IR 05000324/1993015
| ML20035G933 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 04/19/1993 |
| From: | Blake J, Chou R, Lenahan J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20035G927 | List: |
| References | |
| 50-324-93-15, 50-325-93-15, NUDOCS 9304300196 | |
| Download: ML20035G933 (31) | |
Text
,_
.-.
. ~.
,
...
. - _
-
!
LOD Q UNITED STATES
.,
k NUCLEAR REGULATORY COMMisslON
!
o REGION 11
$.
101 MARIETTA STREET, NNV.
i
-
AT LANT A, GEORGI A 30323
'
s, *. o /
APP z ?
!
"
Report Nos.:
50-325/93-15 and 50-324/93-15
\\
Licensee: Carolina Power and Light Company
P. O. Box 1551 Raleigh, NC 27602 l
Docket Nos.:
50-325 and 50-324 License Nos.: DPR-71 and DPR-62 Facility Name: Brunswick 1 and 2
>
Inspection Conducted: March 1-5, 16-19, and 23-26, 1993
'
Inspector / >
-
^Y
." X l.enahan Date Signed i
j CC fat 4-[ fLjJ R. C. Chou (March 1-5 and 16-19)
Date Signed j
Approvedby!
f 3
'
/ Materials Processes Section
.,J. Blake, Chief Date Signed
i Engineering Branch l
Division of Reactor Safety _
!
SUMMARY
.
Scope:
.
This routine, announced inspection was conducted in the areas of repairs to the recirculation system pipe supports and Unit 2 drywell liner, the
miscellaneous structural steel verification program, inspection of masonry
.
walls, short term structural integrity items, evaluation of drilled-in i
anchors, and licensee action on previous inspection findings.
i
Results:
[
In the. areas inspected, violations or deviations were not identified.
j A weakness was identified in the licensee's design verification program -
Paragraph 3.c.
New Inspector Followup items are discussed in paragraphs 2.b and 7..d.
9304300196 930423 PDR ADDCK 05000324 G
!
. -
N
.
_.
_
-
_.. - - _
__
!
.
.
REPORT DETAILS l
1.
Persons Contacted i
!
R. Anderson, Vice-President, Brunswick Nuclear Plant
!
- H. Beane, Quality Control Manager
- S.
Callis, On-Site Licensing Representative
- T. Eason, Quality Control Supervisor
R. Godley, Manager, NRC Compliance
!
- L. Grzeck, Project Engineer, Miscellaneous Steel, Nuclear
Engineering Department (NED)
'
- C. Hinnant, Site Manager
,
T. Jones, Senior Specialist, Regulatory Compliance i
- R. Knott, Principal Engineer, NED
,
- J. Leinenger, Onsite Manager, NED
- W. Levis, Nanager, Regulatory Compliance
A. Lucas, Vice-President, NED
'
- G. Miller, Acting Manager, Technical Support
- W. Monroe, Supervisory Engineer, NED
'
- R. Morgan, Unit 1 Plant Manager
'
- G. Thearling, Senior Specialist, Regulatory Compliance
- J. Titrington, Unit 2 Operations Manager
'
R. Tripp, Supervisory Civil Engineer, NED
- S. Vann, Project Manager, Miscellaneous Steel, NED
C. Warren, Unit 2 Plant Manager
,
E. Willett, Manager, Outage Management and Modifications Other licensee employees contacted during this inspection included engineers, technicians, and administrative personnel.
!
!
'
Other Organizations R. Bizeck, Civil / Structural Consultant, TENERA
- R. Gallagher, Project Manager, Bechtel
- P. Dadlani, Site QA Manager, Bechtel T. Synder, Consultant, TENERA NRC Resident Inspectors
- R. Prevatte, Senior Resident Inspector
.
P. Byron, Resident Inspector
- D. Nelson, Resident Inspector
- Attended March 19 exit interview
- Attended March 26 exit interview
,
- Attended both exit interviews Acronyms and initialisms used in this report are listed in Paragraph 9.
.
_
. _ -
..
,
l
'
2.
Review of Licensee Corrective Actions to Resolve Structural Deficiencies I
In a letter dated July 23, 1992, Serial: NLS 92-160, Subject: Reply to
,
Inspection Report Nos. 50-325/92-12 and 50-324/92-12, the licensee
'
committed to completion of short term corrective actions listed in Enclosure 3 of the letter prior to startup. The inspector examined the short term corrective actions listed below.
,
a.
Item A-2, Walkdown Inspection to Examine STSI Items Affected by Corrosion.
Licensee Commitment Perform a third-party walkdown of non-pipe support, short term structural integrity (STSI) items and pipe supports in areas with high corrosion potential to validate design assumptions. Address any identified deficiencies in accordance with the methodology in Enclosure 2.
Discussion
'
!
The licensee retained EQE Consultants to perform an independent
'
review of pipe supports which were short term qualified, and of non-pipe support STSI items, to validate assumptions used in the STSI reviews. The non-pipe support STSI items will be addressed in closeout of short term corrective action item C-12 in a future
inspection.
The inspector reviewed calculation number ORHR-1004, Summary of-Closure on Pipe Support Corrosion STSI Items, which documents the results of the field walkdown inspections performed by EQE. This calculation included a summary of the review process; the field inspection data sheets; and the results/ conclusions of the field
,
walkdown inspections. The inspector reviewed the inspection data sheets which document the results of the inspections performed by EQE engineers on approximately 170 pipe supports which were short term qualified in May,1992.
EQE engineers identified 14, Unit 2 supports and 49, Unit I supports which exhibited some corrosion.
EQE concluded that the corrosion was not severe enough to affect the STSI evaluation.
However, EQE recommended that six, Unit 2 and ten, Unit I supports which exhibited the highest degree of corrosion be cleaned and reinspected by licensee engineers.
Licensee maintenance personnel removed the rust and corrosion from the 16 supports and licensee engineers reinspected the supports in late May,1992.
Licensee engineers concluded that the corrosion had not degraded the support to the extent that STSI evaluation would be affected.
The majority of the original Unit 2 supports which were short term qualified in May 1992 have been modified and are now long term
'
qualified. The inspector, accompanied by licensee engineers,
.
t 4.
walked down Unit 2 pipe supports which are still short term qualified on the service water and RHR systems, and independently inspected the supports for evidence of corrosion which could potentially affect the STSI evaluations.
Supports examined were as follows: RHR support numbers 2 Ell-18A38, 2E110-21A61, and 2 Ell-118PG78, and service water supports on elevation 50 of the Unit 2 reactor building and in the service water building.
Few of the i
supports exhibited corrosion. The inspector concurred with the conclusion that STSI evaluation for these supports would not be affected by corrosion.
Conclusion The licensee has inspected the Unit 2 pipe supports which were short term qualified and verified that corrosion of the existing supports would not affect the STSI evaluations.
Short term corrective action item A-2 is completed and acceptable for restart of Unit 2.
Field verification of non-pipe support STSI
'
items will be addressed in close-out of item C-12 in a future
inspection.
b.
Item C-1, Miscellaneous Structural Steel Verification Program Licensee Commitment Complete Unit I and Unit 2, Drywell Phase 2, miscellaneous steel
-)
wal kdowns. Complete preliminary bounding load studies. Address
'
repairs, as required in accordance with the methodology in Enclosure 2.
j Discussion Miscellaneous structural steel consists of platforms and other beams / columns which provide personnel access and/or support for piping, electrical raceways and conduits, HVAC ducts, instrumentation, and other equipment not supported from the main building structures. Numerous deficiencies in miscellaneous steel had been identified by either the licensee or NRC, including lack of design calculations, lack of as-built drawings, missing bolts and welds, incorrect size members, undersized welds, missing members, and other construction deficiencies. The licensee retained Bechtel Power Corporation to perform walkdown j
inspections, prepare as-built drawings, and perform design calculations to qualify the miscellaneous steel.
The Bechtel structural steel verification program, which is called the Miscellaneous Steel Verification Program (MSVP), is a two phase project with the purpose of establishing a high confidence that the miscellaneous steel is adequate for operation. The phase 1 program was a walkdown inspection to identify and evaluate any irregularities which could affect the integrity of the structures.
The phase II program involved obtaining detailed field
,
.
measurements to update design documents, prepare as-built drawings, performance of a detailed structural analysis, and preparation of a load tracking program to identify the magnitude
and location of loads.
i l
l The licensee has completed the Phase II program in the Unit 2 drywell, and in the RHR and HPCI rooms in Unit 2.
The Phase I program was completed for the remaining miscellaneous steel in the Unit 2 reactor building. The inspector previously inspected the
.
Bechtel program during inspections documented in NRC Inspection Report numbers 50-325,324/92-20, 92-23, 92-27, 92-33, 92-40, and 93-02.
l These inspections included review of walkdown and design procedures, review of completed walkdown inspection documentation,
,
independent NRC field walkdown inspections, review of design l
calculations, and inspection of completed modifications.
The load bounding studies were reviewed by the NRC Office of Nuclear Reactor Regulation (NRR), as documented in a Safety Evaluation Report (SER), dated October 8, 1992.
l During the current inspection, the inspector reviewed supporting documentation for implementation of corrective action to tighten l
bolts in slip critical connections on beams which had slotted holes in both ends. This work was completed under WR/JO 92-BGSPI (60 beams on El 17); 92-BGSQ1 (7 beams on El 38); 92-BGSR1 (4
)
l I
beams on El 52); and 92-BGSUI (5 beams on El 67). This documentation included the completed work requests, QC inspection
records, and bolt torquing data sheets completed by craft personnel.
'
The inspector reviewed supporting documentation for installation of lubrite plates on the beam seats for two radial beams on the elevation 17 platform steel.
The lubrite plates had not been installed during original construction. The new lubrite plates were installed under PM 92-077.
Installation details are shown on sketch numbers C-1088, sheets 1 and 2 of 2, and C-1088, Sheets I and 2 of 2.
Documentation reviewed included base metal repair data sheets, records for installation of new bolts and clip
!
'
angles, where required, and records for installation of the lubrite plates. No deficiencies were identified.
The inspector also reviewed results of Audits performed by Bechtel i
l Quality Assurance (QA) Personnel. These included audits of field walkdown activities, record keeping, procedure implementation, and completion of calculations.
Corrective action was adequate to address the audit findings.
In addition, the inspector reviewed qualification records fce 8 Bechtel field engineers and 7 Bechtel design engineers.
The inspector concluded that the experience and education levels for these individuals complied with the project requirements.
j
-
.
'
a
'
l The inspector reviewed the final report of the Technical Advisory
.
Committee (TAC), and the Structural Steel Start-Up Report. The
.
Start-Up Report summarized the MSVP, and included justification for acceptability of inaccessible structural steel members. The report also summarized the results of MSVP, including number of irregularities identified, number requiring modification for operability reasons, and recommended modifications to comply with
good engineering practices. All modification required for
'
operability reasons have been implemented. The TAC Report summaries the conclusions and recommendations of the TAC and included acceptance criteria, guidelines for addressing generic irregularities, and structural analysis criteria and methodolugy.
Conclusion The licensee has satisfactorily implemented the miscellaneous structural steel verification program for restart of Unit 2.
Short term corrective action item C-I is completed and acceptable for restart of Unit 2.
The inspector will follow up on the completion of the Phase II program for the remaining structural l
steel in the Unit 2 reactor building in a future inspection. The l
Phase 11 program will be completed by the next refueling outage.
This will be tracked as Inspector Followup Item 324/93-15-01, j
Follow-up on Phase II of MSVP for Unit 2 Reactor Building i
Structural Steel.
c.
Item C-6, Integrity Inspection of Unreinforced Masonry Walls Licensee Commitment Perform an integrity inspection (i.e., for cracks, general condition) of unreinforced masonry walls that are classified as
'
safety-related. Address any identified deficiencies in accordance with the methodology in Enclosure 2.
Discussion
During the design review of masonry walls, the licensee performed field inspections of safety-related unreinforced masonry walls to determine the general condition of the walls, and determine if cracks existed in the walls which could effect the structural integrity of the walls.
Licensee engineers also performed an investigation of the masonry walls to determine if the reinforcing steel (rebar) was installed as shown on the design drawings.
These examinations were performed using non-destructive testing (NDE) equipment, specifically, a "Rebar detector".
The results of the initial NDE tests were inconclusive, but licensee engineers determined, after further investigations, that the rebar was more-or-less installed in the masonry wall as required by the design drawings.
l l
!
l i
.
'4
The inspector reviewed the records summarizing the results of the licensee's inspections of the masonry walls for installed rebar and existence of cracks in the mortar and/or masonry blocks. The inspector, accompanied by a licensee engineer, walked down the following walls, inspected the walls for cracks, and independently verified the presence of rebar using the rebar detector. Walls inspected were as follows:
-
Control building walls containing vertical reinforcing steel-wall numbers IB,10,10, IE, 6D, 70, and 10B.
-
Control building walls containing horizontal reinforcement (duro wall) - wall numbers 2, 4, 9A, 9B, 9C, and 9D.
-
Diesel generator building walls containing vertical reinforcing steel - wall numbers IC, 2A, 2D, 3H, 6A, 9E, and 10A.
The inspector concluded that the rebar was installed in accordance with the details shown on the construction drawings, and met design requirements, although the NDE equipment / method had limitations. The inspector also verified that existing cracks in the masonry walls had been identified by licensee engineers. The inspector reviewed the design calculations for walls which had cracks and verified that the presence of the cracks had been appropriately considered in the design evaluations for these walls.
Conclusion The masonry walls were inspected by licensee engineers in accordance with their commitments. The results of these inspections were considered in the design evaluations of the walls. Short term corrective action item C-6 is complete and acceptable for restart of Units 1 and 2.
d.
Item C-10, Review IE Bulletin 79-02 Inspection Results Licensee Commitment Review the IE Bulletin 79-02 (Pipe Support Anchor) Program to ensure compliance and to ensure methods of inspection used would have detected deficient bolt installation.
Discussion After NRC became aware of the improperly installed concrete expansion anchors (i.e., the Fake bolts) in the diesel generator i
building, masonry block wall, lateral supports, the licensee was questioned by NRC regarding the adequacy of othar concrete i
-
.
,
i I
expansion anchor installations. These installations included pipe supports, conduit and cable tray supports, floor-mounted equipment anchorages, HVAC supports, and structural steel supports.
The licensee had examined concrete expansion anchors in pipe supports in response to IE Bulletin 79-02, Pipe Support Base Plate Design Using Concrete Expansion Anchors.
In order to resolve any questions regarding the adequacy of the expansion anchors for pipe supports, the licensee performed an audit of their IE Bulletin
_
79-02 program. This audit included a review of the field
'
inspection / testing procedure; a review of the pipe support anchor
,
sampling program; a review to determine if the testing procedure
was properly implemented; and a review of the test results.
The results of the audit were transmitted to the NRC\\NRR in CP&L letter, Serial: NS-92-Il8 dated April 15, 1992.
The conclusions
of the audit were that the inspection of pipe support expansion anchors, performed under IEB 79-02 was adequate, and would have
detected improperly installed anchors. NRR reviewed the
licensee's April 15, 1992, letter and requested additional i
information in a letter to CP&L dated April 27, 1992.
The licensee responded to these questions in a letter dated July 16,
,
1992, Serial: NLS-92-136.
Based on these responses, NRR determined that pipe support expansion anchors had been adequately
,
inspected. The inspector also reviewed the results of the
!
license's audit during an inspection documented in NRC Inspection Report number 50-325,324/92-14, and concurred with these findings.
l
'
In order to resolve questions regarding the adequacy of installation of anchors which support other hardware, (e.g.
>
conduit, HVAC, structural steel, etc.), the licensee committed to inspect / test a sample of other concrete expansion anchors. The anchor sampling, testing, and inspecting was performed using Design Guide III.17, NED Design Guide for Inspection of Drilled in Anchors.
,
The inspector reviewed the procedure and sample selection;
,
observed testing / inspection of some anchors; and examined testing / inspection results during inspections documented in NRC
,
Inspection Report numbers 50-325,324/92-20 and 92-23.
The licensee determined that over 12000 expansion anchors had been i
installed in safety-related buildings during original plant
!
construction to support HVAC, conduit, cable tray, floor mounted
!
equipment, and miscellaneous structural steel. Ultrasonic testing
'
was performed on approximately 4300 of the anchors. One short bolt was identified. This was not a " Fake" bolt. The short bolt will be replaced under a routine work request, so that the bolt
has proper thread engagement in the anchor shell. Approximately
'
500 of the anchors were inspected by removing or attempting to remove the installed bolts to verify the presence of the shells.
One missing shell was identified. Some of the bolts could not be
,
l
!
.
.
_
_._
_
_
_ _
-
-
l
1 removed without damaging the installed anchor.
In most of these cases, the bolts were not removed because of the difficulty in replacing a damaged anchor.
The licensee documented the overall results and conclusion of this program in CP&L NED Report number 0MISCB-1001, Drilled-In Anchor
Inspections, dated December, 1992.
During the current inspection, the inspector reviewed the report l
and evaluated the anchor inspection / test data. The data shows that the occurrence of fake bolts was limited to the diesel i
generator wall anchors, although a few improperly installed anchors were identified: one with a short bolt, one with no shell,
$
and one where the shell rotated.
,
l Conclusions The licensee concluded that no safety concerns existed due to potentially deliberate anchor bolt mis-installations. The inspector concurs with this conclusion.
Short term corrective action item C-10 is complete and acceptable
'
for restart of Units 1 and 2.
,
i e.
Item C-II, Service Water Pump Calculations i
licensee Connitment
'
Complete field inspections to assure that calculations supporting interim seismic operability of the service water system pumps are valid. Address any identified deficiencies in accordance with the i
methodology in Enclosure 2.
Discussion Review of the service water pump calculations was performed during an inspection documented in NRC Inspection Report number
'
50 325,324/93-02. An unresolved item, number 325,324/93-02-03,
'
was identified regarding these calculations. The unresolved item was closed out during the current inspection (See paragraph 7.f below.)
Conclusion
!
Based on the results of field inspections performed during the current inspection and inspection 93-02, and review of the
'
calculations, the inspectors concluded that short term corrective i
action item C-Il is complete and acceptable for restart of Units 1 i
and 2.
.
!
!
r e
-
-
.
.
.
.
-
f.
Item C-13, Repairs to Unit 2 Recirculation System Hangers Licensee Commitment:
Complete repairs of recirculation system ring header hangers.
,
Discussion:
After problems were identified at another boiling water reactor, the Licensee performed a review of the recirculation system ring header supports. Their review disclosed the need to reinforce the
'
existing structural supports for 10 spring cans on the
'
recirculation system, one per riser. The licensee added reinforcing gusset plates to strengthen the supports under plant modification PM 91-041, Field Revision 28.
,
The inspectors, accompanied by licensee engineers, examined the 10 spring can pipe supports which had previously been modified under PM 91-041 to correct identified deficiencies. Portions of the supports were examined using the QC verified drawings issued for the completed modifications. Attributes inspected were new weld sizes, new members, new bolts, and dimensions. The inspectors were only able to perform cursory visual inspection of some supports since some supports were partially inaccessible during
'
the inspection. Most new weld sizes exceeded the weld sizes specified on the modification drawings. The supports inspected and the design drawings were as follows:
Item No.
Mark No.
Drawina No.
Rev. No.
HAl 9527-F-1906 B
HA2 9527-F-1906 A
HA3 FP-5719, 5872, 5873 A
HA4 9527-F-1913 A
l
HA7 9527-F-190C A
HB1 9527-F-1906 B
HB3 9527-F-1913 A
HB4 FP-5720, 5874, 5875 A
HB6 9527-F-1906 B
HB7 9527-F-1906 A
No discrepancies were identified.
Conclusion:
Modifications to the recirculation system ring header hangers have been completed on Unit 2.
Item C-13 is complete and acceptable for restart of Unit 2.
>
.
+
l l
i 3.
Short Term Structural Integrity (STSI)
,
STSI items are those identified by licensee personnel, which, after
evaluations by NED, are determined to be operable, although they do not j
meet design criteria established by the FSAR. During previous NRC inspections, questions were raised regarding STSI design criteria; field validation of critical design assumptions; and STSI evaluation techniques.
In order to address these questions, the licensee retained EGE Engineering Consultant, an independent firm, to perform a third party review of STSI items.
The inspectors reviewed STSI design criteria, reviewed the EQE report summarizing the third party review, and independently reviewed selected STSI design calculation. Details of the inspection are as follows:
'
!
a.
Review of STSI Design Criteria
l STSI operability reviews are performed in accordance witn Design Guide 11.20, Civil / Structural Operability Review.
Revision 2 of DG 11.20 was reviewed by NRR, and found to be acceptable, as documented in an SER, dated' October 8, 1992.
l During the current inspection, the inspector reviewed DG II.20,
!
Revision 4, dated October 24, 1992. This review disclosed that the licensee had revised the STSI evaluation methodology for piping systems and had increased the allowable stress values used as acceptance criteria for structural steel from 1.5 times AISC allowable limits to 1.6 times AISC allowable limits.
(Subsequent to the inspection, on April 1, 1993, these revisions i
to DG 11.20 were discussed with NRR personnel at a meeting held at
'
NRC Headquarters at One White Flint in Rockville, Maryland. The i
results of NRR's review will be documented in an SER supplement.)
b.
Third Party Review of STSI Items The inspector reviewed EQE Consultants report number 52171.03-R-001, " Brunswick Nuclear Station Third Party Review of Potential Advance Structural-Mechanical Conditions", Rev. O, dated July 31,
.
1992. The report summarizes the third party review of STSI items, including field validation of design assumptions, and review of communications between the Technical Support organizatica and NED.
The EQE review focused on the current Brunswick Nuclear Plant (BNP) process, from identification of potential adverse conditions through evaluation and final closure. The review included independent assessment of steps in the evaluation of potential non-conforming items, such as analysis, analysis assumptions, l
conclusions in operability determination, documentation, and field i
condition of structures and components.
The scope of review included:
.
_ _
.
_
,
!
,
F h
l
!
-
The current processes and procedures for identification and evaluation of potential adverse conditions.
l
-
A sample of Technical Support Memoranda, Work Requests / Job Orders, Adverse Condition /Non-Conformance Reports, STSI Items, and Engineering Evaluation Reports (EER) related to evaluation of structural-mechanical concerns.
Documentation involved in the identification, evaluation,
-
and disposition of potential adverse conditions.
Site memoranda (Blue Memos) are written for a variety of reasons.
These include requests for information, authorization of funds, solicitation of engineering reviews of unusual plant conditions, i
and communications between the piant and other CP&L organizations.
,
four hundred out of 1400 memoranda written between January 1989, t
and April 1992, addressed structural-mechanical issues.
Forty of
400 were selected for review. The table below summarizes the documents reviewed by EQE.
,
Summary of Documents Reviewed by EQE i
Document Type Population Size Number Reviewed ACR/NRCs
6 Site Memoranda 400
'
STSI 179
WR/J0s 80+
EERs (associated with above)
,
EQE also reviewed CP&L procedures covering activities of the
Nuclear Engineering Department and Brunswick site procedures covering activities of plant operations, technical support, and
,
maintenance, related to requests for NED support in resolution of
'
problems.
I Six recommendations were listed by EQE to improve the processes in the identification, evaluation, and documentation of future conditions:
-
A review should be performed to identify site to engineering memoranda (Blue Memos).
-
Two technical and safety issues identified in this review
!
should receive further evaluation.
-
Supplemental training should be provided to site personnel
!
involved in the processing of potential adverse and non-conforming conditions.
l
,
]
l
._,
-
_.
..
.
.
'
'
.
!
,
-
Supplemental training should be provided to NED staff in the use of site memoranda.
!
Site memoranda identified in this review may be required to
-
be transformed to become part of the component qualification
documentation.
-
NED and DTOP procedures with conflicting criteria for the operability assessments should be reviewed.
The inspector will review implementation of these recommendations
in the close-out of short term corrective action item C-12 in a future inspection.
c.
Review of STSI Calculations
(1)
STSI item C-12, RHR Support Qualification on Isometric
Drawings 002 and 003.
Per the Design Turnover Program (DTOP), discussed in
,
paragraph 7.a. below, CP&L is in the process of reviewing
!
all piping design documents and records to evaluate the
design methods and processes with the established design
!
criteria, specifications, and as-built field conditions for
piping systems.
If the design review shows the piping does i
.
not meet the long term design criteria (for long term
structural integrity), a short term structural integrity review is performed in accordance with DG 11.20. This results in the following actions:
I If a component does not meet the long term design
-
criteria but meets the short term design criteria, a
'
field modification document will be issued for a i
scheduled modification to be implemented by the next refueling outage,
l If a component does not meet the short term design
'
-
criteria, a field modification document will be issued and the component will be fixed before restart.
'
The CP&L DTOP review on RHR Isometric drawings 002 and 003 resulted in 12 supports being identified on Iso. 002 as STSI and 22 supports on Iso. 003 being identified as STSI.
EER 91-0159 was issued to document the long term fixes required
,
for the supports on Iso. 002 while EER 91-0160 was issued to
document the long term fixes required for the supports on
!
Iso. 003. Some of the long term fixes have been completed on these supports. The inspectors reviewed the supporting
,
calculations for EER Nos. 91-0159 and 91-0160 and the support modification drawings. No discrepancies were
identified.
'
i
-
.
..
._
_
.
13 (2)
STSI Item No. 169, fuel Oil Diesel Piping on Iso. 109 Fuel oil diesel piping currently attaches to various non-safety related structural steel platforms. The licensee was required to qualify support of the safety-related fuel oil diesel piping, on the non-safety related platform steel, based on the STSI criteria. A long term fix was designed to support the piping from safety related structures.
CP&L calculation No. ODGB-0002, Rev. O, which covers short term qualification of diesel fuel oil piping supported from a non-safety related, structural steel platform, was randomly selected by the inspector for review. The review included load assumptions, load distributions to the support beams from gratings, input and output of steel member model, and input and output of base plate models.
The inspector's review disclosed the following discrepancies:
The load distribution to the support beam from
-
gratings was incorrect.
The calculation uses load area method to distribute the grating loads to four beams in the rectangular configuration. The calculation assumes four areas divided by drawing two lines from each opposite corners, which results in triangular areas of uniform loads acting on each beam.
Standard design practice for distribution of grating loads is to assume that only the two beams in the longer span (in the normal condition) will carry the total load of the entire rectangular area since grating is only supported in one direction and acts as a simple span beam. The small diameter steel bars in the other direction of the grating serve to connect the long bars and serve as lateral bracing. Those small steel bars (about 1-inch diameter compared to the one-inch deep bars) cannot redistribute loads to the two beams in the short direction. The method used in the calculation reduced the loads to the long beams and added the loads to the short beams.
-
The enveloped input load used for the z-direction in the base plate model (Attachment E) was 101 lbs. The required enveloped input load should be 453 lbs. The coordinate systems for platform members (STRUDL) and base plates (ME-035) were different base plate locations (walls and floors). The shear load increase will slightly affect the bolt interaction ratio since the bolt tension due to the bending moment is a controlling loa i
.
I
.
,
i
However, based on review of the calculation, the inspectors I
concluded that these discrepancies would not affect the overall results/ conclusions. The platform members, base
plates, and base plate thru-bolts or anchor bolts have i
sufficient margin for the above discrepancies.
'
(3)
STSI Item No. 183, Diesel Generator Jacket Water System The DTOP evaluation of the piping on stress Isos for this system l
determined that the piping and pipe supports could not be long l
term qualified.
In addition, a through wall leak was discovered on this line on December 13, 1990. Therefore, the minimum piping
)
wall thickness requirements, and the qualification of piping and supports were the two subjects for STSI item 183.
The licensee generated calculation No. 2-15-34A-278, Rev. O, to investigate a leak in line 2-SW-234-6-157 and determined that the
,
pipe stresses at a single weld were acceptable based on the reduction in the wall thickness.
Several EER's were written to examine, on a continuing basis, the leak locations. The licensee generated stress calculation No. 0-1543A-283 and support I
calculations Nos. PS-DJW-52A to 52F, 53A, and 53B to qualify piping and supports with the STSI criteria. The licensee determined that all the calculations met STSI requirements.
EQE Engineering Consultants reviewed these STSI calculations as part of third party review for STSI items.
EQE's review examined anchorage, support assumptions, loading conditions, modeling assumptions, design input, material strength assumptions, etc.
EQE had the following comments on the calculations:
(1)
Calculation PS-DJW-52C assumes that an angle member is embedded a depth of 16 inches in the wall, but the wall is only 12 inches thick.
(2)
Ultrasonic test (UT) examinations of the piping should be expanded to assess pipe wall thickness at all critical points in the piping system, not just the previous leakage locations.
The licensee performed UT for the above angle embedment lengths and the pipe wall thickness at all critical points.
The test results were acceptable and resolved the EQE concerns.
'
The inspectors randomly selected the support calculations Nos.
PS-DJW-52C, Rev. I and PS-DJW-53B, Rev. O for review.
Each calculation covers one isometric drawing and includes all the supports in the isometric.
During the review of these calculations, the inspectors identified the following discrepancy:
j j
.
15
-
Calculation No. PS-DJW-52C, Rev.1, Page 18 of 40, a 16-inch
embedded angle length was used to calculate the allowable pull out load for the embedded angle instead of the actual 8.575 inch embedded length per Ultrasonic Test Examination.
The embedded angle is still qualified for the applied load by using the actually embedded length since the applied load is small compared to the allowable loads.
This discrepancy was considered minor and the calculations are acceptable. The licensee will revise the calculations to correct i
the discrepancies identified by the inspector and EQE.
The licensee used the seismic coefficients or factors derived for the diesel generator building to qualify the piping and supports l
for the diesel generator tank building. The licensee is currently developing the seismic coefficients for the diesel generator tank building and will revise all the piping and support calculations if the coefficients are significantly different from the coefficients for the diesel generator building.
t The inspectors concluded that STSI No. 183 is acceptable.
However, the errors in the STSI calculations discussed above, and the errors discussed under Unresolved Item 325,324/93-02-03,in paragraph 7.f below are indicative of a weakness in the licensee's design verification program. The inspectors discussed the need for personnel involved in design checking / verification to perform
-
independent verification of the accuracy of design calce,lations and verify the correctness of assumptions made in the design calculations.
Violation or deviations were not identified.
4.
Repairs to Unit 2 Drywell Liner During the inspection documented in NRC Inspection Report No. 50-325, 324/93-02, the inspector identified a problem with corrosion of the
,
drywell liner plate at the intersection of the liner with the elevation 5 concrete floor, around the entire circumference of the drywell. A violation, item No. 325,324/93-02-01, was identified regarding failure of the licensee to measure and evaluate the corrosion. After the corrosion problem was identified by the inspector, the licensee performed extensive inspections of the corrosion, and issued EER No. 93-0173 which evaluated the corrosion and outlined the repairs required to the liner. During inspection 93-02, the inspector reviewed the results of the licensee's investigation of the corrosion, performed field walkdown inspections to evaluate the licensee's corrosion investigations, and examined weld repairs required to restore the liner plate to the required minimum thicknes.
'
,
During the current inspection, the inspector examined the completed coating (painting) of the liner, and the installation of the seal which had been installed after the coatings were completed to cover the expansion joint between the liner plate and concrete floor. No i
deficiencies were identified with the completed coatings or the installed seal.
i The inspector reviewed records documenting surveillance inspections which had been performed by quality control inspectors to monitor the application and curing of the coatings. The QC inspection personnel identified a problem regarding inadequate thickness of the primer coat
on the bare steel liner plate. This problem was corrected by recoating the liner plate with another coat of primer.
'
Based on review of the quality control records, and inspection of completed repairs to the liner plate, the inspector concluded that the Unit 2 liner corrosion problem has been adequately corrected for restart
of Unit 2.
Violations or deviations were not identified.
5.
Design / Construction Concern Concern: During a review of a " Projects in Working" list, the inspector questioned Item 5 on the list which referenced NCR A-89-013, Deviation
)
from Ductwork Support Design Drawings. The " Projects in Working" list
was prepared when the onsite QA group was disbanded and replaced by the Nuclear Assessment Department (NAD).
,
Discussion: The inspector reviewed the nonconformance report (NCR A-89-013) which was written to document problems identified with
HVAC supports in the control building battery rooms. These HVAC
'
ducts / supports had been repaired; the concern on the
,
" projects-in-working" list was the other HVAC duct problems which had not been corrected.
In order to close out NCR A-89-013, the licensee retained EQE
Engineering Consultants to perform a walkdown inspection of HVAC ducts and supports and evaluate nonconforming conditions. The inspector
-
reviewed EQE report number 52029.03-R-002, Evaluation of HVAC Ducts and Supports at BSEP. The overall conclusions of the report were that
,
existing discrepancies did not render the HVAC system inoperable.
although EQE identified some outliers from the Seismic Qualification Utility Group (SQUG), Generic Implementing Procedure, Rev. 2. (GIP-2)
(GIP-2 is the NRR approved procedure for performance of inspections in response to unresolved safety issue USl A-46, Seismic Qualification of Equipment in Operating Plants.)
The licensee initiated calculation OVA-0020, Resolution of NCR A-89-013, to close out the NCR. The inspector concurs with the conclusion of this calculation, and the EQE report, that the deficiencies identified in the l
6
-
.
i
HVAC ducts and supports did not affect operability of the HVAC system;
!
although long term repairs need to be completed to restore design i
margin.
Portions of the HVAC ductwork and supports located in the Unit 2 reactor
,
building, on elevation 20, were replaced during the current outage.
l Other HVAC deficiencies have been identified during the
!
"Hotside/Coldside Walkdowns" and during the " Material Condition l
Walkdowns." These deficiencies were either repaired or subjected to the
screening process in procedure PN-30, Integrated Recovery Methodology.
~
The inspector reviewed randomly-selected, walkdown findings, pertaining l
to HVAC, and verified that they were properly evaluated in accordance with PN-30.
The inspector concluded the HVAC system is acceptable for
'
restart of Unit 2.
Violations or deviations were not identified.
.
'
6.
Licensee Event Reports (LER)
i a.
(0 pen) LER 2-92-006: Control Rod Drive System Scram Discharge
Volume Instrument Line Pipe Supports Were Found Missing.
In August, 1992, during inspection of portions of the scram discharge volume (SDV) piping systems, supports were found to be i
missing from the instrument volume lines in both units. On the i
Unit I piping, one support was found missing (actually it was
installed on the wrong pipe), while another support was incorrect (support was lateral support only, while the drawing called for a l
lateral / vertical support). The problem was documented as Adverse Condition Report (ACR) numbers N 92-112 and B 92-661.
Licensee engineers were able to short term qualify this piping under Design Guide 11.20 for operability.
Inspection of the same piping on Unit 2 disclosed two missing supports. The concrete anchors for the supports were in place, but the two supports were missing. This piping could not be qualified for operability with two missing supports.
This problem was documented in ACR B-92-674. The licensee submitted a written report on this LER to NRC in a letter dated September 21, 1992.
The licensee's corrective actions included replacing the missing supports on both Units, and walkdown inspections of a portion of the control rod drive (CRD) piping to determine if additional supports had been removed, and to identify other non-conforming conditions.
The inspector reviewed the results of the licensee's CRD support reinspection program, and inspected repair of the missing Units 1 and 2 supports during an inspection documented in NRC Inspection Report Nos. 50-325,324/92-4 _
4.
.
,
.
,
The results of the licensee's reinspection were as follows:
-
Unit 1, 179 of 209 supports reinspected, 46 discrepancies
'
identified;
,
-
Unit 2, 206 of 313 supports reinspected, 102 discrepancies identified.
,
Although none of the discrepancias rendered the piping seismically inoperable, the licensee decided to expand the inspection effort to include all CRD supports in both Units.
During the current inspection, the inspector reviewed the results of the licensee's reinspection of the remaining CRD piping supports. A summary of these results is as follows:
-
Unit 1,110 additional supports inspected, discrepancies identified on 36 additional supports, for a total of 82 discrepancies on the 289 Unit I supports;
-
Unit 2,107 additional supports inspected, discrepancies identified on 77 additional supports, for total of 179 discrepancies on the 313 Unit 2 supports.
The majority of the discrepancies involved loose hardware, clearance problems, or dimension / location problems. The discrepancies which have not been corrected and could not be long term qualified were identified as STSI item number 214 for Unit 2 and number 215 for Unit 1.
The licensee was unable to determine why the CRD pipe supports had been removed. After a review of various plant records, the licensee concluded that the supports may have been removed to install new equipment under a plant modification.
Additional corrective actions to resolve this LER included review of maintenance procedures to verify the procedural controls were adequate to control maintenance activities and training of craft
personnel to emphasize that permanent plant equipment could not be
,
altered or removed if it was not specified in a work request
]
(WR/J0). This training is scheduled to be completed by April 30, 1993. The procedure review identified that procedure MAP-004, Modification Work Control, should be revised to add procedural control enhancements for implementation of plant modifications to cover temporary removal of plant equipment. The licensee committed to revise this procedure by March 31, 1993.
LER 2-92-006 will remain open pending completion of the procedure revisions and training. All hardware deficiencies have either be corrected or have been evaluated under the STSI program and will be corrected by the next refueling outage. Corrective action completed to date on this LER is acceptable for restart of Unit 2.
. - -
.
.
b.
(0 pen) LER l-92-012, Emergency Diesel Generator Building Internal Wall Seismic Support Bolting was Defectively Installed During Plant Construction.
In April, 1992, the licensee determined that a significant number of concrete expansion anchors which were required to provide lateral support to interior walls in the diesel generator building had been improperly installed during original construction. This resulted in a determination that several walls were inoperable, which affected critical safety related equipment, resulting in shutdown of both Units. The problem was reported verbally to NRC Region II on April 21, 1992, with written followup reports submitted to NRC in letters, dated May 2?, 1992, June 26, 1992, (Supplement 1) and August 31, 1992 (Supplement 2).
As a result of the deficiencies identified with the diesel generator walls, an extensive program was undertaken by the licensee to evaluate and modify, if required, all masonry walls in safety-related structures in both Units. This resulted in re-evaluation of IE Bulletin 80-11, Masonry Wall Design, and an in-depth review of the design function and design parameters for all masonry walls in proximity of safety related equipment.
In
,
l Enclosure 3 to the July 23, 1992, letter, referenced in paragraph 2, above, the licensee identified short term corrective actions C-2 through C-8 for correction of the structural deficiencies identified with the masonry walls. Closeout of these items, with the exception of item C-6, is documented in NRC Inspection Report numbers 50-325,324/93-02.
Item C-6 is closed in paragraph 2.c, above.
During the structural repairs to the walls, the licensee discovered that the material sealing the expansion joint at the top of the walls (i.e., between the top of the masonry block and the bottom of the reinforced concrete floor slab) was non-fire-rated. The licensee is evaluating methods to repair / replace the expansion joint material. This activity will continue after startup of Unit 2.
Compensatory measures will be used to meet the requirements of the fire protection program using fire watch personnel until the repairs are completed.
This LER will remain open pending completion of repair / replacement of fire stop materials.
Structural repairs are completed and acceptable for restart of Unit 2.
7.
Licensee Action on Previous Inspection Findings.
a.
(0 pen) Inspector Follow-up Item 325,324/92-14-02, Complete Evaluation and repairs to Pipe Supports and Closeout of NCR S-86-021.
..
.
. _ - -
.
... -
- -
..
- - -.
.
=
.
..
l Nonconformance Report (NCR) number S-86-021 was generated due to
discrepancies between the installed pipe supports and the as-built
]
drawings for safety-related piping system.
The as-built drawings
!
were completed as part of IEB 79-14, Seismic analysis for As-Built c
Safety Related Piping Systems. The licensee has initiated a l
Design Turnover Project (DIOP), Phase II, to walkdown and re-I analyze the safety related piping systems and correct the as-built drawings to disposition NCR S-86-021. DTOP has been reviewed by
,
NRC Region II inspectors in 1987 through 1990 as part of the close l
out of IEB 79-14. The licensee originally proposed a date of December 1991 to close out the DTOP design work.
Based on this proposed schedule, IEB 79-14 was closed by NRC. However, budget
,
cuts and reduction in staffing extended the complation date to -
l December 1992.
IFI 325, 324/92-14-02 was identified by the
inspector to track timely closecut of the DTOP Program and to i
follow up on completion of repairs / modifications to pipe supports
.l as a result of DTOP design work.
j
,
During the current inspection, the inspector reviewed the schedule
for completion of the DTOP design work with the responsible CP&L
!
,
principal engineer. The schedule currently calls.for completion
of design work on large bore piping systems by July,1993, and on
small bore piping systems by October,1993. The licensee plans to
!
i complete installation of the required modifications by December, 1994. The completion of the DTOP program and the resulting i
emergent structural repairs is addressed in the licensee's-Brunswick Three-Year Plan (1993-1995) which was submitted to NRC on December 15, 1992.
i l
The IFI will remain open pending closeout of the DTOP program.
,
b.
(0 pen) Inspector Followup Item 50-325,324/92-18-02, Feed Water-l Line Problem Due to Transient i
This item involved recurring support damage and large
.
displacements in feed water lines due to transients. The licensee
,
identified the problem, modified the damaged supports, and
increased the capacity of some supports. However,' support 2EX-
'
9PG-02 was re-damaged after modification. The licensee then decided to re-estimate the transient loads, increase the support-
- i capacity, remodify the damaged supports, and add new supports.to l
'
restrain axial and! lateral loads at the proper locations.
In addition, the licensee decided to review and identify the j
magnitude of potential transients-in all suspect lines.
j During a previous inspection, the inspector also determined that
support deficiencies occurred on other lines adjacent to the feed
!
water lines. The licensee revised EER NO. 92-0025 to Rev. 2 to include the root cause analysis and the corrective' actions to be taken. The licensee also revised stress calculations based on the
.
"
new estimated transient loads.
.
.
_. _, _ _ -
_,
,
_
.
.
I
Support loads based on the stress calculations for existing lateral supports PS-77111 and 2EX0-9PG63 (this support name replaced the previous support named 2EX-9PG62) were approximately
!
10 kips.
These supports were designed and modified to an increased capacity of 20 kips. Two additional supports (one for axial loads and another one for lateral loads) were added to the system to restrain transient loadings. These were support numbers 2EX-9PG64 and 2E-9PG65. A transient load of 18.5 kips was predicted and assumed in the structural dynamic analysis based on the observed movement since the new analysis for the transient loads is not completed yet.
The design capacity for the new supports are two times the estimated actual loads derived from the L
new transient analysis.
The licensee will review the operation of the extraction steam and heater drain systems, and implement new procedure changes, if necessary, to reduce transient loads. A program will also Le developed to determine if other secondary systems are subject to damaging transients.
The licensee has taken adequate steps to correct the feedwater line problem, due to transients, for restart of Unit 2.
This item j
remains open pending review of the following resolutions by NRC:
-
The new transient load which is being developed for the feed
,
water piping.
,
-
Review of any operational procedures changes for systems operation.
!
-
Review of the program being developed to review possible transients in other secondary systems.
,
,
-
A field walkdown during the next refueling outage to review any physical damage on the line after one cycle of operation.
'
c.
(Closed) Violation 324/92-27-02, Phase II Walkdown Deficiencies.
The licensee originally responded to this violation in a letter dated November 25, 1992, Serial BSEP 92-0045. This response was withdrawn by CP&L on December 10, 1992. The licensee resubmitted their response to the Notice of Violation in a letter dated December 22, 1992, Subject: Brunswick Steam Electric Plant, Reply i
to Notice of Violation; Serial: BSEP-92-0055.
i The violation contained five examples of failure to follow j
procedures or inadequate procedures pertaining to performance of
the Ph se Il miscellaneous steel walkdown inspections. The corrective actions are discussed below.
i
.
.-.
-
,
.
'
(1)
Incorrect classification of welds as partial penetration
-
The incorrect classification of the welds was partially
!
attributed to procedure inadequacies. Specifically, Bechtel procedure WDP-002 did not contain criteria for inspection of partial penetration welds. Appendix A of the procedure was revised, in Revision 3, issued October 26, 1992, to emphasize field verification of significant weld attributes,
!
including identification of groove or partial penetration welds.
Training was also conducted for the field welding inspectors (engineers) to cover the revised inspection criteria.
The results of the walkdown inspections performed for the
.
Unit 2 drywell structural steel was reviewed. This review resulted in identification of 117 connections where the original construction details specified partial penetration welds. These 117 connections were reinspected and the results of the reinspection were documented in a Bechtel
memorandum dated December 14, 1992, to R. E. Gallagher, File No. 0308, Subject: Partial Penetration Welds. Of the i
population of 117 welds, 58 were evaluated and found to be acceptable.
The remaining 59 were modified with a fillet i
weld reinforcement.
The inspector examined some of the modified welds during an
'
inspection documented in NRC Inspection Report numbers 50-325,324/92-40. During several previous inspections, the i
inspector also reviewed the revised welding inspections
'
procedure.
'
(2)
Other Errors in Walkdown Inspections This violation example involved failure of walkdown personnel to identify and document a missing bolt in a connection during the Phase II walkdowns. Contributing to
the cause of this violation was failure of the walkdown l
personnel to pay strict attention to details, and an over i
reliance of Bechtel and licensee supervisory personnel on I
use of photographs of connections to review completed structural steel field inspection /walkdown records.
t
i After the violation was identified, additional training was provided to walkdown personnel to emphasize the need to r
reconfirm field data prior to signoff of the inspection data
!
sheets.
In addition, licensee QC inspectors performed field I
walkdown inspections to determine accuracy of the completed
]
field inspection data sheets, and Bechtel field engineers i
field-verified approximately 10 percent of the completed
,
walkdown records.
)
.-
-
--
--
-
- -
.-.
.
.
,
!
!
!
During previous inspections, the inspector identified additional walkdown inspection documentation errors, and reviewed the results of the licensee and Bechtel field verification of structural steel walkdown documentation.
The results of the Bechtel field verification program are documented in a Bechtel Report dated January, 1993, titled, Phase II Verification Programs-Report of Results for Unit 2 Drywell and RHR Areas.
The inspector reviewed the report which concluded that the Phase II walkdown results were acceptable.
The data attached to the reports showed that
'.
the error rate was approximately one percent.
The licensee also retained independent consultant; t: review the Bechtel report and data, and to consider the resuits of errors found by CP&L QC inspectors and NRC in their reviews.
i The inspector reviewed the consultant's data base and conclusions.
These conclusions were as follows:
-
Given the large number of components and the corresponding large number of inspection attributes for each component, some error in the field inspection J
data is inevitable, in spite of the use of a detailed checking process.
-
The weld measurement errors were biased on the conservative side. There were only slight difference i
in weld measurements between various inspectors.
-
The overall error rate was low, and will not affect the final walkdown results.
The inspector concurred with these conclusion.
Based on the operability reviews performed for the irregularities identified to date, the inspector concluded that random errors in the field inspection data would have a negligible effect on results. Also review of the data showed an improvement in the accuracy of the field data due to field walkdown personnel gaining experience on the project.
(3)
Documentation of weld attributes for welds covered with slag.
l This violation example concerned the fact that Bechtel
.
welding engineers failed to note which welds could not be inspected for cracks, lack of fusion, and other irregularities since they were covered with slag.
The inspection records indicated that these inspections had been performed and that the welds were acceptable for these attributes. After this violation was identified, welding engineers were instructed to note weld quality attributes
,
W
.
--,
-
.-
--
L.
.
.
!
which could not be verified as "not obtainable". A review wac made to determine which welds were covered with slag, and the field inspection walkdown data packages were revised to document the presence of slag.
The inspector reviewed the data which listed the slag
-
covered welds. The inspector also reviewed randomly selected Phase II documentation packages and verified that the packages had been revised to include this information.
(4)
Errors in Calculations This violation example concerned the fact that checkers
'
failed to identify significant errors in calculations initiated to evaluate irregularities. After this problem was identified, several training sessions were held for Bechtel design engineers to cover the quality requirements for design calculations.
Several improvements were incorporated in the work processes. An independent, in-depth review was also performed, on the calculations, by senior Bechtel engineers to ensure consistency and accuracy.
In addition the licensee performed a 100 percent review of the calculations.
After these corrective actions were completed, additional reviews of completed calculations were performed by the inspectors du,ing inspections documented in NRC Inspection Report number 50-325,324/92-40 and 93-02.
Some minor errors were identified by the inspectors, but the errors did not affect the calculation results.
(5)
Inadequste Weld losp ection Procedure.
The welding inspection instructions in Bechtel procedure WDP-002 did not comply with referenced Visual Welding Acceptance Criteria (VWAC).
Specific problems identified were permitting inspection of slag covered welds, and acceptance of groove welds with five percent lack of fusion.
.
VWAC requires removal of slag to perform visual inspections
and permits zero percent lack of fusion.
Procedure WDP-002, Revision 3 was issued to correct the identified problems. The provision for acceptance of groove
,
welds with five percent lack of fusion 9:as revised to agree with VWAC zero percent iack of fusion criteria.
(Review of the completed weld inspetion data confirmed that the five
,
percent criteria had not been used previously.) Appendix A to WDP-002 was revised to cover removal of slag for performance of visual inspections. Also Appendix C, Weld Quality Verification Procedure, was added to provide a basis for acceptance of slag covered welds. This procedure used the sampling plan for weld reinspection discussed in VWAC.
'
.
-
-
-
-
-
i
.
l
.'
l
.
The slag was cleaned from the welds on the 64 heaviest loaded connections (highest stressed) and the welds were inspected. The purpose of the weld reinspection program was to determine if weld quality was a concern.
The inspector reviewed the results of the reinspection which verified that weld quality was not a concern with the structural steel.
Later, the sample size was expanded to 102 connections. No discrepancies were identified.
The original inspection data was found to be acceptable.
Based on review of the completed corrective actions, this violation is closed.
d.
(Closed) Unresolved Item 324/92-27-03, Potentially Overspanned Conduits.
During walkdowns in the Unit 2 drywell and reactor building, several concerns were identified by the inspector and licensee engineers regarding overspanned conduits. The overspan condition exceeded the maximum span length criteria specified in CP&L Specification 048-010, Specification for Installation of Seismic Conduit Supports. Discussions with licensee e.gineers disclosed that development of Specification 048-010 was started in 1982 to address concerns with seismic support of conduits. This specification was developed to implement improved design and installation procedures which are consistent with current standards. All modifications installed since 1982 comply with the improved specifications.
It has been recognized that the conduit installation practices in use when the Brunswick plant was constructed do not meet current seismic design standards. The conduit supports will be reviewed during the upcoming USI A-46 walkdown inspections which are scheduled to begin in September, 1993.
This schedule has been approved by HRR. The licensee will utilize SQUG GIP-2 to perform the walkdown inspections.
During the current outage the licensee has also implemented other programs to address conduit support deficiencies.
These include the hotside/coldside walkdown inspections performed in May - June, 1992, and the material condition walkdown inspections currently being performed in accordance with CP&L procedure OSP-92-076, Special Plant Walkdown Procedure.
During the material condition walkdowns, conduits are inspected for overspan using the criteria in GIP-2.
During both walkdown programs, several conduit support deficiencies were identified by the licensee regarding missing and loose hardware, damaged hardware, and overspan. These problems have been, or will be, corrected via trouble tickets (work requests). The licensee has also written an EER, number 92-0290, to address minor conduit
,
.
i
!
!
support deficiencies identified during the material condition walkdown inspection program.
Examples of minor deficiencies include loose or missing anchor bolts; broken, damaged, missing or loose conduit clamps; missing washers; abandoned hangers; damaged supports; and defective welds.
In order to address the specific conduit overspan problems identified by the inspector, the licensee initiated calculation
,
number OSEIS-0006, Conduit and Cable Tray Report.
The inspector reviewed this calculation which was a summary of actions taken to address the overspanned conduit concerns.
Licensee engineers concluded that the overspanned conduits did not cause adverse conditions or result in operability problem. However, modifications were implemented to upgrade the conduits to meet l
current seismic design standards. This work was completed under Plant Modification 91-041.
The inspector walked down the drywell and examined the modifications listed below. Acceptance criteria utilized by the inspector appear in Specification 048-010 and the design drawings (sketches) listed with the conduit modifications. New conduit supports examined were as follows:
-
Conduit Support CS-91-041-443, installed on drywell i
elevation 17, azimuth 58" - 99*, modification sketch l
SK-91041-C-1476, Sheets 1-7 of 7.
-
Conduit Support CS-91-041-440, installed at elevation 47, azimuth 225*, modification sketch SK 91041-C-1473, Sheet 1 of 1.
i
-
Conduit Support CS-91-041-444, installed at elevation 36,
,
J azimuth 82", modification sketch SK 91041-C-1477, Sheet 1 of 1.
-
Conduit Support CS-91-08.1-446, installed at elevation 39, azimuth 60*-82*, modification sketch SK-91041-C-1479, Sheet 1 o f 1.
-
Conduit Support CS-91-041-472, installed at elevation 16'-
6", azimuth 82", modification sketch SK-91041-C-1679, Sheet
1 of 1.
-
Conduit Support CS-91-041-473, installed at elevation 18, j
azimuth 45*, modification sketch SK-91041-C-1680, Sheet 1 of
'
l.
Attributes inspected were conduit support dimension, member sizes, weld sizes, installed conduit support clamps, and configuration.
No deficiencies were identifie *
i
.'
j The inspector concluded that the licensee's actions to address I
conduit overspan problems were acceptable. The licensee has a program to address overspanned conduits and an ongoing process to
,
correct identified deficiencies.
Implementation of this program for identification of overspanned Unit I conduits will be examined prior to restart of Unit 1.
This will be tracked as Inspection Follow-up Item 325/93-15-02, Evaluate Unit 1 Conduits for Potential Overspan Conditions prior to Unit I restart.
e.
(Closed) Violation 325,324/92-33-01, Failure to Take Prompt Corrective Action to Revise Inadequate Weld Inspection Procedure.
The licensee's corrective actions for this violation are stated in a letter to NRC dated January 8, 1993, Subject: Brunswick Steam Electric Plant, Reply to Notice of Violation, Serial: BSEP-92-l 0056. This violation concerned the failure of the licensee to promptly correct deficiencies identified by NRC in the weld inspection criteria specified in Revision 2 of Bechtel procedure WDP-002, Phase II Walkdown Procedure for Reactor Building Miscellaneous Steel and Drywell Platform Steel.
The deficiencies, which resulted in violation 324/92-27-02,were identified on September 17, 1992, but had not been corrected as of October 23, 1992, the last date of the inspection documented in Inspection Report 50-325,324/92-33.
Subsequent to this inspection, Revision 3 of procedure WDP-002 was issued on October 26, 1992.
The inspector reviewed Revision 3 of
procedure WDP-002 and verified that the deficiencies identified in violation item 324/92-27-02 had been corrected.
These corrective
]
actions are discussed in paragraph 7.c, above.
In order to ensure the timeliness of procedure revisions, CP&L
)
authorized implementation of the procedure change notice (PCN)
system in the Bechtel Engineering Department Project Instructions (EDPIs) used on the Brunswick project. The PCN process allows immediate initiation of corrective actions for procedure deficiencies, and prompt implementation. The inspector reviewed EDPI l-01-01 Engineering Department Procedure System, and verified
.
that the PCN process had been implemented in the Bechtel program.
i f.
(Closed) Unresolved Item 325,324/93-02-03, Nuclear Service Water i
Pump - Short Term Qualification Calculations, i
Discrepancies were identified by the inspector during review of STSI SW pump calculations.
The inspector was concerned that the highest stressed bolts on the upper column flange had not been inspected for corrosion. The calculation discrepancies inc.luded no mass input for node point 22; not modeling approximately 14 inches gf pipe; a torsion of 3200 inch-pounds not considered during transformation; and a thrust load of 15,074 pounds not considered in the uplift force check.
\\
,
,
.
The inspectors discussed the discrepancies with the licensee's engineers and design consultant and reviewed the revised calculation.
Revision 7 to Mcdonald Engineering Analysis, Inc.,
calculation number HE-836, " Seismic Stress Analysis of Vertical Pumps" was issued to resolve the discrepancies discussed above.
'
The critical element is still the top column flange bolts which had a computed stress of 26,369 psi against the allowable stress of 27,000 psi (0.977 of the allowable stress). The stress ratio of 0.977 was reduced from 0.994 in the previous revision because the mass used in Revision 6 was higher than the actual mass value furnished by the pump manufacturer. The stress of 26,369 psi was calculated based on the ASME Code,1980 Edition, which did not require inclusion of the axial loads. (The 1975 Edition of the ASME Code requires that the axial load be included in the stress calculations.)
If the axial stress is added to the above stresses, the total stress is 27,598 psi for the Peerless pumps and 26,692 psi for the Johnston pumps. The stress ratio including the axial load is 1.022, which is only 2.2% above the allowable stress. This is considered to be acceptable.
The revised calculation did consider the uplift for the upward thrust load (caused by flow) of 2,400 pounds, with a resultant of i
a net uplift load of 25 pounds, which is negligible.
The 15074
!
pounds is in the downward direction and does not increase stress
on the anchor bolts. Therefore, the uplift condition is not a controlling case for anchor bolt qualification.
During the current inspection, the licensee was performing maintenance on the Unit 1, 1B Nuclear Service Water Pump. The pump motor had been removed and the upper column flange had been
!
removed. The inspector examined the upper column flange bolts
which had been removed from the pump. The stainless steel bolts l
were in good condition and did not exhibit any signs of corrosion or deterioration. The inspector interviewed licensee maintenance personnel regarding the condition of the upper column flange bolts on the remaining Unit I and Unit 2 service water pumps.
Observations of the maintenance personnel were that the condition
of the IB nuclear service water pump column flange bolts were typical of those on the remaining service water pump column flange bolts. They will be confirmed during inspection of further service water pump maintenance or modification activities.
Based on the revised calculation and inspection of the upper column flange bolts, this unresolved item is closed.
- g.
(Closed) Unresolved Item 324/93-02-04 Pipe Support PS-3568.
During the inspection, this support was incorrectly identified as support number PS-3568. The correct number is PS-3563. The inspection concern was the excessive distance the top of the support frame deflected when a small lateral force was applied.
i
?
.-
--
-
~
.
,
.
!
,
The frame did not appear to be rigidly attached to the elevation 80 platform steel. The frame is constructed from 3x3 inch angles and supports two small spring cans.
Subsequent to inspection number 93-02, licensee engineers reviewed calculation number PS B21-206/91-41 and determined that the support was adequate. The inspector also reviewed this calculations which showed that the supports carry vertical dead weight loads of only 52 pounds from one spring can and only 195
,
pounds from the other. However, licensee engineers did not perform a field inspection of the support. At the inspectors i
requested, licensee engineers performed a field walkdown inspection to verify that the frame was attached sufficiently to
-
support these loads.
Licensee engineers, based on the field
,
walkdown inspection, concluded that the frame was adequate. The inspector re-examined the frame and concurred with this conclusion.
8.
Exit Interview The inspection scope and results were summarized on March 19 and 26, 1993, with those persons indicated in paragraph 1.
The inspectors described the areas inspected and discussed in detail the inspection results listed below.
Proprietary information is not contained in this report. Dissenting comments were not received from the licensee.
a.
Inspector Followup Item 324/93-15-01, followup on Phase II of MSVP for Unit 2 Reactor Building Structural Steel. - Paragraph 2.b.
b.
Inspection Follow-up Item 325/93-15-02, Evaluate Unit 1 Conduits For Potential Overspan Conditions prior to Unit I restart. -
Paragraph 7.d.
9.
Acronyms and Initialisms ACR Adverse Condition Report AISC American Institute of Steel Construction
ASME American Society of Mechanical Engineers i
BNP Brunswick Nuclear Plant
'
BSEP Brunswick Steam Electric Plant (see BNP)
CP&L Carolina Power & Light CRD control rod drive
!
DTOP Design Turnover Program EDPI Engineering Department Project Instructions l
EER Engineering Evaluation Reports
!
FSAR Final Safety Analysis Report GIP-2 Generic Implementing Procedure, Rev. 2.
,
HVAC Heating, Ventilation and Air Conditioning
'
IE Inspection and Enforcement LER Licensee Event Reports i
MSVP Miscellaneous Steel Verification Program
,
.
.
.
-
.
,.-
NAD Nuclear Assessment Department NCR Nonconformance Report NDE non-destructive testing NED Nuclear Engineering Department NRR Nuclear Reactor Regulation PCN procedure change notice psi Pounds per square inch QA Quality Assurance QC Quality Control rebar reinforcing steel RHR Residual Heat Removal SDV scram discharge volume SER Safety Evaluation Report SQUG Seismic Qualification Utility Group STSI short term structural integrity TAC Technical Advisory Committee USI unresolved safety issue UT Ultrasonic test
,
VWAC Visual Welding Acceptance Criteria WR/JO work request
,