ML20141H566
| ML20141H566 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 07/22/1997 |
| From: | Gibson G SOUTHERN CALIFORNIA EDISON CO. |
| To: | Howell A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| References | |
| 50-361-97-15, 50-362-97-15, NUDOCS 9707310108 | |
| Download: ML20141H566 (4) | |
Text
. __. _ _ _ _ _ _ _ _
@,sountt4N CAuroRMA i
e EDISON An EDISON INTERNATIONAL Company i
'l l
July 22,1997 i
i Mr. A. T. Howell ill
- Director, Division of Reactor Safety U. S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, Texas 76011-8064 i
j
Dear Mr. Howell:
l
Subject:
Docket Nos. 50-361 and 50-362 l
Alloy 600 Primary Water Stress Corrosion Cracking i
{
San Onofre Nuclear Generating Station, Units 2 and 3 i
On July 3,1997, the NRC held a primary water stress corrosion cracking (PWSCC) i i
inspection exit meeting (Inspection Report 97-15). The inspector identified a potential -
l violaton of 10 CFR 50, Appendix B, Criterion XVI, " Corrective Action." The inspector j
l indicated that the potential violation was for failure to take corrective actions on Alloy j
600 heat number NX7630 RCS nozzles, after PWSCC was identified on two NX7630 I
hot leg nozzles in 1995. Edison does not believe a violation occurred. This letter provides additional information in this regard. Also, Edison requests a meeting, if necessary, to discuss this additional information.
l Criterion XVI requires conditions adverse to quality be "... promptly identified and corrected." Criterion XVI also requires that for significant conditions adverse to quality
... the cause of the condition be determined and corrective actions taken to preclude repetition." In accordance with Criterion XVI, Alloy 600 PWSCC (see Attachment) was 1
identified and documented on nonconformance reports, and corrected by repair or replacement.
PWSCC, which can result in tiny, slowly developing leakage from the RCS, is not a significant condition adverse to quality, which is supported by the following factors:
1)
The combination of plant specific and industry experience and stress analysis f
shows PWSCC grows axially. Axial cracking is not a significant threat to the (h
'i ncale structural integrity, and v/ill result in a small, detectable leak.
2)
PWSCC axial cracks proceed gradually, and slowly increasing leakage permits plant personnel to detect and react to the leakage.
970731o100 970722
\\
l PDR ADOCK oS00o361-1 P. O. Box 128 San Clemente. CA 92674 0128
j-L A. T. Howell July 22,1997 i
i 3)
PWSCC axial cracks may grow to a length of two inches axially without l_
exhibiting unstable crack growth. A Combustion Engineering owners group 1
report (CEOG Task 700) fracture mechanics analysis showed that even with a i
crack two inches long, at normal steady state temperatures and pressures, there i
is a substantial safety factor (>10) against any additional crack growth due to mechanical means.
4 i
4)
Crack detection occurs well before it can grow to an unstable size. Industry and SONGS experience was consistent with this expectation.
i 5)
Independent industry sources concluded there was no immediate safety concern.
i NRC Information Notice 90-10, " Primary Water Stress Corrosion Cracking e
(PWSCC) of Alloy 600," dated February 23,1990.
NUMARC letter to William T. Russell (NRC) regarding Inconel 600, e
dated June 16,1993.
NRC letter from William T. Russell to William Rasin (NEI), NRC response i
e to NUMARC letter, dated November 19,1993.
NUMARC latter from Alex Marion to the NRC, response to previous letter, e
dated January 31,1994.
EPRI TR-103696, "PWSCC of Alloy 600 Materials in Primary System e
Penetrations," dated July 1994, o
NUREG CR-6245, " Assessment of Pressurized Water Reactor Control Rod Drive Mechanism Cracking," dated October 1994.
In the absence of SONGS, industry, or NRC criteria establishing a threshold for replacing an entire heat of material once a failure occurs, the replacement decision must consider all potential factors (e.g., safety significanct ;f condition, ALARA, risk factors for the execution of work, etc.). For example, from an ALARA perspective, the dose rate associated with each hot leg nozzle replacement is approximately 1.7 person-rem per RCS nozzle, and 35 to 40 person-rem would have been expended to replace the remaining NX7630 nozzles.
Because the NRC and industry agreed there was no safety significance to PWSCC which would justify replacement of nozzles, Edison concluded that replacement of nozzles from heat number NX7630 was not necessary at that time and proceeded to:
- 1) qualify a replacement process in house to minimize plant impact; 2) explore other 6
A. T. Howell July 22,1997 alternate replacement processes; 3) continue the inspect and repair progrom through j
Cycle 9 (current outage); and 4) reevaluate our program plan following the Cycle 9 outages.
The inspect and repair program since 1995 has resulted in the detection of PWSCC at San Onofre well before any significant leakage developed (e.g., the recent 3TWO138A nozzle PWSCC found while coming out of the Cycle 9 refueling outage). Edison is assessing the Cycle 9 outage PWSCC inspection results. The 1997 Nuclear Organization Business Plan initiative BP-4-4-1, dated January 15,1997, states that a revised Inconel Strategic Plan will be issued by October 31,1997. This letter i
constitutes a formal Edison commitment to the NRC to complete this plan by October 31,1997.
If a meeting to further discuss PWSCC with you and your staff would be beneficial, please call me to arrange a meeting. Edison appreciates the opportunity to provide this additional information.
s Sincerely, G. T. Gibson Manager, Compliance cc:
J. A. Sloan, NRC Senior Resident inspector, San Onofre Units 2 & 3 M. B. Fields, NRC Project Manager, San Onofre Units 2 & 3 NRC Document Control Desk t
i M,
+
ATTACHMENT Alloy 600 NX7630 Nozzle History i
l i
in 1992 during the Unit 3 Cycle 6 outage, the first indication of PWSCC e
j associated with heat number NX7630 was identified on a pressurizer upper shell nozzle. Edison concluded that all pressurizer upper shell nozzles were the most l
susceptible to PWSCC, based on the high temperature of the location (653 degrees F). These nozzles were all replaced with a less PWSCC susceptible material (inconel 690).
e in June 1993, during the Unit 2 Cycle 7 outage, a second nozzle associated with l
heat number NX7630 evidenced indications of PWSCC. This was the first hot 4
leg nozzle indication of PWSCC. This nozzle was replaced with inconel 690.
e in July 1995, a second NX7630 PWSCC hot leg leak was identified during the Unit 3 Cycle 8 outage. The nozzle was replaced with an inconel-690 replacement.
In October 1995, another Combustion Engineering nuclear power plant e
experienced PWSCC in one nozzle involving heat number NX7630 (this was the only other industry recorded PWSCC for heat number NX7630).
l Table 1 below represents the total hot leg nozzle population, and nozzle PWSCC by heat number, through the SONGS Cycle 8 outages. Subsequent inspections l
performed during the Cycle 9 outages have identified additional indications of nozzle PWSCC in heat numbers NX7630, NX9915, and 9294.
l Table 1 - Hot Leg Nozzle Heat Numbers Through Cycle 8 i.
l Heat Number NX7630 NX9915 9294 7617-4 K259 7760-4 j
UnR 2 10 12 5
5 f
Unit 3 10 7
9 5
1 Total 20 19 14 5
5 1
PWSCC 1 U2 Cycle 7 1 U3 Cycle 8 1 U3 Cycle 8