ML19347B607
| ML19347B607 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 10/03/1980 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML19347B604 | List: |
| References | |
| NUDOCS 8010150436 | |
| Download: ML19347B607 (28) | |
Text
.. -. -. _ - _
~
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t?
Ha.
f G.-
Quadrant Power Tilt K.
Operable i EU The quadrant power tilt per unit is Properly installed in the system and CD' defined as the ratio of the maximum upper capable of performing the intended
]jf excore detector current to the functions in the intended manner as qp average of the upper. excore detector verified by testing and tested at currents or the ratio of the maximum the frequency required by the lower excore detector current to the Technical Specifications.
average. of the lower excore detec tor 3
currents whichever is greater.
If L.
Operating one excore detector is out of service, the three in-service Performing the intended functions units are used in computing the in the intended manner.
a vera ge,
f t.
Operating Cycle H.
Rated Thermal Power The interval between the end of one A steady-state reactor core output of major refueling outage and the end of 4
]
3250 MWt Per unit.
the next subsequent major refueling j
outage per unit.
I.
Reactor Pressure N.
Surveillance Interval The pressure in the steam space of a pressurizer.
Each Surveillance Requirement shall be performed within the specified time interval J.
Refueling Outage with:
}
When Refueling Outage is used to a.
A maximum allowable extension not designate a surveillance interval per uniti to exceed 25% of the surveillance the surveillance will be performed interval; and during the refueling outage or up i
to six months before the refueling b.
A total maximum combined interval j
ou ta ge.
.When a refueling outage time for any three consecutive i
occurs within 8 months of the surveillance intervals not to l
previous refueling outage for a unit, the exceed 3.25 times the specified surveillance testing need not be interval.
performed.
The maximum interval i
between surveillance tests is j
20 months per unit.
I 1
1 i
~
6 l
1 4
SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 7.
Low reactor coolant pump motor frequency:
2 57.0 cps.
8.
Undervoltage to reactor coolant pump motors:
% 70% of normal.
C.
Other reactor trips I
1.
High pressurizer water level:
da 92% of span.
1 2.
Low-low steam generator water level:
2L 10% of narrow range instrument.
3.
Steam feedwater flow mismatch:
5 60% of nominal 100% steam flow rate in coincidence 4
l with low steam generator water level -
2L10% of narrow range instrument span.
4.
Safety Injection - Trip cettings for safety
~
injection are detailed in Section 3.4.
5.
6.
Power range, positive high neutron flux i
rate SE 15% of rated flux in 5 sec.
Powerrange,lnegativehighneutronrate S l-15% of the 7.
flux j
8.
Manual reactor trip.
]
i l
i i
i.
I 12 4
O i
Above 10% power, an automatic reactor trip The Manual Reactor Trip is a redundant channel will. occur if two reactor coolant pumps are to the automatic protective instrumentation
. lost during operation.
Above 60% power, an channels and provides manual reactor trip automatic reactor trip will occur if any capability.
pump is lost or deenergized.
(9)
This latter trip will prevent the minimum value of the DNBR from going below 1.30 during normal operational transients and antici-pated' transients when only three loops are in operation and the overtemperature 6T i
trip setpoint is adjusted to the value specified for four loop operation.
When
]
the overtemperature 6 T trip setpoint is adjusted to the value specified for three loop operation and the loop stop valve is closed in the inactive loop, the-trip at 75% power will prevent the minimum value of the DNBR from going below 1.30 during normal operational transients and anticipated transients when only three 4
loops are in operation.
l 4
.(9)
FSAR Section 14.1.6 i
.I i
i 4
i l
i l
l e
i 24 l
1 l
l 1.
2.
3.
4.
5.
6, tio. Of Minimum Minimum Operator Action Reactor Trip Channel No. Of Channels Operable liegrre Of If Column 3 or 4 Description Per (Jnit Channels To Trit Channelstet Redundancyt+t Can Not Ile Met +
SetpoinL+t
- 1. Manual Peactor Trip 2
1 1
0 Maintain llot Shutdown **
N.A.
2.
Power Hange liigh Flux 25% of Rated Neutro n (Iow set pointi-interlocked with P-10 4
2 3
2 fla a ntain llot shutdown *
- Flux 1094 of Ita ted 3.
Power Range liigh Flux (high set point) 4 2
3 2
Maintain flot Shutdown *
- Neut on Flux 4.
Power Range liiyh Flux Rate 4
2 3
2 Maintain llot Shutdown ** 15% of Hated Neutron Flux /5 sec.
15% of '4ated Neutron
- 5. tiegative Power Hanqe Plux Rate 4
2 3
2 Maintain llot shutdown ** Plux/5 uc.
O 6.
Source Range Neutron Flux-5 Interlocked with P-10 and P-6 2
1 1
0 Haintain slot Shutdown ***
10 counts /see (or CSD if that condition exists) 251 of Rated Neutron 7.
I ntermediat e Range Neutron Flux-Interlocked with P-10 2
1 1
0 Maintain llot Shutdown
- Flux H.
Overtemperature o T, 4 loops 4
2 3
2 Maintain flot Shut down*
- Actuala Th Prograuec'd overt <smperature a T, 3 loops 3
2 3
1 Flaint.ain llot Shutdowr *
- Setpoint 9.
Overpower 6 T, 4 loops 4
2 3
2 Maintain llot Shutdown ** Actual A T 2 Programmed Overpower o T, 3 loops 3
2 3
1 Maintain llot Shutdown ** Setpoint
)
10 Pressuri_er low Pressure -
interlocked with P-7 4
2 3
2 Maintain llot Shutdown ** 1825 psig 11 Pressurizer liigh Pressur e 4
2 3
2 Maintain llut Shutdown ** 2385 psig TABl.E 3.1-1 Reactor Protectaan System - 1.imitinq Operation. Conditions and Setpoints 30 i
l
I.
2.
3.
4.
5.
6.
13 0. Of Minimum Minimum Operator Action Peactor Trip Channel tio. Of Channels Operable Degree Gf If Column 3 or4 Description Per Unit Channels To Trip _
Channelst++
Hedundancytte Can flot P.c flet *
@]tPolfil '+
- 12. Pressurizer fligh Level -
interlocked with P-7 3
2 2
1 Maintain flot Shutdown ** 921 of Span
- 13. Low Primary Coolant Flow -
interlocked with a.
P-7 3 per loop 2 per loop 2 per lonp I
tiaintain tio t Shut down *
- 90% of nominal in 2 loops span b.
P-8 3 per loop 2 per loop 2 por loop 1
Maintain liot Shutdown **
in I loop
- 14. RCP Dus Undervoltage -
interlocked with P-7 1 per bus 2
3 2
Maintain Ilot Shutdown ** 75% of nominal voitnse
- 15. RCP 11us Underfre<guency -
interlocked with P-7 and P-8 1 per bus 2
3 2
flaintain llot shutdown **
57.0 11 2
- 16. RCP Breaker Trip -
interlocked with:
a.
P-7, 4 loops 4 breakers 2
1 2
71aintain Ilot Shutdown *
- N.A.
P-7, 3 loops 3 breakers 1
3 3
Maintain Ilot shutdown **
II, A.
b.
P-8, 4 loops 4 breakers 1
3 2
Maintain llot Shutdown **
H.A.
P-8, 3 loops 3 breakers 0
flA T3A Maintain flot Shutdown **
ti, A.
17 Low Steam Generator Level in 2 per loop 1 per loop i
O Maintain Ilot Shotdown**
25% of narrow coincidence with feed flow-level ranqe p' pan steam flow mismatch 2 par loop 1 per loop 1
0 Maintain tiot Shutdown ** and 0.7 x 10 lbs/hr mirmatcli
- 18. Low-inw Steam Generator 3 pe r loop 2 per loop 2 per loop 1
Maintain Iot Shu t<5wn *
- 10% of narrow Level - interlocked with loop level ranqe span isolation valve position
- 19. Safety injection 2
1 2
1 Maintain Itot Shutdown ** Any safety in-ject actua-tion 20 Turbino Trip-interlocked with P-7 3
2 2
1 Maintain ikit Shutdown *
- tl. A.
- 21. Automatic peactor Trip I.ogic 2
1 2
1 Plaintain tiot Shutdown *
- ra. A.
TABLf: 3.1-1 (Continued)
Reactor Protection System - Limiting operating Conditions and Setpoints 3,
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Reactor Trip Channel Cnannel Channel Channel Description Check Calibration Functional Test Remarks
- 1. ' Manual S/U (1)
(1) If not done in previous week.
2.
Power Range Neutron Flux S
\\
M (1) Heat Balance Calibration (Low Setpoint) l 3.
Power Range Neutron Flux S
D(1)
M (2) Recalibrate 4.
' Power Range Positive Flux l
Q(2)
(3) Compare incore to excore I
M axial imbalance.
Recali-Rate N.A.
EFPM(3) k brate as per specificati n 5.
Power Range Negative Flux k
3.2.2.C.1ifl difference
)I
> l%.
Rate N.A.
M 6.
Source Range Neutron Flux S(l)
N.A.
S/U(2)
(1) Once/ Shift when in service.
4 7.
Intermerliate Range Neutron (2) Not required if performed Flux S (1)
N.A.
S/U(2) within the previous 7 days.
8.
Overtemperature 6T S
R M
9.
Overpower ti T S
R M
10.
Pressurizer Low Pressure S
R M
ll.
Pressurizer High Pressure S
R M
12.
Pressurizer High Level S
R M
13.
Low Primary Coolant Flow S
R M
i 14.
RCP Bus Undervoltage N.A.
R R
15.
RCP Bus 'Inderfrequency N.A.
R R
16.
RCP Breaker Trip N.A.
N.A.
R i
TABLE 4.1-1
}
Reactor Protection System Testing and Calibration Requirements
- I 35 i
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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIRD(ENT 3.2.2 B.
Quadrant Power Tilt Limits 4.2.2' B.
Quadrant Power Tilt Limits 1.
If an indicated quadrant power tilt.
1.
Quadrant power tilt ratio shall be ratio exceeds 1.02, except for calculated and logged along with physics tests, then within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the individual upper'and lower one of the following steps shall be excore calibrated outputs as taken:
c follows:
a.
Correct the tilt, or a.
Once each shift at power levels greater than 50%.
b.
Determine by measurement the b.
Four times a' shift and following core peaking factors and apply a load change of more than 10%
Specification 3.2.2.A, or power at any power level above
~
i 50% if one or both quadrant c;
Restrict core power level so power tilt alarms are inoperable.
ms not to exce,ed full rating I
less 3% for each percent of quadrant power tilt ratio beyond 1.0.
2.
If an indicated quadrant power 2.
Not Applicable tilt ratio exceeds 1.02 for a t
period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without known cause, or if sudden tilt reoccurs intermittently without known cause, the reactor shall be put in the llot Shutdown Condition within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
However, operation below 50% of rated powert for testing and/or correcting the tilt, shall be permitted.
4 48
U
~
~
O O
LIMITINd CONDITION FOR OPERATION QRVEILLANCE REQUIREMENT 1
3.2 3.
Control P.od System operability (per unit)
M.2 3.
control 'cv! " 7 ten onerability (por uni t)
A.
Rod Ttisalignment I,initations A.
Pnd 'tisalionnen t T,ini ta tions 1.
If a full-length control rod is 1.
A rod nalposi tion chock sla ll 1 e out of alionnent with its bank by nade once a nhift usinn both the more than Il2 stens indicated; analon and dioital disnlavs.
then, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, one of the follotring steps shall be taken:
a.
Realion the rod, or b.
notermine by reasurement tha D
b
@g ( gQu i
core peaking factors and D
qi
~
i g
apply Soecification 3.2.2.A, or c.
nostrict power level to 80 of ra ted power.
~
l 2
If the nisalioned control rod in 2
Not anplicable l
not realigned within 0 hcurs, the
~
i rod shall be c'eclared inoperable i
i and the limi ta tions of 3. 2. 3. n apply.
3.
The orovisions of specifications 3.
' Int Applicabla 4
3,2.3.A and 3.2.3.D shall not i
apoly during physics tests in which the contr61 rods are in-tentionally nisaligned.
1 D.
Inoperable Rod I,initations P.
Tnocerable Dod T,initations 1.
An inoperable control rod is a 3.
' Jot ^.policable rod which cannot be rnved by its nochanism or which is c'eclared inoperable by Speci fication 3.2.3.A or 3.2.3.c.
4 1
e j
51
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT i
l 323 23 l
C.
Rod Drop Time C.
Rod Drop Time The individual full length (shutdown Under the conditions of LCO 3 2.3.c.1 and l
and control) rod drop time from the 2.2 3.C.2 the hot rod drop time of full fully withdrawn position shall be length rods shall be demonstrated through
$ 1.8 seconds from beginning of decay measurement prior to reactor criticality:
of stationary gripper coil voltage to dashpot entry with:
- 1. For all rods following each removal of the reactor vessel head,
- 1. T"V6 >~ 530 F, and
- 2. For specifically affected individual
- 2. All reactor coolant pumps operating.
rods following any maintenance on or modification to the control rod drive APPLICABILITY:
Applies to all oper-system which would affect the drop ation with the reactor crJ ' '. cal.
time of those specific rods, and ACTION:
With the drop time of any full
- 3. At least once per refueling outage length rod determined by Surveillance (not to exceed 20 months).
Requirement 4.2 3, to exceed the abose limit restore the rod drop time to within the above limit or declare the rod inoperable per Section 3 2 3A prior to proceeding to criticality.
D.
Inoperable Rod Position Indicator D.
Inoperable Rod Position Indicator Channels Channels
- 1. Not more than one rod position 1.
If a rod position indicator is out indicator channel per control rod of service, then:
group nor two rod position indica-to:' channels per control rod bank
- a. For operation between 50% and shall be permitted to be inoperable 100% of rated power, the at any time, except during hot rod position of the control rod drop timing measurements.
54
LIMITINd CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2.3.D.1 2.
I f the condit.innn of Section 4.2.3.n.l.a shall I,e chart:cd indirectly 3.2.3.D.1 cannot be met the 1 7 ercore detectern and/or reactor shall be broucht to thorrocounlos and/or rnve -
i at least the Ilot Shutdown abic incore detectnrn overy condition within four hours shift: or after.'nv red and the reactor trip break-rotion of the non-indicatinn i
e rn shall rerain open.
rod, erceedinc 12 nteps, whichever occurn first.
b.
nurinn onoration bcInu 50?
of ratori pover, nn snocial 4.
D'in "a rane ters rannitorina in renui red.
i A.
The following DI!!T related para-4.A.l.
Each of the ba raretern linted in me ters shall be maintained wi thin 99ecificatinn 3. . 1. A sball be the limits shown during operatien.
verified to be uithin itn lirit at least once per 12 hnurn.
1.
Reactor Coolan t Synten Tavn POUR Loop:
656f.3*P 2.
The 7enctor Conlan t Syntem total
~
TilREE LOOP:
il flow rato shall be determined to
~
2.
Pressurizer Pressure be within its lirit by rea su renen t FOUR LOOP: g 2220 psia
.at least once por Il rnnthn.
(2205 psio)*
Tl!REE LOOP:
i' l
3.
Reactor Coolant Systen Total Flow Rate POUR LOOP: )350,000Gr*1 TIIREE LOOP:
1-II.
With any of the above pa rano te rn exceeding its limit, restore the parameter to within its limit with-D**]D
]O [l3 }
in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce therral power g
gyJ ]
to less than 5% of rated thornal power within the no::t 4 hourn.
4
- Limit not applicable during either a therra t oower rarp increase in orcess or 5" rated thornal l
power per minute or a thernal pnwer step increase in orcons or 10% rated thernal noser.
f Paraneter limi ts for three loop operation to be entabliqhed prior to onoratinn above D-7 with lens than four loops operating, i
I
When an F measurement is taken, both experimental For normal operation it is not necessary to g
error and manufacturing tolerance must be allowed measure these quantities continuously.
Instead for.
Five percent is the appropriate allowance it has been determined that, provided certain for a full core map taken with the movable incore conditions are observed, the hot channel factor detector flux mapping system and three percent is limits will be met:
these conditions are as the-appropriate allowance for manufacturing follows:
tolerance.
1.
Control rods in a single bank move together InthespecifiedlimitofFNH there is an 8 with no individual rod insertion differing (1) by more than 15 inches from the bank demand percent allowance for uncertainties which means position.
An indicated misalignment limit that normal operation of the core is expected to of + 12 steps, not including instrument result in FNg n 1.55/1.08.
The logic behind the error. precludes a rod misalignment no larger uncertainty in this case is that (a) greater than 15 inches.
With maximum abnormal porturbations in the radial power shape instrumentation error considered the actual (e.g. rod misalignment) affectFyg, in most cases rod misalignment is no more than 24 steps without necessarily affecting Fg, (b) the operator or 15 inches.
has a direct influence on F through movement of rods, andcanlimitittotReaesiredvalue, he 2.
control rod banks are sequences with over-has no direct control over F{n and (c) an error lapping banks as described in Technical in the predictions for radial power shape, which Specification 3.2.
may be detected during startup physics tests can by tighter axial control, 3.
The full length control bank insertion be compensated for in F but compensation forF$gn is less readily available, limits are not violated.
errormustbeallowedfo$g When a measurement of F
is taken, experimental r and 4 percent is the 4.
Axial power distribution control procedures, appropriate allowance for a full core map taken which are given in terms of flux differences with the movable incore detector flux mapping control or additional axial power monitoring
- system, and control bank insertion limits are observed.
Flux difference refers to the Measurements of the hot channel factors are required difference in signals between the top and as part of start-up physics tests and whenever bottom halves of two-scetion excore neutron abnormal power distribution conditions require a detectors.
The flux difference is a measure reduction of core power to a level based on measured of the axial of fset which is defined as the hot channel f act ors.
The incore map taken following difference in normalized power between the initial loading provides confirmation of the basic top and bottom halves of the core.
nuclear design bases including proper loading patterns.
The periodic monthly incore mapping ThepermittedrelaxationinF$gallowsradial provides additional assurance that the nuclear power shape changes with rod insertion limits.
design bases remain inviolate and identify opera-It has been determined that provided the tional anomalies which would, otherwise, affect above conditions 1 through 4 are observed, these bases.
these hot channel factor limits are met.
In Specifications 3.2.2, F
is arbitrarily g
limited for P $ 0.5.
68
+
The nuclear ion chambers located outside a In the event that an LVDT is not in service, reactor vessel measure the flux distribution the ef fects of a malpositioned control rod are of the top and bottom halves of a core.
Core observable on nuclear and process information traverses in a few of the in-core instrument displayed in the control room and by core thermo-thimbles will establish that the excore flux couples and in-core movable detectors.
measurement equipment is properly calibrated.
Operating experience has established that the One inoperable control rod per unit is acceptable excore flux measurement system is of a reliable provided that the power distribution limits are
," design, and that the 10% load reduction, in the met, trip shutdown capability is available, and event of*L'1GCJIfdrLtion delay, is an ultra provided the potential hypothetical ejection of conservative compensation.
the inoperable rod is not worse than the case i
analyzed in the safety analysis report.
The Operating experience at similar PWR plants has rod ejection accident for an isolated fully shown that quadrant power tilts determined by inserted rod will be worse if the residence time 1
monitoring symmetric thermocouples are in very of the rod is long enough to cause significant I
good agreement with quadrant power tilts deter-non-uniform fuel depletion.
The 3 day period mined from power distribution maps using the allowed for the analysis is short compared with Movable Detector System.
the time interval required to achieve a signifi-cant non-uniform fuel depletion.
Operation of one reactor cavity vent fan per unit ensures an adequate flow rate of cooling The rod drop time of 1.8 seconds is based on the 3
air to each NIS Detector (9).
negative reactivity insertion rate used in accident analysis.
(11) i The various control rod assemblies (shutdown banks, control banks A, B, C, D and part-length (1)
FSAR - Figure 3.2.1-8 rods) are each to be moved as a bank, that is, (2)
FSAR - Table 3.2.1-1 with all assemblies in the bank within one step (3)
PSAR - Figure 3.2.1-11 1
(5/8 inch) of the bank position.
Position (4)
PSAR - Chapter 14 indication is provided by two methods:
a digital (5)
PSAR - Secti on 3.1. 2 4
count of actuation pulses which shows the demand (6)
FSAR - Section 3.1.3 position of the banks and a linear position in-(7)
FSAR c.. apter 14, Appendix C i
l dicator (LVDT) which indicates the actual rod (8)
FSAR - Question 3.8
}
position (10).
The rod position indicator (9)
FSAR - Section 9.10.6 i
channel is sufficiently accurate to detect a (10)
FSAR - Section 7.3 misaligned rod 15 inches away froni the demand (11)
FSAR - Figure 14-2 l
position of the bank.
The indicated + 12 step (12)
August 27, 1976 Order for Modification of permissible misalignment provides an enforceable License.
limit below which design distribution is not exceeded.
]
I 72 i
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O O
LIMITINd CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.1.n 4.3.1.F.5 Reports l
A.
Followinn each innervice innocc-tion of stean nonerator tuben, the number of tubes pluened in Q
each stean qenerator shall be
[ g 2' D
cg en en av renorted to the Conmission with-in 15 days.
,av av es n.
The complete results of the steam canerator tube innervice in snec tion shall be included in the 9pecial 9e-oort pursuant to Specification 6.6.3.(c).
The Special Report shall be nubnitted on an annual basin for the perioct in which the innnection was completed.
Thin renort shall include:
1.
! lumber and exten t of tuben in-
,spected.
2.
Location and nercent of wall thickness penetration for each indication of an inperfection.
3.
Identification of tubes plunned.
C.
Renults of stean cenera tor tube in-npections which fall into catencry C-3 and recuire oronnt noti fication of the Conmission shall be reported oursuant to Specifica tion 6.6.2 orior tn renunption of plant nparation.
The written follow up of this rennrt shall provide a descrintion of inventination conducted to doternine the cm:na of the tubo d?oradation and corrective measures taken to prevent re c'i rren ce.
74n
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4 4
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7 4
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LIMITING ' CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 332 Pressurization and System Integrity 4.3 2 Pressurization and Sys tem Integrity A.
Heatup and Cooldown A.
Not Applicable.
The Reactor Coolant System temperature and pressure (with the exception of the pressurizer) shall be limited in accordance with the limit lines shown in Figures 3 3 2-1 and 3.3 2.2 during heatup, cooldown and inservice leak and hydrostatic testing with:
0
- 1. a. A maximum heatup of 100 F in any one hour period.
O
- b. A maximum cooldown of LOO F in any one hour period.
- c. A maximum temperature change of
< 100F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
i 2.
Figures 3.3 2-1 and 3 3.2-2 define limits to assure prevention of non-ductile failure only.
For normal operation other inherent plant characteristics, e.g.,
pump heat addition and pressurizer heater capacity may limit the heatup and cooldown rates that
)
can be achieved over certain pressure-temperature ranges.
- 3. Allowable combinations of pressure and temperature for specified tem-totherightofthelimitlinesshown\\
perature change rates are below and Limit lines for cooldown rates be-tween those presented may be ob-tained by interpolation.
79
LIMITING CONDITION FOR OPERATION SUP.VEIILANCE REQUIREMENT 3.3.2 (Continued).
4.3.2 B.
The limit lines shown in Figures B.
Not Applicable 3.3.2-1 and 3.3.2-2 shall be re-calculated periodically as required, based on results from the material surveillance program.
C.
The secondary side of the steam C.
Not Applicable generator must not be pressurized above 200 psig if the temperature of the pri-mary and secondary coolant is below 700P.
D.
The pressurizer heatup rate shall not D.
Not Applicable exceed'100*F/hr and the pressurizer cooldbwn rato not exceed 200*F/hr.
The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320"P.
E.
Hydrostatic Testing
- E.
Not App-licable 1.
System inservice leak and hydro-tests shall be performed in accordance with the requirements of ASME Boiler and Pressure Vessel Code",Section XI, 1974 Edition, up to and including Summer 1975 Addendum.
80 e
I b
LIMIT 7NG CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.2.
F.
Safety Injection Actuation 4.3.2 F.
Not Applicable
~
1.
If safety injection should occur when a reactor is in the hot shut-down condition or above, the reactor shall remain in the hot shutdown condition until the status of the reactor coolant system integrity is der rmined.
2.
If the inspection and review (Sec.
6.1.G.2.a
and Sec. 6.3) of,the rea". or toolant sys tem integrity determines that:
a.
The injection did not affect reactor coolant system integ-rity the plant may proceed to power operation.
b.
The injection did affect reac' tor coolant system integrity, the reactor shall be placed in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.
In.the event the ECCS is actuated and injects water into the Reactor Coolant System when Tavg 2 350* F,
a Special Report ~shall be prepared and submitted to the Commi'ssion within 90 days describing the circumetances of the actuation and the total a'ccuntiated actuation cycles to date.
G 81
{
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMEN P 3.3.4 4.3.4 D.
Materials Irradiation Surveillance Specimen Inspection (per unit)
Specimen capsules to be used in the reactor vessel material surveillance program shall be withdrawn during t he refueling period either immediately preceeding or following the Effective Full Power Years (EFPY) of unit life as follows:
Withdrawal capsule Schedule Designation (EPPY)
~
T or U 1.2 U or T 3.3 X
5.8 Y
8.3 W,S,V,Z Standby i
105
Basis 4.3.4 The surveillance inspection program has First Ca3sule - At the time when the pre-been developed to comply with Section XI dicted c :iirt of the adjusted reference of the ASME Boiler and Pressure Vessel tempcrature is approxima tely 50o2 or at Code and applicable addenda as required one/ fourth service life, whichever is earlier.
by 10 CFR 50, Section 50 55a(g), except where specific written relief has been Second and Third Capsules - At approximately granted by the NRC pursuant to 10 CFR 50, one/ third and two/ thirds af the time interval Section 50 55a(g)(6)(1).
The design between rirst and fourth capsule withdrawal.
of the plant, state of non-destructive tes ting technology, and access to areas Fourth Capsule - Three/ fourths of service life, to be inspected require such relief.
Fifth Capsule - Standby.
The Reactor Vessel Material Surveillance Program is designed to evaluate the effects The withdrawal schedule using EFPY's of unit of radiation on the fracture toughness of life is based on a lead factor of 2 9, a cap-reactor vessel steel based on the t ra ns -
acity factor of approximately 80% and a ition temperature approach and the frac-service life of 110 years.
ture meche.aics approach.
10 CFR 50, Appendix H, Paragraph Il C.3.C requires that the Reactor Vessel Material Surveillance Program shell provide for the testing of at least five capsules with the following withdrawal schedule:
119
ATTACHMENT 2 Zion Station Units 1 and 2 NRJ Docket Nos. 50-295 and 50-304 Deferred Technical Specification Changes
't i
l l
i
The following list consists of tnose differences between Zion,
Station Technical Specifications (Z-TS) and W-STS identified in Reference (a) that are to be resolved at a later date.
Z-TS W-STS Description 3.2.1.e.1 3/4.1.1.4 Revise to reflect limits on MTC in 3-STS.
3.2.1.D 3/4.1.1.5 Include a minimum temperature for criticality.
3.2.1.F.1.c 3/4.1.2.1&2 Expand to require that the noration flow path also be operatie.
3.2.2.C.1 3/4.3.3.2 Revise to require at least 75% of all incore thimbles to be operable.
3.3.1.A 3/4.4.1.1 Include requirements that at least one RCP or RHR pump be in operation at all times.
3.3.5 3/4.4.7 Revise specification to include all of tne provisions of the g-STS.
3.3.6 3/4.4.8 Expand to incorporate all the provisions shown in the W-STS.
Taole 3.4-1 3/4.3.2 Expand to include instrumentation for Turbine Trip and Feedwater Isolation, Auxiliary Feedwater Actuation, and Loss of Power; also add to Table 4.4-1.
3.8.1.C&D 3/4.5.2 Limit the out of service time to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
3.8.1.G 3/4.5.4&5 c.imit 005 time to one nour.
3.6.2.C&D 3/4.4.5.2 Limit 00S time to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
3.8.6.C&D 3/4.7.3 Limit 00S time to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and one nour respectively.
3.9.5.8 3/4.9.4 Revise per E-STS.
l Z-TS n-STS Description i
3.10.1.A 3/4.6.1.2&3 Inclune the leakage testing acceptance criteria.
1 3.10.2 3/4.6.1.7 Upgrade per the requirements of the W-STS.
3.13.3 3/4.9.4 Include requirement that no fuel nandlins P. allowed unless integrity is maintainea.
3.13.7 3/4.9.10&ll Include a required minimum water level above spent fuel.
3.18 3/4.7.1.4 Revise per W-STS activity requirements if more restrictive.
3.2.2.B.6 3/4.1.3.1 Revise to require reanalysis of the accidents indicated in W-STS Table 3.1-1.
l 3.3.1.B 3/4.4.5 Revise to require all steam generators to be operable.
None 3/4.3.3.3&4 Seismic and meteological instrumentation.
None 3/4.3.3.5 Remote shutdown instrumentation.
None 3/4.6.1.8 Containment ventilation system.
None 3/4.6.5.1 Hydrogen analyzers None 3/4.8.3.1&2 Electrical equipment protection devices.
ATTACHMENT 3 7 ion Station Units 1 and 2 NRC Docket Nos. 50-295 and 50-304 Standard Technical Specification Not Applicaole to Zion Station t
The following list consists of those differences betwen the Zion Station ecnnical Specifications (Z-TS) and W-STS, identifieo in Reference (a), that Commonwealth Edison considers to be incompatible with the design and/or operation of the Zion units.
The basis for these conclusions is discussed in the following pages.
Z-TS W-STS Description 1.
3.3.1.D 3/4.4.4 Revise to specify a maximum pressurizer water volume.
2.
3.3.3.A&B 3/4.4.6.2 Reduce the time to identify leakage to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
3.
4.15.1.9 3/4.8.1.1 Upgrade EDC testing pe r h-STS.
4.
4.15.D 3/4.8.2.3 Upgrade battery testing per W-STS.
5.
3.15.2 3/4.8.1.1 Revise 00S and remaining source testing per W-STS.
6.
None 3/4.3.3.7 Chlorine Detection Systems.
7 None 3/4.7.10 Sealed Source Contamination.
(
I.
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-w.
.. _ _ ~ _
j Basis For Exclusion From Zion Station Tecnnical Specifications 1.
Specification 3.3.1.D requires the reactor to De maintained subtritical until a steam bubble nas been established in the pressurizer.
This prohibits taking the reactor critical when the pressurizer is in a water solid condition.
Therefore this specification is adequate and a technical specification change is unnecessary.
2.
The twenty-four hour period in which to identify leakage is required as explained in the bases for Section 3.3.3.
A four hour period does not allow sufficient time to reduce reactor
~
power to a level where personnel may enter the containment (in
< keeping with ALARA exposures).
There are at least twenty-five valven within dif ferent areas of i
tne auxiliary cuilding which could ;c7 tribute to the RCS leak rate.
Most of tnese valves do not nave individual leakoff lines 4
with separate bullseyes.
In many t a,. e s, temperature
]
measurements of leakoff lines are Pe only method of identifying leakage from these valves.
Rad P: o'.ection requirements to enter tne areas (vertical and horizonta pipe chase, under the RHR Heat Exchanger, etc.) where most P the valves are located would consume most of the allotted four 1ours.
3-5 The aesign of Zion Station does nit correspond to the plant i
design as referenced in the W-S'd, The W-STS are based on a two division, two battery, two DTC. bus plant.
Zion Station is a three division, three batte.v, three D.C.
bus plant. A loss of one D.C. bus at Zion does not result la the same loss of redundancy which occurs in the W standard plant.
The same argument abLve applies to the diesel generators as Zion Station has three diesel generators and three ESF buses.
The testing of the diesel generators is currently being performed in accordance with Regulatory Guide 1.108 upon which the W-STS is basad.
Therefore, for these items the present Zion Technical Spec 1fications conform to the intent of the W-STS and do not require modification.
~
6.
Zion Station's Hypo-cnloride System nas never been used and is inoperable.
There is no chlorine on-site for this system.
At the time this system is put into service, technical specifications will be proposed.
1
-.,.,,,__.m_,-
.________.....u_..._,,__.,.,._._..__
7.
The requirements for possession of source material are specifically stated in the Facility Operatino License, page 2, "The receipt, possession, ano use of source, oy-product, and special nuclear material as authorized by this license will be in accordance with the Commission's regulation in 10 CFR Parts 30, 40 and 70, including 10 CFR Sections 30.33, 40.32, 70.23, and 70.31."
Also, Zion Station conforms to the Illinois Rules and Regulations for Protection Against Radiation, Sections C.22 and C.26, which states the requirements for source material possession.
Since Zion Station already canforms to the above requirements, Commonwealth Edison finds no need for a duplicative specification on Sealed Source Contamination.
71244
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