ML022750524
| ML022750524 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 10/02/2002 |
| From: | Grobe J NRC/RGN-III |
| To: | Myers L FirstEnergy Nuclear Operating Co |
| References | |
| FOIA/PA-2005-0261 IR-02-008 | |
| Download: ML022750524 (36) | |
See also: IR 05000346/2002008
Text
October 2, 2002
Mr. Lew W. Myers
Chief Operating Officer
FirstEnergy Nuclear Operating Company
Davis-Besse Nuclear Power Station
5501 North State Route 2
Oak Harbor, OH 43449-9760
SUBJECT:
DAVIS-BESSE NUCLEAR POWER STATION
NRC AUGMENTED INSPECTION TEAM FOLLOW-UP SPECIAL
INSPECTION REPORT NO. 50-346/02-08(DRS)
Dear Mr. Myers:
On March 12, 2002, the USNRC dispatched an Augmented Inspection Team (AIT) to the
Davis-Besse site in accordance with USNRC Management Directive 8.3, USNRC Incident
Investigation Program. The AIT was chartered to determine the facts and circumstances
related to the significant degradation of the reactor vessel head pressure boundary material.
The AIT developed a sequence of events, interviewed plant personnel, collected and analyzed
factual information relevant to the degraded condition and conducted visual inspections of the
reactor vessel head. The AIT results were summarized for you and your staff during a public
exit meeting on April 5, 2002, and the AIT report was issued on May 3, 2002.
On May 15, 2002, USNRC began a special inspection focused on compliance with USNRC
rules and regulations as they relate to the facts and circumstances associated with the
degradation of the reactor pressure vessel head documented in the AIT report. On August 9,
2002, the USNRC completed this special inspection. The enclosed report documents the
inspection findings which were discussed with you and other members of your staff on August
9, 2002.
Based on this special inspection, ten findings, some apparent violations with multiple examples,
were identified and are documented in the enclosed report. Those findings include: operating
the reactor with prohibited pressure boundary leakage; failure to take effective action to correct
multiple identified safety concerns; inadequacies in the boric acid corrosion control procedure;
failure to effectively implement the boric acid corrosion control procedure and the corrective
action procedure; and multiple examples of inaccurate or incomplete information in letters to the
USNRC or records required by the USNRC to be maintained onsite. Because the USNRCs
determination of the safety significance of the reactor vessel head degradation has not been
finalized and several of these apparent violations remain under review by the USNRC, all of
these findings are currently characterized as unresolved items in the enclosed report.
L. Myers
-2-
In accordance with 10 CFR Part 2.790 of the USNRC's Rules of Practice, a copy of this letter
and its enclosure will be available electronically for public inspection in the USNRC Public
Document Room or from the Publicly Available Records (PARS) component of USNRC's
document system (ADAMS). ADAMS is accessible from the USNRC Web site at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
John A. Grobe, Chairman
Davis-Besse Oversight Panel
Docket No.
50-346
License No.
Enclosure:
USNRC Inspection Report
No. 50-346/02-08(DRS)
cc w/encl:
B. Saunders, President - FENOC
Plant Manager
Manager - Regulatory Affairs
M. OReilly, FirstEnergy
Ohio State Liaison Officer
R. Owen, Ohio Department of Health
Public Utilities Commission of Ohio
President, Board of County Commissioners
Of Lucas County
President, Ottawa County Board of Commissioners
D. Lochbaum, Union of Concerned Scientists
L. Myers
-2-
In accordance with 10 CFR Part 2.790 of the USNRC's Rules of Practice, a copy of this letter
and its enclosure will be available electronically for public inspection in the USNRC Public
Document Room or from the Publicly Available Records (PARS) component of USNRC's
document system (ADAMS). ADAMS is accessible from the USNRC Web site at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
John A. Grobe, Chairman
Davis-Besse Oversight Panel
Docket No.
50-346
License No.
NPF
Enclosure:
USNRC Inspection Report
No. 50-346/02-08(DRS)
cc w/encl:
B. Saunders, President - FENOC
Plant Manager
Manager - Regulatory Affairs
M. OReilly, FirstEnergy
Ohio State Liaison Officer
R. Owen, Ohio Department of Health
Public Utilities Commission of Ohio
President, Board of County Commissioners
Of Lucas County
President, Ottawa County Board of Commissioners
D. Lochbaum, Union of Concerned Scientists
DOCUMENT NAME: G:DRS\\ML022750524.wpd
To receive a copy of this document, indicate in the box: "C" = Copy without attachment/enclosure "E" = Copy with attachment/enclosure "N" = No copy
OFFICE
RIII
RIII
RIII
RIII
NAME
MFarber:sd
ZFalevits
JGrobe for CLipa
RPaul
DATE
10/01/02
10/01/02
10/01/02
10/01/02
OFFICE
RIII
RIII
NAME
CWeil for BClayton
JGrobe
DATE
10/02/02
10/02/02
OFFICIAL RECORD COPY
L. Myers
-3-
ADAMS Distribution:
Chairman Meserve
Commissioner Dicus
Commissioner Diaz
Commissioner McGaffigan
Commissioner Merrifield
W. Travers, EDO
W. Kane, OEDO
S. Rosenberg, OEDO
B. Bonser, OEDO
OCIO
S. Collins, NRR
B. Sheron, NRR
J. Zwolinski, NRR
R. Barrett, NRR
G. Holahan, NRR
B. Bateman, NRR
S. Bajwa, NRR
S. Long, NRR
F. Congel, OE
H. Miller, RI
L. Reyes, RII
E. Merschoff, RIV
AJM
SPS1
RidsNrrDipmIipb
GEG
CST1
C. Ariano (hard copy)
DRPIII
DRSIII
PLB1
JRK1
DB0350
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket No:
50-346
License No:
Report No:
50-346/02-08
Licensee:
FirstEnergy Nuclear Operating Company
Facility:
Davis-Besse Nuclear Power Station
Location:
5501 North State Route 2
Oak Harbor, OH 43449
Dates:
May 15 through August 9, 2002
Inspectors:
J. Gavula, Senior Reactor Inspector
M. Farber, Senior Reactor Inspector
J. Jacobson, Senior Mechanical Engineer
Approved by:
John A. Grobe, Chairman
Davis-Besse Oversight Panel
ii
SUMMARY OF FINDINGS
IR 05000346-02-08, on 05/15-08/09/2002, FirstEnergy Nuclear Operating Company,
Davis-Besse Nuclear Power Station. Augmented Inspection Team Follow-up Special
Inspection.
The report covers a special inspection, by three regional inspectors, that focused on
compliance with USNRC rules and regulations as they relate to the facts and circumstances
associated with the degradation of the reactor pressure vessel head. The USNRCs program
for overseeing the safe operation of commercial nuclear power reactors is described in
NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.
A.
Inspector Identified Findings
Cornerstones: Initiating Events, Barrier Integrity
Significance to be Determined (TBD). The inspectors identified an apparent
violation of Technical Specification Limiting Condition for Operation for Reactor
Coolant System Operational Leakage, paragraph 3.4.6.2, for operation of the
plant with pressure boundary leakage from through-wall cracks in the reactor
coolant system.
This finding is more than minor because the pressure boundary leakage and
resultant cavity in the reactor vessel head represented a loss of the design basis
barrier integrity. The significance of this finding will be determined by the
Significance Determination Process for the issue, which was begun following the
Augmented Inspection Team activities (Section 4OA3.b.1).
Significance to be Determined (TBD). The inspectors identified an apparent
violation involving failure to take adequate corrective action for a continuing
buildup of boric acid deposits on the reactor head.
This finding is more than minor because the corrosion of the reactor head and
the resulting cavity represented a significant loss of the design basis barrier
integrity. The significance of this finding will be determined by the Significance
Determination Process for the issue, which was begun following the Augmented
Inspection Team activities (Section 4OA3.b.2.1).
Significance to be Determined (TBD). The inspectors identified an apparent
violation involving failure to take adequate corrective action for recurrent
accumulations of boric acid on containment air cooler (CAC) fins. These
accumulations resulted in reduced heat removal capability and reduced air flow
through the cooler which was indicated by decreasing plenum pressure.
This finding is more than minor because the corrosion of the reactor head and
the resulting cavity represented a significant loss of the design basis barrier
integrity. The significance of this finding will be determined by the Significance
iii
Determination Process for the issue, which was begun following the Augmented
Inspection Team activities (Section 4OA3.b.2.2).
Significance to be Determined (TBD). The inspectors identified an apparent
violation involving failure to take adequate corrective action for repeated clogging
of radiation element filters although a sample of the filter deposits revealed iron
oxides, radionuclides, and primary chemistry.
This finding is more than minor because the corrosion of the reactor head and
the resulting cavity represented a significant loss of the design basis barrier
integrity. The significance of this finding will be determined by the Significance
Determination Process for the issue, which was begun following the Augmented
Inspection Team activities (Section 4OA3.b.2.3).
Significance to be Determined (TBD). The inspectors identified an apparent
violation involving the failure to follow the corrective action procedure and take
timely corrective action for a condition adverse to quality, in that the licensee
failed to implement a modification to permit complete inspection and cleaning of
the reactor vessel head and CRDM nozzles.
This finding is more than minor because the corrosion of the reactor head and
the resulting cavity represented a significant loss of the design basis barrier
integrity. The significance of this finding will be determined by the Significance
Determination Process for the issue, which was begun following the Augmented
Inspection Team activities (Section 4OA3.b.2.4).
Significance to be Determined (TBD). The inspectors identified a finding
involving failure to complete an identified corrective action for an adverse trend in
This finding is more than minor because the corrosion of the reactor head and
the resulting cavity represented a significant loss of the design basis barrier
integrity. The significance of this finding will be determined by the Significance
Determination Process for the issue, which was begun following the Augmented
Inspection Team activities (Section 4OA3.b.2.5).
Significance to be Determined (TBD). The inspectors identified an apparent
violation involving deficiencies in the licensees Boric Acid Corrosion Control
procedure, NG-EN-00324.
This finding is more than minor because the corrosion of the reactor head and
the resulting cavity represented a significant loss of the design basis barrier
integrity. The significance of this finding will be determined by the Significance
Determination Process for the issue, which was begun following the Augmented
Inspection Team activities (Section 4OA3.b.3.1).
Significance to be Determined (TBD). The inspectors identified an apparent
violation involving multiple examples of failure to follow the boric acid corrosion
control procedure.
iv
This finding is more than minor because the corrosion of the reactor head and
the resulting cavity represented a significant loss of the design basis barrier
integrity. The significance of this finding will be determined by the Significance
Determination Process for the issue, which was begun following the Augmented
Inspection Team activities (Section 4OA3.b.3.2).
Significance to be Determined (TBD). The inspectors identified an apparent
violation involving two examples of failure to follow the stations corrective action
program procedure.
This finding is more than minor because the corrosion of the reactor head and
the resulting cavity represented a significant loss of the design basis barrier
integrity. The significance of this finding will be determined by the Significance
Determination Process for the issue, which was begun following the Augmented
Inspection Team activities (Section 4OA3.b.3.3).
Significance to be Determined (TBD). The inspectors identified an apparent
violation of 10 CFR 50.9 involving multiple examples of information provided to
the Commission or required by the Commissions regulations to be maintained
by the licensee that were not complete and accurate.
Completeness and accuracy in the documents associated with this issue would
have provided an earlier alert to licensee staff and the USNRC about the
problems with control rod drive mechanism nozzle leakage or may have caused
the USNRC to establish a different regulatory position concerning the urgency of
inspections for the reactor pressure vessel head. The significance of this
apparent violation requires additional review as specified in NUREG-1600,
General Statement of Policy and Procedures for USNRC Enforcement (Section
4OA3.b.4).
B.
Licensee Identified Findings
None
1
Report Details
4.
OTHER ACTIVITIES (OA)
4OA3 Event Follow-up (93812)
Background
On March 6, 2002, Davis-Besse personnel notified the USNRC that reactor vessel head
material, adjacent to a control rod drive nozzle, was significantly degraded. An
Augmented Inspection Team (AIT) was chartered and dispatched to the site on
March 12, 2002, to determine the facts and circumstances related to the reactor vessel
head pressure boundary material degradation, and to identify any precursor indications
of this condition. In accordance with USNRC procedures, the AIT charter did not
include the verification of compliance with USNRC rules and regulations, nor the
recommendation of enforcement actions. The AIT concluded its inspection on April 5,
2002, and issued USNRC Inspection Report 50-346/02-03 on May 3, 2002.
a.
Inspection Scope
The purpose of this current inspection effort was to characterize any regulatory issues
revealed during the AITs activities. The inspection scope included a review of the AIT
report, and also encompassed further reviews of licensee activities related to technical
specification and procedural adequacy and compliance, and corrective action adequacy.
This inspection was based on the facts and circumstances discussed in the AIT report
and will not replicate chronologies or technical analyses unless needed to establish
regulatory basis.
In addition to the information in the AIT report, subsequent questions were raised
regarding completeness and accuracy of documents either required by the USNRC to
be maintained by the licensee or submitted to the USNRC. Consequently, as licensee
documents associated with this issue were reviewed for regulatory compliance,
they were concurrently reviewed for completeness and accuracy. Because the risk
significance of the reactor vessel head degradation has not been finalized, and several
of these apparent violations remain under review by the USNRC, all of the findings will
be classified as Unresolved Items in accordance with Manual Chapter 0612, Power
Reactor Inspection Reports.
b.
Findings
b1.
Technical Specification Reactor Coolant System (RCS) Operational Leakage
a.
Introduction
The inspection identified an apparent violation, whose significance is yet to be
determined, involving the Davis-Besse technical specification associated with operation
of the plant with pressure boundary leakage from through-wall cracks in the RCS. This
2
finding had a credible impact on safety and was characterized as an unresolved item
(URI) pending USNRC determination of the significance of the apparent violation.
b.
Description
On February 27 and March 5, 2002, the licensee notified the USNRC that their
evaluation of ultrasonic test data, for axial indications in control rod drive mechanism
nozzles Nos. 1, 2, and 3, confirmed that there was reactor pressure boundary leakage.
Further investigation revealed a cavity adjacent to control rod drive penetration nozzle
No. 3, approximately 5 to 7 inches long and 4 to 5 inches wide. Within this area, the
6.63 inch- thick low-alloy steel head had been corroded away, leaving only the stainless
steel cladding layer on the inside of the reactor vessel head. Based on the length of the
cracks, the amount of boric acid accumulation on the reactor vessel head, and the
extensive corrosion of the reactor vessel head, it is clear that the unit operated well in
excess of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> with pressure boundary leakage.
At the time of plant shutdown, the unidentified primary coolant system leak rate was
approximately 0.2 gpm, within the Technical Specification 3.4.6.2.b limit of 1.0 gpm.
However, Technical Specification 3.4.6.2.a requires that primary coolant operational
leakage shall be limited to No PRESSURE BOUNDARY LEAKAGE when in Modes 1-
4. The associated action requires that the plant be placed in Hot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
and in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Although the time the pressure
boundary leakage began could not be precisely determined, it is clear that the leakage
existed greater than the time frame that would have required plant shutdown.
c.
Analysis
This issue represented a licensee performance deficiency because the licensee had
multiple opportunities over a period of years to identify the leakage; consequently it was
considered a finding. This finding is of more than minor safety significance because the
pressure boundary leakage and resultant cavity in the reactor vessel head represented
a loss of the design basis barrier integrity. Two cornerstones were impacted by this
issue. The Barrier Integrity cornerstone was affected because the through-wall CRDM
cracks compromised the reactor coolant pressure boundary and the Initiating Events
cornerstone was impacted because cracking of the CRDM nozzles resulted in an
increase in the likelihood of a loss of coolant accident (LOCA). The significance of this
finding will be determined by the Significance Determination Process (SDP) for the
issue, which was begun following the AIT activities.
d.
Enforcement
Davis-Besse Technical Specification, Limiting Condition for Operation for Reactor
Coolant System Operational Leakage, paragraph 3.4.6.2, states, in part, that RCS
leakage shall be limited to no pressure boundary leakage, and that with any pressure
boundary leakage, the unit is to be in Cold Shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. This issue is
encompassed within the licensee root cause analysis, conducted for Condition Report
(CR) 2002-01128. There is no current safety concern because the plant is presently
shut down, cooled down, and defueled. Because the safety significance of the apparent
violation has yet to be determined, the noncompliance will be classified as an
3
unresolved item. This will be identified as URI 50-346/02-08-01, Reactor Operation
with Pressure Boundary Leakage.
b.2
Corrective Action
b.2.1 Reactor Head Boric Acid Deposits
a.
Introduction
The inspectors identified an apparent violation, whose significance is yet to be
determined, involving failure to take adequate corrective action for a continuing buildup
of boric acid deposits on the reactor head. This finding had a credible impact on safety
and was characterized as a URI pending USNRC determination of the significance of
the apparent violation.
b.
Description
A series of Potential Condition Adverse to Quality Reports (PCAQR) and Condition
Reports (CR) from 1990 through 2001 tracks recurrent identification of boric acid
deposits on the reactor head. Section XI of the American Society of Mechanical
Engineers (ASME) Boiler and Pressure Vessel Code IWA-5250, requires that the
leakage source and areas of general corrosion be located when boric acid residues are
detected on components. Section 4.3 of the AIT report (IR 50-346/02-03(DRS))
provides a chronology of reactor head inspections which identify boric acid deposits on
the reactor head and how the licensees engineering staff evaluated and dispositioned
each occurrence.
Through much of the early operating years until the mid-1990s, CRDM flanges at
Davis-Besse were prone to developing leaks during the operating cycle. This leakage
was evidenced by boric acid deposits on the flange, the service structure, and on the
reactor head (due to leakage which ran down the nozzles between the nozzles and the
insulation). Beginning with the plants sixth refueling outage (RFO) in 1990, the licensee
began systematically correcting these leaks by replacing the flange gasket with a new
design. As this program was implemented the frequency of CRDM flange leakage was
reduced. By end of the tenth RFO in 1996, all the flanges had the redesigned gasket
installed.
Beginning with the tenth RFO in 1996 and proceeding through the twelfth RFO in 2000,
six PCAQRs and CRs documented the identification of boric acid deposits on the
reactor head and the licensee engineering staffs disposition of the conditions. Each of
these presented an opportunity to identify nozzle leakage. Collectively, they revealed a
focus on CRDM flange leakage as the source of boric acid deposits despite evidence
that the deposits must be from another source. For example, rust-colored deposits,
indicative of iron, could not likely have come from the flanges which were stainless steel.
Corrosion of the split ring nuts associated with the flanges would not have resulted in
the quantity of corrosion products entrained in the deposits on the head. Finally, there
were significant accumulations of boric acid on the head during operating cycles where
CRDM flange leakage was non-existent or considered negligible.
4
c.
Analysis
This issue represented a performance deficiency because the licensee failed to properly
address, either individually or collectively, the continuing accumulation of large amounts
of boric acid on the reactor head, a significant condition adverse to quality. This lack of
adequate corrective action on the licensees part, contributed to the failure to detect
existing through-wall CRDM nozzle cracks.
This finding is more than minor because the corrosion of the reactor head and the
resulting cavity represented a significant loss of the design basis barrier integrity. Two
cornerstones were impacted by this issue. The Barrier Integrity cornerstone was
affected because the through-wall CRDM cracks compromised the reactor coolant
pressure boundary and the Initiating Events cornerstone was impacted because
cracking of the CRDM nozzles resulted in an increase in the likelihood of a LOCA. The
significance of this finding will be determined by the Significance Determination Process
(SDP) for the issue, which was begun following the AIT activities.
d.
Enforcement
10 CFR Part 50, Appendix B, Criterion XVI, states in part, that measures shall be taken
to ensure that conditions adverse to quality such as failures, malfunctions, deficiencies,
deviations, defective material and equipment, and nonconformances are promptly
identified and corrected. Criterion XVI also requires that for significant conditions
adverse to quality, the measures shall assure that the cause of the condition is
determined and that corrective actions are taken to preclude repetition. The failure to
properly address the recurrent accumulation of boric acid deposits on the reactor head,
a significant condition adverse to quality, contributed to the corrosion of the reactor
head. There is no current safety concern because the plant is presently shut down,
cooled down, and defueled. Because the safety significance of the apparent violation
has yet to be determined, the noncompliance will be classified as an unresolved item.
This will be identified as URI 50-346/02-08-02, Reactor Vessel Head Boric Acid
Deposits.
b.2.2 Containment Air Cooler Deposits
a.
Introduction
The inspectors identified an apparent violation whose significance is yet to be
determined involving failure to take adequate corrective action for recurrent
accumulations of boric acid on containment air cooler (CAC) fins. These accumulations
resulted in reduced air flow through the cooler which was indicated by decreasing
plenum pressure. This finding had a credible impact on safety and was characterized
as an unresolved item pending USNRC determination of the significance of the apparent
violation.
b.
Description
The inspectors reviewed one PCAQR and three CRs spanning the period
November 1998 through January 2001. Section 5.3 of the AIT report
5
(IR 50-346/02-03(DRS)) provides a chronology of CAC fouling and how the
licensees engineering staff evaluated and dispositioned each occurrence.
At the onset of the leak, CRDM nozzle leakage created an aerosol of steam, boric acid,
and other contaminants in the air space above the head. The steam and aerosol
particles were picked up by the service structure ventilation system intake and
distributed throughout the containment. The CACs subsequently condensed the steam
and the boric acid plated out on the cooler fins. Later, as the corrosion of the head
progressed, the aerosol consisted of steam, boric acid, and corrosion particles.
Consistent with this, the deposits on the cooler fins changed color from white to red-
brown.
The licensees attempts to address this phenomenon focused on maintaining operability
of the coolers through frequent cleanings of the coolers. Of the four corrective action
documents examined, only one considered the source of the boric acid deposits;
however, no actions to investigate the source were prescribed. Of particular
significance was the licensees evaluation of the July 1999 appearance of rust-colored
deposits. The licensee continued to attribute the boric acid deposits to CRDM flange
leakage; the discoloration of the boric acid was attributed to migration of the surface
corrosion on the CACs into the boric acid and the aging of the boric acid itself.
c.
Analysis
This issue represented a performance deficiency because the licensee failed to properly
address, either individually or collectively, the cause for the recurrent deposition of boric
acid on CAC fins, nor the change in the color of the deposits, although the change was
indicative of carbon steel corrosion. This lack of adequate corrective action on the
licensees part contributed to their failure to detect existing through-wall CRDM nozzle
cracks and the reactor pressure vessel head corrosion.
This finding is more than minor because the corrosion of the reactor head and the
resulting cavity represented a significant loss of the design basis barrier integrity. Two
cornerstones were impacted by this issue. The Barrier Integrity cornerstone was
affected because the through-wall CRDM cracks compromised the reactor coolant
pressure boundary and the Initiating Events cornerstone was impacted because
cracking of the CRDM nozzles resulted in an increase in the likelihood of a LOCA. The
significance of this finding will be determined by the Significance Determination Process
(SDP) for the issue, which was begun following the AIT activities.
d.
Enforcement
10 CFR Part 50, Appendix B, Criterion XVI, states in part, that measures shall be taken
to ensure that conditions adverse to quality such as failures, malfunctions, deficiencies,
deviations, defective material and equipment, and nonconformances are promptly
identified and corrected. Criterion XVI also requires that for significant conditions
adverse to quality, the measures shall assure that the cause of the condition is
determined and that corrective actions are taken to preclude repetition. The failure to
properly address the recurrent deposits of boric acid deposits on the CAC fins, a
significant condition adverse to quality, contributed to the corrosion of the reactor head.
6
There is no current safety concern because the plant is presently shut down, cooled
down, and defueled. Because the safety significance of the apparent violation has yet
to be determined, the noncompliance will be classified as an unresolved item. This will
be identified as URI 50-346/02-08-03, Containment Air Cooler Boric Acid Deposits.
b.2.3 Radiation Element Filter Deposits
a.
Introduction
The inspectors identified an apparent violation, whose significance is yet to be
determined, involving failure to take adequate corrective action for repeated clogging of
radiation element filters, although a sample of the filter deposits revealed iron oxides,
and radionuclides indicative of reactor coolant. This finding had a credible impact on
safety and was characterized as a URI pending USNRC determination of the
significance of the apparent violation.
b.
Description
Starting in May 1999, yellowish-brown material began to accumulate on the radiation
element filters, causing repetitive, degraded performance of the containment radiation
monitors due to low flow. Condition Report 99-0882 issued on May 13, 1999, identified
low flow conditions on radiation element 4597BA. The CR noted that the apparent
cause was boric acid particles collecting on the filter at a very high rate. The licensee
subsequently sent a sample of the material for analysis, and in August 1999, Southwest
Research Institutes Report No. 18-2321-190 concluded that the deposits on the filters
were a powdery iron oxide and were likely corrosion products from an iron-based
component within the system. The licensee initiated Condition Report 1999-1300 on
May 23, 1999, to document this issue, and noted that Plant Engineering was to issue an
Action Plan for the 12 RFO which would include containment walkdowns to identify
possible sources of the rust particles. Sargent and Lundy was subsequently asked to
review the report, and on November 5, 1999, their response letter stated:
The fineness of the iron oxide (assumed to be ferric oxide) particulate
would indicate it probably was formed from a very small steam leak. The
particulate was likely originally ferrous hydroxide in small condensed
droplets of steam and was oxidized to ferric oxide in the air before it
settled on the filters; and the iron oxide does not appear to be coming
from the general corrosion of a bare metal surface in containment or from
steam impingement on a metal surface.
Although the licensee conducted containment entries at power to identify the source
of the apparent steam leak, it was never identified. After the 12th RFO in May 2000,
the radiation element filters continued to clog with corrosion products. CR 01-1110
was issued on April 23, 2001, to document continued clogging of the radiation monitor
filters due to boric acid build-up. Corrective action was to move the sample point.
CR 01-1822 dated July 23, 2001, documented increased frequency of monitor filter
change outs again, due to boric acid clogging. Disposition of this CR was to continue to
change the filters until an upcoming refueling outage could identify the source of RCS
leakage. CR 01-2795, dated October 22, 2001, again identified a high frequency of filter
7
clogging, noting that previous corrective actions were unsuccessful. Corrective action
for this CR was to perform a temporary modification (TM 01-0019) to remove the filter
cartridge.
c.
Analysis
This issue represented a performance deficiency because the licensee failed to take
appropriate corrective action (identify and repair the source of the RCS leakage) for a
significant condition adverse to quality, in that, the filters clogged with a material
indicative of RCS leakage and corrosion products and this continued for more than two
years. It has subsequently been concluded that the material clogging the filters was
from the ongoing corrosion of the reactor vessel head. While some containment
walkdowns were conducted, the source of the RCS leakage was not identified.
Furthermore, the corrective actions implemented for the CRs written on this problem
appeared to focus on the operability of the radiation monitors and not the root cause
(i.e., RCS leakage and corrosion). This lack of adequate corrective action on the
licensees part, contributed to their failure to detect existing through-wall CRDM nozzle
cracks.
This finding is more than minor because it affected the Initiating Events cornerstone
objective in that cracking of CRDM nozzles represented an increase in the likelihood of
a LOCA. The Barrier Integrity cornerstone was also affected in that CRDM cracks
resulted in leakage through the reactor coolant pressure boundary. The significance of
this finding will be determined by the Significance Determination Process (SDP) for the
issue, which was begun following the AIT activities.
d.
Enforcement
10 CFR Part 50, Appendix B, Criterion XVI, requires, in part, that conditions adverse to
quality be promptly identified and corrected. Criterion XVI also requires that for
significant conditions adverse to quality, the measures shall assure that the cause of the
condition is determined and that corrective actions are taken to preclude repetition.
Contrary to this, the licensee failed to correct the condition identified in CR 99-0882,
initiated on May 13, 1999. As of February 16, 2002, the source of the RCS leakage had
not been identified and corrected. There is no current safety concern because the plant
is presently shut down, cooled down, and defueled. Because the safety significance of
this apparent violation has yet to be determined, it will be classified as an unresolved
item. This will be identified as URI 50-346/02-08-04, Radiation Element Filters.
b.2.4 Service Structure Modification
a.
Introduction
An apparent violation whose significance is yet to be determined, was identified for the
failure to implement the corrective action procedure and take prompt corrective action
for a condition adverse to quality. The licensee failed to implement a modification to
permit complete inspection and cleaning of the reactor vessel head and CRDM nozzles.
This finding had a credible impact on safety and was characterized as a URI pending
USNRC determination of risk significance.
8
b.
Description
USNRC Generic Letter 88-05 was issued on March 17, 1988, and notified licensees of
the potential for boric acid degradation of carbon steel reactor pressure boundary
components. On May 28, 1993, BAW issued document BAW 10190P, Safety
Evaluation for BAW Design Reactor Vessel Head Control Rod Drive Mechanism Nozzle
Cracking. This document stated that B&WOG utilities developed plans to visually
inspect the CRDM nozzle area to determine if through wall cracking had occurred. If
any leaks or boric acid crystal deposits were located, an evaluation of the source of the
leak and the extent of any wastage was to be evaluated. The USNRC notified
NUMARC of the results of the USNRC safety evaluation related to this subject on
November 19, 1993. This safety evaluation concluded that there was no current safety
concern for cracking of CRDM penetrations. The conclusion was predicated on the
performance of the visual inspection activities requested in Generic Letter 88-05.
Modification 90-0012 was initiated during RFO 6 on March 21, 1990, to install multiple
access ports in the service structure to permit cleaning and inspection of the reactor
head. At this time, it was noted that boric acid had leaked down from the CRD flanges
and accumulated on the head, thus subjecting the head to potential degradation. This
modification was not performed during the next two refuel outages (RFO 7 and 8) and
was subsequently voided on September 27, 1993, noting that the head had been
successfully inspected with cameras and cleaned during the previous three refuel
outages (utilizing existing small mouse holes holes in the service structure).
Modification 94-0025 was initiated on May 27, 1994, again to install access ports in the
service structure to permit cleaning and inspection of the head. Reasons cited for the
modification were that video inspections (through existing mouse holes) were difficult
and not always adequate and that inspection and cleaning of 100 percent of the head
could not be accomplished without performing the modification. This modification was
not performed during RFO 9 (October 1, 1994) and a decision was made on June 15,
1995, to defer scheduling it for RFO 10 (April 8, 1996) pending further industry
information/investigation concerning actual benefit. During a meeting of the Work
Scope Committee (WSC) on February 20, 1997, a decision was made to reschedule the
modification for RFO 12 (April 1, 2000). Subsequently, during a meeting of the WSC on
September 17, 1998, the modification was again rescheduled, this time, for RFO 13
(February 16, 2002).
Potential Condition Adverse to Quality 96-0551 was initiated on April 21, 1996, to
address the inability to comply with some of the provisions of NG-EN-00324 (Boric Acid
Corrosion Control Program) with respect to inspection and cleaning of the head during
RFO 10. This PCAQR stated, in part, Since the boric acid deposits are not cleaned it is
difficult to distinguish whether the deposits occurred because of the leaking flanges or
the leaking CRDM. This situation represents an adverse trend with the potential for
greater than marginal consequences. The PCAQR further states, The extent of the
inspection was limited to approximately 50 to 60 percent of the head area because of
the restrictions imposed by the location and size of mouse holes. Modification 94-0025
was specified as the corrective action for this PCAQR.
9
c.
Analysis
This issue represented a performance deficiency because the licensee failed to take
corrective action (install the access port modification) for a condition adverse to quality.
As noted above, from the information in Modification 94-0025 and PCAQR 96-0551, it is
apparent that complete inspection and cleaning of the head could not be performed
without installing the access ports. As of February 16, 2002, the modification had not
been performed, the head had not been completely inspected, and the head had not
been completely cleaned. This lack of action on the licensees part, contributed to their
failure to detect existing through-wall CRDM nozzle cracks.
This finding is more than minor because it affected the Initiating Events cornerstone
objective in that cracking of CRDM nozzles represented an increase in the likelihood of
a LOCA. The Barrier Integrity cornerstone was also affected in that CRDM cracks
resulted in leakage through the reactor coolant pressure boundary. Furthermore, the
failure to provide for adequate inspection and cleaning of the head was a contributing
factor to the head degradation. The significance of this finding will be determined by the
SDP for the issue, which was begun following the AIT activities.
d.
Enforcement
10 CFR Part 50, Appendix B, Criterion V, requires, in part, that activities affecting quality
be accomplished in accordance with written procedures. The licensee failed to follow its
corrective action procedure and correct the condition identified on April 21, 1996
(inability to fully inspect the head and CRDM nozzles), in that, as of February 16, 2002,
the corrective action (modification of the service structure) had not been accomplished.
There is no current safety concern because the plant is presently shut down, cooled
down, and defueled. The service structure has since been modified to permit complete
inspection and cleaning of the head. Because the safety significance of this apparent
violation has yet to be determined, it will be classified as an unresolved item. This will
be identified as URI 50-346/02-08-05, Service Structure Modification Delay.
b.2.5 Reactor Coolant System Unidentified Leakage Trend
a.
Introduction
The inspectors identified a finding whose significance is yet to be determined involving
failure to follow the corrective action procedure and complete a prescribed corrective
action for adverse trends in RCS unidentified leakage. This finding had a credible
impact on safety and was characterized as an unresolved item pending USNRC
determination of the significance of the apparent violation.
b.
Description
In 1998, shortly after completing the eleventh RFO, the licensee identified a sharp rise in
RCS unidentified leakage which had been relatively stable at 0.05 gpm. This was
attributed to a temporary modification which bypassed a pressurizer relief valve drain
line and allowed leakage past the relief valves to be vented directly into the containment
atmosphere. This leakage collected in the normal sump and added to the unidentified
10
leakage, which increased to a maximum of 0.8 gpm. During a mid-cycle outage in May
of 1999, the licensee resolved this issue by installing new rupture disks and
reconnecting the drain line. This resulted in a decrease in unidentified leakage.
However, the unidentified leakage returned to levels between 0.15 and 0.25 gpm.
The inspectors reviewed a series of four CRs which demonstrated that the licensee was
aware of the increase in unidentified leakage, and becoming increasingly concerned by
their inability to identify the source. The engineering evaluation into the issue became
more involved with each succeeding CR. CR 2001-2862 contained a detailed evaluation
and a corrective action to develop a containment inspection plan for the forthcoming
RFO. This inspection plan was completed and documented in the CR. It involved
coordination of four actions: the Mode 3 (reactor shutdown, normal operating
temperature and pressure) walkdown, mode 5 (cold shutdown) RCS walkdowns, Boric
acid corrosion control walkdowns, and the ASME VT-2 examinations. The inspectors
determined that proposed corrective actions associated with RCS unidentified leakage
were adequate; however, a key action, the mode 3 walkdown was subsequently
canceled. This significantly reduced the quality of the proposed corrective action to the
point where it was no longer adequate.
c.
Analysis
This issue represented a licensee performance deficiency because elimination of a key
component of what was an adequate proposed corrective action rendered the proposal
inadequate. Consequently, this was considered a finding. This finding was of more
than minor safety significance because the corrosion of the reactor head and the
resulting cavity represented a significant loss of the design basis barrier integrity. Two
cornerstones were impacted by this issue. The Barrier Integrity cornerstone was
affected because the through-wall CRDM cracks compromised the reactor coolant
pressure boundary and the Initiating Events cornerstone was impacted because
cracking of the CRDM nozzles resulted in an increase in the likelihood of a LOCA. The
significance of this finding will be determined by the SDP for the issue, which was begun
following the AIT activities.
d.
Enforcement
The licensee failed to follow the corrective action procedure and implement an effective
corrective action for adverse trends in RCS unidentified leakage. There is no current
safety concern because the plant is presently shut down, cooled down, and defueled.
Because the safety significance of this finding has yet to be determined, it will be
classified as an unresolved item. This will be identified as URI 50-346/02-08-06,
Reactor Coolant System Unidentified Leakage Trend.
11
b.3
Procedures
b.3.1 Procedures Not Appropriate to the Circumstances
a.
Introduction
The inspectors identified an apparent violation whose significance is yet to be
determined, involving deficiencies in the licensees Boric Acid Corrosion Control
procedure, NG-EN-00324. This finding had a credible impact on safety and was
characterized as an unresolved item pending USNRC determination of the significance
of the apparent violation.
b.
Description
The inspectors reviewed the original and subsequent revisions, up to the date of
discovery of the reactor head conditions, of the licensees Boric Acid Corrosion Control
Program, NG-EN-00324. The purpose of the review was to assess the adequacy of the
procedure and determine whether any deficiencies may have contributed to the event.
The inspectors noted the following weaknesses in the procedure:
a.
The procedure had a clear focus on bolted, flanged connections. Seven
of nine principal locations in Section 6.1.1, the definition of an RCS
pressure boundary component in Section 4.9, and the definitions of minor
(Section 4.2), moderate (Section 4.3), and substantial (Section 4.4)
leakage all contained references to bolted connections.
b.
In Section 6.3.4, the procedure directed preparation of a CR, repair tag,
or work order if a detailed inspection was warranted, but guidance,
specifications, or thresholds for initiating a detailed inspection were
inadequate.
c.
The inspectors determined that preparation of a repair tag or work order
in lieu of a CR was inappropriate because it only addressed the symptom
by fixing the leak rather than evaluate why the leak was occurring.
d.
Qualifications for Plant Engineering staff conducting inspections and
evaluations were not addressed. This resulted in inconsistencies in
inspection techniques, observations, recording of results, and
evaluations.
e.
In section 6.3.1.f, the procedure exempted stainless steel or inconel
components from further examination related to boric acid corrosion,
unless the examination was during an ASME Section XI test which might
require a bolting examination. However, there was industry experience
dating back to 1990, including an USNRC Information Notice, identifying
primary water stress corrosion cracking of stainless steel, due to boric
acid attack, as a concern.
12
f.
The procedure did not require maintenance of any documents, such
as checklists or evaluations, although the procedure is quality related
and Davis-Besse Supplemental Procedure Requirements/Guidance,
NG-QS-00120 stated in Attachment 2, Section1.2.b that these
procedures were used to assure safe operation.
The inspectors determined that these weaknesses collectively contributed to the
corrosion of the reactor head, either through narrowing the scope of inspection
or failing to provide adequate instruction for carrying out activities.
c.
Analysis
This issue represented a licensee performance deficiency because the weaknesses in
the procedure contributed to the failure, over a period of years, by the licensees
engineering staff to properly identify and evaluate the leaking CRDM nozzle and the
expanding cavity in the reactor head. Consequently, this was considered a finding. This
finding was of more than minor safety significance because the cavity in the reactor
vessel head represented a loss of the design basis barrier integrity. The significance of
this finding will be determined by the SDP for the issue, which was begun following the
AIT activities.
d.
Enforcement
10 CFR Part 50, Appendix B, Criterion V, states, in part, that activities affecting quality
shall be prescribed by documented instructions, procedures, or drawings, of a type
appropriate to the circumstances and shall be accomplished in accordance with these
instructions, procedures, or drawings. NG-EN-00324, Boric Acid Corrosion Control
Program, Revisions 0 through 2, were classified as a quality procedure under the
licensees procedure administrative system and were not appropriate to the
circumstances in that deficiencies in the procedure contributed to the failure to detect
and address corrosion of the reactor head. There is no current safety concern because
the plant is presently shut down, cooled down, and defueled. Because the safety
significance of the apparent violation has yet to be determined, the noncompliance will
be classified as an unresolved item. This will be identified as URI 50-346/02-08-07,
Inadequate Boric Acid Corrosion Control Program Procedure.
b.3.2 Failure to Follow Boric Acid Corrosion Control Program Procedure
a.
Introduction
The inspectors identified an apparent violation whose significance is yet to be
determined involving multiple examples of failure to follow the boric acid corrosion
control procedure. This finding had a credible impact on safety and was characterized
as an unresolved item pending USNRC determination of the significance of the apparent
violation.
13
b.
Description
The inspectors reviewed a series of PCAQRs and CRs that documented occurrences of
the licensees failure to adhere to the instructions in the Boric Acid Corrosion Control
Program procedure.
PCAQR 96-0551 (in RF010) recorded that a visual inspection of the head
showed boric acid accumulation on the head and that CRDM nozzle No. 67 had
rust-colored deposits where it penetrated the head. The PCAQR further records
that inspection of CRDM nozzle No. 67 flange showed no signs of leakage
during the operating cycle, signifying that the present deposits were the result of
leakage from previous operating cycles. The conclusion drawn from these two
statements was that, as a minimum, boric acid deposits were not removed from
the head, nor was the base metal inspected for corrosion during RF09 as
directed by the boric acid corrosion control program procedure.
PCAQRs 98-0649 and 98-0767 (in RF011) both recorded visual inspections of
the reactor head using a video camera on April 17, 1998, and April 24, 1998,
respectively. Both inspections revealed boric acid residue on the head.
PCAQR 98-0649 focused on CRDM flange D-10 which was determined to have
minor leakage, based on the amount of boric acid on the flange. A review of
unidentified leakage was conducted and average unidentified leakage during the
previous operating cycle was 0.05 gallons per minute. Based on this
information, repair of flange D-10 was deferred. PCAQR 98-0767 recorded that
most of the head area was covered with an uneven layer or boric acid along
with some lumps of boric acid. The color of the layer and the lumps varied from
rust brown to white. The rust or brown color is an indication of the old boric acid
deposits. The conclusion drawn from these two PCAQRs is that boric acid
deposits were left on the head at the end of the tenth RFO and that the base
metal under these deposits was not inspected as directed by the boric acid
corrosion control program procedure.
PCAQR 98-0767 (RF011) records that the reactor head was cleaned as best we
can. Later the PCAQR records that an inspection after the cleaning showed
there were boric acid deposits left on the head after the cleaning. At the end of
RF011, the base metal under these deposits was not inspected as directed by
the boric acid corrosion control program procedure.
CRs 2000-0781, 2000-0782, and 2000-1037 (RF012) were all written to
document the extensive build-up of boric acid residue on the reactor head.
CR 2000-0782 describes the conditions in detail and photographs of the head
near the closure studs were included. Among the corrective actions specified in
the condition report was cleaning boric acid deposits off the head in accordance
with work order 00-001846-000. A video tape made after this cleaning showed
that a thick layer of red/brown boric acid deposits remained around the nozzles
near the center of the head. At the end of the twelfth RFO, the base metal under
these deposits was not inspected as directed by the boric acid corrosion control
program procedure.
14
c.
Analysis
This issue represented a licensee performance deficiency because the recurrent
failures, by the licensees engineering staff, to remove boric acid deposits and inspect
the base metal of the reactor head as directed by the boric acid corrosion control
procedure, resulted in the perpetuation of the CRDM nozzle leak and the development
of the expanding cavity in the reactor head. Consequently, this was considered a
finding. This finding was of more than minor safety significance because the cavity in
the reactor vessel head represented a loss of the design basis barrier integrity. The
significance of this finding will be determined by the SDP for the issue, which was begun
following the AIT activities.
d.
Enforcement
10 CFR Part 50, Appendix B, Criterion V, states, in part, that activities affecting quality
shall be prescribed by documented instructions, procedures, or drawings, of a type
appropriate to the circumstances and shall be accomplished in accordance with these
instructions, procedures, or drawings. The licensees engineering staff failed, on
multiple occasions, to remove boric acid deposits and inspect the base metal of the
reactor head as directed by NG-EN-00324, Revision 2, Boric Acid Corrosion Control
Program. This issue is encompassed within the licensee root cause analysis,
prescribed by CR 2002-01128. There is no current safety concern because the plant is
presently shut down, cooled down, and defueled. Because the safety significance of the
apparent violation has yet to be determined, the noncompliance will be classified as an
unresolved item. This will be identified as URI 50-346/02-08-08, Failure to Follow Boric
Acid Corrosion Control Program Procedure.
b.3.3 Failure to Follow Corrective Action Program Procedure
a.
Introduction
The inspectors identified an apparent violation whose significance is yet to be
determined involving two examples of failure to follow the stations corrective action
program procedure. This finding had a credible impact on safety and was characterized
as an unresolved item pending USNRC determination of the significance of the apparent
violation.
b.
Description
CRs 2000-0782, and 2000-1037 (RF012) were written to document the extensive
build-up of boric acid residue on the reactor head. CR 2000-0782 described the
conditions in detail and photographs of the head near the closure studs were included.
CR 2000-1037 described the analysis (deferring in part to CR 2000-0782) and the
cleaning effort. In both CRs the extent and significance of the boric acid deposits, and
that such build-up was recurrent, were clear. The quantity of boric acid deposits
accumulated was highly unusual, extensive corrective actions were necessary, and an
adverse repetitive trend existed.
15
Attachment 2, Categorization of Condition Report, to NG-NA-00702, Revision 3,
Corrective Action Program, provides guidance and examples for characterization of
condition reports as significant, important, routine, or non-conditions adverse to quality.
Among the examples of significant conditions are:
Issues of collective significance that considered individually may not be
significant, but as a whole indicate problems that warrant root cause
investigation and corrective action to prevent recurrence.
Substantial deviations, deficiencies in construction or design, damage such that
extensive corrective actions are required.
A repetitive or adverse trend exists.
During RF012, the engineering staff was aware of the continuing accumulation of boric
acid on the reactor head; the problem had been documented in PCAQRs and CRs since
1996. Although the quantity of boric acid deposits was a substantial deviation from
acceptable operating conditions for the reactor head, the history of reactor head boric
acid deposits revealed a significant problem, and that a repetitive, adverse trend existed.
Both condition reports were classified as routine.
c.
Analysis
This issue represented a licensee performance deficiency. A proper characterization of
condition reports 2000-0782 and 2000-1037 would have resulted in a formal root cause
evaluation as prescribed by the licensees Root Cause Analysis Reference Guide.
Incorrectly characterizing these CRs as routine resulted in an apparent cause
determination with no required corrective actions to prevent recurrence; an opportunity
to identify the true nature of the leak and the growing cavity in the head was missed.
Consequently, this was considered a finding. This finding was of more than minor
safety significance because the cavity in the reactor vessel head represented a loss of
the design basis barrier integrity. The significance of this finding will be determined by
the SDP for the issue, which was begun following the AIT activities.
d.
Enforcement
10 CFR Part 50, Appendix B, Criterion V, states, in part, that activities affecting quality
shall be prescribed by documented instructions, procedures, or drawings, of a type
appropriate to the circumstances and shall be accomplished in accordance with these
instructions, procedures, or drawings. The licensee failed to properly characterize
CRs 2000-0782 and 2000-1037 as significant conditions adverse to quality, in
accordance with the guidance contained in the licensees corrective action program
procedure. This issue is encompassed within the licensee root cause analysis,
prescribed by CR 2002-01128. There is no current safety concern because the plant is
presently shut down, cooled down, and defueled. Because the safety significance of the
apparent violation has yet to be determined, the noncompliance will be classified as an
unresolved item. This will be identified as URI 50-346/02-08-9, Failure to Follow
Corrective Action Program Procedure.
16
b.4
Completeness and Accuracy of Information
a.
Introduction
In addition to the information in the AIT report subsequent questions were raised
regarding completeness and accuracy of documents either required by the USNRC to
be maintained by the licensee or submitted to the USNRC. Consequently, as licensee
documents associated with this issue were reviewed for regulatory compliance, they
were concurrently reviewed for completeness and accuracy in all material respects as
required by 10 CFR 50.9, Completeness and Accuracy of Information.
b.
Description
The inspectors review of the extensive documentation associated with this issue
revealed a series of examples of information provided to the Commission or required by
the Commissions regulations to be maintained by the licensee that were not complete
and accurate.
1.
The cancellation of Modification 90-012 was based on a statement in a
Document Void Request, approved September 23, 1993: Current inspection
techniques using high-powered cameras preclude the need for inspection ports,
additionally, cleaning of the reactor vessel head during last three outages was
completed successfully without requiring access ports. This statement is
inaccurate because boric acid deposits were left on the head at the end of both
the seventh and eighth refueling outages, the two outages preceding this
statement.
2.
PCAQR 98-0649, dated April 18, 1998, made the statement, Accumulation of
boric acid on the reactor vessel caused by leaking CRDMs has not resulted in
any boric acid corrosion. This was identified through inspections following
reactor vessel head cleaning in past outages. This statement is inaccurate
because areas of the reactor head were not cleaned of boric acid deposits nor
was the base metal under all the deposits inspected.
3.
PCAQR 98-0649 also contained the following statement, Additionally, B&W
documentation discussing CRDM nozzle cracking further stated that boric acid
deposits on the head caused by leaking CRDM flanges would not result in head
corrosion. The PCAQR did not state which B&W document was being
referenced. The inspectors reviewed the following documents:
a.
51-1219275-01, CRDM Leakage Detection Evaluation, December 13,
1993
b.
51-1229638-00, Boric Acid Corrosion Data Summary and Evaluation,
April 15, 1994
c.
BAW-10190P, Safety Evaluation for B&W-Design Reactor Vessel Head
Control Rod Drive Mechanism Nozzle Cracking
17
d.
BAW-10190P, Addendum 1, External Circumference Crack Growth
Analysis for B&W-Design Reactor Vessel Head CRDM Nozzles
e.
BAW-10190P, Addendum 2, Safety Evaluation for Control Rod Drive
Mechanism Nozzle J-Groove Weld
f.
B&W Materials Committee Report 51-1201160-00, Alloy 600 SCC
Susceptibility: Scoping Study of Components at Crystal River 3
g.
B&W Report 51-1218440-00, Alloy PWSCC Time-to-Failure Models
h.
B&W Report 51-1219143-00, CRDM Nozzle Characterization
i.
BAW-2301, B&WOG Integrated Response to USNRC Generic Letter 97-
01, Degradation of Control Rod Drive Mechanism Nozzle and Other
Vessel Closure Head Penetrations
The inspectors did not find that statement in any of those nine documents, nor
did the licensee identify the source document for that statement.
4.
PCAQR 98-0767, dated April 25, 1998, Section 4A, Item F, stated, The boric
acid deposits were removed from the head. This is incorrect information; it has
been acknowledged that the head was not completely cleaned at the end of
eleventh RFO.
5.
Condition report CR 2000-1037, dated April 17, 2000, page 6 of 7, under
Remedial Actions stated, Accumulated boron deposited between the reactor
head and the thermal insulation was removed during the cleaning process
performed under W.O. 00-001846. No boric acid induced damage to the head
surface was noted during the subsequent inspection. This statement was
inaccurate in that the accumulated boric acid was only removed from some
areas of the head, and the subsequent inspection of the head surface for boric
acid induced damage was only for that portion of the head where the boric acid
deposit had been removed.
6.
Work Order 00-001846-000, Clean Boron Accumulation from Top of Reactor
Head and Top of Insulation, dated April 25, 2000, was prepared and issued to
clean the reactor head as directed by the boric acid corrosion control procedure.
The Work Order log stated, work performed without deviation. This was
inaccurate since CRs clearly indicated that boric acid deposits were left on the
head after the cleaning.
7.
QA Audit report AR-00-OUTAG-01, dated July 7, 2000, stated, in part, Boric
Acid Corrosion Control Checklists and Condition Reports were initiated by
inspectors when prudent to document and evaluate boric acid accumulation and
leaks. Boric acid leakage was adequately classified and corrected when
appropriate. Engineering displayed noteworthy persistence in ensuring boric
acid accumulation from the reactor head was thoroughly cleaned. This audit
report contains inaccurate information: (1) the reactor head was not thoroughly
18
cleaned during the outage; (2) a boric acid corrosion control checklist was not
prepared for the boric acid left on the head after the cleaning attempt; and (3)
the boric acid accumulation and leaks were not identified, properly classified, nor
corrected.
8.
Davis-Besse letter, Serial 2731, September 4, 2001, Response to
Bulletin 2001-01, contained the following four inaccuracies:
a.
The response to Item 1.c on page 2 of 19 contained the statement that
the minimum gap being at the dome center of the RPV head where it is
approximately 2 inches, and does not impede a qualified visual
inspection. This is contradicted by statements in several PCAQRs, most
notably 94-0295, which prompted the reintroduction of the service
structure modification, and 96-0551 which clearly stated that inspection
capability at the top of the head was limited. This limitation was caused
by the restricted access to the area through the service structure weep
holes, the curvature of the reactor pressure vessel head, and by the
limited space to manipulate a camera due to the insulation that creates
the two inch gap.
b.
Item 1.d of the Bulletin directed inclusion of a description of any
limitations (insulation or other impediments) to accessibility of the bare
metal of the RPV head for visual examinations. The response was
incomplete in that it did not mention that accessibility to the bare metal of
the reactor head was impeded by the significant accumulations of boric
acid deposits in both the eleventh and twelfth RFOs.
c.
Item 1.d of the bulletin directed a discussion of the findings of vessel
head inspections. The response to this on page 3 for the twelfth RFO
was that inspection of the RPV head/nozzles indicated some
accumulation of boric acid deposits. This was a mischaracterization of
the accumulations as evidenced by the pictures and the video
examination of conditions on the head at the beginning and ending of the
outage.
d.
Additionally on page 3, the response stated, The boric acid deposits
were located beneath the leaking flanges with clear evidence of
downward flow. No visible evidence of nozzle leakage was detected.
This was inaccurate, in that the boric acid deposits were not all located
under leaking flanges and there was no clear evidence of downward flow
for all nozzles. The deposits were not limited to the area beneath the
flanges as implied by that statement and, in fact, the build-up was so
significant that all of the nozzles could not be inspected. There was no
basis for stating that no visible evidence of nozzle leakage was detected.
9.
Davis-Besse letter, Serial 2735, October 17, 2001, Supplemental Response to
Bulletin 2001-01, stated, In May 1996, during a refueling outage, the RPV head
was inspected. No leakage was identified, and these results have been recently
verified by a re-review of the video tapes obtained from that inspection. The
19
RPV head was mechanically cleaned at the end of the outage. Subsequent
inspections of the RPV head in the next two refueling outages (1998 and 2000),
also did not identify any leakage in the CRDM nozzle-to-head areas that could
be inspected. Video tapes taken during these inspections have also been
re-reviewed. These statements were inaccurate in that they implied that the
head was completely cleaned and inspected. The RPV head could not be
completely inspected, as evidenced by PCAQR 96-0551. The RPV head was
not cleaned as evidenced by PCAQR prepared at the start of the 1998 outage
which stated that there were old boric acid deposits on the head.
c.
Analysis
This issue represented a licensee performance deficiency. Completeness and accuracy
of information are essential to the ability of the USNRC to establish a regulatory position
on issues which can affect the health and safety of the general public. Completeness
and accuracy in the documents listed above may have provided an earlier alert to the
licensee staff and the USNRC about the problems with CRDM nozzle leakage or may
have caused the USNRC to establish a different regulatory position concerning the
urgency of inspections for the RPV head. Consequently, this was considered a finding
of more than minor safety significance because the cavity in the reactor vessel head
represented a loss of the design basis barrier integrity.
d.
Enforcement
10 CFR 50.9 requires that information provided to the Commission by a licensee or
information required by statute or by the Commissions regulations, order, or license
conditions maintained by the licensee shall be complete and accurate in all material
respects. The examples listed contain incomplete or inaccurate information material to
the USNRC. The significance of these examples requires additional review as specified
in NUREG-1600, General Statement of Policy and Procedures for USNRC
Enforcement. Because the safety significance of the apparent violation has yet to be
determined, the issue will be classified as an unresolved item and will be identified as
URI 50-346/02-08-10, Completeness and Accuracy of Information.
4OA6 Management Meetings
Exit Meeting Summary
The inspectors presented the inspection results to Mr. L. Myers and other members of
licensee management and staff at the conclusion of the inspection on August 9, 2002.
The licensee acknowledged the information presented. Proprietary information
reviewed and retained by the inspectors was identified.
20
KEY POINTS OF CONTACT
DAVIS-BESSE
D. Baker, LCM(A) Manager
R. Fast, Plant Manager
J. Grabnar, Design Basis Engineering Manager
D. Gudger, Learning Organization Manager
D. Haskins, Human Resources Manager
S. Loehlein, Principal Nuclear Consultant
L. Myers, Site Vice President
L. Pearce, Vice President, Oversight
J. Powers, Engineering Director
P. Roberts, Maintenance Manager
M. Roder, Operations Manager
J. Rogers, Plant Engineering Manager
R. Slyker, Licensing Staff Engineer
H. Stevens, Quality Assurance Manager
G. Wolf, Licensing Staff Engineer
NUCLEAR REGULATORY COMMISSION
J. Grobe, Chairman, Davis-Besse Oversight Panel
C. Lipa, Chief, Reactor Projects Branch 4
S. Thomas, Senior Resident Inspector
LIST OF ACRONYMS USED
Augmented Inspection Team
American Society of Mechanical Engineers
Babcock and Wilcox
Containment Air Cooler
CR
Condition Report
Control Rod Drive Mechanism
Electric Power Research Institute
GL
Generic Letter
gpm
Gallon Per Minute
Loss of Coolant Accident
PCAQR
Potential Conditions Adverse to Quality Report
Public Document Room
RE
Radiation Element
Refueling Outage
Significance Determination Process
Unresolved Item
U. S. Nuclear Regulatory Commission
21
ITEMS OPENED
50-346/2002-08-01
Reactor Operation with Pressure Boundary Leakage
50-346/2002-08-02
Reactor Vessel Head Boric Acid Deposits
50-346/2002-08-03
Containment Air Cooler Boric Acid Deposits
50-346/2002-08-04
Radiation Element Filters
50-346/2002-08-05
Service Structure Modification Delay
50-346/2002-08-06
Reactor Coolant System Unidentified Leakage Trend
50-346/2002-08-07
Inadequate Boric Acid Corrosion Control Program
Procedure
50-346/2002-08-08
Failure to Follow Boric Acid Corrosion Control Program
Procedure
50-346/2002-08-09
Failure to Follow Corrective Action Program Procedure
50-346/2002-08-10
Completeness and Accuracy of Information
22
LIST OF DOCUMENTS REVIEWED
The following is a list of licensee documents reviewed during the inspection, including
documents prepared by others for the licensee. Inclusion on this list does not imply that
USNRC inspectors reviewed the documents in their entirety, but, rather that selected sections
or portions of the documents were evaluated as part of the overall inspection effort. Inclusion
on this list does not imply USNRC acceptance of the document, unless specifically stated in the
inspection report.
Procedures
NG-EN-00324
Boric Acid Corrosion Control, Revisions 1, 2, and 3
NG-NA-00305
Operating Experience Program
NG-NA-00702
Corrective Action Program, Revision 3
DB-PF-00204
ASME Section XI Pressure Testing, Revision 3
DB-OP-01200
Reactor Coolant System Leakage Management, Revision 3
Potential Conditions Adverse to Quality Reports (PCAQR)
1991-0353
Boron on Reactor Vessel Head
1992-0072
CAC Cleaning
1993-0132
Reactor Coolant Leakage from CRD Flange
1994-0295
Improper Closure of the Nozzle Leakage Inspection Commitment
1994-0912
Documents Results of CRDM leakage Video Inspection
1996-0551
Boric Acid on RX Vessel Head
1996-0650
VT-2 Inspection Revealed Evidence of Leakage and Boric Acid Residue
1998-0649
Reactor Vessel Head Boron Deposits
1998-0767
Reactor Vessel Head Inspection Results
1998-0824
CAC Boric Acid Accumulation
1998-1164
Water in RE4597 Sample Lines
1998-1895
CTMT Normal Sump Leakage in Excess of 1 gpm
1998-1965
Water and Boron Accumulation on Filter Cartridges
1998-1980
Potential CAC Fouling
1998-2071
Accumulation of Boric Acid on CTMT Service Water Piping
Condition Reports (CR)
1992-0139
Boron Found on Containment Air Sample Filter
1993-0187
Boric Acid Accumulation on SW Piping
1999-0372
Received Computer PT-RE4597AA/AB High
1999-0510
Low Flow Alarm Observed on RE4597BA While Out of Service for Maintenance
1999-0745
Small Clumps of Boric Acid Present on Wall Opposite of DH108
1999-0861
RE4597AA Sample Lines Were Found to be Full of Water
1999-0928
Increased Frequency of Particulate and Charcoal Filters for RE 4597BA Being
Changed
1999-1300
Analysis of CTMT Radiation Monitor Filters
1999-1614
Due Date of LER Commitment Missed: Boric Acid Control Program Procedure
Change
23
2000-0781
Leakage from CRD Structure Blocked Visual Exam of Reactor Vessel Head
Studs
2000-0782
Inspection of Reactor Flange Indicated Boric Acid Leakage From Weep Holes
2000-0994
RV Head CRDM Nozzle at Location F-10 has Large Pit in Outer Gasket Groove
2000-0995
RV Head CRDM Nozzle Flange at Location D-10 has Extensive Pitting Across
the Outer Gasket Groove. Inner Gasket Also Has Pitting
2000-1037
Inspection of Reactor Head Indicated Accumulation of Boron in Area of the CRD
Nozzle Penetration
2000-1547
CAC Plennum Pressure Drop Following 12 RFO
2000-4138
Frequency for Cleaning Boron From CAC Fins Increased to Interval of
Approximately 8 Weeks
2001-0039
CAC Plenum Pressure Experienced Step Drop
2001-0890
Unidentified RCS Leak Rate Varies Daily by as Much as 100 percent of the
Value
2001-1110
Chemistry is Changing Filters on RE4597BA More Frequently
2001-1822
Frequency of Filter Changes for RE4597BA is Increasing
2001-1857
RCS Unidentified Leakage at .125 to .145 gpm
2001-2769
RE2387 Identified Spiked Above ALERT and High Setpoints
2001-2795
RE4597BA Alarmed on Saturation
2001-2862
Calculated Unidentified Leakage for Reactor Coolant System has Indicated
Increasing Trend
2001-3025
Increase in RCS Unidentified Leakage
2001-3411
Received Equipment Fail Alarm for Detector Saturation on RE4597BA
2002-0685
Loose Boron 1-2" deep 75% Around Circumference of Flange
2002-0846
More Boron Than Expected Found on Top of Head
Modifications
MOD 90-0012
Modification Reactor Closure Head Access Ports
MOD 94-0025
Install Service Structure Inspection Openings
USNRC Generic Communications for Control of Boric Acid Corrosion
IN 86-108
Degradation of Reactor Coolant System Pressure Boundary Resulting
from Boric Acid Corrosion, December 29, 1986
IN 86-108
Supplement 1, April 20,1987
IN 86-108
Supplement 2, November 19, 1987
IN 86-108
Supplement 3, January 5, 1995
Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary
Components in PWR Plants, March,17, 1988
Primary Water Stress Corrosion Cracking (PWSCC) of Inconel 600,
February 23,1990
Boric Acid Corrosion of Charging Pump Casing Caused by Cladding
Cracks, August 30, 1994
Ingress of Demineralizer Resins Increases Potential for Stress Corrosion
Cracking of Control Rod Drive Mechanism Penetrations,
February 14,1996
Degradation of CRDM/CEDM Nozzle and other Vessel Closure Head
Penetrations, April 1, 1997
24
Through-wall Circumferential Cracks of Reactor Pressure Vessel Head
Control Rod Drive Mechanism Penetration Nozzles at Oconee Nuclear
Station, Unit 3, April 30, 2001
Circumferential Cracking of Reactor Pressure Vessel Head Penetration
Nozzles, dated August 3, 2001
Other Documents
RAS02-00132
Probable Cause Summary Report for CR2002-0891, March 22, 2002
BAW-10190P
Safety Evaluation For B&W Design Reactor Vessel Head Control Rod
Drive Mechanism Nozzle Cracking, May 1993
BAW-10190P,
Addendum 1
B&W Owners Group Proprietary, External Circumferential Crack Growth
Analysis for B&W-Design Reactor Vessel Head Control Rod Drive
Mechanism Nozzle Cracking, December 1993
BAW-10190P,
Addendum 2,
B&W Owners Group Proprietary, Safety Evaluation for Control Rod Drive
Mechanism Nozzle J-Groove Weld
BAW-2301
B&W Owners Group Proprietary, B&WOG Integrated Response to
Generic Letter 97-01, July 1997
51-5015818-00
Davis-Besse CRDM Nozzle Heat Information, 2002
51-125825-00
CRDM Nozzle Heat Treatment
51-1218440-00
Alloy PWSCC Time-to-Failure Models
51-1219143-00
CRDM Nozzle Characterization
51-1219275-01
CRDM Leakage Detection Evaluation
51-1229638-00
Boric Acid Corrosion Data Summary and Evaluation
51-1201160-00
Alloy 600 SCC Susceptibility: Scoping Study of Components at Crystal
River 3
Memorandum
Control Rod Drive Nozzle Cracking, May 8, 1996
Root Cause Plan
Dated March 18, 2002.
Intra-Company
Memorandum
Probable Cause Summary Report for CR2002-0891, March 22, 2002
Meeting Minutes
DBPRC Meeting Minutes for MOD 94-0025.
WO 00-001846-00
Work Order - Clean Reactor Head
AR-00-OUTAG-01
QA Audit Report Refueling Outage 12
Serial 2472
Davis-Besse Letter: Response to Generic Letter 97-01, July 28, 1997
Serial 2731
Davis-Besse Letter: Response to Bulletin 2001-01, September 4, 2001
Serial 2735
Davis-Besse Letter: Supplemental Response to Bulletin 2001-01, October
17, 2001
ASME Boiler and Pressure Vessel Code,Section XI, 1986
Edition and 1995 Edition with 1996 Addenda
25
Photographic Records
Video Tape
Davis-Besse Reactor Head Inspection Under Insulation Alloy 600, 12 RFO
Video Tape
Davis-Besse 12 RFO Final Head Inspection
Video Tape
Davis-Besse Reactor Head Cleaning 11 RFO
Video Tape
Davis-Besse Weep Hole Cleaning Nozzle 67, 10 RFO
Video Tape
Davis-Besse Weep Hole Video Inspection 10 RFO
Video Tape
13 RFO Reactor Head Nozzle Remote Visual Inspection
Video Tape
Root Cause Video of Nozzle #3 and Adjacent Nozzles, March 13, 2002 to
March 14, 2002
Video Tape