ML20045F297
| ML20045F297 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 06/29/1993 |
| From: | Pickett D Office of Nuclear Reactor Regulation |
| To: | Phares R ILLINOIS POWER CO. |
| References | |
| TAC-M81044, NUDOCS 9307070174 | |
| Download: ML20045F297 (9) | |
Text
June 29,1993 i
Docket No. 50-461 DISTRIBUTION-Docket'FileJ GHill (2)
NRC & Local PDRs Wanda Jones Mr. Richard F. Phares PDIII-2 p/f.
CGrimes Director - Licensing JRoe ACRS (10) l Clinton Power Station JZwolinski OPA P. O. Box 678 JDyer OC/LFDCB Mail Co'.! V920 CMoore BClayton, RIII Clinton, "'inois 61727 DPickett DHagan OGC RBarrett Dear Mr. Phare RJones
SUBJECT:
CHANGES TO TECHNICAL SPECIFIr;TIONS - CLINTON POWER STATION, oNIl NO. 1 (TAC NO. M81044) f Your letter of June 17, 1991 (U-601848), requested changes to the Bases section of the Clinton Power Station Technical Specifications.
The proposed changes were categorized to accomplish the following:
(1) replace the Maximum i
Average Planar Linear Heat Generation Rate (MAPLHGR) limit factor "0.85" for single recirculation loop operation with a reference to the COR.E OPERATING LIMITS REPORT since this factor is operating cycle specific; (2) correct the number of nozzles in the Residual Heat Removal (RHR) containment spray Train "A"
to reflect the deletion of two spray nozzles; and (3) clarify the Bases F
for the Main Control Room and Standby Gas Treatment system filter train heaters to note that heater test results are corrected for actual bus voltage j
during testing.
l We have found the proposed Bases changes acceptable.
The revised Bases pages and the related overleaf pages are enclosed with this letter and should be used to replace the current pages.
l Sincerely, j
OrigintdSignedBy Chandu Patel for Douglas V. Pickett, Project Manager Project Directorate III-2 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation l
Enclosure-L Bases Pages l
[
p8 g g Ag >,,, h @ y cc w/ enclosure:
.g L
See next page
- See previous concurrence
),
~pf
//o P
D PM:PD32
- BC:SPLB B kCSB BC:SRXB D:PD32 g i
, h2 Mopze RLa gr:rc DPickett CMcCracken RBarrett
/pJoney JDyer i
/t1/93
/M G/tl /93 06/09/93
{ /1I/93 e/[ff93 h /%j/93 l
D CUMENT NAME:
G:CLI81044.TSB g
9307070174 930629 k
PDR ADOCK 05000461 N(
P pog L
M
T Docket No. 50-461 DISTRIBUTION Docket File.
Wanda Jones GHill (2)-
1 NRC & Local PDRs Mr. Richard F. Phares PDIII-2 p/f CGrimes-Director - Licensing-JRoe ACRS (10)
Clinton Power Station JZwolinski OPA P. O. Box 678 Mail Code V920 -
JDyer-OC/LFDCB CMoore BClayton, RIII Clinton, Illinois 61727 DPickett-DHagan-OGC
Dear Mr. Phares:
SUBJECT:
BASES CHANGES TO TECHNICAL SPECIFICATIONS - CLINTON POWER STATION, UNIT NO. 1 (TAC N0. M81044)
Your letter of June 17, 1991 (U-601848), requested changes to the Bases section of the Clinton Power Station Technical Specifications.
The proposed changes were categorized to accomplish' the following:
(1) replace the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit factor "0.85" for single recirculation loop operation with a reference to the CORE OPERATING LIMITS REPORT since this factor is operating cycle specific; (2) correct the' number of nozzles in the Residual Heat Removal (RHR) containment spray Train "A" to reflect the deletion of two spray nozzles; and (3) clarify the Bases for the Main Control Room and Standby Gas Treatment system filter train heaters-to note that heater test results are corrected for actual bus voltage -
during testing.
We have found the' proposed Bases changes acceptable. The revised Bases'pages.
and the related overleaf pages are enclosed with this letter and should be used to replace the current pages.
Sincerely, Douglas V. Pickett, Project Manager Project Directorate 111-2 Division of Reactor Projects Ill/IV/V Office of Nuclear Reactor Regulation
Enclosure:
Bases Pages cc w/ enclosure:
See next page 3V ks Ys/ f LA:PD32 PE:PD32 PM:PD32 BC:SPLB D:PD32 CMoore RLaufer:rc DPickett CMcCracken JDyer
-/ /93 Q ')/93 L/ 7 /93 4 /f/93
/
/93 DOCUMENT NAME: G:CL181044.TSB
June 29, 1993 w
V Docket No. 50-461 DISTRIBUTION Docket File GHill (2)
NRC & Local PDRs Wanda Jones Mr. Richard F. Phares PDill-2 p/f CGrimes Director - Licensing JRoe ACRS (10)
Clinton Power Station JZwolinski OPA P. O. Box 678 JDyer OC/LFDCB Mail Code V920 CMoore BClayton, RIII Clinton, Illinois 61727 DPickett DHagan OGC RBarrett
Dear Mr. Phares:
RJones
SUBJECT:
BASES CHANGES TO TECHNICAL SPECIFICATIONS - CLINTON POWER STATION, UNIT NO. 1 (TAC NO. M81044)
Your letter of June 17, 1991 (U-601848), requested changes to the Bases section of the Clinton Power Station Technical Specifications.
The proposed changes were categorized to accomplish the following:
(1) replace the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit factor "0.85" for single recirculation loop operation with a reference to the CORE OPERATING LIMITS REPORT since this factor is operating cycle specific; (2) correct the number of nozzles in the Residual Heat Removal (RHR) containment spray Train "A" to reflect the deletion of two spray nozzles; and (3) clarify the Bases for the Main Control Room and Standby Gas Treatment system filter train beaters to note that heater test results are corrected for actual bus voltage' during testing.
We have found the proposed Bases changes acceptable.
The revised Bases pages and tha related overleaf pages are enclosed with this letter and should be useo to replace the current pages.
Sincerely, O#na! Signed By* Chandu Patel for
)
Douglas V. Pickett, Project Manager Project Directorate III-2 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Enclosure:
Bases Pages cc w/ enclosure:
See next page
- See previous concurrence'pb fj l
2 P PD PM:PD32
- BC:SPLB B /CSB C:SRXB D:PD32 p CMo. e RLa DPickett CMcCracken RBarrett yJone JDyer k/
/93 hr:rc G/n /93 06/09/93
/ /1 4/93 e//
3 (f/tj/93 DOCUMENT NAME:
G:CL181044.TSB
i
- 9' Mr; Richard'F. Phares Clinton' Power Station j
'lllinois Power Company Unit No. I cc:
.Mr.-J. S. Perry Illinois Department l
Vice President
'of Nuclear Safety Clinton Power Station Office of Nuclear Facility Safety i
Post Office Box 678 1035 Outer Park Drive Clinton, Illinois 61727 Springfield, Illinois '62704 Mr. J. A. Miller Mr. Donald Schopfer Manager Nuclear Station Project Manager Engineering Department Sargent-& Lundy Engineers Clinton Power Station 55 East Monroe Street Post Office Box 678 Chicago, Illinois 60603 Clinton, Illinois 61727 Sheldon Zabel, Esquire Schiff, Hardin & Waite 7200 Sears Tower l
233 Wacker Drive Chicago, Illinois 60606 l
Resident inspector U.S. Nuclear Regulatory Commission RR#3 Box 229 A Clinton, Illinois 61727 Ms. K. K. Berry i
Licensing Services Manager
'i General Electric Company i
175 Curtner Avenue, M/C 382 San Jose, California 95125 Regional Administrator, Region Ill j
799 Roosevelt Road. Building 4 Glen Ellyn, Illinois 60137 Chairman of DeWitt County c/o County Clerk's Office' DeWitt County Courthouse Clinton, Illinois 61727 Mr. Robert Neumann Office of Public Counsel State of Illinois Center-100 W. Randolph Suite 11-300 Chicago, Illinois 60601 l
i e
1
'I
?
POWER DISTRIBUTION LIMITS BASES 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued) b.
Model Chanae q
1.
Core CCFL pressure differential - 1 psi - Incorporate the assumption that flow from the bypass to lower plenum must overcome a 1 psi pressure drop in core.
2.
Incorporate NRC pressure transfer assumption - The assumption used in the SAFE-REFLOOD pressure transfer when the pressure is increasing was changed.
A few of the changes affect the accident calculation irrespective of CCFL.
I These changes are listed below.
a.
Input Chance 1.
Break Areas - The DBA break area was calculated more accurately.
b.
Model Chanae 1.
Improved Radiation and Conduction Calculation - Incorporation of CHASTE 05 for heatup calculation.
A list of the significant plant input parameters to the loss-of-coolant accident analysis is presented in Bases Table B 3.2.1-1.
For plant operation with a single recirculation loop, the MAPLHGR limits specified in the CORE OPERATING LIMITS REPORT are multiplied by the smallest of MAPFAC,, MAPFAC or the numerical factor specified for single recirculation p
loop operation in the CORE OPERATING LIMITS REPORT (Reference 2).
The numerical factor is derived from LOCA analyses initiated from single loop operation to account for earlier boiling transition at the limiting fuel node compared to standard LOCA evaluations.
3 /4. 2. 2 APRM SETPOINTS
[ DELETED]
CLINTON - UNIT 1 B 3/4 2-2 letter dated 6/2793
9 CONTAINMENT SYSTEMS BASES 3/4.6.3 DEPRESSURIZATION SYSTEMS The specifications of this section ensure that the drywell and containment pressure will not exceed the design pressure of 30 psig and 15 psig, respectively, during primary system blowdown from full operating pressure.
The suppression pool water volume must absorb the associated decay and structural sensible heat released during a reactor blowdown from 1040 psia.
Using corservative parameter inputs, the maximum calculated containment pressure during nJ foliowing a design basis accident is below the containment desicr. p :ssure of 15 psig. Similarly the drywell pressure remains below the design ressure of 30 psig. The maximum and minimum water volumes for the suppressicn pool are 150,230 cubic feet and 146,400 cubic feet, respectively.
These values include the water volume of the containment pool, horizontal vents, and weir annulus. Testing in the Mark III Pressure Suppression Test Facility and analysis have assured that the suppression pool temperature will not rise above 185*F for the full range of break sizes.
Should it be necessary to make the suppression pool inoperable, this shall only be done as specified in Specification 3.5.3.
Experimental data indicates that effective steam condensation without excessive load on the containment pool walls will occur with a quencher device and pool temperature below 200*F during relief valve operation.
Specifications have been placed on the envelope of reactor operating conditions to assure the bulk pool temperature does not rise above 185"F in compliance with the containment structural design criteria.
In addition to the limits on temperature of the suppression pool water, operating procedures define the action to be taken in the event a safety-relief valve inadvertently opens or sticks open. As a minimum this action shall include:
(1) use of all available means to close the valve, (2) initiate suppression pool water cooling, (3) initiate reactor shutdown, i
and (4) if other safety-relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open safety relief valve to assure mixing and uniformity of energy insertion to the pool.
The containment spray system consists of two 100% capacity trains, each with two spray rings located at different elevations about the inside circumference of the containment.
RHR A pump supplies one train and RHR pump B supplies the other.
RHR pump C cannot supply the spray system.
Dispersion of the flow of water is effected by 249 nozzles in Train 'A' and 251 nozzles in Train
'B',
enhancing the condensation of water vapor in the containment volume and preventing overpressurization.
Heat rejection is through the RHR heat exchangers. The turbulence caused by the spray system aids in mixing the containment air volume to maintain a homogeneous mixture for H control.
z CLINTON - UNIT 1 B 3/4 6-6 letter dated 6/29/93 i
a
CONTAINMENT SYSTEMS BASES Y
3/4.6.5 DRYWELL POST-LOCA VACUUM REllEF VALVES Drywell vacuum relief valves are provided on the drywell to pass sufficient quantities of gas from the containment to the drywell to prevent an excess negative pressure from developing in the drywell.
3/4.6.6 SECONDARY CONTAINMENT The secondary containment completely encloses the primary containment, except for the upper personnel hatch.
It consists of the fuel building, gas control boundary, and portions of the auxiliary building enclosed by the extension of the gas control boundary and the ECCS cubicles and areas as described in USAR Figure 6.2-132.
The standby gas treatment system (SGTS) is designed to achieve and maintain a negative 1/4" W.G. pressure within the secondary containment following a design basis accident.
This design provides for the capture within the secondary containment of the -
'oactive releases from the primary containment, and their filtration before c'. ease to the atmosphere.
Establishing and maintaining a vacuum in the secondary containment with the standby gas treatment system once per 18 months, along with the surveillance of the doors, hatches, dampers, and valves, is adequate to ensure that there are no violations of the integrity of the secondary containment.
The inleakage values are not verified in the surveillances since no credit for dilution was taken in the dose calculation.
As noted however, adequate drawdown is verified once per 18 months.
The acceptance criteria specified in Figure 4.6.6.1-1 for the drawdown test is based on a computer model, verified by actual performance of drawdown tests, in which the drawdown time determined for accident conditions is adjusted to account for performance of the test during normal plant conditions.
The acceptance criteria indicated per Figure 4.6.6.1-1 is based on conditions corresponding to power operation (with the turbine building ventilation system in operation) and wind speeds less than or equal to 10 mph.
The acceptance criteria for plant conditions other than those assumed will be adjusted as necessary to reflect the conditions which exist during performance of the surveillance test.
The OPERABILITY of the standby gas treatment systems ensures that sufficient iodine removal capability will be available in the event of a LOCA.
The reduction in containment iodine inventory reduces the resulting site boundary radiation doses associated with containment leakage.
The operation of this system and resultant iodine removal capacity are consistent with the assumptions used in the LOCA analyses.
Continuous operation of the system with the heaters OPERABLE for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> during each 31-day period is sufficient to reduce the buildup of moisture on the absorbers and HEPA filters. The specified heater dissipation is based on a bus voltage of 460 volts.
Heater j
test results shall be adjusted to account for actual bus voltage.
A i
3/4.6.7 ATMOSPHERE CONTROL i
l The OPERABILITY of the systems required for the detection and control of hydrogen gas ensures that these systems will be available to maintain the hydrogen con-centration within the containment below its flammable limit during post-LOCA CLINTON - UNIT 1 B 3/4 6-8 letter dated 6/29/93
3/4.7 PLANT SYSTEMS t
BASES 3/4.7.1 SHUTDOWN SERVICE WATER SYSTEM The OPERABILITY of the shutdown service water system ensures that sufficient cooling capacity is available for continued operation of safety-related equip-ment during accident conditions. The redundant cooling capacity of these sys-tems, assuming a single failure, is consistent with the assumptions used in the accident analyses within acceptable limits.
The ultimate heat sink (VHS) specification ensure that sufficient cooling capacity is available for continued operation of safety-related equipment for at least 30 days to pennit safe shutdown and cooldown of the reactor. The surveillance requirements ensure that quantities maintained are consistent with the assumption used in the accident analysis as described in the USAR and the guidance provided in Regulatory Guide 1.27, January 1976.
3/4.7.2 CONTROL ROOM VENTILATION SYSTEM The OPERABILITY of the control room ventilation system ensures that 1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this system and 2) the control room will remain habitable for operations personnel during and following all design basis accident conditions. The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent.
This limitation is consistent with the requirements of General Design Criterion 19 of Appendix "A",
10 CFR 50. Surveillance testing provides assurance that system and component performances continue to be in accordance with performance specifications for Clinton Unit 1, including applicable parts of ANSI N509-1980. Continuous operation of the system with the heaters OPERABLE for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> during each 31 day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters.
The specified heater dissipation is based on a bus voltage of 460 volts.
Heater test results shall be adjusted to account for actual bus vol t age.
3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM The reactor core isolation cooling (RCIC) system is provided to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel without requiring actuation of any of the Emergency Core Cooling System equipment. The RCIC system is conservatively required to be OPERABLE whenever reactor pressure exceeds 150 psig. This pressure is substantially below that for which the low pressure core cooling systems can provide adequate core cooling for events requiring the RCIC system.
The RCIC system specifications are applicable during OPERATIONAL CONDITIONS 1, 2, and 3 when reactor vessel pressure exceeds 150 psig because RCIC is the primary (non-ECCS) source of emergency core cooling when the reactor is pressurized.
j With the RCIC system inoperable, adequate core cooling is assured by the OPERABILITY of the HPCS system and justifies the specified 14 day out-of-service period.
CLINTON - UNIT 1 B 3/4 7-1 letter dated 6/29/93 1
PLANT SYSTEMS 1
BASES 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM (Continued)
The surveillance requirements provide adequate assurance that RCIC will be OPERABLE when required.
Although all active components are testable and full flow can be demonstrated by recirculation during reactor operation, a complete functional test requires reactor shutdown.
The pump discharge piping is main-tained full to prevent water hammer damage.
3/4. 7. 4 SNUBBERS All snubbers are required OPERABLE to ensure that the structural integrity of the reactor coolant system and all other safety related systems is maintained during and following a seismic or other event initiating dynamic loads.
Snubbers are classified and grouped by design and manufacturer but not by size.
For example, mechanical snubbers utilizing the same design features of the 2-kip, 10-kip, and 100-kip capacity manufactured by Company "A" are of the same type.
The same design mechanical snubbers manufactured by Comoany "B" for the purposes of this Technical Specification would be of a different type, as would hydraulic snubbers from either manufacturer.
A list of individual snubbers with detailed information of snubber location, size and system affected shall be available at the plant.
The accessibility of each snubber shall be determined and reviewed by the Facility Review Group e
and approved by the Manager, Clinton Power Station.
The determination shall be based upon the accessibility of the snubber during plant operations (e.g.,
radiation level, temperature, atmosphere, location, etc.).
The addition or deletion of any hydraulic or mechanical snubber shall be made in accordance with Section 50.59 of 10 CFR Part 50.
When a snubber is found inoperable, an engineering evaluation is performed, in addition to the determination of the snubber mode of failure, in order to deter-mine if any safety-related component or system has been adversely affected by the inoperability of the snubber.
The engineering evaluation shall determine whether or not the snubber mode of failure has imparted a significant effect or degradation on the supported component or system.
A representative sample of the installed snubbers will be functionally tested during plant shutdowns at 18 month intervals.
Observed failures of these sample snubbers will require functional testing of additional units.
To provide furtner assurance of snubber reliability, snuobers are visually inspected at the frequencies recommended in NRC Generic Letter 90-09.
1 Hydraulic snubbers and mechanical snubbers may each be treated as a different entity for the above surveillance programs.
The service life of a snubber is evaluated via manufacturer input and informa-tion through consideration of the snubber service conditions and associated installation and maintenance records, i.e., newly installed snubber, seal re-placed, spring replaced, in high radiation area, in high temperature area, CLINTON - UNIT 1 B 3/4 7-2 Amendment No.61 UAN I 51992