IR 05000324/1993002
| ML20044B991 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 03/04/1993 |
| From: | Blake J, Chou R, Lenahan J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20044B986 | List: |
| References | |
| 50-324-93-02, 50-324-93-2, 50-325-93-02, 50-325-93-2, NUDOCS 9303160076 | |
| Download: ML20044B991 (28) | |
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"D UNITED STATES
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NUCLEAR REGULATORY COMMisslON
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REGION 11 R
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,j 101 MARIETTA STREET, N Av.
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'g ATLANTA, GEORGI A 30323
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MAR 4 La93
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Report Nos.:
50-325/93-02 and 50-324/93-02 Licensee: Carolina Power and Light Company P. O. Box 1551 Raleigh, NC 27602
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Docket Nos.:
50-325 and 50-324 License Nos.: DPR-71 and DPR-62 Facility Name: Brunswick I and 2 Inspection Conducted: January 11-15, 25-29, and February 5-12, 1993
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Inspectors:k..cY I
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J. K Lenahan Q. D Date Signed N
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R. th\\Ch p (Jan sky D1-15, and Febru ry 8-12, 1993)
Date Signed Approved by:
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8 - /--f 3-l
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J. J. BhkW Chief Date Signed Materials Processes Section l
4, Engineering Branch l
Division of Reactor Safety i
l SUMMARY
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Scope:
This special announced inspection was conducted in the areas of repairs to masonry walls and seismic instrument racks, the emergency diesel exhaust line
supports and the service water pump, lube water system, pipe supports; the i
hotside/coldside walkdown program; the miscellaneous structural steel
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verification program; investigation and repairs to corrosion of the
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l Unit 2 drywell; evaluation of the service water fer short term structural-integrity; and licensee action on previous inspection findings.
Results:
l In the areas inspected, deviations were not identified, h
A. violation was identified regarding failure to measure and evaluate corrosion-l of the drywell liner plate - paragraph 4.
Weaknesses were identified in housekeeping in the drywell and in the licensee's maintenance program - paragraph 4.
Unresolved items were identified due to discrepancies found in the nuclear service water pump short term qualification calculations - paragraph 6, and an apparent inadequately l
restrained pipe support - paragraph 8.
9303160076 930304 PDR ADOCK 05000324
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REPORT DETAILS
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1.
Persons Contacted
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Licensee Employees
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- J. Bates, Material Condition Walkdown Supervisor, Technical Support '
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- H. Beane, Quality Control Manager
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"*J. Brown, Unit-2 Plant Manager
- S. Callis, Licensing Engineer
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- J. Cowan, Manager, Technical Support and Regulatory Compliance G. Frick, Civil Engineer,' Outage Management and Modifications
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- R. Godley, Manager, NRC Compliance
- L. Grzeck, Project Engineer Miscellaneous. Steel, Nuclear Engineer
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i R. Knott, Principal Engineer, NED
- J. Leininger, Onsite Manager, NED i
- A. '.ucas, Vice-President, NED l
D. McCarthy, Manager, Nuclear Licensing
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- W. Morroe, Supervisory Engineer, NED
- R. Morgan, Unit 1 Plant Manager
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- G. Thearling, Senior Specialist, Regulatory Compliance
R. Tripp, Civil Engineer, NED S. Vann, Miscellaneous Steel Project Manager, NED
- E. Willett, Manager, Outage Management and Modification l
- R. Wojnarawski, Material Condition Walkdown Supervisor, Technical i
Support
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Other licensee employees contacted during this inspection included l
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engineers, technicians, and administrative personnel.
Other Organizations
- R. Bizeck, Civil / Structural Consultant, TENERA
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- R. Gallagher, Project Manager, Bechtel
- T. Synder, Consultant, TENERA l
G. Thomas, Senior Structural Engineer, Bechtel
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NRC Resident Inspectors
- R. Prevatte, Senior Resident Inspector
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- P. Byron, Resident Inspection
D. Nelson, Resident Inspector-l
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- Attended January 29,'1993, exit interview
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- Attended February 12, 1993, exit interview i
- Attended both exit interviews
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Review of Licensee Corrective Actions to Resolve Structural Deficiencies 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-1, Corrosion repairs to service water pump lubrication water piping supports.
Licensee commitment:
Complete corrosion repairs to existing service water lubrication water piping supports.
Discussion:
Questions regarding the effect of corrosion on the structural integrity of the service water lubrication water piping supports were identified by licensee engineers in 1987. During the current outage corroded carbon steel service water lubrication water piping supports were removed and replaced with new stainless steel supports. This work was completed under plant modifications, number 91-047 for Unit I and 90-064 for Unit 2.
The inspector examined the modification documentation packages and walked down the supports listed below and verified that they were installed in accordance with design requirements. Acceptance criteria utilized by the inspector appear in CP&L Specification, BSEP 248-lG7, Installation of Seismic Pipe and HVAC Supports and Miscellaneous Structural-Steel, and the design drawings (Sketches) listed below.
During examination of-the supports the inspector verified support location, support member sizes, weld sizes, anchor bolt size (diameter and end stamp), and method of attachment to piping.
Supports examined and corresponding installation drawings (sketches) were as follows:
-Unit 1 Nuclear SW Pump 1A Support No.
Sketch No.
PS-4317 SK 91047-C-1045, Sheet 1 of 1 PS-4318 SK 91047-C-1055, Sheet 1 of 1 PS-7763 SK 91047-C-1087,' Sheet 1 of 1 PS-4743 SK 91047-C-1080, Sheets 1-3 of 3 PS-4750 SK 91047-C-1106, Sheet 1 of 1 PS-4751 SK 91047-C-1107, Sheet 1 of-1
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-Unit 1 Nuclear SW Pump 1B f
Support No.
Sketch No.
PS-4319 SK 91047-C-1047, Sheet 1 of 1 l
PS-4320 SK 91047-C-1057, Sheet 1 of 1
PS-7764 SK 91047-C-1088, Sheet 1 of 1
PS-4755 SK 91047-C-1075, Sheets 1-3 of 3
PS-4762 SK 91047-C-1126, Sheet 1 of 1
PS-4763 SK 91047-C-1127, Sheet 1 of 1 l-Unit 2 Nuclear SW Pump 2A j
Support No.
Sketch No.
PS-4331 SK 90064-C-1036, Sheet 1 of 1 PS-4332 SK 90064-C-1038, Sheet 1 of 1 PS-7763 SK 90064-C-10xx, Sheet 1 of 1
PS-4803 SK 90064-C-1061, Sheet 1-3 of 3 i
PS-4811 SK 90064-C-1105, Sheet 1 of I f
PS-4810 SK 90064-C-Il04, Sheet 1 of 1 l-Unit 2 Nuclear SW Purup 2B
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Support No.
Sketch No.
l PS-4335 SK 90064-C-1044, Sheet 1 of I f
PS-4336 SK 90064-C-1046, 'neet 1 of 1
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PS-7765 SK 90064-C-1071, Sneet-1 of 1
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PS-4615 SK 90064-C-1063, Sheets 1-3 of 3 i
PS-4823 SK 90064-C-ll25, Sheet 1 of 1 i
PS-4824 SK 90064-C-ll24, Sheet 1 of 1-l t
During the walkdown inspection, the inspector identified a gap of l
1/8 of an inch between support PS-7763 and the Unit'2 SW Pump 2A-i base. The drawings specified no (zero) gap. The licensee issued j
work request (trouble ticket) number WR/JO 93-ACEEI to disposition
the problem. The corrective actions to close' out the trouble
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ticket were completed on January 27, 1993. The inspector re-
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examined the support and. verified that the support was adjusted so l
that no gap existed between the bottom of the support and the pump
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base. Minor discrepancies were identified by the inspector
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regarding incorrect location of~ shims under anchor bolts on support numbers PS 4318 (Unit 1 Pump).and PS 4319 (Unit 1
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i Pump IB).
However.these had no safety significance. All other
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supports were installed within the dimensional tolerances specified in Specification 248-107.
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Conclusion:
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The licensee has completed replacen,ent of the corroded service-water, lubrication water piping supports on Unit I and 2.
Short
term corrective action A-1 is completed and acceptable for restart
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for Units 1 and 2.
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Item A-3, Corrosion repairs of seismic instrument racks Licensee Commitment:
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Address corrosion repairs of seismic instrument racks in accordance with the methodology in Enclosure 2.
Discussion:
The methodology in Enclosure 2 is the licensee's program for evaluating deficiencies to determine if they need to be corrected prior to restart, or corrective action can be delayed until a
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future outage. CP&L procedure PN-30, Integrated Recovery Methodology Specifies the requirements for the operability
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reviews.
In 1987, licensee engineers identified concerns regarding corrosion of the seismic instrument racks, and the effect of the
corrosion on the structural integrity of the racks.
During the j
current outage, the licensee implemented corrective actions to
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replace three Unit 2 instrument racks and completed short term i
repairs to 17 other Unit 2 racks.
The short term repairs included'
addition of lateral braces to some racks and/or installation of new conduit supports and flexible conduits between existing conduits and the instrument racks to prevent seismic interaction between the conduits and racks. Repairs to two other Unit 2 racks, numbers 2-H21-P009 and P010 were in progress during the current inspection.
The inspector examined the structural aspects of the corrosion I
repairs for the racks listed below. The work was performed under
plant modification 92-071, field revision 20. Acceptance criteria
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utilized by the inspector appear in CP&L Specification BSEP 248-107, and the design drawing (sketches) listed below.
Instrument racks examined and controlling installation drawings were as
follows:
-l Rack 2 H21-P014 - Replacement, Sketch number Sk-02071-C-1003, Sheets.I through 38 j
Rack 2 H21-P018 -Replacement, Sketch number Sk-92071-C-1002, Sheets I through 38
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Rack 2 H21-P025 - Modification, Sketch number Sk-92071-C-1025, Sheets 1 through 6
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Rack 2 H21-P036 - Modification, Sketch number Sk-92071-C-1043,
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Sheets 1 through 8 j
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For rack P018, the inspector verified that member size used in
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construction of the racks, weld size, and installation of bolt and l
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miscellaneous hardware were in accordance with appropriate design l
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drawing (sketch) requirements.
For the remaining three racks, the inspector verified configuration of the racks and that support
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member sizes and installation details were in accordance with
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design drawing requirements.
t Conclusions:
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The inspector concluded that the instrument racks were constructed in accordance with requirements shown on the design drawings. The
resident inspector will perform further review of instrument
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installation on the new or modified racks, including calibration
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and post-modification testings.
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Item C-2, Repair to five non-load bearing diesel generator l
building reinforced concrete walls.
Licensee Commitment:
Repair of five reinforced concrete non-load bearing wall panels in
the diesel generator building to restore them to their design
configuration.
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Discussion:
In 1987, licensee engineers identified deficiencies in installation of concrete expansion anchors which support restraining angles (clip angles) at the base, sides, and top of-
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masonry block walls in the= diesel generator building. The purpose
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of the angles is to provide lateral restraint to the walls for
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resisting of seismic loads. The licensee did not perform an
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adequate evaluation of the anchor bolt deficiencies until April,
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1992.
The results of the evaluations showed that the anchor bolt deficiencies involved anchor bolts which were cut and welded to
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the clip angles. These bolts have-been referred to in various documents as inappropriately modified bolts, fraudulent bolts, or.
counterfeit bolts. Additional details regarding the anchor bolt deficiencies are contained in Licensee Event Report number 1-92-012 and NRC Inspection Report number 50-325, 324/92-10 and 50-325,324/92-14. A Notice of Violation / Civil Penalty was issued to the licensee in February,1993 concerning failure to take
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adequate corrective actions regarding the anchor bolt deficiencies.
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Subsequent to investigation of the anchor bolts installed in the
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masonry block wall clip angles, the licensee examined anchor bolts in clip angles restraining five non-keyed reinforced concrete walls in the diesel generator building. This investigation disclosed that the clip angles in these walls also contained counterfeit bolts.
The anchor bolt deficiencies resulted in numerous diesel generator walls being declared inoperable.
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inoperable walls affected operability of vital safety-related
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equipment and resulted in the shut-down of both Units 1 and 2 in April, 1992.
The licensee implemented repairs to the lateral restraints on the -
non-keyed reinforced concrete walls. This work was completed under Emergent Structural Modification 92-011. The repairs consisted of installing concrete anchors in various locations in
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clip angels at the top and bottom of the walls where counterfeit'
anchor bolts had been installed, and installation of new lateral restraint along the sides of the walls. The inspector reviewed the following drawings which specify the repair details:
Sketch No. SK-910ll-C-1000, Sheets 32-36 Wall 9D-1
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Sketch No. SK-910ll-C-1000, Sheets, 37-40 Wall 9D-2 Sketch No. SK-910ll-C-1000, Sheets47-50A Wall 10D
Sketch No. SK-91011-C-1000, Sheets63-66A Wall llD
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Sketch No. SK-910ll-C-1000, Sheets, 79-82A-Wall 120
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The inspector walked down the diesal generator building and
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examined the completed repairs to wall numbers 10D and llD. The inspector verified that repairs had been completed in accordance with design drawing requirements. Attributes inspected included r andomly selected newly installed hardware to verify correct size (diameter and end stamp) of concrete anchors, correct location and i
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spacing of anchors, correct numbers of anchors, and correct size and spacing of new steel plates which provide lateral restraint along the sides of the walls.
Conclusions:
The inspector concluded that the five reinforced-concrete, non-load bearing walls, in the diesel generator building were repaired in accordance with the design drawings requirements. The in progress repair work was examined during an inspection documented in NRC Inspection Report number 50-324,324/92-14. Short term corrective action C-2 is complete and acceptable for restart of i
Units 1 and 2.
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Item C-3, Complete repairs to angle restraints for diesel generator masonry block walls.
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i Licensee Commitment:
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Complete repairs to structural angle restraints for diesel generator building block walls to restore the walls to their
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design configuration.
j Discussion:
Short term corrective action item C-3 involves. replacement of counterfeit anchor bolts in the diesel generator masonry block wall numbers, I through 10. The repair details are specified on
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Sketch number SK-910ll-C-100, Sheets23-27A,.41-46, SI-56A,67-78A, and 83-106B. The inspector reviewed the QC verified
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copies of these sketches and verified that the completed repairs had been inspected and accepted by Quality control (QC) inspection personnel. The inspector examined the completed repairs to these walls and reviewed randomly selected QC inspection records during j
an inspection documented in Inspection Report number 50-
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325,324/92-14.
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Conclusion:
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The walls were repaired in accordance with design drawing
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requirements. Short term corrective action item C-3 is complete and acceptable for restart of Units 1 and 2.
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Item C-4, Design review of non-safety related masonry walls.
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Licensee Commitment:
perform a design review and field inspection review, when l
necessary, of the non-safety masonry walls at the Brunswick Plant
to verify the walls are appropriately classified. Address any i
identified deficiencies in accordance with the methodology ~in j
Enclosure 2.
j Discussion:
During the original IE Bulletin 80-11 analysis, several walls in
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the diesel generator building, control building, and Units 1 and 2
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reactor buildings had been classified as non-safety related.
After the problems with the counterfeit bolts were discovered, the i
licensee performed a comprehensive review of these walls and determined that due to either a change in design _ function, or
installation of safety-related equipment on.or in proximity to
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some non-safety related masonry block walls, several of the walls previously classified as non-safety related in the reactor buildings were reclassified as safety related. All walls previously classified as non-safety related in the diesel generator and control buildings were re-classified as safety related.
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The inspector reviewed the results of the reclassification of the reactor block walls. A total of 68 walls, 30 in the Unit I and 38 j
in the Unit 2 building, are still classified as non-safety
related. A design evaluation was completed on all walls when l
classification was upgraded from non-safety related to safety
related. After completing the design evaluation, license engineers determined that modifications were required to some.
i walls. These are discussed under short term corrective actions l
l C-7 and C-8 (paragraphs 2.g and 2.h, below).
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The inspector walked down the walls on Elevation 98 and 117 in the
Units 1 and 2 reactor building and examined walls on these
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elevations still classified as non-safety related and verified i
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that no safety equipment was located in the proximity of these I
walls.
The inspector previously walked down non-safety related
walls on elevation 20 at the reactor building during an inspection
documented in inspection report numbers 50-325,344/92-45.
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inspector also reviewed the licensce's walkdown inspection data j
for reactor building wall numbers llA,110,15A, ISB,150,150,
15E, and 15F. The walkdown data indicates that there is no safety
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related equipment in the proximity of these walls.
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Discussion with licensee engineers disclosed that wall numbers 15H
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and 151 on elevation 70 of the reactor building are currently
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classified as non-safety related.
The containment atmospheric l
control (CAC) equipment, located in the proximity to these walls
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may be upgraded to safety-related and thus change the
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classification of wall numbers 15H and 15I.. If the classification
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of these walls is changed, the walls will be analyzed, and
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modified if required, to meet FSAR design criteria.
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Conclusion:
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inspection, and during inspection number 325, 324/92-45, the l
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inspector concluded that the classification of masonry block walls
meets FSAR and IEB 80-11 requirements. The re-classification of
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reactor building wall numbers 15H and 151 will be reviewed under short term corrective action item C-6.
Short term corrective action item C-4 is complete and acceptable for restart of Units 1
and 2.
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Item C-5, Remove non-functional masonry wall through-bolts.
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I Licensee Commitment:
Remove accessible non-functional through-bolts and install cover plates over the holes in the diesel generator building walls.
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Discussion:
i During investigation of the counterfeit bolts in the masonry wall
clip angles, the licensee also discovered that some' through-bolts i
supporting the 1/4 inch thick, steel-plate, missile shields on
diesel generator wall numbers 1, 2, 3, 4, 5, 6, 7, 8, 9, and 10
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were also counterfeit. The licensee performed inspections using ultrasonic testing on all accessible masonry wall through-bolts to
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identify any counterfeit through-bolts. Based on results of design evaluation, the licensee decided to remove the counterfeit through-bolts and install cover plates over the holes in the
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missile shield plates. The location of the counterfeit through-
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bolts, and the new cover plates are shown on the following drawings: Sketch SK-010ll-C-1000, Sheets 41A and 42A (wall 1),
Sheets 51A and 52A (wall 2), Sheets 57A and 58A (wall 3), Sheets l
67A and 68B (wall 4), Sheets 73A and 74A (wall 5), Sheets 83A and
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84A (wall 6), Sheet 89A and 90A (wall 7), Sheets 23A and 24A (wall t
8), Sheets 101A and 102A (wall 9), and Sheets 95A and 96A (wall 10).
During an inspection documented in Inspection Report numbers 50-325, 324/92-14, the inspector witnessed the UT testing of the
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through-bolts, reviewed UT data, and observed some of the in
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process work to remove the counterfeit through-bolts and install the new cover plates. During the current inspection, the inspector walked down wall numbers 3, 4, 5, 8, and 10 and verified that the counterfeit through-bolts had been removed and the new cover plates installed in accordance with design drawing requirements. No deficiencies were identified.
Conclusion:
Based on inspections performed during inspection number 50-325, 324/92-14 and the current inspection, the inspector concluded that short term corrective action item C-5 is complete and acceptable for restart of Units 1 and'2.
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Itern C-7, Repair Control Building masonry walls to meet control room habitability requirements.
j Licensee Commitment:
I Complete repairs of upgrading seismic classification of walls i.n
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the control building (elevation 49 foot) that have been determined to be required post-earthquake for control room habitability requirements.
l Discussion:
Cottrol room habitability requirements resulted in re-
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classification of six walls on the west perimeter of the control
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room. The walls required modification to meet design criteria
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requirements. The inspector examined the completed modification i
on three walls during the inspection documented in Inspection Report numbers 50-325,324/92-40.
No deficiencies were identified.
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f Conclusion:
i The modifications to the wall to meet control room habitability requirements have been completed.
This work was inspected during i
inspection number 92-40.
Item C-7 is complete and acceptable for
restart of Units 1 and 2.
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Item C-8, Review IE Bulletin 80-11 Program.
l Licensee Commitment:
l Perform a review of IE Bulletin 80-11 program for the Brunswick Pl ant. The review will address existing masonry wall functions
including missile barrier, tornado barrier, ventilation barrier, I
or other functions for which it is not analyzed.
Addre.ss any i
identified deficiencies in accordance with the methodology in
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Enclosure 2.
l Discussion:
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The review of the IE Bulletin 80-11 program and the counterfeit i
bolt problem resulted in modifications to 65 masonry block walls
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and five reinforced concrete non-load bearing walls.
Six of the l
walls required two modifications. The review of the IEB 80-11
program disclosed that walls required modification due to re-l classification of safety function (i.e. change from non-safety to i
safety related), change in function (e.g. control room i
habitability) and using updated criteria for tornado loads.
The
inspector previously examined ongoing and completed modification i
to the masonry walls during inspections documented in Inspection
Report numbers 325,324/92-14, 92-40 and 92-45.
During the current inspection, the inspector reviewed the wall
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status report which shows all modifications are completed and all-
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walls have been declared operable. The inspector walked down the f
Units 1 and 2 reactor building and the diesel generator building i
and examined the modifications completed under plant modification-
!92-069 to the walls listed in the table below. These
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modifications were required for tornado venting and consis 'd of f
installation of blowout panels, rupture disks, or dampers in the I
wall s.
The inspector reviewed the drawings listed in the table
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below.
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TABLE i
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Masonry Wall Tornado Venting Modifications
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Wall Number i
C - 1000, Sheets 1, 2 Diesel Building 3B
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C - 1001, Sheets 1, 2 Diesel Building 3D
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C - 1002, Sheet 1 Diesel Building 10B
C - 1003, Sheets 1-3 Diesel Building 10B i
C - 1004, sheets 1, 2 Diesel Building 9D
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C - 1005, Sheets 1, 2 Diesel Building 9D
C - 1006, Sheet 1 Diesel Building 9D C - 1007, Sheets 1-3 Diesel Building 90 C - 1010, Sheet 1 Diesel Building 3E C - 1011, Sheets 1-3 Diesel Building 3E C - 1012, sheet 1 Diesel Building 3F
C - 1013, Sheets 1-3 Diesel Building. 3F
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C - 1014, Sheets 1-3 Diesel Building 90 C - 1015 Sheets 1-3 Diesel Building 9C C - 1016, Sheets 1-3 Diesel Building 9C
C - 1017, Sheets 1, 2 Diesel Building 9C
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C - 1018, sheets 1, 2 Diesel Building 9C
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C - 1019, Shee'. I Diesel Building 9C
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C - 1021, Sheet 1 U-l Reactor Building Wall 13B-
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C - 1022, Sheets 1, 2 U-l Reactor Building Wall 138 C - 1023, Sheet 1 U-2 Reactor Building Wall 13B
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C - 1024, Sheets 1, 2 U-2 Reactor Building Wall 13B
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C - 1029, sheet 1 U-1 Reactor Building El.20 Blowout
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C - 1030, Sheet 1 U-2 Reactor Building El.20 Blowout
Window
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- Note: All drawing numbers prefixed by number SK-92069
The inspector concluded that the tornado venting modifications
have been completed in accordance with the requirements shown on
the design drawings. No deficiencies were identified.
l Conclusion-I I
The IE Bulletin 80-11 review, and all masonry wall modifications required as a result of the review, have been completed.
Item C-8 is complete and acceptable for restart of Units 1 and 2.
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Item C-9, Emergency diesel exhaust line modifications.
Licensee Commitment:
Complete long term qualification of the emergency diesel generator exhaust line supports to include tornado loading requirements.
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Discussion:
Re-analysis of the emergency diesel generator exhaust lines for tornado loading requirements showed that extensive modifications t
were required to the existing supports. The modifications
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included removal of some existing supports, modification to-others, installation of new supports, and modification to the
structural steel frames which support the exhaust lines on the roof of the diesel generator building. This work was completed i
under Emergent structural modification 91-011.
The inspector walked down the diesel generator building and
inspected the complete modification to the No. 3 diesel generator
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exhaust line. Acceptance criteria utilized by the inspector were installation details shown on drawing numbers Sk-910ll-C-1113,
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Sheets 1-12, Sk-910ll-C-1057, Sheets 1-3, SK-910ll-C-1066, Sheet 1
and SK-91011-C-1077, Sheet and CP&L Specification 248-107.
During examination of the structural steel frame on the roof of the diesel generator building, the inspector verified new support
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member sizes, weld sizes, bolting details, method of attachment of supports to the exhaust line, and gaps between cupports and
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exhaust the line conformed to drawing requirement and
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specification tolerance.
No deficiencies were identified.
The inspector also made a cursory inspection of the exhaust line supports for diesel generators 1, 2, and 4 and concluded that installation of new or modified supports were complete and
appeared to be in accordance with appropriate drawing requirements.
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Conclusion:
Modifications to the emergency diesel generator exhaust line
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supports have been completed.
Short term corrective action Item C-9 is acceptable for restart of Units 1 and 2.
j Item C-14, Installation of Service water lubrication water pipe
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supports.
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f Licensee Commitment:
Complete design and installation of additional pipe supports for the service-water, lubrication water piping.
t Discussion:
This item was inspected as part of short term corrective action
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A-1, discussed in paragraph 2.a, above.
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Conclusion:
Short term corrective action item C-14 is complete and acceptable for restart of Units 1 and 2.
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Item C-15, Repairs to MCC anchorages.
A Licensee Commitment:
Address seismic repairs of electrical' motor control centers in i
accordance with the methodology in Enclosure 2.
Discussion:
The result of inspection of the motor _ control centers (MCCs) by licensee engineers disclosed that the' anchorage of the MCCs did not comply with the original design drawing requirements. This
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problem was originally discovered in 1989 'and resulted in non-conformance report (NCR) A-89-015.
This NCR was re-examined by licensee engineers during the current outage and resulted in re-
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inspection of all 46 MCCs.
Numerous deficiencies were identified,
including determination that MCC (2-2XB-2) had no anchorage. This MCC was reported as inoperable in Licensee Event Report 92-05 for Unit 2 (Docket No. 0500324) on June 27, 1992.
i The licensee's corrective actions to inspect, evaluate, and repair the MCCs were examined by the inspectors during inspections documented in NRC Inspection Report numbers 50-325,324/92-18,92-i 33, and 92-40.
These inspections also included a review of.
t anchorage of other floor mounted electrical equipment.
Two unresolved items were identified, numbers-325, 324/92-33-02 and 92-18-01.
Unresolved items 325-324/92-33-03 and 324/92-18-01
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have been closed, while unresolved item 325/92-18-01 remains open for Unit 1 pending further review.
Conclusion:
Based on review of the licensee's actions to correct MCC anchorage deficiencies, the inspectors concluded that the short term corrective action item is complete and acceptable for restart of Unit 2.
Item C-15 will: remain open for Unit 1 pending resolution of unresolved item 325/92-18-01.
1.
Item-1, Hot-side walkdown inspection r
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Licensee Commitment:
Perform hot side walkdown inspections. Address any identified-deficiencies in accordance with the methodology in Enclosure 2.
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1 Discussion:
In April, 1992, after the units were shutdown due to structural deficiencies identified with the diesel generator building masonry walls, licensee engineers conducted walkdown inspections of areas which are normally inaccessible (due to high radiation levels)
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when the units are at power. This inspection effort was designated the "Hotside Walkdown." The Unit 2 hotside walkdown inspection program and results were reviewed by NRC inspectors during the current inspection (discussed in parajraph 4, below)
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and during inspections documented in NRC Inspection Report numbers-
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50-325, 324/92-18, 92-20, 92-23, and 92-27.
Conclusion:
The inspector concluded that the Unit 2 hotside side inspections are complete and acceptable for restart of Unit 2.
The licensee is evaluating deficiencies in accordance with CP&L procedure PN-30, Integrated Recovery Methodology.
The results of the PN-30 evaluations for Unit 2 deficiencies will be reviewed by the inspector in a future inspection prior to restart of Unit 2.
The Unit I hotside walkdown inspection results will be reviewed by the inspector prior to restart of Unit 1.
3.
Miscellaneous Steel Verification Project - Units 1 and 2 (37700)
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i The Miscellaneous Steel Verification Project (MSVP) was previously
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inspected and documented in NRC Inspection Report numbers 50-325,
324/92-20, 92-23, 92-27, 92-32 and 92-40.
Details of the MSVP are
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described in these reports.
During the current inspection, the
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inspectors reviewed design calculations and examined repairs to structural steel platforms in the Unit 2 drywell.
Details of the inspection are summarized below:
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a.
Review of Bechtel Design Calculations
The basic goal of the MSVP is to identify any irregularities in
installation of the miscellaneous steel and to determine if the
'
irregularities effect the structural integrity of the miscellaneous steel platforms.
The irregularities will be evaluated for startup (operability) in accordance with FSAR criteria, as supplemented by the licensee, in submittals to the NRC Office of Nuclear Reactor Regulation (NRR). The licensee's design criteria and the overall MSVP have been reviewed and
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accepted by NRR, as documented in a Safety Evaluation Report dated October 8, 1992.
The inspectors reviewed the following Bechtel procedures which
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control evaluation of the irregularities and provide design
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criteria for the' evaluations:
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EDPI-4.90.02, Revision 0, Miscellaneous Steel Verification Program.
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MSVP-C-001, Revision 1, Civil Design Criteria for Irregularity i
Evaluation for the Miscellaneous Steel Verification Program.
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Calculation ORXB-1000, Revision 1, Guidelines for Addressing Generic Irregularities for Reactor Building Miscellaneous Steel.
j The last sentence of paragraph 5.3 of procedure MSVP-C-001, states:
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" Horizontal and Vertical Acceleration Values shall be multiplied by a i
1.2 multi-mode factor (Reference 5)."
Reference 5 is the Generic j
calculation ORXB-1000.
Paragraph 1.2.b of Exhibit E, procedure
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EDPI-4.90.02, Revision 1, states that the acceleration value shall be increased by the 1.2 multi-mode factor when the natural frequency is i
less than ZPA (33 HZ).
The inspector questioned licensee engineers regarding the difference in the procedures regarding use of the multi-mode factor. These discussions disclosed that the statement in j
procedure EDPI-4.90.02 was based on recommendations of the Technical
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Advisory Committee.
Procedure MSVP-C-001 was written based on procedure
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EDPI-4.90.02.
Procedure MSVP-C-0-01 was revised (Revision 2) to clarify the use of the multi-mode factor when natural frequency is less i
then ZPA.
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The inspectors examined the calculations listed in the Table below.
The
types of irregularities evaluated by the calculations are also listed in the Table.
The calculations were reviewed for completeness, accuracy, adherence to design criteria and procedural requirements, and acceptability of calculation methods in accordance with American
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Institute of Steel Construction (AISC) code criteria and good i
engineering practices.
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Table
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Review of Calculation for Evaluation of Walkdown Irregularities
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Item Calculation Revision Walkdown (1)
Type of (2)
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No.
No.
NO.
Package NO.
Irregularities
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1.
2-RB-B-EL.60'-1"
-Loose bolt (P-R/18R-20R)
-Gusset-not installed
_l as designed
2.
2-RB-C-EL.48'-6"
-Welds missing &
i (L-M/22R-23R)
undersized 3.
2-RB-C-EL.10'-0"
-Unstable platform-
(K-L/23R-24R)
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4.
2-RB-B-EL.78'-6"
-Loose bolt (P-S/20R-21R)
5.
2-RB-C-EL.103'-8"
-Corroded base plate (K-L/23R-24R)
-Spalling of Concrete
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6.
2 RIP-1025
2-DW-E-EL.38'-5 1/4"
-Bolt and weld Azimuth 30 -60-)
information not available'
-Clip angle not
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installed as designed
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i 7.
2 RIP-1029
2-DW-E-EL.17'-10 1/4"
-Saw and flame cuts-(Azimuth 351 -9-)
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8.
2 RIP-1043
2-DW-E-EL.17-10 1/4"
-Weld undersized (Azimuth 171 -195-)
-Seat angle missing
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9.
2 RIP-1074 A
2-DW-E-EL.67' 2 1/2"
-Vertical post missing (Azimuth 90 -180-)
i Notes:
(1) Walkdown package number indicates location of structural steel.
i e.g., 2-RB-B-EL.60'-1: (P-R/18R-20R) covers sttoctural steel located'in
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Unit 2 at elevation 60'-1" between column lines P to R and 18R to 20R.
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(2) Calculations are for evaluations of irregularities for restart identified during phase I walkdowns, except for item numbers 6 to 9 which covers irregularities identified during the phase II walkdown in j
the Unit 2 drywell.
j A discrepancy was identified in calculation 2 RIP-1043. On page 6 l
of this calculation, the acceleration factor for the vertical
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direction was selected using a calculated frequency of 59 HZ.
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This resulted in an acceleration of 0.11.
Paragraph 5.3 of Design Criteria No. MSVP-C-001, " Civil Design Criteria for Irregularity Evaluation for the Miscellaneous Steel Verification Program," Rev.
1, states that " Vertical accelerations shall be taken as the peak value from the vertical ground response spectra." The peak vertical acceleration value is 0.23.
On page 6 of this same calculation, the inspector also questioned the calculation of Mx, using a moment arm taken from the edge of the nut / bolt instead of the center of the bolt. As a result of t
the inspector's questions, the calculation was revised using the peak acceleration value and the moment arms from the center to the
bolt. The loads in the original calculation were based on a 40.0
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psf shear load.
Recomputing the loads using the actual dead loads
,
acting on the platform, per procedure MSvP-C-001 paragraph 5.1, the peak vertical acceleration factor, and the moment arm acting from the center of the bolt resulted in calculated stresses within design allowable values. Therefore the calculation discrepancies had no safety significance.
However the licensee elected to repair this connection as a " good engineering practice " The licensee issued Revision 1 to the calculation to reflect the above changes.
The inspector also reviewed calculation numbers 2 RB2-1123-92077 and 2 RB2-1237. Calculation 2RB2-ll23-92077 was issued to address modification of irregularities in walkdown package 2-RB-C-El 10.0
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(K-L/23R-24R) which do not meet operability criteria.
This package concerns a platform which is existing but was not shown on the design drawings.
The analysis resulted in a modification. The modification details are shown on drawing numbers SK-92077-C-1005, Sheets 1 through 6.
Calculation 2RB2-1237 evaluates the results
of load testing performed on single sided or single line fillet welds connecting shear plates to channels.
Based on the testing results, the licensee was able to accept some shear plate connections with single sided or single line fillet welds.
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b.
Examination of Repairs to Miscellaneous Structural Steel Deficiencies (irregularities) identified on structural steel platforms are being repaired under trouble tickets (i.e.
maintenance work orders), or plant modifications.
During the
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inspection documented in NRC Inspection Report Numbers 50-325, 324/92-40, the inspector reviewed trouble tickets issued to repair connections in the Unit 2 drywell.
The trouble tickets covered
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replacement of bolts, installation of missing bolts, or correction of corrosion problems.
The inspector also examined repairs which
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involved reinforcement of welds on shear plates on the elevation l
17 platform. During the current inspection the inspector walked down the elevation 17, 38, 52, 67, and 80 drywell platforms and
examined repairs to the platform completed under plant modification 92-077.
These repairs involved addition to new
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members, installation of shim plates under columns, addition of
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new welds (other than welds on shear plates), and reinforcement of existing members.
Acceptance criteria utilized by the inspector appear on the plant modification drawings and in CP&L Specification BSEP 248-107. The modifications are shown on the following drawings (sketches):
SK-92077-C-1013; C-1029; C-1032 through C-1047; C-1051, C-1055 through C-1059; C-1063 through C-1078, C-1080, C-1082, and C-1083.
During the walkdown, the inspector examined configuration of the modifications, and on a
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random basis checked new member and weld sizes.
No discrepancies were identified.
During the walkdown inspections, the inspector identified a channel at azimuth 0, elevation 38 which had been coped at an opening in the platform to accommodate thermal growth of a large diameter pipe. The inspector reviewed the walkdown package and determined that this irregulatory had not been identified by the Bechtel Field engineers during the Phase II walkdown inspections.
As a result of this finding, Bechtel field engineers inspected all other such openings and identified one other coped channel at a similar location which was not identified in the Phase 11
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walkdown. This problem will be examined further in closeout of violation item 324/92-27-02 in a future inspection.
During the walkdown inspection, the inspector identified a pipe support which appeared to be attached to the toe plate on the elevation 80 platform.
This problem is discussed in paragraph 8 below.
Conclusions:
Based on review of the calculations, the inspectors concluded that calculations were satisfactory although some minor errors were identified.
The discrepancies identified did not affect operability of any connections.
The inspectors also concluded that modifications to the Unit 2 drywell structural steel platforms appeared to have been completed in accordance with the design drawing requirement, although some trouble tickets remained open and will require completion prior to restart of Unit 2.
Additional modifications to Unit 2 structural steel platforms located outside the drywell will be examined-in a future inspection.
Closecut documentation for modifications will also be reviewed.
Violations or deviations were not identified.
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4.
Walkdown Inspection Program Short term corrective action items E-1 and E-2 were previously referred to as the "hotside" and " cold side" walkdown inspections. The licensee now refers to these program as the " Material Condition Walkdowns".
The
inspector reviewed procedure number OSP-92-076, Special Plant Walkdown Procedure, which was developed to control the walkdown inspection.
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These walkdowns are performed by multi-disciplined teams consisting of civil, electrical, and mechanical personnel and the procedure contains detailed instructions for conducting the inspections, inspection criteria, and guidance on evaluation of potential deficiencies. The licensee has completed the material condition inspections under OSP-92-076 in 24 areas in Unit 2.
These area included all the areas previously classified as hotside, i.e., inassessible during plant operation, plus the Unit 1-2 control room overhead areas. Approximately 4300 differences were evaluated.
The inspector walked down the Unit 2 areas listed below and reviewed the results of the licensee's material condition walkdown inspection. Areas examined were as follows:
elevation 5.0 of the drywell, torus, RWCU pump rooms 2A and 2B, Elevation 61 penetration room, and elevation 20 mini steam tunnel. The inspector identified some additional minor discrepancies in the drywell and mini steam tunnel which had not been identified by licensee walkdown personnel.
However the inspector's overall conclusion was that the licensee walkdown personnel had performed thorough, in depth inspections in these areas and had identified the most significant degraded conditions.
The inspector concluded that the results of this inspection program were satisfactory for closecut of short term corrective action item E-1.
During the walkdown in the drywell, the inspector noted that the liner plate had been corroded at the intersection of the liner with the interior Elevation 5 concrete floor, around the entire circumference of the drywell. The material condition walkdown personnel had identified these as a general item regarding excessive corrosion of supports, etc.
on the elevator 5 level.
The inspector questioned licensee engineers regarding the significance of the liner corrosion, and whether it had been identified previously during inspections conducted to comply with Technical Specifications (TS) 4.6.1.4.1 and 4.6.2.1.e;1.
These discussions disclosed that the TS required inspections were performed once every 18 months in accordance with periodic test procedure PT-20.5.1, Primary Containment Inspection.
Licensee engineers indicated that the corrosion had been previously identified and it was not considered a problem since the licensee hao a corrosion monitoring program in the Torus, and the containment was normally purged with nitrogen which would inhibit corrosion.
The inspector reviewed the PT20.5.1 inspection data sheets completed for inspections performed since 1985 in Unit I and since 1984 in Unit 2.
Review of the data sheets disclosed that in the 1987 Unit 1 and 1988 Unit.2 inspections, rust and corrosion or failed coatings were identified on the liner plate.
Procedure PT20.5.1, paragraph VI. c.2 requires that indications, which are defined as ru:;t stains or corrosion, be measured, sketched and described on the data sheets. Paragraph II of PT20.5.1 requires indications be evaluated per procedure ENP-12.
Contrary to this requirement, the licensee did not measure the depth or extent of the
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corrosion in Unit 1 in 1987, 1989, and 1991, and in Unit 2 in 1988, 1990, and 1991.
After questions by the inspector, the licensee initiated an indepth inspection and evaluation of the drywell liner t
corrosion in Unit 2 which disclosed that the depth of corrosion had extended one-half the liner plate thickness in some areas.
The extent and depth of corrosion in Unit 1, had not yet been determined by the licensee as of the current ending inspection date. The licensee issued an EWR to inspect and repair the corroded areas.
The repair program is described in paragraph 5, below.
The failure of the licensee to measure
,'
and evaluate the drywell liner corrosion was identified to the licensee as an example of failure to implement the procedures required by TS 6.8.1.a.
This was identified as Violation item 325,324/93-02-01,
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Failure to Measure and Evaluate Corrosion of the Drywell Liner Plate.
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During the walkdown inspections of the Torus, the inspector also noted
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that extensive corrosion had occurred in some areas of the Torus on the liner plate. The corrosion has primarily occurred in areas where the original protective coatings have been removed to install new supports.
Similar areas were also identified on the drywell liner. Discussions
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with licensee engineers disclosed that painting had not been performed
,
in the drywell-torus for the last 10 to 12 years. The decision not to
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paint in these areas was made by CP&L management because of concerns i
over the qualification of the coatings. The inspector expressed concern regarding the lack of a coating program to protect the liner from i
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corrosion.
The lack of a qualified coating program was identified to j
the licensee as a weakness in the maintenance program.
,
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The inspector, accompanied by the Senior Resident Inspector and the Unit 2 Plant Manager, walked down the drywell and examined the corrosion on
the liner plate, housekeeping in the drywell, and repairs to structural l
steel platforms.
The inspectors noted large numbers of tools, fasteners and other pieces of hardware and debris scattered throughout all
elevations of the drywell.
Licensee management was informed that there was a weakness in their housekeeping programs.
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The inspector also walked down the Unit 2 reactor building and examined i
housekeeping and work in progress. During the walkdown, the inspector observed repairs to service water booster pumps 2B-and 20. During l
examination of an embedded plate which had been removed from a support pedestal adjacent to pump 2B, the inspector noted that seven of eight Nelson Studs which anchor the plate to the concrete had corroded and j
were no longer attached to the plate. Discussions with licensee
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engineers disclosed that an adverse condition report (ACR) had been l
E drafted to investigate and disposition this problem.
The inspector will i
review the results of the licensee's investigation of their problem in a l
future inspection.
Inspector followup item 325, 324/93-02-02, Corrosion
of Nelson Studs on Embed Plates, was identified to track and review this
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problem.
Deviations were not identified.
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i 5.
Repairs to Unit 2 Drywell Liner f
The inspector examined the Unit 2 drywell liner in areas where the
corrosion had occurred and reviewed the licensee inspection data
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i reports.
The inspector also reviewed Special Procedure 05P-93-010,
Drywell Liner Corrosion Examination. This procedure was used to control examination af the drywell liner to quantify the extent and depth of the
!
corrosion. The procedure provides instructions for cleaning of the
[
liner plate, designation of inspection zones, and instructions for
measurement of the base metal plate thickness and depth of pits in i
corroded areas. The plate thickness was measured using Ultrasonic
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Testing (UT) methods in areas adjacent to and unaffected by the corrosion. The depth of corrosion was measured using dental molding.
The inspector reviewed the results of the UT data which established the l
base metal thickness. The inspector also reviewed the results of the licensee's visual inspection which established the most high'.y corroded areas.
These areas were selected to be measured using the dental
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molding material. The inspector walked down the drywell and examined the
!
areas selected by the licensee as those which appeared to be the most i
'
severely corroded.
The inspector concurred with the areas selected, and also identified two others which appeared to be also severely corroded.
!
The inspector subsequently examined the dental molds and the results of
measurements made to determine the depth of the pits.
This data was
,
recorded on data sheets in Attachment B of prrcedure OSP-93-010. The
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data showed that the two areas selected by the inspector were not as i
,
severely curroded as those selected by the licensee.
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l Based on evaluation of the corrosion depth measurements, the licensee
determined that five areas required base metal weld repairs to restore the liner plate to a thickness greater than.the minimum value of 0.20
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inches.
The original liner plate had a nominal thickness of 5/16
inches.
The inspector reviewed Engineering Evaluation, EGR No. 93-0173,
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which documents the evaluation of the corrosion and the corrective
'
actions; including weld repair, coating (painting) of the liner,' and
installation of a seal in the expansion joint between the liner plate i
and concrete floor. The inspector examined the weld process sheet which l
l specified weld areas, cleaning methods, the welding procedure, special
!
requirements to limit maximum interpass temperature to 175 F, and 14DE l
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test requirements. The inspector examined the completed weld repairs of
the five areas, designated,-8-10, 11 + 30, 13 4 2, 14 + 20, and 17 + 44.
- The inspector reviewed the weld data repair steets and the results of the liquid peneirant test performed on the weld repair areas. The data indicated that the weld repair was acceptable.
I Violations or Deviations were not identified.
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6.
>hort Term Qualification of Service Water Pump - Units 1 and 2 (37700)
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During an NRC inspection documented in f4RC Inspection Report 50-325, j
j 324/92-10, the inspectors found that the nuclear service water pumps had j
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been short-tera qualified and seismically degraded for about 11 years.
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The licensee discovered that inappropriate seismic spectra curves had
been used during the original design qualification of the ten nuclear service water pumps and initiated modifications to the pumps. The licensee requalified the pumps under short-term basis due to the problems with using incorrect response spectra. During that inspection,
,
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the inspectors noted that the pump base (sole plate) and hold down bolts
-
were severely corroded and questioned licensee engineers regarding effect of the corrosion on the short term qualification calculations.
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During the current inspection, the inspectors reviewed calculations
'
completed to short-term quality the pumps, walked down the service water
<
building and examined the condition of the pumps and associated j
hardware, and discussed the long term planned corrective actions with
!
licensee engineers.
,
During the walkdown inspection, the inspector noted that the pump base i
(sole plates) and bolts had been cleaned and did not appear to be corroded. The inspectors reviewed ultrasonic inspection (UT) records for i
all ten sole plates which indicated the minimum thickness of the sole i
plates was 0.82 inches. The original steel sole plate thickness was one inch.
The inspectors reviewed the maintenance work requests which j
documented replacement of corroded bolts and nuts in the pump base to
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sole plate mounting for all ten pumps and the pump discharge to electric l
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motor mounting for one pump. These work requests (WR) included WR/JO i
J 92-AMHB1, 92-AMHD1, 92-AMHEl, 92-AMHF1, 92-AMHG1, 92-AMHII, 92-AJSQl, 92-AJSR1, 92-AJYF1, 92-AJYH1, and 92-AJYJ1. The work requests documented repair instructions, work performed, materials used, and bolt
torque data sheets.
The inspectors reviewed portions of the following calculations:
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Mcdonald Engineering Analysis, Inc, Calculation Number ME-836, " Seismic
Stress Analysi.s of Vertical Pumps", Rev. 6; and CP&L Cclculation number
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OSW-0033, " Analysis for Seismic Qualification of SWP Outliners, Rev.1.
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i Calculation Number ME-836 was initiated in 1981 to qualify the pumps under short-term basis while the licensee purchased new pumps and
completed any required modifications. However, the licensee delayed the l
purchase of new pumps, therefore, this calculation was subsequently
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revised six times (current revision is Revision 6) to accommodate new-j information, replacement of materials, degraded conditions, new code
requirements, new computer methodology, new nozzle loads, etc. This I
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calculation covers the seismic, stress, and deflection analysis of the Nuclear Service Water Pumps. United Engineers and Constructors (UN&C)
Design Guide SDG-7 and ASME B&PV Code Section III Class 3 were used for
'
analysis guidance.
' -
The analysis is directed toward verifying both the structural integrity and functional capability of the pump. The ANSYS computer code was used
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for the dynamic analysis since the lateral frequencies were below 33 l
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Hertz. The computer model was a typical lumped mass model at node j
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points wh ch were connected by massless elements. The loads used to
qualify ine structural elements and components were combined from dead l
weights, (including motor, pump, water, pipe, shaft, etc), pressure, i
seismic, transient, and nozzle. This calculation was for qualification
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of the structural elements and components based on the short-term
operation criteria using the Design Base Earthquake (DBE) loads and i
normal loads. Stresses were computed and checked against allowable
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stresses for major components including columns (encased 18" diameter
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pipe), column flange boits, anchor bolts, baseplate, column flange, i
motor hold down bolts, motor support flanges, nozzle, shaft deflection, j
and lube water pipe.
The degraded condition of the components were
considered in the calculation such as the reduced bolt cross-sectional
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areas, and the existing thickness of 0.82" versus the original thickness
'
of one inch for the sole plate. The critical element is the top column
',
flange bolts which had a computed stress of 26,829 psi against the allowable stress 27,000 psi (0.994 of the allowable stress).
.
During the calculation review, the inspectors noted the following discrepancies:
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No mass was input for node point 22 which represents approximately 2.8 feet of water and pipe weights.
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Approximately 14 inches of pipe and water were not modelled.
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(This mass may have been included in the bowl weight.)
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A torsion of My = 3200 inch - pounds transferred from the
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nozzle load was not considered in sole plate anchor bolt l
check on p. 13 of the calculation.
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The thrust load of 15,075 lbs shown on p.16 of the I
calculation was not considered in the anchor bolt check for the sole plate on p.13 of the calculation.
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t The above discrepancies had been discussed with the licensee's engineers I
and design consultant.
Subsequent to the inspection, the inspectors
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discussed the need to revise the current calculation ME-836 to account for the above discrepancies.
The apparent errors in the design
,
calculations have been identified to the licensee as Unresolved item i
50-325, 324/93-02-03, Nuclear Service Water Pump Short Term
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Qualification Calculations.
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The portions of calculation (Attachment D) 0SW-0033 which cover the
,
analysis of the sole plates were reviewed. This calculation showed that i
the plate stress with the actual thickness of 0.82" was below the allowable stress and acceptable.
The computer model STARDYNE was uLed
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in the analysis to compare the results with Calculation ME-836.
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The inspectors questioned licensee engineers regarding condition of the l
bolts in the top column flange. These discussions disclosed that these j
bolts, due to their location were difficult to inspect and that the
results of the inspections may have inconclusive. This matter will be
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reviewed further in a future inspection, along with the licensee's plans and schedule for replacement of the current pumps with new models which will be long term qualified.
Violations or deviations were not identified.
7.
Short Term Structural Integrity (STSI)
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STSI items are those identified by licensee personnel, which, after l
evaluations by NED, are determined to be operable, although they do not l
meet design criteria established by the FSAR.
The operability reviews
are performed in accordance with Design Guide 11.20, civil / structural t
operability review. As a result of questions raised by NRC personnel
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regarding field validation of design assumptions used in STSI
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calculations, the licensee committed to performing a third-party review
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of STSI items to assess analysis assumptions and field conditions.
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Attachment B to Design Guide DG-III.16 provides the scope of work for i
the third party review of STSI activities.
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The inspectors reviewed calculation number 0 MUD-1000, summary of closure on non-pipe support STSI Items, and EQE Engineering Consultants report
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number 52171.03-R-001, Brunswick Nuclear Station Third Party Review of
Potential Adverse Structural Mechanical Conditions. These reports l
summarized the third party review of several STSI items and included
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field validation of design assumptions. The overall conclusion of.the
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various EQE documents included in calculation OMUD-1000, and the
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referenced EQE report, was that all STSI items evaluated were determined
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to be short term qualified, although some discrepancies were identified.
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The inspectors noted that-EQE did not identify the discrepancies in the nuclear service water pump calculations discussed in paragraph' 6, above, i
however EQE did identify a minor error in the torsional moment used in
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some members in the ANSYS computer model, which they concluded would not '
affect the final results.- The inspector will perform' additional review -
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of these reports and perform independent checks of selected STSI-items t
in a future inspection.
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The inspectors reviewed the status of items currently classified as STSI
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discrepancies identified under the design turnover project.
Discussions
with licensee engineers also disclosed that the items identified during
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the walkdown inspections are currently being evaluated for inclusion on the STSI list. The overall status of. STSI items will be reviewed by.the
inspectors prior to restart of Unit 2.
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Violations or deviations were not identified.
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Ji 8.
Inspection of Pipe Supports j
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During the walkdown to inspect completed-structural steel modifications l
in the Unit 2 drywell, the inspector identified a concern with pipe
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support number PS-3568. This support was located at azimuth 225* at-
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elevations 80. The support frame appeared to be attached to the
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platform steel toe plate, and deflected more that six inches when pushed '
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with a slight effort. This frame support two spring cans which support t
two small bore pipes which connect to the reactor vessel.
The inspector questioned the structural integrity of this support and whether it complied with design requirements.
Pending further review of the support by the licensee and NRC, this problem was identified to the
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licensee as unresolved item 324/93-02-04, Pipe Support PS-3568.
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Violations or deviations were not identified.
9.
Licensee Event Reports (LER)
(Closed) LER 2-92-005:
Inoperability of Safety Related Equipment Due to-Seismic Support and Anchorage Deficiencies / MCC 2-2XB-2
On May 27, 1992, during inspection of Unit 2 safety related motor
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control centers, licensee engineers identified deficiencies in the
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anchorage of MCC 2-2XB-2 which resulted in the MCC, being declared inoperable on May 28, 1992.
This problem had existed since original
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plant construction.
The licensee submitted written reports on this LER
to NRC in letters dated June 26, 1992 and October 19, 1992. The
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licensee's corrective action to repair the Unit 2-anchorages on the MCC was examined by the inspectors during inspections documented in inspection report numbers 325, 324/92-33 and 92-40..In the October 19, supplemental report, the licensee committed to revise procedure ~ numbers ENP-12, DG-II.20, and DG-Ill.16 to emphasize the need to thoroughly
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verify as-built records and conduit field inspections prior to making i
The inspector reviewed the current revisions-
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of these procedures and verified that they had been-revised to include these requirements.
The licensee also committed to train appropriate
Technical Support and NED personnel on the revised procedures. The licensee was unable to provide the inspectors with records during the current inspection documenting that this training had been performed.
The licensee made the same commitment to' provide training on these
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revised procedures in LER 2-92-003 and in response to violation 325,
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324/92-11-01. These training records will be reviewed in closecut at the violation.
LER 2-92-005 will be closed since repairs-to the MCCs have been completed.
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10.
Actions on Previous Inspection Findings a.
(Closed) Unresolved Item 324/92-18-01, Welding Qualification of Motor Control Center Panel Anchorages.
This item was originally identified by the inspector because licensee engineers used the
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design drawing information to qualify the anchorage of a motor control center in lieu of using as-built information. After being
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questioned by the inspector, the licensee revised the calculations
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using the actual weld size which was 1/16 to 1/8 inches smaller that the size shown on the design drawing.
The calculation showed that the as-built anchorage for this MCC was adequate.
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f The licensee identified similar concerns related to failure of
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design engineers to adequately verify field conditions and design assumptions when performing operability reviews in Licensee Event Report number 92-003 and 92-005. The Resident Inspector also
identified a violation, number 325, 324/92-11-01, for failure to
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adequately verify actual as-built conditions when performing operab;11ty reviews. The licensee's corrective actions to resolve this problem included revising procedures and training appropriate personnel. These corrective actions will be followed up in closeout of the violation.
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During inspection number 325,324/92-40, the inspectors
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questioned calculation no. 52171-C-003 regarding the use of a four
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percent damping value, the cabinet weights used in the l
calculations, and the use of the corners of the cabinets as pivot points. The inspectors reviewed a comparison of this calculation
with calculation number 84-340-17, ERFIS In-place Testing Results j
- Control Building Equipment Qualification. Calculation 84-340-17
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is a summary of testing performed which verified that the electrical equipment installed at the site meets FSAR criteria and
design requirements. Based on the comparison of data used in the
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two calculations, the inspectors have no further questions j
regarding calculation 52171-C-003. The values used for the weight i
are conservative, the four percent damping value is based on insite test data, and the rigidity of the cabinets justify the use
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of the cabinet corners as pivot points.
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(0 pen) Unresolved Item 325/92-18-01
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Reviews of calculation number 1 RBI-1028 showed that Unit 1 MCC numbers IXA2, IXDB and IXB2 were qualified using existing 1/4 l
diameter shipping bolts. During examination of the anchorage for l
these MCCs, the inspectors noted that some of these bolts didn't
have nuts installed.
Licensee engineers stated that the nu; ware
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not required since they were installed in threaded (" tapped)
I holes in the shipping channels.
Pending further review of the justification for the adequacy of the-installation and
qualification of the 1/4 inch shipping balts, this unresolved item
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will remain open for Unit 1.
In. Unit 2, the shipping bolts were i
replaced with high strength bolts on MCC where they were used for j
anchorage.
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11.
Exit Interview The inspection scope and results were summarized on January 29 and
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February 12, 1993, with those persons indicated in paragraph 1.
The-
.l inspector described the areas inspected and discussed in detail the
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inspection results listed below. Proprietary information is not
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contained in this report. Dissenting comments were not received from
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the licenses
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Violation 325,324/93-02-01, Failure to Measure and Evaluate
'l Corrosion of the Drywell Liner Plate' - Paragraph 4.
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Inspector Followup Item 325, 324/93-02-02, Corrosion of Nelson Studs on.
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Embedded Plates - Paragraph 4.
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Unresolved Item 325, 324/ 93-02-03, Nuclear Service Water Pump Short
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Term Qualification Calculations -L Paragraph 6.
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l Unresolved Item 324/93 02-04, Pipe Support PS 3568 - Paragraph 8.
Subsequent to the inspection, on February 22, 1993, in.a telecon with the NED Onsite Manager, the inspector discussed the need to revise
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calculation number ME-836 to correct the discrepancies discussed in
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paragraph.6 above.
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