ML20217G944
| ML20217G944 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 08/06/1997 |
| From: | Rainsberry J SOUTHERN CALIFORNIA EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-97-01, GL-97-1, NUDOCS 9708080124 | |
| Download: ML20217G944 (7) | |
Text
kOUllllNN { %lllDR\\l4 EDI. SON C2_,
c An I(H$0V l% Tl RN AllON tl
- t eviqwny August 6, 1997 U. S. Nuclear Regulatory Commission i
Attention: Document Control Desk Washington, D.C.
20555 Gentlemen:
Subject:
Docket Nos. 50-361 and 50-362 120 Day Response to Generic Letter 97-01: " Degradation of Control Rod Drive Mechanism Nozzle and Other Yessel Head Penetrations" San Onofre Nuclear Generating Station Units 2 and 3
References:
- 1) "CEOG Response to NRC Generic Letter 97-01, Degradation of CEDM Nozzle and Other Vessel Closure Head Penetrations, CE NPSD-1085," submitted to the Document Control Desk (NRC) by David F. Pilmer for the Combustion Engineering Owners Group (CE0G) on July 25, 1997.
- 2) Letter to Robert E. Link (Wisconsin Electric Power Company) from Alan G. Hansen (NRC) dated March 9,1994;
Subject:
" Acceptance Criteria for Control Rod Drive Mechanism ~
Penetration Inspections at Point Beach Nuclear Plant, Unit 1"
Subject:
Background and Technical Basis for PWR Reactor Vessel Upper Head Penetration Flaw Acceptance Criteria
)!
- 4) Letter to the Document Control Desk (NRC) from Alex Marion (NUMARC) dated January 31, 1994;
Subject:
" Alloy 600 CRDM Penetrations"
/
-This letter provides the Southern California Edison (Edison)'120 day response to Generic Letter (GL) 97-01 for San Onofre Nuclear Generating Station (SONGS)
Units 2 and 3.
The information requested by Generic Letter 97-01 is provided by this letter.
Edison is-committing to perform inspections of the Unit 3 Vessel Head Penetration (VHP) nozzles during the Unit 3 Cycle 10 refueling
- outage, g eoso g g7 g P
' San Onofre NucleaHMdthting 5tation
~
r O. Box 128 San rictuente, CA 92674 0128 g/g 714 368-7420 Wlff
Docur..ent Control Desk BACKGROUND Primary Water Stress Corrosion Cracking (PWSCC) is known to occur in inconel Alloy 600 in the Pressurized Water Reactor (PWR) environment at operating conditions.
Inconel Alloy 600 was used to fabricate the VHPs at SONGS.
There are 91 Control Rod Drive Mechanism (CF;DM ), 'O In-Core Instrumentation (ICl),
and one Reactor Head Vent penetrations at s.dch of SONGS Unit 2 and SONGS Unit 3.
The information requested by Generic Letter 97-01 is provided as follows:
(
RE00ESTED INFORMATION m
- 1. Regarding inspection activities:
1.1 A description of all inspecticas of CRDM nozzle and other VHPs performed to the date of this generic letter, incluu.ng the results cf these inspections.
Pesnonse:
Edison has not performed Eddy Current (EC) or Ultrasonic Testing (UT) examinations on the VHPs.
Visual inspections of the reactor vessel S :ad have been performed on a refueling outage interval at both SONGS Units 2 and 3 as required by ASME Code Section XI, Table IWB-2500-1.
No leakage has been detected from the reactor vessel head as a result of these inspections.
The Boric Acid Inspection Program at SONGS was revised after issuance of Generic letter 88-05 " Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants." This program was enhanced by the addition of the Reactor Coolant System inconel Nozzle Inspection program and procedure which includes the 91 CRDMs and the 10 ICis.
This inspection involves removing accessible insulation from the head allowing bare metal inspections of the 10 ICIs and 24 of the CRDMs.
The bare metal inspections are performed on the re-lphery of the head, where hillside angles are steepest and susceptibility to PWSCC is predicted to be highest.
The remaining VHPs are insputed with the Reactor Head insulation installed.
The inspection procedure was written in April, 1993 and implemented during the Cycle 7 refueling outages at SONGS 2 and 3.
Irs,x tions of the reactor i
vessel head and accessible VHPs have been performed during each
Document Control Desk refueling outage thereafter.
To date no indication of leakage through a VHP nozzle has been identified.
Boric Acid residue has been noted and verified to be spillage from venting CRDM housings and leaking CRDM housing vent valves.
1.2 If a plan has been developed to periodically inspect the CRDM nozzle and other VHPs:
a.
Provide the schedule for first, and subsequent, inspections of the CRDM nozzle and other VHPs, including the technical basis for this schedule.
Response
An integrated inspection program has been developed by Edison in cooperation with the Combustion Engineering OwnersGroup(CE0G).
Edison will perfcnn an EC examination of the VHPs in Unit 3 during the Cycle 10 refueling outage currently scheduled to occur in the spring of 1999.
Any indications will be characterized with UT examination.
The results of this inspection will be made available to the NRC within six months following the completion of the inspection.
The basis for performing this inspection is the output of the timing model developed by the CEOG. The timin9 model is described in detail in the "CE0G Response
~
to NRC Generic Letter 97-01, Degradation of CEDM Nozzle and other Vessel Closure Head Penetrations, CE NPSD-1085,"
Section 2.4 (response to question 1.4).
The schedules for subsequent inspections at Unit 3 or inspections at Unit 2 have not been determined.
These schedules will be bastd on the results of planned inspections and the development of the CE0G and industry integrated inspection plans.
Edison will follow this issue closely and update the inspection schedules as new informat.an becomes available.
b.
Provide the scope for the CRDM nozzle and Other VHP inspections, including the total number of penetrations (and how many will be inspected), which penetrations have thermal sleeves, which are spares, and which are instrument or other penetrations.
m
l Document Control Desk-.-
Response
The scope of the planned inspection at Unit 3 will be 100%,
including all 91 CRDM,10. ICI, and-the Reactor Head Vent penetrations.
None of the_VHPs at SONGS have thermal sleeves, and there are no spares. The 10 ICI nozzles are located on the periphery of the Reactor Head at SONGS 2 and-3.
Edison intends to use the acceptance criteria as accepted by the NRC in a-letter to Point Beach dated March 9,:1994, Reference 2.
_This acceptance criteria wasithe result of letters submitted to the NRC by NUMARC on.
-July 30, 1993, Reference _3, and January 31, 1994, Reference 4.
1.3 lIf a plan has not been developed to. periodically inspect the CRDM nozzle and other VHPs, provide the analysis that supports why no augmented inspection is nucssary.
Resoonse:
An integrated inspection plan has been developed which includes an inspection-at SONGS Unit 3 and at Millstone 2.
Datails of the
-integrated inspection plan are provided in the "rEOG Resr,onse to.
NRC Gener c t.etter 97-01, Degradation of CEDM N9zzie and Other Vessel Closure Head Penetrations, CE hPSD-1085."
1.4-In light of-the degradation of CRDM nozzle and other VHPs described above, provide the. analysis that supports the selected course of-action as listed-in either-1.2 or 1.3, above.
In particular, provide a description of all
-relevant data and/or tests:used to develop crack initiation I
and cran growth models,'the methods and data used to validate these models, the' plant specific _ inputs to these
. models,- and how. these models-substantiate the -
susceptibility evaluation. Also, if-an integrated-industry 1
-inspection program _is being relied on, provide a detailed description of this program.
.Kesponse:
The timing mo' del developed by the CE0G was used to determine which plants have the lowest estimated time remaining to throughwall lJ
O Document Control Desk cracking.
The basis for this timing model and recent results of running the model are discussed in detail in the "CE0G Response to 4
NRC Generic Letter 97-01, Degradation of CEDM Nozzle and other Vessel Closure Head Penetrations, CE NPSD-1085."
- 2. Provide a description of any resin bead intrusions, as described in IN 96-11, that have exceeded the current EPRI PWR Primary Water Chemistry Guidelines recommendations for primary water sulfate levels, including the following information:
2.1 Were the intrusions cation, anion, or mixed bed?
2.2 What were the durations of these intrusions?
l 2.3 Do the plant's RCS (reactor coolant system) water chemistry Technical Specifications follow the EPRI guidelines?
2.4 Identify any RCS chemistry excursions that exceed the plant administrative limits for the following species: sulfates, chlorides or fluorides, oxygen, boron, and lithium.
2.5 Identify any conductivity excursions which may be indicative of resin intrusions.
Provide a technical assessment of each excursion and any follow-up actions.
2.6 Provide an assessment of the potential for any of these intrusiors to result in a significant increase in the probability for IGA (intergranular attack) of VHPs and any associated plan f]r inspections.
Resoonse:
Southern California Edison (Edison) has reviewed the plant historical records to determine if any incident of rt tin ingress similar to those which occurred in 1980 and 1981 at ti? Jose Cabrera (Zorita) plant has occurred at San Onofre Unit 2 and 3.
This data search was structured to identify all resin ii' trusion events into thr pii,'ary coolant system with a magnitude greater than one cubic < (30 liters).
The threshold of one cubic foot was chosen as a conservative lower bound since it represents less than 15'4 of the estimated volume of resin released into the reactor coolant system during the two events at Jose Cabrera.
1
.J
Document Control Desk For the period of plant operation prior to the routine analysis for sulfate in reactor coolant, the data search was based on a review of the plant's reactor coolant chemistry records relative to specific conductance of the reactor coolant. An elevation of a i
28 micros /cmincrementinspecificconductancewasthevalueused as an indicator of cation resin ingress equivalent to a volume of I cubic foot.
Routine analysis for sulfate in reactor coolant was performed for plant operation from January of 1983 to the present. A sulfate concentration in the range of 15 to 17 ppm pek concentration was used as the indicator of cation resin ingress.
This concentration is approximately equivalent to a cation resin ingress volume of I cubic foot.
Had either specific conductance or sulfate increases indicated resin ingress to the magnitude of the threshold quantity identified above, additional data evaluation would have been conducted to look for a corresponding depression in pH or elevation in lithium or Total Organic Carbon as corroborating information of the incident.
In the case of the use of sulfate data as the indicator, specific conductance would also have been included as confirmatory data had a significant in leakage event been identified.
It is unnecessary to review plant records for boron, chlorides, fluorides, ar.d oxygen because these species are not viewed as valid indicators of cation resin ingress and degradation within the primary coolant system of a PWR.
Borate, chloride, and fluoride anions could be associated with the anion portion of mixed bed resin (cation plus anion); however, if mixed bed resin leakage to the RCS occurred, the cation portion of the resin would contain the sulfate indicator described above.
Detectable dissolved oxygen in reactor coolaret, during power operation with appropriate hydrogen in the reactor coolant, could not occur and therefore, could not be associated with resin inleakage.
]
San Onofre Units 2 and 3 have essentially followed the EPRI PWR Primary Water Chemistry Guidelines since March of 1987 and have implemented revisions when issued.
No resin intrusion event was identified at Units 2 or 3 as a result of the historical chemistry data review performed as a l
Document Control Desk result of this generic letter.
The "CE0G Response to NRC Generic Letter 97-01 Degradation of CEDM Nozzle and other Vessel Closure Head Penetrations, CE NPSD-1085, "provides an assessment of the potential for small resin intrusion events to result in a significant increase in the probability for IGA of VHPs.
Currently scheduled inspections will detect cracks that are sulfate induced in the areas adjacent to the partial penetration attachment welds.
If you have any questions or would like additional information, please let me know.
I Sincerely, l
, fjsd<^ 4' For J. L. Rainsberry Manager, Plant Licensing l
cc:
E. W. Merschoff, Regional Administrator, NRC Region IV K. E. Perkins, Jr., Director, Walnut Creek Field Of fice, NRC Region IV J. A. Sloan, NRC Senior Resident Insprctor, San Onofre Units 2 & 3 M. B. Fields, NRC Project Manager, San Onofre Units 2 and 3 State of California County of San Diego Subscribed and sworn before me this 6N dayof)try,1997.
Notary AWN Lwl M\\_.)
/L____/
Carmemone itaos j
Noetry P20c-Capewes tmpococawer r
~
4 MyComm Egees Mar 23.EDI[
. ---------...,