ML20207J940
| ML20207J940 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 07/11/1986 |
| From: | Lawyer L, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20207J936 | List: |
| References | |
| 50-325-OL-86-01, 50-325-OL-86-1, NUDOCS 8607290298 | |
| Download: ML20207J940 (149) | |
Text
._
p floug UNITED STATES Do NUCLEAR REGULATORY COMMISSION
[
REGION 11 n
g j
101 MARIETTA STREET, N.W.
t ATLANTA, GEORGI A 30323
%...../
ENCLOSURE 1 EXAMINATION REPORT 325/0L-86-01 Facility Licensee:
Carolina Power and Light Company 411 Fayetteville, Street Raleigh, NC 27602 Facility Name:
Brunswick Steam Electric Plant Facility Docket Nos:
50-325 and 50-324 Written, simulator and oral examinations were administered at the Brunswick Steam Electric Plant near Southport, North Carolina.
Chief Examiner: r'2//NM i
7//i//C Lawrence L.VLawyer u Date Signed Approved by:
/ k#
7//sh John F. Munr6,' Acting Section Chief Date Signed Summary:
Examinations on May 19-23, 1986 Oral examinations were administered to seven R0 candidates; six of whom passed, and eight SR0 candidates; all of whom passed.
Simulator examinations were administered to eight R0 candidates; seven of whom passed, and eight SR0 candidates; six of whom passed.
Written examinations were administered to seven R0 candidates; four of whom passed, and eight SR0 candidates; all of whom passed.
8607290298 860721 PDR ADOCK 0500 5
i REPORT DETAILS 1.
Facility Employees Contacted:
- G. Barnes, Project Specialist D. Shaw, Operations Training E. Hawkins, Operations Training
- N. Stewart, Senior Specialist, Operations Training J. Allison, Operations Training
- S. Morgan, Senior Specialist, Operations Training J. Keith, Operations Training
- C. Dietz, Plant General Manager
- E. Bishop, Manager, Operations
- J. Moyer, Director of Training 2.
Others Contacted:
- W. Ruland, Senior Resident Inspector
- Attended Exit Meeting 3.
Examiners:
J. Hanek, EG&G T. Bishop, EG&G l
J. Sherman, EG&G M. Spencer, EG&G
- L. Lawyer, NRC
- Chief Examiner 4.
Examination Review Meeting At the conclusion of the written examinations, the examiners provided Mr. Barnes, of your staff, with a copy of the written examination and answer key for review.
Facility comments on the written exam are enclosed as Enclosure 3 to this report.
The following resolutions are provided to these comments and to corrections made during exam grading.
a.
SR0 Exam l
(1) Question 5.13 l
NRC Resolution:
Answer key point breakdown was changed to better reflect grading. Moved [0.5] to after " faster rate than".
l N
6nclosurt 1 2
JUL," { [g86 (2) Question 5.14 NRC Resolution:
Added parenthesis to part's (c) and (d) to exclude those portions of the answer key that were not required.
Added "or" to the answer in part (a) to reflect that either statement is an adequate answer.
(3) Question 5.16(a)
NRC Resolution:
j-Part (a) - Agreed.
The first part of the answer in (a) which states "RPT or TT" was changed to "High Pressure (ATWAS) Recir-culation Pump Trip".
Point value remains unchanged.
Part (b) - Added answer to indicate that "MSIV closure if recirc
~
pumps tripped on LL2" is an alternate answer since, if Level 2 is assumed the cause of the Recirculation Pump trip then MSIV closure l
is an acceptable cause of pressure increasing.
Part (c) - Added parenthesis to part (c) to indicate that portion of answer key not required for full credit.
(4) Question 6.02 NRC Resolution:
Agreed.
IV's will be deleted from answer (c) in future use of this question.
Since (c) is still the only reasonable answer, no other change was required.
(5) Question 6.06 In part (b)ylock Switch (SS)" and "any 2 Added "Ke NRC Resolution:
Agreed.
, answer sheet was changed to of 3" to answer sheet.
accept 50 psig or 25 psig if an explanation of the lower pressure was included.
(6) Question 6.07 NRC Resolution:
The additional answer "both Brunswick units" was accepted. The answer key was changed accordingly.
(7) Question 6.11 NRC Resolution: Modified the point breakdown to indicate a better distribution.
r Lnclosure 1 3
JUL i is (8) Question 6.16 NRC Resolution:
"Less than 50% rod density to 22% power" is an alternate answer to "when in notch control."
The reference material indicates this is when notch control is in effect.
The answer key was changed accordingly.
(9) Question 6.17 NRC Resolution: Agreed. "High Pressure Turbine Exhaust Pressure" or any equivalent answer will be accepted. No change required.
(10) Question 7.01 NRC Resolution:
A0P-22 states " Increase unit generation to the neximum consistent with plant conditions" and " Caution.
The maximum allowable time at (58.4 Hz) is (one minute)".
GP-05 reduces power to 100 MW for normal shutdown.
Since the shutdown required under A0P-22 il of more urgency than implied in choice (d), it will be modified by deleting " reduce load to 100 MWThe inclusion o did prior to future usage.
not make choice (d) wrong. No other change was required.
(11) Question 7.03 NRC Resolution:
Agreed.
The answer sheet was changed to "10.EOP-01 and execute as many steps as possible".
(12) Question 7.04 NRC Resolution:
It was not intended to test the operators memorization and therefore no valve numbers were required.
The question tested the operator's general understanding of the l
subject realignment.
Only the most important aspects, such as order of operation and locked status, are required.
No char.ge required.
(13) Question 7.09(b)
NRC Resolution:
The question and answer are direct quotes from the reference. No change required.
(14) Question 7.12 NRC Resolution:
The referenced TS states that an inoperable control rod should be inserted ranually and disarmed.
That does not infer that insertion and di.sarming renders the rod inoperable.
Furtherniore, the referenced APP does not merely infer the rod is not IMOF, it very specifically states so. No change required.
g 4
JM
.J (15) Question 8.07 NRC Resolution:
Agreed.
Additionally, OP-24 Section 5, CAC establishes that the Containment Atmospheric Control System is used to meet the requirements of TS 3.6.6.3 and therefore selection (a) is the correct answer. The ens,;er sheet was changed accordingly.
(16) Question 8.14 NRC Resolution:
Added alternate answer to part (a) stating "EDG controls will shift to automatic and auto start".
Point value remains unchanged.
b.
R0 Exam (1) Question 1.14 NRC Resolution:
Answer sheet was changed to accept 6.0% + 1 or
-0.5%.
(2) Question 1.15 NRC Resolution:
Added (+ or -10%) to answer No. 2 to allow for small differences due to calculations based on steam tables.
(3) Question 2.01 NRC Resolution:
Same as SR0 Question 6.06.
(4) Question 2.03 NRC Resolution:
Added reactor feed pump speed control lockout relays to answer.
This is a correct answer, not on the answer l
sheet, but in the reference material.
1 (5) Question 2.09 NRC Resolution: Added " Locally" to answer No. 2 and an additional answer, No. 4, to reflect that there are two separate " Emergency Start Buttons".
This was verified by the facility supplied material. Also added "any" to the point value.
(6) Question 2.12 NRC Resolution: The question clearly asked for a " core position".
Of the suggested additional answers, only " center of core TIP tip tube" is a core position.
" Center of core TIP tube" was added to the answer sheet.
5 J Ul.. i isi (7) Question 3.01 NRC Resolution:
Changed point value from 0.5 each for the parameter and setpoint to 0.25 each for parameter and setpoint to correct error. Total point value remains the same.
(8) Question 3.02 NRC Resolution:
Changed answer to allow + or - 10% for numbers in answer.
(9) Question 3.06 NRC Resolution:
Changed answer to allow + or - 5% for numbers in answer.
(10) Question 3.07/3.08 NRC Resolution:
Agreed.
Verified by additional review of facility supplied reference material. Questions were deleted.
(11) Question 3.09 NRC Resolution:
Changed answer (5) for the "45% bypassed" to "both" reactor feed pumps vs. "one or more".
Also changed "and" to "or" and "alann is received LL" to " greater than or equal to 182" for the water level requirement to correct answer to agree with the reference material.
(12) Question 3.10 NRC Resolution:
Deleted the " greater than" and "less than" signs from the No.'s 1, 4, 5 and 6 answers to correct typo's.
(13) Question 4.01 NRC Resolution:
The terminology as used in this question does not clearly distinguish whether it is asking for specific examples or generic classes.
The question, therefore, could be confusing for the candidates.
For this reason the question was deleted.
It should be noted, however, in response to the facility corrinent on R0 exam applicability that NUREG 1021, ES-202 allows the use of questions on administrative procedures in Section 4 of the R0 exam.
(14) Question 4.04 NRC Resolution:
The answer sheet indicates that " pocket dosi-rreter" and " chipper" are not key words.
These words being in parenthesis is an inoication that they are not part of the required answer.
No change required.
Lnt;v ure 1 6
JUL 2 : p (15) Question 4.05 NRC Resolution:
Added "and execute as many of the actions as possible" to answer No. 10.
This made the answer consistent with the Facility comment on Question 7.03.
Also the individual point value was changed to correct a typographical error.
(16) Question 4.06 NRC Resolution:
Added parenthesis around "via SLC" in answer No. I and around "with borax" in answer No. 2.
The question does not solicit these specific answers.
4.
Exit Meeting At the conclusion of the site visit the examiners met with representatives of the plant staff to discuss the results of the examination.
Those individuals who clearly passed the oral examination were identified.
There were no generic weaknesses noted during the oral examination.
The cooperation given to the examiners and the effort to ensure an atmos-phere in the control room conducive to oral examinations was also noted and appreciated.
The licensee did not identify as proprietary any of the material provided to or reviewed by the examiners.
3
. f:
g F3 CLOSURE 2 U.
S.
NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
_gRUNgWIgK_1Pg2___________
REACTOR TYPE:
_gWR-gg4_________________
DATE ADMINISTERED: _gb/0gf12________________
w = - =-,,
_gPENCER _M._____________
f f W/ Y EXAMINER:
1 3
i 4.a k \\s,,
e APPLICANT:
INSIBUCIJgNg_Ig_8EELIC8 nil Use separate paper _for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%.
Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY
% OF APPLICANT'S CATEGORY
__Y6LUE_ _Igl@L
___@CQBE___
_y@LUE__ ______________CQIEggBY_____________
54) 27.00
- 24. ?4,y.
1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
?5 il
_2Z 25__ _-25rt? _ = -
_ _ _ =
________ 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS j ;.3s 24'/3 l l d ire
_2ssZs__ _21 21
________ 3.
INSTRUMENTS AND CONTROLS EL. bS l4. ]4
________ 4.
PROCEDURES - NORMAL, ABNORMAL,
_ M _ _ _-Z E s 1 2 n _ _ --
f EMERGENCY AND RADIOLOGICAL CONTROL M
-issssg2c 100.00
________ TOTALS FINAL GRADE _________________%
All work done on this examination is my own. I have neither M\\A3l'EOHuu'r.y
- r cic givsn nor received aid.
asscreanv s s1saavuas==
e f
e 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATIQN PAGE 2
t IHEBdggyN@ Migs _HE@l_IB@NSEE6_@NQ_ELUlg_ELgy t
QUESTION 1.01 (2.00)
Match each of the power distribution limits E 1,
2,
&3 3 with its associated FAILURE MECHANISM in column A AND its associated LIMITING CONDITION in column B (2.0)
- 1. Linear Heat Generation Rate (LHGR)
- 2. Average Planer Linear Heat Generation Rate (APLHGR)
- 3. Minimum Critical Power Ratio (MCPR)
A-FAILURE MECHANISM B-LIMITING CONDITION
= - - - - -
--=____________
A1. FUEL CLAD CRACKING DUE B1. 1% PLASTIC STRAIN TO LACK OF COOLING CAUSED BY DNB.
A2. FUEL CLAD CRACKING DUE TO B2. PREVENT TRANSITION BOILING HIGH STRESS FROM PELLET EXPANSION A3. GROSS CLAD FAILURE DUE TO B3. LIMIT CLAD TEMP TO 2200 F DECAY HEAT AND STORED HEAT FOLLOWING A LOCA QUESTION 1.02 (2.00)
Closing down on the discharge valve of a motor driven centrifugal pump will cause flow to ____(1) discharge pressure to __(2) __,
and motor amps to____(4)
NPSH (REQUIRED) to
____(3) ____,
(Complete each blank with increase, decrease or remain the-same)
(2.0)
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CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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1.
PRINCIPLES OF_NUgLEAR_PgWER_ELANT_gPERATIgN PAGE 3
2 ISEBdgpyN8digg2_bE81_lB@NgEEB_8Ng_E691g_ELgW QUESTION 1.03 (1.50)
State the predominate mode of heat transfer (conduction, convection, or radiation) for the following:
a.
Center of fuel pellet out to the pellet edge (0.5) b.
Clad surface to the center of the coolant channel (0.5)
Clad surface to coolant under film boiling conditions (0.5) c.
i OUESTION 1.04 (2.50)
For each of the following events, state which coefficient of (2.5) reactivity would act FIRST to change reactivity.
a.
Control rod drop at power b.
SRV opening at power c.
. Loss of shutdown cooling.when shutdown d.
One recirculation pump trips while at 50% power Loss of one f eedwater heater (extraction steam isolated) at power e.
OUESTION 1.05 (2.00)
Following a normal reduction in power from 90% to 70% with recirculation flow, HOW will the following change (increase, decrease, or remain the same) AND WHY:
ASSUME NO CHANGE IN CIRCULATING WATER FLOW The pressure difference between the reactor and the turbine a.
steam chest.
(1,0) b.
Condensate depression at the exit of the condenser.
(1.0) i
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CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATIONt ISEBdQQYN@dlC@t_bE@I,Ig@N@EE6_@ND_ELUID_ELgW QUESTION 1.06 (1.00)
In a subtritical reactor, Keff is increased from.880 to.965.
L Which of the following is the amount of reactivity that was added to the core?
a.
.085 delta k/k b.
.100 delta k / k c.
.125 delta k / k d.
.220 delta k / k QUESTION 1.07 (1.00)
The reactor trips from full power, equilibrium XENON conditions. Twenty-four hours later the reactor is brought critical and power level is main-tained on range 5 of the IRMs for several hours. Which of the following statements is CORRECT concerning control rod motion?
1
- a. Rods will have to be withdrawn due to XENON build-in.
b.
Rods will have to be rapidly inserted since the critical reactor will cause a high rate of XENON burnout.
c.
Rods will have to be inserted since XENON will closely follow its normal decay rate.
- d. Rods will approximately remain as is as the XENON estab-lishes its equilibrium value for this power level.
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
1.
PRINCIPLES OF-NUCLEAR POWER PLANT OPERATION PAGE 5
t IHE8dppyN8dICS _HE8I_IB8NSEE8_8Np_ELUIp_ELgg t
QUESTION 1.08 (1.00)
Concerning control rod worths durir.g a reactor startup from 100% PEAK XENON versus a startup under XENON-FREE conditions, which statement (1.0) is correct?
a.
BOTH control rod worths will be LOWER regardless of core XENON conditions.
b.
CENTRAL control rod worth will be HIGHER during the PEAK XENON startup than during the XENON-FREE startup.
c.
BOTH control rod worths will be the SAME regardless of core Xenon conditions.
d.
PERIPHERAL control rod worth will be HIGHER during the PEAK XENON startup than during the XENON-FREE startup.
QUESTION 1.09 (2.50)
Diven the following plant (Unit 1) conditions:
Reactor power:
- 100%
Reactor pressure:
- 1010 psig Throttle pressure:
'949 psig Maximum combined flow limit:
normal setting Load limit:
normal setting Load set:
normal setting Bypass valve capacity:
25%
Recirc flow controls master manual Bypass valve jack:
0% (closed)
Assume the EHC "A" pressure regulator fails low.
Using attached Figures 1 and 2, explain the effects of the failure on each of the following:
a.
The bypass valves b.
The control valves h
- c. Reactor vessel pressures.vhswe'/ db
^^~
1 e
1T) d.
Throttle pressure ji o.
Reactor power (5 9 0.5 ea = 2.5)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE
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s 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION PAGE 6
t IHEBdgDYN@dlCS _HE@l_lB@NSEEB_@ND_ELUID_ELgW t
QUESTION 1.10 (1.00)
Within normal limits, which of the following correctly describes the Maximum Fraction of Limiting Power Density (MFLPD)?
oa. LHGR-actual / LHGR-li mi t ; must be maintained < 1 b.
LHGR-limit / LHGR-actual ; must be maintained < 1 c.
LHGR-actual / LHGR-limit ; must be maintained > 1 d.
NONE OF THE ABOVE QUESTION 1.11 (2.00) a.
DEFINE " Critical Power".
(1.0) b.
Which one of the following conditions would tend to INCREASE the Critical Power level assuming all other variables remain unchanged?
1.
Inlet subcooling is DECREASED
- 2. Reactor pressure is DECREASED 3.
The axial power peak is RAISED
- 4. Coolant flow rate is DECREASED (1.0)
QUESTION 1.12 (1.00)
Figure 3 of this exam is a graph of temperature coefficient vs.
moderaar temperature.
Two curves have been drawn on this graph.
Identify which curve represents change EARLY IN CORE LIFE and t9hich represents (1.0)
LATE IN CORE LIFE.
(***** CATEGORY 01 CONTINUED ON NEXT PAGE
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1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION PAGE 7
t IHEBdQQYN@dlCS _HE81_IB9NSEEB_@NQ_ELUlp_ELgW t
QUESTION 1.13 (1.00)
Which of the below selections best describes the removal of Samarium-149 from the core.
A.
Decay by alpha emission B.
Decay by Beta emission C.
Burn out by neutrons D.
Fission QUESTION 1.14 (2.00)
The reactor has been operating at high power for several weeks.
a.
How much thermal power, in MEGAWATTS, is being produced immediately following a scram?
After 1 minute?
(1.0) b.
Why does this power NOT INDICATE on the nuclear instrumentation?
(1.0$
QUESTION 1.15 (2.50)
During a cooloown of the reactor vessel from the remote shutdown panel, reactor pressure decreased from 885 PSIG to 595 PSIG in one half hour (30 mins.).
Have your cooldown limits been exceeded?
(Show All Work)
(2.5)
QUESTION 1.16 (2.00)
Unit 1 is operating at 2436 MW (100% of rated), and the reactor scram is set for 118% of rated. The total scram delay time is 10 seconds, measured from the time the scram setpoint is exceeded until sufficient negative reactivity has been added to turn power.
If a nuclear excursion creates a 10 second period, WHAT will be the peak power for the excursion? NOTE:
- 1. Assume NO temperature or void effects.
- 2. Assume a constant period until power turns.
(2.0)
(***** END OF CATEGORY 01 *****
)
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2.
PLANT-DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE S
QUESTION 2.01 (1.50)
With regard to the ADS system:
A.
What are the two methods (using normal control room controls) that
,'lthe operator may use to shut the ADS valves once they have auto initiated and are OPEN7 (1.0)
B.
The ADS valves will remain open until Reactor pressure is about ___________ psig above containment pressure.
(0.5)
QUESTION 2.02 (2.00)
Figure 4 of this exam is a sketch of a hydraulic control unit.
Identify which letter on the sketch points to one of the below (2.0) listed items.
NOTE: NOT ALL ITEMS ARE IDENTIFIED ITEMS 1.
Charging Water Riser 2.
Insert Riser 3.
Outlet Scram Valve 4.
Cooling Water Riser 5.
Inlet Scram Valve 6.
Discharge Riser QUESTION 2.03 (1.25)
List five of the systems or components receiving power from the 4
Uninterruptible Power Supply System.
QUESTION 2.04 (3.00)
List three of the four design bases of the HPCI System.
(3.0)
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1 PAGE 9
2 __P(@NI_DE@l@N_1NC(UDIN@_@@EEIY_@ND_EMEBGENCY_@Y@IEd@
DUESTION 2.05 (1.50)
Regarding the HPCI System and support systems:
A.
What is the driving head supply for the barometric condenser spray?
B.
Where can the barometric condenser condensate pump discharge water?
(two systems or locations required for full credit)
C.
The vacuum pump discharges into what system?
QUESTION 2.06 (1.50)
Starting from the SBGTS inlet valve and stopping at the blower, list the components below in order of flow path repeating items from list as necesssary.
A.
Heater B.
High ef ficiency particulate air filter C.
Moisture seperator D.
Carbon filter E.
Prefilter QUESTION 2.07
(.50)
What supplies the pressure head for gland cooling water to the reactor feed pumps and the heater drain pumps?
(0.5) l i
l
(***** CATEGORY 02 CONTINUED ON NEXT PAGE
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- - = -
1 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 10
, QUESTION 2.08 (1.00)
During single element level control of the feedwater system, the startup level control valve being open allows condensate /feedwater to bypass:
A.
RFP's 2A and 2B B.
FWH's 4A and 4B C.
FWH's 3A and 3B D.
RFP-2A ONLY E.
RFP-2B ONLY QUESTION 2.09 (1.50)
Assuming a diesel generator had been stopped by use of the EMERGENCY STOP BUTTTON, by what three methods can the diesel be restarted?
QUESTION 2.10 (1.00)
List the input signals (reactor power) to Channel B of the Rod DLOCK Monitoring System?
( OTHER THAN THE REFERENCE SIGNAL )
(1.0)
QUESTION 2.11 (1.00)
What is the bases for allowing the Rod Block Monitor System to be automatically bypassed when a peripheral control rod is seleted?
(1.0)
QUESTION 2.12 (1.00)
Which core position is common to all channels of the Transversing (1.0)
Incore Probe System?
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
i 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 11 OUESTION 2.13 (1.00)
. List the emergency bus power supplies to each of the four core spray pumps.
(1.0)
.(Both Units 1 & 2)
QUESTION 2.14 (2.00)
-What-is the minumum flow setpoint for each of the below systems?
(Minumim flow setpoint = That value at which the min flow valve closes)(2.0)
A.
LPCI B.
CORE SPRAY-C.
HPCI D.
RCIC QUESTION 2.15 (1.50)
The following questions concern the Standby Liquid Control System.
(1.5)
- 1. _ Why are the relief valves installed with " flooded" discharge lines?
2.
Why are check valves installed in the discharge lines of each pump?
3.
Which of the below selections best describes the location of the.
discharge line check valves 7 a.
Between valve F003 (discharge isolation) and valve F004 (squib valve) b.
Between valve F003 (discharge isolation) and the pump c.
Between valve F003 (discharge isolation ) and the tee for the relief valve / accumulator d.
down stream of F004 (squib valve)
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CATEGORY 02 CONTINUED ON NEXT PAGE
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PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 12 OUESTION 2.16 (2.00)
The turbine control valve (E41-V9) for the HPCI turbine is maintained CLOSED in standby.
The governor valve (E51-V9) on the RCIC turbine is maintained OPEN in standby.
Why is the position of these two valves (2.0) different?
QUESTION 2.17 (1.00)
The following questions concern the Reactor Protection System.
A.
List the compainion APRM for IRM H.
(0.5)
B.
When is the APRM/IRM (companion) scram automatically bypassed?
(0.5)
QUESTION 2.18 (2.00)
Figure 5 of this exam is a power to flow map.
Identify the lettered lines on the map from the below list.
(2.0)
ITEMS 1.
Jet Pump NPSH Limit Line 2.
Minimum Power Line 3.
APRM Rod Black Line 4.
Design Flow Control Line QUESTION 2.19 (1.00)
List the heating medium for the first stage and second stage reheaters?
( BE SPECIFIC )
(***** END OF CATEGORY 02 *****
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PAGE 13
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3.
INSTRUMENTS AND CONTROLS QUESTION 3.01 (2.00)
List four of the si:t automatic isolation signals for the Reactor Water Cleanup System.
Se.tpoints are required for full credit.
(2.0)
OUESTION 3.02 (2.00)
List four of the five parameters providing trip signals to bypass and/or isolate the AOG Charcoal Absorder System and include the setpoints.
(2.0) t DUESTION 3.03 (2.00)
-List the four requirements and associated setpoints for the Source Range Monitor System to generate a Rod Block.
QUESTION 3.04 (2.00)
A When is the IRM "DOWNSCALE" Rod Block bypassed?
(Two requirements for full credit)
(1.0)
B List the three trip signals for the IRM "INOP" Scram AND (1.0) the requirements to bypass the INOP scram.
QUESTION 3.05 (1.50)
Identify the below as TRUE or FALSE 1.
The RHR Pump Room Fan Cooling Units will start when an RHR pump starts if the control switch is in AUTO.
2.
The Core Spray Pump Room Fan Cooling Units will start when the core spray pump in that room starts if the control switch is in AUTO.
3.
The Reactor Building HVAC supply and exhaust fans automatically l
stop and the dampers remain open on a high pressure in the dry well.
i
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i 3.
INSTRUMENTS AND CONTROLS PAGE 14 OUESTION 3.06 (2.00)
A.
List the parameters and setpoints which cause an EOC-RPT trip.
(1.0)
B.
List the parameters and setpoints which cause an ATWS-RPT trip.
(1.0)
'OUESTION 3.07 (1.00)
List the trip setpoint including direq tod and resultant system response for Unit l's "RECIRC PUMP A SUC IONJREBSURE" alarm.
QUESTION 3.08 (1.00)
List the Trip Setpoint and Fun ort or 2's "RECIRC PUMP B SUCTION
/'
- ' ~ " - '"
(1.0)
PRESSURE" alarm.
/
QUESTION 3.09 (2.25)
The Recirculation flow control system employees the use of two speed limiters.
A.
What is the setpoint of each of the limiters?
(0.5)
B.
When is each of the limiters automatically bypassed?
(1.0)
C.
What does each of the limiters protect against?
(0.75)
QUESTION 3.10 (2.50)
List five of the Group 1 Isolation signals and their respective setpoints.
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
s INSTRUMENT @_AND CONTROL @
PAGE 15 3.
4-GUESTION 3.11 (1.50)
Gervice air compressors A, B,
& C are operating in the below listed arrangement for the pressure control selector switches; Compressor Pressure Control Switch Position A
LOW B
HIGH C
MED The crosstie valve has been Closed but i s now OPENED.
List the LOAD and UNLOAD setpoints for each compressor.
QUESTION 3.12 (2.00)
List four of the " GENERATOR INITIATED" Diesel Generator Trips.
(2.0)
(ASSUME THE DIESEL WAS RUNNING DURING " NORMAL" DIESEL STARTUP)
(Device number not required)
QUESTION 3.13 (1.50)
While the plant is being operated in three-element level control, list the change in reactor vessel level (increase, remain constant, I
or decrease) due to each of the below signal failures.
A.
Fail one main steam line flow instrument HIGH.
I B.
Fail one main feed water flow instrument HIGH.
C.
Fail the selected level instrument HIGH.
QUESTION 3.14 (1.50)
When are the RWM and RSCS systems automatically bypassed, and in each case, what parameter is used to initiate the bypass (1.5) j action (include setpoints)7 1
l
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
c,
-n.
-.m.,
.,-,,,y
-m
..-..w.,_-,,,,,nym.-%
,ww -r,,_
,n-y m e,-.,-.m..,,-.y vw rm ~ ~, - - - -,
PAGE 16 21__lN@lRUMENTS AND_CONTRgLE QUESTION 3.15 (2.00)
The reactor is operating at 90% power when the LOAD REJECT circuitry is actuated.
A.
What actuates the LOAD REJECT circuit?
(1.0)
(Set points required)
B.
What prevents the turbine from overspeeding after the load rej ect?
(0.5)
C.
How is " LOAD REJECT" trip reset?
-( 0. 5)
(***** END OF CATEGORY 03 **
i 4.
PROCEDURES - NORMAL _ABNgBMAl _EMEBGENCY_AND PAGE 17 t
t BBQIgLQGIC@L_CgN189L QUESTION 4.01 (1.00)
List two'of the four authorized uses of " JUMPERS" at BSEP.
(1.0)
(Reference AI-59)
,~;
1 QUESTION 4.02 (1.00)
Per 01-13, " Locked Valve Identification and Locking", STATE the proper method for CONFIRMING valve position.
A.
Turn the valve hand wheel in the OPEN direction; confirm the position by observing that the stem travels in the OPEN direction.
B.
Turn the valve hand wheel in the CLOSED direction; confirm the position by valve position indicator and/or firm tightness.
C.
Turn the valve hand wheel in the DESIRED POSITION direction; confirm the position by valve position indicator and/or firm tightness.
D.
Turn the valve hand wheel in the DESIRED POSITION direction; confirm the valve position by observing that the stem travels in the DESIRED direction.
QUESTION 4.03 (1.00)
_____________ clearance.
(Definition)
Issued to a foreman or lead man who is responsible and cognizant of two or more jobs being performed within the boundary of this clearance.
Fill in the blank with one of the following.
A.
Unit B.
Individual C.
Multiple D.
Master i
i
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
4
_,_....,.,r,.
m_,,,,,.,7.--%,
y_...m,
.,my-
4.
PROCEDURES - NORMAL _ABNQRMAL _EMER@gNCy_ANQ PAGE 18 2
1 68919LggIC9L_C9NIBgL
-QUESTION 4.04
.(1.50)
You enter an area posted ONLY with the following sign:
CAUTION HIGH RADIATION AREA LIST the three (3) authorized methods which can be used to monitor-your exposure.
QUESTION 4.05 (3.75)
A. A fire of unknown orgin breaks out in the Control Room resulting in heavy smoke. The Shift Foreman makes the decision to evacuate the Control Room. As the Unit Control Operator what are 8 of the 10 actions you should take prior to leaving the (2.5)
Control Room (per AOP-32) 7 B. If you could take NO actions prior to leaving the Control Room, what actions should you take outside the Control Room and WHERE would you take them (per AOP-32)7 (1.25)
OUESTION 4.06 (3.00)
A. What are the two entry conditions for procedure EOP-01-LEP-03, (2 3 0.5 ea.)
Alternate Boron Injection 7 B. List four systems that may be used per procedure LEP-03 to (4 3 0.5 ea.)
inject baron.
(*****
CATEGORY 04 CONTINUED ON NEXT PAGE *****)
4.
PROCEDURES - NORMAL _9BNgBM@L _gMgBggNCY_9NQ PAGE 19 1
1 58DigLgglg8L_ggNIggL QUESTION 4.07 (1.50)
OI-1, Operating Principles and Philosophy, defines the terms "shall", "should", and "may" as they apply to BSEP procedures.
(1.5)
Briefly define these three terms per 0I-1.
QUESTION 4.00 (1.00)
You may install an " operator aid" when professionally constructed and approved by......(choose one)
]
A.
NRC licenced operator B.
SF/ SOS C.
2 SRO's D.
Manager-Operations QUESTION 4.09 (1.00)
Per the BSEP Radiological Emergency Plan:
A.
STATE the NORMAL OSC building location.
(0.5)
B.
STATE the NORMAL EOF building location.
(0.5)
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
4.
PROCEDURES - NORMAL _ABNgRMAL _ EMERGENCY _AND PAGE H2O 1
1 88919L991C8L_CgNIBgL QUESTION 4.10 (3.50)
A.
The main turbine is on a roll up (increasing speed) to 1800 RPM.
List FOUR conditions that could occur where the operator should " Trip the turbine" in accordance with GP-03 " Unit Startup and Syncronization".(Setpoints not required)
(2.0)
B. After depressing the 100 rpm speed select push button for a turbine start-up, you should verify valve motion and light indication. Put the following in the order that you would see them per GP-03, Unit Startup and Sychronization.
(1.5) 1.
Intercept valves #1 and 3 - open slowly 2.
Main stop valves #1,3, and 4 - open slowly 3.
Increasing speed light cames on 4.
Main stop valve #2 - begins to open 5.
Control valves - throttle open 6.
Intercept valves #2 & 4 - start to open QUESTION 4.11 (1.00)
EOP-01-UG states in caution #15 to open the SRV's in the following sequence:
A, E,
J, B,
F, D,
G, C,
H.
Explain WHY this sequence and WHY SRV's K and L are not on the list.
QUESTION 4.12 (1.50)
List the minimun level of authority required to approve the below listed procedure changes:
(0.5)
A.
Departure from Established Procedures (1.0)
B.
Deviation from Established Procedures QUESTION 4.13 (2.00)
A.
During the performance of GP-02, how is " Correct SRM/IRM Overlap" demonstrated?
B.
What action and notification is required if correct overlap is not verified?
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
PAGE 21 4___PBQQEgUBE@_;_Ng8M862_8pNgBM861_EMEBgENgY_8ND 68DI96991986_99NI696 OUESTION 4.14 (1.00)
List the minimun number and location of operable SRM's during core (1.0) alterations.
QUESTION 4.15 (1.00)
What is the minimun authority level required for a reactor operator to work more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any seven-day period (not including shift turnover time).
QUESTION 4.16 (1.00)
Concerning use of control rods during plant start up:
After withdrawing a control rod to position 48, an attempt is made to withdraw it again.
What are the two indications that procedure OP-07 instructs the RO to watch for that would indicate the rod is uncoupled.
(1.0)
QUESTION 4.17 (1.50)
List the three MAJOR classifications of " Clearances".
(1.5) y I Tl {
n ty y
(*****
END OF CATEGORY 04
- )l Vj l
(*************
END OF EXAMINATION ***************)
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~
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MODEPaTOR TEMPERATURE COEFFICIE::T versus MODERATO TDiPERATURE I
FIGURE 3 I
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a tt3 Mod ivWt:3ML C31Tu 1/430uld REACTOR POWER VS CORE FLOW OPERATING MAP FIGURE 5 a
EQUATION SHEET f =.ma v = s/t Cycle efficiency = (,1er work cut)/(Energy in) 2 s = V t
- 1/2 at s = mg 3
E = MC (E = 1/2 mv a = (Vf - V )/t A = AM A=Ae 3
3 PE = mgn v = V, + at
- = a/t 1=
In2/t1/2 = 0.693/t1/2 f
2 t
8
- b *1/'M y,y 3p n0 1/2 A=
((c /2I
- II } 3 4
l b
d = 931 sn
- Y Ao c
av t,te o Q = mCpat I=Ie~#
Q = UA AT g
I = I,10-* N '
Pwr.= W d y
TVL = 1.3/u P = P l0,r(t) su HVL = -0.593/u a
P = P e*/'
o SUR = 26.06/T SCR = S/(1 - K,g)
CR = S/(! - K,g x) x CR (1 - K,ff3) = CR II ~ keff2)
SUR = 25a/t* + (a - o)T j
2 T = ( t=/o ) + [(s - o )/ Io ]
M = 1/(! - K,g) = CR /CR,
7 T = 1/(o - a)
M = (1 - K,ffa)/(1 - K,ffj)
T = (a - o)/(Io)
SDM = ( - K,ff)/K,ff L' = 10 seconcs a = (K,g-1)/K,g = $,,g/K,g I = 0.1 seconds-o = [(t*/(T Keg)3 + CI,ff (I + IT)3
/
ll*Id2 =2 2 Id i
l P = (r4V)/(3 x 1010)
Id Id j
g2 2
l r = SN R/hr = (0.5 CE)/d (meters)
R/hr = 6 CE/d2 (f,,g)
Water parameters Miscellaneous Conversions l
1 gal. = 8.345 lem.
I curia = 3.7 x 1010dos 1 ga]. = 3.78 liters 1 kg = 2.21 lem 3 Stu/nr 1 ft3 = 7.48 gal.
1 np = 2.54 x 10 Density = 62.4 lbrp/ft3 1 m = 3.41 x 100 5tu/hr Density = 1 gm/c:9 lin = 2.54 cm Heat of vaporization = 970 Stu/lbm
- F = 9/5'C + 32 Heat of fusion = 144 Stu/lbm
'C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. Hg.
1 BTU = 778 ft-lbf I f t. H O = 0.4335 lbf/in.
2
,, ~..
.,n-
1.
-PRINCIPLES OF NUCLEAR POWER PLANT OPERATION PAGE 22 2
IHEBdgDyNgd]Cg1_HEgI_IggNSEgg_gND_ELUID_ELgW ANSWERS -- BRUNSWICK 1&2
-86/05/19-SPENCER, M.
ANSWER 1.01 (2.00)
FAILURE MECHANISM LIMITING CONDITION 1.
LHGR A2 B1 2.
APLHGR A3 B3 3.
MCPR Al B2 (6 @ 0.33 ae = 2.0)
REFERENCE BRUNSWICK 1 & 2, Student Study Guide, 06-2-A, item 3.0.
ANSWER 1.02 (2.00) 1.
DECREASE 2.
INCREASE 3.
DECREASE (4 @ 0.5 ea = 2.0) 4.
DECREASE REFERENCE BRUNSWICK 1 & 2, Themodynamics Heat Transfer & Fluid Flow, Chapter 7.
ANSWER 1.03 (1.50) c.
Conduction b.
Convection (conduction also accepted if both given) c.
Radiation (3 @ 0.5 ea = 1.5)
REFERENCE BRUNSWICK 1 & 2, Thermodynamics Heat Tranfer and Fluid Flow, Chapter 8.
kGf1C]PY
PAGE 23 1-PRINCIPLES OF NUCLEAR POWER _ PLANT _gPERATIgN2 ISEBdQQXN9dICg2_ DEST TR8NgEg8_8ND_E69]p_ELgW ANSWERS -- BRUNSWICK 1&2
-86/05/19-SPENCER, M.
dLASTt?tw?V rn a.
ANSWER 1.04 (2.50) a.
Doppler or fuel temperature b.
void c.
moderator temperature d.
void (5 9 0.50 ea = 2.5) e.
moderator temperature REFERENCE BRUNSWICK 1 & 2, Student Study Guide, 02-2-A, Chapter 12.
BRUNSWICK 1 & 2, Student Study Guide, 02-2-A, Chapter 12.
ANSWER 1.05 (2.00) a.
Decreases [0.253.
There is less steam flow, therefore, less pressure drop through the main steam lines CO.753.
(1.0) b.
Increases CO.253.
With the same amount of cooling water through the condenser and less heat load, condensate depression (1.0) will increase [0.753.
REFERENCE BRUNSWICK 1 & 2, SSM CONDENSATE AND FEEDWATER SSM MAIN STEAM AND EHC ANSWER 1.06 (1.00)
(1,g) b REFERENCE NUS:
Vol 3,
pp 6.1-3 BRUNSWICK 1 & 2, Student Study Guide, 02-2-A i
l O
1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATIONt PAGE 24 THERdODyNAdlC@t_ HEAT _TRAN@FER_AND_ FLU _I_D FLOW ANSWERS -- BRUNSWICK 1&2
-86/05/19-SPENCER, M.
ANSWER 1.07 (1.00)
(1.0) c REFERENCE BFNP: XENON & SAMARIUM LP,-P.4,12 GGNS: LP OP-NP-514, p.
5-10 BRUNSWICK 1 & 2, Student Study Guide, 02-2-A.
4 ANSWER 1.08 (1.00)
(1.0) d REFERENCE SSM BOOK 2, CH 2-A, SEC 13.7, PG 161 ANSWER 1.09 (2.50) 1
- a. No bypass valve action b.
TCV throttle close due to indicated low pressure then open as backup r.egulator responds
- c. Reactor pressure will increase 10 psig (to *1020)
- d. Throttle pressure will increase 10 psig (to '959) 9.
Reactor power will increase (*1.0%)
(5 9 0.5 ea = 2.5)
REFERENCE BRUNSWICK 1 & 2, Student Study Material, 19-2-B, Figures 13 and 17.
l i
add figures to exam I
I l
.__._,,m
PAGE 25 1.-
PR I NC I PLEg_gF _NUC'_E AR _ P OWER _P L ANT _gP ER AT I QN r IHERMgDYNAM_ICS,'jEAT_TRAN@FER_ANQ_FLUIQ_FLgW t
-86/05/19-SPENCER, M.
ANSWERS -- BRUNSWICK 1&2 ANSWER 1.10 (1.00)
(1.0) a REFERENCE BSEP T.S.
3/4.2.4 LHGR LCO, pg 3/4 2-15 BRJNSWICK 1 & 2, Student Study Guide 06-2-A, page 11.
-l ANSWER 1.11 (2.00)
The bundle power which would cause the onset of transition boiling m.
(1.0) at some point in the assembly.
(1.0) b.
2 REFERENCE BFNP TRANSITION BOILING & ATLAS TESTING LP,P.5-6 GEXL CORRELATION & CRITICAL POWER LP,P.3 l
GGNS MCD, THERMAL LIMITS, P.26,32-33 BRUNSWICK 1 & 2, Student Study Guide, 06-2-A, page 18.
4 ANSWER 1.12 (1.00)
A = Late in core life B = Early in core lif e i
REFERENCE l
BRUNSWICK 1 & 2 -Student Study Guide, 02-2-A, Figure 45.
i ANSWER 1.13 (1.00)
(1.0)
C.
I i
REFERENCE l
BRUNSWICK 1 & 2, Student Study Guide, 02-2-A, page 194.
i-i
--,,_.---~,.n..,.
-,~.---._,,--.-._--,-.._,----,l.--,.,-r.----
1.
PRINCIPLES OF NUCLEAR POWER PLANT __gPgBATIgN PAGE 26 2
IUEgdggyNgdigg1_HggI_IggNSEg8_9Np_ELUlp_ELgW ANSWERS -- BRUNSWICK 1&2
-86/05/19-SPENCER, M.
ANSWER 1.14 (2.00) ie-a.
Immediately:
6% X 2436 MWT = 146.16 MWT(Accept-5,0%~+1 or.5%)
(0.5) 1 minute:
4.0% X 2436 MWT = 97.4 MWT(Accept 4.0% +1 or.5%)
(0.5) b.
The nuclear instrumentation discriminates against the small pulse (1.0) height of the decay gamma.
REFERENCE BRUNSWICK 1 & 2, Student Study Guide, 02-2-A, page 17.
ANSWER 1.15 (2.50)
REFER TO STEAM TABLES:
900 PSIA =532 DEGREES F 610 PSIA =488 DEGREES F 532 F - 488 F = 44 F/ HALF HOUR OR 88F PER OUR (1.5) 3.
88 F DOES NOT EXCEED THE COOLDOWN LIMIT f /0 (0.5)
REFERENCE DRUNSWICK 1 & 2, Steam Tables GP-01, Reactor Startup ANSWER 1.16 (2.00) i Power at setpoint= (118%) (2436 MW) = 2874 MW
(.5) t/T 10sec/10sec Peak Power = (2874MW) e
= 2874 e
= 2874 e (1.0)
(.5)
=7,813.6MW or 7.81GW REFERENCE BRUNSWICK 1 & 2, Student Study Guide, 02-2-A, page 142 i
l l
b t
PAGE 27 A __P(8NI_Dggl@N_ INCLUDING _g8EEIy_8ND_gMg8@gNCy_gygIgMg ANSWERS
- BRUNSWICK 1&2
-86/05/19-SPENCER, M.
4 Y
ANSWER 2.01 (1.50) 9.4 (0.5)
A.
Depressing both timer reset push buttons (Or5)
,,. /,
)
Securing all LP ECCS pumps,f..,f f s.j /
-,(
(0.5)
B.hEOpsig
( + or - 5 psig ) :c j > j.j 2
n / n I (..
11.., i..., p
';/ I i>
REFERENCE BRUNSWICK 1 & 2, BK, RTN 013 BK, HO-14-2/3-F, PAGE 20.
ANSWER 2.02 (2.00) 1.
=A 2.
=B
(
3.
=C (4 0 0.5 ea = 2.0) 4.
=D REFERENCE l
BRUNSWICK 1 & 2, Student Study Material, 09-2-B, page 28 (Fig 2)
ADD FIGURE TO EXAM i
f i
1 i
i h,h, ( ". ?., qR f* TV,
y m
T t.
i A
I dV4 _j Vjhj j
lt 2[__ PLANI _QESl@N_INCLUplN@_@@EEIy_@NQ_EMER@ENCY_@YSIEM@
PAGE 28 ANSWERS
- BRUNSWICK 1&2
-86/05/19-SPENCER, M.
ANSWER 2.03 (1.25) 1.
EHC 2.
Feedwater Control 3.
Stean Jet Air Ejector Logic 4.
Reactor Manual' Control 5.
6.
Turbine Supervisory Instrumentation 7.
Procass Computer 8.
Reactor Level Indication 9.
BOP [ indication of flow, level, pressure, etc of plant systems susch as main stream, feedwater, condensate, service water, service air, RBCCW, TBCCW, condenser, MSR, and Megawatt meters]
4
.r,.i
/ L..t-i,A >
(any 5 9 0.25 ea = 1.25) 10, h..
t-, f, c l p,,,,,
.4,, )
p
/r REFERENCE BRUNSWICK 1 & 2, Student Study Material, 20-2-F, page 26 and 27.
T ANSWER 2.04 (3.00) 1.
Assurance that adequate core cooling takes place to prevent overheating of reactor fuel (max 2200 F) in the event of a small break LOCA.
2.
~ Support a plant shutdown by maintaining sufficient reactor water inventory until the reactor is depressureized to a pressure where the LPCI System or Core Spray System can be placed in operation.
3.
Provide the capability of fulfilling the objectives of the RCIC System in the event that the RCIC system is inoperable.
m.'
4.
Provide automatic operation and capability of startup operation independent of auxiliary AC power, service air, or external cooling water systems.
(any 3 S 1.0 ea = 3.0)
2 __P68NI_pE@l@N_lNC6UQ1N@_@@EEIY_8NQ_EMER@ENCY_@y@IEM@
PAGE 29 t
ANSWERS -- BRUNSWICK 1&2
-86/05/19-SPENCER, M.
REFERENCE BRUNSWICK 1 &2 Student Study Material, 14-2-B, page 2.
ANSWER 2.05 (1.50)
A.
HPCI booster pump B.
1.
Suction of the HPCI pump (0.25) 2.
Clean Radwaste System (0.25)
C.
Standby Gas Treatment System (3 0 0.5 ea = 1.5)
REFERENCE BRUNSWICK 1 & 2, Student Study Material, 14-2-B, page 11.
ANSWER 2.06 (1.50) 1.
C 2.
A 3.
E 4.
B 5.
D 6.
B (6 @ 0.25 ea = 1.5)
REFERENCE BRUNSWICK 1 & 2, Student Study Guide, 15-2-F, pages 6, 7,
& 8.
ANSWER 2.07
(.50)
The condensate booster pumps.
(0.5)
REFERENCE BRUNSWICK 1 & 2, Student Study Guide, 17-2-B, page 3.
I f
i i
PAGE 30 2 __E60NI_DggigN_JNC6Ug1Ng_ggEEIy_gND_EdgRggNCy_gygIgM!
ANSWERG -- BRUNSWICK 162
-96/C5/19-SPENCER, M.
ANSWER 2.00 (1.00)
B.
REFERENCE BRUNSWICK 1 & 2, Student Study Guide, 17-2-B, Figure one.
ANSWER 2.09 (1.50) 1.
pushing the EMERGENCY STOP RESET BUTTON (then normal startup) 2.
pushing the EMERGENCY START BUTTON - tu. dy, (i
43 0 0.5 ea = 1.5) 3.
an auto start signal is received y{ n. d r.>ia y t 6 p,, sa. a.,,,.,,,,., -,:,
REFERENCE BRUNSWICK 1 & 2, Student Study Material, 20-2-D, page 26.
ANSWER 2.10 (1.00)
(2 9 0.5 ea = 1.0)
B and D LPRM Levels REFERENCE BRUNSWICK 1 & 2, Student Study Material, 25-2-E, page 4.
ANSWER 2.11 (1.00)
Testing has shown that a peripheral rod cannot cause thermal (1.0) limits of the core to be exceeded.
REFERENCE BRUNSWICK 1 & 2, Student Study Material, 25-2-E, page 5.
i ANSWER 2.12 (1.00)
C<yhr - e l' - Gr *
/lf /df6 (1.0)
~~
Position 28-29 ce W
,[,
.,m i,-..
,,.,-__.,,-.,-_3--,.
rw,- - -. _ _.
g
PAGE 31 2 __ELONI_pg@lgN_JNCLUDJN@_@SEgly_gNp_gdg895NCY_@YSIgd!
{
ANSWERS -- BRUNSWICK 1&2
-86/05/19-SPENCER, M.
l i
REFERENCE BRUNSWICK 1 & 2, Student Study Material, 75-2-F. pace 9.
l 4
l ANSWER 2.13 (1.00) 1A = El i
1B = E2 2A = E3
-(4 @ 0.25 ea = 1.0) 2D = E4 REFERENCE BRUNSWICK 1 & 2, Student Study Material, 14-2-E. page 10.
ANSWER 2.14 (2.00)
I A.
3000 gpm
(+/- 5 */. )
D.
475 gpm
(+/- 5%)
C.
800 gpm
(+/- 5%)
D.
80 gpm
(+/-
10%)
(4 9 0.5 ea = 2.0) l REFERENCE BRUNSWICK 1 & 2, Student Study Material, 14-2-D, page 25 l
14-2-E, page 10 14-2-B. page 18 l
l 14-2-C, page 27 ANSWER.
2.15 (1.50)
{
i 1.
To prevent evaporation and precipitation within the valves.
2.
To prevent bypass flow from one pump to the other pump in case of relief valve failure.
i (3 @ 0.5 ea = 1.5) f 3.
c l
l i
I
. _, ~ -
2___PL8NI_pgglgN_INCLUplNg_g8FEly_9Np_ EMERGENCY _@y@TEME PAGE 32 j
ANSWERS -- BRUNSWICK 1&2
-86/05/19-SPENCER, M.
REFERENCE Di?UNSWICK 1 & 2, Student Study Material.
14-2-H, pace 7 and figure 1.
ANSWER 2.16 (2.00)
The HPCI system employs a DC powered oil pump to open the control valve upon initiation [1.03.
The RCIC system employs a shaft dirven oil pump to operate the governor therefore, the governor valve must be initially (2.0) open[1.03.
REFERENCE BRUNSWICK 1 & 2, Student Study Material, 14-2-B and 14-2-C.
ANSWER 2.17 (1.00)
(0.5)
A.
APRM F B.
Automatically bypassed when MODE SWITCH not in RUN (0.5)
REFERENCE BRUNSWICK 1 & 2, Student Study Material, 29-2-A, page 36.
ANSWER 2.18 (2.00) 1.
=B 2.
=A 3.
C or E 4.
F REFERENCE BRUNSWICK 1 & 2, Student Study Material, 10-3-A, page 94 (Figure 20)
-n.
.7,-,--.---n,-
, ~,,, - - -., -
m
,--nn..-,~na
v.*
PAGE 33 A __Pb8NI_pgglgN_ lng 6Up1Ng_ggggly_9Np_gME6ggNgy_gygIgMg ANSWERS -- BRUNSWICK 1 ?x2
-86/05/19-SPENCER, M.
ANSWER 2.19 (1.00)
First stage = e:: traction steam f rom the second stage of the high
(
f<
(her pressure turbine.
(
Second stage = steam directly from the main steam system
- reference DRUNSWICK 1 Fx 2, Student Study Material, 18-2-A, page 5.
I t
4 MS"RCCPY i
PAGE 34
_3 __INgIRUMENIg_gNg_CgNTRgLg 2
ANSWERS -- BRUNSWICK 1&2
-86/05/19-SPENCER, M.
' ANSWER 3.01 (2.00) 1.
Low Reactor Water Level LL#2 or 112 inches 2.
Standby Li quid Control System Initiation
.3.
RWCU High Differential Flow 53 gpm 4.
RWCU Area High Temperature 150 F 5.
RWCU Area Ventilation High Temperature Delta temp of 50 F 6.
RWCU Nonregenerative Heat Exchanger Outlet High Temperature 140 F ALL VALUES + OR - 10.0 %
ers-f or parameter 4r5 for setpoint (4 a 0.5 ea = 2.0)
.75
- 0. 3 s
- M
- i. s.
REFERENCE BRUNSWICK 1 & 2, Student Study Material, 11-2-A, page 13 ANSWER 3.02 (2.00) 2%
1.
High hydrogen downstream of the recombiner system 95 %
2.
High condensate drain level 45 F 3.
High moisture downstream of guard bed 14 " from tank top 4.
Gross glycol leak 150 cfm 5.
High-high flow (0.25 for parameter & 0.25 for setpoint) (2.0) kil-Ye Iv < > + cr - /0 $,
REFERENCE BRUNSWICK 1 & 2, Student Study Material, 16-2/3-C, page 24.
I i
i
PAGE 35 3 __INgIRUMENIg_8ND_CgNIRgLE ANSWERS -- BRUNSWICK 1&2
-86/05/19-SPENCER, M.
ANSWER 3.03 (2.00) 1.
SRM Downscale ( < 3 cps ) with IRM's on range 1 or 2.
2.
SRM Upscale 9 ( > 1.0 E5 cps ) with IRM's on range 7 and below.
3.
Detector not full in with SRM reading less than 100 cps and IRM's on range 1 or 2.
(4 9 0.5 ea = 2.0) 4.
SRM Inoperative REFERENCE BRUNSWICK 1 & 2, Student Study Material, 25-2-A, page 28.
ANSWER 3.04 (2.00)
A.
Range 1 E0.53 or Run Mode CO.53 (1.0)
B.
Setpoints 1.
Hi Voltage low 2.
Module unplugged 3.
Function Switch not in Operate (3 0 0.16 ea = 0.5)
BYPASSED -- In the RUN Mode CO.25] with companion APRM on scale CO.25]
(0.5)
REFERENCE BRUNSWICK 1 & 2, Student Study Material, 25-2-B, page 17.
ANSWER 3.05 (1.50) 1.
FALSE 2.
TRUE (3 0 0.5 EA = 1.5) 3.
FALSE REFERENCE BRUNSWICK 1 & 2, Student Study Material, 36-2-A, pages 25, 26, and 31.
PAGE 36 3___JNgIBUMENIS_8Np_CgNIgg(g ANSWERS -- BRUNSWICK 1&2
-86/05/19-SPENCER, M.
ANSWER 3.06 (2.00)
A.
1.
Turbine control valve EHC pressure is less than 500 psig. [0.53 2.
Turbine stop valve is greater than 10 % closed. CO.53 (1.0)
B.
1.
Reactor vessel l evel is less than 112 inches. [0.53 2.
Reactor pressure is greater than 1120 psig. CO.53 (1.0) y c,- - h '7e )
{ $i. vada. >
REFERENCE BRUNSWICK 1 & 2, Student Study Material, 28-2-A, page 51.
-ANCWER -
-3.07 '-
(1.00)
Setpoint = 135 psig ( +/- 6) increasing (0.5) ft Function = a. Cipses RHR,sucti n cooling inboard isolation valve (E41-F009)/ -
I L-L -
b.
C oses Reactor Vessel inboard head spray valve
$51T-FA2-2 F '
(Valve number not required for full credit)
(2 9 0.25 ea = 0.5)
REFERENCE BRUNSWICK 1 & 2, Student Study Material, 10-2-A, page 63.
^NSWEP
-3 r08 C (1.00)
Setpoint = 166 psig
(+
/';-6) i lI l
Function = a.
Closes RiR injWcti alve (E11-F015B)
(1.0)
, 4.U V
.pening of RHR injection valve (E11-F015A) b.
Blocks (Valve numbers not required for full credit)
(2 9 0.25 ea = 0.5)
REFERENCE BRUNSWICK 1 & 2, Student Study Material, 10-2-A, page 63.
1 l
PAGE 37 3 __INgIBUMENIg_ANp_C9NI69L5 ANSWERS -- BRUNSWICK 1&2
-86/05/19-SPENCER, M.
ANSWER 3.09 (2.25)
(0.25)
A.
Maximum of < 28 %
(0.25)
Maximum of 45 %
B.
28 % bypassed = Recirc pump discharge valve is > 90 % open (0.25) and feedwater system flow > 20 %.
(0.25) 0:1}
45 % bypassed = One er -cre-reactor f eed pumps are at > 20 %
(0.25) rated flow 2 /EE". l4 y//4-reactor water levelg el 2r-i w sceived-tL-?
(0.25)
LL- ? [31182 "
required for full credit)
Check at facility for value of LL ?
C.
28 % = To prevent runnung the recirculation pump at high speed with the discharege valve only partially open (overheating)
(0.25) and to provide NPSH protection for the recirculation and jet pumps at low feedwater flow (inadequate subcooling)
(0.25) 45 % = The feedwater system will be able to maintain or recover reactor water level on loss of reactor feed pump.
(0.25)
REFERENCE BRUNSWICK 1 & 2, Student Study Material, 10-2-A, page 28.
4 l
PAGE 38 3 __INgIBUMENIS_6ND_CQNI69LE ANSWERS -- BRUNSWICK 1&2
-86/05/19-SPENCER, M.
ANSWER 3.10 (2.50) 1.
Low-Law reactor water level (LL2)
>.112 inches 2.
Main steam line high radiation 3 X full power background 3.
Main steam line high flow 140% normal UNIT 2 40% in SU yr 4.
Reactor building steam line tunnel high temperature
\\200F ft 5.
Turbine building main steam tunnel high temperature
$200F 7t' 6.
Low condenser vacuum i 7" Hg 7.
Low steam line pressure
[Run ONLY3 825 psig (0.25 for parameter & 0.25 for setpoint)
(10 0 0.25 ea = 2.5) 1 REFERENCE BRUNSWICK 1 & 2, Student Study Material, 17-2-A, page 10.
ANSWER 3.11 (1.50)
COMPRESSOR LOAD UNLOAD A
103 111 B
107 114 C
107 114 (6 0 0.25 ea = 1.5)
(all values +/- 2 psig)
REFERENCE 2
BRUNSWICK 1 & 2, Student Study Material, 21-2-A, page 19.
PAGE 39 3 __INSIByMENIg_6ND_CgNIBQL@
ANSWERS -- BRUNSWICK 1&2
-86/05/19-SPENCER, M.
1 4
ANSWER 3.12 (2.00) 1.
Differential overcurrent 2.
Phase overcurrent with voltage restraint 3.
Loss of excitation 4.
Undervoltage 5.
Reverse power on emergency bus 6.
Reverse power to the generator (any 4 3 0.5 ea = 2.0)
REFERENCE BRUNSWICK 1 & 2, Student Study Material, 20-2-D, page 35.
ANSWER 3.13 (1.50)
A.
Increase B.
Decrease (3 0 0.5 ea = 1.5)
C.
Decrease REFERENCE BRUNSWICK 1 & 2, Student Study Material, 17-2-B, Figure 2.
ANSWER 3.14 (1.50)
SYSTEM WHEN PARAMETER 1RWM 25% POWER (0.5)
STEAM FLOW (0.25)
RSCS 22% POWER (0.5)
FIRST STAGE TURBINE PRESSUE (0.25)
(1.5)
REFERENCE BRUNSWICK 1 & 2, Student Study Material, 27-2-B, page 28.
27-3-C, page 10 & 11.
PAGE 40 3___INgIBUMENIg_8Ng_CgNIggLg ANSWERS -- BRUNSWICK 1&2
-86/05/19-SPENCER, M.
ANSWER 3.15 (2.00)
A.
High pressure turbine exhaust pressure [0.333 compared to stator amps E0.333 indicating a mismatch of 40% E0.33] actuates (1.0) the load reject circuit.
B.
The intercept valves should shut limiting turbine (0.5) speed [0.53.
(0.5)
C.
AUTOMATICALLY REFERENCE BRUNSWICK 1 t< 2, Student Study Material, 19-2-B, pages 31 & 32.
b l
MAHRC3PY
PAGE 41 4___PBggggUggg_;_Nggd863_ggNggdg(2_gdggggNgy_gND 6891969 GIG 86_G9NIgg6 ANSWERS -- BRUNSWICK 1&2
-86/05/19-SPENCER, M.
i.
/.-
,(.
<ANSWE M 4T01-S (1.00) 1.
An electrical component or circuit is out of service, disabling a desired plant component.
2.
The electrical component will be out of service an extended period.
3.
The SF/ SOS determine that it is desirable to reestablish power or control to the desired plant component by installing a jumper or lif ting a lead.
4.
The SF/ SOS determine that it is possible to reestablish power or control to the plant cemponent by installing a jumper or lifting a (any 2 @ 0.5 ea = 1.0) lead.
REFERENCE BRUNSWICK 1 & 2, Administrative Instruction - 59, Rev.
6, page 1.
ANSWER 4.02 (1.00)
B REFERENCE BRUNSWICK 1 & 2, 0I-13, Sect 4.1.1, page 2.
ANSWER 4.03 (1.00)
D REFERENCE BRUNSWICK 1 & 2, AI-58, Section 3.1.6, page 3.
PAGE 42 4 __PBggEgUBEg_;_N98M@L _@gNgBM861_EMEggENgY_8N9 1
689196991986_99NIgg6 ANSWERS -- BRUNSWICK 1&2
-86/05/19-SPENCER, M.
ANSWER 4.04 (1.50) 1)
A radiation monitoring device which continuously indicates the radiation dose. (Pocket Daseimeter) 2)
A radiation monitoring device which continuously integrates the dose rate in the area and alarms when a preset integrated dose is received.
(" Chipper")
3)
An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device.
(3 9 0.5 ea = 1.5)
REFERENCE BRUNSWICK 1 & 2, Radiation Control and Protection, Rev. 18, page 45.
ANSWER 4.05 (3.75)
A.
1.
OBTAIN SHUTDOWN PANEL KEYS 2.
MANUALLY SCRAM THE REACTOR 3.
TRIP THE MAIN TURBINE 4.
- 5. WHEN STEAM FLOW IS < 3.0E6 LB/HR, PLACE MODE SWITCH IN S/D.
6.
TRIP BOTH RECIRC. PUMPS 7.
REDUCE REACTOR PRESSURE TO 700 PSIG WITH BYPASS JACK 8.
PLACE BOOSTER PUMPS IN MAN.
0,10
- 10. ENTER EOP-01o~d<yac le u e&-f c[ 8 4'd-i 42 IA N'
/
(8 9.ar3tr'EA. )
(SPECIFIC SEQUENCE NOT REQUIRED)
B.
1.
Go to the cable spreading room (.25)
- 3. Place RPS alternate feed switch (LG3) to the mid position. (.5)
(1.25)
REFERENCE BRUNSWICK 1 & 2, AOP-32, Rev 7, page 5.
f I
PAGE 43 4:__EBgCgguggg_ _NgBd862_8pNgBd862_gdgBggNCY_8ND B89196991C86_CgNI896 ANSWERS -- BRUNSWICK 1&2
-86/05/19-SPENCER, M.
ANSWER 4.06 (3.00)
(0.5)
A.1. Baron injection is required and the SLC' system is not available.
- 2. The SLC tank is empty and cannot be refilled AND futher (0.5) baron injection is required.
RWCU[viaSLC)
B.1.
condensate syste)m RWCU(with borax 2.
3.
4.
HPCI 5.
RCIC (any 4 S 0.5 ea = 2.0) 6.
CRD REFERENCE BRUNSWICK 1 & 2, EOP-LEP-03 rev 00, pg 3 ANSWER 4.07 (1.50)
A.
SHALL: denotes a requirement B. SHOULD: denotes a recommendation (3 S 0.5 ea = 1.5)
C.
MAY: denotes permission REFERENCE BRUNSWICK 1 & 2, Dill, Rev.13, Sec 2.0, pg 1 ANSWER 4.08 (1.00)
D REFERENCE sec 13, pg 9 BRUNSWICK 1 & 2, DI-1, Operating Principles and Philosophy, PAGE 9 MISSING FROM MATERIAL SENT TO IDAHO CHECK AT FACILITY l
n n
.y,
---+,-----y
.~
~
PAGE 44 4 __PBggEQUEEg_;_NgBd861_8pNgBdg61_EdgBgENgy_8N9 689196991986_99NIBg6 ANSWERS -- BRUNSWICK 1&2
-86/05/19-SPENCER, M.
1 ANSWER 4.09 (1.00)
(0.5)
A.
Service Building (0.5)
B.
Training Building REFERENCE EIH:
63EP-EIP-061-0 BRUNSWICK 1 & 2, Student Study Material, 07-2-V, page 11 & 12.
ANSWER 4.10 (3.50)
A.1.
Turbine shell to rotor diff. expansion indicating in the red band.
2.
Turbine journal bearing high vibration near the critical speeds or any other speeds 3.
High journal bearing metal temp.
4.
High thrust bearing metal temp.
- 5. High diff. temp. between inlet oil and bearing drain oil temp.(4 0 0.5) 6.
Loss of turbine speed control.
(6 0 0.25)
B.
4, 2,
1, 6,
5, 3
REFERENCE BRUNSEICK 1 & 2, GP-03 rev 4, pages 14 & 15.
OP-26 rev 26, pages 32 & 33.
ANSWER 4.11 (1.00)
(0.5)
To distribute heat evenly in the suppression pool.
K & L are not on the list because they discharge near the exhaust (0.5) of HPCI and RCIC REFERENCE BRUNSWICK 1 & 2, EOP-01-UG, Rev 2, caution #15, page 42.
I
. ~.
PAGE 45 4:__PBgCggU6gg_;_Ng6M862_@gNg6M862_gMg6ggNCy_8Ng 889196991C66_C9NIBg6 ANSWERS -- BRUNSWICK 1&2
-86/05/19-SPENCER, M.
ANSWER 4.12 (1.50) or Shift Forman if the Shift Supervisor is A.
Shift Operating Supervisor, unavailable.
[One or the other required for full credit - the Shift Forman answer must include availablity of SOS]
(0.5)
B.
Two members of the plant management staff knowledgeable in the area affected by the procedure E0.53, one of which holds a SRO license on the unit affected and be a supervisor in charge of the shift [0.53 (1.0)
REFERENCE BRUNSWICK 1 & 2, Administrative Procedure, Section 5, pages 5-6 & 5-7.
ANSWER 4.13 (2.00)
A.
Wnen at least three IRM channels in each RPS trip system E0.53 show an ncrease in reading before the first SRM channel reaches 10E5 cps E0.53.
(1.0)
B.
The reactor shall be shut down E0.53 and Nuclear Engineer notifed E0.53 (1.0)
REFERENCE BRUNSWICK 1 & 2, GP-02, rev.
8, page 14.
l ANSWER 4.14 (1.00)
(0.5)
Min Number = 2 l
l Location - One in the quadrant where fuel is being moved CO.25]
l
- One in the adjacent core quadront [0.25]
(0.5) i l
REFERENCE l
BRUNSWICK 1 & 2, GP-07, rev.
4, page 6 l
l
.e u.
4 __PBggEguggg_;_NgBd@62_gBNggd@62_EDEBgENgy_@NE PAGE 46 689196991986_99NIBg6 ANSWERS -- BRUNSWICK 1&2
-86/05/19-SPENCER, M.
ANSWER 4.15 (1.00)
Plant General Manager' REFERENCE BRUNSWICK 1 & 2, Adminstrative Procedure, page 4-18.
ANSWER 4.16 (1.00) 1.
The four rod display indication will go out (0.5) 2.
The " rod overtravel" annunciator will lock in.
(0.5)
REFERENCE BRUNSWICK 1 & 2, BK, OP-07, page 10 ANSWER 4.17 (1.50) 1.
Station 2.
Radwaste (3 0 0.5-ea = 1.5) 3.
Local REFERENCE BRUNSWICK 1 & 2, Administrative Instruction - 58, Rev.
8, page 2.
l j
T n
MASTER CDPY w sree ENCLOSURE 2 U.
S.
NUCLEAR REGULA10RY COMMISSION SENIOR REAC10R OPERATOR LICENSE EXAMINATION FACILITY:
DRUNSWICK 1E2
_____-GE4 DWR REAC10R TYPE:
DATE ADMINISTERED: 06/05/19 EXAMINER:
- LAHYER, S.
APPLICANT:
INSTRUCTIONS TO APPLICANT:
Use separate paper for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question. The passins grade requires at least 70% in each category and a final grade of at least 80%.
Enamination papers will be picked up sin (6) hours after the examination sterts.
% OF CATEGORY
% OF APPLICANT'S CATEGORY VALJE TOTAL SCORE VALUE CATEGORY 27.00 24.83
______-_ 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 27
'S 06
___1_25____[_1__
________ 6.
PLANT SYSTEMS DESIGN, CONTRdL, AND INSTRUMENTATION
_'[8 00_1____
_'[5 75
________ 7.
PROCEDURES - NORMAL, ABNORMAL,
_1__
EMERGENCY AND RADIOLOGICAL CONTROL
_'6.50[______
_'4[_1__7
________ 8.
ADMINISTRATIVE PROCEDURES, 3
CONDITIONS, AND LIMITATIONS 108.75 100.00 TOTALS FINAL GRADE _________________%
All work done on this examination is my oun. I have neither siven nor received aid.
EFetiCEET7s sisRETURE~~~~~~~~~~~~~~
v t
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PACE 2
QUESTION 5.01 (1.00)
Which one of the following statements is N01 true regarding the LHGR (linear heat generation rate) thermal limit?
a.
The LHCR desi3n' limit = 13.4 kw/ft for both 8x8 and P8x8R fuel.
b.
The limit is based on maintaining peak cladding temperature =< 2200 degF.
c.
The LHGR specification assures that the LHGR in any rod is less than the design value even if fuel pellet densification is postulated.
d.
If the limit is exceeded, it could result in fuel clad cracking due to high stress.
QUESTION 5.02 (1.00)
A modcrator is necestcry to slou neutrons down to thermal energies.
Which of the following is the CORRECT reason for operction with thermal inrtead of fast neutrons?
a.
Increased neutron efficiency since thermal neutrons are less likely to leak out of the core than fast neutrons.
b.
Reactors operating primarily on fast neutrons are inherently unstable and have a higher risk of goins prompt critical.
c.
The fission cross section of the fuel is much higher for thermal neutrons than fast neutrons.
d.
Doppler and moderator temperature coefficients become positive as neutron energy increases.
(*****
CATEGORY 05 CONTINUED ON NEXT PAGE
- )
i
I 5.
THEORY OF NUCLEAR PDHER PLANT OPERA 110N, FLUIDS, AND PAGE 3
QUESTION 5.03 (1.00)
Which one of the following statements is CURREC1 concerning the paralleling of electrical systems?
a Although it is desirable to have speed and phase position matched, it is much more important to have voltages matched.
b.
If voltages are not matched at the time the synchronizing switch is closed, there will be VAR flow from the lower voltage source to the higher one.
c.
If the incoming machine is at synchronous speed but out of phase with the runnin3 bus when the breaker is closed, heavy currents will flow to either accelerate or retard the incoming machine.
d.
If the incoming machine is in phase but slightly faster than synchronous speed when paralleled, the system will tend to speed up the incoming machine to synchronous speed.
e.
If the resistances are not matched at the time the synchronizing switch is closed, heavy currents will flow to tend to speed up the incoming machine to synchronous speed.
QUESTION 5.04 (1.00)
The condensate subcooling in a condenser operating at 1 psia with a condensate temperature of 95 desF is approximately:
a.
1.07 de3F.
b.
6.74 desF.
c.
25.3 desF.
d.
102 desF.
l l
e.
196 desF.
l l
l (xxxw* CATEGORY 05 CONTINUED ON NEXT PAGE xxxxx) l l
l f
5.
THEORY OF NUCLEAR POWER PLAN 1 OPERATION, FLUlDS, AND PAGE 4
QUESTION 5 05 (1.00)
Which of the followin3 statements is true regarding intrinsic neutron sources in a shutdown reactor?
a.
Practically all the source neutrons from spontaneous fission come from U-235.
b.
The alpha-neutron source comes from the alpha decay of U-238, U-239 and Plutonium which interact with 0-18 in the moderator to yield neon and a neutron.
c.
The major concentration of source neutrons comes from spontaneous fission of U-238.
d.
The photo-neutron source is less significant at BOL than EOL.
QUESTION 5.06 (1.00)
Adding latent heat to liquid water at saturated conditions will...
a.
increase the temperature of the water.
b.
change the water to steam at the same temperature.
c.
change the water to steam at a slightly hi her temperature.
3 d.
decrease its subcooling by making it boil, e.
increase its subcooling by making it boil.
QUESTION 5.07 (1.00) j The void coefficient of reactivity becomes less negative for which one of the following changes?
I a.
percent void changes from 30 to 40.
b.
average fuel temperature changes from 500 desF to 550 desF.
c.
core age changes from BOC to EOC.
d control rod density chan3es from 20% to 25%.
e.
reactor pressure changes from 980 psi to 1010 psi.
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5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 5
DUESTION 5.08
(.50)
Which one of the graphs on figure 877 (attached) best illustrates the temperature relationships in a parallel flow heat exchanger?
QUESTION 5.09 (3.00)
Following a normal reduction in power from 90% to 70% with recirculation flow, how will the following change (increase, decrease or remain the same) and why?
a.
The pressure difference between the reactor and the turbine steam chest.
b.
Condensate depression at the exit of the condenser.
c.
Final feedwater temperature.
QUESTION 5.10 (1.50)
Reactor startup from cold conditions is in progress.
How would each of the following conditions or events affect the critical rod position (more rod withdrawal, less rod withdrawal, or no significant effect)? No explanation is necessary.
a.
Xe is changing due to extended power operation which was terminated 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> previously.
b.
The reactor head vent is inadvertently closed.
c.
Reactor water clean-up system isolates.
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o 5.
THEORY OF NUCLEAR POWER PLAN 1 OPERATION, FLU 1DS, AND PAGE 6
QUESTION 5.11 (1.50)
The Brunswick containment atmospheres are innerted with nitrogen to limit initial post LOCA oxygen content.
a.
What is the principal source of oxygen in containment following a LOCA?
b.
What is the principal source of hydrogen in containment following a LOCA?
c.
What are the maximum permissible concentrations of hydrogen and oxygen (in volume percent) in containment following a LOCA?
GUESTION 5.12 (1.50)
Using the enclosed Mollier diagram, list the following property values for steem with an enthclpy of 1390 BTU /lbm and an entropy of 1.568 BTU /lbm-desR a.
pressure b.
temperature c.
superheat DUESTION 5.13 (1.00)
The Standby Liquid Control System (SLC) injects a sodium pentaborate Solution of 13.4 weight percent into the reactor coolant at a rate of from 6 to 25 ppm per minute.
Why (i.e.,
what is the basis) is there a minimum rate (6 ppm / min) at which the solution must be injected?
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5.
1HEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PACE 7
QUESTION 5.14 (2.50)
The attached figure 879 represents a transient that could occur at a BWR.
Briefly explain the cause(s) of the following recorder indications (a thru e).
Givent (1) EHC pressure regulator fails to maximum at time t=1.0 min.
(2) No operator actions occur.
(3) Recorder speed = 1 division = 1 minute Note: There may be more than 1 cause for each answer.
a.
Level INCREASE (point A) b.
Reactor power DECREASE (point B) c.
Reactor power DECREASE (point C) d.
Steam flow DECREASE (point D) e.
Pressure FLUCTUATION (region E at time 3 to 6 minutes)
GUESTION 5.15 (1.50) a.
E:: plain how the thermal time constant affects the response of the reactor during normal and transient operations.
In particular, discuss any coefficient affects.
(1.0) b.
For the 8 X 8 fuel in your core, how lon3 is the thermal time constant?
(0.5)
QUESTION 5.16 (3.00)
Attached figures 890 A & B represent a transient that could occur at a BWR.
Briefly explain the cause(s) of the following recorder indications (a thru c).
Given:
1.
Turbine trips from rated conditions at time zero.
2.
No operator actions occur.
3.
Recorder speed = 1 division = 30 seconds.
Note:
There mey be more than 1 cause for each answer.
a.
Why does core flow decrease (Pt 1) and why doesn't it decrease to zero (Pt 2)?
b.
Why does reactor pressure increase (Pt 3) and remain high (Pt 4)?
c.
Why does reactor level decrease initially (Pt 5) and what is causing the peaks in level later (Pt 6)?
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5.
THEORY OF NUCLEAR POWER PLAN 1 OPERATION, FLUIDS, AND PAGE 8
OUESTION 5.17 (1.50)
Attached figure 888 shows a basic closed loop fluid system with its head vs flow plot.
The two pumps are identical, single speed, radial, centrifugal pumps.
Initially, assume pump 1 is operating to supply flow to component 1 as shown.
a.
What is point X on the system head vs flow plot?
b.
Which punip curve, A or B, most accurately shows both pumps operating to supply system flow?
c.
Which way, to the left or to the right, would the system curve shift if component 2 was valved into the system, in addition to component 1?
DUESTION 5 18 (2.50) a.
Define the term BETA with regard to delayed neutrons?
(1.0) b.
When conipating the individual BETA's from thermal fission of U-235, Pu-239 and fast fission of U-238, which BETA is largest?
(0.5) c.
From BOL to EOL, does the core average beta INCREASE, DECREASE or REMAIN THE SAME?
EXPLAIN your answer.
(1.0)
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6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENlATION PAGE 9
QUESTION 6.01 (1.00)
Five rod block monitor (RBM) rod block functions are listed below.
Which one is provided to stop the erroneous withdrawl of the most reactive control rod so that local fuel damage does not result.
a.
either RBM downscele alarm b.
recirculation flow converter comparator alarm c.
any recirculation flow converter upscale or inoperative alarm d.
either RBM inoperative alarm o.
either RBM upscale alarm QUESTION 6.02 (1.00)
The main turbine is at 1800 rpm in preparation for synchronizing the main 92nerator to the grid (ier the 230 kv generator breakers are still open).
Wh t will happen if the 'all valves closed" pushbutton is depressed?
a.
Nothing will happen since the synchronous speed select signal is sealed in.
t b.
The turbine contro1' valves and main stop valves will close, but the intercept valves will remain open.
i c.
All of the control valves (TCVs and ICVs) and main stop valves (MSV) j)-eft +-FtsT will c l o s e.
d.
The control valves (TCVs and IVs) will close, but the main stop valves will remain open.
QUESTION 6.03 (1.00)
Select which one of the followin3 an operator does to increase VARS.
a.
Increase generator speed b.
Increase capacity factor c.
Increase generator voltage d.
Increase generator stator cooling (r**** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
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6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMLN1A11UN PAGE 10 OUESlION 6.04 (1.00)
Which of the following sequences of components correctly reflects the normal HPCI condensate flow path on unit 2?
a.
CST, booster pump, main pump,
'A' FW line upstream of the FW flow detector.
b.
CST, booster pump, main pump,
'A' FW line downstream of the FW flow detector.
c.
CST, main pump, booster pump,
'A' FW line upstream of the FW flow detector.
d.
CST, main pump, booster pump,
'A' FW line downstream of the FW flow detector.
QUESTION 6.05
(.50)
Which AGAF value (P-1 printout) is more conservative?
z.
0.9?
b.
1.01 QUESTION 6.06 (1.50)
With regard to the ADS system:
a.
What are the two methods (using normal contrni room controls) that the operator may use to shut the ADS valves once they have auto l
initiated and are open?
(1.0) b.
The ADS valves will remain open until R>: pressure is about _____ psis above containment pressure.
(0.5) i i
1
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6.
PLANT SYSTEMS DESIGN, CON 1ROL, AND INSTRUMENTATION PAGE 11 QUESTION 6.07 (2.00)
Complete the following statement by filling in the blanks.
If the turbine is operated at rated speed with no load or at loads less than 100 MWE, the LP turbine's diaphragm temperature can decrease to less than 120 desF which may cause severe turbine damage.
______ unit 1 only, both Brunswick units, unit'2 only),the' For
(
turbine will trip if LP rotor 12th stage diaphragm temperature is less than or equal to 120 desF for ______ minutes, turbine speed is greater than ______ rpm, and turbine load is less than ______%.
QUESTION 6.08 (1.50)
Answer the following with regard to the instrument and service air system; a.
Compressors A,B and C have e three-position switch which selects "high-inter-low' modes of operation.
What parameter does "high" and
' lou' refer to?
Which compressor (s) is (are) affected?
Which mode (s) is (are) normally selected?
(1.0) b.
When will the emergency air compressors automatically start?
Include setpoint(s).
(0.5)
QUESTION 6.09 (2.00)
With regard to the vital service water header off the NSW system a.
List all components that receive their coolin3 water supply from the vital service water header?
(1.5) b.
How are corrosion and fouling minimized in the vital service water header?
(0.5)
DUESTION 6.10 (1.00)
The white ' refuel mode one rod permissive light' is energized when what conditions are satisfied?
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6.
PLAN 1 SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 12 GUESTION 6.11 (2.00)
List the four requirements and associated setpoints for the source range conitor system to generate a rod block.
QUESTION 6.12 (1.00)
List the emergency bus power supplies to each of the Unit 1 RHR pumps.
QUESTION 6.13 (3.75)
Assume the feedwater level control system is being operated in 3-element control using reactor level detector channel
'A'.
Reactor power is at 85%,
steady state.
For each of the instrument or control signal failures listed below, state how reactor level will initially respond (increase, decrease or remain constant) and briefly explain why in terms of what is happening in the level control system immediate11y following the failure.
Note: A block diagram of the feedwater level control system is attached.
a.
Channel
'A' reactor level detector signal fails low.
b.
Loss of control signal to
'B' reactor feed pump speed controller.
c.
- B' feedwater line flow signal fails high.
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6.
PLANT SYSTEMS DESIGN, CON 1ROL, AND INSTRUMENTATION PAGE 13 GUESTION 6.14 (1.50)
Given the following conditions and the attached EHC logic diagram:
Reactor power 30%' Senerator output = 30%.
=
Reactor pressure = 934 psis.
Throttle pressure = 929 psis.
Pressure regulator setpoint = 920 psis.
Recite pumps in manual at 28% speed.
Load set = 30 Load limit = 100%, Maximum combined flow = 105%.
Turbine speed = 1800 rpm.
The generator output circuit breakers (OCBs) are suddenly opened manually.
c.
Has a load rejection occurred?
Explain.
(1.0) 6.
Which of the subsystems (eg., pressure control unit, bypass control unit, etc.) of the EHC pressure control and logic system is the first to produce an error si grial when the OCBs open?
No explanation required.
(0.5)
QUESTION 6 15 (2.50)
How does the core spray system piping break detection system confirm the integrity of the core spray piping?
Include in your answer what piping integrity is verified and whye sensing points, nor nial readings
(+ or - dp) and how it changes when a break is detected.
QUESTION 6.16 (2.00)
How does the rod sequence control system (RSCS) determine control rod position and when is each method of position detection used?
QUESTION 6.17 (2.00)
The reactor is operating at 90% power when the load reject circuitry is actuated.
a.
What actuates the load reject circuit?
Set points required.
(1.0) 6.
What prevents the turbine from overspeeding?
(0.5) c.
How is the load reject trip reset?
(0.5)
(*****
END OF CATEGORY 06 *rrrr)
7.
PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 14
~~~~EEDi5L5GiCEE C5 sir 5L QUESTION 7.01 (1.00)
The plant is operatins normally at approximately 85% power.
Which of the following statements describes the proper response to a " Low System Frequency" per AOP-22.0?
a.
When the ' GEN BUS UNDER FRED RELAY" annunciates, promptly reduce generator load to 100 MWe and separate the unit from the grid.
b.
Decrease unit output to the maximum consistent with plant conditions; if frequency increases to 61.5 H: commence a rapid shutdown per GP-05.
c.
Increase recirc pump speed as necessary to maintain unit load as grid frequency decreases.
d.
Increase unit output to maximum consistent with plant conditions; if frequency decreases to 58.4 H: 4-e49ee-Joed ;to -100. MWer and( separ at e the unit from the grid.
QUESTION 7.02 (1.00)
MATCH the following Contamination Levels with the appropriate MINIMAL requirements for wearing PC Gloves
- o.
Alpha Contamiantion 1.
20 dpm / 100 cmucm b.
Beta-Gamma Contamination 2.
200 dpm / 100 cm*cm 3.
1,000 dpm / 100 cm*cm 4.
5,000 dpm / 100 cmucm OUESTION 7.03 (3.00) a.
A fire of unknown crisin breaks out in the control room resulting in heavy smoke.
You make the decision to evacuate the control room.
What are 8 actions you expect the Unit Control Operator to take prior to leaving the control room (per AOP-32)?
(2.0) b.
If you could take no actions prior to leaving the control room, what actions should you take outside the control room to shutdown the reactor and where would you take them (per ADP-32)?
(1.0)
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7.
PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 15 RAU5UEUG5CAE~CUN5EUE~~~~~~~~~~~~~~~~~~~~~~~~
~~~
GUESTION 7.04 (2.00)
What is the Proper order of valve operation when reali3ning the CS pump suctions to the CST vs. the suppression pool suctions?
Indicate locked valve positions as applicable.
QUESTION 7.05 (3.50) e.
On receipt of a valid reactor buildin3 ventilation high radiation alarm (due to spent fuel damage) during refueling operations:
1.
What 2 automatic actions occur per ADP-7.0?
(1.0) 2.
Implementation of AOP-7.0 requires declaration of what emergency action level?
(0.5) b.
What are four abnormal conditions that would be reason to terminate fuel handling operations per fuel handling procedure FH-11?
(2.0)
GUESTION 7.06 (2.00)
AOP-15.0, " Alternate Shutdown Cooling", identifies five conditions which aust be met before that method of shutdown cooling may be used.
What are fcur of those five conditions?
QUESTION 7.07 (2.50)
AOP-32.0, " Plant Shutdown from Outside the Control Room'r states th'at a oinimum of five persons would be required to perform such a shutdown on one unit.
Where would those ' Ave persons be stationed for a remote shutdown on unit 1?
(all stations need not be permanent).
l QUESTION 7.08 (1.00) l There are two End Path Manuals at Brunswick.
State the respective operational occurrences which will put an operator into each EPM.
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7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 16
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GUESTION 7.09 (3.00)
Concerning use of control rods during plant start up; a.
After withdrawing a control rod to position 40, an attempt is made to withdraw it again.
What are the two indications that procedure OP-07 instructs the operator to watch for that would indicate the rod is uncoupled.
b.
If a double notch is detected while attempting notch insertion, what action should the operator take?
Restrict your response to physical actions at the control panel, c.
Why is control rod withdrawl to be avoided above 60% core flow during a plant startup?
GUESTION 7.10 (2.50)
Rescrding operation of the SCGT (standby gas treatment) system, in accordance with procedure OP-10 a.
Upon completion of operation of a SBGT train, why are you cautioned not to place its control switch in the ' standby' position?
(1.0) b.
What automatic action should occur in the reactor building ventilation system when SBGT starts automatically?
(0.5) c.
If the temperature of the operating SBGT train reaches 210 desF while in emergency operation (>180 desF inlet temperature), what does this indicate and, in general terms, what actions, regarding correction of the overtemperature problem, should be considered?
(1.0)
GUESTION 7.11 (2.00)
Briefly e:-: plain the reason (s) for each of the following cautions:
a.
During reactor reject operations with vacuum in the condenser, both reject valves, to radwaste and to condenser, must not be opened simultaneously.
b.
An idle recire loop shall not be restarted unless operating loop flow is less than 50%.
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7.
PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 17
~~~~ d65UL65iEEL E5ETE5L R
OUESTION 7.12 (2.50) c.
What are three parameters or system components you would check to determine the cause of a drifting IN control rod?
(1.5) b.
If the control rod in part "a' was fully inserted and electrically disarmed *
(1.0) 1.
Would the control rod be considered inoperable?
2.
Why would the HCU for that rod not be completely valved out?
QUESTION 7.13 (2.00)
Level Detectors NO36/NO37 are not included in the E0P-01/UG caution (CAUTION 96) concerning high temperatures near the reference les vertical runs.
EXPLAIN WHY these instruments are EXCEPTED from this caution and WHEN, if ever, these instruments would develop excessive insecuracies.
(*****
END OF CATEGORY 07 *****)
8.
ADMINISTRATIVE PROCEDURES, CONDITIONSr AND LIMITATIONS PAGE is GUESTION 8.01 (1.00)
Until the EOF is activated, the Site Emergency Coordinator shall NOT dele 3 ate the responsibility for*
o.
directing the combined activities of plant personnel in the CR, TSC and OSC.
b.
requesting outside emergency assistance.
c.
assessing the emergency condition for possible upgrade in classification.
d.
deciding what protective action recommendations will be made to off-site authorities.
e.
augmenting the on-shift personnelt if required.
QUESTION 8.02 (1.00)
Which one of the following events does NOT require a 1-hour Red Phone Report to the NRC?
c.
Any event that results in manual or automatic actuation of an Engineered Safety Feature.
b.
The initiation of a nuclear plant shutdown required by the plant's technical specification.
c.
Any event that should have resulted in ECCS discharge into the reactor coolant system as a result of a valid signal.
d.
Any event that may have caused e>:posure to the whole body of a'i n
individual to 25 rems or more of raoiation.
e.
Any event that results in failure of an SRV to close.
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8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMIlATIONS PAGE 19 I
GUESTION 0.03 (1.00)
Unit 2 is operating at 75% rated thernial power.
Channe) Functione]
Tests are performed on all of the MSL Radiction Monitorins System 4
channels.
Channels A and D test UNSAT; Channels B and C test SAT.
Maintenance has no estimate of repair time and will not be able to commence troubleshooting and repair for at least 16 - 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.
Which of the followins actions most correctly detail the allowances and/or limitations imposed by the Technical Specifications in this instance?
NOTE: APPLICABLE TS's ARE ENCLOSED FOR REFERENCE a.
Be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> b.
Be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
c.
Place MSL Rad Mon Channel
'A' in the tripped condition within one hour -AND-be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
d.
Place MSL Rad Mon Channel
'A' in the tripped condition within one hour -AND-be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in CO:D SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
GUESTION 8.04 (1.00)
You, as Shift Foreman, must prepare a NEW OWP to remove a system from service for maintenance.
WHO, of the followins, is required i
to approve this OWP for TS interpretation, per 01-10, ' Operations Work Procedures.'
o.
A-second licensed operator b.
PNSC c.
Engineer - Operations d.
The SOS e.
A second licensed senior operator I,
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
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B.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMIlATIONS PAGE 20 QUESTION 8.05 (1.00)
Unit 2 is operating at 92% Rated Thermal Power, with one outstanding LCO:
The HPCI Aux Oil Pump motor has been removed for repairs (It has been out for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />)
Three hours into your shift, a routine surveillance by the instrument shop determines that Channel C-2 of the RCIC Reactor Vessel Water Level - HIGH has a setpoint of 213 inches.
It con not be readjusted by the instrument technician.
Which of the following actions most correctly detail the allowances end/or limitations imposed by the Technical Specifications in this instance.
a.
No new limitations or TS Operational Condition restrictions are initiated.
b.
Place the INOPERABLE channel in the tripped condition within one hout - OR - declate RCIC INOPERADLE.
Power operations may continue.
c.
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 SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
d.
Be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
QUESTION 8.06 (1.00)
The determination is made that an invalid multiple input annunciator condition is being caused by one failed sensor input.
Which of the following steps should be taken per 01-05, ' Abnormal Annunciator Status'?
a.
Remove annuncictor cctd and identify annunciator window with a ' red dot".
b.
Remove annunciator card - Defeat invalid sensor input - Replace annunciator card.
c.
Remove annunciator card - Defeat invalid sensor input - Replace annunciator window with a ' red dot".
d.
Remove existing annunciator card and replace with 'special slow window flash' annunciator card.
Identify window with a ' yellow dot".
e.
Defeat invalid sensor - remove annunciator card.
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8.
ADMINISTRATIVE PROCEDURES, COND1110NSr AND LIMIlAlIONS PAGE 21 00ESTION 8.07 (1.00)
Unit 2 is in COLD SHUTDOWN, the containment is deinerted, and a reactor startup is proceeding with no outstanding deficiencies.
The containment atmosphere dilution (CAD) system becomes inoperable.
It is anticipated that repairs will be complete within two weeks.
With regard to the reactor startup, which of the following actions most correctly details the allowances and/or limitations imposed by the technical specifications?
Note that applicable TSs are enclosed for reference.
a.
Startup activities may continue; Operational Condition 1 may be entered with no restriction on power, but the CAD system must be returned to an operable status within 32 days of exceeding 15% power, b.
Startup activities may continvei Operational condition 1 may be entered with no restriction on power, but the CAD system must be returned to an operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of exceeding 15% power.
c.
Startup activities may continuci Operational Condition 1 may be entered but thermal power is limited to 15%.
d.
Startup activities may continue; Operational Condition 2 may be entered but not exceeded.
G.
Startup activities may continvei however, Operational Condition 4 must be maintained.
(Entry into Operational Condition 5 is acceptable.)
QUESTION 8.08
(.50)
Given:
A safety related component has been declared inoperable in accordance with the Technical Specifications.
It is proposed to install a jumper in the components control circuit.
You have reviewed 10CFR50.59 as directed by the ' Jumper and Wire Removal Approval
- procedure AI-59 and have determined that an unreviewed safety question DOES exist.
TRUE OR FALSE Since this constitutes ' departure' from an established procedure, you may allow the jumper to be installed only if the safety of persons, the reactor, or other equipment is in jeopardy.
QUESTION 8.09 (1.00)
Each high radiation area in which the intensity of radiation is less than or equel to (a) ______ shall be barricaded such that access may be gained only (b)
Fill the blank (a) with the correct number and (b) with a Phrate indierting
'how'.
(rx**i EM EGOPY 00 CON 11NUED ON NT X1 PAGE Fr*rx)
8.
ADMINISTRAllVE PROCEDURES, CONDITIONSr AND LIM 11A110NS PACE 22 GUESTION 8.10
(.50)
Fill in the blank (number and units) as stated in the Brunsuick EERC.
" Extended Annual Dose Limit.
The accumulated annual whole body dose shall not be allowed to exceed ______ without prior approval by the Vice President-Brunswick Nuclear Project.'
OUESTION 8.11 (2.50)
Concerning 01-04, 'LCO evaluation and followup *:
a.
When any system with a technical specification LCO is made or found inoperable, the Shift Foreman on duty shall complete what form?
(0.5) b.
Prior to granting permission to perform routine surveillance testing allowed by the technical specifications, what are four documents or records the shift foreman should check to ensure the testing is not going to be done on a system required to be operable due to the inopertbility or in-test stctus of other ESF systems?
(1.0) c.
If the Shift Foreman decides to permit the surveillance testing (part b above), but must increase the surveillance testing frequency of a related system to comply with the technical specifications, how does he alert the other shifts of the chan3e in surveillance requirements?
(1.0)
GUESTION 8.12 (1.50)
The " Conduct of Operations' section of the plant administrative procedure states that no recorder shall be removed from service without the Shift Forman's permission.
What three things must be marked on a recorder chart that is removed from service?
(***** CATEGORY 00 CONTINUED ON NEXT PAGE
- )
8.
ADMINISTRATIVE PROCEDURES, CONDIl10NS, AND LIMI1AlIONS PAGE 23 DUES 110N 8.13 (2.00)
Regarding the Technical Specifications for the recirculation system a.
Why must the reactor be shutdown if a jet pump is determined to be inoperable?
(Two reasons relative to a DDA LOCA) b.
What is the maximum temperature differential allowed between the reactor coolant in the idle and operating recire loops to permit starting the idle loop?
Why is this limit imposed?
OUESTION 8.14 (2.00)
For each of the following conditions, state whether you would consider the system operable per the TS and briefly state why you determined the system to be so.
a.
An emergency D/G selected to CR manual for testing.
b.
Control of RHR pumps B and D shifted to the Remote Shutdown Panel.
QUESTION 8.15 (3.00) o.
State all situations in which a " departure" from established procedures is justified and list all persons who may authorize it(1 5) b.
State in detail all persons by title who may authorize ' deviations
- from established procedures.
Include any s.pecial qualifications required of those persons.
(.75) c.
What is the difference between a " departure' and a ' deviation' from an established procedure with regard to procedural intent?
(.75)
(*****
CATEGORY 08 CONTINUED ON NEXT PAGE
- )
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMI1ATIONS PAGE 24 QUESTION 8.16 (3.00)
Regarding the procedure for clearances (admin procedure, sect 11);
a.
What are the three major groups of clearances?
(1.5) b.
What is a multiple clearance?
(0.5) c.
Describe the qualifications which are required for a person to hold a clearance.
How will these qualifications normally be verified by the operator on shift?
(1.0)
GUESTION 8.17 (1.00)
The Control Operator is unable to perform one of the pts on the DSR.
How nust he document the non-performance of the PT and how does he reschedule it prior to routing the DSR to the Shift Foreman for review?
DUESTION 8.18
(.50)
Br iefly e:-:pla in the reason for the following caution from 01-13.
- When performing valve checks or line-ups on systems that are normally operated at high temperatures, valves should NOT be positioned on their backseat."
QUESTION 8.19 (1.00)
Core SHUTDOWN MARGIN must be determined by measurement within
.24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to, or during, the first.STARTUP after completing CORE ALTERATIONS.
a.
SDM must show the core to be Subcritical by, at least,
'R 4 ____ % delta k / E'.
b.
Define
'R'.
QUESTION 8.20 (1.00)
I 10CFR20 defines an ' airborne radioactivity area' in two ways.
State either i
one of these definitions.
(***** END OF CATEGORY 08
- )
(*****rrrrrrrr END OF EXAMINATION ***************)
l l
f o ina v8 5/t C/Cle eff1Clency = (Net t:ork cut)/(Energy in) l
,,.n g s = v,:
- 1/2 at E = mC~
<E = 1/2 mv a=(/f - V )/*
A = aN A = A e "I
~
3 g
PE = mgn
- e/t x = an2/t1/2 = 0.693/ti/2 vf = V,
- 4 t 2
1/2'
' b *1 b 3 W=v P
- 0 A=
[(t
) + (t )]
4 ti = 931 am m = V,yAo
-h Q=hh I
- I,e Q = mCoat Q = UA4 T I = I e'"*
g I = I,10**/U L Pwe = w ah f
TVL = 1.3/u sur(t)
P = P 10 HVL = -0.693/v P = P,e*
SUR = 26.06/T SCR = S/(1 - K,ff)
CA, = S/(1 - K,ff,)
CR (1 - K,ffj) = CR (I ~
- ff2)
SUR = 26e/t= + (s - o)T j
2 e
T = (t*/s) + [(s - o)fio]
M = 1/(1 - K,ff) = CR /CR, j
T = s/(o - s)
M = (1 - K,ff,)/(1 - K,ff j)
T = (s - o)/(To)
SOM = (
- K,ff)/K,ff t= 10 secones a = (K,ff-1)/K,ff = AK,ff/K,ff I = 0.1 seconds o = [(1=/(T K,ff)] + [s,ff (1 + IT)]
/
1t*Id Id 2,2 2 l
P = ( teV)/.[3 x 1010)
Id gd jj 22 2
l I = eN R/hr = (0.5 CE)/d (meters) 2 R/hr = 6 CE/d gf,,g)
Water Parameters Miscellaneous Conversions 1 gal. = 1 345 lem.
1 curie = 3.7 x 1010aps I ga;. = 7.78 liters 1 kg = 2.21 lem 3
I fte = 7.48 gal.
1 np = 2.54 x 10 Stu/nr Density = 62.4 Itup/ft3 1 r= = 3.41 x 106 5tu/hr Density = 1 gm/cm3 lin = 2.54 cm Heat of vaporization = 970 Stu/lom
'F = 9/5'C + 32 Heat of fusion = 144 Btu /lbm
'C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. Hg.
1 BTU = 778 ft-lbf i
1 ft. H O = 0.4335 Itf/in, 2
e = 2.718
V3.'me, ft'/lb EMhelpy,8tv$b (ntropy. OttAt a F Water Evep Steam Water Evep Steim Water gvep St:em y'E
'[
t 4,,
A, at seg s,
A
't
'se
's 32 0.08859 0.01602 3305 3305
-0.02 1075.5 1075.5 0.0000 2.1873 2.1873 32 35 0.09991 0.01602 2948 2948 3 00 1073.8 1076 8 0.0061 2.1706 2.1767 35 40 0.12163 0 01602 2446 2446 8 03 1071.0 1079 0 0.0162 2.1432 21594 40 45 0.14744 0 01602 2037.7 2037.8 13.04 1068.1 1081.2 0 0262 2.1164 2.1426 45 i
50 0.17796 0 01602 1704.8 1704.5 18.05 1065 3 1083.4 0 0161 2.0901 2.1262 50 60 0.2561 0.01603 1207.6 1207.6 28.06 1059.7 1067.7 0.0555 2.0391 2.0946 60 70 0.3629 0.01605 868.3 868 4 38.05 10540 1092 1 0.0745 1.9900 2 0645 70 80 0.5068 0.01607 633.3 633.3 48.04 1048.4 1096.4 0.0932 1.9426 2 0359 80 to 0.6981 0.01610 468.1 468.1 58.02 1042.7 1100.8 0.1115 1.8970 2.0086 90 100 0.9492 0.01613 350.4 350.4 68 00 1037.1 1105.1 0.1295 1.8530 1.9825 100 110 1.2750 0.01617 265.4 265.4 77.98 1031.4 1109.3 0.1472 1.8105 1.9577 110 130 1.6927 0.01620 203.25 203.26 87.97 1025.6 1113.6 0.1646 1.7693 1.9339 120 130 2.2230 0.01625 157.32 157.33 97.M 1019.8 1117.8 0.1817 1.7295 1.9112 130 140 2.8892 0.01629 122.98 123.00 107.95 1014.0 1122.0 0.1985 1.6910 1.8895 140 150 3.718 0.01634 97.05 97.07 117.95 1006.2 1126.1 0.2150 1.6536 1.8686 150 160 4.741 0.01640 77.27 77.29 127.M 1002.2 1130.2 0.2313 1.6174 1.4487 160 170 5.993 0.01645 62.04 62.06 137.97 996.2 1134.2 0.2473 1.5822 1.8295 170-les 7.511 0.01651 50.21 50.22 144.00 990.2 1138.2 0.2631 1.5480 1.8111 ISO 190 9.340 0.01657 40.94 40.M 154.04 984.1 1142.1 0.2787 1.5148 1.7934 ISO 200 11.526 0.01664 33.62 33.64 168.09 977.9 1144.0 0.2940 1.4824 1.7764 200 210 14.183 0.01671 27.00 27.82 178.15 971.6 1149.7 0.3091 1.4509 1.7600 210 212 14.696 0.01672 26.78 26.80 100.17 970J 1150.5 0.3121 1.4447 1.7568 212 220 17.186 0.01678 23.13 23.15 188.23 965.2 1153.4 0.3241 1.4201 1.7442 220 230 20.779 0.01685 19.364 19.381 198.33 958.7 1157.1 0.3388 1.3902 1.7290 230 240 24.968 0.0!693 16.304 16.321 208.45 952.1 1160.6 0.3533 1.3609 1.7142 240 ISO 29.825 0.01701 13.802 13.819 218.59 945.4 1164.0 0.3677 1.3323 1.7000 250 260 35.427 0.01709 11.745 11.762 228.76 938 6 1167.4 0.3819 1.3043 1.6842 See 270 41.856 0.01718 10.042 10.060 238.95 931.7 1170.6 0.3960 1.2769 1.6729 270 ISO 49.200 0.01726 S.627 8.644 249.17 924.6 1173.8 0.4098 1.2501 1.4599 ISO 290 57.550 0.01736 7.443 7.460 259.4 917.4 1176.8 0.4236 1.2238 1.6473 290 300 67.005 0.01745 6.448 6.466 269.7 910.0 1179.7 0.4372 1.1979 1.6351 300 310 77.67 0.01755 5.609 5.626 280.0 902.5 1182.5 0.4506 1.1726 1.6232 310 320 89.64 0.01766 4.896 4.914 290.4 894.8 1185.2 0.4640 1.1477 1.6116 320 340 117.99 0.01787 3.770 3.788 311.3 878.8 1190.1 0.4902 1.0990 1.5892 340 360 153.01 0.01811 2.939 2.957 332.3 862.1 1194.4 0.5161 1.0517 1.5678 360 340 195.73 0.01836 2.317 2.335 353.6 844.5 1198.0 0.5416 1.0057 '1.5473 300 400 247.26 0.01864 1.8444 1.8630 375.1 825.9 1201.0 0.5667 0.9607 1.5274 400 420 305.78 0.01894 1.4408 1.4997 396.9 306.2 1203.1 0.5915 0.9145 1.5000 420 440 381.54 0.01926 1.1976 1.2169 419.0 785.4 1204.4 0.6161 0.8729 1.4890 440 440 466.9 0.0196 0.9746 0.9942 441.5 763.2 1204.8 0.6405 0.8299 1.4704 460 480 566.2 0.0200 0.7972 0.8172 464.5 739.6 1204.1 0.6648 0.7871 1.4516 400 500 680.9
- 0.0204 0.4545 0.6749 447.9 714.3 1202.2 0.6890 0.7443 1.4333 500 520 812.5 0.0209 0.5306 0.55M 512.0 687.0 1199.0 0.7133 0.7013 1.4146 520 540 962.8 0.0215 0.4437 0.4651 536.8 657.5 1194.3 0.7378 0.6577 1.3954 540 560 1133.4 0.0221 0.3651 0.3871 562.4 625.3 1187.7 0.7625 0.6132 1.3757 560 580 1326.2 0.0228 0.2994 0.3222 589.1 589.9 1179.0 0.7876 0.5673 1.3550 580 600 1543.2 0.0236 0.2438 0.2675 617.1 550.6 1167.7 0.M 34 0.51M 1.3330 8i00 620 1786.9 0.0247 0.1962 0.2200 646.9 506.3 1153.2 0.84011 0.4689 1.3092 820 640 2059 9 0.0260 0.1543 0.1802 679.1 454.6 1133.7 0.8666 0.4134 1.2421 See 660 2365.7 g 0 0277 0.1166 0.1443 714.9 392.1 1107.0 0.8995 0.3502 1.2498 660 640 2708.6 0.0304 0.0006 0.1112 758.5 310.1 1068.5 0.9365 0.2720 1.2086 400 700 3094.3 0.0366 0.0386 0.0752 822.4 172.7 995.2 0.9901 0.1490 1.1390 700 705.5 3208 2 0.0508 0
0.0500 906.0 0
906.0 1.0612 0
1.0612 705.5 TABLE A.2 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (TEMPERATURE)
A.3
Volume. it'/n Enthalpy. Ste/le Ettripy. Stw/2 a f Energy. Stv/la Pwas.
Temp teter Erep Steam t;tir
- Eve, Steam Weter Even tesem tener Steam P*#.
y6 pole F
s
'g
's f
t A,
s, s,,
s, e,
A A
i v
0.0486 32.018 0.01602 3302.4 3302.4 0.00 1075.5 1075 5 0
2.1872 2.1872 0
1021.3 0.0s46 i
0.10 35.023 0.01602 2945.5 2945 5 3 03 10735 10765 0 0061 2 1705 2.1766 333 1022.3 0.10 O.15 45.453 0.01602 2004.7 20047 13 50 1067.9 1081 4 0 0271 2.1140 2 1411 13.50 1025.7 0.35 0.20 51160 0 01603 1526 3 1526 3 21.22 1063 5 1084 7 0 0422 2 0778 2.1160 21.22 10283 0.20 0.30 64 484 0 01604 1039.7 1039.7 32.54 1C57.1 1089 7 0 0641 2 0165 2.0809 32.54 1032 0 0.30 0.40 72.869 0.01606 792.0 792.1 40.92 1052.4 1093.3 0.0799 1.9762 2.0562 43.92 1034.7 0.40 0.5 79.586 0 01607 641.5 641.5 47.62 1048 6 1096.3 0 0925 1.9446 2.0370 4742 1036.9 0.5 0.6 85.218 0 01609 540.0 S40.1 53 25 1045 5 1098 7 0.1028 1.9186 2.0215 53.24 1038.7 0'6 0.7 90 09 0.01610 466.93 466 94 58 10 1042 7 11008 0.3 1 8966 2.0083 58.10 1040.3 0.7 GA 94.38 0.01611 411.67 411.69 62.39 1040.3 1102.6 0.1117 1.8775 1.9970 62J9 1041.7 0.8 O.9 98.24 0.01612 368.41 368.43 66 24 1038.1 1104.3 0.1264 1A406 1.9870 4624 1042.9 0.9 1.0 101.74 0.01614 333 59 333 60 69.73 1036.1 11058 0.1326 1A455 1.9781 69.73 1044.1 1.0 2.0 126.07 0.01423 173.74 173.76 94.03 1022.1 1116.2 0.1750 1.7450 1.9200 94A3 1051A 2A 3.0 141 47 0.01630 114.71 118.73 109.42 1013.2 1122.6 0.2009 14854 1.8864 10941 1054.7 8.0 4.0 152.96 0.01636 90 63 90 64 120.92 1006.4 1127.3 0.2199 14428 1.8626 120.90 1060.2 4.0 8.0 162.24 0.01641 73.515 73.53 130.20 1000.9 1131.1 0.2349 1.6094 1A443 130.18 1063.1 SA 6.0 170.05 0.01645 41.M7 61 98 138 03 996.2 1134.2 0.2474 1.5820 1A294 1331 10H.4' 4.0 7A 176 84 0.01649 53 634 53.65 144 83 992.1 11M 9 0.2581 1.5587 12168 144Al 1067.4 7A S.0 182.84 0.01653 47.328 47.35 150.87 988.5 1139.3 0 2676 1.5384 12060 15034 1089.2 S0 9.0 ISO 27 0 01654 42.385 42.40 156.30 985.1 1841.4 0.2760 1.5234 1.7964 15428 1070.8 9.0 i
10 193.21 0.01659 36.404 38 42 161.26 982.1 11433 0.2836 1.5043 1.7879 16123 1072J 10 14.696 212.00 0.01672 26.782 26 80 180.17 970.3 1150.5 0.3121 1.4447 1.7968 100.12 1077.6 14.886 i
15 213.03 0.01673 26.274 26.29 181.21 969.7 1150.9 0.3137 1.4415 1.7552 181.16 1977.9 15 20 227.96 0.01683 20.070 20 087 1M.27 960.1 1156.3 0.3358 1.3962 1.7320 19621 IntA 30 30 250.34 0 01701 13.7266 13 744 218.9 945.2 1164.1 0 MS2 1.3313 1.8995 218A 1873 30 40 267.25 0.01715 10 4794 10 497 236.1 933.6 11692 OJ921 1.2844 1.6765 236 4 1082.1 40 l
80 281.02 0.01727 8.4967 8.514 250.2 923.9 1174.1 0.4112 IJ474 16586 290.1 1096.3 Se f
80 292.71 0.01738 7.1542 7.174 262.2 915.4 1177.6 0.4273 1.2147 1A440 262A 108SA 80 l
70 302.93 0.01748 6.1875 4205 272.7 907A 1180.6 0 4411 1.1905 1A316 272.5 1100.2 70 80 3!2.04 0.01757 5 4536
$ 471 232.1 900.9 1153.1 0.4534 1.1675 14208 281.9 1102.1 80 90 320.29 0.01766 4.8777 4.895 290.7 894 6 1185.3 0.4643 1.1470 1.6113 290.4 1103.7 90 100 327.82 0.01774 4.4133 4.431 298.5 888.6 1187.2 0.4743 1.1284 1.6027 298.2 1105.2 100 120 341 27 0.01789 3 7097 3 728 312 6 877A 11934 0 4919 1A960 1.5879 312.2 1107.5 120 140 353 04 0 01803 3 2010 3 219 325.0 868.0 1193 0 0.5071 1.0681 1.5752 324.5 1109.6 140 160 363 55 00;815 2.8155 2.834 336.1 859 0 1195.1 0.5206 1.0435 1.M41 335.5 1111.2 ISO 180 373 08 0.0;S27 2.5129 2.531 346.2 850.7 1196.9 05328 1.0715 1.5543 345A 1112.5 ISO 200 37,1 80 0 01839 2.2689 2.287 355.5 842.8 1198.3 0 5438 1.0016 1.5454 3543 1113.7 380 250 400,97 0 01865 1.8245 1.8432 376.1 825 0 1201.1 0 5679 0 9585 1.5264 375J 1115.8 280 300 417.3F 0 01889 1.5235 1.5427 394 0 808.9 1202.9 0.5882 0.9223 1.5105 392A 1117.2 300 350 421.73 0 01913 1.3064 1.3255 409.8 794 2 12040 0 6055 0 8909 1.4968 408 6 1118.1 350 400 44440 0 0193 1.14162 1.1610 424.2 780 4 12046 0 6217 0.8630 1.4547 422.7 1112 7 400 450 456.28 0 0195 1.01224 1.0318 437.3 767.5 1204 8 0 6360 0A378 1.4738 435.7 1118.9 480 500 46701 0 0195 0 90787 0 9276 449.5 755.1 1204.7 0.6490 0 8148 1.4639 447.7 11183 900 553 47494 0 0199 0 82183 0 8418 460.9 743.3 1204 3 0.6611 0.7936 1.4547 458.9 1118 6 550 603 48523 3 0201 0.74962 0.7698 471.7 732.0 1203 7 0.6723 0.7738 1.4461 409 5 1116.2 500 703
.503 08 0.0205 0.63505 0 6556 491.6 710.2 1201A 0 6928 0 7377 1.4304 488.9 1116.9 700 l 833 51421 0 0209 0.54809 0.5690 5098 689 6 1199 4 0.7111 0.7051 1.4163 506.7 1115.2 880 l
900 Si845 0 0212 047965 05009 526 7 669 7 1196 4 0 7279 06753 1.4032 5232 It13.0 900 1000 5 * *.!,5 0.0216 042435 04463 542.6 f 50 4 1192.9 07434 0.6476 1.3910 53:16 1110.4 1000 1100 515.2d 0.0220 0 373(3 04006 557.5 631.5 11891 0 7573 0.6216 1.3794 553 1 1107.5 1100 1230 l 367.19 0 0223 0.14013 0.362h 571.9 613 0 1184 8 0.7714 0.b969 1.3683 5669 1104.3 1200 1803 577.42 0 0227 0 30722 0.3299 585.6 544.6 !!80 2 0.7843 05733 1.3577 580.1 1100 9 1300 1400 537,07 to 0231 0778/1 0 3018 598 8 576 5 1175.3 0.7966 05507 1.3474 592.9 1037.1 1400 8
1500 SW20 0 0235 02b372 0.27/2 611.7 558 4 1170 1 0 8035 0f283 1.3373 605 7 1093.1 1500 2000 635.80 0.02L 7 01676G 0 1883 672.1 466 2 1138.3 0 86.'t 0 4256 1.7s81 6626 10GS 6 2000 2500 65$ 11 0 02c,6 010209 0 1307 731 7 361.6 1093 3 C 9139 0 3206 1.2345 718.5 1032.9 2500 3000 69h33 0 0343 0 050/3 0.0850 801 8 218.4 1070 3 0 9728 0 1891 1.1619 7822 973.1 3000 3208.2 70LA7 0OSOS 0
0050d 906 0 0
906 0 1.06I2 0
1.0612 875.9 875.9 3208.2 TABLE A.3 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (PRESSURE)
A.4
--,----w m rv s eee
.---.e.'--mwwen-rmrs Pw*
Tonyeestwo, F Abe prose.
(
Se p) 100 200 300 400 900 000 MD 800 900 1000 1100 1200 1300 1400 1500 v 00161 392 5 452.3 511.9 571.5 631.1 690 7 3
4 68 00 1350 2 1195.7 1241.8 1288 6 13361 1384 S (101.74) e 0 129S 2 0509 2.1152 2.1722 2.2237 22708 2.3144 e
0 0161 78 14 9024 102.24 114.21 126 15 13508 150 01 f 6194 173 86 ISS 78 197 10 209 62 221 53 233 4 S
h 68 01 1148 6 1194 8 1241.3 12882 1335 9 13e4 3 1433 6 1883 7 1934 7 19867 1639 6 1893 3 17480 1 (162 24) s 0.1795 1 8716 1.9369 1.9943 2.0460 2 0932 213G9 21776 2 2159 2 2S21 2.28M 2.3194 23503 2.3811 2 e
0 0161 38 84 44 93 S1 03 57.04 63 03 69 00 74 98 80 M S4 91 92 SF 98 84 104 80 110 76 116 72 30 6
68 02 1146 6 11937 1240 6 12 6.8 13355 13840 1433 4 1483 5 1934 6 1546 6 1639 5 1493.3 1747.9 j
(192.21) s 0.1295 1.7928 1AS93 1.9173 1.9692 2.01M 2.0603 2.1011 2 1394 2.1757 2.2101 2.2430 22744 2.3046 2 l
e 0 0161 0 0166 29 899 33 M3 37.985 41.986 45.978 49 964 S3 946 57.926 41 905 65 882 69858 73.833 n S07 18 6
48.04 16809 1192.S 1239 9 1287.3 13352 13838 1433 2 1483 4 1534 S 1586.5 1639 4 1303.2 17474 geo3 4 l
(213.03) e 0.1295 0.2MO 1.8134 18720 1.9242 1.9717 2.01SS 2.0543 2.0946 2.1309 2.1653 2.1982 22297 2.2S99 2.2890 o
00161 0.0146 22.3S4 29428 2SAS7 31 446 34 465 37.458 40 447 43.435 44 420 49 40S S2.308 55.370 58.352 se a 68.05 148 11 1191.4 1239.2 12M.9 1334.9 1383 5 1432 9 1443 2 1534.3 1986.3 1639.3 1883.1 17474 1803.3 (227.M) s 0.1295 0.2940 1.7005 1A397 1A921 1.9397 1.9836 2.0244 2 0628 2.0991 2.13M 2.1465 2.1979 2.2282 22572 e
0.0141 0 0166 11.035 12 424 14.165 15 405 17.195 18 699 20190 21.687 23.lM 24 809 2E183 27.676 29.148 48 6
48.10 148.15 IIM 6 12M 4 1285.0 13336 1382.5 1432.1 1482.5 1533.7 lisSA 1638 8 1992.7 1747.5 1803.0 (267J5) s 0.1295 0.2940 1.89N 1.M00 1A143 1AU4 1.9085 1.9476 1.9800 2.0224 2.0589 2.0009 2.1224 2.1916 2.1807 e
0.0161 0.0156 7.257 83S4 9A00 10 425 11 438 12.444 13ASO 14 Alt 15.452 1E490 17As 18A et n
GS.15 let to 11816 1233.5 1283.2 1332.3 1381.5 1431.3 14812 1533.2 1985 3 1630.4 1g82.4 1747.1 1802A l (292.71) s 0.1296 0.2939 1.6492 1.7134 1.MSI 1A168 1A612 1.9024 1.9410 1.984 2.0130 2.0490 2A766 2.10 l
0.01'41 0 0166 0.0175 6.218 7.018 7.794 S.540 9.319 10.075 10.829 11 581 12331 13AB1 13A2B 14.577
{
00 6 48 21 168.24 269.74 1233.5 1281 3 1330.9 1380.5 1430.5 1481.1 15326 1984.9 1638.0 1402A ' 1746A 1802.5 e
i (312.04) s 0.1295 0.2939 0 4371 1.6790 1.7349 1.7842 13289 1 8702 1.939 1.9454 1.9000 2.0131 2AS46 2AP90 ' 2.1041 e
0.0161 0.0166 0 0175 4 935 S.588 4.216 6A33 7.443 8 050 S.MS 9 298 SASO ISAGS IIA 38 IIAe9 i
ISO n 68.26 168.29 269 77 1227.4 1279.3 1329 6 1379.5 1429.7 1480.4 1532.0 1584.4 1637.6 18D1A 174LS 1902.
(327A2) s 0.12M 0.2939 0.4371 14516 1.7006 1.7586 13034 1A451 13839 1.9205 1.9652 1.9003 2A199 2ASE 3.0794 i e
0.0161 0 01H 0 0175 4 0786 4.4341 S.1437 S.6831 4.1921 67006 7.2000 7.7006 82119 E7130 9J134 9.7130 '
120 A 48.31 168.33 269 81 1224.1 1277.4 1328.1 13784 14288 14798 ISSIA 1943.9 1637.1 ISplJ 1746.2 1802A l
(34127) s 0.1295 0.2939 0 4371 14246 1.4872 1.7376 1.7829 1A244 1.8635 1.9001 1.9349 1.9000 1.9996 I m 00 2.090 e
0 0161 0 0166 0 0175 3 4651 3 95M 4 4119 4450$ S.2995 S.7364 6.1709 4.4036 7A349 7 ASS 2 7AB48 S 140 a 68 37 168 38 269 SS 12208 1275.3 13268 13U.4 1428 0 1479.1 19308 1983 4 1636 7 1830.9 1745.9 180 (353 04) s 0.1295 0 2939 0 4370 1.6085 1.6486 1.71M 1.MS2 1.8071 1A441 1.8828 1.91M 1.9508 1.982$ 2Al29 3.042 e
0.0161 0 0166 0 0175 3 0000 3 4413 3 8480 4.2420 4 6295 S.0132 5394S S.ndl 4.1522 6.S293 '6.9098 7 100 A 68 42 168 42 26989 1217 4 1273.3 1325 4 13M 4 1427.2 1478 4 1$30.3 1582.9 1436.3 1980.5 1746.4 (M3 SS) s 0.1294 0 2938 0 4370 1.5906 1.6522 1.7039 1.7499 1.7919 1A310 1.8678 1.9027 1.9359 1 2 4.1.9900 e
0 0161 0 0166 0 0174 2 6474 3.0433 3A093 3.M21 4.1084 4.4505 4.7907 S.1289 544S7 58014 &l363 4 ISO 4 68 47 164 47 264 94 12138 1271.2 1324 0 1375.3 14263 1477.7 1529 7 1582.4 1435.9 1990.2 17alJ 1 (373.C41 s C.1294 0.2938 04370 1.5743 1 4376 3.4000 1 7362 1.7784 1A176 1.834S 1.8094 1.9227 1 9545 1.98 e
0 0161 0 0166 0 0174 2 3598 2.7247 3.0883 3.3783 3 M15 4 0006 4.3077 4.4128 4.9165 52191 S.52 200 >
68 52 16851 269 N 1210 1 1209.0 13224 1374.3 1425.5 1477.0 15291 1581.9 1635.4 1408 8 174S (35140) s 0 1294 0 2938 0 43G9 1.5593 1A242 14776 1.7239 1.7643 1.8057 1 8426 1A774 1.9109 1A427 1.971 e
0 0141 0 0lM 0 0174 0 0104 2.1904 2 4442 2.8872 2.9410 3.1900 3 4382 3 0837 3.9278 4.1700 4.4131
-a.
ISO 4 68.M 168 63 270 05 3/S.10 I M 3.5 1319.0 1371.6 1423 4 14753 1527.6 1980.6 1634A ISS.9 1744J2 (430 97) s 0.1294 0.2937 0 4368 0.S M 7 1.5951 1.6502 1.6976 1.740S 1.7501 1.8173 latte IASSS lAIF7 1.9482 1 e
0 0161 0 0165 0 3174 0 0186 1.7645 2.0044 22M3 2.4407 2.6509 2 8585 3 0643 3.2008 3.4721 34746 340 a 64 19 IM 74 27u 14 375.15 1257 7 1315 2 1368 9 1421.3 14736 1526.2 1579 4 1633 3 1408 0 1743 (417.35) s 0 1294 0 2S37 0 4737 CS%$
1.5703 1.6214 1.6758 1.7192 1.7591 3.7964 13317 1A6S2 14972 1.927 e
0 0161 0 0166 0 0174 0 0186 1.4913 1.7028 1.8973 2 0332 2 M52 2 444$ 24219 2.7980 2.9730 3.14 350 m C898t 165 85 270 24 375 21 1251 5 13114 IM6.2 1419 2 1471 8 1524.7 1578.2 1632.3 1E87.1 1742 (431.73) o 01293 0 29M 0 43G7 0.5664 1.S483 1.6077 14578 1.7009 1.7431 1.7787 1.8141 1A4D 13798 1.9105 e
0 0161 0 0166 0 0174 0 0162 12N1 1.4763 16497 1.8151 1.9759 2.1339 2.2901 2A490 2.9007 2.75 400 a 69 05 les 97 270 33 375 27 1245 1 1307.4 iM34 1417.0 14701 1523 3 ISM.9 1631.2 1406.2 1741 (444.60) s
$ 1293 0 2935 0 4M6 0 56G3 1.S782 1.5901 1 6406 1.6850 1.72Sb I.7632 1.7908 1A32S 1A647 1.
e 0 0161 0 0166 0 0174 0 0186 0 9919 1.1584 1 3037 1.4397 1.5708 1 6992 1.8254 1.9507 2A746 2.
SCO 4 09 32 Itl l9 270S1 3?5 38 1231.2 12991 13S7.7 1412 7 1466 6 1520 3 1574 4 1629.1 1884 4 17 (457.011 s 0 1292 0 2934 oAM4 0St60 1 4971 1$$95. 1.6823 1GS/S 14990 1.7371 1.7730 1A069 1A393 1 8 TABLE A.4 PROPERTIES OF SUPERHEATED STEAM AND COMPRESSED WATER (TEMPERATURE AND PRESSURE)
A.5 u
' fp T2mperatres, F
];
Closim eat.assel 100 251 300 400 S00 000 700 800 000 1000 1100 3200 1300 1400 1900 00161 0 01M 0 0174 0 0186 0 7944 0 94M I0726 11892 1.3008 14093 15160 16711 1.7252 18?M 19309 400 6 69 58 169 42 270 70 375 49 1215 9 1290 3 1351 8 1408 3 1463 0 1517 4 1571 9 1627.0 1682 6 17388 !?95 6 (486J0)s 0.1292 0J933 0AM2 0.M57 14590 1.5329 15844 16353 1 6769 1 71 % 17517 11859 1.8184 I8494 I8792 e
0.0161 0 0166 0 0174 0 0186 0 0204 0 7928 0 9072 1.0102 1.1078 12023 12948 I.3858 1.4757 I M47 16530 Fee &
69 84 169 65 270 89 375 61 447 93 1781 0 1345 6 14037 1459 4 1514 4 1%94 16248 IM07 1737 7 1794 3 (503.C8) s 01291 0.2932 0 4360 0%% 06889 35090 1.M73 16154 14540 16970 17335 17679 1 00 % I8314 I8617 e
0 0161 0 0!M 0 0174 0 0186 0 0704 0 6774 0 7823 0.8759 0 9631 1 0470 1 1249 1 2093 17825 13669 1.4446 808 4
/0.31 16988 271.07 375 73 487As 1271 1 1339 2 1399.1 1455 A 1511 4 15M 9 1U27 16789 1715 0 1792.9 (5182.)*
0.1290 0 293C 0 4358 O MS2 0.6885 1 4869 1 5484 1.5Mo 16413 16807 1.7175 1 7522 1.7853 1 8164 18M4 e
0.0161 0.0166 0 0174 0 0186 0 C2'J4 05869 O M58 0 7713 0 8504 0 9262 0 9998 1 0720 1.1430 1.2131 1.2825 I
808 6 70 37 17010 211.26 375 84 487 83 1260 6 13327 1394 4 1452 2 1508 5 1M44 1620 6 16771 17M 1 1791 6 (531.M)s 0.1290 0.2929 0.4357 O M49 0 6081 1.4659 1.5311 1.M22 1.6263 1 M62 1.7033 1.7382 13713 1 8028 1 8329 I
e 0 0161 00lM 0.0174 0 0186 0 0204 0 5137 0 6000 0 4875 0.7603 0 8295 0 8964 O M22 1.0766 1.0901 1.1529 38g8 6 70.63 170 33 271.44 375.M 487J9 12491 1325.9 1389.4 1448.5 1504.4 I M I.9 1618 4 1675.3 1732.5 1790 3
($44.ht) s 0.1288 0.2928 0.4355 0.M47 0.4876 1A457 1.5149 1.M77 1.61M I6630 1A905 132M IJS89 1.7905 1.4207 e
0 0141 0 0166 0 0174 0.0185 0 0203 0 4S31 O S440 0 6108 0 6465 0 7505 0 8121 08723 09313 0 98M 1.0468 11st 4 70 90 170.M 271.63 376 08 487.75 1237.3 1318 8 1384 7 1444 3 1502 4 1559.4 1616 3 1673.5 1731.0 !?89.0 1
(SM28) s 0.1289 02927 0.4353 0.5 H 4 0.4872 1A259 1.4996 1.H42 1.4000 1.6410 1.6787 1J141 1.7475 1.7793 1A097 i
i e
0 0141 0.01H 0.0174 0.0185 0 0203 0.4016 0 4905 0.Hll 0 6250 0.4845 0 7418 C.7974 0 8519 0.9059 0.9584 1388 4 71.16 170.78 271A2 374.20 487.72 1224.2 1311.5 1379.7 1440.9 1449 4 1556 9 1614.2 1671.6 1729 4 1787.6 (S67.19) s 0.1288 0.29 N 0.4351 0.6642 OAMS 1A061 1A861 1.M15 1.M83 14298 1 6479 13035 1.7371 1.7681 1.7986 e
0.0141 0 0164 0 0174 0 01SS 0 0203 0.3176 0 4059 0.4712 0.5282 0 5809 0 6311 0 67M 0.7272 0.7737 08195 I
)
1400 4 71.48 17124 272.19 376 44 487 65 11M.1 12961 1369J 1433 2 1493 2 1551.8 1609.9 1648.0 1726.3 17860 (S87A7) s 0.1287 OJ923 0.4348 0.5636 0 4859 IJ652 1AS75 1.5182 1.5670 1.1096 1A484 1.6845 1.7185 1.755 1.7815 e
0.0161 0.0lM 0.0173 0.0185 0 0202 0.02M 0.3415 0 4032 0 4S55 0.5031 0 5482 0 5915 0 8386 067M 0.7153 lass 4 72.21 171 69 272.57 376 69 487.60 616.77 1279.4 1358 5 1425.2 1486.9 IM64 1605.6 1864.3 1723J 1782J (804.87) s 0.1286 0.2921 0.4344 0.M31 0.4851 OA129 1.4312 1.4M8 1.5478 1.5916 1.6312 1.M78 1.7W2 1.7344 1.M67 l
0 0160 0.0165 0 0173 0.0185 0.0202 0 0235 0 2906 0.3900 0.3988 0 4426 0 48N 0.5229 0.StGB 0.S980 0 6743 e
1808 a 72.73 172.15 272.95 376 93 487.M 615.S8 1261.1 1347J 1417.1 1480.6 I M I.1 1401.2 1960.7 1720.1 1779.7 tul/32) s 0.1284 0.2918 0.4341 0.S426 0.6r3 0.8109 1A054 1APG8 1.5302 1.5753 1.4IM 1.4628 1.GSM 1.7204 1.7516 e
0 0160 0.0145 0.0173 0 0184 0.0201 0.0233 0.2488 0.3072 0.3534 0.3942 0 4320 0 4680 0.557 0.SMS 0.5895 2000 4 73 26 172 60 273 32 377.19 487 53 414 48 1240.9 1353 4 1408 7 1447.1 1536 2 1996.9 1457.0 1717.0 1777.1 (635 00) s 0.1263 0.2916 0.4337 0.H21 0 6834 0.8091 1.3794 1.4578 1.5138 1.56C3 1.6014 14391 1.6743 1.707S 1.7389 e
0 0160 C.0165 0.0173 0 0184 0.0200 0.0230 0.1681 0 2293 0 1712 0.3068 0.3390 0 3692 0.3900 04259 OAS28 2500 &
74 57 17334 274 27 377 82 487.50 612.08 1176 1 1303 4 1386.7 14S7.9 1522.9 1S86.9 1447A 1709J 1770.4' (4 8.11) s 0.1200 0 2910 0 4329 0.M09 0.4815 0 8048 1.3076 1.4129 1.4766 1.5269 1.5703 1.0084 1A4E.1.6796 1.7116 e
0 0160 0 0165 0.0172 0 0183 0 0200 0.0228 0 0982 0 1759 0.2161 0.2484 0.2770 0.3033 0.3382 0.3522 c.3753 3000 4 75 53 17t S8 275.22 37847 487.52 610.08 1060.5 1247.0 1363.2 1440.2 1500.4 1574.8 1638.5 1701.4 1M1.5 (695.33) s 0.1277 0.2 % 4 0.4320 0.S$97 0.67M 0.0009 3.lM4 1.3692 1A429 1A976 1.5434 1.M41 1421A ),6l$1 },gggg l
e 0 0160 C 0165 0.0172 0.0183 0.0199 0.0227 0 033S 0.1588 0.1987 0.2301 0.2S76 0.2827 0J065 0.3291 0.3510 l
3200 &
76.4 175.3 275 6 3787 447.5 005.4 000A 1250 9 1353.4 1433.1 1S03.8 1570.3 14343 1698.3 I MlJ r705 08} s C 1274 0 2902 0.4317 0.5592 0.6788 0.7994 0.9708 1.3515 1.4300 1ASM 1.5335 1.9749 1AIN 1.M 77 IJ806 0 0160 3 0164 0.0172 0.0183 0 0199 0.022S 0.0307 0.1364 0.1764 0 2064 0 2326 0.2563 0.2784 0.2995 0.319B e
3500 4 77.2 174.0 276.2 379.1 487.4 608.4 779.4 12246 1338.2 1422 2 1495.5 IMS.3 1429J 1093.6 1767.2 s
0.1274 0.2899 0.4312 0 5545 0 6777 03973 0 9508 1.3242 1All2 1.4709 1.51M 1.M18 1.0002 1.4358 1A081 e
0 0159 0.0164 0.0172 0.0182 0.0198 0 0223 0 0287 0.1052 0.1463 0.1712 0 1994 0.2210 0.2411 0.2801 0.2783 4000 6 78.5 177.2 277.1 3798 487.7 606 5 763 0 1174.3 1311.6 14036 1481.3 1552.2 1619A 1885.7 17S04 s
31271 0.2893 0.4304 0 5573 0 6760 0 7940 0.9M3 1.2754 1.3807 1.4461 1.497G 1.5417 1.5812 1 A177 1.6514 e
0 0159 0 0164 0 0171 0 0181 0.0!M 0 0219 0.0268 0.0591 0.1038 0 1312 0 1529 0.1718 0 1890 0 2050 0.2203 5000 a 81 1.
179 5 2791 381.2 488.1 604 6 744 0 1042.9 1252.9 1364 6 1452.1 1529.1 16039 1870.0 1737.4 s
0.1Mf*0.2M1 0 4287 0.5550 0 6726 03880 0.9153 1.1593 IJ207 IA001 1 ASS 2 1.9081 1.5481 1.M63 1A214 e
0 0159 0.0163 00170 0 01In0 0 0195 00216 0.02S6 0.0397 0.07S7 0.1020 0.1221 0.1391 0.lM4 0.!684 0.1817 6000 4 83.7 181.7 281.0 342 7 #86 602 9 736 1 945.1 1184 8 1323 6 1422.3 1505.9 IM20 1654.2 17242 s
0.1258 0.2670 0 4271 0$$28 0 6693 OJ8M 0 9026 1.0176 1.2615 1.35?4 1.4229 1.4745 1.S194 1.5593 1586.2 e
0.0158 0.0163 0 0170 0 0100 0 0193 0 0713 0 0248 0.0334 0 0573 0 0SI A 0.1004 0.1180 0.1250 0 1424 0.1542 7000 4 84J 184 4 283 0 3s42 489 3 401 3 729 3 901.8 1124.9 1281 7 1392 2 1482.6 1543.1 163R 6 1781.1 a
fil252 0 2859 0.42M 0 5%07 O M63 03/77 0 8926 103SO 12055 1.3 D 1 11904 1.44u4 1.4938 1.53'S l.S735 TABLE A.4 PROPERTIES OF SUPERHEATED STEAM AND COMPRESSED WATER (TEMPERATURE AND PRESSURE) (CONTINUED)
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Density e (Ibsitt*)
PSIA Temp Saturated
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Liquid 1000 2000 2100 2200 2300 2400 2500 3000 32 62.414 62.637 42.846 62.067 82.888 82.909 62.93 82.951 63.056 l
50 62.38 62.55 82.75 62.774 62.798 62.822 82.846 82.87 62.99 100 61.989 82.185 62.371 62.390 42.409 82.427 62.446 62.465 62.559 200 40.118 90.314 60.511 40.53 80.549 80.568 80.587 80.806 80.702 l
300 67.310 57.537 57.767 57.79 57.813 57.838 57.850 57A02 67.998 1
400 63.651 63.903 54.218 54.249 54.28 54.311 64.342 64.373 64.529 410 53.248 53.475 63.79 53.825 63.86 53.89 63.925 63.95 64.11 420 52.798 53.025 53.36 63.40 63.425 63.46 53.50 63.53 63AB 430 52.356 52.575 52.925 62.95 62.99 63.02 63.065 53.09 63.265 440 51.921 52.125 62.42 52.45 52.475 62.51 82.54 62.58 52.275 450 61.546 51.66 52.025 62.085 52.10 62.14 82.175 62.21 82.41 460 51.020 51.175 51.56 51.61 51.84 61.68 61.725 61.78 51J8 j
470 50.505 50.70 51.1 51.14 51.175 51.22 61.25 51.30 51.80 480 50.00 50.20 50.62 50.68 50.7 80.74 80.78 60.825 51A35 4M 49.505 49.685 50.13 50.175 50.22 50.265 50.31 60.35 50 575 500 48.943 49.097 49.618 49.666 48.714 49.782 49.41 49.858 50.000 610 48.31 48.51 49.05 49.101 49.152 49.203 48.254 48.306 49.58 I
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630 47.17 47.29 47.86 47.919 47.978 48.037 48.008 48.156 44.45 540 46.51 47.23 47.296 47.362 47.428 47.494 47.58 47A9 550 45.87 46.59 46.658 46.726 45.794 46862 46.93 47.27 560 45.25 45.92 45.994 46.068 46.142 46.216 48.29 46.06 46.02 570 44.64 45.22 45.30 45.38 45.46 45.54 45.62 540 43.86 44.50 44.586 44.t 72 44.756 44A44 44.93 45.36 550 43.10 43.73 43.825 43.92 44 015 44.11 44.205 44.08 600 42.321 42.913 43.017 43.122 43.226 43.33 43.434 43356 610 41.49 41.96 42.08 42.196 42.314 42.432 42.55 43.14 820 40.552 40.950 41.083 41.217 41.35 41.483 41.816 42J83 41.44 630 1 9.53 40A88 640 38.491 39.38 650 37.31 30.008 660 38.01 38.52 670 34.48 34A38 683 32.744 32.144 690 30.518 TABLE A.6 PROPERTIES OF WATER, DENSITY A.8
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3/a.
APPLICA3:L* Y..............................................
3/4.
?IACT!V!"*? CC: RCL SYS"' EMS 3/a 1-1 3/4.1.1 SH t'I DCWN MARG I N..........................................
3/4.1.2 REACT IV U"! AN0MALIE S.....................................
3/4 1-2 3/4.1.3 CONTROL RODS 3/4 1-3
'i Control Rod Operability..................................
i Control Rod Maximum Scram Insertion Times................
3/4 1-5 Control Rod Average Scram Insertion Times................
3/4 1-6 Four Control Red Group Insertion Times...................
3/4 1-7 Co nt rol Rod Sc ram Ac cumulat o rs...........................
3/4 1-8 3/41-4 Con trol Rod Drive Coupling...............................
Centrol Red Position Indication..........................
3/4 1-11 Control Rod Drive Housing Support........................
3/4 1-13 I
3/4.1.4 CONTROL RCD PROGRAM CONTROLS 3/4 1-14 Rod Worth Minintzer......................................
3/4 1-15 Rod Sequence control System..............................
3/4 1-17 Rod Riock Monitor......... -............................
3/4 1-18 3/4.1.5 STANDST LIQUID CONTROL SYSTEM............................
3/4.2 POWER DISTRIBITTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATIONN RATE..............
3/a 2-1 3-::
3/4.2.2 AF P.M S IT P o : :: S...........................................
I.- 2-9 3/4.2.3 l11 NIM LN CRITICAL POWER RAT
- 0.............................
3'a 2-15 l
3/4.2.4 LINEAR HEAT GENERATION RATE..............................
l 33*.2:S*!* 0X
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i;
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RNDEX 1
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIR PAGE SECTION 3/4.3 INSTRUMENTATION 3/4 3-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION...................
3/4.3.1 ISOLATION !.CTUATION INSTRUMENTATION.........................
3 /4 3-9 3/4.3.2 T24ERCENCY CORE COOLING SYSTEM ACTUATION INSTRLMENTATIO 3/4 3-;;
3/4.3.3 CONTROL ROD WITEDRAWAL BLOCK INSTRUMENTATION................ 3/4 3-39 3/4.3.4 3/4.3.5 MONITORING INSTRUMENTATION Seismic Monitoring Instrumentation.......................
3/4 3-44 Remote Shutdown Monitoring Ins trumentation...............
3/4 3-47 Accidsnt Monitoring Instrumentation......................
3/4 3-50 Source Range Monitors....................................
3/4 3-53 Chic,rine De tection Sy s t em................................
3/4 3-54 Chlo ride Intrusion Monito rs..............................
3 /4 3-55 Fire Detection Instrumentation...........................
3/4 3-59 Radioactive Liquid Ef fluent Monitoring Instrumentation....
3/4 3-61 Radioactive Gaseous Effluent Monitoring Instrumentation...
3/4 3-66 >
RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION 3/4.3.6 ATWS Racirculation Pump Trip System Instrumentation......
3/4 3-78 End-of-Cycle Recirculation Pump Trip System Ins trume nt at ion........................................
3 / 4 3-8 2 REACTOR ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTA 3/4 3-88 3/4.3.7 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM Recir culation Loops......................................
3/ 4 4-1 J e t Pumps................................................
3 / 4 4 - 2 s
Idle Racirculation Loop Startup..........................
3/4 4-3 l
SAFETY / RELIEF VALVES........................................
3/4 4-4 3.4.4.2 l
J 4 Amendment No. 99 V
SRUNSWICK - UNIT 2 l
- - ~ - - - - -
INDEX REQUIREMEN*S LIMI; NG CONOIMONS FOR OPERA!!ON AND SURVE!!. LANCE PAGE SECION REAC*CR COCLANT SYSTEM LEAKAGE 3/4 4-5 3/a.4.3 Leakage ':ecection System.................................
3/4 4-6 Op e r a t ional '.4 aka ge......................................
3/4 4-7 CHEMISTRY................................................
3/4 4-10 3 / 4. 4..
SPECIFIC AC"!VITT........................................
3/ 4.4.5 PRESSURE /~EMPERATTRE LIMITS 3/4 4-13 3 / 4.4.6 Systes...................................
Raactor "co.lan:
3/4 4-13 te a c t o r S t e am 0c = e..................................
3/4 4-19 MAIN SHAM LINE ISCLATION 7AL7ES.......................
3/ 4. 4.7 3/4 4-20 STRU CT"RAL INTEGRI~T.................................
3/4.4.8 EMERCENCY CORE C0 CLING SYSmS_
3/4 5-L 3/4.5 HIGH PRESSURE C00LAST INJECT CN SY5m.................
3/4 5-3 3/ 4. 5.1 AU*0MATIC DEPRESSURI ATION SYSTEM...................
3 / 4. 5.2 LCW PRESSURE COOLING SYSTEMS 3/4 5-4 3/ a. 5. 3 Core Spray Systes........................................
3/4 5-7 Low Pressure Coolane In jec: ion Sys tas....................
3/4 5-9 SiFPRESSIONP00L.........................................
3/ a.5.4 CONTAINMER SYSMS 3/4.6 PRIMARY COE AINMENT 3/4 6-1 3/4.6.1 In t e 5 r i t y............................
Pr13ary CantainaB9nt 3/k 6*1 IASkage.....i........................
Primary Containment 3/4 6=4 Air tack.............................
?timary Containment VI SRUNSWICK - UNIT 2 RETTFE TEG. SPEC 3.
Updated Thru. Amend. 73
i KNDEX LIMITING CONOI!!ONS FOR OPERATION AND SURVEILLANCE REQUIRE.ENTS PAGE
%CTION 3.4.6 CONTAINMENT SYSTEMS (Continued)
Primary Containment Struc tural Int e grity.................
3/466 3/4 6-7 Primary Containment Internal Pressure.............
Primary Containment Average Air Temperature..............
3/4 6-8 3/4.6.2 DEPRESSURI::ATION AND COOLING SYSTEMS 3/4 6-9 S upp r e s s i o n Po o l.........................................
3/4 6-11 S; p p r e s s ion Pool Co o lin g.................................
3/4 6-12 3/4.6.3 PRLMARY CONTAINMENT ISOLATION VALVES.....................
3/4.6.4 VACUUM RELIEF 3/4 6-18 Dryvell - Suppression Pool vacuum Breaker s...............
3/4 6-20 Suppression Pool - Reactor Building Vacum 3reakers......
3/4.6.5 SECONDARY CONTAINMENT 3/4 6-21 Secondary Containment Integrity..........................
Secondary Containment Automatic Isolation Dampers........
3/4 6-22 3/4.6.6 CONTAINMENT ATMOSPHERE CONTROL 3/4 6-25 Standby Gas Treatment System.............................
3/4 6-28 Contain: cent Atmosphere Dilution System...................
3/4 6-29 0xy gen Concentra t ion.....................................
3/4 6-30 l
p s Analyz er S y s t ems.....................................
l 3/4.7 PLUTSYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS 3/4 7-1 f
Residual Haa c Removal Service Wate r Sys tem...............
3/4 7-2
= Service Wacer System.....................................
l l
l SRUNSWICK - UNIT 2 VII RETTPED TECE. SPECS.
Updated Thru. Amend. ;
ENDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE stCTION_
J/4.7 ?LAST SYSTdMS (Continued) 3/4. 7.2 CONTROL ROOM EMERGENCY FILIRATION SYd ii.4.................
3,. ;- 3 3/ 4. 7.3 FLOOD PROTECTION.........................................
jf. ;-o 3/ 4. 7.4 REACTOR CORE ISOLATION COOLING SYSTEM....................
3/. 7 7 3/ 4. 7.5 HYDRAULIC SNUBSERS.......................................
3/a 7-9 l
3/ 4. 7. 6 SEALED SOURCE CONTAMINATION..............................
3/4 7-15 3/4.7.7 FIRE SUPPRESSION SYSTEMS 3/4 7-17 Fire Suppression Water System............................
3/4 7-20 S p ray and / or S p rinkler Sys tems...........................
Systems................................
3/4 7-22 High Pressure CO2 3/4 7-23 Fire Hose Stations.......................................
3/4 7-26 Foaa Systems.............................................
3/4 7-28 3/ 4. 7.8 FIRE BARRIER PdNETRATI0NS................................
3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES 3/4 8-1 Operation of One or Bo th Units...........................
3/4 8-5 i
Shutdown of Both Units....................................
3/4.8.2 ONSITE POWER DISTRIBUTION SYSTE!G
=-
A.C. Distribution - Operation of One or Both Units.......
3/4 8-6 A.C. Distribution - Shutdown of Both Units...............
3/4 8-7 3/4 8-8 D.C. Distribution - Operating............................
l 3/4 8-11 l
D.C. Distribution - Shutdown.............................
3/4 8-12 RPS Elect ric Power Monitoring............................
3/4.9 REFUELING OPERATIONS 3/4 9-1 3/4.9.1 REACTOR MODE SWITCH......................................
BRUNSWICK - UNIT 2 VIII Amendment No.100 J
LIM 1*iIhr. CONDIT10i, /OR OPERATION 1
Limitina Conditions for Operaticn cnd ACTION requirementa, shall he 1.0 t applicable durtnet the OPERATIONAL CONDITIONS or other states specified for each specification.
1.n.2. Adherence to the requirements of the Limiting Condition for Operation and associated ACTION within the specified time interval shall constitute In the event the Limiting Condition for compliance with the specifiestion.
operation is restored prior to expiration of the specified time interval, completion of the ACTION statement is not required.
a Limiting Condition fer Operation and/or associated 1.0 3 In the event be satief ted because of circuestances in excees of ACTION requiresents cannot those addressed in the specifiestion, the unit shall be placed in at least NOT SNUTn0WN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SNUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> imless corrective seasures are completed that permit operation under the permissible ACTION statements for the specified time interval as measured free initial discovery or until the reactor is placed in an OPERATIONAL CONDITION in which the specifiestion is not applicable.
Exceptions to these requirements shall he stated in the individual specifications.
Entry into an OPERATIONAL CONDITION or other specified applicability
- 1. n. I, state shall not be made unless the conditions of the Limiting Condition for without reliance on provisions contained in the ACTION Operation are met statements unless otherwise excepted. This provision shall not prevent passage through OPERATIONAL CONDITIONS required to comply with ACTION requirements.
When a system, subsystem, train, component, or device is determined to 1.0.5 he 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 emerzency power source is OPERABLE; and (2) all of its redundant syates(s),
subsystems (s), train (s), component (s), and device (s) are OPERABLE, or likewise Unless both conditiens (1) satisfy the requirements of this specification.
least NOT SRUTDOWN and (2) are satisfied, the unit shall be placed in at within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in at least COLD SRUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
This specification is not applicable in Conditions 4 or $.
i
=-
e RRUNSWICK - UNIT 2 3/4 0-1 RETTPED TECH. SPECS Updated Thru. Acend l
-___.,_____,,..__________,m.,
l 3/4.3 INSTRUMENTATION _
SEACTOR PROTECTION SYSTEM INSTRUMENTATION 3/4.3.1 LIMITING CONDITION FOR OPERATION As a minimum, the reactor protection system instrumentation chann M
shown in Table 3.3.1-1 shall be OFFRAnfF with REACTOR FROTECTI 3.3.1 RESPONSE TIME as shown it. Table 3.3.1-2.
in Table 2.2.1-1.
As shown in Table 3.3.1-1.
AFFLICARILITY:
ACTION:
With the requirements for the minisua number of OPERAELE chmanals n i
l' satisfied for one trip system, place the inoperable channel (s) and/o a.
system in the tripped condition
- within one hour.
With the requirements for the miniaura sunber of OPERAELE channels satisfied for both trip systems, place at least one. trip systes** in b.
N required by the tripped condition within one hour and take the ACTIO l
Table 3.3.1-1.
l The provisions of Specification 3.0.3 are not applicable in OPERA c.
CONDITION 5.
SUEVEILLANCE REQUIREMENTS Each reactor protection systes instrumentation channe'l shall be demonstrated OPERAELE by the performance of the CHANNEL CHE 4.3.1.1 NAL CALIBRATION and CBANNEL FUNCTIONAL TEST operations during CONDITIONS and at the frequencies shown in Table 4.3.1-1.
14GIC SYSTEM FUNCTIONAL TESTS and simulated automa i
h ll include all channels shall be performed at least once per 18 months as d s a 4.3.1.2 functioning calibration of time delay relays and timers necessary for proper of the trip system.
aEAf'FOR FROTECTION SYSTEM RESPONSE TIME of each reac 4.3.1. E The
?.3.1-2 shall be demonstrated to be within its function of Table h
nd'oam least see pe: 18 months.
such thac noth logic trains are tested at least onc N
hmanals in a l
tiana 13 months where N is the total number of redundant e speci(ie reactor trip function.
- An inoperable chesnel need not be placed in the tripped condit In these cases, the this would cause the Trip Fasetion to occur. inoperable i
shn11 be the ACTION required by Table 3.3.1-1 for that Trip Funct on If more channels are' inoperable in one trip system than in the ot taken.
in the tripped nlace the trin svptea with more inoperable chamaalsbonditi Ameadsent No. 105 3/4 3-1
~~--
taDN5WICE - UNIT 2- - - -
i I
i TABLE 3.1.I-I s as E,
REACTOR PROTECT 10N SYSTEM INSTRIMENTATION li M k
APPLICAulE HLNLMIM IRRIBER 4
j Q OPERATinNAL OPERABLE CIIANNELS ACTION PER TRIP SYSTEM (a)_
]
CONotTisNIS FUNCTtollAl.,tl NIT AND INSTRUNE,NT NLRtEF.R l
N 1.
Intermediate Range Monitures (C51-INM-R60lA,B,C.D.E.F.C.M) 2, 5(b) 3 g
M J
Neut run Flum - High 2
2 a.
j 3, 4 3
1 i
2, 5 b.
Inoperative 2
2 3, 4 i
Average Power Range Monitor 2.
(CSI-APRM-CH.A.B.C.D.E.F) j Ib) 2 3
Neutron Flum - High, 151 2, 5 a.
2 4
b.
Flow Blaced Neutron Flum - High 2
4 Fixed Neutron Flum - High,1201 1
1 u
l 4
c.
2 5
I, 2, 5 d.
tuoperative 2
4 1
e.
Downscale l
(c)
NA 1, 2, 5 f.
LPNH II 2
6 3.
Reactor Vessel Steam Dome Pressure - High I, 2 (B21-PT-Nu23A 5,C,0) l (B21-F1M-N023A-t,B-l,C-1 D-1) 2 6
l 4.
Reactor vessel Water 14 vet -
1, 2 Low, level 1 (521-I.T-Nu t 7 A-1,5-1.C-8,D-1)
(521-I.TM-N017A-1,5-1.C-l, D-1) j "I
4 4
1 5.
Main Steam teolation Valve - Closure I
(B21-F022A,5,C,0 and B21-FU28A,5,C.D) 7 j
I, 2(d) 2
[4 6.
Main Steam Line Radiation - High 5R (Os2 hA-K603A,B,C.D)
.n. sn i
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TABLE 3 3 1-1 (Continued) sEAcr0R PnerterION sTSTEM INrrRUMENTATION ACTION 1
IETf SIUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
In CONDITION 2. be in at least 1
ACTION I In CONDITION 5. suspeed all operations involving Cott ALTERATIONS or positive reactivity changes and fully insert s I
insertable control rods within een hour.
Iack the reactor mode switch in the Shutdown position withis ACTION 2 one hour.
In OPERATIONAL CONDITION 2. be in at Isaat W S'EDT ACTION 3 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
In OPERATIONAL CONDITION 5. suspend all operations involving CORE ALTDATIONS or positive reactivity changes sad fully insert. all insertabla control rods within one hoc.r 5e in at least START-UP within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
AC"*.ON 1 In OPERAn0NAL C::NDITION : or 2, 5e in at issse 30T 3H1'100WN AC!0N 5 wichta 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
In OPTRATIONAL CDNDITION 5. suspend all operations trivolvins l
CORE ALTERATIONS or positive reactisity changes and fully all insertable control rods within one hour.
l insert 8e in at least MOT SHUTDtMt within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
ACION 6 isolation vs17es closef.
Se in START-CP with the 1 min stesa line within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or in at least NOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
ACION
- Initiate a reduction in INERMAL POWER within 15 siautes less than 30% of RAIED TERMAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
ACION 3 at Leest 'l0T 5NUTDOWN SPERAMONAL CDNDITION 1 or 2. be in at 6 1 Pt
- n 2
si:hin ' hours.
'st?.C 7:A; ;- r :T-s : 3r.. i=nedia:elr sne sc '.=' - tecs
-:er.: : 4rs rer:fv : hat si; :.:--; - 's m '".. r. - e -i so
!,1?t?.C:9NAL CONDITION 5. suspend all soerst:sns tr.vol-sing
- s- !=4 m 'i.1v
-82 s.N?.C--"! -r tagitire -geatir.:r
.-ser:
6...r.ser 2:.3..::
1 3/4 3-4 BRUNSWICK - UNIT 2 RETTFED TECE. SPECS.
Updated Thru. Amend.
1
_...-_.._.__...-,__,_.._.___.,__,m_,-.__
TABLE 3.3.1-1 (Continued)_
RIACTOR PROTECTION SYSTEM INSTRUMENTATION O
least MOT SRUTDOWN In OPERATIONAL CONDITION 1 or 2, he in at ACTION 10 -
within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
In OPERATIONAL CONDITION 3 or 4, lock the reactor mode switch in the Shutdown position within ons hour.
In OPERATIONAL (DNDITION 5, suspend all operations involvine CORE ALTERATIONS or positive reactivity i.hanges s d fully all insertable control rods withic one hour.
insert TABLE NOTATIONS A channel may be placed in en inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance without placing the trip system in the tripped s.
condition, provided at least one OPERABLE channel in the sans trip system is monitoring that parameter.
orior to and The " shorting links" shall be removed from the RPS circuitry during the ties any control rod is withdrawn
- and durina shutdown margin b.
demonstrations.
LPRM inputs per An APRM channel is inoperable if there are less than :
level or less than eleven LPRM inputs to an APRM channel.
c.
[
These functions are not required to be OPERABLE when the reactor pressure d.
vessel head is unbolted or removed.
required to be OPERABLE when *RIMARY CONTAINMENT This function is not e.
INTEGRITT is not required.
oc applicable to control rnds remot'ed f.
With any control rod withdrawn.
per Specification 3.9.10.1 or 3.9.10.2.
less than 10% of RATED These functions are bypassed when IMERMAL POWER is 4
THERMAL POWER.
- 1. 0. L1. ' or 3. 0.10. 2.
i required for control rods removed oer Specification
- Not e
3/4 3-5 BRUNSWICK - UNIT 2 RETTFED TECR. SPECS.
Updated Thru. AWmd 7 l
INSTRUMENTATION ISOLATION ACIVATION INSTRUMENTATION 3/4.3.2 LIMITING CONDITION FOR OPERATION The isolation actuation instrumentation channels shown in Table 3.3 shall be OPERABLE with their trip setpoints set consistent with the values 3.3.2 shown in the Trip Setpoint column of Table 3.3.2-2 and with ISOLATION SYSTEM BESPONSE TIME as shown in Table 3.3.2-3.
As show/n in Table 3.3.2-1.
APPLICABILITY:
ACTION:
With an' isolation actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values a.
column of Table 3.3.2-2, declare the channal inoperable and place the inoperable channel in the tripped condition until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
With the requirements for the miniaua number of OPERABLE channels i
not satisfied for one trip system, place the inoperable channel (s) b.
l and/or that trip system in the tripped condition
- within one hour.
With the requirements for the minimum number of OPERABLE channels c
not satisfied for both" trip systems, place at least one trip systes**
c.
in the tripped condition within one hour and take the ACTION required by Table 3.3.2-1.
Tbs provisions of Specification 3.0.3 are not applicable in
~
d.
OPERATIONAL CONDITION 5.
SURVEILLANCE REQUIREMENTS Each isolation actuation instrumentation channe.1 shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL
'4.3.2.1 CALTanat105 and CIANNEL FUNCTIONAL TEST operations during the OPERATIONAL CONDITIONS and at the frequencies'shown in Table 4.3.2-1.
LOGIC SYSTEN FUNCTIONAL TESTS and simulated automatic operation of all ch====1= shall be performed at least once per 18 months and shall include 4.3.2.2 calibration of time delay relays and timers necessary for proper functioning of the trip syntes.
- An inoperable channel naad not be placed in thu tripped condition where In these cases, the this would cause the Trip Function to occur.
inoperable channel shall be restored to OPERABLE status withi= 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
- If more channals are inoperable in one trip system than in the other, taken.
stem with more inoperable chanaals in the tripped 168EIslon,trin s*t when tHe would cause the Trip Fanction to occur.
1 the estep Amendment No. 105 BRUNSWICE - UNIT 2 3/4 3-9
I INSTRUMENTATION e
SURVEILLANCE REQUIREMENTS (Continued)
The ISOLATION SYSTEM RESPONSE TIME of each isolation function sh 4.1 2 3 least once in Table 3.3 2-3 shall be demonstrated to be within itr limit at Each test shall include at least one logic train such that per in months.
both logic chains are tested at least once per 36 months and one channel per function such that all channela are tested at least once every N times 1R months, where N is the total number of redundant channels in a specific innistion function.
l
's BRUNSWICK - UNIT 2 3/4 3-10 RETTFED TECE. SPECS.
Uodated Thru. Amend.
l 1..
.I
.l T.A RI.E. l l. 2. - I.
eg U, j -
I S.t.H.._AT I O_ N. A.t:T.II A.Ti tlN. I NST..N.IIN.E. N..T.A. T..I.ON__
~
[*-l f
Q val.VE 4:NtNirS HININIM litM NER AFFl.lCANLE tirENATED NY til'ENAMt.E OlANIAtl.S OPENAffollAl.
- I' I.2
.TN..I.P. F.i.l p:T.itlN Aille til:ilNilHtKF MI.MNEN
- ll t:N AI (.s.)...
PE.N. T.K.it'.. SYSTEM (b)(c)
Collptfleil ACTION l
4 y
I
-.H.A.NV. t'. Hill' Allalle.8r !!. ell.ATlHN PMl ta 1.
4 m.
Near e..
U. isi. l W.es c o Icvel -
l.
Is w, in. l I 2, e, /, al '
2 1, 2, 1 20 (N?l i r Nell /A 1,N 1 C-l ll.8) 1571 e ril -Inel 14 1,N 1,C-8,15-18 a
f, i
2
- l. 2, 3 20
(?,
i I
2.
I,.w, in-.i l.'
i (M11 sll' No14A l,N l,.and
(
N2B s.F BMaj'A l,N l) s j
(
l s!
(183 0 1.ril-lHl74A -1,N-l estil' I]
05 I l.514 -Ikl2%A 1,N-l) j go li.
Ils yuuI I re eu..ise e Illat e
?, is, /
2 i, 2, 1 20 i Ti t
(8:12 rr IHi i!A.N,4:,ll) 1 I
(4:12 l'lH II.ist2 A -l, N - 8,C-8,D-1 )
h l
1.
M.sillaa I...
- til gh fil)
I 2
l, 2, 1 21 lj a:.
H.nlsi Se e.se i Iese j
l (Ill? HH-Keist lA.N.C.D) i f
2.
I's u nseis. - Inw i
2 1
22 fi
(#2 8 PT-liell%A.N.C,n) h
)
( N 2 8 - PTH-Ittel % A - l. N-l C-l.p-l )
jg 1.
Fl.sw lig gh 1
2/IIne 1
22 i ;.
j g-g-
l (ull rilT-illNHiA,N,C,ll; 182 8 rHT-litNS/A N.C.D; NJl rIIT-MINDMA,N,4:,D; l
NJI ritT-IllM19A.K.C,0) '
yl
- 4 (Ell PIITM-liiNEA 1,N-8,C-l.ll-1; f
l N2 l l*lsTM-IIINBf A - t,N-l,C-[,D-l ;
i 1
i N7I s'uTH-liiNDM A t.1L-1.C-I,D-l ;
N.'I rpTH -898M94 -I,N-t C e,n-I) 1 l
.f
e o
....,s..*.-
- ~
.,,g g,.m.,...
s...
. s.
TABLE 3.3.2-1 (Continued)
ISCLATION AC""ATION INSTEML*f!AUON ACTIONS,
le in at laaet EFT SID00ls"5 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SEUDOWN wi ACTION 10 -r the fellowing 30 hears.
Se is at least STAR 1TP with the amis staan lism isolaties valves ACTIDS 21 -
elesed withia 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> er be is at lasse EDT SEU30WW withis 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and ans COLD SIUTDOWN withis the nest 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
ACTION 22 -
Be in at least STARITF'withia 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
ACTION 13 -
Establish SECONDART CDWZA1:053T INTICEITT with the standby Saa treatamat systes operating within een hour.
Arn0N 26 -
Isolate the reactar water cleanup systes.
Close the af fected system isolation valves and declare :he affectae ACC ON *.S.-
systas inoperable.-
ACCON 26 -
7eri*y power availability es the bus at lass: :nce per *: hours.
Dese:1 vata the shutdown caeling supply and reactar vossal hand spray ACO:N 17 -
Laelaties valves is the closed positian antil the sastar stamm deem pressurs 1.*
withis the specified 1.iaita.
NC"~53
- Asa handling irradiated 'ual is the sec:naa.7 ::::aissant.
See Specification 3.6.3. '. Taele 3.6.3.1-1 !sr ralves is each valve group.
a.
A chanaal any be placed is en isoperatie status for up to 1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> f:=
required surveillar.cs #.:h=ut placing : e :::.; systes is the ::d.ppes 5
esac.1: =n ;rrit:es a: ' ass: =ce :ther :FE?.43*.I ::annel is the sama trip
~
systes is inn:.::::...;
.4: ;ars=etsr.
a... ::ars:Le :.annet teec.::
4L:.. :.1* :ne :.anr.el :e :::: sys:es.
- .2:e:... :.aa :::.::e: :: :.: :. s s:s :::s ses.: ::.s * : e 7 : h= = =:.::.
- : c:.;;.
- :.ese :sses. :.e
.:;ers:.e :.2.r.e. 5 :.. :s res::::: ::
- FEIA3*I s:2:as.rt:.an
- ur: :: ::.e AC*.*:s !ste:::: :y 7.:.e !.;... :::
- ss: 7.: %n::::n s-at; :e ta e..
e 7: ps : e ec..4.::a. ta:. :
. :.s: : a-e. :s ?!?.c
- hannel as :711A3* I ! * :f...4:;.=e-:s t.
A "Ji:h reac::e steaa ;;essurs i !"C :::;.
8 i.
01:ses anly ?.*a'O tu:.e: iso.a:::n ra.ie.
g.
M.
Alars cal 7 RETTFD TICE. SPEC 3.
"d Aus@ Cf.
UW.k 7 Updated Thru Amend.
~ ~ ^ ' " ' ' ' ' '
--vw wm-.,, _ _, _ _
T
' T, '}
L-
~s-L Am..s
(
l i
i, -
h b
i i
Q T. A. pl.E l.1. 2 - 2 IStH.ATitM ACTilATitM INSTulMENTATiull SETPullfrS Q
ll h
h o
i TNIP SETPOINT VAlEE
, h TRIP FINK:TitM ANil III:iTittull.Nr NIMMER lf H
~
I N
l.
,PRINAnY 4:alNTA,Illill.Nr I:isH.ATle,HE
)fi s.
Reactus Ves.a.. l W.al er In:ves t -
> + 162.5 laches*
a
>I 162,5 laches l.
Inw,I. cl I (N2l I TH -Neil / A - l, R-l.C-1, D-1 )
> t,112 laches*
> + 112 incines*
~
~
2.
lasw. 14.. t 7 ph *.
(N21 I fH-tHI!4 A t,N-l and
< 2 pelg l
! l 82 sis-tais !'iA -l, u - l )
b.
11rywell ra i i...ure lii gh
< 2 pelg g l
(C12 PTH u Hl2 A-1,5-8,C-8,D-l)
)
4 y
I.
u.a.II.it I..u
- sligh
~~ :n a tull power background
< 3.5 a felt power c.
Halu :it e.im i.I nu y
/
Fackground (1:12 salt-Klin lA,B,C.H)
~
> 82S pelg
> 325 pelg
'h i
I 2.
Ps.: w sus.-
lasu
( n21 e rH -tMi'e A 1,5-l,C-l,D-1 )
L
< 1401 ni rated flow
< 1401 of rated flow g
I 1.
Flow - ers :li e
(n21. p rH litNe6A-1,5-l,C-l.ll-l ;
l n21 isTH NiMal A-l,5-l,C-l,D-l ;
' l)l m28.HTH-NelGSA-1,5-1.C-l.D-l;
. 1 i
ull infil-NIHl9A-1,R-8,C-8,D-l)
< 4llt of rated flow
~< 40% of rated flow j l 4.
Flow
.si gh
~
l (u21 raiTS-n Mili A-2 ;
l n2l rnTS-Neusin 2; N2 B t hfS Neul8C-2; n21 rhfS-N4Ml4H-2) b u
i r
i e
g 1
e.
li
EMF.RCENCY CORE COOLENG SYSTEMS _
1/a.5 4ICH PRESSURE COOLANT INJECTION SYSTEM 1/4.5 1 LIMITING CONDITION FOR OPERATION 1.5 1 The High Pressure Coolant Injection (tiPCI) system shall be OPERARLE with:
and One OPERABLE high pressure coolant injection pump, a.
suction from the suporession An OPERABLE flow path capable of taking the water to the pressure vessel.
5.
pool and transferring CONDITIONS 1, 2, and 3 with reactor vessel stears does pressure APPLICARILITY:
> 113 psig.
ACTION:
With the RPCI system inoperable, POWER OPERATION any continue and LPCI systems are OPERABLE; restore a.
provided the ADS, CSS, (noperable WPCI system to OPERARLE status within la days the or he 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 SHUT 00WN vithin the..following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
w
'41:h :*e survattiance recuirementa 3f 3:e:ift:2:1m a.t.'
cerfor9e4 1: :he recuired f requencies due to '.cw rese:or 4:e2m are not
- ne orovisions of Soecification a.1.1 is performed
- pressure, aoolicable orovided the appropriate surveillance is adequate to sittin J hours af ter reactor steam pressure cerform the tests.
mRVE:LLANCE RE^UI9EME'.~S WPCI stall 9e demonstetted OPERABLE:
a!;.1 The l
l s.
At '. esse ance per U days 5y:
Verifying that the system piping from the pump discharge valve to the sv9 tem isolation valve is filled with water.
B s
t 1/4 5-1 MRUNSWICK - UNIT 2
' %wT5 SY'O]
RETYPED TECH. SPECS, Cpdated Thru. Amend. '3 I
Il
e.MRn.ENrY CORE COOLING SYSTE_MS
_.SURVEILLANCT. REOUIREMENTS (Continued) each valve ( nanual, power-operated, or automnic) 2.
Verifying that locked, sealed, or otherwise in the flow path that is not secured in position. is in its correct pos it ion.
that the system develops
.1 once per 92 days, by verifying 4250 gym for a system head corresponding to a 5
At least supplied to the flow of at leas t 1000 psig when steam is being reactor pressure >
turbine ac 1000. 7 0,
-14, peig.
2 once per 18 months by:
c.
At least which includes simulated Performing a system functional test its emergency 1.
automatic actuation of the system throughouteach automatic valve in operating sequence and verifying that position. Actual the flow path 4ccuates to its ccrrectinto the reactor vessel is excluded from injection of coolant this test, least a250 rp,
the system develops a flow of at 2.
Verifying that to a reactor pressure of > 165 f or a system head corresponding osix when steam is Vetag supplied to the turbine at 165. + 15.
osiz.
1.
7erifying that the suction for the wFC: svsten is succeatitslly tank to the suppressian transf erred f rom the condensate storage pool en a andensate storage tank low water level signal Jr suopression poo; high water level signal.
I c
3/4 $-2 Mt'NSWICK - UNIT 2
- As. T5'I 34 - o 7 RETTFE3 TICH. SPECS.
Updated Thru. haend. ~S l
l
,,-.,-------,------,.--n--
,. - - - -.. -,. -,. - - ~ - -., - - - - - - - - - - - - -, - - -, - - -, -
I EMERGENCY CORE C00LfNG SYSTEMS
\\
AUTOMATIC DEPRESSURIZATION SYSTEM 3/4.5.2 LIMITING CONDITION FOR OPERATION The Automatic Depressurization System (ADS) shall be OPERABLE with at 3.5.2 least seven OPERABLE ADS valves.
CONDITIONS 1, 2, and 3 with reactor vessel ocean dome pressure APPLICABILITY:
> 113 peig.
ACTION:
With one ADS valve inoperable, POWER OPERATION any continue provided the HPCI, CSS, and LPCI systems are OPERABLE; restore the inoperabla a.
lesse M0T ADS valve to OPERABLE status within 14 days or be in at SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, least NOT SHUTDOWN With two or more ADS valves inoperable, be in at b.
24 hours2.777778e-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 /> and in COLD SHUTDOWN within the next With the Surveillance Requirement of Specification 4.5 2.b not performed at the required interval due to low reactor steam pressure.
c.
the provisions of Specification 4.0.4 are not applicable provided the appropriate surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter reactor vassal steam pressure is adequate to perform the tests.
SURVF.ILLANCE REOUIREMEYr5 4.5.2 The ADS shall be demonstrated OPERABLE at least once per 18 months by:
Performing a system functional test which includes simulated automatic actuation of the system throughout its amargency operating a.
sequence, but excluding actual valve actuation.
Manually opening each ADS valve when the reactor steam does pressure b.
is 1100 psis and observing that either; The control valve or bypass valve position responds accc6...x".*
f 1.
t or i
is a corresponding change in the seasured steam flow.
2.
There I
s 3/4 5-3 SRUNSWICK - UNIT 2 RETTPED TECE."SPECt Updated Thru. Amend 7E f
QwERGENCY COR2 COOLING SYSTEMS 3/t. 5.1 LOW PRESSURE COOL 2NG SYSTEMS CORK SPRAY SYSTEM LIMITING CONDITION POR OPERATION 3.5.3.1 Two independent Core Spray System (CSS) subsystems shall be OPERABLE with each subsystem comorised of:
a.
One pump, and b.
An OPERABLE flow path capable of taking suction from at least one of the following OPFRABLE sources and transferring the water through the spray sparger to the reactor vessel:
1.
In CONDITION 1, 2, or 3, from the suppression pool.
2.
In CONDITION 4 or.9*:
a)
From the suppression oool, or b)
When the suppression pool is inoperable, from the condensate storsze tank :-taining at ' e sst * * *.""" rslions of water.
\\ P oLICARII.ITY : CONDITIONS 1, 2, 1, 4, and $*.
ACTION:
a.
In CONDITION l',
2, or 3:
1.
With one CSS subsystem inoperable, POWER OPERATION any continue provided both LPCI subsystems are OPERABLE; restore the inoperable CSS subsystes to OPERABLE status within 7 days or be in at least MOT SHUTDOWN within tha 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 ho,urs.
With both CSS subsystems inoperable, N in at least MOT SHUT *)CiM
?.
d :5tn '2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 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 />.
s e ser vi svates is not required to be OPERAMLE when the suooression snat 14 1,coe-sh: 4. orovided that the reactor vessel head is removed and y
w' tv '. s f'. coded. the 40ent fuel pool gates ara removed. snd the sater '.vei.s matatainec vichin the limits of Specifications 1.4.3 and 1.*2. 3 MUNSWICK
'INIT 2 3/ 4 5-4 RETTPED TECH. SPECt Updated Thru. Ase:4. 7 I
1 EMERCENCY CORE C00LIhc SYSTEMS _
LIMITING CONDITION FOR OPERATION (Continued)
ACTION (Continued) i the CSS is actuated and injects water into the 1
3.
In the event shall be prepared ard reactor coolant system, a Special Report to Specification 6.9.2 submitted to the Comission pursuant
$ll within 90 days describing the circumstances of the actuation and
]'
the total accumulated actuation cycles to date.
b.
In CONDITION 4 or 5*:
With one CSS subsystem inoperable, operation any continue one LPCI subsystem is OPERABLE within A
- 1..
provided that at least hours; otherwise, suspend all operations that have a potentis' for draining the reactor vessel.
With both CSS subsystese inoperable, operation any continue one LPCI subsystem is OPERABII. and both 2
provided that at least Otherwise, suspend LPCI subsystems are CFERABLE within a hours.
all operations that have a potential for draining the reactor ressel and verify that at least one LPCI subsystas is CFERABLE wi:hin a hours.
applicable.
De provisions of Specification 3.0.3 are not 3.
~
SURVEILLANCE REQUIW.NTS Eacn CSS subsystem shall be demonstrated OPERABLE:
4.5.3.!
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 storase tank sinimum required volume when the condensate storage tank is requiged At,least a.
- o be OP83 TABLE.
b.
At least once per 31 days by:
'tertiving : hat :Me systes pioing from :he pucio discharge val re
- o :ne systes isolation valve is fined si:h water.
l l
l souired to be OPERABLE when the suppression The core scrav systes is the reac:or ressel head is removed and
=
cool ts inoceraole provided : hat fuel pool gates are removed, and the
, he cavity is flooded, :he spentwater level is saintained within the limits of 3pecific 3.4.9.
3/ 4 5-5 BRUNSWICK - ll NIT 2 RETT?ED TECE. SPECS.
Updated Thru. Amend. 'S 1
i e
0 e-
,,n
,w---..,,
n-y,,,,,e,,
EMERCENCY CORE COOLING SYSTEMS Si;R'.T.I* :.ANCE REQUIREMEr5 (Coneinued) each valve (manual, power-operated, or automatic) 2.
Verifying : hat in the flow path that is not locked, sealed, or c:herwise secured in posi:1on, is in its correct position.
c.
At least once per 92 days
- by:
~
1.
Verifying that each CSS pump can be s arted f rom the c:c::r.
room and develops a flow of at least 4625 gym on recir:u.;::.
flow against a system. head corresponding to a reac:ce ces.e.
pressure of > 113 psig.
2.
Performing a CHANNEL CALIBRATION of :he core spray header ;?
instrumentation (E21-dPIS-N004A,5) and verifying che se:;oin: :o be 5, ;tt.5, psid greater than the normal indicated iP.
d.
At least once per 18 months by performing a system functions 1 es:
which includes simulated automatic ac:uation of the syste:
each au:::a: :
1:s emergency operacing sequence and verifying that valve in the flow path actuates to its correct posteion. Ac: 24.
injection of coolant into the reactor vessel is excluded fren -
t cest.
i I
= The survet11ance :es: :scutred by :his '.icense in Accene x A, carsgrt:n
(
flov : esc :f :ne : re sprs 4: 1:d: :ay :e
. 3.3 %4 0.., regar:ing :ne refue;ing vitsge 'Teload 3
- 1 at:nin i
pos:pened during :ne current i
18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> af ter rescorstion if the suooresston chamoer :o iperse'.e 4:a:as but in any case no later chan Novescer 15, 198*.
l 3RL'NSWICK - UNIT 2 3/a 1-6 Amendment No. 96 g
g---
conArmEn sis:Exs O...-
courAneitwr arnosrsrRE ormr10s sferax
, ;i.6 h.hi yyij,jg gg.; _.,
LIMITING CONDITION FOR OPERATION 5.M.G.W.'~CJ3%7. 1.cr.w harm. ~~- led-(M)**% W.' * ^ ~ : '^ '.t.* %fsten
- Y
"-~
' 3.
4.2 the containment' atmos e dilut with:
9 An OPERABLE flow path capable Y supplying nitrogen to the drywell, a.
and b.
A minimum supply of 4350 gallons of liquid nitrogen.
~
AFFLICABILITY:
CONDI!!ON le.
ACTION:
.../
With the CAD system inoperable, restore the CAD system to CPERAB12 status ;
within 31 days or be in at least STAR 2VF within the east 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The i
provisions of Specifitation 3.0.4 are not applicable.
si;IVE!*..;dic REQUIREMENTS 4.6.6.2 The CAD systes shall be demonstrated to be OPERABLE:
At least once per 31 days by verifying that:
h a.
'1.
The system contains a minimum of 4350 gallons of liquid nitrogen, and
~
2.
Each valve (manual, power-operated, or automatic) in the flov path not locked, sealed, or otherwise secured in position', is in l
its correct position.
b.
At least onca par 18 months by:
6 1.
Cycling each power-operated (excluding automatic) valve in the flow path through at least one complete cycle of full travel, and 2.
Verif ying that each automatic valve in the fice path actuates to si nal.
its correct position on a Group 2 and 6 isolation test
(
Wnen oxygen concentration is required to be < !,I per Specification 2.3.6.3.
3RL'NSVICK - UNI 2 3 /l* 6-25 Anendnent No. 55 4
..__n,--
,_w----_
=---:--=---
---m
l
,gg
- 3 6.6.3* The ' primary contciament cteosphere,oryges,eonscitettita shall be less ~ r.
than At by volume au'ing the period from:
- r r
' '.pgg.g.5.4.. -.
..c..TS5.. N%WEs. o.s.r+ii. :n M.
- W % 2 i.. :a..1tithis?44 hours5.092593e-4 days <br />0.0122 hours <br />7.275132e-5 weeks <br />1.6742e-5 months <br /> a WjIr-i-i.e...
G$,D..ws'.
- W.6_ta.5 Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> priot to 's scheduled reduction 'of TRERMAL POWER to,<
~-
b.
15% of RATED THERMAL POWER.4 d ?1.ICABILITY:
CONDITION 1.
ACTION:
With the oxygen concentration in the primary containment exceeding the limit.'
be in at least START-UP within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
~- ;.:;,....,.$g. -
ec. &
)
SURVEILLANCE REOUIREMENTS 1
l
- f. 6 6.3 The oxygen concentration in the primary containment shall be ver* f *v to be within the limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter THERMAL POWER > 15% of RATED TERMAL POWER and at least once per 7 days thereaf ter.
h
- For the period commencing at 0630 on June 29, 1981, a temporary exemption is allowed to operate BSEP-2 in Condition I with containment oxymen concentration exceeding At by volume for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
- 5 I
l l
l l
.=-
i r
BRUNSWICK - UNIT 2 3/4 6-29 RETYPED TECH. SPECS.
Updated Thru. Amend.
pxiT SYST.S_
3/4.7.4 FJMcTOR CORE ISOLAN 0N COOLING STSITi LIMIM NC CONDIn 0N FOR OPERAn 0N h ll be OPERABLZ with Ihe reactor core isolation cooling (RCIC) system s aof autom 3.7.4 an OPERABLE flow path capablesuppression pool and cransferr h
dome OFERAUCNAL CONDIUCNS 1, 2, and 3 with reactor steam APPLICA3ILITY:
pressure greater chan 113 psig.
ACU CN:
inue and the provisions of With the ECIC system inoperable, operation may con EFCI system is 0FRIA3L. E; h
restore the RCIC. system to 0FERABLE scacus within12 hours and reduce r hin the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
80T SHUTDOWN within the next to less chan or equal to 113 psig wit Z
SCRVEFuMC2 RZQUIREMENTS The ECIC system shall be demonscraced OPERA 3LI:
4.7.4 At least once per 31 days by:
the highpoint vents that the system a.
isolacion 7erifying by venting at piping from the pump discharge valve to the systas 1.
- - t'.11sd with water.
vel $
toestic Verifying that each valve, annual, power operated or au herwise secured in the flow path thac is not locked, sealed or ot 2.
in position, is in its correce position.
ECIC pump develops,a h
At least once per 92 days by verifying that t eflow th wich'a ting pressure when b.
system head corresponding to reactor vossal opera 20, a 80 psig.*
steam is being supplied to the turbine ac 1000 +
i t applicabis provided the reactor steam pressure is
- The provisions of Specification 4.0.4 are nosurve111sne f
.sw. so perf orm the cast.
I Assadment No. 94 3/4 7-7 3RUNSWICE - UNIT 2 t
\\
~
PLAN
- SYS D'5_
SU1LVI!1.1.ANCE RIOCIRDtEN*S (C ncinued)
)
lease once per 18 months by:
c.
At Performing a system functional c,est which includes si=ulated and verifying that each 1.
automatic actuacion and restart auccaatic valve in the flow path actuaces to its correct j
h position, but say exclude actual injection of coolanc into c e reactor. vessel.
Verifying that the system will develop a flow of greater chan or lied to equal to 4C0 gym in the cast flow path when steam is supp 2.
a pressure of 150 + 15 peig.*
che curbine at I
7erifying that the suction for the ECIC system is auconscically h
pression transferred from the condensate storage tank to c e sup 3.
pool on a condensate storage tank water level-low signal.
applicable providec c.e
~~~ he provisions of Specificacions 4.0.4 are actsur re111ance cean pressure is adequate to perform the casts.
h is subsequent to a Automatic rescart on a icw vacer level signal whic high water level trip.
l
~
i i
1 i
i I
t l
i Amendment No. 94 1/4 7-3 3RUNSWICK - UNIT 2
,_,_,_._.___.m.
_,7
ELECTRICAL POWER SYSTEMS _
L' e. A 1/4.A.1 A.C.
SOURCES OPERATION OF ONE OR MOTH UNITS LIMITING CONDITION FOR OPERATION the following A.C. electrical power sources shall be h.8.1.1 As a minimum, OPERABLE:
circuits between the off site transmission Two physically independent 1E distribution system, and s.
network and the onsite Class Four separate and independent diesel generators:
b.
1.
Each with a separace:
Engineeunted fuel tank containing a minimum of 100 a) gallons of fuel, Day fuel tank containing a minimum of 22,650 gallons of b) fuel, and c) Fuel transfer pump.
a sininum of .000 2.
With a plant fuel storage tank containing gallons of fuel.
APPLICABILITY : CONDITIONS 1, 2, and 3.
ACTION _:
or one diesel generator of the above With either one offsite circuit required A.C. electrical power soTirees inoperable. demonstrate the a.
Surveillance OPERASILITY of the rossining A.C. sources by. perf orming Requirements 4. A. I. I. l.a and 4. A. I. L.2.a.4 within two hours and at lease two offsite least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereaf ter; restore atto OPERABLE status within 72 circuits and four diesel generators hours 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 COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
and one diesel generator of the above With~one offsite circuit h.
required A.C. electrical pYver sources inoperable, demonstrate the SPERABILITY of the remaining A.C. sources by performing Surveillance d at S.ecuirements *. A.1 1.1.a and 4.R. t. t.2.a.4 within two hours an least one of the ence per 12. hours thereaf ter; restore at inoperstle f ources to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or be in at
- esst 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN least 90T 3Hi MWN within the nextRestore at least two offsica t:.'.9 wine 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
h i
tiesel generators to OPERABLE status within 72 within t e
)
sno hur least ROT SHUTDOWN circuits hours from the time of initial loss or be in at 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD 9HUTDOWN within the following within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
l 3/48-I BRUNSWICK - UNIT 2 PETTFID TEC1. SPR$.
Updated Thru. Amend.
6.
S,
\\
ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
ACTION (Continued)
~
With two of the above required offsite A.C. circuits inoperable, demonstrate the OPERABILITY of f our diesel generators by performing c.
4.8.1 1.2.a.4 within two hours and at least Surveillance Requirement once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereaf ter, unless the diesel generators are least one of the inoperable offsite already operating; restore at least HOT sources to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at
~
SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. With only one offsite source restored, restore at least two of f site circuits to OPERABLE status least 90T within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from time of initial loss or be in at SRUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
With two of the above required diesel generators inoperable, d.
demonstrate the OPERABILITY of the remaining A.C. circuits by performing Surveillance Requirements 4.8.1.1.1.a and 4.8.1.1 2.a.4 within two hours 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 /> thereaf ter; restore three diesel generators to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> l
at least or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD Restore at least & diesel SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
generators to OPERA 3*_E status within 7* hours from time df initial loss or be in at least HOT SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and in COLD SHUT 00WN within the f allowing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
^
s
' SURVEILLANCE REQUIREwENTS circuits between the offsite
..3.1.1.'L Each of the above required independent 1E distribution system shall be:
trannaission network and the onsite Class Determined OPERABLE at least once per 7 days by verifying correct a..'
breaker alignments and indicated power availability, and Demonstrated OPERABLE at least once per 18 months during shutdown by b.
manually transferring unit power supply from the normal circuit the alternate circuit.
Each diesel generator shall se demonstrated OPERABLE:
- .1 1.2 s.
At least ince oer 31 davs on a S MGO:: RED TEST BASIS by:
Verifying the fuel level in t%e engine-sounted fuel tank, 1.
3/4 8-2 BRUNSWICK - UNIT 2 RETTPED TECH. SPEr:
Updated Thru. Amend. 71 -
,m
,,,-.-,,,,.,.__y
, - - -,.. ~ _,,
,_,.---..,,,-,w,._
,,,-,__.m,,,,,-.--...,,..-,_,v-,._-.m.
ELEC*RICAL POWER SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)
Verifying the fuel level in the day fuel tank, 2.
Verifying the fuel transfer puw can ha started and transfers 3.
fuel from the day tank to the engine-ecunted tank, Verifying the diesel starts from ambient condition and 4.
accelerates to at least 514 rys in i 10 seconds,
5.
Verifying the generator is synchronized, loaded to > 1750 kw, and operates for 115 minutes, and Verifying the diesel generator is aligned to provide standby 6.
power to the associated emergency busas.
b.
At least once per 31 days by verifying the fuel level in the plant fuel storage tank.
At least once per 92 days by verifying that a sample of diesel fuel c.
from the fuel storage tank, obtained in accordance with ASTM-D270-1 of ASTM 65, is within the acceptable limits specified in Table D975-74 when checked for viscosi:y, water, and sediment.
d.
At least once per 18 months during shutdown by:
Subjecting the diesel to an inspection in accordance with 1.
procedures prepared in conjunction with its manufacturer's recosunendations for this class of standby service, 2.
Verifying the generator capability to reject a load equal to one core spray pump withcut tripping, Simulatins a loss of offsite power in conjunction with an 3.
emergency core cooling system test signal, and:
i Verifying de-energization of the emergency buses and load a) shedding from the emergency buses.
Verifying the diesel starts f' rom abient condition on the b)* auto-start signal, energizes the emergency buses with permanently connected loads, energizes the auto-connected loads through the load sequence relays, and operates for >
5 minutes while its generator is loaded with the emergency loads.
23,
- For\\he verification of this item scheduled for comoletion by Februarv a onetime-only exemption is allowed to extend this inspection until
- oni,
- bef ore the completion of the Spring 19P.1 cutage, scheduled to commence in March, 1981.
BRUNSWICK - UNIT 2 3/4 8-3 RITYPED TECR. SPECS.
Updated Thru. Amend.
Q--__--__-__
ELECTRICAL PodER SYSTEMS SURVEILLANCE REOUIRIMENTS on the amergency core cooling system test Verifying that engine cvsespeed, 4.
signal, all diesel generator trips except generator dif ferential, low lube oil pressure, reverse power, loss of field, and phase overcurrent with voltage restraint, are automatically bypassed.
Verifying the diesel generator operates for 160 minutes while 5.
loaded to 13500 lot.
the auto-connected loads to each diesel 6.
Verifying that exceed the 2000-hour rating of 3850 kw.
generator do not Verifying that the automatic load sequence relays are OPERABLE with each load sequence time within 10% of the required value.
7.*
- For the verification of this item scheduled for completion by February 23, 1981, a onetime-only exemption is allowed to extend this inspection until "before the completion of the Spring 1981 outage," scheduled to commence March, 1961.
,a T
1 1
3/4 8-4 BRUNS'JICK - UNIT 2 RETTFED TECE. SPEC 5.
Updated Thru. Amend. 7 t
+
ELECTRICAL POWER SYSTEMS SMUTDOWN OF SOTH UNITS LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimus, the following A. I. electrical power sources shall be OPERABLE:
a.
One circuit per Unit between the off site transmission network and the onsite Class 1E distribution system, and Two diesel generators, as r+ mired to operate ECCS systems in b.
accordance with Specifications 3.5.3 1 and 3 5 3.2:
1.
Each with a separate:
a) Engine-counted fuel tank containing a minimum of 100 gallons of fuel, Day fuel tank :ent 11ning a sinimum of 22,650 gallons of b) fuel, and c) Fuel transfer pu=o.
2.
With a fuel storate t:ne ::ntaining a minimum of 37,000 gallons of fuel.
APPLICABILITY: CONDITIONS 4 and 5 ACTION:
With less than the above minimum reau: red A.C. electrical power sources OPERABLE, suspend all operations involving irradiated fuel handling, CORE have the ALTERATIONS, positive reactivity changes, or operations that potential of draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.
f SURVEILLANCE REOUIREMENTS i
The above required A.C. electrical power sources shall be 4.8.1.2 demonstrated OPERABLE per surveillance requirements of Specif t :acions 4.8'.1.1.1 and 4.3.1.1. 2, an:e et for the requirement of 1.8.1.1.2.a.5 l
BRUNSWICK - UNIT 2 3/4 8-5 RETTFED TECH. SPECS.
Updated Thru. Amend. 71
- -~
WSTER CCPY M a s 7~E R 5.
TilE ORY OF UUCLEpp r o pt!.: PLAU1 DPEP M 30i', ILUlbS. Ai'D PAGL 25 ANSWERS -- BRUNS!>lCK 182
-8 6/05 /1't-L AWYER,
S.
ANSWER b.01 (1.00)
(b)
REFERENCE DK, Core parameters and thermal limits, p 1110.
BK,06-A, Table li p 31,p 45, p 9.
ANSWER 5.02 (1.00)
(c)
REFERENCE DPC, FNRE i DK,HO-02-2/3-A, fis 20.
NETRO, p 1.4-1 1/0 K5.25 (1.0/2.1) 269 8/19/05 10 02 02 02
- 6
+0.500 40.500 302 12/17/84 03 01 00 00 04
+0.500
+0.500 ANSWER 5.03 (1.00)
(c)
REFERENCE OC, OP 1106/01, p 10.
i BK,GP-03, p 19.
CR3, power system ops., pp 22-22.
DK, 20-2-C4, p 37.
i 62/0 A4.03 (2.8/2.9) l l
269 8/19/05 10 02 03 01 16
+0.375
+0.375 l
ANSWER 5 04 (1.00)
(b)
REFERENCE Steam Tables l
l I
l h
.,._m
n J
4 5.
THEORY OF NUCLEAR POWER PLAp1 DPERATION, FLUIDS, AND PAGE 26 ANSWERS -- BRUNSWICK 1&2
-86/05/19-LAWYER, S.
ANSWER 5.05 (1.30)
(d)
REFERENCE DK, Reactor theory part II, sheet 44-45.
OK, HO-02-2/3-A, pp 59 & 125.
ANSWER 5.06 (1.00) i (b)
REFERENCE BK, GE TD,HT&FF, p 4-6.
- (1.00)
ANSWER 5.07 (c)
REFERENCE BK, HO-02-2-A, Fiss 45-49.
ANSWER 5.08
(.50)
(b) l REFERENCE BK, FW heaters and drains study guide, p 1.
BK,' Heat transfer (RO), p 7-15.
DK, GE Thermodynamics, p 8-20.
l i-4 f
a
---,----,,,,,,,._-...,.-_.,-.--._,.__.___2,_.,,
5.
THEORY OF UUELEAR POWER PLANT OPERol100, FLUIDS, AND PAGE 27 ANSWERS -- BRUNSWICK 1E2-
-86/05/19-LAWYER, S.
ANSWER 5.00 (3.00) a.
DecreasesEO.25].
There is less steam flowr therefore, less pressure drop through the main steam lines [0.75].
b.
IncreasesEO.253.
With the same amount of cooling water through the condenser and less of a heat load, condensate depression will increase EO.75].
c.
DecreasesEO.25].
Less extraction steam from the turbine'to heat the feedwaterEO.753.
REFERENCE HA, Heat transfer lesson plan, pp 75 & 78.
HA, Nuclear training, p 10.4-11.
BK, GE TDrHTaFF, Chapter 6.
ANSWER 5.10 (1.50) a '. less rod withdrawal b.
no significant effect c.
more rod withdrawal REFERENCE BK, RTN-004 BK,-HO-02-2/3-A, pp 187,198 & 163.
BK, GP-02,.P 14.
ANSWER 5.11 (1.50) a) Radiolytic decomposition of water.
(0.5) b) Metal (fuel cladding) to water reaction.
(0.5) c) o::ysen - 5%
(0.5) hydrogen - 4%
REFERENCE BK, Mitigating core damase study guide, pp 30, 33 and 34.
l
c.
So THEORY OF NUCLEAR POWER PLAN 1 OPERATION, FLU 1DS, AND PAGL 20 ANSWERS -- BRUNSW1CK 1&2
-06/05/19-LAWYER, S.
AMSWER 5.12 (1.50) a.
1000 psia (+200 or -100 psi) b.
800 desF
(+or-50 desF) c.
255 degF
(+or-50 desF)
REFERENCE Mollier diagram ANSWFE 5.13 (1.00)
This ensures that the B10 adds negative reactivity LO551 at a faster rate
'thanspositive reactivity can be added due to reactor cooldown E0.53.
LA U REFERENCE BK, SD-05, p 1.
OK, HO-14-3-H, p 3.
ANSWER 5.14 (2.50) w a.
Swell dve te increased voiding (from pressure decrease).
b.
Due to increased voiding.
Scram (due to turbine triph c.
d.
- Group I isolation (on low steam pressure ( i n R U N m o d e )).
t e.
Effects of HPCI (0.4) and RCIC (0.1) injection.
REFERENCE DK, transient lesson plan, transient HXY-7.
BK, HO-05-2/3-A, p 40.; Fig 15.
ANSWER 5.15 (1.50) a.
This results in a time las between power change and heat dissipation and las time for moderatur temperature and void coefficients to effect power but this causes doppler to dominate even more the early part of transients.
(1.0) b.
5-6 seconds (0.5)
REFERENCE Generl Electric Heat Transfer and Fluid Flow pg 9-102 BK, HO-02-2/3-A, p 213, i
3 l
L 1
5
- T HEnreY- 0F UUCLEAR POWER PLAN 1 OPERATION, FLUIDS, AtID RAGE 29 ANSWERS -- DRUNSHlCK 1&2
-06/05/19-LAWYER, S.
(NE) h 0 0,
. jbc d '"~1
. Y'.
^
ANSPER 5.16
~ )h _ _ 9 3
r
(/afAlLu/ h g~
,s
- +. R P T- - en--T T Natural circulation
- b. DPVs fail to open M58V/// 4 M, M N "C~%)ljadeL a L". f SRVs control pressure at higher levels Void collapse due (to SVs shutting and scraml c.
Level oscillation from SRVs lifting or pressure oscillations j
^
REFERENCE BK, HO-05-2/3-A, p 22.
-ANSWER
'5.17 (1.50) a.
System opercting point.
b.
Curve B.
c.
Right.
REFERENCE HA, Thermodynamics lesson plan, pp 80 & 89.
DK,-GE TD,HT&FF, p 7-115.
ANSWER.
5.18 (2.50) s.
The delayed, neutron fraction is the percentage of fission neutrons that are born delayed.
(1.0) b.
U-238 (0,5)
'c.
Decrease (0.25) As Pu-239 production increases (0.25), and U-235 decreases (0.25) the core average will decrease due to Pu-239's Beta being so much smaller (0.25).
(1.0).
REFERENCE l
NUS Reactor Theory section 11.3 BK, HO-02-2/3-A, pp 132&137.
i l
l n
l l
(
6.
PLANT SYSTEMS DESIGN, CON 1ROL, AND INSTRUhENTA110N PAGL 36 ANSUERS -- CPUNG!11CI' l e.2
-86/05/19-LAWYERr S.
ANSWER 6.01 (1.00)
(e)
REFERENCE BK, HO-25-2/3-E' PP15,16.
ANSWER 6.02 (1.00)
(c)
REFERENCE Dl', GP-03, Rev.
4, p 15.
BK, HO-19-2/3-B, p 16.
ANSWER 6'.00 (1.00)
(c)
REFERENCE DK, GP-03, p 18.
ANSWER 6.04 (1.00)
(b)
REFERENCE BK, RTN 022, SD-19 BKr HO-14-2/3-B, pp 4r5.
ANSWER 6.05
(
.S
)
(a)
REFERENCE l
BKr HO-25-2/3-Dr p 21.
l I
l l
k
l
~,.
,6.
PLANT SYSTEhS DESIGNr CON 1ROL, AND INSTRUMENIA110N pf4GE 31
' ANSWERS --'BRUNSW1CK IE2
-06/05/19-LAWYER, S.
ANSWER 6.06-(1.50) a.
Depressing'both timer. reset PDs, (015)
( 0; 5)
'_d
,Se, curing all LP ECCS pumpsf 1:. f /
r-
,i.-
r
'b.. -50'psis 50 q % u'[q lyV
],, _, j.,,,,
(0.5)
REFERENCE l
.DKr RTN 013 BK, ll0-14-2/3-F, p 20.
ANSWER 6.07 (2.00) a.
Unit 1 o n l y e t-j.44 h.--
/"*b b.
10 +or-10 c.
100 +or-100 di 15 +or-3%
REFERENCE BKr GP-03, p 16.
ANSWER 6.08 (1.50) a.
The "high-inter-low' switch determines the pressure band within which that compressor will load and unload (0.5).
All three compressors normally operate in the " low" pressure band (0.5).
b.
When instrument air pressure decreases to 95 psis (0.5).
REFERENCE DKr SD-46, pp 2-6.
DKr HD-21-2/3-Ar pp 19,20.
ANSWER.
6.09 (2.00) a.
RiiR pump seal Hxs ArB,C and D.
Core spray pump room coolers AED.
b.
Isolation valve mini bypass lines allow trickle flow through the system.
(0.5) g._
~-
=
6.
PLAN 1 SYSTEMS DESIGN, CONTROLr AND INSTRUMEN1ATION PAGE 32 ANSWERS -- BRUNSWICK 1&2
-06/OS/19-LAHYER, S.
REFERENCE i
BKr HO-22-2/3-DEC, p 8.
ANSNER 6.10 (1.00) 1.
Mode' switch in ' refuel'.
4 2.
All rods at 00 (or full in overtravel)
)
(3. No control rod selected for motion -select matrix power off).
REFERENCE i
DM, HO-27-2/3-A, p 17.
4 ANSWER 6.11 (2.00)
(C ')
(.25)
3.
Detector not full in (.25) uith SRM reading less than 100 cps,tand IRMs on renge 1 or 2.(C,125) 6 /2D 4.
SRM (.25) inoperative (.25)
REFERENCE BKr-HO-25-2/3-A, p 28.
t ANSWER 6.12 (1.00) 1C-1E i
1D-2E l
2A-3E 10-4E i
REFERENCE BK, HO-14-2/3-Dr p 14.
I h
i
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1 Y
1
6.
PLANT SYSTEMS DEEIGNr CONTROL, Ar!D INSTRUMENT Al109 P A f, E T1 ANSWERS -- DRUNLWICK 1&2
-06/05/19-LAWYER, S.
ANSWER 6 13 (3.75) a.
Causes reactor level to increase (0.25) due to the leve3 control systen, having a level error, level set S indicated level (0.5) resulting in'an increase in the speed of the reactor feed pump turbines (0.5),
b.
Reactor level should remain constant (0.25) because the turbine speed controller will lock (0.5) the reactor feed pump speed at the speed level demanded at the instant prior to the signal loss (0.5).
c.
Causes reactor level to decrease (0.25) due to the level control system having a steam flow / feed flow error, steam flow < feed flow (0.5) resulting in a decrease in the speed of the reactor feed pump turbines (0.5).
REFERENCE DSEP CD-32-2, p3 D i' 17-2/3-D, p 21.
ANSWER 6.14 (1.50) a.
No. (0.5) Load rejection is caused by a 40% or greater mismatch between crossover pressure ano stator amps (0.5).
b.
When the OCBs open the turbine is virtually unloaded and begins to overspeed.
Tne speed and acceleration unit senses the overspeed and produces a speed error signal.
(0.5)
REFERENCE BK, HO-19-2/3-B, p 31.
ANSWER 6.15 (2.50)
The piping in question is that between the reactor vessel penetration and the chvoud (0.5).
Pressure is sensed above the core plate (.25) and at the sptay sparger (.25).
The indication is normally a negative dp (0.5).
When s bresk occurs, the dp will go positive (0.5).
If a break occurred in the monitored area, full spray flow would not be available for core cooling (0.5).
REFERENCE DK, HO-14-2/3-E, p 9.
i
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6.
PLAPI SYSlEMS DESIGN. CONI POL, AND INSTRUtiENIAll0N PAGE 34 ANSWERS -- BRUNSillCK 1E2
-8 6 / 0 5 /19-L AL'Y E R -
S.
( c'7 n el c W ( h 2~' ?l n ~ & t )
e e
ANSWER 6.16 (2.00) fy - ~ - - - _ ____
RSCCuse[the " full in' or " full out' When rod density ?>50% (0.5) the reedswitch (0.5).
When in notch control 1(0.5) the position indication is atsumed by a circuit counting pulses (settle) generated by the rod movement control switch (0.5).
REFERENCE BK, HO-27-2/3-C, pp S&6.
ANSWER 6.17 (2.00) a.
High pressure turbine exhaust pressure (0.33) compared to stator amps (0.33) indicating a misnistch of 40% (0.33) actuates the load reject.
b.
The intercept valves should shut limiting turbine speed.
(0.5) c.
Automatically.
(0.5)
REFERENCE BK, 110 2 / 3 -B, pp 31,32.
i I-
7.
PDOCEDURES - NOPMAL, ADNORMAL, EMERCENCY AND PAGE 35
-~~~~~~-~~~--~~~-~~-~~~~
~~~~ E5iUE55iCEL 5551R5t R
ANSWERS -- DRUNSWlCK 182
-86/05/19-LAWYER, S.
ANSWER 7.01 (1.00)
~
(d)
REFERENCE BK, AOP-22.0, p 1.
ANSWER 7.02 (1 00) a.
1 b.
3 REFERENCE BK, E&RC manual, Sect.
6.5.3.5, p 44.
ANSWER 7.03 (3.00) a.
1.
obtain shutdown panel keys 2.
manually scram the reactor 3.
trip the main turbine 4.
verify or transfer power to SAT.
5.
(when steam flow is < 3.0E6 lb/hr),-place mode switch in S/D.
6.-trip both recire pumps 7.
reduce reactor pressure to 700 psis with bypass jack 8.
close MSIV (at 700 psis)
P ace booster pumps in manual 9.
l
- 10. e n t e r E D P-1-0 cis,-4 m t'- o ~ ~/ t-il' ' S " ' fi i^
(any 8 9 0.25 ea.)
b.
- 1. so to the Eable spreading room (0.5) 2.
open RPS MG output breakers (0.25)
- 3. place RPS alternate feed suitch (LG3) to the mid position (0.25)
REFERENCE DK, AOP-32, rev 7, p5 ANSHER 7.04 (2.00) 1.
close(.25) and lock (0.25) minimum flow valve (.25)
(0.75)
- 2. close (.25) pump suctions from suppression pool (.25)
(0.50)
- 3. open(.25) and lock (0.25) CST suction valve (.25)
(0.75)
]
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PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 36
-~~~~~~~~~-----~~~~~~~~~
~~~~RE5i5t55iEEL E5ETR5t ANSWERS -- DRUNSWICK 1&2
-86/05/19-LAWYER, S.
REFERENCE
-BK, Op-10' P 16.
ANSWER 7.05 (3.50) a.
1.
Reactor building isolation ( on hi-hi radiation in building ventilation)(0.5) and initiation of SBGT (0.5).
(1.0) 2.
alert (0.5) b.
Indication of criticality Loss of communication between control room and refueling floor Malfunctioning of more than one fuel loading SRM or IRM Accidental dropping or damaging of a fuel element (0.5 each)
(2.0)
REFERENCE DK, ADP-7.0, p 3.
BK, FH-11, pp 4,5.
ANGRER 7.06 (2.00) 1.
All control rods must be fully inserted.
2.
Reactor pressure must be below the SDC interlocks.
3.
RHR SDC cannot be established.
4.
All other methods of SDC (have been attempted and) were unsuccessful.
5.
Further cooldown of the reactor is required.
(4 0 0.5 each)
REFERENCE DK, AOP-15.0, p 3.
ANGWCR 7.07 (2.50) 1.
The remote shutdown panel operator (RD, 20El, vicinity MCC-1XD) 2.
Shift operating supervisor or shift foreman (RB 20' Ele vicinity MCC-1XB)
- 3. MCC-1XDr 1XDB operator (RD, 20' El) 4.
MCC-1XA, 1XDA operator (RB, 20' El)
REFERENCE BK, AOP-32, pp 35,36.
c
7.
PROCEDURES - NORMAL, AUNORMAL, EMERGENCY AND PAGE 37
-~~----~~~-~~~~~~~~~~~~~
~~~~ A515t5EiEELE5nTR5t P
ANSWERS -- BRUNSWICK 1&2
-86/05/19-LAWYER, S.
ANSWER 7.08 (1.00)
EPM 41 - ATWS EPM #2 - successful SCRAM (has occurred)
REFERENCE l
BK, E0P-01/UG, Para 7.0, p 25.
ANSWER 7.09 (3.00) a.
1.
the four rod display indication will go out.
2.
The ' rod overtravel' annunciator will lock in.
b.
Place the " emergency rod in notch override" switch in the " emergency rod in' position (to stop the, double notch).
c.
To avoid PCIOMR violations due to the possibility of double notching REFERENCE DM, OP-07, pp 10,11,7.
ANSWER 7.10 (2.50)
[
a.
The SDGT train will not start automatically under this condition, putting the plant in an LCO.
(1.0) t b.
Should cause isolation of the reactor building ventilation system.(0.5) c.
Fire in the filters (0.5).
Manual initiation of SBGT train deluge valve (0.5).
(1.0)
REFERENCE CK, OP-10, pp 6,9,23.
i ANSWER 7.11 (2.00) a.
Condenser vacuum can be lost via radweste.
b.
Flow induced JP riser vibration.
REFERENCE BK, OP-14, p 10.
BH, OP-2, p 18.
l t
I 1
7.
PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 30
~~---------
~~~~EA5i5t55iEAE c5ETR5t ANSWERS -- BRUNSWICK 1E2
-86/OS/1?-LAWYER.
S.
ANSWER 7.12 (2.50) a.
cooling water pressure or flow scram valve leakage directional control */ 21ve leakage or failure (1.5) b.
1.
No 2.
to maintain cooling water flow to the CRD (1.0)
REFERENCE BK, APP A-5,3-2, p 2.
ANSWER 7.13 (2.00)
These instruments have a cold reference les(0.8)and are uncompensated (0.2).
They are not cffected until D/W tensper etur e exceeds the RPV scturction temperature (1.0).
REFERENCE DK, E0P-01/UG, caution 46, p 30.
L
D._
- ADMINIST R A'1VE PROCEDURES, - CONDIT 10t>S, AND L] h11 AT I0t!D PAGE 39 ANSWERS
- DRUNSWlCK iP2
-06/OS/19-LAWYER, S.
ANSWER 8.01 (1.00)-
(d)
REFERENCE BK, PEP 02.2, p 2.
ANSWER 8.02 (1.00)
(a)
REFERENCE BK, RCI-06.5, pp 5-8.
ANSWER 8.03 (1.00)
(d)
-REFERENCE EIH:
U2 TS, 3.3.1, 3.3.2 BSEP: U2 TS, 3.3.1, 3.3.2 ANSWER 8.04 (1.00)
(b)
REFERENCE BK, 0I-10, Sect D.1, p 4.
ANSWER 8.05 (1.00)
(d)
REFERENCE BSEP: TS 3.0.3, 3 3.7,
& 3.7.4 L
8.
ADMINISTRA11VE PROCEDURES, COND1110NS, AND LIM 11A110NS PAGE 40 ANSWERS -- DRUNSWICK 1&2
-86/05/1Y-LAWYER, S.
ANSWER 8.06 (1.00)
(b)
REFERENCE BK, 01-05, p 3.
ANSWER 8.07 (1.00) er 4% g)
REFERENCE BK, TSs 3.6.6.2, 3.6.6.3, 3.0.4.
c<a e r * / '/ i-. o ANSWER 8.08
(.50)
False REFERENCE BK, AI-59, p 6.
ANSWER 8.09 (1.00) a.
1000 mrem /hr.
b.
in a manner which alerts the individual to the radiation level within the ares.
REFERENCE AI-40; _TS, p 6-31 CK, ERRC, vol VIII, p 36.
ANSWER 8.10
(.50) 5 Rem REFERENCE BK, ERRC, p 15.
A
e
,0.
ADMINISTRATIVE PROCEDURES, CONDIT10PDr AND LIMITA110NS PAGE 41 ANSWERS -- CRUNSWICK 1E2
-06/05/19-LAWYER, S.
ANSWER 8.11 (2.50) a.
E9ent evaluation check sheet.
(0.5)
- b. Jutstanding LCOs or event evaluation check sheets ESF status board Periodic testing in progress Calibrations in progress Preventive maintenance in progress (4 of 5 0 0.25ea)
(1.0) c.
Noting it in the LCO action item section on the DSR cover sheet.
(1.0)
REFERENCE BK, DI-04, pp 6,4,2.
ANSWER 8.12 (1.50) 1.
The initials of the person who removed the recorder from service.
2.
The date and tinie of removal from service.
3.
The reason for the recorder's removal from service.
REFERENCE BK, AP-Vol.
1, p 4-5.
ANSWER 8.13 (2.00)
- a. Increases the blowdown area (0.5)
Eliminates the capability of reflooding the core (0.5) b.
50 degF (0.5).
To prevent undue stress on the vessel no::les and bottom head region (0.5).
REFERENCE BK, Tech Specs bases, p B 3/4 4-1.
i ANSWER 8.14 (2.00) 14 (0.5)8b
/
Allowed for up to two hours for surveillance testing (.5)
- a. Operable b.
Inop LPCI loop B (0.50). RHR-LPCI pumps B&D will not auto start on ECCS initiation signal when transferred to remote SD panel (0.50).
go r,. M M hh M. N M 4 t
we 8.
ADMINISTRATIVE PROCEDUPES, CONDITIONS-AND LIM 11 A110N5 PAGE 4?
ANSWERS -. BRUNSWICK 18.2
-06/05/19-LAWYER, S.
REFERENCE Drunswick RT.M-031, Brunswick RTN Oues 31-5 BK, ADP-32, P4.
ANSWER 8.15 (3.00) a.
Departure from established procedures may be made at the discretion of the Shift Operating Supervisor or the Shift Foreman if the Shift Supervisor is unavailable. (0.75)
Such departures shall be made only in situations where the safety of persons, the reactor, or other equipment is in immediate jeopardy (0.75) 6.
Two members of Plant management staff knowledgeable in the area effected by the procedure, one of which holds a SRO license on the unit affected and is a supervisor in chcrge of the shift.
(0.75) c.
A departure may change the intent of the procedure (while a deviation n.ay not change the intent).
(0.75)
REFERENCE DM, AP-5.3, pp 5-6,7.
ANSWER 8 16 (3.00) a.
1.
Station 2.
Radwaste 3.
Local (0.50 each) b.
A clearance issued to two or more individuals which has the same boundary for work (that may or may not be related).
(0.5) c.
They must have been properly trained in the use of the clearance procedure. (0.5)
By the person displaying their clearance training card. (0.5)
I REFERENCE DK, AI-50, p 2.
ANSWER 8.17 (1.00)
He must enter the reason for non performance on the Comp etion/ Exception l
form (0.5) and tentatively reschedule the PT in red on the rescheduling PT sheet (0 5).
REFERENCE DK, 01-03, p 1.
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- =- -
j3-8.
ADM]NISTRATIVE PROCEDUREF. CONDIT30NS, AND LIM 11A110NS PAGE 43 ANSWERS ~-- BRUNSWICK 1&2
-86/05/19-LAPYER, S.
ANSWER 8.18
(.50)
Thermal expansion (of valve internals on heat up) may cause valve-binding and/or damage.
REFERENCE
'_B K, 0I-13,.p 1.
ANSWER 8.19 (1.00) a.
0.38 (%)
b.
The difference between the calculated value of the maximum core reactivity during the operating cycle (0.25) and calculated DOC reactivity. (0.25)
REFERENCE DSEP! TS, Section B-3/4-1-1 BK, TS, B 3/4 1-1.
ANSWER 8.20 (1.00) 1.
Any room, enclosure, or operating area in which airborne radioactive materials (0.33) composed wholly or partly of licensed materiale(0.33) exist in concentrations in excess of the amounts specified in Appendix B tableI, column 1 of this part. (0.33) or,
- 2. Any room, enclosure, or operating area in which airborne radioact-ive material (0.33) composed wholly or partly of licensed material, (0.33) exists in concentrations which, averaged over the number of hours in any week during which individuals are in the area, exceed 25 percent of the amounts specified in Appendix B, Table I, Column 1 of this part.
REFERENCE 10CFR20.203 DK, EERC, p 38.
i
t.NCLOSURE 3 we
./
BRUNSWICK PLANT 5/19/86 REACTOR OPERATOR EXAMINATION COMMENTS COMMENTS QUESTION NUMBER See attached sheet for accepted power; 6 or 7%
41.14 power should be acceptable.
k.01A Add to possible answers; Keylock Switch to inhibit ADS.
(See attached sheet)
See attached sheet for ADS minimum pressure 50#
2.01B or as little as 25#.
2.12-The exact core location is not a memory item accept
" center of core tip tube", "tip calibration tube",
or position 10 on the tip machine.
3.07/3.08 1.
There is no such Annunciator at BSEP 2.
Question does not specifically ask for the answer given in answer sheet.
3.
Recommend deletion of question.
4.01 1.
SRO Level question 2.
Pg 1 of AI-59 gives 4 items (General)
Pg 2 of AI 59 gives 4 more items (Specific) accept either 4 or a combo of both?
(See attached sheets) 4.04 The Words pocket dosimeter and chipper in the answer kby should not be used as " Key" words.
Brunswick has the RPT Circuitry installed but 5.16 it is not used and has been in bypass for several The graph of the transient was not labeled years.
as to the type of transient, with no other data available it's tough to recognize the problem.
This mod has been installed on U-2, and will be 6.07 active during the start-up of U-2.
Seme candi-dates have been made aware of the dual plant mod and will answer both units. (See Attached Sheets) l~
-6.02 -- - - --
-For l'C" to be _the correct answer the "IV" needs to be removed.
GP-03 and SSM 19-2/3 B both state l
"All CV's and main stop valves cloce".
No mention of IV's intercept stop valves is made in either j
reference. (See Attached Sheets)
I Answer must include " Key Lock" inhibit / auto switch.
6.06 Answer may also be " Cross-Around" Pipe.
Different 6.17 tcrminology--same pipe.
i l
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.D d
QUESTION NUMBER COMMENT!;
7.01 A0P-22 does not reflect any of these answers.
(See Attached Sheet) 7.03 E0P-01 instead of E0P-10, and execute es many items as possible.
7.04 Question asks for Me.norization of an infrequent operation.
7.09 B Concern is for double notch withdrawal.
7.12 T/S Infers Rod is INOP, (3/4 1-3),
APP Infers that it's not INOP.
8.07 Tech Specs 3.6.6.2 pg 3/4 6-28 allows 32 days 3-0-4 is not applicable.
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_ - _ _ _ _ _ _ _ _ _,,, _ -. _, _,.,., _, _ _ _... _ _