ML20127J637
| ML20127J637 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 01/15/1993 |
| From: | Dyer J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20127J639 | List: |
| References | |
| NUDOCS 9301250213 | |
| Download: ML20127J637 (20) | |
Text
.#
'o UNITED bTATES E'
Ig NUCLEAR REGULATORY COMMISSION n
f k
WASHINGTON D. C. 20555 1
\\....+/
[0MMONWEALTH EDISON COMPANY DOCKET NO. 50-373 LASALLE COVNTY STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 89 License No. NPF-11 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment filed by the Commonwealth Edison Company (the licensee) dated April 24, 1991, as supplemented June 2, 1992, and December 1, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulaticas set forth in_iG CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; C.
There is reasonable assurance: (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with' the Commission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical-to the cormon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the enclosure to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-ll is-hereby amended to read as follows:
9301250213 930115 DR ADOCK 0500 3
4
-2 (2)
Technical Soecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 89, and the Environmental Frotection Plan contained in Appendix B, are hereby incorporated in_the license.
The licensee shall operate-the facility in accordance with-the Technical Specifications and the Environmental Protection Plan.
3.
This amendment is effective upon date of issuance to be implemented within 45 days.
FOR THE NUCLEAR REGULATORY COMMISSION NW4 (
&V ames E. Dyer, ' Director Project Directorate III-2 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
January 15, 1993
ATTACHMENT TO LICENSE AMENDMENT NO. 89 FACILITY OPERATING LICENSE NO. NPF-11 DOCKET NO. 50-373 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.
The revised pages are identified.by amendment number and contain a vertical line indicating the area of change.
REMOVE INSERT 3/4 1-4 3/4 1-4 3/4 1-5 3/4 1-5 B3/4 1-2 B3/4 1-2 B3/4 1-3 B3/4 1-3 B3/4 1-4 B3/4 1-4 B3/4 1-5 83/4 1-5 B3/4 1-6
~
ol ed ntr SYSTEM nt niu o
Co rted:
al c ct oin TROL TION inse CON drie OPERA FOR od(s) isciated nd exha EEACTIVITY ust TION ol r asso COND ntr the I
a co r
l rm wate ab e a
TING e
ext in pe dis driv n
r IMI ed) o ur the L
nt niu the ho the TION (Co If With ni *1 eithe : ally, o clo i
r r
sing WN with n lic SHUTDOWN l
a) valv ctric ally valv SHUTDO ab e.
AC 2
es s
by e.
Ele ulicolat oin HOT ot app HOT a
st n
st 1)
Hydr is lea r
at 3re lea r ab e,
l ate in 3.0.4 at 2) w in in pe wis, be at oin o
ab e, be alv e
e l
cific r
v Othe in per o
ne s
r.
Spe ho of ods with o u
b) 12 ab e,
l s
r.
ns u
s ol r lnie days.
visio ontr ain 7
12 ho r
in pe o
o xt The pr 8c dr witnin es ne the alv n
or tha nt lin r.
e i
both v e
ho 50Vv ciated OWN with n with 3
s ore u
With m 12 s
s ur.
i more SHUTD lin 8 ho 12 es u
asso r.
ho with n r ## the HOT c.
o ain with n ext ated ne i
dr With o olate n
in the str or wis, be V ve ciatedWN with n e
nt lin i
demon d
Is e
SD be 1
r Othe SHUTDO shall more asso es nd r ## the HOT 2
o valv n* *, a ne ith o olate in nt ope st o ne W
wis, be nd v be e
Is lea e
a to e,
ain valv o gh at 1
r e
Othe TS dr UIREMEN volu ea u
me ch thr wn ed 2
alve ith rda LANCE i cha ge rifying ch v rm RE r
dsia ea s
each scram days cycling RWM, all walve ving ds ve VEIL 31 ol v mc SUR Ey r
the ontr b
The e pe days el.
al cOPERABL E y:
st o e pe full setpoint of 4.1.3.1 1 92 trav nc n
ctioated a
b r
as e
OPERABL At st cycle w po their lea wer dir str vab e onc of l
a.
lea demon immo lo ve be is Atomplete the ha e
c.
shall mlr **fe od ren e
to b.
c N
of p
REACTIVITY' CONTROL-SYSTEM-LIMITING CONDITION FOR OPERATION (Continued)
ACTION-(Continued)-
f 2.
If the inoperable control rod (s).is inserted:
a)
Within'1 hour-disarm the associated directional control >
l valves
- either:
1)
Electrically, or
[
2)
Hydraulically by closing the drive water and exhaust !
water isolation valves.
b)
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
3.
The provisions of Specification 3.0.4 are'not applicable, c.
With more than 8 control rods inoperable, be in at least HOT SHUTDOUN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, d.
With one or more SDV vent or-drain lines with one valve inoperable, _
1.
Isolate ## the associated line within 7_ days.
2.
Otherwise, te in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; e#
With one or more SDV vent or drain lines with both valves-inoperable,.
1.
Isolate ## the associated line.within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.-
2.
Otherwise, be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS s 4.1.3.1.1 The scram discharge volume drain and vent valves shall be demonstrated OPERABLE by:
a.
At least once per 31. days verifying each valve.to be open**, and b.
At least once per_92 days-cycling each valve through at-least one-complete cycle of full travel.
4.1.3.1.2 When above the,iow power setpoint of-the RWM, all-withdrawn control rods not required to-have their-directional control valves disarmed electrically or hydraulically shall be demonstrated OPERABLE _by movingieach' control rod at least one notch:
a.
At least once per 7 days, and
-b.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when any control rod is immovable _as a:
result of excessive friction or mechanical interference.
- May.be rearmed intermittently, under administrative._ control, to permit testing; associated with restoring the contro1-rod to OPERABLE status.
~
- These valves may be closed intermittently for testing under administrative control.
- Separate Action statement' entry is allowed for each SDV vent and drain line.
-##An isolated line may be unisolated under administrative control to allow draining and venting of the SDV.
LA SALLE -UNIT 1-3/4 1-4 Amendment No. 89 l
. =.
REACTIVITY CONTROL SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.1.3.1.3 All cor. trol rods shall be demonstrated OPERABLE by-performance of Surveillance Requirements 4.1.3.2, 4.1.3.4, 4.1.3.5, 4.1.3.6 and 4.1.3.7.
4.1.3.1.4 The scram discharge volume shall be deterniined OPERABLE by demonstrating the scram discharge volu.ie drain and vent valves OPERABLE at least once per 18 months by verifying that tho' drain and vent valves:
a.
Close within 30 seconds after receipt of a signal for control rods to scram, and b.
Open after the scram signal is reset.
l l
LA SALLE - UNIT 1 3/4 1-5 Amendment No. 89
~
REACTIVITY CONTROL SYSTEMS HASES 3/4.1.3 CONTROL RODS The specification of this section ensure that (1) the minitum SHUTDOWN MARGIN is maintained, (2) the control rod insertion times are consistent with those used in the accident analysis, and (3) the potential effects of the rod drop accident are limited.
The ACTION statements permit variations from the basic requirements but at the same time impose more restrictive criteria for continued operation.
A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimum.
The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.
Damage within the control rod drive mechanism could be a generic problem, therefore with a control rod immovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large
.mber of inoperable control rods.
Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those in the nonfully-inserted position are consistent with the SHUTDOWN MARGIN requirements.
The number of control rods permitted to be inoperable could-be more than the eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor must be shutdown for investigation and resolution of the problem.
The contral rod system is designed to bring the reactor subcritical at a rate fast enough to prevent the MCPR from becoming less than the fuel cladding safety limit during the limiting power transient analyzed in Section 15.0 of the FSAR.
This analysis shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the specifications, provide the required protection and MCPR remains greater than the fuel cladding safety limit, The occurrence of scram times longer then those specified should be viewed a: an ;ndication of a systemic problem with the rod drives and therefore the surveillance interval is educed in order to prevent operation of the reactor for long periods of time with a potentially serious problem.
The SDV vent and drain valves are normally open and discharge any accumulated water in the SDV to ensure that sufficient volume is available at all times to allow a complete scram. During a scram, the SDV vent and drain valves close to contain reactor water. The SDV consists of header piping that connects to each hydraulic control unit (HCU) and drains into an instrument volume.
There are two headers and two instrument volumes, each receiving approximately one half of the control rod drive (CRD) discharges.
The two instrument volumes are connected to a common drain line. The common drain line has two valves in series.
Each header is connected to a common vent line. This common header has two valves in series.
The header piping is sized to receive and contain all the water discharged by the CRDs during a scram.
LA SALLE - UNIT I B 3/4 1-2 Amendment No. 89 m-
b.
REACTIVITY CONTROL SYSTEMS BASES
}/4.1.3 CONTROL RODS (Continued)
The Design Basis Accident and transient analyses assume that all of the control rods are capable of scramming.
The primary function of the SDV is to limit the amount of reactor coolant discharged during a scru. The acceptance criteria for the SDV vent and drain valves are that they operate automatically to:
a.
Close during scram to limit the amount of reactor coolant discharged so that adequate core cooling is maintained and offsite doses remain within the limits of 10 CFR 100; and b.
Open on scram reset to maintain the SDV vent and drain path open such that sufficient volume is available to accept the reactor coolant discharged-during a scram.
The OPERABILITY of all SDV vent and drain valves ensures that, during a scram, the SDV vent and drair, valves will close to contain reactor water discharged into the SDV piping.
Since the vent and drain lines are provided with two valves in series, the single failure of one valve in the open position will-not impair the isolation function of the system. Additionally, the valves are required to be op a to ensure that a path is available for the SDV piping to drain freely at other times.
Isolation of the SDV can also be accomplished by closure of the SDV valves under administrative control. Additionally, the discharge of reactor coolant to the SDV can be terminated by scram reset or closure of the HCU manual isolation valves.
For a bounding leakage case, the offsite doses are well within the limits of 10 CFR 100 and adequate core cooling is maintained.
Note " contained in Specification 3.1.3.1 allows Action Statements d and e to be entered separately for each affected SDV vent and drain line, and Completion Times to be tracked on a per line basis.
For instance, when a vent valve is-declared inoperable, Action d is entered for the vent line and its Completion Time starts.
If a drain valve is subsequently declared inoperable, Action d is entered again for the drain line and a separate Completion Time starts and is tracked for the drain line.
The same is true for both valves inoperable in one line in accordance with Action e, provided the original Completion Time (if any) affecting that line is not exceeded. Also, one line can be in Action d, while the other line is in Action e, provided the applicable Completion Times are met for each line.
Control rods with inoperable accumulators are declared inoperable and.
Specification 3.1.3.1 then applies.
This prevents a pattern of inoperable accumulators that would result in less reactivity insertion on a _ scram than.
has been analyzed even though control rods with inoperable accumulators may still be inserted with nornial drive water pressure. Operability of the accumu-lator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of tne reactors.
LA'SALLE - UNIT 1 B 3/4 1-3 Amendment No. 89
BEACTIVITY CONTROL SYSTEMS BASES 3/4.1.3 CONTROL RODS (Continued) in addition, the automatic CRD charging water header low pressure scram (see Table 2,2.1-1) initiates well before any accumulator loses its full capa-bility to insert the control rod. With this added automatic scram feature, the surveillance of each individual accumulator check valve is no longer necessary to demonstrate adequate stored energy is available for normal scram action.
Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR.
The overtravel position feature provides the only positive means of determining that a rod is properly coupled and therefore this check must be performed prior to achieving criticality after completing CORE ALTERATIONS that could have affected the control rod drive coupling integrity. The subsequent check is performed as & backup to the initial demonstration.
In order to casure that the control rod patterns can be followed and therefore that scher parameters are within their limits, the control rod position indication system must be OPERABLE.
The control rod housing support restricts the outwaro movement of a control rod to less than 3.65 inches in the event of a housing failure.
The amount of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage to the primary coolant system.
The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.
The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.
3/4.1.4 CONTROL R0D PROGRAM CONTROLS u
Control red withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control rod drop accident. The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal.
When THERMAL POWER is greater than 10% of RATED THERMAL POWER, there is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal /gm.
Thus requiring the RWM to be OPERABLE when THERMAL POWER is less than or equal to 10% of RATED THERMAL POWER provides adequate ccntrol.
The RWM provide automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.
The analysis of the rod drop accident is presented in Section 15.4.9 of the FSAR and the techniques of the analysis are presented in a topical report, Reference 1, and two supplements, References 2 and 3.
LA SALLE - UNIT 1 B 3/4 1-4 Amendment No. 89
REACTIVITY CONTROL SYSTEMS BASES 3/4.1.4 CONTROL R0D PROGRAM CONTROLS (Continued)
The RBM is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation.
Two channels are provided.
Tripping one of the channels will block erroneous rod withdrawal.soon enough to prevent fuel damage.
This system backs up the written sequence used by the operator for withdrawal of control rods.
3/4.1.5 STANDBY LIOVID CONTROL SYSTEM The standby liquid control system provides a backup capability for bringing the reactor from full power to a cold, Xenon-free shutdown, assuming that the withdrawn control rods remain fixed in the rated power pattern. To meet this objective it is necessary to inject a quantity of boron which produces a concentration of 660 ppm in the reactor core in approximately 50 to 125 minutes. A normal quantity of 4587 gallons net of solution having a.13.4%
sodium pentaborate concentration is required to meet a shutdown requirement of 3%. There is an additional allowance of 25% in the reactor core to account for imperfect mixing.
The time requirement was selected to override the reactivity insertion rate due to cooldown following the Xenon poison peak and the required pumping rate is 41.2 gpm.
The minimum storage volume of the solution is established to allow for the portion below the pump suction that cannot be inserted and the filling of other piping systems connected to the reactor vessel.
The temperature requirement on the sodium pentaborate solution is necessary to maintain the solubility of the solution as it was initially mixed to the appropriate concentration.
Checking the volume of fluid and the temperature once cach 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is available for injection.
With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable.
Surveillance requirements are established on a frequency that assures a high reliability of the system.
Once the solution is established, boron concentration will not vary unless more boron or water is added, thus a check on the temperature and volume once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is available for use.
Replacement of the explosive charges in the valves at regular intervals will assure that these valves will not fail because of deterioration of the charges.
1.
C. J. Paone, R. C. Stirn and J. A. Woolley. " Rod Drop Accident Analysis for Large BWR's," G. E. Topical Report NED0-10527, March 1972 2.
C. J. Paone, R. C. Stirn and R. M. Young, Supplement I to NED0-10527, July 1972 3.
J. M. Haun, C. J. Paone and R. C. Stirn, Addendum 2, " Exposed Cores,"
Supplement 2 to NED0-10527, January 1973 LA SALLE - UNIT 1 B 3/4 1-5 Amendment No. 89
REACTIVITY CONTROL SYSTEMS BASES 3/4.1.6 ECONOMIC GDIERATION CONTROL SYSTEM Operation.with the economic generation control (EGC) system, at f amatic flow control, is limited to the range of 65% to 100% of rated core flow.
In this flow range and with THERMAL POWER 2 20% of RATED THERMAL POWER, the reactor could safely tolerate a rate of change of load of 8 MWe/s (reference FSAR Section 6.2.4).
Limits within the EGC and the flow control system prevent rates of change greater than approximately 4 HWe/s. When EGC is in operation,_this fact will be indicated on the main control room console.
-1 LA-SALLE - UNIT 1 B 3/4 1-6 Amendment No. 89
a
[gueoy'o UNITED STATES f3
', C NUCLEAR REGULATORY COMMISSION g
WASHINGTON. D. C. 20555 s
E
%....+/
COMMONWEALTH EDIS0N COMPANY DOCKET NO. 50-374 LASALLE COUNTY STATION. UNIT 2 AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No. 74 License No. NPF-18 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment filed by the Commonwealth Edison Company (the licensee) dated April 24, 1991, as supplemented June 2, 1992, and December 1, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; C.
There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the enclosure to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-18 is hereby amerded to read as follows:
L
_ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _. (2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 74, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This amendment is effective upon date of issuance to be implemented within 45 days.
FOR THE NUCLEAR REGULATORY COMMISSION m2f.Qw James E. Dyer, Director Project Directorate 111-2 Division of Reactor Projects - Ill/IV/V
~
Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: January 15, 1993 t
M i
i ATTACHMENT TO LICENSE AMENDMENT NO. 74
~
FACILITY OPERATING LICENSE NO. NPF-18 DOCKET NO. 50-374 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain a vertical line indicating the area of change.
REMOVE INSERT 3/4 1-4 3/4 1-4 i
3/4 1-5 3/4 1-5 B3/4 1-2 83/4 1-2 i
B3/4 1-3 B3/4 1-3 B3/4 1-4 B3/4 1-4 B3/4 1-5 B3/4 1-5 83/4 1-6
REACTIVITY CONTROL SYSTEM
_ LIMITING CONDITION FOR OPERATION (Continued)
ACTION (Continued)-
2.
If the inoperable control rod (s) is inserted; a)
Within I hour disarm the associated directional control valves
- either:
1)
Electrically, or 2)
Hydraulically by. closing the drive water and exhaust water isolation valves, b)
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
3.
The provisions of Specification 3.0.4 are not applicable.
c.
With more than 8 control rods inocerable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
d With one or more SDV vent or drain lines with one valve inoperable, l
1.
Isolate ## the associated line within 7 days.
2.
Otherwise, be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
e#
With one or more SDV vent or drain lines with both-valves inoperable, 1.
Isolate ## the associated line within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
2.
Otherwise, be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The scram discharge volume drain and vent valves shall be demonstrated OPERABLE by:
a.
At least once per 31 days verifying each valve to be open**, and b.
At least once per 92 days cycling each valve through at least one complete cycle of full travel.
4.1.3.1.2 When above the low power
.tpoint of the RWM, all withdrawn control rods-not required to have their directional control valvos disarmed electrically or hydraulically shall be demonstrated OPERABLE oy moving each control rod at least one notch:
a.
At least once per 7 days, and b.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when any control rod is immovable as a result of excessive friction or mechanical interference.
"May be rearmed intermittently,. under administrative control, to permit testing -
associated with restoring the control rod to OPERABLE status.
- These valves may be closed intermittently for testing under admin'istrative control.
- Separate Action statement entry is allowed for each SDV vent and drain line.
- An isolated line may be unisolated under administrative control to allow draining and venting of the SDV.
LA SALLE - UNIT 2 3/4 1-4 Amendment No. 74
REACTIVITY CONTROL SYSTEM SURVEILLANCE REQUIRE)'CiTS (Continued) 4.1.3.1.3 All control rods shall be demonstrated OPERABLE by performance of Surveillance Requirements 4.1.3.2, 4.1.3.4, 4.1.3.5, 4.1.3.6 and 4.1.3.7.
4.1.3.1.4 The scram discharge volume shall be determined OPERABLE by demonstrating the scram discharge volume drain and vent valves OPERABLE at least ence per 18 months by verif ing that the drain and vent valves:
f Close within 30 seconds after receipt of a signal for control a.
rods to scram, and b.
Open after the scram signal is reset.
LA SALLE - UNIT 2 3/4 1-5 Amendment No. 74-4 w
REACTIVITY CONTROL SYSTEMS BASES 3/4.1.3 CONTROL ROM The specification of this section ensure that (1) the minimum SHUTDOWN MARGIN is maintained (2) the control rod insertion times are consistt.nt with those used in the accident analysis, and (3) the potential effects of the rod drop accident are limited.
The ACTION statements permit variations from the basic requirements but at the same time impose more restrictive criteria for continued operation.
A limitation on inoperable rods is set such that the resultant effect on total rcd worth and scram shape will be kept to a minimum.
The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.
Damage within the control rod drive mechanism could be a generic problem, therefore with a control rod immovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods.
Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those in the nonfully-inserted position are consistent with the SHUTDOWN MARGIN requirements.
The number of control rods permitted to be inoperable could be more than the eight allowed by thn specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor must be shutdown for investigation and resolution of the problem.
The control rod system is designed to bring the reactor subtritical at a rate fast enough to prevent the MCPR from becoming %ss i..an the fuel cladding safety limit during the limiting power transient anaiyzea in Section 15.0 of the FSAR.
This analysis shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the specifications, provide the required protection and MCPR remains greater than tle fuel cladding safety limit. The occurrence of scram times longer then those specified should be viewed as an indication of a systemic problem with the rod drives and therefore the surveillance interval is reduced ia order to prevent operation of the reactor for long periods of time with a potentially serious problem.
The SDV vent and drain valves are normally open and discharge any accumulated water in the SDV to ensure that sufficient volume is available at all times to allow a complete scram. During a scram, the SDV vent and drain valves close to contain reactor water. The SDV consists of header piping that connects to each hydraulic control unit (HCU) and drains into an instrument volume.
There are two headers and two instrument volumes, each receiving approximately one half of the control rod drive (CRD) discharges. The two instrument volumes ate connected to a common drain line. The common drain line has two valves in series.
Each header is connected to a common vent line. This common header has two valves in series, fhe header piping is sized to rcceive and contain all the water discharged by the CRDs during a scram.
LA SALLE - UNIT 2 B 3/4 1-2 Amendment No. 74
-~
REACTIVITY CONTROL SYSTEMS BA$fS 3/4.1.3 CONTROL RODS (Continued)
The De:ign Basis Accident and transient analyses assume that all of the control i
rods are capable of scramming.
The primary function of the SDV is to limit the-amount of reactor coolant discharged during a scram.
The acceptance criteria for the SDV vent and drain valves are that they operate automatically to:
a.
Close during scram to limit the amount of reactor coolant discharged so that adequate core cooling is maintained and offsite doses remain within the limits of 10 CFR 100; and b.
Open on scram reset to maintain the SDV vent and drain path open such that sufficient volume is available to accept the reactor coolant discharged during a scram.
The OPERABILITY of all SDV vent and drain valven ensures that, during a scram, the SDV vent and drain valves will close to contain reactor water discharged into the SDV piping.
Since the vent and drain lines are provided with two valves in series, the single failure of one valve in the open position will not impair the isolation function of the system. Additionally, the valves are required to be open to ensure that a path is available for the SDV piping to drain freely at other times.
Isolation of the SDV can also be accomplished by closure of the SDV valves under administrative control. Additionally, the discharge of reactor coolant to the SDV can be terminated by scram reset or closure of the HCU manual isolation valves.
For a bounding leakage case, the offsite doses are well within the limits of 10 CFR 100 and adequate core cooling is maintained.
Note
- contained in Specification 3.1.3.1 allows Action Statements d and e to be entered separately for each affected SDV vent and drain line, and Completion Times to be tracked on a per line basis.
For instance, when a vent valve is declared inoperable, Action d is entered for the vent line and its Completion Time starts.
If a drain valve is subsequently declared inoperable, Action d is entered again for the drain line and a separate Completion Time starts and is-tracked for the drain line. The same is true_for both valves inoperable-in one line in accordance with Action e, provided the original Completion Time (if any affecting that line is not exceeded. Also, one line can be in Action d, whi)le the other line is in Action e, provided the applicable Completion Times are met for each line.
Control rods with inoperable accumulators are declared inoperable and Specification 3.1.3.1 then applies. This prevents a pattern of inoperable accumulators that would result in less reactivity insertion on a scram than has been analyzed even though control rods with inoperable accumulators may still be inserted with normal drive water aressure.
Operability of the accumu-lator ensures that there is a means availa)1e to insert the control rods even under the most unfavorable depressurization of the reactors.
A LA SALLE - UNIT 2 B 3/4 1-3 Amendment No. 74
REACTIVITY CONTRQLSYSTEMS BAsrs 314 l.3 CONTROL RQM (Continued) a In addition, the automatic CRD charging water header low pressure scram (see Table 2.2.1-1) initiates well before any accumulator ',oses its full capa-bility to insert the control rod. With this added automatic scram feature, the surveillance of each individual accumulator check valve is no longer necessary to demonstrate acequate stored energy is available for normal scram action.
Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the TSAR.
The overtravel position feature provides the only positive means of determining that a rod is properly coupled and therefore this check must be performed prior to achieving criticality after completing CORE ALTERATIONS that could have affected the control rod drive coupling integrity.
The subsequent check is performed as a backup to the initial demonstration.
In order to ensure that the control rod patterns can be followed and therefore that other parameters are within their limits, the control rod position indication system must be OPERABLE.
The control rod housing support restricts the outward movement of a control rod to less than 3.65 inches in the event of a housing failure.
The amount of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage to the primary coolant system.
The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.
The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.
3]Ll.4 CONTROL R0() PROGRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control rod drop accident.
The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal. When THERMAL POWER is greater than 10% of RATED THERMAL POWER, there is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal /gm.
Thus requiring the RWM to be OPERABLE when THERMAL POWER is less than or equal to 10% of RATED THERMAL POWER provides adequate control.
The RWM provide automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.
The analysis of the rod drop accident is presented in Section 15.4.9 of the FSAR and the techniques of the analysis are presented in a topical report, Reference 1, and two supplements, References 2 and 3.
1 LA SALLE - UNIT 2 B 3/4 1-4 Amendment No. 74
~ - - _
REACTIVITY CONTROL SYSTQi1
+
PASES 3/4.1.4 CONTROL R00 PROGRAM CONTROLS (Continued)
The RBH is designed to automatically prevent fuel damage in the event of l
erroneous rod withdrawal from locations of high power density during high 30wer operation.
Two channels are provided. Tripping one of the channels will
)1ock erroneous rod withdrawal soon enough to prevent fuel damage. This system backs up the written sequence used by the operator for withdrawal of control rods.
3/4.1.5 STANDBY __L10VID CONTROL SYSI{d The standby liquid control system provides a backup capability for bringing the reactor from full power to a cold, Xenon-free shutdown, assuming that the withdrawn control rods remain fixed in the rated power pattern.
To meet this objective it is necessary to inject 6 quantity of boron which produces a concentration of 660 ppm in the reactor core in approximately 50-to 125 minutes.
A normal quantity of 4587 gallons net of solution having a 13.4%
sodium pentaborate concer.+ ration is required to meet a shutdown requirement of 3%. There is an additional allowance of 25% in the-reactor core to account for imperfect mixing.
The time requirement was selected to override the reactivity insertion rate due to cooldown following the Xenon poison 3eak and the required pumping rate is 41.2 gpm.
The minimum storage volume of tie solution is established to allow for the portion below the pump suction that cannot be inserted and the filling of other piping systems connected to the reactor vessel.
The temperature requirement on the sodium pentaborate solution is necessary to maintain the solubility of the solution as it was initially mixed to the appropriate concentration.
Checking the volume of fluid and the temperature once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is available for injection.
With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable.
Surveillance requirements are established on a frequency that assures a high reliability of the system.
Once the solution is established, boron concentration will not vary unless more boron or water is added, thus a check on the temperature and volume once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is available for use, i
Replacement of the explosive charges in the valves at regular intervals-will assure that these valves will not fail because of deterioration of the charges.
1.
C. J. Paone, R. C. Stirn and J. A. Woolley, " Rod Drop Accident Analysis for Large BWR's," G. E. Topical Report NE00-10527, March 1972 l
2.
C. J. Paone, R. C. Stirn and R. M. Young, Supplement 1 to NE00-10527, July 1972 3.
J. M. Haun, C. J. Paone and R. C. Stirn, Addendum 2, " Exposed Cores,"
Supplement 2 to NE00-10527, January 1973 LA SALLE - UNIT 2 B 3/4 1-5 Amendment No. 74 1
REAC11VITY CONTROL SYSTEMS BASES 2/4.1.6 ECONOMIC GENERATION CONTROL SYSTEM Operation with the economic generation control (EGC) system, automatic flow control, is limited to the range of 65% to 100% of rated core flow.
In this flow rance and with THERMAL POWER 2 20% of RATED THERMAL POWER, the reactor could safely tolerate a rate of change of load of 8 MWe/s (reference FSAR Section 6.2.4).
Limits within the ECC and the flow control system prevent rates of change greater than approximately 4 MWe/s. When EGC is in operation, this fact will be indicated on the main control room console.
1
~
l
~
i L
LA SALLE - UNIT 2 8 3/4 1-6 Amendment No. 74
...