0 016146 21470 / 214 21 85.97 1026.8 11123 0.1611 13774 1.9386 118.8
-
128 9 18927 0 016204 203 25 203 26 87.97 1025 6 l113.6 0.1646 13693 1.9339 1N.8 1228 1 7891 0 016213 192 94 192 95 89.96 1024 5 1114.4 0.1680 13613 1.9293 122.8 i
'
124 8 18901 0 016221 18323 18324 91 %
1023.3 1115.3 01715 1.1533 1.9247 124.5 126 I I9959 0 016229 114 08 114 09 93 96 1022.2 1116.1 0.1749 11453 1.9202 IN.9
,
128 8 2.1068 0016738 16545 16547 95 96 1021.0 1117.0 0 1783 13374 1.9157 IN.O
'
138 O 2 2230 0 016247 157 32 15733 97.96 1019.8 1117.8 01817 11295 1.9112 138.8 132 8 2.3445 0 016256 14964 14966 99 95 1018 7 1118 6 0.1851 13217 1.9068 132.0
'
1348 2 4111 0 016265 14240 14241 101.95 1011.5 1119.5 0.1884 13140 1.9024 Int 13t I 2 6047 0 016214 135 55 135 57 103 95 1016 4 1120 3 0.1918 13063 I.8980 1ES 118 8 27438 0 0167R4 17909 129 11 105 95 10152 1121.1 0.1951 1 6986 1.8937 130.0 149 8 28092 0 016293 122 98 123 00 107.95 1014 0 1122.0 0.1985 1.6910 1.8895 1488 142 8 3 04 tl 0 016303
!!12I 11722 109.95 1012.9 1122.8 0 2018 1.6534 1.8852 142.0 144 I 3 1997 0 016312 Ill14 111 16 Ill 95 10113 1123 6 02051 1.6759 1.0810 144.s 140 8 3 3653 0 016322 106 58 106 59 113 95 1010.5 1124.5 02084 1.6684 1.0769 148.0 14e e 3 MRI 0016317 101 BR 101 10 115 95 10092 1125.3 0.2117 1.6610 1.8727 148 8 190 0 3 7184 0 016343 9705 9707 117.95 1008 2 1126.1 0.2150 1.6536 1.8686 198.0 1528 3 9065 0 016353 9266 9268 119.95 1007.0 1126.9 0.2183 1.6463.l.8646 152.0 1548 4 1025 0 016363 8850 88 52 121 95 1005.8 1127.7 02216 1.6390 1.8606 1540 156 I 4 3068 0 016374 84 56 8457 123 95 1004 6 1128 6 0 2248 1.6318 1.8566 158 9 151 8 4 5197 0 0161R4 R0 R7 R083 125 96 1003.4 1129.4 0 2281 1.6245 1.8526 158.5 10e 8 4 7414 0 016395 7727 1729 127 96 1002.2 1130.2 0.2313 1.6174 1.8487 108.8 182 0 4 9722 0 016406 13 90 13 92 129 96 1001.0 1131.0 0.2345 16103 18448 182.0 1648 5 2124 0 016417 70 70 70 72 131.96 999 8 1131.8 0 2377 1.6032 1 8409 1Kt les I 54623 0 016428 6767 6768 133 97 998 6 1132.6 0 2409 15%I 1.8371 100 O I88 I 5 1723 0 016440 6418 6480 13597 997.4 1133 4 0.2441 1.5892 I.8333 IKO 179 8 5 9926 0 016451 62 04 62 06 137.97 996.2 1134.2 0.2473 1.5822 1.8295 1M8 l
172 8 6 2736 00lR463 59 43 5945 139 98 995 0 1135.0 0 2505 1.5753 1.8258 172.0 174 8 6 5656 0 016414 5695 56 97 141.98 9938 1135.8 0 2531 1.5684 1.8221 174.8
'
litt 6 8690 0 016486 54 59 54 61 143 99 992.6 1136 6 0.2568 1.5616 1.8184 178.9
III 8 7.1840 0 016498 5735 57 36 145 99 991.4 1137.4 0.2600 1.5548 1.8147 178.8 O
~.
-
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.
n q
q
.
.
.:
-
.-
Abs Press Speofic Volume
[nthalpy Entropy Temp lb pei Sal Sal Sal.
Sat.
Sal.
Sat Temp Iahr SqIn.
Iiqmrt Ivan vapor Liquid Evap vapor Liquid Evan Vapor Fahr he h ig h
si sig st I
g va I
p vi veg let s 75110 0 016510 5021 5022 148 00 990 2 1138 2 02631 1.5480 1.8111 lett 132 3 7850 0 016522 48 112 18 189 150 01 989 0 1839 0 0.2662 1.5413 18075 102 0 104 8 8203 0 016534 46 232 46 249 152 01 9878 1139 8 0.2694 1.5346 I8040 184 8 Its t 8 568 0 016547 44 383 44 400 154 02 986 5 1140 5 0 2725 15219 1.8004 let#
Ige s 8 947 0 016559 47 671 42 638 156 03 985 3 1141.3 0 2756 1 5213 1.7%9 ING 198 8 9 340 0 016512 40 94 /
40 957 158 04 984.1 1142.1 02787 1.5148 1.7934 190.0 192 9 9 147 0 016585 39 331 39 354 160 05 982.8 1142 9 0 2818 1.5082 1.7900 192.0 194 0 10 168 0 016598 37 808 37 824 162 05 981 6 1143 7 012848 1.5017 1.7865 194.0 198 9 10 605 0 016611 36 348 36 364 164 06 980.4 1144 4 0.2879 1.4952 1.7831 196.0 IIII 11058 0 016624 34 954 34 970 166 08 979 1 1145 2 0.2910 1.4888 1.7798 194 0 2$4 8 11 526 0 01R637 31 622 33639 168 09 977.9 1146 0 02940 1.4874 1 7764 299.8 294 8 12 512 0 016664 31135 31151 172 11 975 4 1147 5 0 3001 1.4697 17698 294 0 298 8 13 568 0 016691 28 862 28 878 176 14 9778 1149 0 0 3061 1.4511 1.7632 200 0 212 8 14 696 0016719 26 182 26 199 18017 970 3 1150 5 0.3121 1.4447 11568 212 0 218 8 15 901 0 016747 74 R18 74894 184 20 9618 1152.0 0.3181 14323 11505 216.0 titI 17 186 0016715 23 131 23 148 188 23 965 2 1153.4 0324I I4201 11442'
220 0 224I 18 556 0 016805 21 529 il 545 19227 962f 1154.9 0.3300 1.4081 11380 224 e 220 8 20 015 0 016834 20 056 20 013 190 31 960 0 1156.3 0 3359 13%I 11320 220 8 232 I 21 567 0 016864 18101 18718 200 35 957.4 1157.8 0 3417 13842 17260 232.0 236 e 23 216 0 015895 11 454 17 All 204 40 954 8 1159 2 0 3476 13725 11201 235 0 240 0 24 968 0 018928 18 304 18 321 208 45 952.1 1160 6 0 3533 13609 17142 240.0 244I 26 876 0 016958 15 243 15 260 212 50 949 5 1162.0 0 3591 1.3491 1 7085 244.0 240 0 28 196 0 016990 14 264 14 281 21656 946 8 1163.4 03649 I.3379A 17028 248 0 252 8 30 883 0 011012 13 358 13 375 220 62 944.1 1164 7 0 3706 13266 16972 2570 258 8 31 091 00110%
17 570 12 538 224 69 941.4 1166I 03163 1.3154 I6917 256 8 298 g 35 427 0 017089 11 145 11 762 228 76 938 6 1167.4 0 3819 1.30(3 16862 260.0 294 3 37 894 0 017123 11025 11 042 232 83 935 9 18687 0.3876 12933 16808 264.0 200 0 40 500 0 017157 10 358 10 375 236 91 9331 1170 0 0 3932 1.2823 1 6755 260 0 212 8 43 249 0 017193 9 738 9 755 240 99 930 3 1171 3 0 3987 1.2115 1 6702 272.0 218 8 46 147 0 017228 9 167 9 180 24508 927.5 1l72 5 04043 I2607 16650 276 8 200 0 49 200 0 017264 8 627 8 644 249.17 924 6 1173 8 0 4098 12501 16599 200.8 284 8 52 414 0 01130 8 1280 8 1453 253 3 9211 1875 0 0 4154 12395 I6548 284 8 200 9 55 795 0 01134 7 6634 76807 257.4 918 8 1l762 0 4208 I2290 1.6498 288 8 292 0 59 350 0 01738 12301 7 2415 261.5 915 9 1177.4 04263 1.2186 1 6449 292.0 298 8 63 064-0 01/41 6 8759 6 8433 265 6 9130 1178 6 0 4317 12082 I6400 2968
. _ _
.
.
,
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.
Abs Press.
~Sper.ilic Volum, Enthalpy Entropy
'~
Temp Lb per Sal Sat.
Sat.
SRt.
$st.
Sal.
Temp Fahr Sgin.
liquid Evap Vapor Liquid Evap Vapor Liquid Eysp Vapor Fatw v
hr hs h
sg seg s
I i
g a
i vig s
I p
V
.
300 0 67 005 0 01745 6 4483 6 4658 2693 910 0 11793 0 4372 1.1979 1.6351 389 9 304.8 71 119 0 01749 6 0955 6 1830 273 8 901.0 1180 9 0 4426 1.1877 1.6303 304 0 300 e 75 433 0 01753 5 1655 5 7830 278 0 904 0 1182.0 0 4419 1.1776 16256 300 e i
II2 8 79 953 0 01757 54566 5 4742 2821 901 0 11831 0 4533 1.1676 1.6209 312.0 l
318 8 84 688 0 01761 51613 5 1849 286 3 897.9 1184.1 04586 1.1576 1.616 316 0 3NG 09 643 0 01766 4 8961 4 9138 290 4 894 8 1185 2 04640 1.1477 1.6116 329.0 3245 94 826 0 01770 4 6418 4 6595 294 6 891.6 1886 2 0 4692 1.1378 1.6071 324.0 320 g 100 245 001774 4 4030 4 4208 298 7 888 5 1181.2 0.4745 1.1280 1.6025 320 0 j
3320 105 907 0 01779 4 1188 41%6 302 9 885 3 1188 2 0 4798 1.1183 1.5981 332.8
117.
0 01787
388 3 3 878
1 0 90
15 2
.0
-
3448 124 430 0 01192 3 5834 3 6013 315.5 875 5 1191.0 04954 1.0894 15849 344.9
!
MeI 131.143 0 01797 34078 3 4258 319 7 872.2 1191.1 0 5006 1.0199 1.5006 340 O i
352 0 130130 0 01801 3 2423 3 2603 323 9 868 9 l1923 0.5058 1.0705 1.5763 352.0 j
350 0 145 424 0 0lR06 3 OR63 3 1044 328I 865 5 1193 6 0 5110 10611 1.5121 356.9 300.8 151 610 0 01811 2 9392 29573 332 3 862.1 1194 4 0 5161 1.0517 1.5678 300.0 304 0 160 903 0 01816 2 8002 28184 336 5 858 6 1195 2 0 5212 1.0424 1.5637 3H.e 3000 169113 0 01821 2 6691 2 6813 340 8 8551 1195 9 0 5263 1.0332 1.5595 360 0 372 8 177648 0 01826 2 5451 2 5633 345 0 8516 11 %.1 0.5314 1.0240 1.5554 372.0 379 0 186 517 0 01811 2 4219 24462 349 3 8481 1197.4 0 5365 1.0148 1.5513 376.e 300 8 195 729 0 01836 23110 2 3353 353 6 844 5 1l98 0 0.5416 1.0057 1.5473 300 0 3M S 205294 0 01842 22120 2 2304 357.9 840 8 11987 0.5466 0.9966 1.5432 3H.0 300 0 215 220 0 01841 2 1126 2 1311 362.2 S31.2 1199 3 0 5516 0 9876 1.5392 380 B 382.0 225 516 0 03853 2 01M 20369 366.5 833.4 1199.9 05567 0.9786 1.5352 397.0
,
398 9 236193 001R58 1 929?
1 9477 370 8 829 7 1200 4 0 5617 0 96 %
1,5313 396 8 i
408 8 247 259 0 01964 I8444 I8630 375.1 825 9 1201.0 0.5667 0.9607 1.5274 400 9 404 9 258725 001870 11640 11827 379 4 822 0 1201.5 0 5717 09518 1.5234 4M.0
,
400 0 270 600 0 01875 1 6811 1 7064 383 8 818 2 1201.9 0 5766 0 9429 1.5195 480 0
<
412 9 282 894 0 01881 16152 16340 388.1 814.2 1202.4 0 5816 09341 1.5157 412.s elle 295 617 00lR87 1 5461 1 5651 392 5 810 2 1202.8 0.5866 0 9253 1.5118 elle
'
.
l 429 0 300180 001894 14808 14997 396 9 806 2 1203.1 0.5915 09165 1.5000 420 0 424 I 322 391 0 01900 14184 14314 4013 802 2 1203.5 0 5964 0.9017 1.5042 474.0 428 I 336 463 0 01906 13591 13782 4057 798 0 12037 0 6014 0 8990 1.5004 428 e 4320 351 00 0 01913 130266 132179 410 1 7939 1204 0 0 6063 0 8903 1.4966 432 e 4MI 366 03 0 01919 124RR7 126806 414 6 789 7 1204 2 0 6112 08816 1.4928 436 8 440 0 381 54 0 01926 119761 121687 419 0 785 4 1204 4 0 6161 0 8729 1.4890 440 0 4448 397 56 0 01933 1.14874 116806 423 5 781.1 1204 6 0 6210 0 8643 14853 4448 4400 414 09 0 01940 1.10212 112152 428 0 176 7 12043 0 6259 0 8557 1.4815 448 0
<* * * O 43114 0 01941 105164 101111
'12.5 772 3 1204 8 06308 0.8411 1.4778 (.de
_ _ _ _ _ _ _ _ _ _
_ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _.
_ _ _
__
)
-
i
.
Ahs Press Specifer Volume Enthalpy Entropy
~
Temp lb per Sal Sat Sat.
Sat.
Sat.
Sal.
Temp Fahr Sq in l iquirl Ivan Vapor liquid Evap Vapor liquid Evap Vapor Fahr I
p vg v ig,
vg hg h lg h
5:
sig sg I
g 4st t 466 87 0 01961 0 97463 0 99424 441.5 763.2 1204 8 06405 0 8299 1.4704 400.8 464 8 485 56 001969 0935M8 0 95557 446.1 758 6 12043 0 6454 0 8213 1.4667 464.8 est t 504 83 0 01976 0A9885 0 91862 4501 754 0 1204.6 0 6502 0 8127 1.4629 468 0 472 8 524 67 0019R4 0 86145 DM8329 455 2 749 3 1204.5 0.6551 0 8042 1.4592 472.0 als e 54511 00l997 0 R/%R 0R4950 459.9 744 5 1204 3 0 6599 0 7956 1.4555 476 9 480 0 568 15 0 02000 0 79716 / 0 81717 464.5 139 6 1204.1 0 6648 03871 1 4518 40s.8 414 8 587 81 0 02009 076613 0 78622 4691 1343 1203 8 0 66 %
03785 1.4481 484.9 400 I 610 10 007011 0 13641 0 75658 473 8 7293 1203.5 0.6745 03700 1.4444 480.0
'
4t2 8 633 03 0 07026 0 10191 012820 418 5 724 6 1203.1 0 6193 0 7614 1.4407 492.0 498 8 656 61 0 07014 0 6R065 070100 AR3 2 719 5 12023 0 6842 03528 14370 496.0 500 0 680 86 0 02043 0 65448 0 67492 487.9 714 3 1202.2 0.6890 03443 1.4333 500.0 584 8 105 18 0 07053 052938 0 64991 4923 709 0 12013 0 6939 03357 1 42 %
504.9 500 8 731 40 0 02062 0 60530 0 67592 497.5 7037 1201.1 0 6987 01271 1.4258 500.0 512 8 75732 0 02012 058118 060289 502.3 698 2 1200.5 03036 0 7185 1.4221 512.8 510 0 7A4 76 0 07081 0 % 997 0 58079 507.1 6923 II99.8 0 7085 0 7099 1.4183 516.9 529 I 81253 0 02091 0 53864 055956 512.0 6870 1199 0 01133 03013 1.4146 529.9 524 8 84104 0 02102 05!814 0 53916 516 9 681.3 1198.2 03182 0 6926 1.4108 524.8 520 8 870 3I 0 02112 0 49843 0 51955 5218 675 5 1197.3 0 7231 0.6839 1.4070 528 9 532 8 900 34 0 02123 0 47947 0 50070 526 8 669 6 1196 4 0 7280 0 6752 1.4032 532.0 53E O 931.17 0 07134 0 46173 0 48757 5313 663 6 1195.4 07329 0 6665 1.3993 536.0 540 0 962 79 0 02146 644367 0 46513 536 8 657.5 1194.3 0 7378 0 6577 I.3954 548.9 544.8 995 22 0 02157 0 47677 0 44834 541.8 651 3 1193.1 03427 0.6489 1.3915 544.9 540 8 1028 49 0 02169 0 41048 0 43217 546 9 6450 1191.9 0 7476 0.6400 3.3876 548.8 i
5529 1062 59 0 02182 039419 0 41660 552 0 638.5 1190 6 0 7525 0.631I l.3837 552.0 558 8 1091 55 0 07194 0 11966 0 40160 557.2 632.0 1189 2 03575 0 6222 1.3797 556.9 500 0 1133 38 0 02207 0 36507 0 38714 5624 625 3 1187.7 03625 0.6132 1.3757 508 9 544 8 1170 10 0 02221 0 35099 0 31320 567.6 618 5 1186.1 03674 0 6041 1.3716 5640
,
l Ste t 120772 0 02235 0 33741 0 35975 5729 611 5 1184 5 03725 0.5950 1.3675 548 8
,
572 8 1246 26 0 07249 0 37429 0 34678 578 3 604 5 11821 0 1775 0 5859 1.3634 572.0 576 0 1285 74 0 07764 031162 0 33476 583 7 597.2 1180.9 03825 0 5766 1.3592 578.0 900 0 1326.17 0 02279 029937 0 32216 589.1 589.9 1179 0 0 7876 0.5673 1.3550 588.8 584 9 1367 7 0 02295 0 28753 0 31048 594 6 SR2.4 1176 9 01927 05580 1.3507 564.0 500.8 1410 0 0 02311 0 27608 029919 6001 5743 1174 8 03978 0.5485 1.3464 508.9 e
592.0 1453 3 0 02328 0 26499 0 28827 6053 566 8 1172.6 0 8030 0.5390 1.3420 592.9
l.
598 9 14978 0 07345 0 75425 0 27770 611.4 558.8 1870.2 08082 0.5793 1.3315 595 0
"
<
.
O e
.
,
.
.-
Y m
'
i Ahs Press.
Specific Volume Enthalpy Entropy
~~~
.
Temp tb pe:
Sal.
Sal.
Sol.
Sal.
Sal.
Sal.
Temp Fah Sq in iIquid Fvap Vapor Li vid Evep Yapor liquid Evap Vapor Fahr J
v ig_,, _j 8 f
h is h
s, sg, s,
g I
p V
g OOO O 1543 2 0 02364 024384 0 28747 817.1 550 6 1167.7 0 8134 0.5196 1.3330 555.5 884 I 15897 0 02382 0 23374 0 25757 622.9 542.2 1165.1 0.8187 0.5097 1.3284 884.0 898 8 16373 0 02402 0 22194 0 24796 6288 533 6 1162.4 0 8240 0.4997 1.3238 600.8 812 8 16861 0 02422 0 21442 0 23865 634 8 5241 1159.5 0 8294 0.4896 1.3190 812.0 818 5 1735 9 0 07444 0 70516 0 22960 6408 515.6 1156.4 0.8348 0.4794 1.3141 816.0 0 1961 ! 0 22081 646.9 506.3 1153.2 0 0403 0.4689 1.3092 SN.9 Sit t,
1786 9 0 02466 824.5 1839 0 0 02489 0 18737 021226 653.1 4?6.6 1149.8 0.8458 04583 1.3041 824.0 1892 4 0 02514 0178R0 0 20394 659.5 4863 1146.1 0 8514 0.4474 1.2988 820.0 828 8 832 8 19470 0 02539 0 17044 0 19583 665.9 476 4 1142.2 0.8571 0.4364 1.2934 832.0
-
538 8 2007 8 0 07566 016726 018792 672.4 4651 1138.1 0 SE28 0 4251 1.2879 636.9
,
'
848 8 2059 9 0 02595 015427 019021 679.I 454.6 11333 0 8686 0.4134 1.2821 880.0 944 8 2118 3 0 02625 0 14644 017269 685 9 443.1 1129.0 0 8746 0.4015 1.2761 844.8 548 8 2178 1 0 02657 0 13876 0 16534 692 9 431.1 1124 0 0.8806 0.3893 1.2699 548.0 852 8 2239 2 0 02691 0 13124 0 15816 700 0 4181 11187 0 8868 03767 1.2634 552.8 ISe 9 23011 00272R 012187 0 15115 701.4 405 7 1113.1 0 8931 0.3637 1.2567 856.0 000 8 2365 7 0 02768 0 11663 0 14431 714.9 392.1 1107.0 0.8995 0.3502 1.2498 900.0 964 8 2431,1 0 02811 0 10947 013157 722.9 3773 1100.6 0.9064 0.3361 1.2425 064.8 568 8 2498I 002858 010229 0 13087 731.5 362.1 1093.5 0.9137 0 3210 1.2347 tes.e 872 0 2566 6 0 02911 0 09514 0 12424 740 2 3451 1085 9 0.9212 0.3054 1.2266 572.0 175 0 26368 0 02910 0 08799 0 11769 7492 328.5 1077.6 0 9287 0.2892 1.2179 878.8 SM $
2708 8 0 03037 0 08080 0 11117 758.5 310.1 1068.5 0 9365 0.2720 1.2006 0I0.9 Sle t 27821 0 03114 0 07349 0 10463 768 2 290 2 1058.4 0 9447 0.2537', 1.1984 884.8 808 8 28574 0 03204 0 06595 0 09799 778 8 268.2 1047.0 0 9535 0.2337 1.1872 500.5 i
892.0 2934 5 0 03313 0 05797 0 09110 790 5 243.1 1033.6 0.9634 0.2110 1.1744 802.8 598.9 30134 0 03455 0 04916 0 08371 804 4 212.8 1017.2 0 9749 0.1841 1.1591 OM.8
'
790 0 3094 3 0 03662 0 03857 0 07519 822 4 172.7 995.2 0.9901 0.1490 1.1390 MI.0 702 3 3135 5 0 03824 0 03173 0 06997 835 0 1441 9791 1.0006 0.1246 1.1252 782.0 FM I 3171.2 0 04100 0 02192 0 06300 854.2 102.0 956.2 1.0169 0 0816 1.1046 794.0 795 9 3198 3 0 04427 0 01304 0 05730 873 0 61.4 934.4 1.0329 0.0527 1.0856 7e5.0 70547*
37082 0 05078 0 00000 0 05078 906 0
906 0 1.0612 0 0000 1.0612 705.47'
.
l
- f'ritical lemneralute
'
.--
_
_.__
..
-_.
.%
,
.
!
.o
.
.-
l Table 2:
Saturated Steam: Pressure Table
.... - -. - _
... Specific Volume Enthalpy Entropy Abs Press.
Temp Sal Sat.
Sat.
Sat.
Sat.
Sat.
. Abs Press.
th/Sg in.
Fahr liquid Ivan Vapor liquid Evap Vapor liquid Evap Vapor tb/Sg in.
v, v
hg hgg h
s, s gg s
p p
t vi I
i g
g g
190005 32 018 0 016022 3302 4 3302 4 0.0003 1075 5 1075 5 0 0000 2.1872 2.1872 0.00005
,
3 25 59 323 0 016032 1235 5 1235 5 27382 1060.1 1087.4 0 0542 2.0425 2.0967 8.25
'
5 59 79 586 0 016011 641 5 641.5 47.623 1048.6 10 % 3 0 0925 1.9446 2.0370 0.50
10114 0 016136 331 59 333 60 6913 1036I 1105.8 0.1326 1.8455 1.9781 1.8
16224 0 016407 13 515 73 532 130 20 1000 9 1131.1 0 2349 1.6094 1.8443
ft8 193 21 0016592 38404 38 420 161.26 982.1 1143.3 0.2836 1.5043 1.1879 10.8 14 990 212 00 0 016119 26782
/26 199 180.17 970 3 1150 5 0.3121 1.4447 1.7568 14.698
15 e,
213 03 0016126 76214 26 290 181 21
%97 1150.9 0.3137 1.4415 1.7552 15.9 i
29 8 22796 0 016834 20 010 20 087 196 27 9601 1156.3 0 3358 1.3962 1.7320 29.9
,
30 0 250 34 0 011009 13 1266 131436 218.9 945 2 1164.1 0.3682 1.3313 1.6995 30.9
4e e 267 25 0 017151 10 4194 10 4965 236I 933 6 1169.8 0.3921 1.2844 1.6765 48.9 50 0 281 02 0 017274 8 4967 8 5140 250 2 923 9 1174.1 0.4112 1.2474 1.6586 50.0
SG I 29211 0 017383 71562 7.1736 262.2 915 4 1177.6 0.4273 1.2167 1.6440 50 0
78 3 302 93 0 017482 6 1875 6 2050 272.7 907.8 1180 6 04411 1.1905 1.6316 79.8
'
08 I 312 04 0 011573 5 45.16 5 4711 282.1 900.9 1183.1 0.4534 1.1675 1.6208 OO O 99 8 320 28 0 017659 4 8719 4 8953 290 7 894 6 1185 3 0.4643 1.1470 1.6113 90 0 199 9 327 82 0 017740 4 4133 4 4310 298.5 888 6 1187.2 0.4743 1.1284 1.6027 188.5 118.8 334 19 0 01782 4 0306 4 0484 305 8 883.1 1188.9 0 4834 1.1115 1.5950 110.0 129 I 341 27 0 01763 3 1091 3 1275 312.6 877.8 1190 4 0.4919 1.0960 1.5819 120 0 130 0 347.33 0 017 %
3 4364 3 4544 319 0 872 8 1191.7 0.4998 1.0815 1.5813 13e e leg g 353 04 0 01803 3 D10 3 2190 325 0 868 0 1193.0 0 5071 1.0681 1.5752 148.8 150 0 358 43 001809 2 9958 3 0139 330 6 863 4 1194.1 0 5141 1.0554 1.5695 150.8 1st t 363 55 001815 28155 28336 336I 859 0 1195 1 0 5206 1.0435 1.5641 160 e lie a 368.42 0 01821 2 6556 2 6738 341.2 854.8 11 % 0 0 5269 10322 1.5591 17e e Ise 3 373 08 001827 2 5129 25312 346 2 850 7 1196 9 0 5328 1.0215 / 1.5543 100.0 fee s 37753 0 0lR13 2 3847 2 4030 350 9 846 7 1197.6 0 5384 1.0113 1.5498 190 8 298.9 391 90 001839 2 2689 2 2813 355 5 842 8 1198.3 0 5438 10016 1.5454 200.9 III e 385 91 0 01844 716313 218217 359 9 839I 1199 0 0 5490 0.9923 1.5413 219 9 228 8 389 88 0 01850 2 06179 208629 364 2 835 4 1199 6 0 5540 0 9834 1.5374 279.8 230 I 393 10 0 01855 I91991 199846 3683 831.8 1200.1 0 5588 0 9748 1.5336 239 9 248 8 39739 00l860 189909 191169 372.3 828 4 1200 6 0.5634 0.9665 1.5299 240 0 250 0 40097 0 01865 182452 II4317 376I 825 0 1201.1 0 5679 0 9585 15264 250 8 feeI 404 44 0 01810 115548 117418 379 9 821 6 1201.5 0 5722 09508 1.5230 260.9 2rs t 40780 0 01815 169131 171013 383 6 818 3 1201.9 05764 0 9433 15197 278 9
>
200 0 411 01 0 01880 161169 I65049 387.1 8151 1202.3 0.5805 0 9361 1.5166 200 8
&
200 3 414 25 0n18R5-157597 159482 390 6 812 0 1202.6 0 5844 0 9291 1.5135 290 0 300 0 417 35 001889 I52384 154274 394 0 808.9 1202.9 0 5882 0 9223 1.5105 300 0 356 8 431.73 001912 130642 1.32554 409 8 794 2 1204.0 0 6059 0 8909 1.4%8 350 0 age s 444 60 0n1934 1.14167 1.16095 424.2 780 4 1204.6 0 6217 0 8630 1.4847 400 5
___
.__
. _. _ _
_
__
'
-
l
.
.
...
Specific Volume Enthalpy Entropy Abs Press.
Temp Sal Sat.
Sat.
Sat.
Sat.
Sat.
Abs Press.
?
IblSq In.
Fahr Iirluiel ivan Vapot Lit uid Evap Vapor liquid Evap Vapor Lb/Sg in. -
.
i t hg h
sg s g, s,
p g
p i
v, v,,
y, g
g 458 O 456 28-001954 101224 I03119 437.3 7675 1204 8 06360 0.8378 1.4738 450 0 500 0 46701 0 01915 0 90181 0 92162 449.5 7551 12043 0 6490 0 8148 1.4639 Ses e 550 0 416 94 001994 0 81183 0 84117 460 9 743.3 1204.3 0 6611 03936 14547 550 0 800 0 48620 0 02013 0 14962 0 16915 4713 732 0 12033 0 6723 03738 1 4461 seg g 850 0 494 89 0 07032 0 68811 0 10843 4819 720 9 1202.8 0 6828 03552 14381 650.0 199 9 503 04 0 07050 0 63505 065556 491 6 110 2 1201.8 0 6928 33377 1.4304 fee g 750 9 510 84 0 07069 058880 0 60949 500.9 699 8 1200 7 03022 0 7210 1.4232 738 g seet 518 21 0 01081 0 54809 0 568 %
509 8 689 6 1l99 4 03111 0 7051 1.4163 300 g ISO I 525 24 0 02105 0 51191 0 53302 518.4 619 5 1198 0 03197 0 6899 1.4096 35e g 900 e 53195 00?!?3 0 41968 050091 526 7 669 7 11 % 4 03219 0.6753 1.4032 900 e 950 0 538 39 00?!41 0 45064 0 41205 534 7 660 0 11943 03358 06612 1.3970 950 0 Iges B 544 58 0 02159 0 42436 0/45 %
542 6 650 4 1192.9 0 7434 0 6476 1.3910 Iges e 1959 0 550 53 0 02111 0 40041 042224 550I 640 9 1191.0 0 7507 06344 1.3851 1950 e 1100 I t 556 28 0 02195 0 31863 0 40058 557.5 631.5 1189.1 0 7578 0.6216 1.3794 1100 0 1150 0 561 82 0 C"l4 0 35859 0 38013 564 8 622 2 1187.0 03647 06091 13738 1150.0
,
1790 9 567.19 0 0h.12 0 34013 0 36245 511.9 613 0 1184 8 03714 05%9 1.3683 1290 8
.
,
1798 9 572 38 0 02250 0 32306 034556 578 8 603 8 1182.6 01790 0.5850 1.3630 f2508 1300 9 577.42 0 02269 0 30722 0 32991 585 6 594 6 1180 2 03843 0 5733 1.3577 130st 13549 582 32 0 02288 0 29250 0 31537 592.3 585 4 1177 8 03906 0 5620 1.3525 13540 l
14M 8 58707 0 02301 0 21811 0 30118 5988 516 5 1175.3 03966 0.5507 1.3474 1490 8
,
j 14500 59170 0 02327 0 26584 0 28911 605 3 5674 1172.8 0 8026 0 5397 13423 1450 0 1500 I 596 20 0 02346 0 25372 0 21/19 6113 558 4 1870.1 08085
- 05288 1.3373 1590 0 15500 600 59 0 0?366 0 24235 0 26601 618 0 549 4 1167.4 0 8142 0 5182 1.3324 1558.9
)
1600 0 604 87 0 02387 0 23159 0 25545 624 2 540 3 1164.5 0.8199 0 5076 1.3274 1600.0
'
1958 0 609 05 0 02401 0 22143 0 24551 630 4 531.3 1161.6 0 8254 0 4911 13225 1650 9 l
tree s 613 13 0 07478 021178 023607 636.5 522.2 1158 6 0 8309 04867 1.3176 17000 l
1750 0
$1712 0 02450 0 20283 022713 642.5 513.1 1155.6 0 8363 0.4765 1.3128 1750 0 1000 5 02102 0 02412 0 19390 0 21861 648 5 503 8 l152.3 08417 0 4662 1.3079 1000 e 18580 424 83 0 02495 0 18558 021052 654 5 494 6 1149 0 0 8470 0.4561 1.3030 1850 t
'
1999 8 62856 0 02517 0 11161 0 20218 660.4 4852 1145.6 08522 0 4459 1.2981 1990 e itse e 632 22 0 02541 0 16999 0 19540 666 3 415 8 1142.0 0 8574 04358 A 1.2931 19580 2000 0 635 80 0 07565 0 16266 018831 672.1 466 2 1138 3 0 8625 0 4256 12881 2000 9 2l00 8 642 16 0 02615 0 14885 011501 683 8 446 7 14305 0 8727 04053 12180 210e e 2298 e 649 45 0 07669 0 13603 0 16212 695 5 426 7 1122.2 0 8828 0 3848 1.2675 2290 8
,
230s I 65589 0 0/121 0 12406 0 15133 707 2 406 0 1113 2 0 8929 0 3640
!.25a 23009 feet S 66711 0 0/190 0ll?R1 0 14016 719 0 384.8 1I031 0 9031 0 3430 1.2460 24000 j
2900 9 66811 002959 010209 013068 7313 361 6 1093 3 0 9139 0 3206 1.2345 2500 0 i
2000 5 673 91 0 02938 0 09112 012110 744 5 3376 1082.0 0 9247 0 2977 1 2225 2500 8
2100 3 619 53 0 03029 0 08165 011194 157 3 312 3 10693 0 9356 0.2741 1.2097 2700 9 2000e 684 %
0 03134 0 01111 0 10305 710 7 2851 1055 8 0 9468 0 2491 1.1958 2000.0 2900 0 690 22 0 03262 0 06150 0 09420 785 1 2547 1039 8 0 9588 0.2215 1.1803 2900 0 3000 I 695 33 0 03428 0 05013 0 08500 801.8 218 4 1020.3 0 9728 01891 1.1619 3000 0 3100 0 100 28 0 01681 0 03171 0 01452 824 0 169 3 993.3 0 9914 01460 1.1373 3180.8
)
3200 8 705 08 0 04412 0 01191 0 05663 875 5 56.1 931.6 1.0351 0 0482 1.0832 3200 0 l
3700 2'
70541 0Os018 0 00000 005018 906 0
9060 106I2 0 0000 1.0612 3200.2*
l I
f%
,
-.
<
MASTER PAGE
1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,
--- iAERR559RARiCE-sEAi iEAssFEE AR5 FEUi5 FEBR
____________________________________________
ANSWERS -- HOPE CREEK-86/02/24-CRESCENZO, F.
ANSWER 1.01 (2.50)
the o. As the reactor operates during the early part of the cycle, burnable poison depletes more rapidly than the fuel, therefore, control rods must be inserted to hold the power constant.
(1.00)
b. As the control rod density increases, the power producing regions of the core become more undermoderated; the moderator to fuel ratio is decreasing. In effect, as total power production has remained constant but the power producing volume has become smaller, the Because of this operating volume of the ccste has become undermoderated.
(1.50)
effect, the void coefficies.t becomes more negati
~ l e.
~Ti m < d
+a WJ\\
c-
'
' ont Stl(
REFERENCE NMPC Operations Technology vol. I page I-12-6 Hope Creek Lp RXPH19-01 pg.6 LO #4, RXPH28-01 pg.6 and trans. 41 LO #3 RXPH16-01 trans. 42 ANSWER 1.02 (2.00)
a. The half life of samarium is greater than 10 e16 years so it can be considered stable. Because of this, removal of samarium is accomplished by neutron absorption only. Since the production and removal of samarium are functions of neutron flux, the term for neutron flux may be cancelled from the equation for equilibrium samarium thus makin3 it non-dependent on flux / power level.
(1.00)
b. The greatest changes in samarium concentrationc occur durin initial startup. 4i. EUL, 1". ::, **1"deat=*m p g Q
,
(,
,,,.9 (1.00)
REFERENCE NMPC Operations Technology Vol. I, chapter 15 Hope Creek LP RXHP32-01 pg. 7,10 learning obj 2,3,5.
FJC 16 ANSWER 1.03 (2 00)
(0.5)
a. Indicated is higher.
(0.5)
7-10 inches higher (0.5)
b. Indicated level is sensed outside the dryer skirt.
Steam flow through the steam separator / dryer at 100% causes (0.5)
a backpressure of 7-10 inches H2 __
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1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE
--- iAEER559RAsiC5-DEAi isisiFEE As5 FEGi5 FE5s
____________________________________________
ANSWERS -- HOPE CREEK-86/02/24-CRESCENZOr F.
REFERENCE Hope Creek Vessel Instrumention LP ps. 45, learn. obj 9.
FJC 81 ANSWER 1.04 (2.50)
A. 29.9' - 27.9' = 2' Hs absolute (0.25)
2' Hs absolute =.98 psia (0.25)
Tsat for.98 psia = 100 F (0.25)
100 F - 90 F = 10 F condensate depression (0.25)
(1.0)
B. Plant efficiency is reduced (0.5)
C. NPSH increases (0.5)
D. Reduce turbine load, Increase cire water flowe raise condenser pressure (1 required)
(0.5)
REFERENCE Steam Tables, Hope Creek LP HT&T 25 I.O.
FJC184 Ib ANSWER 1.05 (3.00)
8,Y4 Cert e o y
.
o. BETA-EFF = BETA-CORE (I), I= Keff(delayed)/Keff(prompt).i.e. BETA-EFF takes into account that delayed neutrons are born at lower energies than prompt neutrons. The major effect of this is that delayed neutrons are less likely to cause fast fission.
(1.00)
b. Decreases due to the increase in fraction of total core power produced by fission of PU-239 which has a BETA < BETA (U-235,239)
(1.00)
c. Shorter periods as BETA decreases for the same reactivity insertions.
(1 00)
REFERENCE LP RXPH 23-01 ps. 5, 10. Learning objectives 3, 6 (FJC185)
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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE
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____________________________________________
ANSWERS -- HOPE CREEK-86/02/24-CRESCENZO, F.
ANSWER 1.06 (3.00)
cudddes w do e r e-ym es RW=(fluxmm2)SL/px*2, S= control rod span, L= thermal diffusion length.
c. From 0-A, rod worth increases due to the neutron flux and thermal (1.00)
diffusion length increasing.
b. From A-B flux is relatively constant but thermal diffusion lenSth is (1.00)
still increassing due to the heatup.
c. From point B-C, the increase in flux and TDL are at slower rate
,g than the incrqpse in pitch so rod worth decre ses. *t (1.00)
w i lt vse b' -f KAL ( ' #' A k E
id
' " ' rlur eme T
g y s,.,) e REFERENCE uvs LP RXPH31-si pg. 14. Learning objective 46 (FJC186)
ANSWER 1.07 (1.00)
b REFERENCE Steam Tables, Hope Creek LP HT&FT 22 pss 10,12 I.O.
2-1 FJC180 ANSWER 1.08 (2.50)
a. The lowest pressure that the tank could drop to would be the saturation pressure for 60 des. F which is 0.256 psia (0.50)
b. Assumins head loss due to flow is neglisible, the answer is no.
cavitation would not begin until the level drops below 5 ft.
(1.00)
c. Yes. The added pressure of 14.7 psia at the pump suction would allow all of the water to be removed (1.00)
REFERENCE LP FF05-01 ps. 5, Learnin3 Obj. 41 FJC225
.
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1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE
--- isiER55VAARiCi-REsi issniFEE As5 FE5i5 FE5s
____________________________________________
ANSWERS -- HOPE CREEK-86/02/24-CRESCENZO, F.
ANSWER 1.09 (2.50)
c. Using P = Po e to the t/T then P = 75 e to 60/-80 P = 75 e to -0.75 = 35 on Range 4 E1 53 b. On down power transients, the rate of power change is limited by the rate of decay of the longest li.ved precursors,thus retarding the rate of power decrease.[1.03 (55.6 see half life)
REFERENCE SSES Rx Theory, SC023 A-4, Section 9 pg. 4 HOPE CREEK LP RXPH24-01, pg 13 LO #4 FJC256 ANSWER 1.10 (1.00)
Under severe degraded core conditions with vessel pressure remaining high, and the core uncovered, temperatures significantly higher than those calculated from the constant enthalpy line indicate the (1.00)
presence of superheated steam.
REFERENCE Hope Creek M.O.C.D. LP 104 pg. 15 I.O.
FJC264 ANSWER 1.11 (1.00)
CRW=(local flux / ave flux)mx2
,,
It is possible for local power measurements to be,,less than the core average. Withdrawal of control rods under these conditions could cause local flux to increase without causing average flux to increase.
This could result in increasing CRW as the rod is withdrawn. Also, if the highest reading LPRM is bypassed, the gain is required to (1.00)
compensate for the low average.
'
REFERENCE Hope Creek LP 17 pgs. 10,11 I.O.
1, 4.b FJC270
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1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE
--- isiss55isisiEs-sEii isissFEE Es5 FEUi5 FE5s
____________________________________________
ANSWERS -- HOPE CREEK-86/02/24-CRESCENZO, F.
ANSWER 1.12 (2.00)
A.
is the correct answer (0.5).
The larse rod worth is due to the high flux created at the rods location, while the averaSe neutron flux remains very low (1.5).
(2.0)
REFERENCE Hope Creek LP RXPH31 pas. 11, 12 I.O. 4.a FJC288 e
- _. _
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_ _ _ _., _. _ _ _ _
_ _. _ _ _ _
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2.
PLANT DESIGN INCLUDING SAFETY AND ENERGENCY SYSTEMS PAGE
_______________________________________________________
ANSWERS -- HOPE CREEK-86/02/24-CRESCENZO, F.
ANSWER 2.01 (1.00)
d REFERENCE HOPE CREEK LP 06-01 pgs. 23,30 I.0.
4105 LP 05-01 pgs. 32,fis el I.O.42 FJC197 ANSWER 2.02 (3.00)
o.
Will inject EO.253.
Turbine seal leakage resulting in potential air-borne activity in the RCIC room C0.753.
(1.0)
b.
Will inject E0.253.
Pump overheating and seal damase may result durinS low or no flow conditions CO.753.
(1.0)
c.
Will not inject CO.253.
Maximum signal from the flow element will result in the flow controller keeping the turbine speed at minimum [0 753.
(1.0)
REFERENCE LP 30-01, REACTOR CORE ISOLATION COOLING, (RCIC) pgs. 25,40,fis 3, I.O. 4,5 i
I l
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_--_--
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PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
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ANSWERS -- HOPE CREEK-86/02/24-CRESCENZO, F.
t ANSWER 2.03 (3.00)
1. Isolation (due to hi steam flow or equip area temp.) causing steam IV's and suppression pool suction valve (042) to close.
(1.00)
2. Turbine Trip due to isolation (1.00)
3. Vacuum breaker IV's close (due to low team line prgssure and
,F tsel. va lve s (1.00)
hi DW)
p/A 9 3suresa"$b up num REFERENCE HOPE CREEK LP16 pgs. 20,70,71, I.O.47 FJC259 ANSWER 2.04 (1.50)
(0.50)'
a. true b. false (0.50)
c. true (0.50)
REFERENCE HOPE CREEK LP 16 pgs.
4, 72, 79. I.O.41,8,9 FJC260 ANSWER 2.05 (3.00)
o. collected in 55 gallon drums (+0.5) precludes accidental intro.
of boron into any system which can communicate with the R.V.
(0.5)
b. SLC storage tank level decreasins reactor power decreasin3 pump discharSe Pressure squib continuity monitors (indicating lights)
pump running (indicating li3 hts)
squib current flow (back of 9-5 panel)
(5 reqde 0.4 each)
0' b# C',11 REFERENCE I.O. 42 FJC263 Hope Creek LP 23-01 pgs. 10,41
!
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-.-
-
-
-
--e
_a,
- -.,
. _,,, - - - -
, -.. -,. - - - - _,,..,, -
-
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2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
_______________________________________________________
ANSWERS -- HOPE CREEK-86/02/24-CRESCENZO, F.
ANSWER 2.06 (2.00)
1. Reactor water sample valves close 2. Mechanical Vacuum pumps trip if runnin3 3. Reactor scram kmf (4 neededi Jen(sfr[a%
4.
c'-
" - - ' ;teee i sc1 tion-F//
ne (M6 i.e 3 0.5 each)
REFERENCE LP 45, ps. 20 LP 53, P3
LP 22, pg. 61 ANSWER 2.07 (3.00)
1. Demin makeup to SACS on lo level in SACS A surge tank.(0.25)
2. 3X1o level isolates TACS from SACS loop A.
(0.50)
3. Low flow to TACS causes SACS loop B puups to auto start (0.5), TACS supp return valves for SACS B open(0.5). Also causes SW pump D to start (0.5).
4. 3X1o level in SACS B surge tank causes T ACS loop isolation valves to close(0.5).
5. SW pump C starts on low flow to TACS t?T (0.25)
REFERENCE Hope Creek LP 79, pgs. 10, 31. I.O.
Hope Creek LP 80, pss. 34 I.D. 47.s ANSWER 2.08 (1.50)
O. A and B pumps start immediately. C and D start after a 5 second (0.75)
time delay.
b. All pumps will start upon closure of the SDG breaker and bus (0.75)
undervoltage is cleared.
REFERENCE LP 302HC-000.00-028-02
_.
_ _ _ _ _ _ _ _ _ _ _
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- 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
_______________________________________________________
ANSWERS -- HOPE CREEK-86/02/24-CRESCENZO, F.
ANSWER 2.09 (2.00)
WNi SDC
'
' b '"'(
'
not
.
c. Will not d. Will (&reS each)
'
REFERENCE
',
-
RHR LP (28-02)Rev 3 paSes 24, 32, 21, 29, 34 ANSWER 2.10 (1.00)
Monitors static inverter output and switches to back up AC supply when l,0 0 less of inverter output is indicated.
( FreSS REFERENCE LP 66 1E AC power supply ved a llT gD ' O x ek c' O a
t ANSWER 2 11 (3.00)
o. TCV remain at 100% due to load limit d,
p(, g(
(.50)
%g (.50)
b. BPV open 5% due to max combined flow
'nsk k
h *t h
'2 t i (.50)
c. Power decreases due to lower pressure d. Pressure decreases due to BPV (.50)
FINAL (n25)
o. TCV at 100% pocition ( '. 2 5 )
b.
BPV shut c. power lower 1(.25)
d. pressure sliS tly lower (.25)
h REFERENCE Hope Creek LP I.O.
4 3,4,10
.
f a
L.
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2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
_______________________________________________________
ANSWERS -- HOPE CREEK-86/02/24-CRESCENZOr F.
ANSWER 2.12 (1.00)
A check valve installed around one valve to prevent a valve failure inhibiting the desired scram. OR another power source.
(1.00)
REFERENCE Hope Creek LP 22-02 pg. 13 I.O. 43
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-
-.. _ - -
. - - - - -.
.... - -
. -
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-. -
. _ _ -
_ -.
-.
..
. _ _ - - _ _, - - -
~
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3.
INSTRUMENTS AND CONTROLS PAGE
____________________________
ANSWERS -- HOPE CREEK-86/02/24-CRESCENZO, F.
ANSWER 3.01 (2.50)
c.
Alternate Rod Insertion and Recirculation Pump Trip (RPT)
b.
None c.
Feedwater Runback d.
Standby Liq'id Control u'
o.
10 minutes, 53 seconds.
REFERENCE HCGS Exam bank LP 24-0 question 10 and L.P.,
302HC-000.00-024-01, PS 15, 22, 25, a 30 ANSWER 3.02
'(2.00)
,
c.
Causes reactor level to'0ECREASE due to the Level Control System having a STEAM FLOW / FEED FLOW ERROR, STEAM FLOW < FEED FLOW CO.53 p u n p \\ 1-l\\'"'
resultinginaSIGNALtoDECREA(EtheSPEDOfTHEREACT@s FEEDS Yf*
FWhe c ** eTo i nd V -
PUMPS CO.53.
- m N
c
^
1 vom F W
+ o v-b.
Causes reactor level to INCREASE due to the Level Control System having a LEVEL ERROR, with N0 ccmpensating FLOW ERROR [0.53 resulting in a SIGNAL to INCREASE the SPEED OF THE REACTOR FEED PUMPS E0.53 c.
9:::ter 1;v;l shavid RCMAIH CGN3 TANT '-ceva: the
'C" FEED runr g/g u
Cc> rner E^.53 -111 uGCr-UF EG.53.
REFERENCE BF LP 12, pp. 16-19.
HCGS, Rx Water Lvl Cntr1, 302HC-000.00-059-01, pg 10,-11, 8 14 l
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3.
INSTRUMENTS AND CONTROLS PAGE
____________________________
ANSWERS -- HOPE CREEK-86/02/24-CRESCENZO, F.
ANSWER 3.03 (3.00)
(a)
(b)
(c)
ROD GROUP
03
INSERT ERROR 22-51 22-51 18-03 INSERT ERROR 46-55 46-55 46-55 WITHDRAW ERROR 34-27 Blank Blank (.25 each)
or i..
q, g.,
Jge h %, limi h REFERENCE Hope Creek LP 09-01, I.O.6 11
-
j ANSWER 3.04 (1.00)
d REFERENCE Hope Creek LP 14-01 pgs. 9, 10 I.O.
4.b FJC214
i
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- --
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3.
INSTRUMENTS AND CONTROLS PAGE
ANSWERS -- HOPE CREEK-86/02/24-CRESCENZO, F.
ANSWER 3.05 (3.00)
1.
Narrow range O to 60 inches Calibrated hot Used for trip functions 2. Wide Range-150 to +60 inches Calibrated hot Used for trip functions 3. Upset range O to +180 inches Calibrated hot Not used for trips 4.
Shutdown range O to +400 inches Calibrated cold Not used for trip functions 5.
Fuel Zone-150 to + 50 Calibrated cold, no flow.
Not used for trip functions g4g g
p* g,h Q 3
,,. h 6t r<s b y
,
Deet :::ident ":nitv. i. 3.ense
N unique nse, vi.li s es.sisting 1;c 1 4 "+ e "= ant " t i er-Teaperatu m sv-yvnsetud Tui occident c.ditic^.
m De a
_un
+ e 4 r, re tiener (0.16 each)
REFERENCE Hope Creek LP 002-01, pgs. 23-26 I.O.
FJC262 ANSWER 3.06 (3.00)
(ouCCLE 80003'
k o" C r E'
"D c. C;iF S d
'. c ' M
- -
- ^ - "
(0.50)
"
b. All conditions for ADS initiation are not except the 105 see timer.
(ADS logics F,H sealed ini bed not timed out) therefore the valves are closed.
(1 00)
c. Rx press. < 50 psisi LP ECCS pumps stoppedi 1:c;1
- ne--d foit Timer or logic manually reset.
h/ 4 j.., in (1.00)
d. FALSE vh u[
he-t rue
,(v4 h ge,w lo.
(0.50)
- o, REFERENCE Hope Creek LP 29 pgs. 14, 15, 25, figs. 6,7 I.O. 8 FJC265
.
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3.
INSTRUMENTS AND CONTROLS PAGE
____________________________
ANSWERS -- HOPE CREEK-86/02/24-CRESCENZO, F.
ANSWER 3.07 (3.00)
O. Indicated level will increase as variable les pressure increases. (1.00)
b. Indicated level will not change as the variable les pressure will (1.00)
remain constant.
c. Indicated level will increase as reference les and variable les (1.00)
pressures equalize.
REFERENCE Hnpe Creek LP 02 pgs. 46-48, I.O.
FJC266 ANSWER 3.08 (2.00)
c. TRUE b. F*emP T R ' E c.
TRUE (0.5 each)
d.
TRUE REFERENCE H3pe Creek L.P. APRM pgs. 13,14,26,27, table II, figure 9, I.O. 2,4,5,8 ANSWER 3.09 (2.00)
1. Locally at SW structure (0.33)if given control from the control room (0.33)
Pumps A,C can be operated at the {Weumpgreakers(0.33) bytakigsbgr, 2.
' " * -
- '
to emer. tkovr.(0.33)
raw 4 x C y
V"' k w - l y
' * 5 3 3. Pumps B,D canbe operated from the RSP(0.33) if in emer. tkovr.(0.33)
REFERENCE Hope Creek LP 79, pgs. 9, 10, 12. I.O.
44.e, 6 FJC269 ANSWER 3.10 (1.50)
"F" at be d, e (0.25 each)
'I'
at a, c,N(.
h de nk sthse 3 r u n n [g '
wu
&
4-e d
sem L J,,L A <
-
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3.
INSTRUMENTS AND CONTROLS PAGE
____________________________
ANSWERS -- HOPE CREEK-86/02/24-CRESCENZO, F.
REFERENCE Recirc LP (No. 302 HC-000.00-020-01),
Objective 8 ANSWER 3.11 (2.00)
c. The diesel speed changer is used to change load.
(0.50)
b. 1NW" The sync scope should be operating slow in the clockwise (counter clockwise) direction. Close breaker at 5 till 12.
(0.50)
The 1E bus should be running faster with slightly higher voltage than the normal feed bus.
(0.50)
The sync scope will increase in al:;kwrre (counterclockwise) speed (0.50)
S\\ouri C
REFERENCE Hope Creek L.P.
066 I.O.
4 1.6.a Procedure OP-SO.PB-001 pg.S.7 FJC285 Cod O
'N h
On
N
- Q
- t
'"]
es t
?? s 7 r.e.
.
-
..
-.-.y,.
,
, - -
,
. _,.
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4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
-
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~~~~ 56 6L55iEAL 55 sir 5L R
____________________
ANSWERS -- HOPE CREEK-86/02/24-CRESCENZO, F.
ANSWER 4.01 (2.00)
c. To minimize the transient.
(0.50)
b. Max. temp. at which SLC initiation will result in injection of hot shutdown boron weight before the supp. pool reaches the HCTL in an i.e.(assuresshutdownpriortoemergencydepressurization)
(1.50)
ATWS, REFERENCE LP OP-EO.ZZ-102 pg. 9 ANSWER 4.02 (2.25)
o. 90 inches (0.25)
b.
The MSL floods beyond 118 inches which will flood the HPCI/RCIC steam lines. HPCI/RCIC turbine trouble alarms due to flooded steam drain pots. Delay operation of HPCI or RCIC until level drops between levels 2 and 3.
(2.00)
REFERENCE OP-AB.ZZ-117 ANSWER 4.03 (2.25)
a. 3293 MHT (0.25)
b. The average power over any 8 hr period shall not exceed rated.
At no time shall power exceed 102% rated.
(1.00)
c.
Check coolant chemistry, and thermal limits (1.00)
REFERENCE OP-IO.ZZ-006 ANSWER 4.04 (2.00)
1.
Scram condition and power >5% or undetermined (0.50)
2. RPV water lever below -38 inches or undetermined (0.50)
,
'
3. Rx pressure above 1037 psis (0.50)
4. Drywell pressure above 1.68 psis (0.50)
l
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4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
~~~---------------
~~~~ d65UUUUEUAE~UUUTRUL R
____________________
ANSWERS -- HOPE CREEK-86/02/24-CRESCENZO, F.
REFERENCE OP-EO.ZZ-101 ANSWER 4.05 (1.25)
o. 75 REM for life, 25 for equipment (0.50)
b. The site emergency plan must be activated, the e"ra-"-= t':11 Lw
- se::t.J m,
r I;r-ta;it
'-"d
- nd
"*uia""d
"Y t""
dictier pret-n+ inn==^r;-
-
(0.75)
REFERENCE SA-AP.ZZ-024 pg. 46 ANSWER 4.06 (3.00)
s. To maintain an accurate indication of reactor period.
(1.00)
b. Increase pressure setpoint to prevent unplanned depressurization. (1.00)
'
c.
To minimize thermal transients on the reactor vessel. Opening a bypass valve may be necessary to acheive steady feedwater flow. (1.00)
REFERENCE Hope Creek OP-IO-ZZ-003 pg.
9, cautions 3.2.5, 3.2.6 FJC277 ANSWER 4.07 (1.25)
1.
Attempt to reseat the valve via manual switch.
(0.5)
2.
Reduce recire flow and SCRAM when suppression pool temperature exceeds 110 (0.5) or if stuck open 2 minutes.
(0.25)
REFERENCE Hope Creek OP-AB.ZZ-121.
FJC278 ANSWER 4.08 (3.00)
1. To minimize energy addition to the containments.
(1.00)
2. To minimize rad. releast from the primary containment.
(1.00)
3. To maximize injection flow from motor driven pumps (1.00)
AcW f y
bree di tw-h
$NPm4 t c[onupIs Soelb g &y
~~
l l
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4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
~
~~~~ 565 L6556dL C6 iR6L R
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ANSWERS -- HOPE CREEK-86/02/24-CRESCENZO, F.
REFERENCE OP-EO.ZZ.202 Purpose section 1 ANSWER 4.09 (1.00)
o. To prevent CRD mechantsa internal sta) damage from excessive drive water flow. oF f6'o r e tw\\Aq + L,w (0.50)
b. Ensure no trips exist in cther channel to prevent a full scram.
(0.50)
REFERENCE
!! ope Creek OP-SO.SB-001 cautions.
5.3.4, 5.4.2 FJC281 ANSWER 4.10 (3.00)
1. CORE SUBMERGENCE accomplished by maintaining a level above TAF.
(1.0)
2. SPRAY COOLING accomplished by using at least one CS system operating at design conditions.
(1.0)
3. STEAM COOLING accomplished by using SRV to create steam updraft from plenum.dyp.hmuff e i4x4h
.9 (1.0)
i ps,ventory in lower in core reS on or water emig od one s ev open cun on.%
u eve 70e
ig REFERENCE Hope Creek definition of adequate core cooling in E0P's (3.00)
ANSWER 4.11 (1.00)
By maintaining the water level above the minimum for natural cir-(1.0)
culation, adequate coolant mixins and core cooling is assured.
REFERENCE OP-SO.BC-001(0) RHR system procedure, Precaution 3.11.1
.
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4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
~~~~~~~~~~~~~~~~~~~~~~~~
R5656EUU56IE U UTE L
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ANSWERS -- HOPE CREEK-86/02/24-CRESCENZO, F.
ANSNER 4.12 (3.00)
o. Skimmer surge tank makeup valve HV-4660 opens Fuel pool pumps trip RBVS isolates FRVS auto starts (4 rqd 9.5ea)
b. Terminate refuelins activities Evacuate the refuel floor (2 rqd 9.5ea)
REFERENCE Hope Creek procedure OP-AB.ZZ-144
.. _ _
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TEST CROSS REFERENCE PAGE
QUESTION VALUE REFERENCE
________
______
__________
01.01 2.50 FJC0000015 01.02 2.00 FJC0000016 01.03 2.00 FJC0000081 01.04 2.50 FJC0000184 01.05 3.00 FJC0000185 01.06 3.00 FJC0000186
01.07 1.00 FJC0000188 01.08 2.50 FJC0000255 01.09 2.50 FJC0000256 01.10 1.00 FJC0000264 01.11 1.00 FJC0000270 01.12 2.00 FJC0000288
______
25.00 02.01 1.00 FJC0000197 02.02 3.00 FJC0000257 02.03 3.00 FJC0000239
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02.04 1.50 FJC0000260 02.05 3.00 FJC0000263 02.06 2.00 FJC0000267 02.07 3.00 FJC0000271 02.08 1.50 FJC0000272 02.09 2.00 FJC0000273 02.10 1.00 FJC0000275 02.11 3.00 FJC0000286 02.12 1.00 FJC0000287
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25.00 03.01 2.50 FJC0000160 03.02 2.00 FJC0000163 03.03 3.00 FJC0000198 03.04 1.00 FJC0000214 03.05 3.00 FJC0000262 03.06 3.00 FJC0000265 03.07 3.00 FJC0000266 03.08 2.00 FJC0000268 i
03.09 2.00 FJC0000269 03.10 1.50 FJC0000274 03.11 2 00 FJC0000285
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25.00 04.01 2.00 FJC0000165 04.02 2.25 FJC0000172
04.03 2.25 FJC0000174 l
04.04 2.00 FJC0000175
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04.05 1.25 FJC0000179 04.06 3.00 FJC0000277 l
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GUESTION VALUE REFERENCE
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04.07 1.25 FJC0000278 04.08 3.00 FJC0000280 04.09 1.00 FJC0000281 04.10 3.00 FJC0000282
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04.11 1.00 FJC0000289 04.12 3.00 FJC0000290
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NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY *
HOPE CREEK
_________________________
REACTOR TYPE:
BWR-GE4
_________________________
DATE ADMINISTERED: 86/02/24
_________________________
EXAMINER:
KOLONAUSKI/ LANCE APPLICANT:
_.~
__
__
INSTRUCTIONS TO APPLICANT:
__________________________
Use separate paper for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question. The passins grade requires 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.
% OF CATEGORY
% OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY
-----------------------------------
53rar--
23 t-
_ D I__ _ M
___________
________ 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS
'5.00 96,44
_'_______ ______
________ 6.
PLANT SYSTEMS DESIGN, CONTROL,
___________
AND INSTRUMENTATION 202
________ _'_"_;;
________ 7.
PROCEDURES - NORMAL, ABNORMAL, 25.00
-
^^
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EMERGENCY AND RADIOLOGICAL 21 0 23.0 CONTROL
_ff*II___ff!f
________ 8.
ADMINISTRATIVE PROCED.URES,
___________
CONDITIONS, AND LIMITATIONS 95'.S i^^.^0 100.00 TOTALS
________ ______
___________
________
FINAL GRADE _________________%
All work done on this examination is my own. I have neither givan nor received aid.
IEPL5CEUT 5 55G ETUR5~~~~~~~~~~~~~~
I
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
____
______________________________________
______________
GUESTION 5.01 (3.00)
LIST t'te three (3) Limiting Safety System Settings (LSSS)
(3.0)
associated with the APRM system that help maintain the fuel clad integrity safety limits (SL).
For each LSSS listed, include in your answer:
1.
The reactor mode of operation in which the LSSS is active (ie. STARTUP, RUN, etc.), IF applicable.
2. A description of the associated Safety Limit (SL).
-0UESTION 5.02 (1.00)
The steady state MCPR limit given in Tech Specs is multiplied by (1.0)
flow biasing correction factor
"Kf'.
Explain the two (2) purposes for this Kf factor.
QUESTION 5.03 (1.50)
a. List six (6) of the energy terms used in the reactor heat balance (0.9)
equation AND state if the term is an energy INPUT or DUTPUT.
b.
What is the main reason for performing a reactor heat balance?
(0.6)
QUESTION 5.04 (2.00)
a. The thermal utilization factor (f) and the resonnance escape (1.0)
probability (p) are the two factors of the six factor equation for Keft that are influenced the most by the moderator-to-fuel ratio (Nmod/Nfuel).
STATE how each changes (ie., increases, decreases, or remains the same) as this ratio approaches zero.
b. Why are all commercial reactors undermoderated?
(1.0)
!
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THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
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QUESTION 5.05 (3.00)
A Feedwater pump is operating at 75% efficiency.
(3.0)
The water is supplied at 680 psis and exits at 1000 psig.
The mass flow rate through the pump is 2 x 10E6 lbm/hr.
l Find the TEMPERATURE RISE of the water passing through the pump due to the pump's inefficiency.
Recall-density of fluid being pumped
= 62.4 lbm/ft3 1 BTU = 778 ft-lbf Cp = 1 BTU /lbn-desF QUESTION 5.06 (2.00)
The term ' critical power' refers to that bundle power level (2.0)
corresponding to the onset of transition boiling (OTB) somewhere in that bundle.
State how critical power varies (ie. increases, decreases, or is not affected) by each of the following:
a. If coolant mass flow rate increases b.
If reactor pressure increases c.
If local power increases d. If inlet subcooling increases QUESTION 5.07 (1.50)
On a single graph, sketch the following three characteristic (1.5)
curves. Be sure to label the axes and each curve as
'a',
'b',
or
'c'.
a. single speed centrifugal pump b.
two identical centrifugal pumps operating in series c.
two indentical centrifugal pumps operating in parallel.
QUESTI01 5.08 (1.50)
Following a reactor scram from 100% power, explain what happens initially to each of the following parameters (ie., increases, decreases, or remains constant), and WHY.
c. Flow through the core (0.75)
b. Pressure drop in the steam lines (0.75)
(*****
CATEGORY 05 CONTINUED ON NEXT PAGE ummum)
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5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
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DY TM N gg @ g$ M ~ ONuf45 QUESTION 5.09 (1.00)
g g py
,
Which of the'following requires a steam condenser to remove the (1.0)
MOST heat energy in order to condense the steam? (CHOOSE ONE)
c.
one pound of steam at 300 psia b.
two pounds of steam at 600 psia c.
two pounds of steam at 1200 psia d.
one pound of steam at 15 psia QUESTION 5.10 (2.00)
Which of the following situations is correct in most cases?
(2.0)
Explain your choice.
A.
A control rod's worth is greatest when it is fully withdrawn and all other rods remain inserted.
B.
A control rod's worth is greatest when it is fully inserted with all others withdrawn.
QUESTION 5.11 (3.00)
c. Explain the difference between the delayed neutron fraction (BETA-CORE)
and the effective delayed neutron fraction (BETA-EFF).
(1.00)
b. How and WHY does the delayed neutron fraction change through core life (BOL-EOL)?
(1.00)
c.
How does this change in the delayed neutron fraction affect reactor power control?
(1.00)
GUESTION 5.12 (2.50)
s. At the beginning of a fuel cycle, control rod density is approximately 10 to 12% at equilibrium full power. Approximately one third into the cycle, the control rod density is about 15 to 16% at equilibrium full power. Why is there a difference?
(1.00)
b.
What effects does this increase in control rod density have on the void coefficient of reactivity? Explain your answer.
(1.50)
(xxxxx CATEGORY 05 CONTINUED ON NEXT PAGE xxxxx)
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THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
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QUESTION 5.13 (1.00)
Current plant conditions sussest that a severly degraded core condition exists. Several SRV's have been, or are currently, open. The STA reports that the temperature detectors in all the SRV tailpipes are reading ERRONEOUS since the readings are significantly hisher than those calculated from the Mollier diagram. Would you agree with this conclusion?
(1.00)
(Justify your answer.)
i (xxxxx END OF CATEGORY 05 xxxxx)
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6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
GUESTION 6.01 (3.00)
The Redundant Reactivity Control System (RRCS) indentifies and prevents on ATWS condition by initiating an Alternate Rod Insertion (ARI).
a. List ALL conditions (with associated setpoints, if applicable)
(1.0)
that will actuate the ARI logic of the RRCS.
(1.0)
Briefly describe how the ARI function is physically accomplished.
b.
c. Figure i shows the control room panel 10C651 for the RRCS. If (1.0)
after the ARI has automatically initiated, the status light
'ARI READY FOR RESET * illuminates in both logic trains, WHAT THREE CONDITIONS does this signify?
QUESTION 6.02 (2.25)
The Transversing In-Core Probe System (TIP) is used primarily to calibrate the Local Power RanSe Monitors (LPRHs).
a. List two uses the Process Computer has for the TIP scan data.
(1.00)
(0.75)
b. What is the purpose of the TIP purge system?
c. What adverse condition could develop if the TIP purge system were (0.50)
inoperative for an extended period of time?
OUESTION 6.03 (2.00)
List FOUR (4) automatic actions OTHER THAN a Group 1 Isolation, (2.0)
that will occur as a DIRECT result of the Main Steam Line Rad Monitors
'A' and*B' reaching their High-High trip setpoint.
QUESTION 6.04 (3.00)
(1.0)
e. State the purpose of each of the SLC tank heaters.
b. State the three control locations for the SLC pumps and for each (1 0)
location, state whether a local START signal provided to the pump WILL or WILL NOT fire the associated squib valves.
Identify the SLC tank level interlock, identify its setpoint, and (1.0)
c. state how it accomplishes its function.
(xxxxx CATEGORY 06 CONTINUED ON NEXT PAGE xxxxx)
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6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
______________________________________________________
QUESTION 6.05 (1.50)
During a Main Turbine trip, the turbine stop and control valves (1.5)
close very rapidly to protect the turbine from an overspeed condition.
Even with this protection, an overspeed condition may still occur.
c. What system characteristics could cause this to happen?
b. What additional design feature is used for overspeed protection?
OUESTION 6.06 (2.00)
List the four (4) signals which cause Normal Reactor Building (2.0)
Ventilation to shut down and isolate, and automatically start the Recirculation Filtration and Ventilation System.
QUESTION 6.07 (2.25)
The Residual Heat Removal System (RHR) is operating in the SHUTDOWN COOLING Hode. For each item below, describe the effect on the RHR system, including valve responses.
a. Reactor pressure exceeds 90 psis (0.75)
b. Reactor vessel level reaches +12.5'
(0.75)
c. Reactor vessel level reeches -40'
(0.75)
GUESTION 6.08 (2.00)
n. Following an AUTO initiation of RCIC at a reactor pressure of (1.00)
800 psis, reactor pressure decreases to 400 psis. Assumin3 the RCIC system operates as designed, HOW will the RCIC system flow rate be affected by this drop in reactor pressure? (ie. Increase, Decrease, or Not change) Briefly explain your choice.
b.
If the Auto initiation logic fails and RCIC has to be MANUALLY (1.00)
!
!
initiated at a reactor pressure of 800 psis, HOW will the RCIC flow rate be affected if reactor pressure drops to 400 psis?
Again, briefly explain your choice. kSIUME cwMtv in MANM-(
(wih krard )
' darkt)tX M (xxxxx CATEGORY 06 CONTINUED ON NEXT PAGE xxxxx)
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6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
______________________________________________________
GUESTION 6.09 (2.00)
Describe the conditions necessary to cause the following alarms on the Interlock Status Display Module associated with the Refuelin3 Platform.
USE attached Figure 2 for reference.
(0.5)
c. Back Up Hoist Limit b. Rod Block Interlock No. 1 (0.5)
(0.5)
c. Fuel Hoist Interlock d. Brid3e Reverse Stop No. 1 (0.5)
GUESTION 6.10 (1.00)
The Recirculation MG Set Oil System is in its normal lineup for (1.0)
power operation when the running AC oil pump trips. The DC oil pump auto starts when the standby AC pump fails to restore oil pressure above 20 psis. Which of the followin3 correctly refelects the current oil system equipment status?
a. The AC oil pump (s) AND the MG set drive motor have tripped.
b. The AC oil pump (s) have tripped; the MG set scoop tube lockup tripped.
c. The AC oil pump (s) have trippedi the MG set drive motor continues to run.
d. The AC oil pump (s) and the MG set drive motor continue to run.
QUESTION 6.11 (1.00)
SELECT the one statement below which BEST DESCRIBES the operation /
(1.0)
performance of an IRM durins a reactor startup.
a. When the IRM is reading full scale on Range 10, the APRM's should be reading approximately 10% power.
b.
Shifting from Range 4, indicating 75, to Range 5 will result in an indication of 24 on Range 5.
,
c. Reactivity feedback, due to the moderator temperature coefficient, should begin at approximately Range 5.
d. When an IRM channel increases from 25 on Range 2 to 25 on Range 3, j
indication has increased by one decade.
the
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(***** CATEGORY 06 CONTINUED ON NEXT PAGE xxxxx)
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a 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
______________________________________________________
QUESTION 6.12 (3.00)
Following a valid initiation of HPCI (due to high drywell pressure)
('
)
a steam leak just upstream of the steam admission valve develops.
Describe the response of components within the HPCI system, assuming the high drywell signal remains. Limit your answer to include all generated signals, system valve responses, and final HPCI turbine status.
.
(xxxxx END OF CATEGORY 06 xxxxx)
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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
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QUESTION 7.01 (1.00)
When operating one loop of the RHR system in the Shutdown Cooling (1.0)
Mode, and a loss of flow occurs, EXPLAIN WHY reactor level should be raised to 100' on the Shutdown Range.
QUESTION 7.02 (2.50)
List TWO reasons for placing the Mode Switch in SHUTDOWN after a (0.5)
a.
reactor scram has been verified.
b. List the plant conditions neccessary to allow entrance into the Post Scram Recovery operating procedure from:
1. OP-EO.ZZ-101 RPV Control (1.0)
2. OP-EO.ZZ-100 SCRAM (1.0)
Note: Answer each separately.
QUESTION 7.03 (3.00)
Concerning OP-EO.ZZ-202, ' Emergency Depressurization",
a.
What is the reason behind the caution that requires Supression (1.0)
Chamber level to be greater than +78'
before the operator is allowed to use the ADS valves?
b. If the Supression Chamber level is not above
+78',
this procedure (1.0)
lists several alternative systems and components that may be used to depressurize the RPV. List four (4) of these components and/or systems.
c. Assume this emersency procedure is entered, the Supression (1.0)
Chamber level is above
+78',
and less than three (3) SRV/ ADS valves are open.
Why are you instructed to skip the alternate depressurization methods if reactor pressure is less than 50 psis above the supression chamber pressure?
(*****
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7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
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QUESTION 7.04 (2.00)
(2.0)
Hope Creek is in the process of refueling when the crane operator drops a fuel assembly into the storage area. According.to the Irradiated Fuel Damage while Refueling procedure (OP-AB.ZZ-101),
what are the Immediate operator and automatic actions that must be parformed/ verified?
QUESTION 7.05 (1.50)
Consider OP-IO.ZZ-007, " Operations from Hot Standby (MSIVs closed)*!
a. Briefly explain why there should be no vacuum on the Main Condenser prior to removing the steam seals on the Main Turbine.
(0.5)
b. Briefly explain why reactor makeup should be maintained relatively constant during low flow conditions.
(0.5)
c. According to the Hope Creek Tech Specs, what two conditions define (0.5)
Hot Standby?
QUESTION 7.06 (2.00)
While you are on shift, the following alarms come in:
GAS RADW CHAR TRTHNT PNL 00367 RADIATION MONITORING ALARM /TRBL The radweste operator notifies you that the '0FFGAS RADIATION HIGH'
local alarm is also in.
,
c. What immediate operator action > are necessary per OP-AB.ZZ-127, (1.0)
'Off Gas High Radiation'?
'
b. The reactor operator suggests that a break or malfunction of the (1.0)
>
l SJAE may be the cause of this problem. What control room indications would you direct him to monitor, and what would you expect him to
'
j see if this was the case?
!
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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
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QUESTION 7.07 (2.50)
c. List the entry conditions (with set points, if applicable) for (2.0)
the Reactor Pressure Vessel Control procedure (OP-EO.ZZ-101).
(0.5)
b. Caution 10 of this procedure states
'Do not secure or place an ECCS in manual mode unless, by two independent indications, (1) Misoperation in the automatic mode is confirmed, or (2) adequate core cooling is assured.'
How is 'misoperation' defined?
GUESTION 7.08 (3.00)
Answer the following questions regarding procedure OP-IO-ZZ-003,
'Starup From Cold Shutdown to Rated Power *:
a. During a reactor startup, the procedure requires that the 10ee2 and SRM detectors be withdrawn to maintain count rate between 10ee5, however, only 2 detectors may be withdrawn at any one time. (1.00)
What is the purpose of this limitation?
b. During a reactor heatup/ pressurization, reactor pressure increases above the pressure setpoint AND no turbine bypass valves are open.
According to the procedure, what ceatain precaution must (1.00)
be taken prior to initiating corrective action and why?
c. "During low flow conditions, feedwater flow to the reactor should be maintained rela + vely constant...' Explain why this caution exists in and what action the procedure recommends to assist the operator (1.00)
acheiving a steady feedwater flow.
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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
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QUESTION 7.09 (3.00)
During refuelins the followin3 alarms occur:
FUEL POOL LEVEL HI/LO FUEL POOL COOLING SYS LEAKAGE HI FUEL POOL COOLING SYS TROUBLE o. List four automatic actions which would occur (assume pool level continues to decrease)
(2.00)
b. In addition to ensuring that all appropriate automatic actions are completed, what immediate operator actions are necessary?
(1 00)
QUESTION 7.10 (3.00)
According to the E0P's, three viable methods of accomplishing adequate core coolin3 exist. State these three methods in order of preference, briefly describe how each is accomplished, and how adequate core coolin3 can be verified during use of each method (if applicable).
(3.00)
QUESTION 7.11 (1.50)
According to OP-EO.ZZ-207, ' Level / Power Control Procedure', prior to lowerins water level, the operator is instructed to bypass the low water level interlocks for the MSIVs and Primary Containment l
Instrument Gas.
a. Explain why this is done.
(1.0)
b. If the MSIV interlock is bypassed, and the Main Steam Line Rad (0.5)
Monitors exceed three times the normal full power background, will the MSIVs still close?
(xxxxx END OF CATEGORY 07
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ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE
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QUESTION 8.01 (3.00)
Hope Creek is operating at 100% power and has been for the past (3.0)
three days.
A computer P-1 edit indicates the need for a TIP trace in all areas of the core. The Reactor Analyst informs you that 3 of the 5 TIP machines are inoperable and are not repairable for at least one week. You are the Shift Supervisor.
USING THE ATTACHED TECH SPECS-What are your required actions and what is the most limiting LCO?
QUESTION 8.02 (2.00)
,
List the three conditions which must be met in order to make (2.0)
on-the-spot changes to written procedures. - Octord(
to Tech 5yt.Ct QUESTION 8.03 (1.50)
With the plant at 75% powere you are informed by the IRC supervisor (1.5)
that a scram discharge volume (SDV) level switch to RPS has failed the channel function test. What actions must you take according to Tech Specs?
xxxx USE THE ATTACHED SECTIONS OF THE HOPE CREEK TECH SPECS ****
xxx FULLY REFERENCE THE TS SECTIONS THAT YOU USE IN YOUR ANSWER xxx QUESTION 8.04 (2.50)
'
a. List four (4) conditions which require a Radiation Work Permit (1.6)
(RWP) to be issued.
b.
List the Shift Supervisor's responsibilities for an RWP.
(0.9)
QUESTION 8.05 (1.50)
During events that require implementation of the Emergency Plan (0.5)
a. procedures, WHO initially becomes the Emergency Director?
b. List two other individuals who may assume the role of Emergency (1.0)
Director.
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ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE
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QUESTION 8.06 (3.00)
You have just received word from the IRC supervisor that Reactor (3.0)
Pressure Vessel pressure instrument PT-N090K is inoperable.
The cupervisor estimates that it will take 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to replace and recalibrate the in *,rument.
IN ACC3? DANCE WITH THE TECHNICAL SPECIFICATIONS, WHAT ACTIONS MUST BE TAKEN DUE TO THIS INSTRUMENT FAILURE?
(Note: an RPV instrumentation list is attached to this exam as Figure 3.)
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx x
NOTE: USE THE ATTACHED SECTIONS OF THE HOPE CREEK TECH SPECS TO z
x ANSWER THIS DUESTION. FULLY REFERENCE ALL APPLICABLE SECTIONS OF x
x THE T.S. THAT YOU USE TO DEVELOP YOUR ANSWER.
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx 00ESTION 8.07 (2.50)
2. 5" The plant is operatin3 at 75% power. The IRC supervisor informs you L3<t)
that the EOC-RPT trip systems are inoperable due to all turbine control valve fast closure setpoints being out of spec ( <460 psis).
He further informs you that the problem is due to the use of faulty calibration.
test equipment which was used during the last channel Using the attached sectiores of Tech Specs, list all applicable action statements and state which is most limiting.
.
(xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx)
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8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE
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QUESTION 8.08
' 2. 3 01 According to the " Equipment Operational Control' procedure, c. An operator may determine valve position by direct observation (0.5)
of the valve in question.
TRUE or FALSE b. Briefly describe the actions necessary to verify that a manual (0.5)
valve is OPEN.
c. Briefly describe the actions necessary to verify that a manual (0.5)
valve is CLOSED.
According to the " Removal and Raturn of Equipment to Service' procedure, gdj
[ tion should be entered in what two When equipment is removed from service, the appropriate informa-
'O.;}-
logs?
e. Anytime a motor operated valve (MOV) that is required to change (0.5)
positions to fufill a safety a safety function is manually seated, that valve must be declared inoperable.
TRUE or FALSE QUESTION 8.09 (1.50)
According to the Hope Creek Tech Specs, what two guidelines apply (1.5)
to exceedin3 a Surveillance Requirement interval ?
GUESTION 8.10 (3.00)
a. List the four Emergency Classes established by the NRC for (1.0)
l nuclear facilities' in order from least severe to most severe.
b.
State whether or not the following occurrences require one-hour (2.0)
to the NRC. (Answer YES or NO.)
,
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1.
A plant shutdown as required by Tech Specs l
2.
A fire in the Technical Support Center
3.
An EO has discovered that a fire barrier has been nonfunctional for more than 10 days.
4.
Reactor vessel water level reaches -40'
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ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE
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QUESTION 8.11 (2.00)
Ccnsider the following situations a. According to Tech Specs IS IT PERMISSIBLE to go from STARTUP to (1.0)
RUN if IRMs A, B, and C are inoperable? - Exyls WHV.
b. If the same IRMs were found inoperable while in RUN, would you violate any Tech Specs by:
(0.5)
1.
Staying in RUN?
Explain.
(0.5)
2. Placing the mode switch in STARTUP? Explain.
(xxxxx END OF CATEGORY 08 xxxxx)
(x*******xxxxx END OF EXAMINATION xxxxxxxxxxxxxxx)
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1/(1-k)
p = p, et / r M
.-
I Ici - 3.7 x 10 %q N(t) = No e-AT
op = - 1 x 10 5 g/*T c;= (tg+L,) (cred)2 K
(4 avg)
.
s,. - 1 x 10 3 AK/: voids n = v/(1 + d)
,
K P = I e v/(3.7 x 1010)
g = - 4.5 x 10 6 aK/:*T K
t=
(2 s)/18
.
.
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ap = -4.5 x 10 " K/I Power T = 1/p + (p p)/1p,
,
K t= 1/(c-8)
1(c) = Io e-At v=Yg + xvg T1/2 = in(2)/ A E = zhg + (1 x) bg Cp = (CFbase) (ES) (I )
A 5 - x5g + (1-x) Sg Q = 2rt.; at 1 in. - 2.54 ces ap = f L v2 o
D 2sc 1 gal. - 3 785 liters y = 6(/Le 1 kg = 2.205 lb p = kfeff) -1 N = pao /A K(e 1)
17.5E vat ts = 1 ETD/ min
!
CR1 1-I(eff)2 1 psi = 6.895 Ps S (0 0:}
- ' '
1 psi - 2.936 - H ti O (G 4 ;
tl CT.2.
1-r(eff)1 1 psi = 27 68 *
I =.0071
.
0 = He
_
2 x 10, sec 1 =
.
l q = UAAT l
l
-
!
f
l l.:l
i
-
n c; ')
'
P... i.,. k.4
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
______________________________________
____
______________
ANSWERS -- HOPE CREEK-86/02/24-!(OLONAUSKI/LANGE ANSWER 5.01 (3.00)
1.
a. APRM se m at 15%
gW (0.33)
(0.33)
b. Active in TARTUP c. SL-For Rea tor Pressure les than 785 psis or core flow (0.33)
less than 10%
f rated, ta d thermal power must not exceed 25%.
2.
ed thermal power in STARTUP (0.50)
18% and/or (0.66W + 51)% in RUN (0.50)
c.
SL-For Reacto Press e greater than 785 psis and core (0.33)
flow greater an 10% o rated, MCPR must not drop below 1.06.
3.
a.
T=
(FRTP FLPD)
(0.33)
APRM ga'i is adjusted to re d 100% times the CHFLPD when T is les than or equal to 1.0.
b.
No not specified.
c.
'M flow biased simulated therm upscale is compensated (0.33)
hen the existing Power distributio would cause the design LHGR to be exceeded at rated thermal wer.
REFERENCE HC LP HT&T No. 14) Learning Objective No.
2, Page 3.
HC TS LP, LO 1 ANSWER 5.02 (1.00)
1. The Kf factor adjusts the MCPR operating limit for core flows (0.5)
other than rated.
2. The Kf factor assures that the MCPR SL will not be exceeded (0.5)
during the flow increase transient resulting from an MG set speed control failure.
REFERENCE HC LP HT&T No. 14) Learning Objective No. 5, page 5
.
o.4 G0%
A. L55 5 - l6't. therm yeiadr scranw o. 2 ack ta.- whe.w NOT in guts)
SL - 2 s't. rahd. WerumA. yeuse.r w4<m Kx ye<.
wu. < T w p3i)
oA ove cort flow
<,10*/. rakd b. L 5 5 5 - (c. 66W + 61) T Gw - blaud scrwm ; ek@ a.$ til. V '4 o.t
activt - All Madel SL - % HCYtt 7, l.0L % h ynthLM. > ]f 5' yEE*] A *
f-(su > ls't. M d.
O'I C. LIJJ -
ll 3 */. fint A M N ICh o.2 ar,tiut ~
kVM M.ch 0Y S t.,
-
s a.h4L 4.s b,
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
____
_
_________________________________. ___
______________
ANSWERS -- HOPE CREEK-86/02/24-KOLONAUSKI/LANGE ANSWER 5.03 (1.50)
a. INPUTS- 0 Feedwater DUTPUTS - 6 Steam (6 0 0.15 ea)
0 Recire Pump G RWCU(inks F/0)
O CR0 Cooli,n3 Flow 0 Ambient / Radiative O Core dt.gweg (, Joe e/oy b. Heat balances are performed to ensure the accuracy of the (0.6)
nuclear instrumentation. (APRM calibration)
REFERENCE HC LP HT&T No. 35 Reactor Heat Balance; page 4, pages 6-9 ANSWER 5.04 (2.00)
A.
The thermal utili=ation factor, f, INCREASES as the moderator (0.5)
to fuel ratio decreases toward zero.
The resonnance escape probability, p, DECREASES as the moderator (0.5)
to fuel ratio decreases toward zero.
B. An undermoderated core provides stability as power is increased.
(1.0)
As power increases, density decreases, which causes a decrease in Keff for undermoderated cores. Overmoderated cores cause an increase in Keff for a density decrease associated with a power increase.
REFERENCE Ex.lh.
A.
HC LP HT No. 17 page 6 HC LP H T No. 16 Transparencies 6 and 2 B. HC LP T&T No. 17 page 9 and Learnin3 Objective.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
______________________________________
____
HbPbCREEK-86/02/24-KOLONAUSKI/LANGE ANSWERS -
(J<[5 t.
0)
ANSWER 5.05 (0.6)
c. Pump Head-Hp = (Pout-Pin)/
(1000-680)*144/62.4 = 738.5 ft
=
(0.6)
b.
Ideal Work (power)-
Wpeideal = Hp m
= 738.5 (2 x 10E6)/60 = 2.46 x 10E7 ft-lb/ min (0.6)
c. Input Power-Wp, actual = WP, ideal /Np
= 2.46 x 10E7 ft-lb/ min / 0.75
= 3.28 x 10E7 ft-lb/ min (0.6)
d. Power converted to heat-0 = Wpeactual - Wp, ideal
= (3.28-2.46) 10E7 ft-lb/ min = 8.2 x 10E6 ft-lb/ min Convert to BTU:
10,566 BTU / min
=
(0.6)
e. Delta T across pump-delta T = O/mCp 10,566 (60)/2 x 10E6 = 0.32 des F
=
6, pages 11-13 FF LP 03, LO 7 ANSWER 5.06 (2.00)
(0.5)
a. Increases (0.5)
l b. Decreases (0.5)
'
c. Decreases (0.5)
d. Increases REFERENCE HC HT&T No. 11, Learning Objective 2, pages 8 and 9.
- 0.TT cmuoT 61%vS S 4" 'OC**t*
h $(w )wWaMygry
grNg!(t dWudY hr 6 05 qxtity
& = 4 w A G di
- A Ak C Pua*O. Med 2 yrerHeo.
AT:
A %ead tdid 4 aHud.u yn,9 g i
..
Cy 03 Cy o.w {3 nA) 7n N 4 N'F t ETu
5.
. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
____
_
_
______________________________________
______________
ANSWERS -- HOPE CREEK-86/02/24-KOLONAUSKI/LANGE ANSWER 5 07 (1.50)
1'
s (0.5 for each curve)
s, p"*b'
I s
l
-'
Pump i
\\
Discharge
-
- g.
'c'
Head
-
l.a. /
\\
'M i
4 __________________
Capacity REFERENCE HC LP No. FF 06, pages 4, 5.
Transparency 3 ANSWER 5.08 (1.50)
a. Increases (0.25), due to the void collapse when power decreases causing less two phase flow and therefore less two phase flow resistance. (0.5)
b. Decreases (0.25); steam velocity decreases due to the scram, therefore the fluid head (pressure) losses are lower. (0.5)
REFERENCE HC Fluids, FF 03 Bernoulli's Equation, Fluid Friction, Head Loss FF 04 Flow Measurement, Two Phase Flow ANSWER 5.09 (1.00)
b.
REFERENCE Steam Tables e
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
__---------_--
ANSWERS -- HOPE CREEK-86/02/24-KOLONAUSKI/ LANCE ANSWER 5.10 (2.00)
A.
is the correct answer (0.5).
The large rod worth is due to the high flux created at the rods location, while the average nautron flux remains very low (1.5).
(2.0)
8. cam also be rbRed if MWM W A IN CM MY REFERENCE Hope Creek LP RXPH31 pgs. 11, 12 LO 4.a FJC288 ANSWER 5.11 (3.00)
c. BETA-EFF = BETA-CORE (I), I= Keff(delayed)/Keff(prompt).i.e. BETA-EFF takes into account that delayed neutrons are born at lower energies than prompt neutrons. The major effect of this is that delayed neutrons are less likely to cause fast fission.
(1.00)
b.
Decreases due to the increase in fraction of total core power produced by fission of PU-239 which has a BETA < BETA (U-235,239)
(1 00)
c.
Shorter periods as BETA decreases for the same reactivity insertions.
(1.00)
REFERENCE LP RXPH 23-01 pg.
5, 10. Learning objectives 3,
(FJC185)
ANSWER 5.12 (2.50)
a.
As the reactor operates during the early part of the cycle, the burnable poison depletes more rapidly than the fuel, therefore, control rods must be inserted to hold the power constant.
(1.00)
b.
As the control rod density increases, the power producing regions of the core become more undermoderated; the moderator to fuel ratio is decreasing. In effect, as total power production has remained constant but the power producing volume has become smaller, the operating volume of the core has become undermoderated. Because of this effect, the void coefficient becomes more negative.
(1.50)
.
. _.
_
5.
THEORY OF NUCLEAR POWER. PLANT OPERATION, FLUIDS, AND PAGE
--_---_-------
ANSWERS -- HOPE CREEK-86/02/24-KOLONAUSKI/LANGE REFERENCE NMPC Operations Technology vol. I page I-12-6 Hope Creek LP RXPH19-01 pg.6 LO 44, RXPH28-01 pg.6 and trans. 41 LO 63 RXPH16-01 trans. 42 ANSWER 5.13 (1.00)
Under severe de3raded core conditions with vessel pressure remaining high, and the core uncovered, temperatures significantly higher than those calculated from the constant enthalpy line indicate the (1.00)
presence of superheated steam.
REFERENCE FJC264 Hope Creek M.O.C.D.
LP 104 pg. 15 I.O.
.-.
_
_ _ _
. - -.
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
______________________________________________________
ANSWERS -- HOPE CREEK-86/02/24-KOLONAUSKI/LANGE ANSWER 6.01 (3.00)
(1 0)
o. Vessel High Pressure, 1071 psis, OR Vessel Low Level (-38'), OR RRCS Manual Initiation, OR ARI Test Switch turned to TEST (1.0)
b. Eight DC solenoid actuated vent valves blow down the Scram Valve Air Pilot Header.
c. Rx pressure is <1071 psig, Rx water level is >
-38',
and the 30 (1.0)
second time delay has timed out.
REFERENCE HC LP 24, RRCS a. Learning Objective 3, page 19 b.
LO No. 4a, page 14 c. LO No. 6b, paSe 48 ( a,\\{ y m W %. A tt'0.L A W di d ouM r
- 6,ift n' k d 8 " TO C C 8 l O d C d I 0 h O f C O#y*
ANSWER 6.02 (2.25)
/'
y G 0.G1F
.
Lof 5 a. 1. To calculate gain adjustment factors for LPRMs (as detectors-iv.5)
o.s tr
deplete uranium).
'(d 1 2. Calculates substitute data for out of service LPRMs
-
.51
'^
"
b. The TIP purge system maintains the relative humidity in the guide (0 5)
tubes at a constant value, and maintains a dry atmosphere in the drive mechanism and indexer enclosure.
Corrosion products could build up on the helical wrap around the (0.5)
c. drive cable and cause an electrical signal loss.
(
REFERENCE g,wc( /Or M C M W M HC LP 18, TIP System, pages 14 and 20 ygygnyg,
ANSWER 6.03 (2.00)
1. Reactor water sample valves close 2. Mechanical Vacuum pumps trip if running o.C7 3.
Reactor scram l
. C1_.d
>=.1 wiw.-
..ul itun (4 needed; (k/Ieach)
'
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
ANSWERS -- HOPE CREEK-86/02/24-KOLONAUSKI/LANGE pgg y N o p, y m ge3r A Y M s M d o p !C M REFERENCE LP 45, pg. 20 g.g,
> de.tg.
LP 53, pg. 15 LP 22, pg. 61 ANSWER 6.04 (3.00)
c.
10 kW ("o rating heater') - Maintains solution temperature to (0.5)
prevent ecipitation of the sodium pentaborate during operations.
40 k (*. mixing heater") - Used to heat and maintain water temp-(0.5)
er ure during solution preparation.
b Remote Control Panel 10C011; WILL NOT (0.33)
Control Room Panel 10C6513 WILL (0.33)
RRCS Panel; WILL --+ If PtY'"U3I*
(0.33)
c.
The SLC Storage Tank Low Level interlock (0.33) will trip and prevent starting of all SLCS pumps (0.33) when the tank level reaches 2.5" above the centerline of pump outlet connections. (0.33)
M.
e me no M M $a h.
REFERENCE a. LP 23, Learning Objective 4, page 11 b. LP 23, Learning Objective 8, page 17
'
c. LP 23, Learning Objective 7, page 15 ANSWER 6.05 (1.50)
a. An overspeed condition may still occur due to the flashing of (1.0)
the moisture in the moisture separators. gwd (ac.fu.d WaM-( kJ.alttYf 0.5)
b.
The combined intermediate valves. or kittdtr My valvo. (tKtYRChan REFERENCE swen-ntwrn Gdtta )
LP 48, Learning Objective 25, paSe 58 ANSWER 6.06 (2.00)
1. High Drywell Pressure (1.68 psis)
(0.5)
2. Low Rx Water Level (Level 2 or -38')
(0.5)
3.
High Rad on Refuel Floor Ventilation Exhaust (0.5)
4. High Rad on Reactor Buildin3 Ventilation Exhaust (0.5)
REFERENCE LP 40, Learnin3 Objective 8, page 11
_
. _ _ _.
_
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTA1 ION PAGE
ANSWERS -- HOPE CREEK-86/02/24-KOLONAUSKI/LANGE ANSWER 6.07 (2.25)
g g(g-
c. Shutdown cooling suction head ation valves F008, F009 and (0.75)
IN head spray inboard isolat'
lve F022 will auto close. The RHR pump will trip due to ss of suction path.
H2 5"b. Same response a.
(0.75)
-40" c. The shu un cooling return isolation valves HV-F015A, B and the (0.75)
out d head isolation valve F-23 will auto close.
REFERENCE LP 28, page 87 HC Questions 8 and 11 ANSWER 6.08 (2.00)
a. No Change (0.25); the RCIC flow control circuitry will maintain actual flow rate equal to demanded flow rate (0.75).
b.
Increase (0.25); in manual, the RCIC flow control circuitry will maintain turbine speed equal to desired turbine speed. As reactor pressure drops, RCIC flow rate will increase (0.75)
REFERENCE LP 30, page 65 6.07 a-
') 52-y3 i o), SDC initt site y kuevde c( -
ggg y gy y 4, (o g of suctTm yah
[ HV - Foo S, o t (w,og sec suctfen h v - F0 22,2 3 i n. od Mud $ yr isotahm Cl0$l S0c v4.tum vs.lv%
( HV - PC W AJ w,;;.wyu u:'em
. - - -
h-
...,
g;_ u;~ 0-men;=w=a samt a 3 c.
s,w< o.i J 6.
PLANT SYBTEMS DESICH, CONTROL, AND INSTRUMENTATION PAGE
______________________________________________________
ANSWERS -- HOPE CREEK-86/02/24-KOLONAUSKI/LANGE ANSWER 6.09 (2.00)
c. Back Up Hoist Limit:
This lamp lights only if the normal maximum (0.5)
up limit limit fails and the hoist is stopped by the backup hoist limit switch.
b. Rod Block Interlock No. 1: Occurs when a fuel assembly load is on (0.5)
any hoist and refuel switch 41 is activated when the refueling platform is over the vessel.
c.
Fuel Hoist Interlock: Indicates a condition when the platform is (0.5)
over the reactor, a control rod is withdrawn, and the grapple is loaded.
d. Bridge Reverse Stop No. 1: Prohibits bridge travel toward the (0.5)
reactor when a signal from the control room indicates that a control rod is withdrawn, the platform is on a switch indicating that the platform is about to move over the reactor, and a load is on any of the hoists.
REFERENCE GE Refueling Tools Familiarization Hanval
- 2, pages 3-12, 3-13 ANSWER 6.10 (1.00)
Od/Or b.
(1.0)
e.
REFERENCE HC LP 19, pages 22,32 Figure 18, LO 10,13 ANSWER 6.11 (1.00)
d.
(1.0)
REFERENCE HC LP 14-01, pages 9,10 LO 4.b
_.
_
.-
6.
PLANT BYSfEhs DESIGN, CONTROL, AND INSTRUMENTATION PAGE
_,-__- -__--____------_-___--___---____-_____,,
ANSWERS -- HOPti 6 REEK-86/02/24-KOLONAUSKI/LANGE ANSWER 6.12 (3.00)
1. Isolation (due to hi steam flow or equip area temp.) causing steam IV's and suppression pool suction valve (042) to close.
(1.00)
2. Turbine Trip due to isolation (1.00)
3. Vacuum breaker IV's close (due to low steam line pressure and i
(1.00)
h f awefHr Wa.k dJ.ygsf ung & My in 100 ys(3 hi DW) - ont N
"
'
'
REFERENCE HOPE CREEK LP16 pgs. 20,70,71, I.O.47 FJC259
- - _.
--_
_
. - - -
=_ ~-
.-
.-
,
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
~~~~~~~~~~~~~~~~~~~~~~~~
~~~~EI656L665C5L'66UTR6L
____________________
ANSWERS -- HOPE CREEK-86/02/24-KOLONAUSKI/LANGE ANSWER 7.01 (1.00)
By maintaining the water level above the minimum for natural cir-(1.0)
culation, adequate coolant mixin3 and core cooling is assured.
REFERENCE OP-SO.BC-001(0) RHR system procedure, Precaution 3.11.1 MC SO LP, LO 1 3. gyps 3 N&h gg,grcrM*
e ent.
t1 ANSWER 7.02 (2 50)
2.ofja.
1. To backup the scram with add ional RPS contacts (0.25)
2. To keep the HSIV's open by moving out of RUN, and therefore (0.25)
preventing the low pressure isolation.
-
9:8'
o.12 5 0.115-b.
1. Pressure stabilized,and RPV level stabilized above TAF -AND-(0.25)
g All control rods inserted to or past 02, OR
)
(0.25)
M N
,*a.
b.
690 lbs of boron have been injected into the RPV, OR (0.25)
c.
The reactor is shutdown and no baron has been lnjected.
(0.25)
2. RPV level between +12.5" and 54'
(0.33)
-{ vere 100 : RPV Pressure below 1037 psi 3 (0.33)
Reactor shutdown (0.33)
REFERENCE Scram procedure LP, LO 1.3, ps. 6 Post Scram Recovery LP, LO 41.2, ps. 5 p cittr W(L.<. SMckgurb.
d to S.u_ if % (NTENT oF & Y"" h W g t, 2 - p( r t.
Mt.
.
H wdyt W. bdby t cby % poi,,4-clis tyi Mco fw 7. 0 2 b.I :
l-NX h o.33
!
.
j 2. M s h if o 33 h
3. YkS* %(d 0 33 e
s I
l l
l l
r
_.. _
__
_ _..
__.
._
_,
,
_-_
- - - - - - - - - - - - _ _ _ _ _ _ _ _ _ _ _ _
.
(
'
o 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
-
~~~~~~~~~~~~~~~~~~~~~~~~
~~~~ d656L6GiCdL 56UTR6L R
____________________
ANSWERS -- HOPE CREEK-86/02/24-KOLONAUSKI/LANGE 0.'!I ANSWER 7.03 (3.00)
c. If an SRV were opened wit the supression pool level below the (1.0)
chamber and drywell would result. (pressurization of the supression top of the SRV T-quenchers, rapid It is possible that the design k05 pressure for the containment would be exceededy g7 gg An.Ladr, pd. Mt6W (1,o)
b.
Main Condenser MSL Drains ihML C "
'
Head Vent HPCI RCIC SJAE RFPT (0.25 each, 4 needed for full credit)
c.
1. 50 psig is the lowest differential pressure at which an SRV
'O.75)-
will remain fully open with its control switch placed in the
open position.
2. Below 50 psis, RPV Depressurization is considered complete
-- 4 0. 2 5 )
and additional steam paths are not necessary.
c.T REFERENCE OP-EO.ZZ-202 LP a.
pages 6, 7 / b.
page 8 / c.
page 8 ANSWER 7.04 (2.00)
1.
Suspend all refueling operations.
(0.5)
2. Evacuate all unecessary personnel from the Refuel Floor.
(0.5)
3.
Ensure the Reactor Building Ventilation isolates.
(0.5)
4.
Ensure that the Filtration, Recirculation, and Ventilation (0.5)
system (RFVS) auto starts.
REFERENCE Irradiated Fuel Damage While Refueling (OP-AB.ZZ-101)
AB LP, LO 1 ANSWER 7.05 (1.50)
a. This action will prevent pulling cold air in along the Main (0.5)
Turbine shaft.
b. This action will minimize the thermal transients on the RPV.
(0.5)
c. Mode switch in Startup/ Hot Standby, and Reactor Coolant at any (0.5)
temperature
-
-
.
.
I
7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
^^~~ I65UL6G5CdL 66UTR6L
~
~~~~~~~~~~~~~~~~~~~~~~~~
R
____________________
ANSWERS -- HOPE CREEK-86/02/24-KOLONAUSKI/LANGE REFERENCE Operations from Hot Standby procedure I0-007, Cautions 5 2.19.1 and 5.1.8 Hope Creek Tech Specs, Table 1.2 IO LP, LO 1 ANSWER 7.06 (2.00)
c.
1.
Reduce power as necessary to maintain the Offgas activity (0.5)
less than the high high alarm setpoint.
2. If a scram condition is reached, ensure that the reactor (0.5)
scrams and implement OP-EO.ZZ-100, b. Monitor the SJAE for an increase in flow.
(0.5)
Monitor the Main Condenser for a decrease in vacuum.
(0.5)
or mon (to r 5 ) A E 3rd ; N* / h i ^ h
.
REFERENCE OP-AB.ZZ-127 AB LP, LO 1 ANSWER 7.07 (2.50)
c.
1.
Scram condition (0.25) and reactor power >5% or undetermined (0.25).
2. RPV water level below -38'
(0.25) or undetermined (0.25).
3. Reactor pressure above 1037 psis (0.5).
4.
Drywell pressure above 1.68 psis (0.5).
b.
Misoperation means* inappropriate initiation (0.25) or continued oper-ation beyond the trip setpoints (0.25)
,
REFERENCE o. OP-EO.ZZ-101 RPV Control, RPV LP: LO 1.2 b. OP-EO.ZZ-100 Scram LP pa3e 7 ANSWER 7.08 (3.00)
o. To maintain an accurate indication of reactor period.
(1.00)
b. Increase pressure setpoint to prevent unplanned depressurization. (1.00)
c.
o minimize thermal transients on the reactor vessel. Opening a bypass valve may be necessary to acheive steady feedwater flow. (1.00)
L o.5
.. _ _
.
=
-
.
.
.
.
._.
7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
~~~~~~~~~~~~~~~~~~~~~~~~
~~~~ d656L66565E"66 TR6L R
____________________
ANSWERS -- HOPE CREEK-86/02/24-KOLONAUSKI/LANGE REFERENCE Hcpe Creek OP-IO-ZZ-003 pg.
9, cautions 3.2.5, 3 2.6 FJC277 g CMTramL Syb ANSWER 7.09 (3.00)
o. Skimmer surge tank makeup valve HV-4660 opens Fuel pool pumps trip RBVS isolates FRVS auto starts (4 rqd e.5ea)
b. Terminate, refueling activities Evacuate the refuel floor (2 eqd e.5ea)
REFERENCE Hope Creek procedure OP-AB.ZZ-144 ANSWER 7.10 (3.00)
1. CORE SUBMERGENCE accomplished by maintaining a level above TAF.
(1.0)
2. SPRAY COOLING accomplished by using at least one CS system operating (1.0)
whtas Kx pnts ) 700 yt()t at design conditions.
I c te cie:r cidre, fi c.
3. STEAM COOLING accomplished by using3SRV tc (g,o)
4....__.__.
2 _, _.. _ _ ____ ___,
, _ _....
~ ~ ~ ' ' ~~ g [ d 'y U d < i 5'O' yr(
O IAdi yhiM ww ut bcO W.
.....
d
"
REFERENCE (3.00)
Hope Creek definition of adequate core cooling in E0P's ANSWER 7.11 (1.50)
0 16
/
o. Keeping the MSIVs open will allow you to maintain the Main Condenser as a heat sink, and minimize the chance of reaching the Heat Capacity (0.5)
intheSgpgessionChamber.
Temperature Limit (HCTL)
The PCIG low level isolation is bypassed to ensure a continued pneumatic supply to the inboard MSIVs.
(0.5)
(0.5)
b.
Yes
_ _ _ _
.
7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
R 656L55iEAL E5sTRUL'~~~~~~~~~~~~~~~~~~~~~~~
-
~~~~
'
____________________
ANSWERS -- HOPE CREEK-86/02/24-KOLONAUSKI/LANGE REFERENCE OP-EO.ZZ-207 LP pages 9,
1
i I
t i
d
'
!
l
.
. -. - -.
.. -.
. _ _ _. _ _ _ _... _, _ _ _. _ _ _ _ _ _. _ _ _ _. _ _.,,_.___..
,_.. __.. _. _
_ _.. _. _, _ _ _.,..
. _., _ _ _ _ _ -.. - _ _ _ _,.. _. _. _ _.
.
,
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE
__________________________________________________________
ANSWERS -- HOPE CREEK-86/02/24-KOLONAUSKI/LANGE ANSWER 8.01 (3.00)
Since the TIP System provides the computer with the data for calculating the thermal hydraulic limits, we must assume that the curveillance of APLHGR, MCPR, and LHGR cannot be made. When the required monitoring of these limits cannot be met, the following
- 7I cetion statements apply:
.38 (0.fS$
c. Initiate corrective action within 15 minutes.
( 0. 7&) ** *
b. Restore the limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, c. or reduce power to less than 25% RATED THERMAL POWER within (0.79).60 the next four hours.
,,g (0.75).d*
The most limiting LCO is a.
T. S 1 3 4.~1 a p ply 's (o,ga)
"
REFERENCE TS Section 3/4.2, Power Distribution Limits TS LP, LO 2 TJ 337.7 ANSWER 8.02 (2.00)
o. The intent of the original procedure is not altered.
(0.67)
b. The change is approved by two members of the unit management (0.67)
staff, one of which holds an SRO license on the unit.
c. The change is documented and receives the same level of review (0.67)
and approval as the original procedure within 14 days of implemen-tation.
REFERENCE TS 6.8.3 ANSWER 8.03 (1.50)
ylate tho thop chAWb4 l M
.
ir et Iract MOT C""TC0"N The action required by Table 3.3.1-A is to.:
withi.. th:.:: t 12 h mm e - tW. ty(ypcl carditan Wihin 9 h0hrt.
(1.50)
REFERENCE TS 3/4.3.1, RPS Instrumentation t
'
r
-_.
-
-
.,.
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE
__________________________________________________________
ANSWERS -- HOPE CREEK-86/02/24-KOLONAUSKI/LANGE ANSWER 8.04 (2.50)
ggjtty gg, (Ch ek wi h Training Department )
c.
1. Con m'istion levels greater than 10,000dpa/100cm2 2. Main nance of equipment where rad levels are >5 mrem /hr 3. His ad Area entries 4. Un.no n conditions in an area to be entered 5.
utro radiation exposure >5 arem/hr radioactivity requiring the use of respiratory protection 6. Airbor equipmen (4 needed at 0.4 each)
.
(0.9)
b. Get SS responsibilities from training department.
REFERENCE Red Access Control Program (RAC) SA-AP.ZZ-046 (0)
ANSWER 8.05 (1.50)
EDO - Oyr furnayr (0.5)
a. SNSS Oyt 6vpr W,etc..
(1.0)
b. EDO, E R M --- EC Eyan -
Sr. W WWCASar dC-HC. (:tntd b$ty g
REFERENCE a. OP-AP.ZZ-002 (0) 5 2.3 b. ECG Attachments ANSWER 8.06 (3.00)
Due to the fact that this instrument functions as one of the Core Spray permissive signals (0.25), the minimum operable channels per trip system requirement of T.S. Table 3 3.3-1 for the CS system cannot be met for one trip system (0.75).
Therefore, in accordance with TS 3.3.3.b, the action of Table 3.3.3-1 must be declared inop. (1.0)
This action requires that the CS e(S8p>3.5.1.a.$ must be taken.
be taken.
Having CS inop requires that the actions of T.S.
(1.0)
[
REFERENCE I {0dy of CI (hty Hope Creek Technical Specifications TS LP, LO 2
.
< - -
-
--
,
,
- _ - -.
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE
__________________________________________________________
ANSWERS -- HOPE CREEK-86/02/24-KOLONAUSKI/LANGE ANSWER 8.07 (2.50)
g,3 M-re 3.3.4.2.a, 3.3 4.2.e 1 25-44,H 3.3.1.b and Table 3.3.1-1 Action 6 1.
A
-.m_
3.3 ; ;, ; ;,,,,,, v t a 3 ;
(0, ;;
n;t 1;_.
..g-
..
-
Hope Creek Tech Specs Ray 1,4 TV'OE - th h can or a. re.cand yarg REFERENCE vty(Eca.han for a. ihrotHt. valvt.
TS LP, LO 2 0Wrw(54. fatu 2.0 ANSWER 8.08
-'I (0.5)
o. False b. Hove the valve's handwheel in the closed position just enough (0.5)
to verify valve movement, then return to open.
c. Hove the valve's handwheel in the closed direction only.
(0.5)
d.
^ r t i e r-ctetr:.cnt L:3; E7;ir :ent Un /cil:ble Li;t gg
' O. "; )
.
(0 5)
e. True REFERENCE OP-AP.ZZ-109 ' Equipment Operational Control'
o. 5.2.3 b. 5.2.6 c.
5.2.5 OP-AP.ZZ-108 ' Removal and Return of Equipment to Service'
d. 5.1.5.2 e. 5.3.1 ANSWER 8.09 (1.50)
1. A maximum allowable extension not to exceed 25% of the surveil-(0.75)
lance interval.
2. The combined time interval for any 3 consecutive surveillances (0.75)
shall not exceed 3.25 times the specified surveillance interval.
REFERENCE Hope Creek Tech Specs Surveillance Requirements, 4.0.2 a and b.
.-
-
-
.
_
_
.-.
.- -
-
.-.-
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE
ANSWERS -- HOPE CREEK-86/02/24-KOLONAUSKI/LANGE ANSWER 8.10 (3.00)
c. Unusual Event (1.0)
Alert Site Area Emergency General Emergency b.
1. Yes (0.5)
2. Yes (0.5)
3. No (0.5)
4. Yes (0.5)
REFERENCE Emergency Classification Guide SA.AP.ZZ-006 ANSWER 8.11 (2.00)
c. Yes (0.5), following putting RPS trip system A in the tripped position (0.5) as per 3.3.1.a b.
1.
No (0.25), IRMs are not required in Cond 1 and you may stay there (0.25).
No (0.25),
elect ye/ 5:d th:
"'"'S trip syste; ^ logic i r-the 2.
Ves teirped pssiticr- '0:25)
S r.e c 4 + 4 r e t i e r- ?.0.5 i: ;t :pplic:ble.
REFERENCE HC TS, 3/4 0-1, 3/4 3-1 to 3-4 k. 2. M (L(.hm of McVin$ 94 V"4M JWiM M Mft N U " b A TT VCOLAhp -
gg - o m i n J twi-Wy. 9cwox(no.c.hm sb. W 3T14 W I " A '.
gu p he eng L opeh (I2M.r f*r
__.
.. -
.
--
_ _
_.
..
-.
-
.
TEST CROSS REFERENCE PAGE
QUESTION VALUE REFERENCE
________
______
__________
05.01 3.00 BAN 0000001 05.02 1.00 BAN 0000002 05.03 1.50 BAN 0000003 05.04 2.00 BAN 0000004 05.05 3.00 BAN 0000005 05.06 2.00 BAN 0000007 05.07 1.50 BAN 0000011 05.08 1 50 BAN 0000146 05.09 1.00 BAN 0000203 05.10 2.00 BAN 0000213 05.11 3.00 BAN 0000224 05.12 2.50 BAN 0000225 05.13 1.00 BAN 0000226
______
25.00 06.01 3.00 BAN 0000024 06.02 2.25 BAN 0000025 06.03 2.00 BAN 0000026 06.04 3.00 BAN 0000027 06.05 1.50 BAN 0000028 06 06 2.00 BAN 0000030 06.07 2.25 BAN 0000033 06.08 2.00 BAN 0000034 06.09 2.00 BAN 0000035 06.10 1.00 BAN 0000205 06.11 1.00 BAN 0000206 06.12 3.00 BAN 0000228
______
25.00 07.01 1.00 BAN 0000150 07.02 2.50 BAN 0000151 07.03 3.00 BAN 0000186 07.04 2.00 BAN 0000208 07.05 1.50 BAN 0000216 07.06 2.00 BAN 0000218 07.07 2.50 BAN 0000219 07.08 3.00 BAN 0000229
07.09 3.00 BAN 0000230 07.10 3.00 BAN 0000231 07.11 1.50 BAN 0000233
______
25.00 08.01 3.00 BAN 0000014 08.02 2.00 BAN 0000017 08.03 1.50 BAN 0000018 08.04 2.50 BAN 0000020 08.05 1.50 BAN 0000021
..
. - _. _
.-
--
-
.
_ _..
.
TEST CROSS REFERENCE PAGE
OUESTION VALUE REFERENCE
08.06 3.00 BAN 0000148 08.07 2.50 BAN 0000149 08.08 2.50 BAN 0000217 08.09 1.50 BAN 0000220 08.10 3.00 BAN 0000221 08.11 2.00 BAN 0000234
25.00
_-_---
100.00
.
l
l
.
y e
-
-r-y e
-
w-gm-.
e.
wj*
---i,
-
-
e-.,.
. -
- --
. I
-
...
!
.
t
-,
,
.
j Table 1.
Saturated Steam: Temperature Table
__
- -. -.
- -...
l Temp lb per Sal.
Sat.
Sat Sat.
Sat.
Sat.
Temp falu Sg in 1iquid Evap Vapor Lic uid Evap Vapor Liquid Evap Vapor Fahr
'
s I
i p
vg
_ig, vg if h fg h
sg sgg a
v g
32 8 0 08859 0 016022 33047 3304 7 0 0179 1075.5 1075.5 0.0000 2.1873 2.1873 32A
34 0 0 09600 0 016021 3061 9 3061.9 1.996 1014.4 1076 4 0.0041 2 1762 2.1802 34.0
,
36 8 0 10395 0016020 28390 28390 4 008 1073.2 1077.2 0.0081 2.1651 2.1732 MA l
34 0 011249 0 016019 26341 2634 2 6.018 1072.1 1078.1 0.0122 2.1541 2.1663 38 4 40 0 1.12163 0 016019 2445 8 2445 8 8.027 1071.0 1079.0 0.0162 2.1432 2.1594 40.0 42 8 013I43 0 016019 22724 / 2272.4 10 035 1069 8 1079.9 0.0202 2.1325 2.1527 42A 0 14192 0 016019 21128 2112 8 12.041 10681 10803 0.0242 2.1217 2.1459 44A 44 8
-
48 8 015314 0 016020 19657 19657 14.047 1067.6 1081.6 0 0282 2.1111 2.1393 48A 48 0 0 16514 0 016021 1R30 0 1830 0 16 051 1066 4 1082.5 0.0321 2.l006 2.1327 40A l
54 8 0 17796 0016023 1704 8 1704 8 18 054 1065.3 1083.4 0.036l 2.0901 2.1262 50A l
52 0 0 19165 0 016024 15892 1509 2 20 057 1064.2 1084.2 0.0400 2.0798 2.1197 52.0 54 I O20625 0016026 1482 4 1482.4 22.058 1063.1 1065.1 0.0439 2.0695 2.1134 54.0 58 8 0 22183 0016020 1383 6 1383.6 24 059 1061.9 1086.0 0.0478 2.0593 2.1070 58.0
?
58 0 0 23843 0016011 1297 2 1292.2 26 060 10601 1086.9 0.0516 2.0491 2.1000 50 0 l
80 8 0 2561!
0016033 1207.6 1207 6 28 060 10591 10871 0.0555 2.0391 2.0946 80.0 82 I 027494 0 016036 II29 2 II29.2 30 059 1058.5 1088.6 0.0593 2.0291 2.0885 62.0 64 0 0 29497 0 016039 1056 5 1056 5 32.058 1057.4 1089.5 0 0632 2.0192 2.0824 64.0 88 8 0 31626 0 016043 989 0 989.1 34.056 1056.3 1090.4 0.0670 2.0094 2.0764
$$A etI 033889 0 016046 976 5 926 5 36.054 1055.2 1091.2 0.0708 1.9996 2.0704 50A
,
70 0 0 36'32 0 016050 868 3 868 4 38.052 1054.0 1092.1 0.0745 1.9900 2.0645 10A 12 0 0 38844 0 016054 814 3 814 3 40 049 1052.9 1093.0 0.0783 1.9804 2.0587 72.0 14 0 0 41550 0 016058 164 1 7641 42.046 1051.8 1093.8 0 0821 1.9700 2.0529 14.0 76 3 0 44420 0 016063 117 4 717.4 44 043 10501 10943 0.0858 1.9614 2.0472 76.0
,
'
is e 04746I n016067 611 A 613 9 46 040 1049.5 1095 6 0.0095 1.9520 2.0415 70.0
,
00 0 0 50683 0 016072 633 3 633 3 48 037 1048.4 1096.4 0.0932 1.9426 2.0959 08.0 82 8 0 54093 0 016077 595 5 595 5 50 033 10473 1097.3 0.0969 1.9334 2.0303 02.0 MS 057702 0 016082 560 3 560 3 52 029 1046.1 1998 2 0.1006 1.9242 2.0248 M.0 88 0 0 61518 0 016087 227 5 527.5 54 026 10450 1099 0 0.1043 1.9151 2.0193 06.0 00 0 0 65551 0 016093 4% 8 4968 56 022 1043.9 1099.9 0.1079 1.90E0 2.0139 00A 30 0 0 69813 0 016099 4681 468I 58 018 10423 1100 8 0.1115 1.8970 2.0006 08.0
!
02 e 074313 0 016105 441 3 441 3 60 014 1041 6 1101 6 0.1152 1.8081 2.0033 02A 94 0 0 19062 0 016111 416 3 416.3 62 010 10405 1102 5 0 1l88 IA792 1.9900 M.8
>
98 I 084012 0 016111 392 8 392 9 64 006 1039 3 1103.3 01224 1.8704 I9928 95.0 L,
90 0 0 09356 0 016173 370 9 370 9 66 003 10382 1104 2 0 1260 1A617 1.9876 08 8
.-
.-.
--
- _ _ _ - - --. _
_ _
-
.
--
-
.-
l L
,
-
?
Ahs Psess Sperific Volume Enthalpy Entropy Temp tb pe Sal.
Sal.
Sal.
Sal.
Sal.
Saf.
Temp
i alu Sq in liquial Ivap Vapos liquid Evap Vapor liquid Evap Vapor Fahr h l h ig h
s, sgg s
I I
p
'
vi vig vs g
a tes e 094924 0 016130 350 4 350 4 67.999 1037.1 11051 0.1295 1A530 l.9825 1984 102 8 100789 0 016137 331 1 331 1 69 995 1035.9 1105.9 0.1331 1.8444 8.9775 102.0 IN I I06965 0 016144 313 1 313 1 71.992 1034.8 1106.8 0.1366 IA358 I.9725 108.8 les e i1347 0 016151 296 16 29618 13 99 1033 6 1107.6 0.1402 1.8273 19675 188A 100 e 1.2030 0 016158 280 28 200 30 75 98 1032.5 1108.5 0.1437 1.8188 1.9626 1984 lieI 1.2750 0 016165 2F 265 39 71.98 1031.4 1109.3 0.1472 1.8105 I.9577 198A 112 0 13505 0 016113 M : ;7 25138 79 98 1030.2 1110 2 01507 I8021 1.9528 112.8 114.8 1.4299 0 016180 238 21 238 22 81.97 1029I 1111.0 0.1542 13938 1.9480 114A IIII 15133 0 016188 225 84 225 85 83 97 1027.9 1111.9 0.1577 13856 1.9433 118.0 lit e 16009 0 0161 %
214 20 f 214 ?!
85.97 1026.8 1112.7 0.161I I1 774 1.9306 118A 120 0 I6927 0016204 20325 203 26 87.97 1025 6 1113.6 0.1646 1.7693 1.9339 128 i
122 0 17891 0 016213 192 94 19295 89.96 1024 5 1114.4 0.1600 1.7613 3.9293 122A
'
124 8 1 8901 0 016221 18323 183 24 91 %
1023 3 1115 3 0.1715 13533 1.9247 Ing 128 8 I9959 0016229 114 08 114 09 93 96 1022.2 Il16.1 0.1749 13453 1.9202 SES
'
178 8 2.1068 0 016238 15545 16547 95.%
1021.0 1117.0 0 1783 13374 1.9157 IES
!
130 0 2 2230 0 016247 15732 157 33 97.96 1019.8 1117.8 0.1817 13295 1.9112 I3RA 132.0 2.3445 0 018256 149 64 149 66 99 95 10187 1118 6 0.1851 13217 I.9068 132A
'
134 0 2 4111 0 016265 14240 142.41 101.95 1017.5 1119.5 0.1884 13140 1.9G24 Int
'
i 136 I 2 6041 0 016214 135 55 135 57 103 95 1016 4 1820.3 0.1913 13063 1.8900 1ES 130 0 2 1438 0 016284 129 n9 129 11 105 95 1015.2 1121.1 0.1951 1.6906 IA937 IES i
!
140 0 2 8892 0 016293 122 98 123 00 107.95 1014 0 1822.0 0.1985 1.6910 13895 148A
142 8 3 0411 0 016303 1I721 11722 109.95 1012.9 1122.8 02018 1.6534 1.8852 142.0 144 0 3 1997 0 016312 111 74 111 16 111.95 10113 1123 6 0.2051 1.6759 1.8810 144.0 j
I48 8 3 3653 0 016322 106 58 106 59 113 95 1010.5 1124.5 0 2084 1.6684 1 8769 148.8 les e 3 Mel 0 016317 101E8 10130 115 95 1009.3 1125.3 0.2117 1.6610 1.4727 140A
'
ISO O 3 7184 0016343 9705 9707 117.95 1008 2 1126.1 0.2150 1.6536 1.0646 120 152 8 3 9065 0 016353 9266 9268 119 95 1007.0 1126.9 0 2183 1.6463.l.8646 152.0 1548 4 1025 0 016363 8850 88 52 121 95 1005 8 11273 02216 1.6390 1.8606 1R0 i
154 8 4 3068 0 016314 84 56 84 57 123 95 1004.6 11284 0.2248 1.6318 1.8566 IES
158 8 4 5197 0 016384 80 82 80 83 125 96 1003.4 1129.4 0 2281 1.6245 1.8526 ISSA
letI 4 1414 0 016395
?? 27 7/29 127 96 1002.2 1130 2 0.2313 1.6174 1.8487 ISSA j
182 8 4.9722 0 016406 13 90 73 92 129.96 1001.0 1131.0 0.2345 1.6103 13448 102.0
1640 5 2124 0 016417 70 10 70 72 131.96 999 8 1131.8 02377 1.6032 1.8409 IKO
.
188 3 54623 0 016428 6167 6768 13397 998.6 1132.6 0 2409 1.5961 IA371 1ESA l
let I 57223 G016440 64 18 6480 13597 997.4 1133.4 0.2441 1.5N2 IJ333 ISSA
)
lit t 5 9926 0 016451 62 04 62 06 137.97 996.2 1134.2 0.2473 1.5822 1A295 128 172 8 6 2136 0 018463 59 43 5945 139 98 995.0 1135.0 0.2505 1.5753 1A258 172A l
174 8 6 5656 0 016474 56 95 56 97 141.98 993 8 1135.8 0 2537 1.5604 1A221 INA
<
IIIe 68690 0 016486 54 59 54 61 143 99 992.6 1136.6 0.2568 1.5616 1.8184 1R0 lit e 7.1840 0 016498 5235 57 36 145 99 991.4 1137.4 0.2600 I.5548 IA147 128
/
~
@
l
a
.
s
.
o.
Abs Press Specific Volume Enthalpy Entropy Temp tb per Sal Sal Sal Sat.
Sal.
Sal.
Temp Iahr Sq in.
liqmd
[vap Vapor liquid Evap Vapor Liquid Evap Vapor Fahr he h gg h
s, seg st I
va g
I p
vi vtg les s 15110 0 016510 5021 50 22 14800 990 2 11382 0 2631 1.5480 1.8111 100.4 182 8 1850
'0016522 48172 18 189 150 01 989 0 1139 0 03662 1 5413 1.0075 182.0 184 8 8 203 0 016534 46 232 46 249 152 01 987.8 1l39 8 0.2694 I.5346 I8040 IM.0 i
tes t 8 568 0 016547 44 383 44 400 154 02 986 5 1140 5 0 2725 1.5279 1.8004 186.8 s
tes e 8 947 0016559 42 671 42 638 156 03 985 3 1141 3 0.2756 1.5213 13%9 100.8 let t 9 340 0 016572 40 94 [
40 957 158 04 9 84.1 1142.1 02787 1.5148 1.7934 198.8 132 8 9 747 0 016585 39 337 39 354 160 05 982.8 1142.9 0 2818 1.5082 13900 192.0
'
194 0 10 168 0 016598 37 808 37 824 162 05 9816 1843 7 0:2848 1.5017 13865 IM.0 198 9 10 605 0016611 36 348 36 364 164 06 980.4 II44 4 0.2879 1.4952 1.7831 195.0 196 I 11058 0 016624 34 954
!4 970 166 08 979.1 11452 02910 1.4888 I.7798 198.8 264 I 11526 0016637 33 622 33 639 168 09 977.9 1146.0 03940 1.4824 13764 200.0 284 8 12 512 0 016664 31 135 31 151 172 11 975 4 1147.5 0 3001 1.4697 13698 2M.0 288 I 13 568 0016691 28 862 28 818 17614 972 8 1149 0 0 3061 1.4571 13632 200.0 212 8 I4 696 0016719 26 182 26199 18017 970 3 1150 5 0312I 1.4447 13568 212.0 215 0 15 901 0 016747 74 R78 24 894 184 20 967 8 1152.0 03181 1.4323 13505 216.8
!!t I 17 186 0 016775 23 131 23 148 188 23 965 2 1153.4 0 3241 14201 1.7442'
229.8 d
2248 18 556 0 016805 21 529 21545 192 27 962.6 1154.9 03300 1.4081 13380 224.0
228 8 20 015 0016834 20 056 20 073 196 31 960 0 1156.3 0 3359 13961 13320 228 0
2328 21 567 0 016864 18701 18 718 200 35 957.4 1157.8 0 3417 13842 13260 232.0 236 8 23 216 0 016895 17 454 17 471 20440 954 8 1159.2 0 3476 13725 13201 236 0 244 I 24 % 8 0 016926 14 304 16 321 208 45 952.1 1160 6 0 3533 1.3609 13142 244.0 2448 26 826 0 016958 15243 15260 212 50 949 5 1162.0 0 3591 1.3494 13005 244.0 248 I 28 796 0 016990 14 264 14 281 216.56 946 8 1163.4 0 3649 13379d 13028 2488 252 8 30 883 0 0170/2 13 358 13 375 22062 944.1 11643 03706 13266 16972 252.0 254 0 31 091 0 017055 17 570 12 538 224 69 941.4 1166.1 0 3763 13154 16917 254 8 294 0 35 427 0 011089 11145 11762 22816 938 6 1167.4 0 3819 1.30(3 16862 260 0 284 0 37 894 0 017123 11 025 11 042 232 83 935 9 1168 7 0 3876 12933 I6808 264.0 284 8 40 500 0 017157 10 358 10 375 236 91 9331 1170 0 0 3932 1.2823 1 6755 260.0 272.8 43249 0 017193 9138 9 755 24099 930 3 1171 3 0 3987 1.2715 1.6702 272.0
lit I 46147 0 017228 9 162 9 180 245 08 9215 1172 5 04043 1.2607 I6650 276.8 I
200 0 49 200 0 017264 8 627 8 644 249 17 924.6 11738 04098 1.2501 1.6599 200.9 2848 52 414 Ofl130 81280 8 1453 253 3 9213 1175 0 0 4154 1 2395 1.6548 2M.0 280 0 55 795 0 01734 76634 7 6807 2574 918 8 1176 2 0 4208 1.2290 1.6498 290.0
,
292.8 59 350 0 01738 72331 72475 261.5 915 9 1177.4 0 4263 1.2186 1.6449 292.8
?
298 8 63 084-0 01141 6R759 6 8433 265 6 913 0 1178.6 04317 1.2002 16400 296.I i
'
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,"
__ _ _
___
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.
_ _ _ _
_. _
O i
Y O)
,
Abs Press. ' ~
"
~$pecific Yolume Enthalpy Entropy Temp Lb per Sal Sat.
Sat.
Sal.
Sat.
Sal.
Temp Iahr Sq in liquid Evap Vapor Liquid Evap Vapor L6 quid Evap Vapor Fahr I
he h gg h
sg seg s
vs g
a I
p vg vf(
see g 67 005 0 01145 6 4483 6 4658 269.7 910 0 1179.7 04372 1.1979 1.6351 300.9 M48 71119 0Cl149 6 0955 6 1130 273 8 907.0 1180.9 04426 1.1877 1.6303 304 8 388 8 75 433 0 01153 5 7655 5 7830 278 0 9040 1182.0 0.4479 1.1776 1.6256 300.8 312 4 79 953 001157 54566 54142 2821 901.0 1183.1 0.4533 1.1676 l.6209 312.0 als 8 84 688 0 01761 5 1673 5 1849 286.3 897.9 1184.1 0.4586 1.1576 l.616 316.0 3NG 89643 00li66 48%I 4 9138 290 4 894 8 1185 2 0.4640 1.1477 1.6116 320.0
!
324 0 94 826 001770 4 6418 4 6595 294 6 591.6 1186.2 0 4692 1.1378 1.6071 324.0 328 8 100 245 0 01774 4 4030 4 4208 2987 888 5 1181.2 04745 1.1200 1.6025 328.8 j
312 0 105 907 0 01179 4 1788 4 1966 302.9 885 3 1188.2 0.4798 1.1183 1.5981 332.0 til 820 0 01783 3 96Al 3 9859 307.1 882.1 11891 0.4850 1.1006 I.5936 336.0 334 I
.
340 0 117.992 001787 3 7699 3 7878 311 3 878 8 1890.1 0.4902 J.0990 1.5892 344.8 344I 124 430 0 01192 3 5834 3 6013 315 5 875 5 1191.0 0 4954 1.0894 1.5849 344.0 348 8 131.142 0 01797 3 4078 3 4258 319 7 872.2 1191.1 0 5006 1.0799 1.5806 348 e i
352 0 138 138 001801 3 2423 3 2603 323 9 868.9 1192.7 0.5058 1.0705 1.5763 352.8 354 0 145 424 001n06 30863 3.1044 328I 865 5 1l93 6 0 5110 1.0611 1.5121 3M.5
Meg 153.010 0 01811 2 9392 29513 3323 862.1 1194 4 0.5161 1.0517 1.5678 360.0 3s40 160 903 0 01816 2 8002 2 8184 336 5 858 6 1895 2 0 5212 1.0424 1.5637 364.0 Ms0 169 113 001821 2 6691 2 6813 340 8 855I 1l95 9 05263 1.0332 1.5595 368.8 372 8 117 648 0 01876 2 5451 2 5633 345 0 8516 11 % 1 0.5314 1.0240 1.5554 312.0 3t$ 8 186 517 0 01831 24279 24462 349 3 848I 1197.4 0.5365 1.0148 1.5513 376.e 3eeI 195 729 0 01836 2 3110 23353 353 6 844 5 1198 0 0.5416 1.0057 1.5473 380 0 364 e 205 294 0 01842 22120 2 2304 357.9 840 8 1898 7 05466 0.9966 1.5432 384.0 388 O 215 220 0 01847 2 1826 2 1311 362.2 837.2 1199 3 05516 0.9816 1.5392 388.8 302 0 225 516 001853 2 0184 2 0369 366 5 833 4 1199 9 0 5561 0.9786 1.5352 392.0 304 0 236 193 0 01858 I929!
l94??
370 3 829 7 1200 4 0.5617 0.96 %
1s5313 396.8 400 0 247259 001864 I8444 I8630 375.1 825.9 1201.0 0.5667 0.9607 1.5274 400.0 484 0 258125 001870 I1640 11827 3194 822 0 1201.5 0.5717 0.9518 1.5234 404.0 400 I 270 600 0 01815 16871 17064 383 8 818 2 1201.9 0.5766 0.9429 1.5195 400.0 412 0 282 894 0 01881 16152 16340 388.1 814.2 1202.4 0.5816 0.9341 1.5157 412.0
elle 295 617 0 01887 1 5461 15651 392 5 810.2 1202.8 0.5866 0.9253 1.5118 416 8
.
429 0 308 180 0 01894 14808 14997 396 9 806 2 1203.1 0.5915 0.9165 1.5000 429 0 4248 322 391 0 01900 14184 14374 401.3 802.2 1203.5 0 5964 0.9077 1.5042 424.0 428 I 336 463 001906 13591 13782 4057 798 0 1203 7 06014 0 8990 1.5004 428.8
i 432 0 351 00 0 01913 130266 1 32179 410 1 793 9 1204 0 0 6063 0.8903 1.4966 432.0 436 8 366 03 001919 124RH1 126806 414 6 189 7 1204.2 0.6112 0.0816 1.4928 436 O 448 0 38154 0 01926 1.19761 121687 419 0 785 4 1204 4 0.6161 0.8729 1.4890 440 0 4440 397 56 0 01933 1.14874 1.16806 423 5 781.1 1204 6 0 6210 0 8643 14853 444.0 444 I 414 09 0 01940 1.10712 112152 428 0 176 7 1204.7 0 6259 0 8557 1.4815 448 8 a
(*7 0 431 14 0 01947 105764 1 01111
'12 5 712.3 1204 8 0 6308 0.8411 1.4718 (. it
!
!
$8 44873 0 01954 1 01518 1 03472 J 7.0 767.8 1204 8 0 6356 0 8385 I.4741 W40
.
.--
.
.
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.
. _
. _
_.
_ - _ _ _ _ - _ _ - _ _ _ _ _ _ _ _. _ _ _.
./
e i
.
Jo I
Ahs l'ress Specifer Volonie
[nthalpy Entropy lemp lb per Sal Sal Sal.
Sal.
Sal.
Sal.
Temp i alu So lu i eqiut!
Ivan Vapor liquid
[vap Vapor liquid (vap Vapor Fahr i
p vg v ig_
vg hl h ig h
s, sla 5s I
g 468 8 466 87 0 01961 0 97463 0 99424 441.5 763 2 12048 0 6405 0 8299 1.4704 400.0 464 8 485 56 0 01969 0 93588 0 95557 4461 758 6 12043 0 6454 0 0213 1.4667 404.0 4688 504 83 0 01976 0 A9885 0 91862 450 7 754 0 1204.6 0 6502 03127 1.4629 480.0 c
472 8 524 67 0019A4 0 86145 0 88329 455 2 749.3 1204 5
&6551 08042 1.4592 472.0 l
416 e 545ll 001997 087954 On4950 459 9 744 5 1204.3 0 6599 07956 1.4555 4Nt des 8 566 15 002000 0 19716 / 0 81717 464 5 7398 1204.1 0 6644 07871 1.4510 400 0 414 I 587 81 0 02009 076611 0 18622 469I 7343 12038 06696 03705 1.4481 404.0 410 8 610 10 0 07011 0 13641 0 15658 473 8 129 7 1203 5 06745 03700 1.4444 400 0
'
412 8 633 03 0 07026 0 10191 012820 478 5 724 6 1203.1 06793 07614 1.4407 402.0 498 8 6% El 0 07034 0 un65 0 10100 4832 719 5 1202.7 0 6842 03528 1.4370 408.0 580 8 680 86 0 02043 0 65448 0 67492 487.9 714.3 12022 0 6490 0 7443 1.4333 500.0 5e4 8 105 18 0 07053 042938 0 64991 4923 709 0 12013 0 6939 03357 1.4296 504.0 584 8 73I40 0 02062 0 60530 0 62592 497.5 703 7 1201.1 0 6907 03271 1.4258 500.0 512 8 15772 0 02012 0 5R218 060289 502 3 6982 1200 5 03036 07185 1.4221 512.0 5III 144 16 0 07081 0 % 997 0 58019 5071 6923 11918 0 7005 U1099 1.4143 518.0
$20 5 812 53 0 02091 0 53864 0 55956 512 0 6870 1199 0 01133 03013 1.4146 520.0 524 8 841 04 0 02102 0 51814 0 53916 516 9 681 3 1198 2 0 7182 0 6926 1.4100 524.0 578 8 870 31 002112 0 49843 0 51955 5218 675 5 1897.3 0 7231 0 6839 1.4070 578 0 532 8 900 34 0 02123 041947 0 50070 526 8 669 6 11 % 4 0 7280 0 6752 1.4032 532.0 5360 931.11 0 07134 0 46173 0 48757 531 1 663 6 1195 4 0 7329 0 6665 1.3993 536.0 540 I 96219 0 02346 644367 0 46513 536 8 6575 1194 3 0 7378 0 6577 I.3954 540 0 544 8 995 22 0 02157 0 47677 044834 541 8 6513 1193.1 0 7427 06489 1.3915 544.0 548 I 1028 49 0 02169 0 41048 043217 546 9 645 0 1191.9 07416 0.6400 3.3816 See e 552 I 1062 59 0 02182 0 39419 0 41660 552 0 638 5 1190 6 0 1525 0 6311 1.3837 552.0 554 I 1097 55 0 07194 0 17966 0 40160 5572 632.0 11892 0 7575 06222 1.3797 358 9 568 8 1133 38 0 02207 0 36507 038714 5624 625 3 11873 0 7625 0 6132 1.3757 500 0 544 I 1170 10 0 07221 0 35099 0 31320 567.6 618 5 11861 01674 0 6041 1.3716 564 0 588 1 120772 0 02235 0 33141 0 35975 572 9 611 5 1184 5 0 7725 0.5950 1.3675 560 0 572 e 1246 26 0 07249 0 37429 0 34678 518 3 604 5 11823 0 7775 0 5859 1.3634 572.0 576 0 1285 74 0 0??64 0 11162 0 33426 583 7 5911 1180.9 03825 0.5766 1.3592 570.0 584 8 132617 0 02279 0 29937 0 32216 589 1 589.9 1179 0 0 7076 0.5673 1.3550 500.0 584 9 13677 0 02295 0 28153 0 31048 594 6 5824 1176 9 07927 05500 1.3507 504.0 580 8 1t100 0 02311 027608 0 79919 6001 574 7 1174 8 0 7978 0.5485 1.3464 500.0
e 592 8 1453 3 0 02328 0 26199 0 28827 6053 566 8 1172.6 0 8030 0.5390 1.3420 502.0 598 8 14978 0 07345 0 754?$
0 21770 611.4 558 8 1110.2 0 8082 0.5293 1.3375 300.0
"
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_
-
-_
_. _ _ _ _ _ _ _
-_
__.__ __.
.
,
. '.
Y m
.
Ahs Press Specific Volume Enthalpy Entropy
.
femp lb pe Sat.
Sal.
Sal.
Sal.
Sal.
Sat.
Temp Ialu Sq in liquid Ivap Vapor Li vid Evap Vapor liquid Evap Vapor Falw L, _v ig_,_W8 I
h is h
s, s,,
s, g
I P
V g
SeeI 1543 2 0 02364 0 24384 0 28747 617.1 5506 1167.7 08134 0.5196 1.3330 000A 804 0 15897 0 02382 0 23374 0 25757 622.9 542.2 1165.1 08187 0.5097 132M 804.0 884 e 16373 0 02402 0 22394 0 24196 628 8 533.6 1162.4 0 8240 0.4997 1.3238 GABA Si! 8 16861 0 02422 021442 0 23865 6348 5241 1159.5 0 8294 0.4896 13190 812A sie s 1735 9 0 02444 020516 0 22960 6408 515.6 1156.4 0 8348 0.4794 1.3141 888 4 81s t,
1786 9 0 02466 0 1961 0 22021 646.9 5063 1153.2 0.8403 0.4409 13092 83tt 874 9 1839 0 0 02489 0 18737 0 21226 653.1 4*6.6 1149.8 08454 04583 13041 824A
'
828 0 1892 4 0 02514 017880 0 20394 659 5 4863 1846.I 0 8514 0.4414 1.2988 820.0 832 8 1947 0 0 02539 0 17044 019583 665 9 416 4 1142.2 08571 0.4364 1.2934 832.0 838 8 2007 8 0 07566 0 16720 0 18792 672.4 4653 1138.1 0 4628 04251 1.2879 838A
,
848 8 20599 0 02595 0 15427 0 18021 679.1 454.6 1133.7 0.8686 0.4134 1.2821 SetA 844 0 2118 3 0 02625 0 14644 011269 6859 443.1 1129.0 0.8746 0 4015 1.2761 SeeA 848 8 2178 1 0 02657 0 13816 0 16534 692.9 431.1 1124.0 08806 03093 1.2699 sett 852 0 2239 2 0 02691 0 13124 0 15816 700 0 4183 11187 0 8868 0 3767 1.2634 052A 858 9 2301 7 0 02728 012387 015115 7074 405 7 1113.1 0 8931 03637 1.2567 etSA 880 8 23657 0 02768 011663 0 14431 714.9 392.1 1107.0 0 8995 0 3502 1.2498 0010 884 8 2431.1 0 02811 0 10947 0 13757 722.9 3773 1100 6 0.9064 0J361 1.2425 GMA 880 0 24981 0 02858 0 10229 0 13087 731.5 362.1 1093 5 0 9137 03210 1.2347 006.8 872 8 2566 6 0 02911 0 09514 0 12424 740 2 3453 1085 9 0 9212 0 3054 1.2266 872A sis t 2636 8 0 02970 0 08199 0 11769 749 2 328 5 1077.6 0 9287 0.2892 1.2171 SPEA 880 8 27086 0 03037 0 06080 0till7 758.5 3101 1068 5 0 9365 0.2720 1.2006 este 884 8 27821 0 03114 0 01349 0 10463 7682 290.2 1058.4 0 9447 0.2537', 1.19M 884.0 888 8 28574 0 03204 0 06595 0 09199 778 8 268 2 1047.0 0 9535 0 2337 l.1872 Sett 802 0 2934 5 0 03313 0 05197 0 09110 790 5 243.1 1033 6 0. % 34 0.2110 1.1744 802A 898 8 3013 4 0 03455 0 04916 0 08371 804 4 212.8 1017.2 0.9749 0.lMI 1.1591 0010 700 8 3094 3 0 03662 0 03857 0 07519 822.4 1723 995.2 0.9901 0.1490 1.1390 7019 182 8 3135 5 0 03824 0 03173 0 06997 835 0 1443 9793 1.0006 0 1246 1.1252 782A 184 8 3177.2 0 04108 0 02192 0 06300 854 2 102.0 956.2 1.0169 0 0076 1.1046 704.0 198 8 3198 3 0 04427 0 01304 0 05730 873 0 61.4 934.4 1.0329 0.0527 I0856 798 3 195 47'
32082 0 05078 0 00000 0 05078 906 0
906 0 1.0612 00000 1.0612 788.4P
.
' Critical temperature
]
g
'
- -
___
_ - _ _ _ _ _ _ _ _ _
..
_- __
_ _ _
l
e O
Table 2:
Saturated Steam: Pressure Table Specifii volume Enthalpy Entropy Abs Press.
Temp Sal Sat.
Sat.
Sat.
Sat.
Sat.
. Abs Press.
th/Sg in.
Fahr liquid Evan Vapor lic uid Evap Vapor liquid Evap Vapor LblSq In.
h I
g p
It
f s gg
it
'S
'I h
p I
vg v
~
880865 32 018 0 016022 3302 4 3302 4 0 0003 1075 5 1075 5 0 0000 2.1872 2.1872 000005 0 25 59 323 0 016032 1235 5 1235 5 21 382 10601 1087.4 0 0542 2.0425 2.0967 415 8 58 79 586 0 016011 641.5 641.5 47.623 1048 6 1096 3 0 0925 1.9446 2 0370 0.50
101 14 0016136 33159 33360 6913 1036I 1105 8 0 1326 1.5455 1.9181
5I 16224 0016401 13 515 13 532 130 20 1000 9 1131.1 0 2349 1.6094 1.8443
IS I 193 21 0 016592 38404 38 420 161.26 982 I II43 3 0.2836 1.5043 I.7879 M8 14 598 21200 0 016119 26 182 426 799 180 17 970 3 1150 5 0 3121 1.4447 1.1568 14.000 15 e,
213 OJ 0016176 26 214 26 290 18121
%97 1150 9 0 3137 1.4415 1.7552 15.0 28 g 227 96 0 016834 20 070 20 087 1 % 27
%01 1856 3 0 3358 1.3%2 1.7320 26.8 30 s 250 34 0 011009 13 1266 13 7436 218 9 945 2 1164.1 0.3682 1.3313 I.6995 N.8
,
4e e 267 25 0011151 10 4194 10 4965 236I 933 6 1869 8 0 3921 1.2844 1.6765 48.0 58 3 281 02 0017214 8 4967 8 5140 250 2 923 9 1174.I 0.4112 1.2474 1.6586 50.0 le 5 292 11 0 011383 7.1562 71136 2622 915 4 1177.6 0.4273 1.2167 1.6440 tee is 3 302 93 0011482 6 1875 6 2050 272.7 907.8 1180 6 0 4411 1.1905 1.6316 MI se 3 312 04 0 011573 5 4536 5 4111 282.1 900 9 1183.1 0 4534 1.I675 I.6208 88 8 se e 32028 0011659 4 8779 4 8953 290.7 894 6 1185 3 0 4643 1.1410 1.6113 98.8 lese 327 82 0 011740 4 4133 4 4310 298 5 888 6 1187.2 0 4743 1.1284 I.6027 les 8 Ils e 334 79 0 01182 4 0306 4 0484 305 8 883.1 1188 9 0 4834 I.III5 I5950 Ile 8 120 8 34121 0 01189 3 1091 3 7215 312 6 817 8 1190 4 0.4919 1.0960 1.5879 120 0 ~
130 0 347.33 0 01796 3 4364 3 4544 319 0 812 8 1191.7 0 4998 I.0815 1.5813 IM4 les e 353 04 00180)
3 2010 3 2190 325 0 8680 1l93 0 0 5071 1.0681 1.5752 140.0 150 3 35843 001809 29958 30139 330 6 863 4 1l94.1 05141 1.0554 1.5695 154.8 188 0 363 55 0 01815 2 8155 2 8336 3361 859 0 1195 1 0 5206 1.0435 1.5641 less IIII 368 42 0 01821 2 6556 2i:138 341.2 854 8 11 % 0 0 5269 10322 1.5591 litI 188 8 37308 0 018??
2 5129 15312 346 2 8507 11 % 9 0 5328 1.0215 / I.5543 188.8 198 8 31753 0 0lR 13 23841 2 4030 350 9 846 7 1197.6 0 5384 I.0113 1.5498 198 0 20s e 38180 0 01839 22689 2 2813 355 5 542 8 1198 3 0 5438 1.0016 1.5454 2ts 8 211 0 385 91 0 01844 216313 2 18217 359 9 839 1 1199 0 0 5490 0.9923 1.5413 Ile s 228 8 389 88 0 01850 2 06119 2 08629 364 2 835 4 1199 6 0 5540 0 9834 1.5374 229 8 238 8 393 10 0 01855 I91991 199946 368 3 8318 12001 0 5588 09748 1.5336 238 8 240 t 39739 001860 189909 I91769 372.3 828 4 1200 6 0 5634 0.9665 1.5299 248 8 258 8 400 97 0 01865 182452 184317 376 I 825 0 1201.1 0 5679 0 9585 1.5264 ISO 9 288 8 404 44 0 01810 115548 1114I8 319 9 821 6 1201.5 0 5722 0 9508 1.5230 268 8 2III 40180 0 01815 169131 111013 383 6 818 3 1201.9 0 5764 0 9433 1.5197 218 8
>
280 0 411 01 001880 163169 I65049 387.1 8151 1202 3 0.5805 0.9361 1.5166 200 8 a
298 I 414 25 00lflR$-
I51591 859482 390 6 812 0 1202.6 0 5844 0.9291 I.5135 298.8 300 1 41735 0 01889 1.52384 1 54274 39' O 308 9 1202.9 0 5882 0 9223 1.5105 300 0
'
358 8 43173 001912 130642 1.32554 409 8 194 2 1204 0 0 6059 0 8909 1.4964 358 0 ese 3 444 60 0 01914 114162 116095 424 2 780 4 1204 6 0 6217 0.8630 I.4847 448 8
- -
.
Specific Volume
[Ilhalpy
[;Iropy Abs Picst Temp Sal Sal.
Sat.
Sal.
Sal.
Sal.
Abs Press., e P
lb/Sq in Fah liquial Ivap Vapos lic uid Evap Vapor liquid Evap Vapor LblSq in.
.
h, h
s, s,,
s, p
v is g
g v,
g M
p i
vi s
als e 456 28 001954 101224 103119 431 3 167 5 1204 8 06360 0.8378 I.4138 450 0 508 0 46101 0 01915 0 90181 0 91162 449 5 7551 12043 0 6490 0 8143 14639 5ee g 558 0 416 94 001994 0 8/183 0 84111 460 9 143 3 1204 3 0 6611 0 7936 1.4541 SW O sas 8 486 20 0 02013 0 14967 0 16915 411 1 732 0 1203 7 0 6723 0 7738 14461 les O Sit t 494 89 0 0/032 0 68811 0 10843 481.9 720 9 1202.8 0 6828 0 7552 14381 EM S 788 8 503 08 0 07050 0 63s05 0 6 % 56 4916 180 2 12018 0 6928 03377 1.4304 Pos.g 158 I S10 84 0 07069 058880 0 60949 500 9 699 8 1200 7 0 7022 0 7210 1.4232 150 8 seeI 51821 00/061 0 54809 0 568 %
509 8 689 6 1899 4 0 7111 0 7051 14163 see t 850 8 525 24 0 02105 0 51191 0 53302 518 4 619 5 1198 0 03197 0 6899 1.4096 e548 900 I
$3195 0 02123 0 41968 0 50091 526 7 669 7 11 % 4 0 7219 06753 14032 See t 958 I
$38 39 0 07141 0 45064 0 41205 534 7 660 0 11941 03358 06612 1.3910 354 e 1988 8 544 58 0 07159 0 42416 0/45 %
5426 650 4 1897.9 0 7434 0 6416 1.3910 lose 4 18588 550 53 0 02111 0 40041 0 4???4 5501 640 9 1191 0 03507 0 6344 13851 1850 8 lies 8 556 28 0 02195 0 31863 040058 5515 6315 1189.1 0 7578 0.6716 13194 Ilse 0 1850 8 561 82 0 02214 0 35859 0 38013 564 8 6?? 2 1181 0 0 7647 06091 13738 115e 0
'
1200 8 56119 0 07237 0 14013 0 36245 511 9 613 0 1884 8 0 7714 05%9 13683 1200 t 125II 572 38 0 02250 0 12306 0 34556 578 8 603 8 1182 6 0 7780 0 5850 1.3630 1250 8 1388 5 57142 0 02269 0 30122 0 32991 585 6 594 6 1180 2 0 7843 0 5733 1.3571 1300 0 1358 8 582 32 002288 0 29250 0 31531 592.3 585 4 1177 8 07906 0 5620 13525 1350t 1400 1 58101 0 02301 0 27811 0 30118 598 8 516 5 1115 3 03966 0 5507 1.3414 1400 5 145I B 59110 0 02321 0 26584 0 28911 605 3 5674 1172 8 0 8026 0 5397 I3423 1458 8 1588 8 596 20 0 02346 0 25312 0 21/19 611 1 558 4 1170 1 0 8085
' O 5288 13373 1500 8 1558 8 600 59 0 02366 0 24235 0 26601 618 0 549 4 11674 0 8142 0 5182 1 3324 1550 0 llH I 604 81 0 02381 0 73159 0 25545 624 2 5403 1864 5 0 8199 0 5016 13214 1600 8 1658 8 609 05 0 02401 022143 0 24551 630 4 531.3 1161 6 0 8254 0 4911 1 3225 165e 8 1100 0 613 13 0 07478 021118 023601 636 5 52? ?
1158 6 0 8309 04867 1 3116 17e0 8 1758 I 611 12 0 02450 0 20283 0 22113 642 5 513 1 1155 6 0 8363 0 4765 1.3128 1154 8 1888I 821 02 0 02412 0 19390 0 21861 648 5 503 8 115? 3 0 8417 0 4662 1.3019 1888 e 18588 624 81 002495 0 18558 02I052 654 5 494 6 1849 0 0 8470 0 4561 13030 1854 e 1988I 628 56 0 07511 0 11161 0 20718 660 4 4852 1145 6 0 85??
O4459 I2981 190s 8 19588 632 22 0 02541 0 16999 0 19540 666 3 415 8 ll 42 0 0 8574 0 4358 i 12931 1958 e 2ees 8 635 80 0 07565 0 16266 0 18831 612.1 466 2 1138 3 0 8625 0 4256 12881 2000 8 2388 8 642 16 0 01615 0I4885 0 11501 6838 446 7 IJ305 0 8127 0 4053 12180 2104 e flag I 64945 0 07669 0 13603 0 16212 695 5 426 7 1822 2 08878 03848 12616 270s e 2308 8 655 89 0U/1/1 0 12406 0 15133 101 2 406 0 til3 2 0 89?9 0 3640 12569 2300 0 2494 8 667 11 0 0/190 0l17R1 0 14016 119 0 384 8 11037 0 9031 0.3430 12460 2448 8 25H 8 66811 0 02859 0 10709 0 13068 731 3 361 6 1093 3 0 9139 03206 12345 25000 2604 5 613 91 0 02918 0 09112 012110 144 5 3316 1082 0 0 9241 0 2917 11225 2600 0 2IM I 619 53 0 03029 0 08165 011194 1513 3I23 1069 7 0 9356 0 7141 12091 2100 0 2000 8 684 %
0 03134 0 01111 0 10305 110 7 2851 1055 8 0 9468 02491 1.1958 20008 2900 8 690 22 0 03762 0 06158 0 09420 185 1 2547 1039 8 0 9588 01215 1.1803 2900 e 3400 0 695 33 0 03428 0 05013 0 08500 8018 218 4 1020 3 0 9728 01891 1.1619 3000 0 3188 8 700 28 0 036MI 003111 0 01452 824 0 169 3 993 3 0 9914 0.1460 1.1373 3100.8 32H 8 705 08 0 04411 0 01191 0 05663 815 5 56 1 931.6 1 0351 0 0482 1.0832 320e 8 3104 2'
70541 0 0'.n18 O scul 005018 906 0
906 0 10612 0 0000 1.0612 329e.2*
t)
f
.
.-
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..
.
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.-
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.
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i RRCS CHANNEL A CHANNEL A
LOGIC A INDICATIONS LOGIC' B
,
RRCS MANUAL CHANNEL A MANUAL
-
INITIATION OUT OF INITIATION PERMISSIVE SERVICE PERMISSIVE
'
E
RRCS MANUAL MANUAL MANUAL INITIATION INITIATION INITIATION i
ARIREADY ARIREADY
-
FOR RESET ARI FOR RESET I
-
INITIATED
-
ARIRESET ARIRESET RRCS READY RRCS READY
.FOR RESET ARI FOR RESET
,
VALVE OPEN RRCS RESET i
RRCS RESET
'
,
i RRCS
.
FEEDWATER RRCS-c
'-
LOGC A RUN8ACK LOGIC B TROUBLE INITIATED TROUBLE
!
TEST FAULT ESSENTIAL i
l LOGC FAILURE l
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[
,
~
10C651 CONTROLS
.
- FIGURE 1 **
_,
I.
.
GEK-83288
.
N T
/
/
-
-
C k
BACK UP FUEL K)NO AUX TROLLEY ROD BIACK
%
HOIST HOIST HOIST AUX HOIS1 1hTERLOO LIMIT INTERLOCK IhTERLOOf INTERLOCX NO. 1
/
BRIBE
/
N ROD BLOCK BRIDGE FAULT g
lhTERLOCK REVERSE REVERSE 14CK0tTT (
-
NO. 2 S11)P NO.1 STOP NO,2
O l
INTERLOCK STATUS DISPLAY l
C
-
-
Figste 3-6.
Interlock Status Display Module l
- FIGURE 2 **
.
(
l 3-12 (
l
_ __.
.
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- -,
._
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.-
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-
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.
.
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TABLE 2
{
RPV PRESSURE INSTRUMENTATION
.
REACTOR PRESSURE Transmitter Indicator LOC.
SYS.
STPT.
Funbtion PSIG PT-NOO5 PI-R605 650C FDW PT-NOOB PR-R609 650C FDW PT-NO78A PR-R623A 650C 1050 RPS TRIP NS4 ISOLATION
-
PT-N0788 PR-R6238 650C 1050 RPS TRIP
--
NS4 ISOLATION PT-N078C 1050 RPS TRIP
NS4 ISOLATION PT-N0780 1050 RPS TRIP-(
,
NS4 ISOLATION PT-N090A CS/RHR INITIATE -
'
-
PT-bl0908 SS/FHR INITIATE PT-N090E
'
C5/RHR INITIATE
~
-
c.
PT-N090F CS/RHR INITIATE PT-N090J CS INITIATE CS INITIATE PT-N090K
-
PT-N090N CS INITIATE PT-N090P CS INITIATE PT-N403A RRCS ANALOG TRIP MODUE PT-N4038 RRCS ANALOG TRIP MODUE PT-N403E RRCS ANALOG TRIP MODUT RRCS ANALOG TRIP MODUE PT-N403F
-
,
PT-3684A PI-3684A 650C PAM PT-36848 PR-36848 650C
,
!
PT-7853A RHR INTERLOCKS PT-7853 PR-78530 RSP RHR INTERLOCKS
.
PDT-NO32 PDR-R61'3 650C (_
l
- FIGURE 3 **
q
__
-
.
.. - -.
_ _ _ _
-.
-
.
.-
--
.
. _.
__
-
,
.
-
--
.- --
--
.
.
-
.
.
'
k OEFINITIONS SECTION PAGE 1.0 DEFINITIONS 1-1 1.1 ACT!0N.....................................................
,
1-1
'.2 AVERAGE PLANAR EXP05URE....................................
1.3 AVERAGE PLANAR LINEAR HEAT G G ERATIONRATE.................
1-1 1-1 1.4 CHANNEL CALIBRATION.....z.................................
1-1 1.5 CHANNEL CHECK...............:..............................
1-1 1.6 CHANNEL FUNCTIONAL TE5T....................................
1-2
1.1 CORE ALTERATION............................................
.
,
.
1.8 CORE MAXINUM FRACTION OF LIMITING POWER 5ENSITY............
1-2 1-2 1.9 CRITICAL POWER RATI0.......................................
(;.
1.10 05E E Ul m Eni 1-m...................................... i-2 i
'
'
1-2 1 1. n I-Avt uGE Dis m EGRATION aERGy............................
1.12 EMERGENCY CORE Co0 LING SYSTEM (ECCS) RESPONSE' TIME.........
1-2 l
,
1.13 END-OF-CYCLE RECIRCULATION PLMP TRIP $YSTEM RESPONSE TIME..
1-3 l 1-3 l 1.14 FRACTION OF LIMITING POWER DEN 51TY.........................
1-3 l 1.15 FRACTION OF RATED THERMAL P0WER............................
1-l 1 15 FREQUENCY N0TATION........................................
o.
vau wome a aan x o.
o.:o e. nee..s oe en snrm s
1-l 1.
IDENTIFIED LEAKAGE.........................................
'
1-3 l 1.gI50LATION SYSTEM RESPONSE TIME................. ;...........
1-3 l 1.W.,uMITING CONTROL R00 PATTERN...............................
1-3 l 1.(LINEARMEATGENERATIONRATE................................
i-4
-
14toGICsTSTEMruuCTIONALTtst...............................
,
-
- -.__. - -. _..
-.-swn...............................
. _ n,
--
.,
rw. nan-er u.
,wm e.
gs w i ne.
I-4
[ (
g.11 M6n06AD *! TM fs**uc.
- 1-4 H 1.gMININuMCRITICatPOWERRATI0...............................
l-V s,st, ONasrg Deed CAcuursew *taaw *'
I
- ev e i - - s n,.
a
wm-g.-
-,
s
'
Aev,.2,
%=*t. c.*stw
.
..,. -.. - _, -.. - _..,, - -. - - -.,
,_.n___-,__.
,, -,., - -. _ -.. - - -.. - _,.,,,..,,, _. - - - -, _ _ _..,. - -. - -.. - - - - -
--
-
_
- -
.
. _
__
.
INDEX
,
I
\\
DEFINITIONS
.
SECTION PAGE DEFINITIONS (Continued)
s1 1-4 ll 1.y OPERA 8LE - OPERA 81LITY.....................................
1,.25 OPERATIONAL CONDITION - CONDITION..........................
1-4
as
f.nt 0 Nun *~AL Mabe -M*DG-y 1-4 l
14PHYSICSTEST5..............................................
a 1.# PRES 5uRE e0uN0ARygae $.................................
1-4
- * *
!g,s,1 wu a~n*6 s
-
ll
.ae - m s-IJg/RIMARY CONTAlmENT INTEGRITY..............................
-
1-5 ll sr 1.js'RATEDTHERMALP0WER........................................
1-5 l
sc.
1.)t REACTOR PROTECTION SYSTEM RESPONSE TIME....................
,
1-5 U
rr suur 1.)tr REPORTASLE-4GGtHHtfMCE......................................
se 1-5 l
1.)rR00DEM51TY................................................
1.x=== me,:m(seco~o4Y ce a' erat)
r se mcTen. woms IwrEGRITY............................
1-s il (
.
-
-
)
d so 1-6
1 IJS SHUTDOW MARGIN............................................
1-6 I
.,
- - -,
- 1.
STAGGERED TEST SA5!S.......................................
1-6 l
yr
.
1.)5THERMALP0WER..............................................
._,, t
..
.....
.. _.. -..
.,
s..,,,,,...-
-.- -........................................
.
1.kTURSIMEBYPASSSYSTEMRESPONSETIME........................1-7 l
.
,,
1-7 I
1..F1 UNIDENTIFIED LEAKAGE.'......................................
,
'
MW 1-8 I
'
TABLE 1.1. SURVEILLANCE FREQUENCY NOTATION......................
i 1-9 l
TABLE 1. 2. OPERATIONAL C02ITIONS...............................
l
\\
t.1I
$ITE 6** bha 1
,,,, $
~~
l. 4 L COUbe fit ets e N
- ,, g L l.43 JamAct eserce s.(,
-
" f.4 9 U M rsissert o M e n
,'Q
/. 41 m r u nea as na.or rasaresur eysrsn t-7
.
l. f *
VbtTIM
..
Rev.2-
ae
- -..
.i n Wu f
M Mw g---
7g ENA
-
.
,;
!
-
.
.
- - _.
_
_ _________ __-_________ _________________ __- - _
_
_
.
.
.
.
.,
..
.
(
15lu
.
.
.
'
SAFETY LIMITS AND LIMITI M 5AFETY SYSTEM SETTINGS
.
SECTION g
-
.
..
2il 5AFETY LIMIT 5;
,
MRMAE. POWER,LawPressureorLcwF1sw.........'..........
2-1
MIMAL POWC1, Ni gh Pres sure and Hi gh F1 sw................
2-1 Reactor Caot ut System Pressure...........................
2-1 Reactor Vessel histar Leve1................................
2-2
.
2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instroentation Setpoints.......
2-3
.
.
.
.
SASES (
J,
,
.
,
2.1 SAFETY LIMITT,
.
THERMAL POWER, Low Pressure or Low Flow.............l.....
I 2-1 THERMAL POWER, High Pressure and High Flow................
B 2-2
-
.
Reactor Coolant System Pressure...'........................
B 2-5-I Reactor Vessel niater Leve1......................'..........
I 2-5 I
.
.
q 2. 2 LIMITING 5AFETY SYSTEM *5ETTINGS
'
'Reacter Protection System Instrumentation Setpoints........ I 2-6 l
'
.
.
.
.
.
.
.
p
.
-
.
p
_., _ _,,,
,,,
.
-.....
,
het. c* ass
._
a
-. -. _ - - - - -_ -____ _ ___ _ - _- _ - _ - - - -. -
-_
_ - -. _ - ---.
-. --- - -
_
.
.
10 JAN 1983
\\
19 l.1 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE
.
PAGE SECTION
_
3/4 0-1 1/4.0 APPLICA81LITY.............................................
3/4.1 REACTIVITY CONTROL SYSTEMS
.
3/4 1-1 3/4.1.1 SMUTDOWN MARGIN........................................
REACTIVITY AN0MALIES.............................'.
3/4 1-2 3/4.1.2
.
.
3/4.1.3 CONTROL R005
.
3/4 1-3 Control Rod Operability................................
l 3/4 1-6 Control Rod Maximum Sc' ram Insertion Times..............
3/4 le7
.
Control Rod Average Scram Insertion Times..............
3/4 1-8 Four control Rod Group Scram Insertion Times...........
,
3/4 1-9 (
Control Rod Scram Accumulators.........................
i J
Contr'o1 Rod Drive Coup 1tng.............................
3/4 1-U 3/4 1-13 Control Rod Position Indication........................
3/4 1-15 Control Rod Drive Housi ng Support......................
i 3/4.1.4 CONTRCL 200 PROGRAM CONTROLS 3/4 1-16 Rod Worth Minimizer....................................
3/4 1-17 -
Rod Sequence Control 5ystas............................
.
3/4 1-18 Dod Block Monitor......................................
3/4 1-19
.4 s.5 STAN0tY LIQUID CONTROL SY5 TEM..........................
i 3/4.2 POWER DIST113UT10M LIMITS 3/4 2-1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.............
3/4.2.1 3/4 2-f 3 1/4 L *
APRM SETP0!NTS.........................................
3/4 2-EY '-
3/4.1.3 MINIMUM CRITICAL POWERRATI0...........................
3/42ftr l LINkARHEATGENERATIONRATE............................
3/4.2.4
{
iv
- + " " ' r - )
% # T. C A tt w
-_,
_ _. _ _ _
_ _ _. _ _ _ _.
. _. _ _ _ _ _ _.. _ _ _ _ _
.
..
.
.
..
_
_
_ ___ _ _. _ _ _ _ _
_ _ _. _.
_ _.
_
_ _ _ _
__
-
.
-
10 JAN 1983
/
\\
jgEl
'
LINITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS
.
_
_
PAGE SECTION 3/4.3 INSTRUMENTATION
-
,
l 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION............
3/4 3-1
,
l 3/4J.2 ISCLATION ACTUATION INSTRLsENTATION.................. 3/4 3-5
'
3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION r ll INSTRUMENTATION......................................
3/4 3-F
.
3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION
,
w l
ATV5 Recirculation Pump Trip Systes Instrumentation.. 3/4 3-46 Il
'
End-of-Cycle Recireviation Pump Trip System 3/4 34 ll w
Instrumentation......................................
.
t 3/4.3.5 REACTOR CORE ISCLATION C00 LING SYSTEM ACTUATION3/4 346- \\l sy 1NSTRUMENTATION......................................
-
(
con 1=t =0 it0C n NSTRuME m uCN.................... u m ii
,
u4. 3..
.
Y i
3/4.3.7 MONITORING INSTRUMENTATION car ll
-
i Radiation Monitoring Instrumentation.................
3/4 3-57 7s Seismic Monitoring Instrumentation...................
3/4 3-44-ll
se 3/4 3 e ll l
> Meteorological Monitoring Instrumentation............
'
- Remote Shutdown Monitoring Instrumentation........... 3/4 3-7
'
3/4 3 d ll rAccident Monitoring Instrumantetton..................
~
3/4 34 ll
' Source Range Monitors................................
3/4 3d ll
" Traversing In-Core Probe System......................
.
l
- /0M ll S h r' = (cd.' - : - h) ";'.;d = !je r..............
C;,*.if.
",,;... %;; " ;.0^............................
' 't ' ", ll
,
l
,
3/4 3 ll
,
4 Fire Detection Instrumentation.......................
sa '~
3/4 3-06 ll H Ase-Part Detection Svstas........ G h..* h........ A.
e w4 3-sy m..W %4 ME e W.
F /4.3.8 TURRINE OVERSPEED PROTECTION SY5 TEM...
-
-
Framw& Tea /MAmt 'NRanet 'TRW mettM
3/4.3.3 ";." ;Y:T;.": ACTUATION INSTRUMENTATION...............
3/43-99%
s hceMd** */q s 11 yQew hs tNA Wahg I
.
.
r-6 ~ \\s v
"" C: (r.'t)
-
-
tsee eseaw.
__ _._ _. -
_ _ _ ___ _ ___ _
.
.
.
..
.
.
__
_
_ _ _ _ _
..
_ _ _. ___ _.
_
.
.
>
.
.
IEE
-
LIMITING COWITIONS FON OPERATION AND Sutvf!LUWCE REQUIREMENTS
,
'
BElE PAGE
'
3/4.4 REACTOR C00VWT SYSTEM
-
3/4.4.1 RECIRCifLATION SYSTEM
^
Recirculation Laeps..................................
3/4 4-1
Mt M s............................................ 3/4NY
]
tocirculation Pumps..................................
3/4 4-25 141e Recirculation Laep 5tartup......................
3/4 4-p G
,
I 3/4.4.2 SAFETY / RELIEF VALVE 5
Safety / Relief Va1ves.................................
3/4 4 4
'
grvr4 _, _,. __...,._, _, o _.. __ _. _..
pg
,_., m,__
.
_ _ _ _.. -.._.............
_,. -.......
,
.
.
3/4 4.3 REACTOR COOLANT SYSTEM LEAXAGE
-
Lankage Detection Systans............................ 3/448
,
(
-
Operational Laakage..................................
3/44-9-N 3/44-S
!
,. 4 CNENISTRY............................................
-
.
SnCr n C 4CTivm...................................
3/4 44
3/4.4.5
.
3/4.4.6 PRES 5URE/ TEMPERATURE LIMITS I
M i
Reactor Coolant 5ystas...............................
3/44-4-3/44-N
'
Reactor Steam 0eme...................................
3/4 4 4
'
-
v4.4.7 MarN STEAM trNE r$0t4720N vAtvE5.....................
.
STRuCTuRAt m EGR m....:............................ u4 4 4 v4.4..
' stauAL MAT REMhWAL 3/4.4.9 E
.17
,
Not Shutdeun.........................................
3/4 4-95-l
,
3/44M Cold Shutdown........................................
3/4.5 D4;RGENCY Coat COOLING SY37Dt3 3/4.5.1 ECC5 - OPERATING.....................................
3/4 5-1
.
3/4.5.2 ECCS - SMLtig0WN......................................
3/4 5-6 b
3/4.5.3 SUPPRESS ! 0N CNNGE R.................................. 3/4 5-8 i
- ,
......
_...,_
,-
l
.
-
+,wwm
_ -,,.. -,,, -. _ - -, -. - ~ - - -
_ - -
ir--.-r-r--+--r----r-w- - e -
e--zwevt'---- - -
. _ _ _ _ _ _ _ _ _ _ _
. - _.
.
.
.
(
INDEX LIMITING emnITIONS FOR OP(RATION AND SURVi!LLANCE REQUIREMENTS i
'
SECTION PAGE 3/4. 6 CONTAf stNT SYSTEMS
'
3/4.6.1 PRIMARY CONTAINMENT
.
Primary contai nment Integri ty........................ 3/4 6-1 a
Primary Containeer.t Leakag '......................... 3/4 6-t-rs Primary Cantainment Air Locks........ :............... 3/4 6-t-w%
os M51V '_rr;:g. c t :!
5ystes..........................
3/4 6 +
-
-
,
" '
j "--t '- :
Et ;:t: :1 !:t ;:it,..............
- /t :
a s
Drywell and Suppression Chamber Internal Pressure.... 3/4 6-9-l r1 Drywel l Average Ai r Temperature...................... 3/4 6-44 l
.
Drywell and Suppression Chaseer Purge Systas.........
3/4 6 d \\
se a A _=_ l
"-irrj trn'--92F=:E: f r "r:=r'nt'= !j:tr 2/' 5-12-
3/4.5.2 DEPRE55URIZATION SYSTEMS l
.to
+
Suppression Cham 6er..................................
3/4 6-+t
'
Suppression Pool (and Drywell) 5pr,ay.................
3/4 E M l
Supp res si on Peel cool i ng............................. 3/4
" j_:it ' ;;r; n' = C M r M r rt':1 "::n:...
E' 5-?S
'
.
Er 3/4.6.3 PRIMARY CONTA! MENT 15CLATION VALVE 5................. 3/4 6-49
,
l 3/4.6.4 VACUUM RELIEF
Suppression Chamber - Drywell vacuum Breakers........
3/4 6-46 Reacter tutiding - Suppression Chamber Vacuum yr l
6 re a k e rs........................................... 3 / 4 6-46
- ~0"" "7 C*". /T.T anactog smummt. (evet a"* '*d*)
3/4.6.5 Lu cr,.nn. (saweans e-a.=** Mity...................... 3/4 6 d t en d a h,.43 J
-: t, ntegr m a yw=,%[re<
=5
-+ <> ion 40ampersl
- t. Automatic Isolat v1 nxcrj)_rm !. n........................................... 3/4 6-e
.
.
Fih%,ReelmuMian.ead ve,MMM spbg Opjs) 6-49-
/4
{
O r n, __ Trn t n.,nr..........................
..
.,. _
,,,
.
-...
. -,
ON
.
,-------,-----,--~-------w--_--
-.. _,.,. _, -. _..,, - _ _ - _ _.,,
__.____,-.--.___,,.-,,,--.-.-w,----.-
_ --- -. - -
, -, - -.. - -. - - - - - - - -,. -
.
- -
.
.
.
.
-
.
J If!ILE
.
LDu ring CE r-!TIONS FOR OPttATION AND SUtytiLLANCE tt00f tDENT5 Esit SECTION
.
'
CONTA!WENT avaiu6 (Continued)
,
3/4.5.5 PRIMARY CONTA1 MENT ATM35PNERE CONTROL
"- _;? ?D Y b -- r ' r W n Recombiner as l
. htans............................................ 3/4 6-St-l
,,
. _ _ _ _. _, _ _
..
___.,_,_
..
- - _
,.. _... _ _
.. - _.,...........
,,,,
-,. -
.............................................
.,--_
.__ _,,__,. - __,
, n. ___
n,-
r.. p..g w.g-
.'
%"
-7r'***'"
"'
"'#
E "' " W
.
,.,
..............................................
ss
-,. _
trywell and Suppression Chastwr Oxygen Cancentration. 3/4 6-M l I
.
i 3/4.7 PLANT avaivo
-
SERVICE BTU SYSTDts l
3/4.7.1
.
P F W5 A = h m_.__. _. _.. __ ---. Systas. [s..A.c.4... 3/4 7 1 Mm,
-
.
. - - _. _ _ _. - _ _ -
l Plant Service Watar 5ystas...........................
3/4 7-3 vi ti-ta neat 5 i na.................................... u. 7-5 (
.
CONTROL 200M EERGENCY FILTRATION SYSTEM.............
3/4 7-6
'
d 3/4.7.2
.
3/4.7.3 FLOOD PROTECTION..................................... 3/4 7-9
.
M 3/4.7.4 REACTOR CORE ISC LATION COO LING SYSTEM................ 3/4 7-19 54USS ERS...................................
3/4.7.5 SutE050utetCONT=tmTION..........................u47-3
,
u4.7..
-
3/4.7.7 FIRE SUPPRESSION SYSTEMS so
-
Fi re Suppres sion Water 5ystas........................ 3/4 7-at
Spray and/or Sprink k e Systans...................'....
3/4 7 M I
'
v4 7.K
Ca, 5y.t
.........
................................
> > - -
m._
. _ _ _ _
............
,...
.....,,._-............................
Fire Mese Stations.................................. 3/4 7 M
.,_n
__> m m...
u_..
u......
v_ m n _. m >___.,
,.....
.g -..-- - -
y w-
-- - - - - - - - - -............
m V WW 3/4.7.8 FIRE 443EHeSettfHr............................... 3/4 7 l
SAtates. Ps ofaavisees i-
....
6, I
...-...._----.m.-.....-.-...........................
,,
- -_..
,
.,...
..
...,
_ _.
I 3/4.7.14 MAIN TURS!NE SYPA55 5YSTEM...........................
3/4 7-M
.
.
.
-...
haemIA%,
.M
.9 9
.-,
,
__-.,,, - - _ - _
_
,,,--
,----,--,-
-, - - -
- -, - - _ _. - - - - _ - - - ~ - - - _ _ _ _ - _ - -
.
-
-
. -.
- - - -
..
.
.
.
.
10 JAN 1982
.
INDEX s
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS
,
,
.
PAGE
-
.
SECTION
.
3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES A. C. Sources-0perati ng............................... 3/4 8-1 A. C. Sources-Shutdown...........................
3/4.8.2 0.C. SOURCES
,
D.C.
Sources-Operating...............................
3/4 s h j
.I
!
s
0.C. Sources-Shutdown................................
3/4 4-W 4
'i
'
g
'
3/4.8.3 ONSITE POWER DISTRIBUTION SYSTEMS
..
Di s tribution - Operati ng............................. 3/4 8-M p/Y Di stributi on - Shutdown.............................. 3/4 8-E
.
(
\\
d 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES.
.
A.C. C h;.it; !..f i N != j C: t i z at.............
3/' H S r
~
f Prisary Containment Penetration Conductor Overcurrent J u, P rotecti v e De vi ce s................................. 3/4 8 JP-cNee1.3d Protection.-httt*3/.4. SM,,-
zi 1 Overl MotorOperatedValveThMe er.+..+,_... e- -
3f nshe errded alve. n.
Reactor Protection Systes Electric Power Monitoring.. 3/4 8-eess23 3/4.9 REFUELING OPERATIONS
.
.
3/4.9.1 R E 0R IC0E' SWITCH..................................
3/4 9-1
.
!
j
- 3/4.9.2 InsinGPENTAT10N......................................
3/4 9-3 i
3/4.9.3 CkTR0L 800 POSIT 10N................................. 3/4 9-5
"
3/4.9.4 DECAY TIME........................................... 3/4 9-
.
3/4.9.5 ComuM I CAT 10N S....................................... 3/4
.
.
"
3/4.9.5 REFUELING P LATF0RM................................... 3/4 9 me
/
O
-
6 d \\
M van emw
,
, -._ - -..-..... -. -.___.-, _- _ _ _ - - - _ _ _ _
.,.. -., _ _. - - _, - _ _, _ _ _ _
...
c
,, _ _ _ - _ _. _ _ - _ _ -,
-
.___ __.-.
.-
.-
'
<
.
.
.
(
INoEx
-
LIMITING C04 !TIONS FOR OPERATION ANO SURVEILLANCE REQUIREMENTS
.
i SECTION f.s I
REFUELINGOPERATIONS(Centinued)
.
3/4.9.7 CRANE TRAVEL - SPENT FUEL * STORAGE P00L............... 3/4 9 9 3/4.9.8 MTER LEVEL - REACTOR VE55EL......................... 3/4 9-10 3/4.9.9 MTER LEVEL - SPENT FUEL STORAGE P00 L................ 3/4 9-11
'
3/4.9.10 CDNTROL 200 REMVAL Stagle Control Red Remova1...........................
3/4 9-12 i
>
-
Multiple Centrol Rod Remova1.........................
3/4 9-14
'
3/4.9.11 RESIDUAL MEAT RDCVAL AfC COOLANT CIRCULATION
.
Nigh ikter Level..................................... 3/4 9-16 l
Law lister Lave 1...................................... 3/4 9-17 3/4.10 SPECIAL TEST DCEPT! CMS
{
!
>
v4.10.1 Prim RY CONTAr m ENT INTEGR m......................... v4 10-1 v4.10.2 m0 SEQUENCE CONTm L sv57EN.......................... v4 10-2 3/4.10.3 $NUTDOWN MARGIN DOWN5TRATIONS....................... 3/4 10-3
,
3/4.10.4 RECIRCULATION L00PS..................................
3/4 10-4 3/4.10. 5 OXYGEN C0NCDITRATION................................. 3/4 10-5
,
. v4. i o.
TurNim STARTU,5..................................... u4 i G-i
-
.
.
-
!
.
.
.
.P
!
$
.
- m
.....,m
-
i
,
... s
,,
,
hNk
,
J
.
,,. _. - _. _ -, - _ _ _ _ _ _ _ _
,,.._,w
_,.,. _,,, -., _ - _
__.,..m.m--_.__
. - _,.. _ _ _ - - _ - _ _ _ _ _ _. _.. _ _
_
.-.
-
-
_ - -_- - _-_
- - _-.
.
.
.
INDEX (
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE SECTION I*
3/4.11 RADICACTIVE EFFLUENTS s
o 3/4.11.1 LIQUID EFFLUENTS l
Concentrat1nn............................................
3/4 11-1 3/4 11-X g.e Oose.....j..............................................
Liquid (WesteTreatment...................................
3/4 11-#*f 3/4 Il f f *
Liquid Holdup Tanks......................................
O.;;i;;1 Tr; C :nt N.::..........
.....................
3/4 ti 0
.
3/4 11.2 GASEQUS EFFLUENTS 3/4 11-9 l Dose Rate................................................
3/4 11-13 Dose-Noble Gases.........................................
~4 Dose x.d y':.gt p +*um: rte: !:te:
-
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3/4 11-14 3/4 11-15 Gaseous Radweste Treatment...............................
Ventilation Exhaust Treatment............................
3/4 11-16 3/4 11-17,
-}
Explosive Gas Mixture....................................
i 3/4 11 Me#
'
Main Condenser...........................................
3/4 11-Je' /T Mark I or II Containment.................................
-
.
3/4 11.3 SOLID RADIDACTIVE WASTE.................................._,
V.4.11.p as
,
TOT A L D0 5 E.............................................
3/4 11.4 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4 12 1,
3/4.12.1 MONITORING PR0 GRAM......................................
3/4 12-15 3/4.12.2 LAND USE CENSU5.........................................
.
'
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3/4 12-1Y 3/4.12.3 INTERLA80RATORY COMPARISON..............................
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3/4.1 RDCTIVITY w iisui. SYSTDe$
l 3/4.1.1 SIEff00MI 1448 GIN................................. 8 3/4 1-1
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3/4.1.2 MACTIVITY Ancit4 LIES............................. S 3/4 1-1 3/4.1.3 CDerTROL 8005....................................
8 3/4 1-2 3/4.1.4 CONTROL A00 PROGRAM CONTR0LS..................... S 3/4 1-3
3/4.1.5 - stale 8Y UQUID CONTROL 5YSTDt................... 8 3/4 1-4 l
3/4.2 POWER O!!TRIBUT1081 LDef75
.
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l 3/4.2.1 AVERAGE PLANAR UleGR NEAT GDIERATION
.
RATE..........................................
8 3/4 2-1 3/4.2.2 AMIM SETP01h T5.................................. 8 3/4 2-2
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i 3/4.3 INSTartNTATION
'
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l 3/4.3.1 IIDCTOR PROTECTION SYSTD4 INSTituMD(TATION........ B 3/4 3-1 l
3/4.3.2 ISOLATION ACTUATION INSTituMDf7ATION............. 8 3/4 3-2 3/4.3.3 EMERGDeCY CORE C00UlIG SYSTDI ACTUATION
,
DesTRWetTAT10N.................................. B 3/4 3-2
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3/4.3.4 REC 1RCULAT10N PWIP TRIP ACTUATION DesTRsWITAT10N................................. 8 3/4 3-3
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3/4.3.5 ADCTOR Coltf !$0LATICII CD00NG SYSTDI ACTUATION Ut572uMDITAT10N........................ I 3/4 3-4 i
l 3/4.3.6 CONTROL A00 SLOCK De5TapetT AT10ll................................. 8 3/4 3-4
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152
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IEE EUlE"
- INSTRUMENTATION (Continued)
'
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3/4.3.7 NMITIMING INSTRUMENTATION
Radi ati on Monitoring Instrimentatt en............ 8 3/4 3-4 Seismic Monitoring Instruentatten..............
5 3/4 3-4
Noteersl egi cal Moni tori ng lastruentat t en....... S 3/4 3-4
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Assets Shutdown Monitoring Instrumentation...... S 3/4 F5 l
-
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Acci dent Monitori ng Instrimentati on............. t 3/4 F5 Source Range Monitors...........................
8 3/4 3-5
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Treversing In-Care Prete systas.................
B 3/4 F5 l
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3/4.4 atAcTon c00LANT SYSTEM
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B 3/4 4-1
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3/4.4.1 RECIRCULATION SY5 TIM............................
g 3/4 4.i 3/4.4.2 SAFETY /REL!tF VALVt5.........................,,,
3/4.4.3 REACTOR C00LANT SYSTEM LEAKAGE
.
,
.
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I 3/4 4-2 -
Leakage Setection Systans.......................
8 3/4 4-2
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I 0perational Leekage.............................
.
8 3/4 4 2 3/4.4.4 CMEN13TRY.......................................
,
,
S 3/4 4-3 3/4.4.5 SPECIFIC ACTIVITY...............................
5 1/4 4-4
l 3/4.4.6 Ptt3suntntMPERATURE LIMIT 3.....................
l 3/4.4.7 MAIM 57 TAM LINE ISOLATION VALyt5................
8 3/4 4-5
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8 3/4 4 5 l
STRACTURAL INflGR1TY................*............
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3/4,4.s 3 3/4 4-5 (
3/4.4.3 RE51 DUAL MEAT MMOVAL...........................
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sasts PAGE
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SECTION
,
.
3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1/2 ECCS - OPERATING and SMUTD0WN...................
8 3/4 5-i
~
.
Suttatss!ON CWWG ER............................. 8 3/4 5-2 3/4.5.3 3/4.6 CONTAINNENT SYSTEMS 3/4.6.1 PRIMARY CONTAI N NT
-
Primary Containment Integri ty................... 8 3/4 6-1
, Primary Contai nment Laskage..................... 8 3/4 6-1 Primary Contai nment Ai r Lacks.................
8 3/4 6-1
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8 3/4 6-1
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Orywell and Suppression Chamber Internal ( }'
Pr..sure......................................
8 3/4 6-2
Orywell Ave rage Ai r Temperature................. 8 3/4 6-2 l l
Orywell and Suppression Chamber Purge Systas....
8 3/4 6-2 l i
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DEPR155URIZATION SYSTDt5........................ 8 3/4 3/4.6.2 PRIMMtY CONTAlleENT ISOLATION VALYts......'.$..
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3/4.6.3 VACULM RELIEF................................... 8 3/4 3/4.6.4 taAsm suu.wa, poucaat e.easead OZ : ^ ^ 7 Z; " :Z.............................. 8 3/4 6-5 t
1/4.8.5
'
PRIMNtf CONTA!!sENT ATMO5PHERE CONTROL..........
8 3/4 6-6 l
,
l 3/4.8.8
.
3/4.7 Puurf swivis SERVICE wilt 5YSTDts............................ B 1/4 3/4.7.1 CONTROL ROOM DERGENCY FILTRATION SYSTEM........
8 3/4 7-1 3/4.7.2
-
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FLD00 PR0TICTION.........'........................ I 3/4 7-1 i
l 3/4.7.3 REACTOR CORE ISCLATION COOLING SYSTEM...........
8 3/4 7-1
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SETIO88 fgel PLANT SYSTDes (Continued)
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3/4.7.5 SW88ERS........................................
8 3/4 7-2
3/4.7.8 SEALED SOURCE CONTAMINATION.....................
8 3/4 7-3
.
l 3/4.7.7 FIRE SUPPRE5510N SYSTEMS........................
8 1/4 7-4
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FIRE _eme. #waaviews
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3/4.7.8
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3/4.7.10 MM.M TURSINE SYPA55 $Y37EN......................
8 3/4 7-5
3/4.8 ELECTRICAL POWER D u ud
.
,
I 3/4.8.1, 3/4.8.2 and
3/4.8.3 A.C. SOURCES, D.C. SOURCE 5 and ONSITE POWER DISTRISLtTION SY5TEMS..... ~..................... 8 3/4 4-1
.
i 3/4.8.4 ELECTRICAL Equ! PENT PROTECTIVE DEVICES.........
83/48-3'
l
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l (d 3/4.9 REFUELING OPERAT108:5
-
m...i REACTOR =0E $wiTCN............................. 8 m 9-i
l 3/4.9.2 DISTRLMENTAT !0N................................. 8 3/4 9-1 3/4.9.3 CONTROL M0 P0$! TION.....................y......
8 3/4 9 1 3/4.9.4 DECAY T1NE......................................
I 3/4 9-1
I 3/4.9.5 C05UNICATIONS..................................
I 3/4 9-1
-
.
3/4.9.8 REFUELING PLATF38Bt..............................
8 3/4 9-2 i
3/4.9.7 CRANE TRAVEL-SPENT FUEL STORAGE P00L............
8 3/4 9-2
'
.
3/4.9.8 and 3/4.9.9 WTER LEVEL - REACTot VE5SEL
,
and WTER LEVEL - SPENT FUEL STORAGE P00L.......
8 3/4 9 2
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3/4.9.10 CONTROL B00 REMVAL............................. 8 3/4 9-2 3/4.9.11 RESIDUAL NEAT REmvAL Am COOLANT CIRCULATION... 8 1/4 9-2
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M 3/4.10 SPECIAL TEST f1cipTIONS 3/4.10.1 PRIMMtY CONTAIMMENT INTIGA1TY................... S 3/4 10-1 3/4.10.2 20 SEQUDICE CONTROL SYSTEM.....................
t 3/4 10-1
.
3/4.10.3 SWTDohm MARGIN 00eN5TRATIons.................. 8 3/4 10-1 3/4.10.4 RECIRCULATION Lo0PS.............................
3 3/4 10-1
3/4.10.5 OXYGEN CONCDmtAT10N............................ 8 3/4 10-1 3/4.10.8 TRAINING 5TAltTUPS...............................
B 3/410-1
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3/4. n RADICACTIVE EFFLUENTS
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3 3/4 u-1 3/4.u.1 U Qu!D EFFLUENTS........................................
u4. u.
.staus urtu NTs.......................................
. u u-2 3/4.u.3 SOLIO RADI0 ACTIVE tiA5TE.................................
'
8 3/4 u-S
'1lTAL 005E..............................................
S 3/4 11-6 3/4.u.4
.
.
3/4.12 RA0!0 LOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM......................................
8 3/4 12-1 3/4.12.2 LAND USE CENSUS.........................................
8 3/4 12-2 (
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5-1 Eastusten Ares.............................................
F1 Law,eputatten Zone........................................
5. 2 CONTAI M 4T 5-1 Conf 1geretten..............................................
.
5-1 Design Temperature and Pressure............................
Rea.c.b _ 4*d.&.> % (.s. m oaav s e e d
-
5-1
....._.,
... -........................................
5. 3 RDCTOR CO#E
5-4 Fuel Assamm11es............................................
5-4 Centrol And Asseme1tes.....................................
'
S.4 ttACTOR C00LANT SYSTEM l
J oesign,ress..e.n. 1e e,.tu,s............................
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5.5 NETiet0LOGI CAL TOWER (OCAT10N..............................
5.6 Futt STO4AGT PS C ri ti ca 1 1 ty................................................
F5 0rainage...................................................
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F5 Cepecity...................................................
F5 i
S.7 cgus0NENT CYCt!C On TRAN5ttMT LIMIT........................
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ADMINISTRATIVE C0hTROLS SECTION PAGE
'
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6.1 RESPON51SILITY............................................
F1
-
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l 6.2~ ORGANIZATION..............................................
5-1 6.,.1
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5-1 6. 2. 2 uM n srAn..........................................
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$NI FT TEDetI CAL ADVI50R.............................. 6-6
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6.3 UNIT STAFF QUALIFICATIONS.................................
6-6
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6.A m r N r No.................................................
6,r l
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6-7 l
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FUNCTION............................................ 6-7
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Col #051 TION......................................... 6-7
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ALTERNATI$...........................................
6-7
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6-5
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RESPONSIGILITIES...................'................. 6-8
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COMPOSITION......................................... 6-f j
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............................
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'STARTUP REP 0RT.......................................
Al AleluAL MPORTS...................................... 6-B I
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I ADMINISTRATIVE CONTACLS LIST OF FIGURES
PAGE FIGURE
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4-3 6.2.1-1 0FF51TE ORGANIZATION.................................
>
6-4 6.2.2-1 UNIT ORGANIZATION...................................
.
LIST OF TABLES PAGE
-
M 6.2.2-la MINIMJM SHIFT CREW C0fe051 TION - SINGLE UNIT FACILITY.............................................
5-5p i
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3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION
,
3.0.1 Compliance with the Limiting Conditions for Operation' contained in the succeeding Specifications is required during the OPERATIONAL CONDITIONS or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.
~3.0.2 Noncompliance with a Specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTIOR requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, coepletion of the Action requirements is not required.
.
3.0.3 When a Limiting Condition for Operation is not met, except as provided I
in the associated ACTION requirements, within one hour action shall be initi-
.
'
ated to place the unit in an OPERATIONAL CDMDITION in which the Specification
.-
does not apply by placing it, as applicable, in:
-
~
1.
At least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,
2.
At least NOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and
.
3.
At least. COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
,
Where corrective seasures are completed that permit operation under the ACTION
~
,
f requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for
-
Operation. Exceptions to these requirements are stated in the individual Specifications.
l This Specification is not applicable in CPERATIONAL CONDITIONS 4 or 5.
3.0.4 Entry into an OPERATIONAL CONDITION or other specified condition shall not be made unless the conditions for the Limiting Condition for Operation are set without reliance on provisions contained in the ACTION requirements. This provision shall not prevent passage through or to CPERATIONAL CONDITIONS as l
required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual Specifications.
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l 3.0.5 When a system, subsystes, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable limiting condition for Operation provided:
(1) its corresponding normal or>
emergency power source is OPERA 8LE; and (2) all of its redundant system (s),
subsystes(s), train (s), component (s) and device (s) are OPERA 8LE. or likewise satisfy the requirements of this specification. Unless both conditions (1)
and (2) are satisfied, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> action shall be initiated to place the unit in an OPERATIONAL CONDITION fn which the applicable Limiting Condition for Operation does not apply by placing it, as applicable, in:
1.
At least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, 2.
At least HOT SHUTDOW within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and
-
3.
At least COLD SHUTDOW within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
-:
,
This specification is not appitcable in OPERATIONAL CONDITION 4 or 5.
.
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s 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERACE PLANAR LINEAR HEAT GENERATION RATE
.
LIMITING CONDITION FOR OPERATION All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLNGRs) for each type 3.2.1 of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits ll shown in Figures 3.2.1-1. 3.0.'-2,..-4 :..i-2.
APPLICA8ILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 325}% of RATED THEMAL PCWER.
ACTION:
.
With an APLHGR exceeding the limits of Figure 3.2.1-1, 0.0.1-2. er 2.0.'-2, initiate corrective action within 15 minutes and restore APLNGR to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than
.
(253%ofRATEDTHERMALPOWERwithinthenext4 hours.
,
SURVEILLANCE REQUIREMENTS
.
All APLHGRs shall be verified to be equal to or less than the limits l
4.2.1 determined from Figures 3.2.1-1; 3.2.1-0,..t 0.0.1-::
l At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
'
a.
I Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at
-
b.
least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is
!
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operating with allMITING CONTROL 200 PATTERN for APU(R.
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POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS
\\
-
.
LIMITING CON 0! TION FOR OPERATION
.
The APRM flow biased simulated thermal power-gscale scras trip setpoint i
'
3.2.2 g)
(5) and flow biased neutron flux-upscale control rod block trip setpoint (5 shall be established according to the following relationships:
TRIP SETPOINT ALLOWA8LE VALUE
$ 1 (0.65W + 1511%)T 51(0.66W+154j%)T Su i (0.6EW + $45{%)T Sgg 1 (0.66W + (42)%)T are in percent of RATED THERMAL POWER, where: 5 and 5 recirculation flow as,6 percentage of the loop recirculation
,
'
'
W= flow which produces a rated core flow of (100f million 1bs/hr, gl T = Lowest value of the ratio of SFRACTION OF RATED THERMAL divided by the CORE MAXImM FRACTION OF LIMITING POWER DEM51TYKcmtec (f:: f y "!=: :f f=1 '
't =m)., (?.'?) ?:r ("
- ) f=1, ff:fid bj 2: ';**F 21:!=d j
T is applied only if less than L
R si ;
,-
er equal to 1.0.
.
OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or APPLICABILITY:
l l
equal to $25]% ef RATED THERMAL POWER.
ACTION:
With the APRM flow biased simulated thermal power-upscale scram trip setpoint and/or the flow biased neutron flux-upscale control rod block trip setpoint
'
less conservative than the value shown in the Allowable Value coluen for
\\.
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as above determined, initista corrective action within 15 minutes l
b'
5 er 5 and ad%s,t 5 and/or 5 to be consistant with the Trip setpoint values *
within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or redGEe THERMAL POWER to less than 125)% of RATED TH g
POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS I
The (FRTP and the CMFLP01 ('""F) shall be determined, the value of T calculated, and the most recent actual APRM flow biased simulated thermal 4.2.2 power, upscale scres and flow biased neutron flux-escale
..
i s.
At least once per 24 haves, Within 12 heves after completten of a THERMAL POWER increase of at
~
t.
least 1S% of RATED THERMAL POWER, and Initially and lit least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating ll with {CWLPO[ "") greater than er equal to (FRTP),(2 ??)_
c.
greater than the (FITPG (t:f p T) ir t ; m r = = :!:- 3
"With (CMFLPDF ('"**) TTfi.T '", rather than APRM setpoints, the AP94 gain any be adjusted such that the APRMM*3djusting the
""" ;f *^O *
eldings are greater n er equal
,.,; ^ :
.
100% times ICMFLPC) W T). provided that the adjusted APRMAre ding does not exceed 100% cf "T *"rli """S and a notice of adjustment is posted on U
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PC'.ER O!STRIBUTION LIMITS
.
MINIMUM CRITICAL PC'4ER RATIO (0;;i; :?-00Y" 0;t':
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-
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3/4.2.3 LIMITING CCNDITION FOR OPERATION 3.i.3 The MINId'M CRITICAL POWER RATIO (MCPR) shall be equal to or greater
" r :-f ""?". '^E"
limitp,et 'nd'::t:d ::r:
,_r,.._!_crfMCPR[__2. E P r h *S D.9..*5. U $"'i h M.~ k than h ti """9 ih5'r ;?:t'Ei j N UI; (5d5 5 i[ :h$t E IE 5 5?^55d:E5;5:I*'E$5'5-3;2.'.2).
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OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or
.
APPLICA81LITY:
equal to (253% of RATED THERMAL POWER.
,
ACTION:
- '
- 1:tt: ; n tr!; :pt: ' :;:rd!: ;:-
" O Oc ;nd-d' Vj:1 ation 3.3.4.2, operation may continue and the provisions (c.
a hour, Spe
.0.4 are not applicable provided that, with Specifica be equal to or greater than b f and MCPR, R
MCPR is deterni C-RPT inoperable curve).
as shown in Figures 3.2.3-3.2.3-2
$
With the main turbine bypas inop e per Specification 3.7.9, d the provisions of ication 3.0.4 are not ( s.
operation may contina that. yithin one hour, MCPR is de ned to be equal-1 and applicable p and MCPR, as shown in Figures
.
r than both MCRPy
,
to or
..".0-2 by W --?- he'-- W
- =a--=M=
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u w ha.4 6 K With MCPR 1ess than the ;;?f:$1: MCPR limiti;tr '- F.gurep j.'2.3-1 i
A-
-
,
.
2.2.2-2. initiate corrective action within 15 minutes and restore MC within the required limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL power to less than 525)% of RATED THERMAL POWER within the next a hours.
SURVEILLANCE REQUIREMENTS
.
...
'
.
,.
MCPR shall be deterstned to be equal to or greater than the applicable'
l 4.2.3 MCPR limit detarsined free Figure 3.2.3-1:
.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
a.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at b.
- least 151 of RATED THERMAL POWER, and Initially and at least once per 12 houn when the reactor is operating
.
with a LIMITIMG CONTROL 200 PATTERN for MCFR.
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1 0 JAN 592 POWER DISTRIBUTION LIMITS
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'3/4 2 4 LINEAR HEAT GENERATION RATE
..
.
LIMITING CONDITION FOR OPERATION The LINEAR HEAT GENERATION RATE (i.HGR) shall not exceed (13
'
3.2.4 OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or APPLICA8ILITY:
equal to 32!(% of RATED THERMAL POWER.
.
.
ACTION:
With the LNGR of any fuel rod exceeding the limit, initiate corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than (25p of RATED THERMAL POWER within the next l
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
- s SURVEILLANCE REQUIREMENTS
-
.
d 4.2.4 LHGR's shall be dotarsined to be equal to or less than the limit:
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, i
a.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at b.
least 15% of RATED THERMAL POWER, and l
Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating on a LIMITING CONTROL RCD PATTERN for LNGR.
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3/A.3 INSTRUMENTATION
.
'
3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION As a sinious,.the reactor protection system instrumentation channels 3.3.1 shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTIO RESPONSE TIME as shown in Table 3.3.1-2.
APPLICABILITY: As shown in Table 3.3.1-1.
ACTION:
WiththenumberofOPERA&EchannelslessthanrequiredbytheMinimum
.
OPERA 8LE Channels per Trip System neuinment for one trip system, place a.
the inoperable channel (s) and/or that trip system in tne tripped condi-The provisions of Specification 3.0.4 are not tion * within ese hours.
applicable. * P
,
With the number of OPERA 8LE channels less than requind by the Minimus OPERABLE Channels per Trip System requirement for both trip systems, place b.
at least one trip systee** In the tripped condition within.ese hours and
.
6*r take the ACTION required by Table 3.3.1-1.
SURVEILLANCE REQUIREMENTS Each reactor protection system instrumentation channel shall be 4.3.1.1 demonstrated OPERA 8LE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the fnquencies shown in Table 4.3.1.1-1.
LOGIC 1YSTEM FUNCTIONAL TESTS and simulated automatic operation of 4.3.1.2 all channels shall be performed at least once per 18 months.
The REACTOR PROTECTION SYSTEM RESPONSE TIME of each nactor trip functional unit shown in Table 3.3.1-2 shall be demonstrated to be within 4.3.1.3 limit at least once per 18 months. Ea'ch test shall include at least one
-
channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of neundant channels in a l
specific reactor trip system.
,
ipped condition where this l
"An inoperaale enannel need not be placed in these cases, the ir.o,erable channel In tf would cause the Trip Function to occur.
shall be nstored to OPERA 8LE status within Ehours or the ACTI**. required by l Table 3.3.1-1 for that Trip Function shall be taken.
l
- 1f more channels are inoperable in one trip system than in the other, p ace
the trip system with more inoperania channels in the tripped concition,
-
except when this would cause the Trip Function to occur.
.
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APPLICA8tE MINIMim OPERATIONAL OPERA 8L'E CHANNEL 5 II$
CON 0lil0NS_
PER 1 RIP SV51EN (a)
ACTION J;
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.
Intermediate Range Monitors (b);
I 1.
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-
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Reacter Vessel Steam Dome gg)
1 Pre,*.ure - Higle 1. 2 4.
Reacter Vessel Water Level - Low.
-
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1
-
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Main 5tcan line Isolation Valve -
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Closure
.
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p TABLE 3.3.1 _ 1 Continued)
REACTOR PROTECitati SYSTIM INSTRUNENTATION
.
Gj MIN 11tm L
APPtICA8tE OPERA 8tE CHANNELS h'F OPIRA110NAL PER TRIP SYSTEM (a)
ACil0N
"
CONiilll0NS__
FUNCTIONAL t#tti **
5 E.
Main Steam Line Radiation -
1,2(g)
f High 7.- (Primary Containment $ G.,
lil-1. 2(h)
1 Pressure - High 1/3 ScriukOlschargeVolumeWaterbd.Qig 1, 2fg
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Turbine Stop Valve - Closure
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10. Turbine Control Valve Fast. Closure, g gj)
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Valve Trip Systee Oil Pressure - Low
.
11. Reactor Mode Switch Shutdown
-
1. 2
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2 12.
3, 4
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TABLE 3.3.1-1 (Continued)
g REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION A nes t Be in' at least HOT SHUTDOWN within(12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. -
ACTION 1
-
Verify all insertable control rods to be inserted in the core ACTION 2
-
and lock the reactor soce switch in the Shutdown position within one hour.
Suspend all operations involving CORE ALTERATION 5* and insert ACTION 3
-
all insertable control roos within one hour.
Se in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
ACTION 4
-
Se in STARTUP with the main steam line isolation valves closed ACTION 5
-
within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
,
Initista a reduction in THERMAL POWER within 15 minutes and
.
ACTIDM 6 reduce turbine first stage pressure to
'000) ;;f;, ;;_f.: L..
-
0; '" 7.". "Z" ? ::: '2:. '2^ * :" "'.?s0..._....
__..y within 4,1,o h ha uWk by.p.a sdQ 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Verify all insertable control rods to be inserted within one ACTION 7
-
hour.
Lock the reactor mode switch in the Shutdown position within ACTION 3
-
one hour.
Suspend all operations involving CDRE ALTERATION 5*, and insert ACTION 9 all insertable control rods and lock the reactor mode switch in
-
the SMJTDOWN position within one hour.
.
"Escept movement of IRM, SRM or special movable detectors, or replacement of LPRM strings provided SRM instrumentation is OPERA 5LE per Specification 3.9.2.
-
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10 JAN 1983 TABLE 3.3.1-1 (Continued)
g
,
. REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS
+
.
.
(a) A channel may be placed in an inoperable status for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for required surveillance without placing the trip system in the tripped condition provided at least one OPERA 8LE channel in the same trip system is monitoring that parameter.
(b) This function shall be automatically bypassed when the reactor mode switch is in the Run position.
,
,
(c) The " shorting links" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn" and shutdown margin l
demonstrations are being performed per Specification 3.10.3.
(d) The non-coincident 25 reactor trip function logic is such that all channels go to both trip systems. Therefore, when the " shorting links" are removed.
-
the Minimum GPERABLE Channels Per Trip System is 4 APRMS and 6 IRMS.
(e) An APRM channel is inoperable if there are less than 2 LPRM inputs per
'
-
level or less than g LPRM inputs to an APRM channel.
.
(f) This function is not required to be OPERABLE when the reactor pressure l
vessel head is removed per Specification 3.10.1.
,
(g) This function shall be automatically bypassed when the reactor mode switch is not in the Run position.
(h) This function is not required to be OPERABLE when PRIMARY CONTAI MENT INTEGRITY is not required.
(1) With any control rod withdrawn. Not applicable to control rods removed
per Specification 3.9.10.1 er 3.9.10.2.
I
.
(j) This function shall be automatically bypassed when turbine first stage Pr '395 i
at t "Tf' """" ? :::
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Gst admap Te****". k Ps4 is um=J.
Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.
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TAttf 4.3.1,1-1 My g.
REACTOR PROTECTION SYSTEM INSTRt34fMTAY10N SINIVfillANCE REQUltfMENTS
.
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n CHANNEL OPERATIONAL m
i !fn
-
CHANNEL FUNCil0NAL CHANNEL CONDlil0NS FOR 1411CM
l'# ftNICTIONAL UNIT CHECK 1EST CALIBRAfl0N,)
SURVElllANCE REQUIREO g
d:F 1.
Intermediate Bange Monitors:
to a.
Neutron Flum - Higli 5/U,5,p
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'Reacter Vessel Water Level -
Law, Level 3 (59-M 4 fR1 1. 2 l
5.
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- G A
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- Q R
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TAetE 4.3.1.1-1 (Continued).
'
REACTOR P90TECT10N SYSTEN INSTRUMAil0N SURVEILLANCE REQUIREMS W.
.
OPtRATIONAL
- n CHANNEL s
bn CHANNEL FUNCil0NAL CHANNEL CONOli10NS FOR %RilCH l 'e FIRICTIONAL mli CHECK _
TEST CALIRRATION SURVElttANCE REQUIRED
. f,
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-
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- 4 1RI l
l
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11. Reactor Mode Switch Shutdown Posillon M
R MA
-
.
12. Manual Scram'
NA 4t-W NA
'
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g Neutron detectors may be escluded frem CHANNEL'CAtltRAY10el.
The IRM and 5m channels shall be deterefred to everlap for at least {%) decades during each startup
.
w (a)
af ter entering OPERATIONAL CONGITION 2 sad the IM and APM channels shall be determined to everlap I
E (b)
for at least (%$ decades during each controlled shutdown, if not perfereed within the prevleus 7 days.
"JR ";; M i-. Fkr 10 0teMa-. if aat - - M d ;.'t;. h r _ r...:.__ 7 1.,..
,
i This calibration shall consist of the adjustment of the APRM channel to confers to the power values l
-')
calculated by a heat balance during OPERAil0NAL C0fSITION I when THEMAL (d)
'
f Any AP M channel gain adjustment made in compilance ulth Specification 3.2.2 shall not be THERMAL POWER.
POWER.
included in deteretning the absolute difference.
This calibration shall consist of the adjustment af the APM flow biased channel to confers to a i
(e)
calibrated flew signal.
The LPRMs shall be calibrated at least once per 1000 effective. full power hours (EFPN)
(f)
[
Verify measured core flew to be greater than er equal to established core flow at the existing pump sp using the TIP system.
Mj :*c :t, :: ;;g!;d,:")(verifyinglthe6iTsecond h
]g))Thiscalibrationshallcensistof(t':
(
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(h slaulated'thetsal power time constant.1 l
z
,
This fianction is not required to be OPERA 8tE shen the reacter pressure vessel head is removed per
'
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Specificatten 3.10.1.
.
Not appilcable to centrol rods removed per Specification 3.9.10.1 (j) With any central red withdrawn.
er 3.9.10.2.
qs.
Calib.h ivio ud.4 d ie.d once per e d.93 M
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i OEC~6 R I275226
INSTRUMENTATION
__
3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION
3.3.2 The isolation actuation instrumentation channels shown in Table 3.3.2-1 shall be OPERA 8LE with their trip setpoints set consistent with the values shown
in the Trip 5etpoint column of Table 3.3.2-2 and with ISOLATION SYSTEM RESPONSE
~
- .
TIME as shown in Table 3.3.2-3.
'
APPLICA81LITY: As shown in Table 3.3.2-1.
ACTION:
With an isolation actuation instrumentation channel trip satpoint a.
less conservative than the value shown in t.Se C.lo-able values column of Table 3.3.2-2, ceclare the channel in:peraele untti the channel is restored to CPERA8LE status with its trip setpoint adjusted consistant with the Trip Setpoint value.
I b.
With the number of CPERA8LE channels less than required by the Minism CPERA8LE Channels per Trip Systes requirement for one trip systes,
,
l place the inoperable channel (s) and/or that trip systes in the tripped l
.
condition * within one hour.
The provisions of Specification 3.0.4
,
'
are not applicable.
With the number of OPERA 8LE channels less than required by the Minimum
-
c.
i OPERABLE Channels per Trip Systes requirement for both trip systems, place at least one trip systes"* in the tripped condition within one hour and take the ACTION required by Table 3.3.2-1.
.
"An inoperaole enannel need not be placed in the tripped condition where this l
g would cause the Trip Function to occur. In these cases, the inoperacle channel shall be restored to OPERA 8LE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or the ACTION required by Table 3.3.2-1 for that Trip Function shall be taken.
- If more channels are inoperable in one trip systes than in the other, place the trip system with more inoperable channels in the tripped condition, except
,
l when this would cause the Trip Function to occur.
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INSTRLHENTATION
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SURVEILLANCE REQUIREMENTS
.
.
4.3.2.1 Each isolation act eion instrumentation channel shall tie demonstrated CPERABLE by the performance of the CHANNEL CHECK. CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.2.1-1.
4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.
4.3.2.3 The ISOLATION SYSTEM RESPONSE TIME of each isolation trip function j
shown in Table 3.3.2-3 shall be demonstrated to be within its Itait at least Each test shall include at least one channel per trip once per 18 months.
systen such that all channels are tested at least once every N times 18 scnths, where N is the total number of redundant channels in a specific isolation trip system.
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TAatC 3.3.2-1 (Continued)
-
Hg Is0LATIONACTUAll0NiNSTRUMLNTATION n
VALVE ACTUA-
-
TION GROUPS MINIMUM APPLICARLE
.
li
-
C OPERAl[0 BY OPERA 8tE CIIANNE OPERAll0NAL
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O TRIP FUNCTION SIGNAL [d) *PER 1 RIP SYSitM I*
COH0lil0N _
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REActon CORE ISOLATION C00tlNG SYSTEM ISOLATION 1, 2 3
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RClc steam Line & Pressure -HI h
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TABLE 3.3.2-1 (Continued)
-
'
ISOLATION ACTUAT.,10N INSTRUNENTATION VALVE AC104-
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rs Il0N GROUPS NIMIMIM APPLICABLE DPERATED BY OPERA 8tE CHANNE OPERATIONAL I '70
-
SIGNAL (d) PER 1 RIP SYS1EN gjI CON 0lil0N ACTION j
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TABLE 3.3.2-1(Continued)
ht IsotATION ACTUA110N INSTRUNENTATION
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t VALVE ACTUA-
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,Q TION GROUPS MINIMUM APPLICABLE
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1 5 OPERATED BY OPERA 8LECHANNEgjI OPERATIONAL
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CONDITION ACTION i
SIGNAL (d) PER TRIP SYSTEM l
Li TRIP FUNCTI0li
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RHR SYSTEM SHLff00tal C00t1NG N00E 150LAT10N.-
.
.
.
a.
Reactor Vessel Water 2/Qu. (')
1, 2, 3
Level - Low, tevel 3
,
,
'
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Reacter Vessel (RNR Cut-in ft)h&c.ge)
1, 2, 3
I Permissive) Pressure - High
.
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-
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TABLE 3.3.2-1 (Continued)
~
.,
!$0LATION ACTUATION INSTRtNENTATION
. - -
- _.
...
,
.
,
.
.
.. -
ACTTON
--
-.. - - -....
..
..
. _..
....
ACTION 20 Se in at least HDT SHUTWWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and th COLD SHUTDOWN
-
'
within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Se in at least STARTUP with the associated isolation valves ACTION 21
-
closed withfr 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in at least HOT SHUTDOWN within
-
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
-
ACTION 22 -
Se in at least STARTUP within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
l
_. _. _.
-
-.
Close the affected system isolation valves within one hour
~ ~ ~
!
ACTION 23
-
and declare the affected system inoperable.
Restore the manual initiation function to OPERABLE status ACTION 24
-
within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT SHUTDCWN within the next
.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD 5HUTDOWN within the following 74 hours8.564815e-4 days <br />0.0206 hours <br />1.223545e-4 weeks <br />2.8157e-5 months <br />.
Restore the manual initiation functiori to 0PERA8LE status ACTION 25
-
within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or close the affected systea isolation valves within the next hour and declare the affected system incterable.
!
-
-
F.th. hen,Il4 coktien anu REACTOR @JILbWG b hla b.
sb (FRVS)
.
Establish SECONDARY C0"'AlhMEfff IATEGRITY wi the ct--ey ;;s ACTICN 26
-
4eest;;..; ayetes operating within one hour.
Lock the affected systes isolation valves closed within one
ACTION 27
-
,
hour and declare the affected systas inoperable.
NOTES When handling irradiated fuel in the secondary containment and'during
CORE ALTERATIONS and operations with a potential for draining the reactor vessel.
.
%en og hrbid stir vet. 4 yeater b 10 7, que ad[or when Ae.
key loded b Pa55 swthh i in 4t. NCRM Posif.*n -
(a) A channel sai be placed in>an* inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for
/
.
.
required surveillance without placing the trip system in the tripped con-
dition provided at least one other OPERABLE channel in the same trip
'
system is monitoring that parameter.
(b) Also trips and isolates the mechanical vacuum pumps.end-stean-fet-*4c-
.
- $*0$*h--
(c) Also starts the.nandby-ga: in;txt ;y;t% FRVS.
.
(d) (Insert A>
(s.) Sensors arranges per valvf. 9 rove, not per frip sys4em.
(f) Closes only RWCU systas isot.ti n valves HV-foor..S Hv.poov.
l
~
(g) Requires systes steam supply pressure-law cot.ncideny witn ifrywell pressure-high t clog twinine exhnt acwm breakar wtwi.
.
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M
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(kkr)
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=
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[lakr)
4 Temperature - High
.
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KIC'PipeRoutingkrea (i k<) A (L4r)f Toeperature - High i t,. 7 : s n n }.9. s. N i} k 5163 psi)
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M SYSTDI SM30lii C00 Lim leoE 150tAitoll;
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a.
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,
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Any required change Final setpoint te be determined during startup test program.
.*
to this setpoint shall be subeltted to the Commission within 90 days of test completten.
- Initial setpoint.
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TABLE 3.3.2-3
.
150LATION SYSTEM INSTRUMENTATION RESPONSE TIME RESPONSETIME(Seconds)$
TRIP FUNCTION Ia$$ov
ISOLATION
.
^
1.
'
Reactor Vessel Watekevel I a,len! )
L
- 10 a.
1)
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Ls
< c.:7/= (m p g
b bu.k M:..Oh 6 w
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5 m o.g-
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= =..
n.
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l a
Radiation - High(b)
< (1.0)*
(13)(
),,l Main Steam Line
c.
1)
< 1.0)a (1:
2)
Pressure - Low
~ ((0.5)* <J)(a),,
3)
Flow - High
-
i d.
Main Steam Line Tunnel Temperature - High (NA)
-
,
in Steam Line Tunnel A Temperature - High (NA)
e.
f..
ndenser Vacuus - Low (NA
'
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O ell and/or Suppression C
r Radiation - High NA)
-
MA
.
h.
Manual nitiation
<
1.
SECONDARYCONTAMENTISOLATION l
2.
< 13)(a)
h(((13)(a)
I a.
Plant Exhaust lenum Radiation - High(
b.
D:ywell Pressure - High 13)(*)
c.
Reactor Vessel Wa r Level - Low Level 2
<
d.
Refueling Floor Exh st Radiati
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7 (13)(*)
,
,
RA Manual Initiation e.
-
f.
'
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REACTOR WATER CLEANUP SYSTEM LATION 1 (13)(a)(N)
s.
A Flow - High
,
b.
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c.
Heat Exchanger /P Area Ventila on (NA)
Temperature AT High NA
'
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< (13)I")
d.
SLCS Initiati BA i
f.
Manual In ation
_
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REACTORCDdISOLATIONCOOLINGSYSTEMISOLATION (
)
!
4.
< (13)
!
s.
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b.
C Steam Supply Pressure - Low
.
IC Turbine Exhaust Diaphragm Pressure - High T
(NA)
!
c.
IC Equipment Room Temperature.- High d.
RCIC Steam Line Tunnel Temperature - High (NA)
- ~
RCIC Steam Line Tunnel A Temperature - High (NA)
e.
.
.
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I INSERT A TO PG. 3/4 3-37
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Footnotes to Table 3.3.3-1 (cont 'd) :
(g).In divisions 1 and 2, manual initiation is associated with each pump and valve combination; in divisions 3 and 4, manual initiation is associated with each r
pump only.
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TABLE 3.3.3-1 (Continued)
!
'
.
BERGENCY Co#E COOLING SYSTEM..CTUATION INSTRUMENTAT
'
EH95
,
With the numer of OPERA 8LE channels less than required by the i
Minieue 0PERA8LE Channels per Trip Function requirement:
ACTION 30 -
With one channel inoperable, place the inoperable channel l
in the tripped conditten within one heur" er declare the I
a.
I associated system inoperable.
l l
With mere than one channel inoperele, declare the 6.
associated system inoperable.
l
.
,
,
l With the nebetof OPERABLE channels less than nouind by the l
ACTION 31 -
OPERA 8LE Channels per Trip Function requirement, declare
<
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the associated ECC5 inoperable.
i With the neber of OPERA 8LE channels less than requind by the
l l
Minis m CPERA8LE Channels per Trip Function requirement, place ACTION 32 -
t the inoperable channel in the tripped condition within ene hour.
l 7 M L^Z'f'i C.;c h:: S:
- i
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l J.;;~;C^; 2
-
, p ace run. r **ERAaLE Channels per Trip Functi the inope rable cha...-C *" L ;- condition within one hour;
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-*"a within 7 days j
g=m-4he-tegrFali e channel to ortna.:::rr'-t1 :;"
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With the numer of OPERA 8 2 channels less than ret:utred by the
-
Mateue OPERA 8LE Channels per Trip Function requirement, notere l
-
ACTION 34 -
i the inoperable channel to OPERA 8LE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> er
!
!
declare the associated ECCS inoperable.
!
With the naber of OPERA 8 d channels less than required by the ACTION 35 -
Mataue OPERA 8LE Channels per Trip Function requirement:
For one trip systas, place that trip systes in the tripped
conditten within one hour * er declan the NPCI system a.
'
For both trip systans, declare the NPCI system inoperable.
~
l 6.
with the neber of OPERA 8LE charinels less than required by the
,
Mnimum 0PERA8d Channels per Trip Function requirement, place at i
i ACTION 36 -
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ienst one insperele channel in the tripped condition within i
er declare the NPCI system inoperable.
i a
ene hour l
M
.e a m.or.f m RA8 d channeis isss than the Totai Nu. e,
i
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of Channels, declare the associated emergency diesel generator ACrION 37 -
inoperable and taka the ACTION mquired by Specification 3.8.1.1 er 3.8.1.2, as appropriata.
l
With the number of OPERA 8 d channels one less than the
,
i
Muster of Channels, place the inoperable channel in the tripped l
ACTION 38 -
condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />;" speration any then continue until
-
f performance of the next required CHANNEL FUNCTIONAL TEST.
'
"The provisions of Specification 3.0.4 are not applicable.
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3/a.3.4 RECIRCULATION PJer TRIP ACTUATION INSTelp(NTATION l
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ATW5 RECIRCULATION PUMP TRIP SYST S INSTRLpqENTATION '0;;' z :
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LIMITING w=GITION FOR OPERATION The anticipated transient without scras recirculatten puep trip
-
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(ATWS-RPT) system instrumentation channels shown in Table 3.3.4.
3.3.4.1
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!
Setpoint column of Taste 3.3.4.1-2.
l APeLICABILTTY: OPERATIONAL ColeITION 1.
I
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With an ATW5 recirculatten pump trip system instrumentatt'on chan
!
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trip setpoint less conservative than the value shown in the A11rmab f
'
a.
Values column of Table 3.3.4.1-2. declare the channet inoperable u the channel is restored to OPERABLE status with the channel trip
i setpoint adjusted consistent with the Trip Setpoint value.
With the number of OPERAELE channels one less than required by the
Minimum GPERABLE Channels per Trip System D.
!
condition within one hour.
l
With the number of OPERA 8LE channels two er aere less than reql by the Ninfeue OPERA 8LE Channels per Trip system requireme i
c.
trip systas ende dechre. %a %p spbtm %pc aMe.,
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If the ingesmo reactor vessel pressure channe
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With one trip systas inoperable, restore the in
..,,.,._
-
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- the next 4 hoves.
With both trip systans inoperable, restare at le l
e.
the next 4 hoves.
.
.
SURVEILLANCE st@Utita nT5 Each ATW5 recireviation pug trip system instrumeritation ch A EL shall be demonstrated OPERA 8LE by the performance of the C 4.3.4.1.1.
in FUNCTIONAL TEST and CNANNEL CAL 28AAT10N sperations at t
-
'
Table 4.3.4.1-1.
-
f 4.3.4.1.2 LOGIC SYSTS FUNCTIONAL TESTS and eisulated aute all channels shall be performed at least once per 18 months.
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10 JAN 1993 INSTRUMENTATION
_
30-OF-CYtt.E RECIRCULATION W W N
.
LIMITING CONDITION FOR OPERATION
-
The end-of-cycle recirculation pump trip (E0C-RPT) systes E with instrumentation channels shown in Table 3.3.4.2-1 in the Trip 5etpoint 3.3.4.2 h
the(r trip setpoints set consistent with the values s ow
.
ATION PUMP TRIP
.
SYSTEM RESPONSE TIME as shown in Table 3.3.4.2-3.
.
OPERATIONAL CQaQITION 1, when THERMAL POWER is
,
APPLICA81LITY:
equal to 130p of RATED THERMAL POWER.
tion 3:
With an end-of-cycle ncirculation pump trip h
l Allowable Values column of Table 3.3.4.2-2. dec
.
a.
the inoperable until the channel is notand to CPERA8LE stat l
channel setpoint adjusted consistent with the Trip 5etpoint.
by the With the number of CPERA8LE channels one less than requ i
both Minimum CPERA8LE Channels per Trip System requireme
_
i d condition I
b.
trip systees, place the inoperable channel (s) in the tr ppe
"
within one hour.
ired With the number of CPERA8LE channels two or son less th t for one by the Minf aum CPERAILE Channels per Trip Systes n c.
- trip systes and:
If the inoperable channels consist of o
,
!
l 1.
channels in the tripped condition within one hour.
- If the inoperable channels include two i
I 2.
'
system inoperable, i
ten With one trip systes inoperable, motore the inoperabic t
?"?:0". ;;;'. ;: 53
,
tr:
ts OPERA 8LE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> er 't;':';;;'.0.S d.
"e; THERMAL POWER within the next 5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />st.
.
,
RATED i
stem
.
With both trip systees inoperable, restore at least one tr
.
" :0".;; 'c;d ty
.
to OPERA 8LE status within one hour er '
l
-
s.
RATED THERMAL POWER within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
.
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INSTRUMENTATION l
SURVE1LLANCE REQUIREMENTS Each end-of-cycle recirculation pump trip system instrumenta ANNEL channel shall be demonstrated CPERA8LE by the performance of th 4.3.4.2.1 in FUNCTIONAL TEST and CHANNEL CALIBRATION operations at t Table 4.3.4.2.1-1.
LOGIC SYSTEM FUNCTIONAL TESTS and simulated automa a11 channels shall be perforwed at least once per la months.
4.3.4.2.2.
- The END-OF-CYCLE RECIRCULATION PtMP TR be within l each trip function shown in Table 3.3.4.2-3 shall be demonstrated toEa 4.3.4.2.3
its limit at least once per 18 months.
logic of one type of channel input, turbine control valve fast closur tested
,
turbine stop valve closure, such that both types of channel inputs are
- -
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fee-OF-CYCLE RfCIRClitATION Mar TRIP $YSTEM INSTRUENTATION
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MIN 11RSI o
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PER TRIP SYSTEM
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Tertine Step Valve - Closure 2(b)
-
,
Turbine Centrol Velve-Fest closure i
i 2.
!
-
I
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ttlance provided A trip system may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required serve t'
I*I
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that the other trip system is OPERABLE.
(b)This functlen shall be outematically bypassed when tortfae first stage pressure is l
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INSTRUMENTATION r, r p,,.
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.
I TRAVERSING IN-CORE PROBE SYSTEM LIMITING CONDITION FOR OPERATION
.
3.3.7.7.
The traversing in-core probe system shall be OPERABLE with:
a.
Five novable detectors, drives and readout equipment to map the core, and b.
Indexing equipment to allow all five detectors to be calibrated in a common location.
APPLICABILITY: When the traversing in-core probe is used for:
a.
Recalibration of the LPRM detectors, and b.*
Monitoring the APLHGR, LHGR, MCPR, or MFLPD.
ACTION:
With the traversing in-core probe system inoperable, suspend use of the system for the above applicable monitoring or calibration functions. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
.
SURVEILLANCE REQUIREMENTS 4.3.7.7 The traversing in-core probe system shall be demonstrated OPERABLE by normalizing each of the above required detector outputs within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to use for the LPRM calibration function.
,
l
.
"Only the detector (s) in the required seasurement location (s) are required to
'
be OPERABLE.
SEP 3 C ass HOPE CREEK 3/4 3-89 i
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3/4.5 EMERGENCY CORE C00 LING SYSTEMS 3/a.5.1 ECCS - OPERATING LIMITING CONDITION FOR OPERATION
.
The emergency core cooling systems shall be OPERA 8LE with:
3.5.1 The core spray systes 446G3 consisting of two subsystems with each
'
a.
sutsystem comprised of:
emet sem9 l
1.
jTwel CPERA8LE466 pump (sJ, and i
An OPERABLE flo'w path capable of taking suction from the
^
2.
suppression chamber and transferring the water through the spray sparger to the reactor vessel.
The low pressun coolant injection (LPCI) system of the residual
,
b.
heat removal system consisting of e subsystems with each
.
W subsystem comprised of:
-
l Om 1.
ifwe+ 0PERABLE LPCI pump (s), and An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor l
2.
vessel.
The high pressure cooling injection (HPCI) system consisting of:
c.
. 1.
One OPERABLE HPCI pump, and An OPERABLE flow path capable of taking suction from the 2.
supp nssion chamber and transferring the water to the reacter
vessel.
l@
l The automatic depressurization system (ADS) with t hnt (:i ) 64 d.
OPERA 8LE ADS valves.
l OPERATIONAL CONDITION 1, 2*, ** f, and 3*, **,"#
APPLICABILITY:
"The HPCI system is not required to be dPERA8LE when nactor steam dome l
pressure is less than or equal to g psig.
The ADS is not nouind to be OPERA 8LE Twhen nactor steam done pressure is
,,
less than or equal to {1001psig.
- ee special Test Exception 3.10.6.
,
0w,. L.Pt1 sAsp em ch he RHR syde= m3 be brerab\\e W %d
h
% is edged in %s skddow coob3 v^ ode h yearhv vesse(
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Pe=ssue. h less %s %e 1(Mt c4-6 wks.ioe. vd.p4.
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EMERCENCY CORE COOLING SYSTEws LIMITING CONDITION FOR OPERATION (Continued)
ACTION:
For the c'ere spray systes:
'
a.
c.eee se<*a3 1.
With one 466 subsystem inoperable, provided that (at least-ene-bo W ! ; n 5-) ---- LPCI subsystem OPERA 8LE, restore the Cc"SymfinoperacleA45& subsystem to CPEAA8LE status within 7 days or be in at least NOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
covsspeap 2.
With both 4&& subsystems inocerable, be in at least HOT SHUTCChN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
For the LPCI systas:
g,,
,p.,3 we==ee. W.as hweL"'! ; n ' :t e:r :r,[t: e LPCI subsystems inope 1.
With :::
provided_that at leaft one C&&Asutsystem is CPERA8LE, restore
.
the Inoperable LPCI.n--(t) to CPERA8LE status within 7 days or
___
'
A fem
be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD 5HUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
'
i 2.
"iO (:) (nd L.*C! :y:!- cr:::-t!: n! : !!: -' (: e t'
- -
--!. r : nd "n: t: ! n ed ern -tf: n!n ::: :t:-), 5: *- it
..... un, e to m
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..e 4. rm e e m m ruu 4, w 4.
s 2: nt 2' h: :.
6ece 2.\\
With.eae LPCI subsystems;O; _f = inoperable, ;n 'Of 2:" ;;O
'
C: ;_ :j n ; ; : n 0"!1'"LE, restors 2: ' :; et: L*CIlsDrystem to OPERA 8LE status within or be in at least HOT SHUTCCWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.s
,7 n
3'4 With toe-LPCI subsystems :r: f ee inoperacle, be in at least l
Co <-
HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDokh within the
.
next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />."
For the HPCI system, provided the 466,Mthe LPCI system, the A05 and core S
,
i c.
the RCIC system are OPERA 8LE:
With the HPCI system inoperable, restore the HPCI system to 1.
OPERASLE status within 14 days or be in at least NOT SHUTDChN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor steam done pressure l
.
to i tt96t psig within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- o
,
.c l
"Whenever two or nors RHR subsystees are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, saintain reactor coolant temperature
"
as low as practical by use of altarnate heat removal methods.
SeeE ctEtu.
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3/4 5-2
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EMER0ENCY CORE COOLING SYSTEMS LIMTING CON 0! TION FOR OPERATION (c*ontinued)
ACTION: (Continued)
d.
For the ADS:
,,,g qb 1.
With one of the above quired ADS valves inoperable, provided the.HPCI systee, the p and the LPCI systen are OPERA 8LE,.
restore the (noperable A05 valve to 0PERAS*.E status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> l and reduce reactor steam done pressure to 1(1001 psig within the next 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2.
With two or more of the above required ADS valves inoperable, be in at least NOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reacter
'
steam dose pressure to 1)100],psig within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
l In the event an ECCS system is actuated and infects water into the e.
Reactor Coolant Systes, a Special Report shall be prepared and sub-
,'
sitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to data. The current value of the
.
.
useage factor for each arrect.ed safety injection neule shall be
!
provided in this Special Report whenever its value exceeds 0.70.
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EMERGENCY CORE COOLING SYSTEMS i
, SURVEILLANCE REQUIREMENTS
_
4.5.1 The emergency core cooling systems shall be demonstrated OPERABLE by:
At least once perf31 days: cwa em spw,,
s.
1.
For the 4ft,4 the LPCI system, and the HPCI system:
a)
Verifying by venting at the high point vents that the
[
system piping fros,th, pump discharge valve to the system isolati9n valve is filled with water.
b)
Verifyfng that each valve, manual, power operated or automatic, in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct" position.
.
:r W L*C! :y ", ::r' *y'nI:W t (th:) (:t 1:::t :::) LPCI
- ': (:;:r) (:?:::d ef u ;: :-
- j;ts
- 2:y:'
rn:-t'; ::
-
r n : = d ': n W : 1re - r
- as).
2.'1, For the HPCI system, verifying that the HPCI pur.p flow controller I is in the correct tion.
sw hursuanttoSpecification4.0.5:
ews Verifying that,(64 pumps in each subsystem together develop a flow of
!
when test b.
The two 46 1.
or equal to tef psig, tr-n;-- W ; u :
- u r = n :1 prn :ur: ~ 1 at least163501. gpe against a test line pressure of greater than l
A (115) p ';.
l
- '
TN w
x; LPCI pumps in each subsystem tq;th:r developsa flow of l
2. _at least ':1,000) gpa against a test line pressure of > (el psig, l I
n. r n;=d. '.n; u : :::t:r =,,,n,,:1
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3.
The HPCI pump develops a flow of at least 4600-gpm against a l
when steam is being supplied test line pressure of >rtHGG-) psig! psig.""
l t "1"05,
-l'.0, - 20 d-to the tuttiine aM N tooepto,-So
'*
SYS At least once 7pe
-
18 months:
l c.
1.
For the 46&, the LPCI systes, and the HPCI system, performing a system functional test which includes simulated automatic
'
actuation of the systas throughout its emergency operating sequence and verifying that each automatic valve in the flow
-
path actuates to its correct position. Actual injection of
-
coolant into the reactor vessel say be excluded from this test.
.
"Except snat an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present say be in po/sition for another sede of operation.
- The provisions of Specification 4.0.4 are not applicable provided, the surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor staas pressure is adequate to perfore the test.
hM-
- W.ke b.be. deier,m6el bh Pre-c9 S
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EMEMENCY CORE CCOLING SYSTEMS
..
v.,
$','8VEILLANCE RECUIREMENTS (Continued)
2.
For the NPCI system, verifying that:
a)
The system develops a f w of at least G O ") gpa against a test line pressure of psig, :; ::;:- S ; i: - :_..' ' ; ;,--
- ::: :--
gl
--
.;;;;: ;.:::_ : :" ' 155 ::';. when steam ;2*'.. r.'fms ac a
.
- *-
+ @ a.15)psig.**
L _ ;_.: --
'
aoo o
The suction is automatically transferred from the condensate, b)
storage tank to the suppression chamcer on a condensate storage tank water level - low signal and on a suppressien chamb,er - water level high signal.
,
For the ADS:.
At least once per 31 days, performing a CHANNEL FUN TIONAL TEST d.
1.
of the !!"
'!*-- 5- "- r _
--:: fAas system lo(,, essure alarm system.
p -*g e.-e.
=..=4 lude....t
2.
At lea'st one's per 18 months:
Per.f.gtsting a system functional test which includes l
a)..
simulat~ed. automatic actuation of the syst6m throughout its emergency operating sequence, but excluding actual valve
,
actuation.
Manually opening each ADS valve when the reactor steam done pressure is greater than or equal to 100 psig(*=) and l
b)
observing that either:
l 1) -The control valve or bypass valve position responds accordingly, or l
,.,
There is a corresponding change in the measured steam 2)
y,,5 n.:.
4jaJ. d
t1ow.
Performing a CHANNEL CALI5RATICN of theAr~ "-"? ?'a-c)
gas systes low'$fessure alars system and
" - - -
r ; :::-d verifying an alare setpoint of {Ef) + -js} psig on decreasi.ng pressure.
.
.
.
l I
.
!
.
i
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,
,
l
""The provisions of Specification 4.0.4 are not applicaele p
.
-
-
.
{
adequate to perform the test.
I Value 6,
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EFERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS - SHUTOOWN
.
LIMITING CONDITION FOR OPERATION
'
six emr3ew cere. coolas s4syfeas l
43,. mini m;
n y two of the followingshall be OPERABLE:
3.5.2 e.d l
a.Tw [ ore spray system 4Gr) subsystems,with e subsystem comprised of:
e l
core sysy l
(?t h=t :=) 1Two] OPERABLEAGE pump (s), and 1.
-
An OPERABLE flow path capable of taking suction from at least 2.
one of the following water sources and transferring the water l
through the spray sparger to the reactor vessel:
a)
From the suppression chamber, or
-
When the suppression chamber water level is less than the I
b)
limit or is drained, from the condensate storage tank
containing at least M available gallons of water, equivalent to a level) of (275%.
ea d.
y lSs,oco l
r* Low pressure coolant injection (LPCI) system subsystems.with-a-
-,
- ~
'
b. A subsystem comprised of:
1.
At hat One OPERABLE LPCI pump, and
-
An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor ll 2.
s_
vessel.
.
M PLICABILITY: OPERATIONAL CONDITION 4 and 5*.
ACTION:
~
Withoneoftheaboverequiredsubsystem(s) inoperable,restoreat least two subsystem [s) to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or suspend a.
, all operations with a potential for draining the reactor vessel.
,
'
W'ith both of the above required subsystems inoperable, suspend CORE'
ALTIAATIONS and all operations with a potential for draining the b.
Restore at least one subsystem to OPERABLE status reactor vessel.
withiij4hoursorestablishfCCCNCYCGTAbiniINTEGRITYwithi.
N ct " " * 4 ' " Y * * * n)
the neat 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
-
"The ECC5 is not, required to bE OPERABLE provided that the reactor vessel head water level is saintained within the limits of Specification '3.9
-
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}tCPg C RS E V.
3/4 5-6 U:T: (na/4)-
3 0 MAP 1993
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
EMERGENCY CORE C00 LING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.2.1 At least the above required ECCS shall be seecnstrated CPERABLE per l
Surveillance Requirement 4.5.1.
4.5.2.2 The core spray system shall be determinal0PERABLE at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying the condensate storage tank required volume when the condensata storage tank is required to be OPERABLE per Specification 3.5.2.a.2.b). l
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goes CR. tent
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CI T i (%?,/t)
3/4 5-7 3 0 MAP 1993
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__________________ __ ___________________ ____ ____ _____________ _ ________u
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10 JAN 883 EMERGENCY CORE COOLING SYSTEMS l
O 3/a.5.3 SUPPRESSION CHAMBER
-
ERATION
_ LIMITING CONDITION FOR OP The suppression chamber shall be OPERABLE:
3.5.3 In OPERATIONAL CONDITION 1, 2 and 3 with a contained w
)
at least '07,'001 f ta, equivalent toVe-level of G2'0'F74,5 a.
e aa.A.L en AuJed In OPERATIONAL CONDITION (4 and 5' with a c l
us, coo
!
Mexcept that the suppression Int,eco-(--)- f t*,
equivalent to 4)1evel ofchamber level may be less than t 74,P b.
mit or may be drained provided
,
'
l that:
No operations are performed that have a potential for drainin
,
1.
the reactor vessel,
!"
The reactor mode switch is locked in the Shutdo
.
2.
=
availab e(. l position, I'55 ooo The condensate storage tank contains at least4150,,000)
-
gallons of water, equivalent toye-level of TD$, and 3.
m dea.M I
The core spray system is OPERABLE per Specification 3 an OPERABLE flow path capable of taking suction from the 4.
condensate storage tank and transferring the water throug spray sparger to the reactor vessel.
-
OPERATIONAL CONDITIONS 1, 2, 3, 4 and 5*.
APPLICABILITY:
In OPERATIONAL CONDITION 1, 2 or 3 with the suppr ACTION:
i hin level less than the above limit, restore the water level to w t
f TDOWN within the
~
a.
the limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT SH 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
-
'
In OPERATIONAL CONDITION 4 or 5* with the suppre level:less than the above Ifmit or drained and the above tions b.
condibions not satisfied, suspend CORE ALTERATIONS a l and lock the tGrhave a potential for draining the reactor vesseEstablish SECONDA reactor mode switch in the Shutdown position.
CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
I'
that the
- Tae suppression chamber is not required to be OPERABLE p being flooded from reactor vessel head is removed, the cavity is flooded for ed (when the cavity the suppression pool), the spent fuel pool gates are remov h limits of is floodedy, and the water level is maintained within t e s ffHss:ces re p re d s, (
Specification 3.9.8 and 3.9.9 a
5 e tfi d e~
3.&. z,4 fu fmswe t#
See f
3/4 5-8 3*LE,C... Mik,,
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_
_ _
_
-
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EMERGENCY CORE COOLING SYSTEMS P
SURVEILLANCE REQUIREMENTS I
ERABLE by verifying The suppression chamber shall be deterstned CP l to, 4: ;;!'-?!ey
thepater level to be greater than or equa
4.5.3.1 twMedtel C 97at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
"
g
' N.
s.
.
- t ';
- :t ;;;; pr (M) h; r..
bove limit or i
(
With the suppression chamber level less than the at least once 4.
'
I drained in OPERATIONAL CONDITION 4 or 5*, a 4.5.3.2 3.5.3.b to be Verify the required conditions of 5pecification l
a.
satisfied, or VeHfy footnote conditions * to be satisfied.
b.
.
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3/4 5-9 W.ee. cautw 45-N (**/+)-
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