ML20076E475
| ML20076E475 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 08/12/1991 |
| From: | Shiraki C Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20076E478 | List: |
| References | |
| NUDOCS 9108200161 | |
| Download: ML20076E475 (31) | |
Text
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{ *n UNITED STATES-E-
NUCLEAR REGULATORY COMMISSION
' o-I6 WASHINoTON, D C 20666
( v _,/
i 10WA ELECTRIC LIGHT AND POWER COMPANY CENTRAL-IDWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DOCKET NO. 50-331 DUANE ARNOLD ENERGY CENTER AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.174 -
License Nc. OPR-49 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Iowa Electric Light end Power Company, et al., dated December 14, 1990, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and' 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 rules and regulations of the Commission; C,
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted withnut endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; 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 Specifi-cations as-indicated,in the attachment to this license amendnent and paragraph 2.C.(2) of Facility Operating License No. DPR-49 is hereby amended to read as follows:
9108200161 910812 PDR ADOCK 05000331-P PDR
~
6
' (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.174, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
The license amendment is effective as of the date of issuance and shall be implemented within 90 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
~
Clyde Y. Shiroki, Sr. Project Manager Project Directorate 111-3 Division of Reactor Projects !!!/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: August 12, 1991 l
[
,y ATTACHMENT TO LICENSE AMENDMENT NO. 174 FACILITY OPERATING LICENSE NO. DPR-49 DOCKET NO. 50-331 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.
Remove Insert i
i 11 ii 1.0-2 1.0-2 3.4-2 3.4-2 3.4-5 3.4-5 3.5-2 through 3.5-12 3.5-2 through 3.5-12 3.5-15 through 3.5-27 3.5-15 through 3.5-23 3.7-16 3.7-16 3.7-46 3.7-46 U
(
i 1
. ~ --
DAEC-1 TECHNICAL SPECIFICATIONS.
TABLE OF CONTENTS PAGE NO.
I l
1.0 ' Definitions 1.0-1 t
LIMITING SAFETY
_ SAFETY LIMITS SYSTEM SETTING
)
i 1.1 Fuel Cladding Integrity 2.1 1.1-1 1.2 Reactor Coolant System Integrity 2.2 1.2-1 SURVEILLANCE LIMITING CONDITION FOR OPERATION REQUIREMENTS
{
3.1 Reactor Protection System 4.1 3.1-1 l
i 3.2 Protective Instrumentation 4.2 3.2-1 A.
Primary Containment Isolation Functions A
3.2-1 B.
Core and Containment Cooling Systems B
3.2-1 C.
Control Rod Block Actuation C
3.2-2 f
D.
Radiation Monitoring Systems D
3.2-2 E.
Drywell Leak Detection E
3.2-3 F.
Surveillance Information Readouts F
3.2-3 l
G.
Recirculation Pump Trips and Alternate I
Rod Insertion G
3.2-4 l
H.
Accident Monitoring Instrumentation H
3.2-4
[
f 3.3 Reactivity Control 4.3 3.3-1 A..
Reactivity Limitations A
3.3-1 B.
Control Rods-B 3.3-3 i
C.
Scram Insertion Times C
3.3-6 D.
Reactivity Anomalies D
3.3-7
[
f E..
Recirculation Pumps E
3.3-7 3.4 Standby Liquid Control System 4.4 3.4-1 A.
Normal System Availability A
3.4-1 B.
Operation with Inoperable Components 3.4-2 i
C.-
Sodium Pentaborate' Solution C
3.4-2 3.5 Core _-and Containment-Cooling Systems 4.5 3.5-1 A.
Core Spray and LPCI Subsystems A
3.5-1
[
B.
Containment Spray Cooling Capability B
3.5-4
[
l i
Amendment No. JM, 151,174 l
DAEC-1 SURVEILLANCE LIMITING CONDITION FOR OPERATIO_N REQUIREMENTS PAGE NO.
3.5 Core and Containment Cooling-Systems (Continued)-
C.
Residual Heat Removal Service C
3.5-5 Watt.r System l
D.
HPCI Subsystem D
3.5-6 E.
Reactor Core 1 solation Cooling E
3.5-7 Subsystem F.
Automatic Depressurization System F
3.5-9 G.
Minimum Low Pressure Cooling G
3.5-10 l
and Diesel-Generator Availability H.
Maintenance of Filled Discharge Pipe H
3.5-11 1.
Engineered Safeguards Compartments I
3.5-11 Cooling & Ventilation J.
River Water Supply System J
3.5-12 3.6 Primery System Boundary 4.6 3.6-1 A.
Thermal and Pressurization A
3.6-1 Limitations B.
Coolant Chemistry B
3.6-3 C.
Coolant Leakage C
3.6.
D.
3afety and Relief Valves D
3.6-5 E.
Jet Pumps E
3.6-6 F.
Jet Pump Flow Mismatch F
3.6 G.
Structural Integrity G
3.6-8 H,
Shock Suppressors (Snubbers)
H 3.6-10
{
11 Amendment No.174 1
DAEC-1
- 5. OPERABLE-OPERABILITY A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s).
Implicit
-in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).
A verification of OPERABILITY is an administrative check, by examination of appropriate plant reenrds (logs, surveillance test records), to determine that a system, subsystem, train, component or device is not inoperable.
- 6. OPERATING Operating means that a system or component is performing its intended functions in its required manner.
7.
IMMEDIATE Immediate means that the required action will be initiated as soon as practical considering the safe operation of the unit and the importance of the required action.
8.
REACTOR POWER OPERATION Reactcr power operation is any operation with the mode switch in the "Startup" or "Run" position with tne reactor critical and above 1% rated power.
a) SINGLE LOOP OPERATION (SLO): REACTOR POWER OPERATION with only one of the.two recirculation loops in operation.
9, HOT STANDBY CONDITION Hot standby condition means operation with coolant temperature greater than 212 F,-reacto* vessel pressure less than 1055 psig, and the mode switch in the Startun/ Hot Standby position.
- 10. COLD CONDITION Reactor coolant temperature equal to or less than 212*F.
- 11. HOT SHUTDOVN The reactor is in the shutdown mode and the reactor coolant temperature greater than 212*F.
- 12. COLD SHUTDOWN The reactor is in the shutdown mode, the reactor coolant temp. iture equal to or less than 212aF, and the reactor vessel is vented to atmosphere.
1.0-2 Amendment No. J20,174
~
DAEC-1 LIMITfNG CONDITION FOR OPERATION SURVEILLANCE RE0VIREMENT b.
Manually initiate the system to open both explosion actuated valves and conduct flow tests to inject domineralized water through one Standby Liquid Control pump directly into the reactor vessel.
Explode one of three charges manufactured in same batch to verify proper function. Then install the untested charges in the explosion valves.
c.
Prove capability of the sodium-pentaborate storage tank discharge line to convey the minimum pump flow rate of 26.2 gpm.
B.
Operation with Inoperable Components 1.
With an inoperable redundant SLCS component, operation may continue provided that the redundant SLCS components are verified to be OPERABLE: restore the inoperable component to OPERABLE status within 7 days.
C.
Sodium Pentaborate Solution C.
Socium Pentaborate Solution At all times when the Standby Liquid The following tests shall be Control System is required to be performed to verify the operaole the following conditions availability of the liquid Control shaii be met:
Solution:
1.
The ret volume versus concentration of one Liquid Control Solution in 1.
Volume:
Check and record at least the 'iouid control tank shall be once per day.
maintained as required in Figure 3, t. - l.
3.4-2 Amendment No.174
DAEC-1 rule requirements,* only one of the two standby liquid control-pumps is needed for meeting the SLCS design basis. One inoperable pumping circuit does not immediately threaten shutdown capability, and reactor operation can continue while the circuit is being repaired. Assurance that che remaining system will perform its i
intended function and that the long-term average availability of the system is not reduced is obtained for a one-out-of-two system by an allowable equipment out-of-service time of one third of the normal surveillance frequency. This method determines an equipment out-of-service time of ten days. Additional conservatism is i
introduced by reducing the allowable out-of-service time to seven days.
3.
Level indication and alarm indicate whether the solution volume has changed, which might indicate a possible solution concentration change. The test interval has been established in consideration of l
these factors. Temperature and liquid level alarms for the systom are annunciated in-the control room.
The NRC's final rule on Anticipated Transients Without Scram (ATWS),
10 CFR % 50.62, requires that the Standby Liquid Control System (SLCS) be modified to provide an equivalent shutdown capability of 86 gpm at 13.4 weight percent natural boron for a 251 inch I.D. vessel.
For the DAEC, ATWS equivalence is achieved by running both SLCS pumps simul-taneously at a minimum combined flow of 45 gpm at a nominal boron i
concentration of 13% weight percent natural boron. -(NEDC-30859, "Duano Arnold ATWS Assessment," December 1984).
(The equivalence is also met if both pumps supply their minimum tech spec flowrate of 26.2 gpm each with a solution concentration of at least 11.2 weight percent natural boron.) Because ATWS is a very low probability event and is considered i
to be beyond the design basis for the DAEC, the surveillance and LC0 l
requirements need not be more stringent than the original SLCS design basis requirements.
3.4-5 Amendment No. LEI 174
e DAEC-1 LIMITING CONDITION FOR OPERATf0N SURVEILLANCE REQUIREMENT lite Freauenev
.n d.
Pump flow rate-Once/3 months Both loops shall deliver at least 3020 gpa against a system head corresponding to a reactor vessel pressure of 113 psig.
2.
With one Core Spray subsystem inoperable provided the other Core Spray, subsystem, LPCI, and the diesel generators are verified to be OPERABLE,-restore the inoperable Core Spray subsystem to OPERABLE status within 7 days.
l l
3.5-2 Amendment No. J39,J A3,Jg0,174
DAEC-1 LfMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT l
- 3. -
The LPCI Subsystem shall be 3.
LPCI Subsystem Testing shall be l
-OPERABLE whenever irradiated fuel as follows:
is in the reactor vessel, and prior to reactor startup from a 1133 Frecuency I
i COLD CONDITION except as a.
Simulated Annual specified in 3.5.A.4, 3.5.A.5 and Automatic 3.5.G.3 below.
Actuation Test b.
Pump Once/3 months Operability c.
Motor Operated Once/3 months Valve Operability d.
Pump Flow Once/3 months-Rate Three LPCI pumps shall deliver 14,400 gpm against a system head corresponding to a vessel pressure of 20 psig based on individual pump tests, e.
Once per shif t Ms Jalls inspect and verify that RhR valve panel lights and instrumentau on are functioning normally.
4
- With one RHR (LPCI) pump-inoperable, provided the remaining RHR (LPCI) active components, both Core Spray subsystems, the containment spray subsystem, and the diesel generators are verified to be OPERABLE, restore the inoperable RHR (LPCI) pump to OPERABLE: status within 30 days.
5.
With two RHR-(LPCI) pumps inoperable, providing both~ Core Spray subsystems, the containment spray subsystem, and the diesel generators are verified to be OPERABLE, restore at least one RHR (LPCI) pump to OPERABLE status within 7 days.
3.5-3 Amendment No. JM,160,174
e
- -.. - - ~
DAEC-1 i
LIMITING CONDITION FOR OPERATION
-SURVEILLANCE RE0VIREMENT 6.
Othervise, be in at least HOT
[
SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3 B.
Containment Sorav Coolina B.
Containment Sorav Coolina Cacability f
Canability 1.
The suppression pool and drywell spray modes of the residual heat Surveillance of the containment I
removal RHR) system shall be spray loops shall be performed as OPERABLE (with two independent follows:
loops each when the reactor water-temperature is rester than 212 F 1.
During each five year period, an t s speci ted in 3.5.8.2 and air test shall be perforwed on the drywell and suppression pool spray 2.
With one suppression pool spray headers and nozzles, loop and/or one drywell spray loop inoperable restore the inoperable 1000toOP$RABLEstatuswithin30 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following
-24 hours.
3.
With both suppression pool-5 pray loops and/or both drywell spray loops inoperable, restore at least one loop to OPERABLE status within B hours or be in HOT SHUTDOWN ithin the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in w
COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.5-4 Amendment No J0P,,174
DAEC-1 LIMITING CONDITION FOR 'PERATION SVRVEILLANCE REQUIREMENT O
lC.
Residual Heat Removal (RHE}
C.
lyrveillance of the RHR Service a
Service Water System Water System 1.
Surveillance of the RHR service 1.
Except as specified in 3.5.C.2 water system shall be as follows:
3.5.C.3, 3.5.C.4, 3.5.C.5, and 3.5.G.3 below, both RHR service RHR Service Water Subsystem
'I "9
water subsystem loops shall be operable whenever irradiated fuel lits Frecuency is in the reactor vessel and reactor coolant temperature is a.
Pump and Motor Once/3 months greater than 212*F.
P"*t'd V'IV' operability.
b.
Flow Rate after major Test-Each pump RHR service maintenance water pump and every 3 shall deliver months at least 2040 gpm at a TDH of 510 ft. or more.
2.
With one RHR$W pump inoperable, provided the remaining active components of both RHR$W subsystems are verified to be OPERABLE. restore the inoperable pump to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3, With one RHRSW pump in each su5 system inoperable, provided the remaining active components of both RHRSV subsystems and the diesel generators are verified to be OPERABLE, restore at least one inoperable pump to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.5-5 Amendment No. Jpp,,J33,J79,174
DAEC-1
{
LIMIT!NG CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT t
4 With one RHRSW subsystem inoperable, provided the remaining RHRSW subsystem and its associated
^
diesel generator are verified to l
be OPERABLE, restore the L
inoperable system to OPERABLE status with at least one OPERABLE pump within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
L i
i D.
HPCI Subsystem D.
HPCI Subsystem I
t
.l.
The HPCI Subsystem shall be 1.
HPCI Subsystem testing shall be l
OPERABLE whenever there is pedoM as Mows:
irradiated fuel in the reactor h
Freauenev vessel, reactor pressure is
(
greater than 150 psig, and prior a.
ul Annual
(
to reactor startup from a COLD Actuation Test i
CONDITION, except as specified in 3.5.D.2 and 3.5.0,3 below, b.
Pump Operability Once/3 Months I
c.
Motor Operated (ince/3 Months l
Valve Operability d.
At rated reactw Once/3 months pressurn demonstrate ability to deliver rated flow at a disdcego I
pressure greater tnan or equai to that pressure requfred to accomplish pe15el I
injection if vessel pressure were as hign as 1040'psig.
i k
l i
3.5-6 Amendment No. JJE,Jf 3,JfD 174
OAEC-1 LIMITfNG CONDIVION'FOR OPERATION SURVEILLANCE REQUIREMENT-e.
At reactor CU.e/ operating l
e I
pressure of cycle 150 +/- 10 psig demonstrate i
ability to 3
deliver rated
}
flow at a discharge pressure greater than or equal to that pressure required to i
accomplish
[
vessel injection.
1 I
2.
With HPCI inoperable, provided The HPCI pump that both Core Spray subsystems, shall deliver LPCI,_ ADS, and RCIC are verified at least 3000 to be OPERABLE, restore HPCI to
_gp, rop,
OPERABLE status within 14 days >
i or be in at least HOT SHUTDOWN system head j
within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and corresponding reduce reactor _ steam dome pressure to a reactor j
to less than or eaual to 150 psig pressure of j
within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
1040 to 150 psig.
j
't g
f.
Verify that Once/ operating l
t the suction for Cycle i
i the HPCI system l
15 automatically I
transferred from the condensate storage tank _to the suppression pool on a condensate storage tank water level-low l
signal and on a suppression pool water level-high signal.
}
E.
Reactor Core Isolation Coolino E.
lea: tor Core Isolation Coolina l
(RCIC) Subsyltim fRC: C) Subsystem i
1.
RCIC Subsystem testin performed as follows:g shall be j
1.
The RCIC Subsystem shall be OPERABLE whenever there is irradiated fuel-in-the reactor Freauency vessel, the reactor pressure is a.
S mu ated Annual
_Au greater than 150 psig,_and prior gc to reactor startuo from a COLD (and restart)
CONDITION, except as specified in 3.5.E.2 below, b.
Pump Operability Once/3 months j
c.
Motor Operated Once/3 months Valve Operability I
3.5-7 Amendment No. 97 JA3,JB0,174 c
DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT itim Freauency i
d.
At rated reactor Once/3 months l-pressure demonstrate ability to i
deliver rated flow at a discharge pressure greater than or aq: sal to that prissure required to accomplis.h vessel ' jection if vessel pressure wara 's high as 1040 gig.
e.
At reactor Once/ operating l
pressure of cycle 150 1 10 psig demonstrate ability to
+
deliver rated flow at a discharge pressure greater than or equal to that pressure required to accomplish vessel injection.
The RCIC pump shall i
deliver at least 400 gpm for a system head corresponding to 1040 to 150 psig.
f.
Verify that the Once/opertting I
}
suction for the cycle RCIC system is automatically transferred from
+
the condensate starage tank to the suppression pool on a condensate storage tank water level-low signal.
s 3.5-8 Amendment No. JJE,JM,174 T
--~
DALC -
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 2.
With the RCIC system ~ inoperable, provided the HPCI-system is verified to be OPERABLE, restore l
the RCIC system to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor steam i
dome pressure to less than or t
equal to 150 psig within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
F.
Automatic Deoressurization F.
Automatic Georessurization l
System (Aq11 System (AQEt 1.
Once per operating cycle the 1.
The Automatic Depressurization following tests shall be performed Subsystem shall be-OPERABLE on the ADS:
whenever there is irradiated fuel in the reactor vessel and the a.
A simulated automatic actuation
[
reactor pressure is greater than
$[,ftupfome hREhUL$NG 100 psig and prior to a startup OUTAGE.
from a Cold Condition, except as i
specified in 3.5.F.2 below, b.
The ADS Nitrogen Accumulator check valves will be leak tested for a maximum acceptable system leakage rate of 25 sec/ minute.
2, With one ADS valve inopersble, provided that HPCI is ve'ified to be OPERABLE, restore thr.
inoperable ADS valve tr OPERABLE status within 30 days er be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor steam i
dome pressure to less than or equal to 100 psig within the i
following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.
With two or more ADS valves I
inoperable, be in at least HOT SHUTDOWN wit;.in the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor steam. dome pressure to less than or equal to l
100 psi hours g within the following 24 i
t i
I h
l
- 3. 5-9 Amendment No. JJE,J27,J/$,J/3J74 i
l
DAEC 1 LlH! TING CON 0li!0N FOR OPERATION
$URVE!LLANCE REQUIREMENT
'G, Minin g low pregsure_Caglino and G.
MLn,lege. Low Pressure Coolina and l
I Diesel Generator Avaj,hkflily Dji1LLGenerator Avat tability 1.
During any period when one diesel 1.
When it is determined that one i
generator is inoperable, continued diesel generator is inoperable, i
reactor operation is perwissible only during the succeeding seven the remaining diesel generator l
days unless such diesel generator shall be demonstrated to be is sooner made OPERABLE, provided OPERABLE within eight (8) hours i
that the remaining diesel and daily thereafter.
In I
generator ano all low pressure addition, all low pressure core i
c[e a con nm o ng cooling and containment cooling OPERfBLEdiese$generatorare subsystems supported by the OPERABLE.
If this requirement OPERABLE diesel shall be verified cannet be met an orderl $HUTDOWN to be OPERABLE.
shall be initlated and t e reactor shall be in at least HOT $HU100VN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD $HUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
j 2.
Any combination of inoperable components in the core and contain-ment cooling systems shall not defeat the capability of the remaining OPERABLE components to fulfill +,he cooling functions.
l l
3.
When irradiated fuel is in the l
reactor ves;el and the reattor is in the COLD $HUTDOWN Condition or Refuel Mode:
l 4.
I' no work is being performed
[
.nich has the potential for draining the reactor vessel, f
both core spray and RHR systems may be inoperable; or b.
If work ($ being performed which has the potential for draining the reactor vessel, at least two of any combirDtion of core spray
(
and/or RHR (LPCI or shutdown i
cooling modei pumps shall be f
OPERABLE (including the capabi-lity to inject water into the reactor vessel with suction from the suppression pool) except as i
i i
t 3.5-10 Amendment No. E1.139,162,174 5
..---m.--,y..
m,-_,,,.-m..c.,,m.
m.
,._,e
_,,_,,m..
--c...
e-
.-.~,m,
.w,
I (uCC-1 i
l.!MITING CON 0lT10N FOR OPERATION
$URVE!LLANCE RE0VIRLMENT i
l specified in Specification i
3.5.0.3.b(1) and (2), below.
A diosal generator required for i
operatfor, of at least one of L
these pumps shall be OPERABLE.
I (1)Withoneofthetwopumps i
inoperable, restore the inoperable pump to OPERABLE l
status within four hours or susrend all operations with i
a potential for draining the L
reactor vessel.
(2) With both pumps inoperable, suspend all operations with a potential for draining the reactor vessel.
i t
4 During a refueling outage, CORE ALTERATIONS may continue with the
(
suppression pool volume below the r'
minimum values specified in
$pecification 3.7.A.) provided all of the following cenditions are
{
met:
l a.
The reactor head is removed, the cavity is flooded, the spent fuel pool gates are removed and i
spent fuel pool water level is maintained within the limits of l
Specification 3.9.C.
i l
b.
At least one Core Spray pump l
-capable of transferring water to the vessel is OPERABLE with suction aligned to the condensate storage tank (s).
I c.
The condensate storage tanks contain at least 75,000 gallons of water which is available to i
the core spray subsystem.
Condensate storage tank (s) level shall be recorded at least every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, i
-l d.
No work is being performed which has the potential for draining the reactor vessel.
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3.5 10a Amendment No. PJ.Jfl.174 f
i
..... 1
t DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
[
~~
l 5.
If the requirements of Specification 3.5 G.4 cannot be met, suspeno CORE ALTERATIONS.
H.
ntenance of Filled Dischay,gg gg g
E.158 1.
Whenever core sprafC1subsystees'a re LPCIsubsystems,PERA8LE,the H
or RCIC required to be O 1.
The following surveillance discharge piping from the pump requirements shall be adhe ed to, j
discharge of these systems to the to assure that the discharge i
last block valve shall be filled.
piping of the core spray and LPCI the core spray or Lk! piping of f
If the pump discher a.
subsystems
}
depressurires below the system low a.
The pressure switches ehich pressure alarm setpoint while monitor the LPCI and core spray these systems are required to be lines to ensure they are full OPERABLE, the pressure shall be restored within one hour, shall be functionally tested b.
If specification 3.5.H.1 or 3.5.H,].a cannot be met, either l
place the affected-system (s) in b.
The pressure switches which the test mode or declare the r
affected system (s) inoperable and monitor the LPCI and Core Spray enter th> applicable LIMITING lines to ensure they are full
[
CONDITION FOR (VERATION as shall be calibrated once per described in Spet.ification 3.5.A.
operating cycle.
3.5.0 or 3.5.E.
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1.
Eneineered Safeouards Comeartments fagfutered Safeauards Comeartments Coolino ana Ventilation (3011no and Ventilation l
If both unit coolers serving sither the RCIC or HPCI room are The unit coolers for each of the r
out of service,.the associated RCIC, HPCI, Core Spray, and RHR
[
pump shall be Considered pump rooms shall be checked for
[
inoperable for purposes of operability during surveillance f
ons 3.5.0 or 3.5.E as
]
testing of the associated pumps.
If the single unit cooler serving either compartment which houses i
two RHR pumps and a core spray pump l
is out of-service for a period i
greater than seven days, the associated pumps shall be i
considered inoperable for purposes of Specification 3.5 A.
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i 3.5-11 Amer.dinent No.1/3.J/7,174
DAEC 1
\\
LIMITING CONDITION FOR OPERA 110N
$URVEILLANCE REQUIREMENT i
J.
River Water !unnly $vitas J.
River Waler sucolv system 4
1.
Except as specified in 3.5.J.2 1.
River Water supply System Testing:
{
below, at least one pump in each 113tn Frenuency f
river water supply system loop shall be OPERABLE whenever a.
$toulated Once/ operating irradiated fuel is in the reactor automatic cycle vessel and reactor coolant actuation test, j
temperature is greater than 212 F.
b.
Pump and motor Once/3 months l
operated valve t
operability.
c.
Flow Rate Test Each river water After major pump Supply system maintenance and pump shall once per deliver at least 3 months 6000 gpm at TDH i
of 46 ft, or Daily when river
- more, elevation is less than 727 l
feet.
l d.
Operating Pump i
Flow Rate Demonstration Each Operating uelly i
River Water i
Supply System t
Pump shall deliver at least 6000 gpm.
2, With one river water supply loop inoperable, provided the other river water supply loop and its associated diesel genrrator are verified to be OPERABLE, restore at least one-pump in the inoperable loop to OPERABLE status within 7 days or be in at least I
HOT SHUT 00WN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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3.5-12 Amendment No. 70,DP.H3,40,174 i
-i
DAEC 1 Core spray distribution has been %n, in full-scale tests of systems similar in design to that of DAEC to escoed the minimum requirements.
In addition, cooling ef fectiveness has been demonstrated at less than half the rated flow in t,imulated fuel assemblies with heater rods t'o duplicate the decay heat characteristics of irradiated
' fuel.
The accident analysis is additionally conservative in that no credit is taken for spray coolant entering the reactor before the internal pressure has fallen to 150 psig.
The LPCI subsystem is designed to provide emergency cooling to the core by flooding in the event of a loss-of-coolant accident. This system functions in combination with the core spray system to prevent excessive fuel clad temperature. The LPCI subsystem and the core spray subsystem provide adequate coolleg for break areas of approximately 0.2 sausre feet up to and including the double ended recirculation line break without assistance from the high pressure emergency core cooling subsystems.
The allowable repair times are established so that the average risk rate for repair would be no greater than the basic risk rate.
The method and concept are described in Reference 1.
Using the results developed in this reference, the repair period is found to I
be 1/2 the test interval.
This assumes that the core spray subsystems and LPCI constitute a 1 out of 3 system; however, the combined ef fect of any of the two subsystems to limit excessive clad temperatures must also be conside.*ed.
1 I
The survetilance requirements provide adequate assurance that the Core Spray subsystems and the LPCI subsystem will be operable when required.
Should the loss of one LPCI pump occur, a nearly full complement of core and containment spray equipment is available.
The remaining three LPCI pumps and a core spray subsystem will perform the core cooling function.
Because of the availability of the majority of the core cooling equipment, which will be verified to be operable, a thirty day repair period is justified.
If-the LPCI subsystem is not available, at least 1 LPCI pump must be available to fulfill the containment spray function.
The 7 day repair period is set on this basis.
3.5-15 Amendment tio.174
DAEC-1 B & C. Containment Sorav and RHR Service _ Water l
The containment spray subsystem for DAEC consists of 2 loops each with 2 LPCI pumps and 2 RHR service water pumps per loop. The water pumped through the RHR heat enchangers may be diverted to two spray headers in the drywell and one above the suppression pool. The design of these systems is predicated upon use of 1 LpCl, and 2 RHR service water pumps for heat removal af ter a design basis event.
Thus, there are ample spares for margin above the design conditions.
Loss of margin should be avoided and the equipment maintained in a state of operability so a 30-day tut-of-service time is chosen for this equipment.
If one loop is out-of-service. or one pump in each loop is out-of-service, reactor operation is permitted for seven days.
The surveillance requirements provide adequate assurance thtt ther f.ontaineent $ pray subsystem and RHR$W system will be operable when requ'.reo.
i t
Analyses were performed to determine the minimum required flow rate of the RHR Service Water pumps in order to meet the design basis case (Reference 4) and the NUREG-0783 requirements (Reference 5).
(See Section 3.7.A.1 Bases for a discussion of the NUREG requirements).
The results of these analyses justify reducing the required flowrate to 2040 gpm per pump, a 15% reduction in the original 2400 gpm per pump requirement.
f f
3.5-16 Amendment No. Jpf,733,733,174 s
mm-
,y,,._.y_.,,
_em.,,.m_..,,..,._m_..._-.mm.,,w,_..,,,_.y._.,%m<_.w,..,m_..._
r-
_.+.u.u
1 MEC-1 D.
HPCI System I
The HPCI system is provided to assure that the reactor core is adequately cooled to limit fuel clad temperature in the event of a small break in the nuclear system and loss-of-coolant, which does not result in rapid depressurization of the reactor vessel.
The HPCIS permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized. The HPCI$ continues to operate until reactor vessel pressure is below the pressure at which LPCI operation or Core $ pray System operation maintains core cooling.
The capacity of the system is selected to provide this required core cooling, ihe HPCI pump is designed to pump 3000 gpm at reactor pressures between approximately 1135 and 150 psig. Two sources of water are available.
Initially, demineralized water from the condensate storage tank is used instead of in,fecting water from the suppression pool into the reactor.
When the HPC! System begins operation, the reactor depressurites more rapidly than would occur if HPCI was not initiated due to the condensation of steam by the cold fluid pumped into the reactor vessel by the HPCI System. As the reactor vessel pressure continues to decrease, the HPCI flow momentarily reaches equilibrium with the flow through the break. Continued depressurization causes the break flow to decrease below the HPCI flow and the liquid inventory beginsl to rise.
This type of response is typical of the small breaks. The : ore never uncovers and is continuously cooled throughout the transient so that no core damage of any kind occurs for breaks that lie within the capacity range of the HPCI.
s As mentioned in Section 6.3.1 of the Updated FSAR, the ADS provides a single failure proof path for depressurization for postulated transients and accidents.
The RCIC serves as an alternate to the HPCI only for decay heat removal even if the feedwater is assumed to be lost.
The surveillance requirements provide adequate assurance that the HPCI system will be operable when required. With the HPCI system inoperable, adequate core cooling is assured by the operability of the redundant and diversified ADS and both the CS and LPCI systems.
The HPCI out-of-service period of 14 days is based on the 3.5-17 Amendment No. J N, 174
DAEC-1 operability of redundant and diversified low pressure core coollng systems and l the RCIC system, i
The HPCI and RCIC as well as all other Core Standby Cooling Systems must be operable when starting up from a Cold Condition.
It is realized that the HPCI is not designed to operate until reactor pressure exceeds 150 psig and is automatically isolated before the reactor pressure decreases below 100 psig.
It is the intent of this specification to assure that when the reactor is being started up from a Cold Condition, the HPCI is not known to be inoperable.
E.
RCIC $vs.tn l
The RCIC is designed to provide makeup to the nuclear system as part of the planned operation for periods when the main condenser is unavailable. RCIC also serves for decay heat removal when feedwater is lost.
In all other postulated accidents and transients, the ADS provides redundancy for the HPCI.
Based on this, an allowable repair time of 1 month is justified, however, a maximum l
allowable repair time of 14 days is selected for conservatism.
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F.
Automatic Deoressur42ation System (ADS) l The operability of the ADS under all conditions of depressurization of the nuclear system automatically or manually, insures an essential response to station abnormalities.
The nuclear system pressure relief system provides automatic nuclear system depressurization for small breaks in the nuclear system 50 that the low pressure coolant injection (LPCI) and the core spray subsystems can operate to protect the fuel barrier.
Because the Automatic Depressurization System does not provide makeup to the reactor primary vessel, no credit is taken for the steam cooling of the core caused by the system actuation to provide further conservatism to the CSC$,
Performance analysis of the Automatic Depressurization System is considered only with respect to its depressurizing ef fect in conjunction with LPCI and Core Spray and is based on 3 valves. There are four valves in the ADS and each has a capacity of approximately 810,000 lb/hr at a set pressure of 1125 psig.
3.5-18 knendment No JU,174
m DAEC-1 The allowable out-of-service time for one AD$ valve is determined as thirty days because of the redundancy and provided the HPCI system is operable. Therefore, I redundant protection for the core with a small break in the nuclear system is still available.
The A05 test circuit permits continued surveillance on the operable relief valves to assure that they will be available if required.
The Nitrogen supply to the AD$ utilites occumulators and inlet check valves to ensure the operability of the AD$ in the event that a break occurs in the nonseismic portion of the nitrogen supply piping.
The accumulators are sited to allow the A0$ to operate at least 5 times after a period of 100 days post accident with a maximum system leakage rate of 30 sec/ minute. To provide an additional margin of safety, the leakage test allows for a maximum acceptable leakage rate of 2$ sec/ minute.
G.
Minimum low pressure toolina and Diesel Generator Availability l
The purpose of Specification G is to assure that adequate core and containment I tooling equipment (h, as specified in Specifications 3.5. A. 3.5,8 and 3.5.C) I is available at all times, it is during refueling outages that major maintenance is performed and during such time that all low pressure core cooling systems may be out of service.
This specification provides that should this occur, no work will be performed on the primary system which could lead to draining the vessel. This work would include work on certain control rod drive components and recirculation system.
Thus, the specification precludes the e unts which could require core cooling.
If work must be perforwed which has the potential for draining the vessel, $pecification 3.5 G,3.b requires that certain low pressure core cooling subsystems be available and capable of injecting water into the reactor vessel from the suppression pool water supply.
The condensate storage tanks are not considered to be an appropriate water supply as they are not safety related and could provide makeup water for core cooling for only a finite period of time.
The makeup capability of either one core spray pump or one low pressure coolant injection (LPCI) pump is more than double the leakage rate expected from a 3.5-19 Amendment No. JJE.7/7,174
DAEC 1 postulated failure of the control rod velocity limiter section.
Since the system cannot be pressurized during refueling, the potential need for core flooding only entsts and the specified combination of the core spray or the LPCI system can provide this. $pecification 3.8 must also be consulted te determine other requirements for the diesel generators.
To prevent estensive wear and stress en the diesel engines, the diesels are manually started and the speed incrementally increased to synchronous speed.
i H.
h Lnitnance of Filled Discharae Pion l
l i
If the discharge piping of the core spray, LPCI subsystem, HPCI, and RCIC are not filled, a water hammer can develop in this piping when the pump and/or pumps are started.
If a water hammer were to occur at the time at which the system were required, the system would still perform its design function.
- However, to minimize damage to the discharge piping, specification 3.5.H requires that the core spray and LPCI discharge piping pressure be restored within one hour after system depressuritation when the system is required to be operable.
Likewise, for HPCI and RCIC, the discharge piping to the last block valve shall be filled when these systems are required to be operable.
If the discharge piping pressure for the core spray and LpCI subsystems cannot be restored within one hour or the discharge piping for HPCI and RCIC cannot be maintained in a filled condition to the last block valve, the operator is required to perform l
either of the following actions:
- 1) place the af fected system (s) in the test mode which will ensure f. hat the discharge piping is filled with water, or
- 2) declare the affected system (s) inoperable in which case the operator will enter the applicable LCO for the affected systee(s) as defined in Specification 3.5.A (core spray and LPCI), 3.5.0 (HPCI), or 3.5.E(RCIC).
The above actions minimize the possibility of a water hammer and are considered conservative in nature.
3.5-20 Amendment flo. J M.Jf>2, 174
(MEC-1 1.
Enoineered Safecuerds Comeartments Coolino and Ventilation l
t One unit cooler in each pump compartment is capable of providing adequate ventilation flow and cooling.
Engineering analyses indicate that the temperature rise in safeguards compartments without adequate ventilation flow or cooling is such that continued operation of the safeguards equipment or associated auxiliary eouipment cannot be assured.
J.
River Weter Sueolv Svsu s l
Four river water supply pumps in two loops of two pumps each are provided. Both loopt. discharge into the wet pit sump of the RHR and emergency service water system. One river water supply pump is sufficient to supply water to an entire train of RHR and emergency service water pumps, which in turn provide sufficient service water for containment and component cooling after a loss-of-coolant accident.
An additional pump is required to be operable in Specification 3.5.J.1 to provide a completely redundant river water supply for the other RHR and emergency service water train.
Because of the almost continuous operation of the river water supply system during normal operation, two additional pumps, for a total of four, have been installed to provide flexibility in maintenance and operation as well as additional system reliability.
In the event that one river water supply system loop becomes inoperable, plant l operation is restricted to seven days.
l 3.5-21 Amendment No.137,1fD,174
DAEC-1
3.5 REFERENCES
- 1. Jacobs, l.H., Guidelines for Determinino Safe Test Intervals and Rena_ty Times for Enaineered Safeauards, General Electric Company, APED, April,1969 I (APED 5736).
- 2. General Electric Company, lhe GESTR-LOCA and_5AFER Models for the Evaluation of loss-of-Coolant Accident, NEDC-23785-P, October 1984.
- 3. General Electric, Dnat Arnold Enerov_ Canter SAFER /GESTR-LQGA Loss-of-Coolant Accident Analysis, NEDC-31310P, June 1988.
l
- 4. General Electric Company, Analysis of Reduced RHR $ervice W811r Flow at the Dyane Arnold Enerov Center, NEDE-30051-P, January 1983.
- 5. General Electric Company, Dugat Arnold Energy._ Center $ueoression Pool Iemberature Resooalt, NEDC-22082-P, March 1982.
3.5-22 Amendment No. JJE,JA2,JEA,167,174
DAEC-1 I
4.5 BASES Core and Containment Cooling Systems surveillance Frequencies l
The testing interval for the core and contain=7nt cooling systems is based on industry practice, quantitative reliability analysis, judgement and practicality.
The core cooling systems have not been designed to be fully r
testable during operation.
For example, in the case of the HPCI, automatic i
initiation during power operation would result in pumping cold water into the reactor water vessel which is not desirable. Complete ADS testing during power operation causes an undesirable loss-of-coolant inventory.
To increase the i
availability of the core and containment cooling systees, the components which make up the system, i.e., instrumentation, pumps, valves, etc., are tested frequently. The pumps and motor operated injection valves are also tested every three months to assure their operability.
The test intervals are based upon
$ection XI of the ASME Code. A simulated automatic actuation test once per year i combined with frequent tests of the pumps and injection valves is deemed to be adequate testing of these systems.
/
When components and subsystems are out-of-service, overall core and containment I
cooling reliability is maintained by evaluating the operability of the remaining i
equipment.
The degree of evaluation depends on the nature of the reason for the out-of service equipment.
For routine out-of-service periods caused by preventative maintenance, etc., the evaluation may consist of verifying the i
redundant eauipment is not known to be inoperable and applicable surveillance intervals have been satisfied. However, if a failure due to a design deficiency caused the outage, then the evaluation of operability should be thorough enough to assure that a generic problem does not exist.
The RHR valve power bus is not instrumented.
For this reason servelliance recuirements require once per shift observation and verification of lights and instrumentation operability.
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3.5-23 Amendment No. fJ,J/3 JfD,174 i
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e r.
n w
l DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 1
2.a The rtsults of the inplace cold DOP 2.a The tests and sample analysis of l
and halogensted hydrocarbon tests Specification 3.7.B.2 shall be in the flow range of 3600-4000 cfm
[erged ni a
t
,n on HEPA filters and charcoal af ter every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system adsorber banks shall show I 99.9%
operation and following DOP removal and 1 99.9%
significant painting, fire or halogenated hydrocarbon removal, eh ven ati o
e icating with the system.
b.
The results of laboratory carbon b.
Cold 00P testing shall be performed sample analysis shall show < l.0%
after each complete or partial penetration of radioactive methyl replacement of the HEPA filter iodide at 70% R.H.,150+F, 40 + 4 FPM face velocity with an inlet bank or after any structural concentration of 0.5 to 1.5 mgide.
maintenance on the system housing.
/m' inlet concentration methyl io c.
Fans shall be shown to be capable c.
Halogenated hydrocarbon testing of operation from 1800 cfm to the shall be performed after each flow range of 3600-4000 cfm.
complete or partial replacement j
of the charcoal adsorber bank or af ter any structural maintenance on the system housing.
l d.
Each circuit shall be operated with i
the heaters on at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month.
3.
With one train of SGTS inoperable, operation or fuel handling may continue provided the remaining SGTS is verified to be OPERABLE-restoretheinoperableSGTStraln i
to OPERABLE status within 7 days i
or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and suspend fuel handling i
operations.
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3.7-16 Amendment do. 78,J/3,Jpg,Jg3, 174 i
l L
2
DAEC 1 Demonstration of the automatic initiation capability and operability of filter cooling is necessary to assure system performance capability.
Initiating reactor bui: ding isolation and operation of the +andby gas treatment system to maintain at least a 1/4 inch of water vacuum within the secondary containment provides an adequate test of the operation of the reactor building isolation valves, leaktightness of the reactor building and performance of the standby gas treatment
- system, functionally testing the initiating sensors and associated trip channels demonstrates the capability for automatic actuation.
Perfonning these tests prior to refueling will demonstrate secondary containment capability prior to the time the primary containment is opened for refueling.
Periodic testing gives sufficient confidence of reactor building integrity and standby gas treatment system perfonnance capability.
8.
Primary Containment Power Operated Isolation Valves Automatic isolation valves are provided on process piping which penetrates the containment and connunicates with the containment Amendment No. 7f. )W 174 3.7-46