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 N (OL)
Facility Docket No: 50-354 Licensee: Public Service Electric and Gas C 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 / / 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
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.f Strengths noted were as follows: Safety system knowledge was good with exception of item 3. abov . Knowledge of control roon back panel operation had improved significantly since the last examination perio . The candidates were quick to use the procedures to correct off-normal conditions. This also is an improvement since the last examination perio . 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: A weakness in abilities to interpret and apply Technical Specification . Three points were deleted from this section (see paragraph 8) thereby increasing the relative importance of the remaining point . Several candidates delayed completion of this section and were rushed to complete i 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. failure It appears that the candidates were trained to a standard BWR-4 F.W. 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 classe . Personnel Present at Exit Interview:
NRC Personnel Frank Crescenzo
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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 Summary of NRC Comments made at exit interview: The items noted in Paragraph 3 above were discusse The results of a training program inspection documented in Inspection Report 50-354/86-15 were discussed. Summary of facility comments and commitments made at exit interview: Plant management is currently working on providing technical specification interpretation guideline The written examinations were too lon Examiners were professional and competent. 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 startu to this.
2.03 Item 3 not required unless Pressure detector is RPV assumed depressurize upstream of IV' .06 Delete: " Gland Seal steam Clean steam used for isolation". gland seal stea .02 No change for Part Feed flow is detected at individual feed pum .03 Add rod 39-27 to Part Alternate acceptable answe .
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3.05 Delete: "PAM" from answe Not really a range but rather a syste .06 "a" is lower Relay Room Correct answe "d" is delete New logic is tru .10 Delete "f". "3 running" was not defined as before or after. Question invali .01 Changed answer setpoint Reflects current T. S. setpoint .05 Deleted 1.5 point Question was beyond scope of examinatio .03 Same as 2.03 Same as 2.03
6.07 Changed answe Reflects updated materia .12 Same as 2.0 Same as 2.0 .03 Changed answer to reflect Reference provide correct T.S. actio .04 Question delete No reference exist .06 Answer changed to reflect Reflects correct C.S. subsystem ino interpretatio .07 Deleted answer for most This term is not clearly limitin define .08 Question delete Reference changed, no longer valid.
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8.11 Answer changed to allow Correct T.S. interpreta-I hour to trip channe tio Attachments: Written Examination and Answer Key (SRO) Written Examination and Answer Key (RO)
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U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: HOPE CREEK
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REACTOR TYPE: BWR GE4
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DATE ADMINISTERED: 86/02/24
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EXAMINER: CRESCENZO, ;
APPLICANT ____
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INSTRUCTIONS TO APPLICANT:
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Use separate paper for the answers. Write answers on one side onl Staple question sheet on top of the answer sheet 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 start % OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY
________ ______ ___________ ________ ___________________________________
25 0 25.00 PRINCIPLES OF NUCLEAR POWER
___1_0___ ______ ___________ ________
PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 25 0 25 0 PLANT DESIGN INCLUDING SAFETY
___1_0___ ___1_0_ ___________ ________ AND EMERGENCY SYSTEMS 25.00 25.00 INSTRUMENTS AND CONTROLS
________ ______ ___________ ________ ________ PROCEDURES - NORMAL, ABNORMAL,
___ 1____ _25 00__1__
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EMERGENCY AND RADIOLOGICAL CONTROL 100.00 100.00 TOTALS
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FINAL GRADE _________________%
All work done on this examination is my own. I have neither given nor received ai EPPL5CEATI5~555 ETUR5~~~~~~~~~~~~~~
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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 answe (1.50)
GUESTION 1.02 (2.00)
a. The reactivity worth associated with equilibrium samarium concentration 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 occur (1.00)
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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 (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 depressio (0.50)
OUESTION 1.05 (3.00)
a. 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)
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- (3.00)
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4 QUESTION 1.07 (1.00)
' Which of the following requires a steam condenser to remove the most 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 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
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60 des F by heaters. The tank is designed such that it could
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. withstand 15 psi differential pressure in either direction. Assume the vent becomes totally clogged while the pump is in operatio Answer the following question 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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- 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 without cavitation if the vent were open? EXPLAI (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?
(Show calculations) (1~.5)
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b. Explain WHY power decreased at this rat (1.0)
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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 your answe (1.00)
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 gain adjusment necessary? (1.00)
GUESTION 1.12 (2.00)
Which of the followin3 situations is correct in most cases? (2.00)
Explain your choic A. A control rod's worth is greatest when it is fully withdrawn and all other rods remain inserte B. A control rod's worth is greatest when it is fully inserted with all others withdraw .
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PAGE 6 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
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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 provide adequate drivin3 force to the CR0 mechanism to ensure positive Scram insertio ...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 deenergize ...will not prevent their associated CRD's from Scramming if they fail to open, providing that reactor pressure is Sreater than 200 psi ... connect the Scram accumulator to the under-piston port in the Control Rod Driv 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
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with the failed componen Assume NO OPERATOR ACTION, and the component is in the failed condition at the time RCIC receives the auto initiatin3 si3na a. The Barometric Condenser VACUUM PUMP fails to operat (1.00)
b. The MINIMUM FLOW VALVE fails to auto open (STAYS SHUT) when system conditions require it to be ope (1.00)
c. The RCIC PUMP DISCHARGE FLOW ELEMENT OUTPUT SIGNAL (to the RCIC flow controller) is failed at its MAXIMUM outpu (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 (3.00)
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signals, system valve responses, and final HPCI turbine status.
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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 standby mode) (0.50)
b. Following a mechanical overspeed condition on the HPCI turbine, the turbine trip must be reset locally prior to further HPCI operatio (0.50)
c. Following a high suppression pool level, the HPCI suction from 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 poo (0.50)
GUESTION 2.05 (3.00)
With regard to the Standby '. 49 aid Control System (SLC) During the filling of the Standby Liquid Control Storage Tank, some of the liquid everflows through the overflow lin WHERE would the fluid drain to and why? (1.00)
b. If SLC were manually injected, LIST five (5) indications
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that the operator could observe in the control room which would indicate that the SLC system was injectin (Do not include system initiation control switch or manual shutoff valve positions.) (2.00)
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List the FOUR automatic actions that thuuld occur, OTHER THAN a
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Group 1 Isolation, if the Main Steam tine Rad Monitors 'A' and 'B'
j' reach their High-High trip setpoint.-timit your answer to those actions which occur as a DIRECT RESULT of the signa (2.00)
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PAGE 8 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
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 condition SW pumps A,B operating, C,D in standb SACS loop A operating, SACS B in standby,TACS supplied by SACS loop All systems operationali no operator action 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; Off site power availabl (0.75) b. Off site power unavailabl (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 close b. All running RHR pumps tri RHR pump suction valve (F006 A/B) auto close Outboard head spray valve (F023) auto close QUESTION 2.10 (1.00)
What is the function of the static switch in the un-interruptible power supply? (1.00)
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YSTEMS PAGE 9 PLANT DESIGN
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od QUESTION 2.11 (3.00) esm3 EML P's cei'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 onl o. TCV position b. BPV position c. Power Pressur 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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PAGE 10 INSTRUMENTS AND CONTROLS
QUESTION 3.01 (2 50)
Answer the followins questions based upon the situation described calo The RRCS is fully operationa The RRCS receives a reactor High Pressure (1071 psis) signal in both complementary logics It takes 57 of a RRCS channel and remains in for 64 second seconds from the initial reactor High Pressure signal before (2.50)
the APRM levels are downscal WHICH of the four logics integrated into RRCS are actuated T=0 seconds? WHICH logics are actuated at T=10 seconds? WHICH logics are actuated at T=25 seconds? WHICH logics are actuated at T=53 seconds? HOW LONG from T=0 seconds is it before the RRCS can be reset?
QUESTION 3.02 (2.00)
Asssume the FEEDWATER LEVEL CONTROL SYSTEM is being operated in
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3-ELEMENT control usins reactor LEVEL DETECTOR CHANNEL Reactor power is at 85%, STEADY STAT 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 FAILUR .
(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 HIG (1.00) Channel 'A' REACTOR LEVEL detector signal FAILS LO (***** CATEGORY 03 CONTINUED ON NEXT PAGE xxums)
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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 except one rod in each Group ---
22-51 in Group 1, 46-55 in 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 bypasse 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 limi ROD GROUP ________ ________ ________
INSERT ERROR ________ ________ ________
INSERT ERROR ________ ________ ________
WITHDRAW ERROR ________ ________ ________
QUESTION 3.04 (1.00)
SELECT which one of the following best describes the operation /per-formance of an IRM during a reactor startu (1.00)
a. When the IRM is reading full scale on Range 10, the APRM's should
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be reading approximately 10% powe b. Shifting from Range 4, indicating 75, to Range 5, will result in an indication of 24, on Range c. Reactivity feedback, due to the moderator temperature coefficient, should begin at approximately Range d. When an IRM channel increases from 25 on range 2 to .~5 on range 3, j the indication has increased by one decad (***** CATEGORY 03 CONTINUED ON NEXT PAGE whxxx)
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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) 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 of Explain what is occuring. Include in your answer the status of the ADS logics and whether the valves are open or close (1.00) 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 AD (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 INCREASIN (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 DECREASIN (1.00)
c. An accident condition exists whereby actual reactor water level is below the lower instrument tap (variable les tap) and the reference les is flashin (1.00)
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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 tri (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 1 (0.50)
d. While in RUN, with IRM channel A bypassed, APRM channel A fails downscale. This condition should NOT cause a half scra (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 location (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 signa (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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QUESTION 3 11 (2.00)
Answer the following questions concerning the 1E 4.16KV syste c. The diesel generator is operating and its associated breaker closed in parallel with the offsite source. Briefly explain how load (0.50)
changes on the diesel are mad 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 breake . 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 is increased prior to closing the breake (2.00)
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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 minute c. Why is recire runback prior to scram? (0.50)
b. 110 F was chosen because, among other reasons, it is the ' Boron Injection Temperature.' Explain what is meant by this ter (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 (0.25)
injection into the RPV, and verify the reactor has scrammed? 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)
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? According to the same procedure, what checks must be performed followins a thermal power change exceeding 15% of rated? . (1.00)
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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 Authorizatio 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 SRM detectors be withdrawn to maintain count rate between 10ee2 and 10ee5, however, only 2 detectors may be withdrawn at any one tim What is the purpose of this limitation? (1.00)
b. During a reactor heatup/ pressurization, reactor pressure increases above the pressure setpoint AND no turbine bypass valves are ope According to the procedure, what certain precaution must be taken prior to initiatin3 corrective action and why? (1.00) '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 acheiving a steady feedwater flo (1.00)
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QUESTION 4.07 (1.25)
STATE the immediate operator actions for a failed open or stuck open (1.25)
relief valve.
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(xxxxx CATEGORY 04 CONTINUED ON NEXT PAGE xmax*)
. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 17
~~~~
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 required) (3.00)
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 Rang (1.00)
t.
,
(***** CATEGORY 04 CONTINUED OH NEXT PAGE xxxxx)
!
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- . . _..._
. -
.
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. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 18
~~~~ - ------------------------
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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Enthalpy Entropy Abs Pres pecific Volume Sa Temp Sa Sa Sa Sa Temp lb per Sa Liquid Evap Vapor Fahr Liquid Evap Vapor Lic vid Evap Vapor Iabr SqI hg s, sgg sg t h ig t p v g _vit vs if 107 .0000 2.1873 2.1973 3 .5 32 8 0 08859 0 016022 0.0041 2.1762 2.1802 3 .9 1.996 107 .1651 2.1132 3 .008 107 .2 36 0 0 10395 0 016020 2839 0 0 0122 2.1541 2.1663 3 .41 7634 2 6.018 107 .1 38 I O11749 0 016019 0.0162 2.I432 2.1594 4 .027 1071 0 107 .12163 0 016019 2445 8 0 0202 2.1325 2.1527 42.0
10 035 1069 8 107 / 227 .1459 4 .041 10681 10801 44 I - 0 14192 0 016019 108 .1111 2.1393 4 .047 10676 4 .5 0 0321 2.1006 2.1327 0 16514 0 016071 IR10 0 1810 0 i 48 9 108 .0361 2.0901 2.1262 5 .3 I 59 9 0 17796 1589 2 20 057 106 .2 0.0400 2.0798 2.1197 5 .0695 2.1134 5 .058 106 .1 0.0439 54 I O20625 0 016026 1482 4 I482 4 2.1070 56.0
24 059 106 .0593 0 22183 0 016028 1383 6 1383 6 50 0 50 0 26 060 1060 8 108 .0491 2.1006 023843 0 016031 1797 7 1292 2 58 8 10871 0.0555 2.0391 2.0946 8 .6 1207 8 28.060 10591 80 5 025611 105 .0593 2.0291 2.0885 6 .4 108 .0632 2.0192 2 0824 .058 MI 029497 98 .3 109 .0670 2.0094 2.0764 68 0 84 0 0 31626 0016043 989 0 2.0704 68 8 36.054 105 .2 0.0708 1.99 %
88 9 031889 0 016046 926 5 926 5 j
86 .052 105 .1 0.0745 1.9900 2.0645 1 fil 0 36292 0 016050 868 3 1.9804 2.0587 72.5
'
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0 38844 0 016054 814 3 12 0 764 I 42 046 105 .8 0.0821 1.9708 2.0529 1 I l 74 8 0 41550 717 4 717 4 44 043 10501 10941 0.0858 1.%I4 f.0472 7 TII 044420 0 016063 0 0895 1.9520 2.0415 7 n ul6067 671 R 613 9 46.040 104 .6 is t 0 41461 633 3 48.037 104 .4 0 0932 1.9426 2.0959 8 .0 50 033 10473 109 .9334 2.0303 82 5 0 54093 0 016077 595 5 595 5 .029 104 .1006 1.9242 2.0248 M9 057102 0 016082 560 3 560 3 8 .1043 1.9151 2.0193 0 61518 0 016087 227 5 527 5 88 0 0 016093 4968 4%8 56 072 1043 9 109 .1079 1.9060 2.0139 8 .7 1100 8 0 1115 1.8970 2.0086 90 0 90 0 0 69813 0.1152 1.8881 2.0033 9 .8792 1.9980 MO 0 016111 416 3 4163 62 010 1040 5 1102 5 MI O19062 64 006 1039 3 1103 3 0 1724 1.8704 1.9928 9 > 98 I O84072 0 016117 392 8 392 9 310 9 370 9 66 003 10382 1104 2 0 1260 1.8617 1.9876 90 0 0 016173 L 28 0 0 89156
_ _ _ - _ _ _ _ _ _ _ _ _ _ _ _
.
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Specific volume Enthalpy [ntropy Y
Ahs Pres Sa Sa Se Sa Temp Temp lb per Sal Sa liquiil Vapor li uld fysp Vapor liquid Evap Vapor Fahr Iah Sq in Ivan t vg i h ig h8 si sig sg
-
I p v, veg 350 4 350 4 67.999 103 .1295 1.8530 1.9825 18 .1331 1.8444 1.9775 18 .9 lef t 100789 3131 313 1 71.992 1034 8 1106 8 0.1366 1.8358 1.9725 1RO 184 8 1 06 % 5 0 016144 0.1402 18273 1.9675 10 .6 106 8 11347 103 .5 0.1437 1 8188 1.9626 10 R0 28 780 30 7598 196 8 12030 265 39 77.98 103 .3 01472 1 8105 1.9577 120 110 0 1.2750 0 016165 26537 01501 18021 1.9528 11 .2 1110 2 112 8 13505 0 016173 0.1542 13938 1.9480 11 .97 10291 1111.0 1 114 I l4299 111 .9433 11 j 0 016188 225 84 225 85 83 97 1027 9 15133 13774 1.9386
-
IlsI 102 .1611 11 .97 113 e 16009 21470 / 214 21 1025 6 l11 .1646 13693 1.9339 1 .97 128 9 18927 192 95 89.96 1024 5 111 .1680 13613 1.9293 12 i 1 7891 0 016213 192 94 1.1533 1.9247 12 '
1228 18324 91 % 102 .3 01715 124 8 18901 0 016221 18323 0.1749 11453 1.9202 IN.9
,
114 08 114 09 93 96 102 .1 126 I I9959 0 016229 0 1783 13374 1.9157 I .0 111 '
128 8 2.1068 0016738 15733 97.96 101 .8 01817 11295 1.9112 13 O 14966 99 95 1018 7 1118 6 0.1851 13217 1.9068 13 .3445 0 016256 14964 13140 1.9024 Int 14241 101.95 101 .5 0.1884 1348 2 4111 0 016265 14240 13063 I.8980 1ES 135 57 103 95 1016 4 1120 3 0.1918
'
13t I 2 6047 0 016214 135 55 17909 129 11 105 95 10152 112 .1951 1 6986 1.8937 13 R4 1014 0 112 .1985 1.6910 1.8895 1488 28092 0 016293 122 98 123 00 107.95 149 8 112 .6534 1.8852 14 tl 0 016303 !!12I 11722 109.95 101 .6759 1.0810 14 I 3 1997 0 016312 Ill14 111 16 Ill 95 10113 1123 6 14 .5 112 .6684 1.0769 140 8 3 3653 0 016322 0.2117 1.6610 1.8727 148 8 0016317 101 BR 101 10 115 95 10092 112 e e 3 MRI 9705 9707 117.95 1008 2 112 .2150 1.6536 1.8686 19 .6463 .l.8646 15 .95 100 .9 0.2183 1528 3 9065 0 016353 02216 1.6390 1.8606 1540 0 016363 8850 88 52 121 95 100 .7 1548 4 1025 0 2248 1.6318 1.8566 158 9 0 016374 84 56 8457 123 95 1004 6 1128 6 156 I 4 3068 0 2281 1.6245 1.8526 15 R4 R0 R7 R083 125 96 100 .4 151 8 4 5197 113 .2313 1.6174 1.8487 10 .2 10e 8 0 016406 13 90 13 92 129 96 100 .0 0.2345 16103 18448 18 .6032 1 8409 1Kt 0 016417 70 70 70 72 131.96 999 8 113 %I 1.8371 100 O 6767 6768 133 97 998 6 113 les I 54623 0 016428 IKO 0 016440 6418 6480 13597 99 .2441 1.5892 I.8333 I88 I 5 1723 137.97 99 .2 0.2473 1.5822 1.8295 1M8 179 8 5 9926 0 016451 62 04 62 06 00lR463 59 43 5945 139 98 995 0 113 .5753 1.8258 172.0 l 172 8 6 2736 1.5684 1.8221 174.8
'
0 016414 5695 56 97 141.98 9938 113 .5616 1.8184 17 .6 1136 6 0.2568 litt 6 8690 5735 57 36 145 99 99 .4 0.2600 1.5548 1.8147 178.8
- III 8 7.1840 0 016498
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[nthalpy Entropy Abs Press Speofic Volume Sat Temp Sal Sa Sa Sa Temp lb pei Sal Liquid Vapor Fahr vapor Liquid Evap vapor Evan Iahr SqI Iiqmrt Ivan I I p vi veg va he h ig hg si sig st 5022 148 00 990 2 1138 2 02631 1.5480 1.8111 lett let s 75110 0 016510 5021 0.2662 1.5413 18075 102 0 48 112 18 189 150 01 989 0 1839 0 132 3 7850 0 016522 0.2694 1.5346 I8040 184 8 0 016534 46 232 46 249 152 01 9878 1139 8 104 8 8203 154 02 986 5 1140 5 0 2725 15219 1.8004 let#
0 016547 44 383 44 400 ING Its t 8 568 47 671 42 638 156 03 985 3 114 .7%9 Ige s 8 947 0 016559 02787 1.5148 1.7934 19 / 40 957 158 04 98 .1 198 8 9 340 0 016512 0 2818 1.5082 1.7900 19 .8 1142 9 192 9 9 147 0 016585 012848 1.5017 1.7865 19 .2879 1.4952 1.7831 19 .4 1144 4 198 9 10 605 0.2910 1.4888 1.7798 194 0 34 954 34 970 166 08 979 1 1145 2 IIII 11058 0 016624 02940 1.4874 1 7764 29 .9 1146 0 2$4 8 11 526 0 01R637 0 3001 1.4697 17698 294 0 31135 31151 172 11 975 4 1147 5 294 8 12 512 0 016664 0 3061 1.4511 1.7632 200 0 0 016691 28 862 28 878 176 14 9778 1149 0 298 8 13 568 1150 5 0.3121 1.4447 11568 212 0 0016719 26 182 26 199 18017 970 3 212 8 14 696 0.3181 14323 11505 21 R18 74894 184 20 9618 115 .4 0324I I4201 11442' 220 0 titI 17 186 0016715 23 131 11380 224 e 0 016805 21 529 il 545 19227 962f 115 .3300 1.4081 224I 18 556 190 31 960 0 115 %I 11320 220 8 20 015 0 016834 20 056 20 013 23 .4 115 I 21 567 0 016864 18101 18718 204 40 954 8 1159 2 0 3476 13725 11201 235 0 236 e 23 216 0 015895 11 454 17 All 208 45 95 .0 24 968 0 018928 18 304 18 321 240 0 949 5 116 .3491 1 7085 24 I 26 876 21656 946 8 116 I.3379A 17028 248 0 28 196 0 016990 14 264 14 281 240 0 94 % 17 570 12 538 224 69 94 I 03163 1.3154 I6917 256 8 258 8 31 091 228 76 938 6 116 .30(3 16862 26 g 35 427 0 017089 11 145 11 762 935 9 18687 0.3876 12933 16808 26 .2823 1 6755 260 0 40 500 0 017157 10 358 10 375 236 91 9331 1170 0 200 0 0 3987 1.2115 1 6702 27 I2607 16650 276 8 46 147 0 017228 9 167 9 180 24508 92 l72 5 218 8 924 6 1173 8 0 4098 12501 16599 20 .17 200 0 0 4154 12395 I6548 284 8 52 414 0 01130 8 1280 8 1453 253 3 9211 1875 0 284 8 0 4208 I2290 1.6498 288 8 200 9 55 795 0 01134 7 6634 76807 25 l762 915 9 117 .2186 1 6449 29 .5 0 01/41 6 8759 6 8433 265 6 9130 1178 6 0 4317 12082 I6400 2968 298 8 63 064-
. _ _
.
.
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~Sper.ilic Volum, Enthalpy Entropy Abs Pres $s Sa Temp Sal Sa Sa SR Temp Lb per Vapor Fatw liquid Evap Vapor Liquid Evap Vapor Liquid Eysp Fahr Sgi sa I I p .
V i vig vs hr hs i hg sg seg 6 4658 2693 910 0 11793 0 4372 1.1979 1.6351 389 9 300 0 67 005 0 01745 6 4483 6 1830 273 8 90 .1877 1.6303 304 0 30 .0 0 4419 1.1776 16256 300 e 300 e 75 433 54566 5 4742 2821 901 0 11831 0 4533 1.1676 1.6209 312.0 i II2 8 79 953 0 01757 l 84 688 0 01761 51613 5 1849 286 3 89 .1 04586 1.1576 1.616 316 0 318 8 09 643 0 01766 4 8961 4 9138 290 4 894 8 1185 2 04640 1.1477 1.6116 32 NG 0 4692 1.1378 1.6071 32 .6 1886 2 3245 0.4745 1.1280 1.6025 320 0 320 g 100 245 001774 4 4030 j 4 4208 298 7 888 5 118 .1183 1.5981 33 %6 302 9
-
11 .0 3 5834 3 6013 31 .0 04954 1.0894 15849 34 .2 119 .0199 1.5006 340 O
! MeI 131.143 0 01797 868 9 l1923 0.5058 1.0705 1.5763 352.0 i 352 0 130130 0 01801 3 2423 3 2603 323 9 3 1044 328I 865 5 1193 6 0 5110 10611 1.5121 356.9 j 350 0 145 424 0 0lR06 3 OR63 0 01811 2 9392 29573 332 3 86 .0517 1.5678 30 .8 151 610 0 01816 2 8002 28184 336 5 858 6 1195 2 0 5212 1.0424 1.5637 3 .0332 1.5595 360 0 3000 169113 0 01826 2 5451 2 5633 345 0 8516 11 %.1 0.5314 1.0240 1.5554 37 .4 0 5365 1.0148 1.5513 37 l98 0 0.5416 1.0057 1.5473 300 0 300 8 195 729 3M S 205294 0 01842 22120 2 2304 35 .5466 0.9966 1.5432 3 .2 S3 .5392 380 B 300 0
,
38 M 20369 36 .4 119 .9786 1.5352 39 % 1,5313 396 8 i
398 9 236193 001R58 1 929?
408 8 247 259 0 01964 I8444 I8630 37 .0 0.5667 0.9607 1.5274 400 9 001870 11640 11827 379 4 822 0 120 .5234 4 ,
404 9 258725 400 0 270 600 0 01875 1 6811 1 7064 383 8 818 2 120 .5195 480 0 412 9 282 894 0 01881 16152 16340 38 .2 120 .5157 41 <
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elle 295 617 00lR87 1 5461 1 5651 392 5 810 2 120 .5866 0 9253 1.5118 elle .
l 429 0 300180 001894 14808 14997 396 9 806 2 120 .5915 09165 1.5000 420 0 424 I 322 391 0 01900 14184 14314 4013 802 2 120 .9017 1.5042 47 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 78 .10212 112152 428 0 176 7 12043 0 6259 0 8557 1.4815 448 0
<* * * O 43114 0 01941 105164 101111 '1 .8411 1.4778 (.de
_ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ __
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.
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Specifer Volume Enthalpy Entropy 1 Ahs Press Temp Sal Sat Sa Sa Sa Sa Temp lb per Sq in l iquirl Ivan Vapor liquid Evap Vapor liquid Evap Vapor Fahr Fahr I vg hg h lg hg 5: sig sg I p vg v ig ,
44 .2 1204 8 06405 0 8299 1.4704 40 st t 466 87 0 01961 0 97463 0 99424 44 .4667 46 M8 0 95557 754 0 120 .4629 468 0 est t 504 83 0 01976 0A9885 0 91862 4501 455 2 749 3 120 .6551 0 8042 1.4592 47 R4 0 86145 DM8329 0R4950 45 .4555 476 9 als e 54511 00l997 0 R/%R 46 .1 0 6648 03871 1 4518 40 / 0 81717 0 66 % 03785 1.4481 48 .5 0.6745 03700 1.4444 48 I
'
610 10 007011 0 13641 0 75658 418 5 724 6 120 .4407 49 t2 8 633 03 0 07026 0 10191 012820 0 6R065 070100 AR3 2 719 5 12023 0 6842 03528 14370 49 .9 714 3 120 .6890 03443 1.4333 50 % 50 .5 7037 120 .4258 50 .3 698 2 120 .4221 51 .4183 51 A4 76 0 07081 0 % 997 0 58079 50 II9 .0 6870 1199 0 01133 03013 1.4146 52 I 52 !814 0 53916 516 9 68 .2 03182 0 6926 1.4108 524 8 528 9 870 3I 0 02112 0 49843 0 51955 5218 675 5 119 .6839 1.4070 520 8 53 .4032 532 8 900 34 0 02123 53E O 931.17 0 07134 0 46173 0 48757 5313 663 6 119 .3993 53 .5 119 I.3954 54 .8 995 22 0 02157 0 47677 0 44834 54 .1 03427 0.6489 1.3915 544.9 i
540 8 1028 49 0 02169 0 41048 0 43217 546 9 6450 119 .6400 3.3876 54 .5 1190 6 0 7525 0.631I l.3837 55 .2 63 .3797 55 .7 03625 0.6132 1.3757 508 9
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0 35099 0 31320 56 .1 03674 0 6041 1.3716 5640 544 8 1170 10 0 02221
,
0 33741 0 35975 5729 611 5 1184 5 03725 0.5950 1.3675 548 8 l Ste t 120772 0 02235 0 34678 578 3 604 5 11821 0 1775 0 5859 1.3634 57 .2 118 .3592 57 .1 58 .5673 1.3550 58 .17 584 9 1367 7 0 02295 0 28753 0 31048 594 6 SR .3507 56 .8 1410 0 0 02311 0 27608 029919 6001 5743 1174 8 03978 0.5485 1.3464 50 .6 0 8030 0.5390 1.3420 59 e 59 " 0 07345 0 75425 0 27770 61 .8 187 .5793 1.3315 595 0 598 9 14978 <
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Ahs Pres Specific Volume . Enthalpy Entropy Temp tb pe: Sa Sa So Sa Sa Sa Temp Fah Sq in iIquid Fvap Vapor Li vid Evep Yapor liquid Evap Vapor Fahr I p V v ig_,, _j 8 f h is hg s, sg, s, g J 55 .1 550 6 116 .5196 1.3330 1543 2 0 02364 024384 0 28747 88 OOO O 62 .2 116 .8187 0.5097 1.3284 884 I 15897 0 02382 0 23374 0 25757 60 .4 0 8240 0.4997 1.3238 16373 0 02402 0 22194 0 24796 898 8 0 23865 634 8 5241 115 .4896 1.3190 81 .8348 0.4794 1.3141 81 .6 115 .4689 1.3092 S Sit t , 1786 9 0 02466 0 1961 ! 0 22081 64 .3 115 .8458 04583 1.3041 82 .1 4? .8 82 .4474 1.2988 82 R0 0 20394 65 .1 828 8 0.8571 0.4364 1.2934 83 .9 476 4 114 SE28 0 4251 1.2879 63 ,
'
2007 8 0 07566 016726 018792 67 .1 538 8 0 02595 015427 019021 67 .6 11333 0 8686 0.4134 1.2821 88 .4015 1.2761 84 .1 112 .8806 0.3893 1.2699 54 .1 1124 0 548 8 2178 1 0 8868 03767 1.2634 55 .3637 1.2567 85 R 012187 0 15115 70 .1 ISe 9 23011 0 14431 71 .1 110 .8995 0.3502 1.2498 90 .3361 1.2425 06 ,1 0 10947 013157 72 .6 0.9064 964 8 0 02811 te .5 36 .5 0.9137 0 3210 1.2347 568 8 2498I 0.9212 0.3054 1.2266 57 .2892 1.2179 87 .5 107 .5 31 .5 0 9365 0.2720 1.2006 0I SM $ 2708 8 0 03037 0 08080 0 11117 88 .4 0 9447 0.2537', 1.1984 Sle t 27821 0 9535 0.2337 1.1872 50 .2 104 .9634 0.2110 1.1744 80 .1 103 i 89 .1841 1.1591 O .8 101 '
59 .7 99 .9901 0.1490 1.1390 M .0006 0.1246 1.1252 78 .2 0 04100 0 02192 0 06300 85 .0 95 .0169 0 0816 1.1046 79 FM I 1.0329 0.0527 1.0856 7e .4 93 .0612 0 0000 1.0612 705.47'
70547*
.
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- f'ritical lemneralute 3
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l Table 2: Saturated Steam: Pressure Table
. . . . - - . - _
. . . Specific Volume Enthalpy Entropy Abs Pres Temp Sal Sa Sa Sa Sa Sa . Abs Pres th/Sg i Fahr liquid Ivan Vapor liquid Evap Vapor liquid Evap Vapor tb/Sg in.
I p t vi v,i vg hg hgg hg s, s gg sg p
, 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 106 .4 0 0542 2.0425 2.0967 8.25 5 59 79 586 0 016011 641 5 641.5 47.623 104 % 3 0 0925 1.9446 2.0370 0.50 18 10114 0 016136 331 59 333 60 6913 1036I 110 .1326 1.8455 1.9781 .1 0 2349 1.6094 1.8443 50 ft8 193 21 0016592 38404 38 420 161.26 98 .3 0.2836 1.5043 1.1879 1 /26 199 180.17 970 3 1150 5 0.3121 1.4447 1.7568 14.698 1 15 e , 213 03 0016126 76214 26 290 181 21 %97 115 .3137 1.4415 1.7552 1 i 29 8 , 22796 0 016834 20 010 20 087 196 27 9601 115 .3962 1.7320 2 ; 30 0 250 34 0 011009 13 1266 131436 21 .1 0.3682 1.3313 1.6995 3 e e 267 25 0 017151 10 4194 10 4965 236I 933 6 116 .3921 1.2844 1.6765 4 .1 0.4112 1.2474 1.6586 5 *
- SG I 29211 0 017383 71562 7.1736 26 .6 0.4273 1.2167 1.6440 50 0
'
78 3 302 93 0 017482 6 1875 6 2050 27 .8 1180 6 04411 1.1905 1.6316 7 I 312 04 0 011573 5 45.16 5 4711 28 .9 118 .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 29 .2 0.4743 1.1284 1.6027 18 .8 334 19 0 01782 4 0306 4 0484 305 8 88 .9 0 4834 1.1115 1.5950 11 I 341 27 0 01763 3 1091 3 1275 31 .8 1190 4 0.4919 1.0960 1.5819 120 0 130 0 347.33 0 017 % 3 4364 3 4544 319 0 872 8 119 .4998 1.0815 1.5813 13e e leg g 353 04 0 01803 3 D10 3 2190 325 0 868 0 119 .0681 1.5752 14 .1 0 5141 1.0554 1.5695 15 st 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 34 .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 10 fee s 37753 0 0lR13 2 3847 2 4030 350 9 846 7 119 .0113 1.5498 190 8 29 .3 0 5438 10016 1.5454 20 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 27 I 393 10 0 01855 I91991 199846 3683 83 .1 0 5588 0 9748 1.5336 239 9 248 8 39739 00l860 189909 191169 37 .5634 0.9665 1.5299 240 0 250 0 40097 0 01865 182452 II4317 376I 825 0 120 feeI 404 44 0 01810 115548 117418 379 9 821 6 120 .5230 26 rs t 40780 0 01815 169131 171013 383 6 818 3 120 > 200 0 411 01 0 01880 161169 I65049 38 .3 0.5805 0 9361 1.5166 200 8
& 200 3 414 25 0n18R5- 157597 159482 390 6 812 0 120 .5135 290 0 300 0 417 35 001889 I52384 154274 394 0 808.9 120 .5105 300 0 356 8 431.73 001912 130642 1.32554 409 8 794 2 120 .4%8 350 0 age s 444 60 0n1934 1.14167 1.16095 42 .6 0 6217 0 8630 1.4847 400 5
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Enthalpy Entropy Specific Volume Sa Abs Pres Sa Sa Sa Sa Abs Pres Temp Sal Lb/Sg in. -
ivan Vapot Lit uid Evap Vapor liquid Evap Vapor .
? IblSq I Fahr Iirluiel y, it hg g h g sg s g, s, p g p i v, v ,,
1204 8 06360 0.8378 1.4738 450 0 001954 101224 I03119 43 O 456 28- 7551 12043 0 6490 0 8148 1.4639 Ses e 0 01915 0 90181 0 92162 44 .3 120 seg g 550 0 12033 0 6723 03738 1 4461 48620 0 02013 0 14962 0 16915 4713 732 0 14381 65 .8 0 6828 03552 494 89 0 07032 0 68811 1.4304 fee g 850 0 065556 491 6 110 2 120 .4232 738 g 058880 0 60949 50 .4163 300 g 0 54809 0 568 % 509 8 689 6 1l99 4 seet 518 21 0 01081 03197 0 6899 1.4096 35e g 0 51191 0 53302 51 ISO I 525 24 0 02105 0.6753 1.4032 900 e 050091 526 7 669 7 11 % 4 03219 900 e 53195 00?!?3 0 41968 11943 03358 06612 1.3970 950 0 538 39 00?!41 0 45064 0 41205 534 7 660 0 Iges e 950 0 650 4 119 .3910 Iges B 544 58 0 02159 0 42436 0/45 % 542 6 119 .3851 1950 e 1959 0 550 53 0 02111 0 40041 042224 550I 640 9 1.3794 1100 0 0 40058 55 .5 118 .6216 1100 I t 556 28 0 02195 0 31863 13738 115 .0 03647 06091
,
1150 0 561 82 0 C"l4 0 35859 0 38013 1184 8 03714 05%9 1.3683 1290 8
, .
567.19 0 0h.12 0 34013 0 36245 51 .6 01790 0.5850 1.3630 f2508 572 38 0 02250 0 32306 034556 578 8 1798 9 594 6 1180 2 03843 0 5733 1.3577 130st 577.42 0 02269 0 30722 0 32991 585 6 1300 9 585 4 1177 8 03906 0 5620 1.3525 13540 13549 582 32 0 02288 0 29250 0 31537 59 ,
117 .5507 1.3474 1490 8 14M 8 58707 0 02301 0 21811 0 30118 5988 516 5 l
605 3 5674 117 j 14500 59170 0 02327 0 26584 0 28911 187 *05288 1.3373 1590 0 1500 I 596 20 0 02346 0 25372 0 21/19 6113 558 4 155 .4 0 8142 0 5182 1.3324 600 59 0 0?366 0 24235 0 26601 618 0 15500 116 .8199 0 5076 1.3274 160 ) 1600 0 53 .6 0 8254 0 4911 13225 1650 9
'
609 05 0 02401 0 22143 0 24551 630 4 1958 0 52 .3176 17000 tree s 613 13 0 07478 021178 023607 636.5 l 1.3128 1750 0 022713 64 .1 115 .4765
' 1750 0 $1712 0 02450 0 20283 1.3079 1000 e 0 21861 648 5 503 8 l15 .3030 1850 t 021052 654 5 494 6 1149 0 0 8470 0.4561 18580 424 83 0 02495 0 18558 1.2981 1990 e 0 02517 0 11161 0 20218 66 .6 08522 0 4459 1999 8 62856 114 A 1.2931 19580 itse e 632 22 0 02541 0 16999 0 19540 666 3 415 8 466 2 1138 3 0 8625 0 4256 12881 2000 9 2000 0 635 80 0 07565 0 16266 018831 67 e e 2l00 8 642 16 0 02615 0 14885 1.2675 2290 8
,
0 07669 0 13603 0 16212 695 5 426 7 112 e 649 45 !.25a 23009 0 0/121 0 12406 0 15133 707 2 406 0 1113 2 0 8929 0 3640 230s I 65589 1.2460 24000 feet S 66711 0 0/190 0ll?R1 0 14016 719 0 38 I031 0 9031 0 3430 i
013068 7313 361 6 1093 3 0 9139 0 3206 1.2345 2500 0 j 2900 9 66811 002959 010209 2500 8 012110 744 5 3376 108 .2741 1.2097 2700 9 2100 3 619 53 0 03029 0 08165
0 10305 710 7 2851 1055 8 0 9468 0 2491 1.1958 200 e 684 % 0 03134 0 01111 785 1 2547 1039 8 0 9588 0.2215 1.1803 2900 0 2900 0 690 22 0 03262 0 06150 0 09420 218 4 102 .1619 3000 0 3000 I 695 33 0 03428 0 05013 0 08500 80 .3 0 9914 01460 1.1373 318 .1 93 .0351 0 0482 1.0832 3200 0
) 3200 8 705 08 0 04412 0 01191 005018 906 0 00 9060 106I2 0 0000 1.0612 3200.2*
l 3700 2' 70541 0Os018 0 00000 l I , f%
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MASTER PAGE 19 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,
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-86/02/24-CRESCENZO, ANSWERS -- HOPE CREEK ANSWER 1.01 (2.50)
o. 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 constan (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 ccste has become undermoderate Because of this effect, the void coefficies.t becomes more negati ' ~ l (1.50)
~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 (1.00)
samarium thus makin3 it non-dependent on flux / power leve b. The greatest changes in samarium concentrationc occur durin initial startup. 4i. EUL, 1". ::, **1"deat=*m pg Q ,
(, ,,,.9 (1.00)
REFERENCE NMPC Operations Technology Vol. I, chapter 15 FJC 16 Hope Creek LP RXHP32-01 pg. 7,10 learning obj 2,3, ANSWER 1.03 (2 00)
(0.5)
a. Indicated is highe (0.5)
7-10 inches higher b. Indicated level is sensed outside the dryer skir (0.5)
Steam flow through the steam separator / dryer at 100% causes (0.5)
a backpressure of 7-10 inches H2 __
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. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 20
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ANSWERS -- HOPE CREEK -86/02/24-CRESCENZOr REFERENCE Hope Creek Vessel Instrumention LP ps. 45, learn. obj 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. 42 FJC184 ANSWER 1.05 (3.00) 8,Y4 Cert e o y Ib .
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 fissio (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 insertion (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 21 isEis55VsAsiCi- sEAi isEnsFEE Es5 FE5i5 FE5s
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ANSWERS -- HOPE CREEK -86/02/24-CRESCENZO, ANSWER 1.06 (3.00) cudddes w e r e- ym es do RW=(fluxmm2)SL/px*2, S= control rod span, L= thermal diffusion lengt c. From 0-A, rod worth increases due to the neutron flux and thermal diffusion length increasin (1.00)
b. From A-B flux is relatively constant but thermal diffusion lenSth is still increassing due to the heatu (1.00)
c. From point B-C, the increase in flux and TDL are at ,g slower rate than the incrqpse in pitch so rod worth decre se id*t (1.00)
g y s,' .",)'
e rlur w i lt vse b' -f KAL ( ' #'uvs Ak E T
eme REFERENCE 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 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 n cavitation would not begin until the level drops below 5 f (1.00) 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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ANSWERS -- HOPE CREEK -86/02/24-CRESCENZO, 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 presence of superheated stea (1.00)
REFERENCE Hope Creek M.O.C.D. LP 104 pg. 15 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 increas 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 averag '
REFERENCE Hope Creek LP 17 pgs. 10,11 , FJC270
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. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 23
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ANSWERS -- HOPE CREEK -86/02/24-CRESCENZO, 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. FJC288 e
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. PLANT DESIGN INCLUDING SAFETY AND ENERGENCY SYSTEMS PAGE 24
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ANSWERS -- HOPE CREEK -86/02/24-CRESCENZO, 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 C0.75 (1.0) Will inject E0.25 Pump overheating and seal damase may result (1.0)
durinS low or no flow conditions CO.75 Will not inject CO.253. Maximum signal from the flow element will result in the flow controller keeping the turbine speed at minimum [0 75 (1.0)
REFERENCE LP 30-01, REACTOR CORE ISOLATION COOLING, (RCIC) pgs. 25,40,fis 3, I.O. 4,5 i
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. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 25
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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 clos (1.00)
2. Turbine Trip due to isolation (1.00)
3. Vacuum breaker IV's close (due to low team line prgssure and hi DW) p/A 9 3suresa"$b up num ,F tsel. va lve s (1.00)
REFERENCE HOPE CREEK LP16 pgs. 20,70,71, I.O.47 FJC259 ANSWER 2.04 (1.50)
a. true (0.50)'
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 intr of boron into any system which can communicate with the (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)
REFERENCE 0' b# C' ,11 Hope Creek LP 23-01 pgs. 10,41 I.O. 42 FJC263
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PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 26 _______________________________________________________
ANSWERS -- HOPE CREEK -86/02/24-CRESCENZO, ANSWER 2.06 (2.00)
1. Reactor water sample valves close 2. Mechanical Vacuum pumps trip if runnin3 3. Reactor scram 4. c'- " - - ' ;teee i sc1 tion- F// ne (M6 i.e 3 kmf (4 neededi 0.5 each)
Jen(sfr[a%
REFERENCE LP 45, ps. 20 LP 53, P3 15 LP 22, pg. 61 ANSWER 2.07 (3.00) Demin makeup to SACS on lo level in SACS A surge tank.(0.25)
2. 3X1o level isolates TACS from SACS loop (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. Hope Creek LP 80, pss. 34 I.D. 4 ANSWER 2.08 (1.50) A and B pumps start immediately. C and D start after a 5 second (0.75)
time dela b. All pumps will start upon closure of the SDG breaker and bus (0.75)
undervoltage is cleare REFERENCE LP 302HC-000.00-028-02
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- PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 27
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ANSWERS -- HOPE CREEK -86/02/24-CRESCENZO, ANSWER 2.09 (2.00)
. not WNi ' SDC ' *
'b '"'(
c. Will not d. Will (&reS each)
REFERENCE ',-
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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 indicate ( FreSS REFERENCE LP 66 1E AC power supply a
ANSWER 2 11 (3.00) t ved a llT gD ' O x ek c' O o. TCV remain at 100% due to load limit %g d, p( , g( (.50)
b. BPV open 5% due to max combined flow (.50)
c. Power decreases due to lower pressure 'nsk k h *t h '2 t i (.50)
d. Pressure decreases due to BPV (.50)
FINAL o. TCV at 100% pocition (n25) BPV shut ( '. 2 5 )
c. power lower 1(.25)
d. pressure sliS htly lower (.25)
REFERENCE Hope Creek LP I.O. 4 3,4,10
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4 . PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 28
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ANSWERS -- HOPE CREEK -86/02/24-CRESCENZOr ANSWER 2.12 (1.00)
A check valve installed around one valve to prevent a valve failure inhibiting the desired scram. OR another power sourc (1.00)
REFERENCE Hope Creek LP 22-02 pg. 13 I.O. 43
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ANSWERS -- HOPE CREEK -86/02/24-CRESCENZO, ANSWER 3.01 (2.50) Alternate Rod Insertion and Recirculation Pump Trip (RPT) None Feedwater Runback Standby Liq'id u' Control minutes, 53 second 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) , Causes reactor level to'0ECREASE due to the Level Control System having a STEAM FLOW / FEED FLOW ERROR, STEAM FLOW < FEED FLOW CO.53 FEEDS resultinginaSIGNALtoDECREA(EtheSPEDOfTHEREACT@s c c ** eTo i nd V - Yf*
4 1 vom F W p u n p \ 1- l\'"'
PUMPS CO.5 *m N ^
+ o v- FWhe 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 9:::ter 1;v;l shavid RCMAIH CGN3 TANT '-ceva: the 'C" Cc> rner E^.53 -111 uGCr-UF EG.5 u FEED runr g/g REFERENCE BF LP 12, pp. 16-1 HCGS, Rx Water Lvl Cntr1, 302HC-000.00-059-01, pg 10,-11, 8 14 l
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ANSWERS -- HOPE CREEK -86/02/24-CRESCENZO, F.
ANSWER 3.03 (3.00)
(a) (b) (c)
ROD GROUP 03 03 04 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 REFERENCE q , g., Jge h %, limi h 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. FJC214
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. INSTRUMENTS AND CONTROLS PAGE 31
ANSWERS -- HOPE CREEK -86/02/24-CRESCENZO, ANSWER 3.05 (3.00) 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 Shutdown range O to +400 inches Calibrated cold Not used for trip functions Fuel Zone-150 to + 50 Calibrated cold, no flo ,
Not used for trip functions g4g g p* g,h Q 3 Deet :::ident ":nitv. i. 3 .ense , , . h 6t r<s b y
N unique nse, vi.li s es.sisting 1;c 1 4 "+ e "= ant " t i er a-Teaperatu m sv-yvnsetud Tui occident c .ditic^ . m De
_un + e 4 r, re tiener (0.16 each)
REFERENCE Hope Creek LP 002-01, pgs. 23-26 I.O. 46 FJC262 ANSWER 3.06 (3.00) (ouCCLE 80003'
c. C;iF S d '. c ' M ;- " * ^ - " k o" C r E' "D (0.50)
b. All conditions for ADS initiation are not except the 105 see time (ADS logics F,H sealed ini bed not timed out) therefore the valves are close (1 00)
c. Rx press. < 50 psisi LP ECCS pumps stoppedi 1:c;1 ::ne--d foit Timer or logic manually rese h/ 4 j. . , in (1.00)
d. FALSE vh u[ he -t rue ,(v4 h ge,w l *
o, (0.50)
REFERENCE Hope Creek LP 29 pgs. 14, 15, 25, figs. 6,7 I.O. 8 FJC265
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. .. . INSTRUMENTS AND CONTROLS PAGE 32
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ANSWERS -- HOPE CREEK -86/02/24-CRESCENZO, ANSWER 3.07 (3.00) Indicated level will increase as variable les pressure increases. (1.00)
b. Indicated level will not change as the variable les pressure will remain constan (1.00) Indicated level will increase as reference les and variable les (1.00)
pressures equaliz REFERENCE Hnpe Creek LP 02 pgs. 46-48, FJC266 ANSWER 3.08 (2.00)
c. TRUE b . F*emP T R ' E TRUE TRUE (0.5 each)
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)
y V"' k w - l y ' * 5 bytakigsbgr,
'"*- #'
to emer. tkovr.(0.33) raw 4 x C 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. .e, 6 FJC269 ANSWER 3.10 (1.50)
"F" at be d, e
'I' at a, c,N(. (0.25 each)
h de nk sthse 3 r u n n [g ' wu d sem L J ,,L A < - & 4-e l
f I
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. INSTRUMENTS AND CONTROLS PAGE 33
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ANSWERS -- HOPE CREEK -86/02/24-CRESCENZO, 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 loa (0.50)
b. 1NW" The sync scope should be operating slow in the clockwise (counter clockwise) direction. Close breaker at 5 till 1 (0.50)
The 1E bus should be running faster with slightly higher voltage than the normal feed bu (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. FJC285
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. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 34
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ANSWERS -- HOPE CREEK -86/02/24-CRESCENZO, ANSWER 4.01 (2.00)
c. To minimize the transien (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 ATWS, i.e.(assuresshutdownpriortoemergencydepressurization) (1.50)
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 (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 rate At no time shall power exceed 102% rate (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)
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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)
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. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 35
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ANSWERS -- HOPE CREEK -86/02/24-CRESCENZO, 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 perio (1.00)
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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. FJC277 ANSWER 4.07 (1.25)
1. Attempt to reseat the valve via manual switc (0.5) Reduce recire flow and SCRAM when suppression pool temperature exceeds 110 (0.5) or if stuck open 2 minute (0.25)
REFERENCE Hope Creek OP-AB.ZZ-12 FJC278 ANSWER 4.08 (3.00)
1. To minimize energy addition to the containment (1.00)
2. To minimize rad. releast from the primary containmen (1.00)
3. To maximize injection flow from motor driven pumps (1.00)
So AcW f y bree di tw-h elb g &y ~~ $NPm4 t c[onupIs l
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. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 36
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ANSWERS -- HOPE CREEK -86/02/24-CRESCENZO, 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 scra (0.50)
REFERENCE
!! ope Creek OP-SO.SB-001 caution .3.4, 5. FJC281 ANSWER 4.10 (3.00) CORE SUBMERGENCE accomplished by maintaining a level above TA (1.0) SPRAY COOLING accomplished by using at least one CS system operating at design condition (1.0) STEAM COOLING accomplished by using SRV to create steam updraft from water in core reS i on or .9 (1.0)
70e ps,ventory ig in lower od one s ev open cun on.% 7 plenum.dyp.hmuff u e i4x4h eve emig 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 assure REFERENCE OP-SO.BC-001(0) RHR system procedure, Precaution 3.11.1
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. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 37
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ANSWERS -- HOPE CREEK -86/02/24-CRESCENZO, 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 1 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
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25.00 02.01 1.00 FJC0000197 02.02 3.00 FJC0000257
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02.03 3.00 FJC0000239 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 i 03.08 2.00 FJC0000268 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 1 04.03 2.25 FJC0000174 l
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04.04 2.00 FJC0000175 04.05 1.25 FJC0000179 04.06 3.00 FJC0000277 l
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TEST CROSS REFERENCE PAGE 2 GUESTION VALUE REFERENCE
________ ______ __________
04.07 1.25 FJC0000278 04.08 3.00 FJC0000280 04.09 1.00 FJC0000281 04.10 3.00 FJC0000282
<- 04.11 1.00 FJC0000289
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04.12 3.00 FJC0000290
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25.00
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100.00
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U. S. 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 onl Staple question sheet on top of the answer sheet 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 start % OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY
-------- -----------------------------------
53rar-- 23 t-
_ D I__ _ M ___________ ________ THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS
'5.00 96,44
_'_______ ______ ___________ ________ PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 202^^
25.00
________ _'_"_;; PROCEDURES - NORMAL, ABNORMAL,
________ ___________
EMERGENCY AND RADIOLOGICAL CONTROL 21 0 2 _ff*II___ff!f ___________ ________ 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 ai IEPL5CEUTI 5 55G ETUR5~~~~~~~~~~~~~~ THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 2
____ ______________________________________
______________
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 applicabl . 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 facto 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 DUTPU 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 zer b. Why are all commercial reactors undermoderated? (1.0)
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QUESTION 5.05 (3.00)
A Feedwater pump is operating at 75% efficienc (3.0)
The water is supplied at 680 psis and exits at 1000 psi The mass flow rate through the pump is 2 x 10E6 lbm/h l Find the TEMPERATURE RISE of the water passing through the pump due to the pump's inefficienc 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 bundl 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 paralle 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 WH 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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DY TM N QUESTION 5.09 (1.00) gg @ g$ M g~ ONuf45 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) one pound of steam at 300 psia two pounds of steam at 600 psia two pounds of steam at 1200 psia 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 choic A. A control rod's worth is greatest when it is fully withdrawn and all other rods remain inserte 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 answe (1.50)
(xxxxx CATEGORY 05 CONTINUED ON NEXT PAGE xxxxx)
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THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 5
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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?
(Justify your answer.) (1.00)
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PAGE 6 PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION
6.
GUESTION 6.01 (3.00)
The Redundant Reactivity Control System (RRCS) indentifies and prevents on ATWS condition by initiating an Alternate Rod Insertion (ARI).
(1.0)
a. List ALL conditions (with associated setpoints, if applicable)
that will actuate the ARI logic of the RRC (1.0)
b. Briefly describe how the ARI function is physically accomplishe (1.0)
c. Figure i shows the control room panel 10C651 for the RRCS. If 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).
(1.00)
a. List two uses the Process Computer has for the TIP scan dat (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)
(2.0)
List FOUR (4) automatic actions OTHER THAN a Group 1 Isolation, that will occur as a DIRECT result of the Main Steam Line Rad Monitors
'A' and*B' reaching their High-High trip setpoin QUESTION 6.04 (3.00)
(1.0)
e. State the purpose of each of the SLC tank heater (1 0)
b. State the three control locations for the SLC pumps and for each location, state whether a local START signal provided to the pump WILL or WILL NOT fire the associated squib valve (1.0) Identify the SLC tank level interlock, identify its setpoint, and state how it accomplishes its functio (xxxxx CATEGORY 06 CONTINUED ON NEXT PAGE xxxxx) PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 7
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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 conditio Even with this protection, an overspeed condition may still occu 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 Syste 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 response 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.
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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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GUESTION 6.09 (2.00)
Describe the conditions necessary to cause the following alarms on the Interlock Status Display Module associated with the Refuelin3 Platfor USE attached Figure 2 for referenc c. Back Up Hoist Limit (0.5)
b. Rod Block Interlock No. 1 (0.5)
c. Fuel Hoist Interlock (0.5)
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 trippe b. The AC oil pump (s) have tripped; the MG set scoop tube lockup trippe c. The AC oil pump (s) have trippedi the MG set drive motor continues to ru d. The AC oil pump (s) and the MG set drive motor continue to ru QUESTION 6.11 (1.00)
SELECT the one statement below which BEST DESCRIBES the operation / (1.0)
performance of an IRM durins a reactor startu a. When the IRM is reading full scale on Range 10, the APRM's should be reading approximately 10% powe b. Shifting from Range 4, indicating 75, to Range 5 will result in an
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indication of 24 on Range Reactivity feedback, due to the moderator temperature coefficient, should begin at approximately Range 5.
j d. When an IRM channel increases from 25 on Range 2 to 25 on Range 3,
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the indication has increased by one decad (***** CATEGORY 06 CONTINUED ON NEXT PAGE xxxxx)
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QUESTION 6.12 (3.00) 3 Following a valid initiation of HPCI (due to high drywell pressure) (' )
a steam leak just upstream of the steam admission valve develop 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 statu .
(xxxxx END OF CATEGORY 06 xxxxx)
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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 Rang QUESTION 7.02 (2.50)
a. List TWO reasons for placing the Mode Switch in SHUTDOWN after a (0.5)
reactor scram has been verifie b. List the plant conditions neccessary to allow entrance into the Post Scram Recovery operating procedure from: OP-EO.ZZ-101 RPV Control (1.0)
2. OP-EO.ZZ-100 SCRAM (1.0)
Note: Answer each separatel QUESTION 7.03 (3.00)
- Concerning OP-EO.ZZ-202, ' Emergency Depressurization", 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 system 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 ope Why are you instructed to skip the alternate depressurization methods if reactor pressure is less than 50 psis above the supression chamber pressure?
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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QUESTION 7.04 (2.00)
Hope Creek is in the process of refueling when the crane operator (2.0)
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 Turbin (0.5)
b. Briefly explain why reactor makeup should be maintained relatively constant during low flow condition (0.5) According to the Hope Creek Tech Specs, what two conditions define Hot Standby? (0.5)
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.
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c. What immediate operator action > are necessary per OP-AB.ZZ-127, (1.0)
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'Off Gas High Radiation'?
> 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
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would you direct him to monitor, and what would you expect him to j see if this was the case?
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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 tim (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 ope According to the procedure, what ceatain precaution must (1.00)
be taken prior to initiating corrective action and why? "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 flo (xxxxx CATEGORY 07 CONTINUED ON NEXT PAGE xxxxx)
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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 Instrument Gas.
l a. Explain why this is don (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 *****)
- ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 14
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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 day 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 Superviso USING THE ATTACHED TECH SPECS- What are your required actions and what is the most limiting LCO?
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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 issue b. List the Shift Supervisor's responsibilities for an RW (0.9)
QUESTION 8.05 (1.50) During events that require implementation of the Emergency Plan (0.5)
procedures, WHO initially becomes the Emergency Director?
b. List two other individuals who may assume the role of Emergency (1.0)
Directo (xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx)
.
- , , . - - - , , --~_, ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 15 8.
__________________________________________________________
QUESTION 8.06 (3.00)
(3.0)
You have just received word from the IRC supervisor that Reactor 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 *,rumen 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 z x NOTE: USE THE ATTACHED SECTIONS OF THE HOPE CREEK TECH SPECS TO x x ANSWER THIS DUESTION. FULLY REFERENCE ALL APPLICABLE SECTIONS OF *
x THE T.S. THAT YOU USE TO DEVELOP YOUR ANSWE 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 calibratio use of faulty 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 limitin .
(xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx)
.
. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 16
__________________________________________________________
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 questio TRUE or FALSE b. Briefly describe the actions necessary to verify that a manual (0.5)
valve is OPE c. Briefly describe the actions necessary to verify that a manual (0.5)
valve is CLOSE According to the " Removal and Raturn of Equipment to Service' procedure, gdj When equipment is removed from service, the appropriate informa- 'O.;}-
[ tion should be entered in what two 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 sever b. State whether or not the following occurrences require one-hour (2.0)
,
to the NRC. (Answer YES or NO.)
l
! 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 day . Reactor vessel water level reaches -40'
i i
(xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx)
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 17 8.
__________________________________________________________
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 WH b. If the same IRMs were found inoperable while in RUN, would you violate any Tech Specs by:
Explai (0.5)
1. Staying in RUN?
(0.5) Placing the mode switch in STARTUP? Explai (xxxxx END OF CATEGORY 08 xxxxx)
(x*******xxxxx END OF EXAMINATION xxxxxxxxxxxxxxx)
woni wo u y- m -u
,
,, 33
' '
N' 'ER
'
p = p, et / r M *
1/(1-k) .-
I Ici - 3.7 x 101 %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
. .
_
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) (IA)
5 - x5g + (1-x) Sg Q = 2rt.; at 1 in. - 2.54 ces ap = f L ov2 1 gal . - 3 785 liters D 2sc 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
- ' ' *
1 psi - 2.936 - H tiSO(0(G0:}
4 ;
tl C r(eff)1 1 psi = 27 68 *
. I = .0071 0 = He _
. 1 = 2 x 10 , sec l q = UAAT l
l
-
!
f
l
l.:l -
i
'
n c;
')
P... i., . THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 18 ____ ______________________________________
______________
ANSWERS -- HOPE CREEK -86/02/24-!(OLONAUSKI/LANGE ANSWER 5.01 (3.00) a. APRM se m at 15% gW (0.33)
(0.33)
b. Active in TARTUP (0.33)
c. SL- For Rea tor Pressure les than 785 psis or core flow less than 10% f rated, ta d thermal power must not exceed 25%. APRM scram at 15% ed thermal power in STARTUP (0.50) b. Fixed APRM scram f 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 1.0 (FRTP FLPD) (0.33)
3. a. T=
APRM ga'i is adjusted to re d 100% times the CHFLPD when T is les than or equal to No not specifie '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 we REFERENCE HC LP HT&T No. 14) Learning Objective N , Page HC TS LP, LO 1 ANSWER 5.02 (1.00) The Kf factor adjusts the MCPR operating limit for core flows (0.5)
other than rate (0.5)
2. The Kf factor assures that the MCPR SL will not be exceeded during the flow increase transient resulting from an MG set speed control failur REFERENCE HC LP HT&T No. 14) Learning Objective No. 5, page 5
.
A. L55 5 - l6't. therm yeiadr scranw G0%
ack ta.- whe.w NOT in guts) o. 2 SL - 2 s't. rahd. WerumA. yeuse.r w4<m Kx ye<. wu. < T w p3i) oA ove cort flow <,10*/. rakd Gw - blaud scrwm ; ek@ a.$ til. V '4 b. L 5 5 5 - (c. 66W + 61) T
activt - All Madel SL - % HCYtt 7, l.0L % h ynthLM. > ]f 5' yEE*] A *
f-(su > ls't. M O'I LIJJ - ll 3 */. fint A M N ICh ar,tiut ~ kVM M.ch 0Y S t., - s a.h4L 4.s b,
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 19 5.
____ _ _________________________________. ___
______________
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 zer The resonnance escape probability, p, DECREASES as the moderator (0.5)
to fuel ratio decreases toward zer B. An undermoderated core provides stability as power is increase (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 increas REFERENCE Ex.l 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 20
____ ______________________________________
ANSWERS - HbPbCREEK -86/02/24-KOLONAUSKI/LANGE ANSWER 5.05 (J<[50)
c. Pump Head- (0.6)
Hp = (Pout-Pin)/
= (1000-680)*144/62.4 = 738.5 ft Ideal Work (power)- (0.6) Wpeideal = Hp m
= 738.5 (2 x 10E6)/60 = 2.46 x 10E7 ft-lb/ min c. Input Power- (0.6)
Wp, actual = WP, ideal /Np
= 2.46 x 10E7 ft-lb/ min / 0.75
= 3.28 x 10E7 ft-lb/ min d. Power converted to heat- (0.6)
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 e. Delta T across pump- (0.6)
delta T = O/mCp
= 10,566 (60)/2 x 10E6 = 0.32 des F REFERENCE HC LP FF N , pages 11-13 FF LP 03, LO 7 ANSWER 5.06 (2.00)
(0.5) Increases (0.5)
l Decreases
' (0.5) Decreases (0.5) Increases REFERENCE HC HT&T No. 11, Learning Objective 2, pages 8 and .TT cmuoT 61%vS S 4" 'OC**t*
grNg!(t dWudY hr 6 05 qxtity h $(w )wWaMygry
C Pua*O . Med 2 yrerHeo .
& = 4 w A G di - A Ak
.. AT: A %ead tdid 4 aHu yn,9 g i Cy 03 Cy o.w {3N'F nA) 7nt ETu N4 . THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 21
____ _ _ ______________________________________
______________
ANSWERS -- HOPE CREEK -86/02/24-KOLONAUSKI/LANGE ANSWER 5 07 (1.50)
1' s s, (0.5 for each curve)
I s p"*b'
l
-'
Pump i \
Discharge Head 1
- - 'c'
i l.a. / \ 'M 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) REFERENCE Steam Tables e
l l
l THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 22
--------------------------------------
__---------_--
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 REFERENCE MWM W A IN CM MY Hope Creek LP RXPH31 pgs. 11, 12 LO 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 fissio (1.00) 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) Shorter periods as BETA decreases for the same reactivity insertion (1.00)
REFERENCE LP RXPH 23-01 p , 10. Learning objectives 3, 6 (FJC185)
ANSWER 5.12 (2.50) 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 constan (1.00) 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 negativ (1.50)
. . _ . _
PAGE 23 THEORY OF NUCLEAR POWER. PLANT OPERATION, FLUIDS, AND
--------------------------------------
--_---_-------
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 stea REFERENCE FJC264 Hope Creek M.O.C.D. LP 104 pg. 15 I.O. 41
.-.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 24 ______________________________________________________
ANSWERS -- HOPE CREEK -86/02/24-KOLONAUSKI/LANGE ANSWER 6.01 (3.00)
o. Vessel High Pressure, 1071 psis, OR (1 0)
Vessel Low Level (-38'), OR RRCS Manual Initiation, OR ARI Test Switch turned to TEST Eight DC solenoid actuated vent valves blow down the Scram Valve (1.0) Air Pilot Heade Rx pressure is <1071 psig, Rx water level is > -38', and the 30 (1.0) second time delay has timed ou 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)
deplete uranium). o.s tr 5 '^
.51
'(d 1" 2. Calculates substitute data for out of service LPRMs -
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 enclosur Corrosion products could build up on the helical wrap around the (0.5)
drive cable and cause an electrical signal los (
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 3. Reactor scram o.C7
'
. C1_.d >=.1 wiw.- ..ul itun (4 needed; (k/Ieach)
l PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 25
ANSWERS -- HOPE CREEK -86/02/24-KOLONAUSKI/LANGE REFERENCE pgg y N o p , y m ge3r A Y M s M d o p !C M LP 45, pg. 20 g.g, > de.tg .
LP 53, pg. 15 LP 22, pg. 61 ANSWER 6.04 (3.00) kW ("o rating heater') - Maintains solution temperature to (0.5)
prevent ecipitation of the sodium pentaborate during operation k (*. mixing heater") - Used to heat and maintain water temp- (0.5)
er ure during solution preparatio b Remote Control Panel 10C011; WILL NOT (0.33)
Control Room Panel 10C6513 WILL (0.33)
RRCS Panel; WILL --+ If PtY'"U3I* *
(0.33) 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)
REFERENCE e me no M M $a h .
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 The combined intermediate valve .5)
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
_ . _ _ _ . PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTA1 ION PAGE 26
ANSWERS -- HOPE CREEK -86/02/24-KOLONAUSKI/LANGE ANSWER 6.07 (2.25) g g(g-90 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 (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 clos 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
') 52-y3 i o) , SDC initt site y kuevde c( - ggg y gy y 4, (o g of a- 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 h-
. . . , . - - - w,;; .wyu u:'em g;_ u;~ 0-men;=w=a samt a 3 s,w< o.i J PLANT SYBTEMS DESICH, CONTROL, AND INSTRUMENTATION PAGE 27 ;
______________________________________________________
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 switc 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 vesse 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 loade 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 hoist REFERENCE GE Refueling Tools Familiarization Hanval #2, pages 3-12, 3-13 ANSWER 6.10 (1.00) Od/Or b . (1.0)
REFERENCE HC LP 19, pages 22,32 Figure 18, LO 10,13 ANSWER 6.11 (1.00) (1.0)
REFERENCE HC LP 14-01, pages 9,10 LO .- PLANT BYSfEhs DESIGN, CONTROL, AND INSTRUMENTATION PAGE 20
_,-__- -__--____------_-___--___---____-_____, ,
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 clos (1.00)
2. Turbine Trip due to isolation (1.00)
3. Vacuum breaker IV's close (due to low steam line pressure and hi DW) - ont h if awefHr Wa.k dJ.ygsf ung & My in 100 ys(3 (1.00)
N '
" '
REFERENCE HOPE CREEK LP16 pgs. 20,70,71, I.O.47 FJC259
- - _ .
--_ _
. - - - =_ ~- .- .-
,
70 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 29
~~~~~~~~~~~~~~~~~~~~~~~~
~~~~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 assure REFERENCE OP-SO.BC-001(0) RHR system procedure, Precaution 3.1 SO LP, LO 1 gg,grcrM* MC 3. gypst13 N&h e en ANSWER 7.02 (2 50)
1. To backup the scram with add ional RPS contacts (0.25)
2.ofj . To keep the HSIV's open by moving out of RUN, and therefore (0.25)
-
preventing the low pressure isolatio :8' o.12 5 0.115- . Pressure stabilized,and RPV level stabilized above g TAF -AND- (0.25)
,*a. All control rods inserted to or past 02, OR ) (0.25)
MN lbs of boron have been injected into the RPV, OR (0.25) The reactor is shutdown and no baron has been lnjecte (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 .
(NTENT oF & Y"" h W g t, 2 - p( rd t.Mtto. S.u_ if %
.
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 s
e I
l l
l l
r
__ _ _ . . _ ._ _,
_-_ .
- - - - - - - - - - - - _ _ _ _ _ _ _ _ _ _ _ _
,
(
'
o PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 30
- ~~~~~~~~~~~~~~~~~~~~~~~~
~~~~Rd656L6GiCdL 56UTR6L
____________________
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)
top of the SRV T-quenchers, rapid chamber and drywell would result. (pressurization It is possible of thethe that supression design k05 pressure for the containment would be exceededy g7 gg MSL Drains (1,o)
Main Condenser An.Ladr, ihML C " pd. Mt6W '
Head Vent HPCI RCIC SJAE RFPT (0.25 each, 4 needed for full credit) . 50 psig is the lowest differential pressure at which an SRV 'O.75)-
will remain fully open with its control switch placed in the 09 open positio . Below 50 psis, RPV Depressurization is considered complete -- 4 0 . 2 5 )
and additional steam paths are not necessar REFERENCE OP-EO.ZZ-202 LP pages 6, 7 / b. page 8 / c. page 8 ANSWER 7.04 (2.00) Suspend all refueling operation (0.5)
2. Evacuate all unecessary personnel from the Refuel Floo (0.5)
3. Ensure the Reactor Building Ventilation isolate (0.5)
4. Ensure that the Filtration, Recirculation, and Ventilation (0.5)
system (RFVS) auto start 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 shaf b. This action will minimize the thermal transients on the RP (0.5)
c. Mode switch in Startup/ Hot Standby, and Reactor Coolant at any (0.5)
temperature
- - .
.
I PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 31
~ ~~~~~~~~~~~~~~~~~~~~~~~~
^^~~ RI65UL6G5CdL 66UTR6L
____________________
ANSWERS -- HOPE CREEK -86/02/24-KOLONAUSKI/LANGE REFERENCE Operations from Hot Standby procedure I0-007, Cautions 5 2.19.1 and 5. Hope Creek Tech Specs, Table IO LP, LO 1 ANSWER 7.06 (2.00) . Reduce power as necessary to maintain the Offgas activity (0.5)
less than the high high alarm setpoin . 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 flo (0.5)
Monitor the Main Condenser for a decrease in vacuu (0.5)
.
REFERENCE or mon (to r 5 ) A E 3rd ; N* / h i ^ h OP-AB.ZZ-127 AB LP, LO 1 ANSWER 7.07 (2.50) . 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). 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 b. OP-EO.ZZ-100 Scram LP pa3e 7 ANSWER 7.08 (3.00)
o. To maintain an accurate indication of reactor perio (1.00)
b. Increase pressure setpoint to prevent unplanned depressurization. (1.00) o minimize thermal transients on the reactor vessel. Opening a bypass valve may be necessary to acheive steady feedwater flow. (1.00)
05 L ._ .
. . _ _ .
_ = _ - . . . . PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 32
~~~~~~~~~~~~~~~~~~~~~~~~
~~~~Rd656L66565E"66 TR6L
____________________
ANSWERS -- HOPE CREEK -86/02/24-KOLONAUSKI/LANGE REFERENCE Hcpe Creek OP-IO-ZZ-003 p , cautions 3.2.5, 3 FJC277 ANSWER 7.09 (3.00) g CMTramL Syb 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) CORE SUBMERGENCE accomplished by maintaining a level above TA (1.0) SPRAY COOLING accomplished by using at least one CS system operating at design condition I (1.0)
whtasc Kx te pnts cie:r) 700 yt()t fi c .
cidre, 3. STEAM COOLING accomplished by using3SRV tc 2 _ , _ . . _ _ ____ ___, , _ _ . . . . (g,o)
4....__._ O IAdi yhiM ww ut bcO W . . . . . "
~ ~ ~ ' ' ~~ g [ d 'y U d < i 5'O' yr(
REFERENCE Hope Creek definition of adequate core cooling in E0P's (3.00)
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 Temperature Limit (HCTL) intheSgpgessionChambe (0.5)
The PCIG low level isolation is bypassed to ensure a continued pneumatic supply to the inboard MSIV (0.5) Yes (0.5)
_ _ _ _
. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 33
~~~~ -
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R 656L55iEAL E5sTRUL'~~~~~~~~~~~~~~~~~~~~~~~
____________________
ANSWERS -- HOPE CREEK -86/02/24-KOLONAUSKI/LANGE REFERENCE OP-EO.ZZ-207 LP pages 9, 10
5 i
I t
i d
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l
. _ _ . - . - - . _ . . - . . _ _ _ . _ _ _ _ . . . _ , _ _ _ . _ _ _ _ _ _ . _ _ _ _ . _ _ . , ,_.__ ,_.. __ . . _ . _ _ _ . . _ . _ , _ _ _ . , . . . _ . , _ _ _ _ _ - . . - _ _ _ _
, , . . _ . _ . _ . ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 34
__________________________________________________________
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 minute ( 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 />, or reduce power to less than 25% RATED THERMAL POWER within (0.79).60 the next four hour ,,g The most limiting LCO is (0.75).d*
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 33 ANSWER 8.02 (2.00)
o. The intent of the original procedure is not altere (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 uni The change is documented and receives the same level of review (0.67) and approval as the original procedure within 14 days of implemen-tatio REFERENCE TS 6. ANSWER 8.03 (1.50) .
ylate tho thop chAWb4 l M The action required by Table 3.3.1-A is to .: ir et Iract MOT C""TC0"N withi.. th: .:: t 12 h mm e - tW. ty(ypcl carditan Wihin 9 h0hrt . (1.50)
REFERENCE t
TS 3/4.3.1, RPS Instrumentation
'
- . , .
PAGE 35 ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS
__________________________________________________________
ANSWERS -- HOPE CREEK -86/02/24-KOLONAUSKI/LANGE ANSWER 8.04 (2.50) ggjtty gg, c. (Ch ek wi h Training Department )
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 utro radiation exposure >5 arem/hr 6. Airbor radioactivity requiring the use of respiratory protection equipmen . (4 needed at 0.4 each)
(0.9)
b. Get SS responsibilities from training departmen 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 ,et (1.0)
b. EDO, E R M --- EC Eyan - Sr. W WWCASar HC. (:tntd b$ty dC- g REFERENCE a. OP-AP.ZZ-002 (0) 5 ECG Attachments ANSWER 8.06 (3.00)
Due to the fact that this instrument functions as one of the Core Spray permissive signals (0.25), the minimum operable channels per trip system requirement of T.S. Table 3 3.3-1 for the CS system cannot be met for one trip system (0.75).
Therefore, in accordance with TS 3.3.3.b, the action of Table 3.3.3-1 must be declared inop. (1.0)
be taken. This action requires that the CS e(S8p>3.5.1.a.$ must be take Having CS inop requires that the actions of (1.0) [
REFERENCE I {0dy of CI (hty Hope Creek Technical Specifications TS LP, LO 2
.
- -- , , - _ - - .
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 36 8.
__________________________________________________________
ANSWERS -- HOPE CREEK -86/02/24-KOLONAUSKI/LANGE ANSWER 8.07 (2.50) g,3 3.3.4.2.a, 3.3 4. M-re 1 25-44,H 3.3.1.b and Table 3.3.1-1 -.m_ Action 63.3 ; ; , ; ; ,,, , , , v t a 3 ; (0, ;;
n;t 1; ..g- .. A -
REFERENCE Hope Creek Tech Specs Ray 1,4 TV'OE - th h can or a. re.cand yarg TS LP, LO 2 vty(Eca.han for a. ihrotHt. valvt . Wrw(54. fatu ANSWER 8.08 -'I (0.5)
o. False Hove the valve's handwheel in the closed position just enough (0.5)
to verify valve movement, then return to ope Hove the valve's handwheel in the closed direction onl (0.5)
' O . "; ) ^ r t i e r- ctetr:.cnt L:3; E7;ir :ent Un /cil:ble Li;t gg
.
(0 5)
e. True REFERENCE OP-AP.ZZ-109 ' Equipment Operational Control'
o. 5. b. 5. c. 5. OP-AP.ZZ-108 ' Removal and Return of Equipment to Service'
d. 5.1. e. 5. ANSWER 8.09 (1.50)
(0.75) A maximum allowable extension not to exceed 25% of the surveil-lance interva . The combined time interval for any 3 consecutive surveillances (0.75)
shall not exceed 3.25 times the specified surveillance interva REFERENCE Hope Creek Tech Specs Surveillance Requirements, 4.0.2 a and .- - - . _ _ .-. .- - - .-.- ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 37
ANSWERS -- HOPE CREEK -86/02/24-KOLONAUSKI/LANGE ANSWER 8.10 (3.00)
c. Unusual Event (1.0)
Alert Site Area Emergency General Emergency . Yes (0.5) Yes (0.5) No (0.5) 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. b. 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:bl REFERENCE HC TS, 3/4 0-1, 3/4 3-1 to 3-4 k. 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 1 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 2 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
.
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l
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e
I -
!
. . . .
t
- ,
,
.
j Table Saturated Steam: Temperature Table
- - . - . - - . . .
__
Sa Sa Sat Sa Sa Sa Temp Temp lb per
'
l 1iquid Evap Vapor Lic uid Evap Vapor Liquid Evap Vapor Fahr falu Sg in sa I v vg h fg hg sg sgg i p vg _ig, if 33047 3304 7 0 0179 107 .5 0.0000 2.1873 2.1873 32A 32 8 0 08859 0 016022
,
0 016021 3061 9 306 .996 101 .0041 2 1762 2.1802 3 .1651 MA 0016020 28390 28390 4 008 107 .2 0.0081 2.1732 1 36 8 0 10395 38 4 0 016019 26341 2634 2 6.018 107 .1 0.0122 2.1541 2.1663 34 0 011249 l
8.027 107 .0 0.0162 2.1432 2.1594 4 .12163 0 016019 2445 8 2445 8 013I43 0 016019 22724 / 227 .9 0.0202 2.1325 2.1527 42A 42 8 44A 0 016019 21128 2112 8 12.041 10681 10803 0.0242 2.1217 2.1459 44 8 - 0 14192 48A 0 016020 19657 19657 14.047 106 .6 0 0282 2.1111 2.1393 48 8 015314 48 0 0 16514 0 016021 1R30 0 1830 0 16 051 1066 4 108 .0321 2.l006 2.1327 40A 1704 8 18 054 106 .4 0.036l 2.0901 2.1262 50A l 54 8 0 17796 0016023 1704 8 0 19165 0 016024 15892 1509 2 20 057 106 .2 0.0400 2.0798 2.1197 5 l 52 0 54 I O20625 0016026 1482 4 148 .058 106 .1 0.0439 2.0695 2.1134 5 .6 24 059 106 .0 0.0478 2.0593 2.1070 5 .2 26 060 10601 108 .0516 2.0491 2.1000 50 0
? 58 0 0 23843 0 2561! 0016033 120 .0555 2.0391 2.0946 80.0 l 80 8 0.0593 2.0291 2.0885 6 I 027494 0 016036 II29 2 II2 .5 108 .058 105 .5 0 0632 2.0192 2.0824 6 .1 34.056 105 .4 0.0670 2.0094 2.0764 $$A 88 8 0 31626 etI 033889 0 016046 976 5 926 5 36.054 105 .2 0.0708 1.9996 2.0704 50A
,
70 0 0 36'32 0 016050 868 3 868 4 38.052 105 .1 0.0745 1.9900 2.0645 10A 12 0 0 38844 0 016054 814 3 814 3 40 049 105 .0 0.0783 1.9804 2.0587 7 .046 105 .8 0 0821 1.9700 2.0529 14.0
,
'
76 3 0 44420 0 016063 117 4 71 .0858 1.9614 2.0472 76.0
,
is e 04746I n016067 611 A 613 9 46 040 104 .0095 1.9520 2.0415 7 .4 109 .0932 1.9426 2.0959 0 .3 0.0969 1.9334 2.0303 0 MS 057702 0 016082 560 3 560 3 52 029 1046.1 1998 2 0.1006 1.9242 2.0248 .5 54 026 10450 1099 0 0.1043 1.9151 2.0193 0 % 8 4968 56 022 1043.9 109 .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 0 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 41 l88 IA792 1.9900 > 98 I 084012 0 016111 392 8 392 9 64 006 1039 3 110 .8704 I9928 9 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 Sa Sa Sa Sa Sa Sa Temp i alu Sq in liquial Ivap Vapos liquid Evap Vapor liquid Evap Vapor Fahr vig hl h ig hg s, sgg sa I I p '
vi vs tes e 094924 0 016130 350 4 350 4 67.999 103 .1295 1A530 l.9825 1984 100789 0 016137 331 1 331 1 69 995 103 .9 0.1331 1.8444 8.9775 10 IN I I06965 0 016144 313 1 313 1 71.992 103 .8 0.1366 IA358 I.9725 10 les e i1347 0 016151 296 16 29618 13 99 1033 6 110 .1402 1.8273 19675 188A 0 016158 280 28 200 30 75 98 103 .5 0.1437 1.8188 1.9626 1984 100 e 1.2030 1.2750 0 016165 2F 265 39 71.98 103 .3 0.1472 1.8105 I.9577 198A lieI I8021 1.9528 11 M : ;7 25138 79 98 103 .4299 0 016180 238 21 238 22 81.97 1029I 111 .1542 13938 1.9480 114A 11 .9 111 .1577 13856 1.9433 11 IIII 0 0161 % 85.97 102 .7 0.161I I1 774 1.9306 118A lit e 16009 214 20 f 214 ?!
0016204 20325 203 26 87.97 1025 6 111 .1646 1.7693 1.9339 128 120 0 I6927 0.1600 1.7613 3.9293 122A 0 016213 192 94 19295 89.96 1024 5 1114.4 i 122 0 17891 13533 1.9247 Ing
'
0 016221 18323 183 24 91 % 1023 3 1115 3 0.1715 124 8 1 8901 SES
'
I9959 0016229 114 08 114 09 93 96 102 Il1 .1749 13453 1.9202 128 8 0 016238 15545 16547 95.% 102 .0 0 1783 13374 1.9157 IES 178 8 2.1068 15732 157 33 97.96 101 .8 0.1817 13295 1.9112 I3RA
! 130 0 2 2230 0 016247
'
2.3445 0 018256 149 64 149 66 99 95 10187 1118 6 0.1851 13217 I.9068 132A 13 Int
' 2 4111 0 016265 14240 142.41 101.95 101 .5 0.1884 13140 1.9G24 134 0 1ES 136 I 2 6041 0 016214 135 55 135 57 103 95 1016 4 182 .1913 13063 1.8900 i IES 130 0 2 1438 0 016284 129 n9 129 11 105 95 101 .1 0.1951 1.6906 IA937 i
2 8892 0 016293 122 98 123 00 107.95 1014 0 182 .1985 1.6910 13895 148A
! 140 0 112 .6534 1.8852 142.0 4 142 8 3 0411 0 016303 1I721 11722 109.95 101 .95 10113 1123 6 0.2051 1.6759 1.8810 144.0 j I48 8 3 3653 0 016322 106 58 106 59 113 95 101 .5 0 2084 1.6684 1 8769 14 E8 10130 115 95 100 .3 0.2117 1.6610 1.4727 140A
' les e 3 Mel 0 016317 ISO O 3 7184 0016343 9705 9707 117.95 1008 2 112 .2150 1.6536 1.0646 120 3 9065 0 016353 9266 9268 119 95 100 .9 0 2183 1.6463 .l.8646 15 R0 1548 4 1025 0 016363 8850 88 52 121 95 1005 8 11273 02216 1.6390 1.8606 4 3068 0 016314 84 56 84 57 123 95 100 .2248 1.6318 1.8566 IES i 154 8 0 016384 80 82 80 83 125 96 100 .4 0 2281 1.6245 1.8526 ISSA 1 158 8 4 5197
1130 2 0.2313 1.6174 1.8487 ISSA 1 letI 4 1414 0 016395 ?? 27 7/29 127 96 1002.2 j 182 8 4.9722 0 016406 13 90 73 92 129.96 100 .0 0.2345 1.6103 13448 10 .96 999 8 113 .6032 1.8409 IKO 188 3 54623 0 016428 6167 6768 13397 99 .6 0 2409 1.5961 IA371 1ESA
.
let I 57223 G016440 64 18 6480 13597 99 .4 0.2441 1.5N2 IJ333 ISSA l
) lit t 5 9926 0 016451 62 04 62 06 137.97 99 .2 0.2473 1.5822 1A295 128 172 8 6 2136 0 018463 59 43 5945 139 98 99 .0 0.2505 1.5753 1A258 172A
<
174 8 6 5656 0 016474 56 95 56 97 141.98 993 8 113 .5604 1A221 INA l
IIIe 68690 0 016486 54 59 54 61 143 99 99 .6 0.2568 1.5616 1.8184 1R0 lit e 7.1840 0 016498 5235 57 36 145 99 99 .4 0.2600 I.5548 IA147 128 l
/
~
3 @
a . s *
3 . Specific Volume Enthalpy Entropy Abs Press Sa Temp Sal Sal Sal Sa Sa Temp tb per Liquid Evap Vapor Fahr Iahr Sq i liqmd [vap Vapor liquid Evap Vapor he h gg hg s, seg st I I p vi vtg va 14800 990 2 11382 0 2631 1.5480 1.8111 10 les s 15110 0 016510 5021 50 22 150 01 989 0 1139 0 03662 1 5413 1.0075 18 '0016522 48172 18 189 I .8 1l39 8 0.2694 I.5346 I8040 184 8 8 203 0 016534 1.8004 18 s 44 383 44 400 154 02 986 5 1140 5 0 2725 1.5279 i tes t 8 568 0 016547 13%9 10 .2756 1.5213 tes e 8 947 9 8 .1 02787 1.5148 1.7934 19 [ 40 957 158 04
'
let t 9 340 39 354 160 05 98 .9 0 2818 1.5082 13900 19 I :2848 1.5017 13865 194 0 10 168 0 016598 1.7831 19 .4 II44 4 0.2879 1.4952 198 9 10 605 0016611 I.7798 19 !4 970 166 08 97 .4888 196 I 11058 0016637 33 622 33 639 168 09 97 .0 03940 1.4824 13764 20 I 11526 0 3001 1.4697 13698 2 .5 284 8 12 512 0 3061 1.4571 13632 20 I 0312I 1.4447 13568 21 I4 696 0016719 26 182 26199 18017 970 3 1150 5 212 8 13505 21 R78 24 894 184 20 967 8 115 .4323 215 0 d
0 016775 23 131 23 148 188 23 965 2 115 .7442' 22 !!t I 17 186 0 016805 21 529 21545 192 27 96 .9 03300 1.4081 13380 22 .3 0 3359 13961 13320 228 0 1 228 8 20 015 0 016864 18701 18 718 200 35 95 .8 0 3417 13842 13260 23 .2 0 3476 13725 13201 236 0 236 8 23 216 0 016926 14 304 16 321 208 45 95 .3609 13142 24 I 24 % 8 0 016958 15243 15260 212 50 949 5 116 .3494 13005 24 .56 946 8 116 d 13028 248 I 13266 16972 25 /2 13 358 13 375 22062 94 .4 116 .4 0 3819 1.30(3 16862 260 0 294 0 35 427 37 894 0 017123 11 025 11 042 232 83 935 9 1168 7 0 3876 12933 I6808 26 .0 0 017157 10 358 10 375 236 91 9331 1170 0 0 3932 1.2823 1 6755 284 8 40 500 24099 930 3 1171 3 0 3987 1.2715 1.6702 27 .8 43249 0 017193 9138 9 755 46147 0 017228 9 162 9 180 245 08 9215 1172 5 04043 1.2607 I6650 27 lit I I
200 0 49 200 0 017264 8 627 8 644 249 17 92 .2501 1.6599 20 Ofl130 81280 8 1453 253 3 9213 1175 0 0 4154 1 2395 1.6548 2 ,
76634 7 6807 2574 918 8 1176 2 0 4208 1.2290 1.6498 29 .5 915 9 117 .2186 1.6449 29 .8 59 350 0 01141 6R759 6 8433 265 6 913 0 117 .2002 16400 29 i
'
?
w 298 8 63 084-
,"
__ _ _ ___
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _
_ . _
O i
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"
~$pecific Yolume Enthalpy Entropy Abs Press. ' ~
Lb per Sal Sa Sa Sa Sa Sa Temp Temp liquid Evap Vapor Liquid Evap Vapor L6 quid Evap Vapor Fahr Iahr Sq in he h gg hg sg seg sa I I p vg vf( vs see g 67 005 0 01145 6 4483 6 4658 26 .7 04372 1.1979 1.6351 30 M48 71119 0Cl149 6 0955 6 1130 273 8 90 .9 04426 1.1877 1.6303 304 8 75 433 0 01153 5 7655 5 7830 278 0 9040 118 .4479 1.1776 1.6256 30 .0 118 .4533 1.1676 l.6209 31 als 8 84 688 0 01761 5 1673 5 1849 28 .9 118 .4586 1.1576 l.616 31 li66 48%I 4 9138 290 4 894 8 1185 2 0.4640 1.1477 1.6116 32 NG 32 .6 118 .1378 1.6071
! 324 0 328 8 100 245 0 01774 4 4030 j 4 4208 2987 888 5 118 .1200 1.6025 32 .9 885 3 118 .4798 1.1183 1.5981 33 I . til 820 0 01783 3 96Al 3 9859 30 .1 11891 0.4850 1.1006 I.5936 33 .992 001787 3 7699 3 7878 311 3 878 8 189 .4902 J.0990 1.5892 34 .0 0 4954 1.0894 1.5849 34 I 124 430 131.142 0 01797 3 4078 3 4258 319 7 87 .1 0 5006 1.0799 1.5806 348 e 348 8 001801 3 2423 3 2603 323 9 86 .7 0.5058 1.0705 1.5763 352.8 i 352 0 138 138 145 424 001n06 30863 3.1044 328I 865 5 1l93 6 0 5110 1.0611 1.5121 3M.5
354 0 Meg 153.010 0 01811 2 9392 29513 3323 86 .5161 1.0517 1.5678 36 s40 160 903 0 01816 2 8002 2 8184 336 5 858 6 1895 2 0 5212 1.0424 1.5637 36 Ms0 169 113 001821 2 6691 2 6813 340 8 855I 1l95 9 05263 1.0332 1.5595 36 % 1 0.5314 1.0240 1.5554 31 t$ 8 186 517 0 01831 24279 24462 349 3 848I 1197.4 0.5365 1.0148 1.5513 37 eeI 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 35 .9966 1.5432 38 O 215 220 0 01847 2 1826 2 1311 36 .2 1199 3 05516 0.9816 1.5392 38 .9786 1.5352 39 I929! l94?? 370 3 829 7 1200 4 0.5617 0.96 % 1s5313 39 I8444 I8630 37 .9 1201.0 0.5667 0.9607 1.5274 40 I1640 11827 3194 822 0 120 .5717 0.9518 1.5234 40 I 270 600 0 01815 16871 17064 383 8 818 2 120 .5766 0.9429 1.5195 400.0
412 0 282 894 0 01881 16152 16340 38 .2 120 .5816 0.9341 1.5157 41 elle 295 617 0 01887 1 5461 15651 392 5 81 .8 0.5866 0.9253 1.5118 416 8
.
429 0 308 180 0 01894 14808 14997 396 9 806 2 120 .5915 0.9165 1.5000 429 0 4248 322 391 0 01900 14184 14374 40 .2 120 .9077 1.5042 42 I 336 463 001906 13591 13782 4057 798 0 1203 7 06014 0 8990 1.5004 42 ;
i 432 0 351 00 0 01913 130266 1 32179 410 1 793 9 1204 0 0 6063 0.8903 1.4966 43 RH1 126806 414 6 189 7 120 .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 78 .0 a 444 I 414 09 0 01940 1.10712 112152 428 0 176 7 120 .4815 448 8
!
(*7 $80 431 14 0 01947 105764 1 01111 '12 5 71 .8411 1.4718 ( . it
! 44873 0 01954 1 01518 1 03472 J .8 1204 8 0 6356 0 8385 I.4741 W40
. ___ .--
_ - _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ . _ _ _ .
. . . _ . _ . .
i e
./ .
Jo I
Ahs l'ress Specifer Volonie [nthalpy Entropy lemp lb per Sal Sal Sa Sa Sa Sa Temp i eqiut! Ivan Vapor liquid [vap Vapor liquid (vap Vapor Fahr i alu So lu i p vg v ig_ vg hl h ig hg s, sla 5s I 0 97463 0 99424 44 .4704 40 .4667 40 .6 0 6502 03127 1.4629 48 A9885 0 88329 455 2 74 &6551 08042 1.4592 47 A4 0 86145 c
545ll 001997 087954 On4950 459 9 744 5 120 .4555 4Nt l 416 e 464 5 7398 120 .4510 400 0 des 8 566 15 002000 0 19716 / 0 81717 40 I 587 81 0 02009 076611 0 18622 469I 7343 12038 06696 03705 1.4481
'
0 13641 0 15658 473 8 129 7 1203 5 06745 03700 1.4444 400 0 410 8 610 10 0 07011 0 10191 012820 478 5 724 6 120 .4407 40 % El 0 07034 0 un65 0 10100 4832 719 5 120 .4370 40 .9 71 .4333 50 e4 8 105 18 0 07053 042938 0 64991 4923 709 0 12013 0 6939 03357 1.4296 50 .5 703 7 120 .4258 50 I40 0 02062 0 60530 512 8 15772 0 02012 0 5R218 060289 502 3 6982 1200 5 03036 07185 1.4221 51 % 997 0 58019 5071 6923 11918 0 7005 U1099 1.4143 51 III
$20 5 812 53 0 02091 0 53864 0 55956 512 0 6870 1199 0 01133 03013 1.4146 52 .4100 52 .3 0 7231 0 6839 1.4070 578 0 578 8 900 34 0 02123 041947 0 50070 526 8 669 6 11 % 4 0 7280 0 6752 1.4032 53 .11 0 07134 0 46173 0 48757 531 1 663 6 1195 4 0 7329 0 6665 1.3993 53 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 119 .3915 54 I 1028 49 0 02169 0 41048 043217 546 9 645 0 119 .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 55 I 1097 55 0 07194 0 17966 0 40160 5572 63 .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 56 .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 57 ??64 0 11162 0 33426 583 7 5911 118 .5766 1.3592 57 .9 1179 0 0 7076 0.5673 1.3550 50 .3507 50 t100 0 02311 027608 0 79919 6001 574 7 1174 8 0 7978 0.5485 1.3464 50 .5390 1.3420 50 e 592 8 1453 3 0 02328 0 26199 0 28827 6053 566 8 117 " 598 8 14978 0 07345 0 754?$ 0 21770 61 .2 0 8082 0.5293 1.3375 30 _ -
-_
_ . _ _ _ _ _ _ _ -_
__.__ _ ,
.
.'.
Y m
.
Ahs Press Specific Volume . Enthalpy Entropy femp lb pe Sa Sa Sa Sa Sa Sa Temp Ialu Sq in liquid Ivap Vapor Li vid Evap Vapor liquid Evap Vapor Falw I P V L, _v ig_,_W8 I h is hg s, s ,, s, g SeeI 1543 2 0 02364 0 24384 0 28747 61 .7 08134 0.5196 1.3330 000A 15897 0 02382 0 23374 0 25757 62 .2 116 .5097 132M 80 e 16373 0 02402 0 22394 0 24196 628 8 53 .4 0 8240 0.4997 1.3238 GABA Si! 8 16861 0 02422 021442 0 23865 6348 5241 115 .4896 13190 812A sie s 1735 9 0 02444 020516 0 22960 6408 51 .4 0 8348 0.4794 1.3141 888 4 81s t , 1786 9 0 02466 0 1961 0 22021 64 .2 0.8403 0.4409 13092 83tt 4* .8 08454 04583 13041 824A
'
874 9 1839 0 0 02489 0 18737 0 21226 65 .I 0 8514 0.4414 1.2988 82 .2 08571 0.4364 1.2934 83 .4 4653 113 .2879 838A
, 838 8 2007 8 0 02595 0 15427 0 18021 67 .6 113 .8686 0.4134 1.2821 SetA 848 8 20599 844 0 2118 3 0 02625 0 14644 011269 6859 44 .0 0.8746 0 4015 1.2761 SeeA 848 8 2178 1 0 02657 0 13816 0 16534 69 .1 112 .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 7 111 .2567 etSA 23657 0 02768 011663 0 14431 71 .1 110 .2498 0010 880 8 884 8 243 .9 3773 1100 6 0.9064 0J361 1.2425 GMA 880 0 24981 0 02858 0 10229 0 13087 73 .1 1093 5 0 9137 03210 1.2347 00 .2266 872A sis t 2636 8 0 02970 0 08199 0 11769 749 2 328 5 107 .2892 1.2171 SPEA 880 8 27086 0 03037 0 06080 0till7 75 .2720 1.2006 este 0 03114 0 01349 0 10463 7682 29 .4 0 9447 0.2537', 1.19M 88 .0 0 9535 0 2337 l.1872 Sett 802 0 2934 5 0 03313 0 05197 0 09110 790 5 24 . % 34 0.2110 1.1744 802A 898 8 3013 4 0 03455 0 04916 0 08371 804 4 21 .2 0.9749 0.lMI 1.1591 0010 700 8 3094 3 0 03662 0 03857 0 07519 82 .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 8 317 .0 95 .0169 0 0076 1.1046 70 .4 93 .0329 0.0527 I0856 798 3 195 47' 32082 0 05078 0 00000 0 05078 906 0 00 906 0 1.0612 00000 1.0612 788.4P
.
'
' Critical temperature ] g:-
___ _ - _ _ _ _ _ _ _ _ _ .. _- __ _ _ _
- e l
O Table 2: Saturated Steam: Pressure Table Specifii volume Enthalpy Entropy Temp Sal Sa Sa Sa Sa Sa . Abs Pres Abs Pres LblSq I th/Sg i Fahr liquid Evan Vapor lic uid Evap Vapor liquid Evap Vapor h h s gg 5g p p I vg v it 'S 'I It 6 I
f
~
0 0000 2.1872 2.1872 000005 880865 32 018 0 016022 3302 4 3302 4 0 0003 1075 5 1075 5 108 .0425 2.0967 415 0 25 59 323 0 016032 1235 5 1235 5 21 382 10601 0 0925 1.9446 2 0370 0.50 8 58 79 586 0 016011 64 .5 47.623 1048 6 1096 3 0 1326 1.5455 1.9181 38 10 101 14 0016136 33159 33360 6913 1036I 1105 8
13 515 13 532 130 20 1000 9 113 .6094 1.8443 5I 16224 0016401 M8 0 016592 38404 38 420 161.26 982 I II43 3 0.2836 1.5043 I.7879 IS I 193 21 14.000 21200 26 182 426 799 180 17 970 3 1150 5 0 3121 1.4447 1.1568 14 598 0 016119 1 OJ 0016176 26 214 26 290 18121 %97 1150 9 0 3137 1.4415 1.7552 15 e ,
1 % 27 %01 1856 3 0 3358 1.3%2 1.7320 2 g 227 96 0 016834 20 070 20 087
,
250 34 13 1266 13 7436 218 9 945 2 116 .3682 1.3313 I.6995 s 0 011009 4 I 933 6 1869 8 0 3921 1.2844 1.6765 4e e 267 25 0011151 10 4194 250 2 923 9 117 .4112 1.2474 1.6586 5 .1562 71136 2622 915 4 117 .4273 1.2167 1.6440 tee le 5 1.6316 MI 302 93 0011482 6 1875 6 2050 27 .8 1180 6 0 4411 1.1905 is 3 1.I675 I.6208 88 8 se 3 312 04 0 011573 5 4536 5 4111 28 .1 0 4534 4 8953 29 .1410 1.6113 9 se e 32028 0011659 4 8779 4 4133 4 4310 298 5 888 6 118 .1284 I.6027 les 8 lese 327 82 0 011740 Ils e 334 79 0 01182 4 0306 4 0484 305 8 88 I.III5 I5950 Ile 8 3 1091 3 7215 312 6 817 8 1190 4 0.4919 1.0960 1.5879 120 0 ~
120 8 34121 0 01189 347.33 0 01796 3 4364 3 4544 319 0 812 8 119 I.0815 1.5813 IM4 130 0 325 0 8680 1l93 0 0 5071 1.0681 1.5752 14 les e 353 04 00180) 3 2010 3 2190 330 6 863 4 1l9 .0554 1.5695 15 .0435 1.5641 less 188 0 368 42 0 01821 2 6556 2i:138 34 % 0 0 5269 10322 1.5591 litI IIII 188 8 37308 0 018?? 2 5129 15312 346 2 8507 11 % 9 0 5328 1.0215 / I.5543 18 .6 0 5384 I.0113 1.5498 198 0 198 8 31753 0 0lR 13 23841 2 4030 38180 22689 2 2813 355 5 542 8 1198 3 0 5438 1.0016 1.5454 2ts 8 20s e 0 01839 385 91 0 01844 216313 2 18217 359 9 839 1 1199 0 0 5490 0.9923 1.5413 Ile s 211 0 364 2 835 4 1199 6 0 5540 0 9834 1.5374 229 8 228 8 389 88 0 01850 2 06119 2 08629 238 8 393 10 0 01855 I91991 199946 368 3 8318 12001 0 5588 09748 1.5336 238 8 I91769 37 .9665 1.5299 248 8 240 t 39739 001860 189909 184317 376 I 825 0 120 .5264 ISO 9 258 8 400 97 0 01865 182452 1114I8 319 9 821 6 120 .5230 268 8 288 8 404 44 0 01810 115548 111013 383 6 818 3 120 .5197 218 8 2III 40180 0 01815 169131 I65049 38 .5805 0.9361 1.5166 200 8
> 280 0 411 01 001880 163169 I.5135 29 a 298 I 414 25 00lflR$- I51591 859482 390 6 812 0 120 .9291 39' O 308 9 120 .5105 300 0
'
300 1 41735 0 01889 1.52384 1 54274 1.32554 409 8 194 2 1204 0 0 6059 0 8909 1.4964 358 0 358 8 43173 001912 130642 444 60 114162 116095 424 2 780 4 1204 6 0 6217 0.8630 I.4847 448 8 ese 3 0 01914
_ _ _ _ _ _ _ - - _
.
Specific Volume [Ilhalpy [;Iropy Sa Sa Sa Abs Press. , e Sal Sa Sa Abs Picst Temp Vapor liquid Evap Vapor LblSq i lb/Sq in liquial Ivap Vapos lic uid Evap .
P Fah vg is h, g hg s, s ,, s, p M p i vi v, s
1204 8 06360 0.8378 I.4138 450 0 101224 103119 431 3 167 5 14639 5ee g als e 456 28 001954 7551 12043 0 6490 0 8143 0 01915 0 90181 0 91162 449 5 1.4541 SW O 508 0 46101 143 3 1204 3 0 6611 0 7936 001994 0 8/183 0 84111 460 9 14461 les O 558 0 416 94 1203 7 0 6723 0 7738 0 14967 0 16915 411 1 732 0 EM S sas 8 486 20 0 02013 120 .9 720 9 Po Sit t 494 89 0 0/032 12018 0 6928 03377 1.4304 0 63s05 0 6 % 56 4916 180 2 788 8 503 08 0 07050 0 7210 1.4232 150 8 500 9 699 8 1200 7 0 7022 158 I S10 84 0 07069 058880 0 60949 14163 see t 509 8 689 6 1899 4 0 7111 0 7051 51821 00/061 0 54809 0 568 % 1.4096 e548 seeI 518 4 619 5 1198 0 03197 0 6899 525 24 0 02105 0 51191 0 53302 14032 See t 850 8 526 7 669 7 11 % 4 0 7219 06753
$3195 0 02123 0 41968 0 50091 1.3910 354 e 900 I 0 41205 534 7 660 0 11941 03358 06612 958 I $38 39 0 07141 0 45064 0 7434 0 6416 1.3910 lose 4 5426 650 4 189 /45 % 1191 0 03507 0 6344 13851 1850 8 0 40041 0 4???4 5501 640 9 Ilse 0 18588 550 53 0 02111 118 .6716 13194 0 02195 0 31863 040058 5515 6315 lies 8
'
556 28 1181 0 0 7647 06091 13738 115e 0 0 35859 0 38013 564 8 6?? 2 1200 t 1850 8 561 82 0 02214 1884 8 0 7714 05%9 13683 0 07237 0 14013 0 36245 511 9 613 0 1200 8 56119 0 7780 0 5850 1.3630 1250 8 0 34556 578 8 603 8 1182 6 125II 572 38 0 02250 0 12306 0 7843 0 5733 1.3571 1300 0 0 32991 585 6 594 6 1180 2 1388 5 57142 0 02269 0 30122 07906 0 5620 13525 1350t 0 31531 59 .3414 1400 5 0 30118 598 8 516 5 1115 3 1400 1 58101 0 02301 0 27811 0 5397 I3423 1458 8 0 28911 605 3 5674 1172 8 0 8026 145I B 59110 0 02321 0 26584 0 8085 ' O 5288 13373 1500 8 0 21/19 611 1 558 4 1170 1 1588 8 596 20 0 02346 0 25312 0 8142 0 5182 1 3324 1550 0 0 26601 618 0 549 4 11674 1558 8 600 59 0 02366 0 24235 0 8199 0 5016 13214 1600 8 0 25545 624 2 5403 1864 5 llH I 604 81 0 02381 0 73159 0 8254 0 4911 1 3225 165e 8 0 24551 630 4 53 e0 8 0 07478 021118 023601 636 5 52? ?
1100 0 613 13 0 4765 1.3128 1154 8 0 22113 642 5 513 1 1155 6 0 8363 1758 I 611 12 0 02450 0 20283 115? 3 0 8417 0 4662 1.3019 1888 e 0 19390 0 21861 648 5 503 8 1854 e 1888I 821 02 0 02412 494 6 1849 0 0 8470 0 4561 13030 18588 624 81 002495 0 18558 02I052 654 5 O4459 I2981 190s 8 0 20718 660 4 4852 1145 6 0 85??
1988I 628 56 0 07511 0 11161 666 3 415 8 ll 42 0 0 8574 0 4358 i 12931 1958 e 19588 632 22 0 02541 0 16999 0 19540 0 4256 12881 2000 8 0 18831 61 ees 8 635 80 0 07565 0 16266 0 4053 12180 2104 e 0 01615 0I4885 0 11501 6838 446 7 IJ305 0 8127 2388 8 642 16 1822 2 08878 03848 12616 270s e 64945 0 07669 0 13603 0 16212 695 5 426 7 flag I 406 0 til3 2 0 89?9 0 3640 12569 2300 0 655 89 0U/1/1 0 12406 0 15133 101 2 2308 8 384 8 11037 0 9031 0.3430 12460 2448 8 667 11 0 0/190 0l17R1 0 14016 119 0 2494 8 1093 3 0 9139 03206 12345 25000 66811 0 02859 0 10709 0 13068 731 3 361 6 25H 8 3316 1082 0 0 9241 0 2917 11225 2600 0 613 91 0 02918 0 09112 012110 144 5 2100 0 2604 5 3I23 1069 7 0 9356 0 7141 12091 2IM I 619 53 0 03029 0 08165 011194 1513 1.1958 20008 0 10305 110 7 2851 1055 8 0 9468 02491 2000 8 684 % 0 03134 0 01111 1.1803 2900 e 0 09420 185 1 2547 1039 8 0 9588 01215 2900 8 690 22 0 03762 0 06158 1.1619 3000 0 0 08500 8018 218 4 1020 3 0 9728 01891 3400 0 695 33 0 03428 0 05013 1.1373 310 .1460 3188 8 700 28 0 036MI 003111 0 0482 1.0832 320e 8 0 01191 0 05663 815 5 56 1 93 H 8 705 08 0 04411 0 0000 1.0612 329e.2*
0 0'.n18 O scul 005018 906 0 00 906 0 10612 3104 2' 70541 f . t)
.-
.c, , s . . . . . . .. .
.. . . .- . . . _ ._
i RRCS CHANNEL A 1 CHANNEL A LOGIC' B
,
LOGIC A INDICATIONS 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 i RRCS RESET OPEN RRCS RESET
'
,
'-
LOGC A RUN8ACK LOGIC B TROUBLE INITIATED TROUBLE
! TEST FAULT i ESSENTIAL l LOGC FAILURE
'
l
[ ,
~
10C651 CONTROLS
.
- FIGURE 1 **
_, .
GEK-83288
.
- N T / / -
C K)NO AUX TROLLEY ROD BIACK k% BACK UP HOIST FUEL HOIST HOIST AUX HOIS1 1hTERLOO LIMIT INTERLOCK IhTERLOOf INTERLOCX NO. 1
/
N ROD BLOCK BRIDGE BRIBE FAULT /
g REVERSE lhTERLOCK REVERSE 14CK0tTT
- NO. 2 S11)P NO.1 STOP NO,2 (
INTERLOCK STATUS DISPLAY l C O l -
-
Figste 3- Interlock Status Display Module l
- FIGURE 2 **
.
(
l 3-12 (
l
_ _ . . . - , - - , ._
_ _ . . - . - _- .- _ - . _. -
-
. .
- -
-
'
'
. . .
. . . .
.
.
TABLE 2 {
RPV PRESSURE INSTRUMENTATION .
REACTOR PRESSURE Transmitter Indicator LOC. SYS. STP 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
- PT-N090F CS/RHR INITIATE PT-N090J CS INITIATE PT-N090K -
CS INITIATE 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
,
PT-N403F -
RRCS ANALOG TRIP MODUE PT-3684A PI-3684A 650C PAM
, PT-36848 PR-36848 650C
! PT-7853A RHR INTERLOCKS
PDT-NO32 PDR-R61'3 650C (_
l
- FIGURE 3 ** q
__
- _ _ _ . _ _ _ . . - - . _ _ _ _ _ _ _ _ _ - . _ - _ _ . _ _ _ _ _ _ _ _
. _ . __ -
, . - -- .- -- --
. .
-
. .
'
k OEFINITIONS SECTION PAGE 1.0 DEFINITIONS 1-1
, ACT!0N.....................................................
1-1
'.2 1 AVERAGE PLANAR EXP05URE....................................
RATE................. 1-1 1.3 AVERAGE PLANAR LINEAR HEAT G G ERATION 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............................................ .
, .
1-2 1.8 CORE MAXINUM FRACTION OF LIMITING POWER 5ENSITY. . . . . . . . . . . .
1-2 1.9 CRITICAL POWER RATI0.......................................
(;. 1.10 05E E Ul m Eni 1- m ...................................... i-2 i
'
'
1. n I-Avt uGE Dis m EGRATION aERGy............................
1-2 1 1.12 EMERGENCY CORE Co0 LING SYSTEM (ECCS) RESPONSE' TIME.........
1-2 l
,
1.13 END-OF-CYCLE RECIRCULATION PLMP TRIP $YSTEM RESPONSE TIM l 1.14 FRACTION OF LIMITING POWER DEN 51TY......................... 1-3 l 1.15 FRACTION OF RATED THERMAL P0WER. . . . . . . . . . . . . . . . . . . . . . . . . . . .
1-3 l 1 15 vau FREQUENCY 1- l wome a aanN0TATION x ........................................ s .:o e. nee..s oe en snrm o 1- l
' IDENTIFIED LEAKAGE.........................................
1.gI50LATION SYSTEM RESPONSE TIME. . . . . . . . . . . . . . . . . ; . . . . . . . . . . .
1-3 l 1.W.,uMITING CONTROL R00 PATTERN. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 .. l 1-3 l 1.(LINEARMEATGENERATIONRATE................................
i-4 -
14toGICsTSTEMruuCTIONALTtst...............................
,
-
u. ,wm .__. - -. _ . .
gs w i ne. rw. nan-
-.-swn...............................
er
--
..,_ n,
[ ( *
g.11 M6n06AD *! TM fs**u .gMININuMCRITICatPOWERRATI0...............................
I-4 1-4 H l-V s,st, ONasrg Deed CAcuursew *taaw *'
a 5 I -wmev e i - - s n, .
-g.- - ,
s '
Aev, .2,
%=*t. c.*stw
- .
. . , . - . . - _ , - . . - _ . . , , - - . - - - . , ,_.n___-,_ , , - , . , - - . _ - . . - - - . . - _ , . , , , . . , , , _ . - - - - , _ _ _ . . , . - - . - - . . - - - - - --
- _ - - . . _ __
.
INDEX ,
I
\
DEFINITIONS
.
SECTION PAGE DEFINITIONS (Continued)
s1 1-4 1.y OPERA 8LE - OPERA 81LITY..................................... ll as 1-4 0 1,.25 OPERATIONAL CONDITION - CONDITION.......................... *
-y f.nt 0 Nun *~AL Mabe -M*DG 1-4 l 14PHYSICSTEST5..............................................
1-4 a 1.#s PRES 5uRE* *e0uN0ARygae * $.................................
!g ,s,1 .ae wu a~n*6 - m s-
-
-
IJg/RIMARY CONTAlmENT INTEGRITY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ll sr 1-5 ll 1.js'RATEDTHERMALP0WER........................................
s .)t REACTOR PROTECTION SYSTEM RESPONSE TIME.................... 1-5 l ,
rr suur 1-5 U 1.)tr REPORTASLE-4GGtHHtfMCE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
se 1.)rR00DEM51TY................................................
1-5 l r
se mcTen. woms
- ( -
.
) 1.x === me,:m(seco~o4Y ce a' erat)
IwrEGRITY............................ 1-s il
d so 1-6 0 IJS SHUTDOW MARGIN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
. , - - - ,
- STAGGERED TEST SA5!S.......................................
1-6 I yr .
1-6 l 1.)5THERMALP0WER..............................................
.. . . . . . . . _ . . - . . ._,,
., t
- s..,,,,,...-
.
- .- -.. ......................................
1.kTURSIMEBYPASSSYSTEMRESPONSETIME........................
1-7 l
.
'
,
,,
1..F1 UNIDENTIFIED LEAKAGE.'...................................... 1-7 I
'
MW 1-8 I i TABLE 1.1. SURVEILLANCE FREQUENCY NOTATION. . . . . . . . . . . . . . . . . . . . . .
TABLE 1. 2. OPERATIONAL C02ITIONS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
1-9 l l
~~
\ t.1I $ITE 6** bha 1 , ,,, $
l. 4 L COUbe fit ets e N ;,, g JamAct eserce s .(,
L l.43 -
" f.4 9 U M rsissert o M e n m r u nea as na.or rasaresur eysrsn
,'Q
. /. 41 t-7 l. f * VbtTIM
..
Rev.2- 3
- -. . ae f M Mw g--- .i7gn Wu 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
.
. .
.
. 8
.
. p
.
p
-
.
_ . , _ _ , , , ,,,
- ... .. , .
het. c* ass
._ a
- . - . _ - - - - -_ -____ _ ___ _ - _- _ - _ - - - - . -
_ -_ _ - - . _ - ---. - . --- - -
.
.
10 JAN 1983
\
19 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE .
PAGE SECTION
_
3/4 0-1 1/ APPLICA81LITY.............................................
.
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4 1-1 3/4. SMUTDOWN MARGIN........................................
3/4 1-2 3/4. REACTIVITY AN0MALIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . .'.
.
.
.
3/4. 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 i
( Control Rod Scram Accumulators.........................
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. 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 STAN0tY LIQUID CONTROL SY5 TEM. . . . . . . . . . . . . . . . . . . . . . . . . . i 3/4.2 POWER DIST113UT10M LIMITS RATE............. 3/4 2-1 3/4. AVERAGE PLANAR LINEAR HEAT GENERATION 3/4 2-f 3 1/4 L * APRM SETP0!NTS......................................... *
3/4 2-EY '-
RATI0...........................
3/4. MINIMUM CRITICAL POWER 3/42ftr l 3/4. LINkARHEATGENERATIONRATE............................
{
iv
- + " " ' r - )
% # T. C A tt w
-_, _ _ . _ _ _ _
_ _ ___ _ _ . _ _ _ _ _ _ _ _ . _ . _ _ . _ _ _ _ _ __
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l 10 JAN 1983
/
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jgEl
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_
LINITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS
_
PAGE SECTION
-
3/4.3 INSTRUMENTATION ,
REACTOR PROTECTION SYSTEM INSTRUMENTATION............
3/4 3-1 ,
l 3/4. l
'
3/4 ISCLATION ACTUATION INSTRLsENTATION. . . . . . . . . . . . . . . . . . 3/4 3-5 INSTRUMENTATION...................................... 3/4 3- Fr ll 3/4. EMERGENCY CORE COOLING SYSTEM ACTUATION
.
,
3/4. RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION w '
ATV5 Recirculation Pump Trip Systes Instrumentation.. 3/4 3-46 Il l
End-of-Cycle Recireviation Pump Trip System w Instrumentation...................................... 3/4 34 ll .
sy 3/4. REACTOR CORE ISCLATION C00 LING SYSTEM ACTUATION 1NSTRUMENTATION...................................... 3/4 346- \l t
-
( ,
u4. 3. . con 1=t =0 it0C n NSTRuME m uCN.................... u m ii
. Y i 3/4. MONITORING INSTRUMENTATION -
i Radiation Monitoring Instrumentation................. 3/4 3-57 car ll 7s Seismic Monitoring Instrumentation................... 3/4 3-44- ll
se l
> Meteorological Monitoring Instrumentation............ 3/4 3 e ll
- Remote Shutdown Monitoring Instrumentation. . . . . . . . . . . 3/4 3-7
'
'
rAccident Monitoring Instrumantetton.................. 3/4 3 d ll
~
' Source Range Monitors................................ 3/4 34 ll
" Traversing In-Core Probe
.
System...................... 3/4 3d ll l
S h r' = (cd .' - : - h) ";'.;d = !je r. . . . . . . . . . . . . . :/0M ll
,
l C;,*.i ",,;... %;; " ;.0^ ............................
,
' 't ' " , ll
,
3/4 3 ll 4 Fire Detection Instrumentation....................... sa '~
3/4 3-06 ll e w4 3-sy H Ase-Part
-
m. .WDetection%4 ME Svstas........ e G h..* h........ F3 /4.3.8 TURRINE OVERSPEED PROTECTION SY5 8TEM. . .
Framw& Tea s /MAmt 'NRanet 'TRW mettM 3/43-99%
3/4.3.3 ";." ;Y:T;.": ACTUATION INSTRUMENTATION............... hceMd** */q s 11
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.
I yQew hs tNA Wahg v
"" C: (r.'t) -
r- 6 ~ \s-tsee esea __ _ ._ _ . -
_ _ _ _ _ _ .. _ _ _. ___ _
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>
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- IEE
, LIMITING COWITIONS FON OPERATION AND Sutvf!LUWCE REQUIREMENTS
'
'
BElE PAGE 3/4.4 REACTOR C00VWT SYSTEM -
^
3/4. 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. SAFETY / RELIEF VALVE 5
'
Safety / Relief Va1ves................................. 3/4 4 4 7 *
grvr4 _, _ ,. __ _,. . . .-.... ,._ , ... _ , o _ . . ___ __ .___ . .. .
,_. , m ,__
- .._..... . . . . .. . . pg ,
.
. .
3/4 REACTOR COOLANT SYSTEM LEAXAGE
-
Lankage Detection Systans............................ 3/448 N
,
( -
Operational Laakage.................................. 3/44-9-
! . - ,. 4 CNENISTRY............................................ 3/44-S 4 3/4. SnCr n C 4CTivm. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3/4 44
3/4. PRES 5URE/ TEMPERATURE LIMITS I M i Reactor Coolant 5ystas............................... 3/44-4-
'
Reactor Steam 0eme................................... 3/44-N
'
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v4. MarN STEAM trNE r$0t4720N vAtvE5..................... 3/4 4 4 v4. STRuCTuRAt m EGR m ....:............................ u4 4 4 3/4. 'EstauAL MAT REMhWAL
.17 3/4 4-95-
,
l Not Shutdeun......................................... ,
Cold Shutdown........................................ 3/44M 3/4.5 D4;RGENCY Coat COOLING SY37Dt3 3/4. ECC5 - OPERATING..................................... 3/4 5-1 .
3/4. ECCS - SMLtig0WN...................................... 3/4 5-6 i b 3/4. SUPPRESS ! 0N CNNGE R. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 5-8
_ ... ,_
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,-
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- ----+,wwm _ - , , . . - , , , - . _ - - , - . - ~ - - -
. _ _ _ _ _ _ _ _ _ _ _ . - _ .
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( INDEX LIMITING emnITIONS FOR OP(RATION AND SURVi!LLANCE REQUIREMENTS i
'
PAGE SECTION 3/4. 6 CONTAf stNT SYSTEMS
'
3/4. 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.........
se a A _=_ l 3/4 6 d \
"-irrj trn'--92F=:E: f r "r:=r'nt'= !j:tr 2/' 5-12-
l 3/4. DEPRE55URIZATION SYSTEMS
.to 3/4 6-+t
+
'
Suppression Cham 6er..................................
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. PRIMARY CONTA! MENT 15CLATION VALVE 5. . . . . . . . . . . . . . . . . 3/4 6-49 ,
l 3/4. 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 3/4. ;~0"" "7 C*". /T.T anactog smummt. (evet a"* '*d*)
tLu encr, h ,.43 d a.nn . (saweans e-a.=** Mity...................... 3/4 6 d
-: t,J ntegr m a yw=,%[re< =5 -+ <> ion 40ampersl v1
- t. Automatic Isolat .
.
nxcrj)_rm !. n ........................................... 3/4 6- e Fih%,ReelmuMian.ead O r n, __ Trn t n .,n r.......................... /4 ve,MMM spbg Opjs) 6-49- 91
{ .
..
- . . . .,.. -_, ,,,
ON
.
,-------,-----,--~-------w--_-- - . . _ , . , . _ , - . _ . . , , - _ _ - _ _ . , , __.____,-.--.___,,.-,,,--.-.-w,----.- _ --- - . - - , - , - - . . - - . - - - - - - - - , . - .
- - .
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J If!ILE .
LDu ring CE r-!TIONS FOR OPttATION AND SUtytiLLANCE tt00f tDENT5 SECTION . Esit
'
CONTA!WENT avaiu6 (Continued) ,
3/4. 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
-
.
l 3/4.7.1
- . Mm, SERVICE PBTUF W5 A SYSTDts = h m_.__. _ . _. . __ --- . Systas. [s. .A.c.4. . . 3/4 7 1
. - - _ . _ _ _ . - _ _ -
Plant Service Watar 5ystas........................... 3/4 7-3 l
vi ti-ta neat 5 i na. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . u. 7- 5
'
( .
3/4 7-6 .
3/4. CONTROL 200M EERGENCY FILTRATION SYSTEM.............
3/4. FLOOD PROTECTION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-9 M
3/4. REACTOR CORE ISC LATION COO LING SYSTEM. . . . . . . . . . . . . . . . 3/4 7-19
, 3/4. USS ERS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
u4. SutE050utetCONT=tmTION..........................u47-3
-
-
3/4. FIRE SUPPRESSION SYSTEMS so
Fi re Suppres sion Water 5ystas. . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-at I '
Spray and/or Sprink k e Systans...................'.... 3/4 7 M
................................ v4 *
Ca, 5 . . . . . . . . .
> > - - . _ _ _ _ . . . . ........ , . ..
. . . . . ,,._- . ...........................
Fire Mese Stations.................................
. 3/4 7 M u...... .,_n v_ m n _. m >___., __> m m ... u m V WW y w- -- - - - - - - - - - . . . . . . . . . . . .
, .. . .. .g -. .-- - -
SAtates. Ps ofaavisees l 3/4. FIRE 443EHeSettfHr. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7
. .. . 6, i-I
. , . . .
. . . ,
..
_ _ . ...-...._----.m.-. ....- .-...........................
- - , , ,
5YSTEM........................... 3/4 7- M I 7 3/4.7.14 MAIN TURS!NE SYPA55 .
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.M .9 9
- ... haemIA%,
.- ,
,
__- . , , , - - _ - _ _
,,,-- ,----,--,- - , - - - - - , - - _ _ . - - - - _ - - - ~ - - - _ _ _ _ - _ - - _ . _ _ _ _ - _ _ _ _ _ _ - _ _ . - . _ - - - -
..
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.
10 JAN 1982
.
s INDEX
,
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS
, .
. - PAGE SECTION
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3/4.8 ELECTRICAL POWER SYSTEMS 3/4. A.C. SOURCES A. C. Sources-0perati ng. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 8-1 A. C. Sources-Shutdown. . . . . . . . . . . . . . . . . . . . . . . . . . .
3/4. ,
0.C. SOURCES 3/4 s h j I D.C. Sources-Operating...............................
.
!
' s 1 3/4 4- W 4 'i g Sources-Shutdown................................
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3/4. 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. .
ELECTRICAL EQUIPMENT PROTECTIVE DEVICE ~
A.C. C h;.it; r!..f i N != j C: t i z at............. 3/' H S f
J u, Prisary Containment Penetration Conductor Overcurrent P rotecti v e De vi ce s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 8 JP-MotorOperatedValveThMe cNee1.3d Protection.-httt*3/. SM, ,-
1 Overl zi er.+..+,_ ... e- - 3f nshe errded alve. Reactor Protection Systes Electric Power Monitoring.. 3/4 8-eess23 3/4.9 REFUELING OPERATIONS .
.
3/4 9-1 .
j 3/4. R E 0R IC0E' SWITCH.................................. !
3 *
InsinGPENTAT10N...................................... 3/4 9-3 i
3/4. "
3/4. CkTR0L 800 POSIT 10N. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-5 3/4. DECAY TIME . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9- *
.
3/4. ComuM I CAT 10N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9 .
8 . "
me 3/4. REFUELING P LATF0RM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9
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O
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6 d \ M
van emw
, , - ._ - -..-..... - . - .___ .- , _- _ _ _ - - - _ _ _ _ . , . . - . , _ _ . - - _ , - _ _ , _ _ _ _ ...___c , , _ _ _ - _ _ . _ _ -- _ _ - ,
.___ __ .- . .- .-
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l
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- *
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-
INoEx (
LIMITING C04 !TIONS FOR OPERATION ANO SURVEILLANCE REQUIREMENTS
.
i SECTION I REFUELINGOPERATIONS(Centinued) .
- 3/4. CRANE TRAVEL - SPENT FUEL * STORAGE P00L. . . . . . . . . . . . . . . 3/4 9 9 3/4. MTER LEVEL - REACTOR VE55EL. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-10 3/4. MTER LEVEL - SPENT FUEL STORAGE P00 L. . . . . . . . . . . . . . . . 3/4 9- 11
'
3/4.9.10 CDNTROL 200 REMVAL i
Stagle Control Red Remova1........................... 3/4 9-12
-
>
- 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
{ 3>
! 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
,
- TurNim STARTU,5. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . u4 i G-i -
.
. v4. i ;
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!
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.P
!
$
i
,
.
,
- m
... s.....,m
,,
-
, 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 Oose.... .j..............................................
Liquid (WesteTreatment................................... 3/4 11-#*f Liquid Holdup Tanks...................................... 3/4 Il f f *
O.;;i;;1 Tr; C :nt N .::.......... ..................... 3/4 ti 0
.
3/4 11.2 GASEQUS EFFLUENTS Dose Rate................................................ 3/4 11-9 l Dose-Noble Gases......................................... 3/4 11-13
~4 Dose x.d y':.gt p +*um: rte: !:te:
-
.s.. ;h . . isk$l!P- h hh Fem,
. . - .
nd " fi:n.: lid: Oth;r th:n 9: !: C2:::................ 3/4 11-14 Gaseous Radweste Treatment............................... 3/4 11-15 Ventilation Exhaust Treatment............................ 3/4 11-16 Explosive Gas Mixture.................................... 3/4 11-17, -} i Main Condenser...........................................
3/4 11 Me# '
Mark I or II Containment................................. 3/4 11-Je' /T
-
.
V.4.11.p as 3/4 11.3 SOLID RADIDACTIVE WASTE.................................._, ,
3/4 11.4 TOT A L D0 5 E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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......................................... .
'
3/4 12-1Y
.
3/4.12.3 INTERLA80RATORY COMPARISON.............................. ,
x! 6v. . H#s ea'Ek- WM Z ".73-2 ff*
.
.. .. -
, - . . _ , _ _ - - - - - . . - , _ . ,
,- ,, - - - - - . _ _ - - - - - - - - - _ . . . . _ - _ . . . - , , , - - _ -- --
s F
.
.
i .
l (,
15H
'
- 9=s S.$,l
,
.
M 3/ APPt.ICASILITY................,............................ S 3/4 0-1 3/4.1 RDCTIVITY w iisui. SYSTDe$
l 3/4. SIEff00MI 1448 GIN. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 1-1
!
3/4. MACTIVITY Ancit4 LIES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . S 3/4 1-1 3/4. CDerTROL 8005.................................... 8 3/4 1-2 1 3/4. 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 .
"
l .
3/4. AVERAGE PLANAR UleGR NEAT GDIERATION 8 3/4 2-1 RATE..........................................
! 3/4. AMIM SETP01h T5. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 2-2 ,
i ('
d 3/4. ItDIDRM CRITICAL POWER RAT 10. . . . . . . . . . . . . . . . . . . . . B 3/4 2-4
. u 2-5
u = = .AT = = uT ;
-
u4. TE......................
i
'
3/4.3 INSTartNTATION .
. .
l 3/4. IIDCTOR PROTECTION SYSTD4 INSTituMD(TATION........ B 3/4 3-1 l 3/4. ISOLATION ACTUATION INSTituMDf7ATION. . . . . . . . . . . . . 8 3/4 3-2
.
3/4. EMERGDeCY CORE C00UlIG SYSTDI ACTUATION
,
'
,
DesTRWetTAT10N. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 3-2
. t
.
,
. 3/4. REC 1RCULAT10N PWIP TRIP ACTUATION
'
. DesTRsWITAT10N. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 3-3 3/4. ADCTOR Coltf !$0LATICII CD00NG SYSTDI ACTUATION Ut572uMDITAT10N. . . . . . . . . . . . . . . . . . . . . . . . I 3/4 3-4 i l
3/4. CONTROL A00 SLOCK De5TapetT AT10ll. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 3-4
.i
'
[ .
,-
i, f (, . .
.
45-s*s 4awnt4) 'mii l
! net estsA i
!_ -
'
- .- -- - -
. . _ _ _ . - - . _ _. _ _. - __ - . _ - - . . - _-
.
. . .
. .
152 .
-
aAsrs
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IEE
EUlE" *
' INSTRUMENTATION (Continued) ,
3/4. 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
! - Assets Shutdown Monitoring Instrumentation...... S 3/4 F5 l
'
Acci dent Monitori ng Instrimentati on. . . . . . . . . . . . . t 3/4 F5
. Source Range Monitors........................... 8 3/4 3-5 Treversing In-Care Prete systas................. B 3/4 F5
l '
T ' r M ^ _- '-) L^ r ':: ",; M ......... 0 *,'t : ", ,
'
- " ' 't T n' t M --c':r 1 :.....................
- F ire Setecti on Instrimentati on. . . . . . . . . . . . . . . . . . 8 3/4 F(
5 ..... 83 l ( * e-Part i 6. bW DetectionMstas.. ae4 g L....MWmbe/4a-G 3-5 3/4.3.s m!NE OVER5 Pits Pn0TECTION sysTIM............. e 3/4 3-6 l .
g _
%) ven.mtem/m wasine tw wem
"'^^'" S 3/4 3-6 3/4. ""'T !
!
-
h_ah t. (menems ACTUATION th k Me=%hk- INSTRUMENTATION....
3/4.4 atAcTon c00LANT SYSTEM .
.
!
B 3/4 4-1
! 3/4. RECIRCULATION SY5 TIM............................
g 3/4 /4. SAFETY /REL!tF VALVt5.........................,,,
. ,
3/4. REACTOR C00LANT SYSTEM LEAKAGE
. .
Leakage Setection Systans....................... I 3/4 4-2 -
,
I 0perational Leekage............................. 8 3/4 4-2
.
8 3/4 4 2 3/4. CMEN13TRY.......................................
,
,
S 3/4 4-3 3/4. SPECIFIC ACTIVITY...............................
5 1/4 4-4 3/4. Ptt3suntntMPERATURE LIMIT 3. . . . . . . . . . . . . . . . . . . . .
l 3/4. MAIM 57 TAM LINE ISOLATION VALyt5................ 8 3/4 4-5 -
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3/4, INflGR1TY................*............ 8 3/4 4 5 STRACTURAL 3 3/4 4-5
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3/4. RE51 DUAL MEAT MMOVAL. . . . . . . . . . . . . . . . . . . . . . . . . . .
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SECTION 3/4.5 EMERGENCY CORE COOLING SYSTEMS
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8 3/4 5-i
. 3/4.5.1/2 ECCS - OPERATING and SMUTD0WN...................
3/4. Suttatss!ON CWWG ER. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 5-2 3/4.6 CONTAINNENT SYSTEMS 3/4. PRIMARY CONTAI N NT
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Primary Containment Integri ty. . . . . . . . . . . . . . . . . . . 8 3/4 6-1
, Primary Contai nment Laskage. . . . . . . . . . . . . . . . . . . . . 8 3/4 6-1
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- Primary Contai nment Ai r Lacks. . . . . . . . . . . . . . . . . 8 3/4 6-1
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itSIV kW 8 3/4 6-1 -
'_:: ;;s ....A $32tas.....................
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l Orywell and Suppression Chamber Internal
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}' Pr..sure......................................
Orywell Ave rage Ai r Temperature. . . . . . . . . . . . . . . . . 8 3/4 6-2 l 8 3/4 6-2
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Orywell and Suppression Chamber Purge Systas.... 8 3/4 6-2 l 7. E cy C =' nnt, n -w ' - t S m e-9-t*= . n,
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- ,..__ _ ~ ...........................................
3/4. DEPR155URIZATION SYSTDt5. . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 /4. PRIMMtY CONTAlleENT ISOLATION VALYts......'.$..
3/4. VACULM RELIEF. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 taAsm suu.wa, poucaat e.easead t
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1/4. OZ : ^ ^ 7 Z; " :Z.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 6- 5
- 8 3/4 6-6 l ,
l 3/4. PRIMNtf CONTA!!sENT ATMO5PHERE CONTROL.......... .
3/4.7 Puurf swivis 3/4.7.1 SERVICE wilt 5YSTDts. . . . . . . . . . . . . . . . . . . . . . . . . . . . B 1/4 8 3/4 7-1 -
3/4. CONTROL ROOM DERGENCY FILTRATION -
SYSTEM........
PR0TICTION.........'........................ I 3/4 7-1 i
3/4. FLD00 l .
8 3/4 7-1 3/4. REACTOR CORE ISCLATION COOLING SYSTEM...........
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PLANT SYSTDes (Continued)
3/4. SW88ERS........................................ 8 3/4 7-2
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3/4. SEALED SOURCE CONTAMINATION..................... 8 3/4 7-3
l 3/4. FIRE SUPPRE5510N SYSTEMS........................ 8 1/4 7-4
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3/4. FIRE _em ._ ^^ Z#waaviews
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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
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I 3/4.8.1, 3/4.8.2 and *
3/4.8.3 A.C. SOURCES, D.C. SOURCE 5 and ONSITE POWER
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DISTRISLtTION SY5TEMS..... ~ ..................... 8 3/4 4-1 i . 3/4. ELECTRICAL Equ! PENT PROTECTIVE DEVICES......... 83/48-3' l l
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(d *
3/4.9 REFUELING OPERAT108:5 m ...i REACTOR =0E $wiTCN. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 m 9-i l 3/4. DISTRLMENTAT !0N. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 9-1 3/4. CONTROL M0 P0$! TION.....................y...... 8 3/4 9 1 4 3/4. DECAY T1NE......................................
I 3/4 9-1 I 3/4. C05UNICATIONS.................................. I 3/4 9-1
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3/4. REFUELING PLATF38Bt.............................. 8 3/4 9-2 i
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. 3/4. CRANE TRAVEL-SPENT FUEL STORAGE P00L............
8 3/4 9-2
< 3/4.9.8 and 3/4.9.9 WTER LEVEL - REACTot VE5SEL ,
8 3/4 9 2 and WTER LEVEL - SPENT FUEL STORAGE P00L.......
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3/4.9.10 CONTROL B00 REMVAL. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 9-2
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3/4.9.11 RESIDUAL NEAT REmvAL Am COOLANT CIRCULATION... 8 1/4 9-2
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1 m5 e 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
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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 3/4.u.1 U Qu!D EFFLUENTS........................................ 3 3/4 u-1 u4. .staus urtu NTs....................................... . u u-2
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3/4.u.3 SOLIO RADI0 ACTIVE tiA5TE.................................
8 3/4 u-S 3/4. '1lTAL
. 005E.............................................. S 3/4 11-6
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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 N m u.ciarony c = *=rsoN ,.oc . u4 u-2 ( u4. ......................
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, AGE SECTION 5.1 SITY . 3 .
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5-1 Eastusten Ares.............................................
Law ,eputatten Zone........................................
F1 5. 2 CONTAI M 4T 5-1 .
Conf 1geretten..............................................
5-1 Design Temperature and Pressure............................
Rea.c.b _..4*d.&.>
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5-1 5. 3 RDCTOR CO#E
5-4 Fuel Assamm11es............................................
5-4 Centrol And Asseme1tes.....................................
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S.4 ttACTOR C00LANT SYSTEM l
-
J oesign ,ress..e .n. 1e e ,.tu,s............................ .
,ei .e.....................................................
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5.5 NETiet0LOGI CAL TOWER (OCAT10N. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
5.6 Futt STO4AGT PS C ri ti ca 1 1 ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
F5 ,
0rainage...................................................
F5 Cepecity...................................................
S.7 cgus0NENT CYCt!C On TRAN5ttMT LIMIT........................
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ADMINISTRATIVE C0hTROLS SECTION PAGE
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. RESPON51SILITY............................................ -
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6.2~ ORGANIZATION.............................................. 5-1 l 6.,.1 .nsut.............................................. 5-1
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6. 2. 2 uM n srAn. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1
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6.2./ $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 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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6.5 ttVIEW AND AU0!T.......................................... 6-7 l r,Txn,o,w __ eeu_stions esvww cemaims -(reac)
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_ .u.= .awwr w-,. ...__.,........................... 6-7
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FUNCTION ............................................ 6-7 .
Col #051 TION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-7 .
- -
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ALTERNATI$........................................... 6-7
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- IEETINE FREQUDICY . . . . . . . . . . . . . . . . . . . .,r . . . . . . . . . . . . .
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SS$1M...................................~............ 6-5
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RESPONSIGILITIES ...................'................. 6-8 s.ac. a. .. n.cm
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t ADMINISTRATIVE CONTROL 5
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COMPOSITION ......................................... 6-f j
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' -WtTIMA566: . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . =9u0"*
l CONSULTANTS.......................................... Gd gs OFF- Sift hvtew 6deue
=0EG54MS-MEQW6MGV ................................... 6-M-
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6.7 sArtTV Lia r vtotATtoN....................'................ .
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6.8 PROCEDutt$ AND P90 GRAM 5..... ............................ GM
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6.9 atPotTING REQUIREMENTS.................................... ~6-M l '
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'STARTUP REP 0RT.......................................
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- * . . . . . . . 6-B Immt.Y OPERATING MPORT5. . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-H
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6 Is, 6. son SPECIAL REP 0RTS................... ................. 6- M -
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' 8.H RADIATION N iisiiON P90GRAN.............................
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1 9.11 I ADMINISTRATIVE CONTACLS
- LIST OF FIGURES PAGE FIGURE .
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ORGANIZATION................................. 4-3
> 6.2.1-1 0FF51TE
ORGANIZATION................................... 6-4 6.2.2-1 UNIT
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LIST OF TABLES
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PAGE M
6.2.2-la MINIMJM SHIFT CREW C0fe051 TION - SINGLE UNIT 5-5p FACILITY.............................................
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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 me ~
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 require .
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: -
~ 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 />, :
. 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 At least. COLD SHUTDOWN within the subsequent 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ,
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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 Specification This Specification is not applicable in CPERATIONAL CONDITIONS 4 or l 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 Specification .
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JA)s CAT A 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: 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 />, 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
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, At least COLD SHUTDOW within the subsequent 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> :
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This specification is not appitcable in OPERATIONAL CONDITION 4 or .
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s 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERACE PLANAR LINEAR HEAT GENERATION RATE
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LIMITING CONDITION FOR OPERATION 3. All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLNGRs) for each type of fuel as a function of AVERAGE..-4 PLANAR
- . .i-2. EXPOSURE shall not exceed the ll limits shown in Figures 3.2.1-1. 3.0.'-2, APPLICA8ILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 325}% of RATED THEMAL PCWE 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 hour SURVEILLANCE REQUIREMENTS
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4.2.1 All APLHGRs shall be verified to be equal to or less than the limits
..t 0.0.1-:: l determined from Figures 3.2.1-1; 3.2.1-0, 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 />, 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 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 operating with allMITING CONTROL 200 PATTERN for APU( ,
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POWER DISTRIBUTION LIMITS l
\
3/4.2.2 APRM SETPOINTS -
. .
LIMITING CON 0! TION FOR OPERATION i
3.2.2 The APRM flow biased simulated thermal power-gscale scras trip setpoint 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 51(0.66W+154j%)T
$ 1 (0.65W + 1511%)T Sgg 1 (0.66W + (42)%)T Su i (0.6EW + $45{%)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 = divided Lowest value of the ratio of SFRACTION OF RATED THERMAL by the CORE MAXImM FRACTION OF LIMITING POWER DEM51TYKcmtec?
j *) f=1, ff:fid bj 2: ';**F 21:!=d (f::
R si ; f y "!=:'t =m).,
- f f=1 (?.'?)
' ?:r (" T is applied only if less than L
,- er equal to .
OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or *
APPLICABILITY: l equal to $25]% ef RATED THERMAL POWER.
l 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 as above determined, shown initista in theaction corrective Allowable withinValue coluen for 15 minutes b' 5 er 5 and ad%s,t 5 and/or 5 to be consistant with the Trip setpoint values * lg 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 POWER within the next 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SURVEILLANCE REQUIREMENTS I
4.2.2 The (FRTP and the CMFLP01 ('""F) shall be determined, the value of T calculated, and the most recent actual APRM flow biased simulated thermal ..
power, upscale scres and flow biased neutron flux-escale i At least once per 24 haves,
~ Within 12 heves after completten of a THERMAL POWER increase of at 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 ??)_
greater than the (FITPG (t:f p T) ir t ; m r = = :!:- 3 APRM setpoints, the
"With (CMFLPDF
""" ;f *^O('"**) * TTfi .T '", rather than n er equal
, .,; ^ :
AP94 gain any be adjusted such that
.
eldings the are APRMM*3djusting greater the 100% times ICMFLPC) W T). provided that the adjusted APRMAre ding does
,
to """S and a notice of adjustment is posted on
'
not exceed 100% cf "T *"rli U the reactor control pane gepe c.a. tau, , ,. , . . .
g _ ,,, ;- p) 3/42-) s .:.. . .
. . . - , . _ _ _ , - - - - - . , , - _ _ _
. .-
. -
%
.
.
. PC'.ER O!STRIBUTION LIMITS
.
-
MINIMUM CRITICAL PC'4ER RATIO (0;;i; :?-00Y" 0;t': ^}
' 3/4. LIMITING CCNDITION FOR OPERATION 3.i.3 The MINId'M CRITICAL POWER RATIO (MCPR) " rshall
- -f be equal'^E"
""?". to or greater
than h ti """9 limitp,et 'nd'::t:d ::r:
m_
,_r,.._!_crfMCPR[__2
. . _ . . E P r h *S D .9..*5. U $"'i h M.~ k
.
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).
OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or *
APPLICA81LITY:
equal to (253% of RATED THERMAL POWE ,
ACTION:
" O Oc ;nd-d' Vj:1 ;;' ::1:tt: ; n tr!; :pt: ' :;:rd!: ;:-
( ation 3.3.4.2, operation may continue and the provisionsa hour, Spe Specifica .0.4 are not applicable provided that, with R be equal to or greater than b f and MCPR, MCPR is deterni C-RPT inoperable curve).
as shown in Figures 3.2.3- 3.2.3-2 e per Specification 3.7.9, $
With the main turbine bypas inop ( d the provisions of ication 3.0.4 are not ned to be equal operation may contina that. yithin one hour, MCPR is de applicable p . -1 and
,
to or r than both MCRPy and MCPR, as shown in Figures
- =a--=M= eu== 3
..".0-2 by W --?- he'-- W u w h A- -
.
,
i K With MCPR 1ess than the ;;?f:$1: MCPR limiti;tr '- F.gurep j.'2.3-1 2.2.2- initiate limit within the required corrective action within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or within 15 minutes reduce THERMAL powerandto restore less MC than 525)% of RATED THERMAL POWER within the next a hour SURVEILLANCE REQUIREMENTS
.
... . '
,.
4.2.3 MCPR shall be deterstned to be equal to or greater than the applicable' l MCPR limit detarsined free Figure 3.2.3-1:
. At least once per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> * 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 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 MCF ' " *
. . , .
( ,
M ee t c a t E /4 2- R Z 075 ( 7 L' ) 3 0 MAR Sf3 .
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1 0 JAN 592 POWER DISTRIBUTION LIMITS
! '3/4 ..2 4 LINEAR HEAT GENERATION RATE .
LIMITING CONDITION FOR OPERATION
'
3.2.4 The LINEAR HEAT GENERATION RATE (i.HGR) shall not exceed (13 OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or APPLICA8ILITY:
equal to 32!(% of RATED THERMAL POWE .
.
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 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> *
s SURVEILLANCE REQUIREMENTS
- .
4.2.4 LHGR's shall be dotarsined to be equal to or less than the limit:
d 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 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 l
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 operating on a LIMITING CONTROL RCD PATTERN for LNG .
D e
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soweseonr~s connorm mu. uconnsua #~#'**""'**
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.
'
3/A.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3. As a sinious, .the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTIO RESPONSE TIME as shown in Table 3.3.1- APPLICABILITY: As shown in Table 3.3.1- ACTION:
. WiththenumberofOPERA&EchannelslessthanrequiredbytheMinimum OPERA 8LE Channels per Trip System neuinment for one trip system, place the inoperable channel (s) and/or that trip system in tne tripped condi-tion * within ese hours. The provisions of Specification 3.0.4 are not ,
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 .
at least one trip systee** In the tripped condition within.ese 6*rhours and take the ACTION required by Table 3.3.1- SURVEILLANCE REQUIREMENTS 4.3.1.1 Each reactor protection system instrumentation channel shall be 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- .3.1.2 LOGIC 1YSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 month .3.1.3 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 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 l a specific reactor trip syste ipped condition where this l
"An inoperaale enannel need not be placedIn in tf these cases, the ir.o,erable channel would cause the Trip Function to occu 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 take 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 occu .
/
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TAOLE 3.3.1-1 RfACIOR PROTECil0N SYSTEM IN51RtMENTAil0N
.
..
APPLICA8tE MINIMim U$ OPERATIONAL OPERA 8L'E CHANNEL 5 II$
J; FUNCIl0NAL LMit . CON 0lil0NS_ PER 1 RIP SV51EN (a) ACTION Intermediate Range Monitors (b); - 3 I Neutron Flum - High 2 g . * 4(c) d)
I 3 Inoperative 2 Average Power Range Moniter I *I: g ,
! y Neutron flum - Upscale, Setdown
- (c) (d)
~ l Flaw Blased Slaulated thermal - 2 4 '
Power - Upscale 1 2 4 Flued Neutron Flum - Upscale 1
2 I Inoperative 1 2
) (d) -
i
'f(c)
M 1 2 41l
' ( Downscale Reacter Vessel Steam Dome gg) 2 1 Pre,*.ure - Higle 1. 2
- Reacter Vessel Water Level - Lo Level 3 1. 2 Main 5tcan line Isolation Valve - 1g g) 4 Closure
.
\
-
- - - _ _ _ . . . ._
.-- ..
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. .
p TABLE 3.3.1 _ 1 Continued)
- REACTOR PROTECitati SYSTIM INSTRUNENTATION
.
Gj L APPtICA8tE MIN 11tm OPERA 8tE CHANNELS h'F
"
OPIRA110NAL PER TRIP SYSTEM (a) ACil0N CONiilll0NS__
FUNCTIONAL t#tti **
- 5 Main Steam Line Radiation - 1,2(g) 2 f High lil- 1 7.- (Primary Containment $ G., 1. 2(h)
Pressure - High ScriukOlschargeVolumeWaterbd.Qig I./* 1/3 1, 2fg
, =M ;;;! a!+ T .s. & ,.h ,;,ugA, i,315 ) 2/r 1/3 O W.A LJXe(% n
-
5* 1(J) 4(k)
2 Turbine Stop Valve - Closure '
g gj) gg) :
5 10. Turbine Control Valve Fast. Closure, I 2 f Valve Trip Systee Oil Pressure - Low
.
-
11. Reactor Mode Switch Shutdown 2 Position 1. 2 2
l 3, a 3
!
-
2 MInualScram 1. 2 2 8 1 , 4 .
2 9 l
~ O
,
=
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'
.
1 *
s e
_ _ _ _ _ _ _ _
~ . . _
TABLE 3.3.1-1 (Continued)
g REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION A nes t ACTION 1
- 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 2 - Verify all insertable control rods to be inserted in the core and lock the reactor soce switch in the Shutdown position within one hou ACTION 3
- Suspend all operations involving CORE ALTERATION 5* and insert all insertable control roos within one hou ACTION 4 - 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 5
- Se in STARTUP with the main steam line isolation valves closed 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 hour1.388889e-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 reduce turbine first stage pressure to '000) ;;f;, ;;_f.: ACTIDM 6 0; '" 7.". "Z" ? ::: '2:. '2^ * :" "'.?s0 ..._... . __..y within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ,1,o h ha uWk by.p.a sdQ ACTION 7 - Verify all insertable control rods to be inserted within one hou ACTION 3 - Lock the reactor mode switch in the Shutdown position within one hou 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 hou .
"Escept movement of IRM, SRM or special movable detectors, or replacement of LPRM strings provided SRM instrumentation is OPERA 5LE per Specification 3. t 75 ') 3/4 3-4
,
.
_ _ _ _ -. ._ -. - .- _ _ - - . . . . _- __ _ _ _ _ . - . _ . _ _ _ _ _ __ ._ .
_ .. _
_ _ .
i i
. l
,
10 JAN 1983
,
g TABLE 3.3.1-1 (Continued)
. 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 paramete (b) This function shall be automatically bypassed when the reactor mode switch is in the Run positio ,
,
(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 demonstrations are being performed per Specification 3.1 l (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 remove 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 channe (f) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.1 *
l
,
- (g) This function shall be automatically bypassed when the reactor mode switch is not in the Run positio (h) This function is not required to be OPERABLE when PRIMARY CONTAI MENT INTEGRITY is not require (1) With any control rod withdrawn. Not applicable to control rods removed *
I . per Specification 3.9.10.1 er 3.9.1 (j) This function shall be automatically bypassed when turbine first stage pressure is ' '*50) ;:';, :; o at t "Tf' """" ? ::: Pr '395 i
I
- "f'. 5 "". 4o 12.% .Hw% 6sk k.peuuse c'W,".A A "f" .w=Wes uialess e age % -brW@ n .e h.cakt\= a: tem.* 44a ,e4wiveaew4 h (k) Also actuates the EOC-RPT system. T Ela,n ' Tnt * mat At WES tower\*u .Tohdew so%Ge5c RXTED Adramed 4 Lea.uro t3c s M*kedism. c m1 cleiTrt , o, set yoM afi W/. 5i +wrhe, Gst admap Te****". k Ps4 is um= Not required for control rods removed per Specification 3.9.10.1 or 3.9.1 ,
9.ce cessea,
= :T; (n'q 3/43-5
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.
TAttf 4.3.1,1-1 My
, . REACTOR PROTECTION SYSTEM INSTRt34fMTAY10N SINIVfillANCE REQUltfMENTS m n CHANNEL OPERATIONAL i !fn -
CHANNEL FUNCil0NAL CHANNEL *
l'# ftNICTIONAL UNIT g CONDlil0NS FOR 1411CM CHECK 1EST CALIBRAfl0N ,)
d:F SURVElllANCE REQUIREO Intermediate Bange Monitors: to
. Neutron Flum - Higli 5/U,5,p 5 ,W R 2 5 W
,
l R 3, Inoperative NA W NA 2,3,4,5
Upscale, Setdown 5 W 54 3, 5 Flow Slased Slestated
3:'
Thermal Power - Upscale 5,0II8II M ,# Q WIdII*I,5A,(RINI $ 1 '
- Fixed Neutron Flum -
'tu upscale 5 M M-4 WI I, $A 1
, Ineperative NA if-Q NA 1, 2, 3, 5 l J Down' scale 5 WR SA 1{ \
- Reacter Vessel Steam Ocee Pressure - High 35f it-4 JR[ 1, 2 'Reacter Vessel Water Level -
Law, Level 3 (59 -M 4 fR1 1. 2 l Mala Steam Line Iselation Yalve - Closure NA *G A 1 Main Steam Line Radletion -
High 5 *Q gg)
R 1. 2 l JPrlearyContainment)-0.,c 0 )- !
1 Pressure - High J5J -964 ' {R)g 1, 2 i
e
___ _ _ _ _ _ _ _ _ _ _ _ . _ m m .v ..
-.
-
- .
'
TAetE 4.3.1.1-1 (Continued). . REACTOR P90TECT10N SYSTEN INSTRUMAil0N SURVEILLANCE REQUIREMS s ;n CHANNEL OPtRATIONAL bn FUNCil0NAL CHANNEL CONOli10NS FOR %RilCH CHANNEL
- . f, ,
TEST CALIRRATION SURVElttANCE REQUIRED l 'e FIRICTIONAL mli ^
CHECK _
Uf
"
d Scram Discharge Volume WaterLeod.%k t d - m e h tr,i,tw st #4 (rig 1,2,5]'
i 1,s l
u n k. %.& suiwes 40-4 1NI 1
- Tertine Step Valve - Closure 46) M
'
10. Tertine Centrol Valve Fast ,
Closure Valve Trip Systes l 1
!
Oil Pressure - Lou -(5) N A #4 1RI l
.
j 11. Reactor Mode Switch MA
- 1,2,3,4,5 Shutdown Posillon M *
R
.
f 4t-W NA
'
-1,2,3,4,5 NA 12. Manual Scram'
g
.
w (a) Neutron The IRM detectors may be escluded frem CHANNEL'CAtltRAY10e and 5m channels shall be deterefred to everlap for at least {%) decades during each startup I E (b) af ter entering OPERATIONAL CONGITION 2 sad the IM and APM channels shall be determined to everlap for at least (%$ decades during each controlled shutdown, if not perfereed 7 1., . . within the prevleus 7 day r...:.__
-') "JR ";; M i- . Fkr 10 0teMa-. if aat - - M d ;.'t;. h r _ ,
i l
'
(d) This calibration shall consist of the adjustment of the APRM channel to confers to the power values f calculated THERMAL POWE by a heat balance during OPERAil0NAL C0fSITION I when THEMAL POWE Any AP M channel gain adjustment made in compilance ulth Specification 3.2.2 shall not be included in deteretning the absolute differenc (e) This calibration shall consist of the adjustment af the APM flow biased channel to confers to a calibrated flew signa (f) The LPRMs shall be calibrated at least once per 1000 effective . full power hours (EFPN)
using the TIP syste Verify measured core flew to be greater than er equal to established core flow at the existing pump s
[
(
]g))Thiscalibrationshallcensistof(t': Mj :*c :t, :: ;;g!;d,:")(verifyinglthe6iTsecond " h (h z slaulated'thetsal power time constan l rr, (1) This fianction is not required to be OPERA 8tE shen the reacter pressure vessel head is removed per ;**
Specificatten 3.1 .
Not appilcable to centrol rods removed per Specification 3.9.1 (j) With any central red withdraw q er 3.9.1 M Calib.h ivio ud.4 d ie.d once per e d.93
. _- _ - . . _ _ - . ____ _._ -
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__
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- ;.
APPLICA81LITY: As shown in Table 3.3.2- ACTION: With an isolation actuation instrumentation channel trip satpoint 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 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 applicabl With the number of OPERA 8LE channels less than required by the Minimum 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- .
"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 take **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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E -s E4 8 2 7 5 2 2 6 INSTRLHENTATION
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SURVEILLANCE REQUIREMENTS
.
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4.3. 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- .3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 month .3.2.3 The ISOLATION SYSTEM RESPONSE TIME of each isolation trip function j shown in Table 3.3.2-3 shall Each testbeshall demonstrated include at least to beone within channelits Itaitper at trip least once per 18 month 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
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- TAatC 3.3.2-1 (Continued)
Hg Is0LATIONACTUAll0NiNSTRUMLNTATION
- n VALVE ACTUA-li
.
- TION GROUPS MINIMUM APPLICARLE C OPERAl[0 BY OPERA 8tE CIIANNE OPERAll0NAL
,' I* COH0lil0N _ ACfl6N O TRIP FUNCTION . SIGNAL [d) *PER 1 RIP SYSitM
J REActon CORE ISOLATION C00tlNG SYSTEM ISOLATION 1, 2 3 23
- RClc steam Line & Pressure -HI h 6 J1) W)ve . .
. b, Reic SL l_tet. A Piastutt- Il h l'. ia'r (* ,
i aM'O' I, 2,3 A5 (ji, RCIC Steam Supply
4 2/Velvt. I( ) , 1, Pressure - Low-
-
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- TABLE 3.3.2-1 (Continued)
' ISOLATION ACTUAT.,10N INSTRUNENTATION
' VALVE AC104- .
rs Il0N GROUPS NIMIMIM APPLICABLE
-
I '70 OPERA 8tE CHANNE OPERATIONAL
'
j DPERATED BY SIGNAL (d) PER 1 RIP SYS1EN gjI CON 0lil0N ACTION J; TRIP FUNCTION I~
, HIGMPRESSUREC08UNTINJECTIONSYSTEM150tiTION s- jl)/ Valve .) 1, 2, 3 23
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HPCI Turbine Exhaust Olaphrsge 5 d) Pressure - High *
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TABLE 3.3.2-1(Continued)
'
ht IsotATION ACTUA110N INSTRUNENTATION -
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VALVE ACTUA-
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TION GROUPS MINIMUM APPLICABLE
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1 OPERATIONAL 5 OPERATED BY OPERA 8LECHANNEgjI ACTION i
SIGNAL (d) PER TRIP SYSTEM CONDITION l Li TRIP FUNCTI0li .
'
I 1 RHR SYSTEM SHLff00tal C00t1NG N00E 150LAT10N.- . .
. Reactor Vessel Water 1, 2, 3 27
' Level - Low, tevel 3 3 2/Qu. (') , ,
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Reacter Vessel (RNR Cut-in I Permissive) Pressure - High 3 ft)h&c.ge) 1, 2, 3 27 .
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TABLE 3.3.2-1 (Continued)
!$0LATION ACTUATION INSTRtNENTATION . - - - _ . .. .
. , .
,
. . . . . _ . . ACTTON - . . - - - . ... -- . . . -
. . . .
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
' -
ACTION 20 within the next 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ACTION 21 -
Se in at least STARTUP with the associated isolation valves 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 hour2.777778e-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 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> _ . _. .
l ACTION 23 - Close the affected system isolation valves within one hour !
and declare the affected system inoperabl ACTION 24 -
Restore the manual initiation function to OPERABLE status
. 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 hour8.564815e-4 days <br />0.0206 hours <br />1.223545e-4 weeks <br />2.8157e-5 months <br /> ACTION 25 -
Restore the manual initiation functiori to 0PERA8LE status 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 incterabl F.th. hen,Il4 coktien anu REACTOR @JILbWG b hla b . sb (FRVS) .
ACTICN 26 -
Establish SECONDARY C0"'AlhMEfff IATEGRITY wi the ct--ey ;;s 4eest;;..; ayetes operating within one hour.
1 ACTION 27 - Lock the affected systes isolation valves closed within one
,
hour and declare the affected systas inoperabl NOTES
- When handling irradiated fuel in the secondary containment and'during :
CORE ALTERATIONS and operations with a potential for draining the reactor vesse .
- %en og hrbid stir vet. 4 yeater b 10 7, que ad[or when A *
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 4 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 paramete (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% FRV .
(d) (Insert A>
(s.) Sensors arranges per valvf. 9 rove , not per frip sys4e (f) Closes only RWCU systas isot.ti n valves HV-foor ..S Hv.poo l (g) Requires systes steam supply pressure-law cot.ncideny witn ifrywell
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(h) msvat bolatien clases M-foof entw.J anly fellowing enval v cdom<fic initishn 4 4. Rcrc sysice-(i) mv.1 wetation clases Hv-Poes =4 HV-Fei2. %, ad only felf wig in.nud or ademaic. lndAher oF A=. HrcI sysk ,
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' * REK1bt CORE 150tATl0N COOLING SYSTEM 150tATiest og a '
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! - High i 118. 0 Wpo 5127.0 Wad l
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150LATION SYSTEM INSTRUMENTATION RESPONSE TIME RESPONSETIME(Seconds)$
TRIP FUNCTION Ia$$ov
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- ISOLATION .
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C :
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1 . ISOLATION ACTUATION INSTRUNENTATION SURVEILLANCE RE0ulREMEnts
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1 $ *
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CHANNEL CHANNEL FUNCTIONAL * CHANNEL OPERATIONAL CONDITIONS FOR WHICN a
t d[g TRIP FUNCT10N .
CHECK TEST Call 8 RATION SURVEILLANCE REQUIRED
.
. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION
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INSTRUMENTATION
' 3/4. EMERGENCY CORE COOLING SYSTEM ACTUATION INS i
LIMITING CONDITION FOR OPERATION
_
,
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3. The emergency cohe cooling systein (ECCS) actuation instrumentat channels shown in Table 3.3.3-1 shall be OPERA 8LE with 332 their trip set consistent with the values shown in the Trip 5etpoint column 3 .
of
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andJrith EMERGENCY CORE COOLING SYSTIM RE ,
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APPLICA81LITY: As shown in Table 3.3.3- '
M:
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,
SURVEILLANCE REQUIRENENTS f
4.3. OPERABLE Eachby ECC5 actuation the performance of the CHANNEL CHECK, instrumentation TEST and channel CHAN s CHANNEL CALIBRATION operations for the OPERATIONAL C frequencies shown in Table 4.3.3.1- .
4.3. LOGIC $YSTEM FUNCTIONAL TESTS and simulated au ,
.
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all channels shall be performed at least once per la month I*
h Each ,
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. - _ _ _ - - - _ . . - _ _ _ ._. _ _ . -- _ _ .
.
I INSERT A TO PG. 3/4 3-37
.
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 r 3 and 4, manual initiation is associated with each pump only.
i t .
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- _ . _ _ ._ ___ . _ - .- . - .-. ._ .. _ _ _ . -_ __ _ . ___
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i
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'
.
TABLE 3.3.3-1 (Continued)
'
' BERGENCY Co#E COOLING SYSTEM ..CTUATION INSTRUMENTAT
,
EH95 i With the numer of OPERA 8LE channels less than required by the
- ACTION 30 - Minieue 0PERA8LE Channels per Trip Function requirement
- With one channel inoperable, place the inoperable channel l I
I in the tripped conditten within one heur" er declare the associated system inoperable.
l l With mere than one channel inoperele, declare the .
associated system inoperabl ,
,
l l With the nebetof OPERABLE channels less than nouind by the <
l ACTION 31 - OPERA 8LE Channels per Trip Function requirement, declare f M atsum l the associated ECC5 inoperable.
- i With the neber of OPERA 8LE channels less than requind by the l ACTION 32 - Minis m CPERA8LE Channels per Trip Function requirement, place t l
l the inoperable channel in the tripped condition within ene hou h:: S: :;i:2 P '"
J.;;~;C^; 2 k" O, 0; ~ -': 7 M L^Z'f'i C.;c , p ace -
l run. r **ERAaLE Channels per Trip Functi the inope rable cha...-C *" L ;- condition within
~ : -*"a one7hour; within days j
g=m-4he-tegrFali ; S: e channel to ortna.:::rr'-t1
' n;:-9'-:;"
-
'
,
-
=C With the numer of OPERA 8 2 channels less than ret:utred by the l :
- ACTION 34 - Mateue OPERA 8LE Channels per Trip Function requirement, notere 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 inoperabl . For both trip systans, declare the NPCI system inoperabl ~
l with the neber of OPERA 8LE charinels less than required by the
,
i i ACTION 36 - Mnimum 0PERA8d Channels per Trip Function requirement, place at !
i ienst one insperele channel in the tripped condition within i a er declare the NPCI system inoperabl ene hour <
i l
ACrION 37 - M .e a m.or .f m RA8 d channeis isss than the Totai Nu. e ,
of Channels, declare the associated emergency diesel generator inoperable and taka the ACTION mquired by Specification 3.8.1.1
- er 3.8.1.2, as appropriat l ,
l ACTION 38 - With the number of OPERA 8 d channels one less than the Muster of Channels, place the inoperable channel in the tripped-i f '
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 performance of the next required CHANNEL FUNCTIONAL TES t
"The provisions of Specification 3.0.4 are not applicable.
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3/a.3.4 RECIRCULATION PJer TRIP ACTUATION INSTelp(NTATION
>
S l ATW5 RECIRCULATION PUMP TRIP SYST S INSTRLpqENTATION '0;;' z :
! LIMITING w=GITION FOR OPERATION
- .
-
.
.
The anticipated transient without scras recirculatten puep trip i 3.3.4.1 f (ATWS-RPT) system instrumentation channels shown in Table 3. l i OPERA 8LE with their trip setpoints set censistant with values shown i
!
Setpoint column of Taste 3.3.4.1-2.
l I
' APeLICABILTTY: OPERATIONAL ColeITION .
!
.AElD: l
! With an ATW5 recirculatten pump trip system instrumentatt'on chan '
f trip setpoint less conservative than the value shown in the A11rmab
Values column of Table 3.3.4.1-2. declare the channet inoperable u i the channel is restored to OPERABLE status with the channel trip setpoint adjusted consistent with the Trip Setpoint valu * With the number of OPERAELE channels one less than required by the
Minimum GPERABLE Channels per Trip System !
condition within one hou l With the number of OPERA 8LE channels two er aere less than reqli
.
by the Ninfeue OPERA 8LE Channels per Trip system requireme trip systas ende dechre. %a %p spbtm %pc aMe.,
' '
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r:= ;rr! _:t-i ' "f ".; '.-;;. .:;h .. . xu .x;';; ;f c:
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, ace betr l love J:---I and one reactor vessel pressurewithin one hour.
! inoperable channe . '- N trippe e vessel water level
nelude the If the ingesmo reactor vessel pressure channe ,
i
-
g er ..,, .,._ . . .
. _ . . .
, With one trip systas inoperable, restore the *
l in the next 4 hove With both trip systans inoperable, restarel at le .
the next 4 hove .
SURVEILLANCE st@Utita nT5 Each ATW5 recireviation pug trip system instrumeritation A EL ch 4.3.4. shall be demonstrated OPERA 8LE by the performancein-of the C FUNCTIONAL TEST and CNANNEL CAL 28AAT10N sperations at t
'
-
Table 4.3.4.1- f
/
4.3.4.1.2 LOGIC SYSTS FUNCTIONAL TESTS and eisulated aute all channels shall be performed at least once per 18 month g/4
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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) E with systes 3.3. in the Trip 5etpoint
.
instrumentation channels shown h inATION Table PUMP TRIP3.3.4.2-1 the(r trip setpoints set consistent with the values s ow
.
.
SYSTEM RESPONSE TIME as shown in Table 3.3.4.2- OPERATIONAL CQaQITION 1, when THERMAL POWER is ,
APPLICA81LITY:
equal to 130p of RATED THERMAL POWE tion 3: h With an end-of-cycle ncirculation pump l the
.
trip Allowable inoperable until Values the channel columnis of notand Table to 3.3.4.2- CPERA8LEdecl stat channel setpoint adjusted consistent with thebyTrip the 5etpoint.
i
_
With the number of CPERA8LE channels oned condition less boththan requ I Minimum CPERA8LE Channels per Trip System i requireme
"
trip systees, place the inoperable channel (s) in the tr ppe within one hou ired With the number of CPERA8LE channels twot for or one son less tha
by the Minf aum CPERAILE Channels per Trip Systes n
,
trip systes and:
!
l If the inoperable channels consist of o channels in the tripped condition within one hou *
i I
' If the inoperable channels include two system inoperable, i ten
,
With one trip systes inoperable, tr: ?"?:0" motore the inoperabic . ;;;' . ;: 53 t 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;':';;;' . . ,
"e; THERMAL POWER within the next 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> RATED .
.
i stem
.
With both trip systees inoperable, restore " :0" .;; at 'c;dleast ty one tr l to OPERA 8LE status within one hour er ';
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 4.3.4. Each end-of-cycle recirculation pump trip system ANNEL instrumenta in channel shall be demonstrated CPERA8LE by the performance of th FUNCTIONAL TEST and CHANNEL CALIBRATION operations at t Table 4.3.4.2.1- LOGIC SYSTEM FUNCTIONAL TESTS and simulated automa 4.3.4. a11* channels shall be perforwed at least once per la month The END-OF-CYCLE RECIRCULATION PtMP be within l TRI 4.3.4. each trip function shown in Table 3.3.4.2-3 shall be demonstrated *
toEa its limit at least once per 18 month tested
,
logic of one type of channel input, turbine control valve, fast closur
- -
- ':
turbine stop valve closure,
~
'Th: such t'= :11:tt:d that both ':r trn types h:r of :r: channel inputs are at
'
least once per .:r 36 months.hd by txt :t t x:t c:: ;;r 'O r un:.)
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fee-OF-CYCLE RfCIRClitATION Mar TRIP $YSTEM INSTRUENTATION
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l r MIN 11RSI OPERAttECHANNEg)
.df PER TRIP SYSTEM
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l TRIP FINICTICII 2("I 1 Tertine Step Valve - Closure -
l
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2(b)
i i Turbine Centrol Velve-Fest closure
-
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ttlance provided t'
I*I 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
[ that the other trip system is OPERABL l
% (b)This functlen shall be outematically bypassed!"'I^_".
T'" k n ? : 'MZ d "".!!S when tortfae first stage pressure i
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INSTRUMENTATION r, r p, ,. . - . .
- nsu, e t.,. .:,u... , .
I TRAVERSING IN-CORE PROBE SYSTEM LIMITING CONDITION FOR OPERATION .
3.3. The traversing in-core probe system shall be OPERABLE with: Five novable detectors, drives and readout equipment to map the core, and Indexing equipment to allow all five detectors to be calibrated in a common locatio APPLICABILITY: When the traversing in-core probe is used for: Recalibration of the LPRM detectors, and b.* Monitoring the APLHGR, LHGR, MCPR, or MFLP 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 applicabl .
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 OPERABL /4 3-89 SEP 3 C ass HOPE CREEK i
f
.--- - - - _
3/4.5 EMERGENCY CORE C00 LING SYSTEMS 3/a.5.1 ECCS - OPERATING LIMITING CONDITION FOR OPERATION
.
3. The emergency core cooling systems shall be OPERA 8LE with:
' The core spray systes 446G3 consisting of two subsystems with each sutsystem comprised of:
emet sem9 jTwel CPERA8LE466 pump (sJ, and l i
^ An OPERABLE flo'w path capable of taking suction from the suppression chamber and transferring the water through the spray sparger to the reactor vesse , The low pressun coolant injection (LPCI) system of the residual heat removal system consisting of e subsystems with each subsystem comprised of: - W .
Om l 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 vesse The high pressure cooling injection (HPCI) system consisting of:
. One OPERABLE HPCI pump, and An OPERABLE flow path capable of taking suction from the
supp nssion chamber and transferring the water to the reacter vessel.
l l@ The automatic depressurization system (ADS) with t hnt (:i ) 64 OPERA 8LE ADS valve OPERATIONAL CONDITION 1, 2*, ** f, and 3*, **,"# l APPLICABILITY:
"The HPCI system is not required to be dPERA8LE when nactor steam dome l pressure is less than or equal to g psi ,,
The ADS is not nouind to be OPERA 8LE Twhen nactor steam : done pressure is less than or equal to {1001psi ,
- 5ee special Test Exception 3.10.6.
'
- 8 0w,. L.Pt1 sAsp h em ch he RHR syde= m3 be brerab\e W %d
% is edged in %s skddow coob3 v^ ode h yearhv vesse(
,
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.
.
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, ,
__
.
EMERCENCY CORE COOLING SYSTEws LIMITING CONDITION FOR OPERATION (Continued)
ACTION:
' For the c'ere spray systes:
c.eee se<*a3 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 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> covsspeap 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 />, For the LPCI systas: g,, ,p.,3 With :::we ==ee. W.as hweL"'! ; n ' :t e:r :r,[t: e LPCI subsystems inope
___
provided_that at leaft one C&&Asutsystem is CPERA8LE, restore .
the Inoperable LPCI.n--(t) to CPERA8LE status within 7 days or
'
A fem 1 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 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> . "iO (:) (nd L.*C! :y:!- cr:::-t!: n! : !!: -' (: e t' i
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2.\ With6ece.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 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ,7 n 3'4 WithCo <-
toe-LPCI subsystems :r: f ee inoperacle, be in at least l
.
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 />."
,
core S i For the HPCI system, provided the 466,Mthe LPCI system, the A05 and the RCIC system are OPERA 8LE: With the HPCI system inoperable, restore the HPCI system to OPERASLE status within 14 days or be in at least NOT SHUTDChN
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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 to i tt96t psig within the following 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> l
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"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 method SeeE ctEt /4 5-2 0; ; ; (= / ;
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EMER0ENCY CORE COOLING SYSTEMS LIMTING CON 0! TION FOR OPERATION (c*ontinued)
ACTION: (Continued) For the ADS: ,, ,g qb 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
anddays or be reduce in at least reactor HOT steam done SHUTDOWN pressure towithin 1(1001 the psig next within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> l the next 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> . 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 hour2.777778e-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 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 .
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useage factor for each arrect.ed safety injection neule shall be !
provided in this Special Report whenever its value exceeds 0.7 s a
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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,, 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 wate 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" positio *
.
:r W L*C! :y "rn:-t';
- j;ts
- 2:y:'
, ::r' *y'nI:W
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(:;:r) (:t(:?:::d 1:::t :::)
ef uLPCI;: :-
-
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2.'1, For the HPCI system, verifying that the HPCI pur.p flow controller I is in the correct tio ews sw when test hursuanttoSpecification4.0.5:
! Verifying that,(64 The two 46 pumps in each subsystem together develop a flow of at least163501. gpe against a test line pressure : u r =ofngreater
- 1 prnthan
- ur: ~ 1 or equal to tef psig, tr- n;-- W ; u : l
- '
l A (115)w p ';.
l TN x; LPCI pumps in each subsystem tq;th:r developsa flow of I
_at least ':1,000) gpa against a test line pressure of > (el psig, l
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n;= .. , '.n; u : :::t:r =,,,n,,:1
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The HPCI pump develops a flow of at least 4600-gpm against a l when steam is being supplied test line pressure of >rtHGG-)
t "1"05, -l'.0, -psig! 20 psig."" l d- to the '* tuttiine At least once 7pe 18 months:
SYS aM N tooepto,-So -
l . 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 tes "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 operatio **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 tes #W.ke b.be. deier,m6el bh Pre-c9 hM- S nb?_.,E cate n 3;g $.4
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EMEMENCY CORE CCOLING SYSTEMS v ., ..
$','8VEILLANCE RECUIREMENTS (Continued) For the NPCI system, verifying that:
a) The system develops a f w of at least G O ") gpa :against ::: :--
a gl .
test line pressure of psig, :; ::;:- S ; i: - :_..' ' ; ;,--
--
.;;;;: ;.:::_ : :" ' 155 ::';. when steam ;2*'. . r.'fms ac a
- *-
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L _ ;_.: -- +@ aoo a.15)psig.** o b) The suction is automatically transferred from the condensate, storage tank to the suppression chamcer on a condensate storage tank water level - low signal and on a suppressien
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chamb,er - water level high signa For the ADS: . At least once per 31 days, performing a CHANNEL FUN TIONAL TEST of the !!" '!*-- 5- "- r _ --:: fAas system lo(,, essure alarm system. p -*g e.-e. =..=4 lude....t
- At lea'st one's per 18 months:
l a).. Per.f.gtsting a system functional test which includes simulat~ed. automatic actuation of the syst6m throughout its emergency operating sequence, but excluding actual valve ,
actuatio b) Manually opening each ADS valve when the reactor steam done pressure is greater than or equal to 100 psig(*=) and l observing that either:
l 1) -The control valve or bypass valve position responds
,.,
accordingly, or l 2)
There is a corresponding change in the measured steam 4jaJ . d y,,5 n.:.
- t1o ?'a- **
Performing a CHANNEL CALI5RATICN of theAr~ "-"?
" - - -
c)
r ; :::-d gas systes low'$fessure alars system and verifying an alare setpoint of {Ef) + -js} psig on decreasi.ng pressur . .
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""The provisions of Specification 4.0.4 are not applicaele {
p adequate to perform the tes I Value 6, 6, /efer,l,g g , , ,,, ,
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EFERGENCY CORE COOLING SYSTEMS
. 3/4.5.2 ECCS - SHUTOOWN
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LIMITING CONDITION FOR OPERATION m; six emr3ew cere. coolas s4syfeas 43 ,. mini l 3.5.2 n y two of the followingshall be OPERABLE: l a.Tw [ ore spray system 4Gr) subsystems,with e
core sysy e subsystem comprised of:
l l
- (?t h=t :=) 1Two] OPERABLEAGE pump (s), and An OPERABLE flow path capable of taking suction from at least 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 -
b) When the suppression chamber water level is less than the I limit or is drained, from the condensate storage tank :
containing at least M available gallons of water, y equivalent to a level) of (275%.
lSs,oco ea d .
r* - , l Low pressure coolant injection (LPCI) system subsystems.with-a-
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b. A subsystem comprised of:
- 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 s_ vesse .
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
, all operations with a potential for draining the reactor vesse ,
' W'ith both of the above required subsystems inoperable, suspend CORE'
ALTIAATIONS and all operations with a potential for draining the Restore at least one subsystem to OPERABLE status reactor vesse withiij4hoursorestablishfCCCNCYCGTAbiniINTEGRITYwith the neat 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> N ct " " * 4 ' " Y * * * n)
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"The ECC5 is not, required to bE OPERABLE provided that the reactor vessel head water level is saintained within the limits of Specification ' ( }tCPg C RS E /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 Surveillance Requirement 4. l 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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CI T i (%?,/t) 3/4 5-7 3 0 MAP 1993
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10 JAN 883 EMERGENCY CORE COOLING l O SYSTEMS
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3/a.5.3 SUPPRESSION CHAMBER ERATION
_ LIMITING CONDITION FOR OP The suppression chamber shall be OPERABLE:
3. In OPERATIONAL CONDITION 1, 2 and 3 with a contained ) w at least '07,'001 feta, equivalent aa.A.L en AuJed toVe-level of G2'0'F74,5 us, coo l
! In OPERATIONAL CONDITION Mexcept that the(4 and 5' with a c suppression mit or may be drained provided ,
equivalent Int,eco-(--)- f t*, to 4)1evel ofchamber 74,P level may be less than th '
that: l
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3 No operations are performed that have a potential for drainin :
the reactor vessel, !"
The reactor mode switch is locked in the Shutdo
. =
position, I'55 ooo availab e(. l The condensate storage tank contains at least4150,,000) -
gallons of water, equivalent m de toye-level of TD$, and I The core spray system is OPERABLE per Specification 3 an OPERABLE flow path capable of taking suction from the condensate storage tank and transferring the water throug spray sparger to the reactor vesse OPERATIONAL CONDITIONS 1, 2, 3, 4 and 5*.
APPLICABILITY:
ACTION:
In OPERATIONAL CONDITION 1, 2 or 3 with i hin the suppr *
f level less than the above limit, restore the within TDOWN water the level to w t
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24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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
'
In OPERATIONAL CONDITION 4 or 5* with the suppre level:less than the above Ifmit or drained andtions the above condibions not satisfied, suspend CORE ALTERATIONS l and lock the a tGrhave a potential for draining the reactor vesseEstablish SECONDA reactor mode switch in the Shutdown positio I'
CONTAINMENT INTEGRITY within 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> that the
- Tae suppression chamber is not requiredbeing ed flooded to(when be from OPERABLE the cavity p reactor vessel head is removed, the cavityh limits is flooded of for the suppression pool), the spent fuel pool gates are remov is floodedy, and the water level is maintained within t e ( Specification 3.9.8 and 3. fe tfi d e~ 3.&. z ,4 fu fmswe saffHss:ces re p re d s, t# See 3/4 5-8 3*LE,C.w- . . Mik,,
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EMERGENCY CORE COOLING SYSTEMS P
SURVEILLANCE REQUIREMENTS I ERABLE by verifying The suppression chamber l to ,shall 4: ;;!'-?!ey be deterstned CP 8
4.5. thepater level to" be greater than or equa g twMedtel C 97at least once per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> '. . ( ; t ';::t ;;;; pr (M) h; bove limit or i
With the suppression chamber level less than the at least
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I once p 4.5. drained in OPERATIONAL CONDITION 4 or 5*, a to be 3.5. Verify the required conditions of 5pecification l satisfied, or VeHfy footnote conditions * to be satisfie . .
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( W.ee. cautw 3/4 5-9 45-N (**/+)-
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