ML20128M443
| ML20128M443 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 06/11/1985 |
| From: | Munro J, Wilson B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20128M422 | List: |
| References | |
| 50-325-OL-85-01, 50-325-OL-85-1, NUDOCS 8507110730 | |
| Download: ML20128M443 (85) | |
Text
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ENCLOSURE 1 EXAMINATION REPORT 325/0L-85-01 Facility Licensee: Carolina Power and Light Company 411 Fayetteville Street Raleigh, NC 27602 Facility Name: Brunswick Steam Electric Plant Facility Docket Nos.: 50-324 and 50-325 Written, oral and simulator examinations were administered at Brunswick Steam Electric Plant near Southport, North Carolina.
Chief Examiner:
[ M,u, 7 6,[// !K J6hn F. Munft-Date Signed
~
Approved by:
dAAcc Rf9C Bryce A. Wilson, Section Chief Date/ Signed Summary:
Examinations on May 21 - 22, 1985 One complete examination (oral, simulator, and written) was administered to an Instructor Certification candidate who passed.
Two SR0 candidates were adminis-tered written reexaminations; both candidates passed.
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pg
2 REPORT DETAILS 1.
Facility Employees Contacted:
- P. Hopkins, Director - Site Training
- S. Morgan, Senior Specialist - Operator Training D. Shaw, Senior Specialist - Operator Training M. Magill, Reactor Operator - Operations L. Dunlap, Contractor (RTS) Instructor
- Attended Exit Meeting 2.
Examiners:
- J. Munro, NRC
- Chief Examiner 3.
Examination Review Meeting At the conclusion of the written examinations, the examiners met with S. Morgan, and D. Shaw to review the written examination and answer key. The following comment was made by the facility reviewers:
a.
SR0 Exam 1.
Question 5.14(d) - The answer key states " Remain the Same." Fuel centerline temperature will increase slightly due to a loss of (some) feedwater heating (extraction steam loss).
NRC Resolution:
The Brunswick Steam Electric Plant Simulator confirms the accuracy of the above statement during a simulation of question conditions.
The answer key has been changed accordingly.
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.
There were no generic weaknesses (greater than 75 percent of candidates giving incorrect answers to one examination topic) noted during the oral examination.
The cooperation given to the examiner and the effort to ensure an atmosphere in the control room conducive to oral examinations was also noted and appre-l ciated.
l The licensee did not identify as proprietary any of the material provided l
to or reviewed by the examiners.
T A8L E ES-201-7 POWER PLAMI EXAMINATION RESULTS SimMARY SHEET a
ENCLOSURE 2 Total Passed Failed Acitlif Brunswick 1 & 2 Owrali nesults no.
no.
2 no.
2 DAIL or wRITIEN May 21, 1985 Smir Wat" 2
2 100 0
0
- h. Nm wYu can instui k
nNRAn N DATE Of ORALS May 22, 1985
~~ -
- 5. SEM!m IUCH (ITPAM ETM DAIL OF SIMUL ATOR May_22 1985 in=tnrwr artirou.
1 I
1 100 0
0 ExAnintR$
J. Munro IYPE Or EIAM OCMIX) (HOT)
Examination Esaminer's Results Initials RU SPO TYPE NAf1E DOCKE1 NO. CODE I
2 3
4 TOTAL 5
6 7
a TOTAL W
0 5
W D
5
/ / P/ P/ P/
/3yfp /7tia Sosler, L.
55-20226 6
84.6 86.8 92.5 75.9 85.04
/
\\
Reinsburrow, J.
55-8280 5
93.8 86.6 95.3 83.8 89.8
[
Poulk, R.
55-6661 5
86.2 78.5 84.3 81.3 82.5
////7/
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ENCLOSURE 3 U.S. NUCLEAR R[GULA10RY CDMK15510N SENIOR RE AC10R OPERATOR LIC[NS[ [XAMINA110N Facility:
Brunswick Reactor Type: BiF.
Date Administered; 5/Z1/85 taariner.
Munro.
J..
Brockman. K Candidate:
1N51RUC110N510 CANDIDATE:
Use separate paper for the answers.
Write answers on one side only.
5taple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses af ter 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 1 of Candidate's Category value Total Score Value Category
_ 26 24.4 5.
Theory of Nuclear Power Plant Operation.
Fluids, and Thermo-dynamics 28 26.3 6.
Plant Systems Design, Control, and Instrumentation 26.5 24.9 7.
Procedures - Normal, Abnormal. Emergency, and Radiological Control 26 24.4 8.
Administrative Pro-cedures, Conditions,
~
and Limitations 106.5 Totals Final Grade All work done on this examination is my own, I have neither given nor received aid.
Candidate's signature
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l 5.
Theory of Nuclear Power Plant Operation, Fluids, and Thermodynamics 5.1 The fission process in a commercial reactor requires the neutrons that are " born" by fission to be " therma-lized." The interaction in the reactor core which is most efficient in thermalizing neutrons for fission occurs with the... (CHOOSE ONE)
(1.0) a.
OXYGEN atoms in the water molecules b.
BORON atoms in the control rods c.
ZIRCONIUM atoWs in the fuel cladding d.
HYDROGEN atoms in the water molecules 5.2 Which of the following statements best describes the operating characteristics of an LPRM detector?
NOTE: Consider detector operation only.
(1.0) a.
Depletion of the detector's Uranium coating causes both the neutron and the gamma sensitivity to DECREASE with detector age; the resulting neutron to gamma signal ratio remains relatively CONSTANT.
b.
Since the detector functions as an ionization chamber and the Argon gas pressure remains relatively CONSTANT, BOTH the neutron and the gamma sensitivity, as well as the neutron to gamma signal ratio, remain relatively CONTANT as the detector ages.
c.
Depletion of the detector's Uranium coating causes neutron sensitivity to DECREASE, but has an INSIGNIFI-CANT effect on gamma sensitivity; this results in a neutron to gamma signal ratio DECREASE as the detector ages.
l d.
Depletion of the detector's Uranium coating has an insignificant effect on neutron sensitivity, but causes gamma sensitivity to DECREASE; this results in a neutron to gamma signal ratio INCREASE as the detector ages.
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a 2
5.3 As part of the scram procedure, the operator is directed to insert the SRMs and IRMs.
a.
Following a severe LOCA, briefly EXPLAIN how these systems could be used to detect gross core damage (deformation)?
(1.0) b.
Briefly EXPLAIN how these systems could be used to provide a crude indication of water level if level could not be confirmed by normal instrumentation.
(1.0) 5.4 With respect to the reactivity worth of control rods:
A Control Rod will experience its greatest TOTAL worth when... (CHOOSE ONE)
(1.0) a.
...it is INSERTED individually and all other rods are WITHDRAWN.
b.
...it is INSERTED individually and all other rods are INSERTED.
c.
...it is WITHDRAWN individually and all other rods are INSERTED.
d.
...it is WITHDRAWN individually and all other rods are WITHDRAWN.
5.5 Concerning the behavior of Samarium-149, which one of the following statements is TRUE7 (1.0) a.
Once equilibrium Samarium is established, Samarium reactivity does not change regardless of power level changes.
b.
50% equilib-fum Samarium reactivity is equal to 100%
equilibrium Samarium reactivity, c.
Samarium is only removed by radioactive decay.
d.
Samarium is produced by the decay of Iodine.
5.6 a.
After a reactor Scram from power, the shortest STABLE period possible is - 80 seconds.
Briefly EXPLAIN why.
(1.0) b.
Is the INITIAL period IMMEDIATELY following the above described Scram SHORTER or LONGER than the -
80 seconds. Briefly EXPLAIN.
(1.0)
3 5.7 Attached Figure 152 shows a basic closed loop fluid system with its head vs. flow plot (BOLD LINES).
The two pumps are iden'tifical, variable speed, radial, centrifugal pumps.
Pump 1 is initially operating at one-half speed to supply flow to component 1, as shown.
a.
Component 2 is placed into service, thereby increasing the system heat load. Would total power consumption be less by...(CHOOSE ONE)
(1) Doubling the speed of Pump 1 (2) Starting Pump 2 at one-half speed (0.5) b.
Which Pump Curve - A, B, C, or D - most accurately shows BOTH PUMPS (Both Running at one-half speed) operating to supply the system flow?
(0.5) c.
With only Pump 1 operating.at one-half speed and Component 1 in Service - If component 2 were throttled open from its initial position, would the system flow INCREASE, DECREASE, or REMAIN THE SAME7 (COMPONENTS 1 & 2 ARE IDENTICAL!)
(0.5) d.
Given one operating pump, that is changed to a POSITIVE DISPLACEMENT pump.
Is the correct Pump Curve to reflect this Curve A, B, C, or D7 (0.5) 5.8 Which of the following describes the changes to the steam that occurs between the inlet and the outlet of a REAL turbine?
(1.0) a.
Enthalpy DECREASES, Entropy DECREASES, Quality DECREASES b.
Enthalpy INCREASES, Entropy INCREASES, Quality INCREASES c.
Enthalpy CONSTANT, Entropy DECREASES, Quality DECREASES d.
Enthalpy DECREASES, Entropy INCREASES, Quality DECREASES
s 4
5.9 G1 von the following Power History:
m.
w...
%s ss' 1 er.)
(
s.
2.m s'. s. w i;..:..L.4,,',,, c
..... M
..c n....)
Select the most accurate curve displaying the expected XENON transient.
(1.0) u s.
w v
i y
y y
V U
v 5
5 5
Y I
T E
e.. sc w. se se. n w s %.
. *i. i.,. s.
t m e o.... a tg...
s.. %.
b.
C***
v v
v y
y v
y v
v i
y y
y y
7
$0 9
% 0 %.
T&
k.
%Q 0, g I $g %g 4gg %q Ig g
g T %MI (how.%)
e Eg.
Ice a
C.
6 c.
I y
v 3
g I
y v
v y
g b N N N bb TO N N
.g 9 g %gg 4$g 9q
- g T %M C ( bi.w e bI Eg.v.
1sm.
d.
- c. t.
y v
v y
i w
3 g
3 y
y V
V V
5 9
kh kh S.
N $b
.g
'I g kg hg %g 'kg 0
h 0.
k.
hb 5
Ti es t
(%.,.
)
5 5.10 The BWR is designed tcf operate like the RANKINE VAPOR CYCLE, shown below.
Select the statement which is NOT TRUE, as applied to the REAL BWR cycle.
(1.0)
C 1
,....r.
.r.
L Entropw a.
Increasing condenser vacuum (25" changed to 29") INCREASES cycle efficiency, b.
Increasing Condensate Depression, which is required for proper plant equipment performance, INCREASES overall thermo-dynamic efficiency.
c.
Feedwater Heating INCREASES overall thermodynamic efficiency, d.
Feedwater pump pressure increase causes the feedwater to be FURTHER from saturation conditions.
5.11 A steam condenser must remove more heat energy to condense...
(CHOOSE ONE)
(1.0) a.
...five (5) pounds of steam at 15 psia b.
... fifteen (15) pounds of steam at 1000 psia c.
... twenty-five (25) pounds of steam at 2000 psia d.
... fifty (50) pounds of steam at 3000 psia 5.12 STATE for which condition the reactivity coefficient contribution would be MORE NEGATIVE.
Briefly EXPLAIN your choice.
(1.0)
Moderator temperature coefficient for a 75% CONTROL ROD DENSITY.
-OR-Moderator temperature coefficient for a 25% CONTROL R0D DENSITY.
1
6 5.13 Concerning General Electr m's Preconditioning Interim Operating Management Recommendattor.9 (PCIOMR):
a.
Starting with the
- 41 at a threshold of 11.0 kw/ft, a maximum rainp increase is begun at time 0000 and the final desired power of 13.0 kw/ft is achieved at 2000.
At this time, the required soak is performed FOR 10 MINUTES, at which time the load dispatcher directs a power reduction that takes nodal power down to 12.5 kw/ft.
SELECT the valid preconditioned value for this node.
(1.0)
ASSUME THE MAflMUM RAMP RATE IS.10 Kw/ft/hr 1) 11.0 kw/ft 2) 11.8 kw/ft 3) 12.5 kw/ft 4) 13.0 kw/ft b.
After seven hours, the Load Dispatcher directs a return to full power.
SELECT the minimum time required to get back to 13.0 kw/ft, given the above ramp rate.
(1.0) 1)
Immediate (Raise to 13.0 kw/ft, w/o restrictions) 2)
5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 3) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 4) 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> 5.14 STATE how fuel pin centerline temperature will change (INCREASE, DECREASE, or REMAIN THE SAME) with each of the following conditions.
a.
A Feedwater Heater is removed from service causing a reduction in feedwater temperature of 10 deg F.
(0.5) b.
The Pressure Set on EHC is raised by 10 psig.
(0.5) c.
A fuel pellet experiences " swell".
(Gap with clad decreases)
(0.5) d.
A HPCI full flow surveillance is conducted.
(0.5) 5.15 Which of the following valves would cause the greatest energy loss in the fluid going through it (Highest dP)?
(1.0) a.
FULLY OPEN Globe Valve, b.
FULLY OPEN Gate Valve.
c.
FULLY OPEN Ball Valve.
d.
FULLY OPEN Swing-type Check Valve.
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7 5.16 Attached Figure 215 represents a transient that could occur at a BWR.
Given:
(1) One Jet Pump Riser fails at time T=1.2 min.
(2) No operat'or actions occur (3) Recorder Speed = 1 division = 1 minute Briefly EXPLAIN the cause(s) of the following recorder indications:
(2.5)
Note: There may be more than 1 cause for each answer, a.
Level INCREASE (Point A) b.
Reactor Power DECREASE (Point B) c.
Total Steam Flow DECREASE (Point C) d.
Feedwater flow DECREASE (Point D) e.
Reactor Pressure DECREASE (Point E) 5.17 Attached Figure 216 represents a transient that could occur at a BWR.
Given:
(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 Briefly EXPLAIN the cause(s) of the following recorder indications:
(2.5)
Note: 75ere may be more than 1 cause for each answer.
a.
Level INCREASE (Point A) b.
Reactor Power DECREASE (Point B) c.
Reactor Pcwer DECREASE (Point C) d.
Steam Flow DECREASE (Point D) e.
Pressure FLUCTUATION (Point E at times = 3 to 6 minutes) 5.18 A " Periodic NSS Core Performance Log" (P-1) is attached (Fig. 217) for reference. Which statement is most accurately reflected by this printout?
(1.0) a.
Maximum LHGR(s) in the core is 4.92 kw/f t.
b.
Maximum LHGR(s) in the core is 6.06 kw/ft, c.
Maximum LHGR(s) in the core is 10.89 kw/ft, d.
Maximum LHGR(s) in the core is 13.40 kw/ft.
6.
Plant Systems Design, Control, and Instrumentation 6.1 Backup Scram valves provide a redundant means of venting air from the scram pilot valves and scram discharge valves.
These backup
- valves are.... (CHOOSE ONE)
(1.0) a.
...normally energized and will de-energize upon a RPS Scram signal.
b.
... aligned such that two valves in series, one from each RPS trip channel, must actuate to vent the scram air header.
c.
... designed such that both RPS channels must trip in order for any one of the valves to actuate.
d.
... powered from the RPS Buses A and B.
6.2 Trace on Fig. 6.2 (or describe in detail) the flowpath of exhaust water from the CRD mechanism following a normal rod insertion.
Include the specific component (s) or system section(s) the water travels through until it is no longer in the CRD system.
(1.0) 6.3 The Reactor Recirculation Pump seal cartridge assemblies consist of two sets of sealing surfaces and breakdown bushing assemblies.
Failure of the #2 seal assembly at rated conditions would result in... (CHOOSE ONE)
(1.0) a.
...an INCREASE i'n #2 seal cavity pressure from approximately 500 psig to approximately 1000 psig.
b.
...s DECREASE in #2 seal cavity pressure from approximately 500 psig to approximately 0 psig.
c.
...an INCREASE in #1 seal cavity pressure from approximately 500 psig to approximately 1000 psig, d.
...a DECREASE in #1 seal cavity pressure from approximately 500 psig to approximately 0 psig.
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6.4 The plant is operating at 26% power and both recirc pump M/A transfer stations are in manual and set for 28% speed.
The recirc flow A limit annunciator is clear.
For each of the following instances, indicate how the speed of recirc pump "A" would change (i.e., increase, decrease, or remain same) and which component (s) of the control system is limiting.
See Figure 16 for information.
a.
Recirc pump "A" M/A transfer station placed to " Auto" (1.0) b.
Tachometer output feedback signal fails low-contact Y1 opens (1.0) l 6.5 Assume the following initial rod position distribution:
All rods in groups 1 through 3 are fully withdrawn, except for one rod in each group 55 in group 1, 46-55 in group 2, and 18-03 in group 3 - all fully inserted.
All rods in groups 4 through 10 are fully inserted to position 0 except for rod 34-27 in group 4 which is fully withdrawn.
Fill in the following table with the Rod / Rod Group number you would expect to see displayed in each RWM window for both situations below (a&b) upon initialization or relatch of the RWM.
If nothing will be displayed write " BLANK."
(2.0)
RWM (a)
(b)
Window Initial Same as IC Condition but rod 22-55 (IC) withdrawn to 48 Rod Group Insert Error Insert Error Withdraw Error 6.6 Upon a loss of instrument air, how will the following valves fail?
(Closed, open, as is) a.
CRD flow control valve
(.5) b.
F&C SU level control valve
(.5) c.
Scram discharge volume drain valve
(.5)
Assume this is a complete loss of interruptible and non-interruptible instrument air.
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6.7 With the plant operating at 100% power (Unit 1), recirc in master manual, an operator inadvertently increases the " Pressure Set" by 5 psig. Which of the following responses most correctly describes the initial response and final status of the following parameters due to this action?
(1.0)
Assume 1. No operator action
-2. Starting Parameters
- TCVs - 100% steam flow position
- BPVs - 0% steam flow position
- RxPower - 100%
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- RxPressure - 1010 psig
- 3. All EHC control settings are standard Note: All valve %s are % Steam Flow Position.
See Figure 3.3-15 for information.
a.
b.
c.
d.
Initial Response
- TCVs
=82.5%
=82.5%
=82.5%
=100%
- RPVs 0%
=17.5%
=17.5%
=10%
- RxPower Increase 100%
100%
<100%
/*
- RxPressure Increase 1010psig 1010psig
<1010psig Final Status
- TCVs
=100%
=100%
=82.5%
=100%
- BPVs 0%
0%
=17.5%
0%
- RxPower
>100%
100%
>100%
100%
- RxPressure
>1010psig 1010psig
>1010psig 1010psig 6.8 The core contains 124 LPRM detectors in 31 " detector assemblies" l
(stainless steel tubes with 4 LPRM detectors each).
Fill in the blanks.
The " detector assemblies" are (wet, dry) tubes and are 4
(symmetrically, assymmetrically) located in the core.
(1.0) l 0
6.9 Assume that APRM "B" currently has eleven operable LPRM inputs and is reading 65% power. Which of the following indication (s) and/or action (s) will occur as a result of one LPRM (of the eleven remaining LPRM inputs to APRM "B") failing downscale? Assume no operator action.
(1.0)
+
o a.
LPRM downscale alarm - APRM "B" reading <65%
b.
LPRM downscale alarm - APRM "B" reading >65%
c.
LPRM downscale alarm - APRM Inop Alarm - Rod Block -
APRM "B" reading 65%
d.
LPRM downscale alarm - APRM IN0P Alarm - Rod Block -
1/2 scram - APRM "B" Reading 65%
P b
4 6.10 Which of the following axial location sequences correctly describes the axial locations of LPRMs in the core?
(1.0)
Note: All measurements are inches above BAF,
BAF + "A" 9 + 9" + "B" @ + 27," + "C" 0 + 45" + "D" @ + 63" a.
b.
BAF + "A" @ + 18" + "B" @ + 54" + "C" 9 + 90" + "D" @ + 126" c.
BAF + "D" 9 + 9" * "C" 9 + 27" + "B" @ + 45" + "A" @ + 63" d.
BAF + "D" 9 + 18" + "C" @ + 54" + "B" 0 + 90" + "A" 0 + 126" 6.11 The RSCS enforces Group Notch Control from
% rod density to
% reactor power as sensed by (1.0) 6.12 The APRM scram function actually consists of two separate setpoints; i.e.,.66w + 54% and a fixed 120% scram.
t f
a.
Where, specifically, is/are the sensor (s) located which j
measure the variable "w"?
(0.5) b.
While operating at power, one MSIV fails shut resulting in a brief (1 Second) flux spike to 121% power. Which of the two scram setpoints mentioned above (or both) should initiate a reactor scram? Justify your choice.
(1.0) i 1
6.13 Which AGAF value (P-1 Printout) is more conservative?
(0.5) a.
.99 l
b.
1.01 6.14 The main turbine is at 1800 rpm in preparation for synchronizing l
the main generator to the grid (i.e. the 230 kv generator breakers i
are still open). What will happen if the "All Valves Closed" pushbutton is depressed?
(1.0) a.
Nothing will happen since the synchronous speed select signal is sealed in.
i b.
The turbine control valves and main stop valves will close, but the intercept valves will remain open.
c.
All of the control valves (TCVs and IVs) and main stop valves will close.
d.
The control valves (TCVs and IVs) will close, but the main stop valves will remain open.
5 6.15 SELECT which one of the following an operator does to INCREASE VARS.
(1.0) a.
INCREASE Generator Speed b.
INCREASE Capacity Factor c.
INCREASE Generator Voltage d.
INCREASE Generator Stator Cooling 6.16 What components receive their cooling water supply from the vital service water header?
(1.5) 6.17 Which of the following sequences of components correctly reflects the normal HPCI condensate flow path on Unit 27 (1.0)
I a.
CST + Booster Pump + Main Pump + "A" FW Line, upstream of FW Flow detector b.
CST + Booster Pump
CST + Main Pump + Booster Pump + "A" FW Line, upstream of FW flow detector d.
CST + Main Pump + Booster Pump 4 "A" FW Line, downstream of l
FW flow detector l
1 6.18 The RCIC (Reactor Core Isolation Cooling) System is capable of taking a suction from the CST or the suppression pool.
The suppression pool suction valves (F031 and F029) and CST suction i
valve (F010) are interlocked such that... (CHOOSE ONE)
(1.0)
I a.
... the suction will automatically transfer from the CST to I
the suppression pool on high suppression pool water level, b.
... the CST suction valve will automatically open if the suppression pool suction valves are manually closed while in standby mode.
c.
... the CST suction valve will automatically close if the l
suppression pool suction valves are opened.
d.
... the CST suction valve and both suppression pool suction valves will automatically close on a Group V (RCIC) isolation signal.
6 6.19 Which of the following logic signal combinations most correctly detail the complete logic sequence fer the automatic initiation of the RHR system in the LPCI mode?
(1,0) a.
Rx vessel low level (LL #3)
- or -
Drywell high pressure Rx vessel low level (LL #3) with Rx vessel low pressure b.
- or -
Drywell high pressure with Rx vessel low pressure c.
Rx vessel low level (LL #3) with Rx vessel low pressure
- or -
Drywell high pressure Rx vessel low level (LL #3) d.
- or -
Drywell high pressure-with Rx vessel low pressure 6.20 Assume the CS pumps had automatically started in response to a Rx low-low-low level signal.
The operator secures the CS pumps when Rx level is restored to 185".
Which of the following operator actions must be taken to insure proper automatic start of the CS pumps on any subsequent CS initiation signal?
(1.0) a.
No operator actions are required, b.
CS pump control switches must be cycled - i.e., pump re-started and secured.
c.
CS " initiation signal sealed in" reset PB(s) must be depressed.
d.
CS pump control switches must be taken from the stop position to the auto position.
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7 6.'21 The reactor is critical at approximately 5 psig and the
" pressurization" phase of Gp-02 is being performed.
The Normal Control Range GEMAC LIs in the control room read the following
" approximate" values:
GEMAC A (N004 A) 187" GEMAC B (N004 B) 188" GEMAC C (N004 C) 187" a.
The two Emergency System Range (Yarway) control room level indicators should read approximately... (CHOOSE ONE)
(1.0) 1.
.. 150 inches 2.
.. 165 inches 3.
.. 188 inches 4.
.. 210 + inches b.
The Shutdown Vessel Flooding Range control room level indicator should read approximately... (CHOOSE ONE)
(1.0) 1.
.. 150 inches 2.
.. 165 inches 3.
.. 188 inches 4.
.. 210 + inches 6.22 The plant is operating at 100% power in 3-element control when one steam flow input signal is lost to the FWLC system. Which of the following responses describe the correct system / plant response?
Assume no operator action. See Figure 6.22 for information.
(1.0) a.
Reactor water level decreases and stabilizes at a lower level, b.
Reactor water level decreases and initiates a reactor scram.
c.
Reactor water level increases and stabilizes at a higher level.
d.
Reactor water level increases and initiates a turbine trip (w/ scram).
8 6.23 Which one of the following statements correctly describes the operation of the Motor Gear Unit (MGU) in controlling Reactor Feed Pump (RFP) turbine speed?
(1.0) a.
The MGU will control the RFP turbine speed only if its speed demand signal is greater than that from the MSC.
b.
The MGU can be used to control feed flow rate over a turbine speed range of approximately 0-5500 rpm.
c.
The MGU is manually controlled from the control room at either a high or a low speed rate.
d.
The MGU will fail "as is" to prevent a ramp response if it loses its signal from the flow controller.
6.24 Attached Figure 2 :'epicts the UPS power supply line-up and switch contact alignment for Unit 2 UPS with Inverter 2A supplying.
Utilizing Figure 2 as a reference, describe what will automatically occur if inverter output 2A is lost, i.e., what is the new source of UPS power and what switch contacts change positions?
(1.0)
~
f 7.
Procedures - Normal, Abnormil, Emergency and Radiological Control 7.1 a.
SHELL WARMING is in progress. The internal bypass in 2 TSV is slowly opened to raise shell pressure.
With no further operator action a scram results.
Briefly EXPLAIN WHY.
(1.0) 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 tnem per GP-03, Unit Startup and Sychronization.
(1.5) 1.
Intercept valves 1 and 3 - open slowly 2.
Main stop valves 1, 3, 4 - open slowly 3.
Increasing speed light comes on 4.
Main stop valve #2 - begins to open 5.
Control valves - throttle open 6.
Intercept valves #2 and #4 - start to open
~
-c.
When bringing the turbine to rated speed, it is recognized that abnormal vibration exists. What action should be taken?
(0.5) 7.2 With regard to GP-02, Approach to Criticality and Pressurization:
a.
What two actions must be performed if a period of five seconds or less is reached?
(1.0) b.
What action shoulci be taken if a single notch withdrawal results in a period of 20 seconds?
(0.5) 7.3 Complete the blanks of the following ECP CAUTION:
DO NOT SECURE OR PLACE an ECCS OR RCIC in manual mode UNLESS by at least independent indications:
(0.5) 1.
(0.5) 0_R 2.
(0.5)
2 7.4 With regard'to OP-16, RCIC:
a.
When placing RCIC in standby, the sequential steps are as follows:
1.
Verify that the steam supply outboard isola-tion valve, E51-F008 is closed.
2.
Verify that the steam supply inboard isola-tion valve E51-F007 is closed.
3.
Open the supply drain pot drain bypass valve, E51-F054.
4.
Open the steam supply outboard isolation valve, E51-F008.
5.
Slowly throttle open the steam supply inboard isolation valve E51-F007.
Why are you performing these steps, and what are the consequences of opening F007 quickly?
(1.0) b.
When operating RCIC for Rx pressure control, a caution states that opening the bypass to CST valve may cause turbine speed to decrease below 2000 RPM. Why is RCIC operation below 2000 RPM undesirable (2 reasons)?
(1.0) 7.5 Answer the following questions with regard to reactor recirculation pump operational limitations and precautions:
a.
GP-05, " Unit Shutdown," states that recirculation pump operation at a suction pressure below 300 psig should be minimized. Why is this recommendation necessary?
(0.5) b.
When increasing recirc pump speeds with both controllers in MANUAL, their speeds should normally be maintained within The speed differential is limited to
% when below 75% core flow and when above 75% core flow.
(1.5)
a 3
7.6 Answer the following questions with regard to the Reactor Water Cleanup System Operating Procedure (OP-14):
a.
Reactor Coolant temperature is 300*F and a major portion of the RWCU flow is being rejected to the condenser.
Briefly explain gh it is recommended that flow back to w
the reactor vessel be established slowly over a 45 minute period.
(1.0) b.
Op-14 cautions the operator to maintain maximum RWCU System flow and temperature when operating at low power.
Why is this practice recommended?
(1.0) 7.7 An improper RBCCW system lineup could result in possible damage to the pumps and/or heat exchangers as stated in the " CAUTIONS" of the system operating procedure (OP-21). Which of the following lineups / conditions would minimize the likelihood of component damage over an extended operating period?
(1.0) a.
Running one RBCCW pump with two RBCCW heat exchangers.
b.
Running two RBCCW pumps with two RBCCW heat exchangers.
c.
Running two RBCCW pumps with one RBCCW heat exchanger.
d.
Running two RBCCW pumps with three RBCCW heat exchangers.
7.8 GP-07, " Preparation for Core Alterations," cautions the operator to suspend fuel movement in the fuel pool near the fuel pool gates while work is in progress in the reactor cavity.
Briefly explain why this precaution is necessary.
(1.0) 7.9 A LOCA has occurred and a high temperature steam environment exists in the drywell.
Briefly explain why drywell sprays must NOT be initiated in the " unsafe" region of figure #16, "Drywell Spray Initiation Pressure Limit" (attached).
(1.0) 7.10 Unit 2 is operating at 10% power when the "A" CRD pump seizes due to a failed bearing.
Upon starting the "B" o
CRD pump in accordance with A0P-02.1, " Inability to Move Control Rods," it immediately trips on overload and cannot be restarted.
When is a manual reactor scram required per the A0P?
(1.0) a.
If reactor pressure is below 800 psig.
b.
Immediately.
c.
Upon activation of the "CRD HYD TEMP HIGH" annunciator, d.
Upon activation of the "CRD ACCUM LO PRESS /HI LEVEL ALM" annunciator.
4 7.11 Which of the following is a symptom you would expect to see as a result of a " Jet Pump Failure" (A0P-04.4)?
(1.0) a.
Increase in generator megawatt output.
b.
Increase in core thermal power.
c.
Increase in total core flow.
d.
Increase in core plate dp.
7.12 Unit 1 is operating at power when a Safety Relief Valve (SRV) fails and sticks in the open position.
Per A0P-30.0, " Safety Relief Valve Failures," when must a manual reactor scram be initiated?
(1.0) a.
Prior to the automatic reactor scram at 850 psig.
b.
As soon as it is recognized that the SRV will not close.
c.
As soon as suppression pool water temperature reaches 120'F d.
Immediately upon discovering that the SRV is open.
7.la Match the. automatic action in column "A" with the system para-meter in column "B" which will initiate that action.
(2.0)
Column A Column B a.
Service Air header 1.
Service Air header Pressure isolates decreases to 105 psig.
b.
Interrubtable Instrument 2.
Instrument air header Air header isolates pressure decreases to 103 psig.
c.
Standby reactor building 3.
Instrument air header air compressors start pressure decreases to 100 psig.
d.
Air compressors A, B, and 4.
Non-interruptable instrument C start and load air header pressure decreases to 95 psig.
7.14 What are the four abnormal conditions that would be reason to terminate fuel handling operations per Fuel Handling Procedure FH-11?
(2.0) 7.15 A reactor cooldown is in progress and neither RHR loop can be placed in the shudown cooling mode.
In accordance with A0P-15.0, Alternate Shutdown Cooling has been established to remove decay heat and continue the cooldown.
State the condition / status of each of the following components / parameters when operating in this mode of shutdown cooling.
(2.5) a.
MSIV's b.
SRV's c.
RHR loops A and B d.
Rx pressure (with respect to suppression chamber pressure) e.
Rx level (provide numerical value or component reference)
p._
5 7.16 List four of the five entry conditions (parameters'and set-points) for the Containment Control Procedure, E0P-01-CCP.
(2.0) 4 m
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8.0 Administrative Procedures, Conditions, and L1mitations j
i 8.1 Which of the following individuals represents the minimum level of authority required for cancellation j
of an LC0 per 0I-04, LC0 Evaluation and Follow-up?
(1.0) a.
Regulatory Compliance Specialist b.
Control Operator c.
Shift Foreman d.
Shift Operating Supervisor 8.2 Who is the individual charged to perform the preliminary / initial determination of whether an event requires a red phone or prompt report (as defined in the Technical Specifications) per 0I-04, LCO Evaluation and Follow-up?
(1.0) a.
Control Operator b.
Shift Foreman c.
Shift Operating Supervisor d.
Shift Technical Advisor 8.3 Until the EOF is activated, the Site Emergency Coordinator can not delegate the responsibility for... (CHOOSE ONE)
(1.0) a.
... 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.
2 8.4 Which of the following statements correctly describes a Priority 1 work request per the
" Corrective Maintenance" Procedure, MP-147 (1.0) a.
It may be worked on a 24-hr/ day, 7-day / week schedule upon approval by the General Manager.
b.
It must be approved by the Shift Operating Supervisor and the Maintenance Supervisor.
i c.
The WR&A must be completed and signed prior to commencing the maintenance activity.
d.
Priority 1 shall be assigned to all failures of safety-related equipment requiring immediate plant shutdown.
8.5 What is the difference between a DEPARTURE and a DEVIATION from an established procedure?
(1.0) f
--..-.-.7,--
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1 8.6 Procedure 01-01, Operating Principles and Philosophy, states that during the performance of normal evolutions by two persons at different locations, both persons should have a copy of the procedure.
In which of the following situations would it be allowable per 01-01 for only one individual to have a copy of the procedure?
(1.0) a.
During an evolution that requires only a limited number of manipulations by an individual under the direction of the controlling individual.
Only the individual controlling the evolution need have a copy of the procedure.
b.
During an evolution that requires only a limited number of manipulations by an individual under the direction of the controlling individual.
Only-the individual performing the manipulations need have a copy of the procedure.
c.
During an evolution in a contamination area that will be completed within one hour by the individual performing the manipulations irregardless of.the number (of manipulations) involved.
Only the individual controlling the evolution need have a copy of the procedure.
d.
During an evolution in a contamination area that will be completed within one hour by the individual performing the manipulations irregardless of the number (of manipulations) involved. Only the individual performing the manipulations need have a copy of the procedure.
4 8.7 Match the following four emergency classes to the appropriate definition.
(2.0) a.
Unusual Events
,1.
Events are in process or have occurred which involve actual or likely major failures of the plant functions needed for protection of the public.
b.
Alert 2.
Events are in process or have occurred which indicate a potential degradation of the level of safety of the plant.
c.
Site Emergency 3.
Events are in process or have occurred which involve actual or imminent substantial core degra-dation or melting with potential for loss of containment integrity.
d.
General Emergency 4.
Events are in process or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant.
8.8 In accordance with 10 CFR 55, "if a licensee has not been actively performing the functions of an operator or senior operator for a period of months or longer, he shall, prior to resuming activities licensed pursuant to this part, demonstrate to the Commission that his knowledge and understanding of facility operation and administration are satisfactory."
- Fill in the blank with one of the following times.
(1.0) a.
4 b.
6 c.
12 d.
24
5 8.9 In which of the following situations would it be possible to utilize a human red tag in lieu of a properly issued clearance per AI-58, Equipment Clearance Procedure?
( 1,.0) a.
During the processing of a clearance boundary extension for Priority 1 maintenance, if the Clearance Tag Sheet (BSEP 20) will be completed within one hour of the human red tag assignment and it is approved by the Shift Operating Supervisor and the Maintenance Supervisor, b.
During maintenance on equipment that has been rendered safe by the placement of approved wire removal tags, if the maintenance will be complete within one day and it is approved by the Shift Foreman and the Maintenance Supervisor.
c.
During maintenance on Inoperable Safety / Technical Specification related equipment which, if not returned to operability status, will require a plant shutdown within one hour. The approval of the Shift Operating Supervisor or the Shift Foreman is required.
d.
During an emergency situation as
' determined by the Shift Operating Supervisor or the Shift Foreman.
8.10 Clearances may be transferred from one clearance holder to another provided both persons agree to the transfer. Which one of the following individuals must also give his/her consent per AI-58, Equipment Clearance Procedure?
(1.0) a.
Supervisor of original clearance holder.
b.
Applicable unit Control Operator.
c.
Applicable unit Shift Foreman.
d.
Shift Operating Supervisor.
l
6 8.11 clearances.
(definition) - Issued to two or more l
individuals which have the same boundary for work that may may or may not be related.
- Fill in the blank with one of the following terms.
(1.0) l a.
Local b.
Individual c.
Multiple d.
Master 8.12 Unit 2 Technical Specifications define SHUTDOWN MARGIN as...
" SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor would be subcritical assuming..."
List the three (3) conditions which complete the definition of SHUTDOWN MARGIN.
(1.5) 8.13 Given the following conditions on Unit 2:
Refuel Mode Switch 180 F Temperature O psig Pressure 184 inches Level SDC Mode RHR Head bolts to the RPV are DETENSIONED STATE the above described Operational Condition.
(0.5) shall be the qualitative assessment of 8.14 A channel behavior during operation by observation.
This determination shall include, where possible, comparison of the channel indication and/or status with other indication and/or status derived from independent channels measuring the same parameter.
- Fill in the blank with one of the following TS terms.
(1.0) a.
Channel Calibration b.
Channel Check c.
Channel Functional Test d.
Logic System Functional Test 8.15 The APRM Trip Setpoint Formula is (.66W + 54%) T.
What is the definition of variable "T"?
When is it applied to the Formula?
(1.0)
7 8.16 Unit 2 is at 75% rated thermal power, with two outstanding deficiencies:
The Auto - swap of the HPCI suction upon receiving CST low level is determined to be UNSATISFACTORY.
The suction is MANUALLY switched to the Suppression Pool, and the suction to the CST is ISOLATED.
Which of the following actions most correctly detail the allowances and/or limitations imposed by the Technical Specifications in this instance?
(1.0)
NOTE:
Applicable TSs are enclosed for reference.
a.
No new limitations or TS Operational Condition restrictions are initiated by this re-alignment.
b.
Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor steam dome' pressure to less than or equal to 113 psig within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
c.
Be in at least HOT SHUTDOWN within six hours 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 />.
d.
Be in at least HOT SHUTDOWN within six hours 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 />.
=
8 8.17 Unit 2 is at 75% rated thermal power with only one outstanding LCO:
HPCI has been INOP for nine days due to an in progress repair.
Ten minutes into the shift, RHR Pump "A" fails to start twice during the performance of a scheduled surveillance and is declared IN0P.
Which of the following actions most correctly detail the allowances and/or limitations imposed by the Technical Specifications in this instance?
(1.0)
Note:
Applicable TSs are enclosed for reference.
a.
Power Operation may continue for five days; and then, be in at least HOT SHUTDOWN within 4
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 />.
b.
Power Operation may continue for seven days; and then, 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 />.
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 six hours and in COLD SHUTDOWN withi the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
9 8.18 Unit 2 is in COLD SHUTDOWN during a reactor startup with no outstanding deficiencies.
The i
Containment Atmosphere Dilution (CAD) system becomes IN0P.
It is anticipated that repairs will be complete within two weeks.
With regard to the reactor startup, which of the following actions most correctly detail the allowances and/or limitations imposed by the Technical Specifications?
(1.0)
NOTE: 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 continue; Operational 1
Condition 1 may be entered but Thermal Power is limited to 15%.
c.
Startup activities may continue; Operational Condition 2 may be entered but not exceeded.
d.
Startup activities may continue; however, Operational Condition 4 must be maintained.
(Entry into Operational Condition 5 is acceptable.)
m 10 4
8.19 Unit 1 is shutdown and in a long term outage.
Unit 2 has been recently shutdown and placed in Cold Shutdown - cooldown completed last shift.
The shutdown and cooldown of Unit 2 was necessitated by a requirement to drain and visually inspect the Suppression Pool.
The following plant conditions / requirements have 1
been established:
CS system is aligned to the CST with an Operable i
Flow Path capable of transferring water through the spray sparger to the reactor vessel.
Reactor Mode Switch is locked in the Shutdown position.
No operations affecting the reactor vessel or with the possibility of draining the vessel are in progress or planned.
There is only one outstanding deficiency:
DG #4 is INOP due to in progress repairs.
The Shift Operating Super iisor directs that the draining of the Suppression Pool will commence as soon as all of the TS requirements are met.
Which of the following actions most correctly detail the allowances and/or limitations imposed by the Technical Specifications in this instance?
(1.0)
NOTE: Applicable TSs are enclosed for reference.
a.
Commence Suppression Pool draining as soon as practical since all TS LCO requirements are met, b.
Commence Suppression Pool draining only after DG #4 is repaired and declared operabie.
c.
Commence Suppression Pool draining as soon as you insure that No Positive Reactivity changes will occur during this condition.
d.
Commence Suppression Pool draining as soon as you insure that No Positive Reactivity changes will occur and that one LPCI subsystem is operable.
~ - _ _ _
11 1
8.20 Which of the following individuals must perform a Control room board walk down prior to assuming the shift position per 0I-02, Shif t Turnover Checklist?
(1.0) a.
Control Operator b.
Shift Foreman c.
Shift Operating Supervisor d.
All of the above 8.21 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?
(1.0) a.
Remove annunciator card 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 card.
Identi fy 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".
8.22 Briefly explain 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."
(1.0)
~
12 8.23 The following data resulted from the DSR Drywell leakage calcula-tions required to be made every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
The data was taken on Unit 2 during a single day of operation at Operational Condition 1 (The unit has been in Operational Condition 1 for 2 weeks). Only final data is presented; Preliminary calculational data is not supplied.
SHIFTS 00-04 04-08 08-12 Floor Drain Leak Rate 2.52 gpm 4.25 gpm 3.75 gpm Equipment Drain Leak Rate 20.91 gpm 20.58 gpm 21 gpm Leak Rate to Drywell 23.43 gpm 24.83 gpm 24.75 gpm 12-16 16-20 20-24 Floor Drain Leak Rate 4.3 gpm 4.48 gpm 5.2 gpm Equipment Drain Leak Rate 22.25 gpm 24.33 gpm 19.33 gpm Leak Rate to Drywell 26.55 gpm 28.91 gpm 24.53 gpm State the 4 TS Operation Leakage LCO limit (s) applicable (3.0) in this plant condition and Identify any that were exceeded (as indic1ted by the above data).
NOTE: DSR Drywell leakage calculation sheet is provided as information - applicable limits have been deleted.
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DRYW ELL SPRAY INITI ATION PRESSURE LIMIT FIG'JRE #16 BSEP/VOL. VI/EOP-01/UG Page 77 of 93 Rev. 1
e DRYWELL LEAKAGE CALCULATION E
RASE VAll!E LEARACE =
gpm*
SAT SUN MON TUES WED THURS TH1 en((
SUMP FILL TIMER SETTING =
minutes
- MANUALLY PUMP FLOOR DRAIN SUMP - RECORD TIME PUMP o
I~
STOPS (LOW LEVEL TRIP).
j CALCULATE TIME INIIRVAL (MINUTES) SINCE SUMP WAS s
LAST MANUALLY PUMPED.
RECORD INTEGRATOR READING.
1
[3 CAtCULATE DIFFERINCE IN INTEGRATOR READINC FROM TIME SUMP WAS LAST MANUALLY PUMPED (LEAMAGE).
' CALCUlAT[ Fl0OR DRAIN LEAK RATE (DIVIDE LEAKACE BY TIM U lillRyALJ.**
MANUALLY PUMP IQUIPMENT DRAIN SUMP - RECORD TIME PUMP STOPS (LOW LEVEL TRIP).
CALCULATE TIME INTERVA1 (MINUTES) SINCE SUMP WAS LAST MANUALLY PUMPLD.
RECORD INIIGRATOR READING.
CAICULATE DIFFERENCE IN INTECRATOR READING FROM TIME SUMP WAS LASI MANUALLY PUMPED (LEAKACE).
j y
1 LP CALCULATE EQUIPMENT DRAIN LEAR RATE (DIVIDE LEAMAGE l
RY TIM (_INTERYAll.
FLOOR DRAIN LEAM RATE.
(lr Zero, see xxx below)
$2 SM 9#9 EQUIPMENT DRAIN LEAM RATE.
(fr zero, see xxx below).dD,9/ 8/n EMD LEAK RATE TO DRYWELL.
/
/
/
/
/
/
/
1NITIALS: IControI Op_ereigr/Silft fOIemanI l
- NOTE:
esse value leakage and sump rill timer setting will be determined by 16-24 shirt, Friday. Base value leakage will normally be the weekly everage leakage, except arter an outage, in which case the last base value determined at normal operating pressure will be used until the fi rst 24-hour everage leakage can be Af ter determination of the new base value leakage (IIVl),
1 obtained af ter return to normal operating pressure.timor setting to either 350/(BVL + 2) or 150 minuten, i
the floor drain sump rill calculate and readjust Ihese values should be carried forward to the new week's DSR.
d whichever is less.
i l
leakage daily at approximately 0900.
leak calculations requirosi when greater than 212*r cool 4nt temperature and Irradiated fuel in reactor vessel Calculate drywell (conditions I, 2, and 3) - reference - Technical Spectrication 4.4.3.Pa and AOP-01.
Sump At least xxx A channel functional test la required once/31 days per Technical SpecI rication 4.4.3.1.b in cond i tion 1, 2, or 3.
one value other than zero during three consecutive readings constitutes an acceptable runctional test.
wu)
UNIT R
SHiri T)O - 04/
BRUNSWICK STEAM ELECTRIC PLANT DAILY SURVEILLANCE REPORT CONTROL OPERATORS WEEN Of TO I
==
n.
~
OdlVWt LL LL AILAGt CALCutAt8ON I
E BASE VALUE LEAMAGE =
gpm*
SAT SUN Moel TUES WED THUNS I #4 I ns 2
SUMP FILL IIMER SEITING =
minutes
- ugegALLy PUMP [Lo0R ORAIN SUMP. RECORO TIME PUMP o
l I
SIOPS ( LOW LIVEL T*l P).
e j
CALCul. ATE IIME INIIRVAL (MINUIES) Sl80CE SUMP WAS LAST MAleUAtLY PUMPED.
,-e P.
RECORO INIEcRATOR READING.
a o
CAL Cut AT E.DI f IE RE NCE IN INIECRAIOR READINC IROM flME SUMP WAS LASI MANUA1.LY PUMPED (LEAMAGE).
y j
CALCutATE ISOOR DRAIN LEAK RATE (OlVIDE LEARACE J
aY TIME I NI I HyAM. *
J l
CALCUBAIE IIMI S NI E RVAl (MINUTES) SINCE SUMP WAS LASI MANUALLY PUMrtD.
RECORD INi(GRATON READING.
CAICut ATE Dif f ERENCE IN INIECRATOR READING FROM IIME SUMP WAS LASI MANUAlt.Y PUMPED (LEAMAGE).
CAL CULATE EQUIPMENT DRAIN LEAK RATE (OlVf DE LEAMAGE SY TIME INH HVAll.
see xxx bel'ow)
N b4 I
FLOOR DRAIN L[AK RATE.
(1r zero,
[QUIPMENT DHAlN LEAM RATE.
(lF 2ero, see xxx below) 20,W M1 l
- 24. 83 A4 LEAR RATE 10 DRWELL.
INITIALS: IControl Op_erator/Slal f t forgman)
/
/
/
/
/
/
/
timer setting will be determined by 16-24 shirt, Friday. Base value leakage and sump ritt in which case the last base
- $101E:
Base value normally be the weekly everage leakage, except arter en outage,the first 24-tiour everage leakage cain bc leakage will ne used until value determined at normal operating pressure will Af ter determination of the new base value leakage ( Hvi ),
obtained af ter return to normal operating pressure.timor setting to either 350/(HVL + 2) or 150 minutsu, the floor drain sump till calculate and readjust These values should be carried fosverd to the new week's DSR.
whichever is less.
l le Ma~~e dalTy at approximately 0900.
i r rad ia ted fuel in reactor vessel Calculate drywell g
leak calculations requirod wisers greater than 212*F coolint temperature and l
(conditions 1, 2, and 3) - rernrence - Technical Spectrication 4.4.3.Pa and AOP-01.
Sump i
so<
is required once/31 days per Technical Spec t ricat ion Is.4. 3.1.b in condition 1, 2, o r 3.
At least one value other than zero during three consecutive readings constitutes an acceptable functional test.
xxx A channel functional test v*
UNIF 48 SHIFT 04f-R BRUNSWICK STEAM ELICIRIC PLANT DAILY SURVEILLANCE REPORT WEEM OF 10 CONTROL OPERATORS 9
DRYwCLL LEAstACE CALCUIAll088 E
BASE VALUE LEAKAGE =
spe*
SAT Sull MON TUES WED THURS fH8
=
Q Super f ILL TIMER SETilleG =
minutese
<:o useUALLY PUMP FLOOR DRAIN SUMP - RECORD TIME PUMP I
STOPS (LOW LEVEL IRIP).
e j
CALCULATE 11ME INIE RVAL (M191UTES) SileCE SUMP WAS LAST MA800 ALLY PUMPED.
H i
s 11[ CORD INTEGRATOR READ 180G.
O CALCUt AIE DIFFERE81CE IN INTEGRATOR READ 110G FROM IIME SUMP WAS LASI MANUALLY PUMPED (LEAKAGE).
CALCULATE F100R DRAIN LE AK RATE (DIVIDE LEAKAGE BY T lj( 11Lil RVAll. **
MANUALLY PUMP EQUIPMENT DRAIN SUMP - RECORD TIME PUMP STOPS (LOW LEVEL IRIP).
CALCULAIE IIME INTERVAL (MINUIES) SINCE SUMP WAS LASI MANUALLY PUMPLD.
f RECORD INIICRATOR READING.
I CAICULATE DIFFERENCE IN INTEGRATOR READil1G FROM TIME SUMP WAS LAST MANUALLY PUMPED (LEAKAGE).
CALCULATE EQUIPMENI DRAIN LEAK RATE (OlVIDE LEAKAGE i
BY TIME INTERVAll.
i TLOOR D8tAIN LEAK RATE.
( I f zero, see xxx below)
E 7[8#"I EQUIPMENT DRAIN LEAK RATE.
(IF 2ero, see'xxx below) O/ b 8d75 b LEAK RAi[ 10 DRYWELL.
INITIALS: (Cont rol _oggerstor/Stil f t forement
/
/
/
/
/
/
/_ _
- 100TE:
Base value leakage and sump Fill timer setting will be determined by 16-2ae shift, Friday. Base value in which case the last base leakage will norme11y be the weekly everage leakage, except efter en outage, the fi rst 28 -hour everage leakage can be value determined at normel operating pressure will De used until 6
obtelned af ter return to normal operating pressure. Af ter determinetlon of the new base value f eekage (Hyt I, l
calculate and read.just the rioor drain sump fill t i ehe r se t t i ng to e i the r 3'>0/ ( HVL + 2 ) o r 150 m i nu t e t.,
These values should be corried forverd to the new week's DSR.
whichever is sens.
leakage dedfy at approximately 0900.
Calculate drywell lesh calculatiosis require 1 whern greater then 212*F cooldnt temperature and Irradiated fuel in reactor vessel Sump (conditions 1, 2, and 3) - reference - Technical Spectrication as.le.J.2e and AOP-01.
j
,e i
la recluired once/31 days per Technical Spec t rication 8e.as.3.1.b i n cond i t i on 1, 2, o r 3.
At least i
xxx A channel functional test
]
w One value other than Zero during three consecutive readings constitutes en acceptable functional test.
j UNIT 2.__
]
SHIFT Ok-/1 BRUNSWICK STEAM ELECTRIC PLANT
~
DAILY SuhvEILLANCE REPORT C010 TROL OPERATORS WEEK OF TO 1
DRvwELL LEAI AGE CALCut Alleet 4
BASE VALUE LEAKAGE =
gpm*
SAT SUN MON TUES WED THURS fRI l
P8 2
SUMP FILL TIMER SETilleG =
minutes
e CALCULATE TIME INif RVAL (MINUTES) SINCE SUMP WAS LASI MAseUALLY PUMPED.
s s
~~
3 RECORD INTEGRATOR READING.
j i
l 8
CALCULATE DIFFERENCE IN INTEGRATOR REA0lldG FROM 4
IIME SUMP WAS LASI MANUALLY PUMPED (LEAKAGE).
CALCULATE F100R DRAIN LEAK PATE (OlVfDC LEARAGE BY TIME Illi[ RVAl l. **
1 I
MANUALLY PUMP EQUIPME80I DRAIN SUMP - RECORD TIME j
PUMP STOPS (LOW LEVE L TRI P).
CALCULAIE TIME INTERVA1 (MINUTES) SINCE SUMP WAS LAST MANUALLY PUMPEU.
RECORD INTEGRATOR READ 186G.
CAICutAIE DIFFERL90CE IN INTEGRATOR READif0G FROM ilME SUMP WAS LASI MANUAll.Y PUMPED (f.EAMAGE).
y CAL CUL A1E EQUIPMENI DRAIN LEAK RATE (OlVf DE LEAMAGE v
SY TIME INTERJAQ.
FLOOR DRAIN LEAR RATE.
( 1 r Zero, see xxx below)
M. b
)
EQUIPMENT DRAIN LEAK RAIE._ ( I f ze ro, see'xxx below) N2.7$f#1 a7(.. NMBI LEAK RATE TO DRYWELL, INiilALS: 1 CggrgLOgterugtf_ Sit i f t Forgaant
/
/
/
/
/
/
/__
t l
- NOTE:
Base value seekage and sump fill timer setting will be determined by 16-24 shirt, Friday. Base value Isakage will normally be the weekly everage leakege except ef ter en outage, in which case the Inst base the first 24-hour everags leakage enn be j
vols,' determined at normal operating pressure will b used until 4
Af ter determination of the new base value leakage (HVB),
]
-"ad after return to normal operating pressure.timor setting to either 3'>0/(HVL + 2) or 150 minutet.,
obtt the floor drain sump fill celediate and readjustIhese values should be carried forward to the new week's DSR.
whichever is less.
leakage daily at approximately 0900.
Calculate dryweli leek calculations required where greater then 212*f coolint temperature and irradiated fuel in reactor vessel j
j (conditlens 1, 2, and 3) - roterence - Technical Specification 4.4.3.2e and AOP-01.
I Sump i
og At least i s requ i red once/ 31 days pe r Technica l Spec i f ica t ion 4. 4. 3.1. b i n cond i t i on 1, 2, o r 3.
l xxx A channel functionni test One value other than zero during three consecutive readings constitutes an acceptable functional
- test, w
d _
UNil i
SHIFT M-/%
BflustSWICK STEAM ELECTRIC PLANT DAILY SURVEILLAf4CL REPORT CONTROL OPERATORS WEEK Of TO
]
~
we g
i
's DRYWELL LEAKAGE CALCULAll0N BASE VALUE LEAKAGE =
gpm*
SAT SUN MON TUES WED THURS THI "us*:
2 SUNP TILL TIMER SEITileG =
minutes
- gAeIUALLY PUMP ILOOR DRAIN SUMP - RECORD TIME PUMP i
o I
j CALCULATE TIME INilRVAL (MillUTES) SINCE SUMP WAS LAST MANUALLY PUMPED.
RECORD INTEGRATOR READING.
t O
CAICULATE DiffERINCE IN INIECRATOR READING FROM IIME SUMP WAS LASI MANUAlt Y PUMPED (LEAKAGE).
w CALCUBATE IBOOR DRAIN LEAK RATE (DIVIDE LEAKAGE
~
BY TIME INII,H M(1 **
i MANUALLY PUMP EQUl PMf MI DRAIN SUMP - RECORD TIME PUMP STOPS ( LOW LEVE L IRIP).
CALCULAIE TIME INTIRVAl (MINUTES) SINCE SUMP 1
WAS LASI MANUALLY PUMPLO.
i RECORD IMifGRATOR READING.
CAI CUL ATE DI F f f RE NCf IN INIECHATOR READING FROM ilME SUMP WAS LASI HANUALLY PUMPED 'LIANAGE).
y w
CALCULATE EQUIPMENI DRAIN LEAK RAIE (DIVIDE LEAKAGE 8Y TIME 'INilR_ Ell FLOOR DRAIN LEAK RATE.
(if zero, see xxx below)
MMf D EQUIPMENT DRAIN LEAK RAIE.
(if zero, see xxx below) N. M M #1 d.hl LEAK RAIE TO DRYWELL.
l INITIALS: IControl Opersigr/Sillf t foremann
/
/
/
/
/
/
/. _ _.
- NOTE:
Base value leakage and s fill timer setting will be determined by 16-284 shift, friday. Base volun
)
leakage will normally be weekly everage leakage, except af ter an outage, in which case the last base value deteemined at normal operating pressure will De used until the first 24-hour everage laskage ensi be Af ter determination of the new base value leakage (HviI, obtained siter returse to normal operating pressure.timor setting to either 3$0/(llVL + R) or 150 minute..
Llie f loo r d ra in sump f i l l calculate and readjustlinese values should be carried feneverd to the new week's DSR.
whictiever is less, I
leIIiWgIdalTy at iipproximately 0900.
l Calculate drywell leak calculattosis requirent winess greater tisan 212*f cooldnt temperature and Irradiated fuel in reactor vessel l
Sump reference - Technical Specification 4.4.J.?a and AOP-01.
(conditions I, 2, acid 3)
,u i
la requi red once/31 days per Technica l Spec 1 rication is.as.3. I.b in condition I, 2, o r 3.
At least xxx A channel functiunal test test.
one value other than zero durlog titree consecutive readings constitutes an acceptable functional w*
UNil s.
SHIFT
/6 d O BRUNSWICK STEAM ELECTRIC PLANT DAILY SURVEILLANCE REPORT CONTROL OPERATORS WEEK Of TO e.
ORYwtLL LLARACE CALCut At loss s
E BASE VALUE LEAMAGE =
spm*
SAT SUN MON TUES WED THURS f it 8 Super FILL IIMER SElilleG =
minutes
- uMUALLY PUMP FLOOR DRAIN SUMP - RECORO TIME PUMP i
oI STOPS (LOW LEVEL TRIP).
CALCULATE IIME INIfRVAL (MINUTES) SINCE SUMP WAS LAST MA800 ALLY PUMPED.
w RECORO INTEGRATOR READipeG.
s,
~:
O CALCUL ATE DlIFERENCE IN INTEGRATOR READileG FROM ilME SUMP WAS LASI MANUALLY PUMPED (LEAKAGE).
w CALCul AIE It00R DRAIN LEAK RATE (DIVIDE LEAKAGE BY llME INIIRVAtl.**
MAleUALLY PUMP EQUIPMENT DRAIN SUMP - RECORD TIME PUMP STOPS (LOW LEVEL IRIP).
CALCULATE IIME INIFRVAl (MSNUTES) S100CE SUMP WAS LAST MANUALLY PUMPtD.
HLCORD INIICRATOR READING.
CAICutATE DIIFERENCE IN INTEGRATOR READING FROM IIME SUMP WAS LAST MANUALLY PUMPED (LEAKAGE).
y CALCUL AIE EQUIPMENT DRAIN LEAK RATE (DIVIDE LEAKAGE w
Bir TIME I N T E R_y Al l.
see xxx bel'ow) 8.d k#I FLOOR DRAIN LEAK RATE.
( i f zero, EQUIPMENT DRAIN LEAK RAIE.
(if zero, see xxx below) /i N b M b Nb LEAK RAIE TO DRYWELL.
/
/
/
/
/
/
/
i-INiilALS: Icontrol opegelor/Shif t forgment
- IIDIE:
sese value leakage end sump fill timer setting will be determined by 16-24 shif t, Friday. Sese value in which case the last base leakage will normally be the weekly everage leakage, except ef ter en outage,the first 24-hour everage leakage csin be l
value detesmined at normal operating pressure will De used notll Af ter determination of the new base value leakage ( fiVs ),
l obtelned af ter return to normal operating pressure.timor setting to either 350/(tlVL + 2) or l')O minutou, the floor drain sump fill calculate and readjust These values should be carried forverd to the new week's DSR.
whictiever is less.
leakage daily at approximately 0900.
leek calculations requirest when greater then 212*f cool 4nt temperature and I rradiated fuel in reactor vessel Calculate aryweII j
- reference - Ischnical Specification 4.4.3.2e and AOP-01.
Sump i
(conditions 1, 2, and 3)
~
xxx A ch'ennel functional test la required once/31 days per Technical Specification 4.4.3.1.b in condi t ion 1, 2, or 3.
At Inest one value other then zero during three consecutive readings constitutes en acceptable functional test.
l we UNil R
SHIFT MO T2 Y BRUSISWICK STEAM ELECTRIC PL ANT DAILY SURVEILLANCE REPORT WEEK Of TO COIITROL OPERATORS 1
ee o
ATTACHED ARE THE FOLLOWING TECHNICAL SPECIFICATIONS EXCERPTS FOR USE IN ANSWERING QUESTIONS 8.16 thru 8.19 1.
3/4.0 Applicability
- 3.0.1 thru 3.0.5 2.
3/4.5 Emergency Core ' Cooling Systens
- 3.5.1 HPCI
- 3.5.2 ADS
- 3.5.3.1 CS
- 3.5.3.2 LPCI
- 3.5.4 Suppression Pool 3.
3/4.6 Containment Systems
- 3.6.2.1 Suppression Chamber
- 3.6.2.2 Suppression Pool Cooling
- 3.6.6.2 CAD System
- 3.6.6.3 0xygen Concentration 4.
3/4.8 Electrical Power System
- 3.8.1.2 AC Power (5D1
- 3.8.2.2 AC Distribution (SD]
_ - _ - ~ -. - - _. _,. _.. _ _ _.. _ _. _
T/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION
(
i Limiting Conditions for Operation and ACTION requirements shall be 1.0.1 the OPERATIONAL CONDITIONS or other states specified for appiteable during each specification.
1.0.2 dherence to the requirements of the Limiting Condition for operation and associated ACTION within the specified time interval shall constitute If. the event the Limiting Condition for compliance with the specification.
operation is restored prior to expiration of the specified time interval, is not required.
completion of the ACTION statement the event a Limiting Condition for Operction and/or associated 1.013 In be satisfied because of circumstantes in excess of ACTION requirements cannot least ROT those addressed in the specification, the unit shall be placed in at SHilTDOWN 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 following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> unless corrective measures are completed that permit operation under the permissible ACTION statements for the specified time interval as measured from initial discovery or until the reactor is placed in an OPERATIONAL CONDITION in which the specification is not applicable.
F.xceptions to these requirements shall he stated in the individual specifications.
Entry into an OPERATIONAL CONDITION or other specified applicability 3.n.4 state shall not be made unisss the conditions of the Limiting Condition for reliance on provisions contained in the ACTION without Operation are met statements unless otherwise excepted. This provision shall not prevent passage thrnugh OPERATIONAL CONDITIONS required to comply with ACTION requirements.
1.0.5 When a system, subsystem, train, component, or device is determined to inoperable solely because its emergency power source is inoperable, or besolely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided:
(1) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system (s),
subsystems (s), train (s), component (s), and device (s) are OPERABLE, or likewise Unless both conditions (1) satisfy the requirements of this specification.
least HOT SHUTDOWN 1
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 SHUTDOWN 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 5.
3/4 0-1 BRtfNSWICK - UNIT 2 RETYPED TECH. SPECS.
Updated Thru. Acend.
Il
EMERGENCY CORE C00 LINO WMut43$
1/ J.. %
HICH PRESSURE COOLANT INJECTION SYSTEM 1/4.5 1 LIMITING CONDITION FOR OPERATION 3.5 1 The High Pressure Conlant Injection (HPCI) system shall be OPERARLE with:
One OPERARLE hig'h pressure coolant injection pump, and a.
suction from the suppression An OPERABLE flow path capable of taking b.
pool and transferring the water to the pressure vessel.
CONDITION.S 1, 2, and 3 with reactor vessel steam done pressure APPLICARILITY:
> 113 psig.
ACTION:
With the HPCI system inoperable, POWER OPERATION may continue a.
the ADS, CSS, and LPCI systems are OPERABLE; restore provided inoperable NPCI system to OPERARLE status within.14 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 SHUTDOWN within the,following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
recuirements if 3cecifica: ice 1.*.*
With -5e surve t ilance the receired frecuencies due to low reactor g:ean performei tt provisions of Soecification 4.0 4 are not pressure, tne applicable provided the ap'propriate surveillance is performed within 58 hours6.712963e-4 days <br />0.0161 hours <br />9.589947e-5 weeks <br />2.2069e-5 months <br /> af ter reactor steam pressure is adequate to perform the tests.
SURVEILI.ANCE Rent!IREME'!TS The MPCI snall be demonstrsted OPERABLE:
a. i.1 s.
At lesst.ince per 31 days hy:
Verifying that the system piping f rom the pump discharge valve to the system isolation valve is filled with water.
s 1/4 5-1 RRUNSWICK - UNIT 2
%wTY b-O]
RITYPED TECH. SPECS.
Updateo Thru. Amend. 78
v,MERCENCY CORE COOLING WMidsda SURVEILLANCE REOUIREMENTS (Ctntinund) each valve (manual, power-operatcd, cr cutomatic)
N 2.
Verifying that locked, sealed,"or otherwise in the flow path that is not secured in position. is in its correct position.
once per 92 days, by verifying that the system develops a At least 4250 gym for a system head corresponding to a 5
1000 psig when steam is being supplied to the flow of'at least reactor pressure >
turbine at 1000, 720, -; 4, psig.
nnce per 19 months by:
c.
At least Performing a system functional test which includes simulated its emergency 1
" automatic actuation of the system throughouteach automatic valve in operating sequence and verifying that position.
Actual the flow path actuates to its correctinto the reactor vessel is excluded from injection of coolant this test.
least 4250 gpm the system develops a flow of at Verifying that for 'a system head corresponding to a reactor pressure of > 165 2
osin when steam in Niing supplied to the turbine at 165,}[15, psia.
for the RPCI system is automatiestly that the suction the suppression Verifying 3.
transferred from the condensate storage tank to pool on a e.ondensate storage tank low water level signal ar suppression pool high water level signal.
3/4 5-2 MRttNSWICK - UNIT 2
".et T5I 84-07 RE~YP C 'l'ECH. SPECS.
Updated Thru. Amen:. 78
LIMITING CONDITION POR OPERATION The Autoestic Depressurization System (ADS) shall be OPERABLE with at 3 5.2 least seven OPERABLE ADS valves.
CONDITIONS 1, 2, and 3 with reactor vessel steam done pressure APPLICABILITY:
> 113 pois.
ACTION:
With one ADS valve inoperable, POWER OPERATION any continue provided the HPCI, CSS, and LPCI systems are OPERABLE; restore the inoperable s.
least HOT 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 />.
With two or more ADS valves inoperable, be in at least HOT SHUTDOWN b.
within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
With the Surveillance Raquirement of Specification 4.5.2.b not c.
performed at the required interval due to low reactor steam pressure.
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 vessel steam pressure is adequate to perform the tests.
SURVEILLANCE REOUIREMENTS V
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 emergency operating a.
sequence, but excluding actual valve actuation.
Manually opening each ADS valve when the reactor steam done pressure b.
is 1100 psig and observing that either; The control valve or bypass valve position responds accos.1...t'.'.
1 or is a corresponding change in the measured steam flow.
2.
There 3/4 5-3 BRUNSWICK - UNIT 2 RETYPED TECH. SPECS.
Updated Thru. Amend.
l'
)
3 / 4. 5.3 LnW PRESSURE COOLING SYSTEMS 1
i CORE SPRAY SYSTEM f
LIMITING CONDITION FOR OPERATION
- 3. 5. 3.1 Two independent Core Spray System (CSS) subsystess shall be OPERA 51.5 with each subsystem comprised of:
One pump, and a.
i b'
An OPERABLE flow path capable of taking suction from at least one of the following OPERABLE sources and transferring the water through the i
Spray 'sparger to the reactor vessel:
1.
In CONDITION 1, 2, or 3, from the suppression pool.
2.
In CONDITION 4 or P:
a)
From the suppression pool, or b)
When the suppression pool is inoperable, from the condensate storage tank :sntaining at less: 15n,666 gallons of water.
APPLICARILITY:
CONDITIONS 1, 2, 3,'4, and 96 7.
ACTION:
a.
In CnNDITION 1, 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 HOT 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
t 2.
With both CSS subsystems inoperable, be in at least HOT SHUT 30RN wt thin.12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
TSe care scrav system is not required to be OPERABLE when the suopression anal in 1,ooersb;e. stovided that the reactor vessel head is removed and
!Se cavity is f'. coded. the spent fuel pool gates ara removed. and the water level is maintained within the limits of Specifications 3.3.8 and
~
3.5. 0 BRUNSWICK - IINIT 2 3/ 4 5-4 RETYPED TECH. SPECS.
L*pdated Thru. A=end.
\\
I e'
EMERGENCY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
_' ACTION _ (Continued) the CSS is actuated and injects 4 ster into the In the event 3.
reactor coolant system, a Special Report shall.be prepared acd submitted to the Commission pursuant.to Specification 6 9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.
b.
In CONDITION 4 or 5*:
With one CSS subsystem inoperable, operation any continue least one LPCI subsystem is OPERABLE within A
- 1.,
provided that at hours ; otherwise, suspend all operations that have a potentis' for draining the reactor vessel.
With-both CSS subsystems inoperable, operation may continue least one LPCI subsystem is OPERABLE and both 2.
provided that at Otherwise,* suspend LPCI subsystems are OPERABLE within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
all operations that have a potential for draining the reactor vessel and verify that at least one !.PCI subsystem is OPERABLE within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
applicable.
3.
The provisions of Specification 3.0.3 are not
~ '
SURVEILLANCE REOUIREMEYr5 Each CSS subsystem shall be demonstrated OPERABLE:
4.5.3.1 once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying the condensate storage tank minimum required volume when the condensate storage tank is required At least a.
to be OPERABLE.
~
1 b.
At least once per 31 days by:
1.
- 1eri f ying that the system piping from the pumo discharge valve l
fo the system isolation valve is filled with water.
required to be OPERA 3LE when the suppression The core spray system is not the reactor vessel head is re soved and pool is inocerable provided that fuel pool gates are removed, and the the cavity is flooded, the. spent water level is maintained within the limits of 3pecifiestions 3.0.8 and 3 9.9.
3/4 5-5 BRUNSUICX, - UNIT 2 RETTPED TECH. SPECS.
Updated 3ru. Amend. 78 1
f 1
EMERGENCY CORE COOLING SYSTEMS
~
SURVEILLANCE REQUIREMENTS (Continu^d)
(;: -
Verifying that each valve (manual, power-operated, or automatic) 2.
in the flow path that is not locked, sealed, or otherwise secured in position,*is in its correct position.
c.
At least once per 92 days
- by:
1.
Verifying that each CSS pump can be started from the :: :: 7 room and develops a flow of at least 4625 gpm on recir:.14::.
flow against a sy. stem head corresponding to a reactor es_ t pressure of 2, L13 psig.
2.
Performing a CHANNEL CALIBRAT10h of the core spray hea:er ;?
instrumentation (E21-dPIS-N004A,5) and verifying the set;cin: :s be 5, +J.5, psid greater than the normal indicated iP.
least once per 18 months by performing a system functional :ss:
d.
At
':.r:;- :::
which includes simulated automatic actuation of the syste:
its emergency operating sequence and verifying that each autoca:::
valve in the flow path actuates to its correct position.
Ac ua_
injection of coolant into the reactor vessel is excluded from - ;<
i test.
1
- The surveillance test recuired by :his '_icense in Ap end;x A, arsgrich I.
4.5.3.1.C.1, regar:ing :ne flow :est of :r.e :::a sprzy 5; s::= :a:
e
[
pos:poned during the current refueling 19:sge ' ?.e load i
- r. ;; 2 nin a
0 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> af ter restorstion of the suporession chamoer to operas".e 3:2:us but in any case no later than November ! 5, '98.
ia i
3RUNSWICK - UNIT 2 3f. 5-6 Ame nd=e n: No. 96 oo l
LOW PRESSURE COOLANT INX4WKtDPJ WWL4,1 LIMITING CONDITION POR OPERATION hm imie,epdent Iow Pressure Coolant Injection (LPCI) subsystems of
~
3.5.1.2 the tesidual heat removal system shall be OPERAB12 with each subsystes comprised of:
s a.
?.o pumps,.
An 0*ERABLE flow path capable of taking suction from the suppression b.
and transferring the water to the reactor pressure vessel.
poe.
APPLICABILITY: CONDITIONS 1, 2, 3, 4*, and 5*.
ACTION:
In CONDITION 1, 2. or 3:
a.
~41th one LPCI subsystem or one LPCI pump inoperable POVER OPERATION may continue provided both CSS subsystems are OPERABLE; restore the inoperable LPCI subsystem or pump to least NOT SHUTDOWN 1P'.*ABLE status within 7 days 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 SHUTDOWN within the within the next f,13 wing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
least HOT With both LPCI subsystems inoperable, be in at 2.
(-
SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COTE SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
removed With the LPCI system cross-tie valve open or power not 1.
f rom the valve operator, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
the ECCS is actuated and injects water into the N
4.
In the event Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and 9
the total accumulated actuation cycles to date.
l In CONDITION 4* or 5* with one or more LPCI subsystems inoperab. e, take the ACTION required by Specificatrion 3.5.3.J.
The provisions of S.
Specification 3.0.3 are not applicable.
ipplicable when two CSS subsystems are OPERABLF per Specification l
- Not 1.5.1.1.
1 q
3/i 5-1 BRUNSWICK - UNIT 2 RETYPED TECH. SPECS.
Updated Thru. Amend.
SURVEILLANCF, REOUIICMENTS
(,
4 5 3.2 Each 12CI subsystem shall be demonstrated OPERABIA:
Ac' least once per 31s days by:
a.
)
1.
Terifying that the syt, tem piping from the pump discharge valve to the system isolatisa valve is filled ut--
.acer, l
2.
Verifying that each valve (manual, power -operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct no- ::en, and i
3.
Verifying that the subsystem cross-tie valve is closed with power removed from the valve operator.
b.
At least once per 92 days by verifying each pair of LPCI pumps discharging to a common header can be started from the control room and develope a total flow of at least 17,000 gym against a system head corresponding to a reactor vessel pressure of 120 psig.
c.
At least once per 18 months by performing.i 3estas functional test which includes simulated automatic actuatica
- e. ' the system throughout its emergency operating sequence and ver :: -
at each automatic valve in the flow path actuates to its <::.r
-- so=1: ton.
Actual injection of coolant into the reactor vetsn ir su bded from r.bi 4 f~
test.-
4 RRUNS'JICK - UNIT 2 3/4 5-8 kriYPED TECH. SPECS.
Updated Thru. Amend. 7
3/4 5.4
$UPPRESSins eyn LIMITING CONDITION POR OPERATION The suppression pool shall be OPERABLE with a minimum water level >-31 354
~
& aches except the suppression pool any be. inoperab._le.:_.,,.. _.
In OPERATIONAL CONDITT19 4,,provided that:
a.
1 The reactor,od - rei:ch 'is locked in the Shutdown position, and The core spray system is OPERABLE per Specification 3.5.3.1 with 2.
an OPERABLE f'.aw path capable of taking suction from the OPERABLE condensate storage tank and transferring the water through the spray sparger to the reactor vessel.
b.
In OPERATIONAI. C'"":'"'.0N 5, provided that:
The reactor mode switch is locked in the Refuel position, and 1.
The core spray resta s is OPERABLE per Specification 3.5.3.1 with 2.
an OPERABLE P_;e path capable of taking suction frote the OPERABLE -onc.n.:e storage tank, and transferring the. rater through the spray sparger to the reactor vessel, or The reactor vessel head is removed and the cavity is flooded.
3 fuel pool gates are removed, and the water level ts the spent C. -'
maintained within the limits of Specifications 3.9.8 and 3.9.9.
APPLICAR[LITY:
CONDITIONS 1, 2, 3, A, and 5.
ACTION:
In CONDITIONS 1, 2, or 3 with the water level less than the abo se restore the water level to within the limit within I hour or a.
- limit, 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 SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
In CONDITIONS 4 or 5, with the suppression pool inoperable and che h.
satisfied, suspend all operations in the reactor above ennditions not ha
- ressel, all positive reactivity changes, and all operations that a potential f or draining the reactor vessel.
The provisions of Specification 3 0.3 are not acolicable.
1 3/4 5-9 BRUNSWICK - UNIT 2 RETYPED TECH. SPECS.
Updated Thru. Amend. '
r I
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS
.4.5.4.1 The suppression pool.shaL1 be determined OPERABLE by_ verifying ~the water level to be within the limit' et least once per 12 ' hours.
'~
4.5.4.2 The above conditions shall'he verified to be satisfied prior to making the suppression pool inoperable 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 until the suppression pool is restored to OPERABLE status.
G s
I
/
i
)
4 l
RETYPED TECH. SPECS.
Updated Thru. A=end.
BRUSSWICK - UNIT 2 3/
5-19 f
I CONTAINMENT SYSTEMS 3/4.6.'1 DEPRESS!RIZATION SYSTEMS SIPPRESSION CRAMBER N V
"'t
- c r g -!
_7
^ ' " '
LIMITING CONDITION FOR OPERATION 3.6.2.1 The suppression chamber shall be OPERABLE with:
a.
The pool water:
1.
Volume between 87,600 ft aad 89,600 ft3, equivalent to a level 3
(
between -27 inches and -31 inches, and a 2.
Maximum average temperature of 95'T during OPERATIONAL CONDITION 1 or 2, except that the maximum average temperature any be permitted to increase to:
a) 105'F during testing which adds heat to the suppression chamber.
b) 110*F wi:5 THERMAL POWER less than or ecual to 1% of RATED THER.'fAI. POWE?..
c) 1^0'7 wi:h :he :ain s:ca: '.ine isolatien valves closed
(
fo'.lowin; a scram.
Two OPERABLE suppression chamber water temperature instrumentation b.
channels wi:h a minimum of 11 operable RTD inputs per channel.
A :stal leakage from the drywell to the suppression chamber of l'ess c.
than the equivalent leakage through a 1-inch diameter orifice at a
differential pressure of 1 psig.
APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2 and 3.
5 ACTION:
With the suppression chamber water level outside the above limits, a.
restore the water level to within the limits within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in at least HOT SEUTDOWN 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 SEUTDOWN witnin the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
In 'PERATIONAL CONDITION 1 or 2 with the suppression cha ber average water temperature greater than 95'F, restore the average temperature
- o less than or equal to 95'F 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 least HOT f.C!DC".C; sithin 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 SEC~00WN wi:hi the
... :.n; 2-acurs, excep:, as persi::4c abc"e:
T l?'J*.22
'?"!!
3.'.
..a 2,
, e en: ;n, ;.. ]
r CONTAIWENT SYST2MS LIMITING CONDITIONS FOR OPERATION 4Co'tinued)
~
ACTION:
(Continued)
.. -; re
-y n e.
s,
1.~'-Withfthe suppression-M'bhfdh' adds heatr:arerege water Jesperature than 105'? during testing t'o ths suppression chamber, stap all testips which adds heat to the suppression chamber and eastore the 'everage temperature to ides than or least NOT SHUTDOWN equal to 95'F 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 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 SHUTD0WN within the following 2; ours.
2.
With the suppression chamoer average water tecperature graater least one than 11J'F manually scram the reactor and operate at residual heat removal loop in the suppression pool cooling mode.-
3.
With tne.;;pression chamber average water temperature greater than 120'7, depressurise the reactor pressure vessel to less than 200 psig within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
With one supprass on chamber water temperature instrumentation channel inopertele, restore the inoperable channel to OPERABLE status c.
within 7 das -r verify suppression chamber water temperature to be
.ithin tha 11;_:; at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
With both supprassion chamber water temperature instrumentation
(.
d.
least one inoperaole temperature channels inoperaole, restore at instrumentation channel to OPERA 3LE status within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> or be in at least HOT 3H';!;'.'a5 sitnin the next ~ 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and is COLD SHUTD0'w3 within the f:.;: Wing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
With the crywe._-to-suppression chamber bypass leakage in excess of e.
the limit, restore the bypass leakage to within ene limit prior to increasing reactor coolant temperature above 212*T. '
4 SURVEILLANCE REQUIREMENTS
- 4. 6.2.1 The suppression chamber shall be demonstrated OPERASLE:
the By verifying the suppression :nancer water *.*o* u:a ta he sic..
a.
limits at least :nca ;er la hours.
e d
I".
k))] [g[{
e d ows See
.I e
L.
CONTAINMENT SYSIEMS
$URVEILRNCE REQUIREMENTS (Co*tinued)
. n.; a.
~-
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in OFERATIONAL CONDITION 1 or 2 by b.
verifying the suppression chamber average.
esperature to be i
- less them w ;equa14e -95'F, essepts.UP*gA -
eMM m '
)
g"+ ** -
e.
W 6.
7 r:p;pe,.: P.M.5.my. ;
-c
-l I
1.
At least once per 5 minytes during testing wh:.:n ist to
. amber the suppression chamber..by verifying the suppres.. -
^
105'F.
average water temperature to be less chan or ee-2.
At least once per hour when suppression cham er average water temperature is greater than 95'F, by verifying:
a)
Suppression chamber average water tempera.re to be less than or equal to 110*F. and b)
THERMAL POWER to be less than or equal ::.'.1 EATED THERMAL POWER afcar suppression chamber 2rer ;2 water temperature has exceeded 95'F for more than 2:* hours.
3.
At least once per.30 minutes following a scra:
suppression 7
oy chamber average water temperature greater :.;.
i verifying suppression chamber average water :s..a sture less l
chan or equal ; '.'0 7.
By en external risua*. ex::inatica :f selseted e:2 r
- -
- . :coling h
c.
anciosure system suction line penetrations of the suppress:.;n cas.::e.-
prior to taking the peactor f rom COLD SHUTiXNN af ter.. ::ye relief
. :. ; n valve operation wi h :he suppression chamber averan temperature greater :han or equal :c L%'F and rea:: -
- r. _anc system pressure greater than 200 psig.
- i 2:ure d.
By verifying at least two suppression chamber wate:
i.
instrumentation channels OPERABLF. by perf or:ance :-
1.
CHANNEL CHECK at least once par ;* nours.
4 CHANNEL FUNCTIONAL TEST at least once per 31 days, and 2.
-ar 18 months (550 days).
3.
CHA.VNE:. 0AL 32AMON a: least sco with the temperature alars setpoint for *.17. watar temperature less than 2r squal :o 95'F.
(CAC-TE-4'+2 6-2 thr
'3: CAC-IY-4420-1; CAC :t-4 ;o.)
( CAO-U.-4 ' 2 6-L 5 thr : 2 *:
- .:-T
.-11.: :AC-3 -- 26-2) 9 i
e.
Ac *. east once per 13 months by:
l 3.: : t e.
1 risusi taspection af
- ..4 4
n.: rsu L:: ::a::ur ini,:
A. *,. ; E 'l
- 2.
t i
i 3?. : SN:;;;
'2.;;; 2
- ;. o.,,-;
.,, n 3,,3
- o,
,,3 i
I
.. -. _ _ - -. _ _. _ - - _ _ _ ~ -. _. -.. _... _.. - - - -. _.. - _.. _. _ _.. -.. _.. _... _ _ -, - -.... - - - -..
r CONTAI.NMENT SYSTE.4S s
l SURVEILLANCE REQUIREMENTS (Continued) 2.'
Conducting a dryvell-to-suppression chamber bypass leak test at f
an initial differsatial pressure of 1 poig f
h
--the differential pressura does.ast lacrease,and veri ying t a
- 7.g. %.._.--- - f..
1
- eg,, jt.
.y j
O.25 inches of water per minute'Ter a to minute " period.
~~
l l
l f
4
... ~ ~
3 L4 e. ;t ue".c ent 'ic.
r LIMITING CONDITION FOR OPERATION
<r.
i.k.2.2 The supprisch hN EoIe'of the residual heat eveoval (RHR) wysten shall be OPERABLg with two independent cooling loops, each loop consisting heat exchanger.
q _.c m.
l N~'~~",
APPCICAstd wnn non.'
T add 3.~"
4 g
ACTION:
With ona RRR suppression pool cooling loop inoperable, operation may a.
l continue and the provisions of Specification 3.0.4 are not applicable; resto're th,e inoperable loop to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and ir.
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 />.
b.
With both,RHR suppression pool cooling loops inoperable, restore at Lea'st one loop to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 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 SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
1 1
1 SURVEILLANCE REQUIREMENTS 4.6.2.2 The suppression pool cooling mode of the RHR systes shall be demonstrated OPERABLE:
I a.
At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
I b.
At least once per 92 days by verifying that each RHR pump can be started f rom the control room and develops a flow of at least 'U.300 gpm against a system head corresponding to a reactor pressure of
> 20 psig on recirculation flow.
BRUNSWICK - UNIT 2 3/4 6-11 RETYPED TECH. SPECS.
Updated Thru. Amend. 7
\\
i u) l
CONIADMENT STSIEMS r
- CDNEAN88F ATESPERI SIISTION SYSTEM ~
. g ;%gMffggggg:-
LIMITING CONDI!!0N FOR OPERATION
..mm msamt m,r
- g.--m was,. a A w.' gam uer.nas w.cnarwe, a w xr with:
s, An OPERA 33 flow path capable bf supplying nitroget to the drywell, a.
sad b.
A minimum supply of 4350 gallons of liquid nitrogen..
AFFLICA.3ILITY_:
CONDITION 1*.
ACTION:
.-y
.y With the CA.D system inoperable, restore the CAD system to OP11ABM status ;
within 31 days er be in at least frARIUP within the next g hours. The provisions of Specification 3.0.4 are not applicable.
5;;IVII:.* A.NC RICUL't1ME:ES 4.6.6.2 1ha CAD system shall be demonstrated to be OPERA 3LE:
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, er autonatic) in the flow path not locked, sealed, or otherwise secured in position, is in its correct position.
e b.
At least once par 18 months by:
1.
Cycling each power-operated (excluding automatic) valve in the flow path through at least one complete cycle of full travel, l
and 2.
Verif ying that each autonatic valve in the flov,.th actuates to its correct position on a Group 2 and 6 isolaticn test signal.
- 'inen oxygen concentration is required :o be < 4:; per Scecifitati:n 3.e.6.3.
BR::NS*JICK - UNC 2 3/4 6-25 A=endnent :;o. 35 4
LIMITING CONDITION FOR OPERATION O g/.
3 6 6.3* The primary containment steosphere oxygen concentraties shall.be, lass, ;1 M
u%
than 41 by volums 'during the period fromt u._
a.u..
r
' + ' '. a.
- a...Jtfithis?a61 tours after..tnInxAL PoWEn.E ts: ef RATED TWtRMAL POWER,,tej.g S,+
M.d.wse#M 17,. w %.i.tWry,wirewr : T. ' A a w, / g / p % q c.y F C,:W.;ggy
~
b.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to's scheduled reduction 'of THERMAL POWER to <
-T 15: of RATED THERMAL POWER.\\
d?LICABILITY:
CONDITION 1.
ACTION.
With che oxygen concentration in the primary containment exceeding the limit,
5e 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 />.
,, l:'
.".;_,, 4 +
- "[
^'
SURVE1LLANCE REOUIREMENTS
.6 6.3 The oxygen concentration in the primary containment shall be verf**:'
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 THERMAL POWER and at least once per 7 days thereaf ter.
h l
- For the period commencing at 0630 on June 29, 1981, a temporary exemption is l
allowed to operate BSEP-2 in Condition 1 with containment oxygen concentration l
exceeding 4% by volume for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
I O
i l
1 l
BRUNSWICK - UNIT 2 3/4 6-29 RETYPED TECH. SPECS.
Updated Thru. Amend.
u
ELECTRICAL POWER SYSTEMS O-LIMITINGODNh1TIONFOROPERATION
~ ~ - - - - -
3.8.1.2 As a miniana, the following A.C. electrical power sources shall be OPERABLE:
One circuit per Unit betwee. the offsite transmission network and a.
the onsite Class 1E distribution system, and b.
Two diesel generators, as & mired to operate ECCS systems in accordance with Specifications 3.5 3.1 and 3.5.3.2:
1.
Each with a separate:
a) Engine-counted fuel tank containing a miniaua of 100 gallons of fuel, b) Day fuel tank containing a minimum of 22,650 gallons of fuel, and c)
Fuel transfer pump.
2.
With a fuel storage tank containing a miniaua of 37,000 gallons of fuel..
AFFLICABILITY: CONDITIONS 4 and 5.
ACTION:
With less than the above minimum required A.C. electrical power sources OPERABLE, suspend all operations involving irradiated fuel handling, CORE ALTERATIONS, positive reactivity changes, or operations that have the
~
potential of draining the reactor vessel.
The provisions of Specification 9
3.0.3 are not applicable.
SURVEILLANCE REOUIREMENTS 4.8.1.2 The above required A.C. electrical power sources shall be demonstrated OPERABLE per surveillance requirements of Specifications 4.8.1 1.1 and 4.8.1 1 2, except for the requirement of 1.8.1.1 2.a.5 BRUNSWICK - UNIT 2 3 ; '. 8-5 RETYPED TECH. SPECS.
Updated Thru. Amend. 7
~.
ELECTRICAL POWER SYSTEMS A.C. DISTRIBUTION - SHUTDOWN OF SOTH UNITS y',. + g,
s ;,y q,
~
LIMITING CONDITION FOR OPERATION
_.~
~
3.'A.2.' 2 'As
- daisum, owing A.C.
t ca Eses"shall ~ be OPERABLE ~" ~
for each unit but aligned to an OPERAB12 diesel generator, as required to operate ECCS systems in accordance witfi. Specifications 3.5.3.1 and 3 5 3.2:
1 - 4160-vole Energency Bus, 1 - 480-volt Emergency. Bus, 2 - 120-volt A.C. Vital Buses.
AFFLICABILITY: CONDITIONS 4 and 5.
~
ACTIOE:
With less than the above complement of A.C. buses CPE?$.3LE, suspend all operat' ions involving. irradiated fual handling, ~ COT.2.C27.ATIONS, positive reactivity changes, or operations that have the pocannal of draining the reactor vessel. The provisions of Specification 3.0.; are not applicable.
i 1
SURVEILLANCE REOUIREMENTS least once 4.8.2.f The specified A.C. buses shall be decemined OPERABLE at E
and indicated power per 7 days by verifying correct, breaker alignment availabil,ity.
i t
(
BRUNSICK - IJNIT T.
3/1 8-7 RETYPED)TEQl. SPECS.
Updated Thru. Amend. :
r ANSWERS - SECTION 5 5.1 d (1.0)
REF:
EIH:
L-RQ-602 (9)
BSEP:
L/P 02-2/3-A, pp 54, 83 5.2 c (1.0)
REF:
EIH:
GPNT, Vol. II, Chapter 3.E; Nuclear Power Reactor Instrumentation Systems Handbook, Harrer & Beckerly, p 44.
BSEP:
L/P 25-2/3-C, pp 3, 4 5'. 3 a.
By observing the Full-in and Full-out travel lights (the operator could determine if geometric distortion had occurred.
Inability to conduct full' detector movement would indicate that internal misconfiguration had occurred).
(1.0) b.
By observing the neutron level while moving the nuclear instrumentation. A significantly HIGHER (approximately 300 times) count rate would be seen for the UNVOIDED areas of the core as opposed to the VOIDED.
(1.0)
REF:
EIH:
L-RQ-540 (M.8)
BSEP:
L/P 05-2/3-C, p 25; L/P 02-2/3-A 5.4 c (1.0)
REF:
EIH:
GPNT, Vol. VII, Chapter 10.1-78; HNP-2-1001 BSEP:
02-2/3-A, pp 156-159 5.5 b (1.0)
REF:
EIH:
L-RQ-606 (5)
BSEP:
L/P 02-2/3-A, pp 176-177 l
u
r 2
5.6 a.
lambda = Ln 2/T =.693/55.6 =,.0125 sec E -1 1/2 (CALC NOT REQUIRED)
T=1/-lambda = 1/.0125 = - 80 see After the initial prompt drop, power cannot decrease faster than the longest lived delayed neutrons.
(1.0) b.
Shorter [0.5) The initial prompt drop will only be due to prompt neutrons [0.5) -0R-Decay of short lived precursors
[0.5]
(1.0)
REF:
EIH: GPNT, Vol. VII, Chapter 10.1 BSEP:
L/P 02-2/3-A, pp 122, 134, Fig 51 5.7 a.
(2) - Starting Pump 2 (0.5) b.
Curve.B (0.5) c.
INCREASE (0.5) d.
Curve C (0.5)
REF:
Pump Laws 5.8 d (1.0)
REF:
Mollier Diagram (Steam Tables) 5.9 b (1.0)
REF:
EIH: GPNT, Vol. VII, Chapter 10.1-83-86 BSEP:
L/P 02-2/3-A, pp 172-176 5.10 b (1.0)
REF:
EIH: Therrsedynamics L/P, pp 52-56 BSEP:
L/P 04-2/3-E, pp 66-78 5.11 c (1.0)
REF:
Steam Tables t
}
3 5.12 75% CONTROL R0D DENSITY (0.5)
EIH: (The increased Control Rod Density causes greater competition for the thermal neutrons; this necessitates greater pin power for thi same net power output.)
Higher pin power results in a greater Void Fraction which causes a more negative coefficient.
J BSEP: With a greater rod density, a greater number of i
neutrons are " lost" to the control rods (increased leakage).
Thus, a change in rod density affects reactivity more, by allowing increased absorbtion by other fuel bundles.
(Can also explain why low rod density does not have a large reactivity effect, since the leakage to other fuel bundles is already so large.)
(0,5)
REF:
EIH:
Reactor Physics L/P, pp 1._7-9, 10, & 13.
BSEP:
L/P 02-2/3-A, pp 141-143 5.13 a.
2 (1.0) b.
2 (1.0)
REF:
General Electric NEDE 21493 (Rev. 5)
EIH:
GPNT, STA Training Manual, Section 9 BSEP:
L/P 06-2/3-B, p 1-16 5.14 a.
INCREASE b.
INCREASE c.
DECREASE d.
-4EMM#4HIHME-
'Id'M*58 ", 3 se %:% km (0.5 ea) tw ai ws. A u =,. l oo, o..
'e I w a stwo.,., N h e.,, %
T., m. n.cs:,,,
RE F : "d " " "" * " oF W is N w, A, c
.s cular EIH: GPNT, Vol. VII, Chaper 10.2 BSEP: GPC, " Heat Transfer, Thermodynamics, Fluid Flow",
pp 235-241 5.15 a (1.0)
REF:
BSEP:
L/P 04-2/3-B, pp 72-75 5.16 a.
Due to Increased voiding [.25] and flow escaping from the riser to the annulus [.25].
b.
Due to reduced core flow.
c.
Follows steam flow (as EHC closes CVs to control pressure).
d.
Follows steam flow decrease [.25] and level increase [.25].
e.
Follows power reduction.
(0.5 ea.)
REF:
BSEP:
Transient L/P, Transient HXY-7
r 4
5.17 1.
" Swell" due to increased voiding (from pressure decrease).
b.
Due to, increased voiding.
c.
SCRAM due to turbine trip, d.
GROUP I Isolation on low steam pressure (in RUN Mode).
'e.
Effects of HPCI (0.4) and RCIC (0.1) injection.
(0.5 ea.)
REF:
BSEP: Transient L/P, Transient HXY-7 5.18 b (1.0)
REF:
General Electric, NEDE-24810 (Jun 81) s k -
r ANSWERS - SECTION 6 6.1 c (1.0)
Ref:
BSEP SD-03 Fig. 3-3 6.2 See attached trace - Figure 6.2A (1.0)
Ref:
BSEP RTN 2A, Plant Mod 81-291; GE SIL 200 (Suppl. 2) 6.3 b (1.0)
Ref:
BSEP SSM 10-3-A, Figure 5 6.4 a.
Increase (50%) [0.5]; Dual / Master limiter [0.5]
(1.0) b.
Increase [0.5]; scoop tube positioning unit (full range, LS/ mech. stop) [0.5]
(1.0)
Ref:
BSEP SSM 10-3-A, RTN 010 6.5 a.
03 b.
04 22-55 18-03 46-55 46-55 34-27 Blank
(.25 each/2.0)
Ref:
BSEP SSM 27-2-B 6.6 a.
Closed (0.5) b.
As Is (0.5) c.
Closed j
(0.5)
Ref:
BSEP RTN F&C; BSEP NRC Exams 4/83, Requal & 5/84 6.7 a (1.0)
Ref:
BSEP RTN 033, 012; SD 26.2 6.8 - wet (0.5)
- Assymmetrically (0.5)
Ref:
BSEP SSM 25-2-C, RTN 029 6.9 a (1.0)
Ref:
BSEP SSM 25-2-D and 25-2-C, RTN 029 6.10 b (1.0)
Ref:
BSEP SSM 25-2-C m
o 2
6.11 - 50% rod density
(.33)
- to 22% power
(.33) 1
- turbine first stage pressore
(.34)
Ref:
BSEP SSM 27-2-C 6.12 a.
Rec'irc loop flow elements (pump discharge)
(0.5) b.
Only the 120% fixed scram [0.5] because the flow biased scram incorporates a time delay (=6 seconds, representative of fuel temperature transient time) [0.5]
(1.0)
REF: BSEP SSM 25-2-D 6.13 a
.99 (0.5)
Ref:
BSEP SSM 25-2-D 6.14 c (1.0)
Ref:
BSEP GP-03, 50-26.2 6.15 c (1.0)
Ref:
BSEP GP-03 6.16 - RHR pump room cooler (0.5)
- RHR pump seal Hxers (0.5)
- Core Spray Pump Room Coolers (0.5)
Ref:
BSEP SD-43 6.17 b (1.0)
^
Ref:
BSEP RTN 022, SD-19 6.18 c (1.0)
Ref:
BSEP SD-16, 12 6.19 d.
.(1.0)
Ref:
BSEP SSM 14-3-D 6.20 c (1.0)
Ref:
BSEP SSM 14-3-E 6.21 a. - 4 (1.0)
- b. - 3 (1.0)
REF:
BSEP RTN 008, SSM 08-2-A u
r 3
6.22 a (1.0)
Ref: BSEP'RTN 028 6.23 d (1.0)
Ref: BSEP'RTN 026 6.24 - Alternate power source (480/120 VAC transformer)
(0.5)
SSI shuts [.25] and SS2 opens [.25]
(0.5)
Ref: BSEP SSM 20-2-F 9
r l
t
MFit D SCHEM ATC OF WCdlFIED CRO MYDiLMJUC SYSTE.M Figure 6.2 A Cmfo nern.s REACTOR vtSSEL M
1 Omtvt WATER (v6TNCA AW1 4
I O
N I
{ ]
IDRIVE WATEM g
- Yl r
b 123 VALVE I
8 A
~
m (e AA SOUNCAMY Heu vaLvs M
O y
.I A%
k I
N Y
YW Vm 133 121 OIR ECTIONA L 3
CCNTACL SCLENCIO 3
ExMAugT MEAQS VALVI 6
i DNAUST WATEM M_
-M
& CTHEM
)
CTHEM 4 MCLts MCLfs
)
HCtts N
CocuMG WATER p_---________,
i Ik k
EM l
CThER 088VI C:<.
m y
RC WAT.
PCV T
i
-E-DN-m
- acv
- v Osc-m w...m%
c' I
valves & ttice-s l
s
=-_____________ _
_ ___ _ _ g i
u CRo
=
STAaluZ:NG VALVE 3 LEGEMD
& NORMAL PLCW P ATH AEVOSE FLOW CISPER$AL C7 DMAL.:3T l
I i
r o
i ANSWERS - Section 7 7.1 a.
(As the shell warms), first stage pressure inct eases
[0.5]. When shell pressure reaches 155 psig, the reactor scrams on TSV closure at >30% indicated power
[0.5].
(1.0) b.
4,2,1,6,5,3
(.25 ea/1.5) c.
Trip the turbine instantly [0.5] and place on the turning gear.
(0.5)
REF:
- the reactor shall be shutdown (0.5)
- (the reactor SU shall be discontinued until) an assessment can be performed by the NE (and approved by SOS)
(0.5) b.
- control rod inserted to achieve a stable period of >100 seconds (0.5)
REF:
BSEP GP-02 7.3
-2 (0.5)
- Misoperation in the automatic mode is confirmed (0.5)
- Adequate core cooiing is assured (0.5)
- Warming up and pressurizing steamlines (0.5)
- Water hammer in the steam lines (0.5) b.
Operation <2000 RPM should be minimized to ensure adequate oil pressure to operate the governing valve
[0.5] and to prevent exhaust check valve chattering
[0.5].
(1.0)
REF:
BSEP OP-16 7.5 Such operations can shorten seal life.
(0.5) o.
1%, 5%, 10%
(1.5)
REF:
BSEP GP-04, P.4 BSEP GP-05, P.5 u
U f~
2 7.6 a.
To prevent regenerative Hx tube damage [0.5]
due to rapid cooldown rate [0.5].
(1.0) b.
To minimize thermal duty on the feedwater nozzles.
(1.0)
REF:
(1.0)
REF':
BSEP OP-21 7.8 The Fuel Pool gates by themselves provide no shielding for personnel working in the Reactor Cavity Area.
(1.0)
REF:
BSEP GP-07 7.9 Because spraying the drywell may decrease containment pressure below atmospheric at a rate beyond the capacity of the RB-to-suppression chamber vacuum breakers, resulting in negative containment pressures in excess of design.
(1.0)
REF:
REF:
BSEP A0P-02.1 (1.0) 7.11 c REF:
BSEP A0P-04.4 (1.0) 7.12 b REF:
BSEP A0P-30.0 (1.0) 7.13 a.-1 b.-3 c.-4 d.-2 (0.5 ca./2.0)
REF:
BSEP A0P-20.0 7.14 - indications of criticality observed on SRMs
- loss of communications between Control Room and Refuel Floor
- malfunctioning or failure of > one SRM or IRM channel
- accidental dropping or damaging of a fuel element (0.5 ea./2.0)
REF: BSEP FH-11 s_.
r u
o 3
7.15 a.
closed (0.5) b.
1 (or 2) open (0.5) c.
1 RHR loop injecting to Rx
(.25) 1 RHR loop in suppression pool cooling
(.25) d.
107 to 164 psig > suppression chamber pressure (0.5) e.
Rx level to main steam line elevation or >254" (0.5)
REF:
Suppression Po'o1 temperture [.4] > 95*F [.1]
(.5)
DW average temperature [.4] > 135*F [.1]
(.5)
DW pressure [.4] > 2 psig [.1]
(.5)
Suppression Pool level [.4] <-31" [.1]
(.5)
Suppression Pool level [.4] >-27" [.1]
(.5)
(4 req. at.5 ea 2.0)
REF:
Eo u
/
9 ANSWERS - SECTION 8 (1.0) 8.1 c.
REF: BSEP 01-04 8.2 b.
(1.0)
REF:
BSEP 01-04 8.3 d.
(1.0)
REF: BSEP PEP 02.2 8.4 b.
(1.0) s. gewe% p%e-REF: BSEP MP-14 8.5 A departure may change the intent of the proceduce [0.5]
while a deviation may not [0.5].
(1.0)
(1.0)
REF: BSEP 01-01 8.7
- a. - 2 (0.5)
- b. - 4 (0. 5-)
- c. - 1 (0.5)
- d. - 3 (0. 5.)
(1.0)
REF:
10 CFR 55.31(e) 8.9 d.
(.l. 0)
REF: BSEP AI-58 8.10 c.
(1.0)
REF: BSEP AI-58 8.11 c.
(1.0)
REF: BSEP AI-58
r x
=
2 i
\\
1 8.12 -
All control rods capable of insertion are fully inserted except for the analytically determined highest worth rod
[.25] which is assumed to be fully withdrawn [.25]
(0.5)
Cold (68 degrees F)
(0,5)
Xenon-free (0.5) 1 REF: BSEP TS Definitions 8.13 Operational Condition 5 (Refueling)
(0.5)
REF: BSEP TS Table 1.2 8.14 b.
(1.0)
REF:
BSEP TS Definitions 8.15 "T" = lowest value of the ratio'of design TPF divided by the MTPF (obtained for any class of fuel in the core) [0.5]
"T" is applied only if less than or equal to 1.0 [0.5]
(1.0)
(1.0)
REF:
BSEP TSs 3.5.1, 3.5.2, 3.5.3.1 8.17 d.
(1.0)
REF: BSEP TSs 3.0.3,.3.5.1, 3.5.3.2, B 3.0.3 8.18 a.
(1.0)
REF: BSEP TSs 3.6.6.2, 3.6.6.3, 3.0.4 8.19 b.
(1.0)
REF:
BSEP TSs 3.5.4, 3.5.3.1, 3.0.1 8.20 a.
(1.0)
REF:
r
.e ip.
3 4
8.21 b.
(1.0)
REF: BSEP 01-05 8.22 Thermal expansion (of valve internals on heat up) may cause valve binding and/or damage.
(1.0)
REF: BSEP 01-13 8.23 - No Pressure Boundary Leakage
(.5)
- 5 gpm Unidentified leakage [.4] averaged over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period [.1]
(.5)
- 25 gpm total leakage [.4] averaged over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period [.1]
(.5)
- 2 gpm increase in Unidentified leakage [.4] within any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period [.1]
(.5)
- Total leakage exceeded 25 gpm over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (25.50)
(.5)
- Unidentified Leakage increased from 2.52 gpm at 00-04 to a 4.58 gpm and 5.2 gpm rate on shifts 16-20 and 20-24 respectively, thus exceeding 2 gpm increase limit.
(.5)
REF:
BSEP TS 3.4.3.2 h_