ML20127E632
Jump to navigation
Jump to search
| ML20127E632 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 06/05/1985 |
| From: | Joshua Berry, Keller R, Kister H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20127E590 | List: |
| References | |
| 50-293-85-15, NUDOCS 8506240604 | |
| Download: ML20127E632 (88) | |
See also: IR 05000293/1985015
Text
{{#Wiki_filter:,.
*
o
U. S. NUCLEAR REGULATORY COMMISSION REGION I
OPERATOR LICENSING EXAMINATION REPORT
EXAMINATION REPORT NO. 85-15 (OL)
FACILITY DOCKET NO. 50-293
FACILITY LICENSE NO. DPR-35
LICENSEE: Boston Edison Company M/C Nuclear
800 Boylston Street
Boston, Massachusetts 02199
FACILITY: Pilgrim
EXAMINATION DATES: May 13, 1985
CHIEF EXAMINER: o .
[-[[
J. eF , Rea~ct6r ngi er (Examiner) Date
REVIEWED BY: N) M/
R. M. Keller, Chief, Projects Section IC
!Y!W
Date
*
APPROVED BY:
H." B. KisteNr Chief, Projects Branch No. 1
b
'
[[
Bate
SUMMARY:
Operator licensing examinations were conducted at Region I on May 14, 1985.
One Reactor Operator candidate and three Senior Reactor Operator candidates
were administered written examinations. All candidates passed the examinations
with no generic deficiencies noted. -
.
8506240604 850619
PDR
-
ADOCK 05000293
G PDR
d
REPORT DETAILS
TYPE OF EXAMS: Replacement
EXAM RESULTS:
l R0 l SRO l
l Pass / Fail l Pass / Fail l
l l 1
I I I I
l Written Exam l 1/0 l 3/0 l
l 1 1 I
I I I I
l0verall l 1/0 l 3/0 I
I I I i
1. CHIEF EXAMINER: J. Berry
2. OTHER EXAMINERS: F. Crescenzo
,
,, --___-
i
r*
. .
o-
3. Changes Made to Written Exam During Examination Review:
R0 Exam
Question No. Change Reason
4.01 Grader will accept Question wording may have been
full credit if confusing to candidate.
candidate does not
mention E0P-8 as a
means of Rx power
control.
4.03 Delete " perform rapid Reference material was incorrect.
survey" from answer
key.
4.08 Change answer to "D0 Typographical error on answer key.
NOT place RFP sequential
trip...."
.
.
_ _ _ _ . _ _ _ _ _ _ . _ _ .
*
l '
C
SR0 Exam
Question No. Change Reason
5.02 Grader will consider Acceptable alternate answer if
candidates discussion accompanied by appropriate
of decreasing rod discussion.
worth for part "a".
5.07 Change answer to Samarium is a more specific
reflect a buildup of answer.
samarium vice fission
product poisons for
part "a".
5.10 Grader will accept Answer key presents only a
answers other than fraction of correct possible
those listed in key. answers.
6.03 Delete 1 ft/ min. Scenario given in question would
correspond to a level decrease
greater than 1 ft/ min.
7.06 Grader will accept Answer key reflects minimum
setpoints greater than setpoints which may not
those given in answer necessarily be actual setpoints.
key.
7.09 Change answer to "D0 Tyr graphical error on answer key.
NOT place RFP sequential
trip. .."
8.03 Grader will accept Acceptable answer if candidate
prompt notification assumes ECCS setpoint was
per T/S 6.9.B.1.A. exceeded.
8.06 Delete "at the end of Answer key incorrect.
the day" from answer
on part "a".
Attachments:
1. Written Examinations and Answer Keys
,
- _ _ - _ - _ . _ - _ . ..
._Z_______-_..__-. __. _ _ _.2 ._.._._.....__ _ _ _ _ _ __ ._.. :__._ __m
'
r
e.
s'
t'
U. S. NUCLEAR REGULATORY COMMISSION
SENIOR REACTOR OPERATOR LICENSE EXAMINATION
FACILITY: PILGRIM
_________________________
REACTOR TYPE: BNR-GE3
_________________________
DATE ADMINISTERED: 05/05/14
_________________________
EXAMINER: KVAMMEr J.
___________________-_-___
APPLICANT: h _k_ N k_____________
INSTRUCTIONS TO APPLICANf
__________________________ ,
Use separate paper for the answers. Write answers on one side only.
Staple question sheet on top of the answer sheets. Points for each
question are indicated in perentheses after the question. The passing
grade requires at least 70% in each category and a final grade of at
least 00%. E::a min a ti on papers will be picked up si> (6) hourr after
the e::am i.na t ion starts.
% OF
CATEGORY % OF APPLICANT'S CATEGORY
VALUE TOTAL SCORE VALUE CATEGORY
________ ______ ___________ ________ _____________________-_______-____-
'S 0 00 TilEORY OF NUCLEAR PONEP PLANT
_i_i_0___ _'5 ~_1__
_ ___________ ________ 5.
OPERATION'r FLUIDS, AND
TilERM0 DYNAMICS
5 5 PLANT SYSTEMS DESIGNr CONTROL,
_'i__.00_____'i__.00__ ___________ ________ 6.
AND INSTRUMENTATION
I
FROCEDURES - NORMAL, ADNORMAL,
_'5.00[______ _25.00_____ ___________ ________ 7.
EMERGENCY AND RADIOLOGICAL
CONTROL
_ _i____ _ l__ ___________ ________ 8. ADMINISTRATIVE PROCEDURESr
CONDITIONS. AND LIMITATIONS
100.00 100.00 TOTALS
________ ______ ___________ ________
FINAL GRADE _________________%
All work done on this e >:a m i n a t i o n is my own. I have neither
given nor received aid,
5EPl!EC5UTIS ~ 55GU5TUE~~~~~~~~~~~~~~~
___ ______________ .-
_
._o----------,--------~~--------------------
*
.
'
.
.'
5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 2
---- --------------------------------------
-_-_-_--------
DUESTION 5.01 (2.00)
The plant operating at 90% power shuts off extraction steam to the
highest pressure feedwater heater. A visitor, observing that turbine
load increased by 20 MNE after extraction steam was shut eff, con-
cludes that this action has improved the plant's thermodynamic
efficiency. Dc you AGREE or DISAGREE? EXPLAIN using relevant plant
indications to support your position. (2.0)
QUESTION 5.02 (2.50)
One of the factors affecting control rod worth is the condition of
the plant. In the following cases state whether control rod worth
INCREASES or DECREASES and JUSTIFY your answer.
a. Plant condition- cold to hot 1% power (1.25)
b. Plant condition- hot 1% to 100% power (1.25)
GUESTION 5.03 (2.00)
Considering Thermal Limits, is it sufficient to limit merely the '
LHGR's of each individual fuel in any 6' segment of a bundle to
the 13.4 kw/ft limit? EXPLAIN.
NOTE: DO NOT consider MCPR in your answer. (2.0)
.
QUESTION 5.04 (3.00)
In each fuel bundle there are various methods of heat transfer. In
addition, characteristic flow patterns, called flow regimes, exist
along the length of the bundle. Give the name of THRECE of five flow
regimes in the order of their occurence, and describe their
characteristics. (3.0)
( u :rx:x CATEGORY 05 CONTIMUED OM NEXT P AGE :nn*)
.
!
.
1
- _ - - - - _ _ _ _ _ _ _ _ - - _ _ ____-__ - _ _-
_ _ - . _ _ _ - . - - - _ _ _ _ _ _ - - . - - - - _ _ - - - . _ _ _ - - - - _ - - - - _ -- - - - - _ _ - - - - - _ _ _ _ - - - _ _ - - _ _
, _ _ _ - - - _
%.~...__,_~_____-___..___.__,__,_s--_- -- ,-
.
e'
.g 4
t
5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLU 1DSr AND PAGE .3
____ ______________________________________
______________
QUESTION 5.05 (3.25)
Assume that'the reactor is being started up with th'e bulk coolcnt
temperature being less than the saturation temperature. Suddenly
several control rods-malfunction and the reactor bcsins to increase '
~in power level on a short period.
a..Of the Void, Doppler'and Moderator Temperature coefficients
which would coine into ef f ect first, second and third to
lesson the rate of power increase? -(.25)
b.-EXPLAIN your choices of part a.
1.-Assume the operator takes no action.
2. Include a discussion of fuel time constants in your
answer.
3. Assume a scram does not occur. (3.0)
GUESTION 5.06 (2.00)
The attached figure 3.2-11r " Reactor Power VS Core Flow Operating Map"
illustrates how Core Flow changes with respect to Reactor Power on the
Natura1' Circulation Line. EXPLAIN WHY incremental increases in pouer
initially produce very rapid increases in core flow, but eventually
reach a point where further increases produce no increase in core
flow. (2.0)
GUESTION 5.07 (2.25)
With regard to the attached Figure i r 'l(ef f VS EXPOSURE' r EXPLAIN the
reason for'each of the areas labeled (a) through (c). (2.25)
GUESTION 5.08' (2.00)
The Reactor has been operating at 95% power for several days. An
' operator RAPIDLY reduces' reactor power to 60% by reducing the speed
of the recirculation pumps. During the next FEW MINUTEB (2-3 minutes)
the operator. notices that the reactor power slowly incr>aases approximately
,
3%. EXPLAIN the cause'of this effect.
i
:E (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
1
=
..____m_.____u - - _ _ ,_ - . _ _ . _ . _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _
'
s
S
*
.
.*
5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 4
____ _ ______________________________________
______________
QUESTION 5.09 (2.00)
The Reactor has just scrammed from extended full power operation.
Ten (10) hours later cooldown is completer and at that tiine the
Shutdown Margin is measured to be 1%. DESCRIBE the changes, if any,
to the Shutdown Marsin in the next 20 hours.
NOTE: Address in your answer whether a restart is a concern. (2.0)
GUESTION 5.10 (2.00)
a. Explain how the temperature of the Circulating Water System affects
condenser vacuum. (1.0)
b. List three (3) other factors that also affect condenser vacuum. (1.0)
QUESTION 5.11 (2.00)
TRUE OR FALSE:
For a constant reactor period, it takes the SAME AMOUNT OF TIME to
change reactor power from 1% to 5% as it does to change it from
10% to 50%.
EXPLAIN YOU ANSWER FULLY. (2.0)
l (***** END OF CATEGORY 05 *****) l l
.
O
t
. _ . . _ ._.______~~__a-----.--_------- l - __ -_--. ---- _ . - - - - - - -
.
*
'
.
l 8 I
6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 5
______________________________________________________
QUESTION 6.01 (3.00)
a. IDENTIFY FOUR essential loads from the "A Loop' of the Reactor
Building Closed Cooling Water (RSCCW) system. (2.0)
b. During a LOCA how are nonessential loadc on the RDCCW system
isolated? (1.0)
OUESTION 6.02 (1.00)
Which of the below statements properly describes the protective inter-
lock between the turbine and the Nuclear Steam Supply System that is
actuated by the hydraulic thrust bearing wear detector? (1.0)
a. It will energize the high vibration circuitry which energizes
the turbine trip relayr closing the main stop, intermediate stop,
control and intercept valves,
b. It will deenergize the turbine electrical lockout relay which
will deenergize the trip solenoid on the number 3 vacuum trip.
Deeenergizing the number 3 vacuum teip will close the main stopr
intermediate stop and control valves.
c. It will energize the turbine electrical lockout relay which in
turn will energize the trip solenoid on the number 1 vacuum
trip. Energizing the number 1 vacuum trip will close the main
stop, intermediate stop, control and intercept valves.
d. It will deenergize the ETS master trip solenoid which will close '
the main stop, control and MSIV's.
t
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
.. .---._-------..--------- -------- ._-. :- ---- - -- - - - - --- _ - - .- - s
'
.
'
.
.'
6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 6
______________________________________________________
QUESTION 6.03 (3.00)
With regard to the Automatic Depressurization System (ADS)*
a. What are FIVE indications other than annunciators to determine
if a Safety Relief Valve has lifted following a mair, steam line
] isolation from power? (1.25)
b. The reactor is operating at 100% power when a recire flow
detector line breaks inside containment. HPCI ic l out of service,
RCIC fails to start automatically and manually. All signals are
valid for ADS actuation except for the timer not timed out.
1. The operator attempts to reset a valid high dryuell
pressure signal (drywell pressure is 3.0 psis) with
its pushbutton. Will the ADS timer resot? EXPLAIN. (.75)
2. RCIC is now started and water level is raised to above
the lou level setpoint. What effect does this have on
ADS initiation if the timer has NOT t i:n e d out? If the
timer has timed out? Briefly EXPLAIN both. (1.0)
31
QUESTION 6.04 (3.00)
Consider the Source Range and Intermedi:3te Rcnge Monitioring Systems'
a. List the rod block (s) that initiate from the Source Range
Monitoring System. (1.0)
b. List the rod block (s) that initiate from the Intermediate
Range Monitioring System. (1.0)
c. What procedural condition (s) are required for the SRM
retraction? (1.0)
GUESTION 6.05 (3,00)
'
For the below listed Radiation Monitoring Systems identify the t' pe
of detector that is used. (3.0)
,
a. Main steam line radiation
b. Main stack effluent radiation
c. Main condenser air ejector offgas
d. RGCCW Loop (liquid)
' e. Standby gas treatment .
' (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
C
,
e
T
?---.-- . -- --- 4,-_-- i _ _ -.--~ ~ .~-... --.- _...- - ._. .-__ :-- --- -
-_-
'
l .
'
.
.*
6. PLANT SYSTEMS DESIGNr CONTROL, AND INSTRUMENTATION PAGE 7
------------------------------------------------------
QUESTION 6.06 (3.00)
Consider the Rod Block Monitor (RBM) Systemt
a. The average output signal from 'Ehe LPRM detectors is increased by
changing the gain of the RBM averaging amplifier such that the
local core power will be equal to or greater than the average
core power. Give TWO reasons for increasing the gain of the RBM
channel. (2.0) -
b. List THREE ways to bypass the RBM (manual or auto). Include
setpoints if applicable. (1.0)
OUESTION 6.07 (3.00)
The Feedwater Level Control System is being operated in 3-element
control using reactor level detector channel 'A'. Reactor power is
at 85%, steady state. For each of the instrument or control signal
failures listed below, INDICATE HOW REACTOR LEVEL WILL INITIALLY
RESPOND (increaser decreaser or remain constant) and BRIEFLY FXPLAIN
WHY in terms of what is happening in the Level Control System
immediately following the failure.
NOTE: Your answers should include the effects on Reactor
Level, Steam Flow / Feed Flow mismatch, feedwater valve
position. Consider each failure seperately,
s. "B" Feedwater Line Flow signal FAILS HIGil. (1.0)
b. Channel 'A' Reactor Level detector signal FAILS LOW (1.0)
c.-The electrical signal to the "B" Feedwater Regulating
Valve is lost. (1.0)
GUESTION 6.08 (3.00)
The Traversing In-Core Probes (TIPS) are used primarily to calibrate the
Local Power Range Monitors (LPRMs).
a. List and explain at least three (3) operational conditions or times
*
w mn the LPRMs must be calibrated. (2.0)
b. What is the purpose of the Nitrogen Purge System used for the TIPS and
their associated drive mechanisms? What adverse condition can exist if
this purge were inoperative for extended periods? (1.0)
(*2*** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
t.
_ . . _ _ _ _ . _ _ _ . .
_ _ _ _ _ _ _ . .
,
_ _ _ - _ _ _ _ _ _ _ . _ _ _ _ _ . . . _ . _ . _ _ . _ _ . ~ _ . . _ _ _ _ . . _ _ . _ _ _ . . _ _ _ . _ . _ _ _ _
. . _ . , _ , _ _ - _ , .
'
.
*
.
.'
6. PLANT SvSTEMS DESIGNr CONTROLe AND INSTRUMENTATION PAGE 8
______________________________________________________
00CSTTON 6.09 (3.00)
With regard to Low Pressure Coolant Injection (LPCI) mode of the
Residual Heat Removal (RHR) System;
a. What signals cause the initiation of LPCI? (1.0)
b. DESCRIBE the interlocks associated with containment upray
mode of LPCI/RilR operation. (2.0)
-(
(***** END OF CATEGORY 06 xx***)
'
.
_ . . _ _ _ _ _ _ _ _ w
,_______a__________-___-_._----._.-~..-----~l_---- - . _ - -
'
O
'
.
.'
7. PROCEDURES - NORMALr ABNORMAL, EMERGENCY AND PAGE 9
~~~~ - ~~~~~~~~~~~~~~~~~~~~~~~~
RAU50L55iCAL C5NTEUL
____________________
GUESTION 7.01 (2.25)
If durins normal plant power operation, a transient occurs resulting
in the establishment of a " Limiting Control Rod Pattern", WHAT THREE
(3) conditions are required to be met by Off-Normal Procedure 2.4.10
-Limiting Control Rod Patterns? (2.25)
GUESTION 7.02 (2.00)
During 100% power operations, the Off-Gas High radiation alarm is re-
ceived followed shortly by the Off-Gas High High radiation alarm.
a. WHAT AUTOMATIC actions, if any, are associated with these
alarms? (1,0)
b. If the Off-Gas High High alarm CANNOT be cleared, WHAT are
TWO (2) of the IMMEDIATE actions that must be taken? (1.0)
QUESTION 7.03 (3.00)
The station is supplying electrical power with a reactor power of 75%.
You have the followins symptoms in the control room.
1) Loss of power to the ARM's
2) Control rod withdrawal block occurs
3) CRD flow control valves fail close
4) The following instrumentation is without power: CRD Hydraulic,
Jet Pumps, RHR, Core Spray, HPCI, and RWCU.
a. WHAT electrical bus has lost power? (0.5)
b. WHAT will cause the reactor to scram? Include in your
answer WHY THAT scram occurs. (1.0)
c. WHAT actions .are requir ed when the reactor scram does occur?
Include in your answer WHY these actions are taken. (1.5)
GUESTION 7.04 (2.50)
WHAT are FIVE (5) conditions that require the use of a RWP - Padiation
Work Permit? (2.5)
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
,
,
_ _ _ _ _ _ _ _ _ _ _ _ _
J
. _ . _ _ ___
_ _
_ _ _ _ _ _ _ _ _ _ _ - - -
.m . __
.,_ ~ - u______ = _ , - __= - _ _ .a __ - _-_ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _
_
--
.. ~*
*
.
.*
!
7. . PROCEDURES - NORMAL, ABNORMALr EMERGENCY AND PAGE 10
~~~~ ~ ~~~~~~~~~~~~~~~~~~~~~~~~
i
RA656EU556AL EUNTRUL
____________________
1
QUESTION 7.05 (3.00)
In reference to procedure 2.4.143 - Shutdown From Outside Control Room
'
b .Due To Inhabitability of Control Room'
a.. WHERE are the Operators -AND Operating Supervisor directed to
go followins assembly in the 23' 4kv switchgear area?. (1.0)
b.. WHAT is the preferred way to scram the reactor AND WHY is
4 it the preferred method? (1.0) l c .- WHEN should the reactor feedwater pumps be tripped? (1,0)
GUESTION 7.06' (3.00)
-In reference to procedure 2.4.36 - Loss of Condenser Vacuum *
, a. WHAT are FOUR (4) AUTOMATIC actions that will occur on a de-
! creasing condenser. vacuum? Include any applicable setroints.(1.4) L b. WHAT are FOUR (4) of the IMMEDIATE operator actions for a
loss.of: condenser vacuum? Assume steady state operation. (1.6)
QUESTION 7.07 (3.00)
The Control Room'0perator is in the process of initiating shutdown
cooling when the shutdown cooling mode valves (MO-1001-47 and 50) ,
automatically closet
a. WHAT are TWO (2) conditions that would cause the shutdown
,
cooling valves to close? State any assumptions you make. (1.0)
.
b. WHAT are THREE (3) of the IMMEDIATE operator actions on a
Loss of Shutdown Cooling? (1.5) ,
c. If forced c.rculation is lost during cooldown, WHAT should
be done to prevent temperature stratification in the
reactor-vessel? (0.5)
.
$
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) .
1
- !
1 J I i e a
I _ _ . _ . - . . . _ . _ _ . _. _ _ . _ . _ _ _ __ _. __ . . _ . _ _ _ _ _ _ . . _ _ . _ _ _ .- _ _ _ _ . ~ . - . . , _ _ . -
- - --
_ . .
_ . - - , _ . - - - - - - - . . _ - - - ~ - - - - - - - - . . - - - - - - - - - ---
- . _ _
=
.
*
.
.
7. PROCEDURES - NORiiALr ADNORMAL, FMERGENCY AND PAGE 11
~~~~RA5iBE55iEEE E5NTRBE~~~~~~~~~~~~~~~~~~~~~~~~
____________________
QUESTION 7.08 (2.00)
a. In accordance with procedure 2.1.1 "Startup From Cold Shutdount"
under what conditions shall secondary containment integrity be '
maintained? (1.5)
b. You are allowed a certain period of time after placing the reactor
in the RUN mode before the primary containment atmosphere oxygen
is required to be less than 5% by weight. How long is this period? (0.5)
GUESTION 7.09 (3.00)
According to procedure 2.2.96 " Co n ~' .nsa t e and Feedwater Systemr" during
startup of the feedwater pumps:
a. What is the required status of the condensate pumps and controls? (1.00)
b. What precludes startup of the R.F.P. without lubrication? (1.00)
c. How is the oil temperature maintained at 110 degrees? (0.5)
d. When should the second feed pump be placed in service? (0,5)
QUCGTION 7.10 (1.25)
During Reactor heatups and cooldowns, THREE temperatures must be
permanently logged every 15 minutes by PNPS Technical Specifications.
a. What are these temperatures? (.75)
b. When may these logs be discontinued? (.5)
.
(***** END OF CATEGORY 07 *****)
i
i
l
l
I
i
l
I
,
_.._ _
- - . _ ._ _ _-
___._.___ _ _ ___.- __ _ , _, ____ _ __ - _ _ - -- --- -- - --m_ __ - - =
'
,
*
.
.*
8. ADMINISTRATIVE PROCEDURES, CONDITIONSr AND LIMITATIONS PAGE 12
__________________________________________________________
i-
GUESTION 8.01 (2.50)
a. Describe how a Safety Limit and a Limiting Safety System Setting
relate to each other. (1.5)
b. What does the term " Limiting Condition for Operation" mean? (1.0)
QUESTION 8.02 (3.00)
a. In what situation (s) would a temporary change to procedure
implemented by an SRO NOT be authorized? (1.0)
b. A temporary change implemented by SRO may be made provided the
change is approved by two members of the plant management staffr
at least one of whom holds a Senior Reactor Operator's license
on the unit affected.
1. WHO is considered " Plant Management Staff'? (1.0)
2. What other requirement (s) is(are) necessary before a
temporary change implemented by SRO is allowed? (1.0)
QUESTION 0.03 (3.00)
Given the following Reportable Occurrences and the attached Technical
Specifications INDICATE which TYPE of REPORT would be applicable. INDICATE .
which TECHNICAL SPECIFICATIONS STEP is applicable.
a. A procedural inadequacy which when placing the HPCI system in
a standby status prevents the system from reaching full flow
capacity on an initiation. (1.0)
b. A surveillance test of the IRM showed that the scram trip
settings were all at 123/125 of scale.(on all channels) (1.0)
c. An off shore oil spill occurs that results in oil in the Intake. (1.0)
.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
.
t
e
_.a- ___,____-._____--____.__w_.- - _ _ _ _ _ _ ~ _ _ _ ~E Z - ~
.
*
*
.
.*
0. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 13
__________________________________________________________
QUESTION 8.04 (2.00)
What FIVE documents must the off going shift prepare prior to
shift change, according to Procedure 1.3.34- " Conduct of Operations"?
(2.0)
QUESTION 8.05 (1.25)
WHEN must a SCRAM REPORT be filled out and WHO is responsible for
ensuring this is done? BE SPECIFIC (1.25)
OUCCTION 0.06 (2.50)
a. What is the purpose and procedure for installation and removal
of a red tas? (2.0)
b. When is a Red Tag Los utilized? (.5)
OUESTION 8.07 (3.00)
The following data was taken during two days of operation at
approximately 100% rated thermal power. The unit has been at that
power for two weeks.
Identified leakage unidentified leakage
Day 1, 0000-0800 9907 gal. 1480 gal.
0800-1600 10,560 gal. 1008 scl.
1600-2400 10,440 gal. 1824 gal.
Day 2, 0000-0800 9398 gal. 2198 gal.
0800-1600 9600 sal. 2198 gal.
,
1600-2400 9278 gal. 2496 sal.
NOTE! All integrators are inop and coolant leakage rates are being
done manually.
STATE the TECHNICAL SPECIFICATION Coolant Leakage LCO limit (s) applicable
in this plant condition and IDENTIFY any that were exceeded during these
two days.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
-
_ _ _ _ _ _ _ .
--
_ _ _ _ _ _ _ _ _ _ . . _ , . _ _ _ . _ . . . _ _ _ _ _ _ _ _ _ _ ~ . _ _ . . _ _ ~ . _ _ _ - - - - _ _ - - - -
*
.
*
.
.-
O. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 14
__________________________________________________________
GUESTION G.08 (1.75)
On 5/14/85 at 0700 the differential pressure between the drywell
and the supression chamber goes to 1.0 psid due to HPCI testing
and can not be restored above that point even after HPCI testing
is completed. On 5/14/85 at 1000 the quarterly cycling of the scram
discharge volume drain and vent valves commences. At 1930 it is
'
determined that the drain valves is stuck shut and will not reopen.
Give the DATE AND TIME that the reactor MUST be in cold shutdown.
JUSTIFY your answer.
NOTE: The reactor has been operating at rated conditions
GUESTION 8.09 (3.00)
What access control requirements exist for the following hi3h radiation
areas?
a. An intensity greater than 100 mrem /hr but less than 1000 mrem /hr (1.00)
b. An intensity greater than 1000 mrem /hr but less than 10000 mrem /hr(1.00)
c. An intensity greater than 10000 mrem /hr. (1.00)
QUESTION 0.10 (3.00)
a. When according to Technical Specifications is the RPT/ARI sytem
required to be in operation? (1.5)
b. What are the trip setpoints? (1.5)
a
(***** END OF CATE90RY 00 *****)
(************* END OF EXAMINATION ***************)
i
_ _ _ _ _ _ _ _ _ _ _ _ _ - _
w . . ,
. . . . . <
. .
. . , -.-~ __-.._-
,
.- ,
_.__ _
*
. .
s'
Note: Keff for cold, Xenon free, all rods out,
Gadolinia, equilibrium Samarium
1.16 -
(a)
.15 -
1.14 - c)
Ke ff
1.1, -
1.12 -
1.11 -
1.10 . . . . . . . . . .
0 1000 2000 3000 4000 5000 6000 7000 8000 9000 10000
Exposure (MW4/T)
Figure 1. Keff vs. Exposure
,
O
A
R
__
L ___ J u._ . _ - - - - - - _- _ - _ _ . _ - - = -
-
.
,..
...
TEST CROSS REFERENCE- PAGE 1
QUESTION VALUE REFERENCE
________ ______ __________
05.01 2.00 JCK0000410
05.02 2.50 JCK0000413
05.03 2.00 JCK0000,415
05.04 3.00 JCK0000417
05.05 3.25 JCK0000418
05.06 2.00 JCK0000419
05.07 2.25 JCK0000420
05.08 2.00 JCK0000442
05.09 2.00 JCK0000443
05.10 2.00 JCK0000444
05.11 2.00 JCK0000445
______
25.00
-06.01 3.00 JCK0000422
06.02 1.00 JCK0000423
06.03 3.00 JCK0000425
06.04 3.00 JCK0000426
06.05 3.00 JCK0000428
06.06 3.00 JCK0000429
06.07 3.00 JCK0000430
06.08 3.00 JCK0000446
06.09 3.00 JCK0000447
______
25.00
07.01 2.25 JCK0000401
07.02 2.00 JCK0000403
07.03 3.00 JCK0000404
07.04 2.50 JCK0000405
07.05 3.00 JCK0000406
07.06 3.00 JCK0000407
07.07 3.00 JCK0000408
07.08 2.00 JCK0000448
07.09 3.00 JCK0000449
07.10 1.25 JCK0000450
______
25.00
08.01 2.50 JCK0000432
08.02 3.00 JCK0000433
08.03 3.00 JCK0000434
08.04 2.00 JCK0000436
08.05 1.25 JCK0000437
08.06 2.50 JCK0000439
08.07 3.00 JCK0000440
08.08 1.75 JCK0000441
08.09 3.00 JCK0000451
08.10 3.00 JCK0000452
______
25.00
______
______
100.00
,
_ - . . . . _ _ .-.__. _ _ _. l
- - - - - ..
.- _
.
- -
,
.. .
. .
.*
- EQUATION SHEET
-
*
! = ma v = s/t Cycle a#iciency = (.tetwork
cut)/(Energy in)
2
w = ag s = V,c * l/2 at
= x- .
A=Ae*
~
A = U3
KE = 1/2 mv 3 = (Vf - 13 )/t 3
PE = mgn
vf= V, + at * = s/t x = zn2/tjfg = 0.693/t1/2
y,,j -
n0 2
t
1/2*" " [l* m }(b}}
A* 1
4 C(t1/2) * (*bIl
# * 93I '"
m = V,yAo g,ge -D
. .
Q = mCpat
Q = UAa.T
I * I o* ,,, ,
P w = M ,.h I*l o 10-*O.
'
*
Til. = 1.3/s
- sur(t) gyt = .;.533/2
P = P,10
7 = Po e*/I
SUR = 25.06/T SCR = S/(1 - 4 gf)
CR x = S/(1 - Keux)
SUR = 25s/1' + (s - o)T CR j (1 - 4,ff3) = G 2 (I ~ e#2
I
T = ( t*/s ) + ((a - o V ia] M = 1/(1 - 4,g) = CR;/C3 3
'
T = 1/(s - a) M = (? - K ,ffo)/(1 - 4,ff))
T = (a .a)/(Ta) SCM = ( - K,g)/K,g
'
a = (K,g-1)/K,g = r.Kgf /K df t' = 10 secanos
i = 0.1 seconds-I
o = ((&*/(T K,g)] + (a,g/(1 + IT)]
Idli*Id
Id j
2 ,2 g#
2
P = (uV)/(3 x 1010) 22 2
3 = sti R/hr = (0.5 CE)/d (meters)
R/hr = 6 CE/d2 (feet) ,
Watar par?ceters Mises11aneous C nve.sions
I gal. = 8.345 lem. I curie = 3.7 x 1010:33
1ga}.=3.78litars
1 f. = 7.48 gal.
1kg=2.2110m}Stu/hr
I ap = 2.54 x 10
Oensity = 62.4 lbn/ft3 1 m = 3.41 x 100 5tu/hr
Oensity=1gm/cM lin = 2.54 cm
Heat of vacorization = 970 Stu/lem *F = 9/5'C + 32
Heat of fusion = 144 Stu/lem 'C = 5/9 (*F-32)
1 Ata = 14.7 osi = 29.9 in. Hg. 1 BTU = 778 ft-lbf
I ft. H O 2
= 0.4335 lbf/in.
. -
,
9
"
.
- - - - - - - - - - - - - - - - - - - - - - - - - - - - -- -- - - - - - - - - - , . , - -
,-
.C_ a r - M an ' mM 4D 6.%M'
e
, I e
9
- . . 6e w w
PERCENT RATED THERMAL POWER
.
o 5 3 o a $ 8 3 3 8 8 5 3
*
-
1 1 l ll l i i i I I I I I
i
i i
i :- -
3 i
l:
_
.
:
! I mnz
l
I 53>
mn*
I CC
w t- 2
-e I I
>
l 2 ">
l i C >
z 5
n'
l
3 l
E
O w l 8m
c *
..
.g ,
-
E Q OE
m %:
.m::N5'h.hl Oy h
-
ik-
fd Ri. ..% . '. N 2x r-
9 4' I$.hs~. 5g z
a . l 0C m
* $ =
.ny +. -m
W-M.-t, .e: t~ E "
i gi. pe :-
r3
3
.
-
-
e m
1 2F
g
su
_
m r-
2 di O
*
n . . . . 4 W ., g
$ m M-
'
y
b
~
, ** "
E -# .' , - t=
9 = 1
5 + .u2- w - O >
g >
-
g M
m
-
{ :o32
,
-
7
2
h O
C
g
.
m
,ym'j} 3
g
E
:C .g.
o mo
3 e
- n -
p.
o
s
*
,
, a=
,,
.
%
g
m
=
-ii
.cc g :5 - .
z m--
5
.~
o e
=
e
e
,,
-
'" O
r* *C r=
9 m
*4 M ? .'- n
Q e z- -
3
y
o b 1 -
g . . - - m o
C y . .r .:- z
h ? O ^ $: lV $
5 , a 2 :. ,. :. a o
as 1 :D ,j( 2 LN.
'
F
c. o' n - , . r*
3 e Z g .$ '.O.P S D' '. -
E
"
e o r- C . : 2 ' .- . m
, 2 .'!. -
. :. n WCw':g
.
4 ~ z
C 3 ?$z ny,ks m 3g ,-,
z "O
m
Z - - - -
-
oo' _
$ ~~
-
nh
ax
Cb
z~
d z *#
4 m -
Z
Q E
m
2 o2
en ,
"
- r= -4
8"* -=
zO
m U
_
o
l l l l l l l l I l I
I
- . - . -
- - . . - - . . . _ , _ , . . - - - - , - - - - - - - - , . - - -
_ . _ _ - ~ ~ -
_ _ . . _ . - - - ~ . . . . - ----
.
, w........... w... .i tu:s i-iaR UP ERATION i
*
* i E'.'RVHIL!)W- P.'l';!Ii P 3' ':NT
t
3.3 _Fr).CTIVITY CGNT 0'
\
, ! 4.3 REALTIVITY CD.';'70L
g
Ann!!cabiliev:
_
-
_
} Annlicability:
{y
,
:-
Applies to the ope ::ienal status
of the control rod systen. e Applies to the surveilitnce require-
Objective: 1 cents of the centrol rod systee..
l, Chice ive:
-
k
To assurs the ability of the con' ,
trol rod systen to con:rol reac- To verify the ability of i.h: control
tivity.
rod systen to :catrol reacci ity.
.
Specificati:n:
t Specification:
A.
Rescrivitv !.! itatiens A.
e
_
Reactivity Limitations
1. Resetivity harcin - core
loscing 1. Reactivity r_ar-in - cor-
loadinq
: -* The core loadin3 shsil be
li=ited to that which can [ Sufficient control rods shall
be cade subcri:ie:1 in the j be withdrawn followin; a re-
cost rea::ive ::niitica ' fueling outsge wher. core
during the cp-rating cycis alters:1cns were e .rformed '
with the strenges: cperable to deconstrate ith a nar-
A control red in its full- gin of 0.25 percent t.k that
out position and all other .
the core can be nade sub-
operabic rods fully in- c
crit'ic 1 at sny time in th-
{]) -
scrted. !
(
subsequent fuci cycle with
I
the stren;-se oper:bl e con-
trol rod fully wi: drawn
l and all other operable r:ds
y
. fully inser cd.
2. R::ctivity c:r-ir - 1:ccer- 2
,
-able c- 3, 2.
--! :* Ls
,1 Reactivi:v carcin - inene -
. -
able centrol roos
_
g
s. Control rod drives
,
}
which cannot be noved i Each partially or fully with-
,
with centrol rod drive 3
drawn opernble con: el red
pressure shall be con- N shall be exercised one r.otch
sidered in:perablo. If at least once cach week.
c partially or fully h This test shall be perfor:ed
I
l
withdrawn con:rol red l' at least once per 2. Scurs in
drivo cr.nnot bc :oved withJ the event power operatien in
e
--
drive or sers: pressure il continuing with three or = ore
' the reac or shall be 6
;
inoperable centrol rods or in
$ brough: :o a shu:down fi
the event power operatica is
0 c:ndition within 43 hou:s l' continuing with one fully or
15
unless investign:ica n
partia1Jy wi:hdrawn red which
h der:..strates t'.at the D enr.not be coved and for which
* cr;ac of the failure is .'i centrol rod drive techanism
n:: due :3 a failed con- y danage has net been ruied out. j i
gsM * :::1 ::d drive cc:har.is o The surveillance need nar be
::lle: housing, h completed within 21 hears if thei
t.
u
nu2ber of incperable rcJs has
n
F t
l
, k
,
. i Ar.endmunr. I;o. 1 *. 80
__ .
, _ _ . - . . . . - - - - -
.. . . . - -
i
O 4'
.
-
. = ,
.. _ ,. , . . . ~~
_ _ . _ . _ . . . _ _ . ..
- .
,
. . . . . . . . . . . ww . . u u . . .;. roA UPI:PArIO.'i
. .* . l__ st:3y.11.I.t.rC!i s'!!'e.'!Trr
*
.
3.3.A F.UA.'TIVITI C%T'1'2f. 4.3 RP *.CTIVITI CC:r P.01. _
.g b. The c ntrol rod dir:c - '
i tional control valves I been reduced to Ic:s -
for inopwabla centrol than three and if it '
-
'
red: th.d1 bo disarmed has b:en de cns rat.:d
-
electrienlly at:d the that control rod _
control red.a .::h:11 be drive cachanisc colle: "
in such pc.iltic.u that housinr. failiare i.s not ;
Specifics:ica J.!,.A.1 cause of an.i.- ovable -
-
is net. control rod.
1
*
c. Control red drivo: l
which are fully in-
t.crted and electrically
. disar=ed shall not be
considered inoparabis. -
f
.
d.. Control rods with scrr.a l
times greater than
* . .those permitted by l.
.
Specification 3.3.C.3
: --
are inopersble, but l
i .
if they can be noved I
with control rod drite {
pressure they need not i
be disarmed electri- l
' cally.
-
. -
e.
I'
.
Di: ring reactor pcver
operatica, the nu .ber l
.
of incperabic cont:01 ,
rods shall not .: ceed !
U* E C t"'l P E
eight. Specificatica
3.3. A.1 cust ba met , 1. The coupling in:e rity aned2.
at all times.
7 be verified fcr 9ach with-
. e denvn control re:i as fel.::rvs:
j .
B. Control M ds a. When the red i: vithdrsv::
the first time sub:equent
1.- Each control rod chr1.1 bo #
co'.:pl:1 to it* drive or '
completely in:er*.e i :.nd *
the control rod direc-
ti.:.01 or control val.as in: t ritten tat ion. lic rev- r ,
dinnr..cd *:1cetrit:a1Ly. for ir.itiL rod when
i
.
Thic require:en'. c'9c: no t .
ap;.ly in the re.^ael co .- ible, cut:ettent exerci:1:g .
ditie when the rer. tor of t ,,:e
. . . o6 . .. .er ,..%..
la "e.:e1. "Ns ecutral
ro:. 2. ves r.sy be tr~cve. s *
I
.
C - ~.3 f==
,
.a;'eCil'icar.lo inct:wenta .ian rego . e.
*
,'.].A.1 18 ~12 * .
*..
Amendment :o. 14 al
.
M -
9
;__..--
"
_
._-
_ . _ _ . . . ._ _._ . . - --
,
, .* .
e'
CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS
3.3.B Control Rods 4. 3. B Control Rods
,
2. The control rod drive hous- b. When the rod is fully
ing support system shall be withdrawn the first
in place during reactor power time subsequent to each
operation and when the rr.as- refueling outage or af-
tor coolant system is pres- ter maintenance, observe
surized above atomspheric that the drive does not
pressure with fuel in the go to the overtravel
reactor vessel, unless all position.
'
control rods are fully in-
serted and Specification 2. The control rod drive housing
3.3.A.1 is met. support system shall be in-
spected after reassembly and
3. a. No control rods shall the results of the inspec-
be moved when the reac- tion recorded.
cor is below 20% rated
power, except to shut- 3. Prior to control rod with-
down the reactor, unless drawal for startup or inser-
the Rod Worth Minimizer tion to reduce power below
(RWM) is operable. A 20% the operability of the
maximum of two rods may Rod Worth Minimizer (RWM)
be moved below 20% de- shall be verified by:
sign power when the RWM
is inoperable if all a. verifying the correct-
other rods except those ness of the control-
which cannot be moved with rod withdrawal sequence
input to the RWM com-
-
control rod drive pressure
are fully inserted. puter.
b. Control rod patterns and b. performing' the RWM
the sequence of withdrawal computer diagnostic
or insertion shall be as- test.
tablished such that:
c. verifying the annunci-
1) when the reactor is ation of the selection
critical and below errors of at least one
20% design power the out-of-sequence control
maxi =um worth of any rod in each distinct
insequence control RWM group
rod which is not elec-
trically disarmed is d. verifying the red block
less than 0.010 delta function of an out-of-
k. sequence control red
which is withdrawn no
2) and when the reactor more than three notches,
is above 20" design
power the maximum worth
of any control rod, in-
cluding allowance for
a single operator error,
is less than 0.020 delta
'
a) l
Amendment No. 39
i
82 ,
,,
__
!
.- --
l
,
_ _ . _
y ,
'
'..'-
'
.
3.3.5. Control Reds 4.3.8, Control Rods '
.- !
4. Control rods shall not be with- 4. Prior to control rod with-
drawn for startup or refueling drawal for startup or
unless at least two source range
.-
channels have an observed count during refueling, verify
h -
rate equal to or greater, than
three counts per second.
that at least two source
range channels have an
observed count rate of at
5. During operation with limiting least three counts per
control rod patterns, as deter- i
second.
.
mined by the Reactor Engineer. 5.
either: When a limiting control red
pattern exists, an instru-
a. Both RBM channels shall be ment functional test of the
operable: or RBM shall be perfomed
.
prior to withdrawal of the
b. designated red (s) and
Control red withdrawal shall daily there,after.
be blocked: or
c. The operating power level
shall be limited so that
the MCPR will remain above
the Safety Limit MPCR
assuming a single error
that results in complete
withdrawal of any single
operable control rod.
C. Scram Insertion Times .
C. Scram Insertion Times
O 2- Ta av ras scr== 4"= rtioa
. time, based on the deener- t- rolio 4as
..
ch r ru 14as
outage, each operable
,
gization of the scram pilot
valve solenoids as time zero, control rod shall be sub-
of all operable control rods
.
.. -
jected to scram time tests
in the reactor power opera- from the fully withdrawn
position. If testing is
tion condition shall be no not accomplished with the
greater than:
nuclear system pressure
5 Inserted above 950 psig, the
Average Scram
From Fully . measured scram insertion
Insertion time shall be extrapolated
Withdrawn Times (set)
.
to reactor pressures above
10
950 psig using previously
.55 detemined correlations.
<
30 1.275
50
Testing of all operable
2.00 control rods shall be com-
90 3.50 pleted prior to exceeding
40t rated themal power.
*
.
h.) -
Wendment No. 75 68
83
..
- - ~ - - . - - _ _ . . -______-_________ -
._- - . -. . . =
7
.
_ . . . n w. .- ,<.. - -
*m
.* ,
.* - I
l
1
IJXITIN3 CC'CI?!0N FCR ONTIONS SURVETLIAN E RE:UIPDdCf"
'
_
g .3.c Scree Insertion Tice Is .3. C Sere = Tnsertion Ti=e
3.. The average of the scraa 3. At 16 week intervals, 5c?
-
insertion times for the of the control red drives -
three fastest control rods *
shall be tested as in L.3.c.;
,
. of a u groups of four e=n- so that every 32 weeks a u
trol rods in a two by two
array sha2.1 be no greater of the control rods sha n
than: - have been tested. Whenever
Sc5 of the control red drives
have been scran tested, an
(Inserted Avg. Seram evaluation sha n be made to
From Ful.ly Insertion provide reasonable assursnee
Withdrawn Ti=e See,
that proper control rod
10
drive perfer=ance is bei=g
58 maintained.
30 1.35
50 3.12 .
Po 5 30
3 The - '-- scram inser-
-
tion time for 90fa inser- .
tion of any operable con-
trol rod shan not exceed
7.co seconds.,
O m. co=tro1 =od xe=. . Lata =s .>. Co.t==1 ==4 >=====:a===s
-
, .
At au reactors operating pres- once a shift, check the status
sures, a rod accu =ulater any of the pressure a=d level ale.r=.s
be inoperable provided that no for each ace ==.tlater. 7
other control rod in'the nine-
rod square array aroung this .
rod has a: .
*
* 1. Inopershkaaccumulator. , , ,
2. Strestional control valve *
electrically disarmed
while in a non fully ta-
serted posittoa.
3. Scraa insertion time *
sfester than the nazimus
permissible insertion ,
, time.
"
*
d
If*a control rod with as * *
, inoperable secu:rulater is ,
*
'
,
-
inserted " full-in" and les *
direerional control valves
g./ are electrically disarmed,
it shall not be considered
ce have an inoparable 34
occumulater. *
a
-
.. . . _ . . . . . ..
'
~
Amendment No. 65
..
-aw-
---
, . . , -
_ . - .
* ,."
,
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
.-
E. Reactivity Anomalies E. Reactivity Anomalies
I
The reactivity equivalent of During the startup test program
the difference between the and startups following refuel-
(,) actual critical rod configur- ing outages, the critical rod
ation and the expected con- configurations will be compared
figuration during power to the expected configurations
operation shall not exceed 1" at selected operating conditions.
4 K. If this limit is exceed- These comparisons will be used
ed, the reactor will be shut as base data for reactivity
down until the cause nas been monitoring during subsequent
determined and corrective ac- power operation throughout the
tions have been taken if such fuel cycle. At specific power
actions are appropriate. operating conditions, the
critical rod configuration will
be compared to the. configuration
F. If Specifications 3.3.A expected based upon appropriately
through D above cannot be met, , corrected past data. This com-
an orderly shutdown shall be parison will be made at least
initiated and the reactor every full power month,
shall be in the Cold Shutdown
condition within 24 hours.
Specifications 3.3.A through
D above do not apply when
there is no fuel in the
reactor vessel.
G. Scram Discharge Volume G. Scram Discharce Volume
1. The scram discharge 1. The scram discharge volume
volume drain & vent drain and vent valves shall
valves shall be operable be verified open at least once
whenever more than one per month. Each valve shall
operable control rod is be cycled quarterly. These
withdrawn. valves may be closed intermit-
tently for testing under
2. If any of the scram dis- administrative control
charge volume drain or
vent valves are made or 2. During each refueling outage
found inoperable an verify the scram discharge
orderly shutdown shall be volume drain and vent valves;
initiated and the reactor
shall be in Cold Shutdown a) Close within 30 seconds
within 24 hours. after .ceipt of a reactor
sciou signal and
b) Open when the scram is
.
reset.
.
.
Amendment 65 85
.
n
__ _
. ... ~ . . . . . . - - .
.. . _ _ _ . _ _
.
, .
_ . _ . , . ,
, .
, .-
.- .
O .
.
.
<
.
.
(This Page Intentionally Left Blank)
O .
9
.
e
e
?
%
e
s
%
O
e
..
6 GD N S Se Ge e g
' * e
g 4- . _ .
- - - .
._. .__ . _ . . _ . . _ _ . -
- _ _
_
_ - - . . . - . - -
. _ _ . - --. .- -- .
_ .-
_ _ _ _ . _ - _ _ _ . . . . _ _ _ _ _ .
, ,, . _ , . .
7
.
'
3.3 snd 4.3
,B ASES :
A. Reactivity Limitation
1. The core reactivity limitation is a restriction
to be applied principally to the design of new
fuel which may be loaded in the core or into a
particular refueling pattern. Satisfaction of
the limitation can only be demonstrated at the
time of loading and must be such that it will
apply to tne entire subsequent fuel cycle. The
generalized form is that the reactivity of the
core loading will be limited so the core can be
made suberitical by at least R + 0.25:Ak at the
ti=e of the test, with the strongest control rod ,
fully withdrawn and all others fully inserted.
The value of R in Mk is the amount by which the
core reactivity, at any time in the operating
cycle, is calculated to be greater than at the
time of the check; i.e. , the initial loading. R
must be a positive quantity or zero. A core
which contains temporary control or other burn-
able neutron absorbers may have a reactivity
characteristic which increases with core life-
time, goes through a maximum and then decreases
thereafter.
'Ihe value of R is the dif ference between the cal-
culated core reactivity at the beginning of the
operating cycle and the calculated value of core
reactivity any time later in the cycle where it
would be greater than at the beginning. The value
of R shall include the potential shutdown margin loss
assuming full B4 C settling in all inverted poison
tubes present in the core. A new value of R cust
he determined for each full cycle.
The 0.25*Ak in the expression R + 0.25%Ak is
provided as a finite, demonstrable, suberiticality
margin. This margin is demonstrated by full with-
.trawal of the strongest rod and partial withdrawal
of an adjacent rod to a position calculated to in-
aert at least R + 0.25*.ak in reactivity, or by an
insequence, xenon-free cold critical measurement
'I# to demonstrate at least R + 0.25% A k in reactivity with
the most reactive control rod fully withdrawn.
Observation of subcriticality in this condition
assures subcriticality with not only the strongest rod
fully withdrawn but at least an R + 0.25 Ak mar-
gin beyond this.
M
1
*
. ... .
_
*
'8
_ _ _ _ _ - _ _ _ _ _ _ - _ . . -
.
_.
=
_ . , . _ , _ _
_ ._. . _ _ _ . . . .
_ _ _ . . . . -- --
. _
,,.
_
.
. . 3.3 :.ad J . 3
BASE 3: ..
.
o
e
2. .Reacti.vity earsin - inensrable centrol rods,
.Spect:ica:1cn 3.3.A.2 recuires :nat a ra:1 be taken out of service
if it cannct be cioved with drive pressurs. If the rod is fully
inserted and then disar=ed electrically *, it is in a safe positica
" of.:sximum con:ribution to shu: dawn rese:ivity. If it is disarmed
ele:trically in a n=n-f'ully inser cd positien, that position shall
bc consist:n: with the shut.ds-a reactivi'.y limitation s::: d in
Specification 3.3.A.1. This assures that the core can bc :hutdown
at all ti:es with the rc=sining control rods assuming the stronges:
operable control rad does not insert. An allow:bic pattern for
control reds valved ou of service, which shall meet this Specifica: ion.
will be de:er=ined and made availabic to the operator. H e nur.ber of
rods per=i : d to be ineperabie could be many more than the ei2 h:
allowed by the Specification, particularly 1 :: in the opera:Icn cycle;
however, the occurrenc: of core than eight could be indicctive of a
generic control red drive problem and the recctor will be shut dom. s
Also if dams;;c within the centrol red drive mechanism and in particular,
cracks in drive internal housings, cannot be ruled out, then a generic .
,_
'
probics affecting a nt:..ber of drives cannot be ruled out. Circumferential .i
cracks resulting frca stress assisted intergranular corrosion have occurred i,
in the collet hou. sing ci drives at s:veral S'.i'!s. his type of cracking I
could o::ur in a nu .ber of drives and if the cracks propagated until E.
severance of the collet housing occurred, scram could be preven:cd in s
the affe::cd roc's. 1.initing the period of operation with a potentially l ';
severed colle: housin;: and requiring iner:ssed surveillance after d::ceting g
o
one stuck rod will assure tha: the reactor vila not be cperated with Isrge i
number of rods with failed colle: housings. ,,
I
B.. Con' trol Red Withdravai ,
1. Control rod dropout accidents as discussed in th: FSAR can Icad to
significant core da=sge. If coupling integrity is maintained, the
possibility of a rod dropeu ac:iden: is cli inated. n e overtravel
positica feature provides a positive check as only uncoupled drives
may reach this - ositica. Neutren inst:a:entation Icspcase to red
-
movement pro / ides a verification that the red is folicwing its drive.
Absence of such response to drive move:hnt could indicate an uncoupled
.
condition. .
2. De con:rol red housing support restricts the outward : ove=ent of a
'
control rosi to less than 3 inches in the extremely re=ote event of a
housing failure. He a=ount of resetivity which could be added by
this s=sil amount of ::d withdrawal, which is less than a normal
single withdrawal increment, will not contribute to any da= age to
the pricary coolant sysics. % c design
*To dis'ar= the drive electrically, four a phenol type plus connectors are recoved fro.n
the drive insert and withd :wal solenoids r:ndering the rod incapabic of withdrawl.
His procedure is ecut. ale t to valvi: g out of the drivo and is preferred beenuse,
in this conditten. frive sater ccois c..d minial:cs crud accuculation in the drive.
Electrical disar=ing ::es not climinate position indication.
. Ame:rd:cnt No. 14 88
. ... . . . . . . . . . . . . . . _ .... __ .
-
- --
, .
_ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
. . . . _ ._ _ ._. _ . . . .
.. . . . _ _ . . . _ _ _
,_ _
*
l .
,
3.3 and 4.3 ,
-
,em
I SASES:
) basis is given is subscetion 3.5.2 of the FSAR, and the
safety evaluation is given in subsection 3.5.4 This ,
support is r.ot required if the reactor coolant system ,
is at atmospheric presnure since there would then be no
driving force to rapidly eject s drive housinr.. Addi-
tionally, the support is not required if all control
rods are fu;1y inscrecd and if an adequste shutdown
snrgin with one control rod withdrawn has been demon-
. strated, since the reactor would . main subcritical even
in the event of complete ejectien of the strongest
*
control rod.
#
3. In the course of performinn normal st'artup and shutdevn
procedurcs, the reactor operscor follows a pre-specified
sequence for the vitt.drawal or insertion of control
rods. The specificd sequenccs are characteri:cd by
homogeneous, sesttered patterns of control rods selected
for withdrawal or insertion, The maxinum control rod
worths encountered in these patterns are presented in
Figures 3.6-3 of the FSAR. These sequences arc devel- -
2
oped prior to initial operation of the unit to linit
the reactivity worths of individual control rods in the
core, together with the integral rod velocity limiters
, which will limit the velocity during free fall to less
,
than five feet per cecend, limit the potentini
reactivity insertion such that the conseque.cce of a
O. * control rod drop accident vill not exceed s.pcak cal-
culated enthalpy of 2SO eslocies/ Tram generated in the
fuel. The desir.n li=it of 250 calorics /gran is selected
for limitine, pesk enthalpics in UO 3 and is assu=cd to be
the lower threshold at which rapid fuel dispersal and
damaging pressure pulses to the primary system might
occur.
In addition, the calculated radiological consequences of
a control red drop accident should be well within the
guideline values of 10 CFR Part 100. The radiolonical
'
conscquencac of a rod decp accident when the reactor is -
in the hot standby condition at :ero rover is the worst
*
situation for this accident as described in Section
14.6.2 of the FSAR and in Section 4.4 of the Safety
*
Evalustion Report. For this accident, the coct reac-
tive control rod assc bly is sssumed to drop out of the
core 30 minuteu after shutdown, causing 330 fuel rode to
'
exceed a calculated energy input of 170 caloricr/gran
. with the maximum UO, enthalpy input of 220 calorics / gram.
The fuel enthalpy input of 170 calories / gram ic the lower
threshold for fuel cisJdinn perforatten. The ACC Staff'u
estirotes of the radiological conos quences, deceribed in
"
Section 9.4 of the safety r. valuation Report, are well .
within the guidelines of 10 CFR Part 100.
kj/ .
89A
.
e
.
O
O *
s
,
.
e O
-
_ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _
.. -_ - - .
.. - -- . .- =: .. . .. ._ _ __. ...
,
, _ _ _ _ - . _ _
t
1
3.3 and 4.3
4
h_ BASES:
4
When THERMAL POWER is greater than 20% of RATED THERMAL POWE'R, .there is
no possible rod worth which, if dropped at the design rate of the velo-
4
city limiter, could result in a peak enthalpy of 280 cal /gm. Thus re-
" quiring the Rb3 to be OPERABLE when THERMAL POWER is less than or equal
to 20% of RATED THERMAL POWER provides adequate control.
We are therefore requiring as a limiting condition of operation (LCO)
, that the Rod Worth Minimizer (RWM) be operable when the reactor is crit-
ical and below 20% of design power in accordance with Specification
3.3.3.3a so that the maximum in-sequence control rod worth will be lim-
ited to 0.010 delta k as given in Specificaton 3.3.B.3b(1) even' assuming
a single failure of the RWM or an operator error. The Rkh assists and
supplements the operator with an effective backup control rod monitoring
routine that enforces adherence to pre-established startup, shutdown,
and low power level control rod procedures. The RbH computer prevents
the operator from establishing control rod patterns that are not consis-
tent with prestored RkM sequences by initiating appropriate rod select
block, rod withdrawal block, and rod insert block - interlock signals
to the reactor manual control systems rod block circuitry. Reference:
FSAR Section 7.16.4.3. The RkE sequences stored in the computer memory
are based on control rod withdrawal procedures designed to limit the
individual control rod worths to levels given in Specification 3.3.B.3.b.
O Two exceptions to the requirement for RkH operability are permitted.
Control rods may be moved to shutdown the reactor, and up to two control
rods,can be moved provided all other rods, except those which cannot be
+
moved with contral rod drive pressure, are inserted. The first excep-
tion permits the operator to shutdown the reactor in the event the RbH
should become inoperable while the reactor is critical. In this case,
the operator is moving the rods to reduce the reactivity in the crie.
l Outward movement of any control rod is limited to a short adjustus. t ,
and the general sequence of control rod movement is always covtrd a
i
safer pattern during shutdown operations. The second exception permits
,
the control rod drives to be moved when the Rk'M is inoperative provided
that all but two rods are fully inserted except for those control rods
which cannot be moved with control rod drivo pressure.
(
,-
.
,
e
.s .
Amendment No. 39
89B
D +'
- . ., -- .-
.
_. . _. . .: _ . . _ . . _ .
.
_ . . _ _ . . . . . . _ . . . _ _ . . ._ ..- .
.
-
. . . . . . . - . . .
.. .-
.
. 3 3 and h.3
h BASES:
Above 1 @ of desi6n power assuming a single oper'ator
error, it vill not be possible for the maxi =um red '
worth to execed 0.020 delta K in accordance with
Specificatien 3.3.B.3.b(2).
Specification k.3.B.3 require: a sequence of checks
and tests on the RG to verify its operability before
startup and before reducing power to less than 10%
of design power. These checks and tests assure that
the actions of the control operator are always
monitored and blocked when in errer should they lea,d
to a condition which might cause fuel damage during
the control rod drop accident.
Under these specification limits, the maxi =un ener6y
deposition in the fuel and the nu=ber of fuel rods
dacaged resulting from a control red drcp accident,
assuming Technical Specificatica limits en scra=
times (Specification 3.3.c) and red drop velocity (5
feet /second), is established to be below the ecnse-
qucnces calculated by the licensee for the hot stand-
by critical case. Reference: F3AR Secticn 14.6.2 and
Safety Evaluation Section 9.k. Therefore, the as-
h sumptions used by the licensee and the AEC Staff in
estimatin5 the number of failed fuel rods and fuel
dan ge resulting frc= the excursion energy generated
by the red drop accident appear ccnservative within
the LCO.
.
e
e
D
W
e #
.
89c
l l
w e
, _ . . . - - _
_ _ _ _ _ _
,,
-
. ~ . - ..... -._ .- _ . = :. ..
. . - - . . . . . . . -
,
--
. . . _ . _ - . .. . _ . . _ .
*
.
3.3 and 4.3
%
; BASES:
O .
4. The Souroe Rance Monitor (SPM) system performs'
no automatic. safety systcc fbnction; i.e., it
has no scram function. It does previde the oper-
ator with a visual indication of neutron level.
The consequence: of reactivity accidents are
functicns of the initial neutron flux. The
requirement of at least 3 counts per second
assures that any tre.nsient, should it occur,
begins at or above the initial value of 10-0 of
rated power used in the analyses of transients '
from cold conditions. One operable SFl.: channel
would be adequate to monitor the approach to
criticality using ho=o6eneous patterns of scat-
tered control red withdrawal. A =ini=m of
two operable SP2t's are provided as an added con-
servatism.
5 The Rod Block Monitor (RE!) is desi6ned to auto-
matically prevent fuel datage in the event of
erroneous red withdraval from locations of high
power density during high peuer level operatien.
,
Two channels are provided, and one of these may
be bypassed frcm the conscle for naintenance
and/ortesting. Tripping of one of the chcnnels
will block erroneous red withdrawal soon enough -
to prevent fuel da= age. This syste= back.s up
the operator who withdraus con:rcl reds cecording
to written sequences. The specified restric-
tions.with one channel out of serrice conserva-
tively assure that fuel da-"Se will not occur
due to rod withdrawal errors when this conditien
exists.
.
O
O
90
s -
. _ , _ . _ , , - _ , _ _ _ _ _ _ _ , _ _ . - , . _ ._ _. _ . - . . , _ , ..
.
- 2._._-- . . ____ _ _
.u
, , - __ . - _ _ . .
, , , ,
*, - ,
,, .. ,
- -
. . .;, ,. , , .,
.
* *
* *
,
a *
.. . * . .
-
, . . ,. ,
.
.
.
. 3 and 4.3
BASYS: .
*
During reactor operation with certain li=iting
control rod patterns, tae withdrawal of a desig-
naced s1=gla control rod coul .1 result in one or
more fuel rods with .40 R's lass than the Safety *
iddt MOR. During use of such patterns, it is
judged that testing of the R3M system prior to
withdrawal of such rods to assure its operability
will assure that improper withdrawal dosa not
occur. It is the responsibility of the Rasetor
Insinear to identify these limiting pactarns ,
and the designatad rods either when the pactarns
are initially established or as they develop due
to the occurrence of inoperabis control rods in
other than 11=1t1=g patterns. ,
C. Sera = Inser:1en T1=es
The con =ol rod system is designed to bring the reac-
ter suberitical at a rate fast enough to prevent fuel
da= age; i.e. , to prevent the MCR from beco=d.ng less
than the Safety Limit M O R. Analysis of the 'd-4 ting l
power transiant shows that the negative reactivity
ratas result 1=g from the scram with the average responsa
h of all the drives as given in the abova specification,
provida the required protectics, and MOR remains
greater than the Safety 7'-#t MGR.
The scra= ti=es for all control rods will be decar-
ni=ed at the ti=a of each refueling outage. A re *
presentative sampla of concol rods vill be scram tested
dur1=g each cycle as a periodic check against decar1 oration
of the con =ol rod perfor=ance.
.
~
A=end:ent No. 42 .
.)
91
.. .
.
e
.
.
2* *
_ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ , . _ _ _ . _ . _ _ . _ _ _ _ . . . _ . . _ _ . .
_ __ _ _ . _ _ _ _ _ _ _ _ _ _ _
~ . *
,,
* * h = .- e e: .4+ -em-... .. % ==.m. , , , , , ,,
- - - . . . . . . . . _ _,
*
3 ?.end k.3
BASES: .
s
b
.
* 4
p
9
.
4 g
l
.
O
.
.I d ! 5 .> ! t '
.
I
- .
i l , 3
-
l.
t)
l 92 . i l i ! > i .
- . .. . .
l
. - - . . - .. -.-- _ _ _ - _ _ _ - . _ -
- . . . . - - : .: iu=::= n ....:E._ .' * _ . - . - . . . .. -
, , _ . .. ... . _ .
- -
.
3.3 and k.3
% BASES:
C In the analytical treatment of the transients, 2o0 millisecond:
are alloucd between ,a neutron sensor reaching scram point and the
start of ncCative reactivity incertion. This is adequate and
conservative when ecmpared to the typical tino delay of about
210 milliseconds cetimated frc:: scram test results. Approximately
5 120 milli:ecend: later, the control rod motion is e timated to
actually becin. However, 200 millicecond: is con =ervatively
assumed for this time interval in the transient analyses and this
is also included in the allowabic scram insertion time: of
Specificatien 3.3.c."
r
D. control Red Accumulators
Requiring no more than one inoperable accumulator in ~
any nine-rod square array is based on a series of XY
PDQ 4 quarter core calculations of a cold, clean core.
The worst case in a nine-red withdraval sequence re-
sulted in a kerr< 1.0 - other repeating rod sequences
with more rod: withdrawn resulted ine k rr>1.0. At
reactor pressures in excess of 800 peig, even the:e
control red: with inoperable ac:u=ulators will be able
-
to meet required =cras incertien times due to the ac-
tion of reactor pre =sure. In addition, they cay be
O normally incerted using the centrel-red-drive hydraulic
system. Procedural control vill assure that centrol
rods with incperable accumulator: vill be spaced in a
one-in-nine array rather than grouped together.
E. Reactivity Ano=alies
During each fuel cycle excess operative reactivity
varie: as fbel depletes and as any burnable poison in
supplementary control is burned. The cacnitude of this
excess reactivity may be inferred from the critical
rod configuration. As fuel burnup pregressen, anomalous
behavior in the excess reactivity may be detceted by
comparicen of the critical rod pattern at selected base
states to the predicted red inventory at that state.
Pouer operating ba::e conditien: previde the most sensi-
tive and directly interpretabic data relative to core
reactivity. Furthermore, u ing pever cperating ba:e
. conditien: pernits frequent reactivity ccaparicons.
.
93
.
. . . ..
- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
,
. .. -_ . ..- -_ L . - . . .. . . . . . . .
. . . . . .: ..uus
. . .
, ,
;
*
-
*
3.3 and 4.3 .* -
*
* . *
.
*
_5ASES: . *
C '
* Requirinr,a reactivity compart:en ac the specified -
*
frequency assures tJist a co:ipari:en will be made
*
before the core reactivity chante exceeds 17.4K. ,
.
*
,. Deviations in core reactivity r,reater than 17.4K are
*
. hot er;ccted and require thnrnur.h evaluation. Cnc
percent reactivity 11;:it is conr.idered safe since
an inscrtien of ths 'rcactivity inte the core would
.
. not lead to transitnts exceeding desit,n condition: of -
the reactor syste . . .
. .
*
*
d
*
*
.e
0 e
e
(,' .
.
)
1
*
. ,
e
e
G
$
e
e
. g
.
e
e
.
*
O
e
e
o
e
9
.
.
O O
$
er e "g'
0 t
,
e
e
1
9
* e' . . .
.,
,
0,
g
- - , - -. ,_ . _ , - . _ . . , , _ _ . _ _ . _ . . _ . _ _ _ _ . , - _ . _ _ , _ _ _ _ , _ _ _ _ _ __ _ . , _ _ _ _ - . , _ - _ , _ _ . _ _ _ _ - _ _ . - , . _ _ . -,
- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
--------.-..-------------------------------------- --- --
._ _ . . _ _ ..z..r - ".. ; ' .. _ _ _. . __ .
.. _ m
4
.
LIMITING CCN0! TION FOR OPERATION SURVEILLANCE REQUIREMENTS
'
i '
3.6 PRIMARY SYSTEM SOUNOARY 4.6 PRIMARY SYSTEM BOUNDARY
! -
Acolicability: , Aeolicability:
. .
Applies to the operating status of the Applies to the periodic esamination and
reactor coolant system. testing requirements for the reactor
cooling system.
i
!b*ective: Cbiect've:
To assure the Integrity and safe c:er- To determine the c:ncitten Of the rea:::r'
1:*cn cf t e reactor c:claat system. :::la system and the :: era:':n ;# te
.ge,. . g. . 3 7 .,3... . .r i .
.
3.Y.' ' : '
G.L . ..;: ' . *
t
- . 3- 41 a-- 2 ,tiari. -
.
. -
: ;retsurt:a- -
v.
--
.g.. I
-
.. -
-
.
:. e i.e ti. . . , . . ... . : . :. ; ea . .:s 1. : .
tem:4 sture enaage aar.ng normai following temperatures snail :e
'
. j heatus or cooldown shall not exceed permanently logged at least every
ICO'F/hr when averaged over a one- 15 minutes untti the difference
hcur period except when the vessel between any two readings taken over
temceratures are above 450*F. a 45 minute period is less than 5'F.
The shell flange to shell tempera-
ture differential shall not esceed a. Reactor vessel shell adjacent
145'F. to shell flange
b. Reactor vessel snell flange .
c. Recirculation loops A and 5
i 2. The reactor vessel shall not be 2. Reactor vessel shell temoerature and
- pressurtzed for hydrostatic and/or reactor coolant pressure snall be
leakage tests, and critical core permanently logged at least every
operation shall not be conducted 15 minutes whenever the snell tem-
- unless the reactor vessel temperature perature is below 220*F and tne
is above that defined by the reactor vessel is not vented. .
Appropriate curves on Figures 3.6.1
, and 3.6.2. In the event this Test specimens of the reactor vessel i requirement is not met, achieve base, weld and heat affected 2cne ' stable reactor conditions with reactor metal subjected to the highest flu-
vessel temperature above that * defined once of greater than i Nov neutrons
by the appropriate curve and obtain shall be Inttalled in the reactor
j an engineering evaluation to determine vessel adjacent to the vessel wall
the appropriate course of action at tne core aldolane level. The
- to take. specimens and sample program shall !
- conform to the requirements of
!
ASTM E 185-66. Selected
L
, s
,
Amendment No. 82 . 123
_ . _ - _ _ _ _ . _ _ . _ _ _ _ _ _ - _ . . _ _ . _ _ _-. _ . _ _ , _ _ . _ . _ _ _ _ . - , - -
~
'
. - - - - .
-
: '
. .: ' . .: . a . .-_. ..
. ._ . . . . . . , _ _ , n
.
1 . .
LIMITING CON 0! TION FOR OPERATION SURVEILLANCE RECUIREMENTS
3.6.A Thermal and Pressurization 4.6.A Thermal and Pressurization
Limitations (Cont'd) Limitations (Cont'd) 1
i
neutron flux specimens shall be
I
removed at the ' frequency recutred
'
/ by Table 4.6.3 and tested to
emperimentally verify adjustments
- to Figures 3.6.1,and 3.6.2 for
;
predicted NOTT trractation shifts.
3. The reac:or issel head tolting 3. When the react:r vesse' henc :olt-
- st;ds snai*
- : :e uncer tersion Ing s:uds are :enstcre: tec : e
'
ur.less :ne ci :arature of tne reactor is in a C:ld C0ccitt-'t.
,
-
vessel head #1 3*ge tad t*:e head : e reactor vessel 1.'e l l
t
is 4 e,-a- -w s, n't n-- . . g : 2,.
-
' - e.: 4.g r e r :e p ---
.
-
,y .. 2.. -
* **
- .
*
. :.
! .. * r e . .. . a a a t .: . . .. ::c :
*
:- -- '; . . r. ; ; ; 4. . ..: . . - -
'cc: t tail --* :e "ar:ec unless l e 4: .. ..'-a 1**a. tM te-
, -, --- , -
:4- . , : -
n~:
- ,
,
. :, :q~ -*
':5
; c :o . . . .;;s A.e = r:0 .U ' T s . ail :e ;:s :1 4..: ..,,...
of eacn otner.
,
5. The reactor recleculation pumos 5. Prior to start:ng a recirculation
shall not be started unless the pump, the reactor coolant temper-
coolant temperatures between the atures in the dcme and in the
dome and the bottom head drain bottom head drain shall be
are within 145'F. compared and permanently logged.
6. Thermal-Mydraulle Stability
Core thermal power shall not onceed
25% of rated thermal power without
forced recirculation.
8. Coolant Chemistry 8. Coolant Chemistry
1. The reactor coolant system radio- 1. a. A reactor coolant sample shall
activity concentration in water be taken st least every 96
shall not onceed 20 microcurles hours and analyzed for radio-
of total lodine per al of water. activity content,
b. Isotopic analysis of a reactor
coolant sample shall be made
-
,
at least once per month.
2. The reactor coolant water shall 2. During startups and at steaming
not onceed the following limits rates less tnan 100,000 pounds per
with steaming rates less than hour, a sample of reactor coolant
100,000 pounds per hour, encept shall be taken ~every four hours
as specified in 3.6.8.3: and analyzed for chlorine content.
Conductivity .. 2 paho/cm
Chloride ton .. 0.1 ppm
Amendment No. 82 124
"
-
, . . . . .
- . . .
- _ - _ - . .
. . . . . . . . . . . ,
.. - . ..
.
'.- ..
.._ _ . . . _
.
.-
.
9
* .
'
4
7ABLE 2.5.3
RE.2C70R VE33E. "A*E4*al
.: SURVE:'.LA'.CE 24CG4AM !N042aAt. 5" HEC'.'LE
9
. :' .
.- ;*.a. , ;..
. . ..
j i ,.. ,e ,,
.. . . ..
-
i s
+
1
e
! 1 *i . .
,
2 15 6.3 4 10 91
j
*
(approx.) (appron.)
~
3 End of Life 1.4 x 10 136
4 (appron.)
;
,
i
i
-
i i * $
!
4
!
-
, l
.
.
.
Amendment No. 82 , 124A
..
. . . . . -.
. . - - . .-
.
. .
-
11pkT:NCCCNOITIONFOROP!KATION Si'RV!!LLANCI RIOCIRIM!s;$
3.6.3 Coolant Chemiserv (Cont'd) 4.6.3 Coolant chemistrv (Cent'd)
3. For reactor startups and for the 3. a. k*ith steaming rates of 100.000
first 24, hours after placing the pounds per hour or greater, a
reac:or in the power operating reactor coolant sample shall be
condition, the following limits taken at least ever7 96 hours
- .shall not be exceeded, and analyzed for chloride ion
concent.
Conductivity..10 u=ho/:s ,
.
b. When all cos:L:.:us :::du::1ri:7
Chloride 1:n...C.1 ppa =cnitcrs are is::erable. a reac:or
c:o*2:: sa:714 shall be :akas a:
leas: faily and analy: d f::
- * *- '--* ~ *y a:L 13.:rzi s 1:-
-
- *
... - :- .: ,
::ve. . r aa . ::. . ..:
..;4 f *. * : .. . r -
. .:a *. c : at:ee: .
'
. i.- 4 e . s :- - -'
.: . .. ..
1. . . .;C ;: 14 , 2 r . .. . .
'
Conductivi:7.. 10 unho/cm
.
Chloride ion.. 1.0 ppm
J. If Specification 3.6.3 cannot be
set, an orderly shutdown shall be
ini:iated and the reactor shall
be in Iot Shutdove within 24 hrs.
and Cold Shutdown within the next
a hours. *
C. Coolan: Laaksee C. Coolan tankare ,
1. Any ti=e irradiated fuel is in the 1. Reactor coolant systes leakage shall
raae:or vessel and reacter coolant be checked by the sump and air
temperature is above 2110T. reactor sampling system a:d ree:rded at
*
coolant leakage into the p:1=ary least once per day.
containaast frra unidentified .
sources shall not exceed S sys. In
addi:Lon. the total raastor coolant
systes leakage into the primary
containmen: shall not ascoed 25 g;e.
'.
2.* Both the sump and air sampling sys-
tems shall be operable during reac-
ter power operation. Tres and
af:er the doce that one of these
systess is sede or found to be i=o;-
erabia for say ceason. reacter -
.
*
Amendment 'lo. 42
123
- - . -
_
- _ . . - _
._. , _ . - _- _-_ - -. . _ _-.
.
. .
. . . . .
_ . .. . . m
.
. .
'
. . .
.
- g?.gp CC!fDITTON FOR CptRA*"CW .uv a.t' 2'fer At'l"Jt%IMENT
, . _
-
* 3. 8.C Coelant Chemistry (Cass'd) 4.6
i
power operatisa is permissible .
esly during the ausseeding saves *
*
.
i .
days. f ."
'
,
3. If the esaditissa is'1.or 2 above .
, saaset be set, an ordeirly shutdows
+
I
* ekan be tal:1sted and the roastar *
*
, aha n be is a Cs14 Shutiswa Candi- , .
*
tism withia 24 hsars.
i . .
, *
! , 3., 34feev and Re!!af Talves 3. Safet, a-d 1telte! 7:!ves
'
.
. . , . . . . . . .;.... ' -
, ,, . " " *. * * *
;
. .
.. .
.
, , , , ... ..
i .. . .: - .e- -
, 3,u, . ., : .. , . .. . . . . .
.. 4;.ra- ..:s
!
is a. ..t .. *h e s- --2- - " 3* 18 * 2
...m.. 6 3, w- - - -
:f ...
I
'
. ,... ... . . . - .s . a- - ..sM.* d *" -
aza,au. s
.,
asau'be as specut.44 sa spesMii.'s
I
.
easiaa 3.3.
2. If Specificacian 3.4.D.1 is not met.
I
i *
as orderly shutdown shan be is. 3. As least one of the relief / safety
isiated and the reestor seelaat M ves sha1.3.be disasseskled and
pressure shall be bolsw 104 pois ,
188?estad eash refuellas estage. ,
.
within 24 hours. Note: Technical ,
{
*
Specifications 3.4.D.2 - 3.6.D.5 3. Whpever the safety relief valves
j
apply only when two Stage Targer- are required to be operable, the
i
, Beck SRVs are installed.
*
* * * ~ dimeharge pipe temperature of
l . *
3. If the temperature of any safety ensk safety relief valvg, shall be
i legged daily. *
-
relief discharge pipe exceeds 212*F
j -
during seraal reactor power operation 4.
!
Instrusestaties shall be calibrated
~ ~ ~ " " ~for a period of greater than 24 hours,
~
,
as engineering evaluation shall be and aheaked as indicated is Table . -.
4.2.F. - * * " " ~ ~ ~
performed , justifying continued opera-
i
ties for the corresponding tamp, ,
-
isaresses, and a Report shall be S. Notwithstanding the above, as a
Lasued per T.S. Secties 4.9.3.1 which minimum safety relief valves that
j
'
5 ahall address the actions that have , have been in service shall be
been taken or a schedule of actions to tested in the as-found condition
he taken.
4
during both cycle 6 and Cycle 7. ;
4. Any safety relief valve whose dis- *
-
<
eharge pipe temperature saceeds 212'T .
i
for 24 hours or more shall be removed
'
at the seat cold shutdown of 72 heure
!
or more tested is the as-found condi- *
ties, and recalibrated as necessayy i
'
prior to reamstallation. power opera-
tism shall set costiaue beyond 90 days 120
i
; -
.
.
!
.
Amesesent No. 56 '
!
l . . . , . . . . . . . - *" *.*-==* = '
(
( ,
4
-
, - - - - - - . ~ , . ~ , _ , _ _ , , , __,.,cn--n
.
.
. . . :-- . . . - . . .. _ _
.
. un..~.a............ ..n .m..... w .. . . . . . . , . . . . . . . . , ,
*
. 3.8.D. Safety Relief Valves (Cont'd) E. Jet Pumos
from the initial discovery of dis- Whenever there is recirculation
charge pipe temperatures in excess flow with the reacter irt the start-
of 212*F for more than 24 hours up or run modes, jet pume Oper.
*
without prior NRC approval of the ability shall be. checked daily by
engineering evaluation del,ineated verifying that the foliewing cen-
in 3.6.D.3. ,
,
ditions do not oc:ur simu.ltaneously.
. .
. . -
.
5. The limiting conditions of opera- 1. The two rectreulatien loc:s have
tion for the instrumentation that a ficw imbalance of 15: cr =cre
'
monitors tait pipe tamceaature when the sum:s are c:*raten at
are given in Table 3.2.F. tne same speed.
E. Jet 8.: es 2. The indicated value of c:re f6
ra ta v a ri es ' .m t e v a * .e
. .
,
.s..,.,
. .., .. g . . . . s . *. .w. ; , . . . . e .. . . s...,, . ..,..
'*- *
.. ..'..
. ! :;** .
. a; *~
. : I
. '1 '
: = :: . .
.: * '
: * *
:' -
' -~~ss re rss..ng .
:e - 2 -:
'
211
-
' ~
- "- 4* : va 'es
,
ts ir .::n? k'. *
.: .
=
w ; .M r. . . . . . . . .. A f;:".a." . 4 ; . . a ;; .lt.
F. Jet pume riew Mismatc5 F. Jet Pume Flew wis at:M
1. *
Whenever both recirer1'ation pumps Recirculation pumo speeds shall be
are in c:eration. zump scoeds shall checked and logged at least once
be maintained within 10 of each per day.
other when power level is greater *
.
i than SO: and within 15 of each !
other when pcwer level is less ~
than or equal to 80t.
2. If Specification 3.6.F.1 is ex '
caeded intnediate corrective
, action shall be taken. If re- '
circulaticn pump speed mismatch
-
is not corrected within 30 minutes.
.
an orderly shutdown shall be in-
'
itiated and the reactor shall be
in the Cold Shutdown condition
within 24 hours unless the re-
circulatten pumo speed mismat:M
is brought within limits sooner.
4. Structural integrity. G. $tructural InteMty
1. The structural integrity of the The mondestructive ins:ect :ns d
primary system boundary shall be listed in Table 4.6.1 shall te
maintained at the level required perfor ed as specified. The
by the ASME Boiler and Pressure results obtaine~d fren c:f:11ance
< Vessel Code. Section II " Rules with this specification will te
l of Inservice Inspection of Nuclear evaluated after 5 years anc tse
Power Plant Components", 1974 c:nclusions of this evaluatt:n
will be reviewed with at".
Arend. ment No. 71 ...
.o
. _ _ . - - _ _ . - - _ . , _ _ _ _ _ _ _ - . _ _ . _ _ _ _ _ _ _ _ _ , _ . . . _ _ _ _ _ . . _ _ __
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ . _ _ _ _ _ _ . _ _ ___ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
,
.. .
, . _ , . . _ _ . _ . . . _ . . --
- - - . - -
.
, , 4 . .s . w a c nJc Curs ! InCCf* ICV (Cen * c)
.- ,
*
-
.
Edition (AS:E Code, Section XI) . In the interft: until the micicar
system pipin; inspection avsluation icyc1 criteris of the Aste
Boiler and Pressure Vessel Code . Section XI,1974 Edition,, are
C,, 19 sospleted, the applicabic evsiv . tion Icyc1 provisions of the Astt
Boiler and Pressure Vessel Code, Section XI, 1971 Surreter Addenda
shall be used in the Inservice faspection of nucicar pipinC'
-
Components of the primsty systen boundary whose in-service c::amins-
tion. reveals thn absenec'of fisw indications not in execes of the
allowable indication standards of this code are acecptsbic for
continued service. Plant operation with components which have
-
.' in-service exs tination finu indicstion(s) in excess of the
allouable indication standards of the Code shall be su*oject te
NIC approvs1.
r}[, ,
a. Ccmponents whose in-se:vice e:.::rtination reveals fisu
,78 -
* indicatien(s)11 excess of the si'c;ish;e indicati:n
3 = . - f . ~ .' a .t '.pe*= .:g, *
.4.. * .. ..., ,.. r
. :. . : :..
. .. . : ;..- :.
~_ , ( :,, sr. :- ' '
..:s : : ; :- C..
. i .dic .;r.w, ...... be :, e : --
d- u r-s *- :% -. : :. . .-
'
, .' .. .: : ..: . , : .: .?.
ev:.12 .1;.. :.. .'. :': *. ' . . :.:s ;.r::eaurs: :::. .:J
in Appendi:c A. "Evstustion of Fisu Indications,"
of AS*d!; Code, Section XI.
19 l
I
(ii) Prior to the resta.ption of servica, the !!3C shs11
review the analysis aad evaluation and either
.
approve rette.pcica of pisnt operation with the
affected components or require that the cceponent
,
be repaired c: rapiscad. *
, b. For co.iponents appreved for continued service in accordance
with paragraph s, above, res::smination cf the aros contsir.ing
the fisu indicstion(s) shall be conducted durin., each scheduled
successive in-se:vice inspection. An analysis and evaluation
19
l shall be submitted to the GC fellowing esch in-service .
'
inspection. The analysis and evatustion shs11 follow the
procedures outlir.ed in Appendi:: A, . "l:valuati:n of T1sw
Indications," of AC:2 code, Section XI, and shall reference
.
.
prior analyses submitted to the IE to the catent applicabla.
Prior to resurption of service follouing each in-sarvico +
inspection, tha NRC shall reviou the anslysis and evaluation
and either approve resumption of pl. ant operation with the
affected coe;'onants or require that the conycnent be re;stred
or replaced.
.
*
c. Rapsir or replace:sent of ecmponents, includinr, re-exaniastiens,
shall conform with the requirteents of the A31% Cade, Section
XI. In the essa of ropsirs, fisus shs11 he either renoved
or ropsired to the extent neccesary to r.4cet the allouabla
indication sesndards spesificd in A3: Coda, $cction %I.
*
( . . .
.
Amsndment No.19
.
.
.. --
. . . . - . . - . . .
.
.
.. ,
,
.
* .
* * * thMITING CONDITIONS FOR OPERATION
, SURVEII.I.ANCE REQUIREMEYrS
3.6.H High Energy Pipint (outside 4.6.H Hir_h Enerry Ptoine, (outside
containment) containment,1 ,
1. The high energy line sections The inspections listed in Table
identified in Table 4.6.2 shall be 4.6.2 shall be perfor:sd as
I maintained free of visually observable specified to verify the structurst
,
through-wall leaks. integrity of the specified high
energy line sections. The standare
2. If a lesk is detected by the of Section XI of the AS:2 ! . tler
surveillance progran of 4.6.H. and Pressure Vessel C:de, .i74,
efforts to identity the source Article IW3 3000 shall be used
* *
of the lea'5 shall he started.
in thesa inspecticss.
. 4 ... .
,
, : ....:
. , :i...
j .4. . . . .t ; . an :. : .t -
g
g :: :stsetten or *: L.*e lacs .. ! .
*
'
f:2..d to be ! :n a $ *( in ?.i
* '
*
ate.1 n2 L P .- . .
. 5.1, t h a . . . . ... . ....:..:
or tne teactor snail os in a cold
, shutdawn conditien within 48 hours.
4. When the modifications. described
'
*
in FSAR Amendr.ent No. 34, to -
provide protection against high -
energy line breaks outsida of the
pri.ary containmenc have been
completed. Technical Specifications
3.6.H and 4.6.H will no 1casar be
required.
.
.
.
.
.
,
<.
un
19
Amendoent No. 10
.
_ _ . , . - - - , _ , . . - - .---- - - - - - - , - - . - . - , - - - - - . - - . - - - - . . - . . . - - .
- - _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
._ ____.___________;____________________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _
J. *
. _ . . . . . . . . . - - - - - . . - - - - - - - - - - - -
'
' t.!Mi*tNG CONCITIONS 'OR OPERATION
$UR'!Ett,,' U4CE QE:t.i!REw!NTS
3.7'CONTatNMENT SYSTEMS
4.7 CONT 4!WENT SYSTEMS
Ac:llesbility:
Aeolicability:
Acclies to the operating status of the
primary ano seconcary containment systems. Acpttes to the crimary ano see:n:ary
containment integrity.
.
Obiective:
Obiecti ve:
*s assure the integrity of the primary
ara secondary containment systems, to <ertfy the integrity of tne cet arj
ans se: rcary con:stn en:.
5:t::st!-a:
5:ee 'i i-'-a:
A. 2rt 3rv C a!)Ia ea. A, s**-g* C: a * 1 ',a.-s a ?
_
,. . . . .
* ., .
. . ,
. .
. ,
; 4
, .: ;*- ,.
,,;.. -
. ;.
'si ~e c. !J . -
*
.;& . i
: s Orte:ure :.:: e ':n :::t .t!*- : ** t o --t
..;,4 -
9- q g- g*' - 's'*:'****--
.' .: - . -'
*1 ) .
#: *:
.
3::. ;
4. 6:: 1; . ,. * * e ; ' n 1. ' .
**:r 104. fr. . .-
l 4 *. J
3. 7 A.J.
.
"ressicn "Ocl. C"# ::al
tem:erature shall be ::ntinual!j
4.
Minimum water volume . 84.000 ft 8 monitored and also CDserved anQ
loggeo every 5 minutes un:tl tne
b. Manimum water volume - 94.000 ft' heat addition is terminate:.
c. Pavfrum su::ression cool bulk temc. C. Whenever there is inc! cat' n Of
trature curing normal continucus
,
relief valve coerarten wt!s tre
culk temcerature of :ne
cover cceratten shall be <80*F. sucoression cool reacntng 160*r
* esCent as scoctfled in 3.7.A.I.e. or more and the artmary :cClan
c. Mastmum succression cool bulk temp. system crassure greater :9sn 200
erature curing RCIC. *PCI or ADS psig, an eiternal visual es
oceration snall te 190'F. escent nation :f :ne saccress'enacamt. :er
as scoctfleo in 3.7.A.I.e. snali ce concuc:ec te' ore
resuming oc.er Ocerat*:n.
e. In order to continue reactor coner
oceratten, the succressten chamcer d. Whenever there is 'nalca t:n f
cool culk temperature must ce rettef valve cceraticn wl:n :ne
reduced to 180*F within 24 hours, local temcerature of :ne suo.
crossten ecol T-cuenc9er eacntng
f. 200*F or more, an ev:eenal vissal
If the sumoression cool Dulk tenc- esaminatten of tne s.c: cess':a
erature encoeds the limits of enameer snalt te c:nsue:es :ef:re
* Soectftcation 3.7.A.l.d. RCIC. resuming ocwer c: erat'cn.
HPC or aOS testing shall to
terminated and succresstCA Dool e. A disual insceCilCn Of tne
Coollag small be inttlated, su pression cnateer i n ;t e* * :r .
1 If tne suceression cool bulk Including natts line '41'261.
temce*ature curing reactor cower shall ce maae at eacn maj:r
refueling outage.
I
oceratten eiceeds 110*F. tne
l
l
reactor shall De scrammed.
Ameneneae No. 83
1$2
- - _ _ - . _ . . - - _ - - _ . - _ .- -___ .-. - _ _ - - _ _ , . .- - _ . _ _ _
- - - - - - - - - - - - - - - - - - - - - - - -
--
.
.
. . . - . . ... . . - . . . _. --
.
. .
LIMIT!NG CCN0!T!CNS FOR CPERATION SURVE!LLANCE RECUIREMENTS
3.7 C04TatNMENT SYSTEMS (Cont'd) 4.7 CONTAINMENT SYSTEMS (Cont'd)
' h. During reactor isolation f. The pressure differential l
conditions, the reactor between the crywell and
pressure vessel shall be sugeression Chadeer snalt be
decressurtzed to less than ,
recorded at least once each
200 osig at normal cool down shift when the differential
rates if the pool bulk pressure is required,
temperature reaches 120*F.
1. Su::ression chamter water l
l. Offferential pressure betneen level 19all be rec:r:e2 t!
tne drf. ell and sue:ressten least cc:a each snt' .aen
cna :er snail be matitained at :he 3ffferen:tal ;ress6-e
e:ual to or greater : Nan 1.17 15 re;atred.
:stc, etcect as 1 ecified 'n
,
a-
. ., 4
.
, .,
* **
:t si .- .s .
; * S l a : .*. . . ria'".- in :Pe
e. ?::e fal*;w'aj a : - * * * .n . I
I
*9 :t s y3-..
. .
,,
*
*d '60. t3 :. *'
psic 24 nours grtor to a
screduled shuta wn.
k. The differential pressure may
ce reduced to less than 1.17
esta for a maximum of four (4)
nours for maintenance
activities on the differential
pressure control system and
during recutred operacility
testing of the HPCI system, the
relief valves, the RCIC system
and the drywell-sucaression ,
enamter vacuum creaners. .
l. If the specifications of
Item I, acove, cannot be met,
and the differential pressure
cannot be restored within the
subsequent (6) hour period,
an orderly shutdown shall be
initiated and tre reactor shall
De in a cold shutdown condition
'
in twenty-four (24) hours,
m. SuCDression Chamber sater level
sn411 te maintained tetween -6
to -3 Inches on torus level
Instrument whicn corresconds to -
a downcemer suemergence of 3.00
and 3.25 feet respectively.
Amencment No. 83 1523
-
*
.
-. ~_ _ . _ _ _ . _ _ . , _ - . . _ . . . , . _ _ _ _ _ _ . . _ - , . _ _ _ _ _ . _ _ - _ . . _ . _ . . _ -
- . _ --- _ ----
,--__ --------- - ---- ----_--_---_._,--_--------,------_-_----------------------- - - - - - _ _ _ _ _ - - - - _ - - - _ - - - - - - _ - - - - - -
. . _ . ..- . . . _ . . .... .. . . . . . . . . . . . . . . _ . _ . ._ _
_ _
-
.
*
.
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS
'
. Primary containment integrity 2. Intearated Leak Rate Testina
shall be maintained at all times ~
when the reactor is critical a. The primary containment
or when the reactor water tear . " integrity shall be demon-
perature is above 212*F and strated by performing an
fuel is in the reactor vessel Integrated Primary Con-
except while performing "open tainment Leak Test (IPCLT)
vessel" physics tests at power in accordance with either
levels not to exceed 5 Mv(t). Method A or Mett.cd 3. as
follows:
Ve**.ed A ,
,
-
,
,
f s. , ,,. .
,,
L*. ... ..~..' . . . .
*
.
..s ?:: --e'- 4
. ; . r.e u .r .:
.... ..L,ss), e-
.: :1 3
Forform leak rate test prior
to initial unit operation at
the test pressure of 45 psis,
Fe (45), and 23 peig, Fe (23),
to obtain the seasured leak
rates, L. ly.(45)
respective and L. (23),
3. The suppression chamber can be
drained if the conditions as
specified in Sections 3.5.T.3
and 3.5.T.5 of this Technical
Specification are adhered to.
.
.
Amendment No. 39 !$25
_ _ _ _ - _ _ _ . _ _ _ _ _ _ _ _ - _ .
,,
,.
-
_. ._ . ..._..........._..._.;.. . _ . _ , . . L
. . .
SURVEIIINICE EOUE7?:T3
*
k.7.A. Primsrv Centsir.-ent scont'd)
{-
Subsequent lesk rate test: shan be perfor=ed without preli=inary I
leak detection surveys or leak re;nirs i==ediaterly prier to er
. during the test, at an initial pre::ure of approxientely h5 psis
for Method A and 23 psig for Methed B.
Leak repair:, if neces:ary to per=it integrated leak rate tening,
shan be preceded by iceal lesk rate r.escure=ents where ;;: ihlt.
The ic h rest difference, prior to and after repair vhe. cerrte:ed
to the appr:prl: e ter pres: res of k5 p:11; fer Meth:1 A :::. 23
3:ic f r !:sth:d 3 shall be addad te the finsi ir.te;r: 2d its?. rate j
re:ults.
,
. . .. . .. ..
;...
. ....:...:.. -
. ,. .:. .: : * :- .
,. .. :.
The t2:- d=1:1:n :n - na- be la .c :: :.:.:a--s itt. e c . 2 2
-
2
rate t;'s: v eran::, 12: :hr.1*. ':1- -
..4 . :i ad ;or' f
tir.e : tar.f; , by :ei: .c. ; c -
: * -
: r) .:r .
: the :::rting ;:in- (:: :her :.:n-i: : f s ;::.'t:13r.: . ...i*?i 7),
the validity ant. accur:cy cf the lesk rate results.
b. Acceptance Criteria for !7C*."'
Methed A -
The maxi =u= anevable leak rate (1 ) shan not exceed :. 0 veight
percent of the cen:ained air at 45 7psig per 24 hcurs.
,
'
The aucvable operatienal leak rate, L:e(h5), which shall be est
prior to resu=ption of pever cperatien fon:ving a test (eithsr as
measured or follcuing repairs and retect) shall not e:::eed C.75 *e .
Methed 3
Theanevableto:tleakrateig(23)shallnotexceedthelesser
value established as fonovs:
Lg(23) = 1.0 x/.(23) I
I
(le(45) ,
" ' " *
13 (23) / ((k5) $. 1.C
l
. .
Or
Lg(23) = 1.0(P.(23) 1/2
.
Pg(45),,
f- where Pg (23) and F (l45) are mostured units et absoluto pressure.
- The allevable operatier.a1 leak rate, Lgo(23), which ths11 he not
prior to reeu ptton of pucr cierstion ro11avin.1 s teet (ei: hor as
aescured er relieving repair: and estant) ct:su not exceed 0.75 L.( '
'
,
*
*
I
..
153
,__-- - - - . - . - - , - - , . - _ . _ , - . - . . - . .
- - - - - - , - - . - - - . . , . - - - - - - . - - - - - - - -
*
. . . . . . . . . . . . . . _ _ . .'" . _ . . . - . _ . . .
-
l . *
.
4.7.A Primarv Containment (Cont'd)
c. Corrective Action for IPCI.T
.
Methods A and 8
If leak repairs are necessary to meet the allowable operational Icak
rate, the integrated leak rate test need not be repeated provided
local leakage measurements are conducted and the leak rate differences
prior to and af ter repairs. When corrected to the test pressure a id
deducted from the integrated leak rate messuremen:s. yield a leakago
rate value not in excess of the allovable operational leak rate.
d. Tr et se-e*? f )r !P"'_T .
,
. . . .
A. hr c:. it . <- : s ta t:. :- C * . '.: u:' s tr . :. t- - :
leak. :2 .a . .1 be Mri:r cd a': a;;r: . q . . . . . . . . . s .. . .
h e tw. '- t 's .-hu td r ":s f:: i* * ice .. . . ducted a-
tan- i. .
e r t r. s . , * n .: t,t o- .r. t 2 ; .
*
. . /.a . 4 e.:-
: h a l .' J o i .* r f O r".Q s 4 : 1.*. e ..r.s 0 : . E sna!*** * * . it e !*/al ini a.=,
coiacide with ene inservice inspection shutdown period.
,
e. Local Leak Rate Tests (LLRT)
.
jgg, hods A and 8
(1) Pri:rary containment testable penetrations and isolation valves
shall be tested at a pressure t 45 psig, except for the main
eteam line isolation valves which shall be tested at a press are
3,23 pois, each operating cycle. Bolted double-gasketed soais 3
shall be tested whenever the nest is closed af ter being opened
and at least once each operating cycle.
(2) Personnel air lock door semis shall be tested at a pressure
i 10 pois each operating cyclo. 3
l
f. Acceptanco Criteria and Corrective Action for LLRT
Method A
If the total leakago rates listed below are exceeded, repairs and re-
tests shall be perfor.ned to correct the conditions.
(1) Double-gaskoted seals 10% Lg ,(45)
,
l
(2) All testable penetrations and isolation valves 60 Lp (45) 3,17
(3) Any one panotration or isolecion valve except main see,,am line
isolation valves 5.*. Lg ,g43)
l C f 1 (
1$4
. t
I ! I
.%.--.+-.-r-%---, , . - - - - - - , , - . - - - . , - - - . ----.*v---7m,--, --v--- - r --r. ,- -<-e-.---- -m.. --- -,. - .- i
_ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ ______________ _______ -_ -____
.' . _ __..________ _ _L_ _ _ _ _ _ _ _ _ _ _ _- ._3.._ _ _ _ _ _ _ _ _ _ _ _ - _ - _ . . _- _...'
. .. . .- -
-. . ..
_ _.. - -
.
.
-
.
4.7.A Primary Containment (Con'd)
{
(4) Any one main steam line isolation valve 11.5 scf/hr e,23 peig.
Method 8 ., .
If the total leakage rates listed below as adjusted to a test pres-
sure of 23 psig are exceeded, repairs and recasts shall be performed
to correct the condition.
(1) Double-gasketed sesis -
10:.1. (23)
to
'
(2) (a) Testable penetestions and isolation valves 60" L7 (23) l2.*7
*
'
-
'
.
..
. . ..a.... .
.. .. . , ..
sai As : :.: . . n. .; : '. . . . : . : . - -
3 s:: . ... .g
Les'< ::ar : e .* - d -
- ' e t sh ' ! . .:!! .." . t : ;.: t- - -
cf . s ..; ::: :. : ... :..
LLRT(23)adj =LLRTseas x t m(23)
. La(45)
3 Continuous Leak Rate Monitor
When the primary containment is inerted, the containment shall be
continuously monitored for gross leaks;e by reviev of the inerting
'
system eakeup requirements. This monitoring system may be taken
1
out of service for maintenance but shall be returned to service as
.
soon as practicable.
.
h. Drvve11 Surfaces
i
The interior surfsees of the deywell and torus above water line shall
be visually inspected each operating cycle for evidence of deteriora-
- tion.
'
. .
,
4
O
($
, , 155 l
.
.
.. . _
. . - _ _ . . - , . . . - . . - - . _ _ _ _ _ _ - . - . - _ - . - - - . - .
.
. , ._ . . . . . . _ . - . . . . . .. . . . . . . . . . . .
- --
.
'
3.7 . Primary containment 4.7 Primary contafment .
,
3. Pressure Sucoression Chamber - 3. Pressure Sucoression Chamber -
Reaccer Buildina vacuum Breamers Reaccor Building vacuum Breamers
a. Except as specified in 3.7.A.3.b a. The pressure suppression chamber -
below, two pressure suppression reactor building vacu m breakers
chaseer - reactor butiding vacum and associated instroentation
.** breakers shall be operable 'at all including set point shall be checked
times when primary contairment in- for proper operation every three
tegrity as required. The setooint months.
of the differential pressure instru- .
tentation which actuates the pres-
sure suppression chamber - reactor
building breakers shall be 0.5 psig.
b. Frm and after the date
.. .,-
that one of .
<
.....;., . .
:. ..
-
si ': ' 's . . . : e u....
pe m :
'
. t c. e
su:: esc:- .tvs- :a.1 u.- rs :uch
va:- - :.2stea *: 3: na- - '
opera:'s. . :vt:ad t. 4: :.:4 ::.
procacure caos not vtoia:a pri.t.ary
contairment integrity. . .
4. Drywell-Pressure Suoeression 4 Drywe11-Pressure Suoeressien
Ghamoer Vacuum Breaners Chamoer vacuum Greaters
a. idhen primary containment is a. Periodic Operability Tests -
-
required, all drywell-pressure
.
suppression chameer vacum
'
breakers shall be operable -
except during testing and as -
.- -
stated in $pecifications
3.7.A.4.b. c and d, below.
Drywell-pressure suppression .
chamber vacuum breakers shall
be considered operable if:
(1) The valve is demonstrated to (1) Once each month each drywell-pressure
open with the applied force of suppression chamber vacu m breaker
the installed test actuator as shall be exercised and the operability
indicated by the position of the valve and installed position
switches and remote position indicators and alarms verified.
Indicating lights.
(2). The valve shall return by (2) A drywell to suppression chameer dif-
gravity when released after forential pressure decay rate test -
being opened b shall be conducted at least every 3
annual means,to y remote
within 3/32"or months.
of the fully closed position. *
.
(3) Neither of the two position
alem systens which annunciate
en Panel C-7 and Panel 905
when any vacum breaker opening
* *
onceeds 3/32". are in alarm.
Amendment No. 48
"
.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
,
.
. .. . . . . . . . . . . . . . . - . . .
~
e .
.
.
-
.
b. Re: orts of reviews encee:assed by See:icn 6.5.5.7 e. f. 4 ard 9
above, shall be prepared, a: proved, and forwarced to tne Seni:r
Vice President - Nuclear,with a c:oy to the Station Manage *
within 21 days fo11cwing the c:m letion of the review.
c. Audit re:o.ts enc:m:assed by Section 6.5.3.8 above shall te
': - forwarded to the Senior Vice President - Nuclear and t: wt nin
tre
* management posittens res:ensible for the areas audite:
* 30 days after c:e;1stion of the audit.
6.6 RE:*:~13*.! 0:0.RE';*! *:~*0's
The foll: wing ac:t:ns srall be taken in tre event of a re::rta:1e :::.r ta:::
. ; .. g ..,.. . . ....,. .... . . ,.. ..
- .....
.
: * .t ~. . .. . ...
*v ::- :-- a * - ** :- :
3. ~s: :: * - ' *::. e- i
., . ... ..
-
-1v'9-i: :/ ? . ::.; t: s..r* :2- .
*1 2;n'.
-
i.; g::! v ',:- : ,:- -
d
The following acticns snail te taken in the event a Safety Limit is e :'s::::
A. The : revisions of 10 CFR 53.36(:) (1) (1) shall De c: : lied ait".
immediately.
'
The Safe y Li .it Vi:latien small te rocceted :: the :--issi:n : e
8.
Stati:n Manager, anc :: tne NSRAC Chairman immediately.
shall be :recared. The es::*: s s
C. AbeSafety reviewed Limit by Violatien ne CRC. Ee::rThis re::rt small descrite (1) a::!f:a:*e
ete:ar. stances receding the vic14:icn. (2) e"ects of tse vi:'att:a
u::n facility c:e:orents, syste-s or s: ve:Wres, ard (3) c:r- ::'.e
action taken :: ;revent recurreace.
D. The $adety Limit Violati:n Re:ce: small be sub-itted :: t e *:--*ss': .
.
the N5RAC Chatr-an, and the Statten *anager wit .in 14 :ays o' : e
violation. .
6.8 8:CCIOL'8t$
A.
Written prece ures and acministrative colicies small te esta:'*sse:.
imole-ented and eaintained that meet or exceed tne recaire-eats sa
-
rec:-mendations of Se: tion
"A" of US*1RC Regulat:ry Guice 1.33. esce:: as prevt::: 5.1 and 5.3 of AN51 N13.7 - **
117 an:
Accenti:
6.8.8 and 6.8.C teicw.
Each pr:cedare of 6.!. A a::ve. and casaces treaet:. shall :e rev'e 3:': .a:
by the CRC and a::r:ved 3y tne C: Chat ar :r*:r to HP:'e'e*:
l 5.
i Ce"io"iCal*/ as set #i fI" * *'
These rice:Wres
4:Pinistra:1ve pr:cet.res."ev ewe smal' te
l
216 -
l
Amendment No. 30, 44, 19 , 74
. .
. . . -- - . - . - - . - - - . - , . - - _ , - . . --- - - . , - . - - . . - ,, - , , . - - - - - - - . - -
.
' '*- - - --- . . . . . . . . _ . _,
. . ,
NOTE: ORC review and ac:reval of procedures for venders / centra::ces. who have
a QA Program accreved by Boston Edison Cemcany, is not re:uired fer
work perfcmed at the vendcr/ contractor facility.
C. Temporary changes to procedures of 6.8.A above may be made provided:
.
1. The intent of the original procedure is not altered. -
* '
.
. .
*
2. The change is approved by two me-bers of the plant manage ent
staff, at least one of wnom holds a Senior Reactor 0: erat:r's
,
,
license on the unit affected. i
3. Pe change is d::umented, reviewed subst:uently by ne CR", a
a::revec y tne i.RC Chaiman ttnin 7 :ays cf imsle-enta:1cn. j
C. 'ir'
. tt a a
....-
:-::edu m.s to i-cle tet4......., . .
t*e "fre pa : :ti:n 3r:g-19 sna11 :e
., ... ....... ,
. . . . . .
-
-
. . .. .
.
. 14:d ti:- :: .: *:a:*e ** --: ~ .- -
---: *) . .
- - --
4 :n-21 4 : .* a t M s . - e - ' ' **.
-
te : : .
*
..- .... '
:: ; " ;;-
:: t. e : .it-
-d
t-a
--
C 1--
inf:r:2 an: .C.esa :: . : :.u.
A. Routine Re:cets
1. Startt: 8e :-t A sumary re:crt of plant startus and ;cwe*
escalatten testing shall te summitted follcwina (1) recei:t
of an coerating license (2) amenc ent to the Iicense invc1vi99
>
a plannec increase in c:wer level, (3) installation of fuel :nat
has a different design or has been manufactured my a different
fuel su: slier, and (a) modificattens that may have significantly
.
altered tne nuclear, ther-al, or hydraulic cerfernance of :ne
plant. The recort shall acdress eacn of the tests icentifie: in
- the F5AR and shall in general include a description of the
measured values of the a:erating conditiens or characteristics
obtained during the test pr: gram and a c:':arison of tnese va'.es
with design predictions and specift:sti ns. Any corrective act' ns
that were recuired to obtain satisfact:ry estrati n small alsa :e *
described. Any additional scocific detaf15 re:uired in license
conditions based on other c:x.it: ents sna11 he included in nts
report.
Startuo recorts shall be subMited within (1) 90 days (c11: wing
comoletion of the startus test program. (2) 90 days follcwtag
resuration or corrence-ent of c:-: ercial cwer c:erstion, or
3) 9 months follcwing initial criticality, whicnever is ea lies..
.
f the Startuo Re:crt does net c:ver att in-se events (i.e., in-
itial criticality, c:moletion of startue test program, and ess. :-
tien or c:vence ent of co r erical c*er ::eration), su::le e-tary
re: orts sna11 te submitted at least every inree mentns untti all
tnree events have teen c: nieted. .
.
- 2U -
Apen&ent No. 29 M, 46, 74
l
'
.
. - , . - _ . ._ - . _ . . - . __..- .__.- _ . _ . _ . - . - ,,.. -. ,_ ._. - __ %. - .. ~ . , - , - . - _ . - . . - - , _
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.. . , . - . _ . . - - _ .
.
.
.
*
l .
.. ,
2. _M_onthly Operating Report.
shutdown experience and forced reductions in power shall
mitted
Enfercssent
on a monthly basis to Director, Office of Inspection and
"
U.S. Nuclear Regulatory Connission Washington, D.C.
20$i5, to arrive no later than the 15th of each month' following
... the calendar montti covered by the report.
. i.
.
.
The Monthly Operating Report shall include a narrative sumary of
!
,
operating experience that describes the operation of t
3.
_0ceventional Excesure Tabu 11 tion. A tabulation of the nianter of
scation, utility anc other ;ersonnel (including contractors) re-
, ceiving exposures greater than 100 mram/yr and their associatad
man
. esm excesure ic:orcing to weer arf f:b 'unc:1- t, e . s . -* M t -
. . . . . . . . , . . ,..,, . . .
- . . . . .. .s
.
. . ...t .
- --
1. ; .. . .;.1 31 ;.2
.
, ....! - ::::
. :.- . :3 : - - s' - *
3..::: .~,. 2 : s r -:u s e. w : .
1:t:.
,
- -
'*7 :f L2 ;.R 23. Tila
.
. -
n.r: : v .- w: c.:y function., - .
cetry.2r , 'D, :P *"m badge.r.4 . as;;.T.atat cas-1 9 pc
less taar 205 ' :. e *idivic;al total ::s - etc
.ent:. ira 11 e *arei .;ta**'
I:n the aggrap :. not ac::untac
l
.
at.'.15: act of :ne :::si wnole becy dose rece vec
.
frc:n external sources snall be assigned to specific major mark functions.
.
.
.
.
. .
. .
. .
.
. .
,
.
.. .
.
.
.
.
..
.
I
.
-
-
,
*
.
.
'
.
.
F
218(thenextpageis220)
-- .-.
. . . .. . . . . . . . . .
. . . . . . . . . . .
-_ - -
._ __ _ _ - _ _ _ _ _ - . _ _ _ . - ___ - - _ - .-. - __. -__ - _ _ - ._-.
_
. .
e
_ __
. _ , _ . . . . . . _ . . _._ _ . . .
.
.
. .
,
.
3. Reportable Occurrences -
-
Repor:able occurrences, including corrective actions and measures
{ to prevent reoccurrence, shall be reported to the NRC. Supplemental
reports say be required to fully describe finsi resolution of 30
occurrence. In case,of corrected or supplemental reports, a licenses
event report shall be completed and reference shall be made to the
original report date.
1. Precet ?:etifiestion utt5 vrtteen follevue. The types of
events 112 ed bel:V snsil be reper:ec as expedi:1:usit as
possible, bu: vi:hin 21 hours by telessone sai cenitr:sd by
telegrs;n, =sil;rs=, or facsi=ile ::sas=1ssian o the
Director of the sppropris:e Regi nst Office, or his dest ns:e
-, 1 s : c.- -*-**2 f ir-- . ?>.1- - 2 ? ' **-v n: -4 * e i- : . v : .-
-
.
*
... . . . 1. . ... ;: . . . . . . I a..* :..
:! 1 ...;e a .::: ; .. . * .. ::. :-..t
. ;a
'
a li: n. . a s - . *4 :- f:r: -
. *2:* :ss,
.
as nee:ci. : 4:st:::nsi narrs::. .:. atsi :: - "d' 4::
ple:e a::p.: . :::n f ns ::::;=2- :as au:::. 1: in .
a. yailers f :ne resc::: ;r::ac:::n systa= or eener
syste=s subject to lini:ing safety system settings to
initiate the required protective function by the ci=a
a monitored parameter res:hes the setpoin: specified
as the li=iting safe:y systan setting in the technical
t*'
M
specif t:sti:ns or f:11ure to complete the required
*
protectiva function.
Note: Instrument drift discovered as a result of testing
need not be reported under this ites but say be
reportable under ite=s 6.9.3.1.e, 6.9.3.1.f, or
4.9.3.2.a below.
b. Operation of the unit or affected systems when any
parameter or operation subject to a lini:ing condi:1:n
ta less conservative than the least conservative
aspect of the li=1:ing condition for operation established
ta the technical specifications.
.
Note: If specified action is taken when a systen is found
to be operating between the sont conservative and the
least conservative aspects of a list:ing condi:isn for
operation listed in the technical spe:ifications.
,
the liatting condition for operation is not consifered
. to have been vioisted and need not be reported under chts
ites, but ic say be reportable under ites 6.8.8.0.b below.
.
.
.
2:0
_ _ _ _ _ .
-
.. .
*
... ...... _ _ . . . _ _ _
. .
c. Abnor:s1 degradsrien dircovered in fuel cladding, resetor
(' coolant pressure boundary, or primary contsin= cut.
Notc: Lcnkst.c of valve packing or gaskets withih the limits
for identified leakage set forth in technical, spec-
ifiestions need not be reported under this iten.
-
.
d. Reactivity ano=alies, involving disagreecent with I
the predicted value of reactivtty balance under
steady state conditiens during pever operation, grester
than or equsi to 11 ik/k; a :alculated reactivity
balance indicating a shutd:un r.argin less conservs:ive
than s ecified in th: techni:al 1:ecif t:sti:ns; sh:rt-
: ::: . :; -- '- : - r. - -~
:::: :r
. .. -... . . . . . ,
,
:"
. ..
.; ; ;.. e i. . ::.- :- ..:. r.
: . . . ::n; . : . -*
:. :. c:.t: . ..
9 S *'.i:h
e. Fail re er estfunctiun ef :- .T. ::::
prev:nts er : l.:: pr:v- :, .:. :a .. -
of the fun :icn..1 tc tar..sc..:r J ri s::--( s ) w .. ..
cope with accidents analyncd in the SAR.
15
f. Personnel error or procedursi insdccuscy which prevents
or could prevent, by itself, the fulfill ant of the
functional requircrents of cystc=s required to cepc with
( .
accidents snslyr.ed in the SAR.
,
. .
Note: For items 6.9.B.l.e and 6.9.B.l.f reduced redundancy
that does not result in a loss of system function
need not be reported under this sectien but r.ay be
reportable under ite=s 6.9.3.2.b s.sd 6.9.3.2.c bslow.
8- Conditions arising from natursl or msn-made events that,
as a direct result of the event require plant shutdesen,
operstion of safety systces, er other protective reasures
required by technical specifications,
'
h. Errors discovered in the transient or cecident snalyses
or in the =cchods used for such analyses ss described
in the safety analysis report or,in the bases for the
technical specifications that have or could hsyc ~
permitted'rcactor operation in s manner less conservative
-
than assu=cd in the analyses.
~
i. P'crformance of structures, syste=s, or components that
requires remedial action or corrective measures to
prevent operstion in a manner 1ers conserrative tnen
assumed in the accident,anslyscs in the ssiccy snslysis
- report or t echnical specifications bases; cr div::very
during plant life of conditions not specifically con-
sidered in the safety analysis report or techn:eni
specifications that require rent dial action or corrective
,
measures to prevent the ex2u(nce or development of an
unsafe condicion.
11/75
m 221
-
- - - _ _ - _ _ - _ - -
*.
. .... .. . . . . . . . - . . . - . . - - - - -. -- - -
. .
e
.
Note: This item is irrtaded to provida for reporting
of potentially Ccncric problene.
{"
2. Thirty Day Written Reports. The reportable occurrences
discussed beiuw shall be the subject of written reports
,
to the Director of the appropriate Regions 1 Office within
thirty days of occurrence of the event. The written report
shall include, as a minimum, a co=pleted copy of a licensee
event report form. Information provided on the licensee
event report form shall be supple::ented. as needed, by
additiens; narrative material to provide complete
explanstion of the circumstances surrounding the event.
-
. . . :a :.. .t-
: .c.. ;c: .: c :: -
. : . .3
s --
ta s a . - '; . . .: . s ' -
: s,s......-.
. b;r . cch de .o: - :vc - - . c .; -
.
f. .. u - f 9 .
fi.aett- 11 ::2irce:n:c
.
: .~ . n : ? : .- n a. .
b. Ccaditions ;e: ding en operat:ca in c degraded meda
pernitted by a limiting condittoi for operstion or
. pisnt r.hutdown required by a limiting condition for
, operation.
( Note: Routine surveillance testing, instrument calibration,
or preventative maintenance whicn require system
configurations as described in items 6.9.B.2.a and
6.9.B.2.b need not be reported except where test results
themselves reveal a degraded mode as described above,
c. Observed inadequacies in the impler.cntatic.n of admin-
istrative or procedural controls :hich threaten to
cause reduction of dcarce of reduc.dsney provided in
reacter proccction syste.:s or ent,inc-cred caicty featuru
systems.
d. Abnorrsi degrsdation of eveters other than those
specified in iceu 6.9.B.I.c above designed to contain
radioactive material resulting frca the fit sion
process.
Note: Scaled sources or enlibration sources are not inciv/:d
, 'under this item. l.cakar.c of velve packinr or gar.kcts
-
within the limits for identificd leal:sge 'et ferth in
technic al spccifications need cut b.: reported under
thic item.
-
e. Any changes in corporate or sestion organization as described
f* in Section 6.2. la
t
222
.
4 p.
_ _ _ _ _ - _ _ _ _ _
.. .
' --. , .
, ...
_
, -.
.- . . . . . .
i
'"
J. 4
*
- .; - .
.
,
*
C. Unieve Reverti= Reevirements
1. Radioactive Effruent Release Resort
.
A report shall be submitted to the Com=1ssion within 60 days after
January,1 and July,1 of each year specifying the quar.'tity of 'ench
* ef the principal rhdionuclides released to unrestricted areas in
liquid and gessous affluents during the previous 6 months. The
format and content of the report shall be in accordants with
Regulatory cuide 1.21 (Revision 1) dated June. 1971..
,
2. Enviren ental freersw Sata .
a. Annual Repert. A report on the radiological environmen al
surveillar.:n program for the previcus 12 :sonths of operation -
shall be submitted to the Oiratter of the NXC pe:;i:nal Cffi:e
.
:- i - , " 01 t: t - . ?!!!: ef Fu: lear 't.-::r
. . . . .
... .
, 4 .
-
,
.. a:. :s .a :.. . - -
tr. ..
-- -
. : :. : .: . r. . : .t. h2a e- .:: :' :-
.. g: . sa. . na
*
. 2. .;.2 :: -
hr -
,st r. . .:- 2 ::.. : ...c4. ..a..a.
. pert::. i.,s . ;: .:: t
eta sti:.zi =ent- as ::- s n n'. .n e. -
r.:. : . ir: . e .:.. *
..:..- 2. -
.. :a:ts .
v:: : 31 *. . a::: : ri- at
t*i
;ia.t ::::2:.:n :. :.e as tt: ::. .): :e:: .
of t.
also include the results of any land use surveys vnten a!!act
the choice of sample locations. If har=ful effects or evidence .
' of irreversible damage are detected by the sonitoring, the
licensee shall provide an analysis of the problem and a preposed
sourse of action to alleviate the problem. ,
.
Results of all radiological environmental samples shall be
'
suus:arised and tabulated on an annual basis. In the event that
.. -.. . .- .. ..... some results are not available within the 90-day period, the
report shall be submitted, noting and explaining the reasons
.
for the missing results. The missing data shall be submitted
.
as seen as possible in a supplementary report. ,
*
'"
3. Thefollowing event shall be the subject of a written report to the
Director of the NRC Regional Office (with a copy to the Director,
Office of Nuciaar Rasesor Regulation) within 60 days of the occur-
-
rence of the event naasured levels of radioactivity in an environ-
sental sampling medi:n dotarnised to exceed the reporting level
values of Table 6.9.C-1 whan averaged over any calendar quarter
sampling period. When more than one of the radionuclides in Table
6.9.C-1 are detected above lower li=1ts of detection (:*3) in the
sampling sodium, this report shall be submitted if: * -
co teentration (1) * * Concentration (2) + ... g 1.0
Limit Laval (1) LLast Level (1)
- -- When radionuclides other than those in Table 6.9.C-1 are detected
above LLD and are the result of plant effluents, this report shall
be submitted if the potential annual does to an individual is
equal to or greater than quarterly and/or yearly li=its of Tabla
6.9.C-2. This report is not required if the usesured level of
radioactivity was not the result of planc effluentI; however. in
such an event, the condition shall be reported and described in
the Annual Radiological Environmental ?fonitoring Raport.
Amendment No. 57
- 223-
.
._.._ . . . . . . _ . _ _ .. . ..
- ---_ . . . . . . . _.
-- - - - - _ - _ _ _ _
~
_ l _ _ .. . . . - _ -
-.__l-.-_--_'-.- . . _ . ...5......_..___-- ---
)
So THEORY 0F NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 15
_---
--------------------------------------
4
-__----_------
~
ANSWERS -- PILGRIM ,
-85/05/14-KVAMME, J. '
ANSWER 5.01 (2.00)
>
Disagree (.5) Thermodynamic efficiency is a comparison of enersy in f
versus energy out. The increase in generator output resulted from
decreasing that amount of steam flow diverted to the HP FH heater (.5)
This condition requires additional enersy output from the reactor
to raise the FW to the same saturation temperature as before.(.5)
Therefore the thermodynamic efficiency of the plant has decreased.(.5)
-
(2.0)
<
REFERENCE
- Thermodynamic Handout, Section 14
i ANSWER 5.02 (2.50)
a. Rod worth increases.(.5) Increasing coolant temperature decreases
moderator density which increases the thermal diffusion length.
An increased thermal diffusion length increases the probability
, of neutrons interacting with control rods.(.75) (1.25) i- b. Rod worth increases.(.5) As power increases, void fraction
increases. This results in a decrease in moderator density
which results in neutrons travelins farther to become thermalized.
Rod worth is directly proportional to thermal diffusion lensth.(.75)
(1.25)
REFERENCE
Rx Theory, section 31, ps 6-7
ANSWER 5.03 (2.00)
No.(5) Even though each individual pin would be protected from
excessive plastic strain in this case, the decay heat load from
the collection of pins would tend to elevate mid bundle pins to
sceater than 2200 F durins LOCA uncovery conditions.(.75) Therefore
i
a limit (APLHGR) is also placed on the average of the LHGR's of
i
those pins. This limits the decay heat load.(.75). (2.0)
REFERENCE
Heat Trm,nsfer Hendout, Sections 15, 16
-
4
9
e , -- - -- . = , - , , , . - - - - - , . -
. ..
-
. .
. , .
._. ; _ - _ _ - _ - - - - - _ - - .
_ ... . . - . . . - . . . . . .. --- ~ _-_. -_. -
.
5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGC 16
____ _
______________________________________
______________
, ,
ANSWERS -- PILGRIM ,
-85/05/14-KVAMME, J.
i
1 , l
!
ANSHER 5.04 i3.00)
1. SINGLE PHASE FORCED CONVECTION- This pattern is limited to the very 1
bottom of the channel and is characterized by the absence of any
steam bubbles.
t
2. SUBC00 LED BOILING- Bubbles begin to form as nucleate boiling; bubbles
form on the surface but collapse as they enter the bulk coolant stream
since the majority of the coolant is still subcooled.
3. BUBBLE FLOW- This pattern is characterized by nucleate boiling with
!
the bubbles entering the coolant streami the bubbles do not collapse : and this ' bulk boiling'. The bulk temperature is ~ saturation. The , bubbles are not coalescing and have a low quality,
j
'
4. SLUG FLOW- The bubbles now besin coalescing or "growins together',
The bubbles are formed in nucleate boilinc, but now instead of
collapsing in the coolant, the bubbles tend to collect in vapor
sluss.
5. ANNULAR FLOW- The vapor forms an almost continuous phase in the
coolant channel. The slower moving liquid travels along the fuel
element surface and heat transfer takes piace via conduction with
; almost no boilin3 taking place. The vapor travels up the center
of the channel in a continuous stream.
'
'
(Any three required at 1.0 each)
(3.0)
REFERENCE
-
HT XFER, section 9
,
.
I
.
t
-
1
I
I
1
1
- , - --
.. - - . _ , , , - . . - - - . , _ . - . . . . - . . _ . - - - - --,,-
'
, _ _ . _ _ . _ - _ . _ _ _ _ _ _ . - . . . . . - . . . . . . . . - -. . ..... L.._ ._ ..._..____-._2- _-.
.
i
5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 17
----
--------------------------------------
, _---_---_---_-
~
! ANSWERS -- PILGRIM ,
-05/05/14-KVAMME, J.
,
ANSWER 5.05 (3.25)
a. Doppler, Moder;ator Temperature, Void (.25) .
b. As the rods are withdrawn, power level increases. But the additional
heat senerated is not immediately transported to the coolant. The 1
fuel time constant of 8 or 9 seconds slows the rate that the heat
generated in the fuel is conducted into the coolant.(.75) It taken
about 3 time constants or 30 seconds for the total increase in heat
- senerated to be transported to the coolant. So the fuel temperature
- rises first, causing the doppler to be the first effect.(.75) The
l next effect would be the moderator temperature, ac the coolant is i
heated to saturation.(.75) Finally comes the effects of voids,
as the heated senerated in the fuel boils the water flowing
throush the core.(.75)
REFERENCE
Rx Theory, section 26
t
ANSWER 5.06 (2.00)
- Core flow increases due to increase voidins/ buoyancy (1.0) but then
-
stabilizes as increased pressure drop cancels out the natural circulation
driving head.(1.0)
REFERENCE
j Pilgrim Exam Question Bank l l
ANSWER 5.07 (2.25)
er sam wlv e
a. Heff decreases due to the buildup of fission product poisons.(.75)
b. Keff increases due to the burnout of sadolinia and buildup of
Pu-239.(.75)
c. Reff decreases as fuel burns out.(.75) l
4
(2.25)
REFERENCE
>
Px Theory, section 19
)
e
J
-
4
I
I i
*
1
. - . - - . _ , - - _ _ _ _ - . , . , _ _ _ . . _ _ _ _ _ _ . _ ._. . . . . . _ _ _ _ - - - - , . . - - - _ . . - -
... . _ _ _ . _ _ _ _ . _ _ - _________.__.. ~ _ __ ~ . ._... -._-.-.._____.~
i .
5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 18
____
______________________________________
-
______________
,
ANSWERS -- PILGRIM ,
-85/05/14-Kl>AMME, J.
-
ANSWER 5.08 (2.00)
'
The reactor is now producins less steam to so to the turbine. There
will be less extraction steam and reheater drain steam soins to the
feedwater heater.(1.0) Therefore less feedwater heating will occur
- resultins in colder feedwater entering the vessel (.5) which will
cause reactor power to increase about 3% from the positive reactivity
addition (alpha m).(.5)
'
REFERENCE
Reactor Theory Section 26
: Recirculation LP
Hain Trubine LP
ANSWER 5.09 (2.00)
With the reactor shutdown by 1% ac measured at the time of the peak
Xenon, the Shutdown Marsin will decrease as Xenon decays. Since peak
Xenon reactivity is greater than 1% dk/k, a reactor restart would
occur as peak Xenon decays in the next 20 hours.
} _ REFERENCE
i Reactor Theory Section 33
,
ANSWER 5 10 (2.00)
a. The condenser acts as a saturation system.Therefore,the lower the temp- I
erature,the lower the absolute pressure will be or the better the vac-
uum. ( 1.00 )
b. 1. Non-condensable sases . ( 0.33 )
2. Circulatin3 water system * flow rate '
. ( 0.33 ) '
3. Back pressure in the condenser exhaust system. ( 0.33 ) ;
j REFERENCE
l Pilgrim Thermodynamics Student Handout
l l >
e
! '
-
.
!
f I
i
, - - . - . _ . _ ._. _ . . . _ , _ - . _ _ _ _ . _ . _ . _ . _ . _ _ _ _ . _ _ _ _ , _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ . _ _ _ _ ~ . _ _ , _ _ _
______ _
..
.
. . _ - . . - . - - . . . - - . . . . . . - . . - - - . . . . . . . ----.. .--- - --. . . . . . - - - - - - . . - - - . - - ~ ~ . - - - - - - - . - - - -
.
5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 19
---- --------------------------------------
-_----_-------
,
ANSWER 3 -- PILGRIM ,
-85/05/14-KVAMME, J.
ANSWER 5.11 (2.00)
TRUE (0.5)
Usins the equation Power = Power (initial) x in time / period and
solving for time results in the equation:
l
time =Peried x in Power / Power (initial)
From this it can be seen that since 5/1 yields the same value as
50/10, and since all other factors in the equation are equal, the
time is equal (1.5)
REFERENCE
Reactor Theory
.
-
-
. ..
-
. _ _ _ . _ . . . . . . . . - = _.. .-_. _ _ _.__ _-__ . _ . . . . _ . . _ . _ _ _ _ _ _ . . . . _ _ _ - - - . .
.
4
,
6. PLANT SYSTEMS DESIGNr CONTROL,-AND INSTRUMENTATION PAGE 20
______________________________________________________
ANSWERS -- PILGRIM -85/05/14-KVAHMEr J. .
9
-ANSWER -6.01 (3.00)
~
a. 1.A' RHR HTX
2. ' ARC' RHR Pump Mechanical Seal Coolers
3. 'ARB' RHR Pump Area Coolin3 Coils
4. 'ARD' RCIC Pump Area Cooling Coils
5. 'A' Core Spray Pump Thrust Bearing
(4 required G .5-esch)
,
(2.0)
b. Each loop's nonessential loads can be isolated by a single handswitch >
for that loop in the main control room. (1.0)
RSFERENCE
, pr-CCW LP 130 j ANGHER 6.02 (1.00)
- "C"
REFERENCE
Procedure 2 2 99
:
ANSWER 6.03 (3.00)
a. 1. acoustic monitior indicating lights on panel c-171
2. RV temperature recorder on the back panel would increase
water level should be dropping (- l f t/ min )
'
3.
4. torus level would be oscillcting
-
5. torus temperature would increase
6. drywell pressure would increase (5 required R .25 each)
b. 1.no(.25) As lons as there is a valid high drywell pressure signal
the hi h 3 drywell pressure contact will not open and the ADS timer
not reset.(.5) (.75)
2. If the timer has not timed out, the timer will deenersine
and the ADS will not initiate.(.5) If the timer han timed
out, then ADG has initiated and will continue to operate
even with level restored.(.5) (1.0)
t
>
.
.
.
4
. . . _ . _ . . . - _ . . _ _ . _ . . . _ _ , _ ._ . _ _ - _ . .
-
.
.
. .
,_. . ..... ......e. --. . . . - . - . . . . . . . . . . ~ . .. . - . . . . - - . . ._ _-_
.
,
6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 21
______________________________________________________
ANSWERS -- PILGRIM -85/05/14-KVAMME, J.
.
'
REFERENCE
.
ADS LP 423
ANSWER 6.04 (3.00) *
'
a. SRM upscale >10e5 cps OR
3 '
SRM inop OR
SRM downscale < 3 cps OR
i SRM <'100 CPS, detector not full in (.25 each) (1.0)
b. IRM upscale > 100/125 OR
1RM inop OR
IRH downscale < 5/125 OR
, IRM not fully inserted and mode switch not in run (.25 each) (1.0)
c. All IRM ranse switches are on rango 3 or above and the IRM
channels are indicating above the downscale trip or(.5)
SRM channel readings > 100 ces(.5) (1.0)
REFERENCE
LP 964, 65, 66, 67, 60
i
' ANSWER 6.05 (3.00)
-f ~ , a. Ion chamber .
b. Scintillation
c. Ion chamber
d. Scintillation
e. GM tube (.6 each) (3.0) t
<
' REFERENCE
RAD'N MONITORING LP t51,53,57,59,111
i
l
J
, I
'
'
.
s
t
.,
4
_ _ . ,
. . , , . , , . . , ,
-
. . . , , _ . ,. ,--, , _._.,.....__,_._.....,...__.___,.,...._._._.-_.y_m, -
. , _ , . . . _ , . . , _ - ~ . - . . - . , - , , , , ,
. _ _ _ -
,__- _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ - - _ _ . _ , _ - - _ _ _ _ .
_ _ _ _ - ____-_ _ - ____
.. .. _ - ~ ... .
--s......--.........-.-.---._-----..-----.-.----------
.
f
6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 22
______________________________________________________
ANSWERS -- PILGRIM -85/05/14-KVAMME, J.
.
#
ANSWER 6.06 (3.00)
a. 1. To be conservative the LPRM outputs (local core power) are
increased to the averase core power making this output power
level closer to the rod block flow biased trip setpoints. (1.0)
,
2. Several of the bisher reading LPRM's may be bypassed so gain
is increased to compensate for the low average. (1.0) '
b. 1. The RBM is automatically bypassed when its reference APRM goes
,
below 30%.
2. The RBM is byposed when an edge control rod is selected.
3. Con bypass one RSM channel manually by use of the bypass
switch or " joystick'.
(.33 each) (1.0)
REFERENCE
IESSON PLAN 166,60
ANSWER 6.07 (3.00) :
a. 'D' feedwater line flow signal fails high and causes an increnne
in the total feed flow signal, A feed flow / steam flow mismatch is
generated which causes feedwater control valves to close. Since
steam flow is now greater than actual feed flow, reactor vessel
level decreases. (1.0)
b. Channel 'A' reactor level detector signal fails low and causes a l
level error signal to be generated. Since indicated level is lower
than desired level, the feedwater control valves open to increase !
feed flow. This causes a feed flow / steam flow mismatch which tells '
the control valves to close down. The system is more level
dominant and the initial response is to increase reactor vessel ;
level. (1.0) i
c. Loss of signal to a feedwater valve causes the valve to receive
a lockup trip. The 'A' FRV will continue to control level, there- '
fore reactor level will remain tle same. (1,0)
(.25 for re,ponsor .7S for reccon) !
REFERENCE !
RVLC LPt'02 i
:
t
*
<
o
-- . _ _ _ _
- - - - - - - - - - - - - - -
.
,
. . _ . . . . .. ... .... . - .. _ - ....._ . ~ .-.._.. _ . . - _ _ _ . _ - . _ . _ _ . _
.
6. PLANT SYSTEhD DESIGN, CONTROL, AFD INSTRUMENTATIDH PAGE 23
______________________________________________________
ANSWERS -- PILGRIM -85/05/14-KVAMMEr J. .
.
ANSWER 6.00 (3.00)
a. Three required for full credit. ( 0.66 each )
1.
Ateachtbstingplateauduringstar)tuptesting.
2. (Every full power month of operation to recover sensitivity lost due
to detector fuel depletion.
3. When operating mode has changed significantly. iel Losc of FH. heater
4. After refuel activities. Due to refueling the core has been altered
, significantly causing radial and axial flux profiles to change.LPRM
detectors have been changed out and the electronic circuitry has been
idle.
b. To maintain the relative humidity in the guide tubes at a constant value
over the length of the detector travel and maintain a dry atmosphere in
the drive mechanism a index enclosure. ( 0.50 ) If the porse sys.is in-
operative for extended periods,corrision will build up on the helical
wrev around the drive cable,and there will possibly be inculation break-
down resulting in elec. signal loss . ( 0.50 )
REFERENCE
Procedure 2.2.69, TIP System
ANSWER 6.0? (3.00)
a. Hi dnywell pressure 2.5 psis OR (.5)
Lo La Rx water level -49' and less than 400 psis Rx pressure (.3) (1.0)
b. With an auto initiation signal present, high drywell pressure and
core coverage greater than 2/3 you can spray the drywell by
operating the manual override c/s and then open the spray valves.(1.0)
With less than 2/3 coverage and high drywell pressure you must
operate the keylocked 2/3 coverage bypass switch prior to the
manual override c/s.(1.0) (2.0)
REFERENCE
LPCI LP Procedure 2.219
.
-
- - - ,
. ._
.
. - . . . . . _ ...-. - .. -.---- - -_ -..-.... ..-. . . - - - - - _
.
7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 24
~ ------------------------
~~~~R 5U bl65565L 6UUiR5L
____________________
,
ANSWERS -- PILGRIM ,
-85/05/14-KVAMME, J.
ANSWER 7.01 (2.25)
1. Both RBM channels shall be operable OR (.75)
2. Control rod withdrawal shall be blocked OR (.75)
3. The operating power level shall be limited so that MCPR will
remain above 1.00 assumins a single error that results in com-
plete withdrawal of any single operable contro] tod. (.75)
REFERENCE
Docton Edison Procedure 2.4.10, Limiting Control Rod Patterns, ps2, Rev.5
ANSWER 7.02 (2.00)
a. Off-Gac High radiation alarm - No automatic action (0 5)
Off-Gas High High radiation alarm - Initiation of Off-Gas Isota-
tion 15 minute timer. (0.5)
b. Upon isolation of the Off-Gas line-
1. Manually scram the reactor
2. Close the MSIV's
3. Refer to procedure (5.4.1) for Closure of MSIV's on High
Radiation.
(2 required 0 0.5 each) (1.0)
REFERENCE
Doston Edison Procedure 2.4.10, Rapid Increase in Off-Gas Activity,
ps.3, Rev.5
.
ANSWER 7.03 (3.00)
a. Instrument Dus Y-1 (0.5)
b. Low vacuum to.53 as the off sas isolation valves fail closed Co.5](1.0)
c. Feedwater and condensate pumps E.253 must be stopped due to
no minimum flow C.253 and the reactor vessel isolated C.253 to
preserve inventury C.253. Turbine must be tripped E.253 because
the turbine trip from the reactor scram will not occur C.253. (1.5)
.
-
e
_ _ . _ _ . . -
, - _ . . .- _. _, , ,
-- . .
.
, , . . . ~ . . ~ _... . . . . . . . . . . . . . . - - .. ..._-.....~._-..-_._2.
.
7. PROCEDURES - NORMAL, ABNORMALr EMERGENCY AND PAGE 25
~ ~~~~~~~~~~~~~~~~~~~~~~~~
~~~~R565UL65555L 66 sir 6L
____________________
.
ANSWERS -- PILGRIM ,
-85/05/14-KVAMME, J.
REFERENCE
Doston Edison Procedure 5.3.7, Loss of Instrument Power Dus Y-1,
pg.2, Rev.6, and Question Sank
ANSWER 7.04 (2.50)
1. Entry and work in areas having radiation levels equal to or greater
than 100mr/hr.
2. Entry and work in areas having > 22,000 dpm/100 cm.cquared betc
'samma and/or 200 dpm/100 cm. squared alpha activity
3. Handlins of neu or spent _ fuel
4. Removal of radioactive or contaminated material from underwater
storase
5. Openins any process system through which radioactive liquid or
sases may escape to the work area
6. Anytime airborn radioactive material exceeds 25% of 1 MPC
7. At discretion of IIP senior supervisory personnel
D. Cotting or scindins of contaminated material > 1000 dpm/100 cm.
squared smearable and/or 0 1 mc/hr fixed
9. For all activities conducted under the Dy product Meterial License
(5 required 0 0.5 each) (2 5)
REFERENCE - '
Doston Edison Procedure 6.1-022, Radiation Work Permit, ps.2-3, Rev.15
ANSWER 7.05 (3.00)
a. One operator - 23' 4kv switchsear area (.25)
One operator - 37' 4kv switchsear area (.25)
CR0 to RPS MG set room (.25)
Operating Supervisor to Inst Rack 2205 or 2206 (.25)
b. Open the breaker to the APRM's (at the RPS power panele) (0 5)
Preferred because tripping the RPG supply infars a loss of
power and causes undesirable events such as early closure of
MGIV's to take place. (0.5)
c. Two RFP's tripped when unit trips (0.5)
One RFP tripped when level starts to ineraase (0.5)
REFERCNC'E -
Doston Edison Procedure 2.4.143, Ghutdoun from Outcide Control Roome
pg.6-0, Rev.3
-
6
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
. _ _ . _ . ___. . . _ .. . .. _ _ _ _. _ _ _ _ _. _. . . .... . _ _ . _ .. . .. .. __ __
.
.
7. PROCEDURES - NORMALr ABNORMAL, EMERGENCY AND PACE 26
~
~~~~E565UL65IEEL C6UTRdL'~~~~~~~~~~~~~~~~~~~~~~~
____________________
.
ANSWERS -- PILGRIM ,
-BS/05/14-KVAMME, J.
ANSWER 7.06 (3.00)
a. 1. Low vacuum alarm at 26"
2. Reactor scram at 23'
3. Turbine trip at 20'
4. Bypass valves close at 7"
(.25 for action and 0.1 for setpoint) (1.4)
b. 1. Check the steam pressure at SJAE supply
2. Check off sas flow
3. Check the valve line-up to and from SJAE's
4. If valves have closed, establish a normal valve lineup for
the system
5. If steam pressure regulater has failed, operate the bypass to
reestablish the proper steam pressure
6. Check loop seal on after condenser
7. If the problem still e:tists, change over jetu
(4 required 0 0.4 each) (1.6)
REFERENCE
Doston Edison Procedure 2.4 36, Loss of Condenser Vacuum, pg.2-3r Rev.5
ANSWER 7.07 (3.00)
a. 1. When reactor pressure is greater than 100 psia
2. If reactor pressure is less than 100 psig - either high
drywell pressure OR low water level. (counts as two)
3. Primary containment isolation
(2 required at 0.5 each) (1.0)
b. 1. Check reactor vessel temperature
2. Check reactor vessel water level normal
3. Check power available to 4160 v. buses A-5 and A-6 and that
120v. (RPS) power is available to the PCIS logic
4. Determine the cause of the isolation and take corrective
action
(3 required 0 0.5 each) (1.5)
c. Increase reactor water level to +41' or greater (0,5)
RETERENC'E
Doriton Edicon Procedure 2.4.25, Loss of Shutdnun Cooling, pg.2-3 Rev.6
.
.
- _ - . _ - __
. .
. . ... . _ . . . . . . . . . _ _ _ . . _ _ - _ _ . - . _ _ . - . . . . . ~ . . . . - . . . - ~ . - - - . . - . . . . . . . _ . . . _ - - . -
.
m
.
4
7. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 27
~ ------------------------
~~~~ R5656EUU5EAL EUUTR5L
i ____________________
.
ANSWERS -- PILGRIM ,
-85/05/14-l(V AMME , J .
ANSWER 7.08 (2.00)
a. Whenever the reactor is critical or when the Rx water
temperature is above 212 degrees and the head vent closed. (1.5)
b. 24 hours. (0.5)
REFERENCE
Procedure 2 1 1 Rev. 40 pg. 4
i
ANSWER 7.09 (3.00) g, n.T pla c 3;
a. One condensate)punip is required for each feed punip in
servicer Cud > tne sequential trip selector switch to ON unless all
condensate and RFP's are in service.(1.00)
b. The same C/S is used to start the lube oil pump and the R.F.P.
At the first operation of the C/S, the lube oil pump starts. It
spring returns to neutralt and at the second operation, the RFP
is started. (1.00)
c. By throttling the TDCCW oil cooler outlet valve (0.5)
d. Before the first feedpump reaches 90% pump capacity.(or prior
to exceeding 30% reactor power) (0.5)
REFERENCE
. REFERENCE: Procedure 2.2.96 Rev. 15 pp. 9, 10
ANSHER 7.10 (1.25)
a. 1. Rx vessel shell adjacent to shell flange
2. Rx vessel shell flanse
3. Recirculation loops ARB coolant (.25 each) (.75)
b. When the difference between any two readings taken over a 45
minute period is less than 5 F. '(.5)
REFERENCE
Startup from Shutdown Procedure 2.1.1 pg 2-3
.
.
~-
-
? ,
. . . . . . . . . . - ._ _ - - - _ - - '
. - - . - - - . . . . - - - - - - - - - - .- - -- - .... ---
.
l
.
j 8. ADMINISTRATIVE EROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 28
.
- ANSWERS -- PILGRIM -85/05/14-KVAMME, J.
. ,,
. -
, ANSWER 8.01 (2 50)
a. Safety Limits are limits upon important process variables
below which the reasonable maintenance of the cladding and
.
primary systems are assured. LSSS's are settings on instrumentation
l which initiate automatic protective action at a level such that i ,
the Safety Limit will not be exceeded. (1 5)
b. The LCO's specify the minimum acceptable levels of system per-
formance necessary.to assure safe startup and operation. Hhen
these conditions are met, abnormal situations can be safely
l controlled. (1.0) i i REFERENCE
t
'
TS Section 1, pg 1
ANSWER 8.02 (3.00)
1
a. Jumpers are not authorized on safety related circuitry through
use of SRO procedure changes. (1.0)
) b. 1. Any Boston Ed. son Management (non-union) person permenantly
assigned to Pilgrim. (1.0) t
j
'
2. i. The intent of the original procedure is not changed.
ii. The change is documented and reviewed by ORC and approved
j with in 7 days. (1.0)
i
;
l REFERENCE '
! procedure 1.2.4-9 1 t
- ANSWER 8.03 (3.00)
a. prompt (.5) T/S 6.9.B.1.F (.5) (1.0)
I
b. 30 day (.5) T/S 6.9.B.2.A (.5) (1.0)
- c. Prompt (.5) T/S 6.9.B.1.G (.5) (1.0) I
i
REFERENCE
i T/S 6.9 i
T/S T.Jble 3.1.1
i
*
1 ! 1
1
e
4
'
.
. t
. .
l
6
. - . - . . . . - - , - , . . . , , . . . , , . , , , , . , - - . . - _ , . _ _ .
_
_ _ _ _ _ _ . . _ _ - - - . - . . _ . , _ . - _ _ _ - -
_ .I
_ _ _ - _ _ _ _ _ _ _ _ _ _ __ ._- ___ ___ --_ -_ ___ _ _ ___-____ __ ______ _____________-________ _-_ _ _ _ _ _ _ _ _ _ - _ _ - _ _ - _ _ _ _ _ _ _ _ _ .
. . _ . . . . .._.. . - _ _ . - ~ _ . . . . . . . . . . . _ - - . _ . . . . _ . . . . . . . = _ _
,
8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 29
__ _______________________________________________________
ANSWERS -- PILGRIM -85/05/14-KVAMME, J.
.
Off Normal Procedure 5.3.9 page.,1
ANSWER 8.04 (2.00)
1. Emergency Safesvards Equipment Checi. list
2. Watch Engineer /STA Turnover sheet
3. Shift Supervisor Turnover sheet
4. RO Relief Checklist
5. Radweste Turnover Sheet (.4 each) (2.0)
REFERENCE
Procedure ,1.3.34
ANSWER 0.05 (1.25)
.
A seram report shall be prepared whenever the conteal rods ore
sserammed in from a situation where fuel is in the RX vessel
i arid more than one rod is wit,hdrawn.(.75) The Watch Engineer on
duty at the time of the scram is responsible.(.5) (1.25)
REFERENCE
4
Admin Procedure 1.3.9-3
.,
ANSWER 8.06 (2.50)
a. Red tass are applied to protect personnel from injury and equipment
from damage by forbidding the operation of the devices which could
result in equipment being energized, mechanical movement or fluid
flow.(1.0) Red tass are placed on equipment IAW WRP prior to the
besinning of the work.(.5) The tas must be removed upon completion
of the work, er :t ^Se H' t' - r '- t/- whichever occurs
i ,
first.(.5) (2.0)
b. When t are i alled w/o an accompanying WRP. oe dew (.5)
. rep * * by ut. E pser
REFERENCE
4 Tassing Procedure 1.4.5 ,
.
_
. _ - _ -_ . _ _ , _ _ _ , , . _ _ _ _ . - _ _ . _ _ _ _ _ _ _ _ _ _ _ , _
- _ . _ _ ,
.
.
. . . . . . . . . . - . . . . ~ - - - - - - - - - - - - - - - . . . . - . ~ . - - - - . . . , . - .- . . . . - . - ~ ~ - - -
,
D. A9MINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 30
__________________________________________________________ f
ANSWERS -- PILGRIM -05/05/14-KVAMME, J.
.
.
ANSWER 8.07 (3.00)
1. Any time irradiated fuel is in the reactor vessel and reactor
coolant is above 212 F, reactor coolant leakage into the primary
containment from unidentified sources shall not exceed 5 spm. (.75)
2. Total reactor coolant leakage i.ito the prima,ry containment {
shall not exceed 25 spm. (.75) '
Limit 1 above exceeded on day 2 1600-2400; 2496 sal.=5.2 gem (.75)
Limit 2 above exceeded on day 1 1600-2400; 4
10,440 gal + 1824 sal = 12,264 gal [
= 25.55 spm (.75) j
REFERENCE
TECllNICAL SPECIFICATIONS 3.6.C l
1
ANSWER 0.00 (1.75)
Must be in cold shutdown by 1700 on 5/15/85 (.5)
{
Section 3.7.A.1 KEL of T/S allows 4 HRS + 6 HRS + 24 HRS to
cold shutdown on containment dp. (.625) ;
Section 3.3.G T/S allows 24 HRS to cold shutdown on drain valves. (.625)
'
REFERENCE '
TECHNICAL SPECIFICATION 3.3 & 3.7
,
&
+
0
r
!
- i
,
1
. - - - - -
- _ _ _ _ _ _ _ __ _
, _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
. _ . _ , . . _ . . . . . _ - _ _ - . . . ..
. . . . , -
.
8. ADHINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 31
__________________________________________________________
ANSWERS -- PILGRIM -85/05/14-KVAMME, J.
.
.
ANSWER 8.09 (3.00)
'
a. The area shall be baricaded and conspicuosly posted as a High
Radiation Area and entrance thereto shall be controlled by issuance
of a RWP. (1.00)
b. In addition to those requirements stated in part 'e", locked doors
shall be provided to prevent unauthorized entry into such areas.
The keys shall be under the administrative control of the Chief
Radiological Engineer (1.00)
c. An additional lock, keys will be maintained with the watch engineer,
HP rep. present during entry to monitor dose rates, entr> requireu
authori::otion from watch engineer or station manager. (1.00)
REFERENCE
Procedure 6.1-012 Rev. 12
ANSWER 0.10 (3.00)
a. Whenever the reactor mode switch in in the "RUN" po.ition. (1 5)
b. 1. High reactor dome pressure - 1175 psi 3 +- 15 psig (.75)
2. Lo-Lo reactor water level- (-49")(.75) (1.5)
REFERENCE
Tech Specs Sect 3.2.G
.
en
_ . . , ~ - _ - . , _, . _ . . . - - . _ , _ - _ - . ,_
. . . . . . . . . . . . . . - . . . . .. . - . . : . _.u
. .
.
.
U. S. NUCLEAR REGULA10RY COHHISSION
REACTOR OPERATOR LICENCE EXAMINATION *
FACILITY: PILGRIH
____-___-_______----_-___
REACTOR TYPE: BWR-GE3
__--_________ ----_______
DAIE ADMINISTERED: 05/05/14
_________-_-_______--____
EXAMINER: CRESCENZO, F.
APPLICANT: _ _ __ ___ ______
INSTRUCTIONS TO APPLICANT:
___--. __---_-____-____---
Ilse separate paper for the answers. Write answers on one side only.
Staple question sheet on top of the answer sheets. Points for eacii
question are indicated in parentheses after the question. The passing
3rade requires at least 70% in each category and a final grade of at
Icest 80%. E:< amination papers will be picked uP si:< (6) hours arter
the examination starts.
% OF
CAfECORY % OF APPLICANT'S CAfEGORY
VALUE TOTAL SCORE VALUE CATEGORY
*
.. -__-- __-... ----___--__ ---__--_ __---_-_---__---______-______ ..__.
75.00 100.00
.----___ __.--- ----------- -------- 4. PROCEDURES - NORMAL, ADNORnAL.
E'HERGENCY AND RADIOLOGICAL
CONTROL
25.00 100.00 (OTALS
. ____.. ___ .. -____-____. _____ ..
FINAL GRADE _________________%
All vork done on this c:< a m i n a t i o n is my own. I have neither given nor recetved aid.
~
SEPLfCdUT 5~5555diURF~~~~~~~~~~~~~~
.
-
_. . . . . . . . . _ _ _ _ . . . . .. . ........'..~...... . . - . . . ..
-
.w' --
.
.
. ,
,
>
4. PROCEDURES - NORHAL, ABNORMAL, EHERGENCY AND PAGE 2
~~~~R3656E655CEE~C6N5R6L'~~~~~~~~~~~~~~~~~~~~~~~ i
____________________ .
.
OllEC110N 4.01 (2.00)
Concernins E0P-2 'RPV CONTROL, POWER *:
a. What are the entry conditions for E0P-27 (1.00)
b. What methods, other than insertion of control rods, are availabic to
reduce reactor power? (1.00)
GUESTION 4 02 (2 00) i
!
a. According to PNPS Procedure 2.1.9 ' Reactor Recirculation Pump ,
Operation', what temperature requirements must be met to start an *
idle recirculation pump? (1.00) l
b. List two adverse effects of starting a recirculation pump
without having these requirements met. (1.00)
OUESTION 4.03 (3 00)
Concerning PNPS Procedurc 2.4.40 ' Rapid Increase In Orfgas Activity'! I
a. What automatic actions would occur during a high offsas conditions
and when would those actions be initiated? (1.00) '
b. Assuming these automatic actions occurede what immediato
operator actions are required in accordance with the proceduro? (2.00)
i
!
DUESTION 4.04 (2.00) i
i
a. According to procedure 2.4.1.2 ' Stuck or Inoperabic Control Rods', :
what symptoms indicate a stuck or inoperable control rod? (1.00) i
b. What inacdiate actions must be taken for a suspected stock or
* !
inoperable control rod? (1 00)
i
lillFCIION 4 05 (2.50) I
o. Other than alarms, what four symptoms are indicative or a small Icak
(drywell pressure does not cucced 2 psig.) within primary
containment? (1 00) t
b. When would plant shutdoun be initiated for inspection and i
corrective action based on leakage detection system indication? (1.50) l
r
i
I
I
.
!
,
;
l
i
{
l
_ _ _ _ _ _ _ _ _ _ _ _ _ _
. _ . . .. . . . . . - - - . . . . - . . . . . . _ _ . . . . . - . . . . . . _ . ..__ - . . . . . . - _ _ .:.,_ w
.
.
4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 3
~
~~~~R5656L66265L CUUTRUE~~~~~~~~~~~~~~~~~~~~~~~~
-------------------- -
,
OUESTION 4.06 ( 2. !io )
Ucing the attached ' Heat Capacity Level Limit' curve from E0P-06
' Primary Containment Lovel Control' answer the following!
o. What is the minimum water 1cvel for RPV pressure of 406 psis and
pool temperature of 160 do3rees F? (0.5)
b. What is the maximum RPV pressure with 12 feet or water at 154
degrees F? (0.5)
c. If suppresion pool water 1cvel COULD NOT be maintained above the heat
car ty lovel limit, what 2 actions must be takon, and why? (1.5)
HurG1 ION 4.07 (2.00)
a. In accordance with proceduro 2.1.1 'Startup From Cold Shutdown,'
under what conditions shall secondary containment integrity be
maintained? (1.5)
b. You are allowed a cortain period of timo after placing the reactor
in the RUN modo before the primary containment atmosphere nuygen
is required to be less than 5% by weight. How long ts this period? (0.5)
HHECTION 4.08 (3.00)
.
According to procedure 2.2.96 ' Condensate and Feewater System,' during
startup of the foodwater pumps:
a. What is the required status of the condensato pumps and controls? (1 00)
b. What precludes startup of the R.F.P. without lubrication? (1 00)
c. How is the oil temperature maintained at 110 degrecs? (0.5)
.
d. When should the second feed pump be placed in service? (0.5)
4 HEC 1 ION 4.09 (3.00)
What access control requirements exist for the following high radiation
creas?
a. An intensity greater than 100 mrem /hr but 1 css than 1000 mrem /hr (1.00)
b. An intensity greater than 1000 mrom/hr but loss than 10000 mrom/hr(1.00)
c. An intensity greater than 10000 mrom/hr. (1.00)
.
-
k
_ _ _ . __
_ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ , .
- . - - . . . . - . - - . . - . . ..;._ _ __.
.
.
.
4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 4
~~~~Rd656E6556dE"66HTRUE~~~~~~~~~~~~~~~~~~~~~~~~
____________________ .
.
QUESTION 4 10 (1.00)
According to procedure 1.3.4, procedures do not have to be present for
frequently ropcated procedural actions. What two types of procedures
are except from this and must be present and followed step by stop while
teoks are beins performod? (1.00)
OtlE'G110N 4.11 (2.00)
If both stack dilution fans are lost, the associated Of f -Normal
proceduto requires that both standby gas treatment units be placed
1
in service. Why is this necessary? (2.00)
l
.
9
- _ _ - .- - - - . ,__ .- ., -
,- -
.- -- -
- - - - - - -
---- - --- -
..,
-
. . . .. . -.. .. .. . _ .
..
. .. _. . _ . ._
.
.
.
-
l -
es
03
, 2
Q
-
\
. iq
U., ,. -
z. w =-
.. , ,e = <5
.-
$i}
'
s, '
,
'
..
'.
-
i
-z
'E
.. =i
te ;a y
's ._
vo
Uw
.-
** - =z- a
z
..
.
//,. ,., / - -
.
..
:
'
, '.. : 2-
1 ,
c.
2
-
.i-- s
--
;,
--
2 e e r e eer ?
. e a t
,-
t
e
1
e
aat
..
=, _ . - .,
- -
=s
~ - -
t o g
: . , = = a=3e -
a
-
2 z z
-
z z zzz z 8
- z & & &
-
d & & & & & & ,
"
.
/ / / /
s
-
'
//' . /
'
-
-
~ . ,. =
. ;
.
" l *
\ l a 6 y
: l '. I
w
I
i
:
u .-
.
"
.,
2 .
/ l :
< : >
w ;
! u
*
.-
'
-
.
. a
g
-. : : -
=
,
3 '
e
E * .// . O r
'. *
. ,. . ,s - a
.
;3
w
=
o
.
. ,.
/ / ,. * * j E5 == O O 41
- ' f
e w
,.f- ..
, - z
-
* L
s' #p
/' , h
*
. .. / .* -
.*
g,4ll/ -
: .
.
'
A 3
M' c n
'
-'
,::sa / 4 w .i- m',cb
//
.. ..
h' ..
/. f. /
/
'N- ~
.
t:
- -
g :
. I
2 3~ s. *. a. s.
*
,
(4.) 3mnivw3dn11 loo,, NolS$3Wddn1
t ' s
_. . . _._ __.... .- . . . . . . _ - - . , _ . _ _ _ . _ - . . . . . . . :_: ..L. .. _ _ _ .i.
'
. .
O. PROCEDURES - NORMAL, ADNORMAL, ENERGENCY AND PAGF 5
c6N7R5E------------------------
-
~~~~R 5656L55iEAL
____________________ .
ANSWERS -- PILGRIM , -85/05/14-CRESCENZO, F.
ANSWER 4.01 (2.00)
e. A condition exists which requires reactor scram (0 25), AND reactor
i power is abovo 3% (0.25), OR power cannot be determined (0.25), OR
all control rods are not inserted past postion 04. (0.25)
b. Trip reactor recire pumps (0.33), Doron injection (0.33),
Level control E0P-8 (0.33).
REFERENCE
PNPS Procedure No. E0P-02, Rev. O
ANSWER 4.02 (2.00)
a. Do not start an idic pump if delta temp. betwoon loops is crest.cr
than or equal to 50 dos. F (0.5) or if delta temp botucon the vossol
dome and the bottom head drain is 3rcater than 145 dos. F. (0.5)
b. A cold water injection can occur which can cause unusually high
reactivity and a reactor scram.(0.5) Inadequate core flow and high
CRD flow can result in cold water stratification in the vessel.
Starting an idic pump from this condition can cause damase to the
CRD stub tubos from cold water shock.(0.5)
REFERENCE
PNPS Procedure No.2.1.9 'Roactor Recirculation Pump Operation' Rev. 14
ANSWER 4.03 (3.00)
o. Off-Gas High High Radiation Alarm - Initiation of off sas isolation
timer. If the high activity condition still entsts at conclusion of
the 15 minute timer, the holdup line outlet valvo (A03751) and the
holdup lino drain valvo (A03750) will trip closed. (1.00)
b. Manually scram reactor (0.50), Close the MCIV's(0.50), R e f'e r to
the procedure for 'Closuro of MSIV's on High Radiation'(0.50),
Perform rapid survey (0.50).
REFERENCE
PNPD Procedure 2.4.40, Rev. 5
.
* f
.
.~ _ - - - . _ _ - .
-
_ - _ , - - - _ - - - . . . _ , . _ _ .. .,
_ . - - - - _ _ - - - - - - _ - . _ _ _ - - - - - - - - - - - - - -
.
.. ... . . . _ . . . . . _ . . _ _ _ - . _ _ . . . _ . . . _ . . . . . _ _ . _ . . . . _ - . .
.
.
..
4. PROCEDURES - NORMAL, ABHORMAL, EHERGENCY AND PAGE 6
~~~~~~~~~~~~~~~~~~~~~~~~
~~~~R565UL55555E~C N5RUL
____________________ .
ANSWERS -- PILCRIH , -85/05/14-CRESCENZO, F.
ANSWER 4.04 (2.00)
o. Inability to insert or withdraw a control rod partially or fu11y.(0.33)
Slow CRD movement which cannot be adjusted with the CRD module
speed control valves.(0.33) Rod blocks which are the result of power
failures in the RHCS.(0.33)
b. Check power, pressure, and level to assure that the reactor status
is within the enpceted limits.(0.33) Check compliance with technical
specifications.(0.33) Determine if an emergency action IcVel has
been exccoded. (0.33)
REFERENCE
PNPS proceduro 2.4.1 Stuck or Inoperabic Control Rods Rev.4
ANSWER 4.05 (2.50)
c. Execssive sump pump operation due to increased leakase to drywell
equipment drain sump or drywell floor drain sump.(0.25) Abrupt
changes in drywell humidity as indicated on panel C05 recorders.(0.25)
High radiation detected on drywell leak detcetion systone panel
C19 (0 25) Significant drywell pressures changes indicated on
drywell pressure recorders.(0.25)
b. When any leakage detcetion system indicatos,within a 24 hour period
or loss, an increase in rate of unidentified Icakage in onoss of ?
gallons per minute or its equivalent, or when the total unidentified
Icakage attains a rate of 5 sallons por minute or its equivalent
whichever occurs first.
REFERENCE -
PNPS procedure 2.4 14 Leaks Inside The Primary Containment Rev. 5
ANSWER 4.06 (2.50)
o. 11.0 ft. (0 5)
b. 800 psi 3 (0.5)
e. Manually scram reactor,(0.5) depressuri=c RPV using ADS system.(0.5)
this is to ensure the suppresion pool could contain energy released
during a LOCA (0.5)
REFERENCE *
PNPS procedure E0P-06 ' Primary Containment Level Control' Rev. 0
.
. . . . _ _ . . . . . .. _ . . . ....'. . . . . n 2. . .. . u . ... . : L
.
. -
4. PROCEDURES - NORMAL, ABNORMAL, EMERCENCY AND PAGE 7
~~~~ RdDE6EUU56AL"C6HTR6E~~~~~~~~~~~~~~~~~~~~~~~~
____________________ .
ANSWERS -- PILGRIM , -85/05/14-CRESCENZO, F. '
ANSWER 4.07 (2.00)
o. Whencvor the reactor is critical or when the Rx water
tamperature is above 212 do3rees and the head vont closed. (1.5)
b. 74 hours. (0.5)
REFERENCE
Proceduro 2.1.1 Rev. 40 pg. 4
.
ANSWER 4.08 (3.00)
8. One condensato pump is required for cach feed pump in
cervice, and the sequential trip selector switch to ON unless all
condensato and RFP's are in service.(1.00)
b. The same C/S is used to start the lubo oil pump and the R.F.P.
At the first operation of the C/S, the lobe oil pump starts. It
sprin3 returns to n- .al, and at the second operation, the RFP
is started. (1.00)
c. By throttling the TBCCW oil cooler outlet valve (0.5)
d. Before the first foodpump reachos 90% pump capacity.(or prior
1.o execeding 30% reactor power) (0.5)
REFERENCE
RETERENCE! Procedurc 2.2.96 Rev. 15 pp. 9, 10
ANSWER 4.09 (3.00)
o. The area shall be baricaded and conspicuosly posted as a High
Radiation Area and entranco thereto shall be controlled by issuance
or a RWP. (1.00)
b. In addition to those requirements stated in part 'a', locked doors
shall be provided to provent unauthoriced entry into such areas.
The keys shall be under the administrative control of the Chict
Radiolo31 cal Ensincer (1.00)
c. An additional lock, keys will be maintained with the watch engineer,
HP rop. present during ontry to monitor dose rates, entry requires
authorication from watch onsincer or station manager. (1 00)
REFERENCE
Procedure 6.1-012 Rev. 12
.
.
g o
- _ . . _ _ _ ____. _ _ _ __ ___ _ ___
~
., ,
-
. . . . . ,. - _ _ _ _ _ _ _ _ _ _ _ . ______.._.__...;..=___...,;.._.,;_
.
*
,1 -s
. .
4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 8
-------------~~---------
~~~~R A5iBL55iEAE 55siE5L
____________________ .
ANSWERS -- PILCRIM , -85/05/14-CRESCENZO, F.
-
ANSWER 4.10 (1.00)
Operations Checklists designated as OPER forms for plant startup,
chutdowns, and power changes as contained in Volume II, sect. 2 of
the PNPS Operations Manual. (0.5) Surveillance Test Procedures as
contained in Vol. VIII of the PNPS Operations Manual and designated
as either the 8.m or 8 scrics. (0.5)
REFERENCE
Procedure 1.3.4 Rov. 26
ANSWER 4.11 (2.00)
Both standby sas treatment units are placed in service at rated flow
to provide dilution air to reduce the hydrogen concentration in the
stack and maintain suitabic exhaust velocitics at the top of the stack.
(2.00)
REFERENCE
Procedurc 2.4.45 Rev. 5, Ps. 2
<
.
_
--
. .
-
,
. . . - . . . . . ~ . . . . .. . . . . .. . . . . . . . .... _ . . . . _ . . . . _ ,;. ....._;._____.._.
..
...
TEST CROSS REFERENCE PAGE 1
QUESTION VALUE REFERENCE *
------- ------ ----------
'
04.01 2.00 FJC0000001
04.02 2.00- FJC0000002
04.03 3.00 FJC0000003
04.04 2.00 FJC0000005
04.05 2.50 FJC0000004
04.06 2.50 FJC0000007
04.07 2.00 FJC0000000
04.08 3.00 FJC000001b
04.09 3.00 FJC00000tt'
04.10 1.00 FJC0000012
04.11- 2.00 FJC0000014
_-_---
25.00
------
------
75.00
.
-
1
- - _ - _ _ _ _ .
s
}}