05000346/LER-2002-002

LER-2002-002,

Docket Number
Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded
3462002002R00 - NRC Website

Description of Occurrence:

On February 16, 2002, the Davis-Besse Nuclear Power Station (DBNPS) began its thirteenth refueling outage (RFO) that included inspection of the Reactor Pressure Vessel [RPV] head Control Rod Drive Mechanism (CRDM) nozzles [NZL] in accordance with NRC Bulletin 2001-01, "Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles." These inspections were conducted to determine if Primary Water Stress Corrosion Cracking (PWSCC) had occurred that could have caused CRDM nozzle cracking in a circumferential direction (i.e., around the circumference of the nozzle). The inspections consisted of a 100% qualified visual inspection of the RPV head at the CRDM nozzle penetration annulus area and 100% ultrasonic (UT) examination of each of the 69 CRDM nozzles. The qualified visual inspection was inconclusive because of the large amount of boric acid crystal deposits on the RPV head. Subsequent bottom-up blade probe UT examination identified outside diameter (OD) initiated axial flaw indications in the Alloy 600 nozzle for CRDM nozzle #3, which is located near the center of the RPV head (location G-9). On February 27, 2002, at 1330 hours0.0154 days <br />0.369 hours <br />0.0022 weeks <br />5.06065e-4 months <br />, evaluation of the indications on nozzle #3 identified a path for Reactor Coolant System (RCS) [AB] pressure boundary leakage.

Technical Specification 3.4.6.2 states the RCS leakage shall be limited to no pressure boundary leakage.

This was reported within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, at 1542 hours0.0178 days <br />0.428 hours <br />0.00255 weeks <br />5.86731e-4 months <br />, via the Emergency Notification System (ENS) as a non- emergency condition in accordance with 10 CFR 50.72(b)(3)(ii)(A), a condition that resulted in the nuclear power plant, including its principal safety barriers, being seriously degraded. Examination by the UT method later identified pressure boundary leakage indications on CRDM nozzles #1 (location H-8) and #2 (location G-7). In addition, a circumferential flaw indication was identified above the J-groove weld on the OD of CRDM nozzle #2. The circumferential indication on this CRDM nozzle was 34 degrees in length but had not penetrated the nozzle thickness (i.e., approximately 50% through-wall).

These indications were confirmed by top-down rotating probe UT examination. The final UT examination results were provided to the NRC via the ENS in an update to the original notification report on March 5, 2002, at 1921 hours0.0222 days <br />0.534 hours <br />0.00318 weeks <br />7.309405e-4 months <br />.

The following table summarizes and characterizes the number of flaws identified during the UT examinations for the three CRDM nozzles:

CIRCUMFERENTIAL

(through wall) (not through wall) (not through wall) 1 2 3 2 6 2 7 2 2 1 While machining CDRM nozzle # 3 during repair activities, the CRDM nozzle exhibited unexpected movement. To identify the cause of the CRDM nozzle movement, an investigation into the condition of the RPV head surrounding CRDM nozzle #3 was initiated. This investigation included removing the CRDM nozzle from the RPV head and removing boric acid deposits from the top of the RPV head. Upon completing the boric acid removal, a visual examination of the area was conducted which identified a cavity in the RPV head adjacent to CRDM nozzle #3. This condition was reported to the NRC via the FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Davis-Besse Unit Number 1 05000346 YEAR Description of Occurrence: (continued) ENS on March 8, 2002, in accordance with 10CFR 50.72(b)(3)(ii)(A) as a condition that resulted in the nuclear power plant, including its principal safety barriers, being seriously degraded. Ultrasonic thickness measurements of the RPV head in the vicinity of CRDM nozzles #1, #2, and #3 were performed. Follow- up characterization by UT examination indicated degradation of the low alloy steel RPV head material adjacent to the CRDM nozzle. The degraded area was found to extend approximately 6.6 inches from the penetration for CRDM nozzle #3, with a width of approximately 4 to 5 inches at its widest part. The remaining thickness of the RPV head in the degraded area was found to be an average of approximately 0.30 inches. These events are being reported in accordance with 10CFR50.73(a)(2)(ii)(A) as a condition that resulted in the nuclear power plant, including its principal safety barriers, being seriously degraded and 10CFR50.73(a)(2)(i)(B) as a condition prohibited by the plant's Technical Specifications.

The investigation also showed a small area of corrosion where CRDM nozzle #2 penetrated the RPV top head surface. The small area of corrosion at the top of nozzle 2 was found to lie directly over the area of corrosion inside the penetration for CRDM nozzle #2. The investigation also showed evidence of a small leak path where CRDM nozzle #1 penetrated the RPV top head surface. These issues are not reportable, but are included herein for information. Further discussion of these CRDM nozzles is provided in the root cause report, "Significant Degradation of the Reactor Pressure Vessel Head", as was submitted to the NRC on April 18, 2002 (DBNPS Serial Letter Number 1-1270).

Apparent Cause of Occurrence:

On March 8, 2002, the DBNPS management assembled a group of FirstEnergy Nuclear Operating Company personnel and industry experts to investigate the RPV head and CRDM nozzle condition to determine the root cause. This effort determined that:

■ The probable cause of the axial through-wall flaws was PWSCC in the CRDM nozzles due to material susceptibility in the presence of a suitable environment. This is consistent with previous industry experience with PWSCC in Alloy 600 components as was evaluated as part of the NRC Generic Letter 97-01, "Degradation of Control Rod Drive Mechanism Nozzle and Other Closure Head Penetrations.

■ The root cause of the degradation of the RPV head is boric acid corrosion resulting from CRDM nozzle leakage over a significant period of time, and its lack of discovery due to inadequate Boric Acid Corrosion Control Program and Inservice Inspection Program implementation regarding the RPV head. The inadequate implementation of the Boric Acid Corrosion Control Program resulted in the leakage from the CRDM nozzle annulus areas not being identified during plant outages, allowed the plant to return to power with boric acid crystal deposits on the RPV head following the plant outages, and not identifying the degradation of the RPV head base metal during the twelfth ■ Apparent Cause of Occurrence: (continued) ■ RFO in 2000. The root cause report, "Significant Degradation of the Reactor Pressure Vessel Head" was submitted to the NRC on April 18, 2002 (DBNPS letter Serial Number 1-1270).

Analysis of Occurrence:

The average total unidentified leakage from the Reactor Coolant System on February 15, 2002 at the end of the 13th operating cycle was approximately 0.20 gallons per minute (gpm). This is less than the 1 gpm limit permitted by Technical Specification 3.4.6.2. Axial flaws of the type discovered in CRDM nozzles #1, #2, and #3 have been evaluated by the industry and the NRC to not be an immediate safety concern. This evaluation is documented in NRC Generic Letter 97-01, "Degradation of Control Rod Drive Mechanism Nozzle and Other Closure Head Penetrations.

However, this significance assessment is dependent and reliant on effective inspection and detection of leakage from the CRDM nozzle penetration annulus area in a timely manner. The lack of timeliness in the discovery of the leakage from the CRDM nozzle penetration annulus areas resulted in degradation of the low-alloy carbon steel RPV head.

A safety significance assessment of the degraded RPV head was submitted to the NRC on April 8, 2002 (DBNPS letter Serial Number 1-1268). This submittal provided a detailed assessment for the degradation of the RPV head. This evaluation determined that in its degraded condition, structural integrity would have been maintained, based on an average clad thickness of 0.297 inches over a conservative area of degradation, to approximately 5600 pounds per square inch. Evaluation of the minimum clad thickness of 0.24 inches over the conservative area of degradation resulted in structural integrity being maintained to approximately 4600 pounds per square inch. Thus, the as-found RPV head would have functioned to maintain structural integrity during anticipated operational occurrences and postulated accidents.

A deterministic safety assessment was performed and concluded that in the unlikely event of RPV head failure considering the as-found degraded condition: a) adequate core cooling could have been established and maintained for the long term, b) the reactor could have been placed and maintained in a safe shutdown condition, and c) the integrity of the containment would not have been compromised.

In addition, a probabilistic safety assessment concluded, per Regulatory Guide 1.174 guidelines, that there was a small increase in core damage frequency and a very small increase in large early release frequency.

Corrective Actions:

An extent of condition walkdown and inspection of the RCS and structures, systems and components inside of containment is being performed during the current refueling outage to evaluate if boric acid leakage or corrosion in susceptible systems may be occurring.

FACILITY NAME (1) PAGE (3) LER NUMBER (6) DOCKET NUMBER (2) Davis-Besse Unit Number 1 05000346 2002 -- 002 -- � 00 Corrective Actions: (continued) The CRDM nozzles #1, #2, and #3 with pressure boundary leakage, as well as CRDM nozzles #5 (location K-7) and CRDM nozzle #47 (location D-12) discovered with axial flaws that did not penetrate the J-groove weld (i.e. not pressure boundary leakage) and the RPV head degraded area will be repaired prior to the DBNPS return to service following the 13th RFO.

A self-assessment of the Boric Acid Corrosion Control Program and Inservice Inspection Program will be performed to evaluate and correct the deficiencies documented in DBNPS letter Serial Number 1-1270. This activity will be completed by July 1, 2002.

Other contributing causes for this event have been identified in the root cause report, "Significant Degradation of the Reactor Pressure Vessel Head" (DBNPS letter Serial Number 1-1270). The resolution of these issues is being administered in accordance with the requirements of Confirmatory Action Letter Number 3-02-001, dated March 13, 2002.

Failure Data:

There have been no previous occurrences of through-wall axial flaws on CRDM nozzles or RPV head degradation at the DBNPS.

Energy Industry Identification System (EIIS) codes are identified in the text as [XX].