ML20217P917
| ML20217P917 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 04/30/1998 |
| From: | Rainsberry J SOUTHERN CALIFORNIA EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-M99558, TAC-M99559, NUDOCS 9805070136 | |
| Download: ML20217P917 (13) | |
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April 30,1998 l
U.S. Nuclear Regulatory Commission Attention:
Document Control Desk Washington, D.C.
20555 Gentlemen:
Subject:
Docket Nos. 50-361 and 50-362 Use of the Mechanical Nozzle Seal Assembly San Onofre Nuclear Generating Station, Units 2 and 3 (TACNos.M99558andM99559)
References:
1)
Letter from William H. Bateman (NRC) to Dwight E. Nunn (SCE),
dated February 17, 1998;
Subject:
Use of the Mechanical i
Nozzle Seal Assembly for the San Onofre Nuclear Generating Station, Units 2 and 3 (TAC Nos. M99558 and M99559) 2)
Letter from J. L, Rainsberry (SCE) to Document Control Desk (NRC), dated December 12, 1997;
Subject:
Docket Nos. 50-361 and 50-362, Mechanical Nozzle Seal Assembly Code Replacement, Request for Relief from 10 CFR 50.55a, San Onofre Nuclear Generating Station, Units 2 and 3 i
This letter provides additional information to support the long-term use of Mechanical Nozzle Seal Assemblies (MNSAs) at the San Onofre Nuclear Generating Station, Units 2 and 3, as requested by the NRC in the February 17, 1998 letter (Reference 1).
BACKGROUND The February 17, 1998, NRC letter (Reference 1) granted permission for the Southern California Edison Company (SCE) to use MNSAs as an alternate j
ASME Section XI Code repair for an interim period (from the Unit 2 and Unit 3 Cycle 9 mid-cycle outages until the next refueling outages).
Additional information was requested to address long-term effects of degradation mechanisms identified in the Safety Evaluation enclosed in the February 17, 1998 letter (Reference 1).
\\l 9805070136 900430 PDR ADOCK 05000361 P
PDR O
\\A San Onofre Nuclear Generating Station P. O. Box I28 I
San Clemente, CA 92674-0128 714-368-7420
. D i
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, Document Control Desk Unit 2 During the Unit 2 mid-cycle outage, six MNSAs were installed on the following nozzles:
Steam Generator Pressurizer Hot Lea 2PDT-0978-1 2LT-110-1 2TW 139-B 2PDT-0978-2 2LT-110-2 2TE-122-4 The hot leg MNSAs will be removed during the Unit 2 Cycle 10 refueling outage, and the nozzles will be replaced with alloy 690 nozzles. With NRC approval, the two Steam Generator and two Pressurizer MNSAs identified above, a total of four MNSAs, will remain installed.
Unit 3 During the Unit 3 mid-cycle outage, one MNSA was installed on the Pressurizer at 3TE-0101. With NRC approval, this MNSA will remain installed and MNSAs will be installed on the Unit 3 Pressurizer at 3LT-110-1 and 3LT-110-2 during the Unit 3 Cycle 10 refueling outage.
The Pressurizer nozzles in both Units 2 and 3 were not leaking prior to installation of the MNSAs.
The MNSAs were installed as a preventive measure should the nozzle leak in the future.
Alloy 690 Nozzle Inspection (Unit 2)
In accordance with the SCE commitment made in the December 12, 1997, letter to the NRC (Reference 2), the Alloy 690 half-nozzle which was installed in June 1993 was removed during the Unit 2 Cycle 9 Midcycle outage to inspect the annulus for corrosion. Minor surface corrosion was observed with small indications of localized pitting.
The deepest pitting was estimated at l
.008" deep.
Due to the limited nature of the observed corrosion, exposure of I
the base metal to the reactor coolant in the configuration of a nozzle repair or MNSA installation is considered acceptable for the remaining life of the plant.
NRC REQUEST In section 5.0 " Future Inspections and Tests" of the NRC Safety Evaluation provided with Reference 1, the NRC staff indicated a need for additional
, Document Control Desk information to determine the acceptability of the MNSAs to perform their l
l intended safety function beyond the next refueling outages for the San Onofre Nuclear Generating Station units due to:
1) the limited testing of these devices with regard to the effect of long-term exposure to the operating environment, and 2) the uncertainty about degradation mechanisms under long-term operating conditions and their effect on some of the components of the MNSAs.
To address these questions, the staff requested a submittal containing an augmented inspection program and additional information addressing potential long-term degradation and corrosion of the MNSAs. The staff suggested that the submittal include, at a minimum, the following information:
Inspection of bolt and MNSA components for corrosion Long-term integrity of the Grafoil gasket seal under Reactor Coolant System (RCS) pressure and thermal operating conditions (no load relaxation and no galvanic corrosion)
Additional accelerated corrosion tests on all of the materials used to fabricate the subcomponents in the MNSA design.
SCE RESPONSE l
Proposed Inspection Program SCE intends to implement an augmented program of inspections and tests to address long-term corrosion and other degradation of the MNSA, including monitoring to address Stress Corrosion Cracking (SCC) issues.
This program is a combination of two different plans:
i 1.
A visual inspection plan 2.
A disassembly inspection plan
V Document Control Desk,
1.
Visual Inspection Plan Visual inspections will include:
a.
ASME VT-1 and VT-2 examinations as required under our ASME Section XI
- program, b.
Additional visual inspections on every installed MNSA, performed 1
during every refueling outage, or during Midcycle outages, at approximately two-year intervals. Any evidence of degradation, including leakage or corrosion, of the installed MNSA or the surrounding area, will be recorded.
Evidence of leakage from the interface of the vessel wall and the MNSA lower flange, or along the axis of the nozzle, would be pressure boundary leakage and would be cause for further investigation and for timely reporting to NRC.
c.
During visual inspect' ors, feeler gauge measurements of the top plate (anti-ejection de d y g3p.
This will determine if relative movement has occurred between the MNSA and the nozzle. This will indicate if the fasteners have moved, or if the nozzle has separated from the
- pipe, d.
As part of visual inspection, inspection of the condition of the locking tab washers and associated fasteners.
Satisfactory condition is an indication that there has been no loss of preload or load relaxation on the seal.
Documentation of the visual inspections will be included in the Section XI Inservice Inspection Summary Report submitted to the NRC after each refueling outage.
2.
Disassembly and Inspection a.
During the Cycle 10 refueling outage, the two temporary hot leg MNSAs will be removed and the component parts will be completely j
disassembled and inspected for degradation and corrosion.
i l
b.
Af ter every five cycles, one MNSA will be selected for complete disassembly and inspection. The purpose of this inspection is to determine if the nozzle ad leaked, and to monitor the long-term condition of the MNSA components and the surrounding area of the vessel for both galvanic corrosion and boric acid corrosion.
The i
, Document Control Desk structural bolting will be examined for evidence of stress corrosion cracking by ultrasonic and liquid penetrant examination methods.
New bolts and Grafoil gaskets will be used for the re-installation.
The old bolts will either be discarded or returned to stock, depending on examination results. Any degradation of the Grafo11 seal or corrosion of the bolts and MNSA components would be cause for further investigation and for notifying the NRC in a timely manner.
c.
For clarification, all the MNSAs installed in both units constitute
)
one inspection pool.
If a Unit 2 MNSA is selected for disassembly during the Cycle 15 refueling outage, no Unit 3 MNSA will be disassembled.
The next MNSA selected for disassembly could be from either unit.
Results of the inspections will be included in the Section XI ISI Summary Report submitted to the NRC after the appropriate refueling outage.
A discussion of the corrosion and Grafoil seal issues that the NRC raised in the SER follows.
We believe this discussion and the inspection program discussed above adequately address the open issues.
Corrosion In the evaluation of the need for additional accelerated corrosio tests of all MNSA materials and materials that the MNSA will contact in service, SCE considered three types of corrosion: 1) galvanic corrosion of the MNSA, pressurizer, and steam generator materials; 2) stress corrosion cracking of the MNSA bolts; and 3) boric acid corrosion of low alloy carbon steel.
Each type of corrosion is evaluated below:
1.
Galvanic Corrosion Galvanic corrosion occurs because of the difference in electrochemical potential (ECP) between the different parts of a cell.(in this case, the MNSA materials, Grafoil seal, Alloy 600 nozzle, and low alloy steel of the pressurizer or steam generator shell) in a conductive solution (electrolyte).
The part with the highest ECP (least noble) will corrode preferentially.
Specific tests to evaluate galvanic corrosion of the MNSA cell have not been conducted, but the low alloy steel will be the most limiting member of this cell and will corrode preferentially.
The major concern centers on the corrosion occurring in the low alloy steel that is in contact with the Grafoil seal. This particular combination of materials is used in other applications
, Document Control Desk where the low alloy (or carbon) steel is frequently inspected (for example, steam generator secondary side manway and hand hole applications), and there is no history of corrosion problems in these applications.
The MNSA application is similar, and for these reasons significant galvanic corrosion is not expected. Galvanic corrosion tests can not be accelerated. However, in tests conducted in simulated reactor coolant (Reference 3-2) with low alloy steel coupled to a more noble corrosion resistant alloy (Type 304 stainless steel) there was not a significant galvanic effect.
These results provide additional confidence that galvanic corrosion will not be a concern.
It should be noted that the Grade GTJ Grafoil used in the MNSA has been treated with ammonium phosphate to inhibit corrosion.
The corrosion protection provided by this inhibitor is comparable to sacrificial inhibitors such as zinc or aluminum. Union Caibide ran a seven-month corrosion test with Grafoil Grade GTJ placed against Grade 420 stainless steel, which is vulnerable to corrosion, in deionized water. A second sample using uninhibited Grafoil was also tested under the same conditions.
For the sample with the GTJ Grafoil, there was minimal visible pitting with a maximum pit depth of.0007 inches.
For the sample with the uninhibited Grafoil, there was considerable pitting with a maximum pit depth of.0053 inches. While the low alloy steel in the pressurizer, etc. may have a somewhat higher ECP than the 420 stainless steel, it is apparent that the GTJ Grafoil significantly reduces the galvanic corrosion process.
It should also be noted that, in the absence of leakage past the Grafoil seal, the annulus will become stagnant and will not allow replenishment of the boric acid or oxygen.
I However, to monitor any possible galvanic effects, Southern California j
Edison's (SCE's) proposed inspection program will involve the periodic disassembly of a MNSA and visual inspection of the MNSA and the shell materials to identify any galvanic corrosion of the low alloy steel material.
SCE proposes to substitute this inspection program for accelerated laboratory testing.
2.
The bolts attaching the MNSA to the pressurizer, steam generator, or main loop piping are SA453 grade 660 (A-286), a precipitation hardened austenitic l
iron-nickel-chromium alloy.
The alloy was developed for high-temperature applications requiring gnd corrosion resistance and high strength.
A-286 is l
typically used where corrosion resistance similar to that of types 304 and 316 stainless steel is required along with higher strength and fatigue resistance.
A-286 has been used for reactor vessel internals bolts, Control Rod Drive l
i
j l
, Document Control Desk Mechanism (CRDM) parts, reactor coolant pump shaf ts and fasteners, and for external bolting applications.
In many applications, A-286 has generally performed satisfactorily, but there have been some stress corrosion cracking failures of A-286 fasteners immersed in primary coolant (Reference 2-1).
These failures have resulted in concerns about potential SCC failures of all A-286 applications.
Stress corrosion cracking requires the simultaneous presence of three elements:
a) a susceptible material condition b) an aggressive environment c) a tensile stress above some threshold level.
A-286 has been proven in the laboratory and in field service to be susceptible to SCC in a Pressurized Water Reactor (PWR) environment when highly stressed (Reference 2-2,2-3,2-4,2-5).
The Babcock and Wilcox investigations subsequent to internal fasteners cracking at several plants (Reference 2-2, 2-3) indicated that bolts that were hot headed as part of the fabrication process had increased susceptibility to SCC. Most of the A-286 field failures have occurred in hot headed bolts.
The bolts for the MNSA application are machined from heat-treated bar stock in the annealed and aged condition and are less susceptible than hot headed bolts to SCC.
Primary coolant is sufficient to cause SCC in highly stressed A-286. The MNSA bolts are external to the reactor coolant system pressure boundary and are not exposed to reactor coolant under normal conditions.
Under these conditions, A-286 bolts have been in service for more than 10 years in high stress level applications without any reported failures (Reference 2-6).
If the MNSA develops a leak, the bolts may be sprayed with a mixture of borated water and steam.
Since the bolts are hot, conditions for wetting and drying will exist, and thus the accumulation of wet deposition on the bolts may occur. Such leakage will be obvious by the buildup of boric acid deposits and would be detected during the refueling outage or midcycle outage inspection.
Therefore, the time the bolts would be exposed to this environment would be no longer than one fuel cycle.
In tests run by Mager and Wilson, 16 of the 17 samples loaded to yield or higher and tested at 650' F in primary coolant survived to 26,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, at which time the test was ended (Reference 2-7).
Only one sample failed, and that was at 19,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, which is longer than the 24-month fuel cycle. Laboratory tests (Reference 2-1) have indicated that A-286 is resistant to SCC at 482*F in highly concentrated (40%) boric acid solutions.
I
, Document Control Desk.
I In summary, testing in PWR environments and concentrated boric acid solutions and service experience indicate that A-286 bolts in the MNSA application will operate indefinitely without SCC failures under normal conditions.
If the MNSA device leaks, the bolts may be exposed to borated water or steam.
The leaking MNSA will be discovered and repaired as part of the Generic Letter 88-05 walkdown inspections'. The worst case service life for the leaking conditions would be one fuel cycle if the leak occurred at startup.
To further demonstrate the adequacy of the bolts, SCE will periodically inspect bolts from a MNSA as described previously. Given the service and 4
laboratory experience and the planned inspections, additional laboratory testing is not necessary.
ThetestingreferencedwasperformedsubsequenttoNUREG/CR-3604 (Reference 2-8) to address concerns raised with A-286 material.
References:
2-1 J. Gorman, " Materials Handbook for Nuclear Plant Pressure Boundary
'a Applications," TR-109668-S1, December 1997 (Draf t).
2-2 G. O. Hayner, " Babcock and Wilcox Experience with Alloy A-286 Reactor Vessel Internal Bolting," Proceedings: 1986 Workshop'on Advanced High-Strength Materials, paper 5 EPRI NP-6363, May 1989.
2-3 R. S. Piasik, " Stress Corrosion Cracking of Alloy A-286 Bolt Material in Simulated PWR Reactor Environment," Proceedings of the Second International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, ANS, p 18-25, 1986.
2-4 D. E. Powell and J. F. Hall, " Stress-Corrosion Cracking of A-286 Stainless Steel in High Temperature Water," Conference: Improved Technology for Critical Bolting Applications, ASME,1986, pp 15-22.
j 2-5 P. J. Plante and J. P. Miedling, " Degradation of Reactor Coolant Pump Impeller-to-Shaf t Bolting in a PWR," Proceedings of the Fourth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, NACE, 1990, pp 11-45 to 11-56.
Document Control Desk 2-6 J. L. Smith, Jr. and G. L. Garner, " Nickel Chromium Iron Alloys for Nuclear Reactor Vessel Components, Volume 1: Identification and Characterization," EPRI NP-5429 SP, Volume 1, October 1997.
2-7 T. R. Hager and I. L. W. Wilson:
Improved Stress Corrosion Resistance of NiCrfe Alloys," EPRI NP-6908-SD, July 1990.
2-8 NUREG/CR-3604 " Bolting Applications," May 1984.
3.
Boric Acid Corrosion If an Alloy 600 nozzle develops a leak, the MNSA Grafoil seal, the OD surface of the nozzle, and the low alloy steel will be exposed to borated water.
If the MNSA does not leak, the solution in the annulus between the nozzle and shell will be a stagnant solution of borated water or steam with concentrations of boric acid typical of the Reactor Coolant System (RCS)
(relatively weak acidic conditions).
In the absence of leakage past the Grafoil seal, there is no mechanism to increase the concentration of the annulus solution.
For the low alloy steels, exposure to the borated solution will result in only minor general corrosion (several mils per year (mpy) maximum), as indicated by laboratory testing and industry experience.
Laboratory data on the corrosion of low alloy steel in dilute borated solutions are limited, but one study (Reference 3-2) indicated a maximum rate of 1.0 mpy (average of 0.6 mpy) at 500*F in a 2000 ppm B solution in a short-term test.
These values are probably conservative since corrosion in such materials generally follows a parabolic rate law with steady state corrosion conditions being approached after a much longer time.
The same test indicated corrosion rates as high as 7.9 mpy at 100*F under aerated conditions which would only be encountered during plant shutdowns. At San Onofre Unit 3, inspection of a nozzle hole that contained an Alloy 690 half nozzle repair (geometry such that the annulus between the nozzle and pressurizer shell was filled with water or steam) showed essentially no corrosion after four years of service. Other PWRs have operated for up to 30 years with breaches of the cladding on the reactor vessel or other components without notable corrosion.
Thus, the available data indicate corrosion of the low alloy steels within the annulus will be minor and of no significance.
If the MNSA leaks, a mixture of borated water and steam will escape the nozzle and may impinge on the A-286 bolts and 304 stainless steel parts. A buildup of boric acid deposits will make such a leak evident during the boric acid walkdown inspections required by GL 88-05, " Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants." Thus, any j
1
, Document Control Desk significant leakage will exist for only one fuel cycle. The tests described in Reference 3-2 included impingement of a borated steam mixture on corrosion resistant materials similar to A-286. After 2500 hours0.0289 days <br />0.694 hours <br />0.00413 weeks <br />9.5125e-4 months <br /> there was no l
observable corrosion confirming the corrosion resistance of this type of materials. However, low alloy steels exposed to aerated, concentrated boric acid may experience significant corrosion (Reference 3-1, 3-3, 3-4, 3-5, 3-6, 3-7).
The available laboratory data does not adequately represent the geometry of a cracked nozzle in a low alloy steel shell. Therefore, the Combustion Engineering Owners Group (CE0G) funded a test program to evaluate low alloy steel corrosion resulting from leakage through a stress corrosion crack near the J-groove weld in a nozzle. The test included blocks of SA 5338 steel heated to 600-650*F, water from a high temperature test loop with nominal primary side chemical conditions, and leakage from laboratory induced stress corrosion cracks in Alloy 600 tubes welded into the blocks.
Post-test examinations indicated high (up to 2.15 inches per year) corrosion rates in localized areas but overall metal loss rates that were relatively low (1.07 cubic inch / year or less).
The maximum metal loss occurred where the leakage exited the annulus with most of the ID surfaces of the low alloy steel having no corrosion.
The CE0G test results were used to develop and justify inspection recommendations. The approach was to calculate the maximum amount (volume) of material that could be removed at pressurizer heater or nozzle holes without violating the ASME Code shell reinforcement requirements. Two adjacent holes were assumed to suffer corrosion damage such that the remaining undamaged j
ligament was at a minimum volume to maintain ASME requirements. The resulting j
volume of material was divided by the maximum observed corrosion rate to determine the time required to violate the reinforcement requirements.
This conservative value was then reduced by 50% to provide additional conservatism.
For the most limiting nozzle configuration in a CE0G plant, the time required 1
was 7.5 years of operation, and for the most limiting heater sleeve i
configuration, the time was 3.2 years or 1175 days. The recommended inspection interval identified in Reference 3-8 was 1100 days maximum.
The boric acid walkdown inspection that is conducted at each refueling outage at SONGS-2 and SONGS-3 will occur at intervals well below the Reference 3-8 recommended inspection interval.
In summary, boric acid corrosion of the materials of construction for the HNSA and the pressurizer and steam generators has been addressed by CE0G and I
industry investigations. Additional testing is not required to address this issue further.
1
, Document Control Desk
References:
3-1 J. F. Hall, " Corrosion of. Low Alloy Steel Fastener Materials Exposed to Borated Water," Proceedings of the Third International Symposium on Environmental Degradation of Engineering Materials in Nuclear Power Systems - Water Reactors, NACE, 1988, pp 711-722.
3-2 WCAP-7099, " Absorption of Corrosion Hydrogen by A302B Steel at 70"F to 500*F," December 1, 1967.
3-3 J. F. Hall, R. S. Frisk, A. S. O'Neill, R. S. Pathania, and W. B. Neff,
" Boric Acid Corrosion of Carbon and Low Alloy Steels," Proceedings of the Fourth International Symposium on Environmental Degradation of Engineering Materials in Nuclear Power Systems - Water Reactors, NACE, 1990, pp 9-33 to 9-50.
T. P. Magee and J. F. Hall, " Corrosion and Corrosion / Erosion Testing of Pressurizer Shell Material Exposed to Borated Water," CE-NPSD-648-P, April 1991.
3-4 EPRI NP-5769, " Degradation and Failure of Bolting in Nuclear Power Plants."
3-5 T. Esselman and P. K. Shah, " Boric Acid Corrosion Evaluation (BACE)
Program Phase 1 - Task 1 Report," EPRI TR-101108, December 1993.
3-6 A. S 0'Neill and J. F. Hall, " Boric Acid Corrosion of Carbon and Low Alloy Steel Pressure Boundary Components in PWRs," EPRI NP-5985, August, 1988.
3-7 J. F. Hall, "A Survey of the LiterdNre on Low Alloy Steel Fastener Corrosion in PWR Power Plants," EPnl NP-3784, December 1984.
3-8
" Evaluation of Pressurizer Penetrations and Evaluation of Corrosion After Unidentified Leakage Develops," CE NPSD-690-P, January 1992.
Long-Term Integrity of the Grafoil Seal.
The MNSA design makes use of Grafoil (a carbon material product of Union Carbide Corporation) to create a seal between three surfaces: 1) the instrument nozzle penetrating the vessel or pipe, 2) the vessel or pipe itself, and 3) the MNSA restraining components.
t
1
, Document Control Desk The MNSA components are designed to create a pocket that the seal fits into when it is in the compressed condition. The seal is completely encapsulated within this pocket.
The components ~ which compress the seal are designed to form a " metal to metal" joint when the components are "made-up."
Constraint of the seal in this manner alleviates a live-load condition where the loading i
in the seal might vary as operating conditions change.
The MNSA Grafoil seal is always contained in this controlled volume.
One major benefit of Grafoil seals is that they are very forgiving of imperfections in the surface they are sealing against.
The seals tend to
" flow" into the surface defects and still make an effective seal.
For the same reason, the seals are not very sensitive to a small loss of preload when they are confined in the " metal to metal" pocket.
In the MNSA design, the 3eals are compressed to the point where the pressure in them is almost double the pressure they are sealing against.
Some loss of the sealing pressure therefore will not cause the joint to leak.
j Grafoil is almost 100% carbon, and therefore is very inert and is not sensitive to either the radiation or thermal environment where MNSAs are installed. Grafoil does not deteriorate under the environmental conditions encountered in a nuclear application as would seals made from an elastomer material.
Load relaxation will be monitored during every Refueling Outage, at approximately two-year intervals.
Verification that no load relaxation has occurred will be accomplished by visual examination to confirm that the locking tabs are still intact.
Properly installed locking tabs prevent the fastener nuts from loosening, thereby ensuring preload is maintained.
I J
In a report previously provided to the NRC, " Flexible Graphite Non-Asbestos Gasketing Material" by P. S. Petrunich, creep relaxation in Grafoil was found to be statistically insignificant, even in Grafoil that had been irradiated by i
170 megarads.
Likewise, for the bolting material A286 Gr 660 creep does not have to be considered below 800* F.
This is what the temperature limits in
{
the ASME Code are based on.
Since the bolting material and the Grafoil are not subject to creep relaxation, and if there is no evidence of movement in the bolts, it can be
)
, Document Control Desk concluded that no load relaxation will occur during service at pressurizer and steam generator operating temperatures.
The information in this letter provides the additional information requested in the February 17, 1998, NRC letter (Reference 1). As stated in the December 12, 1997, SCE letter, SCE is requesting NRC approval by September 30, 1998, to use the MNSAs permanently on the Pressurizer instrument nozzles and the Steam Generator channel head instrument nozzles.
Because of the SCE delay in providing this response, SCE understands that NRC approval may be similarly delayed.
If you have any questions, please call me.
- interely, x8 kowu o
g, cc:
E. W. Merschoff, Regional Administrator, NRC Region IV K. E. Perkins, Jr., Director, Walnut Creek Field Office, NRC Region IV J. A. Sloan, NRC Senior Resident Inspector, San Onofre Units 2 & 3 J. W. Clifford, NRC Project Manager, San Onofre Units 2 and 3 i