ML20137X770
| ML20137X770 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 02/24/1986 |
| From: | Elin J, Morrill P, Obrien J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20137X767 | List: |
| References | |
| 50-344-OL-86-01, 50-344-OL-86-1, NUDOCS 8603120010 | |
| Download: ML20137X770 (82) | |
Text
.
U. S. NUCLEAR REGULATORY COMMISSION REGION V Examination Report No: 50-344/0L-86-01 Facility:
Trojan Nuclear Plant Docket No:
50-344 Examinations administered at Trojan Nuclear Plant, Rainier, Oregon Chief Examiner:
. /l4(
d'2 b
P. O'Brien, Operator Licensing Examiner Date Signed OO I~N Examiner:
'P. Jl Morrill, Operator Licensing Examiner Date Signed 2 Ai ~
Approved:
[.Elin, Chief,OperationsSection Date Signed Summary:
Examinations on February 11-12, 1986 (Report No. 50-344/0L-86-01)
Written and oral examinations were administered to two Senior Reactor Operator candidates. Both candidates passed all their exams.
h4 P
o V
e REPORT DETAILS 1.
' Examiners:
J. P. O'Brien, RV (Lead Examiner)
P. J. Morrill, RV 2.
Persons attending the exit meetings:
J. P. O'Brien, RV P. J. Morrill, RV W. S. Orser, Plant Manager R. P. Schmitt, Manager of Operations and Maintenance Bud Susee, Operations Supervisor R. L. Russell Assistant Operations Supervisor D. R. Keuter, Manager of Technical Service Steve Nichols, Training Supervisor
- u. W. Ellis, Training Specialist M. C. Peterson. Training Specialist J. A. Bauer, Onsite Regulation Engineer 3.
Written Examination and Facility Review:
Written examinations were administered to two SRO candidates on February 11, 1986.
At the conclusion of the written exam, the facility staff reviewed the exams. The comments made by the staff at the conclusion of the review are included in the attachment (1). These comments were discussed with the staff and appropriate changes were made to the exams prior to the grading of the exams.
4.
Operating Examinations:
OralexaminationsandplantwalkthroughsofthetwoSROcandi!ateswere conducted on February 12, 1986.
No general weaknesses of the training program were noted by the examiners, 5.
Exit Meeting:
On February 12, 1986, the examiners met with the licensee representatives listed in paragraph 2.
Those individuals who clearly passed the operating exam were identified in this meeting.
Further, the changes to the examination standards and examination procedures, that might affect the licensee, and that have been implemented since their last exam were discussed.
1 l
i Resolution of Facility Comments of the February 11, 1986 Trojan Senior Operator Exam Enclosed on the next page are the Facility's Comments on the SRO exam.
Below are the resolutions to these comments:
Question:
Resolution:
5.02 Key did not give point breakdown for the individual areas. Those mentioned by the facility are not part of the what the author had intended for credit. Key amended to indicated credit items and to consider facility comments.
5.03 e.- Accepted, f.- Accepted.
5.07 Wording of the question was awkward, but intent of the question is clear.
Question noted if used for the future.
5.08 Key corrected.
5.10 Pts. distributed:
a.1 - (1.5) a.2 - (0,5) 5.12 Key corrected.
6.01 Key corrected.
6.05 Based on RDC 83-042 review. key corrected to reflect
'b' as the correct response.
l 6.08
- b. - was purposely done as a distractor.
- c. - is typo and corrected.
6.11 Either acceptable, key corrected.
6.12 Key corrected.
7.04
- a. - not accepted.
Inadequate justification,
- b. - comment is noted, no change to the key is required.
8.06 Comment noted. No change to the key required.
8.09 Author tried to make it clear this exception was for normal operations.
Since the complementary exception for emergency procedures was used as a distractor, key was corrected to accept either (b) or (d).
. - ~ -
pimznt i.
REVIEW C0bHENTS FOR SRO EXA'I GIVEN AT *Ilt0JAN ON FEBRUARY 11, 1986, Question Comments 5.02
- Key specifies new operating points, but question does not ask for them.
- Question does not specify to consider each case separately.
Credit should be given for either approach candidate takes.
5.03 e.
Should also allow " film boiling" as answer.
f.
"No" should be an acceptable answer.
5.07 Question asks for a "what" and a "why", but. only one answer is given.
Key is acceptable, question should be modified in future.
Peakforxenonis"2.35%AN/K",not"1.75%AK/K"asspecifiedon 5.08 a.
key.
b.
Generally, we neglect the direct fission yield of Samarium and assume it is all from decay of Promethium. Should delete or redistribute points for answer.
5.10 Since excore NIS detects mostly fast neutron leakage, and boron is a 1/v absorber, lower boron density should have minor effect on NIS reading. Recommend point distribution be changed to emphasize part 1.
5.12 Keyshouldsayieff or Beta Bar Effective instead of Beta Effective.
6.01
- Gemical injection designed as stated but never used in this manner. Key should be reworded as follows: "Gemistry control is provided by batch chemical injection into main feed line and flushed into S/G's using AIW pumps".
- Chemicals used were not asked for.
6.05 (b) is best answer.
Equipment was modified by design change RDC-83-042 (description attached).
6.08 Key is okay.
For the question, on part (b), M0-8811A/B are on the suction of the IUIR pump from the Recirculation Sump, not on the discharge. On part (c), the correct valve numbers are M0-8807A/B not M0-8870A/B.
6.11 Candidates may use monitor number vice noun names.
1.
PERM-9 2.
PERM's - 7/8 3.
PERM-10 4.
PERM-13 5.
PERM-15 6.
PERM-14
Question Comments 6.12 a.-
We normally consider answers 2 and 3 as one arming signal (acceptable answer should be greater than 10% power decrease in less than 2 minutes OR load rejection). Since the question asked for 4 responses, the candidates may list some of the permissives and if correct, credit should be given. Permissives are:
1.
In-Lo TAVG 2.
Condenser Available (vacuum greater than 25" Hg and at least one circulating water pump running).
3.
Io-Lo S/G Level 4.
Bypass Interlock Switches in ON position.
7.04 a.
We disagree that knowing the 10 minute time limit for determining emergency classification is necessary.
b.. Although the Fire Brigade leader is nobnally the Assistant Shift Supervisor, any qualified (icensed operator can be assigned the position.
Key is the expected answer, but credit should not be lost if above is stated.
8.06 Question is poorly worded. Also this question asks essentially the same as question 7.01.a.
8.08
- and note should not be required as part of answer.
8.09 (b) or (d) should be accepted as correct answer.
p hm:nt 2 REVIEW C0bofENTS FOR SRO EXA>l GIVEN AT 1ROJAN ON FEBRUARY 11, 1986 Question
_ Comments 5.02
- Key specifies new operating points, but question does not ask for them.
- Question does not specify to consider each case separately.
Credit should be given for either approach candidate takes.
5.03 e.
Should also allow " film boiling" as answer.
f.
"No" should be an acceptable answer.
5.07 Question asks for a "what" and a "why", but. only one answer is given.
Key is acceptable, question should be modified in future.
5.08 a.
Peak for xenon is "2.35% ok/K", not "1.75% AK/K" as specified on
- key, b.
Generally, we neglect the direct fission yield of Samarium and assume it is all from decay of Promethium.
Should delete or redistribute points for answer.
5.10 Since excore NIS detects mostly fast neutron leakage, and boron is a 1/v absorber, lower boron density should have minor effect on NIS reading. Recommend point distribution be changed to emphasize part 1.
5.12 Keyshouldsayleft or Beta Bar Effective instead of Beta Effective.
6.01
- Chemical injection designed as stated but never used in this manner. Key should be reworded as follows:
" Chemistry control is provided by batch chemical injection into main feed line and flushed into S/G's using AIV pumps".
- Chemicals used were not asked for.
6.05 (b) is best answer.
Equipment was mndified by design change RDC-83-042 (description attached).
6.08 Key is okay.
For the question, on part (b), )f0-8811 A/B are on the suction of the Ri!R pump from the Recirculation Sump, not on the discharge.
On part (c), the correct valve numbers are 510-8807A/B not h!0-8870A/B.
6.11 Candidates may use monitor number vice noun names.
1.
PER>f-9 2.
PERhi's - 7/8 3.
PER>f-10 4.
PERhi-13 5.
PEP 3f-15 6.
PEPJf-14
Quistion Comments 6.12 a.
We normally consider answers 2 and 3 as one arming signal (acceptable answer should be greater than 10% power decrease in less than 2 minutes OR load rejection).
Since the question asked
. for 4 responses, the candidates may list some of the. permissives and if correct, credit should be given.
Permissives are:
1.
Io-Lo TAVG 2.
Condenser Available (vacuum greater than 25" Hg and at least one circulating water pump running).
3.
14-1.0 S/G Level 4.
Bypass Interlock Switches in ON position.
7.04 a.
We disagree that knowing the 10 minute time limit for determining emergency classification is necessary.
b.. Although the Fire Brigade Leader is normally the Assistant Shift Supervisor, any qualified zicensed operator can be assigned the position.
Key is the expected answer, but credit should not be lost if above is stated.
8.06 Question is poorly worded. Also this question asks essentially the same as question 7.01.a.
8.08
- and note should not be required as part of answer.
8.09 (b) or (d) should be accepted as correct answer.
l Mf 5 TGll Ke U.S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION Facility:
TROCAN NPGS Reactor Type: PWR WEC-4 Date Administered:
December 10. 1985 Examiner:
Johnston / o'erten Candidate:
INSTRUCTIONS TO CANDIDATE:
Use separate paper for the answers. Write answers on one side only.
Staple question sheet on top of the answer sheets.
Points for each question are indi-cated in parentheses after the question.
The passing grade requires at least 70% in each category and a final grade of at least 80%.
Examination papers will be picked up six (6) hours after the examination starts.
% of Category
% of Candidate's Category Value Total Score Value Category 25.0 25.0 5.
Theory of Nuclear Power Plant Operation, Fluids, and Thermo-dynamics 25.0 25.0 6.
Plant Systems Design, Control, and Instrumentation 25.0 25.0 7.
Procedures - Normal, Abnormal. Emargency, and Radiological Control 25.0 25.0 8.
Administrative Pro-cedures, Conditions, and Limitations 100.0 Totals Final Grade All work done on this examination is my own, I have neither given nor received aid.
Candidate's Signature s.
s' EQUATION SHEET f = ma v = s/t k (o t)
Cycle efficiency = N 2
w = ag a = v,t + hat E = aC a = (vf - y )/t
~E A = A,'e A = AN g = v, + at KE = my v
PE = agh a = e/c A = In 2/tg = 0.693/tg
} " (t )(tu)
W = vaP q
h*
AE = 931Am (t +t) b 8
Q = pC AT y,7,-Ix p
6=UAAT y,7,-ux
" "f "
I=I 10"*
- SUR(t)'
TVL = 1.3/u P=P 10 e /T HVL = 0.693/u t
P=P SUR = 26.06/T SCR = S/(1 - Keff)
T = 1.44 DT (A'gg )j CR, = S/(1 - K,gg )
o SUR = 26 g,
eff}l - CR Cl ~ Eeff)2
'l (
~
2 T " '(1*/o ) +
[(i ' o)/A,gg ]
o 7 = 1*/ (o - p; M = 1/(1 - K,ff) = CR /CR0 g
~
~#
eff M = (1 - K,gg)0 II ~ Esff)1
- " (E
~l}lEetf " OEefflYeff SDM = (1 - K,gg)/K,g, etf
[1*/TK,gg ] + [E/(1 + A,gg )]
1* = 1 x 10 ' seconds
~
T p=
-I P = I(V/(3 x 1010)
A,gg = 0.1 seconds I = No Id =1d g3 22 WATER PARAMETERS Id =10 g
2 1 gal. - 8.345 lbm R/hr = (0.5 CE)/d (meters)
I gal. = 3.78 liters R/hr = 6 CE/d (feet) 1 ft3 = 7.48 gal.
MISCEl.LANEOUS CONVERSIONS 3
10 Density = 62.4 lbm/ft 1 Curie = 3.7 x 10 dps Density = 1 gn/cm 1 kg = 2.21 lba 3
Heat of salorizations = 970 Etu/lbm I hp = 2.54 x 10 BTU /hr 6
Heat of fusien = 144 Btu /lba 1 Mw = 3.41 x 10 Btu /hr 1 Atm = 14,7 psi = 29.9 in. l~g.
1 Btu = 778 ft-Ibf 2
I ft. H O = 0.4335 lbf/in 1 inch = 2.54 cm 2
F = 9/5 C + 32 C = 5/9 ( F - 32)
Sonicr Operctero Exca SECTION 5 Theory of Nuclear Power Plant Operation, Fluids, and Thermodynamics r
5.01 (1.5)
The reactor is determined to be shutdown by 6*/.
delta K/K with indication in the source range of 30 counts per second.
a) What is the Keff when the reactor is shutdown (0.5) by 6% delta K/K7 b) What would the count rate be if Keff is (0.5) increased to 0.987 c) What would the count rate be if Keff is (0.5) increased to 0.997 5.01 Answers delta K
= 0.06 K
1 -K
= 0.06 K
1 = K + 0.06(K) 1= 1.06(K) a)
KEff = 1/1.06 = 0.94 (0.5) i 1 - K1 CR2 and 0.06 CR2 1 - K2 CR1 0.02 CR1
=
90 cps (0.5) b)
CR2 =
(30)(3)
Converely CR3 = 0.06 30 0.01 180 cps (0.5) c)
CR3 = (30)(6)
=
(Rule of thumb, doubles)
~
Reference:
Glastone and Sesonske, " Nuclear Engineering"..
1
'. 02 (3.0) 5 Refer to FIGURE 5.1, a sketch of a typical (not necessarily Trojan) auxiliary feedwater system utilizing two centrifugal pumps of similar characteristics and capacities. The plot of Volume Flow Rate versus Pressure shows the system with the "A"
auxiliary feed pump in operation as the, initial condition.
a) Show, on the Figure provided, how the curve (s)
(1.0) will change as the PORV opens and reduces Steam Generator pressure by 50%.
b) Show, on the Figure provided, how the curve (s)
(1.0) will change when the discharge' valve for the "A"
pump i s part'.al y shut.
c) What effect will a
decreased temperature of (1.0) the water in the storage tank have on the Net Positive Suction Head of the pump?
5.02 Answers a) attached (1.0) b) attached (1.0) c) Increases (1.0)
ReferencesSection III, Part B, General Physics manual on Heat Transfer and Fluid Flow.
0 2
71GURg 5.1 a
.hf
$ n..
. sie 6
cony >
S/q Chas Donts) 7:::-./3 0%
- 6 A
si...
W Tank i
Am.heergy,98 Powas b...
A,, gb*
h 4....., e... e I
System Operating Curve Ak inliial Operating Point t
/
~
/
\\
"/
Pump Operating Curve (PUMP"A")
\\
Volume Flow Rate (V)
..v Key FIGURE 5.1 a
.hf
$ er..... sio I m
2 sg ATeos. DJu9 s
}g P...O%
/
K
...5 -Af t:-
....,.93. k to...O%.
Af....-
. 3..
I System Operating Curve i
o ai/
t f
\\
initial OperatinD N " Point
/
i l L,.,
0p/
/
Pump Operating
?.
y Curve N "A")
'/
XP.
s v.
Volume Flow Rate (V) s I
5.03 (3.0)
Refer to the FIGURE 5.2 that.follows this page.
The figure is of " Heat Flux" versus
- Temperature Difference between a
Wall and the Bulk Fluid" fcr an operating reactor. Note that there are two curves represented for two pressures (
P1 P2 ).
a) What is the principle type of heat transfer (0.5) that is occuring at pressure Ej and 12953 DIW/Scrit between the w.all and the bulk fluid?
b) What is the principle type of heat transfer (0.5) that is occuring at Pressure E2 and 11953 DIWZUcrit between the wall and the bulk fluid?
c) What pressure will yield a
Lgget fuel (0.5) centerline temperature at 149E3 DIWZUcrit7 d) What pressure will yield a
bl9bfc fuel (0.5) centerline temperature
.at 't9ED SIWZbtrit?
e) What type of heat transfer between the wall and (0.5) the bulk fluid is occuring at Pressure El and 2195D DIUZbcrit?
f) Assuming bulk temperature well below (0.5) saturation, Will decreasing the pressure affect the bulk fluid temperature at a heat flux of 219E5 DIUZBcrit?
5.03 Answers (0.5) for each.
a) Nucleat boiling, b) Convection (other terms may be used).
c) Pressure 1.
d) Pressure 1.
lLM904g 7(s[ team blanketing).
m) Radiant heat transfer N6 -
f) The bull fluid temperature remains constant (in independent of pressure below saturation.)
Reference:
General Physics Nuclear Technology, Section E, Pages 2-144, 2-151, 2-159, and 2-164.
~
t P
/
% D,5
Reference:
General Physics, Volume II, Chapter 4, Sections B & D.
A -2.
i 9
10
4 0 5.11 Refer to Figure 5.6:
a) Sketch, on the Figure, the local radial coolant, (2.0) velocity profile and the local radial temperature profile on the coordinates provided assuming a bulk coolant velocity of 14 ft/sec, and a
peak centerline temperature of 1'/00 degrees F.
5.11 Answer:
a) Attached.
Temperature profile.
(1.0)
Velocity-profile.
(1.O)
Reference:
General Physics, " Heat Transfer, Thermodynamics and Fluid Floes Fundamentals",Section II, Part B, Ch.4.
I i
s 11
FIGURE 5.6 LOCAL R ADI AL TEMPER ATURE AND VELOCITY PROFILES l
I FUEL ROD
\\
l
-l l
r Ai A
fCLADDING CL ADDING m y PELLET PELLET E
}l
}
g COOLANT CHANNEL f
t 2 O.
U l
y Is
~
12 O
i COOLANT VELOCITY PROFILE 4
1700 1600 15OO TEMPERATURE PROFILE 1400 1300 1200
[
o 1100 1000 9OO 800 7OO 600 500
=
FIGURE 5.6 LOC AL R ADI AL TEMPER ATURE AND VEL OCITY PR OFILE S
' FUEL ROD s..
i 1
l A
A fCLADDING CL ADDING g PELLET PELLE1 g
COOLANT CHANNEL 20 o
"~
M 4+ / sat g
1.
(o.s) 12 COOLANT VELOCITY PR IILE 8
g (C.
~
1700 1600 TEMPERATURE PROFILE 1500 1400 1300 1200
[
1100 1000 900 800
/
700 N
/
STO F
_j 600
~
500
.2
5.12 (2.0)
Over the life of a core, the composition of the fuel changes.
a) deu and why does this change in composition (2.0) over core life affect the reactor period for,
equal reactivity additions?
5.12 Answers a)
The decrease in Uranium 235 and e increase of (1.0)
Plutonium 239 cause the Beta ffective to decrease.
Y This BetaAEffective change causes the reactor (1.0) period to decrease.
Reference:
General Physics, Volume II, Chapter 5, Sect.
E.
END OF SECTION 5 e
12
KCTION 6 PLANT SYSTEMB DESION, CONTROL, AND INSTRUPENTATION l
6.01 (4.0)
Concerning the Auxilary Feed Systems
- a. mat plant conditions will cause the two safety-related Auxilary Feedwater Pumps to automatically start?
(1.0)
- b. beist is the NORMAL supply of esator to the safety-related Auxilary Foodwater Pumps?
(0.5)
- c. Nhat is the first ALTERNATE supply of seater to the safety-related Aux 11ery Feedwater Pumps?
(0.5)
- d. How may Oxygen concentration and pH be controlled in the Feed and Condensate system eshen the Auxilary Feedwater system is in operation.
(2.0)
(Briefly describe components and Flowpaths involved) 6.01 Answer (4.0) a.
- 1) Tripping of both main feedwater pumps.
- 2) Low-Low S/G 1evel in a S/G (2 of 3 LTs on S/Gs 1/4)
- 3) Safety Injection Signal.
- 4) Undervoltage on 4.16 kV Essential Bus (Al - turbine; A2 - Diesel)
(4 at 0.25ea.)
b.
Condensate Storage Tank. (0.5) c.
Service Water System. (0.5) d.
Oxygen -An automatic start signal is sent to the shutdown hydrazine pumps, which supply each S/G, when the AFW pumps start (1.0).
pH - The morpholine pump supplies the electric driven AFW pump suction and the CST supply
~
l line to the AFW pumps. (1.0)
-t&
I(/
SD-I-12 AF System
<
- g eferences ##
1
6.02 (3.5)
The Reactor Coolant System is SOLID at 200 Dog.F and 360 Psig.
The overpressure protection switches are in the
" Unblocked" position.
Decay heat removal is providsid by RWt system operation.
Only one MCP is operating.
a.
What is the normal LETDOWN flow path to the VCT in this mode?
(1.0) b.
How is RC8 pressure controlled in this mode?
(0.5) c.
By technical specifications, may a PORV Block Valve be shut in the stated situation?
(0.5) d.
Amoues due ta oporator error or equipeont faiture, the running RCP is lost.
Explain how and ehy the RCS Pressure sould be affected, if a RCP is now started?
( How - 0.5 g tesy - 1.0) 6.02 Answer (3.5) a.
Letdown via the manual /alves downstream of the RHR HX; to the LETDOWN HX; through the Pressure Control Valves (PCV - 131); to the VCT. (1.0) 6.
By the Letdown Pressure Control Valve PCV -131.
(0.5) c.
NO.
(0.5)
(the system must be cool ed down and depressurized and a 3.4 sq.in. vent provided.)
d.
With all RCP's stopped, and RHR valves open to a solid
(0.5)
While the RCP's are stopped, a
quantity of cold seal injection water will accumulate in the pump volume and, upon starting of the pump, will be heated in the steam generator causing the pressure surge.
(1.0)
Reference:
Tech. spec's.; CVCS SD-I-06: RHR SD-I-08; PZR SD-II-04 S
2
6.03 (1.0) mich of the fallowing statements concerning the Containment $ ray Byetem is trues (1.0) a.
The Bodium Hydrowide in the spray facilitates the converting of soluble Iodine into insoluble Iodine?
l b.
Upon reaching the low-low level setpoint in the i
- RMBT, the euction of the spray pumps will be automatically shifted to the containment sump and isolate from the RMBT.
c.
During the injection phase of operation, Optimum spray additive flow, along with proper additive tank concentration, is responsible for maintaining the desired pH
- range, and thus limit corrosion inside of containment.
d.
The Spray Pumps start on a SI signal, but they do not inject into the containment until the discharge valves are opened upon receipt of a high reactor building internal pressure signal on 2/4 channels.
6.03 Answer (1.0)
(c)
(1.0) i
Reference:
SD-I-09 Containment Spray System
.-~
6.04 (1.0)
The Pressurizer pressure and level control system is designed to accommodate without a plant trip.
a.
Ioading or unloading at 10% per minute.
b.
a step load reduction of 50% with automatic rod control and steam dump operation.
I c.
instantaneoue load changes at +/- 20 percent without automatic rod control.
d.
Ioading at 10% per minute and unloading at 15% per minute.
6.04 Answer (1.0)
(b)
Reference:
SD-II-04 PZR pressure and level control l
l l
l 4
6.05 (1.0)
The Reactor Trip Breaker Shunt Coils will energize on 1
m.
only manual trip signals.
b.
any automatic trip signal.
c.
only SI initiated automatic trip signals and manual trip signals.
d..
only RPS and SI initiated automatic trip signals.
6.05 Answer (1.0)
W Reference :
SD-II-03 RPS +@
~
M0 J f w + kpL w
5 L
a.06 (4.0)
Match the following reactw protection system trips with its respective functions Tripes a.Overtemperature delta T b.Loev-Low Steae Beneratw level c.Dverpower delta T d.P weer Range High Flux Rate Trip Functions:
- 1. Protect against rod ejection accident.
- 2. Protects against reactivity excursions, initiated from low pws, eshich are too rapid fw adequate detection.
- 3. Protects the cwe from overpooser conditions that esculd result in excessive kilonoatte per foot and high fuel temperatures.
- 4. Protects the cwe from overposeer conditions that enould result in departure from nucleate boiling and possible damage to the fuel cladding.
5.
Insures that primary heat sink is available f w decay heat removal.
6.06 Answer (4.0) a.
- 4.
b.
- 5.
c.
- 3.
d.
- 1.
l
Reference:
SD-II-03 Reactor Protection System i
i 6
6.07 (1.0) tetich if the following to true concerning the Bource Range channel high voltage cutoff?
(1.0) a.
During a reactor startup either IR channel increasing above the P-6 setpoint will turn off the high voltage.
b.
If one IR channel falle low while at power, the Source Range high voltage will be re-energized.
c.
Teen out of four PR channele above 10 % power will block the high voltage.
d.
During a reactor shutdown either IR channel decreasing below the P-4 setpoint will turn on the high voltage.
6.07 Answer (1.0)
(c)
Reference SD-II-07 Excore NI's l
I I
l 7
6.08 (1.0) mich of the following valves will automatically open by a Safety Injection (51) signal?
a.
Doron injection tank recirculation isolation valves, if shut.(CV 8870 A and 3) b.
fHt pump ontor operated a valves, if shut.
(MO 8811 A and B) c.
1485T to fHt pump isolation valves, if shut and power available. (MO AS12,MO SW70-A and B) 35 0 d.
Baron Injection Tank discharge valves, if shut and power available.
(PE3 5901 A and 3) 6.08 Answer (1.0)
(d)
Reference:
SD-I-07 ECCS d
1 I
9 e
8
6.09 (2.5) m at M Main Steam conditions will automatically a.
result in a Safety Injection Signal?.
(2.5)
(in your answers Name the detector or parameter,.
and state coincidence required, if any.)
6.09 Answer (2.5) 1.
High Steam Line Flow 1/2 detectors on 2/4 main steam lines, coincident with:
a) low steam line pressure on 2/4 detectors or b) low-low T ave (P-12) on 2/4 loops (1.5) 2.
Any one steam line pressure 100 paid lower than the pressure in 2/3 of the other steam lines.(1.0)
Reference SD-II-03 RPS l
9
6.to ca.5>
het will occur to the Service teater System lineup on a safety Injection signal (1.5) 6.10 Answer (1.5) 1.
Standby service water train starts.
(0.5) 2.
Valve relignment to separate Class I from Class II.
(0.5) 3.
Service water is aligned to supply the Emergency DG HXs (0.5)
Reference SD-I-10 p.7 l
l f
(
l l
10
6.11 (1.0)
List FOUR of the seven Continuous Liquid Radiatier)
Monitoring Systems that are provided at Tragen.
~
(1.0) 6.11 Answer
( 1. 0)~
1.
Liquid Rad. Waste Discharge Header (ffe#~i 2.
Component Cooling Water System (2)(Pe<ru s - 7/F) (any 4 -
3.
Steam Generator Sampling Syster(P(Ite -iD) 0.25ea.)
4.
CVCS Normal Letdown (ft1tm -3 %)
- 5.. RCDT ( /d1t* s F )
Containment Sump ( /ptd-/Y) 6.
Reference:
SD-X-09 O A)/
I 11 l
6.12 (2.0)
For~ the following questions assume the plant is at 96 per cent power and normal operations are being conducted.
a.
List the FOLR conditions that will
" ARM" the*
Steae Dueps ?
(1.0) b.
How enould the Steae Duep Systee, in the T Ave Mode, rzepond to a Channel D Cold Leg Narrow Range RTD failing HIGH ?
(1.0) 6.12 Answer (1.5) a.
1.
Selection of STEAM PRESSURE MODE.
2.
Greater than 10 % step decrease in Turbine Power.
- 3. (Greater than 5 % per Min. Turbine)
Load Rejection.
4.
b.( Since none of the above conditions were met, the steam dumps are not " ARMED";
thus although a demand signal is generated by the T ave - T ref mismatch,) the Steam Dumps will remained closed.
(1.0)
Reference:
SD-II-06 SDS
,A mw aou lo Lo -lo 1 4 ft/64wwt 72fV + & ^*"*1
/
- 3. to -to s wa
'I Byfus T*Jd % g g og g 12
6.13 (1.5)
)
a.
W at are the control signal inputs to the Automatic Rod Control System 7 (1.0) b.
Assumming Normal plant operations, met esould be the Automatic Rod Control System's responce to a Channel D Cold Leg Narrose Range failing HISH ?
(0.5)
\\s 6.13 Answer y
a.
1.
Auctioneered Nuclear Power (NIS) 2.
Auctioneered T Ave (3 required
- O.33ea.)
3.
Turbine Impluse Pressure (P imp) b.
Stepping in of the controlling bank at 72 steps / min.
(0.5)
Reference :
SD-II-09 Rod Control t
l 13
Senior Reactor Operators Examination SECTION 7 Procedures - Normal, Abnormal, Emergency and Radiological Control 7.01 (2.5)
Administrative Order AO-3-7 " Post-Trip Review and Permission for Reactor Trip Recovery and Mode Changes" has requiremnets that must be met prior to restarting the reactor after a
a) What are tuttg of the four conditions that must (1.5) be met prior to a restart of the reactor?
b) If any of the conditions above cannot be
- met, (1.0) from whgm must permission be obtained before the reactor can be restarted?
7.01 Answers a) 1.
The cause has been determined for the trip.
2.-The post-trip review has been completed.
- 3. No PRO *s, beyond the event it self.
4.
Permission obtained from management to restart.
(any 3 of the above (0.5) each) b) The Duty General Manager.
(1.0)
Reference:
AO-3-7
" Post-Trip Review and Permission for Reactor Trip Recovery and Mode Changes" 4
1
G i
7.02 (2.0) s Emergency Contingency Action ECA-0.0 " Loss of All AC Power" is entered during an event that has resulted in the loss of all AC power in the plant.
Step 5 in the procedure directs the operator to restore power to the A1 and A2 buspes from the Emergency Diesel Generators.
a)' During Step 5 of the proce' dure the operator (1.0) determines that;an EDG cannot be closed in on one of the busses.
What concern arises if the EDG is permitted tu continue running?
b) If the operator does determine that the EDG can (1.0) be closed in on the bums, but the Shut Down Sequencer does no: automatically load the buss.
What action would you expect the operator to take?
'i.
7.02 Answer:
a) If the buss cannot be loaded the EDG must be (1.0) shut down to prevent damage from loss of cooling.
b) The operator would have to load the EDG by (1.0)
\\
manually closing the loads on the buss.
References ECA-0.0 " Loss of A1) AC Power"
~*
t s
e k
i -
s
\\
9 n
2
7.03 (2.5)
Regarding Procedure ONI-7 " Reactor Control Malfunction".
a) While' moving control bank D rods to verify (1.5) operablity a rod bottom light and a flux tilt alarm are received, and no Reactor trip occurs.
What immediate actions must be taken at this point?
b) The condition above is determined to be a
(1.0) dropped rod.
Maintenance personnel took the last two shifts to repair and correct the problem.
During the recovery of a rod that has been misaligned for an extended pariod, what ia the operator *s major concern regarding rod motionf 7.03 Answer a)
- 1. Stabilize plant parameters.
(0.5)
(If Reactor has tripped refer to El-0.)
- 2. Consult the Emergency Plan.
(0.5)
- 3. Determine quadrant power tilt ratio.
(0.5) and check AFD against TS*s b) The rod must be moved slowly to avoid potential (1.0) large local power increase.
References Procedure DNI-7 pages 3 and 4.
l 3
7.04 (3.0)
While you are standing watch as the, Shift Supervisor an Auxiliary Operator calls in to announce that he sees a fire in the Diesel Generator room, the Control Operator then informs you that he has recieved a smoke alarm indipation in the East Diesel Generator room.
-a) What two considerations are there,for a fire in (1.0) this location as far as determining the classification under the Trojan Emergency plan implementing procedure EP-1
" Emergency Classification"?
b) You dispatch the Firs Brigade to the scene of (2.0) the fire.
Who are the members of the Brigade, (include the number of personnel and their title or job description)?
7.04 Answers a) Location is in a safety related equipment area.
(0.5)
If the fire is not out in 10 minutes.
(0.5) b)
1.
Assistant Shif,t uperv sor - Fire brigade (1.0) leader.(m 84 aM 2.
Plant Operator.
(0.25)
- 3. Plant Operator.
(0.25) 4.
Security Officer.
(0.25) 5.
Security Officer.
(0.25)
Reference:
AO-10-2
" Fire Protection",
" Emergency Classification"
~
4
7.05 (2.0)
You are touring in a controlled area in the plant and you drop your dosimeter, you read the dosimeter and it is reading offscale.
a) What is required of you as far as immediate (1.0) actions are concerned?
b) What will have to be done before you will be (1.0) allowed to re-enter the controlled area?
7.05 Answers a) Immediately proceed to the exit point.
(0.5)
Then Report.to Radiation Protection Office.
(0.5) b) A Personnel Exposure Investigation.
(1.0)
Reference:
Radiation Protection Manual pages 2-10 and 4-4.
O 5
7.06 (2.0)
The plant is about to be brought critical with a calculated Estimated Critical Position (ECP).
a) It is determined during the approach to.
(2.0) criticality that the reactor became critical 100 steps on the step counters before the projected ECP. After initiating a recalculation of the ECP what two other items should be-reverified?
7.06 Answers a) The DRPI and step counters should be rechecked.
(1.0)
The baron concentration should be reverified.
(1.0)
Reference:
GOI-B " Estimated Critical Position" page 2.
i b
6
\\
7.07 (3.0)
The Radiation Protection Manual describes might requirements for when a RWP has to be issued.
a) What are four of those requirements?
(2.0) b) You are on a plant tour investigating an event (1.0) in the plant, you need to gain immediate access to a
High Radiation Area, and there is a
Radiation Protection Technician availiable with a
survey instrument.
Can you enter the High Radiation Area without a specific RWP7 Briefly explain.
7.07 Answer:
Any four (0.5) points each l
a) 1.
Entry into a radiation area, high radiation area, or surface contimination area.
- 2. Entry to or' planned work in an area with potential neutron exposure.
- 3. Maintenance or inspection of equipment with contamination levels in excess of 1,000 DPM/100cc2 beta-gamma or detectable alpha.
- 4. Entry to or planned work in an area where airborne activity is greater than 25% of 10CFR2O limits.
- 5. Work that involves changes (uncovering, opening i
valving, moving,etc) that have the potential of increasing radiation or contamination levels.
- 6. Handling or using radioactive sources.
- 7. Work that involves opening of potentially l
contaminated systems.(ie.5/G's after Tube leak)
- 8. As deemed necessary by the Rad. Protection Dept.
b)
Yes, provided the R.P. Tech. accompanies with the survey instruments.
Ref erence: Procedure AP 305-4 " Radiation Work Permits",
page AP.305-4-1.
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7.03 (3.0)
Regarding DNI-10 " Emergency Boration":
a) What are fgyr of tne five conditions that may (2.0) require emergency boration of the reactor coolant system (RCS)?-
b) If it is determined during the executr'on of the (1.0) immediate actions of ONI-10 that no emergency boration flow exists what alternate source may be used to supply emergency borated water for emergency boration?
7.08 Answers a) Any four (0.5) each 1.
Insufficeint shutdown margin.
2.
Excessive control bank insertion.
- 3. Uncontrolled cooldown of RCS.
4.
Unexplained or uncontrolled reactivity increase.
5.
One or more rod position indicators fail to indicate rods fully inserted after plant shutdown.
b) From the Refueling Water Storage Tank (through (1.0)
MO-112B and MO-112E.)
Reference:
ONI-10 '.' Emergency Boration" pages 1 and 2.
O 8
7.09 (3.O)
During power operation an inadvertant Safety Injection occurs when one of the Pressurizer Poseer Operated Relief Valves opens from unknown cause. The event is diagnosed, and in step 14 of EI-0, it is determined that the cause of the SI wais the stuck open PORV.
m en the open PORV was discovered, the operators shut its block valve.
a) What criteria must be met to allow termination (2.0) of the Safety Injection?
b) The first step of ES-1.1 "SI Termination" is to (1.0) reset SI and CIS.
Automatic initiation, if necessitated, can not reinitiate until what other action is taken?
7.09 Answers:
a)
- 1. RCS pressure stable or increasing.
(0.5)
- 2. Pressurizer level > 5%
(0.5)
- 3. RCS Subcooling > 30 deg.
F.
(0.5)
- 4. Secondary heat sink avail.
(0.5)
(AFW flow > 980 gpm or NR level
> 5% in one S/G) b) The auto reinition will not occur until the (1.0) reactor trip breaker control switches are cycind.
References ES-1.1 "SI Tereination" page 2, EI-1 " Loss of Reactor or Secondary Coolant" page 6.
1 b
4
~
l e
9
7.10 (2.0)
Regarding 01-1-7 "120-V Preferred Instrument Bus Operation":
a) What concern arises when Reactor Coolant System (1.0)
(RCS) temperature is below 553 degrees F.
and shifting power sources for Y11 and Y22?
t b) The procedure contains a precaution to not (1.0) operate an inverter at no-load with high DC input voltage (140-V or greater).
What is the concern associated with this precaution?
7.10 Answers a) A (High Steam Flow) Safety Injection may occur (1.0) from the momentary voltage transient.
b) This precaution is there to ensure the inverter (1.0) is not operated in a mode that would cause overheating.
Reference:
01-1-7 "120-V Preferred Instrument Bus Operation",
page 3.
END OF SECTION 7 1
9 10 1
K CTIEM E
ADMINISTRATIVE PROCE3MES, COPSITIONE, A05 LIMITATIONE 1
B.01 (1.0)
THLE or FALBE 7 Tech Space do not require any action if one control rod is tesovable, provided the immovable rod is within 12 steps of it's group stop counter demand position.
(e See Enclosed Tech Spec) 0.01 Answer (1.0)
False (1.0)
Reference:
TS 3/4.1.3.1 9
h 8
i 1
9.02 (1.0)
Tech If the lowest operating loop Teve drops below Specs allow to restore Teve within the. limit or proceed to place the unit in hot standby.
a.
557 degrees F, 30 minutes b.
557 degrees F, 15 minutes c.
551 degrees F, 30 minutes d.
551 degrees F, 15 minutes 8.02 Answer (1.0)
(d)
Reference : Tech Spec 3.1.1.5 e
en G
2
9.03 (2.0)
Each MPS Pressurizer Pressure - High channel is resguired by Technical Specifications to undergo a channel check on a shift basis (at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />).
Some extensions of the basic interval s e alloosed by the Tech. Spec's.
Records shoes that this esas done ons November 27 at 0000 November 27 at 1500 November 28 at 0000 November 29 at 1400 November 29 at 0800 a.
m et is the maximun alloseable interval beteesen channel check ourveillances 7 (1.0) b.
M en is the next channel check surveillance due?
(1.0).
8.03 Answer (2.0) a.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> + 25% = 1U bguts (1.0) b.
1.
Last done + 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> = Nov 29 at 2000 2.
Last 3 + 3.25 X 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> =
Nov 28 G COOO + 39 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br /> = Nov 29 e 1500 Therfore IT must be done no later than: Ugy 22 g 1Q99 (1.0)
Reference : Tech Spec *s 4.02 3
5.04 (1.0) t#wn a fire supreselon Spray / Sprinkler System is declared inoperable f w a pwtion that protects an area containing redundant safety-related equipeont, the resquired action is tas (pick one)
(1.0) a.
commence a Unit shutdown within one hour.
b.
establish an hourly fire patrol f w the affected area.
c.
establish a
continuous fire watch with backup fire supression equipment in the affected area within one hour.
d.
establish a backup suppression system in one hour or be in HOT WAJTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
8.04 Answer (1.0)
(c)
Reference:
4 4
B.05 (3.5) that are the Containeont Integrity requiremente during CORE ALTERATIONS or anvenent of irradiated fuel within the containment 7
8.05 Answer (3.5) 1.
The equipment door closed and held in place by a minimum of four bolts.
(1.0) 2.
A minimum of one door in each airlock is closed and (1.0) 3.
Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:
a.
closed by an isolation valve, blind flange, or manual valve, (0.75) or b.
be capable of being closed by an OPERABLE automatic containment Ventilation isolation valve.
(0.75)
Reference TS 3.9.4 5
8.06 (4.0)
The Shift Supervisor may initiate a reactor startup after ir a reactor trip, reactor safety limit esas not
- exceeded, appropiate BOI's completed, and the ______
conditions have been mots
( four responces required) 8.06 Answer (4.0) 1.
The cause of the trip has been determined and corrected.
(1.0) 2.
The post trip review has been completed by the Shift Supervisor and Shift Technical Advisor (0.5) with all discrepancies being evaluated and determined not to affect plant safety or startup. (0.5)
(1.0) 3.
There are no possible Reportable Occurrences as a result the reactor trip beyond the reactor protection actuation itself.
(1.0) 4.
Permission obtained from one of the following management personnels a.
Operations Supervisor b.
Assistant Operations Supervisor c.
Plant General Manager or Duty G.M.
(1.0)
Reference:
AO-3-7 G
6
3.07 (1.5)
The plant is in mode 4, preparing for a
routine plant startup.
As a result of routine surveillance, it is' determined that only one Centrifugal Charging pump is* aparable.
The Maintenance Supervisor has told you, that the cause of the failures is knoesn,
parts are on site, and that repairs and operability checks esill be completed in less than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
Explain eshy the startup can or cannot proceed.
(See enclosed Tech Spec's) 0.07 Answer (1.5)
The startup cannot proceed, (0.5) because " entry into an Operating Mode is not allowed if an action statement must be relied upon to do so.
(1.0)
(i e.
you would be in an action statement in mode 3)
Reference:
I I
s l
7
5.OS (2.O)
Provide the sinteue number of individuale required by ISEb. M for the following positions to operate the plant at full power (mode 1)s
- a. _
Senior Operating Licenses (SOL) b.
Operating Licensee (OL) c.
Non-licensed Persons d.
Shift Technical Advisors (STA) 8.08 Answer (2.0) a.
2 **
(0.5) b.
2 (0.5)
- c. 2 (0.5) d.
1 **
(0.5)
( ** notes one of the two required individuals filling the SDL positions may also fill the STA pcsition provided the individual is qualified for both positions, and have a Bachelor's degree).
References Table 6.2-1 and TS 6.2.2 O
O
G.09 (1.0)
Deviations f rom a procedure durin0 DEC981 R1801 MSALEGE8 a.
are not alloesed.
b.
can be made if the orginal intent of the procedure is satisfied.
c.
can be made eith verbal approval of teen licensed operators. (one of eshich is licensed as a ERO).
d.
can be made only in an emergency eden tamediate action is needed to protect the public health and safety.
B.09 Answer (1.0) d.
( 1. 0) og h(/,0)
Reference:
TS 6.0 3 A0-1-1.
@L f
h PuM ON Y' 9
i 3.10 (1.0)
Post-Trip Revisese (PTR) are performed bys a.
Any personnel licensed by the p54.
b.
Shift Supervisor and the mift Technical Advisor.
c.
Whift Technical Advisor.
d.
mift Technical Advisor, only if they hold a valid GRO license.
8.10 Answer (1.0) b.
Reference : A0-3-7.
l 10 4
nn
_ - - - - ~ - - - - -. - - - - -.
w,,----,,--,-
--m,-
l 1
- 3.11 (1.0)
EP-6 defines the Emergency Coordinator as the person who is designated to take charge of all emergency control momeures, and has ultimate authority over all onette activities and personnel.
a.
teto initially fills this role in the early stages of a event?
(0.5)
- b.. teto relieves this first person and has the primary responsibility for this role?
(0.5) 8.11 Answer (1.0) a.
the Shift Supervisor (0.5) b.
the Plant General Manager (0.5) or
( the Duty Plant General Manager)
Reference : EP-6 O
h l
l 11
I 5.12 (1.0)
On most all of the Emergency Supply Inventory
- Lists, various units of Poteesium Iodine Tablets are listed.
Potassius Iodide tablete are made available to e a.
reduce the offacts of Beta radiation to the lens of the eye.
b.
reduce the amount of radioIndine in the thyroid gland.
c.
reduce the effects of neutron radiation to soft body tissue.
I d.
reduce the amount of radiakrypton collected in the bone earron.
9.12 Answer (1.0) b.
l I
i 12
3.13 (3.0) mat are the Technical Specification LIMITS and.the BASES for the following types of leakage from the Reactor Coolant System a
- a. Primary to Secondary (2.0)
- b. Controlled Leakage (1.0) 0.13 Answer (3.0) a.
Primary to Secondary Leakage:
1.
1 GPM through all S/G's (0.25)
Bases: ensures that the dosage contribution from the tube leakage will be limited to a small fraction of the 10 CFR 100 11mits in the event of either a S/G tube rupture or steam line break.
(0.75)
- 2. 500 GPD through any one S/G (0.25)
Bases:
wnsures that the S/G tube integrity is maintained in the event of a main steam line rupture.
(0.75) b.
Controlled Leakage 20 GPM from all reactor coolant pumps 6 GPM from any one pump (0.25)
Bases: ensures in the event of a LOCA, the safety injection flow (borated) will not be less than assumed in the accident analysis. (0.75)
References TG 3.4.6.2 and its bases.
6 13
5.14 (1.0) mich of the folloeing to lEI required of the A.C. and D.C.
Electrical poser sources by Tech Spec *e in mode 5.and & 7 a.
Teen circuito bettesen the of feita trenesission notesork and the onsite ESF Electrical System.
b.
One operable Diesel generator set.
l c.
One energized and operable 125-volt D.C. bus aligned to its associated charger.
d.
One energized and operable 125-volt battery bank.
8.14 Answer (1.0) a.
reference :
TS 3.8.1.2 and 3.8.2.4 i
f 14
3.15 (1.0)
If control power is lost to a preneuriner power operated relief valve (PORV) while in Mode 1, to continue to operates
- a. no action is required by Tech Spec's provided another PORV is operable and all preneuriner code safety valves are operable.
- b. Tech spec's require the associated block valve to be verified open and then its power supply to be removed, if the PtNW is not made operable within one hour.
- c. Tech spec's require the associated block valve to be l
mhut and its power supply to be removed if the PORV is not made operable within one hour.
- d. Tech Epoca requires action to be initiated within one hour to place the plant in at least HOT STAPSBY within the following hour, if the PORY is not made operable.
8.15 Answer (1.0) c.
Reference : TS 3.4.3.2 l
1 j
l i
15 i
l i
l i
3/4 LD4mW CONDITTCNS AR OPERATICN AND SURVEILLANC! REOUIRE{
3/a.0 Appt.ICA8ILITY LIMITTM CCE ITTON FOR OPERATTCN r
3.0.1 i
Lfatting Canettions for operation and ACTION reouirments shall be anglicable ering ce OPERATIONAL M0013 or oe., c=n.itions specified for eacn speciffcatton.
i 3.0.2 Aderence to ee requirements of the Listting Condition for Operation and/or associated ACTICN wiein the specified time intenal snail j
constitute cuof f ance wie the spectff cation.
Candition for Operation is restored prior ta axofration of me specifiedIn ce event tism interval, completton of the ACTICN statment is not required.
3.0.3 In the event a Listting Condition for Operation and/or associated
)
ACTION reevirments cannot be satisfied because of :f reumstancas in excess of these addressed in the speciffcatton, es unt: snail be placed i
in at least NOT STANOSY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, in at least HOT SHUTOOWN wimin A
ee next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and in at least CCL3 SWTUCWN within ee following g
30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> unless corrective measures are causletad that permit ooerstien i
under me persissible ACTION statements for the specified time interval I
as seasured from inttial dise:very or until :ne reactor is placed in a A
l M001 in wnich the soecification is not apolicable.
Excantions to these Q
requirments snail as stated in tne inefvital specifiestions.
3.0.4 Entry into an CPERATICNAL MODE or etner specified aegifcan111ty condition sna11 not be sede unless ce caneiticas of tne Limiting Concition I
for Oceration are met wf thout relf ance on provisions cantained in tne ACTICN statments unless othemise excented.
{
This provision shall not prevent passage tnrougn CPERATICMAL N00E3 as required ts c=maly wita ACTION stataments.
3.0.3 idhen a system, subsystus, train, caponent or device is detarained i
to be inopertale setely because its amargency power scurte is f acceraale, i
i or solely because its normal power sourte fs inoceraale, f t say be i
consicered CPERABLE for the purpose of sattsfying the requiremen.s of its asolfcable Limittag Cone 1 tion for coeration, proviced: (1) its carresnonding normal or ameriency power sourca is CPERA8LE; and (21 all of 1ts recuncant systan(s), suosystan(s), train (s), eomeonent( ) ane
[
l cevice(s) are CPERA8LI, or likewise satisfy the requirments of this scocification.
Unless notn canettions (1) ane (2) era sattsfied, tne i
unit sna11 he placed in at least HOT STANC8Y witnin 1 hcur, in at least HOT SHUT CWN witnin tne next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in at least COL 3 IHUTCCWN witnin tne follcwing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. This specificatien is not amplicanle in MCt3 5 or $.
i
}
TRCJAN-UNIT 1 3/4 C4 Amenc ent 'lo. !3 1
1 Decembe r 11, 1980 4
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)
3/4.0 APPLICA8!LITY SURVE!LLANCE REQUIREMENTS 5
4.0.1 Surveillance Requirements shall be applicable during the OPERA-T!0NAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an indivioual Surveillance Requirement.
4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with:
a.
A maximum allowable extension not to exceed 25% of the sur-ve111ance interval, and b.
A total maximum combined interval time for any 3 consecutive surveillance internals not to exceed 3.25 times me specified surveillance interval.
4.0.3 Perfomance of a Surveillance Requirement within the specified time interval shall constitute compliance with OPERA 81LITY requirements for a Limiting Condition for Cperation and associated ACTION statements unless otherwise required by the specification. Surveillance Require-
/.4 ments do not have to be performed on inoperable equipment.
(]
4.0.4 Entry into en OPERATIONAL MODE or other specifies applicability condition shall not be made unless me Surveillance Requirement (s) associated with 2e Limiting Condition for Operation have been performed within the stated surveillance interval or as 02erwise specified.
The provisions of Specification 4.0.4 are not applicable to the perform-ance of surveillance activities associated with fire protection tech-nical specifications 4.3.3.7.1, 4.3.3.7.2, 4.7.8.1.1, 4.7.8.1.2, 4.7.8.1.3, 4.7.8.3 and 4.7.9 until the completion of the initial sur-ve111ance interval associated with each specification.
4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2 and 3 components shall be applicable as follows:
a.
Inservice inspection of ASME Code Class 1, 2 and 3 compo-nents and inservice testing of ASME Code Class 1, 2 and 3 pumps and valves shall be performed in accordance with Section XI of the ASME 8ciler and Pressure Vessel Code
/\\
and appitcable Addenda as required by 10 CFR 50, Sec-M tion 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50,
- Section 50.55a(g)(6)(1).
i
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TRCJAH-UNIT 1 3/4 0-2 Amendment No. 61 May 8,1981
3/4.0 APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued) b.
Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be appitcable as follows in these Technical Specifications:
ASME Boiler and Pressure Vessel Code and applicable Required frequencies Addenda teminolow for for perfoming inservice inservice testing activities testing activities weekly At least once per / days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days.'
I c.
The provisions of Specification 4.0.2 are applicable to the above required frequencies for perfoming inservice inspection and testing activities.
d.
Perfomance of the above inservice inspection and testing activities shall be in addition to other specified Surveil-lance Requirements.
e.
Nothing in the ASME aoiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.
c l
TROJAN-UNIT 1 3/4 0-3 Amenoment No. 61 May 8,1981 I
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3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MRGIN LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MRGIN shall be 11.6% Ak/k.
i APPLICABILITY: MDES 1. 2*, 3, 4 and 5.
M:
With the SHUTDOWN MRGIN < 1.6% Ak/k inmediately initiate and continue boration at > 30 gpm of 7000 ppm boron or equivalent until the required SHUTDOWN MRilIN is restored.
SURVEILLANCE REQUIREE NTS 4
4.1.1.1.1 The SHUTDOWN MAGIN shall be determined to be 16% Ak/k:
1 l
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Within ona hour after detection of an inoperable control rod (s) s.
A in accordance with Specification 3.1.3.1.a. and at least once Mo per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the rod (s) is inoperable.
If the inoperable. control rod is innovable or untrippable, th's above required SHUTDOWN MAGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the innovable or untrippable control red (s).
8 b.
When in MODES 1 or 2, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that control bank withdrawal is within the limits of Specifica-tion 3.1.3.5.
When in MDE 2ff, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to achieving reactor A
c.
criticality by verifying that the predicted critical control Q90 j
rod position is within the limits of Specification 3.1.3.5.
d.
Prior to initial operation above 5% RATED THERML POWER after each fuel loading, by consideration of the factors of a below, with the control banks at maximum insertion limit of Specification 3.1.3.5.
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- See Special Test Exception 3.10.1.
3 l
- With Xeff 1 0.
1 ffWith X,ff 4 1.0.
4 TROJAN-UNIT 1 3/411 Amendment No. f(, 90 i
j August 2, 1984 j
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CCCS SUCSYSTDIS'- Tava 2. 350*F Lif11 TING C0ftD! TION FOR OPERATION
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3.5.2 Two independent ECCS subsystems shall be OPERACLE with each subsystem comprised of:
One OPERABLE centrifugal char'ing pump.
g a.
l b.
One OPERABLE safety injection pump.
c.
One OPERA 8LE residual heat removal heat exchanger, i
d.
One OPERA 8LE residual heat removal pump, and e.
An OPERA 8LE flow path capable of taking suction from the refueling water storage tank on a safety injection signal and transferring suction to the containment sump during th$ recirculation phase of i
operation.
APPLICA3fLITY: MODES 1, 2 and 3.
ACTION:
~
WktherseECCSsubsystsminoperable,resterotheinoperablesub-a.
system to'0PERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in 110T SHUTD0;!ff 1
1 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b.
In the event the ECCS is actuated and injects water into the j
Reactor Coolant System, a Special Report shall be prepared and
'J submitted to the Consission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.
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i Tr0JAN-UNIT 1 3/4 5-3 l
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{ )S REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVA8LE CONTROL A55EM8 LIES GROUP HEIGHT i
LIMITING CONDITION FOR OPERATION
)
3.1. 3.1 All full length (shutdown and control) rods shall be '0PERA8LE h
~
And positioned within + 12 steps (indicated position) of their group step counter demand positici.
i APPLICA81LITY: MODES 1* and 2*
ACTION:
a.
With one' or,moh full length rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that i
the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in HOT STAND 8Y within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With more than one full length rod inoperable or h
sisali ned from any other rod in its group by more than + 12 steps indicated position), be in NOT STAND 8Y within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
i c.
With one full length rod inoperable or misaligned frcss h,
its group step counter demand height by more than + 12 steps (indicated position), POWER OPERATION,may continue ~provided that within one hour either:
~
1.
The rod is restored to CPERA8LE status within the above alignment requirements, or 2.
The rod is declared inoperable and the $HUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION say then continue provided that:
i a)
An analysis of the potential ejected rod worth is
' performed within 3 days and the rod worth is detar-l sined to be < 0.985 ak at zero power and < 0.215
& at RATED THERMAL PCWER for the remaindir of the fuel cycle, and 1
)
b)
The SHUTDCWN MARGIN requirement of Specification 3.1.1.1 l
1s determined at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and i
f
'See Special Test Exceptions 3.10.2 and 3.10.4.
TROJAN-UNIT 1 3/4 1-18 Amendment No. 7 0 i
Margh 3,1982
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REACTIVITY CONTROL SYSTEMS
(
LIMITING CONDITION FOR OPERATION (Continued) c)
The THERMAL POWER level is reduced to 7 % of RATED THERMAL POWER within one hour and withTn the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the high neutron flux trip setpoint is reduced to < 85% of RATED THERMAL POWER,*6r d)
The remainder of the rods in the group with the inoperable rod are aligned to within + 12 steps of the inoperable rod within one hour whTie maintaining the rod sequence and insertion licits of Figures 3.1-1 and 3.1-2; the THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.5 during subsequent operation ~.
i.
SURVEILLANCE REQUIREMENTS 4.1. 3.1.1 The position of each full length rod shall. be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the Red' Position Deviation Monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
i 4.1.3.1.2 Each full length rod not fully inserted shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days.
~
I iiTROJAN-UNIT 1 3/*4 1-19 Amendment No.70 March 3,1982 O
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REACTIVITY CONTROL SYSTEMS s
POSITION INDICATOR CHANNELS LIMITING CONDITION FOR' OPERATION 3.1.3.2 Control rod position indica' tor channels for control and shutdown k
rods and the demand position indication system shall*6e OPERABLE and
~
~
capable of determining the control rod positions within + 12 steps.
APPLICABILITY: MODES 1 and 2.
ACTION:
a.
With a maximum of one rod position indicator channel per group in-operable either:
1.
Determine the position of the non-indicating rod (s) indirec.tly by the movable incore detectors at least cnce per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and ilmnediately after any motion of the non-indicating rod which exceeds 24 steps in one direction since the last detensination of the rod's position, or 2.-
Reduce THERMAL POWER TO < 50t of RATED THERMAL POWER within
)
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
/
b.
With a' maximum of one demand position indicator per bank incperable either:
1.
Verify that all rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12 steps of each other at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or 2.
Reduce THERMAL POWER to < 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
SURVEILLANCEdEOUIREMENTS 4.1. 3. 2 Each rod position indicator channel shall Le determined to be OPERABLE by verifying the demand position indication system and the rod position indicator channels agree within 10 steps at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the Rod Position Deviation Monitor is inoperable, then cocipare the demand position indication system and the rod position indicator channels at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
TROJAN-UNIT 1 3/4 1-20 Amendment No. 7 0 i
March 3,1982 -
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REACTIVITY CCNTROL SYSTEMS CONTRCL 200 INSERTICN LIMITS LIMITING CCNDITION FOR OPERATICN 3.1.3.5 The control banks shall be limited in physical insertion as shown in Figures 3.1-1 and 3.1-2.
M00'S 1* and 2*f.
APPL!CABILITY:
E ACTION:
With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, either:
a.
Restore the control banks to within the limits within two hours, or b.
Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the group position using the above figures, or
~
c.
Se in HOT STAN08Y within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE RECUIREMENTS s
4.1.3.5 The position of each control bank shall be determined to be within the insertion ilmits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the Rod Insertion Limit MonitoF is inoperable, then verify the individual rod positions at least once per 4 hcurs.
"See Special Test Exceptions 3.10.2 and 3.10.4.
- With X,ff 3,1.0.
TROJAN-UNIT 1 3/4 1-23
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FRACTION OF RATED THERMAL POWER Fig u r e 3.1-I R od Bcnk Insertion Limits Versus Thermal Power Four Loop Op er a t i on h
'!/4124 Amend =en-Mc. H 50 Cetobe-6,1980
l INSTRUENTATION RADI0ACT!YE LIQUID EFFLUENT INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.10 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERA 8LE with their alare/ trip setpoints set to ensure that the Ifmits of Specification 3.11.1.1 are not exceeded.
APPLICA81LITY: As shown in Table 3.3-12.
ACTION:
a.
With a radioactive liquid affluent monitoring instrumen-tation channel alarm / trip setpoint less conservative than a value which will ensure that the limits of 3.11.1.1 are met, famediately suspend the release of radioactive liquid effluents monitored by the affected chanael or declare the channel inoperable.
b.
With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels operable, take the ACTION shown in Table 3.3-12. With the inoperable instrumentation channels not returned to operable status within 30 days, iden-tify the cause of the inoperable channels in the Semiannual Radiological Environmental Report in lieu of any other report.
- c. The provisions of Specifications 3.0J and 3.0.4 are not applicable.
SURVEILLANCE REQUIREE NTS 4.3.3.10.1 The setpoints shall be detamined in accordance with pro-cedures as described in the 00CM and shall be recorded.
4.3.3.10.2 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERA 8LE by performance of the CHANNEL CHECX, SOURCE CHECX, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown 1:3 Table 4.3-8.
TROJAN-UNIT 1 3/4 3-54 Amendment No. 99 December 20, 1984
- - -. =
l E,L,ECTRICAL P02EP. SYSTT:JS_
S,PflT,00'!!!
LntiTI::G CO::DITIC'! FOR OPCRATI0ff As a minimum, the following d.C. electrical power sources shall
~
3.8.1.2 bc OPERABLE:
4 One circuit between the offsite tran:mi::fon network and the a.
onsite ESF Electrical System and b.
One diesel generator set with:
1.
Two diesels driving a cor.non gencretor.
Day tanks containing a minimum of 1370 gallons of fuel.
2.
3.
A fuel storage system containing a mini:num of 33.000 gallons of fuel, and-i 4.
A fuel transfer pump.
s APPLICASILITY: H0 DES 5 and 6.
I
.../
ACTIO ::
j With less than the above minimum required A.C. electrical power sources l
OPEPACLE, suspend all operations involving CORE ALTERATI0i:5 or positiva reactivity changes until the minimum required A.C. electrical power t
~
l sources are restored to CPERAGLE status.
8 f
1 i
SU'WEILLAftCE REGUlr.CiU:T5 t
~
The above required A.C. elcetrical power sources shall be 4.8.1.2 demon:trated OPCPw*t0LE by the [:crfoltonce of each of the Surycillance Rcquircr:cnts of 4.8.1.1.1 and 4.8.1.1.2 cxecpt for requirement i
i
- 4. 8.1.1. 2. a. 6.
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A.C. DISTnicuTIO!! - SI!UTD0*.r!
Lil:IT!!:S CD::01TTO:: FOR OPERAT10*1 3.8.2.2 As a minimum, the following A.C. electrical busses shall be OPERAt:LE and energized from sources of po.cr other than a dicsc1 generator set but cligned to an OPERAti.E diesel generator set:
8 1 '- 41GO volt Emergency Bus 1~- 480 vcit Emergency Bus
[
2 - 120 volt A.C. Vital Busses i
APPLICACILITY: MODES 5 and 6.
~
i ACTIO!!-
!!ith less than th'e above ccmplement of A.C. busses OPERABLE and energi:ed, I
'^\\
establish C0i! Tall:F.E!:T INTEGRITY within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
r en. #
I j
SURVI!!.LA!:CE RE0'l!REi*E :TS r
The specified A.C. busses shall be dotermined OPERASLE and
~
4.8.2.2 j
energi cd from A.C. sources other than the diesel generators at least once par 7 days by verifying correct breaker alignment and indicated power availability.
I I.
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,5jij:,'lilj.I A!:CF P.rCifrS*Tf!TS (Continued),
l
' Z.
The pilot ec11 specific gravity, corrected to 77'F, is
~
> 1.19.
I 3.
The pilot cell' voltage is > 2.00 volts, and 4.
The overall hattery voltage is > 113 volts.
b.
At least once p'er 92 days by verifying that:
The voltage of each connected cell is > 2.00 volts un' der 1.
float charge and has not decreased mere than 0.13 volts fro:s the value observed during the original acceptance i
test, and The specific gravity, corrected to 77'F, of each connected 2.
cell is > 1.19 and has not decreased more than 0.02 from the value observed during the previous test, and 3
,., = =.
3.
Tha electrolyte level of ecch connected cell is between
- i]
- *j the minimum and maximum level indication marks.
At least once per 18 months by verifying that:
c.
The cells, cell plates r$d battery racks show no visual l
1.
indication of physical damage or deterioration.
}
The cell-to-cell and terminaT connections are clean, tight, 2.
free of corrosion and coated with anti-corrosien material, and j
3.
The battery charger will supply at least 197 amperes at E
l 122 volts for at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
.I At least once per 18 months, during shutdown, by verifying d.
that the battery capacity is adequate to supply and maintain in i
OPERADi.E status all of the actunt czargency loads for 30 minutos when the battery is sub,iected to a battery service test.
At least once per 60 months, during shutde.n, by verifying that the battery capacity is at 1 cast C0;; of the manufacturer's a.
This.
rating when subjected to a performance discharge test.
perfor:.ance discharge test shall be performed subsequent to o
the satisfactory cc=pletion of the required battery service j n, test.
- ,J TROM.
- :-t!!!!T 1 3/4 c-9
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