ML20147D190
| ML20147D190 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 02/17/1988 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20147D163 | List: |
| References | |
| NUDOCS 8803030256 | |
| Download: ML20147D190 (10) | |
Text
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UNITED STATES i
NUCLEAR REGULATORY COMMISSION e
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SAFETY EVALVATION BY THE OFFICE OF NIICf. EAR REACTOR REGULATION PFLATED TO AMEND.vFNT NO. 40 TO FACILITY OPEDATING LICFf!SE NPF-35 i
AND AVENDMENT NO. 33 TO FACILITY OPERATING LICFNSE NPF-52 DUKE POWER COMPANY ET AL, DOCKET NOS. 50-413 AND 50 414 CATAWBA NUCLEAR STATION, UNITS i ann 2 INTPODUCTION By letter dated July ?2.1987, and supplemented by letters dated Pay ?6, Auoust 31, October 1, October 30, November 19 and December 14, 1987. Duke Power Company, et al., (the licensee) requested amendments to Facility Operating License Nos.
NPF-35 and NPF-5? for the Catawha Nuclear Station, Units 1 and 2.
The proposed amendments vould revise the Technical Specifications due to changes in the reactor trip system and engineered safety #eatures response times to accomodate tha removal of the Desistance Temperature Device (PTD1 bypass system and the installation of replacement RTDs in thermowells located directly in the hot leg and cold leg piping. This sy fastresponseresistancetemperaturedetectors(RTDs)stemwillusenarrowrance This design modification is to overcome ma.4 r drawbacks of the RTO bypass system which lacked reliability 0
fleakage frrn valve packing nr mechanical,4oints1 and resulted in high radiation doses during the per#ormance of maintenance around the RTD bypass system.
The substance of the charges noticed in the Federal Recister on December ?,1987 and the proposed No Significart Hazarcs determination 4
was not affected by the licensee's letter dated December 14, 1987, which clarified certain aspects of the request.
EVALUATION Currently, the hot and cold leg temperatures are reasured by RTDs inserted into reactor coolant bypass loops. A bypass 1000 from upstream of the steam generator to downstream of the steam generator is used for the hot leg RTDs and a bypass loop from downstream of the reactor coolant pump to upstream of the pump is used for the cold leg RTDs.
The RTDs are located in manifolds and are directly inserted into the reactor coolant bypass loop without l
thermowell s.
Each RTD manifold (one hot leg and one cold leg manifold per reactor conlant loop) contains two narrow-range RTOs: one for protection and control system inputs and one as a spare.
Flow into each hot leg h.vpass loop is provided by three scoops located at 120' intervals around the hot lea pipe perimeter to take account of temperature variation across the pion due to t
hot leg streeming. The action of the coolant pump provides well-mixed coolant in the cold leg bypass using a single tap into the cold leg.
8803030256 080217 PDR ADOCK 05000413 P
2 Each loop's pair of RTDs (one in the hot leg and one in the cold leo) is used stem functions based on the average loop to provide inputs for protection sy/2) and the loop differential temperature temperatures (Tavg = (TH0T + TCOLDj (delta T = Th0T -TCOLD). Protection functions based on these inputs are-overtemperature delta T and overpower delta T reactor trips with their asso-ciated (non-Pretection) rod stop and turbine runback actions, Icw Tavg main feedwater isolation, and low-low Tavg (P-12) steam dump block signals.
Each loop's pair of RTDs is also used to provide inputs for control system functions based on the average loop temperature and the loop differential temperature. Control functions based on these inputs are: turbine loading stop from auctioneered low Tavg; rod, steam dump and pressurizer level con-trol from auctioneered high Tavg; rod insertion limit alanns from auctioneered high delta T and Tavg.
In the proposed modified system, the hot leg temperature inputs from each reactor coolant icop will be developed from three fast response, narrow range RTCs mounted in thermowells located within the three existing RTD bypass nanifold scoops (except for Loop B where two of the three thermowells will be mounted in the scoops, but the third thermowell, because of structural interference, will be located 8.5 inches downstream of the existing scoop in an independent boss).
An outlet port is provided at the end of each scoop dnd the thentcwell is positioned so that the RTD sensing element is located near the middle inlet hole of the scoop.
The objective of this design is to ensure that the temperature sensed by the RTD is close to that of the pre-vious scoop flow.
One RTD per loop will be mounted in a thermowell located at the existing penetration for the bypass loop into the cold leg. Additionally, a new penetration will be added to each cold leg for a spare thermowell-mounted, narrow range RTD.
The RTDs are placed in thermowells to allow replacement without draindown. The thermowells, however, increase the response time.
Each hot leg temperature input for protection system functions will be oeveloped by electronically averaging the signals frem the three new fast response, narrow range RTDs.
This averaged input will replace the single input from the currently installed bot leg RTD.
Each cold leg input for protection system functions will be provided by the new fast response, narrow range RTD which replaces the currently installed cold leg RTD.
In the event of a hot leg RTD failure, the electronics allow a bias developed from historical data for the failed RTD to be manually added via a potentiometer to the remaining two RTD signals in order to obtain an average value comparable to the three-RTD average prior to failure of the one RTD.
If a cold leg RTD fails, the spare cold leg RTD can be used instead. The failure of an RTD would be detected by the Tavg or delta T deviation alann.
Inputs for the control system functiors will be provided, thrcugh isolators, from the average loop temperatures and loop differential temperatures calculated by the protection system. This aspect of the design has not been changed; only the use of three hot leg RTDs instead of one per loop to pro.-
vide an average hot leg toeperature is different.
i The RTD modifications affect plant accident analysis by changing the RTD response time and hot leg temperature measurement uncertainty.
In the licensee's July 22, 1987 submittal, the overall response time of the new themowell RTD hot leg temperature measurement system is given as 7.0 seconos, made up of 5.5 seconds for the RTD thermowell combination and 1.5 seconds for the electronic delays. The increase over the 4.0 second response time for the bypass system was principally oue to slow conduction throuch the themowell.
Because of the increased channel response time, there are no longer delays from the time when fluid conditions in the reactor coolant system (RCS) recuire an overtemperature delta T or overpower delta T reactor trip until the trip actually takes place.
However, as reported in the licensee's submittal of July 22, 1987, the original safety analyses for the bypass RTD system conservatively assumed a response time of 8.0 seconds and this response time was found to be acceptable.
In the supplemental submittal of November 19, 1987, the licensee changed the RTD response time from 7.0 to 8.0 seconds. The 8.0 second response time provided one second of added margin in the analyses.
Pecent tectine at another plant after completion of a similar PTD bypass system removal redification has resulted in response times slightly greater than anticipated. Also, as noted in NUREG-0809 (Reference 1), extensive RTD testing has reveeled degradation of RTD response time with aging.
In accordance with the guidance in NUREG-0809, the licensee in its November 19 1987 subnittal revised Technical Specification (TS) 4.3.1.2 to provide for response time testing of all RTDs once per 18 months.
The testing rethod specified is the Loop Current Step Response (LCSR) method, which is the approved in-situ method for measuring RTD response time.
Since the safety analyses referenced in the licensee's July 22, 1987 submittal found that the 8.0 second response time was acceptable, no additieral analyses are required to justify the preposed revision to 8.0 seconds.
With regard to the effect of the plant modification on the uncertainty of the temperature measurements, the new method of measuring each hot leg temperature
'.ith three themowell RTDs manufactured by the RdF Corporation, usted in place of the RTD bypass system with three scoops, has been analyzed to be sligitly more accurate. The new RTD thermowell with measurement at one point ray have a small streaming error relative to the fomer scoop flow measurement because of a temperature gradient over the 5-inch scoop span. However, this gradient has been calculated to have a small effect.
Also, since possible temperature uncertainties from imbalanced scoop flows are eliminated, the overall result is more reliable.
In addition, since the new method uses three RTDs for each hot leg terperature measurement, it is statistically a more accurate temper-ature measurement than the fomer method which used only one RTD for each hot leg temperature measurement.
Therefore, the current values of nominal setroints for the Catawba Technical Specifications are still valid.
There has been no change in the present RTD te.Tperature deviation alams which include both a Tavg and a delta T deviatien alam.
This alam system compares the Tavo er delta T signals to e pre-set threshold value.
This value is nominally set to + or - 2*F and is adjusted during startup and subsequent operation such that it is just beycnd the range of normal operating variations.
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4-The method to be used by the licensee for calibrating the RTDs at each refueling prior to startup is the Westinghouse recomended RTD cross-calibration method at heatups after each refuelino. This procedure requires multiple measurersents at three or four different temperatures.
To date, Westinghouse has evaluated the data frem over 400 RTDs using this technique, and several repeat tests perforced ore to three years apart have not shown any indication of drift in only one direction. The results of the tests indicate that the RTDs drift less than was assumed for uncertainty calculations for the protection system.
The procedure sensitivity is sufficient to discern a random drift of less than 1.0'F by one or several RTDs.
If a drift is noticed, either the calibration of the resis-tance to voltage converter for the affected RTD would be adjusted to account for the shif t, or, if the drift is appreciable, the RTD would be declared in-cperable and would be replaced.
Since both the old and new methods of coolant temperature measurement have an inherent streaming inaccuracy, accounted for in the staff's safety analyses, it is not appropriate to compare the new wethod to the old method and declare any differences as errors.
It is possible, however, to compare the normalized full power delta T measured before and after the modification.
It is expected that the delta T readings will be very similar once any secondary side measurement errors, such as feedwater flow, have been factored into the power calculation.-
If there were ar.y dramatic differences between the two delta T readings it would indicate that a problem existed with one of the measurement metho,ds.
The licensee will perform a comparison of the temperature indications after the modification with measurements prior to the modification.
The NRC will be notified of the results of this comparison including any explanation of variations larger than expected.
Non-LOCA accident analyses are affected by the plant modificatiers primarily through their effect of increasing RTD response time. Only those events which rely on the Overtemperature and Overpower delta T (OTDT and 0FDT) reactor trips are impacted. The accidents in FSAR Sections 15.1 to 15.6 were examined and the following non-LOCA accidents affected by the longer response time were reanalyzed:
(1) the Uncontrolled Rod Cluster Control Assembly (RCCA) With-drawal; (2) uncontrolled boron dilution at power; and (3) the Steamline Rupture at Power.
The applicant stated that the LOFTRAN computer code was used for the analysis of these events.
The first accident, Uncontrolled PCCA Withdrawal, is described in Section 15.4 ?
i of the FSAR.
For this event, the High Neutron Flux and Overtemperature delta T reactor trips are assumed to provide protection against DNB.
This event was analyzed with the increased time constants and lead / lag changes.
Plots of DNBR versus time were provided which showed that the DNBR criterion was met for this accident.
For the Eeron Dilution at Power event, manual operation, as described in Section 15.4.6 of the FSAR, the tine from initiation of the event to reactor trip is deternined from the Uncontrolled RCCS Withdrawal at Power analysis.
The licensee stated that based upon the results of the Uncontrolled RCCS With-drawal at Fower analysis, the conclusions presented in the FSAR for the Boron
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s Dilution at Power event, manual operation, remain valid, i.e., there is greater than 15 minutes from the time of an alann until the total loss of shutdown margin occurs.
For the Steamline Rupture at Power event the analysis included the increased response time and lead / lag changes.
The analysis showed that the desion basis as described in VCAP-9726-Rev.1, "Reactor Core Response to Excessive secondary Steam Pelease" January 1978, has been net.
The effect of the increase in RTD response time on the FSAR Chapter 15 Loss of load / Turbine Trip event is analyzed for both beginning and end of life cerdi-tions in Section 15.2.3 of the FSt.R.
ho credit for reactor trip on turbine trip is assumed in the safety analyses. Therefore, reactor trips on high pressuri7er pressure, overtemperature delta T, and low-low steam generator water level reactor trips provide the recessary protection for this event during the starting mode. For the Loss of Load / Turbine Trip analyses presented in the FSAR, increased RTD response times were assumed for the Catawba positive moderator temperature coefficient (MTC) safety evaluation.
For all four cases analyzed, reactor trips occurred on either a high pressurizer pressure or low-low steam generator water level signal. An overtemperature delta T signal was never generated prior to reactor trip. Therefore, the analyses currently presented in the Catawba FSAR, based on the positive MTC safety evaluation, have adequately addressed the incieased RTD response time resulting from the RTO bypass elimination.
The impact of the increased RTD response time en the FSAR Chapter 15 non-LOCA accident analyses has been evaluated.
For the events impacted, it was demon-strated that the conclusions presented in the FSAR remain valid.
The eliminatien of the RTD bypass systen impacts the uncertainties associated witt. RCS temperature and flow measurement.
The effect Lf these uncertaintiet on the LOCA evaluation was considered.
The magnitudes of the uncertainties in the RCS inlet and outlet temperatures, thermal design flow rate and the steam generator performance data used in the LOCA analyses are such that the conclusions of the previous analyses will not be affected.
Past sensitivity studies concluded that the inlet temperatJre effect on peak Clad temperature is dependent on break size. As a result of these studies, the LCCA analyses are performed at a nominal value of the inlet temperature without consideration of small uncertainties. The RCS flow rate and steam generator secondary side temperature and pressure are also determined using the loop average temperature (Tavg) output. These nominal values used as inputs to tha analyses are not affected by the RTD bypass elimination.
It is corcluded that the elimination of the RTD bypass piping will not affect the LOCA analyses input and hence, the results of the analyses remain unaffected.
Therefore, the plant design chances due to the RTD bypass elimination are acceptable from a LOCA analysis standpoint without reouiring any reanalysis.
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The RCS flow measurement uncertainty af ter the RTD bypass removal modifications i
was analyzed using the methodology in WCAP-11169 Rev. 1 "RCS Flow Uncertainty for t
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Shearon Harris Unit 1," October 1986.
This analysis used the plant-specific instrumentation for the Catawba Plant.
The results of the analysis indicated i
that the flow measurement uncertainty ws reduced from the current value of 2.1% (not including a 0.1% penalty for feedwater venturf fouling allowance) to 1
a new value of tl.7% (including the cold leg elbow taps and excluding feedwater venteri fouling). Much of this reduced uncertainty ic from the statistical i
i advantage of using three RTDs for the hot leg temperature measurement in the new method rather than the one in the fonner method.
The licensee has chosen hot to request any plant specification changes to take advantage of the reduced flow uncertainty, Since the 2.1% allowance is conservative, its retention is acceptable, The staff's review and evaluation of the plant's instrumentation and controls is based upon Sections 7.2 and 7.3 of the SRP, Those sections state that the objectives of the review are to confinn that the reactor trip and engineered i
f safety features actuation system satisfy the requirements of the acceptance
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criteria and guidelines applicable to the protection system and will perfonn j
their safety function during all plant conditions for which they are required.
I Since the staff's review indicates that the modified system does not functionally i
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change the reactor trio and engineered safety features actuation systems (except i
three hot leg RTDs are utilized instead of just one), the staff's original t
i evaluation conclusions for these systems, as documented in Section 7 of the SER for Catawba Units 1 and 2 (NUREG-0954), remain valid.
Based on this and the 1
licensee's statement that the new hardwara for the RTD bypass elimination has been qualified to WCAP-8587, "Methodology for Qualifying Westinghouse WRD j
Supplied NSSS Safety Related Electrical Equiprent," the staff finds the plant modifications to sliminate the R1D bypass manifold and to install fast response RTDs directly in the reactor coolant system hot and cold legs to be acceptable.
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As a result of the plant modifications and new instrumentation associated with the renoval of the existing RTD bypass manifold and replacement by fast response i
j RTDs, the following changes to the plant's Technical Specifications were i
j proposed:
4 Change 1 -
Include new additional entries for the Total Allowance, Z and Sensor Error for Fungtional Uni Table 2.2-1of"(8.9)""(5.41))"and"(2.65)"respectively.7 Overtempergture d j
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Change 2 - Include new, additional entries for the Total Allowance, I and d
Sensor Error fgr Functiongl Unit 8. Oveppower delta T. in Table 2.2-1 of "(4.9 )," "(1.24 ) " and "(1.7 )" respectively.
Change 3 -
Include new additional entries for Z and Allowable Value for 1
FunctignalUnit12.PgoctorCoolantFlow-low,inTable2.2-1of
"(1.41 )" and "(88.8% )" respectively.
Change 4 - On page 2-4 add a new footnote "# Applicable upon deletion of 1
the RTO Bypass System."
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i in NOTE 1 to Table 2.2-1of"(12),""(1.33)"and,"K Include new additiopal entries for t pndt4 Change 5 -
(h2)"respectively.
p Change 6 - Include a new, aoditional entry, "(3.0t )." for the allowable value for overtemperature delta T contained in NOTE 2 to j
Table 2.2-1.
Change 7 - Include a new, additional entry, "(2.6% )," for the allowable value for overpower delta T contained in NOTE 4 to Table 2.2 1.
Change 8 - On page 2-10 add a new footnote "# Applicable upon deletion of RYO Rypass System."
Chance 9 - On page B 2-5 under "Overtemperature delta T," add "(1) (with the RTD Bypass System installed)" to the first sentence between "to" and "piping." Also add "or (2) (with the RTD Bypass System i
removed) thenral delays associated with the RTDs mounted in therrcwells (about 5 ceconds)," before "and pressure" in the first sentence.
Change 10 - On page B 2-5 under "Overpower delta T," 46d "either" to the second sentence between "for" and "piping." Also add "(with the RTD Bypass System installed), or instrumentation delay associated with the loop temperature detectors (with the RTD Bypass System removed)," between "detectors" and "to" in the second sentence.
Change 11 - On page 3/4 3-1 add a sentence stating that:
"The response time of RTDs associated with the Reactor Trip System shall be demonstrated to be within their limits at least once per 18 months."
Change 12 - Include a new, additioral entry, "(B )." for the response times for Functional Unit 7, Overtemperature delta T, and Functional Unit 8. Overpower delta T. in Table 3.3-2.
1 Change 13 - On page 3/4 3-7 add a new footnote "# Applicable upon deletion of RTO Bypass System.
Change 14 - Include new, additional entries in Table 3.3-4 for the Total Allowance, Z, Sensor Error agd A110wably)Value fog) Functional
" "(0.8 " and Unit 5.g},"Tavg-Low,of"(6.0),""(0.71
"(561"F respectively.
Change 15-Includeanew,additionalentryinTable3.3-4fortheAllowab)e Value for functional Unit 18.c Low-Low Tavg, P-12, of "(550"F )."
Change 16 - On page 3/4 3-36 add a new footnote "# Applicable upon deletion of RTO Bypass System."
Changes 1, 2, 3, 5, 6, 7, 14, and 15 above are new values based on revised instrumentation uncertainties resulting from the bypass manifold elimination.
Thesi new values were calculated using essentially the Westinghouse setpoint methadology as previously approved by the staff for Catawba and for generic 1
use (NUREG-0717 SER for Virgil C. Summer Nuclear Station) as documented in the 1
, licensee's letter dated July 22, 1987 The staff finds these changes acceptable.
Change 11 provides an additional surveillance to verify that the RTDs associated with the Reactor Trip System remain within their limits. On the basis that this change would ensure RTD operability, the staff finds it acceptable.
Changes a, 8, 9,10,13, and 16 are editorial changes necessary to acccenodate the removal of the RTD bypass manifold and the situation where removal of the bypass manifold has been ccepleted on only one of the two units. On the basis that these changes add clarity and conciseness to the plant's technical specifications, the staff finds them acceptable.
Change 12 is acceptable because an RTD response time of 8.0 seconds has been found to be acceptable in previous safety analyses.
The statf has reviewed the fabrication and inspection methods described in WCAP-11308 Pev. 2. RTD Bypass Elimination Report for Catawba Units 1 and ?,
Septerber 1987 for the replacement of the RTD bypass system with the new RTD themowell system. This change requires modifications to the het leg piping, the hot leg scoops, the crossover leg bypass return nozzle, the cold leg piping and the cold leg bypass manifold connection. The new thermowells, caps and penetrations will be fabricated in accordance with the ASNE Code,Section III, Class 1.
The welding will be by approved procedures and in-spected by penetrant testing per the ASME Code Section XI.
In accordance with Article IKA-4000 of Section XI, a hydrostatic test of the new pressure boundary welds will be carried out.
The staff finds that the mechanical safety of the proposed PTD themowells system fabricated, examined and tested as described above is acceptable.
The licensee has estimated the occupational radiation exposure for the RTD bypass modification in the submittal of October 30, 1987. The estimate is based on anticipated stay times for each major subtask and estimated dose rates. The annual estimates per unit are given in the table below.
Manhour Dose Estimate Subtask Estimate (Person-Rem)
(1) Preparation for RTD bypass 33 1.0 modification (2) Shielding Installation /
64 9.6 Removal (3) Renove/ Replace pipes, 417.5 10.4 hangers, electrical interferences, etc...
(4) Modify the RTDs 100 12.0 Total per loop ET U M
Total per unit 2458 132.0
.(4 loops) man-hours person-rem l
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The dose avoided through reduced maintenance and operational requirements is the order of 50 to 100 person-rem per unit per year. Comparing this to the total ore-time dose of 132 person-rem for the RTO replactment operation, a net savings of several thousand person-rem over plant life can be proieeted.
An estimate of the curies of beta and gama radioactivity contained on the RTO cenponents to be removed (piping, insulation, hangers, rupture restraints, valves and instrumentatien) is 5.26 curies per unit.
The expected total volume of this radwaste is 574 cubic feet.
Based on the above and on the licensee's radiation protection and ALARA programs previously evaluated and found to be acceptable in Chapter 12 of the SER, the staff concludes that the RTO bypass removal is acceptable from the radiological viewpoint.
ENVIRONMENTAL CONSIDERATION These arenaments involve changes to the installation or use of facility com-ponents located within the restricted area as defined in 10 CFR Part 20.
The staff has determined that the amendrents involve no significant increase in the amounts, and no sienificant change in the types, of any effluents that may be released offsite and that there is no significant increase in individual or currulative occupational exposures.
The NRC staff has made a detemination that the amendments involve no significant hazards consideration, and there has been no public corrinent on such finding. Accordingly, the mendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact staterent or environmental assessnent need be prepared in connection with the issuance of these arendments.
REFERENCES (1) NUREG-0809, Safety Evaluation Pepe t, Review of Resistance Temperature Detector Time Response Characteristics. August 1981.
(2) NUREG/CR -4928, Degradation of Nuclear Plant Temperature Sensors, June 1987 (3)
K. R. Carr, An Evaluation of Industrial Platinum Resistance Thermometer Temperature - Its Measurerent and Control in Scierce and industry ISA publication, Vol. 4. Part 2, 1972, pages 971-982 (4)
- 8. W. Mangum, the Stability of Small Industrial Platinum Resistance Themometers, Journal of Pesearch of the NBS, Vol. 89, No. 4. July-August 1984, Pages 305-350.
CONCLUSION The Comnission made a proposed detemination that the amenoments involve no sionificant hazards consideration which was published in the Federal Peoister (52 FR 45885) on December 2, 1987 The Ctanission consulted with the state of South Carolina. No public corrents were received, and the state of South Carolina did not have any comments.
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I We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations, and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
(
Principal Contributors:
S. Kirslis, PD!!-3/DRP!/II K. Jabbour, PDII-3/DRP!/II F. Burrows, DEST /SICR H. Balukjian, DEST /SRXB l
Dated:
February 17, 1988 i
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February 17, 1988 J
' AMENDMENT NO.40 TO FACILITY OPERATING LICENSE NPF Catawba Nuclear Station, Unit l'.I AMENDMENT NO. 33 'TO. FACILITY OPERATING LICENSE NPF Catawba Nuclear Station, Unit 2 - l DISTRIBUTION:
6'413/41f NRC PDR Local PDR PRC System NSIC l
PDII-3 Reading M. Rood K. Jabbour S. Kirslis OGC-Bethesda D. Hagan J. Stone E. Jordan L. Harmon W. Jones T. Barnhart (8)
ACRS (10)
GPA/PA ARM /LFMB L. Reyes S. Varga/G. Lainas E. Butcher A
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